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| | number = ML13081A102 | | | number = ML13081A102 |
| | issue date = 06/30/2011 | | | issue date = 06/30/2011 |
| | title = Updated Final Safety Analysis Report, Revision 16. Chapter 15, Accident Analyses. | | | title = Updated Final Safety Analysis Report, Revision 16. Chapter 15, Accident Analyses |
| | author name = | | | author name = |
| | author affiliation = Arizona Public Service Co | | | author affiliation = Arizona Public Service Co |
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| =Text= | | =Text= |
| {{#Wiki_filter:PVNGS UPDATED FSAR CHAPTER 15 ACCIDENT ANALYSES CONTENTS Page | | {{#Wiki_filter:}} |
| : 15. ACCIDENT ANALYSES 15.0-1
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| ==15.0 INTRODUCTION==
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| 15.0-1 15.0.1 CLASSIFICATION OF TRANSIENTS AND ACCIDENTS 15.0-2 15.0.1.1 Format and Content 15.0-2 15.0.1.2 Event Categories 15.0-2 15.0.1.3 Event Frequencies 15.0-3 15.0.1.4 Events and Event Combinations 15.0-4 15.0.1.5 Section Numbering 15.0-10 15.0.1.6 Sequence of Events Analysis 15.0-11 15.0.2 SYSTEMS OPERATION 15.0-14 15.0.2.1 Reactor Protection 15.0-14 15.0.2.2 Engineered Safety Features 15.0-16 15.0.2.3 Control Systems 15.0-16 15.0.2.4 Loss of Off-site Power Following Turbine Trip 15.0-17 15.0.3 CORE AND SYSTEM PERFORMANCE 15.0-21 15.0.3.1 Mathematical Model 15.0-21 15.0.3.2 Initial Conditions 15.0-28 15.0.3.3 Input Parameters 15.0-30 15.0.4 RADIOLOGICAL CONSEQUENCES 15.0-31 15.
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| ==0.5 REFERENCES==
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| 15.0-34 June 2005 15-i Revision 13
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 15.1.1 DECREASE IN MAIN FEEDWATER TEMPERATURE 15.1.1 15.1.1.1 Identification of Causes and Frequency Classification 15.1.1 15.1.1.2 Sequence of Events and System Operations 15.1-1 15.1.1.3 Core and System Performance 15.1-3 15.1.1.4 RCS Pressure Boundary Barrier Performance 15.1-4 15.1.1.5 Containment Performance and Radiological Consequences 15.1.6 15.1.1.6 Conclusions 15.1.6 15.1.2 INCREASE IN MAIN FEEDWATER FLOW 15.1-7 15.1.2.1 Identification of Causes and Frequency Classification 15.1.7 15.1.2.2 Sequence of Events and System Operation 15.1-7 15.1.2.3 Core and System Performance 15.1-8 15.1.2.4 RCS Pressure Boundary Barrier Performance 15.1-10 15.1.2.5 Containment Performance and Radiological Consequences 15.1.11 15.1.2.6 Conclusions 15.1.11 15.1.3 INCREASE IN MAIN STEAM FLOW 15.1-12 15.1.3.1 Identification of Causes and Frequency Classification 15.1-12 June 2011 15-ii Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.1.3.2 Sequence of Events and System Operation 15.1-13 15.1.3.3 Core and System Performance 15.1-20 15.1.3.4 RCS Pressure Boundary Barrier Performance 15.1-28 15.1.3.5 Containment Performance and Radiological Consequences 15.1-29 15.1.3.6 Conclusions 15.1-30 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR ATMOSPHERIC DUMP VALVE 15.1-30 15.1.4.1 Identification of Causes and Frequency Classification 15.1-30 15.1.4.2 Sequence of Events and System Operation 15.1-31 15.1.4.3 Core and System Performance 15.1-37 15.1.4.4 RCS Pressure Boundary Barrier Performance 15.1-45 15.1.4.5 Containment Performance and Radiological Consequences 15.1-47 15.1.4.6 Conclusion 15.1-56 15.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - OPERATING MODES 1 AND 2 15.1-56 15.1.5.1 Identification of Causes and Frequency Classification 15.1-56 June 2011 15-iii Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.1.5.2 Sequence of Events and System Operation 15.1-57 15.1.5.3 Core and System Performance 15.1-66 15.1.5.4 RCS Pressure Boundary Barrier Performance 15.1-85 15.1.5.5 Containment Performance and Radiological Consequences 15.1-87 15.1.5.6 Conclusions 15.1-95 15.1.6 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - OPERATING MODE 3 15.1-96 15.1.6.1 Identification of Causes and Frequency Classification 15.2-96 15.1.6.2 Sequence of Events and System Operation 15.1-96 15.1.6.3 Core and System Performance 15.1-103 15.1.6.4 RCS Pressure Boundary Barrier Performance 15.1-112 15.1.6.5 Containment Performance and Radiological Consequences 15.1.113 15.1.6.6 Conclusions 15.1.113 15.
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| ==1.7 REFERENCES==
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| 15.1.114 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2-1 15.2.1 LOSS OF EXTERNAL LOAD 15.2-1 15.2.1.1 Identification of Event and Causes 15.2-1 June 2011 15-iv Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.2.1.2 Sequence of Events and System Operation 15.2-1 15.2.1.3 Analysis of Effects and Consequences 15.2-1 15.2.1.4 Conclusions 15.2-2 15.2.2 TURBINE TRIP 15.2-2 15.2.2.1 Identification of Event and Causes 15.2-2 15.2.2.2 Sequence of Events and Systems Operation 15.2-3 15.2.2.3 Analysis of Effects and Consequences 15.2-3 15.2.2.4 Conclusions 15.2-4 15.2.3 LOSS OF CONDENSER VACUUM 15.2-4 15.2.3.1 Identification of Causes and Frequency Classification 15.2-4 15.2.3.2 Sequence of Events and System Operation 15.2-5 15.2.3.3 Core and System Performance 15.2-11 15.2.3.4 Reactor Coolant System Barrier Performance 15.2-14 15.2.3.5 Radiological Consequences and Containment Performance 15.2-15 15.2.3.6 Conclusions 15.2-15 15.2.4 MAIN STEAM ISOLATION VALVE CLOSURE 15.2-16 15.2.4.1 Identification of Event and Causes 15.2-16 June 2005 15-v Revision 13
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.2.4.2 Sequence of Events and Systems Operation 15.2-16 15.2.4.3 Analysis of Effects and Consequences 15.2-16 15.2.4.4 Conclusions 15.2-17 15.2.5 STEAM PRESSURE REGULATOR FAILURE 15.2-17 15.2.6 LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2-17 15.2.6.1 Identification of Event and Causes 15.2-17 15.2.6.2 Sequence of Events and Systems Operation 15.2-18 15.2.6.3 Analysis of Effects and Consequences 15.2-19 15.2.6.4 Conclusions 15.2-19 15.2.7 LOSS OF NORMAL FEEDWATER FLOW 15.2-20 15.2.7.1 Identification of Event and Causes 15.2-20 15.2.7.2 Sequence of Events and Systems Operation 15.2-20 15.2.7.3 Analysis of Effects and Consequences 15.2-20 15.2.7.4 Conclusions 15.2-21 15.2.8 FEEDWATER SYSTEM PIPE BREAKS 15.2-22 15.2.8.1 Parametric Analysis for FWLBs 15.2-24 15.2.8.2 Feedwater Line Break Event with Loss of Offsite Power 15.2-32 15.2.8.3 Feedwater Line Break Event With Offsite Power Available and Limiting Single Failure 15.2-48 June 2011 15-vi Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.2.8.4 Feedwater Line Break with LOP and Single Failure for Long Term Cooling 15.2-64 15.
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| ==2.9 REFERENCES==
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| 15.2-80 15.3 DECREASE IN REACTOR COOLANT FLOWRATE 15.3-1 15.3.1 TOTAL LOSS OF REACTOR COOLANT FLOW 15.3-1 15.3.1.1 Identification of Causes and Frequency Classification 15.3-1 15.3.1.2 Sequence of Events and Systems Operation 15.3-1 15.3.1.3 Core and System Performance 15.3-6 15.3.1.4 RCS Pressure Boundary Barrier Performance 15.3-10 15.3.1.5 Conclusions 15.3-12 15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING A FLOW COASTDOWN 15.3-12 15.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER 15.3-12 15.3.4 REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER 15.3-13 15.3.4.1 Identification of Causes and Frequency Condition 15.3-13 15.3.4.2 Sequence of Events and System Operation 15.3-14 15.3.4.3 Core and System Performance 15.3-19 15.3.4.4 RCS Pressure Boundary Barrier Performance 15.3-25 June 2011 15-vii Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.3.4.5 NSSS Response for Control Room Dose Consequences 15.3-30 15.3.4.6 EAB/LPZ Radiological Consequences/Containment Performance 15.3-34 15.3.4.7 Conclusions 15.3-43 15.
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| ==3.5 REFERENCES==
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| 15-3-44 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4-1 15.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL OR LOW (HOT ZERO) POWER CONDITION 15.4-1 15.4.1.1 Identification of Causes and Frequency Classification 15.4-1 15.4.1.2 Sequence of Events and System Operation 15.4-1 15.4.1.3 Core and System Performance 15.4-5 15.4.1.4 Reactor Coolant System Barrier Performance 15.4-10 15.4.1.5 Radiological Consequences and Containment Performance 15.4-12 15.4.1.6 Conclusions 15.4-12 15.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL AT POWER 15.4-12 15.4.2.1 Identification of Causes and Frequency Classification 15.4-12 15.4.2.2 Sequence of Events and System Operation 15.4-13 June 2009 15-viii Revision 15
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.4.2.3 Core and System Performance 15.4-17 15.4.2.4 Reactor Coolant System Barrier Performance 15.4-21 15.4.2.5 Radiological Consequences and Containment Performance 15.4-22 15.4.2.6 Conclusions 15.4-23 15.4.3 SINGLE FULL-STRENGTH CONTROL ELEMENT ASSEMBLY DROP 15.4-23 15.4.3.1 Identification of Causes and Frequency Classification 15.4-24 15.4.3.2 Sequence of Events and System Operation 15.4-25 15.4.3.3 Core and System Performance 15.4-25 15.4.3.4 Reactor Coolant System Barrier Performance 15.4-33 15.4.3.5 Radiological Consequences and Containment Performance 15.4-34 15.4.3.6 Conclusions 15.4-34 15.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP 15.4-35 15.4.4.1 Identification of Event and Causes 15.4-35 15.4.4.2 Sequence of Events and Systems Operation 15.4.35 15.4.4.3 Analysis of Effects and Consequences 15.4-36 15.4.4.4 Results 15.4-38 June 2011 15-ix Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.4.4.5 Conclusions 15.4-38 15.4.5 FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW 15.4-38 15.4.6 INADVERTENT DEBORATION 15.4-39 15.4.6.1 Identification of Event and Causes 15.4-39 15.4.6.2 Sequence of Events and Systems Operation 15.4-39 15.4.6.3 Analysis of Effects and Consequences 15.4-43 15.4.6.4 Results 15.4-49 15.4.6.5 Conclusions 15.4-49 15.4.7 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO THE IMPROPER POSITION 15.4-50 15.4.7.1 Identification of Events and Causes 15.4-50 15.4.7.2 Sequence of Events and System Operation 15.4-50 15.4.7.3 Analysis of Effects and Consequences 15.4-52 15.4.7.4 Conclusion 15.4-54 15.4.8 CONTROL ELEMENT ASSEMBLY EJECTION 15.4-54 15.4.8.1 Identification of Cause and Frequency Classification 15.4-54 15.4.8.2 Sequence of Events and Systems Operation 15.4-55 15.4.8.3 Core and System Performance 15.4-58 June 2011 15-x Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.4.8.4 Reactor Pressure Boundary Barrier Performance 15.4-64 15.4.8.5 Radiological Consequences and Containment Performance 15.4-70 15.4.8.6 Conclusions 15.4-93 15.
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| ==4.9 REFERENCES==
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| 15.4-94 15.5 INCREASE IN RCS INVENTORY 15.5-1 15.5.1 INADVERTENT OPERATION OF THE ECCS 15.5-1 15.5.1.1 Identification of Event and Causes 15.5-1 15.5.1.2 Sequence of Events and Systems Operation 15.5-1 15.5.1.3 Analysis of Effects and Consequences 15.5-1 15.5.1.4 Conclusion 15.5-2 15.5.2 CVCS MALFUNCTION-PRESSURIZER LEVEL CONTROL SYSTEM MALFUNCTION WITH LOSS OF OFFSITE POWER 15.5-2 15.5.2.1 Identification of Event and Causes 15.5-2 15.5.2.2 Sequence of Events and Systems Operation 15.5-3 15.5.2.3 Core and System Performance 15.5-11 15.5.2.4 Primary and Secondary Barrier Performances 15.5-11 15.5.2.5 Radiological Consequences and Containment Performance 15.5-15 15.5.2.6 Conclusion 15.5-16 15.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.6-1 June 2011 15-xi Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY/RELIEF VALVE 15.6-1 15.6.2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT 15.6-1 15.6.2.1 Identification of Causes and Frequency Classification 15.6-1 15.6.2.2 Sequence of Events and Systems Operation 15.6-2 15.6.2.3 Core and System Performance 15.6-5 15.6.2.4 RCS Pressure Boundary and Barrier Performance 15.6-8 15.6.2.5 Radiological Consequences and Containment Performance 15.6-9 15.6.2.6 Conclusions 15.6-12 15.6.3 STEAM GENERATOR TUBE RUPTURE 15.6-13 15.6.3.1 Steam Generator Tube Rupture Without a Loss of Offsite Power 15.6-14 15.6.3.2 Steam Generator Tube Rupture with a Loss of Offsite Power and a Single Failure 15.6-16 15.6.4 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR) 15.6-38 15.6.5 LOSS-OF-COOLANT ACCIDENTS 15.6-39 June 201 15-xii Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.6.5.1 Identification of Event and Causes -
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| Small Break LOCA 15.6-39 15.6.5.2 Sequence of Events and Systems Operation - Small Break LOCA 15.6-39 15.6.5.3 Analysis of Effects and Consequences 15.6-44 15.6.5.4 Identification of Event and Causes -
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| Large Break LOCA 15.6-54 15.6.5.5 Sequence of Events and Systems Operation - Large Break LOCA Dose Calculation 15.6-54 15.6.5.6 Analysis of Effects and Consequences - Large Break LOCA Dose Calculation 15.6-55 15.6.5.7 Conclusions 15.6-57 15.
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| ==6.6 REFERENCES==
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| 15.6-64 15.7 RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7-1 15.7.1 WASTE GAS SYSTEM FAILURE 15.7-1 15.7.1.1 Identification of Event and Causes 15.7-1 15.7.1.2 Sequence of Events and System Operation 15.7-1 15.7.1.3 Analysis of Effects and Consequences 15.7-1 15.7.1.4 Conclusions 15.7-4 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE (RELEASE TO ATMOSPHERE) 15.7-4 June 2011 15-xiii Revision 16
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page 15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING TANK FAILURES 15.7-5 15.7.3.1 Identification of Event and Causes 15.7-5 15.7.3.2 Sequence of Events and System Operation 15.7-5 15.7.3.3 Analysis of Effects and Consequences 15.7-6 15.7.3.4 Conclusions 15.7-6 15.7.4 RADIOLOGICAL CONSEQUENCES OF FUEL HANDLING ACCIDENTS 15.7-7 15.7.4.1 Fuel Handling Accident Outside Containment 15.7-7 15.7.4.2 Fuel Handling Accident Inside Containment 15.7-17 15.7.5 SPENT FUEL CASK DROP ACCIDENT 15.7-18 15.7.6 REFERENCE 15.7-20 APPENDIX 15A RESPONSES TO NRC REQUESTS FOR INFORMATION APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUESNCES OF ACCIDENTS APPENDIX 15C DELETED APPENDIX 15D ANALYSIS METHODS FOR LOSS OF PRIMARY COOLANT FLOW APPENDIX 15E LIMITING INFREQUENT EVENT June 2007 15-xiv Revision 14
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| PVNGS UPDATED FSAR TABLES Page 15.0-0 Single Failures 15.0-7 15.0-1 Chapter 15 Subsection Designation 15.0-10 15.1.3-1 SEQUENCE OF EVENTS FOR THE LIMITING MODERATE FREQUENCY STEAM BYPASS CONTROL SYSTEM MALFUNCTION SAFETY ANALYSIS 15.1-15 15.1.3-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE STEAM BYPASS CONTROL SYSTEM MALFUNCTION SAFETY ANALYSIS 15.1-24 15.1.4-1 SEQUENCE OF EVENTS FOR THE IOSGADVLOP SAFETY ANALYSIS 15.1-33 15.1.4-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE IOSGADVLOP SAFETY ANALYSIS 15.1-40 15.1.4-3 IODINE PARAMETERS AND INITIAL CONDITIONS FOR THE IOSGADV SAFETY ANALYSIS 15.1-41 15.1.4-4 RCS NOBLE GAS SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS 15.1-53 15.1.4-5 RCS NOBLE GAS SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS 15.1-54 15.1.4.6 OFFSITE RADIOLOGICAL DOSES FOR IOSGADVLOP SAFETY ANALYSIS 15.1.54 15.1.5-1 SEQUENCE OF EVENTS FOR THE LIMITING PRE-TRIP MSLB SAFETY ANALYSIS (SLB CASE) 15.1-60 15.1.5-2 SEQUENCE OF EVENTS FOR THE LIMITING POST-TRIP MSLB SAFETY ANALYSIS (SLBFPLOPCASE) 15.1-62 June 2011 15-xv Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.1.5-3 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING PRE-TRIP MAIN STEAM LINE BREAK (SLB CASE) SAFETY ANALYSES 15.1-71 15.1.5-4 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING POST-TRIP MAIN STEAM LINE BREAK (SLBFPLOP CASE) SAFETY ANALYSES 15.1-76 15.1.5-5 CORE PERFORMANCE SAFETY ANALYSIS RESULTS FOR THE LIMITING POST-TRIP MAIN STEAM LINE BREAK (SLBFPLOP CASE) SAFETY ANALYSES 15.1-84 15.1.5-6 RCS IODINE SOURCE TERM FOR THE MSLB OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSES 15.1-91 15.1.5-7 RCS NOBLE GAS SOURCE TERM FOR THE MSLB OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS 15.1-91 15.1.5-8 OFFSITE RADIOLOGICAL DOSE FOR MSLBs OUTSIDE THE CONTAINMENT BUILDING 15.1-95 15.1.6-1 SEQUENCE OF EVENTS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK WITH LOP SAFETY ANALYSIS (RCS TCOLD = 572°F) 15.1-101 15.1.6-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK SAFETY ANALYSES (RCS TCOLD = 572°F) 15.1-105 15.1.6-3 RESULTS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK SAFETY ANALYSES (RCS TCOLD = 572°F) 15.1-109 June 2011 15-xvi Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.2.3-1 Sequence of Events for the LOCV Primary Side Peak Pressure and Fuel Performance (DNBR) Event 15.2-9 15.2.3-2 Sequence of Events for the LOCV Secondary Side Peak Pressure Event 15.2-10 15.2.3-3 Assumed Initial Conditions for LOCV Primary Peak Pressure/DNBR and Secondary Side Peak Pressure Cases 15.2-13 15.2.8-1 Sequence of Events for the Feedwater Line Break Event with Loss of Offsite Power for Peak Pressure and Fuel Performance 15.2-38 15.2.8-2 Assumed Initial Conditions for the Feedwater Line Break Event with Loss of Offsite Power Peak Pressure and Fuel Performance Event 15.2-44 15.2.8-3 Sequence of Events for Feedwater Line Break with Limiting Single Failure and Offsite Power Available 15.2-54 15.2.8-4 Assumed Initial Conditions for Feedwater Line Break with Limiting Single Failure and Offsite Power Available 15.2-61 15.2.8-5 Sequence of Events for Feedwater Line Break with Loss of Offsite Power and Single Failure Event 15.2-70 June 2005 15-xvii Revision 13
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.2.8-6 Assumed Initial Conditions for Feedwater Line Break With Loss of Offsite Power and Single Failure Event 15.2-76 15.3.1.1 Typical Sequence of Events for Total Loss of Reactor Coolant Flow 15.3-5 15.3.1.2 Assumed Initial Conditions for the Total Loss of Reactor Coolant Flow 15.3-9 15.3.4-1 Assumed Initial Conditions for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip Core and System Performance 15.3-21 15.3.4-2 Sequence of Events for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power resulting from Turbine Trip for Core and System Performance 15.3-24 15.3.4-3 Assumed Initial Conditions for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip for Pressure Boundary Performance 15.3-26 15.3.4-4 Sequence of Events for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip for Pressure Boundary Performance 15.3-28 15.3.4-5 Assumed Initial Conditions for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip and a Stuck Open ADV 15.3-31 June 2011 15-xviii Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.3.4-6 Sequence of Events for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip with a Stuck Open ADV 15.3-32 15.3.4-7 Time of SIAS and Secondary Mass Releases for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power and Stuck Open ADV 15.3-34 15.3.4-8 Typical Parameters used in evaluating the Radiological Consequences of a Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power resulting from Turbine Trip 15.3-40 15.3.4-9 Typical Secondary System Mass Release to the Atmosphere for the Single Reactor Coolant Pump Shaft Break with Loss of Offsite Power Resulting from Turbine Trip 15.3-42 15.4.1-1 Sequence of Events for the Subcritical and Hot Zero Power Cases 15.4-4 June 2009 15-xix Revision 15
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.4.1-2 Input Parameters and Initial Conditions for the Limiting Moderate Frequency Uncontrolled CEA Withdrawal Analyses 15.4-8 15.4.2-1 Sequence of Events for the Sequential CEA Withdrawal Event at Full Power 15.4-16 15.4.2-2 Assumptions and Initial Conditions for the Sequential CEA Withdrawal Analysis 15.4-19 15.4.3-1 Sequence of Events for the Single Full-Strength CEA Drop Event 15.4-30 15.4.3-2 Typical Assumptions and Initial Conditions for the Single Full-Strength CEA Drop 15.4-31 15.4.6-1 Assumptions for the Mode 5 Inadvertent Deboration Analysis 15.4-47 15.4.6-2 Assumptions for the Mode 6 Inadvertent Deboration Analysis 15.4-49 15.4.8-1 Typical Sequence of Events for the CEA Ejection Event from Full Power Conditions (Fuel Enthalpy and Temperature Case) 15.4-57 15.4.8-2 Typical Assumptions used for the CEA Ejection Analysis Full Power Beginning of Cycle Initial Conditions (Fuel Enthalpy and Temperature Case) 15.4-60 June 2011 15-xx Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.4.8-3 Typical Initial Reactor States considered for the Typical CEA Ejection Event 15.4-61 15.4.8-4 Typical Assumptions used for the CEA Ejection Analysis for RCS Peak Pressure Event from Full Power Beginning of Cycle Initial Conditions 15.4-67 15.4.8-5 Typical Sequence of Events for CEA Ejection Peak RCS Pressure Event 15.4-69 15.4.8-6 Parameters Used in Evaluating the Radiological Consequences of CEA Ejection - Analyzed Core Power of 3954 MWt with Original Steam Generators 15.4-76 15.4.8-6A Parameters Used in Evaluating the Radiological consequences of a CEA Ejection - Analyzed Core Power of 4070 MWt with Replacement Steam Generators 15.4-81 15.4.8-7 Reactor Coolant Release to Containment and Containment Pressure and Temperature Versus Time - Analyzed Core Power of 3954 MWt with Original Steam Generators 15.4-88 15.4.8-7A Reactor Coolant Release to Containment and Containment Pressure and Temperature Versus Time - Analyzed Core Power of 4070 MWt with Replacement Steam Generators 15.4-89 15.4.8-8 Radiological Consequences of a Control Element Assembly Ejection Accident 15.4-92 June 2011 15-xxi Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.5.2-1 Deleted 15.5.2.2-1 Sequence of Events Power 15.5-10 15.5.2.4-1 Limiting Initial Conditions for PLCSM Peaking Primary and Secondary Pressure 15.5-13 15.5.2.4-2 Initial Condition with Significant Impact on Peak Primary and Secondary Pressure for PLCSM 15.5-14 15.5.2-2 Deleted 15.5.2-3 Deleted 15.6.2-1 Alarms that will be Actuated for the DBLLOCUS Event 15.6-2 15.6.2-2 Sequence of Events for the Double-Ended Break of a Letdown Line Outside Containment Upstream of the Letdown Control Valve 15.6-4 15.6.2-3 Assumed Input Parameters and Initial Conditions for the Double-Ended Break of a Letdown Line Outside Containment Upstream of the Letdown Control Valve 15.6-7 15.6.2-4 Radiological Consequences for the DBLLOCUS 15.6-12 15.3.3-1a Radiological Consequences for the SGTR Event 15.6-15 15.6.3-1 Sequence of Events for the Limiting SGTRLOP Single Failure Event (3990 MWt RTP with RSG) 15.6-25 June 2011 15-xxii Revision 16
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| PVNGS UPDATED FSAR Tables (Cont)
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| Page 15.6.3-2 Parameters used for the Limiting SGTRLOP Single Failure Event (3990 MWt RTP with RSG Case) 15.6-31 15.6.3-3 Radiological Consequences for the Limiting SGTRLOPSF Event 15.6-38 15.6.5-1 Radiological Consequences of a Small Break LOCA 15.6-46 15.6.5-2 Large Break LOCA Radiological Analysis Parameters and Results 15.6-58 15.7.1-1 Assumptions And Radiological Consequences Of Waste Gas System Failure 15.7-2 15.7.3-1 Radiological Consequences Of A Representative Liquid Storage Tank Failure 15.7-6 15.7.4-1 Parameters Used In Evaluating The Radiological Consequences Of A Fuel Handling Accident 15.7-13 15.7.4-2 Deleted 15.7.4-3 Radiological Consequences Of A Fuel Handling Accident Outside Containment 15.7-16 15.7.4-4 Deleted 15.7.4-5 Radiological Consequences Of A Fuel Handling Accident Inside Containment (Without Refueling Purge Isolation) 15.7-18 June 2011 15-xxiii Revision 16
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| PVNGS UPDATED FSAR FIGURES 15.0-1 Sequence of Events - Symbols, Acronyms, and Definitions 15.0-2 CEA Shutdown Worth vs. Position 15.1.3-1 SBCS MALFUNCTION EVENT STEAM FLOW vs. TIME 15.1.3-2 SBCS MALFUNCTION EVENT RCS TEMPERATURE vs. TIME 15.1.3-3 SBCS MALFUNCTION EVENT REACTIVITIES vs. TIME 15.1.3-4 SBCS MALFUNCTION EVENT CORE POWER vs. TIME 15.1.3-5 SBCS MALFUNCTION EVENT CORE AVERAGE HEAT FLUX vs. TIME 15.1.3-6 SBCS MALFUNCTION EVENT WIDE RANGE SG LEVEL vs. TIME 15.1.3-7 SBCS MALFUNCTION EVENT RCS PRESSURE vs. TIME 15.1.3-8 SBCS MALFUNCTION EVENT STEAM GENERATOR PRESSURE vs. TIME 15.1.3-9 SBCS MALFUNCTION EVENT STEAM GENERATOR LIQUID MASS vs. TIME 15.1.3-10 SBCS MALFUNCTION EVENT PRESSURIZER WATER VOLUME vs. TIME 15.1.3-11 SBCS MALFUNCTION EVENT DNBR vs. TIME 15.1.3-12 SBCS MALFUNCTION EVENT MAIN FEEDWATER FLOW vs. TIME 15.1.3-13 SBCS MALFUNCTION EVENT LONG-TERM RCS PRESSURE vs. TIME 15.1.3-14 SBCS MALFUNCTION EVENT LONG-TERM SG PRESSURE vs. TIME 15.1.3-15 SBCS MALFUNCTION EVENT LONG-TERM PRESSURIZER VOLUME vs. TIME 15.1.4-1 IOSGADVLOP EVENT STEAM FLOW vs. TIME June 2011 15-xxiv Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.1.4-2 IOSGADVLOP EVENT RCS HOT LEG TEMPERATURE vs. TIME 15.1.4-3 IOSGADVLOP EVENT RCS COLD LEG TEMPERATURE vs. TIME 15.1.4-4 IOSGADVLOP EVENT RCS AVERAGE TEMPERATURE vs. TIME 15.1.4-5 IOSGADVLOP EVENT REACTIVITIES vs. TIME 15.1.4-6 IOSGADVLOP EVENT CORE POWER vs. TIME 15.1.4-7 IOSGADVLOP EVENT CORE AVERAGE HEAT FLUX vs. TIME 15.1.4-8 IOSGADVLOP EVENT WIDE RANGE SG LEVEL vs. TIME 15.1.4-9 IOSGADVLOP EVENT RCS PRESSURE vs. TIME 15.1.4-10 IOSGADVLOP EVENT STEAM GENERATOR PRESSURE vs. TIME 15.1.4-11 IOSGADVLOP EVENT STEAM GENERATOR LIQUID vs. TIME 15.1.4-12 IOSGADVLOP EVENT PRESSURIZER WATER VOLUME vs. TIME 15.1.4-13 IOSGADVLOP EVENT DNBR vs. TIME 15.1.4-14 IOSGADVLOP EVENT MAIN FEEDWATER FLOW vs. TIME 15.1.5-1 POST-TRIP MSLB EVENT (SLBFPLOP CASE) STEAM FLOW vs. TIME 15.1.5-2 POST-TRIP MSLB EVENT (SLBFPLOP CASE) RCS TEMPERATURE vs. TIME 15.1.5-3 POST-TRIP MSLB EVENT (SLBFPLOP CASE) REACTIVITIES vs. TIME 15.1.5-4 POST-TRIP MSLB EVENT (SLBFPLOP CASE) CORE POWER vs. TIME 15.1.5-5 POST-TRIP MSLB EVENT (SLBFPLOP CASE) CORE AVERAGE HEAT FLUX vs. TIME 15.1.5-6 POST-TRIP MSLB EVENT (SLBFPLOP CASE) RCS PRESSURE vs. TIME June 2011 15-xxv Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.1.5-7 POST-TRIP MSLB EVENT (SLBFPLOP CASE) STEAM GENERATOR PRESSURE vs. TIME 15.1.5-8 POST-TRIP MSLB EVENT (SLBFPLOP CASE) STEAM GENERATOR LIQUID MASS vs. TIME 15.1.5-9 POST-TRIP MSLB EVENT (SLBFPLOP CASE) RCS FLOW RATE vs. TIME 15.1.5-10 POST-TRIP MSLB EVENT (SLBFPLOP CASE) PRESSURIZER WATER VOLUME vs. TIME 15.1.5-11 POST-TRIP MSLB EVENT (SLBFPLOP CASE) REACTOR VESSEL LIQUID LEVEL vs. TIME 15.1.5-12 POST-TRIP MSLB EVENT (SLBFPLOP CASE) SAFETY INJECTION FLOW RATE vs. TIME 15.1.5-13 POST-TRIP MSLB EVENT (SLBFPLOP CASE) MAIN FEEDWATER FLOW vs. TIME 15.1.5-14 POST-TRIP MSLB EVENT (SLBFPOP) CASE) AUXILIARY FEEDWATER FLOW vs. TIME 15.1.5-15 POST-TRIP MSLB EVENT (SLBFPLOP CASE) MACBETH DNBR vs.
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| TIME 15.1.6-1 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) STEAM FLOW vs. TIME 15.1.6-2 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) COLD LEG TEMPERATURES vs. TIME 15.1.6-3 SUBCRICICAL MSLB WITH LOP EVENT (TCOLD = 572°F) RCS TEMPERATURES vs. TIME 15.1.6-4 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F)
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| REACTIVITIES vs. TIME June 2011 15-xxvi Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.1.6-5 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) CORE POWER FRACTION vs. TIME 15.1.6-6 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) HEAT FLUX FRACTION vs. TIME 15.1.6-7 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) RCS PRESSURE vs. TIME 15.1.6-8 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) STEAM GENERATOR PRESSURE vs. TIME 15.1.6-9 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) STEAM GENERATOR LIQUID MASS vs. TIME 15.1.6-10 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F) RCS FLOW RATE vs. TIME 15.1.6-11 SUBCRITICAL MSLB WITH LOP EVENT (TCOLD = 572°F)
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| PRESSURIZER LIQUID VOLUME vs. TIME 15.1.6-12 SUBCRITICAL MSLB EVENTS SHUTDOWN MARGIN CURVES vs.
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| TIME 15.2.3-1 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| CORE POWER VS TIME 15.2.3-2 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| CORE HEAT FLUX VS TIME 15.2.3-3 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| CORE REACTIVITIES VS TIME 15.2.3-4 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| RCS TEMPERATURES VS TIME June 2011 15-xxvii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.3-5 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| RCS PRESSURE VS TIME 15.2.3-6 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| RCS PRESSURE VS TIME 15.2.3-7 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| PRESSURIZER PRESSURE VS TIME 15.2.3-8 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| PRESSURIZER PRESSURE VS TIME 15.2.3-9 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| PRESSURIZER WATER VOLUME VS TIME 15.2.3-10 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| SG PRESSURE VS TIME 15.2.3-11 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| SG LEVEL VS TIME 15.2.3-12 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| SG LIQUID INVENTORY VS TIME 15.2.3-13 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| INTEGRATED STEAM FLOW VS TIME 15.2.3-14 LOCV PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE -
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| DNBR VS TIME 15.2.3-15 LOCV SECONDARY PEAK PRESSURE CASE - CORE POWER VS TIME 15.2.3-16 LOCV SECONDARY PEAK PRESSURE CASE - CORE HEAT FLUX VS TIME 15.2.3-17 LOCV SECONDARY PEAK PRESSURE CASE - CORE REACTIVITIES VS TIME June 2011 15-xxviii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.3-18 LOCV SECONDARY PEAK PRESSURE CASE - RCS TEMPERATURES VS TIME 15.2.3-19 LOCV SECONDARY PEAK PRESSURE CASE - RCS PRESSURE VS TIME 15.2.3-20 LOCV SECONDARY PEAK PRESSURE CASE - PRESSURIZER PRESSURE VS TIME 15.2.3-21 LOCV SECONDARY PEAK PRESSURE CASE - PRESSURIZER WATER VOLUME VS TIME 15.2.3-22 LOCV SECONDARY PEAK PRESSURE CASE - SG PRESSURE VS TIME 15.2.3-23 LOCV SECONDARY PEAK PRESSURE CASE - SG PRESSURE VS TIME 15.2.3-24 LOCV SECONDARY PEAK PRESSURE CASE - SG WATER LEVEL VS TIME 15.2.3-25 LOCV SECONDARY PEAK PRESSURE CASE - SG WATER LEVEL VS TIME 15.2.3-26 LOCV SECONDARY PEAK PRESSURE CASE - SG INVENTORY VS TIME 15.2.3-27 LOCV SECONDARY PEAK PRESSURE CASE - SG INVENTORY VS TIME 15.2.3-28 LOCV SECONDARY PEAK PRESSURE CASE - INTEGRATED STEAM FLOW VS TIME 15.2.8-1 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL RCS FLOW ON RCS PEAK PRESSURE 15.2.8-2 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL RCS FLOW ON PRESSURIZER LEVEL June 2011 15-xxix Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.8-3 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL PRESSURIZER LEVEL ON RCS PEAK PRESSURE 15.2.8-4 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL PRESSURIZER LEVEL ON MAXIMUM PRESSURIZER LEVEL 15.2.8-5 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL GAS GAP CONDUCTANCE ON RCS PRESSURE 15.2.8-6 FEEDWATER LINE BREAK EVENT - EFFECT OF INITIAL GAS GAP CONDUCTANCE ON PRESSURIZER LEVEL 15.2.8-7 FEEDWATER LINE BREAK EVENT - EFFECT OF MODERATOR TEMPERATURE COEFFICIENT ON PEAK RCS PRESSURE 15.2.8-8 FEEDWATER LINE BREAK EVENT - EFFECT OF MODERATOR TEMPERATURE COEFFICIENT ON PRESSURIZER LEVEL 15.2.8-9 FEEDWATER LINE BREAK EVENT - EFFECT OF PSV TOLERANCE ON RCS PRESSURE 15.2.8-10 FEEDWATER LINE BREAK EVENT - EFFECT OF PSV TOLERANCE ON PRESSURIZER LEVEL 15.2.8-11 FEEDWATER LINE BREAK EVENT - EFFECT OF MSSV TOLERANCE ON PRESSURIZER LEVEL 15.2.8-12 FEEDWATER LINE BREAK EVENT - EFFECTS OF INITIAL PRESSURIZER PRESSURE, INITIAL CORE INLET TEMPERATURE AND BREAK SIZE ON PEAK RCS PRESSURE FOR M=0 15.2.8-13 FEEDWATER LINE BREAK EVENT - EFFECTS OF INITIAL PRESSURIZER PRESSURE, INITIAL CORE INLET TEMPERATURE AND BREAK SIZE ON PRESSURIZER LEVEL FOR M=0 June 2011 15-xxx Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.8-14 FEEDWATER LINE BREAK EVENT - EFFECTS OF INITIAL PRESSURIZER PRESSURE, INITIAL CORE INLET TEMPERATURE AND BREAK SIZE ON PEAK RCS PRESSURE FOR M=30,000 LBM 15.2.8-15 FEEDWATER LINE BREAK EVENT - EFFECTS OF INITIAL PRESSURIZER PRESSURE, INITIAL CORE INLET TEMPERATURE AND BREAK SIZE ON PEAK RCS PRESSURE 2
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| FOR BREAK SIZES LESS THAN 0.2FT AND M=30,000 LBM 15.2.8-16 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - CORE POWER VS TIME 15.2.8-17 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - CORE HEAT FLUX VS TIME 15.2.8-18 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - REACTIVITY VS TIME 15.2.8-19 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - RCS PRESSURE VS TIME 15.2.8-20 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - PRESSURIZER PRESSURE VS TIME 15.2.8-21 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - PRESSURIZER WATER VOLUME VS TIME 15.2.8-22 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - SG PRESSURE VS TIME 15.2.8-23 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - SG WATER LEVELS VS TIME June 2011 15-xxxi Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.8-24 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - SG LIQUID INVENTORY VS TIME 15.2.8-25 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - RCS LOOP FLOW VS TIME 15.2.8-26 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - SG STEAM FLOW VS TIME 15.2.8-27 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - BREAK FLOW VS TIME 15.2.8-28 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - BREAK ENTHALPY VS TIME 15.2.8-29 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - PSV FLOW VS TIME 15.2.8-30 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - SURGE FLOW VS TIME 15.2.8-31 FWLB WITH LOP - PRIMARY PEAK PRESSURE/FUEL PERFORMANCE CASE - DNBR VS TIME 15.2.8-32 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - CORE POWER VS TIME 15.2.8-33 FLWB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - CORE HEAT FLUX VS TIME 15.2.8-34 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - REACTIVITIES VS TIME 15.2.8-35 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - RCS TEMPERATURE VS TIME 15.2.8-36 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - PRESSURIZER PRESSURE VS TIME June 2011 15-xxxii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.2.8-37 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - PRESSURIZER WATER VOLUME VS TIME 15.2.8-38 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - SG PRESSURE VS TIME 15.2.8-39 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - SG WATER LEVELS VS TIME 15.2.8-40 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - SG INVENTORIES VS TIME 15.2.8-41 FWLB WITHOUT LOP PLUS SF - PRIMARY PEAK PRESSURE CASE - RCS LOOP FLOW VS TIME 15.2.8-42 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE CORE POWER VS TIME 15.2.8-43 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE UNAFFECTED LOOP RCS TEMPERATURE VS TIME 15.2.8-44 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE RCS PRESSURE VS TIME 15.2.8-45 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE PRESSURIZER PRESSURE VS TIME 15.2.8-46 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE PRESSURIZER WATER VOLUME VS TIME 15.2.8-47 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE SG PRESSURE VS TIME 15.2.8-48 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE SG LIQUID INVENTORIES VS TIME 15.2.8-49 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE AFFECTED SG AFW FLOW VS TIME 15.2.8-51 FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE PSV FLOW VS TIME June 2011 15-xxxiii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.3.1-1 TOTAL LOSS OF FORCED COOLANT FLOW - CORE POWER VS TIME 15.3.1-2 TOTAL LOSS OF FORCED COOLANT FLOW - CORE AVERAGE HEAT FLUX VS TIME 15.3.1-3 TOTAL LOSS OF FORCED COOLANT FLOW - PRESSURIZER PRESSURE VS TIME 15.3.1-4 TOTAL LOSS OF FORCED COOLANT FLOW - RCS TEMPERATURES VS TIME 15.3.1-5 TOTAL LOSS OF FORCED COOLANT FLOW - REACTIVITIES VS TIME 15.3.1-6 TOTAL LOSS OF FORCED COOLANT FLOW - CORE FLOW FRACTION VS TIME 15.3.1-7 TOTAL LOSS OF FORCED COOLANT FLOW - STEAM GENERATOR PRESSURE VS TIME 15.3.1-8 TOTAL LOSS OF FORCED COOLANT FLOW - CORE POWER vs.
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| TIME (0-30 sec.)
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| 15.3.1-9 TOTAL LOSS OF FORCED COOLANT FLOW PRESSURIZER PRESSURE vs. TIME (0-30 sec.)
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| 15.3.1-10 TOTAL LOSS OF FORCED COOLANT FLOW RCS FLOW FRACTION vs. TIME (0-30 sec.)
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| 15.3.1-11 TOTAL LOSS OF FORCED COOLANT FLOW REACTIVITIES vs. TIME(0-30 sec.)
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| 15.3.1-12 TOTAL LOSS OF FORCED COOLANT FLOW SG PRESSURE vs. TIME (0-30 sec.)
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| 15.3.1-13 TOTAL LOSS OF FORCED COOLANT FLOW MINIMUM DNBR vs. TIME 15.3.1-14 TOTAL LOSS OF FORCED COOLANT FLOW REACTIVITY vs. TIME (0-4 sec.)
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| June 2011 15-xxxiv Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.3.4-1 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - CORE POWER VS TIME 15.3.4-2 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - HEAT FLUX VS TIME 15.3.4-3 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - PRESSURIZER PRESSURE VS TIME 15.3.4-4 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - RCS PRESSURE VS TIME 15.3.4-5 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - CORE AVERAGE TEMPERATURE VS TIME 15.3.4-6 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - RCS TEMPERATURE VS TIME 15.3.4-7 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - REACTIVITY VS TIME 15.3.4-8 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - CORE FLOW VS TIME 15.3.4-9 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - STEAM GENERATOR PRESSURE VS TIME 15.3.4-10 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - STEAM GENERATOR PRESSURE VS TIME 15.3.4-11 SINGLE RCP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A TURBINE TRIP - AUXILIARY FEEDWATER FLOW VS TIME June 2011 15-xxxv Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.3.4.12 SINGLE RCP SHAFT BREAK WITH LOSS OF OFF-SITE POWER RESULTING FROM TURBINE TRIP - DNBR VS TIME 15.4.1-1 Uncontrolled Subcritical CEA Withdrawal Core Power vs. Time 15.4.1-2 Uncontrolled Subcritical CEA Withdrawal Core Heat Flux vs. Time 15.4.1-3 Uncontrolled Subcritical CEA Withdrawal RCS Pressure vs. Time 15.4.1-4 Uncontrolled Subcritical CEA Withdrawal Total Reactivity vs. Time 15.4.1-5 Uncontrolled Subcritical CEA Withdrawal Doppler Reactivity vs. Time 15.4.1-6 Uncontrolled Subcritical CEA Withdrawal RCS Temperature vs. Time 15.4.1-7 Uncontrolled Subcritical CEA Withdrawal CETOP DNBR vs. Time 15.4.1-8 Subcritical CEA Withdrawal Subcritical Energy Deposition vs. Time 15.4.1-9 HZP CEA Withdrawal Core Power vs. Time 15.4.1-10 HZP CEA Withdrawal Core Average Heat Flux vs. Time 15.4.1-11 HZP CEA Withdrawal RCS Pressure vs. Time 15.4.1-12 HZP CEA Withdrawal Total Reactivity vs. Time 15.4.1-13 HZP CEA Withdrawal Doppler Reactivity vs. Time 15.4.1-14 HZP CEA Withdrawal RCS Temperature vs. Time 15.4.1-15 HZP CEA Withdrawal CETOP DNBR vs. Time 15.4.1-16 HZP CEA Withdrawal HZP Energy Deposition vs. Time June 2011 15-xxxvi Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.4.2-1 Uncontrolled CEA Withdrawal at Power Core Power vs.
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| Time 15.4.2-2 Uncontrolled CEA Withdrawal at Power Core Heat Flux vs. Time 15.4.2-3 Uncontrolled CEA Withdrawal at Power RCS Pressure vs. Time 15.4.2-4 Uncontrolled CEA Withdrawal at Power DNBR vs. Time 15.4.2-5 Uncontrolled CEA Withdrawal at Power RCS Temperature vs. Time 15.4.2-6 Uncontrolled CEA Withdrawal at Power Steam Generator Pressure vs. Time 15.4.2-7 Uncontrolled CEA Withdrawal at Power Peak Linear Heat Generation vs. Time 15.4.2-8 Uncontrolled CEA Withdrawal at Power Feedwater Enthalpy vs. Time 15.4.2-9 Uncontrolled CEA Withdrawal at Power Feedwater Flow vs. Time 15.4.2.10 Uncontrolled CEA Withdrawal at Power MSSV Flow vs.
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| Time 15.4.2-11 Uncontrolled CEA Withdrawal at Power Total Steam Flow vs. Time 15.4.7-1 Planar Average Power Distribution Corresponding to Maximum FRN Produced by a Fuel Assembly Misloading that is Undetectable During Startup at BOC 15.4.8-1 CEA Ejection Core Power vs. Time 15.4.8-2 CEA Ejection Peak Power Density vs. Time 15.4.8-3 CEA Ejection Core Average Heat Flux vs. Time June 2011 15-xxxvii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.4.8-4 CEA Ejection Peak Hot Channel Heat Flux vs. Time 15.4.8-5 CEA Ejection Hot and Average Channel Fuel and Clad Temperature vs. Time 15.4.8-6 CEA Ejection Reactivity vs. Time 15.4.8-7 CEA Ejection RCS Pressure vs. Time 15.4.8-8 CEA Ejection Pressurizer Pressure vs. Time 15.4.8-9 CEA Ejection Pressurizer Pressure vs. Time 15.4.8-10 CEA Ejection Steam Generator Pressure vs. Time 15.4.8-11 CEA Ejection Steam Generator Pressure vs. Time 15.4.8-12 CEA Ejection MSSV Flow vs. Time 15.5.2-1 DELETED 15.5.2-2 PLCS Malfunction Core Power vs. Time 15.5.2-3 PLCS Malfunction Core Average Heat Flux vs. Time 15.5.2-4 PLCS Malfunction Pressurizer Pressure vs. Time 15.5.2-5 PLCS Malfunction Core Average Coolant Temperatures vs. Time 15.5.2-6 PLCS Malfunction Pressurizer Water Volume vs. Time 15.5.2-7 PLCS Malfunction Steam Generator Water Level vs. Time 15.5.2-8 PLCS Malfunction Steam Generator Pressure vs. Time 15.5.2-9 PLCS Malfunction Total Steam Flow vs. Time 15.5.2-10 PLCS Malfunction Feedwater Flow vs. Time 15.5.2-11 PLCS Malfunction Feedwater Enthalpy vs. Time 15.6.2-1 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Core Power Fraction vs. Time June 2011 15-xxxviii Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.6.2-2 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Core Average Heat Flux vs. Time 15.6.2-3 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Pressurizer Pressure vs. Time 15.6.2-4 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Core Coolant Temperature vs. Time 15.6.2-5 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Steam Generator Pressure vs. Time 15.6.2-6 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Integrated Primary Coolant Discharge vs. Time 15.6.2-7 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Pressurizer Water Level vs. Time 15.6.2-8 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Reactor Coolant System Inventory vs. Time 15.6.2-9 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Steam Generator Water Level vs. Time 15.6.2-10 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Total Steam Flow vs.
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| Time June 2011 15-xxxix Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.6.2-12 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Feedwater Enthalpy vs. Time 15.6.2-13 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Minimum DNBR vs.
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| Time 15.6.2-14 Letdown Line Break, Outside Containment, Upstream of Letdown Line Control Valve Minimum DNBR vs.
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| Time 15.6.3-1 SGTRLOP with Single Failure Event Core Power vs.
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| Time 15.6.3-2 SGTRLOP with Single Failure Event RCS Pressure vs.
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| Time 15.6.3-3 SGTRLOP with Single Failure Event RCS Temperatures Affected Loop vs. Time 15.6.3-4 SGTRLOP with Single Failure Event Upper Head Temperature vs. Time 15.6.3-5 SGTRLOP with Single Failure Event Pressurizer Liquid Volume vs. Time 15.6.3-6 SGTRLOP with Single Failure Event Upper Head Level vs. Time 15.6.3-7 SGTRLOP with Single Failure Event RCS Total Mass vs. Time 15.6.3-8 SGTRLOP with Single Failure Event SG Pressure vs.
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| Time 15.6.3-9 SGTRLOP with Single Failure Event AFW Integrated Flow vs. Time June 2011 15-xl Revision 16
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| PVNGS UPDATED FSAR FIGURES (cont) 15.6.3-10 SGTRLOP with Single Failure Event Tube Leak Rate vs. Time 15.6.3-11 SGTRLOP with Single Failure Event Integrated Tube Leak Flow vs. Time 15.6.3-12 SGTRLOP with Single Failure Event Ruptured Tube Leak Flashing Fraction vs. Time 15.6.3-13 SGTRLOP with Single Failure Event SG Liquid Inventory vs. Time 15.6.3-14 SGTRLOP with Single Failure Event Integrated ADV Flow vs. Time 15.6.3-15 SGTRLOP with Single Failure Event Subcooled Margin vs. Time 15.7.1-1 Sequence of Events Diagram for a Waste Gas Decay Tank Rupture 15.7.3-1 Sequence of Events Diagram for a Radioactive Liquid Tank Rupture 15.7.4-1 Sequence of Events Diagram for a Fuel Handling Accident Outside Containment 15.7.4-2 Sequence of Events Diagram for a Fuel Handling Accident Inside Containment June 2011 15-xli Revision 16
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| This page intentionally blank PVNGS UPDATED FSAR
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| : 15. ACCIDENT ANALYSES
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| | |
| ==15.0 INTRODUCTION==
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| Nuclear power plant safety is evaluated by analyzing the response of the plant to postulated disturbances in process variables, and to postulated malfunctions or failures of equipment. Such analyses provide a significant contribution to the selection of Technical Specification Limiting Conditions for Operation (LCOs), Limiting Safety System Settings (LSSSs),
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| and design specifications for components and systems from the standpoint of public health and safety. Such analyses are also a focal point of the Nuclear Regulatory Commissions (NRC)
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| Operating License reviews.
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| In this chapter, the effects of anticipated process disturbances and postulated component failures are examined to determine their consequences, to evaluate the capability built into the plant to control or accommodate such failures and situations, and to identify any limitations of expected performance. In other words, the Chapter 15 safety analyses are performed to show that, given certain design basis requirements and specifications for Systems, Structures, and Components (SSCs), overall plant response and performance will be acceptable should a Design Basis Event (DBE) occur.
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| The events analyzed herein include Anticipated Operational Occurrences (AOOs), off-design transients that may induce fuel failures above those expected from normal operational occurrences, and postulated accidents of low probability.
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| These analyses include an assessment of the radiological consequences of assumed fission product releases, up to and including the greatest potential hazard from any accident considered credible.
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| June 2005 15.0-1 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.0.1 CLASSIFICATION OF TRANSIENTS AND ACCIDENTS 15.0.1.1 Format and Content The format and content of this UFSAR chapter is structured in accordance with the guidance contained in Chapter 15 of NRC Regulatory Guide 1.70, Revision 3 [Reference 1]. Acceptance criteria for the safety analyses are derived, on a case-by-case basis, from NUREG-75/087 [Reference 2], NUREG-0800 [Reference 3], and/or licensing agreements negotiated with NRC staff, as documented in licensee correspondence and NRC safety evaluations associated with the PVNGS Operating License dockets. Chapter contents are maintained in accordance with 10 CFR 50.71(e) [Reference 4], NRC Regulatory Guide 1.181, Revision 0 [Reference 5], and Nuclear Energy Institute (NEI)
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| Publication 98-03, Revision 1 [Reference 6], as described in UFSAR Section 1.8.
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| 15.0.1.2 Event Categories Each postulated initiating event has been assigned to one of the following categories:
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| * Increased heat removal by the secondary system
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| * Decreased heat removal by the secondary system
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| * Decreased reactor coolant flow
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| * Reactivity and power distribution anomalies
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| * Increase in Reactor Coolant System (RCS) inventory
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| * Decrease in RCS inventory
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| * Radioactive release from a subsystem or component June 2005 15.0-2 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY The assignment of an initiating event to one of these categories is made according to Reference 1, 2, and 3.
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| Although Reference 1 recommends that this chapter include safety analyses for Anticipated Transients Without Scram (ATWS), such analyses are not presented herein. This deviation from regulatory guidance is justified because, following publication of Reference 1, the NRC staff and Nuclear Steam Supply System (NSSS) vendors did not reach final agreement on safety analysis methodologies and acceptance criteria for ATWS events, as documented in Reference 7. In lieu of such an agreement, the NRC promulgated 10 CFR 50.62 [Reference 8],
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| which mandated the installation of diverse plant systems to reduce the risks associated with an ATWS event. PVNGS compliance with the ATWS rule is documented in an NRC safety evaluation report dated October 18, 1990 [Reference 9], and is based on the installation of a Supplementary Protection System (DAFAS), as described in UFSAR Sections 7.2.5 and 7.3.5, respectively.
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| 15.0.1.3 Event Frequencies As noted in UFSAR Chapter 3, PVNGS SSCs are classified according to their importance in preventing and mitigating postulated events, using the classification system described in ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants [Reference 10].
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| ANSI N18.2-1973 divides postulated transients and accidents into four broad categories for design purposes, based on their relative estimated frequency of occurrence. The events analyzed in this chapter are likewise classified into these June 2005 15.0-3 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY four broad categories, or conditions for design, which are as follows:
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| * Normal Operations (Condition I). Condition I occurrences are defined as operations that are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant.
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| * Incidents of Moderate Frequency (Condition II). Condition II occurrences are defined as incidents, any one of which may occur during a calendar year for a particular plant.
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| * Infrequent Events (Condition III). Condition III occurrences are defined as incidents, any one of which may occur during the lifetime of a particular plant.
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| * Limiting Faults (Condition IV). Condition IV occurrences are defined as faults that are not expected to occur, but are postulated nonetheless because their consequences include the potential for the release of significant amounts of radioactive material.
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| 15.0.1.4 Events and Event Combinations The events and event combinations in this chapter are presented with respect to the event-specific acceptance criteria. For each applicable acceptance criterion in an event category, only the limiting event or event combination is presented in analytical detail. As required by Reference 1, qualitative discussions are provided for all other events or event combinations explaining why they are not limiting.
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| For event combinations that require consideration of a single failure, the limiting failures for the NSSS transient and June 2007 15.0-4 Revision 14
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY accident safety analyses in this chapter are selected from those listed in Table 15.0-0. Only low probability, dependent failures (e.g., loss of off-site power following turbine trip) and independent pre-existing failures are considered credible and included in the table. Pre-existing failures are equipment failures existing prior to the event initiation that are not revealed until called upon during the event (e.g., a failure of an auxiliary feedwater pump). High probability, dependent occurrences are always included in the event analysis, if they have an adverse impact (e.g., loss of main feedwater pumps following a loss of electric power).
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| Table 15.0-0 lists a "Loss of Offsite power (LOP) following turbine trip" as a Single Failure to be considered in Safety Analysis. However, the LOP is treated differently depending on the event combination under consideration. The combination of the initiating event with coincident occurrences and single failures changes the event classification and acceptance criteria. For moderate frequency (Condition II) events, the LOP may be treated as the limiting single failure. In that case, the event becomes an infrequent (Condition III) event, with different acceptance criteria. One example is the Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (UFSAR Section 15.1.4). This event by itself is a moderate frequency event. However, once a LOP is taken as the limiting single failure, the event becomes an infrequent event with different acceptance criteria. For Condition IV limiting fault events, such as Loss of Coolant Accidents, Main Steam Line Break, Feedwater Line Break and Steam Generator Tube Rupture, a LOP is treated as a coincident occurrence, along with the limiting fault, and an additional single failure is then postulated. This application of the LOP to safety analysis June 2005 15.0-5 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY ensures that the supporting SSCs can perform their design functions with offsite power unavailable, as required by the General Design Criteria.
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| Analytical assumptions regarding the availability and operation of plant SSCs (e.g., pressurizer heaters and sprays) are described in each event section on a case-by-case basis.
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| Additionally, each safety analysis section describes the analytical credit that has been taken, if any, for administrative controls and procedures, manual equipment operation, or plant operator actions to mitigate an event.
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| June 2005 15.0-6 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY Table 15.0-0 SINGLE FAILURES (Sheet 1 of 3)
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| STEAM BYPASS CONTROL SYSTEM (SBCS)
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| : 1. Failure to modulate open
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| : 2. Failure to quick open
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| : 3. One bypass valve fails to quick close
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| : 4. Excessive steam bypass flow
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| : 5. Failure to generate automatic withdrawal prohibit signal during steam bypass operation
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| : 6. Failure to generate the reactor power cutback signal REACTIVITY CONTROL SYSTEMS
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| : 7. Regulating group(s) fail(s) to insert or withdraw (a)
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| : 8. A single Control Element Assembly (CEA) stuck (a)
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| : 9. A CEA subgroup stuck
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| : 10. Failure to initiate or execute the reactor power cutback
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| : 11. CEAs withdraw upon automatic withdrawal prohibit and/or CEA withdrawal prohibit FEEDWATER CONTROL SYSTEM
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| : 12. Failure of reactor trip override
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| : 13. Failure of high level override
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| : a. Control element drive mechanism does not respond to control signal. Release of CEA(s) on trip is not inhibited.
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| June 2005 15.0-7 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY Table 15.0-0 SINGLE FAILURES (Sheet 2 of 3)
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| TURBINE-GENERATOR CONTROL SYSTEM
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| : 14. Setback without cutback
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| : 15. Failure to modulate the turbine control valves
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| : 16. Failure to setback given a cutback (100% > initial power > 75%)
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| : 17. Failure to setback (75% > initial power > 60%)
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| : 18. Failure to runback (60% > initial power)
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| : 19. Failure to trip the turbine PRESSURIZER PRESSURE CONTROL SYSTEM (PPCS)
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| : 20. Failure of spray control valves to open
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| : 21. Failure of spray control valves to close
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| : 22. Failure of backup heaters to turn on
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| : 23. Failure of backup heaters to turn off PRESSURIZER LEVEL CONTROL SYSTEM (PLCS)
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| : 24. Backup charging pump fails to turn on
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| : 25. Backup charging pump fails to turn off
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| : 26. Letdown flow control valve fails to close
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| : 27. Letdown flow control valve fails to open MAIN FEEDWATER SYSTEM
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| : 28. One Main Feedwater Isolation Valve (MFIV) fails to close
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| : 29. One backflow check valve fails to close June 2005 15.0-8 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY Table 15.0-0 SINGLE FAILURES (Sheet 3 of 3)
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| MAIN STEAM SYSTEM
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| : 30. One Main Steam Isolation Valve (MSIV) fails to close
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| : 31. One Atmospheric Dump Valve (ADV) fails to open
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| : 32. One Main Steam Safety Valve (MSSV) fails to reclose AUXILIARY FEEDWATER SYSTEM
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| : 33. Failure of any one auxiliary feed pump to start EMERGENCY CORE COOLING SYSTEM (ECCS)
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| : 34. Failure of one High Pressure Safety Injection (HPSI) or Low Pressure Safety Injection (LPSI) pump ELECTRICAL POWER SOURCES Loss of offsite power following turbine trip (b) 35.
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| : 36. Failure to achieve fast transfer of a non-Class 1E bus to the startup transformer
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| : 37. Failure of one emergency generator to start, run or load
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| : b. Section 15.0.2.4 describes the loss of off-site power following a turbine trip in more detail, including the time delay between turbine stop valve closure and loss of offsite power.
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| June 2005 15.0-9 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.0.1.5 Section Numbering The safety analyses in this chapter are divided into sections and subsections as described in Table 15.0-1.
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| Table 15.0-1 CHAPTER 15 SUBSECTION DESIGNATION Each event of event combination section number begins with the sequence 15.W.X where:
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| 15 = 15 Safety analyses that are presented in UFSAR Chapter 15 W = 1 Increase in heat removal by the secondary system 2 Decrease in heat removal by the secondary system 3 Decrease in RCS flow rate 4 Reactivity and power distribution anomalies 5 Increase in RCS inventory 6 Decrease in RCS inventory 7 Radioactive release from a subsystem or component X = 1, 2, etc. Event title June 2005 15.0-10 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.0.1.6 Sequence of Events Analysis The Sequence of Events Analysis (SEA) has been performed for each limiting event and event combination, for which detailed safety analysis results are presented in this chapter. The purpose of the SEA is to determine the following:
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| A. The step-by-step sequence of events from event initiation to the final stabilized condition; B. The extent to which normally operating plant instrumentation and controls are assumed to function; C. The extent to which plant and reactor protective systems are required to function; D. The credit taken for the functioning of normally operating plant systems; and E. The operation of engineered safety systems that are required.
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| SEAs have been specifically omitted for those events that, though representing limiting events for their category, do not result in the actuation of safety systems, or for which a detailed, quantitative analysis was not presented. For the safety analyses that appear in this chapter, the primary results of each SEA are presented in a sequence of events table and described in the text. Each sequence of events table presents a chronology of events that may be anticipated to occur during a transient, from event initiation to a final stabilized condition (or until operator action is taken to place the reactor in a safe shutdown condition). The accompanying text provides additional clarification, including information regarding systems operation and a discussion of the effects of postulated single failures. The sequence of events June 2011 15.0-11 Revision 16
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY table and corresponding text fulfill the regulatory guidance contained in Reference 1, by providing a step-by-step chronology and detailed discussion of systems operation for limiting transients.
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| The results of some SEAs are also presented in an optional format for the safety analyses, consisting of two additional tables and a figure. The first of the two optional tables is a matrix that identifies the extent to which normally operating plant systems are assumed to function during a transient. The second table specifies the Reactor Protective System (RPS) and Engineered Safety Features (ESF) that are actuated to accomplish safety functions during the course of the event.
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| The optional figure is a Sequence of Events Diagram (SED), or a simple block diagram that provides a systematic analysis of components that are required to function during a transient.
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| For some of the safety analyses, pertinent information that would otherwise appear on these tables of figure is instead described in the text.
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| The SEDs, together with the chronological list of events and the SEA symbol and acronym drawing (Figures 15.0-1), may be used to trace the actuation and interaction of the systems used to mitigate the consequences of each event. The SED is a block diagram, composed of several success paths that define a set of safety actions leading from the initiating event to the accomplishment of a specific safety function. All of the safety functions used in the SEDs are defined in Figure 15.0-1.
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| A success path may be composed of two branches, one indicated by a solid line, describing the sequence of events that occur in the transient analysis, and the other, indicated by a dotted line, describing an alternative or back-up path to a given June 2009 15.0-12 Revision 15
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY means of accomplishing a safety function. An alternate dotted path is specified if the analysis assumed the action of a non-safety system in achieving a particular safety function. Non-safety systems are indicated by an "NS" in the upper right-hand corner of the system block.
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| The redundancy of a system or component is indicated by a fraction (e.g., 1/2, 2/4) placed beneath the system block. The numerator specifies the number of trains or components required to perform the action, and the denominator specifies the number of trains or components normally available. In cases where no alternate path exists and a single system or component is included in a success path, the symbol "S.F." will be used to indicate that no single active failure will prevent the accomplishment of the safety action.
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| Components or systems that require no active initiation or actuation to perform their function are considered to be passive and are marked as such with a "P" in the lower left-hand corner of the system block. The absence of a passive label implies that a component is considered to be active and must be actively initiated to perform its function.
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| Manual operations performed on a given system or component are indicated by placing an "M" in the lower left-hand corner of the system block. When a manual action is required, the sensed variables necessary to perform the action are shown as inputs and the location of the input signal is shown above the input signal circle.
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| The system setpoint values assumed in the transient analysis, e.g., trip signal setpoints, is noted along the success path.
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| Time delays or the time required to perform an action are shown as a number with square brackets.
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| June 2005 15.0-13 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY All events presented in sequence of events diagrams in the main body of this chapter are shown from event initiation to achievement of the cold shutdown operating mode. Not all events require that the plant be taken to cold shutdown. The SEDs only demonstrate that for any event presented here it is possible to take the plant to cold shutdown by means of the safety actions indicated.
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| 15.0.2 SYSTEMS OPERATION During the course of any event various systems may be called upon to function. Some of these systems are described in Chapter 7 and include those electrical, instrumentation, and control systems designed to perform a safety function (i.e.,
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| those systems that must operate during an event to mitigate the consequences) and those systems not required to perform a safety function (see UFSAR Sections 7.2 through 7.6 and 7.7, respectively).
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| 15.0.2.1 Reactor Protection The Reactor Protective System (RPS) is described in UFSAR Section 7.2.
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| RPS trips credited in the safety analyses, including Core Protection Calculator (CPC) trips, are identified in the sequence of events description for each limiting event and event combination. Analytical trip setpoints are chosen to be consistent with, or conservative with respect to, the RPS trip setpoint allowable values delineated in the PVNGS Technical Specifications and the RPS trip setpoints delineated in UFSAR Section 7.2. Where applicable, analytical trip setpoints are adjusted to account for instrumentation loop uncertainties June 2009 15.0-14 Revision 15
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY derived from design control calculations. The safety analyses also take into consideration the RPS response times associated with the various trip functions.
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| The RPS response time is the sum of the sensor response time and the reactor trip delay time. The sensor response time is defined as the time from when the value of the monitored parameter at the sensor equals or exceeds the RPS trip setpoint, until the sensor output equals or exceeds the trip setpoint. The sensor response is modeled by using a transfer function for the particular sensor used. The reactor trip delay time is defined as the elapsed time from the time the sensor output equals or exceeds the trip setpoint to the time the reactor trip breakers are fully open.
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| As noted in UFSAR Section 3.9.4, the Control Element Assemblies (CEAs) are designed with a maximum drop time of 4.0 seconds, where the drop time is defined as the interval between the time power is removed from the Control Element Drive Mechanism (CEDM) holding coils and the time at which the CEAs reach 90%
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| of their fully inserted positions. For those safety analyses that model CEA insertion following a reactor trip, the 4.0-second CEA drop time is subdivided into two intervals: the holding coil delay time and the CEA insertion time. During the holding coil delay time, which is defined as the time interval between opening of the reactor trip breakers and the time at which the magnetic flux of the CEDM holding coils has decayed enough to allow for CEA motion, the CEAs are assumed to remain in their withdrawn positions. Following expiration of the holding coil delay time, the CEAs are assumed to drop 90% into the core during the remaining CEA insertion time. CEA drop time testing is conducted periodically in accordance with the June 2005 15.0-15 Revision 13
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY PVNGS Technical Specifications and Technical Requirements Manual. Analytical treatment of CEA shutdown reactivity worth versus CEA position is described in UFSAR Section 15.0.3.3.3.
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| 15.0.2.2 Engineered Safety Features The Engineered Safety Feature Actuation Systems (ESFAS) and electrical, instrumentation, and control systems required for safe shutdown are described in UFSAR Sections 7.3 and 7.4, respectively. Analytical ESFAS setpoints and response times are chosen to be consistent with, or conservative with respect to, the setpoint allowable values delineated in the PVNGS Technical Specifications and the response times delineated in UFSAR Section 7.3. Where applicable, analytical ESFAS setpoints are adjusted to account for instrumentation loop uncertainties derived from design control calculations.
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| 15.0.2.3 Control Systems Control and instrumentation systems which may, but are not required to, perform safety functions are described in UFSAR Section 7.7. These include various control systems and the Core Operating Limits Supervisory System (COLSS) which is a monitoring system. In general, normal operation of these control systems is assumed unless lack of operation would make the consequences of the event more adverse. In such cases, the particular control system is assumed to be inoperative, or in the most adverse mode, until the time of operator action.
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| Although these systems are not credited for a safety function, such as mitigation during an event, some safety analyses may credit the normal operation of these systems, consistent with the plant operating procedures, for the purpose of setting June 2005 15.0-16 Revision 13
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY initial conditions for event analysis. For example, the Feedwater Line Break (FWLB) long-term cooling analysis assumes that the Pressurizer Level Control System (PLCS) is initially in normal-automatic operation, with a programmed correspondence between the initial pressurizer water level and RCS loop average temperature. When initial conditions for an event analysis are established in this manner, the values of certain process variables (e.g., temperature, pressure, etc.) may not correspond to their respective Technical Specification limits, and a NRC review may be required to credit the initial normal operation of these systems as an element of methodology for safety analysis.
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| 15.0.2.4 Loss of Off-Site Power Following Turbine Trip The PVNGS off-site and on-site electric power systems are described in UFSAR Sections 8.2 and 8.3, respectively. The PVNGS turbine-generator system is described in UFSAR Section 10.2.
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| During normal plant operations, the Reactor Coolant Pumps (RCPs) are powered from non-Class 1E, 13.8-kV AC busses, which are electrically connected through the unit auxiliary transformer and the isolated phase busses to the main generator. The main generator converts mechanical energy from the main turbine to electrical power. Under normal conditions, a fast bus transfer would be initiated upon tripping of the unit auxiliary transformer output breakers, and alternate supply breakers would close within a few cycles to connect the RCP busses to the startup transformers. The startup transformers, located in the PVNGS switchyard, supply the RCP June 2005 15.0-17 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY busses during plant startup or at other times when the main generator or unit auxiliary transformer is out of service.
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| In the event of a turbine trip during normal plant operations, not involving an electrical fault or underfrequency, the main generator will remain synchronized to the extra high voltage (EHV) grid until residual energy in the turbine is dissipated.
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| The main generator will motor for a short period of time, and will not trip until a sustained reverse power condition exists and the reverse power relay actuates. Reverse power relay actuation will simultaneously trip the generator exciter, the 525-kV breaker and the unit auxiliary transformer output breakers, thereby initiating a fast bus transfer. An analysis of twenty-six PVNGS trips, that occurred between 1990 and 1998, confirms that the time delay between turbine trip (i.e.,
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| turbine stop valve closure) and reverse power relay actuation has varied between 3.969 seconds and 7.55 seconds. Statistical analysis of the data indicates that the time delay would be greater than or equal to three seconds, with a confidence level in excess of 98%. Therefore, in the event of a turbine trip with the RCPs busses connected to the unit auxiliary transformer, the RCPs will receive electrical power for at least three seconds following the turbine trip. Furthermore, a postulated single failure of a breaker to achieve a fast transfer to the backup power supply, which would result in the coastdown of two RCPs, would cause a less rapid loss of flow than the postulated loss of off-site power following a turbine trip (see Table 15.0-0), which would result in the coastdown of all four RCPs.
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| If the main turbine were to trip with the RCP busses connected to the startup transformers, the RCPs would likewise receive June 2005 15.0-18 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY off-site power for at least three seconds following the turbine trip. Reference 11 notes that the loss of a power generating unit on an electrical power grid, such as the loss of a nuclear power plant due to a turbine trip, may generate frequency deviations in the grid, which normally operates at 60 Hz.
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| Under certain conditions the resulting electrical system instability may cause a loss of off-site power to that unit.
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| The degree of instability is characterized by the rate of grid frequency degradation, which is dependent upon the magnitude of the load mismatch and the physical parameters of the grid. The physical response of the grid is dependent upon the available spinning reserve and the stiffness of the grid, that is, the ability to damp out frequency oscillations through load damping. Load shedding may also be utilized to restore the balance between load and power generation and to return the grid frequency to 60 Hz. When corrective actions are not sufficient to avert frequency degradation, loss of off-site power to the plant can occur as a result of that plant tripping offline. Most units are automatically disconnected from the grid between 56 Hz and 58 Hz, to prevent underfrequency damage to plant components. For System 80 plants such as PVNGS, Reference 11 conservatively assumed that a frequency of 57.6 Hz would be the setpoint at which a loss of off-site power occurs.
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| In order to determine the conservative lower bound for the time delay between turbine trip and loss of off-site power, Reference 11 employed the electrical grid system for the Florida Peninsula. This grid can tie into only the Georgia and Alabama grid systems, which can make up only 400 MWe through the transmission lines to Florida. Therefore, the Florida grid becomes an "electrical island" for a generation deficiency caused by the loss of a 1300 MWe unit. On curves of grid June 2005 15.0-19 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY frequency response for this grid system, the effects of a generation deficiency caused by tripping of a System 80 plant were superimposed. Based on this evaluation, a 3.1 second time delay between turbine trip and loss of off-site power was calculated. This time delay is a conservative lower bound because the evaluation assumed:
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| A. No credit for spinning reserve and load shedding; B. The Florida grid "electrical island" conditions (no support from neighboring grid systems);
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| C. Loss of a System 80 plant as a 10% generation loss, which is a much higher percentage than the actual loss (i.e.,
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| less than 3.5%); and D. Loss of off-site power at 57.6 Hz for all System 80 plants.
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| Therefore, for the purpose of simulating the response of PVNGS systems during certain postulated events, a three-second delay may be assumed to occur between a main turbine trip and a loss of off-site power (see Table 15.0-0). The UFSAR Chapter 15 accident analyses that credit this three-second delay include the RCP rotor seizure event, the RCP shaft break event, steam generator tube ruptures, and, beginning with operating Cycle 11, the feedwater line break long-term cooling analysis. This constitutes a credit taken for the functioning of normally operating plant systems, as discussed in UFSAR Section 15.0.1.6.
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| It should be noted, however, that these analyses do not explicitly model a slight decrease in reactor coolant flow that may occur during this three second period, which might result from frequency degradation at the RCP busses. This modeling June 2005 15.0-20 Revision 13
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY assumption is justified by further consideration of the Arizona-New Mexico-California-Southern Nevada grid described in UFSAR Section 8.2.2, to which the PVNGS units are electrically connected. Specifically, this grid can withstand the loss of a PVNGS unit, such as that resulting from a turbine trip, without system instability and with a frequency degradation of less than 0.1 Hz over the three second duration. Assuming a linear relationship between electrical frequency and reactor coolant flow, a frequency degradation of 0.1 Hz would result in only 0.17% reactor coolant flow degradation from full flow. This small amount of flow degradation is less than the conservatisms inherent in the overall uncertainty factors for the CPCs and the Core Operating Limits Supervisory System (COLSS).
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| 15.0.3 CORE AND SYSTEM PERFORMANCE 15.0.3.1 Mathematical Model The NSSS response to various events was simulated using digital computer programs and analytical methods, as described below as well as in individual event analysis sections of this chapter.
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| 15.0.3.1.1 Loss of Flow Analysis Method The method used to analyze events initiated by failures causing a decrease in reactor coolant flowrate is discussed in UFSAR Appendix 15D.
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| 15.0.3.1.2 CEA Ejection Analysis Method The general methodology used to analyze the reactivity and power distribution anomalies associated with CEA ejection events is documented in the Nuclear Steam Supply System (NSSS) vendors Topical Report CENPD-190-A [Reference 12], which was June 2005 15.0-21 Revision 13
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY approved by the NRC for reference in license applications on June 10, 1976. Exceptions to NRC Regulatory Guide 1.77, Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors [Reference 13], are described in UFSAR Section 1.8.
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| Fuel performance (e.g., fuel temperature and enthalpy) is evaluated with STRIKIN-II computer code (see UFSAR Section 15.0.3.1.5). For operating Cycle 10 and earlier cycles, peak RCS pressure was evaluated with the CESEC computer code (see UFSAR Section 15.0.3.1.3.1). Beginning with operating Cycle 11, however, peak RCS pressure is evaluated with the CENTS computer code (see UFSAR Section 15.0.3.1.3.2), as approved by the NRC staff in the safety evaluation report associated with Amendment No. 137 to the PVNGS operating licenses [Reference 14]. For radiological dose assessments associated with postulated CEA ejection events, NSSS analysis codes (e.g.,
| |
| CESEC) are used to estimate the long-term releases from the secondary system until shutdown cooling entry occurs, as described in UFSAR Section 15.4.8.
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| 15.0.3.1.3 NSSS Simulation Computer Programs 15.0.3.1.3.1 CESEC Computer Program NSSS transient simulations, used in long term CEA Ejection radiological consequence evaluations, are performed with the CESEC computer code. The CESEC computer code is described in an April 1974 Topical Report [Reference 15]. The CESEC II and CESEC III versions of the code, which incorporate ATWS model modifications and additional improvements that extend the range of applicability of code June 2007 15.0-22 Revision 14
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY models, are described in Supplements to that Topical Report
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| [References 16, 17, 18, 19, and 20].
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| CESEC computes key system parameters during a transient including core heat flux, pressures, temperatures, and valve actions. A partial list of the dynamic functions included in this NSSS simulation includes: point kinetics neutron behavior, Doppler and moderator reactivity feedback, boron and CEA reactivity effects, multi-node average and hot channel reactor core thermal-hydraulics, reactor coolant pressurization and mass transport, reactor coolant system safety valve behavior, steam generation, steam generator water level, turbine bypass, main steam safety and turbine admission valve behavior, as well as alarm, control, protection, and engineered safety feature systems. The steam turbines, condensers, and their associated controls are not included in the simulation.
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| Steam generator feedwater enthalpy and flowrate are provided as input to CESEC.
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| During the course of execution, CESEC obtains steady-state and transient solutions to the set of equations that mathematically describe the physical models of the subsystems mentioned above.
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| Simultaneous numerical integration of a set of nonlinear, first-order differential equations with time-varying coefficients is carried out by means of a simultaneous solution. As the time variable evolves, edits of the principal systems parameters are printed at prespecified intervals. An extensive library of the thermodynamic properties of uranium dioxide, water, and zircaloy is incorporated into this program.
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| Through the use of CESEC, symmetric and asymmetric plant response over a wide range of operating conditions can be determined.
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| June 2005 15.0-23 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY The CESEC III version of CESEC used in the analyses explicitly models the steam void formation and collapse in the upper head region of the reactor vessel and is documented in Reference 21.
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| Other improvements to this version of CESEC include: a more detailed thermal-hydraulic model that explicitly simulates the mixing in the reactor vessel from asymmetric transients, an RCS flow model that calculates the time dependent reactor coolant mass flow rate in each loop, a wall heat model, a three-dimensional reactivity feedback model, a safety injection tank model, and a primary-to-secondary heat transfer model that calculates the heat transfer for each generator node rather than for a steam generator as a whole.
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| 15.0.3.1.3.2 CENTS Computer Program The CENTS computer program is a computer code developed by the NSSS vendor for the simulation of NSSS transient behavior under normal and abnormal conditions. CENTS is intended to replace the CESEC code originally used to simulate the transient response of the NSSS. The CENTS computer code is documented in Reference 22 and has been approved by the NRC for use in the licensing analyses for PWRs originally designed by Combustion Engineering in Reference 23. The CENTS code approval was subject to five limitations:
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| 1). The CENTS DNBR calculation for determining overall trends in thermal margin should not be used for licensing analyses.
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| 2). The application of CENTS is limited to Combustion Engineering NSSS plants until additional information is submitted and approved.
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| June 2005 15.0-24 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 3). CENTS should not be used for performing LOCA or severe accident licensing analyses.
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| 4). CENTS must use only the point kinetics model in licensing applications.
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| 5). CENTS must not be used for performing CEA ejection licensing analyses. (However, in a safety evaluation report associated with Amendment No. 137 to the PVNGS operating licenses [Reference 14], the NRC staff approved the CENTS computer code for evaluating peak RCS pressure for CEA ejection events. NRC approval was granted on a plant-specific basis for PVNGS, rather than on a generic industry basis. Therefore, beginning with operating Cycle 11, PVNGS CEA ejection licensing analyses utilize the CENTS code for this purpose.)
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| Enhancements to CENTS were made by Westinghouse to more accurately model plant systems and transient behavior of the reactor. These improvements to the CENTS code are documented in Reference 31 and were approved by the NRC in Reference 32.
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| CENTS is a best-estimate code designed to provide a realistic simulation of the neutronics, thermal-hydraulics and plant systems response during transient conditions. CENTS computes key system parameters during a transient including core heat flux, pressures, temperatures, and valve actions. A partial list of the dynamic functions included in this NSSS simulation includes: point kinetics neutron behavior, Doppler and moderator reactivity feedback, boron and CEA reactivity effects, multi-node average and hot channel reactor core thermal-hydraulics, reactor coolant pressurization and mass transport, reactor coolant system safety valve behavior, steam generation, steam generator water level, turbine bypass, main June 2005 15.0-25 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY steam safety and turbine admission valve behavior, as well as alarm, control, protection, and engineered safety feature systems. The steam turbines, condensers, and their associated controls are not included in the simulation. Steam generator feedwater enthalpy and flowrate are provided as input to CENTS.
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| During the course of execution, CENTS obtains steady-state and transient solutions to the set of equations that mathematically describe the physical models of the subsystems mentioned above.
| |
| The RCS model is formulated with five one-dimensional conservation equations. The conservation equations are integrated implicitly by means of a simultaneous solution of the linearized conservation equations. As the time variable evolves, edits of the principal systems parameters are printed at prespecified intervals. An extensive library of the thermodynamic properties of uranium dioxide, water, and zircaloy is incorporated into this program. Through the use of CENTS, symmetric and asymmetric plant response over a wide range of operating conditions can be determined.
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| CENTS uses a more detailed NSSS model than CESEC. The improvements include: the addition of explicit models for determining the nodal solute concentrations and heat loss to the containment, a multi-node versus single node steam generator model, and a non-equilibrium non-homogeneous versus equilibrium homogeneous primary system model.
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| 15.0.3.1.4 COAST Computer Program The COAST computer program is used to calculate the reactor coolant flow coastdown transient for any combination of active and inactive pumps and forward or reverse flow in hot or cold legs. The program is described Reference 24.
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| June 2005 15.0-26 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY The equations of conservation of momentum are written for each of the flow paths of the COAST model assuming unsteady one-dimensional flow of an incompressible fluid. The equation of conservation of mass is written for the appropriate nodal points. Pressure losses due to friction and geometric losses are assumed proportional to the flow velocity squared. Pump dynamics are modeled using a head-flow curve for a pump at full speed and using four-quadrant curves, which are parametric diagrams of pump head and torque on coordinates of speed versus flow, for a pump at other than full speed.
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| 15.0.3.1.5 STRIKIN-II Computer Program The STRIKIN-II computer program is used to simulate the heat conduction within reactor fuel rods and its associated surface heat transfer. The STRIKIN-II program is described in Reference 25.
| |
| The STRIKIN-II computer program provides a single, or dual, closed channel model of a core flow channel to calculate the clad and fuel temperatures for an average or hot fuel rod, and the extent of the zirconium water reaction for a cylindrical geometry fuel rod. STRIKIN-II includes:
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| * Incorporation of all major reactivity feedback mechanisms.
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| * A maximum of six delayed neutron groups.
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| * Both axial (maximum of 20) and radial (maximum of 20) segmentation of the fuel element.
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| * Control rod scram initiation on high neutron power.
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| June 2005 15.0-27 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.0.3.1.6 TORC and CETOP Computer Programs The TORC and CETOP computer programs are used to simulate the fluid conditions within the reactor core region and to calculate fuel pin Departure from Nucleate Boiling Ratio (DNBR). The TORC program is described in References 26 and 27.
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| The CETOP computer program is described in Reference 28.
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| 15.0.3.1.7 Reactor Physics Computer Programs Numerous computer programs are used to produce the input reactor physics parameters required by the NSSS simulation and reactor core programs previously described. These reactor physics computer programs are described in UFSAR Chapter 4.
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| 15.0.3.2 Initial Conditions The events described in this chapter and its appendices have been analyzed over a wide range of initial conditions that encompasses a variety of steady-state operational configurations.
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| In accordance with Reference 1, the most adverse conditions within permitted operating bands for principal process variables have generally been used as initial conditions for the safety analyses. In this context, Reference 1 defines a permitted operating band as the permitted fluctuations in a given parameter or variable, plus any associated uncertainties.
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| However, if the results and conclusions of a safety analysis are insensitive to the initial value chosen for a specific process variable, then a nominal value may instead be used as an initial condition for that variable.
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| June 2005 15.0-28 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY If a process variable is delineated in or controlled by the PVNGS Technical Specifications (e.g., RCS cold leg temperature), the corresponding Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) are typically utilized to define the permitted fluctuations for that variable. The permitted operating band may then be determined by accounting for instrumentation loop uncertainties, which are obtained from design control calculations. This approach ensures consistency between the physical plant, the Technical Specifications, and the safety analyses presented in this UFSAR chapter, as required by 10 CFR 50.36 [Reference 29].
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| If a process variable is not explicitly controlled by the Technical Specifications (e.g., steam generator water level in Mode 1), then the safety analysis guidance contained in Reference 10 may instead be utilized to determine the permitted fluctuations for that variable. Specifically, Reference 10 states that the initial conditions chosen for an analysis should account for the full range of expected normal operating conditions, including the following:
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| * Operating modes (for example, startup, shutdown, loops out of service, refueling);
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| * Systems under manual control considering alarm points, manual action required, and protective overrides; and
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| * Variations in plant parameters with power and core exposure.
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| In such a case, the selection of an initial condition will therefore be consistent with normal plant operating procedures, including control room alarm response procedures and their required operator actions.
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| June 2005 15.0-29 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.0.3.3 Input Parameters 15.0.3.3.1 Doppler Coefficient The fuel temperature coefficient of reactivity (Doppler coefficient) is described in UFSAR Section 4.3. In the safety analyses, the Doppler coefficient is adjusted to account for higher feedback effects in the higher power density core regions, as well as to account for uncertainties in determining the actual fuel temperature reactivity effects. Each analysis utilizes either a more negative or less negative Doppler feedback, in order to produce a more adverse result that is closer to the analytical acceptance criteria.
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| 15.0.3.3.2 Moderator Temperature Coefficient The Moderator Temperature Coefficient (MTC) of reactivity is described in UFSAR Section 4.3. MTC values used in the safety analyses are consistent with the limitations specified in the PVNGS Technical Specifications and the PVNGS Core Operating Limits Reports (COLRs), which vary as a function of both core power level and time in cycle, that is, Beginning-of-Cycle (BOC) to End-of-Cycle (EOC). A conservative MTC value is selected for each analysis on a case-by-case basis.
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| 15.0.3.3.3 Shutdown CEA Reactivity The shutdown reactivity is dependent on the CEA worth available on reactor trip, the axial power distribution, the position of the regulating CEAs, and the time in core life. Please refer to the individual event descriptions in this chapter, to determine the CEA worth that was assumed in each analysis.
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| June 2005 15.0-30 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY Power Dependent Insertion Limits (PDILs), included in the COLR, assure adequate CEA worths are available upon reactor trip.
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| An example of a shutdown reactivity worth versus position curve, for an Axial Shape Index (ASI) of approximately +0.3, is shown in Figure 15.0-2.
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| 15.0.3.3.4 Effective Delayed Neutron Fraction The effective neutron lifetime and delayed neutron fraction are functions of fuel burnup. For each analysis, the values of the neutron lifetime and the delayed neutron fraction are selected consistent with the time in life analyzed.
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| 15.0.3.3.5 Decay Heat Generation Rate Analyses assume decay heat generation based upon an infinite reactor operation at the initial core power level identified for each event.
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| 15.0.4 RADIOLOGICAL CONSEQUENCES Some of the safety analyses presented in this chapter predict that steam or liquid will be released from the RCS or main steam system. Because radioactive material could be present in these discharges, these events are anticipated to result in radiological dose consequences for control room personnel or for the off-site general public. Appendix 15B describes an activity release model that has been used to assess the radiological consequences of certain postulated accidents presented in this chapter. Where applicable, event-specific radiological dose assessment models, which differ from those presented in Appendix 15B, are described on a case-by-case basis in the individual event sections of this chapter.
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| June 2005 15.0-31 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY For radiological event analyses, steam and liquid mass releases to the environment are typically derived from computer code simulations or from alternate calculations. Estimated releases are utilized in the radiological dose analyses, for the purpose of determining whole body and thyroid doses at the Exclusion Area Boundary (EAB), the outer boundary of the Low Population Zone (LPZ), the control room and other required habitable areas. Where applicable, steam and liquid leakage from plant systems, as well as analytical credits taken for automatic actuations (e.g., ESFAS functions) and manual operator actions are described in the individual event sections in this chapter.
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| Unless specified otherwise in the individual event sections in this chapter, the major assumptions used for calculating radiological releases to the environment are as follows:
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| A. The initial RCS activity level is established consistent with the PVNGS licensing basis for the event under consideration. For some events, the event methodology requires that the initial RCS activity be set to the maximum activity due to continuous full power operation with 1% failed fuel. For other events, initial conditions are based on the Technical Specification limit for RCS dose equivalent I-131 specific activity.
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| B. The initial secondary system activity level is equal to 0.1 µCi/gm dose equivalent I-131.
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| C. Primary-to-secondary steam generator tube leakage is included in the calculation of activity releases to the environment from the steam generators. The leakage assumed in the safety analyses is a 1 gpm June 2005 15.0-32 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY primaryto-secondary tube leak (i.e., total leakage for two steam generators).
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| D. For events that require consideration of iodine spiking the following are used:
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| : 1. For iodine spiking generated by the event, the iodine appearance rate is increased by a factor of 500.
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| : 2. For an abnormally high iodine concentration due to a previous iodine spike, a reactor coolant activity of 60 µCi/gm dose equivalent I-131 is assumed.
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| E. Breathing rate (from Table 15B-3).
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| F. Atmospheric dispersion factor (/Q)
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| (from Table 2.3-31)
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| G. For radiological dose calculations performed prior to October 1996, radioiodine dose conversion factors were obtained from TID-14844. For calculations performed after PVNGS Operating License amendments dated October 23, 1996 [Reference 30], however, radioiodine dose conversion factors were obtained from ICRP-30. Both TID-14844 and ICRP-30 factors are shown in UFSAR Appendix 15B, Table 15B-4.
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| June 2005 15.0-33 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY 15.
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| | |
| ==0.5 REFERENCES==
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| : 1. "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,"LWR Edition, NRC Regulatory Guide 1.70, Revision 3, November 1978.
| |
| : 2. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-75/087, as revised through December 31, 1978.
| |
| : 3. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, as revised through July 2000.
| |
| : 4. Maintenance of Records, Making of Reports, 10 CFR 50.71.
| |
| : 5. Content of the Updated Final Safety Analysis Report in Accordance With 10 CFR 50.71(e), NRC Regulatory Guide 1.181, Revision 0, September 1999
| |
| : 6. Guidelines for Updating Final Safety Analysis Reports, Nuclear Energy Institute (NEI) Publication 98-03, Revision 1, June 1999.
| |
| : 7. Anticipated Transients Without Scram for Light Water Reactors, Resolution of Unresolved Safety Issue TAP A-9, Volume 4, NUREG-0460, March 1980.
| |
| : 8. Requirements for Reduction of Risk From Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants, 10 CFR 50.62
| |
| : 9. Compliance With the Anticipated Transients Without Scram (ATWS) Rule - Palo Verde Nuclear Generating Station (PVNGS) Units Nos. 1, 2, and 3 (TAC Nos. 59124, 62698, and 67168), Letter, S. R. Peterson (NRC) to W. F. Conway (PVNGS), October 18, 1990.
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| June 2005 15.0-34 Revision 13
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY
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| : 10. Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, ANSI N18.2-1973, August 1973.
| |
| : 11. Turbine Trip Time Delay, LD-82-40, Letter from A.E.
| |
| Scherer (Director, Nuclear Licensing, Combustion Engineering) to D.G. Eisenhut (Director, Division of Licensing, USNRC), March 31, 1982.
| |
| : 12. "C-E Method for Control Element Assembly Ejection Analysis, CENPD-190-A, January 1976.
| |
| : 13. Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors, NRC Regulatory Guide 1.77, Revision 0, May 1974.
