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| document type = Updated Final Safety Analysis Report (UFSAR)
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 830
| page count = 830
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{{#Wiki_filter:FNP-FSAR-6    6-i REV 21  5/08 6.0  ENGINEERED SAFETY FEATURES  TABLE OF CONTENTS  Page  6.1 GENERAL  .............................................................................................................6.1-1 6.1.1 Safety Features Systems...................................................................................6.1-1 6.1.2 Operational Reliability........................................................................................6.1-2
6.2 CONTAINMENT SYSTEMS...........................................................................................6.2-1
6.2.1 Containment Functional Design.........................................................................6.2-1
6.2.1.1  Design Bases.................................................................................6.2-1 6.2.1.2  System Design...............................................................................6.2-4 6.2.1.3  Design Evaluation..........................................................................6.2-5 6.2.1.4  Containment Testing and Inspection...........................................6.2-32 6.2.1.5  Instrumentation Requirements.....................................................6.2-36 6.2.1.6  Materials......................................................................................6.2-36 6.2.1.7  Heavy Load Safe Load Paths......................................................6.2-37
6.2.2 Containment Heat Removal Systems..............................................................6.2-37
6.2.2.1  Design Bases...............................................................................6.2-37 6.2.2.2  System Design.............................................................................6.2-38 6.2.2.3  Design Evaluation........................................................................6.2-42 6.2.2.4  Testing and Inspection.................................................................6.2-44 6.2.2.5  Instrumentation Requirements.....................................................6.2-46 6.2.2.6  Materials......................................................................................6.2-48
6.2.3 Containment Air Purification and Cleanup  Systems.....................................................................................6.2-48
6.2.3.1  Design Bases...............................................................................6.2-49 6.2.3.2  System Design.............................................................................6.2-55 6.2.3.3  Design Evaluation........................................................................6.2-63 6.2.3.4  Tests and Inspections..................................................................6.2-66 6.2.3.5  Instrumentation Requirements.....................................................6.2-70 6.2.3.6  Materials......................................................................................6.2-72
6.2.4 Containment Isolation System.........................................................................6.2-72
6.2.4.1  Design Bases...............................................................................6.2-72 6.2.4.2  System Design.............................................................................6.2-72 6.2.4.3  Design Evaluation........................................................................6.2-74 6.2.4.4  Tests and Inspections..................................................................6.2-74 6.2.4.5  Materials......................................................................................6.2-75 6.2.5 Combustible Gas Control in Containment...............................................6.2-75 6.2.5.1 Design Bases..........................................................................................6.2-75 6.2.5.2 System Design........................................................................................6.2-78
FNP-FSAR-6  TABLE OF CONTENTS    Page  6-ii REV 21  5/08  6.2.5.3 Design Evaluation....................................................................................6.2-82  6.2.5.4 Tests and Inspections..............................................................................6.2-84 6.2.5.5 Instrumentation Requirements................................................................6.2-85 6.2.5.6 Materials..................................................................................................6.2-85
6.3 EMERGENCY CORE COOLING SYSTEM................................................................6.3-1 6.3.1 Design Bases............................................................................................6.3-1
6.3.1.1 Range of Coolant Ruptures and Leaks.....................................................6.3-1 6.3.1.2 Fission Product Decay Heat . . . . .............................................................6.3-2 6.3.1.3 Reactivity Required for Cold Shutdown.....................................................6.3-2 6.3.1.4 Capability to Meet Functional Requirements.............................................6.3-2
6.3.2 System Design..........................................................................................6.3-3
6.3.2.1 Schematic Piping and Instrumentation Diagrams......................................6.3-3 6.3.2.2 System Components.................................................................................6.3-3 6.3.2.3 Applicable Codes and Classifications......................................................6.3-11 6.3.2.4 Materials Specifications and Compatibility..............................................6.3-11 6.3.2.5 Design Pressures and Temperatures. . ..................................................6.3-12 6.3.2.6 Coolant Quantity .....................................................................................6.3-12 6.3.2.7 Pump Characteristics .............................................................................6.3-13 6.3.2.8 Heat Exchanger Characteristics..............................................................6.3-13 6.3.2.9 ECCS Flow Diagrams..............................................................................6.3-13 6.3.2.10 Relief Valves............................................................................................6.3-13 6.3.2.11 System Reliability ...................................................................................6.3-13 6.3.2.12 Protection Provisions...............................................................................6.3-15 6.3.2.13 Provisions for Performance Testing .......................................................6.3-16 6.3.2.14 Net Positive Suction Head (NPSH).........................................................6.3-16 6.3.2.15 Control of Motor-Operated Isolation Valves............................................6.3-17 6.3.2.16 Motor-Operated Valves and Controls......................................................6.3-18 6.3.2.17 Manual Actions........................................................................................6.3-18 6.3.2.18 Process Instrumentation..........................................................................6.3-18 6.3.2.19 Materials..................................................................................................6.3-18
6.3.3 Performance Evaluation..........................................................................6.3-19
6.3.3.1 Evaluation Model.....................................................................................6.3-19 6.3.3.2 ECCS Performance.................................................................................6.3-19
6.3.3.3 Alternate Analysis Methods.....................................................................6.3-19 6.3.3.4 Fuel Rod Perforations..............................................................................6.3-20
FNP-FSAR-6  TABLE OF CONTENTS    Page  6-iii REV 21  5/08  6.3.3.5 Evaluation Model.....................................................................................6.3-20  6.3.3.6 Fuel Clad Effects.....................................................................................6.3-20 6.3.3.7 ECCS Performance.................................................................................6.3-20 6.3.3.8 Peaking Factors.......................................................................................6.3-20 6.3.3.9 Fuel Rod Perforations..............................................................................6.3-20 6.3.3.10 Conformance with Interim Acceptance Criteria.......................................6.3-20 6.3.3.11 Effects of ECCS Operation on the Core..................................................6.3-20 6.3.3.12 Use of Dual Function Components..........................................................6.3-20 6.3.3.13 Dependence on Other Systems..............................................................6.3-21 6.3.3.14 Lag Times................................................................................................6.3-22 6.3.3.15 Thermal Shock Considerations...............................................................6.3-23 6.3.3.16 Limits on System Parameters..................................................................6.3-23
6.3.4 Tests and Inspections..............................................................................6.3-23 6.3.5 Instrumentation Requirements................................................................6.3-26
6.4 HABITABILITY SYSTEMS..........................................................................................6.4-1
6.4.1 Habitability Systems Functional Design....................................................6.4-1
6.4.1.1 Design Bases............................................................................................6.4-1 6.4.1.2 System Design..........................................................................................6.4-2 6.4.1.3 Design Evaluations....................................................................................6.4-6 6.4.1.4 Testing and Inspection..............................................................................6.4-6 6.4.1.5 Instrumentation Requirement....................................................................6.4-7
6.5 AUXILIARY FEEDWATER SYSTEM..........................................................................6.5-1
6.5.1 Design Bases............................................................................................6.5-1 6.5.2 System Description....................................................................................6.5-2
6.5.2.1 General Description...................................................................................6.5-2 6.5.2.2 Component Description.............................................................................6.5-3 6.5.2.3 System Operation......................................................................................6.5-5
6.5.3 Design Evaluation......................................................................................6.5-7 6.5.4 Tests and Inspection.................................................................................6.5-7 6.5.5 Instrumentation..........................................................................................6.5-8
APPENDIX 6A MATERIALS COMPATIBILITY REVIEW...................................................6A-1
APPENDIX 6B CONTAINMENT PRESSURE ANALYSIS.................................................6B-1 
FNP-FSAR-6  TABLE OF CONTENTS    Page  6-iv REV 21  5/08 APPENDIX 6C CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE  COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM.    (Historical - Prior to December 2007).......................................................6C-1  APPENDIX 6D CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE      COOLING SYSTEM RECIRCULATION SUMP STRAINER DESIGN........6D-1 FNP-FSAR-6  LIST OF TABLES    6-v REV 21  5/08 6.2-1 Principal Containment Design Parameters 6.2-2 Heat Sink Geometric Data 
6.2-3 Initial Conditions for Pressure Analysis 
6.2-4 Heat Sink Thermodynamic Data 
6.2-5 Engineered Safety Features Performance for Containment Pressure Transient Analysis 6.2-6 Containment Pressure Analysis Results for the Spectrum of RCS Break Sizes 
6.2-7 System Parameters, Initial Conditions for Thermal Uprate
6.2-8 Safety Injection Flow - Minimum Safeguards
6.2-9 Safety Injection Flow - Maximum Safeguards
6.2-10 Double-Ended Hot Leg Break, Blowdown Mass and Energy Releases
6.2-11 Plant Data for Blowdown 
6.2-12 Double-Ended Hot Leg Break, Mass Balance
6.2-13 Double-Ended Hot Leg Break, Energy Balance
6.2-14 Double-Ended Pump Suction Break, Blowdown Mass and Energy Releases
6.2-15 Reactor Cavity Release 
6.2-16 Spray Line Break Release 
6.2-17 Surge Line Break Release 
6.2-18 Reactor Cavity Subcompartment Pressure Analysis Summary of Flow Paths and Vent Loss Coefficients 6.2-19 Containment Results for the Design Basis LOCA
6.2-20 Double-Ended Pump Suction Break - Minimum Safeguards, Reflood Mass and Energy Releases 6.2-21 LOCA Chronology of Events 
6.2-22 Subcompartment Differential Pressure Results 
6.2-23 Environmental Conditions for Containment Heat Removal Systems (Deleted) 6.2-24 Component Design Parameters for Containment Spray System and Containment Cooling System FNP-FSAR-6  LIST OF TABLES    6-vi REV 21  5/08 6.2-25 Regulatory Guide 1.52, Section Applicability for the Penetration Room Filtration System  6.2-26 Single Failure Analysis - Containment Spray System  6.2-27 Double-Ended Pump Suction Break - Minimum Safeguards, Blowdown Mass and Energy Releases 6.2-28 Containment Ventilation Systems Component Design Parameters 6.2-29 Spray Evaluation Parameters 
6.2-30 Single Failure Analysis - Penetration Room Filtration System
6.2-31 Containment Isolation Valve Information
6.2-32 Steam Generator Isolation Valve Information 
6.2-33 Electric Hydrogen Recombiner Typical Parameters 
6.2-34 Postaccident Venting System Design Parameters 
6.2-35 Postaccident Sampling System Design Parameters 
6.2-36 Postaccident Mixing System Design Parameters 
6.2-37 Containment Interior Coatings Summary
6.2-38 Containment Penetrations
6.2-39 Containment Isolation Valves 
6.2-40 Steam Generator Isolation Valves 
6.2-41 Containment Pressure/Temperature for 600 gal/min Service Water Flow, 0.003 Fouling Factor 6.2-42 Double-Ended Pump Suction Break - Minimum Safeguards, Principle Parameters During Reflood 6.2-43 Double-Ended Pump Suction Break - Minimum Safeguards, Post-Reflood Mass and Energy Releases 6.2-44 Double-Ended Pump Suction Break, Mass Balance, Minimum Safeguards
6.2-45 Double Ended Pump Suction Break, Energy Balance, Minimum Safeguards
6.2-46 Double-Ended Pump Suction Break - Maximum Safeguards, Reflood Mass and Energy Releases 6.2-47 Double-Ended Pump Suction Break - Maximum Safeguards, Principle Parameters During Reflood 
FNP-FSAR-6  LIST OF TABLES    6-vii REV 21  5/08 6.2-48 Double-Ended Pump Suction Break - Maximum Safeguards, Post-Reflood Mass and Energy Releases  6.2-49 Double-Ended Pump Suction Break, Mass Balance, Maximum Safeguards
6.2-50 Double-Ended Pump Suction Break, Energy Balance, Maximum Safeguards
6.2-51 Double-Ended Hot Leg Break, Sequence of Events
6.2-52 Double-Ended Pump Suction Break - Minimum Safeguards, Sequence of Events
6.2-53 Double-Ended Pump Suction Break - Maximum Safeguards, Sequence of Events
6.2-54 LOCA Mass and Energy Release Analysis, Core Decay Heat Fraction
6.3-1 Emergency Core Cooling System Component Parameters 
6.3-2 ECCS Relief Valve Data 
6.3-3 Sequence of Changeover Operation from Injection to Recirculation (Deleted)
6.3-4 Time Analysis for ECCS Injection/Recirculation Switchover
6.3-5 Materials Employed for Emergency Core Cooling System Components 6.3-6 Normal Operating Status of Emergency Core Cooling 
6.3-7 Single Active Failure Analysis for Emergency Core Cooling System Components 
6.3-8 Maximum Potential Recirculation Loop Leakage External to Containment 
6.3-9 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis 
6.3-10 Emergency Core Cooling System Shared Functions Evaluation
6.5-1 Auxiliary Feedwater System Auxiliary Feedwater Pump Data 6.5-2 Failure Analysis of Auxiliary Feedwater System 
6.5-3 Auxiliary Feedwater System Motor Operated Valve Data 
FNP-FSAR-6  LIST OF FIGURES    6-viii REV 21  5/08 6.2-1 DEPSGB, Minimum ESF 1 AC Pressure vs. Time, PO = 0 PSIG    6.2-2 RSG DEPSG Minimum ESF 1 AC Pressure vs. Time, PO = 3 PSIG 6.2-3 DEHL, Minimum ESF, DBA Short Term Pressure vs. Time, PO = 0 PSIG 6.2-4 RSG DEHLG Minimum ESF, DBA Short Term Pressure vs. Time, PO = +3 PSIG 6.2-5 DECLG Maximum ESF Pressure vs. Time (Deleted)
6.2-6 RSG Pressure vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-6A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-7 RSG Temperature vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-7A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-8 Pressure vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-9 Temperature vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-10 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-10A Pressure vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-11 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-11A Temperature vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-12 Pressure vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-13 Temperature vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-14 Pressure vs. Time Steam Line Full D. E. Break 70% Power
6.2-15 Temperature vs. Time Steam Line Full D. E. Break 70% Power
6.2-16 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-17 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-18 Pressure vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-19 Temperature vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-20 RSG Pressure vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = +3 PSIG 6.2-21  RSG Temperature vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = -1.5 PSIG 6.2-22 RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, PO = -1.5 PSIG 
FNP-FSAR-6  LIST OF FIGURES    6-ix REV 21  5/08 6.2-22A RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, PO = +3 PSIG  6.2-23 RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power, PO = -1.5 PSIG 6.2-23A RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power, PO = +3 PSIG 6.2-24 Pressure vs. Time Steam Line 0.5 ft2 D. E. Break 30% Power 6.2-25 Temperature vs. Time Steam Line 0.5 ft2 D. E. Break 30% Power 6.2-26 Pressure vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = 0 PSIG 6.2.26A Pressure vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = -1.5 PSIG 6.2-27 Temperature vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = 0 PSIG 6.2-27A Temperature vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = -1.5 PSIG 6.2-28 RSG Pressure vs. Time Steam Line 0. 60 ft2 Split 30% Power, PO = -1.5 PSIG 6.2-28A RSG Pressure vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = +3 PSIG 6.2-29 RSG Temperature vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = -1.5 PSIG 6.2-29A RSG Temperature vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = +3 PSIG 6.2-30 RSG Pressure vs. Time Steam Line Full D. E. Break Hot Standby, PO = +3 PSIG 6.2-31 RSG Temperature vs. Time Steam Line Full D. E. Break Hot Standby, PO = -1.5 PSIG 6.2-32 Pressure vs. Time Steam Line 0.2 ft2 D. E. Break Hot Standby 6.2-33 Temperature vs. Time Steam Line 0.2 ft2 D. E. Break Hot Standby 6.2-34 Pressure vs. Time Steam Line 0.1 ft2 D. E. Break Hot Standby 6.2-35 Temperature vs. Time Steam Line 0.1 ft2 D. E. Break Hot Standby 6.2-36 Pressure vs. Time Steam Line 0.30 ft2 Split Hot Standby 6.2-37 Temperature vs. Time Steam Line 0.30 ft2 Split Hot Standby 6.2-38 TS, Equipment Surface Temperature with Uchida Condensing Heat Transfer and Convective Heat Transfer Coefficient of 2 Btu/h/ft2 (Deleted) 6.2-39 DEPSGB Minimum ESF 1 AC P/T Analysis Long-Term Containment Pressure vs. Time (Deleted) 6.2-40 DEPSGB Min ESF DBA Temperature vs. Time, PO = 0 PSIG 
FNP-FSAR-6  LIST OF FIGURES    6-x REV 21  5/08 6.2-41 RSG DEPSG Min ESF DBA Temperature vs. Time, PO = 3 PSIG  6.2-42 Containment Air Cooler Duty vs. Temperature 
6.2-43 Thermal Heat Removal Efficiency of Containment Atmosphere Spray (Deleted)
6.2-44 Residual Heat Exchanger Design Duty Accident Mode 
6.2-45 Mass and Energy Rate vs. Time for LOCA (Deleted)
6.2-46 LOCA Blowdown Mass and Energy Release Rates vs. Time (Deleted)
6.2-47 LOCA Post-Blowdown Mass and Energy Release Rates vs. Time (Deleted)
6.2-48 DEPSG Minimum ESF 1 AC P/T Analysis, Long Term Condensing Heat Transfer Coefficient (RSG) 6.2-49 Short Term Condensing Heat Transfer Coefficient for DBA (Deleted)
6.2-50 Reactor Cavity Model 
6.2-51 Reactor Cavity Block Diagram
6.2-52 Total Horizontal Force vs. Time 
6.2-53 Steam Generator Block Diagram 
6.2-54 Steam Generator Compartment C Differential Pressure vs. Time 
6.2-55 Pressurizer Compartment Pressure Model (Spray Line Break in Lower Compartment) 6.2-56 Pressurizer Compartment Flow Model 
6.2-57 Pressurizer Compartment Spray Line Results 
6.2-58 Node Pressures in Compartments 1 and 2 vs. Time 
6.2-59 Node Pressures in Compartments 3, 4, 5, and 6 vs. Time 
6.2-60 Node Pressures in Compartments 7, 8, 9, and 10 vs. Time 
6.2-61 Node Pressures in Compartments 11, 12, 13, and 14 vs. Time 
6.2-62 Node Pressures in Compartments 15, 16, and 17 vs. Time 
6.2-63 Node Pressures in Compartments 18, 19, 20, and 21, vs. Time 
6.2-64 Node Pressures in Compartments 22, 23, 24, 25, 26, and 27 vs. Time 
6.2-65 Node Pressures in Compartments 28, 29, 30, 31, 32, 33, and 34 vs. Time 
FNP-FSAR-6  LIST OF FIGURES    6-xi REV 21  5/08 6.2-66 Schematic of Reflood Code 19 Element Loop Model for a Pump Suction Break (Deleted)  6.2-67 Core Reflood Correlation (Deleted)
6.2-68 Comparison of Measured and Predicted Carryover Rate Fractions (Deleted)
6.2-69 Inlet Water Temperature vs. Time After End of Blowdown  (Deleted)
6.2-70 Variation in Temperature Rise, Turnaround Time, and Quench Time with Respect to Core Elevation (Deleted) 6.2-71 Energy Balance Model (Deleted)
6.2-72 Reflood Rate and Carryover Fractions vs. Time After End of Blowdown (Deleted)
6.2-73 Flow Through Break vs. Time After End of Blowdown (Deleted)
6.2-74 Water Height vs. Time After End of Blowdown (Deleted)
6.2-75 Post-Reflood Loop Resistance Model (Deleted)
6.2-76 S/G Internal Energy vs. Time After Break (Deleted)
6.2-77 Energy Distribution vs. Time (Deleted)
6.2-78 RSG Temperature Profile Through Containment Wall, PO = +3 PSIG 6.2-79 RHR Heat Exchanger Duty vs. Time, RSG PO = +3 PSIG 6.2-80 Containment Air Cooler Duty vs. Time, RSG PO = +3 PSIG 6.2-81 Minimum Sump pH Following LOCA vs. Time (Deleted)
6.2-82 Minimum Partition Coefficient in the Sump vs. Solution pH (Deleted)
6.2-83 Hydrogen Generation Rate vs. Time in the Lower Compartment 
6.2-84 Isolation Valve Arrangement through 6.2-89
6.2-90 Electric Hydrogen Recombiner 
6.2-91 Electric Hydrogen Recombiner Schematic Diagram (Typical of One Recombiner) 
6.2-92 Lower Compartment Plan 
6.2-93 Section of Lower Reactor Compartment 
6.2-94 Containment Hydrogen Concentration With One Electric Recombiner Started 1 Day after a LOCA 6.2-95 Hydrogen Concentration as a Function of Time in Containment Purge Mode FNP-FSAR-6  LIST OF FIGURES    6-xii REV 21  5/08 6.2-96 Volume Percent Hydrogen vs. Time in the Upper Containment (Unmixed), Outer Periphery (Unmixed), and Bulk Containment (Mixed)  6.2-97 Volume Percent Hydrogen vs. Time in the Lower Compartment 
6.2-98 Hydrogen Generation Rate vs. Time in Outer Periphery and Overall Containment 
6.3-1 Residual Heat Removal Pump Performance Curves 
6.3-2 Charging Pump Performance Curves 
6.3-3 RHR Pump Characteristic Curves 
6.3-4 Containment Spray Pump Characteristic Curves 
FNP-FSAR-6 6.1-1 REV 21  5/08 6.0  ENGINEERED SAFETY FEATURES  6.1 GENERAL  Engineered safety features are structures and equipment required to mitigate design basis accidents including the loss of coolant accident and high energy pipe breaks such as a steam pipe break and a main feedwater pipe break. Engineered safety features are designed to Seismic Category I requirements. They are designed to perform their safety function with complete loss of offsite power. Such equipment is provided with sufficient redundancy that failure of a single component will not result in the loss of the safety function. Engineered safety features fulfill the following safety functions under accident conditions:
A. Protect the fuel cladding. 
B. Ensure containment integrity. 
C. Minimize containment leakage. 
D. Remove fission products from the containment atmosphere. 
The operator action times assumed in this chapter include conservative actions to provide an adequate safety margin for the purpose of nuclear safety system design and nuclear safety analysis of the design basis events. However, they are not intended to serve as a basis for actual operator action times in procedures or training. The assumed time periods are considered in the basis of plant design to permit credit for operator actions. The Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGs) provide a basis for operatior action in response to design basis accidents. 6.1.1 SAFETY FEATURES SYSTEMS The safety features systems provided to satisfy the functions listed above are as follows:
* Containment isolation system (subsection 6.2.4).
* Containment spray system (subsection 6.2.2).
* Containment fan cooler system (subsection 6.2.2).
* Containment air purification and cleanup system (subsection 6.2.3).
* Emergency core cooling system (section 6.3).
* Residual heat removal system (section 5.5).
* Combustible gas control in containment (subsection 6.2.5).
* Penetration room filtration system (section 6.2).
FNP-FSAR-6 6.1-2 REV 21  5/08
* Auxiliary feedwater system (section 6.5).
The fuel cladding is protected by the timely, continuous, and adequate supply of borated water to the reactor coolant system (RCS) and, ultimately, the reactor core. This supply of water is provided by the emergency core cooling system (ECCS). These systems provide high head (centrifugal charging pumps), low head (residual heat removal pumps) injection, and accumulator injection immediately following an incident, and low head/high head recirculation in the long term recovery period. 
The containment integrity is ensured and the containment leakage is minimized by the provision of means for condensing the steam inside the containment, depressurizing the containment following an incident, and maintaining the containment at near atmospheric conditions for an extended period of time. The containment isolation system, spray system, fan cooler system, and the electric hydrogen recombiners provide the means for satisfying these requirements. 
The fission products are removed from the containment atmosphere by the chemical spray additive which enhances the removal of radioactive iodine from the containment atmosphere following an incident. The containment air purification and cleanup systems are provided to meet this function. 
The safety features systems are designed with sufficient redundancy to meet the general design criteria as discussed in sections 3.1, 3.2, and subsection 6.3.2.11. Electrical power for all safety features systems is provided both from offsite sources and from emergency onsite sources as described in sections 8.2 and 8.3, respectively. 
Safety features are separated into two independent trains of equal capability. Either train can handle the entire emergency coolant injection and emergency cooling loads; either train can provide the entire containment isolation, containment cleanup, and containment leakage minimization functions. Each train has an independent onsite and offsite power source. Failure of either train cannot affect the other. 
Some of high and low pressure emergency injection systems use equipment that serves normal functions during normal plant operation or shutdown. Observation of their normal functioning provides monitoring of equipment availability and condition. In cases where equipment is used for emergencies only, systems are designed to permit periodic inspection and tests. 
6.1.2 OPERATIONAL RELIABILITY  Operational reliability is achieved by using proven components and by conducting tests required by the quality control requirements presented in chapter 17.0. All safety features systems are quality items meeting the requirements of 10 CFR 50, Appendix B, and seismically designed as discussed in chapter 3.0. Those safety features essential for post-tornado safety are designed to survive without loss of function the design tornado described in section 3.3. 
Other sections of this report contain additional information on the safety features systems.
Information on seismic requirements is provided in chapters 2.0 and 3.0. Information on the actuation instrumentation of the safety features system is provided in chapter 7.0.
FNP-FSAR-6 6.1-3 REV 21  5/08 Information on functions performed by components of the safety features systems during normal plant operation is provided in chapters 9.0 and 5.0. The safety analysis and demonstration of the ability of the safety features systems to provide adequate protection during accident conditions as provided in chapter 15.0. 
The design bases, design description and evaluation, tests, inspections, and instrumentation for the safety features systems are presented in this chapter. 
[HISTORICAL] [Tests on Liner During Construction    Inspection procedures employed during construction for the liner seam welds, liner fastening, and around penetrations consist of visual inspection, vacuum box soap bubble testing, radiography, dye-penetrant testing, and magnetic particle inspection. 
A. Visual Inspection of Welds All of the welding is visually examined by a technician responsible for welding quality control. The basis for visual quality of welds is as follows:
1. Each weld is uniform in width and size throughout its full length. Each layer of welding shall be smooth and free of slag, cracks, pinholes, and undercut and shall be completely fused to the adjacent weld beads and base metal. In addition, the cover pass is free of coarse ripples, irregular surface, nonuniform head pattern, high crown, and deep ridges or valleys between beads. Peening of welds is not permitted, except for light peening for cleaning purposes.
2. Butt welds are of multipass construction, slightly convex, of uniform height, and have full penetration.
3. Fillet welds are of the specified size, with full throat and legs of uniform length.
B. Soap Bubble Tests All of the welding required for containment integrity is vacuum box soap bubble tested except where the structural configuration or space limitation does not allow. In this test a vacuum box containing a window is placed over the area to be tested and is evacuated to produce at least 5 psi pressure differential. Before the vacuum box is placed over the test area, a soap solution is applied to the weld and any leaks will be indicated by bubbles observed through the window in the box.
C. Radiography Radiography is used as an aid to quality control. The primary purpose of the liner plate and the welds therein is to provide leaktightness integrity to the posttensioned concrete 
containment. Structural integrity of the containment will be provided by the posttensioned concrete and not by the liner plate. Radiography is not recognized as a completely effective method for examining welds to assure leaktightness. Therefore, the maximum benefit expected from radiography in connection with obtaining leaktight welds will be as an aid to quality control. Random radiography of each welder's work will provide verification that the welding is under control and being done in accordance with the previously established and qualified procedures. Additionally, employing random radiography to inspect each welder's work has been proved by past experience to have a positive psychological effect on the improving overall welding workmanship.
For quality control purposes, at least one spot radiograph 12 inches long was taken in the first ten feet of welding completed in the flat, vertical, horizontal, and overhead positions by each welder on liner plate welds. No further welding was permitted until initial radiographic inspection has been satisfactorily completed and the welding found to be acceptable by the Inspector. Thereafter, a minimum of 2 percent of the welding was progressively spot examined as welding is performed, using film 12 in. long, on a random basis to be specified by the inspector, in such a manner that an approximately equal number of spot radiographs was taken from the work of each welder. In addition to the 2 percent radiograph, 18 percent of the welding was nondestructively examined. Under conditions where two or more welders make weld layers in a joint or on the two sides of a double-welded butt joint, one spot examination represented the work for both welders. Where a radiograph discloses welding which did not comply with the minimum quality requirements, as defined in paragraph UW52, Section VIII of ASME code, two additional spots, each 12 inches long, were examined in the same weld seam at locations away from the original spot. The locations of these additional spots were determined by the Inspector as provided for the original spot examination. If two additional spots examined showed welding which met the minimum quality requirements, the entire weld represented by the three radiographs was acceptable. The defective welding disclosed by the first of the three radiographs was removed and repaired by welding.
If either of the two additional spots examined showed welding which did not comply with minimum quality requirements, the entire portion of the seam represented was rejected; or, at the fabricator's option, the entire weld represented was completely radiographed, and defective welding corrected.
The rewelded joints or weld-repaired areas were completely reradiographed and met the weld quality requirements cited above.
D. Dye-Penetrant and Magnetic-Particle Inspection Dye-penetrant and magnetic-particle inspection were used as an aid to quality control. The field welding inspectors used dye-penetrant or magnetic-particle inspection to closely examine welds judged to be of questionable quality on the basis of the initial 
visual inspection. Dye-penetrant or magnetic-particle inspection of liner plate welds were in accordance with Section VIII of the ASME Boiler and Pressure Vessel Code.]   
[HISTORICAL][For Unit 1 there are three 3/8-in. diameter holes between the solid cover plate on the top of the sump screen and the bioshield wall for venting of air during the initial phase of the LOCA when the water level in the sump rises. The slot size varies from approximately 1/4 in. to 1 in. across its length of approximately 3 ft. The potential for debris to enter through this path has been evaluated. The location of the slot near the shield wall was specifically selected to minimize the potential for debris to enter the sump. Since this slot and the vent holes will be under water during the recirculation phase of a 
LOCA, the debris entering through this path will sink to the sump floor due to low approach velocities near the bioshield wall and will not be swept into the opening of the intake pipe.] 
[HISTORICAL]  [Initial tests and the purpose of each test are listed as follows:    A. Component qualification tests - These tests demonstrate the characteristics of materials to be incorporated by the manufacturer into components for a system and ensure that they meet the requirements of procurement specification. The design conditions, which form the basis for these component qualification tests, are presented in table 3.11-1.
B. Component acceptance tests - These tests are factory tests which demonstrate the capability of the components incorporated in the various systems in which they are to operate. For example, fans associated with safeguards systems are tested in the manufacturer's shop to determine their characteristic curves. System valves are tested in the shop to verify effectiveness of seal, opening and closing periods, and the ability of the valve operator to actuate the valve at the maximum anticipated differential pressure.
Test results on actual or similar types of filter assemblies demonstrate their adequacy for this application. The following demonstrative tests are performed: 
A. Radioactive iodine removal efficiency  - a charcoal sample 2 in. in diameter by 2 in. deep is exposed to air flow at 40 ft/min face velocity. The air stream contains concentrations of elemental iodine and methyl iodide, similar to those predicted to occur in the penetration room filters during faulted conditions. Air stream temperature is 150°F, relative humidity 70 percent, and test duration is 12 hours. The efficiency is determined by measuring the activity of iodines upstream and downstream of the sample. Minimum acceptable efficiency is 99.0 percent at the end of 12 hours.
B. Flow resistance test - A module consisting of three absorbine units (six trays), stacked vertically, is capable of filtering 100 ft3/min of air at a pressure drop not exceeding 1.0 in. wg. The actual resistance is recorded and kept available.
C. Leak test - Each filter element is tested for 5 minutes in an air flow of 330 ft3/min containing approximately 20 ppm of Freon 112. Instrumentation is provided to measure the relative upstream and downstream concentrations of Freon 112. A downstream concentration in excess of 0.2 percent of the upstream concentration shall cause rejection of the filter.
D. Carbon lot tests - A sample from each lot of carbon after impregnation will have been subjected to the following tests by the manufacturer and results made a matter of permanent record:
1. Gas life - A bed of carbon 2 in. deep and 2 in. in diameter is tested for iodine collection at a velocity of 40 ft per minute (air at standard conditions). The iodine concentration upstream of the bed is 1000 mg/m3 and the penetration does not exceed 1.0 percent for a period of no less than 12 hours.   
2. Wash test - 250 ml of demineralized water is brought to a minimum boil. Twenty five grams of impregnated carbon is added to the demineralized water and the minimum boil is maintained for 1 minute. After 1 minute the water is decanted from the carbon and analyzed for the impregnate. With a knowledge of initial impregnate loading in the carbon and quantity of impregnate removed by the boiling water, results are reported as percentages of impregnate retained.
3. Ignition temperature - A sample from each lot of carbon is tested for ignition temperature in accordance with the procedure described in USNRC Report DP-1075, "High Temperature Adsorbents for Iodine," by R.C. Milhans.
4. Carbon tetrachloride test - Samples of carbon are tested for carbon tetrachloride adsorption capacity. Testing follows the procedures described in paragraph 6.2 of Military Specification MIL-C-17605.
Systems acceptance tests - Deenergized and energized tests demonstrate the proper mounting of components, proper hookup of circuits and connection, setting of instrumentation and operation of interlocks. Equipment and system performance are monitored and rated.
For the penetration room filtration system, all ducting is given a pneumatic pressure test prior to the installation of the filter elements to assure leak-tight construction. Dimensional tolerances on filter assemblies and frame assemblies are checked to ensure that suitable gasket compression is uniformly achieved on the filter sealing faces. 
A test program is performed after construction tests are completed to demonstrate the following:
A. Proper actuation of control circuitry in both modes. 
B. Proper flow path alignment in both modes. 
C. Leaktightness of each filter assembly. 
D. Verification that a negative pressure is maintained in the spent fuel area with the penetration room filtration system operating in the fuel handling area.
The following tests are performed prior to installation of the filter elements and charcoal bed. A test assembly is installed to simulate filter pressure drop. 
A. Simulate an actuation signal and observe the performance of the system in the LOCA mode.
B. In the LOCA mode, measure the discharge flow from the exhaust fan. At steady state conditions with the penetration room sealed, this corresponds to the penetration room leak rate.
C. In the LOCA mode, verify that the recirculation fan recirculation valve opens on receipt of a differential pressure signal from two out of three differential pressure instruments between the penetration rooms and pressure in the filtration system equipment room.   
D. With the system operating, verify circulation of air within the penetration rooms in the LOCA mode. E. Install the roughing filter and high-efficiency filter. With the systems operating, test leaktightness and performance, using DOP smoke of 0.3 micron mass median diameter.
Penetration should not exceed 0.1 percent.
F. Install the charcoal beds. With the system operating, test the performance, using Freon 112. The test is performed using similar portable equipment described in USNRC Report ORNL-NSIC-65, 1970, by C. A. Burchsted and A. B. Fuller, entitled "Design, Construction, and Testing of High Efficiency Filtration Systems for Nuclear Application" (paragraph 7.5.1, pages 7.8 - 7.9). The testing procedure is in accordance with a paper by D. R. Muhlbaier, "Standardized Non-Destructive Test of Carbon Beds for Reactor Containment Applications," DP-1082, July 1967. Test results must demonstrate removal of 99.5 percent of the Freon 112. The pressure drop is also measured.
G. Simulate a spent fuel pool high radiation signal and observe system performance in the fuel handling mode.
H. Simulate a spent fuel pool low differential pressure and observe system performance in the fuel handling mode.
I. With the system operating in the fuel handling mode, verify that there is a vacuum in the spent fuel pool.
In addition, all instruments are calibrated, alarms, controls, and interlocks checked, and each remotely operated valve is individually stroked to determine its operability and correct performance of indicating lights. 
The inleakage characteristics of the penetration boundary are determined by means of a flowmeter in the supply ducting to the penetration room filtration system filters and a vacuum gauge in the penetration room. With all normally operating ventilation systems in the auxiliary building secured, the internal pressure in the penetration rooms and the exhaust air flow provides the data necessary to ascertain the leaktightness of the joints, partitions, and seals.]
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-1  PRINCIPAL CONTAINMENT DESIGN PARAMETERS  Characteristics Data  Containment design pressure (psig) 54 Containment design temperature (°F) 280 Internal dimensions Cylindrical wall diameter (ft) 130 Cylindrical wall height (ft) 139 Curved dome height (ft) 43.5 Volumes 
Gross internal volume (ft3) 2.35 x 106 Net free internal volume (ft3) 2.0 x 106 Containment design leak rate First 24 h, percent of containment free volume per day 0.15 After first day, percent per day 0.075 
FNP-FSAR-6  REV 21  5/08 TABLE 6.2-2 (SHEET 1 OF 5)  HEAT SINK GEOMETRIC DATA(a)  Heat Sink 1 - Containment Cylinder and Dome 74,908 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Outside Atmosphere
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 0.25 Air Gap 0.00204 Concrete 45.0
Heat Sink 2 - Penetration Plates & Liner Stiffners 3,802 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Outside Atmosphere
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 0.51 Air Gap 0.00204 Concrete 45.0
Heat Sink 3 - Unlined Concrete (excluding reactor support) 60,375 ft2  Exposure  1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Paint 0.019 Surfacer 0.125 Concrete 18.0
FNP-FSAR-6  REV 21  5/08 TABLE 6.2-2 (SHEET 2 OF 5)
Heat Sink 4 - Galvanized Steel (excluding cable trays) 43,320 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Zinc 0.0034 Carbon Steel 0.07 Heat Sink 5 - Painted Carbon Steel  0.5-in. Thickness 95,210 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 0.18 Heat Sink 6 - Painted Carbon Steel  1.0-in. Thickness 25,681 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 0.59 FNP-FSAR-6  REV 21  5/08 TABLE 6.2-2 (SHEET 3 OF 5)
Heat Sink 7 - Painted Carbon Steel  2.0-in. Thickness 8,802 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 1.35 Heat Sink 8 - Painted Carbon Steel  2.0-in. Thickness 3,353 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Paint/Primer 0.0084 Carbon Steel 3.59 Heat Sink 9 - Containment Floor 5,402 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Concrete 108.0
FNP-FSAR-6  REV 21  5/08 TABLE 6.2-2 (SHEET 4 OF 5)
Heat Sink 10 - Refuel Canal Liner 7,894 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Stainless Steel 0.25 Air Gap 0.00204 Concrete 18.0
Heat Sink 11 - Unpainted Stainless Steel 10,116 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Stainless Steel 0.12 
Heat Sink 12 - Galvanized Steel Cable Trays 22,164 ft2  Exposure 
: 1. Containment Atmosphere
: 2. Insulated
Material Thickness (in.)
Zinc 0.0034 Carbon Steel 0.05 
FNP-FSAR-6  REV 21  5/08 TABLE 6.2-2 (SHEET 5 OF 5)
Heat Sink 13 - Reactor Support 2,182 ft2  Exposure  1. Containment Atmosphere - A 150°F source to account for the higher reactor cavity operating temperature 2. Insulated Material Thickness (in.)
Paint 0.019 Surfacer 0.125 Concrete 86.0
_________________ 
(a) An evaluation for these parameters was performed as described in Section 6.2.1.3.13.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-3  INITIAL CONDITIONS FOR PRESSURE ANALYSIS  Characteristics Data  Containment System  Pressure (psia) 13.2 - 17.7 Relative humidity (percent) 50 Inside temperature (°F) 120(a) Outside temperature (°F) 95 Refueling water storage tank water temperature (°F) 110  Accumulator tank water temperature (°F) 120 Service water temperature (°F) 95(b)  Stored Water  Refueling water storage tank (gal) 471,000(c) Three accumulators (ft3) 240       
a. 120°F is the Technical Specifications limit, 127°F was used in the analysis. b. Service water temperature of 97.3°F was used in the analysis. c. A refueling water storage tank delivery capacity of 390,000 gallons was used in the analysis.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-4  HEAT SINK THERMODYNAMIC DATA  MATERIAL PROPERTIES(a)    Thermal Heat  Density Conductivity Capacity Material (lbm/ft3) Btu/h-ft-°F) (Btu/lbm-°F)    Paint (Ameron 66) 162.3 0.50/0.25(b) 0.29 Paint (Ameron 90, 90HS) 160.8 0.38/0.25(b) 0.31 Primer (Dimetcote 6) 196.8 0.63 0.11 Carbon steel 489.0 29.6 0.1096 Concrete 144.0 1.0 0.2292 Surfacer (Ameron 121.2 0.39 0.23 110 AA, 3366/3367)    Zinc 446.0 62.2 0.0942 Stainless steel 488.0 8.6 0.1232 Air 0.069 0.017 0.2095      HEAT TRANSFER COEFFICIENTS 
Surface Value Sink surfaces exposed to containment Modified Tagami atmosphere (LOCA Blowdown)  UCHIDA (LOCA  Reflood & MSLB)
Sump liquid to containment atmosphere Conduction Containment sump and floor to sump Conduction liquid Sink surfaces exposed to outside 2.0 Btu/h-ft2-°F atmosphere 
a. An evaluation was performed for these parameters as described in Section 6.2.1.3.13.
: b. Value for Paint/Primer in combination.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-5 (SHEET 1 OF 2)  ENGINEERED SAFETY FEATURES PERFORMANCE FOR CONTAINMENT PRESSURE TRANSIENT ANALYSIS    Values Used for  Containment Analysis  Maximum Minimum System Operation ESF ESF    Containment spray      Water sources Borated water from    RWST or sump    Initiation Initiated by    Containment Press. High-High-High    Number of lines and  2 1    headers      Number of pumps  2 1  Flowrate, gal/min  2175 2480 (Injection)    per pump  2290 (Recirculation)    Containment air coolers      Initiation Initiated by SIS    Number of units  4 1(b)  Flowrate (air side),  40000 40000    ft3/min per unit      Total design heat  80 x 106 80 x 106(a)    removal at con-        tainment design        temperature,        (Btu/h) per unit      Service water  97.3 97.3  temperature (°F)        RHR/Low pressure safety    injection heat exchangers      Type Horizontal shell    U-tube    Cooling water supply Component cooling    water    Number of units  2 1  Heat transfer area,  4070 3500    ft2 per unit      Overall heat transfer  383 383    coefficient,        Btu/h-ft2-°F FNP-FSAR-6 REV 21  5/08 TABLE 6.2-5 (SHEET 2 OF 2)    Values Used for  Containment Analysis  Maximum Minimum System Operation ESF ESF      Flowrate: Injection 3000 3000    Sump water side, Recirculation 3750 3750    gal/min per unit        Component cooling  4755 4755    water side, gal/min        per unit      Return water point  Primary Primary  loop loop Passive safety injection    system      Capacity, gal each 600      accumulator      Number of accumu-  3 3    lators      Pressure setpoint,  600 600    psig        Active safety injection    system      Initiation Initiated by SIS    High pressure        safety injection:        Number of lines  3 3    Number of pumps  2 1  Flowrate, gal/min  511 511    per pump      Low pressure safety        injection:        Number of lines  3 3    Number of pumps  2 1    Flowrate, gal/min Injection 3000 3000    per pump Recirculation 3750 3750 
a. Value for 600-gal/min service water flow for paragraph 6.2.1.3.12 analysis is 31.2 x 106 at 275 °F. 
: b. Having fewer than 12 coils per containment cooler is acceptable provided that each cooler can adequately remove the containment analysis heat load described in note "a".
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-6  CONTAINMENT PRESSURE ANALYSIS RESULTS FOR THE SPECTRUM OF RCS BREAK SIZES(a)      0.6    DEPSG(b) DEPSG DEPSG(b) PSS(b) DECLG(b) DEHLG  MIN ESF MIN ESF MAX ESF MAX ESF MAX ESF MIN ESF  P0 = 0 psig P0 = +3 psig 4.95 ft2 3 ft2 8.25 ft2 P0 = +3 psig        Peak pressure 38.0 43.8 40.1  40.9 37.6 43.6  (psig)              Time of peak 19.4 552 191.9 194.3 22.3 18.8  pressure (s)              Peak temperature  260 263 264  265 260 264  (°F)              Time of peak 19.4 552 191.9 194.3 22.3 18.7  temperature (s) a. See Table 6.2-41 for MSLB results.
: b. Non-limiting cases, not reanalyzed for power uprate/steam generator replacement, maintained for historical purposes FNP-FSAR-6 REV 21  5/08 TABLE 6.2-7  SYSTEM PARAMETERS  INITIAL CONDITIONS FOR THERMAL UPRATE  PARAMETERS VALUE  Core Thermal Power (MWt) 2830.5  Reactor Coolant System Total Flowrate (lbm/sec) 27250.0  Vessel Outlet Temperature (°F) 619.3  Core Inlet Temperature (°F) 547.1  Vessel Average Temperature (°F) 583.2  Initial Steam Generator Steam Pressure (psia) 817    Steam Generator Design    Model 54F  Steam Generator Tube Plugging (%) 0  Initial Steam Generator Secondary Side Mass (lbm) 121826.1  Assumed Maximum Containment Backpressure (psia) 68.7  Accumulator Water Volume (ft3) per accumulator 1040  N2 Cover Gas Pressure (psia) 600  Temperature (°F) 120  Safety Injection Delay, total (sec)  (from beginning of event) 30.9 
Note: Core Thermal Power, RCS Total Flowrate, RCS Coolant Temperatures, and Steam Generator Secondary Side Mass include appropriate uncertainty and/or allowance. 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-8  SAFETY INJECTION FLOW MINIMUM SAFEGUARDS    RCS PRESSURE TOTAL FLOW    (psig)      (gpm)    INJECTION MODE (REFLOOD PHASE)  0 4411.2 20 4163.4 40 3897.1 60 3603.8 80 3275.0 100 2900.8 120 2190.7 140 1619.5 160 482.7 180 480.0                COLD LEG RECIRCULATION MODE  0 3997.8 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-9  SAFETY INJECTION FLOW MAXIMUM SAFEGUARDS    RCS PRESSURE TOTAL FLOW    (psig)      (gpm)    INJECTION MODE (REFLOOD PHASE)  0 8575.0 20 8094.4 40 7581.5 60 7028.8 80 6425.3 100 5752.0 120 4976.6 140 4327.8 160 3530.3 180 2376.1 COLD LEG RECIRCULATION MODE  0 8575.0 
FNP-FSAR-6  REV 21  5/08 TABLE 6.2-10 (SHEET 1 OF 4)  DOUBLE-ENDED HOT LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec)  (Btu/sec) (lbm/sec) (Btu/sec) 
.00000  .0  .0  .0 .0  .00113 46198.2 29516.1 46195.4 29512.8  .101  40189.0 26064.4 26821.5 17100.0
.201  34372.1 22298.5 23762.2 15065.7
.301  33846.7 21888.6 21200.8 13289.5
.401  32820.3 21206.0 19842.3 12249.4
.501  32023.9 20690.1 18991.0 11540.4  .601  31901.2 20605.1 18375.5 11003.2  .702  31874.4 20599.8 17849.2 10548.9
.801  31502.1 20403.4 17484.0 10215.0
.901  30897.3 20080.6 17160.3 9923.1  1.00 30486.2 19905.6 16866.9 9666.7  1.10 30168.9 19810.0 16638.3 9459.5  1.20 29888.3 19739.6 16448.8 9285.5  1.30 29539.0 19615.6 16339.9 9165.2  1.40 29120.5 19445.2 16294.4 9086.2  1.50 28623.3 19222.2 16299.4 9039.9  1.60 28060.0 18952.1 16335.9 9015.3  1.70 27484.3 18671.5 16392.6 9004.8  1.80 26923.5 18398.8 16463.9 9004.9  1.90 26365.0 18122.1 16540.5 9011.4  2.00 25754.5 17800.6 16619.8 9022.6  2.10 25092.6 17432.7 16698.9 9036.8  2.20 24448.5 17068.0 16775.4 9053.0  2.30 23848.9 16728.9 16847.1 9069.6  2.40 23275.6 16398.4 16908.6 9083.8  2.50 22699.0 16052.7 16957.3 9093.9  2.60 22152.3 15716.1 16990.9 9098.6  2.70 21633.8 15387.8 17010.2 9098.1  2.80 21137.9 15064.3 17014.2 9091.2  2.90 20673.7 14755.2 17004.2 9078.8  3.00 20244.4 14460.2 16980.5 9060.7  3.10 19840.5 14170.6 16943.2 9036.8  3.20 19487.3 13909.9 16893.0 9007.4  3.30 19173.3 13668.3 16831.6 8973.1  3.40 18885.6 13436.2 16758.0 8933.5  3.50 18645.8 13233.3 16673.0 8888.7 FNP-FSAR-6  REV 21  5/08 TABLE 6.2-10 (SHEET 2 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec)  (Btu/sec) (lbm/sec) (Btu/sec)  3.60 18435.9 13045.7 16577.9 8839.6  3.70 18249.1 12868.5 16472.4 8785.7  3.80 18099.7 12715.2 16358.8 8728.4  3.90 17969.9 12572.2 16235.2 8666.6  4.00 17855.3 12436.7 16100.7 8599.8  4.20 17697.1 12216.4 15802.5 8452.7  4.40 17639.5 12063.2 15468.3 8288.4  4.60 17745.0 12020.9 15116.7 8115.9  4.80 17952.2 12017.3 14771.0 7947.5  5.00 18303.0 12077.3 14352.5 7739.9  5.20 18790.2 12199.5 13896.4 7511.5  5.40 13503.2 9607.3 13445.3 7286.1 5.60 14722.7 10243.4 13012.2 7070.8  5.80 14884.5 10185.1 12576.0 6853.4  6.00 14975.8 10228.1 12111.8 6619.8  6.20 15007.9 10189.6 11650.1 6386.0  6.40 15046.6 10177.1 11197.6 6155.7  6.60 15138.4 10141.8 10763.7 5933.6  6.80 15178.5 10074.0 10344.3 5717.9  7.00 15231.5 9998.6 9941.0 5509.8 7.20 14935.3 9813.0 9565.7 5315.9 7.40 15073.7 9819.4 9219.0 5136.9 7.60 15146.3 9794.5 8891.8 4968.1 7.80 15183.5 9756.1 8590.5 4813.0 8.00 15159.8 9685.1 8304.6 4665.8 8.20 15111.2 9601.5 8034.1 4526.8 8.40 15027.5 9502.2 7780.3 4396.7 8.60 14901.2 9383.0 7532.9 4270.1 8.80 14729.3 9243.8 7296.7 4149.5 9.00 14509.7 9083.3 7068.2 4033.5 9.20 14245.5 8903.4 6846.7 3921.6 9.40 13945.5 8708.8 6632.5 3814.1 9.60 13620.4 8505.1 6424.8 3710.5 9.80 13277.4 8295.5 6222.2 3610.3 10.0 12927.4 8085.7 6026.5 3514.2  10.2 12570.6 7875.5 5834.9 3420.9  10.2 12567.5 7873.8 5833.4 3420.2 FNP-FSAR-6  REV 21  5/08 TABLE 6.2-10 (SHEET 3 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec)  (Btu/sec) (lbm/sec) (Btu/sec) 10.4 12210.5 7666.6 5649.8 3331.5  10.6 11856.3 7463.8 5471.9 3246.4  10.8 11500.2 7262.7 5298.1 3164.0  11.0 11151.5 7068.5 5130.8 3085.2  11.2 10805.2 6878.1 4967.8 3009.2 11.4 10465.5 6694.1 4811.0 2936.6 11.6 10130.0 6514.7 4658.5 2866.6 11.8  9791.2 6336.2 4509.6 2798.5  12.0  9437.7 6153.3 4362.5 2731.4  12.2  9059.5 5961.8 4210.8 2661.9  12.4  8654.2 5762.0 4048.8 2588.1  12.6  8228.7 5559.3 3873.8 2509.9  12.8  7800.9 5364.4 3685.7 2428.4  13.0  7369.9 5178.2 3479.8 2341.9  13.2  6952.2 5007.3 3269.6 2255.5  13.4  6533.3 4844.8 3054.5 2167.4  13.6  6122.1 4691.4 2851.2 2082.8  13.8  5711.4 4540.4 2663.8 2000.8  14.0  5304.8 4389.3 2502.3 1925.0  14.2  4906.4 4239.9 2368.1 1856.7  14.4  4513.5 4085.7 2259.7 1796.5  14.6  4070.0 3880.6 2173.8 1744.8  14.8  3665.2 3570.9 2101.9 1697.9  15.0  3386.8 3341.6 2039.7 1656.0  15.2  3100.1 3136.6 1982.4 1618.9  15.4  2792.5 2926.1 1923.6 1584.8  15.6  2470.0 2714.0 1860.2 1553.4  15.8  2150.8 2494.5 1787.6 1521.7  16.0  1957.2 2348.3 1703.4 1489.5  16.2  1810.1 2197.2 1605.9 1458.1  16.4  1699.4 2072.4 1498.2 1431.7  16.6  1574.1 1930.3 1380.6 1403.9  16.8  1458.7 1798.1 1265.4 1374.7  17.0  1362.8 1681.6 1169.2 1338.6  17.2  1258.7 1560.0 1102.7 1295.5  17.4  1160.9 1446.1 1041.3 1242.4  17.6  1081.9 1347.9  984.2 1183.1  17.8  1007.4 1260.6  929.9 1126.9 FNP-FSAR-6  REV 21  5/08 TABLE 6.2-10 (SHEET 4 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec)  (Btu/sec) (lbm/sec) (Btu/sec) 18.0 910.8 1141.2  836.0 1023.3  18.2 814.1 1024.0  707.5  870.1  18.4 715.7  900.0  641.0  791.0  18.6 629.5  793.5  529.7  653.6  18.8 547.0  690.0  403.6  499.9  19.0 458.5  579.1  267.2  332.2  19.2 372.8  471.7  159.1  198.9  19.4 282.1  357.9 97.1  122.4  19.6 207.7  264.6 84.2  107.3  19.8 89.1  114.1  .0  .0  20.0 .0  .0  .0  .0   
*mass and energy exiting from the reactor vessel side of the break.
**mass and energy exiting from the SG side of the break.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-11  PLANT DATA FOR BLOWDOWN    Reactor coolant loops 3 Minimum steam line internal diameter 14 inches Main feedwater isolation valve closing time 30 s Main feedwater control valve closing time 5 s Main steam line isolation valve closing time 10 s Maximum steam line volume between the steam 1180 ft3  generator and the nearest steam line stop  valve  Maximum steam line volume between the faulted 3475 ft3  steam generator stop valves and the steam  line stop valves in the other steam generator  loops  Maximum unisolated feed line volume 202 ft3 Maximum auxiliary feedwater flow to a Varies with  depressurized steam generator steam generator  pressure Time to auxiliary feedwater isolation 1800 s Main feedwater flow Varies Containment pressure setpoint for main steam 19.2 psig  line isolation signal          Air cooler initiation pressure 7.0 psig Air cooler delay from start of accident 92 s 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-12  DOUBLE-ENDED HOT LEG BREAK MASS BALANCE Time (Seconds)  .00  20.00  20.0  Mass (Thousand lbm)      Initial In RCS & ACC  620.08  620.08  620.08 Added Mass Pumped Injection  .00  .00  .00 Total Added  .00  .00  .00      ***TOTAL AVAILABLE***  620.08  620.08  620.08      Distribution Reactor Coolant  416.79  65.09  84.67 Accumulator  203.30  152.90  133.32  Total Contents  620.08  217.99  217.99      Effluent Break Flow  .00  402.08  402.08  ECCS Spill  .00  .00  .00  Total Effluent  .00  402.08  402.08      ***TOTAL ACCOUNTABLE***  620.08  620.07  620.07 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-13  DOUBLE-ENDED HOT LEG BREAK ENERGY BALANCE    Time (Seconds)  .00  20.00  20.0      Energy (Million BTU) Initial Energy In RCS, ACC, S. Gen  673.30  673.30  673.30      Added Energy Pumped Injection  .00  .00  .00  Decay Heat  .00  5.79  5.79  Heat From Secondary  .00  -6.91  -6.91        Total Added  .00  -1.12  -1.12      ***TOTAL AVAILABLE***  673.30  672.19  672.19        Distribution Reactor Coolant  244.82  14.44  16.19  Accumulator  18.20  13.68  11.93  Core Stored  18.93  7.36  7.36  Primary Metal  118. 16  110.31  110.31  Secondary Metal  76.01  74.48  74.48  Steam Generator  197.20  192.31  192.31        Total Contents  673.30  412.59  412.59      Effluent Break Flow  .00  259.11  259.11  ECCS Spill  .00  .00  .00  Total Effluent  .00  259. 11  259.11      ***TOTAL ACCOUNTABLE***  673.30  671.70  671.70 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-14 (SHEET 1 OF 4)  DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)    .00000  .0  .0  .0  .0  .00108 90598.7 48888.7 40349.1 21718.9  .101 40353.3 21793.7 20648.3 11108.4
.202 46482.0 25280.7 22386.6 12050.6
.302 46307.3 25410.3 22661.8 12210.0
.402 46761.9 25935.4 22249.9 12001.2
.502 46296.0 25990.1 21549.4 11632.8
.602 44228.9 25131.5 20917.5 11297.9
.702 44745.5 25709.3 20392.8 11019.3
.801 44635.7 25899.4 19916.9 10765.7
.901 43950.0 25731.1 19498.4 10541.9 1.00 42962.6 25369.1 19151.6 10356.4 1.10 41980.3 24996.5 18887.2 10214.8 1.20 41034.1 24632.9 18699.2 10114.0 1.30 40163.4 24298.2 18560.7 10039.4 1.40 39364.0 23993.8 18435.7  9971.6  1.50 38594.0 23697.1 18316.5  9906.5  1.60 37802.5 23382.0 18214.6  9850.7  1.70 36954.7 23040.0 18126.9  9802.8  1.80 36079.1 22697.6 18027.0  9748.3  1.90 35106.4 22318.3 17885.7  9671.2  2.00 33872.9 21794.1 17715.2  9578.1  2.10 32344.2 21088.9 17549.4  9488.1  2.20 30820.7 20391.9 17349.3  9379.4  2.30 29075.9 19528.2 17091.6  9239.6  2.40 25410.4 17280.8 16797.9  9080.3  2.50 22059.0 15206.0 16471.4  8903.4  2.60 19923.2 13913.5 16092.2  8698.8  2.70 18262.9 12868.1 15821.6  8553.9  2.80 16954.1 12014.8 15567.8  8418.1  2.90 15973.2 11368.2 15302.8  8276.2  3.00 15192.2 10851.6 15017.0  8123.1  3.10 14591.9 10460.8 14790.8  8002.9  3.20 14100.8 10144.9 14596.6  7899.9  3.30 13665.0  9865.9 14406.5  7798.9  3.40 13257.3  9607.6 14230.0  7705.3  3.50 12888.2  9378.1 14119.3  7647.8  3.60 12567.0  9181.8 14054.6  7614.6  3.70 12247.2  8981.6 13891.7  7527.7 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-14 (SHEET 2 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)    3.80 11934.8  8783.5 13724.4  7438.6  3.90 11648.7  8602.6 13561.2  7351.8  4.00 11396.4  8441.0 13396.3  7263.5  4.20 10974.9  8155.0 13058.2  7082.8  4.40 10630.1  7903.0 12794.2  6942.2  4.60 10367.3  7695.6 12572.7  6823.8  4.80 10144.5  7507.1 12312.0  6684.3  5.00 9969.6  7348.6 12100.0  6571.6  5.20 9816.7  7199.9 13057.1  7093.8  5.40 9710.3  7083.2 12743.2  6924.8  5.60 9660.3  7002.2 12615.2  6858.1  5.80 9648.9  6946.2 12420.1  6754.2  6.00 9649.2  6898.6 12299.5  6691.7  6.20 9661.8  6859.0 12185.5  6632.5  6.40 9882.2  6962.3 12043.3  6557.5  6.60 10191.9  7145.0 11990.7  6530.9  6.80 9940.3  7237.9 11911.0  6487.0  7.00 8907.6  6929.3 11760.1  6402.5  7.20 8293.4  6649.5 11602.2  6314.2  7.40 8116.3  6514.4 11455.6  6233.1  7.60 8051.8  6425.4 11314.4  6155.5  7.80 7991.2  6330.6 11160.2  6070.3  8.00 7966.4  6232.7 10991.6  5976.9  8.20 8002.8  6153.3 10830.6  5887.7  8.40 8065.0  6087.9 10677.9  5803.2  8.60 8120.8  6030.9 10524.0  5718.0  8.80 8149.6  5972.1 10369.5  5632.5  9.00 8135.7  5901.3 10219.6  5549.6  9.20 8089.0  5826.5 10074.9  5469.7  9.40 8007.8  5745.1  9929.7  5389.5  9.60 7894.2  5656.3  9786.3  5310.5  9.80 7753.1  5562.2  9647.8  5234.3  10.0  7605.4  5477.1  9509.1  5158.1  10.2  7437.8  5385.7  9364.9  5078.9  10.4  7264.6  5293.5  9226.4  5003.1  10.6  7093.4  5203.6  9086.0  4926.4  10.8  6923.6  5115.3  8946.7  4850.4  11.0  6758.0  5029.1  8809.1  4775.3  11.2  6595.2  4943.5  8671.7  4700.4  11.4  6439.1  4859.7  8536.6  4626.8  11.6  6289.0  4777.2  8402.2  4553.7 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-14 (SHEET 3 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)    11.8  6145.2  4696.4  8269.5  4481.8  12.0  6008.1  4617.9  8138.9  4411.0  12.2  5876.2  4541.4  8010.2  4341.3  12.4  5748.2  4467.9  7882.8  4272.2  12.6  5624.9  4396.3  7759.7  4205.4  12.8  5504.5  4327.1  7637.5  4139.2  13.0  5386.5  4260.6  7517.0  4073.9  13.2  5271.4  4197.3  7398.9  4009.9  13.4  5154.4  4134.4  7213.8  3908.8  13.6  5029.6  4068.0  7067.1  3827.5  13.8  4888.2  3990.5  6878.9  3700.7  14.0  4735.1  3903.5  6840.4  3626.7  14.2  4567.2  3794.9  6612.0  3434.9  14.4  4414.2  3692.8  6562.6  3323.2  14.6  4279.4  3593.3  6813.5  3376.9  14.8  4176.4  3518.4  5778.5  2808.3  15.0  4087.4  3461.3  7270.0  3434.7  15.2  3953.2  3382.4 11154.0  5315.1  15.4  3773.3  3296.0  7340.3  3538.0  15.6  3761.2  3363.0  4436.2  2138.9  15.8  3693.6  3364.8  6754.0  3020.2  16.0  3470.4  3296.9  9836.8  4401.6  16.2  3281.2  3275.4  5828.1  2647.8  16.4  3221.5  3322.6  4708.7  2159.7  16.6  3092.4  3317.4  4129.3  1838.1  16.8  2770.1  3155.6  4691.2  1973.3  17.0  2504.2  3008.1  4954.4  2033.4  17.2  2210.4  2709.2  4419.5  1791.5  17.4  2001.1  2467.9  4204.6  1677.7  17.6  1821.6  2254.3  4260.6  1652.9  17.8  1651.8  2049.3  4635.3  1729.8  18.0  1497.6  1862.4  4545.8  1641.5  18.2  1353.9  1687.2  4397.7  1546.9  18.4  1219.1  1521.5  4220.7  1449.9  18.6  1091.9  1365.7  3948.7  1325.8  18.8 961.6  1204.7  3568.6  1171.0  19.0 840.8  1054.7  3109.8 997.2  19.2 735.9 924.0  2739.8 858.6  19.4 638.6 802.6  2316.7 710.0  19.6 566.7 712.9  1886.5 566.3  19.8 495.0 623.3  1443.6 425.6 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-14 (SHEET 4 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)    20.0 449.1 565.7 984.8 286.2  20.2 405.0 510.5 529.0 152.3  20.4 351.5 443.3 105.7 30.4  20.6 293.9 370.8  .0  .0  20.8 225.7 285.0  .0  .0  21.0 147.8 186.9 126.9 36.8  21.2 88.6 112.3 87.3 25.3  21.4 29.2 37.2  .0  .0  21.6 .0  .0  .0  .0 
*mass and energy exiting the SG side of the break.
** mass and energy exiting the pump side of the break.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-15 (SHEET 1 OF 2)  REACTOR CAVITY RELEASE    Enthalpy Time (s) Flow (lb/s) (Btu/lb)    0.0 0.0 0.0 1.00 x 10-3 7.16 x 103 5.56 x 102 3.04 x 10-3 1.09 x 104 5.55 x 102 5.04 x 10-3 1.33 x 104 5.54 x 102 7.04 x 10-3 1.47 x 104 5.52 x 102 9.08 x 10-3 1.69 x 104 5.51 x 102 1.01 x 10-2 1.72 x 104 5.51 x 102 1.10 x 10-2 1.71 x 104 5.50 x 102 1.31 x 10-2 1.63 x 104 5.48 x 102 1.40 x 10-2 1.72 x 104 5.48 x 102 1.51 x 10-2 1.85 x 104 5.48 x 102 1.70 x 10-2 1.99 x 104 5.47 x 102 1.90 x 10-2 2.01 x 104 5.46 x 102 2.01 x 10-2 2.02 x 104 5.45 x 102 2.11 x 10-2 2.00 x 104 5.44 x 102 2.31 x 10-2 1.99 x 104 5.43 x 102 2.51 x 10-2 1.97 x 104 5.42 x 102 2.70 x 10-2 1.97 x 104 5.41 x 102 2.91 x 10-2 1.98 x 104 5.40 x 102 3.11 x 10-2 2.01 x 104 5.40 x 102 3.31 x 10-2 2.04 x 104 5.39 x 102 3.50 x 10-2 2.08 x 104 5.39 x 102 3.70 x 10-2 2.10 x 104 5.39 x 102 3.91 x 10-2 2.13 x 104 5.38 x 102 4.11 x 10-2 2.14 x 104 5.38 x 102 4.21 x 10-2 2.14 x 104 5.38 x 102 4.31 x 10-2 2.14 x 104 5.37 x 102 4.51 x 10-2 2.11 x 104 5.37 x 102 4.71 x 10-2 2.09 x 104 5.36 x 102 4.92 x 10-2 2.06 x 104 5.36 x 102 5.10 x 10-2 2.04 x 104 5.35 x 102 5.31 x 10-2 2.03 x 104 5.35 x 102 5.50 x 10-2 2.03 x 104 5.35 x 102 6.01 x 10-2 2.04 x 104 5.35 x 102 6.50 x 10-2 2.03 x 104 5.34 x 102 7.01 x 10-2 2.03 x 104 5.34 x 102 7.51 x 10-2 2.02 x 104 5.34 x 102 8.01 x 10-2 1.97 x 104 5.34 x 102 8.50 x 10-2 1.92 x 104 5.33 x 102 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-15 (SHEET 2 OF 2)      Enthalpy Time (s) Flow (lb/s) (Btu/lb)    9.00 x 10-2 1.89 x 104 5.33 x 102 9.50 x 10-2 1.87 x 104 5.33 x 102 1.00 x 10-1 1.85 x 104 5.33 x 102 1.20 x 10-1 1.91 x 104 5.33 x 102 1.25 x 10-1 1.90 x 104 5.33 x 102 1.50 x 10-1 1.71 x 104 5.32 x 102 1.75 x 10-1 1.74 x 104 5.32 x 102 1.81 x 10-1 1.75 x 104 5.32 x 102 2.00 x 10-1 1.72 x 104 5.32 x 102 2.50 x 10-1 1.78 x 104 5.32 x 102 3.00 x 10-1 1.74 x 104 5.32 x 102 3.50 x 10-1 1.78 x 104 5.32 x 102 4.00 x 10-1 1.78 x 104 5.32 x 102 4.50 x 10-1 1.80 x 104 5.32 x 102 4.60 x 10-1 1.80 x 104 5.32 x 102 4.70 x 10-1 1.80 x 104 5.32 x 102 5.00 x 10-1 1.78 x 104 5.32 x 102 5.50 x 10-1 1.78 x 104 5.32 x 102 6.00 x 10-1 1.78 x 104 5.32 x 102 6.50 x 10-1 1.76 x 104 5.32 x 102 7.00 x 10-1 1.74 x 104 5.32 x 102 7.50 x 10-1 1.73 x 104 5.32 x 102 8.00 x 10-1 1.75 x 104 5.32 x 102 8.50 x 10-1 1.76 x 104 5.32 x 102 9.00 x 10-1 1.76 x 104 5.32 x 102 9.50 x 10-1 1.77 x 104 5.32 x 102 1.00 1.78 x 104 5.32 x 102 1.50 1.81 x 104 5.32 x 102 1.90 1.82 x 104 5.33 x 102 2.40 1.82 x 104 5.32 x 102 2.80 1.80 x 104 5.32 x 102 3.00 1.79 x 104 5.32 x 102     
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-16  SPRAY LINE BREAK RELEASE    Enthalpy Time (s) Flow (lb/s) (Btu/lb)    0. 0. 6.42 x 102 0.025 3269 6.41 x 102 0.1 3245 6.39 x 102 0.15 3233 6.39 x 102 0.225 3210 6.39 x 102 0.3 3198 6.39 x 102 0.4 3186 6.39 x 102 0.75 3186 6.38 x 102 0.875 3174 6.38 x 102 1.0 3151 6.38 x 102 1.2 3127 6.38 x 102 1.4 3103 6.38 x 102 1.6 3080 6.38 x 102 1.8 3056 6.38 x 102 2.0 3033 6.38 x 101 2.2 3009 6.38 x 102 2.4 2985 6.38 x 102 2.6 2962 6.38 x 102 2.8 2938 6.38 x 102 3.0 2915 6.38 x 102     
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-17  SURGE LINE BREAK RELEASE    Enthalpy Time (s) Flow (lb/s) (Btu/lb)    0. 0. 692.8 0.025 6463 692.8 0.1 8585 700.9 0.15 8562 701.2 0.2 8569 700.2 0.3 8592 697.5 0.4 8600 695.0 0.5 8581 693.3 0.6 8533 692.7 0.7 8454 693.2 0.8 8352 694.4 0.9 8241 696.0 1.0 8133 697.3 1.2 7923 698.2 1.4 7841 696.3 1.6 7812 692.5 1.8  7789 688.8 2.0 7720 686.1 2.2 7619 684.4 2.4 7501 683.2 2.6 7381 681.1 2.8 7269 679.9 3.0  7167 677.5 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-18 (SHEET 1 OF 3)  REACTOR CAVITY SUBCOMPARTMENT PRESSURE ANALYSIS SUMMARY OF FLOWPATHS AND VENT LOSS COEFFICIENTS    Vent  k  Flowpath Area k k Bend +  (from to) (ft2) Contraction Expansion Friction k C        1 2 14.4 0.04 1.0 0.429 1.47 0.83 1 9 0.3 0.42 1.0  --- 1.42 0.84 1  32 0.3 0.42 1.0  --- 1.42 0.84 29 0.4 0.42 1.0  --- 1.42 0.84 232 0.4 0.42 1.0  --- 1.42 0.84 234 14.118 0.067 1.0  --- 1.07 0.97 34 2.03 0.34 1.0 0.301 1.64 0.78 36 2.03 0.34 1.0 0.301 1.64 0.78 312 1.52 0.32 1.0 0.363 1.68 0.77 319 1.52 0.32 1.0 0.264 1.58 0.79 334 1.14 0.42 1.0  --- 1.42 0.84 45 2.03 --- 1.0 0.214 1.21 0.91 413 0.98 --- 1.0 0.371 1.37 0.85 420 0.98 --- 1.0 0.523 1.52 0.81 57 1.03 0.27 1.0 0.250 1.52 0.81 513 0.80 --- 1.0 0.377 1.38 0.85 520 0.80 --- 1.0 0.510 1.51 0.81 534 0.55 0.42 1.0  --- 1.42 0.84 631 1.03 0.27 1.0 0.330 1.60 0.79 611 1.05 --- 1.0 0.371 1.37 0.85 618 1.05 --- 1.0 0.522 1.52 0.81 634 0.55 0.42 1.0  --- 1.42 0.84 78 0.88 0.37 1.0 0.471 1.84 0.74 714 1.81 --- 1.0 0.360 1.36 0.86 721 1.81 --- 1.0 0.508 1.51 0.81 734 1.12 0.42 1.0  --- 1.42 0.84 815 1.27 --- 1.0 0.366 1.37 0.856 822 1.27 --- 1.0 0.515 1.52 0.812 834 0.57 0.42 1.0  --- 1.42 0.84 833 2.03 --- 1.0 0.310 1.31 0.874 932 0.88 0.37 1.0 0.460 1.83 0.74 916 1.27 --- 1.0 0.366 1.36 0.856 923 1.27 --- 1.0 0.515 1.52 0.812 1017 0.905 --- 1.0 0.374 1.37 0.853 1024 0.905 --- 1.0 0.527 1.53 0.809 1034 0.55 0.42 1.0  --- 1.42 0.84
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-18 (SHEET 2 OF 3)          Vent  k  Flowpath Area k k Bend +  (from to) (ft2) Contraction Expansion Friction k C        1112 0.95 --- 1.0 0.321 1.32 0.87 1116 0.95 --- 1.0 0.447 1.45 0.83 1134 0.65 --- 1.0 0.187 1.19 0.92 1213 0.95 --- 1.0 0.412 1.41 0.84 1234 0.95 --- 1.0 0.183 1.18 0.92 1314 0.95 --- 1.0 0.447 1.45 0.83 1334 1.10 --- 1.0 0.184 1.18 0.92 1415 0.95  --- 1.0 0.542 1.54 0.81 1434 1.13  --- 1.0 0.183 1.18 0.92 1517 0.95  --- 1.0 0.542 1.54 0.81 1534 1.58  --- 1.0 0.181 1.18 0.92 1617 0.95  --- 1.0 0.542 1.54 0.81 1634 1.58  --- 1.0 0.181 1.18 0.92 1734 1.13  --- 1.0 0.183 1.18 0.92 1819 2.17  --- 1.0 0.308 1.31 0.87 1823 2.17  --- 1.0 0.430 1.43 0.84 1825 1.05  --- 1.0 0.534 1.53 0.81 1920 2.17  --- 1.0 0.395 1.39 0.85 1925 1.52  --- 1.0 0.524 1.52 0.81 2021 2.17  --- 1.0 0.430 1.43 0.84 2025 1.78  --- 1.0 0.519 1.52 0.81 2122 2.17  --- 1.0 0.521 1.52 0.81 2126 1.81  --- 1.0 0.520 1.52 0.81 2224 2.17  --- 1.0 0.521 1.52 0.81 2226 2.54  --- 1.0 0.513 1.51 0.81 2324 2.17  --- 1.0 0.521 1.52 0.81 2327 2.54  --- 1.0 0.513 1.51 0.81 2427 1.81  --- 1.0 0.520 1.52 0.81 2526 2.13  --- 1.0 1.010 2.01 0.71 2527 2.13  --- 1.0 1.010 2.01 0.71 2528 4.35  --- 1.0 0.759 1.76 0.75 2627 2.13  --- 1.0 1.010 2.01 0.71 2628 4.35  --- 1.0 0.759 1.76 0.75 2728 4.35  --- 1.0 0.759 1.76 0.75 2829 106.70  --- 1.0 0.350 1.35 0.86 2930 93.31 0.05 1.0 0.070 1.12 0.94 3034 56.66 0.08 1.0 1.198 2.28 0.66 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-18 (SHEET 3 OF 3)          Vent  k  Flowpath Area k k Bend +  (from to) (ft2) Contraction Expansion Friction k C        3134 0.55 0.42 1.0  --- 1.42 0.84 319 2.03  --- 1.0 0.310 1.31 0.874 3116 1.27  --- 1.0 0.366 1.37 0.856 3123 1.27  --- 1.0 0.515 1.52 0.812 3210 2.03  --- 1.0 0.218 1.22 0.906 3217 0.905 --- 1.0 0.374 1.37 0.853 3224 0.905 --- 1.0 0.527 1.53 0.809 3310 1.03 0.34 1.0 0.440 1.78 0.75 3334 0.55 0.42 1.0  --- 1.42 0.84 3315 1.27  --- 1.0 0.366 1.37 0.856 3322 1.27  --- 1.0 0.515 1.5  0.812 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-19  CONTAINMENT RESULTS FOR THE DESIGN BASIS LOCA    Prior to DEPSG DEHL Parameter LOCA At Peak At Peak    Pressures        Time (s)  552 18.8      Steam (psia) 1.03 37.1 37.7  Air (psia) 16.67 21.4 20.6  Total psia 17.70 58.5 58.3  Total gauge (psig) 3.0 43.8 43.6    Temperatures        Time (s)  1252 20.0      Steam and air (°F)      127    263    264  Water in sump (°F)      -    260    256                                        Heat transfer coefficient    (Btu/h-ft2-°F)(a) 0 218 231 
a. Between containment atmosphere and structure.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-20 (SHEET 1 OF 4)  DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)  21.6 .0 .0 .0 .0  22.1 .0 .0 .0 .0  22.2 .0 .0 .0 .0  22.3 .0 .0 .0 .0  22.4 .0 .0 .0 .0  22.5 .0 .0 .0 .0  22.6 55.3  65.3 .0 .0  22.7 31.1  36.7 .0 .0  22.8 33.2  39.2 .0 .0  22.9 39.1  46.1 .0 .0  23.0 45.9  54.2 .0 .0  23.1 49.4  58.3 .0 .0  23.2 55.4  65.3 .0 .0  23.3 59.8  70.6 .0 .0  23.4 64.0  75.5 .0 .0  23.5 68.0  80.2 .0 .0  23.6 71.8  84.7 .0 .0  23.7 75.4  89.0 .0 .0  23.8 79.0  93.2 .0 .0  23.9 82.4  97.2 .0 .0  24.0 85.7 101.1 .0 .0  24.1 88.8 104.9 .0 .0  24.2 91.9 108.5 .0 .0  24.3 94.9 112.0 .0 .0  24.4 97.8 115.5 .0 .0  24.5 100.7 118.8 .0 .0  24.6 103.4 122.1 .0 .0  25.6 127.9 151.0 .0 .0  26.6 148.3 175.1 .0 .0  27.6 165.9 195.9 .0 .0  28.2 321.3 380.3 2733.5 349.6 28.6 421.0 499.0 3803.6 500.4 29.7 453.5 537.9 4080.9 557.9 30.7 443.1 525.5 3984.9 548.9 31.7 460.0 545.7 4164.0 565.5 32.5 451.4 535.3 4085.8 556.9 32.7 449.2 532.8 4066.2 554.7 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-20 (SHEET 2 OF 4)  BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 33.7 438.8 520.4 3970.3 544.3 34.7 428.9 508.5 3877.7 534.2 35.7 419.4 497.2 3788.3 524.4 36.7 410.3 486.4 3702.2 514.9 37.7 401.7 476.0 3619.1 505.8 37.9 400.0 474.0 3602.8 504.0 38.7 393.4 466.1 3538.9 497.0 39.7 385.4 456.7 3461.4 488.5 40.7 377.8 447.6 3386.5 480.3 41.7 370.5 438.9 3314.1 472.3 42.7 363.5 430.5 3243.9 464.5 43.7 356.7 422.4 3175.9 457.0 44.2 353.4 418.5 3142.6 453.4 44.7 350.2 414.7 3109.8 449.8 45.7 343.9 407.2 3045.7 442.7 46.7 337.8 400.0 2983.3 435.8 47.7 332.0 393.0 2922.7 429.1 48.7 326.3 386.3 2863.6 422.5 49.7 320.8 379.7 2806.0 416.1 50.7 315.5 373.4 2749.8 409.9 51.3 312.4 369.7 2716.8 406.2 51.7 310.3 367.3 2695.0 403.8 52.7 253.5 299.8 2021.5 332.6 53.7 324.4 383.8 280.0 153.3 54.7 341.3 404.0 285.3 162.0 55.7 336.8 398.8 283.7 159.7 56.7 332.4 393.5 282.0 157.5 57.7 328.1 388.3 280.4 155.3 58.7 323.7 383.2 278.8 153.2 59.7 319.5 378.1 277.2 151.1 60.7 315.2 373.1 275.6 149.0 61.7 311.0 368.0 274.1 146.9 62.7 306.7 362.9 272.5 144.8 63.7 302.7 358.2 271.0 142.8 64.7 298.7 353.5 269.6 140.9 65.7 294.9 348.9 268.2 139.0 66.5 291.8 345.2 267.1 137.5 66.7 291.0 344.3 266.8 137.1 67.7 287.2 339.8 265.4 135.3 68.7 283.5 335.4 264.1 133.5 69.7 279.8 331.0 262.8 131.8 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-20 (SHEET 3 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 70.7 276.2 326.6 261.5 130.0 71.7 272.6 322.4 260.2 128.3 72.7 269.0 318.2 258.9 126.7 73.7 265.6 314.1 257.7 125.0 74.7 262.1 310.0 256.5 123.4 75.7 258.7 306.0 255.3 121.9 76.7 255.4 302.0 254.1 120.3 77.7 252.1 298.1 253.0 118.8  78.7 248.9 294.3 251.9 117.3  79.7 245.8 290.6 250.8 115.9  80.7 242.6 286.9 249.7 114.5 81.7 239.6 283.2 248.6 113.1 82.7 236.6 279.6 247.6 111.7 84.2 232.2 274.4 246.1 109.8 84.7 230.7 272.7 245.6 109.1 86.7 225.1 266.0 243.7 106.6 88.7 219.7 259.6 241.8 104.2 90.7 214.5 253.5 240.1 102.0 92.7 209.5 247.5 238.4  99.8 94.7 204.7 241.9 236.8  97.8 96.7 200.1 236.5 235.3  95.8 98.7 195.7 231.3 233.9  94.0 100.7 191.6 226.3 232.5  92.2 102.7 187.6 221.6 231.2  90.5 104.7 183.8 217.1 230.0  89.0 105.3 182.7 215.8 229.7  88.5 106.7 180.2 212.9 228.9  87.5 108.7 176.8 208.8 227.8  86.1 110.7 173.5 205.0 226.8  84.8 112.7 170.5 201.4 225.8  83.6 114.7 167.6 198.0 224.9  82.4 116.7 164.9 194.8 224.1  81.4 118.7 162.4 191.8 223.3  80.3 120.7 160.0 189.0 222.5  79.4 122.7 157.8 186.3 221.9  78.5 124.7 155.7 183.9 221.2  77.7 126.7 153.7 181.6 220.6  77.0 128.7 151.9 179.4 220.1  76.3 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-20 (SHEET 4 OF 4)    BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW**  TIME  THOUSAND  THOUSAND  (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 130.1150.8 178.0 219.7  75.8  130.7 150.3 177.5 219.6  75.6 132.7 148.7 175.6 219.1  75.0 134.7 147.3 173.9 218.7  74.5 136.7 146.0 172.4 218.3  73.9 138.7 144.8 171.0 217.9  73.5 140.7 143.7 169.7 217.6  73.0 142.7 142.6 168.5 217.2  72.6 144.7 141.7 167.4 217.0  72.3 146.7 140.9 166.4 216.7  71.9 148.7 140.1 165.5 216.5  71.6 150.7 139.4 164.7 216.2  71.4 152.7 138.8 163.9 216.0  71.1 154.7 138.2 163.2 215.8  70.9 156.7 137.7 162.6 215.7  70.6 158.0 137.4 162.2 215.6  70.5 158.7 137.2 162.0 215.5  70.4  160.7 136.8 161.5 215.4  70.3  162.7 136.4 161.1 215.2  70.1  164.7 136.1 160.7 215.1  70.0  166.7 135.8 160.4 215.0  69.8  168.7 135.6 160.1 214.9  69.7  170.7 135.4 159.9 214.9  69.6  172.7 135.3 159.8 214.8  69.5  174.7 135.2 159.6 214.7  69.5  176.7 135.1 159.5 214.7  69.4  178.7 135.0 159.5 214.6  69.3  180.7 135.0 159.4 214.6  69.3  182.7 135.2 159.6 214.6  69.3  184.7 135.6 160.2 215.3  69.6  186.7 136.1 160.7 216.8  70.0  187.7 136.3 161.0 217.8  70.3
* mass and energy exiting the SG side of the break.
** mass and energy exiting the pump side of the break.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-21  LOCA CHRONOLOGY OF EVENTS  Time  (s) Event  0.0 Pipe ruptures (DEPSG), reactor  depressurization begins.  (a) Mass and energy release modeling. 
62.0 Containment sprays begin operation. 92.0 Air coolers begin operation. 552.0 Containment reaches maximum peak pressure    1252 Sump reaches maximum temperature. 2139 Safety injection water recirculation from the      sump begins as RWST reaches low level.
4256 Containment spray water recirculation from the  sump begins as RWST reaches low-low level.
107 Containment reaches atmospheric pressure (estimate). 
a. See table 6.2-52 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-22 (SHEET 1 OF 2)  SUBCOMPARTMENT DIFFERENTIAL PRESSURE RESULTS  Steam Generator Compartment Compartment Cold Leg Break (psid) Time (s) Design (psid) Pressurizer Surge Line Break (psid)        1, SG-C 33.9 0.42 35 7.3 2, SG-A 22.6 0.42 35 (b) 3, SG-B 19.3 0.60 35 (b)  Pressurizer Compartment  Compartment Spray Line Break (psid) Time (s) Design (psid)      1  9.4 0.14 35 2  3.1 0.10 35 Reactor Cavity  Cold Leg Break  Node No. Volume (ft3) Pressure (psia) Time (s) Design Pressure (psid)      1(a) 67.70 305.48 0.129 667 2(a) 104.12 288.02 0.129 667 3 17.17 19.15 0.600 150 4 7.36 18.83 0.600 150 5 9.46 18.26 0.600 150 6 11.37 21.30 0.600 150 7 14.47 18.40 0.600 150 8 12.41 19.80 0.600 150 9 12.41 57.26 0.135 150 10 10.27 41.65 0.140 150 11 3.68 21.64 0.600 150 12 5.33 18.78 0.600 150 13 6.22 18.21 0.600 150 14 6.35 18.02 0.600 150 15 8.89 19.79 0.600 150 16 8.89 36.50 0.141 150 17 6.35 36.59 0.141 150                      a. Inside penetration at inspection opening.
: b. Only the most limiting subcompartment pressure was re-analyzed.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-22 (SHEET 2 OF 2)  Node No. Volume (ft3) Pressure (psia) Time (s) Design Pressure (psid)      18 8.42 21.36 0.600 150 19 12.19 19.32 0.600 150 20 14.23 18.83 0.600 150 21 14.52 18.86 0.600 150 22 20.32 19.98 0.600 150 23 20.32 26.36 0.148 150 24 14.52 26.29 0.148 150 25 34.25 18.58 0.600 150 26 34.25 18.59 0.600 150 27 34.25 19.69 0.600 150 28 1055.70 16.07 0.598 150 29 2190.60 16.02 0.600 150 30 3603.10 15.98 0.600 150 31 13.00 37.62 0.141 150 32 9.68 58.82 0.135 150 33 13.00 20.20 0.600 150 34 2.0 x 106 15.88 0.600 54 Net Vessel Side Load(b)
Lbf    Time (s) 1.184 x 106 0.12 
b. Reactor vessel support stresses not to exceed design criteria presented in tables 5.2-6 and 5.2-7.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-23 
(This page has been intentionally deleted.) 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-24 (SHEET 1 OF 2)  COMPONENT DESIGN PARAMETERS FOR CONTAINMENT SPRAY SYSTEM AND CONTAINMENT COOLING SYSTEM  Containment Spray Pumps    Type Horizontal Centrifugal Number 2 Pressure (psig) 300 Temperature (°F) 250 Flowrate (each) (gal/min) 2600 Head (ft) 450 Containment Coolers Number 4(a) Pressure (psig) 200 Temperature (°F) 300 Water inlet temperature (°F) 95 Flowrate (normal - high reactor coolant leakage)
(gal/min) 800 
Heat removal rate (normal) (Btu/h) 2.36 x 106 Flowrate (post-LOCA) (gal/min) 2000 (600 for containment analysis)    Heat removal rate (post-LOCA) (Btu/h) 80.0 x 106 (31.2 x 106    for containment    analysis)
Containment Cooler Fans Type Vaneaxial Number 4 Flowrate (high speed) (sf3/min) 80,000 Static head (high speed) (in. wg) 4.75 Horsepower (high speed) (hp) 80 Flowrate (low speed) (sf3/min) 40,000 Static head (low speed) (in. wg) 7.90 Horsepower (low speed) (hp) 105 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-24 (SHEET 2 OF 2)  COMPONENT DESIGN PARAMETERS FOR CONTAINMENT SPRAY SYSTEM AND CONTAINMENT COOLING SYSTEM  Refueling Water Storage Tank    Quantity 1 Volume (gal) 500,000 Design pressure (psig) atmosphere Design temperature (°F) ambient Material stainless steel Piping Pressure (psig) 210 Temperature (°F) 300 Valves Pressure (psig) 210 Temperature (°F) 300 
a. Having fewer than 12 coils per containment cooler is acceptable, provided that each cooler can adequately remove the containment analysis heat load.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-25 (SHEET 1 OF 3)  REGULATORY GUIDE 1.52, REV. 0 SECTION APPLICABILITY FOR THE PENETRATION ROOM FILTRATION SYSTEM  Regulatory Guide Applicability to This Note Section System Index    C.1.a  Yes  1  C.1.b  Yes  -
C.1.c  Yes  -
C.1.d  Yes  -
C.1.e  Yes  -
C.2.a  No  2 C.2.b  No  3 C.2.c  Yes  -
C.2.d  Yes  -
C.2.e  Yes  16 C.2.f  Yes  -
C.2.g  Yes  4 C.2.h  Yes  -
C.2.i  Yes  -
C.2.j  No  6 C.2.k  Yes  -
C.2.l  Yes  -
C.2.m  Yes  -
C.3.a  No  7 C.3.b  Yes  8 C.3.c  Yes  -
C.3.d  Yes  -
C.3.e  Yes  9 C.3.f  Yes  -
C.3.g  Yes  -
C.3.h  Yes  10 C.3.i  Yes  -
C.3.j  No  11 C.3.k  Yes  -
C.3.l  Yes  12 C.3.m  Yes  13 C.3.n  Yes  -
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-25 (SHEET 2 OF 3)  Regulatory Guide Applicability to This Note Section System Index    C.4.a  Yes  -  C.4.b  Yes  -
C.4.c  Yes  14 C.4.d  Yes  -
C.4.e  Yes  -
C.4.f  Yes  -
C.4.g  Yes  -
C.4.h  Yes  15 C.4.i  Yes  -
C.4.j  Yes  -
C.4.k  Yes  -
C.4.l  Yes  -
C.4.m  Yes  -
C.5.a  Yes  -
C.5.b  Yes  17 C.5.c  Yes  17 C.6.a  Yes  -
C.6.b  Yes  - 
NOTES 1. The design basis accident is the postulated 30-day LOCA. 
2. No demister is provided because the unit is located outside the containment and no entrained water droplets are anticipated. No HEPA filters are provided downstream of the charcoals, since radioactive fines carryover is very unlikely. This is true because the charcoal trays are pressure tested at high velocity in the manufacturer's shop prior to delivery, thereby removing fines. Also, during system operation, air is passing through the charcoal at a very low velocity.
3. No physical separation is provided since these units are located in a room where no missiles are postulated.
4. Pressure drops across the prefilters, HEPA, and charcoal filters are instrumented to indicate in the control room. Pressure drops across the HEPA and charcoal filters are instrumented to alarm in the control room. No recording of these signals is provided. Fan loss of flow is also instrumented to signal and alarm in the control room.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-25 (SHEET 3 OF 3)  5. Deleted 6. The size of the engineered safety feature filtration units precludes replacement as a single unit. The unit components are replaced individually.
7. Demisters are not provided. 
8. Electric heaters are used to reduce the relative humidity to 70 percent or less. The use of heating coils to control the relative humidity during DBAs is not credited in the respective DBA dose assessment.
9. Mounting frames for filter and charcoals are constructed of carbon steel coated with an inorganic nuclear grade paint. 
: 10. Internal welds are carbon steel coated with an inorganic nuclear grade paint. 
: 11. The deluge and drain system has been eliminated due to recurring problems experienced at other facilities associated with inadvertent wetting of the absorber. Temperature gauges have been installed to monitor any heat rise in the filter housing.   
: 12. Environmental conditions for systems considered are those specified under outside containment and radioactive area. 
: 13. Duct construction guidelines follow SMACNA in addition to ORNL-NSIC-65. 
: 14. Vacuum breakers are not used. This presents the probability of system leakage from pressure-relieving device leakage or failure.   
: 15. Test probes are not manifolded and are located in readily accessible locations with minimum piping. 
: 16. The accident analyses do not credit the heaters for humidity control.
: 17. Periodic testing to confirm a penetration of less than 0.5% at rated flow. 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-26  SINGLE-FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM  Component Malfunction Comments and Consequences    Spray Nozzles Clogged Large number of nozzles    precludes clogging of a    significant number. Pumps      Containment Fails to start Two pumps provided. Spray pump  Operation of one required. Automatically operated  valves (open on coinci-  dence of two out of four  high-high-high containment  pressure signals or 2/2  manual initiation of  spray system operation  from the control room):    Containment spray Fails to open Two valves provided. pump discharge isolation valve  Operation of one required. Valves operated from  control room for  recirculation Containment Fails to open Two lines in parallel,  sump recirculation  one to each spray pump. isolation  Operation of one required.             
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-27 (SHEET 1 OF 4)  DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW  TIME  THOUSAND  THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)  .00000  .0  .0  .0  .0 .00107 91449.9 49351.4 40349.8 21719.3
.101 40359.3 21797.0 20658.3 11113.7 .202 46617.0 25354.7 22408.0 12062.5
.302 46432.7 25480.6 22666.0 12212.7
.401 46869.1 25996.8 22243.1 11998.0
.501 46368.9 26033.7 21470.2 11589.9
.602 44240.1 25142.7 20750.4 11207.0
.702 44736.7 25706.9 20197.6 10911.5
.801 44554.7 25851.2 19705.5 10648.8
.902 43717.2 25588.5 19292.6 10429.1 1.00 42587.0 25134.1 18975.1 10260.1 1.10 41472.4 24675.8 18706.9 10116.8 1.20 40418.9 24239.3 18519.1 10016.3 1.30 39437.5 23830.9 18414.8 9960.8 1.40 38576.2 23476.8 18363.9 9933.9 1.50 37835.2 23180.4 18319.5 9910.0 1.60 37185.8 22928.6 18249.3 9871.5 1.70 36527.6 22670.7 18165.3 9825.3 1.80 35825.3 22391.2 18102.7 9790.9 1.90 35103.7 22112.9 18055.6 9765.3 2.00 34331.7 21820.3 17973.8 9720.8 2.10 33391.3 21435.8 17816.8 9635.2 2.20 32232.2 20918.3 17637.2 9537.5 2.30 30888.1 20287.8 17468.2 9446.2 2.40 29609.4 19693.0 17287.5 9348.6 2.50 28145.1 18960.7 16966.5 9174.7 2.60 24508.1 16681.4 16630.6 8993.2 2.70 21807.8 15032.8 16330.2 8831.3 2.80 20011.8 13965.9 16044.1 8677.3 2.90 18317.8 12903.2 15779.2 8535.2 3.00 16928.0 12017.7 15521.1 8397.0 3.10 15822.1 11306.9 15270.3 8262.8 3.20 14944.1 10739.5 15046.0 8143.3 3.30 14284.0 10314.6 14844.4 8036.1 3.40 13777.3 9987.8 14652.3 7933.9 3.50 13361.8 9717.3 14467.2 7835.4 3.60 12989.6 9474.7 14205.4 7694.8 3.70 12659.4 9261.3 14028.4 7600.9 3.80 12361.3 9070.4 13847.8 7504.7 3.90 12082.6 8891.7 13720.8 7438.2 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-27 (SHEET 2 OF 4)    BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW  TIME  THOUSAND  THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 4.00 11825.7 8727.7 13627.0 7388.7 4.20 11325.7 8401.6 13270.0 7197.5 4.40 10904.1 8121.2 12960.2 7032.1 4.60 10543.4 7866.4 12763.1 6927.4 4.80 10264.3 7653.1 12479.4 6774.8 5.00 10031.8 7461.6 12247.6 6650.9 5.20 9851.1 7298.9 13134.9 7138.0 5.40 9697.9 7150.0 12990.5 7057.9 5.60 9603.4 7039.6 12669.4 6886.1 5.80 9570.6 6967.1 12529.0 6811.9 6.00 9564.8 6913.1 12367.6 6726.8 6.20 9565.1 6864.3 12249.6 6665.6 6.40 9723.7 6927.8 12117.7 6596.4 6.60 10072.7 7127.9 12039.2 6556.2 6.80 9924.2 7261.9 11949.9 6508.5 7.00 8909.4 6974.7 11798.0 6424.8 7.20 8275.4 6686.1 11627.9 6330.5 7.40 8101.4 6554.2 11465.6 6241.2 7.60 8034.1 6477.0 11318.3 6160.3 7.80 7931.3 6382.2 11155.7 6070.8 8.00 7834.6 6267.4 10977.1 5972.0 8.20 7798.2 6161.1 10806.4 5877.7 8.40 7803.0 6069.0 10644.3 5788.2 8.60 7821.2 5991.1 10481.8 5698.3 8.80 7826.8 5914.7 10319.3 5608.5 9.00 7810.2 5836.0 10164.0 5522.5 9.20 7769.0 5755.2 10011.2 5438.0 9.40 7703.3 5672.7 9860.5 5354.7 9.60 7607.9 5582.5 9713.2 5273.3 9.80 7499.5 5497.8 9572.5 5195.7 10.0  7373.2 5411.9 9424.5 5114.1 10.2  7231.6 5320.2 9282.5 5036.1 10.4  7088.2 5230.3 9141.4 4958.7 10.4  7087.2 5229.7 9140.4 4958.1 10.4  7086.1 5229.1 9139.3 4957.5 10.6  6943.4 5142.6 9001.7 4882.1 10.8  6796.7 5056.6 8864.8 4807.2 11.0  6649.1 4971.8 8728.5 4732.7 11.2  6502.4 4888.3 8594.3 4659.5 11.4  6357.8 4806.1 8460.0 4586.5 11.6  6217.3 4725.9 8329.0 4515.4 11.8  6080.3 4647.4 8198.6 4444.7 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-27 (SHEET 3 OF 4)    BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW  TIME  THOUSAND  THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)  12.0  5947.1 4570.8 8070.2 4375.1 12.2  5817.7 4496.2 7945.7 4307.6 12.4  5689.8 4423.9 7820.6 4239.8 12.6  5566.8 4354.1 7699.8 4174.3 12.8  5445.5 4285.7 7578.7 4108.8 13.0  5327.3 4219.9 7460.5 4044.9 13.2  5207.8 4153.8 7308.6 3962.2 13.4  5077.7 4081.2 7096.6 3847.0 13.6  4933.2 3998.6 6962.0 3763.2 13.8  4774.0 3901.6 6743.9 3609.9 14.0  4615.1 3797.8 6729.5 3543.7 14.2  4468.4 3694.9 6471.6 3337.2 14.4  4349.0 3606.0 6586.0 3319.5 14.6  4246.7 3531.2 6091.1 3005.9 14.8  4153.8 3473.2 6258.5 3014.6 15.0  4048.6 3417.8 6256.0 2966.8 15.2  3947.5 3378.7 5573.0 2607.2 15.4  3844.1 3346.3 5805.5 2657.7 15.6  3719.0 3305.7 5945.1 2683.1 15.8  3608.1 3289.6 5479.0 2453.2 16.0  3486.7 3275.8 5194.7 2300.8 16.2  3362.9 3270.5 5095.5 2229.6 16.4  3220.0 3261.8 5073.6 2193.1 16.6  3025.9 3222.9 4973.3 2127.3 16.8  2725.4 3101.7 4688.8 1986.2 17.0  2481.9 2985.0 4330.8 1817.9 17.2  2199.8 2697.7 4020.6 1669.3 17.4  1987.4 2451.5 3709.4 1515.2 17.6  1804.7 2233.5 3547.4 1413.2 17.8  1636.2 2030.1 3734.9 1439.5 18.0  1478.6 1838.6 4080.5 1520.7 18.2  1330.6 1658.1 4340.7 1567.9 18.4  1196.0 1492.6 4064.5 1432.5 18.6  1065.0 1332.3 3763.4 1297.4 18.8 937.4 1174.4 3425.7 1154.3 19.0 823.9 1033.5 3022.7 994.5 19.2 731.6 918.6 2674.9 858.6 19.4 651.6 818.8 2261.5 708.0 19.6 594.6 748.0 1812.0 554.0 19.8 538.3 677.5 1332.1 398.9
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-27 (SHEET 4 OF 4)    BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW  TIME  THOUSAND  THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)    20.0 483.3 608.7 816.3 240.7 20.2 426.2 537.0 306.4 89.7 20.4 366.0 461.5  .0 .0 20.6 305.4 385.3  .0 .0 20.8 248.8 314.1  .0 .0 21.0 193.1 244.0  .0 .0 21.2 107.4 136.0  .0 .0 21.4 12.3 15.7  .0 .0 21.6 .0  .0  .0 .0 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-28 (SHEET 1 OF 2)  CONTAINMENT VENTILATION SYSTEMS COMPONENT DESIGN PARAMETERS 
Containment Coolers (normal)
Number 4 Pressure (psig) 200 Temperature (°F) 300 Water inlet temperature (°F) 95 Flowrate (each) (gal/min) 800 Heat removal rate (each) (btu/h) 2.36 x 106 Containment Cooler Fans (normal)
Type Vaneaxial Number 4 Flowrate (each) (sft3/min) 80,000 Static head (in. WG) 4.75 Motor horsepower (each) (hp) 80 Containment Recirculation Fans Type Vaneaxial Number 4 Flowrate (each) (sft3/min) 25,000 Static head (in. WG) 0.32 Motor horsepower (each) (hp) 7.5 Control-Rod Mechanism Cooling Fans Type Vaneaxial Number 2 Flowrate (each) (sft3/min) 40,000 Static head (in. WG) 9.0 Motor horsepower (each) (hp) 100 Reactor Cavity Cooling Fans Type Vaneaxial Number 2 Flowrate (each) (sft3/min) 17,000 Static head (in. WG) 2.46 Motor horsepower (each) (hp) 15 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-28 (SHEET 2 OF 2) 
Refueling Water Surface Ventilation Supply Fan Type Vaneaxial Number 1 Flowrate (each) (sft3/min) 7,500 Static head (in. WG) 4.5 Motor horsepower (each) (hp) 15 Refueling Water Surface Ventilation Exhaust Fan Type Vaneaxial Number 1 Flowrate (each) (sft3 min) 22,000 Static head (in. WG) 2.0 Motor horsepower (each) (hp) 15 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-29  SPRAY EVALUATION PARAMETERS    Spray flowrate (gal/min) 2480 (injection)  2290 (recirculation)
Containment sump volume (ft3) 4.92 x 104 Containment sprayed volume (ft3) 1.67 x 106 Minimum spray fall height (ft) 110  Elemental s (h-1) 10.0 (DF < 21)  0.0 (DF > 21)
Methyl s (h-1) 0.0 Particulate s (h-1) 5.4 (injection)  5.0 (recirculation, DF < 50)  0.0 (> 8 h)  0.5 (DF > 50 until 8 h)  pH (Spray injection) 4.5 pH (Spray recirculation) 7.7 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-30  SINGLE FAILURE ANALYSIS - PENETRATION ROOM FILTRATION SYSTEM  Component Malfunction Comments    Fan Fails The other fan and filter system will be available.      Fan discharge valve Fails to open Same as above.
Fan discharge valve Fails to close Check valve will prevent back flow.
Recirculation line valve Fails to open Recirculation fan will operate in the exhaust mode.
Recirculation line valve Fails to close The other system will be available.   
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-32 (SHEET 1 OF 2)  STEAM GENERATOR ISOLATION VALVE INFORMATION Service  Penetra- Valve  Flow  Location  Penetration  Normal    Valve  Valve Item    (No. of  tion Arrange- Direc- Relative to    Line    Valve Position with Position Post LOCA Closure  No. Penetrations) System  Type    ment    tion  Containment Valve Type  Size (in.)  Actuator    Signal Position Power Failure Indicator  Position  Time(s)                1 Main Steam (3) MS III 3 Out Outside Power operated 32 Air to open    SLIAS Open As Is Yes Closed        7        Check  Spring closed                      2 Main Steam MS III 3 Out Outside Gate 3 Air    SLIAS Closed  Closed Yes Closed < 7 (Note 2)  Isolation Valve                Bypass (3)                              3 Steam to MS III 3 Out Outside Stop check 3 Air Remote manual Closed Closed Yes Closed      ---  Auxiliary Feedwater                Pump Turbine Drive (2)                              4 Steam to MS III 3 Out Outside Globe 1 Air        T Open Closed Yes Closed        NA  Aux. Feedwater Pump                Drive Warming Line (2)                              5 Main Steam MS III 3 Out Outside  Globe 6 Air Remote manual Closed Closed Yes Closed      < 35  Atmospheric Relief (3)                              6 Feedwater (3) FW III 4 In Outside  Stop check 14 Electric motor Remote manual Open As is Yes Open        30                7 Auxiliary Feedwater (3) AFW III 4 In Outside  Stop check 4 Electric motor Remote manual Open As is Yes Open        14                                8 Steam Generator MS III 25 Out Outside Globe 2 Air AFPSS Open Closed Yes Closed        < 60  Blowdown (3)                              9 Steam Generator SS III 32 Out Outside Globe 3/8 Air Remote manual Open Closed Yes Closed        < 5  Blowdown Sample (3)        See Note 1 below                      10 Chemical FW III 4 In Outside Globe 1/2 Air        T Open Closed Yes Closed        < 5  Injection (3)               
1. Flow is isolated on AFPSS by valves inside containment. 2. Design requirement only, not operability requirement.
FNP-FSAR-6 REV 21 5/08 TABLE 6.2-32 (SHEET 2 OF 2) 
The "Valve Arrangement" number refers to figures 6.2-84 through 6.2-89. 
The abbreviations used in table 6.2-31 and 6.2-32 are as follows: 
SYSTEM DWS - Demineralized Water System CCS  - Component Cooling System SIS  - Safety Injection System RCS  - Reactor Coolant System WPS - Waste Processing System MS  - Main Steam System FW  - Feedwater System RHRS - Residual Heat Removal System CVCS - Chemical and Volume Control System SS  - Sampling System FHS  - Fuel Handling System RMS -  Radiation Monitoring System H&V  - Heating and Ventilation SA  - Service Air System IA  - Instrument Air System FWS -  Fire Water System SWS - Service Water System AFW - - Auxiliary Feedwater System 
SIGNALS T  - Containment Isolation Actuation Signal, Phase A S  - Safety Injection Signal P  - Containment Isolation Actuation Signal, Phase B CSAS - Containment Spray Actuation Signal SLIAS - Steam Line Isolation Actuation Signal AFPSS - Auxiliary Feedwater Pump Start Signal
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-33  ELECTRIC HYDROGEN RECOMBINER TYPICAL PARAMETERS 
Parameter Value  Power (maximum) 75 kW(a)  Capacity (minimum) 100 sft3/min  Heaters    -Number 5  -Heater surface area/heater 35 ft2  -Maximum heat flux 2850 Btu/h-ft2 or  5.8 Watts/in.2  -Maximum sheath temperature 1550&deg;F  Gas Temperature    -Inlet 80 to 155&deg;F  -In heater section 1150 to 1400&deg;F  Materials    -Outer structure 300-Series S.S.  -Inner structure Inconel-600  -Heater element sheath Incoloy-800  Dimensions    -Height 9 ft  -Width 4.5 ft  -Depth 5.5 ft  Weight 4500 lb 
a. Power can be controlled by SCR. Normal operating power for typical PWR containments is 48.9.
FNP-FSAR-6 REV 21 5/08 TABLE 6.2-34  POSTACCIDENT VENTING SYSTEM DESIGN PARAMETERS 
Parameters Value  Valves  -Design pressure (psig)  150(a) -Design temperature (&deg;F) Inside containment  300(a)  Outside containment  300(a)    Piping  -Design pressure (psig)  150(a) -Design temperature (&deg;F) Inside containment  300(a)  Outside containment  300(a)  HEPA Filter  -Number  1 -Air Flow (sft3/min)  500 -Approximate differential pressure (wg)  1.5 -Maximum differential pressure (loaded) (wg)  4.0 -Design temperature (&deg;F)  180 -Particulate removal efficiency (0.3 micron)  99.97    Charcoal  -Number  1 -Air flow (sft3/min)  500 -Differential pressure (wg)  2.7 -Design temperature (&deg;F)  180 -Charcoal type iodine impregnated-Elemental I2 removal efficiency  99.9 -Organic I2 removal efficiency  99.0 
a. Represents as installed ratings of system piping and valves. Actual design requirements may be substantially lower and may vary throughout the system.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-35  POSTACCIDENT SAMPLING SYSTEM DESIGN PARAMETERS 
Sample Vessel    Number 2 Number required for operation 1 Design pressure (psig) 150 Design temperature (&deg;F) 300 Material of construction Stainless steel    Valves    Design pressure (psig) 150 Design temperature (&deg;F) 300 Piping    Design pressure (psig) 150 Design temperature (&deg;F) 300 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-36  POSTACCIDENT MIXING SYSTEM DESIGN PARAMETERS 
Post-LOCA Containment Mixing Fans:      Type Vaneaxial  Number required 4  Flow (ft3/min) each 7500  Static pressure (in. w.g.) 2.3      Reactor Cavity Hydrogen Dilution Fans:        Unit 1 Type Centrifugal  Number required 2  Flow (ft3/min) each 270  Static pressure (in. w.g.) 126.7    Unit 2 Type Vaneaxial  Number required 2  Flow (ft3/min) each 1570  Static pressure (in. w.g.) 3.26 
FNP-FSAR-6 REV 22  8/09 TABLE 6.2-37  CONTAINMENT INTERIOR COATINGS SUMMARY          Ext. Ext.        Dry  Top Coat  Top Coat  Dry  Surface  Primer/Surfacer  Specific  Recoat  Recoat Specific Surface Area  Type  Manufacturer  Product No. Generic Type  Gravity  Product No. Generic Type  Gravity    (ft2)(a)              Carbon Ameron(b) Dimetcote Inorganic zinc    3.15 Amercoat 66 Epoxy polyamide 2.60 213,750 Steel  No. 6 (D-6)  Amercoat 90 Modified phenolic 2.58            Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58            Carboline Carbozinc 11 Inorganic zinc 4.61 Phenoline 305 Epoxy phenolic 1.73  17,500  SG (CZ-11)  Amercoat 90 Modified phenolic 2.58            4674 (Black)(c) Modified silicone 1.345    ---    ---          ---  2,300    with low chloride        content                4700 Aluminum free  1.28    ---    ---          --- < 2    paint                4674 (Aluminum)(c) Modified silicone 1.54    ---    ---          ---  12,000    aluminum with low        chloride content                Sterling U-475 ERN Epoxy varnish 1.001(d)    ---    ---          ---        28              --- Galvanized Hot dipped zinc 7.15    ---    ---          ---  62,932          Concrete Ameron(b) NU-KLAD Epoxy polyamide 1.95 Amercoat 66 Epoxy polyamide 2.60  80,000  110AA - solid filled  Amercoat 90 Modified phenolic 2.58            Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58            Amercoat 3366 Epoxy surfacer 2.12 Amercoat 90HS Epoxy phenolic 1.72            Amercoat 3367 Epoxy filler 1.80 Amercoat 90HS Epoxy phenolic 1.72   
a. For coating requirements see NMP-MA-011, Nuclear Coatings Program. b. Either system is acceptable for use as original system. c. Generally covered by insulation. d. Wet specific gravity at 75&deg;F. 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 1 OF 7)  CONTAINMENT ISOLATION VALVES (
==Reference:==
Table 6.2-31 [See Note])
No. Location FSAR Figure System P&ID Valve Identification Number      1 ic 6.2-87 (24) D-175038, Sh. 2 Q1E21V049    D-205038, Sh. 2 Q2E21V049  oc 6.2-87 (24) D-175038, Sh. 2 Q1E21V050    D-205038, Sh. 2 Q2E21V050      2 ic 6.2-89 (35) D-175043(a) Q1G31V013    D-205043 Q2G31V013  oc 6.2-89 (35) D-175043(a) Q1G31V012    D-205043 Q2G31V012      3 ic 6.2-87 (23) D-175038, Sh. 2 Q1E21V058    D-205038, Sh. 2 Q2E21V058  oc 6.2-87 (23) D-175038, Sh. 2 Q1E21V059    D-205038, Sh. 2 Q2E21V059      4 ic 6.2-89 (38) D-175037, Sh. 2 Q1B13V037    D-205037, Sh. 2 Q2B13V037  oc 6.2-89 (38) D-175037, Sh. 2 Q1B13V039    D-205037, Sh. 2 Q2B13V039      5 ic 6.2-87 (27) D-175037, Sh. 2 Q1B13V038    D-205037, Sh. 2 Q2B13V038  oc 6.2-87 (27) D-175037, Sh. 2 Q1B13V040    D-205037, Sh. 2 Q2B13V040    D-175037, Sh. 2 Q1B13V110    D-205037, Sh. 2 Q2B13V110      6 ic 6.2-88 (34) D-175037, Sh. 2 Q1B13V054    D-205037, Sh. 2 Q2B13V054  oc 6.2-88 (34) D-175039, Sh. 6 Q1E21V263A,B    D-205039, Sh. 2 Q2E21V263A,B  oc 6.2-88 (34) D-175038, Sh. 2 Q1E11V039A,B    D-175038, Sh. 2 Q1E11V040    D-205038, Sh. 2 Q2E11V039A,B    D-205038, Sh. 2 Q2E11V040      7 ic 6.2-84 (1) D-175042, Sh. 1 Q1G21V064    D-175042, Sh. 1 Q1G21V005        D-205042, Sh. 1(a) Q2G21V064    D-205042, Sh. 1(a) Q2G21V005  oc 6.2-84 (1) D-175042, Sh. 1 Q1G21V006        D-205042, Sh. 1(a) Q2G21V006    D-175042, Sh. 1 Q1G21V950    D-205042, Sh. 1(a) Q2G21V950 
a. This drawing is not presented in the FSAR because the corresponding drawing is applicable to both units. Note: Item numbers correlate with those on Table 6.2-31.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 2 OF 7)
No. Location FSAR Figure System P&ID Valve Identification Number        8 ic 6.2-86 (16) D-175010, Sh. 2 Q1E14V004    D-205010, Sh. 2 Q2E14V004  oc 6.2-86 (16) D-175010, Sh. 2 Q1E14V003    D-205010, Sh. 2 Q2E14V003      9 ic 6.2-85 (5) D-175041 Q1E11V001A,B    D-205041 Q2E11V001A,B  oc See note 2 of FSAR table 6.2-31.      10 ic 6.2-88 (30) D-175038, Sh. 2 Q1E11V042A,B    D-205038, Sh. 2 Q2E11V042A,B  oc 6.2-88 (30) D-175038, Sh. 2 Q1E11V023A,B    D-205038, Sh. 2 Q2E11V023A,B      11 ic  6.2-88 (33) D-175039, Sh. 1 Q1E21V253A,B,C    D-205039, Sh. 1 Q2E21V253A,B,C  oc 6.2-88 (33) D-175039, Sh. 1 Q1E21V254    D-205039, Sh. 1 Q2E21V254      12 ic 6.2-85 (6) D-175039, Sh. 1 Q1E21V249A    D-175039, Sh. 1 Q1E21V213    D-205039, Sh. 1 Q2E21V249A    D-205039, Sh. 1 Q2E21V213  oc 6.2-85 (6) D-175039, Sh. 1 Q1E21V249B    D-205039, Sh. 1 Q2E21V249B      13 ic 6.2-85 (7) D-175039, Sh. 1 Q1E21V119    D-205039, Sh. 1 Q2E21V119  oc 6.2-85 (7) D-175039, Sh. 6 Q1E21V257    D-175039, Sh. 6 Q1E21V258    D-205039, Sh. 2 Q2E21V257    D-205039, Sh. 2 Q2E21V258      14 ic 6.2-86 (15) D-175039, Sh. 1 Q1E21V115A,B,C    D-205039, Sh. 1 Q2E21V115A,B,C  See note 4 of FSAR table 6.2-31. oc 6.2-86 (15) D-175039, Sh. 1      -    D-205039, Sh. 2      -  See note 4 of FSAR table 6.2-31.      15 ic  D-175038, Sh. 3      -    D-205038, Sh. 3      -  See note 5 of FSAR table 6.2-31. oc  D-175038, Sh. 3      -    D-205038, Sh. 3      -  See note 5 of FSAR table 6.2-31.   
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 3 OF 7)
No. Location FSAR Figure System P&ID Valve Identification Number        16 ic 6.2-87 (24) D-175009, Sh. 2 Q1P15SV3104    D-205009, Sh. 2 Q2P15SV3104  oc 6.2-87 (24) D-175009, Sh. 2 Q1P15SV3331    D-205009, Sh. 2 Q2P15SV3331        17 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3103    D-205009, Sh. 1 Q2P15SV3103  oc 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3332    D-205009, Sh. 1 Q2P15SV3332        18 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3765    D-205009, Sh. 1 Q2P15SV3765    D-175009, Sh. 1 Q1P15SV3333    D-205009, Sh. 1 Q2P15SV3333        19 ic 6.2-87 (26) D-175067(b)      -    D-205067(a)(b)      -      20 ic 6.2-87 (10) D-175035, Sh. 1 Q1P18V002    D-205035, Sh. 1 Q2P18V002  oc 6.2-87 (10) D-175035, Sh. 1 Q1P18V001    D-205035, Sh. 1 Q2P18V001      20a  ic 6.2-87 (10) D-205035, Sh. 1 Q2P18V005  oc 6.2-87 (10) D-205035, Sh. 1 Q2P18V004      21 ic 6.2-87 (23) D-175034, Sh. 3 Q1P19V002    D-205034, Sh. 4 Q2P19V002  oc 6.2-87 (23) D-175034, Sh. 2 Q1P19HV3611    D-205034, Sh. 2 Q2P19HV3611      22 ic 6.2-87 (23) D-175010, Sh. 2 Q1E14V001    D-205010, Sh. 2 Q2E14V001  oc 6.2-87 (23) D-175010, Sh. 2 Q1E14HV3657    D-205010, Sh. 2 Q2E14HV3657      23 ic 6.2-85 (11) D-175010, Sh. 2 Q1E14V002    D-205010, Sh. 2 Q2E14V002  oc 6.2-85 (11) D-175010, Sh. 2 Q1E14HV3658    D-175010, Sh. 2 Q2E14HV3658      24 ic 6.2-85 (12) D-175010, Sh. 1 Q1P13V282    D-205010, Sh. 1 Q2P13V282  oc 6.2-85 (12) D-175010, Sh. 2 Q1P13V281    D-205010, Sh. 2 Q2P13V281 
b. This is only a general arrangement drawing.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 4 OF 7)  No. Location FSAR Figure System P&ID Valve Identification Number      24a ic 6.2-85 (12) D-175010, Sh. 1 Q1P13V302    D-205010, Sh. 1 Q2P13V302  oc 6.2-85 (12) D-175010, Sh. 2 Q1P13V301    D-205010, Sh. 2 Q2P13V301      25 ic 6.2-85 (13) D-175010, Sh. 1 Q1P13V283    D-205010, Sh. 1 Q2P13V283  oc 6.2-85 (13) D-175010, Sh. 2 Q1P13V284    D-205010, Sh. 2 Q2P13V284      25a ic 6.2-85 (13) D-175010, Sh. 1 Q1P13V304    D-205010, Sh. 1 Q2P13V304  oc 6.2-85 (13) D-175010, Sh. 2 Q1P13V303    D-205010, Sh. 2 Q2P13V303      26 ic  6.2-89 (39) D-175004, Sh. 1 Q1G21V291    D-175004, Sh. 1 Q1G21HV3376    D-205004, Sh. 1(a) Q2G21V291    D-205004, Sh. 1(a) Q2G21HV3376  oc 6.2-89 (39) D-175004, Sh. 1 Q1G21HV3377    D-205004, Sh. 1(a) Q2G21HV3377      27 ic 6.2-88 (30) D-175003, Sh. 1 Q1P16V206A,B,C,D        D-205003, Sh. 1(a) Q2P16V206A,B,C,D  oc 6.2-88 (30) D-175003, Sh. 1 Q1P16V010A,B,C,D      D-205003, Sh. 1(a) Q2P16V010A,B,C,D    D-175003, Sh. 1 Q1P16V205A,B,C,D    D-205003, Sh. 1(a) Q2P16V205A,B,C,D      28 ic 6.2-85 (9) D-175003, Sh. 1 Q1P16V207A,B,C,D    D-205003, Sh. 1(a) Q2P16V207A,B,C,D  oc 6.2-85 (9) D-175003, Sh. 1  Q1P16V043A,B,C,D    D-205003, Sh. 1(a) Q2P16V043A,B,C,D    D-175003, Sh. 1 Q1P16V044A,B,C,D    D-205003, Sh. 1(a) Q2P16V044A,B,C,D    D-175003, Sh. 1 Q1P16V208A,B,C,D    D-205003, Sh. 1(a) Q2P16V208A,B,C,D      29 ic 6.2-88 (30) D-175002, Sh. 2 Q1P17V083    D-205002, Sh. 2 Q2P17V083  oc 6.2-88 (30) D-175002, Sh. 2 Q1P17V082    D-205002, Sh. 2 Q2P17V082    D-175002, Sh. 2 Q1P17V158    D-205002, Sh. 2 Q2P17V158      30 oc 6.2-86 (21) D-175010, Sh. 1 Q1P23V002A,B    D-205010, Sh. 1 Q2P23V002A,B 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 5 OF 7)
No. Location FSAR Figure System P&ID Valve Identification Number        31 ic  6.2-88 (29) D-175002, Sh. 2 Q1P17V097    D-205002, Sh. 2 Q2P17V097  oc 6.2-88 (29) D-175002, Sh. 2 Q1P17V099    D-205002, Sh. 2 Q2P17V099    D-175002, Sh. 2 Q1P17V155    D-205002, Sh. 2 Q2P17V155      32 ic 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3184    D-205002, Sh. 2 Q2P17HV3184  oc 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3045    D-205002, Sh. 2 Q2P17HV3045      33 ic 6.2-85 (8) D-175002, Sh. 2 Q1P17V159    D-205002, Sh. 2 Q2P17V159  oc 6.2-85 (8) D-175002, Sh. 2 Q1P17HV3095    D-205002, Sh. 2 Q2P17HV3095    D-175002, Sh. 2 Q1P17V153    D-205002, Sh. 2 Q2P17V153      34 ic  6.2-87 (24) D-175002, Sh. 2 Q1P17HV3443    D-205002, Sh. 2 Q2P17HV3443  oc 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3067    D-205002, Sh. 2 Q2P17HV3067    D-175002, Sh. 2 Q1P17V154    D-205002, Sh. 2 Q2P17V154      35 ic 6.2-86 (14) D-175038, Sh. 1 Q1E21V078A,B,C    D-205038, Sh. 1 Q2E21V078A,B,C    D-175038, Sh. 1 Q1E21V079A,B,C    D-205038, Sh. 1 Q2E21V079A,B,C    D-175038, Sh. 1 Q1E21V066A,B,C    D-205038, Sh. 1 Q2E21V066A,B,C  oc 6.2-86 (14) D-175038, Sh. 1 Q1E21V068    D-175038, Sh. 1 Q1E21V072    D-175038, Sh. 1 Q1E21V063    D-205038, Sh. 1 Q2E21V068    D-205038, Sh. 1 Q2E21V072    D-205038, Sh. 1 Q2E21V063      36 ic 6.2-88 (30) D-175038, Sh. 3 Q1E13V002A,B    D-205038, Sh. 3 Q2E13V002A,B  oc 6.2-88 (30) D-175038, Sh. 3 Q1E13V005A,B    D-205038, Sh. 3 Q2E13V005A,B      37 oc 6.2-86 (18) D-175038, Sh. 2 Q1E11V025A,B    D-205038, Sh. 2 Q2E11V025A,B    D-175038, Sh. 2 Q1E11V026A,B    D-205038, Sh. 2 Q2E11V026A,B FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 6 OF 7)
No. Location FSAR Figure System P&ID Valve Identification Number        38 oc 6.2-86 (18) D-175038, Sh. 3 Q1E13V003A,B    D-205038, Sh. 3 Q2E13V003A,B    D-175038, Sh. 3 Q1E13V004A,B    D-205038, Sh. 3 Q2E13V004A,B      39 ic 6.2-85 (39) D-175038, Sh. 2 Q1E21V052    D-205038, Sh. 2 Q2E21V052  oc 6.2-85 (39) D-175038, Sh. 2 Q1E21V091    D-205038, Sh. 2 Q2E21V091      40 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15HV3766    D-205009, Sh. 1 Q2P15HV3766  oc 6.2-87 (24) D-175009, Sh. 1 Q1P15HV3334    D-205009, Sh. 1 Q2P15HV3334      41 ic See note 7 of FSAR table 6.2-31. oc 6.2-86 (19) D-175037, Sh. 2 Q1B13V026B    D-205037, Sh. 2 Q2B13V026B 42 ic 6.2-87 (28) D-175042, Sh. 1 Q1G21V082    D-205042, Sh. 1(a) Q2G21V082  oc 6.2-87 (28)  D-175042, Sh. 1 Q1G21V001    D-205042, Sh. 2(a) Q2G21V001      43 ic  6.2-84 (2) D-175038, Sh. 1 Q1E21V062A,B,C    D-205038, Sh. 1 Q2E21V062A,B,C  oc 6.2-84 (2) D-175038, Sh. 1 Q1E21V016A,B    D-205038, Sh. 1 Q2E21V016A,B      44 ic 6.2-86 (17) D-175038, Sh. 1 Q1E21V076A,B    D-205038, Sh. 1 Q2E21V076A,B  oc 6.2-86 (17) D-175038, Sh. 2 Q1E11V044    D-205038, Sh. 2 Q2E11V044      45 ic 6.2-88 (30) D-175003, Sh. 2 Q1P16V075          D-205003, Sh. 2(a) Q2P16V075  oc 6.2-88 (30) D-175003, Sh. 2 Q1P16V071    D-205003, Sh. 2(a) Q2P16V071    D-175003, Sh. 2 Q1P16V204    D-205003, Sh. 2(a) Q2P16V204      46 ic 6.2-88 (29) D-175003, Sh. 2 Q1P16V081      D-205003, Sh. 2(a) Q2P16V081  oc 6.2-88 (29) D-175003, Sh. 2 Q1P16V072    D-205003, Sh. 2(a) Q2P16V072    D-175003, Sh. 2 Q1P16V203    D-205003, Sh. 2(a) Q2P16V203      47 ic 6.2-87 (23) D-175004, Sh. 1 Q1G21V204          D-205004, Sh. 1(a) Q2G21V204  oc 6.2-87 (23) D-175004, Sh. 1 Q1G21HV3380    D-205004, Sh. 1(a) Q2G21HV3380 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-39 (SHEET 7 OF 7)
No. Location FSAR Figure System P&ID Valve Identification Number        48 ic  6.2-89 (37)  D-175019 Q1E23V022A,B,C,D    D-205019 Q2E23V022A,B,C,D  oc  6.2-89 (37)  D-175019 Q1E23V023A,B      D-205019 Q2E23V023A,B      49 ic  6.2-86 (20)  D-175019 Q1E23V025A,B    D-205019 Q2E23V025A,B  oc  6.2-86 (20)  D-175019 Q1E23V024A,B    D-205019 Q2E23V024A,B      50 ic  6.2-86 (16)  D-175019 Q1E23V003    D-205019 Q2E23V003  oc  6.2-86 (16)  D-175019 Q1E23V002    D-205019 Q2E23V002      51 ic  6.2-87 (23)  D-175047 Q1P11V002    D-205047 Q2P11V002  oc  6.2-87 (23)  D-175047 Q1P11V001    D-205047 Q2P11V001      52 ic  6.2-87 (23)  D-175034, Sh. 1 Q1P19V004      D-205034, Sh. 4(a) Q2P19V004  oc  6.2-87 (23)  D-175034, Sh. 1 Q1P19HV2228    6.2-87 (23A)  D-205034, Sh. 4(a) Q1P19V1099    Q2P19V006      Q2P19V1099 53 ic  6.2-89 (40)  D-206164(b)    ---        oc  6.2-89 (40)  D-206164(b)    --- 
a. This drawing is not presented in the FSAR because the corresponding drawing is applicable to both units. 
: b. This is only a general arrangement drawing.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-40  STEAM GENERATOR ISOLATION VALVES  (
==Reference:==
TABLE 6.2-32 [See Note] )
Valve Item  FSAR  System Identification No. Location  Figure  P&ID Number    1 oc 6.2-84 (3) D-175033 Sh. 1 QV001A,B,C    QV002A,B,C        2 oc 6.2-84 (3) D-175033 Sh. 1 QV003A,B,C,D,E,F          3 oc 6.2-84 (3) D-175033 Sh. 2 Q-N12V001A-A,B-B          4 oc 6.2-84 (3) D-175033 Sh. 2 HV3234A,B          5 oc 6.2-84 (3) D-175033 Sh. 1 PV3371A,B,C          6 oc 6.2-84 (4) D-170117 Sh. 4 Q-N21V001A-B, B-B,C-B        7 oc 6.2-84 (4) D-175007 V0011A,B,C          8 oc 6.2-87 (25) D-175071 Sh. 1 7614A,B,C    D-205071 Sh. 1        9 oc 6.2-88 (32) D-175009 Sh. 2 HV3328, HV3329,    D-205009 Sh. 2 HV3330        10 oc 6.2-84 (4) D-175000 Sh. 1 QV001A,B,C 
Note:  Item numbers correlate with those on Table 6.2-32.
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-41  CONTAINMENT PRESSURE/TEMPERATURE FOR 600 gal/min SERVICE WATER FLOW, 0.003 FOULING FACTOR Power Uprate Case Peak Pressure (psia)  Time (s)  Peak Temp.    (&deg;F)
Time (s)      MSLB CASE 1, P0 = 0.0 58.3 1811 368  60.1 MSLB CASE 1, P0 = -1.5 56.6 1811 383 100.1 MSLB CASE 2 55.7 1811 355 150.1 MSLB CASE 3, P0 = 0.0 55.4 1811 362 170.1 MSLB CASE 3, P0 = -1.5 53.7 1811 370 195.1 MSLB CASE 4 59.9 1821 365 205.1 MSLB CASE 5 59.9 1811 324  70.1 MSLB CASE 6 57.3 1811 331 195.1 MSLB CASE 7 56.7 1811 354 215.1 MSLB CASE 8 61.6 1801 363 200.4 MSLB CASE 9, P0 = 0.0 63.0 400.7 294  75.1 MSLB CASE 9, P0 = 3.0 67.0 400.7 288  80.0 MSLB CASE 10 58.6 1891 313 260.8 MSLB CASE 11, P0 = 0.0 57.3 1831 342 300.8 MSLB CASE 11, P0 = -1.5 56.0 1832 347 340.8 MSLB CASE 12, P0 = 0.0 63.3 1811 359 180.1 MSLB CASE 12, P0 = 3.0 67.1 1811 347 165.1 MSLB CASE 13 61.0 380.7 273 380.7 MSLB CASE 14 43.1 1801 262 760.8 MSLB CASE 15 33.9 2001 302 290.8 MSLB CASE 16 45.3 1331 324 260.8      RSG Case          MSLB CASE 1, P0 = 3.0 59.4 1832 351  87.2 MSLB CASE 1, P0 = -1.5 52.6 1828 367  92.2 MSLB CASE 8, P0 = 3.0 62.2 1498 330 192 MSLB CASE 8, P0 = -1.5 57.2 1503 347 157 MSLB CASE 9, P0 = 3.0 64.5 482 347  87.2 MSLB CASE 9, P0 = -1.5 59.2 1828 363  57.2 MSLB CASE 12, P0 = 3.0 65.4 1518 331 162 MSLB CASE 12, P0 = -1.5 60.3 1518 347 132 MSLB CASE 13, P0 = 3.0 66.7 572 342 87.2 MSLB CASE 13, P0 = -1.5 61.3 573 359 57.2 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-42  DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS PRINCIPLE PARAMETERS DURING REFLOOD FLOODING CORE DOWNCOMER  INJECTION  TIME TEMP RATE CARRYOVER HEIGHT HEIGHT FLOW TOTAL ACCUMULATOR SPILL ENTHALPY (seconds) (degree F) (in/sec) FRACTION (ft) (ft) FRACTION (Pounds Mass per Second) (Btu/lbm)            21.6 181.8 .000 .000  .00  .00 .333 .0 .0 .0 .00 22.3 179.9 26.292 .000  .67 1.47 .000 7724.4 7724.4 .0 89.50 22.5 179.0 29.720 .000 1.02 1.55 .000 7645.9 7645.9 .0 89.50 23.6 178.4 2.675 .300 1.50 4.42 .404 7191.8 7191.8 .0 89.50 24.6 178.5 2.516 .424 1.64 7.02 .440 6874.3 6874.3 .0 89.50 28.2 178.9 3.819 .628 2.00 15.37 .612 5612.0 5612.0 .0 89.50 29.7 178.9 4.533 .672 2.19 15.62 .673 4898.1 4898.1 .0 89.50 30.7 178.9 4.340 .689 2.30 15.62 .670 4739.6 4739.6 .0 89.50 31.7 179.0 4.378 .702 2.42 15.62 .677 4924.6 4469.4 .0 89.36 32.5 179.1 4.269 .709 2.51 15.62 .675 4823.5 4366.1 .0 89.36 37.9 180.0 3.765 .730 3.00 15.62 .659 4236.1 3767.1 .0 89.33 44.2 181.8 3.391 .737 3.50 15.62 .641 3699.8 3221.1 .0 89.31 51.3 184.3 3.083 .739 4.00 15.62 .622 3212.0 2725.5 .0 89.27 52.7 184.8 2.705 .734 4.09 15.62 .579 2434.2 1937.4 .0 89.19 53.7 185.2 3.222 .742 4.15 15.59 .634 487.1 .0 .0 88.00 54.7 185.6 3.274 .742 4.22 15.47 .637 480.7 .0 .0 88.00 59.7 188.0 3.086 .742 4.57 14.88 .633 484.3 .0 .0 88.00 66.5 191.9 2.851 .740 5.00 14.22 .627 488.5 .0 .0 88.00 75.7 198.1 2.575 .738 5.54 13.55 .618 493.2 .0 .0 88.00 84.2 204.2 2.354 .736 6.00 13.13 .608 496.5 .0 .0 88.00 94.7  212.0 2.128 .733  6.52 12.83 .595  499.4 .0 .0  88.00 105.3  219.9 1.947 .732  7.00 12.72 .582  501.5 .0 .0  88.00 118.7  229.5 1.780 .731  7.56 12.80 .567  503.2 .0 .0  88.00 130.1  236.6 1.683 .731  8.00 13.00 .556  504.1 .0 .0  88.00 144.7  244.5 1.605 .733  8.54 13.37 .547  504.9 .0 .0  88.00 158.0  250.7 1.563 .735  9.00 13.79 .543  505.2 .0 .0  88.00 172.7  256.8 1.538 .738  9.50 14.29 .541  505.4 .0 .0  88.00 180.7  259.9 1.531 .741  9.77 14.57 .541  505.5 .0 .0  88.00 187.7  262.4 1.535 .743 10.00 14.82 .543  505.4 .0 .0  88.00 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-43 (SHEET 1 of 2)  DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES    Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      187.8 174.1 216.5 334.1 104.8 192.8 173.9 216.3 334.2 104.6 197.8 172.9 215.1 335.2 104.7 202.8 172.9 215.0 335.2 104.5 207.8 172.1 214.1 336.0 104.5 212.8 171.4 213.2 336.7 104.5 217.8 171.4 213.2 336.7 104.3 222.8 170.7 212.3 337.4 104.3 227.8 169.9 211.4 338.2 104.3 232.8 169.9 211.3 338.2 104.2 237.8 169.2 210.4 339.0 104.2 242.8 168.4 209.4 339.7 104.2 247.8 168.3 209.4 339.8 104.0 252.8 167.6 208.4 340.6 104.0 257.8 166.8 207.4 341.3 104.0 262.8 166.7 207.3 341.4 103.9 267.8 165.9 206.3 342.2 103.9 272.8 165.8 206.2 342.3 103.7 277.8 165.0 205.2 343.1 103.7 282.8 164.1 204.2 344.0 103.8 287.8 164.0 204.0 344.1 103.6 292.8 163.2 202.9 345.0 103.6 297.8 163.0 202.7 345.1 103.5 302.8 162.1 201.7 346.0 103.5 307.8 161.9 201.4 346.2 103.3 312.8 161.1 200.3 347.1 103.4 317.8 160.2 199.2 347.9 103.4 322.8 159.9 198.9 348.2 103.3 327.8 159.0 197.8 349.1 103.3 332.8 158.7 197.4 349.4 103.2 337.8 158.4 197.1 349.7 103.1 342.8 157.5 195.9 350.6 103.1 347.8 157.2 195.5 350.9 103.0 352.8 162.9 202.6 345.2 103.8 357.8 161.8 201.3 346.3 103.9 362.8 161.3 200.6 346.8 103.8 367.8 160.8 200.0 347.3 103.8 372.8 160.2 199.3 347.9 103.7 377.8 159.6 198.5 348.5 103.6 382.8 159.0 197.7 349.1 103.6 387.8 158.3 196.9 349.8 103.6 392.8 157.6 196.1 350.5 103.5 397.8 156.9 195.2 351.2 103.5 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-43 (SHEET 2 of 2)  Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      402.8 156.2 194.3 351.9 103.5  407.8 156.1 194.1 352.0 103.3 412.8 155.4 193.2 352.8 103.3  417.8 154.6 192.3 353.5 103.3  422.8 154.3 191.9 353.8 103.1  427.8 153.4 190.8 354.7 103.1  432.8 153.0 190.3 355.1 103.0  437.8 152.0 189.1 356.1 103.1  442.8 151.5 188.4 356.7 103.0  447.8 150.8 187.6 357.3 103.0  452.8 150.1 186.7 358.0 102.9  457.8 149.7 186.2 358.4 102.8  462.8 148.8 185.1 359.3 102.8  467.8 148.2 184.3 359.9 102.8  472.8 147.5 183.4 360.7 102.7  477.8 147.0 182.8 361.1 102.6  482.8 145.9 181.4 362.2 102.7  487.8 152.0 189.0 356.2 103.3  492.8 151.4 188.3 356.7 103.2  497.8 150.8 187.6 357.3 103.1  502.8 149.9 186.5 358.2 103.1  507.8 149.3 185.7 358.8 103.1  512.8 148.4 184.6 359.7 103.0  517.8 147.6 183.6 360.5 103.0  522.8 72.3 90.0 435.8 123.2  711.2 72.3 90.0 435.8 123.2  711.3 76.6 94.4 431.5 119.3  712.8 76.6 94.4 431.5 119.2 1243.3 76.6 94.4 431.5 119.2 1243.4 67.3 77.4 440.8 45.3 2139.0 59.0 67.9 449.1 46.8 2139.1 59.0 67.9 9.8 8.2 2319.0 58.1 66.9 10.7 8.3 2319.1 58.1 66.9 475.8 95.8 3600.0 51.8 59.7 482.1 96.9 3600.1 42.6 49.0 491.3 86.0 10000.0 31.0 35.6 502.9 88.1 100000.0 16.6 19.1 517.3 90.6 1000000.0 7.1 8.2 526.8 92.3 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-44  DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MINIMUM SAFEGUARDS TIME (SECONDS) .00  21.60  21.60 187.74 711.31 1243.32 3600.00  MASS (THOUSAND LBM) INITIAL IN RCS AND ACC  620.40 620.40 620.40 620.40 620.40 620.40 620.40 ADDED MASS PUMPED INJECTION .00 .00 .00  77.84 343.84 614.17 1765.59 TOTAL ADDED .00 .00 .00  77.84 343.84 614.17 1765.59 ***  TOTAL AVAILABLE*** 620.40 620.40 620.40 698.24 964.24 1234.57 2385.99 DISTRIBUTION REACTOR CO0LANT 417.47  48.24  68.95 130.25 130.25 130.25 130.25 ACCUMULATOR 202.93 155.44 134.73 .00 .00 .00 .00  TOTAL CONTENTS 620.40 203.68 203.68 130.25 130.25 130.25 130.25 EFFLUENT BREAK FLOW .00 416.71 416.71 567.97 833.98 1104.30 2255.73  ECCS SPILL .00 .00 .00 .00 .00 .00 .00 TOTAL EFFLUENT .00 416.71 416.71 567.97 833.98 1104.30 2255.73 ***  TOTAL ACCOUNTABLE  ***  620.40 620.39 620.39 698.22 964.23 1234.55 2385.97 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-45  DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE MINIMUM SAFEGUARDS Time (Seconds) .0021.60 21.60 187.74 711.31 1243.32 3600.00          Energy (Million Btu) Initial Energy In RCS, ACC, S. Gen 675.98675.98 675.98 675.98 675.98 675.98 675.98 Added Energy Pumped Injection .00.00 .00 6.85 30.26 54.05 214.96  Decay Heat .005.39 5.39 21.62 59.57 91.60 203.05  Heat from Secondary .00-5.74 -5.74 -5.74 -3.77 -2.20 -2.20          Total Added .00-.35 -.35 22.74 86.06 143.45 415.81        ***TOTAL AVAILABLE*** 675.98675.63 675.63 698.72 762.04 819.43 1091.80        Distribution Reactor Coolant 244.8010.56 12.41 33.81 33.81 33.81 33.81 Accummulator 18.1613.91 12.06 .00 .00 .00 .00 Core Stored 18.939.53 9.53 4.05 3.90 3.68 2.71  Primary Metal 120.89114.27 114.27 91.33 64.17 53.37 39.48  Secondary Metal 76.0175.71 75.71 68.26 51.06 39.88 29.57  Steam Generator 197.20196.23 196.23 173.56 127.03 99.38 73.89          Total Contents 675.98420.21 420.21 371.01 279.98 230.12 179.45        Effluent Break Flow .00254.95 254.95 321.10 475.47 575.84 900.23  ECCS Spill .00.00 .00 .00 .00 .00 .00 Total Effluent .00254.95 254.95 321.10 475.47 575.84 900.23        ***TOTAL ACCOUNTABLE*** 675.98675.16 675.16 692.11 755.44 805.95 1079.69 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-46 (SHEET 1 OF 5)  DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      21.6 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 .0 .0 .0 .0 22.5 .0 .0 .0 .0 22.6 114.5 135.1 .0 .0 22.7 47.0 55.5 .0 .0 22.8 41.3 48.8 .0 .0 22.9 47.3 55.8 .0 .0 23.0 53.8 63.5 .0 .0 23.1 60.0 70.8 .0 .0 23.2 65.7 77.6 .0 .0 23.3 71.2 84.0 .0 .0 23.4 76.3 90.1 .0 .0 23.5 81.2 95.8 .0 .0 23.5 82.4 97.2 .0 .0 23.6 85.9 101.4 .0 .0 23.7 90.3 106.6 .0 .0 23.8 94.7 111.7 .0 .0 23.9 98.8 116.6 .0 .0 24.0 102.8 121.4 .0 .0 24.1 106.7 126.0 .0 .0 24.2 110.5 130.4 .0 .0 24.3 114.1 134.7 .0 .0 24.4 117.7 138.9 .0 .0 24.5 121.1 143.0 .0 .0 24.6 124.5 147.0 .0 .0 25.6 154.3 182.2 .0 .0 26.6 394.1 466.9 3497.1 447.9 27.4 478.5 567.7 4297.3 574.8 27.7 477.9 567.0 4288.2 576.4 28.7 466.9 553.9 4191.1 567.4 29.7 454.6 539.3 4080.7 556.0 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-46 (SHEET 2 OF 5)
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      30.7 442.5 524.8 3969.8 544.3 31.4 491.3 583.0 4465.5 589.1 31.7 486.5 577.4 4413.9 586.9 32.7 476.0 564.8 4321.5 576.4 33.7 465.8 552.6 4229.1 566.3 34.7 456.0 540.9 4139.8 556.5 35.7 446.6 529.7 4053.7 547.0 36.5 439.4 521.1 3987.0 539.6 36.7 437.7 519.0 3970.6 537.8 37.7 429.1 508.7 3890.6 529.0 38.7 420.9 499.0 3813.4 520.5 39.7 413.0 489.6 3738.8  512.2 40.7 405.5 480.6 3666.7  504.3 41.7 398.3 471.9 3597.1  496.5 42.5 392.7 465.3 3542.9  490.5 42.7 391.3 463.6 3529.6  489.0 43.7 384.6 455.6 3464.2  481.8 44.7 378.1 447.9 3400.8  474.7 45.7 371.9 440.5 3339.2  467.9 46.7 365.8 433.3 3279.4  461.2 47.7 360.0 426.4 3221.3  454.7 48.7 354.4 419.6 3164.7  448.4 49.1 352.2 417.0 3142.5  445.9 49.7 348.9 413.1 3109.6  442.2 50.7 343.6 406.8 3055.8  436.2 51.7 338.4 400.7 3003.4  430.3 52.7 333.4 394.7 2952.3  424.6 53.7 328.6 388.9 2902.4  418.9 54.7 323.8 383.3 2853.5  413.4 55.7 183.9 217.2  647.8  157.9 56.7 183.4 216.7  648.5  157.7 57.7 183.0 216.2  649.4  157.6 58.7 182.5 215.7  650.3  157.4 59.7 182.1 215.2  651.2  157.2 60.7 181.7 214.6  652.1  157.1 61.7 181.3 214.1  653.1  156.9 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-46 (SHEET 3 OF 5) 
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      62.7 180.8 213.6  654.0  156.7 63.7 180.4 213.1  654.9  156.6 64.7 180.0 212.6  655.8  156.4 65.7 179.5 212.1  656.7  156.3 66.6 179.2 211.7  657.6  156.1 66.7 179.1 211.6  657.7  156.1 67.7 178.7 211.1  658.6  155.9 68.7 178.3 210.6  659.5  155.8 69.7 177.9 210.1  660.4  155.6 70.7 177.4 209.6  661.3  155.5 71.7 177.0 209.1  662.2  155.3 72.7 176.6 208.6  663.1  155.2 73.7 176.2 208.1  664.0  155.0 74.7 175.7 207.6  665.0  154.8 75.7 175.3 207.1  665.9  154.7 76.7 174.9 206.6  666.8  154.5 77.7 174.5 206.1  667.7  154.4 78.7 174.1 205.6  668.6  154.2 79.7 173.6 205.1  669.5  154.1  80.7 173.2 204.6 670.5 153.9  81.7 172.8 204.1 671.4 153.8  82.7 172.4 203.6 672.3 153.6  84.7 171.5 202.6 674.2 153.3  86.7 170.7 201.6 676.0 153.0  87.5 170.3 201.2 676.8 152.9  88.7 169.8 200.6 677.9 152.7  90.7 169.0 199.6 679.8 152.4  92.7 168.1 198.6 681.6 152.1  94.7 167.3 197.6 683.5 151.8  96.7 166.4 196.6 685.4 151.5  98.7 165.5 195.5 687.3 151.2 100.7 164.7 194.5 689.2 150.9 102.7 163.8 193.5 691.1 150.6 104.7 162.9 192.4 693.0 150.3 106.7 162.1 191.4 694.9 150.0 108.7 161.2 190.4 696.8 149.7 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-46 (SHEET 4 OF 5) 
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      110.1 160.6 189.6 698.1 149.5 110.7 160.3 189.3 698.7 149.4 112.7 159.4 188.3 700.6 149.1 114.7 158.5 187.2 702.5 148.8 116.7 157.6 186.1 704.3 148.5 118.7 156.7 185.1 706.2 148.2 120.7 155.8 184.0 708.1 147.9 122.7 154.9 183.0 709.9 147.6 124.7 154.0 181.9 711.8 147.3 126.7 153.1 180.8 713.7 147.0 128.7 152.2 179.7 715.5 146.7 130.7 151.3 178.7 717.4 146.4 132.7 150.4 177.6 719.2 146.1 134.7 149.4 176.5 721.1 145.8 136.7 148.5 175.4 722.9 145.4 138.7 147.6 174.3 724.7 145.1 140.7 146.7 173.2 726.6 144.8 142.7 145.8 172.1 728.4 144.5 144.7 144.8 171.0 730.2 144.2 146.7 143.9 169.9 732.0 143.9 148.7 143.0 168.8 733.9 143.6 150.7 142.0 167.7 735.7 143.3 152.7 141.1 166.6 737.5 143.0 154.7 140.2 165.5 739.3 142.7 156.7 139.2 164.4 741.2 142.4 158.7 138.3 163.3 743.0 142.0 160.7 137.4 162.3  744.9 141.9 161.9 137.1 161.9  745.6 141.9 162.7 136.9 161.6  746.0 141.9 164.7 136.3 160.9  747.2 141.8 166.7 135.7 160.2  748.3 141.7 168.7 135.1 159.5  749.4 141.7 170.7 134.5 158.9  750.5 141.6 172.7 134.0 158.2  751.6 141.5 174.7 133.4 157.5  752.7 141.4 176.7 132.8 156.8  753.7 141.3 178.7 132.3 156.2  754.8 141.3 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-46 (SHEET 5 OF 5) 
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      180.7 131.7 155.5  755.9 141.2 182.7 131.1 154.8  756.9 141.1 184.7 130.6 154.2  758.0 141.0 186.7 130.0 153.5  759.1 140.9 188.7 129.5 152.9  760.1 140.8 190.7 128.9 152.2  761.1 140.7 192.1 128.6 151.8  761.9 140.6 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-47    DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS PRINCIPLE PARAMETERS DURING REFLOOD Flooding        Core Downcomer  Injection  Time Temp Rate Carryover Height Height Flow Total Accum Spill Enthalpy Seconds &deg;F in/sec Fraction    (ft)      (ft)    Frac          (Pounds Mass per Second)            Btu/lbm          21.6    183.1 .000 .000  .00  .00 .333 .0 .0 .0 .00 22.3 181.1 22.834 .000  .53 1.85 .000 7698.9 7698.9 .0 89.50 22.5 179.3 28.240 .000 1.06 1.85 .000 7582.0 7582.0 .0 89.50 23.5 178.4 2.939 .303 1.50 5.23 .418 7175.6 7175.6 .0 89.50 24.5 178.4 2.797 .436 1.65 8.77 .451 6865.9 6865.9 .0 89.50 27.4 178.3 5.018 .634 2.01 15.62 .679 5382.1 5382.1 .0 89.50 28.7 178.3 4.633 .673 2.19 15.62 .676 5120.7 5120.7 .0 89.50 30.7 178.4 4.259 .702 2.42 15.62 .670 4800.0 4800.0  .0 89.50 31.4 178.5 4.556 .710 2.50 15.62 .691 5360.3 4479.6 .0 89.25 36.5 179.7 4.036 .732 3.00 15.62 .674 4756.1 3854.9 .0 89.22 42.5 181.9 3.653 .739 3.51 15.62 .659 4224.7 3304.3 .0 89.17 49.1 184.9 3.345 .742 4.01 15.62 .644 3755.9 2820.2 .0 89.13 55.7 188.1 2.267 .729 4.45 15.62 .506 986.7 .0 .0 88.00 56.7 188.6 2.261 .729 4.50 15.62  .506 986.7  .0 .0 88.00 66.6 194.5 2.202 .730 5.00 15.62 .506 986.7 .0  .0 88.00 77.7 202.9 2.137 .732 5.54 15.62 .507 986.7  .0 .0 88.00 87.5 211.2 2.080 .733 6.00 15.62 .507 986.7 .0 .0 88.00 98.7 221.1 2.013 .736 6.51 15.62 .508 986.8  .0 .0 88.00 110.1 230.4 1.945 .738 7.00 15.62  .508 986.8 .0 0 88.00 122.7 239.2 1.870 .740 7.53 15.62  .508 986.9 .0 .0 88.00 134.7 246.4 1.799 .742 8.00 15.62  .509 987.0 .0 .0 88.00 148.7 253.6 1.717 .744 8.53 15.62  .508 987.1 .0 .0 88.00 161.9 259.5 1.642 .745 9.00 15.62  .509 987.2 .0 .0 88.00 176.7 265.1 1.577 .748 9.50 15.62 .511 987.1 .0 .0 88.00 192.1 270.2 1.512 .750 10.00 15.62 .515 987.0 .0 .0  88.00 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-48 (SHEET 1 OF 3)  DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS  POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      192.2 148.6 185.4 841.4 154.6 197.2 148.7 185.5 841.3 154.3 202.2 148.9 185.8 841.1 154.1 207.2 148.4 185.2 841.6 154.0 212.2 148.7 185.6 841.3 153.7 217.2 148.2 185.0 841.8 153.7 222.2 148.5 185.3 841.5 153.4 227.2 148.0 184.7 842.0 153.3 232.2 148.3 185.1 841.7 153.0 237.2 147.8 184.4 842.2 153.0 242.2 148.1 184.8 841.9 152.7 247.2 147.6 184.1 842.4 152.6 252.2 147.8 184.4 842.2 152.4 257.2 148.1 184.7 841.9 152.1 262.2 147.5 184.1 842.5 152.0 267.2 147.8 184.4 842.2 151.8 272.2 147.2 183.7 842.8 151.7 277.2 147.4 184.0 842.6 151.5 282.2 147.6 184.2 842.4 151.2 287.2 147.1 183.5 842.9 151.1 292.2 147.3 183.7 842.7 150.9 297.2 146.7 183.0 843.3 150.8 302.2 146.9 183.2 843.1 150.6 307.2 147.0 183.4 843.0 152.9 312.2 146.4 182.7 843.6 152.9 317.2 146.6 182.9 843.4 152.6 322.2 146.7 183.0 843.3 152.4 327.2 146.8 183.2 843.2 152.1 332.2 146.2 182.4 843.8 152.1 337.2 146.3 182.5 843.7 151.8 342.2 146.4 182.6 843.6 151.6 347.2 146.4 182.7 843.6 151.3 352.2 145.8 181.9 844.2 151.3 357.2 145.8 181.9 844.2 151.1 362.2 145.8 182.0 844.2 150.8 367.2 145.9 182.0 844.1 150.6 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-48 (SHEET 2 OF 3)
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      372.2 145.9 182.0 844.1 150.4 377.2 145.8 182.0 844.2 150.2 382.2 145.1 181.1 844.9 150.1 387.2 145.1 181.0 844.9 149.9 392.2 145.0 180.9 845.0 149.7 397.2 144.9 180.8 845.1 149.5 402.2 144.9 180.8 845.1 149.3 407.2 144.9 180.8 845.1 149.1 412.2 144.9 180.8 845.1 148.8 417.2 144.9 180.7 845.1 148.6 422.2 144.8 180.7 845.2 148.4 427.2 144.7 180.6 845.3 148.2 432.2 144.6 180.5 845.4 148.0 437.2 144.5 180.3 845.5 147.8 442.2 144.4 180.1 845.6 147.6 447.2 144.8 180.7 845.2 147.3 452.2 144.6 180.5 845.4 147.1 457.2 144.4 180.2 845.6 146.9 462.2 144.2 179.9 845.8 149.2 467.2 144.5 180.3 845.5 148.9 472.2 144.2 179.9 845.8 148.7 477.2 144.4 180.2 845.6 148.4 482.2 144.0 179.7 846.0 148.2 487.2 144.2 179.9 845.8 148.0 492.2 144.3 180.0 845.7 147.7 497.2 143.8 179.4 846.2 147.6 502.2 143.8 179.4 846.2 147.3 507.2 143.7 179.3 846.3 147.1 512.2 143.6 179.2 846.4 146.9 517.2 143.9 179.6 846.1 146.5 522.2 143.7 179.2 846.3 146.4 527.2 143.8 179.4 846.2 146.1 532.2 143.4 178.9 846.6 145.9 537.2 143.3 178.9 846.7 145.7 542.2 143.7 179.3 846.3 145.3 547.2 143.4 178.9 846.6 147.5 552.2 143.4 178.9 846.6 147.2 557.2 143.3 178.7 846.7 147.0 562.2 143.4 178.9 846.6 146.7 567.2 143.2 178.7 846.8 146.5 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-48 (SHEET 3 OF 3)
Break Path No. 1 Flow Break Path No. 2 Flow      Time  Thousand  Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)      572.2 143.2 178.7 846.8 146.2 577.2 142.8 178.2 847.2 146.0 582.2 142.8 178.2 847.2 145.7 587.2 143.0 178.4 847.0 145.4 592.2 142.8 178.2 847.2 145.2 597.2 142.6 177.9 847.4 145.0 602.2 142.5 177.8 847.5 144.7 607.2  69.9  87.2 920.2 164.2    792.8 69.9 87.2 920.2 164.2    792.9 75.0 92.6 915.0 161.7    797.2 74.9 92.5 915.1 161.4    1221.1 74.9 92.5 915.1 161.4    1221.2 67.7 77.9 922.3 87.5    1311.6 66.5 76.6 923.5 87.7    1311.7 66.5 76.6 68.4 12.5    1491.6 64.3 74.0 70.6 12.9    1491.7 64.3 74.0 1095.0 166.0    3600.0 52.0 59.8 1107.4 168.2    3600.1 41.2 47.4 1118.1 159.1  10000.0 30.0 34.5 1129.4 160.7 100000.0 16.0 18.4 1143.3 162.7 1000000.0 6.9 7.9 1152.5 164.0 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-49  DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MAXIMUM SAFEGUARDS                Time (Seconds) .00 21.60 21.60 192.12 792.85 1221.12 3600.00          Mass (Thousand lbm)        Initial In RCS and ACC 620.08 620.08 620.08 620.08 620.08 620.08 620.08        Added Mass Pumped Injection .00 .00 .00 157.40 752.04 1176.03 3734.26          Total Added .00 .00 .00 157.40 752.04 1176.03 3734.26        ***TOTAL AVAILABLE*** 620.08 620.08 620.08 777.48 1372.13 1796.12 4354.35        Distribution Reactor Coolant 416.79 47.97 67.66 118.91 118.91 118.91 118.91  Accummulator 203.30 156.65 136.96 .00 .00 .00 .00          Total Contents 620.08 204.61 204.61 118.91 118.91 118.91 118.91        Effluent Break Flow .00 415.46 415.46 649.72 1244.37 1668.36 4226.58  ECCS Spill .00 .00 .00 .00 .00 .00 .00          Total Effluent .00 415.46 415.46 649.72 1244.37 1668.36 4226.58        ***TOTAL ACCOUNTABLE*** 620.08 620.07 620.07 768.63 1363.28 1787.27 4345.49 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-50  DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE MAXIMUM SAFEGUARDS      Time (Seconds) .00 21.60 21.60 192.12 792.85 1221.12 3600.00          Energy (Million Btu) Initial Energy In RCS, ACC, S Gen 673.30 673.30 673.30 673.30 673.30 673.30 673.30        Added Energy Pumped Injection .00 .00 .00 13.85 66.18 103.49 461.34  Decay Heat .00 5.73 5.73 22.33 65.12 90.67 203.33  Heat From Secondary .00 -5.70 -5.70 -5.70 -3.44 -2.25 -2.25          Total Added .00 .03 .03 30.48 127.86 191.91 662.43        ***TOTAL AVAILABLE*** 673.30 673.34 673.34 703.79 801.17 865.22 1335.73        Distribution Reactor Coolant 244.82 10.46 12.22 31.25 31.25 31.25 31.25  Accummulator 18.20 14.02 12.26 .00 .00 .00 .00  Core Stored 18.93 9.68 9.68 4.05 3.90 3.72 2.71  Primary Metal 118.16 111.56 111.56 89.39 61.11 52.35 38.51  Secondary Metal 76.01 75.75 75.75 68.35 49.12 40.00 29.56  Steam Generator 197.20 196.34 196.34 173.69 121.93 99.57 73.76          Total Contents 673.30 417.81 417.81 366.73 267.30 226.88 175.78        Effluent Break Flow .00 255.05 255.05 328.89 525.69 621.47 1145.63  ECCS Spill .00 .00 .00 .00 .00 .00 .00          Total Effluent .00 255.05 255.05 328.89 525.69 621.47 1145.63        ***TOTAL ACCOUNTABLE*** 673.30 672.86 672.86 695.62 792.99 848.35 1321.40 FNP-FSAR-6 REV 21  5/08 TABLE 6.2-51  DOUBLE-ENDED HOT LEG BREAK SEQUENCE OF EVENTS 
Time (sec) Event Description  0.0 Break Occurs, Reactor Trip and Loss of Offsite  Power are assumed  3.0 Low Pressurizer Pressure SI Setpoint -  1715 psia reached by SATAN    11.4 Broken Loop Accumulator Begins Injecting Water    11.6 Intact Loop Accumulator Begins Injecting Water  20.0 End of Blowdown Phase 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-52  DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS SEQUENCE OF EVENTS 
Time (sec) Event Description  0.0 Break Occurs, Reactor Trip and Loss of Offsite  Power are assumed  3.9 Low Pressurizer Pressure SI Setpoint -  1715 psia reached by SATAN    13.4 Broken Loop Accumulator Begins Injecting Water    13.6 Intact Loop Accumulator Begins Injecting Water    21.6 End of Blowdown Phase    30.9 Safety Injection Begins    52.1 Broken Loop Accumulator Water Injection Ends    53.3 Intact Loop Accumulator Water Injection Ends    187.7 End of Reflood Phase  2139.0 Cold Leg Recirculation Begins  1.0E+06 Transient Modeling Terminated 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-53  DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS SEQUENCE OF EVENTS 
Time (sec) Event Description  0.0 Break Occurs, Reactor Trip and Loss of Offsite  Power are assumed  3.9 Low Pressurizer Pressure SI Setpoint -  1715 psia reached by SATAN    13.3 Broken Loop Accumulator Begins Injecting Water    13.5 Intact Loop Accumulator Begins Injecting Water    21.6 End of Blowdown Phase    30.9 Safety Injection Begins    54.7 Broken Loop Accumulator Water Injection Ends    54.9 Intact Loop Accumulator Water Injection Ends    192.1 End of Reflood Phase  1311.6 Cold Leg Recirculation Begins  1.0E+06 Transient Modeling Terminated 
FNP-FSAR-6 REV 21  5/08 TABLE 6.2-54  LOCA MASS AND ENERGY RELEASE ANALYSIS CORE DECAY HEAT FRACTION Time (sec) Decay Heat Generation Rate (Btu/Btu)    10 0.053876 15 0.050401 20 0.048018 40 0.042401 60 0.039244 80 0.037065 100 0.035466 150 0.032724 200 0.030936 400 0.027078 600 0.024931 800 0.023389 1000 0.022156 1500 0.019921 2000 0.018315 4000 0.014781 6000 0.013040 8000 0.012000 10000 0.011262 15000 0.010097 20000 0.009350 40000 0.007778 60000 0.006958 80000 0.006424 100000 0.006021 150000 0.005323 200000 0.004847 400000 0.003770 600000 0.003201 800000 0.002834 1000000 0.002580 
REV 21  5/08 DEPSGB MINIMUM ESF 1 AC PRESSURE VS. TIME PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-1 REV 21  5/08 RSG DEPSG MINIMUM ESF 1 AC PRESSURE VS. TIME, PO = 3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-2 REV 21  5/08 DEHL MINIMUM ESF, DBA SHORT TERM PRESSURE VS. TIME, PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-3 REV 21  5/08 RSG DEHLG MINIMUM ESF, DBA SHORT PRESSURE VS. TIME, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-4 REV 21  5/08 DECLG MAXIMUM ESF PRESSURE VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-5 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-6 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-6A REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-7 REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-7A REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.7 ft2 D.E. BREAK 102% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-8 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.7 ft2 D.E. BREAK 102% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-9 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-10 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-10A REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = - PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-11 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2  D.E. BREAK 102% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-11A REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.528 ft2 SPLIT 102% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-12 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.528 ft2 SPLIT 102% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-13 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-14 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-15 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-16 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-17 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.5 ft2  D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-18 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 70% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-19 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.47 ft2 SPLIT 70% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-20 REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.47 ft2 SPLIT 70% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-21 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-22 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-22A REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-23 REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO =  +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-23A REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 30% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-24 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 30% POWER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-25 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER,PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-26 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-26A REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-27 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-27A REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-28 REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-28A REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = -1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-29 REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-29A REV 21  5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK HOT STANDBY PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-30 REV 21  5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK HOT STANDBY, PO =-1.5 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-31 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.2 ft2 D.E. BREAK HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-32 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.2 ft2 D.E. BREAK HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-33 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.1 ft2 D.E. BREAK HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-34 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.1 ft2 D.E. BREAK HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-35 REV 21  5/08 PRESSURE VERSUS TIME STEAM LINE 0.30 ft2 SPLIT HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-36 REV 21  5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.30 ft2 SPLIT HOT STANDBY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-37 REV 21  5/08 TS, EQUIPMENT SURFACE TEMPERATURE WITH UCHIDA CONDENSING HEAT TRANSFER AND CONVECTIVE HEAT TRANSFER COEFFICIENT OF 2 BTU/HR-ftq  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-38 REV 21  5/08 DEPSGB MINIMUM ESF 1 AC P/T ANALYSIS LONG-TERM CONTAINMENT PRESSURE VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-39 REV 21  5/08 DEPSGB MINIMUM ESF DBA TEMPERATURE VS. TIME PO = 0 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-40 REV 21  5/08 RSG DEPSG MIN ESF DBA TEMPERATURE VS. TIME PO = 3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-41 
