ML17117A370
Text
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 Contai nment 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 Emergenc y 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, P O = 0 PSIG 6.2-2 RSG DEPSG Minimum ESF 1 AC Pressure vs. Time, P O = 3 PSIG
6.2-3 DEHL, Minimum ESF, DBA Short Term Pressure vs. Time, P O = 0 PSIG
6.2-4 RSG DEHLG Minimum ESF, DBA Short Term Pressure vs. Time, P O = +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, P O = +3 PSIG
6.2-6A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, P O = -1.5 PSIG
6.2-7 RSG Temperature vs. Time Steam Line Full D. E. Break 102% Power, P O = +3 PSIG 6.2-7A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, P O = -1.5 PSIG
6.2-8 Pressure vs. Time Steam Line 0.7 ft 2 D. E. Break 102% Power
6.2-9 Temperature vs. Time Steam Line 0.7 ft 2 D. E. Break 102% Power
6.2-10 Pressure vs. Time Steam Line 0.6 ft 2 D. E. Break 102% Power, P O = 0 PSIG
6.2-10A Pressure vs. Time Steam Line 0.6 ft 2 D.E. Break 102% Power, P O = -1.5 PSIG
6.2-11 Temperature vs. Time Steam Line 0.6 ft 2 D. E. Break 102% Power, P O = 0 PSIG
6.2-11A Temperature vs. Time Steam Line 0.6 ft 2 D.E. Break 102% Power, P O = -1.5 PSIG
6.2-12 Pressure vs. Time Steam Line 0.528 ft 2 Split 102% Power
6.2-13 Temperature vs. Time Steam Line 0.528 ft 2 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 ft 2 D. E. Break 70% Power
6.2-17 Temperature vs. Time Steam Line 0.6 ft 2 D. E. Break 70% Power
6.2-18 Pressure vs. Time Steam Line 0.5 ft 2 D. E. Break 70% Power
6.2-19 Temperature vs. Time Steam Line 0.5 ft 2 D. E. Break 70% Power
6.2-20 RSG Pressure vs. Time Steam Line 0.47 ft 2 Split 70% Power, P O = +3 PSIG
6.2-21 RSG Temperature vs. Time Steam Line 0.47 ft 2 Split 70% Power, P O = -1.5 PSIG
6.2-22 RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, P O = -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, P O = +3 PSIG 6.2-23 RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power,
P O = -1.5 PSIG
6.2-23A RSG Temperature vs. Time Steam Line Full D. E. Break 30% Power, P O = +3 PSIG
6.2-24 Pressure vs. Time Steam Line 0.5 ft 2 D. E. Break 30% Power
6.2-25 Temperature vs. Time Steam Line 0.5 ft 2 D. E. Break 30% Power
6.2-26 Pressure vs. Time Steam Line 0.4 ft 2 D. E. Break 30% Power, P O = 0 PSIG
6.2.26A Pressure vs. Time Steam Line 0.4 ft 2 D. E. Break 30% Power, P O = -1.5 PSIG
6.2-27 Temperature vs. Time Steam Line 0.4 ft 2 D. E. Break 30% Power, P O = 0 PSIG
6.2-27A Temperature vs. Time Steam Line 0.4 ft 2 D. E. Break 30% Power, P O = -1.5 PSIG
6.2-28 RSG Pressure vs. Time Steam Line 0. 60 ft 2 Split 30% Power, P O = -1.5 PSIG
6.2-28A RSG Pressure vs. Time Steam Line 0.60 ft 2 Split 30% Power, P O = +3 PSIG
6.2-29 RSG Temperature vs. Time Steam Line 0.60 ft 2 Split 30% Power, P O = -1.5 PSIG
6.2-29A RSG Temperature vs. Time Steam Line 0.60 ft 2 Split 30% Power, P O = +3 PSIG
6.2-30 RSG Pressure vs. Time Steam Line Full D. E. Break Hot Standby, P O = +3 PSIG
6.2-31 RSG Temperature vs. Time Steam Line Full D. E. Break Hot Standby,
P O = -1.5 PSIG
6.2-32 Pressure vs. Time Steam Line 0.2 ft 2 D. E. Break Hot Standby
6.2-33 Temperature vs. Time Steam Line 0.2 ft 2 D. E. Break Hot Standby
6.2-34 Pressure vs. Time Steam Line 0.1 ft 2 D. E. Break Hot Standby
6.2-35 Temperature vs. Time Steam Line 0.1 ft 2 D. E. Break Hot Standby
6.2-36 Pressure vs. Time Steam Line 0.30 ft 2 Split Hot Standby
6.2-37 Temperature vs. Time Steam Line 0.30 ft 2 Split Hot Standby
6.2-38 TS, Equipment Surface Temperature with Uchida Condensing Heat Transfer and Convective Heat Transfer Coefficient of 2 Btu/h/ft 2 (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, P O = 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, P O = 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, P O = +3 PSIG
6.2-79 RHR Heat Exchanger Duty vs. Time, RSG P O = +3 PSIG
6.2-80 Containment Air Cooler Duty vs. Time, RSG P O = +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 contai nment 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 isolati on 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 offsit e 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 featur es 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, vac uum 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 adj acent 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 constructi on, 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 spa ce limitation does not allow. In this test
a vacuum box containing a window is placed o ver 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 th e 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 c ontrol. The primary purpose of the liner plate and the welds therein is to provide leaktight ness 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 complete ly effective method for examining welds to assure leaktightness. Therefore, the maxi mum 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 w ith 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 approxim ately 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 tw o 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 mi nimum 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 loca tions 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 enti re weld represented by the three radiographs was acceptable. The defective welding disclo sed 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 por tion 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 Magne tic-Particle Inspection
Dye-penetrant and magnetic-par ticle inspection were used as an aid to quality control.
