Text

1 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safety Features
	ML23319A068
Person / Time
Site:	Farley
Issue date:	10/31/2023
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML23318A074	List: ML23318A067; ML23318A070; ML23319A062; ML23319A063; ML23319A064; ML23319A065; ML23319A066; ML23319A067; ML23319A068; ML23319A069; ML23319A071; ML23319A073; ML23319A075; ML23319A076; ML23319A077; ML23319A078; ML23319A079; ML23319A080; ML23319A082; ML23319A083; ML23319A084; NL-23-0806, Technical Requirements Manual;
References
	NL-23-0806
	Download: ML23319A068 (1)
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FNP-FSAR-6 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 .......................................................................... 6.2-4 6.2.1.4 Containment Testing and Inspection ........................................... 6.2-33 6.2.1.5 Instrumentation Requirements..................................................... 6.2-36 6.2.1.6 Materials ...................................................................................... 6.2-36 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................................................................. 6.2-45 6.2.2.5 Instrumentation Requirements..................................................... 6.2-47 6.2.2.6 Materials ...................................................................................... 6.2-48 6.2.3 Containment Air Purification and Cleanup Systems..................................................................................... 6.2-49 6.2.3.1 Design Bases............................................................................... 6.2-49 6.2.3.2 System Design............................................................................. 6.2-55 6.2.3.3 Design Evaluation ........................................................................ 6.2-64 6.2.3.4 Tests and Inspections .................................................................. 6.2-67 6.2.3.5 Instrumentation Requirements..................................................... 6.2-71 6.2.3.6 Materials ...................................................................................... 6.2-73 6.2.4 Containment Isolation System ......................................................................... 6.2-73 6.2.4.1 Design Bases............................................................................... 6.2-73 6.2.4.2 System Design............................................................................. 6.2-74 6.2.4.3 Design Evaluation ........................................................................ 6.2-75 6.2.4.4 Tests and Inspections .................................................................. 6.2-75 6.2.4.5 Materials ...................................................................................... 6.2-76 6.2.5 Combustible Gas Control in Containment ............................................... 6.2-76 6.2.5.1 Design Bases .......................................................................................... 6.2-77 6.2.5.2 System Design. ....................................................................................... 6.2-79 6-i REV 31 10/23

FNP-FSAR-6 TABLE OF CONTENTS Page 6.2.5.3 Design Evaluation.................................................................................... 6.2-83 6.2.5.4 Tests and Inspections .............................................................................. 6.2-85 6.2.5.5 Instrumentation Requirements ................................................................ 6.2-86 6.2.5.6 Materials .................................................................................................. 6.2-87 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 ............................................ 6.3-19 6.3.2.16 Motor-Operated Valves and Controls ...................................................... 6.3-19 6.3.2.17 Manual Actions. ....................................................................................... 6.3-19 6.3.2.18 Process Instrumentation .......................................................................... 6.3-19 6.3.2.19 Materials. ................................................................................................. 6.3-19 6.3.3 Performance Evaluation .......................................................................... 6.3-20 6.3.3.1 Evaluation Model ..................................................................................... 6.3-20 6.3.3.2 ECCS Performance ................................................................................. 6.3-20 6.3.3.3 Alternate Analysis Methods ..................................................................... 6.3-20 6.3.3.4 Fuel Rod Perforations.............................................................................. 6.3-21 6-ii REV 31 10/23

FNP-FSAR-6 TABLE OF CONTENTS Page 6.3.3.5 Evaluation Model ..................................................................................... 6.3-21 6.3.3.6 Fuel Clad Effects ..................................................................................... 6.3-21 6.3.3.7 ECCS Performance ................................................................................. 6.3-21 6.3.3.8 Peaking Factors....................................................................................... 6.3-21 6.3.3.9 Fuel Rod Perforations.............................................................................. 6.3-21 6.3.3.10 Conformance with Interim Acceptance Criteria ....................................... 6.3-21 6.3.3.11 Effects of ECCS Operation on the Core .................................................. 6.3-21 6.3.3.12 Use of Dual Function Components .......................................................... 6.3-21 6.3.3.13 Dependence on Other Systems .............................................................. 6.3-22 6.3.3.14 Lag Times ................................................................................................ 6.3-23 6.3.3.15 Thermal Shock Considerations ............................................................... 6.3-24 6.3.3.16 Limits on System Parameters .................................................................. 6.3-24 6.3.4 Tests and Inspections .............................................................................. 6.3-24 6.3.5 Instrumentation Requirements ................................................................ 6.3-27 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 .............................................................................. 6.4-7 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 ................................................................................. 6.5-8 6.5.5 Instrumentation .......................................................................................... 6.5-8 APPENDIX 6A MATERIALS COMPATIBILITY REVIEW ................................................... 6A-1 APPENDIX 6B CONTAINMENT PRESSURE ANALYSIS ................................................. 6B-1 6-iii REV 31 10/23

FNP-FSAR-6 TABLE OF CONTENTS Page 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 6-iv REV 31 10/23

FNP-FSAR-6 LIST OF TABLES 6.2-1 Principal Containment Design Parameters 6.2-2 Heat Sink Geometric Data 6.2-3 Initial Conditions for Pressure Analysis 6.2-4 Heat Sink Thermodynamic Data 6.2-5 Engineered Safety Features Performance for Containment Pressure Transient Analysis 6.2-6 Containment Pressure Analysis Results for the Spectrum of RCS Break Sizes 6.2-7 System Parameters, Initial Conditions for Thermal Uprate 6.2-8 Safety Injection Flow - Minimum Safeguards 6.2-9 Safety Injection Flow - Maximum Safeguards 6.2-10 Double-Ended Hot Leg Break, Blowdown Mass and Energy Releases 6.2-11 Plant Data for Blowdown 6.2-12 Double-Ended Hot Leg Break, Mass Balance 6.2-13 Double-Ended Hot Leg Break, Energy Balance 6.2-14 Double-Ended Pump Suction Break, Blowdown Mass and Energy Releases 6.2-15 Reactor Cavity Release 6.2-16 Spray Line Break Release 6.2-17 Surge Line Break Release 6.2-18 Reactor Cavity Subcompartment Pressure Analysis Summary of Flow Paths and Vent Loss Coefficients 6.2-19 Containment Results for the Design Basis LOCA 6.2-20 Double-Ended Pump Suction Break - Minimum Safeguards, Reflood Mass and Energy Releases 6.2-21 LOCA Chronology of Events 6.2-22 Subcompartment Differential Pressure Results 6.2-23 Environmental Conditions for Containment Heat Removal Systems (Deleted) 6.2-24 Component Design Parameters for Containment Spray System and Containment Cooling System 6-v REV 31 10/23

FNP-FSAR-6 LIST OF TABLES 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 (Deleted) 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 6-vi REV 31 10/23

FNP-FSAR-6 LIST OF TABLES 6.2-48 Double-Ended Pump Suction Break - Maximum Safeguards, Post-Reflood Mass and Energy Releases 6.2-49 Double-Ended Pump Suction Break, Mass Balance, Maximum Safeguards 6.2-50 Double-Ended Pump Suction Break, Energy Balance, Maximum Safeguards 6.2-51 Double-Ended Hot Leg Break, Sequence of Events 6.2-52 Double-Ended Pump Suction Break - Minimum Safeguards, Sequence of Events 6.2-53 Double-Ended Pump Suction Break - Maximum Safeguards, Sequence of Events 6.2-54 LOCA Mass and Energy Release Analysis, Core Decay Heat Fraction 6.3-1 Emergency Core Cooling System Component Parameters 6.3-2 ECCS Relief Valve Data 6.3-3 Sequence of Changeover Operation from Injection to Recirculation (Deleted) 6.3-4 Time Analysis for ECCS Injection/Recirculation Switchover 6.3-5 Materials Employed for Emergency Core Cooling System Components 6.3-6 Normal Operating Status of Emergency Core Cooling 6.3-7 Single Active Failure Analysis for Emergency Core Cooling System Components 6.3-8 Maximum Potential Recirculation Loop Leakage External to Containment 6.3-9 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis 6.3-10 Emergency Core Cooling System Shared Functions Evaluation 6.5-1 Auxiliary Feedwater System Auxiliary Feedwater Pump Data 6.5-2 Failure Analysis of Auxiliary Feedwater System 6.5-3 Auxiliary Feedwater System Motor Operated Valve Data 6-vii REV 31 10/23

FNP-FSAR-6 LIST OF FIGURES 6.2-1 DEPSGB, Minimum ESF 1 AC Pressure vs. Time, PO = 0 PSIG 6.2-2 RSG DEPSG Minimum ESF 1 AC Pressure vs. Time, PO = 3 PSIG 6.2-3 DEHL, Minimum ESF, DBA Short Term Pressure vs. Time, PO = 0 PSIG 6.2-4 RSG DEHLG Minimum ESF, DBA Short Term Pressure vs. Time, PO = +3 PSIG 6.2-5 DECLG Maximum ESF Pressure vs. Time (Deleted) 6.2-6 RSG Pressure vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-6A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-7 RSG Temperature vs. Time Steam Line Full D. E. Break 102% Power, PO = +3 PSIG 6.2-7A RSG Pressure vs. Time Steam Line Full D.E. Break 102% Power, PO = -1.5 PSIG 6.2-8 Pressure vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-9 Temperature vs. Time Steam Line 0.7 ft2 D. E. Break 102% Power 6.2-10 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-10A Pressure vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-11 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 102% Power, PO = 0 PSIG 6.2-11A Temperature vs. Time Steam Line 0.6 ft2 D.E. Break 102% Power, PO = -1.5 PSIG 6.2-12 Pressure vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-13 Temperature vs. Time Steam Line 0.528 ft2 Split 102% Power 6.2-14 Pressure vs. Time Steam Line Full D. E. Break 70% Power 6.2-15 Temperature vs. Time Steam Line Full D. E. Break 70% Power 6.2-16 Pressure vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-17 Temperature vs. Time Steam Line 0.6 ft2 D. E. Break 70% Power 6.2-18 Pressure vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-19 Temperature vs. Time Steam Line 0.5 ft2 D. E. Break 70% Power 6.2-20 RSG Pressure vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = +3 PSIG 6.2-21 RSG Temperature vs. Time Steam Line 0.47 ft2 Split 70% Power, PO = -1.5 PSIG 6.2-22 RSG Pressure vs. Time Steam Line Full D. E. Break 30% Power, PO = -1.5 PSIG 6-viii REV 31 10/23

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

Time (Deleted) 6.2-40 DEPSGB Min ESF DBA Temperature vs. Time, PO = 0 PSIG 6-ix REV 31 10/23

FNP-FSAR-6 LIST OF FIGURES 6.2-41 RSG DEPSG Min ESF DBA Temperature vs. Time, PO = 3 PSIG 6.2-42 Containment Air Cooler Duty vs. Temperature 6.2-43 Thermal Heat Removal Efficiency of Containment Atmosphere Spray (Deleted) 6.2-44 Residual Heat Exchanger Design Duty Accident Mode 6.2-45 Mass and Energy Rate vs. Time for LOCA (Deleted) 6.2-46 LOCA Blowdown Mass and Energy Release Rates vs. Time (Deleted) 6.2-47 LOCA Post-Blowdown Mass and Energy Release Rates vs. Time (Deleted) 6.2-48 DEPSG Minimum ESF P/T Analysis, Long Term Condensing Heat Transfer Coefficient 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 6-x REV 31 10/23

FNP-FSAR-6 LIST OF FIGURES 6.2-66 Schematic of Reflood Code 19 Element Loop Model for a Pump Suction Break (Deleted) 6.2-67 Core Reflood Correlation (Deleted) 6.2-68 Comparison of Measured and Predicted Carryover Rate Fractions (Deleted) 6.2-69 Inlet Water Temperature vs. Time After End of Blowdown (Deleted) 6.2-70 Variation in Temperature Rise, Turnaround Time, and Quench Time with Respect to Core Elevation (Deleted) 6.2-71 Energy Balance Model (Deleted) 6.2-72 Reflood Rate and Carryover Fractions vs. Time After End of Blowdown (Deleted) 6.2-73 Flow Through Break vs. Time After End of Blowdown (Deleted) 6.2-74 Water Height vs. Time After End of Blowdown (Deleted) 6.2-75 Post-Reflood Loop Resistance Model (Deleted) 6.2-76 S/G Internal Energy vs. Time After Break (Deleted) 6.2-77 Energy Distribution vs. Time (Deleted) 6.2-78 RSG Temperature Profile Through Containment Wall, PO = +3 PSIG 6.2-79 RHR Heat Exchanger Duty vs. Time, RSG PO = +3 PSIG 6.2-80 Containment Air Cooler Duty vs. Time, RSG PO = +3 PSIG 6.2-81 Minimum Sump pH Following LOCA vs. Time (Deleted) 6.2-82 Minimum Partition Coefficient in the Sump vs. Solution pH (Deleted) 6.2-83 Hydrogen Generation Rate vs. Time in the Lower Compartment 6.2-84 Isolation Valve Arrangement through 6.2-89 6.2-90 Electric Hydrogen Recombiner 6.2-91 Electric Hydrogen Recombiner Schematic Diagram (Typical of One Recombiner) 6.2-92 Lower Compartment Plan 6.2-93 Section of Lower Reactor Compartment 6.2-94 Containment Hydrogen Concentration With One Electric Recombiner Started 1 Day after a LOCA 6.2-95 Hydrogen Concentration as a Function of Time in Containment Purge Mode 6-xi REV 31 10/23

FNP-FSAR-6 LIST OF FIGURES 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 6-xii REV 31 10/23

FNP-FSAR-6 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).

6.1-1 REV 21 5/08

FNP-FSAR-6

Auxiliary feedwater system (section 6.5).

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

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

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

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

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

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

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

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

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

6.1-2 REV 21 5/08

FNP-FSAR-6 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.

6.1-3 REV 21 5/08

FNP-FSAR-6 6.2 CONTAINMENT SYSTEMS 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2.1.1 Design Bases 6.2.1.1.1 Postulated Accident Conditions The containment, in conjunction with engineered safety features, is designed to withstand the internal pressure and coincident temperature resulting from the energy release of the loss-of-coolant accident (LOCA) associated with 2831 MWt and to limit the site boundary radiation dose to within the guidelines set forth in 10 CFR 100. The containment system functional design meets the NRC acceptance criteria contained in General Design Criteria 16 and 50 of 10 CFR Part 50.

For the original containment analysis, the LOCA was assumed to occur for a range of reactor coolant pipe breaks, up to and including a double-ended break of the largest reactor coolant pipe. For power uprate, steam generator replacement, and NSAL incorporation (PU/RSG/NSAL incorporation), only the limiting breaks were reanalyzed. The design bases also include a simultaneous loss of offsite electrical power and a failure of a single engineered safety feature (ESF).

The postulated accidents considered are as follows:

A. Double-ended pump suction guillotine (DEPSG), maximum ESF on safety injection (SI) flow. (Nonlimiting break not reanalyzed for PU/RSG/NSAL incorporation.)

B. DEPSG, minimum ESF. (RSG/NSAL Incorporation analysis for initial pressure, PO = +3 psig.)

C. 0.6 DEPSG, maximum ESF. (Nonlimiting break not reanalyzed for PU/RSG/NSAL Incorporation.)

D. 3 ft2 pump suction split (PSS), maximum ESF. (Nonlimiting break not reanalyzed for PU/RSG/NSAL Incorporation.)

E. Double-ended cold leg guillotine (DECLG), maximum ESF. (Nonlimiting break not reanalyzed for PU/RSG/NSAL Incorporation.)

F. Double-ended hot leg guillotine (DEHLG), maximum ESF. (Nonlimiting break not reanalyzed for PU/RSG/NSAL Incorporation.)

G. Double-ended hot let guillotine (DEHLG), blowdown phase only. (RSG/NSAL Incorporation analysis for initial pressure, PO = +3 psig.)

6.2-1 REV 31 10/23

FNP-FSAR-6 H. Spectrum of main steam line breaks.

Experience gained by the reactor manufacturer in analyzing several other dry containments has led to the conclusion that the hot leg break results in the highest containment peak pressure during blowdown because it produces the highest blowdown mass and energy release rates.

Since studies have confirmed that there is no reflood peak for the double-ended hot leg (DEHL) break, the analysis of the break extends only out to the end of blowdown. The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break and the addition of the stored energy in the steam generators. As a result, the pump suction break yields the highest energy flowrates during the post-blowdown period by including all of the available energy of the reactor coolant system (RCS) in calculating the releases to the containment.

The spectrum of break sizes together with minimum and maximum ESF have been studied to establish the upper bounds of containment pressure and temperature following the LOCA. The LOCA that results in the peak containment temperature at the end of blowdown is the DEHL break. The LOCA which produces the highest pressure is the DEPSG. A composite of these two LOCA conditions is defined as the design basis accident (DBA). The containment response analyses assume that plant offsite power is lost and a single failure of any active containment or safety-related component occurs simultaneously with the hypothesized pipe rupture. Other postulated simultaneous occurrences such as a seismic event or local pipe break effects are not explicitly evaluated in these sections, except in the event that such occurrences might affect the mass and energy release to the containment. The effects of these other simultaneous occurrences upon the containment structural design are evaluated in section 3.8.

6.2.1.1.2 Post-Accident Energy Sources In order to predict the peak containment pressure following an accident, energy sources are determined by the reactor manufacturer in the calculations of energy and mass release during postulated pipe break events. For reactor coolant pipe ruptures, energy sources include the stored heat energy contained within the primary coolant, SI water, fuel, cladding, reactor vessel internals, reactor vessel, reactor coolant piping, and steam generator secondary coolant and metal. Also considered as an energy source is the fission product decay energy generated within the reactor core. The available energy from the above source is added to the reactor coolant during the course of the event.

6.2.1.1.3 Contribution of Other Engineered Safety Features After an accident, the ESFs, in conjunction with the containment and plant cooling water systems, provide protection for the public and plant personnel from the accidental release of radioactive products from the reactor system. The engineered safeguards function to localize, control, mitigate, and terminate all postulated accidents to ensure that the offsite radiation dose is within the guidelines of 10 CFR 100.

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FNP-FSAR-6 The ESFs consist of the following systems:

A. High-head SI system.

B. Low-head SI system.

C. Containment spray system.

D. Containment cooling system.

E. Penetration room filtration system.

Consistent with criteria concerning loss of plant offsite power, the containment heat removal systems and emergency core cooling systems (ECCSs) are assumed to operate in their minimum heat removal modes. However, due to the redundancy of these systems, the reactor core is adequately cooled so that core integrity is constantly maintained and heat is continuously removed from the core in an orderly and predictable manner. The functional performance of the containment and the ECCSs relies upon the operation of the containment isolation systems as described in subsection 6.2.4. Required isolation operations are assumed for purposes of the containment design evaluation in paragraph 6.2.1.3.

The high- and low-head SI systems inject borated water into the RCS. This provides cooling to limit fuel cladding and core damage and thus minimize fission product release. An adequate shutdown margin is assured regardless of the reactor coolant temperature. The SI system also provides continuous long-term post-accident cooling of the reactor core by recirculating borated water from the containment sump through the RCS.

The containment for each unit is equipped, as follows, with two independent, full capacity systems for cooling the containment atmosphere after the postulated LOCA:

A. The containment spray system supplies borated water to cool the containment atmosphere. The spray system in combination with one of the containment air coolers operating at reduced speed is sized to provide adequate cooling with one of the two containment spray pumps in service on emergency power. The pumps take suction from the refueling water storage tank (RWST). When the RWST reaches low-low level, the suction of the containment spray pumps is aligned to pump water directly from the containment sump back into the containment atmosphere by means of the containment spray nozzles.

Trisodium phosphate is added to the recirculation sump to help reduce airborne iodine activity levels inside containment and to retain the removed iodine in solution in the sump B. The containment cooling system is designed to provide containment atmosphere mixing and cooling. The system design basis is to provide adequate containment cooling from the operation of one train of containment spray and one containment cooler. The system limits the pressure transient inside the containment following a LOCA.

The penetration room filtration system collects and processes ECCS recirculation leakages.

The system limits the environmental activity levels following a LOCA.

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FNP-FSAR-6 6.2.1.1.4 Subcompartment Differential Pressure The original differential pressure analyses for the steam generator compartments were based on a double-ended circumferential rupture of the reactor coolant cold leg. The original differential pressure analysis for the reactor cavity was based on a restrained guillotine break of 100 in.2 in the cold-leg pipe. Currently, Leak-Before-Break (LBB) is applicable to the steam generator subcompartments; therefore, main RCS loop pipes are not postulated to break. The next largest line is the pressurizer surge line (LBB is also approved for the pressurizer surge line but is not credited in this analysis). Differential pressures are maintained below design limits providing adequate venting to the containment.

6.2.1.1.5 Post-Accident Pressure Reduction The parameters affecting the assumed capability for post-accident pressure reduction are discussed in paragraph 6.2.1.3.

6.2.1.1.6 Rejection of Energy to the Outside Environment Parameters affecting the assumed capability to reject energy to the outside environment are discussed in paragraph 6.2.1.1.3.

6.2.1.2 System Design The design of the containment is based upon the mass and energy absorption capacity of the volume contained within the structure. The principal design parameters for the containment are given in table 6.2-1. These values are used for the pressure temperature analyses given in paragraph 6.2.1.3.3.

The design of the reactor and steam generator subcompartments considers applicable thermal, static, seismic, impingement force, and pressure loadings during a LOCA as described in section 3.8.

Materials compatibility considerations are discussed in appendix 6A.

6.2.1.3 Design Evaluation 6.2.1.3.1 Assurance of Containment Leaktightness The containment leakage surveillance system, used to assure containment leaktightness during plant operation, is described in paragraph 6.2.1.4.

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FNP-FSAR-6 6.2.1.3.2 System Capability Analysis A discussion of system capability is given in paragraph 6.2.1.4.

6.2.1.3.3 Containment Pressure Transient Analysis A. Pipe Break Spectrum In the event of a hypothetical LOCA, or main steam line break, the release of the coolant from the rupture area will cause the high pressure, high temperature fluid to rapidly flash to steam and water within the containment. The release of this mass and energy will result in a rise in the pressure and temperature of the containment atmosphere. The rate and magnitude of the pressure increase depend upon the nature, location, and size of the rupture. In order to establish the controlling rupture, a spectrum of primary and secondary coolant breaks is considered. The reactor coolant breaks examined are from a condition of full rated power. Secondary coolant system break analysis considers a spectrum of breaks at different power levels. All of these main steam line breaks allow complete blowdown of one steam generator. These postulated accidents are evaluated to determine their significance in selecting a containment design basis.

The most severe of these accidents is selected as the controlling containment DBA.

B. Initial Conditions and Input Data The containment pressure analysis input data have been based upon the final plant design. A conservative prediction of LOCA consequences has been assured by determining expected values of containment initial conditions and geometric and thermodynamic parameters. A thorough discussion of the input data is given below.

The containment design parameters which determine the net free internal volume, the containment surface areas, and the design pressure and temperature are given in tables 6.2-1 and 6.2-2. As an additional conservatism, the volume occupied by the reactor coolant prior to the LOCA is included as occupied volume rather than free volume.

The initial conditions within the RCS and the containment system prior to accident initiation are given in table 6.2-3. The containment pressure and temperature response analyses are conducted assuming a minimum of available ECCS is in operation. For the LOCA, the containment system is assumed to be at ambient pressure or the maximum pressure permitted under the Technical Specifications, and maximum inside and maximum outside design operating air temperatures to minimize heat transfer during a LOCA. Main steam line break (MSLB) cases were analyzed at varying initial pressure conditions as described below.

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FNP-FSAR-6 The containment heat sink data used in the LOCA analysis are fully described in tables 6.2-2 and 6.2-4. Table 6.2-2 lists the geometry of each heat sink and the way it is modeled for the analysis. An air gap equivalent to a thermal conductance of 100 BTU/(hr-ft2-°F) is postulated for the interface between the containment liner and wall and for the interface between the stainless steel refueling canal liner and the concrete.

Table 6.2-4 lists the material properties and heat transfer coefficients used in the analysis. The coating properties were supplied by the manufacturer. Metal and concrete properties are typical for the temperature range expected. The steel imbedded in the concrete was not considered in the concrete conductivity.

Containment air cooler unit duty, Btu/h, as a function of containment saturation temperature, is given in figure 6.2-42. The RHR heat exchanger is explicitly modeled using the heat exchanger parameters.