REV 21  5/08 CONTAINMENT AIR COOLER DUTY VS. TEMPERATURE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-42 (SHEET 1 OF 2)
REV 21  5/08 CONTAINMENT AIR COOLER DUTY VS. TEMPERATURE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-42 (SHEET 2 OF 2) 
REV 21  5/08 THERMAL HEAT REMOVAL EFFICIENCY OF CONTAINMENT ATMOSPHERE SPRAY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-43 REV 21  5/08 RESIDUAL HEAT EXCHANGER DESIGN DUTY ACCIDENT MODE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-44 REV 21  5/08 MASS &N ENERGY RATE VS TIME FOR DBA  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-45 REV 21  5/08 LOCA BLOWDOWN MASS AND ENERGY RELEASE RATES VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-46 REV 21  5/08 LOCA POST-BLOWDOWN MASS AND ENERGY RELEASE RATES VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-47 REV 21  5/08 DEPSG MIN ESF 1 AD P/T ANALYSIS LONG-TERM CONDENSING HEAT TRANSFER COEFFICIENT (RSG)  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-48 
REV 21  5/08 SHORT TERM CONDENSING HEAT TRANSFER COEFFICIENT FOR DBA  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-49 REV 21  5/08 REACTOR CAVITY MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-50 REV 21  5/08 REACTOR CAVITY BLOCK DIAGRAM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-51 REV 21  5/08 TOTAL HORIZONTAL FORCE VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-52 
REV 21  5/08 STEAM GENERATOR BLOCK DIAGRAM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-53 REV 21  5/08 STEAM GENERATOR COMPARTMENT C DIFFERENTIAL PRESSURE VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-54 REV 21  5/08 PRESSURIZER COMPARTMENT PRESSURE MODEL (SPRAY LINE BREAK IN LOWER COMPARTMENT)  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-55 
REV 21  5/08 PRESSURIZER COMPARTMENT FLOW MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-56 REV 21  5/08 PRESSURIZER COMPARTMENT SPRAY LINE RESULTS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-57 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 1 AND 2 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-58 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 3, 4, 5, AND 6 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-59 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 7, 8, 9, AND 10 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-60 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 11, 12, 13, AND 14 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-61 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 15, 16, AND 17 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-62 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 18, 19, 20, AND 21 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-63 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 22, 23, 24, 25, 26, AND 27 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-64 REV 21  5/08 NODE PRESSURES IN COMPARTMENTS 28, 29, 30, 31, 32, 33, AND 34 VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-65 REV 21  5/08 SCHEMATIC OF REFLOOD CODE 19 ELEMENT LOOP MODEL FOR A PUMP SUCTION BREAK  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-66 REV 21  5/08 CORE REFLOOD CORRELATION  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-67 
THIS FIGURE HAS BEEN DELETED PER REV 15. 
REV 21  5/08 COMPARISON OF MEASURE AND PREDICTED CARRY OVER RATE FRACTIONS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-68 
THIS FIGURE HAS BEEN DELETED PER REV 15. 
REV 21  5/08 INLET WATER TEMPERATURE VS. TIME AFTER END OF BLOWDOWN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-69 REV 21  5/08 VARIATION IN TEMPERATURE RISE, TURNAROUND TIME AND QUENCH TIME WITH RESPECT TO CORE ELEVATION  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-70 REV 21  5/08 ENERGY BALANCE MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-71 REV 21  5/08 REFLOOD RATE AND CARRYOVER FRACTIONS VS. TIME AFTER END OF BLOWDOWN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-72 REV 21  5/08 FLOW THROUGH BREAK VS. TIME AFTER END OF BLOWDOWN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-73 REV 21  5/08 WATER HEIGHT VS. TIME AFTER END OF BLOWDOWN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-74 REV 21  5/08 POST-REFLOOD LOOP RESISTANCE MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-75 REV 21  5/08 S/G INTERNAL ENERGY VS. TIME AFTER BREAK  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-76 REV 21  5/08 ENERGY DISTRIBUTION VS. TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-77 REV 21  5/08 RSG TEMPERATURE PROFILE THROUGH CONTAINMENT WALL PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-78 REV 21  5/08 RHR HX DUTY VS. TIME RSG, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-79 REV 21  5/08 CONTAINMENT AIR COOLING DUTY VS. TIME RSG, PO = +3 PSIG  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-80 REV 21  5/08 MINIMUM SUMP pH FOLLOWING LOCA VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-81 REV 21  5/08 MINIMUM PARTITION COEFFICIENT IN THE SUMP VERSUS SOLUTION pH  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-82 REV 21  5/08 HYDROGEN GENERATION RATE VS. TIME IN  THE LOWER COMPARTMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-83 REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-84 REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-85 REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-86 
Note 1: Containment isolation is provided by the blind flange inside containment. Valve outside  containment for arrangement 26 is shown for completeness only and is not a containment  isolation valve. Note 2: Relief valve shown outside containment for arrangement 24 is applicable to penetration  46 only. The relief valve is classified as a containment isolation valve. REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-87 REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-88 REV 21  5/08 ISOLATION VALVE ARRANGEMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-89 REV 21  5/08 ELECTRIC HYDROGEN RECOMBINER  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-90 REV 21  5/08 ELECTRIC HYDROGEN RECOMBINER  SCHEMATIC DIAGRAM (TYPICAL OF ONE RECOMBINER)  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-91 REV 21  5/08 LOWER COMPARTMENT PLAN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-92 REV 21  5/08 SECTION OF LOWER REACTOR COMPARTMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-93 REV 21  5/08 CONTAINMENT HYDROGEN CONCENTRATION WITH ONE ELECTRIC RECOMBINER STARTED ONE DAY AFTER A LOCA  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-94 REV 21  5/08 HYDROGEN CONCENTRATION AS A FUNCTION OF  TIME IN CONTAINMENT PURGE MODE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-95 REV 21  5/08 VOLUME PERCENT HYDROGEN VS. TIME IN THE UPPER CONTAINMENT (UNMIXED), OUTER PERIPHERY (UNMIXED) AND BULK CONTAINMENT (MIXED)  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-96 REV 21  5/08 VOLUME PERCENT HYDROGEN VS. TIME IN  THE LOWER COMPARTMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-97 
REV 21  5/08 HYDROGEN GENERATION RATE VS. TIME IN  OUTER PERIPHERY AND OVERALL CONTAINMENT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-98 
FNP-FSAR-6    REV 21  5/08  TABLE 6.3-2  ECCS RELIEF VALVE DATA Fluid Inlet  Set Back Psig  Fluid Temp.  &deg;F Pressure Pressure Build-  Description Discharged  Normal    Relieving  (psig)  Constant  Up    Capacity        N2 supply to N2 gas 120 120 700 atm. 0 1500 sf3/min accumulators                                    RHR pumps Water 250 350 600 3 50 20 gal/min discharge        SI line                Accumulator to N2 gas 120 120 700 0 0 1500 sf3/min containment           
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-3  SEQUENCE OF CHANGEOVER OPERATION FROM INJECTION TO RECIRCULATION 
(This table has been deleted.) 
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-4 (SHEET 1 OF 2)  TIME ANALYSIS FOR ECCS INJECTION/RECIRCULATION SWITCHOVER Time Step (s) Volume  Flow Rate From for Constant Remaining In  RWST (gpm) RWST Flow RWST Step  Time (s) During Step    Rate      (gal)          1. Low-level switchover setpoint 0  0 135,716  2. Verify SI reset 10 13,400    3. Direct verification of PRF status 20 13,400    4. Verify CCW flow to RHR heat exchangers  60 13,400    5. Establish only one charging pump in each train 70 13,400    6. Direct verification of recirculation disconnects 80 13,400    7. Stop both RHR pumps 90 13,400 90 115,616  8. Close RWST supply to 'A' RHR pump suction 110 9,000    9. Align containment sump to 'A' RHR pump suction 150 9,000  10. Close RHR to RCS hot legs cross-connect  170 9,000  11. Start 'A' RHR pump  180 9,000  12. Verify 'A' Train LHSI flow 185 9,000  13. Close RWST supply to 'B' RHR pump suction 205 9,000 115 98,366 14. Align containment sump to 'B' RHR pump 245 7,600  15. Close RHR to RCS hot legs cross-connect 265 7,600 FNP-FSAR-6 REV 21  5/08  TABLE 6.3-4 (SHEET 2 OF 2)
Time Step (s) Volume  Flow Rate From for Constant Remaining In  RWST (gpm) RWST Flow RWST Step (continued)  Time (s) During Step    Rate      (gal)        16. Start 'B' RHR pump  275 7,600    17. Verify 'B' Train LHSI flow 280 7,600    18. IF 'A' RHR pump started, THEN align charging pump suction  header isolation valves based on 'B' charging pump status  360 7,600    19. Open RHR supply to 'A' train charging pump suction 380 7,600 175 76,200  20. Verify VCT level 385 6,700 5 75,642  21. Close 'A' train RWST to charging pump header valve 405 7,150    22. IF 'B' RHR pump started, THEN align charging pump suction header isolation valves based on 'B' charging pump status 410 7,150    23. Open RHR supply to 'B' train charging pump suction 430 7,150 45 70,280  24. Verify VCT level 435 6,700    25. Close 'B' train RWST to charging pump header valve 455 6,700    26. Check one charging pump in each train 460 6,700    27. Open charging pump recirculation to RCS cold legs valve 480 6,700    28. Align charging pump discharge header isolation valves based on 'B' charging pump status 560 6,700    29. Verify SI flow 565 6,700 135 55,207   
FNP-FSAR-6  REV 21  5/08  TABLE 6.3-6  NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING 
Number of charging pumps operable 2  Number of residual heat removal pumps operable 2  Number of residual heat exchangers operable 2  Minimum refueling water storage tank volume (gal) 471,000  Boron concentration in refueling water storage 2,300 to tanks (ppm) 2,500    Boron concentration in accumulator (ppm) 2,200 to  2,500  Number of accumulators 3  Normal operating accumulator pressure (psig) band 601 to 649  Nominal accumulator water volume (ft3) 1025(a) 
a. This value includes the liquid volume in the tank plus the liquid volume in the piping measured from the tank to the second check valve. The second check valve is defined as the second check valve from the tank or the first check valve from the reactor coolant system (RCS) loop.
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-7 (SHEET 1 OF 2)  SINGLE ACTIVE FAILURE ANALYSIS FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS SHORT TERM PHASE  Component Malfunction Comments    Accumulator Deliver to broken loop Totally passive system with one accumulator per loop. Evaluation based on one spilling  accumulator    Pump      Centrifugal charging Fails to start  Three provided. Evaluation based on    operation of one    Residual heat removal Fails to start Two provided. Evaluation based on operation    of one    Automatically Operated Valves              Injection line isolation Fails to open Two parallel lines; one valve in either line    required to open              Residual heat removal pumps Fails to close Check valve in series with one gate valve;  suction line to refueling  operation of only one valve required  water storage tank      Centrifugal charging pumps      a. Suction line to refueling Fails to open Two parallel lines; only one valve in either  water storage tank  line is required to open    b. Discharge line to the Fails to close Two valves in series; only one valve required  normal charging path  to close    c. Miniflow line Fails to close Two valves in series; only one valve required    to close  d. Suction from volume Fails to close Two valves in series; only one valve required  control tank  to close FNP-FSAR-6 REV 21  5/08  TABLE 6.3-7 (SHEET 2 OF 2)  LONG TERM PHASE  Component Malfunction Comments    Valves operated from control  room for recirculation      Containment sump recirculation isolation Fails to open Two lines parallel; two valves in either    lines are required to open    Residual heat removal pumps Fails to close Check valve in series with one gate valve;  suction line to refueling water storage tank  operation of either the check or the gate    valve required    Centrifugal charging pump suction line to refueling Fails to close Check valve in series with two parallel  water storage tank  gate valves. Operation of either the check    valve or the gate valves required    Centrifugal charging pump suction line at discharge Fails to open Separate and independent high head injection  of residual heat exchanger  path taking suction from discharge of the    other residual heat exchanger    Pumps      Residual heat removal pump Fails to start Two provided. Evaluation based on operation    of one    Centrifugal charging pump Fails to operate Same as short term phase                Failure of Train B power during switchover from cold leg recirculation  to simultaneous hot and cold leg recirculation results in:
* Residual heat removal discharge valve to hot legs (MOV 8889) Fails to open Align RHR pumps to cold legs, Train A charging  pump to hot legs, and use Train A/Train B Power
* Centrifugal charging pump discharge valve to cold legs (MOV 8803B) Fails to close Transfer Switch (Q1/2R18B037) to apply Train A  power to close MOV 8803B 
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-8  MAXIMUM POTENTIAL RECIRCULATION LOOP LEAKAGE EXTERNAL TO CONTAINMENT    Leakage to Leakage to  Type of Leakage Control and Unit Atmosphere Drain Tank Item Leakage Rate Used in the Analysis (cm3/h) (cm3/h)    Residual heat removal Mechanical seal with leakoff - 0 20  (low head safety injection) 10 cc/hr/seal      Charging pumps Same as residual heat removal 0 60  pump(a)        Flanges:        Pumps Gasket - adjusted to zero 0 0  leakage following any test    Valves bonnet to body 10 drops/min/gauge used 2400 0  (larger than 2 in.) (30 cc/hr). Due to leak      tight flanges on pumps, no    Control valves leakage is assumed to 480 0  atmosphere    Heat exchangers  240 0 Valves - stem leakoffs Back seated double packing 0 50  with leakoff - 1 cc/hr in. stem diameter used.  (See    table 6.3-1.)      Miscellaneous small valves Flanged body packed stems - 600 0  1 drop/min used (3 cm3/h).      Miscellaneous large valves Double packing 1 cm3/h/in. 40 0  (larger than 2 in.) stem diameter used      ________________                                a. Seals are acceptance tested to essentially zero leakage. Due to tandem double seal arrangement and the use of water from the refueling water storage tank as a buffer between the seals, no radioactive leakage from the pumps to the atmosphere is expected.
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-9  EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS  Flow Path Indication of Loss of Flow Path Alternate Flow Path    Low head recirculation  During cold-leg recirculation:    From containment sump to low Accumulation of water in a Via the independent, identical  head injection header via the residual heat removal pump low head flow path utilizing the  residual heat removal pumps compartment or auxiliary second residual heat exchanger  and the residual heat building sump  exchangers High-head recirculation  During hot-leg recirculation:      The high head pumps provide the  required redundancy during this  period    From containment sump to the Accumulation of water in a From containment sump to the  high head injection header residual heat removal pump high head injection headers via  via residual heat removal compartment or the auxiliary alternate residual heat removal  pump, residual heat building sump pump, residual heat exchanger  exchanger and the high head  and the alternate high head  injection pumps  charging pump 
FNP-FSAR-6 REV 21  5/08  TABLE 6.3-10  EMERGENCY CORE COOLING SYSTEM SHARED FUNCTIONS EVALUATION    Normal Operating  Component  Arrangement    Accident Arrangement Refueling water storage Lined up to suction of Lined up to suction of  tank residual heat removal pumps centrifugal charging and    residual heat removal pumps. Valves for realignment of    RWST to charging pumps meet    the single failure criteria Centrifugal charging Lined up for charging service Lined up to high head safety  pumps  injection header. Valves for    realignment meet single    failure criteria    Residual heat removal Lined up to cold legs of Lined up to cold legs of  pumps reactor coolant piping reactor coolant piping    Residual heat exchangers Lined up for residual heat Lined up for residual heat    removal pump operation removal pump operation   
REV 21  5/08 RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-1 REV 21  5/08 CHARGING PUMP PERFORMANCE CURVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-2 REV 21  5/08 TYPICAL RHR PUMP CHARACTERISTIC CURVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-3 REV 21  5/08 CONTAINMENT SPRAY PUMP CHARACTERISTIC CURVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-4 
HISTORICAL] [The following describes pre-operational testing requirements:    1. Tray Type
a. Removal of all iodines with an efficiency of 95.0 percent for a 12 hour continuous flow at 150&deg;F, 70 percent relative humidity, and 40 ft/min face velocity.
b. Each adsorbing unit (2 elements) is capable of filtering 333 ft3/min of air at a pressure drop not exceeding 1.2 in. wg.
c. Each assembled filter unit will be leak tested by the manufacturer. 
d. Each filter will be tested for 5 minutes in airflow of 330 ft3/min containing 20 ppm refrigerant 112. A downstream concentration in excess of 0.2 percent of the upstream concentration will cause rejection of the filter.
2. Bed Type
a. Retaining 99.0 percent minimum of elemental iodine. At relative humidities below 70 percent at 150&deg;F all organic iodines with an 
efficiency of 95 percent for the recirculation filter and 99 percent for the pressurization filter.
b. Each assembled filter unit will be leak tested by the manufacturer.
c. Charcoal adsorbers remove  99 percent of a halogenated hydrocarbon refrigerant test gas.]
FNP-FSAR-6 REV 25  4/14 TABLE 6.5-1  AUXILIARY FEEDWATER SYSTEM AUXILIARY FEEDWATER PUMP DATA 
MOTOR-DRIVEN PUMPS TURBINE-DRIVEN  (DATA PER PUMP) PUMP    Type Horizontal- Horizontal-  Centrifugal Centrifugal    No. of stages 10 7 Design pressure (psig) 1600 1600 Pumping temperature (&deg;F) 95 95 Design flowrate (gal/min) 350 700 Design head (ft) 2845 2835 NPSH required at design (ft) 17 21 Minimum available NPSH (ft) 60 60 Suction pressure range (ft) 45-75 45-75 Shutoff head (ft) 3480 3380 RPM 3600 3960 Bhp required 366 687 Driver horsepower (max) 450 693 Materials:    Casing  SA-217 Gr. WC 9  SA-217 Gr. WE 9  Impeller  SA-296 Gr. CA 15  SA-296 Gr. CA 15  or A-217 Gr. CA 15 or A-217 Gr. CA 15(a)  Shaft  A-276 Tp 410 HT  SA-276 Tp 410 HT  TURBINE DRIVE Type Vertical-Single Stage  Design pressure (psig) 1250 Design temperature (&deg;F) 572 Steam inlet pressure (psig)    Minimum 90  Maximum 1148 Back pressure (psig) 0-10 RPM design/turbine trip 3960/4554 Rated Bhp 687 Governor NEMA Class D Lubrication Forced feed Cooling water Pumped liquid a. Changes are applicable to Unit 2 only.
FNP-FSAR-6 REV 21  5/08 TABLE 6.5-2 (SHEET 1 OF 4)  FAILURE ANALYSIS OF AUXILIARY FEEDWATER SYSTEM  Component Failure                  Comments and Consequences    Motor-driven auxiliary Fails to start on Two motor-driven pumps are provided. One motor-driven pump  feedwater pump automatic signal in conjunction with the turbine-driven pump is sufficient to    meet cooldown requirements for all emergency conditions. One motor-driven pump is sufficient to meet all normal    cooldown requirements Turbine-driven auxiliary Fails to start on Operation of the two motor-driven pumps will provide  feedwater pump automatic signal sufficient flow to meet cooldown requirements for all condi-    tions Turbine-driven pump steam Fails to open on Black- Parallel connections are provided to two main steam lines. inlet isolation valve from out signal One of the two valves must open to supply 100 percent of  main steam header  the turbine steam requirements Steam supply lines to One parallel supply line Check valves installed in each parallel line, upstream of  turbine driven pump broken downstream of inlet the common header connection and below the floor of the    isolation valve in main main steam and feedwater valve room, prevent blowdown through    steam and feedwater the broken line and subsequent loss of steam supply to the    valve room turbine drive Condensate supply Loss of normal supply Water can be supplied to all pumps from the service water    from condensate storage system. Service water supply is separate and redundant    tank Auxiliary feedwater pump Failure of pressure No single failure can prevent the auxiliary feedwater system  discharge line boundary resulting in from providing the minimum required flow. Both manual and    abnormal leakage motor-operated valves are provided for isolating potential    breaks Electrical power supply Failure of power supply Motor-driven pumps are separate and redundant including    bus to components power supplies. One motor-driven pump in conjunction with    associated with one motor the turbine-driven pump will supply the minimum required    driven pump flow for all emergency conditions Motor operated valves in Loss of power All motor operated valves are manually open, fail "as is"  pump discharge piping  on loss of power, and are closed remote manually Air operated flow control Loss of air or loss of Failure modes presented in table 7.3-10, sheet 2  valves in pump discharge 125-V dc power FNP-FSAR-6 REV 21  5/08 TABLE 6.5-2 (SHEET 2 OF 4)
Component Failure Comments and Consequences    Isolation valves Spurious closure of During normal plant operation, these valves are in the  MOV 3350A, B, C motor-operated valve open position and the breakers, which supply power to the    valves, are opened and locked so that no power is supplied to the    valve's motor operator.
Main feedwater line between Failure of a main feedwater Case 1 -  Failure of main feedwater line to steam generator  the containment isolation line with a simultaneous 1A.--The Train B motor-driven pump and the turbine-driven  valve and the steam generator loss of Train A electrical pump start and delivery flow through the restriction orifices  power which limit auxiliary feedwater flow to the faulted steam  generator, thus establishing the minimum required flow to  two intact steam generators. Closing valve MOV 3764E, which  is powered from a Train B electrical power supply, isolates  auxiliary feedwater flow from the motor-driven pump to the  faulted steam generator. This increases flow to the two  intact steam generators, allowing an orderly cooldown to  the cold shutdown condition Case 2 - Failure of main feedwater line to steam generator  1B.--Identical to Case 1 above except that valve MOV 3764B,  which is powered from a Train B electrical power supply,  is closed to isolate auxiliary feedwater flow from the  motor-driven pump to the faulted steam generator Case 3 - Failure of main feedwater line to steam generator  1C.--Identical to Case 1 above except that valve MOV 3764C,  which is powered from a Train B electrical power supply, is  closed to isolate auxiliary feedwater flow from the motor-  driven pump to the faulted steam generator Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A.--The Train A motor-driven pump and the turbine-driven valve and the steam generator loss of Train B electrical pump start and deliver flow through the restriction orifices  power which limit auxiliary feedwater flow to the faulted steam  generator, thus establishing the minimum required flow to  the two intact steam generators. Closing valve MOV 3764A,  which is powered from a Train A electrical power supply,  isolates auxiliary feedwater flow from the motor-driven  pump to the faulted steam generator. This increases flow  to the two intact steam generators, allowing an orderly  cooldown to the cold shutdown condition.
FNP-FSAR-6 REV 21  5/08 TABLE 6.5-2 (SHEET 3 OF 4)
Component Failure Comments and Consequences Case 2 - Failure of main feedwater line to steam generator  1B.--Identical to Case 1 above except that valve MOV 3764D,  which is powered from a Train A electrical power supply,  is closed to isolate auxiliary feedwater flow from the  motor-driven pump to the faulted steam generator.
Case 3 - Failure of main feedwater line to steam generator  1C.--Identical to Case 1 above except that valve MOV 3764F,  which is powered from a Train A electrical power supply,  is closed to isolate auxiliary feedwater flow from the  motor-driven pump to the faulted steam generator.
Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A.--The Train A and Train B motor-driven auxiliary valve and the steam generator loss of the turbine- feedwater pumps start and deliver flow through the  driven auxiliary feedwater pump restriction orifices which limit auxiliary feedwater flow    to the faulted steam generator, thus establishing the    the minimum required flow to the two intact steam    generators. Closing either valve MOV 3764A or MOV 3764E    isolates motor-driven auxiliary feedwater pump flow to the    faulted steam generator. This increases flow to the two    intact steam generators, allowing an orderly cooldown to    the cold shutdown condition.
Case 2 - Failure of main feedwater line to steam generator    1B.--Identical to Case 1 above except that either valve    MOV 3764B or MOV 3764D is closed to isolate auxiliary    feedwater flow to the faulted steam generator.
Case 3 - Failure of main feedwater line to steam generator    1C.--Identical to Case 1 above except that either valve    MOV 3764C or MOV 3764F is closed to isolate auxiliary    feedwater flow to the faulted steam generator.
FNP-FSAR-6 REV 21  5/08 TABLE 6.5-2 (SHEET 4 OF 4)
Component Failure Comments and Consequences    Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator  the containment isolation line with a simultaneous 1A with a simultaneous spurious closure of either motor-  valve and the steam generator spurious closure of a operated valve located in the motor-driven pump discharge    motor operated valve in line to steam generator 1B.--The Train A and Train B motor-    the pump discharge flow driven pumps and the turbine-driven pump start and deliver    path flow through the restriction orifices. The restriction    orifices limit flow to the faulted steam generator, thus    establishing the minimum required flow to the two intact    steam generators. Closing either valve MOV 3764A or    MOV 3764E isolates motor-driven pump flow to the faulted    steam generator and increases motor-driven pump flow    through the open flow path to steam generator 1C, thus    allowing an orderly cooldown to the cold shutdown    condition.      Case 2 - Failure of main feedwater line to steam generator    1A with a simultaneous spurious closure of either motor    operated valve located in the motor-driven pump discharge    line to steam generator 1C.--Identical to Case 1 above    except isolation of the faulted steam generator increases    motor-driven pump flow to steam generator 1B.
Note - For all possible combinations of a faulted steam    generator and a spurious closure of any one of valves    MOV 3764A, B, C, D, E or F, the operator can remote    manually isolate the motor-driven pump flow to the faulted    steam generator, which increases motor-driven pump flow    through the open flow path(s) to the intact steam generators,    thus allowing an orderly cooldown to the cold shutdown    condition.
FNP-FSAR-6 REV 21  5/08 TABLE 6.5-3  AUXILIARY FEEDWATER SYSTEM MOTOR OPERATED VALVE DATA  Motor Control Center  Valve Motor Supplying Electricity  Number to Valve Valve Position (Ref. drawing D-175007) (Ref. drawing D-177001) After Loss of Power    MOV 3209A MCC 1U As is MOV 3209B MCC 1V As is MOV 3210A MCC 1U As is MOV 3210B MCC 1V As is MOV 3216 MCC 1U As is MOV 3350A MCC 1U As is MOV 3350B MCC 1U As is MOV 3350C MCC 1U As is MOV 3764A MCC 1U As is MOV 3764B MCC 1V As is MOV 3764C MCC 1V As is MOV 3764D MCC 1U As is MOV 3764E MCC 1V As is MOV 3764F MCC 1U As is 
FNP-FSAR-6A   
6A-i REV 23  5/11 APPENDIX 6A  MATERIALS COMPATIBILITY REVIEW  TABLE OF CONTENTS          Page  6A.1 DEFINITION OF POSTACCIDENT CONTAINMENT ENVIRONMENTAL CONDITIONS .............................................................................. 6A-1 6A.1.1 Design Basis Accident Temperature-Pressure Cycle...................................... 6A-1
6A.1.2 Design Bases Accident Radiation Environment .............................................. 6A-2
6A.1.3 Design Chemical Composition of the Emergency Core Cooling Solution ....... 6A-2
6A.1.4 Trace Composition of Emergency Core Cooling Solution ............................... 6A-3
6A.2 MATERIALS OF CONSTRUCTION IN CONTAINMENT ............................................ 6A-3 6A.3 CORROSION OF METALS OF CONSTRUCTION IN DESIGN BASIS ECC SOLUTION .............................................................................................. 6A-4
6A.4 CORROSION OF METALS OF CONSTRUCTION BY TRACE CONTAMINANTS IN ECC SOLUTION ....................................................................... 6A-6 6A.4.1 Low Temperature of ECC Solution .................................................................. 6A-6
6A.4.2 Low Chloride Concentration of ECC Solution.................................................. 6A-6
6A.5 CORROSION OF ALUMINUM ALLOYS ..................................................................... 6A-7 6A.6 THE NATURE AND BEHAVIOR OF ALUMINUM CORROSION PRODUCTS IN ALKALINE SOLUTION ...................................................................... 6A-7 6A.6.1 Behavior of Circulating Aluminum Corrosion Products .................................... 6A-9 6A.7 EFFECT OF POSSIBLE CHEMICAL REACTIONS ON IODINE REMOVAL CAPABILITY OF THE CONTAINMENT SPRAY SOLUTION ................. 6A-10
6A.8 COMPATIBILITY OF PROTECTIVE COATINGS WITH POSTACCIDENT ENVIRONMENT ........................................................................................................ 6A-11
6A-9 EVALUATION OF THE COMPATIBILITY OF CONCRETE ECC SOLUTION IN THE POSTACCIDENT ENVIRONMENT .............................................................. 6A-11 FNP-FSAR-6A   
6A-ii REV 23  5/11 LIST OF TABLES  6A-1 Postaccident Containment Temperature Transient Used in the Material Compatibility Review  6A-2 Review of Sources of Various Elements in Containment and their Effects on Materials of Construction 6A-3 Typical Materials of Construction in the Farley Containment
6A-4 Deleted
6A-5 Corrosion of Aluminum Alloys in Alkaline Sodium Borate Solution
6A-6 Deleted
6A-7 Summary of Aluminum Corrosion Product Solubility Data
6A-8 Concrete Specimen Test Data
FNP-FSAR-6A   
6A-1 REV 23  5/11 APPENDIX 6A  MATERIALS COMPATIBILITY REVIEW  6A.1 DEFINITION OF POSTACCIDENT CONTAINMENT ENVIRONMENTAL  CONDITIONS  An evaluation of the suitability of materials of construction for use in the containment has been performed considering the following: 
A. The integrity of the materials of construction of engineered safety features equipment when exposed to postdesign basis accident (DBA) conditions.
B. The effects of corrosion and deterioration products from both engineered safety features (vital equipment) and other (nonvital) equipment, on the integrity and operability of the engineered safety features equipment.
The post DBA environment conditions of temperature, pressure, radiation, and chemical composition are described in the following sections. The time temperature pressure cycle used in the materials evaluation is most conservative, since it considers only partial safeguards operation during the DBA. The spray and core cooling solutions considered herein include both the design chemical compositions and the design chemical compositions contaminated with deterioration products and fission products, which may conceivably be transferred to the solution during recirculation through the various containment safety features systems. 
6A.1.1  DESIGN BASIS ACCIDENT TEMPERATURE-PRESSURE CYCLE  Containment pressure/temperature versus time responses for the various analyzed breaks are shown in figures 6.2-1 through 6.2-41. These figures represent containment environment conditions during and after a postulated accident considering partial safety features operation:
that is, operation with 1 of the 2 spray pumps, 1 of the 4 containment fans, 1 of the 2 residual heat removal pumps, and 1 of the 3 safety injection pumps. 
Table 6A-1 presents the evaluation conditions for Westinghouse supplied material subjected to the containment and the core environment, respectively. For equipment specified by Bechtel and Southern Company Services, Inc., refer to table 3.11-1. 
Material evaluations, to be described, were performed, in general, for the time temperature conditions of table 6A-1 or conservatively considering high temperature conditions for longer periods. The basis for each material evaluation is described with the discussion of its particular suitability. 
FNP-FSAR-6A   
6A-2 REV 23  5/11 6A.1.2  DESIGN BASIS ACCIDENT RADIATION ENVIRONMENT  Evaluation of materials for use in containment included a consideration of the radiation stability requirements for the particular materials application. This evaluation utilized data that were calculated on the basis of a core meltdown and, assuming the following fission product fractional releases, consistent with TID 14844 model:
Noble gases Fractional release 1.0 Halogens Fractional release 0.5 Other isotopes Fractional release 0.01 6A.1.3  DESIGN CHEMICAL COMPOSITION OF THE EMERGENCY  CORE COOLING SOLUTION Farley system designs provide for use of alkaline adjusted boric acid solution as the spray and core cooling fluid. 
Alkaline Sodium Borate Plant designs that utilize the spray solution for fission product iodine removal, as well as containment cooling include provisions for chemical addition to control pH. For Farley trisodium phosphate (TSP) is added to the containment sump. Boric acid solution, containing 2300 to 2500 ppm boron, is pumped from the refueling water storage tank into the core and to the containment by means of the safety injection system pumps, residual heat removal pumps, and spray pumps. The initial pH of the spilled RCS water and containment spray will  be approximately 4.5. Three baskets are located on elevation 105'-6" which contain sufficient TSP so that when their contents dissolve in the water from the RWST, RCS, and accumulators, the resulting containment sump and recirculation (ECCS and spray) systems pH will be between 7.5 and 9.1.
For the purpose of materials evaluation in the design chemistry solution, the following concentration/time relationship was considered:
0  8 hours pH 4.5 Boron 2500 ppm 8 hours 12 months pH 10 Boron 2500 ppm The solutions are considered aerated through the entire exposure period as in the case of pure boric acid spray solution. 
FNP-FSAR-6A   
6A-3 REV 23  5/11 6A.1.4  TRACE COMPOSITION OF EMERGENCY CORE COOLING SOLUTION  During spraying and recirculation, the emergency core cooling (ECC) solution will wash over virtually all the exposed components and structures in the reactor containment. The ECC solution is recirculated through a common sump; hence, any contamination deposited in or leached by the solution from the exposed components and structures will be uniformly mixed in the solution. 
The materials compatibility discussion includes consideration of the effects of trace elements which are identified as conceivably being present in the ECC solution during recirculation. 
To identify the trace elements in containment which may have a  deleterious effect on engineered safety features equipment, one must first establish which elements are potentially harmful to the materials of construction of the safety features equipment and second, ascertain the presence of these elements in forms which can be released to the ECC solution following a design basis accident. Table 6A-2 presents a listing of the major periodic group of elements.
Elements known to be harmful to various metals are noted and potential sources of these elements are identified. 
The concentration of the trace contaminants in the ECC solution will vary with individual plant construction as well as with the chemical composition of the ECC solution itself. 
6A.2 MATERIALS OF CONSTRUCTION IN CONTAINMENT  All materials in containment are reviewed from the standpoint of insuring the integrity of equipment of which they are constructed and to insure that deterioration products of some materials do not aggravate the accident condition. In essence, therefore, all materials of construction in the containment must exhibit resistance to the postaccident environment or, at worst, contribute only insignificant quantities of trace contaminants which have been identified as potentially harmful to vital safeguards equipment. Table 6A-3 lists typical material of construction used in the containment. Examples of equipment containing these materials are included in the table. 
Corrosion testing, described in section 6A.3, showed that of all the metals tested only aluminum alloys and zinc were found  incompatible with the alkaline sodium borate solutions. Aluminum and zinc were observed to corrode at a significant rate, with the generation of hydrogen gas.
Since hydrogen generation can be hazardous to containment integrity a detailed survey was conducted to identify all aluminum and zinc components in containment. 
The as-built aluminum inventory present inside the containment is described in drawing A-508597 (Farley Unit 1) and A-508928 (Farley Unit 2). The drawings also include the mass of metal and exposed surface area of each component used in the calculation of hydrogen generated post-LOCA. The 1100- and the 6000-series aluminum alloys are the major types found in containment. This inventory provides some insight into the range of components which are often fabricated from aluminum. All metals of construction in containment, including aluminum, are compatible with unadjusted boric acid solution under DBA conditions. 
FNP-FSAR-6A   
6A-4 REV 23  5/11 The total analyzed value of zinc inventory considered in the analysis of post-LOCA hydrogen generation is described below. Ample margin was included for each source of zinc in the analysis with respect to the zinc inventory for future addition of zinc inside containment.
Zinc Inventory:
Item Surface Area (ft2)
Zinc Based Paint 298,216  Galvanized Carbon Steel 125,864  Cable Trays  44,328 Since the corrosion rate of zinc is considerably lower than the aluminum, the rate of mass depletion of zinc due to corrosion is lower. Therefore, the thickness and mass of the zinc inventory is not considered in the post-LOCA hydrogen generation analysis.
6A.3 CORROSION OF METALS OF CONSTRUCTION IN DESIGN BASIS  ECC SOLUTION  Emergency core cooling components are austenitic stainless steel and, hence, are quite corrosion resistant to the alkaline sodium borate solution as demonstrated by corrosion tests reported in WCAP-7153(1). The general corrosion rate, for Type 304 and 316 stainless steels, was found to be 0.01 mils/months in pH 10 solution at 200&deg;F. Data on corrosion rates of these materials in the alkaline sodium borate solution have been reported by ORNL(2, 3) to confirm the low values. 
Extensive testing was also performed on other metals of construction found in the reactor containment. Testing was performed on these materials to ascertain their compatibility with the spray solution at design post-accident conditions and to evaluate the extent of deterioration product formation, if any, from these materials. 
Metals tested included zircaloy, Inconel, aluminum alloys, cupronickel alloys, carbon steel, galvanized carbon steel and copper. The results of the corrosion testing of these materials are reported in detail in reference 1. Of the materials tested, only aluminum and zinc were found to be incompatible with the alkaline sodium borate solution. Aluminum corrosion is discussed in section 6A.5. The following is a summary of the corrosion data obtained on various materials of construction exposed for several weeks in aerated alkaline (pH 9.0-9.3) sodium borate solution at 200&deg;F. The exposure condition is considered conservative since the test temperature (200&deg;F) is considerably higher than the long term design basis accident temperature (152&deg;F),
and the pH bounds the long term design basis accident pH. Corrosion of zinc in post-LOCA environment is discussed in section 15.4.1.6.2.
FNP-FSAR-6A   
6A-5 REV 23  5/11  Maximum Observed  Corrosion Rate Material    mil/month Carbon Steel 0.003 Zr-4 0.004
Inconel 718 0.003 Copper 0.015
90 - 10 Cu-Ni 0.02 70 - 30 Cu-Ni 0.006 Galvanized carbon steel 0.051 Brass 0.01 Tests conducted at ORNL(2, 3) also have verified the compatibility of various materials of construction with alkaline sodium borate solution. In tests conducted at 284&deg;F, 212&deg;F, and 130&deg;F, stainless steel, Inconel, cupronickel, Monel and zircaloy-2 experienced negligible changes in appearance and negligible weight loss. 
Corrosion tests at both the Westinghouse Pressurized Water Reactor Division and ORNL have shown copper and copper nickel alloys suffer only slight attack when exposed to the alkaline sodium borate solution at DBA conditions. The corrosion rate of copper, for example, in alkaline sodium borate solution at 200&deg;F is ~0.015 mil/month(1). The corrosion of copper in an alkaline sodium borate environment under spray conditions at 264&deg; and 212&deg;F have been reported by ORNL. Corrosion penetrations of less than 0.02 mil was observed after 24-hour exposure at 284&deg;F (reference 3, table 3-13) and a corrosion rate of less than 0.3 mil per month was observed at 212&deg;C.  (See reference 2, table 3-6.) 
It can be seen therefore that the corrosion of copper in the postaccident environment will have a negligible effect on the integrity of the material. Further, the corrosion product formed during exposure to the solution appears tightly bound to the metal surface and hence will not be released to the ECC solution. 
Consideration was given to possible caustic corrosion of austenitic steels by the alkaline solution. Data presented by Swandby(4) shows that these steels are not subject to caustic stress cracking at the temperature (285&deg;F and below) and 6A-6 caustic concentration (less than 1 weight percent) of interest. The stress cracking boundary temperature as defined by Swandby is considerably above (~80&deg;F) the long term, postaccident design temperature of 152&deg;F.
It should be noted when considering the possibility of caustic cracking of stainless that the sodium hydroxide boric acid solution is a buffer mixture wherein no free caustic exists at the FNP-FSAR-6A   
6A-6 REV 23  5/11 temperatures of interest, even should the solution be concentrated locally through evaporation of water; hence the above consideration is somewhat hypothetical with regard to the Farley postaccident environment. 
6A.4 CORROSION OF METALS OF CONSTRUCTION BY TRACE CONTAMINANTS IN ECC SOLUTION  Of the various trace elements that could occur in the emergency core cooling solution in significant quantities, only chlorine (as chloride) and mercury are adjudged potentially harmful to the materials of construction of the safeguards equipment. 
The use of mercury or mercury bearing items, however, has been restricted in the Farley containment. Most mercury vapor lamps, fluorescent lighting, and instruments that employ mercury for pressure and temperature measurements and for electrical equipment have been prohibited in the containment building. Contamination due to exposure to mercury is possible if one or more temporary underwater lights used in the refueling cavity, transfer canal, and the spent-fuel pool were to fail catastrophically. The lights approved for use in these areas are manufactured by ROS, model HPS-1000, and contain up to 3 mg of mercury each in double encapsulated bulbs. The use of up to twelve of these lights at any one time has been evaluated as acceptable.
The possibility of chloride stress corrosion of austenitic stainless steels has also been considered. It is believed that corrosion by this mechanism will not be significant during the postaccident period for the following reasons: 
6A.4.1  LOW TEMPERATURE OF ECC SOLUTION  The temperature of the ECC solution is reduced after a relatively short period of time (i.e. a few hours) to about 150&deg;F. While the influence of temperature on stress corrosion cracking of stainless steel has not been unequivocally defined, significant laboratory work and field experience indicate that lowering the temperature of the solution decreases the probability of failure. Hoar and Hines(5) observed this trend with austenitic stainless steel in 42 weight percent solutions of MgCl2 with temperature decrease from 310&deg; to 272&deg;F. Staehle and Latanision(6) present data which also shows a decreased probability of failure with decreasing solution temperature from about 392&deg;F to 302&deg;F. Staehler and Latanision(6) also report the data of Warren(7) which showed the significant change with decrease in temperature from 212&deg;F to 104&deg;F. The work of Warren, while pertinent to the present consideration in that it shows the general relationship of temperature to time to failure, is not directly applicable in that the chloride concentration (1800 ppm Cl) believed to have effected the failure was far in excess of reasonable chloride contamination that may occur in the ECC solution. 
FNP-FSAR-6A   
6A-7 REV 23  5/11 6A.4.2  LOW CHLORIDE CONCENTRATION OF ECC SOLUTION  It is anticipated that the chloride concentration of the ECC solution during the postaccident period will be low.
Restrictions in the chloride content of the water used in the postaccident period will not impair system operability. The environment of low chloride concentration, low temperature, and high pH, which will be experienced during the long-term postaccident period, will not be conducive to chloride cracking.  $$[HISTORICAL]$$  $$[Surveillance has been maintained throughout plant construction to ensure that the chloride inventory is maintained at a minimum.]$$
6A.5 CORROSION OF ALUMINUM ALLOYS  Corrosion testing showed that aluminum alloys are not compatible with alkaline borate solution.
The alloys generally corrode fairly rapidly, at the post-accident condition temperatures, with the liberation of hydrogen gas. A number of corrosion tests were conducted in the Westinghouse Pressurized Water Reactor Division laboratories and at ORNL facilities. A review of applicable aluminum corrosion data is given in Table 6A-5. The corrosion rates at the various temperature steps were determined from the  aluminum corrosion rate design curve which was chosen to include essentially all available corrosion data. 
6A.6 THE NATURE AND BEHAVIOR OF ALUMINUM CORROSION PRODUCTS  IN ALKALINE SOLUTION  The corrosion of aluminum in alkaline solution, expected following a design basis accident (DBA), has been shown to proceed with the formation of aluminum hydroxide(12,13,14) and the aluminate ion, as well as with the production of hydrogen gas. 
The expected DBA conditions include the establishment of an alkaline ECC solution having a total volume of liquid of 4.5 x 105 gal after actuation of the engineered safety  features.
As mentioned above, aluminum is known to corrode in alkaline solutions to give a precipitate of Al(OH)3, which in turn can redissolve in an excess of alkali to form a complex aluminate. Van Horn(12) noted that the precipitation of Al(OH)3 begins about pH 4 and is essentially complete at pH 7. A further increase in pH to about 9 causes dissolution of the hydroxide with the formation of the aluminate. 
It can be seen, therefore, that the solubility of aluminum corrosion product is a function of the pH of the environment. Consistent with this, the corrosion of aluminum is also strongly dependent on the solution pH, since when the corrosion products are dissolved from the metal surface, corrosion of the base metal can proceed more freely. 
Aluminum corrosion rate data had been reported in WCAP-7153(1), Table 8. The corrosion rate of aluminum is seen to decrease by a factor of 21 (1/.048) as the pH decreases from 9.3 to 8.3, and by a factor of 83 (1/.032) as the pH decreases from 9.3 to 7.0. Therefore, one must consider both corrosion and the dissolution of the corrosion products at specific reference FNP-FSAR-6A   
6A-8 REV 23  5/11 conditions since the two are directly related. The corrosion reactions that are of interest in the DBA condition here would include the reaction of aluminum in alkaline solution to form aluminum hydroxide: i.e., 
  ++2H3(OH)Al2O2H6Al2 (1)  and dissolution of the hydroxide to form the aluminate, i.e., 
      (2)
A knowledge of the solubility product of the aluminum hydroxide in an alkaline solution allows the determination of the solubility expected for the hydroxide in the DBA environment. 
Deltombe and Pourbaix(15) have determined the solubility product of aluminum hydroxide. Using the value of 2.28 x 10-11 for Ksp, as reported by Deltombe and Pourbaix, the following calculation can be made. 
The solubility of Al(OH)3 is determined from equation 2    [][][][]+=+=+++H2AlO1110x28.2H2AlOspKO2HH2AlO3Al(OH)  at pH = 9.3
  []rmoles/lite210x6.41010x51110x28.22AlO==  Therefore, the solubility of Al(OH)3 in a pH 9.3 solution at 25&deg;C (77&deg;F) is 4.6 x 10-2 moles/liter or 3.0 x 10-2 lb/gal. Expressed as aluminum, the solubility at these conditions is 1.05 x 10-2 lb/gal.
The solubility of the aluminum corrosion products in the post-accident environment is a function of both solution pH and temperature. Plots of the corrosion product solubility are expressed in terms of aluminum versus solution pH for temperatures of 77&deg;F and 150&deg;F. The change in solubility with temperature is found utilizing the relationship of the free energy of formation, temperature, and the solubility product. 
With the knowledge of the reference aluminum corrosion behavior for any specific plant, one can calculate the expected solubility limits for the corrosion reaction. 
For the Farley plant, 4.5 x 105 gal of ECC solution will be present in the containment after actuation of the safety features. The as-built aluminum inventory present inside the containment is described in drawing A-508597 (Farley Unit 1) and A-508928 (Farley Unit 2).
FNP-FSAR-6A   
6A-9 REV 23  5/11 Table 6A-7 presents a summary of the applicable solubility and corrosion parameters for various conditions. The table lists the applicable solubility products (Ksp) and solubilities at the various temperatures and solution pHs together with the soluble aluminum limit for the Farley system at the specific conditions. The last values in the table give the aluminum solubility margin after 100 days corrosion; that is, the soluble Al limit divided by the aluminum corroded. It can be seen  that in all cases, including the low temperature and low pH  conditions, the ECC solution is not expected to be saturated with aluminum corrosion products. Further, within the expected design conditions for temperature  and pH, adequate aluminum solubility margin is available as shown on table 6A-7.
It is concluded therefore, that the corrosion products of aluminum will be in the soluble form during the post accident period considered and, hence, there is no potential for deposition on flow orifices, spray nozzles or other equipment. 
6A.6.1  BEHAVIOR OF CIRCULATING ALUMINUM CORROSION PRODUCTS The solubility of aluminum corrosion products as shown that for the Farley plant, the entire inventory produced after 100 days exposure to the post-DBA condition would remain in solution.
The review also indicates that the ECC solution is only approximately 5.5 percent saturated at 77&deg;F and less than 3 percent saturated at 150&deg;F. 
It is of interest, however, to review the experience of facilities which have operated with insoluble aluminum corrosion products and to relate their conditions with those expected in the post accident environment. 
The most significant experience available to date is that of Griess(16) who operated a recirculating test facility to measure the corrosion resistance of a variety of materials in alkaline sodium borate spray solution. 
Tests were conducted on 1100, 3003, 5052, and 6061 aluminum alloys exposed at 100&deg;C in pH 9.3 sodium borate solution (0.15 M NaOH - 0.28 M H3B03). It was reported that even though the solution contained copious amounts of flocculent aluminum hydroxide, it has no effect on flow through the spray nozzle (0.093-in. orifice). The pH of the solution did not change because of the increase in the corrosion products. 
Griess(a) in describing his observations with regards to aluminum corrosion product deposition potential stated that: 
A. No significant deposition was observed on the cooling coil installed in the solution.
B. No significant deposition was observed on the heated surfaces of the facility. 
C. No significant deposition was observed on isothermal facility surfaces. 
a. Private communication.
FNP-FSAR-6A   
6A-10 REV 23  5/11 The amounts of aluminum corroded to the solution in the tests conducted by Griess at 55&deg;C and 100&deg;C were approximately 4.0 and 18.6 grams, respectively. The concentration of aluminum present in the recirculation stream, therefore, was approximately 0.2 and 1 gram/liter, respectively. This value is about a factor of about 5 above the aluminum concentration expected in the postaccident ECC solution at the Indian Point plant in a pH 9.3 solution after 100 days. 
Hatcher and Rae(17) describe the appearance of turbidity in the  Canadian National Research Experimental Reactor Unit (NRU) reactor and "propose" that deposition of aluminum corrosion products may have occurred on heat exchanger surfaces, although they do not report any specific examination results. Moreover, Hatcher and Rae report no operations problems associated with the presence of aluminum corrosion product turbidity in the NRU reactor. The overall heat transfer coefficient for each NRU reactor heat exchanger was measured after 2 years of full power operation on several occasions and within the limit of accuracy of the measurements, reported at approximately 5 percent, no change in the thermal resistance had been observed. 
It is concluded, therefore, from the work of Griess and Hatcher and Rae, that the deposition of aluminum corrosion products on heat exchangers, surfaces will not be significant in the postaccident environments even for the circumstances of insoluble product formation.
6A.7 EFFECT OF POSSIBLE CHEMICAL REACTIONS ON IODINE REMOVAL  CAPABILITY OF THE CONTAINMENT SPRAY SOLUTION In evaluating the effect of possible chemical reactions on the iodine removal capability of the spray solution, it has been determined that the reaction of aluminum with an alkaline ECC solution is the only reaction occurring in the containment system during a design basis accident (DBA) which has the potential for influencing the chemistry of the ECC solution. The corrosion rate of aluminum and the solubility of the aluminum corrosion products is dependent on the pH and temperature of the alkaline spray solution. Calculations are presented in this review which estimate the mass of aluminum which would be corroded in the Farley containment following a DBA, the mass of aluminum corrosion products which would be formed, and the solubility of these corrosion products in the emergency core cooling solution. As the values in table 6A-7 indicate, there is a conservative aluminum solubility margin in  the ECC solution during DBA conditions. 
In the operation of a test facility to measure the corrosion resistance of a variety of materials in alkaline sodium borate spray solution, the experience of Griess(16) was that the pH of the solution did not change as a result of the buildup of aluminum corrosion products. At concentrations of 0.2 - 1.0 g of aluminum per liter, the test facility experience is representative of the Farley post accident environment, assuming that all of the aluminum in the containment had corroded away and was present in the sump solution. Although no reduction in the sump solution pH is anticipated, the equilibrium sump solution pH of 7.5 exceeds the pH required to assure that iodine is retained in the sump solution. 
FNP-FSAR-6A   
6A-11 REV 23  5/11 6A.8 COMPATIBILITY OF PROTECTIVE COATINGS WITH POSTACCIDENT  ENVIRONMENT The investigation of materials compatibility in the postaccident design basis environment also includes an evaluation of protective coatings for use in containment. 
The results of the protective coatings evaluation presented in WCAP-7198(11) showed that several inorganic zinc, modified phenolics, and epoxy coatings are resistant to an environment of high temperature (320&deg;F maximum test temperature) and alkaline sodium borate. Long term tests included exposure to spray solution at 150&deg;F - 175&deg;F for 60 days, after initially being subjected to the conservative containment temperature transient shown in table 6A-1. The protective coating found to be resistant to the test conditions, that is, exhibited no significant loss of adhesion to the substrate nor formation of deterioration products, comprises virtually all of the protective coatings recommended for use in the containment. Hence, the protective coatings will not add deleterious products to the core cooling solution. 
It should be pointed out that several test panels of the recommended types of protective coatings were exposed for two DBA cycles and showed no deterioration or loss of adhesion with the substrate. In addition, the protective coatings applied to the components of the containment do not function as an integral part of the engineered safeguard features during DBA conditions.
Although the protective coatings are selected for use on the basis of their performance during a DBA, they do not serve as an engineered safety feature to inhibit corrosive attack following a loss-of-coolant accident on the substrates on which they are applied. 
6A.9 EVALUATION OF THE COMPATIBILITY OF CONCRETE ECC SOLUTION  IN THE POSTACCIDENT ENVIRONMENT Concrete specimens were tested in boric acid and alkaline sodium borate solutions at conditions conservatively (320&deg;F maximum and 200&deg;F steady state) simulating the post-DBA environment. 
The purpose of this study was to establish: 
A. The extent of debris formation by solution attack of the concrete surfaces. 
B. The extent and rate of boron removal from the ECC solution through boron concrete reaction.
Tests were conducted in an atmospheric pressure, reflux apparatus to simulate long term exposure conditions and in a high pressure autoclave facility to simulate the DBA short term, high temperature transient. 
Table 6A-8 presents a summary of the data obtained from the concrete boron test series. 
Testing of uncoated concrete specimens in the post accident environment showed that attack by both boric acid and the alkaline boric acid solution is negligible and the amount of FNP-FSAR-6A   
6A-12 REV 23  5/11 deterioration product formation is insignificant. In addition, the boron removal rate from the ECC solution is low. 
FNP-FSAR-6A   
6A-13 REV 23  5/11 REFERENCES 1. Bell, M.J., Bulkowski, J.E. and Picone, L.F.,  "Investigation of Chemical Additives for Reactor Containment Sprays," WCAP-7153, March 1968.  (Westinghouse Proprietary) 2. ORNL Nuclear Safety Research & Development Program Bimonthly Report for July-August 1968, ORNL TM-2368, p. 78.
3. ORNL Nuclear Safety Research & Development Program Bimonthly Report for September-October 1968, ORNL TM-2455, p. 53 4. Swandby, R.K., Chemical Engineer 69, 186 (November 12, 1962).
5. Hoar, T.P., and Hines, J.G., "Stress Corrosion Cracking of  Austenitic Stainless Steel in Aqueous Chloride Solutions," Stress Corrosion Cracking and Embrittlement (ed. W.D. Robertson) John Wieley and Sons, 1956.
6. Latanision, R.M., and Staehle, R.W., Stress Corrosion Cracking of Iron - Nickel Chromium Alloys, Dept. of Metallurgical Engineering, The Ohio State University 7. Warren, D., Proceeding of Fifteenth Annual Industrial Work Conference, Purdue University, May 1960.
8. Edeleanu, C., JISI 173, 1963, 140.
9. Thomas, K.C., et al., "Stress Corrosion of Type 304 Stainless Steel in Chloride Environment," Corrosion, Vol. 20, 1964, p. 89t.   
: 10. Sharfstein, L.R., and Brindley, W.F., "Chloride Stress Corrosion Cracking of Austenitic Stainless Steel - Effect of Temperature and pH," Corrosion, Vol. 14, 1958, p. 588t. 
: 11. Picone, L.F., "Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198, April 1968. (Westinghouse Proprietary) 
: 12. Van Horn, K.C., Aluminum, Vol. I, American Society of Metals, (1967).   
: 13. Sundararajan, J., and Rama Char, T.C., Corrosion 17, 39t, (1961).   
: 14. Cotton, F.A., and Wilkinson, G., Advanced Inorganic Chemistry, Interscience Publishers, (1962).   
: 15. Deltombe, E., and Pourbaix, M., Corrosion 14, 496t, (1958).   
: 16. Griess, J.C., et al., "Corrosion Studies," pp. 76-81, ORNL Nuclear Safety Research and Development Program Bimonthly, July - August 1968, USNRC Report ORNL TM-2368.
FNP-FSAR-6A   
6A-14 REV 23  5/11 17. Hatcher, S.R., and Rae, H.K., Nuclear Sci. and Eng., 10, 316, (1961).   
: 18. Rubin, K., Grover, J. L., Henninger, W. A., and Miller, T. A., "Methodology for Elimination of the Containment Spray Additive," WCAP-11611, Rev. 0, March 1988, (Westinghouse Proprietary) 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-1  POSTACCIDENT CONTAINMENT TEMPERATURE TRANSIENT USED IN THE MATERIAL COMPATIBILITY REVIEW  Time  Interval (s) Temperature (&deg;F)          0 - 300 285    300 - 1000 266  1000 - 2000 234  2000 - 4000 190    >4000 147 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-2 (SHEET 1 OF 2)  REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION    Representative    Group    Elements      Corrosivity of Elements    Sources of Elements      0 H3, Ne, K, Xe No effect on any materials of Fission product release    construction      I a Li, Na, K Generally corrosion inhibitive Li - coolant pH adjusting    properties for steels, and      agent  copper alloys - harmful to Na - spray additive solution,  aluminum      concrete leach product    K  - concrete leach product    II a Mg, Ca, Sr, Ba Generally not harmful to steel Concrete leach products -    or copper base alloys deteriorated insulation      III a Y, La, Ac Not considered harmful in low Fission product release    concentrations      IV a Ti, Zr, Hf Not considered harmful to any Fuel rod cladding, control    materials rod material, alloying    constituent    V a V, Nb, Ta Not considered harmful to any Alloying constituents in    materials low concentration    VI a Cr, Mo, W Not considered harmful to any Alloying constituents in    materials equipment    VII a Mn, Tc, Re Not considered harmful Mn - alloy constituent      VIII Fe, Ni, Cr, Os Fe, Ni, Cr - not harmful to  Fe, Ni, Cr - alloying    any materials constituents. Others have    no identifiable sources FNP-FSAR-6A REV 21  5/08 TABLE 6A-2 (SHEET 2 OF 2)  REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION    Representative    Group  Elements      Corrosivity of Elements    Sources of Elements      I b Cu, Ag, Au Not harmful to any materials Cu present as material of    construction and alloying    constituent    II b Zn, Cd, Hg Hg - harmful to stainless Hg has been entirely excluded        steel, Cu alloys, from use in the containment.        aluminum Cd finish plating on  Zn - unknown components. Zn galvanizing  Cd - unknown and alloying constituent    III b B, A1, Ga, In Not harmful to material B  - neutron poison additive    A1 - materials of construction IV b C, Si, Sn, Pb C, Si, Sn not harmful to Si - concrete leach product    materials. Pb considered Pb - alloy constituent in    harmful to nickel alloys      some brazes  V b N, P, As, Sb, Bi No effect from N unless N  - containment air. Others    ammonia is formed. Others not identified in significant    unknown materials    VI b O, S, Se, Te S possibly harmful to nickel Te - fission product    alloys S  - oils, greases, insulating        materials    VII b F, C1, Br, I F considered potentially C1 - concrete leach product    harmful to zircaloy.      general contamination    C1 potentially harmful to F  - organic materials    stainless steel Br and I, I and Br - fission products    not generally harmful low concentration FNP-FSAR-6A REV 21  5/08 TABLE 6A-3  TYPICAL MATERIALS OF CONSTRUCTION IN THE FARLEY CONTAINMENT  Material Equipment Application  300 series stainless Reactor coolant system, residual steel heat removal loop, spray system,  fan cooler material 400 series stainless Valve materials steel 
Inconel (600, 718) Steam generator tubing, reactor  vessel nozzles, core supports, and  fuel rod grids Galvanized steel Ventilation duct work, CRDM shroud  material, I & C conduit Aluminum Refer to drawing A-508597 for  Farley Unit 1 and A-508928 for  Farley Unit 2 Copper Service water piping, fan cooler  material
70-30 Cu Ni Fan cooler material 90-10 Cu Ni Fan cooler material Carbon steel Component cooling loop, structural  steel, main steam piping, etc Monel Possibly instrument housings Brass Possibly instrument housings Protective coatings General use on carbon steel  structures and equipment,  Inorganic zincs concrete Epoxy Modified phenolics 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-4 
This table has been deleted. 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-5  CORROSION OF ALUMINUM ALLOYS IN ALKALINE SODIUM BORATE SOLUTION Corrosion    Data Temperature Alloy  Test    Rate  Exposure  Point    (&deg;F)    Type  Duration (mg/dm2/h) pH  Condition        Reference            1    275 5025  3 hours    96.2 9 Solution WCAP-7153, Table 9            2    275 5005  3 hours  840 9 Solution WCAP-7153, Table 9            3    200 6061 320 hours    15.4 9.3 Solution WCAP-7153, Table 8        WCAP-7153, Figure 9          4    210 5052  7 days    53.0 9 Solution WCAP-7153, Table 7        WCAP-7153, Figure 8          5    210 5052  2 days    14.0 9 Solution WCAP-7153, Table 5            6    210 5005  2 days    27.1 9 Solution WCAP-7153, Table 5            7    284 5052  1 day    54 9.3 Spray ORNL-TM-2425, Table 3.1          8    284 5052  1 day    31.5 9.3 Solution ORNL-TM-2425, Table 3.1            9    212 6061  3 days  126 9.3 Spray ORNL-TM-2368, Table 3.6          10    212 6061  3 days  110 9.3 Solution ORNL-TM-2368, Table 3.6          11    150 6061  7 days    2.9 9.3 Solution Westinghouse Pressurized        Water Reactor Division        recent data          12    150 5052  7 days    4.2 9.3 Solution Westinghouse Pressurized        recent data 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-6 
This table has been deleted. 
FNP-FSAR-6A REV 21  5/08 TABLE 6A-7  SUMMARY OF ALUMINUM CORROSION PRODUCT SOLUBILITY DATA      Solution Temperature            77&deg;F  150&deg;F    Parameter pH 9.3  pH 8.3 pH 9.3  pH 8.3        Solubility product 2.28 x 10-11  2.28 x 10-11 4.16 x 10-10  4.16 x 10-10    Ksp                Al solubility 1.05 x 10-2  1.05 x 10-3 1.9 x 10-1  1.9 x 10-2    (lb Al/gal)              Soluble Al limit(a) 4.73 x 103  4.73 x 102 8.55 x 104  8.55 x 103    for ECCS (lb)              Al corrosion rate (Not used)  (Not used) 1  0.048    (normalized)              Al corroded after (Not used)  (Not used) 1800  1077  100 days (lb)              Al solubility margin 18  3 47.5  7.9  at 100 days       
a. Solution volume 4.5 x 105  gal.
FNP-FSAR-6A REV 21  5/08 TABLE 6A-8  CONCRETE SPECIMEN TEST DATA    Total  Exposed Initial  Concrete Exposure Surface Weight Specimen  - Boron Period Volume Change Weight  Test No. (days) (in. /gal) (grams) (grams)    Visual examination      1 24 28 - 22.4 560.0 No apparent change      3 28 20 + 21.5 404.0 Light, yellowish, deposit on      specimen      4 72 38 0 641.2 No apparent change - coating      adhesion excellent      5 72 43 - 0.2 769.5 Light, hard deposit on specimen        6 ~4(a) 54 - 601.4 No apparent change - small      amount of sand particles in      test can      7 175 23 + 11.0 457.0 No apparent change      8 175 38 + 26.5 751.0 No apparent change - coating      adhesion excellent      9 ~5(a) 78 + 4.0 702.0 No apparent change - coating      adhesion excellent 
a. These tests were at high temperature DBA transient conditions. All others at 195 - 205&deg;F.
FNP-FSAR-6B   
6B-i REV 21  5/08 APPENDIX 6B  CONTAINMENT PRESSURE ANALYSIS  TABLE OF CONTENTS  Page  6B.1 CONTAINMENT PRESSURE RESPONSE................................................................6B-1 6B.2 CONTAINMENT SUBCOMPARTMENT ANALYSIS...................................................6B-1 
FNP-FSAR-6B   
6B-ii REV 21  5/08 LIST OF TABLES 
6B-1 Node Spacings
6B-2 Thickness of Sensitive Heat Conduction Layer
6B-3 Mesh Spacing in Sensitive Layer to Achieve 0.5% Accuracy
FNP-FSAR-6B   
6B-iii REV 21  5/08 LIST OF FIGURES 
6B-1 Reactor Cavity Block Diagram
6B-2 Steam Generator Cavity Pressurization Analysis
6B-3 Total Horizontal Force vs. Time
6B-4 Reactor Cavity Analysis
FNP-FSAR-6B   
6B-1 REV 21  5/08 APPENDIX 6B  CONTAINMENT PRESSURE ANALYSIS 6B.1 CONTAINMENT PRESSURE RESPONSE  The containment pressure response to a loss-of-coolant accident (LOCA) has been analyzed using the heat sinks as presently designed. The methods and assumptions used in this analysis are described in paragraph 6.2.1. The double-ended pump suction break was originally determined to be the worst case. The analysis for the break showed a peak pressure of 48 psig at 276 s and a maximum temperature of 313&deg;F, at 55 s after the break. Current results are provided in paragraph 6.2.1.3.6.
A summary of the current heat sinks is given in Table 6.2-2. Table 6B-1 provides a table of the original node spacings for original heat sinks. Node spacings for power uprate analyses are generally more fine or comparable to those shown in Table 6B-1. Detailed conservative calculations were performed to determine each heat sink surface area. For additional conservatism, some heat sinks (e.g., all piping in the containment and miscellaneous steel such as some support brackets and rails) were not included in the analysis. 
6B.2 CONTAINMENT SUBCOMPARTMENT ANALYSIS  The following section provides a discussion of the original design prior to application of leak-before-break exclusion of RCS main loop breaks. Current analyses and results are provided in paragraph 6.2.1.3.4.1.
The containment subcompartments analyzed for the pressure response following a LOCA were the reactor cavity and the steam generator annulus (the volume below the steam generator compartments). The pressure transient analysis was performed using a Bechtel computer code which calculates short term pressure and temperature responses. The code conservatively neglects heat transfer and all engineered safety features. A detailed description of the code is provided in appendix 3K, attachment D. 
The model used for the reactor cavity analysis is shown in figures 6B-1, 6B-2, and drawing D-176277. Volumes, vent area, and flow coefficients are also shown in figure 6B-1. Blowdown data was supplied by Westinghouse for the 1 ft2 cold leg break (at 95&deg; az. in drawing D-176277) which is the limiting case for reactor cavity design. The blowdown is split equally between volumes 1 and 2. Insulation in the break region (compartments 1 and 2) is assumed to blow off and completely plug the cold leg penetration at the wagon wheel restraint, as well as the support shoe area ventilation duct. All gaps in the broken leg blowdown restrictor/baffle plate remain completely unobstructed by insulation throughout the transient. In all other places (i.e.,
reactor vessel, nozzle, and pipes for all intact legs) insulation is assumed to remain in place and not crush, leaving the seal ring gap and unbroken leg baffle plate gaps open for ventilation to the containment. The maximum horizontal force was calculated to be 1.4 x 106 lbf. The maximum uplift force was 5.9 x 104 lbf. The force-time history results are shown in figures 6B-3 and 6B-4.
FNP-FSAR-6B   
6B-2 REV 21  5/08 The flow models for the steam generator compartment pressurization analyses are shown in figure 6B-2. Blowdown data were supplied by Westinghouse for a double ended cold leg break in the steam generator compartment C, which is the limiting case. The maximum differential pressure between steam generator compartment C and the containment was found to be 33.9 psia at 0.42 seconds. 
FNP-FSAR-6B REV 21  5/08 TABLE 6B-1 (SHEET 1 OF 5)  NODE SPACINGS  Heat Sink No. 1 - Containment Cylinder and Dome Node Spacing Thickness Material    (in.)      (in.)
Paint 1 x 10-3 2.0 x 10-2 Primer(a) 1 x 10-3 3.0 x 10-3 Carbon steel 6.25 x 10-2 2.5 x 10-1 Concrete region 1 5.0 x 10-2 3.0 Concrete region 2 4.0 x 10-1 6.0 Concrete region 3 1.2 6.0 Concrete region 4 10.0 30.0  Heat Sink No. 2 - Unlined Concrete Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 18.0 x 10-3 Surfacer(a) 1.0 x 10-2 1.25 x 10-1 Concrete region 1 5.0 x 10-2 3.0 Concrete region 2 1.76 x 10-1 3.0 Concrete region 3 6.0 x 10-1 3.0  Heat Sink No. 3 - Outside Reactor Cavity Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 18.0 x 10-3 Surfacer(a) 1.0 x 10-2 1.25 x 10-1 Concrete 5.0 x 10-2 3.0 FNP-FSAR-6B REV 21  5/08 TABLE 6B-1 (SHEET 2 OF 5)  Heat Sink No. 4 - Galvanized Steel Node Spacing Thickness Material    (in.)      (in.)
Zinc 6.7 x 10-4 3.35 x 10-3 Carbon steel 6.5 x 10-3 6.56 x 10-2   
Heat Sink No. 5 - Miscellaneous Steel Less than 0.12 in. Thick 
Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 7.64 x 10-2  Heat Sink No. 6 - Miscellaneous Steel 0.12 to 0.16 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 1.32 x 10-1  Heat Sink No. 7 - Miscellaneous Steel 0.16 to 0.24 in. Thick Node Spacing Thickness Material    (in.)  (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 1.91 x 10-1 FNP-FSAR-6B REV 21  5/08 TABLE 6B-1 (SHEET 3 OF 5)  Heat Sink No. 8 - Miscellaneous Steel 0.24 to 0.30 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 2.55 x 10-1  Heat Sink No. 9 - Miscellaneous Steel 0.30 to 0.40 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 3.38 x 10-1  Heat Sink No. 10 - Miscellaneous Steel 0.40 to 0.50 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 4.92 x 10-1  Heat Sink No. 11 - Miscellaneous Steel 0.50 to 0.625 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 5.76 x 10-1 
FNP-FSAR-6B REV 21  5/08 TABLE 6B-1 (SHEET 4 OF 5)  Heat Sink No. 12 - Miscellaneous Steel 0.625 to 0.75 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 7.24 x 10-1  Heat Sink No. 13 - Miscellaneous Steel 0.75 to 1.0 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.5 x 10-2 9.35 x 10-1  Heat Sink No. 14 - Miscellaneous Steel 1.0 to 1.5 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 2.0 x 10-2 1.43  Heat Sink No. 15 - Miscellaneous Steel Greater than 1.5 in. Thick Node Spacing Thickness Material    (in.)      (in.)
Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 3.5 x 10-2 2.85 FNP-FSAR-6B REV 21  5/08 TABLE 6B-1 (SHEET 5 OF 5)  Heat Sink No. 16 - Stainless Steel Node Spacing Thickness Material    (in.)      (in.)
Stainless steel 5.0 x 10-3 1.68 x 10-1 
____________________
: a. When Amercoat 90 is used as the primer, the average primer thickness will be 5.0 mils. However, the total thickness of primer plus finish coat will not exceed the total thickness of finish coat plus primer (surfacer) listed in the table.
FNP-FSAR-6B REV 21  5/08 TABLE 6B-2  THICKNESS OF SENSITIVE HEAT CONDUCTION LAYER Typical Heat Conduction Time Materials 20 s 100 s 200 s 400 s Concrete A 0.054 ft 0.121 ft 0.170 ft 0.243 ft K =  1.0    Cp = 25.2 Steel 0.200 ft 0.450 ft 0.640 ft 0.906 ft K = 29.6    Cp = 53.6 Inorganic Zinc 0.058 ft 0.130 ft 0.184 ft 0.260 ft Primer K =  1.24    Cp = 27.36   
FNP-FSAR-6B REV 21  5/08 TABLE 6B-3  MESH SPACING IN SENSITIVE LAYER TO ACHIEVE 0.5 PERCENT ACCURACY Typical Accuracy Crossover Time  Materials 20 s 40 s 100 s 200 s 400 s      Concrete A 224 158 100 71 50 K = 1.0 mesh mesh mesh mesh mesh CP = 25.2 ft ft ft ft ft      Steel 60 42 27 19 13 K = 29.6 mesh mesh mesh mesh mesh Cp = 53.6 ft ft ft ft ft      Inorganic 210 148 94 66 47 zinc mesh mesh mesh mesh mesh primer ft ft ft ft ft K = 1.24      Cp = 27.36     
REV 21  5/08 REACTOR CAVITY BLOCK DIAGRAM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-1 
REV 21  5/08 STEAM GENERATOR CAVITY PRESSURIZATION ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-2 REV 21  5/08 TOTAL HORIZONTAL FORCE VERSUS TIME  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-3 REV 21  5/08 REACTOR CAVITY ANALYSIS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-4 FNP-FSAR-6C    6C-i REV 21  5/08 [HISTORICAL (Prior to December 2007)] [APPENDIX 6C CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM  TABLE OF CONTENTS                  Page  6C.I CONTAINMENT SUMP DESCRIPTION -----------------------------------------------6C-1  6C.II ECCS RECIRCULATION MODE TEST PROGRAM ---------------------------------6C-3  6C.III UNIT 1 TESTS-------------------------------------------------------------------------------6C-3  6C.IV UNIT 2 TESTS-------------------------------------------------------------------------------6C-13 
FNP-FSAR-6C    6C-ii REV 21  5/08 LIST OF TABLES 6C-1 Test Conditions for Unit 1 Intake 1  6C-2 Test Conditions for Unit 1 Intakes 2, 3, and 4 
FNP-FSAR-6C 6C-iii REV 21  5/08 LIST OF FIGURES 
[Historical] 6C-1 Typical Arrangement of Containment Sump Suction Line  6C-2 Modeled Areas of ECCS Intakes  6C-3 Hydraulic Model Plan for Intake No. 1 Tests  6C-4 Blockage Test Conditions  6C-5 No. 1 Intake Configuration for Initial Tests  6C-6 Improved Design Intake No. 1  6C-7 Plan of Modeled Area Containing Intakes 2, 3, and 4  6C-8 Containment Sump  6C-10 Photograph of Model  6C-11 Photograph of Model  6C-12 Intakes 2, 3, and 4 Improved Design  6C-13 Blockage Test Conditions for Intakes 2, 3, and 4  6C-14 Photograph of Grating Cage Over Intake 2  6C-15 Plan of Unit 2 Test Facility  6C-16 Section of Unit 2 Test Facility  6C-17 Representative Screen-Grating Structure  6C-18 Grating Cage - Final Design  6C-19 Sump Area of Unit 2  6C-20 Composite Drawing of Unit 2 Sump  6C-21 Composite Drawing of Unit 1 Sump  6C-22 Photo of Unit 2 Grating Cage  6C-23 Photo of Representative Screen-Grating Cage]   
FNP-FSAR-6C 6C-1 REV 21  5/08 [HISTORICAL (Prior to December 2007][APPENDIX 6C  CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM]
Appendix C was made historical in December 2007 following the installation of new containment sump strainers for RHR and CS suction inlets. This was required by Generic Letter (GL) 2004-02, "Potential Impact for Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors."  Appendix 6D has been created to provide a description of the containment sump and the new suction strainers. Appendix 6C contains the design bases for the original containment sumps and is being maintained for historical reference. 
[HISTORICAL] [I. CONTAINMENT SUMP DESCRIPTION  The containment recirculation sump is a collecting reservoir designed to provide an adequate supply of water, with a minimum amount of particulate matter, to the containment spray system (CSS) and the residual heat removal system (RHRS). The containment sump performance meets the NRC acceptance criteria contained in General Design Criteria 35, 36, and 37, and the five NRC acceptance criteria listed below. A. The net positive suction head (NPSH) available to each safety system pump has been shown to provide adequate margin over the required NPSH at limiting runout conditions (see FSAR paragraph 6.3.2.14). B. Housekeeping requirements specified in the quality assurance program and the Technical Requirements Manual. C. The avoidance of materials likely to form debris small enough to pass through sump screens. D. The lack of an apparent mechanism for generating debris large enough to block more than 50 percent of the screen area. E. The ability to monitor and control RHRS status. The design criteria for the containment sumps and sump screens are the following:    A. Separate sumps are provided to serve each of the redundant halves of the ECCS and CSS. The redundant sumps are physically separated from each other and are located outside the missile barrier. The sumps are located on the lowest floor elevation in the containment, exclusive of the reactor vessel cavity. B. The Unit 1 sump intakes are protected by an outer trash rack and a fine mesh inner screen with a steel grating support. The size of the openings in the fine screen take into account the overall operability of the system served.
FNP-FSAR-6C 6C-2 REV 21  5/08  C. A solid plate covers most of the top of each screen structure. This plate can be removed to facilitate inspection of the structure and pump suction intake. The top deck will be fully submerged after a LOCA and completion of safety injection. D. Materials for the grating and screens were selected to avoid degradation during periods of inactivity and operation and have a low sensitivity to adverse effects, such as stress corrosion that may be induced by the chemically reactive spray during loss-of-coolant accident (LOCA) conditions. E. A vortex breaker is provided at the sump intake end of each of the pump suction pipes. F. The sumps are designed to yield low velocities of approach in the vicinity of the sumps to promote the settling out of debris, and to yield negligible pressure drops through the sump screens. Materials inside containment which could cause sump screen blockage post-LOCA have been eliminated or minimized by design. G. The screens and associated structures have been designed to withstand the vibratory motion of seismic events without loss of structural integrity. H. Each pump suction line is installed with a continuous slope from the sump to the pump to assure free venting of air.  (See figure 6C-1.)  There is a sufficient time interval before start of the recirculation phase to allow complete venting of the suction lines (approximately 30 min). I. Field tests have been performed on the pump suction lines for two purposes:  to flush the lines to remove any possible obstructions, and to verify pressure drop calculations made for pump NPSH requirements. The tests were run with the pump startup strainers in place. A typical sump detail drawing prior to modification is shown on figure 6C-8.]  In each of the four pump suction lines from the containment sump, there are two motor-operated gate valves. There is no interdependency between systems or between the redundant portions of the same system. The motor-operated gate valves in the lines from the containment sump to the various pumps are normally closed and remain closed during the injection phase of emergency core cooling system (ECCS) operation. The protective screened structures in the containment sump will be completely submerged at the end of the injection phase and will remain submerged during the recirculation phase. The various parameters (e.g., flowrates, pressure drops, sump levels, etc.) listed in the following sections are from the original ECCS and CSS recirculation mode testing. The ECCS and CSS flowrates and sump levels utilized in the current pump NPSH calculations are within the range of flowrates and sump levels tested in the original sump recirculation tests. The pressure drop across the sump screen, vortex breaker, sump inlet, and suction piping utilized in the current NPSH calculations have been developed from the original sump recirculation test program and the ECCS field tests based on the calculated ECCS and CSS flowrates. Since the current parameters utilized in the NPSH calculations are bounded by those in the original sump recirculation tests, the ECCS and CSS sump intake design will not develop FNP-FSAR-6C 6C-3 REV 21  5/08 II. ECCS RECIRCULATION MODE TEST PROGRAM    A. PURPOSE    The purpose of this hydraulic model study is to document that the ECCS intakes, of the J. M. Farley Nuclear Plant Units 1 and 2 will not develop unacceptable flow reducing or air entraining vortices. The Unit 1 intakes were tested first. The model boundaries were placed remotely from the screen grating structures around the intakes and selected so as to be able to reproduce the flows in the area external to the intakes. Based on the findings from these tests it was concluded that it was not necessary to model the area outside the screen-grating structure for Unit 2. A description of the intakes, the test program and the results and conclusions for each unit are presented in the following sections. III. UNIT I TESTS    A. INTRODUCTION  The emergency core cooling system intakes of Unit 1 are comprised of two 14 in. and two 10 in. vertical inlets located in three intake areas and are designated as intakes 1, 2, 3 and 4, as shown in figure 6C-2. This section presents the results of testing the 14-inch nominal diameter intakes 1 and 2 and the 10-inch nominal diameter intakes 3 and 4 of Unit 1. The tests were conducted to examine NRC's concern relative to the potential occurrence of vortices near or in the intake areas, which could result in loss of pumping capacity or pump failure due to vibration. Such occurrences could reduce pumping capacity by air entrainment and/or by unacceptably high intake head losses. Air entrainment could also produce unbalanced pressures on the pump impeller and cause pump failure because vibration. Therefore, a satisfactory intake design should be free of air entraining vortices and have acceptable intake loss coefficients. Lack of published and documented information relative to effects of the complex flow patterns approaching the intakes, the grating and screens, and the low viscosity of the heated water precluded analytical or empirical predictions as to whether the intake configuration would be free from objectionable vortex action. The plant conditions do not permit inplace testing. Therefore, a hydraulic model was selected to evaluate the adequacy of the intake design with respect to vortices. Drawing D-175200 shows the general features of the containment sump which could affect the flow of water to the sump area. The elevator shaft in the area of the emergency cooling intakes, figure 6C-2, provided a natural model boundary and facilitated the examination of Intake 1 separately from Intakes 2, 3 and 4.
FNP-FSAR-6C 6C-4 REV 21  5/08 Intakes 1 and 2 design flows range from 3000 to 5900 gal/min. The 5900 gal/min corresponds to two residual heat removal (RHR) pumps taking suction through a single sump line. Intakes 3 and 4 have a design flow rate of 3050 gal/min each. The accident condition postulates that under certain conditions flow could approach the intake from both sides. However, for the majority of cases flow from the left (Q1) would exceed the intake 1 flow rate resulting in flow passing this intake toward Intakes 2, 3, and 4.  (See figure 6C-2.)  The calculated minimum and maximum water levels in the containment are 58.3 and 77.1 inches, respectively, above the floor. The maximum containment sump water temperature during recirculation following a postulated LOCA is 212&deg;F at subcooled pressures. A maximum water temperature of 240&deg;F was assumed for the model study. B. THE MODEL  1. General    First, Intake 1 was modeled at a 1:1 undistorted scale within a 25 ft wide, 60 ft long, 12 ft deep concrete tank. Then Intakes 2, 3, and 4 were modeled in the same concrete tank. All columns, restraints, and piping greater than 2 in. diameter were represented in the model.  (See figures 6C-3, 7, 10, and 11.)  The protective screen and grating structure was constructed in accordance with figure 6C-5 and was modified as shown in figures 6C-5, 6, and 12. The screen cloth consisted of 0.120 in. wire with an effective opening of 51.6 percent. The screen was sandwiched between grating of 1-1/4 in. by 3/16 in. bars on 1-3/16 in. centers. Flow baffles were placed at the extremities of the modeled area to insure uniform flow at the model boundaries. Viewing ports were incorporated in the tank to permit observation of flow conditions within the screen area around the intake. Piezometers were installed to measure static pressures inside and outside of the intake screens. Piezometer taps were installed initially at 5 pipe diameters downstream in the Intake 1 pipe and later at 29 pipe diameters downstream in the same intake pipe. They were also installed at 39.6, 36.7, and 25.7 pipe diameters downstream in the Intake 2, 3, and 4 pipes, respectively. Later, an additional tap was installed at 25.6 pipe diameters downstream of Intake 3. The model was capable of being operated at 50 percent above prototype velocities and up to temperatures of 180&deg;-190&deg;F.      2. Scale Selection    The 1:1 scale was chosen in order to test the intakes under conditions which were as close to postulated LOCA conditions as practically possible. The study of fluid dynamics has shown that the parameters which affect vortex formations may be represented by the following dimensionless numbers:
FNP-FSAR-6C 6C-5 REV 21  5/08  a. Weber number, VD2/, which is the ratio of surface tension to inertia forces. b. Froude number gDV, which is a ratio of gravity to inertia forces. c. Reynolds number,VD , which is the ratio of viscous to inertia forces. d. Circulation number, QRVr2, or the similar Kolf number, which characterizes circulation. e. Strouhal number, VDfe, which characterizes the frequency of eddy shedding. The parameters identified in the preceding dimensionless numbers are:    r - radius of inlet, ft. D - characteristic length, ft., e.g., depth or diameter  R - radius of tank or perhaps flow streamline, ft. Q - discharge, ft3/s  V - characteristic velocity, ft/s  fe - frequency of eddy shedding, s-1  g - gravitational acceleration, ft/s2    - surface tension, lb/ft.  - kinematic viscosity, ft2/s  - Density, slugs/ft3    To reproduce exact dynamic and kinematic similarity on a geometrically similar model would require the value of all dimensionless numbers to be the same in model and prototype. The 1:1 scale model, and the test program which followed, permitted tests to be conducted at prototype values of all numbers, but not simultaneously. Conducting tests at prototype discharges and 170&deg;F - 190&deg;F temperatures reproduced the Froude, Circulation and Strouhal numbers with Reynolds and Weber numbers being lower than prototype values. Augmenting the discharges to reproduce prototype Reynolds number yielded Froude, Circulation, Strouhal and Weber numbers in the model which were higher than prototype values.      Since the Froude number involves the principal parameters related to surface flow phenomena, conducting the tests at prototype discharges establishes the surface flow characteristics outside the screen area concurrent with correctly simulated circulation and eddy shedding (Strouhal number) effects. At FNP-FSAR-6C 6C-6 REV 21  5/08 equivalent Froude numbers, the model Weber number was less than the prototype value.      Flow conditions within the screen area are independent of the Froude number and primarily dependent upon the Reynolds and Circulation numbers. Hence, conducting tests at prototype Reynolds numbers permitted examination of conditions within the screen area, concurrent with the Circulation and Weber numbers being greater than prototype values. Based upon the work of Dagget and Keulegan (reference 18), increasing the Circulation number for a constant Reynolds number increases vortex action. Hence it was considered conservative to conduct tests at prototype Reynolds numbers.      Furthermore with Reynolds number equivalence, the model Weber number was greater than the prototype value, which together with the unaugmented flow tests bracketed the prototype Weber number.      Therefore the 1:1 scale model, with tests conducted at and above prototype discharges, reproduced or exceeded the prototype values of the relevant dimensionless numbers. Exceeding prototype values of the dimensionless numbers was considered to produce conservative results. C. THE MODEL TESTING PROGRAM  The tests examined the performance of Intake 1 over the range of flow conditions and water levels given in table 6C-1, for an unblocked condition and for the five postulated blockage conditions shown on figure 6C-4. These conditions were postulated by considering the nature of debris that could reach the screen, and the paths of the flow approaching the screens. Flow directions for Q1 and Q2 are indicated on figure 6C-2. Tests 1 to 6, table 6C-1, were conducted with and without discharges augmented to develop Reynolds numbers equal to, or larger than, prototype values. A preliminary set of runs was also made on Tests 1 to 6 at prototype discharges, without blockage, to:    1. Establish the general performance characteristics of the intake. 2. Observe surface flow conditions at Froude number equivalence between model and prototype. 3. Establish a basis for comparison of surface flow conditions with conditions at augmented discharges. Tests 7 and 8 were to be conducted with and without blockage at prototype discharges and water temperatures of 170&deg;F. There was full prototype equivalence for these two tests. The tests also examined the performance of Intakes 2, 3, and 4 over the range of flow conditions and water levels given in table 6C-2, for an unblocked condition, and for the five postulated blockage conditions shown on figure 6C-13. Flow directions for Q1 and Q2 are indicated on figure 6C-2.
FNP-FSAR-6C 6C-7 REV 21  5/08  Tests 1 to 6 and 8 to 10, table 6C-2, were conducted with and without, blockage and with water temperatures of 180&deg;F or greater and prototype discharges augmented to develop Reynolds numbers equal to or larger than prototype values. Test 7, table 6C-2, was conducted with and without blockage at prototype discharges and water temperatures of 180&deg;F. There was full prototype equivalence for this test. D. MODEL TEST RESULTS    INTAKE 1    1. General  Preliminary tests with the screen grating and Intake 1 design shown in figure 6C-8 indicated that air became trapped underneath the cover plate either during filling or upon coming out of solution due to heating. With an air pocket present, a vortex tended to form beneath the cover plate which immediately withdrew the air into the intake. To minimize the accumulation of trapped air under the plate, the modifications shown in figure 6C-5 were made. The solid cover plate of the screen structure was given a slope of 2 inches over its length and the plate was shortened 1/2 inch to provide a vent slot next to the secondary shield wall. The initial test documentation was made with the intake design of figure 6C-5, for test conditions shown in table 6C-1. These tests indicated that blockage condition 5 created flow conditions within the screen area which generated a horizontally oriented vortex which originated at the secondary shield wall inside the screen and which entered the nearest quadrant of the inlet cruciform. A further modification consisting of the grating skirt shown in Figure 6C-6 was developed to eliminate the penetration of this vortex into the intake. A final series of tests was conducted for the configuration shown in figure 6C-6. The results of the preliminary, initial and final tests are presented below. 2. Preliminary Tests  The preliminary tests were run with unblocked screens at prototype velocities and at velocities increased to produce prototype Reynolds numbers. These tests established that:    a. The proposed design could trap air under the solid cover plate which would lead to the formation of an air core vortex within the screen area that very quickly exhausted the trapped air.      b. There was no vortex formed outside of the screen structure.      c. There was no observable difference between flow patterns at prototype and augmented velocities. Hence, there was no distortion of flow patterns when departing from Froude similitude at augmented flows.
FNP-FSAR-6C 6C-8 REV 21  5/08  3. Initial Documentation Tests  The results from the initial documentation tests are as follows:    a. There were no vortices under any test condition which established an air core from the free surface within the containment area to the screen grating around the intake. Surface depressions in the eye of any eddies or organized circulation did not exceed 1/2 inch in depth.      b. Air introduced artificially under the cover plate above the intake, which represented air trapped during flooding of the containment of coming out of solution, was able to escape through the 1/2 inch vent slot in the solid cover plate for all prototype test conditions. Air could remain trapped below the plate for short periods of time. This air swirled above the intake but no air was drawn into the intake irrespective of the quantity of air forced beneath the plate.      c. No organized circulation or vortices were observed within the screen area around the intake for an unblocked screen, nor for blockage conditions 1 through 4.      Organized circulation did develop for blockage condition 5 with the strength of circulation being a function of the intake flow. The axis of circulation was horizontal and originated near the shield wall, approximately 9 inches to the left of the screen cage center. It curved into the intake quadrant nearest the left side and the shield wall. This condition could first be noticed at intake flows of about 4,000 gal/min. As the flow increased (and thus the pressure in the core of rotation decreased) an intermittent vapor core developed. Above discharges of about 4,300 gal/min a continuous vapor core 1/16 to 1/8 inch in diameter was present. This condition did not result in a measurable increase in intake head loss.      d. The maximum measured head loss across the screen and grating corrected to prototype discharge was 0.09 ft.      The intake loss coefficient was computed from the equation      KhVgVg=2222//  h = pressure drop in feet of water from inside the screen to a pressure tap down stream from the pipe inlet (5.7 feet for Intake 1)  V = average flow velocity in the 14 inch diameter pipe. The intake loss coefficient varied between 0.34 and 0.39 with no trend in the values with blockage cases or flow rates. The consistency of the loss FNP-FSAR-6C 6C-9 REV 21  5/08 coefficients indicated that no discernible flow reduction developed because of circulation or vortex action within the screen area. 4. Final Documentation Tests    An octagonally shaped grating skirt of 1 1/4-in. by 3/16-in. bars on 1 3/16-in. centers was placed around the intake within the screened area to eliminate the horizontal vortex which developed inside of the screen structure during initial tests. The intake design is shown on figure 6C-6.      Since the objectionable vortex action only occurred for blockage condition 5, final tests were only conducted for this case. The full range of test conditions, tests 1 through 8, table 1, were documented at water temperatures of 170&deg;F, or greater.      The test results were as follows:    a. A weak circulation with a horizontal axis was observed at the location where the vapor core developed during initial tests. However, no vapor core formed for any test condition. As noted previously, there were no vortices formed within the screen area for any other blockage condition.      b. The intake loss coefficient, as previously defined, remained between 0.34 and 0.39.      Additional measurements were made utilizing piezometric taps at 29 diameters downstream of the intake to further isolate the full loss of the intake and the elbow, which has a centerline radius of about 1.5 times the pipe diameter. The following formula, which includes the correction for pipe friction, was used:      g2Vg2VhhK22f=    where hf is the computed pressure drop in feet of water due to pipe friction above, based on the estimated pipe surface roughness, height, and other terms as defined before. For intake discharges from 3060 to 7915 gal/min, the loss coefficient for the intake and the elbow ranged from 0.46 to 0.48.      Intakes 2, 3, and 4    The documented tests were conducted with the improved screen and grating and grating cages placed over Intakes 2, 3, and 4 as shown in figures 6C-12 and 14.
FNP-FSAR-6C 6C-10 REV 21  5/08  The full range of test conditions, tests 1 through 10, table 6C-2, with unblocked, and blocked screens were run and documented at water temperatures of 180&deg;F or greater.      The initial test results indicated a substantially higher loss coefficient for intake plus bend for Intake 3 than for Intakes 2 and 4. The loss coefficient was computed from the equation given in subsection D.3.d, above.      An inspection revealed that the model pipe walls of Intake 3 had been severely corroded by the hot water, with the tuberculated pipe indicating protrusions measuring 1/32 to 1/16 inch. An additional pressure tap was installed to record the actual pressure loss over a 9.51-foot section of pipe. The h value of Intake 3 was then calculated from the measured pressure drop values in the pipe.      The test results were as follows:    a. There were no vortices under any test condition which established an air core from the free surface within the containment area to the screen grating around the intakes. Surface depressions in the eye of any eddies or organized circulation did not exceed 1/2 inch in depth.      b. No vortices were observed inside or outside the screen structure.      c. Air introduced artificially under the cover plate above the intake, which represented air trapped during flooding of the containment or coming out of solution, was able to escape through the 1/2-inch vent slot in the cover plate. At augmented discharges, pockets of air would remain trapped below the plate. This air swirled above the intake, but no air was drawn into the intake irrespective of the quantity of air forced beneath the plate. At prototype discharges, this air was able to escape through the 1/2-inch slot, and only a few small bubbles remained.      d. The maximum measured head loss across the screen and grating corrected to prototype discharge of 5900 gal/min for Intake 2 and 3050 gal/min for Intake 3 was 0.14 foot for Intakes 2 and 3 and 0.04 foot for Intake 4 with a discharge of 3050 gal/min.      The intake plus bend loss coefficients varied from      0.36 to 0.46 for Intake 2      0.38 to 0.45 for Intake 3      0.33 to 0.40 for Intake 4,      with no trend in the values with blockage cases or flow rates. The model indicated maximum combined prototype losses due to screen, intake, and bend of FNP-FSAR-6C 6C-11 REV 21  5/08    1.50 ft for Intake 2      1.25 ft for Intake 3      1.01 ft for Intake 4.      The higher combined losses associated with Intakes 2 and 3 were attributed to larger flow per unit area approaching the intakes and the more turbulent approach condition resulting from the proximity of these intakes to the elevator shaft. E. FIELD TEST    Several preoperational tests were performed at the Joseph M. Farley Nuclear Plant, Unit 1 to determine the actual piping resistance of the residual heat removal (RHR) pump sump suction lines. The testing revealed that the maximum expected RHR pump flow during the post-LOCA cold leg recirculation mode with only one RHR pump in operation was 5000 gal/min for Pump A and 4875 gal/min for Pump B. The actual net positive suction head (NPSH) available to each RHR pump from the containment sump was determined to be 18.4 feet at 5000 gal/min without taking credit for subcooling of the water in the containment sump and based upon the most resistive sump piping (25.2 feet elevation head from the sump and 3.9 feet of water above the sump line inlet less 10.7 feet of losses). The NPSH required for the RHR pump is 18.5 feet at 5000 gal/min and 18.0 feet at 4875 gal/min. This indicated that the NPSH available to the RHR pumps under worst case conditions would be marginal during the post-LOCA recirculation phase. Upon the completion of additional tests confirming the resistance of the installed piping system, the RHR system resistance was increased to assure that adequate NPSH is available and that system performance is satisfactory during all operating modes. The system resistance was increased by physically restricting the maximum opening of valves HCV-603A and B on the outlet piping of the RHR heat exchangers and by addition of flow restriction orifices in each of the three cold leg low head safety injection lines. System tests conducted after these modifications show that the maximum flowrate with one pump operating during the cold leg recirculation mode of operation would be approximately 4200 gal/min. The NPSH available for RHR Pumps A and B utilizing simulated recirculation mode plant test data, at this flowrate, is 17.7 feet (25.2 feet elevation head from the sump less 7.5. feet of losses) and 19.2 feet (25.2 feet elevation head from the sump less 6.0 feet of losses) respectively. The NPSH required for the RHR pump is 15.0 feet at 4200 gal/min. Thus, adequate NPSH is assured. These calculations take no credit for water above the containment sump line inlet or for any subcooling of water in the containment sump. Evaluation of the postmodification tests also confirmed that ECCS flows would meet or exceed system requirements during all operating modes. F. SUMMARY AND CONCLUSIONS    The 1:1 scale model of Intake 1, (figures 6C-5 and 6), which was tested at Reynolds numbers equal to or greater than prototype and with circulations which were greater FNP-FSAR-6C 6C-12 REV 21  5/08 than prototype, indicated that the intake will operate without air entraining or flow reducing vortices. The maximum screen grating and intake losses computed from the model test results were 0.09 foot and 0.85 foot respectively at 5900 gal/min. These values were combined with field test data and compared with calculated data used in the NPSH evaluation. The 1:1 scale model of Intakes 2, 3, and 4, (figure 6C-7), which was tested at Reynolds numbers equal to or greater than prototype and with circulations which were greater than prototype, indicated that the intake will operate without air entraining or flow reducing vortices. The maximum losses determined from the model and field tests for each intake are:    Pressure Drop (feet)  Intake      Effect 1 2 3 4      Piping (from field data)(1, 2) 5.32 3.86 6.28 6.89      Inlet (from test data)(1) 0.43 0.37 0.36 0.43          (from test data)(1) 1.49 1.49 2.39 2.39      Screen (from test data) 0.04 0.09 0.09 0.04      Total 7.28 5.81 9.12 9.75  NOTES:    1. Converted to 4200 gal/min base for Intakes 1 and 2 and to 3050 gal/min base for Intakes 3 and 4.      2. Includes additional losses due to 8 feet of test piping for Intakes 1 and 2 and additional losses due to 6 feet of test piping for Intakes 3 and 4. The measured head losses are less than the calculated losses of 8.4 feet for Intakes 1 and 2 and 9.9 feet for Intakes 3 and 4.  (See subsection 6.3.2.14.)    Based on the results of Intake 1 tests, together with Intakes 2, 3, and 4 tests of Unit 1 and on similar work undertaken for other projects, it is the definite opinion of Western Canada Hydraulic Laboratories Ltd. and Bechtel that incorporating a grating cage similar to the above design, (figures 6C-6 and 12), will result in an intake design for all Units 1 and 2 intakes which will operate free from air entraining or flow reducing vortices.
FNP-FSAR-6C 6C-13 REV 21  5/08 IV. UNIT 2 TESTS    A. INTRODUCTION    This section presents the model test program undertaken and the results of these tests to ensure that Joseph M. Farley Unit 2 emergency core cooling and containment spray system recirculation intakes from the containment sump (floor) will operate without effects which could degrade the performance of the pumps in these systems.      Similar tests were conducted for the Farley Unit 1 containment sump intakes in which the containment geometry in the sump areas was modeled at a scale of 1:1 together with flow obstructions such as pipes, supports and valves, etc. around these intakes. These tests revealed that there were no air entraining vortices or flow reducing conditions at these intakes when these intakes were protected with the inner grating cage and the outer screen grating cage structure combination. This is discussed in detail in section II of this appendix.      The tests performed on containment sump intakes of Unit 1 and on other facilities provided strong evidence that the inner grating cage and the outer screen grating cage structure combination employed on Unit 1 was totally effective in destroying vortices ranging from pencil lead size to 1 inch in diameter or greater.      Based on these results, it was concluded that the tests on Unit 2 containment sump intakes can be effectively performed without modeling the containment geometry around the intakes as was done for Unit 1. A more detailed rationale for this approach along with a description of Unit 2 containment sump intakes, discussion of effects which could degrade pump performance, description of test facility and program, and the test result and conclusions are presented in this report. B. DESCRIPTION OF UNIT 2 CONTAINMENT SUMP RECIRCULATION INTAKES    The emergency core cooling and containment spray system recirculation intakes for Unit 2 are comprised of two 14-inch and two 10-inch vertical inlets located in four separate intake areas on the containment floor and are designated as Intakes 1, 2, 3 and 4 as shown in figure 6C-19. Each intake is surrounded by a protective screen grating and grating cage structure as shown in figure 6C-17. The design flows for Intakes 1 and 2, which supply water to the RHR pumps, range from 3000 to 5900 gal/min each. The design flow rates for Intakes 3 and 4, which supply water to containment spray pumps, are 3050 gal/min each.      The calculated minimum and maximum water levels in the containment are 58.3 and 77.1 inches, respectively, above the floor.      The maximum expected containment sump water temperature during recirculation following a postulated LOCA is about 212&deg;F at subcooled pressures.      The flowrates, water depths, water temperature, and the protective screen structure for Unit 2 are identical to Unit 1, except that four separate intake areas are provided for Unit 2 as compared to three intake areas for Unit 1. The elevator shaft in Unit 2 is FNP-FSAR-6C 6C-14 REV 21  5/08 located outside the flow paths approaching the intakes. This will lead to a more uniform flow in the containment sump intake areas than that expected in Unit 1, where the elevator shaft is located in the containment sump intake area. Furthermore, the equipment layout at the Unit 2 containment floor elevation, as shown in figure 6C-20, is not expected to be significantly different from Unit 1, shown in figure 6C-21. C. PROBLEM DEFINITION    Regulatory Guide 1.82 states the position that "Pump intake locations in the sump should be carefully considered to prevent degrading effects, such as vortexing, on the pump performance." Two degrading actions are possible: ingestion of air (a vortex phenomenon), and/or intake entrance losses which are larger than design values used in establishing the required NPSH of the pumps.      Increased entrance loss can develop due to adverse flow approach conditions or free surface and internal vortex action.      1. Factors Causing Increased Entrance Losses      Intake losses are incurred due to contraction and expansion of the flow at the intake. The intake entrance losses are accounted for in the design of pumping systems by calculating the entrance loss based on established intake loss coefficients. Such coefficients are normally based upon measurements taken with uniform flow approaching the intake.      Intake head losses can be increased by high approach velocities, especially at an angle to the pipe axis and/or by strong circulation in the approach flow which results in an increased contraction of the flow at the intake.      Strong circulation can lead to vortex formation with a marked reduction in flow.      A full scale model is capable of indicating any head loss degrading effects for all conditions simulated and tested.      2. Factors Affecting Vortex Creation    Studies of vortex formation have been carried on by several investigators (see references in part II, G). The majority present test results as functions of the intake head loss coefficient, the depth of water at which the air core just penetrates the intake, the circulation numbers at which the air core just penetrates the intake, the Reynolds Number, or some variation of these parameters.      The performance of an intake, as represented by the head loss coefficient K, is usually described (Anwar 1968, Amphlett 1976, Chang 1976) as:      K = F (local geometry, rmax RR, N, W)
FNP-FSAR-6C 6C-15 REV 21  5/08    where    local geometry = f (D, h, b)    HQ.NoReynoldsRadialRR=    NCirculationNoDQWWeberNoQDh===..224    rmax = radius of the tank (or sump) in which      the intake is located = maximum radius      of circulation in the vicinity of the intake      Q    =  discharge      D    =  intake diameter      b    =  height of intake above sump floor          =  density of water          =  surface tension of water          =  surface tension of water      h    =  depth of submergence of intake            = circulation strength = 2Vtr where V  is tangential velocity at radius r. Work by Daggett and Kuelegan (1974) and others have shown that for high Reynolds numbers (RR  >104) and moderate values of circulation (N 2), typical operation ranges for the Farley recirculation intakes, the effects of surface tension and viscosity are relatively small; i.e., W and RR are not important. In this case, the intake performance, and hence the formation of vortices, is a function of three parameters:  the local geometry, the maximum circulation radius, and the strength of circulation of the approaching flow. Each of these factors is discussed in the following sections. D. TEST PROGRAM    1. Rationale      As discussed in subsection C, the intake head losses may be increased by nonuniform flow and/or circulation in the approach flow into the intakes.
FNP-FSAR-6C 6C-16 REV 21  5/08    It is significant to note that ultimately it is the flow condition in the immediate vicinity of the intake pipe that establishes the intake head loss. This condition, and associated head losses, may or may not be affected by the flow conditions removed from the immediate vicinity of the intake pipe.      With an intake that is not protected by a screen grating cage, the flow in the immediate vicinity of the intake pipe will be established by the structural configuration of the containment and affected by the presence of flow obstructions such as valves, piping and restraints. The effect of these flow obstructions will increase with their increased proximity to the intake. Structural members may channelize the approach flow, affecting the approach flow directions and velocities. Channelization can also lead to a general circulation in the vicinity of the intakes, being bounded by the surrounding structures. Eddy-shedding will induce vorticity in the flow which can add to circulation in the vicinity of the intake.      Unquestionably, an intake that is not protected by a screen grating cage can only be tested with full representation of the structural configuration of the containment and valves, piping and restraints. However, the Farley Unit 2 containment sump recirculation intakes are to be covered and protected by a screen grating cage, comprised of a 0.047 inch screen wire with an effective opening of 51.6 percent, sandwiched between two layers of grating. The grating bars will be 1-1/4 inch by 3/16 inch on 1-3/16 inch centers, giving a total effective grating width in the direction of flow of 2-1/2 inch.  (See figure 6C-17.)  The inside grating bars are approximately 2 feet from the intake pipe. Furthermore, an inner grating cage will be placed over the intake pipe.      Due to the proximity of the Farley screen grating cage to the intake pipe, it was concluded that this structure would strongly influence, if not dominate, the approach flow into the intake pipes within the cage. This dominance was observed on full scale mockup tests of the Farley Unit 1 containment sump recirculation intakes where:      a. The grating bars acted as flow straighteners and no angularity or circulation of flow approaching the screen cage, which could lead to the formation of a vortex, was transmitted through the structure, regardless of the angle of approach flow. Flow downstream from the grating exited at right angles to the plane of the grating.      b. The most nonuniform, rotational approach flow to the intake pipe, as evidenced by an air core vortex inside the cage, was developed by a  partial screen blockage configuration. No vortex developed inside the screen cage without blockage.      Hence, it was apparent from the Unit 1 tests that the approach conditions in the immediate vicinity of the intakes (within the screen grating cage) were established by the flow distribution through the screen grating cage. Angularity of flow approaching the outside of the screen grating was removed and any swirl or circulation inside the screen grating cage was due to a nonuniform flow FNP-FSAR-6C 6C-17 REV 21  5/08 distribution through and normal to the plane of the grating. Furthermore, the most adverse velocity distribution inside the screen grating cage could be established by the blockage configuration imposed. These observations lead to the following considerations:      a. In the case of an intake not protected by a screen grating cage, the flow can be channelized at any angle by structural members. In the case of an intake covered by a screen grating cage, the screen grating and the blockage configuration impose the ultimate channelization, and establish the direction of flow normal to the plane of the grating.      b. Irrespective of the structural configuration external to the screen grating cage, and hence irrespective of the approach flow conditions this configuration imposes on an unblocked screen grating structure, there will be a blockage condition which will develop as adverse or a more adverse and potentially a more degrading effect on the intake performance. Hence, this proves that the grating gage will eliminate any vortex potential and would be proof that the potential developed by the external structural arrangement will be eliminated.      c. Because blockage conditions could establish potentially degrading conditions inside the screen grating cage, a grating cage must be incorporated inside the screen grating cage to remove circulation generated within the screen grating cage which could lead to the formation of vortices and/or increased intake head losses.      Thus, based on the experience gained on the full scale mockup tests for Unit 1, the following rationale was applied to the test program for the Unit 2 containment sump recirculation intakes:      a. Ultimately it is the flow condition in the immediate vicinity of the intake pipe that can lead to degrading effects of pump performance.      b. The immediate vicinity of the intakes will be covered by a screen grating cage.      c. If the screen grating cage does not transmit the angularity of circulation of flow outside of the cage, then flow conditions and air core vortex potential within the screen grating cage are established by the blockage conditions imposed (flow distribution), water depth (pressure inside the cage), intake discharge (velocities), and viscosity (fluid shear energy dissipation).      d. The fact the containment may be pressurized does not affect flow conditions. The flow field is established by pressure differentials which would be the same in a closed system irrespective of the air pressure on the water surface.
FNP-FSAR-6C 6C-18 REV 21  5/08    e. If angularity of approach flow is not transmitted through the screen grating cage, then the uniqueness of the flow distribution through an open screen established by the structural configuration and flow obstructions surrounding the screens represents one potential blockage condition.      f. Based on the above, it is necessary to model only the screen grating cage, and all features inside the cage, and demonstrate for postulated flow depths and flow rates that:      i. The screen grating cage will not transmit the angularity or circulation of flows outside the cage.        ii. Under adverse conditions generated by screen blockage, the grating cage over the intake inside the screen grating cage will preclude degrading effects on the performance of the recirculation pumps.      g. Furthermore, since circulation is an essential and necessary feature of a vortex, then irrespective of the strength of circulation, if flow circulation associated with a potential vortex is not transmitted through the screen grating cage, then the vortex formed outside of the cage cannot enter the intake pipe (as discussed in the Final Report on the Davis-Besse Nuclear Power Station ECCS Emergency Pumps and Pump Suction Line Testing, December 15, 1976).      As discussed in subsection C, the formation of vortices is a function of three parameters: the local geometry, the maximum circulation radius, and the strength of circulation of the approaching flow.      Since full scale tests were to be conducted, the local geometry in the immediate vicinity of the intake would be correctly simulated. In addition, since all screen and grating characteristics would be correctly represented, all vortex and flow parameters from the screen grating structure inward to the intake would be correctly simulated, and the intake entrance losses would be correctly measured.      Swirls in the approach flow may vary with respect to the absolute size of the system, strength of circulation, velocity of translation, and travel path. The latter two parameters are of significance since, for a swirl to initiate an intake vortex, it must remain in the vicinity of the intake long enough to organize the circulation in the vicinity of the intake. Hence a stationary circulation directly above the intake becomes the critical case. The system size is of no concern when a 1:1 scale model is used. Thus there are two parameters which must be properly addressed:  the maximum circulation radius (rmax), and the strength of circulation. Experimental evidence indicates that the critical submergence of the intake required to preclude the formation of air entraining vortices increases with both the maximum swirl radius, rmax, (Haindl, 1959) and strength of the initiating swirl (Amphlett, 1976), Daggett and Kuelegan, 1974; Springer & Peterson, 1969; Anwar, 1965).
FNP-FSAR-6C 6C-19 REV 21  5/08    From these experimental data, it can be concluded that a test procedure should include:      a. A stationary circulation directly over the intake.      b. A circulation strength equal to or greater than the largest reasonable values due to the  expected prototype approach flow configuration.      c. A maximum circulation radius, rmax, equal to the largest reasonable value in the prototype, must be said to "bound" the effects developed y the plant geometry and structural members, etc., in the vicinity of the intakes which could lead to a vortex.      2. Objective    The prime objective of the test program was to demonstrate that the Farley Unit 2 containment sump recirculation intakes will not be subjected to degrading effects on pump performance, such as air ingestion or high intake head losses.      Achieving the following fulfilled the prime objective:      a. Documentation of the effectiveness of the grating cage over the intake in straightening the approach flow and removing imposed angularity or circulation which, without the grating cage present, could lead to an air entraining vortex.      b. Documenting that the screen grating cage removed angularity and circulation of approach flow outside of the cage.      Documentation of the effectiveness of the grating cage was achieved by:      i. Imposing on the grating cage, without the screen grating cage over it, a range of circulations, the largest of which was more massive than any circulation that could be developed by the geometry or the structural members of the containment or the presence of flow obstruction such as valves, piping and restraints.        ii. Imposing blockage conditions on the screen grating cage which generated potentially degrading flow conditions within the screen cage, and documenting that those conditions were eliminated by the grating cage.        Documentation of the effectiveness of the screen grating cage was achieved by:      i. Demonstrating that the single layer of grating on the grating cage was effective in removing angularity in the approach flow in the high velocity region close to the intake.
FNP-FSAR-6C 6C-20 REV 21  5/08    ii. Demonstrating that no vortex, angularity or circulation of approach flow passed through the screen grating cage or grating cage when the screen grating cage was subjected to a range of circulation, the largest of which was more massive than any circulation that could be developed by the geometry or the structural members of the containment, or by the presence of flow obstructions such as valves, piping and restraints. E. TEST FACILITY    1. General      The plan view and a section of the experimental facility are shown in figures 6C-15 and 6C-16. Two source sumps, each containing a diffuser, provided the approach flow to the intake area within the concrete tank, (figure 6C-15). A sump floor, which was of 1/8 in. steel plate, was placed 4.5 feet above the tank floor to provide space for the 14 in. diameter intake piping and for an observation tunnel below the steel plate floor.  (See figure 6C-16.)      The flows were distributed and controlled by means of two centrifugal pumps and a flow transmitting network of steel pipes, orifice meters with differential mercury manometers, and control valves. The direction and circulation of the approach flow was controlled by a system of 18-in.-wide vertical directional vanes, which extended over the full depth of the flow.  (See figure 6C-16.)      Two 2,900,000 Btu/h gas heaters were used to heat the water to temperatures in excess of 180&deg;F.      2. Intake Description      A cruciform and reducer section, which was shipped from the project site for use in the experimental facility, was mounted on the intake pipe 6 in. above the steel plate floor. The octagonal grating cage used for the Unit 1 model tests was modified to include a horizontal grating inside the grating gage, 3 in. above the floor, to eliminate potential floor vortices.  (See figure 6C-18.)  The grating cage thus totally encapsulated the intake pipe.      The whole assembly was enclosed by a steel screen grating cage with inside dimensions of 5-ft. x 5-ft. x 2-ft. 5-in. depth.  (See figure 6C-17.)  The floor below the screen grating was of acrylic plastic construction which, together with the portholes in the observation tunnel, permitted observation and lighting of the area inside the screen grating.      3. Test Cases and Procedures      a. Test Cases      The postulated post-LOCA condition and the condition for which the intake was tested are compared below:
FNP-FSAR-6C 6C-21 REV 21  5/08  Postulated for  Containment Sump  post-LOCA            Tested    Minimum water depth (in.) 58.3 24.0 to 58.3 Maximum flow (gal/min) 5900 6574 to 8524 Water temperature (&deg;F) 212 61 to 184 Maximum circulation (ft2/s) 3./for 58.3 in. 8.5 to 10.7  water depth for 58.3 in. water depth Maximum size of 17 18 circulation cell (ft)  Pipe Reynolds Number 4.7 x 106 1.5 to 5.7 x 106  Considerable conservatism was incorporated in the test by:      i. Conducting tests at greater than postulated flow rates.      ii. Conducting tests with screen blockage greater than 50 percent. iii. Conducting tests at less than the minimum postulated water depths.      iv. Conducting tests with a circulation appreciably greater than the maximum value calculated for the plant during LOCA conditions.      v. Augmenting the postulated flows to develop Reynolds numbers in the test facility greater than postulated in the containment.      Furthermore, model scale effects were reduced or eliminated by:      i. Constructing the intake, grating cage, and screen grating cage at a 1:1 scale, thereby eliminating all scale effects introduced by modeling the screen and grating components.      ii. Conducting the tests with water heated to 180&deg;F or greater.      b. Observations and Measurements      All surface flow phenomena were observed from two platform decks. The lower deck was used to make all surface flow observations and to take velocity and temperature measurements. Observations could be made of flow phenomena inside the screen grating through the acrylic plastic cover plate. Video records were also made from this deck. Overhead photos were made from the upper platform deck.        Flow phenomena within the screen grating cage could be observed and recorded on video tape through the portholes in the observation tunnel.
FNP-FSAR-6C 6C-22 REV 21  5/08 Use was made of air bubbles injected into the screen grating area through the acrylic plastic floor for flow visualization. Dyes were used sparingly to preserve water clarity.        Flow measurements were obtained with the calibrated orifices and U-tube mercury manometers. Local velocities were measured with a Gurley propeller meter while surface velocities were obtained with the Gurley meter or from overhead photos of confetti traces.        Pipe intake and screen grating losses were determined from static pressure taps. Taps 1 and 2 were located in the supply sumps and indicated the water surface elevation. Tap 3 consisted of two interconnected taps in the floor inside the screen grating cage to produce an average pressure within the screened area. The mean static head indicated by Tap 3 therefore gave an indication of screen grating losses when compared to the mean water surface elevation from Taps 1 and 2. Two interconnected taps, each on the horizontal diameter, determined the average static pressure inside the intake pipe at each of four locations, at distances of 5.11, 13.11, 21.11, and 30.00 diameters downstream of the intake.      c. Test Procedure      The tests were conducted in two phases.        Phase 1 tests were related to documenting the effect of grating on approach flow conditions external to the screen grating cage.        For this series, without any screen or grating over the intake, given flows were set and the vanes surrounding the intake were adjusted to produce the maximum size vortex. The circulation, vortex size and pressure measurements were taken, together with observations of flow conditions in the immediate vicinity of the intake. This was done for both ambient and heated water. Tests at ambient water temperatures were conducted to facilitate the making of video records of the free surface flow conditions.        Without changing the vane angle, the tests were rerun and the results were documented with the addition of:        i. Cruciform only.        ii. Grating cage and cruciform only.        iii. Screen grating cage and cruciform only.      iv. Screen grating cage, grating cage and  cruciform.
FNP-FSAR-6C 6C-23 REV 21  5/08    Phase 2 tests were related to documenting the effect of the grating cage on adverse flow conditions generated within the screen grating cage.        With the screen grating cage and cruciform in  place, blockage was placed on the screen to produce the largest internal vortices achievable. Pressure measurements were then taken and observations made. The grating cage was then installed and data were recorded for the identical conditions which previously had produced internal vortices.        In summary, test procedures were developed for the Farley Unit 2 recirculation takes which:        i. Modeled all effects of the screen grating cage and grating cage in the immediate vicinity of the intake on a 1:1 basis.        ii. Allowed testing for the effects of the containment geometry and structural members, etc., by subjecting the intake to a range of circulation, the largest of which was greater than will occur in the prototype approach flow.        iii. Demonstrated satisfactory intake performance under unrealistically severe conditions of water depth and circulation.      F. TEST RESULTS    1. Phase 1 Test Results    a. Unprotected Intake      An air entraining vortex was easily formed over the unprotected intake pipe. With a water depth of 58 in. and an intake flow of approximately 7400 gal/min, the air entraining vortex was present intermittently when the flow vanes were aligned radially to the intake. The vortex increased in strength and became stable as the vanes were turned from the radial direction. The maximum vortex occurred with the vanes turned 48 degrees in either direction. The air core diameter of the vortex at the intake was 1.5 to 2 in. with a circulation of 8.5 ft.2/s as compared to a maximum calculated prototype value of 5.4 ft.2/s.        With the vane angle set at 48 degrees, reducing the intake discharge from approximately 7400 gal/min to approximately 5300 gal/min reduced the diameter of the air core at the intake to 1 to 1.25 in.      b. Intake with Cruciform      The cruciform, by itself, did not eliminate air entraining vortices with a water depth of 58 in. The vortices were not as stable as without the FNP-FSAR-6C 6C-24 REV 21  5/08 cruciform; nevertheless, the following air core sizes were observed at the intake:    Water Vane Air Core Flow Temp. Angle Diameter gal/min &deg;F &deg;  in. 7344 to 8457  65 48 1/2 to 3/4    7412 to 8088  65  0 1/8 to 3/8    7018 to 8446 180 48 3/4 to 1-1/2      The circulation for a flow of 8456 gal/min was 9.1 ft2/s.        The intake loss coefficient, K, was 0.69.        The maximum intake loss coefficient, K, for the heated water was 0.73 and the average intake loss coefficient was 0.72.      c. Intake and Cruciform Protected by the Grating Cage      No air entraining vortex penetrated the grating cage with the vanes set at 48 degrees and a water depth of 58 in. The intake flows tested were 7420 gal/min to 8513 gal/min with a water temperature of 119&deg;F to 181&deg;F and 8487 gal/min with a water temperature of 64&deg;F. The flow circulation established by the vanes remained around the grating cage with the water surface depressed approximately 1 in. at the center. Bubbles or particulates in the flow surrounding the cage, which approached at an angle to the cage, were observed to exit at right angles to the plane of the grating.        The average intake loss coefficient was reduced from 0.72 with only the cruciform to 0.65 with the grating cage. The maximum intake loss coefficient was 0.66.        With an intake flow of 8400 gal/min, no air entraining vortex was produced when the water level was lowered from 58 inches to 24 inches.      d. Intake and Cruciform Protected by Screen Grating Cage      No air entraining vortex penetrated the screen grating cage for flows of 6574 to 8487 gal/min, vane angles of 0&deg; and 48&deg;, a water depth of 58 in., and water temperatures of 61&deg;F to 64&deg;F, and 173&deg;F to 180&deg;F. The maximum circulation was 10.7 ft.2/s.