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| : 14. Palo Verde Nuclear Generating Station, Units 1, 2, and 3
| |
| - Issuance of Amendments, RE: Various Administrative Controls (TAC Nos. MB1668, MB1669, and MB1670, Letter, L.R. Wharton (NRC) to G. R. Overbeck (PVNGS), October 15, 2001.
| |
| : 15. CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System, CENPD-107, April 1974 (Proprietary).
| |
| : 16. ATWS Model Modifications to CESEC, CENPD-107, Supplement 1, September 1974 (Proprietary).
| |
| : 17. ATWS Model for Reactivity Feedback and Effect of Pressure on Fuel, CENPD-107, Supplement 2, September 1974 (Proprietary).
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| : 18. ATWS Model Modification to CESEC, CENPD-107, Supplement 3, August 1975.
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| June 2005 15.0-35 Revision 13
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| | |
| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY
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| : 19. ATWS Models Modifications to CESEC, CENPD-107, Supplement 1, November 1975 (Proprietary).
| |
| : 20. ATWS Models Modification to CESEC, CENPD-107, Supplement 4-P, Amendment 1-P, December 1975 (Proprietary).
| |
| : 21. CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System, LD-82-001, Enclosure 1-P to letter from A. E. Scherer to D. G. Eisenhut, December 1981.
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| : 22. Technical Manual for the CENTS Code, CE-NPD 282-P, Volumes 1-3, October 1991 (Proprietary).
| |
| : 23. Acceptance for Referencing of Licensing Topical Report CE-NPD 282-P, Technical Manual for the CENTS Code (TAC No. M82718), Letter, Martin J. Virgilio (NRC) to Mr. S.A.
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| Toelle (ABB-CE), March 17, 1994.
| |
| : 24. COAST Code Description, CENPD-98, April 1973 (Proprietary).
| |
| : 25. STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CENPD-135, April 1974 (Proprietary).
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| STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CENPD-135, Supplement 2, December 1974 (Proprietary).
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| STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CENPD-135, Supplement 4, April 1974 (Proprietary).
| |
| : 26. Combustion Engineering, TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core, CENPD-161-P-A (proprietary), CENPD-161-A (non proprietary),
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| April 1986.
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| June 2011 15.0-36 Revision 16
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| PVNGS UPDATED FSAR ORGANIZATION AND METHODOLOGY
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| : 27. Combustion Engineering, TORC Code - Verification and Simplified Modeling Methods, CENPD-206-P-A (proprietary),
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| CENPD-206-A (non-proprietary), June 1981.
| |
| : 28. Combustion Engineering, CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3, CEN-160(S)-P, September 1981.
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| : 29. Technical Specifications, 10 CFR 50.36.
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| : 30. Issuance of Amendment for the Palo Verde Nuclear Generating Station Unit No. 1 (TAC No. M95880), Unit No. 2 (TAC No. M95881), and Unit No. 3 (TAC No. M95882),
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| Letter, J. W. Clifford (NRC) to J. M. Levine (PVNGS),
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| October 23, 1996.
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| : 31. "Technical Description Manual for the CENTS Code,"
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| WCAP-15996-P, December 2002 (Proprietary).
| |
| : 32. "Final Safety Evaluation for Topical Report WCAP-15996-P, Technical Description Manual for the CENTS Code" (TAC No.
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| MB6982)," Letter, Herbert N. Berkow (NRC) to Mr. Gordon Bischoff (Westinghouse), December 1, 2003.
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| June 2011 15.0-37 Revision 16
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| This page intentionally blank PVNGS UPDATED FSAR 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 DECREASE IN FEEDWATER TEMPERATURE 15.1.1.1 Identification of Causes and Frequency Classification A decrease in main feedwater temperature may be caused by a loss of one or more feedwater heaters. Loss of one of two feedwater heater drain tank pumps interrupts the extraction steam flow through the shell side of one train of high pressure feedwater heaters. No other failure alone would result in the loss of more heaters.
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| A decrease in main feedwater temperature event is classified as an incident of moderate frequency. A decrease in main feedwater temperature event in combination with an additional single failure is classified as an infrequent event.
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| 15.1.1.2 Sequence of Events and System Operation As noted in UFSAR Section 10.4.7, the main feedwater system may be operated in two different modes during normal plant power operations. In the bypass mode of operation, approximately 80%
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| of the total feedwater flow that is delivered to the steam generators passes through two parallel trains of high pressure feedwater heaters, with the remainder of the feedwater flow bypassing these heater trains via an open bypass valve. In the turbo mode of operation, the bypass valve is closed and 100% of the total delivered feedwater flow passes through the high pressure feedwater heater trains.
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| In either mode of operation, the temperature of the feedwater that passes through each parallel train of three high pressure feedwater heaters is increased by approximately 100oF.
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| Feedwater heating is provided by extraction steam from the high pressure turbine and the first and second stage reheaters, June 2011 15.1-1 Revision 16
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| PVNGS UPDATED FSAR which flows through the shell side of the feedwater heaters.
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| Shell side condensate from each train of heaters is drained to a heater drain tank. Heater drain pumps return the condensate to the inlets of the main feedwater pumps, upstream of their respective high pressure feedwater heater trains.
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| Malfunction of a heater drain pump will result closure of the extraction steam isolation valves serving the high pressure feedwater heaters. If extraction steam were to be suddenly and completely lost to one train of three high pressure heaters, the temperature of the feedwater delivered to the steam generators would be reduced by approximately 40oF to 50oF, respectively, if the feedwater system were operating in the bypass or turbo mode at nominal system flow rates.
| |
| A sudden decrease in feedwater temperature would result in a decrease in reactor coolant temperature which, in the presence of a negative Moderator Temperature Coefficient (MTC), would increase core power. A sudden cooldown would likewise cause a decrease in Reactor Coolant System (RCS) and steam generator pressures. Detection of the event could therefore be accomplished by a high reactor power alarm or a steam generator low pressure alarm. If the transient were to result in an approach to Specified Acceptable Fuel Design Limits (SAFDLs),
| |
| trip signals generated by the Core Protection Calculators (CPCs) would ensure that low Departure from Nucleate Boiling Ratio (DNBR) and high Local Power Density (LPD) limits are not exceeded.
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| June 2011 15.1-2 Revision 16
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| PVNGS UPDATED FSAR 15.1.1.3 Core and System Performance A decrease in main feedwater temperature event would not challenge fuel pellet integrity. During the event, any short-term increase in reactor power, or three-dimensional shift in power generation within the core, would not be of sufficient magnitude to raise the linear heat rate above that required to cause fuel centerline melting.
| |
| A decrease in main feedwater temperature event would result in a smaller decrease in RCS temperature than an increase in main steam flow event involving the quick opening of eight Steam Bypass Control System (SBCS) valves or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV) (see UFSAR Section 15.1.4). The smaller RCS cooldown would result in less of a power increase, and hence less of a decrease in the minimum hot channel DNBR during the transient. The minimum hot channel DNBR establishes whether a fuel design limit has been exceeded and therefore whether fuel cladding degradation might be anticipated.
| |
| For the decrease in main feedwater temperature event in combination with a single failure, the parameter of concern is likewise the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distribution effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worst single failure for this event is a Loss of Offsite Power (LOP) following a June 2011 15.1-3 Revision 16
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| | |
| PVNGS UPDATED FSAR turbine trip, which would cause the Reactor Coolant Pumps (RCPs) to coast down and rapidly reduce the coolant flow rate.
| |
| This event, however, would result in a Nuclear Steam Supply System (NSSS) response that is similar to, but less severe than, that caused by the increase in main steam flow event involving the quick opening of eight SBCS valves or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV) in combination with LOP (See UFSAR Section 15.1.3 and 15.1.4). These events result in more severe RCS cooldown that in turn results in more of an increase in power, and hence more of a decrease in the minimum hot channel DNBR. Therefore, the DNBR at the moment RCPs begin to coastdown would be bounded by those events. For this reason, the infrequent decrease in the feedwater temperature event (in combination with a single failure) is bounded by the infrequent event involving the quick opening of eight SBCS valves and the inadvertent opening of steam generator atmospheric dump valve (OSGADV) (in combination with single failure) with respect to the DNBR SAFDL.
| |
| In addition, this event would result in a more benign minimum DNBR than the resulting from the limiting infrequent event that is described in the UFSAR Appendix 15.E. the event described in UFSAR Appendix 15.E establishes a limiting infrequent event, including all incidents of moderate frequency in combination with a single failure, with respect to DNBR degradation, assuming that the DNBR is already at the SAFDL when the isngle failure, LOP, occurs.
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| 15.1.1.4 RCS Pressure Boundary Barrier Performance A decrease in main feedwater temperature event is characterized by an initial cooldown of the primary and secondary systems, and decreasing RCS and steam generator pressures. If the event June 2011 15.1-4 Revision 16
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| PVNGS UPDATED FSAR results in a reactor trip and Main Steam Isolation Signal (MSIS), repressurization of the RCS and steam generators would occur due to decay heat from radionuclides in the core, heat stored in the metal structures of the NSSS, and heat from any operating RCPs. Additionally, if pressurizer pressure decreases below the Safety Injection Actuation Signal (SIAS) setpoint, safety injection flow may also result in repressurization of the RCS. Eventually, however, plant operators would take action to cool down and depressurize the plant to Shutdown Cooling (SDC) entry conditions. This may be accomplished by feeding the steam generators with Auxiliary Feedwater (AFW) flow and by releasing steam through the Atmospheric Dump Valves (ADVs).
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| The subsequent heatup and repressurization of the NSSS would not challenge RCS pressure boundary peak pressure limits.
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| Prior to the operators taking action to cool down the plant, the secondary system peak pressure would be limited by the Main Steam Safety Valves (MSSVs), which have sufficient capacities to relieve the steam that may be generated by NSSS heat sources. Furthermore, if the heat transfer rate from the RCS to the secondary system were degraded for any reason, as might occur when a LOP results in a loss of forced RCS coolant flow, the Pressurizer Safety Valves (PSVs) may also open to limit the RCS peak pressure. Because the maximum allowable lift settings for the MSSVs and PSVs are well below the peak pressure regulatory limits for this event, a decrease in feedwater temperature, or a decrease in feedwater temperature in combination with a single failure, would not challenge the RCS pressure boundary through overpressurization of either the primary or secondary systems.
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| June 2011 15.1-5 Revision 16
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| PVNGS UPDATED FSAR 15.1.1.5 Containment Performance and Radiological Consequences A decrease in feedwater temperature event in combination with an additional single failure (for example, a LOP following turbine trip) is classified as an infrequent event, which may result in limited fuel cladding degradation. Offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. Additionally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A. The offsite and control room radiological dose consequences associated with this infrequent event are bounded by those that may result from an Inadvertent Opening of a Steam Generator Atmospheric Dump Valve with a Loss of Offsite Power (IOSGADVLOP) event, (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E), and are in compliance with regulatory guidelines.
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| 15.1.1.6 Conclusions Evaluation of the decrease in feedwater temperature event shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| * For the moderate frequency decrease in feedwater temperature event (without an additional single failure),
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| fuel cladding integrity will be maintained.
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| June 2011 15.1-6 Revision 16
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| PVNGS UPDATED FSAR
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| * For the infrequent decrease in feedwater temperature event (with an additional single failure), limited fuel cladding degradation may occur. However, offsite and control room radiological dose consequences are bounded by those that may result from an IOSGADVLOP event, (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E), and are in compliance with regulatory guidelines.
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| 15.1.2 INCREASE IN MAIN FEEDWATER FLOW 15.1.2.1 Identification of Causes and Frequency Classification An increase in main feedwater flow to the steam generators may be caused by inadvertent equipment malfunctions in the Feedwater Control System (FWCS), resulting in the opening of feedwater control valves beyond their desired positions, or an increase in feedwater pump speed.
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| An increase in main feedwater flow event is classified as an incident of moderate frequency. An increase in main feedwater flow event in combination with an additional single failure is classified as an infrequent event.
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| 15.1.2.2 Sequence of Events and System Operation During normal power operations, feedwater flow is automatically controlled by the FWCS through main feedwater control valves, which establish steam generator feedwater balancing in conjunction with variable-speed feedwater pump turbine drives. If a hypothetical equipment malfunction were to suddenly increase FWCS demand signals to their maximum output values, the control valves would stroke fully open and the feedwater pumps would accelerate to maximum speed. If the plant were operating at full power when June 2011 15.1-7 Revision 16
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| PVNGS UPDATED FSAR this occurred, the maximum increase in feedwater flow that would result is estimated to be approximately 25% of the nominal main feedwater system flow rate.
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| A sudden increase in main feedwater flow to the steam generators would result in a decrease in reactor coolant temperature which, in the presence of a negative MTC, would increase core power. A sudden increase in feedwater flow would also cause an increase in steam generator water level and a decrease in RCS and steam generator pressures. Detection of the event could therefore be accomplished by a high reactor power alarm, a steam generator low pressure alarm, or a steam generator high water level alarm. If the transient were to result in an approach to SAFDLs, trip signals generated by the CPCs would ensure that low DNBR and high LPD limits were not exceeded. Likewise, if steam generator water level increased significantly, a reactor trip and main steam isolation on high steam generator level would occur, thereby protecting the steam generators from overfilling.
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| 15.1.2.3 Core and System Performance An increase in main feedwater flow event would not challenge fuel pellet integrity. During the event, any short-term increase in reactor power, or three-dimensional shift in power generation within the core, would not be of sufficient magnitude to raise the linear heat rate above that required to cause fuel centerline melting.
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| An increase in main feedwater flow event would result in a smaller decrease in RCS temperature than an increase in main steam flow event involving the quick opening of eight SBCS valves or the inadvertent opening of a steam generator atmospheric dump valve (OSGADV) (see UFSAR Section 15.1.3 and 15.1.4). The June 2011 15.1-8 Revision 16
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| PVNGS UPDATED FSAR smaller RCS cooldown would result in less of a power increase, and hence less of a decrease in the minimum hot channel DNBR during the transient. The minimum hot channel DNBR establishes whether a fuel design limit has been exceeded and therefore whether fuel cladding degradation might be anticipated.
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| For the increase in main feedwater flow event in combination with a single failure, the parameter of concern is likewise the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distribution effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worst single failure for this event is a LOP following a turbine trip, which would cause the RCPs to coast down and rapidly reduce the coolant flow rate. This event, however, would result in a Nuclear Steam Supply System (NSSS) response that is similar to, but less severe than, that caused by the increase in main steam flow event involving the quick opening of eight SBCS valves or an inadvertent opening of a steam generator atmospheric dump valve in combination with LOP (See UFSAR Section 15.1.3 and 15.1.4). The quick opening of eight SBCS valves results in more severe RCS cooldown that in turn results in more of an increase in power, and hence more of a decrease in the minimum hot channel DNBR. Therefore, the DNBR at the moment RCPs begin to coastdown would be bounded by those events. For this reason, the infrequent increase in the feedwater flow event (in combination with a single failure) is bounded by the infrequent event involving the quick opening of eight SBCS valves and inadvertent opening of a steam generator atmospheric dump valve (OSGADV) (in combination with single failure) with respect to the DNBR SAFDL.
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| June 2011 15.1-9 Revision 16
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| PVNGS UPDATED FSAR In addition, this event would result in a more benign minimum DNBR than that resulting from the limiting infrequent event that is described in the UFDAR Appendix 15.E. The event described in UFSAR Appendix 15.E establishes a limiting infrequent event, including all incidents of moderate frequency in combination with a single failure, with respect to DNBR degradation, assuming that the DNBR is already at the SAFDL when the single failure, LOP, occurs.
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| 15.1.2.4 RCS Pressure Boundary Barrier Performance An increase in main feedwater flow event is characterized by an initial cooldown of the primary and secondary systems, decreasing RCS and steam generator pressures, and increasing steam generator water level. If the event results in a reactor trip and MSIS, repressurization of the RCS and steam generators may occur due to decay heat from radionuclides in the core, heat stored in the metal structures of the NSSS, and heat from any operating RCPs. Additionally, if pressurizer pressure decreases below the SIAS setpoint, safety injection flow may also serve to repressurize the RCS. Eventually, however, plant operators would take action to cool down and depressurize the plant to SDC entry conditions. This may be accomplished by feeding the steam generators with AFW flow and by releasing steam through the ADVs.
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| The subsequent heatup and repressurization of the NSSS would not challenge RCS pressure boundary peak pressure limits.
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| Prior to the operators taking action to cool down the plant, the secondary system peak pressure would be limited by the MSSVs, which have sufficient capacities to relieve the steam that may be generated by NSSS heat sources. Furthermore, if the heat transfer rate from the RCS to the secondary system June 2011 15.1-10 Revision 16
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| PVNGS UPDATED FSAR were degraded for any reason, as might occur when a LOP results in a loss of forced RCS coolant flow, the PSVs may also open to limit the RCS peak pressure. Because the maximum allowable lift settings for the MSSVs and PSVs are well below the peak pressure regulatory limits for this event, an increase in main feedwater flow, or an increase in main feedwater flow in combination with a single failure, would not challenge the RCS pressure boundary through overpressurization of either the primary or secondary systems.
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| 15.1.2.5 Containment Performance and Radiological Consequences An increase in main feedwater flow event in combination with an additional single failure (for example, a LOP following turbine trip) is classified as an infrequent event, which may result in limited fuel cladding degradation. Offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. Additionally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A.
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| The offsite and control room radiological dose consequences associated with this infrequent event are bounded by those that may result from an IOSGADVLOP event, (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E),
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| and are in compliance with regulatory guidelines.
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| 15.1.2.6 Conclusions Evaluation of the increase in main feedwater flow event shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained June 2011 15.1-11 Revision 16
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| PVNGS UPDATED FSAR
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| * below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| * For the moderate frequency increase in main feedwater flow event (without an additional single failure), fuel cladding integrity will be maintained.
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| * For the infrequent increase in main feedwater flow event (with an additional single failure), limited fuel cladding degradation may occur. However, offsite and control room radiological dose consequences are bounded by those that may result from an IOSGADVLOP event, (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E), and are in compliance with regulatory guidelines.
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| 15.1.3 INCREASE IN MAIN STEAM FLOW 15.1.3.1 Identification of Causes and Frequency Classification An increase in main steam flow event may be caused by equipment malfunctions or inadvertent operator actions that result in the sudden opening of one or more Steam Bypass Control System (SBCS) valves; the opening of a turbine admission valve beyond its desired position; or the opening of an Atmospheric Dump Valve (ADV). Postulated events that involve the SBCS and turbine admission valves, which are located downstream of the Main Steam Isolation Valves (MSIVs), are addressed in this UFSAR section. Inadvertent openings of ADVs, which are located upstream of the MSIVs, are addressed in UFSAR Section 15.1.4.
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| An increase in main steam flow event is classified as an incident of moderate frequency. An increase in main steam flow event in combination with an additional single failure is classified as an infrequent event.
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| June 2009 15.1-12 Revision 15
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| PVNGS UPDATED FSAR 15.1.3.2 Sequence of Events and System Operation When the plant is operating at full power, main steam flow will increase if a turbine admission valve suddenly opens beyond its desired position, or if one or more SBCS valves suddenly open.
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| The largest possible increase in main steam flow would occur if an equipment malfunction resulted in the simultaneous quick opening of all eight SBCS valves (SBCVs). The maximum allowable capacity of each non-safety-related SBCV is 11% of the Design Steam Rate (DSR), where the DSR is based on the original licensed power level of 3800 MWt. The DSR is less than the nominal steam flow rate at the current licensed Rated Thermal Power (RTP). Given the maximum capacity of each SBCV, main steam flow would increase by less than 88% of the steam flow corresponding the current licensed RTP.
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| Table 15.1.3-1 provides the sequence of events for the limiting moderate frequency increase in main steam flow analysis, involving a quick opening of all eight SBCVs at 100% power. This sequence of events was obtained by simulating the event with the computer codes identified in Section 15.1.3.3. Figures 15.1.3-1 through 15.1.3-12 show the short-term response of key NSSS parameters during the portion of the event that presents the greatest challenge to SAFDLs. Specifically, Figure 15.1.3-11 shows how DNBR approaches and passes through a minimum value shortly after event initiation. Figures 15.1.3-13 through 15.1.3-15 show the long-term response of key parameters prior to the time at which operators are assumed to take control of plant (i.e., 30 minutes after event initiation).
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| June 2009 15.1-13 Revision 15
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| PVNGS UPDATED FSAR The sudden increase in main steam flow (Figure 15.1.3-1) results in a decrease in reactor coolant temperature (Figure 15.1.3-2) which, in the presence of a negative MTC, results in an increase in reactivity (Figure 15.1.3-3), core power (Figure 15.1.3-4), and core heat flux (Figure 15.1.3-5).
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| June 2009 15.1-14 Revision 15
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| PVNGS UPDATED FSAR Table 15.1.3-1 SEQUENCE OF EVENTS FOR THE LIMITING MODERATE FREQUENCY STEAM BYPASS CONTROL SYSTEM MALFUNCTION SAFETY ANALYSIS Time Event 0.00 Eight SBCS valves quick-open 6.49 Steam generator pressure reaches MSIS setpoint 6.72 CPC VOPT reaches reactor trip setpoint 7.47 Reactor trip breakers open 7.72 Turbine trip occurs 8.08 CEAs begin to fall 8.50 Minimum DNBR occurs 12.10 MSIVs close. Flow through SBCS valves stops.
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| 1 112.02 MSSVs open on steam generator 1 1
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| 112.02 MSSVs open on steam generator 2 112.39 Maximum steam generator 2 pressure occurs 1
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| 165.58 MSSVs close on steam generator 1 1
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| 165.58 MSSVs close on steam generator 2 323.52 Pressurizer Pressure reaches SIAS septoint.
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| 363.52 HPSI flow Begins 600.55 Maximum steam generator 1 pressure occurs 665.64 AFW flow delivered to steam generator 1 665.64 AFW flow delivered to steam generator 2 1009.47 AFW flow shutoff reached in steam generator 2.
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| 1011.11 AFW flow shutoff reached in steam generator 1.
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| 1800.00 Plant operators take control of the plant Only the first opening and closing of the MSSVs are documented.
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| 1 June 2011 15.1-15 Revision 16
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| PVNGS UPDATED FSAR Table 15.1.3-1 (Page 2 of 2)
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| SEQUENCE OF EVENTS FOR THE LIMITING MODERATE FREQUENCY STEAM BYPASS CONTROL SYSTEM MALFUNCTION SAFETY ANALYSIS Time Event 1000 Steam generator No. 1 water level recovers to high level AFAS setpoint. AFW flow stopped to steam generator No. 1 1073 MSSVs open on steam generator No. 2 1074 MSSVs open on steam generator No. 1 1114 MSSVs close on steam generator No. 2 1115 MSSVs close on steam generator No. 1 1269 MSSVs open on steam generator Nos. 1 and 2 1279 Steam generator No. 1 water level decreases to low level AFAS setpoint. AFW flow reinitiated to steam generator No. 1 1280 Steam generator No. 2 water level decreases to low level AFAS setpoint. AFW flow reinitiated to steam generator No. 2 1302 MSSVs open on steam generator Nos. 1 and 2 1556 Steam generator No. 2 water level recovers to high level AFAS setpoint. AFW flow stopped to steam generator No. 2 1559 Steam generator No. 1 water level recovers to high level AFAS setpoint. AFW flow stopped to steam generator No. 1 1688 MSSVs open on steam generator Nos. 1 and 2 1728 MSSVs close on steam generator No. 2 1729 MSSVs close on steam generator No. 1 1800 Plant operators take control of plant June 2009 15.1-16 Revision 15
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| PVNGS UPDATED FSAR A sudden increase in main steam flow would also cause a short-term increase in the indicated steam generator water level due to swell effects (Figure 15.1.3-6), and an almost immediate decrease in both RCS and steam generator pressures (Figures 15.1.3-7 and 15.1.3-8, respectively). Despite the short-term increase in indicated steam generator water level, the mass of liquid in both steam generators would actually decrease slightly due to increased steam flow during the early part of the transient (Figure 15.1.3-9). Detection of the event may be accomplished by a high reactor power alarm, an RCS low pressure alarm, or a steam generator low pressure alarm.
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| Rapid cooling of the RCS would increase coolant density and cause a temporary reduction in pressurizer level (Figure 15.1.3-10). When the colder fluid exiting the steam generator reaches the core inlet, the core power will increase, causing the core heat flux to increase, albeit with a delay of a few seconds. This increased heat flux, in conjunction with the decreased RCS pressure, would cause the DNBR to decrease (Figure 15.1.3-11).
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| As the event proceeds, the steam generator pressure decreases and approaches the Low Steam Generator Pressure Trip (LSGPT).
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| At the same time, reactor power increases toward a CPC auxiliary trip, the Variable Overpower Trip (VOPT). A VOPT will occur before the LSGPT for more negative values of MTC; however, for less negative values of MTC, the power increase may not be sufficient to result in the VOPT before the steam generator pressure reaches the LSGPT. The most limiting analysis case, which yielded the lowest DNBR value, was found to occur when the VOPT and LSGPT trips occurred at approximately the same time.
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| June 2009 15.1-17 Revision 15
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| PVNGS UPDATED FSAR Following reactor trip, Control Element Assemblies (CEAs) would fall into the reactor core, rapidly reducing reactivity, core power, and core heat flux. As indicated in Table 15.1.3-1, minimum DNBR is predicted to occur shortly after the CEAs begin inserting into the core.
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| The course of the transient depends on the choice of MTC, because the LSGPT occurs in conjunction with a Main Steam Isolation Signal (MSIS). Thus, depending on which trip signal intervenes, and even how close the LSGPT is to the setpoint, the system responses for the steam system may take one of two different paths.
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| In the case of tripping on the LSGPT, the MSIS accompanying the trip would result in closure of the MSIVs and Main Feedwater Isolation Valves (MFIVs). Closure of these valves would stop the flow of main feedwater to the steam generators (Figure 15.1.3-12) as well as the flow of steam out of the open SBCVs. It would also stop the temperature and pressure decrease in the RCS temporarily and there would be no Safety Injection Actuation Signal (SIAS).
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| If the trip were to occur on the VOPT and the steam generator pressure were far enough above the MSIS setpoint to avoid an immediate MSIS, there would be a turbine trip associated with the VOPT, but the flow through the SBCVs would continue. Thus, in spite of the turbine trip, RCS pressure would continue to decrease as heat loss through the open SBCVs exceeds the post-trip heat generation rate in the NSSS. As the RCS continued to cool down, the pressurizer would eventually empty and pressurizer pressure would decrease to the SIAS setpoint, actuating the safety injection pumps. After a brief time delay, during which the High Pressure Safety Injection (HPSI) pumps reach full speed, safety injection flow would be delivered to June 2009 15.1-18 Revision 15
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| PVNGS UPDATED FSAR the RCS, provided RCS pressure was below the shutoff head of the HPSI pumps. Shortly thereafter, the decreasing steam generator pressure would reach the MSIS setpoint, thereby halting the steam flow from both steam generators and effectively stopping the cooldown.
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| Following the MSIS, the RCS and steam generators would begin to repressurize, due to the heat released by fission product decay, running RCPs, hot metal structures in the NSSS and, in the event of VOPT, the HPSI flow (Figures 15.1.3-13 and 13.1.3-14, respectively). The pressure in the steam generators would continue to rise until the MSSV setpoint is reached. At this point, the MSSVs would open, limiting the pressure and releasing steam generator inventory. As the steam generator pressure dropped from this release, the MSSVs would close and the steam generator pressure would begin rising again. The release of inventory associated with each lifting of the MSSVs would cause the liquid levels in the steam generators to decrease until they reach the Auxiliary Feedwater Actuation Signal (AFAS) setpoint.
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| This would initiate the delivery of AFW flow to the steam generators, which would continue until the high level reset was reached.
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| After initiation of AFW flow to the steam generators, the plant would achieve quasi-steady-state conditions with heat removal provided by cycling of the MSSVs, and makeup water provided by cycling of the AFW system as water levels rise and fall in both steam generators. Pressurizer level would likewise rise and fall as the RCS periodically heats up, then temporarily cools down again with each MSSV lift (Figure 15.1.3-15).
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| Operator action is not credited in the safety analysis until 30 minutes following the SBCS malfunction. At that time, it is assumed that plant operators will take action to initiate a June 2009 15.1-19 Revision 15
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| PVNGS UPDATED FSAR controlled plant cooldown to SDC entry conditions, for example by establishing a steaming path through the ADVs.
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| 15.1.3.3 Core and System Performance An event involving an increase in main steam flow would not challenge fuel pellet integrity. During the event, any short-term increase in reactor power, or three-dimensional shift in power generation within the core, would not be of sufficient magnitude to raise the linear heat rate above that required to cause fuel centerline melting.
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| An increase in main steam flow event results in thermal margin degradation very similar to that experienced in the IOSGADV moderate frequency event with respect to the DNBR SAFDL. The mathematical models, input parameters, initial conditions, and results of the SBCS malfunction safety analysis are described in the subsections below for this moderate frequency event.
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| For the infrequent increase in main steam flow event in combination with a single failure, the parameter of concern is also the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distribution effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worst single failure for this event is a LOP following a turbine trip, which would cause the RCPs to coast down and rapidly reduce the coolant flow rate.
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| The infrequent event involving an increase in main steam flow is initiated by the quick opening of eight SBCVs combined with LOP. This event is bounded by the IOSGADVLOP event with respect to the DNBR SAFDL and dose consequences for those June 2011 15.1-20 Revision 16
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| PVNGS UPDATED FSAR events involving an increase in heat removal by the secondary system.
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| The SBCS malfunction with a LOP is still bounded by the infrequent event described in the UFSAR Appendix 15.E, the Loss of Flow (LOF) from a SAFDL. The LOF from a SAFDL establishes a limiting infrequent event with respect to DNBR degradation for all moderate frequency events in combination with a single failure by assuming that the DNBR is already at the SAFDL when the LOF occurs. The minimum DNBR during the limiting moderate frequency event involving increased main steam flow remains above the DNBR SAFDL and, when combined with LOP following a turbine trip, would yield a minimum DNBR that is higher than that for the LOF from a SAFDL. Hence, the limiting infrequent event for an increase in main steam flow (quick opening of eight SBCVs with a single failure) is bounded with respect to the DNBR by the IOSGADVLOP and the limiting infrequent event evaluated in UFSAR Appendix 15.E.
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| 15.1.3.3.1 Mathematical Models The moderate frequency event for an increase in main steam flow was analyzed with the following mathematical models:
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| * The CENTS computer code was used to simulate the NSSS transient response. The CENTS computer code is described June 2011 15.1-21 Revision 16
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| PVNGS UPDATED FSAR
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| * in UFSAR Section 15.0.3.1.3.2 and in an NSSS vendor (1)(2) topical report.
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| * The CPC FORTRAN computer code was used to simulate CPC reactor trip functions and to predict the time at which the CPC VOPT setpoint would be reached. This time, with appropriate delays for signal processing and opening of the reactor trip breakers, was utilized in CENTS code input. The CPCs are described in UFSAR Section 7.2, and associated algorithms and simulation code are described (3)(4) in NSSS vendor topical reports.
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| * The CETOP-D computer code, which uses the CE-1 Critical Heat Flux (CHF) correlation, was used to calculate the initial and transient DNBR values. CETOP-D was also used to determine initial Power Operating Limit (POL) conditions for this event (see UFSAR Section 15.1.3.3.2 for additional information on POL conditions). The CETOP-D computer code is described in UFSAR Section 4.4 (5)(6)(7) and in NSSS vendor topical reports.
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| 15.1.3.3.2 Input Parameters and Initial Conditions Table 15.1.3-2 summarizes the key input parameters and initial conditions utilized in the safety analysis for the limiting moderate frequency event involving an increase in main steam flow.
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| The following points serve to explain the selection of initial conditions as they appear in Table 15.1.3-2:
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| * During normal plant power operations, the Core Operating Limits Supervisory System (COLSS) monitors various parameters to assist operators in maintaining plant June 2011 15.1-22 Revision 16
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| PVNGS UPDATED FSAR conditions within the Technical Specification Limiting Conditions for Operation (LCOs). When COLSS is in service, it continuously calculates the core power at which the DNBR SAFDL would be reached, based on the measured temperature, pressure, flow, radial peaking factor, and axial power distribution. This COLSS-calculated core power is then divided by a numerical value, called the Required Overpower Margin (ROPM), to yield a DNBR Power Operating Limit (POL). If the measured core power exceeds the calculated POL, a COLSS alarm would alert plant operators to take action as required by Technical Specifications. The DNBR POL therefore serves to preserve thermal margin to accommodate potential operational transients. For this safety analysis, it was assumed that the plant would be operating at a POL condition immediately prior to the event, at the most limiting RCS flow rate. Because the event simulation was initiated from a calculated POL, additional power measurement uncertainties were not added to the initial assumed core power level. Analytical values for the initial core inlet temperature, pressurizer pressure, and RCS coolant flow rate corresponding to the POL were determined with the CETOP-D computer code, for an initial core power of 100% of RTP.
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| June 2009 15.1-23 Revision 15
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| PVNGS UPDATED FSAR Table 15.1.3-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE STEAM BYPASS CONTROL SYSTEM MALFUNCTION SAFETY ANALYSIS Parameter Assumed Value Initial Core Power (% of RTP) 100%
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| Initial core Inlet Temperature (F) 566 Initial Pressurizer Pressure (psia) 2100 Initial RCS Flow Rate (% of Design Rated) 110.4 Initial Pressurizer Water Level (% Narrow Range) 24 Initial Steam Generator Water Level (% Wide Range) 80.7
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| -4 Moderator Temperature Coefficient (/F) -2.25x10 Doppler Fuel Temperature Coefficient Least Negative Delayed Neutron Kinetics EOC Axial Shape Index for DNBR calculation -0.2 CEA Worth at Trip (%) 8.0 2
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| Fuel Rod Gap Conductance (BTU/hr-ft -F) 6100 Total Number of Plugged Steam Generator Tubes 0 Single Failure None June 2011 15.1-24 Revision 16
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| PVNGS UPDATED FSAR
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| * DNBR degradation during the event is sensitive to the initial pressurizer water level and a minimum value was selected for the initial conditions, based on parametric evaluation.
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| * DNBR degradation during the event is not sensitive to the initial steam generator level unless a high steam generator level trip were credited. Since no credit is taken for that trip, it was disabled and nominal values were used for initial steam generator level. The CENTS code automatically calculates the initial steam generator pressure, given steam generator water level, reactor power, and other inputs to the computer code.
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| * The SBCS malfunction event causes a rapid cooldown of the RCS. Using the most negative MTC allowed by the Technical Specifications and COLR would result in the most rapid power increase and the greatest overshoot in power after the VOPT. However, because the heat flux lags behind the power, it does not necessarily produce the greatest decrease in DNBR. As the MTC is made less negative, the power excursion is slowed down. This allows the heat flux to stay closer to the power and it allows the VOPT to shift upwards a small amount. The heat flux at the time of minimum DNBR increases for this event as the MTC is made less negative and the DNBR decreases. As the MTC is made less and less negative, a point is reached at which the LSGPT and the VOPT are reached concurrently. This MTC results in the lowest minimum DNBR. Making the MTC less negative will result in the LSGPT occurring with the power below the VOPT setpoint. This decreases the heat flux and increases the minimum DNBR. The value shown in Table 15.1.3-2 June 2011 15.1-25 Revision 16
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| PVNGS UPDATED FSAR corresponds to the limiting case, for which the VOPT and LSGPT occur at approximately the same time.
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| * The least negative Doppler fuel temperature coefficient curve, which corresponds to Beginning of Cycle (BOC) conditions, was used. Use of the least negative values minimizes the addition of negative reactivity caused by increasing fuel temperature. This is not important for cases with intermediate values of MTC, but for the case with the most negative MTC, this choice will result in a more rapid increase in reactor core power.
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| * End of Cycle (EOC) values were chosen to model delayed neutron kinetics. EOC values enhance the power excursion by minimizing the effect of delayed neutrons on the rate of power increase. This is not important for cases with intermediate values of MTC, but for the case with the most negative MTC, this choice will result in a more rapid increase in reactor core power.
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| * If power generation is shifted toward the bottom of the core, the insertion of negative reactivity following reactor trip will be somewhat delayed until the CEAs have inserted farther into the core. The scram reactivity curve was therefore based on a positive ASI representing a bottom-peaked core. The time versus scram reactivity curve was adjusted to account for a 0.6-second CEA holding coil time delay following opening of the reactor trip breakers, and normalized to model 90% CEA insertion at 4.0 seconds after power is removed from Control Element Drive Mechanism (CEDM) coils, (see UFSAR Section 3.9.4).
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| * The CEA worth at trip represents the minimum SCRAM worth for Hot Full Power (HFP) conditions at BOC, assuming the most June 2011 15.1-26 Revision 16
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| PVNGS UPDATED FSAR reactive CEA remains stuck out of the core following reactor trip. This is more limiting (less negative) than the anticipated scram reactivity worth at other times during the operating cycle for HFP conditions.
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| * The fuel rod gap conductance value was selected in a manner that energy from the fuel would quickly reach the surface of the fuel rod clad. A large value results in a higher heat flux and greater degradation of DNBR during the initial power excursion.
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| * It was assumed that steam generator tubes were not plugged for this safety analysis. This enhances heat transfer from the RCS to the main steam system, which in turn enhances the initial RCS cooldown and maximizes the positive reactivity insertion due to the negative MTC.
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| Additionally, this enhances the decrease in RCS pressure during the cooldown, which serves to degrade DNBR.
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| * For the moderate frequency event involving an increase in main steam flow, an additional single failure was not assumed.
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| For those safety-related Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) setpoints and response times that had a direct effect on acceptance criteria for this event, analytical values were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| June 2011 15.1-27 Revision 16
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| PVNGS UPDATED FSAR 15.1.3.3.3 Results The analysis shows that the calculated minimum DNBR approaches a value of 1.41 at 8.5 seconds following event initiation.
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| This value is greater than the DNBR SAFDL of 1.34. The linear heat rate will not present a credible challenge to fuel centerline melting. Therefore, fuel damage is not predicted to occur for this event.
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| 15.1.3.4 RCS Pressure Boundary Barrier Performance The increased steam flow events are characterized by a cooldown and depressurization of the primary and secondary systems in the short-term. However, this initial cooldown and depressurization is reversed after automatic halting of the steam flow following an MSIS resulting in heat-up and repressurization of primary and secondary systems. A comparison of the RCS pressures and temperatures shows that the RCS temperature and pressure decrease for the increased main steam flow event due to the opening of one SBCV is similar to that for the IOSGADV event described in UFSAR Section 15.1.4 because of the same flow capacity of an ADV and SBCV (11% of the DSR). Opening of more SBCVs would result in a larger cooldown and depressurization of the RCS than the IOSGADV event in the short-term. Additionally, over the longer term, steam flow through the open SBCVs would be halted automatically following an MSIS and controlled heat removal would be achieved by both steam generators, while an IOSGADV event would result in dry-out of one steam generator and long-term controlled heat removal would be achieved through only one steam generator. As a result of the increased cooling from the stuck-open ADV, the RCS would be cooler for the IOSGADV for several minutes post-SCRAM. However, because the heat is being removed through only June 2011 15.1-28 Revision 16
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| PVNGS UPDATED FSAR one steam generator, the RCS temperature and pressure for the IOSGADV would be significantly higher than that for the SBCS malfunction in the long term. Secondary pressures are limited by the MSSV setpoints and will be no greater for the SBCS malfunction than the intact steam generator in the IOSGADV.
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| Based on the above, the maximum primary and secondary pressures for moderate frequency and infrequent events involving a main steam flow increase (inadvertent opening of one or more SBCVs) are bounded by those for moderate frequency and infrequent IOSGADV events, respectively.
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| 15.1.3.5 Containment Performance and Radiological Consequences An increase in main steam flow event in combination with an additional single failure (for example, LOP following turbine trip) is classified as an infrequent event. Offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. Additionally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A. The offsite and control room radiological dose consequences associated with this infrequent event are bounded by those that may result from an IOSGADVLOP event, and are in compliance with regulatory guidelines (see UFSAR Section 15.1.4.5).
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| June 2011 15.1-29 Revision 16
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| PVNGS UPDATED FSAR 15.1.3.6 Conclusions Evaluation of the increase in main steam flow event shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| * For the moderate frequency increase in main steam flow event (without an additional single failure), fuel cladding integrity will be maintained.
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| * For the infrequent increase in main steam flow event (with an additional single failure), limited fuel cladding degradation may occur. However, offsite and control room radiological dose consequences are bounded by those that may result from an IOSGADVLOP event, and are in compliance with regulatory guidelines (see UFSAR Section 15.1.4) 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR ATMOSPHERIC DUMP VALVE 15.1.4.1 Identification of Causes and Frequency Classification An IOSGADV is postulated to occur as a result of an inadvertent operator action or equipment malfunction in the valve control system.
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| An IOSGADV event is classified as an incident of moderate frequency. An IOSGADV event in combination with an additional single failure is classified as an infrequent event.
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| June 2009 15.1-30 Revision 15
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| PVNGS UPDATED FSAR 15.1.4.2 Sequence of Events and System Operation The sequence of events and NSSS response to an IOSGADV event is similar to that described in UFSAR Section 15.1.3 for the increase in main steam flow event involving a sudden opening of the same or one or more SBCS valves. An IOSGADV event, however, would result in a smaller short-term increase in the main steam flow rate than the SBCS malfunction event which evaluates opening of one or more (up to eight) SBCS valves -
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| each with flow capacity equal to that of an ADV. Additionally, over the longer term, steam flow through the open ADV would not be halted automatically following an MSIS, and the affected SG would continue to discharge steam to the atmosphere, eventually drying out. Therefore, the short-term NSSS response to an IOSGADV event would occur more slowly than for the SBCS malfunction event, but long-term controlled heat removal would be achieved through one rather than both steam generators.
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| Table 15.1.4-1 provides the sequence of events for an IOSGADV event in which a single failure involving a LOP after turbine trip has been assumed. This sequence of events was obtained by simulating the event with the computer codes identified in UFSAR Sections 15.1.4.3 and 15.1.4.4. Figures 15.1.4-1 through 15.1.4-14 depict the response of key NSSS parameters during this event.
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| An inadvertent opening of a steam generator ADV causes the main steam flow rate to increase (Figure 15.1.4-1), which results in a decrease in RCS hot leg, cold leg, and average coolant temperatures (Figures 15.1.4-2, 15.1.4-3, and 15.1.4-4, respectively). In the presence of a negative MTC, the decrease in RCS temperature results in an increase in reactivity (Figure 15.1.4-5), core power (Figure 15.1.4-6), and core heat flux (Figure 15.1.4-7). The increase in main steam flow rate would June 2009 15.1-31 Revision 15
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| PVNGS UPDATED FSAR likewise cause a short-term increase in the indicated steam generator water level due to swell effects (Figure 15.1.4-8) and a decrease in both RCS and steam generator pressures (Figures 15.1.4-9 and 15.1.4-10, respectively). Despite the short-term increase in indicated steam generator water level, June 2009 15.1-32 Revision 15
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| PVNGS UPDATED FSAR Table 15.1.4-1 SEQUENCE OF EVENTS FOR THE IOSGADVLOP SAFETY ANALYSIS Time Event 0.0 Inadvertent opening of an SG ADV 187.97 Steam generator pressure reaches reactor trip setpoint 187.97 Steam generator pressure reaches MSIS setpoint 187.97 Turbine trip occurs 187.97 LOP occurs 189.12 Reactor trip breakers open 189.73 CEAs begin to fall 190.64 Minimum DNBR occurs 193.58 MSIVs close. Steam flow from unaffected steam generated halted 239.67 MSSVs open, unaffected SG 255.79 Pressurizer empties 269.74 MSSVs close, unaffected SG 270.46 Pressurizer pressure reaches SIAS setpoint 310.48 HPSI pumps begin injecting into the RCS 1119.50 Affected SG dries out 1800.00 Plant operators close the open ADV on affected SG 1800.00 Plant operators take control of the plant June 2011 15.1-33 Revision 16
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| PVNGS UPDATED FSAR the mass of liquid in both steam generators would not change significantly during the first few minutes of the transient (Figure 15.1.4-11).
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| Detection of the event may be accomplished by a high reactor power alarm, an RCS low pressure alarm, or a steam generator low pressure alarm. Reactor trip may be initiated by the CPCs on VOPT or an approach to the DNBR SAFDL, or by the RPS on a low steam generator pressure trip. As the event proceeds, the steam generator pressure decreases and approaches the Low Steam Generator Pressure Trip (LSGPT). For negative values of MTC, reactor power also increases toward a CPC auxiliary trip, the Variable Overpower Trip (VOPT). A VOPT will occur before the LSGPT for more negative values of MTC; however, for less negative values of MTC, the power increase may not be sufficient to result in the VOPT before the steam generator pressure reaches the LSGPT. The most limiting analysis case, which yielded the lowest DNBR value, was found to occur when the VOPT and LSGPT trips occurred at approximately the same time. Table 15.1.4-1 reflects a trip due to low steam generator pressure, with an MSIS occurring simultaneously at the analysis low steam generator pressure setpoint.
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| A turbine trip would occur following the reactor trip. The IOSGADVLOP case considers that the electric grid becomes unstable and collapses upon the sudden loss of generating capacity from the affected unit. A LOP would occur following the turbine trip. At least three seconds of offsite power is anticipated following the turbine trip (see UFSAR Section 15.0), however, the safety analysis results presented in Table 15.1.4-1 conservatively model a simultaneous reactor trip, turbine trip, and LOP. This conservative assumption is tied to NRC acceptance of the fuel failure analysis convolution June 2011 15.1-34 Revision 16
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| PVNGS UPDATED FSAR methodology. All four RCPs would then begin to coast down as a result of the LOP cutting power to the RCP motors, and the RCS would transition from forced flow to natural circulation conditions.
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| Short-term cooling of the RCS would increase coolant density, causing pressurizer level to decrease and the pressurizer to temporarily empty (Figure 15.1.4-12). The short-term increase in core heat flux and decrease in RCS pressure would, however, effectively reduce coolant subcooling and thereby cause the hot channel DNBR to decrease. DNBR would be degraded further in the first few seconds following the LOP, due to the rapid decrease in RCS flow rate as the RCPs coast down (Figure 15.1.4-13). CEAs will also fall into the reactor core following the reactor trip, rapidly reducing reactivity, core power, and core heat flux. As indicated in Table 15.1.4-1, minimum DNBR is predicted to occur shortly after the CEAs begin inserting into the core.
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| The MSIS would result in closure of the MSIVs, thereby halting the flow of steam from the unaffected or intact steam generator, which serves to maintain secondary system inventory.
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| The MSIS would also result in closure of the MFIVs, stopping the flow of main feedwater to both steam generators (Figure 15.1.4-14). Because the RCS hot leg temperature would be higher than the temperature in the intact steam generator following the MSIS, heat transfer from the RCS would continue to that steam generator. Eventually, the intact steam generator would heat up and repressurize to the lift setting of the first bank of MSSVs. The affected steam generator, however, would continue to blow down and depressurize because its open ADV is located upstream of the MSIVs. The affected steam generator would therefore eventually dry out.
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| June 2011 15.1-35 Revision 16
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| PVNGS UPDATED FSAR The safety analysis predicts that AFW flow would not be automatically delivered to either steam generator. In the case of the unaffected steam generator, the safety analysis shows that closure of the MSIVs and FWIVs effectively preserves secondary system inventory, so that steam generators water level does not decrease to the AFAS setpoint within the first thirty minutes of the event sequence. Additionally, the analysis shows that AFW flow is not automatically delivered to the affected steam generator, even though its water level steadily decreases to the AFAS setpoint. This is due to an AFW Lockout that occurs when the affected steam generators pressure decreases significantly below the pressure in the unaffected steam generator. The AFW Lockout, that is based on the pressure difference between steam generators prevents addition of feedwater to the affected steam generator, and thus a loss of AFW inventory to the environment via the affected steam generator.
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| While blowdown continues through the open ADV on the affected steam generator, heat losses from the RCS to the secondary system would exceed the heat generation rate in the RCS which, after a reactor trip and LOP, would be limited to decay heat from radioactive isotopes in the core and heat released by metallic components and structures in the NSSS. Pressurizer pressure would therefore decrease to the SIAS setpoint. Safety injection flow would be delivered to the RCS, however, only when RCS pressure decreases below the shutoff head of the HPSI pumps.
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| Following dryout of the affected steam generator, heat losses from the RCS to the secondary system would decrease sharply, and the RCS would begin to heat up. RCS repressurization would occur due to safety injection flow as well as the heat released June 2011 15.1-36 Revision 16
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| PVNGS UPDATED FSAR by fission product decay and hot metal structures in the NSSS.
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| The rate of heat transfer to the unaffected steam generator would also decrease as the temperature difference between that steam generator and its associated RCS hot leg decreased. The safety analysis shows that both the PSVs and the first bank of MSSVs would eventually lift to relieve pressure in the NSSS and to provide additional energy removal from the primary and secondary systems.
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| Operator action is not credited in the IOSGADVLOP safety analysis until thirty minutes following event initiation. At that time, it is assumed that plant operators will take action to manually close the ADV and to initiate a controlled plant cooldown to SDC entry conditions.
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| 15.1.4.3 Core and System Performance A moderate frequency IOSGADV event would result in an RCS cooldown that is similar to, but slower than, that caused by an increase in main steam flow event involving the quick opening of eight SBCS valves (see UFSAR Section 15.1.3). However, due to the more limiting MTC value chosen for this event, the reactor trip is delayed until the SGLPT occurs. The results of the IOSGADV are very close to the SBCS malfunction event, even though that event results in a greater increase in main steam flow and a faster RCS cooldown.
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| For the infrequent IOSGADVLOP event, the parameter of concern is likewise the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distribution effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worst single failure June 2011 15.1-37 Revision 16
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| PVNGS UPDATED FSAR for this event is a LOP following a turbine trip, which would cause the RCPs to coast down and rapidly reduce the coolant flow rate. This event results in a NSSS response that is similar to, that caused by the increase in main steam flow event involving the quick opening of eight SBCS valves in combination with LOP (See UFSAR Section 15.1.3). However, even though the quick opening of eight SBCS valves results in more severe RCS cooldown and higher increase in power, the IOSGADVLOP event results in a more limiting minimum hot channel DNBR than other infrequent events involving an increase in heat removal. For this reason, the infrequent IOSGADV event (in combination with a single failure) bounds the infrequent event involving the quick opening of eight SBCS valves (in combination with single failure) with respect to the DNBR SAFDL.
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| 15.1.4.3.1 Mathematical Model The limiting infrequent event involving an increase in heat removal by the secondary system - an IOSGADVLOP - was analyzed with respect to RCS pressure boundary performance with the following mathematical models:
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| * The CENTS computer code was used to simulate the NSSS response, including the predicted time of reactor trip due to low steam generator pressure. The CENTS computer code is described in UFSAR Section 15.0.3.1.3.2 and in an NSSS (1)(2) vendor topical report.
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| * The CETOP-D computer code, which uses the CE-1 CHF correlation, was used to determine the initial DNBR POL conditions for this event (see UFSAR Section 15.1.4.4.2 June 2011 15.1-38 Revision 16
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| PVNGS UPDATED FSAR below). The CETOP-D computer code is described in UFSAR (5)(6)(7)
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| Section 4.4 and in NSSS vendor topical reports.
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| 15.1.4.3.2 Input Parameters and Initial Conditions Table 15.1.4-2 and 15.1.4-3 summarizes the key input parameters and initial conditions utilized in the safety analysis for an IOSGADV and IOSGADVLOP event. The following points serve to explain the selection of initial conditions as they appear in Table 15.1.4-2:
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| * During normal power operations, the Core Operating Limits Supervisory System (COLSS) monitors various parameters to assist operators in maintaining plant conditions within the Technical Specification LCOs. When COLSS is in-service, it continuously calculates the core power at which the DNBR SAFDL would be reached, based on the measured temperature, pressure, flow, radial peaking factor, and axial power distribution. This COLSS-calculated core power is then divided by a numerical value, called the Required Overpower Margin (ROPM), to yield a DNBR Power Operating Limit (POL). If the measured core power exceeds the calculated POL, a COLSS alarm would alert plant operators to take action as required by Technical Specifications. The DNBR POL therefore serves to preserve thermal margin to accommodate potential operational transients. For this safety analysis, it was assumed that the plant would be operating at a POL condition immediately prior to the June 2011 15.1-39 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.4-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE IOSGADV SAFETY ANALYSIS Parameter Value Initial Core Power (% of RTP) 100%
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| Initial Core Inlet Temperature (°F) 566 Initial Pressurizer Pressure (psia) 2100 Initial RCS Flow Rate (% of Design Rated) 110.4 Initial Pressurizer Water Level (% Narrow Range) 24 Initial Steam Generator Water Level (% Wide Range) 86 4
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| Moderator Temperature Coefficient (/°F) -0.20x10-Doppler Fuel Temperature Coefficient Least Negative Delayed Neutron Kinetics BOC Axial Shape Index for DNBR Calculation -0.2 CEA Worth at Trip (%) 8.0 Fuel Rod Gap Conductance (BTU/hr-ft -°F) 2 6100 Total Number of Plugged Steam Generator Tubes 0 Single Failure None event, at the most limiting RCS flow rate. Because the event simulation was initiated from a calculated POL, additional power measurement uncertainties were not added to the initial assumed core power level. Analytical values for the initial core inlet temperature, pressurizer pressure, and RCS coolant flow rate June 2011 15.1-40 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.4-3 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE IOSGADV SAFETY ANALYSIS Parameter Value Initial Core Power (% of RTP) 100%
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| Initial Core Inlet Temperature (°F) 566 Initial Pressurizer Pressure (psia) 2100 Initial RCS Flow Rate (% of Design Rated) 110.4 Initial Pressurizer Water Level (% Narrow Range) 24 Initial Steam Generator Water Level (% Wide Range) 86 Moderator Temperature Coefficient (/°F) -0.40x10-4 Doppler Fuel Temperature Coefficient Least Negative Delayed Neutron Kinetics BOC Axial Shape Index for DNBR Calculation -0.2 CEA Worth at Trip (%) 8.0 Fuel Rod Gap Conductance (BTU/hr-ft -°F) 2 662 Total Number of Plugged Steam Generator Tubes 0 Single Failure LOP June 2011 15.1-41 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM corresponding to the POL were determined with the CETOP-D computer code, for an initial core power of 100% of RTP.
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| * Parametric evaluation determined that the DNBR degradation during the event is sensitive to the initial pressurizer water level, and the minimum pressurizer water level was used.
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| * Parametric evaluation determined that the DNBR degradation during the event is sensitive to the initial steam generator water level, and the maximum initial steam generator water level was used. The CENTS code automatically calculates the initial steam generator pressure, given steam generator water level, reactor power, and other inputs to the computer code.
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| * The IOSGADVLOP event causes a slower cooldown of the RCS than the SBCS malfunction, but the LOP portion of the event causes a heatup of coolant in the core region as the flow rate decreases during RCP coastdown. Using the most negative MTC allowed by the Technical Specifications and COLR would result in the most rapid power increase and the greatest overshoot in power after the VOPT.
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| However, because the heat flux lags behind the power, it does not necessarily produce the greatest decrease in DNBR. As the MTC is made less negative, the power excursion is slowed down. This allows the heat flux to stay close to the power and it allows the VOPT to shift upwards a small amount. The heat flux at the time of minimum DNBR increases for this event as the MTC is made less negative and the DNBR decreases. As the MTC is made less and less negative, a point is reached at which the June 2011 15.1-42 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM LSGPT and the VOPT are reached concurrently. This MTC results in the lowest minimum DNBR.
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| * The least negative Doppler fuel temperature coefficient curve, at BOC, was assumed. Least negative values minimize the addition of negative reactivity caused by increasing fuel temperature.
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| * BOC values were chosen to model delayed neutron kinetics, based on the parametric study described above.
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| * If power generation in the core is shifted toward the bottom, the insertion of negative reactivity following reactor trip will be somewhat delayed until the CEAs have inserted farther into the core. The scram reactivity curve was therefore based on a positive ASI representing a bottom-peaked core. The time versus scram reactivity curve was adjusted to account for a 0.6-second CEA holding coil time delay following opening of the reactor trip breakers, and normalized to model 90% CEA insertion at 4.0 seconds after power is removed from CEDM coils (see UFSAR Section 3.9.4). However, for DNBR calculations, having power at the top of the core would be more limiting and the ASI for the DNBR calculations was based on a negative ASI.
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| * The CEA worth at trip represents the minimum scram worth for HFP conditions at BOC, assuming the most reactive CEA remains stuck out of the core following reactor trip.
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| This is more limiting (less negative) than the anticipated scram reactivity worth at other times during the operating cycle for HFP conditions.
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| June 2011 15.1-43 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * For the fuel rod gas gap conductance, a high value was selected for IOSGADV and a low value was selected for IOSGADVLOP, based on the parametric study described above.
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| * It was assumed that steam generator tubes were not plugged for this safety analysis, based on the parametric study described above.
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| * For the limiting infrequent event analysis, an additional single failure involving a LOP was assumed.
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| For those safety-related Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) setpoints and response times that had a direct effect on acceptance criteria for this event, analytical values were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| 15.1.4.3.3 Results The moderate frequency IOSGADV and infrequent IOSGADVLOP events would not challenge fuel pellet integrity. During these events, any short-term increase in reactor power, or three-dimensional shift in power generation within the core, would not be of sufficient magnitude to raise the linear heat rate above that required to cause fuel centerline melting.
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| The analysis shows that the calculated minimum DNBR for the IOSGADV event approaches a value of 1.40 at 138 seconds following event initiation. This value is greater than the DNBR SAFDL of 1.34.
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| June 2011 15.1-44 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM For the infrequent IOSGADVLOP event (i.e., an IOSGADV event with an additional single failure), limited fuel cladding degradation may occur. However, offsite radiological dose consequences will not exceed a small fraction, or 10%, of 10 CFR Part 100 guideline values. Likewise, control room dose consequences will not exceed the limits specified by GDC 19 of 10 CFR 50 Appendix A.
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| 15.1.4.4 RCS Pressure Boundary Barrier Performance The IOSGADV and IOSGADVLOP events, like the SBCS malfunction event described in UFSAR Section 15.1.3, are characterized by an initial cooldown of the primary and secondary systems, and decreasing RCS and steam generator pressures. However, because the affected steam generator would dry out even following an MSIS, heat removal after an IOSGADV or IOSGADVLOP event must be accomplished through only one intact steam generator, at least until plant operators take action to close the open ADV on the affected steam generator. Therefore, unlike the SBCS malfunction event, heat removal may not be sufficient to prevent repressurization of the RCS and the unaffected steam generators. For this reason, the IOSGADV is the limiting moderate frequency (and limiting infrequent, when combined with single failure) event involving an increase in heat removal by the secondary system with respect to RCS pressure boundary performance.
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| The performance of the RCS pressure boundary is evaluated herein for the limiting infrequent event involving an increase in heat removal by the secondary system with a single failure, an IOSGADVLOP. The conclusions, however, also apply to the moderate frequency IOSGADV event, because peak pressures in June 2011 15.1-45 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM the RCS and secondary systems are limited by PSV and MSSV setpoints and capacities, respectively.
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| 15.1.4.4.1 Mathematical Model The mathematical model is the same as described in Section 15.1.4.3.1.
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| 15.1.4.4.2 Input Parameters and Initial Conditions The input parameters and initial conditions are the same as described in Section 15.1.4.3.2.
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| 15.1.4.4.3 Results The IOSGADVLOP safety analysis shows that pressurizer pressure may increase to the maximum PSV lift setting allowed by the Technical Specifications around the end of the 30-minute simulation. Immediately prior to the PSV lift the peak RCS pressure is calculated to be less than 2560 psia, for both RTPs which is below the regulatory limit for this event (i.e., 110%
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| of the RCS design pressure of 2500 psia, or 2750 psia).
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| The IOSGADVLOP safety analysis, like the SBCS malfunction analysis described in UFSAR Section 15.1.3, also shows that the peak secondary system pressure occurs when the first bank of MSSVs cycle open. The calculated peak secondary system pressure is less than 1305 psia, which is below the regulatory limit for this event (i.e., 110% of the steam generator shell side design pressure of 1270 psia, or 1397 psia).
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| Therefore, the peak RCS and secondary system pressures will not pose a challenge to the RCS pressure boundary, for the limiting moderate frequency and infrequent event, with respect June 2011 15.1-46 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM to pressure limits, involving an increase in heat removal by the secondary system.
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| 15.1.4.5 Containment Performance and Radiological Consequences An inadvertent opening of a steam generator ADV in combination with an additional single failure (i.e., LOP following turbine trip) is classified as an infrequent event. Offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. Additionally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A.
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| Control room radiological assessments for bounding unfiltered inleakage are presented in UFSAR Section 6.4.7. The results presented in that UFSAR section for a postulated Reactor Coolant Pump (RCP) sheared shaft event with a stuck open ADV bound the anticipated control room exposure for the IOSGADVLOP event. The RCP sheared shaft event is predicted to result in a higher percentage of fuel failure than the IOSGADVLOP event which, in combination with a stuck open ADV, would result in a correspondingly higher control room dose than the IOSGADVLOP event.
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| The offsite radiological dose consequences associated with the infrequent IOSGADVLOP event are evaluated in the following UFSAR subsections.
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| 15.1.4.5.1 Mathematical Models For the offsite radiological dose assessment, activity in the RCS is calculated on the basis of the pre-event radioiodine and noble gas activity levels (which are limited by plant Technical June 2011 15.1-47 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Specifications), to which is added the anticipated post-event increase in activity levels due to fuel pin failures. The increase in activity levels due to fuel pin failures is dependent upon the radial peaking factor, which affects the radionuclide inventory in the fuel rod gas gap, as well as the fuel failure fraction, which defines the number of pins that are assumed to release radionuclides to the RCS coolant.
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| Once the activity level in the RCS is determined, the amount of activity carried over to the steam generators by primary-to-secondary leakage is calculated. All of the activity that is contained in or leaked to the affected steam generator within the first 30 minutes of the event is assumed to be released to the atmosphere. At 30 minutes, analytical credit is taken for plant operators closing the open ADV on the affected steam generator, thereby halting the release of radionuclides from that steam generator to the environment.
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| Activity that leaks into the unaffected steam generator is assumed to mix with that steam generators secondary inventory.
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| The level of activity in this generator will therefore increase as the event proceeds. After the operators close the ADV on the affected steam generator, they may begin a controlled cooldown using the unaffected steam generator. The activity released from the unaffected steam generator to the environment may then be determined, based on a steaming rate that will remove decay heat and successfully cool down the NSSS to SDC entry conditions.
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| Based on the activity releases, the thyroid and whole body doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) are calculated as a function of the product of the radial peaking factor (Fr) and fuel failure fraction (FF). The production of Fr and FF that just corresponds to the acceptance June 2011 15.1-48 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM limits, that is a small fraction (10%) of 10 CFR Part 100 guideline values, is calculated from this functional relationship. As long as this calculated product for a reload does not exceed the value corresponding to the acceptance limits, the calculated doses for that reload will not exceed the acceptance limits.
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| 15.1.4.5.2 Input Parameters and Initial Conditions Offsite radiological dose consequences associated with the IOSGADVLOP event were analyzed under the following conditions:
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| : 1. Isotope inventories were based on a core power level of 4070 MWt, or 102% of the RTP of 3990 MWt.
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| : 2. Based on Technical Specification limits, the initial assumed contamination in the NSSS was:
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| * RCS Dose Equivalent (DEQ) I-131: 1.0 µCi/gm
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| * RCS Noble Gas: 100/ µCi/gm
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| * Secondary System DEQ I-131: 0.10 µCi/gm Where is the average of the sum of the average beta and gamma energies per disintegration (in units of MeV), for noble gas isotopes with half lives greater than 15 minutes, weighted in proportion to the concentration of each isotope in the reactor coolant. Tables 15.1.4-3 and 15.1.4-4, respectively, identify the initial RCS iodine and noble gas source terms for the radioisotopes that were included in the analysis. Radioiodine dose (16) conversion factors were set to the ICRP-30 values listed in UFSAR Table 15B-4, and the average beta () and gamma () disintegration energies for each noble gas June 2011 15.1-49 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM isotope were set to the values listed in UFSAR Table 15B-1.
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| : 3. An RCS liquid mass of 555,000 lbm of water was used in the analysis, including 45,000 lbm of water in the pressurizer. Additionally, 4,500 lbm of steam was assumed to be in the pressurizer. Although the RCS may hold more mass, these values were selected to increase the iodine concentration following postulated fuel failures, which conservatively increases offsite dose consequences.Since the PSVs lift for this event, the dose calculation conservatively takes into account the activity released to containment, even though the Reactor Drain Tank is sized to remain intact from the PSV discharge.
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| : 4. A steam generator liquid mass of 160,600 lbm per steam generator was used in the analysis. Although the steam generators may hold more mass, this value was selected to increase the iodine concentration in the affected steam generators, which conservatively increases offsite dose consequences.
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| : 5. A primary-to-secondary leak rate of 0.5 gpm (720 gallons per day) per steam generator was assumed. This is consistent with the PVNGS Technical Specification limit for RCS leakage prior to issuance of Operating License Amendment No. 120(15)
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| : 6. It was assumed that 10% of the iodine and noble gas inventories in the fuel pins were resident in the fuel rod gas gap, and available for release upon clad rupture.
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| June 2011 15.1-50 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 7. All of the activity in the fuel rod gas gap was assumed to be released to the RCS coolant upon fuel pin failure.
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| : 8. All of the iodines associated with leakage to the affected steam generator were assumed to be released to the environment i.e. with a determination factor of 1.0, until the open ADV was closed at 30 minutes into the event sequence. After 30 minutes, the affected steam generator did not contribute further to radiological releases, because all subsequent steaming associated with decay heat removal and controlled plant cooldown were assumed to occur through the unaffected steam generator.
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| : 9. Iodines associated with leakage to the unaffected steam generator are released to the environment during steaming with decontamination factor of 100 since the steam generator inventory, i.e. level, is maintained.
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| : 10. It was assumed that plant operators would not initiate a controlled plant cooldown to SDC entry conditions for at least 30 minutes following event initiation.
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| However, it should be noted that a faster RCS cooldown rate would increase steam releases during the first two hours following the event, which would produce more severe thyroid doses at the EAB. On the other hand, a slower RCS cooldown rate would allow radionuclide concentrations to build up in the secondary system, which would produce more severe 8-hour doses at the LPZ. Therefore, radiological dose calculations were performed using two different cooldown rates:
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| June 2011 15.1-51 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * A maximum Technical Specification cooldown rate of 100 F/hr, initiated at 30 minutes into the event o
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| sequence.
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| * A slower cooldown rate of 40oF/hr, initiated at 30 minutes into the event sequence, which would bring the RCS to SDC entry conditions at approximately 8 hours following event initiation.
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| : 11. Decay heat during the 8-hour period following the event was based on the 1979 ANS decay heat curve, with a 2 uncertainty. Use of a maximum decay heat curve increases the amount of steam released to the environment, thereby resulting in more severe dose consequences.
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| : 12. Although the LOP would cause the RCPs to coast down during an IOSGADVLOP event, it was conservatively assumed that all four RCPs would remain in operation for the radiological dose analysis. Therefore, 26 MWt of RCP heat was added to the 2-hour EAB and 8-hour LPZ dose calculations, which conservatively increased steam releases and offsite doses during the controlled cooldown.
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| : 13. A value of 740,000 BTU/oF was used to represent the specific heat capacity of the RCS, the RCS clad, and the steam generators. Use of this value increases the amount of steam that must be released to the environment during the controlled cooldown.
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| : 14. The /Q atmospheric dispersion factors used in the analysis are the short-term factors shown in UFSAR Table 2.3-31.
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| June 2011 15.1-52 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 15. Although the results of the transient simulation of IOSGADVLOP shows the DNBR remained above SAFDL and thus no fuel pin failures occurred, the radiological dose analysis conservatively assumes a fuel failure fraction of 5.5%.
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| : 16. A radial peaking factor of 1.72, corresponding to the maximum allowable radial peaking factors for PVNGS cores, was used in the analysis.
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| Table 15.1.4-4 RCS IODINE SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Term Isotope (Ci/MWt)
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| I-131 25,100 I-132 38,100 I-133 56,220 I-134 65,760 I-135 51,040 June 2011 15.1-53 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.4-5 RCS NOBLE GAS SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Term Isotope (Ci/MWt)
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| Kr-83m 4,153 Kr-85 440 Kr-85m 13,000 Kr-87 21,540 Kr-88 32,020 Xe-131m 260 Xe-133 56,220 Xe-133m 1,384 Xe-135 53,640 Xe-135m 18,200 Xe-138 49,700 June 2011 15.1-54 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.4.5.3 Results An IOSGADV in combination with an additional single failure (a LOP following turbine trip) is classified as an infrequent event, which may result in limited fuel cladding degradation.
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| Offsite radiological dose consequences are limited to a small fraction or 10%, of 10 CFR Part 100 guideline values.
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| Therefore the radiological limits for the limiting infrequent increase in heat removal by the secondary system event, IOSGADVLOP are 30 Rem for the thyroid and 2.5 Rem for the whole body. Additionally, radiation exposures for control room personnel are subject to the limits specified in GDC 19 of 10 CFR 50 Appendix A.
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| The radiological dose analysis conservatively assumed 5.5% fuel failure in the bounding analysis for dose consequences. The results of the IOSGADVLOP radiological dose analysis for this assumed percentage of failed fuel are shown in Table 15.1.4-5.
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| Table 15.1.4-6 OFFSITE RADIOLOGICAL DOSES FOR IOSGADVLOP SAFETY ANALYSES Thyroid Dose (REM) Whole Body Dose (REM) 0-2 Hour EAD 0-8 Hour LPZ 0-2 Hour EAB 0-8 Hour LPZ 25.1 12.6 1.1 1.1 The dose consequences remain below the acceptance criteria for the IOSGADVLOP event. These results bound core power levels of 3990 MWt and less.
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| June 2011 15.1-55 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.4.6 Conclusion Evaluation of the IOSGADV event shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| * For the moderate frequency IOSGADV event (without an additional single failure), fuel cladding integrity will be maintained.
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| * For the infrequent IOSGADVLOP event (i.e., an IOSGADV event with an additional single failure), limited fuel cladding degradation may occur. However, offsite radiological dose consequences will not exceed a small fraction, or 10%, of 10 CFR Part 100 guideline values.
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| Likewise, control room dose consequences will not exceed the limits specified by GDC 19 of 10 CFR 50 Appendix A.
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| 15.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - OPERATING MODES 1 AND 2 15.1.5.1 Identification of Causes and Frequency Classification A Main Steam Line Break (MSLB) is a postulated break or rupture of a pipe in the main steam system, either inside or outside the containment building.
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| A MSLB is classified as a limiting fault. Protection by design is therefore provided for MSLBs, up to and including the complete severance of a Seismic Category I main steam line June 2011 15.1-56 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM upstream of the containment isolation valves (i.e., Main Steam Isolation Valves).
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| 15.1.5.2 Sequence of Events and System Operation A MSLB is characterized as a cooldown event, because the blowdown of main steam through a pipe break would result in excessive energy removal from the NSSS and a power-to-load mismatch. Additionally, if the MSLB occurred upstream of a Main Steam Isolation Valve (MSIV), the affected steam generator would continue to blow down and dry out following a Main Steam Isolation Signal (MSIS). Long-term controlled heat removal must then be accomplished through the remaining unaffected steam generator.
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| The largest possible MSLB is a double-ended guillotine rupture of a main steam line upstream of an MSIV. The PVNGS steam lines, however, have integral venturi flow restrictors installed in the outlet nozzles of both steam generators. The maximum steam blowdown rate is therefore limited by the cross-sectional throat area of each flow restrictor, which is approximately 1.283 ft2.
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| Two types of analyses are performed for postulated MSLBs that may occur during operating Modes 1 (Power Operation) and 2 (Startup). MSLBs are analyzed for that portion of the accident immediately prior to and during reactor trip, when CEAs begin to fall into the core (henceforth referred to as the pre-trip phase), as well as for that portion of the accident following CEA insertion, when continued cooldown of the NSSS causes moderator density to increase and the reactor again approaches criticality (henceforth referred to as the post-trip phase).
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| There is a greater potential for fuel damage during the pre-June 2011 15.1-57 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM trip phase of a MSLB than during the post-trip phase, because post-trip fission power levels are sufficiently low to prevent a significant degradation in fuel performance. For this reason, the pre-trip analyses are performed for limiting Hot Full Power (HFP) initial conditions. The post-trip analyses, however, are performed for both HFP and Hot Zero Power (HZP) initial conditions, to assess the potential for fuel damage as a result of a Return-to-Power (R-t-P) following postulated MSLBs inside containment.
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| The PVNGS MSLB analyses therefore cover a wide range of initial conditions in Modes 1 and 2. These analyses, which are described in further detail below, are as follows:
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| A. Pre-trip analyses that maximize the potential for a short-term power excursion, a decrease in the hot channel minimum DNBR value, and radiological consequences:
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| : 1. SLB Case - A MSLB outside containment at HFP, in combination with a stuck CEA and with offsite power available. Credit is taken for a Core Protection Calculator (CPC) auxiliary trip, the Variable Over Power Trip (VOPT). There are no credible single failures that might occur during the pre-trip phase of the accident to enhance the power excursion or degrade thermal margin.
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| : 2. SLBLOP Case - A MSLB outside containment at HFP, in combination with a stuck CEA and a coincident Loss of Offsite Power (LOP). Credit is taken for a low RCP shaft speed trip. There are no credible single failures that might occur during the pre-June 2011 15.1-58 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM trip phase of the accident to enhance the power excursion or degrade thermal margin.
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| B. Post-trip analyses that maximize the potential for a R-t-P:
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| : 1. SLBFP Case - A MSLB inside containment at HFP with offsite power available, in combination with a stuck CEA and a single failure of a High Pressure Safety Injection (HPSI) pump. Credit is taken for a CPC VOPT.
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| : 2. SLBFPLOP Case - A MSLB inside containment at HFP with a coincident LOP, in combination with a stuck CEA and a single failure of a HPSI pump. Credit is taken for a low RCP shaft speed trip.
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| : 3. SLBZP Case - A MSLB inside containment at HZP with offsite power available, in combination with a stuck CEA and a single failure of a HPSI pump.
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| Credit is taken for a CPC VOPT.
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| : 4. SLBZPLOP Case - A MSLB inside containment at HZP with a coincident LOP, in combination with a stuck CEA and a single failure of a HPSI pump. Credit is taken for a low RCP shaft speed trip.
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| Detailed analysis of the two pre-trip cases reveals that the SLB case is the limiting pre-trip MSLB safety analysis for PVNGS. This case yields the highest peak power excursion and the lowest hot channel minimum DNBR value, before the CEAs have completed their fall into the reactor core. Likewise, detailed analysis of the four post-trip cases reveals that the SLBFPLOP case is the limiting post-trip MSLB safety analysis for PVNGS.
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| Of the four post-trip cases, the SLBFPLOP yields both the June 2011 15.1-59 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM highest post-trip reactivity value and the highest post-trip fission power.
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| Table 15.1.5-1 SEQUENCE OF EVENTS FOR THE LIMITING PRE-TRIP MSLB SAFETY ANALYSIS (SLB CASE)
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| Time (seconds)
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| RTP 3990 MWt Event 0.00 Double-ended guillotine MSLB (SG#1)occurs outside containment 3.62 CPC VOPT setpoint reached 4.37 Reactor trip breakers open 4.37 Turbine trip occurs 4.98 CEAs begin to fall 5.03 Steam generator level reaches MSIS setpoint 5.37 Peak power occurs 6.00 Hot channel minimum DNBR occurs The sequences of events for the limiting SLB (pre-trip) and SLBFPLOP (post-trip) cases are provided in Tables 15.1.5-1 and 15.1.5-2, respectively. These sequences were obtained by simulating the MSLB events with the computer codes identified in UFSAR Section 15.1.5.3.
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| Because the sequence of events and timing for the pre-trip SLB case is very similar to that provided in UFSAR Section 15.1.3 for the SBCS malfunction safety analysis, figures that depict the short-term response of key NSSS parameters are not provided here for the SLB analysis. However, Figures 15.1.5-1 through 15.1.5-14 are provided to depict the NSSS response for the longer-term SLBFPLOP post-trip analysis.
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| June 2011 15.1-60 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Like the SBCS malfunction and IOSGADV events, a MSLB causes the main steam flow rate to rapidly increase (Figure 15.1.5-1),
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| which results in a power-to-load mismatch and a decrease in RCS temperature (Figure 15.1.5-2). In the presence of a negative MTC, the decrease in RCS temperature results in a short-term increase in reactivity (Figure 15.1.5-3), core power (Figure 15.1.5-4), and core heat flux (Figure 15.1.5-5). The rapid cooldown will also result in an initial decrease in RCS pressure (Figure 15.1.5-6). Blowdown of the affected steam generator results in an initial decrease in steam generator pressures and levels (Figure 15.1.5-7 and 15.1.5-8, respectively) although a very short-term increase in the indicated steam generator water level due to swell effects is also anticipated (Figure 15.1.5-8) which may trigger a MSIS on high steam generator level. Upon MSIS, the MSIVs would close, halting the steam release from the unaffected steam generator.
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| As a consequence, decrease in the pressure and inventory of the unaffected steam generator could stop, while the affected steam generator pressure and level continue to decrease.
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| June 2011 15.1-61 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.5-2 SEQUENCE OF EVENTS FOR THE LIMITING POST-TRIP MSLB SAFETY ANALYSIS (SLBFPLOP CASE)
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| Time Event (seconds) 0.00 Double-ended guillotine MSLB (SG #1) occurs inside containment 0.00 LOP occurs 0.00 RCPs begin to coast down 0.00 SG level reaches MSIS setpoint and FWIVs close 0.65 RCP shaft speed reaches CPC auxiliary trip setpoint 0.95 Reactor trip breakers open 1.55 CEAs begin to fall 5.62 MSIVs closed. Steam flow from steam generator No. 2 halted 13.72 SG Differential Pressure AFW Lockout occurs 19.80 AFW actuation and delivery to SG #2 69.97 SIAS occurs due to low pressurizer pressure 79.05 Pressurizer empties 88.23 Void begins to form in reactor vessel upper head 89.97 One HPSI pump begins injecting into the RCS 171.53 Safety injection boron reaches RCS cold legs 251.01 AFAS cutoff setpoint is reached in SG #2 and AFW flow is terminated 295.41 Maximum post-trip reactivity occurs 341.01 Time of Maximum Return to Power 341.41 MacBeth minimum DNBR occurs 360.81 Steam generator No. 1 dries out 1800 Plant operators take control of the plant June 2011 15.1-62 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Detection of a MSLB may be accomplished by a high reactor power alarm, an RCS or steam generator low pressure alarm, recognition of a power-to-load mismatch, or a high containment pressure alarm (if the MSLB occurs inside containment).
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| The PVNGS MSLB safety analyses credit the CPC VOPT or the RPS Low SG Pressure Trip for the case when offsite power is available (e.g., SLB case), and a CPC low RCP shaft speed trip when a coincident LOP is postulated to occur (e.g., SLBFPLOP case). For the limiting post-trip SLBFPLOP case, the coolant flow rate through the RCS would decrease rapidly following the LOP, as the RCPs coast down and the RCS transitions from forced flow to natural circulation conditions (Figure 15.1.5-9).
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| As explained in UFSAR Section 15.1.5.3 below, assessment of fuel performance degradation differs between the pre-trip and post-trip MSLB safety analyses. For pre-trip cases, consideration is given to the short-term increase in core heat flux and decrease in RCS pressure, which would effectively reduce coolant subcooling and thereby cause the hot channel DNBR to decrease. Also, for pre-trip analyses, the minimum DNBR is predicted to occur as CEAs are falling into the reactor core during the reactor trip. For these pre-trip cases, the minimum DNBR which is computed using the CE-1 Critical Heat Flux (CHF) correlation is compared to the DNBR SAFDL which is based on a statistical combination of uncertainties methodology (see UFSAR Section 4.4.2.2). Table 15.1.5-1 shows that the hot channel minimum DNBR occurs very early in the limiting SLB event sequence.
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| For post-trip analyses, however, fuel degradation cannot be assessed in the same manner, because the applicable range of the CE-1 CHF correlation does not extend to the low RCS flow June 2011 15.1-63 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM rates and low pressures that may occur following a reactor trip. Therefore, as explained in UFSAR Section 15.1.5.3, post-(17)(18) trip analyses utilize the Macbeth CHF correlation to calculate a minimum DNBR value that occurs well after the CEAs have reached the bottom of the core. Table 15.1.5-2 and Figure 15.1.5-14 show that the Macbeth minimum DNBR occurs several minutes into the SLBFPLOP event sequence. Unlike DNBR values calculated with the CE-1 CHF correlation, Macbeth DNBR values are compared to a deterministic limit of 1.30 rather than a statistical one.
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| Although a coincident LOP and loss of forced flow through the RCS may result in higher coolant temperatures in the vicinity of the core, overall the RCS would still continue to cool down while the faulted steam generator dried out. This cooldown would increase RCS coolant density, causing the pressurizer to temporarily empty (Figure 15.1.5-10) and a void to form in the reactor vessel upper head (Figure 15.1.5-11). Pressurizer pressure would eventually decrease to the SIAS setpoint, actuating the safety injection pumps. Even if one HPSI pump failed to start on demand, sufficient safety injection flow would still be delivered to the RCS after the RCS pressure decreased below the shutoff head of the remaining operable HPSI pump (Figure 15.1.5-12). Safety injection flow into the RCS would serve not only to repressurize the system and provide inventory control, but would also deliver soluble boron that would add negative reactivity and slow down an approach to criticality or R-t-P.
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| Low pressure in the affected steam generator would eventually result in an MSIS and closure of the MFIVs, stopping the flow of main feedwater to both steam generators (Figure 15.1.5-13),
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| June 2011 15.1-64 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM and in closure of the MSIVs, thereby halting the flow of steam from the unaffected steam generator, which serves to maintain secondary system inventory.
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| Table 15.1.5-2 indicates that MSIS occurs at time = 0.0 in the event. This is due to the choice of initial steam generator water level. Since the MSIS occurs at time = 0.0, the unaffected SG is isolated from the break and pressurizes to the MSSV setpoints early in the transient. The lockout function prevents the addition of feedwater to the affected steam generator, and thus an unwanted loss of AFW inventory to the environment through the affected steam generator. However, due to the conservative choice of AFAS setpoint, AFAS is actuated in the unaffected SG. AFW continues to supply the unaffected steam generator as needed during the transient. Following dryout of the affected steam generator, decay heat from fission products in the core, and heat released by the hot metal structures of the NSSS, would raise temperatures and repressurize both the RCS and the intact steam generator.
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| However, due to the additional cooling provided by the auxiliary feedwater, neither the main steam system or reactor coolant system pressure boundaries are challenged later in the event sequence.
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| June 2011 15.1-65 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Operator action is not credited in the MSLB safety analyses until 30 minutes following event initiation. At that time, it is assumed that plant operators will take action to initiate a controlled plant cooldown to SDC entry conditions, for example by manually establishing AFW flow and a steaming path through the ADVs associated with the unaffected steam generator.
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| 15.1.5.3 Core and System Performance Because MSLBs result in rapid reactivity insertions and power excursions, they are evaluated with respect to degradation in fuel performance, and the potential for post-trip criticality or R-t-P.
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| For pre-trip MSLB safety analyses, initial conditions are chosen to obtain the most adverse power excursion and fuel performance degradation. Because the hot channel minimum DNBR value provides a measure of fuel performance degradation, pre-trip analyses consider those parameters and conditions that would cause the greatest decrease in the local or hot channel DNBR, such as an increase in local heat flux, an increase in reactor coolant temperature, a decrease in reactor coolant flow rate, and a decrease in reactor coolant pressure.
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| Likewise, for post-trip MSLB safety analyses, initial conditions are chosen to maximize the potential for a R-t-P, as measured by the maximum post-trip reactivity value, the timing of the reactivity insertion, the duration of the reactivity peak, and the maximum post-trip fission power which then translated into minimum local or hot channel DNBR.
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| June 2011 15.1-66 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.5.3.1 Mathematical Models 15.1.5.3.1.1 Pre-Trip Safety Analyses The PVNGS pre-trip MSLB safety analyses utilized the following mathematical models:
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| * The CENTS computer code was used to simulate the NSSS transient response. The CENTS computer code is described in UFSAR Section 15.0.3.1.3.2 and in an NSSS vendor (1)(2) topical report.
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| * The FORTRAN CPC computer code was used to simulate CPC reactor trip functions. Predicted times for reactor trips, with appropriate delays for signal processing and opening of the reactor trip breakers, were utilized in CENTS code input. The CPCs are described in UFSAR Section 7.2, and associated algorithms and simulation (3)(4) code are described in NSSS vendor topical reports.
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| * The CETOP-D computer code, which uses the CE-1 CHF correlation, was used to calculate initial and transient DNBR values. CETOP-D was also used to determine initial Power Operating Limit (POL) conditions. The CETOP-D computer code is described in UFSAR Section 4.4 and in (5)(6)(7)
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| NSSS vendor topical reports.
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| * The TORC computer code, which uses the CE-1 CHF correlation, was used to calculate the minimum DNBR value at the time of minimum DNBR predicted by the CETOP-D code, if CETOP-D predicted a minimum DNBR value below the DNBR SAFDL. Because the models in the CETOP-D code are not as detailed as those in TORC, DNBR predictions from CETOP-D are typically adjusted by penalty factors to June 2011 15.1-67 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM ensure conservatism. Use of the more detailed TORC computer code reduces the need for penalty factors and provides a more accurate prediction of the DNBR value than the CETOP-D code. The TORC computer code is described in UFSAR Section 4.4 and in NSSS vendor topical (10)(11) reports.
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| 15.1.5.3.1.2 Post-Trip Safety Analyses The PVNGS post-trip MSLB safety analyses utilized the following mathematical models:
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| * The CENTS computer code was used to simulate the NSSS transient response. The CENTS computer code is described in UFSAR Section 15.0.3.1.3.2 and in an NSSS vendor (1)(2) topical report.
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| * The FORTRAN CPC computer code was used to simulate CPC reactor trip functions. The CPCs are described in UFSAR Section 7.2, and associated algorithms and simulation (3)(4) code are described in NSSS vendor topical reports.
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| * As noted above, the determination of DNBR for post-trip MSLB analyses requires methods that differ from those used for pre-trip analyses. This is because the verified range of the CE-1 CHF correlation, which is used in the CETOP-D and TORC computer codes, does not extend to lowpressures and low flow rates that may exist in the RCS following a reactor trip. Therefore, the Macbeth (17)(18) correlation is utilized to ascertain the margin to DNB during the post-trip phase. The Macbeth correlation calculates the CHF as a function of mass flux, inlet subcooling, system pressure, heated diameter, and channel June 2011 15.1-68 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM length. Use of a channel heat balance allows the correlation to be converted to a "local conditions" form, thereby allowing the CHF to be determined as a function of height in the hot channel. The effect of non-uniform axial heating may be incorporated by using the method applied by Lee in Reference 19. Because the CEA of greatest reactivity worth is assumed to remain out of the core following the reactor trip, the Macbeth DNBR calculations must account for the high localized power peak that may exist in the core post-trip. The post-trip analyses therefore utilize a maximum fission power dependent core peaking factor, FQ, that is a function of both fission power and coolant flow rate through the core. FQ values are therefore different for forced flow cases (i.e., offsite power available) than for cases in which the RCPs coast down (i.e., a LOP occurs). The Macbeth CHF correlation is also described in the CENTS (1) computer code topical report.
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| 15.1.5.3.2 Input Parameters and Initial Conditions 15.1.5.3.2.1 Pre-Trip Safety Analyses Table 15.1.5-3 summarizes the key input parameters and initial conditions utilized in the limiting pre-trip MSLB safety analyses, SLB case, which were selected to obtain the most adverse power excursion and fuel performance degradation. The following points serve to explain the selection of initial conditions as they appear in Table 15.1.5-3:
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| * For the SLB case, a CPC VOPT auxiliary reactor trip was credited. The initial core power was set to 95% of RTP for this case, thereby allowing the VOPT setpoint to June 2011 15.1-69 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM increase as core power rises early in the simulations.
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| This causes a slight delay in the reactor trip and enhances the initial power excursion. Because the same ROPM value is used at power levels 95% of RTP, initial thermal margin to the DNBR SAFDL is the same at 95% power as at 100% power.
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| * A maximum core inlet temperature was selected because it maximizes the average temperatures in the RCS coolant loops. Maximizing the initial RCS average loop temperature tends to maximize the initial steam generator pressure, and hence maximizes the cooldown rate and reactivity insertion following a MSLB. The values in Table 15.1.5-3 include instrument uncertainty.
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| * The pre-trip analyses are not sensitive to the initial pressurizer pressure. Therefore, a nominal value was used.
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| * A maximum RCS flow rate was used in the CENTS code.
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| CENTS code output was then passed to the CETOP-D code to perform transient DNBR calculations. CETOP-D DNBR calculations were initiated from a Power Operating Limit (POL) corresponding to the maximum RCS flow rate.
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| * The pre-trip analyses are not sensitive to either the initial pressurizer or steam generator water levels.
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| Therefore, nominal values were used.
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| * A MSLB causes a rapid cooldown of the RCS. Therefore, the most negative MTC allowed by the Technical Specifications and COLR was used to maximize the positive reactivity insertion caused by the cooldown.
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| June 2011 15.1-70 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.5-3 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING PRE-TRIP MAIN STEAM LINE BREAK (SLB CASE) SAFETY ANALYSES Assumed Value Parameter RTP 3990 MWt Initial Core Power (% of RTP) 95 Initial Core Inlet Temperature (°F) 566 (a)
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| Initial Pressurizer Pressure (psia) 2250 Initial RCS Flow Rate (% of Design Rated) 116 (a)
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| Initial Pressurizer Water Level (% Narrow Range) 52 (a)
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| Initial Steam Generator Water Level (% Wide 81 Range)
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| -4 Moderator Temperature Coefficient (/°F) -4.4x10 Doppler Fuel Temperature Coefficient BOC Delayed Neutron Kinetics EOC Axial Shape Index for Scram Curve +0.3 CEA Worth at Trip (%) -8.0 2
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| Fuel Rod Gap Conductance (BTU/hr-ft -°F) 6984 Number of Plugged Steam Generator Tubes (Total) 0 Break Size (ft2) 1.283 Loss of Offsite Power No (a) Nominal range values are used since the event is not sensitive to these parameters.
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| June 2011 15.1-71 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * The least negative Doppler fuel temperature coefficient curve, at Beginning of Cycle (BOC), was conservatively assumed. Least negative values minimize the addition of negative reactivity caused by increasing fuel temperature. Therefore, the reactor core may achieve a higher peak power and heat flux during the initial RCS cooldown.
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| * End of Cycle (EOC) values were chosen to model delayed neutron kinetics. EOC values serve to emphasize the initial power excursion by minimizing the effect of delayed neutrons on the rate of power increase.
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| * If power generation in the core is shifted toward the bottom, the insertion of negative reactivity during reactor trip will be delayed until the CEAs have inserted farther into the core. The scram reactivity curve was therefore based on a positive ASI representing a bottom-peaked core. The time versus scram reactivity curve was adjusted to account for a 0.6-second CEA holding coil time delay following opening of the reactor trip breakers, and normalized to model 90% CEA insertion at 4.0 seconds after power is removed from CEDM coils (see UFSAR Section 3.9.4).
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| * The CEA worth at trip represents the minimum scram worth for Hot Full Power (HFP) conditions at BOC, assuming the most reactive CEA remains stuck out of the core following reactor trip. This is more limiting (less negative) than the anticipated HFP scram reactivity worth at other times during the cycle, this is conservative.
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| June 2011 15.1-72 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * The fuel rod gas gap conductance value was selected so that energy from the fuel would quickly reach the surface of the fuel rod clad. This results in a higher heat flux which closely follows core power, and greater degradation of DNBR during the initial power excursion.
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| * It was assumed that steam generator tubes were not plugged for the pre-trip MSLB safety analyses. This enhances the initial rate of heat transfer from the RCS to the main steam system, which in turn enhances the initial RCS cooldown and maximizes the positive reactivity insertion due to the negative MTC.
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| Additionally, this enhances the decrease in RCS pressure during the cooldown, which serves to degrade DNBR.
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| * A large break size was assumed for the MSLB, with steam blowdown limited by the cross-sectional throat area of the flow restrictors in the outlet nozzles of both steam generators. A large break size maximizes the initial cooldown rate and resulting reactivity insertion.