C. Accident Identification and Results Containment pressure/temperature vs. time responses for the various breaks are shown on figures 6.2-1 through 6.2-41. The peak pressures, times of peak pressure, peak temperatures, times of peak temperature, and blowdown energy releases at the times of peak pressure are given in table 6.2-6 for the spectrum of breaks for LOCA. Based on the results presented, the DEHL with minimum ESF produces the maximum containment temperature over the short term while the DEPSG LOCA produces the maximum pressure. Both form the design basis for the short-term. The DEPSG LOCA produces limiting conditions over the post-blowdown period and is considered the design basis for the post-blowdown period. The limiting LOCA was reanalyzed as described in paragraph 6.2.1.3.3, and the resultant peak pressure is below the design pressure of 54 psig, as shown in table 6.2-6.

The blowdown mass and energy release rates as a function of time for the LOCA cases are presented in tables 6.2-10 and 6.2-14.

For the time of peak pressure, a detailed mass and energy balance has been performed on the RCS and containment. These data are given in tables 6.2-19 and 6.2-20. Table 6.2-19 lists the calculated containment pressures, temperatures, and masses for the time of pipe ruptures and the time of peak pressure with the limiting LOCA. Table 6.2-20 gives the energy distribution in the RCS and in the containment at the time of the break and at the time of peak pressure. This table verifies the energy balance during the containment pressurization period since the total RCS energy release equals the net gain in the containment system energy.

The containment condensing heat transfer coefficient vs. time for the DEPSG LOCA case is shown on figure 6.2-48. The initial portion of the curve is the Modified Tagami value with a maximum of 227 Btu/h-ft2-°F at 20 s. After 20 s, 6.2-6 REV 31 10/23

FNP-FSAR-6 the value decays to the Uchida value, which is dependent upon the slowly changing steam air mass ratio in the containment atmosphere.

6.2.1.3.4 Method of Analysis 6.2.1.3.4.1 LOCA Mass and Energy Releases The uncontrolled release of pressurized high temperature reactor coolant, termed a LOCA, will result in release of steam and water into the containment. This, in turn, will result in increases in the local subcompartment pressures and an increase in the global containment pressure and temperature. Therefore, there are both long- and short-term issues reviewed relative to a postulated LOCA that must be considered at the PU/RSG conditions for Farley Units 1 and 2.

The long-term LOCA mass and energy releases are analyzed to approximately 3600 seconds and are utilized as input in the containment integrity analysis, which demonstrates the acceptability of the containment safeguards systems to mitigate the consequences of a hypothetical large break LOCA. The containment safeguards systems must be capable of limiting the peak containment pressure to less than the design pressure and to limit the temperature excursion to less than the Environmental Qualification (EQ) acceptance limits. For this program, Westinghouse generated the mass and energy releases using the March 1979 model, described in references 7 and 27 through 29. The NRC review and approval letter is included with reference 7. Even though this is a first time application of this methodology for Farley Units 1 and 2, it has also been utilized and approved on many plant-specific dockets.

This section discusses the long-term LOCA mass and energy releases analysis. The results of this analysis were provided for use in the containment integrity analysis and EQ reviews.

The short-term LOCA-related mass and energy releases are used as input to the subcompartment analyses, which are performed to ensure that the walls of a subcompartment can maintain their structural integrity during the short pressure pulse (generally

< 3 s) accompanying a high-energy line pipe rupture within that subcompartment. The subcompartments evaluated included the steam generator compartment, the reactor cavity region, and the pressurizer compartment. For the reactor cavity region, the fact that Farley is approved for LBB was used to qualitatively demonstrate that any changes associated with PU/RSG are offset by the LBB benefit of using the smaller RCS nozzle breaks, thus demonstrating that the current licensing bases for these subcompartments remain bounding.

LBB is also applicable to the steam generator subcompartments; therefore, main RCS loop pipes are not postulated to break. The next largest line is the pressurizer surge line, and mass and energy releases as discussed below are used in the PU/RSG analysis of the steam generator subcompartments. For the pressurizer compartment, the critical mass flux correlation utilized in the SATAN computer program (reference 10) was used to conservatively estimate the impact of the changes in RCS temperatures on the short-term releases. The power uprate replacement steam generator (PU/RSG) program evaluation showed that the releases would increase by 18% from the original design basis. The measurement uncertainty recapture (MUR) program evaluation showed that it would be bounded by PU/RSG program. This section discusses the short-term evaluation. The results of this evaluation were provided for use in the pressurizer subcompartment evaluation.

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FNP-FSAR-6 Long-Term LOCA Mass and Energy Releases The mass and energy release rates described in this section form the basis of further computations by Southern Nuclear Operating Company (SNC) to evaluate the containment following the postulated accident. Discussed in this section are the long-term LOCA mass and energy releases for the hypothetical double-ended pump suction (DEPS) rupture and DEHL rupture break cases.

Input Parameters and Assumptions The mass and energy release analysis is sensitive to the assumed characteristics of various plant systems, in addition to other key modeling assumptions. Where appropriate, bounding inputs are utilized and instrumentation uncertainties are included. For example, the RCS operating temperatures are chosen to bound the highest average coolant temperature range of all operating cases, and a temperature uncertainty allowance of (+6.0°F) is then added. Nominal parameters are used in certain instances. For example, the RCS pressure in this analysis is based on a nominal value of 2250 psia plus an uncertainty allowance (+50 psi). All input parameters are chosen consistent with accepted analysis methodology.

Some of the most critical items are the RCS initial conditions, core decay heat, SI flow, and primary and secondary metal mass and steam generator heat release modeling. Specific assumptions concerning each of these items are discussed next. Tables 6.2-7 through 6.2-9 present key data assumed in the analysis.

A core power of 2820.5 MWt representing rated thermal power (RTP) adjusted for calorimetric error was used in the analysis. As previously noted, the use of RCS operating temperatures to bound the highest average coolant temperature range were used as bounding analysis conditions. The use of higher temperatures is conservative because the initial fluid energy is based on coolant temperatures which are at the maximum levels attained in steady-state operation. Additionally, an allowance to account for instrument error and deadband is reflected in the initial RCS temperatures. The selection of 2250 psia as the limiting pressure is considered to affect the blowdown phase results only, since this represents the initial pressure of the RCS. The RCS rapidly depressurizes from this value until the point at which it equilibrates with containment pressure.

The rate at which the RCS blows down is initially more severe at the higher RCS pressure.

Additionally, the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RCS mass available for releases. Thus, 2250 psia plus uncertainty was selected for the initial pressure as the limiting case for the long-term mass and energy release calculations.

The selection of the fuel design features for the long-term mass and energy release calculation is based on the need to conservatively maximize the energy stored in the fuel at the beginning of the postulated accident (i.e., to maximize the core stored energy). The core stored energy was selected to bound the 17 x 17 optimized fuel assembly (OFA) fuel product loaded at Farley Units 1 and 2. The margins in the core stored energy address the thermal fuel model and the 6.2-8 REV 31 10/23

FNP-FSAR-6 associated manufacturing uncertainties and the time in the fuel cycle for the maximum fuel densification. Thus, the analysis very conservatively accounts for the stored energy in the core.

Margin in RCS volume of 3% (which is composed of 1.6% allowance for thermal expansion and 1.4% uncertainty) is modeled.

A uniform steam generator (SG) tube plugging level of 0% is modeled. This assumption maximizes the reactor coolant volume and fluid release by virtue of consideration of the RCS fluid in all SG tubes. During the post-blowdown period the SGs are active heat sources since significant energy remains in the secondary metal and secondary mass that has the potential to be transferred to the primary side. The 0% tube plugging assumption maximizes heat transfer area and, therefore, the transfer of secondary heat across the SG tubes. Additionally, this assumption reduces the reactor coolant loop resistance, which reduces the P upstream of the break for the pump suction breaks and increases break flow. Thus, the analysis very conservatively accounts for the level of SG tube plugging.

Regarding SI flow, the mass and energy release calculation considered configurations/failures to conservatively bound respective alignments. The cases include (a) a Minimum Safeguards case (1 CH/SI and 1 LHSI pump); and (b) a Maximum Safeguards case (2 CH/SI and 2 LHSI pumps).

The following assumptions were employed to ensure that the mass and energy releases are conservatively calculated, thereby maximizing energy release to containment.

1. Maximum expected operating temperature of the RCS (100% full-power conditions).

2. Allowance for RCS temperature uncertainty (+6.0°F).

3. Margin in RCS volume of 3% (which is composed of 1.6% allowance for thermal expansion and 1.4% for uncertainty).

4. Analyzed core power of 2830.5-MWt.

5. Item #4 includes an allowance for a plant specific calorimetric error.

6. Conservative heat transfer coefficients (i.e., steam generator primary/secondary heat transfer and RCS metal heat transfer).

7. Allowance in core stored energy for effect of fuel densification.

8. A margin in core stored energy to bound OFA fuel.

9. An allowance for RCS initial pressure uncertainty (+50 psi).

10. A maximum containment backpressure equal to design pressure (54 psig).

11. Allowance for RCS flow uncertainty (-2.4%).

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FNP-FSAR-6

12. Steam generator tube plugging leveling (0% uniform).

Maximizes reactor coolant volume and fluid release.

Maximizes heat transfer area across the SG tubes.

Reduces coolant loop resistance, which reduces the P upstream of the break for the pump suction breaks and increases break flow.

Limiting releases bounding for both units are provided herein.

Thus, based on the previously discussed conditions and assumptions, a bounding analysis of Farley Units 1 and 2 was made for the release of mass and energy from the RCS in the event of a LOCA.

Description of Analyses The evaluation model used for the long-term LOCA mass and energy release calculations is the March 1979 model described in reference 7. This evaluation model has been reviewed and approved generically by the NRC. The approval letter is included with references 7 and 27 through 29. Even though this is a first time application for Farley Units 1 and 2, it has also been utilized and approved on the plant-specific dockets for other Westinghouse PWRs.

This report section presents the long-term LOCA mass and energy releases generated in support of the Farley Units 1 and 2 uprating program. These mass and energy releases are then subsequently used in the containment integrity analysis.

LOCA M&E Release Phases The containment system receives mass and energy releases following a postulated rupture in the RCS. These releases continue over a time period which, for the LOCA mass and energy analysis, is typically divided into four phases.

1. Blowdown - the period of time from accident initiation (when the reactor is at steady-state operation) to the time that the RCS and containment reach an equilibrium state.

2. Refill - the period of time when the lower plenum is being filled by accumulator and ECCS water. At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum. To conservatively consider the refill period for the purpose of containment mass and energy releases, it is assumed that this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum. This allows an uninterrupted release of mass and energy to containment. Thus, the refill period is conservatively neglected in the mass and energy release calculation.

3. Reflood - begins when the water from the lower plenum enters the core and ends when the core is completely quenched.

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FNP-FSAR-6

4. Post-reflood (Froth) - describes the period following the reflood phase. For the pump suction break, a two-phase mixture exits the core, passes through the hot legs, and is superheated in the SGs prior to exiting the break as steam. After the broken loop SG cools, the break flow becomes two-phase.

Computer Codes The reference 7 mass and energy release evaluation model is comprised of mass and energy release versions of the following codes: SATAN VI, WREFLOOD, FROTH, AND EPITOME.

These codes were used to calculate the long-term LOCA mass and energy releases for Farley Units 1 and 2.

SATAN VI calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation including pressure, enthalpy, density, mass and energy flowrates, and energy transfer between primary and secondary systems as a function of time.

The WREFLOOD code addresses the portion of the LOCA transient where the core reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the ECCS refills the reactor vessel and provides cooling to the core. The most important feature of WREFLOOD is the steam/water mixing model.

FROTH models the post-reflood portion of the transient. The FROTH code is used for the SG heat addition calculation from the broken and intact loop SGs.

EPITOME continues the FROTH post-reflood portion of the transient from the time at which the secondary equilibrates to containment design pressure to the end of the transient. It also compiles a summary of data on the entire transient, including formal instantaneous mass and energy release tables and mass and energy balance tables with data at critical times.

Break Size and Location Generic studies have been performed with respect to the effect of postulated break size on the LOCA mass and energy releases. The double-ended guillotine break has been found to be limiting due to larger mass flowrates during the blowdown phase of the transient. During the reflood and froth phases, the break size has little effect on the releases.

Three distinct locations in the RCS loop can be postulated for pipe rupture for any release purposes:

Hot leg (between vessel and steam generator).

Cold leg (between pump and vessel).

Pump suction (between steam generator and pump).

The break locations analyzed for this program are the double-ended pump suction (DEPS) rupture (10.48 ft2) and the DEHL rupture (9.154 ft2). Break mass and energy releases have 6.2-11 REV 31 10/23

FNP-FSAR-6 been calculated for the blowdown, reflood, and post-reflood phases of the LOCA for the DEPS cases. For the DEHL case, the releases were calculated only for the blowdown. The following information provides a discussion on each break location.

The DEHL rupture has been shown in previous studies to result in the highest blowdown mass and energy release rates. Although the core flooding rates would be the highest for this break location, the amount of energy released from the SG secondary is minimal because the majority of the fluid which exits the core vents directly to containment bypassing the SGs. As a result, the reflood mass and energy releases are reduced significantly as compared to either the pump suction or cold leg break locations where the core exit mixture must pass through the SGs before venting through the break. For the hot leg break, generic studies have confirmed that there is no reflood peak (i.e., from the end of the blowdown period the containment pressure would continually decrease). Therefore, only the mass and energy releases for the hot leg break blowdown phase are calculated and presented in this section of the report.

The cold leg break location has also been found in previous studies to be much less limiting in terms of the overall containment energy releases. The cold leg blowdown is faster than that of the pump suction break, and more mass is released into the containment. However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment. Studies have determined that the blowdown transient for the cold leg is, in general, less limiting than that for the pump suction break. During reflood, the flooding rate is greatly reduced and the energy release rate into the containment is reduced. Therefore, the cold leg break is bounded by other breaks and no further evaluation is necessary.

The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break, and the addition of the stored energy in the steam generators. As a result, the pump suction break yields the highest energy flowrates during the post-blowdown period by including all of the available energy of the RCS in calculating the releases to containment.

Application of Single-Failure Criterion An analysis of the effects of the single-failure criterion has been performed on the mass and energy release rates for each break analyzed. An inherent assumption in the generation of the mass and energy release is that offsite power is lost. This results in the actuation of the emergency diesel generators required to power the SI system. This is not an issue for the blowdown period which is limited by the DEHL break.

Two cases have been analyzed to assess the effects of a single failure. The first case assumes minimum safeguards SI flow based on the postulated single failure of an emergency diesel generator. This results in the loss of one train of safeguards equipment. The other case assumes maximum safeguards SI flow based on no postulated failures that would impact the amount of ECCS flow. The analysis of the cases described provides confidence that the effect of credible single failures is bounded.

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FNP-FSAR-6 Acceptance Criteria for Analyses A large break LOCA is classified as an ANS Condition IV event, an infrequent fault. To satisfy the Nuclear Regulatory Commission (NRC) acceptance criteria presented in the Standard Review Plan section 6.2.1.3, the relevant requirements are as follows:

10 CFR 50, Appendix A.

10 CFR 50, Appendix K, paragraph I.A.

In order to meet these requirements, the following must be addressed:

Sources of energy.

Break size and location.

Calculation of each phase of the accident.

M&E Release Data Blowdown Mass and Energy Release Data The SATAN-VI code is used for computing the blowdown transient. The code utilizes the control volume (element) approach with the capability for modeling a large variety of thermal fluid system configurations. The fluid properties are considered uniform and thermodynamic equilibrium is assumed in each element. A point kinetics model is used with weighted feedback effects. The major feedback effects include moderator density, moderator temperature, and Doppler broadening. A critical flow calculation for subcooled (modified Zaloudek), two-phase (Moody), or superheated break flow is incorporated into the analysis. The methodology for the use of this model is described in reference 7.

Table 6.2-10 presents the calculated mass and energy release for the blowdown phase of the DEHL break. For the hot leg break mass and energy release tables, break path 1 refers to the mass and energy exiting from the reactor vessel side of the break; break path 2 refers to the mass and energy exiting from the SG side of the break.

Table 6.2-14 presents the calculated mass and energy releases for the blowdown phase of the DEPS break. For the pump suction breaks, break path 1 in the mass and energy release tables refers to the mass and energy exiting from the SG side of the break; break path 2 refers to the mass and energy exiting from the pump side of the break.

Reflood Mass and Energy Release Data The WREFLOOD code is used for computing the reflood transient. The WREFLOOD code consists of two basic hydraulic models - one for the contents of the reactor vessel and one for the coolant loops. The two models are coupled through the interchange of the boundary conditions applied at the vessel outlet nozzles and at the top of the downcomer. Additional transient phenomena such as pumped SI and accumulators, reactor coolant pump performance, 6.2-13 REV 31 10/23

FNP-FSAR-6 and SG release are included as auxiliary equations which interact with the basic models as required. The WREFLOOD code permits the capability to calculate variations during the core reflooding transient of basic parameters such as core flooding rate, core and downcomer water levels, fluid thermodynamic conditions (pressure, enthalpy, density) throughout the primary system, and mass flowrates through the primary system. The code permits hydraulic modeling of the two flow paths available for discharging steam and entrained water from the core to the break, i.e., the path through the broken loop and the path through the unbroken loops.

A complete thermal equilibrium mixing condition for the steam and ECCS injection water during the reflood phase has been assumed for each loop receiving ECCS water. This is consistent with the usage and application of the reference 7 mass and energy release evaluation model in the initial docketed analysis for this methodology, e.g., D.C. Cook Docket (reference 8). Even though the reference 7 model credits steam/water mixing only in the intact loop and not in the broken loop, the justification, applicability, and NRC approval for using the mixing model in the broken loop has been documented (reference 8). Moreover, this assumption is supported by test data and is further discussed below.

The model assumes a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two distinct physical processes. The first is a two-phase interaction with condensation of steam by cold ECCS water.

The second is a single-phase mixing of condensate and ECCS water. Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that need be considered. (Any spillage directly heats only the sump.)

The most applicable steam/water mixing test data have been reviewed for validation of the containment integrity reflood steam/water mixing model. These data were generated in 1/3-scale tests (reference 9), which are the largest scale data available and, thus, most clearly simulate the flow regimes and gravitational effects that would occur in a PWR. These tests were designed specifically to study the steam/water interaction for PWR reflood conditions.

A group of 1/3-scale tests corresponds directly to containment integrity reflood conditions. The injection flowrates for this group cover all phases and mixing conditions calculated during the reflood transient. The data from these tests were reviewed and discussed in detail in reference 7. For all of these tests, the data clearly indicate the occurrence of very effective mixing with rapid steam condensation. The mixing model used in the containment integrity reflood calculation is, therefore, wholly supported by the 1/3-scale steam/water mixing data.

Additionally, the following justification is also noted. The post-blowdown limiting break for the containment integrity peak pressure analysis is the pump suction double-ended rupture break.

For this break, there are two flow paths available in the RCS by which mass and energy may be released to containment. One is through the outlet of the SG, the other via reverse flow through the reactor coolant pump. Steam which is not condensed by ECCS injection in the intact RCS loops passes around the downcomer and through the broken loop cold leg and pump and then vents to containment. This steam also encounters ECCS injection water as it passes through the broken loop cold leg; complete mixing occurs and a portion of it is condensed. It is this portion of steam which is condensed that is taken credit for in this analysis. This assumption is justified based upon the postulated break location and the actual physical presence of the 6.2-14 REV 31 10/23

FNP-FSAR-6 ECCS injection nozzle. A description of the test and test results is contained in references 7 and 9.

Tables 6.2-20 and 6.2-46 present the calculated mass and energy releases for the reflood phase of the pump suction double-ended rupture, minimum safeguards, and maximum safeguards cases, respectively.

The transient response of the principal parameters during reflood are given in tables 6.2-42 and 6.2-47 for the DEPS cases.

Post-Reflood Mass and Energy Release Data The FROTH code (reference 10) is used for computing the post-reflood transient. The FROTH code calculates the heat release rates resulting from a two-phase mixture present in the SG tubes. The mass and energy releases that occur during this phase are typically superheated due to the depressurization and equilibration of the broken loop and intact loop SGs. During this phase of the transient, the RCS has equilibrated with the containment pressure, but the SGs contain a secondary inventory at an enthalpy that is much higher than the primary side.

Therefore, there is a significant amount of reverse heat transfer that occurs. Steam is produced in the core due to core decay heat. For a pump suction break, a two-phase fluid exits the core, flows through the hot legs, and becomes superheated as it passes through the SG. Once the broken loop cools, the break flow becomes two-phase. During the FROTH calculation, ECCS injection is addressed for both the injection phase and the recirculation phase. The FROTH code calculation stops when the secondary side equilibrates to the saturation temperature (Tsat) at the containment design pressure; after this point the EPITOME code completes the SG depressurization. (See the subsection Steam Generator Equilibration and Depressurization for additional information.)

The methodology for the use of this model is described in reference 7. The mass and energy release rates are calculated by FROTH and EPITOME until the time of containment depressurization. After containment depressurization (14.7 psia), the mass and energy release available to containment is generated directly from core boiloff/decay heat.

Tables 6.2-43 and 6.2-48 present the two-phase post-reflood mass and energy release data for the pump suction double-ended cases.

Decay Heat Model On November 2, 1978, the Nuclear Power Plant Standards Committee (NUPPSCO) of the American Nuclear Society approved ANS Standard 5.1 (reference 11) for the determination of decay heat. This standard was used in the mass and energy release model for Farley Nuclear Plant Units 1 and 2. Table 6.2-54 lists the decay heat curve used in the Farley SG replacement mass and energy release analysis.

6.2-15 REV 31 10/23

FNP-FSAR-6 Significant assumptions in the generation of the decay heat curve for use in the LOCA mass and energy releases analysis include the following:

1. Decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.

2. Decay heat power from fissioning isotopes other than U-235 is assumed to be identical to that of U-235.

3. Fission rate is constant over the operating history of maximum power level.

4. The factor accounting for neutron capture in fission products has been taken from Table 10 of reference 11.

5. The fuel has been assumed to be at full power for 108 s.

6. The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.

7. Two sigma uncertainty (two times the standard deviation) has been applied to the fission product decay.

Based upon NRC staff review, Safety Evaluation Report (SER) of the March 1979 evaluation model (reference 7), use of the ANS Standard 5.1, November 1979 heat model was approved for the calculation of mass and energy releases to the containment following a LOCA.

Steam Generator Equilibration and Depressurization SG equilibration and depressurization is the process by which secondary side energy is removed from the SGs in stages. The FROTH computer code calculates the heat removal from the secondary mass until the secondary temperature is the saturation temperature (Tsat) at the containment design pressure. After the FROTH calculations, the EPITOME code continues the FROTH calculation for SG cooldown removing SG secondary energy at different rates (i.e., first-and second-stage rates). The first-stage rate is applied until the SG reaches Tsat at the user specified intermediate equilibration pressure, when the secondary pressure is assumed to reach the actual containment pressure. Then the second-stage rate is used until the final depressurization, when the secondary reaches the reference temperature of Tsat at 14.7 psia, or 212°F. The heat removal of the broken loop and intact loop SGs is calculated separately.

During the FROTH calculations, SG heat removal rates are calculated using the secondary side temperature, primary side temperature, and a secondary side heat transfer coefficient determined using a modified McAdams correlation (reference 12). SG energy is removed during the FROTH transient until the secondary side temperature reaches saturation temperature at the containment design pressure. The constant heat removal rate used during the first heat removal stage is based on the final heat removal rate calculated by FROTH. The SG energy available to be released during the first-stage interval is determined by calculating the difference in secondary energy available at the containment design pressure and that at the (lower) user specified intermediate equilibration pressure, assuming saturated conditions. This 6.2-16 REV 31 10/23

FNP-FSAR-6 energy is then divided by the first-stage energy removal rate, resulting in an intermediate equilibration time. At this time, the rate of energy release drops substantially to the second stage rate. The second stage rate is determined as the fraction of the difference in secondary energy available between the intermediate equilibration and final depressurization at 212°F, and the time difference from the time of the intermediate equilibration to the user specified time of the final depressurization at 212°F. With current methodology, all of the secondary energy remaining after the intermediate equilibration is conservatively assumed to be released by imposing a mandatory cooldown and subsequent depressurization down to atmospheric pressure at 3600 s, i.e., 14.7 psia and 212°F.

Sources of Mass and Energy The sources of mass considered in the LOCA mass and energy release analysis are given in tables 6.2-12, 6.2-44, and 6.2-49. These sources are the RCS, accumulators, and pumped SI.

The energy inventories considered in the LOCA mass and energy release analysis are given in tables 6.2-13, 6.2-45, and 6.2-50. The energy sources are listed below.