FNP-FSAR-6C 6C-25 REV 21  5/08    The circulation outside of the screen-grating cage was not transmitted through the structure.      e. Intake and Cruciform Protected by a Grating Cage and Screen Grating Cage      No air entraining vortex penetrated the screen grating cage for flows of 7741 gal/min to 8460 gal/min, vane angle of 48&deg;, a water depth of 58 in., and water temperatures of 64&deg; F and 184&deg;F.        There was no organized circulation inside the unblocked screen grating cage.        The maximum intake loss coefficient was 0.67 and the average coefficient was 0.66.        The maximum screen head loss coefficient Ks with or without blockage was 10.2.        The Ks values indicated a decreasing trend with increasing screen Reynolds number.      2. Phase 2 Tests    The following summarizes the results of the Phase 2 tests:      a. Intake and Cruciform Protected by Screen Grating Cage by Without Grating Cage      Organized circulation could be established within the screen grating cage by selective blockage of the screen.        Air core vortices were established by screen blockage of 61 to 71 percent for intake flows  of 7461 gal/min to 8420 gal/min and water temperatures of 150&deg;F to 177&deg;F. The water depth was 58 in.        Internal vortices could be formed from the floor, inside blockage plates (i.e., simulated walls), and the cover plate on the screen grating cage. One to five vortices could be generated simultaneously depending upon the blockage condition. A smooth surface within the screen grating cage was required to form an internal vortex.        The air core diameter of the internal vortices varied from 1/8 in. to 1/4 in.      The average intake loss coefficient was 0.69 and the maximum coefficient was 0.78. Internal vortices did not increase intake losses.
FNP-FSAR-6C 6C-26 REV 21  5/08    b. Intake and Cruciform Protected by Screen Grating Cage and Grating Cage      Installation of the grating cage over the intake and cruciform completely eliminated all the internal vortices previously generated by the screen blockage and flow conditions discussed in subsection F.2-a.        Flow circulation between the screen grating cage and grating cage, generated by the blockage, was not transmitted through the grating cage as evidenced by observing particulates in the flow.        The average intake loss coefficient with the screen cage, grating cage and cruciform in place was 0.66 and the maximum coefficient was 0.67.      G. SUMMARY AND CONCLUSIONS    The recirculation intake designs to be used for Farley Unit 2 were tested under flow and vortex producing conditions which were potentially more degrading on pump performance than any condition possible in the prototype. The screen grating and inner grating cage were modeled at a 1:1 scale. The following results were obtained:    1. Vortex Action    The screen grating cage will not permit any free surface air entraining vortices to form through which air can be ingested by the intake. Circulation (which is an essential feature of a vortex), or approach flow angularity, were not transmitted through the screen grating cage. The grating used in the screen grating cage was totally effective in eliminating any vortex with air core diameters of 1/8 in. to 2 in., which would have otherwise formed without the presence of the screen grating cage.      Without the inner grating cage, internal vortices could be developed by selective screen blockage. These vortices, which were formed only from smooth surfaces, did not increase intake entrance losses. However, with the grating cage in place as proposed for the Farley Unit 2 design, no internal vortices will develop. Circulation developed within the screen grating cage, which could lead to internal vortices, was not transmitted inside of the grating cage.      2. Head Loss Coefficients    The screen grating cage, grating cage, and cruciform protective design will have a head loss  coefficient for the combined grating cage, intake and 90&deg; pipe bend of 0.67, even with screen blockages in excess of 50 percent.      The maximum measured intake loss coefficients were as follows:      Cruciform    0.73 FNP-FSAR-6C 6C-27 REV 21  5/08    Grating cage and cruciform    0.66    Screen grating cage, grating cage    and cruciform    0.67    Screen grating cage head losses are small with the maximum measured loss coefficient in the model being 10.2.      3. Losses    The maximum losses determined from the model test and calculations for each intake are:  Effect Pressure Drop (feet)  Intake  1 2 3 4      Piping (calculated)(1, 2) 5.89 4.27 6.47 7.09      Inlet (from test data)(1) 0.75 0.75 1.02 1.02        (From test data)(1) 1.48 1.48 2.40 2.40      Screen (from test data) 0.09 0.03 0.03 0.05      Total 8.21 6.53 9.92 10.56    NOTES:    1. Converted to 4200 gal/min base for Intakes 1 and 2 and to 3050 gal/min base for Intakes 3 and 4. 2. These are calculated numbers and will be verified by a field test. However, comparison of the calculated values with the field test data for Unit 1 indicates that the calculated values are conservative (See section 6C.III.F).
FNP-FSAR-6C 6C-28 REV 21  5/08 REFERENCES    1. Addison, H., 1948; "Centrifugal and Other Rotodynamic Pumps," (Chapman and Hall, London.)  2. Akers and Crump, "The Vortex Drip," Journal, Institution of Civil Engineers, p 443, August 1960. 3. Al'Tshul, A. D. and Margonlin, M. S., "Effect of Vortices on the Discharge Coefficient for Flow of a Liquid Through an Orifice" (Translation), Gidrotekhnieheskoe Stroitel'stvi No. 6, p 32, June 1968. 4. Amphlett, M. B., "Air Entraining Vortices at a Horizontal Intake," Report No. OD/7, Hydraulic Research Station, Wallingford, April 1976. 5. Anwar, H. O., "Flow in a Free Vortex", Water Power, April 1965. 6. Anwar, H. O., "Formation of a Weak Vortex", Journal of Hydraulic Research, Vol. 4, No. 1, 1966. 7. Anwar, H. O., "Vortices at Low Head Intakes", Water Power, pp 455-457, November 1967. 8. Anwar, H. O., "Prevention of Vortices at Intakes," Water Power, p 393, October 1968. 9. Anwar, H. O., Discussion of "Effect of Viscosity on Vortex-Orifice Flow," by Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulics Division, ASCE Vol. 95, No. HY 1, Proc. Paper 6323, pp 568-570, January 1969. 10. Berge, J. P., "Enquete sur la Formation de Vortex et Autres Anomalies d'ecoulement dans une enceinte avec ou sans Surface Libre," Societe Hydrotechnique de France-Section Machines - Group de travail No. 10, November 1964. 11. Berge, J. P., "A Study of Vortex Formation and Other Abnormal Flow in a Tank With and Without a Free Surface," La Huille Blanche, Grenoble, France, No. 1, pp 13-40, 1966. 12. Binnie, A. M., and Hockings, G. G., "Laboratory Experiments on Whirlpools," Proceedings, Royal Society, London, Series A. Vol. 194, pp 398-415, September 1948. 13. Brewer, D. "Vortices in Pumps Sumps," The Allen Engineering Review, March 1957. 14. "Review of Literature on Drain Vortices in Cylindrical Tanks," Report TN1342, BHRA, March 1976  15. Denny, D. F., 1953 British Hydromechanics Research Assoc. Report, R. R. 430, Preliminary Report on the Formation of Air entraining Vortices in Pump Section Wells. 16. Denny, D. F. "Experiments with Air in Centrifugal Pumps", British Hydromechanics Research Assoc. Research Report R. R. 465, 1953. 17. Denny, D. F., and Young, G. A. J. "The Prevention of Vortices and Swirl and Intakes", Proceedings, IAHR 7th Congress, Libson, 1957.
FNP-FSAR-6C 6C-29 REV 21  5/08 18. Denny, D. F., "An Experimental Study of Air Entraining Vortices in Pump Sumps", Proceedings of Inst. of Mechanical Engineers, Vol. 170, No.2, 1956. 19. Daggett, L. L. and Keulegan, G. H., "Simulitude Conditions in Free Surface Vortex Formations", Journal of the Hydraulics Division, ASCE, Vol. 100, No. Hy11, Nov. 1974, pp 1565-1581. 20. Donaldson, C. du p., and Sullivan, R. D., "Examination of the Solutions of the Navier-Stokes Equations for a Class of Three-dimensional Vortices, Part I: Velocity Distribution for Steady Motion", Proceedings, Heat Transfer and Fluid Mechanics Institute, Stanford University Press, Calif., 1960, p. 16-30. 21. Einstein, H. A., and Li, H., "Steady Vortex Flow in a Real Fluid," La Huille Blanche, Vol. 10, No. 4, Aug.-Sept., 1955, p. 483-496. 22. Folsom, R. G., "Some Performance Characteristics of Deepwell Turbine Pumps," Technical Memo. No. 6, HP-14, University of California, Pump Testing Laboratories, 1940. 23. Fraser, W. H., "Hydraulic Problems Encountered in Intake Structures of Vertical Wet-Pit Pumps and Methods Leading to Their Solution," Trans. ASME, vol. 75, No. 4, p. 643,  1953. 24. Gordon, J. L., "Vortices at Intakes," Water Power, April 1970, p. 137-138. 25. Cuiton, P., "Cavitation dans les Pompes", La Huille Blanche, Nov., 1962, No. 6. 26. Haindl, K., "Contribution to Air-Entrainment by a Vortex", Paper 16-D International Association for Hydraulic Research, Montrel, 1959. 27. Hattersley, R. T., "Hydraulic Design of Pump Intakes", HY 2, March, 1965, P. 223-249. 28. Hattersley, R. T., "Factors of Inlet Channel Flow affecting the Performance of a Pumping Plant", Report No.23, Water Research Lab., University of New South Wales, Australia, Sept., 1960. 29. Holtorff, G., "The Free Surface and the Conditions of Similitude for a Vortex", La Houille Blache, vol. 19, No. 3, 1964, P. 377-384. 30. Iversen, H. W.,; "Studies of Submergency Requirements of High Specific Speed Pumps", Transactions, ASME, Vol. 75, 1953. 31. Kaufman, Fluid Mechanics, McGraw-Hill, p. 265 and 279. 32. Keulegan, G. H., and Daggett, L. L., "A Note on Gravity Head Viscometer", Miscellaneous Paper H-74-3, United States Army Engineer Waterways Experiment Station, Corps of Engineers, Vicksburg, Miss., Mar., 1974. 33. Kolf, R. C., "Vortex Flow from Horizontal Thin Plate Orifices", thesis presented to the University of Wisconsin, at Madison, Wis., in 1956, in partial fulfillment of the requirements for the degree of Doctor of Philosophy.
FNP-FSAR-6C 6C-30 REV 21  5/08 34. Kolf, R. C., and Zielinski, P. B., "The Vortex Chamber as an Automatic Flow Control Device", Journal of the Hydraulics Division, ASCE, vol. 85, No. HY12, Proc. Paper 2310, Dec. 1959. 35. Lawton, F. L. "Factors Influencing Flow in Large Conduits", Report of the Task Force on Flow in Large Conduits of the Committee on Hydraulic Structures. Transactions, ASCE, Paper 4543, vol. 91 HY 6, November, 1965. 36. Lennart, R., "Flow Problems with Respect to Intakes and tunnels of Swedish Hydro Electric Power Plants", Transactions, of the Royal Institute of technology, Stockholm, Sweden, NR 71, 1953. 37. Lewellen, W. S. "A Solution for Three-Dimensional Vortex Flow with Strong Circulation," J. Fluid Mechanics, vol. 14, 1962. 38. Long, R. R. "A Vortex in an Infinite Fluid," Journal of Fluid Mechanics, vol. 11. 39. Marklund, E. and Pipe, J. A. "Experiments on a Small Pump Suction Well, with Particular Reference to Vortex Formations". Proceedings, The Institution of Mechanical Engineers, vol. 170, 1956. 40. Marklund, E., discussion of "Effect of Viscosity on Vortex-Orifice Flow", by Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulic Division, ASCE, vol. 95, No. HY1, Proc. paper 6323, Jan., 1969, p. 567-568. 41. Messina, J. P., "Periodic Noise in Circulating Water Pumps", Power, Sept. 1971, p. 70-71. 42. McCorquodale, J. A., discussion of "Effect of Viscosity on Vortex-Orifice Flow," By Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulics Division, ASCE, vol. 95, No. HY1, Proc. Paper 7323, Jan., 1969, p. 567-568. 43. McCorquodale, J. A., "Scale Effects in Swirling Flow", Journal of the Hydraulics Division, ASCE, vol. 94, HY1, Disc. by Marco Pica, HY1, Jan. 1969. 44. Pickford, J. A., and Reddy, Y. R. "Vortex Suppression in a Stilling Pond Over Flow", Journal of the Hydraulics Division, ASCE, vol. 100, No. HY11, Nov. 1974, pp 1685-1697. 45. Quick, M. C., "A Study of the Free Spiral Vortex", thesis presented to the University of Bristol, England, in 1961, in partial fulfillment of the requirements for the degree of Doctor of Philosophy. 46. Quick, M. C., "Scale Relationships between Geometrically Similar Free Spiral Vortices", Civil Engineering and Public Works Review, Part 1, September, 1962, Part II, October, 1962, p. 1319. 47. Quick, M. C., "Efficiency of Air Entraining Vortex Formation at Water Intake", Journal of the Hydraulics Div., ASCE, vol., No. 96, HY7, July 1970, p. 1403-1416. 48. Reddy, Y. R., and Pcikford, J. A., "Vortices at Intakes in Conventional Sumps", Water Power, March 1972,p. 108-109 FNP-FSAR-6C 6C-31 REV 21  5/08 49. Rouse, H., and Hsu, H., "On the Growth and Decay of a Vortex Filament", Proceedings, 1st National Congress of Applied Mechanics, 1952, p. 741-746. 50. Richardson, C. A., "Submergence and Spacing of Suction Bells," Water Works and Sewerage Reference and Data, Part 1, Water Supply, p. 25, 1941. 51. Springer, E. K., and Patterson, I. M., "Experimental Investigation of Critical Submergency for Vortexing in a Vertical Cylindrical Tank," ASME Paper 69-FF-49, June 1969. 52. Stepanoff, A. J., 1946 Centrifugal and Axial Flow Pumps, p. 963 (Chapman and Hall, London). 53. Stevens, J. C., discussion of "The Vortex Chamber as an Automatic Flow Control Device", by R. C. Kolf and P. B. Zielinski, Journal of the Hydraulics Division, ASCE, vol. 86, No. HY6, Proc. Paper 2525, June, 1960. 54. Stevens, J. C., and Kolf, R. C., "Vortex Flow Through Horizontal Orifices", Journal of the Sanitary Engineering Division, ASCE, Vol. 83, No. SA6, Proc. Paper  1461, Dec., 1957. 55. Weltmer, W. W. "Proper Suction Intakes Vital for Vertical Circulating Pumps," Power Engineering, vol. 54, No. 6, p. 74, 1950. 56. Young, C.A.H., "Swirl and Vortices at Intakes", Report No. SP 726, British Hydro-Mechanics Research Association, April, 1962. 57. Zelinski, P. B. and Vellimont, J. R., "Effect of Viscosity on Vortex Orifice Flow", vol. 94, HY3, May, 1968, p. 745-751. Disc. on above in Jan. 1969, by Marklund, E., McCorquodale, J. A. and Anwar, H.O.  ]
FNP-FSAR-6C  REV 21  5/08  [Historical]  [TABLE 6C-1  TEST CONDITIONS FOR UNIT 1 INTAKE 1    Water  Discharges - gal/min      Test Depth Q1 Intake 1 Q2 Operating Water Temperature No. In. Prototype Model PrototypeModel PrototypeModel Pumps Prototype Model            1 58.3 3715 5500 4150 6150  -435 -644 1RHR 240 170+            2 77.1 3540 5240 4150 6150  -610 -905 1RHR 240 170+            3 58.3 5500 8140 3000 4440  2500 3700 2 RHR 240 170+            4 77.1 5500 8140 3000 4440  2500 3700 2RHR 240 170+            5 77.1 8750 12,450 4150 6150  4600 6808 1RHR,2S 240 170+            6 77.1 10,500 14,300 5900 8000  4600 6200 2RHR,2S(a) 240 183+            7 77.1 10,600 10,600 4150 4150  6450 6450 2RHR,1S 170 170+            8 77.1 12,900 12,900 4150 4150  8750 8750 2RHR,2S 170 170+                                    RHR = Residual Heat    S = Spray  a. The two RHR pumps taking suction from one inlet]
FNP-FSAR-6C  REV 21  5/08  [Historical]  TABLE 6C-2  TEST CONDITIONS FOR UNIT 1 INTAKES 2, 3, AND 4        ]
REV 21  5/08 [TYPICAL ARRANGEMENT OF CONTAINMENT SUMP SUCTION LINE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-1]
REV 21  5/08 [MODELED AREAS OF ECCS INTAKES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-2]
REV 21  5/08 [HYDRAULIC MODEL PLAN FOR INTAKE NO. 1 TESTS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-3]
REV 21  5/08 [BLOCKAGE TEST CONDITIONS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-4]
REV 21  5/08 [NO. 1 INTAKE CONFIGURATION FOR INITIAL TESTS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-5]
REV 21  5/08 [IMPROVED DESIGN INTAKE NO. 1  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-6]
REV 21  5/08 PLAN OF MODELED AREA CONTAINING INTAKES  2, 3, AND 4  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-7 REV 21  5/08 [CONTAINMENT SUMP  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-8]
REV 21  5/08 [PHOTOGRAPH OF MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-10]
REV 21  5/08 [PHOTOGRAPH OF MODEL  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-11]
REV 21  5/08 [INTAKES 2, 3, AND 4 IMPROVED DESIGN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-12]
REV 21  5/08 [BLOCKAGE TEST CONDITIONS FOR INTAKES 2, 3, 4  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-13]
REV 21  5/08 [PHOTOGRAPH OF GRATING CAGE OVER INTAKE 2  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-14]
REV 21  5/08 [PLAN OF UNIT 2 TEST FACILITY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-15] 
REV 21  5/08 [SECTION OF UNIT 2 TEST FACILITY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-16]
REV 21  5/08 [REPRESENTATIVE SCREEN-GRATING STRUCTURE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-17]
REV 21  5/08 [GRATING CAGE - FINAL DESIGN  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-18] 
REV 21  5/08 [SUMP AREA OF UNIT 2  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-19] 
REV 21  5/08 [COMPOSITE DRAWING OF UNIT 2 SUMP  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-20]
REV 21  5/08 [COMPOSITE DRAWING OF UNIT 1 SUMP  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-21]
REV 21  5/08 [PHOTO OF UNIT 2 GRATING CAGE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-22]
REV 21  5/08 [PHOTO OF REPRESENTATIVE SCREEN - GRATING CAGE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-23]
FNP-FSAR-6D 6D-i REV 22  8/09 APPENDIX 6D  CONTAINMENT SUMP DESCRIPTION AND  EMERGENCY CORE COOLING SYSTEM RECIRCULATION SUMP  STRAINER DESIGN  TABLE OF CONTENTS Page  6D.1 CONTAINMENT SUMP DESCRIPTION -----------------------------------------6D-1 6D.1.1 General Plant System Description--------------------------------------------6D-1
6D.1.2 General Description of New ECCS Strainers Installed------------------6D-2  6D.1.3 Size of New ECCS Strainers Installed---------------------------------------6D-2
6D.2 SUMMARY DESCRIPTION OF APPROACH USED TO SIZE SUMP STRAINERS--------------------------------------------------------------------------------6D-3  6D.2.1 Containment Walkdown---------------------------------------------------------6D-3 6D.2.2 Pipe Break Characterization---------------------------------------------------6D-3
6D.2.3 Debris Generation-----------------------------------------------------------------6D-3  6D.2.4 Latent Debris Accumulation within Containment--------------------------6D-4
6D.2.5 Debris Transport to the Sump-------------------------------------------------6D-4
6D.2.6 Head Loss as a Result of Debris Accumulation---------------------------6D-4  6D.2.7 Debris Source Term Reduction------------------------------------------------6D-5
6D.2.8 Sump Structural Analysis-------------------------------------------------------6D-5 6D.2.9 Upstream Effects of Debris Accumulation----------------------------------6D-5 6D.2.10 Downstream Effects - Components and Systems -----------------------6D-6 6D.2.11 Downstream Effects - Fuel and Vessel-------------------------------------6D-6  6D.2.12 Chemical Effects ------------------------------------------------------------------6D-6 REFERENCES--------------------------------------------------------------------------------------6D-8 
FNP-FSAR-6D 6D-ii REV 22  8/09 LIST OF TABLES  6D-1 Containment Sump Debris Generation Zone of Influence (ZOI)
6D-2 Summary of LOCA Generated Insulation Debris Inside ZOI  6D-3 Debris Generated from Coating Based on ZOI = 4D
6D-4 Latent and Foreign Material Debris used in Analysis 6D-5 Summary of Debris Generated and Transported to Strainer Modules 
FNP-FSAR-6D 6D-iii REV 22  8/09 LIST OF FIGURES  6D-1 Farley Unit 1 Strainer Layout
6D-2 Farley Unit 2 Strainer Layout  6D-3 Vertical Strainer Type
6D-4 Horizontal Strainer Type 6D-5 Postulated Break Locations 6D-6 Typical Arrangement of Containment Sump Suction Line FNP-FSAR-6D 6D-1 REV 22  8/09 APPENDIX 6D  CONTAINMENT SUMP DESCRIPTION AND  EMERGENCY CORE COOLING SYSTEM RECIRCULATION SUMP  STRAINER DESIGN  6D.1 CONTAINMENT SUMP DESCRIPTION 6D.1.1 GENERAL PLANT SYSTEM DESCRIPTION Farley Nuclear Plant (FNP) Units 1 and 2 are Westinghouse three loop Pressurized Water Reactor (PWR) design. The residual heat removal system (RHRS) (low head safety injection), centrifugal charging system (CVCS) (high head safety injection), and containment spray system (CSS) pumps are started following a loss of coolant accident (LOCA). Initially, two RHR, two CVCS, and two CCS pumps take suction from the refueling water storage tank (RWST). When the RWST level reaches the low level setpoint, the RHR pumps are manually stopped and are realigned to take suction from the post-LOCA containment sump. Once the RHR switchover to recirculation is complete, the CVCS pumps take suction from the RHR pump discharge. 
When the RWST level reaches low-low level, the CSS pumps are realigned to take suction from the containment sump. There are four independent suctions (two for RHR and two for CSS) located at el 105 ft-6 in. in the containment, the lowest floor elevation in the containment exclusive of the reactor cavity, and they are located outside the secondary shield wall.
The FNP nuclear steam supply system (NSSS) is a three-loop pressurized water reactor (PWR). The system consists of one reactor pressure vessel (RPV), three steam generators (SGs), three reactor coolant pumps (RCPs), one pressurizer (PZR), and the reactor coolant system (RCS) piping. The NSSS is located inside a bioshield and the reactor cavity. The area inside the bioshield is mostly open at the lowest levels, with the exception of the reactor cavity and surrounding walls in the center, and a concrete wall between the A and C loops. The concrete wall between loops A and C has a walkway against the reactor cavity wall that allows an opening between loops A and C. The outer bioshield walls extend from the containment base elevation of 105 ft-6 in. to el 129 ft-0 in. There are areas of the bioshield walls that are partially open; an inner wall extends from el 105 ft-6 in. to 116 ft-3 in., and an outer wall extends down from el. 129 ft-0 in. to el 115 ft-3 in. at some locations. Above el 129 ft-0 in. smaller "vaults" or "coffins" surround each loop and the associated steam generator and reactor coolant pump. These vaults further narrow around the steam generator at el 155 ft-0 in. and extend up to el 166 ft-6 in.. A separate vault for the pressurizer begins at el 129 ft-0 in. and extends up to el 181 ft-0 in.
The containment recirculation sump is a collecting reservoir designed to provide an adequate supply of water, with a minimum amount of particulate matter, to the CSS and the RHRS. The containment sump performance meets the NRC acceptance criteria contained in General Design Criteria 35, 36, and 37, and the NRC acceptance criteria listed below. A. The net positive suction head (NPSH) available to each safety system pump has been shown to provide adequate margin over the required NPSH at limiting runout conditions (see FSAR paragraph 6.3.2.14).
FNP-FSAR-6D 6D-2 REV 22  8/09 B. Housekeeping requirements specified in the quality assurance program and the Technical Requirements Manual. C. The ability to monitor and control RHRS status.
In each of the four pumps suction lines from the containment sump there are two motor-operated gate valves. There is no interdependency between systems or between the redundant portions of the same system. The motor-operated gate valves in the lines from the containment sump to the various pumps are normally closed and remain closed during the injection phase of emergency core cooling system (ECCS) operation. The protective screened structures in the containment sump will be completely submerged at the end of the injection phase and will remain submerged during the recirculation phase.
6D.1.2 GENERAL DESCRIPTION OF NEW ECCS STRAINERS INSTALLED  FNP contracted with General Electric Company (GE) to provide sump strainers that meet the requirements of GL 2004-02. GE provided FNP with seven horizontal stacked disk strainers (see figure 6D-4) and one vertical stacked disk strainer (see figure 6D-3). The strainers were installed in both Unit 1 and Unit 2. Unit 1 only has the vertical stacked strainer installed on the B-train containment spray pump suction. The strainer plate nominal hole size is 3/32 in. The strainers for FNP Unit 1 and Unit 2 are located outside the biowall between the biowall and CTMT outside wall (see figures 6D-1 and 6D-2). This location protects the strainers from missile impacts.
6D.1.3 SIZE OF NEW ECCS STRAINERS INSTALLED  For Unit 1 the passive strainer solution is shown on figure 6D-1. Each strainer assembly for both RHR strainers and CS Alpha strainer consists of two modular horizontal stacked disk strainer subunits connected to the post-LOCA pump suction through piping. The CS Bravo strainer assembly consists of three modular vertical stacked disk strainer subunits connected to a plenum that assists in directing flow to the post-LOCA pump suction inlet located within the plenum boundary. The RHR strainer assembly, either Alpha or Bravo, is composed of two strainer subunits per sump, each consisting of 22 stacked disks that are 40 in. X 40 in. and provide a total of approximately 878 ft&#xb2; of perforated plate surface area. The CS Alpha strainer assembly consists of one strainer subunit with twenty two 40 in. X 40 in. stacked disks and the other with ten 40 in. X 40 in. stacked disks, providing a total of approximately 638 ft&#xb2; of perforated plate surface area. The CS Bravo strainer assembly is composed of three strainer subunits, each with thirteen 30 in. X 30 in. vertical stacked disks, and provides a total of approximately 389 ft&#xb2; of perforated plate surface area. For Unit 2 the passive strainer solution is shown on figure 6D-2. Each strainer assembly for RHR and CS consists of two modular horizontal stacked disk strainers connected to the sump through piping. The RHR strainer assemblies, both Alpha and Bravo, are composed of two strainers per sump, each consisting of 22 stacked disks that are 40 in. X 40 in. and provide a total of approximately 878 ft&#xb2; of perforated plate surface area. The CS Alpha strainer assembly FNP-FSAR-6D 6D-3 REV 22  8/09 consists of one strainer with twenty two 40 in. X 40 in. stacked disks and the other with ten 40 in. X 40 in. stacked disks, providing a total of approximately 638 ft&#xb2; of perforated plate surface area. The CS Bravo strainer assembly is composed of two strainers, one with ten 40 in. X 40 in. stacked disks and the other with twenty two 30 in. X 30 in. disks, and provides a total of approximately 433 ft&#xb2; of perforated plate surface area.
6D.2 SUMMARY DESCRIPTION OF APPROACH USED TO SIZE SUMP STRAINERS  SNC has performed analysis to determine the susceptibility of the ECCS and CSS recirculation functions for Farley Nuclear Plant to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform to the greatest extent practicable to the NEI 04-07 methodology as approved by the NRC safety evaluation report dated December 6, 2004. Following is a summary description of the analysis areas performed:
6D.2.1 CONTAINMENT WALKDOWN  Walkdown of containment was performed by SNC personnel using the guidance of NEI 02-01.
The information obtained from the walkdown confirmed the insulation that was installed in containment matched the design documentation. Containment walkdowns confirmed the general housekeeping condition of containment was being maintained per plant procedures. 6D.2.2 PIPE BREAK CHARACTERIZATION  Pipe break characterization was performed by Sargent and Lundy of Chicago. The piping runs considered for breaks are the RCS hot legs, the RCS cold legs, RCS interim legs, and all RCS attached energized piping. Breaks in these lines could decrease RCS inventory and result in the ECCS and/or CSS operating in recirculation mode, in which the system pumps would take suction from the containment sumps. Regulatory position 1.3.2.3 of Regulatory Guide 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 3, was used to select the spectrum of breaks for evaluation. A summary of the break locations is provided in figure 6D-5.
6D.2.3 DEBRIS GENERATION  The debris generation analysis was performed by Sargent and Lundy of Chicago. The analysis determined the debris generated based on the NEI guidance and NRC SER of the NEI guidance. The analysis determined the ZOI for each type of material identified inside containment. See table 6D-1 for basis of ZOIs.
Insulation found inside containment that is adversely affected during a LOCA event, was determined to consist of a very small quantity of Tempmat fiber, Transco RMI, and Mirror insulations. Most of the insulation is Transco RMI and Mirror RMI. The amount of Tempmat fiber is very small. See table 6D-2 for summary of debris generated by each break.
FNP-FSAR-6D 6D-4 REV 22  8/09 The limiting break for coatings evaluated for a 4.0D ZOI is also on the intermediate leg of loop B, but at the RCP side of the pipe. Therefore, in order to conservatively maximize the debris available for transport, the maximum insulation debris location (break S2) is combined with the maximum coating debris location. See table 6D-3 for coating debris. Unqualified coatings are also identified in containment walkdown and plant condition reports.
6.D.2.4 LATENT DEBRIS ACCUMULATION WITHIN CONTAINMENT  Programmatic controls are in place at FNP that give bases for the amounts of foreign material and latent debris inside containment remaining below the amounts assumed in the sump analysis. See table 6D-4 for latent and foreign material debris used in the analysis.
6D.2.5 DEBRIS TRANSPORT TO THE SUMP A debris transport analysis estimated the fraction of debris that is transported from debris sources (break locations) to the sump screen. The transport analysis is in accordance with the guidance of NEI 04-07 and the applicable NRC SER. The computational fluid dynamics (CFD) analysis was performed by RWDI Consulting Engineers and Scientists for Sargent and Lundy of Chicago. The CFD modeling techniques used are consistent with the SER, NEI Document number 04-07, and NUREG/CR-6773. CFD analyses of the post-LOCA recirculation flow patterns within the FNP containments were performed to quantify the flow velocities expected inside the secondary shield wall, through the secondary shield wall, outside the secondary shield wall, and near the CS and RHR sumps.
CFD analysis of the post-LOCA recirculation containment flows indicates velocities that will transport debris to the suction strainers. See table 6D-5 for a Summary of Debris Generated and Transported to Strainer Modules.
6D.2.6 HEAD LOSS AS A RESULT OF DEBRIS ACCUMULATION The engineered sump screens installed at FNP are designed to operate in such a way that the thin bed effect does not occur on the sump screen surface. This is due to the small amount of fiber present in the FNP containment. Parametric analyses were performed to estimate the surface area of the engineered screen that meets the FNP head loss criterion for the identified debris inventory. For the limiting break for screen head loss as selected in accordance with NEI 04-07, screens would be fully submerged at the minimum calculated sump levels. The RHR screen height is 44.75 in. above the floor. With leveling shims the height may be increased at points on the screens less than an inch. The minimum calculated water level is 54 in. above the floor elevation which is calculated to occur for the long term and not at the initiation of recirculation. This is largely due to gradual refilling of the area under the reactor vessel and due to conservatively postulated refilling of the SG tubes and the pressurizer. The tallest CS screen is 46.2 in. high; therefore, it may have slightly less submergence. Under this scenario the screens will be fully submerged by no less than 6 in. 
FNP-FSAR-6D 6D-5 REV 22  8/09 A small break LOCA that results in minimum sump level would be one that occurs on top of the pressurizer. This level was not calculated as it is not a limiting break location that results in the highest screen head losses. The connections on the top of the pressurizer are 6 in. in diameter.
Therefore, a break in this location would produce very small amounts of debris. In addition, as compared to the limiting large break location, a small break would result in lower sump flowrates and, therefore, reduced sump debris transport. The resultant reduced RHR flowrates would result in a reduction in both debris bed head loss and a reduction in the NPSH required for the RHR pumps. An SBLOCA clearly does not present a significant challenge to the ECCS sump performance and is bounded by a LBLOCA. Since this is not a limiting break location the screen submergence was not calculated for this break.
As the screens are well covered for the limiting breaks the potential for air injection due to buoyant debris accumulation on top of the strainer is not considered to be plausible. For breaks that may result in some transient uncoverage, RHR flowrates would be reduced. CS screens would be fully covered as the RWST level is drawn down further before CS is placed on recirculation. A vortexing analysis was done for the Farley strainers assuming maximum RHR and CS flowrates. Vortexing was not indicated using the assumption that the strainer has the geometry of an open ended submerged pipe. This conservatively does not account for the complex stacked disc geometry of the strainer which would in effect act as vortex breakers.
6D.2.7 DEBRIS SOURCE TERM REDUCTION  Foreign material (i.e., tags, labels, etc., not qualified for LOCA environmental conditions) may fail following a LOCA and, therefore, can be transported to the sump. Actions have been taken by SNC to ensure that the quantity of foreign material is minimized.
6D.2.8 SUMP STRUCTURAL ANALYSIS  Structural analysis of the engineered passive screen has been completed. SNC has installed an engineered passive strainer on each RHR and CSS containment sump inlet pipe. The screens are located outside the secondary shield wall between the shield wall and the containment wall and, as such, are not exposed to jet impingement or postulated missiles generated from a LOCA event. The screens are of a robust design that support structural and hydraulic load created by the accumulation of debris during the post-LOCA environment. This robust design provides the strength of trash racks and is adequate to protect the screen during a LOCA event.
6D.2.9 UPSTREAM EFFECTS OF DEBRIS ACCUMULATION Evaluations of containment along with review of the CFD model indicate no significant areas will become blocked with debris and hold up water during the sump recirculation phase. As a precautionary measure, SNC modified the reactor cavity drain covers to further reduce the possibility of the drain becoming clogged and trapping a volume of water in the reactor cavity.
FNP-FSAR-6D 6D-6 REV 22  8/09 6D.2.10 DOWNSTREAM EFFECTS - COMPONENTS AND SYSTEMS  The methodologies of NEI 04-07, as modified by the NRC safety evaluation dated December 6, 2004, and WCAP-16406-P, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191," were used to evaluate the downstream effects of debris that is passed by the sump strainer. The only components requiring modification were the safety injection throttle valves. A new flow reducing orifice was installed and the valves were replaced. This modification has been completed on Unit 1 and resulted in 9 of the 12 valves being locked open at a position which produced an internal valve clearance of 110 % of the containment sump screen hole size.
The other 3 valves opening are about 106 % of the hole size. An evaluation to address the acceptability of these valve positions was performed and they were found to be acceptable. The throttle valve replacement on Unit 2 has been delayed until fall 2008. An extension request was approved by the NRC in a letter dated August 29, 2007. This section will be updated later to reflect the installation on the Unit 2 valves.
6D.2.11 DOWNSTREAM EFFECTS - FUEL AND VESSEL  The methodologies of WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Re-circulating Fluid," Revision 0, as modified by NRC staff comments, were used to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling. The evaluation of the impact of chemical deposition on the fuel was performed using the guidance of WCAP-16793-NP with bounding plant parameters. In its supplemental responses to GL 2004-02, submitted to the NRC on February 28, 2008, and April 29, 2008 (see references 13 and 14), SNC concluded there was reasonable assurance that long-term core cooling was demonstrated for Farley Units 1 and 2. 6D.2.12 CHEMICAL EFFECTS  The new strainers installed at FNP have been sized to account for some increase in head loss across the strainer as a result of interaction of the sump water with the debris material as it approached the strainers during recirculation phase. The methodologies of the base model WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191," Revision 0 (reference 10), as modified by NRC safety evaluation report dated December 21, 2007 (reference 11), were used to evaluate the impact of chemical precipitants on the containment sump screens during post-accident recirculation and the resulting effect on available NPSH for the ECCS and CSS pumps. SNC supplemented the chemical effects results with plant-specific test data that demonstrated that the aluminum precipitants do not form until the containment sump temperature drops below 140 &#xba;F (see reference 15). Calculations using the chemical effects testing results and other inputs demonstrated the available NPSH margin for the ECCS and CSS pumps was adequate for the conditions expected during post-accident recirculation. In addition, the results of chemical effects testing were used in the evaluation of downstream effects on fuel and the reactor vessel (refer to 6D.2.11). In its final supplemental response to GL 2004-02, submitted to the NRC on April 29, 2008 (reference 14), SNC concluded that there was reasonable assurance that long-term core cooling was demonstrated for Farley Units 1 and FNP-FSAR-6D 6D-7 REV 22  8/09 2. The details of the chemical effects testing results are documented in GE Report 0000-0056-2976, Containment Sump Passive RHR & CS Strainer System S0100 Hydraulic Sizing Report, Revision 3 (reference 15).
FNP-FSAR-6D 6D-8 REV 22  8/09 REFERENCES  1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," September 13, 2004.
2. Nuclear Energy Institute (NEI) document NEI 04-07 Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology," December 2004.
3. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), "Pressurized Water Reactor Sump Performance Evaluation Methodology," December 6, 2004.
4. Regulatory Guide 1.82, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," Revision 3, November 2003. 5. WCAP-16568-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for DBA-Qualified / Acceptable Coatings," Revision 0. 6. Deleted. 7. Deleted. 8. WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191," Revision 1. 9. NRC SER dated December 20, 2007, Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report (TR) WCAP-16406-P, Revision 1, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191," Pressurized Water Reactor Owners Group. 10. WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191."  11. NRC SER dated December 21, 2007, Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191."  12. WCAP 16793-NP, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Re-circulating Fluid, Revision 0. 13. SNC Letter NL-08-2173 dated February 28, 2008, "Joseph M. Farley Nuclear Plant Supplemental Response to NRC Generic Letter 2004-02."  14. SNC Letter NL-08-0551 dated April 29, 2008, "Joseph M. Farley Nuclear Plant Final Supplemental Response to NRC Generic Letter 2004-02."  15. GE Report 0000-0056-2976 [U-732504], "Containment Sump Passive RHR & CS Strainer System S0100 Hydraulic Sizing Report," Revision 3.
FNP-FSAR-6D REV 21  5/08 TABLE 6D-1 CONTAINMENT SUMP DEBRIS GENERATION ZONE OF INFLUENCE (ZOI)  Debris Constituent  ZOI (Pipe Diameter)  Basis    Transco RMI 2.0D NRC SER Mirror RMI 28.6D NRC SER Temp-Mat Fiber NA All assumed as debris in analysis  Qualified Coatings 4.0D WCAP-16568-P Unqualified Coatings NA NRC SER - All assumed as debris in analysis  Latent Debris NA NRC SER - Conservative value based on plant walkdown  Foreign Materials  (Labels, etc.) NA NRC SER - Conservative value based on plant walkdown FNP-FSAR-6D REV 21  5/08 TABLE 6D-2 SUMMARY OF LOCA GENERATED INSULATION DEBRIS INSIDE ZOI  Break ID Location Transco RMI Foils (ft2) Mirror RMI Foils (ft2) RMI Jacketing (ft2) Temp-Mat (ft3)      S1 Loop C Interim Leg 2054 25527 5795 1      S2* Loop B Interim Leg 2383 35714 8022 1      S3 Loop A  Cold Leg 0 34368 7522 1      S4 (alternate) Loop B Interim Leg 1226 23258 5223 0 
___________________
* S2 is the limiting location. 
FNP-FSAR-6D REV 21  5/08 TABLE 6D-3 DEBRIS GENERATED FROM COATING BASED ON ZOI = 4D  Break Coating Areas (ft2) Coating Volumes (ft3)  Concrete Steel Concrete Steel Interim Leg at SG 200 1332 0.31 1.66 Interim Leg at Mid-span 218 1320 0.34 1.65 *Interim Leg at RCP 523 1091 0.81 1.36 Hot Leg at Primary Wall 294 758 0.46 0.95 Hot Leg at SG 0 1196 0 1.49 Unqualified Coatings NA 1,070 NA 0.535 
________________
* Limiting location for coatings FNP-FSAR-6D REV 21  5/08 TABLE 6D-4 LATENT AND FOREIGN MATERIAL DEBRIS USED IN ANALYSIS  Latent Debris Total (lbm) 200 Fiber  (lbm) 30 Particulate  (lbm) 170 Foreign Material Debris  (ft2) 36.4 FNP-FSAR-6D REV 21  5/08 TABLE 6D-5 SUMMARY OF DEBRIS GENERATED AND TRANSPORTED TO STRAINER MODULES  Debris Type Units Quantity Generated Transport Fraction Quantity at Strainer Modules Fibrous Insulation Debris  Temp-Mat [ft3] 1 1.0 1  Coating Debris in 4D ZOI Modeled as Chips  Concrete Coatings  [ft2; ft3] 523 ; 0.81 0.871 456 ; 0.71 Steel Coatings  [ft2; ft3] 1091 ; 1.36 0.704 768 ; 0.96 Sum [ft2; ft3] 1614 ; 2.18 --- 1224 ; 1.67  Unqualified Coating Debris Modeled as Fines / Chips
* Unqualified Coatings (Actual) [ft2; ft3] 1070 ; 0.535 1.0 1070 ; 0.535  Latent Debris  Latent Fiber (Walkdown) [ft3] 7.8 1.0 7.8 Latent Fiber  (30 lbm)  [ft3] 12.5 1.0 12.5 Latent Particulate (Walkdown) [ft3] 0.63 1.0 0.63 Latent Particulate (170 lbm) [ft3] 1.01 1.0 1.01  Reflective Metal Insulation Debris  Transco Foil [ft2] 2383 0.799 1904 Mirror Foil [ft2] 35714 0.769 27464 Foil Sum [ft2] 38097 --- 29368 RMI Jacketing [ft2] 8022 0.338 2711  Foreign Material  Foreign Material1 (labels, stickers, etc.) [ft2] 36.4 1.0 36.4 
________________________
* Unqualified Coatings were modeled as a mixture of chips and fines. 
Break Name Break ID Elevation Piping S1 31-inch 118'-0" Interim Leg - Loop C S2 31-inch 118'-0" Interim Leg - Loop B S3 27.5-inch 122-9" Cold Leg - Loop A S4 11.19-inch 118'-0" Alternate Break (Interim Leg -Loop B)        REV 21  5/08 POSTULATED BREAK LOCATIONS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6D-5 REFER TO FIGURE 6D-4 FOR CONTINUATION OF SUCTION PIPE CONNECTION. REV 22  8/09 TYPICAL ARRANGEMENT OF  CONTAINMENT SUMP SUCTION LINE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6D-6 FNP-FSAR-7 7-i REV 22  8/09 7.0  INSTRUMENTATION AND CONTROL  TABLE OF CONTENTS 
==7.1 INTRODUCTION==
.........................................................................................................7.1-1 7.1.1 Identification of Safety-Related Systems.........................................................7.1-3 7.1.2 Identification of Safety Criteria........................................................................7.1-4
7.1.2.1 Design Bases..................................................................................7.1-5 7.1.2.2 Independence of Redundant Safety-Related Systems.................7.1-10 7.1.2.3 Physical Identification of Safety-Related Equipment....................7.1-12 7.1.2.4 Conformance to IEEE 317-1971...................................................7.1-12 7.1.2.5 Conformance to IEEE 323-1971...................................................7.1-13 7.1.2.6 Conformance to IEEE 336-1971...................................................7.1-13 7.1.2.7 Conformance to IEEE 338-1971...................................................7.1-13 7.1.2.8 Conformance to Regulatory Guide 1.22.......................................7.1-14 7.1.2.9 Conformance to IEEE 334-1971...................................................7.1-15 7.1.2.10 Conformance to 10 CFR 50.62.....................................................7.1-15 7.1.2.11 Conformance to NUREG-0737.....................................................7.1-15
7.1.3 Detailed Electrical Instrumentation and Control Drawings............................7.1-16
7.1.3.1 Identification and Purpose............................................................7.1-16
7.2 REACTOR TRIP SYSTEM..........................................................................................7.2-1
7.2.1 Description......................................................................................................7.2-1
7.2.1.1 System Description.........................................................................7.2-2 7.2.1.2 Design Bases:  IEEE 279-1971....................................................7.2-14 7.2.1.3 Final System Drawings.................................................................7.2-17
7.2.2 Analysis.........................................................................................................7.2-18
7.2.2.1 Failure Mode and Effects Analysis................................................7.2-18 7.2.2.2 Evaluation of Compliance to Applicable Codes and Standards....7.2-20 7.2.2.3 Specific Control and Protection Interactions.................................7.2-30
7.2.3 Tests and Inspections...................................................................................7.2-34
FNP-FSAR-7 7-ii REV 22  8/09 TABLE OF CONTENTS    7.2.3.1 Inservice Tests and Inspections....................................................7.2-34  7.2.3.2 Periodic Testing of the Nuclear Instrumentation System..............7.2-36 7.2.3.3 Periodic Testing of the Process Analog Channels of the Protection Circuits...............................................................7.2-36 7.2.3.4 Regulatory Guide 1.22..................................................................7.2-37
7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM....................................7.3-1
7.3.1 Description......................................................................................................7.3-1
7.3.1.1 System Description.........................................................................7.3-1 7.3.1.2 Design Bases..................................................................................7.3-6 7.3.1.3 Final System Drawings...................................................................7.3-9
7.3.2 Analysis...........................................................................................................7.3-9 7.3.2.1 Evaluation of Compliance with IEEE 279-1971............................7.3-10 7.3.2.2 Evaluation of Compliance with IEEE 308-1971............................7.3-16  7.3.2.3 Evaluation of Compliance with IEEE 323-1971............................7.3-16 7.3.2.4 Evaluation of Compliance with IEEE 334-1971............................7.3-16 7.3.2.5 Evaluation of Compliance with IEEE 338-1971............................7.3-17 7.3.2.6 Evaluation of Compliance with IEEE 344-1971............................7.3-17 7.3.2.7 Response Time Testing................................................................7.3-17 7.3.2.8 Further Considerations.................................................................7.3-18 7.3.2.9 Summary.......................................................................................7.3-18
7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN......................................................7.4-1 7.4.1 Description......................................................................................................7.4-1
7.4.1.1 Monitoring Indicators.......................................................................7.4-2 7.4.1.2 Controls...........................................................................................7.4-2 7.4.1.3 Essential Services after Incident That Requires Hot Shutdown......7.4-5 7.4.1.4 Equipment and Systems Available for Cold Shutdown...................7.4-6
7.4.2 Analysis...........................................................................................................7.4-7 
FNP-FSAR-7 7-iii REV 22  8/09 TABLE OF CONTENTS  7.5 POSTACCIDENT MONITORING DISPLAY INSTRUMENTATION............................7.5-1 7.5.1 Description......................................................................................................7.5-1 7.5.2 Analysis...........................................................................................................7.5-2  7.5.3 Deleted............................................................................................................7.5-3  7.5.4 Inadequate Core Cooling Monitoring System..................................................7.5-3
7.5.4.1 Reactor Vessel Level......................................................................7.5-3 7.5.4.2 Subcooling Margin Monitor.............................................................7.5-4 7.5.4.3 Core Exit Temperature....................................................................7.5-4
7.5.5 Nuclear Instrumentation..................................................................................7.5-4
7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY...................................................7.6-1
7.6.1 Instrumentation and Control Power Supply System........................................7.6-1
7.6.1.1 Description......................................................................................7.6-1 7.6.1.2 Analysis...........................................................................................7.6-1
7.6.2 Residual Heat Removal Isolation Valves.........................................................7.6-3
7.6.2.1 Description......................................................................................7.6-3 7.6.2.2 Analysis...........................................................................................7.6-4
7.6.3 Refueling Interlocks.........................................................................................7.6-4 7.6.4 Monitoring Combustible Gas in Containment..................................................7.6-6
7.6.4.1 Description......................................................................................7.6-6 7.6.4.2 Analysis...........................................................................................7.6-7
7.6.5 Semiautomatic Backup to Switchover from Injection to Recirculation.............7.6-7 7.6.6 Accumulator Motor-Operated Isolation Valves................................................7.6-7
7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY.............................................7.7-1
7.7.1 Description......................................................................................................7.7-1
7.7.1.1 Reactor Control System..................................................................7.7-3 7.7.1.2 Rod Control System........................................................................7.7-4
FNP-FSAR-7 7-iv REV 22  8/09 TABLE OF CONTENTS    7.7.1.3 Plant Control Signals for Monitoring and Indicating........................7.7-5  7.7.1.4 Plant Control System Interlocks......................................................7.7-9 7.7.1.5 Pressurizer Pressure Control..........................................................7.7-9 7.7.1.6 Pressurizer Water Level Control...................................................7.7-10 7.7.1.7 Steam Generator Water Level Control..........................................7.7-10 7.7.1.8 Steam Dump Control....................................................................7.7-11 7.7.1.9 Incore Instrumentation..................................................................7.7-12 7.7.1.10 Control Board................................................................................7.7-14 7.7.1.11 Boron Concentration Measurement System.................................7.7-15
7.7.2 Analysis.........................................................................................................7.7-17
7.7.2.1 Separation of Protection and Control Systems.............................7.7-18 7.7.2.2 Response Considerations of Reactivity........................................7.7-19 7.7.2.3 Step Load Changes Without Steam Dump...................................7.7-21 7.7.2.4 Loading and Unloading.................................................................7.7-21 7.7.2.5 Load Rejection Furnished by Steam Dump System.....................7.7-21 7.7.2.6 Turbine-Generator Trip with Reactor Trip.....................................7.7-22
7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC)..........................7.8-1
7.8.1 Description......................................................................................................7.8-1
7.8.1.1 System Description.........................................................................7.8-1 7.8.1.2 Equipment Description....................................................................7.8-1 7.8.1.3 Functional Performance Requirements..........................................7.8-3 7.8.1.4 AMSAC Interlocks...........................................................................7.8-3 7.8.1.5 Trip System.....................................................................................7.8-3 7.8.1.6 Isolation Devices.............................................................................7.8-4 7.8.1.7 AMSAC Diversity From the Reactor Protection Systems................7.8-4 7.8.1.8 Power Supply..................................................................................7.8-4 7.8.1.9 Environmental Variations................................................................7.8-4 7.8.1.10 Setpoints.........................................................................................7.8-4
7.8.2 Analysis...........................................................................................................7.8-5 7.8.2.1 Safety Classification/Safety-Related Interface................................7.8-5 7.8.2.2 Redundancy....................................................................................7.8-5 7.8.2.3 Diversity From Existing Trip System...............................................7.8-5
FNP-FSAR-7 7-v REV 22  8/09 TABLE OF CONTENTS    7.8.2.4 Electrical Independence..................................................................7.8-5  7.8.2.5 Physical Separation From the RTS and ESFAS.............................7.8-6 7.8.2.6 Environmental Qualification............................................................7.8-6 7.8.2.7 Seismic Qualification.......................................................................7.8-6 7.8.2.8 Test, Maintenance, and Surveillance Quality Assurance................7.8-6 7.8.2.9 Power Supply..................................................................................7.8-7 7.8.2.10 Testability at Power.........................................................................7.8-7 7.8.2.11 Inadvertent Actuation......................................................................7.8-7 7.8.2.12 Bypass............................................................................................7.8-7 7.8.2.13 Completion of Mitigative Actions Once Initiated..............................7.8-8 7.8.2.14 Manual Initiation..............................................................................7.8-8 7.8.2.15 Information Readout.......................................................................7.8-8 7.8.2.16 Compliance With Standards and Design Criteria............................7.8-9
FNP-FSAR-7 7-vi REV 22  8/09 LIST OF TABLES  7.1-1 List of Schematic Diagrams and Location Drawings for Safety-Related Equipment  7.2-1 List of Reactor Trips 
7.2-2 Protection System Interlocks 
7.2-3 Reactor Trip System Instrument Accuracies 
7.2-4 Trip Correlation 
7.2-5 Reactor Trip System Instrumentation Response Times
7.3-1 Functions Initiated by Engineered Safety Features Actuation System 
7.3-2 Instrumentation Operating Conditions for Engineered Safety Features 
7.3-3 Instrumentation Operating Conditions for Isolation Functions 
7.3-4 Interlocks for Engineered Safety Features Actuation System 
7.3-5 (Deleted)
7.3-6 Failure Mode and Effects Analysis, Service Water System 
7.3-7 Failure Mode and Effects Analysis, Component Cooling Water System 
7.3-8 Failure Mode and Effects Analysis, Control Room and Air Conditioning and Filtration System 7.3-9 Failure Mode and Effects Analysis, Penetration Room Filtration System 
7.3-10 Failure Mode and Effects Analysis, Auxiliary Feedwater System 7.3-11 Failure Mode and Effects Analysis, Emergency Safeguards Pump Room Cooling System 7.3-12 Failure Mode and Effects Analysis, Battery Room Ventilation System 
7.3-13 Failure Mode and Effects Analysis, Battery Room Air Conditioning System 
FNP-FSAR-7 7-vii REV 22  8/09 LIST OF TABLES  7.3-14 Failure Mode and Effects Analysis, Emergency Diesel Generator 7.3-15 Failure Mode and Effects Analysis, Engineered Safety Features Actuation System 
7.3-16 Engineered Safety Features Response Times 
7.5-1 Post Accident Instrumentation 
7.5-2 (Deleted)
7.5-3 Control Room Indicators and/or Recorders Available to the Operator to Monitor Significant Plant Parameters During Normal Operation 7.7-1 Plant Control System Interlocks 
7.7-2 Boron Concentration Measurement System Specifications 
FNP-FSAR-7 7-viii REV 22  8/09 LIST OF FIGURES  7.2-1 Setpoint Reduction Function for Overpower and Overtemperature T Trips  7.2-2 Pressurizer Sealed Reference Leg Level System 
7.2-3 Design to Achieve Isolation Between Channels 
7.3-1 Component Identification ESFAS 
7.6-1 Logic Diagram for Residual Heat Removal System Isolation Valves 
7.6-2 Logic Diagram for Residual Heat Removal System Isolation Valves 
7.6-3 Logic Diagram for Backup to Semiautomatic Switchover Logic from Injection to Recirculation 7.6-4 Functional Block Diagram of Accumulator Isolation Valve 
7.7-1 Simplified Block Diagram of Reactor Control System 
7.7-2 Control Bank Rod Insertion Monitor 
7.7-3 Rod Deviation Comparator 
7.7-4 Block Diagram of Pressurizer Pressure Control System 
7.7-5 Block Diagram of Pressurizer Level Control System 
7.7-6 Block Diagram of Main Feedwater Pump Speed Control System 
7.7-7 Block Diagram of Steam Generator Water Level Control System 7.7-8 Block Diagram of Steam Dump Control System 
7.7-9 Basic Flux Mapping System 
7.7-10 Source-Detector Assembly 
7.7-11 Measurement Unit 
7.7-12 Process Schematic for the Boron Concentration Measurement System 7.7-13 Boron Concentration Measurement System vs Normal Plant Operating Range of Boron Concentrations 7.8-1 Actuation Logic System Architecture 
[HISTORICAL][7.1.3  DETAILED ELECTRICAL INSTRUMENTATION AND CONTROL DRAWINGS    7.1.3.1  Identification and Purpose  A set of volumes containing nonproprietary detailed EI&C drawings has been prepared in accordance with the NRC interim guidelines, pending revisions of the Standard Format. It is entitled, "Joseph M.
Farley Nuclear Plant, Safety-Related Schematic Diagrams and Location Drawings, November 1973," and is in four volumes, FNP-1001, FNP-1002, FNP-1003, and FNP-1004. The supplement furnished detailed information in response to paragraphs 7.2.1.3 and 7.3.1.3 and subsections 7.4.1 and 7.6.1 of the Standard Format. The purpose of the supplement was to facilitate tracing the safety-related signals from sensors to actuating devices. It was submitted with Amendment 27. A list of the submitted EI&C drawings and related FSAR figures is maintained in table 7.1-1 for historical purposes.
FNP-FSAR-7 REV 21  5/08  [HISTORICAL][TABLE 7.1-1 (SHEET 1 OF 33)  LIST OF SCHEMATIC DIAGRAMS AND LOCATION DRAWINGS  FOR SAFETY-RELATED EQUIPMENT 
This table lists drawings which are presented in the FSAR by reference to project drawing numbers or were provided to the NRC in the supplement. 
Submittal    to NRC  Drawing (formerly  Number  AEC)  Title  Equipment Location Number Index 175070 11/01/73 Equipment numbers tabulation Nuclear Instrumentation System (NIS) Block Diagrams and Safeguards Test Cabinet 108D501 11/01/73 Process control block diagram  5655D37 11/01/73 Functional diagrams 5655D49 11/01/73 NIS source range functional block diagram  5655D50 11/01/73 NIS intermediate range functional  block diagram  5655D51 11/01/73 NIS power range functional block diagram  5655D52 11/01/73 NIS auxiliary channels functional block diagram  724D17 11/01/73 Safeguards test cabinet (10 sheets)
Elementary Diagrams, 177000 Series (Includes Related 207000 Series) 177000 11/01/73 Single line electrical    auxiliary system (normal)  177001 11/01/73 Single line electrical    auxiliary system (emergency)  177005 11/01/73 Single line protection and    metering, 4160-V bus 1F  177006 11/01/73 Single line protection and    metering, 4160-V bus 1G  177018 11/01/73 Single line protection and    metering, 4160-V bus 1H  177027 11/01/73 Single line protection and    metering, 4160-V bus 1J  177043 11/01/73 Single line protection and    metering, 4160-V bus 1K  177044 11/01/73 Single line protection and    metering, 4160-V bus 1L  177007 11/01/73 Single line protection and    metering, 600-V load center 1A  177009 11/01/73 Single line protection and    metering, 600-V load center 1C 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 2 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177010 11/01/73 Single line protection and    metering, 600-V load center 1D  177011 11/01/73 Single line protection and    metering, 600-V load center 1E  177012 11/01/73 Single line protection and    metering, 600-V load center 1F  177014 11/01/73 Single line protection and    metering, 600-V load center 1H  177015 11/01/73 Single line protection and    metering, 600-V load center 1J  177045 11/01/73 Single line protection and    metering, 600-V load center 1K  177046 11/01/73 Single line protection and    metering, 600-V load center 1L  177677 11/01/73 Single line protection and    metering, 600-V load center 1R  177678 11/01/73 Single line protection and    metering, 600-V load center 1S  177118 11/01/73 Interlock schematic station    service transformer 1F  177122 11/01/73 Interlock schematic 600-V bus 1A  177024 11/01/73 Single line 120 V-ac vital and    regulated system A  177025 11/01/73 Single line 120 V-ac vital and    regulated system B  177754 11/01/73 Tray and conduit layout, cable    spreading room  177033 11/01/73 Logic diagram diesel 1A auto    start and loading  177032 11/01/73 Logic diagram diesel 1B auto    start and loading  177036 11/01/73 Logic diagram diesel 1C auto    start and loading 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 3 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177037 11/01/73 Logic diagram diesel 2C auto    start and loading  - 11/01/73 One-line diagram dc supply for    diesel generators  207032 11/01/73 Logic diagram  diesel 2B auto    start and loading  177119 11/01/73 Interlock schematic component    cooling water pump 2B  177120 11/01/73 Interlock schematic HHSI pump 2B  177121 11/01/73 Interlock schematic service    water pump 1C  177082 11/01/73 Single line dc distribution    system 1A  177083 11/01/73 Single line dc distribution    system 1B  207000 11/01/73 Single line electrical auxiliary    system (normal 4160 V and    600 V) Unit 2  207001 11/01/73 Single line electrical auxiliary    system (emergency 4160 V and    600 V) Unit 2  207033 11/01/73 Logic diagram diesel 1A auto    start and loading  207036 11/01/73 Logic diagram diesel 1C auto    start and loading  207037 11/01/73 Logic diagram diesel 2C auto    start and loading  177133 11/01/73 Interlock schematic battery    charger 1C  177050 11/01/73 Elementary diagram 600-V LC    bus 1A tie breaker from 600-V    LC bus 1D  177051 01/10/75 Elementary diagram 575-V motor-    operated valve  177052 11/15/74 Elementary diagram 575-V motor-    operated valve  177053 11/15/74 Elementary diagram 575-V motor-    operated valve  177058 11/01/73 Elementary diagram 600-V LC    bus 1C tie breaker  177059 11/01/73 Elementary diagram 600-V LC    bus 1C tie breaker from    600-V LC bus 1E 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 4 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177064 11/01/73 Elementary diagram 600-V LC    bus 1D tie breaker from    600-V LC bus 1A  177070 11/01/73 Elementary diagram 600-V LC    bus 1E tie breaker from    600-V LC bus 1C  177072 11/01/73 Elementary diagram 600-V LC    buses 1D and 1E, including    breaker from bus 1F  177077 11/01/73 Elementary diagram 600-V LC    breakers to battery chargers    1A and 1B  177078 11/01/78 Elementary diagram 600-V LC    breakers to battery charger 1C  177080 11/01/73 Synchronizing diagram 4160-V    emergency buses train A    Units 1 and 2  177081 11/01/73 Synchronizing diagram 4160-V    emergency buses train B    Units 1 and 2  177087 11/01/73 Elementary diagram 600-V LC    buses 1A, 1B, 1C, 1D, and 1E    potential transformer  177089 11/01/73 Elementary diagram 600-V LC    breakers to motor control    centers 1A, 1B, 1F, 1G,    1S, 1U, and 1V  177091 11/01/73 Elementary diagram miscellaneous    relay 177142 11/01/73 Elementary diagram 4160-V    bus 1G incoming breaker from    diesel generator 1B  177143 11/01/73 Elementary diagram 4160-V    bus 1F incoming breaker from    diesel generator 1A  177144 11/01/73 Elementary diagram 4160-V    bus tie from 4160-V bus 1F    to 1KC 1G101L  177145 11/01/73 Elementary diagram 4160-V    bus tie breaker from 4160-V    bus 1F to 1MC 1G101J  177155 11/01/73 Elementary diagram 4160-V    bus 1F incoming startup    transformer 1A 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 5 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177157 11/01/73 Elementary diagram 4160-V bus 1F potential  transformers  177159 11/01/73 Elementary diagram 4160-V bus 1F outgoing station      service transformers 1D      and 1G  177160 11/01/73 Elementary diagram 4160-V bus 1F outgoing station      service transformer 1F  177161 11/01/73 Elementary diagram 4160-V bus 1F incoming startup      transformer 1B  177163 11/01/73 Elementary diagram 4160-V bus 1G potential  transformers  177166 11/01/73 Elementary diagram 4160-V bus 1G outgoing station      service transformer 1F  177167 11/01/73 Elementary diagram 4160-V bus tie breaker 1G to 1J  177168 11/01/73 Elementary diagram 4160-V bus 1G incoming startup      transformer 1A  177169 11/01/73 Elementary diagram 4160-V bus 1G incoming startup      transformer 1B  177170 11/01/73 Elementary diagram 4160-V buses 1F and 1G diff. prot. 177173 11/01/73 Elementary diagram 4160-V bus 1G diff. prot. 177183 11/01/73 Elementary diagram component cooling water pump 4160-V      bus 1C  177184 11/01/73 Elementary diagram component cooling water pump 4160-V      bus 1A  177185 11/01/73 Elementary diagram component cooling water pump 4160-V      bus 1B train A  177186 11/01/73 Elementary diagram auxiliary feedwater pump 4160-V      buses 1A and 1  177187 11/01/73 Elementary diagram component cooling water pump 4160-V      bus 1B train B FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 6 OF 33)    Submittal to NRC    Drawing (formerly    Number    AEC)  Title  177188 11/01/73 Elementary diagram turbine-    driven auxiliary feedwater    pump-starter train A  177189 11/01/73 Elementary diagram turbine-    driven auxiliary feedwater    pump-starter train B  177199 11/01/73 Elementary diagram containment    purge exhaust damper  177204 11/01/73 Elementary diagram containment    purge system isolation    dampers 177206 11/01/73 Elementary diagram containment    post-LOCA air mixing fans  177221 11/01/73 Elementary diagram containment    cooling high speed  177222 11/01/73 Elementary diagram containment    cooling low speed  177224 11/01/73 Elementary diagram boric acid    transfer pumps 1 and 2  177226 11/01/73 Elementary diagram charging/    HHST pump 1B room cooler    fan motor train A  177227 11/01/73 Elementary diagram RHR pump    and containment spray pump    from cooler fan motors  177229 11/01/73 Elementary diagram HHST and    auxiliary feedwater pump room    and common heat exchange cooler    fan motor  177232 11/01/73 Elementary diagram containment    cooler damper motor  177236 11/01/73 Elementary diagram containment    purge supply fan high speed  177237 11/01/73 Elementary diagram containment    purge exhaust fan low speed  177238 11/01/73 Elementary diagram penetration    from exhaust fans 1 and 2  177239 11/01/73 Elementary diagram penetration    room recirculation fans 1 and 2  177240 11/01/73 Elementary diagram boron    injection  tank recirculation    pumps 1 and 2 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 7 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title      177243 11/01/73 Elementary diagram component    cooling water and pump room    cooler fans  177246 11/01/73 Elementary diagram spent fuel    pool air exhaust fans 1 and 2  177253 08/15/74 Elementary diagram phosphate    injection pumps  177259 11/01/73 Elementary diagram radwaste air    exhaust fan 1A motor  177262 11/01/73 Elementary diagram control rod    drive mechanism cooler 1  177263 11/01/73 Elementary diagram control rod    drive mechanism cooling fan    dampers 177270 11/01/73 Elementary diagram control    room filter fan motors  177275 11/01/73 Elementary diagram control room    filter intake dampers  177277 11/01/73 Elementary diagram reactor    cavity H2 dilution A/P    compressors 1A and 1B  177278 11/01/73 Elementary diagram containment    preaccess fan motors  177279 11/01/73 Elementary diagram control    room filter exhaust dampers  177280 11/01/73 Elementary diagram control    room outside air intake    dampers 177281 11/01/73 Elementary diagram penetration    room filter prefilter damper  177282 11/01/73 Elementary diagram refueling    water surface supply and    exhaust fan motors  177283 11/01/73 Elementary diagram penetration    room filter recirculation    damper 177284 11/01/73 Elementary diagram charging/    HHSI pump 1B room cooler fan    train B  177291 11/01/73 Elementary diagram radwaste air    exhaust fan 1B motor  177294 11/01/73 Elementary diagram miscellaneous    auxiliary building sump pump    motors 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 8 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177371 11/01/73 Elementary diagram solenoid    valves, sheet 75    Pressurizer liquid sample    train A    Pressurizer steam sample    train A    Reactor hot leg sample train A    Accumulator sample train A  177372 11/01/73 Elementary diagram solenoid    valves, sheet 76    Pressurizer liquid sample    train B    Pressurizer steam sample    train B    Reactor hot leg sample train B    Accumulator sample train B  177399 11/01/73 Elementary diagram accumulator    discharge valve closed alarm  177400 11/01/73 Elementary diagram accumulator    discharge valve closed alarm  177569 11/01/73 Elementary diagram 575-V motor-    operated valve  177570 11/01/73 Elementary diagram 575-V motor-    operated valve  177572 11/01/73 Elementary diagram 575-V motor-    operated valve  177583 11/01/73 Elementary diagram solenoid    valve, sheet 32    Motor-driven auxiliary    feedwater pump    Auxiliary feedwater bypass  177584 11/01/73 Elementary diagram solenoid    valves, sheet 31    Surge tank discharge to    auxiliary building  177588 11/01/73 Elementary diagram solenoid    valves, sheet 27    Spent fuel exhaust intake  177589 11/01/73 Elementary diagram solenoid    valves, sheet 26    Fuel handling area vent    system Penetration room dampers 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 9 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177590 11/01/73 Elementary diagram solenoid    valves, sheet 22    Turbine-driven auxiliary    feedwater pump discharge  177591 11/01/73 Elementary diagram solenoid    valves, sheet 23    Motor-driven auxiliary    feedwater pump discharge  177592 11/01/73 Elementary diagram solenoid    valves, sheet 24    Auxiliary steam condensate tank  177610 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 7    Reactor coolant pump    component cooling water return    from thermal barrier  177612 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 9  177613 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 10    Containment cooler service    water return    Containment cooler service    water supply  177617 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 14    Service water to blowdown heat    exchange Blowdown heat exchange, letdown    chiller discharge  177618 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 15    Reactor coolant pump    component cooling water    return from oil coolers  177620 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 17    Auxiliary feedwater pump    service water supply    containment leak rate test  177622 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 19    Steam generator feedwater    intake 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 10 OF 33)    Submittal to NRC  Drawing (formerly  Number    AEC)  Title 177623 11/01/73 Elementary diagram 575-V motor-  operated valves, sheet 20    Service water from storage      tank train A  177624 11/01/73 Elementary diagram 575-V motor-  operated valves, sheet 21    Service water from storage      tank train B  177625 11/01/73 Elementary diagram 575-V motor-  operated valves, sheet 22    Component cooling water to      reactor coolant pump  177628 11/01/73 Elementary diagram 575-V motor-      operated valves, sheet 25    Component cooling makeup water    Component cooling water to      spent fuel pool heat exchange Service water to component      cooling water heat exchange    Component cooling water to      residual heat exchange  177627 11/01/73 Elementary diagram 575-V motor-  operated valves,      sheet 24    Auxiliary feedwater to steam      generators 1A, 1B, and 1C  177629 11/01/73 Elementary diagram 575-V motor-      operated valves, sheet 26    Containment cooler service      water bypass    Containment cooler service      water discharge  177630 11/01/73 Elementary diagram 575-V motor-  operated valves, sheet 27    Component cooling water heat      exchange  177632 11/01/73 Elementary diagram 575-V motor-      operated valves, sheet 29    RHR pumps 1 and 1B miniflow  177633 11/01/73 Elementary diagram 575-V motor-  operated valves, sheet 30    Containment cooler discharge 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 11 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177635 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 32    Service water to component    cooling water heat exchange  177636 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 33    Reactor coolant pump motor    cooler service water    discharge 177644 11/01/73 Elementary diagram 575-V motor-    operated valve  177645 11/01/73 Elementary diagram loading    sequencer B1F essential    sequencer 177646 11/01/73 Elementary diagram loading    sequencer B1G essential    sequencer 177647 11/01/73 Elementary diagram essential    loading sequencer B1G breaker    close failure indication  177648 11/01/73 Elementary diagram essential    loading sequencer B1G breaker    close failure indication  177649 11/01/73 Elementary diagram loading    sequencer B1F LOSP    sequencer 177650 11/01/73 Elementary diagram loading    sequencer B1G LOSP    sequencer 177653 11/01/73 Elementary diagram loading    sequencer B1F load shedding    scheme 177654 11/01/73 Elementary diagram loading    sequencer B1G load shedding    scheme 177659 11/01/73 Elementary diagram loading    sequencer B2H load shedding    scheme 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 12 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177660 11/01/73 Elementary diagram loading    sequencer B2J load shedding    scheme 177688 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 47  177689 11/01/73 Elementary diagram 575-V motor-    operated valves, sheet 48  177838 11/01/73 Elementary diagram 575-V motor-    operated valve  177839 11/01/73 Elementary diagram 575-V motor-    operated valve  177840 11/01/73 Elementary diagram 575-V motor-    operated valve  177851 11/01/73 Elementary diagram solenoid    valves, sheet 2    Excess letdown heat exchange    inlet Excess letdown heat exchange    discharge 177852 11/01/73 Elementary diagram solenoid    valves, sheet 3    Surge tank discharge to    auxiliary building  177853 11/01/73 Elementary diagram solenoid    valves, sheet 4    Waste recycle evaporation    discharge and inlet valves  177854 11/01/73 Elementary diagram solenoid    valves, sheet 5    Reactor coolant pump component    cooling 177855 11/01/73 Elementary diagram solenoid    valves, sheet 6    Reactor coolant pump    component cooling water    thermal barrier return  177856 11/01/73 Elementary diagram solenoid    valves, sheet 7    Component cooling heat    exchange service water    discharge 177857 11/01/73 Elementary diagram solenoid    valves, sheet 8 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 13 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title Steam to turbine-driven    auxiliary feedwater pump  177863 11/01/73 Elementary diagram solenoid    valves train A, sheet 14    Main steam isolation valves  177864 11/01/73 Elementary diagram solenoid    valves, sheet 15    Main steam isolation bypass    valve train A  177865 11/01/73 Elementary diagram solenoid    valves, sheet 16    Main steam isolation valve    operator test  177866 11/01/73 Elementary diagram solenoid    valves, sheet 17    Main steam isolation bypass    valve train B  177867 11/01/73 Elementary diagram solenoid    valves, sheet 18    Main steam isolation valves    train B  177205 11/01/73 Elementary diagram spent fuel    pool pumps 1A and 1B  177224 11/01/73 Elementary diagram boric acid    transfer pumps 1A and 1B  177240 11/01/73 Elementary diagram boron    injection tank recirculation    pumps 1A and 1B  177174 11/01/73 Elementary diagram reactor    coolant pumps 1, 2, and 3  177180 11/01/73 Elementary diagram charging/    HHSI pumps 1A and 1C  177181 11/01/73 Elementary diagram charging/    HHSI pump 1B train A  177182 11/01/73 Elementary diagram charging/    HHSI pump 1B train B  177193 11/01/73 Elementary diagram RHR/LHSI    pumps 1A and 1B  177195 11/01/73 Elementary diagram containment    spray pumps 1A and 1B  177107 11/01/73 Elementary diagram pressurizer    heater backup group 1A    (600-V LC emergency bus 1A) 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 14 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177111 11/01/73 Elementary diagram pressurizer    heater backup group 1B    (600-V LC emergency bus 1C)  177364 11/01/73 Elementary diagram solenoid    valve, sheet 35    Letdown line isolation valve    Accumulator fill line isolation    valve Accumulator nitrogen supply    header isolation valve    Accumulator test line to    refueling water  177365 11/01/73 Elementary diagram solenoid    valve, sheet 36    Boron injection tank    recirculation isolation    valve Boron injection recirculation    pump to boron injection tank    isolation valve  177368 11/01/73 Elementary diagram solenoid    valve, sheet 34    Accumulator test line    isolation valve  177309 11/01/73 Elementary diagram boron    injection tank heaters A    and B  177313 11/01/73 Elementary diagram boron    injection surge tank heater  177375 11/01/73 Elementary diagram solenoid    valve, sheet 43    Letdown to demineralizer or    volume control tank valve  177376 11/01/73 Elementary diagram solenoid    valve, sheet 49    Letdown orifice isolation    valve 177377 11/01/73 Elementary diagram solenoid    valve, sheet 50    Letdown orifice isolation    valve 177378 11/01/73 Elementary diagram solenoid    valve, sheet 51 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 15 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title    Letdown orifice isolation    valve 177379 11/01/73 Elementary diagram solenoid    valve, sheet 42    Boric acid filter to boric    acid blender valve  177381 11/01/73 Elementary diagram solenoid    valve, sheet 45    Pressurizer power relief    valve 177382 11/01/73 Elementary diagram solenoid    valve, sheet 48    Pressurizer relief tank to    reactor Makeup water supply isolation    valve Pressurizer relief tank vent    to waste process system    isolation valve  177383 11/01/73 Elementary diagram solenoid    valve, sheet 46    Reactor coolant drain tank    pump discharge valve  177384 11/01/73 Elementary diagram solenoid    valve, sheet 47    Reactor coolant drain tank    vent isolation valve  177508 11/01/73 Elementary diagram solenoid    valve, sheet 53    Waste gas discharge control    valve 177509 11/01/73 Elementary diagram solenoid    valve, sheet 54    Boric acid makeup injection    valve to charging pump    heater 177510 11/01/73 Elementary diagram solenoid    valve, sheet 55    Boric acid dilution injection    valve to volume control    tank 177511 11/01/73 Elementary diagram solenoid    valve, sheet 56 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 16 OF 33)    Submittal to NRC  Drawing (formerly  Number    AEC)  Title Reactor makeup water to boric    acid blender valve  177567 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 41    Reactor coolant pump seal    water return isolation valve  177568 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 42    Containment spray pump to    spray nozzles isolation valve  177569 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 43    RHR system inlet isolation    valve 177570 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 44    RHR system outlet isolation    valve 177571 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 45    Low heat safety injection to    reactor coolant system cross-    over 177572 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 40    RHR system inlet isolation    valve 177585 11/01/73 Elementary diagram solenoid    valve, sheet 30    Letdown line isolation valve  177586 11/01/73 Elementary diagram solenoid    valve, sheet 29    Letdown line isolation valve  177587 11/01/73 Elementary diagram solenoid    valve, sheet 28    Letdown to volume control    tank 177593 11/01/73 Reactor coolant drain tank    pump discharge    Reactor coolant drain tank    vent Pressurizer relief tank 2    supply isolation valve 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 17 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 177602 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 39    Volume control tank outlet    isolation valve  177603 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 38    Refueling water storage tank    to charging pump valve  177604 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 37    Volume control tank outlet    isolation valve  177606 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 3    Charging/safety injection    pumps section heads isolation    valve Charging/safety injection pumps    discharge heater isolation    valve Refueling water storage tank to    RHR pumps 1A and 1B isolation    valve 177607 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 4    Containment sump to RHR pump    1B isolation valve    Containment sump to RHR pump    1A isolation valve    MMSI to reactor coolant system    hot leg    HHSI to reactor coolant system    hot leg    LHSI to reactor coolant system    hot leg    LHSI to reactor coolant system    cold leg  177608 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 5    Charging/safety injection    pumps miniflow isolation    valve 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 18 OF 33)    Submittal to NRC  Drawing (formerly Number    AEC)  Title Charging/safety injection    pumps to reactor coolant    system isolation valve  177609 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 6    Accumulator 1A, 1B, and 1C    discharge 177614 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 11    Boron injection tank outlet    isolation valve    Boron injection tank inlet    isolation valve  177615 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 12    Pressurizer power relief    isolation valve  177631 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 28    Refueling water storage tank    to charging pump  177634 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 31    Reactor coolant pumps seal    water return isolation valve  177637 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 34    Spray additive tank outlet    isolation valve  177638 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 35    Containment spray pump inlet  177639 11/01/73 Elementary diagram 575-V motor-    operated valve, sheet 36    Containment sump outlet valve  177858 11/01/73 Elementary diagram solenoid    valve, sheet 9    Excess letdown isolation    valve 177861 11/01/73 Elementary diagram solenoid    valve, sheet 12    Reactor coolant system normal    charging line 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 19 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title Reactor coolant system    alternate charging line  177362 11/30/73 Elementary diagram solenoid    valve, sheet 77  177373 11/30/73 Elementary diagram solenoid    valve, sheet 78  177374 11/30/73 Elementary diagram solenoid    valve, sheet 79  177523 11/30/73 Elementary diagram solenoid    valve, sheet 80 Elementary Diagrams and Physical Drawings, 172000 Series 172062 02/01/74 Conduit template 600-V    switchgear buses 1H and 1J  172100 02/01/74 Outdoor duct runs general    arrangement 172101 02/01/74 Outdoor electrical duct runs    profile river duct 1A  172102 02/01/74 Outdoor electrical duct runs    profile river duct 1B  172103 02/01/74 Outdoor electrical duct runs    profile service water duct 1A  172104 02/01/74 Outdoor electrical duct runs    profile service water duct 1E  172239 02/01/74 Details and assembly of service    water undervoltage    Detector cabinet service    water battery  172240 02/01/74 Details and assembly of service    water battery fuse boxes  172270 02/01/74 Electrical penetrations of    river water and service water    intake structure  172285 02/01/74 Class 1 cable tray support post  172290 02/01/74 Compression type cable transit    river and service water    intake structures  172292 02/01/74 Class 1 cable tray support    bracket 172328 02/01/73 Bill of material service water    intake structure  172329 02/01/74 Bill of material river water    intake structure 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 20 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172338 02/01/74 Bill of material undervoltage    detector cabinet    Service water intake structure    batteries 172366 02/01/74 125 V-dc distribution cabinet    service water intake    structure train A  172367 02/01/74 125 V-dc distribution cabinet    service water intake    structure train B  172369 02/01/74 120/208-V distribution cabinet    river water intake structure    train A  172370  02/01/74 120/208-V distribution cabinet    river water intake structure    train B  172371 02/01/74 120/208-V distribution cabinet    service water intake    structure train A  172372 02/01/74 120/208-V distribution cabinet    service water structure    train B  172373 02/01/74 Anchor bolt assembly for cable    tray support post  172063 03/01/74 Conduit template 600-V    switchgear buses 1R and 1S  172064 03/01/74 Conduit template 4160-V    switchgear buses 1H and 2H  172065 03/01/74 Conduit template 4160-V    switchgear buses 1J and 2J  172143 03/01/74 Outdoor ducts Class 1 diesel    building to valve boxes and    fuel oil tank  172155 05/15/74 Sections and details Class 1    ducts, diesel building area  172169 03/01/74 Diesel building lightning    protection and roof grounding  172170 03/01/74 Grounding plan, diesel building    to valve boxes  172171 03/01/74 Electrical equipment plan,    diesel building  172172 03/01/74 Electrical sections and details,    diesel building, sheet 1  172195 05/15/74 Diesel building slab 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 21 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172196 05/15/74 Electrical section and details    of conduit below slab,    diesel building  172197 05/15/74 Embedded supports and conduits    in wall  172211 05/15/74 Cable tray layout and    exposed conduit, diesel    building, sheet 1  172212 05/15/74 Cable tray layout and    exposed conduit, diesel    building, sheet 2  172213 05/15/74 Cable tray layout and    exposed conduit, diesel    building, sheet 3  172214 05/15/74 Cable tray layout and    exposed conduit, diesel    building, sheet 4  172230 05/15/74 Enlarged end and cable tray,    partial plan, river intake    structure, sheet 1  172231 05/15/74 Enlarged end and cable tray,    partial plan, river intake    structure, sheet 2  172232 05/15/74 Conduit plan for valve    boxes 172233 05/15/74 Conduit plan valve box river    water supply  172173 03/01/74 Electrical sections and details,    diesel building, sheet 2  172174 03/01/74 Electrical sections and details,    diesel building, sheet 3  172178 03/01/74 Electrical sections and details,    diesel building, sheet 4  172195 03/01/74 Embedded conduit, diesel    building slab  172196 03/01/74 Sections and details of    conduit below slab, diesel    building 172197 03/01/74 Embedded supports and conduit    in walls, diesel building  172203 03/01/74 Diesel building cable tray and    support plan, sheet 1  172211 03/01/74 Cable tray layout, diesel    building, sheet 1 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 22 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172212 03/01/74 Cable tray layout, diesel    building, sheet 2  172213 03/01/74 Cable tray layout, diesel    building, sheet 3  172214 03/01/74 Cable tray layout, diesel    building, sheet 4  172243 03/01/74 Screen enclosure for diesel    generator neutral resistor  172264 03/01/74 Details and assembly Class 1    emergency ventilation station  172265 03/01/74 Details and assembly Class 1    ventilation local control    station 172266 03/01/74 Details and assembly Class 1    heater local control    station 172204 02/01/74 General arrangement cable    tunnel, sheet 1  172205 02/01/74 General arrangement cable    tunnel, sheet 2  172206 02/01/74 General arrangement cable    tunnel, sheet 3  172232 04/05/74 Conduit plan valve boxes,    sheet 1  172233 04/05/74 Conduit plan valve boxes,    sheet 2  172234 04/05/74 Conduit plan valve boxes,    sheet 3  172311 03/01/74 Bill of material cable trays  172312 03/01/74 Bill of material cable tray    supports 172313 03/01/74 Bill of material cable tunnel  172314 03/01/74 Bill of material diesel    building 172340 02/01/74 Details and assembly of    switchgear channels  172384 03/01/74 120/208-V distribution    cabinet diesel 1C  172385 03/01/74 120/208-V distribution    cabinet diesel 2C  172386 03/01/74 120/208-V distribution    cabinet diesel 1-2A  172387 03/01/74 120/208-V distribution    cabinet diesel 1B 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 23 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172700 11/16/73 Main single line diagram    generator and 4160-V    transformer 172701 11/16/73 Single line 4160-V    emergency station service  172202 11/16/73 Single line 4160-V    emergency station service  172204 11/16/73 Single line 600-V emergency    station service  172207 11/16/73 Single line and cable diagram    dc distribution train E    service water building  172708 11/16/73 Single line and cable diagram    dc distribution train E    service water building  172713 11/16/73 Bill of material relay panels    1 through 11  172714 11/16/73 Front view meter and relay    panels 1 through 11  172723 12/21/73 Elementary diagram turbine    auxiliary auto stop trips    and emergency trip and    vacuum reset  172732 11/16/73 Elementary diagram generator    relaying 172741 11/16/73 Elementary diagram fire    protection jockey pump  172744 11/16/73 Wiring diagram DEH valve test    panel junction boxes 1 and 2  172745 11/16/73 Elementary diagram station    service air compressor 1A  172747 01/04/74 Elementary diagram service    water pump 1A  172748 01/04/74 Elementary diagram service    water pump 1B  172749 01/04/74 Elementary diagram service    water pump 1C (bus 1K)  172750 01/04/74 Elementary diagram service    water pump 1C (bus 1L)  172751 01/04/74 Elementary diagram service    water pump 1D  172752 01/04/74 Elementary diagram service    water pump 1E 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 24 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172761 01/04/74 Elementary diagram 4160-V    bus 1H incoming breaker    from diesel generator 1C  172762 11/30/73 Elementary diagram 4160-V    bus 1H feeder breakers to    station service transformers    1H and 1R  172763 01/18/74 Elementary diagram 4160-V    bus 1J incoming breaker    from diesel generator 2C  172764 11/30/73 Elementary diagram 4160-V    bus 1J (emergency) feeder    breaker to station service    transformers 1J and 1S  172765 11/30/73 Elementary diagram 4160-V    bus 1K feeder breaker    station service    transformer 1K  172766 11/30/73 Elementary diagram 4160-V    bus 1E (emergency) feeder    breaker to station    service transformer 1L  172767 01/31/74 Elementary diagram 600-V    buses 1G, 1P, and 1Q    incoming breaker  172768 01/31/74 Elementary diagram feeder    breaker 600-V buses    1G, 1P, and 1Q  172769 01/31/74 Elementary diagram 600-V    buses 1G, 1P, and 1Q bus tie    breaker from bus 1F  172770 11/16/73 Bill of material diesel    generator relay panels  172771 11/16/73 Front view diesel generator    relay panels typical for    1-2A, 1B, 2B, 1C, and 2C  172772 01/31/74 Elementary diagram diesel    generator 1-2A relaying  172773 01/31/74 Elementary diagram diesel    generator 1-2A metering  172774 02/28/74 Elementary diagram diesel    generator 1-2A start, stop,    and shutdown 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 25 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172775 02/28/74 Elementary diagram diesel    generator 1-2A exciter and    miscellaneous controls  172776 01/31/74 Elementary diagram diesel    generator 1B relaying  172777 01/31/74 Elementary diagram diesel    generator 1B metering  172778 02/28/74 Elementary diagram diesel    generator 1B start, stop,    and shutdown  172779 02/28/74 Elementary diagram diesel    generator 1B exciter and    miscellaneous controls  172780 01/31/74 Elementary diagram diesel    generator 1C relaying  172781 01/31/74 Elementary diagram diesel    generator 1C metering  172782 02/15/74 Elementary diagram diesel    generator 1C start, stop,    and shutdown  172783 02/28/74 Elementary diagram diesel    generator 1C exciter and    miscellaneous controls  172784 11/16/73 Elementary diagram generator    and transformer auxiliary    relays 172787 11/16/73 Elementary diagram startup    auxiliary transformers 1A and    1B protective relaying  172791 01/31/74 Elementary diagram diesel    generator 2C relaying  172792 01/31/74 Elementary diagram diesel    generator 2C metering  172793 02/28/74 Elementary diagram diesel    generator 2C start, stop,    and shutdown  172794 02/28/74 Elementary diagram diesel    generator 2C exciter and    miscellaneous controls  172795 11/30/73 Elementary diagram 4160-V    bus 1H feeder breaker to    station service trans-    former 1G 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 26 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172796 11/30/73 Elementary diagram 4160-V    bus 1H differential relaying  172797 11/30/73 Elementary diagram 4160-V    bus 1J differential relaying  172798 11/30/73 Elementary diagram 4160-V    bus 1K differential relaying  172799 11/30/73 Elementary diagram 4160-V    bus 1L differential relaying  172818 01/04/74 Elementary diagram river and    service water motor-operated    and solenoid-operated valves  172825 01/31/74 Elementary diagram 600-V    buses 1H, 1J, 1K, and 1L    auxiliary breakers and    potential transformers,    sheet 1  172826 01/31/74 Elementary diagram 600-V    buses 1H, 1J, 1K, and 1L    feeder breakers and bus tie    breakers, sheet 2  172827 01/31/74 Elementary diagram 600-V    buses 1O, 1P, and 1Q    potential transformers  172828 11/30/73 Elementary diagram 4160-V    bus 1H potential transformer  172829 11/30/73 Elementary diagram 4160-V    bus 1J potential transformer  172830 11/30/73 Elementary diagram 4160-V    buses 1K and 1L potential    transformer 172831 01/31/74 Elementary diagram 600-V    buses 1R and 1S, sheet 1  172832 01/31/74 Elementary diagram 600-V    buses 1R and 1S, sheet 2  172852 01/04/74 Elementary diagram startup    auxiliary transformers    1A and 1B controls  172857 01/04/74 Elementary diagram motor    control center 1K  172858 03/22/74 Elementary diagram motor    control center 1L  172860 03/22/74 Elementary diagram motor    control center 1N 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 27 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172861 03/22/74 Elementary diagram motor    control center 1P  172862 03/22/74 Elementary diagram motor    control center 1S  172863 03/22/74 Elementary diagram motor    control center 1T  172864 03/22/74 Elementary diagram motor    control center 1X  172865 03/22/74 Elementary diagram motor    control center 1Y  172868 11/16/73 Wiring diagram fire protection    engine-driven fire pumps  172869 11/16/73 Elementary diagram motor-    driven fire pump  172870 02/28/74 Single line and cable diagram    fire protection 600-V and    120/208-V distribution    cabinets 172875 01/04/74 Elementary diagram river    water pump 4  172876 01/04/74 Elementary diagram river    water pump 5  172877 01/04/74 Elementary diagram river    water pump 8  172878 01/04/74 Elementary diagram river    water pump 9  172879 01/04/74 Elementary diagram river    water pump 10  172960 01/31/74 Elementary diagram motor-    operated valves diesel    generator cooling  172963 02/15/74 Elementary diagram diesel    generator storage tank fuel    pumps 172973 01/31/74 Elementary diagram diesel    generator 2B relaying  172974 01/31/74 Elementary diagram diesel    generator 2B metering  172975 02/28/74 Elementary diagram diesel    generator 2B start, stop,    and shutdown  172976 02/28/74 Elementary diagram diesel    generator 2B exciter and    miscellaneous controls 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 28 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 6893D82 08/15/74 Elementary diagram reactor trip switchgear  172713 10/16/74 Bill of material for D-172714    front view meter and relay    PNLS, sheet 7  172713 10/16/74 Bill of material for D-172714    front view meter and relay    PNLS, sheet 8  172723 10/16/74 Elementary diagram river    water pumps cooling and lube    water strainers  172732 10/16/74 Elementary diagram generator    relaying 172744 10/16/74 Wiring diagram DEH valve    test panel junction boxes    1 and 2  172761 10/16/74 Elementary diagram 4160-V    bus 1H incoming breaker from    diesel generator 1C  172763 10/16/74 Elementary diagram 4160-V    bus 1J incoming breaker from    diesel generator 2C  172770 10/16/74 Bill of material for C-172771    front view diesel generator    relay PNLS, sheet 4  172825 10/16/74 Elementary diagram 600-V    buses 1H, 1J, 1K, and 1L    incoming breaker and    potential transformer,    sheet 2  172857 10/16/74 Elementary diagram motor    control center 1K (service    water intake structure)  172858 10/16/74 Elementary diagram motor    control center 1L (service    water intake structure)  172860 10/16/74 Elementary diagram motor    control center 1N (diesel    building) 172861 10/16/74 Elementary diagram motor    control center 1P (diesel    building) 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 29 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 172862 10/16/74 Elementary diagram motor    control center 1S (diesel    building) 172863 10/16/74 Elementary diagram motor    control center 1T (diesel    building) 172864 10/16/74 Elementary diagram motor    control center 1X (river    water intake structure)  172865 10/16/74 Elementary diagram motor    control center 1Y (river    water intake structure)  172963 10/16/74 Elementary diagram diesel    generator storage tank fuel    pumps, sheet 1  172963 10/16/74 Elementary diagram diesel    generator storage tank fuel    pumps, sheet 2 Location Drawings, 175000 Series 175055 11/01/73 Equipment location auxiliary    building area plan at    el 155 ft  175056 11/01/73 Equipment location auxiliary    building area plan at    el 139 ft  175057 11/01/73 Equipment location auxiliary    building area plan at    el 121 ft  175059 08/15/74 Equipment location auxiliary    building roof plan at    el 175 ft and above  175061 11/01/73 Equipment location auxiliary    and control building area    plan at el 139 ft  175062 11/01/73 Equipment location auxiliary    and control building  175150 11/01/73 Instrumentation location    containment and fuel handling    area plan at el 105 ft 6 in. 175140 11/01/73 Instrumentation location    auxiliary and control    building area at el 155 ft 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 30 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 175141 11/01/73 Instrumentation location    auxiliary and control    building area plan at    el 139 ft  175142 11/01/73 Instrumentation location    auxiliary and control    building area at el 121 ft  175143 11/01/73 Instrumentation location    auxiliary and control    building area at el 100 ft    and below  175144 11/01/73 Instrumentation location    auxiliary building area    at el 155 ft  175145 11/01/73 Instrumentation location    auxiliary building area at    el 139 ft  175146 11/01/73 Instrumentation location    auxiliary building area at    el 121 ft  175147 11/01/73 Instrumentation location    auxiliary building area at    el 100 ft and below  175148 11/01/73 Instrumentation location    containment and fuel    handling area at    el 155 ft  175149 11/01/73 Instrumentation location    containment and fuel    handling area at    el 129 ft Piping and Instrumentation Drawings, 170000 Series; Instrument Installation Drawings, 170000 Series 170119 11/30/73 P&ID river water system,    sheet 1  170119 11/30/73 P&ID service water system,    sheet 2  170060 08/29/73 P&ID diesel generator fuel    oil supply system (deleted)   
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 31 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 170586 10/16/74 River water system automatic    operations train A  170587 10/16/74 River water system automatic    operations train B  170588 10/16/74 River water pump QSP25P004B  170589 10/16/74 River water pump QSP25P005B  170590 10/16/74 River water pump QSP25P008A  170591 10/16/74 River water pump QSP25P009A  170592 10/16/74 River water pump QSP25P010A  170593 10/16/74 River water motor-operated    valve logic train A  170594 10/16/74 River water system valves    train B  170623 10/16/74 River water lube water    cyclone separator inlet    motor-operated valves  170624 10/16/74 River water hand switch-    operated motor-operated    valves (typical)  170599 10/16/74 Service water pump 1A    train A  170600 10/16/74 Service water pump 1B    train A  170601 10/16/74 Service water pump 1C    train A or B  170602 10/16/74 Service water pump 1D    train B  170603 10/16/74 Service water pump 1E    train B  170604 10/16/74 Service water diesel    generator 2C Unit 1    train B isolation motor-    operated valves  170605 10/16/74 Service water diesel    generator 2C Unit 2    train B isolation motor-    operated valves  170606 10/16/74 Service water diesel    generator 1B Unit 1    train B isolation motor-    operated valves 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 32 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 170607 10/16/74 Service water diesel    generator 1B Unit 2    train B isolation motor-    operated valves  170608 10/16/74 Service water diesel    generator 2B Unit 1    train B isolation motor-    operated valves  170609 10/16/74 Service water diesel    generator 1C Unit 1    train A isolation motor-    operated valves  170610 10/16/74 Service water diesel    generator 1C Unit 2    train B isolation motor-    operated valves  170611 10/16/74 Service water diesel    generator 1-2A Unit 1    train A isolation motor-    operated valves  170612 10/16/74 Service water diesel    generator 1-2A Unit 2    train B isolation motor-    operated valves  170613 10/16/74 Service water diesel    generator building    train B isolation motor-    operated valves  170614 10/16/74 Service water diesel    generator building    train A isolation motor-    operated valves  170615 10/16/74 Service water to turbine    building isolation motor-    operated valves 514 and 516  170616 10/16/74 Service water to turbine    building isolation motor-    operated valves 515 and 517  170617 10/16/74 Service water trains A and B    strainer isolation motor-    operated valves 
FNP-FSAR-7 REV 21  5/08 TABLE 7.1-1 (SHEET 33 OF 33)    Submittal    to NRC  Drawing (formerly  Number    AEC)  Title 170618 10/16/74 Service water trains A and B    emergency recirculation    to pond motor-operated    valves 170619 10/16/74 Service water push button-    operated motor-operated    valves (typical)  170622 10/16/74 Service water lube water    cyclone separator inlet    motor-operated valves  170625 10/16/74 Service water hand switch-    operated motor-operated    valves (typical)  170626 10/16/74 Service water system discharge    backpressure control valves] 
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HISTORICALDuring preliminary startup tests, it will be demonstrated that actual instrument errors and time delays are equal to or less than the values assumed in the accident analyses.
FNP-FSAR-7  REV 21  5/08 [HISTORICAL] [TABLE 7.2-3 (SHEET 1 OF 2)  REACTOR TRIP SYSTEM INSTRUMENT ACCURACIES  Reactor Trip Signal Accuracy Note Power range high neutron +/- 1 percent of full power  flux Intermediate range high +/- 5 percent of full scale (a) neutron flux    +/- 1 percent of full scale (a)  from 10-4 to 10-3  amperes 
Source range high +/- 5 percent of full scale (a) neutron flux Power range high +/- 5 percent (a) positive nuclear  power rate Power range high +/- 5 percent  negative nuclear  power rate Overtemperature T +/- 3.2&deg;F    (a)
Overpower T +/- 2.7&deg;F Pressurizer low +/- 18 psi  pressure Pressurizer high +/- 14 psi  pressure Pressurizer high +/- 2.3 percent of full  water level range P between taps at  design temperature and  pressure 
Low reactor coolant +/- 2.5 percent of full flow (a) flow within range of 70  percent of 100 percent  of full flow Reactor coolant pump +/- 1 percent of relay set  bus undervoltage voltage 
FNP-FSAR-7  REV 21  5/08 TABLE 7.2-3 (SHEET 2 OF 2)  Reactor Trip Signal Accuracy Note Reactor coolant pump +/- 0.1 Hz  bus underfrequency Low-low steam generator +/- 2.9 percent of P signal  water level over pressure range of  600 to 1100 psig  (this does not include EA  and PMA allowances) 
a. Reproducibility.]
FNP-FSAR-7  REV 25  4/14 TABLE 7.2-4 (SHEET 1 OF 3)  TRIP CORRELATION Reactor Trip Accident(a)  Source range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B)  15.2.4-Boron dilution (B) 15.4.6-Rod ejection (B) Intermediate range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B)  15.2.4-Boron dilution (B)  15.4.6-Rod ejection (B)  Power range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal(low setpoint) from a subcritical condition (P)  15.2.4-Boron dilution (P)  15.2.6-Startup of an inactive reactor coolant loop (B) 15.2.10-Excessive heat removal due to  feedwater system malfunction (B)  15.2.11-Excessive load increase (B)  15.4.6-Rod ejection (P) Power range, high flux 15.2.2-Uncontrolled RCCA bank withdrawal(high setpoint) at power (P) 15.2.4-Boron dilution (B) 15.2.6-Startup of an inactive reactor coolant loop (B) 15.2.10-Excessive heat removal due to feed- water system malfunction (B)  15.2.11-Excessive load increase (B)  15.4.6-Rod ejection (P)
Positive neutron flux rate 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B)  15.2.2-Uncontrolled RCCA bank withdrawal at power (P)  15.4.6-Rod ejection (B)
a. (B/P) - Backup/Primary trip designation based on FSAR Chapter 15 analysis.
FNP-FSAR-7  REV 25  4/14 TABLE 7.2-4 (SHEET 2 OF 3)  Reactor Trip Accident Overpower T  15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.4-Boron dilution (B) 15.2.5Partial loss of forced reactor coolant system flow (B)  15.2.10-Excessive heat removal due to  feedwater system malfunction (B)  15.2.11-Excessive load increase (B)  15.4.2-Main steam line break (P - at power)  Overtemperature T 15.2.2-Uncontrolled RCCA bank withdrawal at power (P) 15.2.4-Boron dilution (P) 15.2.5Partial loss of forced reactor coolant system flow (B)  15.2.7-Loss of external electric load and/or turbine trip (P) 15.2.10-Excessive heat removal due to  feedwater system malfunction (B)  15.2.11-Excessive load increase (B)  15.2.12-Accidental depressurization of the  reactor coolant system (P)  15.3.6-Single RCCA withdrawal at power (P) 15.4.2-Feedline break (B) 15.4.3-Steam generator tube rupture (B)
Low primary coolant flow 15.2.5-Partial loss of forced reactor  coolant system flow (P)  15.3.4-Complete loss of forced reactor  coolant system flow (P)  15.4.4-Single reactor coolant pump locked rotor (P)
Reactor coolant pump, under- 15.3.4-Complete loss of forced reactor frequency or undervoltage coolant system flow (B) 
Pressurizer high pressure 15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.7-Loss of external electrical load  and/or turbine trip (P)
FNP-FSAR-7  REV 25  4/14 TABLE 7.2-4 (SHEET 3 OF 3)  Reactor Trip Accident  15.4.2-Feedline break (B) 15.4.4-Single reactor coolant pump locked rotor (B)Pressurizer high water level 15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.4-Boron dilution (B) 15.2.7-Loss of external electrical load  and/or turbine trip (B) 15.2.8-Loss of normal feedwater (B)  15.2.9-Loss of offsite power to the station  auxiliaries (station blackout) (B)  15.2.14-Inadvertent operation of ECCS  during power operation (B)  15.4.2-Feedline break (B)
Pressurizer low pressure 15.2.3-RCCA misalignment (one or more  dropped RCCAs) (B) 15.2.11-Excessive load increase (B)  15.2.12-Accidental depressurization of the  reactor coolant system (B)  15.2.14-Inadvertent operation of ECCS  during power operation (P)  15.3.1-Loss of reactor coolant from small  ruptured pipes or from cracks in  large pipes which actuate emergency core cooling system (small break  LOCA (P) 15.4.2-Main steam line break (B)  15.4.3-Steam generator tube rupture (P)
Low-low steam generator 15.2.7-Loss of external electrical load water level and/or turbine trip (B) 15.2.8-Loss of normal feedwater (P)  15.2.9-Loss of offsite power to the station  auxiliaries (station blackout) (P)  15.4.2-Feedline break (P)
Reactor trip from safety injection 15.4.2-Main steam line break (P - at power) signal (low steam line pressure)
FNP-FSAR-7  REV 25  4/14 TABLE 7.2-5 (SHEET 1 OF 2)  REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES  Functional Unit Response Time(s)  1. Manual reactor trip NA    2. Power range, neutron flux  a. High  0.5 (a)  b. Low  0.5 (a)    3. Power range, neutron flux,  0.65 (a)  high positive rate    4. Not used. 5. Intermediate range, neutron fluxNA    6. Source range, neutron flux NA    7. Overtemperature T  (a)(b)(c)    8. Overpower T ~ (b)(c)    9. Pressurizer pressure-low  2.0    10. Pressurizer pressure-high  1.0    11. Pressurizer water level-highNA    12A. Loss of flow-single loop (above P-8) 1.0    12B. Loss of flow-two loops (above P-7  1.0  and below P-8)    13. Steam generator water level-low-low 2.0    14. Undervoltage-reactor coolant pumpsNA    15. Underfrequency-reactor coolant pumpsNA    16. Turbine trip  a. Low auto stop oil pressureNA  b. Turbine throttle valve closureNA    17. Safety injection input from ESFNA     
FNP-FSAR-7  REV 25  4/14 TABLE 7.2-5 (SHEET 2 OF 2)  Functional Unit Response Time(s)          18. Reactor trip system interlocksNA    19. Reactor trip breakers NA    20. Automatic trip logic NA
a. Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel. 
: b. RTD response time  5.0 s The RTD response time cannot be summed with the channel response times listed in Note (c). 
: c. The following are the required RTS channel response times (encompassing channel electronics/trip logic & breaker/gripper release) for an RTD response time of no greater than 5.0 seconds:
1. Overtemperature T, Tavg input:  2.435 s  2. Overtemperature T, pressurizer pressure input (including sensor):  2.0 s  3. Overtemperature T, nuclear flux input:  2.0 s  4. Overpower T, Tavg input:  2.159 s  5. T input (to both OTT and OPT):  6.159 s  Tavg and T response times include the effect of all transfer functions set to the recommended values. 
REV 21  5/08 SETPOINT REDUCTION FUNCTION FOR OVERPOWER AND OVERTEMPERATURE T TRIPS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-1 REV 21  5/08 PRESSURIZER SEALED REFERENCE LEG LEVEL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-2 REV 21  5/08 DESIGN TO ACHIEVE ISOLATION BETWEEN CHANNELS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-3 
[HISTORICAL][Actuation signal accuracies required for generating the required actuation signals for loss of coolant protection are as follows:  Pressurizer pressure  +/-14 psi (uncompensated)  Actuation signal accuracies required in generating the required actuation signals for steam break protection are given:  1. Steam line pressure 4 percent  2. Steam flow signals  4.5 percent of maximum guarantee flow over pressure range (600 to 1100 psig)  3. Tavg  2&#xba;F   
: 4. Containment pressure 1.8 percent signal of full scale
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-1 (SHEET 1 OF 2)  FUNCTIONS INITIATED BY ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Item Function 1 Reactor trip, provided one has not already been generated by the reactor trip system 
2 Engineered safety features actuation system sequence, which actuates equipment that includes items 2a through 2g and ensures the proper sequencing of engineered safety features power demands on the engineered safety features buses supplied by either preferred or standby power supply 2a Cold leg injection isolation valves, which are opened for injection of borated water by safety injection pumps into the cold legs of the reactor coolant system. The receipt of a safety injection signal by the accumulator motor-operated valves is discussed in paragraph 6.3.2.2.7.
2b Charging pumps, residual heat removal pumps, and associated valving, which provide emergency makeup water to the cold leg of the reactor coolant system following a loss-of-coolant accident 2c Containment air recirculation fans and coolers, which serve to cool the containment and limit the potential for release of fission products from the containment by reducing the pressure following an accident 2d Component cooling pumps and valves 
2e Service water pump and valves, which provide cooling water to the component cooling system heat exchangers and is thus the heat sink for containment cooling 2f Motor-driven auxiliary feedwater pumps and control valves 
2g Penetration room filtration system 
3 Phase A containment isolation, "T" signal, whose function is to prevent fission product release by isolating all nonessential process lines on receipt of the safety injection signal 4 Steam line isolation, to prevent the continuous, uncontrolled blowdown of more than one steam generator and thereby uncontrolled reactor coolant system cooldown 5 Main feedwater line isolation, to limit the energy release for a steam line break and to limit the extent of the reactor coolant system cooldown 6 Emergency diesel start, to ensure backup supply of power to emergency and supporting systems components 7 Control room intake duct isolation, to meet control room occupancy requirements following a loss-of-coolant accident FNP-FSAR-7  REV 21  5/08 TABLE 7.3-1 (SHEET 2 OF 2)
Item Function 
8 Containment spray actuation, "P" signal, which performs the following functions listed as items 8a and 8b 8a Containment spray initiation, which serves to reduce containment pressure and temperature following a loss-of-coolant or a steam break accident 8b Phase B containment isolation initiation, other than safety injection lines which are not closed. The remaining process lines into containment are isolated following a loss of reactor coolant accident or a steam or feedwater line break within containment.   
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-2 (SHEET 1 OF 2)  INSTRUMENTATION OPERATING CONDITIONS FOR ENGINEERED SAFETY FEATURES    Number of  Number of Channels Number Functional Unit Channels (a) to trip      1. Safety Injection      1a. Manual(a) 2 switches 1 switch        1b. Containment pressure high 3 2      1c. Differential pressure high between steam lines 9 (3 per steam line) 2 per steam line and 1/3    comparison between steam lines    1d. Pressurizer low pressure(b) 3 2    1e. Steam line low pressure(c) 3 pressure signals 2    2. Containment Spray      2a. Manual(a) 2 pairs of switches 2 switches per pair    2b. Containment pressure high-high-high 4 2    3. Auxiliary feedwater      3a. Motor driven pumps      3a1 Manual(d) 2 switches (1 switch per pump) 2 switches (1 switch per pump)    3a2 Steam generator water level low-low 3 per steam generator 2/3 in any steam generator    3a3 Safety injection See item 1      3a4 Trip of main feedwater pumps 4 (2 per pump) 2 (1 per pump)
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-2 (SHEET 2 OF 2)        Number of  Number of Channels Number Functional Unit Channels (a) to trip        3a5 AMSAC actuation 1 1    3b Turbine driven pump      3b1 Manual(e) 1 switch 1 switch    3b2 Steam generator water level low-low 3 per steam generator 2/3 in 2/3 steam generators    3b3 RCP bus undervoltage 3 bus 2 bus    3b4 AMSAC actuation 1 1   
a. Each switch actuates both Train A & B. b. Permissible bypass if reactor coolant pressure is less than P-11. c. Permissible bypass if reactor coolant temperature is less than P-12. d. Motor driven AFW pump 1 switch actuates Train A and motor driven AFW pump 2 switch actuates Train B. e. Turbine driven AFW pump switch opens steam admission valves.
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-3 (SHEET 1 OF 2)  INSTRUMENTATION OPERATING CONDITIONS FOR ISOLATION FUNCTIONS    Number of    Number of Channels Number Functional Unit Channels  to Trip      1. Containment Isolation      1a. Safety injection - Phase A See item 1 of table 7.3-2.      1b. Containment pressure high-high-high - Phase B See item 2b of table 7.3-2.      1c. Manual        Phase A (a) 2 1  Phase B See item 2a of table 7.3-2.      2. Steam Line Isolation      2a. Steam flow high coincident with low-low Tavg                Steam flow high 2 per steam line 1 high flow per steam line    on 2/3 steam lines      Low-low Tavg 1 per loop 2/3 low-low Tavg    2b. Steam line low pressure 1 per steam line 2/3 steam lines    2c. Containment pressure high-high 3 2              2d. Manual(a) 1 per loop 1 per loop                                          a. Each switch actuates both Train A & B.
FNP-FSAR-7  REV 21  5/08  TABLE 7.3-3 (SHEET 2 OF 2)      Number of    Number of Channels Number Functional Unit Channels  to Trip        3. Feedwater line isolation      3a. Safety injection See item 1 of table 7.3.2      3b. Steam generator water level high-high 3 per steam generator 2/3 in any steam generator    3c. Low Tavg coincident with reactor trip        Low Tavg 1 per loop 2/3 low Tavg  Reactor trip 2 1    4. Turbine Trip      4a. Safety injection See item 1 of table 7.3-2      4b. Steam generator water level high-high See item 3b See item 3b    4c. Reactor Trip 2 1    5. Steam generator feedwater pump trip (a)      5a. Safety injection See item 1 of table 7.3-2      5b. Steam generator water level high-high See item 3b See item 3b       
a. Train A trips both feedwater pumps.
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-4 (SHEET 1 OF 2)  INTERLOCKS FOR ENGINEERED SAFETY FEATURES ACTUATION SYSTEM    Function Designation Input Performed    P-4(a)  Reactor trip Actuates turbine trip        Prevents opening of main    feedwater valves which    were closed by safety    injection or high steam    generator water level        Allows manual block of    the automatic reactuation    of safety injection        Blocks steam dump control    via load rejection Tavg    controller        Makes steam dump valves    available for either    tripping or modulation      Reactor not Defeats the manual  tripped block preventing    automatic reactuation    of safety injection        Block steam dump control    via plant trip Tavg    P-11  2/3 pressurizer Allows manual block of  pressure below safety injection actuation  setpoint on low pressurizer    pressure signal        Blocks automatic opening    of the power relief valves      2/3 pressurizer Defeats manual block of  pressure above safety injection actuation  setpoint    Opens accumulator motor-operated isolation valves 
___________________  a. See table 7.7-1 for control functions FNP-FSAR-7  REV 21  5/08 TABLE 7.3-4 (SHEET 2 OF 2)    Function Designation Input Performed    P-12  2/3 Tavg below Allows manual block of  setpoint(a) safety injection    actuation on low    steam line pressure        Blocks steam dump    Allows manual bypass    of steam dump block for    the cooldown dump    valves only      2/3 Tavg above Defeats the manual  setpoint block of safety    injection actuation    on low steam line    pressure        Defeats the manual    bypass of steam dump    block    P-14  2/3 steam Closes all feedwater  generator water control valves  level above    setpoint on any Trips all main  steam generator feedwater pumps which    closes the pump    discharge valves        Actuates turbine trip 
__________________  a. This signal, in coincidence with high steam line flow in 2/3 steam lines, actuates steam line isolation.
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-5 
(THIS TABLE INTENTIONALLY DELETED) 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 1 OF 15)  FAILURE MODE AND EFFECTS ANALYSIS, SERVICE WATER SYSTEM  Component Identification  Service Water Pumps Logic Diagram Number          NA          Elementary Number D-172747 through D-172752, Engineering Flow Diagram Number:  D-202747 through D-202752 D-170119 Sh. 1 & D-200013 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    Loss of 4.16-kV Failure of pump to start 4.16-kV bus U/V or bus breaker Redundant train pumps can be  bus power when required or automatic auto trip alarm on emergency started  stoppage when pump is power board  running          Loss of 4.16-kV Interruption of service Breaker automatic trip Two pumps are provided per train; power to motor water supply to one alarm on main control the standby pump will be started due to train board from the main control board automatic    breaker trip        Loss of 125 V-dc Inability of pump to start During testing or observation Two pumps are provided per train; the breaker control on manual or automatic of control switch indicating standby pump will be started from the power to one signal, or trip the breaker lights main control board; a failure to trip a pump when required  breaker may cause loss of bus on one    train, redundant train pumps can be    started manually    Failure of loss of power Inability of both pumps in ESFAS malfunction alarm on Both redundant pumps in other train sequencer or ESS  a train to start main control board or start automatically on loss of  sequencer start automatically periodic testing offsite power; pumps can be started  signal to both  manually from control switches on pumps in one  the main control board train     
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 2 OF 15)  Component Identification  3019A,B,C,D and 3024A,B,C,D  Logic Diagram Number          NA            Elementary Number                  C-177613/D-207613          Engineering Flow Diagram Number:    D-175003 Sh. 1, D-205003 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-610 Two valves in the following Depending on initial valve Valves fail as is; normal position fails to operate grouping would not operate: position and plant operating is open; post-LOCA position is open;  on receipt of MOV 3019 A/B and 3024 A/B or status: operator can open valves  safety injection MOV 3019 C/D and 3024 C/D  signal  a. Computer  a. If valve initially b. Light monitor panel      open, no effect on  c. Periodic testing      system d. Position indication  b. If valve initially closed,      lights at main control      system reduced      board      to 2/4 (minimum        requirements)      Loss of power Two valves in the following Depending on initial valve Position indication lights of main to motor control grouping would not operate: position and plant operating control board will be out center U or V MOV 3019 A/B and 3024 A/B or status:  MOV 3019 C/D and 3024 C/D    a. Computer  a. If valve initially b. Light monitor panel      open, no effect on c. Periodic testing      system d. Loss of valve position  b. If valve initially    at main control board      closed, system        reduced to 2/4        (minimum        requirements)    c. Loss of ability to        close valve      Contacts of relay Valve fails as is; if valve Depending on initial valve Operator can open valves  K-610 fail to  initially closed, system position and plant operating    close on receipt reduced to 3/4 status:  of safety    injection signal  a. Computer  or failure of  b. Light monitor panel  open/close relay  c. Periodic testing  to operate  d. Position indication lights        at main control board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 3 OF 15)  Component Identification  3019A,B,C,D and 3024A,B,C,D (cont.)  Logic Diagram Number          NA          Elementary Number                C-177613/D-207613                    Engineering Flow Diagram Number:    D-175003 Sh. 1, D-205003 Sh. 1      Failure Mode Effect on System Detection of Failure Remarks Thermal overload Valve fails as is; if valve Depending on initial valve relay contacts initially closed, system position and plant operating open reduced to 3/4; loss of status:  ability to close valve    a. Computer  b. Light monitor panel  c. Periodic testing  d. Position indication        lights at main control        board Loss of 120 V-ac Valve fails as is; if valve Depending on initial valve Position indication lights or main  control power initially closed, system position and plant operating control board will be out  reduced to 3/4 status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication at main        control board         
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 4 OF 15)  Component Identification        3131        Logic Diagram Number            NA                Elementary Number    C-177612/D-207612 Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA  fails to valve (3134) operates to position and plant operating position closed; two containment  operate on  effect isolation; operator status: isolation valves in series, one  safety injection can close if initially open  required to operate  signal (normal) a. Computer    b. Light monitor panel  c. Periodic testing  d. Position indication        at main control        board Contacts of ESF Valve fails as is; alternate Depending on initial valve    relay K-604 fail valve (3134) operates to position and plant operating    to close on effect isolation; operator status:  safety injection can close if initially open  signal (normal) a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board Loss of power to Valve fails as is Depending on initial valve motor control  position and plant operating center U  status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board         
REV 15 FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 5 OF 15)  Component Identification      3131 (cont.)        Logic Diagram Number              NA                Elementary Number          C-177612/D-207612  Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position          indication light at          main control board Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:          a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication          at main control          board Failure of  Valve fails as is; operator Depending on initial valve    starter to can close/open valve position and plant operating    operate depending on which starter status:    coil fails      a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 6 OF 15)  Component Identification          3134              Logic Diagram Number            NA              Elementary Number        D-177636/D-207636      Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA  fails to valve (3131) operates to position and plant operating position closed; two containment  operate on effect isolation; operator status: isolation valves in series, one  safety injection can close valve if  required to operate signal initially open (normal) a. Computer  b. Light monitor panel  c. Periodic testing  d. Position indication        at main control        board  Contacts of ESF Valve fails as is; alternate Depending on initial valve  relay K-604 fail valve (3131) operates to position and plant operating    to close on effect isolation; operator status:  safety injection can close valve if  signal initially open (normal) a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board  Loss of power to  Valve fails as is Depending on initial valve    motor control  position and plant operating  center V  status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board         
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 7 OF 15)  Component Identification    3134 (cont.)        Logic Diagram Number            NA              Elementary Number      D-177636/D-207636 Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  Loss of 120 V-ac Valve fails as is Depending on initial valve    control power  position and plant operating    status:          a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position          indication light at          main control board                    Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:          a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication          at main control          board Failure of Valve fails as is; operator Depending on initial valve    starter relay can close/open valve position and plant operating    to operate depending on which starter status:    coil fails      a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication          at main control          board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 8 OF 15)  Component Identification        3135            Logic Diagram Number              NA            Elementary Number      D-177636/D-207636  Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA  fails to valve (QV075) is a check valve position and plant operating position closed; two containment  operate on which operates to effect status: isolation valves in series, one  safety injection isolation; operator can  required to operate signal close valve if initially open a. Computer  (normal) b. Light monitor panel    c  .Periodic testing    d. Position indication        at main control        board Contacts of ESF Valve fails as is; alternate Depending on initial valve  relay K-604 fail valve (QV075) is a check valve position and plant operating    to close on which operates to effect status:  safety injection isolation; operator can  signal close valve if initially open a. Computer  (normal) b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board Loss of power to  Valve fails as is Depending on initial valve    motor control  position and plant operating  center V  status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 9 OF 15)  Component Identification      3135 (cont.)        Logic Diagram Number              NA            Elementary Number      D-177636/D-207636  Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board                  Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board      Failure of Valve fails as is; operator Depending on initial valve    starter relay can close/open valve position and plant operating    to operate depending on which starter status:  coil fails    a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 10 OF 15)  Component Identification          3149            Logic Diagram Number              NA              Elementary Number    C-177617/D-207617              Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-604 Valve 3149 fails as is; Depending on initial valve Normal valve position open; post-LOCA  fails to operator can close position and plant operating position closed operate on valves if initially open status:  safety injection (normal)  signal  a. Computer  b. Light monitor panel  c. Periodic testing  d. Position indication      at main control      board Contacts of ESF Valve 3149 fails as is; Depending on initial valve    relay K-604 fail operator can close position and plant operating    to close on valves if initially open status:  safety injection (normal)  signal  a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board Loss of power to  Valves fail as are Depending on initial valve  motor control  position and plant operating  center U  status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 11 OF 15)  Component Identification      3149 (cont.)        Logic Diagram Number              NA              Elementary Number    C-177617/D-207617    Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board                  Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board      Failure of Valve fails as is; operator Depending on initial valve    starter relay can close/open valve position and plant operating    to operate depending on which starter status:  coil fails    a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 12 OF 15)  Component Identification          3150            Logic Diagram Number            NA              Elementary Number        C-177617/D-207617  Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-604 Valve 3150 fails as is; Depending on initial valve Normal valve position open; post-LOCA  fails to operator can close position and plant operating position closed  operate on valves if initially open status:  safety injection (normal)  signal  a. Computer  b. Light monitor panel  c. Periodic testing  d. Position indication        at main control        board    Contacts of ESF Valve 3150 fails as is; Depending on initial valve    relay K-604 fail operator can close position and plant operating    to close on valves if initially open status:  safety injection (normal)  signal  a. Computer  b. Light monitor panel  c. Periodic testing  d. Position indication        at main control        board Loss of power to  Valves fail as are Depending on initial valve  motor control  position and plant operating center U  status:
a. Computer  b. Light monitor panel  c. Periodic testing  d. Loss of valve position        indication light at        main control board       
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 13 OF 15)  Component Identification      3150 (cont.)        Logic Diagram Number            NA              Elementary Number      C-177617/D-207617  Engineering Flow Diagram Number:    D-175003 Sh. 2, D-205003 Sh. 2      Failure Mode Effect on System Detection of Failure Remarks  Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status:
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position        indication light at        main control board                  Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board      Failure of Valve fails as is; operator Depending on initial valve  starter relay can close/open valve position and plant operating    to operate depending on which starter status:  coil fails    a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication        at main control        board FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 14 OF 15)  Component Identification      3441A,B,C,D        Logic Diagram Number            NA              Elementary Number      D-177633/D-207633    Engineering Flow Diagram Number:    D-175003 Sh. 1, D-205003 Sh. 1      Failure Mode Effect on System Detection of Failure Remarks  ESF relay K-610 Two valves in the following Depending on initial valve Valves fail as is; normal valve  fails to groupings would not operate: position and plant operating position is open; post-LOCA position  operate on MOV 3441 A/B or MOV 3441 C/D status: is open; operator can open valves  receipt of    safety injection a. If valves initially a. Computer signal      open, no effect on b. Light monitor panel      system c. Position indication  b. If valves initially      light at main control      closed, system      board      reduced to 2/4 d. Periodic testing      (minimum        requirements)
Loss of power to Two valves in the following Depending on initial valve    motor control groupings would not operate: position and plant operating    center U or V MOV 3441 A/B or MOV 3441 C/D status:
a. If valves initially a. Computer      open, no effect on b. Light monitor panel      system c. Loss of valve position  b. If valves initially      indication at main        closed, system      control board        reduced to 2/4 d. Periodic testing      (minimum        requirements)  c. Loss of ability to        close valve Contacts of Valve fails as is; if valve Depending on initial valve Operator can open valves  relay K-610 is initially closed, system position and plant operating    fail to close is reduced to 3/4 status:  on receipt of    safety injection  a. Computer signal or  b. Light monitor panel failure of starter  c. Position indication relay to operate      at main control        board  d. Periodic testing FNP-FSAR-7  REV 21  5/08 TABLE 7.3-6 (SHEET 15 OF 15)  Component Identification      3441A,B,C,D (cont.)      Logic Diagram Number              NA              Elementary Number              D-177633/D-207633                    Engineering Flow Diagram Number:    D-175003 Sh. 1, D-205003 Sh. 1      Failure Mode Effect on System Detection of Failure Remarks  Thermal overload Loss of ability to close Depending on initial valve    relay contacts valve; valve fails as is; position and plant operating    open if valve is initially status:    closed, system is    reduced to 3/4 a. Computer    b. Light monitor panel    c. Position indication        at main control        board    d. Periodic testing      Loss of 120 V-ac Loss of ability to close Depending on initial valve    control power valve; valve fails as is; position and plant operating    if valve is initially status:  closed, system is    reduced to 3/4 a. Computer    b. Light monitor panel    c. Loss of valve position        indication at main        control board    d. Periodic testing               