The field welding inspectors used dye-pene trant or magnetic-par ticle inspection to closely examine welds judged to be of questionable quality on the basis of the initial
visual inspection. Dye-penetrant or magne tic-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 venti ng 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 specific ally 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 in to components for a system and ensure that they meet the requirements of procurement specification. The design conditions, which form the basis for these component qualificati on 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 w ith 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, ope ning and closing periods, and the ability of the valve operator to actuate the valve at th e maximum anticipated differential pressure.
Test results on actual or similar types of filter assem blies demonstrate their adequacy for this application.
The following demonstrative tests are performed:
A. Radioactive iodine removal efficiency - a c harcoal 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, sim ilar to those predicte d to occur in the penetration room filters duri ng 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 ft 3/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 ft 3/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/m 3 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 impre gnate. 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 - Deen ergized and energized tests demonstrate the proper mounting of components, proper hookup of circuits and connecti on, setting of instrumentation and operation of interlocks. Equipment and system perf ormance 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 a re checked to ensure that suitabl e gasket compression is uniformly achieved on the filter sealing faces.
A test program is performed after construction t ests 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 ma intained in the spent fuel area with the penetration room filtration system operating in the fuel handling area.
The following tests are performed prior to installati on 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 diff erential pressure and observe syst em performance in the fuel handling mode.
I. With the system operating in the fuel handli ng mode, verify that there is a vacuum in the spent fuel pool.
In addition, all instruments are calibrated, alar ms, controls, and interlocks checked, and each remotely operated valve is individually stroked to determin e 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 va cuum gauge in the penetration room. With all normally operating ventilation system s in the auxiliary building secured, the internal pressure in the penetration rooms and the exhaust ai r 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 (ft
- 3) 2.35 x 10 6 Net free internal volume (ft
- 3) 2.0 x 10 6 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 Exposure
- 1. Containment Atmosphere
- 2. Insulated
Material Thickness (in.)
Paint/Primer 0.0084
Carbon Steel 3.59 Heat Sink 9 - Containment Floor 5,402 ft 2 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 ft 2 Exposure
- 1. Containment Atmosphere
- 2. Insulated
Material Thickness (in.)
Stainless Steel 0.12
Heat Sink 12 - Galvanized Steel Cable Trays 22,164 ft 2 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 ft 2 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 (ft
- 3) 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/ft 3) 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)
Sump liquid to containment atmosphere Conduction
Containment sump and floor to sump Conduction liquid Sink surfaces exposed to outside 2.0 Btu/h-ft 2-°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 ft 3/min per unit Total design heat 80 x 10 6 80 x 10 6(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 ft 2 per unit Overall heat transfer 383 383 coefficient, Btu/h-ft 2-°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 10 6 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 P 0 = 0 psig P 0 = +3 psig 4.95 ft 2 3 ft 2 8.25 ft 2 P 0 = +3 psig Peak pressure 38.0 43.8 40.1 40.9 37.6 43.6 (p s i g) 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 (ft