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| * For the limiting case, a LOP was not assumed to occur, so the reactor trip would be delayed until the CPC VOPT auxiliary trip was received. There are no credible single failures (see UFSAR Table 15.0-0) that would serve to enhance the power excursion or degrade thermal margin during the first few seconds of the pre-trip MSLB simulations; therefore, an additional single failure was not postulated.
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| For those safety-related Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) setpoints and response times that had a direct effect on acceptance June 2011 15.1-73 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM criteria for this event, analytical values were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| 15.1.5.3.2.2 Post-Trip Safety Analyses Table 15.1.5-4 summarizes the key input parameters and initial conditions utilized in the PVNGS post-trip MSLB safety analyses. Because degradation in fuel performance during the post-trip phase of a MSLB can only occur if there is a R-t-P, analytical values were selected to maximize the potential for an approach to criticality or a R-T-P. As noted above, the magnitude of a R-t-P is primarily determined by the maximum post-trip reactivity value, the timing of the reactivity insertion, and the duration of the reactivity peak.
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| The following points serve to explain the selection of initial conditions as they appear in Table 15.1.5-4:
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| * A maximum initial core power, including a 2% power measurement uncertainty, was selected for the HFP cases.
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| Use of a maximum initial core power maximizes the initial core outlet temperature as well as the initial average temperature in the RCS coolant loops. Maximizing the initial core outlet temperature maximizes the initial energy stored in the water and metal of the upper head region of the reactor vessel, and also maximizes the saturation pressure of the liquid in this region.
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| Following a MSLB, when RCS pressure falls below the saturation pressure of the liquid in the upper head region, this stored energy will help vaporize liquid and thereby slow the rate at which the RCS pressure decreases.
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| June 2011 15.1-74 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Selecting the RCP Seal Leakage to be 0.0 gpm also slows the rate at which the RCS pressure decreases. This in turn tends to delay and reduce the rate of safety injection, which minimizes the negative reactivity due to boron at the time of R-t-P. Furthermore, maximizing the initial average temperature of the RCS coolant loops maximizes the initial steam generator pressure, which maximizes the blowdown and rate of energy removal following a MSLB. Increasing the rate of energy removal likewise increases the RCS core inlet temperature cooldown rate which, in the presence of a negative MTC, enhances the positive reactivity insertion due to the cooldown.
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| Maximizing the initial RCS average temperature also causes the cooldown to occur over a more adverse portion of the moderator reactivity function, i.e., the portion having the greatest rate of change of reactivity with temperature.
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| * Maximum initial core inlet temperatures were selected because they maximize the average temperatures in the RCS coolant loops. As explained above, maximizing the initial RCS average loop temperature tends to maximize the initial steam generator pressure, and hence maximizes the cooldown rate and reactivity insertion following a MSLB. The values shown in Table 15.1.5-4 reflect the maximum allowed RCS cold leg temperatures including instrument uncertainty.
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| * A high initial pressurizer pressure and pressurizer water level was selected. This increases transient RCS pressures, thereby delaying and impeding safety injection flow and the delivery of boron to the core region.
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| June 2011 15.1-75 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.5-4 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING POST-TRIP MAIN STEAM LINE BREAK (SLBFPLOP CASE) SAFETY ANALYSES Parameter Assumed Value Initial Core Power (% of RTP) 102 Initial Core Inlet Temperature (°F) 566 Initial Pressurizer Pressure (psia) 2325 Initial RCS Flow Rate (% of Design Rated) 95 Initial Pressurizer Water Level (%) 60 Initial Steam Generator Water Level (% Narrow Range) 96 Auxiliary Feedwater Actuation Setpoint (% Wide Range) 82 Auxiliary Feedwater Cutoff Setpoint (% Wide Range) 99
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| -4 Moderator Temperature Coefficient (/F) -4.4x10 Doppler Fuel Temperature Coefficient EOC Delayed Neutron Kinetics BOC Inverse Boron Worth (ppm/%) -130 Axial Shape Index for Scram Curve +0.6 CEA Worth at Trip (%) -8.75 2
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| Fuel Rod Gap Conductance (BTU/hr-ft -F) 656 Number of Plugged Steam Generator Tubes (Total) 2516 2
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| Break Size (ft ) 1.283 Loss of Offsite Power Yes Additional Single Failure One HPSI pump RCP Seal Leakage (gpm) 0.0 June 2011 15.1-76 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * A minimum initial core flow rate maximizes the rate of core inlet temperature cooldown following a MSLB, which in turn maximizes the reactivity insertion in the presence of a negative MTC cooldown curve.
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| * A high initial steam generator water level corresponding to the high SG level alarm setpoint was selected for both steam generators based on parametric analysis. This increases the cooldown rate following a MSLB. However, it will also result in an MSIS at time = 0.0 and FWIV closure. In addition, the use of a high initial water level in the unaffected SG could prevent level from decreasing to the AFAS low level setpoint during the event.
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| * MTC has a significant effect on the potential for a R-t-P following a MSLB, due to the magnitude of the positive reactivity that results from the RCS cooldown.
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| Therefore, the most negative MTC cooldown curve was used for this analysis, corresponding to End of Cycle (EOC) conditions. That is, because a loss of CEA reactivity worth may occur as the moderator becomes denser during the cooldown, reactivity is adjusted to account for the effects of changes in moderator density. Additionally, the moderator reactivity contribution to core power was based on cold-edge temperature, which is weighted to account for the colder water returning to the RCS from the faulted steam generator, rather than the core bulk or average temperature.
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| June 2011 15.1-77 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * An EOC (most negative) Doppler fuel temperature coefficients curve was used for the post-trip MSLB analyses. Use of a most negative curve adversely affects R-t-P by adding relatively more positive reactivity as the fuel goes from operating temperatures prior to the MSLB, to lower temperatures that occur during the post-MSLB cooldown.
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| * Relative to other parameters, the delayed neutron fraction has a minor effect on the potential for a R-t-P.
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| Beginning of Cycle (BOC) values were conservatively chosen because they result in relatively more delayed neutrons, which delay the rate of decrease in core power post-trip. This in turn extends the duration of a reactivity peak or R-t-P if it occurs.
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| * A maximum value was selected for the inverse boron worth, to reduce the negative reactivity that is inserted as a result of boron injected by the Safety Injection (SI) system. A sweep-out volume of 60.6 cubic feet was used for the SI lines, representing the volume of water that must be displaced before safety injection boron reaches the primary system.
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| June 2011 15.1-78 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * The rate at which negative reactivity is added by CEAs during a reactor trip has little effect on the potential for a post-trip R-t-P. Therefore, the initial axial power distribution is of little importance for post-trip MSLB safety analyses. However, the axial power distribution that was selected was bottom-peaked to delay the full effect of scram reactivity for as long as possible. Additionally, the time versus scram reactivity curve was adjusted to account for a 0.6-second CEA holding coil time delay following opening of the reactor trip breakers, and normalized to model 90% CEA insertion at 4.0 seconds after power is removed from CEDM coils (see UFSAR Section 3.9.4).
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| * The CEA worth at trip represents the minimum allowed scram worth at EOC, assuming the most reactive CEA remains stuck out of the core following the trip. The selection of EOC values is consistent with the selection of the MTC and Doppler values, i.e., they represent the same point in time in an operating cycle.
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| * The post-trip MSLB safety analyses utilized a fuel rod gas gap conductance value which results in a bounding core average effective fuel temperature at full power EOC conditions. A higher full power fuel temperature results in a larger reactivity addition as the core goes from HFP to HZP conditions, which results in a larger scram worth requirement to prevent return to power.
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| June 2011 15.1-79 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * It was assumed that steam generator tubes were plugged for the post-trip MSLB safety analyses. Although this would tend to reduce heat transfer from the RCS to the main steam system, which may reduce the initial RCS cooldown, a parametric evaluation showed that increased tube plugging has a more adverse effect during the post-trip phase. The initial RCS cooldown was maximized by conservatively assuming that the full steam generator heat transfer area would be maintained throughout blowdown of the faulted steam generator, rather than decreasing as steam generator water mass decreased. The heat transfer area was ramped down to zero only after the mass of liquid in the steam generator decreased below 100 lbm. Tube plugging therefore had more of a direct effect on post-trip RCS temperatures and flow rates, especially under natural circulation conditions, which in turn affected post-trip DNBR values and the potential for a R-t-P.
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| * A large break size was assumed for each post-trip MSLB analysis, with steam blowdown limited by the cross-sectional throat area of the flow restrictors in the outlet nozzles of both steam generators. A large break size maximizes the initial cooldown rate and resulting reactivity insertion.
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| June 2011 15.1-80 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * For the limiting case, a LOP was assumed to occur coincident with the MSLB. This assumption results in an early RCP coastdown, which affects both the post-trip minimum DNBR value and the timing and magnitude of a R-t-P. Although a lower core flow rate tends to result in a smaller R-t-P, a lower flow rate has more of a direct effect on the minimum DNBR value than does the magnitude of the R-t-P. Therefore, a LOP coincident with a MSLB yields the greatest potential for degradation in post-trip fuel performance.
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| * For the post-trip MSLB analyses, an additional single failure involving the failure of a HPSI pump to start on demand was assumed (see UFSAR Table 15.0-0). This single failure served to reduce the capacity of the SI system to provide boron to the core region of the RCS, which increased the potential for a R-t-P. Other postulated singles failures, including a single failure of a Main Steam Isolation Valve (MSIV) to close, were determined to be less limiting than the single failure of a HPSI pump.
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| For those safety-related Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESFAS) setpoints and response times that had a direct effect on acceptance criteria for this event, analytical values were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| June 2011 15.1-81 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.5.3.3 Results 15.1.5.3.3.1 Pre-Trip Safety Analyses The limiting pre-trip MSLB safety analysis, for a break that occurs outside containment without a coincident LOP (i.e., the SLB case), shows that core power reached a peak around 115% -
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| 116% of RTP, shortly after the CEAs begin to fall into the core. However, during the event, the short-term excursion in reactor power would not be of sufficient magnitude to raise the linear heat rate above that required to cause fuel centerline melting.
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| The limiting pre-trip MSLB safety analysis also shows that the hot channel minimum DNBR (computed with the CE-1 CHF correlation and CETOP code) was calculated to be 1.33 for 3990 MWt RTP cores. Since CETOP DNBR values are conservative relative to TORC DNBR values, cycle specific Thermal Hydraulic calculations are performed to confirm that the calculated DNBR value remains above the SAFDL. These cycle specific calculations also ensure that the radiological dose consequences presented in Section 15.1.5.5 for this event remain bounding.
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| 15.1.5.3.3.2 Post-Trip Safety Analyses The results of the limiting post-trip MSLB core performance safety analyses (SLBFPLOP) are summarized in Table 15.1.5-5.
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| June 2011 15.1-82 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The SLBFP, SLBZP, and SLBZPLOP cases were determined to be non-limiting with respect to acceptance criteria for post-trip MSLB core performance analyses. That is, Macbeth minimum DNBR values during the low flow, low pressure post-trip phase did not decrease below a deterministic limit of 1.30; post-trip fission power steadily decreased; and the maximum post-trip reactivity achieved in each case indicated that the reactor remained shut down after the CEAs had fully inserted into the core.
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| For the limiting SLBFPLOP case, however, post-trip fission power and reactivity initially decreased, and then began to increase again as the moderator continued to cool down. The maximum fission power and reactivity occur before the affected SG dries out. High pressure safety injection boron ensures that the reactor remains shut down and the minimum DNBR remains above the deterministic limit of 1.30 during the post-trip period.
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| The core linear heat rate during the post-trip phase is dependent upon the post-trip fission power, the decay heat released by radioactive isotopes in the core, and applicable peaking factors. For the limiting SLBFPLOP case, it was determined that the linear heat rate would remain well below that required to cause fuel centerline melting.
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| It is therefore concluded that fuel damage will not occur during the post-trip phase of a postulated MSLB.
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| June 2011 15.1-83 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.5-5 CORE PERFORMANCE SAFETY ANALYSIS RESULTS FOR THE LIMITING POST-TRIP MAIN STEAM LINE BREAK (SLBFPLOP CASE) SAFETY ANALYSES Parameter Results Macbeth Minimum DNBR 2.42 Time of Macbeth Minimum DNBR (sec) 341
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| -2 Maximum Post-Trip Fission Power (% RTP) 1.385x10 Time of Maximum Post-Trip Fission Power 341 (sec)
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| Maximum Post-Trip Reactivity (%) 0.02036 Time of Maximum Post-Trip Reactivity (sec) 295 June 2011 15.1-84 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.5.4 RCS Pressure Boundary Barrier Performance Postulated MSLBs during Modes 1 (Power Operation) and 2 (Startup), like the SBCS malfunction and IOSGADVLOP events described in UFSAR Sections 15.1.3 and 15.1.4, respectively, are characterized by an initial cooldown of the primary and secondary systems, and decreasing RCS and steam generator pressures. Additionally, like an IOSGADVLOP event, the affected steam generator would not be completely secured following an MSIS if the MSLB occurs upstream of an MSIV. If this were to occur, the affected steam generator would eventually dry out, and long-term heat removal would have to be accomplished through the unaffected steam generator.
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| Therefore, for a large MSLB upstream of an MSIV, long-term heat removal via the MSSVs may not be sufficient to prevent re-pressurization of the RCS to the lift setting of the PSVs (see UFSAR Section 15.1.4).
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| In addition to the long term repressurization, the RCS performance is investigated in the short-term for brittle fracture criterion because of low temperature and high pressure conditions that may occur simultaneously due to the rapid cooldown and the high pressure safety injection. This investigation merely consists of comparing the conditions observed during MSLB event with the low temperature overpressurization analyses detailed in UFSAR Chapter 5.
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| 15.1.5.4.1 Mathematical Models The mathematical models that were used to analyze the performance of the RCS pressure boundary are the same as those described in UFSAR Section 15.1.5.3.1.
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| June 2011 15.1-85 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.5.4.2 Input Parameters and Initial Conditions The key input parameters and initial conditions that were used to analyze the performance of the RCS pressure boundary are the same as those described in UFSAR Section 15.1.5.3.2.
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| 15.1.5.4.3 Results Figure 15.1.5.6 and 15.1.5.7 shows the RCS and SG pressure response for the limiting post-trip MSLB, a SLBFPLOP. Due to the early MSIS, the unaffected SG is isolated from the break.
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| Heat up and pressurization occurs, but is turned around by the addition of AFW to the unaffected SG. Later in the event sequence after the AFW cutoff setpoint is reached, the unaffected SG begins to heat up and repressurize again. The peak secondary pressure is limited by the MSSV setpoints, which is below the acceptable design limit (i.e., 110% of the steam generator shell side design pressure of 1270 psia, or 1397 psia). Well before the end of the post-trip MSLB simulation, RCS pressure turns around and begins to increase as a result of safety injection flow, decay heat, and heat released from the hot metal structures that comprise the NSSS. Auxiliary feedwater addition and cooling by the intact SG helps keep the RCS pressure below the PSV setpoint of 2450 psia during the 30 minute transient. Therefore, peak RCS pressure remains below the acceptable design limit for this event (i.e., 110% of the RCS design pressure of 2500 psia, or 2750 psia).
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| June 2011 15.1-86 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM During a MSLB event, combination of low temperature and high pressure does not challenge the RCS integrity due to the brittle fracture since the combination of temperature and pressure remain well within the reactor vessel design that is evaluated in UFSAR Chapter 5. Therefore, the peak RCS and secondary system pressures that may occur following a MSLB in Modes 1 and 2 will be maintained within acceptable design limits.
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| 15.1.5.5 Containment Performance and Radiological Consequences A MSLB is classified as a limiting fault, for which radiological dose consequences are subject to various regulatory limits.
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| Specifically, if fuel failure is postulated to occur, or if the MSLB is assumed to occur following an operational transient that has raised the RCS iodine concentration to the maximum value permitted by Technical Specifications (i.e., a Preaccident Iodine Spike, or PIS case), then offsite radiological doses must not exceed 10 CFR Part 100 guideline values. That is, 2-hour doses at the Exclusion Area Boundary (EAB) and 8-hour doses at the outer boundary of the Low Population Zone (LPZ) would be limited to a thyroid dose of 300 Rem and a whole body dose of 25 Rem. However, if the reactor trip or RCS depressurization following the MSLB is assumed to create an accident-Generated Iodine Spike (GIS) without fuel failure, then offsite dose consequences must not exceed a small fraction, or 10%, of 10 CFR Part 100 guideline values. Finally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A.
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| Control room radiological assessments for bounding unfiltered inleakage are presented in UFSAR Section 6.4.7. The limiting June 2011 15.1-87 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM cases presented in that UFSAR section bound the anticipated control room exposures for postulated MSLB events. For example, the results presented therein for a Steam Generator Tube Rupture (SGTR) with a stuck open ADV bound a MSLB with a PIS or GIS iodine spike. Likewise, a MSLB that is limited to 1% fuel failure is bounded by the RCP sheared shaft event with a stuck open ADV, because the sheared shaft event results in a higher percentage of fuel damage.
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| The offsite radiological dose consequences associated with limiting fault MSLBs are evaluated in the following subsections.
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| 15.1.5.5.1 Mathematical Models The offsite radiological consequences of postulated MSLBs are evaluated for breaks that may occur outside the containment building.
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| Activity in the RCS is calculated on the basis of initial radioiodine and noble gas activity levels, which are limited by plant Technical Specifications, to which is added the increase in activity due to fuel failure or iodine spikes. For postulated fuel failure, the increase in RCS activity is dependent upon the radial peaking factor, which affects the radionuclide inventory in the fuel rod gas gap, as well as the fuel failure fraction, which defines the number of pins that are assumed to release radionuclides to the RCS coolant. For PIS and GIS cases, the increase in RCS activity is determined by analytical iodine spiking factors.
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| Once the activity level in the RCS is determined, the amount of activity carried over to the secondary system by primary-to-secondary leakage is calculated. All of the activity that is June 2011 15.1-88 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM contained in or leaked to the affected steam generator is assumed to be released to the environment.
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| Once primary and secondary activity releases to the environment are quantified, the thyroid and whole body doses at the EAB and LPZ are calculated.
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| 15.1.5.5.2 Input Parameters and Initial Conditions Offsite radiological dose consequences associated with MSLBs were analyzed under the following conditions:
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| : 1. Isotope inventories were based on a core power level of 4070 MWt, or 102% of the RTP of 3990 MWt.
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| : 2. Based on Technical Specification limits, the initial assumed contamination in the NSSS was:
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| * RCS Dose Equivalent (DEQ) I-131: 1.0 µCi/gm
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| * RCS Noble Gas: 100/ µCi/gm
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| * Secondary System DEQ I-131: 0.10 µCi/gm Where E is the average of the sum of the average beta and gamma energies per disintegration (in units of MeV), for noble gas isotopes with half lives greater than 15 minutes, weighted in proportion to the concentration of each isotope in the reactor coolant.
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| Tables 15.1.5-6 and 15.1.5-7, respectively, identify the initial RCS iodine and noble gas source terms for the radioisotopes that were included in the analysis.
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| Radioiodine dose conversion factors were set to the ICRP-30(16) values listed in UFSAR Table 15B-4, and the average beta () and gamma () disintegration energies June 2011 15.1-89 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM for each noble gas isotope were set to the values listed in UFSAR Table 15B-1.
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| : 3. For PIS cases, the initial concentration of DEQ I-131 in the RCS was increased by a factor of 60. For GIS cases, an accident-generated spiking factor of 500 was used to compute the time-dependent RCS iodine concentration.
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| : 4. An RCS liquid mass of 555,000 lbm of water was used in the analysis, including 45,000 lbm of water in the pressurizer. Additionally, 4,500 lbm of steam was assumed to be in the pressurizer. Although the RCS may hold more mass, these values were selected to increase the iodine concentration following postulated fuel failures, which conservatively increases offsite dose consequences.
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| : 5. Since the PSVs may lift for this event, the dose calculation conservatively takes into account activity that might be released to containment, even though the Reactor Drain Tank is sized to remain intact from the PSV discharge.
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| June 2011 15.1-90 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.5-6 RCS IODINE SOURCE TERM FOR THE MSLB OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Term Isotope (Ci/MWt)
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| I-131 25,100 I-132 38,100 I-133 56,220 I-134 65,760 I-135 51,040 Table 15.1.5-7 RCS NOBLE GAS SOURCE TERM FOR THE MSLB OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Term Isotope (Ci/MWt)
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| Kr-83m 4,153 Kr-85 440 Kr-85m 13,000 Kr-87 21,540 Kr-88 32,020 Xe-131m 260 Xe-133 56,220 Xe-133m 1,384 Xe-135 53,640 Xe-135m 18,200 Xe-138 49,700 June 2011 15.1-91 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 6. Steam generator liquid masses ranging from 160,600 lbm of water (bounding low value for a transient) to 310,000 lbm of water (bounding high value for HZP) were considered in the analysis. This range of steam generator liquid masses bounds the masses that would occur during normal operation or during a transient. A minimum value of steam generator liquid mass tends to increase the releases for cases with fuel failures, thereby increasing offsite dose consequences. However, in cases without fuel failure, the releases from the affected steam generator have a much larger contribution and a maximum steam generator, and a maximum liquid mass can increase the offsite dose consequences.
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| : 7. A primary-to-secondary leak rate of 0.5 gpm (720 gallons per day) per steam generator was assumed. This is consistent with the PVNGS Technical Specification limit for RCS leakage prior to issuance of Operating License Amendment No. 120,(15) and conservative with respect to the current Technical Specification limit.
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| : 8. All of the iodines associated with the affected steam generator were assumed to be released to the environment, i.e. with a decontamination factor of 1.0.
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| : 9. Iodines associated with leakage to the unaffected steam generator are released to the environment during steaming with decontamination factor of 100 since the steam generator inventory, i.e. level, is maintained.
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| : 10. A radial peaking factor of 2.0 was used, which conservatively increased the radioisotope inventories that were predicted to reside in the fuel rods.
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| June 2011 15.1-92 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 11. It was assumed that 10% of the iodine and noble gas inventories in the fuel pins were resident in the fuel rod gas gap, and available for release upon clad rupture.
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| : 12. All of the activity in the fuel rod gas gap was assumed to be released to the RCS coolant upon fuel pin failure which was assumed to be 1%.
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| : 13. It was assumed that plant operators would not initiate a plant cooldown to SDC entry conditions for at least 30 minutes following event initiation. However, it should be noted that a faster RCS cooldown rate would increase steam releases during the first two hours following the event, which would produce more severe thyroid doses at the EAB. On the other hand, a slower RCS cooldown rate would allow radionuclide concentrations to build up in the secondary system, which would produce more severe 8-hour doses at the LPZ. Therefore, radiological dose calculations were performed using two different cooldown rates:
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| * A maximum Technical Specifications cooldown rate of 100°F/hr, initiated at 30 minutes into the event sequence.
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| * A slower cooldown rate of 35°F/hr, initiated at 60 minutes into the event sequence, which would bring the RCS to SDC entry conditions at approximately 8 hours following event initiation.
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| : 14. Decay heat following the MSLB was based on the 1979 ANS decay heat curve, with a 2 uncertainty. Use of a maximum decay heat curve increases the amount of steam June 2011 15.1-93 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM released to the environment, thereby resulting in more severe dose consequences.
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| : 15. It was assumed that all four RCPs would remain in operation for the duration of the radiological dose analysis. Therefore, 26 MWt of RCP heat was included in the dose calculations, which conservatively increased steam releases and offsite doses during the controlled cooldown.
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| : 16. A value of 740,000 BTU/°F was used to represent the specific heat capacity of the RCS, the RCS clad, and the steam generators. Use of this value increases the amount of steam that must be released to the environment during the controlled cooldown.
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| : 17. The /Q atmospheric dispersion factors used in the analysis are the short-term factors shown in UFSAR Table 2.3-31.
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| Table 15.1.5-8 shows that the offsite dose consequences of a MSLB outside containment with an accident-Generated Iodine Spike (GIS), will not exceed 10% of the 10 CFR Part 100 guideline values (i.e., 30 Rem thyroid and 2.5 Rem whole body).
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| Likewise, a MSLB outside containment, with either a Preaccident Iodine Spike (PIS) or 1% fuel failure, will not exceed 10 CFR Part 100 guideline values (i.e., 300 Rem thyroid and 25 Rem whole body). The results shown in Table 15.1.5-8 are therefore in compliance with the regulatory guidelines for postulated MSLBs. These results bound the core power levels of 3990 MWt or less.
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| June 2011 15.1-94 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Because the potential for fuel failure is sufficiently limited (see UFSAR Section 15.1.5.3.3), it is also concluded that the core will remain in place and intact with no loss of core cooling capabilities.
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| 15.1.5.5.3 Results Table 15.1.5-8 presents the calculated offsite radiological dose consequences for postulated MSLBs outside containment.
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| Table 15.1.5-8 OFFSITE RADIOLOGICAL DOSES FOR MSLBs OUTSIDE THE CONTAINMENT BUILDING Thyroid Dose (REM) Whole Body Dose (REM)
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| MSLB Fuel Case Failure 0-2 Hour 0-8 Hour 0-2 Hour 0-8 Hour Fraction EAB LPZ EAB LPZ PIS 0% 2.2 1.4 0.022 0.015 GIS 0% 2.5 5.2 0.024 0.048 Fuel Failure (a) 1% 17.7 19.4 0.37 0.36
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| : a. Although fuel damage is not predicted to occur (see UFSAR Section 15.1.5.3.3) the analyses for potential radiological dose consequences assume that the maximum percentage of fuel pins allowed to fail for a MSLB with the break outside containment is 1%.
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| 15.1.5.6 Conclusions Evaluation of postulated MSLBs in plant operating Modes 1 (Power Operation) and 2 (Startup) shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| June 2011 15.1-95 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * If a MSLB results in an accident-Generated Iodine Spike (GIS), offsite radiological dose consequences will not exceed a small fraction, or 10%, of the 10 CFR Part 100 guideline values.
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| * If a MSLB results in 1% failed fuel, or if it occurs with a Preaccident Iodine Spike (PIS), offsite radiological dose consequences will not exceed 10 CFR Part 100 guideline values.
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| * Control room dose consequences following a MSLB will not exceed the limits specified by GDC 19 of 10 CFR 50 Appendix A.
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| 15.1.6 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - OPERATING MODE 3 15.1.6.1 Identification of Causes and Frequency Classification A Main Steam Line Break (MSLB) is a postulated break or rupture of a pipe in the main steam system, either inside or outside the containment building.
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| A MSLB is classified as a limiting fault. Protection by design is therefore provided for MSLBs, up to and including the complete severance of a Seismic Category I main steam line upstream of the containment isolation valves (i.e., Main Steam Isolation Valves).
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| 15.1.6.2 Sequence of Events and System Operation A MSLB is characterized as a cooldown event, because the blowdown of main steam through a pipe break would result in excessive energy removal from the NSSS and a power-to-load mismatch. Additionally, if the MSLB occurred upstream of a June 2011 15.1-96 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Main Steam Isolation Valve (MSIV), the affected steam generator would continue to blow down and dry out following a Main Steam Isolation Signal (MSIS). Long-term controlled heat removal must then be accomplished through the remaining unaffected steam generator.
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| The largest possible MSLB that may occur is a double-ended guillotine rupture of a main steam line upstream of an MSIV.
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| The PVNGS steam lines, however, have integral venturi flow restrictors installed in the steam generator outlet nozzles.
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| The maximum steam blowdown rate is therefore limited by the cross-sectional throat area of a flow restrictor, which is approximately 1.283 ft .
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| 2 Postulated MSLBs that may occur in operating Mode 3 (Hot Standby) are analyzed with respect to fuel performance, as well as to demonstrate the adequacy of shutdown margin requirements.
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| There are four significant differences between the MSLB analyses performed for Modes 1 and 2 (see UFSAR Section 15.1.5) and those performed for Mode 3:
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| * In Mode 3, reactor trip breakers may be either open or closed at the time of event initiation, and CEAs may therefore be either inserted into or withdrawn from the core.
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| * In Mode 3, the reactor would be subcritical at the time of event initiation, with reactivity limited to a keffective that is less than 0.99.
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| * In Mode 3, RCS cold leg temperature at the time of event initiation would be in the range of approximately 350oF (the lower limit of Operating Mode 3) to 572oF.
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| June 2011 15.1-97 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * In Mode 3 with the initial RCS cold leg indicated temperature 485 F, Technical Specifications require that o
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| two trains of Emergency Core Cooling System (ECCS) be operable. Below 485oF, only one train of High Pressure Safety Injection (HPSI) is required to be operable.
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| Application of the Single Failure Criterion therefore results in a total loss of safety injection capability for Mode 3 MSLB analysis initiated at lower cold leg temperatures.
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| Consequently, MSLB safety analysis was performed for a wide range of Mode 3 conditions. For the analysis, initial subcriticalities were determined on the basis of Temperature-Dependent Shutdown Margin (TDSDM) requirements discussed in UFSAR Section 15.1.6.3.3.
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| Samples of the Mode 3 MSLB analysis are as follows:
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| A. An inside containment break initiated at an RCS cold leg temperature of 572oF, with a coincident LOP.
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| B. An inside containment break initiated at an RCS cold leg temperature of 572F, with offsite power available.
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| C. An inside containment break initiated at an RCS cold leg temperature of 572F, with LOP and Steam Generator (SG) tube plugging.
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| D. An inside containment break initiated at an RCS indicated cold leg temperature of 500F, with a coincident LOP.
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| June 2011 15.1-98 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM E. An inside containment break initiated at an RCS indicated cold leg temperature of 485F, with a coincident LOP and no HPSI pump available.
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| F. An inside containment break initiated at an RCS indicated cold leg temperature of 450F, with a coincident LOP and no HPSI pump available.
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| G. An inside containment break initiated at an RCS indicated cold leg temperature of 350F, with a coincident LOP and no HPSI pump available.
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| This safety analysis reveals that the case initiated at an RCS cold leg temperature of 572 F with a coincident LOP and SG tube o
| |
| plugging yields the maximum total reactivity, maximum core power fraction, and maximum heat flux fraction for a postulated Mode 3 MSLB. This case therefore also yields the lowest Macbeth DNBR and highest linear heat rate values, making it the limiting case with respect to the potential for fuel degradation.
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| The sequence of events for this limiting case is provided in Table 15.1.6-1. (For breaks initiated at lower RCS cold leg temperatures, the timing of events may differ, and safety injection may not be credited in the analysis as explained above.) This sequence of events was obtained by simulating the limiting Mode 3 MSLB event with the mathematical models identified in UFSAR Section 15.1.6.3.
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| For this limiting case, the MSLB will initially cause the main steam flow rate to rapidly increase, then gradually decrease as the affected steam generator blows down and depressurizes (Figure 15.1.6-1). The excess steam demand will cool the RCS, June 2011 15.1-99 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM decreasing cold leg temperatures in both RCS loops (Figure 15.1.6-2) as well as the average core inlet and outlet temperatures (Figure 15.1.6-3). In the presence of a negative MTC, the decrease in RCS temperature will result in an increase in reactivity (Figure 15.1.6-4), core power (Figure 15.1.6-5),
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| and core heat flux (Figure 15.1.6-6). Figure 15.1.6-6 shows that the core heat flux may peak twice during the event simulation. The first peak occurs during the first few seconds of the transient, as forced flow through the core rapidly decreases following the LOP, and as MSIVs are closing and slowing the rate of decrease in core inlet temperature. The second peak occurs later, as the rate of increase in moderator reactivity slows, and negative reactivity from safety injection boron effectively stops the increase in core power.
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| Like MSLBs that may occur in Modes 1 and 2, the rapid cooldown following a Mode 3 MSLB will also result in decreasing RCS and steam generator pressures (Figures 15.1.6-7 and 15.1.6-8, respectively). Additionally, although some secondary system inventory may initially be lost from the unaffected steam generator, closure of the MSIVs following an MSIS will serve to retain water in that generator (Figure 15.1.6-9).
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| June 2011 15.1-100 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.1.6-1 SEQUENCE OF EVENTS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK WITH LOP SAFETY ANALYSIS (RCS Tcold = 572 F) o Time Event (seconds) 0.0 Double-ended guillotine MSLB (on SG#1) occurs inside or outside containment 0.0 LOP occurs 0.0 RCPs begins to coast down 6.49 Steam generator pressure reaches MSIS setpoint 6.99 Void begins to form in reactor vessel upper head 11.58 SIAS trip setpoint reached 12.10 All MSIVs closed. Steam flow from steam generator No. 2 halted 41.58 One HPSI pump begins injecting water into the RCS 174.0 Safety injection boron reaches RCS cold legs 212.0 Maximum total reactivity occurs 291.9 Macbeth minimum DNBR occurs 292.0 Maximum core power fraction occurs 295.0 Maximum heat flux fraction occurs 1800.0 Plant operators take control of the plant June 2011 15.1-101 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Detection of a Mode 3 MSLB may be accomplished by an RCS or steam generator low pressure alarm, a steam generator low level alarm, recognition of an excess steam demand, or a high containment pressure alarm (if the MSLB occurs inside containment).
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| For the limiting Mode 3 MSLB, Table 15.1.6-1 does not reflect a reactor trip or CEA insertion following a trip. Although a trip is anticipated in the actual plant, it should be noted that CEAs may already be fully inserted into the core when the plant is in Mode 3. Therefore, for analytical purposes, CEA worth at trip is not explicitly modeled with a time-dependent scram curve, but rather it is accounted for in the initial assumed subcriticality. This analytical practice is utilized for Mode 3 MSLBs because the minimum DNBR and maximum linear heat rate values occur later in the event sequence, after any trippable CEAs have fallen into the core.
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| Table 15.1.6-1 reflects a coincident LOP in the sequence of events. Following a LOP, the coolant flow rate through the RCS would decrease rapidly, as the RCPs coast down and the RCS transitions from forced flow to natural circulation conditions (Figure 15.1.6-10). Although a loss of forced flow may result in locally higher coolant temperatures in the core region, overall the RCS would still continue to cool down while the faulted steam generator dries out. This cooldown would increase RCS coolant density, causing the pressurizer level to drop temporarily (Figure 15.1.6-11) and a void to form in the reactor vessel upper head.
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| Pressurizer pressure would likewise decrease to the SIAS setpoint, actuating any operable safety injection pumps. As June 2011 15.1-102 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM noted above, however, consideration of Technical Specification requirements and the Single Failure Criterion results in only one HPSI pump starting on demand, if the initial cold leg indicated temperature is greater than or equal to 485 F. At o
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| these higher temperatures, safety injection flow into the RCS would serve to deliver soluble boron and add negative reactivity, thereby counteracting the positive reactivity insertion due to the moderator cooldown and Doppler effects.
| |
| At lower initial cold leg temperatures, however, the moderator reactivity insertion is less severe, and soluble boron is not required to halt the increase in reactivity and core power in the presence of adequate CEA worth at trip. For analytical purposes, no safety injection flow is credited for initial cold leg indicated temperatures that are less than or equal to 485oF.
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| Operator action is not credited in the Mode 3 MSLB safety analysis for 30 minutes following event initiation. At that time, however, it is assumed that plant operators would take action to stabilize the plant in a safe shutdown condition.
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| 15.1.6.3 Core and System Performance 15.1.6.3.1 Mathematical Models The PVNGS Mode 3 MSLB safety analysis utilized the following mathematical models:
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| * The CENTS computer code was used to simulate the NSSS transient response. The CENTS computer code is described in UFSAR Section 15.0.3.1.3.2 and in an NSSS vendor (1)(2) topical report.
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| June 2011 15.1-103 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * Because the range of the CE-1 CHF correlation does not extend to low pressures and low flow rates that may exist in the RCS following a Mode 3 MSLB, the Macbeth (17)(18) correlation is utilized to determine the margin to DNB. The Macbeth correlation calculates CHF as a function of mass flux, inlet subcooling, system pressure, heated diameter, and channel length. Use of a channel heat balance allows the correlation to be converted to a "local conditions" form, thereby allowing CHF to be determined as a function of height in the hot channel. The effect of non-uniform axial heating is incorporated by using the method applied by Lee in Reference 19. The Macbeth CHF correlation is also described in the CENTS computer code (1) topical report.
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| 15.1.6.3.2 Input Parameters and Initial Conditions Table 15.1.6-2 summarizes the key input parameters and initial conditions utilized in the PVNGS Mode 3 MSLB safety analyses.
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| The following points serve to explain the selection of initial conditions as they appear in Table 15.1.6-2:
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| * Initial core power level is established by the initial subcriticality corresponding to each RCS cold leg temperature, where the initial subcriticality is assumed to equal the minimum required TDSDM for the reactor trip breakers closed configuration case (see Figure 15.1.6-12). Evaluation of the reactor trip breakers open configuration case is discussed in Section 15.1.6.3.3.
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| June 2011 15.1-104 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * Initial core inlet temperatures were selected to bound the plant configurations, and TDSDM allowed by Technical Specifications in Mode 3.
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| Table 15.1.6-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK SAFETY ANALYSES (RCS TCOLD = 572 F) o Assumed Parameter Values Initial Subcriticality (%) -6.50
| |
| -6 Initial Core Power (% of RTP) 1.54x10 Initial Core Inlet Temperature (°F) 572 Initial Pressurizer Pressure (psia) 1340 Initial RCS Flow Rate (% of Design Rated) 95 Initial Pressurizer Water Level (% Narrow Range) 60 Initial Steam Generator Water Level (Feet) 33 (61% WR)
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| -4 Moderator Temperature Coefficient (/°F) -4.4x10 Doppler Fuel Temperature Coefficient EOC Delayed Neutron Kinetics EOC Inverse Boron Worth (ppm/%) -130 Fuel Rod Gap Conductance (BTU/hr-ft -°F) 2 5755 Number of Plugged Steam Generator Tubes (Total %) 10 Break Size (ft )
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| 2 1.283 Loss of Offsite Power Yes Additional Single Failure HPSI June 2011 15.1-105 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
| |
| * An initial pressurizer pressure of 1340 psia was assumed for the case of 572F cold leg temperature. The SIAS setpoint was set at the saturation pressure of the cold leg indicated temperature. For other temperature cases, a lowest possible Pressurizer pressure and lowest SIAS setpoint (saturation pressure) are used.
| |
| * The initial assumed core flow rate affects the rate of core inlet temperature cooldown following a MSLB, which in turn affects the positive reactivity insertion in the presence of a negative MTC cooldown curve. In accordance with station operating procedures, between 2 and 4 RCPs may be in operation during Mode 3, depending upon the RCS cold leg temperature. Additionally, Technical Specifications allow for all RCPs to be de-energized for up to one hour (per 8-hour period). For comparative purposes, the safety analyses described herein assumed 4 RCPs were in operation at the time of event initiation.
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| For those cases with a coincident LOP, all RCPs were then immediately de-energized at event initiation. For cases with offsite power available, Mode 3 MSLB analysis was performed with various RCP operating configurations. The analysis confirmed that the conclusions described herein remain valid, i.e., the limiting case with regard to fuel performance is the case initiated from an RCS cold leg temperature of 572oF with a coincident LOP, and the shutdown margin requirements for trip breakers closed configuration.
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| June 2011 15.1-106 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * A high Pressurizer level was selected since use of a high level tends to delay SIAS. The analysis conservatively assumes a pressurizer level corresponding to 60% of Pressurizer level control range.
| |
| * The safety analysis performs parametrics on the Pressurizer pressure and steam generator level. High steam generator level increases the stored energy, results in a more severe blowdown, a faster cooldown rate following the MSLB, and a greater reactivity insertion in the presence of a negative MTC cooldown curve. However, this also causes the Pressurizer pressure to drop much lower such that the HPSI pump is able to inject more highly borated water into the core.
| |
| * When the liquid inventory in a steam generator decreased below 5000 lbm as a result of the blowdown, the primary-to-secondary system heat transfer rate was ramped down to zero as the mass decreased to 2500 lbm.
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| * The most negative MTC cooldown curve was used for this analysis, corresponding to End of Cycle (EOC) conditions.
| |
| * An EOC (most negative) Doppler fuel temperature coefficients curve was used for the Mode 3 MSLB analysis.
| |
| Use of EOC values is consistent with the selection of the most negative MTC cooldown curve. Use of a most negative curve adds relatively more positive reactivity as a result of a change in fuel temperature following a MSLB.
| |
| * Relative to other parameters, the delayed neutron fraction has a minor effect on core power following a June 2011 15.1-107 Revision 16
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| | |
| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Mode 3 MSLB. EOC values were chosen, consistent with the use of EOC MTC and Doppler values.
| |
| * A maximum value was selected for the inverse boron worth, to reduce the negative reactivity that would be inserted as a result of boron injected by the Safety Injection (SI) system. The sweep-out volume represents the volume of water that must be displaced before safety injection boron reaches the primary system. The Safety analysis for this event uses a more conservative (larger) sweep-out volume than the required value of 60.6 cubic feet in the SI line from the RCS. Boron injection was credited only for those cases initiated from an RCS cold leg indicated temperature of 485 F or higher (see UFSAR o
| |
| Section 15.1.6.2).
| |
| * The maximum fuel rod gas gap conductance value was selected so that energy from the fuel would quickly reach the surface of the fuel rod clad. This results in a heat flux that closely follows core power, and therefore more thermal margin degradation as power increases.
| |
| * Steam generator tube plugging decreases the heat transfer rate from the RCS to the secondary system and thereby decreases the initial SG pressure. A lower SG pressure promotes an earlier MSIS but delays the SIAS (due to higher pressurizer pressure). The combination effect determines the resulting limiting condition. Based on the parameteric evaluation, it is shown that the limiting case is 10% tube plugged for 3990 MWt configuration.
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| * A large break size was assumed for the analysis, with steam blowdown limited by the cross-sectional throat area June 2011 15.1-108 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM of the flow restrictors in the outlet nozzles of both steam generators. A large break size maximizes the cooldown rate and resulting reactivity insertion.
| |
| * For the Mode 3 MSLB analysis, an additional single failure involving the failure of one HPSI pump to start on demand was assumed (see UFSAR Table 15.0-0). For those cases initiated at an RCS cold leg indicated temperature of 485°F or higher, this single failure served to reduce the capacity of the SI system to provide boron to the RCS. For those cases initiated at lower RCS cold leg temperatures, this failure resulted in a total loss of SI capacity (see UFSAR Section 15.1.6.2). A single failure affecting SI capacity is more limiting with respect to core performance than a postulated failure of an MSIV to close following an MSIS.
| |
| 15.1.6.3.3 Results Key results for the limiting Mode 3 MSLB safety analysis are summarized in Table 15.1.6-3.
| |
| Table 15.1.6-3 RESULTS FOR THE LIMITING SUBCRITICAL MAIN STEAM LINE BREAK SAFETY ANALYSES (RCS TCOLD = 572°F)
| |
| Parameter Results Peak Linear Heat Generation Rate (KW/ft) 12.8 Maximum Core Power (% of RTP) 1.46 Maximum Total Reactivity (%) 0.236 Minimum MacBeth DNBR 1.95 June 2011 15.1-109 Revision 16
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| | |
| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The case initiated at an RCS cold leg temperature of 572 F, o
| |
| with a coincident LOP and SG tube plugging, yielded the largest total reactivity, the largest core power fraction, and the largest heat flux fraction for any of the analyzed Mode 3 MSLBs. Further analysis of this case revealed that, if both the fission power and decay power components were taken into consideration, the peak transient linear heat rate would be less than that required to cause fuel centerline melting.
| |
| Additionally, the calculated Macbeth DNBR remained above the deterministic limit of 1.30. Therefore, it is concluded that fuel clad degradation would not occur following a postulated MSLB in Mode 3.
| |
| As noted previously, the Mode 3 MSLB safety analysis cases were initiated from subcritical conditions, where the initial subcriticality was assumed to be equal to the minimum required TDSDM for an indicated cold leg temperature (see Table 15.1.6-2 and Figure 15.1.6-12). However, as Figure 15.1.6-12 shows, the minimum required TDSDM differs between the trip breakers open configuration and the trip breaker closed configuration.
| |
| Generally, if the reactor trip breakers are open, the minimum SDM required by the Core Operating Limits Report (COLR) is less than that required when the breakers are closed, across the entire range of cold leg temperatures allowed in Mode 3.
| |
| However, it should also be noted that the Technical Specification definition for SDM states that the single CEA of highest reactivity worth is assumed to be fully withdrawn from the core. Therefore, if the CEAs are in a confirmed inserted configuration with reactor trip breakers open, the amount of reactivity by which the plant would actually be subcritical would be the sum of the SDM and the Stuck Rod Worth (SRW).
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| June 2011 15.1-110 Revision 16
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| | |
| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Figure 15.1.6-12 illustrates an example in which the SRW is assumed to be 2.0% .
| |
| Figure 15.1.6-12 shows that the plant must be subcritical by a SDM of at least 4.0% to 6.5% (depending upon the value of Tcold) when reactor trip breakers are closed, or by a SDM of at least 1.0% to 5.0% when reactor trip breakers are open. However, for the breakers open condition, these SDM values are based on the assumption that the most reactive rod is held out of the core. In the trip breakers open configuration, this stuck rod is actually inserted into the core, adding another 2.0% of subcriticality. Therefore, for the trip breakers open configuration, the reactor would actually be subcritical by at least 3.0% to 7.0%
| |
| (depending upon the value of Tcold).
| |
| Further comparison of the curves in Figure 15.1.6-12 shows that, for RCS cold leg temperatures greater than 450oF, the reactor would be less subcritical (i.e., at a higher initial core power level) for the trip breakers closed configuration than for the trip breakers open configuration. Therefore, the Mode 3 MSLB safety analyses described herein for the trip breakers closed configuration (above 450 F), clearly bound the o
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| trip breakers open configuration when the SRW is at least 2.0% .
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| Conversely, for RCS cold leg indicated temperatures less than 450oF, the reactor would be initially less subcritical for the trip breakers open configuration than for the trip breakers closed configuration. Therefore, it is concluded that the trip breakers open configuration likewise bound the trip breakers closed configuration if the SRW is at least 2.0% .
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| June 2011 15.1-111 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Therefore, verification that the SRW is at least 2.0% for reload core designs, ensures that the Mode 3 MSLB safety analysis described herein bounds both the trip breakers open and trip breakers closed configurations, across the full spectrum of Mode 3 cold leg temperatures.
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| 15.1.6.4 RCS Pressure Boundary Barrier Performance 15.1.6.4.1 Mathematical Models The mathematical models that were used to analyze the performance of the RCS pressure boundary are the same as those described in UFSAR Section 15.1.6.3.1.
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| 15.1.6.4.2 Input Parameters and Initial Conditions The key input parameters and initial conditions that were used to analyze the performance of the RCS pressure boundary are the same as those described in Section 15.1.6.3.2.
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| 15.1.6.4.3 Results The safety analysis shows that Mode 3 MSLB is primarily characterized as a cooldown event and that, following an MSIS, both RCS and steam generator pressures will tend to stabilize (see Figures 15.1.6-7 and 15.1.6-8). Because heat sources in the NSSS are not sufficient to raise pressurizer pressure to the minimum PSV lift setting specified in the Technical Specifications, the peak RCS pressure will remain below the acceptable design limit for this event (i.e., 110% of the RCS design pressure of 2500 psia, or 2750 psia). Likewise, pressure in the main steam system will remain below the minimum allowable MSSV lift settings and therefore below the acceptable design June 2011 15.1-112 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM limit for this event (i.e., 110% of the steam generator shell side design pressure of 1270 psia, or 1397 psia).
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| 15.1.6.5 Containment Performance and Radiological Consequences The core performance safety analysis in UFSAR Section 15.1.6.3 indicates that MSLBs in Mode 3 (Hot Standby) will not result in fuel clad degradation. Hence, significant offsite and control room dose consequences are not anticipated for such events.
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| Nonetheless, because MSLBs are classified as limiting faults, their postulated consequences include the potential to release significant amounts of radioactive material to the environment.
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| It is concluded that the radiological consequences of MSLBs that may occur during Mode 3 are conservatively bounded by the results presented in UFSAR Section 15.1.5.5 for MSLBs that may occur in Modes 1 (Power Operation) or 2 (Startup).
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| 15.1.6.6 Conclusions Evaluation of postulated MSLBs in plant operating Mode 3 (Hot Standby) shows that:
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| * Pressure in the RCS will be maintained below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * Pressure in the main steam system will be maintained below 110% of the steam generator shell side design value (i.e., 110% of 1270 psia, or 1397 psia).
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| * Offsite and control room radiological dose consequences are bounded by the results presented in UFSAR Section 15.1.5.5 for MSLBs that may occur in Modes 1 (Power Operation) or 2 (Startup).
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| June 2011 15.1-113 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.
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| | |
| ==1.7 REFERENCES==
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| : 1. Combustion Engineering, Technical Manual for the CENTS Code, CE-NPD 282-P, Volumes 1-3, February 1991. [See also Reference 2 below.]
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| : 2. Nuclear Regulatory Commission, Acceptance for Referencing of Licensing Topical Report CE-NPD 282-P, Technical Manual for the CENTS Code (TAC No. M82718),
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| letter from M. J. Virgilio (NRC) to S. A. Toelle (ABB Combustion Engineering), March 17, 1994.
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| : 3. Combustion Engineering, CPC/CEAC Software Modifications for the CPC Improvement Program, CEN-308-P-A, April 1986.
| |
| : 4. Combustion Engineering, CPC and Methodology Changes for the CPC Improvement Program, CEN-310-P-A, April 1986.
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| : 5. Combustion Engineering, "Responses to First Round Questions on the Statistical Combination of Uncertainties Program: CETOP-D Code Structure and Modeling Methods,"
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| CEN-139(A)-P, March 1981.
| |
| : 6. Combustion Engineering, "Responses to First Round Questions on the Statistical Combination of Uncertainties Program: CETOP-D Code Structure and Modeling Methods,"
| |
| CEN-124(B)-P, Part 2, May 1981.
| |
| : 7. Combustion Engineering, "CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3," CEN-160(S)-P, September 1981.
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| : 8. Intentionally Left Blank.
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| June 2011 15.1-114 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 9. Combustion Engineering, "HERMITE, A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," CENPD-188-A, March 1976.
| |
| : 10. Combustion Engineering, "TORC Code: A Computer Code for Determining the Thermal Margin of a Reactor Core,"
| |
| CENPD-161-P-A (proprietary), CENPD-161-A (non-proprietary),
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| April 1986.
| |
| : 11. Combustion Engineering, "TORC Code Verification and Simplified Modeling Methods," CENPD-206-P-A (proprietary), CENPD-206-A (non-proprietary), June 1981.
| |
| : 12. C-E Methods for Loss of Flow Analysis, CENPD-183, July 1975.
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| : 13. C-E Methods for Loss of Flow Analysis, CENPD-183-A, June 1984.
| |
| : 14. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 26 to Facility Operating License No. NPF-74, Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit No. 3, Docket No. STN-50-530, May 20, 1991.
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| : 15. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 120 to Facility Operating License No. NPF-41, Amendment No. 120 to Facility Operating License No.
| |
| NPF-51, and Amendment No. 120 to Facility Operating License No. NPF-74, Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Docket Nos. STN 50-528, STN 50-529, and STN 50-530, August 5, 1999.
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| June 2011 15.1-115 Revision 16
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| PVNGS UPDATED FSAR INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| : 16. International Commission on Radiation Protection, Publication No. 30, Supplement to Part 1, Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity, 1980.
| |
| : 17. Macbeth, R. V., "An Appraisal of Forced Convection Burn-Out Data," Proceedings of the Institute of Mechanical Engineers, Vol. 180, 1965-66.
| |
| : 18. Macbeth, R. V., "Burn-Out Analysis - Part 5: Examination of Published World Data for Rod Bundles," United Kingdom Atomic Energy Authority, Atomic Energy Establishment Winfrith (AEEW) Report R358, 1964.
| |
| : 19. Lee, D. H., "An Experimental Investigation of Forced Convection Burn-Out in High Pressure Water - Part IV, Large Diameter Tubes at About 1600 psia," A. E. E. W.
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| Report R479, 1966.
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| June 2011 15.1-116 Revision 16
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| PVNGS UPDATED FSAR 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 LOSS OF EXTERNAL LOAD 15.2.1.1 Identification of Event and Causes The loss of external load event is caused by the disconnection of the main generator from the electrical distribution grid.
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| 15.2.1.2 Sequence of Events and Systems Operation A loss of external load generates a turbine trip which results in a reduction in steam flow from the steam generators to the turbine, due to the closure of the turbine stop valves. The steam bypass control system (SBCS) and reactor power cutback system (RPCS) are both normally in the automatic mode and would be available upon turbine trip to accommodate the load rejection without necessitating reactor trip or the opening of main steam safety valves. Should a turbine trip occur with these systems in the manual mode, a complete termination of main steam flow results and reactor trip would occur on high pressurizer pressure. If no credit is taken for immediate operator action, the main steam safety valves will open to limit the secondary pressure increase and provide a heat sink for the nuclear steam supply system (NSSS). The operator can initiate a controlled system cooldown using the SBCS any time after reactor trip occurs.
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| 15.2.1.3 Analysis of Effects and Consequences The results of the loss of load event are no more limiting with respect to reactor coolant system (RCS) pressurization than those of the loss of condenser vacuum (LOCV) event presented in subsection 15.2.3. The LOCV also results in a turbine trip; however, feedwater flow is assumed to terminate following LOCV June 2001 15.2-1 Revision 11
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM whereas it is assumed to ramp down to 5% following the loss of load. This larger reduction in heat removal capability results in a higher peak RCS pressure for the LOCV.
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| Like the LOCV, the departure from nucleate boiling ratio (DNBR) increases during the loss of load due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR.
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| For the loss of load, due to its similarity with the LOCV event, there are no concurrent single failures which, when combined with the loss of external load, result in consequences more severe than the LOCV event with respect to RCS pressurization. The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of flow (LOF) event discussed in subsection 15.3.1. Results of the LOF event are directly applicable to the loss of external load with loss of offsite power following a turbine trip.
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| 15.2.1.4 Conclusions For the loss of load event and the loss of load with a single failure, the RCS pressure remains below 2750 psia thus ensuring primary integrity, and the minimum DNBR remains above the limit thus ensuring fuel cladding integrity.
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| 15.2.2 TURBINE TRIP 15.2.2.1 Identification of Event and Causes A turbine trip may result from a number of conditions which cause the turbine generator control system (TGCS) to initiate a turbine trip signal. A turbine trip initiates closure of the turbine stop valves.
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| June 2001 15.2-2 Revision 11
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.2.2 Sequence of Events and Systems Operation A turbine trip results in a reduction in steam flow from the steam generators to the turbine due to the closure of the turbine stop valves. The SBCS and RPCS are both normally in the automatic mode and would be available upon turbine trip to accommodate the load rejection without necessitating reactor trip or the opening of main steam safety valves. Should a turbine trip occur with these systems in the manual mode, a complete termination of main steam flow results and reactor trip would occur on high pressurizer pressure. If no credit is taken for immediate operator action, the main steam safety valves will open to limit the secondary pressure increase and provide a heat sink for the NSSS. The operator can initiate a controlled system cooldown using the SBCS any time after reactor trip occurs.
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| 15.2.2.3 Analysis of Effects and Consequences The results of the turbine trip event are no more limiting with respect to RCS pressurization than those of the LOCV event presented in subsection 15.2.3. The LOCV also results in a turbine trip; however, feedwater flow is assumed to terminate following LOCV whereas it is assumed to ramp down to 5%
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| following the turbine trip. This larger reduction in heat removal capability results in a larger peak RCS pressure for the LOCV.
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| Like the LOCV, the DNBR increases during the turbine trip due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR for the loss of load. Due to its similarity with the LOCV events, there are no concurrent single failures which when combined with the turbine trip result in consequences more June 2001 15.2-3 Revision 11
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM severe than the LOCV event with respect to RCS pressurization.
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| The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of ac power which initiates the LOF event discussed in subsection 15.3.1. Results of the LOF event are directly applicable to the turbine trip event with a loss of offsite power.
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| 15.2.2.4 Conclusions For the turbine trip event and the turbine trip with a single failure, the RCS pressure remains below 2750 psia thus ensuring primary system integrity, and the minimum DNBR remains above the limit thus ensuring fuel cladding integrity.
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| 15.2.3 LOSS OF CONDENSER VACUUM 15.2.3.1 Identification of Causes and Frequency Classification A loss of condenser vacuum (LOCV) may occur due to, but not limited to, the failure of the circulating water system to supply cooling water to the condenser; failure of the main condenser evacuation system to remove non-condensable gases; failure of a condenser vacuum breaker; excessive in-leakage of air through a turbine gland packing; loss of power (LOP); or rupture of a condenser shell.
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| An LOCV is an Anticipated Operational Occurrence (AOO) and is classified as an incident of moderate frequency.
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| June 2005 15.2-4 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.3.2 Sequence of Events and Systems Operation The LOCV analyses are performed as separate cases for the primary and secondary peak pressure limits, since these events are not mutually conservative. The sequence of events for the moderate frequency LOCV event is presented in Table 15.2.3.1 for the primary peak pressure case and Table 15.2.3.2 for the secondary peak pressure case. The primary peak pressure case also analyzes the fuel integrity (minimum DNBR) for the LOCV event.
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| Condenser pressure will increase following an LOCV, causing a trip of the main turbine and closure of the Turbine Admission Valves (TAVs). The LOCV also causes the feedwater pumps to trip due to low suction pressure and disables the turbine bypass valves. The closure of the turbine stop valves and coastdown of main feedwater pumps result in a Reactor Coolant (Primary) System (RCS) and Main Steam (Secondary) System heat up, and both system pressures increase rapidly. A reactor trip occurs on high pressurizer pressure (HPPT) occurs. The pressure increase in the primary and secondary system are limited by the primary safety valves (PSVs) and main steam safety valves (MSSVs).
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| If the Reactor Power Cutback System (RPCS) and the Steam Bypass Control System (SBCS) are in automatic mode of operation, a reactor power cutback may occur and the Nuclear Steam Supply System (NSSS) may continue to operate at a reduced power level.
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| However, for the LOCV analysis, both the RPCS and SBCS are assumed to be in manual mode and credit is not taken for their functioning. Likewise, the Pressurizer Pressure Control System (PPCS) and Pressurizer Level Control System (PLCS), which may June 2005 15.2-5 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM reduce over pressurization of the RCS, are assumed to be in manual mode and no credit is taken for their functioning.
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| A reactor trip on low steam generator level (LSGLT) could occur immediately following a LOCV, when a steam generator pressure spike causes the steam bubbles in the steam generator to collapse. However, this level trip is not credited in the analysis.
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| An auxiliary feedwater actuation signal (AFAS) on low steam generator level occurs as the plant begins to cooldown and depressurize. The auxiliary feedwater flow is automatically initiated after a time delay and begins to fill the steam generators until a normal level is reached.
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| The LOCV analysis does not credit operator action for the first thirty minutes following the event. Thirty minutes after initiation of the LOCV event, the operators commence a cooldown using the Atmospheric Dump Valves (ADVs).
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| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions are consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS UFSAR delineated in UFSAR Chapter 7.
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| The NRC's Standard Review Plan states that an incident of moderate frequency, such as the loss of condenser vacuum event, should not generate a more serious plant condition without other faults occurring independently. In addition, the Standard Review Plan states that an incident of moderate frequency, in combination with a single active component June 2009 15.2-6 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM failure or single operator error, should not result in the loss of function of any barrier other than the fuel cladding.
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| A LOCV event will cause a termination of feedwater and main steam flow. The analysis does not model the time dependence in detail. Instead, the LOCV is assumed to abruptly and completely terminate both main steam and feedwater flow. In considering the peak pressure criteria, the only mechanisms for mitigation of the RCS and main steam system overpressurization are the PSVs, MSSVs and RCS flow. Table 15.0-0 is used to determine credible single failures for safety analysis. This table indicates that there are no credible failures that can degrade the PSV and MSSV capacity. Technical Specification 3.7.1 places limits on reactor power and variable overpower trip (VOPT) setpoints when one or more MSSVs are inoperable, thereby ensuring secondary system peak pressure remains within 110% of secondary system design pressure. The LOCV is one of the transients analyzed for validating Technical Specification 3.7.1. A decrease in RCS to steam generator heat transfer due to reactor coolant flow coastdown can be caused by a LOP following a turbine trip. A Reactor Coolant Pump (RCP) coastdown results in a reactor trip that is generated by the Core Protection Calculators on RCP speed. Due to the rapid reactor trip, this failure reduces the peak pressure relative to the LOCV itself. The results of the parametric study show that a LOP coinciding or following the High Pressurizer Pressure Trip (HPPT) does not make the primary and secondary side pressures more adverse. In addition, it is assumed that the most reactive control rod fails to insert on scram.
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| June 2005 15.2-7 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Therefore, it was concluded that there is no single failure that would make the maximum primary and secondary pressure more limiting during a LOCV event.
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| A decrease in RCS flow is the only parameter which can significantly reduce the minimum DNBR during a LOCV event. The LOP is the only failure that may affect RCS flow. LOCV by itself, however, produces an increasing RCS pressure which compensates for the elevated RCS temperatures such that the available thermal margin does not degrade before the onset of the LOP. Thus, the overall DNBR degradation experienced during an LOCV event with LOP would be bounded by that of the loss of RCS flow event of UFSAR Section 15.3.1.
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| June 2005 15.2-8 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.3-1 SEQUENCE OF EVENTS FOR THE LOCV PRIMARY SIDE PEAK PRESSURE and FUEL PERFORMANCE (DNBR) EVENT Time (sec) Events 0.00 LOCV, turbine trip, main FW pump trip 0.00 Minimum DNBR occurs 7.05 Pressurizer pressure reaches HPPT setpoint 7.55 Reactor trip breakers open 8.16 Scram CEAs begin falling 8.69 PSVs open 8.99 MSSV Bank 1 opens1 9.60 Maximum RCS pressure reached 11.08 MSSV Bank 2 opens 12.93 PSVs close 13.35 Maximum steam generator pressure occurs 13.36 MSSV Bank 3 opens 16.81 Steam generator water level reaches AFAS Analytical setpoint 19.80 MSSV Bank 3 closes 31.39 MSSV Bank 2 closes 59.98 MSSV Bank 1 closes 62.90 Auxiliary Feedwater (AFW) flow initiated 1798.50 Maximum pressurizer water volume occurs 1800.0 Operator initiates plant cooldown 1
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| Bank 1 MSSVs cycle throughout the 1800 seconds. Only the initial opening and closure are listed in the table.
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| June 2011 15.2-9 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.3-2 SEQUENCE OF EVENTS FOR THE LOCV SECONDARY SIDE PEAK PRESSURE EVENT Time (sec) Event 0.00 LOCV, turbine trip, main feedwater pump trip occur 0.00 Minimum DNBR occurs 4.02 MSSV bank 1 opens2 5.65 MSSV bank 2 opens 6.89 Pressurizer pressure reaches HPPT setpoint 7.32 MSSV bank 3 opens 7.39 Reactor trip breakers open 8.00 Scram CEAs begin falling 9.50 PSVs open 9.83 Maximum RCS pressure occurs 11.50 PSVs close 12.09 Maximum pressurizer water volume 12.56 Steam generator water level reaches AFAS analytical setpoint 13.97 Maximum steam generator pressure occurs 22.62 MSSV bank 3 closes 34.77 MSSV bank 2 closes 58.56 AFW flow initiated 60.10 Pressurizer Pressure reaches SIAS analytical setpoint 64.10 MSSV bank 1 closes 90.10 High Pressure Safety Injection flow initiated 1800.0 Operator initiates plant cooldown 2
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| Bank 1 MSSVs cycle throughout the 1800 seconds. Only the initial opening and closure are listed in the table.
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| June 2011 15.2-10 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.3.3 Core and System Performance A. Mathematical Model The NSSS response to a LOCV was simulated using the CENTS computer program described in UFSAR Section 15.0.3.1.3.2.
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| Parametric studies are performed using key design inputs to maximize primary and secondary side pressures. Inputs to the CENTS code such as moderator reactivity as a function of moderator density, Doppler reactivity as a function of effective fuel temperature, and shutdown rod worth were calculated using the two-dimensional ROCS code discussed in UFSAR Section 4.3.3.1.1.2. Shutdown rod worth assumes that the most reactive control rod fails to insert on scram. Input to the CENTS code may also be calculated using the SIMULATE-3 code discussed in UFSAR Section 4.3.3.1.1.5.
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| The initial and transient DNBR was calculated using the CETOP computer code (see UFSAR Section 4.4 and 15.0.3.1.6), which uses the CE-1 CHF correlation described in reference 2. The LOCV event does not present a challenge to the DNBR SAFDL because the RCS overpressurization tends to increase DNBR.
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| Since there is no power excursion during the transient, the LOCV event does not challenge the peak fuel centerline temperature SAFDL or the limit on linear heat generation rate (21 kW/ft).
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a LOCV from full power conditions are presented in Table 15.2.3-3. The initial June 2005 15.2-11 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM conditions were varied within the ranges of steady state operation configurations (i.e., specified by the Technical Specifications, plant configuration, and design specifications) to determine the set of initial conditions that produce the most adverse consequences following a LOCV.
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| Parameters of interest include initial core inlet temperature, core inlet flow, pressurizer pressure, steam generator level, pressurizer water level, PSV and MSSV tolerances and blowdowns, Moderator Temperature Coefficient (MTC), Fuel Temperature Coefficient (FTC),
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| fuel rod gap conductances, kinetics parameters, LOP, function of PLCS and PPCS and SG tube plugging. Starting from a base case, one parameter at a time is changed to establish the trends for the RCS and steam generator pressure.
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| C. Results The response of key core parameters as a function of time is presented in Figures 15.2.3-1 through 15.2.3-3 and 15.2.3-14 through 15.2.3-17 for this moderate frequency event for the primary peak pressure case and the secondary peak pressure case, respectively. The sudden reduction in steam flow caused by the LOCV leads to a reduction of the primary-to-secondary heat transfer and an increase in RCS temperature. The rapid heatup and volumetric expansion of the reactor coolant results in an increase in pressurizer pressure. When the pressurizer pressure reaches the HPPT setpoint, reactor trip occurs. The CEAs drop into the core initiating the decrease in core power from full power.
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| June 2005 15.2-12 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.3-3 ASSUMED INITIAL CONDITIONS FOR LOCV PRIMARY PEAK PRESSURE/DNBR AND SECONDARY SIDE PEAK PRESSURE CASES Value Parameter Primary Side Secondary Peak Side Peak Pressure/DNBR Pressure Case Case Initial core power (% of RTP) 102 102 Initial core inlet temperature (°F) 555 566 Initial pressurizer pressure (psia) 2100 2100 Initial RCS flow (% design) 116 95 Initial pressurizer water level (%) 24 24 Initial steam generator level (% WR) 60.7 60.7 Moderator Temperature Coefficient 0.0 0.0
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| (/°F)
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| Least Most Fuel Temperature Coefficient negative negative SCRAM delay time (sec) 0.5 0.5 CEA holding coil delay (sec) 0.6 0.6 CEA worth at trip - WRSO (%) 8.0 8.0 2
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| Fuel gap conductance (Btu/hr-ft -°F) 500 500 Plugged tubes (per steam generator) 1258 0 AFW flow (gpm/pump) 650 650 AFW delay time (sec) 46 46 PSV setpoint tolerance +3% +3%
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| PSV blowdown 5% 5%
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| MSSV Setpoint Tolerance +3% +3%
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| MSSV blowdown 5% 5%
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| Pressurizer heaters and sprays Off On Charging and letdown flows (gpm) 139.5/29 46.5/46.5 LOP No No June 2005 15.2-13 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Since there is no power excursion during the transient, the LOCV event does not challenge the peak fuel centerline temperature SAFDL or the limit on linear heat generation rate (21 kW/ft).
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| The minimum DNBR is greater than the DNBR SAFDL value of 1.34 and meets the acceptance criteria of the Standard Review Plan (see Figure 15.2.3-14). Therefore, fuel cladding damage is not predicted for this moderate frequency event.
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| 15.2.3.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate the RCS barrier performance for the limiting moderate frequency event are identical to those described in UFSAR Section 15.2.3.3.A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed to evaluate the RCS barrier performance for the limiting moderate frequency event are identical to those described in UFSAR Section 15.2.3.3.B.
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| C. Results The response of key RCS parameters as a function of time is presented in Figures 15.2.3-4 through 15.2.3-13 and 15.2.3-18 through 15.2.3-28 for this limiting moderate frequency event.
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| The PSVs open and a maximum RCS pressure of 2745 psia is reached, which is less than 110% (2750 psia) of RCS design June 2011 15.2-14 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM pressure (2500 psia). For the secondary side peak pressure case, three banks of MSSVs open and the maximum secondary system pressure of 1390 psia is reached, which is less than 110% (1397 psia) of secondary system design pressure (1270 psia).
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| These primary and secondary side maximum pressures meet the acceptance criteria of Standard Review Plan 15.2.3.
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| 15.2.3.5 Radiological Consequences and Containment Performance LOCV is a moderate frequency event in which no fuel damage occurs. Therefore, radiological consequences are not calculated for this event and containment isolation is not credited.
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| 15.2.3.6 Conclusions For the loss of condenser vacuum event, the maximum RCS pressure remains below 110% of RCS design pressure(2750 psia),
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| thus ensuring primary system integrity. Likewise, the maximum secondary system pressure remains below 110% of design pressure (1397 psia), thus ensuring secondary system integrity.
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| The minimum DNBR remains well above the SAFDL limit, thus ensuring fuel cladding integrity.
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| June 2011 15.2-15 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.4 MAIN STEAM ISOLATION VALVE CLOSURE 15.2.4.1 Identification of Event and Causes The main steam isolation valve closure event is initiated by the closure of all MSIVs due to a spurious closure signal.
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| 15.2.4.2 Sequence of Events and Systems Operation The closure of all MSIVs results in the termination of all main steam flow. The decreased heat removal results in increasing primary and secondary temperatures and pressure. Reactor trip occurs on high pressurizer pressure. The pressure increases in the primary and secondary systems are limited by the pressurizer and steam generator safety valves. The operator can initiate a controlled system cooldown using the steam bypass control system any time after reactor trip occurs.
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| 15.2.4.3 Analysis of Effects and Consequences The results of the MSIV closure event are no more limiting with respect to RCS pressurization than those of the LOCV event presented in subsection 15.2.3. The LOCV also results in the termination of all main steam flow. However, main steam flow is terminated more rapidly during the LOCV since the closure time for the turbine stop valves is much shorter than that for the MSIVs. The faster reduction in heat removal results in a higher peak RCS pressure for the LOCV event.
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| Like the LOCV, the DNBR increases during the MSIV closure event due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR for the MSIV closure event.
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| Due to the similarity with the LOCV event, there are no concurrent single failures which when combined with the MSIV June 2011 15.2-16 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM closure event result in consequences more severe than the LOCV event with respect to RCS pressurization. The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of ac power which initiates the LOF event discussed in subsection 15.3.2. Results of the LOF event are directly applicable to the MSIV closure with loss of offsite power following a turbine trip.
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| 15.2.4.4 Conclusions For the MSIV closure event and the MSIV closure with a single failure, the RCS pressure remains below 2750 psia thus ensuring primary system integrity, and the minimum DNBR remains above the limit thus ensuring fuel clad integrity.
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| 15.2.5 STEAM PRESSURE REGULATOR FAILURE This event does not apply to the CESSAR SYSTEM 80 design and therefore is not presented.
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| 15.2.6 LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6.1 Identification of Event and Causes The loss of nonemergency ac power to the station auxiliaries (LOAC) may result from either a complete loss of the external grid or a loss of the onsite ac distribution system. The LOAC is presented as the initiating event for the four pump loss of flow event discussed in subsection 15.3.1.
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| June 2011 15.2-17 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.6.2 Sequence of Events and Systems Operation When all normal ac power is assumed to be lost to the plant, the turbine stop valves close, and it is assumed that the area of the turbine control valves is instantaneously reduced to zero. Also, the feedwater flow to both steam generators is instantaneously assumed to go to zero. The reactor coolant pumps coast down and the reactor coolant flow begins to decrease. A reactor trip will occur as a result of a low DNBR condition as the flow coastdown begins. The pressure increases in the RCS and steam generators are limited by the pressurizer and steam generator safety valves.
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| The loss of all normal ac power is followed by automatic startup of the standby diesel generators, the power output of which is sufficient to supply electrical power to all necessary engineered safety features systems and to provide the capability of maintaining the plant in a safe shutdown condition. Subsequent to the reactor trip, stored and fission product decay energy must be dissipated by the reactor coolant system and main steam system. In the absence of forced reactor coolant flow, convective heat transfer coolant flow occurs.
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| Initially, residual water inventory in the steam generators is used as a heat sink, and the resultant steam is released to atmosphere by the spring-loaded steam generator safety valves.
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| With the availability of standby diesel power, auxiliary feedwater is automatically initiated on a low steam generator water level signal. Plant cooldown is operator controlled via the atmospheric dump valves.
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| June 2011 15.2-18 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.6.3 Analysis of Effects and Consequences The results of the LOAC event are identical to those of the loss of reactor coolant flow event presented in subsection 15.3.1, and are no more limiting with respect to RCS pressurization than the LOCV event discussed in subsection 15.2.3. During the LOCV event the plant experiences simultaneous losses of steam and feedwater flow and condenser availability. In addition, the plant experiences a complete loss of forced reactor coolant flow during the LOAC event. The loss of forced reactor coolant flow results in an earlier reactor trip for the LOAC event (on low RCP shaft speed) compared to the reactor trip for the LOCV event (on high pressurizer pressure). The earlier trip promotes a less severe primary-to-secondary heat imbalance and hence a lower peak RCS pressure for the LOAC event.
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| The fuel performance for the LOAC is no more limiting than that for the LOF event discussed in subsection 15.3.1. The LOAC is the initiating event for the LOF so the fuel performance results of the LOF event are directly applicable to the LOAC event.
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| 15.2.6.4 Conclusions For the LOAC event and the LOAC with a concurrent single failure, the RCS pressure remains below 2750 psia thus ensuring primary system integrity, and the minimum DNBR remains above the limit thus ensuring fuel cladding integrity.
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| June 2011 15.2-19 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.7 LOSS OF NORMAL FEEDWATER FLOW 15.2.7.1 Identification of Event and Causes The loss of normal feedwater flow (LFW) event may be initiated by losing one or both main feedwater pumps or by a spurious signal being generated by the feedwater control system resulting in a closure of the feedwater control valve(s).
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| 15.2.7.2 Sequence of Events and Systems Operation LFW results in decreasing water level and increasing pressure and temperature in the steam generators. The RCS pressure and temperature also rise until a reactor trip occurs either due to low steam generator water level or high pressurizer pressure.
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| Assuming the SBCS is in the manual mode of operation, termination of main steam flow due to closure of the turbine stop valves following reactor trip temporarily causes steam generator and RCS pressurization. The decrease in core heat rate after insertion of the CEAs in combination with the main steam safety valves opening restores the RCS to a new steady state condition. Auxiliary feedwater flow is automatically initiated on a low steam generator water level, assuring sufficient steam generator inventory for core decay heat removal and cooldown to shutdown cooling entrance conditions.
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| The cooldown is operator controlled using the SBCS and the condenser.
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| 15.2.7.3 Analysis of Effects and Consequences The maximum RCS pressure for the LFW event is less than that for the LOCV event discussed in subsection 15.2.3. The LOCV event results in the termination of main steam flow prior to reactor trip in addition to the total loss of normal feedwater June 2011 15.2-20 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM flow. This additional condition aggravates RCS pressurization by further reducing the rate of primary-to-secondary heat transfer below that of the LFW event.
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| Like the LOCV, the DNBR increases during the LFW event due to the increasing RCS pressure. Thus the initial DNBR is also the minimum DNBR for the LFW event.
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| There are no concurrent single failures which when combined with LFW result in consequences more severe than the LOCV event with respect to RCS pressurization.
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| The limiting single failure with respect to fuel performance is the loss of offsite power following turbine trip. For the LFW event, prior to turbine trip the DNBR increases due to the RCS pressure increase. DNBR then briefly decreases after turbine trip due to the reactor coolant flow coastdown on loss of offsite power. The DNBR decreases similar to the DNBR transient associated with the total loss of reactor coolant flow event shown in subsection 15.3.1; however, the DNBR decrease for LFW is not as severe due to the earlier reactor trip relative to the initiation of the coolant flow coastdown.
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| Therefore, the minimum DNBR remains above the limit.
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| 15.2.7.4 Conclusions For the loss of feedwater flow event and the loss of feedwater flow with a concurrent single failure the RCS pressure remains below 2750 psia thus ensuring primary system integrity, and the minimum DNBR remains above the limit ensuring fuel cladding integrity.
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| June 2011 15.2-21 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.8 FEEDWATER SYSTEM PIPE BREAKS Feedwater line breaks (FWLBs) may occur due to pipe failures in the main feedwater system (FWS). The pipe breaks in the FWS are evaluated to confirm that the reactor coolant system is maintained in a safe status for a range of break areas up to and including double-ended breaks of the largest feedwater line.
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| The FWLB transient that results from the postulated FWS line break is sensitive to the break discharge rate. Therefore, a range of break sizes are evaluated to determine the acceptance of the most limiting Nuclear Steam Supply System (NSSS) response. Depending on the break size, location, and the plant operating conditions at the time of break, the effects of the break could cause a Reactor Coolant System (RCS) heatup (due to reduced feedwater flow to the affected steam generator) or a RCS cooldown (due to excessive energy discharge through the break).
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| In order to discuss the possible effects, FWLBs are categorized as small if the associated discharge flow is within the capacity of the FWS, and otherwise as large. Break sizes that 2
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| are less than or equal to 0.2 ft are considered as small breaks in FWLB analyses. Break locations are identified with respect to the feedwater line reverse flow check valves, which are located between the steam generator feedwater nozzles and the containment penetrations. Closure of these valves to reverse flow from the nearest steam generator maintains the integrity and would limit uncontrolled discharge of that generator in the presence of a break upstream of the valves.
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| June 2011 15.2-22 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Breaks upstream of the check valves can initiate one of the following transients. If the FWS is unavailable following the pipe failure, a total loss of normal feedwater flow (LOFW) results. With the FWS remaining in operation no reduction in feedwater flow occurs for small breaks, while large breaks impose either a partial LOFW or a total LOFW, if the break area is sufficient to discharge the entire feedwater pump flow capacity.
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| In addition to the possibility of partial or total LOFW events, breaks downstream of the check valves have the potential to establish reverse flow from the nearest steam generator (referred to as the "affected" generator) back through the break. Reverse flow occurs whenever the FWS is not operating subsequent to a pipe break or when the FWS is operating but without sufficient capacity to maintain pressure at the break above the steam generator pressure. It is only these breaks developing reverse flow that are of interest in these analyses.
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| Depending on the enthalpy of the reverse flow and the ruptured steam generator's heat transfer characteristics, the reverse flow may induce either a RCS heatup or cooldown. However, excessive heat removal through the break is not considered in this analysis, because the cooldown potential is less than that of the main steam line break events. The maximum break size is smaller for the feedwater line break events than for the Main Steam Line Break (MSLB) event. In addition, MSLBs have a greater potential for discharging high enthalpy fluid due to the location of steam piping above feedwater piping within the steam generator Furthermore, the FWLBs cause an instant reduction in feedwater flow, unlike MSLBs. Since FWLBs can cause a rapid depletion of the affected steam generators June 2011 15.2-23 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM liquid mass, reduced heat transfer capability and a rapid RCS heatup and pressurization, it is the heatup potential that is emphasized in the peak pressure analyses.
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| A general description of the FWLB event follows, assuming a break downstream of the check valves, unavailability of the FWS, and low enthalpy break discharge. The loss of subcooled feedwater flow to both steam generators causes increasing steam generator temperatures and decreasing liquid inventories and water levels. The rising secondary temperatures reduce the primary-to-secondary heat transfer and cause a heatup and pressurization of the RCS. The heatup becomes more severe as the ruptured steam generator experiences a further reduction in its heat transfer capability due to insufficient liquid inventory as the break discharge continues. This initial sequence of events culminates with a reactor trip on high pressurizer pressure, low steam generator water level, or high containment pressure. RCS heatup can continue after the reactor trip due to a total loss of heat transfer in the ruptured steam generator as it empties. Eventually the decreasing core power following reactor trip reduces the core heat rate to the heat removal capacity of the unaffected steam generator.
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| 15.2.8.1 Parametric Analysis for FWLBs Sensitivity studies are used to establish the limiting set of initial operating and transient parameters for the FWLB events with respect to RCS pressurization, long term RCS heat removal capacity of the Auxiliary Feedwater (AFW) system and pressurizer fill using the CENTS computer code. These parameters include break size and steam generator mass interval June 2011 15.2-24 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM over which heat transfer area ramps to zero M, initial core power, initial RCS pressure, initial RCS flow, initial pressurizer liquid volume, pressurizer and main steam safety valve tolerance and blowdown, core physics conditions, fuel rod gap conductance, initial core inlet temperature, initial feedwater enthalpy, and initial steam generator inventory.
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| 15.2.8.1.1 Parametrics on Key Parameters For the parametric study on the key parameters, the FWLB with loss of offsite power (LOP) methodology that considers a coinciding High Pressurizer Pressure Trip (HPPT) and Low Steam Generator Level Trip (LSGLT)(occurring at 5500 lbm liquid mass inventory) with a constant break size is applied and the initial steam generator level is adjusted to match the trips.
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| Selection of the parameters is based on the range specified by the Technical Specifications, plant configuration, and design specifications. Starting from a base case that is based on a set of configuration and assumptions, one parameter at a time is changed. The resulting pressures and levels shown in Figures 15.2.8-1 through 15.2.8-11 are used to establish the trends for the RCS peak pressure and the maximum pressurizer level for each parameter.
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| Initial pressurizer pressure affects the timing of HPPT, and is a parameter that is adjusted to result in coinciding HPPT and the LSGLT. Therefore, the effect of the initial pressurizer pressure on peak RCS and maximum pressurizer level is evaluated together with the initial steam generator inventory, the break size, and the initial temperature. The limiting initial pressurizer pressure is determined in conjunction with the June 2011 15.2-25 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM determination of the limiting break size in UFSAR Section 15.2.8.1.2.
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| A higher initial core inlet temperature increases the initial steam generator pressure and reduces primary to secondary heat transfer. A higher temperature also results in higher RCS energy content to be removed. Increased initial steam generator pressure due to higher temperature results in an earlier opening of the Main Steam Safety Valves (MSSVs), which increases the heat removal by the secondary system during the transient. Depending on the other conditions and time of trip, timing of these competing effects during the transient can make the peak pressure or pressurizer level more adverse or more benign. Therefore, a further study was conducted in conjunction with the break size and initial pressure in UFSAR Section 15.2.8.1.2.
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| 15.2.8.1.2 Parametrics on Limiting Break Size and Steam Generator Heat Transfer Characteristics In order to determine the sensitivity of the RCS pressurization to the ruptured steam generator heat transfer characteristics, the effective heat transfer area was conservatively assumed to decreased linearly (from the design value to zero) as the steam generator liquid mass decreased (from a selected value to zero). The mass interval over which the ramp down is assumed to occur is referred to as M. Decreasing values of M imply a more rapid loss of heat transfer in the ruptured steam generator.
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| After the most limiting key initial parameters are determined as described in UFSAR Section 15.2.8.1.1, the parametric study June 2011 15.2-26 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM on break size and M was performed. In addition to the break size and M, different combinations of the initial pressurizer pressure and initial core inlet temperature are considered since the combinations of these effect the primary to secondary heat transfer characteristics, break flow characteristics and the heat removal by Primary Safety Valves (PSVs) and MSSVs during the transient. The reactor trip is modeled to occur either on high pressurizer pressure trip or steam generator liquid mass. The trip will occur either on the LSGLT for larger break sizes since the steam generator empties faster, or on the HPPT for smaller break sizes since the RCS pressurization is steeper. Effects of initial pressurizer pressure and core inlet temperature are evaluated by starting the event at a minimum steam generator level. Results indicate that the current methodology of matching the HPPT with the LSGLT is limiting.
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| The limiting break size, and initial pressurizer pressure and core inlet temperatures are then evaluated for two heat transfer degradation assumptions per methodology; =0 for FWLBs with LOP, and =30,000 for small FWLBs without LOP and a single failure.
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| M =0 Case:
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| For the instantaneous loss of heat transfer (M =0), the most adverse peak pressure is obtained when the event starts from the maximum initial core inlet temperature and minimum initial pressurizer pressure (Figure 15.2.8-12). The limiting break size is the one that results in simultaneous trips on HPPT and LSGLT, which is chosen to occur when 5500 lbm liquid mass inventory is left in the steam generator.
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| June 2011 15.2-27 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM If the LSGLT occurs much earlier than HPPT, as for larger break sizes, RCS pressure and temperature are lower at the time of trip. In addition, the larger break sizes provide additional heat removal by the energy release from the break. Thus, the complete loss of heat transfer in the affected steam generator has less severe effect for very large breaks since it is off-set by this additional energy removal through the break, resulting in lower RCS peak pressure.
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| If the trip on HPPT occurs much earlier than the LSGLT, as for smaller break sizes, the left over liquid inventory in the affected steam generator provides cooldown of the RCS through the break until the heat transfer is lost at low steam generator inventory. Furthermore, for smaller break sizes, additional cooling becomes available when the PSVs open. For very small break sizes, the cooldown by the PSVs become more effective than the cooldown by the break. The cooling effect of the PSVs can be seen by looking at the results for very small break sizes. Below a certain break size, the cooldown by the leftover steam generator inventory is not significant, and the peak RCS pressure is driven by the PSVs. The peak pressure occurs much earlier than the loss of heat transfer, and is not affected by the break sizes.
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| As illustrated in Figure 15.2.8-12, lower initial pressure results in later simultaneous trips, shifting the peak pressure vs. break size curve to the left. In other words, decreasing initial pressure (or smaller break size) results in higher peak pressure. Lower initial core inlet temperature causes additional delay of the simultaneous trips, resulting in smaller break sizes for matching trips. However, this results in lower peak RCS pressure due to the decreased energy content June 2005 15.2-28 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM of the primary system at lower inlet temperatures. Therefore, for peak pressure determination, the limiting break size is the smallest break size that would result in simultaneous trips for the event starting from highest core inlet temperature and lowest pressurizer pressure.
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| The effect of initial core inlet temperature and initial pressurizer pressure on maximum pressurizer level for different break sizes is shown in Figure 15.2.8-13. Simultaneous HPPT with LSGLT also produces the most adverse pressurizer level.
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| Lower initial pressurizer pressure results in a later trip which provides more time for pressurizer inventory addition before trip. Although the parametric study for M=0 shows that the maximum initial core inlet temperature is more limiting, the pressurizer volume is also increased by decreasing temperatures for smaller break sizes. Since the pressurizer level is evaluated for the long term cooling scenario, the later peaks of maximum pressurizer level are more of a concern.
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| In the long term cooling case, the primary system is cooled by the AFWS, relief from the PSVs, MSSVs, and energy release from the break following the steam generator dryout. For these cooling mechanisms, the lower temperature and smaller break combination results in less heat removal. This is because the higher core inlet temperature (or higher initial steam generator pressure) causes the MSSVs to open early, and larger breaks would discharge more energy. Also, a lower core inlet temperature maximizes the swelling affect due to the density change in the RCS thus maximizing the pressurizer liquid volume increase in the long term. Considering these effects, it is determined that the minimum temperature and minimum break size June 2005 15.2-29 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM would produce most adverse result in terms of the pressurizer level for the long term cooling case.
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| In conclusion, the parametrics have determined that for the M=0 case, the smallest break size that would produce simultaneous HPPT and steam generator dryout combined with the lowest initial core inlet temperature and lowest initial pressurizer pressure is the most limiting case in terms of long term heat removal.
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| M =30000 Case:
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| For the degradation of heat transfer over a mass range corresponding to M =30,000, the most adverse peak pressure is obtained when the event starts from the maximum initial core inlet temperature and maximum initial pressurizer pressure (Figure 15.2.8-14). If the LSGLT (corresponding to 35,000 lbm) occurs much earlier than HPPT, the larger break sizes provide cooldown of the primary by the energy release from the break compensating for the heat transfer degradation. If the trip on HPPT occurs much earlier than the degradation of heat transfer, the cooldown by the PSVs and by the break is adequate enough to limit the RCS pressure prior to the heat transfer degradation.
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| For smaller size breaks, PSVs may open earlier during the heat transfer degradation making the effect of reduced heat transfer insignificant. As illustrated in Figure 15.2.8-14, larger break sizes result in higher peak pressure since the heat transfer degradation is faster and the PSVs open late, shifting the peak pressure vs. break size curve to the right. Competing effects of cooling by the PSVs and the heat transfer degradation can be seen by looking at the results for very small break sizes (Figure 15.2.8-15). Below a certain break June 2005 15.2-30 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM size, the heat transfer degradation is not significant since it is slower than the pressure increase, and the peak RCS pressure is driven by the PSVs. The peak pressure occurs much earlier than the total loss of heat transfer. Also, for small break sizes, the maximum peak pressure occurs when the LSGLT approaches the HPPT. Therefore, the limiting break size is the largest break size that would give coinciding HPPT and LSGLT with the highest initial core inlet temperature and pressurizer pressure. The largest break size for small FWLBs is 0.2 ft ,
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| 2 by definition.
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| Also in Figure 15.2.8-14, the effect of the initial core inlet temperature shows variation by break size. This is due to the additional cooldown provided by the opening of MSSVs. When the initial core inlet temperature is higher, the initial steam generator pressure is higher, which results in earlier opening of MSSVs.
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| The limiting combination of initial core inlet temperature and initial pressurizer pressure on maximum pressurizer level is the same for the M =0 and the M =30000 cases. The effect of lower core inlet temperature and the smaller break size is more noticeable in Figure 15.2.8-15.
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| 15.2.8.1.3 Parametric Study on Single Failure for FWLB without LOP The FWLB without LOP methodology assumes a single failure of one of the two fast bus transfer circuits resulting in a Failure to Fast Bus Transfer (FFBT). Normally, both fast bus transfer circuits are operable. However, there is no restriction on removing one or both circuits from service for a period of time during the normal operations. Therefore, the June 2005 15.2-31 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM limiting single failure from Table 15.0-0 for the FWLB without LOP methodology was investigated. It was determined that there are no credible single failures (see Table 15.0-0) which, in combination with one or both fast bus transfer circuits blocked, would make the consequences of the event more severe.
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| A two Reactor Coolant Pump (RCP) coastdown, assuming both fast bus transfer circuits are in service and a single failure on one of them occurs, is more limiting than either a total LOP (four RCP coastdown) or full fast bus transfer (no RCP coastdown).
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| 15.2.8.2 Feedwater Line Break Event with Loss of Offsite Power Analysis of the limiting FWLB event with a LOP was performed using the CENTS computer code along with several simplifying assumptions which, with respect to RCS overpressurization, conservatively model the break discharge flow and enthalpy and the ruptured steam generator water level and heat transfer.
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| Blowdown of the steam generator nearest the FWLB is modeled assuming frictionless critical flow as calculated by the Henry-Fauske correlation (reference 4). Although the enthalpy of the blowdown depends on the location of the break and fluid conditions within the affected steam generator, it is assumed that saturated liquid is discharged until no liquid remains, at which time saturated steam discharge is assumed.
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| With respect to RCS overpressurization these assumptions result in conservatively high mass, low energy flow from the break, thereby minimizing the ruptured generator heat removal capacity.
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| June 2005 15.2-32 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM No credit is taken for a LSGLT in the affected steam generator until the generator is emptied of liquid. This conservatively delays the time of reactor trip, prolonging the RCS heatup and overpressurization. Additionally, no credit is taken for the high containment pressure trip.
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| 15.2.8.2.1 Identification of Causes and Frequency Classification The FWLB event may occur due to a pipe failure in the FWS.
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| A FWLB with a LOP is classified as a limiting fault event.
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| 15.2.8.2.2 Sequence of Events and Systems Operation The sequence of events for the FWLB with LOP is presented in Table 15.2.8-1.
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| The FWLB event with LOP is initiated by a 0.21 ft break that 2
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| is assumed to occur between the steam generator economizer feedwater nozzle and its associated feedwater line check valve An instantaneous total loss of main feedwater flow to both steam generators is assumed. Critical flow is established from the affected steam generator through the feedwater line break.
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| In addition, it is assumed that the FWS is unavailable and that the break discharge enthalpy remains as saturated liquid until the affected steam generator empties, at which time saturated steam enthalpy is assumed. The loss of subcooled feedwater flow to both steam generators causes increasing steam generator temperatures and decreasing liquid inventories. This reduces the primary-to-secondary heat transfer rate, resulting in increased RCS temperature and pressure.