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 1 OF 17)  FAILURE MODE AND EFFECTS ANALYSIS, COMPONENT COOLING WATER SYSTEM Component Identification  Component Cooling Water Pumps Logic Diagram Number            NA            Elementary Number  D-177183, D-177184, D-177185, D-177187, Engineering Flow Diagram Number:  D-207183, D-207184, D-207185, D-207187 D-175002, Sh. 1, Sh. 2, and Sh. 3,    D-205002, Sh. 1, Sh. 2, and Sh. 3  Failure Mode Effect on System Detection of Failure Remarks    Loss of 4.16-kV Failure of pump to start when 4.16-kV bus U/V or bus  Redundant train pump can be started. power to one required or automatic breaker auto trip alarm  pump stoppage when pump is running on emergency power board                              Loss of 125 V-dc Inability of pump to start on During monthly testing or Three pumps are provided; one pump  breaker control manual or automatic signal, or observation of control is required for normal, hot shutdown,  power trip the breaker when required switch red, green, and or post-LOCA heat removal. A failure        amber lights to trip a breaker may cause loss of bus    on one train, redundant train pump can    be started manually. Failure of loss of Inability of pump to start ESS malfunction alarm Pump can be manually started from  power sequencer upon receipt of automatic on main control board or the main control board  start or ESS start signal periodic testing  sequencer start    signals        Automatic breaker Standby (swing) pump Breaker auto trip alarm Swing pump can be put into service  trip due to automatically starts if on main control board manually; one pump is required for  overcurrent aligned with pump that  normal, hot shutdown, or post-LOCA  tripped  heat removal 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 2 OF 17)  Component Identification        3067          Logic Diagram Number            NA            Elementary Number  F-177851/D-207851    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating closed on loss of instrument air  on receipt of series (3067 and 3443); status:  CIAS phase A only one required to    operate to cause a. Valve position indication  isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fails there are two valves in position and plant operating    to open on series (3067 and 3443); status:  receipt of only one required to  CIAS phase A operate to cause a. Valve position indication  isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at  control power  main control board out    b. Computer    c. Light monitor panel    d. High temperature at    discharge of excess    letdown heat exchanger    (TE-139) 
Solenoid valve Valve remains open; however, Depending on initial valve    3067 fails to there are two valves in position and plant operating    vent (sticky series (3067 and 3443); status:  operator) only one required to    operate to cause a. Valve position indication  isolation at main control board    b. Light monitor panel    c. Computer    d. Periodic testing FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 3 OF 17)
Component Identification        3067 (Cont.)        Logic Diagram Number          NA              Elementary Number  F-177851/D-207851    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System    Detection of Failure Remarks    Loss of  Valve fails closed a. Valve position light at main  instrument air  control board    b. Computer    c. Light monitor panel    d. High temperature at discharge of    excess letdown heat    exchanger (TE-139)    ` 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 4 OF 17)
Component Identification    3095      Logic Diagram Number          NA Elementary Number  F-177851/D-207851    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating closed on loss of instrument air  on receipt of series (3095 and a check status:  CIAS phase A valve); only one required    to operate to cause a. Valve position indication  isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve  relay K-613 fails there are two valves in position and plant operating  to open on series (3095 and a check status:  receipt of  valve); only one required  CIAS phase A to operate to cause a. Valve position indication    isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at  control power  main control board out    b. Computer    c. Light monitor panel    d. High temperature at    discharge of excess    letdown heat exchanger (TE-139)          Solenoid valve Valve remains open; however, Depending on initial valve    3095 fails to there are two valves in position and plant operating    vent (sticky series (3095 and a check status:  operator) valve); only one required to    operate to cause isolation a. Valve position indication      at main control board    b. Light monitor panel    c. Computer    d. Periodic testing FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 5 OF 17)
Component Identification      3095 (cont.)        Logic Diagram Number            NA          Elementary Number  F-177851/D-207851    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    Loss of  Valve fails closed a. Valve position light at  instrument air  main control board    b. Computer    c. Light monitor panel    d. High temperature at    discharge of excess    letdown heat exchanger    (TE-139) 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 6 OF 17)
Component Identification    3096A,B    Logic Diagram Number            NA              Elementary Number  D-177853/D-207853    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-609 Valves remains open; operator Depending on initial valve Valves are normally open and will fail fails to operate can close valves from main position and plant operating closed on loss of instrument air  on receipt of control board status:  safety injection    signal  a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing Contacts of ESF Valves remain open; operator Depending on initial valve    relay K-609 fail can close valves from position and plant operating    to open on main control board status:  receipt of safety    injection signal  a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at  control power  main control board out    b. Computer    c. Light monitor panel Solenoid valve Associated valve remains open; Depending on initial valve    3096A or 3096B other valve will be  position and plant operating    fails to operational status:  vent (sticky    operator)  a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing Loss of  Valve fails closed a. Valve position light at  instrument air  main control board    b. Computer    c. Light monitor panel FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 7 OF 17) `
Component Identification        3045        Logic Diagram Number            NA              Elementary Number  D-177854/D-207854    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-625 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating closed on loss of instrument air;  on receipt of series (3045 and 3184); status: valve operation is not testable with  CIAS phase B only one required to  reactor coolant pumps operating  operate to cause a. Valve position indication  isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-625 fails there are two valves in position and plant operating    to open on series (3045 and 3184); status:  receipt of only one required to  CIAS phase B operate to cause a. Valve position indication  isolation; operator can at main control board  close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails open Valve position lights at main    control power control board      Solenoid valve Valve remains open; however, Depending on initial valve    3045 fails to there are two valves in position and plant operating    vent (sticky series (3045 and 3184); status:  operator) only one required to operate    to cause isolation a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve closes a. Valve position lights at  instrument air  main control board    b. Computer    c. Light monitor panel FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 8 OF 17)
Component Identification    3046        Logic Diagram Number            NA                Elementary Number  C- 177618/D-207618    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA  fails to operate alternate valve 3182 position and plant operating position closed; two containment  on CIAS phase B operates to effect status: isolation valves in series; one  isolation; operator can  required for isolation; valve not  close valve if initially a. Computer testable at power  open (normal) b. Light monitor panel    c. Position indication    lights at main control    board    d. Periodic testing      Contacts of ESF Valve fails as is; Depending on initial valve  relay K-619 alternate valve 3182 position and plant operating  fail to close operates to effect status:  on CIAS phase B isolation; operator can    close valve if initially a. Computer  open (normal) b. Light monitor panel    c. Position indication    at main control    board d. Periodic testing      Loss of power Valve fails as is Depending on initial valve  to motor control  position and plant operating  center U  status:        a. Computer    b. Light monitor panel    c. Loss of valve position    indication light at    main control board    d. Periodic testing                 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 9 OF 17)
Component Identification  3046 (cont.)    Logic Diagram Number            NA                Elementary Number  C-177618/D-207618    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status: 
a. Computer    b. Light monitor panel    c. Loss of valve position    light at main control    board d. Periodic testing             
Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status: 
a. Computer    b. Light monitor panel    c. Position indication    lights at main control    board d. Periodic testing Failure of Valve fails as is; Depending on initial valve  starter relay operator can close or open position and plant operating    to operate valve depending on which status:  starter coil fails    a. Computer    b. Light monitor panel    c. Position indication    lights at main control    board d. Periodic testing 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 10 OF 17)
Component Identification        3052        Logic Diagram Number              NA              Elementary Number  C-177625/D-207625    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA  fails to operate alternate valve is check position and plant operating position closed; two containment  on CIAS phase B valve which operates to status: isolation valves in series; one required to  effect isolation; operator  operate;  valve not testable at power  can close valve if initially a. Computer    open (normal) b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board      Contacts of ESF Valve fails as is; Depending on initial valve  relay K-619 alternate valve is check position and plant operating    fail to close valve which operates to status:  on CIAS phase B effect isolation; operator    can close valve if initially a. Computer  open (normal) b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board      Loss of power Valve fails as is Depending on initial valve  to motor control  position and plant operating  center V  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position    indication light at    main control board           
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 11 OF 17)  Component Identification    3052 (cont.)    Logic Diagram Number              NA              Elementary Number  C-177625/D-207625    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status: 
a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position    indication light at    main control board     
Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status: 
a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board Failure of Valve fails as is; Depending on initial valve  starter relay operator can close/open position and plant operating  to operate valves depending on status:  which starter coil fails    a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board     
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 12 OF 17)
Component Identification      3182          Logic Diagram Number                NA                Elementary Number    D-177610/D-207610      Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operate alternate valve 3046 position and plant operating position closed; two containment on CIAS phase B operates to effect status: isolation valves in series; one required  isolation; operator can  to operate; valve not testable at power  close valve if a. Computer  initially open (normal) b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board      Contacts of ESF Valve fails as is; Depending on initial valve  relay K-619 alternate valve 3046 position and plant operating    fail to close operates to effect status:  on CIAS phase B isolation; operator can    close valve if initially a. Computer  open (normal) b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board      Loss of power Valve fails as is Depending on initial valve  to motor control  position and plant operating  center V  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position    indication light at    main control board           
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 13 OF 17)  Component Identification    3182 (cont.)        Logic Diagram Number                NA              Elementary Number      D-177610/D-207610    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks    Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating    status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Loss of valve position    indication light at    main control board      Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating  open  status:        a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board      Failure of Valve fails as is; Depending on initial valve  starter relay operator can close/open position and plant operating  to operate valves depending on status:  which starter coil fails    a. Computer    b. Light monitor panel    c. Periodic testing    d. Position indication    at main control board 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 14 OF 17)
Component Identification        3184          Logic Diagram Number              NA              Elementary Number  D-177855/D-207855    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-625 Valve remains open; however, Depending on initial valve Valve is normally open and will fails to operate there are two valves in position and plant operating fail closed on loss of instrument  on receipt of series (3184 and 3045); status: air; valve operation is not testable  CIAS phase B only one required to operate  with reactor coolant pumps operating  to cause isolation; operator a. Valve position indication    can close valve from main at main control board  control board b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valves remains open; however, Depending on initial valve    relay K-625 fail there are two valves in position and plant operating    to open on series (3184 and 3045); status:  receipt of only one required to operate  CIAS phase B to cause isolation; operator a. Valve position indication    can close valve from main at main control board  control board b. Light monitor panel    c. Computer    d. Periodic testing      Loss of 125 V-dc Valve open Valve position lights at main  control power  control board      Solenoid valve Valve remains open; however, Depending on initial valve    3184 fails to there are two valves in position and plant operating    vent (sticky series (3184 and 3045); status:  operator) only one required to operate    to cause isolation a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of Valve closes a. Valve position lights at  instrument air  main control board    b. Computer    c. Light monitor panel FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 15 OF 17)
Component Identification        3443          Logic Diagram Number              NA                Elementary Number  D-177374/D-207374    Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valves are normally open and will fail closed fails to operate there are two valves in position and plant operating on loss of instrument air on receipt of series (3443 and 3067);  status:  CIAS phase A only one required to operate    to cause isolation; operator a. Valve position indication    can close valve from main at main control board  control board b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fails there are two valves in position and plant operating    to open on series (3443 and 3067); status:  receipt of only one required to operate  CIAS phase A to cause isolation; operator a. Valve position indication    can close valve from main at main control board  control board b. Light monitor panel    c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at  control power  main control board out    b. Computer    c. Light monitor panel    d. High temperature at    discharge of excess    letdown heat exchanger    (TE-139)      Solenoid valve Valve remains open; however, Depending on initial valve    3443 fails to there are two valves in position and plant operating    vent (sticky series (3443 and 3067); status:  operator) only one required to operate    to cause isolation a. Valve position indication    at main control board    b. Light monitor panel    c. Computer    d. Periodic testing FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 16 OF 17)
Component Identification      3443 (cont.)        Logic Diagram Number            NA                Elementary Number    D-177374/D-207374      Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks    Loss of  Valve fails closed a. Valve position light at  instrument air  main control board    b. Computer    c. Light monitor panel    d. High temperature at    discharge of excess    letdown heat exchanger    (TE-139) 
FNP-FSAR-7  REV 21  5/08 TABLE 7.3-7 (SHEET 17 OF 17)
Component Identification    2229          Logic Diagram Number                NA Elementary Number    D-177853/D-207853      Engineering Flow Diagram Number  D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks      ESF relay K-609 fails to Valve remains open; operator Depending on initial valve position and Valve is normally open and will fail closed on operate on receipt of can close valve from local plant operating status: loss of instrument air safety injection signal control station a. Valve position indication at local    control station    b. Light monitor panel    c. Periodic testing      Contacts of ESF relay Valve remains open; operator Depending on initial valve position and  K-609 fail to open on receipt of safety injection can close valve from local control station plant operating status:  signal  a. Valve position indication at local    control station    b. Light monitor panel    c. Periodic testing      Loss of 125 V-dc control Valve fails closed a. Valve position light at local control  power  station out    b. Light monitor panel        Solenoid valve 2229A fails Associated valve remains Depending on initial valve position and  to vent (sticky operator) open; other valve will be plant operating status:  operational a. Valve position indication at local    control station    b. Light monitor panel    c. Periodic testing      Loss of instrument air Valve fails closed a. Valve position light at local control      station    b. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 1 OF 18)  FAILURE MODE AND EFFECTS ANALYSIS, CONTROL ROOM AND AIR CONDITIONING AND FILTRATION SYSTEM  Component Identification    Filtration Fan      Logic Diagram Number            NA              Elementary Number        D-177270 Sh. 1        Engineering Flow Diagram Numbers:      D-175012 Sh. 1 & D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-614 Motor fails to start; of two filtration a. Control room indication lights  fails to operate fans, only one is required for b. Periodic testing  on CIAS phase A, operation; operator can start motor c. Monitor light box abnormal  Unit 1 or 2 if necessary      Contacts of ESF Motor fails to start; of two filtration a. Control room indication lights  relay K-614 fail to fans, only one is required for b. Periodic testing  close on CIAS operation; operator can start motor c. Monitor light box abnormal  phase A if necessary      Loss of power to 208-V  Motor fails to start a. Control room indication lights  motor control center  b. Periodic testing      1F and 1G        Loss of 120 V-ac Motor fails to start a. Control room indication lights  control power  b. Periodic testing      Thermal overload Motor fails to start a. Control room indication lights  relay contacts open  b. Periodic testing    c. Motor overload trip alarm        in the control room FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 2 OF 18)  Component Identification    Air Conditioning Unit      Logic Diagram Number              NA                Elementary Number            D-177270 Sh. 3              Engineering Flow Diagram Numbers:      D-175012 Sh. 1 & D-205012 Sh. 1 Failure Mode      Effect on System      Detection of Failure Remarks    ESF relay K-614 Motor fails to start; of two air a. Periodic testing  Redundant train will start fails to operate conditioning units, only one is  on CIAS phase A, required for operation;  Unit 1 or 2 operator can start motor    if necessary      Contacts of ESF Motor fails to start; of two air a. Periodic testing  relay K-614 fail to conditioning units, only one is  close on CIAS required for operation;  phase A operator can start motor    if necessary      Loss of power Motor fails to start a. Periodic testing  to 600-V motor    control center    1F & 1G        Loss of 120 V-ac Motor fails to start a. Periodic testing  control power        Thermal overload Motor fails to start a. Periodic testing  relay contacts  b. Motor overload trip alarm  open      in the control room         
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 3 OF 18)  Component Identification      3478A        Logic Diagram Number              NA                Elementary Number    D-177280 Sh. 1      Engineering Flow Diagram Numbers:      D-175012 Sh. 1                                   
Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-614 or its Valve fails as is Depending on initial valve Normal valve position closed;  contact fails to operate  position and plant operating status: post-LOCA position open on control room on CIAS phase A  pressurization fan start signal  a. Computer Redundant valve will open  b. Position indication lights at BOP panel    c. Periodic testing        Contacts of 42X Valve fails as is Depending on initial valve Redundant valve will open relay (Control Room  position and plant operating status:  Pressurization Fan) fails to    operate on CIAS phase A  a. Computer    b. Position indication lights at BOP Panel    c. Periodic testing      Loss of power to Valve fails as is Depending on initial valve  600-V motor  position and plant operating status:  control center    1F  a. Computer    b. Loss of valve position      indication light at BOP Panel    c. Periodic testing FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 4 OF 18)  Component Identification    3478A (cont.)      Logic Diagram Number            NA                Elementary Number      D-177280 Sh. 1          Engineering Flow Diagram Numbers:      D-175012 Sh. 1                                   
Failure Mode Effect on System Detection of Failure Remarks    Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating status:        a. Computer    b. Periodic testing    c. Loss of valve position indication lights at        BOP panel      Thermal overload Valve fails as is Depending on initial valve  relay contacts  position and plant operating status:  open      a. Computer    b. Loss of position indication lights at BOP      panel    c. Periodic testing      Failure of (42) Valve fails as is Depending on initial valve  starter relay  position and plant operating status:  to operate      a. Computer    b. Position indication lights at main control          board    c. Periodic testing FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 5 OF 18)  Component Identification      3478B        Logic Diagram Number            NA              Elementary Number  D-177280 Sh. 1  Engineering Flow Diagram Numbers:      D-205012 Sh. 1                       
Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-614 or its Valve fails as is Depending on initial valve Normal valve position closed;  contact fails to operate  position and plant operating status: post-LOCA position open on control room on CIAS phase A  pressurization fan start signal;  a. Computer Redundant valve will open    b. Periodic testing      c. Position indication at BOP panel      Contact of 42X relay Valve fails as is Depending on initial valve  Redundant valve will open (Control Room Pressuri-  position and plant operating status:  zation Fan) fails to    operate on CIAS phase A  a. Computer    b. Periodic testing    c. Position indication at BOP panel      Loss of power Valve fails as is Depending on initial valve  to 600-V motor  position and plant operating status:  control center    1G  a. Computer    b. Periodic testing    c. Loss of valve position indication light at      BOP panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 6 OF 18)  Component Identification      3478B (cont.)        Logic Diagram Number              NA              Elementary Number              D-177280 Sh. 1          Engineering Flow Diagram Numbers:      D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    Loss of 120 V-ac Valve fails as is Depending on initial valve  control power  position and plant operating status:        a. Computer    b. Periodic testing    c. Loss of valve position indication lights at BOP panel      Thermal overload Valve fails as is Depending on initial valve  relay contacts open  position and plant operating status:        a. Computer    b. Periodic testing    c. Loss of position indication at BOP panel      Failure of (42) Valve fails as is Depending on initial valve  starter relay  position and plant operating status:  to operate      a. Computer    b. Periodic testing    c. Position indication at main control board FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 7 OF 18)  Component Identification        3622          Logic Diagram Number              NA              Elementary Number        D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air  on receipt of series (3622 and 3623);  CIAS phase A only one required to operate a. Valve position indication at BOP panel  to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve  relay K-613 fails there are two valves in position and plant operating status:  to open on series (3622 and 3623);  receipt of only one required to operate a. Valve position indication at BOP panel  CIAS phase A to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at main control board  control power  b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve    3622 fails to there are two valves in position and plant operating status:  vent (sticky series (3622 and 3623);  operator) only one required to operate a. Valve position indication at BOP panel  to cause isolation b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 8 OF 18)  Component Identification      3623        Logic Diagram Number              NA              Elementary Number        D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating status: closed on loss of instrument air  on receipt of series (3623 and 3622);  CIAS phase A only one required to operate a. Valve position indication at BOP panel  to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fails there are two valves in position and plant operating status:  to open on series (3623 and 3622);  receipt of only one required to operate a. Valve position indication at BOP panel  CIAS phase A to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve  3623 fails to there are two valves in position and plant operating status:  vent (sticky series (3623 and 3622);  operator) only one required to operate a. Valve position indication at BOP panel  to cause isolation b. Light monitor panel    c. Computer    d. Periodic testing Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 9 OF 18)  Component Identification      3624        Logic Diagram Number              NA                Elementary Number      D-177373          Engineering Flow Diagram Numbers:      D-205012 Sh. 1                                    Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-606 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating status: closed on loss of instrument air  on receipt of series (3624 and 3625);  CIAS phase A only one required to operate    to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-606 fail there are two valves in position and plant operating status:  to open on series (3624 and 3625);  receipt of only one required to operate a. Valve position indication at BOP panel  CIAS phase A to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve  3624 fails to there are two valves in position and plant operating status:  vent (sticky series (3624 and 3625);  operator) only one required to operate a. Valve position indication at BOP panel  to cause isolation b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 10 OF 18)  Component Identification        3625          Logic Diagram Number              NA              Elementary Number        D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1                                  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air  on receipt of series (3625 and 3624);  CIAS phase A only one required to operate a. Valve position indication at BOP panel  to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fail there are two valves in position and plant operating status:  to open on series (3625 and 3624);  receipt of only one required to operate a. Valve position indication at BOP panel  CIAS phase A to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve    3625 fails to there are two valves in position and plant operating status:  vent (sticky series (3625 and 3624);  operator) only one required to operate a. Valve position indication at BOP panel  to cause isolation b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 11 OF 18)  Component Identification      3626        Logic Diagram Number                NA              Elementary Number      D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating closed on loss of instrument air  on receipt of series (3626 and 3627); status:  CIAS phase A only one required to operate    to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fails there are two valves in position and plant operating status:  to open on series (3626 and 3627);  receipt of only one required to operate a. Valve position indication at BOP panel  CIAS phase A to cause isolation; operator b. Light monitor panel  can close valve from main c. Computer  control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve    3626 fails to there are two valves in position and plant operating    vent (sticky series (3626 and 3627); status:  operator) only one required to operate    to cause isolation a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 12 OF 18)  Component Identification        3627        Logic Diagram Number              NA                Elementary Number        D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail  fails to operate there are two valves in position and plant operating status: closed on loss of instrument air  on receipt of series (3627 and 3626);  CIAS phase A only one required to operate    to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main control board b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-613 fails there are two valves in position and plant operating status:  to open on series (3627 and 3626);  receipt of only one required to operate  CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve    3627 fails to there are two valves in position and plant operating status:  vent (sticky series (3627 and 3626);  operator) only one required to operate a. Valve position indication at BOP panel  to cause isolation b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at  instrument air      BOP panel    b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 13 OF 18)  Component Identification        3628          Logic Diagram Number              NA              Elementary Number        D-177373              Engineering Flow Diagram Numbers:      D-205012 Sh. 1                                  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-605 fails to operate Valve remains open; however, Depending on initial valve Valve is normally open and will fail  on receipt of CIAS phase A there are two valves in position and plant operating status: closed on loss of instrument air  series (3628 and 3629);    only one required to operate    to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-605 fails there are two valves in position and plant operating status:  to open on series (3628 and 3629);  receipt of only one required to operate  CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel      Solenoid valve3628 fails to Valve remains open; however, Depending on initial valve    vent (sticky operator) there are two valves in position and plant operating status:  series (3628 and 3629);    only one required to operate    to cause isolation a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at  instrument air      BOP panel    b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 14 OF 18)  Component Identification      3629        Logic Diagram Number              NA              Elementary Number        D-177373            Engineering Flow Diagram Numbers:      D-205012 Sh. 1                                    Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-605 fails to operate Valve remains open; however, Depending on initial valve Valve is normally open and will fail  on receipt of CIAS phase A there are two valves in position and plant operating status: closed on loss of instrument air  series (3629 and 3628);    only one required to operate    to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; however, Depending on initial valve    relay K-605 fails there are two valves in position and plant operating status:  to open on series (3629 and 3628);  receipt of only one required to operate  CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel  can close valve from main b. Light monitor panel  control board c. Computer    d. Periodic testing      Loss of 125 V-dc control power Valve fails closed a. Valve position light at BOP panel    b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve    3628 fails to there are two valves in position and plant operating    vent (sticky series (3629 and 3628); status:  operator) only one required to operate    to cause isolation a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 15 OF 18)  Component Identification      3649A          Logic Diagram Number              NA                Elementary Number          D-177883            Engineering Flow Diagram Numbers:      D-175012 Sh. 1                                    Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 fails to operate Valve remains open; operator Depending on initial valve Valve is open only for smoke purge;  on receipt of CIAS phase A can close valve from main control board position and plant operating status: will fail closed on loss of instrument air      a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; operator Depending on initial valve    relay K-613 fail to open on can close valve from main control board position and plant operating status:  receipt of CIAS phase A      a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of 125 V-dc control power Valve fails closed a. Valve position light at BOP panel    b. Computer    c. Light monitor panel      Solenoid valve3649A fails to Valve remains open; however, valve is Depending on initial valve    vent (sticky operator) normally closed except for smoke purge position and plant operating status:        a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of instrument air Valve fails closed a. Valve position light at BOP panel    b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 16 OF 18)  Component Identification        3649B          Logic Diagram Number              NA                Elementary Number          D-177883            Engineering Flow Diagram Numbers:      D-175012 Sh. 1                                    Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 fails to operate Valve remains open; operator Depending on initial valve Valve is open only for smoke purge and on receipt of CIAS phase A can close valve from main control board position and plant operating status: will fail closed on loss of instrument air      a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; operator Depending on initial valve    relay K-613 fail can close valve from main position and plant operating    to open on control board status:  receipt of    CIAS phase A  a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel  control power  b. Computer    c. Light monitor panel      Solenoid valve Valve remains open; however, Depending on initial valve    3649B fails to valve is normally closed position and plant operating    vent (sticky except for smoke purging status:  operator)      a. Valve position indication at BOP panel    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at BOP panel  instrument air  b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 17 OF 18)  Component Identification        3649C          Logic Diagram Number              NA                Elementary Number          D-177883            Engineering Flow Diagram Numbers:      D-175012 Sh. 1                                  Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-613 Valve remains open; operator Depending on initial valve Valve is open only for smoke purge and  fails to operate can close valve from main control board position and plant operating status: will fail closed on loss of instrument air on receipt of CIAS phase A      a. Valve position indication at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Valve remains open; operator Depending on initial valve    relay K-613 fail can close valve from main position and plant operating    to open on control board status:  receipt of    CIAS phase A  a. Valve position indication at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of 125 V-dc Valve fails closed a. Valve position light at main control board  control power  b. Computer    c. Light monitor panel      Solenoid valve 3696 fails to Valve remains open; however, the valve is Depending on initial valve    vent (sticky operator) normally closed except for smoke purging position and plant operating status:        a. Valve position indication at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of  Valve fails closed a. Valve position light at  instrument air      main control board    b. Computer    c. Light monitor panel FNP-FSAR-7 REV 21  5/08 TABLE 7.3-8 (SHEET 18 OF 18)  Component Identification  Control Room Pressurization Fan  Logic Diagram Number              NA                Elementary Number        D-177280 Sh. 3                        Engineering Flow Diagram Numbers:      D-175012 Sh. 1 and D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-614 Motor fails to start; of two a. Control room indication lights    fails to operate pressurization fans, only one b. Periodic testing  on CIAS phase A, is required for operation; c. Light monitor panel  Units 1 or 2 operator can start motor if necessary          Contacts of ESF Motor fails to start; of two a. Control room indication lights    relay K-614 fail to pressurization fans, only one b. Periodic testing  close on CIAS is required for operation; c. Light monitor panel  phase A operator can start motor if necessary          Loss of power to Motor fails to start a. Loss of both control room  600-V motor      indication lights  control center  b. Periodic testing          Loss of 120 V-ac Motor fails to start a. Loss of both control room  control power      indication lights    b. Periodic testing          Thermal overload Motor fails to start a. Control room indication  relay contacts open      lights    b. Periodic testing    c. Motor overload trip alarm        in the control room FNP-FSAR-7 REV 21  5/08 TABLE 7.3-9 (SHEET 1 OF 4)  FAILURE MODE AND EFFECTS ANALYSIS, PENETRATION ROOM FILTRATION SYSTEM  Component Identification      Exhaust Fan          Logic Diagram Number              NA                Elementary Number          D-177238/D-207238              Engineering Flow Diagram Numbers:      D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-626 Motor fails to start a. Light monitor panel    fails to operate  b. Indication lights at    on CIAS phase B      main control board    c. Periodic testing      Contacts of ESF Motor fails to start a. Light monitor panel    relay K-626 fail to  b. Indication lights at    close on CIAS      main control board  phase B  c. Periodic testing      Loss of power to Motor fails to start a. Light monitor panel  600-V motor  b. Indication lights at  control center      main control board  A  c. Periodic testing      Loss of 120 V-ac Motor fails to start a. Light monitor panel  control power  b. Indication lights at        main control board    c. Periodic testing      Thermal overload Motor fails to start a. Light monitor panel  relay contacts  b. Indication lights at  open      main control board    c. Periodic testing    d. Motor overload trip        alarm at main control        board 
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-9 (SHEET 2 OF 4)  Component Identification      Recirculation Fan          Logic Diagram Number                NA                Elementary Number      D-177239/D-207239            Engineering Flow Diagram Numbers:      D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-626 Motor fails to start a. Light monitor panel    fails to operate  b. Indication lights at  on CIAS phase B      main control board    c. Periodic testing      Contacts of ESF Motor fails to start a. Light monitor panel    relay K-626 fail to  b. Indication lights at    close on CIAS      main control board  phase B  c. Periodic testing        Loss of power to Motor fails to start a. Light monitor panel  600-V motor  b. Indication lights at  control center      main control board  A  c. Periodic testing      Loss of 120 V-ac Motor fails to start a. Light monitor panel  control power  b. Indication lights at        main control board    c. Periodic testing      Thermal overload Motor fails to start a. Light monitor panel  relay contacts  b. Indication lights at  open      main control board    c. Periodic testing    d. Motor overload trip        alarm at main control        board 
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-9 (SHEET 3 OF 4)  Component Identification        3362A,B          Logic Diagram Number                NA                  Elementary Number      D-177281/D-207281    Engineering Flow Diagram Numbers:      D-175022 Sh. 1 and D-205022 Sh. 1 
Failure Mode Effect on System Detection of Failure Remarks    ESF relay K-626 train A Associated valve remains closed Depending on initial valve Valve is normally closed  or B fails to operate on receipt  position and plant operating    of CIAS phase B  status:        a. Valve position indication at main control          board    b. Light monitor panel    c. Computer    d. Periodic testing      Contacts of ESF Associated valve remains closed Depending on initial valve    relay K-626 train A or B  position and plant operating status  fail to close on receipt of  :  CIAS phase B  a. Valve position indication          at main control board    b. Light monitor panel    c. Computer    d. Periodic testing      Loss of power Associated valve fails a. Loss of valve position light at    to 600-V motor control centers as is; the other valve      main control board  U-3362A and V-3362B will open b. Computer    c. Light monitor panel      Loss of 120 V-ac Associated valve fails Loss of valve position light    control power as is at main control board      Thermal overload Associated valve fails Depending on initial valve    relay contacts open as is position and plant operating status:        a. Valve position indication        at main control board    b. Light monitor panel    c. Computer    d. Periodic testing FNP-FSAR-7 REV 21  5/08 TABLE 7.3-9 (SHEET 4 OF 4)  Component Identification      3362A,B (cont.)          Logic Diagram Number                NA                Elementary Number      D-177281/D-207281          Engineering Flow Diagram Numbers:      D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks    Failure of Valve fails as is Depending on initial valve  starter relay  position and plant operating  to operate  status:        a. Valve position indication        at main control board    b. Light monitor panel    c. Computer    d. Periodic testing 
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-11 (SHEET 1 OF 5)  FAILURE MODE AND EFFECTS ANALYSIS, EMERGENCY SAFEGUARDS PUMP ROOM COOLING SYSTEM  Component Identification Containment Spray Pump Room Coolers  Logic Diagram Number          NA          Elementary Number        D-177227 Sh. 2, D-207227 Sh. 1                Engineering Flow Diagram Numbers:        D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode    Effect on System    Detection of Failure  Remarks Loss of 600-V Motor stops a. Fan fault alarm Two spray pumps provided; each unit  motor control      at main control board is provided with its own cooler and center A or  b. Loss of red/green  fan; emergency core cooling system  B        indication lights at analysis is based upon operation        BOP panel of one pump  c. Monitor light box        abnormal      Loss of 120 V-ac Motor stops a. Fan fault alarm  control power      at main control board    b. Loss of red/green          indication lights  at        BOP panel    c. Monitor light box        abnormal      Thermal overload Motor stops a. Fan fault alarm  contacts open      at main control board    b. Loss of red/green          indication lights  at        BOP panel    c. Monitor light box        abnormal     
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-11 (SHEET 2 OF 5)  Component Identification Residual Heat Removal Pump Room Coolers  Logic Diagram Number          NA          Elementary Number        D-177227 Sh. 1, D-207227 Sh. 1                      Engineering Flow Diagram Numbers:      D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode    Effect on System  Detection of Failure            Remarks Loss of 600-V Motor stops a. Fan fault alarm at Two RHR pumps provided; each unit motor control      main control board is provided with its own cooler and center A or  b. Loss of red/green fan; emergency core cooling system  B      indication lights at analysis is based upon operation          BOP panel of one pump  c. Monitor light box        abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm at  control power      main control board    b. Loss of red/green        indication lights at        BOP panel    c. Monitor light box        abnormal Thermal overload Motor stops a. Fan fault alarm at  contacts open      main control board    b. Loss of red/green        indication lights at        BOP panel    c. Monitor light box        abnormal     
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-11 (SHEET 3 OF 5)  Component Identification Charging/HHSI Pump Room Cooler Fan Motors  Logic Diagram Number          NA          Elementary Numbers  D-177226, D-177229 Sh. 2, D-177284,        Engineering Flow Diagram Numbers:    D-207226, D-207229, D-207284  D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode    Effect on System  Detection of Failure            Remarks Loss of 600-V Motor stops a. Fan fault alarm at Three charging/HHSI pumps provided;  motor control      main control board each unit is provided with its own center A or B  b. Loss of red/green cooler and fan; emergency core  train dedicated pump rooms      indication lights at cooling system analysis is based  have corresponding train      BOP panel upon operation of one pump  dedicated room coolers;  c. Monitor light box  swing pump room cooler can      abnormal  be aligned to either train A or    Train B, depending on    pumps alignment Loss of 120 V-ac Motor stops a. Fan fault alarm at  control power      main control board    b. Loss of red/green        indication lights at        BOP panel    c. Monitor light box        abnormal Thermal overload Motor stops a. Fan fault alarm at  contacts open      main control board    b. Monitor light box        abnormal     
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-11 (SHEET 4 OF 5)  Component Identification Component Cooling Water Pump Room Coolers  Logic Diagram Number          NA          Elementary Number      D-177243 Sh. 1, D-207243 Sh. 1                              Engineering Flow Diagram Numbers:      D-175011 Sh. 3, D-205011 Sh. 3       
Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm Three component cooling water pumps  motor control      at main control board provided; there are two coolers for center  b. Loss of red/green the three pumps; emergency core        indication lights at cooling system analysis is based upon        BOP panel operation of one pump  c. Monitor light box        abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm  control power      at main control board    b. Loss of red/green        indication lights at        BOP panel    c. Monitor light box        abnormal Thermal overload Motor stops a. Fan fault alarm  contacts open      at main control board    b. Monitor light box        abnormal     
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-11 (SHEET 5 OF 5)  Component Identification Motor-Driven Auxiliary Feedwater Pump Room Coolers  Logic Diagram Number          NA          Elementary Number        D-177229 Sh. 1, D-207229 Sh. 1                                        Engineering Flow Diagram Numbers:        D-175011 Sh. 3,  D-205011 Sh. 3 Failure Mode Effect on System  Detection of Failure          Remarks Loss of 600-V Motor stops a. Fan fault alarm at Two MD auxiliary feedwater pumps  motor control      main control board provided; each unit is provided center A or  b. Loss of red/green with its own cooler and fan; B      indication lights at emergency core cooling system        BOP panel analysis is based upon operation    c. Monitor light box of one pump        abnormal                    Loss of 120 V-ac Motor stops a. Fan fault alarm at      control power      main control board    b. Loss of red/green          indication lights at          BOP panel    c. Monitor light box        abnormal      Thermal overload Motor stops a. Fan fault alarm at  contacts open      at main control board    b. Loss of red/green        indication lights at        BOP panel    c. Monitor light box        abnormal     
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-12  FAILURE MODE AND EFFECTS ANALYSIS, BATTERY ROOM VENTILATION SYSTEM  Component Identification Battery Room Exhaust Fan  Logic Diagram Number          NA          Elementary Number          D-177265 Sh. 1, D-207265 Sh. 1        Engineering Flow Diagram Numbers:      D-175014 Sh. 1, D-205014 Sh. 1 Failure Mode    Effect on System  Detection of Failure            Remarks Loss of 208-V Fan stops a. Loss of red/green indication One exhaust fan is provided for  motor control      lights at BOP Panel each battery room; one battery is  center A or B  required during post-LOCA operation;    effect of loss of exhaust fan on    hydrogen accumulation is discussed    in paragraph 9.4.2.3.4      Loss of 120 V-ac Fan stops a. Loss of red/green indication  control power      lights at BOP Panel     
Thermal overload Fan stops a. Fan operating light out  contacts open       
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-13  FAILURE MODE AND EFFECTS ANALYSIS, BATTERY ROOM AIR CONDITIONING SYSTEM  Component Identification Battery Room Cooler  Logic Diagram Number          NA          Elementary Number              NA                  Engineering Flow Diagram Number  NA   
See Table 9.4-6. 
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-14 (SHEET 1 OF 3)  FAILURE MODE AND EFFECTS ANALYSIS, EMERGENCY DIESEL GENERATOR Component Identification          Diesel Generator Supply Breaker          Elementary Number    D-172761, D-177142, and D-177143   
Failure Mode    Effect on System  Detection of Failure            Remarks Loss of 125 V-dc Loss of ability to tie diesel Loss of dc annunciator in Redundant diesel generator will be  control power generator to bus when control room started  necessary 
Failure of 2AJX Loss of ability to tie diesel Incomplete sequence Redundant diesel generator will be  contacts to generator to bus when annunciator started close in necessary  emergency Failure of  Loss of ability to tie diesel Indicating light on control Redundant diesel generator will be  59/81X contacts generator to bus when board started  necessary 
Mechanical or Loss of ability to tie diesel Breaker position indicating Redundant diesel generator will be  electrical generator to bus when lights in control room started  failure of necessary  breaker   
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-14 (SHEET 2 OF 3)  Component Identification Diesel Generator Start, Stop, and Shutdown Controls  Elementary Numbers      D-172774, D-172778, D-172782 Failure Mode    Effect on System  Detection of Failure            Remarks Loss of 125 V-dc Loss of ability to start Annunciator and loss of Redundant diesel generator will be  control power diesel generator in emergency indicating lights on board started Failure of start None; redundant starting Testing  contact in  circuit will start diesel  diesel starting    circuit A or B    to close in    emergency Failure of a None; redundant starting Testing  relay in circuit will start diesel  starting    circuit A or B Failure of Loss of ability to stop Diesel running light in Diesel can be stopped manually  signal contact diesel from control room control room  or relay in    diesel stop    circuit Failure of Diesel would not shut down Observation of diesel All safety features are cut out  contact or when trouble occurred failure except overspeed and low oil  relay in    pressure during emergency diesel  operation shutdown    circuit FNP-FSAR-7 REV 21  5/08 TABLE 7.3-14 (SHEET 3 OF 3)  Component Identification Diesel Generator Excitation and Miscellaneous Controls  Elementary Numbers          D-172775, D-172779, D-172783         
Failure Mode    Effect on System  Detection of Failure            Remarks Failure of Diesel generator may not Observation of voltage and Redundant generator can be used governor pick up load or may drop frequency on board  control load during load    fluctuations 
Failure of Improper voltage output Observation of meter on Redundant generator can be used excitation from generator board  circuit Failure of Improper voltage output Observation of meter on Redundant generator manual voltage auto voltage from generator board control can be used 
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 1 OF 6)  FAILURE MODE AND EFFECTS ANALYSIS, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Train A  Component Malfunction                                  Comment A1 Fail logic zero Prevents manual reset and block of safety injection train A Fail logic one Safety injection from train A not affected if failure occurs before  safety injection is called for at M1; safety injection signal will    be removed upon reactor trip indicated by P-4, safeguards sequencer must latch in and continue sequence If failure occurs after safety injection is called for at M1, safety injection signal will be reset; no effect provided safeguards    sequencer latches in A2 Fail logic zero Prevents reset of safety injection if one of the four inputs to O4 is still calling for safety injection Fail logic one Automatic safety injection actuation will be prevented; manual safety injection is still possible A3 Fail logic zero Automatic safety injection actuation train A will be prevented; manual safety injection train A is still possible Fail logic one Spurious safety injection train A; no direct reactor trip A4 Fail logic zero Prevents high containment pressure safety injection actuation train A if called for A5 Fail logic zero Prevents low steam line pressure; safety injection actuation in train A if called for Fail logic one Spurious safety injection; reactor trip and steam line isolation in train A A6 Fail logic zero Prevents steam line isolation on high steam line flow coincident with low-low Tavg (train A only)
Fail logic one Spurious steam line isolation in train A A7 Fail logic zero Prevents steam line isolation on high steam line flow coincident with low-low Tavg (train A only)
Fail logic one Spurious steam line isolation (train A) if false logic one output of A7 occurs coincident with low-low Tavg FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 2 OF 6)  Component Malfunction                                  Comment    A8 Fail logic zero Prevents train A safety injection actuation by low steam line pressure Fail logic one Spurious safety injection and steam line isolation (train A) if false logic one output of A8 occurs coincident with low-low Tavg A9 Fail logic zero Prevents low-low Tavg and low steam line pressure safety injection block Fail logic one Prevents low steam line pressure safety injection and steam line isolation actuation in train A if called for A10 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious train A safety injection and reactor trips A11 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-1, P-3) indicated by 2/3 logic at A12 A12 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-1, P-2) indicated by 2/3 logic at A11 A13 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 2 Fail logic one Spurious safety injection train A and reactor trip A14 Fail logic zero Loss of protection against high steam line differential pressure, i.e., low pressure loop 2 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-2, P-3) indicated by 2/3 logic at A15 A15 Fail logic zero Loss of protection against high steam line differential pressure, i.e., low pressure loop 2 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-1, P-2) indicated by 2/3 logic at A14 A16 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection train A and reactor trip FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 3 OF 6)  Component Malfunction                                  Comment    A17 Fail logic zero Loss of protection against high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-1, P-3) indicated by 2/3    logic at A18 A18 Fail  logic zero Loss of protection against high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-2, P-3) indicated by 2/3    logic at A17 A19 Fail logic zero Prevents train A low pressurizer pressure safety injection and reactor trip actuation if called for Fail logic one Spurious safety injection and reactor trip A20 Fail logic zero Prevents train A low pressurizer pressure safety injection and reactor trip      Fail logic one Spurious safety injection and reactor trip train A if not blocked    A21 Deleted A22  Deleted      A23 Fail logic zero Prevents pressurizer safety injection block Fail logic one Prevents train A low pressurizer pressure safety injection and reactor trip actuation if called for A24 Fail logic zero Allows train A safety injection blocks when no block is called for Fail logic one Prevents pressurizer safety injection block   
FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 4 OF 6)  Component Malfunction                                  Comment    A25 Fail logic zero Prevents steam line safety injection block of low steam line pressure; also prevents train A steam line isolation due  to high steam line flow coincident with low-low Tavg Fail logic one Allows operator to block safety injection whether or not block should be allowed; if safety injection is not blocked  and false output occurs coincident with low steam line pressure, spurious safety injection would result Spurious steam line isolation would occur if high steam line flow (2/3) occurs O1 Fail logic zero Prevents safety injection actuation train A Fail logic one Spurious safety injection; no direct reactor trip O2 Fail logic zero Prevents reset of safety injection if one of the four inputs to O4 is still calling for safety injection Fail logic one Safety injection actuation from train A not affected if failure occurs after safety injection is called for at M1;  if failure occurs before safety injection actuation, train A safety injection can be spuriously blocked (if P-4 is also logic one)
O3 Fail logic zero Prevents manual safety injection actuation train A Fail logic one Spurious reactor trip and safety injection train A O4 Fail logic zero Automatic safety injection actuation train A will be prevented; manual safety injection train A is still possible Fail logic one Spurious reactor trip; spurious safety injection train A if safety injection has not been manually blocked O5 Fail logic zero Loss of protection in train A against high steam line flow in loop 1; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 O6 Fail logic zero Loss of protection in train A against high steam line flow in loop 2; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 O7 Fail logic zero Loss of protection in train A against high steam line flow in loop 3; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 5 OF 6)  Component Malfunction                                  Comment    O8 Deleted not necessary O9 Fail logic zero Prevents safety injection block and allows spurious safety injection and steam line isolation if in  coincidence with low steam line pressure Fail logic one Blocks safety injection (train A) if low steam line pressure occurs coincident with low-low Tavg; safety injection  is not prevented if low steam line pressure occurs alone (not in coincidence with low-low Tavg)
O10 Fail logic zero Prevents train A high steam line differential pressure safety injection and reactor trip actuation Fail logic one Spurious safety injection and reactor trip O11 Deleted 
O12 Fail logic zero Prevents pressurizer safety injection block Fail logic one Blocks pressurizer safety injection if failure occurs coincident with P-11 (-1)
O13 Fail logic zero Prevents safety injection from actuating reactor trip Fail logic one Spurious reactor trip O14 Fail logic zero Prevents steam line isolation actuation of high steam line flow coincident with low-low Tavg and prevents steam line isolation    of low steam line pressure when safety injection is called for Fail logic one Spurious steam line isolation actuation N1 Fail logic zero Safety injection will be blocked at A3 although no attempt to reset has taken place Fail logic one Prevents safety injection block at A3 when resetting N2 Fail logic zero Blocks safety injection actuation (train A) of low steam line pressure when no block is called for Fail logic one Fails to block safety injection when block is called for N3 Fail logic zero Prevents steam line safety injection block Fail logic one Prevents manual reset of steam line safety injection block control; allows continuous block FNP-FSAR-7 REV 21  5/08 TABLE 7.3-15 (SHEET 6 OF 6)  Component Malfunction                                  Comment    N4 Fail logic zero Blocks train A safety injection actuation due to low pressurizer pressure coincident with low pressurizer level  when no block is called for Fail logic one Prevents manual block N5 Fail logic zero Prevents operator block of safety injection and reactor trip train A when block should occur Fail logic one Allows operator block of safety injection and reactor trip train A when block should not be allowed N6 Fail logic zero Prevents pressurizer safety injection block Fail logic one Prevents manual reset of pressurizer safety injection block control; allows continuous block TD1 Fail logic zero Prevents manual reset and block Fail logic one Short time delay Allows resetting of safety injection train A before safety injection sequence time delay has been completed Constant output Allows a manual block and reset at any time M1 Fail logic zero Prevents safety injection train A Fail logic one Spurious safety injection train A; prevents reset of safety injection signal Bistable Any one bistable Protection ensured by operation of other bistable inputs in the same system (coincidence changed from 2/3 to 1/2, etc.) inputs to input fails logic  solid state one  protection Any one bistable Protection ensured by operation of other bistable inputs in the same system (coincidence changed from 2/3 to 2/2, etc.)  input to solid  state protection  fails logic zero   
REV 21  5/08 COMPONENT IDENTIFICATION ESFAS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 1 OF 4)
REV 21  5/08 COMPONENT IDENTIFICATION ESFAS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 2 OF 4)
REV 21  5/08 COMPONENT IDENTIFICATION ESFAS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 3 OF 4)
REV 21  5/08 COMPONENT IDENTIFICATION ESFAS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 4 OF 4)
FNP-FSAR-7    7.4-1 REV 23  5/11 7.4  SYSTEMS REQUIRED FOR SAFE SHUTDOWN The functions necessary for safe shutdown are available from instrumentation channels that are associated with the major systems in both the primary and secondary sides of the nuclear steam supply system. These channels are normally aligned to serve a variety of operational functions, including startup and shutdown as well as protective functions. In achieving a safe shutdown, benefit is taken from many of these systems and equipment having multiple functions, and as such there are no specifically identifiable safe shutdown systems per se. However, prescribed procedures for securing and maintaining the plant in a safe condition can be instituted by appropriate alignment of selected nuclear steam supply systems. The discussion of these systems and the applicable codes, criteria, and guidelines are to be found in other sections of the report. In addition, the alignment of shutdown functions associated with the engineered safety features, which are invoked under postulated limiting fault situations, is discussed in chapter 6 and section 7.3. 
The instrumentation and control functions which are identified as being required for maintaining safe shutdown of the reactor are by definition the minimum required under nonaccident conditions.  (Control room inaccessibility as well as offsite power interruptions during a hot shutdown are considered as incidents.) 
These functions will permit the necessary operations that will: 
A. Prevent the reactor from achieving criticality in violation of the plant technical specifications.
B. Provide an adequate heat sink such that design and safety limits are not exceeded.
7.4.1  DESCRIPTION  The designation of systems that can be used for safe shutdown depends on identifying those systems which provide the following capabilities for maintaining a safe shutdown: 
A. Boration with related charging and letdown. 
B. Adequate supply for auxiliary feedwater. 
C. Residual heat removal. 
These systems and the associated instrumentation and controls provisions are identified in the following lists. The identification of the monitoring indicators (paragraph 7.4.1.1) and controls (paragraph 7.4.1.2) are those necessary for maintaining a hot shutdown. The essential services for the capabilities necessary for maintaining a hot shutdown are listed in paragraph 7.4.1.3, with the equipment and services available for a cold shutdown identified in paragraphs 7.4.1.4 and 7.4.1.5. 
See subsection 7.1.4 and Table 7.1-1 for a list of supplemental drawings. 
FNP-FSAR-7    7.4-2 REV 23  5/11 Periodic testing of remote shutdown system instrumentation and controls is conducted in accordance with the Technical Specifications.
7.4.1.1  Monitoring Indicators  The characteristics of these indicators, which are provided outside as well as inside the control room, are described in Section 7.5. The necessary indicators are as follows: 
A. Water level indicator for each steam generator. 
B. Pressure indicator for each steam generator. 
C. Pressurizer water level indicator. 
D. Pressurizer pressure indicator. 
E. Reactor coolant loop 1 hot leg temperature. 
F. Reactor coolant loop 1 cold leg temperature. 
G. Neutron flux.
H. Condensate storage tank level. 
7.4.1.2  Controls  7.4.1.2.1  General Considerations    A. The turbine is tripped.  (Note that this can be accomplished at the turbine as well as in the control room.)
B. The reactor is tripped.  (Note that this can be accomplished at the reactor trip switchgear as well as in the control room.)
C. All automatic systems continue functioning.  (These are discussed in Sections 7.2 and 7.7.)
D. For equipment having motor controls outside the control room (which duplicate the functions inside the control room), the controls will be provided with a selector switch which transfers control of the switchgear from the control room to a selected local station. Placing the local selector switch in the local operating position will give an annunciating alarm in the control room and will turn off the motor control position lights on the control room panel.   
FNP-FSAR-7    7.4-3 REV 23  5/11 7.4.1.2.2  Pumps    A. Auxiliary Feedwater Pumps 
Auxiliary feedwater pumps (electric) will start automatically on the loss of both main feedwater pumps. Start/stop motor controls positioned locally (and inside the control room) as well as handwheel control for the valves are provided. It is noted below that emergency power is available from the diesels which can be started locally and that the loads such as valves and pumps will be sequenced as necessary.
B. Charging and Boric Acid Transfer Pumps 
Start/stop motor controls are provided for these pumps. The controls for the charging and boric acid pumps are positioned locally (and in the control room).
C. Service Water Pumps 
These pumps, by means of the onsite power system, will start automatically following a loss of normal electrical power. Start/stop motor controls located outside and inside the control room will be provided.
D. Component Cooling Water Pumps 
These pumps, energized from the diesel generator, start automatically following a loss of normal electrical power. Start/stop controls located outside and inside the control room are provided.
E. Instrument Air Compressors 
These compressors start automatically on low air pressure. 
F. Reactor Containment Fan Cooler Units 
Start/stop motor controls with a selector switch are provided for the fan motors. The controls are located outside and inside the control room.
G. Control Room Ventilation Unit Including the Control Room Air Inlet Dampers 
A start/stop switch located outside the control room is provided for this unit(s). Also a control to close the inlet air damper(s) is provided. These controls duplicate functions inside the control room.
7.4.1.2.3  Diesels  These units start automatically following a loss of normal ac power. However, manual controls for diesel startup are also provided locally at the diesel generators (as well as in the control room), and loading is sequenced automatically. 
FNP-FSAR-7    7.4-4 REV 23  5/11 7.4.1.2.4  Valves    A. Charging Flow Control Valves 
Manual control with a selector switch outside the control room is provided for the charging line flow control valves. This control duplicates functions available in the control room.
B. Letdown Orifice Isolation Valves 
Open/close controls with a selector switch for the letdown orifice isolation valves are grouped with the controls for the charging flow control valve. These controls duplicate functions that are inside the control room.
C. Auxiliary Feedwater Control Valves 
Manual control is provided in the auxiliary feedwater pump area that duplicates functions that are inside the control room. A handwheel is also provided for each valve.
D. Atmospheric Steam Relief Valves 
Atmospheric relief valves are automatically controlled. Manual control is provided locally and inside the control room for the atmospheric relief valves. A handwheel is also provided for each valve.
E. Auxiliary Feedwater Pump Speed Control 
Manual speed control (mechanical device) is provided locally and in the control room for the steam-driven auxiliary feedwater pump.
F. Pressurizer Heater Control 
On/off control with selector switches are provided for two backup heater groups. The heater group will be connected to separate buses, such that each can be connected to separate diesels in the event of loss of outside power.
The control is grouped with the charging flow controls and duplicates functions available in the control room. 
It is noted that the instrumentation and controls listed in subsections 7.4.1.1 and 7.4.1.2 for achieving and maintaining a safe shutdown are available in the event an evacuation of the control room is required. Cable routing of key instrumentation loops will allow the plant to be brought to hot standby from the hot shutdown panel with the loss of either the cable spreading room, control room, or a cable chase. These controls and instrumentation channels, with the equipment and services identified in subsections 7.4.1.3 and 7.4.1.4 which are available for both hot and cold shutdown, identify the potential capability for cold shutdown of the reactor subsequent to a control room evacuation through the use of suitable procedures. Therefore, the applicable requirements of 1971 General Design Criterion 19 are met. 
FNP-FSAR-7    7.4-5 REV 23  5/11 7.4.1.3  Essential Services after Incident That Requires Hot Shutdown A. Auxiliary feedwater pumps which start automatically within 1 min for blackout condition.  (See chapter 10.)
B. Reactor containment air recirculation fans and coolers.  (See chapter 6.) 
C. Diesel generators, loaded within 1 min.  (See chapter 8.) 
D. Lighting in the areas of plant required during this condition.  (See subsection 9.5.3.)
E. Pressurizer heaters.  (See chapter 5.) 
F. Communication network (see subsection 9.5.2) to be available for prompt use between feedwater pumps area and the following areas:
1. Feedwater source from outside. 
2. Charging pump. 
3. Boric acid transfer pump. 
4. Diesel generator. 
5. Switchgear room. 
6. Steam relief valves. 
G. Boric acid transfer pumps.  (See chapter 9.) 
H. Charging pumps.  (See chapter 9.) 
I. Service water pumps.  (See chapter 9.) 
J. Component cooling pumps.  (See chapter 9.) 
K. Instrument air compressors.  (See chapter 9.) 
L. Control room ventilation unit and air inlet damper.  (See chapter 9.) 
FNP-FSAR-7    7.4-6 REV 23  5/11 7.4.1.4  Equipment and Systems Available for Cold Shutdown A. Reactor coolant pump.  (See chapter 5.)(a)
B. Auxiliary feedwater pumps.  (See chapter 10.) 
C. Boric acid transfer pump.  (See chapter 9.) 
D. Charging pumps.  (See chapter 9.) 
E. Service water pumps.  (See chapter 9.) 
F. Containment fans.  (See chapter 6.) 
G. Control room ventilation.  (See chapter 9.) 
H. Component cooling pumps.  (See chapter 9.) 
I. Residual heat removal pumps.  (See chapter 5.)(a)
J. Motor control center and switchgear sections associated with above loads. 
K. Controlled steam release and feedwater supply.  (See section 7.7 and chapter 10.) 
L. Boration capability.  (See chapter 9.) 
M. Nuclear instrumentation system (source range and intermediate range).  (See sections 7.2 and 7.7.)(a)
N. Reactor coolant inventory control (charging and letdown).  (See chapter 9.) 
O. Pressurizer pressure control including opening control for pressurizer relief valves (heaters and spray).  (See chapter 5.)(a)
In addition, the safety injection signal trip circuit must be defeated and the accumulator isolation valves closed.(a)  The performance of the emergency core cooling system under these  conditions was evaluated. Conditions during plant cooldown were divided into the following four 
a. Instrumentation and controls for these systems may require some modification in order that their functions may be performed from outside the control room. Note that the reactor plant design does not preclude attaining the cold shutdown condition from outside the control room.
An assessment of plant conditions can be made on a long term basis (a week or more) to establish procedures for making the necessary physical modifications to instrumentation and control equipment in order to attain cold shutdown. During such time the plant could be safely maintained at hot shutdown condition. 
Detailed procedures to be followed in effecting cold shutdown from outside the control room are best determined by plant personnel at the time of the postulated incident.
FNP-FSAR-7    7.4-7 REV 23  5/11 phases:  (1) from operating reactor coolant pressure to 1900 psig, (2) from 1990 to 1000 psig, (3) from 1000 to 400 psig, and (4) from 400 psig to cold shutdown. The break size used in the analysis was determined using the moderate energy line break criteria identified in Branch Technical Positions APCSB 3-1 and MEB 3-1. Based on the analysis, the available emergency core cooling system can cool the core under plant cooldown conditions and, therefore, meets the NRC acceptance criteria, as applicable, contained in 10 CFR 50.46 and 10 CFR 50, Appendix K.
7.4.2  ANALYSIS  Hot shutdown is a stable plant condition, automatically reached following a plant shutdown. The hot shutdown condition can be maintained safely for an extended period of time either automatically or manually. In the unlikely event that access to the control room is restricted, the plant can be safely kept at a hot shutdown until the control room can be reentered by the use of the monitoring indicators and the controls listed in paragraphs 7.4.1.1 and 7.4.1.2. These indicators and controls are provided outside and inside the control room. The safety evaluation of the maintenance of a shutdown with these systems and associated instrumentation and controls has included consideration of the accident consequences that might jeopardize safe shutdown conditions. The accident consequences that are germane are those that would tend to degrade the capabilities for boration, adequate supply for auxiliary feedwater, and residual heat removal. 
The results of the accident analyses are presented in chapter 15. Of these the following produce the most severe consequences that are pertinent: 
A. Uncontrolled boron dilution. 
B. Loss of normal feedwater. 
C. Loss of external electrical load and/or turbine trip. 
D. Loss of all ac power to the station auxiliaries (station blackout). 
It is shown by these analyses that safety is not compromised by these incidents, with the associated assumptions being that the instrumentation and controls indicated in paragraphs 7.4.1.1 and 7.4.1.2 are available to control and/or monitor shutdown. These available systems will allow a maintenance of hot shutdown even under the accident conditions listed above which would tend toward a return to criticality or a loss of heat sink. 
FNP-FSAR-7    7.5-1 REV 21  5/08 7.5  POSTACCIDENT MONITORING DISPLAY INSTRUMENTATION 7.5.1  DESCRIPTION  Table 7.5-1 lists the instrumentation provided to the operator to perform necessary functions, assess plant conditions, and verify system performance during accident situations. Listed below are the five classifications of variables that have been identified to provide this instrumentation.
Type A: Those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events.
Type B: Those variables to be monitored that provide to the control room operator information to assess the process of accomplishing or maintaining critical safety functions.
Type C: Those variables to be monitored that provide to the control room operator information to monitor (1) the extent to which parameters, which have the potential for causing a breach of the primary reactor containment or fuel cladding, have exceeded the design basis values, or (2) that the in-core fuel clad, the reactor coolant system pressure boundary or the primary reactor containment may have been breached.
Type D: Those variables that provide information to indicate the operation of individual safety systems and other systems important to safety.
Type E: Those variables that are to be monitored as required for use in determining the magnitude of the release of radioactive materials and in continuously assessing such releases.
These variables are subdivided into three categories which define the qualification requirements of the instrumentation. Table 7.5-1 identifies the variable category identified in the R. G. 1.97 Compliance Report. The qualification and configuration requirements and the Farley Evaluation Criteria for specific R. G. 1.97 requirements are described in the Design and Qualification Review Criteria section of the R. G. 1.97 Compliance Report.
The instrumentation channels that provide the information for the variables listed in Table 7.5-1, are powered as described in the R. G. 1.97 Compliance Report, and are energized from the onsite electrical power supplies as described in chapter 8. 
Table 7.5-3 lists the information available to the operator for monitoring conditions in the reactor, in the reactor coolant system, and in the containment and process systems throughout all normal operating conditions of the plant, including anticipated operational occurrences. 
Post-accident monitoring instrumentation is discussed in the Technical Specifications.
FNP-FSAR-7    7.5-2 REV 21  5/08 Containment hydrogen monitoring instrumentation surveillance is discussed in FSAR section 16.1.
7.5.2  ANALYSIS  With the issuance of Regulatory Guide 1.97, Alabama Power Company performed a comprehensive review and issued a R. G. 1.97 Compliance Report documenting Farley's commitment to R. G. 1.97.
The display instrumentation for postaccident monitoring enables the required manual functions to be performed following a Condition II, III, or IV event to provide the necessary information to maintain the plant at a safe hot shutdown or to proceed to a cold shutdown condition consistent with the technical specification limits. Other design criteria used in the display system are given at the end of this section. 
All commitments concerning recording, separation, qualification, and redundancy are provided in the R. G. 1.97 Compliance Report.
For postaccident scenarios (see table 7.5-1), sufficient duplication of information is provided to ensure that the minimum information required will be available. The information is part of the operational monitoring of the plant which is under surveillance by the operator during normal plant operation. This is functionally arranged on the control board to provide the operator with ready understanding and interpretation of plant conditions. Comparisons between duplicate information channels or between functionally related channels will enable the operator to readily identify a malfunction in a particular channel. 
It is noted that there is a degree of functional redundancy between those display channels that are required for postaccident monitoring and many other diverse instrumentation channels which are also located on the main control board. For example, after the actuation of safety injection, the residual heat removal pump flow, high head (charging) pump flow, and spray pump flow can be verified by their respective flow channels. The transmitters for these flow channels are outside the containment. In addition, the containment sump level is continuously read out on the main control board. This information provides a diverse means for checking refueling water storage tank level data obtained from the safety-related display information. 
Channel separation is provided between sensors and the process cabinets. From the process cabinets to the main control board, the interconnecting circuits meet the separation requirements between safety trains, with two channels being associated with one train. 
The design criteria used in the display system are listed below: 
A. Range and accuracy requirements are determined through the analyses of postaccident conditions as described in chapter 15. The display system meets the following requirement: the range of the readouts extends over the maximum expected range of the variable being measured.
B. Power for the display instruments is obtained from the instrumentation and control power supply system. This system is described in section 7.6 and complies with FNP-FSAR-7    7.5-3 REV 21  5/08 paragraph 5.4 of the Institute of Electrical and Electronics Engineers standard 308.
C. Those channels determined to provide useful information in charting the course of events are recorded.
7.5.3  DELETED    7.5.4  INADEQUATE CORE COOLING MONITORING SYSTEM  The inadequate core cooling monitoring system (ICCMS) is a safety grade processing and display system which meets the NRC requirements to provide the capability to monitor the approach to, existence of, and recovery from potential reactor core inadequate core cooling situations. The requirements addressed by the ICCMS are defined in paragraph II.f.2 of NUREG 0737, "Clarification of TMI Action Plan" and Generic Letter 82-28. Inadequate core cooling monitoring requirements are met by measuring and displaying margin to saturation, reactor vessel water level above the core, and core exit temperatures. 
7.5.4.1  Reactor Vessel Level  The redundant heated junction thermocouple (HJTC) probes are described in paragraph 4.4.5.5. Redundant processors are located in the control room. Redundant level indication is provided on a reactor vessel mimic display on the main control board. The mimic indicates covered or uncovered for each of the eight heated junctions for each probe. The mimic background shows the elevation of each sensor and its location in the reactor vessel in relation to major components and penetrations.
The redundant signal processors, one per HJTC probe, monitor the HJTC probe thermocouples, control power to the HJTC probe heaters, and drive the level displays. The processor HJTC calculations are as follows:
The differential temperature (T) is calculated from the temperature values for the unheated junction (Tu) and the heated junction (TH) thermocouple inputs. T is equal to TH minus Tu, and that T is compared against a low T setpoint (25&deg;F). If T is less than the low T setpoint the corresponding error number is set. A low T error indicates that there is a loss of heater power or a heater controller malfunction. There are two heater controllers per channel. Each heater controller is connected to four heaters in series. If one heater fails open then all the rest of the heaters will be turned off. This will cause either all the odd numbered Ts or even numbered Ts to have a low T error. T or Tu is used to determine percent level for both the head area and the plenum area. A sensor is considered uncovered whenever T or Tu is greater than 200&deg;F or 700&deg;F, respectively. Five-degree dead bands exist in both the T and Tu setpoints for uncovered sensors to prevent cycling.
FNP-FSAR-7    7.5-4 REV 21  5/08 A maximum TH and a maximum T are selected and are used to calculate separate setpoint signals for the heater controllers. The minimum of the two heater controller setpoint signals is selected and sent to each of the heater controllers. The TH and T heater controller setpoint signals are reduced at a constant rate, when their respective TH and T values increase above a predetermined value. The TH and T heater controller setpoints will decrease until they equal zero at a second predetermined setpoint.
7.5.4.2  Subcooling Margin Monitor  The subcooling margin monitor (SMM) provides continuous, redundant indication of the margin to saturated conditions in the reactor coolant system (RCS). The SMM inputs are RCS hot leg and cold leg temperatures from loop RTDs, core exit thermocouple temperature, RCS wide range pressure, and pressurizer pressure. The margin to saturation, displayed in degrees F, is the difference between the measured RCS temperature and the saturation temperature. The highest RCS loop temperature and the highest core exit thermocouple temperature, excluding upper head thermocouples, are used to calculate margins to saturation. The lowest pressure value is used to calculate the saturation temperature. The control board SMM display has a switch to select margin to saturation indication based on RCS loop RTD temperature or core exit thermocouple temperature. 
7.5.4.3  Core Exit Temperature  Core exit temperature is continuously indicated on redundant control board displays. The chromel-alumel thermocouples in the vessel measure temperature at the flow exit of selected fuel assemblies and locations within the reactor vessel head plenum. 
The redundant displays each normally indicate the temperature of the hottest thermocouple for that channel. The operator can interrogate the display to indicate the temperature of any individual thermocouple or the highest temperature in each core quadrant. 
7.5.5  NUCLEAR INSTRUMENTATION  In addition to the Westinghouse nuclear instrumentation system that is described in section 7.2 and whose indications are listed in table 7.5-3, an independent channel of Gamma-Metrics nuclear instrumentation is provided to satisfy alternate shutdown requirements. 
The Gamma-Metrics channel provides neutron flux indication at the hot shutdown panel and the control room via isolated outputs. A fission chamber detector measures neutron flux from shutdown to full power. Detector sensitivity is 10-2 to 1010 nv. The following displays are provided on the main control board and the hot shutdown panel:
FNP-FSAR-7    7.5-5 REV 21  5/08  Source Range 0.1 to 105 counts/s    Source Range Startup Rate-1 to 7 decades/min. Log Power Level 10-8 to 100-percent power FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 1 OF 16)  POST ACCIDENT INSTRUMENTATION TYPE A VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION      VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY      Plant Specific 1 Information required for operator action 1 RCS Pressure (wide range) 1    2 RCS Hot Leg Temperature (wide range) 1    3 RCS Cold Leg Temperature (wide range) 1    4 Steam Generator Level (wide range) 1    5 Steam Generator Level (narrow range) 1    6 Pressurizer Level 1    7 Containment Pressure (normal range) 1    8 Main Steam Line Pressure 1    9 Refueling Water Storage Tank Level 1    10 Containment Water Level 1    11 Condensate Storage Tank Level 1    12 Auxiliary Feedwater Flow 1    15 Core Exit Temperature 1    132 Core Subcooling Monitor 2 
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 2 OF 16)  TYPE B VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Reactivity Control        Neutron Flux 1 Function detection; accomplishment of mitigation 17 Neutron Flux (intermediate range) 1        Control Rod Position 3 Verification 1009 Control Rod Position 3        RCS Soluble Boron Concentration 3 Verification 1017 Post Accident Sample 3  RCS Cold Leg Water Temperature 3 Verification 3 RCS Cold Leg Temperature (wide range) 1      Core Cooling        RCS Hot Leg Water Temperature 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 2 RCS Hot Leg Temperature (wide range) 1        RCS Cold Leg Water Temperature 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 3 RCS Cold Leg Temperature (wide range) 1 
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 3 OF 16)  TYPE B VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY        RCS Pressure 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 1 RCS Pressure (wide range) 1      Core Exit Temperature 3 Verification 15 Core Exit Temperature 1        Coolant Inventory 1 Verification; accomplishment of mitigation; 18 Reactor Water Level 1      Degrees of Subcooling 2 Verification and analysis of plant conditions 132 Core Subcooling Monitor 2      Maintaining Reactor Coolant System Integrity        RCS Pressure 1 Function detection; accomplishment of mitigation 1 RCS Pressure (wide range) 1      Containment Sump Water Level (narrow range) 2 Function detection; accomplishment of mitigation; verification 111 Reactor Cavity Sump Level 2 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 4 OF 16)  TYPE B VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY        Containment Sump    Water Level (wide range) 1 Function detection; accomplishment of mitigation; verification 10 Containment Water Level 1      Containment Pressure 1 Function detection; accomplishment of mitigation; verification 7 Containment Pressure (normal range) 1      Maintaining Containment Integrity        Containment                Isolation Valve Position (excluding check valves) 1 Accomplishment of isolation 19 Containment Isolation Valve Position 1          Containment Pressure 1 Function detection; accomplishment of mitigation; verification 7 Containment Pressure (normal range) 1       
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 5 OF 16)  TYPE C VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Fuel Cladding        Core Exit                    Temperature 1 Detection of potential for breach; accomplishment of mitigation; long-term surveillance  15 Core Exit Temperature 1        Radioactivity Concentration or Radiation Level in Circulating Primary Coolant 1 Detection of breach 14 Primary Coolant Radioactivity Concentration 1        Analysis of Primary Coolant (gamma spectrum) 3 Detail analysis; accomplishment of mitigation; verification; long-term surveillance 1017 Post Accident Sample 3      Reactor Coolant Pressure Boundary        RCS Pressure 1 Detection of potential for or actual breach; accomplishment of mitigation; long-term surveillance  1 RCS Pressure (wide range) 1         
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 6 OF 16)  TYPE C VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY        Containment              Pressure 1 Detection of breach; accomplishment of mitigation; verification; long-term surveillance  7 Containment Pressure (normal range) 1        Containment Sump Water Level (narrow range) 2 Detection of breach; accomplishment of mitigation; verification; long-term surveillance  111 Reactor Cavity Sump Level 2        Containment Sump    Water Level (wide range) 1 Detection of breach; accomplishment of mitigation; verification; long-term surveillance 10 Containment Water Level 1        Containment Area Radiation 3 Detection of breach; verification  13 Containment Radiation (high range) 1        Effluent Radioactivity -
Noble Gas Effluent from Condenser Air Removal System Exhaust 3 Detection of breach; verification  120 Condenser SJAE Radiation 2 
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 7 OF 16)  TYPE C VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Containment      RCS Pressure 1 Detection of potential for breach; accomplishment of mitigation 1 RCS Pressure (wide range)      1      Containment Hydrogen Concentration 1 Detection of potential for breach; accomplishment of mitigation; long-term surveillance 1006 Containment Hydrogen Concentration 3      Containment Pressure 1 Detection of potential for or an actual breach; accomplishment of mitigation  16 Containment Pressure (extended range) 1      Containment Effluent Radioactivity - Noble Gases from Identified Release Points 2 Detection of breach; accomplishment of mitigation; verification 121 Plant Vent Effluent Radiation 2      Effluent Radioactivity -
Noble Gases (from buildings or areas where penetrations and hatches are located, e. g.,
secondary containment and auxiliary buildings and fuel handling buildings that are in direct contact with primary containment) 2 Indication of breach 121 Plant Vent Effluent Radiation 2 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 8 OF 16)  TYPE D VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Residual Heat Removal (RHR) or Decay Heat Removal System        RHR System Flow 2 To monitor operation  101 RHR/LHSI Flow 2        RHR Heat Exchanger          Outlet Temperature 2 To monitor operation and for analysis 114 RHR HX Discharge Temperature 2      Safety Injection Systems        Accumulator Tank Level and Pressure 2 To monitor operation  125 1018 Accumulator Tank Pressure Accumulator Tank Level 2 3        Accumulator Isolation Valve Position 2 Operation status  126 Accumulator Tank Isolation Valve Position2        Boric Acid Charging Flow 2 To monitor operation  102 Boric Acid Flow 2        Flow in HPI System 2 To monitor operation  103 HHSI Flow 2        Flow in LPI System 2 To monitor operation  101 RHR/LHSI Flow 2        Refueling Water Storage Tank Level 2 To monitor operation  9 Refueling Water Storage Tank Level 1      Primary Coolant System        Reactor Coolant Pump        Status 3 To monitor operation 1011 RCP Motor Current 3 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 9 OF 16)  TYPE D VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY        Primary System Safety        Relief Valve Positions (including PORV and code valves) or Flow Through or Pressure in Relief Valve Lines 2 Operation status; to monitor for loss of coolant 127 128 Pressurizer PORV Position Pressurizer Safety Valve Position 2 2        Pressurizer Level 1 To ensure proper operation of pressurizer 6 Pressurizer Level  1  Pressurizer Heater Status 2 To determine operating status 130 112 Pressurizer Heater Breaker Position Pressurizer Pressure 2 2        Quench Tank Level 3 To monitor operation 1002 Pressurizer Relief Tank level 3        Quench Tank Temperature 3 To monitor operation 1004 Pressurizer Relief Tank Temperature  3  Quench Tank Pressure 3 To monitor operation 1007 Pressurizer Relief Tank Pressure 3      Secondary System (Steam Generator)        Steam Generator Level 1 To monitor operation 4 Steam Generator Level (wide range) 1        Steam Generator Pressure 2 To monitor operation 8 Main Steam Line Pressure 1          Safety/Relief Valve              Positions or Main Steam Flow 2 To monitor operation 104 Main Steam Flow 2          Main Feedwater Flow 3 To monitor operation 1001 Main Feedwater Flow 3 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 10 OF 16)  TYPE D VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Auxiliary Feedwater or Emergency Feedwater System        Auxiliary or Emergency        Feedwater Flow 2 To monitor operation 12 Auxiliary Feedwater Flow 1        Condensate Storage Tank  Water Level 1 To ensure water supply for auxiliary feedwater (can be Category 3 if not primary source of AFW. Then whatever is primary source of AFW should be listed and should be Category 1) 11 Condensate Storage Tank Level 1      Containment Cooling Systems        Containment Spray Flow 2 To monitor operation 105 Containment Spray Flow 2        Heat Removal by the          Containment Fan Heat Removal System 2 To monitor operation 115  116 133 Temperature of Service Water to Aux. Bldg
CTMT Cooler Service Water Outlet Temperature
Service Water Flow to CTMT Coolers 2  2 2      Containment Atmosphere Temperature 2 To indicate accomplishment of cooling 117 Containment Atmosphere Temperature 2 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 11 OF 16)  TYPE D VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY        Containment Sump Water  Temperature 2 To monitor operation 118 RHR HX Inlet Temperature 2      Chemical and Volume Control System        Makeup Flow - In 2 To monitor operation 106 110 Charging Line Flow RCP Seal Injection Flow 2 2  Letdown Flow- Out 2 To monitor operation 107 Letdown Flow 2        Volume Control Tank Level 2 To monitor operation 113 Volume Control Tank Level 2      Cooling Water System        Component Cooling Water Temperature to ESF System 2 To monitor operation 119 Component Cooling Water Heat Exchanger Discharge Temperature 2        Component Cooling Water Flow to ESF Sys 2 To monitor operation 108 CCW HX Inlet Flow 2      Radwaste Systems        High-level Radioactive Liquid Tank Level 3 To indicate storage volume 1003 Radioactive Liquid Tank Levels 3  Radioactive Gas Holdup Tank Pressure 3 To indicate storage capacity 1008 Waste gas Decay Tank Pressure 3      Ventilation Systems        Emergency Ventilation Damper Position 2 To indicate damper status 129 HVAC Emergency Damper Position 2 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 12 OF 16)  TYPE D VARIABLES R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Power Supplies        Status of Standby Power    and Other Energy Sources Important to Safety (electric, hydraulic, pneumatic) (voltages, currents, pressures) 2 To indicate system status 131 Emergency Power Status 2 
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 13 OF 16)  TYPE E VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO.
DESCRIPTION  CATEGORY      Containment Radiation        Containment Area                Radiation - High Range 1 Detection of significant releases; release assessment; long-term surveillance; emergency plan actuation 13 Containment Radiation (high range) 1      Area Radiation        Radiation Exposure rate (inside buildings or areas where access is required to service equipment important to safety) 3 Detection of significant releases; release assessment; long-term surveillance122 1005 Accessible Area Radiation Portable Plant/Environs Radiation 2 3      Airborne Radioactive Materials Released from Plant        Noble Gases and Vent Flow Rate        Containment or Purge Effluent 2 Detection of significant releases; release assessment  Not Applicable, see Common Plant vent    Reactor Shield Building Annulus (if in design) 2 Detection of significant releases; release assessment  Not Applicable, not in design   
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 14 OF 16)  TYPE E VARIABLES R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Auxiliary Building (including any building containing primary system gases, e.
g., waste gas decay tank) 2 Detection of significant releases; release assessment; long-term surveillance Not Applicable, see Common Plant Vent        Condenser Air Removal System Exhaust 2 Detection of significant releases; release assessment 120 Condenser SJAE Radiation 2      Common Plant Vent or Multi-purpose Vent Discharging Any of Above Releases (if containment purge is included) 2 Detection of significant release; release assessment; long-term surveillance121 109 Plant Vent Effluent Radiation Plant Vent Stack Flow  2 2      Vent From Steam Generator Safety Relief Valves or Atmospheric Dump Valves 2 Detection of significant releases; release assessment 104 123 124 Main Steam Flow Main Steam Effluent Radiation TDAFW Effluent Radiation 2 2
2      All Other Identified Release Points 2 Detection of significant releases; release assessment; long-term surveillance Not Applicable         
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 15 OF 16)  TYPE E VARIABLES  R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Particulates and Halogens      All Identified Plant Release Points (except steam generator safety relief valves or atmospheric steam dump valves and condenser air removal system exhaust). Sampling with Onsite Analysis Capability 3 Detection of significant releases; release assessment; long-term surveillance 1012 Particulates and Halogens Sampling (Vent Stack)  3      Environs Radiation and Radioactivity      Airborne Radiohalogens and Particulates (portable sampling with onsite analysis capability) 3 Release assessment; analysis 1013 Airborne Radiohalogens and Particulates (Environs) 3      Plant and Environs Radiation (portable instrumentation) 3 Release assessment; analysis 1005 Portable Plant/Environs Radiation 3      Plant and Environs Radioactivity (portable instrumentation) 3 Release assessment; analysis 1019 Portable Plant/Environs Radioactivity (Gamma-ray Spectrometer) 3      Meteorology      Wind Direction 3 Release assessment 1014 Wind Direction 3      Wind Speed 3 Release assessment 1015 Wind Speed 3      Estimation of Atmospheric Stability 3 Release assessment 1016 Estimation of Atmospheric Stability 3 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-1 (SHEET 16 OF 16)  TYPE E VARIABLES R.G. 1.97 VARIABLES  FNP POSITION  VARIABLE  CATEGORY  PURPOSE VARIABLE NO. DESCRIPTION  CATEGORY      Accident Sampling Capability (Analysis Capability On Site)      Primary Coolant and Sump -Gross Activity
-Gamma Spectrum
-Boron Content
-Chloride Content
-Dissolved Hydrogen or Total Gas -Dissolved Oxygen
-pH 3 Release assessment; verification; analysis 1017 Post Accident Sample 3      Containment Air -Hydrogen Content
-Oxygen Content
-Gamma Spectrum 3 Release assessment; verification; analysis 1010  Post Accident Sample - CTMT Air 3 FNP-FSAR-7  REV 21  5/08 TABLE 7.5-2 
  (This table has been deleted.)   
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-3 (SHEET 1 OF 6)  CONTROL ROOM INDICATORS AND/OR RECORDERS AVAILABLE TO THE OPERATOR TO MONITOR SIGNIFICANT PLANT PARAMETERS DURING NORMAL OPERATION    No. of        Channels    Indicated      Indicator/  Parameter Available    Range    Accuracy      Recorder    Location      Notes      Notes              Nuclear Instrumentation              Source range                Count rate 2 1 to 106 +/-7% of the linear Both channels indicated;  Control board One 2-pen record-    counts/s full scale analog either may be selected  er is used to re-    voltage for recording  cord any of the        8 nuclear chan-        nels (2 source        range, 2        intermediate        range, and 4        power range).          Startup rate 2 -0.5 to 5.0 +/-7% of the linear Both channels indicated  Control board      decades/min full scale analog        voltage          Intermediate range                Flux level 2 8 decades of +/-7% of the linear Both channels indicated;      neutron flux full scale analog either may be selected      (corresponds to voltage and +/-3% for recording      0 to full scale of the linear full      analog voltage) scale voltage in      overlapping the the range of 10-4      source range by to 10-3 A      2 decades              Startup rate 2 -0.5 to 5.0 +/-7% of the linear Both channels indicated Control board    decades/min full scale analog        voltage     
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-3 (SHEET 2 OF 6)    No. of        Channels    Indicated      Indicator/  Parameter Available    Range    Accuracy      Recorder    Location      Notes        Power range          Uncalibrated ion 4 0 to 120% of +/-1% of full span All 8 current signals NIS racks in    chamber current  full power  indicated control room    (top and bottom  current        uncompensated ion          chambers)                Calibrated ion 4 0 to 120% of +/-2% of full power These 8 current signals are      chamber current  full power  available for selectable      (top and bottom    trending by the operator      uncompensated ion          chambers)                                                                                        Upper and lower ion 4 -50 to +50% +/-3% of full power Diagonally opposed; Control board    chamber current    any 2 of the 4      difference    channels may be se-        lected for recording        at the same time        using intermediate        range recorder            Average flux of the 4 0 to 120% of +/-3% of full power All 4 channels Control board    top and bottom ion  full power for indication; indicated; any 2 of the      chamber  +/-2% for recording 4 channels may be        recorded using source        range recorder            Average flux of the 4 0 to 200% of +/-2 to 120% of All 4 channels Control board    top and bottom ion  full power full power; recorded      chambers  +/-6 to 200% of        full power     
FNP-FSAR-7  REV 21  5/08 TABLE 7.5-3 (SHEET 3 OF 6)    No. of        Channels    Indicated      Indicator/  Parameter Available    Range    Accuracy      Recorder    Location      Notes Flux difference of 4 -30 to +30% +/-4% All 4 channels  Control board    the top and bottom    indicated      ion chambers Reactor Coolant System Tavg (measured) 1 per 530&deg; to 630&deg;F +/-4&deg;F All channels indicated Control board    loop T (measured) 1 per 0 to 150% of +/-4% of full All channels indicated Control board    loop full power T power T Tcold or 1 Thot, 0&deg;F to 700&deg;F +/-4% Both channels recorded Control board    Thot (measured, 1 Tcold        wide range) per loop Overpower T 1 per 0 to 150% of +/-4% of full All channels indicated Control board  setpoint loop full power T power T Overtemperature 1 per 0 to 150% of +/-4% of full All channels indicated Control board  T setpoint loop full power T power T Pressurizer 5 1700 to +/-20 psi All channels indicated Control board  pressure  2500 psig Pressurizer 3 Entire distance +/-3.5% of P All channels indicated; Control board 2-pen recorder level  between taps at 2250 psia 1 channel is selected  used; second      for recording  pen records        reference level        signal Primary coolant 3 per 0 to 120% of Repeatability of All channels indicated Control board  flow loop rated flow +/-4% of full flow Reactor coolant 1 per 0 to 1200 amps - All channels indicated Control board One channel for pump bus amperes loop    each bus System pressure 2 0 to 3000 psig +/-4% All channels indicated Control board  wide range    and recorded FNP-FSAR-7  REV 21  5/08 TABLE 7.5-3 (SHEET 4 OF 6)    No. of        Channels Indicated Indicated    Parameter Available Range Accuracy Recorder Location Notes        Reactor Control System              Demanded rod speed 1 0 to 75 +/-2% 1 channel is indicated Control board    steps/min            Median Tavg 1 530&deg;F to 630&deg;F +/-4&deg;F 1 channel is recorded Control board                      Tref 1 530&deg;F to 630&deg;F +/-4&deg;F 1 channel is recorded Control board        Control rod position      If system not      available, borate      and sample ac-        cordingly        Number of steps 1 per 0 to 231 steps(a) +/-1 step Each group is indi- Control  board These signals are of demanded rod group  cated  used in conjunc- withdrawal      tion with the      full-length rod        measured position      signals to detect      deviation of any      individual rod      from the demanded      position; a      deviation will      actuate an alarm              Full-length rod 1 per 0 to 228 steps(b) +/-4 steps at full Each rod position Control board  measured position rod  accuracy; +/-8 is indicated      steps at 1/2      accuracy             
a. Fully withdrawn position can be varied from 225 to 231 steps to reduce RCCA wear. The NRC acceptance criteria regarding the range associated with the fully withdrawn RCCA position are that the fully withdrawn position selected for use throughout each cycle will be evaluated as part of the reload safety evaluation process to verify that sufficient margin exists in the safety analyses to bound the related effects. b. Digital Rod Position Indication (DRPI) system maximum indication is 228 steps.
FNP-FSAR-7  REV 21  5/08  TABLE 7.5-3 (SHEET 5 OF 6)  No. of      Channels  Indicated Indicated  Parameter Available Range Accuracy Recorder Location Notes          Control rod bank 4 0 to 100%(a) +/-3% of total All 4 control rod Control board 1 channel for    position  bank travel bank positions are  each control    recorded along with  rod; an alarm and    the low-low limit  annunciator are    alarm for each  actuated when the    bank  rod control bank to be      withdrawn reaches      withdrawal limit, when      any rod control bank      reaches the low      insertion limit,      and when any rod      control bank      reaches the low-      low insertion      limit Containment System Containment pressure 4 -5 to 65 psig +/-3% All 4 channels Control board      indicated Feedwater and Steam Systems Auxiliary feedwater 1 per 0 to 400 +/-3% All channels Control board 1 channel to measure flow - Unit 1 steam gal/min    the flow to each  line    steam generator Auxiliary feedwater 1 per 0 to 400 +/-3% All channels Control board 1 channel to measure flow - Unit 2 steam gal/min    the flow to each  line    steam generator Steam generator 3 per +6.2 to -11.5 ft from +/-3% of level All channels indicated; Control board  level (narrow range) steam nominal full load (hot) channels used for    generator level  control are recorded Steam generator 1 per +6.2 to -41.7 ft from +/-5% of level All channels recorded Control board  level (wide range) steam nominal full load (cold)      generator level                          a. One-hundred percent is the fully withdrawn position.
FNP-FSAR-7  REV 21  5/08  TABLE 7.5-3 (SHEET 6 OF 6)    No. of        Channels  Indicated Indicated  Parameter Available Range Accuracy Recorder Location Notes        Main feedwater 2 per 0 to 120% of +/-5% All channels indicated; Control board  flow steam maximum calcu-  channels used for    generator lated flow  control are recorded          Magnitude of 1 per 0 to 100% of +/-1.5% All channels indicated Control board 1 channel for signal control- main, valve opening    each main feed- ing main 1 per    water control  bypass    valve;open/shut        indication is        provided in        control room        for each        main feedwater        control valve        Steam flow 2 per 0 to 120% of +/-5.5% All channels indicated; Control board Accuracy is  steam maximum cal-  channels used for  equipment  generator culated flow  control are recorded  capability;        however, abso-        lute accuracy        depends on        applicant cali-        bration against        feedwater flow        Steam line 3 per 0 to 1200 psig +/-4% All channels indicated Control board  pressure loop            Steam dump 1 0 to 85% max- +/-1.5% 1 channel is indicated Control board Open/shut indi- modulate  mum calculated    cation is pro- signal  steam flow    vided in the        control room        for each steam        dump valve        Turbine impulse 2 0 to 120% of +/-3.5% Both channels indicated Control board Open/shut indi- chamber pressure  maximum calcu-    cation is pro-    lated turbine    vided in the    load    control room for      each turbine        stop valve 
REV 21  5/08 LOGIC DIAGRAM FOR RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-1 
REV 21  5/08 LOGIC DIAGRAM FOR RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-2 REV 21  5/08 LOGIC DIAGRAM FOR BACKUP TO  SEMIAUTOMATIC SWITCHOVER LOGIC FROM  INJECTION TO RECIRCULATION  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-3 REV 21  5/08 FUNCTIONAL BLOCK DIAGRAM OF  ACCUMULATOR ISOLATION VALVE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-4 
FNP-FSAR-7 REV 21  5/08 TABLE 7.7-1 (SHEET 1 OF 2)  PLANT CONTROL SYSTEM INTERLOCKS  Designation Derivation Function      C-1 1/2 neutron flux Blocks automatic and  (intermediate range) manual control rod  above setpoint withdrawal    C-2 1/4 neutron flux Blocks automatic and  (power range) above manual control rod  setpoint withdrawal    C-3 2/3 overtemperature Blocks automatic and  T above setpoint manual control rod  withdrawal      C-4 2/3 overpower T above Blocks automatic and  setpoint manual control rod  withdrawal      C-5 1/1 turbine impulse Blocks automatic control  chamber pressure below rod withdrawal  setpoint    C-7 1/1 time derivative Makes steam dump valves  (absolute value) of available for either  turbine impulse chamber tripping or modulation  pressure (decrease only)  above setpoint   
FNP-FSAR-7 REV 21  5/08 TABLE 7.7-1 (SHEET 2 OF 2)
Designation Derivation Function    C-9 Any condenser pressure Blocks steam dump to  above setpoint condenser    or All circulation water  pump breakers open    C-11 1/1 bank D control rod Blocks automatic rod  position above setpoint withdrawal    C-20 Two-of-two turbine impulse Arms AMSAC; below  chamber pressure above setpoint, blocks AMSAC  setpoint (Generated in AMSAC; see  section 7.8.)  Control  grade only. P-4(a) Reactor Trip Closes main feedwater    valves on low Tavg  below setpoint      Blocks steam dump control  via load rejection Tavg  controller      Makes steam dump valves    available for either    tripping or modulation    Reactor not tripped Block steam dump control  via plant trip Tavg  controller 
                (a) See table 7.3-4 for safety functions.
FNP-FSAR-7 REV 21  5/08 TABLE 7.7-2  BORON CONCENTRATION MEASUREMENT SYSTEM SPECIFICATIONS  Operating Conditions Line voltage:  120 V-AC (+/-10 percent); 60 Hz (+/-1 percent)
Pressure:  15 to 225 psig (sample)
Temperature:  70&deg;F to 130&deg;F (sample)
Sample flowrate:  0 to 0.4 gal/min
Ambient temperature:  60&deg;F to 105&deg;F
Relative humidity:  to 95 percent
Radiation levels:  <2 mr/h at 24 in. from all tank surfaces Reading time:  Variable depending on boron concentration; maximum time for 5000 ppm is approximately 5 min Accuracy Accuracy Boron ppm of Water Standard Deviation 0 - 1800 ppm +/-10 ppm 1800 - 5000 ppm +/-1.25 percent Drift:  Less than 10 ppm/week REV 21  5/08 SIMPLIFIED BLOCK DIAGRAM OF REACTOR CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-1 REV 21  5/08 CONTROL BANK ROD INSERTION MONITOR  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-2 
REV 21  5/08 ROD DEVIATION COMPARATOR  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-3 REV 21  5/08 BLOCK DIAGRAM OF PRESSURIZER  PRESSURE CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-4 
REV 21  5/08 BLOCK DIAGRAM OF PRESSURIZER LEVEL CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-5 
REV 21  5/08 BLOCK DIAGRAM OF MAIN FEEDWATER PUMP SPEED CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-6 REV 21  5/08 BLOCK DIAGRAM OF STEAM GENERATOR WATER LEVEL CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-7 
REV 21  5/08 BLOCK DIAGRAM OF STEAM DUMP CONTROL SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-8 REV 21  5/08 BASIC FLUX MAPPING SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-9 REV 21  5/08 SOURCE-DETECTOR ASSEMBLY  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-10 REV 21  5/08 MEASUREMENT UNIT  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-11 
REV 21  5/08 PROCESS SCHEMATIC FOR THE BORON CONCENTRATION MEASUREMENT SYSTEM  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-12 REV 21  5/08 BORON CONCENTRATION MEASUREMENT SYSTEM VS NORMAL PLANT OPERATING RANGE OF BORON CONCENTRATIONS  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-13 FNP-FSAR-7 7.8-1 REV 21  5/08  7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) 7.8.1 DESCRIPTION
7.8.1.1  System Description  The ATWS (anticipated transient without scram) mitigation system actuation circuitry (AMSAC) provides a backup to the reactor trip system (RTS) and ESF actuation system (ESFAS) for initiating turbine trip and auxiliary feedwater flow in the event of an anticipated transient (e.g., in the complete loss of main feedwater). The AMSAC is independent of and diverse from the RTS and ESFAS, with the exception of the final actuation devices. The AMSAC equipment, with the exception of the output isolation relays, is classified as control-grade equipment. It is a highly-reliable, microprocessor-based, single-train system powered by a non-Class 1E source.
The AMSAC continuously monitors level in the steam generators, which is an anticipatory indication of a loss of heat sink, and initiates certain functions when the level drops below a predetermined setpoint for at least a preselected time and for two of the three steam generator levels. These initiated functions are the tripping of the turbine, the initiation of auxiliary feedwater, and isolation of the steam generator blowdown and sample lines.
The AMSAC is designed to be highly reliable, resistant to inadvertent actuation, and easily maintained. Reliability is assured through the use of internal redundancy and continual self testing by the system. Inadvertent actuations are minimized through the use of internal redundancy and majority voting at the output stage of the system. The time delay on low steam generator level and the coincidence logic used also minimize inadvertent actuations.
The AMSAC automatically performs its actuations when above a preselected power level (determined using turbine impulse chamber pressure) and remains armed sufficiently long after that pressure drops below the setpoint to ensure that its function will be performed in the event of a turbine trip.
7.8.1.2  Equipment Description  The AMSAC consists of a single train of equipment located primarily in a seismically qualified cabinet. The output isolation relays, however, are located in two separate qualified wall-mounted cabinets.
The design of the AMSAC is based on the industry standard Intel multibus format, which permits the use of various readily available, widely used microprocessor cards on a common data bus for various functions.
The AMSAC consists of the following:
FNP-FSAR-7 7.8-2 REV 21  5/08  A. Steam Generator (SG) Level Sensing AMSAC utilizes the SG level signals as measured with three differential pressure-type level transmitters, measuring the level for each of the main steam generators as shown on drawing U-166237.
B. Turbine Impulse Pressure AMSAC also utilizes the turbine impulse pressure signal for measuring pressure in the turbine, as shown on drawing U-166245.
C. System Hardware The system hardware consists of two primary systems: the actuation logic system (ALS) and the test/ maintenance system (T/MS).
1. Actuation Logic System
The ALS monitors the analog and digital inputs, performs the functional logic required, provides actuation outputs to trip the turbine and initiate auxiliary feedwater flow, and provides status information to the T/MS. The ALS consists of three groups of input/output (I/O) modules, three actuation logic processors (ALPs), two majority voting modules, and two output relay panels. The I/O modules provide signal conditioning, isolation, and test features for interfacing the ALS and the T/MS.
Conditioned signals are sent to three identical ALPs for analog-to-digital conversion, setpoint comparison, and coincidence logic performance.
Each of the ALPs perform identical logic calculations using the same inputs, and derive component actuation demands, which are then sent to the majority voting modules. The majority voting modules perform a two-out-of-three vote on the ALP demand signals. These modules drive the relays providing outputs to the existing turbine trip and auxiliary feedwater initiation circuits.
A simplified block diagram of the AMSAC ALS architecture is presented on figure 7.8-1.
2. Test/Maintenance System
The test/maintenance system provides the AMSAC with automated and manual testing as well as a maintenance mode. Automated testing is the continuously performed self checking done by the system during normal operation. ALS status is monitored by the T/MS and sent to the plant computer and the main control board. Manual testing of the system by the computer services staff can be performed on-line to provide assurance that the ALS system is fully operational. The maintenance mode permits the computer services staff, under administrative control, to modify channel setpoints, channel status, and timer values and to initiate channel calibration.
FNP-FSAR-7 7.8-3 REV 21  5/08    The T/MS consists of a test/maintenance processor, a digital-to-analog conversion board, a memory board, expansion boards, a self-health board, digital output modules, a test/maintenance panel, and a portable terminal/printer.
D. Equipment Actuation
The output relay panels provide component actuation signals through isolation relays, which then drive the final actuation circuitry as shown on drawings U-166244 and U-166245 for initiation of auxiliary feedwater and for turbine trip.
7.8.1.3  Functional Performance Requirements  Analyses have shown that the most limiting ATWS event is a loss of feedwater event without a reactor trip. AMSAC performs the mitigative actuations of automatically initiating auxiliary feedwater, tripping the turbine, and isolating steam generator blowdown and sampling lines.
These are initiated in order to ensure a secondary heat sink following an anticipated transient (ANS Condition II) without a reactor trip, in order to limit core damage following an anticipated transient without a reactor trip and to ensure that the energy generated in the core is compatible with the design limits to protect the reactor coolant pressure boundary by maintaining the reactor coolant pressure to within ASME stress level C.
7.8.1.4  AMSAC Interlocks  A single interlock, designated as C-20, is provided to allow for the automatic arming and blocking of the AMSAC (drawing U-166245). The system is blocked at sufficiently low reactor power levels when the actions taken by the AMSAC following an ATWS need not be automatically initiated. Turbine impulse chamber pressure in a two-out-of-two logic scheme is used for the blocking function. Turbine impulse chamber pressure above the setpoint will automatically defeat any block, i.e., will arm the AMSAC. Dropping below this setpoint will automatically block the AMSAC. Removal of the C-20 permissive is automatically delayed for a predetermined time. The operating status of the AMSAC is displayed on the main control board.
7.8.1.5  Trip System  The SG level and turbine impulse chamber pressure inputs are used by AMSAC to determine trip demand. Signal conditioning is performed on the transmitter output and used by each of the ALPs to derive a component actuation demand. If two of the three steam generators have a low level at a power level greater than the C-20 permissive, a trip demand signal is generated following a time delay. This signal drives output relays for performing the necessary mitigative actions.
FNP-FSAR-7 7.8-4 REV 21  5/08  7.8.1.6  Isolation Devices AMSAC is independent of the RTS and ESFAS. The AMSAC inputs for measuring narrow range steam generator water level are derived from existing transmitters and channels within the process protection system. Connections to these channels are made downstream of Class-1E isolation devices located within the process protection cabinets. These isolation devices ensure that the existing protection system continues to meet all applicable safety criteria by providing isolation. Buffering of the AMSAC outputs from the safety-related final actuation device circuits is achieved through qualified relays. A credible fault occurring in the nonsafety-related AMSAC will not propagate through and degrade the RTS and ESFAS.
7.8.1.7  AMSAC Diversity from the Reactor Protection Systems  Equipment diverse from the RTS and ESFAS is used in the AMSAC to prevent common mode failures that might affect the AMSAC and the RTS or ESFAS. The AMSAC is a digital, microprocessor-based system with the exception of the analog SG level and turbine impulse pressure transmitter inputs, whereas the reactor trip system utilizes an analog based protection system. Also, where similar components are utilized for the same function in both AMSAC and the reactor trip system, the components used in AMSAC are provided by a different manufacturer.
Common mode failure of identical components in the analog portion of the RTS that results in the inability to generate a reactor trip signal will not impact the ability of the digital AMSAC to generate the necessary mitigative actuations. Similarly, a postulated common mode failure affecting analog components in ESFAS, affecting its ability to initiate auxiliary feedwater, will not impact the ability of the digital based AMSAC to automatically initiate auxiliary feedwater.
7.8.1.8  Power Supply  The AMSAC power supply is a dedicated uninterruptible power supply (UPS) which is independent from the RTS power supplies and is backed by batteries which are independent from the existing batteries which supply the RTS.
7.8.1.9  Environmental Variations  The AMSAC equipment is located in a controlled environment such that variations in the ambient conditions are minimized.
7.8.1.10 Setpoints  The AMSAC makes use of two setpoints in the coincidence logic in order to determine if mitigative functions are required. Water level in each steam generator is sensed to determine if a loss of secondary heat sink is imminent. The low-level setpoint is selected in such a manner that a true lowering of the level will be detected by the system. The normal small variations in steam generator level will not result in a spurious AMSAC signal.
FNP-FSAR-7 7.8-5 REV 21  5/08  The C-20 permissive setpoint is selected in order to be consistent with ATWS investigations showing that the mitigative actions performed by the AMSAC need not be automatically actuated below a certain power level. The maximum allowable value of the C-20 permissive setpoint is defined by these investigations.
To avoid inadvertent AMSAC actuation on the loss of one main feedwater pump, AMSAC actuation is delayed by a defined amount of time. This will ensure the reactor protection system will provide the first trip signal.
To ensure that the AMSAC remains armed sufficiently long to permit its function in the event of a turbine trip, the C-20 permissive is maintained for a preset time delay after the turbine impulse chamber pressure drops below the setpoint. The setpoints and the capability for their modification in the AMSAC are under administrative control.
7.8.2 ANALYSIS  7.8.2.1  Safety Classification/Safety-Related Interface  The AMSAC is not safety related, therefore, it need not meet the requirements of IEEE-279-1971. The AMSAC has been implemented such that the RTS and ESFAS continue to meet all applicable safety-related criteria. The AMSAC is independent of the RTS and ESFAS. The isolation provided between the RTS and the AMSAC and between the ESFAS and the AMSAC by the isolator modules and the isolation relays, respectively, ensures that applicable safety- related criteria are met for the RTS and the ESFAS.
7.8.2.2  Redundancy  System redundancy has not been provided. Since AMSAC is a backup nonsafety-related system to the redundant RTS, redundancy is not required. To ensure high system reliability, portions of the AMSAC have been implemented as internally redundant, such that a single failure of an input channel or ALP will neither actuate nor prevent actuation of the AMSAC.
7.8.2.3  Diversity from the Existing Trip System  Diverse equipment has been selected in order that common cause failures affecting both the RTS and the AMSAC or both the ESFAS and the AMSAC will not render these systems inoperable simultaneously. A more detailed discussion of the diversity between the RTS and the AMSAC and between the ESFAS and the AMSAC is presented in paragraph 7.8.1.7.
7.8.2.4  Electrical Independence  The AMSAC is electrically independent of the RTS and ESFAS with the exception of the final actuation devices. Qualified isolation devices are provided to isolate the nonsafety AMSAC FNP-FSAR-7 7.8-6 REV 21  5/08  circuitry from the safety-related actuation circuits of the auxiliary feedwater system as discussed in paragraph 7.8.1.6.
7.8.2.5  Physical Separation from the RTS and ESFAS  AMSAC is, by necessity, physically separated from the existing protection system hardware.
The two trains of AMSAC outputs are provided from separate wall-mounted enclosures outside of the cabinet.
7.8.2.6  Environmental Qualification  Equipment related to the AMSAC is designed to operate under conditions resulting from anticipated operational occurrences for the respective equipment location. The AMSAC equipment, with the exception of the isolation devices, is not designated as safety- related equipment and, therefore, is not required to be qualified as safety related per the requirements of IEEE Standard 279-1971, "IEEE Standard for Criteria for Protection Systems for Nuclear Power Generating Stations."  The safety-related AMSAC output isolation devices are located in a mild environment.
7.8.2.7  Seismic Qualification  It is required that only the isolation devices comply with seismic qualification. The AMSAC output isolation device is qualified in accordance with a program that was developed to implement the requirements of IEEE Standard 344-1975, "IEEE Standard for Seismic Qualification of Class 1E Electrical Equipment for Nuclear Power Generating Stations."
7.8.2.8  Test, Maintenance, and Surveillance Quality Assurance  NRC Generic Letter 85-06, "Quality Assurance Guidance for ATWS Equipment that is not Safety Related," requires quality assurance procedures commensurate with the nonsafety-related classification of the AMSAC. The quality controls for the AMSAC are, at a minimum, consistent with existing plant procedures or practices for nonsafety-related equipment.
Design of the AMSAC followed procedures relating to equipment procurement, document control, and specification of system components, materials and services. In addition, specifications also define quality assurance practices for inspections, examinations, storage, shipping, and tests as appropriate to a specific item or service.
A computer software verification program and a firmware validation program have been implemented commensurate with the nonsafety-related classification of the AMSAC to ensure that the system design requirements implemented with the use of software have been properly implemented and to ensure compliance with the system functional, performance, and interface requirements.
FNP-FSAR-7 7.8-7 REV 21  5/08  System testing is completed prior to the installation and operation of the AMSAC as part of normal factory acceptance testing and the validation program. Periodic testing is performed automatically through use of the system automatic self-checking capability and manually under administrative control via the AMSAC test/maintenance panel.
7.8.2.9  Power Supply  Power to the AMSAC is from a battery-backed, dedicated uninterruptible power supply independent of the power supplies for the RTS and ESFAS. The station battery supplying power to the AMSAC is independent of those used for the RTS and ESFAS. The AMSAC is an energize-to-actuate system capable of performing its mitigative functions with a loss-of-offsite power.
7.8.2.10 Testability at Power  The AMSAC is testable at power. This testing is done via the system test/maintenance panel. The capability of the AMSAC to perform its mitigative actuations is bypassed at a system level while in the test mode. Total system testing is performed as a set of three sequential, partial, overlapping tests. The first of the tests checks the analog input portions of the AMSAC in order to verify accuracy. Each of the analog input modules is checked separately. The second test checks each of the ALPs separately to verify that the appropriate coincidence logic is sent to the majority voter. The last test exercises the majority voter and the integrity of the associated output relays. The majority voter and associated output relays are tested by exercising all possible input combinations to the majority voter. The integrity of each of the output relays is checked by confirming continuity of the relay coils without operating the relays. The capability to individually operate the output relays, confirm integrity of the associated field wiring, and operate the corresponding isolation relays and final actuation devices at plant shutdown is provided.
7.8.2.11 Inadvertent Actuation  The AMSAC has been designed such that the frequency of inadvertent actuations is minimized.
This high reliability is ensured through use of three redundant ALPs and a majority voting module. A single failure in any of these modules will not result in a spurious AMSAC actuation.
In addition, a two-out-of-three low-steam generator level coincidence logic and a time delay have been selected to further minimize the potential for inadvertent actuations.
7.8.2.12 Bypass  7.8.2.12.1 Maintenance Bypasses  The AMSAC is blocked at the system level during maintenance, repair, calibration, or test.
While the system is blocked, the bypass condition is indicated in the main control room.
FNP-FSAR-7 7.8-8 REV 21  5/08  7.8.2.12.2 Operating Bypasses  The AMSAC has been designed to allow for operational bypasses with the inclusion of the C-20 permissive. Above the C-20 setpoint, the AMSAC is automatically unblocked (i.e., armed); below the setpoint, the system is automatically blocked. The operating status of the AMSAC is indicated in the main control room via a bypass and permissive panel window.
7.8.2.12.3 Indication of Bypasses  Whenever the mitigative capabilities of the AMSAC are bypassed or deliberately rendered inoperable, this condition is indicated in the main control room. In addition to the operating bypass, any manual maintenance bypass is indicated via the AMSAC general warning sent to the main control room.
7.8.2.12.4 Means for Bypassing  A permanently installed system bypass selector switch is provided to bypass the system. This is a two-position selector switch with "NORMAL" and "BYPASS" positions. At no time is it necessary to use any temporary means, such as installing jumpers or pulling fuses, to bypass the system.
7.8.2.13 Completion of Mitigative Actions Once Initiated  The AMSAC mitigative actions go to completion as long as the coincidence logic is satisfied and the time delay requirements are met. If the flow in the feedwater lines is reinitiated before the timer expires and the SG water level increases to above the AMSAC low setpoint, the coincidence logic will no longer be satisfied and the actuation signal disappears. If the coincidence logic conditions are maintained for the duration of the time delay, the mitigative actions go to completion. The auxiliary feedwater initiation and the turbine trip signals are latched in at the activated component level through the existing circuits. Deliberate operator action is then necessary to terminate auxiliary feedwater flow, clear the turbine trip signal using the main control board turbine trip reset switch, and proceed with the reopening of the turbine stop valves.
7.8.2.14 Manual Initiation  Manual initiation of the AMSAC is not provided. The capability to initiate the AMSAC mitigative functions manually, i.e., initiate auxiliary feedwater, trip the turbine, and isolate steam generator blowdown and sampling lines, exists at the main control board independent of AMSAC.
7.8.2.15 Information Readout  The AMSAC has been designed such that the operating and maintenance staffs have accurate, complete, and timely information pertinent to the status of the AMSAC. A system level general FNP-FSAR-7 7.8-9 REV 21  5/08  warning alarm is indicated in the control room. Diagnostic capability exists from the test/maintenance panel to determine the cause of any unanticipated inoperability or deviation.
7.8.2.16 Compliance with Standards and Design Criteria  The AMSAC meets the NRC acceptance criteria contained in 10 CFR 50.62 and the quality assurance requirements contained in NRC Generic Letter 85-06. The AMSAC also complies with the generic designs presented in WCAP-10858-P-A, which have been determined to be acceptable by the NRC for meeting the requirements of 10 CFR 50.62. In addition, the time delay design for the AMSAC associated with the C-20 permissive signal is consistent with Revision 1 to WCAP-10858-P-A, which has been accepted by the NRC.
REV 21  5/08 ACTUATION LOGIC SYSTEM ARCHITECTURE  JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.8-1 
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Revision as of 10:56, 28 April 2018