- 3) per accumulator 1040 N 2 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 ft 3 generator and the nearest steam line stop valve Maximum steam line volume between the faulted 3475 ft 3 steam generator stop valves and the steam line stop valves in the other steam generator loops Maximum unisolated feed line volume 202 ft 3 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 10 3 5.56 x 10 2 3.04 x 10-3 1.09 x 10 4 5.55 x 10 2 5.04 x 10-3 1.33 x 10 4 5.54 x 10 2 7.04 x 10-3 1.47 x 10 4 5.52 x 10 2 9.08 x 10-3 1.69 x 10 4 5.51 x 10 2 1.01 x 10-2 1.72 x 10 4 5.51 x 10 2 1.10 x 10-2 1.71 x 10 4 5.50 x 10 2 1.31 x 10-2 1.63 x 10 4 5.48 x 10 2 1.40 x 10-2 1.72 x 10 4 5.48 x 10 2 1.51 x 10-2 1.85 x 10 4 5.48 x 10 2 1.70 x 10-2 1.99 x 10 4 5.47 x 10 2 1.90 x 10-2 2.01 x 10 4 5.46 x 10 2 2.01 x 10-2 2.02 x 10 4 5.45 x 10 2 2.11 x 10-2 2.00 x 10 4 5.44 x 10 2 2.31 x 10-2 1.99 x 10 4 5.43 x 10 2 2.51 x 10-2 1.97 x 10 4 5.42 x 10 2 2.70 x 10-2 1.97 x 10 4 5.41 x 10 2 2.91 x 10-2 1.98 x 10 4 5.40 x 10 2 3.11 x 10-2 2.01 x 10 4 5.40 x 10 2 3.31 x 10-2 2.04 x 10 4 5.39 x 10 2 3.50 x 10-2 2.08 x 10 4 5.39 x 10 2 3.70 x 10-2 2.10 x 10 4 5.39 x 10 2 3.91 x 10-2 2.13 x 10 4 5.38 x 10 2 4.11 x 10-2 2.14 x 10 4 5.38 x 10 2 4.21 x 10-2 2.14 x 10 4 5.38 x 10 2 4.31 x 10-2 2.14 x 10 4 5.37 x 10 2 4.51 x 10-2 2.11 x 10 4 5.37 x 10 2 4.71 x 10-2 2.09 x 10 4 5.36 x 10 2 4.92 x 10-2 2.06 x 10 4 5.36 x 10 2 5.10 x 10-2 2.04 x 10 4 5.35 x 10 2 5.31 x 10-2 2.03 x 10 4 5.35 x 10 2 5.50 x 10-2 2.03 x 10 4 5.35 x 10 2 6.01 x 10-2 2.04 x 10 4 5.35 x 10 2 6.50 x 10-2 2.03 x 10 4 5.34 x 10 2 7.01 x 10-2 2.03 x 10 4 5.34 x 10 2 7.51 x 10-2 2.02 x 10 4 5.34 x 10 2 8.01 x 10-2 1.97 x 10 4 5.34 x 10 2 8.50 x 10-2 1.92 x 10 4 5.33 x 10 2 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 10 4 5.33 x 10 2 9.50 x 10-2 1.87 x 10 4 5.33 x 10 2 1.00 x 10-1 1.85 x 10 4 5.33 x 10 2 1.20 x 10-1 1.91 x 10 4 5.33 x 10 2 1.25 x 10-1 1.90 x 10 4 5.33 x 10 2 1.50 x 10-1 1.71 x 10 4 5.32 x 10 2 1.75 x 10-1 1.74 x 10 4 5.32 x 10 2 1.81 x 10-1 1.75 x 10 4 5.32 x 10 2 2.00 x 10-1 1.72 x 10 4 5.32 x 10 2 2.50 x 10-1 1.78 x 10 4 5.32 x 10 2 3.00 x 10-1 1.74 x 10 4 5.32 x 10 2 3.50 x 10-1 1.78 x 10 4 5.32 x 10 2 4.00 x 10-1 1.78 x 10 4 5.32 x 10 2 4.50 x 10-1 1.80 x 10 4 5.32 x 10 2 4.60 x 10-1 1.80 x 10 4 5.32 x 10 2 4.70 x 10-1 1.80 x 10 4 5.32 x 10 2 5.00 x 10-1 1.78 x 10 4 5.32 x 10 2 5.50 x 10-1 1.78 x 10 4 5.32 x 10 2 6.00 x 10-1 1.78 x 10 4 5.32 x 10 2 6.50 x 10-1 1.76 x 10 4 5.32 x 10 2 7.00 x 10-1 1.74 x 10 4 5.32 x 10 2 7.50 x 10-1 1.73 x 10 4 5.32 x 10 2 8.00 x 10-1 1.75 x 10 4 5.32 x 10 2 8.50 x 10-1 1.76 x 10 4 5.32 x 10 2 9.00 x 10-1 1.76 x 10 4 5.32 x 10 2 9.50 x 10-1 1.77 x 10 4 5.32 x 10 2 1.00 1.78 x 10 4 5.32 x 10 2 1.50 1.81 x 10 4 5.32 x 10 2 1.90 1.82 x 10 4 5.33 x 10 2 2.40 1.82 x 10 4 5.32 x 10 2 2.80 1.80 x 10 4 5.32 x 10 2 3.00 1.79 x 10 4 5.32 x 10 2
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 10 2 0.025 3269 6.41 x 10 2 0.1 3245 6.39 x 10 2 0.15 3233 6.39 x 10 2 0.225 3210 6.39 x 10 2 0.3 3198 6.39 x 10 2 0.4 3186 6.39 x 10 2 0.75 3186 6.38 x 10 2 0.875 3174 6.38 x 10 2 1.0 3151 6.38 x 10 2 1.2 3127 6.38 x 10 2 1.4 3103 6.38 x 10 2 1.6 3080 6.38 x 10 2 1.8 3056 6.38 x 10 2 2.0 3033 6.38 x 10 1 2.2 3009 6.38 x 10 2 2.4 2985 6.38 x 10 2 2.6 2962 6.38 x 10 2 2.8 2938 6.38 x 10 2 3.0 2915 6.38 x 10 2
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) (ft 2) 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 10 17 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) (ft 2) 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 27 28 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) (ft 2) 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-ft 2-°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.