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| June 2005 15.2-33 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The affected steam generator is assumed to instantaneously lose all heat transfer capability due to total depletion of its liquid inventory by boil off and break discharge flow. This initiates a rapid heatup and pressurization of the RCS and depressurization of the steam generators. A reactor trip occurs on high pressurizer pressure, which is coincident with a trip signal (LSGLT) on low steam generator water level. A turbine trip on reactor trip occurs followed by a LOP and closure of the turbine admission valves (TAVs). The closing of the TAVs leaves the feedwater line break as the only steam discharge path. This results in steam generator pressurization which reduces the primary-to-secondary temperature difference and heat transfer rate, thus continuing the RCS heatup. In addition, the loss of reactor coolant flow following the LOP decreases the rate of heat removal in the steam generator tubes, resulting in a significant reduction of heat removal from the RCS.
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| The reduction of primary side heat removal causes RCS volumetric expansion that results in compression of the pressurizer steam volume due to insurge flow through the pressurizer surge line, which increases the pressurizer pressure above the primary safety valve (PSV) setpoint. When the pressurizer pressure reaches the PSV setpoint, the PSVs open. The maximum RCS pressure remains below 120% of design pressure. The primary side heatup rate lowers due to the decrease in core decay heat flux, which results in a decrease in RCS pressure.
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| The MSSVs open, limiting secondary side pressure and stabilizing the temperature. This allows a greater heat transfer rate to the unaffected steam generator. The June 2005 15.2-34 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM unaffected steam generator maximum pressure, remains below 120%
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| of design pressure.
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| As a result of steaming through the MSSVs, the unaffected steam generator water level decreases and initiates an auxiliary feedwater actuation signal (AFAS). During the transient, the pressurizer water volume remains below the PSV nozzle elevation.
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| The secondary side pressure decreases due to a cooldown attributed to auxiliary feedwater (AFW), PSV and MSSV action, and steam flow from the reverse direction through the affected steam generator and out the feedwater line break. A main steam isolation signal (MSIS) is generated on a low steam generator pressure, which closes the main steam isolation valves (MSIVs),
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| and isolates the unaffected steam generator from the affected steam generator. After the MSIVs close, the pressure difference between the affected and unaffected steam generators increases and reaches the AFW P-lockout setpoint. At this time, the AFW is fully diverted to the unaffected steam generator, restoring its water level. The unaffected steam generator re-pressurizes, causing a reduction in heat transfer and subsequent primary system heatup and pressurization. The primary-to-secondary heat and pressure imbalance is eliminated shortly after the re-opening of the MSSVs. The NSSS enters a quasi-steady state with a gradual cooldown and depressurization due to decreasing core decay heat generation. After 1800 seconds the operators initiate a controlled cooldown to shutdown cooling entry conditions, using the atmospheric dump valves (ADVs).
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| June 2005 15.2-35 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The FWLB with LOP analysis conservatively assumes operator action is delayed until 30 minutes after the occurrence of the event.
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| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions are consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| The NSSS is protected during this transient by the primary safety valves (PSVs) and the following trips:
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| * Steam Generator Low Level
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| * Steam Generator Low Pressure
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| * High Pressurizer Pressure
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| * Low Departure from Nucleate Boiling Ratio (DNBR)
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| * High Containment Pressure
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| * Variable Overpower Trip Depending upon the initial conditions, any one of these trips may terminate this transient. The NSSS is also protected by MSIVs, feedwater line check valves, MSSVs, and the AFWS, all of which serve to maintain the integrity of the secondary heat sink following reactor trip.
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| In considering the peak pressure criteria for this event and a postulated worst single active component failure in a system required to mitigate the transient, UFSAR Table 15.0-0 was used. As a result of the evaluation method applied to this analysis, the only mechanisms for mitigation of the RCS and June 2009 15.2-36 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM main steam pressurization are the PSVs, MSSVs, and RCS flow.
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| The RCS flow and MSSVs influence the RCS-to-steam generator heat transfer rate.
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| There are no credible failures that can degrade the PSV or MSSV capacity. Technical Specifications place limits on reactor power and variable overpower trip (VOPT) setpoints when one or more MSSVs are inoperable, thereby ensuring primary and secondary system peak pressure remains within applicable maximum pressure limits. The FWLB event is one of the transients analyzed for validating Technical Specification 3.7.1. A decrease in RCS-to-steam generator heat transfer due to reactor coolant flow coastdown can be caused by a FFBT following turbine trip or by a LOP following turbine trip (i.e., two RCP or four RCP coastdown, respectively). In this analysis a LOP is assumed to occur following a turbine trip, which results in a four pump coastdown. In addition, it is assumed that the most reactive control rod fails to insert on scram. Therefore, for the FWLB event with a reactor trip followed by a turbine trip and a LOP, there is no credible single failure to make the event consequences more adverse with respect to primary peak pressure.
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| The FWLB long term cooling event presented in UFSAR Section 15.2.8.4 evaluates the single failure of one auxiliary feedwater pump that results in reduced secondary side heat removal capacity.
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| June 2005 15.2-37 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.8-1 SEQUENCE OF EVENTS FOR THE FEEDWATER LINE BREAK EVENT WITH LOSS OF OFFSITE POWER FOR PEAK PRESSURE and FUEL PERFORMANCE Time (sec)
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| RTP RTP Events 3876 3990 MWt MWt 0.0 0.00 FWLB occurs. Complete loss of normal feedwater to both steam generators occurs 24.41 27.90 Pressurizer pressure reaches HPPT setpoint 24.59 28.11 Dryout of affected steam generator (5,000 lbm of 3
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| liquid inventory). AFAS generated in affected SG 24.91 28.40 Reactor trip breakers open 24.91 28.40 Turbine trip occurs 24.91 28.40 LOP occurs 25.51 29.00 Scram CEAs begin falling 26.11 29.55 PSVs open 27.92 31.43 Maximum RCS pressure 30.88 30.11 MSSV bank 1 opens on unaffected steam generator 31.30 30.43 MSSV bank 1 opens on affected steam generator
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| --- 33.78 MSSV bank 2 opens on unaffected steam generator 32.11 35.78 PSVs close 33.95 33.88 Peak secondary pressure occurs 35.29 38.92 Maximum liquid volume of pressurizer
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| --- 44.93 MSSV bank 2 closes on unaffected steam generator 46.01 52.79 MSSV bank 1 closes on affected steam generator 47.09 55.52 MSSV bank 1 closes on unaffected steam generator
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| < 1800 < 1800 Long-term automatic plant system actions and NSSS response to this transient are similar to the long-term cooling FWLB event 1800.0 1800.0 Operator initiates plant cooldown 3
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| Although AFAS is assumed to occur when the steam generator reaches dryout, no AFW flow is delivered to the steam generators during the 60 seconds of the transient analyzed because of the 46-second delay time, nor is a MSIS or AFW lockout generated. The responses of these systems are similar to Long-Term Cooling FWLB transient presented in UFSAR Section 15.2.8.4.
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| June 2005 15.2-38 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.8.2.3 Core and System Performance A Mathematical Model The NSSS response to the FWLB with LOP was simulated using the CENTS computer code described in UFSAR Section 15.0.3.1.3.2. Inputs to the CENTS code such as moderator reactivity as a function of moderator density, Doppler reactivity as a function of effective fuel temperature, and shutdown rod worth were calculated using the two-dimensional ROCS code discussed in UFSAR Section 4.3.3.1.1.2. The shutdown rod worth assumes that the most reactive control rod fails to insert on scram. Input to the CENTS code may also be calculated using the SIMULATE-3 code discussed in UFSAR Section 4.3.3.1.1.5.
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| The DNBR for the core hot channel was calculated using the CETOP computer code (see UFSAR Sections 4.4 and 15.0.3.1.6) which uses the CE-1 critical heat flux (CHF) correlation (Reference 2). Transient dependent input to CETOP such as RCS pressure, coolant flowrate through the core, core inlet temperature, and core average heat flux are obtained from the transient response predicted by CENTS.
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| The methodology for FWLB with LOP applies to a whole spectrum of feedwater line breaks, occurring with a LOP resulting from a turbine trip and a limiting single failure. FWLB event with LOP and a single failure is subject to ASME Service Level C pressure limit (120% of design pressure) due to the very low estimated frequency of occurrence.
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| June 2005 15.2-39 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The analysis methods address the influence of the four major controlling parameters:
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| * discharge enthalpy
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| * discharge flow
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| * low water level trip condition in the ruptured steam generator
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| * heat transfer characteristics of the ruptured steam generator.
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| The principal conservative assumptions and analytical methods utilized in the analysis of this event include:
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| * Conservative estimation of the break flow and enthalpy, i.e. discharge of saturated liquid until steam generator is dry (steam generator is considered dry when 5000 lbm or less of liquid inventory is left).
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| * Delay of heat transfer degradation in the affected steam generator until the liquid inventory is depleted and then assuming an instantaneous loss of heat transfer.
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| * Initializing key parameters such that a reactor trip occurs on high pressurizer pressure coinciding with LSGLT, which is delayed until liquid mass inventory in the affected steam generator is depleted.
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| * Delaying the AFAS until liquid mass inventory in the affected steam generator is depleted.
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| June 2005 15.2-40 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a FWLB with LOP from full power conditions are presented in Table 15.2.8-2. The input parameters and initial conditions are based on the parametric studies discussed in UFSAR Section 15.2.8.1, selected to maximize the consequences of the FWLB.
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| The input parameters used in this analysis include:
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| * Maximum initial core power - Maximum core power maximizes the heat content of the primary system and the amount of energy to be removed by the secondary system. This results in a larger heat up and pressurization of the primary and secondary systems.
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| * Maximum initial core inlet temperature - This maximizes the amount of energy to be removed from the primary system following a loss of heat sink, and increases the initial steam generator pressure. This results in a reduced primary to secondary heat transfer, and a higher RCS peak pressure.
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| * Maximum initial RCS flow - For a given core power and core inlet temperature, maximum RCS flow results in a lower hot leg temperature and thus a lower steam generator temperature. This results in decreased break flow and enthalpy. The decreased break flow and enthalpy (i.e., decreased heat removal through the break) prior to trip increases the primary system pressurization.
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| June 2005 15.2-41 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * Minimum initial pressurizer pressure - Sensitivity studies show that the minimum pressurizer pressure and maximum initial core inlet temperature result in the limiting RCS pressurization.
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| * Minimum Initial Core Average Fuel Rod Gap Conductance
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| - Minimum fuel rod gap conductance delays the heat transfer from the fuel to the reactor coolant. This increases the energy content of the primary system after trip, resulting in higher primary and secondary peak pressures.
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| * Maximum PSVs opening setpoint - This delays the pressure relief of primary system and heat removal through the PSVs, thus maximizing the peak primary pressure.
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| * Minimum initial pressurizer volume - This delays the HPPT due to steam cushioning effect, thereby resulting in a higher primary peak pressure.
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| * Minimum initial steam generator level - A lower initial steam generator inventory results in earlier degradation in heat transfer, earlier emptying of the affected and intact steam generators and a higher RCS pressure.
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| * Minimum initial feedwater enthalpy - This minimizes the heat removal capability of the affected steam generator, thereby contributing to primary side heatup and results in higher RCS pressure.
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| * Least negative (most positive) moderator temperature coefficient (MTC) - This reduces the negative June 2005 15.2-42 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM reactivity insertion into the core due to coolant heat up during the event, thus resulting in a slower decrease in power and higher heat content of the primary.
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| * Maximum number of plugged steam generator tubes -
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| Increasing the number of plugged steam generator tubes decreases the heat transfer from primary to secondary side due to the reduced steam generator heat transfer surface area. This contributes to RCS heatup and pressurization.
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| * Limiting break size - The limiting FWLB with LOP event break size is determined by parametric study discussed in UFSAR Section 15.2.8.1.2.
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| C. Results The response of key core parameters as a function of time following a FWLB and LOP break are provided in Table 15.2.8-1 and Figures 15.2.8-16 through 15.2.8-18 and 15.2.8-31.
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| The FWLB transient DNBR values are dependent upon variations in RCS pressure, core inlet temperature, core flow and core heat flux. The immediate loss of feedwater results in increased RCS temperature and pressure. The increase in RCS temperature and pressure is initially gradual since the steam flow has not been immediately interrupted. The increasing RCS pressure tends to increase the DNBR value, however, the increasing RCS temperature tends to stabilize the DNBR and result in a rather flat trace as depicted in figure 15.2.8-31. The LOP and coastdown of the RCPs significantly reduces flow June 2005 15.2-43 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.8-2 ASSUMED INITIAL CONDITIONS FOR THE FEEDWATER LINE BREAK EVENT WITH LOSS OF OFFSITE POWER PEAK PRESSURE AND FUEL PERFORMANCE EVENT Value Parameter RTP RTP 3876 3990 MWt MWt Initial core power(% of RTP) 102 102 Initial core inlet temperature (°F) 562 566 Initial pressurizer pressure, psia 2100 2100 Initial RCS flow, (% of design) 116 116 Initial pressurizer level (%) 24 24 Initial steam generator level (% of WR) 61.6 60.7 Initial feedwater enthalpy (Btu/lbm) 397 428 Moderator Temperature Coefficient (/°F) 0.0 0.0 Least Least Fuel Temperature Coefficient negative negative SCRAM delay time (sec) 0.5 0.5 CEA holding coil delay (sec) 0.6 0.6 CEA worth of trip-WRSO (%) 8.0 8.0 Fuel rod gap conductance (Btu/hr-ft2-°F) 500 500 Plugged tubes per Steam Generator 1750 1258 PSV setpoint tolerance +3% +3%
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| PSV Blowdown 5% 5%
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| MSSV setpoint tolerance +3% +3%
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| MSSV Blowdown 5% 5%
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| Single Failure None None LOP Yes Yes 2
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| Feedwater pipe break area (ft ) 0.207 0.139 June 2005 15.2-44 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM to the core. This in conjunction with rising coolant temperature, positive MTC, reactivity feedbacks, and core power result in overcoming the effect of increased pressure on DNBR. The DNBR sharply decreases for a few seconds and reaches its lowest value, but remains above the Specified Acceptable Fuel Designed Limit (SAFDL) for DNBR of 1.34. The decrease in core heat flux after trip, increase in RCS pressure and stabilizing RCS temperature, results in a sharp increase in DNBR value.
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| Since there is no power excursion during the transient, the FWLB event does not challenge the peak fuel centerline temperature SAFDL or the limit on linear heat generation rate (21 kW/ft).
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| The FWLB with LOP transient minimum DNBR (see Figure 15.2.8-31) is greater than the DNBR SAFDL value of 1.34, and therefore meets the acceptance criteria documented in Reference 3. Fuel cladding damage does not occur for this limiting fault event.
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| 15.2.8.2.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate RCS barrier performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.2.3.A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed in the computer codes to evaluate RCS barrier June 2005 15.2-45 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.2.3.B.
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| C. Results The response of key RCS parameters as a function of time is presented in Figures 15.2.8-19 through 15.2.8-30 for this limiting fault event.
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| The FWLB with LOP assumes a loss of main feedwater to both steam generators resulting in a reduction of steam generator water inventory, pressurization of the secondary side, and a resulting heatup and pressurization of the primary side. The primary side heatup causes volumetric expansion and increase in pressurizer water level and pressure. A reactor trip occurs on a HPPT followed by a concurrent LOP on turbine trip. The turbine trip causes TAV closure. This reduction in primary to secondary heat transfer causes a rapid heatup of primary side coolant and the PSVs open to limit pressure. The MSSVs open to limit secondary side pressure.
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| For 3876 MWt, the maximum RCS pressure reaches 2749 psia, which is less than 120% (3000 psia) of RCS design pressure (2500 psia). The maximum secondary system pressure is 1317 psia, which is less than 120% (1524 psia) of the secondary system design pressure (1270 psia).
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| For 3990 MWt, the maximum RCS pressure reaches 2778 psia, which is less than 120% (3000 psia) of RCS design pressure (2500 psia). The maximum secondary system pressure is 1354 psia, which is less than 120% (1524 psia) of the secondary system design pressure (1270 psia).
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| June 2005 15.2-46 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The maximum primary and secondary system pressures for this event meet the limiting pressure acceptance criteria of the Standard Review Plan.
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| As explained in UFSAR Section 15.2.8, FWLB is analyzed as a heat-up transient, and the cooldown potential of a FWLB is less than that of a MSLB. Therefore, the potential of reactor vessel being subject to brittle fracture (GDC 35) is bounded by the MSLB.
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| 15.2.8.2.5 Radiological Consequences and Containment Performance Fuel damage is not predicted for this limiting fault event.
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| During this event, three sources of radioactivity contribute to the site boundary dose; the initial activity in the steam generator inventory, the activity associated with primary-to-secondary leakage from the steam generator tubes and releases from the reactor drain tank. These sources are assumed to be at 0.1 µCi/gm for the initial steam generator inventory, and at 1.0 µCi/gm dose equivalent I-131 for the other sources, respectively. Analysis methodologies are the same as those used in UFSAR Section 15.1.5.5 for the limiting fault MSLB event. The dose analysis for FWLB with LOP uses bounding MSLB release rates, determined independently of break size.
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| Assuming all of the radioactivity is released to the atmosphere, the offsite dose due to the feedwater line break with loss of offsite power results in no more than 3.1 REM two-hour inhalation thyroid dose at the exclusion area boundary (EAB) and 1.7 REM eight hour inhalation thyroid dose at the low population zone (LPZ) boundary. Whole body doses do not exceed 0.1 REM for either two-hour EAB or eight-hour LPZ. Since no June 2007 15.2-47 Revision 14
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM fuel failure is predicted, no containment isolation is credited. In addition, the control room dose is bounded by the control room dose analyzed for events in UFSAR Section 6.4.7.3.
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| 15.2.8.2.6 Conclusions For the FWLB with a LOP resulting from a turbine trip, the maximum RCS pressure remains below 120% (3000 psia) of design pressure, thus ensuring primary system integrity. Likewise, the maximum secondary system pressure remains below 120% (1524 psia) of design pressure.
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| The minimum DNBR remains above the SAFDL limit, thereby ensuring fuel cladding integrity. All dose consequences are well within the 10CFR100 guidelines.
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| 15.2.8.3 Feedwater Line Break Event With Offsite Power Available and Limiting Single Failure The FWLB with LOP event presented in UFSAR Section 15.2.8.2 shows that the limiting break size, when combined with the LOP, produces the maximum primary and secondary pressures below 120%
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| of the design values. The NRC has stated (Reference 3) that 120% of design maximum pressure criterion is appropriate for FWLB combined with the LOP. However, the NRC also stated in reference 3, that it must be shown that small break loss of feedwater inventory events with the limiting single failure and offsite power available meet the maximum pressure criterion of 110% of the design value. In order to demonstrate compliance with this criterion, a conservative analysis method was developed for the FWLB, with a break size area less than or 2
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| equal to 0.2 ft with offsite power available. This is also referred to as a "small" feedwater line break event because of June 2007 15.2-48 Revision 14
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM the analysis method assumed for the steam generator heat transfer characteristics.
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| 15.2.8.3.1 Identification of Causes and Frequency Classification The feedwater line break with a single failure and offsite power available may occur due to a pipe failure in the FWS.
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| The FWLB with a limiting single failure and offsite power available is classified as a limiting fault event.
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| 15.2.8.3.2 Sequence of Events and Systems Operation The sequence of events presented in Table 15.2.8-3 summarizes the important plant system responses for the FWLB with a limiting single failure and offsite power available.
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| The FWLB with single failure and offsite power available is initiated by a 0.20 ft2 break or less that is assumed to occur between the steam generator economizer feedwater nozzle and its associated feedwater line check valve and results in an instantaneous total loss of main feedwater flow to both steam generators. Critical flow is established from the affected steam generator through the feedwater line break. In addition, it is assumed that the FWS is unavailable and that the break discharge enthalpy remains as saturated liquid until the affected steam generator empties, at which time saturated vapor enthalpy is assumed. The loss of subcooled feedwater flow to both steam generators causes increasing steam generator temperatures and decreasing liquid inventories. This causes a reduction in the primary-to-secondary heat transfer rate and an increase RCS temperature and pressure.
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| June 2005 15.2-49 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM As a result of heat transfer degradation due to insufficient water inventory in the affected steam generator a rapid heatup and pressurization of the RCS occurs, generating a HPPT signal that is coincident with a LSGLT signal. The reactor trip breakers open followed by an assumed instantaneous turbine trip and closure of TAVs. The closure of the TAVs leaves the feedwater line break as the only steam discharge path. This results in steam generator pressurization and reduction in primary-to-secondary heat transfer rate, which causes a RCS heatup. Immediately following turbine trip, a FFBT is assumed and results in the coastdown of two RCPs. The two-pump loss of reactor coolant flow decreases the rate of heat removal in the steam generator tubes resulting in a significant reduction of heat removal from the RCS.
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| The reduction of primary side heat removal causes RCS volumetric expansion that results in compression of the pressurizer steam volume due to insurge flow through the pressurizer surge line that increases the pressurizer pressure above the PSV setpoint. When the pressurizer pressure reaches the PSV setpoint, the PSVs open and offset the pressurization by releasing steam. The maximum RCS pressure remains below 110% of design pressure. The primary side heatup rate lowers due to the decrease in core decay heat flux, which results in a decrease in RCS pressure.
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| The MSSVs open, stabilizing the secondary side temperature.
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| This allows a greater heat transfer rate to the unaffected steam generator. The maximum unaffected steam generator pressure remains below 110% of design pressure.
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| As a result of steaming through the MSSVs, the unaffected steam generator water level decreases and initiates an AFAS.
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| June 2005 15.2-50 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM The secondary side pressure decreases due to a cooldown attributed to AFW, PSV and MSSV action, and steam flow from the reverse direction through the affected steam generator and out the feedwater line break. A MSIS is generated on low steam generator pressure, which closes the MSIVs, and isolates the unaffected steam generator from the affected steam generator and the feedwater line break. After the MSIVs close, the pressure difference between the affected and unaffected steam generators increases and reaches the AFW P lockout setpoint.
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| At this time, the AFW is fully diverted to the unaffected steam generator, restoring its water level. The unaffected steam generator re-pressurizes, causing a reduction of heat transfer and subsequent primary system heatup and pressurization. The primary-to-secondary heat and pressure imbalance is eliminated shortly after the re-opening of the MSSVs. The NSSS enters a quasi-steady state with a gradual cooldown and depressurization due to decreasing core decay heat generation. After 1800 seconds the operators initiate a controlled cooldown to shutdown cooling entry conditions, using the ADVs.
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| The analysis of the FWLB with a single failure and offsite power available conservatively assumes operator action is delayed until 30 minutes after the occurrence of the initiating event.
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| Analytical setpoints and response times associated with the RPS trip functions and ESFAS functions are consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| The NSSS is protected during the transient by the PSVs and the following trips:
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| June 2005 15.2-51 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * Steam Generator Low Level
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| * Steam Generator Low Pressure
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| * High Pressurizer Pressure
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| * Low Departure from Nucleate Boiling Ratio (DNBR)
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| * High Containment Pressure
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| * Variable Overpower Trip.
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| Depending upon the particular initial conditions, any one of these trips may terminate this transient. The NSSS is also protected by MSIVs, feedwater line check valves, MSSVs, and the AFWS which serve to maintain the integrity of the secondary heat sink following reactor trip.
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| In considering the peak pressure criteria for this event and a postulated active single component failure in a system required to mitigate the transient, Table 15.0-0 was used. As a result of the evaluation method applied to this analysis, the only mechanisms for mitigation of the RCS and main steam pressurization are the PSVs, MSSVs, and RCS flow. The RCS flow and MSSVs influence the RCS-to-steam generator heat transfer rate.
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| Table 15.0-0 indicates that there are no credible failures that can degrade the PSV or MSSV capacity. Technical specification 3.7-1 places limits on reactor power and VOPT setpoints when one or more MSSVs are inoperable, thereby ensuring primary and secondary system peak pressure remains within 110% of system design pressure. A decrease in RCS-to-steam generator heat transfer due to reactor coolant flow coastdown can be caused by a FFBT following turbine trip or LOP following turbine trip June 2005 15.2-52 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (i.e., two RCP or four RCP coastdown, respectively). Because offsite power is assumed to be available for this analysis, a fast bus transfer will occur following turbine trip if the transfer buses are available. Assuming both transfer buses are available, a FFBT is assumed following the turbine trip, which results in the coastdown of two RCPs in diagonally opposite loops. It has been determined by parametric analysis in UFSAR Section 15.2.8.1.3 that this plant configuration is limiting for the event. There is no other credible single failure, besides FFBT, to make the event consequences more adverse with respect to primary peak pressure. In addition, it is assumed that the most reactive control rod fails to insert on scram.
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| The FWLB with LOP long term cooling event presented in UFSAR Section 15.2.8.4 evaluates the single failure of one AFW pump that results in reduced secondary side heat removal capacity.
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| 15.2.8.3.3 Core and System Performance A. Mathematical Model The computer codes used to simulate the NSSS and core thermal-hydraulic response to the FWLB with single failure and offsite power available are the same as those described in UFSAR Section 15.2.8.2.3.A.
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| The methodology used in the analysis of FWLB with single failure and offsite power available, (small break loss of feedwater inventory events), is the same as that described and applied in UFSAR Section 15.2.8.2, with the exception of the treatment of steam generator heat transfer and reactor trip on steam generator low water level.
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| June 2005 15.2-53 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.8-3 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK WITH LIMITING SINGLE FAILURE AND OFFSITE POWER AVAILABLE Time (sec)
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| RTP RTP Event 3876 3990 MWt MWt FWLB occurs. A complete loss of normal feedwater to 0.00 0.00 both steam generators occurs 18.53 18.14 Pressurizer pressure reaches HPPT setpoint 19.03 18.64 Reactor Trip breakers open 19.03 18.64 Turbine Trip occur and FFBT 19.63 19.24 Scram CEAs begin falling 21.24 20.81 PSVs open 21.87 21.48 Maximum RCS pressure occurs 22.54 21.04 MSSV bank 1 opens on unaffected steam generator 22.56 21.04 MSSV bank 1 opens on affected steam generator 24.60 22.68 MSSV bank 2 opens on unaffected steam generator 24.84 22.78 MSSV bank 2 opens on affected steam generator 24.62 24.51 PSVs close 25.72 25.64 Peak secondary pressure occurs Affected steam generator dries out. AFAS is 27.36 26.46 4 generated in affected steam generator 30.31 28.00 Maximum liquid volume of pressurizer occurs 31.05 32.14 MSSV bank 2 closes on affected steam generator 31.48 32.63 MSSV bank 2 closes on unaffected steam generator 38.33 40.08 MSSV bank 1 closes on affected steam generator 40.75 42.74 MSSV bank 1 closes on unaffected steam generator Long-term automatic plant system actions and NSSS
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| < 1800 < 1800 response to this transient are similar to the long-term cooling FWLB event 1800.0 1800.0 Plant cooldown is initiated 4
| |
| Although AFAS is assumed to occur when the steam generator reaches dryout, no AFW flow is delivered to the steam generator during the transient because of the 46 second delay time assumed nor is a MSIS or AFW lockout generated. The response of these systems are similar to Long-Term Cooling FWLB transient presented in UFSAR Section 15.2.8.4.
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| June 2005 15.2-54 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Predictions of steam generator heat transfer and level behavior are based on the model documented in references 5 through 8. As discussed below, this model is conservative when applied to the small break loss of feedwater inventory events.
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| Steam Generator Heat Transfer RCS pressurization is largely a function of the rate of heat transfer decrease by the affected steam generator as its inventory is depleted. UFSAR Section 15.2.8.1.2 documents the sensitivity of RCS pressurization to steam generator heat transfer behavior. The study verified that RCS pressurization is maximized by underestimating the affected steam generator liquid mass corresponding to the initiation of heat transfer degradation (i.e.,
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| overestimating the rate of heat transfer decrease). The original methodology took a simplistic and clearly conservative approach by assuming heat transfer degradation was instantaneous upon steam generator dryout.
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| However, this approach is modified in order to more realistically model the behavior for small breaks.
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| A gradual heat transfer reduction is expected as the steam generator tubes are exposed to increasing void fractions, which force the tubes from the normal nucleate boiling heat transfer regime into transition boiling and eventually into liquid deficient heat transfer.
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| Transition boiling is anticipated when the local void fraction exceeds 0.9. Liquid deficient heat transfer develops when local qualities approach 0.9. Under full power conditions and utilizing the steam generator model documented in references 3, 5 and 6, the onset of these June 2005 15.2-55 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM heat transfer regimes corresponds to steam generator liquid inventories of approximately 70,000 lbm and 35,000 lbm, respectively, for the System 80 design. However, the referenced model conservatively ignores the transition boiling regime, thereby delaying heat transfer degradation until fluid conditions correspond to liquid deficient heat transfer. Therefore, the modified treatment of steam generator heat transfer behavior is conservative, since it underestimates the liquid mass associated with the initiation of heat transfer degradation.
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| Steam Generator Low Water Level Trip As discussed in UFSAR Section 15.2.8.2, the original loss of feedwater inventory event method credited the LSGLT in the affected steam generator only after its liquid inventory had been depleted. This assured conservative treatment of low level trip even if the loss of feedwater inventory event caused rapid steam generator depressurization (i.e., large breaks) and consequent swelling of the downcomer level due to flashing of the downcomer liquid. However, for sufficiently small breaks the steam generator pressure remains constant or increases prior to reactor trip and no downcomer level swell will occur due to flashing. Therefore, in the analysis of small break loss of feedwater inventory events, the LSGLT is modeled to occur with a larger liquid inventory remaining.
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| For the System 80 design steam generators, the low level trip setpoint corresponds to a downcomer liquid level of approximately 24 feet above the tube sheet and a liquid inventory of over 70,000 lbm under full power conditions June 2005 15.2-56 Revision 13
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| | |
| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (based on the reference steam generator model). However, the analysis of small break loss of feedwater inventory events conservatively delays low level trip until heat transfer degradation begins with approximately 35,000 lbm of liquid remaining in the affected steam generator.
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| The methodology was developed to meet the requirement that the analysis is of the FWLB with single failure and offsite power available will not result in exceeding 110%
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| of primary and secondary system design pressures.
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| The method of analyses includes parametrics (sensitivity studies) to establish the limiting initial operating and transient parameters and break sizes with respect to RCS overpressurization during the small feedwater line break event (UFSAR Section 15.2.8.1).
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| The conservative assumptions made in this FWLB with single failure and offsite power available analysis include:
| |
| * Conservative estimation of the break flow and enthalpy by assuming discharge of saturated liquid until the steam generator is dry (steam generator is considered dry when 5000 lbm or less of liquid inventory remains).
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| * Conservative delay in crediting a reactor trip generated on low level by the affected steam generator. The LSGLT is delayed until the liquid mass inventory is at or below 35000 lbm.
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| * Conservative gradual heat transfer reduction. The affected steam generator heat transfer is gradually reduced until liquid deficient heat transfer begins at June 2005 15.2-57 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM approximately 35000 lbm. This conservative treatment of steam generator heat transfer behavior effectively underestimates the liquid mass associated with the initiation of heat transfer degradation at which time the LSGLT is credited.
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| * The transient is initialized so that a reactor trip occurs on the HPPT simultaneously with a LSGLT signal (when 35000 lbm remains in affected steam generator).
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| * The AFAS is delayed until the liquid mass inventory in the affected steam generator is depleted.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a FWLB with single failure and offsite power available from full power conditions are presented in Table 15.2.8-4. The input parameters and initial conditions were selected in order to maximize the consequences of the FWLB with single failure and offsite power available.
| |
| The following input parameters used in the analysis are discussed:
| |
| * Maximum initial core power - Maximum core power maximizes the heat content of the primary system and maximizes the amount of energy to be removed by the secondary system. This results in a larger heat up and pressurization of the primary and secondary systems.
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| * Maximum initial core inlet temperature - This maximizes the amount of energy to be removed from the primary system following a loss of heat sink, and increases the June 2005 15.2-58 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM initial steam generator pressure. This results in a reduced primary to secondary heat transfer, and a higher RCS peak pressure.
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| * Maximum initial RCS flow - For a given core power and core inlet temperature, maximum RCS flow results in a lower hot leg temperature and thus a lower steam generator temperature. This results in decreased break flow and enthalpy. The decreased break flow and enthalpy prior to trip (i.e., decreased heat removal through the break) increases the primary system pressurization.
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| * Maximum (adjusted) initial pressurizer pressure -
| |
| Sensitivity studies show that the maximum pressurizer pressure and maximum initial core inlet temperature result in the maximum RCS pressurization.
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| * Minimum Initial Core Average fuel rod gap conductance
| |
| - Minimum fuel rod gap conductance delays the heat transfer from the fuel to the reactor coolant. This increases the energy content of the primary system after trip, resulting in higher primary and secondary peak pressures.
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| * Maximum PSVs opening setpoint - This delays the pressure relief of primary system and heat removal through the PSVs, thus maximizing the peak primary pressure.
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| * Minimum initial pressurizer volume - This delays the HPPT due to steam cushioning effect, thereby resulting in a higher primary peak pressure.
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| June 2005 15.2-59 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
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| * Minimum initial steam generator level - A lower initial steam generator inventory results in earlier degradation in heat transfer, earlier emptying of the affected and unaffected steam generators and a higher RCS pressure.
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| * Minimum initial feedwater enthalpy - This minimizes the affected steam generators heat removal capability, thereby contributing to primary side heatup and results in higher RCS pressure.
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| * Least negative (most positive) MTC - This reduces the negative reactivity insertion into the core due to coolant heat up during the event, thus resulting in a slower decrease in power and higher heat content of the primary.
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| * Maximum number of plugged steam generator tubes -
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| Increasing the number of plugged steam generator tubes decreases the heat transfer from primary to secondary side due to the reduced steam generator heat transfer surface area. This contributes to RCS heatup and pressurization.
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| * Limiting break size - The largest break size applicable to the methodology for a FWLB with single failure and 2
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| offsite power available is defined as 0.2 ft . The limiting FWLB with single failure and offsite power available event break size is determined by parametric study presented in UFSAR Section 15.2.8.1 to be the 2
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| largest break in the spectrum, i.e., 0.2 ft .
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| June 2005 15.2-60 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.8-4 ASSUMED INITIAL CONDITIONS FOR FEEDWATER LINE BREAK WITH LIMITING SINGLE FAILURE AND OFFSITE POWER AVAILABLE Value Parameter RTP RTP 3876 3990 MWt MWt Initial core power (% of RTP) 102 102 Initial core inlet temperature (°F) 562 566 Initial pressurizer pressure (psia) 2212 2286 Initial RCS flow (% of design) 116 116 Initial pressurizer level (%) 24 24 Initial steam generator level (% of WR) 61.6 60.7 Initial feedwater enthalpy (Btu/lbm) 397 428 Moderator Temperature Coefficient (/°F) 0.0 0.0 Least Least Fuel Temperature Coefficient negative negative CEA worth of trip-WRSO (%) 8.0 8.0 2
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| Fuel rod gap conductance (Btu/h-ft -°F) 500 500 Plugged tubes 1750 1258 PSV setpoint tolerance +3% +3%
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| PSV Blowdown 5% 5%
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| MSSV setpoint tolerance +3% +3%
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| MSSV Blowdown 5% 5%
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| Single Failure FFBT FFBT LOP No No 2
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| Feedwater pipe break area, ft 0.20 0.20 June 2005 15.2-61 Revision 13
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| | |
| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM C. Results The response of key core parameters as a function of time following a FWLB with a limiting single failure and offsite power available for a 0.20 ft break are provided 2
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| in table 15.2.8-3 and figures 15.2.8-32 through 15.2.8-34.
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| The FWLB transient DNBR and fuel centerline temperature or LHGR discussion in UFSAR Section 15.2.8.2.3.C is applicable to this event. The minimum DNBR calculated in UFSAR Section 15.2.8.2.3 for the peak pressure FWLB with LOP event that is initiated from low RCS pressure and high RCS temperature with the four pump coastdown on LOP is more limiting for degradation of DNBR. Therefore, it is concluded that fuel clad degradation would not occur following a FWLB with a single failure and offsite power available.
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| 15.2.8.3.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate RCS barrier performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.3.3.A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed in the computer codes to evaluate RCS barrier performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.3.3.B.
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| June 2005 15.2-62 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM C. Results The response of key RCS parameters as a function of time is presented in Figures 15.2.8-35 through 15.2.8-41 for this limiting fault event.
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| The FWLB assumes a loss of main feedwater to both steam generators resulting in a reduction of steam generator water inventory, pressurization of the secondary side, and a resulting heatup and pressurization of the primary side.
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| The primary side heatup causes volumetric expansion and an increase in pressurizer water level and pressure. A reactor trip occurs on the HPPT along with a turbine trip and TAV closure. Following the turbine trip, the FFBT occurs and results in the coastdown of two RCPs resulting in a reduction of heat removal from the RCS. This reduction in primary to secondary heat transfer causes a rapid heatup of primary side coolant which causes the PSVs to open to limit the RCS pressure increase. The MSSVs open to limit the secondary side pressure increase.
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| For 3876 MWt, the maximum RCS pressure reaches 2684 psia, which is less than 110% (2750 psia) of RCS design pressure (2500 psia). The maximum secondary system pressure reaches 1353 psia, which is less than 110% (1397 psia) of the secondary system design pressure (1270 psia).
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| For 3990 MWt, the maximum RCS pressure reaches 2706 psia, which is less than 110% (2750 psia) of RCS design pressure (2500 psia). The maximum secondary system pressure reaches 1366 psia, which is less than 110% (1397 psia) of the secondary system design pressure (1270 psia). The maximum primary and secondary system pressures for this June 2005 15.2-63 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM event meet the limiting pressure acceptance criteria of Reference 3.
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| 15.2.8.3.5 Radiological Consequences and Containment Performance Fuel damage is not predicted for this limiting fault event.
| |
| The dose consequences for this event are no more limiting then the dose consequence assessment presented in UFSAR Section 15.2.8.2.5.
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| 15.2.8.3.6 Conclusions For the FWLB with a limiting single failure and offsite power 2
| |
| available, with a break size less than or equal to 0.2 ft , the maximum RCS pressure remains below 110% of design pressure (2750 psia), thus ensuring primary system integrity. Likewise, the maximum secondary system pressure remains below 110% of design pressure (1397 psia).
| |
| The minimum DNBR remains above the SAFDL limit, thereby ensuring fuel cladding integrity. All dose consequences are within the 10CFR100 guidelines.
| |
| 15.2.8.4 Feedwater Line Break with LOP and Single Failure for Long Term Cooling The FWLB with LOP and SF transient is the most limiting transient with respect to long-term RCS heat removal capability. It bounds the FWLB without LOP and SF event discussed in 15.2.8.3. The adequacy of the auxiliary feedwater (AFW) system capacity to remove decay heat from the primary side following a FWLB event when considered with a LOP and a June 2005 15.2-64 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM single failure (failure of one of the two AFW pumps to start) is demonstrated in this section. This section also demonstrates that the largest increase in pressurizer water level occurs during this transient and remains below the primary safety valve (PSV) inlet nozzles. As a result, it demonstrates that part of UFSAR Chapter 18.II.D TMI Requirements regarding PSV operability are met; since the water level in the pressurizer remains below the PSV inlet nozzles and only steam is discharged through the PSVs. PSV design requirements are discussed in UFSAR Chapter 5.
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| 15.2.8.4.1 Identification of Causes and Frequency Classification A decrease in heat removal by the secondary system may be caused by a feedwater line break (FWLB) in the main feedwater system (FWS). The FWLB with a single failure and loss of offsite power is classified as a limiting fault.
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| 15.2.8.4.2 Sequence of Events and Systems Operation The sequence of events for the FWLB with a LOP resulting from turbine trip and a single failure (the active failure of an AFW pump) is presented in Table 15.2.8-5. This sequence of events was obtained by simulating the FWLB event with the computer codes identified in Section 15.2.8.3.A.
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| The postulated event is initiated by a pipe break that is assumed to occur between the steam generator economizer feedwater nozzle and its associated feedwater line check valve and results in an instantaneous total loss of main feedwater flow to both steam generators. Critical flow is established from the affected steam generator through the feedwater line June 2005 15.2-65 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM break. In addition, it is assumed that the break discharge enthalpy corresponds to that of saturated liquid until the affected steam generator empties, at which time saturated vapor enthalpy is assumed. The loss of subcooled feedwater flow to both steam generators causes increasing steam generator temperatures and decreasing liquid inventories, that reduce the primary-to-secondary heat transfer rate and increase RCS temperature and pressure.
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| The affected steam generator is assumed to instantaneously lose all heat transfer capability due to total depletion of its liquid inventory. As a result of steaming through the break, both the affected steam generator and the unaffected steam generator water level decreases and initiates an auxiliary feedwater actuation signal (AFAS). No credit is taken for the AFAS or the low steam generator level trip until steam generator dryout occurs. A reactor trip occurs on high pressurizer pressure, which is coincident with the affected steam generator dryout. Depending on the feedwater line break size, a containment pressure trip may also occur for an inside containment FWLB. In order to account for the containment pressure trip condition and coincident containment isolation signal (CIAS), safety injection actuation signal (SIAS) and main steam isolation signal (MSIS), it is assumed that a containment pressure trip condition and CIAS/SIAS/MSIS occur at the same time as the high pressurizer pressure trip. The turbine trips following reactor trip and a loss of offsite power (LOP) occurs three seconds after turbine trip (see UFSAR Section 15.0.2.4). The closing of the turbine admission valves leaves the feedwater line break as the only steam discharge path. The loss of reactor coolant flow following the LOP June 2005 15.2-66 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM decreases the rate of heat transfer in the steam generator tubes, resulting in a reduction o f heat removal from the RCS.
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| This exacerbates the heatup and pressurization of the RCS and steam generators.
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| The reduction of primary side heat removal causes RCS volumetric expansion that results in compression of the pressurizer steam volume due to insurge flow through the pressurizer surge line, which increases the pressurizer pressure above the pressurizer safety valve (PSV) opening setpoints. The main steam safety valves (MSSVs) also open, limiting secondary side pressure and stabilizing temperature.
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| This increases the heat transfer rate to the unaffected steam generator, which results in a decrease in RCS pressure.
| |
| One charging pump load sequences to the diesel generator and provides full flow after LOP and SIAS (see footnote h to UFSAR Table 8.3-3). The main steam isolation valves (MSIVs) close after the MSIS, isolating the unaffected steam generator from the break. After the MSIVs close, the reduction in heat transfer causes the unaffected SG to repressurize, the primary system to heat up the repressurize and additional pressurizer insurge occurs. The opening of the PSVs also results in additional pressurizer insurge. The pressure difference between the affected and unaffected steam generator reaches the auxiliary feedwater P-lockout setpoint prior to any auxiliary feedwater flow being initiated.
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| The auxiliary feedwater is fully diverted to the unaffected steam generator, restoring its water level. The secondary side pressure fluctuates around the MSSV opening and closing setpoints as they cycle. The PSVs will likewise cycle to June 2005 15.2-67 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM relieve RCS pressure. The maximum liquid volume attained in the pressurizer during the FWLB event remains below the inlets to the PSV nozzles when they are open (the PSV nozzles are at 3
| |
| an elevation equivalent to 99.4% level and 1738 ft pressurizer volume). Thus, the pressurizer does not go solid at any time and RCS pressure control is maintained. The transient is terminated at 1800 seconds, when operators initiate a controlled cooldown, such as by using ADVs, to shutdown cooling entry conditions. The operator can take action to isolate the affected steam generator and refill the unaffected steam generator by manual control of AFW any time after the reactor trip occurs. However, the FWLB with LOP and Single Failure analysis does not credit any operator action for the first 30 minutes of the transient.
| |
| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.
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| The NSSS is protected during this transient by the primary safety valves (PSVs) and the following trips:
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| * Steam Generator Low Level
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| * Steam Generator Low Pressure
| |
| * High Pressurizer Pressure
| |
| * Low Departure from Nucleate Boiling Ratio (DNBR)
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| * High Containment Pressure
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| * Variable Overpower Trip.
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| June 2009 15.2-68 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Depending upon the initial conditions, any one of these trips may terminate this transient. The NSSS is also protected by main steam isolation valves (MSIVs), feedwater line check valves, main steam safety valves (MSSVs), and the auxiliary feedwater system (AFWS) which serve to maintain the integrity of the secondary heat sink following reactor trip.
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| The peak pressure criteria for the FWLB with LOP and a postulated active single component failure (UFSAR, Table 15.0-0) in a system required to control the transient in accordance with the NRCs Standard Review Plan (SRP) are presented in Section 15.2.8.2.2 and 15.2.8.2.3 for the FWLB with LOP and FWLB with single failure and offsite power available, respectively.
| |
| For the long term cooling aspect of the FWLB event, the mechanisms to mitigate the primary and secondary heatup and pressurization and to provide a heat sink for decay heat are the PSVs, MSSVs, RCS flow, and the AFW capacity. There is no credible single failure that can degrade the PSV and MSSV capacity, and the rationale for the degradation of the RCS flow as a result of the LOP is the same as that discussed in Sections 15.2.8.2.2 and 15.2.8.2.3. The only active single failure that can reduce the long-term secondary side heat removal capacity is the failure of one of the two auxiliary feedwater pumps to start (Table 15.0-0).
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| In this analysis it is demonstrated that the FWLB with a loss of offsite power and the failure of one AFW pump to start provides adequate decay heat removal so that no loss of core cooling would result. It is further demonstrated in the analysis that the pressurizer does not go solid or pass water through the PSVs. Hence, no loss of control in the RCS pressure boundary occurs.
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| June 2005 15.2-69 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM TABLE 15.2.8-5 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK WITH LOSS OF OFFSITE POWER AND SINGLE FAILURE EVENT Time (sec.) Event RTP RTP 3876 3990 MWt MWt FWLB occurs; A complete loss of normal feedwater 0.0 0.0 to both steam generators occurs 26.1 29.4 AFAS generated in unaffected steam generator 1
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| 26.8 29.9 Pressurizer pressure reaches HPPT setpoint HPPT signal generated; SIAS/CIAS/MSIS signal 26.8 29.9 generated 26.8 30.0 PSVs open Affected steam generator dries out; AFAS generated 27.0 30.2 in affected steam generator 27.3 30.4 Reactor trip breakers open 27.3 30.4 Turbine trip occurs 27.5 30.7 Maximum RCS pressure occurs 27.9 31.0 Scram CEAs begin falling 30.3 33.4 LOP occurs 32.4 35.6 MSIVs close 34.6 35.6 MSSV bank 1 opens (first time) 34.9 38.3 PSVs close 36.2 41.5 AFW lockout occurs 37.8 37.2 MSSV bank 2 opens 37.8 41.9 Steam generator pressure reaches a maximum 50.3 58.1 MSSV bank 2 closes 70.3 73.5 One charging pump re-starts 73.0 76.2 AFW initiated to unaffected steam generator 78.4 85.9 MSSV bank 1 closes 487.0 --- PSVs open 488.6 --- PSVs close 1792.0 1800.0 Maximum liquid volume of pressurizer occurs 1800.0 1800.0 Operator initiates plant cooldown 1
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| The HPPT is coincident with the LSGLT.
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| June 2005 15.2-70 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.8.4.3 Core and System Performance A. Mathematical Model The NSSS response to the long-term heat removal FWLB with LOP and Single Failure event was simulated using the CENTS computer code described in UFSAR Section 15.0.3.
| |
| Reactivity/Physics related data were provided to the CENTS computer code via the same computer codes and methods discussed in Section 15.2.8.2.3.A.
| |
| The minimum DNBR for the core hot channel for the FWLB with LOP was calculated with the CETOP computer code (described in UFSAR Section 4.4) and is discussed in Section 15.2.8.2.3.
| |
| The method of analysis includes parametrics (sensitivity studies) used to establish the limiting initial operating and transient parameters and break sizes with respect to long term heat removal and pressurizer fill, as discussed in Section 15.2.8.1. The method of analysis is the same as the FWLB with LOP methodology discussed in Section 15.2.8.2 in addition to the following conservative assumptions:
| |
| * The single failure of one of the two safety-related auxiliary feedwater pumps.
| |
| * All auxiliary feedwater which is diverted to the affected steam generator is not credited for heat removal.
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| * The maximum value within the allowable range is assumed for auxiliary feedwater temperature.
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| June 2005 15.2-71 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM
| |
| * CIAS/SIAS/MSIS are initiated on high containment pressure at the time of reactor trip. This methodology does not change the timing of the reactor trip, or the methodology for matching the HPPT with dryout of the affected steam generator. Parametric analysis determined that the time of reactor trip is the most adverse time to initiate the CIAS/SIAS/MSIS. This methodology assumes that a CIAS/SIAS/MSIS occurs on high containment pressure, simultaneously with the high pressurizer pressure trip. Early MSIS (before SG low pressure occurs) is conservative with respect to pressurizer level criteria since it eliminates the unaffected SG cooldown through the break early in the transient. There is no effect on peak RCS pressure due to early MSIS, since the peak pressure occurs before the closure of the main steam isolation valves (MSIVs).
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| A SIAS causes the charging pumps to load sequence to the diesel generator after LOP, depending on demand from the PLCS. This is a conservative assumption for pressurizer fill criteria since it adds inventory to the RCS.
| |
| * The pressurizer Level Control System is in the automatic mode with the plant operated on program Tavg at the start of the transient. This methodology provides justification for using the nominal cold leg temperature as the initial cold leg temperature assumption for the event.
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| June 2005 15.2-72 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a FWLB with LOP and Single Failure event are summarized in Table 15.2.8-6. The parameters and conditions were selected in order to demonstrate adequacy of auxiliary feedwater capacity for primary side decay heat removal and to determine the maximum pressurizer water level for PSV operability. A full spectrum of break areas based on parametrics were considered up to a break size of the combined area of flow distribution nozzles in the feedwater ring in establishing the limiting break size.
| |
| The input parameters used in this analysis include:
| |
| * Maximum initial core power - Maximum core power maximizes the heat content of the primary system and the amount of energy to be removed by the secondary system. This results in a larger heat up and pressurization of the primary and secondary systems and pressurizer level.
| |
| * Initial core inlet temperature - As determined by parametric analysis, a lower core inlet temperature results in a higher transient pressurizer level.
| |
| However, the initial core inlet temperature is chosen based on the assumption that the plant is operated on program Tavg, corresponding to the pressurizer level control system program setpoint at hot full power conditions, at the start of the transient.
| |
| * Minimum initial RCS flow - For a given power and core inlet temperature, a lower RCS flow results in a higher June 2005 15.2-73 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM core outlet temperature. This maximizes the energy stored in the RCS and the energy to be removed by the secondary system, resulting in higher pressurizer level.
| |
| * Minimum Initial Pressurizer Pressure - Parametric analysis shows that a minimum initial pressurizer pressure results in a maximum pressurizer water level during the long-term heat removal FWLB.
| |
| * Minimum Initial Core Average Gap - Minimum gap conductance delays the heat transfer from the fuel to the reactor coolant. This increases the energy content of the primary system after trip, resulting in higher primary and secondary peak pressures and higher pressurizer level.
| |
| * Least negative (most positive) moderator temperature coefficient (MTC) - This reduces the negative reactivity insertion into the core due to coolant heat up during the event, thus resulting in a slower decrease in power and higher heat content of the primary.
| |
| * Minimum Pressurizer Safety Valves (PSVs) opening setpoint - Earlier opening of the PSVs increases the surge line flow into the pressurizer, thus increasing the pressurizer level.
| |
| * Maximum initial pressurizer liquid level - Parametric study shows that initiating the transient from the maximum initial pressurizer level has no significant June 2005 15.2-74 Revision 13
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| | |
| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM sensitivity. However, starting from this level results in the maximum pressurizer level during the transient.
| |
| * Minimum initial steam generator level - Parametric study shows that a minimum initial steam generator inventory results in earlier degradation in heat transfer and earlier emptying of the affected and intact steam generators. This increases the RCS pressurization and pressurizer level.
| |
| * Minimum initial feedwater enthalpy - This minimizes the heat removal capability of the affected steam generator, which results in higher RCS temperature and pressure and a greater demand on AFW decay heat removal capacity.
| |
| * Maximum number of plugged steam generator tubes - The parametric study shows that asymmetric tube plugging results in maximum pressurizer level. Increasing the number of plugged steam generator tubes decreases the heat transfer from primary to secondary side due to the reduced steam generator heat transfer surface area.
| |
| This contributes to RCS heatup and pressurization and greater demand on AFW decay heat removal capacity.
| |
| * Limiting break size - The limiting FWLB long term cooling event break size is determined by parametric study discussed in Section 15.2.8.1.2.
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| June 2005 15.2-75 Revision 13
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| | |
| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Table 15.2.8-6 ASSUMED INITIAL CONDITIONS FOR FEEDWATER LINE BREAK WITH LOSS OF OFFSITE POWER AND SINGLE FAILURE EVENT Value RTP RTP Parameter 3876 3990 MWt MWt Initial core power (% of RTP) 102 102 Initial average RCS temperature, Tavg (at 100% power and maximum pressurizer 583.0 585.6 level) (°F)
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| Initial pressurizer pressure (psia) 2100 2100 Initial RCS flow (% design) 95 95 Initial pressurizer water level (%) 59% 59%
| |
| Initial steam generator water level 25% (70% WR) 25% (69% WR)
| |
| (%NR)
| |
| Moderator Temperature Coefficient
| |
| -4 0.0 0.0 (x10 /°F)
| |
| Least Least Fuel Temperature Coefficient negative negative Kinetics Maximum Maximum CEA worth of trip-WRSO (%) 8.0 8.0 Scram delay time (sec) 0.5 0.5 CEA Holding Coil Delay time (sec) 0.6 0.6 2
| |
| Fuel rod gap conductance (Btu/h-ft -°F) 500 500 Plugged steam generator tubes 1000/2500 0/1258 PSV Opening Setpoint (psia) 2475 2475 PSV setpoint tolerance -1% -1%
| |
| PSV Blowdown 14.2% 14.2%
| |
| MSSV setpoint tolerance +3% +3%
| |
| MSSV Blowdown 5% 5%
| |
| Single Failure One AFW Pump One AFW Pump LOP Yes Yes 2
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| Feedwater line break area, ft 0.24 0.23 June 2011 15.2-76 Revision 16
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| | |
| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM C. Results The response of key core parameters as a function of time following a FWLB with a loss of offsite power and a single failure of an AFW pump is the same as those discussed in Section 15.2.8.2.3.C, since that event documents the fuel performance for the FWLB events.
| |
| The FWLB transient DNBR is discussed in Section 15.2.8.2.3.C and depicted in Figure 15.2.8-31. The FWLB transient DNBR and LHGR discussion in Section 15.2.8.2.4.C is applicable to FWLB long term cooling events. The minimum DNBR versus time as shown on this figure remains above the SAFDL throughout the transient. The minimum DNBR calculated in Section 15.2.8.2.4 for the peak pressure FWLB with LOP event that is initiated from low RCS pressure and high RCS temperature with the four pump coastdown on LOP is more limiting for degradation of DNBR.
| |
| Therefore, it is concluded that fuel clad degradation would not occur following a long term FWLB with loss of offsite power and a single failure.
| |
| 15.2.8.4.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate RCS barrier performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.4.3.A.
| |
| B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed in the computer codes to evaluate RCS barrier June 2005 15.2-77 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.4.3.B.
| |
| C. Results The response of key RCS parameters as a function of time is presented in Figures 15.2.8-42 through 15.2.8-51 for this limiting fault event.
| |
| The limiting peak pressure FWLB events are discussed in Sections 15.2.8.2 and 15.2.8.3.
| |
| For FWLB with LOP and a single failure, auxiliary feedwater actuation in the affected steam generator is delayed until affected steam generator dry-out. A main steam isolation signal on high containment pressure isolates the unaffected steam generator from the break early in the transient. Following the isolation, AFW delivery increases the level in the unaffected steam generator (Figure 15.2.8-48 and 50). The AFW flow provides sufficient inventory for heat removal to occur through the MSSVs such that RCS pressure control is maintained by the PSVs, and the RCS converges to a quasi steady state prior to 1800 seconds (Figure 15.2.8-44 and 46). Operator action may be taken at 1800 seconds. This demonstrates the adequacy of RCS decay heat removal with the AFW system during the FWLB which satisfies SRP 10.4.9 and 15.2.8.
| |
| Throughout the transient (Figure 15.2.8-46) the pressurizer water level remains below the PSV inlet nozzle 3
| |
| location of 99.4% level or 1738 ft volume, any time the PSVs are open, and only steam is discharged as required by June 2005 15.2-78 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM UFSAR Chapter 5B and 18.II.D for meeting the NUREG-0737 Requirements.
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| 15.2.8.4.5 Radiological Consequences and Containment Performance Fuel damage is not predicted for this limiting fault event.
| |
| The dose consequences for this event are no more limiting than the dose consequence assessment presented in section 15.2.8.2.5.
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| 15.2.8.4.6 Conclusions The auxiliary feedwater capacity is adequate to provide removal of the core decay heat until operator action is taken 30 minutes after event initiation.
| |
| The maximum pressurizer water level remains below the PSV inlet nozzles and only steam is discharged, thereby satisfying NUREG-0737 Requirements presented in UFSAR Chapter 5B and 18.II.D.
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| June 2005 15.2-79 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.
| |
| | |
| ==2.9 REFERENCES==
| |
| : 1. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.2.8, Feedwater System Pipe Breaks Inside and Outside Containment (PWR)," NUREG-0800 Rev. 1, July 1981.
| |
| : 2. "CE Critical Heat Flux - Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Space Grids,"
| |
| CENPD-162-A, September 1976 (Proprietary).
| |
| : 3. "Safety Evaluation Report Related to the Final Design Approval of the Combustion Engineering Standard Nuclear Steam Supply System (CESSAR)," NUREG-0852, Section 15E.3.2
| |
| : 4. Henry, R.E. and H. K. Fauske, "The Two Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes," Journal of Heat Transfer, Transactions of the ASME, May 1971.
| |
| : 5. ATWS Model Modification to CESEC, CENPD-107, Supplement 1 (Section 3.0), September 1974.
| |
| : 6. ATWS Model Modification to CESEC, CENPD-107, Supplement 1, Amendment 1-P (Section 3.3), November 1975.
| |
| : 7. ATWS Model Modification to CESEC, CENPD-107, Supplement 3 (Sections 240.8, 240.11, and 240.9),
| |
| August 1975.
| |
| : 8. ATWS Model Modification to CESEC, CENPD-107, Supplement 4 (Section 1.6, 1.8, and 4.2), December 1975.
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| June 2005 15.2-80 Revision 13
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| PVNGS UPDATED FSAR 15.3 DECREASE IN REACTOR COOLANT FLOWRATE 15.3.1 TOTAL LOSS OF REACTOR COOLANT FLOW 15.3.1.1 Identification of Causes and Frequency Classification A complete loss of forced reactor coolant flow (LOF) may result from the simultaneous loss of electric power to all four reactor coolant pumps (RCPs). The only limiting credible failure, which can result in a simultaneous loss of power to the four RCPs, is the complete loss of offsite power.
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| An LOF event is an Anticipated Operational Occurrence (AOO) and is classified as an incident of moderate frequency.
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| 15.3.1.2 Sequence of Events and System Operation A loss of electric power to all four reactor coolant pumps produces a reduction of coolant flow through the reactor core that causes an increase in core average coolant temperature, system pressure, and a decrease in margin to DNB. A total loss of forced reactor coolant flow will produce a minimum DNBR more adverse than any partial loss of forced reactor coolant flow event that involves the loss of electrical power to three or less RCPs. This is because the reactor will trip at the same time for both cases, however the partial loss of flow has a slower flow coastdown.
| |
| If credit is not taken for a reactor trip on turbine trip, then reactor protection is provided by a core protection calculator (CPC) generated trip initiated when any one of the four RCP shaft speeds drops to 95 percent of normal speed. The credited CPC trip ensures that the event induced minimum DNBR value will remain above the Specified Acceptable Fuel Design Limit (SAFDL) for DNBR.
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| June 2005 15.3-1 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE The combination of loss of primary heat sink (due to loss of offsite power causing a loss of load on turbine, turbine trip and closure of turbine admission valves) with a reduction of reactor coolant flow results in an increase in RCS pressure that is limited by the primary safety valves(PSVs).
| |
| The steam bypass control system also becomes unavailable due to loss of offsite power, which results in a loss of condenser vacuum and termination of main feedwater to the steam generators. This sequence of system interactions leads to the opening of the main steam safety valves (MSSVs) which limits the secondary side pressure and removes heat stored in the core and the RCS.
| |
| The sequence of events for this moderate frequency LOF event are presented in Table 15.3.1-1. A low voltage on the 4.16 kV safety buses generates an undervoltage signal which starts the emergency diesel generators. The non-safety buses are automatically separated from the safety buses and all loads are shed (except for load centers). After each diesel generator set has attained operating voltage and frequency, its output breaker closes connecting it to its safety bus. Engineered safety feature equipment is then loaded in sequence onto this bus.
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| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System functions were consistent with, or conservative with respect to, numerical values delineated in UFSAR Sections 7.2 and 7.3.
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| Following reactor trip and total loss of forced reactor coolant flow, stored and core decay heat removal occurs by means of June 2009 15.3-2 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE natural circulation through the core with the steam generators providing primary to secondary side heat transfer.
| |
| An auxiliary feedwater actuation signal (AFAS) occurs as the steam generator levels decrease due to the pressure relief and mass discharge during cycling of MSSVs. Actuation of auxiliary feedwater system (AFW) system at a specific time (46 seconds after the AFAS is generated) has no impact on the event DNBR.
| |
| Plant operators may initiate cooldown 30 minutes after the event induced reactor trip occurs by utilizing the AFWS and atmospheric dump valves (ADVs).
| |
| The Standard Review Plan (Reference 1) states that an incident of moderate frequency, such as the loss of forced coolant flow, should not generate a more serious plant condition without other faults occurring independently. Furthermore, the Standard Review Plan states that an incident of moderate frequency, in combination with a single active component failure, or single operator error, should not result in the loss of function of any barrier other than the fuel cladding.
| |
| The loss of offsite power event plus a single failure will not result in a lower DNBR than that calculated for the loss of offsite power event alone. For decreasing reactor coolant flow events, the major parameter of concern is the minimum hot channel DNBR. This parameter establishes whether a fuel design limit has been violated and thus whether fuel damage might be anticipated. Those factors which cause a decrease in local DNBR are:
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| * increasing coolant temperature
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| * decreasing coolant pressure June 2005 15.3-3 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE
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| * increasing local heat flux (including radial and axial power distribution effects)
| |
| * decreasing coolant flow.
| |
| For the total loss of RCS flow event, the minimum DNBR occurs during the first few seconds of the transient and the reactor is tripped by the CPCs on low RCP shaft speed. Therefore, any single failure that would result in a lower DNBR during the transient would have to effect at least one of the above parameters during the first few seconds of the event. None of the single failures listed in table 15.0-0 will have any affect on the transient minimum DNBR during this period.
| |
| Additionally, none of the single failures listed in table 15.0-0 will have any effect on the peak primary system pressure. Nor will the loss of offsite power make unavailable any systems whose failure could affect the calculated peak pressure. For example, a failure of the steam dump and bypass system to modulate or quick open and a failure of the pressurizer spray control valve to open involve systems (steam dump and bypass system and pressurizer pressure control system) assumed to be in the manual mode as a result of the loss of offsite power and, hence, unavailable for at least 30 minutes.
| |
| For the reasons stated above, the loss of offsite power event with a single failure is no more adverse than the loss of offsite power event in terms of the minimum DNBR and peak primary system pressure.
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| June 2005 15.3-4 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.1-1 TYPICAL SEQUENCE OF EVENTS FOR TOTAL LOSS OF REACTOR COOLANT FLOW Time(sec) 3876 MWt 3990 MWt Event RTP RTP 0.0 0.0 Loss of offsite power occurs 0.0 0.0 Turbine trip, Diesel generator starting signal, RCPs coast down and main feedwater is lost 0.6 0.6 Low RCP shaft speed trip condition 0.9 0.9 Trip breakers open 1.5 1.5 CEAs begin to drop 2.80 2.85 Minimum DNBR occurs 4.35 4.10 PSVs open (first occurrence) 4.75 4.65 Maximum RCS pressure 8.95 6.70 MSSVs open (first occurrence) 6.85 7.10 PSVs close (last occurrence) 14.7 10.1 Maximum steam generator pressure 791.25 548.1 Low water level AFAS setpoint reached in steam generator 1 837.35 594.2 AFW begins entering steam generators 831.25 1203.75 MSSVs close (last occurrence) 1800.0 1800.0 Operator initiates plant cooldown June 2009 15.3-5 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.3.1.3 Core and System Performance A. Mathematical Model Several computer codes were employed to evaluate core and system performance for this moderate frequency event. The HERMITE computer code (see reference 2 and UFSAR Appendix 15D) is used to determine the reactor core response during the postulated RCS flow coastdown. The HERMITE LOF simulation, which includes thermal hydraulic data is transferred to the CETOP computer code (which uses the CE-1 CHF correlation that is described in UFSAR Sections 4.4 and 15.0.3) in order to determine thermal hydraulic conditions at time of minimum DNBR. The thermal hydraulic conditions at time of minimum DNBR are then input to the TORC computer code which also uses the CE-1 CHF correlation to calculate the value of the minimum DNBR during the LOF transient. The thermal margin to DNB for the event is calculated using the TORC computer code.
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| The CENTS computer code (see UFSAR Section 15.0.3.1.3.2) is used to simulate the Nuclear Steam Supply System (NSSS) response to the total loss of reactor coolant flow event.
| |
| B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a total loss of RCS flow are selected to minimize DNBR during the transient and are presented in Table 15.3.1-2.
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| The set of initial conditions selected for the analysis presented in this section is one of a very large number of combinations within the reactor operating space of steady state operational configurations, that would provide the June 2009 15.3-6 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE minimum thermal margin required by the core operating limit supervisory system (COLSS). The COLSS (described in UFSAR Section 7.7) computes a Power Operating Limit that ensures the thermal margin available in the core is greater than that required to maintain a calculated LOF event minimum DNBR value that is equal to or greater than the DNBR SAFDL.
| |
| Parameters were chosen in a manner that minimizes reactor bulk saturation, which results in less void reactivity as determined in HERMITE, since minimizing negative feedback results in more adverse consequences and therefore, presents the limiting postulated LOF event. Results of parametrics show that for any axial power distribution, the most limiting Required Overpower Margin is attributed to the following: a) a minimum gap conductance, which delays the core heat flux decrease after reactor trip and results in a later DNBR turn-around and a lower RCS flow at the time of minimum DNBR. b) a maximum RCS pressure and minimum core inlet temperature, which removes the reactor core from bulk saturation conditions and minimizes void reactivity effects. c) a maximum core flow, which removes the core from bulk saturation conditions and minimizes void feedback effects.
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| The four pump LOF transient is characterized by the flow coastdown curve that bounds the coastdowns observed during startup testing. The consequences following a total LOF initiated from any one of these combinations of conditions would be no more adverse than those presented herein.
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| A bounding number of plugged steam generator tubes was assumed in the LOF analysis. The flow coastdown June 2009 15.3-7 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE associated with the plugged steam generator tubes four-Pump LOF is more conservative since it causes a more rapid decrease in the RCS flow.
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| As shown in Table 15.3.1-2, the control element assemblies (CEAs) begin to drop into the core after loss of electrical power to the RCPs and after a conservative delay time that includes the largest possible delay times for sensor delays, CPC response time, and control element drive mechanism (CEDM) coil decay time.
| |
| C. Results The typical response of key parameters as a function of time is presented in Figures 15.3.1-1 to 15.3.1-14 for this moderate frequency event. The loss of offsite power causes the plant to experience a simultaneous turbine trip, loss of main feedwater, condenser inoperability, and a four RCP coastdown. As a result of the RCP coastdown, the CPC generates a trip signal and the CEAs start to drop into the core after a short conservative delay time.
| |
| Since there is no power excursion during the transient, the LOF event does not challenge the linear heat generation rate limit of 21 kw/ft and, consequently, the fuel temperature remains below the SAFDL.
| |
| The minimum DNBR is greater than the DNBR SAFDL value of 1.34 (see Figure 15.3.1-13) and meets the acceptance criteria of the Standard Review Plan.
| |
| Therefore, fuel cladding damage is not predicted for this moderate frequency event.
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| June 2009 15.3-8 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.1-2 ASSUMED INITIAL CONDITIONS FOR THE TOTAL LOSS OF REACTOR COOLANT FLOW Value Parameter RTP RTP 3876 MWt 3990 MWt Core power level (% of rated) 102 102 Core inlet coolant temperature (°F) 548 548 Pressurizer pressure (psia) 2325 2325 Core mass flow (% of design) 116 116 Moderator temperature coefficient (/°F) -0.20E-4 -0.20E-4 Fuel temperature coefficient Least Least negative negative CEA worth for trip-WRSO (%) -8.0 -8.0 Minimum Radial power peaking factor1 1.28 1.28 Fuel rod gap conductance (Btu/hr-ft2-°F) 500 500 Number of plugged SG tubes 1750 1258 Trip Delay Times (sec)
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| : a. Time for CPCs to detect low pump speed 0.60 0.60
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| : b. CPC delay to generate trip signal 0.30 0.30
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| : c. CEA holding coil delay 0.60 0.60 TOTAL 1.50 1.50 Note: The transient is insensitive to the pressurizer and steam generator levels. Nominal values were used in the analysis.
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| 1 This value corresponds to the lower limit on radial peaking for the "RANGE" trip in the CPC.
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| June 2009 15.3-9 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.3.1.4 RCS Pressure Boundary Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate fission product barrier performance (other than fuel cladding) for this moderate frequency event are identical to those described in UFSAR Section 15.3.1.3.A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions relevant to barrier performance for this moderate frequency event are similar to those presented in Table 15.3.1-2 of UFSAR Section 15.3.1.3.B. The only differences are core inlet temperature and core mass flow. The maximum core inlet temperature is used to maximize the peak primary and secondary pressure.
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| For secondary peak pressure cases, the parametric results indicate that the 3990 MWt configuration is not sensitive to core flow. However, for 3876 MWt configuration, minimum core flow results in the highest secondary pressure.
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| The PSVs were modeled to maximize primary pressure. The maximum allowable setpoints (as allowed by Technical Specification 3.4.10) were used (2475 + 3% tolerance).
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| The MSSVs were also modeled to maximize pressure. The maximum allowable setpoints (as allowed by Technical Specification 3.7.1) were used (setpoint + 3% tolerance).
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| June 2009 15.3-10 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE C. Results The typical response of key parameters as a function of time is presented in Figures 15.3.1-1 to 15.3.1-12 for this moderate frequency event. The figures are representative of the transient.
| |
| The loss of steam flow due to closure of the turbine stop valves results in a rapid increase in the steam generator pressure. A sharp reduction in primary to secondary heat transfer follows, which, in conjunction with the loss of forced reactor coolant flow, causes a rapid heatup of the primary coolant. The primary safety valves (PSVs) open and cycle several times, and slightly later the main steam safety valves (MSSVs) open and cycle several times.
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| For 3876 MWt, the maximum RCS pressure is 2649 psia, which is less than 2750 psia (110% of RCS system design pressure of 2500 psia). The maximum secondary-system pressure is 1348 psia, which is less than 1397 psia (110% of secondary system design pressure of 1270 psia).
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| For 3990 MWt, the maximum RCS pressure is 2670 psia, which is less than 2750 psia (110% of RCS system design pressure of 2500 psia). The maximum secondary-system pressure is 1349 psia, which is less than 1397 psia (110% of secondary system design pressure of 1270 psia).
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| These values meet the acceptance criteria of the Standard Review Plan (Reference 1).
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| June 2009 15.3-11 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.3.1.5 Conclusions The minimum DNBR remains above the SAFDL limit, thereby ensuring fuel cladding integrity. The initial margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin.
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| The maximum RCS and secondary system pressures remain within 110% of their design values following the total LOF event.
| |
| Radiological consequences for this event are bounded. The consequences are the result of normal RCS releases at design source terms and are negligible. This event would not result in any releases of radioactive material above that of a normal reactor trip.
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| 15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING A FLOW COASTDOWN This event is categorized as a boiling water reactor event in NRC Standard Review Plan 15.3.2 and will therefore not be analyzed.
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| 15.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER A single reactor coolant pump (RCP) rotor seizure can be caused by seizure of the upper or lower thrust-journal bearings. A single reactor coolant pump rotor seizure with loss of offsite power (LOP) is classified as a limiting fault.
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| The sequence of events, system operations and plant response for the single RCP rotor seizure are almost identical to those of a single RCP shaft break. Both events cause a rapid drop in core flow to the three pump RCS flow configuration.
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| June 2009 15.3-12 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE The difference is that for the rotor seizure event, the reactor is tripped by the Core Protection Calculator (CPC) on a low RCP speed condition, whereas for the shaft break event, the reactor is tripped by the Reactor Protective System (RPS) on a steam generator differential pressure low flow trip. The seized rotor, having the greater resistance to the Reactor Coolant System (RCS) flow, has a slightly faster coastdown. The RCP shaft break allows a freewheeling coastdown of the impeller with the RCP motor continuing to rotate. The RCS flow coastdown is slightly slower, but because the RCP motor and shaft continue to turn the speed signals to the CPC do not decrease. Protection for this event is delayed until the RPS trip is generated.
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| Both the seized rotor and sheared shaft events were assessed with the LOP and it was found that the RCP shaft break resulted in slightly more fuel failure and higher radiological dose than the seized rotor event. Therefore, the results of the single RCP sheared shaft are more limiting than the seized rotor event.
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| 15.3.4 REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER 15.3.4.1 Identification of Causes and Frequency Classification A single reactor coolant pump sheared shaft could be caused by mechanical failure of the pump shaft. This is assumed to result from a manufacturing defect in the shaft. A single active failure of an Atmospheric Dump Valve (ADV) to close is assumed upon opening of the ADVs after reactor trip. This ADV is assumed to remain open for the duration of the event.
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| June 2009 15.3-13 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Following a turbine trip, a Loss of Offsite Power (LOP) caused by a complete loss of the external electrical grid, triggered by the turbine trip is assumed. See UFSAR Section 15.0.2.4 for more information regarding the potential LOP following a turbine trip.
| |
| A single RCP sheared shaft is classified as a limiting fault.
| |
| 15.3.4.2 Sequence of Events and Systems Operation The shearing of the RCP shaft causes the core flow rate to rapidly decrease to a value that would occur with three reactor coolant pumps operating. The reduction in primary coolant flow rate causes an increase in the average coolant temperature in the core, a corresponding reduction in the margin to Departure from Nucleate Boiling (DNB) that may result in some fuel pins experiencing DNB, and an increase in the primary system pressure. Reactor protection is provided by a trip generated by rapid flow reduction that causes a pressure differential
| |
| () across the steam generator primary side in the affected loop to decrease to a value below the trip setpoint.
| |
| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions are consistent with, or conservative with respect to, numerical values delineated in UFSAR Sections 7.2 (table 7.2-1) and 7.3.
| |
| The RPS trip conservatively assumes the largest possible delay time for sensor delay, calculation period, Control Element Drive Mechanism (CEDM) dead time, and CEDM coil decay time.
| |
| The RPS Steam Generator differential pressure trip is single failure proof.
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| June 2009 15.3-14 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE The sequence of events for this limiting fault incident is presented for each evaluation performed. Approximately 3 seconds following turbine trip, an assumed LOP causes a loss of AC power to the onsite loads due to grid instability (see UFSAR Section 15.0.2.4 for more details on 3 second delay). This results in a simultaneous loss of feedwater flow, condenser unavailability, and a coastdown of all RCPs. Approximately 12 seconds after the LOP occurs, the diesel generators start providing power to the two plant 4.16 kV safety buses.
| |
| The pressurizer can assist (but is not credited) in the control of the RCS pressure and volume changes during the transient by compensating for the initial expansion of the RCS fluid. The combination of loss of primary heat sink (due to LOP, which causes a loss of load on turbine and closure of turbine admission valves) with the reduction of reactor coolant flow, results in an increase in RCS pressure.
| |
| The unavailability of the steam bypass control system (SBCS) due to the LOP results in an increase in secondary pressure.
| |
| If no operator action is taken to open the ADVs, the RCS pressure increase is limited by the primary safety valves (PSVs) and the main steam safety valves (MSSVs) limit the secondary side pressure and remove heat stored in the core and the RCS.
| |
| The reactor heat removal takes place by means of natural circulation in the RCS, following the coastdown of the undamaged RCPs. The steam generators provide primary to secondary heat transfer. The water level in each of the steam generators begins decreasing immediately after the loss of main feedwater flow, and an auxiliary feedwater actuation signal (AFAS) is generated on low water level in a steam generator.
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| June 2009 15.3-15 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE The AFAS setpoint is first reached in the steam generator in the unaffected loop. This leads to the startup of the auxiliary feedwater (AFW) pumps.
| |
| For radiological evaluation it is assumed that the operators open the ADVs after reactor trip. Once the ADVs are opened, one valve is assumed to remain stuck open. This results in the eventual generation of a Main Steam Isolation Signal (MSIS) on low steam generator pressure. Once the main steam isolation valves are closed further blowdown of the unaffected steam generator is prevented. AFW is automatically terminated to the affected steam generator as a result of a high differential pressure signal between steam generators. Thirty minutes from the time of shaft shear, the operator is assumed to override the AFW lockout and divert all of the AFW flow to the affected steam generator, covering the tops of the U-tubes after 90 minutes. The operator then initiates cooldown of the RCS by using the ADVs and the AFWS on the unaffected steam generator, while maintaining the level on the affected steam generator.
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| The process of feeding with the AFWS and releasing steam with the ADVs continues until shutdown cooling entry conditions are reached. The operator may let the ESFAS regulate the feedwater flow by issuing and withdrawing AFAS-1 and/or AFAS-2 signals down to cold shutdown entry conditions. See UFSAR Section 10.4.9 for details of the AFW systems (interface requirements are given in UFSAR Section 5.1.4).
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| For the core and system performance evaluation, the major parameter of concern is the minimum hot channel Departure from Nucleate Boiling Ratio (DNBR). This parameter establishes whether fuel design limit has been violated and thus whether June 2009 15.3-16 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE fuel damage could be anticipated. The factors that cause a decrease in local DNBR are:
| |
| * increasing coolant temperature
| |
| * decreasing coolant pressure
| |
| * increasing local heat flux (including radial and axial power distribution effects)
| |
| * decreasing coolant flow For the single RCP shaft break event, the minimum DNBR occurs during the first one to four seconds. Therefore, any single failure that would result in a lower DNBR during the transient would have to affect at least one of the above parameters during the first one to four seconds of the event.
| |
| The single failures that have been postulated are listed in Table 15.0-0. Most of these failures affect the secondary system, and during the first one to four seconds they do not affect the primary system parameters that determine the DNBR.
| |
| The analysis does not credit non-safety related components for any mitigating purposes. The only failures, that could affect the RCS behavior during this interval, are:
| |
| * a loss of normal AC
| |
| * a failure of the pressurizer level control system
| |
| * a failure of the pressurizer pressure control system
| |
| * a failure of the reactor regulating system The loss of normal AC power, which is assumed to occur three seconds after turbine trip, results in loss of power to the RCP, the condensate pumps, the circulating water pumps, the June 2009 15.3-17 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE pressurizer pressure and level control systems, the reactor regulating system, and the feedwater control system.
| |
| Loss of function of the condensate and circulating water pumps and the feedwater control system initially affects only the secondary system, and thus does not affect DNBR in the first one to four seconds of the transient. Loss of power to the reactor regulating system and pressurizer level and pressure control systems renders those systems unavailable. This unavailability will have no significant impact on DNBR during the first one to four seconds. Loss of power to the RCPs is the only potentially significant failure with regard to DNBR that results from a loss of AC. However, as a result of a three second delay between the time of turbine trip and the time of loss of offsite power (see UFSAR section 15.0.2.4),
| |
| there is no effect on minimum DNBR. Failure of the pressurizer level control, pressure control, or reactor regulating systems has minimal affect on any of the four factors that determine DNBR during the first one to four seconds of the event. Thus, none of the single failures listed in Table 15.0-0 will result in a more adverse transient minimum DNBR than that predicted for the single RCP shaft break event.
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| For pressure boundary performance evaluations, there is no single failure in addition to the LOP which results in more limiting peak RCS or secondary side pressures.
| |
| For radiological evaluations, a single active failure of an ADV to close is assumed once the ADVs are opened and this ADV is assumed to remain open for the duration of the event. The stuck open ADV is assumed to cause all of the iodine contained in the affected steam generator to be released to the atmosphere. Thus, this failure in combination with the LOP June 2009 15.3-18 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE maximizes the radiological consequences of the single RCP shaft break event. None of the other single failures listed in Table 15.0-0 in combination with a loss of AC will yield more severe radiological consequences.
| |
| 15.3.4.3 Core and System Performance A. Mathematical Model Several computer codes were employed to evaluate core and system performance for the sheared shaft limiting fault event. The transient core response was simulated using the HERMITE computer code (Reference 2) and the CETOP computer code (described in UFSAR Section 4.4 and 15.0.3.1.6) to generate the limiting core thermal hydraulic conditions at the time of minimum DNBR. The time of occurrence and the value of the minimum DNBR were calculated by the CETOP code.
| |
| The thermal-hydraulic code, TORC (described in UFSAR Section 15.0.3.1.6), was used to calculate DNBR values at various integrated radial peaking factors (Fr) to form data pairs. TORC output was used to determine fuel failure using the statistical convolution technique (see UFSAR section 15.4.8.3.C).
| |
| B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a single RCP sheared shaft are presented in Table 15.3.4-1. These initial conditions result in the most adverse core performance.
| |
| June 2009 15.3-19 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE UFSAR Appendix 15D describes the RCP coastdown methodology.
| |
| The flow coastdown is shown in Figure 15.3.4-8. The flow coastdown curve was developed using the methodology in Appendix 15D.
| |
| June 2009 15.3-20 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.4-1 ASSUMED INITIAL CONDITIONS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP CORE AND SYSTEM PERFORMANCE Parameter Value Core Power Level 100%
| |
| Core Average Heat Flux Maximum Core inlet coolant temperature deg F 548 Core inlet pressure psia 2415 Core mass flow (% of design) 116 Moderator Temperature Coefficient -0.20x10-4
| |
| /deg F Fuel Temperature Coefficient Least Negative CEA worth for trip - WRSO (%) -8.0 Maximum radial peaking factor 2.0 Fuel rod gap conductance (Btu/hr-ft2- Minimum deg F)
| |
| Kinetics Parameters BOC Axial Power Distribution -0.3 Loss of Offsite Power Yes The limiting initial conditions selected for the analysis have the core as far from bulk saturation conditions as possible and yet represents reasonable initial plant conditions from an operational standpoint. The results of parametrics show that the sheared shaft event initiated from top peaked initial conditions, away from saturation conditions lead to the most conservative transient simulations. Since the minimum DNBR for this event occurs so quickly in the transient (< 3 seconds), the analysis is June 2009 15.3-21 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE rather insensitive to the system responses when compared to the initial core parameters.
| |
| That is why the core parameters are discussed in detail in this section. The following analysis initial conditions tend to maximize the calculated fuel damage:
| |
| * Maximum rated core power (from a Power Operating Limit): Maximum allowable power results in more fuel failure.
| |
| * Minimum core inlet temperature: Maximizes initial core wide subcooling.
| |
| * Maximum core inlet pressure: Maximizes initial core wide subcooling.
| |
| * Maximum RCS flow: Maximizes initial core wide subcooling.
| |
| * Full power core average heat flux: Maximum Core Average Heat Flux results in more fuel failure.
| |
| * Top peaked power distribution (most negative Axial Shape Index limit): Since minimum DNBR occurs prior to CEA insertion, top peaked power distribution results in more limiting DNBR values.
| |
| * Maximum radial peaking factor (Fr): A maximum unrodded peaking factor promotes core wide subcooling and corresponds to a more limiting pin power distribution.
| |
| An unrodded pin power distribution also results in a larger number of failed fuel pins.
| |
| June 2009 15.3-22 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE
| |
| * Most positive (least negative) Moderator Temperature Coefficient: Increases the positive reactivity insertion due to moderator temperature feedback during the flow coastdown.
| |
| * Least negative (most positive) Fuel Temperature Coefficient (FTC): FTC has minimal impact on the analysis due to the event being short (< 4 seconds).
| |
| Any reactivity feedback from the fuel in this period has a benign effect mainly because the fuel temperature does not change significantly during the time of interest.
| |
| * Maximum delayed neutron fraction: A maximum delayed neutron fraction, , consistent with beginning of cycle conditions delays the core power decrease after reactor trip.
| |
| * Slower CEA drop time (scram position versus time) and CEDM Coil Delay time: Delays the core power decrease after reactor trip which results in a later DNBR turnaround and a lower flow at the time of minimum DNBR.
| |
| * Minimum gap conductance: Delays the core heat flux decrease after trip: resulting in a later DNBR turnaround and a lower flow at time of minimum DNBR.
| |
| June 2009 15.3-23 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE C. Sequence of Events The Sequence of Events for the Core and System Performance Analysis is shown in Table 15.3.4-2: below.
| |
| Table 15.3.4-2 SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP FOR CORE AND SYSTEM PERFORMANCE Time Event (sec) 0.0 Reactor Coolant Pump Shaft Break occurs 2.5 Minimum DNBR Occurs 2.5 Reactor trip on low RCS flow, based on SG 2.5 Reactor Trip Breakers Open, Turbine Generator Trip 3.1 CEAs begin to drop into the core 5.5 Loss of Offsite Power Occurs 10.0 Event Terminated D. Results During the first few seconds of the transient, the combination of decreasing flow rate and increasing RCS temperature results in a decrease in the DNBR of the fuel pins. The transient minimum DNBR is below the Specified Acceptable Fuel Design Limit (SAFDL) for DNBR. Figure 15.3.4-12 shows the variation of the minimum DNBR with time. The negative CEA reactivity inserted after reactor trip causes a rapid power and heat flux decrease, which June 2011 15.3-24 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE causes DNBR to increase. The amount of predicted failed fuel is determined with the statistical convolution technique (see UFSAR Section 15.4.8.3.C). The limiting fuel failure is discussed in UFSAR Section 15.3.4.6
| |
| ("EAB/LPZ Radiological Consequences and Containment Performance") and as long as the Fr (radial peaking factor) and fuel failure combination results in a bounding (2-hour site boundary) thyroid dose of less than 260 REM, the consequences remain within 10 CFR 100 guideline values.
| |
| DNB Propagation is evaluated by verifying that the bounding fuel clad strain evaluation is still applicable.
| |
| The absolute minimum time in DNB required to reach NRC imposed strain limit of 29.3% is 4.5 seconds. This minimum time is based on the following conditions:
| |
| * Fuel Rod to RCS Differential Pressure < 1200 psid
| |
| * Local Heat Flux < 0.7E6 BTU/hr-ft2
| |
| * Local Mass Flux > 1.4 E6 lbm/hr-ft2
| |
| * Local Quality > -0.1 The analysis verifies the local conditions are within the specified ranges and that the overall time in DNB is less than 4.5 seconds. Therefore, DNB propagation will not occur.
| |
| 15.3.4.4 RCS Pressure Boundary Barrier Performance A. Mathematical Model The CENTS computer code (see UFSAR Section 15.0.3.1.3.2) was used to simulate the secondary and Nuclear Steam Supply System thermal hydraulic response to a single RCP shaft break with a LOP resulting from turbine trip June 2011 15.3-25 Revision 16
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE B. Input Parameters and Initial Conditions The input parameters and initial conditions relevant to barrier performance for this limiting fault event are presented in Table 15.3.4-3.
| |
| Table 15.3.4-3 ASSUMED INITIAL CONDITIONS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP FOR PRESSURE BOUNDARY PERFORMANCE Parameter Value Core Power Level 102%
| |
| Core inlet coolant temperature deg F 548 Pressurizer Pressure (psia) 2325 Core mass flow (% of design) 116 Moderator Temperature Coefficient -0.18x10-4
| |
| /deg F Fuel Temperature Coefficient Least Negative CEA worth for trip - WRSO (%) -8.0 Fuel rod gap conductance (Btu/hr-ft2- Minimum deg F)
| |
| Kinetics Parameters BOC AFAS Setpoint (%WR) 20%
| |
| MSSV Setpoints Minimum MSSV Tolerance +3%
| |
| PSV Setpoints Maximum PSV Tolerance +3%
| |
| Number of plugged SG Tubes 1258 Loss of Offsite Power Yes Single Failure None June 2009 15.3-26 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE The PSVs were modeled to maximize primary pressure. The maximum allowable setpoints (as allowed by Technical Specification 3.4.10) were used (2475 + 3%).
| |
| The MSSVs were also modeled to maximize primary and secondary pressure. The maximum allowable setpoints (as allowed by Technical Specification 3.7.1) were used (setpoint + 3% tolerance).
| |
| June 2009 15.3-27 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE C. Sequence of Events The Sequence of Events for the pressure boundary performance analysis are shown in Table 15.3.4-4 below.
| |
| Table 15.3.4-4 SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM A
| |
| TURBINE TRIP FOR PRESSURE BOUNDARY PERFORMANCE Time Event (sec) 0.0 Reactor Coolant Pump Shaft Break occurs 1.7 Main Steam Isolation Signal on low SG pressure 2.5 Reactor trip on low RCS flow, based on SG 2.5 Reactor Trip Breakers Open, Turbine Generator Trip 3.1 CEAs begin to drop into the core 5.5 Loss of Offsite Power Occurs 6.4 PSVs open 6.6 Peak RCS Pressure occurs 7.8 PSVs close 17.2 Peak SG pressure occurs 17.2 MSSVs open first time 72.7 MSSVs closeb 807.0 AFAS generated in SG #1 853.0 AFW flow delivered to SG #1 964.3 AFW Reset in SG #1b 1301.9 AFAS generated in SG #2 1301.9 AFW flow delivered to SG #2 c
| |
| 1373.2 AFW Reset in SG #2 1800 Operators begin plan cooldown. Event Terminated
| |
| : a. The exact diesel generator sequencing time is not critical for this event and was not specifically modeled since the diesel generator will be available to power AFW A well before AFAS is reached.
| |
| : b. MSSVs cycle open and closed throughout the transient.
| |
| : c. AFW actuation may occur and reset more than once.
| |
| June 2009 15.3-28 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE D. Results The typical transient response of key NSSS parameters as a function of time is presented on Figures 15.3.4-1 to 15.3.4-12 for this limiting fault event.
| |
| The shearing of the RCP shaft causes a reactor trip to occur on a steam generator low RCS flow trip that results in a reactor trip followed by a concurrent turbine trip, which causes turbine admission valve closure.
| |
| Furthermore, a reduction of flow in the affected RCS loop is compounded by the occurrence of a LOP three seconds later. The steam bypass control system, condenser and main feedwater system become unavailable, resulting in a rapid increase in secondary side pressure and temperature.
| |
| The reduction in primary-to-secondary heat transfer causes a rapid heatup of the primary side coolant. No operator action is assumed for 30 minutes. The PSVs open to limit pressure and slightly later the main steam safety valves open.
| |
| The RCS pressure reaches a maximum of 2614 psia (see Figure 15.3.4-4), which is less than 2750 psia (110% of RCS system design pressure of 2500 psia). The secondary system pressure reaches a maximum of 1303 psia (see Figures 15.3.4-9 and 15.3.4-10), which is less than 1397 psia (110% of secondary design pressure of 1270 psia).
| |
| These event primary and secondary pressure values meet the acceptance criteria of the Standard Review Plan.
| |
| June 2009 15.3-29 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.3.4.5 NSSS Response for Control Room Dose Consequences A NSSS response evaluation is performed to determine the time of Control Room Essential Filtration Actuation and secondary mass releases due to the MSSVs and the stuck open ADV for input into the control room dose analysis presented in UFSAR 6.4.7.3.
| |
| A. Mathematical Model The CENTS computer code (see UFSAR Section 15.0.3.1.3.2) was used to simulate the secondary and Nuclear Steam Supply System (NSSS) thermal hydraulic response to a single RCP shaft break with a LOP resulting from turbine trip.
| |
| June 2009 15.3-30 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE B. Input Parameters and Initial Conditions Input parameters and initial conditions are established to maximize secondary side releases and delay the Safety Injection Actuation Signal, which starts the control room essential filtration. Initial conditions are shown below in Table 15.3.4-5.
| |
| Table 15.3.4-5 ASSUMED INITIAL CONDITIONS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP AND A STUCK OPEN ADV Parameter Value Core Power Level 102%
| |
| Core inlet coolant temperature deg F 548 Pressurizer Pressure (psia) 2325 Core mass flow (% of design) 116 Moderator Temperature Coefficient /deg F -4.4x10-4 CEA worth for trip - WRSO (%) -8.0 SIAS Setpoint (psia) 1750 Initial SG Level (%WR) 42%
| |
| MSIS Setpoint (psia) 1005 AFAS Setpoint (%WR) 20%
| |
| AFW SG DP Lockout Setpoint (psid) 130 MSSV Setpoints Minimum MSSV Tolerance -3%
| |
| Loss of Offsite Power Yes Single Failure Stuck Open ADV June 2009 15.3-31 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE C. Sequence of Events The Sequence of Events for the NSSS Response for Control Room Dose are shown in Table 15.3.4-6 below.
| |
| Table 15.3.4-6 SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP WITH A STUCK OPEN ADV Time Event (sec) 0.0 Reactor Coolant Pump Shaft Break occurs 2.5 Reactor trip on low RCS flow, based on SG 2.5 Reactor Trip Breakers Open, Turbine Generator Trip 3.1 CEAs begin to drop into the core 5.5 Loss of Offsite Power Occurs 7.4 Main Steam Safety Valves Open 15.3 Steam Generator Water Level Reaches AFAS setpoint in SG #1 18.8 Steam Generator Water Level Reaches AFAS setpoint in SG #2 61.4 Auxiliary Feedwater Initiated to SG #1 61.4 Auxiliary Feedwater Initiated to SG #2 63.9 Main Steam Safety Valves Close 122.5 Operator initiates plant cooldown by opening one ADV on each steam generator. ADV on SG #1 instantly fails full open 152.5 MSIS actuated on low SG pressure 172.4 SG causes AFAS lockout on SG #1, AFW to SG #1 terminated 242.5 Operator shuts ADV on SG #2 396.9 AFAS reset on high SG level in SG #2 290.3 SIAS Setpoint reached on low pressurizer pressure 330.3 SI flow initiated 447.0 SG #1 reaches dryout conditions 1800. Operators take manual control of AFW and begin filling SG #1. AFW in SG #2 is terminated.
| |
| 5291.5 SG #1 level reaches the physical location of the top of the SG U-tubes (14% NR) 5400.4 SG #1 level reaches the top of the SG U-tubes considering EOP Uncertainties (43% NR) 5401.4 Event Terminated.