Joseph M. Farley Nuclear Plant, Updated Final Safety Analysis Report, Revision 27, Chapter 6 Through Chapter 7
ML17117A370
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Issue date: 04/20/2017
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FNP-FSAR-6 6-i REV 21 5/08 6.0 ENGINEERED SAFETY FEATURES TABLE OF CONTENTS Page 6.1 GENERAL .............................................................................................................6.1-1 6.1.1 Safety Features Systems...................................................................................6.1-1 6.1.2 Operational Reliability........................................................................................6.1-2

6.2 CONTAINMENT SYSTEMS...........................................................................................6.2-1

6.2.1 Containment Functional Design.........................................................................6.2-1

6.2.1.1 Design Bases.................................................................................6.2-1 6.2.1.2 System Design...............................................................................6.2-4 6.2.1.3 Design Evaluation..........................................................................6.2-5 6.2.1.4 Containment Testing and Inspection...........................................6.2-32 6.2.1.5 Instrumentation Requirements.....................................................6.2-36 6.2.1.6 Materials......................................................................................6.2-36 6.2.1.7 Heavy Load Safe Load Paths......................................................6.2-37

6.2.2 Containment Heat Removal Systems..............................................................6.2-37

6.2.2.1 Design Bases...............................................................................6.2-37 6.2.2.2 System Design.............................................................................6.2-38 6.2.2.3 Design Evaluation........................................................................6.2-42 6.2.2.4 Testing and Inspection.................................................................6.2-44 6.2.2.5 Instrumentation Requirements.....................................................6.2-46 6.2.2.6 Materials......................................................................................6.2-48

6.2.3 Containment Air Purification and Cleanup Systems.....................................................................................6.2-48

6.2.3.1 Design Bases...............................................................................6.2-49 6.2.3.2 System Design.............................................................................6.2-55 6.2.3.3 Design Evaluation........................................................................6.2-63 6.2.3.4 Tests and Inspections..................................................................6.2-66 6.2.3.5 Instrumentation Requirements.....................................................6.2-70 6.2.3.6 Materials......................................................................................6.2-72

6.2.4 Containment Isolation System.........................................................................6.2-72

6.2.4.1 Design Bases...............................................................................6.2-72 6.2.4.2 System Design.............................................................................6.2-72 6.2.4.3 Design Evaluation........................................................................6.2-74 6.2.4.4 Tests and Inspections..................................................................6.2-74 6.2.4.5 Materials......................................................................................6.2-75 6.2.5 Combustible Gas Control in Containment...............................................6.2-75 6.2.5.1 Design Bases..........................................................................................6.2-75 6.2.5.2 System Design........................................................................................6.2-78

FNP-FSAR-6 TABLE OF CONTENTS Page 6-ii REV 21 5/08 6.2.5.3 Design Evaluation....................................................................................6.2-82 6.2.5.4 Tests and Inspections..............................................................................6.2-84 6.2.5.5 Instrumentation Requirements................................................................6.2-85 6.2.5.6 Materials..................................................................................................6.2-85

6.3 EMERGENCY CORE COOLING SYSTEM................................................................6.3-1 6.3.1 Design Bases............................................................................................6.3-1

6.3.1.1 Range of Coolant Ruptures and Leaks.....................................................6.3-1 6.3.1.2 Fission Product Decay Heat . . . . .............................................................6.3-2 6.3.1.3 Reactivity Required for Cold Shutdown.....................................................6.3-2 6.3.1.4 Capability to Meet Functional Requirements.............................................6.3-2

6.3.2 System Design..........................................................................................6.3-3

6.3.2.1 Schematic Piping and Instrumentation Diagrams......................................6.3-3 6.3.2.2 System Components.................................................................................6.3-3 6.3.2.3 Applicable Codes and Classifications......................................................6.3-11 6.3.2.4 Materials Specifications and Compatibility..............................................6.3-11 6.3.2.5 Design Pressures and Temperatures. . ..................................................6.3-12 6.3.2.6 Coolant Quantity .....................................................................................6.3-12 6.3.2.7 Pump Characteristics .............................................................................6.3-13 6.3.2.8 Heat Exchanger Characteristics..............................................................6.3-13 6.3.2.9 ECCS Flow Diagrams..............................................................................6.3-13 6.3.2.10 Relief Valves............................................................................................6.3-13 6.3.2.11 System Reliability ...................................................................................6.3-13 6.3.2.12 Protection Provisions...............................................................................6.3-15 6.3.2.13 Provisions for Performance Testing .......................................................6.3-16 6.3.2.14 Net Positive Suction Head (NPSH).........................................................6.3-16 6.3.2.15 Control of Motor-Operated Isolation Valves............................................6.3-17 6.3.2.16 Motor-Operated Valves and Controls......................................................6.3-18 6.3.2.17 Manual Actions........................................................................................6.3-18 6.3.2.18 Process Instrumentation..........................................................................6.3-18 6.3.2.19 Materials..................................................................................................6.3-18

6.3.3 Performance Evaluation..........................................................................6.3-19

6.3.3.1 Evaluation Model.....................................................................................6.3-19 6.3.3.2 ECCS Performance.................................................................................6.3-19

6.3.3.3 Alternate Analysis Methods.....................................................................6.3-19 6.3.3.4 Fuel Rod Perforations..............................................................................6.3-20

FNP-FSAR-6 TABLE OF CONTENTS Page 6-iii REV 21 5/08 6.3.3.5 Evaluation Model.....................................................................................6.3-20 6.3.3.6 Fuel Clad Effects.....................................................................................6.3-20 6.3.3.7 ECCS Performance.................................................................................6.3-20 6.3.3.8 Peaking Factors.......................................................................................6.3-20 6.3.3.9 Fuel Rod Perforations..............................................................................6.3-20 6.3.3.10 Conformance with Interim Acceptance Criteria.......................................6.3-20 6.3.3.11 Effects of ECCS Operation on the Core..................................................6.3-20 6.3.3.12 Use of Dual Function Components..........................................................6.3-20 6.3.3.13 Dependence on Other Systems..............................................................6.3-21 6.3.3.14 Lag Times................................................................................................6.3-22 6.3.3.15 Thermal Shock Considerations...............................................................6.3-23 6.3.3.16 Limits on System Parameters..................................................................6.3-23

6.3.4 Tests and Inspections..............................................................................6.3-23 6.3.5 Instrumentation Requirements................................................................6.3-26

6.4 HABITABILITY SYSTEMS..........................................................................................6.4-1

6.4.1 Habitability Systems Functional Design....................................................6.4-1

6.4.1.1 Design Bases............................................................................................6.4-1 6.4.1.2 System Design..........................................................................................6.4-2 6.4.1.3 Design Evaluations....................................................................................6.4-6 6.4.1.4 Testing and Inspection..............................................................................6.4-6 6.4.1.5 Instrumentation Requirement....................................................................6.4-7

6.5 AUXILIARY FEEDWATER SYSTEM..........................................................................6.5-1

6.5.1 Design Bases............................................................................................6.5-1 6.5.2 System Description....................................................................................6.5-2

6.5.2.1 General Description...................................................................................6.5-2 6.5.2.2 Component Description.............................................................................6.5-3 6.5.2.3 System Operation......................................................................................6.5-5

6.5.3 Design Evaluation......................................................................................6.5-7 6.5.4 Tests and Inspection.................................................................................6.5-7 6.5.5 Instrumentation..........................................................................................6.5-8

APPENDIX 6A MATERIALS COMPATIBILITY REVIEW...................................................6A-1

APPENDIX 6B CONTAINMENT PRESSURE ANALYSIS.................................................6B-1

FNP-FSAR-6 TABLE OF CONTENTS Page 6-iv REV 21 5/08 APPENDIX 6C CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM. (Historical - Prior to December 2007).......................................................6C-1 APPENDIX 6D CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION SUMP STRAINER DESIGN........6D-1 FNP-FSAR-6 LIST OF TABLES 6-v REV 21 5/08 6.2-1 Principal Containment Design Parameters 6.2-2 Heat Sink Geometric Data

6.2-3 Initial Conditions for Pressure Analysis

6.2-4 Heat Sink Thermodynamic Data

6.2-5 Engineered Safety Features Performance for Containment Pressure Transient Analysis 6.2-6 Containment Pressure Analysis Results for the Spectrum of RCS Break Sizes

6.2-7 System Parameters, Initial Conditions for Thermal Uprate

6.2-8 Safety Injection Flow - Minimum Safeguards

6.2-9 Safety Injection Flow - Maximum Safeguards

6.2-10 Double-Ended Hot Leg Break, Blowdown Mass and Energy Releases

6.2-11 Plant Data for Blowdown

6.2-12 Double-Ended Hot Leg Break, Mass Balance

6.2-13 Double-Ended Hot Leg Break, Energy Balance

6.2-14 Double-Ended Pump Suction Break, Blowdown Mass and Energy Releases

6.2-15 Reactor Cavity Release

6.2-16 Spray Line Break Release

6.2-17 Surge Line Break Release

6.2-18 Reactor Cavity Subcompartment Pressure Analysis Summary of Flow Paths and Vent Loss Coefficients 6.2-19 Containment Results for the Design Basis LOCA

6.2-20 Double-Ended Pump Suction Break - Minimum Safeguards, Reflood Mass and Energy Releases 6.2-21 LOCA Chronology of Events

6.2-22 Subcompartment Differential Pressure Results

6.2-23 Environmental Conditions for Containment Heat Removal Systems (Deleted) 6.2-24 Component Design Parameters for Containment Spray System and Containment Cooling System FNP-FSAR-6 LIST OF TABLES 6-vi REV 21 5/08 6.2-25 Regulatory Guide 1.52, Section Applicability for the Penetration Room Filtration System 6.2-26 Single Failure Analysis - Containment Spray System 6.2-27 Double-Ended Pump Suction Break - Minimum Safeguards, Blowdown Mass and Energy Releases 6.2-28 Containment Ventilation Systems Component Design Parameters 6.2-29 Spray Evaluation Parameters

6.2-30 Single Failure Analysis - Penetration Room Filtration System

6.2-31 Containment Isolation Valve Information

6.2-32 Steam Generator Isolation Valve Information

6.2-33 Electric Hydrogen Recombiner Typical Parameters

6.2-34 Postaccident Venting System Design Parameters

6.2-35 Postaccident Sampling System Design Parameters

6.2-36 Postaccident Mixing System Design Parameters

6.2-37 Containment Interior Coatings Summary

6.2-38 Containment Penetrations

6.2-39 Containment Isolation Valves

6.2-40 Steam Generator Isolation Valves

6.2-41 Containment Pressure/Temperature for 600 gal/min Service Water Flow, 0.003 Fouling Factor 6.2-42 Double-Ended Pump Suction Break - Minimum Safeguards, Principle Parameters During Reflood 6.2-43 Double-Ended Pump Suction Break - Minimum Safeguards, Post-Reflood Mass and Energy Releases 6.2-44 Double-Ended Pump Suction Break, Mass Balance, Minimum Safeguards

6.2-45 Double Ended Pump Suction Break, Energy Balance, Minimum Safeguards

6.2-46 Double-Ended Pump Suction Break - Maximum Safeguards, Reflood Mass and Energy Releases 6.2-47 Double-Ended Pump Suction Break - Maximum Safeguards, Principle Parameters During Reflood

FNP-FSAR-6 LIST OF TABLES 6-vii REV 21 5/08 6.2-48 Double-Ended Pump Suction Break - Maximum Safeguards, Post-Reflood Mass and Energy Releases 6.2-49 Double-Ended Pump Suction Break, Mass Balance, Maximum Safeguards

6.2-50 Double-Ended Pump Suction Break, Energy Balance, Maximum Safeguards

6.2-51 Double-Ended Hot Leg Break, Sequence of Events

6.2-52 Double-Ended Pump Suction Break - Minimum Safeguards, Sequence of Events

6.2-53 Double-Ended Pump Suction Break - Maximum Safeguards, Sequence of Events

6.2-54 LOCA Mass and Energy Release Analysis, Core Decay Heat Fraction

6.3-1 Emergency Core Cooling System Component Parameters

6.3-2 ECCS Relief Valve Data

6.3-3 Sequence of Changeover Operation from Injection to Recirculation (Deleted)

6.3-4 Time Analysis for ECCS Injection/Recirculation Switchover

6.3-5 Materials Employed for Emergency Core Cooling System Components 6.3-6 Normal Operating Status of Emergency Core Cooling

6.3-7 Single Active Failure Analysis for Emergency Core Cooling System Components

6.3-8 Maximum Potential Recirculation Loop Leakage External to Containment

6.3-9 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis

6.3-10 Emergency Core Cooling System Shared Functions Evaluation

6.5-1 Auxiliary Feedwater System Auxiliary Feedwater Pump Data 6.5-2 Failure Analysis of Auxiliary Feedwater System

6.5-3 Auxiliary Feedwater System Motor Operated Valve Data

FNP-FSAR-6 LIST OF FIGURES 6-viii REV 21 5/08 6.2-1 DEPSGB, Minimum ESF 1 AC Pressure vs. Time, PO = 0 PSIG 6.2-2 RSG DEPSG Minimum ESF 1 AC Pressure vs. Time, PO = 3 PSIG 6.2-3 DEHL, Minimum ESF, DBA Short Term Pressure vs. Time, PO = 0 PSIG 6.2-4 RSG DEHLG Minimum ESF, DBA Short Term Pressure vs. Time, PO = +3 PSIG 6.2-5 DECLG Maximum ESF Pressure vs. Time (Deleted)

6.2-6 RSG Pressure vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-6A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-7 RSG Temperature vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-7A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-8 Pressure vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-9 Temperature vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-10 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-10A Pressure vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-11 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-11A Temperature vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-12 Pressure vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-13 Temperature vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-14 Pressure vs. Time Steam Line Full D. E. Break 70% Power

6.2-15 Temperature vs. Time Steam Line Full D. E. Break 70% Power

6.2-16 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-17 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-18 Pressure vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-19 Temperature vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-20 RSG Pressure vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = +3 PSIG 6.2-21 RSG Temperature vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = -1.5 PSIG 6.2-22 RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, PO = -1.5 PSIG

FNP-FSAR-6 LIST OF FIGURES 6-ix REV 21 5/08 6.2-22A RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, PO = +3 PSIG 6.2-23 RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power, PO = -1.5 PSIG 6.2-23A RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power, PO = +3 PSIG 6.2-24 Pressure vs. Time Steam Line 0.5 ft2 D. E. Break 30% Power 6.2-25 Temperature vs. Time Steam Line 0.5 ft2 D. E. Break 30% Power 6.2-26 Pressure vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = 0 PSIG 6.2.26A Pressure vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = -1.5 PSIG 6.2-27 Temperature vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = 0 PSIG 6.2-27A Temperature vs. Time Steam Line 0.4 ft2 D. E. Break 30% Power, PO = -1.5 PSIG 6.2-28 RSG Pressure vs. Time Steam Line 0. 60 ft2 Split 30% Power, PO = -1.5 PSIG 6.2-28A RSG Pressure vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = +3 PSIG 6.2-29 RSG Temperature vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = -1.5 PSIG 6.2-29A RSG Temperature vs. Time Steam Line 0.60 ft2 Split 30% Power, PO = +3 PSIG 6.2-30 RSG Pressure vs. Time Steam Line Full D. E. Break Hot Standby, PO = +3 PSIG 6.2-31 RSG Temperature vs. Time Steam Line Full D. E. Break Hot Standby, PO = -1.5 PSIG 6.2-32 Pressure vs. Time Steam Line 0.2 ft2 D. E. Break Hot Standby 6.2-33 Temperature vs. Time Steam Line 0.2 ft2 D. E. Break Hot Standby 6.2-34 Pressure vs. Time Steam Line 0.1 ft2 D. E. Break Hot Standby 6.2-35 Temperature vs. Time Steam Line 0.1 ft2 D. E. Break Hot Standby 6.2-36 Pressure vs. Time Steam Line 0.30 ft2 Split Hot Standby 6.2-37 Temperature vs. Time Steam Line 0.30 ft2 Split Hot Standby 6.2-38 TS, Equipment Surface Temperature with Uchida Condensing Heat Transfer and Convective Heat Transfer Coefficient of 2 Btu/h/ft2 (Deleted) 6.2-39 DEPSGB Minimum ESF 1 AC P/T Analysis Long-Term Containment Pressure vs. Time (Deleted) 6.2-40 DEPSGB Min ESF DBA Temperature vs. Time, PO = 0 PSIG

FNP-FSAR-6 LIST OF FIGURES 6-x REV 21 5/08 6.2-41 RSG DEPSG Min ESF DBA Temperature vs. Time, PO = 3 PSIG 6.2-42 Containment Air Cooler Duty vs. Temperature

6.2-43 Thermal Heat Removal Efficiency of Containment Atmosphere Spray (Deleted)

6.2-44 Residual Heat Exchanger Design Duty Accident Mode

6.2-45 Mass and Energy Rate vs. Time for LOCA (Deleted)

6.2-46 LOCA Blowdown Mass and Energy Release Rates vs. Time (Deleted)

6.2-47 LOCA Post-Blowdown Mass and Energy Release Rates vs. Time (Deleted)

6.2-48 DEPSG Minimum ESF 1 AC P/T Analysis, Long Term Condensing Heat Transfer Coefficient (RSG) 6.2-49 Short Term Condensing Heat Transfer Coefficient for DBA (Deleted)

6.2-50 Reactor Cavity Model

6.2-51 Reactor Cavity Block Diagram

6.2-52 Total Horizontal Force vs. Time

6.2-53 Steam Generator Block Diagram

6.2-54 Steam Generator Compartment C Differential Pressure vs. Time

6.2-55 Pressurizer Compartment Pressure Model (Spray Line Break in Lower Compartment) 6.2-56 Pressurizer Compartment Flow Model

6.2-57 Pressurizer Compartment Spray Line Results

6.2-58 Node Pressures in Compartments 1 and 2 vs. Time

6.2-59 Node Pressures in Compartments 3, 4, 5, and 6 vs. Time

6.2-60 Node Pressures in Compartments 7, 8, 9, and 10 vs. Time

6.2-61 Node Pressures in Compartments 11, 12, 13, and 14 vs. Time

6.2-62 Node Pressures in Compartments 15, 16, and 17 vs. Time

6.2-63 Node Pressures in Compartments 18, 19, 20, and 21, vs. Time

6.2-64 Node Pressures in Compartments 22, 23, 24, 25, 26, and 27 vs. Time

6.2-65 Node Pressures in Compartments 28, 29, 30, 31, 32, 33, and 34 vs. Time

FNP-FSAR-6 LIST OF FIGURES 6-xi REV 21 5/08 6.2-66 Schematic of Reflood Code 19 Element Loop Model for a Pump Suction Break (Deleted) 6.2-67 Core Reflood Correlation (Deleted)

6.2-68 Comparison of Measured and Predicted Carryover Rate Fractions (Deleted)

6.2-69 Inlet Water Temperature vs. Time After End of Blowdown (Deleted)

6.2-70 Variation in Temperature Rise, Turnaround Time, and Quench Time with Respect to Core Elevation (Deleted) 6.2-71 Energy Balance Model (Deleted)

6.2-72 Reflood Rate and Carryover Fractions vs. Time After End of Blowdown (Deleted)

6.2-73 Flow Through Break vs. Time After End of Blowdown (Deleted)

6.2-74 Water Height vs. Time After End of Blowdown (Deleted)

6.2-75 Post-Reflood Loop Resistance Model (Deleted)

6.2-76 S/G Internal Energy vs. Time After Break (Deleted)

6.2-77 Energy Distribution vs. Time (Deleted)

6.2-78 RSG Temperature Profile Through Containment Wall, PO = +3 PSIG 6.2-79 RHR Heat Exchanger Duty vs. Time, RSG PO = +3 PSIG 6.2-80 Containment Air Cooler Duty vs. Time, RSG PO = +3 PSIG 6.2-81 Minimum Sump pH Following LOCA vs. Time (Deleted)

6.2-82 Minimum Partition Coefficient in the Sump vs. Solution pH (Deleted)

6.2-83 Hydrogen Generation Rate vs. Time in the Lower Compartment

6.2-84 Isolation Valve Arrangement through 6.2-89

6.2-90 Electric Hydrogen Recombiner

6.2-91 Electric Hydrogen Recombiner Schematic Diagram (Typical of One Recombiner)

6.2-92 Lower Compartment Plan

6.2-93 Section of Lower Reactor Compartment

6.2-94 Containment Hydrogen Concentration With One Electric Recombiner Started 1 Day after a LOCA 6.2-95 Hydrogen Concentration as a Function of Time in Containment Purge Mode FNP-FSAR-6 LIST OF FIGURES 6-xii REV 21 5/08 6.2-96 Volume Percent Hydrogen vs. Time in the Upper Containment (Unmixed), Outer Periphery (Unmixed), and Bulk Containment (Mixed) 6.2-97 Volume Percent Hydrogen vs. Time in the Lower Compartment

6.2-98 Hydrogen Generation Rate vs. Time in Outer Periphery and Overall Containment

6.3-1 Residual Heat Removal Pump Performance Curves

6.3-2 Charging Pump Performance Curves

6.3-3 RHR Pump Characteristic Curves

6.3-4 Containment Spray Pump Characteristic Curves

FNP-FSAR-6 6.1-1 REV 21 5/08 6.0 ENGINEERED SAFETY FEATURES 6.1 GENERAL Engineered safety features are structures and equipment required to mitigate design basis accidents including the loss of coolant accident and high energy pipe breaks such as a steam pipe break and a main feedwater pipe break. Engineered safety features are designed to Seismic Category I requirements. They are designed to perform their safety function with complete loss of offsite power. Such equipment is provided with sufficient redundancy that failure of a single component will not result in the loss of the safety function. Engineered safety features fulfill the following safety functions under accident conditions:

A. Protect the fuel cladding.

B. Ensure containment integrity.

C. Minimize containment leakage.

D. Remove fission products from the containment atmosphere.

The operator action times assumed in this chapter include conservative actions to provide an adequate safety margin for the purpose of nuclear safety system design and nuclear safety analysis of the design basis events. However, they are not intended to serve as a basis for actual operator action times in procedures or training. The assumed time periods are considered in the basis of plant design to permit credit for operator actions. The Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGs) provide a basis for operatior action in response to design basis accidents. 6.1.1 SAFETY FEATURES SYSTEMS The safety features systems provided to satisfy the functions listed above are as follows:

  • Containment isolation system (subsection 6.2.4).
  • Containment fan cooler system (subsection 6.2.2).
  • Containment air purification and cleanup system (subsection 6.2.3).
  • Combustible gas control in containment (subsection 6.2.5).

FNP-FSAR-6 6.1-2 REV 21 5/08

The fuel cladding is protected by the timely, continuous, and adequate supply of borated water to the reactor coolant system (RCS) and, ultimately, the reactor core. This supply of water is provided by the emergency core cooling system (ECCS). These systems provide high head (centrifugal charging pumps), low head (residual heat removal pumps) injection, and accumulator injection immediately following an incident, and low head/high head recirculation in the long term recovery period.

The containment integrity is ensured and the containment leakage is minimized by the provision of means for condensing the steam inside the containment, depressurizing the containment following an incident, and maintaining the containment at near atmospheric conditions for an extended period of time. The containment isolation system, spray system, fan cooler system, and the electric hydrogen recombiners provide the means for satisfying these requirements.

The fission products are removed from the containment atmosphere by the chemical spray additive which enhances the removal of radioactive iodine from the containment atmosphere following an incident. The containment air purification and cleanup systems are provided to meet this function.

The safety features systems are designed with sufficient redundancy to meet the general design criteria as discussed in sections 3.1, 3.2, and subsection 6.3.2.11. Electrical power for all safety features systems is provided both from offsite sources and from emergency onsite sources as described in sections 8.2 and 8.3, respectively.

Safety features are separated into two independent trains of equal capability. Either train can handle the entire emergency coolant injection and emergency cooling loads; either train can provide the entire containment isolation, containment cleanup, and containment leakage minimization functions. Each train has an independent onsite and offsite power source. Failure of either train cannot affect the other.

Some of high and low pressure emergency injection systems use equipment that serves normal functions during normal plant operation or shutdown. Observation of their normal functioning provides monitoring of equipment availability and condition. In cases where equipment is used for emergencies only, systems are designed to permit periodic inspection and tests.

6.1.2 OPERATIONAL RELIABILITY Operational reliability is achieved by using proven components and by conducting tests required by the quality control requirements presented in chapter 17.0. All safety features systems are quality items meeting the requirements of 10 CFR 50, Appendix B, and seismically designed as discussed in chapter 3.0. Those safety features essential for post-tornado safety are designed to survive without loss of function the design tornado described in section 3.3.

Other sections of this report contain additional information on the safety features systems.

Information on seismic requirements is provided in chapters 2.0 and 3.0. Information on the actuation instrumentation of the safety features system is provided in chapter 7.0.

FNP-FSAR-6 6.1-3 REV 21 5/08 Information on functions performed by components of the safety features systems during normal plant operation is provided in chapters 9.0 and 5.0. The safety analysis and demonstration of the ability of the safety features systems to provide adequate protection during accident conditions as provided in chapter 15.0.

The design bases, design description and evaluation, tests, inspections, and instrumentation for the safety features systems are presented in this chapter.

[HISTORICAL] [Tests on Liner During Construction Inspection procedures employed during construction for the liner seam welds, liner fastening, and around penetrations consist of visual inspection, vacuum box soap bubble testing, radiography, dye-penetrant testing, and magnetic particle inspection.

A. Visual Inspection of Welds All of the welding is visually examined by a technician responsible for welding quality control. The basis for visual quality of welds is as follows:

1. Each weld is uniform in width and size throughout its full length. Each layer of welding shall be smooth and free of slag, cracks, pinholes, and undercut and shall be completely fused to the adjacent weld beads and base metal. In addition, the cover pass is free of coarse ripples, irregular surface, nonuniform head pattern, high crown, and deep ridges or valleys between beads. Peening of welds is not permitted, except for light peening for cleaning purposes.

2. Butt welds are of multipass construction, slightly convex, of uniform height, and have full penetration.

3. Fillet welds are of the specified size, with full throat and legs of uniform length.

B. Soap Bubble Tests All of the welding required for containment integrity is vacuum box soap bubble tested except where the structural configuration or space limitation does not allow. In this test a vacuum box containing a window is placed over the area to be tested and is evacuated to produce at least 5 psi pressure differential. Before the vacuum box is placed over the test area, a soap solution is applied to the weld and any leaks will be indicated by bubbles observed through the window in the box.

C. Radiography Radiography is used as an aid to quality control. The primary purpose of the liner plate and the welds therein is to provide leaktightness integrity to the posttensioned concrete

containment. Structural integrity of the containment will be provided by the posttensioned concrete and not by the liner plate. Radiography is not recognized as a completely effective method for examining welds to assure leaktightness. Therefore, the maximum benefit expected from radiography in connection with obtaining leaktight welds will be as an aid to quality control. Random radiography of each welder's work will provide verification that the welding is under control and being done in accordance with the previously established and qualified procedures. Additionally, employing random radiography to inspect each welder's work has been proved by past experience to have a positive psychological effect on the improving overall welding workmanship.

For quality control purposes, at least one spot radiograph 12 inches long was taken in the first ten feet of welding completed in the flat, vertical, horizontal, and overhead positions by each welder on liner plate welds. No further welding was permitted until initial radiographic inspection has been satisfactorily completed and the welding found to be acceptable by the Inspector. Thereafter, a minimum of 2 percent of the welding was progressively spot examined as welding is performed, using film 12 in. long, on a random basis to be specified by the inspector, in such a manner that an approximately equal number of spot radiographs was taken from the work of each welder. In addition to the 2 percent radiograph, 18 percent of the welding was nondestructively examined. Under conditions where two or more welders make weld layers in a joint or on the two sides of a double-welded butt joint, one spot examination represented the work for both welders. Where a radiograph discloses welding which did not comply with the minimum quality requirements, as defined in paragraph UW52,Section VIII of ASME code, two additional spots, each 12 inches long, were examined in the same weld seam at locations away from the original spot. The locations of these additional spots were determined by the Inspector as provided for the original spot examination. If two additional spots examined showed welding which met the minimum quality requirements, the entire weld represented by the three radiographs was acceptable. The defective welding disclosed by the first of the three radiographs was removed and repaired by welding.

If either of the two additional spots examined showed welding which did not comply with minimum quality requirements, the entire portion of the seam represented was rejected; or, at the fabricator's option, the entire weld represented was completely radiographed, and defective welding corrected.

The rewelded joints or weld-repaired areas were completely reradiographed and met the weld quality requirements cited above.

D. Dye-Penetrant and Magnetic-Particle Inspection Dye-penetrant and magnetic-particle inspection were used as an aid to quality control. The field welding inspectors used dye-penetrant or magnetic-particle inspection to closely examine welds judged to be of questionable quality on the basis of the initial

visual inspection. Dye-penetrant or magnetic-particle inspection of liner plate welds were in accordance with Section VIII of the ASME Boiler and Pressure Vessel Code.]

[HISTORICAL][For Unit 1 there are three 3/8-in. diameter holes between the solid cover plate on the top of the sump screen and the bioshield wall for venting of air during the initial phase of the LOCA when the water level in the sump rises. The slot size varies from approximately 1/4 in. to 1 in. across its length of approximately 3 ft. The potential for debris to enter through this path has been evaluated. The location of the slot near the shield wall was specifically selected to minimize the potential for debris to enter the sump. Since this slot and the vent holes will be under water during the recirculation phase of a

LOCA, the debris entering through this path will sink to the sump floor due to low approach velocities near the bioshield wall and will not be swept into the opening of the intake pipe.]

[HISTORICAL] [Initial tests and the purpose of each test are listed as follows: A. Component qualification tests - These tests demonstrate the characteristics of materials to be incorporated by the manufacturer into components for a system and ensure that they meet the requirements of procurement specification. The design conditions, which form the basis for these component qualification tests, are presented in table 3.11-1.

B. Component acceptance tests - These tests are factory tests which demonstrate the capability of the components incorporated in the various systems in which they are to operate. For example, fans associated with safeguards systems are tested in the manufacturer's shop to determine their characteristic curves. System valves are tested in the shop to verify effectiveness of seal, opening and closing periods, and the ability of the valve operator to actuate the valve at the maximum anticipated differential pressure.

Test results on actual or similar types of filter assemblies demonstrate their adequacy for this application. The following demonstrative tests are performed:

A. Radioactive iodine removal efficiency - a charcoal sample 2 in. in diameter by 2 in. deep is exposed to air flow at 40 ft/min face velocity. The air stream contains concentrations of elemental iodine and methyl iodide, similar to those predicted to occur in the penetration room filters during faulted conditions. Air stream temperature is 150°F, relative humidity 70 percent, and test duration is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The efficiency is determined by measuring the activity of iodines upstream and downstream of the sample. Minimum acceptable efficiency is 99.0 percent at the end of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Flow resistance test - A module consisting of three absorbine units (six trays), stacked vertically, is capable of filtering 100 ft3/min of air at a pressure drop not exceeding 1.0 in. wg. The actual resistance is recorded and kept available.

C. Leak test - Each filter element is tested for 5 minutes in an air flow of 330 ft3/min containing approximately 20 ppm of Freon 112. Instrumentation is provided to measure the relative upstream and downstream concentrations of Freon 112. A downstream concentration in excess of 0.2 percent of the upstream concentration shall cause rejection of the filter.

D. Carbon lot tests - A sample from each lot of carbon after impregnation will have been subjected to the following tests by the manufacturer and results made a matter of permanent record:

1. Gas life - A bed of carbon 2 in. deep and 2 in. in diameter is tested for iodine collection at a velocity of 40 ft per minute (air at standard conditions). The iodine concentration upstream of the bed is 1000 mg/m3 and the penetration does not exceed 1.0 percent for a period of no less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. Wash test - 250 ml of demineralized water is brought to a minimum boil. Twenty five grams of impregnated carbon is added to the demineralized water and the minimum boil is maintained for 1 minute. After 1 minute the water is decanted from the carbon and analyzed for the impregnate. With a knowledge of initial impregnate loading in the carbon and quantity of impregnate removed by the boiling water, results are reported as percentages of impregnate retained.

3. Ignition temperature - A sample from each lot of carbon is tested for ignition temperature in accordance with the procedure described in USNRC Report DP-1075, "High Temperature Adsorbents for Iodine," by R.C. Milhans.

4. Carbon tetrachloride test - Samples of carbon are tested for carbon tetrachloride adsorption capacity. Testing follows the procedures described in paragraph 6.2 of Military Specification MIL-C-17605.

Systems acceptance tests - Deenergized and energized tests demonstrate the proper mounting of components, proper hookup of circuits and connection, setting of instrumentation and operation of interlocks. Equipment and system performance are monitored and rated.

For the penetration room filtration system, all ducting is given a pneumatic pressure test prior to the installation of the filter elements to assure leak-tight construction. Dimensional tolerances on filter assemblies and frame assemblies are checked to ensure that suitable gasket compression is uniformly achieved on the filter sealing faces.

A test program is performed after construction tests are completed to demonstrate the following:

A. Proper actuation of control circuitry in both modes.

B. Proper flow path alignment in both modes.

C. Leaktightness of each filter assembly.

D. Verification that a negative pressure is maintained in the spent fuel area with the penetration room filtration system operating in the fuel handling area.

The following tests are performed prior to installation of the filter elements and charcoal bed. A test assembly is installed to simulate filter pressure drop.

A. Simulate an actuation signal and observe the performance of the system in the LOCA mode.

B. In the LOCA mode, measure the discharge flow from the exhaust fan. At steady state conditions with the penetration room sealed, this corresponds to the penetration room leak rate.

C. In the LOCA mode, verify that the recirculation fan recirculation valve opens on receipt of a differential pressure signal from two out of three differential pressure instruments between the penetration rooms and pressure in the filtration system equipment room.

D. With the system operating, verify circulation of air within the penetration rooms in the LOCA mode. E. Install the roughing filter and high-efficiency filter. With the systems operating, test leaktightness and performance, using DOP smoke of 0.3 micron mass median diameter.

Penetration should not exceed 0.1 percent.

F. Install the charcoal beds. With the system operating, test the performance, using Freon 112. The test is performed using similar portable equipment described in USNRC Report ORNL-NSIC-65, 1970, by C. A. Burchsted and A. B. Fuller, entitled "Design, Construction, and Testing of High Efficiency Filtration Systems for Nuclear Application" (paragraph 7.5.1, pages 7.8 - 7.9). The testing procedure is in accordance with a paper by D. R. Muhlbaier, "Standardized Non-Destructive Test of Carbon Beds for Reactor Containment Applications," DP-1082, July 1967. Test results must demonstrate removal of 99.5 percent of the Freon 112. The pressure drop is also measured.

G. Simulate a spent fuel pool high radiation signal and observe system performance in the fuel handling mode.

H. Simulate a spent fuel pool low differential pressure and observe system performance in the fuel handling mode.

I. With the system operating in the fuel handling mode, verify that there is a vacuum in the spent fuel pool.

In addition, all instruments are calibrated, alarms, controls, and interlocks checked, and each remotely operated valve is individually stroked to determine its operability and correct performance of indicating lights.

The inleakage characteristics of the penetration boundary are determined by means of a flowmeter in the supply ducting to the penetration room filtration system filters and a vacuum gauge in the penetration room. With all normally operating ventilation systems in the auxiliary building secured, the internal pressure in the penetration rooms and the exhaust air flow provides the data necessary to ascertain the leaktightness of the joints, partitions, and seals.]

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-1 PRINCIPAL CONTAINMENT DESIGN PARAMETERS Characteristics Data Containment design pressure (psig) 54 Containment design temperature (°F) 280 Internal dimensions Cylindrical wall diameter (ft) 130 Cylindrical wall height (ft) 139 Curved dome height (ft) 43.5 Volumes

Gross internal volume (ft3) 2.35 x 106 Net free internal volume (ft3) 2.0 x 106 Containment design leak rate First 24 h, percent of containment free volume per day 0.15 After first day, percent per day 0.075

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-2 (SHEET 1 OF 5) HEAT SINK GEOMETRIC DATA(a) Heat Sink 1 - Containment Cylinder and Dome 74,908 ft2 Exposure

1. Containment Atmosphere
2. Outside Atmosphere

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.25 Air Gap 0.00204 Concrete 45.0

Heat Sink 2 - Penetration Plates & Liner Stiffners 3,802 ft2 Exposure

1. Containment Atmosphere
2. Outside Atmosphere

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.51 Air Gap 0.00204 Concrete 45.0

Heat Sink 3 - Unlined Concrete (excluding reactor support) 60,375 ft2 Exposure 1. Containment Atmosphere

2. Insulated

Material Thickness (in.)

Paint 0.019 Surfacer 0.125 Concrete 18.0

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-2 (SHEET 2 OF 5)

Heat Sink 4 - Galvanized Steel (excluding cable trays) 43,320 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Zinc 0.0034 Carbon Steel 0.07 Heat Sink 5 - Painted Carbon Steel 0.5-in. Thickness 95,210 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.18 Heat Sink 6 - Painted Carbon Steel 1.0-in. Thickness 25,681 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.59 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-2 (SHEET 3 OF 5)

Heat Sink 7 - Painted Carbon Steel 2.0-in. Thickness 8,802 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 1.35 Heat Sink 8 - Painted Carbon Steel 2.0-in. Thickness 3,353 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 3.59 Heat Sink 9 - Containment Floor 5,402 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Concrete 108.0

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-2 (SHEET 4 OF 5)

Heat Sink 10 - Refuel Canal Liner 7,894 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Stainless Steel 0.25 Air Gap 0.00204 Concrete 18.0

Heat Sink 11 - Unpainted Stainless Steel 10,116 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Stainless Steel 0.12

Heat Sink 12 - Galvanized Steel Cable Trays 22,164 ft2 Exposure

1. Containment Atmosphere
2. Insulated

Material Thickness (in.)

Zinc 0.0034 Carbon Steel 0.05

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-2 (SHEET 5 OF 5)

Heat Sink 13 - Reactor Support 2,182 ft2 Exposure 1. Containment Atmosphere - A 150°F source to account for the higher reactor cavity operating temperature 2. Insulated Material Thickness (in.)

Paint 0.019 Surfacer 0.125 Concrete 86.0

_________________

(a) An evaluation for these parameters was performed as described in Section 6.2.1.3.13.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-3 INITIAL CONDITIONS FOR PRESSURE ANALYSIS Characteristics Data Containment System Pressure (psia) 13.2 - 17.7 Relative humidity (percent) 50 Inside temperature (°F) 120(a) Outside temperature (°F) 95 Refueling water storage tank water temperature (°F) 110 Accumulator tank water temperature (°F) 120 Service water temperature (°F) 95(b) Stored Water Refueling water storage tank (gal) 471,000(c) Three accumulators (ft3) 240

a. 120°F is the Technical Specifications limit, 127°F was used in the analysis. b. Service water temperature of 97.3°F was used in the analysis. c. A refueling water storage tank delivery capacity of 390,000 gallons was used in the analysis.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-4 HEAT SINK THERMODYNAMIC DATA MATERIAL PROPERTIES(a) Thermal Heat Density Conductivity Capacity Material (lbm/ft3) Btu/h-ft-°F) (Btu/lbm-°F) Paint (Ameron 66) 162.3 0.50/0.25(b) 0.29 Paint (Ameron 90, 90HS) 160.8 0.38/0.25(b) 0.31 Primer (Dimetcote 6) 196.8 0.63 0.11 Carbon steel 489.0 29.6 0.1096 Concrete 144.0 1.0 0.2292 Surfacer (Ameron 121.2 0.39 0.23 110 AA, 3366/3367) Zinc 446.0 62.2 0.0942 Stainless steel 488.0 8.6 0.1232 Air 0.069 0.017 0.2095 HEAT TRANSFER COEFFICIENTS

Surface Value Sink surfaces exposed to containment Modified Tagami atmosphere (LOCA Blowdown) UCHIDA (LOCA Reflood & MSLB)

Sump liquid to containment atmosphere Conduction Containment sump and floor to sump Conduction liquid Sink surfaces exposed to outside 2.0 Btu/h-ft2-°F atmosphere

a. An evaluation was performed for these parameters as described in Section 6.2.1.3.13.

b. Value for Paint/Primer in combination.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-5 (SHEET 1 OF 2) ENGINEERED SAFETY FEATURES PERFORMANCE FOR CONTAINMENT PRESSURE TRANSIENT ANALYSIS Values Used for Containment Analysis Maximum Minimum System Operation ESF ESF Containment spray Water sources Borated water from RWST or sump Initiation Initiated by Containment Press. High-High-High Number of lines and 2 1 headers Number of pumps 2 1 Flowrate, gal/min 2175 2480 (Injection) per pump 2290 (Recirculation) Containment air coolers Initiation Initiated by SIS Number of units 4 1(b) Flowrate (air side), 40000 40000 ft3/min per unit Total design heat 80 x 106 80 x 106(a) removal at con- tainment design temperature, (Btu/h) per unit Service water 97.3 97.3 temperature (°F) RHR/Low pressure safety injection heat exchangers Type Horizontal shell U-tube Cooling water supply Component cooling water Number of units 2 1 Heat transfer area, 4070 3500 ft2 per unit Overall heat transfer 383 383 coefficient, Btu/h-ft2-°F FNP-FSAR-6 REV 21 5/08 TABLE 6.2-5 (SHEET 2 OF 2) Values Used for Containment Analysis Maximum Minimum System Operation ESF ESF Flowrate: Injection 3000 3000 Sump water side, Recirculation 3750 3750 gal/min per unit Component cooling 4755 4755 water side, gal/min per unit Return water point Primary Primary loop loop Passive safety injection system Capacity, gal each 600 accumulator Number of accumu- 3 3 lators Pressure setpoint, 600 600 psig Active safety injection system Initiation Initiated by SIS High pressure safety injection: Number of lines 3 3 Number of pumps 2 1 Flowrate, gal/min 511 511 per pump Low pressure safety injection: Number of lines 3 3 Number of pumps 2 1 Flowrate, gal/min Injection 3000 3000 per pump Recirculation 3750 3750

a. Value for 600-gal/min service water flow for paragraph 6.2.1.3.12 analysis is 31.2 x 106 at 275 °F.

b. Having fewer than 12 coils per containment cooler is acceptable provided that each cooler can adequately remove the containment analysis heat load described in note "a".

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-6 CONTAINMENT PRESSURE ANALYSIS RESULTS FOR THE SPECTRUM OF RCS BREAK SIZES(a) 0.6 DEPSG(b) DEPSG DEPSG(b) PSS(b) DECLG(b) DEHLG MIN ESF MIN ESF MAX ESF MAX ESF MAX ESF MIN ESF P0 = 0 psig P0 = +3 psig 4.95 ft2 3 ft2 8.25 ft2 P0 = +3 psig Peak pressure 38.0 43.8 40.1 40.9 37.6 43.6 (psig) Time of peak 19.4 552 191.9 194.3 22.3 18.8 pressure (s) Peak temperature 260 263 264 265 260 264 (°F) Time of peak 19.4 552 191.9 194.3 22.3 18.7 temperature (s) a. See Table 6.2-41 for MSLB results.

b. Non-limiting cases, not reanalyzed for power uprate/steam generator replacement, maintained for historical purposes FNP-FSAR-6 REV 21 5/08 TABLE 6.2-7 SYSTEM PARAMETERS INITIAL CONDITIONS FOR THERMAL UPRATE PARAMETERS VALUE Core Thermal Power (MWt) 2830.5 Reactor Coolant System Total Flowrate (lbm/sec) 27250.0 Vessel Outlet Temperature (°F) 619.3 Core Inlet Temperature (°F) 547.1 Vessel Average Temperature (°F) 583.2 Initial Steam Generator Steam Pressure (psia) 817 Steam Generator Design Model 54F Steam Generator Tube Plugging (%) 0 Initial Steam Generator Secondary Side Mass (lbm) 121826.1 Assumed Maximum Containment Backpressure (psia) 68.7 Accumulator Water Volume (ft3) per accumulator 1040 N2 Cover Gas Pressure (psia) 600 Temperature (°F) 120 Safety Injection Delay, total (sec) (from beginning of event) 30.9

Note: Core Thermal Power, RCS Total Flowrate, RCS Coolant Temperatures, and Steam Generator Secondary Side Mass include appropriate uncertainty and/or allowance.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-8 SAFETY INJECTION FLOW MINIMUM SAFEGUARDS RCS PRESSURE TOTAL FLOW (psig) (gpm) INJECTION MODE (REFLOOD PHASE) 0 4411.2 20 4163.4 40 3897.1 60 3603.8 80 3275.0 100 2900.8 120 2190.7 140 1619.5 160 482.7 180 480.0 COLD LEG RECIRCULATION MODE 0 3997.8

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-9 SAFETY INJECTION FLOW MAXIMUM SAFEGUARDS RCS PRESSURE TOTAL FLOW (psig) (gpm) INJECTION MODE (REFLOOD PHASE) 0 8575.0 20 8094.4 40 7581.5 60 7028.8 80 6425.3 100 5752.0 120 4976.6 140 4327.8 160 3530.3 180 2376.1 COLD LEG RECIRCULATION MODE 0 8575.0

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-10 (SHEET 1 OF 4) DOUBLE-ENDED HOT LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec)

.00000 .0 .0 .0 .0 .00113 46198.2 29516.1 46195.4 29512.8 .101 40189.0 26064.4 26821.5 17100.0

.201 34372.1 22298.5 23762.2 15065.7

.301 33846.7 21888.6 21200.8 13289.5

.401 32820.3 21206.0 19842.3 12249.4

.501 32023.9 20690.1 18991.0 11540.4 .601 31901.2 20605.1 18375.5 11003.2 .702 31874.4 20599.8 17849.2 10548.9

.801 31502.1 20403.4 17484.0 10215.0

.901 30897.3 20080.6 17160.3 9923.1 1.00 30486.2 19905.6 16866.9 9666.7 1.10 30168.9 19810.0 16638.3 9459.5 1.20 29888.3 19739.6 16448.8 9285.5 1.30 29539.0 19615.6 16339.9 9165.2 1.40 29120.5 19445.2 16294.4 9086.2 1.50 28623.3 19222.2 16299.4 9039.9 1.60 28060.0 18952.1 16335.9 9015.3 1.70 27484.3 18671.5 16392.6 9004.8 1.80 26923.5 18398.8 16463.9 9004.9 1.90 26365.0 18122.1 16540.5 9011.4 2.00 25754.5 17800.6 16619.8 9022.6 2.10 25092.6 17432.7 16698.9 9036.8 2.20 24448.5 17068.0 16775.4 9053.0 2.30 23848.9 16728.9 16847.1 9069.6 2.40 23275.6 16398.4 16908.6 9083.8 2.50 22699.0 16052.7 16957.3 9093.9 2.60 22152.3 15716.1 16990.9 9098.6 2.70 21633.8 15387.8 17010.2 9098.1 2.80 21137.9 15064.3 17014.2 9091.2 2.90 20673.7 14755.2 17004.2 9078.8 3.00 20244.4 14460.2 16980.5 9060.7 3.10 19840.5 14170.6 16943.2 9036.8 3.20 19487.3 13909.9 16893.0 9007.4 3.30 19173.3 13668.3 16831.6 8973.1 3.40 18885.6 13436.2 16758.0 8933.5 3.50 18645.8 13233.3 16673.0 8888.7 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-10 (SHEET 2 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 3.60 18435.9 13045.7 16577.9 8839.6 3.70 18249.1 12868.5 16472.4 8785.7 3.80 18099.7 12715.2 16358.8 8728.4 3.90 17969.9 12572.2 16235.2 8666.6 4.00 17855.3 12436.7 16100.7 8599.8 4.20 17697.1 12216.4 15802.5 8452.7 4.40 17639.5 12063.2 15468.3 8288.4 4.60 17745.0 12020.9 15116.7 8115.9 4.80 17952.2 12017.3 14771.0 7947.5 5.00 18303.0 12077.3 14352.5 7739.9 5.20 18790.2 12199.5 13896.4 7511.5 5.40 13503.2 9607.3 13445.3 7286.1 5.60 14722.7 10243.4 13012.2 7070.8 5.80 14884.5 10185.1 12576.0 6853.4 6.00 14975.8 10228.1 12111.8 6619.8 6.20 15007.9 10189.6 11650.1 6386.0 6.40 15046.6 10177.1 11197.6 6155.7 6.60 15138.4 10141.8 10763.7 5933.6 6.80 15178.5 10074.0 10344.3 5717.9 7.00 15231.5 9998.6 9941.0 5509.8 7.20 14935.3 9813.0 9565.7 5315.9 7.40 15073.7 9819.4 9219.0 5136.9 7.60 15146.3 9794.5 8891.8 4968.1 7.80 15183.5 9756.1 8590.5 4813.0 8.00 15159.8 9685.1 8304.6 4665.8 8.20 15111.2 9601.5 8034.1 4526.8 8.40 15027.5 9502.2 7780.3 4396.7 8.60 14901.2 9383.0 7532.9 4270.1 8.80 14729.3 9243.8 7296.7 4149.5 9.00 14509.7 9083.3 7068.2 4033.5 9.20 14245.5 8903.4 6846.7 3921.6 9.40 13945.5 8708.8 6632.5 3814.1 9.60 13620.4 8505.1 6424.8 3710.5 9.80 13277.4 8295.5 6222.2 3610.3 10.0 12927.4 8085.7 6026.5 3514.2 10.2 12570.6 7875.5 5834.9 3420.9 10.2 12567.5 7873.8 5833.4 3420.2 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-10 (SHEET 3 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 10.4 12210.5 7666.6 5649.8 3331.5 10.6 11856.3 7463.8 5471.9 3246.4 10.8 11500.2 7262.7 5298.1 3164.0 11.0 11151.5 7068.5 5130.8 3085.2 11.2 10805.2 6878.1 4967.8 3009.2 11.4 10465.5 6694.1 4811.0 2936.6 11.6 10130.0 6514.7 4658.5 2866.6 11.8 9791.2 6336.2 4509.6 2798.5 12.0 9437.7 6153.3 4362.5 2731.4 12.2 9059.5 5961.8 4210.8 2661.9 12.4 8654.2 5762.0 4048.8 2588.1 12.6 8228.7 5559.3 3873.8 2509.9 12.8 7800.9 5364.4 3685.7 2428.4 13.0 7369.9 5178.2 3479.8 2341.9 13.2 6952.2 5007.3 3269.6 2255.5 13.4 6533.3 4844.8 3054.5 2167.4 13.6 6122.1 4691.4 2851.2 2082.8 13.8 5711.4 4540.4 2663.8 2000.8 14.0 5304.8 4389.3 2502.3 1925.0 14.2 4906.4 4239.9 2368.1 1856.7 14.4 4513.5 4085.7 2259.7 1796.5 14.6 4070.0 3880.6 2173.8 1744.8 14.8 3665.2 3570.9 2101.9 1697.9 15.0 3386.8 3341.6 2039.7 1656.0 15.2 3100.1 3136.6 1982.4 1618.9 15.4 2792.5 2926.1 1923.6 1584.8 15.6 2470.0 2714.0 1860.2 1553.4 15.8 2150.8 2494.5 1787.6 1521.7 16.0 1957.2 2348.3 1703.4 1489.5 16.2 1810.1 2197.2 1605.9 1458.1 16.4 1699.4 2072.4 1498.2 1431.7 16.6 1574.1 1930.3 1380.6 1403.9 16.8 1458.7 1798.1 1265.4 1374.7 17.0 1362.8 1681.6 1169.2 1338.6 17.2 1258.7 1560.0 1102.7 1295.5 17.4 1160.9 1446.1 1041.3 1242.4 17.6 1081.9 1347.9 984.2 1183.1 17.8 1007.4 1260.6 929.9 1126.9 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-10 (SHEET 4 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 18.0 910.8 1141.2 836.0 1023.3 18.2 814.1 1024.0 707.5 870.1 18.4 715.7 900.0 641.0 791.0 18.6 629.5 793.5 529.7 653.6 18.8 547.0 690.0 403.6 499.9 19.0 458.5 579.1 267.2 332.2 19.2 372.8 471.7 159.1 198.9 19.4 282.1 357.9 97.1 122.4 19.6 207.7 264.6 84.2 107.3 19.8 89.1 114.1 .0 .0 20.0 .0 .0 .0 .0

  • mass and energy exiting from the reactor vessel side of the break.
    • mass and energy exiting from the SG side of the break.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-11 PLANT DATA FOR BLOWDOWN Reactor coolant loops 3 Minimum steam line internal diameter 14 inches Main feedwater isolation valve closing time 30 s Main feedwater control valve closing time 5 s Main steam line isolation valve closing time 10 s Maximum steam line volume between the steam 1180 ft3 generator and the nearest steam line stop valve Maximum steam line volume between the faulted 3475 ft3 steam generator stop valves and the steam line stop valves in the other steam generator loops Maximum unisolated feed line volume 202 ft3 Maximum auxiliary feedwater flow to a Varies with depressurized steam generator steam generator pressure Time to auxiliary feedwater isolation 1800 s Main feedwater flow Varies Containment pressure setpoint for main steam 19.2 psig line isolation signal Air cooler initiation pressure 7.0 psig Air cooler delay from start of accident 92 s

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-12 DOUBLE-ENDED HOT LEG BREAK MASS BALANCE Time (Seconds) .00 20.00 20.0 Mass (Thousand lbm) Initial In RCS & ACC 620.08 620.08 620.08 Added Mass Pumped Injection .00 .00 .00 Total Added .00 .00 .00 ***TOTAL AVAILABLE*** 620.08 620.08 620.08 Distribution Reactor Coolant 416.79 65.09 84.67 Accumulator 203.30 152.90 133.32 Total Contents 620.08 217.99 217.99 Effluent Break Flow .00 402.08 402.08 ECCS Spill .00 .00 .00 Total Effluent .00 402.08 402.08 ***TOTAL ACCOUNTABLE*** 620.08 620.07 620.07

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-13 DOUBLE-ENDED HOT LEG BREAK ENERGY BALANCE Time (Seconds) .00 20.00 20.0 Energy (Million BTU) Initial Energy In RCS, ACC, S. Gen 673.30 673.30 673.30 Added Energy Pumped Injection .00 .00 .00 Decay Heat .00 5.79 5.79 Heat From Secondary .00 -6.91 -6.91 Total Added .00 -1.12 -1.12 ***TOTAL AVAILABLE*** 673.30 672.19 672.19 Distribution Reactor Coolant 244.82 14.44 16.19 Accumulator 18.20 13.68 11.93 Core Stored 18.93 7.36 7.36 Primary Metal 118. 16 110.31 110.31 Secondary Metal 76.01 74.48 74.48 Steam Generator 197.20 192.31 192.31 Total Contents 673.30 412.59 412.59 Effluent Break Flow .00 259.11 259.11 ECCS Spill .00 .00 .00 Total Effluent .00 259. 11 259.11 ***TOTAL ACCOUNTABLE*** 673.30 671.70 671.70

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-14 (SHEET 1 OF 4) DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) .00000 .0 .0 .0 .0 .00108 90598.7 48888.7 40349.1 21718.9 .101 40353.3 21793.7 20648.3 11108.4

.202 46482.0 25280.7 22386.6 12050.6

.302 46307.3 25410.3 22661.8 12210.0

.402 46761.9 25935.4 22249.9 12001.2

.502 46296.0 25990.1 21549.4 11632.8

.602 44228.9 25131.5 20917.5 11297.9

.702 44745.5 25709.3 20392.8 11019.3

.801 44635.7 25899.4 19916.9 10765.7

.901 43950.0 25731.1 19498.4 10541.9 1.00 42962.6 25369.1 19151.6 10356.4 1.10 41980.3 24996.5 18887.2 10214.8 1.20 41034.1 24632.9 18699.2 10114.0 1.30 40163.4 24298.2 18560.7 10039.4 1.40 39364.0 23993.8 18435.7 9971.6 1.50 38594.0 23697.1 18316.5 9906.5 1.60 37802.5 23382.0 18214.6 9850.7 1.70 36954.7 23040.0 18126.9 9802.8 1.80 36079.1 22697.6 18027.0 9748.3 1.90 35106.4 22318.3 17885.7 9671.2 2.00 33872.9 21794.1 17715.2 9578.1 2.10 32344.2 21088.9 17549.4 9488.1 2.20 30820.7 20391.9 17349.3 9379.4 2.30 29075.9 19528.2 17091.6 9239.6 2.40 25410.4 17280.8 16797.9 9080.3 2.50 22059.0 15206.0 16471.4 8903.4 2.60 19923.2 13913.5 16092.2 8698.8 2.70 18262.9 12868.1 15821.6 8553.9 2.80 16954.1 12014.8 15567.8 8418.1 2.90 15973.2 11368.2 15302.8 8276.2 3.00 15192.2 10851.6 15017.0 8123.1 3.10 14591.9 10460.8 14790.8 8002.9 3.20 14100.8 10144.9 14596.6 7899.9 3.30 13665.0 9865.9 14406.5 7798.9 3.40 13257.3 9607.6 14230.0 7705.3 3.50 12888.2 9378.1 14119.3 7647.8 3.60 12567.0 9181.8 14054.6 7614.6 3.70 12247.2 8981.6 13891.7 7527.7 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-14 (SHEET 2 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 3.80 11934.8 8783.5 13724.4 7438.6 3.90 11648.7 8602.6 13561.2 7351.8 4.00 11396.4 8441.0 13396.3 7263.5 4.20 10974.9 8155.0 13058.2 7082.8 4.40 10630.1 7903.0 12794.2 6942.2 4.60 10367.3 7695.6 12572.7 6823.8 4.80 10144.5 7507.1 12312.0 6684.3 5.00 9969.6 7348.6 12100.0 6571.6 5.20 9816.7 7199.9 13057.1 7093.8 5.40 9710.3 7083.2 12743.2 6924.8 5.60 9660.3 7002.2 12615.2 6858.1 5.80 9648.9 6946.2 12420.1 6754.2 6.00 9649.2 6898.6 12299.5 6691.7 6.20 9661.8 6859.0 12185.5 6632.5 6.40 9882.2 6962.3 12043.3 6557.5 6.60 10191.9 7145.0 11990.7 6530.9 6.80 9940.3 7237.9 11911.0 6487.0 7.00 8907.6 6929.3 11760.1 6402.5 7.20 8293.4 6649.5 11602.2 6314.2 7.40 8116.3 6514.4 11455.6 6233.1 7.60 8051.8 6425.4 11314.4 6155.5 7.80 7991.2 6330.6 11160.2 6070.3 8.00 7966.4 6232.7 10991.6 5976.9 8.20 8002.8 6153.3 10830.6 5887.7 8.40 8065.0 6087.9 10677.9 5803.2 8.60 8120.8 6030.9 10524.0 5718.0 8.80 8149.6 5972.1 10369.5 5632.5 9.00 8135.7 5901.3 10219.6 5549.6 9.20 8089.0 5826.5 10074.9 5469.7 9.40 8007.8 5745.1 9929.7 5389.5 9.60 7894.2 5656.3 9786.3 5310.5 9.80 7753.1 5562.2 9647.8 5234.3 10.0 7605.4 5477.1 9509.1 5158.1 10.2 7437.8 5385.7 9364.9 5078.9 10.4 7264.6 5293.5 9226.4 5003.1 10.6 7093.4 5203.6 9086.0 4926.4 10.8 6923.6 5115.3 8946.7 4850.4 11.0 6758.0 5029.1 8809.1 4775.3 11.2 6595.2 4943.5 8671.7 4700.4 11.4 6439.1 4859.7 8536.6 4626.8 11.6 6289.0 4777.2 8402.2 4553.7 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-14 (SHEET 3 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 11.8 6145.2 4696.4 8269.5 4481.8 12.0 6008.1 4617.9 8138.9 4411.0 12.2 5876.2 4541.4 8010.2 4341.3 12.4 5748.2 4467.9 7882.8 4272.2 12.6 5624.9 4396.3 7759.7 4205.4 12.8 5504.5 4327.1 7637.5 4139.2 13.0 5386.5 4260.6 7517.0 4073.9 13.2 5271.4 4197.3 7398.9 4009.9 13.4 5154.4 4134.4 7213.8 3908.8 13.6 5029.6 4068.0 7067.1 3827.5 13.8 4888.2 3990.5 6878.9 3700.7 14.0 4735.1 3903.5 6840.4 3626.7 14.2 4567.2 3794.9 6612.0 3434.9 14.4 4414.2 3692.8 6562.6 3323.2 14.6 4279.4 3593.3 6813.5 3376.9 14.8 4176.4 3518.4 5778.5 2808.3 15.0 4087.4 3461.3 7270.0 3434.7 15.2 3953.2 3382.4 11154.0 5315.1 15.4 3773.3 3296.0 7340.3 3538.0 15.6 3761.2 3363.0 4436.2 2138.9 15.8 3693.6 3364.8 6754.0 3020.2 16.0 3470.4 3296.9 9836.8 4401.6 16.2 3281.2 3275.4 5828.1 2647.8 16.4 3221.5 3322.6 4708.7 2159.7 16.6 3092.4 3317.4 4129.3 1838.1 16.8 2770.1 3155.6 4691.2 1973.3 17.0 2504.2 3008.1 4954.4 2033.4 17.2 2210.4 2709.2 4419.5 1791.5 17.4 2001.1 2467.9 4204.6 1677.7 17.6 1821.6 2254.3 4260.6 1652.9 17.8 1651.8 2049.3 4635.3 1729.8 18.0 1497.6 1862.4 4545.8 1641.5 18.2 1353.9 1687.2 4397.7 1546.9 18.4 1219.1 1521.5 4220.7 1449.9 18.6 1091.9 1365.7 3948.7 1325.8 18.8 961.6 1204.7 3568.6 1171.0 19.0 840.8 1054.7 3109.8 997.2 19.2 735.9 924.0 2739.8 858.6 19.4 638.6 802.6 2316.7 710.0 19.6 566.7 712.9 1886.5 566.3 19.8 495.0 623.3 1443.6 425.6 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-14 (SHEET 4 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 20.0 449.1 565.7 984.8 286.2 20.2 405.0 510.5 529.0 152.3 20.4 351.5 443.3 105.7 30.4 20.6 293.9 370.8 .0 .0 20.8 225.7 285.0 .0 .0 21.0 147.8 186.9 126.9 36.8 21.2 88.6 112.3 87.3 25.3 21.4 29.2 37.2 .0 .0 21.6 .0 .0 .0 .0

  • mass and energy exiting the SG side of the break.
    • mass and energy exiting the pump side of the break.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-15 (SHEET 1 OF 2) REACTOR CAVITY RELEASE Enthalpy Time (s) Flow (lb/s) (Btu/lb) 0.0 0.0 0.0 1.00 x 10-3 7.16 x 103 5.56 x 102 3.04 x 10-3 1.09 x 104 5.55 x 102 5.04 x 10-3 1.33 x 104 5.54 x 102 7.04 x 10-3 1.47 x 104 5.52 x 102 9.08 x 10-3 1.69 x 104 5.51 x 102 1.01 x 10-2 1.72 x 104 5.51 x 102 1.10 x 10-2 1.71 x 104 5.50 x 102 1.31 x 10-2 1.63 x 104 5.48 x 102 1.40 x 10-2 1.72 x 104 5.48 x 102 1.51 x 10-2 1.85 x 104 5.48 x 102 1.70 x 10-2 1.99 x 104 5.47 x 102 1.90 x 10-2 2.01 x 104 5.46 x 102 2.01 x 10-2 2.02 x 104 5.45 x 102 2.11 x 10-2 2.00 x 104 5.44 x 102 2.31 x 10-2 1.99 x 104 5.43 x 102 2.51 x 10-2 1.97 x 104 5.42 x 102 2.70 x 10-2 1.97 x 104 5.41 x 102 2.91 x 10-2 1.98 x 104 5.40 x 102 3.11 x 10-2 2.01 x 104 5.40 x 102 3.31 x 10-2 2.04 x 104 5.39 x 102 3.50 x 10-2 2.08 x 104 5.39 x 102 3.70 x 10-2 2.10 x 104 5.39 x 102 3.91 x 10-2 2.13 x 104 5.38 x 102 4.11 x 10-2 2.14 x 104 5.38 x 102 4.21 x 10-2 2.14 x 104 5.38 x 102 4.31 x 10-2 2.14 x 104 5.37 x 102 4.51 x 10-2 2.11 x 104 5.37 x 102 4.71 x 10-2 2.09 x 104 5.36 x 102 4.92 x 10-2 2.06 x 104 5.36 x 102 5.10 x 10-2 2.04 x 104 5.35 x 102 5.31 x 10-2 2.03 x 104 5.35 x 102 5.50 x 10-2 2.03 x 104 5.35 x 102 6.01 x 10-2 2.04 x 104 5.35 x 102 6.50 x 10-2 2.03 x 104 5.34 x 102 7.01 x 10-2 2.03 x 104 5.34 x 102 7.51 x 10-2 2.02 x 104 5.34 x 102 8.01 x 10-2 1.97 x 104 5.34 x 102 8.50 x 10-2 1.92 x 104 5.33 x 102 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-15 (SHEET 2 OF 2) Enthalpy Time (s) Flow (lb/s) (Btu/lb) 9.00 x 10-2 1.89 x 104 5.33 x 102 9.50 x 10-2 1.87 x 104 5.33 x 102 1.00 x 10-1 1.85 x 104 5.33 x 102 1.20 x 10-1 1.91 x 104 5.33 x 102 1.25 x 10-1 1.90 x 104 5.33 x 102 1.50 x 10-1 1.71 x 104 5.32 x 102 1.75 x 10-1 1.74 x 104 5.32 x 102 1.81 x 10-1 1.75 x 104 5.32 x 102 2.00 x 10-1 1.72 x 104 5.32 x 102 2.50 x 10-1 1.78 x 104 5.32 x 102 3.00 x 10-1 1.74 x 104 5.32 x 102 3.50 x 10-1 1.78 x 104 5.32 x 102 4.00 x 10-1 1.78 x 104 5.32 x 102 4.50 x 10-1 1.80 x 104 5.32 x 102 4.60 x 10-1 1.80 x 104 5.32 x 102 4.70 x 10-1 1.80 x 104 5.32 x 102 5.00 x 10-1 1.78 x 104 5.32 x 102 5.50 x 10-1 1.78 x 104 5.32 x 102 6.00 x 10-1 1.78 x 104 5.32 x 102 6.50 x 10-1 1.76 x 104 5.32 x 102 7.00 x 10-1 1.74 x 104 5.32 x 102 7.50 x 10-1 1.73 x 104 5.32 x 102 8.00 x 10-1 1.75 x 104 5.32 x 102 8.50 x 10-1 1.76 x 104 5.32 x 102 9.00 x 10-1 1.76 x 104 5.32 x 102 9.50 x 10-1 1.77 x 104 5.32 x 102 1.00 1.78 x 104 5.32 x 102 1.50 1.81 x 104 5.32 x 102 1.90 1.82 x 104 5.33 x 102 2.40 1.82 x 104 5.32 x 102 2.80 1.80 x 104 5.32 x 102 3.00 1.79 x 104 5.32 x 102

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-16 SPRAY LINE BREAK RELEASE Enthalpy Time (s) Flow (lb/s) (Btu/lb) 0. 0. 6.42 x 102 0.025 3269 6.41 x 102 0.1 3245 6.39 x 102 0.15 3233 6.39 x 102 0.225 3210 6.39 x 102 0.3 3198 6.39 x 102 0.4 3186 6.39 x 102 0.75 3186 6.38 x 102 0.875 3174 6.38 x 102 1.0 3151 6.38 x 102 1.2 3127 6.38 x 102 1.4 3103 6.38 x 102 1.6 3080 6.38 x 102 1.8 3056 6.38 x 102 2.0 3033 6.38 x 101 2.2 3009 6.38 x 102 2.4 2985 6.38 x 102 2.6 2962 6.38 x 102 2.8 2938 6.38 x 102 3.0 2915 6.38 x 102

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-17 SURGE LINE BREAK RELEASE Enthalpy Time (s) Flow (lb/s) (Btu/lb) 0. 0. 692.8 0.025 6463 692.8 0.1 8585 700.9 0.15 8562 701.2 0.2 8569 700.2 0.3 8592 697.5 0.4 8600 695.0 0.5 8581 693.3 0.6 8533 692.7 0.7 8454 693.2 0.8 8352 694.4 0.9 8241 696.0 1.0 8133 697.3 1.2 7923 698.2 1.4 7841 696.3 1.6 7812 692.5 1.8 7789 688.8 2.0 7720 686.1 2.2 7619 684.4 2.4 7501 683.2 2.6 7381 681.1 2.8 7269 679.9 3.0 7167 677.5

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-18 (SHEET 1 OF 3) REACTOR CAVITY SUBCOMPARTMENT PRESSURE ANALYSIS SUMMARY OF FLOWPATHS AND VENT LOSS COEFFICIENTS Vent k Flowpath Area k k Bend + (from to) (ft2) Contraction Expansion Friction k C 1 2 14.4 0.04 1.0 0.429 1.47 0.83 1 9 0.3 0.42 1.0 --- 1.42 0.84 1 32 0.3 0.42 1.0 --- 1.42 0.84 29 0.4 0.42 1.0 --- 1.42 0.84 232 0.4 0.42 1.0 --- 1.42 0.84 234 14.118 0.067 1.0 --- 1.07 0.97 34 2.03 0.34 1.0 0.301 1.64 0.78 36 2.03 0.34 1.0 0.301 1.64 0.78 312 1.52 0.32 1.0 0.363 1.68 0.77 319 1.52 0.32 1.0 0.264 1.58 0.79 334 1.14 0.42 1.0 --- 1.42 0.84 45 2.03 --- 1.0 0.214 1.21 0.91 413 0.98 --- 1.0 0.371 1.37 0.85 420 0.98 --- 1.0 0.523 1.52 0.81 57 1.03 0.27 1.0 0.250 1.52 0.81 513 0.80 --- 1.0 0.377 1.38 0.85 520 0.80 --- 1.0 0.510 1.51 0.81 534 0.55 0.42 1.0 --- 1.42 0.84 631 1.03 0.27 1.0 0.330 1.60 0.79 611 1.05 --- 1.0 0.371 1.37 0.85 618 1.05 --- 1.0 0.522 1.52 0.81 634 0.55 0.42 1.0 --- 1.42 0.84 78 0.88 0.37 1.0 0.471 1.84 0.74 714 1.81 --- 1.0 0.360 1.36 0.86 721 1.81 --- 1.0 0.508 1.51 0.81 734 1.12 0.42 1.0 --- 1.42 0.84 815 1.27 --- 1.0 0.366 1.37 0.856 822 1.27 --- 1.0 0.515 1.52 0.812 834 0.57 0.42 1.0 --- 1.42 0.84 833 2.03 --- 1.0 0.310 1.31 0.874 932 0.88 0.37 1.0 0.460 1.83 0.74 916 1.27 --- 1.0 0.366 1.36 0.856 923 1.27 --- 1.0 0.515 1.52 0.812 1017 0.905 --- 1.0 0.374 1.37 0.853 1024 0.905 --- 1.0 0.527 1.53 0.809 1034 0.55 0.42 1.0 --- 1.42 0.84

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-18 (SHEET 2 OF 3) Vent k Flowpath Area k k Bend + (from to) (ft2) Contraction Expansion Friction k C 1112 0.95 --- 1.0 0.321 1.32 0.87 1116 0.95 --- 1.0 0.447 1.45 0.83 1134 0.65 --- 1.0 0.187 1.19 0.92 1213 0.95 --- 1.0 0.412 1.41 0.84 1234 0.95 --- 1.0 0.183 1.18 0.92 1314 0.95 --- 1.0 0.447 1.45 0.83 1334 1.10 --- 1.0 0.184 1.18 0.92 1415 0.95 --- 1.0 0.542 1.54 0.81 1434 1.13 --- 1.0 0.183 1.18 0.92 1517 0.95 --- 1.0 0.542 1.54 0.81 1534 1.58 --- 1.0 0.181 1.18 0.92 1617 0.95 --- 1.0 0.542 1.54 0.81 1634 1.58 --- 1.0 0.181 1.18 0.92 1734 1.13 --- 1.0 0.183 1.18 0.92 1819 2.17 --- 1.0 0.308 1.31 0.87 1823 2.17 --- 1.0 0.430 1.43 0.84 1825 1.05 --- 1.0 0.534 1.53 0.81 1920 2.17 --- 1.0 0.395 1.39 0.85 1925 1.52 --- 1.0 0.524 1.52 0.81 2021 2.17 --- 1.0 0.430 1.43 0.84 2025 1.78 --- 1.0 0.519 1.52 0.81 2122 2.17 --- 1.0 0.521 1.52 0.81 2126 1.81 --- 1.0 0.520 1.52 0.81 2224 2.17 --- 1.0 0.521 1.52 0.81 2226 2.54 --- 1.0 0.513 1.51 0.81 2324 2.17 --- 1.0 0.521 1.52 0.81 2327 2.54 --- 1.0 0.513 1.51 0.81 2427 1.81 --- 1.0 0.520 1.52 0.81 2526 2.13 --- 1.0 1.010 2.01 0.71 2527 2.13 --- 1.0 1.010 2.01 0.71 2528 4.35 --- 1.0 0.759 1.76 0.75 2627 2.13 --- 1.0 1.010 2.01 0.71 2628 4.35 --- 1.0 0.759 1.76 0.75 2728 4.35 --- 1.0 0.759 1.76 0.75 2829 106.70 --- 1.0 0.350 1.35 0.86 2930 93.31 0.05 1.0 0.070 1.12 0.94 3034 56.66 0.08 1.0 1.198 2.28 0.66 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-18 (SHEET 3 OF 3) Vent k Flowpath Area k k Bend + (from to) (ft2) Contraction Expansion Friction k C 3134 0.55 0.42 1.0 --- 1.42 0.84 319 2.03 --- 1.0 0.310 1.31 0.874 3116 1.27 --- 1.0 0.366 1.37 0.856 3123 1.27 --- 1.0 0.515 1.52 0.812 3210 2.03 --- 1.0 0.218 1.22 0.906 3217 0.905 --- 1.0 0.374 1.37 0.853 3224 0.905 --- 1.0 0.527 1.53 0.809 3310 1.03 0.34 1.0 0.440 1.78 0.75 3334 0.55 0.42 1.0 --- 1.42 0.84 3315 1.27 --- 1.0 0.366 1.37 0.856 3322 1.27 --- 1.0 0.515 1.5 0.812 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-19 CONTAINMENT RESULTS FOR THE DESIGN BASIS LOCA Prior to DEPSG DEHL Parameter LOCA At Peak At Peak Pressures Time (s) 552 18.8 Steam (psia) 1.03 37.1 37.7 Air (psia) 16.67 21.4 20.6 Total psia 17.70 58.5 58.3 Total gauge (psig) 3.0 43.8 43.6 Temperatures Time (s) 1252 20.0 Steam and air (°F) 127 263 264 Water in sump (°F) - 260 256 Heat transfer coefficient (Btu/h-ft2-°F)(a) 0 218 231

a. Between containment atmosphere and structure.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-20 (SHEET 1 OF 4) DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 21.6 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.2 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 .0 .0 .0 .0 22.6 55.3 65.3 .0 .0 22.7 31.1 36.7 .0 .0 22.8 33.2 39.2 .0 .0 22.9 39.1 46.1 .0 .0 23.0 45.9 54.2 .0 .0 23.1 49.4 58.3 .0 .0 23.2 55.4 65.3 .0 .0 23.3 59.8 70.6 .0 .0 23.4 64.0 75.5 .0 .0 23.5 68.0 80.2 .0 .0 23.6 71.8 84.7 .0 .0 23.7 75.4 89.0 .0 .0 23.8 79.0 93.2 .0 .0 23.9 82.4 97.2 .0 .0 24.0 85.7 101.1 .0 .0 24.1 88.8 104.9 .0 .0 24.2 91.9 108.5 .0 .0 24.3 94.9 112.0 .0 .0 24.4 97.8 115.5 .0 .0 24.5 100.7 118.8 .0 .0 24.6 103.4 122.1 .0 .0 25.6 127.9 151.0 .0 .0 26.6 148.3 175.1 .0 .0 27.6 165.9 195.9 .0 .0 28.2 321.3 380.3 2733.5 349.6 28.6 421.0 499.0 3803.6 500.4 29.7 453.5 537.9 4080.9 557.9 30.7 443.1 525.5 3984.9 548.9 31.7 460.0 545.7 4164.0 565.5 32.5 451.4 535.3 4085.8 556.9 32.7 449.2 532.8 4066.2 554.7 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-20 (SHEET 2 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 33.7 438.8 520.4 3970.3 544.3 34.7 428.9 508.5 3877.7 534.2 35.7 419.4 497.2 3788.3 524.4 36.7 410.3 486.4 3702.2 514.9 37.7 401.7 476.0 3619.1 505.8 37.9 400.0 474.0 3602.8 504.0 38.7 393.4 466.1 3538.9 497.0 39.7 385.4 456.7 3461.4 488.5 40.7 377.8 447.6 3386.5 480.3 41.7 370.5 438.9 3314.1 472.3 42.7 363.5 430.5 3243.9 464.5 43.7 356.7 422.4 3175.9 457.0 44.2 353.4 418.5 3142.6 453.4 44.7 350.2 414.7 3109.8 449.8 45.7 343.9 407.2 3045.7 442.7 46.7 337.8 400.0 2983.3 435.8 47.7 332.0 393.0 2922.7 429.1 48.7 326.3 386.3 2863.6 422.5 49.7 320.8 379.7 2806.0 416.1 50.7 315.5 373.4 2749.8 409.9 51.3 312.4 369.7 2716.8 406.2 51.7 310.3 367.3 2695.0 403.8 52.7 253.5 299.8 2021.5 332.6 53.7 324.4 383.8 280.0 153.3 54.7 341.3 404.0 285.3 162.0 55.7 336.8 398.8 283.7 159.7 56.7 332.4 393.5 282.0 157.5 57.7 328.1 388.3 280.4 155.3 58.7 323.7 383.2 278.8 153.2 59.7 319.5 378.1 277.2 151.1 60.7 315.2 373.1 275.6 149.0 61.7 311.0 368.0 274.1 146.9 62.7 306.7 362.9 272.5 144.8 63.7 302.7 358.2 271.0 142.8 64.7 298.7 353.5 269.6 140.9 65.7 294.9 348.9 268.2 139.0 66.5 291.8 345.2 267.1 137.5 66.7 291.0 344.3 266.8 137.1 67.7 287.2 339.8 265.4 135.3 68.7 283.5 335.4 264.1 133.5 69.7 279.8 331.0 262.8 131.8 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-20 (SHEET 3 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 70.7 276.2 326.6 261.5 130.0 71.7 272.6 322.4 260.2 128.3 72.7 269.0 318.2 258.9 126.7 73.7 265.6 314.1 257.7 125.0 74.7 262.1 310.0 256.5 123.4 75.7 258.7 306.0 255.3 121.9 76.7 255.4 302.0 254.1 120.3 77.7 252.1 298.1 253.0 118.8 78.7 248.9 294.3 251.9 117.3 79.7 245.8 290.6 250.8 115.9 80.7 242.6 286.9 249.7 114.5 81.7 239.6 283.2 248.6 113.1 82.7 236.6 279.6 247.6 111.7 84.2 232.2 274.4 246.1 109.8 84.7 230.7 272.7 245.6 109.1 86.7 225.1 266.0 243.7 106.6 88.7 219.7 259.6 241.8 104.2 90.7 214.5 253.5 240.1 102.0 92.7 209.5 247.5 238.4 99.8 94.7 204.7 241.9 236.8 97.8 96.7 200.1 236.5 235.3 95.8 98.7 195.7 231.3 233.9 94.0 100.7 191.6 226.3 232.5 92.2 102.7 187.6 221.6 231.2 90.5 104.7 183.8 217.1 230.0 89.0 105.3 182.7 215.8 229.7 88.5 106.7 180.2 212.9 228.9 87.5 108.7 176.8 208.8 227.8 86.1 110.7 173.5 205.0 226.8 84.8 112.7 170.5 201.4 225.8 83.6 114.7 167.6 198.0 224.9 82.4 116.7 164.9 194.8 224.1 81.4 118.7 162.4 191.8 223.3 80.3 120.7 160.0 189.0 222.5 79.4 122.7 157.8 186.3 221.9 78.5 124.7 155.7 183.9 221.2 77.7 126.7 153.7 181.6 220.6 77.0 128.7 151.9 179.4 220.1 76.3 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-20 (SHEET 4 OF 4) BREAK PATH NO.1 FLOW* BREAK PATH NO.2 FLOW** TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 130.1150.8 178.0 219.7 75.8 130.7 150.3 177.5 219.6 75.6 132.7 148.7 175.6 219.1 75.0 134.7 147.3 173.9 218.7 74.5 136.7 146.0 172.4 218.3 73.9 138.7 144.8 171.0 217.9 73.5 140.7 143.7 169.7 217.6 73.0 142.7 142.6 168.5 217.2 72.6 144.7 141.7 167.4 217.0 72.3 146.7 140.9 166.4 216.7 71.9 148.7 140.1 165.5 216.5 71.6 150.7 139.4 164.7 216.2 71.4 152.7 138.8 163.9 216.0 71.1 154.7 138.2 163.2 215.8 70.9 156.7 137.7 162.6 215.7 70.6 158.0 137.4 162.2 215.6 70.5 158.7 137.2 162.0 215.5 70.4 160.7 136.8 161.5 215.4 70.3 162.7 136.4 161.1 215.2 70.1 164.7 136.1 160.7 215.1 70.0 166.7 135.8 160.4 215.0 69.8 168.7 135.6 160.1 214.9 69.7 170.7 135.4 159.9 214.9 69.6 172.7 135.3 159.8 214.8 69.5 174.7 135.2 159.6 214.7 69.5 176.7 135.1 159.5 214.7 69.4 178.7 135.0 159.5 214.6 69.3 180.7 135.0 159.4 214.6 69.3 182.7 135.2 159.6 214.6 69.3 184.7 135.6 160.2 215.3 69.6 186.7 136.1 160.7 216.8 70.0 187.7 136.3 161.0 217.8 70.3

  • mass and energy exiting the SG side of the break.
    • mass and energy exiting the pump side of the break.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-21 LOCA CHRONOLOGY OF EVENTS Time (s) Event 0.0 Pipe ruptures (DEPSG), reactor depressurization begins. (a) Mass and energy release modeling.

62.0 Containment sprays begin operation. 92.0 Air coolers begin operation. 552.0 Containment reaches maximum peak pressure 1252 Sump reaches maximum temperature. 2139 Safety injection water recirculation from the sump begins as RWST reaches low level.

4256 Containment spray water recirculation from the sump begins as RWST reaches low-low level.

107 Containment reaches atmospheric pressure (estimate).

a. See table 6.2-52 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-22 (SHEET 1 OF 2) SUBCOMPARTMENT DIFFERENTIAL PRESSURE RESULTS Steam Generator Compartment Compartment Cold Leg Break (psid) Time (s) Design (psid) Pressurizer Surge Line Break (psid) 1, SG-C 33.9 0.42 35 7.3 2, SG-A 22.6 0.42 35 (b) 3, SG-B 19.3 0.60 35 (b) Pressurizer Compartment Compartment Spray Line Break (psid) Time (s) Design (psid) 1 9.4 0.14 35 2 3.1 0.10 35 Reactor Cavity Cold Leg Break Node No. Volume (ft3) Pressure (psia) Time (s) Design Pressure (psid) 1(a) 67.70 305.48 0.129 667 2(a) 104.12 288.02 0.129 667 3 17.17 19.15 0.600 150 4 7.36 18.83 0.600 150 5 9.46 18.26 0.600 150 6 11.37 21.30 0.600 150 7 14.47 18.40 0.600 150 8 12.41 19.80 0.600 150 9 12.41 57.26 0.135 150 10 10.27 41.65 0.140 150 11 3.68 21.64 0.600 150 12 5.33 18.78 0.600 150 13 6.22 18.21 0.600 150 14 6.35 18.02 0.600 150 15 8.89 19.79 0.600 150 16 8.89 36.50 0.141 150 17 6.35 36.59 0.141 150 a. Inside penetration at inspection opening.

b. Only the most limiting subcompartment pressure was re-analyzed.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-22 (SHEET 2 OF 2) Node No. Volume (ft3) Pressure (psia) Time (s) Design Pressure (psid) 18 8.42 21.36 0.600 150 19 12.19 19.32 0.600 150 20 14.23 18.83 0.600 150 21 14.52 18.86 0.600 150 22 20.32 19.98 0.600 150 23 20.32 26.36 0.148 150 24 14.52 26.29 0.148 150 25 34.25 18.58 0.600 150 26 34.25 18.59 0.600 150 27 34.25 19.69 0.600 150 28 1055.70 16.07 0.598 150 29 2190.60 16.02 0.600 150 30 3603.10 15.98 0.600 150 31 13.00 37.62 0.141 150 32 9.68 58.82 0.135 150 33 13.00 20.20 0.600 150 34 2.0 x 106 15.88 0.600 54 Net Vessel Side Load(b)

Lbf Time (s) 1.184 x 106 0.12

b. Reactor vessel support stresses not to exceed design criteria presented in tables 5.2-6 and 5.2-7.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-23

(This page has been intentionally deleted.)

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-24 (SHEET 1 OF 2) COMPONENT DESIGN PARAMETERS FOR CONTAINMENT SPRAY SYSTEM AND CONTAINMENT COOLING SYSTEM Containment Spray Pumps Type Horizontal Centrifugal Number 2 Pressure (psig) 300 Temperature (°F) 250 Flowrate (each) (gal/min) 2600 Head (ft) 450 Containment Coolers Number 4(a) Pressure (psig) 200 Temperature (°F) 300 Water inlet temperature (°F) 95 Flowrate (normal - high reactor coolant leakage)

(gal/min) 800

Heat removal rate (normal) (Btu/h) 2.36 x 106 Flowrate (post-LOCA) (gal/min) 2000 (600 for containment analysis) Heat removal rate (post-LOCA) (Btu/h) 80.0 x 106 (31.2 x 106 for containment analysis)

Containment Cooler Fans Type Vaneaxial Number 4 Flowrate (high speed) (sf3/min) 80,000 Static head (high speed) (in. wg) 4.75 Horsepower (high speed) (hp) 80 Flowrate (low speed) (sf3/min) 40,000 Static head (low speed) (in. wg) 7.90 Horsepower (low speed) (hp) 105 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-24 (SHEET 2 OF 2) COMPONENT DESIGN PARAMETERS FOR CONTAINMENT SPRAY SYSTEM AND CONTAINMENT COOLING SYSTEM Refueling Water Storage Tank Quantity 1 Volume (gal) 500,000 Design pressure (psig) atmosphere Design temperature (°F) ambient Material stainless steel Piping Pressure (psig) 210 Temperature (°F) 300 Valves Pressure (psig) 210 Temperature (°F) 300

a. Having fewer than 12 coils per containment cooler is acceptable, provided that each cooler can adequately remove the containment analysis heat load.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-25 (SHEET 1 OF 3) REGULATORY GUIDE 1.52, REV. 0 SECTION APPLICABILITY FOR THE PENETRATION ROOM FILTRATION SYSTEM Regulatory Guide Applicability to This Note Section System Index C.1.a Yes 1 C.1.b Yes -

C.1.c Yes -

C.1.d Yes -

C.1.e Yes -

C.2.a No 2 C.2.b No 3 C.2.c Yes -

C.2.d Yes -

C.2.e Yes 16 C.2.f Yes -

C.2.g Yes 4 C.2.h Yes -

C.2.i Yes -

C.2.j No 6 C.2.k Yes -

C.2.l Yes -

C.2.m Yes -

C.3.a No 7 C.3.b Yes 8 C.3.c Yes -

C.3.d Yes -

C.3.e Yes 9 C.3.f Yes -

C.3.g Yes -

C.3.h Yes 10 C.3.i Yes -

C.3.j No 11 C.3.k Yes -

C.3.l Yes 12 C.3.m Yes 13 C.3.n Yes -

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-25 (SHEET 2 OF 3) Regulatory Guide Applicability to This Note Section System Index C.4.a Yes - C.4.b Yes -

C.4.c Yes 14 C.4.d Yes -

C.4.e Yes -

C.4.f Yes -

C.4.g Yes -

C.4.h Yes 15 C.4.i Yes -

C.4.j Yes -

C.4.k Yes -

C.4.l Yes -

C.4.m Yes -

C.5.a Yes -

C.5.b Yes 17 C.5.c Yes 17 C.6.a Yes -

C.6.b Yes -

NOTES 1. The design basis accident is the postulated 30-day LOCA.

2. No demister is provided because the unit is located outside the containment and no entrained water droplets are anticipated. No HEPA filters are provided downstream of the charcoals, since radioactive fines carryover is very unlikely. This is true because the charcoal trays are pressure tested at high velocity in the manufacturer's shop prior to delivery, thereby removing fines. Also, during system operation, air is passing through the charcoal at a very low velocity.

3. No physical separation is provided since these units are located in a room where no missiles are postulated.

4. Pressure drops across the prefilters, HEPA, and charcoal filters are instrumented to indicate in the control room. Pressure drops across the HEPA and charcoal filters are instrumented to alarm in the control room. No recording of these signals is provided. Fan loss of flow is also instrumented to signal and alarm in the control room.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-25 (SHEET 3 OF 3) 5. Deleted 6. The size of the engineered safety feature filtration units precludes replacement as a single unit. The unit components are replaced individually.

7. Demisters are not provided.

8. Electric heaters are used to reduce the relative humidity to 70 percent or less. The use of heating coils to control the relative humidity during DBAs is not credited in the respective DBA dose assessment.

9. Mounting frames for filter and charcoals are constructed of carbon steel coated with an inorganic nuclear grade paint.

10. Internal welds are carbon steel coated with an inorganic nuclear grade paint.
11. The deluge and drain system has been eliminated due to recurring problems experienced at other facilities associated with inadvertent wetting of the absorber. Temperature gauges have been installed to monitor any heat rise in the filter housing.
12. Environmental conditions for systems considered are those specified under outside containment and radioactive area.
13. Duct construction guidelines follow SMACNA in addition to ORNL-NSIC-65.
14. Vacuum breakers are not used. This presents the probability of system leakage from pressure-relieving device leakage or failure.
15. Test probes are not manifolded and are located in readily accessible locations with minimum piping.
16. The accident analyses do not credit the heaters for humidity control.
17. Periodic testing to confirm a penetration of less than 0.5% at rated flow.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-26 SINGLE-FAILURE ANALYSIS - CONTAINMENT SPRAY SYSTEM Component Malfunction Comments and Consequences Spray Nozzles Clogged Large number of nozzles precludes clogging of a significant number. Pumps Containment Fails to start Two pumps provided. Spray pump Operation of one required. Automatically operated valves (open on coinci- dence of two out of four high-high-high containment pressure signals or 2/2 manual initiation of spray system operation from the control room): Containment spray Fails to open Two valves provided. pump discharge isolation valve Operation of one required. Valves operated from control room for recirculation Containment Fails to open Two lines in parallel, sump recirculation one to each spray pump. isolation Operation of one required.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-27 (SHEET 1 OF 4) DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS BLOWDOWN MASS AND ENERGY RELEASES BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) .00000 .0 .0 .0 .0 .00107 91449.9 49351.4 40349.8 21719.3

.101 40359.3 21797.0 20658.3 11113.7 .202 46617.0 25354.7 22408.0 12062.5

.302 46432.7 25480.6 22666.0 12212.7

.401 46869.1 25996.8 22243.1 11998.0

.501 46368.9 26033.7 21470.2 11589.9

.602 44240.1 25142.7 20750.4 11207.0

.702 44736.7 25706.9 20197.6 10911.5

.801 44554.7 25851.2 19705.5 10648.8

.902 43717.2 25588.5 19292.6 10429.1 1.00 42587.0 25134.1 18975.1 10260.1 1.10 41472.4 24675.8 18706.9 10116.8 1.20 40418.9 24239.3 18519.1 10016.3 1.30 39437.5 23830.9 18414.8 9960.8 1.40 38576.2 23476.8 18363.9 9933.9 1.50 37835.2 23180.4 18319.5 9910.0 1.60 37185.8 22928.6 18249.3 9871.5 1.70 36527.6 22670.7 18165.3 9825.3 1.80 35825.3 22391.2 18102.7 9790.9 1.90 35103.7 22112.9 18055.6 9765.3 2.00 34331.7 21820.3 17973.8 9720.8 2.10 33391.3 21435.8 17816.8 9635.2 2.20 32232.2 20918.3 17637.2 9537.5 2.30 30888.1 20287.8 17468.2 9446.2 2.40 29609.4 19693.0 17287.5 9348.6 2.50 28145.1 18960.7 16966.5 9174.7 2.60 24508.1 16681.4 16630.6 8993.2 2.70 21807.8 15032.8 16330.2 8831.3 2.80 20011.8 13965.9 16044.1 8677.3 2.90 18317.8 12903.2 15779.2 8535.2 3.00 16928.0 12017.7 15521.1 8397.0 3.10 15822.1 11306.9 15270.3 8262.8 3.20 14944.1 10739.5 15046.0 8143.3 3.30 14284.0 10314.6 14844.4 8036.1 3.40 13777.3 9987.8 14652.3 7933.9 3.50 13361.8 9717.3 14467.2 7835.4 3.60 12989.6 9474.7 14205.4 7694.8 3.70 12659.4 9261.3 14028.4 7600.9 3.80 12361.3 9070.4 13847.8 7504.7 3.90 12082.6 8891.7 13720.8 7438.2 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-27 (SHEET 2 OF 4) BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 4.00 11825.7 8727.7 13627.0 7388.7 4.20 11325.7 8401.6 13270.0 7197.5 4.40 10904.1 8121.2 12960.2 7032.1 4.60 10543.4 7866.4 12763.1 6927.4 4.80 10264.3 7653.1 12479.4 6774.8 5.00 10031.8 7461.6 12247.6 6650.9 5.20 9851.1 7298.9 13134.9 7138.0 5.40 9697.9 7150.0 12990.5 7057.9 5.60 9603.4 7039.6 12669.4 6886.1 5.80 9570.6 6967.1 12529.0 6811.9 6.00 9564.8 6913.1 12367.6 6726.8 6.20 9565.1 6864.3 12249.6 6665.6 6.40 9723.7 6927.8 12117.7 6596.4 6.60 10072.7 7127.9 12039.2 6556.2 6.80 9924.2 7261.9 11949.9 6508.5 7.00 8909.4 6974.7 11798.0 6424.8 7.20 8275.4 6686.1 11627.9 6330.5 7.40 8101.4 6554.2 11465.6 6241.2 7.60 8034.1 6477.0 11318.3 6160.3 7.80 7931.3 6382.2 11155.7 6070.8 8.00 7834.6 6267.4 10977.1 5972.0 8.20 7798.2 6161.1 10806.4 5877.7 8.40 7803.0 6069.0 10644.3 5788.2 8.60 7821.2 5991.1 10481.8 5698.3 8.80 7826.8 5914.7 10319.3 5608.5 9.00 7810.2 5836.0 10164.0 5522.5 9.20 7769.0 5755.2 10011.2 5438.0 9.40 7703.3 5672.7 9860.5 5354.7 9.60 7607.9 5582.5 9713.2 5273.3 9.80 7499.5 5497.8 9572.5 5195.7 10.0 7373.2 5411.9 9424.5 5114.1 10.2 7231.6 5320.2 9282.5 5036.1 10.4 7088.2 5230.3 9141.4 4958.7 10.4 7087.2 5229.7 9140.4 4958.1 10.4 7086.1 5229.1 9139.3 4957.5 10.6 6943.4 5142.6 9001.7 4882.1 10.8 6796.7 5056.6 8864.8 4807.2 11.0 6649.1 4971.8 8728.5 4732.7 11.2 6502.4 4888.3 8594.3 4659.5 11.4 6357.8 4806.1 8460.0 4586.5 11.6 6217.3 4725.9 8329.0 4515.4 11.8 6080.3 4647.4 8198.6 4444.7

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-27 (SHEET 3 OF 4) BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 12.0 5947.1 4570.8 8070.2 4375.1 12.2 5817.7 4496.2 7945.7 4307.6 12.4 5689.8 4423.9 7820.6 4239.8 12.6 5566.8 4354.1 7699.8 4174.3 12.8 5445.5 4285.7 7578.7 4108.8 13.0 5327.3 4219.9 7460.5 4044.9 13.2 5207.8 4153.8 7308.6 3962.2 13.4 5077.7 4081.2 7096.6 3847.0 13.6 4933.2 3998.6 6962.0 3763.2 13.8 4774.0 3901.6 6743.9 3609.9 14.0 4615.1 3797.8 6729.5 3543.7 14.2 4468.4 3694.9 6471.6 3337.2 14.4 4349.0 3606.0 6586.0 3319.5 14.6 4246.7 3531.2 6091.1 3005.9 14.8 4153.8 3473.2 6258.5 3014.6 15.0 4048.6 3417.8 6256.0 2966.8 15.2 3947.5 3378.7 5573.0 2607.2 15.4 3844.1 3346.3 5805.5 2657.7 15.6 3719.0 3305.7 5945.1 2683.1 15.8 3608.1 3289.6 5479.0 2453.2 16.0 3486.7 3275.8 5194.7 2300.8 16.2 3362.9 3270.5 5095.5 2229.6 16.4 3220.0 3261.8 5073.6 2193.1 16.6 3025.9 3222.9 4973.3 2127.3 16.8 2725.4 3101.7 4688.8 1986.2 17.0 2481.9 2985.0 4330.8 1817.9 17.2 2199.8 2697.7 4020.6 1669.3 17.4 1987.4 2451.5 3709.4 1515.2 17.6 1804.7 2233.5 3547.4 1413.2 17.8 1636.2 2030.1 3734.9 1439.5 18.0 1478.6 1838.6 4080.5 1520.7 18.2 1330.6 1658.1 4340.7 1567.9 18.4 1196.0 1492.6 4064.5 1432.5 18.6 1065.0 1332.3 3763.4 1297.4 18.8 937.4 1174.4 3425.7 1154.3 19.0 823.9 1033.5 3022.7 994.5 19.2 731.6 918.6 2674.9 858.6 19.4 651.6 818.8 2261.5 708.0 19.6 594.6 748.0 1812.0 554.0 19.8 538.3 677.5 1332.1 398.9

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-27 (SHEET 4 OF 4) BREAK PATH NO.1 FLOW BREAK PATH NO.2 FLOW TIME THOUSAND THOUSAND (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 20.0 483.3 608.7 816.3 240.7 20.2 426.2 537.0 306.4 89.7 20.4 366.0 461.5 .0 .0 20.6 305.4 385.3 .0 .0 20.8 248.8 314.1 .0 .0 21.0 193.1 244.0 .0 .0 21.2 107.4 136.0 .0 .0 21.4 12.3 15.7 .0 .0 21.6 .0 .0 .0 .0

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-28 (SHEET 1 OF 2) CONTAINMENT VENTILATION SYSTEMS COMPONENT DESIGN PARAMETERS

Containment Coolers (normal)

Number 4 Pressure (psig) 200 Temperature (°F) 300 Water inlet temperature (°F) 95 Flowrate (each) (gal/min) 800 Heat removal rate (each) (btu/h) 2.36 x 106 Containment Cooler Fans (normal)

Type Vaneaxial Number 4 Flowrate (each) (sft3/min) 80,000 Static head (in. WG) 4.75 Motor horsepower (each) (hp) 80 Containment Recirculation Fans Type Vaneaxial Number 4 Flowrate (each) (sft3/min) 25,000 Static head (in. WG) 0.32 Motor horsepower (each) (hp) 7.5 Control-Rod Mechanism Cooling Fans Type Vaneaxial Number 2 Flowrate (each) (sft3/min) 40,000 Static head (in. WG) 9.0 Motor horsepower (each) (hp) 100 Reactor Cavity Cooling Fans Type Vaneaxial Number 2 Flowrate (each) (sft3/min) 17,000 Static head (in. WG) 2.46 Motor horsepower (each) (hp) 15 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-28 (SHEET 2 OF 2)

Refueling Water Surface Ventilation Supply Fan Type Vaneaxial Number 1 Flowrate (each) (sft3/min) 7,500 Static head (in. WG) 4.5 Motor horsepower (each) (hp) 15 Refueling Water Surface Ventilation Exhaust Fan Type Vaneaxial Number 1 Flowrate (each) (sft3 min) 22,000 Static head (in. WG) 2.0 Motor horsepower (each) (hp) 15

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-29 SPRAY EVALUATION PARAMETERS Spray flowrate (gal/min) 2480 (injection) 2290 (recirculation)

Containment sump volume (ft3) 4.92 x 104 Containment sprayed volume (ft3) 1.67 x 106 Minimum spray fall height (ft) 110 Elemental s (h-1) 10.0 (DF < 21) 0.0 (DF > 21)

Methyl s (h-1) 0.0 Particulate s (h-1) 5.4 (injection) 5.0 (recirculation, DF < 50) 0.0 (> 8 h) 0.5 (DF > 50 until 8 h) pH (Spray injection) 4.5 pH (Spray recirculation) 7.7

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-30 SINGLE FAILURE ANALYSIS - PENETRATION ROOM FILTRATION SYSTEM Component Malfunction Comments Fan Fails The other fan and filter system will be available. Fan discharge valve Fails to open Same as above.

Fan discharge valve Fails to close Check valve will prevent back flow.

Recirculation line valve Fails to open Recirculation fan will operate in the exhaust mode.

Recirculation line valve Fails to close The other system will be available.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-32 (SHEET 1 OF 2) STEAM GENERATOR ISOLATION VALVE INFORMATION Service Penetra- Valve Flow Location Penetration Normal Valve Valve Item (No. of tion Arrange- Direc- Relative to Line Valve Position with Position Post LOCA Closure No. Penetrations) System Type ment tion Containment Valve Type Size (in.) Actuator Signal Position Power Failure Indicator Position Time(s) 1 Main Steam (3) MS III 3 Out Outside Power operated 32 Air to open SLIAS Open As Is Yes Closed 7 Check Spring closed 2 Main Steam MS III 3 Out Outside Gate 3 Air SLIAS Closed Closed Yes Closed < 7 (Note 2) Isolation Valve Bypass (3) 3 Steam to MS III 3 Out Outside Stop check 3 Air Remote manual Closed Closed Yes Closed --- Auxiliary Feedwater Pump Turbine Drive (2) 4 Steam to MS III 3 Out Outside Globe 1 Air T Open Closed Yes Closed NA Aux. Feedwater Pump Drive Warming Line (2) 5 Main Steam MS III 3 Out Outside Globe 6 Air Remote manual Closed Closed Yes Closed < 35 Atmospheric Relief (3) 6 Feedwater (3) FW III 4 In Outside Stop check 14 Electric motor Remote manual Open As is Yes Open 30 7 Auxiliary Feedwater (3) AFW III 4 In Outside Stop check 4 Electric motor Remote manual Open As is Yes Open 14 8 Steam Generator MS III 25 Out Outside Globe 2 Air AFPSS Open Closed Yes Closed < 60 Blowdown (3) 9 Steam Generator SS III 32 Out Outside Globe 3/8 Air Remote manual Open Closed Yes Closed < 5 Blowdown Sample (3) See Note 1 below 10 Chemical FW III 4 In Outside Globe 1/2 Air T Open Closed Yes Closed < 5 Injection (3)

1. Flow is isolated on AFPSS by valves inside containment. 2. Design requirement only, not operability requirement.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-32 (SHEET 2 OF 2)

The "Valve Arrangement" number refers to figures 6.2-84 through 6.2-89.

The abbreviations used in table 6.2-31 and 6.2-32 are as follows:

SYSTEM DWS - Demineralized Water System CCS - Component Cooling System SIS - Safety Injection System RCS - Reactor Coolant System WPS - Waste Processing System MS - Main Steam System FW - Feedwater System RHRS - Residual Heat Removal System CVCS - Chemical and Volume Control System SS - Sampling System FHS - Fuel Handling System RMS - Radiation Monitoring System H&V - Heating and Ventilation SA - Service Air System IA - Instrument Air System FWS - Fire Water System SWS - Service Water System AFW - - Auxiliary Feedwater System

SIGNALS T - Containment Isolation Actuation Signal, Phase A S - Safety Injection Signal P - Containment Isolation Actuation Signal, Phase B CSAS - Containment Spray Actuation Signal SLIAS - Steam Line Isolation Actuation Signal AFPSS - Auxiliary Feedwater Pump Start Signal

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-33 ELECTRIC HYDROGEN RECOMBINER TYPICAL PARAMETERS

Parameter Value Power (maximum) 75 kW(a) Capacity (minimum) 100 sft3/min Heaters -Number 5 -Heater surface area/heater 35 ft2 -Maximum heat flux 2850 Btu/h-ft2 or 5.8 Watts/in.2 -Maximum sheath temperature 1550°F Gas Temperature -Inlet 80 to 155°F -In heater section 1150 to 1400°F Materials -Outer structure 300-Series S.S. -Inner structure Inconel-600 -Heater element sheath Incoloy-800 Dimensions -Height 9 ft -Width 4.5 ft -Depth 5.5 ft Weight 4500 lb

a. Power can be controlled by SCR. Normal operating power for typical PWR containments is 48.9.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-34 POSTACCIDENT VENTING SYSTEM DESIGN PARAMETERS

Parameters Value Valves -Design pressure (psig) 150(a) -Design temperature (°F) Inside containment 300(a) Outside containment 300(a) Piping -Design pressure (psig) 150(a) -Design temperature (°F) Inside containment 300(a) Outside containment 300(a) HEPA Filter -Number 1 -Air Flow (sft3/min) 500 -Approximate differential pressure (wg) 1.5 -Maximum differential pressure (loaded) (wg) 4.0 -Design temperature (°F) 180 -Particulate removal efficiency (0.3 micron) 99.97 Charcoal -Number 1 -Air flow (sft3/min) 500 -Differential pressure (wg) 2.7 -Design temperature (°F) 180 -Charcoal type iodine impregnated-Elemental I2 removal efficiency 99.9 -Organic I2 removal efficiency 99.0

a. Represents as installed ratings of system piping and valves. Actual design requirements may be substantially lower and may vary throughout the system.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-35 POSTACCIDENT SAMPLING SYSTEM DESIGN PARAMETERS

Sample Vessel Number 2 Number required for operation 1 Design pressure (psig) 150 Design temperature (°F) 300 Material of construction Stainless steel Valves Design pressure (psig) 150 Design temperature (°F) 300 Piping Design pressure (psig) 150 Design temperature (°F) 300 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-36 POSTACCIDENT MIXING SYSTEM DESIGN PARAMETERS

Post-LOCA Containment Mixing Fans: Type Vaneaxial Number required 4 Flow (ft3/min) each 7500 Static pressure (in. w.g.) 2.3 Reactor Cavity Hydrogen Dilution Fans: Unit 1 Type Centrifugal Number required 2 Flow (ft3/min) each 270 Static pressure (in. w.g.) 126.7 Unit 2 Type Vaneaxial Number required 2 Flow (ft3/min) each 1570 Static pressure (in. w.g.) 3.26

FNP-FSAR-6 REV 22 8/09 TABLE 6.2-37 CONTAINMENT INTERIOR COATINGS SUMMARY Ext. Ext. Dry Top Coat Top Coat Dry Surface Primer/Surfacer Specific Recoat Recoat Specific Surface Area Type Manufacturer Product No. Generic Type Gravity Product No. Generic Type Gravity (ft2)(a) Carbon Ameron(b) Dimetcote Inorganic zinc 3.15 Amercoat 66 Epoxy polyamide 2.60 213,750 Steel No. 6 (D-6) Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58 Carboline Carbozinc 11 Inorganic zinc 4.61 Phenoline 305 Epoxy phenolic 1.73 17,500 SG (CZ-11) Amercoat 90 Modified phenolic 2.58 4674 (Black)(c) Modified silicone 1.345 --- --- --- 2,300 with low chloride content 4700 Aluminum free 1.28 --- --- --- < 2 paint 4674 (Aluminum)(c) Modified silicone 1.54 --- --- --- 12,000 aluminum with low chloride content Sterling U-475 ERN Epoxy varnish 1.001(d) --- --- --- 28 --- Galvanized Hot dipped zinc 7.15 --- --- --- 62,932 Concrete Ameron(b) NU-KLAD Epoxy polyamide 1.95 Amercoat 66 Epoxy polyamide 2.60 80,000 110AA - solid filled Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58 Amercoat 90 Modified phenolic 2.58 Amercoat 3366 Epoxy surfacer 2.12 Amercoat 90HS Epoxy phenolic 1.72 Amercoat 3367 Epoxy filler 1.80 Amercoat 90HS Epoxy phenolic 1.72

a. For coating requirements see NMP-MA-011, Nuclear Coatings Program. b. Either system is acceptable for use as original system. c. Generally covered by insulation. d. Wet specific gravity at 75°F.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 1 OF 7) CONTAINMENT ISOLATION VALVES (

Reference:

Table 6.2-31 [See Note])

No. Location FSAR Figure System P&ID Valve Identification Number 1 ic 6.2-87 (24) D-175038, Sh. 2 Q1E21V049 D-205038, Sh. 2 Q2E21V049 oc 6.2-87 (24) D-175038, Sh. 2 Q1E21V050 D-205038, Sh. 2 Q2E21V050 2 ic 6.2-89 (35) D-175043(a) Q1G31V013 D-205043 Q2G31V013 oc 6.2-89 (35) D-175043(a) Q1G31V012 D-205043 Q2G31V012 3 ic 6.2-87 (23) D-175038, Sh. 2 Q1E21V058 D-205038, Sh. 2 Q2E21V058 oc 6.2-87 (23) D-175038, Sh. 2 Q1E21V059 D-205038, Sh. 2 Q2E21V059 4 ic 6.2-89 (38) D-175037, Sh. 2 Q1B13V037 D-205037, Sh. 2 Q2B13V037 oc 6.2-89 (38) D-175037, Sh. 2 Q1B13V039 D-205037, Sh. 2 Q2B13V039 5 ic 6.2-87 (27) D-175037, Sh. 2 Q1B13V038 D-205037, Sh. 2 Q2B13V038 oc 6.2-87 (27) D-175037, Sh. 2 Q1B13V040 D-205037, Sh. 2 Q2B13V040 D-175037, Sh. 2 Q1B13V110 D-205037, Sh. 2 Q2B13V110 6 ic 6.2-88 (34) D-175037, Sh. 2 Q1B13V054 D-205037, Sh. 2 Q2B13V054 oc 6.2-88 (34) D-175039, Sh. 6 Q1E21V263A,B D-205039, Sh. 2 Q2E21V263A,B oc 6.2-88 (34) D-175038, Sh. 2 Q1E11V039A,B D-175038, Sh. 2 Q1E11V040 D-205038, Sh. 2 Q2E11V039A,B D-205038, Sh. 2 Q2E11V040 7 ic 6.2-84 (1) D-175042, Sh. 1 Q1G21V064 D-175042, Sh. 1 Q1G21V005 D-205042, Sh. 1(a) Q2G21V064 D-205042, Sh. 1(a) Q2G21V005 oc 6.2-84 (1) D-175042, Sh. 1 Q1G21V006 D-205042, Sh. 1(a) Q2G21V006 D-175042, Sh. 1 Q1G21V950 D-205042, Sh. 1(a) Q2G21V950

a. This drawing is not presented in the FSAR because the corresponding drawing is applicable to both units. Note: Item numbers correlate with those on Table 6.2-31.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 2 OF 7)

No. Location FSAR Figure System P&ID Valve Identification Number 8 ic 6.2-86 (16) D-175010, Sh. 2 Q1E14V004 D-205010, Sh. 2 Q2E14V004 oc 6.2-86 (16) D-175010, Sh. 2 Q1E14V003 D-205010, Sh. 2 Q2E14V003 9 ic 6.2-85 (5) D-175041 Q1E11V001A,B D-205041 Q2E11V001A,B oc See note 2 of FSAR table 6.2-31. 10 ic 6.2-88 (30) D-175038, Sh. 2 Q1E11V042A,B D-205038, Sh. 2 Q2E11V042A,B oc 6.2-88 (30) D-175038, Sh. 2 Q1E11V023A,B D-205038, Sh. 2 Q2E11V023A,B 11 ic 6.2-88 (33) D-175039, Sh. 1 Q1E21V253A,B,C D-205039, Sh. 1 Q2E21V253A,B,C oc 6.2-88 (33) D-175039, Sh. 1 Q1E21V254 D-205039, Sh. 1 Q2E21V254 12 ic 6.2-85 (6) D-175039, Sh. 1 Q1E21V249A D-175039, Sh. 1 Q1E21V213 D-205039, Sh. 1 Q2E21V249A D-205039, Sh. 1 Q2E21V213 oc 6.2-85 (6) D-175039, Sh. 1 Q1E21V249B D-205039, Sh. 1 Q2E21V249B 13 ic 6.2-85 (7) D-175039, Sh. 1 Q1E21V119 D-205039, Sh. 1 Q2E21V119 oc 6.2-85 (7) D-175039, Sh. 6 Q1E21V257 D-175039, Sh. 6 Q1E21V258 D-205039, Sh. 2 Q2E21V257 D-205039, Sh. 2 Q2E21V258 14 ic 6.2-86 (15) D-175039, Sh. 1 Q1E21V115A,B,C D-205039, Sh. 1 Q2E21V115A,B,C See note 4 of FSAR table 6.2-31. oc 6.2-86 (15) D-175039, Sh. 1 - D-205039, Sh. 2 - See note 4 of FSAR table 6.2-31. 15 ic D-175038, Sh. 3 - D-205038, Sh. 3 - See note 5 of FSAR table 6.2-31. oc D-175038, Sh. 3 - D-205038, Sh. 3 - See note 5 of FSAR table 6.2-31.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 3 OF 7)

No. Location FSAR Figure System P&ID Valve Identification Number 16 ic 6.2-87 (24) D-175009, Sh. 2 Q1P15SV3104 D-205009, Sh. 2 Q2P15SV3104 oc 6.2-87 (24) D-175009, Sh. 2 Q1P15SV3331 D-205009, Sh. 2 Q2P15SV3331 17 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3103 D-205009, Sh. 1 Q2P15SV3103 oc 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3332 D-205009, Sh. 1 Q2P15SV3332 18 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15SV3765 D-205009, Sh. 1 Q2P15SV3765 D-175009, Sh. 1 Q1P15SV3333 D-205009, Sh. 1 Q2P15SV3333 19 ic 6.2-87 (26) D-175067(b) - D-205067(a)(b) - 20 ic 6.2-87 (10) D-175035, Sh. 1 Q1P18V002 D-205035, Sh. 1 Q2P18V002 oc 6.2-87 (10) D-175035, Sh. 1 Q1P18V001 D-205035, Sh. 1 Q2P18V001 20a ic 6.2-87 (10) D-205035, Sh. 1 Q2P18V005 oc 6.2-87 (10) D-205035, Sh. 1 Q2P18V004 21 ic 6.2-87 (23) D-175034, Sh. 3 Q1P19V002 D-205034, Sh. 4 Q2P19V002 oc 6.2-87 (23) D-175034, Sh. 2 Q1P19HV3611 D-205034, Sh. 2 Q2P19HV3611 22 ic 6.2-87 (23) D-175010, Sh. 2 Q1E14V001 D-205010, Sh. 2 Q2E14V001 oc 6.2-87 (23) D-175010, Sh. 2 Q1E14HV3657 D-205010, Sh. 2 Q2E14HV3657 23 ic 6.2-85 (11) D-175010, Sh. 2 Q1E14V002 D-205010, Sh. 2 Q2E14V002 oc 6.2-85 (11) D-175010, Sh. 2 Q1E14HV3658 D-175010, Sh. 2 Q2E14HV3658 24 ic 6.2-85 (12) D-175010, Sh. 1 Q1P13V282 D-205010, Sh. 1 Q2P13V282 oc 6.2-85 (12) D-175010, Sh. 2 Q1P13V281 D-205010, Sh. 2 Q2P13V281

b. This is only a general arrangement drawing.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 4 OF 7) No. Location FSAR Figure System P&ID Valve Identification Number 24a ic 6.2-85 (12) D-175010, Sh. 1 Q1P13V302 D-205010, Sh. 1 Q2P13V302 oc 6.2-85 (12) D-175010, Sh. 2 Q1P13V301 D-205010, Sh. 2 Q2P13V301 25 ic 6.2-85 (13) D-175010, Sh. 1 Q1P13V283 D-205010, Sh. 1 Q2P13V283 oc 6.2-85 (13) D-175010, Sh. 2 Q1P13V284 D-205010, Sh. 2 Q2P13V284 25a ic 6.2-85 (13) D-175010, Sh. 1 Q1P13V304 D-205010, Sh. 1 Q2P13V304 oc 6.2-85 (13) D-175010, Sh. 2 Q1P13V303 D-205010, Sh. 2 Q2P13V303 26 ic 6.2-89 (39) D-175004, Sh. 1 Q1G21V291 D-175004, Sh. 1 Q1G21HV3376 D-205004, Sh. 1(a) Q2G21V291 D-205004, Sh. 1(a) Q2G21HV3376 oc 6.2-89 (39) D-175004, Sh. 1 Q1G21HV3377 D-205004, Sh. 1(a) Q2G21HV3377 27 ic 6.2-88 (30) D-175003, Sh. 1 Q1P16V206A,B,C,D D-205003, Sh. 1(a) Q2P16V206A,B,C,D oc 6.2-88 (30) D-175003, Sh. 1 Q1P16V010A,B,C,D D-205003, Sh. 1(a) Q2P16V010A,B,C,D D-175003, Sh. 1 Q1P16V205A,B,C,D D-205003, Sh. 1(a) Q2P16V205A,B,C,D 28 ic 6.2-85 (9) D-175003, Sh. 1 Q1P16V207A,B,C,D D-205003, Sh. 1(a) Q2P16V207A,B,C,D oc 6.2-85 (9) D-175003, Sh. 1 Q1P16V043A,B,C,D D-205003, Sh. 1(a) Q2P16V043A,B,C,D D-175003, Sh. 1 Q1P16V044A,B,C,D D-205003, Sh. 1(a) Q2P16V044A,B,C,D D-175003, Sh. 1 Q1P16V208A,B,C,D D-205003, Sh. 1(a) Q2P16V208A,B,C,D 29 ic 6.2-88 (30) D-175002, Sh. 2 Q1P17V083 D-205002, Sh. 2 Q2P17V083 oc 6.2-88 (30) D-175002, Sh. 2 Q1P17V082 D-205002, Sh. 2 Q2P17V082 D-175002, Sh. 2 Q1P17V158 D-205002, Sh. 2 Q2P17V158 30 oc 6.2-86 (21) D-175010, Sh. 1 Q1P23V002A,B D-205010, Sh. 1 Q2P23V002A,B

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 5 OF 7)

No. Location FSAR Figure System P&ID Valve Identification Number 31 ic 6.2-88 (29) D-175002, Sh. 2 Q1P17V097 D-205002, Sh. 2 Q2P17V097 oc 6.2-88 (29) D-175002, Sh. 2 Q1P17V099 D-205002, Sh. 2 Q2P17V099 D-175002, Sh. 2 Q1P17V155 D-205002, Sh. 2 Q2P17V155 32 ic 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3184 D-205002, Sh. 2 Q2P17HV3184 oc 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3045 D-205002, Sh. 2 Q2P17HV3045 33 ic 6.2-85 (8) D-175002, Sh. 2 Q1P17V159 D-205002, Sh. 2 Q2P17V159 oc 6.2-85 (8) D-175002, Sh. 2 Q1P17HV3095 D-205002, Sh. 2 Q2P17HV3095 D-175002, Sh. 2 Q1P17V153 D-205002, Sh. 2 Q2P17V153 34 ic 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3443 D-205002, Sh. 2 Q2P17HV3443 oc 6.2-87 (24) D-175002, Sh. 2 Q1P17HV3067 D-205002, Sh. 2 Q2P17HV3067 D-175002, Sh. 2 Q1P17V154 D-205002, Sh. 2 Q2P17V154 35 ic 6.2-86 (14) D-175038, Sh. 1 Q1E21V078A,B,C D-205038, Sh. 1 Q2E21V078A,B,C D-175038, Sh. 1 Q1E21V079A,B,C D-205038, Sh. 1 Q2E21V079A,B,C D-175038, Sh. 1 Q1E21V066A,B,C D-205038, Sh. 1 Q2E21V066A,B,C oc 6.2-86 (14) D-175038, Sh. 1 Q1E21V068 D-175038, Sh. 1 Q1E21V072 D-175038, Sh. 1 Q1E21V063 D-205038, Sh. 1 Q2E21V068 D-205038, Sh. 1 Q2E21V072 D-205038, Sh. 1 Q2E21V063 36 ic 6.2-88 (30) D-175038, Sh. 3 Q1E13V002A,B D-205038, Sh. 3 Q2E13V002A,B oc 6.2-88 (30) D-175038, Sh. 3 Q1E13V005A,B D-205038, Sh. 3 Q2E13V005A,B 37 oc 6.2-86 (18) D-175038, Sh. 2 Q1E11V025A,B D-205038, Sh. 2 Q2E11V025A,B D-175038, Sh. 2 Q1E11V026A,B D-205038, Sh. 2 Q2E11V026A,B FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 6 OF 7)

No. Location FSAR Figure System P&ID Valve Identification Number 38 oc 6.2-86 (18) D-175038, Sh. 3 Q1E13V003A,B D-205038, Sh. 3 Q2E13V003A,B D-175038, Sh. 3 Q1E13V004A,B D-205038, Sh. 3 Q2E13V004A,B 39 ic 6.2-85 (39) D-175038, Sh. 2 Q1E21V052 D-205038, Sh. 2 Q2E21V052 oc 6.2-85 (39) D-175038, Sh. 2 Q1E21V091 D-205038, Sh. 2 Q2E21V091 40 ic 6.2-87 (24) D-175009, Sh. 1 Q1P15HV3766 D-205009, Sh. 1 Q2P15HV3766 oc 6.2-87 (24) D-175009, Sh. 1 Q1P15HV3334 D-205009, Sh. 1 Q2P15HV3334 41 ic See note 7 of FSAR table 6.2-31. oc 6.2-86 (19) D-175037, Sh. 2 Q1B13V026B D-205037, Sh. 2 Q2B13V026B 42 ic 6.2-87 (28) D-175042, Sh. 1 Q1G21V082 D-205042, Sh. 1(a) Q2G21V082 oc 6.2-87 (28) D-175042, Sh. 1 Q1G21V001 D-205042, Sh. 2(a) Q2G21V001 43 ic 6.2-84 (2) D-175038, Sh. 1 Q1E21V062A,B,C D-205038, Sh. 1 Q2E21V062A,B,C oc 6.2-84 (2) D-175038, Sh. 1 Q1E21V016A,B D-205038, Sh. 1 Q2E21V016A,B 44 ic 6.2-86 (17) D-175038, Sh. 1 Q1E21V076A,B D-205038, Sh. 1 Q2E21V076A,B oc 6.2-86 (17) D-175038, Sh. 2 Q1E11V044 D-205038, Sh. 2 Q2E11V044 45 ic 6.2-88 (30) D-175003, Sh. 2 Q1P16V075 D-205003, Sh. 2(a) Q2P16V075 oc 6.2-88 (30) D-175003, Sh. 2 Q1P16V071 D-205003, Sh. 2(a) Q2P16V071 D-175003, Sh. 2 Q1P16V204 D-205003, Sh. 2(a) Q2P16V204 46 ic 6.2-88 (29) D-175003, Sh. 2 Q1P16V081 D-205003, Sh. 2(a) Q2P16V081 oc 6.2-88 (29) D-175003, Sh. 2 Q1P16V072 D-205003, Sh. 2(a) Q2P16V072 D-175003, Sh. 2 Q1P16V203 D-205003, Sh. 2(a) Q2P16V203 47 ic 6.2-87 (23) D-175004, Sh. 1 Q1G21V204 D-205004, Sh. 1(a) Q2G21V204 oc 6.2-87 (23) D-175004, Sh. 1 Q1G21HV3380 D-205004, Sh. 1(a) Q2G21HV3380 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-39 (SHEET 7 OF 7)

No. Location FSAR Figure System P&ID Valve Identification Number 48 ic 6.2-89 (37) D-175019 Q1E23V022A,B,C,D D-205019 Q2E23V022A,B,C,D oc 6.2-89 (37) D-175019 Q1E23V023A,B D-205019 Q2E23V023A,B 49 ic 6.2-86 (20) D-175019 Q1E23V025A,B D-205019 Q2E23V025A,B oc 6.2-86 (20) D-175019 Q1E23V024A,B D-205019 Q2E23V024A,B 50 ic 6.2-86 (16) D-175019 Q1E23V003 D-205019 Q2E23V003 oc 6.2-86 (16) D-175019 Q1E23V002 D-205019 Q2E23V002 51 ic 6.2-87 (23) D-175047 Q1P11V002 D-205047 Q2P11V002 oc 6.2-87 (23) D-175047 Q1P11V001 D-205047 Q2P11V001 52 ic 6.2-87 (23) D-175034, Sh. 1 Q1P19V004 D-205034, Sh. 4(a) Q2P19V004 oc 6.2-87 (23) D-175034, Sh. 1 Q1P19HV2228 6.2-87 (23A) D-205034, Sh. 4(a) Q1P19V1099 Q2P19V006 Q2P19V1099 53 ic 6.2-89 (40) D-206164(b) --- oc 6.2-89 (40) D-206164(b) ---

a. This drawing is not presented in the FSAR because the corresponding drawing is applicable to both units.

b. This is only a general arrangement drawing.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-40 STEAM GENERATOR ISOLATION VALVES (

Reference:

TABLE 6.2-32 [See Note] )

Valve Item FSAR System Identification No. Location Figure P&ID Number 1 oc 6.2-84 (3) D-175033 Sh. 1 QV001A,B,C QV002A,B,C 2 oc 6.2-84 (3) D-175033 Sh. 1 QV003A,B,C,D,E,F 3 oc 6.2-84 (3) D-175033 Sh. 2 Q-N12V001A-A,B-B 4 oc 6.2-84 (3) D-175033 Sh. 2 HV3234A,B 5 oc 6.2-84 (3) D-175033 Sh. 1 PV3371A,B,C 6 oc 6.2-84 (4) D-170117 Sh. 4 Q-N21V001A-B, B-B,C-B 7 oc 6.2-84 (4) D-175007 V0011A,B,C 8 oc 6.2-87 (25) D-175071 Sh. 1 7614A,B,C D-205071 Sh. 1 9 oc 6.2-88 (32) D-175009 Sh. 2 HV3328, HV3329, D-205009 Sh. 2 HV3330 10 oc 6.2-84 (4) D-175000 Sh. 1 QV001A,B,C

Note: Item numbers correlate with those on Table 6.2-32.