1. RCS water.

2. Accumulator water (all three inject).

3. Pumped SI water.

4. Decay heat.

5. Core stored energy.

6. RCS metal (includes SG tubes).

7. SG metal (includes transition cone, shell, wrapper, and other internals).

8. SG secondary energy (includes fluid mass and steam mass).

9. Secondary transfer of energy (feedwater into and steam out of the SG secondary).

The energy reference points are as follows:

1. Available Energy: 212°F; 14.7 psia.

2. Total Energy Content: 32°F; 14.7 psia.

The mass and energy inventories are presented in these balance tables at the following times, as appropriate:

1. Time zero (initial conditions).

2. End of blowdown time.

3. End of refill time.

4. End of reflood time.

5. Time of broken loop SG equilibration to pressure setpoint.

6. Time of intact loop SG equilibration to pressure setpoint.

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FNP-FSAR-6

7. Time of full depressurization (3600 s).

In the mass and energy release data presented, no Zirc-water reaction heat is specifically presented because the clad temperature does not rise high enough for the rate of the Zirc-water reaction heat to be of any significance.

The sequence of events for the LOCA mass and energy release analysis are shown in tables 6.2-51 through 6.2-53.

Conclusions The consideration of the various energy sources in the long-term mass and energy release analysis provides assurance that all available sources of energy have been included in this analysis. Thus, the review guidelines presented in Standard Review Plan Section 6.2.1.3 have been satisfied. The results of this analysis were provided for use in the containment integrity analysis.

Short-Term LOCA Mass and Energy Releases Purpose An evaluation was conducted to determine the effect of a MUR program on the PU/RSG program results for short-term LOCA-related mass and energy releases that support subcompartment analyses discussed in section 6.2 of the Farley FSAR. From the FSAR, a double-ended circumferential rupture of the reactor coolant cold leg forms the basis for the steam generator compartments, a 100 in2 reactor vessel inlet break forms the basis for the reactor cavity region, and both a spray line break and a surge line break were considered for the pressurizer compartment. This evaluation addressed the impact of MUR and other relevant issues on the current licensing basis for these four breaks.

Discussion and Evaluation The subcompartment analysis is performed to ensure that the walls of a subcompartment can maintain their structural integrity during the short pressure pulse (generally < 3 s) which accompanies a high-energy line pipe rupture within the subcompartment. The magnitude of the pressure differential across the walls is a function of several parameters, which include the blowdown M&E release rates, the subcompartment volume, vent areas, and vent flow behavior.

The blowdown M&E release rates are affected by the initial RCS temperature conditions. Since short-term releases are linked directly to the critical mass flux, which increases with decreasing temperatures, the short-term LOCA releases would be expected to increase due to any reductions in RCS coolant temperature conditions. Short-term blowdown transients are characterized by a peak mass and energy release rate that occurs during a subcooled condition; thus the Zaloudek correlation, which models this condition, is currently used in the short-term LOCA mass and energy release analyses with the SATAN computer program.

This correlation was used to conservatively evaluate the impact of the changes in RCS temperature conditions due to MUR on the short-term releases. The evaluation concluded that 6.2-18 REV 31 10/23

FNP-FSAR-6 the RCS temperature conditions due to the PU/RSG program were bounding so the current analysis of record short-term LOCA mass and energy releases are also bounding.

Any changes in RCS volume, SG liquid/steam mass and volume, and differences in units have no effect on the releases because of the short duration of the postulated accident. Any volumetric changes are small and have no impact on the subcompartment model. Therefore, the only change that needs to be addressed for this program is the decreased RCS coolant temperatures.

For this MUR, a RCS pressure of 2250 psia, a vessel outlet temperature of 605.0°F, and a vessel/core inlet temperature of 530.6°F were considered. Considering the temperature uncertainty of 6°F and the current licensing basis initial conditions, the following RCS temperature ranges were used for the evaluations herein:

Vessel Outlet 616.9°F to 597.8°F

Vessel/Core Inlet 563.1°F to 524.6°F A pressure uncertainty of 50 psi was also included in the evaluation.

Based upon the results of the MUR evaluation, the current design basis LOCA-related mass and energy releases, including the spray line and the surge line releases, would be bounded by the 1.18 factor that was determined for the PU/RSG program over the original design basis mass and energy releases due to RCS temperature effects.

Per reference 13, Farley is approved for LBB. LBB eliminates the dynamic effects of postulated primary loop pipe ruptures from the design basis. This means that the current breaks (a double-ended circumferential rupture of the reactor coolant cold leg break for the SG compartments and a 144-in2 reactor vessel inlet break for the reactor cavity region) no longer have to be considered for the short-term effects. Since the RCS piping has been eliminated from consideration, the large branch nozzles must be considered for design verification. This includes the surge line, accumulator line, and the RHR line. These smaller breaks, which are outside the cavity region, would result in minimal asymmetric pressurization in the reactor cavity region. Additionally, compared to the large RCS double-ended ruptures, the differential loadings are significantly reduced. For example, the peak break compartment pressure can be reduced by a factor of > 2, and the peak differential across an adjacent wall can be reduced by a factor of > 3 if the nozzle breaks are considered. Therefore, since Farley is approved for LBB, the decrease in mass and energy releases associated with the small RCS nozzle breaks, as compared to the larger RCS pipe breaks, more than offsets the increased releases associated with decreased RCS initial coolant temperatures. The SG and pressurizer subcompartments were re-analyzed for PU/RSG using the pressurizer surge line increased mass and energy blowdown data. The current licensing basis subcompartment analyses that consider breaks in the RCS remain bounding.

6.2-19 REV 31 10/23

FNP-FSAR-6 Results and Conclusions The short-term LOCA-related mass and energy releases discussed in section 6.2 of the Farley FSAR have been reviewed to assess the effects associated with the Farley Steam Generator Replacement Project. Results show that the current design basis spray line and surge line releases would increase 18% from the original design basis analysis due to RCS temperature effects. Therefore, the mass rates (lb/s) in FSAR tables 6.2-16 and 6.2-17 have been multiplied by 1.18. The results of this evaluation were used in the steam generator and pressurizer subcompartment structural analyses and are shown in table 6.2-22. Since Farley is approved for LBB, the decrease in mass and energy releases associated with the smaller RCS nozzle breaks, as compared to the larger RCS pipe breaks, more than offsets the increased releases associated with decreased RCS initial coolant temperatures. The current licensing basis subcompartment analyses that consider breaks in the primary loop RLS piping (i.e., SG subcompartment and reactor cavity region), therefore, remain bounding.

Evaluation of Impact on LOCA Mass and Energy Release Analysis of Closing MOV8887A/

MOV8887B and MOV8888B/MOV8888A During Quarterly RHR Pump Inservice Testing During the performance of quarterly RHR pump inservice testing, MOV8887B and MOV8888A are closed when the A RHR pump is tested, and MOV8887A and MOV8888B are closed when the B RHR pump is tested. The closure MOV8887A/MOV8887B and MOV8888B/MOV8888A affects both trains of RHR, since the ECCS flow analysis assumes that this flow path is open to provide an even flow distribution to all three cold legs regardless of which RHR pump is operating.

The LOCA mass and energy (M&E) release analysis assumes the failure of an emergency diesel generator, which results in the loss of one train of ECCS (one charging/HHSI pump and one RHR/LHSI pump), one containment spray train, and one containment cooling train. The LOCA M&E release analysis assumes the minimum flow provided by one train of ECCS (one charging/HHSI pump and one RHR/LHSI pump). The RHR pump that is tested is declared inoperable during quarterly RHR pump testing and the Technical Specification Actions are entered. No additional failures are required to be considered when the Actions are entered, since the Actions only allow operation to continue for the duration of the Completion Time, prior to restoring the affected equipment to operable status. Therefore, the reduced ECCS flows available during the testing of an RHR pump are based on two charging pumps and one RHR pump.

The ECCS flow that would be provided during RHR pump testing is approximately 2.6 lbm/s less than the ECCS flow that is assumed in the current design basis LOCA M&E analysis. The ECCS flow reduction results in a penalty (increase) of approximately 0.03% in the integrated energy released during the first hour of the transient. An increase of this magnitude is judged to be insignificant with regard to the containment response during the transient. Additionally, both trains of containment spray and containment cooling systems would be available to further reduce the containment response during the transient.

Therefore, it can be concluded that there is no adverse impact on the long-term LOCA M&E releases and containment integrity analyses. The short-term LOCA M&E releases are not impacted since the ECCS is not actuated during the short-term transients.

6.2-20 REV 31 10/23

FNP-FSAR-6 NRC Acceptance of PWROG-17034-P-A: Evaluation of the WCAP-10325-P-A Westinghouse LOCA Mass and Energy Release Methodology PWROG-17034-P-A (reference 29) was submitted to the NRC to justify the continued use of the previously approved LOCA M&E analysis methodology contained in the WCAP-10325-P-A (reference 7). The justification was needed as the properties of the reactor coolant system material used in WCAP-10325 were found to have non-conservatively lower values compared to the current American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. In order to justify the continued use of WCAP-10325, the PWROG performed sensitivity analysis with and without changing material properties, and by comparing the resulting M&E and containment response analysis with the analysis of another NRC approved M&E methodology which did have the correct material properties. The NRC has accepted this topical report (reference 28, ADAMS Accession No. ML20133P693), and it is now incorporated into the Farley Unit 1 and Unit 2 LOCA Containment Pressure/Temperature Response Analysis of Record.

Nuclear Safety Advisory Letter (NSAL) Incorporation NSAL-14-2, NSAL-11-5, and NSAL-06-6 are incorporated in the mass and energy releases of the limiting cases (DEPSG MIN ESF and DEHLG). The containment response for the limiting cases is re-analyzed with the mass and energy releases incorporating the NSALs.

6.2.1.3.4.2 Containment Pressure Analysis. The containment pressure analyses to determine the limiting LOCA were performed using the GOTHIC computer program that was developed for the purpose of transient analysis of atmospheres in multicompartment containments of water-cooled nuclear power plants. The use of GOTHIC for FNP containment pressure analyses was reviewed and approved by the NRC (reference 26).

The GOTHIC model predicts both the pressure and temperature within the containment regions and the temperatures in the containment structures. It is assumed that separate blowdown and core thermal behavior studies have been made by the nuclear steam supply system (NSSS) manufacturer to determine mass and/or energy input rates from sources such as: the release of reactor coolant, chemical reactions, and decay energy and sensible heat release which may cause heating or boiloff of residual water in the reactor vessel or superheating of steam as it passes through the reactor system and enters the containment through the postulated point of RCS rupture.

The GOTHIC model treats the containment and the heat transfer surfaces following a LOCA.

Included in this model are ESF and analytical techniques to enable calculation of their effects upon the containment. Several options have been incorporated in the model to facilitate use of these features.

GOTHIC calculates a pressure-time transient with stepwise iteration between the thermodynamic state points. The iterations are based on the laws of the conservation of mass and energy together with their thermodynamic relationships. Superposition of heat input functions is assumed so that any combination of coolant release, decay heat generation, and 6.2-21 REV 31 10/23

FNP-FSAR-6 sensible heat release can be used with appropriate ESF features to determine the containment pressure-time history associated with a LOCA.

The program uses a three-phase containment model consisting of the containment atmosphere (vapor phase), the sump (liquid phase), and falling drops (drop phase). Mass and energy are transferred between the regions by boiling, condensation, or liquid dropout. Evaporation is limited to 8% of the condensed steam. Heat transfer between the sump liquid and atmosphere vapor regions is modeled using the GOTHIC internal interfacial heat transfer model. Heat transfer in this mode is very small, so the inclusion of this mode of heat transfer is of no significance to the results of the analysis. Each region is assumed homogeneous, but a temperature difference can exist between regions. Moisture condensed in the vapor region during a time increment will agglomorate into drops which will fall into the liquid region.

Noncondensible gases are included in the vapor region. Thermodynamic state points of steam in the saturated and superheated state are based on models developed for the COBRA-NC and COBRA-TRAC codes. The models are described in reference 31.

6.2.1.3.5 Special Pressure Reduction Containment Concepts This section is not applicable.

6.2.1.3.6 Long-Term Containment Performance The results of the long-term limiting LOCA case (DEPSG) have been evaluated to verify the ability of the ECCS and containment heat removal system (CHRS) to keep the reactor vessel flooded and maintain the containment below the design conditions following a LOCA, as described in paragraphs 6.2.1.3.12, 6.2.1.3.13, and 6.2.1.3.14. This evaluation has been based upon conservatively assumed performance of the ESFs. The containment fan cooling units are actuated at 21.7 psia (Containment High-1 Safety Analysis Limit). Since High-1 occurs prior to diesel start, a 115-s delay is assumed from the start of the accident. This 115-s delay includes 15 s for LOSP signal generation and diesel start, a 2-s High-1 signal delay, and 98 s for the containment cooler service water inlet motor-operated valve (MOV) to stroke fully open. The containment spray system is actuated at 44.7 psia (Containment High-3 Safety Analysis Limit) with a 62-s delay from the start of the accident since High-3 also occurs before the fastest expected diesel loading. This time includes 15 s for LOSP signal generation and diesel start, 12 s for the spray discharge MOV to stroke open, and 35 s to fill the spray discharge header.

The ESF sequencer and spray pump start times are bounded by the MOV stroke time.

The residual decay heat rate is shown in figure 15.1-6 as watts per watt at maximum NSSS power level, 2785 MWt. Sensible heat remaining in the primary and secondary systems at the end of the post reflood failure is added to the reactor vessel water, as listed in table 6.2-14.

These criteria aid in assuring a conservative prediction of the third containment pressure peak, which occurs during sump water recirculation, and demonstrate the ability of the CHRS to maintain pressure effectively at a fraction of that value occurring during blowdown.

The containment pressure time response for the limiting LOCA case out to 105 s is shown in figure 6.2-2 for the minimum safeguards performance mode outlined in table 6.2-5. The 6.2-22 REV 31 10/23

FNP-FSAR-6 maximum pressure of 45 psig occurs at 876 s for the DEPSG break. A DEPSG break initial peak pressure is followed by a decrease in containment pressure during refill. At 2139 s, SI water from the sump begins to recirculate as the RWST reaches low level. The containment continues to depressurize until 4316 s, when containment spray water begins to recirculate from the sump as the RWST reaches low-low level. During sump water recirculation, a third pressure peak occurs due to steam evolution from the reactor because of boiloff of the hotter core injection water. The containment atmosphere and sump temperatures versus time are given in figure 6.2-41 for the DEPSG. The peak atmosphere temperature of 265°F occurs at 875 s. The maximum sump water temperature of 259°F occurs at 1370 s. The DEHL break has a peak atmosphere temperature of 264°F at 18.2 s.

The temperature response of the containment structure is shown in figure 6.2-78 at several points in time for the limiting LOCA. The containment wall liner plate reaches a maximum temperature of approximately 255°F at 1000 s.

6.2.1.3.7 Accident Chronology A chronology of events with time for the limiting LOCA case (DEPSG) is given in table 6.2-21 from the time of pipe rupture to 2.6 x 106 s when accident calculations were terminated. At that time, the containment pressure is 5.4 psig. It is assumed that time equals zero at the start of the LOCA.

6.2.1.3.8 Energy Balance An energy balance for the limiting LOCA is given in table 6.2-45.

6.2.1.3.9 Post-LOCA Parameters This section contains plots of various post-LOCA parameters as a function of time. The heat generation rate from core decay heat is shown in figure 15.1-6. The heat removal rate from the RHR heat exchanger and from the containment air cooler is shown in figures 6.2-79 and 6.2-80, respectively. The containment pressure/temperature vs. time profiles are shown in figures 6.2-1 through 6.2-5 and figure 6.2-40. Refer to paragraph 6.2.1.3.14 for subsequent evaluations regarding post-accident containment performance.

6.2.1.3.10 Containment Subcompartment Analysis The NRC acceptance criteria associated with compartment design are based on their acceptance of the analysis techniques described in this section which demonstrate that compartment calculated peak pressures for postulated pipe ruptures do not exceed compartment design pressures. Postulated breaks in the reactor coolant loop (RCL), except for branch line connections, have been eliminated from the structural design basis for both Unit 1 and 2, as allowed by the revision of GDC-4. The elimination of these breaks is a result of application of LBB technology.

6.2-23 REV 31 10/23

FNP-FSAR-6 A. Steam Generator Compartments Short-term differential pressurization across the secondary shield walls following a LOCA in the SG compartments was calculated with the Bechtel computer program COPDA. This program is described in attachment D of appendix 3K.

The compressible flow option was used.

The block diagram for the break in SG C compartment is shown in figure 6.2-53.

All major obstructions such as columns, tanks, and the SGs are accounted for in the volume and vent calculations, and the results are reduced by 10 percent to allow for minor obstructions such as cable trays and small pipes. Flow through the reactor cavity is neglected. Flow coefficients are calculated as for long passageways, since in each case L/D > 2. Entrance, exit, head, and friction losses are included. Each loss factor at cross-sectional areas greater than the minimum are scaled to the minimum by the square of the area ratio. Thus, 2

A min K total = i K i Ai Then the flow coefficient is calculated by 1

C=

K total The peak differential pressures for the cold leg breaks are listed in table 6.2-22.

The highest pressure (33.9 psid) is in the C compartment, since the refueling canal support and the incore instrumentation tunnel extension impede the outflow. The differential pressure for this case is plotted in figure 6.2-54. The blowdown mass and energy flowrates are listed in table 6.2-10. All pressures are below the allowable, as shown in table 6.2-22.

The hot leg break in the C compartment was also analyzed. Since the resulting peak is substantially lower than that for the cold leg break, the other compartments were not analyzed. The pump suction break yields lower flowrates than the cold leg breaks, so these were not analyzed either.

Application of LBB to the SG subcompartments eliminates the large RCS loop breaks. SG subcompartment C was reanalyzed for the RSG conditions with the pressure surge line mass and energy blowdown data provided in table 6.2-17. The results of this reanalysis confirm that the original analyses remain bounding, as shown in table 6.2-22.

6.2-24 REV 31 10/23

FNP-FSAR-6 B. Reactor Cavity The breaking of the RCL at the reactor nozzle will result in differential loadings on the reactor vessel caused by a discharge of steam and liquid at the postulated breakpoint. The mass rate of flow of the discharge fluid will be a function of:

1. The opening area of the break.

2. The internal energy of the discharging fluid.

The loading imposed on the vessel as a result of the break will be influenced by:

1. Vent openings from the break compartment to the reactor cavity compartments.

2. Thrust loads imparted by the fluid discharge from the nozzle at the postulated breakpoint

3. Vent areas from the break compartments to the containment in a direction away from the reactor cavity

4. The transient movement of vapor from the break as it circumvents the reactor vessel.

Opening area of the postulated break:

The double-ended guillotine break occurs at the weld connecting the cold leg pipe to the nozzle. Pipe movement is restrained so that the break area is < 100 in.2 A baffle plate is attached to the cavity annulus. Thus, most of the flow is through the inspection holes, which are open under normal operating conditions.

The blowdown area is then a function of pipe movement, vessel movement, and reactor coolant pump movement. Blowdown area is effectively minimized by providing a pipe restraint on all the legs of the RCL, located at a point as close as possible to the postulated breakpoint. The calculated discharge area accounting for vessel movement, pump movement, and restrained pipe whip is 100 in2 or less. The calculated discharge area accounting for the relative motion of the broken pipe end as determined from the reactor pressure vessel and RCL structural analyses is 85 in2. (See appendix 3M.)

See drawings D-176275, D-176277, D-176278, D-176279, D-206275, D-206277, D-206278, and D-206279 for a physical layout describing the cavity area. For the physical layout of the area under the reactor, see drawings D-176107 and D-206107.

Internal energy of the fluid:

The internal energy of the fluid was determined by analyzing the thermodynamic properties of the entire reactor loop as it exists at the time of the postulated break. A cold leg break was found to be the most severe case for fluid discharge and was therefore used in the pressurization analysis. In addition, a break area of 100 in.2 was used to determine the blowdown data, thereby making the data 6.2-25 REV 31 10/23

FNP-FSAR-6 conservative when compared to the actual calculated opening of 85 in.2. The blowdown data for a 100 in.2 break are listed in table 6.2-15.

Thrust loads imparted by the fluid discharge from the nozzle at the postulated break point:

These loads were determined by the rate of fluid discharge and the energy released during discharge.

Vent areas from the break area to the containment:

Venting from the break compartment to the inspection plug opening was considered as a factor in the final calculations. The inlet nozzle at 215° was selected as the break location because that penetration has the smallest inspection plug opening, approximately 2033 in.2 Selection of this location, when coupled with cold leg blowdown data, presents the most conservative case for analysis of reactor cavity pressure response.

Transient movement of vapor from the break compartment as it circumvents the vessel:

Loads are imparted on the reactor vessel by virtue of the fact that vapor movement around the vessel is not instantaneous.

Cavity Model:

The reactor cavity was subdivided into compartments at all significant flow restrictions, such as hot and cold leg nozzles. Compartment 1 is below the nozzle centerline in the penetration at 215° and is outside the blowdown restrictor plate. Compartment 2 is directly above the nozzle centerline and includes the inspection plug opening to the containment. Blowdown is assumed to split evenly between compartments 1 and 2.

The annulus compartmentalization is illustrated in figure 6.2-50. In this model, insulation in the broken leg penetration is assumed to blow off and completely plug the penetration at the pipe restraint as well as the vessel support ventilation duct. All gaps in the broken leg blowdown restrictor plate remain unobstructed throughout the transient to allow maximum pressurization of the cavity. In all other areas, insulation is assumed to remain in place and uncrushed, and the insulation volume is deducted from compartment volumes and vent areas. A summary of all compartment volumes is provided in table 6.2-22.

Flow coefficients are calculated as for the SG compartments (paragraph A.

above). Table 6.2-18 summarizes individual loss components for each flowpath as well as effective vent areas and calculated flow coefficients. Figure 6.2-51 is a block diagram of the reactor cavity model. It illustrates all compartment net free volumes, flow paths, and related vent areas and flow coefficients.

6.2-26 REV 31 10/23

FNP-FSAR-6 The Bechtel computer code COPDA was used for the analysis. (See attachment D, appendix 3K.) The program uses a Moody multiplier of 0.6.

Moody flow formulations were used except when flow was subcritical, in which case homogeneous frozen flow equations were applied.

Results:

Maximum and design pressures and time to peak are listed for all compartments in table 6.2-22. A two-dimensional axisymmetric finite element analysis was done to check the acceptability of the reactor cavity under the nonuniform differential pressure loading, and the results verify that allowable loads are not exceeded. All other subcompartments are also below the design pressure limit. The time-dependent pressure histories for all nodes are shown graphically in figures 6.2-58 through 6.2-65. The net pressure vessel loading history is illustrated graphically in figure 6.2-52. The asymmetric pressure results in a maximum horizontal loading of 1.2 x 106 lb and a maximum vertical loading of 2.12 x 104 lb on the reactor pressure vessel.

Nodalization Sensitivity Study:

A nodalization sensitivity study was performed to determine the minimum number of nodes required to conservatively predict the maximum pressure in the reactor cavity. A base case and two additional models were analyzed. The mass and energy release for all three cases of the sensitivity study correspond to a 144 in.2 break in the loop 3 cold leg.

The nodalization in the base case, which is a 31-compartment model, was indentical to the model shown in figure 6.2-50 except that compartment 31 was combined with compartment 9, compartment 32 was combined with compartment 10, and compartment 33 was combined with compartment 8.

In model B, a 27-compartment model, nodes were combined directly above and below the nozzles between 25° and 145° to form single compartments. This decreased the number of nodes both circumferentially and axially. In model C, a 38-compartment model, this region was modeled with twice the number of nodes used in the base case model. In addition, compartment 3 (adjacent to the blowdown restrictor plate inside the cavity) was divided into four subcompartments. Blowdown from compartments 1 and 2 enters the annulus through these nodes.

For the base case, the maximum horizontal force was calculated to be 1.4 x 106 lbs and the maximum uplift force was calculated to be 5.9 x 104 lb. For both models B and C, the results show the maximum horizontal force to be 1.42 x 106 lb, or within 1.5% of the base case. Maximum uplift force for both cases was approximately 6.0 x 104 lb. Based on these results it is concluded that the compartment boundaries should properly be placed at flow restrictions and the addition of arbitrary compartments will not significantly affect the results.

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FNP-FSAR-6 C. Pressurizer The pressurizer compartment has been analyzed for the double ended rupture of any line in this compartment. The pressurizer spray line break is the controlling one, the rupture of which results in a peak compartment pressure differential of 9.4 psid. The blowdown data for the spray line break are listed in table 6.2-16.

