WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, Chapter 6.0 - Engineered Safety Features
| ML16203A375 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/31/2016 |
| From: | Wolf Creek |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16203A351 | List:
|
| References | |
| WM 16-0008 | |
| Download: ML16203A375 (676) | |
Text
WOLF CREEK TABLE OF CONTENTS CHAPTER 6.0 ENGINEERED SAFETY FEATURES
Section Page
6.0 ENGINEERED SAFETY FEATURES 6.1-1
6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1-3
6.1.1 METALLIC MATERIALS 6.1-3
6.1.1.1 Materials Selection and Fabrication 6.1-3 6.1.1.2 Composition, Compatibility, and Stability of Containment and Core Spray Coolants 6.1-6
6.1.2 ORGANIC MATERIALS 6.1-7 6.1.3 POST-ACCIDENT CHEMISTRY 6.1-8 6.
1.4 REFERENCES
6.1-8
6.2 CONTAINMENT SYSTEMS 6.2-1
6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1
6.2.1.1 Containment Structure 6.2-1 6.2.1.2 Containment Subcompartments 6.2-15 6.2.1.3 Mass and Energy Release Analyses for 6.2-21 Postulated Loss-of-Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for 6.2-29 Postulated Secondary Systems Pipe
Ruptures Inside Containment 6.2.1.5 Minimum Containment Pressure Analysis 6.2-40 for Performance Capability Studies on
Emergency Core Cooling System 6.2.1.6 Tests and Inspections 6.2-42 6.2.1.7 Instrumentation Requirements 6.2-42
6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-42
6.2.2.1 Containment Spray System 6.2-42 6.2.2.2 Containment Cooling System 6.2-56
6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN 6.2-63 6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2-63
6.2.4.1 Design Bases 6.2-63 6.2.4.2 System Description 6.2-66 6.2.4.3 Safety Evaluation 6.2-69 6.2.4.4 Tests and Inspections 6.2-70 6.2.4.5 Instrumentation Application 6.2-71
6.0-i Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Page
6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-72
6.2.5.1 Design Bases 6.2-73 6.2.5.2 System Design 6.2-74 6.2.5.3 Safety Evaluations 6.2-83 6.2.5.4 Tests and Inspections 6.2-87 6.2.5.5 Instrumentation Requirements 6.2-88
6.2.6 CONTAINMENT LEAKAGE TESTING 6.2-89
6.2.6.1 Containment Integrated Leakage Rate Test 6.2-89 (Type A Test) 6.2.6.2 Containment Penetration Leakage Rate Tests 6.2-92 (Type B Tests) 6.2.6.3 Containment Isolation Valve Leakage Rate 6.2-93 Tests (Type C Tests) 6.2.6.4 Scheduling and Reporting of Periodic Tests 6.2-96 6.2.6.5 Special Testing Requirements 6.2-96
6.
2.7 REFERENCES
6.2-96
6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1
6.3.1 DESIGN BASES 6.3-1
6.3.1.1 Safety Design Basis 6.3-1 6.3.1.2 Power Generation Design Basis 6.3-2
6.3.2 SYSTEM DESCRIPTION 6.3-3
6.3.2.1 General Description 6.3-3 6.3.2.2 Equipment and Component Descriptions 6.3-4 6.3.2.3 Applicable Codes and Construction Standards 6.3-17 6.3.2.4 Material Specifications and Compatibility 6.3-18 6.3.2.5 System Reliability 6.3-18 6.3.2.6 Protection Provisions 6.3-23 6.3.2.7 Provisions for Performance Testing 6.3-23 6.3.2.8 Manual Actions 6.3-23
6.3.3 SAFETY EVALUATION 6.3-25 6.3.4 TESTS AND INSPECTIONS 6.3-37
6.3.4.1 ECCS Performance Tests 6.3-37 6.3.4.2 Reliability Tests and Inspections 6.3-38
6.0-ii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Page
6.3.5 INSTRUMENTATION REQUIREMENTS 6.3-41
6.3.5.1 Temperature Indication 6.3-41 6.3.5.2 Pressure Indication 6.3-42 6.3.5.3 Flow Indication 6.3-43 6.3.5.4 Level Indication 6.3-43 6.3.5.5 Valve Position Indication 6.3-44
6.3.6 REFERENCE 6.3-45
6.4 HABITABILITY SYSTEMS 6.4-1
6.4.1 DESIGN BASES 6.4-1
6.4.1.1 Safety Design Bases 6.4-1 6.4.1.2 Power Generation Design Bases 6.4-3
6.4.2 SYSTEM DESIGN 6.4-3
6.4.2.1 Definition of Control Room Envelope 6.4-3 6.4.2.2 Ventilation System Design 6.4-3 6.4.2.3 Leaktightness 6.4-4 6.4.2.4 Interaction With Other Zones and Pressure-Containing Equipment 6.4-4 6.4.2.5 Shielding Design 6.4-5
6.4.3 SYSTEM OPERATIONAL PROCEDURES 6.4-5 6.4.4 DESIGN EVALUATIONS 6.4-5 6.4.5 TESTS AND INSPECTIONS 6.4-6 6.4.6 INSTRUMENTATION REQUIREMENT 6.4-7 6.4.7 REFERENCE 6.4-8
6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5-1
6.5.1 ENGINEERED SAFETY FEATURE (ESF) FILTER SYSTEMS 6.5-1
6.5.1.1 Design Basis 6.5-1 6.5.1.2 System Design 6.5-2 6.5.1.3 Safety Evaluation 6.5-3 6.5.1.4 Tests and Inspections 6.5-3 6.5.1.5 Instrumentation Requirements 6.5-3 6.5.1.6 Materials 6.5-3
6.5.2 CONTAINMENT SPRAY SYSTEM 6.5-3
6.5.2.1 Design Basis 6.5-4 6.5.2.2 System Design 6.5-4
6.0-iii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Page
6.5.2.3 Safety Evaluation 6.5-8 6.5.2.4 Tests and Inspections 6.5-10 6.5.2.5 Instrumentation Requirements 6.5-11 6.5.2.6 Materials 6.5-12
6.5.3 FISSION PRODUCT CONTROL SYSTEMS 6.5-13
6.5.3.1 Primary Containment 6.5-13 6.5.3.2 Secondary Containment 6.5-14
6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM 6.5-14 6.
5.5 REFERENCES
6.5-14
App. 6.5A IODINE REMOVAL MODELS FOR THE CONTAINMENT SPRAY SYSTEM 6.5A-1
6.5A.1 PARTICULATE IODINE MODEL 6.5A-2 6.5A.2 ELEMENTAL IODINE MODEL 6.5A-3 6.5A.3 REFERENCES 6.5A-10
6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS 6.6-1
6.6.1 COMPONENTS SUBJECT TO INSPECTION 6.6-1 6.6.2 ACCESSIBILITY 6.6-2 6.6.3 EXAMINATION TECHNIQUES AND PROCEDURES 6.6-3 6.6.4 INSPECTION INTERVALS 6.6-3 6.6.5 EXAMINATION CATEGORIES AND REQUIREMENTS 6.6-4 6.6.6 EVALUATION OF EXAMINATIONS 6.6-4 6.6.7 SYSTEM PRESSURE TEST 6.6-5 6.6.8 AUGMENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURE 6.6-5
6.0-iv Rev. 29 WOLF CREEK
WOLF CREEK
TABLE OF CONTENTS (Continued)
LIST OF TABLES
Number Title
6.1-1 ESF Materials of Construction
6.1-2 Design Comparison to Regulatory Positions of
Regulatory Guide 1.54, Revision 0, Dated June 1973, Titled "Quality Assurance Requirements for Protective
Coatings Applied to Water-Cooled Nuclear Power
Plants"
6.1-3 Containment Components-Coating Schedule
6.1-4 Design Comparison to Regulatory Positions of
Regulatory Guide 1.44, Revision 0, Date May 1973, Titled "Control of the Use of Sensitized Stainless
Steel"
6.1.5 Design Comparison to Regulatory Positions of
Regulatory Guide 1.37, Revision 0, Dated March 1973, Titled "Quality Assurance Requirements for Cleaning
of Fluid Systems and Associated Components of Water-
Cooled Nuclear Power Plants"
6.1-6 Design Comparison to Regulatory Positions of
Regulatory Guide 1.36, Revision 0, Dated February
1973, Titled, "Nonmetallic Thermal Insulation for
Austenitic Stainless Steel"
6.1-7 Design Comparison to Regulatory Positions of
Regulatory Guide 1.50, Revision 0, Dated May 1973, Titled "Control of Preheat Temperatures for Welding
of Low-Alloy Steel"
6.1-8 Design Comparison to Regulatory Positions of
Regulatory Guide 1.71, Revision 0, Dated December
1973, Titled, "Welder Qualification for Areas of
Limited Accessibility"
6.1-9 Design Comparison to Regulatory Positions of
Regulatory Guide 1.31, Revision 3, Dated April 1978,
Titled, "Control of Ferrite Content on Stainless
Steel Weld Metal"
6.1-10 Table of Lubricants Inside Containment
6.0-v Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.1-1 Spectrum of Postulated Loss-of-Coolant Accidents
6.2.1-2 Principal Containment Design Parameters
6.2.1-3 Engineered Safety Features Design Parameters for
Containment Analysis
6.2.1-4 Containment Passive Heat Sink Parameters
6.2.1-5 Containment and Reactor Coolant System Initial
Conditions for Containment Analysis
6.2.1-6 Chronology of Events DEPSG Break W/Min SI
6.2.1-7 Chronology of Events DEPSG Break W/Max SI
6.2.1-8 Comparative Results: Summary of Results of
Containment Pressure and Temperature Analysis for
the Spectrum of Postulated Accidents
6.2.1-9 Containment Mass and Energy Balance DEPSG with
Minimum Safety Injection
6.2.1-10 Containment Mass and Energy Balance DEPSG Break with
Maximum Safety Injection
6.2.1-11 Additional Mass and Energy Release-LOCA
6.2.1-12 Reactor Cavity Cold Leg 150 Square Inch Break Break
Mass Flow and Energy Flow
6.2.1-13 Hot Leg Longitudinal Split Break 763 Square Inches
Break Mass Flow and Energy Flow
6.2.1-14 Limited Area Circumferential Break Pump Suction 436
Square Inches Break Mass Flow and Energy Flow
6.2.1-15 Limited Area Circumferential Break-Cold Leg 236
Square Inches Break Mass Flow and Energy Flow
6.2.1-16 Pressurizer Surge Line Double-Ended Guillotine Break
Break Mass Flow and Energy Flow
6.2.1-17 Reactor Cavity Pressure-Temperature Analysis
Subcompartment Nodal Description
6.2.1-18 Reactor Cavity Analysis Subcompartment Vent Path
Description
6.0-vi Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.1-19 Volumes and Vent Areas of Waterbag Compartments
6.2.1-20 Vent Areas and L/As for Waterbag Compartment
6.2.1-21 Reactor Cavity Analysis Coefficients for
Determination of Forces and Moments on the Reactor
Pressure Vessel
6.2.1-22 Steam Generator Loop Compartment Analysis
6.2.1-23 Steam Generator Loop Compartment Analysis
6.2.1-24 Steam Generator Loop Compartment Analysis Force
Coefficients for Steam Generator
6.2.1-25 Steam Generator Loop Compartment Analysis Force
Coefficients on Reactor Coolant Pump
6.2.1-26 Pressurizer Compartment Analysis
6.2.1-27 Pressurizer Compartment Analysis
6.2.1-28 Blowdown Mass and Energy Release Double-Ended Pump
Suction Guillotine
6.2.1-29 Blowdown Mass and Energy Release 0.6 Double-Ended
Pump Suction Guillotine
6.2.1-30 Blowdown Mass and Energy Release Three-Foot Squared
Pump Suction Split
6.2.1-31 Blowdown Mass and Energy Release Double-Ended Hot
Leg Guillotine
6.2.1-32 Blowdown Mass and Energy Release Double-Ended Cold
Leg Guillotine
6.2.1-33 Reflood Mass and Energy Release Double-Ended Pump
Suction Guillotine (Minimum Safety Injection)
6.2.1-34 Reflood Mass and Energy Release Double-Ended Pump
Suction Guillotine (Maximum Safety Injection)
6.2.1-35 Post Blowdown Mass and Energy Release 0.6 Double-
Ended Pump Suction Guillotine
6.2.1-36 Post-Blowdown Mass and Energy Release Three-Foot
Squared Pump Suction Split
6.0-vii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.1-37 Post Blowdown Mass and Energy Release Double-Ended
Hot Leg Guillotine
6.2.1-38 Post-Blowdown Mass and Energy Release Double-Ended
Cold Leg Guillotine
6.2.1-39 Post-Reflood Mass and Energy Release Double-Ended
Pump Suction Guillotine (Minimum Safety Injection)
6.2.1-40 Post-Reflood Mass and Energy Release Double-Ended
Pump Suction Guillotine (Maximum Safety Injection)
6.2.1-41 Depressurization Energy Double-Ended Pump Suction
Guillotine (Minimum Safety Injection)
6.2.1-42 Depressurization Energy Double-Ended Pump Suction
Guillotine (Maximum Safety Injection)
6.2.1-43 Reactor Coolant System Mass and Energy Balance
Double-Ended Pump Suction Guillotine (Minimum Safety
Injection)
6.2.1-44 Reactor Coolant System Mass and Energy Balance
Double-Ended Pump Suction Guillotine (Maximum Safety
Injection)
6.2.1-45 Primary Coolant System Mass and Energy Balance 0.6
Double-Ended Pump Suction Guillotine
6.2.1-46 Reactor Coolant System Mass and Energy Balance
Three-Foot-Squared Pump Suction Split
6.2.1-47 Reactor Coolant System Mass and Energy Balance
Double-Ended Hot Leg Guillotine
6.2.1-48 Reactor Coolant System Mass and Energy Balance
Double-Ended Cold Leg Guillotine
6.2.1-49 Principal Reflood Parameters Transients Double-Ended
Pump Suction (Minimum Safety Injection)
6.2.1-50 Principal Reflood Parameters Transients Double-Ended
Pump Suction (Minimum Safety Injection)
6.2.1-51 Bases for Analysis
6.2.1-52 Safety Injection Flow Rate Versus Backpressure
6.0-viii Rev. 0 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.1-53 19-Element Reflood Model
6.2.1-54 Mass and Energy Release Hydraulic Characteristics
for Post-Reflood (One Intact Loop) at 130.7 Seconds
6.2.1-55 Mass and Energy Release Hydraulic Characteristics
for Post-Reflood (Broken Loop) at 130.7 Seconds
6.2.1-56 Spectrum of Secondary System Pipe Ruptures Analyzed
6.2.1-56A Time Sequence of Events for the Steamline Break Mass and Energy Releases to Containment
6.2.1-57 Specific Plant Design Input for MSLB Mass and
Energy Release Analysis
6.2.1-57A Mass and Energy Release Data for Case 10 - Peak Calculated Containment Pressure for MSLB
6.2.1-57B Mass and Energy Release Data for Class 1 - Peak Calculated Containment Temperature for MSLB
6.2.1-57C Containment Fan Cooler Performance Data
6.2.1-58 Summary of Results for MSLB Containment Pressure -
Temperature Analysis
6.2.1-59 Sequence of Events for Case 10 Peak Calculated Containment Pressure Case for MSLB
6.2.1-60 Sequence of Events for Case 1 Peak Calculated Containment Temperature for MSLB
6.2.1-61 This Table Deleted
6.2.1-62 This Table Deleted
6.2.1-63 Wolf Creek BELOCA Mass and Energy Release Data Used for COCO Calculations
6.2.1-64 Deleted 6.2.1-65 Large-Break LOCA Containment Data Used for Pressure Calculations 6.2.1-66 Large-Break LOCA Containment Heat Sink Data Used for Containment Pressure Calculations
6.0-ix Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.2-1 Comparison of the Recirculation Sump Design with
Each of the Positions of Regulatory Guide 1.82
6.2.2-2 Containment Heat Removal Systems Component Design
Parameters
6.2.2-3 Summary of Accident Chronology for Containment
Spray System for Loss-of-Coolant Accident
6.2.2-4 Spray Injection Phase Duration
6.2.2-5 Containment Spray System Single-Failure Analysis
6.2.2-6 Water Sources and Water Losses Which Contribute to
the Water Level Within the Reactor Building
Following a Large LOCA
6.2.2-6a Water Sources and Water Losses Which Contribute to
the Water Level Within the Reactor Building
Following a Main Steam Line Break
6.2.2.7 Input and Results of NPSH Analysis
6.2.2-8 Containment Air Cooling System
6.2.2-9 Sump Screen and Approach Velocities for LOCA and
MSLB Conditions
6.2.4-1 Listing of Containment Piping Penetrations
6.2.4-2 Design Comparison to Regulatory Guide 1.141 Revision
0, Dated May 1978, Titled Containment Isolation
Provisions for Fluid Systems
6.2.5-1 Design Data for Containment Hydrogen Control System
Components
6.2.5-2 Summary of Assumptions Used for Hydrogen Generation
from Radiolysis
6.2.5-3 Parameters Used to Determine Hydrogen Generation
6.2.5-4 Deleted
6.2.5-5 Single Failure Analysis Containment Hydrogen Control
System
6.0-x Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.2.5-6 Comparison of the Design to Regulatory Positions of
Regulatory Guide 1.7, Revision 2, Dated November, 1978, Titled "Control of Combustible Gas
Concentrations in Containment Following a Loss-of-
Coolant Accident"
6.3-1 Emergency Core Cooling System Component Parameters
6.3-2 Emergency Core Cooling System Relief Valve Data
6.3-3 Motor-Operated Isolation Valves in the Emergency
Core Cooling System
6.3-4 Materials Employed for Emergency Core Cooling System
Components
6.3-5 Failure Mode and Effects Analysis - Emergency Core
Cooling System - Active Components
6.3-6 Single Active Failure Analysis for Emergency Core
Cooling System Components
6.3-7 Emergency Core Cooling System Recirculation Piping
Passive Failure Analysis Long-Term Phase
6.3-8 Sequence of Changeover Operation from Injection to
Recirculation
6.3-9 Emergency Core Cooling System Shared Functions
Evaluation
6.3-10 Normal Operating Status of Emergency Core Cooling
System Components for Core Cooling
6.3-11 RWST Outflow (Large Break) - No Failures
6.3.12 RWST Outflow (Large Break) - Worst Single Failure
(9)
6.4-1 Comparison of the Design to Regulatory Positions of
Regulatory Guide 1.78, Dated June, 1974, Titled
"Assumptions for Evaluating the Habitability of a
Nuclear Power Plant Control Room During a Postulated
Hazardous Chemical Release"
6.4-2 Comparison of the Design to Regulatory Positions of
Regulatory Guide 1.95, Revision 1, Dated January, 1977, Titled "Protection of Nuclear Power Plant
Control Room Operators Against an Accidental
Chlorine Release"
6.0-xi Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Number Title
6.5-1 ESF Filtration Systems Input Parameters to Chapter
15.0 Accident Analysis
6.5-2 Input Parameters and Results of Spray Iodine Removal
Analysis
6.5-3 Spray Additive Subsystem Design Parameters
6.5-4 Spray Additive Subsystem Single Failure Analysis
6.5-5 Containment Spray System Fluid Chemistry
6.0-xii Rev. 29
WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) Title Drawing #*
6.2.1-1 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-2 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-3 0 0.6 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-4 0 3.0 Square Foot Pump Suction Split Break, Maximum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-5 0 Double-Ended Hot Leg Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-6 0 Double-Ended Cold Leg Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Pressure vs. Time 6.2.1-7 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-8 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-9 0 0.6 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-10 0 3.0 Square Foot Pump Suction Split Break, Maximum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-11 0 Double-Ended Hot Leg Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-12 0 Double-Ended Cold Leg Guillotine Break Maximum Safety Injection, 2 Air Coolers, Containment Vapor Temperature vs. Time 6.2.1-13 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Condensing Heat Transfer Coefficient vs. Time 6.2.1-14 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Condensing Heat Transfer Coefficient vs. Time 6.2.1-15 0 Containment Air Cooler Duty Curve, Heat Removal Rate vs. Temperature 6.0-xiii Rev. 17 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) TitleDrawing #*
6.2.1-16 1 Reactor Decay Power 6.2.1-16 2 Reactor Decay Power 6.2.1-17 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Sump Temperature vs. Time 6.2.1-18 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Sump Temperature vs. Time 6.2.1-19 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Total Air Cooler Heat Removal Rate vs. Time 6.2.1-20 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Total Air Cooler Heat Removal Rate vs. Time 6.2.1-21 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, RHR Heat Exchanger Heat Removal vs. Time 6.2.1-22 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, RHR Heat Exchanger Heat Removal Rate vs. Time 6.2.1-23 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Heat Sink Total Heat Transfer Rate vs. Time 6.2.1-24 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers, Heat Sink Total Heat Transfer Rate vs. Time 6.2.1-25 0 Double-Ended Pump Suction Guillotine Break Minimum Safety Injection, 2 Air Coolers, Energy Inventory vs. Time 6.2.1-26 0 Double-Ended Pump Suction Guillotine Break Maximum Safety Injection, 2 Air Coolers Energy Inventory vs. Time 6.2.1-27 0 Reactor Cavity Analysis Nodalization Scheme -
Elevation View 6.2.1-28 0 Reactor Cavity Analysis Nodalization Scheme -
Level 1 6.2.1-29 0 Reactor Cavity Analysis Nodalization Scheme -
Level 2 6.2.1-30 0 Reactor Cavity Analysis Nodalization Scheme -
Level 3 6.2.1-31 0 Reactor Cavity Analysis Nodalization Scheme -
Level 4 6.0-xiv Rev. 17 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) TitleDrawing #*
6.2.1-32 0 Reactor Cavity Analysis Nodalization Scheme -
Level 5 6.2.1-33 0 Elevation on CL Reactor Bldg., Nodalization Scheme for Compartment Level 6 6.2.1-34 0 Reactor Cavity Pressure - Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 3 6.2.1-35 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 2 6.2.1-36 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 4 6.2.1-37 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 5 6.2.1-38 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 6 6.2.1-39 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Compartment Pressures on Level 1 6.2.1-40 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. Cold-Leg Break, Horizontal Force Components on RPV 6.2.1-41 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. in Cold-Leg Break - Total Moments 6.2.1-42 0 Reactor Cavity Pressure-Temperature Analysis, 150 Sq. In. in Cold-Leg Break 6.2.1-43 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 1 6.2.1-44 0 Steam Generator Loop Compartment Analysis Nodalization Scheme - Level 1 6.2.1-45 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 2 6.2.1-46 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 2 6.2.1-47 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 2A 6.2.1-48 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 3 6.2.1-49 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 3 6.0-xv Rev. 17 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) TitleDrawing #*
6.2.1-50 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 4 6.2.1-51 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 4 6.2.1-52 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 5 6.2.1-53 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 5 6.2.1-54 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 6 6.2.1-55 0 Steam Generator Loop Compartment Analysis, Nodalization Scheme - Level 6 6.2.1-56 0 Steam Generator Loop Compartment Analysis, Cold-Leg Break, Abs. Pressures Near the Break Compartment 6.2.1-57 0 Steam Generator Loop Compartment Analysis, Abs.
Pressures Near the Break, Compartments Cold-Leg Break 6.2.1-58 0 Steam Generator Loop Compartment Analysis, Cold-Leg Break, Vertical and E-W and N-S Forces on Sg 6.2.1-59 0 Steam Generator Loop Compartment Analysis, Loads on the RCP, 236 Sq. In. Cold-Leg Break 6.2.1-60 0 Steam Generator Loop Compartment Analysis, 236 In.2 Cold-Leg Break, Direction of Peak Horizontal Forces on Reactor Coolant, Pump and Steam Generator 6.2.1-61 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, Absolute Pressures Near the Break 6.2.1-62 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, N-S Component of Horizontal Force on SG 6.2.1-63 0 Steam Generator Loop, Compartment Analysis, 436 In.2 Pump Suction Line Break, E-W Component of Horizontal Force SG 6.2.1-64 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, Vertical Force on SG 6.2.1-65 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, N-S Component of Horizontal Force on RCP 6.0-xvi Rev. 17 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) TitleDrawing #*6.2.1-66 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break E-W Component of Horizontal Force on RCP 6.2.1-67 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, Vertical Force on RCP 6.2.1-68 0 Steam Generator Loop Compartment Analysis, 436 In.2 Pump Suction Line Break, Direction of Peak Horizontal Forces on Reactor Coolant Pump and Steam Generator 6.2.1-69 0 Steam Generator Loop Compartment Analysis, 736 In.2 Hot-Leg Break, Absolute Pressure Near the Break 6.2.1-70 0 Steam Generator Loop Compartment Analysis, 763 In.2 Hot-Leg Break, Horizontal Forces on SG 6.2.1-71 0 Steam Generator Loop Compartment Analysis, 763 In.2 Hot-Leg Break, Vertical Forces on SG 6.2.1-72 0 Steam Generator Loop Compartment Analysis, 763 in.2 Hot-Leg Break, N-S Component of Horizontal Force on RCP 6.2.1-73 0 Steam Generator Loop Compartment Analysis, 763 in.2 Hot-Leg Break, E-W Component of Horizontal Force on RCP 6.2.1-74 0 Steam Generator Loop Compartment Analysis, 763 in.2 Hot-Leg Break, Vertical Force on RCP 6.2.1-75 0 Steam Generator Loop Compartment Analysis, 763 in.2 Hot-Leg Break, Direction of Peak Horizontal Forces on Reactor Coolant Pump and Steam Generator 6.2.1-76 0 Pressurizer Compartment Analysis Nodalization Scheme -Elevation View 6.2.1-77 0 Flow Diagram Pressurizer Compartment Analysis 6.2.1-78 0 Pressurizer Compartment Analysis, Pressurizer Surge Line Break, Absolute Pressures Below the Pressurizer 6.2.1-79 0 A Simplified Schematic of the Wolf Creek Containment 6.2.1-80 0 Wolf Creek GOTHIC Containment Model for MSLB Events 6.2.1-81 0 Containment Pressure, Vapor Temperature and Sump Water Temperature Response to a Postulated MSLZB - Case 10 Scenario 6.2.1-82 0 Containment Pressure, Vapor Temperature and Sump Water Temperature Response to a Postulated MSLB - Case 1 Scenario 6.0-xvii Rev. 29
WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) TitleDrawing #*6.2.1-83 0 Heat Transfer Coefficient vs Time, Limiting Containment Pressure Scenario - MSLB Case 10 6.2.1-84 0 Heat Transfer Coefficient vs Time, Limiting Containment Temperature Scenario - MSLB Case 1 6.2.1-85 0 Surface Temperature vs. Time for Representative Materials Inside Containment Following MSLB, Case 7 6.2.1-86 0 Analysis vs. Calculated containment Back Pressure 6.2.1-87 0 DELETED 6.2.2-1 0 Containment Spray System M-12EN01 6.2.2-2 1 Containment Spray System Header Arrangement M-13EN03 6.2.2-2 2 Containment Spray System Header Arrangement M-13EN04 6.2.2-2 3 Containment Spray System Header Arrangement M-13EN05 6.2.2-3 1 Recirculation Sump Strainer 6.2.2-4 0 CSS Area Coverage at Operating Deck of Containment 6.2.2-5 0 CSS Pump Performance Curve 6.2.2-6 0 Typical Detail of Fusible Link Plates on Containment Air Cooler 6.2.2-7 0 Expected Internal Air Flow Patterns in Containment Post LOCA 6.2.4-1 1 Containment Penetrations (P-1) 6.2.4-1 2 Containment Penetrations (P-2) 6.2.4-1 3 Containment Penetrations (P-3) 6.2.4-1 4 Containment Penetrations (P-4) 6.2.4-1 5 Containment Penetrations (P-5) 6.2.4-1 6 Containment Penetrations (P-6) 6.2.4-1 7 Containment Penetrations (P-7) 6.2.4-1 8 Containment Penetrations (P-8) 6.2.4-1 9 Containment Penetrations (P-9) 6.2.4-1 10 Containment Penetrations (P-10) 6.2.4-1 11 Containment Penetrations (P-11) 6.2.4-1 12 Containment Penetrations (P-12) 6.2.4-1 13 Containment Penetrations (P-13) 6.2.4-1 14 Containment Penetrations (P-14) 6.2.4-1 15 Containment Penetrations (P-15) 6.2.4-1 16 Containment Penetrations (P-16) 6.2.4-1 17 Containment Penetrations (P-21) 6.2.4-1 18 Containment Penetrations (P-22) 6.2.4-1 19 Containment Penetrations (P-23) 6.2.4-1 20 Containment Penetrations (P-24) 6.2.4-1 21 Containment Penetrations (P-25)
6.0-xviii Rev. 29 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) Title Drawing #*
6.2.4-1 22 Containment Penetrations (P-26) 6.2.4-1 23 Containment Penetrations (P-27) 6.2.4-1 24 Containment Penetrations (P-28) 6.2.4-1 25 Containment Penetrations (P-29) 6.2.4-1 26 Containment Penetrations (P-30) 6.2.4-1 27 Containment Penetrations (P-32) 6.2.4-1 28 Containment Penetrations (P-34) 6.2.4-1 29 Containment Penetrations (P-39) 6.2.4-1 30 Containment Penetrations (P-40) 6.2.4-1 31 Containment Penetrations (P-41) 6.2.4-1 32 Containment Penetrations (P-43) 6.2.4-1 33 Containment Penetrations (P-44) 6.2.4-1 34 Containment Penetrations (P-45) 6.2.4-1 35 Containment Penetrations (P-48) 6.2.4-1 36 Containment Penetrations (P-49) 6.2.4-1 37 Containment Penetrations (P-51) 6.2.4-1 38 Containment Penetrations (P-52) 6.2.4-1 39 Containment Penetrations (P-53) 6.2.4-1 40 Containment Penetrations (P-54) 6.2.4-1 41 Containment Penetrations (P-55) 6.2.4-1 42 Containment Penetrations (P-56) 6.2.4-1 42a Containment Penetrations (P-56) 6.2.4-1 42b Containment Penetrations (P-57) 6.2.4-1 43 Containment Penetrations (P-58) 6.2.4-1 43a Containment Penetrations (P-59,91) 6.2.4-1 44 Containment Penetrations (P-62) 6.2.4-1 45 Containment Penetrations (P-63) 6.2.4-1 45a Containment Penetrations (P-64) 6.2.4-1 46 Containment Penetrations (P-65) 6.2.4-1 47 Containment Penetrations (P-66) 6.2.4-1 48 Containment Penetrations (P-67) 6.2.4-1 49 Containment Penetrations (P-69) 6.2.4-1 50 Containment Penetrations (P-71) 6.2.4-1 51 Containment Penetrations (P-73) 6.2.4-1 52 Containment Penetrations (P-74) 6.2.4-1 53 Containment Penetrations (P-75) 6.2.4-1 54 Containment Penetrations (P-76) 6.2.4-1 55 Containment Penetrations (P-78) 6.2.4-1 56 Containment Penetrations (P-79) 6.2.4-1 57 Containment Penetrations (P-80) 6.2.4-1 58 Containment Penetrations (P-82) 6.2.4-1 59 Containment Penetrations (P-83) 6.2.4-1 60 Containment Penetrations (P-84)
6.0-xix Rev. 29 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) Title Drawing #*
6.2.4-1 61 Containment Penetrations (P-85) 6.2.4-1 62 Containment Penetrations (P-86) 6.2.4-1 63 Containment Penetrations (P-87) 6.2.4-1 64 Containment Penetrations (P-88) 6.2.4-1 65 Containment Penetrations (P-89) 6.2.4-1 66 Containment Penetrations (P-92) 6.2.4-1 67 Containment Penetrations (P-93) 6.2.4-1 68 Containment Penetrations (P-95) 6.2.4-1 69 Containment Penetrations (P-97) 6.2.4-1 69a Containment Penetrations (P-97) 6.2.4-1 69b Containment Penetrations (P-98) 6.2.4-1 70 Containment Penetrations (P-99) 6.2.4-1 70a Containment Penetrations (P-99) 6.2.4-1 71 Containment Penetrations (P-101) 6.2.4-1 71a Containment Penetrations (P-101) 6.2.4-1 72 Containment Penetrations (P-103 & 104) 6.2.4-1 73 Containment Penetrations (V-160) 6.2.4-1 74 Containment Penetrations (V-161) 6.2.4-2 0 Steam Generator and Associated Systems as a Barrier to the Release of Radioactivity Post LOCA 6.2.5-1 0 Containment Hydrogen Control System M-12GS01 6.2.5-2 0 Hydrogen Volume Concentration in Containment With One Recombiner Operating at One Day 6.2.5-3 0 Hydrogen Generation in Containment 6.2.5-4 0 Hydrogen Accumulation in Containment 6.2.5-5 0 Hydrogen Volume Concentration in Containment Assuming No Preventive Action Taken 6.2.5-6 0 Hydrogen Volume Concentration in Containment with Purging After 4 Days 6.2.5-7 0 Aluminum Corrosion Rates 6.2.5-8 0 Zinc Corrosion Rates 6.2.5-9 0 Temperature Profile Used for Adjusting Corrosion Rates 6.2.6-1 0 Containment Integrated Leak Rate Test M-12GP01 6.3-1 1 Borated Refueling Water Storage System M-12BN01 6.3-1 2 High Pressure Coolant Injection System M-12EM01 6.3-1 3 High Pressure Coolant Injection System M-12EM02 6.3-1 4 Accumulator Safety Injection M-12EP01 6.3-2 1 Emergency Core Cooling System Process Flow Diagram 6.3-2 2 Emergency Core Cooling System Process Flow Diagram 6.0-xx Rev. 17 WOLF CREEK CHAPTER 6 - LIST OF FIGURES
- Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure # Sheet(s) Title Drawing #*
6.3-3 0 Typical Residual Heat Removal Pump Performance Curve 6.3-4 0 Typical Centrifugal Charging Pump Performance Curve 6.3-5 0 Typical Safety Injection Pump Performance Curve 6.3-6 0 Gate Valve Assembly 6.3-7 0 RWST Levels and Volumes 6.4-1 1 Typical Detail Sealing of Piping Penetration Through Cont. Rm. Fl. or Wall 6.4-1 2 Typical Detail Sealing of Ductwork Penet. Through Cont. Rm. Fl. or Wall 6.4-1 3 Typical Detail Sealing of Cable Tray Penet. Through Cont. Rm. Fl. or Wall 6.5-1 0 Capacity Curve 15215-1C-304SS-6.3 Whirljet Nozzle 6.5-2 0 Spatial Droplet Size Distribution of 15215-1C-304SS-6.3 Whirljet Spray Nozzle 6.5-3 0 Particle Size vs. Pressure 15215-1C-304-S-6.3 Whirljet Spray Nozzle 6.5-4 0 Spray Envelope Reduction Factor 6.5-5 0 Containment Sump PH with Nominal Eductor Flow for One Eductor and Two Eductor Operation
6.0-xxi Rev. 28 WOLF CREEK NRC QUESTIONS PERTAINING TO CHAPTER 6.0 Section Question Number Section Question Number6.4 450.00 6.4 450.01 6.4 450.02 6.4 450.03 6.4 450.04 6.4 450.05 6.4 450.06 6.5.2 450.07 6.0-xxii Rev. 17 WOLF CREEK CHAPTER 6.0 ENGINEERED SAFETY FEATURES Engineered safety features (ESF) are those safety-related systems and components designed to directly mitigate the consequences of a design basis accident by:
- a. Protecting the fuel cladding
- b. Ensuring the containment integrity
- c. Limiting fission product releases to the environment within the guideline values of 10 CFR, Part 100 The limiting design basis accidents which are discussed and analyzed in Chapter 15.0 and Section 6.3 are:
- a. Loss-of-coolant accident (LOCA)
- b. Main steam line break (MSLB)
- c. Steam generator tube rupture
- d. Fuel handling accident
(Items a and b are also discussed in Section 6.2)
The engineered safety features consist of the following systems:
- a. Containment (Section 6.2.1)
- b. Containment heat removal (Section 6.2.2)
- c. Containment isolation (Sections 6.2.4 and 6.2.6)
- d. Containment combustible gas control (Section 6.2.5)
- e. Emergency core cooling (Section 6.3)
- f. Fission product removal and control systems (Section 6.5)
- g. Emergency HVAC and filtration (Section 9.4)
- h. Control room habitability (Section 6.4)
- i. Auxiliary feedwater (Section 10.4.9)
The containment is provided to contain radioactivity following a LOCA. 6.1-1 Rev. 13 WOLF CREEK The containment spray system, in conjunction with the containment fan coolers
and the emergency core cooling system, was designed to remove sufficient heat
from the containment atmosphere following a LOCA or main steam line break
inside the containment to rapidly reduce the containment pressure and
temperature and maintain them at acceptably low levels.
The containment spray system was also designed to minimize the iodine and
particulate fission product inventories in the containment atmosphere resulting
from a postulated LOCA.
Containment isolation is provided to minimize leakage from the containment.
Steam line and feedwater line isolation is provided to minimize the heat
removal from the reactor coolant system and prevents excessive blowdown of a
steam generator following a postulated main steam line rupture. Steam line
isolation also prevent excessive radioactivity release following a steam
generator tube rupture. The containment purge isolation capability is provided
to reduce the radioiodine released following a fuel handling accident inside
the containment.
The emergency core cooling system (ECCS), consisting of accumulator tanks, safety injection pumps, RHR pumps, and centrifugal charging pumps, is provided
for emergency core cooling to limit fuel damage following a LOCA or main steam
line break.
An emergency exhaust system is provided to reduce the radioiodine released
following a fuel handling accident outside the containment and to filter ECCS
leakage outside the containment following a LOCA.
The auxiliary feedwater system provides an adequate amount of feedwater into
the steam generators to prevent a pressure transient which could cause a loss
of reactor coolant through the pressurizer relief valves and a possible
uncovering of the reactor core following a main steam line break or loss of the
main feedwater system.
Other safety-related systems are identified in Section 3.2. Because of the
importance of safety-related systems to the health and safety of the general
public, special precautions are taken to ensure high quality in the components
and in the system design and to ensure reliable and dependable operation.
6.1-2 Rev. 29 WOLF CREEK 6.1 ENGINEERED SAFETY FEATURE MATERIALS This section provides a discussion of the materials used in the fabrication of engineered safety feature components and of the material interactions that could potentially impair the operation of the ESF.
6.1.1 METALLIC MATERIALS
6.1.1.1 Materials Selection and Fabrication Information on the selection and fabrication of the materials in the engineered safety features of the plant, such as the emergency core cooling systems, the
containment heat removal systems, the containment combustible gas control
system, and the containment spray system, is provided below. Materials for use
in the ESF are selected for their compatibility with the reactor coolant system
and containment spray solutions, as required by Section III of the ASME Boiler
and Pressure Vessel Code, Articles NC-2160 and NC-3120.
6.1.1.1.1 Specifications for Principal Pressure-Retaining Materials All pressure-retaining material in the engineered safety feature systems' components complies with the corresponding material specification permitted by ASME Section III, Division 1.
The material specifications for pressure-retaining material in each component of the engineered safety feature systems will meet the requirements of Article NC-2000 of ASME Section III, Class 2, for quality group B and Article ND-2000
of ASME Section III, Class 3, for quality group C components. Containment
penetration materials will meet the requirements of Article NE-2300 of ASME Section III, Division I. Table 6.1-1 includes the specifications for the
principal pressure-retaining components.
6.1.1.1.2 Engineered Safety Feature Materials of Construction The engineered safety feature materials that would be exposed to the emergency core cooling water and containment sprays following a LOCA are indicated in
Table 6.1-1. These materials are chosen to be compatible with the core cooling
and spray solutions. Additional information concerning metallic materials'
compatibility with post-LOCA conditions is provided in Reference 1. 6.1-3 Rev. 0 WOLF CREEK In order to keep materials within the containment that are subject to corrosion to a minimum, the following restrictions are placed on the use of zinc, aluminum, and mercury in the containment:
- a. Aluminum is severely attacked by the alkaline containment spray solution. This reaction may result in the loss of structural integrity and the generation of gaseous hydrogen. The use of aluminum
in the containment is minimized.
- b. Boric acid reacts with zinc, oxidizing it and liberating hydrogen gas. The use of zinc (galvanized materials and paint) in the containment is minimized to reduce the generation of hydrogen.
- c. The use of mercury and mercuric compounds is minimized inside the containment because of its corrosive effects
on stainless steel, NiCrFe alloy 600, and alloys
containing copper. The amount of mercury associated with
plant lighting and control switches, etc., is negligible.
Table 6.2.5-3 is a list of the amounts of aluminum and zinc which are in the containment and which could potentially be exposed to a corrosive environment.
These materials are listed by the system or component in which they are used, and an estimate of their expected corrosion rate is given. Aluminum or zinc is not used in any safety-related item where exposure to the spray solution is possible.For other materials which could come in contact with containment sprays, tests have been performed and are detailed in Reference 2. These tests have shown that no significant amount of corrosion products is produced from these materials.
Many coatings which are in common industrial use may deteriorate in the post-accident environment and contribute substantial quantities of foreign solids
and residue to the containment sump. Consequently, protective coatings used
inside the containment in significant quantities are demonstrated to withstand
the design basis accident conditions and are designed to meet the criteria
given in ANSI N101.2 (1972), "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities," and are in compliance with Regulatory Guide 1.54, "Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," as indicated in Table 6.1-2. Some small
items may be painted or coated using common industrial practice but the
paint/coating is not in sufficient quantity to cause any 6.1-4 Rev. 0 WOLF CREEK clogging problems for the sump strainer. Any precipitation of appreciable size that occurs either settles out prior to reaching the sump strainer or is trapped by the sump filter strainer. The strainer opening size 0.045 inch is smaller than the line piping, the RHR heat exchanger tubes, the spray nozzles, and clearances in the reactor core. Therefore, particles which could
potentially cause blockage are filtered out. Refer to Section 6.2.2.1 for a
discussion of the sump design and consideration given to strainer clogging.
For each containment component, a complete list of the surface coatings, the
dry film thickness, and the surface area covered is presented in Table 6.1-3.
6.1.1.1.3 Integrity of Safety-Related Components
The following information is provided to demonstrate that the integrity of the safety-related components is maintained during all stages of component
manufacturing:
- a. Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel, is complied with to the extent specified in Table 6.1-4 for the purpose of avoiding significant sensitization and stress corrosion cracking in austenitic stainless steel components of the engineered safety features.
- b. Cleaning and contamination protection of austenitic stainless steel components of the engineered safety features complies with Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel, as described in
Table 6.1-4. Regulatory Guide 1.37, Quality Assurance
Requirements for Cleaning of Fluid Systems and Associated
Components of Water-Cooled Nuclear Power Plants, is complied with to the extent specified in Table 6.1-5.
- c. Cold worked austenitic stainless steel material with 0.2-percent offset yield strengths greater than 90,000 psi
are not used in components that are part of the
engineered safety features.
- d. The selection, procurement, testing, storage, and installation of all nonmetallic thermal insulation assure
that the leachable concentrations of chloride, fluoride, sodium, and silicate are in accordance with Regulatory
Guide 1.36, Nonmetallic Thermal Insulation for Austenitic
Stainless Steel. Compliance with Regulatory Guide 1.36
is discussed in Table 6.1-6. 6.1-5 Rev. 20 WOLF CREEK
- e. With regard to the preheat temperature used for welding low alloy steels, the recommendations of Regulatory Guide
1.50, Control of Preheat Temperatures for Welding of Low Alloy Steel, were followed, as discussed in Table 6.1-7.
- f. The recommendations of Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility, are followed as discussed in Table 6.1-8.
- g. In order to determine the RT NDT for the steam and feedwater system materials, the guidelines in NRC Branch Technical Position MTEB 5-2 Section 1.1, Article 4 were followed.
The applied test methods and acceptance criteria for all materials used in the steam and feedwater systems, with the exception of the steam generators, comply completely with ASME Code Section III, Article NC-2310 of the Winter 1974 Addenda for fracture toughness of ferritic materials
used in Class 2 components. The applied test methods and
acceptance criteria for all Class 2 steam generator
materials comply with the requirements of ASME Code
Section III 1971 Edition through Summer 1973 Addenda.
6.1.1.1.4 Control of Stainless Steel Welding Regulatory Guide 1.31, Control of Stainless Steel Welding, as supplemented by Branch Technical Position MTEB 5-1, is complied with to the extent specified in
Table 6.1-9 for the purpose of avoiding fissuring in austenitic stainless steel
welds that are part of the engineered safety features.
6.1.1.2 Composition, Compatibility, and Stability of Containment and Core Spray Coolants The information given below is provided on the composition, compatibility, and stability of the core cooling water and the containment sprays on the
engineered safety features.
6.1.1.2.1 Control of pH During a Loss-of-Coolant Accident
A description of the method of establishing containment spray and recirculation sump pH following a LOCA is included in Sections 6.2.2 and 6.5. The resultant
basic pH range of 8.5-9.0 is not conducive to stress-corrosion cracking in austenitic stainless steels. Hydrogen evolution is discussed in Section 6.2.5, Combustible Gas Control in Containment. 6.1-6 Rev. 0 WOLF CREEK 6.1.1.2.2 Engineered Safety Feature Coolant Storage The borated water supply for the containment sprays and emergency core cooling system is drawn from the refueling water storage tank. As described in Section 6.3, the refueling water storage tank is fabricated of stainless steel and is not subject to significant corrosive attack by the tank's contents. Spray additive (NaOH) for the containment spray system is stored in a stainless steel tank to prevent corrosive attack by the 30 weight percent (nominal) sodium
hydroxide solution.
The accumulator tanks which store borated water for the accumulator safety injection system are made of carbon steel and are clad with stainless steel to ensure that they are resistant to corrosion.
6.1.2 ORGANIC MATERIALS
Use of organic material inside the containment is kept to a minimum.
The amount of lubricants inside the containment which is subject to being released to the containment is listed in Table 6.1-10. The lubricants, such as
those needed for the reactor coolant pumps and hydraulic snubbers, are, however, totally enclosed and not open to the containment atmosphere.
Table 6.1-3 is a coating schedule for the containment which indicates the type of paint and compliance with Regulatory Guide 1.54.
Protective coatings covered by Regulatory Guide 1.54 which are applied to surfaces within the containment have been tested to demonstrate that they will
remain intact during postulated LOCA conditions. The tests are performed by an
independent laboratory and show that no significant decomposition, radiolytic or pyrolytic failures will occur during a DBA.
Where the surface area and application type do not dictate special coatings, the coatings are evaluated by generic-type and formulation information. Paint chip formation is controlled by limiting the thickness of nonqualified coatings
to a point where there is insufficient tensile strength in a removed film to
form a chip. 6.1-7 Rev. 0 WOLF CREEK 6.1.3 POST-ACCIDENT CHEMISTRY Following a main steam line break or design basis LOCA, sodium hydroxide and boric acid solutions will be present in the containment sumps. Figure 6.5-5 represents the time-history of the pH of the aqueous phase in the containment sump. Table 6.5-5 indicates the quantities of sodium hydroxide and boric acid that will be present in the containment after an accident. The pH control reduces the probability of chloride stress corrosion cracking on stainless
steel and attack on aluminum fittings.
6.
1.4 REFERENCES
- 1. Whyte, D. D. and Picone, L. F., "Behavior of Austenitic Stainless Steel in Post Hypothetical Loss-of-Coolant
Environment," WCAP-7798-L (Proprietary), November 1971
and WCAP-7803 (Non-Proprietary), December 1971.
- 2. Picone, L. F., "Evaluation of Protective Coatings for use in Reactor Containment," WCAP-7198-L (Proprietary), April
1968 and WCAP-7825 (Non-Proprietary), December 1971.
3 Caplan, J. S., "The Application of Preheat Temperatures after Welding Pressure Vessel Steels," WCAP-8577 (Non-
Proprietary), September 1975. 6.1-8 Rev. 0 WOLFCREEKTABLE6.1-1ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingSafetyInjectionSystems-IncludesResidualHeatRemovalandCVCSSystemsRefuelingwaterstorage6.3NoNoIII-2SA240,Type304;N/A tankSA312,Type304;SA1182,F304;SA479,Type304Accumulator6.3YesNoIII-2SA533withSSCladChemicallycuredepoxyor modified phenolicepoxyHighheadsafetyin-6.3NoYesIII-2 jectionpump CasingSA351,GradeCF8N/AorCF8M,SA182,F304orF316 ImpellerA296CA40N/A ShaftA276410N/AResidualheatremoval5.4.7/6.3NoYesIII-2
pump CasingSA182,F304N/A ImpellerA296CA40N/A ShaftA276410N/ARev.0 WOLFCREEKTABLE6.1-1(Sheet2)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingResidualheatremovalheatexchanger5.4.7/6.3NoYesIII-2 ShellSA240andSA312,N/AType304 TubesSA213,Type304;N/ASA249,Type304TubeSheetsSA182,F304;N/ASA240,Type304;SA516,Grade70withSSCladdingRecirculationvalveencapsulation6.3NoNoIII-2SA240,Type304;Carbozinc11SA312,Type304;forcarbonSA182,F304;SA285steelskirtGradeCBoroninjectiontank6.3NoYesIII-2SA351,GradeCF8A;N/ASA240,Type304Centrifugalcharging9.3.4NoYesIII-1SA182,F304N/A pumpContainmentSpraySystem Containmentspraypump6.2.2NoYesIII-2 CasingSA182,F304N/A ImpellerA487,CB6MMN/ARev.0 WOLF CREEK TABLE 6.1-1 (Sheet 3)
ESF MATERIALS OF CONSTRUCTION Internally Externally Exposed to
Exposed to Containment
Applicable Containment DBA Design Protective
Item Section Environment Environment Code Specification Coating Shaft A 276, Type 410, N/A
Condition T Containment spray 6.2.2 No No III-2 SA 240, Type 304 N/A additive tank Containment spray 6.2.2 No Yes III-2 additive eductor Body SA 182, Type 304 N/A (Body)
Insert SA 564, Type 630 N/A (Insert)
Containment spray 6.2.2 Yes Yes III-2 header and nozzles Header SA 312, Type 304 N/A or SA 376, Type 304 Nozzles SA 351, Type 304 N/A Containment 6.2.2 Yes Yes III-2 Type 304 SS N/A recirculation sump strainer Recirculation valve 6.2.2 No No III-2 SA 240, Type 304; Carbozinc 11
encapsulation SA 312, Type 304; for carbon
SA 182, F 304; steel skirt
SA 285, Grade C Rev. 20 WOLFCREEKTABLE6.1-1(Sheet4)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingAuxiliaryFeedwaterSystemMotor-drivenauxiliaryfeedwaterpump10.4.9NoNoIII-3 CasingSA217,WC9Mfrs.Std.
ImpellerA296,CA6NMN/A ShaftA276,Type410,ConditionTN/ATurbine-drivenauxiliary10.4.9NoNoIII-3 feedwaterpump CasingSA217,WC9Mfrs.Std.
ImpellerA297,CA6NMN/A ShaftA276,Type410,ConditionTN/AAuxiliaryfeedwaterpump10.4.9NoNoMS turbine CasingA216,WCBMfrs.Std.
RotorAISI4140N/ARev.0 WOLFCREEKTABLE6.1-1(Sheet5)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingMainFeedwaterSystemPortionofsystempipingandinstrumentation10.4.7YesNoIII-2SA333,Grade6Carbozinc11Isolationvalve10.4.7NoNoIII-2SA216,WCBN/AMainSteamSystemPortionofsystempipingandinstrumentation10.3YesNoIII-2SA155,KCF-70Carbozinc11Isolationvalve10.3NoNoIII-2SA216,WCBN/AContainmentandPipingPenetrationsContainmentpiping6.2.4YesYes/NoIII-2SA155,KCF-70CLCarbozinc11 penetrationSA333,Grade6ContainmentpenetrationSeeASMEIIIN/Aisolationvalves6.2.4YesYes/NoIII-2Class2ValvesContainmentpenetrationpipingbetweenisolation6.2.4YesYes/NoIII-2SeeASMEIIICarbozinc11 valvesClass2PipingorN/AContainmentliner6.2.4YesN/AIII,SA285,GradeACarbozinc11Div2 (Prop)
Sec.
3,000Rev.0 WOLFCREEKTABLE6.1-1(Sheet6)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingContainmentCoolingSystemContainmentcoolerfan6.2.2/9.4YesYesN/AA283ModifiedHousingconeandbellphenolicepoxyContainmentcoolercoils6.2.2/9.4YesNoIII-3SB111,Alloy706;N/AB152,Alloy110;SB 466,Alloy706;A526Containmentcooler6.2.2/9.4YesYesN/AA500B,A570,Modified housingGradeDphenolicepoxyContainmentcooler6.2.2/9.4YesNoNEMACarbonsteel,Modified fanmotorcopperphenolicepoxyHydrogenmixingfan6.2.2/9.4YesYesN/ACarbonSteelModifiedphenolicepoxyHydrogenmixingfan6.2.2/9.4YesNoNEMACarbonsteel,Mfrs.Std.
motor copperContainmentHydrogenControlSystemElectricrecombiner6.2.5YesYesNEMAA240,Type304N/ARev.0 WOLFCREEKTABLE6.1-1(Sheet7)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingHydrogenanalyzer6.2.5NoYesN/ATubing(includingSA213,Type304orN/A coolers)316 FittingsSA479,Type316,N/ASA182,Type316PipingandvalvesASMEIIIClass13.9.3III-1PipingYesYesSA312,Type304,N/A seamlessValvesYesYesSA182,F316N/ASA351,GradeCP8or
CF8MASMEIIIClass23.9.3III-2PipingYesYesSA312,Type304,N/AseamlessorweldedSA155,KC-70,C1.1,weldedCarbozinc11 SA155,KCP-70SA106,GradesBandC,seamlessCarbozinc11 SA333,Grade6,seamlessorweldedCarbozinc11Rev.0 WOLFCREEKTABLE6.1-1(Sheet8)ESFMATERIALSOFCONSTRUCTION InternallyExternallyExposedtoExposedtoContainmentApplicableContainmentDBADesignProtectiveItemSection Environment Environment Code Specification CoatingValvesYesYesSA182,F316N/ASA351,GradeCP8orN/A CF8MSA216,WCBASMEIIIClass33.9.3III-3PipingYesNoSA312,Type304,N/Aseamlessorwelded SA155,KC-70,C1.1,Carbozinc11
welded SA106,GradeB,Carbozinc11
seamlessValvesYesNoSA182,F316N/ASA351,GradeCF8or CF8MSA216,WCBCarbozinc11Rev.0 WOLF CREEK TABLE 6.1-2 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containment1. ANSI N101.4-1972 should be 1. Complies. 1, 2, 3 and 4. NSSS equipment located used in conjunction with ANSI in the containment building is separated N45.2-1971, "Quality Assurance into four categories to identify the Program Requirements for Nuclear applicability of this regulatory guide
Power Plants." to various types of equipment. These
categories of equipment are as follows: 2. Subdivision 2.7 of ANSI N101- 2. Complies.
4-1972 states that when references
are made to other standards, these references shall imply the most a. Category 1 - Large equipment recent or current editions of the b. Category 2 - Intermediate equipment referenced standards. The specific c. Category 3 - Small equipment applicability or acceptability of d. Category 4 - Insulated/stainless referenced standards will be steel equipment
covered separately in other regula-
tory guides, where appropriate.
A discussion of each equipment category
follows:3. Subdivision 1.1.2 of ANSI 3. Complies, except N101.4-1972 states that quality that for certain a. Category 1 - Large Equipment assurance, as covered by this applications within standard, comprises all those the containment, The Category 1 equipment consists planned and systematic actions where the coating of the following:
necessary to provide specified is not necessary documentation and adequate con- for the protection (1) Reactor coolant system supports fidence that shop or field of the component, (2) Reactor coolant pumps (motor and coating work for nuclear a quality assurance motor stand)
Rev. 17 WOLF CREEK TABLE 6.1-2 (Sheet 2)
DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containmentfacilities will perform satis- program was not (3) Accumulator tanks factorily in service. This applied. In those (4) Refueling machine statement should not be inter- applications, the preted as implying that the end coating was reviewed Since this equipment has a large product of quality assurance actions to assure that surface area and was procured from is the production of specified docu- there were no long- only a few vendors, it was possible mentation. The term "quality assurance," term detrimental to implement tight controls over as used in ANSI N101.4-1972, should effects. these items. Stringent requirements be considered to comprise all those were specified for protective coatings planned and systematic actions on this equipment through the use of necessary to provide adequate a painting specification in the confidence that shop or field coating procurement documents. This specifi-work for nuclear facilities will cation defined requirements for:
perform satisfactorily in service.
In this connection, it is (1) Preparation of vendor procedures
emphasized that records and documents listed in Subdivisions (2) Use of specific coatings systems
7.4 through 7.8 and included which are qualified to ANSI N101.2 in the standard, are suggested forms only. Alternate documen- (3) Surface preparation
tation consistent with the requirements of Appendix B to (4) Application of the coating systems
10 CFR Part 50 is also con-in accordance with the paint sidered acceptable.
manufacturer's instructions Rev. 17 WOLF CREEK TABLE 6.1-2 (Sheet 3)
DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containment4. Sections 3 and 4 of ANSI 4. Complies (5) Inspections and nondestructive
N101.4-1972 delineate quality examinations assurance requirements for coating materials and surface (6) Exclusive of certain materials
preparation of substrates.
Cleaning materials used with (7) Identification of all noncon-stainless steel would not be formances
compounded from or treated with chemical compounds con- (8) Certifications of compliance
taining elements that could contribute to corrosion, The vendor's procedures were subject intergranular cracking, or to review by engineering personnel, stress corrosion cracking. and the vendor's implementation of Examples of such chemical the specification requirements was compounds are those con- monitored during quality assurance taining chlorides, fluorides, surveillance activities.
lead, zinc, copper, sulfur, or mercury where such elements This system of controls provides are leachable or where they assurance that the protective coatings could be released by breakdown will properly adhere to the base metal of the chemical compounds under during prolonged exposure to a post-expected environmental condi- accident environment present within tions (e.g., by radiation). the containment building.
This limitation is not in-tended to prohibit the use of b. Category 2 - Intermediate Equipment
trichlorotrifluoroethane, Military Specification The Category 2 equipment consists of the following:
Rev. 17 WOLF CREEK TABLE 6.1-2 (Sheet 4)
DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containment MIL-C-81302b, for cleaning or degreasing of austenitic (1) Seismic platform and tie rods stainless steel provided (2) Reactor internals lifting rig adequate removal is assured. (3) Head lifting rig (4) Electrical cabinets Since these items were procured from a large number of vendors, and individually have very small
surface areas, it was not practical
to enforce the complete set of
stringent requirements which are
applied to Category 1 items.
Another painting specification was
used in these procurement docu-
ments. This specification defined
to the vendors the requirements for: (1) Use of specific coating systems which are qualified to ANSI N101.2 (2) Surface preparation (3) Application of the coating systems in accordance with the paint manu-facturer's instructions Rev. 17 WOLF CREEK TABLE 6.1-2 (Sheet 5)
DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containment The vendor's compliance with the
requirements was also checked during
quality assurance surveillance
activities in the vendor's plant.
These measures of control provide
a high degree of assurance that
the protective coatings will adhere
properly to the base metal and with-
stand the postulated accident envi-
ronment within the containment building. c. Category 3 - Small Equipment Category 3 equipment consists of the
following:
(1) Transmitters (2) Alarm horns (3) Small instruments
(4) Valves (5) Heat exchanger supports These items were procured from several different vendors and were painted by
the vendor in accordance with con-
ventional industry practices. Because
the total exposed surface area is very
small, Westinghouse did not specify
further requirements.
Rev. 17 WOLF CREEK TABLE 6.1-2 (Sheet 6)
DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.54 REVISION 0, DATED JUNE 1973, TITLED "QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components Inside Containment Inside Containmentd. Category 4 - Insulated or Stainless Steel Equipment Category 4 equipment consists of the following: (1) Steam generators - covered with wrapped insulation (2) Pressurizer - covered with wrapped insulation (3) Reactor pressure vessel - covered with rigid reflective
insulation (4) Reactor cooling piping - stainless steel (5) Reactor coolant pump casings - stainless steel Since Category 4 equipment is insulated or is stainless steel, no painted sur-
face areas are exposed within the con-
tainment. Therefore, this regulatory
guide is not applicable for Category 4
equipment.
Rev. 17 WOLF CREEK TABLE 6.1-3 CONTAINMENT COMPONENTS - COATING SCHEDULE Uncoated Category Item/Type/ Description R.G. 1.54 Q coating MFRS. STD. COAT Stainless
Galvanized
Insulation Generic Type (1) Estimated Total Film Thickness (mils) Shop Applied Field Applied Estimated Area (Square Feet) Carbon steel Containment-dome X Inorganic zinc 2-4 X Touch-up 31,000 (4) liner-plate Containment-walls (8) X Inorganic zinc 2-4 X Touch-up 59,000 (4) Structural steel Heavy support steel X Inorganic zinc 2-4 X Touch-up 182,300 (4) Miscellaneous steel X Inorganic zinc 2-4 X Touch-up 16,500 (5) Gratings X 43,700 Elevator Metal siding X 8,500 Tanks and pools Accumulator tanks X Epoxy 4-5 X Touch-up 5,200 Refueling pool X N/A Reactor coolant drain tank X N/A Carbon steel pipe, Pipe X X N/A hangers, valves, Pipe X Inorganic zinc 2-4 X Touch-up 9,100 and supports Pipe supports X Inorganic zinc 2-4 X Touch-up 25,500 Valves and valve actuators X Alkyd/red oxide 2.5-4 X Touch-up 3,500 Mechanical Polar crane X Inorganic zinc 4-7 X 36,700 Equipment Pumps (RCPs) X Epoxy 2-4 X Touch-up 3,000 (including driver) Fans and fan hous- X Epoxy 7.5-11 X 1,200 Ings (carbon steel) X Epoxy 7.5-11 X 400 HVAC ducting X 14,000 (6) HVAC ducting X N/A Steam generators X 15,200 Hydrogen recombiners X N/A Containment coolers X Epoxy 6-10 X 5,400 Containment coolers X 1,100 Heat exchangers X Epoxy 2-4 X Touch-up 300 Internals lifting device X Epoxy 8-12 X 2,000 Containment Tool Room Cabinets X Enamel 1-2 X 850 Rev. 25 WOLF CREEK TABLE 6.1-3 (Sheet 2)
Uncoated Category Item/Type/ Description R.G. 1.54 Q coating MFRS. STD. COAT Stainless
Galvanized
Insulation Generic Type (1) Estimated Total Film Thickness (mils) Shop Applied Field Applied Estimated Area (Square Feet) Electrical Motor control centers X Alkyd/red oxide 1-2.5 X 500 Terminal boxes X 600 Control panels X Epoxy 1.75-3 X 1,000 Raceways, conduit, cable trays, and supports X 38,400 (6)
Cable Rack Assemblies X Thermoset Acrylic 0.5-0.8 X 123(7) Concrete and masonry Floor, cove, and wainscot X Epoxy (2) 12 X 12,900 (3)
NOTES: (1) Generic coating systems acceptable for containment use were selected from suppliers who are prequalified to
Bechtel standards and test criteria. Other coating systems may be shown to be acceptable based on individual
analyses.
(2) Concrete, if painted, was painted with epoxy surfacer or epoxy coating system.
(3) The wainscot extends 12 inches above the floor and was painted the same as described in Note 2, then top
coated with 8 to 10 mils of epoxy-based paint.
(4) Top coated with epoxy at the Wolf Creek Plant. Touch-up of the top coat is a client option.
(5) Includes approximately 900 square feet of nonconforming coating on the maintenance truss which is protected by
screen cover.
(6) Estimated area includes a limited amount of unqualified touch-up coating.
(7) Unqualified coating of cable rack assemblies added as part of simplified head modification.
(8) Containment Normal Sumps and Incore Instrumentation Sump are coated with non-qualified EPOXY coating.
Rev. 27 WOLF CREEK TABLE 6.1-4 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.44, REVISION 0, DATED MAY 1973, TITLED "CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL" Regulatory Guide Position on Non- Position on 1.44 Position NSSS Components NSSS Components Unstabilized, austenitic stain-
less steel of the AISI Type 3XXX
series used for components that
are part of (1) the reactor
coolant pressure boundary, (2)
systems required for reactor
shutdown, (3) systems required
for emergency core cooling, and
(4) reactor vessel internals that
are relied upon to permit
adequate core cooling for any
mode of normal operation or under
credible postulated accident
conditions should meet the
following:
- 1. Material should be suitable 1. Complies. 1. Complies, as discussed protected against containments in Section 5.2.3.4.1.
capable of causing stress
corrosion cracking during
fabrication, shipment, storage, construction, testing, and
operation of components and
systems.2. Material from which com- 2. Complies. 2. Complies, as discussed ponents and systems are to be in Section 5.2.3.4.2.
fabricated should be solution
heat treated to produce a
nonsensitized condition in
the material.
- 3. Nonsensitization of the 3. All austenitic stain- 3 Complies, as discussed in material should be verified less steels were Section 5.2.3.4.3.
using ASTM A 262-70, "Rec- furnished in the solu-
commended Practices for Detecting tion annealed and water-
Attack in Stainless Steel," quenched condition.
Practices A or E, or another Since susceptibility to
method that can be demonstrated stress corrosion crack-
to show nonsensitization ing in this condition
is austenitic stainless steel. is minimal, testing was
Test specimens should be selected not performed.
from material subject to
each different heat treatment
practice and from each heat.
Rev. 0 WOLF CREEK TABLE 6.1-4 (Sheet 2)
Regulatory Guide Position on Non- Position on 1.44 Position NSSS Components NSSS Components
- 4. Material subject to sen- 4. During fabrication and 4. Complies, as discussed
sitizing temperature in the installation, austenitic in Section 5.2.3.4.4.
range of 800ø to 1500øF, stainless steels were not
subsequent to solution heat permitted to be exposed to
treating in accordance with temperatures in the range of
Subparagraph C.2. above and 800ø-1500øF. except for
testing in accordance with welding. Welding practices
Subparagraph C.3. above, should were controlled to minimize
be L Grade material; that is, sensitization, as discussed
it should not have a carbon in Position 5 below.
content greater than 0.03
percent. Exceptions are:
- a. Material exposed to rector coolant which has a
controlled concentration of
less than 0.10 ppm dissolved
oxygen at all temperatures
above 200øF during normal
operation, or
- b. Material in the form of castings or weld metal with a
ferrite content of at least 5
percent; or
- c. Piping in the solution annealed condition whose exposure
to temperatures in the range of
800ø to 1500øF has been limited to
welding operations, provided it is
of sufficiently small diameter so
that in the event of a credible
postulated failure of the piping
during normal reactor operation, the reactor can be shut down and
cooled down in an orderly manner, assuming makeup is provided by the
reactor coolant makeup system
only.5. Material subjected to 5. Heat treatment of austenitic 5. Complies, as discussed sensitizing temperatures in the stainless steel in the Section 5.2.3.4.5.
range of 800ø to 1500øF during heat temperature range 800ø to 1500ø
treating or processing other than F was not permitted. Hot
welding, subsequent to solution bending of austenitic stainless
heat treating in accordance with steel piping was performed.
Rev. 0 WOLF CREEK TABLE 6.1-4 (Sheet 3)
Regulatory Guide Position on Non- Position on 1.44 Position NSSS Components NSSS Components Subparagraph C.2. above, and at the solution annealing
testing in accordance with temperature, followed by an
Subparagraph C.3. above, immediate water quenching. If
should be retested in hot bending was performed at
accordance with Subparagraph some temperature other than
C.3. above, to demonstrate the solution annealing
that it is not susceptible to temperature, the pipes were
intergranular attack, except re-solution annealed and water
that retest is not required quenched. Since sensitization
for: was avoided, testing to
determine susceptibility to
- a. Cast metal or weld metal intergranular attack is not
with a ferrite content of 5 performed.
percent or more: or
- b. Material with a carbon content of 0.03 percent or
less that is subjected to
temperatures in the range of
800ø to 1500øF for less than 1
hour or c. Material exposed to special processing, provided
the processing is properly
controlled to develop a
uniform product and provided
that adequate documentation
exists of service experience
and/or test data to
demonstrate that the
processing will not result in
increased susceptibility to
intergranular stress
corrosion.
Specimens for the above retest should be taken from
each heat of material and
should be subjected to a
thermal treatment that is
representative of the
anticipated thermal conditions
that the production material
will undergo.
- 6. Welding practices and, if 6. Welding practices were 6. Complies, as discussed necessary, material composi- controlled to minimize in Section 5.2.3.4.4. and
tion should be controlled to sensitization in the 5.2.3.4.5.
avoid excessive sensitization heat-affected zone of
Rev. 5 WOLF CREEK TABLE 6.1-4 (Sheet 4)
Regulatory Guide Position on Non- Position on 1.44 Position NSSS Components NSSS Components of base metal heat-affected unstabilized austenitic zones of weldments. An stainless steels, as
intergranular corrosion test, described below.
such as specified in Sub-
paragraph C.3 above, should a. Weld Heat Input
be performed for each
welding procedure to be Heat input during weld-
used for 0.03 percent. ing was controlled by
limiting the size of elec-
trodes for the shielded
metal arc and gas tungsten
arc processes and the bead
thickness for submerged arc
welding. Other welding
processes were not permitted.
- b. Interpass Temperatures..
Interpass temperatures during welding were controlled so as
not to exceed 350øF.
- c. Ferrite Content
Stainless steel welding materials were furnished with
a ferrite content in the range
of 8 to 25 percent for type
308 and 308L welding materials
and 5 to 15 percent for type
316, 316L, 309, and 309L
welding materials. Additional
discussion regarding
compliance to Regulatory Guide
1.31 is provided in Table
6.1-9.
- d. Postweld Heat Treatment
Postweld heat treatment at temperatures in excess of
350øF was not permitted unless
a full-solution anneal and
water quench was performed.
The above welding practices were sufficient to ensure that
unacceptable sensitization of
the base metal heat affected
does not occur; therefore, the
intergranular corrosion
testing was not performed.
Rev. 0 WOLFCREEKTABLE6.1-5DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.37,REVISION0,DATEDMARCH1973,TITLED"QUALITYASSURANCEREQUIREMENTSFORCLEANINGOFFLUIDSYSTEMSANDASSOCIATEDCOMPONENTSOFWATER-COOLEDNUCLEARPOWERPLANTS"RegulatoryGuide1.37PositionWCGSPositionTherequirementsandrecommendationsforon-sitecleaningofmaterialsandcomponents, cleannesscontrol,andpreoperationalclean-ingandlayupofwater-coolednuclearpowerplantfluidsystemsthatareincludedinANSI N45.2.1-1973,"CleaningofFluidSystemsand AssociatedComponentsDuringConstruction PhaseofNuclearPowerPlants,"aregenerallyacceptableandprovideanadequatebasisforcomplyingwiththepertinentqualityassurance requirementsofAppendixBto10CFRPart50, subjecttothefollowing:1.Subdivision1.5ofANSIN45.2.1,1.Complies.1973statesthatotherdocumentsrequiredto beincludedasapartofthestandardareeitheridentifiedatthepointofreferenceor describedinSection10ofthestandard.The specificapplicabilityoracceptabilityof theselisteddocumentshasbeenorwillbe coveredseparatelyinotherregulatory guidesorinCommissionregulations,where
appropriate.2.Althoughsubdivision1.2ofANSIN45.2.1-1973statesthattherequirements promulgatedapplyduringtheconstruction phaseofanuclearpowerplant,manyof therequirementsandrecommendationscon-tainedinthestandardarealsoappropriate tocleaningoffluidsystemsandassociated componentsduringtheoperationphaseofa nuclearpowerplantandtheyshouldbeusedwhenapplicable.Inthisregard,however, itshouldbeparticularlynotedthatdecon-taminationandcleanupofradioactivelycon-taminatedsystemsandcomponentsarenot addressedbyANSIN45.2.1-1973.Theseoper-ationswillbeconsideredseparatelyin futureregulatoryguides.Rev.0 WOLFCREEKTABLE6.1-5(Sheet2)RegulatoryGuide1.37PositionWCGSPosition3.Subdivision3.2ofANSIN45.2.1-3.Complies.1973statesthattheselectionofthewaterqualityforaspecificapplicationshallbemadebytheorganizationresponsibleforthecleaningoperationsunlessotherwise specifiedinthepurchasedocument.The waterqualityforfinalflushesoffluid systemsandassociatedcomponentsshouldbeatleastequivalenttothequalityoftheoperatingsystemwater.4.Section5ofANSIN45.2.1-19734.Complies.states,inpart,thatlowsulfur,low fluorine,and/orlowchlorinecompounds maybeusedonausteniticstainlesssteels andthatlowsulfurandlowleadcompoundsmaybeusedonnickel-basealloys.Chemical compoundsthatcouldcontributetointer-granularcrackingorstress-corrosioncrack-ingshouldnotbeusedwithausteniticstain-lesssteelandnickel-basealloys.Examples ofsuchchemicalcompoundsarethosecontain-ingchlorides,fluorides,lead,zinc,copper, sulfur,ormercurywheresuchelementsare leachableorwheretheycouldbereleasedby breakdownofthecompoundsunderexpected environmentalconditions(e.g.,byradiation).