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-41 CONTAINMENT PRESSURE/TEMPERATURE FOR 600 gal/min SERVICE WATER FLOW, 0.003 FOULING FACTOR Power Uprate Case Peak Pressure (psia) Time (s) Peak Temp. (°F)

Time (s) MSLB CASE 1, P0 = 0.0 58.3 1811 368 60.1 MSLB CASE 1, P0 = -1.5 56.6 1811 383 100.1 MSLB CASE 2 55.7 1811 355 150.1 MSLB CASE 3, P0 = 0.0 55.4 1811 362 170.1 MSLB CASE 3, P0 = -1.5 53.7 1811 370 195.1 MSLB CASE 4 59.9 1821 365 205.1 MSLB CASE 5 59.9 1811 324 70.1 MSLB CASE 6 57.3 1811 331 195.1 MSLB CASE 7 56.7 1811 354 215.1 MSLB CASE 8 61.6 1801 363 200.4 MSLB CASE 9, P0 = 0.0 63.0 400.7 294 75.1 MSLB CASE 9, P0 = 3.0 67.0 400.7 288 80.0 MSLB CASE 10 58.6 1891 313 260.8 MSLB CASE 11, P0 = 0.0 57.3 1831 342 300.8 MSLB CASE 11, P0 = -1.5 56.0 1832 347 340.8 MSLB CASE 12, P0 = 0.0 63.3 1811 359 180.1 MSLB CASE 12, P0 = 3.0 67.1 1811 347 165.1 MSLB CASE 13 61.0 380.7 273 380.7 MSLB CASE 14 43.1 1801 262 760.8 MSLB CASE 15 33.9 2001 302 290.8 MSLB CASE 16 45.3 1331 324 260.8 RSG Case MSLB CASE 1, P0 = 3.0 59.4 1832 351 87.2 MSLB CASE 1, P0 = -1.5 52.6 1828 367 92.2 MSLB CASE 8, P0 = 3.0 62.2 1498 330 192 MSLB CASE 8, P0 = -1.5 57.2 1503 347 157 MSLB CASE 9, P0 = 3.0 64.5 482 347 87.2 MSLB CASE 9, P0 = -1.5 59.2 1828 363 57.2 MSLB CASE 12, P0 = 3.0 65.4 1518 331 162 MSLB CASE 12, P0 = -1.5 60.3 1518 347 132 MSLB CASE 13, P0 = 3.0 66.7 572 342 87.2 MSLB CASE 13, P0 = -1.5 61.3 573 359 57.2 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-42 DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS PRINCIPLE PARAMETERS DURING REFLOOD FLOODING CORE DOWNCOMER INJECTION TIME TEMP RATE CARRYOVER HEIGHT HEIGHT FLOW TOTAL ACCUMULATOR SPILL ENTHALPY (seconds) (degree F) (in/sec) FRACTION (ft) (ft) FRACTION (Pounds Mass per Second) (Btu/lbm) 21.6 181.8 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.3 179.9 26.292 .000 .67 1.47 .000 7724.4 7724.4 .0 89.50 22.5 179.0 29.720 .000 1.02 1.55 .000 7645.9 7645.9 .0 89.50 23.6 178.4 2.675 .300 1.50 4.42 .404 7191.8 7191.8 .0 89.50 24.6 178.5 2.516 .424 1.64 7.02 .440 6874.3 6874.3 .0 89.50 28.2 178.9 3.819 .628 2.00 15.37 .612 5612.0 5612.0 .0 89.50 29.7 178.9 4.533 .672 2.19 15.62 .673 4898.1 4898.1 .0 89.50 30.7 178.9 4.340 .689 2.30 15.62 .670 4739.6 4739.6 .0 89.50 31.7 179.0 4.378 .702 2.42 15.62 .677 4924.6 4469.4 .0 89.36 32.5 179.1 4.269 .709 2.51 15.62 .675 4823.5 4366.1 .0 89.36 37.9 180.0 3.765 .730 3.00 15.62 .659 4236.1 3767.1 .0 89.33 44.2 181.8 3.391 .737 3.50 15.62 .641 3699.8 3221.1 .0 89.31 51.3 184.3 3.083 .739 4.00 15.62 .622 3212.0 2725.5 .0 89.27 52.7 184.8 2.705 .734 4.09 15.62 .579 2434.2 1937.4 .0 89.19 53.7 185.2 3.222 .742 4.15 15.59 .634 487.1 .0 .0 88.00 54.7 185.6 3.274 .742 4.22 15.47 .637 480.7 .0 .0 88.00 59.7 188.0 3.086 .742 4.57 14.88 .633 484.3 .0 .0 88.00 66.5 191.9 2.851 .740 5.00 14.22 .627 488.5 .0 .0 88.00 75.7 198.1 2.575 .738 5.54 13.55 .618 493.2 .0 .0 88.00 84.2 204.2 2.354 .736 6.00 13.13 .608 496.5 .0 .0 88.00 94.7 212.0 2.128 .733 6.52 12.83 .595 499.4 .0 .0 88.00 105.3 219.9 1.947 .732 7.00 12.72 .582 501.5 .0 .0 88.00 118.7 229.5 1.780 .731 7.56 12.80 .567 503.2 .0 .0 88.00 130.1 236.6 1.683 .731 8.00 13.00 .556 504.1 .0 .0 88.00 144.7 244.5 1.605 .733 8.54 13.37 .547 504.9 .0 .0 88.00 158.0 250.7 1.563 .735 9.00 13.79 .543 505.2 .0 .0 88.00 172.7 256.8 1.538 .738 9.50 14.29 .541 505.4 .0 .0 88.00 180.7 259.9 1.531 .741 9.77 14.57 .541 505.5 .0 .0 88.00 187.7 262.4 1.535 .743 10.00 14.82 .543 505.4 .0 .0 88.00

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-43 (SHEET 1 of 2) DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 187.8 174.1 216.5 334.1 104.8 192.8 173.9 216.3 334.2 104.6 197.8 172.9 215.1 335.2 104.7 202.8 172.9 215.0 335.2 104.5 207.8 172.1 214.1 336.0 104.5 212.8 171.4 213.2 336.7 104.5 217.8 171.4 213.2 336.7 104.3 222.8 170.7 212.3 337.4 104.3 227.8 169.9 211.4 338.2 104.3 232.8 169.9 211.3 338.2 104.2 237.8 169.2 210.4 339.0 104.2 242.8 168.4 209.4 339.7 104.2 247.8 168.3 209.4 339.8 104.0 252.8 167.6 208.4 340.6 104.0 257.8 166.8 207.4 341.3 104.0 262.8 166.7 207.3 341.4 103.9 267.8 165.9 206.3 342.2 103.9 272.8 165.8 206.2 342.3 103.7 277.8 165.0 205.2 343.1 103.7 282.8 164.1 204.2 344.0 103.8 287.8 164.0 204.0 344.1 103.6 292.8 163.2 202.9 345.0 103.6 297.8 163.0 202.7 345.1 103.5 302.8 162.1 201.7 346.0 103.5 307.8 161.9 201.4 346.2 103.3 312.8 161.1 200.3 347.1 103.4 317.8 160.2 199.2 347.9 103.4 322.8 159.9 198.9 348.2 103.3 327.8 159.0 197.8 349.1 103.3 332.8 158.7 197.4 349.4 103.2 337.8 158.4 197.1 349.7 103.1 342.8 157.5 195.9 350.6 103.1 347.8 157.2 195.5 350.9 103.0 352.8 162.9 202.6 345.2 103.8 357.8 161.8 201.3 346.3 103.9 362.8 161.3 200.6 346.8 103.8 367.8 160.8 200.0 347.3 103.8 372.8 160.2 199.3 347.9 103.7 377.8 159.6 198.5 348.5 103.6 382.8 159.0 197.7 349.1 103.6 387.8 158.3 196.9 349.8 103.6 392.8 157.6 196.1 350.5 103.5 397.8 156.9 195.2 351.2 103.5 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-43 (SHEET 2 of 2) Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 402.8 156.2 194.3 351.9 103.5 407.8 156.1 194.1 352.0 103.3 412.8 155.4 193.2 352.8 103.3 417.8 154.6 192.3 353.5 103.3 422.8 154.3 191.9 353.8 103.1 427.8 153.4 190.8 354.7 103.1 432.8 153.0 190.3 355.1 103.0 437.8 152.0 189.1 356.1 103.1 442.8 151.5 188.4 356.7 103.0 447.8 150.8 187.6 357.3 103.0 452.8 150.1 186.7 358.0 102.9 457.8 149.7 186.2 358.4 102.8 462.8 148.8 185.1 359.3 102.8 467.8 148.2 184.3 359.9 102.8 472.8 147.5 183.4 360.7 102.7 477.8 147.0 182.8 361.1 102.6 482.8 145.9 181.4 362.2 102.7 487.8 152.0 189.0 356.2 103.3 492.8 151.4 188.3 356.7 103.2 497.8 150.8 187.6 357.3 103.1 502.8 149.9 186.5 358.2 103.1 507.8 149.3 185.7 358.8 103.1 512.8 148.4 184.6 359.7 103.0 517.8 147.6 183.6 360.5 103.0 522.8 72.3 90.0 435.8 123.2 711.2 72.3 90.0 435.8 123.2 711.3 76.6 94.4 431.5 119.3 712.8 76.6 94.4 431.5 119.2 1243.3 76.6 94.4 431.5 119.2 1243.4 67.3 77.4 440.8 45.3 2139.0 59.0 67.9 449.1 46.8 2139.1 59.0 67.9 9.8 8.2 2319.0 58.1 66.9 10.7 8.3 2319.1 58.1 66.9 475.8 95.8 3600.0 51.8 59.7 482.1 96.9 3600.1 42.6 49.0 491.3 86.0 10000.0 31.0 35.6 502.9 88.1 100000.0 16.6 19.1 517.3 90.6 1000000.0 7.1 8.2 526.8 92.3 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-44 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MINIMUM SAFEGUARDS TIME (SECONDS) .00 21.60 21.60 187.74 711.31 1243.32 3600.00 MASS (THOUSAND LBM) INITIAL IN RCS AND ACC 620.40 620.40 620.40 620.40 620.40 620.40 620.40 ADDED MASS PUMPED INJECTION .00 .00 .00 77.84 343.84 614.17 1765.59 TOTAL ADDED .00 .00 .00 77.84 343.84 614.17 1765.59 *** TOTAL AVAILABLE*** 620.40 620.40 620.40 698.24 964.24 1234.57 2385.99 DISTRIBUTION REACTOR CO0LANT 417.47 48.24 68.95 130.25 130.25 130.25 130.25 ACCUMULATOR 202.93 155.44 134.73 .00 .00 .00 .00 TOTAL CONTENTS 620.40 203.68 203.68 130.25 130.25 130.25 130.25 EFFLUENT BREAK FLOW .00 416.71 416.71 567.97 833.98 1104.30 2255.73 ECCS SPILL .00 .00 .00 .00 .00 .00 .00 TOTAL EFFLUENT .00 416.71 416.71 567.97 833.98 1104.30 2255.73 *** TOTAL ACCOUNTABLE *** 620.40 620.39 620.39 698.22 964.23 1234.55 2385.97 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-45 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE MINIMUM SAFEGUARDS Time (Seconds) .0021.60 21.60 187.74 711.31 1243.32 3600.00 Energy (Million Btu) Initial Energy In RCS, ACC, S. Gen 675.98675.98 675.98 675.98 675.98 675.98 675.98 Added Energy Pumped Injection .00.00 .00 6.85 30.26 54.05 214.96 Decay Heat .005.39 5.39 21.62 59.57 91.60 203.05 Heat from Secondary .00-5.74 -5.74 -5.74 -3.77 -2.20 -2.20 Total Added .00-.35 -.35 22.74 86.06 143.45 415.81 ***TOTAL AVAILABLE*** 675.98675.63 675.63 698.72 762.04 819.43 1091.80 Distribution Reactor Coolant 244.8010.56 12.41 33.81 33.81 33.81 33.81 Accummulator 18.1613.91 12.06 .00 .00 .00 .00 Core Stored 18.939.53 9.53 4.05 3.90 3.68 2.71 Primary Metal 120.89114.27 114.27 91.33 64.17 53.37 39.48 Secondary Metal 76.0175.71 75.71 68.26 51.06 39.88 29.57 Steam Generator 197.20196.23 196.23 173.56 127.03 99.38 73.89 Total Contents 675.98420.21 420.21 371.01 279.98 230.12 179.45 Effluent Break Flow .00254.95 254.95 321.10 475.47 575.84 900.23 ECCS Spill .00.00 .00 .00 .00 .00 .00 Total Effluent .00254.95 254.95 321.10 475.47 575.84 900.23 ***TOTAL ACCOUNTABLE*** 675.98675.16 675.16 692.11 755.44 805.95 1079.69 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-46 (SHEET 1 OF 5) DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 21.6 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 .0 .0 .0 .0 22.5 .0 .0 .0 .0 22.6 114.5 135.1 .0 .0 22.7 47.0 55.5 .0 .0 22.8 41.3 48.8 .0 .0 22.9 47.3 55.8 .0 .0 23.0 53.8 63.5 .0 .0 23.1 60.0 70.8 .0 .0 23.2 65.7 77.6 .0 .0 23.3 71.2 84.0 .0 .0 23.4 76.3 90.1 .0 .0 23.5 81.2 95.8 .0 .0 23.5 82.4 97.2 .0 .0 23.6 85.9 101.4 .0 .0 23.7 90.3 106.6 .0 .0 23.8 94.7 111.7 .0 .0 23.9 98.8 116.6 .0 .0 24.0 102.8 121.4 .0 .0 24.1 106.7 126.0 .0 .0 24.2 110.5 130.4 .0 .0 24.3 114.1 134.7 .0 .0 24.4 117.7 138.9 .0 .0 24.5 121.1 143.0 .0 .0 24.6 124.5 147.0 .0 .0 25.6 154.3 182.2 .0 .0 26.6 394.1 466.9 3497.1 447.9 27.4 478.5 567.7 4297.3 574.8 27.7 477.9 567.0 4288.2 576.4 28.7 466.9 553.9 4191.1 567.4 29.7 454.6 539.3 4080.7 556.0 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-46 (SHEET 2 OF 5)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 30.7 442.5 524.8 3969.8 544.3 31.4 491.3 583.0 4465.5 589.1 31.7 486.5 577.4 4413.9 586.9 32.7 476.0 564.8 4321.5 576.4 33.7 465.8 552.6 4229.1 566.3 34.7 456.0 540.9 4139.8 556.5 35.7 446.6 529.7 4053.7 547.0 36.5 439.4 521.1 3987.0 539.6 36.7 437.7 519.0 3970.6 537.8 37.7 429.1 508.7 3890.6 529.0 38.7 420.9 499.0 3813.4 520.5 39.7 413.0 489.6 3738.8 512.2 40.7 405.5 480.6 3666.7 504.3 41.7 398.3 471.9 3597.1 496.5 42.5 392.7 465.3 3542.9 490.5 42.7 391.3 463.6 3529.6 489.0 43.7 384.6 455.6 3464.2 481.8 44.7 378.1 447.9 3400.8 474.7 45.7 371.9 440.5 3339.2 467.9 46.7 365.8 433.3 3279.4 461.2 47.7 360.0 426.4 3221.3 454.7 48.7 354.4 419.6 3164.7 448.4 49.1 352.2 417.0 3142.5 445.9 49.7 348.9 413.1 3109.6 442.2 50.7 343.6 406.8 3055.8 436.2 51.7 338.4 400.7 3003.4 430.3 52.7 333.4 394.7 2952.3 424.6 53.7 328.6 388.9 2902.4 418.9 54.7 323.8 383.3 2853.5 413.4 55.7 183.9 217.2 647.8 157.9 56.7 183.4 216.7 648.5 157.7 57.7 183.0 216.2 649.4 157.6 58.7 182.5 215.7 650.3 157.4 59.7 182.1 215.2 651.2 157.2 60.7 181.7 214.6 652.1 157.1 61.7 181.3 214.1 653.1 156.9 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-46 (SHEET 3 OF 5)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 62.7 180.8 213.6 654.0 156.7 63.7 180.4 213.1 654.9 156.6 64.7 180.0 212.6 655.8 156.4 65.7 179.5 212.1 656.7 156.3 66.6 179.2 211.7 657.6 156.1 66.7 179.1 211.6 657.7 156.1 67.7 178.7 211.1 658.6 155.9 68.7 178.3 210.6 659.5 155.8 69.7 177.9 210.1 660.4 155.6 70.7 177.4 209.6 661.3 155.5 71.7 177.0 209.1 662.2 155.3 72.7 176.6 208.6 663.1 155.2 73.7 176.2 208.1 664.0 155.0 74.7 175.7 207.6 665.0 154.8 75.7 175.3 207.1 665.9 154.7 76.7 174.9 206.6 666.8 154.5 77.7 174.5 206.1 667.7 154.4 78.7 174.1 205.6 668.6 154.2 79.7 173.6 205.1 669.5 154.1 80.7 173.2 204.6 670.5 153.9 81.7 172.8 204.1 671.4 153.8 82.7 172.4 203.6 672.3 153.6 84.7 171.5 202.6 674.2 153.3 86.7 170.7 201.6 676.0 153.0 87.5 170.3 201.2 676.8 152.9 88.7 169.8 200.6 677.9 152.7 90.7 169.0 199.6 679.8 152.4 92.7 168.1 198.6 681.6 152.1 94.7 167.3 197.6 683.5 151.8 96.7 166.4 196.6 685.4 151.5 98.7 165.5 195.5 687.3 151.2 100.7 164.7 194.5 689.2 150.9 102.7 163.8 193.5 691.1 150.6 104.7 162.9 192.4 693.0 150.3 106.7 162.1 191.4 694.9 150.0 108.7 161.2 190.4 696.8 149.7 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-46 (SHEET 4 OF 5)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 110.1 160.6 189.6 698.1 149.5 110.7 160.3 189.3 698.7 149.4 112.7 159.4 188.3 700.6 149.1 114.7 158.5 187.2 702.5 148.8 116.7 157.6 186.1 704.3 148.5 118.7 156.7 185.1 706.2 148.2 120.7 155.8 184.0 708.1 147.9 122.7 154.9 183.0 709.9 147.6 124.7 154.0 181.9 711.8 147.3 126.7 153.1 180.8 713.7 147.0 128.7 152.2 179.7 715.5 146.7 130.7 151.3 178.7 717.4 146.4 132.7 150.4 177.6 719.2 146.1 134.7 149.4 176.5 721.1 145.8 136.7 148.5 175.4 722.9 145.4 138.7 147.6 174.3 724.7 145.1 140.7 146.7 173.2 726.6 144.8 142.7 145.8 172.1 728.4 144.5 144.7 144.8 171.0 730.2 144.2 146.7 143.9 169.9 732.0 143.9 148.7 143.0 168.8 733.9 143.6 150.7 142.0 167.7 735.7 143.3 152.7 141.1 166.6 737.5 143.0 154.7 140.2 165.5 739.3 142.7 156.7 139.2 164.4 741.2 142.4 158.7 138.3 163.3 743.0 142.0 160.7 137.4 162.3 744.9 141.9 161.9 137.1 161.9 745.6 141.9 162.7 136.9 161.6 746.0 141.9 164.7 136.3 160.9 747.2 141.8 166.7 135.7 160.2 748.3 141.7 168.7 135.1 159.5 749.4 141.7 170.7 134.5 158.9 750.5 141.6 172.7 134.0 158.2 751.6 141.5 174.7 133.4 157.5 752.7 141.4 176.7 132.8 156.8 753.7 141.3 178.7 132.3 156.2 754.8 141.3 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-46 (SHEET 5 OF 5)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 180.7 131.7 155.5 755.9 141.2 182.7 131.1 154.8 756.9 141.1 184.7 130.6 154.2 758.0 141.0 186.7 130.0 153.5 759.1 140.9 188.7 129.5 152.9 760.1 140.8 190.7 128.9 152.2 761.1 140.7 192.1 128.6 151.8 761.9 140.6

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-47 DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS PRINCIPLE PARAMETERS DURING REFLOOD Flooding Core Downcomer Injection Time Temp Rate Carryover Height Height Flow Total Accum Spill Enthalpy Seconds °F in/sec Fraction (ft) (ft) Frac (Pounds Mass per Second) Btu/lbm 21.6 183.1 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.3 181.1 22.834 .000 .53 1.85 .000 7698.9 7698.9 .0 89.50 22.5 179.3 28.240 .000 1.06 1.85 .000 7582.0 7582.0 .0 89.50 23.5 178.4 2.939 .303 1.50 5.23 .418 7175.6 7175.6 .0 89.50 24.5 178.4 2.797 .436 1.65 8.77 .451 6865.9 6865.9 .0 89.50 27.4 178.3 5.018 .634 2.01 15.62 .679 5382.1 5382.1 .0 89.50 28.7 178.3 4.633 .673 2.19 15.62 .676 5120.7 5120.7 .0 89.50 30.7 178.4 4.259 .702 2.42 15.62 .670 4800.0 4800.0 .0 89.50 31.4 178.5 4.556 .710 2.50 15.62 .691 5360.3 4479.6 .0 89.25 36.5 179.7 4.036 .732 3.00 15.62 .674 4756.1 3854.9 .0 89.22 42.5 181.9 3.653 .739 3.51 15.62 .659 4224.7 3304.3 .0 89.17 49.1 184.9 3.345 .742 4.01 15.62 .644 3755.9 2820.2 .0 89.13 55.7 188.1 2.267 .729 4.45 15.62 .506 986.7 .0 .0 88.00 56.7 188.6 2.261 .729 4.50 15.62 .506 986.7 .0 .0 88.00 66.6 194.5 2.202 .730 5.00 15.62 .506 986.7 .0 .0 88.00 77.7 202.9 2.137 .732 5.54 15.62 .507 986.7 .0 .0 88.00 87.5 211.2 2.080 .733 6.00 15.62 .507 986.7 .0 .0 88.00 98.7 221.1 2.013 .736 6.51 15.62 .508 986.8 .0 .0 88.00 110.1 230.4 1.945 .738 7.00 15.62 .508 986.8 .0 0 88.00 122.7 239.2 1.870 .740 7.53 15.62 .508 986.9 .0 .0 88.00 134.7 246.4 1.799 .742 8.00 15.62 .509 987.0 .0 .0 88.00 148.7 253.6 1.717 .744 8.53 15.62 .508 987.1 .0 .0 88.00 161.9 259.5 1.642 .745 9.00 15.62 .509 987.2 .0 .0 88.00 176.7 265.1 1.577 .748 9.50 15.62 .511 987.1 .0 .0 88.00 192.1 270.2 1.512 .750 10.00 15.62 .515 987.0 .0 .0 88.00

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-48 (SHEET 1 OF 3) DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 192.2 148.6 185.4 841.4 154.6 197.2 148.7 185.5 841.3 154.3 202.2 148.9 185.8 841.1 154.1 207.2 148.4 185.2 841.6 154.0 212.2 148.7 185.6 841.3 153.7 217.2 148.2 185.0 841.8 153.7 222.2 148.5 185.3 841.5 153.4 227.2 148.0 184.7 842.0 153.3 232.2 148.3 185.1 841.7 153.0 237.2 147.8 184.4 842.2 153.0 242.2 148.1 184.8 841.9 152.7 247.2 147.6 184.1 842.4 152.6 252.2 147.8 184.4 842.2 152.4 257.2 148.1 184.7 841.9 152.1 262.2 147.5 184.1 842.5 152.0 267.2 147.8 184.4 842.2 151.8 272.2 147.2 183.7 842.8 151.7 277.2 147.4 184.0 842.6 151.5 282.2 147.6 184.2 842.4 151.2 287.2 147.1 183.5 842.9 151.1 292.2 147.3 183.7 842.7 150.9 297.2 146.7 183.0 843.3 150.8 302.2 146.9 183.2 843.1 150.6 307.2 147.0 183.4 843.0 152.9 312.2 146.4 182.7 843.6 152.9 317.2 146.6 182.9 843.4 152.6 322.2 146.7 183.0 843.3 152.4 327.2 146.8 183.2 843.2 152.1 332.2 146.2 182.4 843.8 152.1 337.2 146.3 182.5 843.7 151.8 342.2 146.4 182.6 843.6 151.6 347.2 146.4 182.7 843.6 151.3 352.2 145.8 181.9 844.2 151.3 357.2 145.8 181.9 844.2 151.1 362.2 145.8 182.0 844.2 150.8 367.2 145.9 182.0 844.1 150.6 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-48 (SHEET 2 OF 3)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 372.2 145.9 182.0 844.1 150.4 377.2 145.8 182.0 844.2 150.2 382.2 145.1 181.1 844.9 150.1 387.2 145.1 181.0 844.9 149.9 392.2 145.0 180.9 845.0 149.7 397.2 144.9 180.8 845.1 149.5 402.2 144.9 180.8 845.1 149.3 407.2 144.9 180.8 845.1 149.1 412.2 144.9 180.8 845.1 148.8 417.2 144.9 180.7 845.1 148.6 422.2 144.8 180.7 845.2 148.4 427.2 144.7 180.6 845.3 148.2 432.2 144.6 180.5 845.4 148.0 437.2 144.5 180.3 845.5 147.8 442.2 144.4 180.1 845.6 147.6 447.2 144.8 180.7 845.2 147.3 452.2 144.6 180.5 845.4 147.1 457.2 144.4 180.2 845.6 146.9 462.2 144.2 179.9 845.8 149.2 467.2 144.5 180.3 845.5 148.9 472.2 144.2 179.9 845.8 148.7 477.2 144.4 180.2 845.6 148.4 482.2 144.0 179.7 846.0 148.2 487.2 144.2 179.9 845.8 148.0 492.2 144.3 180.0 845.7 147.7 497.2 143.8 179.4 846.2 147.6 502.2 143.8 179.4 846.2 147.3 507.2 143.7 179.3 846.3 147.1 512.2 143.6 179.2 846.4 146.9 517.2 143.9 179.6 846.1 146.5 522.2 143.7 179.2 846.3 146.4 527.2 143.8 179.4 846.2 146.1 532.2 143.4 178.9 846.6 145.9 537.2 143.3 178.9 846.7 145.7 542.2 143.7 179.3 846.3 145.3 547.2 143.4 178.9 846.6 147.5 552.2 143.4 178.9 846.6 147.2 557.2 143.3 178.7 846.7 147.0 562.2 143.4 178.9 846.6 146.7 567.2 143.2 178.7 846.8 146.5 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-48 (SHEET 3 OF 3)

Break Path No. 1 Flow Break Path No. 2 Flow Time Thousand Thousand (seconds) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 572.2 143.2 178.7 846.8 146.2 577.2 142.8 178.2 847.2 146.0 582.2 142.8 178.2 847.2 145.7 587.2 143.0 178.4 847.0 145.4 592.2 142.8 178.2 847.2 145.2 597.2 142.6 177.9 847.4 145.0 602.2 142.5 177.8 847.5 144.7 607.2 69.9 87.2 920.2 164.2 792.8 69.9 87.2 920.2 164.2 792.9 75.0 92.6 915.0 161.7 797.2 74.9 92.5 915.1 161.4 1221.1 74.9 92.5 915.1 161.4 1221.2 67.7 77.9 922.3 87.5 1311.6 66.5 76.6 923.5 87.7 1311.7 66.5 76.6 68.4 12.5 1491.6 64.3 74.0 70.6 12.9 1491.7 64.3 74.0 1095.0 166.0 3600.0 52.0 59.8 1107.4 168.2 3600.1 41.2 47.4 1118.1 159.1 10000.0 30.0 34.5 1129.4 160.7 100000.0 16.0 18.4 1143.3 162.7 1000000.0 6.9 7.9 1152.5 164.0

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-49 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MAXIMUM SAFEGUARDS Time (Seconds) .00 21.60 21.60 192.12 792.85 1221.12 3600.00 Mass (Thousand lbm) Initial In RCS and ACC 620.08 620.08 620.08 620.08 620.08 620.08 620.08 Added Mass Pumped Injection .00 .00 .00 157.40 752.04 1176.03 3734.26 Total Added .00 .00 .00 157.40 752.04 1176.03 3734.26 ***TOTAL AVAILABLE*** 620.08 620.08 620.08 777.48 1372.13 1796.12 4354.35 Distribution Reactor Coolant 416.79 47.97 67.66 118.91 118.91 118.91 118.91 Accummulator 203.30 156.65 136.96 .00 .00 .00 .00 Total Contents 620.08 204.61 204.61 118.91 118.91 118.91 118.91 Effluent Break Flow .00 415.46 415.46 649.72 1244.37 1668.36 4226.58 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 415.46 415.46 649.72 1244.37 1668.36 4226.58 ***TOTAL ACCOUNTABLE*** 620.08 620.07 620.07 768.63 1363.28 1787.27 4345.49

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-50 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE MAXIMUM SAFEGUARDS Time (Seconds) .00 21.60 21.60 192.12 792.85 1221.12 3600.00 Energy (Million Btu) Initial Energy In RCS, ACC, S Gen 673.30 673.30 673.30 673.30 673.30 673.30 673.30 Added Energy Pumped Injection .00 .00 .00 13.85 66.18 103.49 461.34 Decay Heat .00 5.73 5.73 22.33 65.12 90.67 203.33 Heat From Secondary .00 -5.70 -5.70 -5.70 -3.44 -2.25 -2.25 Total Added .00 .03 .03 30.48 127.86 191.91 662.43 ***TOTAL AVAILABLE*** 673.30 673.34 673.34 703.79 801.17 865.22 1335.73 Distribution Reactor Coolant 244.82 10.46 12.22 31.25 31.25 31.25 31.25 Accummulator 18.20 14.02 12.26 .00 .00 .00 .00 Core Stored 18.93 9.68 9.68 4.05 3.90 3.72 2.71 Primary Metal 118.16 111.56 111.56 89.39 61.11 52.35 38.51 Secondary Metal 76.01 75.75 75.75 68.35 49.12 40.00 29.56 Steam Generator 197.20 196.34 196.34 173.69 121.93 99.57 73.76 Total Contents 673.30 417.81 417.81 366.73 267.30 226.88 175.78 Effluent Break Flow .00 255.05 255.05 328.89 525.69 621.47 1145.63 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 255.05 255.05 328.89 525.69 621.47 1145.63 ***TOTAL ACCOUNTABLE*** 673.30 672.86 672.86 695.62 792.99 848.35 1321.40 FNP-FSAR-6 REV 21 5/08 TABLE 6.2-51 DOUBLE-ENDED HOT LEG BREAK SEQUENCE OF EVENTS

Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 3.0 Low Pressurizer Pressure SI Setpoint - 1715 psia reached by SATAN 11.4 Broken Loop Accumulator Begins Injecting Water 11.6 Intact Loop Accumulator Begins Injecting Water 20.0 End of Blowdown Phase

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-52 DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS SEQUENCE OF EVENTS

Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 3.9 Low Pressurizer Pressure SI Setpoint - 1715 psia reached by SATAN 13.4 Broken Loop Accumulator Begins Injecting Water 13.6 Intact Loop Accumulator Begins Injecting Water 21.6 End of Blowdown Phase 30.9 Safety Injection Begins 52.1 Broken Loop Accumulator Water Injection Ends 53.3 Intact Loop Accumulator Water Injection Ends 187.7 End of Reflood Phase 2139.0 Cold Leg Recirculation Begins 1.0E+06 Transient Modeling Terminated

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-53 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS SEQUENCE OF EVENTS

Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 3.9 Low Pressurizer Pressure SI Setpoint - 1715 psia reached by SATAN 13.3 Broken Loop Accumulator Begins Injecting Water 13.5 Intact Loop Accumulator Begins Injecting Water 21.6 End of Blowdown Phase 30.9 Safety Injection Begins 54.7 Broken Loop Accumulator Water Injection Ends 54.9 Intact Loop Accumulator Water Injection Ends 192.1 End of Reflood Phase 1311.6 Cold Leg Recirculation Begins 1.0E+06 Transient Modeling Terminated

FNP-FSAR-6 REV 21 5/08 TABLE 6.2-54 LOCA MASS AND ENERGY RELEASE ANALYSIS CORE DECAY HEAT FRACTION Time (sec) Decay Heat Generation Rate (Btu/Btu) 10 0.053876 15 0.050401 20 0.048018 40 0.042401 60 0.039244 80 0.037065 100 0.035466 150 0.032724 200 0.030936 400 0.027078 600 0.024931 800 0.023389 1000 0.022156 1500 0.019921 2000 0.018315 4000 0.014781 6000 0.013040 8000 0.012000 10000 0.011262 15000 0.010097 20000 0.009350 40000 0.007778 60000 0.006958 80000 0.006424 100000 0.006021 150000 0.005323 200000 0.004847 400000 0.003770 600000 0.003201 800000 0.002834 1000000 0.002580

REV 21 5/08 DEPSGB MINIMUM ESF 1 AC PRESSURE VS. TIME PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-1 REV 21 5/08 RSG DEPSG MINIMUM ESF 1 AC PRESSURE VS. TIME, PO = 3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-2 REV 21 5/08 DEHL MINIMUM ESF, DBA SHORT TERM PRESSURE VS. TIME, PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-3 REV 21 5/08 RSG DEHLG MINIMUM ESF, DBA SHORT PRESSURE VS. TIME, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-4 REV 21 5/08 DECLG MAXIMUM ESF PRESSURE VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-5 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-6 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-6A REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-7 REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 102% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-7A REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.7 ft2 D.E. BREAK 102% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-8 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.7 ft2 D.E. BREAK 102% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-9 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-10 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-10A REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = - PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-11 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 102% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-11A REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.528 ft2 SPLIT 102% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-12 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.528 ft2 SPLIT 102% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-13 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-14 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-15 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-16 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.6 ft2 D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-17 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-18 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 70% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-19 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.47 ft2 SPLIT 70% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-20 REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.47 ft2 SPLIT 70% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-21 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-22 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-22A REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-23 REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK 30% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-23A REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 30% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-24 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.5 ft2 D.E. BREAK 30% POWER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-25 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER,PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-26 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-26A REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-27 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.4 ft2 D.E. BREAK 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-27A REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-28 REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-28A REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = -1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-29 REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE 0.60 ft2 SPLIT 30% POWER, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-29A REV 21 5/08 RSG PRESSURE VERSUS TIME STEAM LINE FULL D.E. BREAK HOT STANDBY PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-30 REV 21 5/08 RSG TEMPERATURE VERSUS TIME STEAM LINE FULL D.E. BREAK HOT STANDBY, PO =-1.5 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-31 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.2 ft2 D.E. BREAK HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-32 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.2 ft2 D.E. BREAK HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-33 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.1 ft2 D.E. BREAK HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-34 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.1 ft2 D.E. BREAK HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-35 REV 21 5/08 PRESSURE VERSUS TIME STEAM LINE 0.30 ft2 SPLIT HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-36 REV 21 5/08 TEMPERATURE VERSUS TIME STEAM LINE 0.30 ft2 SPLIT HOT STANDBY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-37 REV 21 5/08 TS, EQUIPMENT SURFACE TEMPERATURE WITH UCHIDA CONDENSING HEAT TRANSFER AND CONVECTIVE HEAT TRANSFER COEFFICIENT OF 2 BTU/HR-ftq JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-38 REV 21 5/08 DEPSGB MINIMUM ESF 1 AC P/T ANALYSIS LONG-TERM CONTAINMENT PRESSURE VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-39 REV 21 5/08 DEPSGB MINIMUM ESF DBA TEMPERATURE VS. TIME PO = 0 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-40 REV 21 5/08 RSG DEPSG MIN ESF DBA TEMPERATURE VS. TIME PO = 3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-41

REV 21 5/08 CONTAINMENT AIR COOLER DUTY VS. TEMPERATURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-42 (SHEET 1 OF 2)

REV 21 5/08 CONTAINMENT AIR COOLER DUTY VS. TEMPERATURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-42 (SHEET 2 OF 2)

REV 21 5/08 THERMAL HEAT REMOVAL EFFICIENCY OF CONTAINMENT ATMOSPHERE SPRAY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-43 REV 21 5/08 RESIDUAL HEAT EXCHANGER DESIGN DUTY ACCIDENT MODE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-44 REV 21 5/08 MASS &N ENERGY RATE VS TIME FOR DBA JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-45 REV 21 5/08 LOCA BLOWDOWN MASS AND ENERGY RELEASE RATES VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-46 REV 21 5/08 LOCA POST-BLOWDOWN MASS AND ENERGY RELEASE RATES VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-47 REV 21 5/08 DEPSG MIN ESF 1 AD P/T ANALYSIS LONG-TERM CONDENSING HEAT TRANSFER COEFFICIENT (RSG) JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-48

REV 21 5/08 SHORT TERM CONDENSING HEAT TRANSFER COEFFICIENT FOR DBA JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-49 REV 21 5/08 REACTOR CAVITY MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-50 REV 21 5/08 REACTOR CAVITY BLOCK DIAGRAM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-51 REV 21 5/08 TOTAL HORIZONTAL FORCE VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-52

REV 21 5/08 STEAM GENERATOR BLOCK DIAGRAM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-53 REV 21 5/08 STEAM GENERATOR COMPARTMENT C DIFFERENTIAL PRESSURE VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-54 REV 21 5/08 PRESSURIZER COMPARTMENT PRESSURE MODEL (SPRAY LINE BREAK IN LOWER COMPARTMENT) JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-55

REV 21 5/08 PRESSURIZER COMPARTMENT FLOW MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-56 REV 21 5/08 PRESSURIZER COMPARTMENT SPRAY LINE RESULTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-57 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 1 AND 2 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-58 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 3, 4, 5, AND 6 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-59 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 7, 8, 9, AND 10 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-60 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 11, 12, 13, AND 14 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-61 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 15, 16, AND 17 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-62 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 18, 19, 20, AND 21 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-63 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 22, 23, 24, 25, 26, AND 27 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-64 REV 21 5/08 NODE PRESSURES IN COMPARTMENTS 28, 29, 30, 31, 32, 33, AND 34 VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-65 REV 21 5/08 SCHEMATIC OF REFLOOD CODE 19 ELEMENT LOOP MODEL FOR A PUMP SUCTION BREAK JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-66 REV 21 5/08 CORE REFLOOD CORRELATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-67

THIS FIGURE HAS BEEN DELETED PER REV 15.

REV 21 5/08 COMPARISON OF MEASURE AND PREDICTED CARRY OVER RATE FRACTIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-68

THIS FIGURE HAS BEEN DELETED PER REV 15.

REV 21 5/08 INLET WATER TEMPERATURE VS. TIME AFTER END OF BLOWDOWN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-69 REV 21 5/08 VARIATION IN TEMPERATURE RISE, TURNAROUND TIME AND QUENCH TIME WITH RESPECT TO CORE ELEVATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-70 REV 21 5/08 ENERGY BALANCE MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-71 REV 21 5/08 REFLOOD RATE AND CARRYOVER FRACTIONS VS. TIME AFTER END OF BLOWDOWN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-72 REV 21 5/08 FLOW THROUGH BREAK VS. TIME AFTER END OF BLOWDOWN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-73 REV 21 5/08 WATER HEIGHT VS. TIME AFTER END OF BLOWDOWN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-74 REV 21 5/08 POST-REFLOOD LOOP RESISTANCE MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-75 REV 21 5/08 S/G INTERNAL ENERGY VS. TIME AFTER BREAK JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-76 REV 21 5/08 ENERGY DISTRIBUTION VS. TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-77 REV 21 5/08 RSG TEMPERATURE PROFILE THROUGH CONTAINMENT WALL PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-78 REV 21 5/08 RHR HX DUTY VS. TIME RSG, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-79 REV 21 5/08 CONTAINMENT AIR COOLING DUTY VS. TIME RSG, PO = +3 PSIG JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-80 REV 21 5/08 MINIMUM SUMP pH FOLLOWING LOCA VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-81 REV 21 5/08 MINIMUM PARTITION COEFFICIENT IN THE SUMP VERSUS SOLUTION pH JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-82 REV 21 5/08 HYDROGEN GENERATION RATE VS. TIME IN THE LOWER COMPARTMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-83 REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-84 REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-85 REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-86

Note 1: Containment isolation is provided by the blind flange inside containment. Valve outside containment for arrangement 26 is shown for completeness only and is not a containment isolation valve. Note 2: Relief valve shown outside containment for arrangement 24 is applicable to penetration 46 only. The relief valve is classified as a containment isolation valve. REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-87 REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-88 REV 21 5/08 ISOLATION VALVE ARRANGEMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-89 REV 21 5/08 ELECTRIC HYDROGEN RECOMBINER JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-90 REV 21 5/08 ELECTRIC HYDROGEN RECOMBINER SCHEMATIC DIAGRAM (TYPICAL OF ONE RECOMBINER) JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-91 REV 21 5/08 LOWER COMPARTMENT PLAN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-92 REV 21 5/08 SECTION OF LOWER REACTOR COMPARTMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-93 REV 21 5/08 CONTAINMENT HYDROGEN CONCENTRATION WITH ONE ELECTRIC RECOMBINER STARTED ONE DAY AFTER A LOCA JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-94 REV 21 5/08 HYDROGEN CONCENTRATION AS A FUNCTION OF TIME IN CONTAINMENT PURGE MODE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-95 REV 21 5/08 VOLUME PERCENT HYDROGEN VS. TIME IN THE UPPER CONTAINMENT (UNMIXED), OUTER PERIPHERY (UNMIXED) AND BULK CONTAINMENT (MIXED) JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-96 REV 21 5/08 VOLUME PERCENT HYDROGEN VS. TIME IN THE LOWER COMPARTMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-97

REV 21 5/08 HYDROGEN GENERATION RATE VS. TIME IN OUTER PERIPHERY AND OVERALL CONTAINMENT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.2-98

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-2 ECCS RELIEF VALVE DATA Fluid Inlet Set Back Psig Fluid Temp. °F Pressure Pressure Build- Description Discharged Normal Relieving (psig) Constant Up Capacity N2 supply to N2 gas 120 120 700 atm. 0 1500 sf3/min accumulators RHR pumps Water 250 350 600 3 50 20 gal/min discharge SI line Accumulator to N2 gas 120 120 700 0 0 1500 sf3/min containment

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-3 SEQUENCE OF CHANGEOVER OPERATION FROM INJECTION TO RECIRCULATION

(This table has been deleted.)

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-4 (SHEET 1 OF 2) TIME ANALYSIS FOR ECCS INJECTION/RECIRCULATION SWITCHOVER Time Step (s) Volume Flow Rate From for Constant Remaining In RWST (gpm) RWST Flow RWST Step Time (s) During Step Rate (gal) 1. Low-level switchover setpoint 0 0 135,716 2. Verify SI reset 10 13,400 3. Direct verification of PRF status 20 13,400 4. Verify CCW flow to RHR heat exchangers 60 13,400 5. Establish only one charging pump in each train 70 13,400 6. Direct verification of recirculation disconnects 80 13,400 7. Stop both RHR pumps 90 13,400 90 115,616 8. Close RWST supply to 'A' RHR pump suction 110 9,000 9. Align containment sump to 'A' RHR pump suction 150 9,000 10. Close RHR to RCS hot legs cross-connect 170 9,000 11. Start 'A' RHR pump 180 9,000 12. Verify 'A' Train LHSI flow 185 9,000 13. Close RWST supply to 'B' RHR pump suction 205 9,000 115 98,366 14. Align containment sump to 'B' RHR pump 245 7,600 15. Close RHR to RCS hot legs cross-connect 265 7,600 FNP-FSAR-6 REV 21 5/08 TABLE 6.3-4 (SHEET 2 OF 2)

Time Step (s) Volume Flow Rate From for Constant Remaining In RWST (gpm) RWST Flow RWST Step (continued) Time (s) During Step Rate (gal) 16. Start 'B' RHR pump 275 7,600 17. Verify 'B' Train LHSI flow 280 7,600 18. IF 'A' RHR pump started, THEN align charging pump suction header isolation valves based on 'B' charging pump status 360 7,600 19. Open RHR supply to 'A' train charging pump suction 380 7,600 175 76,200 20. Verify VCT level 385 6,700 5 75,642 21. Close 'A' train RWST to charging pump header valve 405 7,150 22. IF 'B' RHR pump started, THEN align charging pump suction header isolation valves based on 'B' charging pump status 410 7,150 23. Open RHR supply to 'B' train charging pump suction 430 7,150 45 70,280 24. Verify VCT level 435 6,700 25. Close 'B' train RWST to charging pump header valve 455 6,700 26. Check one charging pump in each train 460 6,700 27. Open charging pump recirculation to RCS cold legs valve 480 6,700 28. Align charging pump discharge header isolation valves based on 'B' charging pump status 560 6,700 29. Verify SI flow 565 6,700 135 55,207

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-6 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING

Number of charging pumps operable 2 Number of residual heat removal pumps operable 2 Number of residual heat exchangers operable 2 Minimum refueling water storage tank volume (gal) 471,000 Boron concentration in refueling water storage 2,300 to tanks (ppm) 2,500 Boron concentration in accumulator (ppm) 2,200 to 2,500 Number of accumulators 3 Normal operating accumulator pressure (psig) band 601 to 649 Nominal accumulator water volume (ft3) 1025(a)

a. This value includes the liquid volume in the tank plus the liquid volume in the piping measured from the tank to the second check valve. The second check valve is defined as the second check valve from the tank or the first check valve from the reactor coolant system (RCS) loop.

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-7 (SHEET 1 OF 2) SINGLE ACTIVE FAILURE ANALYSIS FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS SHORT TERM PHASE Component Malfunction Comments Accumulator Deliver to broken loop Totally passive system with one accumulator per loop. Evaluation based on one spilling accumulator Pump Centrifugal charging Fails to start Three provided. Evaluation based on operation of one Residual heat removal Fails to start Two provided. Evaluation based on operation of one Automatically Operated Valves Injection line isolation Fails to open Two parallel lines; one valve in either line required to open Residual heat removal pumps Fails to close Check valve in series with one gate valve; suction line to refueling operation of only one valve required water storage tank Centrifugal charging pumps a. Suction line to refueling Fails to open Two parallel lines; only one valve in either water storage tank line is required to open b. Discharge line to the Fails to close Two valves in series; only one valve required normal charging path to close c. Miniflow line Fails to close Two valves in series; only one valve required to close d. Suction from volume Fails to close Two valves in series; only one valve required control tank to close FNP-FSAR-6 REV 21 5/08 TABLE 6.3-7 (SHEET 2 OF 2) LONG TERM PHASE Component Malfunction Comments Valves operated from control room for recirculation Containment sump recirculation isolation Fails to open Two lines parallel; two valves in either lines are required to open Residual heat removal pumps Fails to close Check valve in series with one gate valve; suction line to refueling water storage tank operation of either the check or the gate valve required Centrifugal charging pump suction line to refueling Fails to close Check valve in series with two parallel water storage tank gate valves. Operation of either the check valve or the gate valves required Centrifugal charging pump suction line at discharge Fails to open Separate and independent high head injection of residual heat exchanger path taking suction from discharge of the other residual heat exchanger Pumps Residual heat removal pump Fails to start Two provided. Evaluation based on operation of one Centrifugal charging pump Fails to operate Same as short term phase Failure of Train B power during switchover from cold leg recirculation to simultaneous hot and cold leg recirculation results in:

  • Residual heat removal discharge valve to hot legs (MOV 8889) Fails to open Align RHR pumps to cold legs, Train A charging pump to hot legs, and use Train A/Train B Power
  • Centrifugal charging pump discharge valve to cold legs (MOV 8803B) Fails to close Transfer Switch (Q1/2R18B037) to apply Train A power to close MOV 8803B

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-8 MAXIMUM POTENTIAL RECIRCULATION LOOP LEAKAGE EXTERNAL TO CONTAINMENT Leakage to Leakage to Type of Leakage Control and Unit Atmosphere Drain Tank Item Leakage Rate Used in the Analysis (cm3/h) (cm3/h) Residual heat removal Mechanical seal with leakoff - 0 20 (low head safety injection) 10 cc/hr/seal Charging pumps Same as residual heat removal 0 60 pump(a) Flanges: Pumps Gasket - adjusted to zero 0 0 leakage following any test Valves bonnet to body 10 drops/min/gauge used 2400 0 (larger than 2 in.) (30 cc/hr). Due to leak tight flanges on pumps, no Control valves leakage is assumed to 480 0 atmosphere Heat exchangers 240 0 Valves - stem leakoffs Back seated double packing 0 50 with leakoff - 1 cc/hr in. stem diameter used. (See table 6.3-1.) Miscellaneous small valves Flanged body packed stems - 600 0 1 drop/min used (3 cm3/h). Miscellaneous large valves Double packing 1 cm3/h/in. 40 0 (larger than 2 in.) stem diameter used ________________ a. Seals are acceptance tested to essentially zero leakage. Due to tandem double seal arrangement and the use of water from the refueling water storage tank as a buffer between the seals, no radioactive leakage from the pumps to the atmosphere is expected.

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-9 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS Flow Path Indication of Loss of Flow Path Alternate Flow Path Low head recirculation During cold-leg recirculation: From containment sump to low Accumulation of water in a Via the independent, identical head injection header via the residual heat removal pump low head flow path utilizing the residual heat removal pumps compartment or auxiliary second residual heat exchanger and the residual heat building sump exchangers High-head recirculation During hot-leg recirculation: The high head pumps provide the required redundancy during this period From containment sump to the Accumulation of water in a From containment sump to the high head injection header residual heat removal pump high head injection headers via via residual heat removal compartment or the auxiliary alternate residual heat removal pump, residual heat building sump pump, residual heat exchanger exchanger and the high head and the alternate high head injection pumps charging pump

FNP-FSAR-6 REV 21 5/08 TABLE 6.3-10 EMERGENCY CORE COOLING SYSTEM SHARED FUNCTIONS EVALUATION Normal Operating Component Arrangement Accident Arrangement Refueling water storage Lined up to suction of Lined up to suction of tank residual heat removal pumps centrifugal charging and residual heat removal pumps. Valves for realignment of RWST to charging pumps meet the single failure criteria Centrifugal charging Lined up for charging service Lined up to high head safety pumps injection header. Valves for realignment meet single failure criteria Residual heat removal Lined up to cold legs of Lined up to cold legs of pumps reactor coolant piping reactor coolant piping Residual heat exchangers Lined up for residual heat Lined up for residual heat removal pump operation removal pump operation

REV 21 5/08 RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-1 REV 21 5/08 CHARGING PUMP PERFORMANCE CURVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-2 REV 21 5/08 TYPICAL RHR PUMP CHARACTERISTIC CURVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-3 REV 21 5/08 CONTAINMENT SPRAY PUMP CHARACTERISTIC CURVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6.3-4

HISTORICAL] [The following describes pre-operational testing requirements: 1. Tray Type

a. Removal of all iodines with an efficiency of 95.0 percent for a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> continuous flow at 150°F, 70 percent relative humidity, and 40 ft/min face velocity.

b. Each adsorbing unit (2 elements) is capable of filtering 333 ft3/min of air at a pressure drop not exceeding 1.2 in. wg.

c. Each assembled filter unit will be leak tested by the manufacturer.

d. Each filter will be tested for 5 minutes in airflow of 330 ft3/min containing 20 ppm refrigerant 112. A downstream concentration in excess of 0.2 percent of the upstream concentration will cause rejection of the filter.

2. Bed Type

a. Retaining 99.0 percent minimum of elemental iodine. At relative humidities below 70 percent at 150°F all organic iodines with an

efficiency of 95 percent for the recirculation filter and 99 percent for the pressurization filter.

b. Each assembled filter unit will be leak tested by the manufacturer.

c. Charcoal adsorbers remove 99 percent of a halogenated hydrocarbon refrigerant test gas.]

FNP-FSAR-6 REV 25 4/14 TABLE 6.5-1 AUXILIARY FEEDWATER SYSTEM AUXILIARY FEEDWATER PUMP DATA

MOTOR-DRIVEN PUMPS TURBINE-DRIVEN (DATA PER PUMP) PUMP Type Horizontal- Horizontal- Centrifugal Centrifugal No. of stages 10 7 Design pressure (psig) 1600 1600 Pumping temperature (°F) 95 95 Design flowrate (gal/min) 350 700 Design head (ft) 2845 2835 NPSH required at design (ft) 17 21 Minimum available NPSH (ft) 60 60 Suction pressure range (ft) 45-75 45-75 Shutoff head (ft) 3480 3380 RPM 3600 3960 Bhp required 366 687 Driver horsepower (max) 450 693 Materials: Casing SA-217 Gr. WC 9 SA-217 Gr. WE 9 Impeller SA-296 Gr. CA 15 SA-296 Gr. CA 15 or A-217 Gr. CA 15 or A-217 Gr. CA 15(a) Shaft A-276 Tp 410 HT SA-276 Tp 410 HT TURBINE DRIVE Type Vertical-Single Stage Design pressure (psig) 1250 Design temperature (°F) 572 Steam inlet pressure (psig) Minimum 90 Maximum 1148 Back pressure (psig) 0-10 RPM design/turbine trip 3960/4554 Rated Bhp 687 Governor NEMA Class D Lubrication Forced feed Cooling water Pumped liquid a. Changes are applicable to Unit 2 only.

FNP-FSAR-6 REV 21 5/08 TABLE 6.5-2 (SHEET 1 OF 4) FAILURE ANALYSIS OF AUXILIARY FEEDWATER SYSTEM Component Failure Comments and Consequences Motor-driven auxiliary Fails to start on Two motor-driven pumps are provided. One motor-driven pump feedwater pump automatic signal in conjunction with the turbine-driven pump is sufficient to meet cooldown requirements for all emergency conditions. One motor-driven pump is sufficient to meet all normal cooldown requirements Turbine-driven auxiliary Fails to start on Operation of the two motor-driven pumps will provide feedwater pump automatic signal sufficient flow to meet cooldown requirements for all condi- tions Turbine-driven pump steam Fails to open on Black- Parallel connections are provided to two main steam lines. inlet isolation valve from out signal One of the two valves must open to supply 100 percent of main steam header the turbine steam requirements Steam supply lines to One parallel supply line Check valves installed in each parallel line, upstream of turbine driven pump broken downstream of inlet the common header connection and below the floor of the isolation valve in main main steam and feedwater valve room, prevent blowdown through steam and feedwater the broken line and subsequent loss of steam supply to the valve room turbine drive Condensate supply Loss of normal supply Water can be supplied to all pumps from the service water from condensate storage system. Service water supply is separate and redundant tank Auxiliary feedwater pump Failure of pressure No single failure can prevent the auxiliary feedwater system discharge line boundary resulting in from providing the minimum required flow. Both manual and abnormal leakage motor-operated valves are provided for isolating potential breaks Electrical power supply Failure of power supply Motor-driven pumps are separate and redundant including bus to components power supplies. One motor-driven pump in conjunction with associated with one motor the turbine-driven pump will supply the minimum required driven pump flow for all emergency conditions Motor operated valves in Loss of power All motor operated valves are manually open, fail "as is" pump discharge piping on loss of power, and are closed remote manually Air operated flow control Loss of air or loss of Failure modes presented in table 7.3-10, sheet 2 valves in pump discharge 125-V dc power FNP-FSAR-6 REV 21 5/08 TABLE 6.5-2 (SHEET 2 OF 4)

Component Failure Comments and Consequences Isolation valves Spurious closure of During normal plant operation, these valves are in the MOV 3350A, B, C motor-operated valve open position and the breakers, which supply power to the valves, are opened and locked so that no power is supplied to the valve's motor operator.

Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A.--The Train B motor-driven pump and the turbine-driven valve and the steam generator loss of Train A electrical pump start and delivery flow through the restriction orifices power which limit auxiliary feedwater flow to the faulted steam generator, thus establishing the minimum required flow to two intact steam generators. Closing valve MOV 3764E, which is powered from a Train B electrical power supply, isolates auxiliary feedwater flow from the motor-driven pump to the faulted steam generator. This increases flow to the two intact steam generators, allowing an orderly cooldown to the cold shutdown condition Case 2 - Failure of main feedwater line to steam generator 1B.--Identical to Case 1 above except that valve MOV 3764B, which is powered from a Train B electrical power supply, is closed to isolate auxiliary feedwater flow from the motor-driven pump to the faulted steam generator Case 3 - Failure of main feedwater line to steam generator 1C.--Identical to Case 1 above except that valve MOV 3764C, which is powered from a Train B electrical power supply, is closed to isolate auxiliary feedwater flow from the motor- driven pump to the faulted steam generator Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A.--The Train A motor-driven pump and the turbine-driven valve and the steam generator loss of Train B electrical pump start and deliver flow through the restriction orifices power which limit auxiliary feedwater flow to the faulted steam generator, thus establishing the minimum required flow to the two intact steam generators. Closing valve MOV 3764A, which is powered from a Train A electrical power supply, isolates auxiliary feedwater flow from the motor-driven pump to the faulted steam generator. This increases flow to the two intact steam generators, allowing an orderly cooldown to the cold shutdown condition.

FNP-FSAR-6 REV 21 5/08 TABLE 6.5-2 (SHEET 3 OF 4)

Component Failure Comments and Consequences Case 2 - Failure of main feedwater line to steam generator 1B.--Identical to Case 1 above except that valve MOV 3764D, which is powered from a Train A electrical power supply, is closed to isolate auxiliary feedwater flow from the motor-driven pump to the faulted steam generator.

Case 3 - Failure of main feedwater line to steam generator 1C.--Identical to Case 1 above except that valve MOV 3764F, which is powered from a Train A electrical power supply, is closed to isolate auxiliary feedwater flow from the motor-driven pump to the faulted steam generator.

Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A.--The Train A and Train B motor-driven auxiliary valve and the steam generator loss of the turbine- feedwater pumps start and deliver flow through the driven auxiliary feedwater pump restriction orifices which limit auxiliary feedwater flow to the faulted steam generator, thus establishing the the minimum required flow to the two intact steam generators. Closing either valve MOV 3764A or MOV 3764E isolates motor-driven auxiliary feedwater pump flow to the faulted steam generator. This increases flow to the two intact steam generators, allowing an orderly cooldown to the cold shutdown condition.

Case 2 - Failure of main feedwater line to steam generator 1B.--Identical to Case 1 above except that either valve MOV 3764B or MOV 3764D is closed to isolate auxiliary feedwater flow to the faulted steam generator.

Case 3 - Failure of main feedwater line to steam generator 1C.--Identical to Case 1 above except that either valve MOV 3764C or MOV 3764F is closed to isolate auxiliary feedwater flow to the faulted steam generator.

FNP-FSAR-6 REV 21 5/08 TABLE 6.5-2 (SHEET 4 OF 4)

Component Failure Comments and Consequences Main feedwater line between Failure of a main feedwater Case 1 - Failure of main feedwater line to steam generator the containment isolation line with a simultaneous 1A with a simultaneous spurious closure of either motor- valve and the steam generator spurious closure of a operated valve located in the motor-driven pump discharge motor operated valve in line to steam generator 1B.--The Train A and Train B motor- the pump discharge flow driven pumps and the turbine-driven pump start and deliver path flow through the restriction orifices. The restriction orifices limit flow to the faulted steam generator, thus establishing the minimum required flow to the two intact steam generators. Closing either valve MOV 3764A or MOV 3764E isolates motor-driven pump flow to the faulted steam generator and increases motor-driven pump flow through the open flow path to steam generator 1C, thus allowing an orderly cooldown to the cold shutdown condition. Case 2 - Failure of main feedwater line to steam generator 1A with a simultaneous spurious closure of either motor operated valve located in the motor-driven pump discharge line to steam generator 1C.--Identical to Case 1 above except isolation of the faulted steam generator increases motor-driven pump flow to steam generator 1B.

Note - For all possible combinations of a faulted steam generator and a spurious closure of any one of valves MOV 3764A, B, C, D, E or F, the operator can remote manually isolate the motor-driven pump flow to the faulted steam generator, which increases motor-driven pump flow through the open flow path(s) to the intact steam generators, thus allowing an orderly cooldown to the cold shutdown condition.

FNP-FSAR-6 REV 21 5/08 TABLE 6.5-3 AUXILIARY FEEDWATER SYSTEM MOTOR OPERATED VALVE DATA Motor Control Center Valve Motor Supplying Electricity Number to Valve Valve Position (Ref. drawing D-175007) (Ref. drawing D-177001) After Loss of Power MOV 3209A MCC 1U As is MOV 3209B MCC 1V As is MOV 3210A MCC 1U As is MOV 3210B MCC 1V As is MOV 3216 MCC 1U As is MOV 3350A MCC 1U As is MOV 3350B MCC 1U As is MOV 3350C MCC 1U As is MOV 3764A MCC 1U As is MOV 3764B MCC 1V As is MOV 3764C MCC 1V As is MOV 3764D MCC 1U As is MOV 3764E MCC 1V As is MOV 3764F MCC 1U As is

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6A-i REV 23 5/11 APPENDIX 6A MATERIALS COMPATIBILITY REVIEW TABLE OF CONTENTS Page 6A.1 DEFINITION OF POSTACCIDENT CONTAINMENT ENVIRONMENTAL CONDITIONS .............................................................................. 6A-1 6A.1.1 Design Basis Accident Temperature-Pressure Cycle...................................... 6A-1

6A.1.2 Design Bases Accident Radiation Environment .............................................. 6A-2

6A.1.3 Design Chemical Composition of the Emergency Core Cooling Solution ....... 6A-2

6A.1.4 Trace Composition of Emergency Core Cooling Solution ............................... 6A-3

6A.2 MATERIALS OF CONSTRUCTION IN CONTAINMENT ............................................ 6A-3 6A.3 CORROSION OF METALS OF CONSTRUCTION IN DESIGN BASIS ECC SOLUTION .............................................................................................. 6A-4

6A.4 CORROSION OF METALS OF CONSTRUCTION BY TRACE CONTAMINANTS IN ECC SOLUTION ....................................................................... 6A-6 6A.4.1 Low Temperature of ECC Solution .................................................................. 6A-6

6A.4.2 Low Chloride Concentration of ECC Solution.................................................. 6A-6

6A.5 CORROSION OF ALUMINUM ALLOYS ..................................................................... 6A-7 6A.6 THE NATURE AND BEHAVIOR OF ALUMINUM CORROSION PRODUCTS IN ALKALINE SOLUTION ...................................................................... 6A-7 6A.6.1 Behavior of Circulating Aluminum Corrosion Products .................................... 6A-9 6A.7 EFFECT OF POSSIBLE CHEMICAL REACTIONS ON IODINE REMOVAL CAPABILITY OF THE CONTAINMENT SPRAY SOLUTION ................. 6A-10

6A.8 COMPATIBILITY OF PROTECTIVE COATINGS WITH POSTACCIDENT ENVIRONMENT ........................................................................................................ 6A-11

6A-9 EVALUATION OF THE COMPATIBILITY OF CONCRETE ECC SOLUTION IN THE POSTACCIDENT ENVIRONMENT .............................................................. 6A-11 FNP-FSAR-6A

6A-ii REV 23 5/11 LIST OF TABLES 6A-1 Postaccident Containment Temperature Transient Used in the Material Compatibility Review 6A-2 Review of Sources of Various Elements in Containment and their Effects on Materials of Construction 6A-3 Typical Materials of Construction in the Farley Containment

6A-4 Deleted

6A-5 Corrosion of Aluminum Alloys in Alkaline Sodium Borate Solution

6A-6 Deleted

6A-7 Summary of Aluminum Corrosion Product Solubility Data

6A-8 Concrete Specimen Test Data

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6A-1 REV 23 5/11 APPENDIX 6A MATERIALS COMPATIBILITY REVIEW 6A.1 DEFINITION OF POSTACCIDENT CONTAINMENT ENVIRONMENTAL CONDITIONS An evaluation of the suitability of materials of construction for use in the containment has been performed considering the following:

A. The integrity of the materials of construction of engineered safety features equipment when exposed to postdesign basis accident (DBA) conditions.

B. The effects of corrosion and deterioration products from both engineered safety features (vital equipment) and other (nonvital) equipment, on the integrity and operability of the engineered safety features equipment.

The post DBA environment conditions of temperature, pressure, radiation, and chemical composition are described in the following sections. The time temperature pressure cycle used in the materials evaluation is most conservative, since it considers only partial safeguards operation during the DBA. The spray and core cooling solutions considered herein include both the design chemical compositions and the design chemical compositions contaminated with deterioration products and fission products, which may conceivably be transferred to the solution during recirculation through the various containment safety features systems.

6A.1.1 DESIGN BASIS ACCIDENT TEMPERATURE-PRESSURE CYCLE Containment pressure/temperature versus time responses for the various analyzed breaks are shown in figures 6.2-1 through 6.2-41. These figures represent containment environment conditions during and after a postulated accident considering partial safety features operation:

that is, operation with 1 of the 2 spray pumps, 1 of the 4 containment fans, 1 of the 2 residual heat removal pumps, and 1 of the 3 safety injection pumps.

Table 6A-1 presents the evaluation conditions for Westinghouse supplied material subjected to the containment and the core environment, respectively. For equipment specified by Bechtel and Southern Company Services, Inc., refer to table 3.11-1.

Material evaluations, to be described, were performed, in general, for the time temperature conditions of table 6A-1 or conservatively considering high temperature conditions for longer periods. The basis for each material evaluation is described with the discussion of its particular suitability.

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6A-2 REV 23 5/11 6A.1.2 DESIGN BASIS ACCIDENT RADIATION ENVIRONMENT Evaluation of materials for use in containment included a consideration of the radiation stability requirements for the particular materials application. This evaluation utilized data that were calculated on the basis of a core meltdown and, assuming the following fission product fractional releases, consistent with TID 14844 model:

Noble gases Fractional release 1.0 Halogens Fractional release 0.5 Other isotopes Fractional release 0.01 6A.1.3 DESIGN CHEMICAL COMPOSITION OF THE EMERGENCY CORE COOLING SOLUTION Farley system designs provide for use of alkaline adjusted boric acid solution as the spray and core cooling fluid.

Alkaline Sodium Borate Plant designs that utilize the spray solution for fission product iodine removal, as well as containment cooling include provisions for chemical addition to control pH. For Farley trisodium phosphate (TSP) is added to the containment sump. Boric acid solution, containing 2300 to 2500 ppm boron, is pumped from the refueling water storage tank into the core and to the containment by means of the safety injection system pumps, residual heat removal pumps, and spray pumps. The initial pH of the spilled RCS water and containment spray will be approximately 4.5. Three baskets are located on elevation 105'-6" which contain sufficient TSP so that when their contents dissolve in the water from the RWST, RCS, and accumulators, the resulting containment sump and recirculation (ECCS and spray) systems pH will be between 7.5 and 9.1.

For the purpose of materials evaluation in the design chemistry solution, the following concentration/time relationship was considered:

0 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> pH 4.5 Boron 2500 ppm 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 12 months pH 10 Boron 2500 ppm The solutions are considered aerated through the entire exposure period as in the case of pure boric acid spray solution.

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6A-3 REV 23 5/11 6A.1.4 TRACE COMPOSITION OF EMERGENCY CORE COOLING SOLUTION During spraying and recirculation, the emergency core cooling (ECC) solution will wash over virtually all the exposed components and structures in the reactor containment. The ECC solution is recirculated through a common sump; hence, any contamination deposited in or leached by the solution from the exposed components and structures will be uniformly mixed in the solution.

The materials compatibility discussion includes consideration of the effects of trace elements which are identified as conceivably being present in the ECC solution during recirculation.

To identify the trace elements in containment which may have a deleterious effect on engineered safety features equipment, one must first establish which elements are potentially harmful to the materials of construction of the safety features equipment and second, ascertain the presence of these elements in forms which can be released to the ECC solution following a design basis accident. Table 6A-2 presents a listing of the major periodic group of elements.

Elements known to be harmful to various metals are noted and potential sources of these elements are identified.

The concentration of the trace contaminants in the ECC solution will vary with individual plant construction as well as with the chemical composition of the ECC solution itself.

6A.2 MATERIALS OF CONSTRUCTION IN CONTAINMENT All materials in containment are reviewed from the standpoint of insuring the integrity of equipment of which they are constructed and to insure that deterioration products of some materials do not aggravate the accident condition. In essence, therefore, all materials of construction in the containment must exhibit resistance to the postaccident environment or, at worst, contribute only insignificant quantities of trace contaminants which have been identified as potentially harmful to vital safeguards equipment. Table 6A-3 lists typical material of construction used in the containment. Examples of equipment containing these materials are included in the table.

Corrosion testing, described in section 6A.3, showed that of all the metals tested only aluminum alloys and zinc were found incompatible with the alkaline sodium borate solutions. Aluminum and zinc were observed to corrode at a significant rate, with the generation of hydrogen gas.

Since hydrogen generation can be hazardous to containment integrity a detailed survey was conducted to identify all aluminum and zinc components in containment.

The as-built aluminum inventory present inside the containment is described in drawing A-508597 (Farley Unit 1) and A-508928 (Farley Unit 2). The drawings also include the mass of metal and exposed surface area of each component used in the calculation of hydrogen generated post-LOCA. The 1100- and the 6000-series aluminum alloys are the major types found in containment. This inventory provides some insight into the range of components which are often fabricated from aluminum. All metals of construction in containment, including aluminum, are compatible with unadjusted boric acid solution under DBA conditions.

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6A-4 REV 23 5/11 The total analyzed value of zinc inventory considered in the analysis of post-LOCA hydrogen generation is described below. Ample margin was included for each source of zinc in the analysis with respect to the zinc inventory for future addition of zinc inside containment.

Zinc Inventory:

Item Surface Area (ft2)

Zinc Based Paint 298,216 Galvanized Carbon Steel 125,864 Cable Trays 44,328 Since the corrosion rate of zinc is considerably lower than the aluminum, the rate of mass depletion of zinc due to corrosion is lower. Therefore, the thickness and mass of the zinc inventory is not considered in the post-LOCA hydrogen generation analysis.

6A.3 CORROSION OF METALS OF CONSTRUCTION IN DESIGN BASIS ECC SOLUTION Emergency core cooling components are austenitic stainless steel and, hence, are quite corrosion resistant to the alkaline sodium borate solution as demonstrated by corrosion tests reported in WCAP-7153(1). The general corrosion rate, for Type 304 and 316 stainless steels, was found to be 0.01 mils/months in pH 10 solution at 200°F. Data on corrosion rates of these materials in the alkaline sodium borate solution have been reported by ORNL(2, 3) to confirm the low values.

Extensive testing was also performed on other metals of construction found in the reactor containment. Testing was performed on these materials to ascertain their compatibility with the spray solution at design post-accident conditions and to evaluate the extent of deterioration product formation, if any, from these materials.

Metals tested included zircaloy, Inconel, aluminum alloys, cupronickel alloys, carbon steel, galvanized carbon steel and copper. The results of the corrosion testing of these materials are reported in detail in reference 1. Of the materials tested, only aluminum and zinc were found to be incompatible with the alkaline sodium borate solution. Aluminum corrosion is discussed in section 6A.5. The following is a summary of the corrosion data obtained on various materials of construction exposed for several weeks in aerated alkaline (pH 9.0-9.3) sodium borate solution at 200°F. The exposure condition is considered conservative since the test temperature (200°F) is considerably higher than the long term design basis accident temperature (152°F),

and the pH bounds the long term design basis accident pH. Corrosion of zinc in post-LOCA environment is discussed in section 15.4.1.6.2.

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6A-5 REV 23 5/11 Maximum Observed Corrosion Rate Material mil/month Carbon Steel 0.003 Zr-4 0.004

Inconel 718 0.003 Copper 0.015

90 - 10 Cu-Ni 0.02 70 - 30 Cu-Ni 0.006 Galvanized carbon steel 0.051 Brass 0.01 Tests conducted at ORNL(2, 3) also have verified the compatibility of various materials of construction with alkaline sodium borate solution. In tests conducted at 284°F, 212°F, and 130°F, stainless steel, Inconel, cupronickel, Monel and zircaloy-2 experienced negligible changes in appearance and negligible weight loss.

Corrosion tests at both the Westinghouse Pressurized Water Reactor Division and ORNL have shown copper and copper nickel alloys suffer only slight attack when exposed to the alkaline sodium borate solution at DBA conditions. The corrosion rate of copper, for example, in alkaline sodium borate solution at 200°F is ~0.015 mil/month(1). The corrosion of copper in an alkaline sodium borate environment under spray conditions at 264° and 212°F have been reported by ORNL. Corrosion penetrations of less than 0.02 mil was observed after 24-hour exposure at 284°F (reference 3, table 3-13) and a corrosion rate of less than 0.3 mil per month was observed at 212°C. (See reference 2, table 3-6.)

It can be seen therefore that the corrosion of copper in the postaccident environment will have a negligible effect on the integrity of the material. Further, the corrosion product formed during exposure to the solution appears tightly bound to the metal surface and hence will not be released to the ECC solution.

Consideration was given to possible caustic corrosion of austenitic steels by the alkaline solution. Data presented by Swandby(4) shows that these steels are not subject to caustic stress cracking at the temperature (285°F and below) and 6A-6 caustic concentration (less than 1 weight percent) of interest. The stress cracking boundary temperature as defined by Swandby is considerably above (~80°F) the long term, postaccident design temperature of 152°F.

It should be noted when considering the possibility of caustic cracking of stainless that the sodium hydroxide boric acid solution is a buffer mixture wherein no free caustic exists at the FNP-FSAR-6A

6A-6 REV 23 5/11 temperatures of interest, even should the solution be concentrated locally through evaporation of water; hence the above consideration is somewhat hypothetical with regard to the Farley postaccident environment.

6A.4 CORROSION OF METALS OF CONSTRUCTION BY TRACE CONTAMINANTS IN ECC SOLUTION Of the various trace elements that could occur in the emergency core cooling solution in significant quantities, only chlorine (as chloride) and mercury are adjudged potentially harmful to the materials of construction of the safeguards equipment.

The use of mercury or mercury bearing items, however, has been restricted in the Farley containment. Most mercury vapor lamps, fluorescent lighting, and instruments that employ mercury for pressure and temperature measurements and for electrical equipment have been prohibited in the containment building. Contamination due to exposure to mercury is possible if one or more temporary underwater lights used in the refueling cavity, transfer canal, and the spent-fuel pool were to fail catastrophically. The lights approved for use in these areas are manufactured by ROS, model HPS-1000, and contain up to 3 mg of mercury each in double encapsulated bulbs. The use of up to twelve of these lights at any one time has been evaluated as acceptable.

The possibility of chloride stress corrosion of austenitic stainless steels has also been considered. It is believed that corrosion by this mechanism will not be significant during the postaccident period for the following reasons:

6A.4.1 LOW TEMPERATURE OF ECC SOLUTION The temperature of the ECC solution is reduced after a relatively short period of time (i.e. a few hours) to about 150°F. While the influence of temperature on stress corrosion cracking of stainless steel has not been unequivocally defined, significant laboratory work and field experience indicate that lowering the temperature of the solution decreases the probability of failure. Hoar and Hines(5) observed this trend with austenitic stainless steel in 42 weight percent solutions of MgCl2 with temperature decrease from 310° to 272°F. Staehle and Latanision(6) present data which also shows a decreased probability of failure with decreasing solution temperature from about 392°F to 302°F. Staehler and Latanision(6) also report the data of Warren(7) which showed the significant change with decrease in temperature from 212°F to 104°F. The work of Warren, while pertinent to the present consideration in that it shows the general relationship of temperature to time to failure, is not directly applicable in that the chloride concentration (1800 ppm Cl) believed to have effected the failure was far in excess of reasonable chloride contamination that may occur in the ECC solution.

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6A-7 REV 23 5/11 6A.4.2 LOW CHLORIDE CONCENTRATION OF ECC SOLUTION It is anticipated that the chloride concentration of the ECC solution during the postaccident period will be low.

Restrictions in the chloride content of the water used in the postaccident period will not impair system operability. The environment of low chloride concentration, low temperature, and high pH, which will be experienced during the long-term postaccident period, will not be conducive to chloride cracking. $$[HISTORICAL]$$ $$[Surveillance has been maintained throughout plant construction to ensure that the chloride inventory is maintained at a minimum.]$$

6A.5 CORROSION OF ALUMINUM ALLOYS Corrosion testing showed that aluminum alloys are not compatible with alkaline borate solution.

The alloys generally corrode fairly rapidly, at the post-accident condition temperatures, with the liberation of hydrogen gas. A number of corrosion tests were conducted in the Westinghouse Pressurized Water Reactor Division laboratories and at ORNL facilities. A review of applicable aluminum corrosion data is given in Table 6A-5. The corrosion rates at the various temperature steps were determined from the aluminum corrosion rate design curve which was chosen to include essentially all available corrosion data.

6A.6 THE NATURE AND BEHAVIOR OF ALUMINUM CORROSION PRODUCTS IN ALKALINE SOLUTION The corrosion of aluminum in alkaline solution, expected following a design basis accident (DBA), has been shown to proceed with the formation of aluminum hydroxide(12,13,14) and the aluminate ion, as well as with the production of hydrogen gas.

The expected DBA conditions include the establishment of an alkaline ECC solution having a total volume of liquid of 4.5 x 105 gal after actuation of the engineered safety features.

As mentioned above, aluminum is known to corrode in alkaline solutions to give a precipitate of Al(OH)3, which in turn can redissolve in an excess of alkali to form a complex aluminate. Van Horn(12) noted that the precipitation of Al(OH)3 begins about pH 4 and is essentially complete at pH 7. A further increase in pH to about 9 causes dissolution of the hydroxide with the formation of the aluminate.

It can be seen, therefore, that the solubility of aluminum corrosion product is a function of the pH of the environment. Consistent with this, the corrosion of aluminum is also strongly dependent on the solution pH, since when the corrosion products are dissolved from the metal surface, corrosion of the base metal can proceed more freely.

Aluminum corrosion rate data had been reported in WCAP-7153(1), Table 8. The corrosion rate of aluminum is seen to decrease by a factor of 21 (1/.048) as the pH decreases from 9.3 to 8.3, and by a factor of 83 (1/.032) as the pH decreases from 9.3 to 7.0. Therefore, one must consider both corrosion and the dissolution of the corrosion products at specific reference FNP-FSAR-6A

6A-8 REV 23 5/11 conditions since the two are directly related. The corrosion reactions that are of interest in the DBA condition here would include the reaction of aluminum in alkaline solution to form aluminum hydroxide: i.e.,

++2H3(OH)Al2O2H6Al2 (1) and dissolution of the hydroxide to form the aluminate, i.e.,

(2)

A knowledge of the solubility product of the aluminum hydroxide in an alkaline solution allows the determination of the solubility expected for the hydroxide in the DBA environment.

Deltombe and Pourbaix(15) have determined the solubility product of aluminum hydroxide. Using the value of 2.28 x 10-11 for Ksp, as reported by Deltombe and Pourbaix, the following calculation can be made.

The solubility of Al(OH)3 is determined from equation 2 [][][][]+=+=+++H2AlO1110x28.2H2AlOspKO2HH2AlO3Al(OH) at pH = 9.3

[]rmoles/lite210x6.41010x51110x28.22AlO== Therefore, the solubility of Al(OH)3 in a pH 9.3 solution at 25°C (77°F) is 4.6 x 10-2 moles/liter or 3.0 x 10-2 lb/gal. Expressed as aluminum, the solubility at these conditions is 1.05 x 10-2 lb/gal.

The solubility of the aluminum corrosion products in the post-accident environment is a function of both solution pH and temperature. Plots of the corrosion product solubility are expressed in terms of aluminum versus solution pH for temperatures of 77°F and 150°F. The change in solubility with temperature is found utilizing the relationship of the free energy of formation, temperature, and the solubility product.

With the knowledge of the reference aluminum corrosion behavior for any specific plant, one can calculate the expected solubility limits for the corrosion reaction.

For the Farley plant, 4.5 x 105 gal of ECC solution will be present in the containment after actuation of the safety features. The as-built aluminum inventory present inside the containment is described in drawing A-508597 (Farley Unit 1) and A-508928 (Farley Unit 2).

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6A-9 REV 23 5/11 Table 6A-7 presents a summary of the applicable solubility and corrosion parameters for various conditions. The table lists the applicable solubility products (Ksp) and solubilities at the various temperatures and solution pHs together with the soluble aluminum limit for the Farley system at the specific conditions. The last values in the table give the aluminum solubility margin after 100 days corrosion; that is, the soluble Al limit divided by the aluminum corroded. It can be seen that in all cases, including the low temperature and low pH conditions, the ECC solution is not expected to be saturated with aluminum corrosion products. Further, within the expected design conditions for temperature and pH, adequate aluminum solubility margin is available as shown on table 6A-7.

It is concluded therefore, that the corrosion products of aluminum will be in the soluble form during the post accident period considered and, hence, there is no potential for deposition on flow orifices, spray nozzles or other equipment.

6A.6.1 BEHAVIOR OF CIRCULATING ALUMINUM CORROSION PRODUCTS The solubility of aluminum corrosion products as shown that for the Farley plant, the entire inventory produced after 100 days exposure to the post-DBA condition would remain in solution.

The review also indicates that the ECC solution is only approximately 5.5 percent saturated at 77°F and less than 3 percent saturated at 150°F.

It is of interest, however, to review the experience of facilities which have operated with insoluble aluminum corrosion products and to relate their conditions with those expected in the post accident environment.

The most significant experience available to date is that of Griess(16) who operated a recirculating test facility to measure the corrosion resistance of a variety of materials in alkaline sodium borate spray solution.

Tests were conducted on 1100, 3003, 5052, and 6061 aluminum alloys exposed at 100°C in pH 9.3 sodium borate solution (0.15 M NaOH - 0.28 M H3B03). It was reported that even though the solution contained copious amounts of flocculent aluminum hydroxide, it has no effect on flow through the spray nozzle (0.093-in. orifice). The pH of the solution did not change because of the increase in the corrosion products.

Griess(a) in describing his observations with regards to aluminum corrosion product deposition potential stated that:

A. No significant deposition was observed on the cooling coil installed in the solution.

B. No significant deposition was observed on the heated surfaces of the facility.

C. No significant deposition was observed on isothermal facility surfaces.

a. Private communication.

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6A-10 REV 23 5/11 The amounts of aluminum corroded to the solution in the tests conducted by Griess at 55°C and 100°C were approximately 4.0 and 18.6 grams, respectively. The concentration of aluminum present in the recirculation stream, therefore, was approximately 0.2 and 1 gram/liter, respectively. This value is about a factor of about 5 above the aluminum concentration expected in the postaccident ECC solution at the Indian Point plant in a pH 9.3 solution after 100 days.

Hatcher and Rae(17) describe the appearance of turbidity in the Canadian National Research Experimental Reactor Unit (NRU) reactor and "propose" that deposition of aluminum corrosion products may have occurred on heat exchanger surfaces, although they do not report any specific examination results. Moreover, Hatcher and Rae report no operations problems associated with the presence of aluminum corrosion product turbidity in the NRU reactor. The overall heat transfer coefficient for each NRU reactor heat exchanger was measured after 2 years of full power operation on several occasions and within the limit of accuracy of the measurements, reported at approximately 5 percent, no change in the thermal resistance had been observed.

It is concluded, therefore, from the work of Griess and Hatcher and Rae, that the deposition of aluminum corrosion products on heat exchangers, surfaces will not be significant in the postaccident environments even for the circumstances of insoluble product formation.

6A.7 EFFECT OF POSSIBLE CHEMICAL REACTIONS ON IODINE REMOVAL CAPABILITY OF THE CONTAINMENT SPRAY SOLUTION In evaluating the effect of possible chemical reactions on the iodine removal capability of the spray solution, it has been determined that the reaction of aluminum with an alkaline ECC solution is the only reaction occurring in the containment system during a design basis accident (DBA) which has the potential for influencing the chemistry of the ECC solution. The corrosion rate of aluminum and the solubility of the aluminum corrosion products is dependent on the pH and temperature of the alkaline spray solution. Calculations are presented in this review which estimate the mass of aluminum which would be corroded in the Farley containment following a DBA, the mass of aluminum corrosion products which would be formed, and the solubility of these corrosion products in the emergency core cooling solution. As the values in table 6A-7 indicate, there is a conservative aluminum solubility margin in the ECC solution during DBA conditions.

In the operation of a test facility to measure the corrosion resistance of a variety of materials in alkaline sodium borate spray solution, the experience of Griess(16) was that the pH of the solution did not change as a result of the buildup of aluminum corrosion products. At concentrations of 0.2 - 1.0 g of aluminum per liter, the test facility experience is representative of the Farley post accident environment, assuming that all of the aluminum in the containment had corroded away and was present in the sump solution. Although no reduction in the sump solution pH is anticipated, the equilibrium sump solution pH of 7.5 exceeds the pH required to assure that iodine is retained in the sump solution.

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6A-11 REV 23 5/11 6A.8 COMPATIBILITY OF PROTECTIVE COATINGS WITH POSTACCIDENT ENVIRONMENT The investigation of materials compatibility in the postaccident design basis environment also includes an evaluation of protective coatings for use in containment.

The results of the protective coatings evaluation presented in WCAP-7198(11) showed that several inorganic zinc, modified phenolics, and epoxy coatings are resistant to an environment of high temperature (320°F maximum test temperature) and alkaline sodium borate. Long term tests included exposure to spray solution at 150°F - 175°F for 60 days, after initially being subjected to the conservative containment temperature transient shown in table 6A-1. The protective coating found to be resistant to the test conditions, that is, exhibited no significant loss of adhesion to the substrate nor formation of deterioration products, comprises virtually all of the protective coatings recommended for use in the containment. Hence, the protective coatings will not add deleterious products to the core cooling solution.

It should be pointed out that several test panels of the recommended types of protective coatings were exposed for two DBA cycles and showed no deterioration or loss of adhesion with the substrate. In addition, the protective coatings applied to the components of the containment do not function as an integral part of the engineered safeguard features during DBA conditions.

Although the protective coatings are selected for use on the basis of their performance during a DBA, they do not serve as an engineered safety feature to inhibit corrosive attack following a loss-of-coolant accident on the substrates on which they are applied.

6A.9 EVALUATION OF THE COMPATIBILITY OF CONCRETE ECC SOLUTION IN THE POSTACCIDENT ENVIRONMENT Concrete specimens were tested in boric acid and alkaline sodium borate solutions at conditions conservatively (320°F maximum and 200°F steady state) simulating the post-DBA environment.

The purpose of this study was to establish:

A. The extent of debris formation by solution attack of the concrete surfaces.

B. The extent and rate of boron removal from the ECC solution through boron concrete reaction.

Tests were conducted in an atmospheric pressure, reflux apparatus to simulate long term exposure conditions and in a high pressure autoclave facility to simulate the DBA short term, high temperature transient.

Table 6A-8 presents a summary of the data obtained from the concrete boron test series.

Testing of uncoated concrete specimens in the post accident environment showed that attack by both boric acid and the alkaline boric acid solution is negligible and the amount of FNP-FSAR-6A

6A-12 REV 23 5/11 deterioration product formation is insignificant. In addition, the boron removal rate from the ECC solution is low.

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6A-13 REV 23 5/11 REFERENCES 1. Bell, M.J., Bulkowski, J.E. and Picone, L.F., "Investigation of Chemical Additives for Reactor Containment Sprays," WCAP-7153, March 1968. (Westinghouse Proprietary) 2. ORNL Nuclear Safety Research & Development Program Bimonthly Report for July-August 1968, ORNL TM-2368, p. 78.

3. ORNL Nuclear Safety Research & Development Program Bimonthly Report for September-October 1968, ORNL TM-2455, p. 53 4. Swandby, R.K., Chemical Engineer 69, 186 (November 12, 1962).

5. Hoar, T.P., and Hines, J.G., "Stress Corrosion Cracking of Austenitic Stainless Steel in Aqueous Chloride Solutions," Stress Corrosion Cracking and Embrittlement (ed. W.D. Robertson) John Wieley and Sons, 1956.

6. Latanision, R.M., and Staehle, R.W., Stress Corrosion Cracking of Iron - Nickel Chromium Alloys, Dept. of Metallurgical Engineering, The Ohio State University 7. Warren, D., Proceeding of Fifteenth Annual Industrial Work Conference, Purdue University, May 1960.

8. Edeleanu, C., JISI 173, 1963, 140.

9. Thomas, K.C., et al., "Stress Corrosion of Type 304 Stainless Steel in Chloride Environment," Corrosion, Vol. 20, 1964, p. 89t.

10. Sharfstein, L.R., and Brindley, W.F., "Chloride Stress Corrosion Cracking of Austenitic Stainless Steel - Effect of Temperature and pH," Corrosion, Vol. 14, 1958, p. 588t.
11. Picone, L.F., "Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198, April 1968. (Westinghouse Proprietary)
12. Van Horn, K.C., Aluminum, Vol. I, American Society of Metals, (1967).
13. Sundararajan, J., and Rama Char, T.C., Corrosion 17, 39t, (1961).
14. Cotton, F.A., and Wilkinson, G., Advanced Inorganic Chemistry, Interscience Publishers, (1962).
15. Deltombe, E., and Pourbaix, M., Corrosion 14, 496t, (1958).
16. Griess, J.C., et al., "Corrosion Studies," pp. 76-81, ORNL Nuclear Safety Research and Development Program Bimonthly, July - August 1968, USNRC Report ORNL TM-2368.

FNP-FSAR-6A

6A-14 REV 23 5/11 17. Hatcher, S.R., and Rae, H.K., Nuclear Sci. and Eng., 10, 316, (1961).

18. Rubin, K., Grover, J. L., Henninger, W. A., and Miller, T. A., "Methodology for Elimination of the Containment Spray Additive," WCAP-11611, Rev. 0, March 1988, (Westinghouse Proprietary)

FNP-FSAR-6A REV 21 5/08 TABLE 6A-1 POSTACCIDENT CONTAINMENT TEMPERATURE TRANSIENT USED IN THE MATERIAL COMPATIBILITY REVIEW Time Interval (s) Temperature (°F) 0 - 300 285 300 - 1000 266 1000 - 2000 234 2000 - 4000 190 >4000 147

FNP-FSAR-6A REV 21 5/08 TABLE 6A-2 (SHEET 1 OF 2) REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION Representative Group Elements Corrosivity of Elements Sources of Elements 0 H3, Ne, K, Xe No effect on any materials of Fission product release construction I a Li, Na, K Generally corrosion inhibitive Li - coolant pH adjusting properties for steels, and agent copper alloys - harmful to Na - spray additive solution, aluminum concrete leach product K - concrete leach product II a Mg, Ca, Sr, Ba Generally not harmful to steel Concrete leach products - or copper base alloys deteriorated insulation III a Y, La, Ac Not considered harmful in low Fission product release concentrations IV a Ti, Zr, Hf Not considered harmful to any Fuel rod cladding, control materials rod material, alloying constituent V a V, Nb, Ta Not considered harmful to any Alloying constituents in materials low concentration VI a Cr, Mo, W Not considered harmful to any Alloying constituents in materials equipment VII a Mn, Tc, Re Not considered harmful Mn - alloy constituent VIII Fe, Ni, Cr, Os Fe, Ni, Cr - not harmful to Fe, Ni, Cr - alloying any materials constituents. Others have no identifiable sources FNP-FSAR-6A REV 21 5/08 TABLE 6A-2 (SHEET 2 OF 2) REVIEW OF SOURCES OF VARIOUS ELEMENTS IN CONTAINMENT AND THEIR EFFECTS ON MATERIALS OF CONSTRUCTION Representative Group Elements Corrosivity of Elements Sources of Elements I b Cu, Ag, Au Not harmful to any materials Cu present as material of construction and alloying constituent II b Zn, Cd, Hg Hg - harmful to stainless Hg has been entirely excluded steel, Cu alloys, from use in the containment. aluminum Cd finish plating on Zn - unknown components. Zn galvanizing Cd - unknown and alloying constituent III b B, A1, Ga, In Not harmful to material B - neutron poison additive A1 - materials of construction IV b C, Si, Sn, Pb C, Si, Sn not harmful to Si - concrete leach product materials. Pb considered Pb - alloy constituent in harmful to nickel alloys some brazes V b N, P, As, Sb, Bi No effect from N unless N - containment air. Others ammonia is formed. Others not identified in significant unknown materials VI b O, S, Se, Te S possibly harmful to nickel Te - fission product alloys S - oils, greases, insulating materials VII b F, C1, Br, I F considered potentially C1 - concrete leach product harmful to zircaloy. general contamination C1 potentially harmful to F - organic materials stainless steel Br and I, I and Br - fission products not generally harmful low concentration FNP-FSAR-6A REV 21 5/08 TABLE 6A-3 TYPICAL MATERIALS OF CONSTRUCTION IN THE FARLEY CONTAINMENT Material Equipment Application 300 series stainless Reactor coolant system, residual steel heat removal loop, spray system, fan cooler material 400 series stainless Valve materials steel

Inconel (600, 718) Steam generator tubing, reactor vessel nozzles, core supports, and fuel rod grids Galvanized steel Ventilation duct work, CRDM shroud material, I & C conduit Aluminum Refer to drawing A-508597 for Farley Unit 1 and A-508928 for Farley Unit 2 Copper Service water piping, fan cooler material

70-30 Cu Ni Fan cooler material 90-10 Cu Ni Fan cooler material Carbon steel Component cooling loop, structural steel, main steam piping, etc Monel Possibly instrument housings Brass Possibly instrument housings Protective coatings General use on carbon steel structures and equipment, Inorganic zincs concrete Epoxy Modified phenolics

FNP-FSAR-6A REV 21 5/08 TABLE 6A-4

This table has been deleted.

FNP-FSAR-6A REV 21 5/08 TABLE 6A-5 CORROSION OF ALUMINUM ALLOYS IN ALKALINE SODIUM BORATE SOLUTION Corrosion Data Temperature Alloy Test Rate Exposure Point (°F) Type Duration (mg/dm2/h) pH Condition Reference 1 275 5025 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 96.2 9 Solution WCAP-7153, Table 9 2 275 5005 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 840 9 Solution WCAP-7153, Table 9 3 200 6061 320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> 15.4 9.3 Solution WCAP-7153, Table 8 WCAP-7153, Figure 9 4 210 5052 7 days 53.0 9 Solution WCAP-7153, Table 7 WCAP-7153, Figure 8 5 210 5052 2 days 14.0 9 Solution WCAP-7153, Table 5 6 210 5005 2 days 27.1 9 Solution WCAP-7153, Table 5 7 284 5052 1 day 54 9.3 Spray ORNL-TM-2425, Table 3.1 8 284 5052 1 day 31.5 9.3 Solution ORNL-TM-2425, Table 3.1 9 212 6061 3 days 126 9.3 Spray ORNL-TM-2368, Table 3.6 10 212 6061 3 days 110 9.3 Solution ORNL-TM-2368, Table 3.6 11 150 6061 7 days 2.9 9.3 Solution Westinghouse Pressurized Water Reactor Division recent data 12 150 5052 7 days 4.2 9.3 Solution Westinghouse Pressurized recent data

FNP-FSAR-6A REV 21 5/08 TABLE 6A-6

This table has been deleted.

FNP-FSAR-6A REV 21 5/08 TABLE 6A-7 SUMMARY OF ALUMINUM CORROSION PRODUCT SOLUBILITY DATA Solution Temperature 77°F 150°F Parameter pH 9.3 pH 8.3 pH 9.3 pH 8.3 Solubility product 2.28 x 10-11 2.28 x 10-11 4.16 x 10-10 4.16 x 10-10 Ksp Al solubility 1.05 x 10-2 1.05 x 10-3 1.9 x 10-1 1.9 x 10-2 (lb Al/gal) Soluble Al limit(a) 4.73 x 103 4.73 x 102 8.55 x 104 8.55 x 103 for ECCS (lb) Al corrosion rate (Not used) (Not used) 1 0.048 (normalized) Al corroded after (Not used) (Not used) 1800 1077 100 days (lb) Al solubility margin 18 3 47.5 7.9 at 100 days

a. Solution volume 4.5 x 105 gal.

FNP-FSAR-6A REV 21 5/08 TABLE 6A-8 CONCRETE SPECIMEN TEST DATA Total Exposed Initial Concrete Exposure Surface Weight Specimen - Boron Period Volume Change Weight Test No. (days) (in. /gal) (grams) (grams) Visual examination 1 24 28 - 22.4 560.0 No apparent change 3 28 20 + 21.5 404.0 Light, yellowish, deposit on specimen 4 72 38 0 641.2 No apparent change - coating adhesion excellent 5 72 43 - 0.2 769.5 Light, hard deposit on specimen 6 ~4(a) 54 - 601.4 No apparent change - small amount of sand particles in test can 7 175 23 + 11.0 457.0 No apparent change 8 175 38 + 26.5 751.0 No apparent change - coating adhesion excellent 9 ~5(a) 78 + 4.0 702.0 No apparent change - coating adhesion excellent

a. These tests were at high temperature DBA transient conditions. All others at 195 - 205°F.

FNP-FSAR-6B

6B-i REV 21 5/08 APPENDIX 6B CONTAINMENT PRESSURE ANALYSIS TABLE OF CONTENTS Page 6B.1 CONTAINMENT PRESSURE RESPONSE................................................................6B-1 6B.2 CONTAINMENT SUBCOMPARTMENT ANALYSIS...................................................6B-1

FNP-FSAR-6B

6B-ii REV 21 5/08 LIST OF TABLES

6B-1 Node Spacings

6B-2 Thickness of Sensitive Heat Conduction Layer

6B-3 Mesh Spacing in Sensitive Layer to Achieve 0.5% Accuracy

FNP-FSAR-6B

6B-iii REV 21 5/08 LIST OF FIGURES

6B-1 Reactor Cavity Block Diagram

6B-2 Steam Generator Cavity Pressurization Analysis

6B-3 Total Horizontal Force vs. Time

6B-4 Reactor Cavity Analysis

FNP-FSAR-6B

6B-1 REV 21 5/08 APPENDIX 6B CONTAINMENT PRESSURE ANALYSIS 6B.1 CONTAINMENT PRESSURE RESPONSE The containment pressure response to a loss-of-coolant accident (LOCA) has been analyzed using the heat sinks as presently designed. The methods and assumptions used in this analysis are described in paragraph 6.2.1. The double-ended pump suction break was originally determined to be the worst case. The analysis for the break showed a peak pressure of 48 psig at 276 s and a maximum temperature of 313°F, at 55 s after the break. Current results are provided in paragraph 6.2.1.3.6.

A summary of the current heat sinks is given in Table 6.2-2. Table 6B-1 provides a table of the original node spacings for original heat sinks. Node spacings for power uprate analyses are generally more fine or comparable to those shown in Table 6B-1. Detailed conservative calculations were performed to determine each heat sink surface area. For additional conservatism, some heat sinks (e.g., all piping in the containment and miscellaneous steel such as some support brackets and rails) were not included in the analysis.

6B.2 CONTAINMENT SUBCOMPARTMENT ANALYSIS The following section provides a discussion of the original design prior to application of leak-before-break exclusion of RCS main loop breaks. Current analyses and results are provided in paragraph 6.2.1.3.4.1.

The containment subcompartments analyzed for the pressure response following a LOCA were the reactor cavity and the steam generator annulus (the volume below the steam generator compartments). The pressure transient analysis was performed using a Bechtel computer code which calculates short term pressure and temperature responses. The code conservatively neglects heat transfer and all engineered safety features. A detailed description of the code is provided in appendix 3K, attachment D.

The model used for the reactor cavity analysis is shown in figures 6B-1, 6B-2, and drawing D-176277. Volumes, vent area, and flow coefficients are also shown in figure 6B-1. Blowdown data was supplied by Westinghouse for the 1 ft2 cold leg break (at 95° az. in drawing D-176277) which is the limiting case for reactor cavity design. The blowdown is split equally between volumes 1 and 2. Insulation in the break region (compartments 1 and 2) is assumed to blow off and completely plug the cold leg penetration at the wagon wheel restraint, as well as the support shoe area ventilation duct. All gaps in the broken leg blowdown restrictor/baffle plate remain completely unobstructed by insulation throughout the transient. In all other places (i.e.,

reactor vessel, nozzle, and pipes for all intact legs) insulation is assumed to remain in place and not crush, leaving the seal ring gap and unbroken leg baffle plate gaps open for ventilation to the containment. The maximum horizontal force was calculated to be 1.4 x 106 lbf. The maximum uplift force was 5.9 x 104 lbf. The force-time history results are shown in figures 6B-3 and 6B-4.

FNP-FSAR-6B

6B-2 REV 21 5/08 The flow models for the steam generator compartment pressurization analyses are shown in figure 6B-2. Blowdown data were supplied by Westinghouse for a double ended cold leg break in the steam generator compartment C, which is the limiting case. The maximum differential pressure between steam generator compartment C and the containment was found to be 33.9 psia at 0.42 seconds.

FNP-FSAR-6B REV 21 5/08 TABLE 6B-1 (SHEET 1 OF 5) NODE SPACINGS Heat Sink No. 1 - Containment Cylinder and Dome Node Spacing Thickness Material (in.) (in.)

Paint 1 x 10-3 2.0 x 10-2 Primer(a) 1 x 10-3 3.0 x 10-3 Carbon steel 6.25 x 10-2 2.5 x 10-1 Concrete region 1 5.0 x 10-2 3.0 Concrete region 2 4.0 x 10-1 6.0 Concrete region 3 1.2 6.0 Concrete region 4 10.0 30.0 Heat Sink No. 2 - Unlined Concrete Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 18.0 x 10-3 Surfacer(a) 1.0 x 10-2 1.25 x 10-1 Concrete region 1 5.0 x 10-2 3.0 Concrete region 2 1.76 x 10-1 3.0 Concrete region 3 6.0 x 10-1 3.0 Heat Sink No. 3 - Outside Reactor Cavity Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 18.0 x 10-3 Surfacer(a) 1.0 x 10-2 1.25 x 10-1 Concrete 5.0 x 10-2 3.0 FNP-FSAR-6B REV 21 5/08 TABLE 6B-1 (SHEET 2 OF 5) Heat Sink No. 4 - Galvanized Steel Node Spacing Thickness Material (in.) (in.)

Zinc 6.7 x 10-4 3.35 x 10-3 Carbon steel 6.5 x 10-3 6.56 x 10-2

Heat Sink No. 5 - Miscellaneous Steel Less than 0.12 in. Thick

Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 7.64 x 10-2 Heat Sink No. 6 - Miscellaneous Steel 0.12 to 0.16 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 1.32 x 10-1 Heat Sink No. 7 - Miscellaneous Steel 0.16 to 0.24 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 1.91 x 10-1 FNP-FSAR-6B REV 21 5/08 TABLE 6B-1 (SHEET 3 OF 5) Heat Sink No. 8 - Miscellaneous Steel 0.24 to 0.30 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 2.55 x 10-1 Heat Sink No. 9 - Miscellaneous Steel 0.30 to 0.40 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 5.0 x 10-3 3.38 x 10-1 Heat Sink No. 10 - Miscellaneous Steel 0.40 to 0.50 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 4.92 x 10-1 Heat Sink No. 11 - Miscellaneous Steel 0.50 to 0.625 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 5.76 x 10-1

FNP-FSAR-6B REV 21 5/08 TABLE 6B-1 (SHEET 4 OF 5) Heat Sink No. 12 - Miscellaneous Steel 0.625 to 0.75 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.0 x 10-2 7.24 x 10-1 Heat Sink No. 13 - Miscellaneous Steel 0.75 to 1.0 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 1.5 x 10-2 9.35 x 10-1 Heat Sink No. 14 - Miscellaneous Steel 1.0 to 1.5 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 2.0 x 10-2 1.43 Heat Sink No. 15 - Miscellaneous Steel Greater than 1.5 in. Thick Node Spacing Thickness Material (in.) (in.)

Paint 1.0 x 10-3 2.0 x 10-2 Primer(a) 1.0 x 10-3 3.0 x 10-3 Steel 3.5 x 10-2 2.85 FNP-FSAR-6B REV 21 5/08 TABLE 6B-1 (SHEET 5 OF 5) Heat Sink No. 16 - Stainless Steel Node Spacing Thickness Material (in.) (in.)

Stainless steel 5.0 x 10-3 1.68 x 10-1

____________________

a. When Amercoat 90 is used as the primer, the average primer thickness will be 5.0 mils. However, the total thickness of primer plus finish coat will not exceed the total thickness of finish coat plus primer (surfacer) listed in the table.

FNP-FSAR-6B REV 21 5/08 TABLE 6B-2 THICKNESS OF SENSITIVE HEAT CONDUCTION LAYER Typical Heat Conduction Time Materials 20 s 100 s 200 s 400 s Concrete A 0.054 ft 0.121 ft 0.170 ft 0.243 ft K = 1.0 Cp = 25.2 Steel 0.200 ft 0.450 ft 0.640 ft 0.906 ft K = 29.6 Cp = 53.6 Inorganic Zinc 0.058 ft 0.130 ft 0.184 ft 0.260 ft Primer K = 1.24 Cp = 27.36

FNP-FSAR-6B REV 21 5/08 TABLE 6B-3 MESH SPACING IN SENSITIVE LAYER TO ACHIEVE 0.5 PERCENT ACCURACY Typical Accuracy Crossover Time Materials 20 s 40 s 100 s 200 s 400 s Concrete A 224 158 100 71 50 K = 1.0 mesh mesh mesh mesh mesh CP = 25.2 ft ft ft ft ft Steel 60 42 27 19 13 K = 29.6 mesh mesh mesh mesh mesh Cp = 53.6 ft ft ft ft ft Inorganic 210 148 94 66 47 zinc mesh mesh mesh mesh mesh primer ft ft ft ft ft K = 1.24 Cp = 27.36

REV 21 5/08 REACTOR CAVITY BLOCK DIAGRAM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-1

REV 21 5/08 STEAM GENERATOR CAVITY PRESSURIZATION ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-2 REV 21 5/08 TOTAL HORIZONTAL FORCE VERSUS TIME JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-3 REV 21 5/08 REACTOR CAVITY ANALYSIS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6B-4 FNP-FSAR-6C 6C-i REV 21 5/08 [HISTORICAL (Prior to December 2007)] [APPENDIX 6C CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM TABLE OF CONTENTS Page 6C.I CONTAINMENT SUMP DESCRIPTION -----------------------------------------------6C-1 6C.II ECCS RECIRCULATION MODE TEST PROGRAM ---------------------------------6C-3 6C.III UNIT 1 TESTS-------------------------------------------------------------------------------6C-3 6C.IV UNIT 2 TESTS-------------------------------------------------------------------------------6C-13

FNP-FSAR-6C 6C-ii REV 21 5/08 LIST OF TABLES 6C-1 Test Conditions for Unit 1 Intake 1 6C-2 Test Conditions for Unit 1 Intakes 2, 3, and 4

FNP-FSAR-6C 6C-iii REV 21 5/08 LIST OF FIGURES

[Historical] 6C-1 Typical Arrangement of Containment Sump Suction Line 6C-2 Modeled Areas of ECCS Intakes 6C-3 Hydraulic Model Plan for Intake No. 1 Tests 6C-4 Blockage Test Conditions 6C-5 No. 1 Intake Configuration for Initial Tests 6C-6 Improved Design Intake No. 1 6C-7 Plan of Modeled Area Containing Intakes 2, 3, and 4 6C-8 Containment Sump 6C-10 Photograph of Model 6C-11 Photograph of Model 6C-12 Intakes 2, 3, and 4 Improved Design 6C-13 Blockage Test Conditions for Intakes 2, 3, and 4 6C-14 Photograph of Grating Cage Over Intake 2 6C-15 Plan of Unit 2 Test Facility 6C-16 Section of Unit 2 Test Facility 6C-17 Representative Screen-Grating Structure 6C-18 Grating Cage - Final Design 6C-19 Sump Area of Unit 2 6C-20 Composite Drawing of Unit 2 Sump 6C-21 Composite Drawing of Unit 1 Sump 6C-22 Photo of Unit 2 Grating Cage 6C-23 Photo of Representative Screen-Grating Cage]

FNP-FSAR-6C 6C-1 REV 21 5/08 [HISTORICAL (Prior to December 2007][APPENDIX 6C CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION MODE TEST PROGRAM]

Appendix C was made historical in December 2007 following the installation of new containment sump strainers for RHR and CS suction inlets. This was required by Generic Letter (GL) 2004-02, "Potential Impact for Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors." Appendix 6D has been created to provide a description of the containment sump and the new suction strainers. Appendix 6C contains the design bases for the original containment sumps and is being maintained for historical reference.

[HISTORICAL] [I. CONTAINMENT SUMP DESCRIPTION The containment recirculation sump is a collecting reservoir designed to provide an adequate supply of water, with a minimum amount of particulate matter, to the containment spray system (CSS) and the residual heat removal system (RHRS). The containment sump performance meets the NRC acceptance criteria contained in General Design Criteria 35, 36, and 37, and the five NRC acceptance criteria listed below. A. The net positive suction head (NPSH) available to each safety system pump has been shown to provide adequate margin over the required NPSH at limiting runout conditions (see FSAR paragraph 6.3.2.14). B. Housekeeping requirements specified in the quality assurance program and the Technical Requirements Manual. C. The avoidance of materials likely to form debris small enough to pass through sump screens. D. The lack of an apparent mechanism for generating debris large enough to block more than 50 percent of the screen area. E. The ability to monitor and control RHRS status. The design criteria for the containment sumps and sump screens are the following: A. Separate sumps are provided to serve each of the redundant halves of the ECCS and CSS. The redundant sumps are physically separated from each other and are located outside the missile barrier. The sumps are located on the lowest floor elevation in the containment, exclusive of the reactor vessel cavity. B. The Unit 1 sump intakes are protected by an outer trash rack and a fine mesh inner screen with a steel grating support. The size of the openings in the fine screen take into account the overall operability of the system served.

FNP-FSAR-6C 6C-2 REV 21 5/08 C. A solid plate covers most of the top of each screen structure. This plate can be removed to facilitate inspection of the structure and pump suction intake. The top deck will be fully submerged after a LOCA and completion of safety injection. D. Materials for the grating and screens were selected to avoid degradation during periods of inactivity and operation and have a low sensitivity to adverse effects, such as stress corrosion that may be induced by the chemically reactive spray during loss-of-coolant accident (LOCA) conditions. E. A vortex breaker is provided at the sump intake end of each of the pump suction pipes. F. The sumps are designed to yield low velocities of approach in the vicinity of the sumps to promote the settling out of debris, and to yield negligible pressure drops through the sump screens. Materials inside containment which could cause sump screen blockage post-LOCA have been eliminated or minimized by design. G. The screens and associated structures have been designed to withstand the vibratory motion of seismic events without loss of structural integrity. H. Each pump suction line is installed with a continuous slope from the sump to the pump to assure free venting of air. (See figure 6C-1.) There is a sufficient time interval before start of the recirculation phase to allow complete venting of the suction lines (approximately 30 min). I. Field tests have been performed on the pump suction lines for two purposes: to flush the lines to remove any possible obstructions, and to verify pressure drop calculations made for pump NPSH requirements. The tests were run with the pump startup strainers in place. A typical sump detail drawing prior to modification is shown on figure 6C-8.] In each of the four pump suction lines from the containment sump, there are two motor-operated gate valves. There is no interdependency between systems or between the redundant portions of the same system. The motor-operated gate valves in the lines from the containment sump to the various pumps are normally closed and remain closed during the injection phase of emergency core cooling system (ECCS) operation. The protective screened structures in the containment sump will be completely submerged at the end of the injection phase and will remain submerged during the recirculation phase. The various parameters (e.g., flowrates, pressure drops, sump levels, etc.) listed in the following sections are from the original ECCS and CSS recirculation mode testing. The ECCS and CSS flowrates and sump levels utilized in the current pump NPSH calculations are within the range of flowrates and sump levels tested in the original sump recirculation tests. The pressure drop across the sump screen, vortex breaker, sump inlet, and suction piping utilized in the current NPSH calculations have been developed from the original sump recirculation test program and the ECCS field tests based on the calculated ECCS and CSS flowrates. Since the current parameters utilized in the NPSH calculations are bounded by those in the original sump recirculation tests, the ECCS and CSS sump intake design will not develop FNP-FSAR-6C 6C-3 REV 21 5/08 II. ECCS RECIRCULATION MODE TEST PROGRAM A. PURPOSE The purpose of this hydraulic model study is to document that the ECCS intakes, of the J. M. Farley Nuclear Plant Units 1 and 2 will not develop unacceptable flow reducing or air entraining vortices. The Unit 1 intakes were tested first. The model boundaries were placed remotely from the screen grating structures around the intakes and selected so as to be able to reproduce the flows in the area external to the intakes. Based on the findings from these tests it was concluded that it was not necessary to model the area outside the screen-grating structure for Unit 2. A description of the intakes, the test program and the results and conclusions for each unit are presented in the following sections. III. UNIT I TESTS A. INTRODUCTION The emergency core cooling system intakes of Unit 1 are comprised of two 14 in. and two 10 in. vertical inlets located in three intake areas and are designated as intakes 1, 2, 3 and 4, as shown in figure 6C-2. This section presents the results of testing the 14-inch nominal diameter intakes 1 and 2 and the 10-inch nominal diameter intakes 3 and 4 of Unit 1. The tests were conducted to examine NRC's concern relative to the potential occurrence of vortices near or in the intake areas, which could result in loss of pumping capacity or pump failure due to vibration. Such occurrences could reduce pumping capacity by air entrainment and/or by unacceptably high intake head losses. Air entrainment could also produce unbalanced pressures on the pump impeller and cause pump failure because vibration. Therefore, a satisfactory intake design should be free of air entraining vortices and have acceptable intake loss coefficients. Lack of published and documented information relative to effects of the complex flow patterns approaching the intakes, the grating and screens, and the low viscosity of the heated water precluded analytical or empirical predictions as to whether the intake configuration would be free from objectionable vortex action. The plant conditions do not permit inplace testing. Therefore, a hydraulic model was selected to evaluate the adequacy of the intake design with respect to vortices. Drawing D-175200 shows the general features of the containment sump which could affect the flow of water to the sump area. The elevator shaft in the area of the emergency cooling intakes, figure 6C-2, provided a natural model boundary and facilitated the examination of Intake 1 separately from Intakes 2, 3 and 4.

FNP-FSAR-6C 6C-4 REV 21 5/08 Intakes 1 and 2 design flows range from 3000 to 5900 gal/min. The 5900 gal/min corresponds to two residual heat removal (RHR) pumps taking suction through a single sump line. Intakes 3 and 4 have a design flow rate of 3050 gal/min each. The accident condition postulates that under certain conditions flow could approach the intake from both sides. However, for the majority of cases flow from the left (Q1) would exceed the intake 1 flow rate resulting in flow passing this intake toward Intakes 2, 3, and 4. (See figure 6C-2.) The calculated minimum and maximum water levels in the containment are 58.3 and 77.1 inches, respectively, above the floor. The maximum containment sump water temperature during recirculation following a postulated LOCA is 212°F at subcooled pressures. A maximum water temperature of 240°F was assumed for the model study. B. THE MODEL 1. General First, Intake 1 was modeled at a 1:1 undistorted scale within a 25 ft wide, 60 ft long, 12 ft deep concrete tank. Then Intakes 2, 3, and 4 were modeled in the same concrete tank. All columns, restraints, and piping greater than 2 in. diameter were represented in the model. (See figures 6C-3, 7, 10, and 11.) The protective screen and grating structure was constructed in accordance with figure 6C-5 and was modified as shown in figures 6C-5, 6, and 12. The screen cloth consisted of 0.120 in. wire with an effective opening of 51.6 percent. The screen was sandwiched between grating of 1-1/4 in. by 3/16 in. bars on 1-3/16 in. centers. Flow baffles were placed at the extremities of the modeled area to insure uniform flow at the model boundaries. Viewing ports were incorporated in the tank to permit observation of flow conditions within the screen area around the intake. Piezometers were installed to measure static pressures inside and outside of the intake screens. Piezometer taps were installed initially at 5 pipe diameters downstream in the Intake 1 pipe and later at 29 pipe diameters downstream in the same intake pipe. They were also installed at 39.6, 36.7, and 25.7 pipe diameters downstream in the Intake 2, 3, and 4 pipes, respectively. Later, an additional tap was installed at 25.6 pipe diameters downstream of Intake 3. The model was capable of being operated at 50 percent above prototype velocities and up to temperatures of 180°-190°F. 2. Scale Selection The 1:1 scale was chosen in order to test the intakes under conditions which were as close to postulated LOCA conditions as practically possible. The study of fluid dynamics has shown that the parameters which affect vortex formations may be represented by the following dimensionless numbers:

FNP-FSAR-6C 6C-5 REV 21 5/08 a. Weber number, VD2/, which is the ratio of surface tension to inertia forces. b. Froude number gDV, which is a ratio of gravity to inertia forces. c. Reynolds number,VD , which is the ratio of viscous to inertia forces. d. Circulation number, QRVr2, or the similar Kolf number, which characterizes circulation. e. Strouhal number, VDfe, which characterizes the frequency of eddy shedding. The parameters identified in the preceding dimensionless numbers are: r - radius of inlet, ft. D - characteristic length, ft., e.g., depth or diameter R - radius of tank or perhaps flow streamline, ft. Q - discharge, ft3/s V - characteristic velocity, ft/s fe - frequency of eddy shedding, s-1 g - gravitational acceleration, ft/s2 - surface tension, lb/ft. - kinematic viscosity, ft2/s - Density, slugs/ft3 To reproduce exact dynamic and kinematic similarity on a geometrically similar model would require the value of all dimensionless numbers to be the same in model and prototype. The 1:1 scale model, and the test program which followed, permitted tests to be conducted at prototype values of all numbers, but not simultaneously. Conducting tests at prototype discharges and 170°F - 190°F temperatures reproduced the Froude, Circulation and Strouhal numbers with Reynolds and Weber numbers being lower than prototype values. Augmenting the discharges to reproduce prototype Reynolds number yielded Froude, Circulation, Strouhal and Weber numbers in the model which were higher than prototype values. Since the Froude number involves the principal parameters related to surface flow phenomena, conducting the tests at prototype discharges establishes the surface flow characteristics outside the screen area concurrent with correctly simulated circulation and eddy shedding (Strouhal number) effects. At FNP-FSAR-6C 6C-6 REV 21 5/08 equivalent Froude numbers, the model Weber number was less than the prototype value. Flow conditions within the screen area are independent of the Froude number and primarily dependent upon the Reynolds and Circulation numbers. Hence, conducting tests at prototype Reynolds numbers permitted examination of conditions within the screen area, concurrent with the Circulation and Weber numbers being greater than prototype values. Based upon the work of Dagget and Keulegan (reference 18), increasing the Circulation number for a constant Reynolds number increases vortex action. Hence it was considered conservative to conduct tests at prototype Reynolds numbers. Furthermore with Reynolds number equivalence, the model Weber number was greater than the prototype value, which together with the unaugmented flow tests bracketed the prototype Weber number. Therefore the 1:1 scale model, with tests conducted at and above prototype discharges, reproduced or exceeded the prototype values of the relevant dimensionless numbers. Exceeding prototype values of the dimensionless numbers was considered to produce conservative results. C. THE MODEL TESTING PROGRAM The tests examined the performance of Intake 1 over the range of flow conditions and water levels given in table 6C-1, for an unblocked condition and for the five postulated blockage conditions shown on figure 6C-4. These conditions were postulated by considering the nature of debris that could reach the screen, and the paths of the flow approaching the screens. Flow directions for Q1 and Q2 are indicated on figure 6C-2. Tests 1 to 6, table 6C-1, were conducted with and without discharges augmented to develop Reynolds numbers equal to, or larger than, prototype values. A preliminary set of runs was also made on Tests 1 to 6 at prototype discharges, without blockage, to: 1. Establish the general performance characteristics of the intake. 2. Observe surface flow conditions at Froude number equivalence between model and prototype. 3. Establish a basis for comparison of surface flow conditions with conditions at augmented discharges. Tests 7 and 8 were to be conducted with and without blockage at prototype discharges and water temperatures of 170°F. There was full prototype equivalence for these two tests. The tests also examined the performance of Intakes 2, 3, and 4 over the range of flow conditions and water levels given in table 6C-2, for an unblocked condition, and for the five postulated blockage conditions shown on figure 6C-13. Flow directions for Q1 and Q2 are indicated on figure 6C-2.