10 7 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 (ft 3) 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 (ft 3) 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 10 6 15.88 0.600 54
Net Vessel Side Load (b)
Lbf Time (s) 1.184 x 10 6 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 10 6 Flowrate (post-LOCA) (gal/min) 2000 (600 for containment analysis)
Heat removal rate (post-LOCA) (Btu/h) 80.0 x 10 6 (31.2 x 10 6 for containment analysis)
Containment Cooler Fans
Type Vaneaxial
Number 4 Flowrate (high speed) (sf 3/min) 80,000 Static head (high speed) (in. wg) 4.75 Horsepower (high speed) (hp) 80 Flowrate (low speed) (sf 3/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 pr esents 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 10 6 Containment Cooler Fans (normal)
Type Vaneaxial
Number 4 Flowrate (each) (sft 3/min) 80,000 Static head (in. WG) 4.75 Motor horsepower (each) (hp) 80 Containment Recirculation Fans
Type Vaneaxial
Number 4 Flowrate (each) (sft 3/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) (sft 3/min) 40,000 Static head (in. WG)
9.0 Motor
horsepower (each) (hp) 100 Reactor Cavity Cooling Fans
Type Vaneaxial
Number 2 Flowrate (each) (sft 3/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) (sft 3/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) (sft 3 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 (ft
- 3) 4.92 x 10 4 Containment sprayed volume (ft
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 sft 3/min Heaters -Number 5 -Heater surface area/heater 35 ft 2 -Maximum heat flux 2850 Btu/h-ft 2 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
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 (sft 3/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 (sft 3/min) 500 -Differential pressure (wg) 2.7 -Design temperature (°F) 180 -Charcoal type iodine impregnated-Elemental I 2 removal efficiency 99.9 -Organic I 2 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 (ft 3/min) each 7500 Static pressure (in. w.g.)
2.3 Reactor
Cavity Hydrogen Dilution Fans
- Unit 1 Type Centrifugal Number required 2 Flow (ft 3/min) each 270 Static pressure (in. w.g.)
126.7 Unit 2 Type Vaneaxial Number required 2 Flow (ft 3/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 (ft 2)(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 NM P-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, P 0 = 0.0 58.3 1811 368 60.1 MSLB CASE 1, P 0 = -1.5 56.6 1811 383 100.1 MSLB CASE 2 55.7 1811 355 150.1 MSLB CASE 3, P 0 = 0.0 55.4 1811 362 170.1 MSLB CASE 3, P 0 = -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, P 0 = 0.0 63.0 400.7 294 75.1 MSLB CASE 9, P 0 = 3.0 67.0 400.7 288 80.0 MSLB CASE 10 58.6 1891 313 260.8 MSLB CASE 11, P 0 = 0.0 57.3 1831 342 300.8 MSLB CASE 11, P 0 = -1.5 56.0 1832 347 340.8 MSLB CASE 12, P 0 = 0.0 63.3 1811 359 180.1 MSLB CASE 12, P 0 = 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, P 0 = 3.0 59.4 1832 351 87.2 MSLB CASE 1, P 0 = -1.5 52.6 1828 367 92.2 MSLB CASE 8, P 0 = 3.0 62.2 1498 330 192 MSLB CASE 8, P 0 = -1.5 57.2 1503 347 157 MSLB CASE 9, P 0 = 3.0 64.5 482 347 87.2 MSLB CASE 9, P 0 = -1.5 59.2 1828 363 57.2 MSLB CASE 12, P 0 = 3.0 65.4 1518 331 162 MSLB CASE 12, P 0 = -1.5 60.3 1518 347 132 MSLB CASE 13, P 0 = 3.0 66.7 572 342 87.2 MSLB CASE 13, P 0 = -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 P O = 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, P O = 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, P O = 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, P O = +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, P O = +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, P O = -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, P O = +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, P O = -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 ft 2 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 ft 2 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 ft 2 D.E. BREAK 102% POWER, P O = 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 ft 2 D.E. BREAK 102% POWER, P O = -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 ft 2 D.E. BREAK 102% POWER, P O = - 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 ft 2 D.E. BREAK 102% POWER, P O = -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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 SPLIT 70% POWER, P O = +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 ft 2 SPLIT 70% POWER, P O = -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, P O = -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, P O = +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, P O = -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, P O = +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 ft 2 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 ft 2 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 ft 2 D.E. BREAK 30% POWER,P O = 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 ft 2 D.E. BREAK 30% POWER, P O = -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 ft 2 D.E. BREAK 30% POWER, P O = 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 ft 2 D.E. BREAK 30% POWER, P O = -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 ft 2 SPLIT 30% POWER, P O = -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 ft 2 SPLIT 30% POWER, P O = +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 ft 2 SPLIT 30% POWER, P O = -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 ft 2 SPLIT 30% POWER, P O = +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 P O = +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, P O =-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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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 ft 2 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-ft q 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 P O = 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 P O = 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 P O = +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, P O = +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, P O = +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 N 2 supply to N 2 gas 120 120 700 atm. 0 1500 sf 3/min accumulators RHR pumps Water 250 350 600 3 50 20 gal/min discharge SI line Accumulator to N 2 gas 120 120 700 0 0 1500 sf 3/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 (ft
- 3) 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 ch eck 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 (cm 3/h) (cm 3/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 cm 3/h). Miscellaneous large valves Double packing 1 cm 3/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 ft 3/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 ft 3/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 reci rculation 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-dri ven pumps are provided. One motor-driven pump feedwater pump automatic signal in conjunction wi th 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 va lves 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 Fa ilure 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 clos ure 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-d riven 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 moto r-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 s purious closure of either motor- valve and the steam generator spurious closure of a oper ated 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 spuri ous 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. Thes e 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, respec tively. 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 c onsidering 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 chemic al 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 injecti on 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.