Thislimitationisnotintendedtoprohibit theuseoftrichlorotrifluoroethanewhich meetstherequirementsofMilitarySpecifica-tionMil-C-81302bforcleaningordegreasing ofausteniticstainlesssteelprovidedtheprecautionsofsubdivision7.3(4)ofANSI N45.2.1-1973areobserved.5.Section5ofANSIN45.2.1-19735.Complies.states,inpart,thatoperationssuchas grindingandweldingwhichgenerateparticu-latemattershouldbecontrolled.Adequate controloftoolsusedinabrasiveworkopera-tionssuchasgrinding,sanding,chipping,or wirebrushingshouldbeprovided.Specifi-cally,toolswhichcontainmaterialsthat couldcontributetointergranularcrackingor stress-corrosioncrackingorwhich,because ofprevioususagemayhavebecomecontami-natedwithsuchmaterials,shouldnotbeused onsurfacesofcorrosion-resistantalloys.Rev.0 WOLFCREEKTABLE6.1-5(Sheet3)RegulatoryGuide1.37PositionWCGSPositionExamplesofsuchmaterialsarelistedinRegulatoryPosition4.6.Subdivision1.4ofANSIN45.2.1-19736.Complies.suggeststheuseofASTMA262-68orASTMA393-63fordetectionofintergranularprecipi-tationofchromiumcarbidesincorrosion-resistantalloys.ASTMA393-63hasbeenwithdrawnbyASTMandisnolongerconsideredavalidtest.Rev.0 WOLFCREEKTABLE6.1-6DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.36,REVISION0,DATEDFEBRUARY1973,TITLED"NONMETALLICTHERMALINSULATIONFORAUSTENITICSTAINLESSSTEEL"RegulatoryGuide1.36PositionWCGSPositionThelevelsofleachablecontaminantsinnonme-tallicinsulationmaterialsthatcomeintocon-tactwithausteniticstainlesssteelsoftheAmericanIron&SteelInstitute(AISI)Type3XXseriesusedinfluidsystemsimportanttosafety shouldbecarefullycontrolledsothatstress-corrosioncrackingisnotpromoted.Inpartic-ular,theleachablechloridesandfluoridesshouldbeheldtothelowestpracticablelevels.Insulationfortheaboveapplication shouldmeetthefollowingconditions:1.Allinsulatingmaterialsshouldbe1.Complies.manufactured,processed,packaged,shipped, stored,andinstalledinamannerthatwill limit,tothemaximumextentpractical,chlorideandfluoridecontaminationfrom externalsources.2.QualificationTest:Eachtypeof2.Complies.insulatingmaterialshouldbequalifiedby themanufacturerorsupplierforuseby:a.Anappropriatetesttoreasonablyassurethattheinsulationformulationdoes notinducestresscorrosion.Twoacceptable testsare:(1)ASTMC692-71,"StandardMethodforEvaluatingStressCorrosionEffect ofWicking-TypeThermalInsulationsonStainless Steel"(DanaTest).Thematerialshouldbe rejectedifmorethanoneoffivespecimenscrack;and(2)RDTM12-lT,"TestRequire-mentsforThermalInsulatingMaterialsfor UseonAusteniticStainlessSteel,"Section5 (KnollsAtomicPowerLaboratory(KAPL)Test).
Thematerialshouldberejectedifmorethan oneoffourspecimenscrack.Rev.0 WOLF CREEK TABLE 6.1-6 (Sheet 2)
Regulatory Guide 1.36 Position WCGS Position
- b. Chemical analysis to determine the ion concentrations of leachable chloride, fluoride, sodium, and silicate. Insulating material that is not demonstrated by the analysis to be within the acceptable region
of Figure l of this guide should be rejected.
This analysis should also be used as a com-
parison basis for the production test specified in C.3 below.
- 3. Production Test: A representative 3. Complies, sample from each production lot of insulation except that the material to be used adjacent to, or in contact representative with, austenitic stainless steels used in sample will be fluid systems important to safety should be chemically chemically analyzed to determine leachable analyzed as in chloride, fluoride, sodium, and silicate C.2.b above.
ion concentrations as in C.2.a above. The lot should be accepted only if:
- a. The analysis shows the material to be within the acceptable region of Figure 1;
and b. Neither the sum of chloride plus fluoride ion concentrations nor the sum of
sodium plus silicate ion concentrations de-
termined by this analysis deviates by more
than 50 percent from the values determined
on the sample used to qualify the insulation
in C.2 above.
- 4. Requalification: When a change is 4. Complies.
made in the type, nature, or quality of the
ingredients, the formulation, or the manu-
facturing process, the insulation material
should be requalified by repeating the tests
described in C.2 above.
Rev. 16 WOLFCREEKTABLE6.1-7DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.50REVISION0,DATEDMAY1973,TITLED,"CONTROLOFPREHEATTEMPERATURESFORWELDINGOFLOW-ALLOYSTEEL"RegulatoryGuidePositiononNon-Positionon1.50PositionNSSSComponentsNSSSComponentsWeldfabricationforlowWestinghouseconsidersthisalloysteelcomponentsRegulatoryGuideapplicableshouldcomplywiththefab-onlytoASMEIII,Class1ricationrequirementsspec-components.ifiedinSectionIIIandSectionIXoftheASMEB&PV Codesupplementedbythe
following:1.Theprocedurequalifi-1.Paragraph1.awas1.a.Complies,forClass1cationshouldrequirethat:compliedwithwhenimpactcomponents.testing,inaccordancea.AminimumpreheatandwithASMEBoilerandamaximuminterpasstempera-PressureVesselCode,turebespecified.SectionIII,Subarticle2300,wasrequired.WhenImpacttestingwasnotrequired,specificationofamaximuminterpasstemperatureintheweldingprocedureswasnotnecessaryinordertoassurethattheotherrequiredmechanicalpropertiesoftheweldare
met.b.Theweldingprocedures1.b.Complies1.b.ForClass1components,bequalifiedattheminimumweldingproceduresarepreheattemperature.qualifiedwithinthe preheattemperaturerangesrequiredbySectionIXoftheASMEBoilerand PressureVesselCode.2.Forproductionwelds,the2.Compilesforpressure2.Complianceisdiscussed preheattemperatureshouldbevesselswithnominalthick-InSection6.1.4, maintaineduntilapostweldnessesgreaterthan1inch.Reference3.
heattreatmenthasbeenper-Maintenanceofpreheatformed.beyondcompletionofweldingRev.Ountilpostweldheattreatment WOLFCREEKTABLE6.1-7(Sheet2)RegulatoryGuidePositiononNon-Positionon1.50PositionNSSSComponentsNSSSComponents(PWHT)wasnotrequired forthinnersections, sinceexperiencehas indicatedthatdelayed crackingintheweld orheataffectedzone (HAZ)isnotaproblem.3.Productionweldingshould3.Currentusageoflow3.Complies,forClass1bemonitoredtoverifythatalloysteelinpiping,components.thelimitsonpreheatandpumps,andvalvesis interpasstemperaturesareminimalandisnormally maintained.limitedtoClass3con-struction.Whenlowalloy steelpiping,pumps,and valveswereused,preheat wasmaintaineduntilwelding iscomplete,butnotuntil postweldheattreatment (PWHT)wasperformed,since theconditionswhichcause delayedcrackinginthe weldorheataffectedzone (HAZ)werenotpresent.4.Intheeventthat4.Complies.4.Complies,forClass1regulatorypositions components.C.1,C.2,andC.3,above,arenotmet, theweldissubjectto rejection.However, thesoundnessoftheweldmaybeverifiedRev.Obyanacceptableexaminationprocedure.
WOLFCREEKTABLE6.1-8DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.71REVISION0,DATEDDECEMBER1973,TITLED,WELDERQUALIFICATIONFORAREASOFLIMITEDACCESSIBILITY"RegulatoryGuidePositiononNon-Positionon1.71PositionNSSSComponentsNSSSComponentsWeldfabricationandrepair forwroughtlow-alloyor othermaterialssuchas staticandcentrifugal castingsandbimetallic jointsshouldcomplywith thefabricationrequire-mentsspecifiedinSection IIIandSectionIXsupple-mentedbythefollowing:1.Theperformancequalifi-1.Performancequalifi-1.Performancequalificationcationshouldrequiretestingcationsforpersonnelforrequalification)of ofthewelderundersimulatedwhoweldunderconditionswelderforareasoflimited accessconditionswhenphysicaloflimitedaccesswereaccessibilitywasnotrequired.conditionsrestrictthewelder'smaintainedinaccordanceExperienceshowsthatcurrentaccesstoaproductionweldwiththeapplicablere-shoppracticesproducehigh tolessthan30to35cm(12quirementsofASMEqualitywelds.Inaddition, to14inches)inanydirectionSectionsIIIandIX.theperformanceofnondestruc-fromthejoint.Additionally,respon-tiveexaminationsprovidessiblesitesupervisorsfurtherassuranceofaccept-wererequiredtoassignableweldquality.Limited onlythemosthighlyaccessibilityqualification skilledwelderstoforrequalification)in limited-accesswelding.excessofASMECode,Section Ofcourse,weldingIIIorIXrequirementswasan conductedinareasofundulyrestrictiverequirement limitedaccesswasforcomponentfabrication, subjectedtothewherethewelder8physical requirednondestruc-positionrelativetothe tivetesting,andnoweldsiscontrolled.
waiverorrelaxation ofexaminationmethodsoracceptancecriteriabecauseofthelimited accesswaspermitted.2.Requalificationis
required:2.Requalificationwasre-2.SeeresponsetoIabove.a.Whensignificantlyquired:whenanyofdifferentrestrictedaccess-theessentialvariables abilityconditionsoccur,orofASMESectionIXarechanged,orwhenanyRev.0 authorizedinspector questionstheability WOLFCREEKTABLE6.1-8(Sheet2)RegulatoryGuidePositiononNon-Positionon1.71PositionNSSSComponentsNSSSComponentsb.Whenanyoftheoftheweldertoperformessentialweldingvariablessatisfactorilythere-listedinSectionIXarequirementsofASMESec-changed.tionIIIorIX.3.Productionwelding3.Productionweldingwas3.Seeresponseto1above.shouldbemonitoredandmonitoredandwelding adherencetoweldingqualificationswere qualificationrequire-certifiedinaccordance mentsshouldbecertified.with(1)and(2)above.Rev.O WOLF CREEK TABLE 6.1-9 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.31, REVISION 3, DATED APRIL 1978, TITLED, "CONTROL OF FERRITE CONTENT ON STAINLESS STEEL WELD METAL" Requirements of this regulatory guide are applied to production weld (full penetration pressure boundary welds)
which could be subject to microfissures due to low delta ferrite content of the deposited weld metal, in
austenitic stainless steel ASME Section III, Class 1 and 2 components, and core support structures.
Regulatory Guide Position on Position on 1.31 Position Non-NSSS Components NSSS Components
- 1. Verification of Delta 1. Portions of the non- 1. Field welding of Ferrite Content of Filler NSSS components con- NSSS components Is Materials form to the require- done in accordance ments of Revision 3 with Revision 3 of Prior to production usage, the of this regulatory this regulatory delta ferrite content of test guide. guide.
weld deposits from each lot and each heat of weld filler metal The remainder of the Section 5.2.3.4.6 procured for the welding of non-NSSS components, describes the extent austenitic stainless steel core fabricated prior to the of compliance to the support structures and Class 1 implementation of Revi- NRC interim position and 2 components should be veri- sion 3 of this regulatory on Revision 1 of this fled for each process to be used guide, conform to the re- regulatory guide of In production. quirements specified In the NSSS supplied and the PSAR position on the fabricated components.
It is not necessary to make NRC interim position on delta ferrite determination Revision 1 of this regu-for SFA-5.4 type 16-8-2 weld latory guide.
metal or for filler metal used for weld metal cladding. Delta The requirements of the ferrite determinations for con- PSAR include magnetic sumable Inserts, electrodes, rod testing of randomly or wire filler metal used with selected production the gas tungsten arc welding welds made from wire process, and deposits made with whose delta ferrite the plasma arc welding process content was determined may be predicted from their from constitution diagrams.
chemical composition using an applicable constitutional dia- Revision 3 requirements gram to demonstrate compliance. include the requirement Delta ferrite verification to determine the delta should be made for all other ferrite content of the processes by tests using mag- weld wire by magnetic netic measuring devices on un- tests on undiluted test diluted weld deposits. For pads. Production weld submerged arc welding processes, testing is not required the verification tests for each by Revision 3 of this Rev. 23 wire and flux combination may regulatory guide WOLF CREEK TABLE 6.1-9 (Sheet 2)
Regulatory Guide Position on Position on
1.31 Position Non-NSSS Components NSSS Components be made on a production weld
or simulated production weld.
All other delta ferrite weld filler verification tests
should be made on weld pads
that contain undiluted layers
of weld metal.
- 2. Ferrite Measurement 2. Compiles where magnetic 2. See Section 5.2.3.4.6 testing was performed to Appendix A to this guide verify the weld filler contains extracts from a material as described future edition of the in 1. above.
American Welding Society's AWS A5.4, "Specification
for Corrosion-Resisting
Chromium and Chromium-Nickel
Steel Covered Welding
Electrodes," which de-
scribes a procedure for
pad preparation and
ferrite measurement.
The NRC staff considers
this procedure acceptable
for use with covered
electrodes.
- 3. Instrumentation 3. Compiles where magnetic 3. See Section 5.2.3.4.6 testing was performed to The weld pad should be verify the weld filler examined for ferrite con- material as described tent by a magnetic mea- in 1. above. When suring Instrument which production weld testing has been calibrated against was performed to support a Magnegage In accordance chemical composition with American Welding similar Instrumentation Society Specifications AWS requirements were met.
A4.2-74, "Procedures for
Calibrating Magnetic
Instruments to Measure
the Delta Ferrite Content
of Austenitic Stainless Rev. O Steel Weld Metal." The WOLF CREEK TABLE.E 6.1-9 (Sheet 3)
Regulatory Guide Position on Position on
1.31 Position Non-NSSS Components NSSS Components Magnegage should have
been previously cali-
brated in accordance with AWS A4.2-74 using
primary standards as
defined therein.
- 4. Acceptability of Test Results 4. Complies. 4. See Section 5.2.3.4.6
Weld pad test results
showing an average Ferrite
Number From 5 to 20 Indicate
that the filler metal is acceptable for production welding of Class 1 and 2
austenitic stainless steel
components and core support
structures.
The upper limit of 20 may be
waived for (a) welds that do
not receive postweld stress
relief heat treatment or
welds for which such postweld
stress relief treatment is
conducted at temperatures
less than 900 F, (b) welds
that are given a solution annealing heat treatment, and (c) welds that employ
consumable inserts.
- 5. Quality Assurance 5. Complies. 5. See Section 5.2.3.4.6
She applicable provisions
of 10 CFR Part 50, Appendix P, should be used in verifying
compliance with requirements
for delta ferrite as described
herein.
Rev. 0 WOLF CREEK TABLE 6.1-10 TABLE OF LUBRICANTS INSIDE CONTAINMENT Equipment Lubricant Type Quantity (4)Reactor coolant pumps (1) Oil 265 gal Polar crane (2) Wire rope lube 24 lb Gear lube 10 lb Graphite lube sticks 8 lb Miscellaneous hoists and cranes (2) Wire rope lube 1 lb Miscellaneous fans (3) Oil/grease Neg Miscellaneous pumps (3) Oil/grease Neg Steam generator hydraulic Hydraulic fluid Neg snubbers (3)NOTES: (1) Assumes lube oil from one RC pump spills into sump.
(2) Assumes 10 percent will be washed off by containment spray. (3) Motors, bearings, and snubbers are enclosed.
(4) Quantity subject to be released into the containment.
Rev. 19 WOLF CREEK 6.2 CONTAINMENT SYSTEMS The containment systems include the containment, the containment heat removal
systems, the containment isolation system, and the containment combustible gas
control system.
The design basis accident (DBA) is defined as the most severe of a spectrum of
hypothetical loss-of-coolant accidents (LOCA). The ability of the containment
systems to mitigate the consequences of a DBA depends upon the high reliability
of these systems. This section provides the design criteria and evaluations to
demonstrate that these systems function within the specified limits throughout
the unit operating lifetime.
6.2.1 CONTAINMENT FUNCTIONAL DESIGN
A physical description of the containment and the design criteria relating to construction techniques, static loads, and seismic loads is provided in Section
3.8. This section pertains to those aspects of containment design, testing, and evaluation that relate to the accident mitigation function.
6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases
The safety design basis for the containment is that the containment must withstand the pressures and temperatures of the DBA without exceeding the
design leakage rate, as required by 10 CFR 50, Appendix A, General Design
Criterion 50, and that, in conjunction with the other containment systems and
the other engineered safety features, the release of radioactive material
subsequent to a DBA does not result in doses in excess of the guideline values
specified in 10 CFR 100. The radiological consequences of the DBA are
presented in Section 15.6.
- a. Assumed Accident Conditions
For the purpose of determining the design pressure requirements for the containment structure and the
containment internal structures, the following
simultaneous occurrences were assumed:
- 1. The postulated reactor coolant system pipe rupture, as listed in Table 6.2.1-1, was assumed to be
concurrent with the loss of offsite power and the
worst single active failure. No two pipe breaks were
assumed to occur simultaneously or consecutively.
6.2-1 Rev. 0 WOLF CREEK For design loadings on the systems used to mitigate the consequences of a
postulated reactor coolant system pipe rupture, a safe shutdown earthquake was
assumed.
- 2. The postulated secondary system pipe rupture, as
identified in Section 6.2.1.4, was assumed concurrent
with the loss of offsite power and the worst single
active failure. No two pipe breaks were assumed to
occur simultaneously or consecutively.
- 3. The postulated inadvertent operation of a containment
heat removal system is considered a low probability
event and was not considered to be concurrent with
any other event.
The calculated maximum containment structure internal and external pressures
are listed in Table 6.2.1-2. These calculated pressures are based on the
conservative analyses described in Section 6.2.1.1.3 and demonstrate that a
substantial margin exists (approximately 20 percent on internal pressure and 10
percent on external pressure) between the calculated maximum pressure and the
design pressures.
The calculated maximum pressures on the containment internal structures (e.g.
primary and secondary shield walls) are listed in Table 6.2.1-2. These
pressures are based on the conservative analyses described in Section 6.2.1.2.
The loads on the internal structures were calculated using the differentials
between the maximum calculated subcompartment pressures and 14.7 psia, the
pressure of the containment atmosphere at the time of peak subcompartment
pressure.
- b. Sources and Amounts of Mass and Energy Released
The sources and amounts of mass and energy released for
the postulated reactor coolant system pipe ruptures and
secondary system pipe ruptures are discussed in Sections
6.2.1.3 and 6.2.1.4, respectively.
- c. Effects of the ESFs as Heat Removal Systems
The effects of the ECCS as an energy removal system are
discussed in the determination of the mass and energy
release data provided in Section 6.2.1.3. The only
additional effect of this system considered is the long-
term heat removal capability of the residual heat removal
heat exchangers. In performing the containment design
6.2-2 Rev. 0 WOLF CREEK evaluation in Section 6.2.1.1.3, single failures of the
ECCS are assumed to be consistent with the mass and
energy release data assumptions for the break analyzed.
The effects of the containment heat removal systems, as
active energy removal systems, have been considered in
the containment design evaluation in Section 6.2.1.1.3.
The most stringent single active failure of these systems
is assumed to be consistent with the mass and energy
release data assumptions for each break analyzed. The
total heat removed by each of the containment heat
removal systems up to the time of the calculated peak
containment pressure is listed in Table 6.2.1-8. The
design bases of the containment heat removal systems are discussed in Section 6.2.2.
The functional performance of the containment and the
ECCS also rely upon the operation of the containment
isolation system, as described in Section 6.2.4. Required
isolation operations are assumed for purposes of the
containment design evaluation in Section 6.2.1.1.3.
- d. Parameters Affecting Capability for Post-Accident
Pressure Reduction
The principal parameters which affect post-accident
pressure reduction are 1) the heat absorbed by the heat
sinks inside the containment, 2) the heat removed by the
containment air coolers, and 3) the heat transferred to the containment sump by the containment spray system.
A conservative amount of heat sink material has been
calculated, and its heat absorption capability has been
considered in the containment design evaluation in
Section 6.2.1.1.3. The parameters describing the heat
sinks credited with heat absorption are provided in Table
6.2.1-4.
The pressure reduction capability of the containment air
coolers and the containment spray system consider the
parameters provided in Table 6.2.1-3. The assumed start
time of these active heat removal systems considers a
diesel start time of 12 seconds, load sequencing times, and the maximum startup time of the systems.
6.2-3 Rev. 0 WOLF CREEK In support of case c, large break LOCA (DECLG C D = 0.6, Maximum SI with 12-second diesel generator start) of
Table 15.6-10, an evaluation of the assumptions used in
the LOCA and MSLB containment pressurization calculations, with respect to the full functioning times
of the containment spray system and the containment air
coolers, was performed. The evaluation shows that the
containment pressurization calculations for both LOCA and
MSLB provided sufficient margin so that a 12-second
diesel generator start time does not change the assumed
full functioning times of the containment spray and the
containment air coolers. Therefore, additional LOCA and
MSLB containment pressurization calculations are not
required for case c of Table 15.6-10, since this case is bounded by the previously performed containment calculations.
- e. Parameters Affecting Heat Removal from the Containment
Heat is transferred from the containment to the outside
environment via the fan coolers and residual heat removal
heat exchangers through the component cooling water and
essential service water systems and released to the
ultimate heat sink. A small amount of heat is also
transferred through the containment wall and dome to the
outside atmosphere.
The component cooling water system is described in
Section 9.2.2, the essential service water system is described in Section 9.2.1, and the ultimate heat sink is described in Section 9.2.5.
Single failures in systems which remove energy from the
containment are considered to be consistent with the
single failures assumed in the development of the mass
and energy release data. The energy removal capability
of the containment air coolers, the containment spray
system, and the residual heat removal system consider the
parameters provided in Table 6.2.1-3. The long-term
energy inventories and total heat transferred to the
various containment heat removal mechanisms, as a
function of time, are diagrammed in Figures 6.2.1-25 and
6.2.1-26 for the double-ended pump suction guillotine
(DEPSG) break with minimum safety injection and DEPSG
break with maximum safety injection cases, respectively.
6.2-4 Rev. 0 WOLF CREEK
- f. Bases for Containment Depressurization Rate
To meet the containment safety design basis of limiting
the release of radioactive material subsequent to a DBA so that the doses are within the guideline values
specified in 10 CFR 100, the containment pressure is
reduced to less than 50 percent of the peak calculated
pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident. Chapter 15.0
contains the assumptions used in the analysis of the
offsite radiological consequences of the accident.
- g. Bases for Minimum Containment Pressure Used in ECCS
Performance Studies
The minimum containment pressure transient used in the analysis of the emergency core cooling system's
capability is based on the conservative overestimated
heat removal capability and pressure reduction capability
of the containment structures and the containment systems
and on the conservative reactor coolant system thermal
analysis provided in Section 15.6. The determination
and evaluation of the minimum containment pressure
transient are provided in Section 6.2.1.5.
6.2.1.1.2 Design Features
The principal containment and containment subcompartment design parameters are
provided in Table 6.2.1-2. General arrangement drawings for the reactor
containment are provided in Figures 1.2-9 through 1.2-18. Simplified arrangement drawings illustrating the nodalization model used for the containment subcompartment analyses are provided in Figures 6.2.1-27 through
6.2.1-33, 6.2.1-43 through 6.2.1-55, and 6.2.1-76.
- a. Missile and Pipe Whip Protection
Missile shield considerations are described in Section
3.5. The structural design of the containment and the
containment subcompartments is discussed in Section 3.8.
The designed structural strength considers the effects of
pipe whip and jet forces, as discussed in Section 3.6.
- b. Codes and Standards
The codes, standards, and guides applied in the design of
the containment structure and the containment internal
structures are identified in Section 3.8.
6.2-5 Rev. 15 WOLF CREEK
- c. Inadvertent Operation of the Containment Spray System
The design external pressure load on the reactor
containment is provided in Table 6.2.1-2. The lowest calculated internal pressure is also provided in Table
6.2.1-2, and is the result of an assumed inadvertent
actuation of the containment spray system. The analysis
performed to determine the lowest calculated internal
pressure following an inadvertent actuation of the
containment spray system is provided in Section
6.2.1.1.3. At least a 10-percent margin exists between
the lowest calculated internal pressure and the design
external pressure load.
- d. Entrapment of Recirculation Water
Locations within the reactor containment which may entrap
spray water and subtract from the water inventory
considered to be available in the containment sump are
identified in Section 6.2.2.1. The effect of this
potential water loss is considered in determining the net
positive suction head available to the RHR and
containment spray pumps. Any special provisions which
reduce the amount of the entrapped water are discussed in
Section 6.2.2.1.
- e. Normal Operation of Systems Which Control the Containment
Environment
The functional capability and frequency of operation of the systems provided to maintain the containment and
subcompartment atmospheres within prescribed pressure, temperature, and humidity limits during normal operation
are discussed in Sections 6.2.2.2 and 9.4.6.
6.2.1.1.3 Design Evaluation
- a. Analysis of Postulated Ruptures
In the event of a LOCA in the containment, much of the
released reactor coolant will flash to steam. This
release of mass and energy raises the temperature and
pressure of the atmosphere within the containment. The
severity of the temperature and pressure peaks depends
upon the nature, size, and location of the postulated
rupture.
6.2-6 Rev. 0 WOLF CREEK In order to identify the worst case, the spectrum of hypothetical accidents listed in Table 6.2.1-1 has been
analyzed. The analytical model and computer code
designed to predict containment pressure and temperature
responses following the accidents are described in item
- b. of this section.
A summary of the results of the containment pressure and
temperature analysis for this spectrum of postulated
accidents is provided in Table 6.2.1-8. The peak
containment pressure calculated resulted from the assumed
(DEPSG) break with minimum safety injection and with the
worst single failure being the loss of one emergency
diesel.
The calculated containment pressure and temperature responses as a function of time for the spectrum of
postulated accidents are illustrated in Figures 6.2.1-1
through 6.2.1-12.
To assess the impact of the plant uprating on the containment integrity
analyses, the containment pressure and temperature responses following a
postulated LOCA are re-analyzed at uprated power (3565 MWt) conditions using
the mass and energy releases data that are re-generated using the Westinghouse
1979 M&E release model (Ref. 26). The analyses are performed for a spectrum of
possible pipe break sizes and locations at rerated conditions to assure that
the worst case has been identified.
The results of the containment integrity analyses at rerated conditions show
that the peak calculated containment pressure and temperature following a postulated LOCA are less limiting than the peak containment pressure and temperature calculated in the original WCGS licensing basis analyses. Since
the long term LOCA mass and energy releases utilized in the original licensing
basis containment integrity analyses were already based upon an NSSS power of
3579 MWt, the original licensing basis containment analysis information
presented in the USAR Section 6.2.1 remains unchanged as a result of the rerate
analysis. The peak containment pressure and temperature calculated in the
original WCGS licensing basis analyses are shown in Figures 6.2.1-1 and 6.2.1-
- 7.
- b. Computer Codes for Analyses of Containment Response to
The temperature and pressure conditions in the
containment due to a spectrum (including break size and
location) of postulated loss-of-coolant accidents have
been analyzed by the CONTEMPT-LT/28 computer code
(Ref. 24). CONTEMPT-LT/28 is the most recent code in a
series of computer programs developed by the Idaho
National Engineering Laboratory for the USNRC to analyze
the transient behavior of LWR containment systems. The
licensee's application of CONTEMPT-LT/28 for the
calculation of the containment environment response to a
postulated primary or secondary pipe break has been
reviewed and accepted by the NRC (Ref. 25).
6.2-7 Rev. 12 WOLF CREEK CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. Models are provided to describe containment heat removal systems that include fan cooler systems, spray systems and residual heat removal systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the primary system can have a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. The program user defines which compartments are used, specifies input mass and energy additions, defines heat structure and leakage systems, and describes the time advancement and output control. The mass and energy release data used in the analysis are developed and described in Section 6.2.1.3.
The heat removal due to containment air cooler operation is simulated in the CONTEMPT-LT/28 code by specifying input values from a curve of heat removal rate versus containment atmosphere temperature. This curve is based upon the cooling coil thermal-physical design and is given in Figure 6.2.1-15. The fan coolers assumed start time is provided in Tables 6.2.1-6 and 6.2.1-7 for the DEPSG breaks with minimum safety injection and with maximum safety injection, respectively. This start time is based upon the diesel start time of 12 seconds, the loading sequence, and the startup time of the system.
The parameters describing the containment air coolers are given in Table 6.2.1-3.
- c. Initial Conditions
Initial conditions used for the containment analysis are listed in Table 6.2.1-5.
The initial containment conditions were selected based on the range of the normal expected conditions within the containment with consideration given to maximizing the calculated peak containment pressure. Parametric studies have been performed to determine the effects of varying these initial containment conditions (Ref. 1). The results of these studies showed that varying the initial containment conditions over a wide range of values changes the calculated peak pressure by less than 1 percent. Therefore, the initial containment conditions are relatively unimportant parameters with respect to the containment pressure temperature analysis. Nevertheless, in order to account for initial temperature instrument uncertainty, the limiting case (DEPSG break with minimum safety injection) was evaluated with a bounding initial containment temperature of 130 F.
The conservatisms in the assumptions made with respect to the containment heat removal systems and the emergency core cooling system operability are discussed in Sections 6.2.2 and 6.3, respectively.
6.2-8 Rev. 29 WOLF CREEK
- d. Results of the Failure Mode and Effects Analysis
Single active failures have been considered in the emergency core cooling system and in the containment heat removal systems with respect to maximizing energy release to the containment and minimizing the heat removal from the containment. The criteria used to determine the worst single active failure was the calculated peak containment pressure. Therefore, single active failures in the containment heat removal systems were considered consistent with the mass and energy release data determined by the corresponding common mode failure in the emergency core cooling system.
The worst calculated peak containment pressure was the result of a double-ended reactor coolant pump suction guillotine break with minimum performance of the emergency core cooling system and the containment heat removal systems.
- e. Containment Design Parameters
The principal containment design parameters are provided in Table 6.2.1-2.
- f. Engineered Safety Features Design Parameters
The engineered safety features design parameters used in the containment analysis are listed in Table 6.2.1-3.
The parameters identified as full capacity were used when no failure was assumed to affect the operation of that system, and the parameters identified of minimum capacity were used when a single failure was assumed to affect the operation of that system. The containment air cooler duty curve per air cooler used in the analysis is given in Figure 6.2.1-15. The limiting case (DEPSG break with minimum safety injection) modeled the containment air cooler duty curve given in Table 6.2.1-57C.
The containment air cooler duty curve per air cooler used in the remaining analyses is given in Figure 6.2.1-15.
- g. Results of Postulated Accidents Analyzed
A summary of the results of the containment pressure temperature analyses for the spectrum of postulated accidents is tabulated in Table 6.2.1-8.
- h. Secondary System Pipe Rupture Containment Analysis
A complete discussion of secondary system pipe ruptures inside the containment with respect to the containment pressure and temperature responses is provided in Section
6.2.1.4.
6.2-9 Rev. 29 WOLF CREEK
- i. Containment Passive Heat Sinks
With respect to the modeling of the containment passive
heat sinks for the heat transfer calculations used in the containment pressure temperature analysis, the method
discussed in Reference 24 provides the justification for
- 1) the computer mesh spacing used for concrete, steel,
and steel-lined concrete heat sinks, 2) the steel-
concrete interface resistance used for the steel-lined
concrete heat sinks, and 3) the heat transfer
correlations used in the heat transfer calculations.
The specific passive heat sinks considered in the
containment pressure temperature analysis and their parameters are listed in Table 6.2.1-4. Figures 6.2.1-13 and 6.2.1-14 show the condensing heat transfer
coefficient as a function of time for the DEPSG with
minimum safety injection and DEPSG with maximum safety
injection cases, respectively.
Zero heat transfer is specified at the outside surface of
the containment exposed to the earth, and between the
containment sump liquid and the containment atmosphere
within the containment.
- j. Analysis of Inadvertent Operation of a Containment Heat
Removal System
Inadvertent actuation of the containment spray system results in the lowest calculated containment internal pressure.
As discussed in Section 6.2.2.1, the containment spray
system can only be actuated in two ways, either
automatically upon receipt of two-out-of-four
containment high pressure signals or manually from
the control room.
Section 7.3.8 discusses the engineered safety features
actuation system and demonstrates that the system design
precludes a single active or passive failure from
inadvertently actuating the containment spray system.
Manual actuation of the containment spray system can only
be accomplished by the reactor operator deliberately
switching on two switches on the main control board. The
main control board is designed with physical separation
of these switches to prevent accidental actuation of the
spray system. Thus, inadvertent actuation of the sprays
6.2-10 Rev. 6 WOLF CREEK is precluded by design, and only a deliberate actuation
of the containment spray system could result in the
reactor building being sprayed.
Although precluded by design, inadvertent actuation of
the containment spray system has been assumed, and the
resultant reduction in the containment pressure has been
calculated. The postulated inadvertent actuation of the
containment spray system is assumed, concurrent with the
following conservative containment and environmental
conditions:
Summer Winter
Initial containment temperature,°F 120 100 Initial containment pressure, psia 14.7 14.7
Initial containment relative
humidity, % 100 100
Containment spray flow rate, gpm
(per train) 3,900* 3,900
- RWST water temperature,°F 60 37
- Runout flow rates for the containment spray system.
Actuation of the containment spray system could be
postulated under any set of containment and environmental conditions. However, no consistent set of realistic conditions can categorically be selected as the
worst case initial condition to be used in the
containment pressure analysis. These assumed initial
conditions are defined as limiting in that these
conditions 1) represent the largest differences in the
containment ambient temperature and the RWST temperature
and 2) the 100-percent humidity case maximizes the
amount of mass transferred out of the containment
atmosphere.
Using Henry's law of partial pressures and the Ideal Gas
Law and assuming that the inadvertent operation of the
containment spray system will reduce the containment
vapor temperature to coincide with that of the RWST water
being sprayed, the maximum reduction in the containment
pressure is provided in Table 6.2.1-2.
6.2-11 Rev. 0 WOLF CREEK The containment design external pressure load is provided
in Table 6.2.1-2, and shows a minimum of 10-percent
margin above the maximum reduction in the containment
pressure calculated by the above-described method. Thus, corrective action by the operator is not required to
ensure that containment integrity is maintained.
The control room operator is notified that the
containment spray system is operating through the
following means:
- 1. The containment spray actuation annunciator light is
on, and an audible alert alarm is sounded.
- 2. The running status light of the containment spray pumps is on.
- 3. The open status lights of the containment spray
system isolation valves is on.
- 4. The containment normal sump and the incore
instrumentation tunnel level indicators and level
alarms are actuated.
- 5. The flow indicators for the discharge of the
containment spray pumps indicate flow in the
containment spray pumps.
- 6. The NPIS computer visually informs the operator that the containment spray system is actuated.
- k. Accident Chronology
The chronology of events occurring after a DEPSG break
with minimum safety injection is given in Table 6.2.1-6.
The chronology of events after a DEPSG break with maximum
safety injection is given in Table 6.2.1-7.
- 1. Mass and Energy Balances
A mass and energy balance for the reactor coolant system, steam generators, and the safety injection system is
provided in Section 6.2.1.3.2 and shows the distribution
of energy prior to the accident, at the end of the
blowdown phase, at the end of the core reflood phase, and
at the end of the post-reflood phase.
6.2-12 Rev. 15 WOLF CREEK A mass and energy balance for the reactor and containment
systems for the DEPSG break with minimum safety
injection and DEPSG break with maximum safety injection
are provided in Tables 6.2.1-9 and 6.2.1-10, respectively. These tables provide the distribution of
energy at the following times:
- 1. Prior to the accident
- 2. Blowdown peak pressure
- 3. End of blowdown
- 4. Peak containment pressure
- 5. End of reflood
- 6. Approximately one day after recirculation
- m. Long-Term Cooling Following a LOCA
The long-term system behavior during various LOCAs has
been evaluated to verify the ability of the ECCS and the
containment heat removal systems to keep the reactor
vessel flooded and maintain the containment below design
conditions for all times following a LOCA. This
evaluation is based on the conservative predictions of
the performance of these engineered safety features
consistent with the single failures assumed for each accident analyzed. The heat generation rate from shutdown fissions, heavy isotope decay, and fission
product decay is provided in Figure 6.2.1-16.
The containment pressure and temperature transients for
the DEPSG break with minimum safety injection up to 10 6 seconds are shown in Figures 6.2.1-1 and 6.2.1-7, respectively. These figures demonstrate the containment
systems' capability of rapidly reducing the containment
pressure and temperature and maintaining those parameters
to acceptably low values. The containment pressure and
temperature transients for the DEPSG break with maximum
safety injection up to 10 6 seconds are shown in Figures 6.2.1-2 and 6.2.1-8, respectively. For all other
accidents analyzed, the pressure and temperature
transients are provided for 10 3 seconds. These transients demonstrate similar characteristics to the
large DEPSG break transients discussed above and, since
the performance of the containment heat removal systems
should be similar, long-term cooling is ensured.
6.2-13 Rev. 0 WOLF CREEK The sump temperature transients for the DEPSG break with minimum safety injection and the DEPSG break with maximum safety injection are provided in Figures 6.2.1-17 and 6.2.1-18, respectively.
The energy removal rates for the containment fan coolers, the RHR heat exchangers, and the containment passive heat sinks for the DEPSG break with minimum safety injection and the DEPSG break with maximum safety injection as a function of time are shown in Figures 6.2.1-19 through
6.2.1-24.
The containment system energy inventory as a function of time is plotted for the DEPSG break with minimum safety injection and maximum safety injection in Figures 6.2.1-25 and 6.2.1-26, respectively. All mechanisms of energy removal from and transfer within the containment are addressed in these figures. Included are the vapor energy, sump energy, energy contained in heat sinks, total energy removed from the containment by fan coolers and by the residual heat removal system, and net energy transferred by sprays from the containment vapor to the sump.
For the DBA at the time of the calculated peak containment pressure, the vapor energy is 310.8 x 10 6 Btu, the energy deposited in the sump is 82.1 x 10 6 Btu, the containment passive heat sinks have absorbed 81.0 x 10 6 Btu, 2.0 x 10 6 Btu have been removed by the containment fan coolers, 3.6 x 10 6 Btu have been transferred from the containment vapor to the sump via the containment sprays, and no energy has been removed by the RHR system. Safety injection is switched to the recirculation mode at 1,509 seconds, and the containment sprays are switched to the recirculation mode at 3,227 seconds after the accident.
- n. Accumulator Nitrogen Release
Table 6.2.1-11 provides the nitrogen release rate from the accumulators following the discharge of their liquid volumes. The added mass and associated energy of this nitrogen release are accounted for in the LOCA analysis.
6.2-14 Rev. 29 WOLF CREEK
- o. Normal Containment Ventilation System Evaluation
The functional capability of the normal containment
ventilation systems to maintain the temperature, pressure, and humidity in the containment and
containment subcompartments is discussed in Sections
6.2.2.2 and 9.4.6.
- p. Post-Accident Monitoring
Instrumentation for post-accident monitoring is discussed
in Section 7.5.
6.2.1.2 Containment Subcompartments 6.2.1.2.1 Design Basis
Subcompartments within the containment, principally the reactor cavity, the steam generator loop compartments, and the pressurizer compartment, are
designed to withstand the transient differential pressures and jet impingement
forces of a postulated pipe break. Venting of these chambers maintains the
differential pressures within the structural limits. In addition, restraints
on the reactor coolant pipes, reactor vessel, steam generators, etc., are
designed so that neither pipe whip nor vessel upset forces threaten the
integrity of the subcompartments or of the containment structure.
Analysis of the pressure transients in the reactor cavity, steam generator
compartment, and pressurizer compartment has been performed to verify the adequacy of the structural design of these structures under accident conditions. The following is a synopsis of the pipe breaks analyzed:
- a. For the reactor cavity subcompartment analysis, the
design basis break is a double-ended cold leg guillotine
break within the reactor cavity. Pipe restraints
are employed to limit the break flow area to less than
150 square inches.
- b. For the steam generator loop compartments, the design
basis break is a steam generator inlet elbow longitudinal
split with a break flow area of 763 square inches, a
double-ended steam generator outlet nozzle break
restrained to a break flow area of 436 square
inches, and a double-ended reactor coolant pump outlet nozzle break restrained to a break flow area of 236 square inches.
6.2-15 Rev. 0 WOLF CREEK
- c. The pressurizer compartment is divided into two
compartments: 1) the pressurizer vault and 2) the
pressurizer surge line compartment.
The design basis break for these subcompartments is the
double-ended pressurizer surge line break. In addition
to this break, the pressurizer spray line break and the
three break cases from the steam generator loop
compartment analysis were considered in the selection of
the design analysis break. In all cases, the pressures
in the pressurizer compartment were substantially lower
than those resulting from the pressurizer surge line
break.
6.2.1.2.2 Design Features
All design features provided for alleviating pressure buildup within the
subcompartments are discussed in the subcompartment design evaluation in
Section 6.2.1.2.3. Reference 2 describes the design features which limit the
movement of the pipe after the postulated break.
6.2.1.2.3 Design Evaluation
- a. Mass and Energy Release Rate Transient Model
The computer programs used to develop the mass and energy
release transients for subcompartment pressurization
analyses are described in Reference 3. Tables 6.2.1-12
through 6.2.1-16 provide tabulations of the mass and energy release rates versus time for the spectrum of breaks analyzed.
- b. Subcompartment Pressure Analyses Model
The COPDA computer code (Ref. 4) employs a finite
difference technique to solve the time dependent
equations for the conservation of mass, energy, and
momentum to perform the subcompartment analyses. This
code and the assumptions inherent to it are described
fully in Reference 5.
- 1. Reactor Cavity Rupture Analysis
The design break is a double-ended cold leg break
that is postulated to occur at the reactor vessel
nozzle to cold leg weld. The reactor cavity design
and the reactor coolant loop design are such that the
6.2-16 Rev. 0 WOLF CREEK analyses provided in Reference 2 verify that no break
need be considered in the reactor cavity pipe
penetrations (see Section 3.6). Pipe movement, and
thus break flow area, is mechanically restrained to limit the effective break size to less than 150
square inches. The mass and energy release data used
in the pressure response analysis are presented in
Table 6.2.1-12.
The reactor cavity nodalization model is shown in
Figures 6.2.1-27 through 6.2.1-33. As may be seen in
the figures, node boundaries were chosen wherever
significant restrictions to flow occurred. Thus, node boundaries were selected at the reactor coolant loop nozzles, the neutron detector wells, the shield support platform at the reactor vessel flange, the
nozzle support ledge, and the incore instrumentation
support platform. Since all restrictions to flow
were considered, further nodalization was neither
necessary nor appropriate.
The model included water-filled shield bags located at the reactor vessel flange. The purpose of the bags was to provide shielding at the top of the reactor cavity annulus and yet not prevent the venting of mass and energy from the cavity to the containment in the event of a LOCA. These bags have been replaced with a Permanent Reactor Cavity
Seal Ring. (The considered effect of bags breaking, as described below, is for analysis purposes only.)
The bag material and bag life are based on expected
containment environmental conditions and dose rates.
The shielding features are discussed in Section 12.3.2.2.1. The neutron shield water bags were designed to completely cover the reactor cavity
opening with a minimum water depth of 12 inches.
For the original design, bags were designed to
rupture when pressure was exerted on them sufficient
to lift them from their support platform. For this
analysis the bags were assumed to remain intact and
rigid until they reached a height of 1.0 foot above
the support platform before rupturing. The top
boundaries of the nodes below the shield bags were
then assumed to occur at the bottom surface of the
6.2-17 Rev. 12 WOLF CREEK bags. Thus, the volumes, vent areas, and 1/a's, which are the inertial terms in the momentum equation
associated with these compartments were treated as
dynamic parameters in the COPDA code, varying with the changing height of the bags. Volumes, vent
areas, flow coefficients, and 1/a's were calculated
to obtain a conservative estimate of pressures near
the break.
In the analysis, the insulation on the reactor cavity
wall was conservatively assumed to remain intact and
to be pressed flush against the wall. The insulation
below the shield support platform is constructed of a
material that, on impact, will break apart into small crushable pieces and a powder. This insulation does not present any obstruction to flow in the event of a
LOCA, nor will the pieces formed act as missiles. The
insulation around the reactor coolant piping was
assumed to block any possible venting through the
piping penetrations; thus no credit was assumed for
these vent areas. In the lower region of the reactor
cavity, venting through the incore instrumentation
support platform was assumed to be blocked by
insulation. The design of the reactor cavity
precludes blockage by insulation within the reactor
cavity annulus. Tables 6.2.1-17 and 6.2.1-19 show
the volume of each subcompartment as well as the
initial conditions prior to the postulated accident.
The vent areas, 1/a's, and head loss coefficients used in the calculation of the flow coefficients are given in Tables 6.2.1-18 and 6.2.1-20 for all of the
flowpaths between subcompartments.
The homogeneous frozen flow option was employed in
the analysis. This flow option is described in
Reference 5. The resulting peak pressure for each
subcompartment is listed in Table 6.2.1-17. The
complete pressure histories for all of the
subcompartments are shown in Figures 6.2.1-34 through
6.2.1-39.
The subcompartment pressures, when applied to the
projected areas of the subcompartments on the reactor
vessel, yield the force on the vessel. Time-
histories of the horizontal and vertical forces and
the upending moment imposed on the reactor vessel by
the asymmetric pressurization of the reactor cavity
are presented in Figures 6.2.1-40 through 6.2.1-42.
The force and moment coefficients for each
subcompartment are given in Table 6.2.1-21.
6.2-18 Rev. 1 WOLF CREEK
- 2. Steam Generator Loop Compartments
The steam generator loop compartment is subjected to
double-ended breaks of the pump suction line, the cold leg, the hot leg, a longitudinal split of the
hot leg, and double-ended branch line breaks. All
double-ended breaks are mechanically restrained so
that the largest breaks in the hot leg, cold leg, and
pump suction are 763 in 2 , 236 in 2 , and 436 in 2 , respectively. These three breaks envelope all
postulated breaks within the steam generator loop
compartment. These breaks were analyzed, using the
same 59-node model, to determine the maximum
pressures on the walls of the compartment and on the enclosed equipment, i.e., the steam generator, the reactor coolant pump, and the pressurizer. The
blowdown data for the three breaks are given in
Tables 6.2.1-13 through 6.2.1-15. The nodalization
model for the analyses is given in Figures 6.2.1-43
through 6.2.1-55. Only breaks in loop 4 were
analyzed, since this loop has the smallest vent area
directly to the remainder of the containment due to
the presence of the pressurizer, and thus results in
the highest pressures.
To ensure conservative design of the loop compartment
walls and the equipment supports, the loads
calculated for loop 4 were applied to the other three
steam generator loop compartments by appropriate translation and rotation of the force vector axes.
The volumes of the subcompartments, as well as the
initial conditions prior to the transient, are given
in Table 6.2.1-22.
As with the reactor cavity analysis, the node
boundaries were selected wherever significant
restrictions to flow occurred. A sensitivity study
was performed in which the number of nodes in the
steam generator compartment was varied. The
resulting forces on the compartment walls and on the
equipment in all cases were less than the forces
calculated with the 59-node model. Therefore, it was
assumed that the nodalization employed in the
original model was both adequate and conservative.
All major obstructions, such as columns, pumps, tanks, grating, and the steam generators, were
considered in the calculation of the subcompartment
volumes and vent areas. In addition, the values for
6.2-19 Rev. 20 WOLF CREEK volume were reduced by 5 percent to allow for minor
obstructions, such as cable trays, supports, and
various piping. The principal obstructions within
the steam generator loop compartments were the reactor coolant pumps and the steam generators. Flow
through the reactor cavity was neglected. The flow
coefficients associated with the flow paths were
calculated in the same manner as for the reactor
cavity. The head loss coefficients used in the
calculation of the flow coefficients, as well as the
vent areas and 1/a's for each flowpath, are listed in
Table 6.2.1-23.
The fluid flow from one subcompartment to another was calculated, using the homogeneous frozen flow option in the analysis. The peak pressures for each
subcompartment are listed in Table 6.2.1-22. The
complete pressure histories for those subcompartments
near the break for each of the three break cases
analyzed are shown in Figures 6.2.1-56, 6.2.1-57, 6.2.1-61, and 6.2.1-69. When the subcompartment
pressures were applied to their projected areas on
the steam generator and the reactor coolant pump, the
forces were determined on these pieces of equipment.
The forces on the reactor coolant pump and the steam
generator are shown in Figures 6.2.1-58, 6.2.1-59, 6.2.1-62 through 6.2.1-67, and 6.2.1-70 through
6.2.1-74. The coefficients used to calculate the
forces are given in Tables 6.2.1-24 and 6.2.1-25.
The component and resultant forces on the steam
generator and reactor coolant pump for the three
breaks analyzed are illustrated in Figures 6.2.1-60, 6.2.1-68, and 6.2.1-75.
- 3. Pressurizer Vault
The pressurizer vault is subjected to a pressurizer
spray line break, a pressurizer surge line break, and
a reactor coolant loop break. The pressurizer surge
line compartment located directly below the
pressurizer vault is subject to a pressurizer surge
line break and reactor coolant pipe break within the
steam generator compartment adjacent to the
pressurizer vault. Analyses showed that the worst
postulated break for both the pressurizer vault and
the surge line compartment was the double-ended
pressurizer surge line break. The mass and energy
release data for this case are given in
Table 6.2.1-16.
6.2-20 Rev. 0 WOLF CREEK In the model, the pressure is relieved through large
vents in the top of the pressurizer vault, and through
the surge line compartment, out into the steam generator
loop compartment and then up to the remainder of the containment. Figure 6.2.1-76 provides a simplified
elevation view of the pressurizer vault, and Figure
6.2.1-77 shows a schematic diagram of the flow model.
The subcompartment volumes along with the peak calculated
pressures and the design pressures in the pressurizer
vault and the surge line compartment are given in Table
6.2.1-26. The pressure histories of those
subcompartments directly below the pressurizer are given
in Figure 6.2.1-78. Table 6.2.1-27 summarizes the head loss coefficients used to calculate the flow coefficients and the vent areas and 1/a's for all of the flow paths.
- c. Nodalization Model Adequacy
The determination of nodalization models used for the
subcompartment analysis is adequate and based on the
following criteria:
- 1. The models are physically representative of the
geometry investigated.
- 2. The models are of adequate detail to consider all
significant obstructions and flow losses.
- 3. The selection of nodal boundaries and volumes reflect the conservative theoretical thermo and fluid dynamic
application.
A determination that these criteria are met is based on
previously performed developmental subcompartment
analysis, Bechtel experience in the performance of other
PWR subcompartment analyses, and comparisons with
information in the public domain (such as NUREG/CR-1199, and NUREG-0609).
6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents The containment system receives mass and energy releases following a postulated
rupture of the reactor coolant system (RCS). These releases continue through
blowdown and post-blowdown. The release rates are calculated for pipe failure at three distinct locations: 1) hot leg, 2) pump suction, and 3) cold legs.
6.2-21 Rev. 0 WOLF CREEK Because of the pressure in the RCS before the postulated rupture, the mass and
energy flows rapidly from the RCS to the containment. As the water exits from
the rupture, a portion of it flashes into steam due to the pressure and
temperature in the containment, compared to the pressure and temperature of the RCS. The blowdown reduces the pressure in the RCS.
During the reflood phase, these breaks have the following different
characteristics. For a cold leg pipe break, all of the fluid which leaves the
core must vent through a steam generator and becomes superheated. However, relative to breaks at other locations, the core flooding rate (and, therefore, the rate of fluid leaving the core) for cold leg breaks is low because all the
core vent paths include the resistance of the reactor coolant pump. For a hot
leg pipe break, the vent path resistance is relatively low, which results in a
high core flooding rate, but the majority of the fluid which exits the core bypasses the steam generators in venting to the containment. The pump suction break combines the effects of the relatively high core flooding rate, as in the
hot leg break, and steam generator heat addition, as in the cold leg break. As
a result, the pump suction break yields the highest energy flow rates during
the post-blowdown period. The spectrum of breaks analyzed includes the largest
cold and hot leg breaks, reactor inlet and outlet, respectively, and a range of
pump suction breaks from the largest to a 3.0 ft 2 break.
Because of the phenomena of reflood, as discussed above, the pump suction break
location is the limiting case, with the double-ended pump suction break being
the most limiting. This conclusion is supported by studies of smaller hot leg
breaks which have been shown on similar plants to be less severe than the
double-ended hot leg. Cold leg breaks, however, are lower both in the blowdown
peak and in the reflood pressure rise. Thus, an analysis of smaller pump
suction breaks is representative of the spectrum of break sizes.
The LOCA analysis calculational model is typically divided into three phases, which are: 1) blowdown, which includes the period from accident occurrence (when the reactor is at steady state full power operation) to the time when
zero break flow is first calculated, 2) refill, which is from the end of
blowdown to the time the emergency core cooling system (ECCS) fills the vessel
lower plenum, and 3) reflood, which begins when water starts
6.2-22 Rev. 0 WOLF CREEK moving into the core and continues until the end of the transient. For the
pump suction break, consideration is given to a possible fourth phase; that is, froth boiling in the steam generator tubes after the core has been quenched.
For a description of the calculational model used for the mass and energy release analysis, see Reference 3.
6.2.1.3.1 Mass and Energy Release Data
- a. Blowdown Mass and Energy Release Data
Tables 6.2.1-28 through 6.2.1-32 present the calculated
mass and energy releases for the blowdown phase of the
various breaks analyzed with the corresponding break
size.
- b. Reflood Mass and Energy Release Data
The lower vessel plenum is assumed to refill immediately
following blowdown, hence the refill phase is skipped.
Tables 6.2.1-33 through 6.2.1-38 present the calculated
mass and energy releases for the reflood phase of the
various breaks analyzed along with the corresponding
safeguards assumption (maximum or minimum).
c Dry Steam Post-Reflood Mass and Energy Release Data
The calculated mass and energy releases for the post-
reflood phase with dry steam are provided in the reflood
mass and energy release tables (Tables 6.2.1-35 through 6.2.1-38) after the end of the 10-foot entrainment occurs. These tables correspond to the hot leg, cold
leg, and small pump suction breaks analyzed.
- d. Two-Phase Post-Reflood Mass and Energy Release Data
Tables 6.2.1-39 and 6.2.1-40 present the two-phase
(froth) mass and energy release data for a double-ended
pump suction break, using minimum and maximum safeguards
assumptions, respectively. The following procedure was
followed to account for the depressurization, equilibration, and decay heat mass and energy releases to
106 seconds.
6.2-23 Rev. 0 WOLF CREEK
- 1. Depressurization Energy Release
The froth mass and energy release was initially
tabulated based on a reference temperature for heat stored in the steam generator metal and secondary
fluid of saturation at the containment design
backpressure of 60.0 psig. Additional two-phase mass
and energy releases become available due to the
energy within the steam generators, as the
containment depressurizes to atmospheric conditions.
- 2. Depressurization (Two-Phase Mixture)
Tables 6.2.1-41 and 6.2.1-42 show the available depressurization energy of the steam generators above atmospheric pressure (14.7 psia).
This energy is brought out in two stages. In the
first, the sources above are brought into equilibrium
with the actual containment pressure. The rate for
this phase is set by the froth calculation models.
In the second, the sources give up additional energy
as the containment pressure decreases. The rate for
this stage is set by the containment depressurization
rate.
The depressurization mass and energy release rates
can be determined if the depressurization time is
known. The depressurization time was estimated by choosing a conservatively low value that would maximize the depressurization mass and energy release
rates to the containment (3,600 seconds for normal
dry containment). First, a containment pressure
calculation was performed, neglecting the
depressurization energy release. For this case, the
containment will depressurize faster and, hence, a
conservative depressurization time is calculated.
The second containment pressure calculation is made, utilizing depressurization time with the procedure
for calculating depressurization mass and energy
release rates described in this section.
The steam generator depressurization mass and energy
release rate from the broken and intact loops were
calculated and added to the initial mass and energy
releases, which were based on a containment back
pressure of 60.0 psig, described below.
6.2-24 Rev. 0 WOLF CREEK
- 3. Broken Loop Steam Generator - Equilibration Stage
The amount of energy in the steam generator is
directly proportional to pressure and, hence, the fraction to be brought out equals the difference
between reference pressure (60.0 psig) and the actual
containment pressure divided by reference pressure.
Since the broken loop steam generator is in
equilibrium with the reference pressure of 60.0 psig
prior to the beginning of froth, a conservative value
for steam generator heat release was assumed. A rate
of 100,000 Btu/sec would release all of the available
energy in 209 seconds. This value is conservative.
- 4. Broken Loop Steam Generator - Depressurization Stage
The amount of energy to be brought out is the
original amount of energy remaining in the broken
loop steam generator given in Tables 6.2.1-41 and
6.2.1-42, less what is brought out to reach
equilibrium. The heat addition rate is this amount
divided by the assumed depressurization time. The
mass boiloff rate is this rate divided by latent
heat. The energy addition rate is the boiloff rate
times saturated vapor enthalpy.
- 5. Intact Loop Steam Generator - Equilibration Stage
The same procedure as for the broken loop is used here. However, metal and core energy is lumped with the steam generator energy for this calculation. The
fraction to be brought out to attain equilibrium
equals the difference between the reference
containment pressure and the actual containment
pressure divided by the reference value. The rate of
addition to the containment is 90.0 lb/sec at 1,034
seconds. This cools the steam generator and metal at
37,576 Btu/sec. Thus, the duration of the extension
of the post-reflood table is the fraction times the
available energy divided by the rate of cooling.
This was not extended beyond recirculation because
the continued condensation effect is implicit in
these numbers and should change after recirculation.
6.2-25 Rev. 0 WOLF CREEK
- 6. Intact Loop Steam Generator - Depressurization Stage
Again the procedure used here is the same as the
broken loop case except that the decay heat should be added to the heat addition rate, which was not
included in the initial post-reflood tables. The
amount of energy to be brought out is the original
energy remaining given in Tables 6.2.1-41 and 6.2.1-
42, less what is brought out to reach equilibrium.
The heat addition rate is this amount divided by the
depressurization time. The mass boiloff rate is this
rate divided by latent heat. The energy addition
rate is the boiloff rate times saturated vapor
enthalpy. Beyond the equilibration stage, the mass boiloff rate due to decay heat is added to the depressurization mass boiloff rates. This rate is
the decay heat rate divided by latent heat, including
ECCS water subcooling prior to recirculation. The
corresponding energy addition rate is the boiloff
rate times saturated vapor enthalpy.
The continued condensation benefit is not implicit in
these numbers, and thus this calculation may extend
beyond recirculation.
- 7. Decay Heat
Figure 6.2.1-16 presents the decay heat which is used
for the depressurization calculation.
- 8. Post-Recirculation Energy Release
Recirculation for the maximum safety injection case occurs at 849 seconds, which is during the broken loop depressurization stage. During this stage, the
energy release is a function of the recirculated
safety injection water sensible heat, the reactor
decay heat (Figure 6.2.1-16), the broken loop depressurization heat release, and the actual
containment pressure. The intact loop equilibration
and depressurization releases are accounted for in
the same way.
6.2-26 Rev. 6 WOLF CREEK Recirculation for the minimum safety injection case
occurs at 1,509 seconds, which is during the intact
loop depressurization stage. During this stage, the
energy release is a function of the recirculated safety injection water sensible heat, the reactor
decay heat, the broken and intact loop
depressurization heat releases, and the actual
containment pressure.
Following the end of broken and intact loop
depressurization to the end of the transient, the
energy release is a function of decay heat and
sensible heat only. End of depressurization occurs
at 3,772 and 3,775 seconds for the maximum and minimum safety injection cases, respectively.
6.2.1.3.2 Energy Sources
The sources of energy considered in the LOCA mass and energy release analysis
are given in the energy balance tables (Tables 6.2.1-43 through 6.2.1-48).
These energy sources are:
- a. RCS, accumulators, and pumped safety injection sensible
heat
- b. Decay heat
- c. Core stored energy
- d. Thick and thin metal energy
- e. Steam generator energy
The energy balance tables show the initial energy distribution and the energy
distribution at end-of-blowdown (EOB), end-of-entrainment (EOE), end-of-froth (EOF), and end-of-froth intact loops (EOFIL) for the two-phase post-reflood
analyses. For the dry steam post-reflood analyses, the energy distribution at
an assumed recirculation time of 1,500 seconds is given instead of EOF and
EOFIL.
The methods and assumptions used to release the various energy sources are
given in Reference 3.
The following items ensure that the core energy release is conservatively
analyzed for maximum containment pressure.
6.2-27 Rev. 0 WOLF CREEK
- a. Core power level of 3,636 MWt (102 percent of ultimate
core power level)
- b. Allowance in temperature for instrument error and dead band (+4 F)
- c. Margin in volume (1.4 percent)
- d. Allowance in volume for thermal expansion (1.6 percent)
- e. Margin in core power associated with use of engineered
safeguards design rating (ESDR)
- f. Allowance for calorimetric error (2 percent of ESDR)
- g. Conservatively modified coefficients of heat transfer
- h. Allowance in core-stored energy for effect of fuel
densification
- i. Margin in core-stored energy (+15 percent)
- j. Maximum calculated operating temperature (627.3 F) with
above assumptions
6.2.1.3.3 Description of Blowdown Model
A description of the model used to determine the mass and energy released from
the RCS during the blowdown phase of a postulated LOCA is provided in Reference
- 3. All significant correlations are discussed.
6.2.1.3.4 Description of Core Reflood Model
A description of the model used to determine the mass and energy released from
the RCS during the reflood phase of a postulated LOCA is provided in Reference
- 3. All significant correlations are discussed. Transients of the principal
parameters during reflood are given in Tables 6.2.1-49 and 6.2.1-50 for the
limiting case pump suction breaks with maximum and minimum safeguards.
6.2.1.3.5 Description of Long-Term Cooling Model
The calculational procedure used to determine the mass and energy released
during the post-reflood phase of a postulated LOCA is described in Reference 3.
6.2-28 Rev. 0 WOLF CREEK 6.2.1.3.6 Single Failure Analysis
The effect of single failures of various ECCS components on the mass and energy
releases is included in these data. The two analyses for the DEPSG breaks bound this effect.
No single failure is assumed in determining the mass and energy releases for
the maximum safeguards case. For the minimum safeguards case, the single
failure assumed is the loss of one emergency diesel. This failure results in
the loss of one pumped safety injection train. The analysis of both maximum
and minimum safeguards cases ensures that the effect of all credible single
failures is bounded.
6.2.1.3.7 Metal-Water Reaction In the mass and energy release data presented here, no Zr-H 2 O reaction heat was considered because the clad temperature did not rise high enough for the rate
of the Zr-H 2 O reaction to be of any significance.
6.2.1.3.8 Reactor Coolant System Mass and Energy Balance
Reactor coolant system mass and energy balances are tabulated for hot leg, cold
leg, and pump suction breaks in Tables 6.2.1-43 through 6.2.1-48.
6.2.1.3.9 Additional Information Required for Confirmatory
Analysis
System parameters and hydraulic characteristics needed to perform confirmatory
analysis are provided in Tables 6.2.1-51 through 6.2.1-55.
6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary Pipe Ruptures Inside Containment Steam line ruptures occurring inside a reactor containment structure may result
in significant releases of high energy fluid to the containment environment, possibly resulting in high containment temperatures and pressures. The high pressures and temperatures can result in failure of any equipment which is not qualified to perform its function in an adverse environment. This could degrade the effectiveness of the protection system in mitigating the consequences of the steamline rupture. In addition, the containment structure is designed to withstand a limited internal pressure (i.e., 60 psig). Thus, an associated containment response analysis may be performed to demonstrate that the conditions inside the containment during a steamline rupture do not violate the existing environmental qualification (EQ) envelopes, and to demonstrte that the containment design pressure is not exceeded.
Consistent with the NRC-approved methodology documented in Reference 6, the analyses are performed to maximize the amount of mass and energy released to the containment. The releases following a steamline rupture are dependent upon many possible configurations of the plant steam system and containment designs, as well as the plant operating conditions and the size of the rupture. There are competing effects as power and break size change, and thus multiple cases are typically analyzed. Therefore, the steamline break event is analyzed for a spectrum of pipe break sizes and various plant conditions from hot standby to 102% of full power (i.e., re-rated power of 3579 MWt). Break sizes are considered beginning with the full double-ended break and decreasing in area until no water entrainment is calculated to occur. The spectrum of powers and breaks analyzed is listed in Table 6.2.1-56.
6.2-29 Rev. 22 WOLF CREEK 6.2.1.4.1 Significant Parameters Affecting Steam Line Break Mass and Energy Releases
There are four major factors that influence the release of mass and energy
following a steam line break: steam generator fluid inventory, primary to
secondary heat transfer, protective system operation, and the state of the secondary fluid blowdown. The following is a list of those plant variables
which determine the influence of each of these factors:
- a. Plant power level
- b. Main feedwater system design
- c. Auxiliary feedwater system design
- d. Postulated break type, size, and location
- e. Availability of offsite power
- f. Safety system failures
- g. SG reverse heat transfer and reactor coolant system metal
heat capacity
- h. Steam Generator Fluid Mass
- i. MSIV and MFIV Closure Times
- j. Safety Injectin System
- k. Protection System Actuations
- l. Operator Response Time
6.2-30 Rev. 22 WOLF CREEK The following is a discussion of each of these variables.
6.2.1.4.1.1 Plant Power Level
Steam line breaks can be postulated to occur with the plant in any operating condition ranging from hot standby to full power. Since steam generator mass
decreases with increasing power level, breaks occurring at lower power
generally result in a greater total mass release to the plant containment.
However, because of increased energy storage in the primary plant, increased
heat transfer in the steam generators, and the additional energy generation in
the nuclear fuel, the energy release to the containment from breaks postulated
to occur during power operation may be greater than for breaks occurring with
the plant in a hot standby condition. Additionally, steam pressure and the
dynamic conditions in the steam generators change with increasing power and
have significant influence on both the rate of blowdown and the amount of moisture entrained in the fluid leaving the break following a steambreak event.
Because of the opposing effects of changing power level on steam line break
releases, no single power level can be singled out as a worst case initial
condition for a steam line break event. Therefore, several different power
levels spanning the operating range as well as the hot standby condition have
been analyzed.
6.2.1.4.1.2 Main Feedwater System Design The rapid depressurization which occurs following a rupture may result in large
amounts of water being added to the steam generators through the main feedwater
system. Rapid closing isolation valves are provided in the main feedwater
lines to limit this effect. Also, the piping layout downstream of the isolation valves affects the volume in the feedwater lines that cannot be
isolated from the steam generators. As the steam generator pressure decreases, some of the fluid in this volume will flash into the steam generator, providing
additional secondary fluid which may exit out the rupture.
The feedwater addition which occurs prior to closing of the feedwater line
isolation valves influences the steam generator blowdown in several ways.
First, the rapid addition increases the amount of entrained water in large-
break cases by lowering the bulk quality of the steam generator inventory.
Secondly, because the water entering the steam generator is subcooled, it lowers the steam pressure, thereby reducing the flow rate out of the break.
6.2-31 Rev. 22 WOLF CREEK Finally, the increased flow rate causes an increase in the heat transfer rate from the primary to secondary system, resulting in greater energy being
released out the break. Since these are competing effects on the total mass
and energy release, no "worst case" feedwater transient can be defined for all
plant conditions. In the results presented, the worst effects of each variable
have been used. For example, moisture entrainment for each break is
calculated assuming conservatively small feedwater additions so that the
entrained water is minimized. Determination of total steam generator
inventory, however, is based on conservatively large feedwater additions.
The effects of any flashing of the feedwater trapped between the steam
generator and the isolation valves is included in the analyses. The failure of
the MFIV on the faulted loop results in additional fluid being added to the
faulted steam generator. The quantity of the additional fluid to be released
is based on the volume between the isolation valve and the main feedwater
control valve (MFCV) on the faulted loop. Thus, the mass added to the faulted
steam generator from both the pumped main feedwater flow and the feedline
flashing will be larger with a failure of a feedwater isolation valve.
6.2.1.4.1.3 Auxiliary Feedwater System Design
Within the first minute following a steam line break, the auxiliary feed system
is initiated on any one of several protection system signals. Addition of
auxiliary feedwater to the steam generators increases the secondary mass
available for release to the containment, as well as increases the heat
transferred to the secondary fluid. The effects on steam generator mass are
maximized by assuming auxiliary feed flow to the faulted steam generator
starting from the time a safety injection signal is initiated on low steamline
pressure or high containment pressure and continuing until manually stopped by
the plant operator.
The maximum auxiliary feedwater flow delivered to the faulted steam generator
represents the most limiting single failure from the perspective of mass and
energy releases following a postulated steamline break. Failure of one
protection train is assumed so that only one motor and the turbine driven AFW
pump are operating during the transient. To maximize the AFW flowrate to the
ruptured steam generator, the control valve on the discharge side of the
operating motor driven AFW pump feeding the faulted generator is assumed to
fail in the wide open position.
6.2.1.4.1.4 Postulated Break Type, Size, and Location
- a. Postulated Break Type
Two types of postulated pipe ruptures are considered in evaluating steam line breaks.
First is a split rupture in which a hole opens at some point on the side of the steam pipe or steam header but does not result in a complete severance of the pipe. A single, distinct break area is fed uniformly by all steam generators until steam line isolation occurs. The blowdown flow rates from the individual steam generators are interdependent, since fluid coupling exists between all steam lines. Because flow limiting orifices are provided in each steam generator, the largest possible split rupture can have an effective area prior to isolation that is no greater than the throat area of the flow restrictor times the number of plant primary coolant loops. Following isolation, the effective break area for the steam generator with the broken line can be no greater than the flow restrictor throat area.
6.2-32 Rev. 29 WOLF CREEK The second break type is the double-ended guillotine rupture in which the steam pipe is completely severed and
the ends of the break displace from each other.
Guillotine ruptures are characterized by two distinct
break locations, each of equal area but being fed by different steam generators. The largest possible
guillotine rupture can have an effective area per steam
generator no greater than the throat area of one
steamline flow restrictor.
The type of break influences the mass and energy releases
to containment by altering both the nature of the steam
blowdown from the piping in the steam plant and the
effective break area fed by each steam generator prior to
steam line isolation. For example, a double-ended rupture in a pipe having a cross-sectional area of 2.4 square feet would appear as a 1.4-square-foot rupture to
a single steam generator feeding one end of the break, but would appear as a 0.8-square-foot rupture to each of
the steam generators feeding the other end of the break.
- b. Postulated size
Break area is also important when evaluating steam line
breaks. It controls the rate of releases to the
containment as well as exerts significant influence on
the steam pressure decay and the amount of entrained
water in the blowdown flow. The data presented in this
section include releases for three break areas at each of
five initial power levels. Included are two double-ended and one split rupture, as follows:
- 1. A full double-ended pipe rupture downstream of the
steam line flow restrictor. For this case, the
actual break area equals the cross-sectional area of
the steam line, but the blowdown from the steam
generator with the broken line is controlled by the
flow restrictor throat area (1.4 square feet). The
reverse flow from the intact steam generators is
controlled by the smaller of the pipe cross section, the steam stop valve seat area, or the total flow
restrictor throat area in the intact loops. The
reverse flow has been conservatively assumed to be
controlled by the flow restrictors in each of the
intact loop steam generators. Actually, the combined
flow from the three steam generators must pass
through an 18-inch (1.42 square feet) line, which
would greatly restrict the flow.
- 2. A small double-ended rupture having an area just
larger than the area at which water entrainment
ceases. Entrainment is assumed in the forward
direction only. Dry steam blowdown is assumed to
occur in the reverse direction.
6.2-33 Rev. 7 WOLF CREEK
- 3. A split break that represents the largest break which
neither generates a steam line isolation signal from
the primary protection equipment nor results in
moisture entrainment. Steam and feedwater line isolation signals are generated by high containment
pressure signals for these cases. Being a split
rupture, the effective area seen by the faulted steam
generator will increase by a factor of 4, following
steam line isolation. Conceivably, moisture
entrainment could occur at that time. However, since
steam line isolation for these breaks generally does
not occur before 20-60 seconds, it is conservatively
assumed that the pressure has decreased sufficiently
in the affected steam generator to preclude any moisture carryover.
- 4. A break representing the largest double-ended rupture
for which only dry steam blowdown occurs need not be
presented. Studies (Ref. 7) have shown that this
break size is typically smaller than the largest
split break (no entrainment) for which blowdown for
the split rupture will be more severe than the no-
entrainment DER at any given power level.
- c. Postulated Break Location
Break location affects steam line blowdowns by virtue of
the pressure losses which would occur in the length of
piping between the steam generator and the break. The effect of the pressure loss is to reduce the effective break area seen by the steam generator. Although this
would reduce the rate of blowdown, it would not
significantly change the total release of energy to the
containment. Therefore, piping loss effects have been
conservatively ignored in all blowdown results, except in
the small double-ended ruptures in which moisture
entrainment occurs. The effects of pipe friction are
conservatively assumed to be sufficiently large in this
case to prevent moisture entrainment in the reverse flow, thus minimizing water relief to the containment.
6.2.1.4.1.5 Availability of Offsite Power
Loss of offsite power following a steamline rupture would result in tripping of the RCPs, main feedwater pumps, and a possible delay of AFW initiation due to emergency diesel generator starting delays. Each of these occurrences aids in mitigating the effects of the steamline break releases by either reducing the fluid inventory available to feed the blowdown or reducing the energy transferred from the primary coolant system to the steam generators.
The effects of the assumption of the availability of offsite power has been
enveloped in the analysis. Loss of offsite power has been assumed where it
delays the actuation of the containment heat removal systems (i.e., containment
sprays and containment air
6.2-34 Rev. 22 WOLF CREEK coolers) due to the time required to start the emergency diesel generators.