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| June 2009 15.3-32 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE D. Results The shearing of the RCP shaft causes a reactor trip to occur on a steam generator , low-RCS-flow trip that results in a reactor trip followed by a concurrent turbine trip, which causes turbine admission valve closure.
| |
| Furthermore, a reduction of flow in the affected RCS loop is compounded by the occurrence of a LOP three seconds later. The steam bypass control system, condenser and main feedwater system become unavailable, resulting in a rapid increase in secondary side pressure and temperature.
| |
| The reduction in primary-to-secondary heat transfer causes a rapid heatup of the primary side coolant. After reactor trip the operator opens ADVs on both steam generators.
| |
| Once the ADVs are opened, one valve is assumed to remain stuck open. This results in the eventual generation of a Main Steam Isolation Signal (MSIS) on low steam generator pressure. Once the main steam isolation valves are closed, further blowdown of the unaffected steam generator is prevented. AFW is automatically terminated to the affected steam generator as a result of a high differential pressure signal between steam generators.
| |
| Thirty minutes from the time of shaft shear, the operator overrides the AFW lockout and diverts all of the AFW flow to the affected steam generator, covering the tops of the U-tubes after 90 minutes.
| |
| June 2009 15.3-33 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.4-7 TIME OF SIAS AND SECONDARY MASS RELEASES FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER AND STUCK OPEN ADV Time of Total MSSV Total ADV ADV Release SIAS Release Release for 600 (sec) (lbm) from SG #1 seconds (lbm) from SG #1 (lbm) 4 5 5 290.3 12.238x10 4.196x10 1.108x10 15.3.4.6 EAB/LPZ Radiological Consequences/Containment Performance A. Physical Model and Assumptions To evaluate the consequences of the single reactor coolant pump shaft break with a LOP event, it is assumed that the condenser is not available for the entirety of the transient. After reactor trip occurs, an MSIS is generated, the steam generators are isolated and the pressure in the steam generators rises quickly to the MSSV setpoint. Sometime after the reactor trip, the operators will open an ADV on each steam generator to stop the MSSVs from cycling. When they open the ADVs, one is assumed to cycle to a fully-opened position and stick there. At this point, the operators close the ADV on the unaffected steam generator. The exact timing of these operator actions does not affect the calculated offsite doses since the dose calculation makes the simplifying assumption that the iodine contained in the affected steam generator is released to the atmosphere at the initiation of the event.
| |
| At thirty minutes the operators divert all of the AFW flow June 2009 15.3-34 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE to the affected steam generator until the tops of the U-tubes are covered and then proceeds with the controlled cooldown using the ADVs on the unaffected steam generator and the AFWS while maintaining the level in the affected steam generator.
| |
| The U-tubes for the unaffected steam generator remain covered by water throughout the event so a Decontamination Factor (DF) of 100 is used for releases from this steam generator.
| |
| Peak containment pressure is not calculated for these events and would be bounded by the LOCA and MSLB events.
| |
| Containment integrity will not be challenged. In the case of the sheared shaft event, the actual time the PSV is opened is less than 3 seconds.
| |
| B. Calculational Methods and Parameters Even for identical core average and hot channel conditions for a given transient event, the number of fuel pins that experience DNB, will vary from cycle to cycle. The calculated amount of fuel failure is sensitive to fuel loading pattern (i.e. pin power distribution). Therefore, it would be difficult to bound the calculated fuel failure of all future reloads based upon any one transient response. In order to accommodate the potential variability between cycles, the calculated dose was expressed as a function of the product of the radial peaking factor and the fuel failure fraction. The values of this product just corresponding to the limits (i.e.,
| |
| thyroid doses of 240 REM for 3876 MWt and 260 REM for 3990 MWt) were determined.
| |
| June 2009 15.3-35 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE A cycle-specific analysis ensures that the product of the failure (based upon actual core loading) and the maximum cycle radial peaking factor do not exceed these limit products. As long as the product for the cycle is less than the product corresponding to the respective limit, the doses for the cycle will be less than the limits and the radiological consequences for a 2-hour, site-boundary thyroid doses will be within 10 CFR 100 guideline values.
| |
| Since an ADV is assumed to stick open, the containment barrier is not credited. Operators take action to close all system boundary valves to minimize flow/discharge to the environment.
| |
| The major assumptions, parameters, and inputs to calculational methods used to evaluate the radiological consequences of the single RCP shaft break are presented in Table 15.3.4-8. Additional clarification is provided as follows:
| |
| : 1. The RCS equilibrium activity is based on long term operation at 102% of Rated Thermal Power (RTP) with a technical specification limit on primary activity, expressed in Dose Equivalent I-131 (DEQ I-131) of 1.0µCi/gm.
| |
| The RCS activity is calculated to determine the total amount of activity leaked into the secondary system during the duration of the accident due to a 0.5 gal/min primary-to-secondary leak per steam generator. The primary-to-secondary leakage of 1 gal/min is assumed to continue to the steam generators for the entire event. The activity in the June 2009 15.3-36 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE fuel clad gap is 10% of the iodines and 10% of the noble gases accumulated in the fuel at the end of core life, assuming continuous full power operation.
| |
| All of the activity in the fuel gap for fuel rods that are calculated to experience DNB is assumed to be uniformly mixed with the reactor coolant. This assumption is consistent with Regulatory Guide 1.77.
| |
| : 2. The steam generator equilibrium activity is assumed to be 0.1 µCi/gm DEQ I-131 prior to the accident.
| |
| This is the technical specification limit for steam generator activity.
| |
| : 3. Offsite power is not available. When the operators open the ADVs following the reactor trip, one is assumed to cycle to a fully-opened position and stick there. At this point, the operators close the ADV on the unaffected steam generator. The exact opening time of the stuck open ADV does not matter for the calculated EAB or LPZ doses since the simplified dose calculation model assumes all of the iodine contained in the affected steam generator is released to the atmosphere instantaneously at the initiation of the event. At 1800 seconds, the operators begin flooding the affected steam generator and cover the tubes by 5400 seconds. The operators control the cooldown using the ADVs and the AFWS on the unaffected steam generator, while maintaining the level in the affected steam generator.
| |
| : 4. For the fluid leaked from primary to secondary, iodine is assumed to be released to the atmosphere June 2009 15.3-37 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE with a DF of 1.0 in the affected steam generator when the tubes are uncovered. The DF is increased to 100 when the tubes are re-covered at about 90 minutes.
| |
| : 5. Mass releases are based on the heat that must be removed from the RCS and they include the decay heat and the stored heat in the primary coolant and in the metal masses. Table 15.3.4-9 shows the releases for this analysis.
| |
| : 6. No credit for radioactive decay in transit to dose point is assumed.
| |
| : 7. The atmospheric dispersion factor used in this analysis is listed in Table 2.3-31.
| |
| : 8. The mathematical model used to analyze the activity released during the course of the accident is described in UFSAR Section 15.0.4 ("Radiological Consequences").
| |
| : 9. Since the PSVs lift for this event, the dose calculation conservatively takes into account the activity released to containment, even though the Reactor Drain Tank is sized to remain intact from the PSV discharge.
| |
| The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the single RCP shaft break with a LOP are as follows:
| |
| : 1. A conservative primary-to-secondary leakage of 1 gpm is used. This corresponds to 1440 gallons per day (gpd).
| |
| Operation with a primary-to-secondary leak of 1 gpm is not allowed. Technical Specification 3.4.14 limits primary-June 2009 15.3-38 Revision 15
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE to-secondary leakage through any one steam generator to less than 150 gpd.
| |
| : 2. The meteorological conditions assumed to be present at the site during the course of the accident are based on 5%
| |
| level /Q values. Meteorological conditions will be less severe 95% of the time. This results in conservative values of atmospheric dispersion calculated for the EAB or LPZ outer boundary. Furthermore, no credit has been taken for the transit time required for activity to travel from the point of release to the EAB or LPZ outer boundary.
| |
| : 3. The dose calculations conservatively use the maximum cooldown rate of 100°F/hr allowed by Technical Specifications. This approach was independent of whether the charging system could compensate for the accompanying rate of shrinkage.
| |
| : 4. The dose calculations use a conservatively low steam generator liquid mass constant value of 160,600 lbm for the intact steam generators. This lower value is conservative as it will increase the steam generator DEQ I-131 concentration from which releases to the site boundary evolve.
| |
| June 2009 15.3-39 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.4-8 TYPICAL PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP (Sheet 1 of 2)
| |
| Parameter Value A. Data and assumptions used to evaluate the radioactive source term
| |
| : a. Power level (% RTP) 102
| |
| : b. Percent of fuel calculated to experience 17.0 at FR=1.72 2
| |
| DNB and fail (%).
| |
| : c. Reactor coolant activity before event, 1.0 based on Technical Specifications (µCi/gm)
| |
| : d. Secondary system activity before event 0.1
| |
| (µCi/gm)
| |
| : e. Minimum primary system liquid inventory 510,000 (excluding mass in pressurizer), lbm
| |
| : f. Minimum steam generator inventory, lbm per 160,600 steam generator B. Data and assumptions used to estimate activity released from the secondary system
| |
| : a. Primary to secondary leak (gpm) 0.5 per SG
| |
| : b. Total mass release through the MSSVs and ADVs based on boiling off inventory to remove decay heat and stored heat.
| |
| * Decay heat based on 1979 ANS Standard with a 2 uncertainty.
| |
| * Specific heat for RCS metal masses. Maximum C. Atmospheric dispersion factors (sec/m )
| |
| 3 2.3 x 10
| |
| -4
| |
| * EAB (0-2 hours) 6.4 x 10
| |
| -5
| |
| * LPZ (0-8 hours) 2 These values were used in the Control Room Dose calculation and are more restrictive than those used for the EAB dose. As long as the product of the Fr and fuel failure fraction (= 29.24%) does not increase, the Control Room Dose calculations will remain bounding and the EAB dose will remain within acceptance criteria.
| |
| June 2009 15.3-40 Revision 15
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.4-8 TYPICAL PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP (Sheet 2 of 2)
| |
| Parameter Value D. Health Physics Parameters
| |
| : 1. Dose conversion assumptions Refer to UFSAR Section 15.0.4
| |
| : 2. Control room design parameters Refer to Appendix 15B and UFSAR Section 6.4 E. Percent of core fission products assumed to be 10 available for release to reactor coolant F. Iodine DF for the unaffected steam generator 100 G. Iodine partition coefficient for the affected 1 steam generator for first 90 minutes 3
| |
| H. Credit for radioactive decay in transit to No dose point I. Loss of offsite power Yes J. RCS Iodine and Noble Gas Source Terms after event initiation.
| |
| Isotope Ci/MWt I-131 2.51 x 10 4
| |
| I-132 3.81 x 10 4
| |
| I-133 5.62 x 10 4
| |
| I-134 6.57 x 10 4
| |
| I-135 5.10 x 10 4
| |
| Kr-83m 4.15 x 10 3
| |
| Kr-85 4.40 x 10 2
| |
| Kr-85m 1.30 x 10 4
| |
| Kr-87 2.15 x 10 4
| |
| Kr-88 3.20 x 10 4
| |
| Xe-131m 2.60 x 10 2
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| Xe-133m 1.38 x 10 3
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| Xe-133 5.62 x 10 4
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| Xe-138 4.97 x 10 4
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| 3 The U-tubes in the affected steam generator are assumed to be re-covered by water after this time.
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| June 2009 15.3-41 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE Table 15.3.4-9 TYPICAL SECONDARY SYSTEM MASS RELEASE TO THE ATMOSPHERE FOR THE SINGLE REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Integrated Primary-to-Secondary Secondary System Mass Release Time Leakage (lbm)
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| (gallons) 5 2 hr 120 9.54 x 10 6
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| 8 hr 480 2.42 x 10 June 2009 15.3-42 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.3.4.7 Conclusions The maximum RCS and secondary side pressures due to a single RCP shaft break in combination with a LOP resulting from turbine trip remain less than 110% of their design values.
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| In the event of a single RCP shaft break or rotor seizure event, the 2-hour EAB doses will be less than 260 REM for a RTP of 3990 MWt. This exposure is within 10 CFR 100 limits. The NRC changed the acceptance criteria for this event in CESSAR SER Supplement 2 as a result of the assumptions made (LOP, ADV sticks open). The acceptance criteria were changed from the guidelines in the Standard Review Plan (Reference 1) to the limits given in 10 CFR Part 100.
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| June 2009 15.3-43 Revision 15
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT FLOWRATE 15.
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| | |
| ==3.5 REFERENCES==
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| : 1. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Sections 15.3.3 -
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| 15.3.4: Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break," NUREG-0800 Rev. 1, July 1981.
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| : 2. "HERMITE A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," CENPD-188, March 1976 (Proprietary).
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| June 2009 15.3-44 Revision 15
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| PVNGS UPDATED FSAR 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL OR LOW (HOT ZERO) POWER CONDITION 15.4.1.1 Identification Cause and Frequency Classification An uncontrolled withdrawal of control element assemblies (CEAs) is postulated to occur as a result of a single failure in the control element drive mechanism (CEDM), control element drive mechanism control system, reactor regulating system or as a result of operator error.
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| These initiating events are Anticipated Operational Occurrences (AOOs), as discussed in Table 3.9-1 and are classified as incidents of moderate frequency.
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| The uncontrolled CEA withdrawal (CEAW) from subcritical and low (hot zero) power conditions are presented in this section.
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| 15.4.1.2 Sequence of Events and Systems Operation The withdrawal of CEAs from subcritical or low (hot zero) power conditions adds reactivity to the reactor core, causing both the core power level and the core heat flux to increase, followed by corresponding increases in reactor coolant temperatures and reactor coolant system (RCS) pressure. The withdrawal of CEAs also produces a time-dependent redistribution of core power. These transient variations in core thermal parameters may result in the systems approach to the specified acceptable fuel design limits (SAFDLs) and RCS and secondary system pressure limits, thereby requiring the protective action of the reactor protective system (RPS).
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| The reactivity insertion rate accompanying the uncontrolled CEA withdrawal is dependent primarily upon the CEA withdrawal rate and the CEA worth since, at subcritical or Hot Zero Power (HZP)
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| June 2009 15.4-1 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES conditions, the normal reactor feedback mechanisms do not occur until power generation in the core is large enough to cause changes in the fuel and moderator temperatures. The reactivity insertion rate determines the rate of approach to the fuel design limits. Based on the reactivity insertion rate and the system initial conditions, the limiting moderate frequency uncontrolled CEAW transient is terminated by a high logarithmic power level trip (HLPT) for the subcritical initial condition, or an RPS variable overpower trip (VOPT) for the HZP initial condition. Depending on the reactivity insertion rate and the system initial conditions, the CPC VOPT, CPC Low DNBR, or high LPD trips may terminate the transient, as well.
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| The uncontrolled CEA withdrawal from subcritical or HZP conditions causes subcritical multiplication to increase core power resulting in a reactor trip. The brief power excursion which results in a slight increase in RCS temperature, is terminated when the CEAs begin to insert. The RCS pressure remains below the pressurizer pressure safety relief valve setpoint. The secondary side pressure increases slightly following reactor trip and is limited by the steam generator safety valves. The atmospheric dump valves are used to cool the RCS to shutdown cooling entry conditions. The feedwater flow rate is operated in manual mode and is very low because it matches the steam flow rates. RCS heat is removed via the steam bypass control system until the shutdown cooling system (SCS) is manually actuated at a time when the RCS temperature and pressure have been reduced to approximately 350°F and 400 psia.
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| The SCS provides sufficient cooling flow to cool the RCS to cold shutdown.
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| June 2009 15.4-2 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The sequence of events for the limiting moderate frequency CEA withdrawal transient from subcritical and HZP conditions are presented in Table 15.4.1-1. Analytical setpoints and response times associated with the RPS trip functions are consistent with, or conservative with respect to, numerical values delineated in UFSAR Section 7.2. A conservative CEA coil decay time of 0.6 seconds was used in simulating the uncontrolled CEAW transients.
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| June 2009 15.4-3 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.1-1 SEQUENCE OF EVENTS FOR SUBCRITICAL AND HOT ZERO POWER CASES Time (seconds) Event Subcritical HZP 0.0 0.0 CEAs begin uncontrolled withdrawal 71.86 24.89 Core power reaches RX trip setpoint
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| (% RTP) 72.36 25.34 Trip breakers open and CEA withdrawal stops 72.96 25.94 CEAs begin to drop 72.97 25.95 Peak power reached (% RTP) 73.16 26.18 Minimum DNBR occurs 73.17 26.19 Peak heat flux reached (% RTP) 73.75 30.12 Withdrawn CEAs are assumed to be fully inserted to its original position for the Subcritical event while all the CEAs except the most Worth Rod Stuck Out (WRSO) are assumed fully inserted for HZP case June 2009 15.4-4 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1.3 Core and System Performance A. Mathematical Model Several computer codes are employed to evaluate core and system performance for the limiting uncontrolled CEAW moderate frequency transients. The CENTS computer code (see UFSAR Section 15.0.3.1.3.2) is used to simulate the Nuclear Steam Supply System (NSSS) response to these events by modeling the neutronics, thermal hydraulics and plant systems during transient conditions. The CETOP computer code (see UFSAR Section 4.4 and 15.0.3.1.6), uses thermal hydraulic and heat flux data from CENTS to simulate fluid conditions within the reactor core in order to calculate the time and numerical value of the fuel pin minimum departure from nucleate boiling ratio (DNBR).
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| A steady state peak linear heat rate of 21 kW/ft has been established as the Limiting Safety System Setting (LSSS) to prevent fuel centerline melting during normal steady state operation. Following design basis AOOs, the transient linear heat rate may exceed 21 kW/ft provided the fuel centerline melt temperature is not exceeded. However, if the transient linear heat rate does not exceed 21 kW/ft, then the fuel centerline melt temperature is also not exceeded. The calculated transient value of Linear Heat Generation Rate (LHGR) exceeds the nominal steady state LSSS of 21 kW/ft for a short period of time during the transient. Therefore, a hand calculation is made on the amount of energy rise and deposition in the fuel, i.e., an adiabatic deposited energy calculation is performed. This is done to ensure June 2009 15.4-5 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES that the fuel temperature (i.e., fuel enthalpy) remains below the melting point and no fuel failure occurs.
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| B. Input Parameters and Initial Conditions Important input parameters and initial conditions used to analyze the NSSS response to a CEA withdrawal from subcritical and HZP conditions are delineated in Table 15.4.1-2. These parameters have been determined to comprise the limiting set of conditions from which an uncontrolled CEA withdrawal could be initiated from subcritical or low HZP conditions and produce the limiting moderate frequency events. A maximum initial core gap conductance is used to minimize Doppler feedback and maximize core power.
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| Parametric analyses have indicated that the lowest initial power and the highest reactivity insertion rates result in the highest peak core power and the most limiting thermal and hydraulic conditions for DNBR. For the subcritical event, the initial subcritical power level that results from a conservative neutron source strength is assumed. The initial minimum power, 5E-10% RTP, is calculated based on this source strength and the subcriticality imposed by the withdrawn bank, and subcriticality multiplication. For the low power event, the lowest initial power is determined based on the High Log Power Trip (HLPT) bypass removal setpoint.
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| The maximum reactivity insertion rates are based on the maximum CEA withdrawal rate of the CEA drive system, 30 inches/min. For the subcritical event, only the June 2009 15.4-6 Revision 15
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES withdrawal of regulating CEAs are evaluated since the PVNGS Technical Specifications requires adequate shutdown margin to prevent going critical by withdrawal of shutdown CEAs, and the out-of-sequence withdrawal of CEAs would result in immediate trip by CPCs due to high penalty applied. In addition, plant startup procedures instructs sequential withdrawal of CEAs in the following order, first the shutdown CEAs, then the part strength CEAs, and finally regulating CEAs. Based on the calculated maximum CEA worths, and the maximum CEA withdrawal rate, the reactivity insertion has a maximum expected rate of 2.835E-4 delta rho/sec for a CEA withdrawal from subcritical condition. The low (HZP) power maximum reactivity insertion rate is determined by performing a parametric study on the CEA bank worth, CEA positions, and axial power shapes to establish bounding bank worth and bounding axial shape. Based on the bounding bank worth and the bounding axial shape, bounding differential rod worths are determined using the HERMITE code. The maximum reactivity insertion rate of 1.7E-4 delta rho/sec is used corresponding to the bounding differential rod worth and the maximum withdrawal speed of 30 inches/min for a CEA withdrawal from low power conditions.
| |
| For the subcritical CEA withdrawal (CEAW), the High Log Power Trip (HLPT) occurs at a setpoint of 0.1% RTP. For the CEAW at low power conditions, the analysis RPS VOPT occurs at a setpoint of 11%. The evaluation used a 12%
| |
| trip setpoint to conservatively envelope the RPS VOPT band and increase rates described in UFSAR Section 7.2.
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| June 2009 15.4-7 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.1-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE LIMITING MODERATE FREQUENCY UNCONTROLLED CEA WITHDRAWAL ANALYSES VALUE PARAMETER Subcritical Low (Hot Zero)
| |
| Analysis Power Analysis
| |
| -8 -5 Initial core power (% of RTP) 5.0x10 1.9x10 Initial core inlet temperature (°F) 572 572 Initial pressurizer pressure (psia) 1785 2100 a
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| Initial RCS flow (lbm/sec) 43278 43278
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| -4 -4 MTC (/°F) 0.5x10 0.5x10 FTCb Least negative Least negative Maximum peaking factor (Fq) 13.8 10.5
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| -4 -4 Maximum reactivity insertion rate 5.67x10 3.4x10
| |
| (/inch) c CEA worth at trip , (%) -1.60 -6.5 (WRSO) e Trip Setpoint (% of RTP) d 0.1 11.0 SCRAM delay (sec) 0.5 0.45 CEA holding coil delay (sec) 0.6 0.6 Fuel rod gap conductance 6530 6530 (Btu/hr-ft2-°F)
| |
| : a. This corresponds to 95% of the original design flow of 164.0 Mlbm/hr.
| |
| : b. The fuel temperature coefficients used are found in the unit- and cycle-specific analyses.
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| : c. For Subcritical CEAW only the withdrawn CEA is reinserted whereas for the Hot Zero Power case, the withdrawn CEA along with all the other CEAs except for the Worst Rod Stuck Out (WRSO) is reinserted.
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| : d. For the CEAW from subcritical, the trip setpoint is for the HLPT.
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| For the CEAW from HZP, the trip setpoint is for the RPS analog VOPT.
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| : e. Note in the computer runs this is conservatively set at 12%.
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| June 2009 15.4-8 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The initial conditions that result in the most rapid approach to the fuel design limits for the subcritical and HZP events are a core inlet temperature of 572°F, based on the Technical Specification in Mode 1 and operating procedures limiting temperature in lower modes, and minimum RCS flow based on four pump operation. These yield reduced heat removal, resulting in higher fuel temperatures and lower DNBRs.
| |
| A minimum pressurizer pressure was used for the subcritical and HZP event conditions to obtain a conservative minimum DNBR and to minimize the initial negative reactivity feedback margin for greater fuel enthalpy production. The most positive moderator temperature coefficient (MTC) of +0.5 x 10 /°F is
| |
| -4 assumed for this analysis. Also, the least negative in cycle Doppler coefficient was used for both uncontrolled CEA withdrawal transients.
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| C. Results The responses of key parameters as a function of time are presented in Figures 15.4.1-1 through 15.4.1-16, typical for the uncontrolled CEAW transient from subcritical condition and the HZP condition.
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| The uncontrolled CEA withdrawal from a subcritical condition resulted in a reactor trip on HLPT. The minimum DNBR calculated for this event that was initiated from the conditions of Table 15.4.1-2, was 2.24 which is greater than the design limit of 1.34.
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| June 2009 15.4-9 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The peak LHGR was calculated to be 56.20 kW/ft, which exceeds the LSSS but the highest fuel centerline temperature reached was 1570°F, which is less than the fuel melt temperature which is based on burnup and erbia content.
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| The uncontrolled CEAW from HZP conditions resulted in a reactor trip on the RPS analog VOPT. The minimum DNBR calculated for this event, initiated from the conditions of Table 15.4.1-2, was 1.67 which is greater than the design limit of 1.34.
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| The calculated peak LHGR was 40.52 kW/ft, which exceeds the LSSS, but the fuel centerline temperature was bounded by 2600°F, which is less than the fuel melt temperature which is based on burnup and erbia content.
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| Therefore, fuel and/or cladding damage is not predicted for these limiting moderate frequency events and the acceptance criteria delineated in Section 15.4.1 of the Standard Review Plan (Reference 1) are met.
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| 15.4.1.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate fission product barrier performance for the limiting moderate frequency events are described in UFSAR Section 15.4.1.3.A.
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| June 2009 15.4-10 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed to evaluate fission product barrier performance for the limiting moderate frequency events are described in UFSAR Section 15.4.1.3.B.
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| C. Results The response of key parameters as a function of time is presented in Figures 15.4.1-1 through 15.4.1-16 for these limiting moderate frequency events. The peak RCS pressure for the CEA withdrawal from a subcritical condition presented in Figure 15.4.1-3 is less than that of the CEA withdrawal from the HZP condition presented in Figure 15.4.1-11. The calculated peak RCS pressure was 1881 psia for the CEAW from subcritical (see Figure 15.4.1-3) and 2225 psia for the CEAW from HZP (see Figure 15.4.1-11), both of which are less than the design limit of 2750 psia.
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| The calculated secondary side peak pressures were 1246 psia for the CEAW from subcritical and 1260 psia for the CEAW from HZP, both of which are less than the design limit of 1397 psia.
| |
| It should be noted that these peak pressures were obtained from the cases for which the initial and transient conditions were selected to maximize heat transfer degradation and fuel centerline temperature for demonstration of this event being not a peak pressure event because of the small amount of heat transferred to the RCS from fuel during the transient.
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| Furthermore, the peak pressure is not a SRP review June 2009 15.4-11 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES criterion for this event, however, the evaluation results are reported herein.
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| 15.4.1.5 Radiological Consequences and Containment Performance Fuel damage is not predicted for the limiting moderate frequency uncontrolled CEA withdrawal events, and therefore there are no radiological consequences from these events.
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| These events would not result in any release of radioactive material above that of a normal reactor trip.
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| 15.4.1.6 Conclusions For the postulated events involving an uncontrolled CEA withdrawal from subcritical or HZP conditions, the PVNGS design meets the relevant requirements of Standard Review Plan (Reference 1).
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| 15.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL AT POWER 15.4.2.1 Identification of Causes and Frequency Classification An uncontrolled CEA withdrawal (CEAW) at power is assumed to occur as a result of a single failure in the control element drive mechanism (CEDM), control element drive mechanism control system, the reactor regulating system, or as a result of operator error.
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| June 2009 15.4-12 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES An uncontrolled CEAW from power conditions is an Anticipated Occupational Occurrence (AOO), as discussed in Table 3.9-1 and is classified as an incident of moderate frequency.
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| 15.4.2.2 Sequence of Events and Systems Operation The uncontrolled withdrawal of a Control Element Assembly (CEA) from full power conditions adds reactivity to the core, causing both the core power level and the core heat flux to increase, followed by corresponding increases in reactor coolant temperatures and reactor coolant system (RCS) pressure. The withdrawal of CEAs also produces a time-dependent redistribution of core power. These transient variations in core thermal parameters may result in an approach to the specified acceptable fuel design limits (SAFDLs) on DNBR and fuel centerline melt temperature, thereby requiring the protective action of the reactor protective system (RPS).
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| The net reactivity insertion rate accompanying the uncontrolled CEAW is dependent upon the CEA withdrawal rate and reactivity feedback mechanisms present at the time of the CEAW from full power conditions. The net reactivity insertion rate determines the rate of approach to the fuel design limits. Depending on the reactivity insertion rate and the system initial conditions, the uncontrolled CEAW transient from full power is terminated by a Core Protection Calculator (CPC) Variable Overpower Trip (VOPT) or the High Pressurizer Pressure Trip (HPPT).
| |
| Table 15.4.2-1 gives a sequence of events from the time the CEAs start to withdraw until the operator initiates a cooldown of the Nuclear Steam Supply System (NSSS). This typical sequence of events was obtained by simulating the event with June 2009 15.4-13 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES the computer codes identified in UFSAR Sections 15.4.2.3 and 2
| |
| 15.4.2.4. Figures 15.4.2-1 through 15.4.2-11 depict the response of key NSSS parameters during this event. The withdrawal of CEAs causes a positive reactivity change, resulting in an increase in the core power and core heat flux (Figures 15.4.2-1 and 15.4.2-2, respectively). Following the generation of a turbine trip on reactor trip, main feedwater flow reduces to 5% of nominal, full flow. The steam bypass control system (SBCS) is assumed to be in manual mode with all the bypass valves closed, resulting in the main steam safety valves (MSSVs) opening (Figure 15.4.2-10) to limit secondary system pressure and remove stored heat transferred from the core and the RCS. After the reactor trip and turbine trip, there is a brief power mismatch between the primary and secondary sides of the steam generator until the MSSVs open, resulting in an increase in RCS pressure and temperature (Figures 15.4.2-3 and 15.4.2-5, respectively). The RCS pressure (Figure 15.4.2-3) remains below the PSV setpoint. The analysis does not credit the actuation of the pressurizer pressure control system and level control systems. However, for RCS control and recovery following the opening of the MSSVs, the pressurizer heaters are adjusted to maintain pressure around 2100 psia. The operator initiates cooldown 30 minutes following the initiation of the event utilizing the main feedwater and the SBCS.
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| 2 Figures are typical representation of the transient.
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| June 2009 15.4-14 Revision 15
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.2-1 does not reflect the event time line beyond 1800 seconds. The cooldown proceeds with the operator reducing the main steam isolation actuation setpoint to prevent the inadvertent generation of a Main Steam Isolation Signal (MSIS).
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| When steam pressure decreases to a point where the main feedwater pumps can no longer be used, the operator secures the main pumps. Cooldown is continued by utilizing one auxiliary feedwater pump SBCS until the shutdown cooling system is manually actuated at a time when the RCS temperature and pressure have been reduced to approximately 350°F and 400 psia.
| |
| Analytical setpoints and response times associated with the RPS trip functions were consistent with, or conservative with respect to, numerical values delineated in UFSAR Section 7.2.
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| A conservative CEA coil decay time of 0.6 seconds was used in simulating the uncontrolled CEAW at power transient.
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| June 2009 15.4-15 Revision 15
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.2-1 SEQUENCE OF EVENTS FOR THE SEQUENTIAL CEA WITHDRAWAL EVENT AT FULL POWER Time (sec)
| |
| RTP RTP Event 3876 3990 MWt MWt 0.00 0.00 CEAs begin withdrawing 12.62 12.92 CPC reactor trip signal generated 13.37 13.67 Reactor trip breaker open 13.37 13.67 Turbine trip occurs 13.40 13.70 Maximum core power occurs
| |
| ~13.5 ~13.5 Minimum DNBR occurs 13.97 14.27 Scram CEAs begin to drop into core 14.10 14.60 Maximum core average heat flux occurs 16.80 17.20 Maximum pressurizer pressure occurs 25.00 23.50 Maximum secondary pressure occurs 25.10 23.60 MSSV bank 1 opens and begins to cycle open and closed 1800.0 1800.0 Operator initiates cooldown June 2009 15.4-16 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.2.3 Core and System Performance A. Mathematical Model The NSSS response to a CEA group withdrawal at full power conditions was simulated using the CENTS computer code described in UFSAR Section 15.0.3.1.3.2.
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| The thermal margin on DNBR in the reactor core was simulated using the CETOP computer code (described in UFSAR Sections 4.4 and 15.0.3.1.6) with the CE-1 CHF correlation that is also described in UFSAR Section 4.4. If the calculated transient value of Linear Heat Generation Rate (LHGR) exceeds the conservative, steady-state limit of 21 kW/ft for a short period of time during the transient, an additional, conservative, hand calculation is performed to confirm that the fuel temperature remains below the melting point. The SAFDL requires the calculated fuel temperature not exceed the fuel melting temperature, but states that showing the LHGR remains below 21 kW/ft guarantees no fuel melting. The fuel temperature is calculated based on the amount of energy deposited in the fuel over time. This is done to ensure that the fuel temperature (i.e., fuel enthalpy) remains below the melting point and no fuel failure occurs.
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| B. Input Parameters and Initial Conditions The assumptions and input parameters that are unique to this event analysis are discussed below and are listed in Table 15.4.2-2.
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| June 2009 15.4-17 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES These initial conditions (i.e., radial power peak, core flow, and inlet temperature) were chosen such that a reactor trip on low DNBR is actuated prior to or at the same time as the HPPT or the VOPT would be initiated. The selection of these parameters in this manner minimizes the hot channel minimum DNBR.
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| The initial conditions and NSSS characteristics used in this analysis yield the minimum DNBR for any CEA group withdrawal incident. Parametric studies were performed on core inlet temperature, fuel rod gap conductance, and core flow. The studies indicated that minimum DNBR during the CEA withdrawal is most sensitive to initial core inlet temperature. Thus, the minimum allowable core inlet temperature was assumed. The minimum initial pressurizer pressure, which has a negligible impact on the event was selected to avoid a HPPT actuation. Thus, the conditions chosen yield the minimum DNBR for a CEA withdrawal at power.
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| The maximum power level from which the withdrawal is initiated was assumed to be 102% of licensed power.
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| Minimum DNBR during the CEA withdrawal is more sensitive to high initial power levels. The initial core average axial power distribution for this analysis is a shape characterized by an axial shape index equal to -0.3. This distribution is assumed because it minimizes the DNBR.
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| June 2009 15.4-18 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.2-2 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE SEQUENTIAL CEA WITHDRAWAL ANALYSIS Value Parameter RTP RTP 3876 MWt 3990 MWt Initial core power (% of RTP) 102 102 Initial core inlet temperature (°F) 548 548 Initial pressurizer pressure (psia) 2100 2100 Initial RCS flow (% of design) 95 95 Moderator temperature coefficient (/°F) 0.0 0.0 Doppler fuel temperature coefficient4 Least Least negative negative Kinetics5 Minimum Minimum Maximum radial peaking, (Fr) 2.0 2.0 Differential reactivity insertion 0.008 0.008
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| (%/in)
| |
| CEA withdrawal speed (inches/min) 30.0 30.0 CEA worth at trip (%) 8.0 8.0 Fuel rod gap conductance (Btu/hr-ft2-°F) 6527 6527 Number of Plugged Steam Generator Tubes 0 0 Single failure None None Loss of Offsite Power (LOP) No No 4
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| The fuel temperature coefficient used is found in the unit- and cycle-specific analyses.
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| 5 The kinetics parameters used are found in the unit- and cycle-specific analyses.
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| June 2009 15.4-19 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Other input parameters that are important to this analysis are moderator temperature coefficient (MTC) and fuel temperature coefficient (FTC) of reactivity.
| |
| The MTC assumed in this analysis corresponds to beginning-of-life core conditions. This MTC has the smallest impact on retarding the rate of change of power, coolant temperature, and DNBR. A FTC corresponding to beginning-of-life conditions was used in the analysis, since this FTC causes the least amount of negative reactivity change for mitigating the transient increases in core power, heat flux, and the reactor coolant temperatures.
| |
| The regulating CEA position from which the CEA withdrawal is initiated corresponds to 25% insertion of the first regulating bank. This particular insertion was selected based on the calculated CEA worth and associated uncertainties to produce the worst transient. A corresponding maximum differential worth of 0.008% per inch of rod motion was conservatively assumed in the present analysis. This corresponds to a maximum reactivity withdrawal rate of 0.4 x 10-4 per second based on the maximum CEA withdrawal speed of 30 inches per minute, including all uncertainties.
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| All the control systems, except the SBCS, were assumed to be in the automatic mode since these systems have no impact on the minimum DNBR obtained during the transient. The SBCS is assumed to be in manual mode because this minimizes DNBR during the transient.
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| June 2009 15.4-20 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES C. Results The dynamic behavior of key NSSS parameters following a CEAW at power are presented in Figures 15.4.2-1 to 15.4.2-11.
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| The minimum DNBR calculated by CETOP is 1.57 for both power levels (3876 MWt and 3990 MWt), which is greater than the DNBR SAFDL of 1.34, and occurs at 13.5 seconds into the transient as shown in Figure 15.4.2-4.
| |
| The peak LHGR reached during this transient is less than 15 kW/ft for both power levels (3876 MWt and 3990 MWt), as shown in Figure 15.4.2-7. This computed LHGR is well below that for the LSSS of 21 kW/ft and, as discussed above, the peak fuel temperatures during this transient are below that of centerline melt.
| |
| Therefore, the results of the uncontrolled CEAW from full power conditions show that for the limiting event, the acceptance criteria for the DNBR SAFDL and fuel centerline melt temperature limit are met.
| |
| 15.4.2.4 Reactor Coolant System Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate fission product barrier performance for this moderate frequency event are the same as those described in UFSAR Section 15.4.2.3.A.
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| June 2009 15.4-21 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES B. Input parameters and Initial Conditions The input parameters and initial conditions that were employed to evaluate fission product barrier performance for this moderate frequency event are the same as those described in UFSAR Section 15.4.2.3.B.
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| C. Results The uncontrolled CEAW from full power results in an increase in RCS pressure and the secondary pressure.
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| For 3876 MWt, the maximum pressure is 2274 psia (see Figure 15.4.2-3), which is below the primary side limit of 2750 psia (110% of the design pressure of 2500 psia). The secondary side pressure reaches 1228 psia (see Figure 15.4.2-6), which is below the secondary side limit of 1397 psia (110% of the design pressure of 1270 psia).
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| For 3990 MWt, the maximum pressure is 2280 psia (see Figure 15.4.2-3), which is below the primary side limit of 2750 psia (110% of the design pressure of 2500 psia). The secondary side pressure reaches 1229 psia (see Figure 15.4.2-6), which is below the secondary side limit of 1397 psia (110% of the design pressure of 1270 psia).
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| Figure 15.4.2-10 gives the MSSVs flow versus time for the uncontrolled CEAW from full power.
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| 15.4.2.5 Radiological Consequences and Containment Performance Fuel cladding degradation is not predicted for this moderate frequency event, and therefore there are no calculated offsite dose radiological consequences for this event. This event would June 2009 15.4-22 Revision 15
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES not result in any releases of radioactive material above that of a normal reactor trip.
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| 15.4.2.6 Conclusions Evaluation of the moderate frequency uncontrolled CEAW from full power shows that:
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| * The fuel cladding integrity will be maintained with the minimum DNBR remaining above the SAFDL, the maximum LHGR remaining below the value that causes peak centerline melt temperature.
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| * The RCS pressure remains below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * The secondary side pressure remains below 110% of its design value (i.e., 110% 0f 1270 psia, or 1397 psia).
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| * Fuel cladding degradation is not anticipated and there are no radiological consequences resulting from the event. This event would not result in any releases of radioactive material above that of a normal reactor trip.
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| * For the postulated uncontrolled CEAW from full power, the PVNGS design meets the relevant requirements of the Standard Review Plan.
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| 15.4.3 SINGLE FULL-STRENGTH CONTROL ELEMENT ASSEMBLY DROP The 4-finger Control Element Assembly (CEA) drops are ensured acceptable results by the initial thermal margin preserved by the Limiting Conditions of Operation (LCOs), and do not rely upon CEA position (i.e., power distribution) factors contained June 2007 15.4-23 Revision 14
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES within the calculations in the Core Protection Calculators (CPCs).
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| The CEA position-related penalty factors for downward deviations of 12-fingered CEAs are calculated such that the CPCs will provide a trip when necessary. A part-strength Power Dependent Insertion Limit (PDIL) also restricts the part-strength CEA insertion to less than 25% for power levels greater than 50%. From these initial conditions, the part-strength single or subgroup drop inserts only negative reactivity (similar to a full-strength single or subgroup drop event). For CEA subgroup drops, the CEA position-related penalty factors for downward deviations are used by the CPCs to provide a trip when necessary.
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| 15.4.3.1 Identification of Causes and Frequency Classification A single full-strength CEA drop (FSCEAD) results from an interruption in the electrical power to the control element drive mechanism (CEDM) holding coil of a single full-strength CEA. This interruption can be caused by a holding coil failure or loss of power to the holding coil. The limiting case is the FSCEAD that does not cause a reactor trip to occur but results in an approach to the Specified Acceptable Fuel Design Limit (SAFDL) on the Departure from Nucleate Boiling Ratio (DNBR).
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| The FSCEAD event is an Anticipated Operational Occurrence (AOO) as discussed in Table 3.9-1 and is classified as an incident of moderate frequency.
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| June 2011 15.4-24 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.3.2 Sequence of Events and Systems Operation Table 15.4.3-1 presents a chronological list of events that occur during the FSCEAD transient, from initiation to the attainment of steady state conditions.
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| The transient is initiated by the release and subsequent drop of a single full-strength CEA. This initiates a reduction in core power and a primary to secondary side power to load mismatch. This mismatch results in a cooldown of the RCS due to excess heat removal by the secondary system. In the presence of a negative Moderator Temperature Coefficient (MTC),
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| the cooldown adds positive reactivity and the core power tends to return to its pre-drop level.
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| The resultant increase in the hot pin radial peaking factor coupled with a return to initial power (following a temporary power depression) results in a power distribution distortion that increases with time as xenon redistributes and a minimum DNBR that remains above the DNBR SAFDL value of 1.34 at 900 seconds following the drop event.
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| By 900 seconds the operator is assumed to have reduced power if the CEA has not been re-aligned. Operation at reduced power is allowed for a limited period to allow the CEA to be re-aligned.
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| 15.4.3.3 Core and System Performance A. Mathematical Model Hand calculations are performed to verify acceptable results for a FSCEAD. This is acceptable since the major effect considered to degrade thermal margin comes from the radially distorted power. A maximum radial distortion factor including 15 minutes of Xenon June 2011 15.4-25 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES redistribution resulting from a FSCEAD is obtained from the physics calculation. The ratio of pre- and post- drop radial distortion is converted to the equivalent power ratio (the required margin) by the quasi partial derivative of the Power Operating Limit (POL) with respect to radial distortion factor. A bounding value of the POL partial derivative within LCO parameters is used to maximize the required margin, which must be reserved by COLSS, CPCS or other LCOs. Bounding partial derivatives were developed by varying one input parameter while the remaining parameters were kept unchanged. As long as the assumptions used in the development of these derivatives remain valid, their values will remain unchanged. For the derivation of the quasi partial derivatives of the POL with respect to radial distortion factor, it is assumed that the coolant inlet temperature, core flow, Fr, and pressure are at their initial pre-transient values. This is conservation because the decrease in DNBR in the transient caused by decreasing RCS pressure is more than offset by the decreasing coolant temperature and reduced core average power.
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| The same methodology is used to analyze a subgroup CEA drop when both CEA calculators (CEACs) are out of service. These margin analyses (calculation of cycle specific distortion factors to ensure they are bounded by the assumed distortion values) are performed each cycle as part of the reload analysis.
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| June 2011 15.4-26 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The CETOP-D computer code (see UFSAR Section 15.0.3.1.6) which uses the CE-1 critical heat flux correlation (see UFSAR Section 4.4) is used to calculate the thermal margin preserved by the TS LCOs for RCS pressure, temperature, flow and ASI.
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| Several computer codes are employed to create a typical sequence of events. The CENTS computer code (see UFSAR Section 15.0.3.1.3.2) is used to simulate the Nuclear Steam Supply System (NSSS) response to this event. The CETOP-D computer code (see UFSAR Section 15.0.3.1.6) which uses the CE-1 critical heat flux correlation (see UFSAR Section 4.4) is used to calculate the equivalent power change corresponding to the axial and radial power distortion when the minimum DNBR is kept unchanged at the SAFDL.
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| B. Input Parameters and Initial Conditions Hand Calculation Methodology The initial conditions are set by the thermal margin reserved in COLSS or CPCS via TS 3.2.4. Since COLSS and CPCS perform an online calculation of DNBR and use measured input values, there are infinite combinations of power, pressure, temperature, coolant flow rate, radial peaking factors, and axial power distribution for any given thermal margin requirement. However TS 3.2.4 ensures a minimum thermal margin which can be converted into a maximum allowable radial distortion which is then verified to be conservative for each core design.
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| June 2011 15.4-27 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Sequence of Events For the purposes of creating a sequence of events, Table 15.4.3-2 lists the assumptions and initial conditions used for the FSCEAD event. The initial conditions of power, pressure, temperature, and coolant flow rate were typical values. The axial power distribution was a limiting shape, and the radial peaking factor was chosen such that a minimum initial thermal margin was obtained. This was done so that the DNBR would be minimized.
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| The negative reactivity inserted by a dropped CEA causes the power to initially decrease everywhere in the core. With no reactor trip, the coolant inlet temperature and pressure will gradually decrease.
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| Concurrently, the radial peaking factor will increase to an asymptotic post drop value. The decreasing coolant temperature combined with the negative doppler and moderator temperature coefficients causes a positive reactivity insertion which brings the core back to the initial power.
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| There is no single power level or plant configuration COLSS In Service (IS)/Out of Service (OOS) and CEACSs IS/OOS that is clearly most limiting. Rather, all conditions and power levels must be considered. To generate a typical sequence of events, the heat flux is based on the 95% power conditions and the asymptotic radial peaking factor existing at that time. This particular power was chosen based on its historically limiting condition for other events and precedence in licensing submittals.
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| June 2011 15.4-28 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES For this event, the choice of mode for the reactor regulating system is inconsequential because there would be no regulating bank motion if the system were in manual mode; and in the automatic mode, the CEA withdrawal prohibit, actuated on the CEAC based rod deviation, prevents the motion of any regulating bank that could cause the CPC calculated minimum DNBR to approach the DNBR SAFDL.
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| June 2011 15.4-29 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.3-1 SEQUENCE OF EVENTS FOR THE SINGLE FULL-STRENGTH CEA DROP EVENT Time (sec) Event 0.0 CEA begins to drop into core 4.0 CEA reaches fully inserted position 4.05 Core power level reaches minimum and begins to increase due to reactivity feedback 30.0 Minimum pressurizer pressure 50.0 Core power returns to maximum value 900.0 Minimum DNBR is reached 900.0 Operator begins a power reduction if the dropped CEA is not re-aligned 6
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| 6 The COLR requires that the power reduction begin at 10 minutes (600 seconds). The effects of xenon on radial peaking at 15 minutes (900 seconds) are used to bound the peaking augmentation.
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| June 2011 15.4-30 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.3-2 TYPICAL ASSUMPTIONS AND INITIAL CONDITIONS FOR THE SINGLE FULL-STRENGTH CEA DROP Value Parameter Core thermal power (% of RTP) 95 Thermal margin 117%
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| Axial Shape -2.0 ASI Initial core inlet temperature (°F)7 558 Initial pressurizer pressure (psia)8 2250 Initial RCS flow rate, (lbm/sec)9 48086 MTC (°/F) -4.4 x 10-4 Fuel Temperature Coefficient (FTC)10 Least negative Dropped CEA worth (%) -0.15 Pre CEA drop radial peaking distortion factor 1.6150 Post CEA drop radial peaking distortion factor 1.8146 CEA drop radial peaking distortion factor @ 15 1.8609 mins CEA drop time (sec) 4.0 Fuel rod gap conductance (Btu/hr-ft2-°F) 1620 Number of plugged steam generator tubes 0 Single failure None Loss of offsite power No 7
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| Since the purpose of the transient analysis was to provide a representative sequence of events, the inlet temperature was set to a representative value.
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| 8 Since the purpose of the transient analysis was to provide a representative sequence of events, the pressure was set to a representative value.
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| 9 Since the purpose of the transient analysis was to provide a representative sequence of events, the RCS flow was set to a representative value. The flow represents the actual value used in the runs.
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| 10 The fuel temperature coefficient used is found in the unit- and cycle-specific analyses.
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| June 2011 15.4-31 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES C. Results Table 15.4.3-1 presents the sequence of events for the full-strength CEA drop event initiated at the conditions described in Table 15.4.3-2.
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| A minimum CE-1 DNBR of greater than the DNBR SAFDL is obtained at 900 seconds, as determined from the initial radial power peaking increase following CEA drop plus 15 minutes of xenon redistribution at the final coolant conditions. At this time the operator will take action to reduce power in accordance with the Technical Specifications, if the misaligned CEA has not been realigned.
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| The fuel centerline melt temperature is not exceeded if the transient Linear Heat Generation Rate (LHGR) does not exceed 21 kW/ft. The limiting initial power is 95%.
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| Based on the bounding radial peaking distortion, the maximum allowable initial LHGR that could exist as an initial linear heat condition without exceeding 21.0 kW/ft during this transient exceeds 15 kW/ft (21 kW/ft/Maximum Distortion Factor = maximum allowable initial LHGR). This amount of margin is assured because the linear heat rate LCO is based on the more limiting allowable LHGR for LOCA.
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| Therefore, the results of the FSCEAD analysis show that for the limiting event, the acceptance criteria for the DNBR SAFDL and peak fuel centerline temperature limit are met.
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| June 2011 15.4-32 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.3.4 Reactor Coolant System Barrier Performance A. Mathematical Model CENTS and CETOP computer codes (see UFSAR Sections 15.0.3.1.3.2 and 15.0.3.1.3.6) are used to simulate the Nuclear Steam Supply System (NSSS) response to this event to create a typical sequence of events.
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| The computer codes that were employed to evaluate fission product barrier performance for this moderate frequency event are the same as those described in UFSAR Section 15.4.3.3.A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions that were employed to evaluate fission product barrier performance for this moderate frequency event are the same as those described in Section 15.4.3.3.B.
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| C. Results The barrier performance parameters following a FSCEAD would be less adverse than those following the CEA withdrawal events from subcritical, HZP, or at power (see UFSAR Sections 15.4.1 and 15.4.2).
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| This event is initiated with a nominal pressurizer pressure of 2250 psia. The RCS pressure decreases as a result of the FSCEAD and remains well below the 110%
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| primary side design limit of 2750 psia. The secondary side pressure also remains below the 110% secondary side design limit value of 1397 psia.
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| June 2011 15.4-33 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The single FSCEAD event does not result in a reactor and turbine trip and therefore, there are no resultant event related steam releases to the atmosphere.
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| 15.4.3.5 Radiological Consequences and Containment Performance Fuel cladding degradation is not predicted for this moderate frequency event, and therefore there are no offsite dose radiological consequences for this event. This event would not result in any releases of radioactive material above that of a normal reactor trip.
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| 15.4.3.6 Conclusions Evaluation of the moderate frequency FSCEAD event shows that:
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| * The fuel cladding integrity will be maintained with the minimum DNBR remaining above the SAFDL, the maximum fuel centerline temperature remaining below the fuel melt temperature.
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| * The RCS pressure remains below 110% of its design value (i.e., 110% of 2500 psia, or 2750 psia).
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| * The secondary side pressure remains below 110% of its design value (i.e., 110% 0f 1270 psia, or 1397 psia).
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| * Fuel cladding degradation is not anticipated and there are no radiological consequences resulting from the event.
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| * For the postulated single full-strength CEA drop event initiated from the Technical Specification LCOs, the PVNGS design meets the relevant requirements of the Standard Review Plan (Reference 1).
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| June 2011 15.4-34 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP 15.4.4.1 Identification of Event and Causes The startup of an inactive reactor coolant pump (SIRCP) is presented here with respect to potential loss of subcriticality. This event is also evaluated with respect to Reactor Coolant System (RCS) pressure and fuel performance criteria.
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| Administrative procedures govern the starting of RCPs and reduce the effects of RCP starts.
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| 15.4.4.2 Sequence of Events and Systems Operation SIRCP can either raise or lower core average coolant temperature. The average temperature can be lowered by increased heat transfer to the steam generators, caused by increased core coolant flow and by colder primary system water in the steam generators being forced into the core. The core average temperature can be raised by increased heat transfer from the steam generators to the RCS, as a result of increased core coolant flow and by hotter primary system water in the steam generators being forced into the core.
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| The SIRCP event which lowers the core average temperature (the cooldown event), combined with a negative isothermal temperature coefficient (ITC), produces a positive reactivity insertion. The SIRCP event which increases core average temperature (the heatup event), combined with a positive ITC, produces an increase in RCS pressure and a positive reactivity insertion.
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| June 2011 15.4-35 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.4.3 Analysis of Effects and Consequences SIRCP can cause either a heatup or cooldown of the primary system depending on the primary to secondary T.
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| SIRCP was examined in Modes 3 through 6, since plant operation with less than four RCPs running is only permitted in these modes.
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| A. Mathematical Models The reactivity added to the core during a heatup or cooldown SIRCP event was determined using conservative isothermal temperature coefficients (ITCs) with a maximum uncertainty applied. These ITCs were used with the maximum core temperature increase or decrease to determine the maximum reactivity inserted during SIRCP. This reactivity insertion is compared to the total amount of subcriticality.
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| B. Input Parameters and Initial Conditions The initial conditions considered for this event ranged from a positive to a negative temperature difference between the secondary and primary system.
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| Assuming primary system temperature higher than the secondary temperatures (a positive temperature difference) would result in cooling down the RCS.
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| Assuming secondary system temperature initially higher than the primary temperature (a negative temperature difference) would result in heating up the RCS.
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| Cooling the RCS would increase reactivity if there is a negative ITC. Heating the RCS would increase reactivity and RCS pressure if there is a positive ITC.
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| June 2011 15.4-36 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES To conservatively calculate the reactivity added to the core during SIRCP, the most negative or positive ITCs are used with uncertainties applied in the most conservative direction. The initial core average moderator temperature during SIRCP is assumed to be at the temperature corresponding to the most positive ITC for the heatup event, or the most negative ITC for the cooldown event.
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| The following assumptions are made:
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| : 1. Prior to SIRCP all reactor coolant pumps are off.
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| Normally at least one RCP must be running (or one shutdown cooling train during shutdown cooling operation). The Technical Specifications allow operation without any pumps running for up to one hour. This assumption maximizes the change in temperature during SIRCP.
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| : 2. Following SIRCP the core average temperature either (1) drops to the temperature of the coldest steam generator, for the cooldown event, or (2) increases to the temperature of the hottest steam generator, for the heatup event. This conservatively bounds the maximum change in core temperature that can occur during this event, by ignoring coolant mixing that would occur in the reactor coolant system.
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| June 2011 15.4-37 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.4.4 Results The results show that the maximum temperature change during SIRCP when used with the most conservative ITCs does not result in a loss of subcriticality.
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| When the RCS is above the conditions requiring low temperature overpressure (LTOP) protection, the SIRCP event that results in a heatup of the RCS will not result in a peak pressure greater than 110% of design pressure. While the RCS is in the LTOP mode, the shutdown cooling system (SCS) relief valves will prevent violation of RCS integrity limits. (See section 5.2 for a general discussion of RCS integrity.)
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| Since subcriticality is not lost during the event, there is no increase in heat flux and therefore no decrease in minimum departure from nucleate boiling ratio (DNBR).
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| 15.4.4.5 Conclusions The SIRCP does not result in a loss of subcriticality. The increase in pressure during this event will not result in peak pressures above the applicable limits. There is no increase in core heat flux and therefore no decrease in minimum DNBR.
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| 15.4.5 FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW This event is not applicable to pressurized water reactors and, therefore, is not included in this FSAR.
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| June 2011 15.4-38 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.6 INADVERTENT DEBORATION 15.4.6.1 Identification of Event and Causes The Inadvertent Deboration (ID) event is presented here with respect to the time available for operator corrective action prior to the reactor achieving criticality. Fuel integrity is not challenged by this event.
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| The ID event may be caused by improper operator action or by a failure in the boric acid makeup flow path, which reduces the flow of borated water to the charging pump suction. Either cause can produce a boron concentration of the charging flow which is below the concentration of the reactor coolant. The ID event is classified as an incident of moderate frequency as defined in Reference 1 of UFSAR Section 15.0.
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| This evaluation shows that Mode 5 (cold shutdown) with the Reactor Coolant System (RCS) drained down results in the least time available for detection and termination of an ID event.
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| The combination of lowered RCS volume and three operating charging pumps results in a small dilution time constant and the fastest dilution rate, and therefore yields the shortest time interval between initiation of an ID event and the reactor achieving criticality.
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| Since RCS boron concentration is maintained under strict procedural controls, the probability of a sustained and erroneous dilution due to operator error is very low.
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| 15.4.6.2 Sequence of Events and Systems Operation An ID event occurs when charging flow into the RCS has a lower boron concentration than the fluid within the RCS. The resulting decrease in RCS boron concentration adds positive June 2011 15.4-39 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES reactivity to the core. Assuming dilution continues at the maximum possible rate with unborated charging flow, the operator has at least 15 minutes in Modes 1 through 5, and 30 minutes in Mode 6, between the receipt of an alarm from the Boron Dilution Alarm System (BDAS) and the reactor achieving criticality. The mechanism to notify the operator of an ID event when the BDAS is inoperable is chemical surveillance of the RCS boron concentration, in accordance with the unit-specific Core Operating Limits Report (COLR).
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| The success path for reactivity control is as follows: The operator is alerted to a decrease in the RCS boron concentration either through a high neutron flux alarm on the startup flux channel, sampling, reactor make-up flow rate, or boric acid flow rate. The operator turns off the charging pump(s) in order to halt further dilution. Next, the operator increases the RCS boron concentration by implementing the emergency boration procedure.
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| When the reactor is critical (Modes 1 and 2) an ID event will result in a slow increase in core power and RCS temperature.
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| This event is slower than other reactivity excursions analyzed (e.g., CEA withdrawals), and the reactor will trip in time to prevent violation of any safety limit. This trip ensures a second dilution period (Mode 3 or lower with All Rods In (ARI) or in an N-1 configuration), during which the operator must be notified of any ongoing deboration at least 15 minutes before the reactor achieves criticality. Therefore, Modes 1 and 2 do not have to be analyzed further with respect to an ID event.
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| If the reactor is subcritical with trippable CEAs withdrawn from the core (All Rods Out (ARO) in Modes 3 through 5), the high log power trip must be active. Shutdown Margin (SDM) is June 2011 15.4-40 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES assumed to be at the minimum value consistent with Technical Specification limits. If the operator is not notified of the ID event with sufficient time to prevent criticality, the high log trip will actuate and insert the withdrawn CEAs into the core. The trip will alert the operator to the ID event and ensure a second dilution period during which the operator will have at least 15 minutes to respond before the reactor achieves criticality. Thus ARO configurations in Modes 3 through 5 do not need to be analyzed further with respect to an ID event.
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| If the reactor is subcritical without trippable CEAs (ARI or N-1 configuration in Modes 3 through 5), an ID event will result in degradation of SDM and an increase in subcritical multiplication. SDM is assumed to be at the minimum value consistent with Technical Specification limits. The operator must be notified of the ID event at least 15 minutes before the reactor achieves criticality, so that corrective actions may be implemented. Either the BDAS, or chemical surveillance of RCS boron concentration (when BDAS is not operable), will alert the operator to the ID event.
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| If the reactor is in Mode 6, an ID event will result in degradation of SDM. In this instance the operator must be notified of the ID event at least 30 minutes before complete loss of subcriticality. The mechanisms to notify the operator are the same as above for Modes 3 through 5 (BDAS or chemical sampling). In Mode 6, the CEAs may be totally removed from the core, thus the ARO configuration is the most limiting and is the only configuration analyzed.
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| The indications and/or alarms available to alert operators to an ID event in each of the operational modes are outlined below.
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| June 2011 15.4-41 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES A. In Modes 1 and 2, a high power trip or, for some sets of conditions, a high pressurizer pressure trip in Mode 1 or a high log power trip in Mode 2 will provide indication of any boron dilution event. Furthermore, a high TAVE alarm may also occur prior to trip.
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| B. In Modes 3 and 4 with CEAs withdrawn, the high logarithmic power level trip and pre-trip alarm, and a high neutron flux alarm (BDAS alarm) will provide an indication to alert the operator of an inadvertent boron dilution.
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| C. In Modes 3, 4, and 5 with CEAs fully inserted except the worst rod stuck out and in Mode 6, a high neutron flux alarm (BDAS alarm) will provide indication of any boron dilution event.
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| D. In Mode 5 with the RCS partially drained for system maintenance, the primary coolant volume available for mixing consists of only the volume of the reactor vessel up to the level of the coolant legs and the volume of the shutdown cooling system. Similarly, in Modes 4 or 5 when cooling the RCS with the shutdown cooling system, the active volume may consist of only the volume of the reactor vessel (excluding the upper head region) and the volume of the shutdown cooling system. In these conditions a high neutron flux alarm (BDAS alarm) will provide indication of any boron dilution event.
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| Operational procedure guidelines, in addition to these indications and/or alarms, will assure detection and June 2011 15.4-42 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES termination of an ID event before the reactor achieves criticality.
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| 15.4.6.3 Analysis of Effects and Consequences The time interval between the onset of an ID event and the reactor achieving criticality may be calculated for each possible set of initial conditions (operating mode, mixing volume, charging flow, SDM, and, if applicable, stuck rod worth). These time intervals are conservatively translated into required boron concentration surveillance intervals in the COLR, for use when BDAS is inoperable.
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| A. Mathematical Model Assuming complete mixing of boron in the RCS, the rate of change of boron concentration during dilution is described by the following equation:
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| (1)
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| Where: M = RCS mass C = RCS boron concentration t = time W = Charging mass flow rate of unborated water and dC/dt is maximized by maximizing W and minimizing M.
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| Assuming: W = constant, equal to the maximum possible value; and June 2011 15.4-43 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES M = constant, equal to the minimum value occurring during the boron dilution incident, the solution of equation (1) can be written (2)
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| Where: C(t) = Boron concentration at time t C0 = Initial boron concentration at time t = 0 T = M / W = Boron dilution time constant The time Tcrit that is required to dilute to criticality (from the start of the ID event at time t = 0) is given by:
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| (3)
| |
| Where: Ccrit = Critical boron concentration Furthermore, the relationship between the alarm time and the BDAS alarm setpoint (i.e., the SRM ratio, or the ratio of the source range flux signal at a particular time to the initial source range flux signal) is given by:
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| ICRR(C0 CCRIT )
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| (4)
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| ICRR(C0e a / T CCRIT )
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| Where: SRMratio = BDAS alarm setpoint (limiting value of 2.2)
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| June 2011 15.4-44 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES ICRR= Inverse count rate ratio which is an empirically determined fraction of the difference between the measured boron concentration and the calculated critical boron concentration.
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| B. Input Parameters and Initial Conditions It is assumed that the ID event proceeds at the maximum possible rate. For this to occur, all charging pumps must be on, the reactor makeup water pump must be on, letdown flow must be diverted from the volume control tank, and a failure in the boric acid makeup water flow path (e.g., flow control valve FV-210Y failing in the closed position) must terminate borated water flow to the charging pump suction.
| |
| Specific input parameters and initial conditions for Mode 5 with the RCS partially drained, and for Mode 6, are described in the sections below.
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| C. Mode 5 Drained Down Boron Dilution Event Evaluation of ID events initiated during each of the six plant operational modes (defined in the Technical Specifications) shows that Mode 5 (cold shutdown) in the drained-down configuration results in the shortest available time for detection and termination of the event. Therefore, the initial conditions and analysis parameters are chosen for the cold shutdown operational mode to minimize the interval from initiation of dilution to the time at which criticality is reached. The following are the analysis assumptions for the Mode 5 ID event:
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| June 2011 15.4-45 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES
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| : 1. Complete mixing of boron within the RCS is assumed.
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| : 2. The event is initiated at the ARI condition with reactor trip breakers open.
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| : 3. The primary coolant volume, including only the volumes for Mode 5 drained conditions as 3
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| described above, is 4500 ft . A conservatively low reactor coolant mass was assumed by using the cold RCS internal volume. Assuming a coolant temperature of 210 F, the Technical Specification o
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| upper limit for cold shutdown, the resulting mass is 269,461 lbm.
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| : 4. All three charging pumps are assumed to be operating at a rate of 45 gpm per pump, for a total of 135 gpm. The corresponding mass flow rate, assuming cold liquid flow, is 18.78 lbm/sec.
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| : 5. The critical boron concentration (with ARI) and the IBW are 1150 ppm and 85.6 ppm/%,
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| respectively, including uncertainties for the cold shutdown conditions. For Mode 5 ARI with the RCS drained to the middle of the hot legs, a value of 3.0% (1.0% Technical Specification SDM + 2.0% Stuck Rod Worth) is used for initial subcriticality. The initial subcritical boron concentration is found by adding the product of the IBW and the initial subcriticality (i.e., 3.0% ) to the critical boron concentration. The resulting initial boron June 2011 15.4-46 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES concentration in the Mode 5 ARI configuration is therefore 1407 ppm. Thus the change in boron concentration from 3.0% subcritical to a critical condition is 247 ppm.
| |
| The parameters discussed above are summarized in table 15.4.6-1.
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| Table 15.4.6-1 ASSUMPTIONS FOR THE MODE 5 INADVERTENT DEBORATION ANALYSIS Parameter Assumed Value Cold RCS volume (excluding pressurizer and 4,500 surge line), ft3 RCS mass (excluding pressurizer and surge 269,461 line), lbm Volumetric charging rate, gpm 135 Mass charging rate, lbm/sec 18.78 Dilution time constant, T, sec 14,351 Initial boron concentration, Co, ppm 1,407 Critical boron concentration, Ccrit ppm 1,150 D. Mode 6 Boron Dilution Event If the reactor is in Mode 6 an ID will result in degradation of SDM. In this instance the operator must be notified of the ID event at least 30 minutes before the reactor achieves criticality. In Mode 6, the CEAs may be totally removed from the core.
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| Therefore, the ARO configuration is the most limiting and is the only configuration analyzed. The following are the analysis assumptions for the ID event:
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| June 2011 15.4-47 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES
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| : 1. Complete mixing of boron within the RCS is assumed.
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| : 2. The event is initiated at the ARO condition.
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| : 3. The primary coolant volume is conservatively set to the values assumed for Mode 5 drained conditions. A coolant temperature of 135° F is assumed for Mode 6, however, with a resulting mass of 276,583 lbm.
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| : 4. All three charging pumps are assumed to be on at a rate of 45 gpm per pump, for a total of 135 gpm.
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| The corresponding mass flow rate, assuming cold liquid flow, is 18.78 lbm/sec.
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| : 5. Initial and critical boron concentrations are conservatively established at 3000 ppm (the minimum refueling boron concentration allowed by the COLR) and 2204 ppm, respectively.
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| The parameters discussed above are summarized in table 15.4.6-2.
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| June 2011 15.4-48 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.6-2 ASSUMPTIONS FOR THE MODE 6 INADVERTENT DEBORATION ANALYSIS Parameter Assumed Value Cold RCS volume (excluding pressurizer and 4,500 surge line), ft3 RCS mass (excluding pressurizer and surge 276,583 line), lbm Volumetric charging rate, gpm 135 Mass charging rate, lbm/sec 18.78 Dilution time constant, T, sec 14,731 Initial boron concentration, Co, ppm 3,000 Critical boron concentration, Ccrit ppm 2,204 15.4.6.4 Results Using the conservative parameters described above in equations (3) and (4), sufficient time is available to assure the detection and termination of an ID event. Numerous indications of improper operation and the high neutron flux alarm on the startup flux channel will alert the operator of an ID event with at least 15 minutes (30 minutes in mode 6) remaining before the core becomes critical.
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| 15.4.6.5 Conclusions The ID event will result in acceptable consequences.
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| Sufficient time is available for the operator to detect and to terminate an ID event if it occurs. Fuel integrity is not challenged during this event.
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| June 2011 15.4-49 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.7 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO THE IMPROPER POSITION 15.4.7.1 Identification of Events and Causes The inadvertent loading of a fuel assembly into the improper position event is initiated by interchanging two fuel assemblies. The likelihood of an error in core loading is considered to be extremely remote because of the strict procedural control used during core loading.
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| 15.4.7.2 Sequence of Events and System Operation The fuel enrichment within a fuel assembly is identified by a coded serial number marked on the exposed surface of the top end plate of the fuel assembly. This serial number is used as a means of positive identification for each assembly in the plant. A tag board is provided in the main control room showing a schematic representation of the reactor core and spent fuel storage area. During the period of core loading, the location of each fuel assembly, and source will be shown on this tag board by a tag carrying its identification number.
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| The tag board in the main control room will be constantly updated by a designated member of the reactor engineering staff whenever a fuel assembly is being moved. The reactor engineering representative will be in constant communication with each area where this is occurring. All core alterations shall be observed and directly supervised by either a licensed senior reactor operator or a senior reactor operator limited to fuel handling who has no other concurrent responsibilities during this operation. Fuel assemblies will not be moved unless these lines of communication are available. In addition June 2011 15.4-50 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES to these precautions, annual inventories of the spent fuel and new fuel storage areas will be performed in addition to post refueling reactor core mapping. These inventories will be used as the basis for setting up the tag board for use during fuel movement. At the completion of core loading, the exposed surfaces of the top end plates are inspected to verify that all assemblies are correctly located. These precautions are included in the core loading procedures which are to be reviewed by appropriate plant personnel.
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| If, in spite of the extreme precautions described above, a fuel misloading does occur, the consequences depend on the types and locations of the fuel assemblies that have been interchanged.
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| The misloading of a fuel assembly may affect the core power distribution only slightly, for example, if assemblies of similar enrichments and reactivities are misloaded.
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| Alternatively, if assemblies having very different enrichments or reactivities are misloaded the core power distribution may be affected enough so that core performance would be degraded.
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| In the unlikely event that two assemblies of different enrichments would be interchanged, some misloadings would be detected using ex-core control channel detectors and the reactivity computer during the low power physics testing.
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| During the initial start-up of PVNGS Units 1 and 2, a symmetry check was performed in which the reactivity worths of symmetrically located CEAs were compared against one another.
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| The interchange of two or more fuel assemblies with greatly different K destroys the octant symmetry of the core flux distribution and would thus produce significant variations in the worths of symmetrically located CEAs. This asymmetry would be corroborated by symmetry checks performed for other June 2011 15.4-51 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES symmetric rod groups, thereby confirming and possibly even locating a fuel assembly misload.
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| Starting with Unit 3 Cycle 1, individual rod worth symmetry testing is no longer performed. Reg. Guide 1.68, Rev. 0 requires that we perform flux distribution measurement (flux symmetry testing).
| |
| Incore Flux Symmetry measurements are performed and compared to predictions. Deviations between the measurements and predictions exceeding the specified criteria could be indicative of a fuel misloading event or other fuel design related errors.
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| Thus most of the fuel assembly misloadings that can be postulated are easily detectable both during the rod symmetry checks and during power range operation. However, there are a small number of misloadings which are undetectable during the rod symmetry testing or even early in the cycle with in-core instrumentation during power range operation. Of this small class the worst case is the interchange of a shimmed assembly with an unshimmed one at the center of the core. This case, although not detectable at BOL, would cause local power peaking as the shims burn out.
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| 15.4.7.3 Analysis of Effects and Consequences Several single assembly interchanges of this type were postulated and investigated using the fine-mesh neutronics methods discussed in section 4.3. Most were shown to be detectable when estimates of the symmetric rod worths were calculated. Of those misloads which were not conclusively demonstrated to be detectable during startup at BOC1, the June 2011 15.4-52 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES interchange of assemblies 9 and 50 was shown to result in the highest FRN value (1.72) during subsequent full answer operation over the first cycle. The associated power distribution shown in figure 15.4.7-1 has a calculated minimum DNBR of 1.48.
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| Since this is greater than the minimum acceptable DNBR, no clad failure is expected to occur.
| |
| The above analysis results are for cycle 1 for Units 1,2, and
| |
| : 3. Although the analysis looked at several different misloads and reasonably determines the worst undetectable case, it was not a comprehensive perturbation analysis which determined the single most bounding case. With subsequent reloads and changes in core parameters (e.g., enrichment limits, DNBR limits, fuel mechanical design changes) and new fuel management patterns, reanalyzing a comprehensive set of peturbations would not change the conclusions in the original analysis. Therefore, the reference analysis remains a valid representative case for current fuel management and demonstrates the general detectability of worst case misloads with in-core instrumentation. This is because the changes in fuel management do not impact the basis of the original Cycle 1 illustrative analysis. That is, the path by which two assemblies appear to be neutronically similar at beginning of cycle (e.g. by fuel pellet changes, enrichment changes, burnable absorber changes) has never been material to the analysis. The detection is unaffected by final design changes since it relies on the differences in the reactivities of the misloaded fuel assemblies. The detection method relies on the observation of anomalous local power effects identified with the incore detector system.
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| June 2011 15.4-53 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Furthermore, even though these misloads may not be detected during startup at BOC, it is very probable that the anomaly would be detected early in the cycle before the maximum FRN value is attained. This is because this type of interchange (i.e., shimmed with unshimmed) tends to produce an increasingly distorted power distribution which would alert the reactor engineer to the possibility of a fuel misloading.
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| 15.4.7.4 Conclusion The inadvertent misloading of a fuel assembly into the improper position events have been analyzed and shown to be highly improbable. The fuel handling procedures and core instrument system more than adequately assure that there is not possibility of a misloaded fuel assembly event proceeding to a point that would fail fuel, and hence meets the 10CFR 100 requirements.
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| 15.4.8 CONTROL ELEMENT ASSEMBLY EJECTION 15.4.8.1 Identification of Cause and Frequency Classification A Control Element Assembly (CEA) ejection (CEAE) event is postulated to occur as a result of a mechanical failure that causes an instantaneous circumferential rupture of the control element drive mechanism (CEDM) housing or its associated nozzle. This results in the reactor coolant system pressure ejecting the CEA and drive shaft to the fully withdrawn position.
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| The CEDM housings are capable of withstanding throughout their design life all normal operating loads including the steady state and transient operating conditions specified for the June 2011 15.4-54 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES reactor vessel. Hence, the occurrence of such a failure is considered to be incredible, and the CEAE is classified as a limiting fault event.
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| 15.4.8.2 Sequence of Events and Systems Operation The sequence of events that occur during the fuel performance aspect of the CEAE initiated from full power Beginning-of-Cycle (BOC) conditions is presented in Table 15.4.8-1. Likewise, the sequence of events that occur during the peak pressure aspect of the CEAE is presented in Table 15.4.8-5.
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| The postulated mechanical failure of the CEDM causes the ejection of a CEA which adds positive reactivity to the core that results in a rapid increase in reactor core power for a short period of time. This power excursion is terminated by the combination of delayed neutron and Doppler feedback effects. Closely following the CEAE, reactor shutdown is initiated by a core protection calculator (CPC) and/or reactor protective system (RPS) variable overpower trip (VOPT) on high neutron power. The reactor power decreases rapidly as the shutdown CEAs drop into the reactor core.
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| The reactor core is therefore protected against severe fuel damage by restrictions on CEA patterns and/or power dependent insertion limits during operation; and by a reactor trip.
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| These factors combine to limit the acceptable values for fuel enthalpy, fuel and clad temperatures, and reactor coolant system (RCS) and secondary side pressures during the transient.
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| The sequence of events that occurs during the CEA ejection initiated from full power BOC conditions for the peak RCS pressure event is presented in Table 15.4.8-5.
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| June 2011 15.4-55 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The limiting secondary steam releases for the CEAE event are based on a full-power, BOC analysis , with the sequence of 11 events summarized in Table 15.4.8-1. These steam releases are applied to the CEAE event radiological consequences assessment presented in UFSAR Section 15.4.8.5. The analysis assumed that the ejected CEA results in a hole at the top of the reactor vessel head. In the analysis, secondary side steaming was maximized by assuming a loss of offsite power (LOP) following the turbine trip. This caused a 4 pump Reactor Coolant Pump (RCP) coastdown, main feedwater pump trip, loss of condenser vacuum and loss of the steam bypass control system. This results in increased primary and secondary side pressures due to decreased heat removal by the steam generators with the subsequent opening of the primary and secondary safety valves to relieve pressure and dissipate energy. A safety injection actuation signal (SIAS) was generated, adding additional boron to the core by means of HPSI pumps. Subsequently, the reduced reactor power following the reactor trip, in addition to the postulated break in the primary system, caused the RCS pressure and temperature to decrease to below that of the steam generators. The analysis also assumed that operator action was delayed until 30 minutes after event initiation. Plant cooldown was accomplished by using the Auxiliary Feedwater (AFW) system in conjunction with the atmospheric dump valves (ADVs) until shutdown cooling entry conditions were reached.
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| 11 The system response for the calculation of the limiting steam releases used the CESEC computer code (see UFSAR Section 15.0.3.1.3.1).
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| June 2011 15.4-56 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-1 TYPICAL SEQUENCE OF EVENTS FOR THE CEA EJECTION EVENT FROM FULL POWER CONDITIONS (FUEL ENTHALPY AND TEMPERATURE CASE)
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| Time Event (sec) 0.00 Mechanical failure of CEDM causes CEA to eject 0.04 Core power reaches CPC VOPT analysis setpoint 0.05 CEA is fully ejected 0.07 Maximum core power occurs 0.79 CPC VOPT signal is generated 0.79 Trip breakers open 0.79 Turbine trip occurs 1.39 CEAs begin to drop 3.33 Maximum clad surface temperature in the hot node occurs 3.33 Maximum fuel centerline temperature in the hot node occurs
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| ~5.57 CEAs fully inserted; core power reduced to below 10% of full power June 2011 15.4-57 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Section 15.4.8 of the Standard Review Plan (Reference 1) does not require the evaluation of the 4 pump RCP coastdown for this event with respect to fuel performance. However, Regulatory Guide 1.77, Appendix B, indicates that release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power.
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| 15.4.8.3 Core and System Performance A. Mathematical Model The Nuclear Steam Supply System (NSSS) response to a CEAE was simulated using the method of analysis described in Reference 3. The procedure outlined in Figure 2.1 of Reference 3 was used to determine the energy deposition in the fuel rod. The number of fuel pins predicted to experience departure from nucleate boiling was calculated using the STRIKIN-II computer program described in UFSAR Section 15.0.3.1.5 with the CE-1 correlation described in UFSAR Chapter 4.4. A matrix relating the initial and ejected CEA peaking factors to a pin census edit was obtained from Step 6 of the C-E synthesis method (Reference 3) and used to calculate the number of fuel pins experiencing DNB.
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| Further conservatism was introduced by assuming that clad failure occurs when fuel rods experience DNB.
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| The CENTS computer code described in UFSAR Section 15.0.3.1.3.2 was used to determine the peak RCS and secondary side pressures and the overall NSSS response to the event. The inputs to CENTS were selected so that the ejected rod power excursion that resulted, June 2011 15.4-58 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES maximized the time-dependent energy deposition into the RCS.
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| B. Input Parameters and Initial Conditions Important input parameters and initial conditions used to analyze the NSSS response to a CEAE are delineated in Table 15.4.8-2. A spectrum of initial reactor states (including conditions characteristic of the beginning and end of the fuel cycle) was considered.
| |
| Table 15.4.8-3 lists the initial CEA Bank configurations considered as separate initial reactor states and gives the maximum worth for a CEA ejected from the state as well as the maximum post-ejection radial peaking enhancement factor. The initial conditions for the principal process variables were varied within the reactor operating space of steady state operational configurations to determine the set of conditions that produce the most adverse consequences following a CEAE. Various combinations of initial core inlet temperature, core inlet flow rate, pressurizer pressure, and axial power distribution were considered. The initial pressurizer and steam generator water levels, as controlled within the operating space, have an insignificant effect on the consequences of the CEAE analysis.
| |
| For all cases analyzed, an axial power distribution was chosen to maximize the energy content in the hottest fuel pellet. The remaining parameters were chosen based on the results shown in Chapter 4 of Reference 3. These parameters were varied in the most adverse direction until a COLSS power operating limit was achieved.
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| June 2011 15.4-59 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-2 TYPICAL ASSUMPTIONS USED FOR THE CEA EJECTION ANALYSIS FULL POWER BEGINNING OF CYCLE INITIAL CONDITIONS (FUEL ENTHALPY AND TEMPERATURE CASE)
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| Parameters Values(12)
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| Initial core power (% of RTP)13 102 Initial Core inlet temperature (°F) 569 Initial RCS flow rate (lbm/sec)14 43277 Initial pressurizer pressure (psia) 2100 Moderator temperature coefficient (/°F) 0.0 Maximum Ejected CEA worth (%) 0.131 Fuel Temperature Coefficient15 Least negative Kinetics 16 Minimum Pre-ejection 3-D fuel pin peaking factor 2.005 Post-ejection 3-D fuel pin peaking factor 3.8095 CEA worth at trip, WRSO (%) 5.5 Fuel Rod gap conductance (Btu/hr-ft2-°F) Minimum17 Postulated time to eject CEA (sec) 0.05 CEA coil delay time (sec) 0.6 Axial shape index -0.2 12 Values are for the bounding CEAE analysis.
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| 13 Only the 3990 MWt RTP case was simulated for maximizing the fuel centerline temperature and this bounds the 3876 MWt RTP case.
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| 14 This corresponds to 95% of the original design flow.
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| 15 The fuel temperature coefficients used are found in the unit- and cycle-specific analyses.
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| 16 The kinetics parameters used are found in the unit- and cycle-specific analyses.
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| 17 The STRIKIN-II code, using the FATES runs, solves the 1-D, radial heat conduction equation for each axial region along the hot rod. The conduction model explicitly represents the gap region and dynamically calculates the gap conductance in each axial region. This results in the smallest gap conductance so that heat transfer to the coolant is minimized.
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| June 2011 15.4-60 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-3 TYPICAL INITIAL REACTOR STATES CONSIDERED FOR THE TYPICAL CEA EJECTION EVENT Ejected Rod Ejected Radial Initial Rod Configuration Worth, % Peaking Factor Bank 5 inserted (95% power) 0.131 1.9 Banks 4 & 5 inserted (50% power) 0.300 2.4 Banks 3, 4 & 5 inserted (20% power) 0.404 2.7 June 2011 15.4-61 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The key input parameters and initial conditions used in analysis of the CEA ejection peak RCS pressure event is presented in UFSAR Section 15.4.8.4.B. The assumptions and initial conditions used in the analysis that determined the CEA ejection secondary side steam releases are discussed in UFSAR Section 15.4.8.5.
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| C. Results A spectrum of initial reactor states, shown in Table 15.4.8-3, was analyzed to show that each case met the criteria established in Regulatory Guide 1.77. All cases resulted in a radial average fuel specific enthalpy less than 280 cal/gram at the hottest axial location of the hot fuel pin. The case that resulted in the greatest potential for offsite dose consequences (i.e., the case resulting in the largest number of postulated fuel failures) was identified as the case initiated from full power beginning of cycle (BOC) initial conditions.
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| The following paragraphs describe this event in detail. Table 15.4.8-1 contains the sequence of events that occur during a CEAE initiated from full power.
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| Refer to Table 15.4.8-2 for the initial conditions and assumptions used for this analysis. Figures 15.4.8-1 through 15.4.8-5 show the reactor core power, peak core power density, core average heat flux, peak hot channel heat flux, and clad and fuel temperatures during the significant portion of transient.
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| June 2011 15.4-62 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Ejection of a CEA causes the core power to increase rapidly due to the almost instantaneous addition of positive reactivity. However, the rapid increase in core power is terminated by a combination of Doppler feedback and delayed neutron effects. This increase in power results in a high power trip and the reactor power begins to decrease as the CEAs enter the core.
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| Reactivity effects are shown in Figure 15.4.8-6.
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| In the hot channel, the increase in heat flux is such that DNB is calculated to occur, resulting in:
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| * a rapid decrease in the surface heat transfer coefficient
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| * a rapid decrease in heat flux
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| * a rapid increase in clad temperature.
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| The heat flux continues to decrease for the remainder of the transient.
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| The calculated radially averaged fuel enthalpy and fuel centerline enthalpy of the hottest fuel pellet for the limiting case remains below the criterion of 280 cal/gm and the fuel centerline temperature is less than the fuel melt temperature. These results show that the CEAE accident will not result in a radial average fuel enthalpy greater than 280 cal/gm at any axial location in any fuel rod, and that no fuel rod exceeds fuel centerline melt temperature.
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| NRC Regulatory Guide 1.77 recommends that the onset of DNB be used as the basis for predicting clad failure for the postulated CEA ejection event. For PVNGS, the June 2011 15.4-63 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES number of fuel rods that experience DNB is calculated with a statistical convolution technique, which is discussed in UFSAR Section 15E.3.3 and described in Reference 4. The statistical convolution technique involves the summation, over the reactor core, of the number of fuel rods with a specific DNBR value, multiplied by the probability of DNB at that DNBR value. However, to provide a conservative assessment of radiological consequences, a bounding number of fuel rods is assumed to suffer clad failure in the 3954 MWt evaluation, as shown in Table 15.4.8-6. In the 4070 MWt evaluation, the allowable clad failure percentage is a function of the product of clad failure percentage and maximum fuel radial peaking factor as discussed in Section 15.4.8.5D. This limitation on clad failure is compared to a unit cycle specific fuel pin census performed for each reload analysis using the statistical convolution technique in order to predict that the number of fuel rods that experience DNB will result in less than assumed fuel rod failure in dose calculations.
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| 15.4.8.4 Reactor Pressure Boundary Barrier Performance The CEAE peak RCS pressure event initiated from full power BOC conditions is presented in this section.
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| The reactor coolant discharged through the CEA break to containment and the steaming mass release through the MSSVs and ADVs are discussed in UFSAR Section 15.4.8.5.
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| June 2011 15.4-64 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES A. Mathematical Model The CENTS computer code described in UFSAR Section 15.0.3.1.3.2 was used to determine the RCS and secondary side peak pressures and the overall NSSS response to the CEAE event.
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| The CESEC computer code described in UFSAR Section 15.0.3.1.3.1 was used in determining the barrier performance aspect of the CEAE analysis that deals with secondary side releases to atmosphere that are used in the radiological consequence UFSAR Section 15.4.8.5.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions used in determining barrier performance for the peak RCS pressure during the CEA ejection accident from full power BOC conditions are presented in Table 15.4.8-4.
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| The following assumptions were made in the analysis:
| |
| * Initial conditions for the key process variables were varied within the ranges of steady state operational configurations including the uncertainties to determine the set of initial conditions and input variables that would produce the most adverse consequences.
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| * It was conservatively assumed that there was no pressure boundary breach or leakage in the CEDM area of the reactor vessel head and no pressure reduction caused by the failure of the control element mechanism housing for the primary peak pressure case.
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| June 2011 15.4-65 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES
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| * Only the high pressurizer pressure trip (HPPT) was credited. Although the CPC or RPS VOPT trip may occur on high neutron power much earlier than the HPPT making the event more benign, no credit was taken for these trips.
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| June 2011 15.4-66 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-4 TYPICAL ASSUMPTIONS USED FOR THE CEA EJECTION ANALYSIS FOR RCS PEAK PRESSURE EVENT FROM FULL POWER BEGINNING OF CYCLE INITIAL CONDITIONS Value Parameters RTP RTP 3876 MWt 3990 MWt Initial core power (% of RTP) 102 102 Initial core inlet temperature (°F) 548 548 Initial RCS flow (lbm/sec) 52845 52845 18 Initial pressurizer pressure (psia) 2100 2100 Initial pressurizer water level (ft) 11.4 11.4 Initial steam generator water level (ft) 24.5 25.7 MTC (/°F) 0.0 0.0 Fuel Temperature Coefficient 19 Least Least negative negative Kinetics 20 Minimum Minimum SCRAM worth at Trip, N-2 (%) 5.5 5.5 Fuel rod gap conductance (Btu/hr-ft -°F) 2 6984 6984 Ejected CEA worth (%)
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| 21 0.157 0.157 Postulated CEAE time (sec) 0.05 0.05 SCRAM delay time (sec) 0.75 0.75 CEA holding coil delay time (sec) 0.6 0.6 Plugged steam generator tubes 0 0 PSV Tolerance +3% +3%
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| MSSV Tolerance +3% +3%
| |
| Single Failure None None LOP No No 18 This corresponds to 116% of original design flow.
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| 19 The fuel temperature coefficient used is found in the unit- and cycle-specific analyses.
| |
| 20 The kinetics parameters used are found in the unit- and cycle-specific analyses.
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| 21 The ejected rod worth is limited to 0.131 % from the fuel enthalpy and temperature case (see Table 15.4.8-2) and use of 0.157 % is conservative for the peak pressure case.
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| June 2011 15.4-67 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES
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| * The CEAE was assumed to result in almost immediate Turbine Admission Valve (TAV) closure (valve closes in 0.2 seconds). In addition, main feedwater was ramped to zero flow in 1.0 seconds.
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| * The operator may cool the NSSS by using manual operation of the AFW system and the ADVs anytime after the trip occurs. However, no credit is taken for the operator action for the first 30 minutes of the event.
| |
| The assumptions and initial conditions used in determining the secondary side steaming releases for the CEAE event are summarized in Table 15.4.8-6.
| |
| C. Results The sequence of events for CEAE peak RCS and main steam pressures for barrier performance is shown in Table 15.4.8-5. The CEAE with a postulated turbine trip and loss of main feedwater results in a peak RCS pressure of 2682 psia for the 3876 MWt case and 2702 psia for the 3990 MWt case (see Figure 15.4.8-7) that is eventually stopped by the PSVs and by the HPPT. The reactor trip causes the closure of the turbine admission valves, which causes a rapid rise in secondary-side pressure to peak pressures of 1348 psia for the 3876 MWt case and 1349 psia for the 3990 MWt case (see Figure 15.4.8-10). The MSSVs open to relieve secondary-side pressure and dissipate energy. Typical NSSS pressure responses to the CEAE transient are presented in Figures 15.4.8-7 through 15.4.8-11 .
| |
| 22 22 Figures 15.4.8-7, 15.4.8-8 and 15.4.8-10 for RCS and Secondary pressures are from the CENTS runs using the PSV opening area of 0.21602 ft2, which is the main purpose of this event simulation. Figures 15.4.8-9 and 15.4.8-11 are from a CENTS run at 3876 MWt using a slightly larger PSV area of 0.0301 ft2. The figures represent typical responses for the transient simulation.