FNP-FSAR-6C 6C-7 REV 21 5/08 Tests 1 to 6 and 8 to 10, table 6C-2, were conducted with and without, blockage and with water temperatures of 180°F or greater and prototype discharges augmented to develop Reynolds numbers equal to or larger than prototype values. Test 7, table 6C-2, was conducted with and without blockage at prototype discharges and water temperatures of 180°F. There was full prototype equivalence for this test. D. MODEL TEST RESULTS INTAKE 1 1. General Preliminary tests with the screen grating and Intake 1 design shown in figure 6C-8 indicated that air became trapped underneath the cover plate either during filling or upon coming out of solution due to heating. With an air pocket present, a vortex tended to form beneath the cover plate which immediately withdrew the air into the intake. To minimize the accumulation of trapped air under the plate, the modifications shown in figure 6C-5 were made. The solid cover plate of the screen structure was given a slope of 2 inches over its length and the plate was shortened 1/2 inch to provide a vent slot next to the secondary shield wall. The initial test documentation was made with the intake design of figure 6C-5, for test conditions shown in table 6C-1. These tests indicated that blockage condition 5 created flow conditions within the screen area which generated a horizontally oriented vortex which originated at the secondary shield wall inside the screen and which entered the nearest quadrant of the inlet cruciform. A further modification consisting of the grating skirt shown in Figure 6C-6 was developed to eliminate the penetration of this vortex into the intake. A final series of tests was conducted for the configuration shown in figure 6C-6. The results of the preliminary, initial and final tests are presented below. 2. Preliminary Tests The preliminary tests were run with unblocked screens at prototype velocities and at velocities increased to produce prototype Reynolds numbers. These tests established that: a. The proposed design could trap air under the solid cover plate which would lead to the formation of an air core vortex within the screen area that very quickly exhausted the trapped air. b. There was no vortex formed outside of the screen structure. c. There was no observable difference between flow patterns at prototype and augmented velocities. Hence, there was no distortion of flow patterns when departing from Froude similitude at augmented flows.

FNP-FSAR-6C 6C-8 REV 21 5/08 3. Initial Documentation Tests The results from the initial documentation tests are as follows: a. There were no vortices under any test condition which established an air core from the free surface within the containment area to the screen grating around the intake. Surface depressions in the eye of any eddies or organized circulation did not exceed 1/2 inch in depth. b. Air introduced artificially under the cover plate above the intake, which represented air trapped during flooding of the containment of coming out of solution, was able to escape through the 1/2 inch vent slot in the solid cover plate for all prototype test conditions. Air could remain trapped below the plate for short periods of time. This air swirled above the intake but no air was drawn into the intake irrespective of the quantity of air forced beneath the plate. c. No organized circulation or vortices were observed within the screen area around the intake for an unblocked screen, nor for blockage conditions 1 through 4. Organized circulation did develop for blockage condition 5 with the strength of circulation being a function of the intake flow. The axis of circulation was horizontal and originated near the shield wall, approximately 9 inches to the left of the screen cage center. It curved into the intake quadrant nearest the left side and the shield wall. This condition could first be noticed at intake flows of about 4,000 gal/min. As the flow increased (and thus the pressure in the core of rotation decreased) an intermittent vapor core developed. Above discharges of about 4,300 gal/min a continuous vapor core 1/16 to 1/8 inch in diameter was present. This condition did not result in a measurable increase in intake head loss. d. The maximum measured head loss across the screen and grating corrected to prototype discharge was 0.09 ft. The intake loss coefficient was computed from the equation KhVgVg=2222// h = pressure drop in feet of water from inside the screen to a pressure tap down stream from the pipe inlet (5.7 feet for Intake 1) V = average flow velocity in the 14 inch diameter pipe. The intake loss coefficient varied between 0.34 and 0.39 with no trend in the values with blockage cases or flow rates. The consistency of the loss FNP-FSAR-6C 6C-9 REV 21 5/08 coefficients indicated that no discernible flow reduction developed because of circulation or vortex action within the screen area. 4. Final Documentation Tests An octagonally shaped grating skirt of 1 1/4-in. by 3/16-in. bars on 1 3/16-in. centers was placed around the intake within the screened area to eliminate the horizontal vortex which developed inside of the screen structure during initial tests. The intake design is shown on figure 6C-6. Since the objectionable vortex action only occurred for blockage condition 5, final tests were only conducted for this case. The full range of test conditions, tests 1 through 8, table 1, were documented at water temperatures of 170°F, or greater. The test results were as follows: a. A weak circulation with a horizontal axis was observed at the location where the vapor core developed during initial tests. However, no vapor core formed for any test condition. As noted previously, there were no vortices formed within the screen area for any other blockage condition. b. The intake loss coefficient, as previously defined, remained between 0.34 and 0.39. Additional measurements were made utilizing piezometric taps at 29 diameters downstream of the intake to further isolate the full loss of the intake and the elbow, which has a centerline radius of about 1.5 times the pipe diameter. The following formula, which includes the correction for pipe friction, was used: g2Vg2VhhK22f= where hf is the computed pressure drop in feet of water due to pipe friction above, based on the estimated pipe surface roughness, height, and other terms as defined before. For intake discharges from 3060 to 7915 gal/min, the loss coefficient for the intake and the elbow ranged from 0.46 to 0.48. Intakes 2, 3, and 4 The documented tests were conducted with the improved screen and grating and grating cages placed over Intakes 2, 3, and 4 as shown in figures 6C-12 and 14.

FNP-FSAR-6C 6C-10 REV 21 5/08 The full range of test conditions, tests 1 through 10, table 6C-2, with unblocked, and blocked screens were run and documented at water temperatures of 180°F or greater. The initial test results indicated a substantially higher loss coefficient for intake plus bend for Intake 3 than for Intakes 2 and 4. The loss coefficient was computed from the equation given in subsection D.3.d, above. An inspection revealed that the model pipe walls of Intake 3 had been severely corroded by the hot water, with the tuberculated pipe indicating protrusions measuring 1/32 to 1/16 inch. An additional pressure tap was installed to record the actual pressure loss over a 9.51-foot section of pipe. The h value of Intake 3 was then calculated from the measured pressure drop values in the pipe. The test results were as follows: a. There were no vortices under any test condition which established an air core from the free surface within the containment area to the screen grating around the intakes. Surface depressions in the eye of any eddies or organized circulation did not exceed 1/2 inch in depth. b. No vortices were observed inside or outside the screen structure. c. Air introduced artificially under the cover plate above the intake, which represented air trapped during flooding of the containment or coming out of solution, was able to escape through the 1/2-inch vent slot in the cover plate. At augmented discharges, pockets of air would remain trapped below the plate. This air swirled above the intake, but no air was drawn into the intake irrespective of the quantity of air forced beneath the plate. At prototype discharges, this air was able to escape through the 1/2-inch slot, and only a few small bubbles remained. d. The maximum measured head loss across the screen and grating corrected to prototype discharge of 5900 gal/min for Intake 2 and 3050 gal/min for Intake 3 was 0.14 foot for Intakes 2 and 3 and 0.04 foot for Intake 4 with a discharge of 3050 gal/min. The intake plus bend loss coefficients varied from 0.36 to 0.46 for Intake 2 0.38 to 0.45 for Intake 3 0.33 to 0.40 for Intake 4, with no trend in the values with blockage cases or flow rates. The model indicated maximum combined prototype losses due to screen, intake, and bend of FNP-FSAR-6C 6C-11 REV 21 5/08 1.50 ft for Intake 2 1.25 ft for Intake 3 1.01 ft for Intake 4. The higher combined losses associated with Intakes 2 and 3 were attributed to larger flow per unit area approaching the intakes and the more turbulent approach condition resulting from the proximity of these intakes to the elevator shaft. E. FIELD TEST Several preoperational tests were performed at the Joseph M. Farley Nuclear Plant, Unit 1 to determine the actual piping resistance of the residual heat removal (RHR) pump sump suction lines. The testing revealed that the maximum expected RHR pump flow during the post-LOCA cold leg recirculation mode with only one RHR pump in operation was 5000 gal/min for Pump A and 4875 gal/min for Pump B. The actual net positive suction head (NPSH) available to each RHR pump from the containment sump was determined to be 18.4 feet at 5000 gal/min without taking credit for subcooling of the water in the containment sump and based upon the most resistive sump piping (25.2 feet elevation head from the sump and 3.9 feet of water above the sump line inlet less 10.7 feet of losses). The NPSH required for the RHR pump is 18.5 feet at 5000 gal/min and 18.0 feet at 4875 gal/min. This indicated that the NPSH available to the RHR pumps under worst case conditions would be marginal during the post-LOCA recirculation phase. Upon the completion of additional tests confirming the resistance of the installed piping system, the RHR system resistance was increased to assure that adequate NPSH is available and that system performance is satisfactory during all operating modes. The system resistance was increased by physically restricting the maximum opening of valves HCV-603A and B on the outlet piping of the RHR heat exchangers and by addition of flow restriction orifices in each of the three cold leg low head safety injection lines. System tests conducted after these modifications show that the maximum flowrate with one pump operating during the cold leg recirculation mode of operation would be approximately 4200 gal/min. The NPSH available for RHR Pumps A and B utilizing simulated recirculation mode plant test data, at this flowrate, is 17.7 feet (25.2 feet elevation head from the sump less 7.5. feet of losses) and 19.2 feet (25.2 feet elevation head from the sump less 6.0 feet of losses) respectively. The NPSH required for the RHR pump is 15.0 feet at 4200 gal/min. Thus, adequate NPSH is assured. These calculations take no credit for water above the containment sump line inlet or for any subcooling of water in the containment sump. Evaluation of the postmodification tests also confirmed that ECCS flows would meet or exceed system requirements during all operating modes. F. SUMMARY AND CONCLUSIONS The 1:1 scale model of Intake 1, (figures 6C-5 and 6), which was tested at Reynolds numbers equal to or greater than prototype and with circulations which were greater FNP-FSAR-6C 6C-12 REV 21 5/08 than prototype, indicated that the intake will operate without air entraining or flow reducing vortices. The maximum screen grating and intake losses computed from the model test results were 0.09 foot and 0.85 foot respectively at 5900 gal/min. These values were combined with field test data and compared with calculated data used in the NPSH evaluation. The 1:1 scale model of Intakes 2, 3, and 4, (figure 6C-7), which was tested at Reynolds numbers equal to or greater than prototype and with circulations which were greater than prototype, indicated that the intake will operate without air entraining or flow reducing vortices. The maximum losses determined from the model and field tests for each intake are: Pressure Drop (feet) Intake Effect 1 2 3 4 Piping (from field data)(1, 2) 5.32 3.86 6.28 6.89 Inlet (from test data)(1) 0.43 0.37 0.36 0.43 (from test data)(1) 1.49 1.49 2.39 2.39 Screen (from test data) 0.04 0.09 0.09 0.04 Total 7.28 5.81 9.12 9.75 NOTES: 1. Converted to 4200 gal/min base for Intakes 1 and 2 and to 3050 gal/min base for Intakes 3 and 4. 2. Includes additional losses due to 8 feet of test piping for Intakes 1 and 2 and additional losses due to 6 feet of test piping for Intakes 3 and 4. The measured head losses are less than the calculated losses of 8.4 feet for Intakes 1 and 2 and 9.9 feet for Intakes 3 and 4. (See subsection 6.3.2.14.) Based on the results of Intake 1 tests, together with Intakes 2, 3, and 4 tests of Unit 1 and on similar work undertaken for other projects, it is the definite opinion of Western Canada Hydraulic Laboratories Ltd. and Bechtel that incorporating a grating cage similar to the above design, (figures 6C-6 and 12), will result in an intake design for all Units 1 and 2 intakes which will operate free from air entraining or flow reducing vortices.

FNP-FSAR-6C 6C-13 REV 21 5/08 IV. UNIT 2 TESTS A. INTRODUCTION This section presents the model test program undertaken and the results of these tests to ensure that Joseph M. Farley Unit 2 emergency core cooling and containment spray system recirculation intakes from the containment sump (floor) will operate without effects which could degrade the performance of the pumps in these systems. Similar tests were conducted for the Farley Unit 1 containment sump intakes in which the containment geometry in the sump areas was modeled at a scale of 1:1 together with flow obstructions such as pipes, supports and valves, etc. around these intakes. These tests revealed that there were no air entraining vortices or flow reducing conditions at these intakes when these intakes were protected with the inner grating cage and the outer screen grating cage structure combination. This is discussed in detail in section II of this appendix. The tests performed on containment sump intakes of Unit 1 and on other facilities provided strong evidence that the inner grating cage and the outer screen grating cage structure combination employed on Unit 1 was totally effective in destroying vortices ranging from pencil lead size to 1 inch in diameter or greater. Based on these results, it was concluded that the tests on Unit 2 containment sump intakes can be effectively performed without modeling the containment geometry around the intakes as was done for Unit 1. A more detailed rationale for this approach along with a description of Unit 2 containment sump intakes, discussion of effects which could degrade pump performance, description of test facility and program, and the test result and conclusions are presented in this report. B. DESCRIPTION OF UNIT 2 CONTAINMENT SUMP RECIRCULATION INTAKES The emergency core cooling and containment spray system recirculation intakes for Unit 2 are comprised of two 14-inch and two 10-inch vertical inlets located in four separate intake areas on the containment floor and are designated as Intakes 1, 2, 3 and 4 as shown in figure 6C-19. Each intake is surrounded by a protective screen grating and grating cage structure as shown in figure 6C-17. The design flows for Intakes 1 and 2, which supply water to the RHR pumps, range from 3000 to 5900 gal/min each. The design flow rates for Intakes 3 and 4, which supply water to containment spray pumps, are 3050 gal/min each. The calculated minimum and maximum water levels in the containment are 58.3 and 77.1 inches, respectively, above the floor. The maximum expected containment sump water temperature during recirculation following a postulated LOCA is about 212°F at subcooled pressures. The flowrates, water depths, water temperature, and the protective screen structure for Unit 2 are identical to Unit 1, except that four separate intake areas are provided for Unit 2 as compared to three intake areas for Unit 1. The elevator shaft in Unit 2 is FNP-FSAR-6C 6C-14 REV 21 5/08 located outside the flow paths approaching the intakes. This will lead to a more uniform flow in the containment sump intake areas than that expected in Unit 1, where the elevator shaft is located in the containment sump intake area. Furthermore, the equipment layout at the Unit 2 containment floor elevation, as shown in figure 6C-20, is not expected to be significantly different from Unit 1, shown in figure 6C-21. C. PROBLEM DEFINITION Regulatory Guide 1.82 states the position that "Pump intake locations in the sump should be carefully considered to prevent degrading effects, such as vortexing, on the pump performance." Two degrading actions are possible: ingestion of air (a vortex phenomenon), and/or intake entrance losses which are larger than design values used in establishing the required NPSH of the pumps. Increased entrance loss can develop due to adverse flow approach conditions or free surface and internal vortex action. 1. Factors Causing Increased Entrance Losses Intake losses are incurred due to contraction and expansion of the flow at the intake. The intake entrance losses are accounted for in the design of pumping systems by calculating the entrance loss based on established intake loss coefficients. Such coefficients are normally based upon measurements taken with uniform flow approaching the intake. Intake head losses can be increased by high approach velocities, especially at an angle to the pipe axis and/or by strong circulation in the approach flow which results in an increased contraction of the flow at the intake. Strong circulation can lead to vortex formation with a marked reduction in flow. A full scale model is capable of indicating any head loss degrading effects for all conditions simulated and tested. 2. Factors Affecting Vortex Creation Studies of vortex formation have been carried on by several investigators (see references in part II, G). The majority present test results as functions of the intake head loss coefficient, the depth of water at which the air core just penetrates the intake, the circulation numbers at which the air core just penetrates the intake, the Reynolds Number, or some variation of these parameters. The performance of an intake, as represented by the head loss coefficient K, is usually described (Anwar 1968, Amphlett 1976, Chang 1976) as: K = F (local geometry, rmax RR, N, W)

FNP-FSAR-6C 6C-15 REV 21 5/08 where local geometry = f (D, h, b) HQ.NoReynoldsRadialRR= NCirculationNoDQWWeberNoQDh===..224 rmax = radius of the tank (or sump) in which the intake is located = maximum radius of circulation in the vicinity of the intake Q = discharge D = intake diameter b = height of intake above sump floor = density of water = surface tension of water = surface tension of water h = depth of submergence of intake = circulation strength = 2Vtr where V is tangential velocity at radius r. Work by Daggett and Kuelegan (1974) and others have shown that for high Reynolds numbers (RR >104) and moderate values of circulation (N 2), typical operation ranges for the Farley recirculation intakes, the effects of surface tension and viscosity are relatively small; i.e., W and RR are not important. In this case, the intake performance, and hence the formation of vortices, is a function of three parameters: the local geometry, the maximum circulation radius, and the strength of circulation of the approaching flow. Each of these factors is discussed in the following sections. D. TEST PROGRAM 1. Rationale As discussed in subsection C, the intake head losses may be increased by nonuniform flow and/or circulation in the approach flow into the intakes.

FNP-FSAR-6C 6C-16 REV 21 5/08 It is significant to note that ultimately it is the flow condition in the immediate vicinity of the intake pipe that establishes the intake head loss. This condition, and associated head losses, may or may not be affected by the flow conditions removed from the immediate vicinity of the intake pipe. With an intake that is not protected by a screen grating cage, the flow in the immediate vicinity of the intake pipe will be established by the structural configuration of the containment and affected by the presence of flow obstructions such as valves, piping and restraints. The effect of these flow obstructions will increase with their increased proximity to the intake. Structural members may channelize the approach flow, affecting the approach flow directions and velocities. Channelization can also lead to a general circulation in the vicinity of the intakes, being bounded by the surrounding structures. Eddy-shedding will induce vorticity in the flow which can add to circulation in the vicinity of the intake. Unquestionably, an intake that is not protected by a screen grating cage can only be tested with full representation of the structural configuration of the containment and valves, piping and restraints. However, the Farley Unit 2 containment sump recirculation intakes are to be covered and protected by a screen grating cage, comprised of a 0.047 inch screen wire with an effective opening of 51.6 percent, sandwiched between two layers of grating. The grating bars will be 1-1/4 inch by 3/16 inch on 1-3/16 inch centers, giving a total effective grating width in the direction of flow of 2-1/2 inch. (See figure 6C-17.) The inside grating bars are approximately 2 feet from the intake pipe. Furthermore, an inner grating cage will be placed over the intake pipe. Due to the proximity of the Farley screen grating cage to the intake pipe, it was concluded that this structure would strongly influence, if not dominate, the approach flow into the intake pipes within the cage. This dominance was observed on full scale mockup tests of the Farley Unit 1 containment sump recirculation intakes where: a. The grating bars acted as flow straighteners and no angularity or circulation of flow approaching the screen cage, which could lead to the formation of a vortex, was transmitted through the structure, regardless of the angle of approach flow. Flow downstream from the grating exited at right angles to the plane of the grating. b. The most nonuniform, rotational approach flow to the intake pipe, as evidenced by an air core vortex inside the cage, was developed by a partial screen blockage configuration. No vortex developed inside the screen cage without blockage. Hence, it was apparent from the Unit 1 tests that the approach conditions in the immediate vicinity of the intakes (within the screen grating cage) were established by the flow distribution through the screen grating cage. Angularity of flow approaching the outside of the screen grating was removed and any swirl or circulation inside the screen grating cage was due to a nonuniform flow FNP-FSAR-6C 6C-17 REV 21 5/08 distribution through and normal to the plane of the grating. Furthermore, the most adverse velocity distribution inside the screen grating cage could be established by the blockage configuration imposed. These observations lead to the following considerations: a. In the case of an intake not protected by a screen grating cage, the flow can be channelized at any angle by structural members. In the case of an intake covered by a screen grating cage, the screen grating and the blockage configuration impose the ultimate channelization, and establish the direction of flow normal to the plane of the grating. b. Irrespective of the structural configuration external to the screen grating cage, and hence irrespective of the approach flow conditions this configuration imposes on an unblocked screen grating structure, there will be a blockage condition which will develop as adverse or a more adverse and potentially a more degrading effect on the intake performance. Hence, this proves that the grating gage will eliminate any vortex potential and would be proof that the potential developed by the external structural arrangement will be eliminated. c. Because blockage conditions could establish potentially degrading conditions inside the screen grating cage, a grating cage must be incorporated inside the screen grating cage to remove circulation generated within the screen grating cage which could lead to the formation of vortices and/or increased intake head losses. Thus, based on the experience gained on the full scale mockup tests for Unit 1, the following rationale was applied to the test program for the Unit 2 containment sump recirculation intakes: a. Ultimately it is the flow condition in the immediate vicinity of the intake pipe that can lead to degrading effects of pump performance. b. The immediate vicinity of the intakes will be covered by a screen grating cage. c. If the screen grating cage does not transmit the angularity of circulation of flow outside of the cage, then flow conditions and air core vortex potential within the screen grating cage are established by the blockage conditions imposed (flow distribution), water depth (pressure inside the cage), intake discharge (velocities), and viscosity (fluid shear energy dissipation). d. The fact the containment may be pressurized does not affect flow conditions. The flow field is established by pressure differentials which would be the same in a closed system irrespective of the air pressure on the water surface.

FNP-FSAR-6C 6C-18 REV 21 5/08 e. If angularity of approach flow is not transmitted through the screen grating cage, then the uniqueness of the flow distribution through an open screen established by the structural configuration and flow obstructions surrounding the screens represents one potential blockage condition. f. Based on the above, it is necessary to model only the screen grating cage, and all features inside the cage, and demonstrate for postulated flow depths and flow rates that: i. The screen grating cage will not transmit the angularity or circulation of flows outside the cage. ii. Under adverse conditions generated by screen blockage, the grating cage over the intake inside the screen grating cage will preclude degrading effects on the performance of the recirculation pumps. g. Furthermore, since circulation is an essential and necessary feature of a vortex, then irrespective of the strength of circulation, if flow circulation associated with a potential vortex is not transmitted through the screen grating cage, then the vortex formed outside of the cage cannot enter the intake pipe (as discussed in the Final Report on the Davis-Besse Nuclear Power Station ECCS Emergency Pumps and Pump Suction Line Testing, December 15, 1976). As discussed in subsection C, the formation of vortices is a function of three parameters: the local geometry, the maximum circulation radius, and the strength of circulation of the approaching flow. Since full scale tests were to be conducted, the local geometry in the immediate vicinity of the intake would be correctly simulated. In addition, since all screen and grating characteristics would be correctly represented, all vortex and flow parameters from the screen grating structure inward to the intake would be correctly simulated, and the intake entrance losses would be correctly measured. Swirls in the approach flow may vary with respect to the absolute size of the system, strength of circulation, velocity of translation, and travel path. The latter two parameters are of significance since, for a swirl to initiate an intake vortex, it must remain in the vicinity of the intake long enough to organize the circulation in the vicinity of the intake. Hence a stationary circulation directly above the intake becomes the critical case. The system size is of no concern when a 1:1 scale model is used. Thus there are two parameters which must be properly addressed: the maximum circulation radius (rmax), and the strength of circulation. Experimental evidence indicates that the critical submergence of the intake required to preclude the formation of air entraining vortices increases with both the maximum swirl radius, rmax, (Haindl, 1959) and strength of the initiating swirl (Amphlett, 1976), Daggett and Kuelegan, 1974; Springer & Peterson, 1969; Anwar, 1965).

FNP-FSAR-6C 6C-19 REV 21 5/08 From these experimental data, it can be concluded that a test procedure should include: a. A stationary circulation directly over the intake. b. A circulation strength equal to or greater than the largest reasonable values due to the expected prototype approach flow configuration. c. A maximum circulation radius, rmax, equal to the largest reasonable value in the prototype, must be said to "bound" the effects developed y the plant geometry and structural members, etc., in the vicinity of the intakes which could lead to a vortex. 2. Objective The prime objective of the test program was to demonstrate that the Farley Unit 2 containment sump recirculation intakes will not be subjected to degrading effects on pump performance, such as air ingestion or high intake head losses. Achieving the following fulfilled the prime objective: a. Documentation of the effectiveness of the grating cage over the intake in straightening the approach flow and removing imposed angularity or circulation which, without the grating cage present, could lead to an air entraining vortex. b. Documenting that the screen grating cage removed angularity and circulation of approach flow outside of the cage. Documentation of the effectiveness of the grating cage was achieved by: i. Imposing on the grating cage, without the screen grating cage over it, a range of circulations, the largest of which was more massive than any circulation that could be developed by the geometry or the structural members of the containment or the presence of flow obstruction such as valves, piping and restraints. ii. Imposing blockage conditions on the screen grating cage which generated potentially degrading flow conditions within the screen cage, and documenting that those conditions were eliminated by the grating cage. Documentation of the effectiveness of the screen grating cage was achieved by: i. Demonstrating that the single layer of grating on the grating cage was effective in removing angularity in the approach flow in the high velocity region close to the intake.

FNP-FSAR-6C 6C-20 REV 21 5/08 ii. Demonstrating that no vortex, angularity or circulation of approach flow passed through the screen grating cage or grating cage when the screen grating cage was subjected to a range of circulation, the largest of which was more massive than any circulation that could be developed by the geometry or the structural members of the containment, or by the presence of flow obstructions such as valves, piping and restraints. E. TEST FACILITY 1. General The plan view and a section of the experimental facility are shown in figures 6C-15 and 6C-16. Two source sumps, each containing a diffuser, provided the approach flow to the intake area within the concrete tank, (figure 6C-15). A sump floor, which was of 1/8 in. steel plate, was placed 4.5 feet above the tank floor to provide space for the 14 in. diameter intake piping and for an observation tunnel below the steel plate floor. (See figure 6C-16.) The flows were distributed and controlled by means of two centrifugal pumps and a flow transmitting network of steel pipes, orifice meters with differential mercury manometers, and control valves. The direction and circulation of the approach flow was controlled by a system of 18-in.-wide vertical directional vanes, which extended over the full depth of the flow. (See figure 6C-16.) Two 2,900,000 Btu/h gas heaters were used to heat the water to temperatures in excess of 180°F. 2. Intake Description A cruciform and reducer section, which was shipped from the project site for use in the experimental facility, was mounted on the intake pipe 6 in. above the steel plate floor. The octagonal grating cage used for the Unit 1 model tests was modified to include a horizontal grating inside the grating gage, 3 in. above the floor, to eliminate potential floor vortices. (See figure 6C-18.) The grating cage thus totally encapsulated the intake pipe. The whole assembly was enclosed by a steel screen grating cage with inside dimensions of 5-ft. x 5-ft. x 2-ft. 5-in. depth. (See figure 6C-17.) The floor below the screen grating was of acrylic plastic construction which, together with the portholes in the observation tunnel, permitted observation and lighting of the area inside the screen grating. 3. Test Cases and Procedures a. Test Cases The postulated post-LOCA condition and the condition for which the intake was tested are compared below:

FNP-FSAR-6C 6C-21 REV 21 5/08 Postulated for Containment Sump post-LOCA Tested Minimum water depth (in.) 58.3 24.0 to 58.3 Maximum flow (gal/min) 5900 6574 to 8524 Water temperature (°F) 212 61 to 184 Maximum circulation (ft2/s) 3./for 58.3 in. 8.5 to 10.7 water depth for 58.3 in. water depth Maximum size of 17 18 circulation cell (ft) Pipe Reynolds Number 4.7 x 106 1.5 to 5.7 x 106 Considerable conservatism was incorporated in the test by: i. Conducting tests at greater than postulated flow rates. ii. Conducting tests with screen blockage greater than 50 percent. iii. Conducting tests at less than the minimum postulated water depths. iv. Conducting tests with a circulation appreciably greater than the maximum value calculated for the plant during LOCA conditions. v. Augmenting the postulated flows to develop Reynolds numbers in the test facility greater than postulated in the containment. Furthermore, model scale effects were reduced or eliminated by: i. Constructing the intake, grating cage, and screen grating cage at a 1:1 scale, thereby eliminating all scale effects introduced by modeling the screen and grating components. ii. Conducting the tests with water heated to 180°F or greater. b. Observations and Measurements All surface flow phenomena were observed from two platform decks. The lower deck was used to make all surface flow observations and to take velocity and temperature measurements. Observations could be made of flow phenomena inside the screen grating through the acrylic plastic cover plate. Video records were also made from this deck. Overhead photos were made from the upper platform deck. Flow phenomena within the screen grating cage could be observed and recorded on video tape through the portholes in the observation tunnel.

FNP-FSAR-6C 6C-22 REV 21 5/08 Use was made of air bubbles injected into the screen grating area through the acrylic plastic floor for flow visualization. Dyes were used sparingly to preserve water clarity. Flow measurements were obtained with the calibrated orifices and U-tube mercury manometers. Local velocities were measured with a Gurley propeller meter while surface velocities were obtained with the Gurley meter or from overhead photos of confetti traces. Pipe intake and screen grating losses were determined from static pressure taps. Taps 1 and 2 were located in the supply sumps and indicated the water surface elevation. Tap 3 consisted of two interconnected taps in the floor inside the screen grating cage to produce an average pressure within the screened area. The mean static head indicated by Tap 3 therefore gave an indication of screen grating losses when compared to the mean water surface elevation from Taps 1 and 2. Two interconnected taps, each on the horizontal diameter, determined the average static pressure inside the intake pipe at each of four locations, at distances of 5.11, 13.11, 21.11, and 30.00 diameters downstream of the intake. c. Test Procedure The tests were conducted in two phases. Phase 1 tests were related to documenting the effect of grating on approach flow conditions external to the screen grating cage. For this series, without any screen or grating over the intake, given flows were set and the vanes surrounding the intake were adjusted to produce the maximum size vortex. The circulation, vortex size and pressure measurements were taken, together with observations of flow conditions in the immediate vicinity of the intake. This was done for both ambient and heated water. Tests at ambient water temperatures were conducted to facilitate the making of video records of the free surface flow conditions. Without changing the vane angle, the tests were rerun and the results were documented with the addition of: i. Cruciform only. ii. Grating cage and cruciform only. iii. Screen grating cage and cruciform only. iv. Screen grating cage, grating cage and cruciform.

FNP-FSAR-6C 6C-23 REV 21 5/08 Phase 2 tests were related to documenting the effect of the grating cage on adverse flow conditions generated within the screen grating cage. With the screen grating cage and cruciform in place, blockage was placed on the screen to produce the largest internal vortices achievable. Pressure measurements were then taken and observations made. The grating cage was then installed and data were recorded for the identical conditions which previously had produced internal vortices. In summary, test procedures were developed for the Farley Unit 2 recirculation takes which: i. Modeled all effects of the screen grating cage and grating cage in the immediate vicinity of the intake on a 1:1 basis. ii. Allowed testing for the effects of the containment geometry and structural members, etc., by subjecting the intake to a range of circulation, the largest of which was greater than will occur in the prototype approach flow. iii. Demonstrated satisfactory intake performance under unrealistically severe conditions of water depth and circulation. F. TEST RESULTS 1. Phase 1 Test Results a. Unprotected Intake An air entraining vortex was easily formed over the unprotected intake pipe. With a water depth of 58 in. and an intake flow of approximately 7400 gal/min, the air entraining vortex was present intermittently when the flow vanes were aligned radially to the intake. The vortex increased in strength and became stable as the vanes were turned from the radial direction. The maximum vortex occurred with the vanes turned 48 degrees in either direction. The air core diameter of the vortex at the intake was 1.5 to 2 in. with a circulation of 8.5 ft.2/s as compared to a maximum calculated prototype value of 5.4 ft.2/s. With the vane angle set at 48 degrees, reducing the intake discharge from approximately 7400 gal/min to approximately 5300 gal/min reduced the diameter of the air core at the intake to 1 to 1.25 in. b. Intake with Cruciform The cruciform, by itself, did not eliminate air entraining vortices with a water depth of 58 in. The vortices were not as stable as without the FNP-FSAR-6C 6C-24 REV 21 5/08 cruciform; nevertheless, the following air core sizes were observed at the intake: Water Vane Air Core Flow Temp. Angle Diameter gal/min °F ° in. 7344 to 8457 65 48 1/2 to 3/4 7412 to 8088 65 0 1/8 to 3/8 7018 to 8446 180 48 3/4 to 1-1/2 The circulation for a flow of 8456 gal/min was 9.1 ft2/s. The intake loss coefficient, K, was 0.69. The maximum intake loss coefficient, K, for the heated water was 0.73 and the average intake loss coefficient was 0.72. c. Intake and Cruciform Protected by the Grating Cage No air entraining vortex penetrated the grating cage with the vanes set at 48 degrees and a water depth of 58 in. The intake flows tested were 7420 gal/min to 8513 gal/min with a water temperature of 119°F to 181°F and 8487 gal/min with a water temperature of 64°F. The flow circulation established by the vanes remained around the grating cage with the water surface depressed approximately 1 in. at the center. Bubbles or particulates in the flow surrounding the cage, which approached at an angle to the cage, were observed to exit at right angles to the plane of the grating. The average intake loss coefficient was reduced from 0.72 with only the cruciform to 0.65 with the grating cage. The maximum intake loss coefficient was 0.66. With an intake flow of 8400 gal/min, no air entraining vortex was produced when the water level was lowered from 58 inches to 24 inches. d. Intake and Cruciform Protected by Screen Grating Cage No air entraining vortex penetrated the screen grating cage for flows of 6574 to 8487 gal/min, vane angles of 0° and 48°, a water depth of 58 in., and water temperatures of 61°F to 64°F, and 173°F to 180°F. The maximum circulation was 10.7 ft.2/s.

FNP-FSAR-6C 6C-25 REV 21 5/08 The circulation outside of the screen-grating cage was not transmitted through the structure. e. Intake and Cruciform Protected by a Grating Cage and Screen Grating Cage No air entraining vortex penetrated the screen grating cage for flows of 7741 gal/min to 8460 gal/min, vane angle of 48°, a water depth of 58 in., and water temperatures of 64° F and 184°F. There was no organized circulation inside the unblocked screen grating cage. The maximum intake loss coefficient was 0.67 and the average coefficient was 0.66. The maximum screen head loss coefficient Ks with or without blockage was 10.2. The Ks values indicated a decreasing trend with increasing screen Reynolds number. 2. Phase 2 Tests The following summarizes the results of the Phase 2 tests: a. Intake and Cruciform Protected by Screen Grating Cage by Without Grating Cage Organized circulation could be established within the screen grating cage by selective blockage of the screen. Air core vortices were established by screen blockage of 61 to 71 percent for intake flows of 7461 gal/min to 8420 gal/min and water temperatures of 150°F to 177°F. The water depth was 58 in. Internal vortices could be formed from the floor, inside blockage plates (i.e., simulated walls), and the cover plate on the screen grating cage. One to five vortices could be generated simultaneously depending upon the blockage condition. A smooth surface within the screen grating cage was required to form an internal vortex. The air core diameter of the internal vortices varied from 1/8 in. to 1/4 in. The average intake loss coefficient was 0.69 and the maximum coefficient was 0.78. Internal vortices did not increase intake losses.

FNP-FSAR-6C 6C-26 REV 21 5/08 b. Intake and Cruciform Protected by Screen Grating Cage and Grating Cage Installation of the grating cage over the intake and cruciform completely eliminated all the internal vortices previously generated by the screen blockage and flow conditions discussed in subsection F.2-a. Flow circulation between the screen grating cage and grating cage, generated by the blockage, was not transmitted through the grating cage as evidenced by observing particulates in the flow. The average intake loss coefficient with the screen cage, grating cage and cruciform in place was 0.66 and the maximum coefficient was 0.67. G. SUMMARY AND CONCLUSIONS The recirculation intake designs to be used for Farley Unit 2 were tested under flow and vortex producing conditions which were potentially more degrading on pump performance than any condition possible in the prototype. The screen grating and inner grating cage were modeled at a 1:1 scale. The following results were obtained: 1. Vortex Action The screen grating cage will not permit any free surface air entraining vortices to form through which air can be ingested by the intake. Circulation (which is an essential feature of a vortex), or approach flow angularity, were not transmitted through the screen grating cage. The grating used in the screen grating cage was totally effective in eliminating any vortex with air core diameters of 1/8 in. to 2 in., which would have otherwise formed without the presence of the screen grating cage. Without the inner grating cage, internal vortices could be developed by selective screen blockage. These vortices, which were formed only from smooth surfaces, did not increase intake entrance losses. However, with the grating cage in place as proposed for the Farley Unit 2 design, no internal vortices will develop. Circulation developed within the screen grating cage, which could lead to internal vortices, was not transmitted inside of the grating cage. 2. Head Loss Coefficients The screen grating cage, grating cage, and cruciform protective design will have a head loss coefficient for the combined grating cage, intake and 90° pipe bend of 0.67, even with screen blockages in excess of 50 percent. The maximum measured intake loss coefficients were as follows: Cruciform 0.73 FNP-FSAR-6C 6C-27 REV 21 5/08 Grating cage and cruciform 0.66 Screen grating cage, grating cage and cruciform 0.67 Screen grating cage head losses are small with the maximum measured loss coefficient in the model being 10.2. 3. Losses The maximum losses determined from the model test and calculations for each intake are: Effect Pressure Drop (feet) Intake 1 2 3 4 Piping (calculated)(1, 2) 5.89 4.27 6.47 7.09 Inlet (from test data)(1) 0.75 0.75 1.02 1.02 (From test data)(1) 1.48 1.48 2.40 2.40 Screen (from test data) 0.09 0.03 0.03 0.05 Total 8.21 6.53 9.92 10.56 NOTES: 1. Converted to 4200 gal/min base for Intakes 1 and 2 and to 3050 gal/min base for Intakes 3 and 4. 2. These are calculated numbers and will be verified by a field test. However, comparison of the calculated values with the field test data for Unit 1 indicates that the calculated values are conservative (See section 6C.III.F).

FNP-FSAR-6C 6C-28 REV 21 5/08 REFERENCES 1. Addison, H., 1948; "Centrifugal and Other Rotodynamic Pumps," (Chapman and Hall, London.) 2. Akers and Crump, "The Vortex Drip," Journal, Institution of Civil Engineers, p 443, August 1960. 3. Al'Tshul, A. D. and Margonlin, M. S., "Effect of Vortices on the Discharge Coefficient for Flow of a Liquid Through an Orifice" (Translation), Gidrotekhnieheskoe Stroitel'stvi No. 6, p 32, June 1968. 4. Amphlett, M. B., "Air Entraining Vortices at a Horizontal Intake," Report No. OD/7, Hydraulic Research Station, Wallingford, April 1976. 5. Anwar, H. O., "Flow in a Free Vortex", Water Power, April 1965. 6. Anwar, H. O., "Formation of a Weak Vortex", Journal of Hydraulic Research, Vol. 4, No. 1, 1966. 7. Anwar, H. O., "Vortices at Low Head Intakes", Water Power, pp 455-457, November 1967. 8. Anwar, H. O., "Prevention of Vortices at Intakes," Water Power, p 393, October 1968. 9. Anwar, H. O., Discussion of "Effect of Viscosity on Vortex-Orifice Flow," by Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulics Division, ASCE Vol. 95, No. HY 1, Proc. Paper 6323, pp 568-570, January 1969. 10. Berge, J. P., "Enquete sur la Formation de Vortex et Autres Anomalies d'ecoulement dans une enceinte avec ou sans Surface Libre," Societe Hydrotechnique de France-Section Machines - Group de travail No. 10, November 1964. 11. Berge, J. P., "A Study of Vortex Formation and Other Abnormal Flow in a Tank With and Without a Free Surface," La Huille Blanche, Grenoble, France, No. 1, pp 13-40, 1966. 12. Binnie, A. M., and Hockings, G. G., "Laboratory Experiments on Whirlpools," Proceedings, Royal Society, London, Series A. Vol. 194, pp 398-415, September 1948. 13. Brewer, D. "Vortices in Pumps Sumps," The Allen Engineering Review, March 1957. 14. "Review of Literature on Drain Vortices in Cylindrical Tanks," Report TN1342, BHRA, March 1976 15. Denny, D. F., 1953 British Hydromechanics Research Assoc. Report, R. R. 430, Preliminary Report on the Formation of Air entraining Vortices in Pump Section Wells. 16. Denny, D. F. "Experiments with Air in Centrifugal Pumps", British Hydromechanics Research Assoc. Research Report R. R. 465, 1953. 17. Denny, D. F., and Young, G. A. J. "The Prevention of Vortices and Swirl and Intakes", Proceedings, IAHR 7th Congress, Libson, 1957.

FNP-FSAR-6C 6C-29 REV 21 5/08 18. Denny, D. F., "An Experimental Study of Air Entraining Vortices in Pump Sumps", Proceedings of Inst. of Mechanical Engineers, Vol. 170, No.2, 1956. 19. Daggett, L. L. and Keulegan, G. H., "Simulitude Conditions in Free Surface Vortex Formations", Journal of the Hydraulics Division, ASCE, Vol. 100, No. Hy11, Nov. 1974, pp 1565-1581. 20. Donaldson, C. du p., and Sullivan, R. D., "Examination of the Solutions of the Navier-Stokes Equations for a Class of Three-dimensional Vortices, Part I: Velocity Distribution for Steady Motion", Proceedings, Heat Transfer and Fluid Mechanics Institute, Stanford University Press, Calif., 1960, p. 16-30. 21. Einstein, H. A., and Li, H., "Steady Vortex Flow in a Real Fluid," La Huille Blanche, Vol. 10, No. 4, Aug.-Sept., 1955, p. 483-496. 22. Folsom, R. G., "Some Performance Characteristics of Deepwell Turbine Pumps," Technical Memo. No. 6, HP-14, University of California, Pump Testing Laboratories, 1940. 23. Fraser, W. H., "Hydraulic Problems Encountered in Intake Structures of Vertical Wet-Pit Pumps and Methods Leading to Their Solution," Trans. ASME, vol. 75, No. 4, p. 643, 1953. 24. Gordon, J. L., "Vortices at Intakes," Water Power, April 1970, p. 137-138. 25. Cuiton, P., "Cavitation dans les Pompes", La Huille Blanche, Nov., 1962, No. 6. 26. Haindl, K., "Contribution to Air-Entrainment by a Vortex", Paper 16-D International Association for Hydraulic Research, Montrel, 1959. 27. Hattersley, R. T., "Hydraulic Design of Pump Intakes", HY 2, March, 1965, P. 223-249. 28. Hattersley, R. T., "Factors of Inlet Channel Flow affecting the Performance of a Pumping Plant", Report No.23, Water Research Lab., University of New South Wales, Australia, Sept., 1960. 29. Holtorff, G., "The Free Surface and the Conditions of Similitude for a Vortex", La Houille Blache, vol. 19, No. 3, 1964, P. 377-384. 30. Iversen, H. W.,; "Studies of Submergency Requirements of High Specific Speed Pumps", Transactions, ASME, Vol. 75, 1953. 31. Kaufman, Fluid Mechanics, McGraw-Hill, p. 265 and 279. 32. Keulegan, G. H., and Daggett, L. L., "A Note on Gravity Head Viscometer", Miscellaneous Paper H-74-3, United States Army Engineer Waterways Experiment Station, Corps of Engineers, Vicksburg, Miss., Mar., 1974. 33. Kolf, R. C., "Vortex Flow from Horizontal Thin Plate Orifices", thesis presented to the University of Wisconsin, at Madison, Wis., in 1956, in partial fulfillment of the requirements for the degree of Doctor of Philosophy.

FNP-FSAR-6C 6C-30 REV 21 5/08 34. Kolf, R. C., and Zielinski, P. B., "The Vortex Chamber as an Automatic Flow Control Device", Journal of the Hydraulics Division, ASCE, vol. 85, No. HY12, Proc. Paper 2310, Dec. 1959. 35. Lawton, F. L. "Factors Influencing Flow in Large Conduits", Report of the Task Force on Flow in Large Conduits of the Committee on Hydraulic Structures. Transactions, ASCE, Paper 4543, vol. 91 HY 6, November, 1965. 36. Lennart, R., "Flow Problems with Respect to Intakes and tunnels of Swedish Hydro Electric Power Plants", Transactions, of the Royal Institute of technology, Stockholm, Sweden, NR 71, 1953. 37. Lewellen, W. S. "A Solution for Three-Dimensional Vortex Flow with Strong Circulation," J. Fluid Mechanics, vol. 14, 1962. 38. Long, R. R. "A Vortex in an Infinite Fluid," Journal of Fluid Mechanics, vol. 11. 39. Marklund, E. and Pipe, J. A. "Experiments on a Small Pump Suction Well, with Particular Reference to Vortex Formations". Proceedings, The Institution of Mechanical Engineers, vol. 170, 1956. 40. Marklund, E., discussion of "Effect of Viscosity on Vortex-Orifice Flow", by Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulic Division, ASCE, vol. 95, No. HY1, Proc. paper 6323, Jan., 1969, p. 567-568. 41. Messina, J. P., "Periodic Noise in Circulating Water Pumps", Power, Sept. 1971, p. 70-71. 42. McCorquodale, J. A., discussion of "Effect of Viscosity on Vortex-Orifice Flow," By Paul B. Zielinski and James R. Villemonte, Journal of the Hydraulics Division, ASCE, vol. 95, No. HY1, Proc. Paper 7323, Jan., 1969, p. 567-568. 43. McCorquodale, J. A., "Scale Effects in Swirling Flow", Journal of the Hydraulics Division, ASCE, vol. 94, HY1, Disc. by Marco Pica, HY1, Jan. 1969. 44. Pickford, J. A., and Reddy, Y. R. "Vortex Suppression in a Stilling Pond Over Flow", Journal of the Hydraulics Division, ASCE, vol. 100, No. HY11, Nov. 1974, pp 1685-1697. 45. Quick, M. C., "A Study of the Free Spiral Vortex", thesis presented to the University of Bristol, England, in 1961, in partial fulfillment of the requirements for the degree of Doctor of Philosophy. 46. Quick, M. C., "Scale Relationships between Geometrically Similar Free Spiral Vortices", Civil Engineering and Public Works Review, Part 1, September, 1962, Part II, October, 1962, p. 1319. 47. Quick, M. C., "Efficiency of Air Entraining Vortex Formation at Water Intake", Journal of the Hydraulics Div., ASCE, vol., No. 96, HY7, July 1970, p. 1403-1416. 48. Reddy, Y. R., and Pcikford, J. A., "Vortices at Intakes in Conventional Sumps", Water Power, March 1972,p. 108-109 FNP-FSAR-6C 6C-31 REV 21 5/08 49. Rouse, H., and Hsu, H., "On the Growth and Decay of a Vortex Filament", Proceedings, 1st National Congress of Applied Mechanics, 1952, p. 741-746. 50. Richardson, C. A., "Submergence and Spacing of Suction Bells," Water Works and Sewerage Reference and Data, Part 1, Water Supply, p. 25, 1941. 51. Springer, E. K., and Patterson, I. M., "Experimental Investigation of Critical Submergency for Vortexing in a Vertical Cylindrical Tank," ASME Paper 69-FF-49, June 1969. 52. Stepanoff, A. J., 1946 Centrifugal and Axial Flow Pumps, p. 963 (Chapman and Hall, London). 53. Stevens, J. C., discussion of "The Vortex Chamber as an Automatic Flow Control Device", by R. C. Kolf and P. B. Zielinski, Journal of the Hydraulics Division, ASCE, vol. 86, No. HY6, Proc. Paper 2525, June, 1960. 54. Stevens, J. C., and Kolf, R. C., "Vortex Flow Through Horizontal Orifices", Journal of the Sanitary Engineering Division, ASCE, Vol. 83, No. SA6, Proc. Paper 1461, Dec., 1957. 55. Weltmer, W. W. "Proper Suction Intakes Vital for Vertical Circulating Pumps," Power Engineering, vol. 54, No. 6, p. 74, 1950. 56. Young, C.A.H., "Swirl and Vortices at Intakes", Report No. SP 726, British Hydro-Mechanics Research Association, April, 1962. 57. Zelinski, P. B. and Vellimont, J. R., "Effect of Viscosity on Vortex Orifice Flow", vol. 94, HY3, May, 1968, p. 745-751. Disc. on above in Jan. 1969, by Marklund, E., McCorquodale, J. A. and Anwar, H.O. ]

FNP-FSAR-6C REV 21 5/08 [Historical] [TABLE 6C-1 TEST CONDITIONS FOR UNIT 1 INTAKE 1 Water Discharges - gal/min Test Depth Q1 Intake 1 Q2 Operating Water Temperature No. In. Prototype Model PrototypeModel PrototypeModel Pumps Prototype Model 1 58.3 3715 5500 4150 6150 -435 -644 1RHR 240 170+ 2 77.1 3540 5240 4150 6150 -610 -905 1RHR 240 170+ 3 58.3 5500 8140 3000 4440 2500 3700 2 RHR 240 170+ 4 77.1 5500 8140 3000 4440 2500 3700 2RHR 240 170+ 5 77.1 8750 12,450 4150 6150 4600 6808 1RHR,2S 240 170+ 6 77.1 10,500 14,300 5900 8000 4600 6200 2RHR,2S(a) 240 183+ 7 77.1 10,600 10,600 4150 4150 6450 6450 2RHR,1S 170 170+ 8 77.1 12,900 12,900 4150 4150 8750 8750 2RHR,2S 170 170+ RHR = Residual Heat S = Spray a. The two RHR pumps taking suction from one inlet]

FNP-FSAR-6C REV 21 5/08 [Historical] TABLE 6C-2 TEST CONDITIONS FOR UNIT 1 INTAKES 2, 3, AND 4 ]

REV 21 5/08 [TYPICAL ARRANGEMENT OF CONTAINMENT SUMP SUCTION LINE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-1]

REV 21 5/08 [MODELED AREAS OF ECCS INTAKES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-2]

REV 21 5/08 [HYDRAULIC MODEL PLAN FOR INTAKE NO. 1 TESTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-3]

REV 21 5/08 [BLOCKAGE TEST CONDITIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-4]

REV 21 5/08 [NO. 1 INTAKE CONFIGURATION FOR INITIAL TESTS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-5]

REV 21 5/08 [IMPROVED DESIGN INTAKE NO. 1 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-6]

REV 21 5/08 PLAN OF MODELED AREA CONTAINING INTAKES 2, 3, AND 4 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-7 REV 21 5/08 [CONTAINMENT SUMP JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-8]

REV 21 5/08 [PHOTOGRAPH OF MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-10]

REV 21 5/08 [PHOTOGRAPH OF MODEL JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-11]

REV 21 5/08 [INTAKES 2, 3, AND 4 IMPROVED DESIGN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-12]

REV 21 5/08 [BLOCKAGE TEST CONDITIONS FOR INTAKES 2, 3, 4 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-13]

REV 21 5/08 [PHOTOGRAPH OF GRATING CAGE OVER INTAKE 2 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-14]

REV 21 5/08 [PLAN OF UNIT 2 TEST FACILITY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-15]

REV 21 5/08 [SECTION OF UNIT 2 TEST FACILITY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-16]

REV 21 5/08 [REPRESENTATIVE SCREEN-GRATING STRUCTURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-17]

REV 21 5/08 [GRATING CAGE - FINAL DESIGN JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-18]

REV 21 5/08 [SUMP AREA OF UNIT 2 JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-19]

REV 21 5/08 [COMPOSITE DRAWING OF UNIT 2 SUMP JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-20]

REV 21 5/08 [COMPOSITE DRAWING OF UNIT 1 SUMP JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-21]

REV 21 5/08 [PHOTO OF UNIT 2 GRATING CAGE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-22]

REV 21 5/08 [PHOTO OF REPRESENTATIVE SCREEN - GRATING CAGE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6C-23]

FNP-FSAR-6D 6D-i REV 22 8/09 APPENDIX 6D CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION SUMP STRAINER DESIGN TABLE OF CONTENTS Page 6D.1 CONTAINMENT SUMP DESCRIPTION -----------------------------------------6D-1 6D.1.1 General Plant System Description--------------------------------------------6D-1

6D.1.2 General Description of New ECCS Strainers Installed------------------6D-2 6D.1.3 Size of New ECCS Strainers Installed---------------------------------------6D-2

6D.2 SUMMARY DESCRIPTION OF APPROACH USED TO SIZE SUMP STRAINERS--------------------------------------------------------------------------------6D-3 6D.2.1 Containment Walkdown---------------------------------------------------------6D-3 6D.2.2 Pipe Break Characterization---------------------------------------------------6D-3

6D.2.3 Debris Generation-----------------------------------------------------------------6D-3 6D.2.4 Latent Debris Accumulation within Containment--------------------------6D-4

6D.2.5 Debris Transport to the Sump-------------------------------------------------6D-4

6D.2.6 Head Loss as a Result of Debris Accumulation---------------------------6D-4 6D.2.7 Debris Source Term Reduction------------------------------------------------6D-5

6D.2.8 Sump Structural Analysis-------------------------------------------------------6D-5 6D.2.9 Upstream Effects of Debris Accumulation----------------------------------6D-5 6D.2.10 Downstream Effects - Components and Systems -----------------------6D-6 6D.2.11 Downstream Effects - Fuel and Vessel-------------------------------------6D-6 6D.2.12 Chemical Effects ------------------------------------------------------------------6D-6 REFERENCES--------------------------------------------------------------------------------------6D-8

FNP-FSAR-6D 6D-ii REV 22 8/09 LIST OF TABLES 6D-1 Containment Sump Debris Generation Zone of Influence (ZOI)

6D-2 Summary of LOCA Generated Insulation Debris Inside ZOI 6D-3 Debris Generated from Coating Based on ZOI = 4D

6D-4 Latent and Foreign Material Debris used in Analysis 6D-5 Summary of Debris Generated and Transported to Strainer Modules

FNP-FSAR-6D 6D-iii REV 22 8/09 LIST OF FIGURES 6D-1 Farley Unit 1 Strainer Layout

6D-2 Farley Unit 2 Strainer Layout 6D-3 Vertical Strainer Type

6D-4 Horizontal Strainer Type 6D-5 Postulated Break Locations 6D-6 Typical Arrangement of Containment Sump Suction Line FNP-FSAR-6D 6D-1 REV 22 8/09 APPENDIX 6D CONTAINMENT SUMP DESCRIPTION AND EMERGENCY CORE COOLING SYSTEM RECIRCULATION SUMP STRAINER DESIGN 6D.1 CONTAINMENT SUMP DESCRIPTION 6D.1.1 GENERAL PLANT SYSTEM DESCRIPTION Farley Nuclear Plant (FNP) Units 1 and 2 are Westinghouse three loop Pressurized Water Reactor (PWR) design. The residual heat removal system (RHRS) (low head safety injection), centrifugal charging system (CVCS) (high head safety injection), and containment spray system (CSS) pumps are started following a loss of coolant accident (LOCA). Initially, two RHR, two CVCS, and two CCS pumps take suction from the refueling water storage tank (RWST). When the RWST level reaches the low level setpoint, the RHR pumps are manually stopped and are realigned to take suction from the post-LOCA containment sump. Once the RHR switchover to recirculation is complete, the CVCS pumps take suction from the RHR pump discharge.

When the RWST level reaches low-low level, the CSS pumps are realigned to take suction from the containment sump. There are four independent suctions (two for RHR and two for CSS) located at el 105 ft-6 in. in the containment, the lowest floor elevation in the containment exclusive of the reactor cavity, and they are located outside the secondary shield wall.

The FNP nuclear steam supply system (NSSS) is a three-loop pressurized water reactor (PWR). The system consists of one reactor pressure vessel (RPV), three steam generators (SGs), three reactor coolant pumps (RCPs), one pressurizer (PZR), and the reactor coolant system (RCS) piping. The NSSS is located inside a bioshield and the reactor cavity. The area inside the bioshield is mostly open at the lowest levels, with the exception of the reactor cavity and surrounding walls in the center, and a concrete wall between the A and C loops. The concrete wall between loops A and C has a walkway against the reactor cavity wall that allows an opening between loops A and C. The outer bioshield walls extend from the containment base elevation of 105 ft-6 in. to el 129 ft-0 in. There are areas of the bioshield walls that are partially open; an inner wall extends from el 105 ft-6 in. to 116 ft-3 in., and an outer wall extends down from el. 129 ft-0 in. to el 115 ft-3 in. at some locations. Above el 129 ft-0 in. smaller "vaults" or "coffins" surround each loop and the associated steam generator and reactor coolant pump. These vaults further narrow around the steam generator at el 155 ft-0 in. and extend up to el 166 ft-6 in.. A separate vault for the pressurizer begins at el 129 ft-0 in. and extends up to el 181 ft-0 in.

The containment recirculation sump is a collecting reservoir designed to provide an adequate supply of water, with a minimum amount of particulate matter, to the CSS and the RHRS. The containment sump performance meets the NRC acceptance criteria contained in General Design Criteria 35, 36, and 37, and the NRC acceptance criteria listed below. A. The net positive suction head (NPSH) available to each safety system pump has been shown to provide adequate margin over the required NPSH at limiting runout conditions (see FSAR paragraph 6.3.2.14).

FNP-FSAR-6D 6D-2 REV 22 8/09 B. Housekeeping requirements specified in the quality assurance program and the Technical Requirements Manual. C. The ability to monitor and control RHRS status.

In each of the four pumps suction lines from the containment sump there are two motor-operated gate valves. There is no interdependency between systems or between the redundant portions of the same system. The motor-operated gate valves in the lines from the containment sump to the various pumps are normally closed and remain closed during the injection phase of emergency core cooling system (ECCS) operation. The protective screened structures in the containment sump will be completely submerged at the end of the injection phase and will remain submerged during the recirculation phase.

6D.1.2 GENERAL DESCRIPTION OF NEW ECCS STRAINERS INSTALLED FNP contracted with General Electric Company (GE) to provide sump strainers that meet the requirements of GL 2004-02. GE provided FNP with seven horizontal stacked disk strainers (see figure 6D-4) and one vertical stacked disk strainer (see figure 6D-3). The strainers were installed in both Unit 1 and Unit 2. Unit 1 only has the vertical stacked strainer installed on the B-train containment spray pump suction. The strainer plate nominal hole size is 3/32 in. The strainers for FNP Unit 1 and Unit 2 are located outside the biowall between the biowall and CTMT outside wall (see figures 6D-1 and 6D-2). This location protects the strainers from missile impacts.

6D.1.3 SIZE OF NEW ECCS STRAINERS INSTALLED For Unit 1 the passive strainer solution is shown on figure 6D-1. Each strainer assembly for both RHR strainers and CS Alpha strainer consists of two modular horizontal stacked disk strainer subunits connected to the post-LOCA pump suction through piping. The CS Bravo strainer assembly consists of three modular vertical stacked disk strainer subunits connected to a plenum that assists in directing flow to the post-LOCA pump suction inlet located within the plenum boundary. The RHR strainer assembly, either Alpha or Bravo, is composed of two strainer subunits per sump, each consisting of 22 stacked disks that are 40 in. X 40 in. and provide a total of approximately 878 ft² of perforated plate surface area. The CS Alpha strainer assembly consists of one strainer subunit with twenty two 40 in. X 40 in. stacked disks and the other with ten 40 in. X 40 in. stacked disks, providing a total of approximately 638 ft² of perforated plate surface area. The CS Bravo strainer assembly is composed of three strainer subunits, each with thirteen 30 in. X 30 in. vertical stacked disks, and provides a total of approximately 389 ft² of perforated plate surface area. For Unit 2 the passive strainer solution is shown on figure 6D-2. Each strainer assembly for RHR and CS consists of two modular horizontal stacked disk strainers connected to the sump through piping. The RHR strainer assemblies, both Alpha and Bravo, are composed of two strainers per sump, each consisting of 22 stacked disks that are 40 in. X 40 in. and provide a total of approximately 878 ft² of perforated plate surface area. The CS Alpha strainer assembly FNP-FSAR-6D 6D-3 REV 22 8/09 consists of one strainer with twenty two 40 in. X 40 in. stacked disks and the other with ten 40 in. X 40 in. stacked disks, providing a total of approximately 638 ft² of perforated plate surface area. The CS Bravo strainer assembly is composed of two strainers, one with ten 40 in. X 40 in. stacked disks and the other with twenty two 30 in. X 30 in. disks, and provides a total of approximately 433 ft² of perforated plate surface area.

6D.2 SUMMARY DESCRIPTION OF APPROACH USED TO SIZE SUMP STRAINERS SNC has performed analysis to determine the susceptibility of the ECCS and CSS recirculation functions for Farley Nuclear Plant to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform to the greatest extent practicable to the NEI 04-07 methodology as approved by the NRC safety evaluation report dated December 6, 2004. Following is a summary description of the analysis areas performed:

6D.2.1 CONTAINMENT WALKDOWN Walkdown of containment was performed by SNC personnel using the guidance of NEI 02-01.

The information obtained from the walkdown confirmed the insulation that was installed in containment matched the design documentation. Containment walkdowns confirmed the general housekeeping condition of containment was being maintained per plant procedures. 6D.2.2 PIPE BREAK CHARACTERIZATION Pipe break characterization was performed by Sargent and Lundy of Chicago. The piping runs considered for breaks are the RCS hot legs, the RCS cold legs, RCS interim legs, and all RCS attached energized piping. Breaks in these lines could decrease RCS inventory and result in the ECCS and/or CSS operating in recirculation mode, in which the system pumps would take suction from the containment sumps. Regulatory position 1.3.2.3 of Regulatory Guide 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 3, was used to select the spectrum of breaks for evaluation. A summary of the break locations is provided in figure 6D-5.

6D.2.3 DEBRIS GENERATION The debris generation analysis was performed by Sargent and Lundy of Chicago. The analysis determined the debris generated based on the NEI guidance and NRC SER of the NEI guidance. The analysis determined the ZOI for each type of material identified inside containment. See table 6D-1 for basis of ZOIs.

Insulation found inside containment that is adversely affected during a LOCA event, was determined to consist of a very small quantity of Tempmat fiber, Transco RMI, and Mirror insulations. Most of the insulation is Transco RMI and Mirror RMI. The amount of Tempmat fiber is very small. See table 6D-2 for summary of debris generated by each break.

FNP-FSAR-6D 6D-4 REV 22 8/09 The limiting break for coatings evaluated for a 4.0D ZOI is also on the intermediate leg of loop B, but at the RCP side of the pipe. Therefore, in order to conservatively maximize the debris available for transport, the maximum insulation debris location (break S2) is combined with the maximum coating debris location. See table 6D-3 for coating debris. Unqualified coatings are also identified in containment walkdown and plant condition reports.

6.D.2.4 LATENT DEBRIS ACCUMULATION WITHIN CONTAINMENT Programmatic controls are in place at FNP that give bases for the amounts of foreign material and latent debris inside containment remaining below the amounts assumed in the sump analysis. See table 6D-4 for latent and foreign material debris used in the analysis.

6D.2.5 DEBRIS TRANSPORT TO THE SUMP A debris transport analysis estimated the fraction of debris that is transported from debris sources (break locations) to the sump screen. The transport analysis is in accordance with the guidance of NEI 04-07 and the applicable NRC SER. The computational fluid dynamics (CFD) analysis was performed by RWDI Consulting Engineers and Scientists for Sargent and Lundy of Chicago. The CFD modeling techniques used are consistent with the SER, NEI Document number 04-07, and NUREG/CR-6773. CFD analyses of the post-LOCA recirculation flow patterns within the FNP containments were performed to quantify the flow velocities expected inside the secondary shield wall, through the secondary shield wall, outside the secondary shield wall, and near the CS and RHR sumps.

CFD analysis of the post-LOCA recirculation containment flows indicates velocities that will transport debris to the suction strainers. See table 6D-5 for a Summary of Debris Generated and Transported to Strainer Modules.

6D.2.6 HEAD LOSS AS A RESULT OF DEBRIS ACCUMULATION The engineered sump screens installed at FNP are designed to operate in such a way that the thin bed effect does not occur on the sump screen surface. This is due to the small amount of fiber present in the FNP containment. Parametric analyses were performed to estimate the surface area of the engineered screen that meets the FNP head loss criterion for the identified debris inventory. For the limiting break for screen head loss as selected in accordance with NEI 04-07, screens would be fully submerged at the minimum calculated sump levels. The RHR screen height is 44.75 in. above the floor. With leveling shims the height may be increased at points on the screens less than an inch. The minimum calculated water level is 54 in. above the floor elevation which is calculated to occur for the long term and not at the initiation of recirculation. This is largely due to gradual refilling of the area under the reactor vessel and due to conservatively postulated refilling of the SG tubes and the pressurizer. The tallest CS screen is 46.2 in. high; therefore, it may have slightly less submergence. Under this scenario the screens will be fully submerged by no less than 6 in.

FNP-FSAR-6D 6D-5 REV 22 8/09 A small break LOCA that results in minimum sump level would be one that occurs on top of the pressurizer. This level was not calculated as it is not a limiting break location that results in the highest screen head losses. The connections on the top of the pressurizer are 6 in. in diameter.

Therefore, a break in this location would produce very small amounts of debris. In addition, as compared to the limiting large break location, a small break would result in lower sump flowrates and, therefore, reduced sump debris transport. The resultant reduced RHR flowrates would result in a reduction in both debris bed head loss and a reduction in the NPSH required for the RHR pumps. An SBLOCA clearly does not present a significant challenge to the ECCS sump performance and is bounded by a LBLOCA. Since this is not a limiting break location the screen submergence was not calculated for this break.

As the screens are well covered for the limiting breaks the potential for air injection due to buoyant debris accumulation on top of the strainer is not considered to be plausible. For breaks that may result in some transient uncoverage, RHR flowrates would be reduced. CS screens would be fully covered as the RWST level is drawn down further before CS is placed on recirculation. A vortexing analysis was done for the Farley strainers assuming maximum RHR and CS flowrates. Vortexing was not indicated using the assumption that the strainer has the geometry of an open ended submerged pipe. This conservatively does not account for the complex stacked disc geometry of the strainer which would in effect act as vortex breakers.

6D.2.7 DEBRIS SOURCE TERM REDUCTION Foreign material (i.e., tags, labels, etc., not qualified for LOCA environmental conditions) may fail following a LOCA and, therefore, can be transported to the sump. Actions have been taken by SNC to ensure that the quantity of foreign material is minimized.

6D.2.8 SUMP STRUCTURAL ANALYSIS Structural analysis of the engineered passive screen has been completed. SNC has installed an engineered passive strainer on each RHR and CSS containment sump inlet pipe. The screens are located outside the secondary shield wall between the shield wall and the containment wall and, as such, are not exposed to jet impingement or postulated missiles generated from a LOCA event. The screens are of a robust design that support structural and hydraulic load created by the accumulation of debris during the post-LOCA environment. This robust design provides the strength of trash racks and is adequate to protect the screen during a LOCA event.

6D.2.9 UPSTREAM EFFECTS OF DEBRIS ACCUMULATION Evaluations of containment along with review of the CFD model indicate no significant areas will become blocked with debris and hold up water during the sump recirculation phase. As a precautionary measure, SNC modified the reactor cavity drain covers to further reduce the possibility of the drain becoming clogged and trapping a volume of water in the reactor cavity.

FNP-FSAR-6D 6D-6 REV 22 8/09 6D.2.10 DOWNSTREAM EFFECTS - COMPONENTS AND SYSTEMS The methodologies of NEI 04-07, as modified by the NRC safety evaluation dated December 6, 2004, and WCAP-16406-P, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191," were used to evaluate the downstream effects of debris that is passed by the sump strainer. The only components requiring modification were the safety injection throttle valves. A new flow reducing orifice was installed and the valves were replaced. This modification has been completed on Unit 1 and resulted in 9 of the 12 valves being locked open at a position which produced an internal valve clearance of 110 % of the containment sump screen hole size.

The other 3 valves opening are about 106 % of the hole size. An evaluation to address the acceptability of these valve positions was performed and they were found to be acceptable. The throttle valve replacement on Unit 2 has been delayed until fall 2008. An extension request was approved by the NRC in a letter dated August 29, 2007. This section will be updated later to reflect the installation on the Unit 2 valves.

6D.2.11 DOWNSTREAM EFFECTS - FUEL AND VESSEL The methodologies of WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Re-circulating Fluid," Revision 0, as modified by NRC staff comments, were used to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling. The evaluation of the impact of chemical deposition on the fuel was performed using the guidance of WCAP-16793-NP with bounding plant parameters. In its supplemental responses to GL 2004-02, submitted to the NRC on February 28, 2008, and April 29, 2008 (see references 13 and 14), SNC concluded there was reasonable assurance that long-term core cooling was demonstrated for Farley Units 1 and 2. 6D.2.12 CHEMICAL EFFECTS The new strainers installed at FNP have been sized to account for some increase in head loss across the strainer as a result of interaction of the sump water with the debris material as it approached the strainers during recirculation phase. The methodologies of the base model WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191," Revision 0 (reference 10), as modified by NRC safety evaluation report dated December 21, 2007 (reference 11), were used to evaluate the impact of chemical precipitants on the containment sump screens during post-accident recirculation and the resulting effect on available NPSH for the ECCS and CSS pumps. SNC supplemented the chemical effects results with plant-specific test data that demonstrated that the aluminum precipitants do not form until the containment sump temperature drops below 140 ºF (see reference 15). Calculations using the chemical effects testing results and other inputs demonstrated the available NPSH margin for the ECCS and CSS pumps was adequate for the conditions expected during post-accident recirculation. In addition, the results of chemical effects testing were used in the evaluation of downstream effects on fuel and the reactor vessel (refer to 6D.2.11). In its final supplemental response to GL 2004-02, submitted to the NRC on April 29, 2008 (reference 14), SNC concluded that there was reasonable assurance that long-term core cooling was demonstrated for Farley Units 1 and FNP-FSAR-6D 6D-7 REV 22 8/09 2. The details of the chemical effects testing results are documented in GE Report 0000-0056-2976, Containment Sump Passive RHR & CS Strainer System S0100 Hydraulic Sizing Report, Revision 3 (reference 15).

FNP-FSAR-6D 6D-8 REV 22 8/09 REFERENCES 1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," September 13, 2004.

2. Nuclear Energy Institute (NEI) document NEI 04-07 Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology," December 2004.

3. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), "Pressurized Water Reactor Sump Performance Evaluation Methodology," December 6, 2004.

4. Regulatory Guide 1.82, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," Revision 3, November 2003. 5. WCAP-16568-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for DBA-Qualified / Acceptable Coatings," Revision 0. 6. Deleted. 7. Deleted. 8. WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191," Revision 1. 9. NRC SER dated December 20, 2007, Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report (TR) WCAP-16406-P, Revision 1, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191," Pressurized Water Reactor Owners Group. 10. WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191." 11. NRC SER dated December 21, 2007, Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191." 12. WCAP 16793-NP, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Re-circulating Fluid, Revision 0. 13. SNC Letter NL-08-2173 dated February 28, 2008, "Joseph M. Farley Nuclear Plant Supplemental Response to NRC Generic Letter 2004-02." 14. SNC Letter NL-08-0551 dated April 29, 2008, "Joseph M. Farley Nuclear Plant Final Supplemental Response to NRC Generic Letter 2004-02." 15. GE Report 0000-0056-2976 [U-732504], "Containment Sump Passive RHR & CS Strainer System S0100 Hydraulic Sizing Report," Revision 3.

FNP-FSAR-6D REV 21 5/08 TABLE 6D-1 CONTAINMENT SUMP DEBRIS GENERATION ZONE OF INFLUENCE (ZOI) Debris Constituent ZOI (Pipe Diameter) Basis Transco RMI 2.0D NRC SER Mirror RMI 28.6D NRC SER Temp-Mat Fiber NA All assumed as debris in analysis Qualified Coatings 4.0D WCAP-16568-P Unqualified Coatings NA NRC SER - All assumed as debris in analysis Latent Debris NA NRC SER - Conservative value based on plant walkdown Foreign Materials (Labels, etc.) NA NRC SER - Conservative value based on plant walkdown FNP-FSAR-6D REV 21 5/08 TABLE 6D-2 SUMMARY OF LOCA GENERATED INSULATION DEBRIS INSIDE ZOI Break ID Location Transco RMI Foils (ft2) Mirror RMI Foils (ft2) RMI Jacketing (ft2) Temp-Mat (ft3) S1 Loop C Interim Leg 2054 25527 5795 1 S2* Loop B Interim Leg 2383 35714 8022 1 S3 Loop A Cold Leg 0 34368 7522 1 S4 (alternate) Loop B Interim Leg 1226 23258 5223 0

___________________

  • S2 is the limiting location.

FNP-FSAR-6D REV 21 5/08 TABLE 6D-3 DEBRIS GENERATED FROM COATING BASED ON ZOI = 4D Break Coating Areas (ft2) Coating Volumes (ft3) Concrete Steel Concrete Steel Interim Leg at SG 200 1332 0.31 1.66 Interim Leg at Mid-span 218 1320 0.34 1.65 *Interim Leg at RCP 523 1091 0.81 1.36 Hot Leg at Primary Wall 294 758 0.46 0.95 Hot Leg at SG 0 1196 0 1.49 Unqualified Coatings NA 1,070 NA 0.535

________________

  • Limiting location for coatings FNP-FSAR-6D REV 21 5/08 TABLE 6D-4 LATENT AND FOREIGN MATERIAL DEBRIS USED IN ANALYSIS Latent Debris Total (lbm) 200 Fiber (lbm) 30 Particulate (lbm) 170 Foreign Material Debris (ft2) 36.4 FNP-FSAR-6D REV 21 5/08 TABLE 6D-5 SUMMARY OF DEBRIS GENERATED AND TRANSPORTED TO STRAINER MODULES Debris Type Units Quantity Generated Transport Fraction Quantity at Strainer Modules Fibrous Insulation Debris Temp-Mat [ft3] 1 1.0 1 Coating Debris in 4D ZOI Modeled as Chips Concrete Coatings [ft2; ft3] 523 ; 0.81 0.871 456 ; 0.71 Steel Coatings [ft2; ft3] 1091 ; 1.36 0.704 768 ; 0.96 Sum [ft2; ft3] 1614 ; 2.18 --- 1224 ; 1.67 Unqualified Coating Debris Modeled as Fines / Chips
  • Unqualified Coatings (Actual) [ft2; ft3] 1070 ; 0.535 1.0 1070 ; 0.535 Latent Debris Latent Fiber (Walkdown) [ft3] 7.8 1.0 7.8 Latent Fiber (30 lbm) [ft3] 12.5 1.0 12.5 Latent Particulate (Walkdown) [ft3] 0.63 1.0 0.63 Latent Particulate (170 lbm) [ft3] 1.01 1.0 1.01 Reflective Metal Insulation Debris Transco Foil [ft2] 2383 0.799 1904 Mirror Foil [ft2] 35714 0.769 27464 Foil Sum [ft2] 38097 --- 29368 RMI Jacketing [ft2] 8022 0.338 2711 Foreign Material Foreign Material1 (labels, stickers, etc.) [ft2] 36.4 1.0 36.4

________________________

  • Unqualified Coatings were modeled as a mixture of chips and fines.

Break Name Break ID Elevation Piping S1 31-inch 118'-0" Interim Leg - Loop C S2 31-inch 118'-0" Interim Leg - Loop B S3 27.5-inch 122-9" Cold Leg - Loop A S4 11.19-inch 118'-0" Alternate Break (Interim Leg -Loop B) REV 21 5/08 POSTULATED BREAK LOCATIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6D-5 REFER TO FIGURE 6D-4 FOR CONTINUATION OF SUCTION PIPE CONNECTION. REV 22 8/09 TYPICAL ARRANGEMENT OF CONTAINMENT SUMP SUCTION LINE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 6D-6 FNP-FSAR-7 7-i REV 22 8/09 7.0 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS

7.1 INTRODUCTION

.........................................................................................................7.1-1 7.1.1 Identification of Safety-Related Systems.........................................................7.1-3 7.1.2 Identification of Safety Criteria........................................................................7.1-4

7.1.2.1 Design Bases..................................................................................7.1-5 7.1.2.2 Independence of Redundant Safety-Related Systems.................7.1-10 7.1.2.3 Physical Identification of Safety-Related Equipment....................7.1-12 7.1.2.4 Conformance to IEEE 317-1971...................................................7.1-12 7.1.2.5 Conformance to IEEE 323-1971...................................................7.1-13 7.1.2.6 Conformance to IEEE 336-1971...................................................7.1-13 7.1.2.7 Conformance to IEEE 338-1971...................................................7.1-13 7.1.2.8 Conformance to Regulatory Guide 1.22.......................................7.1-14 7.1.2.9 Conformance to IEEE 334-1971...................................................7.1-15 7.1.2.10 Conformance to 10 CFR 50.62.....................................................7.1-15 7.1.2.11 Conformance to NUREG-0737.....................................................7.1-15

7.1.3 Detailed Electrical Instrumentation and Control Drawings............................7.1-16

7.1.3.1 Identification and Purpose............................................................7.1-16

7.2 REACTOR TRIP SYSTEM..........................................................................................7.2-1

7.2.1 Description......................................................................................................7.2-1

7.2.1.1 System Description.........................................................................7.2-2 7.2.1.2 Design Bases: IEEE 279-1971....................................................7.2-14 7.2.1.3 Final System Drawings.................................................................7.2-17

7.2.2 Analysis.........................................................................................................7.2-18

7.2.2.1 Failure Mode and Effects Analysis................................................7.2-18 7.2.2.2 Evaluation of Compliance to Applicable Codes and Standards....7.2-20 7.2.2.3 Specific Control and Protection Interactions.................................7.2-30

7.2.3 Tests and Inspections...................................................................................7.2-34

FNP-FSAR-7 7-ii REV 22 8/09 TABLE OF CONTENTS 7.2.3.1 Inservice Tests and Inspections....................................................7.2-34 7.2.3.2 Periodic Testing of the Nuclear Instrumentation System..............7.2-36 7.2.3.3 Periodic Testing of the Process Analog Channels of the Protection Circuits...............................................................7.2-36 7.2.3.4 Regulatory Guide 1.22..................................................................7.2-37

7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM....................................7.3-1

7.3.1 Description......................................................................................................7.3-1

7.3.1.1 System Description.........................................................................7.3-1 7.3.1.2 Design Bases..................................................................................7.3-6 7.3.1.3 Final System Drawings...................................................................7.3-9

7.3.2 Analysis...........................................................................................................7.3-9 7.3.2.1 Evaluation of Compliance with IEEE 279-1971............................7.3-10 7.3.2.2 Evaluation of Compliance with IEEE 308-1971............................7.3-16 7.3.2.3 Evaluation of Compliance with IEEE 323-1971............................7.3-16 7.3.2.4 Evaluation of Compliance with IEEE 334-1971............................7.3-16 7.3.2.5 Evaluation of Compliance with IEEE 338-1971............................7.3-17 7.3.2.6 Evaluation of Compliance with IEEE 344-1971............................7.3-17 7.3.2.7 Response Time Testing................................................................7.3-17 7.3.2.8 Further Considerations.................................................................7.3-18 7.3.2.9 Summary.......................................................................................7.3-18

7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN......................................................7.4-1 7.4.1 Description......................................................................................................7.4-1

7.4.1.1 Monitoring Indicators.......................................................................7.4-2 7.4.1.2 Controls...........................................................................................7.4-2 7.4.1.3 Essential Services after Incident That Requires Hot Shutdown......7.4-5 7.4.1.4 Equipment and Systems Available for Cold Shutdown...................7.4-6

7.4.2 Analysis...........................................................................................................7.4-7

FNP-FSAR-7 7-iii REV 22 8/09 TABLE OF CONTENTS 7.5 POSTACCIDENT MONITORING DISPLAY INSTRUMENTATION............................7.5-1 7.5.1 Description......................................................................................................7.5-1 7.5.2 Analysis...........................................................................................................7.5-2 7.5.3 Deleted............................................................................................................7.5-3 7.5.4 Inadequate Core Cooling Monitoring System..................................................7.5-3

7.5.4.1 Reactor Vessel Level......................................................................7.5-3 7.5.4.2 Subcooling Margin Monitor.............................................................7.5-4 7.5.4.3 Core Exit Temperature....................................................................7.5-4

7.5.5 Nuclear Instrumentation..................................................................................7.5-4

7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY...................................................7.6-1

7.6.1 Instrumentation and Control Power Supply System........................................7.6-1

7.6.1.1 Description......................................................................................7.6-1 7.6.1.2 Analysis...........................................................................................7.6-1

7.6.2 Residual Heat Removal Isolation Valves.........................................................7.6-3

7.6.2.1 Description......................................................................................7.6-3 7.6.2.2 Analysis...........................................................................................7.6-4

7.6.3 Refueling Interlocks.........................................................................................7.6-4 7.6.4 Monitoring Combustible Gas in Containment..................................................7.6-6

7.6.4.1 Description......................................................................................7.6-6 7.6.4.2 Analysis...........................................................................................7.6-7

7.6.5 Semiautomatic Backup to Switchover from Injection to Recirculation.............7.6-7 7.6.6 Accumulator Motor-Operated Isolation Valves................................................7.6-7

7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY.............................................7.7-1

7.7.1 Description......................................................................................................7.7-1

7.7.1.1 Reactor Control System..................................................................7.7-3 7.7.1.2 Rod Control System........................................................................7.7-4

FNP-FSAR-7 7-iv REV 22 8/09 TABLE OF CONTENTS 7.7.1.3 Plant Control Signals for Monitoring and Indicating........................7.7-5 7.7.1.4 Plant Control System Interlocks......................................................7.7-9 7.7.1.5 Pressurizer Pressure Control..........................................................7.7-9 7.7.1.6 Pressurizer Water Level Control...................................................7.7-10 7.7.1.7 Steam Generator Water Level Control..........................................7.7-10 7.7.1.8 Steam Dump Control....................................................................7.7-11 7.7.1.9 Incore Instrumentation..................................................................7.7-12 7.7.1.10 Control Board................................................................................7.7-14 7.7.1.11 Boron Concentration Measurement System.................................7.7-15

7.7.2 Analysis.........................................................................................................7.7-17

7.7.2.1 Separation of Protection and Control Systems.............................7.7-18 7.7.2.2 Response Considerations of Reactivity........................................7.7-19 7.7.2.3 Step Load Changes Without Steam Dump...................................7.7-21 7.7.2.4 Loading and Unloading.................................................................7.7-21 7.7.2.5 Load Rejection Furnished by Steam Dump System.....................7.7-21 7.7.2.6 Turbine-Generator Trip with Reactor Trip.....................................7.7-22

7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC)..........................7.8-1

7.8.1 Description......................................................................................................7.8-1

7.8.1.1 System Description.........................................................................7.8-1 7.8.1.2 Equipment Description....................................................................7.8-1 7.8.1.3 Functional Performance Requirements..........................................7.8-3 7.8.1.4 AMSAC Interlocks...........................................................................7.8-3 7.8.1.5 Trip System.....................................................................................7.8-3 7.8.1.6 Isolation Devices.............................................................................7.8-4 7.8.1.7 AMSAC Diversity From the Reactor Protection Systems................7.8-4 7.8.1.8 Power Supply..................................................................................7.8-4 7.8.1.9 Environmental Variations................................................................7.8-4 7.8.1.10 Setpoints.........................................................................................7.8-4

7.8.2 Analysis...........................................................................................................7.8-5 7.8.2.1 Safety Classification/Safety-Related Interface................................7.8-5 7.8.2.2 Redundancy....................................................................................7.8-5 7.8.2.3 Diversity From Existing Trip System...............................................7.8-5

FNP-FSAR-7 7-v REV 22 8/09 TABLE OF CONTENTS 7.8.2.4 Electrical Independence..................................................................7.8-5 7.8.2.5 Physical Separation From the RTS and ESFAS.............................7.8-6 7.8.2.6 Environmental Qualification............................................................7.8-6 7.8.2.7 Seismic Qualification.......................................................................7.8-6 7.8.2.8 Test, Maintenance, and Surveillance Quality Assurance................7.8-6 7.8.2.9 Power Supply..................................................................................7.8-7 7.8.2.10 Testability at Power.........................................................................7.8-7 7.8.2.11 Inadvertent Actuation......................................................................7.8-7 7.8.2.12 Bypass............................................................................................7.8-7 7.8.2.13 Completion of Mitigative Actions Once Initiated..............................7.8-8 7.8.2.14 Manual Initiation..............................................................................7.8-8 7.8.2.15 Information Readout.......................................................................7.8-8 7.8.2.16 Compliance With Standards and Design Criteria............................7.8-9

FNP-FSAR-7 7-vi REV 22 8/09 LIST OF TABLES 7.1-1 List of Schematic Diagrams and Location Drawings for Safety-Related Equipment 7.2-1 List of Reactor Trips

7.2-2 Protection System Interlocks

7.2-3 Reactor Trip System Instrument Accuracies

7.2-4 Trip Correlation

7.2-5 Reactor Trip System Instrumentation Response Times

7.3-1 Functions Initiated by Engineered Safety Features Actuation System

7.3-2 Instrumentation Operating Conditions for Engineered Safety Features

7.3-3 Instrumentation Operating Conditions for Isolation Functions

7.3-4 Interlocks for Engineered Safety Features Actuation System

7.3-5 (Deleted)

7.3-6 Failure Mode and Effects Analysis, Service Water System

7.3-7 Failure Mode and Effects Analysis, Component Cooling Water System

7.3-8 Failure Mode and Effects Analysis, Control Room and Air Conditioning and Filtration System 7.3-9 Failure Mode and Effects Analysis, Penetration Room Filtration System

7.3-10 Failure Mode and Effects Analysis, Auxiliary Feedwater System 7.3-11 Failure Mode and Effects Analysis, Emergency Safeguards Pump Room Cooling System 7.3-12 Failure Mode and Effects Analysis, Battery Room Ventilation System

7.3-13 Failure Mode and Effects Analysis, Battery Room Air Conditioning System

FNP-FSAR-7 7-vii REV 22 8/09 LIST OF TABLES 7.3-14 Failure Mode and Effects Analysis, Emergency Diesel Generator 7.3-15 Failure Mode and Effects Analysis, Engineered Safety Features Actuation System

7.3-16 Engineered Safety Features Response Times

7.5-1 Post Accident Instrumentation

7.5-2 (Deleted)

7.5-3 Control Room Indicators and/or Recorders Available to the Operator to Monitor Significant Plant Parameters During Normal Operation 7.7-1 Plant Control System Interlocks

7.7-2 Boron Concentration Measurement System Specifications

FNP-FSAR-7 7-viii REV 22 8/09 LIST OF FIGURES 7.2-1 Setpoint Reduction Function for Overpower and Overtemperature T Trips 7.2-2 Pressurizer Sealed Reference Leg Level System

7.2-3 Design to Achieve Isolation Between Channels

7.3-1 Component Identification ESFAS

7.6-1 Logic Diagram for Residual Heat Removal System Isolation Valves

7.6-2 Logic Diagram for Residual Heat Removal System Isolation Valves

7.6-3 Logic Diagram for Backup to Semiautomatic Switchover Logic from Injection to Recirculation 7.6-4 Functional Block Diagram of Accumulator Isolation Valve

7.7-1 Simplified Block Diagram of Reactor Control System

7.7-2 Control Bank Rod Insertion Monitor

7.7-3 Rod Deviation Comparator

7.7-4 Block Diagram of Pressurizer Pressure Control System

7.7-5 Block Diagram of Pressurizer Level Control System

7.7-6 Block Diagram of Main Feedwater Pump Speed Control System

7.7-7 Block Diagram of Steam Generator Water Level Control System 7.7-8 Block Diagram of Steam Dump Control System

7.7-9 Basic Flux Mapping System

7.7-10 Source-Detector Assembly

7.7-11 Measurement Unit

7.7-12 Process Schematic for the Boron Concentration Measurement System 7.7-13 Boron Concentration Measurement System vs Normal Plant Operating Range of Boron Concentrations 7.8-1 Actuation Logic System Architecture

[HISTORICAL][7.1.3 DETAILED ELECTRICAL INSTRUMENTATION AND CONTROL DRAWINGS 7.1.3.1 Identification and Purpose A set of volumes containing nonproprietary detailed EI&C drawings has been prepared in accordance with the NRC interim guidelines, pending revisions of the Standard Format. It is entitled, "Joseph M.

Farley Nuclear Plant, Safety-Related Schematic Diagrams and Location Drawings, November 1973," and is in four volumes, FNP-1001, FNP-1002, FNP-1003, and FNP-1004. The supplement furnished detailed information in response to paragraphs 7.2.1.3 and 7.3.1.3 and subsections 7.4.1 and 7.6.1 of the Standard Format. The purpose of the supplement was to facilitate tracing the safety-related signals from sensors to actuating devices. It was submitted with Amendment 27. A list of the submitted EI&C drawings and related FSAR figures is maintained in table 7.1-1 for historical purposes.

FNP-FSAR-7 REV 21 5/08 [HISTORICAL][TABLE 7.1-1 (SHEET 1 OF 33) LIST OF SCHEMATIC DIAGRAMS AND LOCATION DRAWINGS FOR SAFETY-RELATED EQUIPMENT

This table lists drawings which are presented in the FSAR by reference to project drawing numbers or were provided to the NRC in the supplement.

Submittal to NRC Drawing (formerly Number AEC) Title Equipment Location Number Index 175070 11/01/73 Equipment numbers tabulation Nuclear Instrumentation System (NIS) Block Diagrams and Safeguards Test Cabinet 108D501 11/01/73 Process control block diagram 5655D37 11/01/73 Functional diagrams 5655D49 11/01/73 NIS source range functional block diagram 5655D50 11/01/73 NIS intermediate range functional block diagram 5655D51 11/01/73 NIS power range functional block diagram 5655D52 11/01/73 NIS auxiliary channels functional block diagram 724D17 11/01/73 Safeguards test cabinet (10 sheets)

Elementary Diagrams, 177000 Series (Includes Related 207000 Series) 177000 11/01/73 Single line electrical auxiliary system (normal) 177001 11/01/73 Single line electrical auxiliary system (emergency) 177005 11/01/73 Single line protection and metering, 4160-V bus 1F 177006 11/01/73 Single line protection and metering, 4160-V bus 1G 177018 11/01/73 Single line protection and metering, 4160-V bus 1H 177027 11/01/73 Single line protection and metering, 4160-V bus 1J 177043 11/01/73 Single line protection and metering, 4160-V bus 1K 177044 11/01/73 Single line protection and metering, 4160-V bus 1L 177007 11/01/73 Single line protection and metering, 600-V load center 1A 177009 11/01/73 Single line protection and metering, 600-V load center 1C

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 2 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177010 11/01/73 Single line protection and metering, 600-V load center 1D 177011 11/01/73 Single line protection and metering, 600-V load center 1E 177012 11/01/73 Single line protection and metering, 600-V load center 1F 177014 11/01/73 Single line protection and metering, 600-V load center 1H 177015 11/01/73 Single line protection and metering, 600-V load center 1J 177045 11/01/73 Single line protection and metering, 600-V load center 1K 177046 11/01/73 Single line protection and metering, 600-V load center 1L 177677 11/01/73 Single line protection and metering, 600-V load center 1R 177678 11/01/73 Single line protection and metering, 600-V load center 1S 177118 11/01/73 Interlock schematic station service transformer 1F 177122 11/01/73 Interlock schematic 600-V bus 1A 177024 11/01/73 Single line 120 V-ac vital and regulated system A 177025 11/01/73 Single line 120 V-ac vital and regulated system B 177754 11/01/73 Tray and conduit layout, cable spreading room 177033 11/01/73 Logic diagram diesel 1A auto start and loading 177032 11/01/73 Logic diagram diesel 1B auto start and loading 177036 11/01/73 Logic diagram diesel 1C auto start and loading

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 3 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177037 11/01/73 Logic diagram diesel 2C auto start and loading - 11/01/73 One-line diagram dc supply for diesel generators 207032 11/01/73 Logic diagram diesel 2B auto start and loading 177119 11/01/73 Interlock schematic component cooling water pump 2B 177120 11/01/73 Interlock schematic HHSI pump 2B 177121 11/01/73 Interlock schematic service water pump 1C 177082 11/01/73 Single line dc distribution system 1A 177083 11/01/73 Single line dc distribution system 1B 207000 11/01/73 Single line electrical auxiliary system (normal 4160 V and 600 V) Unit 2 207001 11/01/73 Single line electrical auxiliary system (emergency 4160 V and 600 V) Unit 2 207033 11/01/73 Logic diagram diesel 1A auto start and loading 207036 11/01/73 Logic diagram diesel 1C auto start and loading 207037 11/01/73 Logic diagram diesel 2C auto start and loading 177133 11/01/73 Interlock schematic battery charger 1C 177050 11/01/73 Elementary diagram 600-V LC bus 1A tie breaker from 600-V LC bus 1D 177051 01/10/75 Elementary diagram 575-V motor- operated valve 177052 11/15/74 Elementary diagram 575-V motor- operated valve 177053 11/15/74 Elementary diagram 575-V motor- operated valve 177058 11/01/73 Elementary diagram 600-V LC bus 1C tie breaker 177059 11/01/73 Elementary diagram 600-V LC bus 1C tie breaker from 600-V LC bus 1E

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 4 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177064 11/01/73 Elementary diagram 600-V LC bus 1D tie breaker from 600-V LC bus 1A 177070 11/01/73 Elementary diagram 600-V LC bus 1E tie breaker from 600-V LC bus 1C 177072 11/01/73 Elementary diagram 600-V LC buses 1D and 1E, including breaker from bus 1F 177077 11/01/73 Elementary diagram 600-V LC breakers to battery chargers 1A and 1B 177078 11/01/78 Elementary diagram 600-V LC breakers to battery charger 1C 177080 11/01/73 Synchronizing diagram 4160-V emergency buses train A Units 1 and 2 177081 11/01/73 Synchronizing diagram 4160-V emergency buses train B Units 1 and 2 177087 11/01/73 Elementary diagram 600-V LC buses 1A, 1B, 1C, 1D, and 1E potential transformer 177089 11/01/73 Elementary diagram 600-V LC breakers to motor control centers 1A, 1B, 1F, 1G, 1S, 1U, and 1V 177091 11/01/73 Elementary diagram miscellaneous relay 177142 11/01/73 Elementary diagram 4160-V bus 1G incoming breaker from diesel generator 1B 177143 11/01/73 Elementary diagram 4160-V bus 1F incoming breaker from diesel generator 1A 177144 11/01/73 Elementary diagram 4160-V bus tie from 4160-V bus 1F to 1KC 1G101L 177145 11/01/73 Elementary diagram 4160-V bus tie breaker from 4160-V bus 1F to 1MC 1G101J 177155 11/01/73 Elementary diagram 4160-V bus 1F incoming startup transformer 1A

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 5 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177157 11/01/73 Elementary diagram 4160-V bus 1F potential transformers 177159 11/01/73 Elementary diagram 4160-V bus 1F outgoing station service transformers 1D and 1G 177160 11/01/73 Elementary diagram 4160-V bus 1F outgoing station service transformer 1F 177161 11/01/73 Elementary diagram 4160-V bus 1F incoming startup transformer 1B 177163 11/01/73 Elementary diagram 4160-V bus 1G potential transformers 177166 11/01/73 Elementary diagram 4160-V bus 1G outgoing station service transformer 1F 177167 11/01/73 Elementary diagram 4160-V bus tie breaker 1G to 1J 177168 11/01/73 Elementary diagram 4160-V bus 1G incoming startup transformer 1A 177169 11/01/73 Elementary diagram 4160-V bus 1G incoming startup transformer 1B 177170 11/01/73 Elementary diagram 4160-V buses 1F and 1G diff. prot. 177173 11/01/73 Elementary diagram 4160-V bus 1G diff. prot. 177183 11/01/73 Elementary diagram component cooling water pump 4160-V bus 1C 177184 11/01/73 Elementary diagram component cooling water pump 4160-V bus 1A 177185 11/01/73 Elementary diagram component cooling water pump 4160-V bus 1B train A 177186 11/01/73 Elementary diagram auxiliary feedwater pump 4160-V buses 1A and 1 177187 11/01/73 Elementary diagram component cooling water pump 4160-V bus 1B train B FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 6 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177188 11/01/73 Elementary diagram turbine- driven auxiliary feedwater pump-starter train A 177189 11/01/73 Elementary diagram turbine- driven auxiliary feedwater pump-starter train B 177199 11/01/73 Elementary diagram containment purge exhaust damper 177204 11/01/73 Elementary diagram containment purge system isolation dampers 177206 11/01/73 Elementary diagram containment post-LOCA air mixing fans 177221 11/01/73 Elementary diagram containment cooling high speed 177222 11/01/73 Elementary diagram containment cooling low speed 177224 11/01/73 Elementary diagram boric acid transfer pumps 1 and 2 177226 11/01/73 Elementary diagram charging/ HHST pump 1B room cooler fan motor train A 177227 11/01/73 Elementary diagram RHR pump and containment spray pump from cooler fan motors 177229 11/01/73 Elementary diagram HHST and auxiliary feedwater pump room and common heat exchange cooler fan motor 177232 11/01/73 Elementary diagram containment cooler damper motor 177236 11/01/73 Elementary diagram containment purge supply fan high speed 177237 11/01/73 Elementary diagram containment purge exhaust fan low speed 177238 11/01/73 Elementary diagram penetration from exhaust fans 1 and 2 177239 11/01/73 Elementary diagram penetration room recirculation fans 1 and 2 177240 11/01/73 Elementary diagram boron injection tank recirculation pumps 1 and 2

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 7 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177243 11/01/73 Elementary diagram component cooling water and pump room cooler fans 177246 11/01/73 Elementary diagram spent fuel pool air exhaust fans 1 and 2 177253 08/15/74 Elementary diagram phosphate injection pumps 177259 11/01/73 Elementary diagram radwaste air exhaust fan 1A motor 177262 11/01/73 Elementary diagram control rod drive mechanism cooler 1 177263 11/01/73 Elementary diagram control rod drive mechanism cooling fan dampers 177270 11/01/73 Elementary diagram control room filter fan motors 177275 11/01/73 Elementary diagram control room filter intake dampers 177277 11/01/73 Elementary diagram reactor cavity H2 dilution A/P compressors 1A and 1B 177278 11/01/73 Elementary diagram containment preaccess fan motors 177279 11/01/73 Elementary diagram control room filter exhaust dampers 177280 11/01/73 Elementary diagram control room outside air intake dampers 177281 11/01/73 Elementary diagram penetration room filter prefilter damper 177282 11/01/73 Elementary diagram refueling water surface supply and exhaust fan motors 177283 11/01/73 Elementary diagram penetration room filter recirculation damper 177284 11/01/73 Elementary diagram charging/ HHSI pump 1B room cooler fan train B 177291 11/01/73 Elementary diagram radwaste air exhaust fan 1B motor 177294 11/01/73 Elementary diagram miscellaneous auxiliary building sump pump motors

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 8 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177371 11/01/73 Elementary diagram solenoid valves, sheet 75 Pressurizer liquid sample train A Pressurizer steam sample train A Reactor hot leg sample train A Accumulator sample train A 177372 11/01/73 Elementary diagram solenoid valves, sheet 76 Pressurizer liquid sample train B Pressurizer steam sample train B Reactor hot leg sample train B Accumulator sample train B 177399 11/01/73 Elementary diagram accumulator discharge valve closed alarm 177400 11/01/73 Elementary diagram accumulator discharge valve closed alarm 177569 11/01/73 Elementary diagram 575-V motor- operated valve 177570 11/01/73 Elementary diagram 575-V motor- operated valve 177572 11/01/73 Elementary diagram 575-V motor- operated valve 177583 11/01/73 Elementary diagram solenoid valve, sheet 32 Motor-driven auxiliary feedwater pump Auxiliary feedwater bypass 177584 11/01/73 Elementary diagram solenoid valves, sheet 31 Surge tank discharge to auxiliary building 177588 11/01/73 Elementary diagram solenoid valves, sheet 27 Spent fuel exhaust intake 177589 11/01/73 Elementary diagram solenoid valves, sheet 26 Fuel handling area vent system Penetration room dampers

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 9 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177590 11/01/73 Elementary diagram solenoid valves, sheet 22 Turbine-driven auxiliary feedwater pump discharge 177591 11/01/73 Elementary diagram solenoid valves, sheet 23 Motor-driven auxiliary feedwater pump discharge 177592 11/01/73 Elementary diagram solenoid valves, sheet 24 Auxiliary steam condensate tank 177610 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 7 Reactor coolant pump component cooling water return from thermal barrier 177612 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 9 177613 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 10 Containment cooler service water return Containment cooler service water supply 177617 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 14 Service water to blowdown heat exchange Blowdown heat exchange, letdown chiller discharge 177618 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 15 Reactor coolant pump component cooling water return from oil coolers 177620 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 17 Auxiliary feedwater pump service water supply containment leak rate test 177622 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 19 Steam generator feedwater intake

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 10 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177623 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 20 Service water from storage tank train A 177624 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 21 Service water from storage tank train B 177625 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 22 Component cooling water to reactor coolant pump 177628 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 25 Component cooling makeup water Component cooling water to spent fuel pool heat exchange Service water to component cooling water heat exchange Component cooling water to residual heat exchange 177627 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 24 Auxiliary feedwater to steam generators 1A, 1B, and 1C 177629 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 26 Containment cooler service water bypass Containment cooler service water discharge 177630 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 27 Component cooling water heat exchange 177632 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 29 RHR pumps 1 and 1B miniflow 177633 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 30 Containment cooler discharge

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 11 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177635 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 32 Service water to component cooling water heat exchange 177636 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 33 Reactor coolant pump motor cooler service water discharge 177644 11/01/73 Elementary diagram 575-V motor- operated valve 177645 11/01/73 Elementary diagram loading sequencer B1F essential sequencer 177646 11/01/73 Elementary diagram loading sequencer B1G essential sequencer 177647 11/01/73 Elementary diagram essential loading sequencer B1G breaker close failure indication 177648 11/01/73 Elementary diagram essential loading sequencer B1G breaker close failure indication 177649 11/01/73 Elementary diagram loading sequencer B1F LOSP sequencer 177650 11/01/73 Elementary diagram loading sequencer B1G LOSP sequencer 177653 11/01/73 Elementary diagram loading sequencer B1F load shedding scheme 177654 11/01/73 Elementary diagram loading sequencer B1G load shedding scheme 177659 11/01/73 Elementary diagram loading sequencer B2H load shedding scheme

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 12 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177660 11/01/73 Elementary diagram loading sequencer B2J load shedding scheme 177688 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 47 177689 11/01/73 Elementary diagram 575-V motor- operated valves, sheet 48 177838 11/01/73 Elementary diagram 575-V motor- operated valve 177839 11/01/73 Elementary diagram 575-V motor- operated valve 177840 11/01/73 Elementary diagram 575-V motor- operated valve 177851 11/01/73 Elementary diagram solenoid valves, sheet 2 Excess letdown heat exchange inlet Excess letdown heat exchange discharge 177852 11/01/73 Elementary diagram solenoid valves, sheet 3 Surge tank discharge to auxiliary building 177853 11/01/73 Elementary diagram solenoid valves, sheet 4 Waste recycle evaporation discharge and inlet valves 177854 11/01/73 Elementary diagram solenoid valves, sheet 5 Reactor coolant pump component cooling 177855 11/01/73 Elementary diagram solenoid valves, sheet 6 Reactor coolant pump component cooling water thermal barrier return 177856 11/01/73 Elementary diagram solenoid valves, sheet 7 Component cooling heat exchange service water discharge 177857 11/01/73 Elementary diagram solenoid valves, sheet 8

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 13 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title Steam to turbine-driven auxiliary feedwater pump 177863 11/01/73 Elementary diagram solenoid valves train A, sheet 14 Main steam isolation valves 177864 11/01/73 Elementary diagram solenoid valves, sheet 15 Main steam isolation bypass valve train A 177865 11/01/73 Elementary diagram solenoid valves, sheet 16 Main steam isolation valve operator test 177866 11/01/73 Elementary diagram solenoid valves, sheet 17 Main steam isolation bypass valve train B 177867 11/01/73 Elementary diagram solenoid valves, sheet 18 Main steam isolation valves train B 177205 11/01/73 Elementary diagram spent fuel pool pumps 1A and 1B 177224 11/01/73 Elementary diagram boric acid transfer pumps 1A and 1B 177240 11/01/73 Elementary diagram boron injection tank recirculation pumps 1A and 1B 177174 11/01/73 Elementary diagram reactor coolant pumps 1, 2, and 3 177180 11/01/73 Elementary diagram charging/ HHSI pumps 1A and 1C 177181 11/01/73 Elementary diagram charging/ HHSI pump 1B train A 177182 11/01/73 Elementary diagram charging/ HHSI pump 1B train B 177193 11/01/73 Elementary diagram RHR/LHSI pumps 1A and 1B 177195 11/01/73 Elementary diagram containment spray pumps 1A and 1B 177107 11/01/73 Elementary diagram pressurizer heater backup group 1A (600-V LC emergency bus 1A)

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 14 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177111 11/01/73 Elementary diagram pressurizer heater backup group 1B (600-V LC emergency bus 1C) 177364 11/01/73 Elementary diagram solenoid valve, sheet 35 Letdown line isolation valve Accumulator fill line isolation valve Accumulator nitrogen supply header isolation valve Accumulator test line to refueling water 177365 11/01/73 Elementary diagram solenoid valve, sheet 36 Boron injection tank recirculation isolation valve Boron injection recirculation pump to boron injection tank isolation valve 177368 11/01/73 Elementary diagram solenoid valve, sheet 34 Accumulator test line isolation valve 177309 11/01/73 Elementary diagram boron injection tank heaters A and B 177313 11/01/73 Elementary diagram boron injection surge tank heater 177375 11/01/73 Elementary diagram solenoid valve, sheet 43 Letdown to demineralizer or volume control tank valve 177376 11/01/73 Elementary diagram solenoid valve, sheet 49 Letdown orifice isolation valve 177377 11/01/73 Elementary diagram solenoid valve, sheet 50 Letdown orifice isolation valve 177378 11/01/73 Elementary diagram solenoid valve, sheet 51

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 15 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title Letdown orifice isolation valve 177379 11/01/73 Elementary diagram solenoid valve, sheet 42 Boric acid filter to boric acid blender valve 177381 11/01/73 Elementary diagram solenoid valve, sheet 45 Pressurizer power relief valve 177382 11/01/73 Elementary diagram solenoid valve, sheet 48 Pressurizer relief tank to reactor Makeup water supply isolation valve Pressurizer relief tank vent to waste process system isolation valve 177383 11/01/73 Elementary diagram solenoid valve, sheet 46 Reactor coolant drain tank pump discharge valve 177384 11/01/73 Elementary diagram solenoid valve, sheet 47 Reactor coolant drain tank vent isolation valve 177508 11/01/73 Elementary diagram solenoid valve, sheet 53 Waste gas discharge control valve 177509 11/01/73 Elementary diagram solenoid valve, sheet 54 Boric acid makeup injection valve to charging pump heater 177510 11/01/73 Elementary diagram solenoid valve, sheet 55 Boric acid dilution injection valve to volume control tank 177511 11/01/73 Elementary diagram solenoid valve, sheet 56

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 16 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title Reactor makeup water to boric acid blender valve 177567 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 41 Reactor coolant pump seal water return isolation valve 177568 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 42 Containment spray pump to spray nozzles isolation valve 177569 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 43 RHR system inlet isolation valve 177570 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 44 RHR system outlet isolation valve 177571 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 45 Low heat safety injection to reactor coolant system cross- over 177572 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 40 RHR system inlet isolation valve 177585 11/01/73 Elementary diagram solenoid valve, sheet 30 Letdown line isolation valve 177586 11/01/73 Elementary diagram solenoid valve, sheet 29 Letdown line isolation valve 177587 11/01/73 Elementary diagram solenoid valve, sheet 28 Letdown to volume control tank 177593 11/01/73 Reactor coolant drain tank pump discharge Reactor coolant drain tank vent Pressurizer relief tank 2 supply isolation valve

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 17 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 177602 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 39 Volume control tank outlet isolation valve 177603 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 38 Refueling water storage tank to charging pump valve 177604 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 37 Volume control tank outlet isolation valve 177606 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 3 Charging/safety injection pumps section heads isolation valve Charging/safety injection pumps discharge heater isolation valve Refueling water storage tank to RHR pumps 1A and 1B isolation valve 177607 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 4 Containment sump to RHR pump 1B isolation valve Containment sump to RHR pump 1A isolation valve MMSI to reactor coolant system hot leg HHSI to reactor coolant system hot leg LHSI to reactor coolant system hot leg LHSI to reactor coolant system cold leg 177608 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 5 Charging/safety injection pumps miniflow isolation valve

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 18 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title Charging/safety injection pumps to reactor coolant system isolation valve 177609 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 6 Accumulator 1A, 1B, and 1C discharge 177614 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 11 Boron injection tank outlet isolation valve Boron injection tank inlet isolation valve 177615 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 12 Pressurizer power relief isolation valve 177631 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 28 Refueling water storage tank to charging pump 177634 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 31 Reactor coolant pumps seal water return isolation valve 177637 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 34 Spray additive tank outlet isolation valve 177638 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 35 Containment spray pump inlet 177639 11/01/73 Elementary diagram 575-V motor- operated valve, sheet 36 Containment sump outlet valve 177858 11/01/73 Elementary diagram solenoid valve, sheet 9 Excess letdown isolation valve 177861 11/01/73 Elementary diagram solenoid valve, sheet 12 Reactor coolant system normal charging line

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 19 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title Reactor coolant system alternate charging line 177362 11/30/73 Elementary diagram solenoid valve, sheet 77 177373 11/30/73 Elementary diagram solenoid valve, sheet 78 177374 11/30/73 Elementary diagram solenoid valve, sheet 79 177523 11/30/73 Elementary diagram solenoid valve, sheet 80 Elementary Diagrams and Physical Drawings, 172000 Series 172062 02/01/74 Conduit template 600-V switchgear buses 1H and 1J 172100 02/01/74 Outdoor duct runs general arrangement 172101 02/01/74 Outdoor electrical duct runs profile river duct 1A 172102 02/01/74 Outdoor electrical duct runs profile river duct 1B 172103 02/01/74 Outdoor electrical duct runs profile service water duct 1A 172104 02/01/74 Outdoor electrical duct runs profile service water duct 1E 172239 02/01/74 Details and assembly of service water undervoltage Detector cabinet service water battery 172240 02/01/74 Details and assembly of service water battery fuse boxes 172270 02/01/74 Electrical penetrations of river water and service water intake structure 172285 02/01/74 Class 1 cable tray support post 172290 02/01/74 Compression type cable transit river and service water intake structures 172292 02/01/74 Class 1 cable tray support bracket 172328 02/01/73 Bill of material service water intake structure 172329 02/01/74 Bill of material river water intake structure

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 20 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172338 02/01/74 Bill of material undervoltage detector cabinet Service water intake structure batteries 172366 02/01/74 125 V-dc distribution cabinet service water intake structure train A 172367 02/01/74 125 V-dc distribution cabinet service water intake structure train B 172369 02/01/74 120/208-V distribution cabinet river water intake structure train A 172370 02/01/74 120/208-V distribution cabinet river water intake structure train B 172371 02/01/74 120/208-V distribution cabinet service water intake structure train A 172372 02/01/74 120/208-V distribution cabinet service water structure train B 172373 02/01/74 Anchor bolt assembly for cable tray support post 172063 03/01/74 Conduit template 600-V switchgear buses 1R and 1S 172064 03/01/74 Conduit template 4160-V switchgear buses 1H and 2H 172065 03/01/74 Conduit template 4160-V switchgear buses 1J and 2J 172143 03/01/74 Outdoor ducts Class 1 diesel building to valve boxes and fuel oil tank 172155 05/15/74 Sections and details Class 1 ducts, diesel building area 172169 03/01/74 Diesel building lightning protection and roof grounding 172170 03/01/74 Grounding plan, diesel building to valve boxes 172171 03/01/74 Electrical equipment plan, diesel building 172172 03/01/74 Electrical sections and details, diesel building, sheet 1 172195 05/15/74 Diesel building slab

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 21 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172196 05/15/74 Electrical section and details of conduit below slab, diesel building 172197 05/15/74 Embedded supports and conduits in wall 172211 05/15/74 Cable tray layout and exposed conduit, diesel building, sheet 1 172212 05/15/74 Cable tray layout and exposed conduit, diesel building, sheet 2 172213 05/15/74 Cable tray layout and exposed conduit, diesel building, sheet 3 172214 05/15/74 Cable tray layout and exposed conduit, diesel building, sheet 4 172230 05/15/74 Enlarged end and cable tray, partial plan, river intake structure, sheet 1 172231 05/15/74 Enlarged end and cable tray, partial plan, river intake structure, sheet 2 172232 05/15/74 Conduit plan for valve boxes 172233 05/15/74 Conduit plan valve box river water supply 172173 03/01/74 Electrical sections and details, diesel building, sheet 2 172174 03/01/74 Electrical sections and details, diesel building, sheet 3 172178 03/01/74 Electrical sections and details, diesel building, sheet 4 172195 03/01/74 Embedded conduit, diesel building slab 172196 03/01/74 Sections and details of conduit below slab, diesel building 172197 03/01/74 Embedded supports and conduit in walls, diesel building 172203 03/01/74 Diesel building cable tray and support plan, sheet 1 172211 03/01/74 Cable tray layout, diesel building, sheet 1

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 22 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172212 03/01/74 Cable tray layout, diesel building, sheet 2 172213 03/01/74 Cable tray layout, diesel building, sheet 3 172214 03/01/74 Cable tray layout, diesel building, sheet 4 172243 03/01/74 Screen enclosure for diesel generator neutral resistor 172264 03/01/74 Details and assembly Class 1 emergency ventilation station 172265 03/01/74 Details and assembly Class 1 ventilation local control station 172266 03/01/74 Details and assembly Class 1 heater local control station 172204 02/01/74 General arrangement cable tunnel, sheet 1 172205 02/01/74 General arrangement cable tunnel, sheet 2 172206 02/01/74 General arrangement cable tunnel, sheet 3 172232 04/05/74 Conduit plan valve boxes, sheet 1 172233 04/05/74 Conduit plan valve boxes, sheet 2 172234 04/05/74 Conduit plan valve boxes, sheet 3 172311 03/01/74 Bill of material cable trays 172312 03/01/74 Bill of material cable tray supports 172313 03/01/74 Bill of material cable tunnel 172314 03/01/74 Bill of material diesel building 172340 02/01/74 Details and assembly of switchgear channels 172384 03/01/74 120/208-V distribution cabinet diesel 1C 172385 03/01/74 120/208-V distribution cabinet diesel 2C 172386 03/01/74 120/208-V distribution cabinet diesel 1-2A 172387 03/01/74 120/208-V distribution cabinet diesel 1B

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 23 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172700 11/16/73 Main single line diagram generator and 4160-V transformer 172701 11/16/73 Single line 4160-V emergency station service 172202 11/16/73 Single line 4160-V emergency station service 172204 11/16/73 Single line 600-V emergency station service 172207 11/16/73 Single line and cable diagram dc distribution train E service water building 172708 11/16/73 Single line and cable diagram dc distribution train E service water building 172713 11/16/73 Bill of material relay panels 1 through 11 172714 11/16/73 Front view meter and relay panels 1 through 11 172723 12/21/73 Elementary diagram turbine auxiliary auto stop trips and emergency trip and vacuum reset 172732 11/16/73 Elementary diagram generator relaying 172741 11/16/73 Elementary diagram fire protection jockey pump 172744 11/16/73 Wiring diagram DEH valve test panel junction boxes 1 and 2 172745 11/16/73 Elementary diagram station service air compressor 1A 172747 01/04/74 Elementary diagram service water pump 1A 172748 01/04/74 Elementary diagram service water pump 1B 172749 01/04/74 Elementary diagram service water pump 1C (bus 1K) 172750 01/04/74 Elementary diagram service water pump 1C (bus 1L) 172751 01/04/74 Elementary diagram service water pump 1D 172752 01/04/74 Elementary diagram service water pump 1E

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 24 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172761 01/04/74 Elementary diagram 4160-V bus 1H incoming breaker from diesel generator 1C 172762 11/30/73 Elementary diagram 4160-V bus 1H feeder breakers to station service transformers 1H and 1R 172763 01/18/74 Elementary diagram 4160-V bus 1J incoming breaker from diesel generator 2C 172764 11/30/73 Elementary diagram 4160-V bus 1J (emergency) feeder breaker to station service transformers 1J and 1S 172765 11/30/73 Elementary diagram 4160-V bus 1K feeder breaker station service transformer 1K 172766 11/30/73 Elementary diagram 4160-V bus 1E (emergency) feeder breaker to station service transformer 1L 172767 01/31/74 Elementary diagram 600-V buses 1G, 1P, and 1Q incoming breaker 172768 01/31/74 Elementary diagram feeder breaker 600-V buses 1G, 1P, and 1Q 172769 01/31/74 Elementary diagram 600-V buses 1G, 1P, and 1Q bus tie breaker from bus 1F 172770 11/16/73 Bill of material diesel generator relay panels 172771 11/16/73 Front view diesel generator relay panels typical for 1-2A, 1B, 2B, 1C, and 2C 172772 01/31/74 Elementary diagram diesel generator 1-2A relaying 172773 01/31/74 Elementary diagram diesel generator 1-2A metering 172774 02/28/74 Elementary diagram diesel generator 1-2A start, stop, and shutdown

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 25 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172775 02/28/74 Elementary diagram diesel generator 1-2A exciter and miscellaneous controls 172776 01/31/74 Elementary diagram diesel generator 1B relaying 172777 01/31/74 Elementary diagram diesel generator 1B metering 172778 02/28/74 Elementary diagram diesel generator 1B start, stop, and shutdown 172779 02/28/74 Elementary diagram diesel generator 1B exciter and miscellaneous controls 172780 01/31/74 Elementary diagram diesel generator 1C relaying 172781 01/31/74 Elementary diagram diesel generator 1C metering 172782 02/15/74 Elementary diagram diesel generator 1C start, stop, and shutdown 172783 02/28/74 Elementary diagram diesel generator 1C exciter and miscellaneous controls 172784 11/16/73 Elementary diagram generator and transformer auxiliary relays 172787 11/16/73 Elementary diagram startup auxiliary transformers 1A and 1B protective relaying 172791 01/31/74 Elementary diagram diesel generator 2C relaying 172792 01/31/74 Elementary diagram diesel generator 2C metering 172793 02/28/74 Elementary diagram diesel generator 2C start, stop, and shutdown 172794 02/28/74 Elementary diagram diesel generator 2C exciter and miscellaneous controls 172795 11/30/73 Elementary diagram 4160-V bus 1H feeder breaker to station service trans- former 1G

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 26 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172796 11/30/73 Elementary diagram 4160-V bus 1H differential relaying 172797 11/30/73 Elementary diagram 4160-V bus 1J differential relaying 172798 11/30/73 Elementary diagram 4160-V bus 1K differential relaying 172799 11/30/73 Elementary diagram 4160-V bus 1L differential relaying 172818 01/04/74 Elementary diagram river and service water motor-operated and solenoid-operated valves 172825 01/31/74 Elementary diagram 600-V buses 1H, 1J, 1K, and 1L auxiliary breakers and potential transformers, sheet 1 172826 01/31/74 Elementary diagram 600-V buses 1H, 1J, 1K, and 1L feeder breakers and bus tie breakers, sheet 2 172827 01/31/74 Elementary diagram 600-V buses 1O, 1P, and 1Q potential transformers 172828 11/30/73 Elementary diagram 4160-V bus 1H potential transformer 172829 11/30/73 Elementary diagram 4160-V bus 1J potential transformer 172830 11/30/73 Elementary diagram 4160-V buses 1K and 1L potential transformer 172831 01/31/74 Elementary diagram 600-V buses 1R and 1S, sheet 1 172832 01/31/74 Elementary diagram 600-V buses 1R and 1S, sheet 2 172852 01/04/74 Elementary diagram startup auxiliary transformers 1A and 1B controls 172857 01/04/74 Elementary diagram motor control center 1K 172858 03/22/74 Elementary diagram motor control center 1L 172860 03/22/74 Elementary diagram motor control center 1N

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 27 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172861 03/22/74 Elementary diagram motor control center 1P 172862 03/22/74 Elementary diagram motor control center 1S 172863 03/22/74 Elementary diagram motor control center 1T 172864 03/22/74 Elementary diagram motor control center 1X 172865 03/22/74 Elementary diagram motor control center 1Y 172868 11/16/73 Wiring diagram fire protection engine-driven fire pumps 172869 11/16/73 Elementary diagram motor- driven fire pump 172870 02/28/74 Single line and cable diagram fire protection 600-V and 120/208-V distribution cabinets 172875 01/04/74 Elementary diagram river water pump 4 172876 01/04/74 Elementary diagram river water pump 5 172877 01/04/74 Elementary diagram river water pump 8 172878 01/04/74 Elementary diagram river water pump 9 172879 01/04/74 Elementary diagram river water pump 10 172960 01/31/74 Elementary diagram motor- operated valves diesel generator cooling 172963 02/15/74 Elementary diagram diesel generator storage tank fuel pumps 172973 01/31/74 Elementary diagram diesel generator 2B relaying 172974 01/31/74 Elementary diagram diesel generator 2B metering 172975 02/28/74 Elementary diagram diesel generator 2B start, stop, and shutdown 172976 02/28/74 Elementary diagram diesel generator 2B exciter and miscellaneous controls

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 28 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 6893D82 08/15/74 Elementary diagram reactor trip switchgear 172713 10/16/74 Bill of material for D-172714 front view meter and relay PNLS, sheet 7 172713 10/16/74 Bill of material for D-172714 front view meter and relay PNLS, sheet 8 172723 10/16/74 Elementary diagram river water pumps cooling and lube water strainers 172732 10/16/74 Elementary diagram generator relaying 172744 10/16/74 Wiring diagram DEH valve test panel junction boxes 1 and 2 172761 10/16/74 Elementary diagram 4160-V bus 1H incoming breaker from diesel generator 1C 172763 10/16/74 Elementary diagram 4160-V bus 1J incoming breaker from diesel generator 2C 172770 10/16/74 Bill of material for C-172771 front view diesel generator relay PNLS, sheet 4 172825 10/16/74 Elementary diagram 600-V buses 1H, 1J, 1K, and 1L incoming breaker and potential transformer, sheet 2 172857 10/16/74 Elementary diagram motor control center 1K (service water intake structure) 172858 10/16/74 Elementary diagram motor control center 1L (service water intake structure) 172860 10/16/74 Elementary diagram motor control center 1N (diesel building) 172861 10/16/74 Elementary diagram motor control center 1P (diesel building)

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 29 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 172862 10/16/74 Elementary diagram motor control center 1S (diesel building) 172863 10/16/74 Elementary diagram motor control center 1T (diesel building) 172864 10/16/74 Elementary diagram motor control center 1X (river water intake structure) 172865 10/16/74 Elementary diagram motor control center 1Y (river water intake structure) 172963 10/16/74 Elementary diagram diesel generator storage tank fuel pumps, sheet 1 172963 10/16/74 Elementary diagram diesel generator storage tank fuel pumps, sheet 2 Location Drawings, 175000 Series 175055 11/01/73 Equipment location auxiliary building area plan at el 155 ft 175056 11/01/73 Equipment location auxiliary building area plan at el 139 ft 175057 11/01/73 Equipment location auxiliary building area plan at el 121 ft 175059 08/15/74 Equipment location auxiliary building roof plan at el 175 ft and above 175061 11/01/73 Equipment location auxiliary and control building area plan at el 139 ft 175062 11/01/73 Equipment location auxiliary and control building 175150 11/01/73 Instrumentation location containment and fuel handling area plan at el 105 ft 6 in. 175140 11/01/73 Instrumentation location auxiliary and control building area at el 155 ft

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 30 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 175141 11/01/73 Instrumentation location auxiliary and control building area plan at el 139 ft 175142 11/01/73 Instrumentation location auxiliary and control building area at el 121 ft 175143 11/01/73 Instrumentation location auxiliary and control building area at el 100 ft and below 175144 11/01/73 Instrumentation location auxiliary building area at el 155 ft 175145 11/01/73 Instrumentation location auxiliary building area at el 139 ft 175146 11/01/73 Instrumentation location auxiliary building area at el 121 ft 175147 11/01/73 Instrumentation location auxiliary building area at el 100 ft and below 175148 11/01/73 Instrumentation location containment and fuel handling area at el 155 ft 175149 11/01/73 Instrumentation location containment and fuel handling area at el 129 ft Piping and Instrumentation Drawings, 170000 Series; Instrument Installation Drawings, 170000 Series 170119 11/30/73 P&ID river water system, sheet 1 170119 11/30/73 P&ID service water system, sheet 2 170060 08/29/73 P&ID diesel generator fuel oil supply system (deleted)

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 31 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 170586 10/16/74 River water system automatic operations train A 170587 10/16/74 River water system automatic operations train B 170588 10/16/74 River water pump QSP25P004B 170589 10/16/74 River water pump QSP25P005B 170590 10/16/74 River water pump QSP25P008A 170591 10/16/74 River water pump QSP25P009A 170592 10/16/74 River water pump QSP25P010A 170593 10/16/74 River water motor-operated valve logic train A 170594 10/16/74 River water system valves train B 170623 10/16/74 River water lube water cyclone separator inlet motor-operated valves 170624 10/16/74 River water hand switch- operated motor-operated valves (typical) 170599 10/16/74 Service water pump 1A train A 170600 10/16/74 Service water pump 1B train A 170601 10/16/74 Service water pump 1C train A or B 170602 10/16/74 Service water pump 1D train B 170603 10/16/74 Service water pump 1E train B 170604 10/16/74 Service water diesel generator 2C Unit 1 train B isolation motor- operated valves 170605 10/16/74 Service water diesel generator 2C Unit 2 train B isolation motor- operated valves 170606 10/16/74 Service water diesel generator 1B Unit 1 train B isolation motor- operated valves

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 32 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 170607 10/16/74 Service water diesel generator 1B Unit 2 train B isolation motor- operated valves 170608 10/16/74 Service water diesel generator 2B Unit 1 train B isolation motor- operated valves 170609 10/16/74 Service water diesel generator 1C Unit 1 train A isolation motor- operated valves 170610 10/16/74 Service water diesel generator 1C Unit 2 train B isolation motor- operated valves 170611 10/16/74 Service water diesel generator 1-2A Unit 1 train A isolation motor- operated valves 170612 10/16/74 Service water diesel generator 1-2A Unit 2 train B isolation motor- operated valves 170613 10/16/74 Service water diesel generator building train B isolation motor- operated valves 170614 10/16/74 Service water diesel generator building train A isolation motor- operated valves 170615 10/16/74 Service water to turbine building isolation motor- operated valves 514 and 516 170616 10/16/74 Service water to turbine building isolation motor- operated valves 515 and 517 170617 10/16/74 Service water trains A and B strainer isolation motor- operated valves

FNP-FSAR-7 REV 21 5/08 TABLE 7.1-1 (SHEET 33 OF 33) Submittal to NRC Drawing (formerly Number AEC) Title 170618 10/16/74 Service water trains A and B emergency recirculation to pond motor-operated valves 170619 10/16/74 Service water push button- operated motor-operated valves (typical) 170622 10/16/74 Service water lube water cyclone separator inlet motor-operated valves 170625 10/16/74 Service water hand switch- operated motor-operated valves (typical) 170626 10/16/74 Service water system discharge backpressure control valves]

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HISTORICALDuring preliminary startup tests, it will be demonstrated that actual instrument errors and time delays are equal to or less than the values assumed in the accident analyses.