Offsite power has been assumed to be available where it maximizes the mass and
energy released from the break due to 1) the continued operation of the reactor
coolant pumps which maximizes the energy transferred from the reactor coolant system to the steam generators and 2) continued operation of the feedwater
pumps and actuation of the auxiliary feedwater system which maximizes the steam
generator inventories available for release.
6.2.1.4.1.6 Safety System Failures
In addition to assuming a loss of offsite power, the following single active
failures were considered:
- a. Loss of one emergency diesel
- b. Failure of one main steam isolation valve
- c. Failure of one main feedwater isolation valve
The loss of one diesel results in the loss of one train of each of the
containment heat removal systems. As discussed in Section 6.2.1.4.3.3, this is
the most severe single active failure.
The effect of a main steam isolation valve failure is to provide additional
fluid which may be released to the containment via the break. This results
from the blowdown of all the steam piping between the break location and the
isolation valves in the intact loops. The effect of the failure of the MSIV and the associated bypass valve on the faulted loop is considered. It should be noted that closure of the faulted loop MSIV does not terminate the break flow from the faulted steam generator, since the limiting break is postulated to be located between the steam generator and the MSIV. However, the faulted loop MSIV and the associated bypass valve do isolate the break from the remainder of the steamline and the other steam generators. If the faulted loop MSIV and the associated bypass valve fail to close, blowdown from multiple steam generators is prevented by the closure of the corresponding MSIV for each intact steam generator. But failure of the MSIV and the associated bypass valve does increase the unisolable steamline volume containing steam which will be released to the containment.
The failure of a main feedwater isolation valve results in additional fluid
being released to the containment following a main steam line break. The
additional fluid to be released is the volume between the isolation valve and
the feedwater control valve.
6.2.1.4.1.7 Steam Generator Reverse Heat Transfer and Reactor
Coolant System Metal Heat Capacity
Once steam line isolation is complete, those steam generators in the intact
steam loops become sources of energy which can be transferred to the steam
generator with the broken line. This energy transfer occurs via the primary
coolant. As the primary plant cools, the temperature of the coolant flowing in
the steam generator tubes drops below the temperature of the secondary fluid in
the intact units, resulting in energy being returned to the primary coolant.
This energy is then available to be transferred to the steam generator with the broken steamline.
Similarly, the heat stored in the metal of the reactor coolant piping, the reactor vessel, and the reactor coolant pumps is transferred to the primary coolant as the plant cooldown progresses. This energy also is available to be transferred to the steam generator with the broken line.
6.2-35 Rev. 22 WOLF CREEK The effects of both the reactor coolant system metal and the reverse steam generator heat transfer are included in the results presented in this document.
6.2.1.4.1.8 Steam Generator Fluid Mass
A maximum initial steam generator mass in all the steam generators was used in
all of the analyzed cases. The use of a high initial steam generator mass
maximizes the steam generator inventory available for release to containment.
The initial mass has been calculated as the value corresponding to the
programmed level (i.e., 50% narrow-range span) plus 10% to account for the SG
water level uncertainties, plus 10% to account for mass uncertanties. For
split breaks, the mass in the unisolable steam line volume is also included in
the initial faulted loop SG mass.
6.2.1.4.1.9 MSIV and MFIV Closure Times
A MSIV/MFIV stroke time delay of 15 seconds was conservatively assumed in the
analyses of these steamline break events. Note: The actual analysis
assumption consists of a total delay of 17 seconds, which includes a 2-second
allowance for signal processing delays.
6.2.1.4.1.10 Safety Injection System
Minimum safety injection system (SIS) flowrates corresponding to the failure of
one SI system train are assumed in this analysis. A minimum SI flow is
conservative since the reduced boron addition maximizes a return to power
resulting from the RCS cooldown. The higher power generation increases heat
transfer to the secondary side, maximizing steam flow out of the break. The
delay time to achieve full SI flow is assumed to be 27 seconds for this
analysis with offsite power available.
6.2.1.4.1.11 Protection System Actuations
The protection systems available to mitigate the effects of a MSLB accident
inside containment include reactor trip, safety injection, steamline isolation, and feedwater isolation. The setpoints used are conservative values with
respect to the plant-specific values delineated in the Technical Specification
Bases.
For the full double-ended rupture MSLB at all power levels and certain small
double-ended ruptures at high power levels, the first protection system signal
is low steamline pressure (2-of-3 channels per loop, lead/lag compensated in
each channel) in any loop that initiates safety injection and steamline
isolation; the SI signal produces a reactor trip signal. Feedwater system
isolation and AFW actuation occur as a result of the SI signal.
For the split breaks at all power levels and certain small double-ended
ruptures at median to low power levels, the steamline break protection function
typically relies on the high containment pressure signals for reactor trip and feedline and steamline isolations. Specifically, a safety injection signal is generated on a hi-1 (6 psig) containment pressure signal, and a steamline isolation signal is generated on a hi-2 (20 psig) containment pressure signal.
The timing of these signals must be determined iteratively with the containment response analysis and then modeled in LOFTRAN using "manual" actuation input parameters.
6.2-36 Rev. 29 WOLF CREEK 6.2.1.4.1.12 Operator Response Time
As long as AFW is being delivered to the faulted steam generator, the steamline
break mass and energy release to containment will continue. Operator action is
credited to re-align the AFW system to terminate the flow to the faulted steam generator, while continuing to feed the intact steam generators. A 20 minutes
operator action time confirmed by simulator scenario measurements, is credited
in this analysis. Actual termination of auxiliary feedwater flow to the
affected steam generator due to operator action is expected to occur prior to
600 seconds (10 minutes), as discussed in USAR Section 10.4.9.
6.2.1.4.2 Description of Blowdown Model
The MSLB mass and energy releases have been performed, based upon the NRC-
approved methodology documented in Reference 6. The system transient that provides the break flows and enthalpies of the steam release through the steam line break has been analyzed with the LOFTRAN code (Reference 7). Blowdown
mass and energy releases determined include the effects of core power
generation, main and auxiliary feedwater additions, engineered safeguards
systems, reactor coolant system thick metal heat storage, and reverse steam
generator heat transfer. The specific plant design input which was assumed is
proved for each case in Table 6.2.1-57. Table 6.2.1-57A and 6.2.1-57B provide
the mass and energy release data for the cases which resulted in the highest
temperature and pressure, respectively.
6.2.1.4.3 Containment Response Analysis
The GOTHIC computer code (Reference 28) was used to determine the containment
responses following the postulated main steam line breaks. The following
assumptions were made to obtain these responses.
6.2.1.4.3.1 Initial Conditions
The initial containment conditions are the same as those used in the
containment response analysis for the postulated reactor coolant system pipe
ruptures (see Table 6.2.1-5).
6.2.1.4.3.2 Input Parameters and Assumptions
- 1. The mass and energy release data used to determine the containment response for the spectrum of steamline breaks are calculated using the
LOFTRAN code (Ref. 7), along with the assumptions and models described
in Section 6.2.1.4.1.
- 2. Loss of offsite power is assumed as it delays the actuation of the containment heat removal systems (i.e., containment sprays and containment air coolers) due to the time required to start the emergency diesel generators.
- 3. Loss of one emergency diesel generator, associated with the loss of offsite power, is assumed. As a result, only one train of the containment heat removal systems (i.e., containment sprays and containment air coolers) is operable.
- 4. The heat removal capacity of the containment fan coolers is degraded uniformly by 20% based on their actual performance capability determined by the fan cooler vendor, shown in Table 6.2.1-57C.
6.2-37 Rev. 23 WOLF CREEK 5. The heat removal capability of fan coolers is not credited until a total response time of at least 70 seconds has elapsed. This response time considered the time interval between the time of steamline break initiation/LOOP and the time full containment cooling system air and
safety grade cooling water flow is established. Purging and filling of
the voids that are expected to reside in the fan coolers and cooling
water pipe lines as a result of the drain down scenario associated with
LOOP is also accounted for.
- 6. The containment spray pump performance is assumed to be degraded by 5%.
This results in a reduction of the spray injection flowrate from the
calculated flowrate of 3086 gpm to 2931.7 gpm.
- 7. If the containment pressure reaches the containment Hi-3 pressure setpoint (30 psig, including uncertainty) before 27 seconds, full flow
spray is conservatively assumed to occur at 60 seconds, accounting for
time to attain operating speed and design flow of the containment spray
pump and fill up the spray lines. Note: The load sequencer applies
power to containment spray pumps at 27 seconds. Otherwise, the
containment spray injection starts 30 seconds after the containment
pressure reaches the actuation setpoint (i.e., containment Hi-3
pressure). The 30 seconds time delay accounts for the spray pump
startup and spray line filling.
- 8. The surface area for the liquid pool is assumed to the 0 ft 2 in order to neglect the heat transfer from the vapor region to liquid region.
6.2.1.4.3.3 Description of Analysis Methods
A simplified schematic of the Wolf Creek containment, along with the GOTHIC
containment model for the MSLB, is shown in Figure 6.2.1-79 and Figure 6.2.1-
80, respectively. The model is comprised of three volumes representing the
containment volume, the outside air and a separate volume representing the fan
cooler ducts. The containment (Volume 1) is modeled with a single lumped
parameter node. Two boundary conditions (1F and 2F) are used to represent the
sources of mass and energy from the break and the spray injection system, respectively. Flow paths connect the boundary conditions to the containment
volume. Fourteen heat sinks, a fan cooler component and a volumetric fan are
also shown.
The direct heat transfer coefficient set is used for the GOTHIC calculation, along with the diffusion layer model (DLM) mass transfer correlation, for all
of the internal heat sinks in the Wolf Creek containment MSLB evaluation model.
The DLM is used to calculate condensation mass transfer between the heat sinks
and the atmosphere. The DLM model is described in Reference 28 and the
qualification for use in containment design basis analyses are described in
Reference 30.
6.2.1.4.3.4 Containment Pressure-Temperature Results
The containment pressure and temperature response to a postulated MSLB has been
analyzed, based on the developed GOTHIC model, for the 16 cases. The peak
calculated containment pressure and temperature for each case is presented in
Table 6.2.1-58. The full double-ended MSLB at 25% power (Case 10) and the full double-ended MSLB at the 102% power (Case 1), are found to result in the
highest containment peak pressure and temperature, respectively. The sequence
of events following a postulated main steam line break is listed in Tables
6.2.1-59 and 6.2.1-60 for the worst pressure and temperature cases, respectively. Figures 6.2.1-81 and 6.2.1-82 show the calculated containment
pressure, vapor temperature, and sump water temperature for these two limiting
cases.
6.2-38 Rev. 29 WOLF CREEK As illustrated in Figure 6.2.1-79, case 10, full double-ended MSLB at 25%
power, results in a peak pressure of 52.85 psig. This case represents the peak calculated containment pressure for the spectrum of breaks analyzed. The condensing heat transfer coefficient versus time for this case is provided in
Figure 6.2.1-83.
It is important to note that the peak calculated pressure is coincident with
the termination of the auxiliary feedwater flow to the affected steam
generator, which was assumed to occur at 1,200 seconds (20 minutes). Actual
termination of auxiliary feedwater flow to the affected steam generator due to
operator action is expected to occur prior to 600 seconds (10 minutes), as
discussed in Section 10.4.9. In all cases, the peak calculated containment pressure demonstrates considerable margin below the containment design
pressure.
As illustrated in Figure 6.2.1-82, case 1, full double-ended rupture at 102-
percent power, results in a peak vapor temperature of 364.9°F. This case represents the peak calculated containment vapor temperature for the spectrum
of breaks analyzed. The condensing heat transfer coefficient versus time for
this case is provided in Figure 6.2.1-84.
For the spectrum of breaks analyzed, the calculated containment vapor
temperature for some cases exceeds the specified containment design temperature
of 320 F for a short period of time. The 320 F containment design temperature
is the design temperature for safety-related equipment and instrumentation
located within the containment and not the maximum temperature allowed for the
containment atmosphere vapor.
It is important to note that the original containment analysis using CONTEMPT-LT/28 showed the calculated peak containment temperature was 386.5 F. Since the re-calculated peak containment temperature is less than the CONTEMPT-LT/28 analysis result, which was utilized in the current analysis of record for the
equipment surface temperatures, no revised equipment surface temperature
analysis is necessary. The existing equipment surface temperatures described
below remain valid and the temperature profiles presented in USAR Figures
3.11(B)-7 and 3.11(B)-7A for the equipment environmental qualification remain
bounding.
Figure 6.2.1-85 provides plots of surface temperature versus time for various
representative materials within the containment. These curves are calculated
using a model based on the acceptable methodology for safety-related component
thermal analysis discussed in Appendix B of Reference 8, in conjunction with
CONTEMPT-LT/28 analysis for the case resulting in the highest material surface
temperatures. These figures clearly show that the actual equipment
temperatures, following a postulated secondary system break, are well below
their design temperatures and are, in fact, approximated more closely by the
containment vapor saturation temperature.
Cables located inside the containment are qualified to higher temperatures (340
to 385 F) than their surfaces are expected to experience as shown in Figure 7A
of the NUREG-0588 submittal. The calculated temperature for each type of cable
is below the qualification temperature; however, due to the low mass to surface
area ratios for cables, the calculated jacket/cable surface temperatures exceed
the containment vapor saturation temperature.
6.2-39 Rev. 29 WOLF CREEK 6.2.1.4.4 Results of Postulated Feedwater Line Breaks Inside
Containment
The main feedwater addition is generally below the steam generator water level;
therefore, main feedwater line break (MFLB) scenarios always commence with two-
phase blowdowns. The enthalpy of the blowdown is less than the enthalpy of
saturated steam at the secondary-side operating pressures. As a result, the
long-term integrated energy released following an MFLB is bounded by the long-
term integrated energy released following an MSLB. It is expected that MFLB
cases would not produce peak containment pressure or temperature conditions as
severe as MSLB cases; therefore, MFLB cases are not considered for long-term
containment pressure and temperature analyses.
6.2.1.4.5 Additional Information Required for Confirmatory
Analysis
No additional information is deemed necessary for the performance of
confirmatory analyses.
6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System
For PWR plants, there is a direct dependence of core flooding rate on
containment pressure following a design-basis loss-of-coolant accident; i.e.,
the core flooding rate will increase with increasing containment pressure. A
decrease in containment pressure tends to result in a decreased core inlet
flooding rate and an increased peak cladding temperature (PCT). Therefore, Appendix K to 10 CFR Part 50 requires that the containment pressure used to
evaluate the performance capability of a PWR ECCS does not exceed a pressure
calculated conservatively for that purpose. It further requires that the
calculation include the effects of operation of all installed pressure-reducing
systems and processes. Therefore, the operation of all ESF containment heat
removal systems operating at maximum heat removal capacity; i.e., with all
containment spray trains operating at maximum flow conditions and all emergency
fan cooler units operating, are assumed to insure a conservatively low
containment backpressure for the ECCS performance evaluation.
The containment backpressure used for the BELOCA ASTRUM Uncertainty Analysis is calculated using the methods and assumptions discussed in Section 15.6.5.3.
Input parameters, including the containment initial conditions, net free
containment volume, passive heat sink materials, thicknesses, and surface
areas, and starting time and number of containment cooling systems used in the
analysis, are described in the following paragraphs.
6.2.1.5.1 Mass and Energy Release Data
The Mass and Energy releases used in the minimum Containment Pressure calculation were generated in the WCOBRA/TRAC Reference Transient simulation.
This data is shown in Table 6.2.1-63 (discussed in Section 15.6.5.3.2). The Table 6.2.1-63 mass and energy releases are taken from the 'Reference Transient' case of Section 15.6.5.3.1, which did not include the fuel TCD modeling. The conservatively low containment backpressure from this COCO study is bounding since the core stored energy increase when explicitly modeling fuel TCD, which would tend to increase energy released through the break and hence increase containment pressure.
6.2-40 Rev. 29 WOLF CREEK 6.2.1.5.2 Initial Containment Internal Conditions The initial values used in the analysis are provided in Table 6.2.1-65.
These containment initial conditions are representatively low values
anticipated during normal full power operation.
6.2.1.5.3 Containment Volume
The volume used in the analysis was 2.7 x 10 6 ft 3 6.2.1.5.4 Active Heat Sinks
The containment spray system and containment air coolers operate to remove heat
from the containment.
Pertinent data for these systems which were used in the analysis are presented
in Table 6.2.1-65.
The sump temperature was not used in the analysis because the maximum peak
cladding temperature occurs prior to initiation of the recirculation phase for
the containment spray system. In addition, heat transfer between the sump
water and the containment vapor space was not considered in the analysis.
6.2.1.5.5 Steam-Water Mixing
Water spillage rates from the broken loop accumulator are determined as part of
the core reflooding calculation and are included in the containment code (COCO)
calculational model.
6.2.1.5.6 Passive Heat Sinks
The passive heat sinks used in the analysis, with their thermo-physical
properties, are given in Table 6.2.1-66. The passive heat sinks and
thermophysical properties were derived in compliance with Branch Technical
Position CSB 6-1, "Minimum Containment Pressure Model for PWR ECCS Performance
Evaluation."
6.2.1.5.7 Heat Transfer to Passive Heat Sinks
The inputs to the containment pressure calculation are skewed in order to obtain a conservative (low) pressure transient. For example, the Tagami correlation is increased by a factor of 5 to obtain the maximum condensing heat transfer coefficient at the end of blowdown. The condensing heat transfer coefficients used for heat transfer to the steel Containment structures are included in the Containment calculation model. The COCO calculated Containment Pressure and the WCOBRA/TRAC calculated Containment Pressure for the PCT/CWO limiting case are shown in Figure 6.2.1-86.
6.2-41 Rev. 29 WOLF CREEK 6.2.1.6 Tests and Inspections Refer to Sections 6.2.6 and 6.6
6.2.1.7 Instrumentation Requirements
Instrumentation is provided to actuate the engineered safety features and to
monitor the containment temperature, pressure, and sump level. Design details
and logic of the instrumentation are discussed in Sections 7.1, 7.2, 7.3, and 7.5.
6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS
The functional performance objective of the containment heat removal system, as
an engineered safety features system, is to reduce the containment temperature
and pressure following a LOCA or main steam line break (MSLB) accident by
removing thermal energy from the containment atmosphere. These cooling systems
also serve to limit offsite radiation levels by reducing the pressure
differential between the containment atmosphere and the external environment, thereby diminishing the driving force for the leakage of fission products from the containment to the environment. The containment heat removal systems
include the residual heat removal system discussed in Sections 5.4.7, 6.2.1, and 6.3, the containment spray system (CSS) discussed in Section 6.2.2.1, and
the containment cooling system discussed in Section 6.2.2.2.
6.2.2.1 Containment Spray System 6.2.2.1.1 Design Bases
6.2.2.1.1.1 Safety Design Bases
SAFETY DESIGN BASIS ONE - The CSS is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, or external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The CSS is designed to remain functional after a SSE or to perform its intended function following the postulated hazard of a pipe break (GDC-3 and 4).
SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC-38). SAFETY DESIGN BASIS FOUR - The active components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection of components at appropriate times specified in the ASME Boiler and Pressure Vessel Code, Section XI (GDC-39 and 40).
SAFETY DESIGN BASIS FIVE - The CSS is designed and fabricated to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The capability of isolating components or piping is provided so that the CSS safety function is not compromised. This includes isolation of components to deal with leakage or malfunctions (GDC-38).
SAFETY DESIGN BASIS SEVEN - The containment isolation valves in the system are selected, tested, and located in accordance with the requirements of GDC-54 and 56 and 10 CFR 50, Appendix J, Type A testing.
6.2-42 Rev. 22 WOLF CREEK SAFETY DESIGN BASIS EIGHT - The CSS, in conjunction with the containment fan cooler system and the emergency core cooling system, is designed to be capable
of removing sufficient heat and subsequent decay heat from the containment
atmosphere following the hypothesized LOCA or MSLB to maintain the containment
pressure below the containment design pressure. Section 6.2.1 provides the assumptions as to sources and amounts of energy considered and the analysis of
the containment pressure transient following a LOCA or MSLB accident inside the
containment (GDC-38).
SAFETY DESIGN BASIS NINE - The CSS remains operable in the accident environment.
SAFETY DESIGN BASIS TEN - The containment spray water does not contain
substances which would be unstable in the thermal or radiolytic environment of
the LOCA or cause extensive corrosive attack on equipment.
SAFETY DESIGN BASIS ELEVEN - The CSS is designed so that adequate net positive
suction head (NPSH) exists at the suction of the containment spray pumps during
all operating phases, in accordance with Regulatory Guide 1.1.
SAFETY DESIGN BASIS TWELVE - The CSS is designed to prevent debris which could
impair the performance of the containment spray pumps, valves, eductors, or
spray nozzles from entering the recirculation piping. Design is in accordance
with Regulatory Guide 1.82, as discussed in Table 6.2.2-1.
6.2.2.1.1.2 Power Generation Design Bases The CSS has no power generation design bases.
6.2.2.1.2 System Design 6.2.2.1.2.1 General Description The CSS, shown schematically in Figure 6.2.2-1, consists of two separate trains
of equal capacity, each independently capable of meeting the design bases.
Each train includes a containment spray pump, spray header and nozzles, spray
additive eductor, valves, and the necessary piping, instrumentation, flushing
connections, and controls. The containment spray additive tank supplies 30 weight percent (nominal) sodium hydroxide to both trains. The refueling water
storage tank supplies borated injection water to the containment spray system.
Each train takes suction from separate containment recirculation sumps during
the recirculation phase.
The CSS provides a spray of cold or subcooled borated water, adjusted with NaOH, from the upper regions of the containment to reduce the containment
pressure and temperature during either a LOCA or MSLB inside the containment.
Each CSS pump discharges into the containment atmosphere through an independent
spray header. The spray headers are located in the upper part of the reactor
building to allow maximum time for the falling spray droplets to reach thermal
equilibrium with the steam-air atmosphere. The condensation of the steam by
the falling spray results in a reduction in containment pressure and temperature. Each spray train provides adequate coverage to meet the design
requirements with respect to both containment heat removal and iodine removal.
Further discussion of the iodine removal function of the CSS is provided in
Section 6.5.2.
In the CSS, only the containment recirculation sumps and the spray headers, nozzles, and associated piping and valves are located within the containment.
The remainder of the system is located within the auxiliary building, separated
from that portion in the containment by motor-operated isolation valves.
6.2-43 Rev. 22 WOLF CREEK During the recirculation phase, leakage outside of the containment is detected
with the auxiliary building radiation indicators and alarms, temperature
alarms, and auxiliary building sump alarms. The motor-operated isolation
valves in each train assure train isolation capability in the event of leakage during the recirculation phase. Leakage detection within the auxiliary
building is discussed in Section 9.3.3.
6.2.2.1.2.2 Component Description
Mechanical components of the CSS, except those in the spray additive subsystem, are described in this section. Description of the mechanical components in the
spray additive subsystem is provided in Section 6.5.2. Component design
parameters are given in Table 6.2.2-2.
Each component in the CSS is designed and manufactured to withstand the environmental effects, including radiation, found in Table 3.11(B)-2.
CONTAINMENT SPRAY PUMPS - The two CS pumps are the vertical centrifugal type, driven by electric induction motors. The motors have open drip-proof
enclosures and are provided with adequate insulation which allows continuous
operation of a 100-percent-rated load at 50 C ambient. Power for these motors
is supplied from the Class IE 4,160-Volt busses. Power supply availability is
discussed in Section 8.3.
The pump motors are specified to have the capability of starting and
accelerating the driven equipment, under load, to a design point running speed
within 4 seconds, based on 75 percent of the rated motor voltage. The pumps
are designed to withstand a thermal transient from 37°F to 300°F occurring in
10 seconds, which exceeds the severity of the transient occurring when pump suction is switched from the RWST to the containment sump.
The shaft seals on the pumps are reliable, easy to maintain, and compatible
with the fluids to be circulated. They are designed to operate at a
temperature of 300°F, which exceeds the maximum temperature to which they will
be exposed following an accident.
The containment spray pumps are designed to handle the runout flow associated
with the startup transient, when minimal discharge head is applied.
CONTAINMENT SPRAY HEADER AND NOZZLES - Each containment spray header contains
197 hollow cone nozzles, each capable of the design flow and differential
pressure given in Table 6.2.2-2. These nozzles have a 7/16-inch spray orifice.
The nozzles produce a drop size distribution, as described in Figure 6.5-2, at
system design conditions. Special tests performed on the spray nozzles are
discussed in Section 6.5.2.2.2. The
6.2-44 Rev. 0 WOLF CREEK spray solution is completely stable and soluble at temperatures of interest in
the containment and, therefore, does not precipitate or otherwise interfere
with nozzle performance. The nozzles of each header are oriented to provide
greater than 90-percent area coverage at the operating deck of the reactor building. The area coverage at the operating deck (based on the calculated
post-LOCA containment saturation temperature) is provided in Table 6.5-2 for
various nozzle orientations. The containment spray envelope reduction factor
as a function of post-LOCA containment saturation temperature is provided in
Figure 6.5-4. The spray header design, nozzle spacing, and orientation are
shown in Figure 6.2.2-2. The containment spray header and nozzles are designed
to withstand the impulse of a water hammer at the commencement of flow.
CONTAINMENT RECIRCULATION SUMPS - The two containment recirculation sumps are collecting reservoirs from which the containment spray pumps and the residual heat removal pumps separately take suction after the contents of the refueling water storage tank have been expended. The sumps are located as far as feasible from the reactor coolant system piping and components which could become sources of debris. Thermal insulation used inside containment will be a significant source of debris. The majority of insulation is removable fiberglass blanket type enclosed in a stainless steel jacket with quick-release latches. Limited quantities of other types of insulation are used in widely dispersed locations. Insulation other than removable fiberglass blanket type has been evaluated to ensure that it will not be subject to degradation under a design basis accident or, if in a few dispersed locations the insulation should degrade under DBA conditions, the debris generated as a result of the degradation is trapped by the building components so that the debris will not adversely affect the performance of the sump. The strainer arrangement consisting of stacked modules with fine mesh perforated plates completely surrounds the inlet piping to prevent floating debris and high-density particles from entering. Sources of debris, as indicated above, are physically remote from the recirculation sumps. Debris generated as a result of a LOCA will either be retained in an area such as the reactor cavity or refueling pool or must follow a tortuous path to reach the recirculation sump strainers.
Figure 6.2.2-3 shows the stacked module arrangement.
However, the strainers have been evaluated to meet the intent of Regulatory Guide 1.82. To limit any possible vortexing, vortex breakers are placed in the suction lines from containment sumps to the containment spray pumps.
Additionally, the strainers have been evaluated for the possibility of vortexing and found to be acceptable. The suction lines from the containment sumps to the containment spray pumps are sloped to assure switchover capability. These lines, up to and including the isolation valve, are encased in guard piping.
6.2-45 Rev. 20 WOLF CREEK REFUELING WATER STORAGE TANK - The refueling water storage tank (RWST) is an
austenitic stainless steel tank containing borated water at a concentration of
2,400 to 2,500 ppm boron. The design parameters are given in Table 6.2.2-2.
The tank is an atmospheric storage tank vented directly to the atmosphere.
Thermal insulation and heating are provided to prevent the tank contents from
freezing. A manway is provided for tank internal inspection. Tank level
indication and high and low level alarms are also provided. Additional
information is provided in Section 6.3.
VALVES - CSS motor-operated valves are capable of being operated from the
control room. All valves are purchased with seats capable of limiting through
leakage to less than 2 cubic centimeters per hour per nominal inch of pipe
diameter. This is demonstrated as required by the valve purchase specifications prior to installation. Those valves with leakage criteria are tested as described in section 6.2.6. If leakage exceeds the criteria, maintenance is performed to reduce the leakage. Other MOVs without specific
leakage criteria are tested as described in section 3.9(B).6. Gate and Globe
valves are provided with backseats.
Encapsulation - The containment spray system suction lines from the containment
recirculation sumps are each provided with a single gate isolation valve
outside the containment. The piping from the sump up to and including the
valve and its motor operator is enclosed in an encapsulation arrangement which
is leaktight at the containment design pressure. A seal is provided so that
the ambient inside the encapsulation is not connected directly to the
containment sump or containment atmosphere. A single passive or active failure
in the sump lines or in the encapsulation arrangement does not provide a path
for leakage to the environment.
PIPING - The piping of each spray header contains a test connection. Air can
be introduced into this connection to verify spray nozzle flow. Check valves
immediately upstream of each spray ring header prevent system contamination due
to pressurization in the containment and provide containment isolation backup
protection.
6.2-46 Rev. 20 WOLF CREEK A containment spray pump test line between the pumps' discharges and the RWST
and lines between each pumps suction and discharge are installed for periodic testing.
6.2.2.1.2.3 System Operation
The CSS has two phases of operation, which are initiated sequentially following system actuation; they are the injection phase and the recirculation phase.
INJECTION PHASE - The CSS is actuated either manually from the control room or
on the coincidence of two-out-of-four containment Hi-3 pressure signals.
Both containment spray pumps start and the motor-operated spray ring header
isolation valves open to begin the injection phase. The same coincident signal
opens the motor-operated additive eductor suction valves to the sodium
hydroxide tank. A summary of the accident chronology for the containment spray
system is provided in Table 6.2.2-3 for the injection phase of a LOCA and MSLB inside the containment, respectively.
The containment spray pump inlet nozzle, located at El. 1,970, takes suction
from the RWST, located at El. 2,000'-6", through locked open valves.
Approximately 95 percent of the pump discharge is directed to the containment
spray ring headers. These headers are located at elevations up to
approximately 2203 feet, the highest practical level to maximize iodine removal (discussed in Section 6.5.2).The headers are located outside of and above the
internal containment structures which serve as missile barriers and are thereby
protected from missiles generated during a LOCA or MSLB. The remaining portion
of the containment spray pump discharge is bypassed through the spray additive
eductors where it is used as the motive flow to draw the spray additive
solution from the containment spray additive tank and direct it to the
containment spray pump suction. The containment spray additive tank supplies
the spray additive solution to the eductor through a motor-operated valve.
Further discussion of the operation of the spray additive subsystem is provided in Section 6.5.2.2.3. If the level in the NaOH tank reaches low-low prior to
switching to the recirculation phase, the spray additive tank isolation valves
are automatically closed to terminate the flow of spray additive solution and
prevent N 2 from being drawn into the pump suction.
On coincidence of two-out-of-four low-low-1 level signals from the RWST level
transmitters, the emergency core cooling system (ECCS) pumps switch suction to
the containment recirculation sump, as described in Section 6.3.2. Switchover for the spray pumps is manually initiated when the low-low-2 level in the RWST
is reached. The low-low-2 level alarm ensures that the system piping remains
full of water and that adequate NPSH for the spray pumps is maintained. The
RWST low-low-2 level alarms and level indicators inform the operator of the
need to make this switchover.
6.2-47 Rev. 23 WOLF CREEK The time length of the containment spray injection phase is given in Table 6.2.2-4. These times are based on the minimum RWST volume and are given for
credible combinations of minimum and maximum containment spray and ECCS
operation and runout flow rates of these pumps. The containment spray additive
design flow rate is given in Table 6.5-2.
RECIRCULATION PHASE - The recirculation phase initiated by the operator
manually shifting containment spray pump suction from the RWST to the
containment recirculation sump. The accident chronology for the containment
spray system for the recirculation phase of a LOCA is provided in Table 6.2.2-
- 3.
The RWST suction line valves remain open during the switchover to the
recirculation phase to preclude the loss of supply to the containment spray
pumps in the highly unlikely event that the isolation valve in the recirculation line is delayed in opening. The operator then remote manually closes the motor-operated valves in the RWST suction lines. If the
predetermined amount of spray additive defined in Section 6.5.2 has been added, a permissive signal from the spray additive tank level switches allows the
operator to remote manually close the motor-operated valves in the spray
additive supply lines to the containment spray additive eductor. If this
minimum level in the spray additive tank has not been reached, the valves
cannot be manually closed.
The suction line from the containment recirculation sump to the spray pump is a
sloped line which precludes air from entering the system. The single valve in
the containment sump recirculation line for the containment spray pump is
encapsulated and located outside the containment. The flow paths from the
spray pumps are the same as in the injection phase. Check valves are provided
in the recirculation sump suction lines to prevent the establishment of a flow path between the RWST and the containment sump.
Containment spray in the recirculation mode maintains an equilibrium
temperature between the containment atmosphere and the recirculation sump
water. The length of time that the CSS operates during the recirculation phase
is determined by the operator. The spray cannot be terminated until completion
of the injection phase.
6.2-48 Rev. 20 WOLF CREEK 6.2.2.1.3 Safety Evaluation
Safety evaluations are numbered to correspond to the safety design basis.
SAFETY EVALUATION ONE - The safety-related portions of the CSS are located in
the reactor and auxiliary buildings. These buildings are designed to withstand
the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and
3.8 provide the basis for the adequacy of the structural design of these
buildings.
SAFETY EVALUATION TWO - The safety-related portions of the CSS are designed to
remain functional after a SSE. Sections 3.7(B).2 and 3.9(B) provide the design
loading conditions that were considered. Section 3.6 provides the hazards analysis to assure that the system performs its intended function.
SAFETY EVALUATION THREE - There are two spray system trains with complete
redundancy of active components. Each train is capable of providing full
design flow and cooling. In the event of the failure of a pump, valve, actuation system, or any other component in one train, the other train would be
unaffected. To assure that a single failure will neither initiate a spurious
containment spray nor prevent the activation of a necessary component, the
containment spray pumps and containment header valves are actuated by the
independent containment spray actuation signal (CSAS). The containment spray
additive tank and refueling water storage tank are common to the two trains.
Redundant level indication for each of these tanks is provided. No power-
operated valve is installed in the common suction lines from the tanks so that
it is impossible for an active failure to disable both trains during the
injection phase. Single failure analysis for the CSS is given in Table 6.2.2-5 and for the spray additive subsystem in Table 6.5-4.
The emergency power supply pump room cooling and control and instrumentation
systems serving one train are independent of comparable supporting systems for
the other train. All vital power can be supplied from either onsite or offsite
power systems, as described in Chapter 8.0. Minimum availability of the CSS is
discussed in Technical Specifications.
SAFETY EVALUATION FOUR - The CSS is initially tested with the program given in
Chapter 14.0. Functional testing is done in accordance with Section 6.2.2.1.4.
6.2-49 Rev. 20 WOLF CREEK Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the CSS.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portion of
this system and supporting systems. Section 6.2.2.1.2.2 shows that safety-
related components meet the design and fabrication codes given in Section 3.2.
All the power supplies and the control functions necessary for the safe
function of the CSS are Class IE, as described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - Section 6.2.2.1.2.1 describes provisions made to
identify and isolate leakage or malfunction and to isolate the nonsafety-
related portions of the system.
SAFETY EVALUATION SEVEN - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability.
SAFETY EVALUATION EIGHT - As shown by the containment analysis and the
description of the analytical methods and models given in Section 6.2.1, the
containment spray system, in conjunction with the emergency core cooling system
and the containment fan coolers, is capable of removing sufficient heat energy
and subsequent decay heat from the containment atmosphere following the
hypothesized LOCA and MSLB inside the containment to maintain the containment
pressure below the design pressure. Curves showing sump temperature, heat
generation rates, heat removal rates of the containment heat removal systems, and containment total pressure, vapor pressure, and temperature as a function
of time for minimum engineered safety features performance are also given in
Section 6.2.1.
During the injection phase, all pressure transient analyses take credit for a spray system capable of delivering borated 100°F spray water at the design flow
rate. For the design basis LOCA and MSLB accident, credit is taken for spray
flow initiation within 60 seconds.
A minimum water volume of 394,000 gallons is maintained in the RWST to ensure
that, after a LOCA, sufficient water is injected for emergency core cooling and
for rapidly reducing the containment pressure and temperature. In addition, this volume ensures that sufficient water is available in the containment sump
to permit recirculation flow to the core and the containment and to meet the
NPSH requirements of the residual heat removal and containment spray pumps and
assures that a sufficient water volume is available in the RWST to allow for
manual switchover of the containment spray pumps.
6.2-50 Rev. 20 WOLF CREEK For the recirculation phase, while the safety injection system pumps are still
operating after a LOCA, containment pressure transient analysis in Section
6.2.1 assumes residual heat removal by heat exchangers, as described in Section
5.4.7. Credit is taken for heat removal from heat exchangers during the recirculation phase based on a tube side inlet temperature equal to the
recirculation sump temperature, which is given in Section 6.2.1 as a function
of time after the accident.
Each spray header train provides a minimum of 90-percent area coverage at the
operating deck, as demonstrated in Figure 6.2.2-4. Area coverage by these
spray nozzles varies as a function of saturation temperature. The design basis
coverage for the nozzles at various orientations is provided in Table 6.5-2 and
is based on the calculated containment saturation temperature. Figure 6.5-4
provides the curve of the containment spray envelope reduction factor to determine the design basis coverage. The minimum of 90-percent area coverage at the operating deck is used as a layout guide for the location of the spray
nozzles on the containment spray headers to assure 100-percent volumetric
coverage above the operating floor of the containment. Physical obstructions, such as the containment polar crane, are not considered to impede the spray
coverage due to the extreme turbulence created by the containment air coolers, the spray within the containment, and the blowdown resulting from the
postulated rupture. Thus, the header layout coupled with the extreme
turbulence assures the validity of a one-region model above the operating deck
for accident dose calculations (see Chapter 15.0).
Discussion of the volume of containment covered by the sprays is provided in
Section 6.5.2.
SAFETY EVALUATION NINE - That part of the CSS located inside the containment is designed to remain operable in the containment accident environment described in Section 3.11(B). The material compatibility of the containment spray system
in contact with the post-accident recirculation fluids is discussed in Section
6.1. That part of the CSS located in the auxiliary building is designed to
remain operable in the auxiliary building accident environment described in
Section 3.11(B).
SAFETY EVALUATION TEN - The basic borate spray solution is stable under the
anticipated LOCA thermal and radiolytic conditions. The borate solution is
chemically compatible with components with which it may come into contact. The
use of materials which react with sodium hydroxide to release hydrogen (principally zinc and aluminum) has been minimized in equipment located inside the containment. An analysis of hydrogen generation following a LOCA is given in Section 6.2.5.
6.2-51 Rev. 23 WOLF CREEK SAFETY EVALUATION ELEVEN - System piping size and layout provides adequate NPSH
to the containment spray pump during all anticipated operating conditions, in
accordance with Regulatory Guide 1.1. In calculating available NPSH, the
conservative assumption has been made that the water in the containment sump after a design basis LOCA is a saturated liquid, and no credit has been taken
for anticipated subcooling. That is, although NPSH = elevation head +
(containment pressure - liquid vapor pressure) - suction line losses, the (containment pressure - liquid vapor pressure) term has been assumed to be
zero. Calculated NPSH exceeds required NPSH by at least 10 percent. The
recirculation piping penetrating the containment sumps is nearly horizontal to
minimize vortexing. In addition, a vortex breaker is provided in the inlet of
the piping from the sump.
In calculating the water level within the reactor building which contributes to the NPSH available to the containment spray pumps at the beginning of its recirculation phase, consideration has been given to the potential mechanisms
of water loss within the reactor building. These water loss mechanisms include
water present in the vapor phase, water loss to compartments below El. 2,000, water loss above El. 2,000, and water loss due to wetted surfaces. Tables
6.2.2-6 and 6.2.2-6a identify each water source which releases water to the
reactor building and its associated mass and each potential water loss
mechanism and the volume of water not assumed to contribute to the water level
within the containment for a large LOCA and a MSLB, respectively. The static
head available to contribute to the NPSH of the pump, suction line losses, and
the minimum NPSH available are also given in Table 6.2.2-7. The CSS pump NPSH
versus flow is shown in Figure 6.2.2-5. The reduction in water level due to
potential water loss mechanisms is considered in the calculated NPSH available.
SAFETY EVALUATION TWELVE - Recirculation sump construction provides straining down to 0.045-inch strainer hole size to prevent entrained particles in excess of that size from entering the containment recirculation sump and containment spray system suction piping. Restrictions in the reactor core channels and ECCS throttle valves are the minimum restrictions and, therefore, the basis of the strainer hole opening size.
Since the containment spray pumps are designed to operate with entrained
particles up to 1/4 inch in diameter and the minimum constriction size in the
spray nozzles is 7/16 inch, this strainer hole size is adequate to assure proper system operability.
6.2-52 Rev. 20 WOLF CREEK Each strainer provides sufficient NPSH to the ECCS pumps to maintain recirculation cooling during an event.
The sump curb does not allow flow into the sump below 6 inches above the concrete floor level surrounding the sump. This arrangement leaves ample depth for buildup of high-density debris without affecting sump performance.
Additionally, the velocity of recirculated fluids approaching the curb will be between 0.01 and 0.08 fps for all modes of operation following a LOCA or MSLB, and thus a low velocity settling region for high-density particles is provided.
Table 6.2.2-9 provides flow velocities at several times and locations for a large LOCA and an MSLB.
Any debris which eludes the curb passes into the sump through the 0.045 inch perforated plate and will be drawn into the suction piping for the containment spray and residual heat removal systems. Such debris is small enough to pass through any restrictions in the ECCS throttle valves, the Containment Spray
System, or the reactor vessel channels, and will eventually be pumped back into
the containment.
A comparison of the containment recirculation sump design features with each of the positions of Regulatory Guide 1.82, "Sump for Emergency Core Cooling and
Containment Spray Systems," is provided in Table 6.2.2-1.
6.2.2.1.4 Tests and Inspections
Testing and inspection of components of the CSS, except those in the spray
additive subsystem, are discussed in this section. Testing and inspection of
components in the spray additive subsystem are discussed in Section 6.5.2.4.
Each containment spray pump has a shop test to generate complete performance curves. The test includes verifying total differential developed head (TDH), efficiency and brake horsepower for various flow rates. An NPSH test for
various flow rates was performed on one pump. A shop thermal transient
analysis, from ambient temperature to 350 F in 10 seconds, has been performed
on the CSS pump. Results of that analysis assure that the design is suitable
for the switchover from the injection to the recirculation phase.
6.2-53 Rev. 20 WOLF CREEK The strainer configuration on the containment recirculation sumps is shop tested to verify that all design requirements are adequately met.
The spray nozzles' design parameters were verified with prototype tests in the
vendor's shop. Results of those test are provided in Section 6.5.2.2.2.
PREOPERATIONAL TESTING - Instruments are calibrated prior to system preoperational testing. Alarm functions are checked for operability and limits
during preoperational testing. The flow paths and flow capacities of all
components are verified during preoperational tests.
The functional test of the ECCS, described in Section 6.3, demonstrates proper
transfer to the emergency diesel generator power source in the event of a loss
of power. 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 functionally tested separately.
The spray header nozzle performance is verified during the preoperational
testing by blowing air through the nozzles and observing the movement of the
telltales.
The objectives of preoperational testing are to:
- a. Demonstrate that the system is adequate to meet the design pressure and temperature conditions. Components are tested in conformance with applicable codes.
- b. Demonstrate that the spray nozzles in the containment spray header are clear of obstructions by passing air through them, utilizing test connections.
- c. Verify that the proper sequencing of valves and pumps occurs on initiation of the CSS and demonstrate the proper operation of remotely operated valves.
- d. Verify the operation of the spray pumps. Each spray pump is operated at full flow to verify that it meets the design curve generated during shop testing. Both design point and runout flow rates are utilized to verify that the pump performance is within design. In addition, each spray pump is operated at minimum flow, which is directed back to the refueling water storage tank. A flow orifice is provided to regulate minimum flow to that required for routine testing.
6.2-54 Rev. 20 WOLF CREEK The sump strainers have been evaluated for vortex formation, air ingestion and
void fraction, and the results were determined to be acceptable. In addition, head loss testing was performed on the strainers. Data from these tests
together with known pressure drops across suction lines and valves (determined using standard engineering calculations) verified that the available net
positive suction head is adequate.
Further details of each test which was performed are discussed in Chapter 14.0.
OPERATIONAL TESTING - The CSS is designed to permit periodic determination of
proper system operability, as specified in the Technical Specifications. The
objectives of operational testing are to:
- a. 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.
- b. Verify the operation of the spray pumps. Each pump is run at full flow and the flow is directed back to the pump suction or the RWST.
To assure the structural and leaktight integrity of components, the operability
and performance of the active components, and the operability of the system as
a whole, the system is periodically tested up to the last isolation valve
before the containment penetration. The testing is accomplished by using a recirculation line back to the RWST, or a test line between each pumps suction and discharge, which allows a flow path to achieve full flow testing. During the full flow test alignment, some flow can be directed back to the RWST as needed for a heat sink. Sodium hydroxide is not sent to the RWST so the eductor subsystem is to be tested by other means, as discussed in Section 6.5.2. All instrumentation will also be periodically checked and calibrated.
The CSS actuation is verified as follows:
6.2-55 Rev. 23 WOLF CREEK
- a. A containment spray actuation signal (CSAS) subchannel is
actuated during normal operation to start the containment
spray pump.
- b. A separate CSAS slave relay is actuated during normal
reactor operation to ensure the opening of the containment
header valves. The CSS pump is not operating.
A visual inspection is performed to verify that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the
containment sump and cause restriction of the pump suctions during LOCA
conditions. Visual inspections are performed:
- a. for all accessible areas of the containment prior to establishing containment operability, and
- b. at least once daily of the areas affected within containment by containment entry and during the final entry when containment
operability is established.
6.2.2.1.5 Instrumentation Requirements
The CSS instrumentation was designed to facilitate automatic operation, remote
control, and continuous indication of system parameters. Discussion of
instrumentation in the spray additive subsystem is provided in Section 6.5.2.5.
The containment has redundant analog level channels for sump recirculation with
indication and alarms in the control room.
These circuits aid the operator in determining the presence and rate of increase of the sump water level.
All system motor-operated valves have position indication provided in, and are
operable from, the control room. This allows the operator to continuously
monitor system status and remotely operate valves, as necessary. Details of
the design and logic of the instrumentation are discussed in Chapter 7.0.
6.2.2.1.6 Materials
The CSS is constructed primarily of corrosion-resistant austenitic stainless
steel and contains none of the restricted materials discussed in Section
6.1.1.1.2.
Construction materials for components in the CSS, except for components in the
spray additive subsystem, are provided in Table 6.2.2-2. Discussion of
construction materials for components in the spray additive subsystem is
provided in Table 6.5-3.
Further discussion of the materials associated with the CSS, including
containment spray fluid chemistry, is given in Section 6.5.2.6.
6.2.2.2 Containment Cooling System
The containment cooling system (CtCS), in conjunction with the containment HVAC
systems described in Section 9.4.6, functions during normal plant operation to
maintain a suitable atmosphere for equipment located within the containment.
Subsequent to a DBA
6.2-56 Rev. 20 WOLF CREEK within the containment, the containment cooling system provides a means of
cooling the containment atmosphere to reduce pressure and thus reduce the
potential for containment leakage of airborne and gaseous radioactivity to the
environment.
6.2.2.2.1 Design Bases
6.2.2.2.1.1 Safety Design Bases
The CtCS, excluding the system ductwork downstream of the cooler discharge
plenum, is safety related and required to function following a DBA to achieve
and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The CtCS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, or external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The CtCS is designed to remain functional after a
safe shutdown earthquake or to perform its intended function following a
postulated hazard, such as a fire, internal missile, or pipe break (GDC-3 and
4).
SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a
single active component failure coincident with the loss of offsite power (GDC-
38).
SAFETY DESIGN BASIS FOUR - Active components are capable of being tested during
plant operation. Provisions are made to allow for inservice inspection of
components at appropriate times specified in the ASME Boiler and Pressure Vessel Code, Section XI (GDC-39 and 40).
SAFETY DESIGN BASIS FIVE - The CtCS is designed and fabricated to codes
consistent with the quality group classification assigned by Regulatory Guide
1.26 and the seismic category assigned by Regulatory Guide 1.29. The power
supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The capability of isolating components, systems, or
piping is provided, if required, so that the system's safety function is not
compromised. This includes the bypassing of the nonsafety-related ductwork
portions of the system.
SAFETY DESIGN BASIS SEVEN - The CtCS, in conjunction with the CSS, is capable
of removing sufficient heat energy and subsequent decay heat from the
containment atmosphere following the LOCA or MSLB accident to maintain the
containment pressure below design values.
6.2-57 Rev. 0 WOLF CREEK Section 6.2.1, Containment Functional Design, provides the assumptions as to
sources and amounts of energy considered and the analyses of the containment
pressure transient following a LOCA or an MSLB accident inside the containment.
Actual containment fan cooler system parameters are such that those used in the analyses are equal to or more conservative than the actual containment fan
cooler system capability.
SAFETY DESIGN BASIS EIGHT - The containment coolers, including the fan/motor
combination, remain operable in the accident environment.
SAFETY DESIGN BASIS NINE - The containment coolers, in conjunction with
Essential Service Water System (ESWS), provides sufficient heat energy to
maintain the ESWS inlet trash racks from being blocked with frazil ice.
6.2.2.2.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The containment cooling system, operating
in conjunction with the containment heating, ventilating, and air-conditioning
system described in Section 9.4.6, is designed to limit the ambient containment
air temperature during normal plant operation to 120°F. During normal plant
operations, the hydrogen mixing fans are designed to provide sufficient air
flow through the steam generator compartments so that a suitable environment
for the equipment in the steam generator compartment can be maintained.
6.2.2.2.2 System Description
6.2.2.2.2.1 General Description
The containment cooling system provides cooling by recirculation of the containment air across air-to-water heat exchangers. The bulk of this cooled air is supplied to the lower regions of the steam generator compartments. The
remaining air is supplied to the instrument tunnel and at each level (operating
floor and below) of the containment outside the secondary shield wall. The air
supplied to each steam generator compartment is drawn upwards through the
compartments by the hydrogen mixing fans and discharged into the upper
elevations of the containment.
6.2.2.2.2.2 Component Description
Design parameters for the major components of the containment cooling system
are provided in Table 6.2.2-2.
CONTAINMENT COOLER FAN - The containment cooler fans are located vertically in
the bottom of the cooler housing. Fans are vaneaxial fans with two-speed
motors. The fans and motors are designed for high-speed operation during
normal plant operations and for low-speed operation under post-LOCA conditions.
6.2-58 Rev. 11 WOLF CREEK CONTAINMENT COOLER HOUSING/DISCHARGE PLENUM - The containment cooler housing
and discharge plenums are constructed of structural steel framework and
galvanized steel coverings.
The containment cooler housing, including the section of ductwork containing
the fusible link plates, is designed to sustain a differential pressure of 2
psi during pressure transients associated with accident conditions. An
analysis which was performed to establish the differential pressure across the
cooler housing indicates the maximum differential to be less than 0.1 psi (2.8
in. w.g.) under accident conditions. Ductwork was not considered in the
analysis since it is designed to separate from the cooler by action of the
fusible link plates. The fusible link plates are steel plates which are hinged
to the ductwork and held in a closed position by the fusible links (typical
detail is shown in Figure 6.2.2-6). The plates employ a release mechanism so that after fusion of the links the plates release from the ductwork. The fusible links are designed to release at a temperature of approximately 160°F.
The open area vacated by the plates exceeds the cross-sectional area of the
fan, thus providing an unrestricted flow path.
6.2.2.2.2.3 System Operation
NORMAL OPERATION - Containment coolers are operated as required to provide
containment cooling capabilities, of approximately 9.2 x 10 6 Btu/hr. In hot weather, four coolers are normally operated at high speed to maintain proper air flow distribution. During cold weather some fans may be operated at slow
speed or switched off, but in all conditions at least one fan cooler in each
train is maintained in operation in Modes 1 through 4 to assure heat input to the ESW system adequate to prevent frazil ice from blocking the ESW intakes
during certain winter weather conditions. The coolers are normally operating
with service water providing flow to the coils. The coil heat removal
capabilities were designed assuming a tube fouling factor of 0.002.
Condensate from the fan cooler coils is collected and measured to detect leaks
into the containment atmosphere, as discussed in Section 5.2.5.
PLANT SHUTDOWN/REFUELING - The containment coolers may be operated during
shutdown/refueling operations to provide supplemental air distribution within the containment. The containment cooler fans may be operated at low speed to reduce noise levels within the containment during this mode of operation. The
coolers may be operated with the service water to provide supplemental cooling
or without service water for supplemental heating by utilizing the motor heat
load.
6.2-59 Rev. 11 WOLF CREEK CONTAINMENT INTEGRATED LEAK RATE TESTING - The containment coolers may be
operated during containment integrated leak rate testing (ILRT) to control
containment temperature. The coolers are operated with service water to
provide cooling and without service water to provide heating, by utilizing the motor heat load, during the test procedure. The fans are operated at low
speeds during this elevated pressure condition to prevent motor overload.
POSTACCIDENT OPERATION - Following an SIS, the fans are designed to start
automatically in slow speed if not already running. If running in high (normal) speed, the fans automatically shift to slow speed. Assuming loss of
offsite power, the containment cooler fans are started 45 seconds after
generation of the SIS.
To compensate for the reduced air flow over the coils and to maximize heat removal, the cooling water flow through the cooling coils for each unit is automatically increased from 925 gpm to 1,000 gpm upon receipt of a SIS. The
fusible link plates open to allow unrestricted flow through the air coolers.
Under design accident conditions, each containment cooler is capable of
removing at least the amount of heat assumed in the containment P/T response
analysis as shown on Figure 6.2.1-15. The coil heat removal capabilities were
designed assuming a tube fouling factor of 0.002.
The fan can be operated from the control room at any time, but cannot be
manually operated at high speed if a containment high pressure signal is in
effect in order to prevent motor overload.
The postaccident air-distribution system is designed to discharge the air from
each unit through the opening left by the fusible link plate. The fusible link
plates are steel plates which are hinged to the ductwork and held in a closed position by the fusible links. The plates employ a release mechanism, using counterbalance weights to ensure that after fusion of the links the plates will
release from the ductwork without the aid of the fan head and against the
pressure differential established during the pressure transient. The fusible
links are designed to release at a temperature of approximately 160°F. The
open area vacated by the plates approximately equals the cross-sectional area
of the fan, thus providing an unrestricted flow path.
Under design conditions, it is assumed that the existing ductwork is restricted
so that all the air is discharged through this opening. Under these
conditions, the throw is approximately 100 feet. Thus, the discharge from the
units is well beyond their intake regions, preventing any short circuiting.
The air streams drop off toward the end of the throw and tend to settle toward
the bottom of the containment due to the slightly lower temperatures and the
air flow patterns established by
6.2-60 Rev. 15 WOLF CREEK natural convection caused by post accident conditions inside the containment (Ref. 26). The volume of air recirculated in one hour by the combined air
flows of one train of the containment coolers is approximately three times the
containment free volume. These air flow patterns and recirculation volumes provide adequate circulation and, therefore, sufficient post accident mixing of
the containment atmosphere.
6.2.2.2.3 Safety Evaluation
Safety evaluations are numbered to correspond to the safety design bases in
Section 6.2.2.2.1.
SAFETY EVALUATION ONE - The safety-related portions of the containment cooling
system are located in the reactor building. This building is designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of the structural design of
these buildings.
SAFETY EVALUATION TWO - The safety-related portions of the containment cooling
system are designed to remain functional after a SSE. Sections 3.7(B).2 and
3.9(B) provide the design loading conditions that were considered. Sections
3.5 and 3.6 provide the hazards analyses to assure that a post accident safe
shutdown, as outlined in Section 7.4, can be achieved and maintained.
SAFETY EVALUATION THREE - The system description for the containment cooling
system shows that complete redundancy is provided and, as indicated by Table
6.2.2-8, no single failure will compromise the system's safety functions. All
vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0.
SAFETY EVALUATION FOUR - The containment cooling system is initially tested
with the program given in Chapter 14.0. Periodic inservice functional testing
is done in accordance with Section 6.2.2.2.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the containment cooling system.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group
classification and seismic category applicable to the safety-related portion of
this system and supporting system. All the power supplies and control
functions necessary for safe function of the containment cooling system are
Class IE, as described in Chapters 7.0 and 8.0.
6.2-61 Rev. 19 WOLF CREEK SAFETY EVALUATION SIX - Section 6.2.2.2.2.3 describes provisions made to allow
the bypassing of the nonsafety-related ductwork portions of the system.
SAFETY EVALUATION SEVEN - As shown by the containment analysis and the description of the analytical methods and models given in Section 6.2.1, the
containment cooling system, in conjunction with the containment spray system, is capable of removing sufficient energy and subsequent decay heat from the
containment atmosphere following the hypothesized LOCA or MSLB accident inside
the containment to maintain the containment below the design pressure. Both
analyses assume the single failure which results in the minimum containment
cooling capability.
Curves showing sump temperature, heat generation rates, heat removal rates of
the containment heat removal systems, and containment total pressure, vapor pressure, and temperature as a function of time for minimum engineered safety features performance are given in Section 6.2.1. The containment cooler heat
removal rates as a function of containment temperature and pressure are given
in Figure 6.2.1-15. This data has been furnished by American Air Filter and is
supported by their topical report (Ref. 10). A constant essential service
water temperature of 95°F at the coil inlet has been assumed. This is the
maximum conservatively calculated temperature that would exist at any time
during the accident. The assumptions used in calculating this temperature are
discussed in Section 9.2.5.
SAFETY EVALUATION EIGHT - The containment cooler fan/motor combination is
qualified to operate during the DBA, in accordance with IEEE-334, 1974.
Section 6.2.2.2.2.2 provides the basis for the assumption of structural
integrity of the cooler housing and discharge plenum during a DBA. American
Air Filter (Ref.10) demonstrates the compatibility of the housing and plenum materials with the DBA environment.
SAFETY EVALUATION NINE - As described in Section 9.2.1.2.2.3 and Table 9.2-25, the containment cooling system supports the ESWS by providing part of the heat
energy needed to maintain the ESWS inlet trash racks from being blocked with
6.2.2.2.4 Tests and Inspections
Preoperational Chapter 14.0. One containment cooler fan is tested in
accordance with AMCA Standard testing is described in Test Code 211, "Certified
Rating for Air-Moving Devices."
The analytical data used to predict coil performance for both normal and DBA
conditions are based upon the tests and data in Reference 10.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
6.2-62 Rev. 10 WOLF CREEK 6.2.2.2.5 Instrumentation Applications
Each containment cooler is monitored for leaving air temperature via the plant computer. Each containment cooler motor is monitored for vibration. In addition, containment air temperature will also be monitored in the area of
each containment cooler intake. Direct control room indication is provided for
the inlet air temperatures. The leaving air temperature can be displayed in
the control room via the plant computer.
Each containment cooler fan is operable from the control room.
6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN
Based on the fission product removal and control systems discussed in Section
6.5 and the radiological consequences analyzed in Chapter 15.0 following a
LOCA, no secondary containment is required for WCGS.
6 2.4 CONTAINMENT ISOLATION SYSTEM
The containment isolation system allows the normal or emergency passage of fluids through the containment boundary while preserving the ability of the
boundary to minimize the release of fission products following a LOCA or fuel
handling accident within the containment.
6.2.4.1 Design Bases 6.2.4.1.1 Safety Design Bases
SAFETY DESIGN BASIS ONE - The containment isolation system is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The containment isolation system is designed to
remain functional after a safe shutdown earthquake and to perform its intended
function following the postulated hazards of fire, internal missiles, or pipe
breaks (GDC-3 and 4).
6.2-63 Rev. 22 WOLF CREEK SAFETY DESIGN BASIS THREE - The containment isolation system is designed and
fabricated to codes consistent with the quality group classification assigned
by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide
1.29. The power supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS FOUR - Piping systems penetrating the primary reactor
containment are provided with leak detection, isolation, and containment
capabilities having redundancy, reliability, and performance capabilities which
reflect the importance to safety of isolating these piping systems. Such
piping systems are designed with a capability to periodically test the
operability of the isolation valves and associated apparatus and to determine
if valve leakage is within acceptable limits (GDC-54).
SAFETY DESIGN BASIS FIVE - Each line that is part of the reactor coolant pressure boundary and that penetrates the primary reactor containment is
provided with containment isolation valves as follows:
- a. One locked closed isolation valve inside and one locked
closed isolation valve outside the containment; or
- b. One automatic isolation valve inside and one locked
closed isolation valve outside the containment; or
- c. One locked closed isolation valve inside and one
automatic isolation valve outside the containment. A
simple check valve is not used as the automatic isolation
valve outside the containment; or
- d. One automatic isolation valve inside and one automatic isolation valve outside the containment. A simple check
valve is not used as the automatic isolation valve
outside the containment; or
- e. Some other defined bases that meet the intent of
containment isolation as an alternative to a through d
above.
Isolation valves outside the containment are located as close to the
containment as practical and, upon loss of actuating power, automatic isolation
valves are designed to take the position that provides the greater safety (GDC-
55).
6.2-64 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS SIX - Each line that connects directly to the containment
atmosphere and penetrates the primary reactor containment is provided with
containment isolation valves as follows:
- a. One locked closed isolation valve inside and one locked
closed isolation valve outside the containment; or
- b. One automatic isolation valve inside and one locked
closed isolation valve outside the containment; or
- c. One locked closed isolation valve inside and one
automatic isolation valve outside the containment. A
simple check valve is not used as the automatic isolation
valve outside the containment; or
- d. One automatic isolation valve inside and one automatic
isolation valve outside the containment. A simple check
valve is not used as the automatic isolation valve
outside the containment; or
- e. Some other defined bases that meet the intent of
containment isolation, as an alternative to a through d
above.
Isolation valves outside the containment are located as close to the
containment as practical and, upon loss of actuating power, automatic isolation
valves are designed to take the position that provides greater safety (GDC-56).
SAFETY DESIGN BASIS SEVEN - Each line that penetrates the primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere h as:
- a. At least one containment isolation valve which is either
automatic, locked closed, or capable of remote manual
operation; or
- b. Some other defined bases that meet the intent of
containment isolation as an alternative to a above.
Valves are outside the containment and located as close to the containment as
practical. A simple check valve is not used as the automatic isolation valve.
For a closed system, the design is commensurate with quality group B (GDC-57).
6.2-65 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS EIGHT - The containment isolation system, in conjunction
with other plant features, serves to minimize the release of fission products
generated following a LOCA or fuel handling accident within the containment.
6.2.4.1.2 Power Generation Design Basis
The containment isolation system has no power generation design basis.
6.2.4.2 System Description 6.2.4.2.1 General Description
Each piping system which penetrates the containment is provided with containment isolation features which serve to minimize the release of fission
products following a LOCA or fuel handling accident. Provisions are made to
allow for passage of emergency fluid through the boundary following a
postulated accident. Figure 6.2.4-1 provides the arrangement for each piping
penetration, along with design information and justification of how the
appropriate General Design Criteria are met. NRC SRP 6.2.4 and Regulatory
Guide 1.141 provide acceptable alternative arrangements to the explicit
arrangements given in GDC-55, 56, and 57. Each penetration is provided with a
redundant barrier so that in the event that a single failure is postulated and
one barrier does not perform as intended the containment integrity is maintained. Table 6.2.4-1 lists each penetration under the appropriate GDC and provides a reference to the section that describes the system of which the
containment penetration is an integral part.
Piping penetration sleeves have been assigned numbers P-1 through P-17 and P-
21 through P-104. Numbers P-18, 19, and 20 were not utilized. The fuel
transfer tube was assigned to P-17; however, this is not a true piping
penetration since it utilizes a blind flange which serves as the containment
boundary and is subject to Type B testing. Penetrations 36 and 68 have been
assigned to outage activities. They utilize a bolted flange closure and are
subject to Type B testing. The remainder of the "P" numbers between 1 and 104
not appearing on Figure 6.2.4-1 are spare sleeves to which closure heads have
been permanently attached, as shown in Figure 3.8-47. These penetration
sleeves include P-31, 33, 35, 37, 38, 46, 47, 50, 60, 61, 72, 77, 81, 94, 96, 100, and 102. The leaktight integrity of the sleeve and closure head is verified during the periodic Type A tests.
For those systems which have automatic isolation valves or for which remote
manual isolation is provided, Section 6.2.4.5 describes the vital power supply
and associated actuation system.
6.2-66 Rev. 6 WOLF CREEK Two phases of valve actuation are considered in Figure 6.2.4-1. The actuation
signal which occurs directly as a result of the event initiating containment
isolation is designated as the primary actuation signal. The primary valve
position is a consequence of the primary actuation signal. If a change in valve position is required at any time following primary actuation, a
secondary actuation signal is generated which places the valve in the
secondary position.
The closure times for automatic isolation valves are provided in Figure 6.2.4-
- 1. The containment purge system provides a direct path between the
containment and outside atmospheres. As described in Section 9.4, the 18-inch
4,000 cfm minipurge lines may be open during normal plant operation and during
shutdown condition and are provided with isolation valves capable of three-
second closure. The 36-inch 20,000 cfm purge lines are open only during a shutdown condition and are provided with an isolation valve capable of 10-second closure. An analysis of the radiological consequences and the effect
on the containment backpressure due to the release of containment atmosphere
are discussed in Section 6.2.1.5 and Chapter 15.0.
In the event of a LOCA, the secondary shield wall prevents any missiles or jet
impingement from damaging or degrading the performance capability of
containment isolation. Sections 3.5 and 3.6 discuss in detail the missiles and
pipe break effects, and Section 3.8 discusses the internal structures, including the secondary shield wall. The operators for all power-operated
containment isolation valves inside the containment are located above the
maximum water level, following a LOCA. In addition, lines associated with
those penetrations which are considered closed systems inside the containment
are protected from the effects of a LOCA.
Provisions are made to ensure that closure of the containment isolation valves is not inhibited by entrapped debris in the valve body. For the majority of
the systems, the fluid is demineralized water; thus quality does not affect
valve operation. For containment purge lines, screens are provided in the
lines upstream of the isolation valves. For the containment sump lines, including the emergency sump, a provision is provided to prevent large debris
from entering the system.
Some other defined bases for containment isolation are provided in NRC SRP
6.2.4 and Regulatory Guide 1.141. Compliance with Regulatory Guide 1.141 is
provided to the extent specified in Table 6.2.4-2. For the ECCS and
containment spray system penetrations, the acceptability of the alternative
arrangement relies upon provisions for the detection of possible leakage from
these
6.2-67 Rev. 19 WOLF CREEK lines outside the containment. Section 9.3.3 describes the leak detection
provisions that have been made in the plant drainage system. Other provisions, such as containment water level and system flow, temperature, and pressure
instrumentation, may be used by the operator.
In addition to containment isolation, Figure 6.2.4-1 also contains systems
which are required for post-LOCA mitigation. Since these systems, such as the
ECCS, perform additional safety-related functions, they are associated with
engineered safety features and are so indicated on Figure 6.2.4-1. Because
these systems are required to operate for post-LOCA mitigation and because they
are closed systems external to the containment, the length of the piping
between the containment and the system outside the isolation valves is not
shown.
6.2.4.2.2 Component Description
Codes and standards applicable to the piping and valves associated with
containment isolation are listed in Table 3.2-1. Containment penetrations are
classified as quality group B and seismic Category I.
Section 3.11 provides the post-LOCA environment that is used to qualify the
operability of power-operated isolation valves located inside the containment.
The containment penetrations are designed to meet the stress requirements of
NRC BTP MEB 3-1 and the classification and inspection requirements of NRC BTP
APCSB 3-1, as described in Section 3.6. Section 3.8 discusses the interface
between the piping system and the containment liner.
6.2.4.2.3 System Operation During normal operation, many penetrations are not isolated. Lines which are
not required for the passage of emergency fluids are automatically isolated
upon receipt of isolation signals, as discussed in Sections 6.2.4.5 and 7.0.
Other open lines to the containment can be isolated subsequent to the LOCA by
remote-manual operation when dictated by the emergency system functional
requirements. Lines not in use during power operation are normally closed, and
remain closed under Technical Specification, administrative control during
reactor operation; refer to Section 6.2.4.4 for a further discussion.
6.2-68 Rev. 0 WOLF CREEK Upon detection of high radioactivity indicative of a fuel handling accident
during refueling, the isolation valves in the containment purge system are
closed to minimize any fission product release to the environment.
6.2.4.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in
Section 6.2.4.1.1.
SAFETY EVALUATION ONE - The piping and valves associated with the containment
isolation system are located in the reactor and auxiliary buildings. These
buildings are designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena.
Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of
the structural design of these buildings.
SAFETY EVALUATION TWO - The piping and valves associated with the containment
isolation system are designed to remain functional after a safe shutdown
earthquake. Sections 3.7(B).2, 3.9(B), and 3.9(N) provide the design loading conditions that were considered. Sections 3.5 and 3.6 provide the hazards analyses to assure that a post accident safe shutdown, as outlined in Section
7.4, can be achieved and maintained.
SAFETY EVALUATION THREE - Section 3.2 delineates the quality group
classification and seismic category applicable to the safety-related portion
of this system and supporting systems. Figure 6.2.4-1 shows that the
components meet the design and fabrication codes given in Section 3.2. All the
power supplies and control functions necessary for the safe function of the
containment isolation system are Class IE, as described in Chapters 7.0 and
8.0.
SAFETY EVALUATION FOUR - Figure 6.2.4-1 shows the arrangement for each line
penetrating the containment and provides the design information that
demonstrates that GDC-54 is met. Leak detection capabilities are discussed in Section 9.3.3 and in the system descriptions associated with the applicable penetrations. Tests and inspections for piping penetrations are discussed in
Sections 6.2.4.4 and 6.2.6.
SAFETY EVALUATION FIVE - Figure 6.2.4-1 shows the arrangement and justifies
compliance with the intent of GDC-55 for lines that are part of the reactor
coolant pressure boundary and that penetrate the primary reactor containment.
A list of penetrations subject to GDC-55 is provided in Table 6.2.4-1.
6.2-69 Rev. 19 WOLF CREEK SAFETY EVALUATION SIX - Figure 6.2.4-1 shows the arrangement and justifies
compliance with the intent of GDC-56 for lines that are connected directly to
the containment atmosphere and penetrate the primary reactor containment. A
list of penetrations subject to GDC-56 is provided in Table 6.2.4-1.
SAFETY EVALUATION SEVEN - As indicated in Table 6.2.4-1, there are no
penetrations which are subject to GDC-57. Note that the containment
penetrations associated with the steam generators are not subject to GDC-57, since the containment barrier integrity is not breached. The boundary or
barrier against fission product leakage to the environment is the inside of the
steam generator tubes, the outside of the steam generator shell, and the
outside of the lines emanating from the steam generator shell side. Figure
6.2.4-2 shows the arrangement and justifies compliance with containment
isolation.
As shown in Section 18.2.11.2, several portions of the main steam lines are
considered essential and do not receive an automatic signal to close. These
include the atmospheric relief valves (PV-01, 02, 03, and 04) which receive no
signal and the steam supply line isolation valves (HV-05 and 06) to the AFW
pump turbines which open on AFAS.
SAFETY EVALUATION EIGHT - Sections 6.2.2, 6.5, and 9.4 and Chapter 15.0 provide
an evaluation that demonstrates that the containment isolation system, in
conjunction with other plant features, serves to minimize the release of
fission products generated following a LOCA or fuel handling accident inside
the containment.
6.2.4.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. The system associated
with each penetration is in continuous use or is periodically in use, which
demonstrates the system performance and structural and leaktight integrity of its components.
All manual valves which serve as containment isolation valves are locked or
sealed closed. The manual valves in the process lines are subject to the
surveillance requirements of the Technical Specifications. Manual valves
serving as vents, drains, and test connections within the isolation valve
envelope are subject to administrative procedures to ensure that they are in
the proper position. Since each manual valve is locked or sealed closed, the
design meets the recommendations of SRP 6.2.4, Section II.3.f.
6.2-70 Rev. 13 WOLF CREEK The containment isolation system is testable through the operational sequence
that is postulated to take place following an accident, including operation of
applicable portions of the protection system and the transfer between normal
and standby power sources.
The piping and valves associated with the containment penetration are designed
and located to permit preservice and inservice inspection in accordance with
ASME Section XI, as discussed in Section 6.6.
Each line penetrating the containment is provided with testing features to
allow containment leakrate tests in accordance with 10 CFR 50, Appendix J, as
discussed in Section 6.2.6.
6.2.4.5 Instrumentation Application The generation of a CIS-A, SLIS, CIS-B, or CPIS which isolates the appropriate
containment isolation valves is described in Section 7.3.
The CPIS serves to isolate the containment purge in the event of a fuel-
handling accident or LOCA.
The CIS-A, SLIS, and CIS-B serve to actuate the containment isolation system
following a LOCA. A CIS-A signal actuates all power-operated valves which can
be immediately closed, since doing so will not increase the potential for
damage to the containment equipment, or which are not required to be open for
the operation of essential equipment post accident.
SLIS signal actuates appropriate power-operated valves based on system functional requirements, as discussed in the appropriate system description.
As described in Section 9.2.2 and shown on Figure 9.2-15, Sheet 3, CIS-B
isolates component cooling water system (CCWS) to the components located within
the containment. The CCWS is a seismically designed closed loop system both
inside and outside of the containment. A hazards analysis of the system has
ensured that the system boundary will remain intact following a LOCA or high
energy line break.
Since the CCWS penetrations are classified as essential penetrations (refer to
Section 18.2.11.2), isolation of the system is not provided until cooling to
the RCPS is no longer warranted. During the short time period following an
accident, passive failures are not postulated, and the pressure boundary would
remain intact until a CIS-B is received. Also, the radiation monitor on the
6.2-71 Rev. 20 WOLF CREEK CCWS surge tank closes the vent valve on high radiation (refer to Section
9.2.2.5) thus preventing release of radioactivity to the auxiliary building.