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| June 2011 15.4-68 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-5 TYPICAL SEQUENCE OF EVENTS FOR CEA EJECTION PEAK RCS PRESSURE EVENT Time (sec)
| |
| RTP RTP Event 3876 3990 MWt MWt 0.00 0.00 Mechanical failure of CEDM causes CEA to eject 0.05 0.05 CEA fully ejected 0.07 0.07 Maximum core power 19.70 19.60 HPPT reached 20.45 20.35 HPPT reactor trip, turbine trip, main feedwater trip 21.05 20.95 Scram CEAs begin to drop 21.74 21.62 PSVs open 22.22 22.16 Peak RCS pressure occurs 24.56 24.71 PSVs close 25.34 24.81 MSSVs open 27.95 27.89 Maximum steam generator pressure occurs 32.80 32.70 Steam generator level drops to auxiliary feedwater actuation signal setpoint
| |
| < 1800 < 1800 Long-term automatic plant system actions and NSSS response to this transient are similar to the control element assembly withdrawal at power 1800.0 1800.0 Operator initiates cooldown June 2011 15.4-69 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES The maximum RCS pressure is less than 120% (3000 psia) of RCS design pressure (2500 psia). The maximum primary pressure for this event meets the limiting pressure acceptance criteria of the Standard Review Plan (Reference 1).
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| 15.4.8.5 Radiological Consequences and Containment Performance A. Mathematical Model The number of fuel pins predicted to experience departure from nucleate boiling was calculated using the STRIKIN-II computer program described in section 15.0 with the CE-1 correlation described in chapter 4.
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| A matrix relating the initial and ejected CEA peaking factors to a pin census edit is obtained from Step 6 of the C-E synthesis method and is used to calculate the number of fuel pins experiencing DNB. The time-dependent energy deposition in the NSSS was determined from the above analysis and input into the CESEC III computer program to determine the overall NSSS response to this event.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions used for the fuel evaluation portion of a CEA ejection analysis are delineated in UFSAR Tables 15.4.8-2, 15.4.8-3 and 15.4.8-4.
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| For all cases analyzed, an axial power distribution was chosen to maximize the energy content in the hottest fuel pellet. The remaining parameters were chosen based on the results shown in Chapter 4 of June 2011 15.4-70 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES reference 3. These parameters were varied in the most adverse direction until a COLSS power operating limit was achieved.
| |
| C. Results The spectrum of initial reactor states contained in table 15.4.8-3 was analyzed to show that each case met the criteria established in Regulatory Guide 1.77.
| |
| All cases resulted in a radial average fuel specific enthalpy less than 280 cal/gram at the hottest axial location of the hot fuel pin. The case that resulted in the greatest potential for offsite dose consequences (i.e., the case resulting in the largest number of postulated fuel failures) was identified as the case initiated from full power (FP) beginning of cycle (BOC) initial conditions. The following paragraphs describe this event in detail. Refer to table 15.4.8-4 for the initial conditions and assumptions used for this analysis.
| |
| Figures 15.4.8-1 through 15.4.8-5 show the reactor power, heat flux, and clad and fuel temperatures during the significant portion of transient. Table 15.4.8-1 contains the sequence of events that occur during a CEA ejection initiated from full power BOC initial conditions.
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| Ejection of a CEA causes the core power to increase rapidly due to the almost instantaneous addition of positive reactivity. However, the rapid increase in core power is terminated by a combination of Doppler feedback and delayed neutron effects. This increase June 2011 15.4-71 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES in power results in a high power trip and the reactor power begins to decrease as the CEAs enter the core.
| |
| Reactivity effects are shown in figure 15.4.8-6.
| |
| In the hot channel, the increase in heat flux is such that DNB is calculated to occur, resulting in:
| |
| * a rapid decrease in the surface heat transfer coefficient
| |
| * a rapid decrease in heat flux
| |
| * a rapid increase in clad temperature.
| |
| The transient behavior of the NSSS following a postulated CEA ejection is as follows. The steam generator pressure increases rapidly due to the closure of the turbine control valve following reactor and turbine trip. The steam bypass control system is inoperable on loss of offsite power and therefore is unavailable.
| |
| Subsequently, the reduced reactor power following the reactor trip, in addition to the postulated break in the primary system, causes the RCS pressure and temperature to decrease.
| |
| The steam generator pressure decreases slowly until the main steam safety valves close. The total mass released through the safety valves is approximately 164,160 lbm in the 3954 MWt analysis and 165,528 lbm in the 4070 MWt analysis.
| |
| NRC Regulatory Guide 1.77 recommends that the onset of DNB be used as the basis for predicting clad failure for the postulated CEA ejection event. For PVNGS, the June 2011 15.4-72 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES number of fuel rods that experience DNB is calculated with a statistical convolution technique, which is discussed in UFSAR Section 15E.3.3 and described in Reference 4. The statistical convolution technique involves the summation, over the reactor core, of the number of fuel rods with a specific DNBR value, multiplied by the probability of DNB at that DNBR value. However, to provide a conservative assessment of radiological consequences, 19% of the fuel rods are assumed to suffer clad failure in the 3954 MWt analysis, as shown in Table 15.4.8-6. This assumed value is greater than the percentage of fuel rods that the statistical convolution technique predicts will experience DNB.
| |
| The activity released to the containment (through the ruptured CEDM pressure housing), is assumed to be mixed instantaneously throughout the containment and is available for leakage to the atmosphere. Activity released to the containment building is the activity in primary coolant that is discharged through the CEA break. The activity in the primary coolant consists of primary coolant concentration prior to the accident and fuel-clad gap activity from the fuel rods that experience DNB. Activity is released from the containment building through the power access purge until the purge system release path is isolated by a Containment Isolation Actuation Signal (CIAS) due to a low pressurizer pressure for the 3954 MWt analysis and due to high radiation levels at the purge monitors for the 4070 MWt analysis. Following isolation of the June 2011 15.4-73 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES containment power access purge, airborne activity in containment is released via containment structural leakage.
| |
| The activity released from the secondary system is the activity released to the atmosphere from the main steam safety valves and from the atmospheric dump valves during cooldown.
| |
| Another source of activity release to the environment is due to ESF recirculation leakage outside the containment building, which is assumed to start at 20 minutes after the event. The iodine activity concentration in the recirculating water is determined assuming that 50% of the total reactor coolant system iodine activity is diluted in a containment sump water volume composed of the combined minimum water volumes of the refueling water tank, the reactor coolant system, and the safety injection tanks. The fraction of radioactive iodine in the leakage water that becomes airborne and available for release is based on the flashing fraction.
| |
| Assumptions and parameters that were unique to the evaluation of a CEA ejection event are itemized in table 15.4.8-6. The following paragraphs provide additional clarification to some of the items contained in the table.
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| June 2011 15.4-74 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Activity available for release from containment at time zero.
| |
| The activity available for leakage from containment is based on the following Regulatory Guide 1.77, Appendix B assumptions:
| |
| : 1. The activity in the fuel clad gap is 10% of the iodines and 10% of the noble gases accumulated in the fuel at the end of core life (infinite cycle length is assumed for short lived isotopes (i.e., all isotopes other than Kr-85) per TID 14844 methodology), assuming continuous maximum full power operation. The ORIGEN computer code was used to obtain source term activities for long-lived isotopes (such as I-129 and Kr-85) using a conservative burnup as summarized in Tables 15.4.8-6 (for the 3954 MWt analysis) and 15.4.8-6A (for the 4070 MWt analysis). All of the activity in the fuel gap for fuel rods that are calculated to experience DNB is assumed to be instantaneously available for release from containment.
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| June 2011 15.4-75 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS (Sheet 1 of 5)
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| Parameter Value A. Data and assumptions used to evaluate the radioactive source term
| |
| : 1. General
| |
| : a. Power level, MWt 3954
| |
| : b. Burnup infinite Short lived fission (TID 14844) product (I, Xe, Kr other than Kr-85)
| |
| Long lived fission 44.93 GWD/MTU product (Kr-85) (EOL)(ORIGEN)
| |
| : c. Fuel assumed to 19 experience DNB, %
| |
| : d. Fuel calculated to 0.0 experience incipient centerline melt, %
| |
| : e. Maximum fuel radial peaking 2.0 factor
| |
| : f. Secondary system activity 0.1 before start of the event, uCi/gm I-131 Dose equivalent
| |
| : g. Primary system liquid 571,776 inventory, lbm
| |
| : h. RCS activity before start uCi/gm of the event I-131 3.0 I-132 0.83 I-133 4.4 I-134 0.52 I-135 2.5 June 2011 15.4-76 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS (Sheet 2 of 5)
| |
| Parameter Value A.1 h. RCS activity before start of the uCi/gm event (contd)
| |
| Kr-83m 0.013 Kr-85 6.1 Kr-85m 1.3 Kr-87 1.0 Kr-88 2.8 Kr-89 0.076 Xe-131m 5.9 Xe-133m 0.34 Xe-133 360 Xe-135m 0.74 Xe-135 7.7 Xe-137 0.17 Xe-138 0.63 B. 1. Dose Conversion Factor for Rem/Ci iodine inhalation (Thyroid) are based on ICRP 30, Rem/Ci I-131 1.08+6 I-132 6.44+3 I-133 1.80+5 I-134 1.07+4 I-135 3.13+4
| |
| : 2. Whole body and beta skin Dose Table 15B-2 Conversion Factors for all other isotopes are based on Reg. Guide 1.109.
| |
| C. Metrological data (based on 1986 through 1991 weather data).
| |
| : 1. EAB X/Q, 0-2 hr. sec/m3 Table 2.3-31
| |
| : 2. LPZ X/Q, sec/m3 0-8 hr Table 2.3-31 8-24 hr 24-96 hr 96-720 hr June 2011 15.4-77 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS (Sheet 3 of 5)
| |
| Parameter Value D. Data and assumptions used to estimate Containment release
| |
| : 1. Containment leakage
| |
| : a. Containment net volume, ft3 2.62E+6
| |
| : b. Containment leak rate, % vol containment 0-24 hr 0.1 24 hr - 30 day 0.05
| |
| : 2. Gap activity (of core inventory)
| |
| Iodine 10 Noble gases 10
| |
| : 3. Activity discharged to containment, 100
| |
| % RCS activity
| |
| : 4. Core gap activity available for Ci release I-131 9.92E+07 I-132 1.50E+08 I-133 2.22E+08 I-134 2.59E+08 I-135 2.01E+08 Kr-83M 1.64E+07 Kr-85 1.36E+06 Kr-85M 5.14E+07 Kr-87 8.50E+07 Kr-88 1.26E+08 Kr-89 1.63E+08 Xe-131M 1.02E+06 Xe-133M 5.45E+06 Xe-133 2.22E+08 Xe-135M 7.19E+07 Xe-135 2.11E+08 Xe-137 2.10E+08 Xe-138 1.96E+08 June 2011 15.4-78 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS (Sheet 4 of 5)
| |
| Parameter Value
| |
| : 5. Activity release from secondary system
| |
| : a. Primary to secondary 1.0 leak rate, gal/min
| |
| : b. Steam mass released through 164,160 MSSVs, lbm
| |
| : c. Steam mass released through 1,144,000 ADVs
| |
| : d. Partition factor 1
| |
| Iodine 0.01 Noble gases 1 E. Power access purge parameters
| |
| : 1. Number of valves 2
| |
| : 2. Nominal size of valves, inch 8
| |
| : 3. Time to isolation (start event to isolation), sec 77
| |
| : 4. Total RCS mass discharge to 45,742 containment for 77 sec, lbm F. ESF leakage parameters
| |
| : 1. Total volume of water in ESF sumps 6.98E+04 3
| |
| post event, ft
| |
| : 2. Fraction of RCS activity retained by ESF sumps, %
| |
| : a. Iodine 50
| |
| : b. Noble gases 0.0 1
| |
| Justification for the iodine partition factor of 0.01 is provided in Westinghouse Letter LTR-OA-02-86.
| |
| June 2011 15.4-79 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS (Sheet 5 of 5)
| |
| Parameter Value F. ESF leakage parameters (contd)
| |
| : 3. RAS actuation time 20 (conservative), min
| |
| : 4. Safety injection system 3000 leakage, ml/hr
| |
| : 5. Flashing fraction for 10 iodine, %
| |
| (a)
| |
| G. Control room parameters
| |
| : 1. Control room essential HVAC
| |
| : a. Normal HVAC Isolation time 119 (CPIAS), sec (a)
| |
| Refer to UFSAR Section 6.4 and Appendix 15.B for parameters related to control room volume and operation of the essential HVAC system and to UFSAR Appendix 15.B for control room dispersion coefficients, occupancy factors and breathing rate. The bounding unfiltered infiltration rate to the control room is presented in UFSAR Section 6.4.7.
| |
| June 2011 15.4-80 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS (Sheet 1 of 5)
| |
| Parameter Value A. Data and assumptions used to evaluate the radioactive source term
| |
| : 1. General
| |
| : a. Power level, MWt 4070
| |
| : b. Burnup infinite Short lived fission (TID 14844) product (I, Xe, Kr other than Kr-85)
| |
| Long lived fission 70.00 GWD/MTU product (Kr-85) (EOL)(ORIGEN)
| |
| : c. Fuel calculated to 0.0 experience incipient centerline melt, %
| |
| : d. Secondary system activity 0.1 before start of the event, uCi/gm I-131 Dose equivalent
| |
| : e. Primary system liquid 606,083 inventory, lbm
| |
| : f. RCS activity before start uCi/gm of the event I-131 3.0 I-132 0.83 I-133 4.4 I-134 0.52 I-135 2.5 June 2011 15.4-81 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS (Sheet 2 of 5)
| |
| Parameter Value A.1 f. RCS activity before start of the uCi/gm event (contd)
| |
| Kr-83m 0.013 Kr-85 6.1 Kr-85m 1.3 Kr-87 1.0 Kr-88 2.8 Kr-89 0.076 Xe-131m 5.9 Xe-133m 0.34 Xe-133 360 Xe-135m 0.74 Xe-135 7.7 Xe-137 0.17 Xe-138 0.63 B. 1. Dose Conversion Factor for Rem/Ci iodine inhalation (Thyroid) are based on ICRP 30, Rem/Ci I-131 1.08+6 I-132 6.44+3 I-133 1.80+5 I-134 1.07+4 I-135 3.13+4
| |
| : 2. Whole body and beta skin Dose Table 15B-2 Conversion Factors for all other isotopes are based on Reg. Guide 1.109.
| |
| C. Metrological data (based on 1986 through 1991 weather data).
| |
| : 1. EAB X/Q, 0-2 hr. sec/m3 Table 2.3-31
| |
| : 2. LPZ X/Q, sec/m3 0-8 hr Table 2.3-31 8-24 hr 24-96 hr 96-720 hr June 2011 15.4-82 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS (Sheet 3 of 5)
| |
| Parameter Value D. Data and assumptions used to estimate Containment release
| |
| : 1. Containment leakage
| |
| : a. Containment net volume, ft3 2.62E+6
| |
| : b. Containment leak rate, % vol containment 0-24 hr 0.1 24 hr - 30 day 0.05
| |
| : 2. Gap activity (of core inventory)
| |
| Iodine 10 Noble gases 10
| |
| : 3. Activity discharged to containment, 100
| |
| % RCS activity
| |
| : 4. Core activity available for release Ci (based on 4070 MWt power level)
| |
| I-131 1.02E+08 I-132 1.55E+08 I-133 2.29E+08 I-134 2.68E+08 I-135 2.08E+08 Kr-83m 1.69E+07 Kr-85 1.79E+06 Kr-85m 5.28E+07 Kr-87 8.77E+07 Kr-88 1.30E+08 Kr-89 1.69E+08 Xe-131m 1.06E+06 Xe-133m 5.63E+06 Xe-133 2.29E+08 Xe-135m 7.39E+07 Xe-135 2.18E+08 Xe-137 2.17E+08 Xe-138 2.02E+08 June 2011 15.4-83 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS (Sheet 4 of 5)
| |
| Parameter Value
| |
| : 5. Activity release from secondary system
| |
| : a. Primary to secondary 1.0 leak rate, gal/min
| |
| : b. Steam mass released through 165,528 MSSVs, lbm
| |
| : c. Steam mass released through 1,260,000 ADVs
| |
| : d. Partition factor 1
| |
| Iodine 0.01 Noble gases 1 E. Power access purge parameters
| |
| : 1. Number of valves 2
| |
| : 2. Nominal size of valves, inch 8
| |
| : 3. Time to isolation (start event to isolation), sec 48 F. ESF leakage parameters
| |
| : 1. Total volume of water in ESF sumps 7.023E+04 3
| |
| post event, ft
| |
| : 2. Fraction of RCS activity retained by ESF sumps, %
| |
| : a. Iodine 50
| |
| : b. Noble gases 0.0 1
| |
| Justification for the iodine partition factor of 0.01 is provided in Westinghouse Letter LTR-OA-02-86.
| |
| June 2011 15.4-84 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-6A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A CEA EJECTION ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS (Sheet 5 of 5)
| |
| Parameter Value F. ESF leakage parameters (contd)
| |
| : 3. RAS actuation time 20 (conservative), min
| |
| : 4. Safety injection system 3000 leakage, ml/hr
| |
| : 5. Flashing fraction for 10 iodine, %
| |
| (a)
| |
| G. Control room parameters
| |
| : 1. Control room essential HVAC
| |
| : a. Normal HVAC Isolation time 90 (CPIAS), sec (a)
| |
| Refer to UFSAR Section 6.4 and Appendix 15.B for parameters related to control room volume and operation of the essential HVAC system and to UFSAR Appendix 15.B for control room dispersion coefficients, occupancy factors and breathing rate.
| |
| The bounding unfiltered infiltration rate to the control room is presented in UFSAR Section 6.4.7.
| |
| June 2011 15.4-85 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES
| |
| : 2. The nuclide inventory of the fraction of fuel which reaches or exceeds the initiation temperature of fuel melting at any time during the transient is to be calculated; 100% of the noble gases and 25% of the iodines in this fraction of fuel are assumed to be instantaneously available for release from the containment.
| |
| None of the fuel was calculated to reach or exceed initiation temperature for fuel melting.
| |
| Activity release from the secondary system.
| |
| Activity released from the secondary system is based upon the secondary activity initially in the steam generators plus primary activity resulting from a 1 gpm steam generator tube leak. A steam generator decontamination factor of 100 is applied to radioactive iodine in the 1 gpm primary to secondary leakage. Supporting documentation for applying a steam generator decontamination factor of 100 to radioactive iodine is provided in Reference 7.
| |
| Activity released in secondary system steam includes activity in steam released through the Main Steam Safety Valves (MSSVs) and the Atmospheric Dump Valves (ADVs). Main steam is released via the MSSVs for 30 minutes following accident initiation. From 30 minutes after the accident initiation until shutdown cooling is established, main steam releases are via the ADVs. The mass of steam released through the MSSVs and the ADVs is given in Table 15.4.8-6 for the June 2011 15.4-86 Revision 16
| |
| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 3954 MWt analysis and Table 15.4.8-6A for the 4070 MWt analysis. The Westinghouse (Old CE) CESEC III computer code was used to determine MSSV and ADV secondary system steam releases. It is indicated in Reference 8 that the CESEC III computer code underestimates the decay heat which will cause an under prediction of the steam release. MSSV and ADV steam releases were adjusted to compensate for the CESEC III under prediction of steam release. Steam releases were also adjusted to reflect RSG and power uprate for the 4070 MWt analysis.
| |
| Reactor coolant system activity after event.
| |
| The RCS activity after the event was based on the assumptions given above. The reactor coolant activity after the event is equal to the reactor coolant activity prior to the event plus the increase in activity due to fuel clad gap activity from the fraction of the fuel that experiences DNB.
| |
| June 2011 15.4-87 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-7 REACTOR COOLANT RELEASE TO CONTAINMENT AND CONTAINMENT PRESSURE AND TEMPERATURE VERSUS TIME ANALYZED CORE POWER OF 3954 MWt WITH ORIGINAL STEAM GENERATORS Reactor Coolant Blowdown Containment Pressure and Temperature Time Leak Rate Time P T (sec) (lbm/sec) (sec) (psia) (°F) 0.0 0.00E+00 0.0 14.2 100.0 4.0 1.12E+03 4.4 14.4 103.4 8.0 1.08E+03 9.4 14.8 110.3 12.0 1.05E+03 14.4 15.3 116.6 16.0 1.02E+03 19.4 15.6 122.5 20.0 9.88E+023 24.4 16.0 127.9 24.0 9.62E+02 29.4 16.3 132.9 28.0 9.38E+02 34.4 16.6 137.5 32.0 9.31E+02 39.4 16.9 141.8 36.0 9.29E+02 44.4 17.2 146.0 40.0 9.28E+02 49.4 17.5 150.0 44.0 9.27E+02 59.4 18.0 157.0 48.0 9.26E+02 69.4 18.5 163.4 52.0 9.25E+02 79.4 19.0 169.2 56.0 9.23E+02 89.4 19.4 174.6 June 2011 15.4-88 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-7A REACTOR COOLANT RELEASE TO CONTAINMENT AND CONTAINMENT PRESSURE AND TEMPERATURE VERSUS TIME ANALYZED CORE POWER OF 4070 MWt WITH REPLACEMENT STEAM GENERATORS Reactor Coolant Blowdown Containment Pressure and Temperature Time Leak Rate Time P T (sec) (lbm/sec) (sec) (psia) (°F) 0.0 0.00E+00 0.0 14.2 100.0 4.0 1.17E+03 4.4 14.4 104.3 8.0 1.13E+03 9.4 14.9 111.8 12.0 1.09E+03 14.4 15.2 117.7 16.0 1.06E+03 19.4 15.6 122.6 20.0 1.03E+03 24.4 15.9 126.5 24.0 1.00E+03 29.4 16.1 129.9 28.0 9.75E+02 34.4 16.4 132.9 32.0 9.59E+02 39.4 16.6 135.5 36.0 9.56E+02 44.4 16.8 137.9 40.0 9.55E+02 49.4 17.0 140.1 44.0 9.53E+02 59.4 17.4 144.1 48.0 9.50E+02 69.4 17.8 147.5 52.0 9.47E+02 79.4 18.1 150.6 56.0 9.44E+02 89.4 18.4 153.3 June 2011 15.4-89 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Although it is unlikely that the entire radioactivity noted in table 15.4.8-6 would be instantly released from the core and the RCS, measures have been incorporated in the PVNGS design to keep offsite doses below 10CFR100 limits should such a release take place. Credit for iodine removal by sprays has not been assumed. Leakage from recirculation components outside the containment, as well as containment leakage and containment purge releases, have been assumed.
| |
| Table 15.4.8-8 presents the estimated offsite doses at the exclusion area and low population zone boundaries.
| |
| D.
| |
| | |
| == Conclusions:==
| |
| Radiological Consequences Table 15.4.8-8 presents offsite doses at the exclusion area and low population zone boundaries.
| |
| Analyzed Core Power of 3954 MWt The EAB and LPZ radiological consequences of a CEA Ejection accident are presented in Table 15.4.8-8.
| |
| Thyroid and whole body doses in Table 15.4.8-8 are representative of the parameters presented in Table 15.4.8-6 for a reactor core power level of 3954 MWt with original steam generators. Control room doses are due to a CEA Ejection accident are addressed in Section 6.4.7.3.
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| June 2011 15.4-90 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Analyzed Core Powe of 4070 MWt CEA Ejection analyses for a reactor core power level of 4070 MWt with replacement steam generators determine allowable combinations of accident generated failed fuel percentage (Ff) and fuel radial peaking factor (Fr) such that the total design basis offsite dose values, presented in Table 15.4.8-8, are maintained.
| |
| The limiting dose associated with a CEA Ejection accident is the control room dose. It is indicated in Section 6.4.7.3 and Table 6.4.7-1 that the CEA Ejection accident is the limiting accident for the control room dose. The limiting product of Ff and Fr for the control room accident is 0.30. Fuel cycle characteristics are controlled such that the product of Ff and Fr does not exceed 0.30 following a design basis CEA Ejection accident. This will ensure that the offsite doses are within the design basis dose values presented in Table 15.4.8-8.
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| June 2011 15.4-91 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4.8-8 RADIOLOGICAL CONSEQUENCES OF A CONTROL ELEMENT (a)
| |
| ASSEMBLY EJECTION ACCIDENT (c)
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| Thyroid Dose Whole Body Dose Dose (rem) (rem)
| |
| Exclusion area boundary (EAB) 2-hour consequences Containment contribution 70.2 2.20 (purge system and boundary)
| |
| ESF contribution leakage 0.059 2.59 x 10-4 Secondary contribution 6.68 1.51 (b)
| |
| Total 77 3.71 Low population zone (LPZ) 30-day consequences Containment contribution 148.0 0.83 (purge system and boundary)
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| ESF contribution leakage 0.05 1.2 x 10-5 Secondary contribution 2.20 0.47 Total(b) 151 1.30
| |
| : a. Assumes no credit for containment sprays or non-ESF HVAC filtration.
| |
| : b. Values have been rounded up.
| |
| : c. The bounding control room thyroid dose is given in Section 6.4.7.
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| June 2011 15.4-92 Revision 16
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| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.8.6 Conclusions The rupture of a CEDM nozzle or housing and the subsequent ejection of a CEA will not result in a radial average fuel enthalpy greater than 280 cal/gram at any axial location in any fuel rod. The fuel centerline temperature will be less than fuel melt temperature. For dose consequences, refer to UFSAR Section 15.4.8.5 for details.
| |
| The peak RCS pressure for the CEAE event is less than Service Limit C, 3000 psia (120% of the design pressure of 2500 psia),
| |
| as defined in the ASME code.
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| For the postulated event involving a CEAE, the PVNGS design meets the relevant requirements of the Standard Review Plan (Reference 1).
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| June 2011 15.4-93 Revision 16
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| | |
| PVNGS UPDATED FSAR REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.
| |
| | |
| ==4.9 REFERENCES==
| |
| : 1. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.4.8 -
| |
| Spectrum of Rod Ejection Accidents (PWR) and Section 15.4.8, Appendix A - Radiological Consequences of a Control Rod Ejection Accident (PWR)," NUREG-0800 Rev. 1, July 1981.
| |
| : 2. Intentionally deleted.
| |
| : 3. "C-E Method for Control Element Assembly Ejection Analysis," CENPD-190-A, January 1976.
| |
| : 4. "C-E Methods for Loss of Flow Analysis," CENPD-183-A, June 1984.
| |
| : 5. "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Regulatory Guide 1.109, Revision 1, October 1977.
| |
| : 6. DiNunno, J.J. et al., Calculation of Distance Factors for Power and Test Reactor Sites, TID-14844, March 23, 1962.
| |
| : 7. Westinghouse Letter No. LTR-OA-02-86, Transmittal of Justification Documentation Factor for PVNGS CEA Ejection Analysis, dated June 17, 2002.
| |
| : 8. Westinghouse (Old ABB-CE), TID-97-005, CESEC III Error Notification C-97-002, dated April 8, 1997.
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| June 2011 15.4-94 Revision 16
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| | |
| PVNGS UPDATED FSAR 15.5 INCREASE IN RCS INVENTORY 15.5.1 INADVERTENT OPERATION OF THE ECCS 15.5.1.1 Identification of Event and Causes The inadvertent operation of the emergency core cooling system (ECCS) is assumed to actuate the two high pressure safety injection (HPSI) pumps and open the corresponding discharge valves. This operation occurs as a result of a spurious signal to the system or operator error.
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| 15.5.1.2 Sequence of Events and Systems Operation Inadvertent operation of the ECCS is only of consequence when it occurs below the HPSI pump shutoff head pressure. Above that pressure there will be no injection of fluid into the system.
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| Below the HPSI pump shutoff head pressure, when the shutdown cooling system is isolated the HPSI flow will increase reactor coolant system (RCS) inventory and pressure until the pressure reaches the pump shutoff head pressure. During shutdown cooling system operation the increase in RCS inventory and pressure will be mitigated by the shutdown cooling system relief valves.
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| 15.5.1.3 Analysis of Effects and Consequences Plant operation above the HPSI pump shutoff head pressure will not be impacted by the inadvertent operation of the ECCS. Below the HPSI pump shutoff head pressure when the shutdown cooling system is isolated, there will be an RCS inventory and pressure increase. This increase will be terminated when the pressure rises above the shutoff head pressure. Due to the pressure increase caused by this transient at low RCS temperatures, there is an approach to the brittle fracture limits of the RCS.
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| Examination of the Technical Specifications RCS pressure-June 2001 15.5-1 Revision 11
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| | |
| PVNGS UPDATED FSAR temperature limitations shows that the brittle fracture limits will not be violated for this transient. Should the ECCS inadvertently actuate during shutdown cooling operation, the shutdown cooling relief valves will mitigate the pressure transient so that the temperature-pressure limits are not exceeded. The shutdown cooling relief valves can be isolated when the RCS temperature is above the pressure-temperature limits for brittle fracture of the RCS (See UFSAR Section 5.2.2.11).
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| 15.5.1.4 Conclusion The peak pressurizer pressure reached during the inadvertent operation of the ECCS is well within 110% of design pressure.
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| Additionally, the pressure-temperature limits for brittle fracture of the RCS are not violated by this transient. The fuel integrity is not challenged by this event.
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| 15.5.2 CVCS MALFUNCTION-PRESSURIZER LEVEL CONTROL SYSTEM MALFUNCTION WITH LOSS OF OFFSITE POWER 15.5.2.1 Identification of Event and Causes All events and events plus single failures which cause an increase in RCS inventory as a result of failure or misoperation of the CVCS were examined with respect to the RCS pressure and fuel cladding performance. The Pressurizer Level Control System (PLCS) malfunction was the most limiting event and is more limiting than the UFSAR Chapter 15.5.1 event.
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| The failure of the PLCS is an Anticipated Operational Occurrence (AOO) and is classified as an incident of moderate frequency. In combination with a single failure, the PLCS malfunction is an infrequent event.
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| June 2007 15.5-2 Revision 14
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| PVNGS UPDATED FSAR 15.5.2.2 Sequence of Events and Systems Operation The PLCS Malfunction analyses are performed as separate cases for the primary and secondary peak pressure limits, since these events are not mutually conservative. The sequence of events is presented in Table 15.5.2.2-1.
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| When in the automatic mode, the PLCS responds to changes in pressurizer level by changing charging and letdown flows to maintain the program level. Normally, one charging pump is running with two charging pumps available for automatic startup when a low level setpoint is reached. If the pressurizer level controller fails low or the level setpoint generated by the reactor regulating system fails high, a low level signal can be transmitted to the controller. In response, the controller will start all the charging pumps and close the letdown control valve to its minimum opening, resulting in the maximum mass addition to the RCS.
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| The event initiates at the most limiting conditions allowed per Technical Specifications. The simulated malfunction of the PLCS results in the reduction of letdown to a minimum and the increase in charging to a maximum. The mismatch in charging and letdown will result in a slow, continuous in surge of RCS coolant into the pressurizer. The RCS liquid entering the pressurizer will displace the steam above the liquid. The steam space being compressed will result in increasing RCS pressure. With the Pressurizer Pressure Control System (PPCS) in manual, the pressurizer sprays will not open and the RCS pressure will increase. The event proceeds until a reactor trip occurs on high pressurizer pressure (HPPT).
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| June 2007 15.5-3 Revision 14
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| PVNGS UPDATED FSAR If the Steam Bypass Control System (SBCS) is in automatic mode of operation, primary and secondary peak pressures should remain below relief valve setpoints. However, for the PLCS malfunction analysis, the SBCS is assumed to be in manual mode and credit is not taken for their functioning. The pressure increase in the primary and secondary systems are limited by the Primary Safety Valves (PSVs) and the Main Steam Safety Valves (MSSVs). The maximum RCS pressure occurs in the first two to five seconds following reactor trip.
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| As steam and energy is relieved in the Steam Generators (S/Gs) through the MSSVs, the level in the S/Gs will drop. An auxiliary feedwater actuation signal (AFAS) on low steam generator level occurs. The auxiliary feedwater flow is automatically initiated after a time delay and supplies water from the condensate storage tank (CST) to the Steam Generators.
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| The PLCS malfunction analysis does not credit operator action for the first thirty minutes following the event. Thirty minutes after initiation of the event, the operators stabilize the plant by securing excess charging and/or balance letdown with charging.
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| The operators commence a cooldown using the safety grade Atmospheric Dump Valves (ADVs) and/or non-safety grade SBCS depending on system availability.
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| Analytical setpoints and response times associated with the Reactor Protective System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions are consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS UFSAR delineated in UFSAR Chapter 7.
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| June 2009 15.5-4 Revision 15
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| PVNGS UPDATED FSAR The NRCs Standard Review Plan states that an incident of moderate frequency, such as the PLCS malfunction event, should not generate a more serious plant condition without other faults occurring independently. In addition, the Standard Review Plan states that an incident of moderate frequency, in combination with a single active component failure or single operator error, should not result in the loss of function of any barrier other than the fuel cladding.
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| The PLCS malfunction with pressurizer sprays in manual or off causes a reactor trip on high pressurizer pressure. The limiting 1
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| criteria for this event are peak pressure and secondary pressure which occur within the first two to five seconds after the high pressure reactor trip. Therefore, any single failure which would result in a higher RCS pressure during the transient would have to have an affect during the first two to five seconds following reactor trip.
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| The single failures that have been postulated are listed in table 15.0-0. The failures which affect the RCS behavior during this interval are:
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| * Loss of Normal AC
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| * Failure of the Pressurizer Pressure Control System
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| * Failure of the Steam Bypass Control System
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| * Failure of the Reactor Regulating System
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| * Failure of the Feedwater Control System.
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| 1 Primary integrity can not be compromised (i.e., neither can the primary go solid nor the PSVs pass water)
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| June 2007 15.5-5 Revision 14
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| PVNGS UPDATED FSAR Although these non-class systems would not normally be credited, the difference is the loss of RCP flow. Decreased RCP flow has the effect of decreasing heat transfer to the secondary and therefore would by itself increase primary pressure; however, loss of power to the RCPs reduces the RCP head and has a larger effect and reduces peak RCS pressure. Computer simulation has determined that a loss of power is not limiting. The de-energizing of the RCPs has the effect of reducing peak pressure in the primary and secondary. De-energizing the RCPs reduces pump flow and pump head. The CENTS code explicitly models the RCS pressures in each node and has determined that the maximum flow with RCPs running results in higher peak primary and secondary pressure. Consequently, peak RCS pressure occurs with offsite power being available and RCPs running. Peak secondary pressure also occurs with RCPs running which maximizes the heat transfer to the secondary.
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| Table 15.0-0 is used to determine credible single failures for safety analysis. This table indicates that there are no credible failures that can degrade the PSV and MSSV capacity. Technical Specification 3.7.1 places limits on reactor power and variable overpower trip (VOPT) setpoints when one or more MSSVs are inoperable, thereby ensuring secondary system peak pressure remains within 110% of secondary system design pressure. The LOCV is one of the transients analyzed for validating Technical Specification 3.7.1. A decrease in RCS to steam generator heat transfer due to reactor coolant flow coastdown can be caused by a LOP following a turbine trip. However, the results of the parametric study show that a LOP coinciding or following the High Pressurizer Pressure Trip (HPPT) does not make the primary and secondary side pressures more adverse. In addition, it is assumed that the most reactive control rod fails to insert on scram.
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| June 2007 15.5-6 Revision 14
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| PVNGS UPDATED FSAR Other single failures were examined such as operator action or equipment failure resulting in an inadvertent dilution; however, none were identified as credible. No credible single failure has been identified that makes the consequences worse than as specified under the limiting conditions described for this event.
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| This is similar to conclusion in UFSAR 15.2.3.2 where no credible single failure was identified.
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| Therefore, it was concluded that there is no single failure that would make the maximum primary and secondary pressure more limiting than the LOCV event.
| |
| Regarding the approach to the fuel design limit, the major parameter of concern is the minimum hot channel Departure from Nucleate Boiling Ratio (DNBR). The major factors which cause a decrease in local DNBR are:
| |
| * Increasing Coolant Temperature
| |
| * Decreasing Coolant Flow
| |
| * Decreasing RCS Pressure, and
| |
| * Increasing Local Heat Flux (including radial and axial power distribution effects).
| |
| A decrease in RCS flow is the only parameter which can significantly reduce the minimum DNBR during the event. The PLCS malfunction with pressurizer sprays in manual or off causes a reactor trip. The heat flux starts to drop in the first five seconds. The single failures that have been postulated are listed in table 15.0-0. The failures which affect the RCS behavior during this interval are:
| |
| * Loss of Normal AC
| |
| * Failure of the Pressurizer Pressure Control System
| |
| * Failure of the Steam Bypass Control System June 2007 15.5-7 Revision 14
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| | |
| PVNGS UPDATED FSAR
| |
| * Failure of the Reactor Regulating System
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| * Failure of the Feedwater Control System.
| |
| The loss of normal AC power results in Loss of Power to the:
| |
| * Reactor Coolant Pump,
| |
| * Condensate Pumps,
| |
| * Circulating Water Pumps,
| |
| * Pressurizer Pressure and Level Control System,
| |
| * Reactor Regulating System,
| |
| * Feedwater Control System, and
| |
| * Steam Bypass Control System.
| |
| The effect of losing normal ac power on the PLCS malfunction is as follows. Loss of power to the condensate and circulating water pumps and the feedwater control system initially effects only the secondary system and thus does not affect DNBR in the first five seconds of the transient. Loss of power to the reactor regulating system pressurizer level and pressure control systems renders those systems inoperable. Failure of the pressurizer pressure control system or reactor regulating system cannot appreciably affect any of the major factors which determine DNBR during the first five seconds of the event.
| |
| This inoperability will have no significant impact on DNBR during the first five seconds. Thus, none of the single failures listed in table 15.0-0 will result in a lower DNBR than that predicted for the PLCS malfunction with a loss of offsite power following turbine trip. Loss of power to the reactor coolant pumps is the only significant failure with regard to DNBR which results from a loss of normal ac power.
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| No single failure was identified from table 15.0-0 which would have a significant effect on DNBR prior to the reactor trip.
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| June 2007 15.5-8 Revision 14
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| | |
| PVNGS UPDATED FSAR Therefore, any single failure which would result in a lower DNBR during the transient would have to affect at least one of the above parameters during the first five seconds of the event. The LOP is the only failure that may affect RCS flow.
| |
| PLCS malfunction by itself, however, produces an increasing RCS pressure which compensates for the elevated RCS temperatures such that the available thermal margin does not degrade before the onset of the LOP. Thus, the overall DNBR degradation experienced during an PLCS malfunction event with LOP would be bounded by that of the loss of RCS flow event of UFSAR Section 15.3.1.
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| Table 15.5.2.2-1 presents a chronological sequence of events for the peak primary case which occur during PLCS malfunction from the initial malfunction until the operator stabilizes the plant and initiates plant cooldown.
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| June 2009 15.5-9 Revision 15
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| | |
| PVNGS UPDATED FSAR Table 15.5.2.2-1 Sequence of Events Time Event (sec) 0 Charging flow maximized and letdown flow minimized a
| |
| 296.1 Pressurizer pressure reaches reactor trip analysis setpoint 296.1 High pressurizer pressure trip signal generated 296.1 Turbine trip 296.6 Trip breakers open 297.2 Control rods start inserting 299.0 PSVs open b
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| 299.4 Maximum RCS pressure 300.6 MSSVs first open and continue to cycle during the event c
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| 301.1 PSVs close 308 Maximum steam generator pressure 357 AFW enters the steam generators 1800 Operator initiates plant cooldown
| |
| : a. The time of event initiation is highly dependent on initial conditions (e.g., initial pressurizer level) that have little impact on peak primary or secondary pressure.
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| The relative times from reactor trip until breakers opening, control rods inserting, PSVs opening and peak pressure are reasonably invariant.
| |
| : b. Maximum RCS pressure includes RCP and elevation head in addition to pressurizer pressure.
| |
| : c. When and how often PSVs cycle is highly dependent on initial conditions (e.g., initial pressurizer level RCS temperature) that have little impact on peak primary or secondary pressure.
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| June 2009 15.5-10 Revision 15
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| | |
| PVNGS UPDATED FSAR 15.5.2.3 Core and System Performance A. Mathematical Model The nuclear steam supply system (NSSS) response to PLCS malfunction with a reactor trip and a turbine trip was simulated using the CENTS computer program described in subsection 15.0.3.
| |
| B. Input Parameters and Initial Conditions This event was not analyzed for fuel failure. Fuel failure as a result of DNB or peak linear heat rate is not of concern for this event. Pressure is increasing in this event and no power peaking or low RCS flow would occur that would not already be bounded by loss of flow.
| |
| C. Results Since this transient causes an increase in RCS pressure due to an increase in primary coolant inventory, the DNBR increases. Therefore, the acceptance criterion regarding fuel performance is met.
| |
| 15.5.2.4 Primary and Secondary Barrier Performance A. Mathematical Model The nuclear steam supply system (NSSS) response to PLCS malfunction with a reactor trip and a turbine trip was simulated using the CENTS computer program described in subsection 15.0.3.
| |
| B. Input Parameters and Initial Conditions The initial conditions are set conservatively with respect to allowable TS limits, plant design, operating procedures, and instrument uncertainties.
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| June 2007 15.5-11 Revision 14
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| | |
| PVNGS UPDATED FSAR The initial conditions were varied within the ranges of steady state operation configurations (i.e.,
| |
| specified by the Technical Specifications, plant configuration, and design specifications) to determine the set of initial conditions that produce the most adverse consequences.
| |
| Parameters of interest include initial core inlet temperature, core inlet flow, pressurizer pressure, pressurizer water level, steam generator level, Moderator Temperature Coefficient (MTC), Fuel Temperature Coefficient (FTC), fuel rod gap conductances, kinetics parameters, LOP, and SG tube plugging. Starting from a base case, one parameter at a time is changed to establish the trends for the RCS and steam generator pressure.
| |
| For peak primary pressure and peak secondary pressure, neither the net charging rate nor the initial level have a significant impact. Although not discussed in detail, the PLCS malfunction with the sprays in automatic was analyzed and determined to be not limiting with respect to overfilling the pressurizer.
| |
| For this scenario, total charging flow due to all three pumps is 132 gal/min. Although a conservative bleedoff of 2 gpm per pump could be credited, no credit for this bleedoff has been taken. The minimum letdown flow is 30 gal/min.
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| June 2009 15.5-12 Revision 15
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| | |
| PVNGS UPDATED FSAR Table 15.5.2.4-1 Limiting Initial Conditions for PLCSM Peak Primary and Secondary Pressure Value Parameter Primary Secondary Peak Peak Pressure Pressure Initial core power (% of RTP) 102 a
| |
| Initial core inlet temperature (°F) 566 b
| |
| Initial pressurizer pressure (psia) 2325 Initial RCS flow (% design) 116 95 Initial pressurizer water level (%) 23.9 Pressurizer heaters On Pressurizer sprays Off Charging and letdown flows (gpm) 135/30 b.
| |
| MTC (/°F) -0.0002 b.
| |
| Maximum b.
| |
| FTC (/°F) Most negative
| |
| : b. b.
| |
| Prompt neutron lifetime (l*) Min Max Fuel gap conductance (Btu/hr-ft2-°F) 500 SCRAM delay time (sec) 0.5 CEA holding coil delay (sec) 0.6 CEA worth at trip - WRSO (%) 8.0 Plugged tubes (per steam generator) 0 Initial steam generator level (% WR) 28.22 28.22 AFW flow (gpm/pump) 650 AFW delay time (sec) 46 PSV setpoint tolerance +3%
| |
| PSV blowdown 5%
| |
| MSSV Setpoint Tolerance +3%
| |
| MSSV blowdown 5%
| |
| LOP No
| |
| : a. The sensitivity of this parameter from minimum to maximum on peak primary pressure is much smaller than the uncertainties.
| |
| : b. The sensitivity of this parameter from minimum to maximum on peak primary and secondary pressure is much smaller than uncertainties. Use of nominal values or other changes would have a negligible effect.
| |
| June 2009 15.5-13 Revision 15
| |
| | |
| PVNGS UPDATED FSAR Table 15.5.2.4-2 Initial Conditions with Significant Impact on Peak Primary and Secondary Pressure for PLCSM Value Parameter Peak Pressure Primary Secondary Initial core power (% of RTP) 102 Initial RCS flow (% design) 116 95 Fuel gap conductance (Btu/hr-ft2-°F) 500
| |
| Maximum SCRAM delay time (sec) 0.5 CEA holding coil delay (sec) 0.6 CEA worth at trip - WRSO (%) 8.0 Plugged tubes (per steam generator) 0 PSV setpoint tolerance +3%
| |
| Inlet temperature (°F) 566 MSSV Setpoint Tolerance +3%
| |
| C. Results The dynamic behavior of NSSS parameters for a PLCS malfunction with a reactor and turbine trip turbine trip is presented in figure 15.5.2-2 to 15.5.2-11. Note that the peak secondary pressure occurs for maximum primary temperature so that the figures represent both peak and primary pressure and peak secondary transient.
| |
| Failure of the PLCS causes an increase in reactor coolant system inventory initiated by the startup of the third charging pump coupled with the decrease in letdown flow to its minimum.
| |
| With the PPCS in the manual mode and the proportional sprays turned off, increase in RCS inventory results in a pressurizer pressure increase to the reactor trip analysis setpoint.
| |
| June 2009 15.5-14 Revision 15
| |
| | |
| PVNGS UPDATED FSAR Since the SBCS is conservatively assumed in the manual mode and the rate of closure of the turbine stop valves is faster than the rate of control rod insertion, pressurizer pressure increases which opens the primary safety valves. Peak primary pressure is 2681 psia which is less than 110% of the design pressure (2500 psia) or 2750 psia.
| |
| The unavailability of the SBCS causes the steam generator pressure to increase, causing the MSSVs to open. The decreasing core power and the safety valves function to limit the peak steam generator pressure to 1382 psia which is less than 110% of the design pressure (1270 psia) or 1397 psia.
| |
| At 1800 seconds, the operator stabilizes the plant and initiates plant cooldown using ADVs or SBCS.
| |
| This event has also been evaluated for peak pressurizer level (overfill) to assess the impact on the operability of the pressurizer safety valves. However, the limiting transient is discussed in section 18.II.D in accordance with NUREG-0737.
| |
| 15.5.2.5 Radiological Consequences and Containment Performance PLCSM is a moderate frequency event in which no fuel damage occurs. As noted above, the steam is discharged by the PSVs.
| |
| That steam is directed to the Reactor Drain Tank (RDT). If a second lift of the PSVs occurs, the RDT rupture disk will rupture. The steam would be released to the containment.
| |
| Since fuel failure will not occur as a result of not exceeding any Specified Acceptable Fuel Design Limits (SAFDLs), the dose and effluents will be controlled to the limits specified in 10CFR20. Therefore, radiological consequences are not calculated for this event and containment isolation is not credited.
| |
| June 2009 15.5-15 Revision 15
| |
| | |
| PVNGS UPDATED FSAR 15.5.2.6 Conclusion For the PLCS malfunction following turbine trip event, the maximum RCS pressure remains below 110% of RCS design pressure (2750 psia), thus ensuring primary system integrity. Likewise, the maximum secondary system pressure remains below 110% of design pressure (1397 psia), thus ensuring secondary system integrity.
| |
| Since this transient causes an increase in RCS pressure due to an increase in primary coolant inventory, the DNBR increases.
| |
| Therefore, the acceptance criterion regarding fuel performance is met.
| |
| June 2009 15.5-16 Revision 15
| |
| | |
| PVNGS UPDATED FSAR 15.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY/RELIEF VALVE The inadvertent opening of a pressurizer safety valve event as described in NRC Standard Review Plan 15.6.1 is evaluated in the emergency core cooling systems analyses (section 6.3).
| |
| 15.6.2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT 15.6.2.1 Identification of Causes and Frequency Classification Direct release of reactor coolant may result from a break or leak outside containment in a letdown line, instrument line, or sample line. The double-ended break of the Chemical and Volume Control System (CVCS) letdown line outside containment upstream of the letdown line control valve (DBLLOCUS) was selected for this analysis because it is the largest line, and results in the largest release of reactor coolant outside the containment.
| |
| The single active failure of an isolation valve was not considered in the analysis because the letdown line includes two isolation valves in series situated inside the containment. Hence, failure of one isolation valve does not make the consequences of the event more severe.
| |
| A letdown line break can range from a small crack in the piping to a complete double-ended break. The cause of the event may be attributed to corrosion which forms etch pits, or to fatigue cracks resulting from vibration or inadequate welds.
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| June 2005 15.6-1 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The letdown line break is classified as a limiting fault event whose occurrence is not expected, but is postulated because its consequences include the potential for release of significant amounts of radioactive material.
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| 15.6.2.2 Sequence of Events and Systems Operation A double-ended break of the letdown line outside containment upstream of the letdown line control valve releases primary fluid to the auxiliary building as shown in figure 15.6.2-6.
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| This discharge rate is more than twice the maximum expected letdown flow. The event will set off a number of alarms.
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| Table 15.6.2-1 lists the alarms that would be noted by the reactor operator in the control room.
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| Table 15.6.2-1 ALARMS THAT WILL BE ACTUATED FOR THE DBLLOCUS EVENT
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| : 1. Regenerative heat exchanger high exit temperature alarm
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| : 2. Letdown line low pressure alarm (downstream of the break)
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| Letdown line process radiation monitor loop low flow 3.
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| alarms
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| : 4. Auxiliary building high radiation alarm
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| : 5. Auxiliary building high temperature alarm
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| : 6. Pressurizer level deviation alarm
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| : 7. Auxiliary building sump high level alarm
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| : 8. Pressurizer low level alarm June 2007 15.6-2 Revision 14
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The first three alarms listed in table 15.6.2-1 (the regenerative heat exchanger (RHX) high exit temperature alarm, the letdown line low pressure alarm, and the low flow alarm in the process radiation monitor loop) will immediately alert the operator after the initiation of the event. The high RHX outlet temperature alarm, in addition to sounding the alarm, also initiates isolation of the letdown line by closing one of the two letdown line isolation valves inside the containment.
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| However, no credit is taken for this isolation action in the analysis. Secondly, the high radiation and high temperature alarms in the auxiliary building are expected to be triggered within a few seconds after the event initiation. Thirdly, the pressurizer level deviation alarm is expected to alert the operator approximately one minute after the initiation of the event. Next, the auxiliary building sump high level alarm is expected to be triggered within a few minutes after the initiation of the event. Lastly, the pressurizer low level alarm will occur in approximately 10 minutes after the initiation of the event.
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| For most nominal starting configurations, the makeup system will be able to restore VCT level before the VCT low level alarm is actuated.
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| The analysis assumes that the operator isolates the letdown line 10 minutes after the first three alarms resulting from the DBLLOCUS, thereby terminating any further release of primary flow to the auxiliary building. Subsequently, the operator is assumed to take appropriate steps for a controlled reactor shutdown. The assumption of operator action within 10 minutes after the first few alarms are triggered is based on an ANS/ANSI Standard (Reference 1). The 10 minutes is the minimum June 2007 15.6-3 Revision 14
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY time for the letdown line break event category that shall elapse from the time of the alarm until operator actions can be considered for initiation of safety functions.
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| Table 15.6.2-2 presents a chronological sequence of events which occur following a double-ended break of the letdown line until the operator takes action to terminate the primary system fluid loss 10 minutes after the initiation of the event.
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| Table 15.6.2-2 SEQUENCE OF EVENTS FOR THE DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT UPSTREAM OF THE LETDOWN CONTROL VALVE Time (sec)
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| Event 3876 3990 MWt MWt 0 0 Letdown line rupture occurs setting off alarms listed in table 15.6.2-1 87.2 88.0 Pressurizer backup heaters turn on 243.9 245.4 Third charging pump starts 489.8 492.4 Pressurizer heaters turn off due to low pressurizer level 600 600 Operator isolates the letdown line break and takes steps for a controlled shutdown of the reactor The operator diagnoses the event based on alarms specified in table 15.6.2-1, and generates a manual reactor trip after isolating the letdown line. The control element assemblies (CEAs) fall into the core to provide a negative reactivity insertion. The boron concentration is adjusted to ensure that a proper negative reactivity shutdown margin is achieved prior to cooldown. The boron concentration is adjusted by June 2007 15.6-4 Revision 14
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY manually controlling the chemical and volume control system (CVCS).
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| The turbine automatically trips on the manual reactor trip. The steam bypass control system (SBCS) automatically actuates and opens the steam bypass valves to dump steam to the condenser.
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| The main feedwater control system (FWCS) responds to the reactor trip and generates a reactor trip override signal which reduces feedwater flow to a value commensurate with the decay heat load.
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| The plant cooldown is controlled by manual operation of the SBCS. The main feedwater pumps are manually controlled and continue to supply feedwater until the operator starts the auxiliary feedwater pump and secures the main feedwater pumps.
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| The shutdown cooling system (SCS) is manually actuated when reactor coolant system (RCS) temperature and pressure have been reduced to approximately 350F and 400 psia. This system provides sufficient cooling flow to cool the RCS to cold shutdown.
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| 15.6.2.3 Core and System Performance A. Mathematical Model The nuclear steam supply system (NSSS) response to a DBLLOCUS was simulated with the CENTS computer program (see UFSAR Section 15.0.3). The analysis assumes critical flow through the break and accounts for letdown line losses and for operation of the pressurizer pressure control system (PPCS) and pressurizer level control system (PLCS).
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| The CETOP-D computer code (see UFSAR Section 4.4), which uses the CE-1 CHF correlation, was used to calculate the initial June 2005 15.6-5 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY conditions corresponding to the DNBR SAFDL. CETOP-D was also used to calculate transient DNBR values and thermal-hydraulic conditions at the time of the minimum DNBR.
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| B. Input Parameters and Initial Conditions Table 15.6.2-3 lists the assumptions and initial conditions used for the letdown line break analysis. Conditions were chosen to maximize the primary system mass release into the auxiliary building atmosphere for DBLLOCUS. This, in turn, leads to the most conservative predictions of radiological releases.
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| The initial conditions and NSSS characteristics used in this analysis of the maximum total radiological release for the letdown line break were based on parametric studies. The parameters evaluated were initial core inlet temperature, initial power level, initial pressurizer pressure, initial core inlet flow rate, initial pressurizer liquid inventory, MTC, and friction loss (K-Factor) effect. The maximum total mass release is obtained when the transient is initiated with the following parameters: the maximum core power, minimum allowed core inlet temperature, maximum core flow, a maximum initial pressurizer pressure and water level and most negative MTC (see table 15.6.2-3). However, for dose calculations, it was determined that maximum core inlet temperature results in higher flashing fractions (lower mass release) and a higher overall integrated release for the dose calculation. Hence, the higher inlet temperature condition is the most limiting event for the dose results even though it results in a smaller mass release.
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| June 2005 15.6-6 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY All control systems are assumed to be in the automatic mode to maximize the total primary mass release. The break is assumed to be the full cross sectional area (double-ended) pipe break 2
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| (0.01556 ft ).
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| Table 15.6.2-3 ASSUMED INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT UPSTREAM OF THE LETDOWN CONTROL VALVE Parameter Assumed Value Initial Core Power Level (102% of Rated), MWt 3954 (3876 MWt core) 4070 (3990 MWt core)
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| Initial Core Inlet Temperature, °F 568 Initial Pressurizer Pressure, psia 2325 Initial RCS Flow (116% of design), 10 lbm/hr 190.2 6
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| Initial Pressurizer Water Level, % 59 Moderator Temperature Coefficient (MTC), -4.2 10 /°F
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| -4 Fuel Temperature (Doppler) Coefficient (BOC) Least negative Kinetics Parameters (EOC) minimum CEA Worth at Trip (most reactive CEA fully no reactor trip withdrawn), %
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| Hot Spot Gap Conductance, BTU/hr-ft -°F 2
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| 5755 Number of Plugged Tubes per Steam Generator 0 Break Size (double-ended), ft 0.01556 2
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| June 2005 15.6-7 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY C. Results The responses of key parameters as a function of time are presented in figures 15.6.2-1 to 15.6.2-13 for the DBLLOCUS.
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| The decrease in the primary system mass causes the pressurizer pressure to decrease about 200 psi during the 10 minute transient. Also, during the same time period the pressurizer level decreases about 12.6 feet to a new level of about 10.7 feet above the lower tap (approximately 24%).
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| Ten minutes into the transient the operator isolates the letdown line, terminating the release of primary fluid outside the containment. The amount of reactor coolant released into the auxiliary building is shown in figure 15.6.2-6. Some time shortly after the termination of the primary system mass release, the operator manually trips the reactor.
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| The minimum departure from nucleate boiling ratio (DNBR) remains above the SAFDL value of 1.34 during the transient (see figure 15.6.2-13).
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| 15.6.2.4 RCS Pressure Boundary and Barriers Performance A. Mathematical Model The mathematical model used for evaluation of barrier performance is described in section 15.6.2.3 A.
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| B. Input Parameters and Initial Conditions The input parameters and initial conditions relevant to barrier performance for the limiting fault event are the same as those presented in table 15.6.2-3 for a letdown line break from full power conditions.
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| June 2005 15.6-8 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY C Results At 10 minutes into the letdown line break transient with PPCS and PLCS in operation, the breached line is isolated. The total primary system coolant released outside the containment building is shown in figure 15.6.2-6.
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| The double-ended break of a letdown line outside containment upstream of the letdown line control valve results in gradual depressurization of the reactor coolant system. The secondary side pressure does not increase above its initial condition value during the transient and remains well below 110% of design (1397 psia) ensuring integrity of the main steam system.
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| 15.6.2.5 Radiological Consequences and Containment Performance A. Mathematical Model The DBLLOCUS event is indicated by several non-Q1E alarms listed in table 15.6.2-1. The first three alarms are expected to take place immediately following initiation of the event. Ten minutes after the initiating event, the letdown line is isolated by the reactor operator.
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| The methodology to determine the most adverse dose results includes multiplying the amount of primary coolant released into the auxiliary building by the maximum flashing fraction.
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| The flashing fraction is based on the enthalpy and pressure of the primary coolant at the break location. Both the mass released and enthalpy values are obtained from the CENTS code.
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| The methodology ensures that the multiplication of the flashing fraction by the mass released remains bounded by previous analyses.
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| June 2005 15.6-9 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The mathematical model used to analyze the activity released during the course of the accident is described in section 15.0.4 (Radiological Consequences) and control room doses are discussed in section 6.4.7 (Bounding System Unfiltered Air Inleakage for Radiological Design).
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| B. Assumptions and Parameters The letdown line break outside containment results in the discharge of radioactivity to the environment. Worst case or conservative assumptions are:
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| : 1. The initial activity level of the primary coolant is assumed to be 3.81 µCi/gm dose equivalent I-131(DEQ I-131) as calculated using ICRP-30 iodine dose conversion factors.
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| This corresponds to the maximum equilibrium value with 1%
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| failed fuel.
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| : 2. A concurrent iodine activity spike with a spiking factor of 500 for the GIS is assumed to occur coincident with initiation of the transient.
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| : 3. The quantity of steam (coolant times maximum flashing fraction) released outside containment is maximized by assuming the most adverse initial conditions and by assuming critical flow through the break.
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| : 4. The blowdown decontamination factor (DF) is calculated to determine how much iodine contained in the released primary mass is assumed to be airborne. This is based on the fraction of primary fluid that flashes to steam in the auxiliary building, based on the enthalpy of the escaping fluid.
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| June 2005 15.6-10 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY
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| : 5. No credit is taken for the retention within the auxiliary building and filtration system.
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| : 6. No credit is taken for ground deposition of the activity that escapes the auxiliary building or of decay in transit to the exclusion area boundary.
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| : 7. The meteorological conditions assumed to be present at the site during the course of the accident are based on /Q values which are expected to be conservative 95% of the time. This condition results in the poorest values of atmospheric dispersion calculated for the EAB or low population zone (LPZ) outer boundary. Furthermore, no credit has been taken for the transit time required for activity to travel from the point of release to the EAB or LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions are conservative. The PVNGS site specific /Q value used for this analysis is discussed in section 2.3.4.
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| : 8. Parametrics performed in the dose calculation concluded that all 3 charging pumps in operation prior to initiation of the line break resulted in the highest doses. The higher assumed charging flow maximizes the Iodine source in the primary and bounds all possible modes of operation.
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| C. Results The radiological consequences resulting from the occurrence of a postulated letdown line rupture have been conservatively analyzed using assumptions and models described in the preceding subsections. The thyroid inhalation dose and whole body dose have been analyzed for the two-hour dose at the June 2005 15.6-11 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY exclusion area boundary. The two-hour thyroid inhalation dose and whole body dose values remain less than a small fraction of 10 CFR 100 guidelines (10% of 10 CFR 100 limits) as listed in table 15.6.2-4. The containment barrier is not challenged since no releases to containment occur during this event.
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| Table 15.6.2-4 RADIOLOGICAL CONSEQUENCES FOR THE DBLLOCUS Target Organ, Evaluation Period, Dose SRP Limit Location (REM) (REM)
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| Thyroid, 0-2 hrs at EAB 10.1 30 Whole Body, 0-2 hrs at EAB 0.021 2.5 15.6.2.6 Conclusions The double-ended break of letdown line outside containment upstream of the letdown line control valve results in gradual depressurization of the reactor coolant system. The peak secondary pressure does not increase during the transient and remains below 110% of design ensuring integrity of the main steam system. The minimum DNBR remains above the SAFDL value, thereby ensuring fuel cladding integrity. During the 600 second duration of the transient the amount of coolant released is shown in figure 15.6.2-6. The amount of coolant released multiplied by the maximum flashing fraction remains bounded by the assumed dose calculation value of 10,255 lbm. The event generated iodine spikes (GIS) radiological releases results in a two-hour thyroid inhalation dose and whole body dose that are a small fraction of 10 CFR 100 guidelines (10% of 10 CFR 100 limits).
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| June 2005 15.6-12 Revision 13
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY For the postulated event involving a letdown line break from full power conditions, the PVNGS design conforms with the applicable requirements of 10 CFR Part 50 Appendix A, General Design Criteria 55 (Reactor Coolant Pressure Boundary Penetrating Containment), as described in the NRC Standard Review Plan (Reference 2).
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| 15.6.3 STEAM GENERATOR TUBE RUPTURE The Steam Generator Tube Rupture (SGTR) accident is a penetration of the barrier between the Reactor Coolant System (RCS) and the main steam system, and results from a double ended guillotine break of a steam generator U-tube.
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| A SGTR is classified as a limiting fault event, whose occurrence is not expected during the lifetime of the plant, however, the event is postulated because the consequences include the potential for the release of significant amounts of radioactivity to the environment (i.e., an ANSI N18.2-1973 Condition IV event).
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| Acceptance criteria for SGTR safety analyses are established on the basis of radiological dose consequences, whose acceptance limits vary with the analytical assumptions used for the SGTR event combination under consideration (e.g., availability of offsite power, iodine spiking, single failure, etc.).
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| SGTR event combinations are also evaluated to ensure that Emergency Operating Procedure (EOP) mitigation strategies provide sufficient direction to plant operators to prevent the occurrence of steam generator overfill. These evaluations are performed in accordance with 10 CFR 50, Appendix B, Criterion III (Design Control) processes. Prevention of steam generator overfill is not an acceptance criterion for SGTR analyses in NCR Standard Review Plan (NUREG-0800) Section 15.6.3.
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| June 2011 15.6-13 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.3.1 Steam Generator Tube Rupture Without a Loss of Offsite Power The SGTR with a LOP and an additional Single Failure (SF) presented in UFSAR Section 15.6.3.2 bounds all SGTR events, including the SGTR without a LOP, with respect to offsite dose consequences. The LOP causes the plant to experience a loss of turbine load, loss of normal feedwater flow and loss of condenser vacuum that results in direct steaming to the atmosphere via stuck-open Atmospheric Dump Valve (ADV), the most limiting SF, and the Main Steam Safety Valves (MSSVs).
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| Based on the loss of the condenser vacuum and the direct steaming through the stuck-open ADV, the SGTR with LOP and a single failure is more limiting than the SGTR with respect to radiological consequences.
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| Additionally, operator action in conformance with Emergency Operating Procedures (EOPs) will prevent overfilling of the steam generators following a SGTR event. Steam generator level control is afforded primarily by controlling the delivery of feedwater (or auxiliary feedwater) to the steam generators, and by releasing steam through the Steam Bypass Control System (SBCS) or the Atmospheric Dump Valves (ADVs).
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| However, SRP Section 15.6.3 requires that the following two criteria must be met for the SGTR with no LOP:
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| * For the SGTR event with a pre accident iodine spike, the calculated dose should not exceed the 10 CFR 100 limits, i.e., 300 REM for the thyroid.
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| * For the SGTR event with an accident generated iodine spike, the calculated dose should not exceed a small part of 10 CFR 100, i.e., 30 REM to the thyroid.
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| June 2011 15.6-14 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY In order to evaluate the dose consequence for the SGTR event, explicit modeling of the transient is not performed. The NSSS response for the SGTR with LOP event presented in UFSAR Section 15.6.3.2 was conservatively assumed to apply for the SGTR event. The dose calculation for the loss of offsite power event was modified with the assumption that the start of the cooldown at 2081 seconds (as depicted in Table 15.6.3-1), is conducted with the Steam Bypass Control System (SBCS) to the condenser instead of the ADVs for the SGTR with LOP event. The activity released to the environment is through the condenser with a DF of 100. With the exception of the performance of radiological consequences, the details outlined below for the SGTR with LOP event including the assumptions applies for the SGTR event. Thus only the radiological consequences are reported below.
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| 15.6.3.1.1 Radiological Consequences A. Results The reported values for the 2-hour EAB and the 8-hour LPZ thyroid inhalation doses for the PIS and the GIS cases are presented in Table 15.6.3-1a.
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| Table 15.6.3-1a RADIOLOGICAL CONSEQUENCES FOR THE SGTR EVENT Event Case Evaluation Period & Dose Location (REM) 0-2 hrs at EAB 2.9 GIS 0-8 hrs at LPZ 0.9 0-2 hrs at EAB 6.2 PIS 0-8 hrs at LPZ 1.8 June 2011 15.6-15 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.3.1.2 Conclusions The radiological releases calculated for the SGTR event were demonstrated to be within the SRP Section 15.6.3 guidelines.
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| Specifically, the calculated dose for the GIS does not exceed 30 REM (small part of 10 CFR 100 limit) for the thyroid.
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| 15.6.3.2 Steam Generator Tube Rupture With a Loss of Offsite Power and a Single Failure 15.6.3.2.1 Identification of Causes and Frequency Classification The SGTR with a LOP and a SF event is initiated by the rupture of a steam generator tube, resulting in a failure of the barrier between the RCS and the main steam system. It employs the conservative assumptions of the Standard Review Plan as described in Reference 2 (e.g., loss of offsite power, accident meteorology, iodine spiking, etc.). However, it also assumes that the challenge to the plant is enhanced by actions and failures beyond those postulated by Part 15.6.3 of the Standard Review Plan (Reference 9), as described below.
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| The analysis of the event postulates that the operators open the ADVs on both steam generators, at which time the ADV on affected steam generator runs to the full open position and sticks full open for the duration of the transient. This ADV is presumed to remain full open despite the availability of control systems that would close the ADV as well as a hand wheel which could be used by the operators to manually close the ADV.
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| June 2011 15.6-16 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY A SGTRLOPSF is classified as a limiting fault event, whose occurrence is not expected during the lifetime of the plant, however, the event is postulated because the consequences include the potential for the release of significant amounts of radioactive materials. These releases cannot result in radiological doses that exceed the 10 CFR 100 limits.
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| 15.6.3.2.2 Sequence of Events and Systems Operation Integrity of the barrier between the RCS and main steam system is significant from a radiological release standpoint. The radioactivity from the leaking steam generator tube mixes with the shell-side water in the affected steam generator. Prior to turbine trip, the radioactive water is transported through the turbine to the condenser as steam, where the non-condensable radioactive materials are released via the condenser air removal pumps. Following the reactor trip and turbine trip followed by the postulated LOP, the condenser is unavailable.
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| As a result, the radioactive fluid is released through the MSSVs or ADVs.
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| Table 15.6.3-1 provides a sequence of events for the SGTRLOPSF, from the initiation of the break to the attainment of SDC entry conditions.
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| The double-ended break of a steam generator tube results in a primary-to-secondary leak rate that exceeds the capacity of the charging pumps. The automatic operation of the PLCS reduces the letdown flow to a minimum value less than a minute into the event.
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| June 2011 15.6-17 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY As a result of the inventory loss, pressurizer level (See Figure 15.6.3-5) and RCS mass gradually decreases from its initial value (See Figure 15.6.3-7). The RCS pressure also decreases (See Figure 15.6.3-2). The backup heaters are energized by the action of the PPCS to mitigate against further depressurization and the PLCS starts the third charging pump.
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| Following the rupture of the steam generator tube, coolant begins leaking from the RCS into the steam generators (See Figure 15.6.3-10) and the RCS continues to depressurize (See Figure 15.6.3-2). The decrease in RCS pressure typically results in a CPC trip on margin to hot-leg saturation. The auxiliary trip in the CPC (Primary Pressure out of Analyzed Range - Low) and the Low Pressurizer Pressure Trip in the Plant Protection System could also occur. Sensitivity studies for trip times showed an early trip produces more adverse radiological consequences, principally because the early trip results in the ADVs being opened sooner by the operators. A manual trip at 100 seconds was assumed in order to bound automatic trip time, estimated to occur somewhat later in the transient.
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| The reactor trip causes the power to drop rapidly (See Figure 15.6.3-1). It also causes the turbine to trip and the main feedwater flow to drop rapidly to zero. When the reactor and turbine trips occur, with the SBCS in manual mode, the secondary system begins to pressurize (See Figure 15.6.3-8) 1 until the MSSVs open , mitigating further secondary system pressure increases and releasing steam directly to the atmosphere. The LOP occurs three seconds after the turbine trip and the plant loses the turbine load, normal feedwater flow, forced reactor coolant flow, condenser vacuum and steam June 2011 15.6-18 Revision 16
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY generator blowdown capability. Heat removal is initially achieved by steaming directly to atmosphere through the MSSVs and ADVs (See Figure 15.6.3-14) and AFW flow (See Figure 15.6.3-9), which is automatically initiated to both steam generators. Non-condensable material released from the condenser and the steam releases directly to the atmosphere from the MSSVs contribute to the dose consequences.
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| Based on the standard post-trip procedure for PVNGS, as the pressure in the steam generators rises above the allowable range, the operator opens one ADV in each steam generator after the trip to minimize cycling of the MSSVs. Earlier opening of the ADVs results in more adverse dose consequences since it increases the release to the atmosphere. The time between the trip and operator action to open the ADVs is assumed to be 2 minutes to bound operating experience and simulator scenarios.
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| The ADV on the affected steam generator is assumed to go to a full open position, causing an increased blowdown of the affected steam generator.
| |
| 1 Figure 15.6.3-8 does not display the peaks of the secondary pressure spikes just after the LOP that result in the opening of the MSSVs. This is a result of a lower frequency of data recording (50 second interval) for plot files for the long-term response and the fact that these pressure spikes are only about 20 seconds long.
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| June 2011 15.6-19 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The continued decrease in RCS and pressurizer pressure results in a SIAS, and HPSI flow to the RCS begins. The pressurizer empties due to the primary-to-secondary leakage and post-trip RCS liquid shrinkage. Decreasing steam generator pressure due to flow through the ADVs results in a MSIS on low steam generator pressure being generated. The MSIS results in the closure of the MSIVs, causing a pressure differential to grow between the two steam generators as the flow through the fully-open ADV on the affected steam generator drives its pressure down faster than that of the unaffected steam generator due to the lower flow through the partially-open ADV on that steam generator. As the pressure differential increases between the two steam generators, the AFW to the affected steam generator is terminated by the lockout.
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| The LOP and the RCS flow coastdown result in the reduction of flow to the reactor vessel upper head region. This region becomes thermal-hydraulically de-coupled from the rest of the RCS, and due to flashing caused by the depressurization and boil off from the metal structure-to-coolant heat transfer, voids begin to form in this region (See Figure 15.6.3-6) as the saturation pressure falls below the fluid temperature (See Figure 15.6.3-4). HPSI flow delivery to the RCS halts further RCS shrinkage and depressurization and the voids eventually collapse.
| |
| The remainder of the transient is determined by the diagnostic and mitigating operator actions, which are based on the EOP instructions for SGTR and the Functional Recovery Procedures.
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| Timing those operator actions is consistent with or conservative to the times described in ANSI/ANS-N58.8-1984 (Reference 3).
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| June 2011 15.6-20 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The operator first diagnoses the excessive steam demand and closes the ADVs to prevent excessive cooldown. It is assumed that the ADV on the affected steam generator sticks open while the ADV on the unaffected steam generator closes. Diagnosis of a SGTR with Excess Steam Flow is facilitated by any or all of the following monitors and alarms:
| |
| * Rise in Condenser Off-Gas Monitor or alarm
| |
| * Rise in Steam Generator Blowdown Monitor or alarm
| |
| * Rise in Main Steam Line Monitor or alarm
| |
| * Rise in Main Steam Line N-16 Monitor or alarm
| |
| * Rise in activity in Steam Generator liquid sample
| |
| * Mismatch between feed flow and steam flow The major post-trip EOP operator actions considered are discussed below in detail.
| |
| A. Manual Reactor Trip Although the reactor trip by CPCs on margin to hot-leg saturation or on low pressure, or by RPS on low pressurizer pressure, is expected due to RCS depressurization during a SGTR event, an operator action to manually trip the reactor at an earlier time is assumed in order to bound the timing of automatic reactor trips, which are estimated to occur later in the transient. This is an adverse operation action, principally because the earlier trip results in the ADVs being opened sooner by the operators.
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| June 2011 15.6-21 Revision 16
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| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY B. Opening of ADVs In order to preclude a direct challenge to the MSSVs, the operators open the ADVs (on both steam generators) after the reactor trip, as instructed in the procedures, to relieve the pressure on the steam generators since, because of the LOP, the SBCS is not available. This is also an adverse action since an earlier opening of the ADVs results in increased release to the atmosphere, and thus, more adverse dose consequences. A bounding value of 2 minutes for the time after trip to manually open the ADVs is assumed based on operating experience and simulator scenarios.
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| C. Diagnose the Event and Stabilize the Plant The operators first diagnose the Excess Steam Demand (ESD) and take action to close the ADVs to prevent excessive cooldown.
| |
| This action is assumed to occur approximately seven minutes after the reactor trip and is consistent with expected operator action to ensure adequate RCS heat removal. The analysis assumes that a diagnosis of a SGTR with a continued ESD caused by the stuck-open ADV, will take approximately 15 minutes after indication of double ended guillotine break of a steam generator tube by any of the alarms and monitors listed above, after which operator actions follow guidance from the appropriate PVNGS procedure.
| |
| D. Functional Recovery Strategy At 15 minutes post-trip, the operators are assumed to override the AFAS on the affected steam generator, which has blocked flow to the affected steam generator based on the pressure difference between the two steam generators, and establish June 2011 15.6-22 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY dedicated flow from both AFW pumps to the affected steam generator until the steam generator level recovers above 40%
| |
| NR. This action is consistent with the procedural strategy in response to indications of a SGTR with an ESD (due to uncontrolled steaming to atmosphere from the affected steam generator). The total AFW flow specified in the EOP for a SGTR with an ESD is between 1360 and 1600 gpm. The lower value was used to delay covering the steam generator tubes thereby maximizing the radiological releases when AFW is in manual.
| |
| E. Post-Tube Covering Strategy After the affected steam generator level is above 40% NR, the operators are assumed to initiate a conservatively low AFW flow of 500 gpm to the unaffected steam generator. In accordance with the EOPs, the operator maintains the affected steam generator level between 40% and 60% NR, thus covering the U-tubes, for the remainder of the event by adjusting AFW flow as necessary. The operator is assumed to shift the control of heat removal to the unaffected steam generator when the affected steam generator U-tubes are covered.
| |
| F. Cooldown and Depressurize RCS to SDC Entry Conditions The cooldown and depressurization of the RCS is predominantly due to the stuck-open ADV. However, the analysis assumes operator actions to minimize the cooldown and depressurization to remain within the restriction of EOP guidelines.
| |
| Specifically, adequate SCM is maintained by utilization of HPSI pumps and pressurizer class back-up heaters.
| |
| June 2011 15.6-23 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY In addition to maintaining adequate subcooling, the operator is simultaneously responsible for assuring adequate RCS inventory is maintained. Specifically, the EOPs require the operator to retain specified levels in the pressurizer and the upper head before throttling back the HPSI flow. Accordingly, the pressurizer level and the SCM (See Figure 15.6.3-15) in the analysis are maintained above the level required by the EOPs.
| |
| Although SDC entry conditions may be reached in less than 8 hours, the event was simulated for 8 hours to maximize the dose consequences.
| |
| June 2011 15.6-24 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-1 SEQUENCE OF EVENTS FOR THE LIMITING SGTRLOP SINGLE FAILURE EVENT (3990 MWt RTP with RSG) (Page 1 of 2)
| |
| Time Event (sec) 0 SGTR occurs 43 Letdown control valve throttled to minimum 79 Backup pressurizer heaters energized 100 Manual reactor trip 100 Reactor trip breakers open 100 Turbine trip occurs 100.6 SCRAM CEAs begin falling 102 MSSVs open 104 LOP occurs 105 Maximum steam generator pressure 109 Steam generator level reaches AFAS setpoint in unaffected steam generator 110 Steam generator level reaches AFAS setpoint in affected steam generator 155 AFW initiated to unaffected steam generator 156 AFW initiated to affected steam generator 162 MSSVs close 220 Operator initiates plant cooldown by opening one ADV on each steam generator. The ADV on one steam generator (affected) instantaneously opens fully 245 Pressurizer pressure reaches SIAS setpoint 245 SI flow initiated with no delay 251 MSIS actuation, secondary pressure June 2011 15.6-25 Revision 16
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| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-1 SEQUENCE OF EVENTS FOR THE LIMITING SGTRLOP SINGLE FAILURE EVENT (3990 MWt RTP with RSG) (Page 2 of 2)
| |
| Time Event (sec) 268 AFAS 1 lockout on high 317 Voids begin to form in the upper head 520 Operator shuts ADV on the unaffected steam generator to prevent excessive cooldown 608 AFAS 2 reset on high steam generator level 847 Voids collapsed in the upper head 1000 Operator overrides the lockout and initiates dedicated AFW flow of 1360 gpm to affected steam generator 1900 Operator opens pressurizer head vent 2716 HPSI flow throttled to maintain SCM less than the limit 2779 With level in the affected steam generator above the top of U-tubes, the operator secures AFW flow to affected steam generator and initiates AFW to unaffected steam generator 12610 Class back-up heaters energized to maintain target harsh SCM criteria 26014 ADV opened in the unaffected steam generator in preparation of approaching SDC entry conditions 26260 SDC entry conditions reached in the affected loop 28800 Operator activates SDC system June 2011 15.6-26 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.3.2.3 Core and System Performance A. Mathematical Model The thermal-hydraulic response of the NSSS to the SGTRLOPSF was simulated using the CENTS computer code (described in UFSAR Section 15.0.3.1.3.2). The features incorporated in the analytical model include the following:
| |
| * secondary releases from both the MSSVs and ADVs
| |
| * early operator action for manual trip
| |
| * early operator action to open the ADVs
| |
| * a series of operator actions to cover the steam generator tubes
| |
| * time delays for operator functional recovery actions
| |
| * delay in reaching shutdown cooling (chosen to maximize 8-hour steam release)
| |
| B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a SGTRLOPSF are listed in Table 15.6.3-2. The initial conditions for several process variables were varied parametrically in order to determine the values or assumptions that would produce the most adverse radiological consequences.
| |
| The initial condition choices that would contribute to a higher calculated radiological release to atmosphere are discussed below.
| |
| * Maximum rated core power - Maximizes the initial heat content of the primary system and maximizes the energy removed from the secondary system. This leads to an increased heat up and June 2011 15.6-27 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY pressurization of the primary and secondary systems, which increases the secondary system releases.
| |
| * Maximum core inlet temperature - Maximizes the initial secondary system pressure, which increases the steaming through the MSSVs, and maximizes the amount of heat that must be removed during the 2- and 8-hour cooldown intervals by steam releases.
| |
| * Minimum RCS flow rate - Maximizes the temperature differential across the core, which maximizes the energy that must be removed by steaming through the steam generators.
| |
| This increases the activity releases through the MSSVs and ADVs.
| |
| * Maximum pressurizer pressure - Maximum initial pressurizer pressure increases the leak rate from the RCS to the affected steam generator, which increases the releases from the secondary system.
| |
| * Initial steam generator mass - Low initial steam generator water mass contributes to more adverse radiological consequences by allowing the specific activity in the steam generators to increase more rapidly because of the leak.
| |
| * Minimum scram worth at hot full power - Minimizes the rate of decreasing core power after the reactor trip and therefore increases the heat load to be removed by secondary system releases.
| |
| * Safety Injection - Maximizing the HPSI flow will result in higher RCS pressures and increased leakage to the affected steam generator. The SIAS setpoint was set high to provide an early delivery of HPSI flow. No delay time was applied to June 2011 15.6-28 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY this signal. In addition, two HPSI pumps were assumed to be available, thus maximizing the flow delivered to the RCS upon SIAS signal.
| |
| The SGTRLOPSF transient is not sensitive to the values of MTC and FTC, with regards to radiological consequences, as there are no changes in the fuel or moderator temperatures prior to reactor trip. The most negative MTC and the least FTC were used.
| |
| The major assumptions regarding systems operation during the event are summarized below.
| |
| * After reactor trip, the main feedwater flow is ramped down to zero in one second.
| |
| * Subsequent beneficial operator actions are delayed by times that are also consistent with the ANSI Standard (Reference 3).
| |
| * MSSV setpoints are set at their lowest lift pressure in order to make the MSSVs open and release steam to the atmosphere early.
| |
| * After the LOP and RCPs coast down, charging pumps are de-energized, letdown is isolated and heaters and sprays are lost.
| |
| * The LOP also causes loss of condenser vacuum, loss of steam bypass control valves, and loss of forced flow (RCP coastdown).
| |
| * When the ADVs are opened on both steam generators, one of the ADVs is assumed to fail wide open. Most of the cooldown comes about by steaming from the affected steam generator, June 2011 15.6-29 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY with the unaffected steam generator being used to control cooldown to SDC entry conditions.
| |
| * The AFW system is activated at 20% level wide range and shuts off at 30% level wide range prior to operator action.
| |
| * Two AFW pumps are assumed to be available to supply feedwater to either steam generator. No credit is taken for the third 1E-powered AFW train.
| |
| * Two HPSI pumps are assumed to be available subsequent to the generation of a SIAS. The SIAS comes from a low pressurizer pressure signal and has no time delay, so that it results in maximum HPSI flow.
| |
| * The pressurizer 250 kW class 1E backup heaters are cycled as necessary to maintain adequate SCM.
| |
| * The pressurizer head vent system is manually controlled by the operators, as necessary, to depressurize the RCS during the event.
| |
| * The SBCS is assumed to be in the manual mode initially.
| |
| The event is simulated so that the SDC entry conditions are reached at about 8 hours after initiation of the double-ended rupture of the steam generator U-tube.
| |
| June 2011 15.6-30 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-2 PARAMETERS USED FOR THE LIMITING SGTRLOP SINGLE FAILURE EVENT (3990 MWt RTP with RSG Case)
| |
| PARAMETER Value Initial core power (% of RTP) 102 Initial core inlet temp (°F) 568 Initial pressurizer pressure (psia) 2325 Initial RCS flow (% of design) 95 Initial pressurizer level (%) 53 Initial steam generator level (% WR) 41
| |
| -4 MTC (/°F) -4.0x10 FTC Least negative Kinetics Minimum CEA worth at trip - WRSO (%) -8.0 2
| |
| Fuel rod gap conductance (Btu/hr-ft -°F) 518 Plugged steam generator tubes 0 SGTR break location at the tube sheet Single failure Stuck-Open ADV (at full open position)
| |
| LOP Yes C. Results Table 15.6.3-1 presents a typical sequence of events for the SGTRLOPSF event. Typical transient response of key NSSS parameters as a function of time is presented in Figures 15.6.3-1 to 15.6.3-15 for this limiting fault event.
| |
| The calculated transient minimum DNBR is greater than the DNBR SAFDL value of 1.34. Therefore, fuel cladding damage is not predicted for the limiting fault SGTRLOPSF event.
| |
| June 2011 15.6-31 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.3.2.4 RCS Pressure Boundary Barrier Performance A. Mathematical Model The computer codes that were employed to evaluate fission product barrier performance (other than fuel cladding) for this limiting fault event are identical to those described in UFSAR Section 15.6.3.2.3.
| |
| B. Input Parameters and Initial Conditions The input parameters and initial condition relevant to barrier performance for this limiting fault event are the same as those presented in Table 15.6.3-2 of UFSAR Section 15.6.3.2.3.
| |
| C. Results Due to depressurization of the primary system, the RCS pressures during the event do not exceed the initial pressure which is less than 110% (2,750 psia) of RCS system design pressure (2,500 psia). The secondary pressures reach maxima around 1300 psia just after the turbine trip and opening of the MSSVs. Thus, the maximum pressure is less than 110% (1,397 psia) of secondary design pressure (1,270 psia).
| |
| Steam generator overfill does not occur because of the single failure of the stuck open ADV on the affected steam generator, and manual control of the ADV on the unaffected steam generator.
| |
| 15.6.3.2.5 Containment Performance and Radiological Consequences A SGTRLOPSF is classified as a limiting fault. Offsite radiological dose consequences are limited to 10 CFR Part 100 guideline values. Additionally, radiation exposure for control June 2011 15.6-32 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY room personnel are subject to the limits specified in General Design Criterion (GDC) 19 of 10 CFR 50 Appendix A.
| |
| Control room radiological assessments for bounding unfiltered in-leakage are presented in UFSAR Section 6.4.7. The evaluation of offsite radiological dose consequences associated with the SGTRLOPSF event is discussed below.
| |
| Peak containment pressure is not calculated for this event and would be bounded by the Design Basis Accidents, Loss-of-Coolant Accident and the Main Steam Line Break events (see UFSAR Section 6.2). Since this event results in the depressurization of the primary system, the pressurizer safety valves do not lift. The impact of the releases by the pressurizer vent system to the containment is small enough that the impact of the radiological release from the containment to the atmosphere has negligible impact on the EAB, LPZ or Control Room doses.
| |
| A. Mathematical Model The mathematical model employed in the evaluation of the radiological consequences resulting from the SGTRLOPSF is based on the general modeling in UFSAR Section 15.0.4 and is described below.
| |
| The SGTRLOPSF predicts that steam or liquid will be released from the RCS or main steam system and radioactive material will be present in these discharges. As a result, the SGTRLOPSF is anticipated to result in radiological dose consequences for the off-site general public. Appendix 15B describes a generic activity release model for assessing the radiological consequences of postulated accidents.
| |
| To analyze the radiological consequences of the SGTRLOPSF, the steam releases to the environment are extracted from the CENTS June 2011 15.6-33 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY simulations of the event. Estimated releases are utilized in the radiological dose analyses, for the purpose of determining thyroid doses at the EAB and at the outer boundary of the LPZ.
| |
| The evaluation of the radiological consequences of the SGTRLOPSF assumes a double-ended, guillotine break at the hot side tube sheet of a steam generator U-tube while the reactor is operating at full power. Occurrence of the accident leads to an increase in contamination of the secondary system due to reactor coolant leakage through the ruptured tube. An early reactor trip is assumed, which causes the turbine to trip and leads to a LOP three seconds after the turbine trip. Prior to LOP, the activity is released to the environment through the condenser. Following the closure of turbine admission valves, the steam generator pressure increases rapidly, resulting in opening of MSSVs and steam discharge as well as activity release through the MSSVs. Venting from the affected steam generator by the MSSVs continues until the closure of MSSVs when the secondary system pressure drops below the MSSV blowdown setpoint. Two minutes after the trip, the operator partially opens one ADV on each steam generator to minimize further challenges to the MSSVs and to stabilize RCS temperature. At this point, one of the ADVs (on affected steam generator) is assumed to open fully and remain open for the remainder of the transient.
| |
| For most of the event, the heat extraction occurs through the affected steam generator while the unaffected steam generator releases very little steam. The open ADV dumps the inventory of the affected steam generator and the leakage from the primary loop flashes and is released directly. When operator action to divert all AFW to the affected steam generator is June 2011 15.6-34 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY taken, the inventory begins to increase until the U-tubes are eventually covered by liquid. At this point the release rate drops dramatically as the iodine is scrubbed by the water.
| |
| Boil-off of the inventory in the affected steam generator, which is maintained by the operators, and controlled cooling by feeding and bleeding of the unaffected steam generator provides the cooling for the event.
| |
| The tube leak in the affected steam generator is minimized when primary-to-secondary pressure differential is reduced with a target of keeping it within 50 psid, in accordance with EOP guidance. For this event, the SDC entry conditions are reached in the affected loop prior to 8 hours. However, the transient is simulated for 8 hours in order to maximize dose consequences.
| |
| The analysis of the radiological consequences of a SGTRLOPSF considers the most severe release of secondary activity as well as primary system activity leaked from the tube break. The inventory of iodine and noble gas fission product activity available for release to the environment is a function of primary-to-secondary coolant leakage rate, the iodine spiking factor, the initial condition of the fuel in the core and the mass of steam discharged to the environment. Conservative assumptions are made for all these parameters.
| |
| B. Input Parameters and Initial Conditions The assumptions and parameters used to determine the activity releases and offsite doses for a SGTRLOPSF are discussed below.
| |
| : 1. Accident doses are calculated for two different iodine spiking assumptions: (a) an event-Generated Iodine Spike June 2011 15.6-35 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY (GIS) coincident with the initiation of the event and (b) a Pre-accident Iodine Spike (PIS).
| |
| : 2. Technical Specification limits for the initial primary system (1.0 µCi/gm) and secondary system activity (0.1
| |
| µCi/gm) concentrations are assumed. Transient primary system specific activity is calculated using the dilution from HPSI flow.
| |
| : 3. A spiking factor of 500 is employed for the GIS at the time of event initiation.
| |
| : 4. A CVCS purification efficiency of 100% is assumed based on the bounding purification flow rate of 150 gpm.
| |
| -7 -1
| |
| : 5. The I-131 decay constant is 9.97 x 10 sec .
| |
| : 6. For the PIS condition, a PIS factor of 60 for the primary system activity concentration is employed.
| |
| : 7. Total allowable primary-to-secondary leakage of 1 gpm is conservatively assumed to be in the unaffected steam generator for the duration of the transient, instead of 0.5 gpm per steam generator.
| |
| : 8. In the unaffected steam generator, the primary-to-secondary is released to the atmosphere with the Decontamination Factor (DF) of 100.
| |
| : 9. In the affected steam generator, the portion of the leaking primary fluid that flashes to steam upon entering to the steam generators is assumed to be released to the atmosphere with a DF of 1.0, while unflashed portion is assumed to mix with the steam generator inventory and released to the atmosphere with a DF of 100.
| |
| June 2011 15.6-36 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The portion of the leaking primary fluid that flashes to steam is calculated based on the enthalpy of the leak (See Figure 15.6.3-12). During periods of U-tube uncovery, the flashing fraction is set to 1.0.
| |
| : 10. The atmosphere dispersion factors employed in the
| |
| -4 3 -5 analyses are 2.3 x 10 sec/m for the EAB and 6.4 x 10 3
| |
| sec/m for the LPZ.
| |
| : 11. The Dose Conversion Factors (DCFs) for dose equivalent Iodine are derived from ICRP-30 (Reference 4).
| |
| : 12. The primary-to-secondary leakage through the tube rupture (See Figure 15.6.3-11), secondary mass inventory (See Figure 15.6.3-13), and secondary system releases from ADVs (See Figure 15.6.3-14) and MSSVs are calculated from the transient simulation of the event.
| |
| : 13. Prior to LOP, the activity is released to the environment through the condenser with a DF of 100.
| |
| C. Results The reported values for the 2-hour EAB and the 8-hour LPZ thyroid inhalation doses for the PIS and the GIS cases are presented in Table 15.6.3-3. The calculated EAB and LPZ doses are within the values of 10 CFR 100. These results bound PVNGS Units operating at a RTP of 3990 MWt or less.