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 (ft
- 2)
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.
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°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 condi tions 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 l ong 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 buffe r 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 meas urements 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 MgCl 2 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.
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 t he 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 10 5 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.,
++2 H 3(OH)Al2O 2H6Al 2 (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 K sp , as reported by Deltombe and Pourbaix, the following calculation can be made.
The solubility of Al(OH) 3 is determined from equation 2
[][][][]+=+=+++H 2 AlO 1110x28.2 H 2 AlO sp K O 2HH 2 AlO 3 Al(OH) at pH = 9.3
[]rmoles/lite 210x6.4 1010x5 1110x28.2 2 AlO== 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 10 5 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 (K sp) and solubilities at the various temperatures and solution pHs together with the sol uble 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 al uminum 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 H 3 B0 3). It was reported that even though the solution contained copious amounts of flocculent al uminum 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 soluti on in the tests conducted by Griess at 55°C and 100°C were approximately 4.0 and 18.6 grams, res pectively. 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 circumstanc es 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 t hat 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°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 te st 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.
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
(°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 inhi bitive 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 har mful 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/dm 2/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 K sp 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 10 3 4.73 x 10 2 8.55 x 10 4 8.55 x 10 3 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 10 5 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
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 ft 2 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 10 6 lbf. The maximum uplift force was 5.9 x 10 4 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 compar tment 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 collecti ng 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 containm ent sump performance m eets 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) av ailable 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 fo rm 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. Th e size of the openings in the fine screen take into account the overall oper ability 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 st ructure 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 indu ced by the chemically reactive spray during loss-of-coolant accident (LOCA) conditions.
E. A vortex breaker is provided at the su mp intake end of each of the pump suction pipes. F. The sumps are designed to yield low vel ocities of approach in the vicinity of the sumps to promote the settling out of debr is, 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 structu res 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 th e 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 requirem ents. 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 system s or between the redundant portions of the same system.
The motor-operated gate valves in the lines from th e containment sump to the various pumps are normally closed and remain closed during the inj ection phase of emergency core cooling system (ECCS) operation. The protective screened st ructures 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, su mp 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 curren t 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 th e 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 mode l 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 findi ngs from these tests it was concluded that it was not necessary to model the area outside the screen-g rating 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, wh ich could result in loss of pumping capacity or pump failure due to vibration. Such occu rrences 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 im peller and cause pump failure because vibration. Therefore, a satisfactory intake design should be free of air entraining vortices and have acceptable in take loss coefficients.
Lack of published and documented information relative to effects of the complex flow patterns approaching the intakes, the grati ng and screens, and the low viscosity of the heated water precluded analytical or empiri cal predictions as to whether the intake configuration would be free from objectionabl e vortex action. The plant conditions do not permit inplace testing. Therefore, a hy draulic 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 exa mination of Intake 1 separately from Intakes 2, 3 and 4.
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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 (R HR) 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 cond itions 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 ma ximum 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 te mperature 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 wa s 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 with in the screen area around the intake.
Piezometers were installed to measure st atic pressures inside and outside of the intake screens. Piezometer taps were in stalled 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 c onditions 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:
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6C-5 REV 21 5/08 a. Weber number, VD 2/, which is the ratio of surface tension to inertia forces.
- b. Froude number gD V , 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, VDf e , which characterizes the frequency of eddy shedding.
The parameters identified in the p receding 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, ft 3/s V - characteristic velocity, ft/s fe - frequency of eddy shedding, s
-1 g - gravitational acceleration, ft/s 2 - surface tension, lb/ft.
- kinematic viscosity, ft 2/s - Density, slugs/ft 3 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 mode l, 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 th e 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, conc urrent 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 e quivalence, the model Weber number was greater than the prototype value, whic h 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 th e prototype values of the relevant dimensionless numbers. Exceeding prototype values of the dimensionless numbers was considered to pr oduce conservative results.
C. THE MODEL TESTING PROGRAM The tests examined the performance of In take 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 Q 1 and Q 2 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 t han, 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 performan ce 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 Q 1 and Q 2 are indicated on figure 6C-2.
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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 a ccumulation 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 origi nated 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 ai r core vortex within the screen area that very quickly exhausted the trapped air.
- b. There was no vortex formed out side 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.
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6C-8 REV 21 5/08 3. Initial Documentation Tests The results from the initial documen tation 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 in to 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 le ft 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 th e 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 he ad loss across the screen and grating corrected to prototype discharge was 0.09 ft.
The intake loss coefficient was computed from the equation KhVg Vg=2 2 2 2// 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 fo rmed 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 u tilizing piezometric taps at 29 diameters downstream of the intake to further is olate 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 co rrection for pipe friction, was used:
g2 Vg2 Vhh K 2 2 f= where h f 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 dischar ges 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.