As described in Section 9.3.3, Class IE level indication is provided in the
auxiliary building sumps to help identify any liquid leakage from the CCW system. Figure 7.2-1, Sheet 8, shows the actuation logic for CIS-B. The
pressure transmitters which actuate CIS-B also actuate the containment spray
system. Diversity for CIS-B is provided in the logic for manual actuation of
containment spray, which, when manually actuated, also automatically actuates
CIS-B.
For those valves for which automatic closure is not desired, based on the
system safety function, remote-manual operation is available from the control
room.
Containment isolation valves equipped with power operators and which are automatically actuated may also be controlled individually by positioning hand
switches in the control room. Except as noted below, containment isolation
valves cannot be repositioned via hand switches in the control room when the
automatic containment isolation signal is present. Reset of the automatic
signal is required to permit remote manual control of a containment isolation
valve. Containment isolation valves that require repositioning for post-event
monitoring or sampling are provided with device level manual overrides which
permit valve repositioning when the automatic isolation signal is reset. The
device manual override is described in Section 7.3.5. Containment isolation
valves with power operators are provided with open/closed indication, which is
displayed in the control room. The valve mechanism also provides a local, mechanical indication of valve position.
All power supplies and control functions necessary for containment isolation are Class lE, as described in Chapters 7.0 and 8.0.
6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT
10 CFR 50.44 was revised in 2003. The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates the requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that the hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage.
In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant beyond design-basis accidents. With the elimination of the design-basis LOCA hydrogen release, hydrogen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2.
The hydrogen control system (HCS) is an engineered safety feature which serves
to control combustible gas concentrations in the containment. The HCS consists
of redundant hydrogen recombiners, a redundant hydrogen monitoring system and a
backup hydrogen purge system. The HCS in conjunction with the provisions for hydrogen mixing satisfies GDC-41. A redundant hydrogen mixing subsystem is
also provided but is not required to assure adequate hydrogen mixing.
Sources of hydrogen gas in containment are as follows:
- a. Metal-water reaction involving the zirconium fuel
cladding and the reactor coolant
6.2-72 Rev. 21 WOLF CREEK
- b. Radiolytic decomposition of the post-LOCA emergency
cooling solutions (oxygen also evolves in this process)
- c. Corrosion of metals and paints by solutions used for emergency core cooling or containment spray
6.2.5.1 Design Bases 6.2.5.1.1 Safety Design Bases
Portions of the HCS are safety related and are required to function following a LOCA.
SAFETY DESIGN BASIS ONE - The HCS is capable of withstanding the effects of
natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, or
external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The HCS is designed to remain functional after a SSE
or a pipe break in containment (LOCA, steam line break, etc. (GDC-3 and 4).
SAFETY DESIGN BASIS THREE - Component redundancy is provided so that safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC-44).
SAFETY DESIGN BASIS FOUR - The HCS is designed and fabricated to codes
consistent with the quality group classification assigned by Regulatory Guides
1.7 and 1.26 and the seismic category assigned by Regulatory Guides 1.7 and
1.29. The power supply and control functions are in accordance with Regulatory
Guides 1.7 and 1.32.
SAFETY DESIGN BASIS FIVE - The capability of isolating components, systems, or
piping is provided, if required, so that the system's safety function is not
compromised. This includes the isolation of components to deal with leakage or
malfunctions and to isolate nonsafety-related portions of the system.
SAFETY DESIGN BASIS SIX - The containment isolation valves in the system are selected, tested, and located in accordance with the requirements of 10 CFR 50, Appendix A, General Design Criteria 54 and 56 and 10 CFR 50, Appendix J, Type C
testing.
SAFETY DESIGN BASIS SEVEN - The HCS is capable of maintaining the containment
hydrogen concentration below 4.0 volume percent, as required by Regulatory
Guide 1.7.
6.2-73 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS EIGHT - The hydrogen purge subsystem serves as a backup to
the hydrogen recombiners and is capable of venting and purging the containment
atmosphere in order to maintain the hydrogen concentration below 4.0 volume
percent following a LOCA. With the purge system operating, the doses at the exclusion area boundary and the low population zone outer boundary does not
exceed the guideline values of 10 CFR 100. Except for the containment
penetration and associated isolation valves, the purge subsystem is not
redundant or seismic Category I, as allowed by Regulatory Guide 1.7.
SAFETY DESIGN BASIS NINE - The containment design and hydrogen mixing
provisions ensure adequate mixing of the containment atmosphere in order to
eliminate stagnant pockets and prevent stratification of the hydrogen-air
mixture.
SAFETY DESIGN BASIS TEN - The hydrogen monitoring subsystem is designed to inform the operator of the hydrogen concentration inside the containment and to
provide periodic samples of the post-LOCA containment atmosphere to be analyzed
for hydrogen and/or oxygen and other substances, if required.
SAFETY DESIGN BASIS ELEVEN - The HCS is designed with provisions for periodic
inspection and testing of all safety-related components (GDC-42 and 43).
6.2.5.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS - The hydrogen mixing subsystem provides
continual mixing of the containment air during normal plant operation. The
containment penetrations in the hydrogen monitoring subsystem are closed during
normal plant operation. The remainder of the HCS performs no function during
normal plant operations.
6.2.5.2 System Design 6.2.5.2.1 General Description
The total system for control of combustible hydrogen concentrations in the containment following a LOCA, shown schematically in Figures 6.2.5-1 and 9.4-6, consists of a hydrogen monitoring subsystem that provides containment
atmosphere samples, hydrogen mixing provisions which assure a nearly uniform
hydrogen concentration in the containment atmosphere, electric (thermal)
hydrogen recombiners which provide the primary means of reducing containment
hydrogen concentrations, and a hydrogen purge subsystem which is used as a
backup system to the recombiners.
6.2-74 Rev. 8 WOLF CREEK The hydrogen monitoring, mixing and recombiner subsystems are designed to meet
seismic Category I requirements and the single failure criterion, as defined in
Section 3.1. The function of the hydrogen mixing fans including the mixing
provisions following a LOCA are discussed in Section 6.2.5.3, Safety Design
Basis Nine Evaluation and Section 6.2.5.2.2.2, Hydrogen Mixing Provisions.
Generation of hydrogen is discussed in Section 6.2.5.2.3.
Those portions of the HCS that are exposed to the post-accident environment are
located within containment except for portions of the hydrogen monitoring
system. Leakage outside the containment is detected with the auxiliary
building radiation indicators and alarms. The solenoid-operated isolation
valves in each train ensure train isolation capability in the event of leakage.
6.2.5.2.2 Component Description
Design data for major components of the HCS are presented in Table 6.2.5-1.
Codes and standards applicable to this system are listed in Table 3.2-1.
6.2.5.2.2.1 Hydrogen Recombiner Subsystem
License Amendment No. 157 was issued by the NRC on January 31, 2005 and deleted
the Technical Specification requirements for the hydrogen recombiners. The recombiners and their associated controls are still installed in the plant in a deenergized condition.
Each recombiner subsystem consists of a control panel located in the control
building, a control switch located on the main control board, a power supply
cabinet located in the control building, and a recombiner located on the
operating deck of the containment. There are no moving parts or controls
inside the containment. Heating of air within the unit causes air flow by
natural convection. The recombiner is a completely passive device.
The power supply cabinet located in the control building contains an isolation
transformer, plus a controller, to regulate the power supply to the recombiner.
This equipment is not exposed to the post-LOCA environment. The controls for
the power supply are located in the control room and are manually actuated.
Each hydrogen recombiner consists of the following design features:
- a. A preheater section, consisting of a shroud placed around the central heaters to take advantage of heat conduction through the central walls, for preheating incoming air
- b. An orifice plate to regulate the rate of air flow through the unit
- c. A heater section, consisting of four banks of metal-sheathed electric resistance heaters, to heat the air flowing through it to hydrogen-oxygen recombination temperatures
6.2-75 Rev. 29 WOLF CREEK
- d. An exhaust chamber which mixes and dilutes the hot
effluent with containment air to lower the temperature
of the discharge stream
- e. An outer enclosure to protect the unit from impingement
- f. No need for external services except electrical power
Containment atmosphere is heated within the recombiner in a vertical duct, causing it to rise by natural convection. As it rises, replacement air is
drawn through intake louvers downward through a preheater section which will
temper the air and lower its relative humidity. The preheated air then flows
through an orifice plate, sized to maintain a 100-scfm flow rate, to the heater section. The air flow is heated to a temperature above 1,150°F, the reaction temperature for the hydrogen-oxygen reaction. Any free hydrogen present reacts
with atmospheric oxygen to form water vapor. After passing through the heater
section, the flow enters a mixing section which is a louvered chamber where the
hot gases are mixed and cooled with containment atmosphere before the gases are
discharged directly into the containment. The air-discharge louvers are
located on three sides of the recombiner. To avoid short-circuiting of
previously processed air, no discharge louvers are located on the intake side
of the recombiner.
Tests have verified that the hydrogen-oxygen recombination is not a catalytic
surface effect associated with the heaters (see Section 6.2.5.4), but occurs
due to the increased temperature of the process gases. As the phenomenon is
not a catalytic effect, saturation of the unit cannot occur.
Two recombiners are provided to meet the requirements for redundancy and independence. Each recombiner is powered from a separate Class IE 480-V load
center described in Chapter 8.0 and is provided with a separate power panel and
control panel. No interdependency exists between this system and the other
safety-related subsystems.
The unit is manufactured of corrosion-resistant, high-temperature material.
The electric hydrogen recombiner uses commercial-type electric resistance
heaters sheathed with Incoloy-800, which is an excellent corrosion-resistant
material for this service. The recombiner heaters operate at significantly
lower power densities than similar heaters used in commercial practice.
6.2-76 Rev. 0 WOLF CREEK Operation of the recombiner is performed manually from a switch on the main
control board or from a control panel located in the control room. The power
panel for the recombiner contains an isolation transformer plus a controller to
regulate power into the recombiner. This equipment is not exposed to the post-LOCA environment. For equipment test and periodic checkout, a thermocouple
readout instrument is also provided in the control panel for monitoring
temperatures in the recombiner.
6.2.5.2.2.2 Hydrogen Mixing Provisions
The containment design is such that mixing, adequate to prevent the formation
of hydrogen pockets, is assured without reliance on the hydrogen mixing fans.
The hydrogen mixing subsystem shown in Figure 9.4-6 is provided for
defense-in-depth and consists of four mixing fans which enhance the uniform mixing of the containment atmosphere. When operating, these fans draw air from the steam generator compartments and discharge it toward the upper regions of
the containment. This complements the air patterns established by the initial
blowdown, the natural convection, the containment air coolers, which take
suction from above the operating floor level and discharge to the lower regions
of the containment, and the containment sprays which cool the air and cause it
to drop to lower elevations. Two speed, hydrogen mixing fans are provided for
additional mixing when operating. The design flow rate of the hydrogen mixing
fans (high-speed operation) is based on air distribution requirements during
normal operation when a containment air cooler is taken out of service. The
design flow rate of the hydrogen mixing fans (low-speed operation) exceeds, with a large factor of safety, the air distribution requirements to ensure
uniform concentrations of hydrogen throughout the containment. An elevation
drawing showing the expected air flow patterns is provided in Figure 6.2.2-7.
Further information is contained in Sections 6.2.2.2 and 6.2.5.3.
6.2.5.2.2.3 Hydrogen Monitoring Subsystem
License Amendment No. 157 was issued by the NRC on January 31, 2005 relocated the Technical Specification requirements for the hydrogen monitors to the Technical Requirements Manual. The NRC has indicated that since the monitors are required to diagnose the course of beyond design-basis accidents, the monitors must be maintained capable of diagnosing beyond design-basis accidents. The NRC has also indicated that the monitors no longer meet the definition of Category 1 in Regulatory Guide 1.97 and that the monitors may be categorized as Category 3 Regulatory Guide 1.97 instrumentation.
Each redundant hydrogen monitoring train in the hydrogen monitoring subsystem
consists of a hydrogen analyzer and two associated sample lines with isolation
valves inside and outside the containment. These sampling lines are designed
to be free of water traps (runs where liquid could accumulate), and are equipped with sufficient heat tracing to prevent condensation of the sample
being supplied to the analyzers.
After the sample has been analyzed, it is returned to the containment. The
analyzers are located in accessible areas outside of the containment. The
hydrogen monitoring subsystem pressure boundary outside the containment is in
accordance with the criteria of Regulatory Guide 1.26, quality group B.
Solenoid-operated isolation valves are provided to obtain samples from two
locations within the containment for each train. One sampling
6.2-77 Rev. 21 WOLF CREEK point is above the main operating level near the intake of the containment air coolers, and the other is near the post-LOCA water level in the containment
recirculation sumps. The operator may select either of these sampling points
from the main control room.
The operation of the hydrogen gas analyzer is based on the measurement of
thermal conductivity of the gaseous containment atmosphere sample. The thermal
conductivity of the gas mixture changes proportionally to the changes in the
concentration of the individual gas constituents of the mixture. The thermal
conductivity of hydrogen is far greater (approximately seven times the thermal
conductivity of air) than any other gases or vapors expected to be present.
The operation of the hydrogen monitoring subsystem is not limited due to
radiation, moisture, or temperature expected at the equipment location. The
equipment qualification testing, including radiation exposure, aging and
vibration, satisfies IEEE Standards 323-1974 and 344-1975.
6.2.5.2.2.4 Hydrogen Purge Subsystem
The hydrogen purge subsystem utilizes the Emergency Exhaust System to perform
its functions. The emergency exhaust system is described in Section 9.4.3.
The isolation valve is the only moving part located inside the containment.
The hydrogen purge subsystem is designed to vent containment atmosphere at a
rate of 100 scfm.
The hydrogen purge subsystem has one penetration through which the containment
air is vented and filtered. This purge line is located in a missile-protected
area, and draws air from well-ventilated areas of the containment in a manner
which prevents either spray or sump water from entering the pipe. As indicated
in Section 6.2.5.3, venting would not be initiated before 4 days after a LOCA, therefore, no separate air supply line is needed. Makeup air is available through the instrument air penetration; and, if this penetration is unavailable by the time purging would be necessary, an air bottle can be connected to a
number of available penetrations. Should it be necessary to use this backup
system, operational considerations and site meteorology would determine the
timing and duration of the purges. In any case, sufficient purging would be
performed to maintain the hydrogen concentration in the containment atmosphere
below 4 volume percent.
6.2-78 Rev. 20 WOLF CREEK 6.2.5.2.3 Hydrogen Generation
Hydrogen is generated within the containment by various mechanisms, as
described below.
- a. Radiolytic Hydrogen Generation
Water is decomposed into hydrogen and oxygen by the absorption of energy emitted by nuclides contained in the fuel and those intimately mixed with the LOCA water. The quantity of hydrogen that is produced by radiolysis is a function of both the energy of ionizing radiation absorbed by the LOCA water and the net hydrogen radiolysis yield, G(H 2), pertaining to the particular physical-chemical state of the irradiated water.
Evidence indicates that the net hydrogen yield from the radiolysis of pure water is 0.44-0.45 molecule per 100 eV of absorbed energy when the gaseous radiolysis products are continuously purged from the water. In the presence of reactive solutes and water in the absence of gas purging of the solution, significant recombination of the products of radiolysis can occur, thereby reducing the net hydrogen yield. However, in accordance with Regulatory Guide 1.7, a value of 0.5 molecule/100 eV has been assumed for the net yield of hydrogen from radiolysis of all LOCA water.
The assumptions given in Regulatory Guide 1.7 were used to determine the fission product distribution after the accident. This distribution is assumed to be intantaneous after the accident, and hydrogen production is assumed to begin immediately. Fifty percent of the halogens and 1 percent of the solids are assumed to be released from the fuel and intimately mixed with the water in the sump. All noble gas activity is released from the fuel and is present in the containment atmosphere. The decay energy was calculated using the method of Standard Review Plant section 6.2.5, Appendix A, which is based on two year reactor operation. Table 6.2.5-2 gives a summary of the remaining assumptions made in the analysis.
- b. Zirconium-Water Reaction
One of the major sources of hydrogen immediately following a LOCA is due to metal-water reaction. The extent of the metal-water reaction depends strongly on the course of events assumed for the LOCA and the effectiveness of the emergency core cooling systems. The extent of metal-water reaction is evaluated in accordance with the assumptions of Regulatory Guide 1.7.
6.2-79 Rev. 29 WOLF CREEK Zirconium reacts with steam according to the reaction:
Zr + 2H 2 O --> ZrO 2 + 2H 2 The hydrogen gas evolved from this reaction is calculated
to be:
2 lb-mole H 2/lb-mole Zr 0.022 lb-moles H 2 =
91.22 lb Zr/lb-mole Zr lb Zr
The emergency core cooling system (ECCS) is designed to
remove core heat at a rate that will prevent the fuel rods from heating to the point where significant Zr-H 2 O reactions will take place. The LOCA analysis shows that with the passive accumulators and the active elements of
high-pressure and low-pressure safety injection, less
than 0.1 percent of the zirconium cladding will react
with water to generate hydrogen (Section 15.6.5). In
analyzing postaccident hydrogen generation, it has been
conservatively assumed that 5 percent (50 times the
calculated amount) of the total mass of zirconium in
Zircaloy-4 fuel cladding reacts. For the estimated
54,000 pounds of zirconium metal in the active portion of
the core, this amounts to 2,700 pounds of zirconium
reacting. The total hydrogen generation from this source
is then estimated to be 59.20 lb-moles H
- 2. c. Corrosion of Metals and Paints in the Containment
Hydrogen is formed by corrosion of metals in the
containment. The significant portion of this source of
hydrogen is from the corrosion of zinc and aluminum.
Table 6.2.5-3 gives the quantity of each material allowed
in the containment. Figure 6.2.5-9 shows the temperature
used in the corrosion calculation. Table 6.1-3 provides
the qualification information for coating materials used
inside containment.
Zinc in the containment is in two forms: zinc base paint and in galvanized steel. The containment, during the injection phase, is sprayed with a borated solution adjusted a pH between 9.0 and 11.0 with sodium hydroxide addition in operation, while a minimum pH of 4.0 could be experience in one of the spray trains in the event of a single failure in the spray additive subsystem. During the recirculation phase, the pH of the spray is calculated to be greater than 8.5. The corrosion rates for aluminum and zinc in this environment are given in Figures 6.2.5-7 and 6.2.5-8.
Accelerated rates are used for the higher temperatures early in the accident as requested in Regulatory Guide 1.7.
6.2-80 Rev. 13 WOLF CREEK The surface areas for the corrosion of metals and paints
are assumed constant throughout the analysis.
- d. Insignificant Sources of Hydrogen
During normal operation of the plant, hydrogen is
dissolved in the primary system water. The concentration
of hydrogen in primary coolant ranges as shown in Table
5.2-5. The total amount of hydrogen in the primary
system has been calculated to be insignificant.
Table 6.1-10 identifies the quantity of organic
lubricants found inside containment. The quantity of
electrical cable insulation inside the containment is less than 50,000 pounds.
If it is assumed that the above organic materials, excluding coatings (that were already included in the
analyses above), can be considered as unsaturated
hydrocarbons, Reference 12
indicates that they would have a G value for hydrogen of
1 molecule per 100 eV of energy absorbed and a G value
for methane of .01 to .4 molecules per 100 eV of energy
absorbed. The integrated DBA dose that this material
could be subjected to would be <3.0 x 10 7 Rads over a 1-year period following an accident.
Applying these conservative assumptions, approximately
1.7 lb-moles of hydrogen and approximately .7 lb-moles of methane could be potentially released from these
sources over the 1-year period.
This quantity of hydrogen is not considered to be a
significant contribution compared to the sources
identified in Figure 6.2.5-4, and is not included in the
evaluation. Likewise, the small amount of methane that
might be produced is not considered a significant
contributor to combustibility.
The quantities of organic lubricants given in Table 6.1-10 are those quantities subject to be released into
the containment. Due to the environmental qualification
requirements for the cable insulation used inside
6.2-81 Rev. 21 WOLF CREEK containment, it is expected to essentially maintain its mechanical stability and not contribute any debris that might reach the containment sump.
After a LOCA, hydrogen is also generated by noble gas radiolysis. Calculations show that this total amount of hydrogen is insignificant when compared with the sources discussed in a, b, and c above.
6.2.5.2.4 System Operation
6.2.5.2.4.1 Normal Operation
Except for testing and the normal use of the hydrogen mixing subsystem, as
discussed in Section 9.4.6, the system is not normally operated.
6.2.5.2.4.2 Accident Operation 10 CFR 50.44 was revised in 2003 and Revision 3 to Regulatory Guide 1.7 was issued in May 2003. The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates the requirements for hydrogen control systems to mitigate such a release. License Amendment No. 157 was issued by the NRC on January 31, 2005 and deleted the Technical specification requirements for the hydrogen recombiners.
The HCS is normally on standby and is initiated manually from the control room following a LOCA. After a LOCA, sufficient emergency power is available to
handle the load required to operate the electric (thermal) hydrogen
recombiners. Hence, the electric recombiners are turned on when the presence
of Hydrogen is detected (even though they are not required at this early point
in time) in order to keep the hydrogen concentration as low as practicable.
The electric hydrogen recombiner subsystem is to be started when the presence
of hydrogen is detected after a LOCA. However, inadvertent actuation
immediately after a LOCA will not damage the recombiners in any manner, nor
will their capability to perform their design function be impaired. The
electric (thermal) recombiners are completely passive devices. The recombiners
heat the containment hydrogen-air atmosphere that is introduced into the
recombiner to a temperature greater than 1,150°F, causing the recombination of
H2 and O2 to occur. Hence, the hydrogen volume percent is reduced. The air is
then passed to a mixing chamber, in the top of the recombiner, where the hot
air is mixed with the cooler containment air to discharge it back into the
containment at a temperature of approximately 50°F above ambient. Section
6.2.5.3, Safety Evaluation Seven, demonstrates that the recombination rate is
sufficient so that the volume percent of hydrogen is maintained at less than
3.0 volume percent.
The hydrogen purge subsystem is not required at any time unless failure of both
recombiners results in a hydrogen concentration of 3.0 volume percent. In such
a case, the purge subsystem will be manually initiated.
6.2-82 Rev. 29 WOLF CREEK Although not required to function following indication of an accident
condition, each hydrogen mixing fan will be automatically started or switched
from high speed to low speed by an SIS. The hydrogen mixing fans are designed
to withstand the pressure transients associated with a design basis LOCA and remain functional. The initial blowdown, natural convection, containment
coolers and the containment sprays provide mixing of the containment post-LOCA
atmosphere without reliance on the hydrogen mixing fans. The function of the
hydrogen mixing fans including the mixing provisions following a LOCA are
discussed in Sections 6.2.5.3 and 6.2.5.2.2.2. The operation of the
containment sprays and containment coolers is described in Section 6.2.2.2.
The hydrogen monitoring subsystem is normally closed to the containment
atmosphere. Following a LOCA, a CIS-A signal assures that the isolation
valves, located in each sample line penetrating the containment, are closed.
The operator will manually open the isolation valves after a LOCA and initiate hydrogen sampling. Once initiated, hydrogen analyzers provide a continuous
measurement of hydrogen concentration within 30 minutes. Individual valve
control switches are provided in the main control room with a provision for
remote manual bypass as described in Section 7.3.8. Containment atmosphere
samples, maintained in the vapor phase, are brought to the analyzer, which
measures the concentration of hydrogen. From the analyzer, the sample is
returned to the containment atmosphere. The hydrogen analyzer system is
designed with the capability to obtain an accurate sample 30 minutes after
initiation of safety injection.
6.2.5.3 Safety Evaluations Safety evaluations are numbered to correspond to safety design bases.
SAFETY EVALUATION ONE - The safety-related portions of the HCS are located in the reactor, auxiliary, and control buildings. These buildings are designed to
withstand the effects of earthquakes, tornadoes, hurricanes, floods, external
missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of the structural design of
these buildings.
SAFETY EVALUATION TWO - The safety-related portions of the HCS are designed to
remain functional after a SSE. Sections 3.7(B).2 and 3.9(B) provide the design
loading conditions that were considered. Section 3.6 provides a hazards
analysis which assures protection of the HCS and piping following a postulated LOCA or MSLB.
SAFETY EVALUATION THREE - Section 6.2.5.2 demonstrates that the required
redundancy is provided and, as indicated by Table 6.2.5-5, no single failure
can compromise the system's safety functions. All vital power can be supplied
from either onsite or offsite power systems, as described in Chapter 8.0.
6.2-83 Rev. 8 WOLF CREEK SAFETY EVALUATION FOUR - Section 3.2 delineates the quality group
classification and seismic category applicable to the safety-related portion of
this system and supporting systems. Table 6.2.5-1 shows that the components
meet the design and fabrication codes given in Section 3.2. All the power supplies and control functions necessary for safe functioning of the HCS are
Class IE, as described in Chapters 7.0 and 8.0. Comparison of the design to
Regulatory Guide 1.7 positions is provided in Table 6.2.5-6.
SAFETY EVALUATION FIVE - Section 6.2.5.2.1 describes the provisions made to
identify and isolate leakage or malfunction and to isolate the nonsafety-
related portions of the system.
SAFETY EVALUATION SIX - Sections 6.2.4 and 6.2.6 provide the safety evaluation
for the system containment isolation arrangement and testability.
SAFETY EVALUATION SEVEN - Since only one of the two completely separate
recombiner systems is required, a single active or passive failure does not
prevent the recombiners from fulfilling the design function.
Inadvertent actuation of the recombiners immediately after a LOCA will not
damage the recombiners in any manner nor is their capability to perform their
design function be hindered or impaired. Figure 6.2.5-2 shows the hydrogen
volume concentration versus time within the containment as a result of one
recombiner starting 1 day following a LOCA.
Tests have verified that recombination is not a catalytic surface effect, but
that it occurs due to the increased temperature of the process gases (see
Section 6.2.5.4). Since the phenomenon is not a catalytic effect, poisoning of
the unit by fission products or containment spray will not occur. The heater-recombiner section consists of four vertically stacked assemblies of electric heaters. Each assembly contains individual heating elements. Since the
temperature of each assembly results from the contribution of 60 individual
heaters, failure of a few heaters will not affect the efficiency of the
recombiner.
Only the recombiners are located in the containment. All auxiliary equipment
associated with the recombiners is located outside the containment. The
recombiners are designed to withstand, without impairment of function, exposure
to the design temperature and pressure transient in the containment and are
resistant to the chemical and radiation environment of the post-
6.2-84 Rev. 11 WOLF CREEK LOCA containment environment. The auxiliary equipment located in the control
building is designed to withstand, without impairment to function, the exposure
to the post-LOCA control building environment.
The hydrogen generation rate and hydrogen accumulation within the containment, as a function of time, are given in Figures 6.2.5-3 and 6.2.5-4, respectively.
The hydrogen concentration in the containment is given in Figure 6.2.5-5, assuming that no preventive action is taken.
The recombiners are located in the containment so that they process a flow of
containment air containing hydrogen at a concentration which is generally
typical of the average concentration throughout the containment.
The recombiners are located away from the high velocity air streams, such as could emanate from the fan cooler exhaust ports.
SAFETY EVALUATION EIGHT - In the extremely unlikely event that a LOCA occurs
and the redundant recombiners fail to function properly, a purge subsystem may
be utilized to control the hydrogen concentration inside the containment.
Since the purging of any amount of containment atmosphere is undesirable, the
operation of the purge system would be initiated only when it has been
determined that the recombiners are inoperable and only if samples taken from
the containment atmosphere indicate that a hydrogen content of 3.0 volume
percent has been attained.
The concept of purging allows considerable operational flexibility and, in
practice, the specific mode of operation would be determined by the actual
hydrogen generation rate and hydrogen concentration in the containment atmosphere, the amount of airborne activity in the containment, and the prevailing meteorological conditions.
Calculations, assuming no operation of the recombiners, show that the hydrogen
concentration will reach 3 percent at 4 days. A 100-scfm purge initiated at
that time would reduce the hydrogen concentration below the 3-percent level.
The effect of the purge on the hydrogen volume concentration is shown in Figure
6.2.5-6.
6.2-85 Rev. 12 WOLF CREEK SAFETY EVALUATION NINE - Reference 26 provides a complete description of
containment mixing and concludes that sufficient air flow is provided to ensure
proper hydrogen mixing without reliance upon the hydrogen mixing fans following
a LOCA.
The open containment design provides for communication of the areas above and
below the operating deck and the areas inside the secondary shield wall.
Natural convection is a major contribution to the mixing design during all
phases following a LOCA; however, other mechanisms assist the mixing process.
Blowdown and steam release assist the mixing prior to the actuation of the
containment air coolers and containment sprays. During the injection phase and
the initial phases of recirculation from the containment sump, the containment
sprays and air coolers assist the natural circulation mixing of the containment
atmosphere. During the very long term, the natural circulation alone is adequate to prevent localized accumulations of hydrogen from exceeding combustible limits; however, two containment air coolers continue to operate to
enhance mixing and cool the containment.
The hydrogen mixing subsystems as designed enhances the mixing provisions
described in Reference 26 for defense-in-depth when operating. The operating
hydrogen mixing fans increase the rate of air mixing in the containment by
drawing air from the steam generator loop compartments and exhausting it into
the upper containment air space where mixing occurs in the turbulence created
by natural convection and by the operation of the containment sprays and
containment air coolers. Although the hydrogen mixing fans are not required to
operate Post-LOCA, each pair of hydrogen mixing fans is completely redundant, and powered from independent Class 1E power sources. Continued plant operation
is allowed if the fans are out of service since other mixing provisions are
available for adequate mixing. Further discussion of the mixing fans normal operation design bases is provided in Section 9.4.6.
SAFETY EVALUATION TEN - The hydrogen monitoring subsystem is designed to take
air samples from a total of four locations (two for each redundant train)
inside the containment. These samples are analyzed, and the results are
indicated in the control room.
The hydrogen monitor and associated sample lines, located outside the
containment, are considered to be an extension of the containment pressure
boundary, and, therefore, are designed to withstand the pressure, temperature, and humidity transients associated with the design basis LOCA.
6.2-86 Rev. 8 WOLF CREEK SAFETY EVALUATION ELEVEN - The HCS was initially tested with the program given
in Chapter 14.0. Periodic inservice functional testing and inspection are done
in accordance with Section 6.2.5.4.
6.2.5.4 Testing and Inspections
10 CFR 50.44 was revised in 2003 and Revision 3 to Regulatory Guide 1.7 was issued in May 2003. The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates the requirements for hydrogen control systems to mitigate such a release. License Amendment No. 157 was issued by the NRC on January 31, 2005 and deleted the Technical Specification requirements for the hydrogen recombiners. Testing and Inspection of the equipment is no longer performed.
The analytical and test program for the electric recombiner includes proof-of-
principle tests and full-scale prototype tests.
The proof-of-principle tests and prototype tests have been completed, and the
results of these tests were submitted to the NRC in References 14 through 22.
Results of the proof-of-principle tests show that hydrogen-oxygen reaction
occurs at air temperatures of about 1,150°F or above, and no detectable
hydrogen was found in the effluent gases. Tests demonstrated that the
recombination reaction occurred due to increased gas temperature and not to a
catalytic surface effect.
A full-size prototype recombiner was constructed for testing. Hydrogen tests
were conducted by erecting a steel building around the recombiner to permit
simulation of a plant containment environment. During tests, a hydrogen air
mixture was introduced into the building and the recombiner was operated.
Hydrogen-air mixtures ranged from 0.6 to 4.0 volume percent. In all cases, the
recombiner performed satisfactorily and no detectable hydrogen was found in the
effluent gases. Tests were also performed with containment spray containing
sodium tetraborate (2,500 ppm boron as boric acid adjusted to a pH of 10 with
sodium hydroxide) impinging on the recombiner from spray nozzles mounted in the
upper part of the building. In another test, steam was injected into the
simulated containment. In all cases, the recombiner performed satisfactorily.
Tests were conducted to show the effect of air currents on the recombiner.
These tests also indicated satisfactory performance. During the test program, the power, sheath temperatures, and air temperatures in the recombiner, as well
as containment hydrogen concentration and hydrogen concentration at the exit of
the recombiner, were measured. These tests have shown satisfactory performance
for all conditions of interest.
Upon installation, the recombiners are energized and brought up to temperature.
If a recombiner's temperature exceeds 1,150°F, it is considered operable and
capable of performing its design function. No hydrogen is present during the
test, since the proof-of-principle tests and prototype tests indicate that
recombination occurred solely because of the increased temperature.
6.2-87 Rev. 29 WOLF CREEK The performance of the hydrogen gas analyzer is periodically verified by
comparing the response of the thermal conductivity instrument to a known sample
of reference gas.
Nondestructive examination is performed on the components of the hydrogen
monitoring subsystem and the hydrogen purge subsystem. Periodic inservice
testing of all fans, valves, and instrumentation is performed.
6.2.5.5 Instrumentation Requirements 6.2.5.5.1 Hydrogen Recombiner Subsystem
Controls for operation of the hydrogen recombiners are provided in the control room. A manual control station is provided for each train to regulate power to
the heaters in the associated recombiner. The controller maintains the correct
power input to bring the recombiner above the threshold temperature for the
recombination process. The controller setting is adjusted to accommodate
variations in the containment temperature, pressure, and hydrogen concentration
in the post-LOCA environment. The system is designed to conform to the
applicable portions of IEEE 279, 323, 344, and 383 and is powered from a Class
IE source.
Proper recombiner operation is assured by measuring the power input to the heaters from a station outside the containment. The proper air flow through the recombiners is achieved through the use of an orifice plate built into each
unit.
For convenience in testing and conducting periodic checkout of the recombiners, temperature indicators are provided. The temperature indicators are not
required to assure proper operation of the recombiner during post-LOCA
conditions.
6.2.5.5.2 Hydrogen Mixing Provisions
Although not required to assure adequate mixing, the operation of the hydrogen
mixing fans is actuated automatically upon receipt of a safety injection
signal. Control switches and indicator lights for the four hydrogen mixing
fans are located in the control room. The system is designed to conform to the applicable portions of IEEE 279 and 334 and is powered from a Class 1E source.
The function of the hydrogen mixing fans including the mixing provisions
following a LOCA are discussed in Sections 6.2.5.3 and 6.2.5.2.2.2.
6.2.5.5.3 Hydrogen Purge Subsystem
Operation of the hydrogen purge subsystem is manually initiated from the
control room. Instrumentation requirements of the hydrogen purge subsystem are
described in more detail in Section 9.4.3.
6.2-88 Rev. 8 WOLF CREEK The line penetrating the primary reactor containment is provided with power-
operated isolation valves with position indicators and controls in the control
room to allow operator control during post-LOCA operation. A complete
discussion of the isolation valve provisions is presented in Section 6.2.4.
6.2.5.5.4 Hydrogen Monitoring Subsystem
A hydrogen analyzer is provided for periodic sampling of the containment
atmosphere following a design basis event. The hydrogen analyzer has a readout
scale of 0 to 10 percent. The output signal of the analyzer is indicated at
the analyzer mounting location and recorded and alarmed in the main control
room. In addition to the high hydrogen alarm, each analyzer provides
malfunction alarms, including low sample flow, low temperature, and loss of
power. The displays provided are described further in Section 7.5.
6.2.6 CONTAINMENT LEAKAGE TESTING
The reactor containment, containment penetrations, and containment isolation
barriers are designed to permit periodic leakage rate testing as required by 10
CFR 50, Appendix A, General Design Criteria 52, 53, and 54. 10 CFR 50, Appendix J, outlines the containment leakage test requirements and establishes
the acceptance criteria for such tests. The objective of the leakrate testing
is to ensure that the leakage from the containment is within the limits set by
Technical Specifications.
Compliance with 10 CFR 50 Appendix J, Types A, B, and C, testing is discussed
in Sections 6.2.6.1, 6.2.6.2, and 6.2.6.3.
6.2.6.1 Containment Integrated Leakage Rate Test (Type A Test)
The containment was designed with an allowable leakage rate (La) of 0.20
percent (weight percent) of containment free air volume per day for the first
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The maximum leakage rate occurs at the calculated peak containment pressure bounding a LOCA, Pa = 48 psig. The actual leakage rate is determined
by using the methods and requirements of 10 CFR 50, Appendix J, Option B for
Type A tests as administered by the Containment Leakage Rate Testing Program.
The acceptance criteria specified in Appendix J for the Integrated Leakage Rate
Test (ILRT or Type A test) includes a margin for possible deterioration of the
containment leakage integrity during the service intervals between tests. The
as left leakage rate prior to the first startup after performing an ILRT is
required to be less than 0.75 La for the overall Type A leakage rate. At all
other times between tests the acceptance criteria is based on an overall Type A leakage rate limit of less than or equal to 1.0 La.
6.2-89 Rev.15 WOLF CREEK 6.2.6.1.1 ILRT Pretest Requirements
The containment integrated leakage rate test complements local leakage rate
tests. Local leakage rate tests, in which potential leakage paths through the containment boundary, i.e. containment penetrations, are subjected to test
conditions similar to those occurring during the integrated leakage rate test, allow detection and correction of leak paths through the containment without
pressurizing the entire containment structure. These local leakage rate tests
are the Type B and C tests described in Sections 6.2.6.2 and 6.2.6.3.
The significant ILRT and system alignment requirements are as follows.
The visual examination of containment concrete surfaces outside containment and
steel liner plate inside containment to fulfill the requirement of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements and frequency specified by ASME Section XI code IWE/IWL. Except
where relief has been authorized.
Closure of containment isolation valves for Type A testing is accomplished by
normal means and without adjustment. Alternative methods of valve closure may
be used provided they are documented and are equivalent to normal means. If a
leakage path requires isolation to successfully complete an ILRT, its local
leakage rate, measured after the ILRT, is added to the Type A test results.
Normal and accident positions for each containment isolation valve are shown on
Figure 6.2.4-1.
Portions of fluid systems that are part of the reactor coolant pressure
boundary and are open indirectly to the containment atmosphere due to the LOCA
conditions and are, therefore, extensions of the boundary of the containment are opened or vented to the containment atmosphere during the ILRT. Venting the reactor coolant system to the containment atmosphere fulfills this
requirement. Figure 6.2.4-1 contains the applicable GDC or other defined
criteria for the isolation valve arrangements provided.
Pathways that are open to the containment atmosphere under post-LOCA conditions
are vented to the containment atmosphere during the ILRT. Pathways are
considered open to the containment atmosphere if the system fluid is drained or
driven off by the LOCA. This includes portions of pathways inside or outside
containment that penetrate the containment and may rupture inside containment
under a LOCA conditions. Vented pathways are also drained of fluid inside the
containment, between the containment isolation valves and outside the
containment to expose the pathways to post-LOCA differential pressure. For
pathways not vented and drained, their local leakage rate is added to the Type
A test results as directed by the Containment Leakage Rate Testing Program.
In exception to the above, for planning and scheduling purposes, or ALARA
considerations, pathways which are local leakage rate testable may be left
isolated and fluid filled, not vented and drained. The as-found and as-left
leakage rate for all pathways that are not vented and drained will have been
determined within the previous 24 calendar months of the time that the Type A
test is performed. For pathways not vented and drained, their local leakage
rate is added to the Type A test results as directed by the Containment Leakage
Rate Testing Program.
6.2-90 Rev. 20 WOLF CREEK In exception to the above, for pathways in systems which are required for proper conduct of the ILRT, or to maintain the plant in a safe condition during
the ILRT, (e.g. essential service water lines to the containment air coolers)
they may remain operable in their normal mode and need not be vented or
drained. However, if these pathways are potential post-LOCA leakage pathways, their local leakage rate is added to the Type A test results as directed by the
Containment Leakage Rate Testing Program.
Isolation, repair or adjustment to a leakage pathway that may affect the
leakage rate through that pathway is permitted prior to or during the Type A
test provided that the pathways are local leakage rate testable and that the
Type A test results are corrected.
For portions of the pathways outside containment that are designed to Seismic
Category 1 and at least Safety Class 2, are not vented. These systems (e.g.
ECCS and Containment Spray) are normally filled with water and operating under post-accident conditions therefore they are also not drained. The containment
isolation valves in these systems are closed if the associated subsystem is not
operating. Normally, under post-accident conditions, a water seal is present
inside the inner isolation valve during the long-term period. Should
operational leakage exist outside the containment and the isolation valves
leak, the containment sump water level (Elevation 2003'-10") will ensure that
the water in the piping system will provide a water seal on the outside
containment isolation valves which are located at Elevations 2002'-0" and
~1993'. These water seals ensure that containment air does not leak into the
auxiliary building. These systems outside the containment are in their post
accident alignment for the Type A test. These penetrations are identified in
Section 6.2.6.3 and Figure 6.2.4-1.
The steam generator tubes and shell and the associated piping systems passing through the containment liner are considered to be an extension of the containment. Therefore, the secondary side of the steam generator and
connecting systems are not vented to the containment atmosphere. During the
Type A test, the secondary side of the steam generators are vented outside of
the containment to ensure the most conservative test configuration. The
systems associated with the secondary side of the steam generator are
identified in Figure 6.2.4-1.
Pressurized gas lines inside containment are depressurized. Gas pressurized
containment leakage pathways outside containment are isolated during the ILRT
and vented between the outboard containment isolation valve and the pressurized
test boundary isolation valve.
Pressurized components within containment (e.g. reactor coolant drain tank, pressurizer relief tank, accumulator tanks) are vented during the ILRT. This
is done to protect the tanks from the external pressure of the test and to
preclude leakage to or from the tanks which would detract from the accuracy of
the test results.
The containment hydrogen monitors and associated sample lines, located outside
containment, are considered to be extensions of the containment pressure
boundary. For the ILRT, the containment isolation valves are normally open and
the monitors and sample lines outside containment tested.
6.2.6.1.2 ILRT Test Method
Figure 6.2.6-1 shows the general equipment arrangement for a Type A test. For
penetrations which are exempt from Type B or C tests, as noted in Figure 6.2.4-
1, the leakage testing requirement of Appendix J is accomplished by the Type A
test.
6.2-91 Rev. 20 WOLF CREEK The Containment Leakage Rate Testing Program embraces the requirements of 10 CFR 50, Appendix J, Option B.
Documents which provide the methods and requirements for integrated leakage
rate testing and for the Containment Leakage Rate Testing Program include:
- a. 10 CFR, Appendix J, Option B. "Primary Reactor Containment Leakage Testing for water-cooled Power Reactors" (Option B: Performance-
Based Requirements)
- b. Regulatory Guide 1.163, September 1995. "Performance Based Containment Leak Test Program"
- c. Industry Guideline NEI 94-01, revision 0. "Nuclear Energy Institute Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J"
- d. Standard ANSI / ANS 56.8-1994. "Containment System Leakage Testing Requirements"
- e. Standard ANS / N45.4-1972. "Leakage Rate Testing of Containment Structures for Nuclear Reactors"
- f. Bechtel Topical Report BN-TOP-1, revision 1. "Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures
for Nuclear Power Plants"
- g. Amendment No. 152 dated March 17, 2004.
For ILRT performance in general, containment dry bulb temperature, pressure, and dew point temperature are periodically measured and recorded. These data
are analyzed as they are taken so that the containment leakage rate and its
statistical significance is known as the test progresses. Once the leak rate
has been found with sufficient accuracy, a known additional leak is imposed and
the measurements are continued, giving additional verification of the leakage
rate.
6.2.6.2 Containment Penetration Leakage Rate Tests (Type B Tests)
Each of the following containment penetrations are tested with a Type B test.
- a. Personnel access hatches (refer to Section 3.8.2)
- b. Equipment hatch (refer to Section 3.8.2)
- c. Fuel transfer tube (refer to Section 3.8.2)
- d. Electrical penetrations (refer to Section 8.3)
- e. Penetration 34, containment pressurization line
6.2-92 Rev. 18 WOLF CREEK
- f. Penetration 51, ILRT pressurization pressure sensing line
- g. Penetrations 36 and 68, Access hatches for outage
activities.
These penetrations are provided with double seal closures and connections to
allow for pressurization between the seals, except for Eddy Current Cable
Penetrations P-36 and P-68 which are single seal closures used during outage
activities. Each penetration is designed to withstand the calculated peak
containment pressure while maintaining its seal. Personnel hatches have
provisions for test clamps for support of the internal closure during testing.
In addition, Penetrations 34 and 51, containment pressurization line and
pressure sensing lines for the ILRT pressurization system, are also Type B
tests. The test pressure for Type B tests is the calculated peak pressure for the
containment, Pa. The combined leakage rate for all Type B and C tests must be
less than 0.6 L a (maximum allowable leakage rate). The individual leakage rates and testing performed on the Type B penetrations are described in the
Containment Leakage Rate Testing Program.
The test equipment utilized to perform the Type B tests is the same equipment
used for Type C tests. The test equipment is described in Section 6.2.6.3.
The test procedure is the same as the one used for Type C tests.
Type B tests are performed in accordance with Appendix J to 10 CFR 50, with the
following addition: an additional test method may be used. This method
measures the air flow rate to maintain the test volume at a constant pressure.
6.2.6.3 Containment Isolation Valve Leakage Rate Tests (Type C tests)
Figure 6.2.4-1 lists all valves which are associated with the penetrating
piping systems. Figure 6.2.4-1 also indicates the containment isolation valves
which are to be subjected to a Type C test. The following criteria were used to determine which containment isolation valves are local leak tested.
- a. The penetrating system provides a direct connection
between the inside and outside atmospheres of the
containment under normal operation.
- b. The system is isolated by containment isolation valves
which close automatically to effect containment isolation
in response to a CIS signal.
- c. The system is not an engineered safety feature system consisting of a closed piping system outside of the
containment.
6.2-93 Rev. 12 WOLF CREEK The lines serving engineered safety feature systems which consist of closed
piping systems outside the containment have isolation valves which were not
leak tested. All of these lines will initially open or will be opened during
some phase following a LOCA. Valves which are closed initially or closed at some time following a LOCA are positioned to effect proper system operation and
not to effect a barrier against release of containment atmosphere. Should the
valves leak slightly when closed, the fluid seal within the pipe and the closed
piping system outside the containment would preclude release of containment
atmosphere to the environs. Engineered safety features in this classification
penetrate the containment at penetrations numbered P-13, 14, 15, 16, 21, 27, 48, 49, 52, 66, 79, 82, 87, 88 and 89. The containment pressure transmitters
are designed to meet the requirements of Regulatory Guide 1.11 and are
described in Chapter 7.0. These lines have no isolation valves and rely on
closed systems both inside and outside of the containment to preclude the release of the containment atmosphere. These lines penetrate the containment at penetrations 103 and 104. The integrity of these closed systems are
verified during the periodic Type A tests.
As noted in Section 6.2.4.3, the valves associated with the piping systems
connected to the secondary side of the steam generators isolate the steam
generators and are not considered containment isolation valves and are, therefore, not leak tested. All portions of the secondary side of the steam
generators are considered an extension of the containment. These systems
penetrate the containment shell at penetrations P-1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 83, 84, 85, and 86. As shown on Figure 6.2.4-2 the water level in
all steam generators are maintained above the tubes following a LOCA to
preclude the entrance of containment atmosphere into the secondary side of the
steam generators. This requirement has been included in the Emergency
Operating Instructions.
The test equipment to be used during the Type C tests will consist of a
connection to an air supply source, a pressure regulator, a pressure gauge, a
flow indicator, and associated valving.
Isolation valves are positioned to their post-accident position by the normal
method with no accompanying adjustments. Fluid systems are properly drained
and vented with the valves aligned to provide a test volume and atmospheric air
back pressure on the isolation valve(s) being tested.
6.2-94 Rev. 13 WOLF CREEK The test volume is pressurized to the test pressure Pa, as specified in the
Technical Specifications. The pressure regulator(s) maintain the test volume
at a minimum of Pa. The air flow rate into the test volume is recorded, as is
the pressure reading. These records are utilized to determine the leakage rate in cubic centimeters per minute.
For larger test volumes, a pressure decay method may be utilized to determine
the leakage rate.
The total leakage rate for Type B and C tests must be less than 0.6 L
- a.
The criteria for determining the direction in which the test pressure is
applied to the isolation valves are as follows:
Gate Valves Parallel disc a. Test in the DBA direction.
- b. Testing can be performed between the discs if a test connection or
drain is provided in the valve
design.
Flexible wedge a. Test in the DBA direction.
- b. Testing can be performed between
the wedge sections if a test
connection or drain is provided
in the valve design.
Solid wedge a. Test in the DBA direction.
Globe Valves If the DBA flow direction is over the disc (flow to close), the valve may be tested in the reverse direction.
However, if the DBA flow direction is
under the disc (flow to open), then the valve must be tested in this direction.
Butterfly Valves Test in either direction.
Flanges Test in either direction.
Testing of the isolation valves in the nonaccident pressure direction, as
allowed above, is as conservative or more conservative than testing them in the
accident pressure direction.
6.2-95 Rev. 12 WOLF CREEK 6.2.6.4 Scheduling and Reporting of Periodic Tests
Type A, B, and C tests are conducted at the intervals specified in the
Containment Leakage Rate Testing Program.
Test results are retained for internal and external review.
The preoperational test report contains a schematic of the leak measuring
system, instrumentation used, supplemental test method, test program, and
analysis and interpretation of the leakage test data for the Type A test.
6.2.6.5 Special Testing Requirements WCGS does not have a subatmospheric containment or a secondary containment, hence there are no special testing requirements beyond those delineated in
Sections 6.2.6.1 through 6.2.6.4.
6.
2.7 REFERENCES
- 1. Bechtel Power Corporation, "Performance and Sizing of Dry
Pressure Containments," Topical Report No. BN-TOP-3, (Rev.
4), October 1977.
- 2. "Pipe Breaks for the LOCA Analysis of the Westinghouse
Primary Coolant Loop," WCAP-8082-P-A (Proprietary) and WCAP-
8172-A (Non-Proprietary), January 1975.
- 3. Shepard, R.M., et al, "Westinghouse Mass and Energy Release
Data for Containment Design," WCAP-8264-P-A, Rev. 1, (Proprietary) and WCAP-8312-A, Rev. 2 (Non-Proprietary),
August 1975.
- 4. Bechtel Power Corporation "COPDA, Compartment Pressure Design
Analysis," (Bechtel Computer Code), 1973.
- 5. Bechtel Power Corporation, "Subcompartment Pressure and
Temperature Transient Analysis," Topical Report No. BN-TOP-4, (Rev. 1), October 1977.
- 6. Land, R. E., "Mass and Energy Releases Following a Steam Line
Rupture," WCAP-8822 (Proprietary) and WCAP-8860 (Non-Proprietary), September 1976.
- 7. WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
- 8. NUREG-0588, "Interim Staff Position on Environmental
Qualification of Safety-Related Electrical Equipment,"
December 1979.
6.2-96 Rev. 10 WOLF CREEK
- 9. Deleted
- 10. Topical Report AAF-TR-7101, "Design and Testing of Fan Cooler-Filter Systems for Nuclear Applications"; February 20, 1972; American Air Filter Co., Inc.; Louisville, KY.
- 11. Topical Report OCF-1, "Nuclear Containment Insulation System," August 1977, Owens-Corning Fiberglas Corporation, Lenexa, KS.
- 12. WAPD-PT 24, "Fission Product Decay Energy" (December 1961).
- 13. TID 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," J. J. DiNunno, F. D. Anderson, R. E.
Baker, R. L. Waterfield; March 23, 1962; Division of Licensing and Regulation, USAEC, Washington, D. C.
- 14. Wilson, J. F., "Electrical Hydrogen Recombiner for Water Reactor Containments," WCAP-7709-L (Proprietary), July 1971, and WCAP-7820 (Non-Proprietary) December 1971.
- 15. Wilson, J. F., "Electric Hydrogen Recombiner for PWR Containments - Final Development Report," WCAP-7709-L, Supplement 1 (Proprietary), April 1972, and WCAP-7820, Supplement 1 (Non-Proprietary), May 1972.
- 16. Wilson, J. F., "Electric Hydrogen Recombiner for PWR Containments - Equipment Qualification Report," WCAP-7709-L, Supplement 2 (Proprietary), September 1973, and WCAP-7820, Supplement 2 (Non-Proprietary), October 1973.
- 17. Wilson, J. F., "Electric Hydrogen Recombiner for PWR Containments - Long-Term Tests," WCAP-7709-L, Supplement 3 (Proprietary), January 1974, and WCAP-7820, Supplement 3 (Non-Proprietary), February 1974.
- 18. Wilson, J. F., "Electric Hydrogen Recombiner for PWR Containments," WCAP-7709-L, Supplement 4 (Proprietary), April 1974, and WCAP-7820, Supplement 4 (Non-Proprietary), May
1974.
- 19. Wilson, J. F., "Electric Hydrogen Recombiner Special Tests,"
WCAP-7709-L, Supplement 5 (Proprietary) and WCAP-7820, Supplement 5 (Non-Proprietary), December 1975.
- 20. Wilson, J. F., "Electric Hydrogen Recombiner IEEE 323-1974 Qualification," WCAP-7709-L, Supplement 6 (Proprietary) and WCAP-7820, Supplement 6 (Non-Proprietary), October 1976.
6.2-97 Rev. 29 WOLF CREEK
Supplemental Test Number 2," WCAP-7709-L, Supplement 7
(Proprietary), August 1972, and WCAP-7820, Supplement 7 (Non-
Proprietary), October 1977.
- 22. Wilson, J. F., "Electric Hydrogen Recombiner Qualification
Testing for Model B," WCAP-9346, July 1978.
- 23. Kircher, J. R. and Bowman, R. E., "Effects of Radiation on
Materials and Components," 1964.
- 24. D.W. Hargroves et. al., "CONTEMPT-LT/28-A Computer Program
for Predicting Containment Pressure-Temperature Response to a
Loss-of-Coolant Accident", NUREG/CR-0255, March, 1979.
- 25. Nuclear Regulatory Commission, Docket No. 50-482, "Wolf Creek
Generating Station - Amendment No. 50 to Facility Operating
License No. NPF-42 (TAC No. 80714)" Relating to Tech. Spec.
section 4.6.2.3, "Containment Cooling System," November 4,
1991.
- 26. WCAP-8264-P-A, Rev. 1, "Westinghouse LOCA Mass and Energy
Release Model for Containment Design - March 1979 Version,"
April 1979.
- 27. SLNRC 85-8, dated February 19, 1985; SNUPPS letter to the
NRC, Docket No. STN 50-482, Wolf Creek Technical
Specifications including Justification for deletion of
Technical Specification 3/4.6.4.3, Hydrogen Mixing Fans, February 1985.
- 28. NAI 8907-02, Revision 17, "GOTHIC Containment Analysis Package User Manual," Version 7.2a(QA), January 2006
- 29. NAI 8907-06, Revision 16, "GOTHIC Containment Analysis Package Technical Manual," Version 7.2a(QA), January 2006
- 30. NRC Letter from Anthony C. McMurtray (NRC) to Thomnas Coutu (NMC), Enclosure 2, Safety Evaluation, September 29, 2003.
6.2-98 Rev. 22 WOLF CREEK TABLE 6.2.1-1 SPECTRUM OF POSTULATED LOSS-OF-COOLANT ACCIDENTS
- 1. Double-ended pump suction guillotine (DEPSG) break, with minimum safety injection.
- 2. DEPSG with maximum safety injection.
- 3. 0.6 DEPSG - with maximum safety injection.
- 4. 3 ft 2 pump suction split with maximum safety injection.
- 5. Double-ended hot leg guillotine break with maximum safety
injection.
- 6. ouble-ended cold leg guillotine break with maximum safety
injection
Rev. 22 WOLF CREEK TABLE 6.2.1-2 PRINCIPAL CONTAINMENT DESIGN PARAMETERS
Containment design internal pressure 60 psig
Containment peak calculated internal pressure
LOCA 47.8 psig MSLB 52.9 psig Containment design external pressure load 3.0 psid
Containment calculated external pressure 2.72 psid
Containment design temperature 320°F Containment peak calculated vapor temperature
LOCA 307.2°F MSLB 364.9°F Peak calculated equipment temperature-MSLB See Figure 6.2.1-85
Internal dimensions
Cylindrical wall diameter 140 ft
Cylindrical wall height 135 ft
Curved dome height above spring line 70 ft
Volume Net free internal volume 2.50x10 6 ft 3 Containment design leak rate
First 24 hrs 0.20 percent free
vol/day
After 1 day 0.10 percent free
vol/day
Containment
Internal Compartments:
Reactor cavity design pressure See Table 6.2.1-17
Reactor cavity calculated
pressure See Table 6.2.1-17
Steam generator loop compartment
design pressure See Table 6.2.1-22
Steam generator loop compartment
calculated pressure See Table 6.2.1-22
Pressurizer vault design pressure See Table 6.2.1-26
Pressurizer vault calculated
pressure w/surge line break See Table 6.2.1-26
Pressurizer surge line compartment
design pressure See Table 6.2.1-26
Pressurizer surge line compartment
calculated pressure w/surge
line break See Table 6.2.1-26
Rev. 29 WOLF CREEK TABLE 6.2.1-3 ENGINEERED SAFETY FEATURES DESIGN PARAMETERS USED AS INPUT TO THE CONTAINMENT ANALYSIS Full Minimum Capacity Capacity ECCS Passive safety injection system Number of accumulators 4 4 Nominal Pressure, psig 602-648 602-648
Nominal Liquid volume, ft3 850/accumulator 850/accumulator
Active safety injection systems High-pressure system injection Number of lines 4 4
Number of centrifugal charging 2 1 pumps Intermediate pressure safety injection Number of lines 4 4
Number of safety injection pumps 2 1
Low-pressure safety injection
Number of lines 4 4
Number of RHR pumps 2 l Total injection flow rate, lbm/sec 1,401 586
Total recirculation flow rate, gpm 9,600 4,800
Containment heat removal systems
Containment spray system
Number of lines 2 1 Rev. 12 WOLF CREEK TABLE 6.2.1-3 (Sheet 2)
Full Minimum
Capacity Capacity
Number of pumps 2 1
Number of headers 2 1
Injection flow rate, gpm *2932/pump *2932/pump
Recirculation flow rate, gpm 3434/pump 3434/pump
Containment air coolers
Number of units 4 2
Duty per cooler (See Figure 6.2.1-15)**
Air-side flow rate, acfm 69,400 69,400 through fan during DBA conditions.
RHR Heat Exchangers
Type Shell and U-type
Number 2 1
Effective heat transfer
coefficient times heat
transfer area, Btu/hr F 2.3 x 10 6/unit 2.3 x 10 6/unit Primary side flow through
RHR heat exchanger, lb/hr 2.34 x 10 6/unit 2.34 x 10 6/unit Secondary side flow
through RHR heat
exchangers, lb/hr 3.8 x 10 6/unit 3.8 x 10 6/unit Source of cooling water Component cooling water
Flow begin, sec, minimum 849 1,509
Component cooling Water Heat Exchangers
Type Shell and straight tube
- For containment integrity analyses, an injection flow rate of 2932 gpm/pump, representing a 5% degradation of the delivered flow rate of 3086 gpm/pump, is
used.
- The limiting case (DEPSG break with minimum safety injection) modeled the containment air cooler duty curve given in Table 6.2.1-57C.
Rev. 29 WOLF CREEK TABLE 6.2.1-3 (Sheet 3)
Full Minimum Capacity Capacity Number 2 1 Effective heat transfer coefficient times heat transfer area, Btu/hr
°F 5.86 x 10 8/unit 5.86 x 1O 8/unit Primary side flow through CCW heat exchangers, lb/hr 3.8 x 10 6/unit 3.8 x 10 6/unit Secondary side flow through
CCW heat exchangers, lb/hr (1) 3.68 x 10 6/unit 3.68 x 10 6/unit Source of cooling water Essential Service Water Temperature of cooling 95 95 water, max, °F (1) The essential service water flow to the CCW heat exchanger(s) may be reduced to as low as 3.58X10 6/unit in order to ensure that the design bases cooling water flows are provided to the remaining ESW system components.
Rev. 4 WOLFCREEKTABLE6.2.1-4CONTAINMENTPASSIVEHEATSINKPARAMETERSThermophysicalPropertiesVolumetricHeatThermalCapacityConductivityMaterialBtu/ft 3-FBtu/hrftFEpoxypaint49.90.97Inorganic21.70.63zincpaint Stainlesssteel53.98.40Carbon54.328.35 steelConcrete30.030.80 Zinccoating40.964.8 Air0.01450.017 ValueHeatTransferCoefficientContainmentatmosphere"ModifiedTagami"toheatsinksurfacesContainmentatmospheretocontainmentsumpwater0Containmentsumpwatertocontainmentfloor0Linergapconductance20Btu/hr-ft 2-FContainmentwallstooutsideatmosphere2.0Btu/hr-ft 2-FPassiveHeatSinkDescriptionContainmentwalls GeometrySlabSurfacearea,ft 2 58807Composition,ftEpoxypaint0.00033Inorganiczincpaint0.00033Rev.0 WOLFCREEKTABLE6.2.1-4(Sheet2)Carbonsteel0.02083Airgap0.00085 Concrete4.00000Boundaryconditions-Linerplateexposedtocontainmentatmosphere;outsideexposedtotheoutsideatmosphereContainmentDomeGeometrySlabSurfacearea,ft 2 30806Composition,ftEpoxypaint0.00033Inorganiczincpaint0.00033 Carbonsteel0.02083 Airgap0.00085 Concrete3.00000Boundaryconditions-Linerplateexposedtocontainmentatmosphere;outsideexposedtotheoutsideatmosphereUnlinedConcreteGeometrySlabSurfacearea,ft 2 65831Composition,ftConcrete1.72000Boundaryconditions-Onesideexposedtocontainmentatmosphere;theothersideinsulated.StainlessSteelLinedConcreteGeometrySlabSurfacearea,ft 2 7197Composition,ftStainlesssteel0.02083Airgap0.00085 Concrete2.00000Boundaryconditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.Rev.0 WOLFCREEKTABLE6.2.1-4(Sheet3)GalvanizedSteelLinedConcreteGeometrySlabSurfacearea,ft 2 6679Composition,ftZinccoating0.00011Carbonsteel0.00529 Airgap0.00085 Concrete1.34300BoundaryConditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.StainlessSteelGeometrySlabSurfacearea,ft 2 18648Composition,ftStainlesssteel0.01792Boundaryconditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.GalvanizedSteelGeometrySlabSurfacearea,ft 2 68451Composition,ftZinccoating0.00011Carbonsteel0.00783Boundaryconditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.Carbonsteel-unpainted GeometrySurfacearea,ft 2 1769Composition,ftCarbonsteel.0208BoundaryConditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.Rev.0 WOLFCREEKTABLE6.2.1-4(Sheet4)CarbonSteel-PaintedGeometrySlabSurfacearea,ft 2 13450Composition,ftEpoxypaint0.00033Inorganiczincpaint0.00033 Carbonsteel(00.125inthick)0.00696Surfacearea,ft 2 84088Composition,ftEpoxypaint0.00033Inorganiczincpaint0.00033 Carbonsteel(0.1250.25in.thick)0.01667Surfacearea,ft 2 40471Composition,ftEpoxypaint0.00033 Inorganiczincpaint0.00033 Carbonsteel(0.250.5in.thick)0.02817Surfacearea,ft 2 24306Composition,ftEpoxypaint0.00033Inorganiczincpaint0.00033 Carbonsteel(0.51.0in.thick)0.05900Surfacearea,ft 2 11932Composition,ftEpoxypaint0.00033 Inorganiczincpaint0.00033 Carbonsteel(0.52.5in.thick)0.11192Surfacearea,ft 2 7804Composition,ftEpoxyPaint0.00033 Inorganiczincpaint0.00033 Carbonsteel(>2.5in.thick)0.27892Boundaryconditions-Onesideexposedtocontainmentatmosphere,theothersideinsulated.Rev.0 WOLF CREEK TABLE 6.2.1-5 CONTAINMENT AND REACTOR COOLANT SYSTEM INITIAL CONDITIONS FOR CONTAINMENT ANALYSIS Reactor coolant system (at overpower of 102-percent engineered
safeguards ratings)
Reactor core power level, MWt 3636
Average coolant temperature, °F 595.0
Mass of reactor coolant, lb 504.52 x 10 3
Reactor coolant energy, Btu(1) 304.90 x lO 6
pressure, psia
Containment
Free volume, ft 3 2.5 x 10 6
Pressure, psia 14.7
Atmosphere temperature, °F 130 (2)
Outside atmosphere temperature, °F 120 (3)
Relative humidity, percent 50
Stored water
Refueling water storage tank, gal 394,000 (4)
Refueling water temperature, °F 100
Essential service water temperature, °F 95
Accumulators (4) capacity, lbs 210,300
(1) All energies relative to 32 °F.
(2) The limiting cases (DEPSG break with minimum safety injection for the LOCA analyses and Case 1, Case 10, Case 13, and Case 16 for the MSLB analyses) modeled a temperature of 1309 °F. the remaining analyses modeled a temperature of 120 °F.
(3) In order to address a CONTEMPT code limitation, the limiting LOcA analysis (DEPSG break with minimum safety injection), the outside air temperature was conservatively modeled as 130 °F.
(4) This is the minimum volume maintained in the RWST.
Rev. 29 WOLF CREEK TABLE 6.2.1-6 CHRONOLOGY OF EVENTS DEPSG BREAK W/MIN SI Event Time (Sec)
Break occurs 0
Peak containment pressure during
blowdown (41.2 psig) 18.0
Primary system blowdown complete 20.86
Accumulator injection begins 20.86
Charging pump injection begins 20.86
Safety injection pump injection begins 20.86
RHR pump injection begins 20.86
Containment spray injection begins 60.0 Peak temperature reached (307.2 o F) 60.0 Containment fan coolers begin removing heat 70.0*
Accumulators empty 92.5
Peak containment pressure reached(47.8 psig) 130.0 Reflood complete 130.7
ECCS recirculation 1509.0
Containment spray recirculation 3227.0
End of steam generator energy release 3775.0
Containment pressure less than 50 percent
of peak calculated pressure 5500.0
End of analysis 10 6
- An analysis performed for Generic Letter 96-06 indicates that the
water side of the containment coolers may not completely fill and
pressurize until 65 seconds after a design basis event coincident
with a loss of offsite power. The original calculation of
containment peak temperature and pressure assumed that the
containment coolers would begin removing heat from the containment
at the design rate at 60 seconds after the event. The containment
pressure/temperature profile was recalculated using 70 seconds as
the assumption for when the containment coolers would begin
removing heat from the containment environment. There was
negligible effect on the pressure/temperature profile as a result
of the additional 10 second delay.
Rev. 29 WOLF CREEK TABLE 6.2.1-7 CHRONOLOGY OF EVENTS DEPSG BREAK W/MAX SI Event Time (Sec)
Break occurs 0 Peak containment pressure during
blowdown of (41.2) psig 18.0 Primary system blowdown complete 20.86 Accumulator injection begins 20.86
Charging pump injection begins 20.86
Safety injection begins 20.86
RHR pump injection begins 20.86 Containment fan coolers begin to remove heat 60.0*
Containment spray injection begins 60.0
Peak temperature reached (301.7 o F) 60.0 Accumulators empty 92.1
Peak containment pressure reached (45.9 psig) 120.0
Reflood complete 124.0
ECCS recirculation 849.0
Containment pressure less than 50 percent of peak calculated pressure 1050.0 Containment spray recirculation 1667.0 End of steam generator energy release 3772.0
End of analysis 10 6*A calculation of the peak containment pressure and temperature was performed for the limiting minimum S.I. case using 70 seconds
for the assumption of when the containment coolers begin removing
heat for the containment (see Table 6.2.1-6). There was
negligible change in the peak containment pressure and
temperature. Therefore, the calculation for the less limiting
case of DEPSG Break with Maximum S.I. was not revised.