| |
| June 2011 15.6-37 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-3 RADIOLOGICAL CONSEQUENCES FOR THE LIMITING SGTRLOPSF EVENT Event Case Evaluation Period & Dose Location (REM) 0-2 hrs at EAB 182 GIS 0-8 hrs at LPZ 125 0-2 hrs at EAB 294 PIS 0-8 hrs at LPZ 91 15.6.3.2.6 Conclusions The dynamic behavior of important NSSS parameters during a typical event was presented in Figures 15.6.3-1 through 15.6.3-15. The radiological releases calculated for the limiting SGTR event (SGTR with a loss of offsite power and a fully stuck open ADV) were demonstrated to be within the 10 CFR 100 guidelines.
| |
| The RCS and secondary system pressures were shown to be below 110% of the design pressure limits, thus assuring the integrity of these systems.
| |
| Additionally, it was demonstrated that there would be no violation of the fuel thermal limits, since the minimum DNBR remains above the DNBR SAFDL value throughout the duration of the event.
| |
| 15.6.4 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR)
| |
| Not applicable.
| |
| June 2011 15.6-38 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.5 LOSS-OF-COOLANT ACCIDENTS Refer to subsection 6.3.3 and paragraphs 6.2.1.5, 6.3.3.2.1, and 6.3.3.3 for loss-of-coolant accident (LOCA) performance evaluations of reactivity control, reactor heat removal, primary system integrity, and secondary system integrity. Also refer to the above sections for LOCA analysis of effects and consequences as they pertain to releases from the primary, secondary, and safety injection systems.
| |
| The auxiliary feedwater system is described in subsection 10.4.9.
| |
| A SIAS will actuate control room and fuel building essential ventilation systems. See sections 6.4 and 9.4 for details and sections 18.II.E.1.1 and 18.II.E.1.2 for reevaluation of the auxiliary feedwater system with respect to TMI lessons learned.
| |
| A CIAS or CPIAS will terminate the containment power access purge, as described in section 9.4.
| |
| 15.6.5.1 Identification of Event and Causes - Small Break LOCA Refer to subsection 6.3.3.
| |
| 15.6.5.2 Sequence of Events and Systems Operation - Small Break LOCA Seven breaks were analyzed to characterize the radiological consequences. The spectrum of seven SBLOCA break sizes included 2
| |
| span of the limiting break size for ECCS (0.05 ft ) with most limiting peak cladding temperature, to break size with no core 2 or uncovery (0.01 ft less), to a 1 inch diameter break 2
| |
| (0.005 ft ). These analyses are specifically done for radiological assessment of SBLOCA and include the effect of ZIRLO fuel, which is bounding for both ZIRLO and Zircaloy-4 clad fuel.
| |
| June 2011 15.6-39 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.5.2.1 Evaluation Model The C-E computer codes CEFLASH - 4AS for primary coolant system thermal hydraulics, CONTRANS2 for containment pressure evaluation, and Bechtel computer code LOCADOSE for dose assessment are used for this Evaluation Model.
| |
| 15.6.5.2.2 Release Pathways The release of radioisotopes is postulated through the following pathways.
| |
| * Through containment leakage which results from release of primary coolant to the containment from the ruptured pipe. This pathway includes:
| |
| * unfiltered discharge through the power access purge lines until such time as the valves are closed due to generation of CIAS or CPIAS,
| |
| * Leakage through the containment structure at Technical Specification leak rates.
| |
| June 2011 15.6-40 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY
| |
| * Release of contaminated steam from the secondary system.
| |
| This pathway includes:
| |
| * Release of steam contaminated by leakage of primary coolant to the secondary side. The primary-to-secondary leakage is assumed to be at a rate of 1 gpm, and
| |
| * Release from the secondary system inventory
| |
| [feedwater] at Technical Specification concentration of 0.1 uCi/gm Dose Equivalent I-131.
| |
| * Release of contaminated sump inventory from leakage of ESF components outside containment during recirculation phase.
| |
| 15.6.5.2.3 Description of Results The results of these analyses show that all SBLOCA transients achieved containment isolation and containment spray actuation before core uncovery, that is before the possibility of large radioactive release resulting from fuel cladding damage that core 2
| |
| uncovery may cause. The 0.03 ft break was determined to be the smallest size break that would exhibit cladding rupture behavior.
| |
| Table 15.6.5-1 sheet one provides a summary of calculated times to core uncovery, CIAS, CSAS and core recovery. This information is used in the dose assessment evaluation to determine the magnitude of source term and duration of release from containment power access purge prior to containment isolation.
| |
| The following are assumptions used to evaluate the RCS and containment behavior during SBLOCA:
| |
| June 2011 15.6-41 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY
| |
| * ALL breaks are in RCS cold leg; potential hot leg breaks are less limiting.
| |
| * Loss of offsite power at time of reactor trip.
| |
| * Isolation of Main Steam and Main feed at time of reactor trip.
| |
| * RCP trip on loss of offsite power.
| |
| * SIAS on low pressurizer pressure.
| |
| * Single failure: Loss of one diesel
| |
| * Auxiliary feedwater actuation to maintain RCS heat removal.
| |
| * Containment power purge is operating at start of this event and these valves would be isolated at initiation of CIAS (for hydraulic analysis).
| |
| * SI spillage from broken cold leg is assumed to have insignificant effect on containment responses.
| |
| * No operator action has been assumed.
| |
| Radiological consequences associated with a spectrum of small break LOCAs have been evaluated using computer code LOCADOSE in accordance with the guidelines of Regulatory Guide 1.4, Regulatory Guide 1.77, and SRP Section 15.6.5. This evaluation is partitioned in to two groups: break sizes that would result in core uncovery and breaks that would not result in core uncovery. The limiting break's thyroid dose due to inhalation and whole-body gamma dose due to immersion are presented in 2
| |
| table 15.6.5-1. Two break sizes were evaluated 0.03 ft and 2
| |
| 0.005 ft Fuel-clad failure was postulated for the first case, 2
| |
| fuel-clad rupture was not predicted for the 0.005 ft break.
| |
| Doses are calculated using ICRP-30 iodine inhalation dose conversion factors and regulatory guide 1.109 dose conversion factors for all other path ways and isotopes.
| |
| June 2011 15.6-42 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY The following are a list of assumptions used to evaluate the consequences of the spectrum of SBLOCA,
| |
| * Core power was set at an elevated level of 102% of uprated licensed power (3990 MWt).
| |
| * The core activity level was based on the "bounding" source term using TID-14844 methodology.
| |
| * The initial primary system activity level was based on the maximum activity in the reactor coolant due to continuous full power operation with 1% failed fuel, with iodine at a pre-existing iodine spike level of 60 uCi/gm DEQ I-131.
| |
| * Fuel-clad failure was assumed for the case where core 2
| |
| uncovering was observed [0.03 ft break].
| |
| 2
| |
| * For the case where fuel failure was postulated [0.03 ft break]; all gaseous constituents in the fuel-clad gap were released into the primary coolant. The amount of activity accumulated in the fuel-clad gap was assumed to be 10% of the core iodine and noble gases [Reg. Guide 1.77].
| |
| * The offsite [EAB & LPZ] and the control room atmospheric dispersion parameters [X/Qs] were based on the updated site specific meteorological data [refer to section 2.3-31],
| |
| * The containment power access purge closed within 8 seconds (8 seconds includes: instrument response time, ESF loop delay time and valve closure time) upon receipt of a containment isolation actuation signal (CIAS) or CPIAS. Time to generate an isolation signal is conservatively set to the elapsed time from the initiating event to time for containment pressure to reach 5 PSIG.
| |
| June 2011 15.6-43 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY
| |
| * Control room essential filtration system was activated within 50 seconds upon receipt of a SIAS.
| |
| * Spray water was delivered to the containment atmosphere in 92 seconds upon receipt of a containment spray actuation signal (CSAS). The time for the containment pressure to reach a CSAS setpoint ranged from 282 to 10,000 seconds depending upon the size of the break.
| |
| * Containment leakage rate was set at value of 0.1 vol% per day for the first 24 hours, and at half of that rate thereafter per regulatory guide 1.4.
| |
| * Primary-to-secondary leakage rate was set at value of 1.0 gpm (total). Duration of 3 hours was assumed for the break size of 2 2 0.03 ft . For 0.005 ft break a value of 8 hours was assumed based on natural circulation cooldown.
| |
| * ESF component leakage rate during recirculation phase was conservatively set at a constant rate of 3,000 ml/hr. This pathway only applies to the fuel-clad failure case because the smaller break does not result in sump recirculation.
| |
| 15.6.5.3 Analysis of Effects and Consequences For the break for which fuel cladding failure was postulated 2
| |
| (0.03 ft ), the main dose contributor was the primary fluid that leaked to the secondary side, following the release of 100 percent of the gas gap activity into the primary.
| |
| 2 For a 0.005 ft break, fuel-clad rupture was not predicted.
| |
| Although the source term was based on design RCS activity level, with pre-existing iodine spike, a relatively higher thyroid dose for control room was noted which was due to extended release June 2011 15.6-44 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY through the mini-purge and a fairly long isolation time for the control room.
| |
| For all cases analyzed (refer to table 15.6.5-1), the offsite doses remained bounded by the large break LOCA doses that are presented in table 15.6.5-2 and the control room doses were below GDC 19 exposure limits.
| |
| June 2011 15.6-45 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA 1
| |
| Summary of Calculated Time vs. Break Size React CSAS Break size CIAS on Core Core or on 10 2 5 PSIG uncovery recovery (ft ) Trip PSIG (sec) (sec) (sec)
| |
| (sec) (sec) 0.07 75 155 282 610 2480 0.05 119 217 442 772 2970 0.03 256 360 1190 1400 5000 0.01 2596 970 6770 No uncovery No uncovery 0.008 3826 1200 9370 No uncovery No uncovery 0.006 6242 1565 >10000 No uncovery No uncovery 0.005 8945 1860 >10000 No uncovery No uncovery 1
| |
| Duration provided is approximate from the outputs of CEFLASH and CONTRANS2.
| |
| June 2011 15.6-46 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value A. Source Term Data 1 Core Activity (curies):
| |
| I-131 1.02E+08 I-132 1.55E+08 I-133 2.29E+08 I-134 2.68E+08 I-135 2.08E+08 Kr-83m 1.69E+07 Kr-85 1.79E+06 Kr-85m 5.28E+07 Kr-87 8.77E+07 Kr-88 1.30E+08 Kr-89 1.69E+08 Xe-131m 1.06E+06 Xe-133m 5.63E+06 Xe-133 2.29E+08 Xe-135m 7.39E+07 Xe-135 2.18E+08 Xe-137 2.17E+08 Xe-138 2.02E+08 2 RCS specific activity concentration prior to event: uCi/gm I-131 3.0 I-132 0.83 I-133 4.4 I-134 0.52 I-135 2.5 Kr-83m 0.013 Kr-85 6.1 Kr-85m 1.3 Kr-87 1.0 Kr-88 2.8 Kr-89 0.076 Xe-131m 5.9 Xe-133m 0.34 Xe-133 360 Xe-135m 0.74 Xe-135 7.7 Xe-137 0.17 Xe-138 0.63 June 2011 15.6-47 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value 3 Fuel-Clad failure 2
| |
| 0.03 ft break Yes 2
| |
| 0.005 ft break No 4 Activity accumulated in the fuel-gap, in percent of core:
| |
| Iodine 10%
| |
| Noble Gases 10%
| |
| 5 Iodine composition:
| |
| Elemental 91%
| |
| Particulate 5%
| |
| Organic 4%
| |
| 6 Percent of the accumulated fission prod- 100%
| |
| ucts in the fuel-gap that would be released into the primary coolant due to event-induced fuel-clad failure B. Containment Power Access Purge (Mini-Purge) Data 7 Source Terms Iodine [assuming pre-existing iodine 60 uCi/gm spike, Dose Equivalent I-131]
| |
| Noble gases RCS normal 8 Purge valve type Butterfly Purge valve size, inch 8 2
| |
| Number of valves (0.03 ft break) 2 9 Effective purge flow rates, cfm Calculated by LOCADOSE 2
| |
| 0.03 ft break 0.005 ft2 break (Maximum) 2,200 10 Total containment power access purge isolation time [duration of release to environment in sec]
| |
| Break sizes 2
| |
| : 1. 0.03 ft 265 2
| |
| : 2. 0.005 ft 1869 June 2011 15.6-48 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value C. Containment Leakage Data 11 Source Terms 0.03 ft2 breaks RCSnormal+
| |
| Gap activity 0.005 ft2 break RCSnormal with iodine spike 12 Percent of the discharged primary cool-ant activity which is released to the con-tainment atmosphere [airborne]:
| |
| Iodine 25%
| |
| Noble gases 100%
| |
| 3 13 Containment net free volume, ft 2.62E+6 14 Containment leak rate, vol.%/day 0-24 hr 0.1
| |
| > 24 hr 0.05 3
| |
| 15 Containment region volumes, ft :
| |
| Main spray region 2.27E+6 Auxiliary spray region 2.00E+5 Unsprayed region 1.50E+5
| |
| -1 16 Transfer rate between sprayed and 3.3 hr unsprayed regions, in terms of unsprayed volume change per hour [8,250 cfm]
| |
| 17 Air transfer rates between the contain-ment regions, cfm:
| |
| Main sprayed and unsprayed regions 7,582 auxiliary sprayed and unsprayed regions 668 18 Total instrumentation and pump 33 response time (w/LOP) for containment spray pump (DG start, ESFAS, sequencer and pump response time) sec-onds June 2011 15.6-49 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value 19 Iodine removal by spray, during injec-tion phase:
| |
| Main sprayed region: Coefficients
| |
| -1 Elemental 19.6 hr Organic 0
| |
| -1 Particulate 0.32 hr Auxiliary sprayed region:
| |
| -1 Elemental 6.05 hr Organic 0
| |
| -1 Particulate 0.09 hr Spray elemental-iodine decontamination DF Factor 6.51 20 Elemental iodine removal by plate-out
| |
| [wall deposition]: Coefficients
| |
| -1 Main sprayed region 2.14 hr
| |
| -1 Auxiliary sprayed region 14.4 hr
| |
| -1 unsprayed region 14.4 hr Elemental-iodine decontamination fac- DF Tor 100 21 Duration of containment leakage 30 days D. Primary-to-Secondary Leakage Data 22 Primary-to-secondary leak rate (steam 1 gpm generator tube leakage), total 23 Source Terms 2
| |
| 0.03 ft breaks RCSnormal+
| |
| Gap activity 2
| |
| 0.005 ft break RCSnormal with iodine spike 24 Duration of leakage 2
| |
| 0.03 ft breaks 3 hrs 2
| |
| 0.005 ft break 8 hrs 25 Steam Generator Partition Factors Iodine 0.01 noble gases 1 June 2011 15.6-50 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value E. Secondary Steam Release Data 26 Source Terms 0.10 uCi/gm Secondary coolant activity concentra- DEQ I-131 tion prior to onset of the event 27 Total mass release through the main 374,400 lbm steam safety valves (MSSVs) and (total second-through the atmospheric dump valves ary volume)
| |
| (ADVs)
| |
| Steam generator iodine partition factor 1.0 F. ESF Recirculation Data 3
| |
| 28 Sump volume (ft ),
| |
| 3 Reduced RWT volume of 400,000 gal 53,476 ft 3
| |
| RCS volume, including Pzr & CVCS 9,177 ft 3
| |
| Safety injection tanks [SITs] 4 x 1,750 ft 29 Sump activity of iodine, as a percent 50%
| |
| post-accident reactor coolant activity 30 Recirculation start time 20 minutes, post accident 31 Credit assumed for radioactive decay of yes iodine prior to recirculation 32 Total ESF component leakage rate [two 3,000 ml/hr trains]
| |
| 33 Percent of the iodine in the leaked water 10%
| |
| which is assumed to become volatile
| |
| [flashing fraction]
| |
| 34 Duration of ESF leakage 30 days (b)
| |
| G. Control Room Data - refer to section 6.4.7 H. Transport Data 3
| |
| 35 EAB X/Q, 0-2 hr, sec/m 2.3E-4 3
| |
| LPZ X/Q, sec/m :
| |
| 0-8 hr 6.4E-5 8-24 hr 4.8E-5 24-96 hr 2.6E-5 96-720 hr 1.1E-5 June 2011 15.6-51 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of input parameters for radiological consequences Parameter Value 3
| |
| 36 Offsite Breathing Rates, m /sec:
| |
| 0-8 hr 3.47E-4 8-24 hr 1.75E-4
| |
| > 24 hr 2.32E-4 37 Credit for depletion of the effluent Not Assumed plume of radioactive iodine due to depo-sition on the ground 38 Credit for radiological decay in transit Not Assumed I. Dose Calculation Data 39 Dose Conversion Factors (DCFs): NRC-ICRP-Inhalation Thyroid DCFs, rem/Ci 30 I-131 1.08E+6 I-132 6.44E+3 I-133 1.80E+5 I-134 1.07E+3 I-135 3.13E+4 40 Immersion [Beta Skin & Whole-Body] Reg. Guide 1.109 DCFs June 2011 15.6-52 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-1 (contd)
| |
| RADIOLOGICAL CONSEQUENCES OF A SMALL BREAK LOCA Summary of Input Parameters for Radiological Consequences EAB (rem) (0-2hr) LPZ (rem)(0-30day) a Location Control room (rem)(0-30day)
| |
| Whole Body Whole Body Whole (b)
| |
| Dose Contributor Thyroid Thyroid Beta skin Thyroid Body 2
| |
| Break size 0.03 ft. Limiting break with fuel rupture Containment Air Lkg:
| |
| Mini-Purge 4.30E-3 8.89E+0 1.20E-3 2.47E+0 7.72E-4 1.61E-2 3.78E+0 Containment Bldg. 2.53E-1 6.76E+0 1.90E-1 9.55E+0 1.27E-1 2.62E+0 1.10E+0 Steam Release:
| |
| Primary-Sec. Lkg 3.84E+0 1.60E+1 1.24E+0 5.75E+0 7.87E-1 1.52E+1 6.11E-1 Secondary Steam 3.41E-4 1.46E+0 9.48E-5 4.07E-1 1.99E-5 1.47E-4 1.74E+0 ESF Comp. Lkg 7.05E-4 1.58E-1 1.22E-03 1.03E+0 4.49E-5 8.55E-4 1.03E-1 Total 4.10E+0 3.33E+1 1.43E+0 1.92E+1 9.15E-1 1.78E+1 7.33E+0 2
| |
| Break size 0.005 ft. Limiting event without fuel rupture Containment Air Lkg:
| |
| Mini-Purge 1.18E-2 1.35E+1 3.28E-3 3.76E+0 2.93E-3 6.21E-2 1.42E+1 Containment Bldg. 2.41E-5 2.07E-2 1.38E-4 1.21E-1 8.32E-5 2.01E-3 1.30E-2 Steam Release:
| |
| Primary-Sec. Lkg 5.01E-4 2.08E-2 3.47E-4 1.64E-2 3.73E-4 8.42E-3 7.81E-3 Secondary Steam 3.41E-4 1.46E+0 9.48E-5 4.07E-1 1.99E-5 1.47E-4 1.74E+0 Total 1.27E-2 1.50E+1 3.86E-3 4.31E+0 3.41E-3 7.27E-2 1.59E+1
| |
| : a. Whole body doses do not included contribution from direct/ scatter shine (containment, outside cloud and piping/filter dose). SBLOCA direct and scatter doses are bounded by Large LOCA see table 15.6.5-2.
| |
| : b. The bounding leakage and thyroid dose is given in section 6.4.7.
| |
| June 2011 15.6-53 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.6.5.4 Identification of Event and Causes - Large Break LOCA Refer to subsection 6.3.3.
| |
| 15.6.5.5 Sequence of Events and Systems Operation - Large Break LOCA Dose Calculation Containment power access purge through an 8-inch penetration will be terminated within 12 seconds after generation of CIAS CPIAS as described in subsection 6.5.3.1.
| |
| A SIAS will initiate a switch to the filtered recirculation and filtered makeup mode of control room ventilation as discussed in section 6.4. A SIAS will initiate filtered ventilation of the lower region (below 100-foot elevation) of the auxiliary building as discussed in section 9.4. Since recirculation loop equipment and piping for safety injection and containment sprays in the auxiliary building is located below the 100-foot elevation, leakage from active recirculation equipment is filtered prior to release to the environment.
| |
| For the limiting large break LOCA, the reactor trip will result in a turbine trip, and a subsequent loss of offsite power will result in the loss of main feedwater flow. As result of the loss of feedwater and due to fast depressurization of RCS as result of the initiating event, high containment pressure would generate a CIAS and then MSIS actuation. At this time the primary coolant and containment environment would be at lower pressure than the secondary side of the steam generators and the secondary would become a heat source for the primary system.
| |
| The steam generators would then quickly depressurize due to containment spray system and possible use of ADVs. During this time the leakage would change direction from the primary coolant June 2011 15.6-54 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY to the secondary side of the steam generators. This would result in a leakage path from containment to the environment.
| |
| 15.6.5.6 Analysis of Effects and Consequences - Large Break LOCA Dose Calculation It is assumed that there is a preexisting RCS iodine spike of 60 µCi/cc dose equivalent I-131 since there will not be any fuel cladding rupture within the 20 seconds after initiation of the large break LOCA. This activity is instantaneously mixed with the containment atmosphere and available for release via the power access purge. The containment airborne radioactivity inventory will be affected by four factors: leakage, radioactive decay, plateout and sprays. No credit has been taken for spray removal of organic iodine. Refer to section 6.5 for a discussion of spray effectiveness. It is assumed that the containment leaks at the maximum rates allowed by the Technical Specifications, i.e., 0.1 vol %/d for the first 24 hours and half of that rate thereafter. This leakage, when combined with initial releases, will result in potential doses offsite and in the control room. Contribution from the containment leakage through the steam generators is evaluated.
| |
| The offsite and control room doses are calculated with a single failure of a GDC 57 valve or a stuck open ADV. The leakage from containment environment through the primary system to the secondary system is conservatively assumed to be containment atmosphere at post large break LOCA conditions. The flowrate is calculated as that produced by leaking through a steam generator tube fault equivalent in size to that which would allow the 1 gpm RCS liquid at normal operating plant conditions. This rate is used for the first 24 hours. The June 2011 15.6-55 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY leakage rate is assumed to be half that for the duration of the analysis. The analysis does not take any credit for iodine partition factors or operator action to flood the steam generator. The doses are listed in table 15.6.5-2.
| |
| Additionally, there will be exposure offsite and in the control room from the filtered release of recirculation leakage. The calculated leakage is based on the containment sump inventory as per table 15.6.5-2. The doses from recirculation releases are listed in table 15.6.5-2. The total combined doses to an individual offsite and to control room operators following a postulated large break LOCA are also presented in table 15.6.5-2.
| |
| The release of radioisotopes, due to LOCA, is postulated through the following pathways:
| |
| 1- The containment leakage which results from the release of primary coolant to the containment from the postulated break. This pathway includes: (1) unfiltered discharge through the power access [mini-purge] intake and exhaust lines until such time the valves are closed due to generation of a CIAS, (2) leakage through the containment structure at Technical Specification leak rates, and (3) release of containment air through depressurized secondary system.
| |
| 2- release of contaminated sump inventory due to leakage from ESF systems outside containment during the recirculation phase. This pathway includes: (1) ESF component leakage outside containment, and (2) back-leakage of recirculating sump water to refueling water storage tank during long term cooling, post RAS due to check valve CH-V-305 and CH-V-306 leakage or any other June 2011 15.6-56 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY SI system leakage to the refueling water storage tank
| |
| [IN 91-56].
| |
| In addition to the above contributors, control room doses are evaluated for radiological exposure due to direct dose from containment, ESF piping, shine from outside cloud and dose due to accumulation iodine on essential control room HVAC filtration filters. Table 15.6.5-2 provides detailed information on key parameters used to evaluate the consequences of LOCA and a summary of integrated doses at different locations.
| |
| 15.6.5.7 Conclusions Based on use of very conservative assumptions regarding spray effectiveness, containment ventilation, and leakage, as well as conservative fuel failure models, the offsite doses presented in table 15.6.5-2 for LOCA are substantially below 10CFR100 limits.
| |
| June 2011 15.6-57 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 Large break LOCA radiological analysis parameters and results Parameter Value (3954 MWt) Value (4070 MWt)
| |
| Source Term Data 1 Core Activity (curies) Ci Ci I-131 9.92E+07 1.02E+08 I-132 1.51E+08 1.55E+08 I-133 2.22E+08 2.29E+08 I-134 2.60E+08 2.68E+08 I-135 2.02E+08 2.08E+08 Kr-83m 1.64E+07 1.69E+07 Kr-85 1.37E+06 1.79E+06 Kr-85m 5.14E+07 5.28E+07 Kr-87 8.50E+07 8.77E+07 Kr-88 1.27E+08 1.30E+08 Kr-89 1.64E+08 1.69E+08 Xe-131m 1.03E+06 1.06E+06 Xe-133m 5.46E+06 5.63E+06 Xe-133 2.22E+08 2.29E+08 Xe-135m 7.20E+07 7.39E+07 Xe-135 2.12E+08 2.18E+08 Xe-137 2.10E+08 2.17E+08 Xe-138 1.97E+08 2.02E+08 2 RCS specific activity concentration prior to event: uCi/gm uCi/gm I-131 60 DEQ I-131 60 DEQ I-131 I-132 --- ---
| |
| I-133 --- ---
| |
| I-134 --- ---
| |
| I-135 --- ---
| |
| Kr-83m 0.013 0.013 Kr-85 6.1 6.1 Kr-85m 1.3 1.3 Kr-87 1.0 1.0 Kr-88 2.8 2.8 Kr-89 0.076 0.076 Xe-131m 5.9 5.9
| |
| \ Xe-133m 0.34 0.34 Xe-133 360 360 Xe-135m 0.74 0.74 Xe-135 7.7 7.7 Xe-137 0.17 0.17 Xe-138 0.63 0.63 June 2011 15.6-58 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 (contd)
| |
| Large break LOCA radiological analysis parameters and results Parameter Value 3954(MWt) Value (4070 MWt) 3 Primary coolant weight 571,776lbm 606,083 lbm 3 3 Primary coolant specific volume @ 70F 0.01605 ft /lbm 0.01605 ft /lbm 4 Iodine composition:
| |
| Elemental, Organic, Particulate 91%, 4%, 5% 91%, 4%, 5%
| |
| Containment Data 3 3 5 Containment Net Free Volume 2.62E+6 ft 2.62E+6 ft 6 Initial Pressure 16.7 psia 16.7 psia Initial Temperature 120 °F 120 °F Post-LOCA Peak Pressure 66.1 psia 74.7 psia Post-LOCA Peak Temperature 251 °F 308°F Power Access Purge (mini-purge) Model 7 Source Terms Iodine, (Dose Equivalent I-131) 60 uCi/gm 60 uCi/gm Noble gases RCS Normal RCS Normal 8 Purge valve type Butterfly Butterfly Purge valve size, inch 8 8 Number of release flow paths 2 2 9 Containment power access purge total 12 sec 12 sec isolation time [duration of release to environment]
| |
| 10 Percent of the primary coolant mass 100% 100%
| |
| released to the containment during the first 12 seconds 11 Source Terms [fraction of core activity Containment Leakage Model initially airborne in the containment]:
| |
| Iodines 25% of core 25% of core Noble gases 100% of core 100% of core 12 Containment leak rate, vol.%/day 0-24 hr 0.1 0.1
| |
| >24 hr 0.05 0.05 13 Containment air leak rate through the 0.9 cfm 0.9 cfm depressurized secondary system, cfm June 2011 15.6-59 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 (contd)
| |
| Large break LOCA radiological analysis parameters and results Parameter Value (3954 MWt) Value (4070 MWt) 14 Duration of containment leakage 30 days 30 days 3
| |
| 15 Containment region volumes, ft :
| |
| Main spray region 2.27E+6 2.27E+6 Auxiliary spray region 0.20E+6 0.20E+6 Unsprayed region 0.15E+6 0.15E+6 total [containment net free volume] 2.62E+06 2.62E+06
| |
| -1 -1 16 Transfer rate between sprayed and 3.3 hr 3.3 hr unsprayed regions, in terms of [8,250 cfm] [8,250 cfm]
| |
| unsprayed Volume change per hour 17 Air transfer rated between the contain-ment regions, cfm:
| |
| Main sprayed and unsprayed regions 7,582 7,582 Auxiliary sprayed and unsprayed 668 668 regions 18 Spray flow start time, second
| |
| - Time to reach High-High containment <1 sec <1 sec pressure setpoint [to generate CSAS]
| |
| - Total instrument response time 33 sec 33 sec
| |
| - Time to fill spray header 58 sec 58 sec total 92 sec 92 sec
| |
| [assumes loss of offsite power]
| |
| 19 Spray Iodine Removal Coefficients, s, (during injection phase):
| |
| Main sprayed region: Coefficients Coefficients
| |
| -1 -1 Elemental 19.6 hr 19.6 hr Organic 0 0
| |
| -1 -1 Particulate 0.32 hr 0.32 hr Auxiliary sprayed region:
| |
| -1 -1 Elemental 6.05 hr 6.05 hr Organic 0 0
| |
| -1 -1 Particulate 0.09 hr 0.09 hr Spray elemental-iodine decontamination DF DF factor 6.51 6.51 20 Removal of elemental iodine by plate-out [wall deposition], p: Coefficients Coefficients Main sprayed region 2.16 hr-1 2.14 hr-1 Auxiliary sprayed region 14.6 hr-1 14.4 hr-1 unsprayed region 14.6 hr-1 14.4 hr-1 Elemental-iodine decontamination fac- DF DF tor 100 100 ESF Recirculation Leakage Model 21 Source Term: Sump activity of iodine, 50% of core 50% of core as a percent of total core activity June 2011 15.6-60 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 (contd)
| |
| Large break LOCA radiological analysis parameters and results Parameter Value (3954 MWt) Value (4070 MWt) 3 22 Sump volume (ft ):
| |
| 3 3 RWT volume 5.35E+04 ft 5.35E+04 ft 3 3 RCS volume, including PZR & CVCS 9.18E+03 ft 9.73E+03 ft 3 3 Safety injection tanks [SITs] 7.16E+03 ft 7.00E+03 ft 3 3 total, Input to LOCADOSE 6.98E+04 ft 7.023E+04 ft 23 Recirculation start time 20 minutes, 20 minutes, post accident post accident 24 Credit assumed for radioactive decay of yes yes iodine prior to recirculation [20 minutes] [20 minutes]
| |
| 25 Total ESF component leakage rate 3,000 ml/hr 3,000 ml/hr 26 Percent of iodine in the leaked water 10% 10%
| |
| which is assumed to become volatile
| |
| [flashing fraction]
| |
| 27 Fuel building/low aux essential filtra-tion Filter efficiency, 2 inch Charcoal (iodine):
| |
| Elemental 95% 95%
| |
| Organic 95% 95%
| |
| Particulate 95% 95%
| |
| 28 Duration of ESF leakage 30-days 30-days 29 Partition coefficient of iodine in RWT 1000 1000 RWT Backleakage Model 30 Dilution volume RWT volume 1.15E+5 1.15E+5 Fuel building Volume 7.45E+5 7.45E+5 Control Room Data refer to section 6.4.7 (b)
| |
| Transport Data 3
| |
| 31 EAB X/Q, 0-2 hr, sec/m 2.3E-4 2.3E-4 3
| |
| LPZ X/Q, sec/m :
| |
| 0-8 hr 6.4E-5 6.4E-5 8-24 hr 4.8E-5 4.8E-5 24-96 hr 2.6E-5 2.6E-5 96-720 hr 1.1E-5 1.1E-5 3
| |
| 32 Offsite Breathing Rated, m /sec:
| |
| 0-8 hr 3.47E-4 3.47E-4 8-24 hr 1.75E-4 1.75E-4
| |
| > 24 hr 2.32E-4 2.32E-4 June 2011 15.6-61 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 Large break LOCA radiological analysis parameters and results Parameter Value (3954 MWt) Value (4070 MWt) 33 Credit for depletion of the effluent Not Assumed Not Assumed plume of radioactive iodine due to depo-sition on the ground 34 Credit for radiological decay in transit Not Assumed Not Assumed Dose Calculation Data 35 Thyroid Inhalation DCFs,-rem/Ci-ICRP-30 ICRP-30 I-131 1.08E+6 1.08E+6 I-132 6.44E+3 6.44E+3 I-133 1.80E+5 1.80E+5 I-134 1.07E+3 1.07E+3 I-135 3.13E+4 3.13E+4 36 Immersion [Beta Skin & Whole-Body] LOCADOSE LOCADOSE DCFs [Reg. Guide [Reg. Guide 1.109] 1.109]
| |
| June 2011 15.6-62 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.5-2 (contd)
| |
| Radiological Consequences of Large leak LOCA c
| |
| 0-2 hr EAB, rem 30-day LPZ, rem 30-day Control room, rem Whole- Whole- Whole-Body Beta Contributor Thyroid Thyroid (b)
| |
| Body Body Thyroid Skin Power Access [mini] 1.03 4.96E-04 2.81E-01 1.38E-04 5.40E-01 8.75E-05 1.68E-03 Purge (1.09) (5.25E-04) (3.02E-01) (1.46E-04) (3.88E-01) (9.26E-05) (1.78E-03)
| |
| Containment Lkg 37.6 2.693 9.22E+01 1.939 5.95 1.10 1.72E+01 (38.8) (2.52) (9.52E+01) (1.90) (6.14) (1.10) (1.71E+00)
| |
| Containment Lkg via 18.6 1.33 4.57E+01 9.59E-01 2.94 5.46E-01 8.50 Depressurized Secondary (19.2) (1.25) (4.71E+01) (9.39E-01) (3.04) (5.42E-01) (8.44)
| |
| System ESF Component Leakage 1.53 6.84E-03 9.93 1.18E-02 5.38E-01 4.09E-04 8.07E-03 (1.57) (6.99E-03) (10.2) (1.21E-02) (5.50E-01) (4.19E-04) (8.26E-03)
| |
| RWT Back-Leakage 2.16E-02 1.10E-04 7.54 4.67E-03 5.08E-01 5.23E-05 1.05E-a
| |
| [@ 43 gpm] (2.34E-02) (1.14E-04) (7.72) (4.81E-03) (3.36E-01) (5.38E-05) (1.08E-03)
| |
| Contribution from NA nil NA nil NA 0.071 NA contain- (nil) (nil) (0.10) ment direct shine Contribution from out NA nil NA nil NA 0.205 NA side cloud to control- (nil) (nil) (0.215) room Contribution from NA nil NA nil NA 0.061 NA control (nil) (nil) (0.065) room essential filtration system Total 58.75 4.03 155.66 2.91 10.48 1.99 25.69 (60.63) (3.77) (160.50) (2.86) (10.45) (2.03) (25.53)
| |
| : a. Contribution from IN 91-56
| |
| : b. The bounding leakage and thyroid dose is given in section 6.4.7.
| |
| : c. Values shown are for licensed power of 3954 MWt, values in parentheses are for licensed power of 4070 MWt.
| |
| June 2011 15.6-63 Revision 16
| |
| | |
| PVNGS UPDATED FSAR DECREASE IN REACTOR COOLANT INVENTORY 15.
| |
| | |
| ==6.6 REFERENCES==
| |
| : 1. "Time Response Design Criteria for Safety-Related Operator Actions," ANS 58.8, ANSI N660, Rev. 2, 1981.
| |
| : 2. "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-75/087, December 31, 1978.
| |
| : 3. Time Response Design Criteria for Safety - Related Operator Actions, ANSI-N660/ANS-58.8, 1984.
| |
| : 4. Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity, International Commission on Radiological Protection (ICRP), Publication 30, Supplement to Part 1, 1980.
| |
| : 5. /Q Based on 1986-1991 Meteorological Data, 13-NC-XX-0204, Rev. 01, A. Karimi, July 30, 1996.
| |
| : 6. Calculation of Distance Factors for Power and Test Reactor Sites, TID 14844, DiNunno, J.J., et al, March 1962.
| |
| : 7. PVNGS Technical Specification, Amendment Nos. 75, 61, and 47 for Units 1, 2, and 3 respectively, May 16, 1994.
| |
| : 8. PVNGS Technical Specification, Amendment Nos. 109, 101, and 81 for Units 1, 2, and 3 respectively, October 23, 1996.
| |
| : 9. Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System CESSAR System 80, Section 15.3.7, Steam Generator Tube Rupture, NUREG-0852, Supplement 2, September 1983.
| |
| June 2011 15.6-64 Revision 16
| |
| | |
| PVNGS UPDATED FSAR 15.7 RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 WASTE GAS SYSTEM FAILURE 15.7.1.1 Identification of Event and Causes The most limiting waste gas system accident is defined as an uncontrolled release to the atmosphere of the contents of one waste gas decay tank. The gaseous radwaste system is described in section 11.3.
| |
| This accident is considered a limiting fault and is analyzed to define the worst consequences of a gaseous release that could result from any malfunction in the gaseous radwaste system.
| |
| The accident as described assumes a combined failure of the waste gas decay tank and of the normal (non-ESF) radwaste building ventilation (filtration) system described in section 9.4.
| |
| 15.7.1.2 Sequence of Events and System Operation A sequence of events diagram for this accident is provided as figure 15.7.1-1. The event is characterized as a rapid release of the contents of a single waste gas decay tank to the environment (Puff model). It is postulated that the tank contains its maximum inventory and that no action is taken to mitigate the consequences of the event.
| |
| 15.7.1.3 Analysis of Effects and Consequences The instantaneous release of the waste gas decay tank inventory will result in the radiological consequences shown in table 15.7.1-1. No credit has been taken for control room essential HVAC System.
| |
| June 2001 15.7-1 Revision 11
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7.1-1 ASSUMPTIONS AND RADIOLOGICAL CONSEQUENCES OF WASTE GAS SYSTEM FAILURE Parameter Value
| |
| : 1. System parameters RCS mass
| |
| [lbm] Control room air volume 571,776
| |
| [ft3] 1.61E+05
| |
| : 2. Control room normal air handling unit: maximum outside 1,200 (a) air flow rate [cfm]
| |
| : 3. Control room in-leakage 10 (ingress/ egress)[scfm]
| |
| : 4. Control room essential air no credit handling unit (AHU) and recirculating charcoal filter unit
| |
| : 5. Iodine species fractions (after discharge) 0.91 elemental 0.04 organic 0.05 particulate
| |
| : 6. Atmospheric dispersion Refer to factor section 3
| |
| (/Q)[sec/m ] 2.3.4
| |
| : 7. Control room accident Refer to
| |
| /Q[sec/m ]
| |
| 3 Appendix 15B,table B-5
| |
| : 8. 3 Breathing rate [m /sec]
| |
| 0-720 hour offsite 3.47E-04 0-720 hour control room 3.47E-04
| |
| : 9. Iodine removal efficiency due to RCS degassing by gas-stripper iodines 0.1 noble gases 1.0
| |
| : 10. Pre-holdup ion exchanger decontamination factor (DF) 10 iodines 1 noble gases
| |
| : 11. Primary coolant letdown 140 flow Rate through the gas stripper
| |
| [gpm]
| |
| June 2003 15.7-2 Revision 12
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7.1-1 ASSUMPTIONS AND RADIOLOGICAL CONSEQUENCES OF WASTE GAS SYSTEM FAILURE Parameter Value
| |
| : 12. Control room accident /Q for adjacent unit [sec/m3]
| |
| 0-8 hr 4.80E-04 8-24 hr 3.41E-04 24-96 hr 1.37E-04 96-720 hr 2.76E-05
| |
| : 13. Source term, Max GRS Activity [Ci]:
| |
| Kr-83m 1.53E-01 Kr-85m 9.54E+01 Kr-85 1.59E+03 Kr-87 3.01E+00 Kr-88 1.00E+02 Xe-131m 1.50E+03 Xe-133m 7.94E+01 Xe-133 8.95E+04 Xe-135m 7.01E-08 Xe-135 1.07E+03 Xe-138 6.16E-09 I-131 7.57E+00 I-132 1.81E-01 I-133 8.72E+00 I-134 2.08E-03 I-135 2.76E+00
| |
| : a. The bounding inleakage is given in Section 6.4.7.
| |
| June 2003 15.7-3 Revision 12
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7.1-1 ASSUMPTIONS AND RADIOLOGICAL CONSEQUENCES OF WASTE GAS SYSTEM FAILURE Parameter Value Results rem EAB (0-2hr)
| |
| Thyroid 7.85E-01 Whole Body 2.19E-01 LPZ (0-30 days)
| |
| Thyroid Whole Body 2.18E-01 6.11E-02 Control room (effected unit)(0-30 day)
| |
| Thyroid Whole Body 5.22E-02 1.42E-01 Control room (adjacent unit)(0-30 day)
| |
| Thyroid 1.61E-02 Whole Body 2.18E-02 15.7.1.4 Conclusions As noted in table 15.7.1-1, the radiological consequences are less than 1% of 10CFR100 limits even assuming coincident failure of the normal (non-ESF) radwaste building ventilation system.
| |
| 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE (RELEASE TO ATMOSPHERE)
| |
| The most limiting liquid waste system leak or failure would be the failure of the representative outside liquid storage tank.
| |
| Refer to subsection 15.7.3.
| |
| June 2001 15.7-4 Revision 11
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING TANK FAILURES 15.7.3.1 Identification of Event and Causes The most limiting radioactive liquid tank failure would be the uncontrolled release of liquid from a representative outside liquid storage tank. The representative outside liquid storage tank is a hypothetical tank that bounds all possible outside liquid storage tank ruptures under the absolute worst case conditions. The tank is assumed to contain the Technical Specification maximum allowable curie content, with the mixture of isotopes representative of the isotopic mixture present in RCS fluid at 600 EFPD, no gas stripping, and with 1% failed fuel. The use of the RCS source terms are intended to remove any uncertainty based on operational considerations in the liquid processing systems that feed the outside liquid storage tanks that are part of the chemical and volume control system (CVCS) as described in subsection 9.3.4, or are part of the liquid radwaste system (LRS) as described in section 11.2. In addition to use of the RCS isotopic inventory mixture, noble gases are included in the representative outside liquid storage tank inventory and are available for release. This conservative practice will bound all possible outside liquid storage tank ruptures under the worst case conditions.
| |
| A nonmechanistic failure that instantaneously releases 100% of the tank's contents to the environment is postulated.
| |
| 15.7.3.2 Sequence of Events and System Operation A sequence of events diagram for this accident is provided as figure 15.7.3-1. The event is characterized as a rapid release of the representative outside liquid storage tank contents to June 2001 15.7-5 Revision 11
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT the environment. It is postulated that no action is taken to mitigate the consequences of the event. Dose modeling methodology is described in subsection 15B.6.5.
| |
| 15.7.3.3 Analysis of Effects and Consequences The radiological consequences of the representative liquid storage tank rupture are described in table 15.7.3-1. The instantaneous release of the representative liquid storage tank gaseous inventory will result in radiological consequences less than waste gas system failure as described section 15.7.1.3.
| |
| Table 15.7.3-1 RADIOLOGICAL CONSEQUENCES OF A REPRESENTATIVE LIQUID STORAGE TANK FAILURE Parameter Value Results Rem EAB (0 - 2 hr) Thyroid 1.184E-02 EAB (0 - 2 hr) Whole Body 4.594E-03 LPZ (0 - 8 hr) Thyroid 3.293E-03 LPZ (0 - 8 hr) Whole Body 1.278E-03 15.7.3.4 Conclusions As noted in table 15.7.3-1 the radiological consequences of gaseous release from a Representative Liquid Storage Tank are less than 1% of 10CFR100 limits even though nonmechanistic instantaneous failure of the tank was postulated.
| |
| June 2001 15.7-6 Revision 11
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.4 RADIOLOGICAL CONSEQUENCES OF FUEL HANDLING ACCIDENTS The fuel handling accident is considered to occur at two locations at PVNGS: outside the containment building in the fuel building, and inside the containment. The events at each location are independent of each other. The failure modes for fuel handling equipment inside and outside containment are non-mechanistic and therefore the initiating events in the two buildings are independent. The analyses described herein are based on movement of only one fuel assembly.
| |
| 15.7.4.1 Fuel Handling Accident Outside Containment 15.7.4.1.1 Identification of Event and Causes The fuel handling accident that is considered results from the dropping of a single fuel assembly during fuel handling.
| |
| 15.7.4.1.2 Sequence of Events and Systems Operation A sequence of events diagram for this accident is provided as figure 15.7.4-1. The radiation monitoring system (RMS),
| |
| described in section 11.5, will provide prompt notification of high airborne radiation levels in the fuel building which may develop as a result of a fuel handling accident. Additionally, the safety-related monitors of the RMS will initiate generation of the fuel building essential ventilation actuation signal June 2007 15.7-7 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT (FBEVAS). Engineered safety features equipment functions following a FBEVAS are described in section 9.4. The FBEVAS logic is described in section 7.3.
| |
| By reducing building exhaust rates and initiating exhaust filtration, ESF actions will substantially reduce potential offsite radiological exposures in the event of a fuel handling accident.
| |
| 15.7.4.1.3 Analysis of Effects and Consequences If a dropped assembly were damaged to the extent that one or more fuel rods were broken, the accumulated fission gases and iodines in the fuel rod gaps would be released to the surrounding water. Release of the solid fission products in the fuel would be negligible because of the low fuel temperature during refueling. The fuel assemblies are stored within the spent fuel racks resting on the bottom of the spent fuel pool. The tops of the racks extend above the tops of the stored fuel assemblies. A dropped fuel assembly could not strike more than one fuel assembly in the storage (1) rack. Impact could occur only between the ends of the involved fuel assemblies, with the lower end fitting of the dropped fuel assembly impacting against the upper end fitting of the stored fuel assembly.
| |
| A. Vertical Fuel Assembly Drop Analytical methods used to calculate the impact velocity and the resulting impact stress in Zircaloy-4 and ZIRLO fuel rod cladding for the vertical drop are described below.
| |
| The analysis of the fuel assembly vertical drop employed a summation of the forces acting on the fuel assembly in the vertical direction to determine the equation of June 2007 15.7-8 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT motion of the fuel assembly. The resulting equation of motion is given below:
| |
| Fvert = M x a = F d + Fb - F w where:
| |
| M = mass of a fuel assembly a = acceleration Fd = drag force of a fuel assembly 2
| |
| (drag coefficient x (velocity) )
| |
| _______
| |
| : 1. In Unit 2 Spent Fuel Pool Storage Location A38, Element P2F003 and surrounding support apparatus are undergoing long term storage. Due to structural damage and additional height provided by the support apparatus, Element P2F003 protrudes 2.59 inches above the upper surface of the storage rack. This condition makes it possible for an element dropped near vertically to strike one element seated in storage location and then rotate to strike the top of Element P2F003.
| |
| June 2003 15.7-9 Revision 12
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Fb = buoyant force of a fuel assembly Fw = weight (dry) of a fuel assembly The analysis assumed the fuel assembly drop distance was sufficient for the fuel assembly to reach its terminal velocity (acceleration equals zero in the above equation), thus making the results conservative or applicable for any drop height. For this worst case, the terminal velocity, and therefore the assumed impact velocity of the fuel assembly, is 254.4 and 240 inches per second for Zircaloy-4 and ZIRLO fuel, respectively, and the resulting stress in the fuel rod cladding is 24,000 psi for Zircaloy-4 and 22,320 psi for ZIRLO.
| |
| The equation employed in calculating the above impact stress in the fuel rod clad is as follows:
| |
| Si = Vi (E/)
| |
| where:
| |
| Si = impact stress Vi = impact velocity E= modulus of elasticity
| |
| = specific volume The yield stress of the fuel rod cladding is 49,000 psi for Zircaloy-4 and 81,785 psi for ZIRLO. This is the minimum yield stress value for unirradiated fuel and is conservative for irradiated fuel. Thus, for the fuel assembly vertical drop, the impact stresses which result from absorbing the kinetic energy of the drop are below the yield stress of the clad for both the Zircaloy-4 and ZIRLO fuel assembly, and no fuel rod failures will occur.
| |
| June 2005 15.7-10 Revision 13
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT B. Horizontal Fuel Assembly Drop Horizontal impact of a fuel assembly could result from a dropped fuel assembly falling in the horizontal position, or from a vertical fuel assembly rotating to the horizontal position. This event assumes that all fuel rods in the dropped fuel assembly fail, which is consistent with Regulatory Guide 1.25 assumptions. Thus, the dose analysis is bounding for the Zircaloy-4 and ZIRLO fuel assembly horizontal drop analyses.
| |
| C. Input Parameters and Initial Conditions for the Radiological Analysis For the radiological consequences of a fuel handling accident evaluation, cladding failure was assumed to occur for all fuel rods in an assembly. The reactor was assumed to operate at a power level of 3990 MWt, and the earliest time at which a spent fuel assembly can be moved is considered to be 72 hours after shutdown.
| |
| Assumptions and parameters used in evaluating the fuel handling accident are listed in table 15.7.4-1. The calculational methods and assumptions described in Regulatory Guide 1.25 apply since the dropped fuel assembly meets all of the requirements of Regulatory Guide 1.25, except as discussed below.
| |
| Based on the plant specific fuel cycle design and core, the maximum fuel rod discharge pressure would exceed the Regulatory Guide 1.25 limit of 1200 psig. Therefore, a site specific methodology for addressing fuel rod pressure has been developed and is implemented herein with prior NRC approval. This methodology (Reference 3) determines the peak assembly average fuel pin pressure rather than June 2007 15.7-11 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT the maximum fuel pin pressure and demonstrates that it is less than the Regulatory Guide 1.25 limit of 1200 psig.
| |
| The peak assembly average fuel pin pressure is determined by first determining the maximum fuel rod pressure based on worst case fuel cycle parameters using computer code FATES 3B. The fuel pin internal gas pressure is based on the FATES 3B analysis for the hot rod in the hot assembly and is performed and verified for each core reload.
| |
| Pressure in a fuel rod in the spent fuel pool or the refueling pool is obtained by adjusting the FATES 3B results for the temperature of the water in the spent fuel pool. This methodology is independent of individual fuel rods (or fuel assembly since all rods in an assembly are assumed to be at maximum pressure) and fuel rod core location and results in a peak rod internal pressure for the fuel type analyzed by FATES 3B.
| |
| Each type of fuel is analyzed in the same manner. Then the maximum peak assembly average fuel pin pressure is calculated by averaging the pressure for the different types of fuel pins in an assembly.
| |
| Sum of Individual Pin Pressures Average Pin Pressure =
| |
| Number of Pins in Assembly
| |
| <1200 psig June 2007 15.7-12 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7.4-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT Reg. Guide 1.25 Calculational Parameter Assumptions Assumption Fuel Assy. Data: (1)
| |
| Radial peaking factor 1.65 1.7 Burnup, MWD/MTU 25,000 70,000 Max. fuel pin pressure, psig 1,200 Peak assembly average (average pressure for 236 fuel pin pressure is fuel rods in peak assembly) < 1200 when fuel is (2) being moved Decay time, hours None 72 Number of failed pins All fuel rods in an All fuel rods in assembly assembly (236)
| |
| Fraction of fission product gases contained in the gap region of fuel rods, %
| |
| Kr-85 30 30(3)
| |
| Xe-133 10 16(3)
| |
| Other noble gases 10 15(3)
| |
| Iodine 10 15(3)
| |
| Iodine gap chemical composition, %
| |
| Organic iodine compounds 0.25 0.25 Inorganic iodine compounds 99.75 99.75 Percentage of gap activity released to pool 100 100 Peak assembly shutdown gap Not quantified I-131------- 1.08E+05 activity, for key isotopes beyond 100% of gap Xe-131m----- 1.12E+03 (Ci): inventory Xe-133------ 2.58E+05 Kr-85------- 3.78E+03 Minimum water depth, Fuel 23 22.5 Pool surface to top of damaged fuel rods, feet Fuel Building filter efficiencies, %
| |
| Filter eff., organic iodine 70 70 Filter eff., inorganic iodine 90 90 Pool decontamination factors:
| |
| Iodine 100 100 Noble gases 1 1 June 2007 15.7-13 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT Reg. Guide 1.25 Calculational Parameter Assumptions Assumption Composition of iodine in atmosphere above pool, %:
| |
| Organic iodine compounds 25 25 Inorganic iodine compounds 75 75 Time for radioactive material to escape from fuel bldg.,
| |
| hours 99.9% in 2 hours >99.9% in 2 hours Filter system effluent Typically none; All activity escaping dilution factor direct passage to the water surface in exhaust system. Fuel Bld./Cont.
| |
| (Evaluated on immediately &
| |
| individual basis.) homogeneously mixed throughout building 3
| |
| volume of 7.36E+05 ft Atmospheric diffusion factors Interim factors provided until site
| |
| --Inside Containment meteorological data See Table 15B-5 obtained; licensee
| |
| --Outside Containment directed to use site 3 2.24E-04 sec/m specific data when available
| |
| : 1) Fuel data valid only for oxide fuels with:
| |
| a) Highest power assy. peak linear power density no greater than 20.5 kW/ft.;
| |
| b) Highest power assy. center-line operating temp. <4,500°F.
| |
| : 2) FATES 3B computer code.
| |
| : 3) Methodology conservatively assumed to be bounding for Zircaloy-4 and ZIRLO fuel; exceeds values calculated by ANSI/ANS 5.4-1982.
| |
| June 2007 15.7-14 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT The releases from the spent fuel pool are calculated in accordance with Regulatory Guide 1.25. The top of the highest (2) damaged fuel rod would be at elevation 115 feet. The spent fuel pool low water level alarm and high water level alarm setpoints are at elevations 137 feet 6 inches and 138 feet 2 inches, respectively. As such, there is a nominal 23 feet of water over the damaged fuel pins. Even if the pool water level were at the low level alarm point, there would be a minimum of 22 feet 6 inches of water over the damaged fuel pins.
| |
| Gap activities at reactor shutdown in the fuel assembly damaged as a result of a fuel handling accident in accordance with Regulatory Guide 1.25 are given in table 15.7.4-1. This activity is assumed to be instantaneously released to the fuel building subsequent to evolution from the spent fuel pool.
| |
| ________
| |
| 2 In Unit 2 Spent Fuel Pool Storage Location A38, Element P2F003 and surrounding support apparatus are undergoing long term storage. Due to structural damage and additional height provided by the support apparatus, Element P2F003 protrudes 2.59 inches above the upper surface of the storage rack. In the most conservative case of an element falling onto and being supported by this Element P2F003, the highest point on the supported element would be at elevation 115' 5.18". An administratively controlled minimum Spent Fuel Pool water level of 138' ensures at least 22'6" of water coverage of all parts of the dropped element. This is only required during fuel movement within the administrative restricted area surrounding fuel element P2F003.
| |
| June 2007 15.7-15 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT The fuel building essential ventilation system will actuate on high radiation, which would minimize the releases to the environment.
| |
| The offsite doses at the exclusion area boundary and low population zones for a fuel handling accident outside contain-ment are presented in table 15.7.4-3.
| |
| 15.7.4.1.4 Conclusions As noted in table 15.7.4-3, the radiological consequences of a fuel handling accident outside containment are within SRP limits and well within (25%) 10CFR100 limits, and are in all cases bounded by the postulated fuel handling accident inside containment doses. Control room doses are discussed in Section 6.4.7.3. The operator doses are within the limits set by 10CFR50 Appendix A GDC 19.
| |
| Table 15.7.4-3 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT OUTSIDE CONTAINMENT Dose (rem)
| |
| Thyroid Whole Body
| |
| -1 2-hour 25.1 2.88 x 10 exclusion area boundary
| |
| -1 30-day 7.9 1.11 x 10 low popula-tion zone June 2007 15.7-16 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.4.2 Fuel Handling Accident Inside Containment 15.7.4.2.1 Identification of Event and Causes Refer to paragraph 15.7.4.1.1.
| |
| 15.7.4.2.2 Sequence of Events and Systems Operation A sequence of events diagram for this accident is provided as figure 15.7.4-2. The RMS, described in section 11.5, will provide prompt notification of high airborne radiation levels in the containment which may develop as a result of a fuel handling accident.
| |
| 15.7.4.2.3 Analysis of Effects and Consequences The analysis presented in paragraph 15.7.4.1.3 for a fuel handling accident outside containment is also applicable to this section and as paragraph 15.7.4.1.3 assumes acceleration is zero at the terminal velocity. Thus, the consequences of a fuel handling accident in the containment are no more severe in terms of activity released from the fuel element than a fuel handling accident outside containment.
| |
| The releases from the refueling pool are calculated in accordance with the assumptions of Regulatory Guide 1.25.
| |
| It is also assumed that, for the postulated accident, this activity is instantaneously released to the containment and released to the outside atmosphere within two hours.
| |
| The 2-hour doses at the exclusion area boundary and 30 day low population zone doses are presented in table 15.7.4-5.
| |
| June 2007 15.7-17 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.4.2.4 Conclusions As shown in the results of the analyses in table 15.7.4-5, the radiological consequences of a fuel handling accident inside containment are within SRP limits and well within (25%)
| |
| 10CFR100 limits. Control room doses are discussed in Section 6.4.7.3. The operator doses are within the limits set by 10CFR50 Appendix A GDC 19.
| |
| Table 15.7.4-5 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT (WITHOUT REFUELING PURGE ISOLATION)
| |
| Dose (rem)
| |
| Thyroid Whole Body
| |
| -1 2-hour exclusion 74.7 4.16 x 10 area boundary
| |
| -1 30-day low 20.8 1.16 x 10 population zone 15.7.5 SPENT FUEL CASK DROP ACCIDENT The probability of fuel handling accidents in the fuel building that result from dropping either a TSC/TFR containing spent fuel or other heavy load from the single failure proof Cask Handling Crane is sufficiently small that they are not credible events, and therefore do not require analysis. The Cask Handling Crane, the TSC, the TFR, and the associated lifting devices used for dry fuel storage handling in the fuel building meet the applicable criteria of NUREG-0612 Section 5.1.6 (Single-Failure-Proof Handling Systems). Transport of loaded June 2007 15.7-18 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT casks to the ISFSI storage location is performed within the bounds of the NAC-UMS Certificate of Compliance (CoC) (Docket no, 72-1015) and the NAC-UMS FSAR. Refer to the NAC-UMS and the ISFSI 72.212 Evaluation Report for details of fuel handling accidents during these operations. Interlocks and procedural and administrative controls involved in fuel handling are described in subsection 9.1.4.
| |
| June 2007 15.7-19 Revision 14
| |
| | |
| PVNGS UPDATED FSAR RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 15.
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| | |
| ==7.6 REFERENCES==
| |
| : 1. "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x16 PWR Fuel,"
| |
| CEN-386-P-A, August 1992.
| |
| : 2. "Report on the Implementation of a 1-Pin Burnup Limit of 60 MWD/kgU at PVNGS," CEN-427-(V)-P, November 1995.
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| : 3. Technical Specification for Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments RE:
| |
| Internal Fuel Pin Pressure (TAC Nos. MC0620, MC0621, and MC0622), NRC letter dated September 27, 2004 (Amendment 153).
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| June 2007 15.7-20 Revision 14
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| PVNGS UPDATED FSAR APPENDIX 15A RESPONSES TO NRC REQUESTS FOR INFORMATION
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| PVNGS UPDATED FSAR PVNGS UPDATED FSAR CONTENTS Page Question 15A.1 DELETED 15A-1 Question 15A.2 (NRC Question 450.10) (15.6.2) 15A-1 Question 15A.3 (NRC Question 450.11) (15.7.3) 15A-1 Question 15A.4 (NRC Question 460.19) (15.7) 15A-1 Question 15A.5 (NRC Question 450.13) (15.6.3) 15A-2 Question 15A.6 (NRC Question 450.14) (15.6.3) 15A-3 Question 15A.7 (NRC Question 450.15) (15.6.3) 15A-3 Question 15A.8 (NRC Question 450.16) (15.6.3) 15A-4 Question 15A.9 (NRC Question 450.17) (15.6.2) 15A-5 Question 15A.10 (NRC Question 450.19) (15.7.4) 15A-6 Question 15A.11 (NRC Question 440.32) (15.0) 15A-8 Question 15A.12 (NRC Question 440.33) (15.0) 15A-8 Question 15A.13 (NRC Question 440.34) (15.0) 15A-9 Question 15A.14 (NRC Question 440.35) (15.0) 15A-9 Question 15A.15 (NRC Question 440.36) (15.0) 15A-9 Question 15A.16 (NRC Question 440.37) (15.0) 15A-10 Question 15A.17 DELETED 15A-10 Question 15A.18 (NRC Question 440.39) (15.0) 15A-10 Question 15A.19 (NRC Question 440.40) (15.0) 15A-11 Question 15A.20 (NRC Question 440.41) (15.0) 15A-11 Question 15A.21 (NRC Question 440.42) (15B) 15A-11 Question 15A.22 (NRC Question 440.43) (15B) 15A-14 June 2005 15A-i Revision 13
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| PVNGS UPDATED FSAR CONTENTS (cont)
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| Page Question 15A.23 (NRC Question 440.44) (15.0) 15A-14 Question 15A.24 DELETED 15A-15 Question 15A.25 (NRC Question 440.46) (15.0) 15A-15 Question 15A.26 DELETED 15A-16 Question 15A.27 (NRC Question 440.48) (15.0) 15A-16 Question 15A.28 (NRC Question 440.49) (15.0) 15A-16 Question 15A.29 DELETED 15A-17 Question 15A.30 DELETED 15A-17 Question 15A.31 DELETED 15A-17 Question 15A.32 DELETED 15A-17 Question 15A.33 DELETED 15A-17 Question 15A.34 (NRC Question 440.55) (15.6) 15A-17 Question 15A.35 DELETED 15A-18 Question 15A.36 (NRC Question 440.58) (15.6.3) 15A-18 Question 15A.37 (NRC Question 440.59) (15.6.3) 15A-18 Question 15A.38 (NRC Question 440.60) (15.6.3) 15A-18 Question 15A.39 (NRC Question 440.61) (15.6.3) 15A-19 Question 15A.40 (NRC Question 440.62) (15.6.3) 15A-19 Question 15A.41 (NRC Question 440.63) (15.6.3) 15A-19 Question 15A.42 (NRC Question 440.64) (15.0) 15A-20 Question 15A.43 (NRC Question 440.65) (15.2) 15A-20 Question 15A.44 (NRC Question 440.66) (15A) 15A-21 June 2005 15A-ii Revision 13
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| PVNGS UPDATED FSAR CONTENTS (Cont)
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| Page Question 15A.45 (NRC Question 440.67) (15.3) 15A-22 Question 15A.46 (NRC Question 440.68) (15.0) 15A-22 Question 15A.47 (NRC Question 440.69) (15.5) 15A-23 Question 15A.48 (NRC Question 440.70) (15.0) 15A-24 Question 15A.49 (NRC Question 440.71) (15D) 15A-24 Question 15A.50 DELETED 15A-25 Question 15A.51 (NRC Question 440.73) (15D) 15A-25 Question 15A.52 (NRC Question 440.74) (15D) 15A-25 Question 15A.53 (NRC Question 440.75) (15.6) 15A-26 Question 15A.54 (NRC Question 440.76) (15.8) 15A-26 Question 15A.55 (NRC Question 440.82) (15.0) 15A-27 Question 15A.56 (SGTR Question 1) (15.6.3) 15A-29 Question 15A.57 (SGTR Question 2) (15.6.3) 15A-32 Question 15A.58 (SGTR Question 3) (15.6.3) 15A-32 Question 15A.59 (SGTR Question 1) (15.6.3) 15A-43 Question 15A.60 (SGTR Question 2) (15.6.3) 15A-45 Question 15A.61 (SGTR Question 3) (15.6.3) 15A-46 Question 15A.62 (SGTR Question 4) (15.6.3) 15A-47 June 2005 15A-iii Revision 13
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| PVNGS UPDATED FSAR TABLES Page 15A-1 Radiological Consequences of the Steam Generator Tube Rupture with a Loss of Offsite Power and Fully Stuck Open ADV 15A-31 15A-2 Sequence of Events for a Steam Generator Tube Rupture with a Loss of Offsite Power and Fully Stuck Open ADV 15A-36 June 2005 15A-iv Revision 13
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| | |
| PVNGS UPDATED FSAR FIGURES 15A-1 Control Schematic of an Atmospheric Dump Valve 15A-2 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Core Power vs Time 15A-3 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve RCS Pressure vs Time 15A-4 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Core Coolant Temperature vs Time 15A-5 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Upper Head Temperature vs Time 15A-6 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Pressurizer Water Volume vs Time 15A-7 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Liquid Volume Above Top of Hot Leg vs Time 15A-8 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve RCS Liquid Mass vs Time 15A-9 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Steam Generator Pressure vs Time 15A-10 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Integrated AFW Flow to Affected SG vs Time June 2001 15A-v Revision 11
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| PVNGS UPDATED FSAR FIGURES (Cont) 15A-11 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Tube Leak Rate vs Time 15A-12 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Integrated Leak Flow vs Time 15A-13 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Fraction of Leak Flashed vs Time 15A-14 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Steam Generator Mass vs Time 15A-15 Steam Generator Tube Rupture with Loss of Offsite Power and a Fully Stuck Open Atmospheric Dump Valve Integrated ADV Flow vs Time June 2001 15A-vi Revision 11
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| PVNGS UPDATED FSAR QUESTION 15A.1 DELETED QUESTION 15A.2 (NRC Question 450.10) (15.6.2)
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| In evaluating the double-ended break of the letdown line outside containment, provide the following:
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| (1) summary of primary system's iodine activity, including the potential increase in iodine release rate (iodine spiking) above the equilibrium value during the accident and its effect on the accident doses; and (2) valve closure time and maximum permissible leakage rate of the letdown line isolation valve.
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| RESPONSE: The response will be provided on the CESSAR docket.
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| QUESTION 15A.3 (NRC No. 450.11) (15.7.3)
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| In evaluating the radioactive liquid waste system leak or failure, provide data, assumptions and methodology used in analyzing the radiological consequences of fission gases released to the atmosphere.
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| RESPONSE: The response is provided in revised paragraph 15.7.3.2.
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| QUESTION 15A.4 (NRC Question 460.19) (15.7)
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| We are currently evaluating the liquid radwaste tank failure accident for Palo Verde Nuclear Generating Station (PVNGS)
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| Units 1, 2, and 3. Based on our evaluation, we have concluded that the tank most likely to result in the highest levels of concentrations in the release in the event of tank failure is the concentrate monitor tank which is located in the radwaste building at elevation 100 feet (on grade) and has a capacity of June 2005 15A-1 Revision 13
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| | |
| PVNGS UPDATED FSAR 4000 gallons when filled to 80% of its capacity (see engineering drawing 13-P-OOB-003, FSAR, PVNGS for location).
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| 8 Our calculations show that either a dilution factor of 3 x 10 or a transit time of approximately 718 years will be required in order to ensure that the radionuclide concentrations at the applicable location are well below the 10CFR Part 20 limits.
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| Please indicate whether such values can be expected at the applicable location for PVNGS and if not what dilution factor and transit time can be expected at the applicable location.
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| RESPONSE: The concentrate monitor tanks are located inside the radwaste building. In the event of tank failure, all leakage would be contained by curbs and floor drains and pumped (by sump pumps) for holdup and/or processing. Accordingly, no ground contamination would occur.
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| The concentrate monitor tank gaseous inventory is expected to be very small (refer to table 12.2-5). Thus, radiological impact due to concentrate tank failure is projected to be less than that of the refueling water tank (RWT) as noted in subsections 15.7.2 and 15.7.3. The RWT has the highest liquid and gaseous inventories and concentrations of any outside tank. Dilution factors and transit times for the perched water and regional aquifer for failure of the RWT are presented in paragraph 2.14.13.3.
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| QUESTION 15A.5 (NRC Question 450.13) (15.6.3)
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| The radiological consequences of the steam generator tube rupture accident are presented in CESSAR Section 15.6.3 for only the exclusion area boundary. This does not fully comply with the guidelines of 10CFR Part 100 which requires a June 2005 15A-2 Revision 13
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| | |
| PVNGS UPDATED FSAR radiological consequence evaluation be performed at a low population zone for the duration of the accident. It is the staff position that CE provide the assumed LPZ envelope and provide the estimated radiological consequences at the assumed LPZ boundary for the duration of the accident.
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| RESPONSE: The response will be provided on the CESSAR docket.
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| QUESTION 15A.6 (NRC Question 450.14) (15.6.3)
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| The amount of steam released from the affected and unaffected steam generators presented in FSAR subsection 15.6.3 are based upon the assumption of recovery of offsite power and condenser during the accident. This is contrary to the staff position of assuming that offsite power is lost for the duration of the accident and therefore the condenser is never available for steam dump. Because the staff uses the applicant's values or curves of the steam release estimates in its radiological consequence calculations, it is not possible to complete our review until appropriate steam release values are received.
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| Provide the estimates of steam release to the environment assuming that the condenser is not available for the accident duration.
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| RESPONSE: The response will be provided on the CESSAR docket.
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| QUESTION 15A.7 (NRC Question 450.15) (15.6.3)
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| The analysis of the radiological consequences presented in FSAR subsection 15.6.3 does not appear to consider the steam release pathway occurring from the steam jet air ejectors (SJAEs) to the environment prior to assumed loss of the condenser.
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| Because the condenser is available for a significant period June 2005 15A-3 Revision 13
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| PVNGS UPDATED FSAR prior to condenser trip and the primary to secondary leakage is greatest prior to the loss of the condenser, it is the staff position that the steam release pathway through the SJAEs to the environment needs to be considered in the radiological consequence analysis. Provide the amount of steam released through the SJAEs to the environment prior to the loss of the condenser.
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| RESPONSE: The PVNGS design does not include SJAEs. Air is removed via the condenser air removal system. This pathway is monitored for radioactivity prior to release and is filtered by charcoal and HEPA filters in the event of high effluent radioactivity. (Refer also to CESSAR Section 15.6.3)
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| QUESTION 15A.8 (NRC Question 450.16) (15.6.3)
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| The radiological consequence evaluation provided in CESSAR Section 15.6.3 are based upon assumptions which vary greatly from previous staff practice in the following areas:
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| : 1) The evaluations were performed using values which are less than the proposed technical specification limits for normal operation.
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| This is not acceptable to the staff because the technical specifications define the operating envelope under which the plant can operate without restriction. Analyses using values less than the Technical Specifications do not verify that at the Technical Specification limits the plant would operate safely and that the radiological consequences would not exceed the staff acceptance criteria on radiological exposures.
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| June 2005 15A-4 Revision 13
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| | |
| PVNGS UPDATED FSAR
| |
| : 2) The radiological consequence analysis of a steam generator tube rupture makes no mention of iodine spiking in the CESSAR document, and the Palo Verde docket uses a spiking factor of only 100.
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| The staff position on an acceptable spiking factor is provided in SRP Section 15.6.3. This section states that a spiking factor of 500 times the normal release rate of iodine from the fuel should be used.
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| : 3) The iodine transport in the steam generator is determined using CENPD-180 and its supplement.
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| The staff position on iodine transport in the steam generator is defined in SRP Section 15.6.3 and is that the iodine transport should be determined using the methods and models described in NUREG-0409. The CESSAR and Palo Verde dockets do not discuss the differences in the methods used to those proposed in NUREG-0409.
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| Based upon the above discussion, the staff position is that the applicant provide an analysis of the radiological consequences of the steam generator tube rupture which assumes operation at the proposed technical specification values and describes the differences between the models used and those presented in the staff Standard Review Plan.
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| RESPONSE: The response will be provided on the CESSAR docket.
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| QUESTION 15A.9 (NRC Question 450.17) (15.6.2)
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| In the evaluation of the double-ended break of the letdown line outside containment-upstream of the letdown line control valve (CESSAR Section 15.6.2), the staff has calculated the dose in June 2005 15A-5 Revision 13
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| PVNGS UPDATED FSAR accordance with SRP 15.6.2. The result shows the EAB, 0-2 hours, thyroid dose to be 85 rems. This value is more than twice the acceptable limit of 30 rems as defined in SRP 15.6.2.
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| The staff position is that the maximum equilibrium fission product concentration given in the technical specification be reduced from 4.7 µci/cc to the standard 1 µci/cc. This measure will correspondingly reduce the dose to within acceptable levels. State your intent regarding compliance with our position.
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| RESPONSE: The response will be provided on the CESSAR docket.
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| QUESTION 15A.10 (NRC Question 450.19) (15.7.4)
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| In order to complete our evaluation of the fuel handling accident analysis we request that you provide the following information:
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| (1) Location of RMS detector used to isolate containment refueling purge system and air flow transit time between detector and valve based on normal flowrates.
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| (2) Specify if the RMS detector used to isolate containment refueling purge system is ESF grade and redundant and, if so, include location of redundant detector.
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| (3) Location of RMS detector used to isolate fuel handling building in the event of a fuel handling accident and air flow transit line between detector and damper based on normal flowrates.
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| (4) Specify if RMS detector used to isolate fuel handling building, in the event of a fuel handling accident, June 2005 15A-6 Revision 13
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| PVNGS UPDATED FSAR is ESF grade and redundant and, if so, include location of redundant detector.
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| RESPONSE: As noted in paragraphs 15.7.4.1 and 15.7.4.2, offsite doses due to fuel handling accidents will be a small fraction of 10CFR100 limits even without isolation or filtration of containment or fuel building exhausts.
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| Protective action is not required. The design, however, does include radiation monitors to sense the occurrence of an accident and initiate protective action.
| |
| : 1) The PAPA area radiation monitors isolate the containment refueling purge upon high radiation.
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| They are located just outside the containment between the refueling purge exhaust ducting and the power access purge ducting as shown on engineering drawing 13-N-RAR-004. Air flow transit time is less than 1/4 second.
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| : 2) There are two redundant, ESF-grade monitors, 13-J-SQA-RU-37 and 13-J-SQB-RU-38. Refer to section 11.5.
| |
| : 3) Area radiation monitor 13-J-SQA-RU-31 is located on the east wall of the fuel building adjacent to the spent fuel pool. The detector will register the evolution of airborne radioactivity from the pool within 1/4 second.
| |
| : 4) 13-J-SQA-RU-31 is ESF grade. The redundant ESF monitor has two channels, low range and high range (13-J-SQB-RU-145 and 13-J-SQB-RU-146). It is located just below the roof level and samples the exhaust downstream of the dampers. The June 2005 15A-7 Revision 13
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| PVNGS UPDATED FSAR sample transit time is less than 10 seconds between duct and monitor.
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| QUESTION 15A.11 (NRC Question 440.32) (15.0)
| |
| Expand table 15.0-6, the list of single failure considered in transient and accident analyses, to include the following:
| |
| : 1. One primary safety valve stuck closed
| |
| : 2. One secondary safety valve fail to open or fail to close
| |
| : 3. Loss of offsite power
| |
| : 4. failure of one diesel to operate (for the events with loss of offsite power being treated as a consequential result of the event).
| |
| : 5. failure to achieve fast transfer RESPONSE: The response was provided on the CESSAR docket.
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| See CESSAR FSAR Responses to NRC Questions.
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| NOTE: Subsequent to the docketed response, it was determined that the single failure of a fast bus transfer was no longer a limiting single failure for any of the UFSAR Chapter 15 accident analyses. As noted in UFSAR Section 15.0.2.4, this single failure is bounded by the postulated loss of offsite power following a turbine trip.
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| QUESTION 15A.12 (NRC Question 440.33) (15.0)
| |
| For all analyses of transients with concurrent single failures, provide a reference to the sensitivity study which shows that the failure selected is the worst case single failure.
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| RESPONSE: The response was provided on the CESSAR docket.
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| See CESSAR FSAR Responses to NRC Questions.
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| June 2005 15A-8 Revision 13
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| PVNGS UPDATED FSAR QUESTION 15A.13 (NRC Question 440.34) (15.0)
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| Confirm that during the preoperational or startup test phase you intend to verify the valve discharge rates and response times (such as opening and closing times for main feedwater, auxiliary feedwater, turbine and main steam isolation valves, and steam generator and pressurizer relief and safety valves) to show that they have been conservatively modeled in the chapter 15 analyses.
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| RESPONSE: PVNGS intends to verify response times during a preoperational test to show that they have been conservatively modeled in chapter 15 analyses as described in CESSAR chapter 14 for valves within the CESSAR scope, and in PVNGS FSAR chapter 14 for valves outside the CESSAR scope.
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| QUESTION 15A.14 (NRC Question 440.35) (15.0)
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| The method that you have used for calculating the amount of failed fuel after an accident has not been approved. It is our position that fuel failures be recalculated using the criteria that any fuel rod which has a CE-1 DNBR less than the minimum DNBR value determined in section 4.4 fails. Radiological consequences should be calculated accordingly.
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| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
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| QUESTION 15A.15 (NRC Question 440.36) (15.0)
| |
| Verify that for each transient analyzed in chapter 15, if operator action is not discussed then no operator action is required. In particular, consider events in which the ECCS is June 2005 15A-9 Revision 13
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| | |
| PVNGS UPDATED FSAR actuated or RCP trip would be required based on present procedures.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
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| QUESTION 15A.16 (NRC Question 440.37) (15.0)
| |
| For each accident, discuss nonsafety grade equipment which was assumed to operate and could result in the transient becoming more severe or verify that no nonsafety grade equipment operating would produce a more severe transient. For example, the pressurizer heaters being energized for a transient resulting in high RCS pressures could tend to worsen the effects of the transient. Likewise, pressurizer spray could be detrimental for a transient resulting in low RCS pressure.
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| RESPONSE: The response was provided on the CESSAR docket.
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| See CESSAR FSAR Responses to NRC Questions.
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| QUESTION 15A.17 DELETED QUESTION 15A.18 (NRC Question 440.39) (15.0)
| |
| One of the key parameters in LOCA analyses is peak clad temperature. For non-LOCA transients, minimum DNBR (departure from nucleate boiling ratio) is of primary importance. For those transients analyzed in chapter 15 of the FSAR, provide graphical output of the DNBR as a function of time.
| |
| RESPONSE: The response was provided on the CESSAR docket.
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| See CESSAR FSAR Responses to NRC Questions.
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| June 2005 15A-10 Revision 13
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| PVNGS UPDATED FSAR QUESTION 15A.19 (NRC Question 440.40) (15.0)
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| As part of the CESEC review, the NRC intends to perform audit evaluations of feedwater line breaks, steam line breaks, and large- and small-break LOCAs (as part of the FSAR and TMI Action Plan Item II.K.3.30 and II.K.3.31 reviews). In order to perform these audits, we require the following data, as outlined in the "PWR Information Request Package."
| |
| RESPONSE: The response will be provided on the CESSAR docket.
| |
| QUESTION 15A.20 (NRC Question 440.41) (15.0)
| |
| The current CESEC model does not properly account for steam formation in the reactor vessel. Therefore, for all events in which (a) the pressurizer is calculated to drain into the hot leg, or (b) the system pressure drops to the saturation pressure of the hottest fluid in the system during normal operation, we require the applicant to reanalyze these events with an acceptable model or otherwise justify the acceptability of Palo Verde chapter 15 analyses conclusions performed with CESEC.
| |
| RESPONSE: The response was provided in CESSAR Amendment 6.
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| QUESTION 15A.21 (NRC Question 440.42) (15B)
| |
| Figure 15B-19 shows the primary system pressure exceeding 110%
| |
| of the design pressure. This figure also indicates a substantial pressure differential between the pressurizer and reactor vessel. The standard review plans typically limit the pressurization of the RCS to 110% of the design pressure.
| |
| However, the ASME pressure vessel code permits exceeding the 110% limit to approximately 120% for very low probability June 2005 15A-11 Revision 13
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| | |
| PVNGS UPDATED FSAR events. The NRC will accept the limiting pressurization transient (i.e., feedwater line break) as calculated for System 80 if we can be assured that the analysis performed is conservative and that a small break in the feedwater line is a very low probability event.
| |
| As such, we request the following information be provided:
| |
| (1) Verification of CESEC to predict pressurization transients. This should include the developed pressure differential across the pressurizer surge line.
| |
| (2) Demonstrate that the probability of a small break in the feedwater system is not significantly more probable than the large break. Include the consideration of ancillary line breaks.
| |
| (3) Section 15B.3 references a sensitivity study for RCS overpressurization transient to plant initial conditions.
| |
| Provide the results to this study in graphical form.
| |
| Specifically, include DNBR and pressure as a function of time.
| |
| (4) It is expected that increasing the break area for a feedwater line break would increase the degree of primary system pressurization. A larger break area should result in an earlier loss of heat sink and corresponding higher decay heat for system pressurization. Figure 15B-1 indicates that the limiting feedwater line break is not a double-ended guillotine break (1.4 square feet), but a 0.2-square foot break. Provide greater details as to why this occurs. Is this behavior considered realistic or a consequence of a modeling assumption? Provide additional graphical explanations, including heat transfer coefficient, heat flow, secondary side inventory, all June 2005 15A-12 Revision 13
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| | |
| PVNGS UPDATED FSAR secondary side flow rates, and any additional data required to demonstrate the reasons for the 0.2-square foot break begin the limiting break size.
| |
| (5) Figure 15B-10 provides the relationship between the maximum RCS pressure to initial steam generator inventory.
| |
| Provide additional information which explains in detail functional behavior of this curve.
| |
| Page 15B-5 states: "...the initial RCS pressure can be adjusted to provide simultaneous reactor trip signals from high pressurizer pressure and low water level in the intact steam generator and hence the plateau of maximum RCS pressure." Provide greater details of the analyses and assumptions made in order to achieve coincident trip signals from the pressurizer and SG.
| |
| (6) For figure 15B-11 (and page 15B-6), how does raising the degree of feedwater subcooling increase the maximum RCS pressure? It would appear that raising the degree of subcooling would result in a larger heat sink, and, therefore, a lower peak pressure.
| |
| (7) What decay heat model does CESEC use? Does this model assume infinite irradiation?
| |
| (8) Provide details of the core and steam generator heat transfer models used in CESEC.
| |
| (9) Utilizing a one-node representation of the steam generator secondary side, how is the low liquid level trip analyzed?
| |
| (10) Provide verification of the CESEC pressurizer model for pressurization transients (resulting in the opening of a safety valve or PORT) with data from experiments and operating plant transients. Of interest is level and June 2005 15A-13 Revision 13
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| | |
| PVNGS UPDATED FSAR pressure as a function of time. Document the assumptions made in analyzing these tests.
| |
| (11) Document the sensitivity of a feedwater line break with and without loss of offsite power.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions and CESSAR Amendment 6.
| |
| QUESTION 15A.22 (NRC Question 440.43) (15B)
| |
| For the feedwater line break analysis, provide the pressurizer liquid and mixture level as a function of time.
| |
| Provide detailed plots for the following parameters during the initial 50 seconds of the transient:
| |
| : 1. Pressurizer pressure
| |
| : 2. Surge line flow
| |
| : 3. Pressurizer mixture level
| |
| : 4. Pressurizer safety valve flow and quality RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.23 (NRC Question 440.44) (15.0)
| |
| We require additional information regarding the steam generator behavior during a feedwater line break. Provide the steam generator secondary side coolant inventory, mixture level, heat transfer coefficients, energy removed by each steam generator (Btu) and secondary side flow as a function of time.
| |
| It is our understanding that the limiting heat transfer modeling technique utilized in CESEC assumes an approximately June 2005 15A-14 Revision 13
| |
| | |
| PVNGS UPDATED FSAR constant heat transfer coefficient between the primary and secondary systems until all the liquid mass in the secondary system is depleted (i.e., WM = 0). It is not clear why the limiting modeling technique was not the case where the heat transfer was degraded as the secondary side inventory began uncovering the tubes. Please explain.
| |
| Discuss differences in the steam generator secondary heat transfer modeling between a feedwater line break and a steam line break.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.24 DELETED QUESTION 15A.25 (NRC Question 440.46) (15.0)
| |
| Accidents resulting in containment isolation also isolate the component cooling water to the reactor coolant pumps. This can potentially lead to RCP seal damage which may result in a LOCA.
| |
| Address the time available for the operators to restore the coolant to the seals. Has consideration been given to not isolating component cooling water to the RCP seals on containment isolation? If pump seal integrity cannot be maintained, evaluate the consequential failure of the pump seals for the limiting accident.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| June 2005 15A-15 Revision 13
| |
| | |
| PVNGS UPDATED FSAR QUESTION 15A.26 DELETED QUESTION 15A.27 (NRC Question 440.48) (15.0)
| |
| Provide a list of transients which result in opening of the pressurizer safety valves.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.28 (NRC Question 440.49) (15.0)
| |
| The staff has been informed that the CESEC-III computer program is best suited to analyze transients which void the upper head of the reactor vessel. As such, we request that the following information be provided:
| |
| (1) Documentation of the CESEC-III code. As part of the documentation, address the differences between the different versions of CESEC (I, II, and III).
| |
| (2) Provide comparative analyses with the different versions of the CESEC programs (used for licensing) to demonstrate the adequacy of previous analyses.
| |
| (3) Provide verification of CESEC-III against plant and experimental data for pressurization and depressurization transients (such as the AN0-2 experiments and the St. Lucie I cooldown experience).
| |
| (4) For those transients which result in primary system voiding, provide graphical output of the upper head mixture level as a function of time. Discuss operator actions/guidelines for detecting and mitigating primary system void formation.
| |
| June 2005 15A-16 Revision 13
| |
| | |
| PVNGS UPDATED FSAR (5) Show, by analysis or otherwise, that the allowable cooling rate (for cold shutdown conditions) will not result in primary system voiding.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.29 DELETED QUESTION 15A.30 DELETED QUESTION 15A.31 DELETED QUESTION 15A.32 DELETED QUESTION 15A.33 DELETED QUESTION 15A.34 (NRC Question 440.55) (15.6)
| |
| For small-break LOCAs, containment isolation may occur. It is our understanding that component cooling water to the RCP seals will be isolated upon containment isolation. Demonstrate that the RCP seals will remain intact and maintain the pressure boundary for the duration of the accident. Address expected RCP operation. If seal integrity cannot be maintained, seal failure must be assumed. Discuss the maximum seal leakage rates based on operating experience. If the consequences of seal failure are assumed to be covered by the analyzed break spectrum, justify the differences in the break locations from the locations analyzed.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| June 2005 15A-17 Revision 13
| |
| | |
| PVNGS UPDATED FSAR QUESTION 15A.35 DELETED QUESTION 15A.36 (NRC Question 440.58) (15.6.3)
| |
| The analysis for a steam generator tube rupture does not address tube leakage in the unaffected steam generator.
| |
| Provide an interface requirement for the allowable steam generator tube leakage and reference the technical specification limit. Confirm the analyses were performed using this allowable limit or provide justification why this leakage term can be excluded from the analyses.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.37 (NRC Question 440.59) (15.6.3)
| |
| The analysis for a steam generator tube rupture is for a double-ended rupture. Provide the analyses used to determine that this is the limiting ease. If a partial area break is considered, such that the steam generator relief valves open at a longer time into the transient is more primary coolant leaked to the secondary and out the SRVs, resulting in an increased dose rate.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.38 (NRC Question 440.60) (15.6.3)
| |
| SRP 15.6.3 acceptance criteria requires that this event be analyzed with a concurrent loss of offsite power. Provide an analysis for the limiting case which includes a concurrent loss of offsite power.
| |
| June 2005 15A-18 Revision 13
| |
| | |
| PVNGS UPDATED FSAR RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.39 (NRC Question 440.61) (15.6.3)
| |
| For the SGTR event, what prevents steam from the affected steam generator being used to drive the steam-driven auxiliary feedwater pump and exhausted to the environment? If operator action is required, confirm that no credit for operator action was given for 30 minutes, consider with your assumption for isolation of the affected steam generator. If credit was given for operator action in less than 30 minutes, provide justification why this credit can be given, or reanalyze the event assuming steam from the faulted steam generator is used to drive the steam-driven AFW pump and is exhausted to the environment.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.40 (NRC Question 440.62) (15.6.3)
| |
| Provide a description of the CESEC model used to model the CVCS from the reactor coolant system to the break point. Include a description of the environmental conditions at the break point (pressure, enthalpy, break flow model used).