The blowdown model and flow model are shown in figures 6.2-55 and 6.2-56.

The results are shown on figure 6.2-57.

The pressurizer surge line extends through an open penetration in the floor of the pressurizer compartment to the bottom of the pressurizer, and is enclosed by the pressurizer compartment volume.

The blowdown from a break in the pressurizer surge line, given by table 6.2-17, at the pressurizer nozzle will be vented directly into the lower compartment. This lower compartment is designed to withstand the differential pressures resulting from the DBA.

Postulated breaks in the pressurizer surge line have been eliminated from the structural design basis for both Unit 1 and 2 through the application of LBB technology.

D. Secondary Shield Annulus The steam lines penetrate the containment vessel at el 138 ft, which is below the top of the secondary shield walls. An analysis was made to determine if local high pressures could occur in the relatively confined space. Steam flows around the annulus, upward and directly to the containment upper volume.

At no time do differential pressures across the walls exceed 0.25 psid.

6.2.1.3.11 Main Steam Line Ruptures Inside Containment A. Introduction Steam line ruptures occurring inside the containment structure may result in significant pressure and temperature transients. In order to determine the rupture which results in the worst case, a complete spectrum of ruptures was analyzed for power uprate. This extensive analysis was necessary due to the number of variables on the determination of the blowdown data. The five most limiting power uprate containment pressure and temperature response cases were reanalyzed for RSG and the measurement uncertainty recapture (MUR) power uprate (up to 2841 MWt, which includes 10 MWt RCP heat). These five cases span 0 to 100% nominal full-load MUR uprated power and three break sizes as noted below. A summary of the plant particular data utilized for this analysis is given in table 6.2-11.

6.2-28 REV 31 10/23

FNP-FSAR-6 After the blowdown was determined, a case-by-case containment analysis was performed using the GOTHIC computer code. As described in paragraph 6.2.1.3.12, GOTHIC is a containment pressure/temperature transient analysis code. A more detailed description is contained in reference 22.

The LOCA model heat sinks were used in the containment analyses of steam line ruptures. The pertinent information and data for these are given in tables 6.2-2 and 6.2-4.

The following list of ruptures was analyzed:

Case 1: Full double-ended rupture at 100% MUR uprated power.

(Reanalyzed for RSG at -1.5- and +3.0-psig initial pressure.)

Case 2: 0.7 ft2 double-ended rupture at 102% power.

Case 3: 0.6 ft2 double-ended rupture at 102% power, at both 0 and

-1.5-psig initial pressure.

Case 4: 0.528 ft2 split rupture at 102% power.

Case 5: Full double-ended rupture at 70% power.

Case 6: 0.6 ft2 double-ended rupture at 70% power.

Case 7: 0.5 ft2 double-ended rupture at 70% power.

Case 8: 0.22 ft2 split rupture at 70% MUR uprated power.

(Reanalyzed for RSG with a 0.22 ft2 at -1.5- and +3.0-psig initial pressure.)

Case 9: Full double-ended rupture at 30% MUR uprated power.

(Reanalyzed for RSG at -1.5- and +3.0-psig initial pressure.)

Case 10: 0.5 ft2 double-ended rupture at 30% power.

Case 11: 0.4 ft2 double-ended rupture at 30% power.

Case 12: 0.33 ft2 split rupture at 30% MUR uprated power.

(Reanalyzed for RSG with a 0.33 ft2 rupture at -1.5- and

+3.0-psig initial pressure.)

Case 13: Full double-ended rupture at hot standby. (Reanalyzed for RSG at -1.5- and +3.0-psig initial pressure.)

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FNP-FSAR-6 Case 14: 0.2 ft2 double-ended rupture at hot standby.

Case 15: 0.1 ft2 double-ended rupture at hot standby.

Case 16: 0.3 ft2 split rupture at hot standby.

All steam line rupture cases indicated with an asterisk (*) have been analyzed in support of the SG replacement and the MUR power uprate at both Farley Nuclear Plant units. These represent the limiting cases with respect to the containment pressure and temperature responses. The other 11 cases were analyzed in support of the power uprate for Farley Nuclear Plant. A discussion of the steam line rupture mass and energy releases is presented in FSAR paragraph 6.2.1.3.11, item B.

B. Mass and Energy Releases Following a Main Steam Line Rupture Inside Containment The mass and energy release data for each case are determined using the methodology documented in reference 23. The LOFTRAN code (reference 24) was used in the analysis documented for the Farley Nuclear Plant power uprate.

In support of the SG replacement and the MUR uprated power, the analysis methodology for the limiting 5 cases (see FSAR paragraph 6.2.1.3.11, item A) was supplemented using the model documented in reference 25.

C. Evaluation of Effects of Various Single Failures The method of determining the blowdown assumes no failure in steam or feedwater isolation. This blowdown is used in conjunction with minimum containment spray and fan coolers to allow for failure of a diesel generator.

Steam line isolation failure is not postulated, since there are two redundant swing disc trip valves in each steam line. However, since these valves stop flow only in the forward direction, the mass/energy release to containment as calculated by the LOFTRAN code (reference 24) was modified to include the entire steam piping volume downstream of the isolation valves for the other SGs, including the steam line header and steam dump piping. For the steam line rupture cases analyzed in support of the SG replacement, those indicated with an asterisk (*) in FSAR paragraph 6.1.2.3.11, item A, the mass/energy released to containment as calculated by the RETRAN model (reference 25) includes the contents from the main steam header piping downstream of the postulated break location.

For a complete description of the functions that provide the necessary protection against a steam pipe rupture refer to FSAR paragraph 15.4.2.1.1, items A through D.

D. Pressure Temperature Results The pressure/temperature results of the analysis are illustrated in figures 6.2-6 through 6.2-37. The highest pressure obtained was 53.4 psig for a full 6.2-30 REV 31 10/23

FNP-FSAR-6 double-ended rupture at 0-percent power. The highest temperature reached was 367°F for a double-ended rupture at 102-percent power.

Equipment Temperature Transient A transient main steam line break (MSLB) containment thermal analysis was performed for the worst case of the spectrum of breaks for each type of Class 1E component inside containment to determine the peak component surface temperatures. The methodology used to calculate the equipment surface temperatures was based on the NRC staff's approved assumptions discussed in Appendix B of NUREG-0588. These analyses show that the peak surface temperature resulting from the MSLB environment do not exceed the qualification temperature for each type of component. The methodology and results of the calculations of the equipment surface temperatures are reported in the docketed Farley response to NUREG-0588 and I&E Bulletin 79-01B.

The above results show that the maximum pressure in the containment is below the containment design pressure at 54 psig. In addition, the components covered by I&E Bulletin 79-01B and NUREG-0588 and required for safe shutdown and accident mitigation maintain their environmental qualification for the resulting temperature and pressure profiles inside the containment as determined by the above analysis.

E. The containment design meets the NRC acceptance criteria contained in IE Bulletin 80-04 related to the issue of containment overpressurization resulting from a main steam line rupture with continued feedwater addition. Considering all possible sources of water, there is no potential for containment overpressurization because the main feedwater system is isolated and auxiliary feedwater system flow restrictors limit flow to the affected SG. Also, the auxiliary feedwater system pumps are protected from the effects of runout flow and, therefore, can be expected to carry out their intended function during a main steam line rupture event.

6.2.1.3.12 Reduced Service Water Flow Containment Response Evaluations of the service water system have indicated that under certain conditions for component failures and loss of offsite power (LOSP), the service water flow to the containment air coolers may be reduced below the original design basis flow. The effect of this reduction in service water supply to the containment air coolers was included in the LOCA analysis previously described.

6.2.1.3.13 Containment Pressure/Temperature (P/T) Evaluations Subsequent to Power Uprate As a result of calculating new RWST level uncertainties, evaluating net positive suction head for the ECCS/CSS pumps, and evaluating changes to the ECCS/CSS switchover sequence (from 6.2-31 REV 31 10/23

FNP-FSAR-6 injection to sump recirculation), several parameter changes have been evaluated to determine the impact on the P/T analysis. The parameter changes which were evaluated include revised RWST deliverable volumes, RHR interruption times, time of ECCS/CSS switchover to sump recirculation, and increased CS flowrate.

The changes to RWST delivered volumes, RHR interruption times, and time of ECCS/CSS switchover affect only the long-term portion of the LOCA cases (MSLB is not affected).

Additionally, increased spray flowrates do not adversely affect the containment P/T analysis since minimum spray flow is conservative. Since the LOCA peak pressure occurs before ECCS/CSS switchover and the parameter changes only affect the long-term portion of the transient, the LOCA peak pressure, as given in the power uprate containment P/T analysis, remained valid. However, for the long-term portion of the LOCA cases, increases in the pressure and temperature response were evaluated. The evaluation demonstrated the acceptability of the results.

Re-analysis for the NSAL incorporation uses conservative values for switchover to recirculation, and the long-term effects on pressure and temperature are accounted for in the limiting cases that were re-analyzed (Table 6.2-6 and Figures 6.2-2 and 6.2-41).

6.2.1.3.14 Containment Pressure/Temperature Evaluations Subsequent to Steam Generator Replacement An evaluation was performed for the revised heat sinks in table 6.2-2. More recent information from the vendor also resulted in minor changes to coating thermodynamic properties as reflected in table 6.2-4. Additional changes involve the actuation delays associated with the containment spray and fan cooler ESFs. Table 7.3-16 has been appropriately revised to reflect the reassessed spray response time. The fan cooler delay of 115 s assumed in the containment pressure/temperature analysis is conservative with respect to the 87-s service water delay in table 7.3-16. The DEHL and DEPS LOCA cases, the limiting main steam line break cases (Cases 1 and 13), and a nonlimiting MSLB case (Case 9) were reanalyzed with these revised inputs. In order to offset adverse impacts from these changes, minimum spray flow was recalculated and increased from 2175 gal/min during injection and recirculation modes to 2480 gal/min and to 2290 gal/min in injection and recirculation modes, respectively. The LOCA limiting cases re-analyzed for NSAL incorporation include these changes. The MSLB peak pressures remained bounded by prior results. The peak MSLB temperature increased 0.2°F which is indiscernible in the transient figures. MSLB pressures were slightly higher early in the transients (~0.3 psi) due to revised spray and fan cooler response times but remained bounded by the LOCA cases at this point in the transient. The revised profiles were evaluated for impact on the Environmental Qualification program and it was determined that the existing P/T profiles remain valid. Figures for the reanalyzed breaks were revised. Based on the observed sensitivities, the remaining figures remain bounding or are negligibly affected.

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FNP-FSAR-6 6.2.1.4 Containment Testing and Inspection 6.2.1.4.1 Preoperation Testing 6.2.1.4.1.1 Integrated Test. Upon completion of the containment and installation of all penetrations, an integrated leakage rate test was performed to verify that the potential leakage rate from the containment is maintained within acceptable values.

The integrated leakage rate tests consist initially of a preoperational test at the peak calculated accident pressure of 48.0 psig, as well as at least one at a lower pressure.

The total allowable leak rate is not more than 0.15 percent by weight of the contained atmosphere per day at 48.0 psig. It has been demonstrated that with good quality control during construction, this is a reasonable requirement. The basis for the performance of the integrated leakage rate test is "Containment Structures for Nuclear Power Plants," Revision 1, November 1, 1972.

The initial leak rate test of the containment and its penetrations were conducted at 100 percent of peak calculated accident pressure and 50 percent of peak calculated accident pressure.

Values of containment ambient dry bulb temperature and relative humidity were recorded during the test period for correction of data as required. The test establishes the capability of the containment to contain the pressure for which it was designed, at a leak rate not exceeding that specified.

The test measurement system utilized for the initial leak rate test of the containment is a packaged portable unit designed for use on Units 1 and 2. This portable unit is also used for the periodic leak rate test. The instruments are calibrated prior to each periodic leak rate test. It is anticipated that these instruments will be available for the lifetime of the plant; however, spare instruments of each type are provided.

For further details, see the Technical Specifications.

6.2.1.4.1.2 Local Tests. Prior to initial startup, penetrations and isolation valves were leak tested to verify that the potential leakage is within acceptable values.

The local tests will be performed at a pressure of 48.0 psig and in accordance with the Technical Specifications.

6.2.1.4.2 Postoperational Leakage Tests Periodic leak rate tests of the containment, penetrations, and isolation valves are conducted to verify their continued leaktight integrity.

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FNP-FSAR-6 The first postoperational integrated leakage rate tests were conducted at 50 percent of peak calculated accident pressure while the penetrations and isolation valves postoperational tests were conducted at 48.0 psig.

Postoperational leakage tests are currently conducted in accordance with the Containment Leakage Rate Testing Program. The test frequencies and acceptance criteria are specified in this program.

The Containment Leakage Rate Testing Program complies with the NRC acceptance criteria contained in GDC 52, 53, and 54 and implements the requirements of 10 CFR 50 Appendix J, Option B; the guidance in Nuclear Energy Institute (NEI) 94-01, Revisions 2A and 3A with their associated limitations and conditions; and the technical methods and techniques for performing containment leakage tests in ANSI/ANS 56.8-2002.

The ILRT performance criterion (La) is 0.15 weight % per day. The NRC acceptance criteria regarding airlock leakage tests are also in the Containment Leakage Rate Testing Program.

6.2.1.4.3 Containment Materials Inspection, Testing, and Surveillance A. Tests to Ensure Liner Integrity The following tests were/are performed:

1. Construction tests during the erection of the containment liner.

2. Preoperational tests after the erection of the containment complete with liner, electrical and piping penetrations, equipment hatch, and personnel lock, but before reactor operation.

3. Postoperational leakage tests will be performed at periodic intervals for the life of the plant.

4. Inservice inspection of the metallic liner and the pressure retaining concrete structure of the containment will be performed at periodic intervals for the life of the plant as discussed in paragraph 3.8.1.7, Testing and Inservice Surveillance Requirements.

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

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

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FNP-FSAR-6

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

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

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

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

C. Radiography Radiography is used as an aid to quality control. The primary purpose of the liner plate and the welds therein is to provide leaktightness integrity to the posttensioned concrete containment. Structural integrity of the containment will be provided by the posttensioned concrete and not by the liner plate.

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

For quality control purposes, at least one spot radiograph 12 inches long was taken in the first ten feet of welding completed in the flat, vertical, horizontal, and overhead positions by each welder on liner plate welds. No further welding was permitted until initial radiographic inspection has been satisfactorily completed and the welding found to be acceptable by the Inspector.

Thereafter, a minimum of 2 percent of the welding was progressively spot examined as welding is performed, using film 12 in. long, on a random basis to be specified by the inspector, in such a manner that an approximately equal number of spot radiographs was taken from the work of each welder. In addition to the 2 percent radiograph, 18 percent 6.2-35 REV 31 10/23

FNP-FSAR-6 of the welding was nondestructively examined. Under conditions where two or more welders make weld layers in a joint or on the two sides of a double-welded butt joint, one spot examination represented the work for both welders. Where a radiograph discloses welding which did not comply with the minimum quality requirements, as defined in paragraph UW52,Section VIII of ASME code, two additional spots, each 12 inches long, were examined in the same weld seam at locations away from the original spot. The locations of these additional spots were determined by the Inspector as provided for the original spot examination. If two additional spots examined showed welding which met the minimum quality requirements, the entire weld represented by the three radiographs was acceptable. The defective welding disclosed by the first of the three radiographs was removed and repaired by welding.

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

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

D. Dye-Penetrant and Magnetic-Particle Inspection Dye-penetrant and magnetic-particle inspection were used as an aid to quality control.

The field welding inspectors used dye-penetrant or magnetic-particle inspection to closely examine welds judged to be of questionable quality on the basis of the initial visual inspection. Dye-penetrant or magnetic-particle inspection of liner plate welds were in accordance with Section VIII of the ASME Boiler and Pressure Vessel Code.]

6.2.1.5 Instrumentation Requirements The containment atmosphere is continuously monitored by four pressure transmitters outside the containment, which are connected to the containment atmosphere. The ESF actuation design details and logics associated with these pressure transmitters are discussed in section 7.3. Instrumentation also monitors the containment to identify any pressure boundary leakage in the RCS. The leakage detection system is discussed in subsection 5.2.7.

Airborne radiation inside the containment is monitored by the containment monitoring system, which is discussed in section 11.4.

Liquid level indication is provided in the containment sump, as discussed in section 7.3.

6.2.1.6 Materials Materials used in or on ESF systems were chosen so that any decomposition products of the materials will not interfere with safe operation of ESFs.

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FNP-FSAR-6 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS Three systems are provided to reduce containment atmosphere temperature and pressure and/or to remove heat from the containment under post-accident conditions. These are the low-head SI/RHRS, the containment spray system, and the containment cooling system. The design, operation, and reliability of the low-head SI/RHRS are discussed in section 6.3; the remaining two systems are discussed below.

6.2.2.1 Design Bases Design environmental conditions for the containment heat removal systems are given in table 3.11-1. Design parameters for system components are discussed in paragraph 6.2.2.2.

The sources and amounts of energy that were considered in sizing the heat removal systems are listed in table 6.2-20. The containment heat removal systems are in conformance with the NRC acceptance criteria contained in General Design Criteria 38, 39, and 40 and Regulatory Guide 1.1.

6.2.2.1.1 Containment Spray System The function of the containment spray system is to spray water into the containment atmosphere, when appropriate, in the event of a LOCA to ensure that containment peak pressure is below its design value. This protection is afforded for pipe breaks up to and including the hypothetical instantaneous double-ended circumferential rupture of an RCL as analyzed in subsection 6.2.1. Although the water in the reactor vessel after a LOCA is quickly subcooled by the ECCS, the containment spray system design was based on the conservative assumption that the core residual heat would be released to the containment as steam.

The two redundant trains of the containment spray system have been designed to provide sufficient heat removal capacity to prevent exceeding containment design pressure for all piping breaks. Assuming that the water in the RWST is at a temperature of 110°F, the containment atmospheric heat removal capability associated with the spray from one containment spray train will initially be 2 x 108 Btu/h at initial spray actuation during the injection phase.

The containment spray system is designed to operate over an extended period of time and under environmental conditions existing following a RCS failure.

6.2.2.1.2 Containment Cooling System The containment cooling system has been designed to remove heat which will be released to the containment atmosphere during any MSLB or LOCA up to and including the double-ended rupture of the largest system pipe. This is accomplished by one of four containment air coolers.

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FNP-FSAR-6 The experimental heat transfer data for the containment air coolers are based on the following documents which have been submitted to the NRC:

A. "Cooling Coil Thermal and Structural Capacity Evaluation for the Palisades Plant of Consumers Power Company," American Air Filter Company, Inc., Report R-1003 (Palisades Docket Number 50-255).

B. Topical Report on "Design and Testing of Fan Cooler/Filter Systems For Nuclear Applications," AAF-TR-7101 February 20, 1972.

Test requirements are certified by the cooler manufacturer.

6.2.2.2 System Design Design parameters for the heat removal system components are presented in table 6.2-24.

Individual system designs are discussed below.

6.2.2.2.1 Containment Spray System The containment spray system P&ID shown in D-175038, sheet 3 and D-205038, sheet 3 consists of two pumps, spray ring headers and nozzles, valves, and piping. During the initial (injection) phase of operation, water from the RWST is used for containment spraying. During the later (recirculation) phase of operation, water for containment spraying is recirculated from the containment sump.

Detection of leakage from the containment spray system involves the use of sump level instrumentation and floor drain tank level instrumentation. The spray pumps are in rooms that contain a sump. If the alarm in the control room indicates a high sump level, the sump contents would be pumped by duplex sump pumps to the waste holdup tank. Any leak in one of the pump rooms would be detected and the leak isolated.

A leak in the spray system piping in a corridor, pipe chase, or penetration room is indicated by an increase in the floor drain tank level.

Whenever one of the spray headers is isolated during the process of system leak detection and isolation, flow indication in the operating header assures the operator that sufficient spray flow is being returned to the containment.

The two redundant and independent containment spray trains that comprise the containment spray system, including required auxiliary systems, are designed so that a single active failure during the injection phase or a single active or passive failure during the recirculation phase following a RCS failure does not result in loss of the protective function.

The containment spray system is designed to accommodate the 1/2 safe shutdown earthquake (1/2 SSE) within applicable codes stress limits and to withstand the SSE without rupture or loss of function.

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FNP-FSAR-6 FNP was originally designed as follows: Between the edge of the containment mat and the auxiliary building, each containment spray suction line is surrounded by a concentric guard pipe.

Three annular seal rings are installed around the guard pipe at each end to minimize any postulated bypass leakage along the outside of the guard pipe. The concentric guard pipes enter the auxiliary building in a trench within the pipe penetration room. Therefore, any postulated airborne leakage that escapes outside the concentric guard pipes is processed by the penetration room filtration system, and any postulated water leakage is processed by the waste disposal system. Within the auxiliary building, each suction line is located within a protective chamber. This chamber is designed to withstand the containment design pressure in addition to the head of water present in the containment sump at the end of the injection phase.

From the protective chamber to the first motor-operated isolation valve inside the auxiliary building, each suction line is surrounded by a concentric guard pipe. The first motor-operated isolation valve is surrounded by a watertight closure. This arrangement, which is the same as that used with the suction lines in the low-head SI system, ensures that, in the unlikely event of leakage from the suction pipe during long term recirculation, the integrity of the recirculation system is not impaired and public safety is not hazarded.

The NRC subsequently approved removal of these protective chambers and portions of the concentric guard pipes from the containment spray and low-head SI fluid systems based on demonstration that the design of the piping and first isolation valve in each suction line precludes a breach of piping integrity, and the design of the plant provides the capability to detect and terminate leakage from the valve shaft or bonnet seals on these valves. The pipes enter the auxiliary building within the pipe penetration room. Therefore, any postulated airborne leakage that escapes from the valve shaft or bonnet seals on these valves is processed by the penetration room filtration system, and any postulated water leakage is processed by the waste disposal system. In addition, alarms in the main control room alert operators of leakage to allow identification of the source of the leakage and isolation of the leakage. These features, which are the same as that used with the suction lines in the low-head SI system, ensure that, in the unlikely event of leakage from the valve shaft or bonnet seals on these valves during long term recirculation, the leak can be isolated and public safety is not hazarded.

Adequate net positive suction head (NPSH) is available to the containment spray pump suctions at all times during both the injection and the recirculation phases of operation, assuming the most adverse combination of flowrate, sump water temperature, and containment pressure.

The margin between available and required NPSH during the recirculation phase is based upon conservative calculations that account for the head loss due to debris accumulation on the sump strainers. These calculations neglect any containment pressure in excess of the TS minimum operating containment pressure prior to the LOCA accident when the containment sump temperature is lower than the saturation temperature corresponding to the TS minimum operating containment pressure. For sump temperatures above the saturation temperature at the TS minimum operating containment pressure, containment pressure in excess of the saturation pressure corresponding to the sump water temperature is conservatively neglected.

The available NPSH for the containment spray pumps during the recirculation phase is provided in subsection 6.3.2.14.

Descriptions of the containment spray system components are presented below.

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FNP-FSAR-6 Refueling Water Storage Tank The RWST serves as a source of emergency borated cooling water for injection. It is normally used to fill the refueling canal for refueling operations. However, during all other plant operating periods, it is aligned to the suction of the residual heat removal pumps and the containment spray pumps. The capacity of the tank is 66,850 ft3. The tank is fabricated from stainless steel and is designed and constructed in accordance with Code ASME III, Class 2. Water in the tank is borated to a concentration which assures reactor shutdown by at least 10% k/k, when all RCC assemblies are inserted and the core cooled down for refueling.

Containment Spray Pumps The containment spray pumps are horizontal centrifugal type driven by electric motors. Each motor is powered from an individual emergency bus.

The pumps are designed to perform at rated capacity against a total head composed of containment design pressure, nozzle elevation head, and the line and nozzle pressure losses.

Pump design parameters are presented in table 6.2-24.

Pump room coolers are used to maintain air temperatures in the pump rooms at or below 104°F during normal operation. Refer to table 9.4-6A for post-DBA room temperatures. The cooler fans are interlocked to start with the respective pump motors. The Seismic Category I portion of the service water system is used as the cooling medium for the pump room coolers.

Containment spray pump room coolers are discussed in paragraph 9.4.2.1.9.

A gas accumulation monitoring and trending process for the Farley Unit 1 and 2 ECCS (CVCS and RHR) and containment spray systems has been established to meet the requirements of NRC Generic Letter 2008-01.