Rev. 15 WOLF CREEK TABLE 6.2.1-8 COMPARATIVE RESULTS:
SUMMARY
OF RESULTS OF CONTAINMENT PRESSURE AND TEMPERATURE ANALYSIS FOR THE SPECTRUM OF POSTULATED ACCIDENTS Accident 1 2 3 4 5 6 Break location Pump suction PS PS PS Hot Leg Pump Discharge (PS)
Break type Double-ended DEG 0.6 DEG Split DEG DEG
guillotine
(DEG)
Break size 10.24 ft 2 10.24 ft 2 6.14 ft 2 3 ft 2 9.18 ft 2 8.25 ft 2 Safety injection min max max max max max
Containment sprays min max max max max max
Containment fan coolers min min min min min min Peak pressure, psig 47.8 45.9 45.4 46.0 41.7 38.4 Time to peak pressure, sec 130.0 120.0 115.0 140.0 17.0 15.0 Peak temperature, F 307.2 301.7 302.7 287.0 265.8 274.1 Time to peak temperature, sec 60.0 60.0 60.0 60.0 17.0 60.0 Energy released to containment
at time of peak pressure, 10 6 Btu 449.28 452.9 445.74 435.21 324.5 309.9 Energy absorbed by passive heat sinks
at time of peak pressure, 10 6 Btu 81.0 80.9 78.0 83.4 25.6 22.2 Energy in vapor region at
time of peak pressure, 10 6 Btu 310.8 297.2 294.15 297.8 274.28 254.25 Energy in sump water at
time to peak pressure, x10 6 Btu 82.1 92.5 90.94 97.87 50.52 54.1 Energy removed by containment
fan coolers up to the time of
peak pressure, x10 6 Btu 1.99 1.27 1.26 1.94 0.0 0.0 Energy removed by containment sprays
up to time of peak pressure, x10 6 Btu 3.60 6.23 6.19 9.57 0.0 0.0 Rev. 29 WOLF CREEK TABLE 6.2.1-9 CONTAINMENT MASS AND ENERGY BALANCE DEPSG BREAK WITH MINIMUM SAFETY INJECTION
ENERGY BALANCE (X 10 6 Btu)
Blowdown End of End of Peak Beginning of 1 Day into
Initial Peak Blowdown Reflood Pressure Recirculation Recirculation
0 sec 18 sec 21 sec 130 sec 130 sec 1509 sec 87,000 sec
Containment atmosphere 24.1 274.3 271.3 310.8 310.8 262.7 45.2 Containment sump 0.0 51.2 54.5 82.1 82.1 441.9 376.2 Heat sinks 0.0 19.8 25.5 81.0 81.0 168.3 181.5 Reactor vessel 0.0 0.0 0.0 0.0 0.0 42.5 22.9
Total Energy Remaining 24.1 345.3 351.3 473.9 473.9 915.4 625.8
Initial energy 24.1 24.1 24.1 24.1 24.1 66.6 66.6 Energy added from
primary system 0.0 321.0 327.4 448.6 448.6 842.8 3376.5
Energy added by
sprays 0.0 0.0 0.0 1.9 1.9 39.7 87.2
Energy added by
N 2 discharge 0.0 0.0 0.0 0.2 0.2 0.2 0.2 Heat removed by RHR 0.0 0.0 0.0 0.0 0.0 0.0 2003.0
Heat removed by air
coolers 0.0 0.0 0.0 2.0 2.0 43.1 1094.1 Heat removed to out-
side atmosphere
via heat sinks 0.0 ~0.0 ~0.0 ~0.0 ~0.0 ~0.0 0.2
Total Energy Remaining 24.1 345.1 351.5 472.8 472.8 906.2 499.7 Rev. 29 WOLF CREEK
TABLE 6.2.1-9 (Sheet 2)
MASS BALANCE (X 10 3 lbm)
Blowdown End of End of Peak Beginning of 1 Day into
Initial Peak Blowdown Reflood Pressure Recirculation Recirculation
0 sec 18 sec 21 sec 130 sec 130 sec 1509 sec 87,000 sec
Containment atmosphere 7.9 233.5 230.9 265.8 265.8 222.5 26.8 Containment sump 0.0 271.2 284.9 455.0 455.0 1864.8 2694.1 Reactor coolant 0.0 0.0 0.0 0.0 0.0 114.0 180.5
Total Mass 7.9 504.7 515.8 720.8 720.8 2201.3 2901.4
Initial mass 7.9 7.9 7.9 7.9 7.9 121.9 121.9 Mass released from
primary system
during injection
phase 0.0 496.5 507.9 684.7 684.7 1492.9 1492.9
Mass added by sprays
during injection
phase 0.0 0.0 0.0 28.4 28.4 583.2 1283.0
Total Mass 7.9 504.4 515.8 721.0 721.0 2198.0 2897.8
Rev. 29 WOLF CREEK TABLE 6.2.1-10 CONTAINMENT MASS AND ENERGY BALANCE DEPSG BREAK WITH MAXIMUM SAFETY INJECTION ENERGY BALANCE (X 10 6 Btu) Blowdown End of Peak End of Beginning of 1 Day into Initial Peak Blowdown Pressure Reflood Recirculation Recirculation
0 sec 18 sec 21 sec 120 sec 124 sec 849 sec 87,000 sec Containment atmosphere 22.5 271.4 268.0 297.2 296.3 205.0 44.9
Containment sump 0.0 51.4 54.7 92.5 94.4 417.0 367.6
Heat sinks 0.0 20.9 26.9 80.9 81.9 136.7 189.9
Reactor vessel 0.0 0.0 0.0 0.0 0.0 41.2 17.8 Total Energy Remaining 22.5 343.7 349.6 470.6 472.6 799.9 620.2
Initial energy 22.5 22.5 22.5 22.5 22.5 63.7 63.7 Energy added from
primary system 0.0 321.0 327.4 444.6 450.3 727.0 3283.1
Energy added by
sprays 0.0 0.0 0.0 3.3 3.5 43.5 88.5
Energy added by
N 2 discharge 0.0 0.0 0.0 0.2 0.2 0.2 0.2 Heat removed by RHR 0.0 0.0 0.0 0.0 0.0 0.0 2204.3
Heat removed by air
coolers 0.0 0.0 0.0 1.3 1.4 18.1 574.9
Heat removed to out-
side atmosphere
via heat sinks 0.0 ~0.0 ~0.0 ~0.0 ~0.0 0.1 9.8 Total Energy Remaining 22.5 343.5 350.0 469.3 475.1 816.2 646.5 Rev. 6 WOLF CREEK TABLE 6.2.1-10 (Sheet 2)
MASS BALANCE (X 10 3 lbm) Blowdown End of Peak End of Beginning of 1 Day into Initial Peak Blowdown Pressure Reflood Recirculation Recirculation
0 sec 18 sec 21 sec 120 sec 124 sec 849 sec 87,000 sec Containment atmosphere 6.2 235.3 227.4 253.3 252.4 171.6 28.4
Containment sump 0.0 272.4 286.6 538.1 547.2 2174.8 2986.6
Reactor coolant 0.0 0.0 0.0 0.0 0.0 123.2 182.1 Total Mass 6.2 502.7 514.0 790.9 799.6 2469.6 3197.1 Initial mass 6.2 6.2 6.2 6.2 6.2 129.4 129.4 Mass released from
primary system
during injection
phase 0.0 496.5 507.9 746.9 756.9 1772.2 1772.2
Mass added by sprays
during injection
phase 0.0 0.0 0.0 48.6 51.8 639.1 1301.7 Total Mass 6.2 502.7 514.1 801.7 814.9 2540.7 3203.3 Rev. 6 WOLFCREEKTABLE6.2.1-11ADDITIONALMASSANDENERGYRELEASE-LOCAAccumulatorNitrogenReleaseFollowingAccumulatorEmptyTimeMassTemp(sec)(Lbs/sec)(F)000 Accumulatorempty00 Accumulatorempty937470Accumulatorempty+5937470Accumulatorempty+500 10 6 00Rev.0 WOLFCREEKTABLE6.2.1-12REACTORCAVITYCOLDLEG150SQUAREINCHBREAKBREAKMASSFLOWANDENERGYFLOWTimeMassFlowEnergyFlowAverageEnthalpy (sec)(lb/sec)(Btu/sec)(Btu/lb)0.000000.0.0.000.002511.2479195E+047.0036421E+06561.23 0.005001.6637344E+049.3378033E+06561.260.007511.9176510E+041.0762552E+07561.240.010042.1386323E+041.2003236E+07561.00 0.012532.2520092E+041.2626109E+07560.66 0.015022.2352620E+041.2517510E+07560.00 0.017512.4950199E+041.3983657E+07560.460.020052.5459121E+041.4257657E+07560.020.022522.4866287E+041.3909298E+07559.36 0.025012.4633466E+041.3769194E+07558.96 0.027542.5014356E+041.3978759E+07558.83 0.030082.5119803E+041.4032344E+07558.62 0.032572.5524819E+041.4258325E+07558.61 0.035112.6140547E+041.4604555E+07558.69 0.037612.6562706E+041.4841209E+07558.720.040082.6938656E+041.5052055E+07558.75 0.042532.7292354E+041.5250487E+07558.78 0.045012.7388755E+041.5301809E+07558.69 0.047562.7236107E+041.5211357E+07558.50 0.050052.7085068E+041.5122646E+07558.34 0.052572.6940928E+041.5038569E+07558.21 0.055132.6836674E+041.4977666E+07558.10 0.057602.6764042E+041.4935072E+07558.03 0.060062.6614323E+041.4848672E+07557.92 0.062532.6285228E+041.4660363E+07557.74 0.065082.5851165E+041.4412680E+07557.53 0.067502.5511843E+041.4219705E+07557.38 0.070032.5402939E+041.4158179E+07557.34 0.072542.5559335E+041.4247626E+07557.43 0.075062.5806312E+041.4388536E+07557.56 0.077502.5940979E+041.4465034E+07557.610.080052.5870657E+041.4424514E+07557.56 0.082532.5607302E+041.4274260E+07557.43 0.085092.5192299E+041.4038094E+07557.24 0.087522.4717011E+041.3768031E+07557.03 0.090072.4223797E+041.3488405E+07556.82 0.092542.3793468E+041.3244805E+07556.66 0.095062.3440760E+041.3045481E+07556.53 0.097562.3189290E+041.2903588E+07556.45 0.100112.3013785E+041.2804787E+07556.40Rev.0 WOLFCREEKTABLE6.2.1-12(Sheet2)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb)0.105142.2833488E+041.2703935E+07556.370.110032.2965013E+041.2779727E+07556.49 0.115042.3456540E+041.3058989E+07556.730.120132.3888539E+041.3304130E+07556.930.125162.3987351E+041.3359886E+07556.96 0.130012.3910924E+041.3315992E+07556.90 0.135072.3636693E+041.3160000E+07556.76 0.140132.3175308E+041.2898561E+07556.560.145032.2889044E+041.2736857E+07556.460.145022.2822906E+041.2699974E+07556.46 0.155022.2774884E+041.2673177E+07556.45 0.160002.2529730E+041.2534576E+07556.36 0.165092.2238877E+041.2370534E+07556.26 0.170122.2140066E+041.2315260E+07556.24 0.175082.2176390E+041.2336414E+07556.29 0.180132.2268559E+041.2389108E+07556.350.185082.2406351E+041.2467322E+07556.42 0.190062.2609861E+041.2582689E+07556.51 0.195072.2289889E+041.2712888E+07556.61 0.200102.2992641E+041.2799181E+07556.66 0.212512.3099985E+041.2859073E+07556.67 0.225152.2766403E+041.2669465E+07556.50 0.237532.2665482E+041.2612627E+07556.47 0.250082.2917330E+041.2755851E+07556.60 0.262602.2788777E+041.2685208E+07556.52 0.275092.2335385E+041.2425851E+07556.33 0.287572.2347390E+041.2433610E+07556.38 0.300012.2887445E+041.2739650E+07556.62 0.312652.2969350E+041.2750924E+07556.58 0.325072.2697713E+041.2630559E+07556.47 0.337592.2741599E+041.2655926E+07556.51 0.350222.2756910E+041.2665639E+07556.510.362602.2730072E+041.2648997E+07556.49 0.375132.2668840E+041.2614348E+07556.46 0.387502.2657310E+041.2607927E+07556.46 0.400292.2719561E+041.2643147E+07556.49 0.412572.2702061E+041.2633128E+07556.47 0.425242.2816315E+041.2697746E+07556.52 0.437622.2857841E+041.2721116E+07556.53 0.450102.2765073E+041.2668196E+07556.48 0.462562.2711434E+041.2637794E+07556.45 0.475852.2739781E+041.2654050E+07556.47 0.487652.2789504E+041.2682156E+07556.49 0.500022.2892916E+041.2739682E+07556.54 0.525062.2783476E+041.2678118E+07556.46 0.550032.2749255E+041.2659461E+07556.48 0.575142.2873354E+041.2729382E+07556.52 0.600182.2801817E+041.2688886E+07556.49 0.625182.2913270E+041.2752101E+07556.54Rev.0 WOLFCREEKTABLE6.2.1-12(Sheet3)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb)0.650072.2870850E+041.2727855E+07556.570.675132.2860966E+041.2722588E+07556.52 0.700392.2933255E+041.2763494E+07556.550.725012.2903970E+041.2746812E+07556.530.750052.2969196E+041.2783878E+07556.57 0.775122.2934406E+041.2764040E+07556.55 0.800052.2951385E+041.2773815E+07556.56 0.825052.2974247E+041.2786780E+07556.570.850012.2994030E+041.2797983E+07556.580.875102.2988790E+041.2794993E+07556.58 0.900022.2996781E+041.2799550E+07556.58 0.925242.3003880E+041.2803606E+07556.58 0.950172.3022905E+041.2814416E+07556.59 0.975142.3022883E+041.2814383E+07556.59 1.000032.3021832E+041.2813804E+07556.59Rev.0 WOLFCREEKTABLE6.2.1-13HOTLEGLONGITUDINALSPLITBREAK763SQUAREINCHESBREAKMASSFLOWANDENERGYFLOWTimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb).000000.0.0.00.001018.1033415E+035.2479074E+06647.62
.002021.4891822E+049.6436977E+06647.58.003003.3538763E+042.1729309E+07647.89.004024.3750401E+042.8341351E+07647.80
.005024.8666372E+043.1512215E+07647.52
.006015.0589146E+043.2737434E+07647.12
.007025.1006945E+043.2992058E+07646.82.008035.0895242E+043.2914153E+07646.70.009025.0885216E+043.2910703E+07646.76
.010015.1218730E+043.3131758E+07646.87
.011005.1876999E+043.3560988E+07646.93
.012025.2726932E+043.4110885E+07646.93
.013035.3551829E+043.4641668E+07646.88
.014035.4244736E+043.5085029E+07646.79
.015035.4734699E+043.5395489E+07646.67.016005.4991203E+043.5553987E+07646.54
.017005.5031540E+043.5572926E+07646.41
.018025.4890352E+043.5475352E+07646.29
.019005.4638748E+043.5308193E+07646.21
.020035.4323329E+043.5101741E+07646.16
.021015.4037895E+043.4917074E+07646.16
.022015.3815910E+043.4775539E+07646.19
.023045.3696178E+043.4701830E+07646.26
.024005.3694077E+043.4704850E+07646.34
.025005.3796572E+043.4775974E+07646.43
.026025.3990369E+043.4906106E+07646.52
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.030035.5093564E+043.5634737E+07646.80
.031055.5368834E+043.5815396E+07646.85.032045.5628834E+043.5985533E+07646.90
.033015.5870866E+043.6144992E+07646.94
.034035.6116586E+043.6306019E+07646.97
.035045.6352566E+043.6460386E+07647.00
.036045.6574044E+043.6605007E+07647.03
.037055.6788352E+043.6744693E+07647.05
.038015.6980592E+043.6869767E+07647.06Rev.0 WOLFCREEKTABLE6.2.1-13(Sheet2)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb).039035.7166900E+043.6990763E+07647.07.040025.7333293E+043.7098501E+07647.07
.041045.7482392E+043.7194704E+07647.06.042035.7606029E+043.7274227E+07647.05.043045.7708024E+043.7339466E+07647.04
.044025.7784192E+043.7387841E+07647.03
.045055.7840728E+043.7423358E+07647.01
.046025.7873443E+043.7443488E+07646.99.047015.7889007E+043.7452485E+07646.97.048055.7888115E+043.7450818E+07646.95
.049025.7873180E+043.7440175E+07646.93
.050035.7845754E+043.7421521E+07646.92
.051035.7807767E+043.7396135E+07646.91
.052015.7759239E+043.7364084E+07646.89
.053015.7703737E+043.7327910E+07646.89
.054015.7642796E+043.7288847E+07646.90.055015.7580205E+043.7249159E+07646.91
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.057015.7470363E+043.7182530E+07646.99
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.060015.7665585E+043.7335873E+07647.46
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.080015.6604661E+043.6685846E+07648.11
.081015.6411574E+043.6561466E+07648.12
.082015.6177799E+043.6410684E+07648.13
.083025.5901391E+043.6232262E+07648.15
.084005.5597445E+043.6035854E+07648.16
.085005.5257094E+043.5815753E+07648.17Rev.0 WOLFCREEKTABLE6.2.1-13(Sheet3)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb).086015.4895819E+043.5581993E+07648.17.087025.4507445E+043.5330892E+07648.18
.088025.4119131E+043.5079676E+07648.19.089025.3735948E+043.4832294E+07648.21.090015.3370360E+043.4596380E+07648.23
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.093025.2396247E+043.3967350E+07648.28.094025.2128103E+043.3794433E+07648.30.095025.1897012E+043.3645411E+07648.31
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.100025.1098817E+043.3132545E+07648.40
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.120045.2014096E+043.3765100E+07649.15
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.195095.0862614E+043.3094170E+07650.66
.200055.0849271E+043.3095949E+07650.86
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.310064.9217645E+043.2120448E+07652.62
.320054.8987294E+043.1974617E+07652.71Rev.0 WOLFCREEKWOLFCREEKTABLE6.2.1-13(Sheet4)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb).330094.8990341E+043.1979432E+07652.77.340034.9108943E+043.2058107E+07652.80
.350064.9137581E+043.2078127E+07652.82.360034.9028820E+043.2009935E+07652.88.370034.8894983E+043.1926330E+07652.96
.380134.8812191E+043.1875826E+07653.03
.390024.8757250E+043.1842653E+07653.09
.400094.8677019E+043.1792462E+07653.13.410124.8567941E+043.1723202E+07653.17.420044.8477201E+043.1665706E+07653.21
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.450034.8439546E+043.1644850E+07653.29
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0.240071.7091493E+041.1530055E+07674.61 0.242531.7044464E+041.1499260E+07674.66 0.245091.6981075E+041.1457784E+07674.740.247521.6909137E+041.1410815E+07674.830.250001.6816993E+041.1350678E+07674.95 0.252561.6725891E+041.1291225E+07675.07 0.255051.6633374E+041.1230921E+07675.20 0.257511.6542717E+041.1171824E+07675.33 0.260051.6466594E+041.1122207E+07675.44 0.262511.6403107E+041.1080847E+07675.53 0.265041.6352876E+041.1048104E+07675.610.267511.6321725E+041.1027787E+07675.65 0.270091.6296620E+041.1011400E+07675.69 0.272651.6276846E+041.0998473E+07675.71 0.275091.6255893E+041.0984763E+07675.74 0.277511.6228524E+041.0966874E+07675.78 0.280031.6185426E+041.0938740E+07675.84 0.282531.6127011E+041.0900616E+07675.92 0.285121.6053276E+041.0852567E+07676.03 0.287551.5970294E+041.0798521E+07676.16 0.290011.5898962E+041.0752073E+07676.28 0.292561.5832910E+041.0709102E+07676.38 0.295091.5829868E+041.0707076E+07676.38 0.297511.5825312E+041.0704075E+07676.39 0.300091.5808551E+041.0693108E+07676.41 0.302651.5808165E+041.0692807E+07676.41 0.305171.5803688E+041.0689818E+07676.410.307531.5753769E+041.0657243E+07676.49 0.310061.5744724E+041.0651197E+07676.47 0.312561.5753012E+041.0656385E+07676.47 0.315011.5760820E+041.0661304E+07676.44 0.317611.5737615E+041.0646068E+07676.47 0.320141.5735895E+041.0644859E+07676.47 0.322731.5734566E+041.0643956E+07676.47 0.325091.5732595E+041.0642618E+07676.47 0.327531.5729905E+041.0640794E+07676.47 0.330241.5726744E+041.0638378E+07676.47 0.332511.5723019E+041.0636121E+07676.47 0.335261.5718891E+041.0633317E+07676.47 0.337521.5715349E+041.0630912E+07676.47 0.340031.5711537E+041.0628319E+07676.47 0.342711.5707870E+041.0625820E+07676.46 0.345061.5704632E+041.0623609E+07676.46 0.347701.5701190E+041.0621254E+07676.46Rev.0 WOLFCREEKTABLE6.2.1-16(Sheet4)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb)0.350251.5698641E+041.0619896E+07676.460.352611.5695247E+041.0617177E+07676.46 0.355161.5692136E+041.0615038E+07676.460.357621.5688914E+041.0612827E+07676.450.360111.5685456E+041.0610463E+07676.45 0.362591.5681810E+041.0607961E+07676.45 0.365181.5677911E+041.0605294E+07676.45 0.367581.5674219E+041.0602767E+07676.450.370341.5670037E+041.0599906E+07676.440.372671.5666639E+041.0597581E+07676.44 0.375041.5663437E+041.0595386E+07676.44 0.377571.5660250E+041.0593197E+07676.44 0.380271.5657086E+041.0591016E+07676.44 0.382771.5654322E+041.0589105E+07676.43 0.385161.5651744E+041.0587321E+07676.43 0.387541.5649144E+041.0585526E+07676.430.390271.5646820E+041.0583914E+07676.43 0.392621.5644725E+041.0582456E+07676.42 0.395221.5642789E+041.0581100E+07676.42 0.397601.5641070E+041.0579890E+07676.42 0.400041.5639449E+041.0578741E+07676.41 0.402571.5637925E+041.0577652E+07676.41 0.405051.5636591E+041.0576689E+07676.41 0.407631.5635413E+041.0575828E+07676.40 0.410131.5634337E+041.0575027E+07676.40 0.412511.5633562E+041.0574432E+07676.39 0.415041.5632975E+041.0573960E+07676.39 0.417751.5632551E+041.0573584E+07676.38 0.420011.5632366E+041.0573381E+07676.38 0.422681.5632315E+041.0573254E+07676.37 0.425021.5632376E+041.0573212E+07676.37 0.427811.5632537E+041.0573215E+07676.360.430271.5632708E+041.0573240E+07676.35 0.432671.5632877E+041.0573264E+07676.35 0.435331.5633025E+041.0573264E+07676.34 0.437521.5633896E+041.0573227E+07676.34 0.440151.5633088E+041.0573125E+07676.33 0.442571.5632964E+041.0572955E+07676.32 0.445231.5632658E+041.0572657E+07676.32 0.447561.5632172E+041.0572247E+07676.31 0.450191.5631366E+041.0571613E+07676.31 0.452871.5630249E+041.0570770E+07676.30 0.455171.5629249E+041.0569885E+07676.30 0.457571.5627623E+041.0568848E+07676.29 0.460221.5635881E+041.0567590E+07676.29 0.462891.5623874E+041.0566152E+07676.28 0.466141.5622119E+041.0564900E+07676.28 0.467771.5619955E+041.0563361E+07676.27Rev.0 WOLFCREEKTABLE6.2.1-16(Sheet5)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb)0.470361.5617741E+041.0561792E+07676.270.472881.5615520E+041.0560223E+07676.26 0.475031.5613565E+041.0558844E+07676.260.477571.5611133E+041.0557133E+07676.260.480111.5608643E+041.0555384E+07676.25 0.482621.5606180E+041.0553588E+07676.25 0.485121.5603492E+041.0551775E+07676.24 0.487581.5600916E+041.0549973E+07676.240.490091.5598319E+041.0548154E+07676.240.492731.5595641E+041.0546276E+07676.23 0.495051.5593323E+041.0544647E+07676.23 0.497861.5590573E+041.0542711E+07676.22 0.500291.5588315E+041.0541119E+07676.22 0.510191.5579889E+041.0535146E+07676.20 0.520411.5571914E+041.0529450E+07676.18 0.530131.5564820E+041.0524389E+07676.170.540251.5557270E+041.0518995E+07676.15 0.550441.5550075E+041.0513839E+07676.13 0.560291.5544285E+041.0509628E+07676.11 0.570431.5539811E+041.0506279E+07676.09 0.580351.5537036E+041.0504053E+07676.07 0.590121.5535838E+041.0502863E+07676.04 0.600131.5535793E+041.0502401E+07676.01 0.610301.5536152E+041.0502172E+07675.98 0.620231.5535777E+041.0501463E+07675.95 0.630271.5533739E+041.0499646E+07675.93 0.640281.5529494E+041.0496323E+07675.90 0.650091.5523642E+041.0491943E+07675.87 0.660101.5516674E+041.0486820E+07675.84 0.670061.5509627E+041.0481657E+07675.82 0.680021.5503246E+041.0476939E+07675.79 0.690431.5497613E+041.0472703E+07675.760.700031.5493333E+041.0469413E+07675.74 0.710081.5489729E+041.0466564E+07675.71 0.720621.5486462E+041.0463919E+07675.68 0.730471.5483628E+041.0461579E+07675.65 0.740061.5481145E+041.0459476E+07675.63 0.750581.5478712E+041.0457353E+07675.60 0.760391.5476578E+041.0456447E+07675.57 0.770461.5474438E+041.0453521E+07675.53 0.780031.5472213E+041.0451554E+07675.50 0.790341.5469339E+041.0449119E+07675.47 0.800221.5465740E+041.0446217E+07675.44 0.810041.5461251E+041.0442584E+07675.41 0.820131.5455070E+041.0478084E+07675.38 0.830541.5447988E+041.0432852E+07675.36 0.840131.5441170E+041.0427849E.07675.33 0.850571.5433935E+041.0422535E+07675.30Rev.0 WOLFCREEKTABLE6.2.1-16(Sheet6)TimeMassFlowEnergyFlowAverageEnthalpy(sec)(lb/sec)(Btu/sec)(Btu/lb)0.860101.5427803E+041.0418005E+07675.270.870641.5421686E+041.0413445E+07675.25 0.880491.5416519E+041.0409555E+07675.220.890291.5411736E+041.0405923E+07675.190.900621.5496860E+041.0402204E+07675.17 0.910431.5402219E+041.0398661E+07675.14 0.920141.5397541E+041.0395092E+07675.11 0.930321.5392543E+041.0391286E+07675.090.940131.5387606E+041.0387534E+07675.060.950201.5382356E+041.0383562E+07675.03 0.960441.5376745E+041.0379341E+07675.00 0.970221.5371025E+041.0375073E+07674.98 0.980231.5364719E+041.0370408E+07674.95 0.990001.5358118E+041.0365561E+07674.92 1.009241.5350854E+041.0360257E+07674.90 1.050091.5319038E+041.0336848E+07674.771.100201.5296400E+041.0318842E+07674.64 1.150271.5270563E+041.0299964E+07674.50 1.200661.5244981E+041.0280607E+07674.36 1.250221.5225611E+041.0265425E+07674.22 1.300011.5284826E+041.0249226E+07674.08 1.350091.5181682E+041.0231482E+07673.94 1.400431.5158958E+041.0214058E+07673.80 1.450081.5135654E+041.0196263E+07673.66 1.500521.5108628E+041.0175997E+07673.52 1.550211.5082453E+041.0156373E+07673.39 1.600011.5056220E+041.0136713E+07673.26 1.650411.5029192E+041.0116506E+07673.12 1.700531.5003251E+041.0097044E+07672.99 1.750421.4978791E+041.0078568E+07672.86 1.800001.4954574E+041.0060259E+07672.72 1.850041.4910174E+041.0027829E+07672.551.900381.4937494E+041.0045448E+07672.50 1.950601.4846071E+049.9803806E+06672.26 2.000321.4821804E+049.9621501E+06672.13Rev.0 WOLFCREEKTABLE6.2.1-17REACTORCAVITYPRESSURE-TEMPERATUREANALYSISSUBCOMPARTMENTNODALDESCRIPTION CalculatedPeakDesignPeakInitialConditionsPressurePressureDesignNodeVolumeTemp.Press.Humid.DifferentialDifferentialMarginNo.(ft3)(F)(psia)(%)(psig)(psig)%155.512014.7100101.95131.522.5255.812014.710098.88131.524.8357.212014.710046.36131.564.7456.012014.710024.86131.581.1 555.612014.710020.70131.584.3 656.012014.710020.89131.584.1 757.212014.710023.16131.582.4856.012014.710039.00131.570.39174.912014.710081.70131.537.910172.712014.710080.89131.538.5 11171.812014.710043.97131.566.6 12173.012014.710025.21131.580.8 13175.012014.710020.74131.584.2 14173.012014.710020.91131.584.1 15171.812014.710023.49131.582.1 16173.012014.710038.66131.570.6 1748.312014.710030.59104.770.81866.212014.710030.70104.770.71948.312014.710020.29104.780.6 2066.212014.710018.79104.782.1 2148.312014.710017.89104.782.9 2266.212014.710018.54104.782.3 2348.312014.710020.19104.780.7 2466.212014.710030.23104.771.1 2557.212014.710019.4287.977.9 2657.312014.710019.1387.978.2 2757.212014.710016.5987.981.1 2857.312014.710015.2287.982.7 2957.212014.710015.0687.982.9 3057.312014.710015.3287.982.6 3157.212014.710016.5987.981.1 3257.312014.710019.1387.978.2 331944.712014.710010.2460.082.9 341802.712014.71007.8660.086.9 351306.512014.71007.7060.087.2 36566.712014.71005.9460.090.1 371291.112014.71002.8660.095.2 38252.412014.71003.8860.093.5 39252.412014.71003.1560.094.8 40252.412014.71002.8160.095.3 413883.112014.71002.2560.096.3Rev.0 WOLFCREEKWOLFCREEKTABLE6.2.1-17(Sheet2)
CalculatedPeakDesignPeakInitialConditionsPressurePressureDesignNodeVolumeTemp.Press.Humid.DifferentialDifferentialMarginNo.(ft3)(F)(psia)(%)(psig)(psig)%4212.3*12014.710065.4467.02.34311.9*12014.710064.9467.03.14411.6*12014.710036.1467.046.14515.0*12014.710020.6467.069.2 4612.4*12014.710016.7567.075.0 4715.0*12014.710018.0667.073.0 4811.6*12014.710020.1967.069.94911.9*12014.710032.5067.051.5 502.486x10612014.71001.9760.096.7*Initialvolumeofneutronshieldcompartment.SeeTable6.2.1-19forneutronshieldbagcompartmentvolumeandventareaasafunctionof height.Rev.0 WOLFCREEKTABLE6.2.1-18REACTORCAVITYANALYSISSUBCOMPARTMENTVENTPATHDESCRIPTION Choked/time-VentFromToUnchokedVentAreaL/ATermHEADLOSS--------KPathNo.Node Node Flow (ft2)(ft-1)Friction Turning Expansion Contraction Total112unchoked1.053.83270.13780.051.00.411.5978219choked/.007-.00913.540.36630.04170.01.00.281.3217 323choked/.003-1.004.001.27000.09440.051.00.331.4744 4210unchoked14.260.34780.02890.01.00.271.2989 534.095-1.00choked/.017-.0451.932.16290.09440.051.00.41.54446311unchoked14.980.33110.02800.01.00.251.278 745unchoked2.651.76730.13780.051.00.3751.5628 8412unchoked14.430.34370.02860.01.00.261.2886956unchoked1.053.83270.13780.051.00.411.597810513unchoked13.200.37580.04170.01.00.291.33171167unchoked4.001.27000.09440.051.00.331.4744 12614unchoked14.430.34370.02860.01.00.261.2886 1378choked/.025-.0451.932.16290.09440.051.00.41.5444 14715unchoked14.980.33110.02800.01.00.251.278 15816unchoked14.430.34370.02860.01.00.261.2886 16823choked/.023-.0551.622.84600.13060.01.00.411.5406 17824unchoked3.351.43190.07770.01.00.381.4577 18910unchoked16.870.38470.04130.051.00.151.2413 19916choked/.005-1.0016.870.38470.04130.051.00.151.2413 20942choked/.035-1.0019.730.30410.08020.01.00.0251.1052 211011choked/.005-1.0018.210.36540.03880.051.00.111.1988 221043choked/.055-1.0019.260.31550.08220.01.00.0251.1072 231112.095-1.0018.210.36540.03880.051.00.111.1988 choked/.023-.045241144choked/.055-1.0018.800.31910.08420.01.00.031.1142 251213unchoked16.870.38470.04130.051.00.151.2413 261245choked/.065-.07519.260.31150.08220.01.00.0251.1072 271314unchoked16.870.38470.04130.051.00.151.2413 281346choked/.075-.08519.730.30410.08020.01.00.0251.1052 291415unchoked18.210.36540.03880.051.00.111.1988 301447choked/.075-.08519.260.31150.08220.01.00.0251.1072 311516choked/.03-.05518.210.36540.03880.051.00.111.1988321548unchoked18.800.31910.08420.01.00.031.1142331649choked/.055-1.0019.260.31150.08220.01.00.0251.1072 341718unchoked4.440.76560.14330.051.00.311.5033 351724unchoked4.440.76560.14330.051.00.311.5033 361725unchoked7.441.10080.08750.01.00.01.0875 371819unchoked4.440.76560.14330.051.00.311.5033 381826unchoked7.441.10080.08750.01.00.01.0875 391920unchoked4.440.76560.14330.051.00.311.5033 401927unchoked7.441.10080.08750.01.00.01.0875 412021unchoked4.440.76560.14330.051.00.311.5033 422028unchoked7.441.10080.08750.01.00.01.0875Rev.0 WOLFCREEKTABLE6.2.1-18(Sheet2)
Choked/time-VentFromToUnchokedVentAreaL/ATermHEADLOSS--------KPathNo.Node Node Flow (ft2)(ft-1)Friction Turning Expansion Contraction Total432122unchoked4.440.76560.14330.051.00.311.5033442129unchoked7.441.10080.08750.01.00.01.0875 452223unchoked4.440.76560.14330.051.00.311.5033 462230unchoked7.441.10080.08750.01.00.01.0875 472324unchoked4.440.76560.14330.051.00.311.5033 482331unchoked7.441.10080.08750.01.00.01.0875 492432unchoked7.441.10080.08750.01.00.01.0875 502526unchoked5.050.93440.11150.051.00.221.3815 512532unchoked5.050.93440.11150.051.00.221.3815 522533unchoked8.021.17210.09320.01.00.01.0932 532627unchoked5.050.93440.11150.051.00.221.3815542633unchoked8.021.17210.09320.01.00.01.0932552728unchoked5.050.93440.11150.051.00.221.3815 562733unchoked8.021.17210.09320.01.00.01.0932 572829unchoked5.050.93440.11150.051.00.221.3815 582833unchoked8.021.17210.09320.01.00.01.0932 592930unchoked5.050.93440.11150.051.00.221.3815 602933unchoked8.021.17210.09320.01.00.01.0932 613031unchoked5.050.93440.11150.051.00.221.3815 623033unchoked8.021.17210.09320.01.00.01.0932 633132unchoked5.050.93440.11150.051.00.221.3815 643133unchoked8.021.17210.09320.01.00.00.0932 653233unchoked8.021.17210.09320.01.00.00.0932 663334unchoked116.950.30910.00.01.00.31.3 673337unchoked17.421.10050.00.171.00.331.505 683435unchoked233.050.03860.00.01.00.0751.075 693536unchoked79.880.07920.00.01.00.221.22 703637unchoked74.250.2349-----------------------ORIFICE---------------------
713638unchoked29.180.18390.00.01.00.11.1 723738unchoked39.250.1529-----------------------ORIFICE---------------------
733739unchoked39.250.1529-----------------------ORIFICE---------------------
743740unchoked39.250.1529-----------------------ORIFICE---------------------
753741unchoked71.500.20980.0370.01.00.01.037763750unchoked82.660.23290.0780.01.00.031.108773839unchoked29.180.17140.00.01.00.11.1 783940unchoked29.180.17140.00.01.00.11.1 794041unchoked29.180.33120.00.01.00.11.1 804150unchoked98.680.16210.0450.01.00.181.225 814243unchoked*0.0320.0321.00.51.564 824249.095-.645*0.0320.0321.00.51.564
.055-.065 choked/.007-.035834250choked/.007-1.000*0.00.01.00.01.0 844344choked/.007-.03*0.0320.0321.00.51.564 854350choked/.007-.1000*0.00.01.00.01.0 864445choked/.065-.075*0.0320.0321.00.51.564 874450choked/.017-.095*0.00.01.00.01.0Rev.0 WOLFCREEKTABLE6.2.1-18(Sheet3)
Choked/time-VentFromToUnchokedVentAreaL/ATermHEADLOSS--------KPathNo.Node Node Flow (ft2)(ft-1)Friction Turning Expansion Contraction Total884546unchoked*0.0320.0321.00.51.564894550choked/.03-.075*0.00.01.00.01.0 904647unchoked*0.0320.0321.00.51.564 914650choked/.045-.075*0.00.01.00.01.0 924748choked/.075-.085*0.0320.0321.00.51.564 934750choked/.045-.075*0.00.01.00.01.0 944849unchoked*0.0320.0321.00.51.564 954850choked/.035-.075*0.00.01.00.01.0 964950choked/.017-.055*0.00.01.00.01.0 9718choked/.003-1.002.651.76730.13780.051.00.3751.5628 98117choked/.003-1.061.622.84600.13060.01.00.411.540699124choked/.003-1.003.521.36330.07640.01.00.381.4564100217choked/.003-1.001.463.15150.15470.01.00.411.5647101218choked/.003-1.003.521.36330.07640.01.00.381.4564 102318unchoked3.351.43190.07770.01.00.381.4577 103319.145-1.001.463.15150.15470.01.00.411.5647 choked/.021-0.55104419unchoked1.622.84600.13060.01.00.411.5406 105420unchoked3.351.43190.07770.01.00.381.4577 106520unchoked3.521.36330.07640.01.00.381.4564 107521unchoked1.622.84600.13060.01.00.411.5406 108621unchoked1.463.15150.15470.01.00.411.5647 109622unchoked3.521.36330.07640.01.00.381.4564 110722unchoked3.351.43190.07770.01.00.381.4577 111723unchoked1.463.15150.15470.01.00.411.5647SeeTablesofVolumesandVentAreasof"WaterbagCompartments"andVentAreasandL/Asfor"WaterbagCompartments" (Tables6.2.1-19and6.2.1-20)*SeeTableofVentAreasandL/Asfor"WaterbagCompartments"(Table6.2.1-20)Rev.0 WOLFCREEKTABLE6.2.1-19VOLUMESANDVENTAREASOFWATERBAGCOMPARTMENTSHeightVentAreatoContainmentVolume(ft)CmptNo.(ft 2)(ft 3)0.0424.70512.30434.592711.90444.49311.60453.74015.00464.69112.40 473.74015.00 484.49311.60 494.592711.901.0426.19050.96435.88649.55 445.58348.24 459.32462.40 4615.69151.06 478.32462.40 485.58348.24495.88649.551.0014238.660147.614337.650143.68 4436.640139.84 4547.400180.90 4638.660147.71 4747.400180.90 4836.640139.84 4937.650143.681.1004238.660147.614337.650143.68 4436.640139.84 4547.400180.90 4638.660147.714747.400180.90 4836.640139.84 4937.650143.68Rev.0 WOLFCREEKTABLE6.2.1-20VENTAREASANDL/AsFORWATERBAGCOMPARTMENTSFlowpath42to5042to4342to4943to5043to44TimeVentVentVentVentVent(sec)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)0.04.705+005.356-010.00.04.593+005.487-010.00.00254.705+005.356-011.174-066.629+060.04.593+005.487-010.0 0.00604.709+005.352-014.935-031.577+030.04.596+005.483-010.0 0.01204.780+005.251-011.273-016.115+016.196-031.256+034.679+005.386-016.970-031.117+03 0.01805.147+004.896-015.947-011.309+016.953-021.120+024.992+005.049-017.797-029.983+01 0.02906.190+004.071-012.914+002.672+006.375-011.221+015.886+004.281-017.162-011.087+01 0.03503.866+016.518-024.314+001.804+001.334+005.835+003.765+016.693-021.500+005.189+00 0.05803.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.08003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.10003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.20003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.30003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+000.40003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+000.50003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.60003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.70003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.80003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 0.90003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00 1.00003.866+016.518-024.314+001.804+004.314+001.804+003.765+016.693-024.314+001.804+00Flowpath44to5044to4545to5045to46TimeVentVentVentVent(sec)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)0.04.493+005.609-010.03.740+008.262-010.00.00254.493+005.609-010.03.740+008.262-010.0 0.00604.493+005.609-010.03.740+008.262-010.0 0.01204.497+005.604-014.534-061.717+063.740+008.262-010.0 0.01804.538+005.553-015.440-031.431+033.748+008.245-012.227-063.419+06 0.02904.909+005.134-011.490-015.224+013.952+007.819-012.168-023.591+020.03505.364+004.698-013.840-012.027+014.287+007.208-019.657-028.061+010.05803.664+016.878-022.994+002.600+008.003+003.861-011.594+004.882+00 0.08003.664+016.878-024.314+001.804+004.740+016.519-024.192+001.897+00 0.10003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.20003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.30003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.50003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.50003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.60003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.70003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.80003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 0.90003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00 1.00003.664+016.878-024.314+001.804+004.740+016.519-024.314+001.804+00L/Atermisnotapplicablewhenventarea=0.0Rev.0 WOLFCREEKTABLE6.2.1-20(Sheet2)Flowpath46to5046to4747to5047to48TimeVentVentVentVent(sec)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)0.04.691+005.372-010.03.740+008.262-010.00.00254.691+005.372-010.03.740+008.262-010.0 0.00604.691+005.372-010.03.740+008.262-010.0 0.01204.691+005.372-010.03.740+008.262-014.331-061.797+06 0.01804.691+005.372-012.277-063.419+063.740+008.262-012.355-023.305+02 0.02904.752+005.303-012.168-023.591+023.774+008.189-011.034-017.532+01 0.03504.962+005.079-019.657-028.061+013.887+007.949-011.657+004.698+00 0.05809.163+002.750-011.594+004.882+006.099+005.066-014.314+001.804+00 0.08003.866+016.518-024.192+001.857+004.740+016.519-024.314+001.804+00 0.10003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.20003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+000.30003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+000.40003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.50003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.60003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.70003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.80003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 0.90003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00 1.00003.866+016.518-024.314+001.804+004.740+016.519-024.314+001.804+00Flowpath48to5048to4949to50TimeVentVentVent(sec)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)Area(ft 2)L/A(ft-1)0.04.493+005.609-010.04.593+005.487-010.00254.493+005.609-010.04.593+005.487-01 0.00604.493+005.609-010.04.593+005.487-01 0.01204.493+005.609-011.788-064.354+064.597+005.482-01 0.01804.496+005.606-014.226-031.842+034.639+005.432-01 0.02904.563+005.522-011.210-016.436+015.017+005.023-010.03504.678+005.387-013.181-012.447+015.481+004.598-010.05805.583+004.514-012.651+002.936+003.765+016.693-02 0.08003.664+016.878-024.314+001.804+003.765+016.693-02 0.10003.664+016.878-024.314+001.804+003.765+016.693-02 0.20003.664+016.878-024.314+001.804+003.765+016.693-02 0.30003.664+016.878-024.314+001.804+003.765+016.693-02 0.40003.664+016.878-024.314+001.804+003.765+016.693-02 0.50003.664+016.878-024.314+001.804+003.765+016.693-02 0.60003.664+016.878-024.314+001.804+003.765+016.693-02 0.70003.664+016.878-024.314+001.804+003.765+016.693-02 0.80003.664+016.878-024.314+001.804+003.765+016.693-02 0.90003.664+016.878-024.314+001.804+003.765+016.693-02 1.00003.664+016.878-024.314+001.804+003.765+016.693-02L/Atermisnotapplicablewhenventarea=0.0Rev.0 WOLFCREEKTABLE6.2.1-21REACTORCAVITYANALYSISCOEFFICIENTSFORDETERMINATIONOFFORCESANDMOMENTSONTHEREACTORPRESSUREVESSELLevelCmpt.X-ForceY-ForceUpliftForceMomentAboutX-axisMomentAboutY-axisNo.No.(in.2)(in.2)(in.2)(ft-in.2)(ft-in.2)142*Ao=289.91Ao=-289.910.0SeenotebelowSeenotebelowAv=869.74Av=-869.740.0SeenotebelowSeenotebelow43*Ao=289.91Ao=289.910.0SeenotebelowSeenotebelowAv=869.74Av=869.740.0SeenotebelowSeenotebelow44*Ao=120.08Ao=120.080.0SeenotebelowSeenotebelowAv=360.24Av=360.240.0SeenotebelowSeenotebelow45*Ao=-120.08Ao=120.080.0SeenotebelowSeenotebelowAv=-360.24Av=360.240.0SeenotebelowSeenotebelow46*Ao=-289.91Ao=289.910.0SeenotebelowSeenotebelowAv=-869.74Av=869.740.0SeenotebelowSeenotebelow47*Ao=-289.91Ao=-289.910.0SeenotebelowSeenotebelowAv=-869.74Av=-869.740.0SeenotebelowSeenotebelow48*Ao=-120.08Ao=-120.080.0SeenotebelowSeenotebelowAv=-360.24Av=-360.240.0SeenotebelowSeenotebelow49*Ao=120.08Ao=-120.080.0SeenotebelowSeenotebelowAv=360.24Av=-360.240.0SeenotebelowSeenotebelow296305.84-2611.97-946.98-12,689.8312,332.62106222.552662.99-852.0012,383.8312,934.18112568.016199.73-757.0228,926.325722.57 12-2516.996283.03-852.0029,800.40-5116.21 13-6305.842611.97-946.9812,689.83-12,332.62 14-6222.55-2662.99-852.00-12,383.83-12,934.1815-2568.01-6199.73-757.02-28,926.32-5722.57 162516.99-6283.03-852.00-29,800.405116.21313058.33-1266.811715.184668.5520,128.9223017.941291.551620.20-4062.1919,254.9331245.483006.871525.22-7859.048428.514-1220.743047.271620.20-8460.69-8823.62 5-3058.331266.811715.18-4668.55-20,128.92 6-3017.94-1291.551620.204062.19-19,254.93 7-1245.48-3006.871525.227859.04-8428.5181220.74-3047.271620.208460.698823.62Rev.0 WOLFCREEKTABLE6.2.1-21(Sheet2)LevelCmpt.X-ForceY-ForceUpliftForceMomentAboutX-axisMomentAboutY-axisNo.No.(in.2)(in.2)(in.2)(ft-in.2)(ft-in.2)4177869.41-68.6882.2-305.6335,018.87185613.085515.9682.224,546.0224,978.2119-68.687869.4282.235,018.92-305.63 20-5515.965613.0882.224,978.21-24,546.02 21-7869.4168.6882.2305.63-35,018.8722-5613.08-5515.9682.2-24,546.02-24,978.212368.68-7869.4282.2-35,018.92305.63245515.96-5613.0882.2-24,978.2124,546.02 256633.91-57.900.0-257.6529,520.91264731.834649.950.020,692.2921,056.6327-57.906633.920.029,520.95-257.65 28-4649.954731.830.021,056.62-20,692.30 29-6633.9157.900.0257.65-29,520.9130-4731.83-4649.950.0-20,692.29-21,056.63 3157.90-6633.920.0-29,520.95257.65324649.95-4731.830.0-21,056.6220,692.30330.00.028,727.170.00.0 340.00.00.00.00.0 350.00.00.00.00.0 360.00.00.00.00.0 370.00.00.00.00.0 380.00.00.00.00.0 390.00.00.00.00.0 400.00.00.00.00.0 410.00.00.00.00.0 Contain-ment50---37,627.21--Rev.0 WOLFCREEKTABLE6.2.1-21(Sheet3)Note:Ingeneral,theforceonthereactorpressurevesselduetopressureincompartmentiatanytimeisgivenbyF(t)=P(t)A,iiiwhereF(t)=forceonRPVduetocompartmentiattimet(lb), ifP(t)=pressureincompartmentiatt(lb/in.
2),fromFigures6.2.1-34throughi6.2.1-39fandA=areaofprojectionofcompartmentionRPVperpendiculartothedirectioniofforce,obtainedfromtheappropriatecolumnabove(in.
2).Forcompartmentsdenotedby*,theforceisfunctionoftheheightofwater-filledneutronshieldbags.Priortodisintegrationofaneutronshieldbag,theforceisgivenbyh(t)F(t)=P(t)x[Ao+Avxi]+P(t)x[Avx(H-h(t)-T)],iiii1.0'ciiwhereAo=initialprojectedareaofcompartmentionRPV;notafunctionofshieldibagheight(in.2)Av=variableprojectedareaofcompartmenti,basedona1.0ftmaximumbagitraveldistancepriortodisintegrationofbag(in.
2)h(t)=distancetraveledbyshieldbagattimet,(ft),determinedbytheiequationsofrigidbodymechanicsP(t)=containmentpressure(compartment50)inlb F/in.2attimet,obtainedcfromFigure6.2.1-39H=maximumheightofcompartmentprojectionofRPVafterbagdisintegration;thisheightrepresentsthedistancefromthetopoftheneutronshieldsupportplatformtothetopoftheRPVheadbolts(3.5ft),andT=shieldbagthickness(1.0ft).Followingdisintegrationofashieldbag,theforceisgivenbyF(t)=P(t)x[Ao+(AvxH)]iiiiwhereallparametersweredefinedabove.Rev.0 WOLFCREEKTABLE6.2.1-21(Sheet4)MomentsaretakenaboutthenozzlecenterlineatEl.2014'-6",withdirectionshowninFigures6.2.1-28through6.2.1-32.Ingeneral,themomentabouttheaxisofrotationduetopressureincompartmentiisgivenbyM(t)=P(t)C,iiiwhereM(t)=momentduetocompartmentiattimet(ft-lb), ifP(t)wasdefinedabove, iandC=projectedareaofcompartmentionRPVtimesappropriatemomentiarm(ft-in.2)Forcompartmentsdenotedby*,themomentisafunctionofneutronshieldbagheight.Priortodisintegrationofashieldbag,themomentduetocompartmentiish(t)givenbyM(t)=P(t)x[Ao+Avxi]x[7.1'+h(t)]iiii1.02+P(t)x[Avx(H-h(t)-T)]x[7.1'+(h(t
)+T+H)],cii2whereallparametersweredefinedabove.MomentsaboutX-axisareduetothey-directionforce.Therefore,theappropriatevaluesofAoandAvshouldbeused.Followingdisintegrationofashieldbag,themomentsduetocompartmentsdenotedby*aregivenbyM(t)=P(t)x[Ao+(AvxH)]x
[7.1'+H]iiii2whereallparametersweredefinedabove.Rev.0 WOLF CREEK TABLE 6.2.1-22 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS Net Volume Peak Pressure c Time to Peak Break Design Pressure c Node a (ft 3) (psig) Pressure (sec) Case b (psig) 1 3962.5 8.911 9.800 x 10-2 1 24.53 2 545.9 9.368 9.550 x 10-2 1 24.53 3 828.1 9.895 5.550 x 10-2 1 24.53 4 2452.8 7.522 5.500 x 10-2 1 24.53 5 1957.1 15.864 3.700 x 10-2 1 24.53
6 826.8 12.746 8.300 x 10-2 2 24.53 7 231.7 27.321 e 5.700 x 10-3 3 24.53 8 2299.5 17.463 1.800 x 10-2 1 24.53 9 4075.4 10.903 9.450 x 10-2 1 24.53/13.03 f 10 3452.2 13.219 1.600 x 10-2 1 24.53/13.03 11 3294.4 8.868 4.800 x 10-2 1 13.03
12 8144.3 8.397 1.000 x 10-1 1 13.03
13 7912.9 3.475 1.000 x 10-1 1 13.03 14 17788.0 - 1.000 x 10-1 1 -
d 15 23994.0 1.532 1.000 x 10-1 1 24.53/13.03 16 2.5 x 10 6 - 1.000 x 10-1 1 -
17 1677.5 8.667 9.950 x 10-2 1 24.53 18 295.2 9.407 5.550 x 10-2 1 24.53
19 184.7 9.058 7.400 x 10-2 1 24.53 20 78.1 10.385 6.000 x 10-2 1 24.53 21 734.4 10.754 5.950 x 10-2 1 24.53
22 278.6 10.231 8.650 x 10-2 1 24.53
23 639.0 12.078 3.300 x 10-2 1 24.53 24 1303.4 10.202 3.000 x 10-2 1 24.53/13 03 25 1165.1 9.984 9.200 x 10-2 1 24.53/13.03 26 1167.7 8.464 1.000 x 10-1 1 13.03 27 2976.2 8.282 1.000 x 10-1 1 13.03 28 1385.1 6.160 8.700 x 10-2 1 13.03 29 10860.2 1.638 1.000 x 10-1 1 24.53/13.03
30 865.3 8.047 9.600 x 10-2 1 17.96
31 2208.9 7.218 9.750 x 10-2 1 17.96
32 1679.5 12.049 6.600 x 10-2 1 17.60
33 3152.0 10.210 8.050 x 10-2 1 17.60
34 7706.7 9.307 9.300 x 10-2 1 17.60
35 12006.6 8.987 9.300 x 10-2 1 11.79
36 4206.6 6.765 9.850 x 10-2 1 11.79
37 25571.4 1.792 1.000 x 10-1 1 17.60
38 1578.0 5.610 9.150 x 10-2 1 14.63
39 1862.0 8.236 9.550 x 10-2 1 17.60
40 1920.6 8.254 8.500 x 10-2 1 17.60
41 1920.6 7.723 9.750 x 10-2 1 17.60
42 1862.0 8.054 9.150 x 10-2 1 17.60 43 4008.7 1.154 1.000 x 10-1 1 17.60
44 3824.0 1.150 1.000 x 10-1 1 17.60
45 1621.8 4.662 8.650 x 10-2 1 14.63 Rev. 0 WOLF CREEK TABLE 6.2.1-22 (Sheet 2) a Net Volume Peak Pressure c Time to Peak Break Design Pressure cNode (ft3) (psig) Pressure (sec) Case b (psia) 46 896.9 5.650 8.700 x 10
-2 1 11.79 47 979.4 5.520 8.050 x 10
-2 1 11.79 48 979.4 5.562 8.050 x 10
-2 1 11.79 49 896.9 5.489 8.100 x 10
-2 1 11.79 50 2011.7 0.702 1.000 x 10
-1 1 11.79 51 1904.3 0.705 1.000 x 10
-1 1 11.79 52 4543.7 1.842 9.900 x 10
-2 1 14.63 53 2234.9 - 7.550 x 10
-2 1 -
d 54 2305.4 - 7.900 x 10
-2 1 -
d 55 2305.4 - 7.900 x 10
-2 1 -
d 56 2234.9 - 8.200 x 10
-2 1 -
d 57 4811.4 - 1.000 x 10
-1 1 -
d 58 4595.6 - 1.000 x 10
-1 1 -
d 59 2601.5 9.825 8.750 x 10
-2 1 17.60 NOTES: a. Initial conditions for all nodes are identical: Temp = 120
°F, press. = 14.7 psia, and relative humidity = 50%
- b. Break cases: l = 763 in.
2 hot leg split 2 = 436 in.
2 double-ended pump suction line break 3 = 236 in.
2 double-ended cold leg break c. These are diffential pressures between the compartment and the remainder of the containment (Node 16).
- d. The compartments where no peak or design pressure is given are considered to be part of the containment with no walls between them
and the containment on which a pressure differential could be exerted.
- e. Structural model considered average pressure load over element (see nodes 3 and 7, Figure 6.2.1-43). Hence, resultant pressure on effected element does not exceed design pressure of 24.53 psig.
- f. Structural model divided at this node. Design pressure higher on effected half (24.53 psig), lower on noneffected half (13.03 psig).
Rev. 0 WOLFCREEKTABLE6.2.1-23STEAMGENERATORLOOPCOMPARTMENTANALYSISNodesVentAreaHeadLossCoefficientsFlowFromTo(Ft 2)K contraction K expansion K friction Coefficient/a12205.870.321.00.01580.8700.0447l380.750.401.00.0520.8300.2238 l1621.000.441.00.0000.8300.3667 117207.000.051.00.02280.9660.0476 23126.230.091.00.010.9500.0454 26105.00.121.00.01420.9380.081921917.350.321.00.1050.8380.68213442.000001.00.080.9800.500 3765.000.201.00.0250.9030.0779 31833.960.281.00.0550.8700.34854726.300.001.00.1030.9500.92904818.300.441.00.0870.8090.312841586.600.001.00.0870.9600.2944 56270.700.021.00.0100.9850.0516 59177.200.021.00.0270.9800.0960 523100.620.081.00.0290.9500.094 68224.910.101.00.0110.9500.0322 62241.910.201.00.0600.8910.282478103.600.1251.00.0240.9330.07367206.20.351.00.2020.8001.909810189.00.0501.00.0250.9600.079282191.840.221.00.03520.8900.1289910382.730.021.00.0110.9850.0286 911177.200.021.00.0270.9800.1096 924210.000.081.00.0190.9540.04551012190.000.051.00.0370.9600.09871025168.130.081.00.0220.9530.07041112266.500.021.00.0190.9800.04921126182.760.041.00.0150.9740.05711213247.300.271.00.0660.8650.0561216102.000.381.00.02430.8430.18621227477.660.051.00.0120.9700.02091314127.2250.151.00.1280.8850.3868 1315131.000.171.00.0660.8650.1764 132846.250.411.00.1150.8100.2560 1516204.000.381.00.2160.7900.1155 15291334.000.051.00.0090.9700.009171831.680.401.00.0840.8200.125171959.850.421.00.0410.8450.1118 1730187.700.031.00.0110.9800.0298 181937.340.221.00.0190.8980.1156182013.000.361.00.02460.8500.1986183245.670.101.00.0470.9330.2976193217.350.321.00.1050.8380.7883 202130.200.401.00.04250.8330.2615 205919.030.171.00.0750.8960.7141212250.880.281.00.0210.8770.0822Rev.0 WOLFCREEKTABLE6.2.1-23(Sheet2)NodesVentAreaHeadLossCoefficientsFlowFromTo(Ft 2)K contraction K expansion K friction Coefficient/a212582.900.031.00.0360.9690.2015215964.450.081.00.0400.9400.2109222372.300.151.00.0130.9270.1316225916.160.201.00.0980.8770.8410222334.340.081.00.1110.9160.39572425150.840.031.00.0120.9790.1086 242635.150.281.00.0470.8680.2895 2434210.000.081.00.0190.9540.065 252782.900.031.00.0360.9700.2516 2534207.550.081.00.0220.9600.0655 262770.620.151.00.0500.9130.1250 2635182.760.041.00.9740.9740.0744 271640.200.381.00.8430.8360.4747 272899.390.271.00.0660.8650.1430 2735477.600.051.00.0120.9700.0285 282998.940.171.00.0660.8650.4061 2836216.000.001.00.001.00.0629 291680.400.381.00.2070.7940.2944 29371334.000.051.00.0090.9700.0102 301610.50.441.00.000.8330.9606 3031147.000.101.00.0230.9440.0663 311621.000.441.00.000.8330.3939 3138102.000.121.00.03530.9300.1489 3259215.300.121.00.0430.9270.0456 3334248.000.271.00.02240.8800.0366 3339147.860.141.00.030.9250.1185 3359415.120.001.00.001.000.0284 3435248.000.271.00.02240.8800.064 3440149.300.081.00.0280.9500.1172 3441149.300.081.00.0280.9500.1172 35160.00-----
3536300.300.021.00.01610.9800.0753542147.860.371.00.03790.8430.11853637300.300.021.00.0540.9650.132 3743313.600.101.00.01370.9480.0558 3744300.600.101.00.0230.9440.0582 3845102.000.001.00.02660.9870.113 3940131.300.301.00.02530.8690.0233 3946109.100.231.00.0340.8900.1062 4047119.000.201.00.02850.9030.0974 4142131.300.301.00.02530.8690.2330 4148119.000.201.00.02850.9030.0974 4249109.100.231.00.0340.8900.1062 4350244.500.201.00.0200.9050.0474 4451231.500.231.00.0200.8940.0501 4552102.000.001.00.03610.9820.1481 464760.000.351.00.0500.8450.055 4653109.100.231.00.0340.8880.1195Rev.0 WOLFCREEKTABLE6.2.1-23(Sheet3)NodesVentAreaHeadLossCoefficientsFlowFromTo(Ft 2)K contraction K expansion K friction Coefficient/a4748125.900.121.00.04180.9280.18964754119.000.201.00.02850.9030.1096484960.000.351.00.0500.8450.0554855119.000.201.00.02850.9030.10964956109.100.231.00.0340.8880.11955051125.900.121.00.04180.9280.18765057244.500.201.00.05340.9040.05335158231.450.231.00.0230.8930.05635354160.850.301.00.0260.8680.02745455405.420.001.00.001.000.0344 5556160.850.301.00.0260.8680.02745758405.420.001.00.001.000.0632192229.000.271.00.0310.8770.2093232435.150.281.00.0470.8680.28952333100.620.081.00.0290.9500.1351 5216219.430.001.00.01.00.0494 5316183.560.001.00.01.00.0474 5416173.920.001.00.01.00.0500 5516173.920.001.00.01.00.0500 5616183.560.001.00.01.00.04745716372.240.001.00.01.00.02345816360.4960.001.00.01.00.0241Rev.0 WOLFCREEKTABLE6.2.1-24STEAMGENERATORLOOPCOMPARTMENTANALYSISFORCECOEFFICIENTSFORSTEAMGENERATORForceinE-WForceinN-SNodeDirectionDirectionUpliftForce5-2251.36-2141.193593.46265.60-1179.781283.6 82830.43-1023.383454.859-2254.312254.313712.23101409.972090.032804.8216905.60-2496.80------22648.00-2878.40------
23-5492.80-5224.00------24-5500.005500.00------253440.005099.20------346206.70-32208.78------35-6206.7032208.78------
395075.58-26115.665612.8440-5075.5826115.664390.24463430.96-17651.92------47-3430.9617651.92------535781.17-29743.49-13689.8 54-5781.1729743.49-10707.7Rev.0 WOLFCREEKTABLE6.2.1-25STEAMGENERATORLOOPCOMPARTMENTANALYSISFORCECOEFFICIENTSONREACTORCOOLANTPUMPForceinE-WForceinN-SNodeDirection DirectionUpliftForce2-9311.58-10347.01929.33310404.0-10404.02042.82 6-10273.75519.621543.4577356.743048.371021.4181829.4412241.121663.76184082.04-4082.0419-3653.86-4060.16 202886.781196.0421717.794803.022-4031.02165.90321311.4-25029.30-3949.4559-1311.425023.27-4221.82Rev.0 WOLF CREEK TABLE 6.2.1-26 PRESSURIZER COMPARTMENT ANALYSIS Net Peak Time to Design Volume Pressure Peak Pressure Pressure
Node a (ft 3) (psig) (sec) (psig) 1 3962 8.9 0.056 24.53 2 1374 6.7 0.04 24.53 3 2453 0.9 0.5 24.53 4 1677 13.6 0.014 24.53 5 480 5.7 0.055 24.53 6 865 14.7 0.016 17.96 7 2209 12.0 0.028 17.96 8 1578 9.7 0.067 14.63 9 1622 7.9 0.055 14.63 10 4544 1.1 0.5 14.63 11 2.6 x 10 6 0.5 a Initial conditions for all nodes are identical. Temp = 120
°F, press. = 14.7 psia, and relative humidity = 50%.