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.41 (NRC Question 440.63) (15.6.3)
| |
| Discuss the single failure assumed for these analyses. What analyses/evaluations were performed to justify that the single failures chosen were the most limiting?
| |
| June 2005 15A-19 Revision 13
| |
| | |
| PVNGS UPDATED FSAR RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.42 (NRC Question 440.64) (15.0)
| |
| In this section, you have selected the turbine trip without a single failure as the limiting reactor coolant system pressure and the limiting radiological release event for the moderate frequent event category in the decreased heat removal by secondary system group. However, these limiting cases were not selected by a qualitative comparison of similar initiating events specified in SRP 15.2.1 through SRP 15.2.7 (e.g., loss of external load, turbine trip, loss of condenser vacuum, steam pressure regulator failure, loss of normal AC power and loss of normal feedwater flow). Provide a qualitative analysis in the FSAR for each of the initiating events in the same group per the SRP, and identify the limiting cases for the group.
| |
| Provide a detail quantitative analysis for each of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.43 (NRC Question 440.65) (15.2)
| |
| In this section, you have provided the loss of condenser vacuum with a fast transfer failure and technical specification steam generator tube leakage as the limiting RCS pressure and the limiting radiological release event for the limiting fault event category in the decreased heat removal by secondary system group. Although, these limiting cases may be the candidates for the limiting cases for the infrequent event category in the group, they were not selected by a qualitative June 2005 15A-20 Revision 13
| |
| | |
| PVNGS UPDATED FSAR comparison of similar initiating events plus a single failure specified in SRP 15.2.1 through 15.2.7. Provide a qualitative analysis in the FSAR for each of the initiating event plus a single failure in the same group per the SRP, and identify the limiting cases for the group. Provide a detailed quantitative analysis for each of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release. Confirm that the results of the analyses meet the acceptance criteria for these events per SRP 15.2.1.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| NOTE: Subsequent to the docketed response, it was determined that the single failure of a fast bus transfer was no longer a limiting single failure for any of the UFSAR Chapter 15 accident analyses. As noted in UFSAR Section 15.0.2.4, this single failure is bounded by the postulated loss of offsite power following a turbine trip.
| |
| QUESTION 15A.44 (NRC Question 440.66) (15A)
| |
| Provide tabulations of the sequence of events, disposition of normally operating systems, utilization of safety systems, and a transient curve of primary system pressure for the total loss of primary coolant flow event. Also provide an analysis of the total loss of primary coolant flow with a single failure event.
| |
| Confirm that the results of these analyses meet the acceptance criteria for these events per SRP 15.3.1.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| June 2005 15A-21 Revision 13
| |
| | |
| PVNGS UPDATED FSAR QUESTION 15A.45 (NRC Question 440.67) (15.3)
| |
| In subsection 15.3.5 you have provided the single reactor coolant pump shaft seizure with loss of offsite power following turbine trip and with technical specification tube leakage as the limiting RCS pressure and radiological release event for the limited fault event category. This postulated event is classified as an infrequent event per SRP 15.3.3. Confirm that the results of the analysis meet the acceptance criteria for these events per SRP 15.3.3, using the criteria stated in Question 440.35 to calculate the amount of failed fuel in this event. State the amount of failed fuel in the results of the analysis. Radiological consequences should be calculated accordingly.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions and CESSAR Amendment 6.
| |
| QUESTION 15A.46 (NRC Question 440.68) (15.0)
| |
| Provide results of an analysis of the reactor coolant pump shaft break as required by SRP 15.3.4 for staff review. The event should consider loss of offsite power following turbine trip and with technical specification steam generator tube leakage. The criteria stated in Question 440.35 should be used for the calculation of the amount of failed fuel for this event. State the amount of failed fuel in the results of the analysis. Radiological consequences should be calculated accordingly. Confirm that the results of the analysis meet the acceptance criteria for these events per SRP 15.3.4 which classifies this event as an infrequent event.
| |
| June 2005 15A-22 Revision 13
| |
| | |
| PVNGS UPDATED FSAR RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.47 (NRC Question 440.69) (15.5)
| |
| In this section, you have provided the pressurizer level control system malfunction (PLCSM) with a fast transfer failure and the PLCSM with a loss of offsite power at turbine trip with Technical Specification steam generator tube leakage as the limiting RCS pressure and radiological release event for the limiting fault event category in the increase in reactor coolant system inventory group. However these limiting cases were not selected by a qualitative comparison of similar initiating events plus a single failure specified in SRP 15.5.1 (e.g., inadvertent operation of high pressure ECCS or a malfunction of the CVCS). Provide a qualitative analysis in the FSAR for each of the initiating events (with and without a single active failure) in the same group per the SRP, and identify the limiting cases for the group. Provide a detailed quantitative analysis for each of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release. Confirm that the results of the analyses meet the acceptance criteria for these events per SRP 15.5.1.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions NOTE: Subsequent to the docketed response, it was determined that the single failure of a fast bus transfer was no longer a limiting single failure for any of the UFSAR Chapter 15 accident analyses. As noted in UFSAR Section 15.0.2.4, this single failure is bounded by the postulated loss of offsite power following a turbine trip.
| |
| June 2005 15A-23 Revision 13
| |
| | |
| PVNGS UPDATED FSAR QUESTION 15A.48 (NRC Question 440.70) (15.0)
| |
| Provide tabulations of the sequence of events, disposition of normally operating systems, utilization of safety systems, and all necessary transient curves for the startup of an inactive reactor coolant pump event. The comparison to peak RCS pressure acceptance criteria should be included in the analysis. Also provide the results of an analysis of this event with a single failure. Confirm that the results of these analyses meet the acceptance criteria for these events per SRP 15.4.4.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.49 (NRC Question 440.71) (15D)
| |
| You have provided, in CESSAR Appendix 15D, the results of an inadvertent boron dilution event without a single failure under plant cold shutdown conditions. This information is not sufficient. You should provide results of analyses for all possible boron dilution events under various plant operational modes (e.g., refueling, startup, power operation, hot standby and cold shutdown). Also provide the results of analyses of these events with a single failure. Confirm that the results of these analyses meet the acceptance criteria for these events per SRP 15.5.1. In particular, the available times per operator action between time of alarm and time of loss of shutdown margin should be shown to meet the SRP guidelines.
| |
| The results of the analyses should be presented in the FSAR including tabulations of sequence of events, disposition of normally operating systems, utilization of safety systems, and all necessary transient curves for the events.
| |
| June 2005 15A-24 Revision 13
| |
| | |
| PVNGS UPDATED FSAR In your analysis, indicate for all modes of operation what alarms would identify to the operators that a boron dilution event was occurring. Consider the failure of the first alarm.
| |
| Provide the time interval from this alarm to when the core would go critical. If a second alarm is not provided, show that the consequences of the most limiting unmitigated boron dilution event meet the staff criteria and are acceptable.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.50 DELETED QUESTION 15A.51 (NRC Question 440.73) (15D)
| |
| Several recent LERs indicate there has been a deficiency in the inadvertent boron dilution analysis at some plants. Provide an analysis of the dilution event when the RCS is drained to the hot leg.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.52 (NRC Question 440.74) (15D)
| |
| Recently, an operating PWR experienced a boron dilution incident due to inadvertent injection of NaOH into the reactor coolant system while the reactor was in a cold shutdown condition. Discuss the potential for a boron dilution incident caused by dilution sources other than the CVCS.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| June 2005 15A-25 Revision 13
| |
| | |
| PVNGS UPDATED FSAR QUESTION 15A.53 (NRC Question 440.75) (15.6)
| |
| Discuss the transient resulting from a break of an ECCS injection line. In particular, describe the flow splitting which will occur in the event of a single failure and verify that the amount of flow actually reaching the core is consistent with the assumptions used in the analysis.
| |
| RESPONSE: The response was provided on the CESSAR docket.
| |
| See CESSAR FSAR Responses to NRC Questions.
| |
| QUESTION 15A.54 (NRC Question 440.76) (15.8)
| |
| The NRC is currently considering what actions may be necessary to reduce the probability and consequences of anticipated transients without scram (ATWS). Until such time as the Commission determines what plant modifications are necessary, we have generally concluded that pressurized water plants can continue to operate because the risk from anticipated transient without scram events in a limited time period is acceptably small. However, in order to further reduce the risk from anticipated transient without scram events during the interim period before completing the plant modifications determined by the Commission to be necessary, we have required that the following actions be taken:
| |
| : 1. Develop emergency procedures to train operators to recognize anticipated transient without scram events, including consideration of scram indicators, rod position indicators, flux monitors, pressurizer level and pressure indicators, pressurizer relief valve and safety valve indicators, and any other alarms annunciated in the control room with emphasis on alarms not processed through the electrical portion of the reactor scram system.
| |
| June 2005 15A-26 Revision 13
| |
| | |
| PVNGS UPDATED FSAR
| |
| : 2. Train operators to take actions in the event of an anticipated transient without scram, including consideration of manually scramming the reactor by using the manual scram button, prompt actuation of the auxiliary feedwater system to assure delivery to the full capacity of this system, and initiation of turbine trip. The operator should also be trained to initiate boration by actuation of the high pressure safety injection system to bring the facility to a safe shutdown condition.
| |
| Describe how you will meet the above requirements, and provide a schedule for submittal of the ATWS procedures for staff review.
| |
| RESPONSE: Procedures will be developed to cover emergencies and off-normal events. These procedures will provide sufficient guidance to ensure that correct action is taken by the operator. ATWS events will be covered in these procedures. PVNGS will provide training on ATWS events and emergency and off-normal procedures.
| |
| Sufficient information will be provided so that the operator can determine if his actions are effective.
| |
| Should the operator's actions not be effective, the procedure will contain additional action that can be taken by the operator to ensure the parameter and/or condition is restored to acceptable values.
| |
| Procedures will be available for NRC review at least 60 days prior to fuel load.
| |
| QUESTION 15A.55 (NRC Question 440.82) (15.0)
| |
| Section 15D.2.2.2 of the CESSAR System 80 FSAR states that the loss of instrument air event impact on the plant systems and June 2005 15A-27 Revision 13
| |
| | |
| PVNGS UPDATED FSAR components will be addressed in the applicant's FSAR. Discuss the loss of instrument air for Palo Verde showing that it meets the appropriate acceptance criteria for a moderate frequency event. Causes and potential systems interactions should be addressed and the loss of instrument air should be considered during all phases of reactor operation. Also, present your plans and capability for preoperational or startup tests to substantiate the analyses.
| |
| RESPONSE: The nitrogen supply system will support the required ESF air-supplied components normally supplied from instrument air system for one hour on loss of instrument air. This is accomplished by providing an automatic control valve connecting the nitrogen system to the instrument air system. Depletion of the nitrogen system will not affect any safety-related systems.
| |
| The following list shows the systems which would be affected on loss of instrument air and depletion of the nitrogen supply. Also shown is the position of the air (or nitrogen supply valve upon depletion of the nitrogen supply.
| |
| SYSTEM VALVE POSITION RCP Seal Injection Open RCS Letdown Closed CVCS Charging Closed Boric Acid Concentrator Closed Suction to RDPs Closed Gas Stripper Closed Pressurizer Sprays Closed Steam Bypass to Condenser Closed June 2005 15A-28 Revision 13
| |
| | |
| PVNGS UPDATED FSAR Main Steam Line Drains Closed Nitrogen Charging to SITs Closed Sulfuric Acid to ESPs Closed Letdown Hx Cooling Water Closed Turbine Cooling Water Open Normal Chilled Water Closed Auxiliary Steam System Closed Instrument air loss would not incapacitate any safety-related systems or equipment needed for safe shutdown. It would affect the above systems by fail-safe closing or opening (as indicated) of air-operated valves upon air failure and depletion of nitrogen supply.
| |
| (a)
| |
| QUESTION 15A.56 (SGTR Question 1) (15.6.3)
| |
| In the SGTR analysis for Palo Verde units submitted by your letter dated January 27, 1984, the acceptability of the radiological consequences is heavily dependent on the operator's action on controlling the cooldown rate. It is assumed in the analysis that the operator has to open one ADV in each steam generator at a 10.5% opening position to ensure a maximum cooldown rate of 75F. The staff notes that the ADVs have no device to limit their opening to the assumed 10.5%, and other calculations have shown that an opening of slightly less than 12% would result in exceeding the 10CFR Part 100 guideline values. Also, there are no specific limits in either the technical specifications or procedures to restrict opening of the ADV to less than the 10.5% assumed. There is only the
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated April 27, 1984.
| |
| June 2005 15A-29 Revision 13
| |
| | |
| PVNGS UPDATED FSAR maximum cooldown rate limit of 75F/hr, a value that we believe is difficult for the operator to determine during a complicated event like the SGTR. Discuss what positive measures will be taken to ensure that the assumed ADV opening position and cooldown rate will not be exceeded.
| |
| RESPONSE: The January 27, 1984, analysis of the SGTR with the loss of offsite power and the failure of the stuck open ADV event assumed that, once the operator identified and pursued isolating the affected steam generator, all auxiliary feedwater flow ceased to that generator. This approach was chosen to maximize radiological consequences pursuant to direction from the Regulatory Staff. This arbitrary restriction results in the hypothetical radiological consequences being very sensitive to valve opening position because of the subsequent tube uncovery.
| |
| Further review of this nondesign basis accident calculation has indicated that it is unnecessarily conservative to assume tube uncovery. Accordingly, the PVNGS Emergency Operating Procedures will include direction to feed the affected steam generator in order to keep the tubes covered and maintain the iodine partition coefficient. This is not a deviation from the CE Emergency Procedure Guidelines (CEN-152). Rather, it is an additional consideration to be used to mitigate the consequences of a SGTR, and provides substantial benefits for the instances where the ruptured steam generator cannot be isolated from the atmosphere (e.g., stuck open ADV). This multiple failure event, SGTR and a fully stuck open ADV, was not contemplated by CEN-152, just as it is not considered by the NRC's Standard Review Plan.
| |
| June 2005 15A-30 Revision 13
| |
| | |
| PVNGS UPDATED FSAR This modification to the PVNGS Procedure will be incorporated before fuel load of PVNGS Unit 1. Training of the operators will commence soon after approval of the procedure modification, and will require approximately 3 months to train all of the Unit 1 shifts. This training should be complete by initial criticality. Training will include simulator time and will emphasize the reduction of offsite releases and the potential for overfill of the affected steam generator.
| |
| Including this additional procedure into the analysis leads to a revised 0 to 2 hour thyroid dose of 200 rem including a fully (100%) open atmospheric dump valve and a preexisting iodine spike. This is the highest dose (refer to table 15A-1 for the complete dose results) and is well within Part 100 criteria.
| |
| It should be noted that a stuck, fully open ADV is not considered a credible event as there is no single failure that can cause the valve to run full open and stay there.
| |
| Refer also to the response provided for Question 15A.58.
| |
| Table 15A-1 RADIOLOGICAL CONSEQUENCES OF THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV Offsite Doses, Rem Location GIS PIS
| |
| : 1. Exclusion area boundary 40 200 0-2 hours thyroid
| |
| : 2. Low population zone outer 20 41 boundary 0-8 hours thyroid June 2005 15A-31 Revision 13
| |
| | |
| PVNGS UPDATED FSAR (a)
| |
| QUESTION 15A.57 (SGTR Question 2) (15.6.3)
| |
| The SGTR analysis also assumes a cooldown rate of 30F per hour at 30 minutes after the attempted closing of the affected steam generator ADV. Describe how the operator monitors the plant conditions to prevent the cooldown rate exceeding 30F/hr during this time period.
| |
| RESPONSE: The long-term cooldown rate of 30F per hour was chosen for the January 27, 1984, analysis so that shutdown cooling conditions were reached 8 hours after the event.
| |
| This maximized the 0-8 hour dose. A more rapid, or a slower, cooldown would release less steam from the ruptured steam generator.
| |
| This assumption is not used in the revised analysis presented in the response to Question 15A.56.
| |
| (a)
| |
| QUESTION 15A.58 (SGTR Question 3) (15.6.3)
| |
| Since the ADVs at Palo Verde do not have upstream block valves, there would be virtually no way of isolating a stuck open ADV.
| |
| The staff believes from an overall plant safety standpoint, Palo Verde should install block valves upsteam of the ADVs, per the interface requirement stated in the CESSAR System 80 FSAR.
| |
| Discuss your technical justification for a lack of the block valves, especially in light of industry experience suggesting that stuck open steam system valves are not an uncommon occurrence.
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated April 27, 1984.
| |
| June 2005 15A-32 Revision 13
| |
| | |
| PVNGS UPDATED FSAR Additionally, the Palo Verde SGTR analysis should either assume an ADV stuck in the full open position, or the applicant should provide positive assurance that the ADV cannot be opened beyond the assumed 10.5%.
| |
| RESPONSE: The analysis presented in the response to Question 15A.56 assumes an ADV stuck in the full open position. Nevertheless, this is not considered a credible single failure.
| |
| The PVNGS ADVs are air-operated hydraulic valves. The valves are spring loaded to fail closed on loss of air.
| |
| Additionally, they may be closed by air or by an integral handwheel, if necessary. In order for the valve to open, an air supply must be provided. Two parallel sets of fail closed, three-way solenoid valves (four total) provide the air supply. In the closed position, the valves isolate the air supply and bleed air off of the ADV. The solenoid valves are powered by two channels of essential dc power.
| |
| Each set of solenoid valves is controllable from the Control Room. Closure of any one set of valves is sufficient to terminate the air supply and close the ADV.
| |
| The control schematic is provided as figure 15A-1.
| |
| Should all four solenoid valves fail by remaining energized, the operator can regulate the air supply by using the valve positioner and controller. These will also be able to close the ADV. In short, for the ADV to open and remain open, there must be six failures involving two channels of dc power. This is considered an incredible event.
| |
| Mechanical binding of the valve was also considered. In order to remain open, the valve would need to seize up so June 2005 15A-33 Revision 13
| |
| | |
| PVNGS UPDATED FSAR firmly that neither air pressure, spring nor manual handwheel operation would be able to close the valve.
| |
| This would result in the valve sticking at the operating range. As noted in the January 27, 1984, analysis, offsite dose exposure is less than 150 rem even with the tubes uncovered. Under the revised analysis of Question 15A.56, with the tubes covered, the dose is 41 rem (table 15A-1).
| |
| The analysis of the steam generator tube rupture with a loss of offsite power and a fully stuck open atmospheric dump valve (ADV) follows:
| |
| Identification of Event and Causes This transient is similar to that described in CESSAR Appendix 15D. It assumes that the plant is challenged by a steam generator tube rupture that includes additional events and failures beyond those postulated by the NRC Standard Review Plan 15.6.3. In addition to the conservative assumptions of the SRP (loss of offsite power, accident meteorology, iodine spiking, etc.), this analysis postulates that the operators open an ADV on the affected steam generator and that it both runs to the full open position and that it sticks full open for the duration of the transient. The ADV is presumed to remain open despite the availability of two redundant and independent safety grade valve control systems and a manual handwheel to close the ADV.
| |
| Sequence of Events and Systems Operation Table 15A-2 presents a chronological list of events that are assumed to occur during the steam generator tube rupture event with a loss of offsite power from the time June 2005 15A-34 Revision 13
| |
| | |
| PVNGS UPDATED FSAR of the double-ended rupture of a steam generator U-tube to the attainment of shutdown cooling entry conditions.
| |
| The C-E Emergency Procedure Guidelines, CEN-152, contain guidance to the operator for controlling a steam generator tube rupture. Recognizing that the coincident occurrence of the limiting (conservative) assumptions of the SRP is unlikely, CEN-152 proposes that, should offsite power and the steam bypass control system be unavailable, the operator opens an ADV on each steam generator (ruptured or not) in order to preclude a challenge to the main steam safety valves (MSSVs). This action presupposes that the ADVs are reliable and can be closed after the RCS is cooled to a temperature which precludes a challenge to the MSSVs. It also presupposes that the MSSVs have not opened. However, due to the coincident conservative assumptions of the SRP, the MSSVs open early in the transient. Furthermore, Palo Verde procedures are oriented towards diagnosing the event and stabilizing the plant prior to initiating cooldown.
| |
| Because of the PVNGS emphasis on proper diagnosis prior to operator action, it is unlikely that the operator would open the ADV once the diagnosis indicated an SGTR.
| |
| Nevertheless, this scenario assumes that once an operator diagnoses a SGTR, he opens an ADV (as suggested in CEN-152). To recover from this scenario, the plant specific Palo Verde Steam Generator Tube Rupture Procedure includes direction to the operator to maintain steam generator level such that the steam generator tubes are covered.
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| June 2005 15A-35 Revision 13
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| | |
| PVNGS UPDATED FSAR Table 15A-2 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 1 of 4)
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| Time Setpoint Success Path (Sec) Event or Value or Comment 0.0 Tube rupture occurs -- --
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| 40 Third charging pump -0.75 Primary system started, feet below integrity program level 40 Letdown control valve -0.75 Primary system throttled back to integrity minimum flow, feet below prgram level 47 CPC hot leg saturation -- Reactivity control trip signal generated 47.15 Trip breakers open -- Reactivity control 48 Turbine trip -- Secondary system integrity 51 Loss of offsite power -- --
| |
| 52 LH main steam safety 1265 Secondary system valves open, psia integrity 52 RH main steam safety 1265 Secondary system valves open, psia integrity 56 Maximum steam generator 1330 --
| |
| pressures both steam generators, psia 121 Steam generator water 25 Secondary system level reaches auxiliary integrity feedwater actuation signal (AFAS) analysis setpoint in unaffected generator, percent wide range level June 2005 15A-36 Revision 13
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| | |
| PVNGS UPDATED FSAR Table 15A-2 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 2 of 4)
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| Time Setpoint Success Path (Sec) Event or Value or Comment 122 AFAS generated -- --
| |
| 131 Steam generator water 25 Secondary system level reaches AFAS integrity analysis setpoint in the affected generator, percent wide range level 132 AFAS generated -- --
| |
| 167.0 Auxiliary feedwater -- Secondary system initiated to unaffected integrity steam gnerator 177.0 Auxiliary feedwater -- Secondary system initiated to affected integrity steam generator 460 Operator initiates plant -- Removal heat cooldown by opening one removal ADV on each SG - ADV of the affected SG instantaneously opens fully 484 Pressurizer empties -- --
| |
| 513 MSIS actuation, 919 Secondary system secondary pressure, integrity psia 535 Automated isolation of 185 Secondary system AFW to affected SG, integrity P SGs, psi 581 Pressurizer pressure 1578 Reactivity control reaches safety injection actuation signal (SIAS) analaysis setpoint, psia June 2005 15A-37 Revision 13
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| | |
| PVNGS UPDATED FSAR Table 15A-2 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 3 of 4)
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| Time Setpoint Success Path (Sec) Event or Value or Comment 581 Safety injection -- --
| |
| actuation signal generated 581 Safety injection flow -- Reactivity control initiated 655 Operator overrides the -- --
| |
| AFW isolation signal and starts feeding the affected SG with AFW 775 Operator takes manual -- --
| |
| control of the AFW system, feeds affected SG with both AFW pumps 895 Operator shuts the ADV -- --
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| of the unaffected steam generator 1015 Operator initiates -- --
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| auxiliary spray to the pressurizer 1385 Level in the affected SG 71.5 --
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| above the top of U-tubes, percent wide range 2040 Pressurizer level, 50 --
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| percent 2400 Operator controls HPSI 20 --
| |
| flow, backup pres-surizer heater output, and auxiliary spray flow to control RCS pressure and sub-cooling, F June 2005 15A-38 Revision 13
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| | |
| PVNGS UPDATED FSAR Table 15A-2 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 4 of 4)
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| Time Setpoint Success Path (Sec) Event or Value or Comment 28,800 Shutdown cooling entry 400/350 --
| |
| conditions are reached RCS pressure, psia/
| |
| temperature, F 28,800 Operator activates -- --
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| shutdown cooling system June 2011 15A-39 Revision 16
| |
| | |
| PVNGS UPDATED FSAR As is evident, the multiple failure scenario being postulated is not internally consistent. However, for analytical purposes, the sequence of events described in table 15A-2 serves to bound the scenario by projecting the adverse operator action (full opening of the ADV on the ruptured generator) and the nonmechanistic ADV failure to occur at the earliest possible time consistent with ANSI Standard N660. Subsequent beneficial operator actions are delayed by times that are also consistent with the ANSI standard.
| |
| Accordingly, an analytical model was developed from the bounding assumptions. The model features include:
| |
| * secondary releases from both the MSSVs and ADVs
| |
| * early operator action to open the ADVs
| |
| * one potential series of operator actions to cover the S/G tubes
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| * time delays for operator recovery action
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| * delay in reaching shutdown cooling (chosen to maximize 8-hour steam release)
| |
| The disposition of normally operating systems for the SGTR event are the same as those presented in Table 15D-2 of CESSAR. The utilization of safety systems during the event is the same as that presented in Table 15D-3 of CESSAR.
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| The major assumptions regarding systems operation during the event are summarized below.
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| June 2005 15A-40 Revision 13
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| | |
| PVNGS UPDATED FSAR
| |
| : 1) The auxiliary feedwater system (AFWS) is activated at 25% level wide range and shuts off at 30% level wide range prior to operator action.
| |
| : 2) Two AFW pumps are assumed to be available to supply feedwater to either steam generator. No credit is taken or the third 1E powered AFW train. An AFW flowrate of 650 gallons per minute per pump is assumed to be delivered to the steam generators at a SG pressure of 1270 psia.
| |
| : 3) The response times of ADVs, MSIVs, AFW control valves, and AFW flow isolation valves are assumed to be instantaneous.
| |
| : 4) After the loss of offsite power subsequent to reactor trip, no credit is taken for charging. One charging pump is assumed available for auxiliary spray in the pressurizer.
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| : 5) Two high pressure safety injection (HPSI) pumps are assumed to be available subsequent to the generation of a safety injection actuation signal.
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| RADIOLOGICAL CONSEQUENCES The physical model is the same as that discussed in CESSAR Section 15D.3.2 except that the ADV of the affected steam generator opens fully. In order to reduce the radiological releases, the operator takes appropriate actions to recover the U-tubes of the affected steam generator. Actions assumed in this analysis included overriding the automatic isolation of AFW flow to the affected steam generator and diverting the flow of both AFW pumps of the affected steam generator.
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| June 2005 15A-41 Revision 13
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| | |
| PVNGS UPDATED FSAR The assumptions and conditions employed for the evaluation of radiological releases are the same as those discussed in CESSAR Section 15D.3.2.B with the exceptions of assumptions 7, 9, and 10. They are:
| |
| : 7. During the period when the water level in the affected steam generator is above the top of the U-tubes, that portion of the leaking primary fluid which flashes to steam upon entering the steam generator is assumed to be released to the atmosphere with a decontamination factor (DF) of 1.0. The portion of the leaked fluid that does not flash, mixes with the liquid in the steam generator and is released to the atmosphere with a DF of 100. During the period when the water level is below the top of the U-tubes, it is assumed that all the leaking primary fluid escapes to the atmosphere with a DF of 1.0. No credit is taken for the presence of steam separators and dryers which would retain a part of the escaping primary liquid in the steam generator.
| |
| : 9. The 0-2 hour and 2-8 hour primary-to-secondary leakage through the rupture are 285,400 lbm and 516,700 lbm, respectively.
| |
| : 10. The atmospheric dispersion factors employed in the
| |
| -4 3 analyses are: 3.1 x 10 sec/m for the exclusion area
| |
| -5 3 boundary and 5.1 x 10 sec/m for the low population zone.
| |
| The mathematical model is as described in CESSAR Section 15D.3.2.C.
| |
| The 2-hour exclusion area boundary (EAB) and the 8-hour low population zone (LPZ) boundary inhalation doses for June 2005 15A-42 Revision 13
| |
| | |
| PVNGS UPDATED FSAR both the GIS and the PIS are presented in table 15A-1.
| |
| The calculated EAB and LPZ doses are well within the acceptance criteria.
| |
| CONCLUSIONS The dynamic behavior of important NSSS parameters during the event is presented in figures 15A-2 through 15A-15.
| |
| The radiological releases calculated for the SGTR event with a loss of offsite power and a fully stuck open ADV are well within the 10CFR100 guidelines. The RCS and secondary system pressures are well below the 110% of the design pressure limits, thus assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs, since the minimum DNBR remains above the 1.19 value throughout the duration of the event.
| |
| (a)
| |
| QUESTION 15A.59 (SGTR Question 1) (15.6.3)
| |
| In the response to Question 15A.56, it states that feeding the affected steam generator is not a deviation from CEN-152. Our position is that since the Palo Verde SG isolation strategy is different from the approved generic CEN-152 strategy, this is a deviation which should be justified.
| |
| RESPONSE: The PVNGS procedure for mitigating a steam generator tube rupture (SGTR) has been augmented to include direction to the operator to feed the affected steam generator in order to keep the tubes covered and maintain the iodine partition coefficient. APS acknowledges that this is a deviation from CEN-152, and our justification is provided below.
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated November 28, 1984.
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| June 2005 15A-43 Revision 13
| |
| | |
| PVNGS UPDATED FSAR By directing the operator to feed the affected steam generator, to cover the tubes, the iodine partition coefficient is maintained. By maintaining the iodine partition co-efficient, offsite doses will be reduced.
| |
| With this new procedural consideration, offsite doses will be reduced for any postulated SGTR. That is, a reduction in offsite doses would be realized for those scenarios which do not include a full open atmospheric dump valve failure.
| |
| This modification to the PVNGS procedure will be incorporated before fuel load of PVNGS Unit 1. Training of the operators will commence soon after approval of the procedure modification, and will require approximately 3 months to train all of the Unit 1 shifts. This training should be complete by initial criticality. Training will include simulator time and will emphasize the reduction of offsite releases and the potential for overfill of the affected steam generator.
| |
| We believe this justifies the deviation from CEN-152, for the PVNGS SGTR procedure.
| |
| As shown in figure 15A-4, the operator action to feed the affected steam generator in combination with the stuck open ADV may result in a primary system cooldown rate in excess of the Technical Specification maximum cooldown rate of 100F per hour. Although this operator action may result in exceeding a Technical Specification requirement, the action is justified by the need to maintain offsite doses as low as possible during a SGTR event.
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| June 2005 15A-44 Revision 13
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| | |
| PVNGS UPDATED FSAR (a)
| |
| QUESTION 15A.60 (SGTR Question 2) (15.6.3)
| |
| For a steam generator tube rupture, initially the secondary side of the affected steam generator will be fed by both the primary-to-secondary side leak and feedwater. This influx of water creates the potential for overfilling the steam generator. Discuss the information available to the operator to prevent overfilling the steam generator and the sensitivity of the time period from the start of the accident to the time when there could be an overfill problem, assuming the operator does not take any action to prevent overfilling. Alternately, show that the consequences of SG overfill are not significant.
| |
| RESPONSE: For a SGTR with loss of offsite power and a fully stuck open ADV, the steaming rate through the stuck open ADV is significantly higher than the primary-to-secondary side leak during the entire reported period of the transient.
| |
| Therefore, the influx of leak flow from the primary system alone will not create the potential for overfilling the steam generator. The operator, after taking manual control of the auxiliary feedwater system, first raises the affected SG level above the top of the U-tubes. Thereafter, the auxiliary feedwater flow is throttled to maintain the level above the top of the SG U-tubes at about 7l.5% wide range.
| |
| This level prevents any overfilling of the affected steam generator. The operator will continuously rely on the SG level measurements (12 Class 1E indicators per SG) for information all through the transient to keep the level below the acceptable limit of about 90% wide range in both generators. There are audio and visual alarms that actuate
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated November 28, 1984.
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| June 2005 15A-45 Revision 13
| |
| | |
| PVNGS UPDATED FSAR when it appears that the SG is being overfilled. It is these alarms and MSIS actuation which would provide overfill protection assuming no operator action. If at anytime there is a concern regarding SG overfill, the auxiliary feedwater to the affected steam generator will be temporarily terminated.
| |
| Combustion Engineering Emergency Procedure Guidelines (CEN-152) provides suggestions regarding steam generator overfill during a SGTR event mitigation. These suggestions include draining to radwaste system and steaming the generator.
| |
| (a)
| |
| QUESTION 15A.61 (SGTR Question 3) (15.6.3)
| |
| Evaluate the sensitivity of the time period the SG tubes could be uncovered to the increase in radiological consequences.
| |
| Relate this study to the amount of tube uncovery without credit for manual SG level control.
| |
| RESPONSE: In the analysis of a SGTR with loss of offsite power and a fully stuck open ADV, the first operator action taken to recover the level in the affected SG was assumed to occur 2 minutes after isolation of the auxiliary feedwater flow to the affected SG. The action consisted of overriding the auxiliary feedwater isolation signal in order to start feeding the affected steam generator again. Two minutes subsequent to this action the operator takes manual control of the auxiliary feedwater system and starts feeding the affected steam generator with both auxiliary feedwater pumps. The actions were taken to raise the level in the
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated November 28, 1984.
| |
| June 2005 15A-46 Revision 13
| |
| | |
| PVNGS UPDATED FSAR affected SG above the top of the U-tubes as quickly as allowed by the emergency operating procedure since the magnitude of the offsite radiation dose is sensitive to the duration of SG tube uncovery.
| |
| In order to limit the doses within acceptable limits the operator needs to take timely actions. The current analysis assumed the operator takes manual control of the auxiliary feedwater system approximately 5 minutes after opening the ADV on each SG or 2 minutes after overriding the auxiliary feedwater system isolation signal. Calculations performed indicate that to limit the offsite doses to 10CFR100 guidelines, the operator will need to take manual control of the auxiliary feedwater system no later than approximately 12 minutes after opening of the ADV on each generator. The time interval between overriding the auxiliary feedwater isolation signal and taking manual control of the system is again 2 minutes. Therefore, within the constraints and conservatisms inherent in the current model, the operator can delay taking manual control of the SG level by approximately 12 minutes after the opening of the ADVs, and still limit the offsite doses to 10CFR100 guidelines.
| |
| (a)
| |
| QUESTION 15A.62 (SGTR Question 4) (15.6.3)
| |
| In your discussion of the steam generator tube rupture (Appendix); it states that 460 seconds (about 7-1/2 minutes) is the earliest possible time that the operator can take an adverse action. The bases for this statement was given as
| |
| : a. Letter from G. W. Knighton, NRC, to E. E. Van Brunt, Jr.,
| |
| APS, dated November 28, 1984.
| |
| June 2005 15A-47 Revision 13
| |
| | |
| PVNGS UPDATED FSAR reference to ANSI Standard N660. Since the purpose of ANSI Standard N660 is not to determine the earliest time for operator to take "adverse" actions, our position is that inadequate bases have been provided to justify that the operator could not have opened the ADV earlier than 460 seconds. Therefore, determine the radiological consequences of the steam generator tube rupture with loss of offsite power assuming the operator opens an ADV on each steam generator at the earliest time possible that would result in the maximum radiological consequences.
| |
| RESPONSE: The largest contribution of the offsite dose during the event occurs in the time period between the opening of the ADVs and the time of recovery of the affected SG level above the U-tubes. This time period is greatly influenced by the inventory in the affected steam generator at the time that the ADVs are opened. In the current analysis the auxiliary feedwater flow to the affected SG is actuated on low SG level at about 177 seconds. Thereafter, the level is maintained in the narrow band between 25 and 30% wide range by the automatic operation of the auxiliary feedwater system. The SG level in the affected generator will be higher than 25% wide range prior to the auxiliary feedwater system actuation. Hence, opening of the ADVs at a prior time (that is, before 177 seconds) results in the inventory in the affected steam generator being higher than that calculated for the current analysis at the time the ADVs were opened. This means quicker recovery of the level in the affected steam generator since the inventory will be less depleted than for the current analysis.
| |
| Opening the ADV at an earlier time, when primary system pressure is higher, also causes increased primary-to-June 2005 15A-48 Revision 13
| |
| | |
| PVNGS UPDATED FSAR secondary flow through the postulated tube rupture. This offsets the level effects described above. Therefore, the overall impact on offsite doses is expected to be minimal.
| |
| Analysis has verified that the most limiting offsite dose (preexisting iodine spike) is increased by less than 5%.
| |
| For the offsite dose with an event generated iodine spike, analysis has verified an increase of approximately 8%. This assumes that the sequence of events between the opening of the ADVs and the operator taking manual control of the auxiliary feedwater system is the same for both cases.
| |
| Therefore, even if the operator was to open the ADVs at a time prior to that assumed in the analysis, resulting offsite doses would still be within 10CFR100 guidelines.
| |
| June 2005 15A-49 Revision 13
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| | |
| This page intentionally blank PVNGS UPDATED FSAR APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS
| |
| | |
| PVNGS UPDATED FSAR PVNGS UPDATED FSAR CONTENTS Page 15B.1 INTRODUCTION 15B-1 15B.2 ASSUMPTIONS 15B-1 15B.3 WHOLE-BODY GAMMA AND BETA SKIN DOSE 15B-2 15B.4 THYROID INHALATION DOSE 15B-5 15B.5 CONTROL ROOM DOSE 15B-7 15B.6 ACTIVITY RELEASE MODELS 15B-10 15B.6.1 GENERAL EQUATION 15B-10 15B.6.2 THE MODEL FOR CONTAINMENT LEAKAGE 15B-12 15B.6.3 THE MODEL FOR RECIRCULATION LOOP LEAKAGE 15B-16 15B.6.4 THE MODEL FOR THE FUEL HANDLING ACCIDENT IN THE FUEL BUILDING WITH ESF SAFEGUARDS ACTUATION 15B-17 15B.6.5 OTHER ACCIDENT MODELS 15B-18 15B.7 REFERENCES 15B-20 June 2001 15B-i Revision 11
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| | |
| PVNGS UPDATED FSAR TABLES Page 15B-1 Radionuclide Parameters 15B-2 15B-2 Whole-Body Gamma and Beta Skin Dose Conversion Factors 15B-4 15B-3 Breathing Rates 15B-6 15B-4 Iodine Dose Conversion Factors 15B-6 15B-5 Atmospheric Dispersion Factors 15B-9 15B-6 Control Room Essential Ventilation System Parameters 15B-11 June 2001 15B-ii Revision 11
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| | |
| PVNGS UPDATED FSAR FIGURES 15B-1 Containment Leakage Dose Model 15B-2 ESF Room Leakage Dose Models 15B-3 Other Accident Dose Model June 2001 15B-iii Revision 11
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| | |
| PVNGS UPDATED FSAR PVNGS UPDATED FSAR APPENDIX 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION This appendix identifies the models used to calculate control room and offsite radiological doses, not calculated in CESSAR, that would result from releases of radioactivity due to various postulated accidents.
| |
| 15B.2 ASSUMPTIONS The following assumptions are basic to the model for the whole-body dose due to immersion in a cloud of radioactivity and to the model for the thyroid dose due to inhalation of radio-activity:
| |
| A. All radioactive releases are treated as ground level releases regardless of the point of discharge.
| |
| B. The dose receptor is a standard man, as defined by the International Commission on Radiological Protection (ICRP), (reference 1).
| |
| C. No credit is taken for cloud depletion by ground deposition and radioactive decay during transport to the exclusion area boundary (EAB) or the outer boundary of the low population zone (LPZ).
| |
| D. Radionuclide data, including decay constants and decay energies presented in table 15B-1, are taken from references 2 through 6.
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| June 2001 15B-1 Revision 11
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| | |
| PVNGS UPDATED FSAR APPENDIX 15B Table 15B-1 RADIONUCLIDE PARAMETERS Decay Average Nuclide Constant MeV/Disintegration MeV/Disintegration (sec-1) (gamma) (beta)
| |
| I-131 9.97E-7 0.381 0.194 I-132 8.42E-5 2.333 0.519 I-133 9.21E-6 0.608 0.403 I-134 2.2E-4 2.529 0.558 I-135 2.91E-5 1.635 0.475 Kr-83m 1.05E-4 0.002 0.037 Kr-85m 4.29E-5 0.159 0.253 Kr-85 2.05E-9 0.002 0.251 Kr-87 1.51E-4 0.793 1.324 Kr-88 6.78E-5 1.950 0.375 Kr-89 3.63E-3 1.712 1.001 Xe-131m 6.81E-7 0.02 0.143 Xe-133m 3.66E-6 0.0416 0.190 Xe-133 1.52E-6 0.0454 0.135 Xe-135m 7.38E-4 0.432 0.095 Xe-135 2.11E-5 0.247 0.316 Xe-137 3.02E-3 0.194 1.642 Xe-138 8.15E-4 1.183 0.606 15B.3 WHOLE-BODY GAMMA AND BETA SKIN DOSE The whole-body gamma dose delivered to an offsite dose receptor is calculated by assuming the receptor to be immersed in a hemispherical radioactive cloud that is infinite in all directions above the ground plane; i.e., a semi-infinite cloud.
| |
| The concentration of radioactive material within this cloud is uniform and equal to the maximum centerline ground level June 2001 15B-2 Revision 11
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| | |
| PVNGS UPDATED FSAR APPENDIX 15B concentration that would exist in the cloud at the appropriate distance from the point of release.
| |
| The gamma dose to an offsite receptor due to gamma radiation for a given time period is:
| |
| D wb = /Q * i DCFwbi
| |
| * Qi (1) where:
| |
| Dwb = whole-body dose to an offsite receptor from gamma radiation, (rem)
| |
| /Q = site atmospheric dispersion factor effective during the time period at the point of exposure, 3
| |
| (s/m )
| |
| DCFwbi = whole body dose conversion factor for the semi-3 infinite cloud model for nuclide i, (rem-m /Ci-s).
| |
| (See table 15B-2)
| |
| Qi = total activity of nuclide i released during the time period, (Ci)
| |
| The gamma dose to the control room personnel is calculated assuming a finite hemispherical cloud model. The gamma dose due to gamma radiation in the control room for a given time period is:
| |
| (CRVOL)0.338 (IQi)(3600)(CRO)
| |
| D wb =
| |
| 1173
| |
|
| |
| i DCFwbi (CRVOL)(0.02832)
| |
| (2) where:
| |
| Dwb = whole-body gamma dose to control room personnel from gamma radiation, (rem)
| |
| CRO = the control room occupancy factor 1 June 2001 15B-3 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B Table 15B-2 WHOLE-BODY GAMMA AND BETA SKIN IMMERSION DOSE CONVERSION FACTORS Beta Skin DCF Whole-Body Gamma DCF Radionuclide (rem - m3/Ci - sec) (rem - m3/Ci - s)
| |
| I-131 3.17E-2 8.72E-2 I-132 1.32E-1 5.13E-1 I-133 7.35E-2 1.55E-1 I-134 9.23E-2 5.32E-1 I-135 1.29E-1 4.21E-1 Kr-83m 0 2.40E-6 Kr-85 4.24E-2 5.102E-4 Kr-85m 4.62E-2 3.7E-2 Kr-87 3.08E-1 1.88-1 Kr-88 7.51E-2 4.65E-1 Kr-89 3.2E-1 5.26E-1 Xe-131m 1.508E-2 2.89E-3 Xe-133m 3.15E-2 7.95E-3 Xe-133 9.69E-3 9.32E-3 Xe-135m 2.25E-2 9.89E-2 Xe-135 5.89E-2 5.74E-2 Xe-137 3.86E-1 4.50E-2 Xe-138 1.31E-1 2.80E-1 3600 = conversion factor, s/h 3 3 0.02832 = conversion factor, ft /m 3
| |
| CRVOL = control room volume, ft IQi = total integrated activity for nuclide i in control room for the time period, (Ci-hr)
| |
| DCFwbi = the semi-infinite cloud whole-body gamma dose 3
| |
| conversion factor for nuclide i, (rem-m /Ci-s).
| |
| (See table 15B-2)
| |
| June 2001 15B-4 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B (CRVOL)0.338 The expression is a geometrical correction factor 1173 (7) to ratio a finite cloud to infinite cloud .
| |
| The beta skin dose to control room personnel is calculated assuming 2
| |
| a tissue depth of 7 mg/cm . The beta skin dose to control room personnel for a given time period is:
| |
| CRO D s =
| |
| (CRVOL)(0.02832)
| |
|
| |
| i Dsi
| |
| * IQi (3) where:
| |
| D si = the beta skin dose conversion factor for nuclide i, 3
| |
| (rem-m /Ci-h). (See table 15B-2 for factor) and all other parameters are as previously defined.
| |
| 15B.4 THYROID INHALATION DOSE The thyroid dose to an offsite receptor for a given time period is obtained from the following expression:
| |
| D = x / Q
| |
| * Bi * (Qi
| |
| * DCFi) (4) where:
| |
| D = thyroid inhalation dose, (rem) x/Q = site atmospheric dispersion factor during the 3
| |
| time period, (s/m )
| |
| 3 B = breathing rate during the time period, (m /s).
| |
| (See table 15B-3)
| |
| Qi = total activity of nuclide i released during time period, (Ci)
| |
| DCFi = thyroid dose conversion factor for nuclide i, June 2001 15B-5 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B (rem/Ci inhaled). (See table 15B-4)
| |
| The radionuclide data are given in table 15B-1. The atmospheric dispersion factors used in the analysis of the environmental consequences of accidents are given in section 2.3.
| |
| Breathing rates and dose conversion factors for radioactive iodines required for computing thyroid inhalation doses are tabulated in Tables 15B-3 and 15B-4, respectively.
| |
| Table 15B-3 (a)
| |
| BREATHING RATES 3
| |
| Time After Accident m /s 0 to 8 hours 3.47(-04) 8 to 24 hours 1.75(-04) 1 to 30 days 2.32(-04)
| |
| : a. From Regulatory Guide 1.4 Table 15B-4 IODINE DOSE CONVERSION FACTORS Rem -Thyroid/Curie Inhaled Iodine Isotope a b TID-14844 ICRP-30 I-131 1.48(+06) 1.08(+06)
| |
| I-132 5.35(+04) 6.44(+03)
| |
| I-133 4.00(+05) 1.80(+05)
| |
| I-134 2.50(+04) 1.07(+03)
| |
| I-135 1.24(+05) 3.13(+04)
| |
| : a. See reference 8
| |
| : b. See reference 9 June 2001 15B-6 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B 15B.5 CONTROL ROOM DOSE During the course of an accident, control room personnel may receive doses from the following sources:
| |
| A. Direct whole-body gamma dose from the radioactivity present in the containment building B. Direct whole-body gamma dose from the radioactive cloud surrounding the control room C. Whole-body gamma, thyroid inhalation, and beta skin doses from the airborne radioactivity present in the control room.
| |
| In calculating the exposure to control room personnel, occupancy factors were obtained from reference 7 as follows:
| |
| 0 to 24 hours: occupancy factor = 1 1 to 4 days: occupancy factor = 0.6 4 to 30 days: occupancy factor = 0.4 The dose model for each of the radiation sources is discussed below:
| |
| A. Direct whole-body gamma dose from the radioactivity present in the containment building (direct containment dose).
| |
| Time integrated (0 to 30 days) radionuclide concentra-tions in the containment are calculated. For conservatism, no credit is taken for reduction of the containment activity by means other than radioactive decay. The containment is modeled by an equivalent volume cylindrical source having a diameter of 146 feet and a height of 155 feet. The radioactivity present in the containment is assumed to be uniformly June 2001 15B-7 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B distributed in the cylindrical source. Shielding is provided by the 4-foot concrete containment walls, 120 feet of air separating the containment building from the control building, and 2-foot thick control room walls.
| |
| No credit is taken for any shielding that would be provided by the auxiliary building.
| |
| B. Direct whole-body gamma dose from the radioactive cloud surrounding the control room (outside cloud dose).
| |
| Leakage from the containment building, or any building, will result in the formation of a radioactive plume. For conservatism it is assumed that this plume forms a cloud surrounding the control room. Gamma radiation from this cloud, although attenuated, can penetrate the control room roof and walls resulting in a whole-body gamma dose to control room personnel. The radius of the cloud is computed using a mass balance of the radioactivity released due to leakage and the volume of the cloud; therefore, the radioactive cloud is time variant and expands for the duration of the accident.
| |
| 3 Radioactivity concentration (Ci/m ) in the radio-active cloud surrounding the control room is the product of the building leak rate (Ci/s) and the control room 3
| |
| atmospheric dispersion factor, /Q (s / m . Exclusion area boundary and low population zone /Qs are presented in section 2.3. A tabulation of control room
| |
| /Qs is presented in table 15B-5.
| |
| Credit is taken for concrete shielding provided by the control room walls and ceiling.
| |
| June 2001 15B-8 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B Table 15B-5 ATMOSPHERIC DISPERSION FACTORS (1986 - 1991) 3 Time Period Control Room /Q (s/m )
| |
| 0 to 8 hours 1.56(-3) 8 to 24 hours 1.08(-3) 1 to 4 days 4.15(-4) 4 to 30 days 1.03(-4)
| |
| C. Dose from the airborne radioactivity present in the control room (occupancy dose).
| |
| Airborne radioactivity will be drawn into the control room due to the intake of outside air required to maintain a positive pressure in the control room.
| |
| This contributes to the whole-body gamma, thyroid inhalation, and beta skin doses. The major parameters of the control room ventilation system are presented in table 15B-6.
| |
| The whole-body gamma dose is computed using a finite cloud model. The calculational model is an equivalent volume hemisphere of 42-foot radius.
| |
| A thyroid inhalation dose results from the radioactive iodine present in the control room. The control room habitability system, designed to remove iodine from the air, is described in table 15B-6.
| |
| June 2001 15B-9 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B 15B.6 ACTIVITY RELEASE MODELS 15B.6.1 GENERAL EQUATION The activity released from a postulated accident is calculated by using the following matrix equation for each isotope and each specie of iodine:
| |
| dA
| |
| + C A = S; Initial Condition A(to) = A o (5) dt Q = L x AI where:
| |
| A(T) = i (a (t))
| |
| ai = the activity in the ith node, (Ci)
| |
| C = (Cij) matrix Cij = the transfer rate from the ith node to the jth
| |
| -1 node, (s )
| |
| S = (Si) vector Si = the production rate in the ith node (Ci/sec)
| |
| Q = the activity released to the environment over the time period to to ti, (Ci)
| |
| L = (" i) matrix "i = the leak rate from the ith node to environment
| |
| ( /sec)
| |
|
| |
| t1 AI = A (t) dt (Ci-sec) to June 2001 15B-10 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B Table 15B-6 (a)
| |
| CONTROL ROOM ESSENTIAL VENTILATION SYSTEM PARAMETERS Parameter Assumption Number of emergency ventilation systems 1 operating Maximum filtered intake rate, (SCFM) 1,000 Unfiltered intake rate, (SCFM) 0 Unfiltered intake for egress/ingress See Section (SCFM) 6.4.7 Intake clean filter efficiency Iodine, elemental,% 95 Iodine, organic,% 95 Iodine, particulate% 99 Minimum Recirculation rate, standard 25,740 ft3/min Recirculation cleanup filter efficiency Iodine, elemental, % 95 Iodine, organic. % 95 Iodine, particulate % 99 Leak rate, standard ft3/min (out 1,010 leakage)
| |
| Control room volume, standard ft3 161,000
| |
| : a. There are two completely redundant emergency control room ventilation systems.
| |
| For a more detailed description of this system, refer to subsection 9.4.2. The dose model employed in this analysis is consistent with the thyroid inhalation model discussed in section 15B.4.
| |
| The beta skin dose model is consistent with the "infinite hemispherical cloud" model described in section 15B.3.
| |
| June 2003 15B-11 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B Each node represents a volume where activity can be accumulated.
| |
| The environment and the control room are each represented by a node. To ensure that the system of differential equations has constant coefficients, the time scale is broken up into time intervals over which all parameters are constant. Thus, all coefficients and sources are assumed to be representable by step functions.
| |
| The matrix equation is solved using matrix techniques. The particular solution is obtained by Gaussian elimination. The homogenous solution is obtained by solving for the eigenvectors and the eigenvalues of the coefficient matrix C. They are determined by using QR transformation techniques.
| |
| The following sections describe how the coefficient matrix and the source vector are calculated for the different accident calculations.
| |
| 15B.6.2 THE MODEL FOR CONTAINMENT LEAKAGE The model for LOCA containment leakage is shown in figure 15B-1.
| |
| The system of differential equations for estimating the released activity is as follows:
| |
| dA 1
| |
| + dA1 - L21A2 - L31A3 - L41A4 = 0 (6a) dt dA 2
| |
| + (d + s + L21 + L24) A2 - L42A4 = 0 (6b) dt dA 3
| |
| + (d + s + L31 + L34)A3 - L43A4 = 0 (6c) dt dA 4
| |
| - L24A2 - L34A3 + (d + L41 + L42 + L43)A4 = 0 (6d) dt June 2001 15B-12 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B dA 5 X
| |
| X
| |
| - (Lu+(1-fL)Lf) L21A2 - (Lu+(1-fL)Lf)L31 dt Q Q X
| |
| (Lu+(1-fL)Lf) L41A4 + (Lf + Lu + frRc + d) A5 = 0 (6e)
| |
| Q t1 Q = t0 (L21A2 + L31A3 + L41A4) dt (7) where:
| |
| A1(t) = activity in the environment, (Ci)
| |
| A2(t) = activity in the sprayed main region of the containment, (Ci)
| |
| A3(t) = activity in the auxiliary sprayed region of the containment (Ci)
| |
| A4(t)=activity in the unsprayed region of the containment, (Ci)
| |
| A5(t) = activity in the control room, (Ci) d
| |
| -1
| |
| = radioactive decay constant, (s )
| |
| T21 -1 L21 = , (s )
| |
| (100)(24)(3600)
| |
| T21 = leak rate from the main sprayed volume to the environment, (%/day)
| |
| T31 -1 L31 = , (s )
| |
| (100)(24)(3600) 31 = leak rate from the auxiliary sprayed volume to the environment, (%/day)
| |
| T41 -1 L41 = , (s )
| |
| (100)(24)(3600)
| |
| T41 = leak rate from the unsprayed volume to the environment (%/day)
| |
| June 2001 15B-13 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B s
| |
| -1
| |
| = the spray removal constant, (s )
| |
| T24 -1 L24 = , (s )
| |
| (V2)(60)
| |
| T24 = transfer rate from the main sprayed region to the 3
| |
| unsprayed region, (ft /min) 3 V2 = volume of the main sprayed region, (ft )
| |
| T42 -1 L42 = , (s )
| |
| (V4)(60)
| |
| T42 = transfer rate from the unsprayed region to the 3
| |
| sprayed region, (ft /min) 3 V4 = volume of the unsprayed region, (ft )
| |
| T34 -1 L34 = , (s )
| |
| (V3)(60)
| |
| T34 = transfer rate from the auxiliary sprayed 3
| |
| region to the unsprayed region (ft /min) 3 V3 = volume of the auxiliary sprayed region (ft )
| |
| T43 -1 L43 = , (s )
| |
| (V4)(60)
| |
| T43 = transfer rate from the unsprayed region 3
| |
| to the auxiliary sprayed region, (ft /min)
| |
| Tu (0.3048)3 3 Lu = , (m /s) 60 Tu = unfiltered inleakage into the control room, 3
| |
| ft /min)
| |
| Tf (0.3048)3 3 Lf = , (m /s) 60 June 2001 15B-14 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B f = filtered air intake rate into the control room, 3
| |
| (ft /min) fL = filter efficiency of the filters on the intake units x/Q = atmospheric dispersion factor for the control room, 3
| |
| (s/m )
| |
| Tr -1 Rc = , (s )
| |
| (Vc)(60)
| |
| TR = filtered recirculation rate in the control room, 3
| |
| (ft /min) 3 Vc = control room free volume, (ft )
| |
| fR = filter efficiency of the filter on the recircula-tion unit Q = activity released to the environment, (Ci)
| |
| The coefficient matrix is:
| |
| C =
| |
| d -L21 -L31 -L41 0 0 d+s+L21+L24 0 -L42 0 0 0 d+s+L31+L34 -L43 0 0 -L24 -L34 d+L41+L42+L43 0 0 -x/Q(Lu+(1-fL)Lf)L21 -x/Q(Lu+(1-fL)Lf)L31 -x/Q(Lu+(1-fL)Lf)L41 Lf+Lu+frRc+d After solving for A(t), the integrated activity in each node can then be calculated.
| |
| June 2001 15B-15 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B From the integrated activity, the offsite doses and the doses to the operators in the control room can be calculated using the dose models given in sections 15B.3 and 15B.4.
| |
| 15B.6.3 THE MODEL FOR RECIRCULATION LOOP LEAKAGE The model for LOCA leakage in recirculation loops outside containment is shown in figure 15.B-2. The activity released due to the operational leakage of the engineered safety feature (ESF) components during the recirculation mode of the postulated LOCA is calculated from the following equations:
| |
| dA 1
| |
| + dA1 - (1-f)L21 A2 = 0 (8a) dt dA 2
| |
| + (+d+L21) A2 = S2 (8b) dt Q = tt1o (1-f) L21 A2 dt (9) where:
| |
| A1 = the activity in the environment, (Ci)
| |
| A2 = the activity in the ESF component rooms, (Ci) d = decay constant, (s )
| |
| -1 L21 = filtered leak rate to the environment, (ESF room vol/s) f = filter efficiency of the filters on the ESF room purge units A oTs S2 = P
| |
| * Vs Ao = activity in the recirculation water, (Ci)
| |
| P = iodine partition factor June 2001 15B-16 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B 3
| |
| Ts = twice the maximum operational leak rate, (cm /s) 3 Vs = total volume of recirculation water, (cm )
| |
| Q = activity released to the environment, (Ci)
| |
| The coefficient matrix is:
| |
| C = d -(1-f)L21 O (d + L21)
| |
| The source vector is O
| |
| S =
| |
| S2 15B.6.4 THE MODEL FOR THE FUEL HANDLING ACCIDENT IN THE FUEL BUILDING WITH ESF SAFEGUARDS ACTUATION The model for the release of activity from the fuel building during a postulated fuel handling accident is shown in figure 15B-3. The activity released to the environment is estimated from the following equations:
| |
| d 1
| |
| + d 1 (1-f)L21A2 = 0 (10a) dt d 2
| |
| + ( d + L21) 2 = 0 (10b) dt Q = tt1o L21 2dt (11) where:
| |
| A1 = activity in the environment, (Ci)
| |
| A2 = activity in the fuel building atmosphere, (Ci)
| |
| June 2001 15B-17 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B d
| |
| -1
| |
| = decay constant, (s )
| |
| -1 L21 = purge rate to the environment, (s )
| |
| f = filter efficiency of the filters on the ventilation unit Q = activity released to the environment, (Ci)
| |
| The resultant coefficient matrix is:
| |
| d -(1-f)L21 C=
| |
| O (d + L21) 15B.6.5 OTHER ACCIDENT MODELS Other accidents can be conservatively modeled as simulated instantaneous releases to the environment. This is simulated as a large transfer rate to the environment. The model is shown in figure 15B-3. The system of differential equations is:
| |
| d 1
| |
| + d 1 L21 2 = 0 (12a) dt d 2
| |
| + ( d + L 21) 2 = 0 (12b) dt Q = tt1o L21 2 dt (13) where:
| |
| A1 = activity in the environment, (Ci)
| |
| A2 = activity to be released to the environment, (Ci) d
| |
| -1
| |
| = decay constant, (s )
| |
| -1 L21 = very large transfer rate to the environment, (s )
| |
| June 2001 15B-18 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B Q = activity released to the environment, (Ci)
| |
| The resultant coefficient matrix is:
| |
| d -L21 C=
| |
| O (d + L21)
| |
| June 2001 15B-19 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15B 15B.7 REFERENCES
| |
| : 1. "Report of ICRP Committee II, Permissible Dose for Internal Radiation (1959)," Health Physics, 3, pp 30, 146-153, 1960.
| |
| : 2. Martin, M. J. and Blichert-Toft, P. H., Radioactive Atoms, Auger-Electron, L-, B-, q-, and X-Ray Data, Nuclear Data Tables A8, 1, 1970.
| |
| : 3. Martin, M. J., Radioactive Atoms - Supplement 1, ORNL-4923, August 1973.
| |
| : 4. Bowman, W. W. and MacMurdo, K. W., "Radioactive Decay Gammas, Ordered by Energy and Nuclide," Atomic Data and Nuclear Data Tables 13, 89, 1974.
| |
| : 5. Meek, M. E. and Gilbert, R. S.,"Summary of Gamma and Beta Energy and Intensity Data," NEDO-12037, January 1970.
| |
| : 6. Lederer, C. M., Hollander, J. M., and Perlman, I., Table of the Isotopes, 6th Edition, March 1968.
| |
| : 7. Murphy, K. G. and Campe, K. M., "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," Thirteenth AEC Air Cleaning Conference.
| |
| : 8. Di Nunno, J. J., et al, "Calculation of Distance Factors for Power and Test Reactor Sites," TID 14844, March 1962.
| |
| : 9. "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity", International Commission on Radiological Protection (ICRP), Publication 30, Supplement to Part 1, 1980.
| |
| June 2001 15B-20 Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15C APPENDIX 15C DELETED June 2005 Revision 13
| |
| | |
| Intentionally Left Blank PVNGS UPDATED FSAR APPENDIX 15D ANALYSIS METHODS FOR LOSS OF PRIMARY COOLANT FLOW
| |
| | |
| PVNGS UPDATED FSAR PVNGS UPDATED FSAR CONTENTS Page 15D.1 INTRODUCTION 15D-1 15D.2 COMPUTER CODES 15D-1 15D.2.1 DATA TRANSFER 15D-1 15D.2.2 COAST 15D-2 15D.2.3 CESEC III OR CENTS 15D-3 15D.2.4 HERMITE 15D-4 15D.2.5 CETOP 15D-5 15D.3 COMPARISON WITH PREVIOUS METHODS 15D-6 15D.4 ANALYSIS ASSUMPTIONS 15D-7 15D.5 REFERENCES 15D-8 June 2003 15D-i Revision 12
| |
| | |
| PVNGS UPDATED FSAR FIGURES 15D-1 Data Transfer Between Computer Codes for the ST-LOF Method June 2001 15D-ii Revision 11
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D ANALYSIS METHODS FOR LOSS OF PRIMARY COOLANT FLOW 15D.1 INTRODUCTION This appendix provides a description of methods used in the analysis of the nuclear steam supply system (NSSS) response to a loss of primary coolant flow (LOF) event. An LOF could occur as a result of a loss of electrical power to the four reactor coolant pumps (RCPs). The conclusions and results presented in subsection 15.3.1 were obtained using the methods described here. This method will hereafter be referred to as Space-Time Kinetics LOF (ST-LOF).
| |
| The computer codes used in the ST-LOF method are described in section 15D.2. The principal time-dependent parameters calculated are the primary coolant flow rate, reactor core power, hot bundle heat flux, and limiting channel departure from nucleate boiling ratio (DNBR). A comparison of present analysis methods with previous methods is given in section 15D.3. The analysis assumptions are listed in section 15D.4.
| |
| 15D.2 COMPUTER CODES 15D.2.1 DATA TRANSFER Given the postulated initiating event, the COAST code is used to compute the core inlet volumetric flow rate as a function of time. These data are then input to the CESEC or CENTS code which predicts the overall system response. CESEC or CENTS calculates plant protection system responses and valve actuations for assessing the long term consequences of the LOF.
| |
| CESEC or CENTS also computes the time-dependent core inlet mass flux, core inlet coolant temperature, and reactor coolant system (RCS) pressure (note that no credit is taken for June 2003 15D-1 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D for pressure increases when computing the DNBR transient effects). These parameters can later be used as input to the HERMITE code. For those cases where a reactor trip occurs so rapidly that only the coolant flow rate changes, CESEC or CENTS is bypassed and the flow coastdown is input directly to HERMITE.
| |
| HERMITE is used to predict the reactor core response during a LOF. HERMITE calculates the transient core power, core average heat flux, and hot bundle heat flux. The time-dependent core average heat flux along with the core inlet coolant mass flux, core inlet coolant temperature, and RCS pressure are input to the CETOP code. This code computes the limiting channel coolant conditions and the limiting channel DNBR. Figure 15D-1 depicts the transfer of data between the computer codes used.
| |
| The entire data transfer highlighted in figure 15D-1 is not repeated for each cycle reload since nothing in the early steps changes from cycle to cycle. The later steps (specifically, HERMITE and CETOP executions) are repeated with each cycle because of changes in fuel parameters (e.g., core average heat flux and hot bundle heat flux).
| |
| 15D.2.2 COAST The COAST code is used in the same manner as described in (1)
| |
| CENPD-183. COAST analyzes reactor coolant flow under any combination of active and inactive pumps in a two-loop, four pump plant. The equation of conservation of momentum is written for each of the flow paths of the COAST model assuming unsteady one-dimensional flow of an incompressible fluid. The equation of conservation of mass is written for the appropriate nodal points. Pressure losses due to friction, bends, and shock losses are assumed proportional to the flow velocity June 2003 15D-2 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D squared. Pump dynamics are modeled using a head flow curve for a pump at full speed and using four quadrant curves, which are parametric diagrams of pump head and torque on coordinates of speed versus flow, for a pump at other than full speed.
| |
| The COAST code has been verified by comparison to measurements taken during the initial startup test program of the PVNGS units. A further description of COAST is contained in (2)
| |
| CENPD-98.
| |
| 15D.2.3 CESEC III OR CENTS The CESEC III or CENTS code is used to determine the long term response of the NSSS to primary coolant flow reductions resulting from postulated LOF events. The CESEC III or CENTS code may also be used to predict the change in core inlet coolant temperature if this parameter changes before the time of minimum DNBR.
| |
| CESEC III or CENTS computes key system parameters during a transient including core heat flux, pressures, temperatures, and valve actions. A partial list of the dynamic functions included in this NSSS simulation includes:
| |
| * point kinetics neutron behavior
| |
| * Doppler and moderator reactivity feedback
| |
| * boron and control element assembly (CEA) reactivity effects
| |
| * multi-node average channel reactor core thermal hydraulics
| |
| * reactor coolant pressurization and mass transport
| |
| * RCS safety valve behavior
| |
| * steam generation
| |
| * steam generator water level June 2003 15D-3 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D
| |
| * main steam bypass
| |
| * secondary safety and turbine valve behavior
| |
| * alarm, control, protection, and engineered safety feature system actions.
| |
| Initial steam generator feedwater enthalpy and flow rate are typically set to match the initial power for transient simulations. For a further description of CESEC III or CENTS, see section 15.0.
| |
| 15D.2.4 HERMITE One application of the HERMITE code is to determine the reactor core response during postulated LOF events. HERMITE can accept as input the transient boundary conditions of coolant flow rate, inlet coolant temperature, RCS pressure, and CEA position. In this application, HERMITE solves the few-group, space- and time-dependent neutron diffusion equation including feedback effects of fuel temperature, coolant temperature, coolant density, and control element motion for a one-dimensional average fuel bundle. The fuel temperature model explicitly represents the pellet, gap, and clad regions of an average fuel pin and representative hot bundle fuel pin. The hot bundle fuel pin power density is related to the average fuel pin power density by time-dependent planar radial power peaking factors. For the calculation of heat flux, heat conduction equations are solved by a finite difference method.
| |
| Continuity and energy conservation equations are solved in order to determine the coolant temperature and density for the average and hot bundles. A further description of HERMITE is (3) given in CENPD-188-A.
| |
| June 2003 15D-4 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D The hot bundle fuel pin power density is equal to the core average fuel pin power density multiplied by the planar radial power peaking factor, F(xy)(z). For times prior to the insertion of CEAs, and for regions of the core that the CEAs have not entered, the F(xy)(z) is equal to a conservatively chosen initial value. As the CEAs pass a plane of the core, the radial power peaking factor of that plane is increased as a function of time from the initial value. The increase in F(xy)(z) calculated by HERMITE is limited so that the power at a hot spot within a given plane of the core will not rise faster than for the average of that plane. If the power in the average channel of the plane has fallen since the last time step, the F(xy)(z) increase is limited so that the power in the hot spot for that plane does not increase.
| |
| The synthesis of the axial power distribution and the planar radial power peaking factors provides a conservative representation of the hottest fuel assembly during the LOF transient, including maximum three dimensional power peaking effects. This technique yields a conservative prediction of the minimum DNBR which can occur as a result of the LOF transient.
| |
| 15D.2.5 CETOP The CETOP code uses the CE-1 critical heat flux correlation (4) described in CENPD-162-P to calculate the limiting channel DNBR transient. CETOP receives the core average fuel bundle heat flux, core inlet coolant mass flux, core inlet coolant temperature, and RCS pressure at selected times during the LOF transient. The code is used to perform static calculations of the axial coolant enthalpy distribution and DNBR at these June 2003 15D-5 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D times. No credit is taken for RCS pressure increases in calculating the DNBR. The CETOP code is also discussed in section 15.0.
| |
| 15D.3 COMPARISON WITH PREVIOUS METHODS (3)
| |
| CENPD-183, Appendix A describes the methodology used to predict the consequences of postulated LOF events for many previous Combustion Engineering NSSS designs. This section summarizes the fundamental differences between the ST-LOF method and that described in CENPD-183.
| |
| The primary difference between these methods is in the calculation of the core power. The CENPD-183 method uses the QUIX code to compute reactivity as a function of CEA position assuming the neutron flux and delayed neutron precursors are in equilibrium. Combining CEA position versus time data with the reactivity versus CEA position data produces the time-dependent reactivity function which is input to the CESEC or CENTS point kinetics equations.
| |
| The ST-LOF method uses HERMITE to calculate the core power directly from CEA position versus time. HERMITE calculates the time-dependent neutron flux in one dimension (axially) with the few group diffusion equation explicitly accounting for fission, absorption, and transport cross-section variations.
| |
| Other differences exist in the calculation of the hot channel heat flux. In the CENPD-183 methodology, it is assumed that the hot channel normalized heat flux decay is equivalent to the core average normalized heat flux decay for computing the time of minimum DNBR. Furthermore, it is assumed that the axial heat flux distribution is constant in time. The minimum DNBR June 2003 15D-6 Revision 12
| |
| | |
| PVNGS UPDATED FSAR APPENDIX 15D value calculated with the CENPD-183 methodology assumes no decay of the hot channel heat flux.
| |
| In the ST-LOF method it is assumed that the hot bundle normalized power decay is equivalent to the core average normalized power decay prior to the insertion of the CEAs. As the CEAs are inserted in the core, the planar radial peaking factors are increased so that the hot channel power decreases less rapidly than core average power for the rodded planes.
| |
| The hot bundle and core average axial heat flux distributions are each time-dependent. The minimum DNBR value calculated with the ST-LOF method is based on the decay heat flux calculated by HERMITE at the time of minimum DNBR.
| |
| CENPD-183 describes both static and dynamic methods for computing the DNBR. The ST-LOF method uses the static method for calculating the DNBR as described in CENPD-183, Appendix A except that CETOP is used in place of COSMO.
| |
| 15D.4 ANALYSIS ASSUMPTIONS A number of conservative assumptions are made in the LOF analysis. These assumptions are:
| |
| * RCS pressure increase during the transient is not credited for in DNBR calculations
| |
| * a conservative (most positive) moderator temperature coefficient (MTC) is assumed
| |
| * a conservative time of the pump speed trip is assumed
| |
| * a conservative (minimum) scram bank reactivity rod worth is assumed.
| |
| June 2003 15D-7 Revision 12
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| | |
| PVNGS UPDATED FSAR APPENDIX 15D 15D.5 REFERENCES
| |
| : 1. "Loss of Flow - CE Methods for Loss of Flow Analysis,"
| |
| CENPD-183, July 1975 (Proprietary).
| |
| : 2. "COAST Code Description," CENPD-98, April 1973 (Proprietary).
| |
| : 3. "HERMITE Space-Time Kinetics," CENPD-188-A, July 1976.
| |
| : 4. "CE Critical Heat Flux - Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Space Grids,"
| |
| CENPD-162-P, April 1975 (Proprietary).
| |
| June 2001 15D-8 Revision 11
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| | |
| PVNGS UPDATED FSAR APPENDIX 15E LIMITING INFREQUENT EVENT
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| PVNGS UPDATED FSAR APPENDIX 15E CONTENTS Page 15E.1 IDENTIFICATION OF CAUSES AND FREQUENCY CLASSIFICATION 15E-1 15E.2 SEQUENCE OF EVENTS AND SYSTEM OPERATIONS 15E-1 15E.3 CORE AND SYSTEM PERFORMANCE 15E-2 15E.4 RCS PRESSURE BOUNDARY BARRIER PERFORMANCE 15E-10 15E.5 CONTAINMENT PERFORMANCE AND RADIOLOGICAL CONSEQUENCES 15E-10 15E.6 CONCLUSIONS 15E-17 15E.7 REFERENCES 15E-17 June 2005 15E-i Revision 13
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| PVNGS UPDATED FSAR APPENDIX 15E TABLES Page 15E-1 PARAMETERS USED FOR THE LIMITING INFREQUENT EVENT 15E-8 15E-2 MINIMUM DNBR VERSUS RADIAL PEAKING FACTOR FOR THE LIMITING INFREQUENT EVENT CORE PERFORMANCE SAFETY ANALYSIS 15E-10 15E-3 RCS IODINE SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS 15E-16 15E-4 RCS NOBLE GAS SOURCE TERM FOR THE IOSGADV OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS 15E-16 15E-5 RADIOLOGICAL CONSEQUENCE OF LIMITING INFREQUENT EVENT RADIOLOGICAL ANALYSIS (FR = 2.0, 10% FUEL FAILURE) 15E-17 June 2005 15E-ii Revision 13
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| PVNGS UPDATED FSAR APPENDIX 15E FIGURES 15E-1 DNBR vs. TIME June 2005 15E-iii Revision 13
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| PVNGS UPDATED FSAR PVNGS UPDATED FSAR APPENDIX 15E APPENDIX 15E LIMITING INFREQUENT EVENTS 15E.1 IDENTIFICATION OF CAUSES AND FREQUENCY CLASSIFICATION This event is a composite event that is evaluated to bound all infrequent events, including Anticipated Operational Occurrences (AOOs) in combination with a single active failure, with respect to the degradation in the Departure from Nucleate Boiling Ratio (DNBR). In general, all combinations of infrequent events including AOOs with single active failures need to be evaluated. To avoid evaluating all the potential AOOs as initiating events and single failures, the composite event assumes that an unspecified initiating event degrades all the thermal margin preserved by the COLSS and brings the core conditions to the DNBR SAFDL. Therefore, this event is initiated by any AOO (moderate frequency events). The composite event, by definition, is an infrequent event and includes moderate frequency events in combination with an active single failure.
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| Note that this composite event bounds all infrequent events including AOOs with single failure with respect to the DNBR degradation. The limiting AOO with respect to Fuel Centerline Melt Temperature (or Peak Linear Heat Rate) is CEA Bank Withdrawal event that is evaluated in UFSAR Chapter 15.4.
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| 15E.2 SEQUENCE OF EVENTS AND SYSTEM OPERATIONS The composite event assumes that an unspecified initiating event degrades all the thermal margin preserved and brings the core conditions to the DNBR SAFDL. This assumption is June 2005 15E-1 Revision 13
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| | |
| PVNGS UPDATED FSAR APPENDIX 15E conservative since the AOOs are specifically analyzed to ensure that the SAFDLs are not violated and the necessary thermal margin is preserved by the LCOs. The active single failure is assumed to further aggravate the DNBR degradation. The most limiting active single failure for DNBR degradation is determined to be LOP, which in turn, result in the coastdown of all four RCPs. Therefore, the most limiting infrequent event with respect to the DNBR degradation can be described as The Loss of Flow (LOF) from the SAFDL.
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| The composite event then is simply modeled as a LOF, using the LOF methodology (References 1 and 2), with the conditions at the beginning of the flow coastdown corresponding to SAFDL conditions. Starting from SAFDL conditions, the LOP results in an RCP coastdown that leads almost immediately to a reactor trip by the Core Protection Calculator (CPC) DNBR function, as the reduction in Reactor Coolant System (RCS) flow degrades DNBR below the initial SAFDL conditions. Within ~3.5 seconds of event initiation (~3 seconds after reactor trip), the local and average core heat fluxes have decreased sufficiently so that no pins remain in DNB. Hence, DNB propagation is not predicted to occur. Figure 15E-1 provides the transient DNBR response for the event.
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| 15E.3 CORE AND SYSTEM PERFORMANCE A set of initial conditions corresponding to the DNBR SAFDL was calculated with the CETOP-D code. This is a bounding assumption, since the CPC DNBR trip will provide a trip prior to the core conditions reaching the DNBR SAFDL conditions with a very high probability. The SAFDL conditions include an assumed, pre-existing power of 116%, representing the undefined June 2009 15E-2 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E limiting AOO. The LOF methodology models only the core average and the hot channel using the HERMITE computer code (Reference 3).
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| The core average and hot channel response to the LOF event from these initial conditions was simulated using the 1-D HERMITE code.
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| The transient DNBR values were calculated using the CETOP-D code, which uses the CE-1 CHF correlation. Input parameters and initial conditions were selected to maximize the DNBR degradation. Using the conditions at the time of minimum DNBR, a more accurate DNBR is calculated using the more detailed TORC code for several different values of radial peaking.
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| Although a LOP would not occur for at least three seconds following a turbine trip, this evaluation conservatively assumes a coincident turbine trip and LOP. The RCP coastdown leads to a CPC DNBR reactor trip. RCS flow coastdown degrades DNBR below the initial SAFDL conditions. DNBR degradation is terminated when the mitigating effects of the SCRAM Control Element Assembly (CEA) insertion dominate the flow coastdown.
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| 15E.3.1 Mathematical Models The Limiting Infrequent Event was analyzed with respect to core performance with the following mathematical models:
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| * The one-dimensional HERMITE space-time computer code was used to calculate the core average and hot channel response to a LOF event from the DNBR SAFDL. The HERMITE computer code is described in UFSAR Section 4.3 and in an NSSS vendor topical report (Reference 3). HERMITE was also used to determine the boron concentration at the Moderator Temperature Coefficient (MTC) value selected for this analysis.
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| June 2009 15E-3 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E
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| * The CETOP-D computer code, which uses the CE-1 Critical Heat Flux correlation, was used to calculate the initial and transient DNBR values. The CETOP-D computer code is described in UFSAR Section 4.4 and in NSSS vendor topical reports (References 4, 5, and 6). CETOP-D was also used to determine initial Power Operating Limit (POL) conditions for this event (see UFSAR Section 15.1.3.3.2 for additional information on POL conditions).
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| * The TORC computer code, which uses the CE-1 CHF correlation, was used to calculate the minimum DNBR value using the conditions corresponding to the time minimum DNBR was predicted by the CETOP-D code. The TORC computer code is described in UFSAR Section 4.4 and in NSSS vendor topical reports (References 8 and 9).
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| * Because the models in the CETOP-D code are not as detailed as those in TORC, DNBR predictions from CETOP-D are typically adjusted by penalty factors to ensure conservatism. Use of the more detailed TORC computer code removes the requirement for penalty factors and provides a more accurate prediction of the DNBR value than the CETOP-D code.
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| * Fuel failure calculations use a statistical convolution technique that is described in Reference 1. This technique involves the summation, over the reactor core, of the number of fuel pins at a specific DNBR value, multiplied by the probability of DNB occurring at that DNBR value.
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| June 2011 15E-4 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E 15E.3.2 Input Parameters and Initial Conditions Table 15E-1 summarizes the key input parameters and initial conditions utilized in the core performance safety analysis for the limiting infrequent event. Since the average NSSS response is not applicable for this event, the parameters that are not input to the HERMITE or CETOP/TORC are not listed in the table.
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| The following assumptions are made in this analysis:
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| * The Rated Thermal Power (RTP) was set to the maximum, 3990 MWt.
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| * The most limiting event would occur from full power operation. The initial power level used in the core performance safety analysis was increased to 116% of RTP (to represent the ROPM for an unspecified limiting AOO).
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| * The initial core inlet temperature, and pressurizer pressure are selected to maximize the DNBR degradation, and RCS flow rate was determined with the CETOP-D code, corresponding to DNBR SAFDL conditions at the assumed initial core power. For the purpose of computing SAFDL conditions, the radial peaking factor, FR, was set to a maximum value when obtaining the SAFDL conditions.
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| * The LOF involves a reduction in reactor coolant flow rate, which decreases the coolant mass flux and increases the coolant temperature in the core region. Use of a negative MTC value during this heatup would add negative reactivity, which in turn would tend to reduce reactor power and core heat flux. Therefore, for conservatism, the least negative MTC allowed by the Technical June 2009 15E-5 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E
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| * Specifications at Hot Full Power (HFP) was used in the analysis.
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| * The least negative Doppler fuel temperature coefficient curve, at Beginning of Cycle (BOC), was assumed. Least negative values minimize the addition of negative reactivity caused by increasing fuel temperature.
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| * BOC values were chosen to model delayed neutron kinetics.
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| The delayed fraction is larger at BOC values and results in a slower power response. This delays the decrease in core power during the flow coastdown and following the reactor trip causing the heat flux decreasing more slowly and causes a later turn-around of DNBR. Since the flow is decreasing with time, delaying the heat flux results in a lower flow rate at the time of minimum DNBR.
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| * If power generation in the core is shifted toward the bottom, the insertion of negative reactivity following reactor trip will be somewhat delayed until the CEAs have inserted farther into the core. The scram reactivity curve was therefore based on a positive ASI representing a bottom-peaked core. The time versus scram reactivity curve was adjusted to account for a 0.6-second CEA holding coil time delay following opening of the reactor trip breakers, and normalized to model 90% CEA insertion at 4.0 seconds after power is removed from the Control Element Drive Mechanism (CEDM) coils (see UFSAR Section 3.9.4).
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| In addition, the insertion of reactivity was delayed to account for the response time of the Reactor Protective System (RPS). This delay accounts for the time interval between when the CPCs would detect a low DNBR condition to June 2009 15E-6 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E
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| * the time at which electrical power to the CEDM coils would be interrupted.
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| * The CEA worth at trip represents the minimum scram worth for HFP conditions at BOC, assuming the most reactive CEA remains stuck out of the core following reactor trip.
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| This is more limiting (less scram worth) than the anticipated scram reactivity worth at other times during the operating cycle for HFP conditions.
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| * Parametric studies with the HERMITE code for a LOF event show that, for any given axial power distribution, the most limiting ROPM occurs with lower core inlet temperatures, higher pressures, higher core flow rates and lower fuel rod pellet-to-clad gap conductance. A low value for gap conductance, which delays the decay of the heat flux, was therefore selected for this analysis, corresponding to the maximum core average linear heat rate.
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| * For the limiting infrequent event, an additional single active component failure involving a LOP was modeled for the core performance safety analysis.
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| * There was no operator action for the first 30 minutes of the event.
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| June 2005 15E-7 Revision 13
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| PVNGS UPDATED FSAR APPENDIX 15E TABLE 15.E-1 PARAMETERS USED FOR THE LIMITING INFREQUENT EVENT PARAMETER Value Rated Thermal Power (MWt) 3990 Initial core power (% of RTP) 116 Initial core inlet temp (°F) 548 Initial pressurizer pressure (psia) 2325 Initial RCS flow (% of design) 108.8 MTC (/°F) -0.2E-4 FTC Least negative Kinetics Maximum CEA worth at trip - WRSO (%) -8.0 2
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| Fuel rod gap conductance (Btu/hr-ft -°F) Minimum Local Plugged SG tubes N/A Single failure None LOP Yes NOTE 1: The Local Minimum Fuel Rod Gap conductance (Hgap) is determined using the FATES code and is documented in a reload analyses calculation.
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| June 2009 15E-8 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E 15E.3.3 Results The Standard Review Plan provides a specific acceptance criterion for all AOOs with a single failure as:
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| An incident of moderate frequency in combination with any single active component failure, or single operator error, should not result in loss of function of any barrier other than the fuel cladding. A limited number of fuel rod cladding perforations are acceptable.
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| The safety analysis shows that the calculated minimum DNBR would approach a value of 1.17 at approximately 2.1 seconds following the LOP, which is well below the DNBR SAFDL of 1.34.
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| Figure 15E-1 depicts a representative hot channel DNBR transient for this limiting event. Within ~3.5 seconds, local and average core heat flux has decreased enough such that no pins experiencing DNB remain. Hence, DNB propagation is not predicted to occur.
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| The radial peaking factor used to calculate the minimum DNBR was 1.9243 for this limiting case. This corresponds to an initial peaking of 1.91 and reflects the peaking increase in the heat fluxes caused primarily by the coolant heat-up during the flow coast-down. Additional TORC cases were run to assess the sensitivity of the minimum DNBR value to changes in the radial peaking factor. The results of these additional cases, all of which used the same thermal-hydraulic conditions at the time of minimum DNBR predicted by CETOP-D, are shown in Table 15E-2.
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| June 2009 15E-9 Revision 15
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| PVNGS UPDATED FSAR APPENDIX 15E Table 15E-2 TYPICAL MINIMUM DNBR VERSUS RADIAL PEAKING FACTOR FOR THE LIMITING INFREQUENT EVENT CORE PERFORMANCE SAFETY ANALYSIS Radial Peaking Factor, FR, at the Time of Minimum DNBR Minimum DNBR 1.9243 1.17 1.7 1.44 1.5 1.74 1.3 2.11 Minimum DNBR versus FR data pairs similar to Table 15E-2 are calculated in cycle-specific reload analyses to provide a prediction of the DNBR propagation and amount of fuel failure for any proposed reload core design pin census (i.e., the distribution of power generation among fuel pins in a core).
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| Such fuel failure calculations use a statistical convolution technique that is described in Reference 2. This technique involves the grouping of fuel rods with respect to radial peaking factors; calculating the minimum DNBR in each radial pealing group; and then determining the probability of experiencing DNB corresponding to each minimum DNBR value. The number of fuel rods damaged within a radial peaking group is given by the number of fuel rods in that group, multiplied by the probability of experiencing DNB at that groups minimum DNBR value. Finally, summing up the damaged fuel rods in all radial peaking groups yields the total number of fuel rods damaged in the core.
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| June 2011 15E-10 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E NRC approval is based upon the use of conservative analytical assumptions, including a hot channel flow factor that does not exceed 73% of the core average assembly inlet flow. A hot channel flow factor of 70%, which includes additional voluntary conservatism beyond that required by the PVNGS licensing basis, was utilized to obtain the minimum DNBR versus Fr data pairs in Table 15E-2.
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| 15E.4 RCS PRESSURE BOUNDARY BARRIER PERFORMANCE The purpose of the analysis of this event is to bound DNBR degradation for infrequent events, including AOOs with active single failure. Thus, it does not address the reactor pressure boundary performance. Reactor pressure boundary barrier performance for infrequent events and AOOs with or without a single failure is addressed in their respective UFSAR sections.
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| 15E.5 CONTAINMENT PERFORMANCE AND RADIOLOGICAL CONSEQUENCES Offsite radiological consequences for limiting infrequent event were calculated for 2 hours at the Exclusion Area Boundary (EAB) and for 8 hours for the Low Population Zone (LPZ). The offsite dose calculation assumed 10% fuel failure to bound future fuel cycles.
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| The release path for iodine and noble gas activity consisted of releases by the MSSVs and controlled steaming through the Atmospheric Dump Valves (ADVs) on both steam generators during the cooldown. Due to a LOP, the condenser was unavailable and MSSVs and ADVs were employed to remove decay heat and cool down the RCS.
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| June 2011 15E-11 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E It was assumed that plant operators would not initiate a plant cooldown to SDC entry conditions for at least 30 minutes following event initiation.
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| Since this is an infrequent event, offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. Additionally, radiation exposures for control room personnel are subject to the limits specified in General Design Criterion 19 of 10 CFR 50 Appendix A.
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| Control room radiological assessments for bounding unfiltered in-leakage are presented in UFSAR Section 6.4.7. The results presented in that UFSAR section for a postulated RCP Sheared Shaft event with a stuck open ADV bound the anticipated control room exposure for the limiting infrequent event including AOOs with an active single failure. The RCP Sheared Shaft event was predicted to result in a higher percentage of fuel failure than the limiting infrequent event and, in combination with a stuck open ADV, the RCP Sheared Shaft event would result in a correspondingly higher control room dose than the limiting infrequent event.
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| The offsite radiological dose consequences associated with the limiting infrequent event are evaluated below.
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| 15E.5.1 Mathematical Models For the offsite radiological dose assessment, activity in the RCS was calculated on the basis of the pre-event radioiodine and noble gas activity levels (which are limited by Plant Technical Specifications), to which was added the anticipated post-event increase in activity levels due to fuel pin failures. The increase in activity levels due to fuel pin June 2011 15E-12 Revision 16
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| | |
| PVNGS UPDATED FSAR APPENDIX 15E failures is dependent upon the radial peaking factor, which affected the radionuclide inventory in the fuel rod gap, as well as the fuel failure fraction, which defined the number of pins that release radionuclides to the RCS coolant.
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| Once the activity level in the RCS was determined, the amount of activity carried over to the steam generators by primary-to-secondary leakage was calculated. Activity that leaks into the steam generators was assumed to mix with that steam generators secondary inventory. The level of activity in the generator increased as the event proceeded. The activity released from the steam generators to the environment was determined, based on a steaming rate that removed decay heat and the stored heat in the plant to successfully cool down the NSSS to SDC entry conditions. Once activity releases were quantified, the thyroid and whole body doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) were calculated as a function of the radial peaking factors and fuel failure fraction.
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| 15E.5.2 Input Parameters and Initial Conditions Offsite radiological dose consequences associated with the limiting infrequent event were analyzed under the following conditions:
| |
| : 1. Isotope inventories were based on a core power level of 102% of RTP.
| |
| : 2. Based on Technical Specification limits, the initial assumed contamination in the NSSS was:
| |
| * RCS Dose Equivalent (DEQ) I-131 1.0 µCi/gm
| |
| * RCS Noble Gas 100/ µCi/gm June 2011 15E-13 Revision 16
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| | |
| PVNGS UPDATED FSAR APPENDIX 15E
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| * Secondary System DEQ I-131 0.10 µCi/gm where is the average of the sum of the average beta ()
| |
| and gamma () energies per disintegration (in units of MeV), for noble gas isotopes with half lives greater than 15 minutes, weighted in proportion to the concentration of each isotope in the reactor coolant. Tables 15E-3 and 15E-4, respectively, identify the initial RCS iodine and noble gas source terms for the radioisotopes that were included in the analysis. Radioiodine dose conversion factors were set to the ICRP-30 (Reference 10) values listed in Table 15B-4, and the average and disintegration energies for each noble gas isotope were set to the values listed in Table 15B-1.
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| : 3. A RCS liquid mass of 555,000 lbm of water was used in the analysis, including 45,000 lbm of water in the pressurizer. Additionally, 4,500 lbm of steam was assumed to be in the pressurizer. Although the RCS may hold more mass, these values were selected to increase the iodine concentration following postulated fuel failures, which conservatively increases offsite dose consequences.
| |
| : 4. A steam generator liquid mass of 160,600 lbm per steam generator was used in the analysis. Although the steam generators can hold more mass, this value was selected to increase the iodine concentration in the unaffected steam generator, which conservatively increases offsite dose consequences.
| |
| : 5. A primary-to-secondary leak rate of 0.5 gpm (720 gallons per day) per steam generator was assumed. This is June 2011 15E-14 Revision 16
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| | |
| PVNGS UPDATED FSAR APPENDIX 15E consistent with the PVNGS Technical Specification limit for RCS leakage prior to issuance of Operating License Amendment No. 120 (Reference 11).
| |
| : 6. It was assumed that 10% of the iodine and noble gas inventories in the fuel pins were resident in the fuel rod pellet-to-clad gap, and available for release upon clad rupture.
| |
| : 7. All of the activity in the fuel rod gap was assumed to be released to the RCS coolant upon fuel pin failure.
| |
| : 8. Iodines associated with leakage to the steam generators were assumed to be released to the environment during steaming with a Decontamination Factor (DF) of 100.
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| : 9. It was assumed that plant operators would not initiate a plant cooldown to SDC entry conditions for at least 30 minutes following event initiation. However, it should be noted that a faster RCS cooldown rate would increase steam releases during the first two hours following the event, which would produce more severe thyroid doses at the EAB.
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| On the other hand, a slower RCS cooldown rate would allow radionuclide concentrations to build up in the secondary system, which would produce more severe 8-hour doses at the LPZ. Therefore, radiological dose calculations were performed using two different cooldown rates:
| |
| * A maximum Technical Specification cooldown rate of 100°F/hr, initiated at 30 minutes into the event sequence.
| |
| * A slower cooldown rate of 40oF/hr, initiated at 30 minutes into the event sequence, which would bring June 2011 15E-15 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E the RCS to SDC entry conditions at approximately 8 hours following event initiation.
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| : 10. Decay heat during the 8-hour period following the event was based on a 1979 ANS decay heat curve increased by an amount corresponding to the 2 of the uncertainty.
| |
| : 11. A value of 740,000 BTU/oF was used to represent the specific heat capacity of the RCS, the RCS clad and the steam generators. Use of a large value increases the amount of steam that must be released to the environment during the cooldown.
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| : 12. The /Q atmospheric dispersion factors used in the analysis are the short-term factors shown in Table 2.3-31.
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| : 13. Since the PSVs lift for this event, the dose calculation conservatively takes into account the activity released to containment, even though the Reactor Drain Tank is sized to remain intact from the PSV discharge.
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| 1 June 2011 15E-16 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E Table 15E-3 RCS IODINE SOURCE TERM FOR THE IOSGADVLOP OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Isotope Term(Ci/MWt)
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| I-131 25,100 I-132 38,100 I-133 56,220 I-134 65,760 I-135 51,040 Table 15E-4 RCS NOBLE GAS SOURCE TERM FOR THE IOSGADV OFFSITE RADIOLOGICAL DOSE SAFETY ANALYSIS Source Term Isotope (Ci/MWt)
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| Kr-83m 4,153 Kr-85 440 Kr-85m 13,000 Kr-87 21,540 Kr-88 32,020 Xe-131m 260 Xe-133 56,220 Xe-133m 1,384 Xe-135 53,640 Xe-135m 18,200 Xe-138 49,700 June 2011 15E-17 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E 15E.5.3 Results The results of the limiting infrequent event radiological dose analysis are shown in Table 15E-5 for radial peaking of 1.72 and a fuel failure fraction of 10.0%. These results bound RTP of 3990 MWt or less.
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| Table 15E-5 RADIOLOGICAL CONSEQUENCE OF LIMITING INFREQUENT EVENT RADIOLOGICAL ANALYSIS (FR = 1.72, 10% FUEL FAILURE)
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| Thyroid Dose (REM) Whole Body Dose (REM) 2 Hour EAB 8 Hour LPZ 2 Hour EAB 8 Hour LPZ 4.5 11.7 1.58 1.81 15E.6 CONCLUSIONS The limiting infrequent event (i.e., AOO with active single failure) results in a limited number of fuel pins predicted to be in DNB for a few seconds. DNB propagation is not predicted to occur. Offsite doses remained below the acceptance criteria for this category of event. Specifically, a small fraction of 10 CFR Part 100 guidelines (i.e., 30 REM thyroid, 2.5 REM whole body).
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| 15E.7 REFERENCES
| |
| : 1. C-E Methods for Loss of Flow Analysis, CENPD-183, July 1975.
| |
| : 2. C-E Methods for Loss of Flow Analysis, CENPD-183-A, June 1984.
| |
| : 3. Combustion Engineering, "HERMITE, A Multi-Dimensional Space-Time Kinetics Code for PWR Transients,"
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| CENPD-188-A, March 1976.
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| June 2011 15E-18 Revision 16
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| PVNGS UPDATED FSAR APPENDIX 15E
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| : 4. Combustion Engineering, "Responses to First Round Questions on the Statistical Combination of Uncertainties Program: CETOP-D Code Structure and Modeling Methods," CEN-139(A)-P, March 1981.
| |
| : 5. Combustion Engineering, "Responses to First Round Questions on the Statistical Combination of Uncertainties Program: CETOP-D Code Structure and Modeling Methods," CEN-124(B)-P, Part 2, May 1981.
| |
| : 6. Combustion Engineering, "CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3," CEN-160(S)-P, September 1981.
| |
| : 7. Intentionally Left Blank.
| |
| : 8. Combustion Engineering, "TORC Code: A Computer Code for Determining the Thermal Margin of a Reactor Core,"
| |
| CENPD-161-P-A (proprietary), CENPD-161-A (non-proprietary), April 1986.
| |
| : 9. Combustion Engineering, "TORC Code Verification and Simplified Modeling Methods," CENPD-206-P-A (proprietary), CENPD-206-A (non-proprietary), June 1981.
| |
| : 10. International Commission on Radiation Protection, Publication No. 30, Supplement to Part 1, Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity, 1980.
| |
| : 11. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 120 to Facility Operating License No.
| |
| NPF-41, Amendment No. 120 to Facility Operating License No. NPF-51, and Amendment No. 120 to Facility Operating License No. NPF-74, Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Docket Nos. STN 50-528, STN 50-529, and STN 50-530, August5, 199530, August 5, 1999.
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| June 2011 15E-19 Revision 16
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