Spray Nozzles The spray nozzles, which are of the hollow cone design, are not subject to clogging by particles

< 1/4 in. in size, and produce a drop-size spectrum with a mean diameter of < 700 microns when operating at the design pressure differential of 40 psi. The spray nozzles are stainless steel and have a 3/8-in. diameter orifice, which is larger than the 3/32-in. openings in the Unit 1 and 2 post-LOCA containment sump strainers.

[HISTORICAL][For Unit 1 there are three 3/8-in. diameter holes between the solid cover plate on the top of the sump screen and the bioshield wall for venting of air during the initial phase of the LOCA when the water level in the sump rises. The slot size varies from approximately 1/4 in. to 1 in. across its length of approximately 3 ft. The potential for debris to enter through this path has been evaluated. The location of the slot near the shield wall was specifically selected to minimize the potential for debris to enter the sump. Since this slot and the vent holes will be under water during the recirculation phase of a LOCA, the debris entering through this path will sink to the sump floor due to low approach velocities near the bioshield wall and will not be swept into the opening of the intake pipe.] Therefore, the hydraulic performance of the CS system will not be adversely affected.

6.2-40 REV 31 10/23

FNP-FSAR-6 There is a vent slot and, for Unit 1, three 3/8-in.-diameter holes between the solid cover plate on the top of the sump screen and the bioshield wall for venting of air during the initial phase of the LOCA when the water level in the sump rises. The slot size varies from approximately 1/4 in. to 1 in. across its length of approximately 3 ft. The potential for debris to enter through this path has been evaluated. The location of the slot near the shield wall was specifically selected to minimize the potential for debris to enter the sump. Since this slot and the vent holes will be under water during the recirculation phase of a LOCA, the debris entering through this path will sink to the sump floor due to low approach velocities near the bioshield wall and will not be swept into the opening of the intake pipe. Therefore, the hydraulic performance of the CS system will not be adversely affected.

Containment Recirculation Sump See appendix 6D.

Motors, Valves, and Piping Motors, valves, and piping for the containment spray system are designed in accordance with the specifications for these components in the ECCS, as discussed in section 6.3.

Electrical Supply Details of the emergency bus power source are discussed in subsection 8.3.1.

6.2.2.2.2 Containment Cooling System The containment cooling system consists of four containment air coolers, each with a one-third cooling capacity during normal operation, up to four units will be operating. Each air cooler consists of a fan and finned tube coil supplied by water from the service water system. During outage periods, auxiliary cooling may be supplied to one of the four containment air coolers in Unit 2 to enhance containment heat removal.

As the post-accident containment atmosphere, which consists of a steam-air mixture, is circulated through the bank of cooling coils, it is cooled and a portion of the steam is condensed.

The capacity of one cooler in conjunction with one containment spray train is adequate to maintain pressure and temperature below peak calculated LOCA conditions.

Cooling surfaces are constructed in accordance with TEMA guidelines. Headers for coolers are designed and fabricated to the requirements of the ANSI B31.7. The service water system piping inside the containment is designed to ASME Section III.

Fan motors conform to standards of NEMA, IEEE, and ANSI. A supply prototype fan motor has demonstrated capability to supply design flow of steam air mixture through the cooling coils.

Dropout plates with release mechanisms actuated by fusible links are provided at the discharge of the containment coolers. These plates fall away to uncouple the cooler discharge from the distribution ductwork after a LOCA, thereby reducing discharge backpressure and allowing less restricted flow through the coolers. The fusible links used with the dropout plate release 6.2-41 REV 31 10/23

FNP-FSAR-6 mechanisms are constructed from suitable materials and are of a proven design shown to be reliable by extensive past commercial usage in critical service for fire protection. They are designed, manufactured, and tested in accordance with the applicable provisions of the following codes and standards:

Fire Protection Handbook.

Underwriters Laboratories' Standard for Fusible Links for Fire Protection Service, UL33-1968.

Underwriters Laboratories' Standard of Safety, UL555-1970.

Factory Mutual Approval Guide, 1971.

6.2.2.3 Design Valuation Descriptions of the analytical methods and models used to assess the performance capability of the containment heat removal systems are provided in subsection 6.2.1. Curves describing the calculated response of pertinent system and containment parameters following a LOCA are presented in subsection 6.2.1. Additional design evaluations are presented below.

6.2.2.3.1 Containment Spray System The containment spray system is designed to limit the effects of postblowdown energy additions to the containment during the injection phase in the event of a LOCA assuming no credit for core cooling by the ECCS. The accident analysis presented in subsection 6.2.1 assumes that containment cooling capability is reduced to one of two containment spray pumps and one out of four fan coolers. This is the minimum equipment available considering the single-failure criterion in the emergency power, the containment spray, and the containment cooling system.

The minimum fall path of water droplets is approximately 110 ft, which is the distance from the lowest spray ring to the operating deck. Detailed calculation of the heatup of spray drops in post-accident containment atmospheres by Parsly(2) show that drops of all sizes encountered in the containment spray reach equilibrium in a fraction of their residence time.

A failure analysis has been made on all active components of the system during the injection phase, and on all active and passive components during the recirculation phase, to show that the failure of any single component will not prevent fulfilling the design function. This analysis is summarized in table 6.2-26.

Concerning the containment sumps, there are several different flow paths which a particle might follow in circulating through the different systems that must use the sumps as a source. The maximum particle size that could be passed through the post-LOCA sump strainer screen is <

3/32 in. The different flow paths are:

A. A particle of sufficiently small size which has passed through one of the fine mesh screens may flow into one of the containment spray pump intakes, through 6.2-42 REV 31 10/23

FNP-FSAR-6 the spray pump, and be discharged through one of the spray nozzles into the containment. The particle may or may not pass through one of the containment spray eductors; however, the eductor size is sufficiently large for the largest particle to pass through. See paragraph 6.2.2.2.1 for a further description of the spray system components.

B. A particle of sufficiently small size which has passed through one of the fine mesh sump screens may flow into one of the low-head SIS pump intakes, through the low-head SIS pump, and through one of the residual heat exchangers; from there it may follow one of the following paths:

1. It may flow directly into the RCL (hot leg or cold leg).

2. It may flow into the suction of the high-head SIS pumps and through one of these pumps. From there, it may flow into the seal injection filter, which supplies seal water to the reactor coolant pumps and is considered an alternate SI path post LOCA; or it may flow directly into the RCL (hot leg or cold leg).

The effects of debris ingested through the containment sump strainer during operation of the containment spray system in recirculation mode were evaluated for Generic Letter 2004-02, Potential Impact of Debris on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors. The Generic Letter Response (including RAI responses and Supplemental Content Guide) includes evaluations of potential erosive wear, abrasion, and blockage of flow paths within containment spray system components. The Generic Letter 2004-02 evaluations conclude that system design is adequate to maintain containment spray function during recirculation mode operation.

Details of the design evaluations for the effects of debris on the function of the containment spray system are provided in Appendix 6D.

6.2.2.3.2 Containment Cooling System The air coolers, with their associated piping, valves, and instrumentation are located outside the secondary shield walls and above the maximum possible post-accident water level to provide protection against missiles and flooding.

Each containment cooler is equipped with a dropout plate, held shut during normal operation by fusible links, which disconnects the cooler discharge from the containment ductwork after a LOCA. During operation, the load on the fusible links is supplied by springs in order to minimize the effect upon the load of thermal growth of the release mechanism and any elongation of the fusible link just prior to separation. Positive means are provided to assure that each dropout plate is released after separation of the fusible links. The fusible links are specified to release at a temperature of 135°F. Following a LOCA, the actuation of the dropout plates and the reduction to low speed decreases the fan motor capacity to approximately 40,000 sft3/min at 7.9-in. WG static pressure with a power requirement of approximately 105 hp.

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FNP-FSAR-6 Following a LOCA, the SIS water in the core rapidly falls below saturation temperature.

Nevertheless, the containment cooling system design is based on the assumption that core residual heat appears as steam in the containment.

The criteria for the selection of the design service lifetime for the air coolers in the accident environment is based primarily on the service lifetime for the fan motor. The motor service lifetime is based on the criteria in IEEE Report NSG/TCS/SC2-A, Proposed Guide for Qualification Tests for Class IE Motors Installed Within the Containment of Nuclear Fueled Generating Stations.

Coil design and arrangement provide for rapid drainage of large quantities of condensed steam, thereby preventing loss of cooling capacity and maintaining cooling water temperatures below the boiling point.

Service water flow to the containment air cooling coils is unmodulated, reducing the probability of a failure due to a modulating valve or controller malfunction. Each of the four cooling coils has a separate branch supply and return run through the containment wall with appropriate isolation valves as discussed in subsection 6.2.4.

A radiation monitor is provided in the service water discharge from the containment air coolers to detect and prevent the release of radioactive service water to the environment.

High reliability is maintained through careful quality control and assurance procedures and by general arrangement of equipment to provide access for inspection and maintenance.

Components are designed to operate and withstand the post-accident environment.

Failure of a cooling unit is more than compensated for by the operation of either spray pump. If a containment air cooling unit coil should leak excessively, adequate alternate systems for containment heat removal are provided to meet the design requirements.

A coil rupture is detected by flow monitoring equipment and, as shown in the service water P&ID, drawings D-175003, sheet 1; D-175003, sheet 2; D-175003, sheet 3; D-175003, sheet 4; D-205003, sheet 1; D-205003, sheet 2; D-205003, sheet 3; and D-205003, sheet 4, isolated.

Leakage into the containment would have a negligible effect upon water boron concentration. A relief valve is provided on each coil to prevent excessive tube pressure due to heat buildup in the isolated portion after isolation. Because service water is being constantly circulated through the cooling coils, tube clogging during an accident is highly unlikely.

The heat sink for the containment air cooling units is the service water system. Adequate redundancy is provided in this system to assure continued availability in the event of a single failure. Failure of an inlet or outlet valve to a containment air cooling coil is detected by flow reduction. The containment sprays provide adequate backup cooling capacity.

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FNP-FSAR-6 6.2.2.4 Testing and Inspection 6.2.2.4.1 Containment Spray System Components of the spray system are inspected periodically to demonstrate system readiness, as discussed in response to GDC 39 in section 3.1. The pressure containing systems are inspected for leaks from pump seals, valve packings, flanged joints, and safety valves during system testing. The components of the system outside the containment are accessible for leaktightness inspection during periodic flow tests.

Proper functioning of the containment spray system will be demonstrated by periodic tests, as discussed in the response to GDC 40 in section 3.1.

Preoperational testing is performed to:

A. Demonstrate that the system is adequate to meet the design pressure conditions.

Outside the containment, the piping welds are subjected to radiographic inspection and/or part hydrostatic testing. Inside the containment, the spray header welds are subjected to 100% radiographic inspection.

B. Demonstrate that the spray nozzles in the containment spray header are clear of obstructions by passing air through the test connections.

C. Verify that the proper sequencing of valves and pumps occurs on initiation of the containment spray signal, and demonstrate the proper operation of remotely operated valves.

D. Verify the operation of the spray pumps. Each pump is run at shutoff and the miniflow directed through the normal path back to the RWST. During this time, the pump flow and discharge pressure should be recorded.

Periodic postoperational testing is performed to:

A. Verify that each automatic containment spray valve in the flow path that is not locked, sealed, or otherwise secured in position actuates to the correct position and each containment spray pump starts automatically on an actual or simulated actuation signal.

B. Verify each containment spray pumps developed head at the flow test point is greater than or equal to the required developed head.

With the procedures used to clean the containment spray system prior to completion of installation of the spray nozzles - system flush to remove particulate matter - and with the use of corrosion resistant materials of construction, it is not considered credible that a significant clogging of any of the nozzles would occur that would reduce the effectiveness of the system.

For this reason, the airflow through the nozzles has been tested during the preoperational 6.2-45 REV 31 10/23

FNP-FSAR-6 testing program and will be verified every 20 years to indicate that plugging of the nozzles has not occurred.

Airflow is delivered to the nozzles, through the headers, via test lines. Either balloons or streamers are deployed at the nozzle openings or hot air with an infrared camera is used to verify that sufficient airflow is established to the nozzles.

The sequence for the transfer of ESF loads to alternate power sources is discussed in subsection 8.3.1.

6.2.2.4.2 Containment Cooling System The containment cooling system is periodically tested and inspected as discussed in the response to GDC 39 and 40 in section 3.1.

Fans are tested and rated in accordance with the standards of the Air Moving Conditioning Association (AMCA).

Containment air cooling unit coil performance has been verified by factory test of a small coil section tested under the normal and LOCA conditions. Containment air cooling unit fan motors are certified by the manufacturer to operate during and/or subsequent to a LOCA. The certification is based on a qualification test on a prototype full size Class IE motor higher in horsepower and equal in design and construction to that of the actual purchased fan motor.

Test results taken from the prototype motor are interpolated into the size and capacity of the actual purchased motor. The test sequence simulating the design service conditions is aging, seismic forces, and containment environment. The motor manufacturer has conformed with detailed test procedures based on environmental conditions shown in table 3.11-1, Category A.

The fusible links are performance rated by Underwriters Laboratories. Two hundred fifty tests of actuation and tests to verify speed of operation demonstrate the acceptability of the link design in accordance with the provisions of Underwriters Laboratories' Standard of Safety UL 555-1970.

The load in the fusible links after they have been installed in the containment air cooling unit housing is supplied by adjustable compression springs designed to minimize the effect of thermal growth in the links which could possibly elongate them, thereby affecting their ability to permit fast opening of the plates at the moment of separation.

Provisions are made to allow inplace testing of fusible links to demonstrate that they perform their function following a LOCA.

The testing of the containment cooling system as well as its components demonstrates the initial capability of equipment and systems. Written test procedures establish minimum acceptance values for all tests. A recording of test results is useful in enabling detection of subsequent faulty performance.

6.2-46 REV 31 10/23

FNP-FSAR-6 Periodic tests of the activation circuitry and the system components can be conducted during normal plant operation assuring, in this way, a reliable performance upon demand throughout the plant lifetime.

Component qualification tests have been performed that demonstrate the characteristics of materials to be incorporated into components for the system by the manufacturer and ensure that the requirements of the procurement specification have been met.

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

Systems acceptance tests consist of deenergized and energized tests which demonstrate the proper mounting of components, proper hookup of circuits and connections, setting of instrumentation, and operation of interlocks. Equipment and system performance are monitored and rated.

6.2.2.5 Instrumentation Requirements 6.2.2.5.1 Containment Spray System Instrumentation and associated analog and logic channels employed for initiation of the containment spray system are discussed in section 7.3.

The injection phase of operation of the spray system is actuated either manually from the control room (2/2 logic) or on coincidence of two sets out of four high-high-high containment pressure signals. This signal starts the containment spray pumps and opens the discharge valves to the spray headers. When the RWST is exhausted, the recirculation phase of operation is manually initiated by the operator.

The following describes the instrumentation that is used for monitoring the containment spray system during normal or post-LOCA operation:

RWST level - Two channels of level instrumentation are provided on the RWST. Both channels additionally provide a high, low, low-low, and Technical Specification minimum level alarm which annunciate in the control room.

Containment sump water level - Two level indicators provide control room indication of containment sump level.

Containment pressure - Four channels provide containment pressure indication and high, high-high, and high-high-high containment pressure annunciations in the control room.

6.2-47 REV 31 10/23

FNP-FSAR-6 Pump discharge pressure - A pressure indicator is located in each containment spray pump discharge line. Readout is local.

Pump discharge flow - A flow indicator is located in each containment spray pump discharge line. Readout is provided in the control room.

6.2.2.5.2 Containment Cooling System Instrumentation and associated analog and logic channels employed for initiation of the containment cooling system are discussed in section 7.3.

The high-speed winding for each dual-speed fan motor receives power from a normal bus, and the low-speed winding receives power from an emergency bus. Electrical interlocks prevent the connection of both power supplies simultaneously. During normal operation, up to four units will be operating. On receipt of an SI signal, with offsite power available, the operating fans will be electrically switched from high to low speed. One unit per train could then be placed in the standby mode by the operator. On receipt of an SI signal with a loss of offsite power (LOSP/SI),

any running fans will trip (load shed signal or bus UV signal), and only two fans, as selected on the MCB train selector switches, will automatically start in slow speed. Additional fans may be manually started as desired, if diesel capacity and plant conditions permit. If failure of the electrical interlocks were to allow the fan motor to operate at high speed during an accident, the high-speed winding power supply would be disconnected by the overload trip due to the excessive current required to handle the high density steam air mixture. The fan motor would then be operated at low speed if required.

The following describes the instrumentation that is used to monitor the containment air cooling system during normal or accident conditions.

Cooling water exit temperature - Cooling water exit temperature is monitored.

Remote indication is provided in the control room.

Cooling water flow - Remote indication of cooling water supply and return flow is provided and low return flow is annunciated in the control room.

Cooling water radiation level - Cooling water return flow is monitored for radiation level. Indication and high-level annunciation are provided in the control room.

6.2.2.6 Materials A. Decomposition Products Materials used in or on ESF systems have been chosen so that the decomposition products, if any, of each material will not interfere with safe operation of any ESF. The estimated amount of material or chemicals that would 6.2-48 REV 31 10/23

FNP-FSAR-6 interfere with the safe operation of this system is zero. There are no commercial name materials included in this system which would cause degradation of system performance.

B. Materials Compatibility Materials compatibility of the containment spray system is discussed in paragraph 6.2.3.6.

6.2.3 CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEMS Post-accident ECCS recirculation fluid pH control system operates in conjunction with the penetration room filtration system. The penetration room filtration system is provided to limit the radiological offsite consequences resulting from a LOCA. The containment preaccess filter system and the containment purge system provide ventilation and filtration to allow access to the containment after shutdown. The containment minipurge system provides continuous ventilation and filtration of the containment atmosphere so as to allow periodic occupation of the containment during normal power operation. The control rod drive mechanism (CRDM) cooling system, the reactor cavity cooling system, the refueling water surface ventilation system, and the containment air cooling system provide ventilation and cooling for equipment located inside the containment.

6.2.3.1 Design Bases 6.2.3.1.1 ECCS Recirculation Sump pH Control System The ECCS recirculation sump pH control system is designed to ensure that offsite doses from an accident are within the limits of 10 CFR 100 by providing the following:

A. Scrubbing action in the containment atmosphere to reduce airborne iodine activity levels.

B. Retention of the iodine in solution.

C. The ECCS recirculation sump pH control system is designed to operate over a short period of time and under the environmental conditions existing after an accident as presented in table 3.11-1.

6.2.3.1.2 Penetration Room Filtration System The primary purpose of the penetration room filtration system is to limit release to the environment of radioisotopes from ECCS leakage into the penetration room under accident conditions. Although not credited in accident analyses, the penetration room filtration system also filters radioisotopes from containment leaks into the penetration area following a DBA. The system performs the following safety functions:

6.2-49 REV 31 10/23

FNP-FSAR-6 A. Exhausting of the penetration room atmosphere to the environment through particulate and charcoal filters which remove airborne radioactivity and maintain a slightly negative pressure within the penetration room area. This negative pressure ensures inleakage to the penetration room area, which helps to minimize the release of radioactivity to the environment.

B. Processing the air from the fuel handling area following a fuel handling accident.

Post-LOCA, the flow through each filter train may either be exhausted to the plant vent or partially recirculated to the penetration room area and ECCS pump rooms depending on the alignment of the PRF system. Recirculation is provided as a feature that provides multipass filtration of long-term ECCS recirculation and containment leakage which may reduce post-LOCA releases to the environment through fission product cleanup. The recirculated flow when ducted through the penetration area boundary will provide dilution and mixing of airborne radioactivity within the boundary. Recirculation cleanup of air through the penetration rooms is not credited in the accident analysis.

The penetration rooms are formed adjacent to the outside surface of the containment by enclosing the area around all penetrations, except as noted below. The penetration room boundary is shown in figures 1.2-2 through 1.2-7 and encloses a volume of 328,000 ft3. The only penetrations which do not pass within the penetration room are the following:

Containment leak rate test connections.

Refueling tube.

Equipment hatch, which contains a doubled gasketed closure.

Personnel access lock.

Auxiliary access lock.

Main steam and feedwater penetrations.

The containment leak rate test connections and the refueling tubes are equipped with blind flanges which are opened only during shutdown. The equipment hatch, personnel access lock, and auxiliary access lock can be tested during normal operation and are not considered sources of significant leakage. The main steam and feedwater penetrations are not considered a source of significant leakage because they are welded to the containment liner plate.

In analyzing the possibility of containment isolation valve leakage bypassing the penetration room filtration system, the penetrations can be divided into the following groups (the numbers in parentheses refer to the item numbers listed in table 6.2-31):

A. Penetrations for systems open to the containment atmosphere after a LOCA but which have components within the penetration room boundary.

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FNP-FSAR-6

1. RHR loop-in/low-head SI (10).

2. Reactor coolant pump seal water supply (14).

3. Containment spray lines (36).

4. High-head SI lines (35).

5. Containment sump recirculation lines (9, 37, and 38).

6. High-head to RCS cold-leg injection line (43).

7. Low-head SI line (44).

Potential leakage from the components of these systems is contained within the penetration room boundary.

B. Penetrations for systems which operate after a LOCA and are not open to the containment atmosphere, but which extend beyond the penetration room boundary.

1. Service water lines to and from the containment air coolers (27 and 28).

Potential leakage from these lines can be analyzed as follows:

a. The pressure in the supply line to the coolers is approximately 50 psig. Any leakage, even for the short-term pressure transient, is into the containment. Isolation valves can be closed to isolate the supply line if necessary.

b. The pressure in the return lines from the coolers is approximately 15 psig. Any leakage into the service water system is detected by the radiation monitor discussed in chapter 11, and isolation valves are closed to isolate the return line.

C. Penetrations for systems which are completely isolated by containment isolation valves after a LOCA and which form closed loops within the penetration room boundary.

1. Charging pump suction relief valve discharge to pressurizer relief tank (6).

2. Containment differential pressure instrument (8).

3. RHR loop-out (9).

4. Containment pressure instrument (15).

5. Containment sump pumps discharge (26).

6.2-51 REV 31 10/23

FNP-FSAR-6

6. Containment sump pump sample recirculation line (47).

7. Post-LOCA containment sampling system (48 and 49).

Potential leakage from these systems is contained within penetration room boundary, with the exception of backleakage to the reactor water storage tank vent past the isolation valves of the containment sump pumps discharge. This leakage pathway is included in the evaluation of LOCA dose consequences in section 15.4.

D. Penetrations for systems which are completely isolated by containment isolation valves after a LOCA but pass through the penetration room into other portions of the auxiliary building. They are connected to Seismic Category I systems and are water filled in the post-accident condition.

1. Pressurizer relief tank makeup (5).

2. Normal charging line (13).

3. Pressurizer steam sample line (16).

4. Pressurizer liquid sample line (17).

5. Hot leg sample line (18).

6. Reactor coolant pump cooling water supply and return (29 and 31).

7. Reactor coolant pump thermal barrier cooling water return (32).

8. Excess letdown heat exchanger and RC drain tank heat exchanger component cooling water supply and return (33 and 34).

9. Accumulator sample line (40).

10. Pressurizer pressure deadweight generator (41).

11. Service water to and from reactor coolant pump motor air coolers (45 and 46).

12. Normal letdown line (11).

13. Excess letdown and seal water (12).

Potential leakage from these systems is considered in the same manner as potential leakage from the penetrations in group E below.

E. Penetrations for systems which are completely isolated by containment isolation valves after a LOCA but pass through portions of the auxiliary building other than 6.2-52 REV 31 10/23

FNP-FSAR-6 the penetration room and are either connected to nonseismic Category I systems or are not water filled in the post-accident condition.

1. Accumulator test line (1).

2. Refueling cavity supply (2).

3. Nitrogen supply to accumulators (3).

4. Nitrogen supply to pressurizer relief tank (4).

5. Reactor coolant drain tank drain (7).

6. Deleted.

7. Deleted.

8. Fuel transfer tube (19).

9. Service air (20).

10. Instrument air (21).

11. Containment air sampling system (22 and 23).

12. Containment purge supply and exhaust (24 and 25).(a)

13. Leak rate test lines (30).

14. Accumulator makeup line (39).

15. Reactor coolant drain tank vent (42).

16. Post-LOCA containment venting (50).

17. Demineralized water (51).

As stated in subsection 6.2.4, all containment isolation valves which are closed in the post-LOCA condition are individually tested for leaktightness as required by the Containment Leakage Rate Testing Program. The allowable rate of isolation valve leakage is defined and limited by the Containment Leakage Rate Testing Program. Any leakage past the isolation valves of the systems in groups D and E above has the potential to leak into the auxiliary building beyond the penetration room boundary.