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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 subs tantially 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 measu red 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 irrespec tive 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 he ad 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 bl ockage cases or flow rates. The model indicated maximum combined prot otype 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 a ssociated with Intakes 2 and 3 were attributed to larger flow per un it area approaching the intakes and the more turbulent approach condition resulti ng 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 t hat 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 increa sed to assure that adequate NPSH is available and that system performance is sa tisfactory 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 co ld leg low head safety injection lines.
System tests conducted after these modificati ons show that the maximum flowrate with one pump operating during the cold leg reci rculation 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, ade quate NPSH is assured. These calculations take no credit for water above the containment su mp 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 prototyp e 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 co mputed 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 prototyp e 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 m odel 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 ca lculated 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 project s, it is the definite opinion of Western Canada Hydraulic Laboratories Ltd. and Bechte l 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 f ree from air entraining or flow reducing vortices.
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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 emerg ency core cooling and containment spray system recirculation intakes from the containm ent 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 m odeled 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 gr ating cage and the outer screen grating cage structure combination employed on Unit 1 wa s 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 wit hout 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 containmen t 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, r ange 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 wat er levels in the containment are 58.3 and 77.1 inches, respectively, above the floor.
The maximum expected containment su mp water temperature during recirculation following a postulated LOCA is about 212°F at subcooled pressures.
The flowrates, water depths, water temp erature, and the protecti ve screen structure for Unit 2 are identical to Unit 1, except that f our separate intake areas are provided for Unit 2 as compared to three in take 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 containm ent sump intake area.
Furthermore, the equipment layout at the Unit 2 containment fl oor 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 whic h 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 cont raction 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 norma lly 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 str ong 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 b een carried on by several investigators (see references in part II, G). The majority p resent 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 rep resented by the head loss coefficient K, is usually described (Anwar 1968, Amphlett 1976, Chang 1976) as:
K = F (local geometry, rmax RR, N , W)
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6C-15 REV 21 5/08 where local geometry = f (D, h, b)
H Q.No Reynolds Radial R R= NCirculationNo D QWWeberNo Q D h===..2 24 r max = 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 (R R >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 paramet ers: 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, th e intake head losses may be increased by nonuniform flow and/or circulation in the approach flow into the intakes.
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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 establis hes 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 c an 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 w ill 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 with in 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 c onfiguration 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 t hat 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 gr ating cage to remove circulation generated within the screen grati ng 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 e ffects 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 ma y be pressurized does not affect flow conditions. The flow field is estab lished by pressure differentials which would be the same in a closed syst em irrespective of the air pressure on the water surface.
FNP-FSAR-6C
6C-18 REV 21 5/08 e. If angularity of approach flo w 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 n ecessary to model only the screen grating cage, and all features inside the c age, 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 circulati on is an essential and necessary feature of a vortex, then irrespective of the stre ngth 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 Fi nal 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 conduc ted, 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 struct ure inward to the intake would be correctly simulated, and the intake entran ce 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, fo r 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 indica tes 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 sa id 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 effectiven ess 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 c age, 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 conditi ons 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 removi ng 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 contro lled 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 m ounted 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 gra ting 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 (ft 2/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 10 6 1.5 to 5.7 x 10 6 Considerable conservatism was inco rporated 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 th e 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 inject ed 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 th e 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 w ith 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 vide o 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 larg est 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 procedu res were developed for the Farley Unit 2 recirculation takes which:
- i. Modeled all effects of th e 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 whic h was greater than will occur in the prototype approach flow.
iii. Demonstrated satis factory 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 ea sily 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 vort ex 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 o ccurred 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 de grees, reducing the intake discharge from approximately 7400 gal/min to approx imately 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 followi ng 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 ft 2/s. The intake loss coefficient, K, was 0.69.
The maximum intake loss coefficie nt, 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 grati ng 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 low ered 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 coefficie nt was 0.67 and the average coefficient was 0.66.
The maximum screen head loss coefficient K s with or without blockage was 10.2. The K s values indicated a decreasing trend with increasing screen Reynolds number.
- 2. Phase 2 Tests The following summarizes the resu lts 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 ge nerated 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 co efficient was 0.69 and the maximum coefficient was 0.78. Internal vo rtices 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 vorti ces 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 coefficie nt 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 u sed 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, inter nal vortices could be developed by selective screen blockage. These vortices, which w ere formed only from smooth surfaces, did not increase intake entrance losses. How ever, 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 transmitte d 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 th e 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 calc ulated 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., "Simulit ude 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 Charact eristics 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 Re quirements 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 Th in Plate Orifices", thesis presented to the University of Wisconsin, at Madison, Wis., in 1956, in par tial 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 Institu te 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 Vort ex", 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 Vo rtex 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., "Exp erimental 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 V ital for Vertical Circulating Pumps," Power Engineering , vol. 54, No. 6, p. 74, 1950.