Rev. 7 WOLF CREEK TABLE 6.2.1-27 PRESSURIZER COMPARTMENT ANALYSIS Nodes Vent Area Head Loss Coefficients Flow
From To (Ft 2) K contraction K expansion K friction Coefficient l/a 1 2 286.62 0.25 1.0 0.0134 0.89 0.034
1 4 207.00 0.05 1.0 0.0228 0.966 0.0476
2 3 42.00 0.25 1.0 0.080 0.0828 0.500 2 5 51.31 0.27 1.0 0.160 0.838 0.5153
2 11 170.00 0.28 1.0 0.039 0.8707 0.001
3 11 131.2 0.00 1.0 0.00 1.00 0.001 4 5 91.53 0.30 1.0 0.0253 0.869 0.0866
4 6 187.70 0.03 1.0 0.011 0.980 0.0298
5 11 105.02 0.00 1.0 0.00 1.00 0.001
6 7 147.00 0.10 1.0 0.023 0.944 0.0663
6 11 10.5 0.44 1.0 0.00 0.833 0.9606
7 8 102.00 0.12 1.0 0.0353 0.93 0.1489
7 11 21.00 0.44 1.0 0.00 0.833 0.3939
8 9 102.00 0.00 1.0 0.0266 0.987 0.113
9 10 102.00 0.00 1.0 0.0361 0.982 0.1481
10 11 219.43 0.00 1.0 0.00 1.00 0.001
Rev. 7 WOLFCREEKTABLE6.2.1-28BLOWDOWNMASSANDENERGYRELEASEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINETimeMassEnergy(seconds)(1000lbm)(1,000,000Btu)0.0000.000.000.0503.892.21 0.20015.458.800.30023.5013.410.40030.8417.64 0.50037.8321.71 0.65047.9827.66 0.75054.5831.570.90064.2237.321.00070.4941.08 1.30188.4651.91 1.600105.2962.17 1.800115.6868.58 2.500148.1188.87 3.500189.17114.99 5.000242.34149.326.500290.88180.66 8.001334.86209.02 9.501374.62234.8110.502396.87250.31 12.500433.77276.59 14.001455.47292.88 15.001467.12302.00 16.501482.94312.84 18.000495.37320.34 19.000502.43323.85 20.500506.67325.89 20.856506.70325.92Rev.0 WOLFCRERKTABLE6.2.1-28(Sheet2)TimeMassRateEnergyRate(seconds)(1000lbm/sec)(1,000,000Btu/sec)0.00077.72144.2130.02577.72144.213 0.12576.91043.8360.25080.60246.1800.35073.54142.382 0.45069.90140.642 0.57567.63239.687 0.70065.96039.0510.82564.33338.3570.95062.49937.464 1.15059.84836.093 1.45156.14834.217 1.70051.97632.042 2.15046.33428.997 3.00041.06426.123 4.25035.44322.8815.75032.36020.895 7.25129.30318.895 8.75126.50117.19110.00222.23115.493 11.50118.46513.148 13.25114.46210.857 14.50111.6539.126 15.75110.5507.229 17.2508.2864.996 18.5007.0593.509 19.7502.8331.364 20.6780.0610.079 20.8570.0000.000Rev.0 WOLFCREEKTABLE6.2.1-29BLOWDOWNMASSANDENERGYRELEASE0.6DOUBLE-ENDEDPUMPSUCTIONGUILLOTINETimeMassEnerqy(seconds)(1000lbm)(1,000,000btu)0.0000.000.000.0502.641.50 0.20011.946.820.30018.2310.420.45027.3115.65 0.55033.1419.04 0.70041.6524.00 0.85049.6228.690.95054.7431.711.20066.9138.93 1.50081.3347.53 1.80095.6856.12 2.000105.1161.80 3.500168.31100.71 5.000218.08132.35 6.500262.48160.668.000302.91186.42 9.500339.96210.03 11.002372.70231.78 12.501399.88250.49 14.501430.39271.98 16.002448.91285.58 17.501464.12297.17 19.500481.91309.26 21.000493.72315.95 22.500504.40320.93 24.000508.24322.75 24.500508.26322.78Rev.0 WOLFCREEKTable6.2.1-29(Sheet2)TimeMassRateEnergyRate(seconds)(1000lbm/sec)(1,000,000Btu/sec)0.00052.70630.0410.02552.70630.041 0.12561.88935.3460.25062.83236.0180.37560.66134.952 0.50058.27733.807 0.62556.71233.117 0.77553.23731.3040.90051.12330.1921.07548.65328.854 1.35048.08128.661 1.65047.81128.636 1.90047.15128.375 2.75042.13825.941 4.25033.17421.093 5.75029.60218.8737.25026.95517.174 8.75024.69815.74110.25121.80214.480 11.75218.12912.481 13.50115.25810.745 15.25212.3349.062 16.75210.1497.734 18.5018.9016.048 20.2507.8714.455 21.7507.1183.323 23.2502.5611.211 24.2500.0510.066 24.5000.0000.000Rev.0 WOLFCREEKTABLE6.2.1-30BLOWDOWNMASSANDENERGYRELEASETHREEFOOTSQUAREDPUMPSUCTIONSPLITTimeMassEnergy(seconds)(1000lbm)(1,000,000Btu)0.0000.000.000.0501.440.81 0.25010.235.810.40016.739.520.55023.1013.19 0.75031.2717.94 0.90037.1021.36 1.20048.1627.901.60062.0136.151.90071.7942.02 3.500115.5968.71 5.500162.2997.73 7.500204.87124.00 9.500245.37148.7811.503283.46172.28 13.500316.82193.5915.503348.83213.88 18.502390.06240.86 21.002417.60259.87 23.002435.91273.06 26.000458.29290.02 28.000469.29299.10 30.000478.16306.00 32.001484.84310.84 34.000491.97314.50 35.500494.67315.91 37.228497.25316.89 38.000497.86317.10 Rev.0 WOLFCREEKTABLE6.2.1-30(Sheet2)TimeMassRateEnergyRate(seconds)(1000lbm/sec)(1,000,000btu/sec)0.00028.73416.2600.02528.23416.260 0.15043.91824.949 0.32543.29824.751 0.47542.50024.4720.65040.84123.738 0.82538.87222.790 1.05035.87421.799 1.40034.64820.647 1.75032.57719.528 2.70027.37416.684 4.50032.35214.513 6.50021.29213.136 8.50020.24812.39010.50119.02211.73312.50116.69610.66814.50115.98910.131 17.00213.7468.998 19.75211.0197.604 22.0029.1546.593 24.5017.4625.65s 27.0005.5044.540 29.0004.4333.451 31.0003.3412.419 33.0003.5671.831 34.7501.8000.939 36.3641.4890.568 37.6140.7930.272 38.0000.0000.000 Rev.0 WOLFCREEKTABLE6.2.1-31BLOWDOWNMASSANDEMERGYRELEASEDOUBLE-ENDEDHOTLEGGUILLOTINETimeMassEnergy(seconds)(1000lbm)(1,000,000Btu)0.0000.000.000.0503.872.55 0.15012.438.210.25020.0313.10.35027.2717.89 0.50037.6424.58 0.60044.2528.83 0.70050.6732.950.85060.0138.960.95166.0942.87 1.20180.8352.37 1.40092.2959.76 1.700108.8270.43 1.900119.3077.22 3.000169.93110.34 4.000209.60136.425.500265.32172.74 6.500302.15196.29 8.001354.04229.45 9.002382.82248.7410.502419.34273.82 12.001447.87294.31 13.500467.98310.12 15.001480.26320.28 16.000485.24324.53 17.501488.67327.88 18.500488.91328.12 18.766488.92328.12Rev.0 WOLFCREEKTABLE6.2.1-31(Sheet2)TimeMassRateEnergyRate(seconds)(1000lbm/sec)(1,000,000btu/sec)0.00077.09150.8760.02577.09150.876 0.10085.60056.509 0.20076.04249.828 0.30072.53047.1240.42568.98044.489 0.55066.16742.541 0.65064.30841.333 0.77562.17339.989 0.90160.48338.951 1.07658.96737.985 1.30057.47237.033 1.55055.11735.591 1.80052.35033.945 2.45046.03230.103 3.50039.65926.0774.75037.15024.215 6.00036.82923.552 7.25134.58522.101 8.50228.72319.257 9.75224.35816.72711.25119.03813.66912.75013.40810.546 14.2518.1866.771 15.5014.9834.249 16.7512.2872.233 18.0000.2400.238 18.6330.0110.000 18.7660.000Rev.0 WOLFCREEKTABLE6.2.1-32BLOWDOWNMASSANDENERGYRELEASEDOUBLE-ENDEDCOLDLEGGUILLOTINETimeMassEnergy(seconds)(1000lbm)(1,000,000Btu)0.0000.000.000.0503.041.73 0.15011.056.300.30024.4213.930.40033.2518.97 0.50041.9423.93 0.65054.7731.26 0.75063.1736.060.85071.4040.771.00083.5747.74 1.20099.5356.90 1.500122.4770.13 1.700137.0778.59 2.000157.5490.49 3.000214.71123.92 4.000261.31151.635.501323.53189.31 6.501360.97212.45 8.000408.12242.38 9.000432.94258.4610.500462.97278.96 12.001480.65294.01 13.000486.89300.70 14.500495.37307.84 15.500502.88311.89 17.000512.50316.19 18.500516.71317.84 19.000516.71317.84Rev.0 WOLFCREEKTABLE6.2.1-32(Sheet2)TimeMassRateEnergyRate(seconds)(1000lbm/sec)(1,000,000Btu/sec)0.00060.51334.3970.02560.51334.397 0.10080.29845.8190.22589.09950.8700.35088.17550.338 0.45086.99449.665 0.57585.58748.880 0.70083.92347.9620.80082.34047.0960.92581.10146.451 1.10079.78045.805 1.35076.47644.091 1.60072.95042.250 1.85068.29239.698 2.50057.16533.435 3.50046.59627.7044.75141.44725.102 6.00137.43523.132 7.25131.46019.970 8.50024.82016.087 9.75020.01713.66411.25111.78410.029 12.5016.2416.598 13.7505.6584.756 15.0007.5114.052 16.2506.4142.866 17.7502.8021.100 18.7500.0060.008 19.0000.0000.000Rev.0 WOLFCREEKTABE6.2.1-33REFLOODMASSANDENERGYRELEASEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINE(MINIMUMSAFETYINJECTION)SteamReleaseWaterRelease(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)20.90.00.00.00.021.20.10.10.00.0 21.8336.8438.10.00.022.4417.8543.40.00.025.5812.71055.80.00.0 2991135.71471.00.00.0 3091181.61529.33377.7297.2 39.61091.81402.72420.1713.045.01088.21392.11951.8171.850.01050.51338.21736.7152.8 67.8916.41151.8995.687.6 74.5914.11143.7805.770.9 94.6771.7953.00.00.0100.0696.7857.50.00.0 110.7565.0692.20.00.0 130.7445.5541.90.00.0Rev.0 WOLFCREEKTABLE6.2.1-34REFLOODMASSANDENERGYRELEASEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINE(MAXIMUMSAFETYINJECTION)SteamReleaseWaterRelease Time(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)20.90.00.00.00.021.20.20.20.00.0 21.7383.2498.50.00.022.3417.1542.50.00.025.2876.01138.00.00.0 29.41151.01490.84321.4380.3 34.01164.31502.13693.0325.0 44.51052.61346.52733.2240.550.01027.51308.62476.5217.967.8950.41194.61715.4151.0 74.0881.81103.41574.3138.5 87.2840.21042.01430.7125.9 92.1785.2970.7264.423.3 94.1798.5986.2233.220.5100.0771.6949.5278.024.5 124.3632.0376.4457.040.2Rev.0 WOLFCREEKTABLE6.2.1-35POST-BLOWDOWNMASSANDENERGYRELEASE0.6DOUBLE-ENDEDPUMPSUCTIONGUILLOTINESteamReleaseWaterRelease Time(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)24.50.00.00.00.024.80.10.10.00.0 25.4384.2499.00.00.025.9386.4501.80.00.028.9890.71155.40.00.0 33.11118.91446.90.00.0 33.51120.81448.94148.3365.1 37.71181.71522.33624.8319.042.71060.11359.73115.3274.250.01072.11366.82590.1227.9 53.41076.81369.02360.7207.7 77.0858.81072.51596.3140.5 89.8864.31069.51433.4126.1 96.6762.9940.1344.230.3100.0746.3917.6254.322.4 126.2679.8462.2448.939.5126.2163.6198.8455.740.1 200.0143.8174.81155.2101.7 500.0104.1126.41194.9105.21000.081.198.51217.9107.2 1500.072.087.41227.0108.0 1500.086.9105.51212.1336.1 2000.080.697.81218.5337.9 5000.061.774.81237.3343.110000.050.561.21248.5346.2 20000.041.550.11257.5348.7 50000.032.138.61266.9351.3100000.025.331.51272.7352.91000000.012.614.91286.4356.7Entrainmentendsat125.16seconds.Rev.0 WOLFCREEKTABLE6.2.1-36POST-BLOWDOWNMASSANDENERGYRELEASETHREEFOOTSQUAREDPUMPSUCTIONSPLITSteamReleaseWaterRelease Time(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)38.00.00.00.00.038.40.10.10.00.0 39.0322.0418.20.00.039.4358.6465.70.00.042.7795.01031.40.00.0 47.11055.81365.80.00.0 48.01118.21445.43651.5321.3 50.01102.31422.43461.1304.656.61030.81323.02880.0253.478.6937.51183.31819.0160.1 84.5875.91101.01683.0148.1 96.8838.61045.31496.2131.7100.0808.31005.4897.279.0 103.3776.3963.5235.520.7 131.3668.5816.6425.037.4 139.0615.0748.3479.042.2139.0160.9195.7485.842.8 200.0144.9176.21160.5102.1 500.0104.6127.21200.8105.71000.081.398.91224.1107.7 1500.071.987.41233.6108.6 1500.086.9105.61218.5337.9 2000.080.697.91224.9339.6 5000.061.674.71243.9344.910000.050.461.01255.1348.0 20000.041.550.21263.9350.5 50000.032.138.61273.4353.1100000.026.131.51279.2354.71000000.012.614.91292.9358.5Entrainmentendsat138.99seconds.Rev.0 WOLFCREEKTABLE6.2.1-37POST-BLOWDOWNMASSANDENERGYRELEASE:DOUBLE-ENDEDHOTLEGGUILLOTINESteamReleaseWaterRelease Time(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)18.80.00.00.00.019.1.2.00.00.0 19.31255.0250.10.00.019.5649.5329.40.00.020.0801.3359.50.00.0 24.93175.5952.40.00.0 28.32515.2624.70.00.0 29.63809.91068.20.00.035.12527.0624.21783.7157.050.02699.4755.21419.7124.9 65.32726.8839.91045.892.0 68.81717.4490.51032.690.9 69.52313.0709.70.00.0 74.11373.9446.60.00.0 79.31707.1636.50.00.0 85.1850.4363.70.00.085.1178.8213.00.00.0100.0173.1206.11188.9104.6200.0143.3170.61218.7107.2 500.0103.9123.71258.1110.71000.081.296.71280.8112.71500.071.985.71290.1113.5 1500.086.9103.51275.1353.6 2000.080.593.91281.5355.3 5000.061.673.41300.3360.610000.050.560.11311.5363.6 20000.041.549.41320.5366.1 50000.032.138.21329.9368.81000000.012.514.91349.4374.2Entrainmentendsat85-11seconds.Rev.0 WOLFCREEKTABLE6.2.1-38POST-BLOWDOWNMASSANDENERGYRELEASEDOUBLE-ENDEDCOLDLEGGUILLOTINESteamReleaseWaterRelease Time(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)19.00.00.00.00.019.40.50.70.00.0 19.40.50.70.00.020.0183.6239.70.00.020.8215.4281.20.00.0 22.0263.5344.00.00.0 27.0463.4604.90.00.0 29.0420.3548.50.00.030.0446.5582.64791.3421.636.6457.5596.54340.6382.0 48.6430.3560.43772.1331.9 50.0430.5560.53496.6307.7 61.7433.5563.8845.074.4100.0391.8507.6871.276.7 116.9373.5483.2888.878.2 131.2376.1485.9899.579.2200.0312.2401.1968.985.3 252.9267.0342.01020.389.8 252.9132.6169.91022.189.9 500.0104.0133.01284.3113.01000.081.2103.71307.1115.0 1500.071.991.71316.3115.8 1500.086.9110.81301.4360.8 2000.080.5102.51307.7362.6 5000.061.677.91326.6367.810000.050.563.31337.8370.9 20000.041.551.41346.8373.4 50000.032.138.91356.2376.01000000.012.614.9137s.7381.4Entrainmentendsat252.91seconds.Rev.0 WOLFCREEKTABLE6.2.1-39Post-RefloodMassandEnergyReleaseDouble-EndedPumpSuctionGuillotine(MinimumSafetyInjection)TimeSteamReleaseWaterRelease(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)130.7330.7390.5255.672.3174.6330.7390.5255.672.3174.6209.8247.8376.5104.6 185.7204.6241.7381.7106.1 210.7198.9235.0387.4107.7 240.7192.7227.7393.6109.4265.7188.0222.1398.3110.7295.7176.9209.0409.4113.8 325.7172.2203.4414.1115.1 355.7167.7198.1418.6116.3 380.7164.1193.9422.2117.3 410.7160.0189.0426.3118.5 435.7156.7185.1429.6119.4 465.7152.8180.5433.5120.5490.7149.6176.7436.7121.3 520.7145.9172.3440.4122.4 550.7142.2168.0444.1123.4 580.7138.6163.7447.7124.4 605.7135.6160.2450.7125.2 635.7132.1156.0454.2126.2 660.7134.2158.5452.1125.6 690.7130.5154.2455.8126.6 715.7127.5150.5458.8127.5 745.7123.8146.2462.5128.5 775.7124.8147.4461.5128.2 805.7120.9142.8465.4129.3 830.7117.6138.9468.7130.2 860.7118.0139.4468.3130.1 920.7113.6134.2472.7131.3 975.7108.9128.7477.4132.61035.7105.8124.9480.5133.51323.7105.8124.9480.5133.5 1323.7109.1127.2477.2132.5 1400.0107.7125.5478.6132.9 1509.0106.3123.9480.0133.3Rev.0 WOLFCREEKTABLE6.2.1-39(Sheet2)TimeSteamReleaseWaterRelease(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)1509.0Seenotebelow102.1Seenotebelow2000.0Seenotebelow92.1Seenotebelow3775.0Seenotebelow79.7Seenotebelow 3775.0Seenotebelow54.0Seenotebelow 5000.0Seenotebelow49.2Seenotebelow10000.0Seenotebelow40.0Seenotebelow20000.0Seenotebelow33.2Seenotebelow50000.0Seenotebelow24.5Seenotebelow100000.0Seenotebelow19.8Seenotebelow 200000.0Seenotebelow16.0Seenotebelow 500000.0Seenotebelow11.1Seenotebelow1000000.0Seenotebelow8.2SeenotebelowNOTE:Followingswitchovertorecirculationmodeat1509.0seconds,thereleasesareafunctionofthereactordecayheat,thesafetyinjectionsystemflowrate(4800gpm),the steamgeneratordepressurizationenergyreleases,andthe containmentsaturationpressure.Theenergyreleasesabove representtherateofadditionofenergytotheprimary systemduringrecirculationfromdecayheatand depressurization.Steamandwaterreleasestothe containmentarecalculatedasfunctionsoftimeusingthe procedureoutlinedinsection3.2.4ofBN-TOP-3,Rev.4 (Ref.1)Rev.0 WOLFCREEKTABLE6.2.1-40POST-REFLOODMASSANDENERGYRELEASEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINE(MAXIMUMSAFETYINJECTION)TimeSteamReleaseWaterRelease(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)124.3257.3303.71143.7231.2172.4257.3303.71143.7231.2172.4142.3168.21258.7254.1 179.3141.0166.61260.0252.7 204.3137.0161.91264.0252.2 234.3132.9157.01268.1251.3259.3129.9153.51271.1245.2289.3126.8149.81274.2244.2 319.3124.1146.51276.9243.1 349.3121.6143.61279.4241.9 374.3119.7141.41281.3240.9 404.3117.7139.01283.3239.6 434.3115.8136.81285.2238.2 464.3114.0134.71287.0236.7489.3112.7133.11288.3235.5 519.3111.1131.31289.9233.9 549.3109.7129.61291.3232.3 579.3108.4128.01292.6230.6 604.3107.4126.81293.6229.1 634.3106.2125.41294.8227.3 659.3105.2124.31295.8225.8 689.3104.2123.01296.8223.9 714.3103.3122.01297.7226.6 744.3102.3120.91298.7224.5 774.3101.4119.81299.6222.3 804.3100.5118.71300.5224.1 829.399.8117.91301.2222.0 849.099.3117.21301.7220.3Rev.0 WOLFCREEKTABLE6.2.1-40(Sheet2)TimeSteamReleaseWaterRelease(seconds)(lbm/sec)(1000Btu/sec)(lbm/sec)(1000Btu/sec)849.0Seenotebelow93.5Seenotebelow1000.0Seenotebelow90.5Seenotebelow1489.9Seenotebelow80.5Seenotebelow 1489.9Seenotebelow95.5Seenotebelow 2000.0Seenotebelow85.8Seenotebelow 3772.0Seenotebelow73.5Seenotebelow3772.0Seenotebelow54.0Seenotebelow5000.0Seenotebelow49.2Seenotebelow10000.0Seenotebelow40.0Seenotebelow 20000.0Seenotebelow33.2Seenotebelow 50000.0Seenotebelow24.5Seenotebelow100000.0Seenotebelow19.8Seenotebelow 200000.0Seenotebelow16.0Seenotebelow 500000.0Seenotebelow11.1Seenotebelow1000000.0Seenotebelow8.2SeenotebelowNOTE:Followingswitchovertorecirculationmodeat849.0seconds,thereleasesareafunctionofthereactordecayheat,thesafetyinjectionflowrate(9600gpm),thesteamgenerator depressurizationenergyreleases,andthecontainment saturationpressure.Theenergyreleasesaboverepresentthe rateofadditionofenergytotheprimarysystemduring recirculationfromdecayheatanddepressurization.Steamand waterreleasestothecontainmentarecalculatedasfunctions oftimeusingtheprocedureoutlinedinsection3.2.4of BN-TOP-3,Rev.4(Ref.1).Rev.0 WOLF CREEK TABLE 6.2.1-41 DEPRESSURIZATION ENERGY DOUBLE-ENDED PUMP SUCTION GUILLOTINE
(MINIMUM SAFETY INJECTION)
Depressurization Output The post-reflood energy release from the steam generators is terminated when the steam generators are in equilibrium
with the referenced containment design pressure (74.7 psia, 60.0 psig or 307.3
°F). This leaves the following energy stored in the system (above 14.7 psia or 212.0
°F): Energy Remaining (1,000,000 Btu)
Broken loop steam generator 20.9 Unbroken loop steam generator 62.8
Metal energy (thin and thick) 20.8 Core stored 1.8 Total available energy 106.3 Rev. 0 WOLF CREEK TABLE 6.2.1-42 DEPRESSURIZATION ENERGY DOUBLE-ENDED PUMP SUCTION GUILLOTINE
(MAXIMUM SAFETY INJECTION)
Depressurization Output The post-reflood energy release from the steam generators is terminated when the steam generators are in equilibrium with
the referenced containment design pressure (74.7 psia, 60.0 psig or 307.3
°F). This leaves the following energy stored in the system (above 14.7 psia or 212.0
°F): Energy Remaining (1,000,000 Btu)
Broken loop steam generator 20.9 Unbroken loop steam generator 62.8
Metal energy (thin and thick) 21.1 Core stored 1.8 Total available energy 106.6 Rev. 0 WOLFCREEKTABLE6.2.1-43REACTORCOOLANTSYSTEMMASSANDENERGYBALANCEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINE(MINIMUMSAFETYINJECTION)MassBalanceTime(seconds)0.0020.86130.69130.691035.69Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54739.54AddedMassPumpedinjection0.000.0054.7754.77585.37Totaladded0.000.0054.7754.77585.37 Totalavailable739.54739.54794.31794.311324.91 DistributionReactorcoolant504.5240.00113.99113.99113.99 Accumulator235.02192.840.000.000.00 Totalcontents739.54232.84113.99113.99113.99 EffluentBreakflow0.00506.70597.00597.00725.51 ECCSspill0.000.0083.3283.32485.51 Totaleffluent0.00506.70680.32680.321210.92TotalAccountable739.54739.54794.31794.311324.81Rev.0 WOLFCREEKTABLE6.2.1-43(Sheet2)EnergyBalanceTime(seconds)0.0020.86130.69130.691035.69Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50874.50steamgeneratorAddedEnergyPumpedinjection0.000.004.824.8251.51Decayheat0.007.0624.5324.53117.15 Heatfromsecondary0.00-0.61-0.61-0.6116.91 Totaladded0.006.4528.7328.73185.57 Totalavailable874.50880.95903.24903.241060.08 DistributionReactorcoolant304.9011.0222.9122.9122.91 Accumulator20.6816.970.000.000.00 Corestored26.976.035.095.095.09 Thinmetal19.9716.789.849.849.84 Thickmetal33.0333.0328.0628.0628.06 Steamgenerator468.96471.21390.46390.46283.94 Totalcontents874.50555.03456.35456.35349.82 EffluentBreakflow0.00325.92439.56439.56591.43ECCSspill0.000.007.337.33118.82 Totaleffluent0.00325.92446.89446.89710.25TotalAccountable874.50880.95903.24903.241060.08Rev.0 WOLFCREEKTABLE6.2.1-44REACTORCOOLANTSYSTEMMASSANDENERGYBALANCEDOUBLE-ENDEDPUMPSUCTIONGUILLOTINE(MAXIMUMSAFETYINJECTION)MassBalanceTime(seconds)0.0020.86124.30124.301034.30Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54739.54AddedMassPumpedinjection0.000.00131.19131.191406.10Totaladded0.000.00131.19131.191406.10 Totalavailable739.54739.54870.73870.732145.64 DistributionReactorcoolant504.5240.00123.16123.16123.16 Accumulator235.02192.840.000.000.00 Totalcontents739.54232.84123.16123.16123.16 EffluentBreakflow0.00506.70597.78597.78695.89 ECCSspill0.000.00149.79149.791326.59 Totaleffluent0.00506.70747.57747.572022.48TotalAccountable739.54739.54870.73870.732145.64Rev.0 WOLFCREEKTABLE6.2.1-44(Sheet2)EnergyBalanceTime(seconds)0.0020.86124.30124.301034.30Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50874.50steamgeneratorAddedEnergyPumpedinjection0.000.0011.5411.54123.74Decayheat0.007.0623.6323.63117.04 Heatfromsecondary0.00-0.61-0.61-0.6116.91 Totaladded0.006.4534.5634.56257.68 Totalavailable874.50880.95909.06909.061132.18 DistributionReactorcoolant304.9011.0223.6923.6923.69 Accumulator20.6816.970.000.000.00 Corestored26.976.035.095.095.09 Thinmetal19.9716.789.849.849.84 Thickmetal33.0333.0328.3028.3028.30 Steamgenerator468.96471.21389.63389.63284.07 Totalcontents874.50555.03456.55456.55350.99 EffluentBreakflow0.00325.92439.33439.33555.27ECCSspill0.000.0013.1813.18225.93 Totaleffluent0.00325.92452.51452.51781.20TotalAccountable874.50880.95909.06909.061132.18Rev.0 WOLFCREEKTABLE6.2.1-45PRIMARYCOOLANTSYSTEMMASSANDENERGYBALANCE0.6DOUBLE-EMDEDPUMPSUCTIONGUILLOTINEMassBalanceTime(seconds)0.0020.86126.161500.00Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54AddedMassPumpedinjection0.000.00130.401915.04Totaladded0.000.00130.401915.04Totalavailable739.54739.54869.942654.58
DistributionReactorcoolant504.5243.19126.36123.36 Accumulator235.02188.090.000.00 Totalcontents739.54231.27126.36126.36 EffluentBreakflow0.00508.26597.65728.47 ECCSspill0.000.00145.921799.74 Totaleffluent0.00508.26743.572528.21TotalAccountable739.54739.54869.942654.57Rev.0 WOLFCREEKTABLE6.2.1-45(Sheet2)EnergyBalanceTime(seconds)0.0024.50126.161500.00Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50steamgeneeatorPumpedinjection0.008.1124.23155.29Heatfromsecondary0.00-20.26-20.26-20.26Totaladded0.00-12.1515.44303.55 Totalavailable874.50862.35889.941178.05Reactorcoolant304.9011.7324.4024.40Accumulator20.6816.550.000.00 Corestored26.975.095.095.09 Thinmetal19.9716.529.849.84 Thickmetal33.0333.0328.3716.34 Steamgenerator468.96456.65376.89372.62Totalcontents874.50539.57444.59428.29Breakflow0.00322.78432.51591.38ECCSspill0.000.0012.84158.38 Totaleffluent0.00322.78445.35749.76TotalAccountable874.50862.35889.941178.05Rev0 WOLFCREEKTABLE6.2.1-46REACTORCOOLANTSYSTEMMASSANDENERGYBALANCETHREE-FOOT-SQUAREDPUMPSUCTIONSPLITMassBalanceTime(seconds)0.0038.00138.991500.00Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54AddedMassPumpedinjection0.000.00131.091907.83Totaladded0.000.00131.091907.83 Totalavailable739.54739.54870.632647.37 DistributionReactorcoolant504.5275.42158.58158.58 Accumulator235.02166.260.000.00 Totalcontents739.54241.68158.58158.58 EffluentBreakflow0.00497.86582.35711.55 ECCSspill0.000.00129.711777.24 Totaleffluent0.00497.86712.052488.79TotalAccountable739.54739.54870.632647.37Rev.0 WOLFCREEKTABLE6.2.1-46(Sheet2)EnergyBalanceTime(seconds)0.0038.00138.991500.00Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50steamgeneratorAddedEnergyPumpedinjection0.000.0011.54167.89Decayheat0.0013.2128.65157.92 Heatfromsecondary0.00-30.26-30.26-30.26 Totaladded0.00-17.049.93295.55 Totalavailable874.50857.46884.431170.05 DistributionReactorcoolant304.9017.5830.2630.26 Accumulator20.6814.630.000.00 Corestored26.976.325.095.09 Thinmetal19.9715.669.849.84 Thickmetal33.0333.0328.4016.35 Steamgenerator468.96453.14376.02371.64 Totalcontents874.50540.36449.61433.18 EffluentBreakflow0.00317.10423.40580.48ECCSspill0.000.0011.41156.40 Totaleffluent0.00317.10434.82736.87TotalAccountable874.50857.46884.431170.05Rev.0 WOLFCREEKTABLE6.2.1-47REACTORCOOLANTSYSTEMMASSANDENERGYBALANCEDOUBLE-ENDEDHOTLEGGUILLOTINEMassBalanceTime(seconds)0.0018.7785.111500.00Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54AddedMassPumpedinjection0.000.0096.322023.39Totaladded0.000.0096.322023.39 Totalavailable739.54739.54835.862762.93 DistributionReactorcoolant504.5259.29135.07143.18 Accumulator235.02191.330.000.00 Totalcontents739.54250.62135.07143.18 EffluentBreakflow0.00488.92644.63782.27 ECCSspill0.000.0056.161837.48 Totaleffluent0.00488.92700.792619.75TotalAccountable739.54739.54835.862762.93Rev.0 WOLFCREEKTABLE6.2.1-47(Sheet2)EnergyBalanceTime(seconds)0.0018.7785.111500.00Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50steamgeneratorAddedEnergyPumpedinjection0.000.008.48178.06Decayheat0.007.3118.56155.62 Heatfromsecondary0.00-.61-.61-.61 Totaladded0.006.7026.42333.07 Totalavailable874.50881.20900.921207.57 DistributionReactorcoolant304.9015.0226.9427.66 Accumulator20.6816.840.000.00 Corestored26.977.015.095.09 Thinmetal19.9716.939.849.84 Thickmetal33.0333.0329.8216.34 Steamgenerator468.96464.26450.49449.23 Totalcontents874.50553.08522.18508.15 EffluentBreakflow0.00328.12373.80537.72ECCSspill0.000.004.94161.70 Totaleffluent0.00328.12378.74699.42TotalAccountable874.50881.20900.921207.57Rev.0 WOLFCREEKTABLE6.2.1-48REACTORCOOLANTSYSTEMMASSANDENERGYBALANCEDOUBLE-ENDEDCOLDLEGGUILLOTINEMassBalanceTime(seconds)0.0019.00252.911500.00Mass(1000lb)
AvailableInitialRCSandaccumulator739.54739.54739.54739.54AddedMassPumpedinjection0.000.00321.432052.74 Totaladded0.000.00321.432052.74 Totalavailable739.54739.541060.972792.27 DistributionReactorcoolant504.5226.70113.74113.74 Accumulator235.02137.370.000.00 Totalcontents739.54164.07113.74113.74 EffluentBreakflow0.00516.71600.98713.06 ECCSspill0.0058.75346.251965.48 Totaleffluent0.00575.46947.212678.54TotalAccountable739.54739.541060.972792.27Rev.0 WOLFCREEKTABLE6.2.1-48(Sheet2)EnergyBalanceTime(seconds)0.0019.00252.911500.00Energy(1,000,000Btu)
AvailableInRCS,accumulator,and874.50874.50874.50874.50steamgeneratorAddedEnergyPumpedinjection0.000.0028.29180.64Decayheat0.005.5839.09153.84 Heatfromsecondary0.00-0.61-0.61-0.61 Totaladded0.004.9766.77333.87 Totalavailable874.50879.47941.271208.37 DistributionReactorcoolant304.907.3320.4620.46 Accumulator20.6812.090.000.00 Corestored26.9713.605.095.09 Thinmetal19.9716.919.849.84 Thickmetal33.0333.0324.1816.34 Steamgenerator468.96473.50424.42413.67 Totalcontents874.50556.46484.00465.40 EffluentBreakflow0.00317.84426.80570.01ECCSspill0.005.1730.47172.96 Totaleffluent0.00323.01457.27742.97TotalAccountable874.50879.47941.271208.37Rev.0 WOLFCREEKTABLE6.2.1-49PRINCIPALREFLOODPARAMETERSTRANSIENTSDOUBLE-ENDEDPUMPSUCTION(MINIMUMSAFETYINJECTION)
Flooding InjectionCoreDowncomerTimeTemperatureRateCarryoverHeightHeightFlowTotalAccumulatorSpillEnthalpy(seconds)(F)(in/sec)Fraction (ft)(ft)Fraction (ft 3/sec)(ft 3/sec)(ft 3/sec)(Btu/lbm)0.00307.890.0000.0000.000.000.2500.00.00.088.000.26301.8152.8230.0000.620.120.267146.6137.10.088.00 0.34298.6262.2080.0001.01-0.030.419146.1136.60.088.00 0.92294.723.5520.3471.500.970.655140.2131.00.088.00 1.31294.173.4670.4221.571.890.659137.0127.90.088.00 1.41294.033.4670.4381.592.130.660136.2127.10.088.00 1.51293.903.4670.4541.602.360.661135.4126.30.088.00 4.61289.074.3390.6812.008.740.673112.3104.20.088.009.02279.875.6960.7592.5015.170.67883.877.20.088.0010.01277.595.7980.7662.6116.000.67879.373.054.688.0013.64270.025.5620.7803.0016.000.67971.064.447.388.00 18.74261.185.3470.7873.5016.000.68061.654.838.888.00 24.11253.515.1670.7884.0016.000.68253.946.931.988.00 29.68246.864.9990.7884.5016.000.68347.640.326.288.00 35.43241.054.8350.7875.0016.000.68442.334.921.788.00 41.33235.954.6720.7865.5016.000.68538.130.518.188.00 47.41231.394.5090.7846.0016.000.68634.626.915.488.00 53.67227.284.3460.7836.5016.000.68831.924.113.488.00 60.13223.524.1820.7817.0016.000.68929.821.811.988.00 66.79220.074.0180.7807.5016.000.69028.120.011.088.00 70.01218.533.9400.7797.7416.000.69027.519.310.788.00 71.60217.803.9020.7787.8516.000.6918.20.00.088.00 73.70216.883.7480.7778.0015.670.6908.30.00.088.00 81.32214.073.2570.7728.5014.730.6908.60.00.088.00 89.86211.682.8400.7669.0014.040.6898.80.00.088.00 99.36209.672.4980.7609.5013.590.6889.00.00.088.00109.83207.992.2290.75410.0013.380.6879.10.00.088.00Rev.0 WOLFCREEKTABLE6.2.1-50PRINCIPALREFLOODPARAMETERSTRANSIENTSDOUBLE-ENDEDPUMPSUCTION(MINIMUMSAFETYINJECTION)
Flooding InjectionCoreDowncomerTimeTemperatureRateCarryoverHeightHeightFlowTotalAccumulatorSpillEnthalpy(seconds)(F)(in/sec)Fraction (ft)(ft)Fraction (ft3/sec)(ft3/sec)(ft3/sec)(Btu/lbm)0.00307.090.0000.0000.00.000.2500.00.00.088.000.25301.7955.2520.0000.620.160.267159.8137.20.088.00 0.33298.4365.7970.0001.030.000.507159.3136.70.088.00 0.86294.643.7640.3501.501.010.656153.1131.10.088.00 1.41293.803.6430.4541.602.430.662148.3126.60.088.00 4.37288.844.5540.6812.009.180.674124.2104.40.088.00 8.57279.455.8970.7592.5016.000.67894.677.669.488.0013.17269.595.5640.7803.0016.000.67983.466.059.788.00 18.27260.745.3480.7863.5016.000.68073.956.151.088.00 23.63253.105.1680.7884.0016.000.68266.048.044.088.00 29.20246.484.9990.7884.5016.000.68359.641.238.388.00 34.94240.724.8350.7875.0016.000.68454.435.733.788.00 40.84235.654.6710.7865.5016.000.68550.131.130.288.00 46.92231.144.5080.7846.0016.000.68646.727.427.588.00 53.18227.074.3450.7836.5016.000.68844.024.525.488.00 59.64223.364.1800.7817.0016.000.68941.922.124.088.00 66.30219.954.0160.7797.5016.000.69040.220.223.188.00 71.28217.633.8950.7787.8616.000.69020.20.03.688.00 73.19216.783.8500.7788.0016.000.69120.20.03.888.00 80.33213.843.6820.7768.5016.000.69220.50.04.788.00 87.73211.093.5140.7749.0016.000.69320.70.05.788.00 95.42208.513.3440.7719.5016.000.69320.90.06.688.00103.44206.103.1710.76910.0016.000.69421.00.07.588.00Rev.0 WOLFCREEKTABLE6.2.1-51BASESFORANALYSISPlantmodel4loop,12ftcoreCorepower,licenseapplication,MWt3411 Ultimatecorepowerrating,MWt3565 Nominalinlettemperature,F560.0Nominaloutlettemperature,F618.6 Steampressure,psia1000 Rodarray17x17 Totalaccumulatormass,lbm235,020Accumulatortemperature,F120 Assumedcontainmentdesignpressure,psia74.7 AssumedRWSTtemperature,F120Pumpedinjection(assumedforfroth)Minimum,lb/sec586.3Maximum,lb/sec1401RPVvolumebelowbreak,ft 3 2959Rev.0 WOLFCREEKTABLE6.2.1-52SAFETYINJECTIONFLOWRATEVERSUSBACKPRESSUREMinimumMaximumPressureFlowRateFlowRate(psia)(ft 3/sec)(ft 3/sec)14.711.2626.13114.78.03519.84214.72.2577.7411014.71.5302.730Rev.0 TABLE6.2.1-5319-ELEMENTREFLOODMODEL (a)UnbrokenBrokenLoopLoopAreaFormEquivalentHydraulicElementArea(ft 2)(ft 2)Factor(K)Length(ft)Diameter(ft)1.Hotlegnozzle4.5913.770.21160.02.422.Hotlegpiping4.5913.770.05300.02.423.Steamgeneratorinletplenum19.9059.709.2980.02.584.Steamgeneratortubes11.34334.032.98157.820.050675.Steamgeneratoroutletplenum19.9059.704.1420.02.58 6.Crossoverlegpiping5.2415.720.05290.02.587.Pump(forward)4.12512.3(b)0.02.292 8.Coldlegpiping4.1212.30.03320.02.29 9.Coldleginletnozzle(c)-12.30.60680.02.2910.Arounddowncomer(estimate)-10.000.0110.04.00 11.Coldleginletnozzle-4.120.60680.02.2912.Coldlegpiping-4.120.03320.02.2913.Pump(reverse)-4.125(b)0.02.29214.Crossoverlegpiping-5.240.05290.02.58 15.Steamgeneratoroutletplenum-19.904.1420.02.58 16.Steamgeneratortubes-11.3432.98157.820.05067 17.Steamgeneratorinletplenum-19.909.9280.02.58 18.Hotlegpiping-4.590.05300.02.42(a)The19-elementmodelincludeselementsforboththebrokenandunbroken(asanexampleifthereis1brokenloopelementthereare18unbrokenloopelements,foratotalof19elements).(b)Theanalysisaccountsfortransientpumpresistancesduetopumpcoastdown.(c)Thepatharoundthedowncomerisspecifiedonlytoprovidealoopreferencepointforpressureattopofdowncomer.Thefrictionalpressuredropdataareestimatedandprovidenegligiblepressuredrop.Rev.0 WOLFCREEKTABLE6.2.1-54MASSANDENERGYRELEASEHYDRAULICCHARACTERISTICSFORPOST-REFLOOD(ONEINTACTLOOP)AT130.7SECONDS SteamMassSteamItemFlowAreaVelocityDensityHeightP ElevationP FrictionNumberDescription (lb/sec)(ft 2)(ft/sec)(lb/ft 3)(ft)K (psf)(psf)1Downcomer----57.116.0-913.0-2Core154.051.28.780.7315.712.0--187.9-3Upperplenum154.0152.15.910.6022.74.0--90.7-4Steamgenerator2.620.50.730.2244.38.0--354.4-inletplenumAvailableP 280.05Steamgenerator58.811.3--0.327*1.5-0.5-tubes6LoopP58.84.5--0.130-13.9-283.5(steamgeneratorpluspump)P+PLoopTubes 284.0*Densityintubesbasedonmassbalancedqualityinsteadofvoidfractioncorrelation.Rev.0 WOLFCREEKTABLE6.2.1-55MASSANDENERGYRELEASEHYDRAULICCHARACTERISTICSFORPOST-REFLOOD(BROKENLOOP)AT130.7SECONDSSteamLiquidMassMassSteamItemFlowFlowAreaVelocityDensityHeightPElevationPFrictionNumberDescription (lb/sec)(lb/sec)(ft 2)(ft/sec)(lb/ft 3)(ft)K (psf)(psf)1Coldleg(pump)75.0365.04.5--1.0-14.0-2095.72Downcomer-----57.116.0-913.0-3Core154.0-51.28.780.7315.712.0--187.9-4Upperplenum154.0-152.15.910.6022.74.0--90.7-5Steamgenerator146.3-20.541.661.000.28.0--1.4-inletplenum AvailableP 2728.7(a)(b)6Steamgenerator146.30.011.3--0.17157.8-0.0-tubes7Hotleg(steam146.30.04.5--0.171-4.5-431.1 generator)P+PHotLegTubes 431.1(a)Densitybasedonamassbalancedqualityandincludeseffectofnegativehead(H=1/2averagelengthoftubes).(b)Steamgeneratortubeheightisassumedequaltotheheattransferlength(sincetheeffectiveheadisnegligibleandisareasonbleassumption).Rev.0 WOLF CREEK TABLE 6.2.1-56 Spectrum of Main Steamline Ruptures Analyzed Case # Power Level
(%)* Break Size (ft 2) Break Type Remarks 1 102 Full Double-ended
- 2 102 0.6 Double-ended 3 102 0.8 Split 4 75 Full Double-ended
- 5 75 0.55 Double-ended 6 75 0.84 Split 7 50 Full Double-ended
- 8 50 0.45 Double-ended 9 50 0.80 Split 10 25 Full Double-ended
- 11 25 0.33 Double-ended 12 25 0.66 Split 13 0 full Double-ended
- 14 0 0.20 Double-ended 15 0 0.40 Split 16** 0 0.40 Split MSIV Failure
- The power % is scaled to reference the re-rated power of 3579 MWt ** Same as Case 15, except additional failure of the MSIV on faulted loop has been taken into consideration. *** An MSIV failure is conservatively accounted for by combining the LOFTRAN results with the hand-calculated initial blowdown.
Rev. 22 WOLF CREEK TABLE 6.2.1-56A Time Sequence of Events for the Steamline Break Mass and Energy Releases to Containment Case Rx Trip Signal SI Signal Steamline Isolation Signal SI Actuation (sec) Feedwater Isolation (sec) Steamline Isolation (sec) SG Tube Uncovery (sec) SG Dryout (sec) 1 SI LSP LSP 1.389 18.389 18.389 188.0 276.6 2 SI LSP LSP 1.991 8.991 18.991 290.8 580.2 3 OPDT Hi-1 Cont P Hi-2 Cont P 18.7 35.7 86.7 280.2 473.6 4 SI LSP LSP 1.209 18.209 18.209 158.4 300.7 5 SI LSP LSP 2.800 19.800 19.800 330.2 911.2 6 SI Hi-1 Cont P Hi-2 Cont P 16.8 33.8 84.7 285.0 600.8 7 SI LSP LSP 1.133 18.133 18.133 164.2 396.7 8 SI Hi-1 Cont P Hi-2 Cont P 14.7 31.7 79.5 481.2 545.7 9 SI Hi-1 Cont P Hi-2 Cont P 16.7 33.7 89.3 315.6 337.4 10 SI LSP LSP 1.115 18.115 18.115 175.2 192.0 11 SI Hi-1 Cont P Hi-2 Cont P 18.8 35.8 125.5 698.2 788.7 12 SI Hi-1 Cont P Hi-2 Cont P 19.1 36.1 108.2 402.8 443.2 13 SI LSP LSp 1.168 18.168 18.168 200.0 217.0 14 SI Hi-1 Cont P Hi-2 Cont P 29.7 46.7 219.9 1480.2 1744.0 15 SI Hi-1 Cont P Hi-2 Cont P 30.7 47.7 192.7 761.0 832.5 16 SI Hi-1 Cont P Hi-2 Cont P 30.7 47.7 192.7 824.0 895.0 SI - Safety Injection LSP - Low Steamline Pressure OPT - Overpower Delta T Hi-1 Cont P - Hi-1 (6 psig) Containment Pressure Hi-2 Cont P - Hi-2 (20 psig) Containment Pressure Rev. 22
WOLF CREEK TABLE 6.2.1-57 Specific Plant Design Input for MSLB Mass & Energy Release Analysis
Case 1 2 3 4 5 6 7 8 Initial steam generator inventory 1 , lbm 117728 119532 117728 130632 130632 132494 144079 144079 Initial average temperature, F 595.53 595.53 595.53 587.05 587.05 587.05 579.20 579.20 Initial pressurizer
water volume, ft 3 1026.63 1026.63 1026.63 891.68 891.68 891.68 766.72 766.72 Initial feedwater
enthalpy, Btu/lbm 428.32 428.32 428.32 390.00 390.00 390.00 349.42 349.42 Safety injection water
enthalpy 2 , Btu/lbm 555.918 555.918 555.918 555.918 555.918 555.918 555.918 555.918 Break area, ft 2 1.4 0.6 0.80 1.4 0.55 0.84 1.4 0.45 Feedwater isolation
time, sec 18.389 18.991 35.7 18.209 19.8 33.8 18.133 31.7 Steamline isolation
time, sec 18.389 18.991 86.7 18.209 19.8 84.7 18.133 79.5 Maximum AFW flow rate
to the affected SG 3 , gpm 1360 1360 1360 1360 1360 1360 1360 1360 Termination of AFW
addition, sec 1200 1200 1200 1200 1200 1200 1200 1200
Rev. 22
WOLF CREEK TABLE 6.2.1-57 (sheet 2)
Case 9 10 11 12 13 14 15 16 Initial steam generator inventory 1 , lbm 146008 161098 161098 163106 188398 188398 190497 210702 Initial average temperature, F 579.20 571.35 371.35 571.35 563.5 563.5 563.5 563.5 Initial pressurizer
water volume, ft 3 766.72 641.76 641.76 641.76 516.81 516.81 516.81 516.81 Initial feedwater
enthalpy, Btu/lbm 349.42 300.43 300.43 300.43 76.98 76.98 76.98 76.98 Safety injection water
enthalpy 2 , Btu/lbm 555.918 555.918 555.918 555.918 555.918 555.918 555.918 555.918 Break area, ft 2 0.80 1.4 0.33 0.66 1.4 0.20 0.40 0.40 Feedwater isolation
time, sec 33.7 18.115 35.8 36.1 18.168 46.7 47.7 47.7 Steamline isolation
time, sec 89.3 18.115 125.5 108.2 18.168 219.9 192.7 192.7 Maximum AFW flow rate
to the affected SG 3 , gpm 1360 1360 1360 1360 1360 1360 1360 1360 Termination of AFW
addition, sec 1200 1200 1200 1200 1200 1200 1200 1200
Rev 22
WOLF CREEK TABLE 6.2.1-57 (sheet 3)
Note:
- 1) For split breaks, the initial mass of the unisolable portion of the steamline is added to the initial mass of the faulted steam generator.
- 2) Represents the enthalpy at the regions downstream of the boron injection tank (BIT) instead of the water source from the refueling water storage tank (RWST). Mixing of the relatively colder water from the RWST with the hotter water residing in the BIT has been accounted for.
- 3) Corresponds to the situation that the faulted steam generator depressurizes to atmospheric condition and the intact steam generators are maintained at 1200 psia.
Rev. 22 WOLF CREEK TABLE 6.2.1-57A Steamline Break Mass and Energy Release Rates, Case 1, 102% Power, Full Double-Ended Break Time (sec) Mass Release Rate (lbm/sec)
Energy Release Rate (E+6 Btu/sec)
Enthalpy (Btu/lbm) 0 0 0 0 0.2 8857 10.56 1192 0.4 8710 10.39 1192 0.6 8632 10.3 1193 0.8 8515 10.16 1193 1 8443 10.08 1194 1.4 8258 9.862 1194 1.8 8122 9.704 1195 2 8051 9.622 1195 3.2 7633 9.133 1197 3.4 8378 10.03 1197 3.6 8356 10 1197 3.8 8337 9.979 1197 4 8317 9.957 1197 5 8107 9.71 1198 6 7935 9.508 1198 8 7645 9.166 1199 10 7351 8.82 1200 14 6564 7.888 1202 15 6359 7.645 1202 18 5847 7.035 1203 18.4 5783 6.959 1203 18.6 1625 1.956 1204 18.8 1616 1.946 1204 19 1608 1.936 1204 20 1568 1.888 1204 25 1380 1.662 1204 30 1228 1.479 1204 35 1128 1.358 1204 40 1065 1.282 1204 50 998.5 1.202 1204 60 965.4 1.162 1204 70 944.7 1.137 1204 75 936.7 1.127 1203 100 913.5 1.099 1203 125 904 1.088 1203 150 899.4 1.082 1203 175 897.2 1.079 1203 180 897 1.079 1203 185 896.7 1.079 1203 188 896.6 1.079 1203 190 875.4 1.053 1203 195 787.5 0.9466 1202 200 696.9 0.8369 1201 205 611.8 0.7337 1199 Rev. 22 WOLF CREEK TABLE 6.2.1-57A (sheet 2)
Steamline Break Mass and Energy Release Rates, Case 1, 102% Power, Full Double-Ended Break Time (sec) Mass Release Rate (lbm/sec)
Energy Release Rate (E+6 Btu/sec)
Enthalpy (Btu/lbm) 210 525.7 0.6294 1197 215 442.8 0.5288 1194 220 369.3 0.4399 1191 225 310.6 0.369 1188 230 266 0.3153 1185 240 215.2 0.2542 1181 250 194.6 0.2295 1179 260 187.2 0.2207 1178 275 184.2 0.217 1178 300 183.5 0.2162 1178 450.2 183.5 0.2162 1178 600.2 183.5 0.2162 1178 900.2 183.5 0.2162 1178 1200 183.5 0.2162 1178 Rev. 22 WOLF CREEK TABLE 6.2.1-57B Steamline Break Mass and Energy Release Rates, Case 16, Hot Zero Power, 0.40 ft 2 Split Break With MSIV Failure Time (sec) Mass Release Rate (lbm/sec)
Energy Release Rate (E+6 Btu/sec)
Enthalpy (Btu/lbm) 0 0 0 0 0.2 957.8 1.136 1186 0.6 954.3 1.132 1186 1 951.1 1.128 1186 2 943.4 1.119 1186 4 929 1.103 1187 6 916.2 1.088 1188 8 904.6 1.075 1188 10 894 1.063 1189 15 870.1 1.035 1190 20 848.8 1.011 1191 25 830.4 0.9896 1192 30 816.4 0.9733 1192 32 811 0.9671 1193 32.6 809.3 0.9651 1193 32.8 808.8 0.9646 1193 35 804.9 0.9601 1193 40 802.8 0.9576 1193 50 782.3 0.9339 1194 60 759.6 0.9076 1195 75 726.6 0.869 1196 100 676.3 0.8102 1198 125 635.2 0.7618 1199 150 616.6 0.7399 1200 175 612.4 0.7349 1200 190 612.2 0.7346 1200 192.6 612.2 0.7347 1200 192.8 612.2 0.7347 1200 195 598.3 0.7183 1201 200 569.9 0.6847 1201 205 547.8 0.6585 1202 210 530.1 0.6374 1202 220 503.8 0.6061 1203 225 494.1 0.5945 1203 230 486.2 0.5851 1203 240 474.5 0.571 1204 250 466.6 0.5616 1204 275 456.4 0.5494 1204 300 452.6 0.5449 1204 450 450 0.5418 1204 600 451.4 0.5435 1204 750 453 0.5453 1204 800 453.4 0.5458 1204 824 453.5 0.546 1204 830 440.2 0.53 1204 Rev. 22 WOLF CREEK TABLE 6.2.1-57B (sheet 2)
Steamline Break Mass and Energy Release Rates, Case 16, Hot Zero Power, 0.40 ft 2 Split Break With MSIV Failure Time (sec) Mass Release Rate (lbm/sec)
Energy Release Rate (E+6 Btu/sec)
Enthalpy (Btu/lbm) 835 418.4 0.5038 1204 840 387.1 0.4662 1204 845 349.3 0.4207 1204 850 310.4 0.3738 1204 855 276 0.3322 1204 860 248.5 0.2989 1203 865 227.8 0.2738 1202 870 212.5 0.2553 1202 875 201.4 0.2418 1201 880 193.2 0.232 1200 885 187.4 0.2249 1200 890 183.2 0.2198 1200 900 178.1 0.2136 1199 1200 173.1 0.2076 1199 Rev. 22 WOLF CREEK TABLE 6.2.1-57C Containment Fan Cooler Performance Data: Based on 69400 CFM Air Flow with ESW Flowrate =1000 GPM per Cooler and 95 F ESW Water Temperature Inlet Gas Temperature Actual Heat Removal Rate Actual Heat Removal Rate 20% Degradation (See Note)
( F) (Btu/hr) (Btu/sec) (Btu/sec) 125 8,606,664 2,390.74 1,923.99 131 10,653,612 2,959.34 2,368.74 153 18,851,040 5,236.40 4,180.52 190 34,021,272 9,450.35 7,594.62 220 46,670,472 12,964.02 10,417.78 253 60,678,684 16,855.19 13,519.12 277 70,891,380 19,692.05 15,701.33 288 75,580,104 20,994.47 15,701.33 Note: The actual heat removal rates are based on AL-6XN replacement coils. The 20% degradation values are from the original CAC equipment specifications, reduced by 20%.
Rev. 29 WOLF CREEK TABLE 6.2.1-58 MSLB Peak Pressure and Temperature Results
Case Power Level/Break Type Peak Pressure (psig) Time of Peak
Pressure (sec) Peak Temperature
( F) Time of Peak Temperature (sec) 1 102%/Full DER 50.59 1202 364.9 18 2 102%/0.60 ft 2 DER 38.86 1202 296.6 298 3 102%/0.80 ft 2 Split 46.59 1202 303.8 186 4 75%/Full DER 48.66 1202 357.0 18 5 75%/0.55 ft 2 DER 39.50 1202 295.3 322 6 75%/0.84 ft 2 Split 45.57 1202 304.1 192 7 50%/Full DER 49.41 1202 356.8 18 8 50%/0.45 ft 2 DER 42.22 1202 293.4 62 9 50%/0.80 ft 2 Split 46.83 1202 303.4 202 10 25%/Full DER 52.85 1202 362.4 18 11 25%/0.33 ft 2 DER 43.61 1202 272.2 62 12 25%/0.66 ft 2 Split 44.34 1202 299.3 242 13 0%/Full DER 50.98 1202 363.8 18 14 0%/0.20 ft 2 DER 39.82 1204 263.3 1202 15 0%/0.40 ft 2 Split 48.69 1202 289.5 384 16 0%/0.40 ft 2 Split w/MSIV Failure 52.69 1200 293.8 364
Rev. 29 WOLF CREEK TABLE 6.2.1-59 Sequence of Events for Case 10 Peak Calculated Containment Pressure Case for MSLB Time (sec) Event
0.0 Break occurs, blowdown from all four steam generators
1.1 Steamline pressure setpoint for isolation of main feedwater lines and main steam lines reached (360 psig) 14.6 Containment pressure setpoint for actuation of containment sprays reaches (30 psig) 18.0 Peak containment vapor tenmperature of 362.4°F is reached 18.1 Main feedwater line isolation valves closed 18.1 Main steam line isolation valves closed, blowdown from broken loop steam generator and unisolated steam piping only 60.0* Containment sprays start 175.2 Steam generator tube uncovery occurs 192.0 Dryout occurs, blowdown equals auxiliary feedwater addition rate 225.0** Containment air coolers are assumed to begin removing heat.
1200.0 Auxiliary feedwater addition is terminated
1202.0 Peak containment pressure of 52.85 psig is reached
- The containment pressure reaches the containment Hi-3 pressure setpoint (30 psig, including uncertainty) before 27 seconds.
Therefore, full flow containment spray is conservatively assumed to occur at 60 seconds, accounting for time to attain operating speed and design flow of the containment spray pump and fill up the spray lines.
- This response time considered an extended time interval between the time of steamline break initiation/LOOP and the time full containment cooling system air and safety grade cooling water flow is establiished. The delay was increased to bound all potential water hammer modification designs. Following the completion of the water hammer modification design, the actual delay time is less than 70 seconds; however, the use of the 225 second delay time results in conservative results.
Rev. 29 WOLF CREEK TABLE 6.2.1-60 Sequence of Events for Case 1 Peak Calculated Containment Temperature Case for MSLB Time (sec) Event 0.0 Break occurs, blowdown from all four steam generators
1.4 Low steamline pressure setpoint for isolation of main steam lines and main feedwater lines reached
14.9 Containment pressure setpoint for actuation of containment sprays reaches (30 psig) 18.0 Peak containment vapor temperature of 364.9 F is reached 18.4 Main steam line isolation valves closed, blowdown from broken loop steam generator and unisolated steam piping only
18.4 Main feedwater line isolation valves closed
60.0* Containment sprays start
188.0 Steam Generator tube uncovery occurs 225.0** Containment air coolers are assumed to begin removing heat
276.6 Dryout occurs, blowdown equals auxiliary feedwater addition rate 1200.0 Auxiliary feedwater addition is terminated
1202.0 Peak containment pressure of 50.59 psig is reached
- The containment pressure reaches the containment Hi-3 pressure setpoint (30 psig, including uncertainty) before 27 seconds.
Therefore, full flow containment spray is conservatively assumed to
occur at 60 seconds, accounting for time to attain operating speed
and design flow of the containment spray pump and fill up the spray
lines.
- This response time considered an extended time interval between the time of steamline break initiation/LOOP and the time full containment cooling system air and safety grade cooling water flow is established. The delay was increased to bound all potential water hammer modification designs. Following te completion of the water hammer modification design, the actual delay time is less tan 70 seconds; however, the use of the 225 second delay time results in conservative results.
Rev. 29 WOLF CREEK TABLE 6.2.1-61 Deleted Rev. 22 WOLF CREEK TABLE 6.2.1-62 Deleted Rev. 22 WOLF CREEK TABLE 6.2.1-63 BELOCA Mass & Energy Release Data Used for COCO Calculations Time (s) RCP Side Mass Flow Rate (lb/s) RCP Side Energy Flow (BTU/s)
Vessel Side Mass Flow Rate (lb/s) Vessel Side Energy Flow (BTU/s) 0.0 9408.2 5207067 -9.1 0 0.5 25172.1 13846789 53074.0 29215435 1 24638.8 13749897 48465.2 26653226 2 19413.5 11344687 34907.8 19207958 3 13086.7 8113011 28349.9 15669619 4 9175.6 6485948 25311.3 14108456 5 6978.9 5595382 23164.5 13075017 6 6359.8 5279076 21123.0 12184754 7 6005.0 5033120 19415.4 11397310 8 5676.4 4758261 17638.7 10547493 9 5447.0 4483483 16415.4 9776754 10 5393.8 4274180 13807.0 8618746 15 2438.7 2241777 8577.4 4531724 20 474.6 572078 6000.7 1682727 25 67.8 86101 134.2 34519 30 28.9 37264 -29.8 0 35 45.9 59001 6.5 6154 40 94.1 119331 1043.0 136641 45 95.4 121295 5296.1 786287 50 157.0 195982 1056.9 585688 55 95.9 121869 948.4 382474 60 68.6 87743 405.6 259988 65 60.7 77789 344.5 222160 70 55.8 71421 143.3 113647 75 53.7 68810 120.2 99563 80 56.0 71753 125.7 101808 85 54.3 69502 116.4 96356 90 55.2 70574 123.4 101204 Rev. 29 WOLF CREEK TABLE 6.2.1-63 (Sheet 2)
BELOCA Mass & Energy Release Data Used for COCO Calculations 95 58.9 74926 137.5 109960 100 65.8 83084 201.2 138477 110 65.4 82633 200.1 138558 120 61.3 77626 177.1 119744 130 75.7 93618 459.3 213458 140 80.5 97380 444.1 200998 150 61.9 77554 243.3 116303 160 66.9 83458 370.5 172567 170 71.3 88216 338.8 183997 180 68.2 84812 566.5 198845 190 70.1 86088 576.8 209728 200 66.0 82585 438.0 188509 210 75.3 89834 444.0 205182 220 67.1 82774 719.9 228167 230 58.7 73455 386.2 157530 240 62.9 76765 376.8 176575 250 48.1 60381 290.1 142306 260 60.8 72829 308.5 152940 270 56.0 69259 876.4 237684 280 51.7 64261 499.9 174101 290 53.0 65427 210.8 117241 300 53.5 65749 256.0 130689 310 44.8 55949 240.5 109860 320 46.2 57372 206.2 118458 330 51.5 62442 230.4 132012 340 48.5 60013 195.2 129474 350 51.9 63371 628.2 178219 Rev. 29 WOLF CREEK TABLE 6.2.1-63 (Sheet 3)
Deleted Rev. 29 WOLF CREEK TABLE 6.2.1-63 (Sheet 4)
Deleted Rev. 29 WOLF CREEK TABLE 6.2.1-63 (Sheet 5)
Deleted Rev. 29 WOLF CREEK TABLE 6.2.1-64 Deleted Rev. 29 WOLF CREEK TABLE 6.2.1-65 Large-Break LOCA Containment Data Used for Pressure Calculations Maximum Containment Net Free Volume 2,700,000 ft 3 Initial Conditions Minimum initial containment pressure at full power operation 14.7 psia Minimum initial containment temperature at full power operation 90°F Minimum RWST temperature (for Safety Injection) 37.0°F Minimum Temperature outside containment
-60.0°F Minimum Initial containment spray temperature 37.0°F Spray System Maximum number of containment spray pumps in operation 2 Minimum Post-accident containment spray system initiation delay time 15 seconds (with offsite power) 25 seconds (with LOOP)
Maximum spray system flow from all containment spray pumps 7,754 gal/min.
Recirculation Spray Not Applicable Fan Coolers (1) Maximum number of containment fan coolers in operation 4 Minimum fan cooler initiation delay time 35 seconds (with offsite power) 45 seconds (with LOOP)
Containment Purge Maximum number of containment purge lines OPEN at onset of transient 2 Maximum containment purge valve closure time 5 seconds Containment purge valve diameter 18 in. NOTES: (1) The containment fan cooler performance is determined based on:
a) Essential Service Water (ESW) is entering the containment fan cooler at 1000 gpm**.
b) Steam/air mixture leaving the containment fan cooler at a flowrate of 69,400 acfm.
c) ESW temperature of 33°F, with an zero internal fouling factor.
Rev. 29 WOLF CREEK TABLE 6.2.1-66 Large-Break LOCA Containment Heat Sink Data Used for Containment Pressure Calculations Structural Heat Sinks Wall (feet) T Air (°F) Area (ft 2) Height (ft) T initial (°F) 1. 0.000167 paint 0.021 carbon steel 4.0 concrete
-60.0 64919.0 10.0 90.0 2. 0.000167 paint 0.021 carbon steel 3.0 concrete
-60.0 34129.0 10.0 90.0 3. 0.000167 paint 1.5 concrete 0.021 carbon steel, 10.0 concrete 90.0 13538.0 10.0 90.0 4. 1.0 concrete 90.0 8564.0 10.0 90.0 5. 2.0 concrete 90.0 43497.0 10.0 90.0 6. 2.5 concrete 90.0 17061.0 10.0 90.0 7. 0.000167 paint, 0.021 carbon steel, 2.0 concrete 90.0 7821.0 10.0 90.0 8. 0.021 stainless steel, 2.0 concrete 90.0 8708.0 10.0 90.0 9. 0.0001083 zinc coating, 0.005 carbon steel, 2.0 concrete 90.0 8081.0 10.0 90.0 10. 0.0001083 zinc coating, 0.0104 carbon steel 90.0 186183.0 10.0 90.0 11. 0.000167 paint, 0.0104 carbon steel 90.0 17746.0 10.0 90.0 12. 0.000167 paint, 0.0208 carbon steel 90.0 114205.0 10.0 90.0 13. 0.000167 paint, 0.0417 carbon steel 90.0 49101.0 10.0 90.0 14. 0.000167 paint, 0.0833 carbon steel 90.0 31372.0 10.0 90.0 15. 0.000167 paint, 0.1667 carbon steel 90.0 5631.0 10.0 90.0 16. 0.000167 paint, 0.3333 carbon steel 90.0 8355.0 10.0 90.0 17. 0.000167 paint, 0.6667 carbon steel 90.0 189.0 10.0 90.0 18. 0.000167 paint, 1.333 carbon steel 90.0 157.0 10.0 90.0 Rev. 29 WOLF CREEK TABLE 6.2.1-66 (Sheet 2)
Large-Break LOCA Containment Heat Sink Data Used for Containment Pressure Calculations
- 19. 0.08333 carbon steel 90.0 261.0 10.0 90.0 20. 0.1667 carbon steel 90.0 863.0 10.0 90.0 21. 0.333 carbon steel 90.0 522.0 10.0 90.0 22. 0.6667 carbon steel 90.0 424.0 10.0 90.0 23. 1.333 carbon steel 90.0 141.0 10.0 90.0 24. 0.0104 stainless steel 90.0 7827.0 10.0 90.0 25. 0.0208 stainless steel 90.0 5976.0 10.0 90.0 26. 0.0417 stainless steel 90.0 8166.0 10.0 90.0 27. 0.08333 stainless steel 90.0 1321.0 10.0 90.0 28. 0.1667 stainless steel 90.0 20.0 10.0 90.0 Notes: The heat sinks thermal conductivities (BTU/hr/ft/
o F) are: Paint: 0.63 Carbon steel: 30 Stainless steel: 10 Zinc coating: 65 Concrete: 1.2 Rev. 29 WOLF CREEK TABLE 6.2.2-1 COMPARISON OF THE RECIRCULATION SUMP DESIGN WITH EACH OF THE POSITIONS OF REGULATORY GUIDE 1.82 Regulatory Guide 1.82 Position Recirculation Sump Design1. A minimum of two sumps should be provided, each with sufficient capacity to serve one of the redundant halves of the ECCS and CS systems.
Two sumps are provided, and each has sufficient
capacity to serve one of the redundant halves of
the ECCS and CS systems. 2. The redundant sumps should be physically separated from each other and from high energy piping systems by structural barriers, to the
extent practical, to preclude damage to the sump intake filters by whipping pipes or high-
velocity jets of water or steam.
The redundant sumps are physically separated from each other and from high energy piping. 3. The sumps should be located on the lowest floor elevation in the containment exclusive of the reactor vessel cavity. As a minimum, the sump
intake should be protected by two screens: (1)
an outer trash rack and (2) a fine inner screen.
The sump screens should not be depressed below
the floor elevation.
The sumps are located in El.2000, which is the lowest floor elevation in the Reactor Building, exclusive of the reactor cavity. The strainers are made out of steel with structural steel members supporting the top and bottom. The perforated plate is structurally stiffer than the wire screen and does not need trash racks to protect it from damage. The sump strainers are installed in the recirculation sump pits and extend approximately one foot above the Reactor Building floor. The intent is met.
Rev. 20 WOLF CREEK TABLE 6.2.2-1 (Sheet 2)
Regulatory Guide 1.82 Position Recirculation Sump Design
- 4. The floor level in the vicinity of the coolant sump location should slope gradually down away
from the sump.
The floor is level in the vicinity of the sump.
However, a 6-inch concrete curb is provided which
prevents the high-density particles from reaching
the sumps. The intent is met.
- 5. All drains from the upper regions of the reactor building should terminate in such a manner that
direct streams of water, which may contain
entrained debris, will not impinge on the filter
assemblies.
All drains in the upper regions of the reactor building are terminated in such a manner that direct streams of water which may contain entrained
debris will not impinge on the filter assemblies
- 6. A vertically mounted outer trash rack should be provided to prevent large debris from reaching the fine inner screen. The strength of the trash
rack should be considered in protecting the
inner screen from missiles and large debris.
Each sump strainer has approximately 3300 ft 2 of effective surface area that can accomodate the
amount of debris generated and carried to the sumps
following a debris-generating event. The sumps and
strainers are located outside the Secondary Shield
Wall, which protects them from missiles. The
intent is met.
- 7. A vertically mounted fine inner screen should be provided. The design coolant velocity at the
inner screen should be approximately 6 cm/sec
(0.2 ft/sec). The available surface area used
in determining the design coolant velocity
should be based on one-half of the free surface
area of the fine inner screen to conservatively
account for partial blockage. Only the vertical
screens should be considered in determining
available surface area.
The strainers are installed in the sump pit with
each strainer consisting of 72 modules stacked in a
four by four matrix. The approach velocity of the
recirculation coolant flow at the sump strainer
face is less than 0.006 ft/sec. The intent is met.
In addition, in accordance with Generic Letter 2004-02 requirements,a mechanistic analysis has been performed to assess the potential adverse effects of post-accident debris blockage and operation with debris-laden fluids to impede or prevent the recirculation functions of the ECCS and CSS following postulated accidents for which the recirculation of these systems is required. The methodology for this analysis is consistent with that documented in NEI 04-07.
Rev. 22 WOLF CREEK WOLF CREEK TABLE 6.2.2-1 (Sheet 3)
Regulatory Guide 1.82 Position Recirculation Sump Design8. A solid top deck is preferable, and the top deck should be designed to be fully submerged after a LOCA and completion of the safety injection.
The strainers consist of individual modules stacked on top of each other. The top of each module on the top layer contains a perforated plate. The strainers extend approximately one foot above the Reactor Building floor. Therefore, they will be submerged following a Large Break LOCA. For the small break LOCA, a small portion of the upper modules will not be submerged. The intent is met. 9. The trash rack and screens should be designed to withstand the vibratory motion of seismic events without loss of structural integrity.
The strainers are designed to be seismic Category I.10.The size of openings in the fine screen should be based on the minimum restrictions found in
systems served by the sump. The minimum restriction should take into account the overall operability of the system served.
The strainers have a nominal 0.045" hole size. The strainers protect the downstream equipment by removing material from the flow stream that potentially could cause damage. The performated hole size effectively removes particles larger than 0.045" from the fluid stream. This protects the reactor core channels, safety injection valves and other equipment from clogging.
Rev. 20 WOLF CREEK WOLF CREEK TABLE 6.2.2-1 (Sheet 4)
Regulatory Guide 1.82 Position Recirculation Sump Design 11.Pump intake locations in the sump should be carefully considered to prevent degrading effects, such as vortexing on the pump performance.
The pump intake location in the sump is horizontal to limit any degrading effects due to vortexing.
12.Materials for trash racks and screens should be selected to avoid degradation during periods of inactivity and operation and should have a low sensitivity to adverse effects, such as stress-
assisted corrosion, that may be induced by the
chemically reactive spray during LOCA
conditions.
The strainers are made out of stainless steel that has a low sensitivity to corrosion.
13.The trash rack and screen structure should include access openings to facilitate inspection of the structure and pump suction intake.
The inspection of the liner plate in the recirculatin sumps will be remotely done via cameras, boro-scope, mirrors or other devices. The intent is met.
14.Inservice inspection requirements for coolant sump components (trash racks, screens, and pump suction inlets) should include the following:
Inservice inspection requirements consist of visual
examination during every scheduled refueling
downtime.a. Coolant sump components should be inspected during every refueling period downtime, and b. The inspection should be a visual examination of the components for evidence of structural distress or corrosion.
Rev. 20 WOLF CREEK TABLE 6.2.2-2 CONTAINMENT HEAT REMOVAL SYSTEMS COMPONENT DESIGN PARAMETERS Containment Spray Pumps Type Vertical centrifugal
Quantity 2 Design pressure, psig 450 Design temperature, F 300
Motor, hp 500 Service factor 1.15 Start time, sec 4
Design flow rate, gpm 3,165/3,750 (injection/recirculation)
Design head, ft 464/400 (injection/recirculation)
NPSH available, ft See Table 6.2.2-7
Material in contact with fluid Stainless steel
Design codes
Pump ASME Section III, Class 2 Motor NEMA, IEEE 323, 334, 344 Seismic design Category I
Containment Spray Nozzles
Type Whirljet, hollow cone spray nozzles Design flow per nozzle at 15.2 gpm 40 psi P Number of nozzles 197/header Material Stainless steel Rev. 1 WOLF CREEK TABLE 6.2.2-2 (Sheet 2)
Design code ASME Section III, Class 2
Seismic design Category I
Refueling Water Storage Tank
Quantity 1
Type Vertical
Assured Water Volume, gal 394,000
Design temperature, F 120
Design pressure, psig Atmospheric
Material Stainless steel
Design code ASME Section III, Class 2
Seismic design Category I
Containment Spray System Piping
Material Stainless steel
Design code ASME Section III, Class 2
Seismic design Category I
Containment Air Coolers
Quantity 4
Type Draw-through
Duty Btu/hr each
Normal 3.68 x 10 6 (Note 1) (Note 5)
Post Accident Figure 6.2.1-15
Air side flow (normal/accident), acfm each 140,000/69,400
Static pressure (normal/accident),
in. w.g. 3.76/2.38
Water flow (normal/accident), gpm each 925/1,000
Rev. 29 WOLF CREEK TABLE 6.2.2-2 (Sheet 3)
Inlet water temp (normal/accident), F 90/95 Leaving water temp (normal/accident), F 101/244 (277) (Notes 1,2,3, &4)
Inlet air temp (normal/accident), F 120/277 (Note 2)
Leaving air temp (normal/accident), F 100/274 (Notes 1 & 2)
Type of fan Vaneaxial
Arrangement 4
Motor horsepower (normal/accident), hp 150/75
Motor rpm (normal/accident) 1,200/600
Fouling factor 0.002
Material (tube) Cu-Ni
Material (header) Cu-Ni
Design Code ASME Section III, Class 3 Seismic Design Category I Containment Spray System Isolation Valve Encapsulation Tank Manufacturer Richmond Eng.
Quantity 2
Height ft-in. 10 - 9
Diameter, ft-in. 4 - 0
Design pressure, psig 75
Design temperature, F 400
Material Austenitic Stainless Steel Codes and standards ASME Section III, Class 2 Seismic Category I Rev. 11 WOLF CREEK TABLE 6.2.2-2 (Sheet 4)
Note 1 Design value based on 0.002 fouling factor, specified flow rate and inlet temperature, and no tubes plugged or coils removed from service.
Actual heat transfer performance will be sufficient to limit containment
temperature to 120F during normal operation. Corresponding outlet
temperatures may vary.
Note 2 Design value based on 0.002 fouling factor, specified flow rate and inlet temperature, no tubes plugged or coils removed from service, and a
gas inlet temperature of 277F. 277F bounds the saturation temperature
corresponding to the maximum post accident vapor pressure. Actual heat
transfer performance will be sufficient to limit Actual heat transfer
performance (and corresponding outlet temperatures) will be sufficient
to limit the maximum peak containment pressure obtained from a steam
line break event to 48.9 psig.
Note 3 DELETED Note 4 277 F could occur Post-LOCA with clean (unfouled) coil tubes.
Note 5 Duty Btu/hr in normal conditions for Containment Air Cooler "D" cooler is 3.67 x 10
- 6.
Rev. 29 WOLF CREEK TABLE 6.2.2-3
SUMMARY
OF ACCIDENT CHRONOLOGY FOR CONTAINMENT SPRAY SYSTEM FOR LOSS-OF-COOLANT ACCIDENT Injection Phase Time (Sec) Action
0.0 Event, SI signal, and start diesel generators.
3.0 Containment pressure reaches Hi-1 containment pressure setpoint (6 psig), assuming worst case LOCA or MSLB
inside the containment. Time includes instrument lag
time.
9.0 Hi-3 containment pressure setpoint (30 psig)* attained.
12.0 Diesel generators attain rated speed and voltage, including actuation instrument lag time.
12.0 Sequencer energizes motor control centers to open motor-operated valves in spray additive tank discharge
and spray header isolation valves. Maximum valve opening time is 15 seconds.
27.0 Sequencer applies power to containment spray pumps.
27.0 Slowest spray header motor-operated isolation valves reach full open position.
31.0 Containment spray pumps attain operating speed and design flow.
<60.0 Flow is delivered to the containment.
- When containment pressure drops below 3 psig, reset containment spray actuation signal (CSAS).
NOTES: The worst case LOCA inside the containment is assumed to occur at time zero.
Using conservative analyses, spray flow will be delivered to all spray nozzles within 25 seconds after the spray pump starts; however, 33 seconds is assumed for conservatism.
- Actual setpoint for Hi-3 is 27 psig. The safety analysis uses 30 psig for conservatism.
Rev. 15 WOLF CREEK
TABLE 6.2.2-3 (Sheet 2)
Recirculation Phase Time (Minutes) Action 0.0 Upon reaching RWST Level LO LO-2, initiate opening HV-1 and HV-7 0.8 Verify valves HV-1 & HV-7 are open and initiate closing BN-HV-3 & BN-HV-4 Note: Time 00 begins when RWST Level reaches LOLO-2 (Annunciator 00-047C is LIT) or RWST Level is <12% on indicators BN LI-930, BN LI-931, BN LI-932, BN LI-933.
Rev. 23 WOLFCREEKTABLE 6.2.2-3 (Sheet 3)
SUMMARY
OF ACCIDENT CHRONOLOGY FOR CONTAINMENT SPRAY FORMAIN STEAM LINE BREAK WITH OFFSITE POWER AVAILABLE(CASE 7 AND CASE 9) (1)Time (sec)ActionCase 7Case 9 0 0Break occurs, blowdown from all steam generators. 4.2 15.3 (2)Containment pressure Hi-1 setpoint reached. Initiate SI, CIS-A, feedline isolation, etc. Since offsite power is available, th e loadsequencer starts and provides power to the CSS containment isolation valve immediately and 15 seconds later power is supplied t othe containment spray pump. (The CSS components do not actuate until CSMS is generated by a containment Hi-3 pressure signal). 95.0 144.0Containment pressure Hi-3 setpoint reached. CSAS generated which simultaneously open the containment isolation valves andstarts the spray pumps. 99.0 148.0Containment spray pumps reach operating speed. The flow rate has rapidly increased toward runout conditions as flow fills pipe.The resistance of the partially open containment isolation valve rapidly decreases as the circular wedge arises. 110.0 159.0Containment isolation valve reaches the first open position. Runout flow rates are conservatively assumed as flow continues to fillthe spray headers which offer little flow resistance. 125.0 174.0All air is vented from the last spray nozzle as the headers become water solid. The system flow rate rapidly reduc es from runoutconditions to the design flow rate as the nozzles impose the design pressure drop shown on Figure 6.5-1. 1800 1800Mass and energy addition to the containment ends, containment pressure reduces. Containment spray may be terminate d.(1)Table 6.2.1-58 provides information on 16 steam line breaks analyzed for containment pressure and temperature analyses and includes the times at which Hi-1 and Hi-3 containment pressures are reached for each case.(2)As described in Section 15.1.5, low steam line pressure could initiate safety injection sooner than Hi-1; however, the use of Hi-1 is conservative.Rev. 6 WOLF CREEK TABLE 6.2.2-4
SPRAY INJECTION PHASE DURATION
Operator Action Time Length Of Case Flow Condition Single Failure For Spray Switchover Injection (Min.)
Remarks 1 Two trains ECCS None 20 seconds after receipt 22.50 Refer to Table 6.3-11.
Two trains spray of the lo-lo-2 alarm 2 Two trains ECCS RHR/RWST 20 seconds after receipt 21.44 Refer to Table 6.3-12.
Two trains spray Valve fails to of the lo-lo-2 alarm close 3 Two trains ECCS One spray train 30 seconds after receipt 43.7 One train spray fails of the lo-lo-2 alarm 4 Two trains spray One train of 30 seconds after receipt 53.2 ECCS one-train flow One train ECCS ECCS pumps as- of the lo-lo-2 alarm rates are as follows:
sumed to fail RHR 5100 gpm
SI 660 gpm
CC 550 gpm 5 Two trains spray Ctmt spray sump 30 seconds after receipt 26.4 Operator shuts down one Two trains ECCS valve fails to of the lo-lo-2 alarm spray train to protect
open the pump. Rev. 16 WOLFCREEKTABLE6.2.2-5CONTAINMENTSPRAYSYSTEMSINGLE-FAILUREANALYSISComponentMalfunction CommentsContainmentFailstostartTwopumpsprovided;spraypumpoperationofone required.ContainmentFailstoopenTwopumpsprovided, each spraypumpwithaseparatedis-dischargeisolationchargeisolationpumpvalve*valve;operationof one required.ContainmentsumpFailstoopenTwolinesinparallel,recirculationoneeachspraypump; isolationvalveoperationofone required.*Opensoncoincidenceoftwo-out-of-fourHi-3containmentpressure signals.Rev.0 WOLF CREEK TABLE 6.2.2-6 WATER SOURCES AND WATER LOSSES WHICH CONTRIBUTE TO THE WATER LEVEL WITHIN THE REACTOR BUILDING FOLLOWING A LARGE LOCA Water Sources Min. Max. Reactor coolant inventory, lbm 504,520 504,520 Accumulator tanks inventory, lbm 200,300 220,100 Initial atmosphere water vapor, lbm 732 12,245 Containment spray additive solution, lbm 17,540 39,190
RWST, lbm at:
Initiation of ECCS switchover 1,964,000 2,181,000 Containment spray switchover 2,766,000 3,036,000 Long-term recirculation 2,954,000 3,209,000 Total at: Initiation of ECCS switchover, ft 3 44,925 51,051 Containment spray switchover, ft 3 57,605 65,472 Long-term recirculation, ft 3 59,584 66,627 Water Losses Below El 2,000 ft, ft 3 15,685 15,685 Water remaining in RCS, ft 3 1,901 2,651 Trenches below El 2,000 ft, ft 3 147 176 Trenches below El 2,001 ft-4 in., ft 3 100 120 Miscellaneous wetted surfaces, ft 3 582 2,328 Upending pit, ft3 158 159 Water vapor, ft 3 at: Initiation of ECCS switchover 1,563 4,817 Containment spray switchover 813 4,134 Long-term recirculation 198 813
Total at: Initiation of ECCS switchover, ft 3 20,136 25,936 Containment spray switchover, ft 3 19,386 25,253 Long-term recirculation, ft 3 18,771 21,932 Accumulation Volume Available for Buildup From El 2,000 ft to El 2001 ft-4 in., ft 3 9,898 9,948 From El 2,001 ft-4 in. to El. 2,005 ft-4 in., 11,599 11,838 ft3/ft Results Elevation of water at:
Initiation of ECCS switchover 2,002'-1" 2,003'-2" Containment spray switchover 2003'-2 1/2" 2,004'-6" Long-term recirculation 2003' 8" 2,004'-8"
Rev. 23 WOLF CREEK TABLE 6.2.2-6a WATER SOURCES AND WATER LOSSES WHICH CONTRIBUTE TO THE WATER LEVEL WITHIN THE REACTOR BUILDING FOLLOWING A MAIN STEAM LINE BREAK
Water Sources Min. Max. Blowdown + Aux Feedwater mass, lbm (1) 146,510 624,700 Containment spray additive solution, lbm 17,540 39,190
RWST, lbm at:
Initiation of ECCS switchover 2,048,700 2,765,700 Containment spray switchover 2,766,000 3,036,000 Long-term recirculation 2,954,000 3,209,000 Net Volume added excluding Primary side loss due to shrinkage: Initiation of ECCS switchover, ft 3 34,252 47,747 Containment spray switchover, ft 3 46,192 62,186 Long-term recirculation, ft 3 49,239 63,427
Water Losses
- a. Primary side loss due to shrinkage,lbm at: Initiation of ECCS switchover 79,277 79,277 Containment spray switchover 79,277 79,277 Long-term recirculation 79,277 79,277 b. Other losses, ft 3 Below Elevation 2,000 ft 15,685 15,685 Trenches below Elevation 2,000 ft 147 176 Trenches below Elevation 2,001 ft-4 in. 100 120 Miscellaneous wetted surfaces 582 2,328 Upending pit 158 159 Water vapor at:
Initiation of ECCS switchover 1,563 4,817 Containment spray switchover 813 4,134 Long-term recirculation 198 813
Total (b), ft 3 at: Initiation of ECCS switchover 18,235 23,285 Containment spray switchover 17,485 22,602 Long-term recirculation 16,870 19,281 Accumulation Volume Available for Buildup
From Elevation 2,000 ft to Elevation 9,898 9,948 2,001 ft-4 in., ft 3 From Elevation 2,001 ft-4 in. to Elevation 2,005 ft - 4 in., ft 3/ft 11,599 11,838
Rev. 23 WOLF CREEK TABLE 6.2.2-6a (Sheet 2)
Results (1) Min. Max. Elevation of water at:
Initiation of ECCS switchover 2,001'-5" 2,003'-5/16" Containment spray switchover 2,002'-5 11/16" 2,004'-4" Long-term recirculation 2,003'-3/16" 2,004'-6"
(1) The AFW mass addition and the resultant maximum flood depths are based on operator action at 20 minutes to terminate AFW flow to the affected steam generator.
Rev. 23 WOLF CREEK TABLE 6.2.2-7 INPUT AND RESULTS OF NPSH ANALYSIS
Containment Spray Pumps
Static head available (MSLB) 31 ft 3/16 in.
Pump elevation (discharge centerline) 1971 ft 3/4 in.
Suction line losses @ 3,950 gpm 9.56 ft
Available NPSH @ 3,950 gpm (3) 20.1 ft
Required NPSH @ 3,950 gpm 16.5 ft (from Figure 6.2.2-5)
Residual Heat Removal Pumps
NPSH Reference Elevation (2) 1972.07 ft.
Static head available (LOCA)(1) 30.015 ft
Suction line losses @ 4,760 gpm 3.945 ft
Available NPSH @ 4,760 gpm (3) 23.79 ft
Required NPSH @ 4,760 gpm 21.01 ft
(1) Large LOCA conditions are provided for the RHR pumps since the flow rates, line losses, and NPSH required are greater than those associated
with an MSLB wherein the RCS pressure remains above the RHR shutoff head
at switchover to recirculation.
(2) NPSH reference elevation is 3 3/8 inches above the discharge centerline.
(3) Includes 1.724 ft. total head loss across the sump strainer with both the Spray Pump and RHR Pump running in Recirculation, and a 0.56 ft.
allowance for EDG frequency uncertainities.
Rev. 25 WOLFCREEKTABLE6.2.2-8CONTAINMENTAIRCOOLINGSYSTEMComponentMalfunction CommentsContainmentcoolerHousingfailure,Oneunitoutofhousingairbypassesservice.Threeunitscoilsarefunctional.*CoolingcoilsLossofonetrainTwounitsoutofofessentialser-ofservice.Re-vicewaterdundanttrain(twocoolers)is functional.Lossofoneemer-Twounitsoutofgencydieselservice.Redundanttrain(twocoolers)is
functional.FanFailstostartatOneunitoutofhalfspeedservice.Threeunitsarefunc-tional.*FusiblelinkplatesFailstoopen,Oneunitoutofpartialtocom-service.Threepletelossofoneunitsarefunc-cooler,dependingtional.