The design environmental conditions for the penetration room filtration system components are presented in table 3.11-1.

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FNP-FSAR-6 6.2.3.1.3 Containment Preaccess Filtration and Purge Systems The containment preaccess filtration and purge systems are designed to do the following:

A. Provide sufficient air circulation and filtering throughout all containment areas to permit safe and continuous access to the containment within 2 h after reactor shutdown, assuming defects exist in no more than 1% of the fuel rods.

B. The minipurge system will provide sufficient purge air circulation so as to allow up to 6 h per week occupancy of the containment during power operation.

C. The containment purge system will normally provide purging of the containment after cold shutdown (prohibited in Modes 1 through 4). The minipurge system may be in use during reactor operation, reactor shutdown, and defueled.

Interlocks prevent simultaneous operation of the minipurge system and the containment purge system.

D. The containment purge system and the minipurge system provide for purging of the containment to the vent stack for dispersion to the environment.

E. The isolation valves on the minipurge system are capable of containment purge isolation within 5 s in the event an accident is detected in the containment. The containment purge system isolation valves are closed and deactivated during Modes 1 through 4. During modes 5 and 6, these valves are designed to close within 5 s upon receipt of a high radiation signal and containment ventilation isolation signal.

Total releases due to containment purging are in accordance with requirements of the applicable portions of the Code of Federal Regulations, Title 10. For a discussion of plant releases due to containment purging, refer to subsection 11.3.6.

a. Containment purge supply and exhaust lines contain a third isolation valve and a normally open bleed valve. As shown in drawings D-175010, sheet 1 and D-175010, sheet 2, any leakage past the first two isolation valves is vented into the penetration room. Similarly, any leakage past the minipurge supply and exhaust isolation valves is vented into the penetration room.

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FNP-FSAR-6 6.2.3.1.4 Containment Ventilation Systems The containment ventilation systems, consisting of the containment air cooling system, the CRDM cooling system, the reactor cavity cooling system, and the refueling water surface ventilation system, are designed to perform the following functions:

A. Remove the normal heat loss from all equipment and piping in the containment during plant operation and to limit the maximum average ambient temperature to 120°F.

B. Provide for depressurization of the containment following an accident. The post-accident design and operating criteria are detailed in subsection 6.2.2.

C. Provide a reliable supply of cooling air to the CRDMs.

D. Remove water vapor above the reactor cavity pool during refueling operations to improve visibility of fuel elements within the pool, reduce the heat stress on personnel, and minimize personnel inhalation exposure during shutdown.

E. Provide a minimum containment ambient temperature of 60°F during reactor shutdown.

F. Provide cooling air for the primary shield, the enclosed nuclear instrumentation detectors, and the reactor vessel supports.

The systems are designed to satisfy the containment ventilation and cooling requirements during normal operation. In certain cases where ESF functions are also served by the systems, component sizing is determined from the heavier duty specifications based upon accident conditions and requirements.

6.2.3.2 System Design 6.2.3.2.1 ECCS Recirculation Sump pH Control System The ECCS recirculation fluid pH control system consists of trisodium phosphate baskets located in the recirculation sump area of containment. The initial spray water consists of borated water from the RWST. As the containment is sprayed during injection, the recirculation sump water level rises, eventually immersing the baskets and dissolving the trisodium phosphate contents.

Once the contents are dissolved through the fine mesh basket screen sides and bottom, the recirculation sump water pH is raised to an equilibrium value of 7.0 or greater. The containment spray pumps will circulate approximately 5 percent of the flow through their bypass line. Refer to D-175038, sheet 3 and D-205038, sheet 3 for configuration.

When the RWST reaches low-low level, recirculation spray flow is initiated for iodine removal.

The operator can remotely initiate recirculation flow by use of either of the spray pumps. This mode of operation is continued for a period of at least 2 h following the accident in order to continue iodine removal from the containment atmosphere.

6.2-55 REV 31 10/23

FNP-FSAR-6 Coatings used in the containment meet ANSI 101.2-1972. This requirement is not applicable to paints procured for the touch-up or coating of off-the-shelf items or equipment having manufacturers standard finishes. Most types of paint used in the containment are listed by manufacturers' designation in paragraph 3.8.1.6.6. Table 6.2-37 provides a listing of most containment coatings including dry specific gravity and surface area covered. The quality assurance program for paint applications is discussed in chapter 17.

6.2.3.2.2 Penetration Room Filtration System The penetration room filtration system is shown in D-175022, while the conformance of the penetration room filtration system to NRC acceptance criteria contained in Regulatory Guide 1.52 is presented in table 6.2-25.

The system consists of two full-capacity redundant fans and filter systems which share a common vent. The fans and filters are located in the penetration room filtration system equipment room.

The design flowrate through each filter train is 5000 sft3/min. The exhaust fan has a capacity of 500 sft3/min., and the recirculation fan capacity is 4500 sft3/min. These design flowrates are based on the following sizing criteria:

(Note: The assumed inleakage used in sizing the fan capacities was not achieved during plant startup. Higher inleakage rates are acceptable due to conservative system sizing and the systems ability to maintain a slightly negative pressure.)

A. The exhaust flowrate is equivalent to the penetration room boundary inleakage; i.e., the sum of all possible inleakages when a pressure of -1.5-in. WG is maintained within the penetration room boundary. This is an important parameter because it determines the relationship between the amount of filter effluent discharged to the environment and the amount recirculated to the penetration room.

Minimizing the penetration room inleakage increases the system effectiveness.

Evaluation of the proposed structure joints and pipe and cable tray seals indicates a potential inleakage rate of < 10 percent of the penetration room boundary volume per day. However, for estimating the exhaust fan capacity, it has been conservatively assumed that, with a -1.5-in. WG pressure, the inleakage is 100% of the penetration room volume per day. This inleakage is equivalent to 250 sft3/min. Each exhaust fan has been conservatively designed to provide twice this flowrate.

B. The design ratio of recirculation flow to exhaust flow is 9 to 1. No recirculation credit has been taken in calculating the consequences of an accident.

Each system consists of a prefilter, high efficiency particulate air (HEPA) filter, charcoal filter, heating coil, and two fans.

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FNP-FSAR-6 Both the recirculation and exhaust fans may be used to hold the penetration room area and ECCS pump rooms or the spent fuel pool area at a negative pressure with respect to the penetration room filtration equipment room following a LOCA or FHA, respectively.

During normal operations, the spent fuel pool area is served by its normal heating, ventilating, and filtration system. Whenever irradiated fuel is in the spent fuel storage pool, one penetration room filtration system train is aligned to the spent fuel pool area. Both penetration room filtration system trains are aligned to the spent fuel pool area during movement of fuel or heavy loads over the fuel. During maintenance activities, the penetration room filtration system may be aligned to allow the spent fuel ventilation system to exhaust penetration room areas in order to provide temporary ventilation. Under accident conditions, the penetration room filtration system can be aligned to operate under either of two modes, namely the fuel handling accident (FHA) mode or the post-LOCA operating mode. While the penetration room filtration system is not credited in the FHA dose analysis (subsection 15.4.5), its operation in FHA mode is described below.

Upon receiving an actuation signal initiated by either high radiation or low flow in the normal spent fuel pool ventilation system, the penetration room filtration system is automatically placed in its fuel handling accident alignment by starting the penetration room filtration system fans and isolating the normal spent fuel pool ventilation system. The spent fuel pool room is automatically isolated by the closure of redundant isolation dampers located in the normal spent fuel pool HVAC supply and exhaust ductwork. Movement of new fuel over spent fuel with the spent fuel area roof new fuel access hatch open creates the potential for a fuel handling accident with a release pathway which bypasses the radiation monitors in the exhaust duct, and consequently a bypass of the PRF. This configuration is specifically evaluated in subsection 15.4.5 as being bounded by the design basis accident FHA. Both Train A and B penetration room filtration systems start based on input from their respective instruments.

Operators may manually shut down one train after the autostart of both penetration room filtration system trains. Train A and B penetration room filtration systems share a common duct from the spent fuel pool area to the penetration room filtration system mechanical equipment room.

Upon receiving a containment isolation actuation system (CIAS) phase B signal from the protection system, both Train A and B penetration room filtration system fans energize and the suction dampers to the penetration room area open automatically. The flow through each filter train may either be exhausted to the plant vent or partially recirculated to the penetration room area with the remainder exhausted to the plant vent stack. The recirculation damper may be positioned automatically or manually to maintain negative pressure control of the penetration room area. The penetration room filtration system recirculation and exhaust discharge dampers to the plant vent stack are interlocked with their respective fans to open and close automatically whenever the fan starts or stops. A backdraft damper is in series with the discharge dampers to prevent backflow through the filtration unit. Before post-LOCA emergency core cooling recirculation, the penetration room filtration system is placed in its post-LOCA alignment by manually securing spent fuel pool area suction line valves closed.

6.2-57 REV 31 10/23

FNP-FSAR-6 Fans - The exhaust and recirculation fans are closed impeller centrifugal fans. The fan motor sizes are 1.75 hp for the exhaust fan and 20 hp for the recirculation fan. Fan motors conform to standards of NEMA, IEEE, and ANSI.

Filters - The filters are composite units consisting of prefilter section, absolute filter section, and impregnated charcoal bed filter section. Each section is designed as follows:

A. The prefilters are designed to have a mean efficiency of 85% when tested in accordance with the National Institute of Standards and Technology Discoloration Test Method.

B. The HEPA filters are designed to be capable of removing 99.97% minimum of particulate matter 0.3 micron or larger in size. This particulate filter is water and fire resistant.

C. The charcoal filter is an impregnated, activated carbon bed that is designed to be capable of removing elemental iodine with an efficiency of 99.0% and, at relative humidities below 70%, all iodines, including organic, with an efficiency of 99.0 %.

The prefilter, absolute, and charcoal filters are designed for a nominal flowrate of 1000 ft3/min per 4 ft2 face area and are sized with the assumption that 100% iodines from ECCS recirculation leakage will pass through the penetration room filtration system and will also be held up on the filters. The total amount of carbon in each penetration room filtration train is approximately 774 lb.

The HEPA filters used in the penetration room filtration system are designed and manufactured in accordance with the requirements of USNRC Health and Safety Bulletin No. 306 (Military Specification MIL-F-51068C, June 8, 1970, Filter, Particulate, High-Efficiency, Fire-Resistant),

and USNRC Health and Safety Bulletin No. 297. The charcoal is activated, coconut base charcoal, impregnated and retained in horizontal trays of nominal 2-in. bed depth in accordance with AACC-CS-IT, 1968. Humidity control consists of a heater which energizes when the penetration room filtration system fans are operating.

Ducting - The ducting upstream of the fans takes suction at the higher elevation of the penetration room boundary, and the recirculation ducts discharge air into the lower elevation of the boundary.

The ducts upstream of the fans are designed to operate under a pressure of in. WG. The ducts from the discharge of the fans to the vent stack are designed for 2-in. WG positive pressure. The recirculation ducts are designed to operate under a pressure of in. WG.

Valves - The valves at the discharge of all fans are pneumatically operated, two-position, fail-open butterfly valves. The valves in the ducting of the filter trains and in the suction and recirculation lines are motor-operated valves.

The penetration room filtration system instrumentation is in accordance with the requirements of IEEE 279.

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FNP-FSAR-6 Three differential pressure transmitters measure the differential pressure between the penetration rooms and pressure in the penetration room filtration system equipment room. The recirculation fan continues to operate in the exhaust mode until the pressure in the penetration room boundary reaches a predetermined set point. The analysis of the combined system (fans vs. inleakage) indicates a setpoint of in. WG pressure to be used for switching to recirculation operation. A two-out-of-three arrangement of differential pressure switch signals actuates the recirculation line butterfly valve to open when the setpoint is reached. A two-out-of-three differential pressure signal of -2 in. and a recirculation line valve open signal is indicated in the control room. A permit switch allows the recirculation line valve to open only if the recirculation fan is operating.

After automatic post-LOCA valve alignment, the system valves may be closed or throttled by the operators to affect ALARA as long as negative pressure is maintained. The flow from the exhaust and recirculation fans may be exhausted to the plant vent or aligned to partially recirculate flow to the penetration room boundary with the remaining exhausted by the plant vent.

In the event of control air failure, the penetration room filtration system air operated valves fail open (in the exhaust mode), except the spent fuel pool suction valves to the penetration room filtration system, which fail closed on loss of control air. In the event of control air failure during fuel handling operation, air accumulators provide backup air supply. The suction valves are manually closed remotely during post-LOCA operation.

If both systems start following an accident, one can be manually shut down and placed in the standby mode. The standby system is manually restarted if the operating system should fail.

The following conservative assumptions were used to calculate an upperbound value of heat generation rate inside the penetration room filtration system charcoal filters versus time after a LOCA:

A. Regulatory Guide 1.4 source terms.

B. Leak rate from the containment into the penetration room volume of 0.15 containment volume percent per day.

C. Instantaneous deposition of 100% leaked iodines onto the charcoal filters.

D. Heat deposition in the filters from 100% of all betas and 50% of all gammas.

E. No spray removal of contaminants inside containment.

Using these assumptions, a peak filter heat generation rate of 163 W is calculated to occur at 11.5 days after a LOCA. Further calculations indicate that natural convection is more than sufficient to remove this heat load from an isolated filter, while maintaining filter temperatures acceptably low.

The penetration room filtration system mechanical room is heated by area heaters located in the filter area to ensure that warm, moist air cannot be processed by relatively cooler filters, thereby 6.2-59 REV 31 10/23

FNP-FSAR-6 resulting in moisture condensation and accumulation inside the charcoal filters. Electric heating coils are also provided at the upstream side of each unit to maintain the entering air below 70 percent relative humidity. The heating coils are provided for defense-in-depth since no credit is taken for reduction of relative humidity in the accident analysis. Temperatures in the emergency pump rooms within the penetration room boundary are controlled by pump room coolers in each room.

Details of the emergency power sources are discussed in subsection 8.3.1.

6.2.3.2.3 Containment Preaccess Filtration and Purge Systems The containment preaccess filtration system is designed to reduce activity levels in the containment atmosphere prior to routine personnel access at power or in advance of a scheduled reactor shutdown. The system, as shown in drawings D-175010, sheet 1 and D-175010, sheet 2 draws contaminated air from the containment across a filter assembly which consists of a prefilter, HEPA filter, and a charcoal filter. The air then passes through the system fan and is discharged back into the containment. Design system flowrate is 10,000 ft3/min through each of two filter subsystems.

The containment purge system is independent of any other system and includes provisions to supply and exhaust air from the containment. The system is shown in drawings D-175010, sheet 1 and D-175010, sheet 2; D-205010, sheet 1 and D-205010, sheet 2. The supply system includes an outside air connection to prefilters, heating coils, a fan, a duct system, and a supply penetration with three butterfly valves in series for tight shutoff. The exhaust system includes an exhaust penetration with three butterfly valves in series, a duct system, a filter bank with prefilters, HEPA and charcoal filters, and an exhaust fan. The exhaust system includes a two-speed fan so that purging can be performed at half or full flowrate. The full flowrate is 48,500 ft3/min. The quick closing purge isolation valves are capable of closing within 5 s, during Modes 5 and 6, upon receipt of a high radiation signal from the purge discharge radiation monitors, and upon receipt of a containment ventilation isolation signal.

NRC acceptance criteria associated with the design of the containment purge system include the capability to detect high radioactivity conditions in containment prior to opening any of the purge valves. High radioactive conditions inside containment would be detected prior to opening the purge valves by means of the containment atmosphere particulate radioactivity monitor (R-11) and the containment atmosphere gaseous radioactivity monitor (R-12). Should prescribed high radiation levels exist, an audible annunciation would be provided in the control room.

The containment minipurge system is designed to maintain radioactivity levels in the containment consistent with occupancy requirements with continuous system operation. The preaccess filtration system is not required for this occupancy but is available for use in minimizing the need for containment purging.

The operation of the minipurge system is independent of the operation of the containment purge system, although there are common ductwork and common filters. The system is shown in drawing D-175010, sheet 1 and D-175010, sheet 2. The supply system uses the containment 6.2-60 REV 31 10/23

FNP-FSAR-6 purge supply filter and ducting. A separate minipurge supply fan, located in the penetration room, operates in an 18-in. duct system which reduces to 8 in. and bypasses the first purge isolation valve (48 in.) outside the containment. The outside air is discharged through an 18-in.

duct which reduces to 8 in. and is connected to the 48-in. purge duct. The minipurge supply system has two isolation valves in series for tight shutoff. The exhaust system has an exhaust connection inside the containment and an exhaust fan similar to arrangement to the supply system. Exhaust air passes through the containment purge exhaust filter. The exhaust system has two isolation valves in series.

The operation of the exhaust fan is dependent upon the operation of the supply fan. An interlock from the supply fan is installed in the control circuit of the exhaust fan such that the exhaust fan runs automatically when the supply fan is running. However, the exhaust fan may be operated independent of the supply fan, when the main purge supply valves or the minipurge supply fan is out of service by defeating the interlock between the two fans. During this mode of operation the plant administrative controls will be in place to ensure that the containment pressure is maintained within the Technical Specification limits.

The minipurge supply fan provides a flowrate of approximately 2850 ft3/min and the exhaust fan provides a flowrate of approximately 2650 ft3/min. The 8-in. isolation valves are similar in design to the 48-in. containment purge isolation valves and they are capable of closing in 5 s after receipt of a containment ventilation isolation signal or a high radiation signal from the purge discharge radiation monitors.

The plant can operate either in the minipurge mode, with the 48-in. containment isolation valves closed in the supply and exhaust ducts during plant Modes 1 through 6 and defueled, or in the full-purge mode, with the 48-in. containment isolation valves open and the 8-in. minipurge isolation valves closed during plant Modes 5 and 6 or defueled. Interlocks prevent operation in both modes simultaneously.

The containment purge system and minipurge system are designed to meet the NRC acceptance criteria for containment isolation requirements found in Standard Review Plan (SRP) Section 6.2.4, Revision 1; Branch Technical Position CSB 6-4, Revision 1; SRP Section 3.9.3, Revision 1; and NUREG-0737, item II.E.4.2. In addition, the safety signals to all purge and ventilation isolation valves meet the following criteria:

A. The overriding (i.e., the signal is still present but is blocked in order to perform a function contrary to the signal) of one type of safety actuation signal (e.g.,

radiation) must not cause the blocking of any other type of safety actuation signal (e.g., pressure) to the isolation valves.

B. Sufficient physical features (e.g., key lock switches) are provided to facilitate adequate administrative controls.

C. The system-level annunciation of the overridden status is provided for every safety system impacted when any override is active.

D. Diverse signals should be provided to initiate isolation of the containment ventilation system. Specifically, containment high radiation, SI actuation, and 6.2-61 REV 31 10/23

FNP-FSAR-6 containment high pressure should automatically initiate containment ventilation isolation.

E. The instrumentation and control systems provided to initiate containment ventilation isolation should be designed and qualified as safety-grade equipment.

F. The overriding or resetting (i.e., the signal has come and gone, and the circuit is being cleared in order to return it to the normal condition) of the isolation actuation signal should not cause the automatic reopening of any isolation/purge valve.

Other NRC acceptance criteria include:

G. The radioactivity released during normal operation will be within the limits of column 1, table II, Appendix B to 10 CFR 20.1 - 20.601; the total radioactivity released following an accident results in calculated offsite doses less than the guideline values in 10 CFR 100; and the systems meet the applicable requirements of General Design Criterion 56.

H. The resilient material valve seals of the 48-in. and the 8-in. containment purge supply and exhaust isolation valves shall be replaced at least once per 9 years, in accordance with the Technical Requirements Manual.

The temperature and the humidity of the air stream through the preaccess and the purge systems are the same as for the containment atmosphere and are maintained by the containment cooling and ventilating systems.

Fans - The fan used in each half-capacity subsystem of the containment preaccess filtration system is of the vaneaxial type, with a design flowrate of 10,000 sf3/min each. Fan motors are 25 hp each. Both the exhaust fan and the supply fan in the containment purge system are two-speed centrifugal fans with design flowrates of 50,000/25,000 sf3/min each.

Fan motors are 60 and 125 hp for the supply and exhaust fans, respectively. Both the minipurge supply and exhaust fans are centrifugal fans with rated flowrates of 5000 sf3/min each; however, the supply fan will operate at approximately 2850 ft3/min and the exhaust fan will operate at approximately 2650 ft3/min with the 8-in. minipurge valves. Motors for these fans are 15 hp each. They are designed in accordance with the applicable portions of AMCA 99-67, Standards Handbook; AMCA 210-67, Test Codes for Air Handling Devices; and AMCA 211A-65, Certified Rating Program for Air Moving Devices.

Filters - The filters are composite units consisting of prefilter sections, absolute filter section, and impregnated charcoal bed filter section. The prefilter section and the absolute (HEPA) filter section are as described in paragraph 6.2.3.2.2. The charcoal filters are impregnated, activated carbon beds that are designed to be capable of removing, at relative humidities below 70%, all iodines with an efficiency of at least 95%. The prefilters and the absolute and charcoal filters are designed for a nominal flowrate of 1000 ft3/min per 4-ft2 face area. Carbon weights are approximately 1875 lb and 7400 lb for the preaccess filtration and purge systems, respectively.

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FNP-FSAR-6 The containment purge exhaust filters are seismically qualified. Details of the emergency power sources are discussed in chapter 8.

6.2.3.2.4 Containment Ventilation Systems The containment ventilation systems flow diagram is shown in drawings D-175010, sheet 1; D-175010, sheet 2; D-205010, sheet 1; and D-205010, sheet 2. The containment air cooling fans, CRDM cooling fans, and reactor vessel support cooling fans are direct driven units, each with standby units for redundancy. Each system is provided with pressure differential switches to verify the existence of airflow in the associated ducting.

The emergency functioning of the containment air coolers is discussed in subsection 6.2.2.

During normal operation, the four coolers take suction from the operating level at el 155 ft and discharge into a common header which distributes the cooled air to the lower regions of the containment through distribution ductwork. Each air cooling unit consists of the following equipment arranged so that, during normal operation, air flows through the assembly in the following sequence: inlet screen, cooling coil, fan, and a discharge header which is common to all units. Containment cooler and cooler fan design parameters are presented in table 6.2-28.

The containment air cooling units are located at the 155-ft elevation outside the secondary shield area. The shielded location makes inspection of the equipment possible at power under controlled access conditions and immediately after a hot shutdown. The containment cooling and ventilating functions are augmented by the containment recirculation fans, which take suction from the containment dome and discharge downward toward el 155 ft. Design parameters for these four fans are presented in table 6.2-28.

The containment air cooling units, in conjunction with the containment recirculation fans, provide mixing of the containment atmosphere during normal operation to augment heat removal and maintain uniform temperature distributions throughout the containment volume. A portion of the containment air cooler discharge is ducted directly to the reactor cavity to provide cooling to the cavity and to the instrumentation and equipment within.

The CRDM cooling system consists of fans and ducting to draw air through the CRDM shroud and eject it to the main containment atmosphere. One hundred-percent redundancy is provided by a standby fan.

In the event of a failure of the CRDM cooling system, the airflow to the CRDMs may be lost.

Loss of airflow to the CRDM would increase the operating coil temperatures. In the event of high CRDM temperatures, the control rods fall to the safe position, i.e., inserted into the core.

Design parameters for the CRDM cooling system fans are presented in table 6.2-28.

The reactor vessel support cooling system, consisting of two 100% capacity fans and ducting, is arranged to cool the reactor vessel supports by drawing air through the supports. One hundred percent redundancy of all active components is provided.

Reactor cavity cooling system fan design parameters are presented in table 6.2-28.

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FNP-FSAR-6 The refueling water surface ventilation system is utilized when necessary during refueling operations to remove water vapor above the refueling canal, thereby improving the visibility of the fuel elements within the pool and reducing the heat stress on personnel working in the vicinity of the refueling canal. This is accomplished by having a supply fan draw air from the containment atmosphere and supply it above the water surface. This air mixes with the water vapor emanating from the canal. The exhaust fan draws air from the opposite side of the canal and exhausts to the containment where it is diluted and filtered before being discharged through the plant vent. Refueling water surface ventilation system supply and exhaust fan design parameters are presented in table 6.2-28.