FNP-FSAR-7 REV 21 5/08 [HISTORICAL] [TABLE 7.2-3 (SHEET 1 OF 2) REACTOR TRIP SYSTEM INSTRUMENT ACCURACIES Reactor Trip Signal Accuracy Note Power range high neutron +/- 1 percent of full power flux Intermediate range high +/- 5 percent of full scale (a) neutron flux +/- 1 percent of full scale (a) from 10-4 to 10-3 amperes

Source range high +/- 5 percent of full scale (a) neutron flux Power range high +/- 5 percent (a) positive nuclear power rate Power range high +/- 5 percent negative nuclear power rate Overtemperature T +/- 3.2°F (a)

Overpower T +/- 2.7°F Pressurizer low +/- 18 psi pressure Pressurizer high +/- 14 psi pressure Pressurizer high +/- 2.3 percent of full water level range P between taps at design temperature and pressure

Low reactor coolant +/- 2.5 percent of full flow (a) flow within range of 70 percent of 100 percent of full flow Reactor coolant pump +/- 1 percent of relay set bus undervoltage voltage

FNP-FSAR-7 REV 21 5/08 TABLE 7.2-3 (SHEET 2 OF 2) Reactor Trip Signal Accuracy Note Reactor coolant pump +/- 0.1 Hz bus underfrequency Low-low steam generator +/- 2.9 percent of P signal water level over pressure range of 600 to 1100 psig (this does not include EA and PMA allowances)

a. Reproducibility.]

FNP-FSAR-7 REV 25 4/14 TABLE 7.2-4 (SHEET 1 OF 3) TRIP CORRELATION Reactor Trip Accident(a) Source range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B) 15.2.4-Boron dilution (B) 15.4.6-Rod ejection (B) Intermediate range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B) 15.2.4-Boron dilution (B) 15.4.6-Rod ejection (B) Power range, high flux 15.2.1-Uncontrolled RCCA bank withdrawal(low setpoint) from a subcritical condition (P) 15.2.4-Boron dilution (P) 15.2.6-Startup of an inactive reactor coolant loop (B) 15.2.10-Excessive heat removal due to feedwater system malfunction (B) 15.2.11-Excessive load increase (B) 15.4.6-Rod ejection (P) Power range, high flux 15.2.2-Uncontrolled RCCA bank withdrawal(high setpoint) at power (P) 15.2.4-Boron dilution (B) 15.2.6-Startup of an inactive reactor coolant loop (B) 15.2.10-Excessive heat removal due to feed- water system malfunction (B) 15.2.11-Excessive load increase (B) 15.4.6-Rod ejection (P)

Positive neutron flux rate 15.2.1-Uncontrolled RCCA bank withdrawal from a subcritical condition (B) 15.2.2-Uncontrolled RCCA bank withdrawal at power (P) 15.4.6-Rod ejection (B)

a. (B/P) - Backup/Primary trip designation based on FSAR Chapter 15 analysis.

FNP-FSAR-7 REV 25 4/14 TABLE 7.2-4 (SHEET 2 OF 3) Reactor Trip Accident Overpower T 15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.4-Boron dilution (B) 15.2.5Partial loss of forced reactor coolant system flow (B) 15.2.10-Excessive heat removal due to feedwater system malfunction (B) 15.2.11-Excessive load increase (B) 15.4.2-Main steam line break (P - at power) Overtemperature T 15.2.2-Uncontrolled RCCA bank withdrawal at power (P) 15.2.4-Boron dilution (P) 15.2.5Partial loss of forced reactor coolant system flow (B) 15.2.7-Loss of external electric load and/or turbine trip (P) 15.2.10-Excessive heat removal due to feedwater system malfunction (B) 15.2.11-Excessive load increase (B) 15.2.12-Accidental depressurization of the reactor coolant system (P) 15.3.6-Single RCCA withdrawal at power (P) 15.4.2-Feedline break (B) 15.4.3-Steam generator tube rupture (B)

Low primary coolant flow 15.2.5-Partial loss of forced reactor coolant system flow (P) 15.3.4-Complete loss of forced reactor coolant system flow (P) 15.4.4-Single reactor coolant pump locked rotor (P)

Reactor coolant pump, under- 15.3.4-Complete loss of forced reactor frequency or undervoltage coolant system flow (B)

Pressurizer high pressure 15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.7-Loss of external electrical load and/or turbine trip (P)

FNP-FSAR-7 REV 25 4/14 TABLE 7.2-4 (SHEET 3 OF 3) Reactor Trip Accident 15.4.2-Feedline break (B) 15.4.4-Single reactor coolant pump locked rotor (B)Pressurizer high water level 15.2.2-Uncontrolled RCCA bank withdrawal at power (B) 15.2.4-Boron dilution (B) 15.2.7-Loss of external electrical load and/or turbine trip (B) 15.2.8-Loss of normal feedwater (B) 15.2.9-Loss of offsite power to the station auxiliaries (station blackout) (B) 15.2.14-Inadvertent operation of ECCS during power operation (B) 15.4.2-Feedline break (B)

Pressurizer low pressure 15.2.3-RCCA misalignment (one or more dropped RCCAs) (B) 15.2.11-Excessive load increase (B) 15.2.12-Accidental depressurization of the reactor coolant system (B) 15.2.14-Inadvertent operation of ECCS during power operation (P) 15.3.1-Loss of reactor coolant from small ruptured pipes or from cracks in large pipes which actuate emergency core cooling system (small break LOCA (P) 15.4.2-Main steam line break (B) 15.4.3-Steam generator tube rupture (P)

Low-low steam generator 15.2.7-Loss of external electrical load water level and/or turbine trip (B) 15.2.8-Loss of normal feedwater (P) 15.2.9-Loss of offsite power to the station auxiliaries (station blackout) (P) 15.4.2-Feedline break (P)

Reactor trip from safety injection 15.4.2-Main steam line break (P - at power) signal (low steam line pressure)

FNP-FSAR-7 REV 25 4/14 TABLE 7.2-5 (SHEET 1 OF 2) REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES Functional Unit Response Time(s) 1. Manual reactor trip NA 2. Power range, neutron flux a. High 0.5 (a) b. Low 0.5 (a) 3. Power range, neutron flux, 0.65 (a) high positive rate 4. Not used. 5. Intermediate range, neutron fluxNA 6. Source range, neutron flux NA 7. Overtemperature T (a)(b)(c) 8. Overpower T ~ (b)(c) 9. Pressurizer pressure-low 2.0 10. Pressurizer pressure-high 1.0 11. Pressurizer water level-highNA 12A. Loss of flow-single loop (above P-8) 1.0 12B. Loss of flow-two loops (above P-7 1.0 and below P-8) 13. Steam generator water level-low-low 2.0 14. Undervoltage-reactor coolant pumpsNA 15. Underfrequency-reactor coolant pumpsNA 16. Turbine trip a. Low auto stop oil pressureNA b. Turbine throttle valve closureNA 17. Safety injection input from ESFNA

FNP-FSAR-7 REV 25 4/14 TABLE 7.2-5 (SHEET 2 OF 2) Functional Unit Response Time(s) 18. Reactor trip system interlocksNA 19. Reactor trip breakers NA 20. Automatic trip logic NA

a. Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

b. RTD response time 5.0 s The RTD response time cannot be summed with the channel response times listed in Note (c).
c. The following are the required RTS channel response times (encompassing channel electronics/trip logic & breaker/gripper release) for an RTD response time of no greater than 5.0 seconds:

1. Overtemperature T, Tavg input: 2.435 s 2. Overtemperature T, pressurizer pressure input (including sensor): 2.0 s 3. Overtemperature T, nuclear flux input: 2.0 s 4. Overpower T, Tavg input: 2.159 s 5. T input (to both OTT and OPT): 6.159 s Tavg and T response times include the effect of all transfer functions set to the recommended values.

REV 21 5/08 SETPOINT REDUCTION FUNCTION FOR OVERPOWER AND OVERTEMPERATURE T TRIPS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-1 REV 21 5/08 PRESSURIZER SEALED REFERENCE LEG LEVEL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-2 REV 21 5/08 DESIGN TO ACHIEVE ISOLATION BETWEEN CHANNELS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.2-3

[HISTORICAL][Actuation signal accuracies required for generating the required actuation signals for loss of coolant protection are as follows: Pressurizer pressure +/-14 psi (uncompensated) Actuation signal accuracies required in generating the required actuation signals for steam break protection are given: 1. Steam line pressure 4 percent 2. Steam flow signals 4.5 percent of maximum guarantee flow over pressure range (600 to 1100 psig) 3. Tavg 2ºF

4. Containment pressure 1.8 percent signal of full scale

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-1 (SHEET 1 OF 2) FUNCTIONS INITIATED BY ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Item Function 1 Reactor trip, provided one has not already been generated by the reactor trip system

2 Engineered safety features actuation system sequence, which actuates equipment that includes items 2a through 2g and ensures the proper sequencing of engineered safety features power demands on the engineered safety features buses supplied by either preferred or standby power supply 2a Cold leg injection isolation valves, which are opened for injection of borated water by safety injection pumps into the cold legs of the reactor coolant system. The receipt of a safety injection signal by the accumulator motor-operated valves is discussed in paragraph 6.3.2.2.7.

2b Charging pumps, residual heat removal pumps, and associated valving, which provide emergency makeup water to the cold leg of the reactor coolant system following a loss-of-coolant accident 2c Containment air recirculation fans and coolers, which serve to cool the containment and limit the potential for release of fission products from the containment by reducing the pressure following an accident 2d Component cooling pumps and valves

2e Service water pump and valves, which provide cooling water to the component cooling system heat exchangers and is thus the heat sink for containment cooling 2f Motor-driven auxiliary feedwater pumps and control valves

2g Penetration room filtration system

3 Phase A containment isolation, "T" signal, whose function is to prevent fission product release by isolating all nonessential process lines on receipt of the safety injection signal 4 Steam line isolation, to prevent the continuous, uncontrolled blowdown of more than one steam generator and thereby uncontrolled reactor coolant system cooldown 5 Main feedwater line isolation, to limit the energy release for a steam line break and to limit the extent of the reactor coolant system cooldown 6 Emergency diesel start, to ensure backup supply of power to emergency and supporting systems components 7 Control room intake duct isolation, to meet control room occupancy requirements following a loss-of-coolant accident FNP-FSAR-7 REV 21 5/08 TABLE 7.3-1 (SHEET 2 OF 2)

Item Function

8 Containment spray actuation, "P" signal, which performs the following functions listed as items 8a and 8b 8a Containment spray initiation, which serves to reduce containment pressure and temperature following a loss-of-coolant or a steam break accident 8b Phase B containment isolation initiation, other than safety injection lines which are not closed. The remaining process lines into containment are isolated following a loss of reactor coolant accident or a steam or feedwater line break within containment.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-2 (SHEET 1 OF 2) INSTRUMENTATION OPERATING CONDITIONS FOR ENGINEERED SAFETY FEATURES Number of Number of Channels Number Functional Unit Channels (a) to trip 1. Safety Injection 1a. Manual(a) 2 switches 1 switch 1b. Containment pressure high 3 2 1c. Differential pressure high between steam lines 9 (3 per steam line) 2 per steam line and 1/3 comparison between steam lines 1d. Pressurizer low pressure(b) 3 2 1e. Steam line low pressure(c) 3 pressure signals 2 2. Containment Spray 2a. Manual(a) 2 pairs of switches 2 switches per pair 2b. Containment pressure high-high-high 4 2 3. Auxiliary feedwater 3a. Motor driven pumps 3a1 Manual(d) 2 switches (1 switch per pump) 2 switches (1 switch per pump) 3a2 Steam generator water level low-low 3 per steam generator 2/3 in any steam generator 3a3 Safety injection See item 1 3a4 Trip of main feedwater pumps 4 (2 per pump) 2 (1 per pump)

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-2 (SHEET 2 OF 2) Number of Number of Channels Number Functional Unit Channels (a) to trip 3a5 AMSAC actuation 1 1 3b Turbine driven pump 3b1 Manual(e) 1 switch 1 switch 3b2 Steam generator water level low-low 3 per steam generator 2/3 in 2/3 steam generators 3b3 RCP bus undervoltage 3 bus 2 bus 3b4 AMSAC actuation 1 1

a. Each switch actuates both Train A & B. b. Permissible bypass if reactor coolant pressure is less than P-11. c. Permissible bypass if reactor coolant temperature is less than P-12. d. Motor driven AFW pump 1 switch actuates Train A and motor driven AFW pump 2 switch actuates Train B. e. Turbine driven AFW pump switch opens steam admission valves.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-3 (SHEET 1 OF 2) INSTRUMENTATION OPERATING CONDITIONS FOR ISOLATION FUNCTIONS Number of Number of Channels Number Functional Unit Channels to Trip 1. Containment Isolation 1a. Safety injection - Phase A See item 1 of table 7.3-2. 1b. Containment pressure high-high-high - Phase B See item 2b of table 7.3-2. 1c. Manual Phase A (a) 2 1 Phase B See item 2a of table 7.3-2. 2. Steam Line Isolation 2a. Steam flow high coincident with low-low Tavg Steam flow high 2 per steam line 1 high flow per steam line on 2/3 steam lines Low-low Tavg 1 per loop 2/3 low-low Tavg 2b. Steam line low pressure 1 per steam line 2/3 steam lines 2c. Containment pressure high-high 3 2 2d. Manual(a) 1 per loop 1 per loop a. Each switch actuates both Train A & B.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-3 (SHEET 2 OF 2) Number of Number of Channels Number Functional Unit Channels to Trip 3. Feedwater line isolation 3a. Safety injection See item 1 of table 7.3.2 3b. Steam generator water level high-high 3 per steam generator 2/3 in any steam generator 3c. Low Tavg coincident with reactor trip Low Tavg 1 per loop 2/3 low Tavg Reactor trip 2 1 4. Turbine Trip 4a. Safety injection See item 1 of table 7.3-2 4b. Steam generator water level high-high See item 3b See item 3b 4c. Reactor Trip 2 1 5. Steam generator feedwater pump trip (a) 5a. Safety injection See item 1 of table 7.3-2 5b. Steam generator water level high-high See item 3b See item 3b

a. Train A trips both feedwater pumps.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-4 (SHEET 1 OF 2) INTERLOCKS FOR ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Function Designation Input Performed P-4(a) Reactor trip Actuates turbine trip Prevents opening of main feedwater valves which were closed by safety injection or high steam generator water level Allows manual block of the automatic reactuation of safety injection Blocks steam dump control via load rejection Tavg controller Makes steam dump valves available for either tripping or modulation Reactor not Defeats the manual tripped block preventing automatic reactuation of safety injection Block steam dump control via plant trip Tavg P-11 2/3 pressurizer Allows manual block of pressure below safety injection actuation setpoint on low pressurizer pressure signal Blocks automatic opening of the power relief valves 2/3 pressurizer Defeats manual block of pressure above safety injection actuation setpoint Opens accumulator motor-operated isolation valves

___________________ a. See table 7.7-1 for control functions FNP-FSAR-7 REV 21 5/08 TABLE 7.3-4 (SHEET 2 OF 2) Function Designation Input Performed P-12 2/3 Tavg below Allows manual block of setpoint(a) safety injection actuation on low steam line pressure Blocks steam dump Allows manual bypass of steam dump block for the cooldown dump valves only 2/3 Tavg above Defeats the manual setpoint block of safety injection actuation on low steam line pressure Defeats the manual bypass of steam dump block P-14 2/3 steam Closes all feedwater generator water control valves level above setpoint on any Trips all main steam generator feedwater pumps which closes the pump discharge valves Actuates turbine trip

__________________ a. This signal, in coincidence with high steam line flow in 2/3 steam lines, actuates steam line isolation.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-5

(THIS TABLE INTENTIONALLY DELETED)

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 1 OF 15) FAILURE MODE AND EFFECTS ANALYSIS, SERVICE WATER SYSTEM Component Identification Service Water Pumps Logic Diagram Number NA Elementary Number D-172747 through D-172752, Engineering Flow Diagram Number: D-202747 through D-202752 D-170119 Sh. 1 & D-200013 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 4.16-kV Failure of pump to start 4.16-kV bus U/V or bus breaker Redundant train pumps can be bus power when required or automatic auto trip alarm on emergency started stoppage when pump is power board running Loss of 4.16-kV Interruption of service Breaker automatic trip Two pumps are provided per train; power to motor water supply to one alarm on main control the standby pump will be started due to train board from the main control board automatic breaker trip Loss of 125 V-dc Inability of pump to start During testing or observation Two pumps are provided per train; the breaker control on manual or automatic of control switch indicating standby pump will be started from the power to one signal, or trip the breaker lights main control board; a failure to trip a pump when required breaker may cause loss of bus on one train, redundant train pumps can be started manually Failure of loss of power Inability of both pumps in ESFAS malfunction alarm on Both redundant pumps in other train sequencer or ESS a train to start main control board or start automatically on loss of sequencer start automatically periodic testing offsite power; pumps can be started signal to both manually from control switches on pumps in one the main control board train

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 2 OF 15) Component Identification 3019A,B,C,D and 3024A,B,C,D Logic Diagram Number NA Elementary Number C-177613/D-207613 Engineering Flow Diagram Number: D-175003 Sh. 1, D-205003 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-610 Two valves in the following Depending on initial valve Valves fail as is; normal position fails to operate grouping would not operate: position and plant operating is open; post-LOCA position is open; on receipt of MOV 3019 A/B and 3024 A/B or status: operator can open valves safety injection MOV 3019 C/D and 3024 C/D signal a. Computer a. If valve initially b. Light monitor panel open, no effect on c. Periodic testing system d. Position indication b. If valve initially closed, lights at main control system reduced board to 2/4 (minimum requirements) Loss of power Two valves in the following Depending on initial valve Position indication lights of main to motor control grouping would not operate: position and plant operating control board will be out center U or V MOV 3019 A/B and 3024 A/B or status: MOV 3019 C/D and 3024 C/D a. Computer a. If valve initially b. Light monitor panel open, no effect on c. Periodic testing system d. Loss of valve position b. If valve initially at main control board closed, system reduced to 2/4 (minimum requirements) c. Loss of ability to close valve Contacts of relay Valve fails as is; if valve Depending on initial valve Operator can open valves K-610 fail to initially closed, system position and plant operating close on receipt reduced to 3/4 status: of safety injection signal a. Computer or failure of b. Light monitor panel open/close relay c. Periodic testing to operate d. Position indication lights at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 3 OF 15) Component Identification 3019A,B,C,D and 3024A,B,C,D (cont.) Logic Diagram Number NA Elementary Number C-177613/D-207613 Engineering Flow Diagram Number: D-175003 Sh. 1, D-205003 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks Thermal overload Valve fails as is; if valve Depending on initial valve relay contacts initially closed, system position and plant operating open reduced to 3/4; loss of status: ability to close valve a. Computer b. Light monitor panel c. Periodic testing d. Position indication lights at main control board Loss of 120 V-ac Valve fails as is; if valve Depending on initial valve Position indication lights or main control power initially closed, system position and plant operating control board will be out reduced to 3/4 status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 4 OF 15) Component Identification 3131 Logic Diagram Number NA Elementary Number C-177612/D-207612 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA fails to valve (3134) operates to position and plant operating position closed; two containment operate on effect isolation; operator status: isolation valves in series, one safety injection can close if initially open required to operate signal (normal) a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve fails as is; alternate Depending on initial valve relay K-604 fail valve (3134) operates to position and plant operating to close on effect isolation; operator status: safety injection can close if initially open signal (normal) a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power to Valve fails as is Depending on initial valve motor control position and plant operating center U status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

REV 15 FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 5 OF 15) Component Identification 3131 (cont.) Logic Diagram Number NA Elementary Number C-177612/D-207612 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; operator Depending on initial valve starter to can close/open valve position and plant operating operate depending on which starter status: coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 6 OF 15) Component Identification 3134 Logic Diagram Number NA Elementary Number D-177636/D-207636 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA fails to valve (3131) operates to position and plant operating position closed; two containment operate on effect isolation; operator status: isolation valves in series, one safety injection can close valve if required to operate signal initially open (normal) a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve fails as is; alternate Depending on initial valve relay K-604 fail valve (3131) operates to position and plant operating to close on effect isolation; operator status: safety injection can close valve if signal initially open (normal) a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power to Valve fails as is Depending on initial valve motor control position and plant operating center V status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 7 OF 15) Component Identification 3134 (cont.) Logic Diagram Number NA Elementary Number D-177636/D-207636 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status: a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; operator Depending on initial valve starter relay can close/open valve position and plant operating to operate depending on which starter status: coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 8 OF 15) Component Identification 3135 Logic Diagram Number NA Elementary Number D-177636/D-207636 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-604 Valve fails as is; alternate Depending on initial valve Normal valve position open; post-LOCA fails to valve (QV075) is a check valve position and plant operating position closed; two containment operate on which operates to effect status: isolation valves in series, one safety injection isolation; operator can required to operate signal close valve if initially open a. Computer (normal) b. Light monitor panel c .Periodic testing d. Position indication at main control board Contacts of ESF Valve fails as is; alternate Depending on initial valve relay K-604 fail valve (QV075) is a check valve position and plant operating to close on which operates to effect status: safety injection isolation; operator can signal close valve if initially open a. Computer (normal) b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power to Valve fails as is Depending on initial valve motor control position and plant operating center V status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 9 OF 15) Component Identification 3135 (cont.) Logic Diagram Number NA Elementary Number D-177636/D-207636 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; operator Depending on initial valve starter relay can close/open valve position and plant operating to operate depending on which starter status: coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 10 OF 15) Component Identification 3149 Logic Diagram Number NA Elementary Number C-177617/D-207617 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-604 Valve 3149 fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operator can close position and plant operating position closed operate on valves if initially open status: safety injection (normal) signal a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve 3149 fails as is; Depending on initial valve relay K-604 fail operator can close position and plant operating to close on valves if initially open status: safety injection (normal) signal a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power to Valves fail as are Depending on initial valve motor control position and plant operating center U status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 11 OF 15) Component Identification 3149 (cont.) Logic Diagram Number NA Elementary Number C-177617/D-207617 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; operator Depending on initial valve starter relay can close/open valve position and plant operating to operate depending on which starter status: coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 12 OF 15) Component Identification 3150 Logic Diagram Number NA Elementary Number C-177617/D-207617 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-604 Valve 3150 fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operator can close position and plant operating position closed operate on valves if initially open status: safety injection (normal) signal a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve 3150 fails as is; Depending on initial valve relay K-604 fail operator can close position and plant operating to close on valves if initially open status: safety injection (normal) signal a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power to Valves fail as are Depending on initial valve motor control position and plant operating center U status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 13 OF 15) Component Identification 3150 (cont.) Logic Diagram Number NA Elementary Number C-177617/D-207617 Engineering Flow Diagram Number: D-175003 Sh. 2, D-205003 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; operator Depending on initial valve starter relay can close/open valve position and plant operating to operate depending on which starter status: coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 14 OF 15) Component Identification 3441A,B,C,D Logic Diagram Number NA Elementary Number D-177633/D-207633 Engineering Flow Diagram Number: D-175003 Sh. 1, D-205003 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-610 Two valves in the following Depending on initial valve Valves fail as is; normal valve fails to groupings would not operate: position and plant operating position is open; post-LOCA position operate on MOV 3441 A/B or MOV 3441 C/D status: is open; operator can open valves receipt of safety injection a. If valves initially a. Computer signal open, no effect on b. Light monitor panel system c. Position indication b. If valves initially light at main control closed, system board reduced to 2/4 d. Periodic testing (minimum requirements)

Loss of power to Two valves in the following Depending on initial valve motor control groupings would not operate: position and plant operating center U or V MOV 3441 A/B or MOV 3441 C/D status:

a. If valves initially a. Computer open, no effect on b. Light monitor panel system c. Loss of valve position b. If valves initially indication at main closed, system control board reduced to 2/4 d. Periodic testing (minimum requirements) c. Loss of ability to close valve Contacts of Valve fails as is; if valve Depending on initial valve Operator can open valves relay K-610 is initially closed, system position and plant operating fail to close is reduced to 3/4 status: on receipt of safety injection a. Computer signal or b. Light monitor panel failure of starter c. Position indication relay to operate at main control board d. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-6 (SHEET 15 OF 15) Component Identification 3441A,B,C,D (cont.) Logic Diagram Number NA Elementary Number D-177633/D-207633 Engineering Flow Diagram Number: D-175003 Sh. 1, D-205003 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks Thermal overload Loss of ability to close Depending on initial valve relay contacts valve; valve fails as is; position and plant operating open if valve is initially status: closed, system is reduced to 3/4 a. Computer b. Light monitor panel c. Position indication at main control board d. Periodic testing Loss of 120 V-ac Loss of ability to close Depending on initial valve control power valve; valve fails as is; position and plant operating if valve is initially status: closed, system is reduced to 3/4 a. Computer b. Light monitor panel c. Loss of valve position indication at main control board d. Periodic testing

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 1 OF 17) FAILURE MODE AND EFFECTS ANALYSIS, COMPONENT COOLING WATER SYSTEM Component Identification Component Cooling Water Pumps Logic Diagram Number NA Elementary Number D-177183, D-177184, D-177185, D-177187, Engineering Flow Diagram Number: D-207183, D-207184, D-207185, D-207187 D-175002, Sh. 1, Sh. 2, and Sh. 3, D-205002, Sh. 1, Sh. 2, and Sh. 3 Failure Mode Effect on System Detection of Failure Remarks Loss of 4.16-kV Failure of pump to start when 4.16-kV bus U/V or bus Redundant train pump can be started. power to one required or automatic breaker auto trip alarm pump stoppage when pump is running on emergency power board Loss of 125 V-dc Inability of pump to start on During monthly testing or Three pumps are provided; one pump breaker control manual or automatic signal, or observation of control is required for normal, hot shutdown, power trip the breaker when required switch red, green, and or post-LOCA heat removal. A failure amber lights to trip a breaker may cause loss of bus on one train, redundant train pump can be started manually. Failure of loss of Inability of pump to start ESS malfunction alarm Pump can be manually started from power sequencer upon receipt of automatic on main control board or the main control board start or ESS start signal periodic testing sequencer start signals Automatic breaker Standby (swing) pump Breaker auto trip alarm Swing pump can be put into service trip due to automatically starts if on main control board manually; one pump is required for overcurrent aligned with pump that normal, hot shutdown, or post-LOCA tripped heat removal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 2 OF 17) Component Identification 3067 Logic Diagram Number NA Elementary Number F-177851/D-207851 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating closed on loss of instrument air on receipt of series (3067 and 3443); status: CIAS phase A only one required to operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating to open on series (3067 and 3443); status: receipt of only one required to CIAS phase A operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at control power main control board out b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139)

Solenoid valve Valve remains open; however, Depending on initial valve 3067 fails to there are two valves in position and plant operating vent (sticky series (3067 and 3443); status: operator) only one required to operate to cause a. Valve position indication isolation at main control board b. Light monitor panel c. Computer d. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 3 OF 17)

Component Identification 3067 (Cont.) Logic Diagram Number NA Elementary Number F-177851/D-207851 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of Valve fails closed a. Valve position light at main instrument air control board b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139) `

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 4 OF 17)

Component Identification 3095 Logic Diagram Number NA Elementary Number F-177851/D-207851 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating closed on loss of instrument air on receipt of series (3095 and a check status: CIAS phase A valve); only one required to operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating to open on series (3095 and a check status: receipt of valve); only one required CIAS phase A to operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at control power main control board out b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139) Solenoid valve Valve remains open; however, Depending on initial valve 3095 fails to there are two valves in position and plant operating vent (sticky series (3095 and a check status: operator) valve); only one required to operate to cause isolation a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 5 OF 17)

Component Identification 3095 (cont.) Logic Diagram Number NA Elementary Number F-177851/D-207851 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of Valve fails closed a. Valve position light at instrument air main control board b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139)

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 6 OF 17)

Component Identification 3096A,B Logic Diagram Number NA Elementary Number D-177853/D-207853 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-609 Valves remains open; operator Depending on initial valve Valves are normally open and will fail fails to operate can close valves from main position and plant operating closed on loss of instrument air on receipt of control board status: safety injection signal a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valves remain open; operator Depending on initial valve relay K-609 fail can close valves from position and plant operating to open on main control board status: receipt of safety injection signal a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at control power main control board out b. Computer c. Light monitor panel Solenoid valve Associated valve remains open; Depending on initial valve 3096A or 3096B other valve will be position and plant operating fails to operational status: vent (sticky operator) a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at instrument air main control board b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 7 OF 17) `

Component Identification 3045 Logic Diagram Number NA Elementary Number D-177854/D-207854 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-625 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating closed on loss of instrument air; on receipt of series (3045 and 3184); status: valve operation is not testable with CIAS phase B only one required to reactor coolant pumps operating operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-625 fails there are two valves in position and plant operating to open on series (3045 and 3184); status: receipt of only one required to CIAS phase B operate to cause a. Valve position indication isolation; operator can at main control board close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc Valve fails open Valve position lights at main control power control board Solenoid valve Valve remains open; however, Depending on initial valve 3045 fails to there are two valves in position and plant operating vent (sticky series (3045 and 3184); status: operator) only one required to operate to cause isolation a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of Valve closes a. Valve position lights at instrument air main control board b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 8 OF 17)

Component Identification 3046 Logic Diagram Number NA Elementary Number C- 177618/D-207618 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operate alternate valve 3182 position and plant operating position closed; two containment on CIAS phase B operates to effect status: isolation valves in series; one isolation; operator can required for isolation; valve not close valve if initially a. Computer testable at power open (normal) b. Light monitor panel c. Position indication lights at main control board d. Periodic testing Contacts of ESF Valve fails as is; Depending on initial valve relay K-619 alternate valve 3182 position and plant operating fail to close operates to effect status: on CIAS phase B isolation; operator can close valve if initially a. Computer open (normal) b. Light monitor panel c. Position indication at main control board d. Periodic testing Loss of power Valve fails as is Depending on initial valve to motor control position and plant operating center U status: a. Computer b. Light monitor panel c. Loss of valve position indication light at main control board d. Periodic testing

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 9 OF 17)

Component Identification 3046 (cont.) Logic Diagram Number NA Elementary Number C-177618/D-207618 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Loss of valve position light at main control board d. Periodic testing

Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status:

a. Computer b. Light monitor panel c. Position indication lights at main control board d. Periodic testing Failure of Valve fails as is; Depending on initial valve starter relay operator can close or open position and plant operating to operate valve depending on which status: starter coil fails a. Computer b. Light monitor panel c. Position indication lights at main control board d. Periodic testing

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 10 OF 17)

Component Identification 3052 Logic Diagram Number NA Elementary Number C-177625/D-207625 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operate alternate valve is check position and plant operating position closed; two containment on CIAS phase B valve which operates to status: isolation valves in series; one required to effect isolation; operator operate; valve not testable at power can close valve if initially a. Computer open (normal) b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve fails as is; Depending on initial valve relay K-619 alternate valve is check position and plant operating fail to close valve which operates to status: on CIAS phase B effect isolation; operator can close valve if initially a. Computer open (normal) b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power Valve fails as is Depending on initial valve to motor control position and plant operating center V status: a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 11 OF 17) Component Identification 3052 (cont.) Logic Diagram Number NA Elementary Number C-177625/D-207625 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status:

a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status:

a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; Depending on initial valve starter relay operator can close/open position and plant operating to operate valves depending on status: which starter coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 12 OF 17)

Component Identification 3182 Logic Diagram Number NA Elementary Number D-177610/D-207610 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-619 Valve fails as is; Depending on initial valve Normal valve position open; post-LOCA fails to operate alternate valve 3046 position and plant operating position closed; two containment on CIAS phase B operates to effect status: isolation valves in series; one required isolation; operator can to operate; valve not testable at power close valve if a. Computer initially open (normal) b. Light monitor panel c. Periodic testing d. Position indication at main control board Contacts of ESF Valve fails as is; Depending on initial valve relay K-619 alternate valve 3046 position and plant operating fail to close operates to effect status: on CIAS phase B isolation; operator can close valve if initially a. Computer open (normal) b. Light monitor panel c. Periodic testing d. Position indication at main control board Loss of power Valve fails as is Depending on initial valve to motor control position and plant operating center V status: a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 13 OF 17) Component Identification 3182 (cont.) Logic Diagram Number NA Elementary Number D-177610/D-207610 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status: a. Computer b. Light monitor panel c. Periodic testing d. Loss of valve position indication light at main control board Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating open status: a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board Failure of Valve fails as is; Depending on initial valve starter relay operator can close/open position and plant operating to operate valves depending on status: which starter coil fails a. Computer b. Light monitor panel c. Periodic testing d. Position indication at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 14 OF 17)

Component Identification 3184 Logic Diagram Number NA Elementary Number D-177855/D-207855 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-625 Valve remains open; however, Depending on initial valve Valve is normally open and will fails to operate there are two valves in position and plant operating fail closed on loss of instrument on receipt of series (3184 and 3045); status: air; valve operation is not testable CIAS phase B only one required to operate with reactor coolant pumps operating to cause isolation; operator a. Valve position indication can close valve from main at main control board control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valves remains open; however, Depending on initial valve relay K-625 fail there are two valves in position and plant operating to open on series (3184 and 3045); status: receipt of only one required to operate CIAS phase B to cause isolation; operator a. Valve position indication can close valve from main at main control board control board b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc Valve open Valve position lights at main control power control board Solenoid valve Valve remains open; however, Depending on initial valve 3184 fails to there are two valves in position and plant operating vent (sticky series (3184 and 3045); status: operator) only one required to operate to cause isolation a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of Valve closes a. Valve position lights at instrument air main control board b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 15 OF 17)

Component Identification 3443 Logic Diagram Number NA Elementary Number D-177374/D-207374 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valves are normally open and will fail closed fails to operate there are two valves in position and plant operating on loss of instrument air on receipt of series (3443 and 3067); status: CIAS phase A only one required to operate to cause isolation; operator a. Valve position indication can close valve from main at main control board control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating to open on series (3443 and 3067); status: receipt of only one required to operate CIAS phase A to cause isolation; operator a. Valve position indication can close valve from main at main control board control board b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at control power main control board out b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139) Solenoid valve Valve remains open; however, Depending on initial valve 3443 fails to there are two valves in position and plant operating vent (sticky series (3443 and 3067); status: operator) only one required to operate to cause isolation a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 16 OF 17)

Component Identification 3443 (cont.) Logic Diagram Number NA Elementary Number D-177374/D-207374 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks Loss of Valve fails closed a. Valve position light at instrument air main control board b. Computer c. Light monitor panel d. High temperature at discharge of excess letdown heat exchanger (TE-139)

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-7 (SHEET 17 OF 17)

Component Identification 2229 Logic Diagram Number NA Elementary Number D-177853/D-207853 Engineering Flow Diagram Number D-175002 Sh. 2, D-205002 Sh. 2 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-609 fails to Valve remains open; operator Depending on initial valve position and Valve is normally open and will fail closed on operate on receipt of can close valve from local plant operating status: loss of instrument air safety injection signal control station a. Valve position indication at local control station b. Light monitor panel c. Periodic testing Contacts of ESF relay Valve remains open; operator Depending on initial valve position and K-609 fail to open on receipt of safety injection can close valve from local control station plant operating status: signal a. Valve position indication at local control station b. Light monitor panel c. Periodic testing Loss of 125 V-dc control Valve fails closed a. Valve position light at local control power station out b. Light monitor panel Solenoid valve 2229A fails Associated valve remains Depending on initial valve position and to vent (sticky operator) open; other valve will be plant operating status: operational a. Valve position indication at local control station b. Light monitor panel c. Periodic testing Loss of instrument air Valve fails closed a. Valve position light at local control station b. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 1 OF 18) FAILURE MODE AND EFFECTS ANALYSIS, CONTROL ROOM AND AIR CONDITIONING AND FILTRATION SYSTEM Component Identification Filtration Fan Logic Diagram Number NA Elementary Number D-177270 Sh. 1 Engineering Flow Diagram Numbers: D-175012 Sh. 1 & D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-614 Motor fails to start; of two filtration a. Control room indication lights fails to operate fans, only one is required for b. Periodic testing on CIAS phase A, operation; operator can start motor c. Monitor light box abnormal Unit 1 or 2 if necessary Contacts of ESF Motor fails to start; of two filtration a. Control room indication lights relay K-614 fail to fans, only one is required for b. Periodic testing close on CIAS operation; operator can start motor c. Monitor light box abnormal phase A if necessary Loss of power to 208-V Motor fails to start a. Control room indication lights motor control center b. Periodic testing 1F and 1G Loss of 120 V-ac Motor fails to start a. Control room indication lights control power b. Periodic testing Thermal overload Motor fails to start a. Control room indication lights relay contacts open b. Periodic testing c. Motor overload trip alarm in the control room FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 2 OF 18) Component Identification Air Conditioning Unit Logic Diagram Number NA Elementary Number D-177270 Sh. 3 Engineering Flow Diagram Numbers: D-175012 Sh. 1 & D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-614 Motor fails to start; of two air a. Periodic testing Redundant train will start fails to operate conditioning units, only one is on CIAS phase A, required for operation; Unit 1 or 2 operator can start motor if necessary Contacts of ESF Motor fails to start; of two air a. Periodic testing relay K-614 fail to conditioning units, only one is close on CIAS required for operation; phase A operator can start motor if necessary Loss of power Motor fails to start a. Periodic testing to 600-V motor control center 1F & 1G Loss of 120 V-ac Motor fails to start a. Periodic testing control power Thermal overload Motor fails to start a. Periodic testing relay contacts b. Motor overload trip alarm open in the control room

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 3 OF 18) Component Identification 3478A Logic Diagram Number NA Elementary Number D-177280 Sh. 1 Engineering Flow Diagram Numbers: D-175012 Sh. 1

Failure Mode Effect on System Detection of Failure Remarks ESF relay K-614 or its Valve fails as is Depending on initial valve Normal valve position closed; contact fails to operate position and plant operating status: post-LOCA position open on control room on CIAS phase A pressurization fan start signal a. Computer Redundant valve will open b. Position indication lights at BOP panel c. Periodic testing Contacts of 42X Valve fails as is Depending on initial valve Redundant valve will open relay (Control Room position and plant operating status: Pressurization Fan) fails to operate on CIAS phase A a. Computer b. Position indication lights at BOP Panel c. Periodic testing Loss of power to Valve fails as is Depending on initial valve 600-V motor position and plant operating status: control center 1F a. Computer b. Loss of valve position indication light at BOP Panel c. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 4 OF 18) Component Identification 3478A (cont.) Logic Diagram Number NA Elementary Number D-177280 Sh. 1 Engineering Flow Diagram Numbers: D-175012 Sh. 1

Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status: a. Computer b. Periodic testing c. Loss of valve position indication lights at BOP panel Thermal overload Valve fails as is Depending on initial valve relay contacts position and plant operating status: open a. Computer b. Loss of position indication lights at BOP panel c. Periodic testing Failure of (42) Valve fails as is Depending on initial valve starter relay position and plant operating status: to operate a. Computer b. Position indication lights at main control board c. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 5 OF 18) Component Identification 3478B Logic Diagram Number NA Elementary Number D-177280 Sh. 1 Engineering Flow Diagram Numbers: D-205012 Sh. 1

Failure Mode Effect on System Detection of Failure Remarks ESF relay K-614 or its Valve fails as is Depending on initial valve Normal valve position closed; contact fails to operate position and plant operating status: post-LOCA position open on control room on CIAS phase A pressurization fan start signal; a. Computer Redundant valve will open b. Periodic testing c. Position indication at BOP panel Contact of 42X relay Valve fails as is Depending on initial valve Redundant valve will open (Control Room Pressuri- position and plant operating status: zation Fan) fails to operate on CIAS phase A a. Computer b. Periodic testing c. Position indication at BOP panel Loss of power Valve fails as is Depending on initial valve to 600-V motor position and plant operating status: control center 1G a. Computer b. Periodic testing c. Loss of valve position indication light at BOP panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 6 OF 18) Component Identification 3478B (cont.) Logic Diagram Number NA Elementary Number D-177280 Sh. 1 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks Loss of 120 V-ac Valve fails as is Depending on initial valve control power position and plant operating status: a. Computer b. Periodic testing c. Loss of valve position indication lights at BOP panel Thermal overload Valve fails as is Depending on initial valve relay contacts open position and plant operating status: a. Computer b. Periodic testing c. Loss of position indication at BOP panel Failure of (42) Valve fails as is Depending on initial valve starter relay position and plant operating status: to operate a. Computer b. Periodic testing c. Position indication at main control board FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 7 OF 18) Component Identification 3622 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air on receipt of series (3622 and 3623); CIAS phase A only one required to operate a. Valve position indication at BOP panel to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating status: to open on series (3622 and 3623); receipt of only one required to operate a. Valve position indication at BOP panel CIAS phase A to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at main control board control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3622 fails to there are two valves in position and plant operating status: vent (sticky series (3622 and 3623); operator) only one required to operate a. Valve position indication at BOP panel to cause isolation b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 8 OF 18) Component Identification 3623 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air on receipt of series (3623 and 3622); CIAS phase A only one required to operate a. Valve position indication at BOP panel to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating status: to open on series (3623 and 3622); receipt of only one required to operate a. Valve position indication at BOP panel CIAS phase A to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3623 fails to there are two valves in position and plant operating status: vent (sticky series (3623 and 3622); operator) only one required to operate a. Valve position indication at BOP panel to cause isolation b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 9 OF 18) Component Identification 3624 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-606 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air on receipt of series (3624 and 3625); CIAS phase A only one required to operate to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-606 fail there are two valves in position and plant operating status: to open on series (3624 and 3625); receipt of only one required to operate a. Valve position indication at BOP panel CIAS phase A to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3624 fails to there are two valves in position and plant operating status: vent (sticky series (3624 and 3625); operator) only one required to operate a. Valve position indication at BOP panel to cause isolation b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 10 OF 18) Component Identification 3625 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air on receipt of series (3625 and 3624); CIAS phase A only one required to operate a. Valve position indication at BOP panel to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fail there are two valves in position and plant operating status: to open on series (3625 and 3624); receipt of only one required to operate a. Valve position indication at BOP panel CIAS phase A to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3625 fails to there are two valves in position and plant operating status: vent (sticky series (3625 and 3624); operator) only one required to operate a. Valve position indication at BOP panel to cause isolation b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 11 OF 18) Component Identification 3626 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating closed on loss of instrument air on receipt of series (3626 and 3627); status: CIAS phase A only one required to operate to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating status: to open on series (3626 and 3627); receipt of only one required to operate a. Valve position indication at BOP panel CIAS phase A to cause isolation; operator b. Light monitor panel can close valve from main c. Computer control board d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3626 fails to there are two valves in position and plant operating vent (sticky series (3626 and 3627); status: operator) only one required to operate to cause isolation a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 12 OF 18) Component Identification 3627 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; however, Depending on initial valve Valve is normally open and will fail fails to operate there are two valves in position and plant operating status: closed on loss of instrument air on receipt of series (3627 and 3626); CIAS phase A only one required to operate to cause isolation; operator a. Valve position indication at BOP panel can close valve from main control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-613 fails there are two valves in position and plant operating status: to open on series (3627 and 3626); receipt of only one required to operate CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3627 fails to there are two valves in position and plant operating status: vent (sticky series (3627 and 3626); operator) only one required to operate a. Valve position indication at BOP panel to cause isolation b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at instrument air BOP panel b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 13 OF 18) Component Identification 3628 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-605 fails to operate Valve remains open; however, Depending on initial valve Valve is normally open and will fail on receipt of CIAS phase A there are two valves in position and plant operating status: closed on loss of instrument air series (3628 and 3629); only one required to operate to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-605 fails there are two valves in position and plant operating status: to open on series (3628 and 3629); receipt of only one required to operate CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve3628 fails to Valve remains open; however, Depending on initial valve vent (sticky operator) there are two valves in position and plant operating status: series (3628 and 3629); only one required to operate to cause isolation a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at instrument air BOP panel b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 14 OF 18) Component Identification 3629 Logic Diagram Number NA Elementary Number D-177373 Engineering Flow Diagram Numbers: D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-605 fails to operate Valve remains open; however, Depending on initial valve Valve is normally open and will fail on receipt of CIAS phase A there are two valves in position and plant operating status: closed on loss of instrument air series (3629 and 3628); only one required to operate to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Contacts of ESF Valve remains open; however, Depending on initial valve relay K-605 fails there are two valves in position and plant operating status: to open on series (3629 and 3628); receipt of only one required to operate CIAS phase A to cause isolation; operator a. Valve position indication at BOP panel can close valve from main b. Light monitor panel control board c. Computer d. Periodic testing Loss of 125 V-dc control power Valve fails closed a. Valve position light at BOP panel b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3628 fails to there are two valves in position and plant operating vent (sticky series (3629 and 3628); status: operator) only one required to operate to cause isolation a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 15 OF 18) Component Identification 3649A Logic Diagram Number NA Elementary Number D-177883 Engineering Flow Diagram Numbers: D-175012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 fails to operate Valve remains open; operator Depending on initial valve Valve is open only for smoke purge; on receipt of CIAS phase A can close valve from main control board position and plant operating status: will fail closed on loss of instrument air a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valve remains open; operator Depending on initial valve relay K-613 fail to open on can close valve from main control board position and plant operating status: receipt of CIAS phase A a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc control power Valve fails closed a. Valve position light at BOP panel b. Computer c. Light monitor panel Solenoid valve3649A fails to Valve remains open; however, valve is Depending on initial valve vent (sticky operator) normally closed except for smoke purge position and plant operating status: a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of instrument air Valve fails closed a. Valve position light at BOP panel b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 16 OF 18) Component Identification 3649B Logic Diagram Number NA Elementary Number D-177883 Engineering Flow Diagram Numbers: D-175012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 fails to operate Valve remains open; operator Depending on initial valve Valve is open only for smoke purge and on receipt of CIAS phase A can close valve from main control board position and plant operating status: will fail closed on loss of instrument air a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valve remains open; operator Depending on initial valve relay K-613 fail can close valve from main position and plant operating to open on control board status: receipt of CIAS phase A a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at BOP panel control power b. Computer c. Light monitor panel Solenoid valve Valve remains open; however, Depending on initial valve 3649B fails to valve is normally closed position and plant operating vent (sticky except for smoke purging status: operator) a. Valve position indication at BOP panel b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at BOP panel instrument air b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 17 OF 18) Component Identification 3649C Logic Diagram Number NA Elementary Number D-177883 Engineering Flow Diagram Numbers: D-175012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-613 Valve remains open; operator Depending on initial valve Valve is open only for smoke purge and fails to operate can close valve from main control board position and plant operating status: will fail closed on loss of instrument air on receipt of CIAS phase A a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Valve remains open; operator Depending on initial valve relay K-613 fail can close valve from main position and plant operating to open on control board status: receipt of CIAS phase A a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of 125 V-dc Valve fails closed a. Valve position light at main control board control power b. Computer c. Light monitor panel Solenoid valve 3696 fails to Valve remains open; however, the valve is Depending on initial valve vent (sticky operator) normally closed except for smoke purging position and plant operating status: a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of Valve fails closed a. Valve position light at instrument air main control board b. Computer c. Light monitor panel FNP-FSAR-7 REV 21 5/08 TABLE 7.3-8 (SHEET 18 OF 18) Component Identification Control Room Pressurization Fan Logic Diagram Number NA Elementary Number D-177280 Sh. 3 Engineering Flow Diagram Numbers: D-175012 Sh. 1 and D-205012 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-614 Motor fails to start; of two a. Control room indication lights fails to operate pressurization fans, only one b. Periodic testing on CIAS phase A, is required for operation; c. Light monitor panel Units 1 or 2 operator can start motor if necessary Contacts of ESF Motor fails to start; of two a. Control room indication lights relay K-614 fail to pressurization fans, only one b. Periodic testing close on CIAS is required for operation; c. Light monitor panel phase A operator can start motor if necessary Loss of power to Motor fails to start a. Loss of both control room 600-V motor indication lights control center b. Periodic testing Loss of 120 V-ac Motor fails to start a. Loss of both control room control power indication lights b. Periodic testing Thermal overload Motor fails to start a. Control room indication relay contacts open lights b. Periodic testing c. Motor overload trip alarm in the control room FNP-FSAR-7 REV 21 5/08 TABLE 7.3-9 (SHEET 1 OF 4) FAILURE MODE AND EFFECTS ANALYSIS, PENETRATION ROOM FILTRATION SYSTEM Component Identification Exhaust Fan Logic Diagram Number NA Elementary Number D-177238/D-207238 Engineering Flow Diagram Numbers: D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-626 Motor fails to start a. Light monitor panel fails to operate b. Indication lights at on CIAS phase B main control board c. Periodic testing Contacts of ESF Motor fails to start a. Light monitor panel relay K-626 fail to b. Indication lights at close on CIAS main control board phase B c. Periodic testing Loss of power to Motor fails to start a. Light monitor panel 600-V motor b. Indication lights at control center main control board A c. Periodic testing Loss of 120 V-ac Motor fails to start a. Light monitor panel control power b. Indication lights at main control board c. Periodic testing Thermal overload Motor fails to start a. Light monitor panel relay contacts b. Indication lights at open main control board c. Periodic testing d. Motor overload trip alarm at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-9 (SHEET 2 OF 4) Component Identification Recirculation Fan Logic Diagram Number NA Elementary Number D-177239/D-207239 Engineering Flow Diagram Numbers: D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks ESF relay K-626 Motor fails to start a. Light monitor panel fails to operate b. Indication lights at on CIAS phase B main control board c. Periodic testing Contacts of ESF Motor fails to start a. Light monitor panel relay K-626 fail to b. Indication lights at close on CIAS main control board phase B c. Periodic testing Loss of power to Motor fails to start a. Light monitor panel 600-V motor b. Indication lights at control center main control board A c. Periodic testing Loss of 120 V-ac Motor fails to start a. Light monitor panel control power b. Indication lights at main control board c. Periodic testing Thermal overload Motor fails to start a. Light monitor panel relay contacts b. Indication lights at open main control board c. Periodic testing d. Motor overload trip alarm at main control board

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-9 (SHEET 3 OF 4) Component Identification 3362A,B Logic Diagram Number NA Elementary Number D-177281/D-207281 Engineering Flow Diagram Numbers: D-175022 Sh. 1 and D-205022 Sh. 1

Failure Mode Effect on System Detection of Failure Remarks ESF relay K-626 train A Associated valve remains closed Depending on initial valve Valve is normally closed or B fails to operate on receipt position and plant operating of CIAS phase B status: a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Contacts of ESF Associated valve remains closed Depending on initial valve relay K-626 train A or B position and plant operating status fail to close on receipt of  : CIAS phase B a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing Loss of power Associated valve fails a. Loss of valve position light at to 600-V motor control centers as is; the other valve main control board U-3362A and V-3362B will open b. Computer c. Light monitor panel Loss of 120 V-ac Associated valve fails Loss of valve position light control power as is at main control board Thermal overload Associated valve fails Depending on initial valve relay contacts open as is position and plant operating status: a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing FNP-FSAR-7 REV 21 5/08 TABLE 7.3-9 (SHEET 4 OF 4) Component Identification 3362A,B (cont.) Logic Diagram Number NA Elementary Number D-177281/D-207281 Engineering Flow Diagram Numbers: D-175022 Sh. 1 and D-205022 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks Failure of Valve fails as is Depending on initial valve starter relay position and plant operating to operate status: a. Valve position indication at main control board b. Light monitor panel c. Computer d. Periodic testing

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-11 (SHEET 1 OF 5) FAILURE MODE AND EFFECTS ANALYSIS, EMERGENCY SAFEGUARDS PUMP ROOM COOLING SYSTEM Component Identification Containment Spray Pump Room Coolers Logic Diagram Number NA Elementary Number D-177227 Sh. 2, D-207227 Sh. 1 Engineering Flow Diagram Numbers: D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm Two spray pumps provided; each unit motor control at main control board is provided with its own cooler and center A or b. Loss of red/green fan; emergency core cooling system B indication lights at analysis is based upon operation BOP panel of one pump c. Monitor light box abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm control power at main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal Thermal overload Motor stops a. Fan fault alarm contacts open at main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-11 (SHEET 2 OF 5) Component Identification Residual Heat Removal Pump Room Coolers Logic Diagram Number NA Elementary Number D-177227 Sh. 1, D-207227 Sh. 1 Engineering Flow Diagram Numbers: D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm at Two RHR pumps provided; each unit motor control main control board is provided with its own cooler and center A or b. Loss of red/green fan; emergency core cooling system B indication lights at analysis is based upon operation BOP panel of one pump c. Monitor light box abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm at control power main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal Thermal overload Motor stops a. Fan fault alarm at contacts open main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-11 (SHEET 3 OF 5) Component Identification Charging/HHSI Pump Room Cooler Fan Motors Logic Diagram Number NA Elementary Numbers D-177226, D-177229 Sh. 2, D-177284, Engineering Flow Diagram Numbers: D-207226, D-207229, D-207284 D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm at Three charging/HHSI pumps provided; motor control main control board each unit is provided with its own center A or B b. Loss of red/green cooler and fan; emergency core train dedicated pump rooms indication lights at cooling system analysis is based have corresponding train BOP panel upon operation of one pump dedicated room coolers; c. Monitor light box swing pump room cooler can abnormal be aligned to either train A or Train B, depending on pumps alignment Loss of 120 V-ac Motor stops a. Fan fault alarm at control power main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal Thermal overload Motor stops a. Fan fault alarm at contacts open main control board b. Monitor light box abnormal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-11 (SHEET 4 OF 5) Component Identification Component Cooling Water Pump Room Coolers Logic Diagram Number NA Elementary Number D-177243 Sh. 1, D-207243 Sh. 1 Engineering Flow Diagram Numbers: D-175011 Sh. 3, D-205011 Sh. 3

Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm Three component cooling water pumps motor control at main control board provided; there are two coolers for center b. Loss of red/green the three pumps; emergency core indication lights at cooling system analysis is based upon BOP panel operation of one pump c. Monitor light box abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm control power at main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal Thermal overload Motor stops a. Fan fault alarm contacts open at main control board b. Monitor light box abnormal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-11 (SHEET 5 OF 5) Component Identification Motor-Driven Auxiliary Feedwater Pump Room Coolers Logic Diagram Number NA Elementary Number D-177229 Sh. 1, D-207229 Sh. 1 Engineering Flow Diagram Numbers: D-175011 Sh. 3, D-205011 Sh. 3 Failure Mode Effect on System Detection of Failure Remarks Loss of 600-V Motor stops a. Fan fault alarm at Two MD auxiliary feedwater pumps motor control main control board provided; each unit is provided center A or b. Loss of red/green with its own cooler and fan; B indication lights at emergency core cooling system BOP panel analysis is based upon operation c. Monitor light box of one pump abnormal Loss of 120 V-ac Motor stops a. Fan fault alarm at control power main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal Thermal overload Motor stops a. Fan fault alarm at contacts open at main control board b. Loss of red/green indication lights at BOP panel c. Monitor light box abnormal

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-12 FAILURE MODE AND EFFECTS ANALYSIS, BATTERY ROOM VENTILATION SYSTEM Component Identification Battery Room Exhaust Fan Logic Diagram Number NA Elementary Number D-177265 Sh. 1, D-207265 Sh. 1 Engineering Flow Diagram Numbers: D-175014 Sh. 1, D-205014 Sh. 1 Failure Mode Effect on System Detection of Failure Remarks Loss of 208-V Fan stops a. Loss of red/green indication One exhaust fan is provided for motor control lights at BOP Panel each battery room; one battery is center A or B required during post-LOCA operation; effect of loss of exhaust fan on hydrogen accumulation is discussed in paragraph 9.4.2.3.4 Loss of 120 V-ac Fan stops a. Loss of red/green indication control power lights at BOP Panel

Thermal overload Fan stops a. Fan operating light out contacts open

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-13 FAILURE MODE AND EFFECTS ANALYSIS, BATTERY ROOM AIR CONDITIONING SYSTEM Component Identification Battery Room Cooler Logic Diagram Number NA Elementary Number NA Engineering Flow Diagram Number NA

See Table 9.4-6.

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-14 (SHEET 1 OF 3) FAILURE MODE AND EFFECTS ANALYSIS, EMERGENCY DIESEL GENERATOR Component Identification Diesel Generator Supply Breaker Elementary Number D-172761, D-177142, and D-177143

Failure Mode Effect on System Detection of Failure Remarks Loss of 125 V-dc Loss of ability to tie diesel Loss of dc annunciator in Redundant diesel generator will be control power generator to bus when control room started necessary

Failure of 2AJX Loss of ability to tie diesel Incomplete sequence Redundant diesel generator will be contacts to generator to bus when annunciator started close in necessary emergency Failure of Loss of ability to tie diesel Indicating light on control Redundant diesel generator will be 59/81X contacts generator to bus when board started necessary

Mechanical or Loss of ability to tie diesel Breaker position indicating Redundant diesel generator will be electrical generator to bus when lights in control room started failure of necessary breaker

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-14 (SHEET 2 OF 3) Component Identification Diesel Generator Start, Stop, and Shutdown Controls Elementary Numbers D-172774, D-172778, D-172782 Failure Mode Effect on System Detection of Failure Remarks Loss of 125 V-dc Loss of ability to start Annunciator and loss of Redundant diesel generator will be control power diesel generator in emergency indicating lights on board started Failure of start None; redundant starting Testing contact in circuit will start diesel diesel starting circuit A or B to close in emergency Failure of a None; redundant starting Testing relay in circuit will start diesel starting circuit A or B Failure of Loss of ability to stop Diesel running light in Diesel can be stopped manually signal contact diesel from control room control room or relay in diesel stop circuit Failure of Diesel would not shut down Observation of diesel All safety features are cut out contact or when trouble occurred failure except overspeed and low oil relay in pressure during emergency diesel operation shutdown circuit FNP-FSAR-7 REV 21 5/08 TABLE 7.3-14 (SHEET 3 OF 3) Component Identification Diesel Generator Excitation and Miscellaneous Controls Elementary Numbers D-172775, D-172779, D-172783

Failure Mode Effect on System Detection of Failure Remarks Failure of Diesel generator may not Observation of voltage and Redundant generator can be used governor pick up load or may drop frequency on board control load during load fluctuations

Failure of Improper voltage output Observation of meter on Redundant generator can be used excitation from generator board circuit Failure of Improper voltage output Observation of meter on Redundant generator manual voltage auto voltage from generator board control can be used

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 1 OF 6) FAILURE MODE AND EFFECTS ANALYSIS, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Train A Component Malfunction Comment A1 Fail logic zero Prevents manual reset and block of safety injection train A Fail logic one Safety injection from train A not affected if failure occurs before safety injection is called for at M1; safety injection signal will be removed upon reactor trip indicated by P-4, safeguards sequencer must latch in and continue sequence If failure occurs after safety injection is called for at M1, safety injection signal will be reset; no effect provided safeguards sequencer latches in A2 Fail logic zero Prevents reset of safety injection if one of the four inputs to O4 is still calling for safety injection Fail logic one Automatic safety injection actuation will be prevented; manual safety injection is still possible A3 Fail logic zero Automatic safety injection actuation train A will be prevented; manual safety injection train A is still possible Fail logic one Spurious safety injection train A; no direct reactor trip A4 Fail logic zero Prevents high containment pressure safety injection actuation train A if called for A5 Fail logic zero Prevents low steam line pressure; safety injection actuation in train A if called for Fail logic one Spurious safety injection; reactor trip and steam line isolation in train A A6 Fail logic zero Prevents steam line isolation on high steam line flow coincident with low-low Tavg (train A only)

Fail logic one Spurious steam line isolation in train A A7 Fail logic zero Prevents steam line isolation on high steam line flow coincident with low-low Tavg (train A only)

Fail logic one Spurious steam line isolation (train A) if false logic one output of A7 occurs coincident with low-low Tavg FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 2 OF 6) Component Malfunction Comment A8 Fail logic zero Prevents train A safety injection actuation by low steam line pressure Fail logic one Spurious safety injection and steam line isolation (train A) if false logic one output of A8 occurs coincident with low-low Tavg A9 Fail logic zero Prevents low-low Tavg and low steam line pressure safety injection block Fail logic one Prevents low steam line pressure safety injection and steam line isolation actuation in train A if called for A10 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious train A safety injection and reactor trips A11 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-1, P-3) indicated by 2/3 logic at A12 A12 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 1 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-1, P-2) indicated by 2/3 logic at A11 A13 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 2 Fail logic one Spurious safety injection train A and reactor trip A14 Fail logic zero Loss of protection against high steam line differential pressure, i.e., low pressure loop 2 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with low (P-2, P-3) indicated by 2/3 logic at A15 A15 Fail logic zero Loss of protection against high steam line differential pressure, i.e., low pressure loop 2 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-1, P-2) indicated by 2/3 logic at A14 A16 Fail logic zero Partial protection for high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection train A and reactor trip FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 3 OF 6) Component Malfunction Comment A17 Fail logic zero Loss of protection against high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-1, P-3) indicated by 2/3 logic at A18 A18 Fail logic zero Loss of protection against high steam line differential pressure lost, i.e., low pressure loop 3 Fail logic one Spurious safety injection and reactor trip if false output occurs coincident with high (P-2, P-3) indicated by 2/3 logic at A17 A19 Fail logic zero Prevents train A low pressurizer pressure safety injection and reactor trip actuation if called for Fail logic one Spurious safety injection and reactor trip A20 Fail logic zero Prevents train A low pressurizer pressure safety injection and reactor trip Fail logic one Spurious safety injection and reactor trip train A if not blocked A21 Deleted A22 Deleted A23 Fail logic zero Prevents pressurizer safety injection block Fail logic one Prevents train A low pressurizer pressure safety injection and reactor trip actuation if called for A24 Fail logic zero Allows train A safety injection blocks when no block is called for Fail logic one Prevents pressurizer safety injection block

FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 4 OF 6) Component Malfunction Comment A25 Fail logic zero Prevents steam line safety injection block of low steam line pressure; also prevents train A steam line isolation due to high steam line flow coincident with low-low Tavg Fail logic one Allows operator to block safety injection whether or not block should be allowed; if safety injection is not blocked and false output occurs coincident with low steam line pressure, spurious safety injection would result Spurious steam line isolation would occur if high steam line flow (2/3) occurs O1 Fail logic zero Prevents safety injection actuation train A Fail logic one Spurious safety injection; no direct reactor trip O2 Fail logic zero Prevents reset of safety injection if one of the four inputs to O4 is still calling for safety injection Fail logic one Safety injection actuation from train A not affected if failure occurs after safety injection is called for at M1; if failure occurs before safety injection actuation, train A safety injection can be spuriously blocked (if P-4 is also logic one)

O3 Fail logic zero Prevents manual safety injection actuation train A Fail logic one Spurious reactor trip and safety injection train A O4 Fail logic zero Automatic safety injection actuation train A will be prevented; manual safety injection train A is still possible Fail logic one Spurious reactor trip; spurious safety injection train A if safety injection has not been manually blocked O5 Fail logic zero Loss of protection in train A against high steam line flow in loop 1; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 O6 Fail logic zero Loss of protection in train A against high steam line flow in loop 2; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 O7 Fail logic zero Loss of protection in train A against high steam line flow in loop 3; logic at A7 changed to 2/2 Fail logic one Logic at A7 changed to 1/2 FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 5 OF 6) Component Malfunction Comment O8 Deleted not necessary O9 Fail logic zero Prevents safety injection block and allows spurious safety injection and steam line isolation if in coincidence with low steam line pressure Fail logic one Blocks safety injection (train A) if low steam line pressure occurs coincident with low-low Tavg; safety injection is not prevented if low steam line pressure occurs alone (not in coincidence with low-low Tavg)

O10 Fail logic zero Prevents train A high steam line differential pressure safety injection and reactor trip actuation Fail logic one Spurious safety injection and reactor trip O11 Deleted

O12 Fail logic zero Prevents pressurizer safety injection block Fail logic one Blocks pressurizer safety injection if failure occurs coincident with P-11 (-1)

O13 Fail logic zero Prevents safety injection from actuating reactor trip Fail logic one Spurious reactor trip O14 Fail logic zero Prevents steam line isolation actuation of high steam line flow coincident with low-low Tavg and prevents steam line isolation of low steam line pressure when safety injection is called for Fail logic one Spurious steam line isolation actuation N1 Fail logic zero Safety injection will be blocked at A3 although no attempt to reset has taken place Fail logic one Prevents safety injection block at A3 when resetting N2 Fail logic zero Blocks safety injection actuation (train A) of low steam line pressure when no block is called for Fail logic one Fails to block safety injection when block is called for N3 Fail logic zero Prevents steam line safety injection block Fail logic one Prevents manual reset of steam line safety injection block control; allows continuous block FNP-FSAR-7 REV 21 5/08 TABLE 7.3-15 (SHEET 6 OF 6) Component Malfunction Comment N4 Fail logic zero Blocks train A safety injection actuation due to low pressurizer pressure coincident with low pressurizer level when no block is called for Fail logic one Prevents manual block N5 Fail logic zero Prevents operator block of safety injection and reactor trip train A when block should occur Fail logic one Allows operator block of safety injection and reactor trip train A when block should not be allowed N6 Fail logic zero Prevents pressurizer safety injection block Fail logic one Prevents manual reset of pressurizer safety injection block control; allows continuous block TD1 Fail logic zero Prevents manual reset and block Fail logic one Short time delay Allows resetting of safety injection train A before safety injection sequence time delay has been completed Constant output Allows a manual block and reset at any time M1 Fail logic zero Prevents safety injection train A Fail logic one Spurious safety injection train A; prevents reset of safety injection signal Bistable Any one bistable Protection ensured by operation of other bistable inputs in the same system (coincidence changed from 2/3 to 1/2, etc.) inputs to input fails logic solid state one protection Any one bistable Protection ensured by operation of other bistable inputs in the same system (coincidence changed from 2/3 to 2/2, etc.) input to solid state protection fails logic zero

REV 21 5/08 COMPONENT IDENTIFICATION ESFAS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 1 OF 4)

REV 21 5/08 COMPONENT IDENTIFICATION ESFAS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 2 OF 4)

REV 21 5/08 COMPONENT IDENTIFICATION ESFAS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 3 OF 4)

REV 21 5/08 COMPONENT IDENTIFICATION ESFAS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.3-1 (SHEET 4 OF 4)

FNP-FSAR-7 7.4-1 REV 23 5/11 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN The functions necessary for safe shutdown are available from instrumentation channels that are associated with the major systems in both the primary and secondary sides of the nuclear steam supply system. These channels are normally aligned to serve a variety of operational functions, including startup and shutdown as well as protective functions. In achieving a safe shutdown, benefit is taken from many of these systems and equipment having multiple functions, and as such there are no specifically identifiable safe shutdown systems per se. However, prescribed procedures for securing and maintaining the plant in a safe condition can be instituted by appropriate alignment of selected nuclear steam supply systems. The discussion of these systems and the applicable codes, criteria, and guidelines are to be found in other sections of the report. In addition, the alignment of shutdown functions associated with the engineered safety features, which are invoked under postulated limiting fault situations, is discussed in chapter 6 and section 7.3.

The instrumentation and control functions which are identified as being required for maintaining safe shutdown of the reactor are by definition the minimum required under nonaccident conditions. (Control room inaccessibility as well as offsite power interruptions during a hot shutdown are considered as incidents.)

These functions will permit the necessary operations that will:

A. Prevent the reactor from achieving criticality in violation of the plant technical specifications.

B. Provide an adequate heat sink such that design and safety limits are not exceeded.

7.4.1 DESCRIPTION The designation of systems that can be used for safe shutdown depends on identifying those systems which provide the following capabilities for maintaining a safe shutdown:

A. Boration with related charging and letdown.

B. Adequate supply for auxiliary feedwater.

C. Residual heat removal.

These systems and the associated instrumentation and controls provisions are identified in the following lists. The identification of the monitoring indicators (paragraph 7.4.1.1) and controls (paragraph 7.4.1.2) are those necessary for maintaining a hot shutdown. The essential services for the capabilities necessary for maintaining a hot shutdown are listed in paragraph 7.4.1.3, with the equipment and services available for a cold shutdown identified in paragraphs 7.4.1.4 and 7.4.1.5.

See subsection 7.1.4 and Table 7.1-1 for a list of supplemental drawings.

FNP-FSAR-7 7.4-2 REV 23 5/11 Periodic testing of remote shutdown system instrumentation and controls is conducted in accordance with the Technical Specifications.

7.4.1.1 Monitoring Indicators The characteristics of these indicators, which are provided outside as well as inside the control room, are described in Section 7.5. The necessary indicators are as follows:

A. Water level indicator for each steam generator.

B. Pressure indicator for each steam generator.

C. Pressurizer water level indicator.

D. Pressurizer pressure indicator.

E. Reactor coolant loop 1 hot leg temperature.

F. Reactor coolant loop 1 cold leg temperature.

G. Neutron flux.

H. Condensate storage tank level.

7.4.1.2 Controls 7.4.1.2.1 General Considerations A. The turbine is tripped. (Note that this can be accomplished at the turbine as well as in the control room.)

B. The reactor is tripped. (Note that this can be accomplished at the reactor trip switchgear as well as in the control room.)

C. All automatic systems continue functioning. (These are discussed in Sections 7.2 and 7.7.)

D. For equipment having motor controls outside the control room (which duplicate the functions inside the control room), the controls will be provided with a selector switch which transfers control of the switchgear from the control room to a selected local station. Placing the local selector switch in the local operating position will give an annunciating alarm in the control room and will turn off the motor control position lights on the control room panel.

FNP-FSAR-7 7.4-3 REV 23 5/11 7.4.1.2.2 Pumps A. Auxiliary Feedwater Pumps

Auxiliary feedwater pumps (electric) will start automatically on the loss of both main feedwater pumps. Start/stop motor controls positioned locally (and inside the control room) as well as handwheel control for the valves are provided. It is noted below that emergency power is available from the diesels which can be started locally and that the loads such as valves and pumps will be sequenced as necessary.

B. Charging and Boric Acid Transfer Pumps

Start/stop motor controls are provided for these pumps. The controls for the charging and boric acid pumps are positioned locally (and in the control room).

C. Service Water Pumps

These pumps, by means of the onsite power system, will start automatically following a loss of normal electrical power. Start/stop motor controls located outside and inside the control room will be provided.

D. Component Cooling Water Pumps

These pumps, energized from the diesel generator, start automatically following a loss of normal electrical power. Start/stop controls located outside and inside the control room are provided.

E. Instrument Air Compressors

These compressors start automatically on low air pressure.

F. Reactor Containment Fan Cooler Units

Start/stop motor controls with a selector switch are provided for the fan motors. The controls are located outside and inside the control room.

G. Control Room Ventilation Unit Including the Control Room Air Inlet Dampers

A start/stop switch located outside the control room is provided for this unit(s). Also a control to close the inlet air damper(s) is provided. These controls duplicate functions inside the control room.

7.4.1.2.3 Diesels These units start automatically following a loss of normal ac power. However, manual controls for diesel startup are also provided locally at the diesel generators (as well as in the control room), and loading is sequenced automatically.

FNP-FSAR-7 7.4-4 REV 23 5/11 7.4.1.2.4 Valves A. Charging Flow Control Valves

Manual control with a selector switch outside the control room is provided for the charging line flow control valves. This control duplicates functions available in the control room.

B. Letdown Orifice Isolation Valves

Open/close controls with a selector switch for the letdown orifice isolation valves are grouped with the controls for the charging flow control valve. These controls duplicate functions that are inside the control room.

C. Auxiliary Feedwater Control Valves

Manual control is provided in the auxiliary feedwater pump area that duplicates functions that are inside the control room. A handwheel is also provided for each valve.

D. Atmospheric Steam Relief Valves

Atmospheric relief valves are automatically controlled. Manual control is provided locally and inside the control room for the atmospheric relief valves. A handwheel is also provided for each valve.

E. Auxiliary Feedwater Pump Speed Control

Manual speed control (mechanical device) is provided locally and in the control room for the steam-driven auxiliary feedwater pump.

F. Pressurizer Heater Control

On/off control with selector switches are provided for two backup heater groups. The heater group will be connected to separate buses, such that each can be connected to separate diesels in the event of loss of outside power.

The control is grouped with the charging flow controls and duplicates functions available in the control room.

It is noted that the instrumentation and controls listed in subsections 7.4.1.1 and 7.4.1.2 for achieving and maintaining a safe shutdown are available in the event an evacuation of the control room is required. Cable routing of key instrumentation loops will allow the plant to be brought to hot standby from the hot shutdown panel with the loss of either the cable spreading room, control room, or a cable chase. These controls and instrumentation channels, with the equipment and services identified in subsections 7.4.1.3 and 7.4.1.4 which are available for both hot and cold shutdown, identify the potential capability for cold shutdown of the reactor subsequent to a control room evacuation through the use of suitable procedures. Therefore, the applicable requirements of 1971 General Design Criterion 19 are met.

FNP-FSAR-7 7.4-5 REV 23 5/11 7.4.1.3 Essential Services after Incident That Requires Hot Shutdown A. Auxiliary feedwater pumps which start automatically within 1 min for blackout condition. (See chapter 10.)

B. Reactor containment air recirculation fans and coolers. (See chapter 6.)

C. Diesel generators, loaded within 1 min. (See chapter 8.)

D. Lighting in the areas of plant required during this condition. (See subsection 9.5.3.)

E. Pressurizer heaters. (See chapter 5.)

F. Communication network (see subsection 9.5.2) to be available for prompt use between feedwater pumps area and the following areas:

1. Feedwater source from outside.

2. Charging pump.

3. Boric acid transfer pump.

4. Diesel generator.

5. Switchgear room.

6. Steam relief valves.

G. Boric acid transfer pumps. (See chapter 9.)

H. Charging pumps. (See chapter 9.)

I. Service water pumps. (See chapter 9.)

J. Component cooling pumps. (See chapter 9.)

K. Instrument air compressors. (See chapter 9.)

L. Control room ventilation unit and air inlet damper. (See chapter 9.)

FNP-FSAR-7 7.4-6 REV 23 5/11 7.4.1.4 Equipment and Systems Available for Cold Shutdown A. Reactor coolant pump. (See chapter 5.)(a)

B. Auxiliary feedwater pumps. (See chapter 10.)

C. Boric acid transfer pump. (See chapter 9.)

D. Charging pumps. (See chapter 9.)

E. Service water pumps. (See chapter 9.)

F. Containment fans. (See chapter 6.)

G. Control room ventilation. (See chapter 9.)

H. Component cooling pumps. (See chapter 9.)

I. Residual heat removal pumps. (See chapter 5.)(a)

J. Motor control center and switchgear sections associated with above loads.

K. Controlled steam release and feedwater supply. (See section 7.7 and chapter 10.)

L. Boration capability. (See chapter 9.)

M. Nuclear instrumentation system (source range and intermediate range). (See sections 7.2 and 7.7.)(a)

N. Reactor coolant inventory control (charging and letdown). (See chapter 9.)

O. Pressurizer pressure control including opening control for pressurizer relief valves (heaters and spray). (See chapter 5.)(a)

In addition, the safety injection signal trip circuit must be defeated and the accumulator isolation valves closed.(a) The performance of the emergency core cooling system under these conditions was evaluated. Conditions during plant cooldown were divided into the following four

a. Instrumentation and controls for these systems may require some modification in order that their functions may be performed from outside the control room. Note that the reactor plant design does not preclude attaining the cold shutdown condition from outside the control room.

An assessment of plant conditions can be made on a long term basis (a week or more) to establish procedures for making the necessary physical modifications to instrumentation and control equipment in order to attain cold shutdown. During such time the plant could be safely maintained at hot shutdown condition.

Detailed procedures to be followed in effecting cold shutdown from outside the control room are best determined by plant personnel at the time of the postulated incident.

FNP-FSAR-7 7.4-7 REV 23 5/11 phases: (1) from operating reactor coolant pressure to 1900 psig, (2) from 1990 to 1000 psig, (3) from 1000 to 400 psig, and (4) from 400 psig to cold shutdown. The break size used in the analysis was determined using the moderate energy line break criteria identified in Branch Technical Positions APCSB 3-1 and MEB 3-1. Based on the analysis, the available emergency core cooling system can cool the core under plant cooldown conditions and, therefore, meets the NRC acceptance criteria, as applicable, contained in 10 CFR 50.46 and 10 CFR 50, Appendix K.

7.4.2 ANALYSIS Hot shutdown is a stable plant condition, automatically reached following a plant shutdown. The hot shutdown condition can be maintained safely for an extended period of time either automatically or manually. In the unlikely event that access to the control room is restricted, the plant can be safely kept at a hot shutdown until the control room can be reentered by the use of the monitoring indicators and the controls listed in paragraphs 7.4.1.1 and 7.4.1.2. These indicators and controls are provided outside and inside the control room. The safety evaluation of the maintenance of a shutdown with these systems and associated instrumentation and controls has included consideration of the accident consequences that might jeopardize safe shutdown conditions. The accident consequences that are germane are those that would tend to degrade the capabilities for boration, adequate supply for auxiliary feedwater, and residual heat removal.

The results of the accident analyses are presented in chapter 15. Of these the following produce the most severe consequences that are pertinent:

A. Uncontrolled boron dilution.

B. Loss of normal feedwater.

C. Loss of external electrical load and/or turbine trip.

D. Loss of all ac power to the station auxiliaries (station blackout).

It is shown by these analyses that safety is not compromised by these incidents, with the associated assumptions being that the instrumentation and controls indicated in paragraphs 7.4.1.1 and 7.4.1.2 are available to control and/or monitor shutdown. These available systems will allow a maintenance of hot shutdown even under the accident conditions listed above which would tend toward a return to criticality or a loss of heat sink.

FNP-FSAR-7 7.5-1 REV 21 5/08 7.5 POSTACCIDENT MONITORING DISPLAY INSTRUMENTATION 7.5.1 DESCRIPTION Table 7.5-1 lists the instrumentation provided to the operator to perform necessary functions, assess plant conditions, and verify system performance during accident situations. Listed below are the five classifications of variables that have been identified to provide this instrumentation.

Type A: Those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events.

Type B: Those variables to be monitored that provide to the control room operator information to assess the process of accomplishing or maintaining critical safety functions.

Type C: Those variables to be monitored that provide to the control room operator information to monitor (1) the extent to which parameters, which have the potential for causing a breach of the primary reactor containment or fuel cladding, have exceeded the design basis values, or (2) that the in-core fuel clad, the reactor coolant system pressure boundary or the primary reactor containment may have been breached.

Type D: Those variables that provide information to indicate the operation of individual safety systems and other systems important to safety.

Type E: Those variables that are to be monitored as required for use in determining the magnitude of the release of radioactive materials and in continuously assessing such releases.

These variables are subdivided into three categories which define the qualification requirements of the instrumentation. Table 7.5-1 identifies the variable category identified in the R. G. 1.97 Compliance Report. The qualification and configuration requirements and the Farley Evaluation Criteria for specific R. G. 1.97 requirements are described in the Design and Qualification Review Criteria section of the R. G. 1.97 Compliance Report.

The instrumentation channels that provide the information for the variables listed in Table 7.5-1, are powered as described in the R. G. 1.97 Compliance Report, and are energized from the onsite electrical power supplies as described in chapter 8.

Table 7.5-3 lists the information available to the operator for monitoring conditions in the reactor, in the reactor coolant system, and in the containment and process systems throughout all normal operating conditions of the plant, including anticipated operational occurrences.

Post-accident monitoring instrumentation is discussed in the Technical Specifications.

FNP-FSAR-7 7.5-2 REV 21 5/08 Containment hydrogen monitoring instrumentation surveillance is discussed in FSAR section 16.1.

7.5.2 ANALYSIS With the issuance of Regulatory Guide 1.97, Alabama Power Company performed a comprehensive review and issued a R. G. 1.97 Compliance Report documenting Farley's commitment to R. G. 1.97.

The display instrumentation for postaccident monitoring enables the required manual functions to be performed following a Condition II, III, or IV event to provide the necessary information to maintain the plant at a safe hot shutdown or to proceed to a cold shutdown condition consistent with the technical specification limits. Other design criteria used in the display system are given at the end of this section.

All commitments concerning recording, separation, qualification, and redundancy are provided in the R. G. 1.97 Compliance Report.

For postaccident scenarios (see table 7.5-1), sufficient duplication of information is provided to ensure that the minimum information required will be available. The information is part of the operational monitoring of the plant which is under surveillance by the operator during normal plant operation. This is functionally arranged on the control board to provide the operator with ready understanding and interpretation of plant conditions. Comparisons between duplicate information channels or between functionally related channels will enable the operator to readily identify a malfunction in a particular channel.

It is noted that there is a degree of functional redundancy between those display channels that are required for postaccident monitoring and many other diverse instrumentation channels which are also located on the main control board. For example, after the actuation of safety injection, the residual heat removal pump flow, high head (charging) pump flow, and spray pump flow can be verified by their respective flow channels. The transmitters for these flow channels are outside the containment. In addition, the containment sump level is continuously read out on the main control board. This information provides a diverse means for checking refueling water storage tank level data obtained from the safety-related display information.

Channel separation is provided between sensors and the process cabinets. From the process cabinets to the main control board, the interconnecting circuits meet the separation requirements between safety trains, with two channels being associated with one train.

The design criteria used in the display system are listed below:

A. Range and accuracy requirements are determined through the analyses of postaccident conditions as described in chapter 15. The display system meets the following requirement: the range of the readouts extends over the maximum expected range of the variable being measured.

B. Power for the display instruments is obtained from the instrumentation and control power supply system. This system is described in section 7.6 and complies with FNP-FSAR-7 7.5-3 REV 21 5/08 paragraph 5.4 of the Institute of Electrical and Electronics Engineers standard 308.

C. Those channels determined to provide useful information in charting the course of events are recorded.

7.5.3 DELETED 7.5.4 INADEQUATE CORE COOLING MONITORING SYSTEM The inadequate core cooling monitoring system (ICCMS) is a safety grade processing and display system which meets the NRC requirements to provide the capability to monitor the approach to, existence of, and recovery from potential reactor core inadequate core cooling situations. The requirements addressed by the ICCMS are defined in paragraph II.f.2 of NUREG 0737, "Clarification of TMI Action Plan" and Generic Letter 82-28. Inadequate core cooling monitoring requirements are met by measuring and displaying margin to saturation, reactor vessel water level above the core, and core exit temperatures.

7.5.4.1 Reactor Vessel Level The redundant heated junction thermocouple (HJTC) probes are described in paragraph 4.4.5.5. Redundant processors are located in the control room. Redundant level indication is provided on a reactor vessel mimic display on the main control board. The mimic indicates covered or uncovered for each of the eight heated junctions for each probe. The mimic background shows the elevation of each sensor and its location in the reactor vessel in relation to major components and penetrations.

The redundant signal processors, one per HJTC probe, monitor the HJTC probe thermocouples, control power to the HJTC probe heaters, and drive the level displays. The processor HJTC calculations are as follows:

The differential temperature (T) is calculated from the temperature values for the unheated junction (Tu) and the heated junction (TH) thermocouple inputs. T is equal to TH minus Tu, and that T is compared against a low T setpoint (25°F). If T is less than the low T setpoint the corresponding error number is set. A low T error indicates that there is a loss of heater power or a heater controller malfunction. There are two heater controllers per channel. Each heater controller is connected to four heaters in series. If one heater fails open then all the rest of the heaters will be turned off. This will cause either all the odd numbered Ts or even numbered Ts to have a low T error. T or Tu is used to determine percent level for both the head area and the plenum area. A sensor is considered uncovered whenever T or Tu is greater than 200°F or 700°F, respectively. Five-degree dead bands exist in both the T and Tu setpoints for uncovered sensors to prevent cycling.

FNP-FSAR-7 7.5-4 REV 21 5/08 A maximum TH and a maximum T are selected and are used to calculate separate setpoint signals for the heater controllers. The minimum of the two heater controller setpoint signals is selected and sent to each of the heater controllers. The TH and T heater controller setpoint signals are reduced at a constant rate, when their respective TH and T values increase above a predetermined value. The TH and T heater controller setpoints will decrease until they equal zero at a second predetermined setpoint.

7.5.4.2 Subcooling Margin Monitor The subcooling margin monitor (SMM) provides continuous, redundant indication of the margin to saturated conditions in the reactor coolant system (RCS). The SMM inputs are RCS hot leg and cold leg temperatures from loop RTDs, core exit thermocouple temperature, RCS wide range pressure, and pressurizer pressure. The margin to saturation, displayed in degrees F, is the difference between the measured RCS temperature and the saturation temperature. The highest RCS loop temperature and the highest core exit thermocouple temperature, excluding upper head thermocouples, are used to calculate margins to saturation. The lowest pressure value is used to calculate the saturation temperature. The control board SMM display has a switch to select margin to saturation indication based on RCS loop RTD temperature or core exit thermocouple temperature.

7.5.4.3 Core Exit Temperature Core exit temperature is continuously indicated on redundant control board displays. The chromel-alumel thermocouples in the vessel measure temperature at the flow exit of selected fuel assemblies and locations within the reactor vessel head plenum.

The redundant displays each normally indicate the temperature of the hottest thermocouple for that channel. The operator can interrogate the display to indicate the temperature of any individual thermocouple or the highest temperature in each core quadrant.

7.5.5 NUCLEAR INSTRUMENTATION In addition to the Westinghouse nuclear instrumentation system that is described in section 7.2 and whose indications are listed in table 7.5-3, an independent channel of Gamma-Metrics nuclear instrumentation is provided to satisfy alternate shutdown requirements.

The Gamma-Metrics channel provides neutron flux indication at the hot shutdown panel and the control room via isolated outputs. A fission chamber detector measures neutron flux from shutdown to full power. Detector sensitivity is 10-2 to 1010 nv. The following displays are provided on the main control board and the hot shutdown panel:

FNP-FSAR-7 7.5-5 REV 21 5/08 Source Range 0.1 to 105 counts/s Source Range Startup Rate-1 to 7 decades/min. Log Power Level 10-8 to 100-percent power FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 1 OF 16) POST ACCIDENT INSTRUMENTATION TYPE A VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Plant Specific 1 Information required for operator action 1 RCS Pressure (wide range) 1 2 RCS Hot Leg Temperature (wide range) 1 3 RCS Cold Leg Temperature (wide range) 1 4 Steam Generator Level (wide range) 1 5 Steam Generator Level (narrow range) 1 6 Pressurizer Level 1 7 Containment Pressure (normal range) 1 8 Main Steam Line Pressure 1 9 Refueling Water Storage Tank Level 1 10 Containment Water Level 1 11 Condensate Storage Tank Level 1 12 Auxiliary Feedwater Flow 1 15 Core Exit Temperature 1 132 Core Subcooling Monitor 2

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 2 OF 16) TYPE B VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Reactivity Control Neutron Flux 1 Function detection; accomplishment of mitigation 17 Neutron Flux (intermediate range) 1 Control Rod Position 3 Verification 1009 Control Rod Position 3 RCS Soluble Boron Concentration 3 Verification 1017 Post Accident Sample 3 RCS Cold Leg Water Temperature 3 Verification 3 RCS Cold Leg Temperature (wide range) 1 Core Cooling RCS Hot Leg Water Temperature 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 2 RCS Hot Leg Temperature (wide range) 1 RCS Cold Leg Water Temperature 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 3 RCS Cold Leg Temperature (wide range) 1

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 3 OF 16) TYPE B VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY RCS Pressure 1 Function detection; accomplishment of mitigation; verification; long-term surveillance 1 RCS Pressure (wide range) 1 Core Exit Temperature 3 Verification 15 Core Exit Temperature 1 Coolant Inventory 1 Verification; accomplishment of mitigation; 18 Reactor Water Level 1 Degrees of Subcooling 2 Verification and analysis of plant conditions 132 Core Subcooling Monitor 2 Maintaining Reactor Coolant System Integrity RCS Pressure 1 Function detection; accomplishment of mitigation 1 RCS Pressure (wide range) 1 Containment Sump Water Level (narrow range) 2 Function detection; accomplishment of mitigation; verification 111 Reactor Cavity Sump Level 2 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 4 OF 16) TYPE B VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Containment Sump Water Level (wide range) 1 Function detection; accomplishment of mitigation; verification 10 Containment Water Level 1 Containment Pressure 1 Function detection; accomplishment of mitigation; verification 7 Containment Pressure (normal range) 1 Maintaining Containment Integrity Containment Isolation Valve Position (excluding check valves) 1 Accomplishment of isolation 19 Containment Isolation Valve Position 1 Containment Pressure 1 Function detection; accomplishment of mitigation; verification 7 Containment Pressure (normal range) 1

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 5 OF 16) TYPE C VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Fuel Cladding Core Exit Temperature 1 Detection of potential for breach; accomplishment of mitigation; long-term surveillance 15 Core Exit Temperature 1 Radioactivity Concentration or Radiation Level in Circulating Primary Coolant 1 Detection of breach 14 Primary Coolant Radioactivity Concentration 1 Analysis of Primary Coolant (gamma spectrum) 3 Detail analysis; accomplishment of mitigation; verification; long-term surveillance 1017 Post Accident Sample 3 Reactor Coolant Pressure Boundary RCS Pressure 1 Detection of potential for or actual breach; accomplishment of mitigation; long-term surveillance 1 RCS Pressure (wide range) 1

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 6 OF 16) TYPE C VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Containment Pressure 1 Detection of breach; accomplishment of mitigation; verification; long-term surveillance 7 Containment Pressure (normal range) 1 Containment Sump Water Level (narrow range) 2 Detection of breach; accomplishment of mitigation; verification; long-term surveillance 111 Reactor Cavity Sump Level 2 Containment Sump Water Level (wide range) 1 Detection of breach; accomplishment of mitigation; verification; long-term surveillance 10 Containment Water Level 1 Containment Area Radiation 3 Detection of breach; verification 13 Containment Radiation (high range) 1 Effluent Radioactivity -

Noble Gas Effluent from Condenser Air Removal System Exhaust 3 Detection of breach; verification 120 Condenser SJAE Radiation 2

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 7 OF 16) TYPE C VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Containment RCS Pressure 1 Detection of potential for breach; accomplishment of mitigation 1 RCS Pressure (wide range) 1 Containment Hydrogen Concentration 1 Detection of potential for breach; accomplishment of mitigation; long-term surveillance 1006 Containment Hydrogen Concentration 3 Containment Pressure 1 Detection of potential for or an actual breach; accomplishment of mitigation 16 Containment Pressure (extended range) 1 Containment Effluent Radioactivity - Noble Gases from Identified Release Points 2 Detection of breach; accomplishment of mitigation; verification 121 Plant Vent Effluent Radiation 2 Effluent Radioactivity -

Noble Gases (from buildings or areas where penetrations and hatches are located, e. g.,

secondary containment and auxiliary buildings and fuel handling buildings that are in direct contact with primary containment) 2 Indication of breach 121 Plant Vent Effluent Radiation 2 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 8 OF 16) TYPE D VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Residual Heat Removal (RHR) or Decay Heat Removal System RHR System Flow 2 To monitor operation 101 RHR/LHSI Flow 2 RHR Heat Exchanger Outlet Temperature 2 To monitor operation and for analysis 114 RHR HX Discharge Temperature 2 Safety Injection Systems Accumulator Tank Level and Pressure 2 To monitor operation 125 1018 Accumulator Tank Pressure Accumulator Tank Level 2 3 Accumulator Isolation Valve Position 2 Operation status 126 Accumulator Tank Isolation Valve Position2 Boric Acid Charging Flow 2 To monitor operation 102 Boric Acid Flow 2 Flow in HPI System 2 To monitor operation 103 HHSI Flow 2 Flow in LPI System 2 To monitor operation 101 RHR/LHSI Flow 2 Refueling Water Storage Tank Level 2 To monitor operation 9 Refueling Water Storage Tank Level 1 Primary Coolant System Reactor Coolant Pump Status 3 To monitor operation 1011 RCP Motor Current 3 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 9 OF 16) TYPE D VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Primary System Safety Relief Valve Positions (including PORV and code valves) or Flow Through or Pressure in Relief Valve Lines 2 Operation status; to monitor for loss of coolant 127 128 Pressurizer PORV Position Pressurizer Safety Valve Position 2 2 Pressurizer Level 1 To ensure proper operation of pressurizer 6 Pressurizer Level 1 Pressurizer Heater Status 2 To determine operating status 130 112 Pressurizer Heater Breaker Position Pressurizer Pressure 2 2 Quench Tank Level 3 To monitor operation 1002 Pressurizer Relief Tank level 3 Quench Tank Temperature 3 To monitor operation 1004 Pressurizer Relief Tank Temperature 3 Quench Tank Pressure 3 To monitor operation 1007 Pressurizer Relief Tank Pressure 3 Secondary System (Steam Generator) Steam Generator Level 1 To monitor operation 4 Steam Generator Level (wide range) 1 Steam Generator Pressure 2 To monitor operation 8 Main Steam Line Pressure 1 Safety/Relief Valve Positions or Main Steam Flow 2 To monitor operation 104 Main Steam Flow 2 Main Feedwater Flow 3 To monitor operation 1001 Main Feedwater Flow 3 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 10 OF 16) TYPE D VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Auxiliary Feedwater or Emergency Feedwater System Auxiliary or Emergency Feedwater Flow 2 To monitor operation 12 Auxiliary Feedwater Flow 1 Condensate Storage Tank Water Level 1 To ensure water supply for auxiliary feedwater (can be Category 3 if not primary source of AFW. Then whatever is primary source of AFW should be listed and should be Category 1) 11 Condensate Storage Tank Level 1 Containment Cooling Systems Containment Spray Flow 2 To monitor operation 105 Containment Spray Flow 2 Heat Removal by the Containment Fan Heat Removal System 2 To monitor operation 115 116 133 Temperature of Service Water to Aux. Bldg

CTMT Cooler Service Water Outlet Temperature

Service Water Flow to CTMT Coolers 2 2 2 Containment Atmosphere Temperature 2 To indicate accomplishment of cooling 117 Containment Atmosphere Temperature 2 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 11 OF 16) TYPE D VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Containment Sump Water Temperature 2 To monitor operation 118 RHR HX Inlet Temperature 2 Chemical and Volume Control System Makeup Flow - In 2 To monitor operation 106 110 Charging Line Flow RCP Seal Injection Flow 2 2 Letdown Flow- Out 2 To monitor operation 107 Letdown Flow 2 Volume Control Tank Level 2 To monitor operation 113 Volume Control Tank Level 2 Cooling Water System Component Cooling Water Temperature to ESF System 2 To monitor operation 119 Component Cooling Water Heat Exchanger Discharge Temperature 2 Component Cooling Water Flow to ESF Sys 2 To monitor operation 108 CCW HX Inlet Flow 2 Radwaste Systems High-level Radioactive Liquid Tank Level 3 To indicate storage volume 1003 Radioactive Liquid Tank Levels 3 Radioactive Gas Holdup Tank Pressure 3 To indicate storage capacity 1008 Waste gas Decay Tank Pressure 3 Ventilation Systems Emergency Ventilation Damper Position 2 To indicate damper status 129 HVAC Emergency Damper Position 2 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 12 OF 16) TYPE D VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Power Supplies Status of Standby Power and Other Energy Sources Important to Safety (electric, hydraulic, pneumatic) (voltages, currents, pressures) 2 To indicate system status 131 Emergency Power Status 2

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 13 OF 16) TYPE E VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO.

DESCRIPTION CATEGORY Containment Radiation Containment Area Radiation - High Range 1 Detection of significant releases; release assessment; long-term surveillance; emergency plan actuation 13 Containment Radiation (high range) 1 Area Radiation Radiation Exposure rate (inside buildings or areas where access is required to service equipment important to safety) 3 Detection of significant releases; release assessment; long-term surveillance122 1005 Accessible Area Radiation Portable Plant/Environs Radiation 2 3 Airborne Radioactive Materials Released from Plant Noble Gases and Vent Flow Rate Containment or Purge Effluent 2 Detection of significant releases; release assessment Not Applicable, see Common Plant vent Reactor Shield Building Annulus (if in design) 2 Detection of significant releases; release assessment Not Applicable, not in design

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 14 OF 16) TYPE E VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Auxiliary Building (including any building containing primary system gases, e.

g., waste gas decay tank) 2 Detection of significant releases; release assessment; long-term surveillance Not Applicable, see Common Plant Vent Condenser Air Removal System Exhaust 2 Detection of significant releases; release assessment 120 Condenser SJAE Radiation 2 Common Plant Vent or Multi-purpose Vent Discharging Any of Above Releases (if containment purge is included) 2 Detection of significant release; release assessment; long-term surveillance121 109 Plant Vent Effluent Radiation Plant Vent Stack Flow 2 2 Vent From Steam Generator Safety Relief Valves or Atmospheric Dump Valves 2 Detection of significant releases; release assessment 104 123 124 Main Steam Flow Main Steam Effluent Radiation TDAFW Effluent Radiation 2 2

2 All Other Identified Release Points 2 Detection of significant releases; release assessment; long-term surveillance Not Applicable

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 15 OF 16) TYPE E VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Particulates and Halogens All Identified Plant Release Points (except steam generator safety relief valves or atmospheric steam dump valves and condenser air removal system exhaust). Sampling with Onsite Analysis Capability 3 Detection of significant releases; release assessment; long-term surveillance 1012 Particulates and Halogens Sampling (Vent Stack) 3 Environs Radiation and Radioactivity Airborne Radiohalogens and Particulates (portable sampling with onsite analysis capability) 3 Release assessment; analysis 1013 Airborne Radiohalogens and Particulates (Environs) 3 Plant and Environs Radiation (portable instrumentation) 3 Release assessment; analysis 1005 Portable Plant/Environs Radiation 3 Plant and Environs Radioactivity (portable instrumentation) 3 Release assessment; analysis 1019 Portable Plant/Environs Radioactivity (Gamma-ray Spectrometer) 3 Meteorology Wind Direction 3 Release assessment 1014 Wind Direction 3 Wind Speed 3 Release assessment 1015 Wind Speed 3 Estimation of Atmospheric Stability 3 Release assessment 1016 Estimation of Atmospheric Stability 3 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-1 (SHEET 16 OF 16) TYPE E VARIABLES R.G. 1.97 VARIABLES FNP POSITION VARIABLE CATEGORY PURPOSE VARIABLE NO. DESCRIPTION CATEGORY Accident Sampling Capability (Analysis Capability On Site) Primary Coolant and Sump -Gross Activity

-Gamma Spectrum

-Boron Content

-Chloride Content

-Dissolved Hydrogen or Total Gas -Dissolved Oxygen

-pH 3 Release assessment; verification; analysis 1017 Post Accident Sample 3 Containment Air -Hydrogen Content

-Oxygen Content

-Gamma Spectrum 3 Release assessment; verification; analysis 1010 Post Accident Sample - CTMT Air 3 FNP-FSAR-7 REV 21 5/08 TABLE 7.5-2

(This table has been deleted.)

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 1 OF 6) CONTROL ROOM INDICATORS AND/OR RECORDERS AVAILABLE TO THE OPERATOR TO MONITOR SIGNIFICANT PLANT PARAMETERS DURING NORMAL OPERATION No. of Channels Indicated Indicator/ Parameter Available Range Accuracy Recorder Location Notes Notes Nuclear Instrumentation Source range Count rate 2 1 to 106 +/-7% of the linear Both channels indicated; Control board One 2-pen record- counts/s full scale analog either may be selected er is used to re- voltage for recording cord any of the 8 nuclear chan- nels (2 source range, 2 intermediate range, and 4 power range). Startup rate 2 -0.5 to 5.0 +/-7% of the linear Both channels indicated Control board decades/min full scale analog voltage Intermediate range Flux level 2 8 decades of +/-7% of the linear Both channels indicated; neutron flux full scale analog either may be selected (corresponds to voltage and +/-3% for recording 0 to full scale of the linear full analog voltage) scale voltage in overlapping the the range of 10-4 source range by to 10-3 A 2 decades Startup rate 2 -0.5 to 5.0 +/-7% of the linear Both channels indicated Control board decades/min full scale analog voltage

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 2 OF 6) No. of Channels Indicated Indicator/ Parameter Available Range Accuracy Recorder Location Notes Power range Uncalibrated ion 4 0 to 120% of +/-1% of full span All 8 current signals NIS racks in chamber current full power indicated control room (top and bottom current uncompensated ion chambers) Calibrated ion 4 0 to 120% of +/-2% of full power These 8 current signals are chamber current full power available for selectable (top and bottom trending by the operator uncompensated ion chambers) Upper and lower ion 4 -50 to +50% +/-3% of full power Diagonally opposed; Control board chamber current any 2 of the 4 difference channels may be se- lected for recording at the same time using intermediate range recorder Average flux of the 4 0 to 120% of +/-3% of full power All 4 channels Control board top and bottom ion full power for indication; indicated; any 2 of the chamber +/-2% for recording 4 channels may be recorded using source range recorder Average flux of the 4 0 to 200% of +/-2 to 120% of All 4 channels Control board top and bottom ion full power full power; recorded chambers +/-6 to 200% of full power

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 3 OF 6) No. of Channels Indicated Indicator/ Parameter Available Range Accuracy Recorder Location Notes Flux difference of 4 -30 to +30% +/-4% All 4 channels Control board the top and bottom indicated ion chambers Reactor Coolant System Tavg (measured) 1 per 530° to 630°F +/-4°F All channels indicated Control board loop T (measured) 1 per 0 to 150% of +/-4% of full All channels indicated Control board loop full power T power T Tcold or 1 Thot, 0°F to 700°F +/-4% Both channels recorded Control board Thot (measured, 1 Tcold wide range) per loop Overpower T 1 per 0 to 150% of +/-4% of full All channels indicated Control board setpoint loop full power T power T Overtemperature 1 per 0 to 150% of +/-4% of full All channels indicated Control board T setpoint loop full power T power T Pressurizer 5 1700 to +/-20 psi All channels indicated Control board pressure 2500 psig Pressurizer 3 Entire distance +/-3.5% of P All channels indicated; Control board 2-pen recorder level between taps at 2250 psia 1 channel is selected used; second for recording pen records reference level signal Primary coolant 3 per 0 to 120% of Repeatability of All channels indicated Control board flow loop rated flow +/-4% of full flow Reactor coolant 1 per 0 to 1200 amps - All channels indicated Control board One channel for pump bus amperes loop each bus System pressure 2 0 to 3000 psig +/-4% All channels indicated Control board wide range and recorded FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 4 OF 6) No. of Channels Indicated Indicated Parameter Available Range Accuracy Recorder Location Notes Reactor Control System Demanded rod speed 1 0 to 75 +/-2% 1 channel is indicated Control board steps/min Median Tavg 1 530°F to 630°F +/-4°F 1 channel is recorded Control board Tref 1 530°F to 630°F +/-4°F 1 channel is recorded Control board Control rod position If system not available, borate and sample ac- cordingly Number of steps 1 per 0 to 231 steps(a) +/-1 step Each group is indi- Control board These signals are of demanded rod group cated used in conjunc- withdrawal tion with the full-length rod measured position signals to detect deviation of any individual rod from the demanded position; a deviation will actuate an alarm Full-length rod 1 per 0 to 228 steps(b) +/-4 steps at full Each rod position Control board measured position rod accuracy; +/-8 is indicated steps at 1/2 accuracy

a. Fully withdrawn position can be varied from 225 to 231 steps to reduce RCCA wear. The NRC acceptance criteria regarding the range associated with the fully withdrawn RCCA position are that the fully withdrawn position selected for use throughout each cycle will be evaluated as part of the reload safety evaluation process to verify that sufficient margin exists in the safety analyses to bound the related effects. b. Digital Rod Position Indication (DRPI) system maximum indication is 228 steps.

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 5 OF 6) No. of Channels Indicated Indicated Parameter Available Range Accuracy Recorder Location Notes Control rod bank 4 0 to 100%(a) +/-3% of total All 4 control rod Control board 1 channel for position bank travel bank positions are each control recorded along with rod; an alarm and the low-low limit annunciator are alarm for each actuated when the bank rod control bank to be withdrawn reaches withdrawal limit, when any rod control bank reaches the low insertion limit, and when any rod control bank reaches the low- low insertion limit Containment System Containment pressure 4 -5 to 65 psig +/-3% All 4 channels Control board indicated Feedwater and Steam Systems Auxiliary feedwater 1 per 0 to 400 +/-3% All channels Control board 1 channel to measure flow - Unit 1 steam gal/min the flow to each line steam generator Auxiliary feedwater 1 per 0 to 400 +/-3% All channels Control board 1 channel to measure flow - Unit 2 steam gal/min the flow to each line steam generator Steam generator 3 per +6.2 to -11.5 ft from +/-3% of level All channels indicated; Control board level (narrow range) steam nominal full load (hot) channels used for generator level control are recorded Steam generator 1 per +6.2 to -41.7 ft from +/-5% of level All channels recorded Control board level (wide range) steam nominal full load (cold) generator level a. One-hundred percent is the fully withdrawn position.

FNP-FSAR-7 REV 21 5/08 TABLE 7.5-3 (SHEET 6 OF 6) No. of Channels Indicated Indicated Parameter Available Range Accuracy Recorder Location Notes Main feedwater 2 per 0 to 120% of +/-5% All channels indicated; Control board flow steam maximum calcu- channels used for generator lated flow control are recorded Magnitude of 1 per 0 to 100% of +/-1.5% All channels indicated Control board 1 channel for signal control- main, valve opening each main feed- ing main 1 per water control bypass valve;open/shut indication is provided in control room for each main feedwater control valve Steam flow 2 per 0 to 120% of +/-5.5% All channels indicated; Control board Accuracy is steam maximum cal- channels used for equipment generator culated flow control are recorded capability; however, abso- lute accuracy depends on applicant cali- bration against feedwater flow Steam line 3 per 0 to 1200 psig +/-4% All channels indicated Control board pressure loop Steam dump 1 0 to 85% max- +/-1.5% 1 channel is indicated Control board Open/shut indi- modulate mum calculated cation is pro- signal steam flow vided in the control room for each steam dump valve Turbine impulse 2 0 to 120% of +/-3.5% Both channels indicated Control board Open/shut indi- chamber pressure maximum calcu- cation is pro- lated turbine vided in the load control room for each turbine stop valve

REV 21 5/08 LOGIC DIAGRAM FOR RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-1

REV 21 5/08 LOGIC DIAGRAM FOR RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-2 REV 21 5/08 LOGIC DIAGRAM FOR BACKUP TO SEMIAUTOMATIC SWITCHOVER LOGIC FROM INJECTION TO RECIRCULATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-3 REV 21 5/08 FUNCTIONAL BLOCK DIAGRAM OF ACCUMULATOR ISOLATION VALVE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.6-4

FNP-FSAR-7 REV 21 5/08 TABLE 7.7-1 (SHEET 1 OF 2) PLANT CONTROL SYSTEM INTERLOCKS Designation Derivation Function C-1 1/2 neutron flux Blocks automatic and (intermediate range) manual control rod above setpoint withdrawal C-2 1/4 neutron flux Blocks automatic and (power range) above manual control rod setpoint withdrawal C-3 2/3 overtemperature Blocks automatic and T above setpoint manual control rod withdrawal C-4 2/3 overpower T above Blocks automatic and setpoint manual control rod withdrawal C-5 1/1 turbine impulse Blocks automatic control chamber pressure below rod withdrawal setpoint C-7 1/1 time derivative Makes steam dump valves (absolute value) of available for either turbine impulse chamber tripping or modulation pressure (decrease only) above setpoint

FNP-FSAR-7 REV 21 5/08 TABLE 7.7-1 (SHEET 2 OF 2)

Designation Derivation Function C-9 Any condenser pressure Blocks steam dump to above setpoint condenser or All circulation water pump breakers open C-11 1/1 bank D control rod Blocks automatic rod position above setpoint withdrawal C-20 Two-of-two turbine impulse Arms AMSAC; below chamber pressure above setpoint, blocks AMSAC setpoint (Generated in AMSAC; see section 7.8.) Control grade only. P-4(a) Reactor Trip Closes main feedwater valves on low Tavg below setpoint Blocks steam dump control via load rejection Tavg controller Makes steam dump valves available for either tripping or modulation Reactor not tripped Block steam dump control via plant trip Tavg controller

(a) See table 7.3-4 for safety functions.

FNP-FSAR-7 REV 21 5/08 TABLE 7.7-2 BORON CONCENTRATION MEASUREMENT SYSTEM SPECIFICATIONS Operating Conditions Line voltage: 120 V-AC (+/-10 percent); 60 Hz (+/-1 percent)

Pressure: 15 to 225 psig (sample)

Temperature: 70°F to 130°F (sample)

Sample flowrate: 0 to 0.4 gal/min

Ambient temperature: 60°F to 105°F

Relative humidity: to 95 percent

Radiation levels: <2 mr/h at 24 in. from all tank surfaces Reading time: Variable depending on boron concentration; maximum time for 5000 ppm is approximately 5 min Accuracy Accuracy Boron ppm of Water Standard Deviation 0 - 1800 ppm +/-10 ppm 1800 - 5000 ppm +/-1.25 percent Drift: Less than 10 ppm/week REV 21 5/08 SIMPLIFIED BLOCK DIAGRAM OF REACTOR CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-1 REV 21 5/08 CONTROL BANK ROD INSERTION MONITOR JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-2

REV 21 5/08 ROD DEVIATION COMPARATOR JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-3 REV 21 5/08 BLOCK DIAGRAM OF PRESSURIZER PRESSURE CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-4

REV 21 5/08 BLOCK DIAGRAM OF PRESSURIZER LEVEL CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-5

REV 21 5/08 BLOCK DIAGRAM OF MAIN FEEDWATER PUMP SPEED CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-6 REV 21 5/08 BLOCK DIAGRAM OF STEAM GENERATOR WATER LEVEL CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-7

REV 21 5/08 BLOCK DIAGRAM OF STEAM DUMP CONTROL SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-8 REV 21 5/08 BASIC FLUX MAPPING SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-9 REV 21 5/08 SOURCE-DETECTOR ASSEMBLY JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-10 REV 21 5/08 MEASUREMENT UNIT JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-11

REV 21 5/08 PROCESS SCHEMATIC FOR THE BORON CONCENTRATION MEASUREMENT SYSTEM JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-12 REV 21 5/08 BORON CONCENTRATION MEASUREMENT SYSTEM VS NORMAL PLANT OPERATING RANGE OF BORON CONCENTRATIONS JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.7-13 FNP-FSAR-7 7.8-1 REV 21 5/08 7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) 7.8.1 DESCRIPTION

7.8.1.1 System Description The ATWS (anticipated transient without scram) mitigation system actuation circuitry (AMSAC) provides a backup to the reactor trip system (RTS) and ESF actuation system (ESFAS) for initiating turbine trip and auxiliary feedwater flow in the event of an anticipated transient (e.g., in the complete loss of main feedwater). The AMSAC is independent of and diverse from the RTS and ESFAS, with the exception of the final actuation devices. The AMSAC equipment, with the exception of the output isolation relays, is classified as control-grade equipment. It is a highly-reliable, microprocessor-based, single-train system powered by a non-Class 1E source.

The AMSAC continuously monitors level in the steam generators, which is an anticipatory indication of a loss of heat sink, and initiates certain functions when the level drops below a predetermined setpoint for at least a preselected time and for two of the three steam generator levels. These initiated functions are the tripping of the turbine, the initiation of auxiliary feedwater, and isolation of the steam generator blowdown and sample lines.

The AMSAC is designed to be highly reliable, resistant to inadvertent actuation, and easily maintained. Reliability is assured through the use of internal redundancy and continual self testing by the system. Inadvertent actuations are minimized through the use of internal redundancy and majority voting at the output stage of the system. The time delay on low steam generator level and the coincidence logic used also minimize inadvertent actuations.

The AMSAC automatically performs its actuations when above a preselected power level (determined using turbine impulse chamber pressure) and remains armed sufficiently long after that pressure drops below the setpoint to ensure that its function will be performed in the event of a turbine trip.

7.8.1.2 Equipment Description The AMSAC consists of a single train of equipment located primarily in a seismically qualified cabinet. The output isolation relays, however, are located in two separate qualified wall-mounted cabinets.

The design of the AMSAC is based on the industry standard Intel multibus format, which permits the use of various readily available, widely used microprocessor cards on a common data bus for various functions.

The AMSAC consists of the following:

FNP-FSAR-7 7.8-2 REV 21 5/08 A. Steam Generator (SG) Level Sensing AMSAC utilizes the SG level signals as measured with three differential pressure-type level transmitters, measuring the level for each of the main steam generators as shown on drawing U-166237.

B. Turbine Impulse Pressure AMSAC also utilizes the turbine impulse pressure signal for measuring pressure in the turbine, as shown on drawing U-166245.

C. System Hardware The system hardware consists of two primary systems: the actuation logic system (ALS) and the test/ maintenance system (T/MS).

1. Actuation Logic System

The ALS monitors the analog and digital inputs, performs the functional logic required, provides actuation outputs to trip the turbine and initiate auxiliary feedwater flow, and provides status information to the T/MS. The ALS consists of three groups of input/output (I/O) modules, three actuation logic processors (ALPs), two majority voting modules, and two output relay panels. The I/O modules provide signal conditioning, isolation, and test features for interfacing the ALS and the T/MS.

Conditioned signals are sent to three identical ALPs for analog-to-digital conversion, setpoint comparison, and coincidence logic performance.

Each of the ALPs perform identical logic calculations using the same inputs, and derive component actuation demands, which are then sent to the majority voting modules. The majority voting modules perform a two-out-of-three vote on the ALP demand signals. These modules drive the relays providing outputs to the existing turbine trip and auxiliary feedwater initiation circuits.

A simplified block diagram of the AMSAC ALS architecture is presented on figure 7.8-1.

2. Test/Maintenance System

The test/maintenance system provides the AMSAC with automated and manual testing as well as a maintenance mode. Automated testing is the continuously performed self checking done by the system during normal operation. ALS status is monitored by the T/MS and sent to the plant computer and the main control board. Manual testing of the system by the computer services staff can be performed on-line to provide assurance that the ALS system is fully operational. The maintenance mode permits the computer services staff, under administrative control, to modify channel setpoints, channel status, and timer values and to initiate channel calibration.

FNP-FSAR-7 7.8-3 REV 21 5/08 The T/MS consists of a test/maintenance processor, a digital-to-analog conversion board, a memory board, expansion boards, a self-health board, digital output modules, a test/maintenance panel, and a portable terminal/printer.

D. Equipment Actuation

The output relay panels provide component actuation signals through isolation relays, which then drive the final actuation circuitry as shown on drawings U-166244 and U-166245 for initiation of auxiliary feedwater and for turbine trip.

7.8.1.3 Functional Performance Requirements Analyses have shown that the most limiting ATWS event is a loss of feedwater event without a reactor trip. AMSAC performs the mitigative actuations of automatically initiating auxiliary feedwater, tripping the turbine, and isolating steam generator blowdown and sampling lines.

These are initiated in order to ensure a secondary heat sink following an anticipated transient (ANS Condition II) without a reactor trip, in order to limit core damage following an anticipated transient without a reactor trip and to ensure that the energy generated in the core is compatible with the design limits to protect the reactor coolant pressure boundary by maintaining the reactor coolant pressure to within ASME stress level C.

7.8.1.4 AMSAC Interlocks A single interlock, designated as C-20, is provided to allow for the automatic arming and blocking of the AMSAC (drawing U-166245). The system is blocked at sufficiently low reactor power levels when the actions taken by the AMSAC following an ATWS need not be automatically initiated. Turbine impulse chamber pressure in a two-out-of-two logic scheme is used for the blocking function. Turbine impulse chamber pressure above the setpoint will automatically defeat any block, i.e., will arm the AMSAC. Dropping below this setpoint will automatically block the AMSAC. Removal of the C-20 permissive is automatically delayed for a predetermined time. The operating status of the AMSAC is displayed on the main control board.

7.8.1.5 Trip System The SG level and turbine impulse chamber pressure inputs are used by AMSAC to determine trip demand. Signal conditioning is performed on the transmitter output and used by each of the ALPs to derive a component actuation demand. If two of the three steam generators have a low level at a power level greater than the C-20 permissive, a trip demand signal is generated following a time delay. This signal drives output relays for performing the necessary mitigative actions.

FNP-FSAR-7 7.8-4 REV 21 5/08 7.8.1.6 Isolation Devices AMSAC is independent of the RTS and ESFAS. The AMSAC inputs for measuring narrow range steam generator water level are derived from existing transmitters and channels within the process protection system. Connections to these channels are made downstream of Class-1E isolation devices located within the process protection cabinets. These isolation devices ensure that the existing protection system continues to meet all applicable safety criteria by providing isolation. Buffering of the AMSAC outputs from the safety-related final actuation device circuits is achieved through qualified relays. A credible fault occurring in the nonsafety-related AMSAC will not propagate through and degrade the RTS and ESFAS.

7.8.1.7 AMSAC Diversity from the Reactor Protection Systems Equipment diverse from the RTS and ESFAS is used in the AMSAC to prevent common mode failures that might affect the AMSAC and the RTS or ESFAS. The AMSAC is a digital, microprocessor-based system with the exception of the analog SG level and turbine impulse pressure transmitter inputs, whereas the reactor trip system utilizes an analog based protection system. Also, where similar components are utilized for the same function in both AMSAC and the reactor trip system, the components used in AMSAC are provided by a different manufacturer.

Common mode failure of identical components in the analog portion of the RTS that results in the inability to generate a reactor trip signal will not impact the ability of the digital AMSAC to generate the necessary mitigative actuations. Similarly, a postulated common mode failure affecting analog components in ESFAS, affecting its ability to initiate auxiliary feedwater, will not impact the ability of the digital based AMSAC to automatically initiate auxiliary feedwater.

7.8.1.8 Power Supply The AMSAC power supply is a dedicated uninterruptible power supply (UPS) which is independent from the RTS power supplies and is backed by batteries which are independent from the existing batteries which supply the RTS.

7.8.1.9 Environmental Variations The AMSAC equipment is located in a controlled environment such that variations in the ambient conditions are minimized.

7.8.1.10 Setpoints The AMSAC makes use of two setpoints in the coincidence logic in order to determine if mitigative functions are required. Water level in each steam generator is sensed to determine if a loss of secondary heat sink is imminent. The low-level setpoint is selected in such a manner that a true lowering of the level will be detected by the system. The normal small variations in steam generator level will not result in a spurious AMSAC signal.

FNP-FSAR-7 7.8-5 REV 21 5/08 The C-20 permissive setpoint is selected in order to be consistent with ATWS investigations showing that the mitigative actions performed by the AMSAC need not be automatically actuated below a certain power level. The maximum allowable value of the C-20 permissive setpoint is defined by these investigations.

To avoid inadvertent AMSAC actuation on the loss of one main feedwater pump, AMSAC actuation is delayed by a defined amount of time. This will ensure the reactor protection system will provide the first trip signal.

To ensure that the AMSAC remains armed sufficiently long to permit its function in the event of a turbine trip, the C-20 permissive is maintained for a preset time delay after the turbine impulse chamber pressure drops below the setpoint. The setpoints and the capability for their modification in the AMSAC are under administrative control.

7.8.2 ANALYSIS 7.8.2.1 Safety Classification/Safety-Related Interface The AMSAC is not safety related, therefore, it need not meet the requirements of IEEE-279-1971. The AMSAC has been implemented such that the RTS and ESFAS continue to meet all applicable safety-related criteria. The AMSAC is independent of the RTS and ESFAS. The isolation provided between the RTS and the AMSAC and between the ESFAS and the AMSAC by the isolator modules and the isolation relays, respectively, ensures that applicable safety- related criteria are met for the RTS and the ESFAS.

7.8.2.2 Redundancy System redundancy has not been provided. Since AMSAC is a backup nonsafety-related system to the redundant RTS, redundancy is not required. To ensure high system reliability, portions of the AMSAC have been implemented as internally redundant, such that a single failure of an input channel or ALP will neither actuate nor prevent actuation of the AMSAC.

7.8.2.3 Diversity from the Existing Trip System Diverse equipment has been selected in order that common cause failures affecting both the RTS and the AMSAC or both the ESFAS and the AMSAC will not render these systems inoperable simultaneously. A more detailed discussion of the diversity between the RTS and the AMSAC and between the ESFAS and the AMSAC is presented in paragraph 7.8.1.7.

7.8.2.4 Electrical Independence The AMSAC is electrically independent of the RTS and ESFAS with the exception of the final actuation devices. Qualified isolation devices are provided to isolate the nonsafety AMSAC FNP-FSAR-7 7.8-6 REV 21 5/08 circuitry from the safety-related actuation circuits of the auxiliary feedwater system as discussed in paragraph 7.8.1.6.

7.8.2.5 Physical Separation from the RTS and ESFAS AMSAC is, by necessity, physically separated from the existing protection system hardware.

The two trains of AMSAC outputs are provided from separate wall-mounted enclosures outside of the cabinet.

7.8.2.6 Environmental Qualification Equipment related to the AMSAC is designed to operate under conditions resulting from anticipated operational occurrences for the respective equipment location. The AMSAC equipment, with the exception of the isolation devices, is not designated as safety- related equipment and, therefore, is not required to be qualified as safety related per the requirements of IEEE Standard 279-1971, "IEEE Standard for Criteria for Protection Systems for Nuclear Power Generating Stations." The safety-related AMSAC output isolation devices are located in a mild environment.

7.8.2.7 Seismic Qualification It is required that only the isolation devices comply with seismic qualification. The AMSAC output isolation device is qualified in accordance with a program that was developed to implement the requirements of IEEE Standard 344-1975, "IEEE Standard for Seismic Qualification of Class 1E Electrical Equipment for Nuclear Power Generating Stations."

7.8.2.8 Test, Maintenance, and Surveillance Quality Assurance NRC Generic Letter 85-06, "Quality Assurance Guidance for ATWS Equipment that is not Safety Related," requires quality assurance procedures commensurate with the nonsafety-related classification of the AMSAC. The quality controls for the AMSAC are, at a minimum, consistent with existing plant procedures or practices for nonsafety-related equipment.

Design of the AMSAC followed procedures relating to equipment procurement, document control, and specification of system components, materials and services. In addition, specifications also define quality assurance practices for inspections, examinations, storage, shipping, and tests as appropriate to a specific item or service.

A computer software verification program and a firmware validation program have been implemented commensurate with the nonsafety-related classification of the AMSAC to ensure that the system design requirements implemented with the use of software have been properly implemented and to ensure compliance with the system functional, performance, and interface requirements.

FNP-FSAR-7 7.8-7 REV 21 5/08 System testing is completed prior to the installation and operation of the AMSAC as part of normal factory acceptance testing and the validation program. Periodic testing is performed automatically through use of the system automatic self-checking capability and manually under administrative control via the AMSAC test/maintenance panel.

7.8.2.9 Power Supply Power to the AMSAC is from a battery-backed, dedicated uninterruptible power supply independent of the power supplies for the RTS and ESFAS. The station battery supplying power to the AMSAC is independent of those used for the RTS and ESFAS. The AMSAC is an energize-to-actuate system capable of performing its mitigative functions with a loss-of-offsite power.

7.8.2.10 Testability at Power The AMSAC is testable at power. This testing is done via the system test/maintenance panel. The capability of the AMSAC to perform its mitigative actuations is bypassed at a system level while in the test mode. Total system testing is performed as a set of three sequential, partial, overlapping tests. The first of the tests checks the analog input portions of the AMSAC in order to verify accuracy. Each of the analog input modules is checked separately. The second test checks each of the ALPs separately to verify that the appropriate coincidence logic is sent to the majority voter. The last test exercises the majority voter and the integrity of the associated output relays. The majority voter and associated output relays are tested by exercising all possible input combinations to the majority voter. The integrity of each of the output relays is checked by confirming continuity of the relay coils without operating the relays. The capability to individually operate the output relays, confirm integrity of the associated field wiring, and operate the corresponding isolation relays and final actuation devices at plant shutdown is provided.

7.8.2.11 Inadvertent Actuation The AMSAC has been designed such that the frequency of inadvertent actuations is minimized.

This high reliability is ensured through use of three redundant ALPs and a majority voting module. A single failure in any of these modules will not result in a spurious AMSAC actuation.

In addition, a two-out-of-three low-steam generator level coincidence logic and a time delay have been selected to further minimize the potential for inadvertent actuations.

7.8.2.12 Bypass 7.8.2.12.1 Maintenance Bypasses The AMSAC is blocked at the system level during maintenance, repair, calibration, or test.

While the system is blocked, the bypass condition is indicated in the main control room.

FNP-FSAR-7 7.8-8 REV 21 5/08 7.8.2.12.2 Operating Bypasses The AMSAC has been designed to allow for operational bypasses with the inclusion of the C-20 permissive. Above the C-20 setpoint, the AMSAC is automatically unblocked (i.e., armed); below the setpoint, the system is automatically blocked. The operating status of the AMSAC is indicated in the main control room via a bypass and permissive panel window.

7.8.2.12.3 Indication of Bypasses Whenever the mitigative capabilities of the AMSAC are bypassed or deliberately rendered inoperable, this condition is indicated in the main control room. In addition to the operating bypass, any manual maintenance bypass is indicated via the AMSAC general warning sent to the main control room.

7.8.2.12.4 Means for Bypassing A permanently installed system bypass selector switch is provided to bypass the system. This is a two-position selector switch with "NORMAL" and "BYPASS" positions. At no time is it necessary to use any temporary means, such as installing jumpers or pulling fuses, to bypass the system.

7.8.2.13 Completion of Mitigative Actions Once Initiated The AMSAC mitigative actions go to completion as long as the coincidence logic is satisfied and the time delay requirements are met. If the flow in the feedwater lines is reinitiated before the timer expires and the SG water level increases to above the AMSAC low setpoint, the coincidence logic will no longer be satisfied and the actuation signal disappears. If the coincidence logic conditions are maintained for the duration of the time delay, the mitigative actions go to completion. The auxiliary feedwater initiation and the turbine trip signals are latched in at the activated component level through the existing circuits. Deliberate operator action is then necessary to terminate auxiliary feedwater flow, clear the turbine trip signal using the main control board turbine trip reset switch, and proceed with the reopening of the turbine stop valves.

7.8.2.14 Manual Initiation Manual initiation of the AMSAC is not provided. The capability to initiate the AMSAC mitigative functions manually, i.e., initiate auxiliary feedwater, trip the turbine, and isolate steam generator blowdown and sampling lines, exists at the main control board independent of AMSAC.

7.8.2.15 Information Readout The AMSAC has been designed such that the operating and maintenance staffs have accurate, complete, and timely information pertinent to the status of the AMSAC. A system level general FNP-FSAR-7 7.8-9 REV 21 5/08 warning alarm is indicated in the control room. Diagnostic capability exists from the test/maintenance panel to determine the cause of any unanticipated inoperability or deviation.

7.8.2.16 Compliance with Standards and Design Criteria The AMSAC meets the NRC acceptance criteria contained in 10 CFR 50.62 and the quality assurance requirements contained in NRC Generic Letter 85-06. The AMSAC also complies with the generic designs presented in WCAP-10858-P-A, which have been determined to be acceptable by the NRC for meeting the requirements of 10 CFR 50.62. In addition, the time delay design for the AMSAC associated with the C-20 permissive signal is consistent with Revision 1 to WCAP-10858-P-A, which has been accepted by the NRC.

REV 21 5/08 ACTUATION LOGIC SYSTEM ARCHITECTURE JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 FIGURE 7.8-1

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