- upondegreeofrestrictionin ductworksystem
- Consistsoftheredundanttrain(twocoolers)andtheremainingfunctionalcoolerassociatedwiththemalfunctioningunit.Rev.0 WOLF CREEK TABLE 6.2.2-9 SUMP STRAINER APPROACH VELOCITIES FOR LOCA AND MSLB CONDITIONS FLOW FLOOD DEPTH (1) RATE FLOW VELOCITIES - FPS OPERATIONAL PHASE/MODE Min Max gpm Approach to Sump Area (3) At Strainer Surface (5)LARGE LOCA - At ECCS Switchover 2002-1 2003-0 4800 0.07 0.003 - At Ctmt. Spray Switchover 2003-3 2004-4 8750 0.08 0.006 - During Long-Term Cooling (6) 2003-8 2004-5 4800 0.04 0.003 MSLB - At ECCS SwItchover 2001-8 2002-B 1200 0.02 0.001 - At Ctmt. Spray Switchover 2002-9 2003-11 5150 0.06 0.003 - During Long-Term Cooling (7) 2003-3 2004-1 1200 0.01 0.001 NOTES:
(1) Flood depths (minimum and maximum) for each operational mode or phase are taken from Tables 6.2.2-6 and 6.2.2-6a.
(2) Flow velocities are based on the minimum and maximum flood depths. Minimum depths are used for NPSH available calculations and maximum depths are for worst-case flooding analysis.
(3) The sump area approach velocity is based on a point 6 inches in front of the curb. This velocity more accurately describes the maximum velocity associated with debris settlement.
(4) Deleted (5) This flow velocity is based upon the surface area of the sump strainer exposed to flow.
(6) The velocities for long-term cooling following a LOCA assume that the contaimnent spray system operation Is terminated since the cooling function is completed at switchover and iodine removal has been accomplished.
(7) The velocities for long-term cooling following an MSLB assume that contaimnent spray system operation is terminated and the RCS pressure is at 400 psig which is above the shutoff head of the RHR pumps. As noted on Table 6.2.2-6a, isolation of auxiliary feedwater to the broken loop occurs at 10 minutes which terminates
blowdown to the containment. Long-term recovery from an MSLB will be through cooldown using the normal RHR
suction from the primary loop hot legs. Once flow is established from the primary loop suction from the sump
will not be required.
Rev. 20 WOLF CREEK TABLE 6.2.4-1 LISTING OF CONTAINMENT PIPING PENETRATIONS Penetration Section
Number Service Number
Listing of Penetrations Under Category GDC-55
P-21 RHR hot leg injection 5.4.7/6.3
P-22 RCP-B seal water supply 5.4
P-23 CVCS letdown 9.3.4
P-24 RCP seal water return 5.4
P-27 RHR cold leg injection loops 3 and 4 5.4.7/6.3
P-39 RCPC, seal water supply 5.4
P-40 RCPD, seal water supply 5.4
P-41 RCP-A, seal water supply 5.4
P-48 SI pump - B, discharge to hot legs
1 and 4 6.3
P-49 SI pumps crosstie to cold legs 1, 2, 3, and 4 6.3
P-52 RHR pump suction from hot leg loop 4 5.4.7/6.3
P-64 RC loop and pressurizer liquid sample 9.3.2/18.2.3
P-69 Pressurizer vapor sample 9.3.2
P-79 RHR pump suction from hot leg loop 1 5.4.7/6.3
P-80 CVCS charging 9.3.4
P-82 RHR pump discharge to hot leg loops
1 and 2 5.4.7/6.3
P-87 SI pump A discharge to hot leg loops 2 & 3 6.3
P-88 Boron injection supply to cold leg
loops 1, 2, 3, and 4 6.3
P-93 RC loop liquid sample 9.3.2/18.2.3
P-95 Accumulator liquid sample 9.3.2
P-59 Reactor Vessel Level Indication System 18.2.13.2
P-91 Reactor Vessel Level Indication System 18.2.13.2
Rev. 29 WOLF CREEK TABLE 6.2.4-1 (Sheet 2)
LISTING OF CONTAINMENT PIPING PENETRATIONS Penetration Section
Number Service Number
Listing of Penetrations Under Category GDC-56
P-13 Containment recirculation sump to
containment spray pump 6.2.2
P-14 Containment recirculation sump to RHR
pump suction 5.4.7/6.3
P-15 Containment recirculation sump to RHR
pump suction 5.4.7/6.3
P-16 Containment recirculation sump to
containment spray pump 6.2.2
P-25 Reactor make-up water supply 9.2.7
P-26 Reactor coolant drain tank discharge 11.2
P-28 ESW supply to containment air coolers 6.2.2
P-29 ESW return from containment air coolers 6.2.2
P-30 Instrument air 9.3.1
P-32 Containment sump pump discharge 9.3.3
P-34 Containment ILRT test line 6.2.6
P-43 Auxiliary steam-decontamination 12.3
P-44 Reactor coolant drain tank vent 11.2
P-45 Accumulator nitrogen supply 6.3
P-51 ILRT pressure test line 6.2.6
P-53 FPC and clean-up, refueling pool supply 9.1.3
P-54 FPC and clean-up, refueling pool suction 9.1.3
P-55 FPC and clean-up, refueling pool skimmer 9.1.3
P-56 Post-LOCA hydrogen analyzer return 6.2.5
P-56 Containment atmosphere monitor GT-RE-31 9.4.6
return
P-57 TO RCDT (former PASS sampling return) 18.2.3 Rev. 29 WOLF CREEK TABLE 6.2.4-1 (Sheet 3)
LISTING OF CONTAINMENT PIPING PENETRATIONS Penetration Section Number Service Number Listing of Penetrations Under Category GDC-56 (Continued)
P-58 Accumulator fill line from SI pumps 6.3 P-62 Pressurizer relief tank nitrogen supply 5.4 P-63 Service air supply 9.3.1
P-65 Hydrogen purge 6.2.5
P-66 Containment spray supply from pump B 6.2.2
P-67 Fire protection 9.5.1
P-71 ESW supply to containment coolers 6.2.2
P-73 ESW return from containment coolers 6.2.2
P-74 CCW supply 9.2.2
P-75 CCW return 9.2.2
P-76 Cooling water thermal barrier return 9.2.2
P-78 Drain line from steam generator 10.4.8
P-89 Containment spray supply from pump A 6.2.2
P-92 ECCS test line return 6.3 P-97 Post-LOCA hydrogen analyzer return 6.2.5
P-97 Containment atmosphere monitor GT-RE-32 9.4.6 return P-98 Breathing Air Supply 9.5.10 P-99 Post-LOCA hydrogen analyzer supply 6.2 P-99 Containment atmosphere monitor GT-RE-31 9.4.6
supply Rev. 13 WOLF CREEK TABLE 6.2.4-1 (Sheet 4)
LISTING OF CONTAINMENT PIPING PENETRATIONS Penetration Section Number Service Number Listing of Penetrations Under Category GDC-56 (Continued)
P-101 Post-LOCA hydrogen analyzer supply 6.2 P-101 Containment atmosphere monitor GT-RE-32 9.4.6 supply P-103 Containment pressure sensing monitor 6.3/9.4
P-104 Containment pressure sensing monitor 6.3/9.4
V-160 Containment purge 9.4
V-161 Containment purge 9.4 Rev. 0 WOLFCREEKTABLE6.2.4-2DESIGNCOMPARISONTOREGULATORYGUIDE1.141REVISION0,DATEDAPRIL1978,TITLEDCONTAINMENTISOLATIONPROVISIONSFORFLUIDSYSTEMSRegulatoryGuide1.141PositionsWCGSC.REGULATORYPOSITIONTherequirementsandrecommendationsforFigure6.2.4-1showsthearrangementcontainmentisolationoffluidsystemsandjustifiescompliancewiththeintentthatpenetratetheprimarycontainmentofGDC-55,56,and57.Guidelinesprovided oflight-water-cooledreactorsasspeci-byRegulatoryGuide1.11,ANSIN271-1976,fiedtoANSIN271-1976,"ContainmentNRCSRPs6.2.4and6.2.6,andthisguide IsolationProvisionsforFluidSystems,"arethebasesforcompliance.
aregenerallyacceptableandprovidean adequatebasisforcomplyingwiththepertinentcontainmentisolationrequire-mentsofAppendixAto10CFRPart50,subjecttothefollowing:1.Section3.64ofANSIN271-1976states:1.CompliesasshownforPenetrationsP-1,"Theclosedsystemshallbeleaktested2,3,4,5,6,7,8,9,10,11,12,83,84, inaccordancewith5.3ofthisstandard85,and86.
unlessitcanbeshownbyinspection thatsystemintegrityisbeingmaintained forthosesystemsoperatingatapressure equaltoorabovethecontainmentdesign pressure."Thisexceptiontosystem leaktestingisalsoapplicableto closedsystemsinsidethecontainment.2.Section4.2.3ofANSIN271-19762.CompliesasdescribedinSection6.2.4.5.states:"Sealedclosedisolationvalves areunderadministrativecontrolsanddo notrequirepositionindicationintheRev.0 WOLFCREEKTABLE6.2.4-2(Sheet2)RegulatoryGuide1.141Positionscontrolroomforvalvestatus."Sincethe containmentisolationvalvesarecompo-nentsofthecontainmentisolationsystem, whichisanengineered-safety-feature system,allpower-operatedvalvesshould havepositionindicationinthecontrol
room.3.Section9.2.5ofANSIN271-19763.CompliesasdescribedinSection7.3.states:"Diversityinmeansofactua-tionofautomaticisolationvalvesshouldbeconsideredtoprecludecommon modefailure."TheNRCstaff'sposition isthatthereshouldbediversityinthe parameterssensed(i.e.,typesofisola-tionsignals)fortheinitiationof containmentisolation.4.Section4.4.8ofANSIN271-19764.Complies.givesgeneraldesignrequirementsfor closedsystems.Inaddition,allbranch linesandtheirisolationvalvesinclosedsystemsbothinsideandoutsidethecontainmentshouldmeetthedesign criteriaofSection3.5orSection3.6.7ifthebranchlinesconstituteoneofthe containmentisolationbarriers.5.InSection4.6.3ofANSIN271-1976,5.CompliesasdescribedinSection3.11.referenceismadetoRegulatoryGuide1.7, "ControlofCombustibleGasConcentrations inContainmentFollowingaLoss-of-Coolant Accident,"forguidanceindetermining radiationexposuresforaloss-of-coolantaccident.MoreappropriateguidanceisgiveninRegulatoryGuide1.89,"Qualifi-cationofClassIEEquipmentforNuclear PowerPlants."Rev.0 WOLFCREEKTABLE6.2.4-2(Sheet3)RegulatoryGuide1.141Positions6.Section4.14ofANSIN271-19766.Complies.states:"Thepipingbetweenisolation barriersorpipingwhichformspartof isolationbarriersshallmeettherequire-mentsof3.7andapplicablerequirements forisolationbarriers."Pipingbetween isolationbarriersshouldmeettheappli-cablerequirementsofSection3.5or Section3.7.Rev.0 WOLF CREEK TABLE 6.2.5-1 DESIGN DATA FOR CONTAINMENT HYDROGEN CONTROL SYSTEM COMPONENTS Hydrogen Recombiners Quantity 2 per unit Power (each), max/min, kW 75/50
Capacity (each), minimum, scfm 100
Heaters (per recombiner)
Number 4 banks Maximum heat flux, Btu/hr-ft 2 2,850 Maximum sheath temperature, F 1,550 Gas temperatures
Inlet, F 80-155
Outlet of heater section, F 1,150 to 1,450 Exhaust Approx 50 F above ambient
Materials
Outer structure Type 300 series SS
Inner structure Incoloy 800
Heater element sheath Incoloy 800
Weight, lbs 4,500
Codes and standards ASME Sect. IX, UL, NEMA, NFPA, IEEE
279, 308, 323, 344, and 383, ANS Safety
Class 2 Hydrogen Analyzer
Quantity 2 per unit Type Thermal conductivity
Range 0-10 volume percent Channel Accuracy 6 percent of full scale Valves (isolation)
Quantity 10
Type Solenoid-operated
gate valve
Tubing material Stainless steel
Codes and standards IEEE 279, 323, 344, 383, NEMA, ANS Safety Class 2 Rev. 19 WOLF CREEK TABLE 6.2.5-1 (Sheet 2)
Hydrogen Mixing Fans (see note below)
Quantity 4 Type Vaneaxial
Arrangement/AMCA class 4/II
Air flow (normal/accident),
cfm each 85,000/42,500
Static pressure (normal/accident),
in. w.g. each 0.71/0.50 Brake horsepower (normal/accident),
hp each 26.6/9.3
Motor horsepower (normal/accident),
hp each 50/25
Motor rpm (normal/accident) 900/450
Codes and standards (Motor) IEEE Std 334
(Motor) NEMA
(Fan) AMCA
ANS Safety Class 2 NOTE: The hydrogen mixing fans are not required to operate following an accident. Rev. 8 WOLFCREEKTABLE6.2.5-2
SUMMARY
OFASSUMPTIONSUSEDFORHYDROGENGENERATIONFROMRADIOLYSISa.Theaveragefuelexposureis600fullpowerdaysat3,636 MWt.b.Aninsignificantquantityofhydrogenisgeneratedduetotheradiolysisfromthenoblegasisotopes.c.TheguidelinessetforthinRegulatoryGuide1.7were followed:1.100percentofthenoblegasesisreleasedtothe atmosphere.2.50percentofthehalogensand1percentofthesolidspresentinthecoreareintimatelymixedwiththecoolantwater.3.G(H 2)is0.5molecule/100eV.4.G(O 2)is0.25molecule/100eV.5.Thefollowingpercentageoffissionproductradiationenergyisabsorbedbythecoolant:PercentageRadiationTypeLocationofSource0%BetaFuelrods100%BetaCoolant10%GammaFuelrods100%GammaCoolantRev.0 WOLF CREEK TABLE 6.2.5-3 PARAMETERS USED TO DETERMINE HYDROGEN GENERATION Plant power level, MWt 3,636 MWt Containment free volume, ft 3 2.5 x 1O 6 Containment temperature at accident, F 120 F
Weight of zirconium, lb 54,000
Hydrogen generated zirconium-water
reaction, lb-moles 59.20
Corrodible metals Aluminum, zinc
Surface Area of Aluminum, ft 2 1,114*
Surface Area of Zinc, ft 2 550,000*
Surface Weight Area (lbs)
(ft 2)Initial Inventory of Aluminum in Containment*
HVAC dampers 119 32 Source, intermediate and power range detectors 244 83
Control rod drive mechanism
connectors 193 42
Miscellaneous hydraulic
valves 230 86 Rod position indicators 151 81 Flux map drive system 205 88
Refueling machine 28 5
Contingency (NSSS) 250 85
Containment atmosphere
control filtration train 34 20
Polar crane 134 99
Control valves on
containment cooler
standpipes 57 13
RPI system connectors 66 36
Control rod drive connectors
@ intermediate connection
panel 56 12
Reactor cavity cooling fans 200 14
Reactor building elevator, aux. monorail and hoist 25 6
RTD adapter plates (4 total)
and containment jib cranes 1 negligible
Resistor temperature
detectors (13 total) 2 1 Surface Area (ft 2)Initial Inventory of Zinc in Containment*
Inorganic zinc based paint 347,177 (includes containment liner plate, equipment, painted structural steel, conduit, etc.)
Galvanized surfaces (includes 194,244 equipment, cable trays, conduit, etc.)
- Surface areas of aluminum and zinc greater than the actual inventories were used in the hydrogen generation calculation to allow for future additions.
The current inventory is maintained in a station engineering calculation.
Rev. 16 WOLFCREEKTABLE6.2.5-4Thistableisdeleted.TheinformationpreviouslyinthistableisnowinFigures6.2.5-7,6.2.5-8,and6.2.5-9Rev.8 WOLFCREEKTABLE6.2.5-5SINGLEFAILUREANALYSISCONTAINMENTHYDROGENCONTROLSYSTEMComponentMalfunction ConsequencesHydrogenrecombinerRecombinerfailstoRedundantre-subsystemoperateproperlycombineravailableHydrogenanalyzerAnalyzerfailstoRedundantanalyzersubsystemoperateand/oranwithseparateisolationvalvesamplinglines failstoopenavailableHydrogenmixingWithlossofoneTworedundant,full-subsystemtrainofpower,capacitymixingfanstwofansfailtoavailable,powered operatefromanindependentClassIEbusRev.0 WOLFCREEKTABLE6.2.5-6COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.7,REVISION2,DATEDNOVEMBER,1978,TITLED"CONTROLOF COMBUSTIBLEGASCONCENTRATIONSINCONTAINMENTFOLLOWINGALOSS-OF-COOLANTACCIDENT"RegulatoryGuide1.7PositionWCGSPosition1.Eachboilingorpressurizedlight-1.Complies.waternuclearpowerreactorfueledwith uraniumoxidepelletswithincylindricalzircaloycladdingshouldhavethecapabi-litytomeasurethehydrogenconcentration inthecontainment,mixtheatmospherein thecontainment,andcontrolcombustible gasconcentrationswithoutrelyingonpurgingand/orrepressurizationofthecontainmentatmospherefollowingaLOCA.2.Thecontinuouspresenceofredundant2.Equipmentiscombustiblegascontrolequipmentatthepermanently sitemaynotbenecessaryprovideditisinstalled.
availableonanappropriatetimescale.
However,appropriatedesignandproce-duralprovisionsshouldbemadeforits use.Theseprovisionsshouldinclude considerationofshieldingrequirements topermitaccesstotheareawherethe mobilecombustiblegascontrolsystem willbecoupledupandpermitthe couplingoperationtobeexecuted.In addition,centralizedstoragefacil-itiesthatwouldservemultiplesites maybeused,providedthesefacilities includeprovisionssuchasmaintenance, protectivefeatures,testing,and transportationforredundantunitsto aparticularsite.Rev.0 WOLFCREEKTABLE6.2.5-6(Sheet2)RegulatoryGuide1.7PositionWCGSPosition3.Combustiblegascontrolsystems3.Complies.andtheprovisionsformixing,mea-suring,andsamplingshouldmeetthedesign,qualityassurance,redundancy,energysource,andinstrumentation requirementsforanengineeredsafety feature.Inaddition,thesystem itselfshouldnotintroducesafetyproblemsthatmayaffectcontainmentintegrity.Thecombustiblegascon-trolsystemshouldbedesignated SeismicCategoryI(seeRegulatory Guide1.29,"SeismicDesignClassi-fication"),andtheGroupBquality standardsofRegulatoryGuide1.26, "QualityGroupClassificationsandStandardsforWater-,Steam-,and Radioactive-Waste-ContainingCom-ponentsofNuclearPowerPlants,"
shouldbeapplied.4.Allwater-cooledpowerreactors4.Complies.Seeshouldalsohavetheinstalledcapabi-Section lityforacontrolledpurgeofthe6.2.5.2.2.4.
containmentatmospheretoaidin cleanup.Thepurgeorventilation systemmaybeaseparatesystemor partofanexistingsystem.Itneed notberedundantorbedesignated SeismicCategoryI(seeRegulatory Guide1.29),exceptinsofaras portionsofthesystemconstitute partoftheprimarycontainmentboundaryorcontainfilters.5.Theparametervalueslistedin5.CompliesbyTable1oftheguideshouldbeusedusing incalculatinghydrogenandoxygenRegulatory gasconcentrationsincontainmentsGuide1.7 andevaluatingdesignsprovidedtoparameter controlandtopurgecombustiblevalues.
gasesevolvedinthecourseof loss-of-coolantaccidents.These valuesmaybechangedonthebasis ofadditionalexperimentalevidence andanalyses.Rev.0 WOLFCREEKTABLE6.2.5-6(Sheet3)RegulatoryGuide1.7PositionWCGSPosition6.Materialswithinthecontainment6.Complies.thatwouldyieldhydrogengasduetoTable6.2.5-3 corrosionfromtheemergencycoolingprovidesorcontainmentspraysolutionsshouldthesourcebeidentified,andtheiruseshouldinventories.
belimitedasmuchaspractical.Rev.0
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 1:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,ContainmentPressurevs.Time
50 40 @ 30 -w a: ::l (/) (/) w f 20 10
- 0 1E-01 45.9 PSIG @ t = 120 SECONDS 1E+00 1E+01 Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-2 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION.
2 AIR COOLERS. CONTAINMENT PRESSURE vs. TIME 1E+02 1E+03 1E+04 1E+05 1E+06 TIME (SECONDS)
- 1E+07 *
- 50 40 B 30 (/) 0.. -w a: :::> (/) (/) w f 20 10 0 1E-01 1E+00
- 45.4 PSIG @ 115 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANAL VSIS REPORT FIGURE 6.2.1-3 0.6 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, CONTAINMENT PRESSURE vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS)
- 1E+04 50 40 G' 30 (/) 0.. -w cc Gl) w f 20 10
- 0 1E-01 1E+00 46.0 PSIG @ 140 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-4 3.0 SQUARE FOOT PUMP SUCTION SPLIT BREAK, MAXIMUM SAFETY INJECTION 2 AIR COOLERS CONTAINMENT vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS)
- 1E+04 *
- *
- 50 40 5 3o (/) 0.. -w a: ::> (/) (/) w g: 20 10 0 1E-01 1E+00 41.7 PSIG@ 17 SEC Rev. 6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-5 DOUBLE-ENDED HOT LEG GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION 2 AIR COOLERS CONTAINMENT PRESSURE vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS) 1E+04 50 40 (5 30 (/) a.. w cc :::> (/) (/) w g: 20 10 38.4 PSIG @ 15 SEC Rev. 6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-6 DOUBLE-ENDED COLD LEG GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION 2 AIR COOLERS CONTAINMENT PRESSURE vs. TIME 0 1E-01 1E+00 1E+01 1E+02 1E+03 1E+04 TIME (SECONDS)
- _. _____
- __ j
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 7:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,ContainmentVaporTemperaturevs.Time
350 300 C) 250 w a -w 0: :l 0: w 200 w t-150 100 1E-01
- 301.7 DEG. F @ t = 60 SECONDS Rev.6 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-8 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, CONTAINMENT VAPOR TEMPERATURE vs. TIME 1E+00 1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 TIME (SECONDS)
- 1E+07 *
- 350 300 (!) 250 w a -w a: ::J 1-a: w 200 w 1-150 100 1E-01 1E+00
- 302.7 DEG. F@ t = 60 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-9 0.6 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, CONTAINMENT VAPOR TEMPERATURE vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS)
- 1E+04 u: C) w a -w a: :::> 1-<( a: w Q_ ::E w 1-300 250 200 150 100 1E-01
- 1E+00 287 DEG. F @ t = 60 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-10 3.0 SQUARE FOOT PUMP SUCTION SPLIT BREAK, MAXIMUM SAFETY INJECTION 2 AIR COOLERS, CONTAINMENT VAPOR TEMPERATURE vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS)
- 1E+04 *
- u: C) w a -w a: :::> 1-<( a: w Q.. w 1-300 250 200 150 100 1E-01 1E+00
- 265.8 DEG. F@ t = 17 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-11 DOUBLE-ENDED HOT LEG GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION 2 AIR COOLERS, CONTAINMENT VAPOR TEMPERATURE vs. TIME 1E+01 1E+02 1E+03 TIME (SECONDS)
- 1E+04
-LL' . C) w a -w a: :::l 1-<{ a: w 0.. :E w 1-300 250 200 150 100 1E-01
- 1E+00 274.1 DEG. F@ t = 60 SEC Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-12 DOUBLE-ENDED COLO LEG GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION.
2 AIR COOLERS, CONTAINMENT VAPOR TEMPERATURE vs. TIME 1E+01 1E +02 1E+03 TIME (SECONDS)
- 1E+04
- REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 13:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,CondensingHeatTransferCoefficientvs.Time
250 -200 u. I I a= :::> !z w u H: w 0 0 a= w u. en z <( .a= 1-J: 150 100 50 0 1E-01
- 1E+OO 1E+01 Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-14 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, CONDENSING HEAT TRANSFER COEFFICIENT vs. TIME 1E+02 1E+03 1E+04 1E+05 1E+06 TIME (SECONDS)
- 1E+07 * ,,
- 10,000,000 20,000,000 30,000,000 40,000,000 50,000,000 60,000,000 70,000,000 80,000,000050100150200250300350400450HEAT REMOVAL RATE (BTU/HR) CONTAINMENT SATURATION TEMPERATURE (DEG. F) FIGURE 6.2.1-15 CONTAINMENT AIR COOLER DUTY CURVE, HEAT REMOVAL RATE vs. TEMPERATURE WOLF CREEK UPDATED SAFETY ANALYSIS REPORT DESIGN PERFORMANCE PER COOLER BASED ON ESW FLOW RATE OF 1000 GPM PER COOLER AND 69,400 ACFM GAS FLOW RATE. ACTUAL MAY BE LESS DUE TO ISOLATION OF LEAKS OR OTHER FACTORS, BUT WILL ALWAYS EQUAL OR EXCEED THE ANALYSIS VALUES PERFORMANCE ASSUMED IN CONTAINMENT P/T ANALYSIS 5 2 IQ-1 FISSION PRODUCT DECAY 5 U-238 CAPTURE DECAY 2 IQ-3 IQ-1 2 5 2 CREEK I NOTES: I. APPROPRIATE UNCERTAINTIES ARE ALREADY INCLUDED.
- 2. UP TO 103 SECONDS INFINITE OPERATION DATA COINCIDES WITH THE DATA SHOWN. RESIDUAL FISSIONS FOR q% SHUTDOWN 5 O
2 SHUTDOWN 5 ( 2
..
..
=. =.
FIG U E _ :? *: _1 -l6 REACTOR DECAY POWER -... . (
"\
,: Rev. 0 . . (SHEET. 1 j .
NOTES: I. APPROPRIATE UNCERTAINTIES ARE ALREADY INCLUDED.
5 2 a::: IQ-2 G.. ..... 5 ""' I= ""' 3t -2 Cl) ..... ..... ""' 3t 10-3 5 2 5 2 5 10 6 2 5 10 7 2 5 2 5 TIME AFTER SHUTDOWN (SECONDS)
- ->
- c '":* FIGURE 6.2.1-16 POWER' ' ,, ... , . ' Rev. 0 1 (SHEET 2)
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 17:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,SumpTemperaturevs.Time
- 240 220
(!) w 0 -w *a: :::> a: :::E D.. :::E 180 :::> 160 (/) 140 120 1E+00 1E+01
- 1E+02 1E+03 1E+04 TIME (SECONDS)
Rev.6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-18 DOUBLE-ENDED PUMP SUCTION GUillOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, CONTAINMENT SUMP TEMPERATURE vs. TIME 1E+05 1E+06
- 1E+07 L.....------------J
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 19:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,TotalAirCoolerHeatRemovalRatevs.Time
- << :I: -::l 1-e w 1-< a: ..J < > 0 ::! w a: 1-:I: a: w ..J 0 0 0 !!: < 1E+08 8E+07 6E+07 4E+07 2E+07 OE+OO 1E+00 *
- WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-20 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, TOTAL AIR COOLER HEAT REMOVAL RATE vs. TIME Rev.6 1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07 TIME (SECONDS)
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 21:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,RHRHeatExchangerHeatRemovalRatevs.Time
- ir :I: -::J ... m -::J c a: w (!) z < :I: u X w ... < w :I: a: :I: a: 2.5E+08 2E+08 1.5E+08 1E+08 5E+07 OE+OO 1E+00 1E+01 *
- WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-22 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, RHR HEAT EXCHANGER HEAT REMOVAL RATE vs. TIME Rev.6 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07 TIME (SECONDS)
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 23:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,HeatSinkTotalHeatTransferRatevs.Time
- co 0 + w ..... )( iC J: -:::>> w ._ c( a: a: w LL (I) z c( a: ._ ._ J: 120 100 80 60 40 20 0 (20) 1E+00 *
- WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-24 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, HEAT SINK TOTAL HEAT TRANSFER RATE vs. TIME Rev.6 1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07 TIME (SECONDS)
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 25:DoubleEndedPumpSuctionGuillotineBreakMinimumSafetyInjection,2AirCoolers,EnergyInventoryvs.TimeTOTALENERGY ATMOSENERGY SUMPENERGY HEATREMOVEDBYRHRHX HEATREMOVEDBYAIRCOOLERS HEATREMOVEDBYSPRAYS NETHEATTRANSFERREDTOHEATSINKS 40 20 0 20 40 60 80100 1E011E+001E+011E+021E+031E+041E+05 ENERGYINVERNTORY(10 7BTU)TIME(SECONDS)
- 100 80 I' 60 0 + w )( 5 ti 40 -> a: 0 w 20 > > (!) a: w z 0 w (20) (40) Rev. 6 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-26 DOUBLE-ENDED PUMP SUCTION GUILLOTINE BREAK MAXIMUM SAFETY INJECTION, 2 AIR COOLERS, ENERGY INVENTORY vs. TIME CONTAINMENT ATMOSPHERE ENERGY 1E+00 1E+01
- TOTAL ENERGY " 1E+02 1E+03 TIME (SECONDS)
HEAT REMOVED BY RHR HX ""-HEAT REMOVED BY SPRAYS 1E+04 * . 1E+05 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 6.2.1-27 Reactor Cavity Analysis Nodalization Scheme Elevation View Rev. 1 ----
--
NEUTRON SHIELDING AND l7 SUPPORT STRUCTURES 1 2 ..... .
.. . ........ . *.
FOAMGLASOR EQUI\,1.
v1 ......
- INSULATION
""::!--.f-.
I r-4---.J I . ..1. .I -...
.. ------: *)-1-ffi-J-i \ '+') I.L r 2014.5' '+" r 2011.o* .1"!1 4\ 2002.1' 5 1994.6' 6 .;..t . ..
- 1970.5' .* ..
- ..;. ....
' ..... ,,._..,. ' ... f '.I tl :' ***
WOLF CREEK FOAMGLAS EQUIV.
+X PLAN AT EL. 2021"*7 1/2" I \+Y Rev. 0 ... FIGURE 6. 2.1-2'8 REACTOR CAVITY NODALIZATION SCHEME -LtVEL 1 1 *REACTOR BUILDING PLAN AT EL. 2014-6" WOLF CREEK Rev. 0 lfOLP CR UPDATED SAI'Bifr 'l" r.
- IS .REPORT. FIGURE '§ :2: 1-29 *,.
CAVIT-Y ANALYSIS
-LEVEL 2 WOLF CREEK 270° 10° REACTOR REACTOR VESSEL BUILDING REACTOR BUILDING Rev. 0 WOLF CREEK UPDA'l'BD SA!'l!:,'l'Y Alli.LYSIS' REPOR'l' . ' FIGURE-6.2.1-30 REM>TOR "CAVITY AN>ALYSIS NODAI!IlATION., SCHE-ME -LPVEL -.'3
+Y WOLF CREEK +X PLAN EL. 2001'*10" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT -fiGURE REACTOR CAVITY ANALYSIS N 0 D A CfZ'A T ION Sc H £'M E -L EVE L 4 < *
---**
-*** ,. ...... < **
+Y WOLF CREEK +X PLAN EL. 2001'-10" Rev. 0 WOLF CREEK OPDA"l'ED SAFETY-ANALYSIS REPOR"l' FIGURE 6.2.1:--32 REACTOR CAVITY ANALYSIS NODALIZATION SCHEME -LEVEL 5 ' ... " ..
'! . . .. . .... * ** ...... . .,.... . .. , . . . . . . .
- WOLF CREEK EL.
41 Rev. 0 J. WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1_-33 ELEVATION ON C 1 REACTOR BLDG., NODALIZATIQN SCHEME FOR COMPARTMENT LEVEL 6 WOLF CREEK ... ----------------------------------------------------------------------------------------------, -0 N -0 C! 8 -0 0 -o CXl en 0.. w a: ::::> 8 0.. ci co 0 N 3 WOLF CREEK ANALYSIS FIGURE 6.2.1-34 REACTOR CAVITY PRESSURE -TEMPERATURE ANALYSIS, 150 SQ. IN COlD-LEG BREAK, COMPARTMENT PRESSURES ON LEVEL 3 Rev. 0 4 COMPARTMENT 7 COMPARTMENT 6 .._--COMPARTMENT 5 l_ ________ L_ ________
.00 .10 .20 . . . TIME (SECONDS) iii !:!:-w a: ::> w a: Q.. 0 0 WOLF CREEK
-g 0 N -g 0 0 8 0 co 0 0 0 co 0 q 0 N WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-35 REACTOR CAVITY TEMPERATURE ANALYSIS, 150 SQ. IN COLD-LEG COMPARTMENT PRESSURES LEVEL 2 Rev. 0 COMPARTMENT 9 COMPARTMENT10 COMPARTMENT 11 COMPARTMENT15
.60 . 70 .80 .90 1.00 .00 .10 .20 .30 .40 .50 TIME (SECONDS)
WOLF CREEK --COMPARTMENT 17 COMPARTMENT 23 WOLF CREEK UPDATED SAFETY ANALYSIS R£PORT FIGURE 6.2.1-36 REACtOR CAVITY TEMPERATURE ANALYSIS, 150 SQ. IN. COLD-LEG BRE'AK, COMPARTMENT PRESSURES QN LEVEL 4 Rev. 0 ENT20 COMPARTMENT 22 co .00 .10 20 .30 .50 .60 .70 .80 .90 1.00 TIME (SECONDS)
WOLF CREEK 0
..,., 8 <(ai _N (/) 2: w a:. :::1 (/) (/) w a: Q. g u:;: ..... 0 ---COMPARTMENT
- 25. COMPARTMENT 28 WOLF CREEK UPDATED SAFE'l'Y*ANALYSIS REPORT FIGURE 6.2.1-37 REACTOR CAVITY TEMPERATURE ANALYSIS.
150 SQ. IN. COLD-LEG BREAK. COMPARTMENT PRESSURES ON LEVEL 5 Rev. 0 ENT32 COMPARTMENT 31 __________
._ _________
._ ________
.... rio .to .20 .30 ,40 .50 .60 .70 .BO .90 1.00 TIME (SECONDS)
WOLF CREEK re 8. QO ... WOLF CREEK . DPDA'rBD SAFE'rr AB&r.rsrs RBPORT FIGURE 6.2.1-38 RiACTOR CAVITY TEMPERATUR£ ANAlYSIS, 150 SQ. IN. . COLD-LEG BREAK,. COMPARTMENT
- PRESSURES oN-LEVEL 6 Rev. 0 COMPARTMENT
...; 30 .40 .50 .60 . -.oo .10 .20 . TIME (SECONDS)
WOLF CREEK &. 0 CX) & 0 .... 8 g 8 g & 0 N .10 .20 .30 .40 .50 TIME (SECONDS)
WOLF CREEK ANALYSIS FIGURE 6.2.1-39 REACTOR CAVITY TEMPERATURE ANA(YSIS.
150 SQ. IN. CoLD-LEG BREAK, COMPARTMENT PRESSURES ON LEVEL 1 Rev. 0 --COMPARTMENT 49 .60 .70 .80 .90 1.00 WOLF CREEK 2ao.oo---------------------------------------------------, t 0
- u) III ...J w (.) a: 240.00 200.00 l 160.00 120.00 80.00 40.00 . 20 .30 FORCE IN THE X DIRECTION WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-40 REACTOR CAVITY TEMPERATURE
- ANALYSIS, 150 SQ. IN. COLD-LEG BREAK, HORIZONTAL FORCE COMPONENTS ON RPV ' Rev. 0 FORCE IN THEY DIRECTION
.40 .50 .60 .70 .80 .90 TIME (SECONDS) 1.00 WOLF CREEK 120.00--------------------------------------------------
100.00 80.00' 1\. ' MOMENT ABOUT Y AXIS
- > ; 60.00 1--40.00 20.00 MOMENT ABOUT X AXIS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-41 REACTOR CAVITY TEMPERATURE ANALYSIS, 150 SQ. IN. IN COLD-LEG BREAK -TOTAL
.00 .10 20 .30 .40 .50 .60 .70 .80 .90 1.00 TIME (SECONDS)
WOLF CREEK 60.00------------------------------------------------------------------------------------------------------., .....
- 50.00 40.00
_. w 0 a: 0 IL _. j: a: 20.00 w > 10.00 . 00 Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-42 REACTOR CAVITY ANALYSIS, 150 SQ. IN . IN BREAK TIME (SECONDS)
WOLF CREEK -N_., .
.***' . 'i ". \'.\ ________
.. 0 -NODE NUMBER -------NODE BOUNDARY ... **a:. LEVEL 1 EL. 2001'-4" TO 2018'-4" I I I I 0 --------Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT F.IGURE 6.2.1-43 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 1 . *. *;.: ......
'1 .*.
WOLF CREEK ----,r----------. ..
,., .. : ... c.* 0 @ .. *-.:,;: ..
________ ..,
CONTAINMENT LEVEL 1 NODENUMBER EL. 2001'-4" TO 2018'-4" ----NODE BOUNDARY Rev. 0 WOLF CREEK OPDA'l'ED SAFE'l'Y ANALYSIS REPOR'l' FIGURE 6.2.1-44 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS NODALIZAJION SCHEME -LEVEL 1 WOLF CREEK . 0 \ \ \ \ ") ... , LEVEL 2 EL. 2018'-4" TO 2025'-0" I I I I @ 0 ..... .. -..
.. . ' . *. :' --D ----------@ Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.f-45 STEAM GENERATOR LOOP. COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 2 WOLF CREEK ---0 **
- .. *) i*.J
@CONTAINMENT LEVEL 2 EL. 2018'-4" TO 2025'.0" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-46 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -. LEVEL 2 WOLF CREEK .,,. *:::* .....
..... LEVEL 2A PRESSURIZER COMPT. ONLY 'EL. 2025'-0" TO 2029'-6 1/4" Rev. 0 WOLF CREEK ANALYSIS FIGURE 6.2.1:-47 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 2A
._ ____ a.,.. ___ ,.._._ .*. , ,.).*
.... : .. .... p -------WOLF CREEK 0 :1: _ _ _, I!,,---.. 4 .. ,.,
i,*! .. I. 'At. *r *' 1-'--* .& ( -------LEVEL 3 EL. 2025'0" TO 2045'6" (EL. 2029'-0 1/4" TO 2044'-0" IN PRESSURIZER COMPT.) Rev. 0 WOLF CREEK UPDATED ANALYSIS REPORT FIGURE 6.2.1-48' STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 3 WOLF CREEK 0 LEVEL 3 EL. 2025'-Q" TO 2045'-6" .... _ ------0 FIGURE 6.2.1-49 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 3 ' I I LEVEL4 I -2045'-6" TO
. (EL.2044'-6"T02QOO'.()"INPRESSURIZERCOMPT.)
_____________
_ I Rev. 1 UPDAT£0 SAFETY ANALYSIS REPORT Figure 6. 2.1-50 Steam Generator Loop Compartment Analysis Nodalization Scheme -Level 4 I WOLF CREEK ...
.. -.
<,;
' * -Ia'
- LEVEL4 EL. 2045'-6" TO Rev. 0 WOLP CREEK UPDA'fBD SUTn'* A!t&LYSIS RBPOH FIGlJRT 6.-2. 1-51 ... STEAM GENERATOR LOOP COMPARTMENT ANALYSIS.
NODAtiZATION SCHEME -LEvEL 4 I IDLF CREEK I I @ ; ... 0 LEVEL 5 EL. 2060'-0" TO 2068'-8" Rev. 0 WOLP CREBK UPDA'l'BD SAPE'l'Y ANALYSIS REPOR'l' *FIGURE. 6.
STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEVEL 5-.
WOLF CREEK LEVEL 5 EL. 2060'-0" TO 2068'-8" Rev. 0 WOLF CUEI.t UPDA'l'BD SAPB'l'Y .ARU.YSIS RBPOR'l' FIG U R E
- 1 :-53 STEAM GENERATOR lOOP COMPARTMENT
- ANALYSIS, SCHEME -LEVEL 5 WOLF CREEK LEVEL6 EL. 2068'-8" TO 2086' ..() 3/4" (EL. 2068'8" TO 2090'4" IN PRESSURIZER COMPT.) Rev. 0 WOL CRBBK UPDA.-r-BD SAFB!' . SIS* RBPOR!' FIGURE 6.2-1-54 STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODALIZATION SCHEME -LEV£{ '6 .
I CREEK ! '5a'
.. .._ ______ _, **
LEVEL6 EL. 2068' -8" TO 2086' -0 3/4" Rev. 0 WOLP CRBBI: UPDATED SAFETY ARALYSIS REPORT FIGURE 6.2.1-55 \r, ' ' '::: STEAM GENERATOR LOOP COMPARTMENT ANALYSIS, NODAtizt{T*ION StHEM£ -'.
"*
w a: ::> CJ) CJ) W. a: O-w :> _, 0 CJ) !D <( j i I f:s {gl_!005 S_!!C -----
I Rev. 1 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 6. 2 .1-56 Steam Generator Loop Compartment Analysis Cold Leg Break Abs. Pressures Near The Break Compartments
- I I 1 I, CREEK UPDATED SAFETY ANALYSIS REPORT Figure 6.2.1-57 Steam Generator Loop Compartment Analysis Abs. Pressures Near The Break Oompartments-Oo1d Leg Break *Al>r**-* **.sol -.so\ --.9or
- TIME(SECONDS)
- 10-11 ! ' . 1.90 1 WOLF CREEK 200DO*r------------------------------------------------------------------------------
160.00 120.00 80.00 40.00 . . 00 -40.00 ---
c.uH .. ... .. OPDA!'BD SA!'InT.;.AialwYSIS RBPOft i.-t!,"f58
,. ct :.,-:, i\: i..." STEAM G£NMRA't6R*
ENT NALYSIS, 8REAK, VERTICAL
[ ""-\.f AMDL'N-5 f"'ORCES OW SG; I I Rev. 0 I 125.4 @I .095sec "'--..,.VERTICAL FORCE --E-W HORIZONTAL FORCE --...,N-S HORIZONTAL FORCE -80.00'---------"L-------"'L---------l---------L-------J..------L-------l...-------'-------..L..------J
.00 .10 .20 .30 .40 .50 .60 .70 .80 .90 1.00 .!IME (SEC) *10*1
.....
- c 0.. () a: w :I: 1-z 0 (/) w (), a:. 0' u.l CREEK 280 .oo r-----------------;-:;:::::::;;::;::;;, __ ;;::wo;;tr.;=;
2;:::;;;;;;;;;;:::====;;:;:=::;;i---, -UPDA-rBD sU'IWY. SSIS UPOn'-FIGURE 6:2. f-59 240.00 200.00 >
- Rev. 0 160.00 N-S HORIZONTAl FORCE 120.00 80.00 40.00 TIME (SECOND) *10-1 I I* I
- Fx=1.114 x 10S1bt! Fy=1.653 X 1051bf' FT=1.993 X 1051bf WOI..J!" CREEK UPDATED SAJ!"ETY ANALYSIS REPORT Figure 6.2.1-60 Steam Generator 2 Loop Compartment Analysis, 236 in. Cold Leg Break Direction of Peak Horizontal Forces on Reactor Coolant Pump and Steam Generator Rev. 1 --l I ]Fx=7.141 X 10 4 1bfl ___ "" jFy=5.026 X 10 4 1bfl IF,-=8.733 X 1o41bf' --!REACTOR PRESSURE ____ _.. VESSEL f. /.. I ? I """ I .,
<( c;; w a: :::> (/) (/) w a: n. 24.0 20.0 16.0 COMPARTMENT 2 12.0 1 WOLF CREEK ! Rev. 0 WOLF CREEK UPDATED SAFETY AHALY&IS REPORT Fis"uR-E -6. 2': 1-61 STEAM GENERATOR COMPARTMENT ANALYSIS, 436 .. 1N. PUMP SUCTION LINE BREAK *. ABSOLUTt*PRESSURES f'IEAR THE 'BREAK 0 0 .01 .02 .03 .04 .05 .06 .07 .08 .09 .1 0 TIME (SEC) L It) -0.2 0 ,... )( u. co ...1 w u a: 0 -0.4 u. -0.6 -0.8 O.G1 0.02 0.03 WOLF CREEK WOLF CRBBit.
- OPDM'BO SAFB'l'Y AMA£:!SIS RBPOR'!
1 STEAK GENERATOR
... CDMPARTMENT
>PUMP SUCTION LINE N-S COM'f>ONE#T OF HORIZONTAL FORC-E ON SG -0.9435 X 10 5 lbf @0.055SEC Rev. 0 0.04 0.05 0.06 0.07 0.08 TIME (SEC) 0.09 l v 0 ..... X u. lXI ...J" w u a: 0 u. 4.0 3.0 2.0 1.0. I 5.1 X 1o4 lbf @ O.Q18 SEC WOLF CREEK WOL!' CRBBK UPDATED SA!'BTY BBPORT FIGURE 6. 2. *1-63: STEAM GEN'ERATOR.
- COMPARTMENT
- ANALYSIS, 436 IN.
- PUMP SueT I ON OF HORIZONTAL FORCE SG Rev. 0 -1*0 0 l_0-----1 L.._0 ____ 2..l_0-----: 3:l:.o:-----.
TIME (SEC) x 1 o-2 1.0 0 -X LL aJ _. .. w (.) a: 0 LL WOLF CREEK 2.10 ,--------------------------------.
1.80 1.596 X 105 lbf @0.09SEC 1.50 120 .900 .600 .300 0 .o1 .02 .03 .04 .05 .06 .07 .08 .09 .1 0 TIME (SEC) I Rev. 0 WOLF CREEK .UPDATED SAFETY ARALYSIS.RBPORT fiGURE 6.2:1-64 STEAM G£NE.RATOR COMPARTMENT ANAlYSIS,'
436 IN *. >PuMP SucTION LIN( BREA.K,. VERT!CAL . FORCE ON SG -' l ID 0 ..... X -u.. lXI ..J w () a: 0 u.. WOLF CREEK 0.8,----------------------------------,
0.0 .01 .02 .03 .04 .05 .06 TIME (SEC) .07 .08 .09 .10 I Rev. 0 WQLP* CRBBK UPDA'l"BD SUftY. ANALYSIS RBPOH FIGURE '6. 2. 1-65 STEAM GENERATOR COMPARTMENT 436 IN. PUMP SUCTION LINE,'HR'EAK, N*-S COMPONENT DF HORIZONTAL.
FORCE ON RCP l 10 0 .... X u. w (.) a: 0 u. WOLF CREEK 2.10r-------------------------------
.300 M m M .07 .08 .09 .10 TIME (SEC) j Rev. 0 I WQLF_CREEK UPDATED SAF8t¥1AHALYSIS REPORT --, .
6.2.1-66 STEAM to!i.
f>ARTM ENT ANAlYSIS.
436S1IN PUMP SUCT.I_ON tiN'E-*£RE.AK f;;.W COM:RtJN&NT OF
- HORIZONTAL FORCE ON RCP l
It) 0 .... X LL. a:l ....1 w u a: 0 LL. WOLF CREEK 1.40 r----------------------------------, 1.20 1.00 .800 .600 .400 .200 0.0. .01 .02 0.6206 X 10 5 1bf @0.025SEC TIME (SEC) Rev. 0 WOLF CREEK UPDATED SAFE!'Y ... UALYSIS REPOR'l' STEAM GENERAJQR COMPARTMENT . ANAL YSI:L 436 IN.* PUMP SUCTION LINE BREAK, VERTICAL FORCE ON RCP .10 Fx*1.417 x toslbf Fy .. O.B31 F:r-1.643-; ,os ibf WOLF CREEK Fy .. WOL JIB.BIC . . ... OPDA'!'BD SAPBft*H.ALYSIS RBPOR'J' 2,.;1-68 . STEAM C(}MPARTMENT ANALYSIS, 436'1iN. PUMP SUCTI'ON LINE
'oF PE:AK HORIZONtAL.
FORCES o*N REACTOR MP Si'AM ENERATOR STEAM GENERATOR Rev. 0 Fx*9.435 x to4 lbf Fy=3.112 X to41bf FT:.S.935
)( 104 lbf REACTOR PRESSURE VESSEL WOLF CREEK 32.0 29.0 26.0 w a: ::> w 23.0 a: a.. 20.0 17.0 COMPARTMENT 8 COMPARTMENT 9 .07 .08 .09 TIME (SEC) Rev. 0 WOLI' CUBit . OPDA!'BD SA!'ftY UALYSIS RBPOR'r FIG;tfRE 6. 2. 1-69 , STEAM GENERATOR 200P COMPARTMENT ANALYSIS, 736 IN *. HoT-lEG BREAK, ABSOLUTE PRESSURE NEAR THE
.10 l I.C) 0 ... X u.. fXI ...J w u a: 0 u.. WOLF CREEK 1.40 .----------------------------------------
.900 .400 -.100 -.600 -1.10 -1.60 E-W 1.027 X 105 lbf @0.020SEC
-1.496 1o5 ibt N*S @ 0.040 SEC. -2.1o,....., __
__ __.. ___ ........ ___ _._ ___
___ '--__ ....... ___ __,_ ___ -'-___ .., *o .01 .02 .03 .04 .05 TIME (SEC) .06 .07 .08 .09 1 Rev. 0 I . _ WOLP _ CREBK UPDATED SAPB!'fY UALYSIS RBPOH FIGURE £.2.1-70 .10 STEAM COMPARTMENT 763 lNL_ 8otLLEG BREAK, HORIZONTAL FORCES ON SG l 1.0 0 .... u. a:l ...1 w (.) a: 0 u. I WOLF CREEK 2Bor---------------------------------------------------------------, 2.40 .05 .06 .08 TIME (SEC) WOLF CREEK 2.262 x 1 os Jbt @0.090SEC
' .09 Rev. 0 --UPDATED SAFs-rY QALYSIS REPORT* FrGURE 6.2:1-11 STEAM tJENERATOR
- ANALYSIS, 7S3 IN. ,:Hoi-LEG BREAK, VER-TICAL FoRcrs *-oN SG ."';. . "' ' .. : .10 l It) 0 _. X u: a:l ...J w u a: 0 I.L WOLF CREEK 2.40 1.60 BO .40 2.313 X 105Jbf @0.010SEC
.03 .04 .05 .06 .07 .08 .09 .10 TIME (SEC) Rev. 0 WOLE CRB:R -**-*
SAPE!T AMALYSIS STEAM GEttfRATOR 200P COMPARTMENT ANALYSIS, 763 -1N HOT-LEG BREAK, N-S COMPONENT OF HORI ZOtH AL FORCE RCP-1.0 0 ..... 0.0 -1.4 LL Ol ..J -2.8 a: 0 LL* -.42 -.56 WOLF CREEK -Q.5473 X 1051bf @0.040SEC TIME (SEC) Rev. 0 W01'4 CRBBI.t UPDA'l'BD SA!"B'l'Y ADLYSIS RBPOR'l' ** fiGURE 6.2.1-73 STEAM GENERATOR.
200P ANALYSIS, 763 IN. HOT-LEG BREAK, E-W COMPONENT Of HORIZONTAL FORCE ON RCP
- l LO 0 ... u.. ID ..J w (.) a: 0 u.. , WOLF CREEK 1.40,------------------------------
TIME (SEC} I Rev. 0 ..,._ WOLP CRQK UPDA'l'D SAPB'l'Y ARALYSIS REPOR'l' FIGURE 6.2.1-74 STEAM CbMPARTMENT ANALYSIS, 763 IN.
BREAK, VERTICAL FORCE ON RCP l
Fx"'2.313 X 1051bf Fy=0.547 X 1051bf FT=2.327 X 10 5 1bf CREEK Fy WOLF CREEK UPDAT&D-SAFETY ANALYSIS REPOR' .. -.FIGURE *6.2.1-75 STEAM Rfro*R P --*cQ_M PA RTM ENT I. ANALYSIS;, 763 HJ. HOT-LEG BREAK, DIRE,CTiONr' of. PEAK Ho)U FORCES ON REA&TOR .COOlANT PUMP AN STEAM GENE*RAT'OR Rev. 0 Fx=1.495 X 10 5 1bf Fy=0.044 X 10 5 lbf F,..=1.496 X 1051bf L REACTOR PRESSURE VESSEL DD 10 2068'-8" 2044'-6" _...._ __ 7 6 2029'-6 1/4" 2025'-0" 2018'-4" 1 2001'-4" WOLF CREEK 2090'-4" WOLP CREEK UPDATED SAPBTY ANALYSIS REPORT FIGURE 6.2.1-76 -.. PRE.SSURIZER COMPARTMENT ANALYSIS SCHEME -ELEVATION
--. . VIEW STEAM GENERATOR REACTOR COOLANT PUMP REACTOR VESSEL Rev. 0 WOLF CREEK @ 4 , r .... 0 @ ..... @ ..... .. ...... t> .. 0 @ ... -0 .. 0 . @ -G) . .. ....... FOR NODE VOLUMES AND VENT PATH AREAS, FLOW COEFFICENTS AND 1/a's, REFER TO TABLES 6.2.1-26 AND 6.2.1-27 Rev. 0 WOLF -CREEK --UPDATED REPORT FIGURE-1-77 FLOV DIAGRAM PRESSURIZER COMPARTMENT ANALYSIS w 0:: ::> w 0:: 0.. WOLF CREEK 28.0 -------------------------------------------------, 26.0 24.0 20.0 NODE1 18.0\ 16.0 Rev. 0 WOLP CREEK DPDA!"BD SAPftY RBPOR!' '1 '"" **FIGURE 6.2.1-78 PRESSURIZER COMPARTMENT ANALYSIS, PRESSURIZER SURGE ,LINE BREAK, A B S 0 L U T E PR ES S U ltE S > 8 E L 0 W T H E PR*£S S R ' .
L' 0.0 o. . TIME (SEC)
- a
- 8 I I I I I I I I I 1 I I I I I I I I I __ J WOLFCREEK REV. 22 UPDATED SAFETY ANALYSIS REPORT FIGURE A SIMPLIFIED SCHEMATIC OF THE WOLF CREEK I I I I_ I I I I I I I I I I I I I I I I I I I I I I j2-, I I I : : : I I I I I I I I I I I I I I _J I I WOLF CREEK REV .22 UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-80 WOLF CREEK GOTHIC MODEL FOR MSLB EVENTS REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 81:ContainmentPressure,VaporTemperatureandSumpWaterTemperatureResponseToAPostulatedMSLB-Case10Scenario REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 82:ContainmentPressure,VaporTemperatureandSumpWaterTemperatureResponseToAPostulatedMSLB-Case1Scenario
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 83:HeatTransferCoefficientvsTime,LimitingContainmentPressureScenario-MSLBCase10
REV.29 WOLFCREEKUPDATEDSAFETYANALYSISREPORTFIGURE6.2.1 84:HeatTransferCoefficientvsTime,LimitingContainmentTemperatureScenario-MSLBCase1 300 -250 u.. . (!) w c -w a: :::> 1-200 <( a: w a.. w 1-150 MOTOR OPERATED VALVES ELECTRONIC DIFFERENTIAL PRESSURE TRANSMITTER Rev.6 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-85 SURFACE TEMPERATURE vs. TIME FOR REPRESENTATIVE MATERIALS INSIDE CONTAINMENT FOLLOWING MSLB CASE 1 100 0 50 100 150 200 250 300 350 TIME (SECONDS)
- *
- REV. 2 9 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.1-86 Analysis vs. Calculated Containment Backpressure
00 8" 2' 4' J --ll 1----I I 1----8' I 11'-6" TYP 6' I _I I 10& A llfll Ill I rf_ REACTOR BLDG. I ' ' " I ' ' I .... ' I ' *' I ' A 7l ----!------1-----T-G 0 _J Q) a:: 0 l-Si w a:: 0 1-00 ' N L{) '------00 I FLOOR EL. 2000'-0" OP OF T LOW EL. ER PLENUM 1993'-0%" 1:11 1-. jllj ... I ' *-11 8 11 SUMP r.-lu " ........
.. TB' §1!AE5! § AG4 ...... E--'-<1 l=§AE411=
'E: AG3 e-----:1 Ill I -E:----;1 "l' E:----;1 i::i:IAE31 AF31t;:
- j;;;;i:JAH2!;;;
E---3 IIU ---il
§8AE2! AF2e ---;!
-
F8,AF1!E:
--1=---------/ !---1----6" 12" 18" EL. 1992'-0" SUPPORT STRUCTURES NOT SHOWN FOR CLARITY SECTION A <I '
' I I. I l -<II -TOP OF UPPER PLE EL. 1995'-NUM 2" SUCTION Ll NE TO PUMPS II All SUMP REV .20 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.2-3 RECIRCULATION SUMP STRAINER ARRANGEMENT WOLF CREEK 270" 180" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.2-4 CSS AREA COVERAGE AT OPERATING DECK OF CONTAINMENT l
0 8 0 ,.. X Iii w LL 8 0 <( w J: ..J <( ... 0 ... 4 2 0 BHP HEAD NPSHR 1 2 GALLONS PER MINUTE x 1 000 -3 4 WOLF CREEK a: w 8000 Q. w 0 a: 4000 J: w 200 <( a: m ... w w LL J: <<< Q. z 0 UPDATED SAFETY ANALYSIS REPORT FIGURE 8.2.2-5 CSS PUMP PERFORMANCE CURVE REV. 0 ELEVATION NO SCALE IDLF CREEK e UNK(7YR) G¥7HM *.NG"AVY H/FII'IFH4/It.E H/Nq,EIS SECTION "A-A" NO SCALE WOLP CREEK H/Nt$e aPON4E lf't/Ail6111f' Rev. 0 SAPErY ANALYSIS FIGURE 6.2.2-6 TYPICAL DETAIL OF FUSIBLE LINK PLATES ON CONTAINMENT AIR COOLER l WOLF CREEK i il REACTOR BUILDING REV.19 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.2-7 EXPECTED INTERNAL AIR FLOW PATTERN IN CONTAINMENT POST LOCA WlVE NO. LINE/ VALVE SIZE, IN. INSIDE/ OUTSIDE CONT. ASSOCIATED WITH A SAIFETY FEATURES SYS. YES[!) NO 0 FLUID CONTAINED:
STEAM LENGTH OF PIPING TO OUTERMOST ISOLATION VAILVE: 40.0ft N'PLICPBLE GDC NO. GENERAIL COMMENTS:
NONE . 1111--llf
... ITMTICIIIII
,..aa. MD 111711 Till l'riJW CllllllUmiM Alii .no lu.IIC"ftl
--..... .,.. ** .,.
- lltrr ....... on..,._," 011111111111
....... *-MDDUCT LIAUIII TD -.. ......,111111
........ .... .... ... IA'IIII wu.&-IHI WI* -O* Will 1.-IIWIArtlll
'IICIII1MI
.,.... ..............
.... M ..... M JIIIUIII ...,_1, IHI ITIAII -INmlll **L 1m111111 ro 1MI v111r ...........
1'1CI -*-ft. lOIII Pal 1111 -lftoiiiPUIW,IIIIo_,'MIIIIo THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VAIL VES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VAlVES. NORMAL FLOW DIRECTION VM.VE TYPE L r VALVE OPERATOR Dll DLI POWER SOURCE DBB DLB PRIMARY ACTUATION SIGNAL DBB SECONDARY MAXIMUM ACTUATION CLOSURE SIGNAL TIME CSEC.) + DBD NORM AIL VALVE POSITION APPENDIX J REQUIREMENT SHUTDOWN FAIL PRIMARY CONTAINMENT PENETRATION NO. P-1 DESCRIPTION:
MAIN STEAM LINE REFERENCE SECTIONCSl 10 * .3 REV.1.3 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 1 OF 74 WLVE NO. LINE/ VALVE SIZE, IN. ASSOCIATED WITH A SAFETY INSIDE/ OUTSIDE CONT. FEATURES SYS.
NO 0 FLUID CONTAINED:
STEAM LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 37.9ft APPLICABLE GDC NO. NONE
GENERAL COMMENT
S' 1111 CIIIVT,.._IIT
..
A'ND Wlnl nil ti'Uit Gl .. lllftllll .Ill ...... --.. _..,, -.........
f-IAIIIIIII 1111-IIY
- 1101 -MIIID. Till 10U110M1Y 011 MIIMU Mlllilll'
- -** MGOUa'l' Lao\IWII TO ...... _lilY II Till III.DI 011 1MI .............
_ , ... -Jill lVI' -IIP"'III a-. IIIA ... TM MIIM ntl ............
IIIILI& M .._ Dll ,.._. ,. ... ,, Till ITUII --11111 ... " lmi!IGI 10 Till VliiiT 11MVIICII 1111 COIIDI-TI
'0'111 roll ntl ............
'l ..... lftllll. THIS PENETR.&.TION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORMAL FLOW DIRECTION VALVE TYPE L r Dll OLII VALVE OPERATOR POWER SOURCE DBB DLB PRIM PRY ACTUATION SIGNAL DBD SECONDAAY MAXIMUM ACTUATION CLOSURE SIGNAL TIME CSEC.l NORMAL V451 v.on VALVE POSITION APPENDIX J ECONDARY REQUIREMENT SHUTDOWN FAIL PRIMJIRY CONTAINMENT PENETRATION NO. P-2 DESCRIPTION:
MAIN STEAM LINE REFERENCE SECTIONCSl 10.3 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 2 OF 74
. -.. -**-.. -*. ----.. ----.---.--------.-------------------
- -----------* -* -----.-----------
---*
.--.----------
.. --------.--
---*-.. -----*-* *----. -----*---------------
- ----* --*-----1 i VN...VE NO. UNE/ VN...VE SIZE, IN. INSIDE/ OUTSIDE CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES!!] NO 0 FLUID CONTAINED:
STEAM LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 37.9ft APPLICABLE GDC NO.
GENERAL COMMENT
S:
NONE THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORMN... FLOW DIRECTION VN...VE TYPE L r VN...VE OPERATOR POWER SOURCE DBB DL8 PRIMARY ACTUATION SIGNN... 088 SECONDARY ACTUATION CLOSURE SIGNN... TIME (SEC.l + DBD NORMAL VALVE POSITION APPOOIX J REQUIREMENT SHUTDOWN FAIL PRIMARY CONTAINMENT PENETRATION NO. P-3 DESCRIPTION:
MAIN STEAM LINE REFERENCE SECTION!Sl 10.3 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 3 OF 74 VN_VE NO. UNE/ VALVE ASSOCIATED WITH A SAFETY INSIDE/ OUTSIDE CONT. FEATURES SYS.
NO 0 FLUID CONTAINED:
STEAM LENGTH OF PIPING TO OUTERMOST ISOLATION VAI..VE* 40.0ft APPLICABLE GDC NO. NONE GENERAl..
COMMENTS*
THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VAL YES. NORMAL FLOW DIRECTION VALVE TYPE VALVE OPERATOR L r DLI POWER SOURCE DBB DLB PRNARY ACTUATION SIGNAL DBD SECONDARY MAXIMW ACTUATION CLOSURE SIGNAL TIME !SEC.l NORt.AAI..
VAI..VE POSITION APPENDIX J ECONDARY REQUIREMENT SHUTDOWN FAIL PRIMARY CONTAINMENT PENETRATION NO. P-4 DESCRIPTION*
MAIN STEAM LINE REFERENCE SECTION!Sl 10.3 REY.13 WOLF CREBK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 4 OF 74 VALVE LINE/ INSIDE/ NORMAL VALVE OUTSIDE FLOW NO. SIZE , IN. CONT. DIRECTION ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!] NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 15ft AIPPLICABLE GDC NO. NONE
GENERAL COMMENT
S:
THE CONTAINMENT PENETRATIONS ASSOCI* ATEO WITH THE STEAM GENERATORS ARE NOT SUBJECT TO GDC-57 , SINCE THE CON* TAINMENT BARRIER INTEGRITY IS NOT BREACHED.
THE BOUNDARY OR BARRIER AGAINST FISSION PRODUCT LEAKAGE TO THE ENVIRONMENT IS THE INSIDE OF THE STEAM GENERATOR TUBES AND THE OUT* SIDE OF THE LINES EMANATING FROM THE STEAM GENERATOR SHELLS. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. ASSETS AEVD120. AEV0121, AEV0122 , AND AEV0123 HAVE VALVE DISCS , HINGE PINS AND BUSHINGS REMOVED. VALVE TYPE POWER PRIMARY VALVE ACTUATION OPERATOR SOURCE SIGNAL V-332 ELB CXl N ,., > .... 0 ,., > SECONDARY ACTUATION SIGNAL v EBB 0 N N > ... "' > MAXIMUM CLOSURE TIME CSECJ 5: y:-p EBD VALVE POSITION APPENDIX J NORMAL SHUTDOWN AMMONIA AND / < '\. HYDRAZINE INJECTION FAIL PRIMARY REQUIREMENT MAIN FEEDWATER SYSTEM CONTAINMENT PENETRATION NO. P-5 DESCRIPTION:
MAIN FEEDWATER LINE REFERENCE SECTIONCSl 10.4.7 REV.25 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 5 OF 74 VALVE LINE/ INS()E! NO. VALVE OUTSIDE SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 13.4 FT AIPPLICASL E GDC NO. NONE
GENERAL COMMENT
S:
THE CONTAINMENT PENETRATIONS ASSOCIATED WITH THE STEAM GENERATORS ARE NOT SUBJECT TO GDC-57 , SINCE THE CONTAINMENT BARRIER INTEGRITY IS NOT BREACHED. THE BOUNDARY OR BARRIER AGAINST FISSION PRODUCT LEAKAGE TO THE ENVIRONMENT IS THE INSIDE OF THE STEAM GENERATOR TUBES AND THE SIDE OF THE LINES EMANATING FROM THE STEAM GENERATOR SHELLS. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS. NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT I SOLATION VALVES. ASSETS AEV0120, AEV0121 , AEV0122, AND AEV0123 HAVE VALVE DISCS, HINGE PINS AND BUSHINGS REMOVED. NORMAL VALVE VALVE POWER FLOW TYPE OPERATOR SOURCE DIRECTION V-329 EBB A ELB EBB PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME ISECJ .1:. v , > >, ELB L I , N "' N N > ::> .n EBB N p , > EDB VALVE POSITION NORMAL SHUTDOWN FAIL AMMONIAANO
/ ( HYOAAZINE INJECTION APPENDIX J PRIMARY SECONDARY REQUIREMENT CONTAINMENT PENETRATION NO. P-6 DESCRIPTION:
MAIN FEEDWATER LINE REFERENCE SECTION<Sl 10.4. 7 REV.25 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 6 OF 74 VALVE LINE/ INSIDE/ NO. VALVE OUTSIDE SIZE. IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!! NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 13.4 FT APPLICABLE GDC NO. NONE
GENERAL COMMENT
S: THE CONTAINMENT PENETRATIONS ASSOCI* ATED WITH THE STEAM GENERATORS ARE NOT SUBJECT TO GDC-57 , SINCE THE CONTAINMENT BARRIER INTEGRITY IS NOT BREACHED.
THE BOUNDARY OR BARRIER AGAINST FISSION PRODUCT LEAKAGE TO THE ENVIRONMENT IS THE INSIDE OF THE STEAM GENERATOR TUBES AND THE OUT* SIDE OF THE LINES EMANATING FROM THE STEAM GENERATOR SHELLS. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. ASSETS AEV0120, AEV0121, AEV0122, AND AEV0123 HAVE VALVE DISCS, HINGE PINS AND BUSHINGS REMOVED. NORMAL VALVE VALVE POWER rLOW TYPE OPERATOR SOURCE DIRECTION ELB EBB PRIMARY SECONDARY MAXIMUM VALVE POSITION APPEND I X J ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME CSECJ NORMAL SHUTDOWN FAIL PRIMARY SECONDARY REQUIREMENT v .n 0 EBB N "' N ::> .:c. p EDB AMMONIA AND < HYDAAZINE IN J ECTION MAIN FEEDWATEA SYST EM CONTAINMENT PENETRATION NO. P-7 DESCRIPTION
- MAIN FEEDWATER LINE REFERENCE SECTIONCS1 10.4. 7 REV.25 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 7 OF 74 VALVE LINE/ INSIDE/ NO. VALVE OUTSIDE SIZE, IN. CONL ASSOCIATED WITH A SAIFETY FEATURES SYS. YES [!I NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 15.2 FT APPLICABLE GDC NO. NONE
GENERAL COMMENT
S: THE CONTAINMENT PENETRATIONS ASSOCI* ATEO WITH THE STEAM GENERATORS ARE NOT SUBJECT TO GOC-57 , SINCE THE CONTAINMENT BARRIER INTEGRITY IS NOT BREACHED.
THE BOUNDARY OR BARRIER AGAINST FISSION PRODUCT LEAKAGE TO THE ENVIRONMENT IS THE INSIDE OF THE STEAM GENERATOR TUBES AND THE OUT* SIDE OF THE LINES EMANATING FROM THE STEAM GENERATOR SHELLS. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS. NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. ASSETS AEV0120, AEV0121, AEV0122, AND AEV0123 HAVE VALVE DISCS , HINGE PINS AND BUSHINGS REMOVED. NORMAL VALVE VALVE POWER FLOW DIRECTION TYPE OPERATOR SOURCE EBB ELB L c ELB_..,* ... EBB PRIMARY SECONDARY MAXIMUM VALVE POSITION APPENDIX J ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME ISECJ NORMAL SHUTDOWN FAIL PRIMARY REQUIREMENT
.5:. AMMONIA AND / < "\. :g , , "' FV-45 v HYDAAZINE INJECTION
.... "' "' > "' "' . ';' > EBB p EBD CONTAINMENT PENETRATION NO. P-8 DESCRIPTION:
MAIN FEEDWATER LINE REFERENCE SECTION<Sl 10.4.7 REV.25 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 8 OF 74 V/>LVE LINE/ INSIDE/ NO. VI>LVE OUTSIDE SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 8.5 FT APPLICABLE GDC NO. NONE
GENERAL COMMENT
S*
lltE CONTAINMENT PI!NETRATIDNI AIIOCI* ATBO WITH THIITBAM O!NERATDRI ARE NOT lua.IECT TD IIDC-117, IIMCE THE TAINIIINT IAIIIII!II INTIOIIITY II NOT IIIBACHI!O.
THI IOUNDAIIY 011 1411111111 AOAINn FilliON PIIODUCT LEAKAGE TO THE BNVIIIDNMENT II THE INIIDE OF THI ITI!AM QINIIIATOA TU8EI AND THE DUT* lilliE OP THI LINII IMANATIIVO FIIDM THI ITIAM GINIIIA'TOIIIHILLI.
THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORM/>L V/>LVE V/>LVE POWER FLOW TYPE OPERATOR SOURCE DIRECTION ST!.AM GEN. ) D z w t: 1:1 z Ill ... 5 TUBE BHEET DRAIN V-037 PRIMARY SECONDARY t.tAXIMUM VALVE POSITION N'PENDIX J ACTUATION ACTUATION CLOSURE SECONDARY REQUIREMENT SIGNAL SIGNAL TIME <SEC.l NORMAL SHUTDOWN FAIL PRIMARY NUCLEAR "'" SEE P*B3 ---SAMPLING ,/". DBB .TUBING ...
i!i ! V-034 DBB 2,p DBD L r V.(X!II CONTAINMENT PENETRATION NO. P-9 DESCRIPTION:
SLOWDOWN LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTION<Sl 10.4.8 REV. 1.3 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 9 OF 74 Vf.lVE LINE/ INSIDE/ Vf.lVE OUTSIDE NO. SIZE, IN. CON1. ASSOCIATED WITH A SAFETY FEATURES SYS. YESI!J NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 5.8 FT APPLICABLE GOG NO. NONE
GENERAL COMMENT
S:
THII CONTAINIIIIINT PINITIIATIO. Aftll Willi lliE ITEAM CJEN&RATOIII Alii NOT IUaiiCT TD OIICII7, IINCI THI CON* TAINMUIT IARIIIIII IN111DIIITV II NOT 8111ACHICI.
TMI BOUNDARY OR IIAIIIIIIII AQAinT ,.a!ON PRODUCT L.EIIIt.AI!E TD THE ENYIIIDIIMiiNT II THE INIIDE OF TH& 811!AM OENIIIATOII TU-AND THE OUT* IIDI OF THE LIN&I EMANATIND PIIOM TNI ITEAM OENERATOIIIHI LU. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VAJLVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORJ.!f.l V.ALVE POWER FLOW Vf.lVE DIRECTION TYPE OPERATOR SOURCE STEAMGEN:
} '" A z Ill I1J 0 IIC t: Iii z :*,-,r-SLOWDOWN LINE TUllE SHEET DRAIN V.C04 PRIMARY SECONDARY MAXIMUM VALVE POSITION APPENDIX J ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME <SEC.l NORMAL SHUTDOWN FAIL PRIMARY REQUIREMENT NlllllEAR "': SEE P*B4 ---
I" DSS nJBING ... l O'ca V-001 DSS """ 2,P DBD L r V-002 CONTAINMENT PENETRATION NO. P-10 DESCRIPTION:
SLOWDOWN LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTION<Sl 10.4.8 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 10 OF 74 V/>L VE LINE/ INSIDE/ VALVE OUTSDE NO. SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YESl!J NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 5.3 FT APPLICABLE GDC NO. NONE
GENERAL COMMENT
S' THE CONTAINMINT PINITIIATIDNI ATiiD WITH THI m.u. QINIRATDIQI AilE NOT IUJ,IECT '10 C!DC-117, IlNCE 1111 CON* TAIJIMINT IAIIIIIIR INTEGRITY IS NQT IIII!AOHIII.