6.2.3.3 Design Evaluation 6.2.3.3.1 ECCS Recirculation Sump pH Control System The spray system, by virtue of the large surface area provided between the droplets and the containment atmosphere, affords an excellent removal mechanism for fission products postulated to be dispersed in the containment atmosphere. Radioiodine, in its various forms, is the fission product of primary concern in the evaluation of a LOCA. The major benefit of the containment spray is its capacity to absorb molecular and particulate iodine from the containment atmosphere. To enhance the capacity to retain this iodine, the recirculation solution is adjusted to an alkaline pH which promotes iodine hydrolysis to nonvolatile forms.

Values for the spray removal half-life of iodine in a typical containment are on the order of minutes. This makes the spray system a very efficient fission product removal system in comparison to such alternatives as charcoal filtration systems.

6.2.3.3.1.1 Containment Spray Iodine Removal Model. The spray iodine removal model is based on data obtained from measurements of the performance of Spray Engineering Company nozzle 1713 for the range of pressures expected to be encountered in the containment for a LOCA. A conservative elemental iodine spray removal coefficient, as shown in table 6.2-29, is used based on the data provided in references 19 through 21. The limiting decontamination factor (DF), calculated in accordance with reference 21 using a partition coefficient based on NUREG-0800, is reached at about 24 min.

A particulate iodine spray removal coefficient is calculated by the method outlined in reference 20. Removal efficiency for particulate iodine is assumed to decrease by a factor of 10 upon reaching a DF of 50, and to cease at 8 h.

6.2.3.3.1.2 Deposition Iodine Removal Model. The deposition removal coefficient for elemental iodine is calculated based on the guidance of reference 20. Deposition is assumed to occur only on galvanized surfaces and surfaces coated with zinc based paint or epoxy paint, as listed in table 6.2-2. The deposition removal rate of iodine is assumed to continue at the initial 6.2-64 REV 31 10/23

FNP-FSAR-6 rate until a DF of 100 is achieved for the containment atmosphere, and then at a reduced rate until a DF of 1000 is achieved, as shown in table 6.2-29.

6.2.3.3.1.3 Spray Performance Evaluation.

Injection Phase Operation The spray iodine removal analysis is based on the assumptions that:

A. Only one of two spray pumps is operating.

B. The iodine removal constants are calculated based on conservative containment parameters and test data from references 19 and 21.

The performance of the spray system was conservatively evaluated at the peak temperature and pressure resulting from a double-ended rupture of the RCS with no credit taken for the subcooling of the ECCS. These pressure and temperature conditions were used resulting in a minimum spray flowrate of 2480 gal/min during injection and 2240 gal/min during recirculation for the single spray pump assumed to be operating.

Iodine removal constants for the spray system, calculated with the model described and with the above mentioned assumptions, are shown in table 6.2-29.

Recirculation Phase The containment spray system is operated in the injection mode until the RWST low-low level setpoint is reached. The spray system is then operated in the recirculation mode, taking suction from the containment sump.

Reevolution of Iodine Any reevolution of dissolved iodine from the sump to the containment atmosphere is dependent on the concentration gradient between the liquid and vapor phases, pH, and temperature. The assumptions used minimize the amount of TSP dissolved in the sump, and maximize the temperature and volume of water delivered to the sump.

With this set of assumptions, the containment recirculation sump conditions (temperature and pH) are such that the required partition coefficient to maintain the calculated DF in the vapor phase is exceeded in the spray recirculation phase.

6.2.3.3.2 Penetration Room Filtration System The penetration room filtration system consists of two complete and separate fan and filter systems, each with 100-percent capacity. Although the system actuation signal automatically starts both fan and filter systems, one system can be placed in the standby mode by the operator. Motor or fan failure in the operating system is annunciated in the control room. The 6.2-65 REV 31 10/23

FNP-FSAR-6 standby system is then manually started. Upon loss of offsite power, the entire system is automatically connected to the emergency diesel generators.

The ducting conveying unfiltered air upstream of the filter trains is at negative pressure, barring the possibility of outleakages of contaminants.

Filter components receive factory and field tests to ensure against bypass and confirm specified efficiencies.

The penetration room filtration system produces a slightly negative pressure gradient inside the penetration room boundary by exhausting air from the highest point within the boundary, the electrical penetration room. Communicative paths between the various rooms within the penetration room boundary are connected by a series of pipe sleeves of various sizes to serve as flow paths from the penetration room boundary lower levels. Pressure differentials between the highest and lowest points within the penetration room boundary are small and allow for boundary pressure to be sensed at a single point and displayed in the control room, with high and low boundary pressure alarmed. Pressure detection within the penetration room boundary is at the location of least negative pressure, the RHR heat exchanger room, which is located at the lowest level within the boundary.

The RHR heat exchanger room is maintained at a slightly negative pressure with the penetration room filtration system partially recirculating to the penetration room boundary with the remainder exhausted to the plant vent stack. If the penetration room filtration system exhausts all air from the boundary to the vent stack, much lower negative boundary pressures can be maintained. The penetration room boundary is located primarily below grade and is completely enclosed within the auxiliary building; therefore, no unfiltered leakage from wind effects are considered.

The penetration room filtration system ensures that offsite radiation exposures resulting from post-LOCA ECCS recirculation leakage are within the requirements of 10 CFR 50.67.

A single-failure analysis for the penetration room filtration system is given in table 6.2-30.

The penetration room model used in the LOCA analysis is given in paragraph 15.4.1.10.

6.2.3.3.3 Containment Preaccess Filtration and Purge Systems An evaluation of the capability of the containment preaccess filtration, minipurge, and purge systems to maintain offsite effluent concentrations during normal operation within established guidelines is included in subsection 11.3.6. Radiation monitors isolate the minipurge and full purge system supply and discharge in the event of high discharge activity levels, such as might be experienced following a fuel handling accident inside the containment.

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FNP-FSAR-6 6.2.3.3.4 Containment Ventilation Systems The post-accident operation of the containment air coolers is evaluated in paragraph 6.2.2.3.

The normal operation of the containment air coolers and the operation of the remaining containment ventilation systems are not required to reduce accident doses or maintain offsite effluent concentrations during normal operation within established guidelines.

6.2.3.4 Tests and Inspections 6.2.3.4.1 ECCS Recirculation Sump pH Control System The functional and periodic post-operational testing of the containment spray system is described in paragraph 6.2.2.4.1. A test signal simulating the containment spray signal is used to demonstrate the operation of the spray system up to the isolation valves on the pump discharge. The isolation valves are closed for the test. These isolation valves are checked separately.

ECCS recirculation sump pH control is a passive system which consists of three baskets located in the recirculation sump area of containment. Each basket has level marks visible from the outside which indicate the acceptable range in trisodium phosphate volume for the basket for the trisodium phosphate compound being used (currently Na3PO4

12H2O

1/4NaOH). An equivalent amount of trisodium phosphate compound with a different chemical formula may be used. When equivalent compounds are used, the allowable weights/volumes may be different; however, the equivalent amount of trisodium phosphate compound must raise the pH of the recirculating solution into the range of 7.0 to 10.5. A visual inspection of the baskets is performed each refueling outage to verify structural integrity and level for the baskets. The intent of the associated surveillance requirements is to verify containment of the trisodium phosphate. Therefore, broken, crimped, or oxidized screen mesh is acceptable as long as the contents are contained. Also, lumps/caking is an analyzed condition.

6.2.3.4.2 Penetration Room Filtration System Fans are tested and rated in accordance with the standards of the Air Moving Conditioning Association (AMCA).

The penetration room filtration system, as well as its components, are tested prior to startup and periodically during operation. Written test procedures establish minimum acceptance values for all tests. A recording of test results and instrumentation that alarms at the main control board enables early detection of faulty performance.

The penetration room filtration system is provided with testing facilities to demonstrate system operability.

Periodic tests of the activation circuitry and the system components can be conducted during normal plant operation ensuring a reliable performance upon demand for the life of the plant.

6.2-67 REV 31 10/23

FNP-FSAR-6 Periodic testing of the penetration room filtration system actuation instrumentation is conducted in accordance with the Technical Specifications.

[HISTORICAL] [Initial tests and the purpose of each test are listed as follows:

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

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

Test results on actual or similar types of filter assemblies demonstrate their adequacy for this application.

The following demonstrative tests are performed:

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

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

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

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

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

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FNP-FSAR-6

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

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

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

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

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

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

A. Proper actuation of control circuitry in both modes.

B. Proper flow path alignment in both modes.

C. Leaktightness of each filter assembly.

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

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

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

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

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

6.2-69 REV 31 10/23

FNP-FSAR-6 D. With the system operating, verify circulation of air within the penetration rooms in the LOCA mode.

E. Install the roughing filter and high-efficiency filter. With the systems operating, test leaktightness and performance, using DOP smoke of 0.3 micron mass median diameter.

Penetration should not exceed 0.1 percent.

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

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

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

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

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

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

Periodic tests following acceptance - These tests demonstrate that the system performance capability is in accordance with the Technical Specifications. Operability of the penetration room filtration system is checked by periodically starting the fans and exercising the valves. For purposes of periodically testing the retentive capability of the charcoal filter, test canisters are placed in the filter housing in locations which allow the canisters to be subjected to the same air currents as the beds. These are periodically removed and tested at the charcoal supplier's laboratory.

Provision is made for periodic inspection and testing of the penetration room filtration system.

Methods of testing the system are per guidance from ASME N510-1989. Reference FSAR table 9.4-18 for conformance positions to ASME N510-1989. Inspection of the filters for gasket deterioration is performed periodically. Routine testing of the filters and adsorbers can be 6.2-70 REV 31 10/23

FNP-FSAR-6 performed during scheduled refuelings. Permanently installed inspection and sampling ports are provided and adequate space is provided adjacent to the filter housings for test equipment.

Testing of the inleakage characteristics of the penetration rooms is performed following major modifications or repair to the penetration rooms' boundary.

6.2.3.4.3 Containment Preaccess Filtration and Purge Systems The radiation monitors and actuation circuits associated with isolation of the containment purge supply and discharge in the event of high discharge activity levels are tested and inspected in accordance with the provisions of IEEE Standard 279-1971. Periodic testing of the radiation monitors and actuation circuits is discussed in the Technical Specifications. The testing and inspection of the isolation valves associated with this function are discussed in paragraph 6.2.4.4. The remaining portions of the containment preaccess filtration system are not required to reduce the radiological consequences of an accident. The portion of the purge exhaust up to the discharge of the filter is Seismic Category I to assure the pressure boundary integrity of the penetration room filtration system, containment isolation, and radiation detection.

6.2.3.4.4 Containment Ventilation Systems The containment ventilation systems are not required to reduce the radiological consequences of an accident.

6.2.3.5 Instrumentation Requirements 6.2.3.5.1 ECCS Recirculation Sump pH Control System The system is passive and requires no instrumentation.

6.2.3.5.2 Penetration Room Filtration System When aligned in the fuel handling mode, both trains will be started automatically by an actuation signal from the fuel handling area, initiated by either high radiation or low flow in the spent fuel pool exhaust system. When aligned in the normal operating mode, both trains of the penetration room filtration system are automatically started upon receipt of a containment phase "B" isolation actuation signal.

The following is displayed in the control room:

A. Position indication of all fan discharge valves and the recirculation line valves.

B. Temperature indication of the charcoal filter beds. High temperature is annunciated.

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FNP-FSAR-6 C. Differential pressure across each filter train.

D. Differential pressure across each fan.

E. Fan status for all fans.

F. Radiation level of exhaust air.

G. Differential pressure between the penetration rooms and the PRF equipment room. High penetration room pressure is annunciated.

H. Penetration room temperature.

In addition, all alarms are annunciated in the control room.

The actuating circuits and instrumentation for the penetration room filtration system will meet the criteria for a protection system as set forth in IEEE Standard, Criteria for Protection Systems for Nuclear Power Generating Stations, IEEE Standard 279-1971.

6.2.3.5.3 Containment Preaccess Filtration and Purge Systems The containment minipurge system operates during power operation to continuously purge the containment atmosphere. Particulate and gas monitor indications of the purge exhaust activity levels are used to determine routine releases from the containment. Radiation monitors in the purge exhaust duct are designed to IEEE Standard 279-1971.

During power operation, the containment particulate and gas monitor indications help determine the desirability of using the preaccess filtration system in addition to the minipurge system for preaccess cleanup.

The full containment purge system operates only during reactor shutdown. Releases from the plant vent are continuously monitored during this period with a radioactive particulate and gas monitor.

6.2.3.5.4 Containment Ventilation Systems Instrumentation applications associated with post-accident operation of the containment air coolers are discussed in paragraph 6.2.2.5.2. The remaining aspects of the operation of the containment ventilation systems are not safety related and are manually actuated. During normal operation, temperature indications verify the proper operation of the containment air cooling system, the CRDM cooling system, and the reactor vessel support cooling system. The refueling water surface ventilation system is used only during shutdown.

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FNP-FSAR-6 6.2.3.6 Materials Materials information regarding paints used within containment, including type and manufacturer's designation as well as surface areas covered, are provided in table 6.2-37.

Materials considerations are further described below:

A. Decomposition Products Materials used in or on ESFs have been chosen so that the decomposition products of the materials will not interfere with safe operation of engineered safety features.

B. Materials Compatibility Spray pumps initially provide borated water from the RWST with an approximate pH of 4.5. After dissolving the trisodium phosphate compound in the baskets, an equilibrium sump pH between 7.0 and 10.5 is attained. Studies have indicated no adverse effects on austenitic stainless steel stress corrosion crack initiation as a result of this short duration exposure to low pH spray. Sprays will be maintained after the equilibrium sump pH is reached to wash the low-pH solution from the containment materials. The carbon steel surfaces of the trisodium phosphate baskets are coated in accordance with the guidance in ANSI 101.2 (1972), Protective Coatings (Paints) for Light Water Reactor Containment Facilities.

6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2.4.1 Design Bases The containment isolation system is in conformance with the NRC acceptance criteria contained in General Design Criteria 54, 55, 56, and 57 and Regulatory Guide 1.11. The general design basis governing isolation valve requirements is as follows:

Leakage through all fluid penetrations not serving accident consequence limiting systems is minimized by a double barrier so that no single credible failure or malfunction of an active component results in loss of isolation or intolerable leakage. The installed double barriers take the form of closed piping systems, both inside and outside the containment, and various types of isolation valves.

Containment isolation in nonessential process lines occurs coincident with SI actuation signal.

Valves which isolate penetrations that are directly open to the containment, such as the purge valves and sump drain valves, are included in this group. Isolation of valves in essential process lines, such as reactor coolant pump cooling, occurs coincident with containment spray actuation signal. The closure signal for each containment isolation valve is shown in table 6.2-31. The closure signal for each SG isolation valve is shown in table 6.2-32. The conditions required for generation of containment isolation signals are outlined in section 7.3.

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FNP-FSAR-6 In accordance with GDC 54, fluid lines penetrating the containment are provided with isolation capability as follows:

Type I Each line connecting directly to the RCSs has two containment isolation valves. One valve is located inside the containment and is either an automatic valve, locked closed, or a check valve depending on the direction of normal flow. The second valve is located outside the containment and is either an automatic valve or is locked closed.

Type II Each line connecting directly to the containment atmosphere has two containment isolation valves. One valve is located inside the containment and is either an automatic valve, is locked closed, or is a check valve, depending on the direction of normal flow. The second valve is located outside the containment and is either an automatic valve or is locked closed.

Type III Each line that is not directly connected to the RCS nor is open to the containment atmosphere has at least one containment isolation valve located outside the containment. This valve is either an automatic valve or is locked closed, or is capable of remote manual operation.

The above isolation valve arrangements have been established to conform with GDC 55, 56, and 57. Fluid instrument lines have been classified in accordance with the design basis given above. The specific classification of each isolation valve arrangement is given in table 6.2-31.

The classification of each SG isolation valve is shown in table 6.2-32. Table 6.2-31 (sheet 6) delineates special valve arrangements for meeting the General Design Criteria.

Simple check valves are not used as isolation valves outside containment.

6.2.4.2 System Design Tables 6.2-31 and 6.2-32 list each fluid penetration and its isolation valves.

In addition to the containment isolation signals shown in tables 6.2-31 and 6.2-32, all automatic valves are provided with hand switches in the main control room for remote manual operation.

All valves outside the containment are provided with means for manual closure.

Inside the containment, containment isolation valves are located outside the missile barrier.

Outside containment, containment isolation valves are located in the penetration rooms as close as practical to the containment wall. (Any exceptions to this are specified in table 6.2-31.)

The valve closure times for all containment isolation valves listed in table 6.2-31 are shown in the table. Motor operators for valves installed within the containment have been designed and qualified for the post-LOCA environment as described in subsection 3.1.1.

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FNP-FSAR-6 Each containment penetration is part of a particular system which is listed in tables 6.2-31 and 6.2-32. The design conditions and applicable codes for the containment and SG isolation valves are described in their respective system sections.

The radiation dosage design requirements for the containment isolation valves are given in table 3.11-1.

The piping and valves forming the isolation system associated with each penetration are designed as Seismic Category I and have been analyzed for the dynamic effects of the SSE in addition to the normal operating pressure and temperature conditions. The main steam piping system has also been analyzed for the case in which one of the isolation valves is inadvertently closed when the line is at full flow; this condition does not adversely affect the integrity of the isolation system.

6.2.4.3 Design Evaluation The containment isolation system is designed for leaktightness and reliability of operation. The isolation scheme and valve types for each penetration have been selected in consideration of normal plant operating requirements as well as the isolation function.

Where two electric motor-operated or solenoid-operated valves are provided for isolation of a Type I or Type II line, these valves are actuated from separate power supplies. All solenoid actuated valves will close if electrical power to the solenoid is lost. Where air-operated valves are provided as containment isolation valves, loss of instrument air pressure from the valve operator closes the valve.

The motor-operated valves listed in table 6.2-31 will remain in the as-is position upon power failure. The use of motor operated or manual valves which fail as is upon loss of actuating power in lines penetrating the containment are based upon the consideration of what valve position ensures the greatest plant safety. Furthermore, each of these valves that fail as is are provided with redundant backup valves to ensure that no single failure will prevent the system as a whole from performing its isolation function, (e.g., a check valve inside the containment and motor-operated valve outside the containment or two motor-operated valves in series, each powered from a separate safeguards bus).

All piping and valves relied upon to perform a containment function are classified as Safety Class 2a or better. No single failure, active or passive, will prevent the system as a whole from performing its isolation function.

6.2.4.4 Tests and Inspections All isolation valves were shop tested for leakage and reliability of operation by the manufacturer.

Gate, globe, and check valve leakage did not exceed 3 cm3/in. of nominal valve size.

Diaphragm valves were tested to zero leakage requirements. The butterfly valves for the containment purge supply and return penetration were shop tested to zero leakage in accordance with MSS-SP-67 requirements for Type I valves.

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FNP-FSAR-6 Each valve may be tested after installation to ensure its operability and leaktightness in the piping system. All piping systems penetrating the containment have been provided with test connections and test vents, or have other provisions to allow periodic leak testing as required by Appendix J to 10 CFR 50. Throughout plant life, the containment isolation valves can be tested periodically. Valves which cannot be tested during plant operation are tested during scheduled shutdowns.

Table 6.2-38 lists the containment penetrations and the type of tests required.

Type C testing of the residual heat removal sump suctions and the containment spray sump suctions is not required. The justification for this is that these valves are required to be open at some time after the accident to effect immediate and long-term core cooling. Type C testing of the residual heat removal suctions from the reactor coolant system hot legs shares a common suction line with the residual heat removal sump suctions. These systems are closed systems outside containment, and are water filled during normal operation and following a LOCA. These systems are designed and constructed to ASME III, Class 2 and Seismic Category I requirements and, as such, they do not constitute a potential containment atmosphere leak path during or following a LOCA with a single-active failure of a system component. Should the valves leak slightly when closed, the fluid seal within the pipe or the closed piping system outside/inside containment would preclude release of containment atmosphere to the environs.

Table 6.2-39 lists the containment isolation valves, their location with respect to the containment, appropriate referenced drawings and/or FSAR figures, and the valve identification number. Table 6.2-40 lists the same information for the SG isolation valves. For the valves listed in table 6.2-39 and associated with table 6.2-38, which must be Type C tested to comply with Appendix J to 10 CFR 50, the fluid to be used will be in accordance with the Containment Leakage Rate Testing Program.

Testing requirements associated with containment isolation valves are discussed in the Technical Specifications.

6.2.4.5 Materials Material compatibility is discussed in appendix 6A.

6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT Following a DBA, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction, and radiolysis of aqueous solution in the core and sump. To ensure that the hydrogen concentration is maintained at a safe level, redundant recombiners are provided along with a backup post-accident venting system. A mixing system is provided to maintain accumulation below the flammability limit and a sampling system is provided to monitor the containment atmosphere.

6.2-76 REV 31 10/23

FNP-FSAR-6 The combustible gas control system is in conformance with the NRC acceptance criteria contained in General Design Criteria 41, 42, and 43 and Regulatory Guide 1.7. Post-LOCA hydrogen concentration in the containment would not reach the lower flammability limits of 4.0 volume percent, assuming that one electric hydrogen recombiner is started 1 day after the start of LOCA (see figure 6.2-94). However, no credit is taken for venting of the containment atmosphere.

6.2.5.1 Design Bases 6.2.5.1.1 Electric Hydrogen Recombiners The following design bases apply to the electric hydrogen recombiners:

A. The recombiners are designed to sustain all normal loads as well as accident loads including seismic loads and temperature and pressure transients from a design basis LOCA.

B. The recombiners are protected from damage by missiles or jet impingement from broken lines.

C. The recombiners are located away from high velocity air streams such as could emanate from fan cooler exhaust ports, or they are protected from direct impingement of such high velocity air streams by suitable barriers such as walls or floors.

D. The recombiners are designed for a lifetime of 40 years, consistent with that of the plant.(a)

E. All materials used in the recombiners are selected to be compatible with the environmental conditions inside the reactor containment during normal operation or during accident conditions.

F. The recombiners are powered from separate electrical power trains and can be powered from an emergency power source if necessary.

G. Process capacity is such that the containment hydrogen concentration will not exceed 4 volume percent based on the NRC TID release model as indicated in NRC Regulatory Guide 1.7: "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident."

a. The renewed operating licenses authorize an additional 20-year period of extended operation for both FNP units, resulting in a plant operating life of 60 years. In accordance with 10 CFR Part 54, appropriate aging management programs and activities have been initiated to manage the detrimental effects of aging to maintain functionality during the period of extended operation (see chapter 18).

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FNP-FSAR-6 6.2.5.1.2 Post-Accident Venting System The following design bases apply to the post-accident venting system:

A. The venting system is a Seismic Category I system, and is designed to sustain all normal loads as well as temperature and pressure transients from the design basis LOCA.

B. The venting system is protected from damage by missiles or jet impingement from broken pipes.

C. The venting inlet ducts are located in a well ventilated part of the containment.

D. The system is designed for a lifetime consistent with that of the reactor plant.

E. All materials in the venting system are selected to be compatible with the environmental conditions inside the reactor containment during normal operation or during accident conditions.

F. Process capacity such that the containment hydrogen concentration will not exceed 4 volume percent based on the NRC TID release model as indicated in NRC Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident."

6.2.5.1.3 Post-Accident Combustible Gas Sampling System The following design bases apply to the post-accident combustible gas sampling system:

A. The sampling system is a Seismic Category I system and is designed to sustain all normal loads as well as temperature and pressure transients from the design basis LOCA.

B. The sampling system is protected from damage by missiles or jet impingement from broken pipes.

C. The sampling system will provide samples to determine representative hydrogen concentrations inside the containment from at least two independent locations.

Hydrogen analyzers will be provided to continuously monitor the hydrogen concentration of the containment atmosphere following a LOCA. The system will have the capability to provide grab samples.

D. The combustible gas monitoring system is described in subsection 7.6.4.

E. All materials in the sampling system are selected to be compatible with the environmental conditions inside the reactor containment during normal operation or during accident conditions.

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FNP-FSAR-6 6.2.5.1.4 Post-Accident Mixing System The following design bases apply to the post-accident mixing system:

A. The mixing system is a Seismic Category I system and is designed to sustain all normal loads as well as temperature and pressure transients from a design basis LOCA.

B. The mixing system components are protected from damage by missiles or jet impingement from broken pipes.

C. The system is designed for maintenance free operation for a period of 100 days following a LOCA.

D. All materials in the post-accident mixing system are selected to be compatible with the environmental conditions inside the containment during normal operation or after an accident.

E. The active components are redundant, and the electric cables and instrument lines are separated so that no single failure can incapacitate the entire system.

F. The active components are powered from separate power supply trains and can be powered from an emergency power source if necessary.