- 56. Young, C.A.H., "Swirl and Vortices at Inta kes", 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 Q 1 Intake 1 Q 2 Operating Water Temperature No. In. Prototype Model Prototype Model Prototype Model 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 CONDITIO NS 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 consis ts 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 (ft 2) Mirror RMI Foils (ft 2) RMI Jacketing (ft 2) Temp-Mat (ft 3) 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 (ft
- 2) Coating Volumes (ft
- 3) 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 (lb m) 200 Fiber (lb m) 30 Particulate (lb m) 170 Foreign Material Debris (ft
- 2) 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 [ft 3] 1 1.0 1 Coating Debris in 4D ZOI Modeled as Chips Concrete Coatings
[ft 2; ft 3] 523 ; 0.81 0.871 456 ; 0.71 Steel Coatings
[ft 2; ft 3] 1091 ; 1.36 0.704 768 ; 0.96 Sum [ft 2; ft 3] 1614 ; 2.18
--- 1224 ; 1.67 Unqualified Coating Debris Modeled as Fines / Chips
- Unqualified Coatings (Actual)
[ft 2; ft 3] 1070 ; 0.535 1.0 1070 ; 0.535 Latent Debris Latent Fiber (Walkdown)
[ft 3] 7.8 1.0 7.8 Latent Fiber (30 lb m) [ft 3] 12.5 1.0 12.5 Latent Particulate (Walkdown)
[ft 3] 0.63 1.0 0.63 Latent Particulate (170 lb m) [ft 3] 1.01 1.0 1.01 Reflective Metal Insulation Debris Transco Foil
[ft 2] 2383 0.799 1904 Mirror Foil
[ft 2] 35714 0.769 27464 Foil Sum [ft 2] 38097 --- 29368 RMI Jacketing
[ft 2] 8022 0.338 2711 Foreign Material Foreign Material 1 (labels, stickers, etc.) [ft 2] 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 LOCATION S 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 Locati on 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 Analys is, Auxiliary Feedwater System
7.3-11 Failure Mode and Effects Analysis, Emer gency 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, Engi neered 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 Recor ders 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 Generat or 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 Bor on 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 intermed iate 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 Synchroni zing diagram 4160-V emergency buses train A Units 1 and 2 177081 11/01/73 Synchroni zing 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
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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
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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
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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
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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
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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
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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)
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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
]
)(1 1 1 1 1 1 1 2 6 5 4 6 6 3 3 5 4 I sss s s s
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 e j ection (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/o r 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-A ccidental 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 reacto r coolant system flow (P) 15.3.4-Complete loss of forced reacto r coolant system flow (P) 15.4.4-Single reactor coolant pump locked rotor (P)
Reactor coolant pump, unde r- 15.3.4-Complete loss of forced reacto r 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 r otor (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-A ccidental 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 generato r 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 tri p N A 2. Power ran g e, neutron flux
- a. Hi g h 0.5 (a) b. Low 0.5 (a) 3. Power ran g e, neutron flux, 0.65 (a) hi g h p ositive rate
- 4. Not used.
- 5. Intermediate ran g e, neutron flux N A 6. Source ran g e, neutron flux N A 7. Overtemperature T (a)(b)(c) 8. Overpower T ~ (b)(c) 9. Pressurizer p ressure-low 2.0 10. P r essurizer p ressure-hi g h 1.0 11. Pressurizer water level-hi g hN A 12A. Loss of flow-sin g le loo p (above P-8) 1.0 12B. Loss of flow-two loo p s (above P-7 1.0 and below P-8
) 13. Steam g enerator water level-low-low 2.0 14. Undervolta g e-reactor coolant p um p sN A 15. Underfre q uenc y-reactor coolant p um p sN A 16. Turbine tri p a. Low auto sto p oil p ressure N A b. Turbine throttle valve closure N A 17. Safet y in j ection in p ut from ESF N A
FNP-FSAR-7 REV 25 4/14 TABLE 7.2-5 (SHEET 2 OF 2)
Functional Unit Response Time(s)
- 18. Reactor tri p s ystem interlocks N A 19. Reactor tri p breakers N A 20. A utomatic tri p lo g ic N A
- 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 ge nerating the required actuation signals for loss of coolant protection are as follows:
Pressurizer pressure
+/-14 psi (uncompensated)
Actuation signal accuracies required in generati ng 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. T avg 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 engineer ed 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 c oolant 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 R edundant 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 se rvice Breaker automatic trip Tw o 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 Tw o pumps are provided per train; the breaker control on manual or automat ic of control switch indicating standby pump will be st arted from the power to one signal, or trip the breaker light s 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 cont rol board or start automatically on loss of sequencer start automatically periodic test ing 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 oper ate: position and plant operating is open; post-LO CA 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:
pos ition 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 O perator can open valves K-610 fail to initially closed, syst em 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 operat ing 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 pl ant operating position closed; two containment operate on effect isolation; operator status: isolati on 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 isolat ion; 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 positi on 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 positi on 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; oper ator Depending on initial valve starter to can close/open valve pos ition 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 pl ant operating position closed; two containment operate on effect isolation; operator status: isolati on valves in series, one safety injection can close va lve 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 isolat ion; 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 positi on 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 positi on 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; operat or 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; oper ator 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 positi on 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 posit ion 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; oper ator 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 positi on 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 posit ion 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; oper ator 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 init ial valve Normal valve position open; post-LOCA fails to operator can close positi on 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 positi on 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 posit ion 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; oper ator 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 operat e: 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 operat e: 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 D epending on initial valve Oper ator can open valves relay K-610 is initially closed, syst em 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 automat ic 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 requir ed switch red, green, and or post-LO CA 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 ma lfunction 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 lo ss 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 clos ed 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. Va lve 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 lo ss 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 positi on 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 fr om 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 valv es in position and plant operating cl osed 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 r eactor 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 Valv e 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 init ial 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 isolati on; 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 positi on 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 posit ion 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 init ial 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 is olation; 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 positi on 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 posit ion 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 clos e/open position and pl ant 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 init ial valve Normal valve position open; post-LOCA fails to operate alternate valve 3046 position and pl ant 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 pos ition and plant operating fail to close operates to effect status: on CIAS phase B isolati on; 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 positi on 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 posit ion 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, Dependi ng 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 wi th 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 Valv e 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 in itial valve Valves are normally open and will fail closed fails to operate there are two valves in posit ion and plant operating on lo ss 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 clos ed 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 f ail closed on operate on receipt of can close valve from local plant operating status:
lo ss 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 r equired 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 operati ng 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 D epending 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 operati ng 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 pl ant 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; oper ator 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; oper ator 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 pl ant 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; oper ator 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 pl ant 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; oper ator 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; oper ator 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 pl ant 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 cl osed 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 los s 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 cl osed 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 D epending 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 los s 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 cl osed position and plant operating vent (sticky except fo r 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 in itial valve Valve is open only for smoke purge and fails to operate can close valve from main control board posit ion and plant operating status: will fail closed on loss of instr ument 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 pl ant 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 operat ion; c. Light monitor panel Units 1 or 2 operator can start motor if necessary Contacts of ESF Motor fails to start; of tw o a. Control room indication lights relay K-614 fail to pressurization fans, only one
- b. Periodic testing close on CIAS is required for operat ion; 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 clos ed 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 fa ils Depending on initial valve relay contacts open as is posit ion 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 al arm Two spray pumps provided; each unit motor control at main contro l 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 f an; 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 C ooling 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 contro l 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 al arm at Two MD auxiliary feedwater pumps motor control main contro l 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 Dies el 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 Indica ting 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 li ghts 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 Generat or 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 Annuncia tor 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 r unning 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 Observati on 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 Generat or 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 Observ ation 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 Observat ion of meter on Redundant generator can be used excitation from generator board circuit
Failure of Improper voltage output Observati on 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 la tch in and continue sequence If failure occurs after safety injection is called for at M1, safety injection signal will be reset; no effect provided safeg uards 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 pos sible Fail logic one Spurious safety injecti on 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; r eactor 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 T avg (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 T avg (train A only)
Fail logic one Spurious steam line isolation (train A) if false logic one output of A7 occurs coincident with low-low T avg 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 T avg A9 Fail logic zero Prevents low-low T avg and low steam line pressure safety injection block Fail logic one Prevents low steam line pressure safety injecti on and steam line isolation actuat ion 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 iso lation due to high steam line flow coincident with low-low T avg Fail logic one Allows operator to block safety injection whether or not block should be allowed; if safety injection is not bl ocked 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 pos sible 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 T avg; safety injection is not prevented if low steam line pressure occurs alone (not in coincidence with low-low T avg)
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 inje ction 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 T avg 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 le vel 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 operati on 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 se lected 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.
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 aux iliary 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 automatica lly 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 t hat 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 syst ems 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 rela ted 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 requirement s 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 (T H) thermocouple inputs. T is equal to T H minus T u , 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 eit her all the odd numbered Ts or even numbered Ts to have a low T error. T or T u is used to determine percent level for both the head area and the plenum area. A sensor is considered uncovered whenever T or T u is greater than 200°F or 700°F, respectively.
Five-degree dead bands exist in both the T and T u setpoints for uncovered sensors to prevent cycling.
FNP-FSAR-7 7.5-4 REV 21 5/08 A maximum T H 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 T H and T heater controller setpoint signals are reduced at a constant rate, when their respective T H and T values increase above a predetermined value. The T H 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 10 10 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 10 5 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 C oolant 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.,
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;
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 surveillance 122 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 surveillance 121 109 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 10 6 +/-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 t he 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 i ndication; 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 T cold or 1 T hot, 0°F to 700°F
+/-4% Both channels recorded Control board T hot (measured, 1 T cold 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 c hannel is indicated Control board steps/min Median Tavg 1 530°F to 630°F +/-4°F 1 channel is recorded Control board T ref 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.
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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
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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) mi tigation 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-tr ain 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 actuat ions 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:
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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 prov ides the AMSAC with automated and manual testing as well as a maintenance mode. Automated testing is the
continuously performed self checki ng 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.
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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 ac tuations 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 t he 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.
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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 syst em 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 us ed 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.
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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 perfo rmed 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. Si nce 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, physica lly 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-rela ted 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.
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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 mitigativ e 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.
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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 indica ted via the AMSAC general warning sent to the main control room.
7.8.2.12.4 Means for Bypassing A permanently installed system by pass selector switch is provi ded 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 feedw ater 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 feedwat er, 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