6.2.5.2 System Design The total system for control of combustible H2 concentrations in the containment following a LOCA consists of a sampling system that provides containment atmosphere samples, electric hydrogen recombiners as the primary means of reducing H2 containment concentrations, a venting system which is used as a backup system to the recombiners, and a mixing system to maintain concentrations in the lower containment compartments below the lower flammability limit.

6.2.5.2.1 Electric Hydrogen Recombiners Redundant electrical recombiners, as shown on figure 6.2-90 and drawings D-175019 and D-205019, are located inside the containment upper compartment for controlling hydrogen concentrations following a DBA. The recombiners meet safety features requirements and the controls are located outside the containment. The recombiner units are located in the containment such that they process a flow of containment air containing hydrogen at a concentration which is generally typical of the average concentration throughout the containment.

To meet the requirements for redundancy and independence, two recombiners are provided.

Each recombiner is provided with a separate power panel and control panel and each is 6.2-79 REV 31 10/23

FNP-FSAR-6 powered from a separate bus. Each can be switched to the emergency power source if necessary. There is no interdependency between this system and the other ESFs.

Containment atmosphere is circulated by natural convection through a recombiner where hydrogen is removed by heating to a temperature sufficient to cause recombination with the containment oxygen.

The recombiner consists of a thermally insulated vertical metal duct with electric resistance metal sheathed heaters provided to heat a continuous flow of containment air (containing hydrogen) up to a temperature which is sufficient to cause a reaction between hydrogen and oxygen.

The recombiner is provided with an outer enclosure to keep out water coming from the containment spray system. The recombiner consists of an inlet preheater section, a heater recombination section, and a mixing chamber.

The unit is manufactured primarily of corrosion resistant, high temperature material for major structural components, except for the base which is carbon steel. The electric hydrogen recombiner uses conventional type electric resistance heaters sheathed with Incoloy-800, which is an excellent corrosion resistant material for this service. These heaters are designed to operate with sheath temperatures equal to those used in certain commercial heaters; however, these recombiner heaters operate at significantly lower power densities than is commercial practice.

Air is drawn into the recombiner by natural convection and passes first through the preheater section. This section consists of a shroud placed around the central heater section to take advantage of heat conduction through the walls to preheat the incoming air. This accomplishes the dual functions of reducing heat losses from the recombiner and of preheating the air.

The warmed air passes through an orifice plate and then enters the electric heater section where it is heated to between 1150°F - 1400°F, causing recombination to occur. Tests have verified that the recombination is not a catalytic surface effect associated with the heaters but occurs due to the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, saturating of the unit by fission products will not occur. The heater section consists of five assemblies of electric heaters stacked vertically. Each assembly contains individual heating elements. Operation of the unit is virtually unaffected should a few individual heating elements fail to function properly. Table 6.2-33 gives the recombiner design parameters.

The recombiner, power supply panel, and control panel are shown schematically in figure 6.2-91. The power panel for the recombiner is located in the auxiliary building, and contains an isolation transformer plus an SCR controller to regulate power into the recombiner.

This equipment is not exposed to the post-LOCA environment.

The control panel is located in the control room. To control the recombination process, the correct power input which will bring the recombiner above the threshold temperature for recombination will be set on the controller. The controller setting will be set at the control panel to maintain the recombiner air temperature between 1150°F and 1400°F. This power setting 6.2-80 REV 31 10/23

FNP-FSAR-6 and recombiner air temperature range will cover variations in containment temperature, pressure, and hydrogen concentration in the post-LOCA environment. For an equipment test and periodic checkout, a thermocouple readout instrument is also provided in the control panel for monitoring temperatures in the recombiner.

Results of testing a prototype of the electric hydrogen recombiner are given in WCAP-7820, Supplement 3, "Electrical Hydrogen Recombiner for Water Reactor Containments," March 1974.

A power setting of approximately 48.9 kW for the prototype recombiners was determined to be satisfactory for PWR containments. However, production recombiners and power supplies have been designed to achieve up to 75 kW to provide operational flexibility for containments and design conditions which differ from the prototype testing.

6.2.5.2.2 Post-Accident Venting System The post-accident containment venting system, as shown on drawings D-175019 and D-205019, consists of a supply line through which hydrogen free air can be admitted to the containment, and an exhaust line through which hydrogen bearing gases may be vented from the containment. The gases are filtered through HEPA and charcoal filters to limit discharge of particulates and iodine. Piping and valving in the exhaust line are Safety Class 2A starting inside the containment, proceeding up to and including the venting filtration unit outside the containment. Equipment and piping beyond the filter unit are Safety Class 2B. Design parameters are given in table 6.2-34.

The supply line in the containment is located in a missile protected area and is terminated so as to prevent either spray or sump water from entering the pipe. The exhaust line in the containment is located in a missile protected area and is terminated in well ventilated areas in a manner which prevents either spray or sump water from entering the pipe.

The containment hydrogen concentration and hydrogen generation rate are determined every 24 h after the initiation of the accident. When the projected hydrogen concentration for the next 24-h period has increased to 3.5 volume percent, containment pressurization is initiated. Using plant air compressors, hydrogen free air will be pumped into the containment until the required containment pressure is reached. The air supply will then be stopped and the supply line isolated. The addition of air to pressurize the containment will lower the hydrogen concentration and permit the containment to be isolated until sampling indicates that the next-day projected hydrogen concentration will be above 3.5 volume percent when venting is necessary.

Venting is then started by opening the motor-operated valves in the line and adjusting the manually controlled flow control valve to obtain the required flow. The flowrate required to maintain the hydrogen concentration less than 4 volume percent of the containment volume by venting the containment is determined from the containment hydrogen concentration and the hydrogen generation rate. When the operation is complete, the vent line is isolated by the valve outside the containment. The motor-operated valve inside the containment will also be closed.

This process of containment pressurization followed by venting may be repeated as necessary to maintain the hydrogen concentration below 4 volume percent.

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FNP-FSAR-6 6.2.5.2.3 Post-Accident Combustible Gas Sampling System The post-accident sampling system is shown schematically on drawings D-175019 and D-205019 and comprises two independent trains consisting of sample inlet lines, a removable sample container, a hydrogen analyzer, and a return line routed to the containment. Piping and equipment from the innermost containment isolation valves outward are Safety Class 2A.

Piping and equipment inside the innermost containment isolation valves are Safety Class 2B.

Design parameters are given in table 6.2-35.

The supply lines in the containment are located in missile protected areas and are terminated so as to prevent either spray or sump water from entering the pipe. The sample exhaust line in the containment is located in a missile protected area and is terminated in a well ventilated area in a manner which prevents either spray or sump water from obstructing the pipe.

A containment air sample can be taken from either of four independent locations within the containment. Sample point locations are as noted on drawings D-175019 and D-205019. Each of the sample lines is routed independently inside the containment up to and including motor operated globe valves, which serve as containment isolation barriers. Immediately downstream of these valves the four lines are joined into two and are routed through two separate containment penetrations, one penetration for each train. Outside the containment, the sample flows through a motor-operated isolation valve, through a line which includes pressure indication, and through a hydrogen analyzer. Upstream of the analyzer there is a removable sample vessel, located in parallel with the sample discharge line, which can be used to provide grab samples. The discharge line passes through a motor-operated isolation valve located outside the containment. The line is routed through a containment penetration and discharges into the containment atmosphere through a motor-operated globe valve serving as a containment isolation valve.

6.2.5.2.4 Post-Accident Containment Mixing System Drawings D-175019 and D-205019 include the P&ID for the post-LOCA containment mixing system.

Four fans are provided to circulate containment atmosphere among the lower containment compartments, the upper containment volume above the operating floor, and the reactor cavity.

These fans are placed on the hatch covers above the three reactor coolant pumps.

Of these four fans, two are powered from each of two separate and redundant emergency power trains. These fans take suction from the equipment compartments and the lower plenum-like volume over the containment sump and discharge upward, thereby establishing flow downward around the periphery of the containment, through the lower containment volume, and upward out of the lower containment volume through the SG compartments. A labyrinth at each fan suction will protect the fans and motors from missiles that might be generated within the secondary missile shield inside containment.

Two additional fans ventilate the reactor cavity to ensure that this volume is available for the dilution of containment hydrogen and to maintain hydrogen concentrations in this volume in 6.2-82 REV 31 10/23

FNP-FSAR-6 equilibrium with that of the remainder of the containment. These fans discharge into the reactor cavity through a circular header embedded in the cavity wall at an elevation approximately coincident with that of the lower reactor vessel head. This flow is discharged from the cavity upward around the reactor vessel and outward through the incore instrument chase. The two fans, one each of which will be powered from one of two separate and redundant emergency power trains, will take suction from the periphery of the containment just below the operating floor. These fans and motors are located outside the secondary missile shield inside the containment.

The post-LOCA mixing fans and the reactor cavity hydrogen dilution fans are designed to Seismic Category I requirements. The design capacities of these fans are given in table 6.2-36.

The internal structures of the containment building have been designed to provide vertical cylindrical compartments around each SG which project upward from a lower plenum covering much of the containment sump. (See figures 6.2-92 and 6.2-93.) Following a LOCA, the SGs and reactor coolant piping (short term) and the sump fluid (long term) represent heat sources that establish and maintain natural convective flows upward out of the lower containment volume through the SG compartments. This, coupled with the generally lower air temperatures provided by the containment air coolers around the periphery of the containment, will establish downward flow from the upper containment volume through grating located around the periphery of the containment. This convective flow into and out of the lower containment, coupled with the mixing action of the sprays and containment air coolers above the operating floor, will complement the action of the post-LOCA mixing fans. The undersides of all floors inside the containment have been sloped to augment this natural convection. Inverted pockets that might tend to encourage stagnant accumulations of containment atmosphere have been avoided by design.

To determine the design capacities of the post-LOCA mixing fans, the containment was sectionalized into specific compartments as shown in figures 6.2-92 and 6.2-93. The lower compartment was defined as the summation of compartments 2, 3, 4, 5, 6, and 8.

Figures 6.2-96 and 6.2-97 give the calculated volume percent of hydrogen in each compartment as a function of time assuming no intercompartmental mixing. Figures 6.2-98 and 6.2-83 give the hydrogen generation rates as a function of time. For each compartment, the ventilation sweep rate required to maintain the compartmental hydrogen concentration below 3.5 volume percent was calculated. The post-LOCA mixing fans were selected so that they have a design flow that exceeds the calculated required sweep rate for the lower compartments.

6.2.5.3 Design Evaluation 6.2.5.3.1 Electric Hydrogen Recombiners The prediction of hydrogen generation following a LOCA is shown in figures 6.2-83 and 6.2-98, which demonstrate that the hydrogen production rate decreases with time after the accident. As discussed in paragraph 15.4.1.6.5, and as can be determined from these figures, the total hydrogen accumulation can exceed the lower flammability limit of 4 volume percent and control measures are necessary to prevent hydrogen accumulation to this limit. The electric 6.2-83 REV 31 10/23

FNP-FSAR-6 recombiner provides the means to prevent unsafe levels of hydrogen concentration from being reached in the containment following a LOCA.

For the purpose of showing that the electric recombiner is capable of maintaining safe hydrogen concentrations, analysis was performed using the NRC Regulatory Guide No. 1.7 Model. The result is shown in figure 6.2-94. The Regulatory Guide No. 1.7 Model is based upon assuming a fission product activity release specified in TID-14844 and the values for post-accident hydrogen generation specified in this guide. The results using the Westinghouse model are also shown on this figure.

Each electric recombiner is capable of continually processing a minimum of 100 sft3/min (standard conditions of 68°F and 1 ATM) of containment atmosphere. All of the hydrogen contained in the processed atmosphere is converted to steam, thus reducing the overall containment hydrogen concentration. The hydrogen concentration in the containment was calculated for the models described above based on a recombiner capability of processing 93 sft3/min of containment atmosphere at modeled conditions of 32°F and 1 ATM. This calculation shows that the maximum hydrogen concentration will be much less than the lower flammability limit of 4 volume percent if the recombiner is started 1 day following the accident.

Therefore, one of these units meets the design criterion of maintaining a safe hydrogen concentration with considerable margin, and the second unit provides the redundance of a system of equal capability on a redundant power supply.

The peak hydrogen concentration is indicative of that time when the amount of hydrogen being generated is equal to the amount of hydrogen being reprocessed. Since the production rate of hydrogen decreases with increasing time following the accident, once this peak has been reached, the recombiner will be processing hydrogen at a faster rate than it is being produced.

This will result in an overall reduction of the hydrogen concentration inside the containment.

Thus, once the peak has been reached, the electric recombiner provides a continually increasing margin between the containment hydrogen concentration and the lower flammability limit of 4 volume percent.

The recombiners are protected from damage by missiles. The unit is designed to sustain all normal loads as well as accident loads such as seismic loads and temperature and pressure transients from a LOCA.

The recombiners are manually initiated devices and, although IEEE-279 is not a design requirement, they do meet the single failure, separation testability, qualification, and manual initiation criteria of IEEE-279.

6.2.5.3.2 Post-Accident Venting System As a backup system to the electric recombiners for controlling combustible gas concentrations in the containment following a LOCA, the venting system has a process capacity such that H2 concentration will not exceed 4 volume percent. This process capacity permits intermittent operation of the system as deemed necessary by the operator. A curve of hydrogen 6.2-84 REV 31 10/23

FNP-FSAR-6 concentration as a function of time for the containment purge mode of operation is provided in figure 6.2-95.

The venting system is designed to be protected from missiles, to sustain all normal as well as accident loads without loss of function.

An analysis of the offsite doses due to operation of the venting system is presented in section 15.4.

6.2.5.3.3 Post-Accident Combustible Gas Sampling System The post-accident containment sampling system is designed to obtain post-accident containment atmosphere grab samples for analysis, to monitor the hydrogen concentration in the containment atmosphere following a LOCA, and to provide a method of monitoring hydrogen recombiner operation. Samples may be obtained from any of four independent locations within the containment in the containment dome, above the containment cooler units, below hydrogen recombiner No. 1B, and in a lower compartment. Locating sample points at different elevations in all quadrants of the containment provides the capability to obtain a representative sample of the containment atmosphere.

The post-accident containment sampling system is protected from missiles and is designed to sustain all normal and accident loads without loss of function. All sample lines are terminated so as to prevent spray or sump water from entering the line. All sample lines within the containment are sloped to enhance drainage of any accumulated moisture and all potential water pocket areas are avoided by design.

6.2.5.3.4 Post-Accident Mixing System In order to maintain compartment hydrogen accumulations below the lower flammability limit, the units are sized to limit the hydrogen accumulation to 3.5 volume percent based on NRC Regulatory Guide No. 1.7 assumptions.

The system can be operated continuously or intermittently depending on containment conditions and is designed to be protected from missiles and to sustain all normal as well as accident loads without loss of function.

6.2.5.4 Tests and Inspections 6.2.5.4.1 Electric Hydrogen Recombiners The electric hydrogen recombiners have undergone many tests as part of the Westinghouse development program. These tests encompassed the initial analytical studies, laboratory proof of principle tests, and full scale prototype testing. The full scale prototype tests included the effects of:

6.2-85 REV 31 10/23

FNP-FSAR-6

Varying hydrogen concentrations.

Alkaline spray atmosphere.

Steam effects.

Convection currents.

A detailed discussion of these tests is given in WCAP-7820, Supplement 1.

Post-operational tests and inspections are performed in accordance with the Technical Specifications and the Technical Requirements Manual. Testing will be performed by inputting power to the recombiner and observing the thermocouple readout instruments on the control panel. Inspections will be performed to verify the integrity of all electrical connections to assure the capability of the recombiner to perform its function.

6.2.5.4.2 Post-Accident Venting and Sampling Systems Test connections are provided to allow testing of the systems. Testing will periodically demonstrate that the systems will perform their intended functions.

Periodic testing of the post-accident venting system is based on testing guidance from ASME N510-1989. Field surveillance testing, as provided in ASME N510-1989, is based on nuclear filtration units designed and installed to the requirements of ASME N509-1989. Since the post-accident venting filter units were designed to earlier standards, these filter units do not possess all of the design features required to provide for field testing as described in ASME N510-1989. Conformance to ASME N510-1989 is clarified in table 9.4-20.

6.2.5.4.3 Post-Accident Containment Mixing System To demonstrate that this system will perform as required, tests will be performed in accordance with the Technical Specifications. Pressure instrumentation is provided to demonstrate that fan developed head has not deteriorated. Inspections will be performed to verify the integrity of all electrical connections to assure the capability of the fans to perform their function.

Post-accident mixing system fans' and motors' capability to operate during LOCA is certified by the manufacturer. The qualification test is the same as the containment air cooling units qualification test which is described under paragraph 6.2.2.4.2.

6.2.5.5 Instrumentation Requirements The recombiners will be started manually after a LOCA, and the mixing system is started automatically upon receipt of an SI signal. The sampling system will be used in obtaining 6.2-86 REV 31 10/23

FNP-FSAR-6 containment atmosphere samples that will indicate when the recombiners or the venting system should be actuated. This sample can be taken from any of four well ventilated locations within the containment.

The recombiners do not require any instrumentation inside the containment for proper operation after a LOCA.

6.2.5.6 Materials 6.2.5.6.1 Electric Recombiner The materials of construction for the electric recombiner are selected for their compatibility with the post-LOCA environment.

The major structural components are manufactured from 300-Series stainless steel except for the base which is carbon steel. Incoloy-800 is used for the heater sheaths and Inconel-600 for other parts such as the heat duct, which operates at high temperature.

There are no radiolytic or pyrolytic decomposition products from these materials. The carbon steel base of the recombiner unit is coated with a paint that satisfies the requirements of ANSI 101.2 (1972), "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities."

Testing of the electric recombiner in the post-LOCA spray environments is reported in WCAP-7820 (Supplement 1).

6.2.5.6.2 Post-Accident Venting and Sampling System The materials of construction of these systems are chosen as to their location relative to the reactor containment. All components inside containment, and up to and including the containment isolation valves outside are fabricated of stainless steel. There are no pyrolytic or radiolytic decomposition products from this material; thus, these systems have no effect on any other safety feature system.

6.2.5.6.3 Post-Accident Containment Mixing System All components of this system are fabricated from stainless steel. There are no pyrolytic or radiolytic decomposition products of this material, and this system has no effect on any other safety feature system.

6.2-87 REV 31 10/23

FNP-FSAR-6 REFERENCES

1. Strikovich, M. A., et al., Atomnoya Energiya, Volume 17, No. 1, pp. 45-49, (Translation in UDE - 621.039.562.5), July 1964.

2. Parsly, L. F., Jr., "Design Considerations of Reactor Containment Spray Systems - Part VI," ORNL-TM-2412, Part 6, 1969.

3. Parsly, L. F., Jr., "Design Considerations of Reactor Containment Spray Systems - Part VII," ORNL-TM-2412, Part 7, 1970.

4. Eggleton, A. E. J., "A Theoretical Examination of Iodine Water Partition Coefficient,"

AERE (R) - 4887, 1967.

5. Deleted

6. Burchsted, C. A., and Fuller, A. B., Design, Construction, and Testing of High Efficiency Air Filtration Systems for Nuclear Application," ORNL-NSIC-65.

7. Westinghouse LOCA Mass and Energy Release Model for Containment Design -

March 1979 Version, WCAP-10325-P-A, May 1983 (Proprietary), WCAP-10326-A (Nonproprietary).

8. Docket No. 50-315, Amendment No. 126, Facility Operating License No. DPR-58 (TAC No. 7106), for D. C. Cook Nuclear Plant Unit 1, June 9, 1989.

9. EPRI 294-2, Mixing of Emergency Core Cooling Water with Steam: 1/3-Scale Test and Summary, WCAP-8423, Final Report, June 1975.

10. Westinghouse Mass and Energy Release Data for Containment Design, WCAP-8264-P-A, Rev. 1, August 1975 (Proprietary), WCAP-8312-A, (Nonproprietary).

11. ANSI/ANS-5.1 1979, American National Standard for Decay Heat Power in Light Water Reactors, August 1979.

12. McAdam, W. H., Heat Transmission, McGraw-Hill, 3rd Edition, 1954, p. 172.

13. NL-91-0873, Joseph M. Farley Nuclear Plant Leak-Before Break Approval, 9/13/91.

14. Deleted

15. Schmitt, R. C., Binham, G. E., and Norberg, J. A., "Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment," Idaho Nuclear Corp., NTIS TID-4500 1970.

16. "Performance and Sizing of Dry Pressure Containment," BN-TOP-3, Rev. 3, Bechtel Power Corporation, August 1975.

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17. Chapman, A. J., "Heat Transfer," MacMillan, 1967.

18. EPRI/NAI 8907-02, Rev. 6, GOTHIC Containment Analysis Package User Manual, Version 5.

19. 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.

20. U. S. Nuclear Regulatory Commission, Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels, NUREG/CR-0009, 10/78.

21. U. S. Atomic Energy Commission, A Review of Mathematical Models for Predicting Spray Removal of Fission Products in Reactor Containment Vessels, WASH-1329, June 15, 1974.

22. EPRI/NAI 8907-06, Rev. 5, GOTHIC Containment Analysis Package Technical Manual, Version 5.

23. Mass and Energy Releases Following a Steam Line Rupture, WCAP-8822 (Proprietary), WCAP-8860 (Nonproprietary), September 1976; Supplement 1 -

Calculations of Steam Superheat in Mass/Energy Releases Following a Steamline Rupture, WCAP-8822-S1-P-A (Proprietary), WCAP-8860-S1-A (Nonproprietary),

September 1986; Supplement 2 - Impact of Steam Superheat in Mass/Energy Releases Following a Steamline Rupture for Dry and Sub-Atmospheric Containment Designs, WCAP-8822-S2-P-A (Proprietary), WCAP-8860-S2-A (Nonproprietary), September 1986.

24. Burnett, T. W. T., et al, LOFTRAN Code Description, WCAP-7907-P-A (Proprietary),

WCAP-7907-A (Non-Proprietary), April 1984.

25. Huegel, D. S., et al., RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, WCAP-14882 (Proprietary),

June 1997.

26. Safety Evaluation for Amendment No. 137 to NPF-2 and Amendment No. 129 to NPF-8, April 29, 1998 (Power Uprate).

27. Letter from Herbert N. Berkow, Director (NRC) to James A. Gresham (Westinghouse),

Acceptance of Clarifications of Topical Report WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version (TAC No. MC7980), October 18, 2005. [ADAMS Accession No. ML052660242]

28. U.S. Nuclear Regulatory Commission, U.S. Nuclear Regulatory Commission Verification Letter for the Pressurized Water Reactors Owners Group Topical Report PWROG-17034, Revision 0, Evaluation of the WCAP-10325-P-A Westinghouse LOCA Mass & Energy Release Methodology (EPID L-2018-TOP-0013), May 15, 2020.

[ADAMS Accession No. ML20133P693]

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29. PWR Owners Group, Evaluation of the WCAP-10325-P-A Westinghouse LOCA Mass &

Energy Release Methodology, PWROG-17034-P-A, Revision 0, March 2018.

30. GOTHIC Thermal Hydraulic Analysis Package User Manual, Version 8.1 (QA),

September 2014.

31. GOTHIC Thermal Hydraulic Analysis Package Technical Manual, Version 8.1 (QA),

September 2014.

6.2-90 REV 31 10/23

FNP-FSAR-6 TABLE 6.2-1 PRINCIPAL CONTAINMENT DESIGN PARAMETERS Characteristics Data Containment design pressure (psig) 54 Containment design temperature (°F) 280 Internal dimensions Cylindrical wall diameter (ft) 130 Cylindrical wall height (ft) 139 Curved dome height (ft) 43.5 Volumes Gross internal volume (ft3) 2.35 x 106 Net free internal volume (ft3) 2.0 x 106 Containment design leak rate First 24 h, percent of containment free volume per day 0.15 After first day, percent per day 0.075 REV 21 5/08

FNP-FSAR-6 TABLE 6.2-2 (SHEET 1 OF 5)

HEAT SINK GEOMETRIC DATA(a)

Heat Sink 1 - Containment Cylinder and Dome 74,908 ft2 Exposure

1. Containment Atmosphere

2. Outside Atmosphere Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.25 Air Gap 0.00204 Concrete 45.0 Heat Sink 2 - Penetration Plates & Liner Stiffners 3,802 ft2 Exposure

1. Containment Atmosphere

2. Outside Atmosphere Material Thickness (in.)

Paint/Primer 0.0084 Carbon Steel 0.51 Air Gap 0.00204 Concrete 45.0 Heat Sink 3 - Unlined Concrete (excluding reactor support) 60,375 ft2 Exposure

1. Containment Atmosphere