THE BOUNDA!IY 011 IAIIIIIIII AQAINSI' PraiON PIIDIIUCT LIAICAQI TO 1111 INYIIIONMINT II THE INIIDI OF THE SI'IAM QIINERA'ION TUIE. AND THI OUT* IIDI DF THE LINEI UIIANATING PROM THI ITIAM QINERATDR IHILL& THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORMAL VM.VE FLOW VM.VE TYPE OPERATOR DIRECTION ) '" / B BLOWDOWN LINE POWER PRIMARY SECONDARY I.W(II.IJM VALVE POSITION ACTUATION ClOSURE APPENDIX J ACTUATION SOURCE REQUIREMENT SIGNAL SIGNAL TIME !SEC.> NORM,bL SHUTDOWN FAIL PRIMARY V.OIS NUCLEAR oC:I '" SEE P*85 ... z .. --""f DBB* TUBING !l !C' ... z ! V.012 V*OIB DBB "" 2Df--& DBD L r V*013 8 S.G. BLOWDOWN FLASH TANK P\JMPS CONTAINMENT PENETRATION NO. P-11 DESCRIPTION:
BLOWDOWN LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTION($)
10.4.6 REV. 13 WOLF CREEX UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 11 OF 74 VM.VE UNEI INSIDE/ VM.VE OUTSIDE NO. SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES I!) NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VI>LVE* 6.8 FT APPLICABLE GDC NO. NONE
GENERAL COMMENT
S*
THE CONTAINMINT P!NITIIATIOIII AIIOCI* AT!D WITH THI ITIAM OENERATOIII Alii NOT IUINECT TO IlNCE THI TAINMINT IIARIIIIR ITIITICIIIITY It NOT .REACHIII.
THI IOUNDAR'I 011 IIARIIIIR ACIAIIIIT P-ION PRODUCT LIAICACII TO THE ENYIIIONMINT II THI INIIDI OF THE STIAM CIINIRATOR TUBES AND Tit! OUT* 1101 OF THE LINEIIIIIIANATING PROM THI IT! AM OINIIIATOA
.. ELUL THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORMM. VM.VE POWER FLOW VM.VE DIRECTION TYPE OPERATOR SOURCE STEAM GEN. ) / c z B !: z 'Jltl TUBE SHEET DRAIN V-'lll PRIMARY SECONDARY MAXIMUM VALVE POSITION APPENDIX J ACTUATION CLOSURE ACTUATION SIGNAL SIGN& TIME CSEC.I NORMAL SHUTDOWN FAIL PRIMARY REQUIREMENT NUCLEAR " SEE P-88 ... --SAMPLING A, DBB TUBING ! "
a n V-023 D88 2,¥ DBD L Y*191 Y.(J27 PVMPS CONT AINt.IENT PENETRATION NO. P-12 DESCRIPTION:
SLOWDOWN LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTIONCS) 10.4.8 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CDNT AINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 12 OF 74 -----*---------------------------------------------------------------------------------------------------------------
Vf!lVE LINE/ INSIDE/ NO. Vll.VE OUTSIDE SIZE, IN. CONT. ENHV-07 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION V/>l...VE:
N/A APPLICABLE GDC NO. 56 GENERAL COMt.AENTS:
Tl!ll P!NETRATION II AISDCIATIID WITH THE CONTAINMENT .IIAV IIVITe.., WHICH II REOUIRiiD TQ lllmOATB THI COral* QUINCEI QP A LCICA, A IINCILE IIEIIIOTE*
MANUAL IIIILATION II I'RCIVIOID, LOCAT* liD WITHIN A WAT!IITIGHT COMPARTMINT DUTBIDI THI CONTAINIWIINT, I'CIII GREAT* lA IIVITEM lleLIAIILITV, A IINIILI ACTIVI OR PAIIIYE FAILUAI CAN IE OATID IINCI THE llftTWM II CLOIIiO OUT* liD& THI CONTAINMINT AND II DEIICIN&D "liD C0115TRUCT!D COMMINEUIIATE WITH THE DEIION AND CONITIIUCTIGN OP THB CONT,.,INMENT.
LOCAL t!IITINII OP THE VALVI 011 'nf! CLOII!O IIVITIM OUTBID& THI OONTAIN* MINT II NOT R!QUIIIIIIIIINCI THIIYITIM II OI'IIIATI!D AND lll.liCTID DUIIINII NORMAL PLANT OPIIIATION TQ AIIUJII THAT THI INTEGRITY II lElNO MAINo TAINED, NORMf!l V/>l..VE V/>l..VE POWER FLOW DIRECTION TYPE OPERATOR SOURCE OUT GATE MOTOR 4 I CONT. RECIRC, SUMP PRIMARY SECONOARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGN/>l..
SIGNf!l TIME tSEC.l NORMAL CIS-A REt.1/MAN 30 CLOSED HV 7 L ... *** I . r V*008 HCB '-P TCilD HCD VALVE POSITION IIPPENDIX J SHUTDOWN FAIL PRIMARY REQUIREMENT CLOSED AS IS CONT. SPRAY ) CLOSED OPEN A CONTAINMENT PENETRATION NO. P-13 DESCRIPTION:
RECIRCULATION LINE CONTAINt.AENT SPRAY SYSTEM REFERENCE SECTIONtSl 6.2.2 REV. 13 WOLF CI\EEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 13 OF 74 VALVE LINE/ INSIDE/ NO. VALVE OUTSIDE SIZE, IN. CONT. EJHV-8811B 14/14 OUTSIDE EJV-224 I OUTSIDE EJV-225 I OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!] NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
THIS PENETRATION IS ASSOCIATED WITH THE RHR SYSTEM. RHR IS REQUIRED TO MITIGATE THE CONSEQUENCES OF A LOCA. A SINGLE REMOTE-MANUAL TION IS PROVIDED, LOCATED WITHIN A WATERTIGHT COMPARTMENT OUTSIDE THE CONTAINMENT, FOR GREATER SYSTEM RELIABILITY.
A SINGLE ACTIVE OR PASSIVE FAILURE CAN BE ACCOMMODATED SINCE THE SYSTEM IS CLOSED OUTSIDE THE TAINMENT AND IS DESIGNED AND STRUCTED COMMENSURATE WITH THE DESIGN AND CONSTRUCTION OF THE TAINMENT.
LOCAL TESTING OF THE RHR VALVE OR THE CLOSED SYSTEM OUTSIDE THE TAINMENT IS NOT REQUIRED SINCE THE SYSTEM IS OPERATED AND INSPECTED DURING NORMAL PLANT OPERATION TO ASSURE THAT THE INTEGRITY IS BEING MAINTAINED.
NORMAL VALVE VALVE POWER FLOW TYPE OPERATOR SOURCE DIRECTION OUT GATE MOTOR 4 N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A CONT. RECIRC. SUMP PRIMARY ACTUATION SIGNAL REM/MAN N/A N/A ECD Vl SECONDARY MAXIMUM VALVE POSITION APPENDIX J ACTUATION CLOSURE :sECONDARY REQUIREMENT SIGNAL TIME <SEC. NORMAL SHUTDOWN FAIL PRIMARY SIS8cRWST -LO N/A CLOSED CLOSED N/A N/A CLOSED CLOSED N/A N/A CLOSED CLOSED 1.1 V224 V225 RHR SHUTDOWN '\. HCB ECB p TC&D HCD AS IS CLOSED OPEN A N/A N/A CLOSED N/A N/A CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-14 DESCRIPTION:
RECIRCULATION LINE RESIDUAL HEAT REMOVAL SYSTEM REFERENCE SECTIONISl 5.4.7 8c 6.3 WOLF CREEK REV.28 UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.4-1 CONTAINMENT PENETRATIONS PAGE 14 OF 74 VALVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. EJHV-8811A 14/14 OUTSIDE EJV-220 I OUTSIDE EJV-222 I ¥4 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!] NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
THIS PENETRATION IS ASSOCIATED WITH THE RHR SYSTEM. RHR IS REQUIRED TO MITIGATE THE CONSEQUENCES OF A LOCA. A SINGLE REMOTE-MANUAL TION IS PROVIDED, LOCATED WITHIN A WATERTIGHT COMPARTMENT OUTSIDE THE CONTAINMENT, FOR GREATER SYSTEM RELIABILITY.
A SINGLE ACTIVE OR PASSIVE FAILURE CAN BE ACCOMMODATED SINCE THE SYSTEM IS CLOSED OUTSIDE THE TAINMENT AND IS DESIGNED AND STRUCTED COMMENSURATE WITH THE DESIGN AND CONSTRUCTION OF THE TAINMENT.
LOCAL TESTING OF THE RHR VALVE OR THE CLOSED SYSTEM OUTSIDE THE TAINMENT IS NOT REQUIRED THE SYSTEM IS OPERATED AND INSPECTED DURING NORMAL PLANT OPERATION TO ASSURE THAT THE INTEGRITY IS BEING MAINTAINED.
NORMAL VALVE VALVE POWER FLOW DIRECTION TYPE OPERATOR SOURCE OUT GATE MOTOR 1 NIA GLOBE MANUAL NIA NIA GLOBE MANUAL NIA CONT. RECIRC. SUMP PRIMARY SECONDARY ACTUATION ACTUATION SIGNAL SIGNAL REM/MAN SIS AND RWST-LO NIA NIA N/A NIA ECD '*-ECB L MAXIMUM CLOSURE TII.IE ISEC.l N/A NIA NIA 1.1 V222 V220 ECB VALVE POSITION APPENDIX J NORMAL SHUTDOWN FAIL CLOSED CLOSED AS IS CLOSED CLOSED NIA CLOSED CLOSED NIA RHR SHUTDOWN SEE P-79 ) >-HCB g ____L____M
>-,-PRIMARY SECONDAR't REQUIREI.IENT CLOSED OPEN A CLOSED NIA NIA CLOSED N/A NIA SUCTION CONTAINMENT PENETRATION NO. P-15 DESCRIPTION:
RECIRCULATION LINE RESIDUAL HEAT REMOVAL SYSTEM REFERENCE SECTION!Sl 5.4. 7 8c 6.3 r-----------------------------------------, TC&D HCD WOLF CREEK. REV.28 UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.4-1 CONTAINMENT PENETRATIONS PAGE 15 OF 74 Vft.VE LINU INSIDE/ VPJ..VE OUTSIDE NO. SIZE, IN. CONT. ENHV-01 2/12 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
nlll PINITI'IATION II AIIOCI .. TI!D WITH THI CONTAIIIM!NT
- rtAV IVITEM. WHICH II fiiQUIIIIO TO MITIGATE THI CONII* QUIINCIS OP A !.DCA. A IIIIGIL! RIMOTI* MANUAL lllllLATIOfl 18 PRIWIDID, LOCAT* 10 WITHIN A II"TERTIGHT COMPAR"IfftNT OUTIIOI THI CONTAINMENT, FOR GREAT* lfiiYITIIII "ILIABJLIT'I', A liNGLE ACTIYii OR PAIIIVB FAILURE Clllll BE ACCOMMO* D"T!D IINIII THI IIV.,.EM II CL-11 OUT* IIDE THE CDNTAINMI!NT MID 18 DIIIGINEII AND CONITRUCTED CDIIIIIII!tiiURATE WITH THI DEIIIIN AND CONSTI'IUCTION OP THE CDNTii*INMINT.
LOCAL T!ITINGI OF THE YALV& OR "I'HI OLOIED IVITEM OUT"IIDI THII CONTAIN* IIIENT II NOT IIIOUIRiiD IINCI Till IYIITEM If DPIIIATID AND 111-CTED DURING NORMAL PLANT DPIIIATION TD QIUIII THAT THE INTEGRITY II BEING MAIN* TAINEII, NORMAL Vft.VE VALVE POWER FLOW DIRECTION TYPE OPERATOR SOURCE OUT GATE lAO TOR 1 I CONT. IIICIIIC.IUMI' PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNft. TIME CSEC.l NORMN.. CIS-A REM/MAN 30 CLOSED ... ,v L .... .... I ., \1.(102 r HCB TCIIDII HCD V N.. VE POSITION N'PENDIX J SHUTDOWN CLOSED FAIL PRIMARY SECONDARY REQUIREMENT I>S IS CLOSED OPEN A CONT. SPRAY ' f PUMP A CONTAINMENT PENETRATION NO. P-16 DESCRIPTION=
RECIRCULATION LINE CONTAINMENT SPRAY SYSYTEM REFERENCE SECTION($)
6.2.2 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 16 OF 74 VALVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. EJHCV-8825
¥41¥4 INSIDE EJHV-8840 10/10 OUTSIDE EJV-056 111 OUTSIDE EJV-124 111 INSIDE EJV-122 ¥4 INSIDE EJV-118,120 111 INSIDE EJV -175,6,7 ,8 ¥41¥4 INSIDE EJV-218 ¥.!1¥4 OUTSIDE EJV-219 ¥41¥.! OUTSIDE EJ-8841A 616 INSIDE EJ-8841B 6/6 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!] NOD FLUID CONTAINED:
WATER LENGTH Of PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THIS PENETRATION IS ASSOCIATED WITH THE RESIDUAL HEAT REMOVAL SYSTEM. WHICH IS REQUIRED TO MITIGATE THE CONSEQUENCES OF A LOCA. A CHECK VALVE IS PROVIDED FOR EACH BRANCH LINE INSIDE THE CONTAINMENT, AND A REMOTE*MANUAL ISOLATION VALVE IS PROVIDED OUTSIDE THE CONTAINMENT.
A SINGLE ACTIVE OR PASSIVE FAILURE CAN BE ACCOMMODATED SINCE THE SYS TEM IS CLOSED OUTSIDE THE MENT AND IS DESIGNED AND ED COMMENSURATE WITH THE DESIGN AND CONSTRUCTION OF THE MENT. LEAKAGE DETECTION FROM THIS LINE OUTSIDE THE CONTAINMENT IS VIDED, AS DESCRIBED IN SECTION 9.3.3. LOCAL TESTING OF THE VALVES OR THE CLOSED SYSTEM OUTSIDE THE MENT IS NOT REQUIRED SINCE THE SYSTEM IS OPERATED AND INSPECTED DURING MAL PLANT OPERATION TO ASSURE THAT THE INTEGRITY IS BEING MAINTAINED.
NOTE* ALL VENTS,DRAINS AND FLOW POINTS AS INDICATED BELOW. BCD NORMAL VALVE VALVE POWER FLOW DIRECTION TYPE OPERATOR SOURCE IN GLOBE AIR 1 IN GATE MOTOR 4 N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A IN CHECK N/A N/A IN CHECK N/A N/A 8840 TC&D PUMP'S PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME ISEC.l NORMAL CIS-A NONE 10 CLOSED NONE REM/MAN N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED BCD v -'V S!! v BCB D VALVE POSITION APPENDIX J SHUTDOWN fAIL CLOSED CLOSED CLOSED AS IS CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A CLOSED N/A HCV-8821 .:i: ........ .........
--IICII BCD PRIMARY REQUIREMENT CLOSED N/A A CLOSED OPEN A CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A CLOSED N/A N/A CLOSED N/A N/A CLOSED OPEN A CLOSED OPEN A '\. / SIS TEST LINES SIS HOT LECi CONTAINMENT PENETRATION NO. P-21 DESCRIPTION:
HOT LEG INJECTION RESIDUAL HEAT REMOVAL SYSTEM REFERENCE SECTIONISl 5.4. 7/6.3 WOLF CREEK REV.28 UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 17 OF 74 VPJ..VE LINE/ INSIDE/ NO. VPJ..VE OUTSIDE SIZE, IN. CONT. BBHV-8351B 212 OUTSIDE BBV-354 1/1 OUTSIDE BBV-246 %!1'. OUTSIDE BBV148 212 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 11.7fL AIPPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THIS PENETRATION PIPING HAS A HIGH PRESSURE WATER INFLOW WHICH ClUD'ES THE NEED FOR ISOLATION OF THIS PENETRATION.
THE CVCS CHARGING PUMPS-SUPPL V REACTOR COOLANT PUMP SEAL INJECTION WATER, AND THERE IS A POTENTIAL FOR DAMAGE TO THE REACTOR COOLANT PUMP IF UNDESIRED ISOLATION SHOULD OCCUR. THE tSOLATION CAN BE AFFECTED BY REMOTE-MANUAL CLOSURE OF THE OPERATED VALVE BY THE OPERATOR AFTER THE CHARGING PUMPS COMPLETE THEIR SAFETY FUNCTION.
NORMPJ.. VALVE VALVE POWER FLOW TYPE SOURCE DIRECTION OPERATOR IN GLOBE MOTOR 4 N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A IN CHECK N/A N/A CHAIIGING PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNPJ.. SIGNPJ.. TIME ISECJ NORMAL NONE REM/MAN N/A OPEN N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A OPEN TC&V HV , BCD BCD VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED CLOSED OPEN FAIL AS IS N/A N/A N/A PRIMARY SECONDARY REQUIREMENT OPEN CLOSED c CLOSED N/A N/A CLOSED N/A N/A OPEN CLOSED c REACTOR COOLANT PUMP B CONTAINMENT PENETRATION NO. P-22 DESCRIPTION:
RCP SEAL WATER SUPPLY REACTOR COOLANT SYSTEM REFERENCE SECTIONISJ 5.0 REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 18 OF 74 VN..VE LINE/ INSIDE! NORMAL VN..VE VALVE POWER NO. VALVE OUTSIDE FLOW TYPE OPERATOR SOURCE SIZE, IN. CONT. DIRECTION BGHV-8160 3/3 INSIDE OUT GLOBE AIR 1 BGV-.36.3 111 OUTSIDE N/A GLOBE MANUAL N/A BGHV-8152 3/3 OUTSIDE OUT GLOBE AIR 4 ASSOCIATED WITH A SAIFETY FEATURES SYS. YESD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 11.6ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S*
REGEN. NONE HEATEXCH.
PRIMARY SECOND.ARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNN.. TIME CSEC.l NORMAL CIS-A NONE 10 OPEN N/A NONE N/A CLOSED CIS-A NONE 10 OPEN ECD VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN FAIL PRIMARY REQUREMENT CLOSED CLOSED N/A c N/A CLOSED N/A N/A CLOSED CLOSED N/A c LETDOWN CONTAINMENT PENETRATION NO. P-
2.3 DESCRIPTION
NORMAL LETDOWN CHEMICAL S. VOLUME CONTROL SYSTEM REFERENCE SECTIONCSl 9.3.4 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONT AJNMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 19 OF 74 _____________________________
_j VALVE LINE/ INSIDE/ NORMAL VALVE POWER VALVE OUTSIDE FLOW VALVE NO, TYPE SOURCE SIZE, IN. CONT. DIRECTION OPERATOR BGHV-8112 212 INSIDE OUT GLOBE MOTOR 1 BGV-135 %1¥4 INSIDE IN CHECK N/A N/A BGHV-8100 212 OUTSIDE OUT GLOBE MOTOR 4 BGV-457 1/1 OUTSIDE N/A GLOBE MANUAL N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NO [9 FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 12.6ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
NONE TC&D PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME CSEC.) CIS-A NONE 10 N/A N/A N/A CIS-A NONE 10 N/A N/A N/A TC&V HCD (if (/) L HCD HCB t HCD NORMAL OPEN CLOSED OPEN CLOSED VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN CLOSED FAIL AS IS N/A AS IS N/A PRIMARY SECONDARY REQUIREMENT CLOSED N/A c CLOSED N/A c CLOSED N/A c CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-24 DESCRIPTION:
PCP-SEAL WATER RETURN CHEMICAL & VOLUME CONTROL SYSTEM REFERENCE SECTIONCS) 9.3.4 REV. 19 WOLF CREEK UPDATED SAFETY ANAL Y§l§ REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 20 OF 74 L ____________________________________________________________________________________________
_
VN..VE LINE/ INSIDE/ NORM Pl.. VN..VE POWER V/>J..VE OUTSIDE FLOW VN..VE NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SOURCE BLHV-8047 OUTSIDE IN DIAPHRAGM AIR 4 BLV-054 111 OUTSIDE N/A GLOBE MANUAL N/A BL-8046 3/3 INSIDE IN CHECK N/A N/A ASSOCIATED WITH A SAIFETY FEATURES SYS. YESO FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 12.2ft APPLICABLE HCO GDC NO. 56
GENERAL COMMENT
S:
NONE PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL Tlt.IE !SEC.l NORMAL CIS-A NONE 10 OPEN N/A N/A N/A CLOSED N/A N/A N/A OPEN I Mal s HCB HCD TCIID VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN FAIL CLOSED N/A N/A PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c ,T., .. a CONTAINMENT PENETRATION NO. P-25 DESCRIPTION:
REACTOR MAKEUP WATER REACTOR MAKEUP WATER SYSTEM REFERENCE SECTION!Sl 9.2.7 REV.11 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 21 OF 74 VAJ..VE LINE/ INSIJE/ NORMAL VALVE VALVE POWER VALVE OUTSIDE FLOW NO. SIZE,IN. CONT. DIRECTION TYPE OPERATOR SOURCE HBHV-7176 3/3 INSIDE OUT DIAPHRAGM AIR 1 HBHV-7136 3/3 OUTSIDE OUT DIAPHRAGM AIR 4 HBV-419 1/1 OUTSIDE N/A GLOBE MANUAL N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!) FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 11.911 APPLICABLE GDC NO. 56 REACTOR COOLANT
GENERAL COMMENT
S:
NONE ---------------------*---------*---
PRINARY SECONDARY MAXt.1UM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TINE (SEC.) CIS-A NONE 10 CIS-A NONE 10 N/A N/A N/A L VALVE POSITION APPENDIX J NORMAL SHUTDOWN FAIL PRIMARY REQUIREMENT OPEN OPEN CLOSED CLOSED N/A c OPEN OPEN CLOSED CLOSED N/A c CLOSED CLOSED N/A CLOSED N/A N/A RECYCLE AREA OF CHANGE ' HOLDUP TANK CONTAINMENT PENETRATION NO. P-26 DESCRIPTION*
REACTOR COOLANT DRAIN TANK DISCHARGE LIQUID RADWASTE SYSTEM REFERENCE SECTION
11.2 REV. 15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 22 OF 74 ,_, _____________
j 2 VALVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. EJHV-8809B 10/10 OUTSIDE EJV-058 111 OUTSIDE EJHCV-8890B Y4tY4 INSIDE EJV-086 1/1 INSIDE EJV-0 8 8,090 Y 41¥4 INSIDE EP-8818C 6/6 INSIDE E JY-7!L£l'L*180 Y4tY4 INSIDE EP-8818D 6/6 INSIDE EJV-166 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH O F PIPING TO OUTERMOST ISOLATION V N. VE: N/A APPLICABLE GDC NO. 55 GENERN. CO M MENTS: THII l't!N ETI'ATION II AIIIIOCI AT E D WI TH TH E l'liii DUAL HEAT R I MOVA L SYSTE M. -ICH 15 II Ell UifiED TO MITICIAT&
THE CONUQU I NC Q O P A L OCA. A CHEC K V ALVE 18 PIIOY IDE D 1'011 IACM IIIIAIICif LloiE I ... IIE TH E CONTA-EN T , AND A R IMOR.uiNUAL ISOLATION V ALVE IS PROVIDED DUTIIIIE TH E CONT.--NT. A -L! ACTIV E 0 1' PASai V E I'AILUI'E CAN K IIINC I! T HE mTI'!II OUTSID E T H E IIENT AN D 18 DOKIII ED AND CONaTIIUCT*
&D WITH THE Dti&KIN A ND a.sT II UCTION O F THE CDNTIIUfT. LOCAL T UT INO Of THE VALV I OR TH E CLOI!D S YSTI!M DUTIID! T1lll! COfiiTAIN*
MINT II NOT II E OUII'ED .NCii T .. sYSTEM 18 OI'IIIA TE D AND IN SPECTED DUI'ING NDIINAL PLANT llrERATIOIII TO AAURE THA T TH E IN T!GI UTY IS BE ING NA INT A W. 1 0. NOTE: N.L VENTS, DRAINS AND FLOW POINTS AS INDICATED BELOW. BCD NORWAL VALVE POWER FLOW VALVE TYPE SOURCE DIRECTION OPERATOR IN GATE MOTOR 4 N/A GLOBE MANUN. N/A IN GLOBE AIR 4 N/A GLOBE MANUN. N/A N/A GLOBE MANUN. N/A IN CHECK N/A N/A N/A GLOBE MAN UN. N/A IN CHECK N/A N/A N/A GLOBE MAN UN. N/A c u I --t"'*l*,.... PRit.CARY SECONDARY WAXIWUW ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIWE <SEC.l NONE REM/MAN N/A N/A N/A N/A C I S-A NONE 13 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A [0> ...
z ::::; "' "' w ID u V-086 V N. VE POSITION APPENDIX J NORMN. SHUTDOWN OPEN OPEN CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED O PEN CLOSED CLOSED CLOSED OPEN CLOSED CLOSED v .,8C v FAIL AS IS N/A CLOSED N/A N/A N/A N/A N/A N/A P RIMARY REQUI R E WENT OPEN CLOSED A CLOSED N/A N/A CLOSED N/A A C LOSED N/A N/A CLOSED N/A N/A O PEN CLOSED A CLOSED N/A N/A OPEN CLOSED A CLOSED N/A N/A M:IOOLD LEG ) 1..001' 3 CONTAINMENT PENETRATION NO. P-27 DESCRIPTION
- C O LO LEG INJECTI O N RESIOUN. HEAT REMOV N. SYSTEM REFERENCE SECTION<Sl 5.4.7 S. 6.3 REV.27 W OL F CREEK U PDA TE D S A F E TY ANA L YSI S R EPOR T CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 23 OF 74 WlVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. EFHV-32 4/14 OUTSIDE EFHV-34 4/14 INSIDE EFV 278 /1 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 17.2ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S*
TH. PENETRATION II AI8DCIATID WITH THI ElliNTIAL SERVICE WATER IVITEM,. WHICH II REQUIRiiD TO MITIGATE THE CDNIEQU!NCD DP A LOCA. A REMDTii* MANUAL POMR-OPERATID VALVI II LOCATED INIIDI, AND A REMOTE-MANUAL POWIIII*DPERATED VAl.VI lll.DCATED OUT* 1101 THI CONTAINMENT.
THEIE VAl.VI!I ARI! POWERED PROM THI! lAME POWER IDURCE FOR CREATER &Va. TEM REl.IABil.ITV.
A liNGLE ACTIVE DR PAIIIVE FAILURE CAN II .ACCOMMODATED IINCI THI IYITI!M II A CLDIED IYSTIM INIIDE THE CONTAINMENT, WHICH II Dl!* IIONED AND CONITRUCTI!D IN ANCE WITH AIMI IICTIDN Ill, CLAII 3 REQUIREMENTS.
THii I.ENTIAL SERVICE WATER LINEI ARE NOT VENTED DR DRAIN* I!D DUIII:INO A TYI'I! A TilT IlNCE THE AIR CDDLERI MAV II! REQUIRED TO CODL THI CONTAINMENT.
A TVPE C TESr II &0. NORM.bl VALVE POWER FLOW VALVE DIRECTION TYPE OPERATOR SOURCE IN BUTTERFLY MOTOR 4 IN BUTTERFLY MOTOR 4 N/A GATE MANUAL N/A SERVICE WATER PRIM MY SECONDAAY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME CSEC.l NORMAL SIS REM/MAN N/A OPEN SIS REM/MAN N/A OPEN N/A N/A N/A CLOSED HV .. HBC HCB HBB 0 VALVE POSITION APPENDIX J SHUTDOWN OPEN OPEN CLOSED HV ,. FAIL AS IS AS IS N/A PRIMARY REQUIREMENT OPEN CLOSED c OPEN CLOSED c CLOSED N/A N/A CONTAINMENT
} ) AI" COOLER CONTAINMENT PENETRATION NO. P-28 DESCRIPTION' ESW TO CONTAINMENT AIR COOLER ESSENTIAL SERVICE WATER SYSTEM REFERENCE SECTIONCSl 6.2.2 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 24 OF 74 VN...VE LINE/ INSIDE/ NO. VM.VE OUTSIDE SIZE, IN. CONT. EFHV-46 INSIDE EFV-279 INSIDE EFHV-50 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!g NOD FLUID CONTAINED:
WATER LENGTH Of PIPING TO OUTERMOST ISOLATION V/>J...VE*
N/A APPLICI!BLE GDC NO. 56 GENERAIL COMMENTS:
THII PINITIIIATICIN II .-.aciATID WITH TMI _.-riAL *IIVIII WATIII ft'ln!M. WHICH II IIICIUIIIID TO MITIIA'lll TMI -UINCU 0' A LCII:otl.
A -* MANUAL I'OWI-TED VA&.VI II LOCIATID 1-1 AND A IIIIIOr-UAL I'OWIII.fiPIIIADD YALVI II LCICATID DUr* 1101 TMI CONT-IIT.
'Ill-VALVU Alii I'IIWIMD *-TMI --II..,_ POll OIIIATIIIIfto Till IIILWIILI7Y, A -LI AC'I'IVI 011 '-lVI PAILUIII CAIIDII __,.TID .NCI TMI -* A -D l'ni'IM ..... TMI CDNTA-1!111, WHat II 01* .aiii!D MD --ID IN -Do ANC1 M1H -* .rcTICIN Ul, OLAII I III!QUIIII-1111'1.
TMI _,.TIAL tiiiYICI!
WATIII Lilla Alii IIGT ¥1-011 DIIAIIID --A TYPI A 1UI' IINCI Till Alii CDOLIIIIIIIAY
- lmiUDIID TOCDCIIL THI CDNTA.._, A TWI 0 Tin' IIPIIII'a-D.
NORMAL VN...VE FLOW TYPE DIRECTION OUT BUTTERFLY N/A GATE OUT BUTTERFLY FROM CONTAINMENT VN...VE POWER OPERATOR SOURCE MOTOR 4 MANU/>J...
N/A MOTOR 4 HV D----1 HBC HCB HBB AIR COOLER I I ............
> PRIIII'fiY SECONDI'fiY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNN... SIGNAL TIME <SEC.I NORM/>J...
SIS REM/MAN N/A OPEN N/A N/A N/A CLOSED SIS REM/MAN N/A OPEN HV 50 L r D V />J... VE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN fAIL AS IS N/A AS IS PRIMARY REQUIREMENT OPEN CLOSED c CLOSED N/A N/A OPEN CLOSED c CONTAINMENT PENETRATION NO. P-29 DESCRIPTION:
ESW FROM CONTAINMENT COOLERS ESSENTI/>J...
SERVICE WATER SYSTEM REFERENCE SECTION<Sl 6.2.2 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 25 OF 74 VM.VE LINE/ INSDE/ NORL4M. VI>J...VE NO. VALVE OUTSIDE FLOW TYPE SIZE, IN. CONT. DIRECTION KArV-29 IVz/2 OUTSIDE IN GLOBE KAV-218 *Vtt!Vt OUTSIDE N/A GLOBE KAV-204 1Vzt1V, INSIDE IN CHECK ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NOI!J FLUID CONTAINED' AIR LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE' 7.911 APPLICABLE GDC NO. 56
GENERAL COMMENT
S' INSTIIUMDIT NONE AIR VM.VE POWER OPERATOR SCM<<:E AIR 1 MAINUAL N/A N/A N/A HCD PRIMARY SECONDARY MAXIMUM VALVE POSITION 1\PPENDIX J ACTUATION ACTUATION CLOSURE SECONDAR' REQUIID.IENT SIGNAL SIGNAL TIME ISEC.l NORMAL SHUTDOWN FAIL PRIMARY CIS-A REM/MAIN 5 OPEN OPEN CLOSED CLOSED OPEN c N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A (c1 N/A N/A N/A OPEN OPEN N/A CLOSED OPEN c AREA OF CHAINGE---
HCD A ' HCD _j V-204 ., HCB --, BUILDING HCB HCD TC t CONTAINMENT PENETRATION NO. P-JO DESCRIPTION' INST. AIR AIND H2 CONTROL MAKEUP AIR COMPRESSED AIR SYSTEM REFERENCE SECTIONCSl 9.3.1 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 26 OF 74 I ---------------------------*
-***------------------*-----*-*___]
' 'oo-oO-Oo-Oo-0 0-0 0----oo--o-o 0--0-0--*o-o --0 0----0 0----00--0-0--
- o-0--00----o 0----0 0--0----oo-o--0 0----0 0----oo--0-0--------0 0-0--0---------------------------0-------------------------------0-----------oo-------------------------, I i i I I : l I i 1 l I I I I WLVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. LFFV-95 6/6 INSIDE LFFV-96 6/6 OUTSIDE LFV-093 /1 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!) FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* 18.3ft APPLICABLE CDC NO. 56
GENERAL COMMENT
S:
VAL VI! FV-IIIOPIIQONLV WIIIN ONI OF THE CONTAINMINT lUMP PU-ARI OI'EIIATINCI, nil! CONTROL GRADI IIGNAL TO DPIIII THI VALYIIIIII£GATID I'IHIN A I'RD'111CTIDII CIRADE CII--A IIGNAL II fltmCIIVID NORMAL VALVE VALVE POWER FLOW DIRECTION TYPE OPERATOR SOlAlCE OUT GATE MOTOR 1 OUT GLOBE AIR 4 N/A GLOBE MANUAL N/A PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME !SEC.) CIS-A NONE 30 CIS-A NONE 4 N/A N/A N/A FV Ill V AI... VE POSITION APPENDIX J NORMAL SHUTDOWN OPEN OPEN SEE NOTES SEE NOTES CLOSED CLOSED I'Y-88 FAIL />S IS CLOSED N/A PRIMARY SECONDAR't REQUIREMENT CLOSED N/A c CLOSED N/A c CLOSED N/A N/A FLDOII CONTAINMENT PENETRATION NO. P-32 DESCRIPTION*
CONTAINMENT SUMP PUMP DISCHARGE FLOOR AND EQUIPMENT DRAINAGE SYSTEM REFERENCE SECTION(Sl 9.3.3 REV.13 WOLF Cl\EEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 27 OF 74 i 0 L _______________________
o _______________________________________
o _____________________________________________________________
J i ! I i I i I I I : I I i ' I I I ! I ---------------------------------------------------------------------------------------------------------
'
---.--------.-------------
- ---.----.. ----.-----
.. ----.-----.---.----------.--------
.. -------.---------.----.---------.--------------.-------.--------------------------------.---.----. -------.------.----------.-----.--------.i ! I i ! I ! l i l ! VILVE LINE/ INSIDE/ NORMIL Vf.I..VE POWER PRIMARY SECONDARY MAXIMUM VALVE POSITION N'PENDIX J ! I VILVE OUTSIDE FLOW Vf.I..VE ACTUATION ACTUATION CLOSURE NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SMCE SIGNIL SIGWL TIME CSEC.) NORMAL SHUTDOWN FAIL PRIMARY :>ECONDAR'r REQUIREMENT
! ! GPV-010 /1 OUTSIDE N/A GLOBE MANUAL N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A FLANGES 6/6 BOTH N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A B i i I I i ! ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO I!) FLUID CONT AINEO: AIR LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
PI.ANG&IARIIIIMOV&DONLY DUlliNG .. IIP&::IIIMANI:E OF 'IYPII A TilT II HBD TC 0 L HBB u CONTAINMENT PENETRATION NO. P-34 DESCRIPTION:
CONTAINMENT PRESSURIZATION LINE ILRT PRESSURIZATION SYSTEM REFERENCE SECTIONCSl 6.2.6 REV.13 WOLF CQ.EEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 28 OF 74 I I I I i I l I I -*----------------------------------------------------------------------------------------------------------------------------------------J VALVE LINE! INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONL BBHV-8351C 212 OUTSIDE BBV-356 111 OUTSIDE BBV-24 7 r.1 r. OUTSIDE BBV-178 212 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NOQ9 FLUID CONTAINED' WATER LENGTH OF PIPING TO OUTERMOST I SOL A TION VALVE' 17.5 APPLICABLE CDC NO. 55
GENERAL COMMENT
S' THIS PEN£TRATfON PIPING HAS A HIGH PRESSURE W.O.TER INFLOW WHICH CLUDES THE NEED FOR AUTOMAiiC ISOLATION OF THIS PENE"TRA TION. THE CVCS CHARGING PUMPS SUPPLY REACTOR COOLANT PUMP SEAL INJECTION WATER. A.NO THERE IS A POTENTrAL FOR DAMAGE TO THE REACTOR C00LA,NT PUMP IF UNDESIRED ISOLATION SHOULD OCCUR. THE ISOLATION CAN BE AFFECTEO BY REMOTE-MANUAL CLOSURE OF THE MOTOR* OPERATED VAl.V.I::
B-Y l"HE OPERATOR AFTER THE CHARGING PUMPS COMPLETE THEIR SAFETY FUNCTION.
NORMAL VALVE FLOW TYPE DIRECTION IN GLOBE N/A GLOBE N/A GLOBE IN CHECK PUMPS POWER PRIMARY VALVE ACTUATION SOURCE OPERATOR SIGNAL MOTOR 4 NONE MANUAL N/A N/A MANUAL N/A N/A N/A N/A N/A HV 51C .... BCB :> L..::..:....l,_A
--+-t _____,I TC&D BCD SECONDARY MAXIMUM ACTUATION CLOSURE SIGNAL TIME ISECJ NORMAL REM/MAN N/A OPEN N/A N/A CLOSED N/A N/A CLOSED N/A N/A OPEN BCD t 1!l .... _j BCB _j V*178 VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED CLOSED OPEN \1 FAIL AS IS N/A N/A N/A PRIMARY f:>ECONDARY REQUIREMENT OPEN CLOSED c CLOSED N/A N/A CLOSED N/A N/A OPEN CLOSED c CONTAINMENT PENETRATION NO. P-39 DESCRIPTION' RCP -SEAL WATER SUPPLY REACTOR COOLANT SYSTEM REFERENCE SECTIONCSJ 5.0 REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 29 OF 74 VALVE LINE/ INSIDE! VALVE OUTSIDE NO. SIZE, IN. CONT. BBHV-83510 212 OUTSIDE BBV-358 1/1 OUTSIDE BBV-248 1'41 1'4 OUTSIDE BBV-208 212 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO N00 FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 17.5ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THIS PENETRATION PIPING HAS A HIGH PRESSURE WATER INFLOW WHICH CLUDES TH£ NEED FOR AUTOMATIC ISOLATION OF iHIS THE CVCS CHARG lNG PUMPS SUPPL V REACTOR COOLANT PUMP SEAL INJE:CTION WATER, AND THERE IS A POTEN"fiAL FOR DAMAGE TO THE REACTOR COOLANT PtJMP JF UNOESIREO ISOlATION SHOULD OCCUR. THE ISOLATION CAN BE AFFECTED BY REMOTE-MANUAL CLOSURE OF THE OPERATED VALVE BY THE OPERATOR AFTER THE CHARGING PUMPS COMPLETE THEIR SAFETY FUNCTION.
NORMAL VALVE POWER FLOW VALVE TYPE SOURCE DIRECTION OPERATOR IN GLOBE MOTOR 4 N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A IN CHECK N/A N/A CHARGING PUMPS PRIMARY SECONDARY ACTUATION ACTUATION SIGNAL SIGNAL NONE REM/MAN N/A N/A N/A N/A N/A N/A H 8361 MAXIMUM VALVE POSIT ION APPENDIX J CLOSURE TIME ISEC.l NORMAL SHUTDOWN N/A OPEN OPEN N/A CLOSED CLOSED N/A CLOSED CLOSED N/A OPEN OPEN BCD TC&V Ul _j _j BCB V*208 BCD FAIL AS IS N/A N/A N/A PRIMARY REQUIREMENT OPEN CLOSED c CLOSED N/A N/A CLOSED N/A N/A OPEN CLOSED c PUMPD CONTAINMENT PENETRATION NO. P-4D DESCRIPTION:
RCP -SEAL WATER SUPPLY REACTOR COOLANT SYSTEM REFERENCE SECTION
5.0 REV 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 30 OF 74 VALVE LINE! INSIDE/ VALVE OUTSIDE NO. SIZE,IN. CONT. BBHV-8351A 212 OUTSIDE BBV-352 1/1 OUTSIDE BBV-245 7'41 :y. OUTSIDE BBV-118 2/2 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO N00 FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 20.0ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THIS PENETRATrON PIPING HAS A HIGH PRESSURE WATER INFLOW WHICH CtUDES THE NEED FOR AUTOMATIC ISOLATION OF THts PENETRATION.
THE CVCS CHARGING PUMPS SUPPLY REACTOR COOLANT PUMP SEAL INJECTION WATER, A.ND THERE IS A POTENTIAl FOR DAMAGE TO THE REACTOR COOLANT PUMP IF UNDESIRED ISOLATION SHOUlD OCCUR. THE ISOLATION CAN BE AFFECTED BY REMOTE-MANUAL CLOSURE OF THE OPERATED VALVE BY THE OPERATO AFTER THE CHARGING PUMPS COMPLETE THEIR SAFETY FUNCTION.
NORMAL VALVE POWER FLOW VALVE DIRECTION TYPE OPERATOR SOURCE IN GLOBE MOTOR 4 N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A IN CHECK N/A N/A PUMPS PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME <SEC.l NORMAL NONE REM/MAN N/A OPEN N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A OPEN BCD t HV 51C TC&V_j 18/U BCB _j V*178 BCD VALVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED CLOSED OPEN "V FAIL AS IS N/A N/A N/A PRIMARY SECONDARY REQUIREMENT OPEN CLOSED c CLOSED N/A N/A CLOSED N/A N/A OPEN CLOSED c CONTAINMENT PENETRATION NO. P-41 DESCRIPTION:
RCP -SEAL WATER SUPPLY REACTOR COOLANT SYSTEM REFERENCE SECTION
5.0 REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 31 OF 74
- -
- :.
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-:. :-:-_:-=----=--=---
--=---:_ -:-:..:-. .:-:.:-_-:-:..:.
- --:-_ _-_-:-: _-_ _-:-_-: _-:-: _-_ =-: .:--:-...:..
-=-:..: .:.:-.: _-:-_:_ ::.:::-_:
.::-:. ---=-=-
I I ' I i l I I I . I I LINE/ IIIISIDEI NORMf.l PRIMARY MAXIMUM I 1 Vf.LVE VM..VE POWER SECONDARY V/'LVE POSITION APPENDIX J VALVE OUTSIDE FLOW VALVE ACTUATION ACTUATION CLOSURE NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SOURCE SIGNf.l SIGNAL TIME CSECJ NORM/'L SHUTDOWN FAIL PRIMARY SECONDARY REQUIREMENT I , ! I HDV-016 2/2 OUTSIDE IN GLOBE MANUAL N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A c i I i I I I i I I i ! i I I I I I ! i HDV-023 OUTSIDE HDV-017 1212 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO I!) FLUID CONTAINED:
STEAM LENGTH OF PIPING TO OUTERMOST ISOLATION VAILVE: 14.9ft APPLICABLE GDC NO. 56 GENERAIL COMMENTS-NONE N/A GLOBE MANUAL IN GLOBE MANUAL HID N/A N/A N/A N/A CLOSED N/A N/A N/A N/A CLOSED Hll Hll HID _j t V-011 --, HBB I V417 TC HBD N/A N/A CLOSED N/A CLOSED N/A N/A CLOSED N/A c RII.HEAD WAIHDOWN AIII!A CONTAINMENT PENETRATION NO. P-43 DESCRIPTION:
DECONTAMINATION STEAM DECONTAMINATION SYSTEM REFERENCE SECTIONCSl 12.3 REV.13 WOLF CREEK UPDATED SAFBTY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 32 OF 74 I I I I i I I ------------------------------------------------------------------------------------------------------------------------------------------J
- ---------------------------------------------------
i ! I i l I ! ! I i 1 I VN..VE LINE/ INSIDEJ NORMAL VN..VE VALVE POWER PRIMARY SECONDAAY MAXIMUM V/>LVE POSITION APPENDIX J VALVE OUTSIDE FLOW ACTUATION ACTUATION CLOSURE NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SOIA1CE SIGNAL SIGNAl.. TIME (SEC.) NORM/>L SHUTDOWN FAIL PRIMARY SECONDAR't REQUIREMENT I ' ! ! HBHV-7126
¥41 ¥4 INSIDE OUT DIAPHRAGM AIR 1 CIS-A NONE 10 OPEN OPEN CLOSED CLOSED N/A c HBHV-7150
¥..1 ¥4 OUTSIDE OUT DIAPHRAGM AIR 4 CIS-A NONE 10 OPEN OPEN CLOSED CLOSED N/A c 1 i I HBV-420 ¥..1 ¥4 OUTSIDE N/A GLOBE MANUAL N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A i I i I I I I I I i
- I I I I I ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NO [!J FLUID CONT AINED* GAS LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 12.Jft APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
NONE MV-7121 DRAIN TANK L r ! TC 43 HIID HV-71150 HBB HBD COMI'R&SSOR CONTAINMENT PENETRATION NO. P-44 DESCRIPTION:
R.C.D. TANK VENT LINE LIQUID RADWASTE SYSTEM REFERENCE SECTIQN(Sl 11.2 REV.1J WOLF Cl\EEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 33 OF 74 i I I ---------*---*----------------------------*---------------*----------------------------*---*-------------------*---------*---------*--*---J I ' I I I
---.. -----------------------------------------.--.-----------------------------------.---------------------------------
.----
! i I ! I I I ! i I i I I I I i I . i I ! ! ! I ! 1 I i I I I I I I Vf>lVE LINE/ INSIDE/ V,4LVE OUTSIDE NO. SIZE, IN. CONT. EPV-046 111 INSIDE EPV-043 f4/f4 OUTSIDE EPHV-8860 111 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO I!) FLUID CONTAINED:
GAS LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 13.0ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
NONE NORM,4L V,4LVE VALVE FLOW DIRECTION TYPE OPERATOR IN CHECK N/A N/A GLOBE MANUAL IN GLOBE AIR TANKS POWER PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SOlllCE SIGN,4L SIGN,4L !SECJ NORMAL N/A N/A N/A N/A CLOSED N/A N/A N/A N/A CLOSED 4 CIS-A NONE 10 CLOSED HV.acl VAlVE POSITION APPENDIX J SHUTDOWN CLOSED CLOSED CLOSED FAIL N/A N/A CLOSED PRIMARY sECONDARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CONTAINMENT PENETRATION NO. P-45 DESCRIPTION:
NITROGEN SUPPLY LINE ACCUMULATOR SAFETY INJECTION SYSTEM REFERENCE SECTION!Sl 6.3 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 34 OF 74 i I I ------------------------------------------------------------------------------------------------------------------------------------J
! :.
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- . =----:-_:-_-__ --::_:_-::
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- _-:: _-_:::_-:
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- . .::-:__ ::_:-:-:._
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-::_-_ :::__:_-:-:_
- ._ :-::_-:__:.:-_.::
_-:-:.__-_
=-=-::-_ :_::-: _-:_-:.:-
i I I I ! 1 I LINE/ INSIDE/ NORM.4L PRIMARY SECONDARY MAXIMUM V/>LVE POSITION I I I ' Vlt.VE V.4LVE POWER APPENDIX J V,4LVE OUTSIDE FLOW VM.VE ACTUATION ACTUATION CLOSURE NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SMCE SIGN,4L SIGN.4L TIME !SECJ NORM/>L SHUTDOWN FAIL PRIMARY REQUIREMENT
! ! i i EMHV-8802B 4/4 OUTSIDE IN GATE MOTOR 4 NONE REM/MAN N/A CLOSED CLOSED AS IS CLOSED OPEN A EMHV-8824
¥41 Y4 INSIDE our GLOBE AIR 1 CIS-A NONE 10 CLOSED CLOSED CLOSED CLOSED N/A A I I EMV-OOJ 2/2 INSIDE IN CHECK N/A N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED OPEN A i I EMV-004 1212 INSIDE IN CHECK N/A N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED OPEN A
¥41 7'4 INSIDE EMV-217, 169 EMV-170, 172 111 INSIDE EMV-059 1/1 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES I!] NOD FLUID CONTAINED' WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VM.VE: N/A APPLICABLE GDC NO. 55 GENER/>L COMMENTS' THII PI!NITIIATION II ...-ciCIATED WITH TH! HIGH I'I'IE!Ialll COOLANT IN.IIICTION IVITEM, WHICH lllliCIUIRID TO MITIGATE THI CONIEIIUENCEI liP A LOCA. A CHECK VALVE II PROVIDED FIIR EAOH 111/lNCH LINE INIIDI THE CONTAINMiNT, AND A IIEMOT!*MANUAL IIIILATIIIN VALVI II PIIGVIgiQ OU'IliiDI 1MI CGNTAIIIMINT, AIIIICILI!
ACTIVI 011 PAll lVI PAILURI CAll II ACCOMMIIDAn!D IINCII THE IVI'n!M II CI.OIED DU1SIDI THE CONTAIIIMINT AND It OIIIGNED AND CGNITIIUet'iD CCIIMIN* IUIIATI Wmt Ttll DEIIIiiN AND CONITIIUO.
TION IIF THII CONTAIN!IENT.
LEAKAGE II!TI!CTION FROM THII LIN& OUTIIIIIIllti CONTA-ENT II PRDVID!O, AIIIDCR .. ID lltii!CTIDN I.U, LCIIAL TilTING OP THE VALVII 1111 THI CLCIED IVITIIII OUTIIDI THI cotiTAINo MENT II NOT RICWIIIEIIIIIIC!
THIIYIIT!M II OPIIIATI!II MD INIPKTID IIUIIIIIQ NIIRMAL PLANT OPEIIATION TO AllURE THAT THE II 81111111 MAIIITAIII*
ED. NOTE* ALL VENTS, AND FLOW POINTS AS INDICATED BELOW. N/A GLOBE MANU/>L N/A N/A N/A GLOBE MANU/>L N/A N/A N/A GLOBE MANU/>L N/A N/A CCB "" SIS PUMP I BCB t A BCD N/A N/A N/A N/A N/A N/A ..._ BCB CLOSED CLOSED N/A CLOSED CLOSED N/A CLOSED CLOSED N/A V*C81 BCI D FP FP CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A N/A V.aa3 RCS ) BCA HOT LEG*! CONTAINMENT PENETRATION NO. P-48 DESCRIPTION' HOT LEG INJECTION HIGH PRESSURE SAFETY INJECTION SYSTEM REFERENCE SECTION!Sl 6.3 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT P BCB !,'. CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 1 PAGE 35 OF 74 i BCD ! I I ------------------------------------------------------------------------------------------------------------------------------------J VALVE UNE/ INSIDE/
VALVE OUTSIDE FLOW NO. SIZE, IN. CONT. DIRECTION EMHV-8835 4/4 OUTSIDE EMHV-8823 INSIDE EPV-020 12 INSIDE EPV-010 2/2 INSIDE EPV-040 /2 INSIDE EPV-030 212 INSIDE EM_y067 111 OUTSIDE EMV-068 thru 3/4-3/4 EMV-075 EMV-21B /1 INSIDE EM¥162 thru 168 ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTNNED' WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE' N/A .APPLICABLE GDC NO. 55
GENERAL COMMENT
S' THIS PENETRATION IS ASSOCIATED WITH THE HIGH PRESSURE COOLANT INJECTION SYSTEM, WHICH IS REQUIRED TO MITIGATE THE CONSEQUENCES OF A LOCA. A CHECK VALVE IS PROVIDED FOR EACH BRM'JCH LlfJE INSIDE THE COtJT AINMENT AND A REMOTE-MM'iUAL ISOLATION VALVE IS PROVIDED OUTSIDE THE CONTAINMENT.
A SINGLE ACTIVE OR PASSIVE F AlLURE CAN BE ACCOMMODATED SINCE THE SYSTEM IS CLOSED OUTSIDE THE CONTAINMENT M'JD IS DESIGNED AND CONSTRUCTED SURATE WITH THE DESIGN M'JD TION OF THE COtJT NNMENT. LEAKAGE DETECTION FROM THIS LINE OUTSIDE THE CDNT AINMENT IS PROVIDED.
AS DESCRIBED IN SECTION 9.3.3. LOCAL TESTir<G OF THE VALVES OR THE CLOSED SYSTEM OUTSIDE THE COtJT MENT IS NOT REQUIRED SINCE THE SYSTEM IS OPERATED AND INSPECTED DURING NORMAL PLANT OPERATION TO ASSURE THAT THE INTEGRITY IS BEING MAINTNNED.
NOTE: ALL VENTS, DRAINS AND FLOW POINTS AS INDICATED BELOW. BCD. IN OUT If< IN IN N/A N/A N/A VAlVE VAlVE POWER PRIMARY TYPE OPERATOR SOURCE ACTUATION SIGNPJ.. GATE MOTOR 4 GLOBE AIR 1 CIS-A CHECK WA N/A fJ/A CHECK N/A N/A N/A CHECK N/A N/A N/A CHECK N/A WA N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A fJ/A GLOBE MANUAL N/A N/A "' t;)w :c "' >----'
"' HV 835 >
- I -SECONDARY MAXIMUM ACTUATION CLOSURE SIGNAL TIME <SEC.) REM/MAN N/A NONE 10 N/A WA N/A WA N/A N/A N/A N/A N/A WA N/A N/A N/A N/A 1 2 FE 980 D FP VALVE POSITION APPENDIX J NORMAL SHUTDOWN OPEN OPEN CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED ,_,,*
D FAIL AS IS CLOSED N/A N/A N/A N/A N/A N/A N/A PRIMARY REQUIREMENT OPEN CLOSED A CLOSED N/A A OPEN CLOSED A OPEN CLOSED A OPEN CLOSED A OPEN CLOSED A CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A N/A 2 z 0 ;:: CONT NNMENT PENETRATION NO. P-49 DESCRIPTION:
COLD LEG INJECTION HIGH PRESSURE COOLANT INJECTION SYSTEM REFERENCE SECTION ($) 6.3 REV. 14 WOLF CRIEEK UPDATED SAFETY ANALYSIS REPORT cmn AINMENT FIGURE 6.2.4-1 <PAGE 36 OF 74l r-----* --------* ---------* -------------------------
* -------------------------*
- -------------------
- ------------------------------------------
- --------*----*
I VN-VE LINE/ INSIDE/ VN-VE OUTSIDE NO. SIZE, IN. CONT. GPV-011 /1 OUTSIDE GPV-012 111 OUTSIDE FLANGES 111 BOTH ASSOCIATED WITH A SN'ETY FEATURES SYS. YESO NO I!) FLUID CONT AINEO: AIR LENGTH OF PIPING TO OUTERMOST ISOLATION VN..VE: N/A .APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
FLANBEB ARE REMOVED ONLY DURING PERFORMANCE OF TYPE A TEST NORMN-VN-VE POWER PRIMARY FLOW VALVE ACTUATION DIRECTION TYPE OPERATOR SOOlCE SIGNN-N/A GLOBE MANUAL N/A N/A N/A GLOBE MANUAL N/A N/A N/A N/A N/A N/A N/A HBB I SECONDARY MAXIMUM ACTUATION CLOSURE SIGNN-TIME ISEC.l NORMN.. N/A N/A CLOSED N/A N/A CLOSED N/A N/A N/A _j VN..VE POSITION APPENDIX J SHUTDOWN CLOSED CLOSED N/A R FAIL N/A N/A N/A PRIMARY REQUIREMENT CLOSED N/A N/A CLOSED N/A N/A N/A N/A B CONTAINMENT PENETRATION NO. P-51 DESCRIPTION:
PRESSURE SENSING LINES ILRT PRESSURIZATION SYSTEM REFERENCE SECTION!Sl 6.2.6 REV.1.3 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 37 OF 74 ! I I ) I j I I I . : L _______________________________________________________________________________________________________________________________________
j VILVE UNE/ INSDE/ VILVE OUTSIDE NO. SIZE, IN. CONT. EJHV-67018 12/12 INSIDE EJ-67068 fJ/3 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!g NOD FLUID CONTAINED' WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A AIPPUCABLE GDC NO. 55
GENERAL COMMENT
S' THE RI!IIDUAI.
HIAT RIMOVAL IY!ITEM IUCITION UNE FROM THE RI!AClOR COOLANT fY8TI!M CONTAINITftO NORMAL* L Y CLOIIO, POWER-OPI!RATII)
RIMI)TII MANUAL VALVEI IN IERIEI ,_.,. THE OONTAINMI!NT.
THE VALVII ARI AL1D IN'fiiRLOCICEO TO PllliVENT THEil PROM DING INADVERTENTLY MENT IIIOLATION
- AIIURIO IY 8'1'8TEM .aLATION VALVII CLOEIT TO THE TAINMENT AND THI CLOfiO IMITIM OUT* eRIE THE CONTA.-ENT, INHICH m DIIIGN* EO AND COIIITIIUCTI!O -ENIURATI WITH THE DEIRCIN AND IXJNIITRUilTION OP THE CONTAINM&NT, LIAIIAOE DIITEC. TIDN FROM THII LINE OUTIIDE THE TAINMINT
- PROVIDED, AI DEICRIBEO
"'IECTIONII.II.I.
LOCAl. TEITINO OF THE VALVE OR THE CLDIIEO S'I'STCM OUTIIIIII 111& MENT II NOT RECWIIIED IINCI Till lftTIM
- OPIRATEO MD INIII'ICRO DURING NORMAL PLANT OPUATION TO -.r111 THAT THE INTEQArt'l' II BEING ED. NORMIL VILVE POWER FLOW VILVE DIRECTION TYPE OPERATOR SO\JlCE OUT GATE MOTOR 1 N/A RELIEF N/A N/A BCD D PRIMARY ACTUATION SIGNIL REM/MAN N/A BCO AREA OF CHANGE SECONDARY IAAXIMUM ACTUATION CLOSURE SIGNIL TIME CSECJ NORMAL NONE N/AY CLOSED N/A N/A CLOSED RC:S PRESSURIZER
,.,.,. IIELI&FTANK L r V/JJ..VE POSITION APPENDIX J SHUTDOWN FAIL PRIMARY REQUREt.IENT OPEN AS IS CLOSED N/A A CLOSED N/A CLOSED N/A A ---F £ RECIRC. SUMP SEE I'ENETRATION P*14 TcaY ECD CONTAINMENT PENETRATION NO. P-52 DESCRIPTION:
RHR SHUTDOWN LINES RESIDUAL HEAT REMOVAL SYSTEM REFERENCE SECTION(Sl 5.4. 7 lo 6.3 REV. 15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT l---*-------*-
CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 38 OF 74
r ------.----------.--------------.---------------------. .L.. ---.-----------------------------------------------------------------------------------------------------.----------------------------------------------------------------I I . I i ! I ! VN...VE LINE/ INSIDE/ NORMAL VN.VE POWER PRIMARY VALVE OUTSIDE FLOW VALVE ACTUATION NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SOURCE SIGNN. ECV-084 6/6 INSIDE IN GATE MANUAL N/A NIA ECV-085 111 OUTSIDE N/A GLOBE MANUAL N/A N/A ECV-083 6/6 OUTSIDE IN GATE MANUAL N/A N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!) FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* 6.1 FT APPLICABLE GDC NO. 56 FIE FUELING GENERAIL COMMENTS:
POOL NONE SECONDARY ACTUATION SIGNAL N/A N/A N/A MAXIMUM V .AL VE POSITION N'PENDIX J CLOSURE TIME lSECJ NORMAL SHUTDOWN NIA CLOSED OPEN N/A CLOSED CLOSED N/A CLOSED OPEN HCD wt-" . L HCB FAIL N/A N/A N/A HCB Hctl PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CONTAINMENT PENETRATION NO. P-53 DESCRIPTION*
CLEANUP RETURN FUEL POOL COOLING AND CLEANUP SYSTEM REFERENCE SECTIONCSl 9.1.3 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT I ! i I i I I ' I I I ; ! i I I I I ! ! i I i I ! ' I I I i ! j I I I I ! ! i I i I ! ' CONTAINMENT PENETRATIONS I FIGURE 6.2.4-1 1 I PAGE 39 OF 74 l l . I --------------------------------------------------------------------------------------------------------------------------_1 VN..VE LINE/ INSIDE/ NORMAL VN..VE POWER VN..VE OUTSIDE FLOW VALVE NO. TYPE SOURCE SIZE, IN. CONT. DIRECTION OPERATOR ECV 087 6/6 INSIDE OUT GATE MANUAL N/A ECV-086 111 OUTSIDE N/A GLOBE MANUAL N/A ECV-088 6/6 OUTSIDE OUT GATE MANUAL N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VN...VE: 6.1 FT .APPLICABLE GDC NO. 56
GENERAL COMMENT
S-NONE .. &FUELING POOL PRit.lARY SECONDARY t.lAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNN.. TIME (SEC.l NORMAL N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED HCD L HCB v.<JII1 T I VN...VE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN TC
- F.oJL N/A N/A N/A PRIMARY ECONDARY REQUIREt.lENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c FUEL POOL COOLING LOOP CONTAINMENT PENETRATION NO. P-54 DESCRIPTION:
REFUELING POOL CLEANUP LINE FUEL POOL COOLING 8! CLEANUP SYSTEM REFERENCE SECTION<Sl 9.1.3 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 40 OF 74 ------*-*-----------*-*-------------*-*-----------*-*-------------*-*-----------*-*-------------*-*-----------*-*-------------*-*-----------*__..J VN...VE LINE/ INSIDE/ VAJ....VE OUTSIDE NO. SIZE, IN. CONT. ECV-095 3/3 INSIDE ECV-094 111 OUTSIDE ECV-096 3/3 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NO (<9 FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION V,AJ...VE:
APPLICABLE GDC NO. GENER,AJ...
COMMENTS*
NONE 6.1 FT 56 NORMN... VAlVE POWER PRIMARY FLOW VAlVE ACTUATION DIRECTION TYPE OPERATOR SOURCE SIGNAL OUT GATE MANUAL N/A N/A N/A GLOBE MANUAL N/A N/A OUT GATE MANUAL N/A N/A V-093 V-120 V-092 SECONDARY MAXIMUM ACTUATION CLOSURE SIGNN... TIME (SECJ NORM,AJ... N/A N/A CLOSED N/A N/A CLOSED N/A N/A CLOSED HCD s fJl V*DI!i r TCScD ------, V,AJ...VE POSITION .APPENDIX J SHUTDOWN OPEN CLOSED OPEN TC * > FAJL N/A N/A N/A v ..... HCI HCD s PRIMARY ECONDARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c FUEL POOL IIKIMMER PUMf' CONTAINMENT PENETRATION NO. P-55 DESCRIPTION:
REFUELING POOL SKIMMER LINES FUEL POOL COOLING 8c CLEANUP SYSTEM REFERENCE SECTIONCSl 9.1.3 REV. 13 WOLF Cl\EEK UPDATED SAFETY ANALYSIS 1\EPOI\T CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 410F 74 i ------------------------------------------------------------------------------------------------------------------------------------
__________
j UPDATED SAFETY ANALYSIS REPORT WOLF CREEK CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 VALVE POSITION TIME (SEC.)
CLOSURE MAXIMUM SIGNAL ACTUATION SECONDARY SIGNAL ACTUATION PRIMARY SOURCE POWER OPERATOR VALVE TYPE VALVE DIRECTION FLOW NORMAL CONT.OUTSIDE INSIDE/SIZE, IN.VALVE LINE/NO.VALVE 1/1 1/1 1/1 INSIDE OUTSIDE OUTSIDE IN IN N/A GATE GATE GLOBE MANUAL 4 4 N/A CIS-A CIS-A N/A REM/MAN REM/MAN N/A 5 5 N/A CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED N/A CLOSED CLOSED CLOSED OPEN OPEN N/A NORMAL SHUTDOWN FAIL PRIMARY SECONDARY FEATURES SYS. YES NO ASSOCIATED WITH A SAFETY ISOLATION VALVE: N/A LENGTH OF PIPING TO OUTERMOST GDC NO. 56 APPLICABLE FLUID CONTAINED: CONT. ATM X GSHV-9 GSHV-8 GSV-032
REQUIREMENT APPENDIX J A,C A,C N/A
GENERAL COMMENT
S:
PAGE 42 OF 74 SOLENOID SOLENOID 2 GEN ANALYZER THE HYDRO-HCB HV 9 ATMOSPHERE CONTAINMENT TO 29 REV. 29 REFERENCE SECTION(S) 6.2.5 CONTAINMENT HYDROGEN CONTROL SYSTEM H SAMPLE RETURN DESCRIPTION:
CONTAINMENT PENETRATION NO. P-56 29 VN..VE LINE/ INSIDE/ NO. VI<J..VE OUTSIDE SIZE, IN. CONT. GSHY-38 111 OUTSIDE GSHY-39 111 INSIDE GSY 058 11 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO FLUID CONTAINED:
CONT. ATM LENGTH OF PIPING TO OUTERMOST ISOLATION 7.1 N>PLICABLE GDC NO. 56 COMMENTS:
NONE NORMN.. VI<J..VE FLOW TYPE DIRECTION IN GATE IN GATE NIA GLOBE ) CONTAINMENT ATMOSPHERE MONITOR GT*RE-31 VI<J..VE OPERATOR SOLENOID SOLENOID MANUAL ' POWER PRII.1.ARY SECONDARY MAXIMUM SOURCE ACTUATION ACTUATION CLOSURE SIGN.'L. SIGNN.. TIME CSEC.l NORMAL 1 CIS-A REM/MAN 5 OPEN 4 CIS-A REM/MAN 5 OPEN N/A N/A N/A N/A CLOSED HV 38 .....
_J HCB TC.--HCD HCD POSITION APPENDIX J SHUTDOWN OPEN OPEN CLOSED "1?'.t FAIL PRIMARY REQUIREMENT CLOSED CLOSED OPEN c CLOSED CLOSED OPEN c N/A CLOSED N/A N/A TO CONTAINMENT ATMOSPHERE CONTAINMENT PENETRATION NO. P-56 DESCRIPTION:
SAMPLE RETURN CONTAINMENT ATMOSPHERE MONITOR REFERENCE SECTIONCSl 9.4.6 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 42o OF 74 I i I -----**--*-. ---*----.
* ---------.-.--
.. ----. *----* *--------*-.---* ----* *----* ---------*-.. --.-----*----* *----*-----.. -**-*--.. -**-.. -**-* *-.. -.. ---.. -.. ----.. ----. *-.. -**-.* -. *--* --.--.----
.. ----*--.. -**-*. -.. -* *-. *-*. -**-. *-. J UPDATED SAFETY ANALYSIS REPORT WOLF CREEK CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 VALVE POSITION TIME (SEC.)
CLOSURE MAXIMUM SIGNAL ACTUATION SECONDARY SIGNAL ACTUATION PRIMARY SOURCE POWER OPERATOR VALVE TYPE VALVE DIRECTION FLOW NORMAL CONT.OUTSIDE INSIDE/SIZE, IN.VALVE LINE/NO.VALVE 1/1 1/1 1/1 1/1 OUTSIDE OUTSIDE INSIDE OUTSIDE IN IN IN OUT GLOBE GLOBE CHECK GLOBE N/A MANUAL 1 4 N/A N/A CIS-A CIS-A N/A N/A REM/MAN REM/MAN N/A N/A 5 5 N/A N/A CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED N/A N/A CLOSED CLOSED CLOSED CLOSED OPEN OPEN OPEN N/A NORMAL SHUTDOWN FAIL PRIMARY SECONDARY FEATURES SYS. YES NO ASSOCIATED WITH A SAFETY ISOLATION VALVE: 8.1 LENGTH OF PIPING TO OUTERMOST GDC NO. 56 APPLICABLE X SJV-111 SJV-114
REQUIREMENT APPENDIX J C C C N/A
GENERAL COMMENT
S:
PAGE 42b OF 74 SOLENOID SOLENOID SJHV-131 SJHV-132
& CONTAINMENT SUMP FLUID CONTAINED: REACTOR COOLANT V-114 BCB BCD BCB BCD BCB BCD CLOSED BCD BCB REV. 29 REFERENCE SECTION(S) 18.2.3
POST ACCIDENT SAMPLING SYSTEM SAMPLE RETURN DESCRIPTION:
CONTAINMENT PENETRATION NO. P-57 29
.:_:-: _-:_---: _ _: -=---:_-:-_ :-_ :-: =-----.:_ _---:_-=--=-::._::::_:
---=---:_:-:_-=-
=----_-::_.=_-=-::.
_-__:::-:_
- _ _--:-_:-_
=-=--=-:-:-_ -=------:-:-
_-__ --::_ :-_--:-::
_-_
_-:_.:_ :-: _-_:::-_ =-
_-_ _--::_ :-_-::-___-_
= _-:_---: _ _:-=---=-=-_-:--_:-_::
_:.
-=--! I VN..VE LINE/ INSIDU NORMN.. VN..VE POWER VALVE OUTSIDE FLOW VALVE NO. SIZE, IN. CONT. OIRECTION TYPE OPERATOR SO!JlCE EMV-006 /1 INSIDE IN CHECK N/A N/A EMV-162 o/.41¥4 OUTSIDE N/A GLOBE MANUAL N/A EMV-123 o/.41¥4 OUTSIDE N/A GLOBE MANUAL N/A EMHV-6668 111 OUTSIDE IN GLOBE AIR 4 ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO I!) FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 7.6 FT />PPLICABLE GDC NO. 56
GENERAL COMMENT
S:
SAFETY NONE INJECTION PUMPS HV ... PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNN.. SIGN!t. TIME !SEC.l N/A N/A N/A N/A N/A N/A N/A N/A N/A CIS-A NONE 5 v CCD
(/] VALVE POSITION APPENDIX J NORMAL SHUTDOWN CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED V-008 FAIL N/A N/A N/A CLOSED PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A c CONTAINMENT PENETRATION NO. P-56 DESCRIPTION:
ACCUMULATOR FILL LINE HIGH PRESSURE COOLANT INJECTION SYSTEM REFERENCE SECTION!Sl 6.3 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 1 , PAGE 43 OF 74 ! ' l I ' I . , L ___________________________________________
- --*-----*---*-----*---------*---*-----*---*-*-------*------------------*------------*--------j VALVE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. N/A N/A N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NO 129 FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 55
GENERAL COMMENT
S*
HYDRAULIC SENSORS PROVI[lE. LATION OF RCS FROM THE CAPIL* LARY TUBING. THE CAPILLARY TUBING AND THE LIS'S SERVE AS THE SECOND BOUNDARY.
THIS RANGEMENT IS SIMILAR TO THAT PROVIDED FOR THE CONTAINMENT PRESSURE TRANSMITTERS SHOWN ON SHEET 72 OF 74. NORMAL VALVE POWER FLOW VALVE TYPE SOURCE DIRECTION OPERATOR N/A N/A N/A N/A RV HEAD SEAL TABLE ---i RV HEAD PRIMARY SECOND.ARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME CSEC.l NORMAL N/A N/A N/A N/A L__ VALVE POSITION PPPENDIX J SHUTDOWN N/A IS F!IJL N/A PRIMARY ECONDAR REQUIREMENT N/A N/A A CONTAINMENT PENETRATION NO. P-59, 91 DESCRIPTION:
RVLIS SAMPLE LINE REACTOR COOLANT SYSTEM REFERENCE SECTION
18.2.13.2 REV. 11 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 43o OF 74
.-----------------
- --*-----*----------.
--* --------.,--.-
.. -------.. -------------------------------------------
.. -.. -.. -----------------------.-.------------------------------.-------
1 ' ! i ! l 1
- I ' ' ! ! ! VALVE LINE/ INS1DE/ VALVE OUTSIDE NO. SIZE, N. CONT. BBHV-8026 111 INSIDE BBHV-8027 111 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO(!] FLUID CONT AI NED* GAS LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* 12.3ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
NO It& NOOMAL VALVE FLOW VALVE DIRECTION TYPE OPERATOR BOTH DIAPHRAGM AIR BOTH DIAPHRAGM AIR l'ltliiUIIIZIII ) ) ltEUIFTNIIC POWER PRIM MY SECOND MY MAXI ACTUATION ACTUATION CLOSURE SOURCE SIGNAL SIGNAL TIME CSECJ NORMAL 1 CIS-A NONE 10 CLOSED 4 CIS-A NONE 10 CLOSED HV*1021 L r V AI... VE POSITION NlPENliX J SHUTDOWN OPEN OPEN HV.eDZ7 FAIL CLOSED CLOSED PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A c GAIEOUI ) ) RADWAITI!
IVBTEM CONTAINMENT PENETRATION NO. P-62 DESCRIPTION:
PRESSURIZER PURGE S. VENT LINE REACTOR COOLANT SYSTEM REFERENCE SECTION<Sl 5.0 REV. 11 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4:-1 PAGE 44 OF 74 L-----------------------------------------------------------------------------------------------------------------------------------
Vl>l._VE LINE/ NSIDE/ NORt.IAL Vl>l._VE VALVE POWER NO. VALVE OUTSIDE FLOW TYPE OPERATOR SOURCE SIZE, IN. CONT. DIRECTION KAV-118 4/4 OUTSIDE IN GLOBE MANUAL N/A KAV-163 111 OUTSIDE N/A GLOBE MANUAL N/A KAV-039 4/4 INSIDE IN CHECK N/A N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO FLUID CONTAINED:
AIR LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 8.6ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
NONE V*111 PRIMAAY SECONOAAY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNN.. SIGNAL Tt.lE (SEC.) NORMAL N/A N/A N/A CLOSED N/A N/A N/A CLOSED N/A N/A N/A CLOSED _j V431 -, -t HBB ' ¥ > TcaD HBD V AI... VE POSITION APPENDIX J SHUTDOWN CLOSED CLOSED CLOSED FAIL N/A N/A N/A PRIMARY SECONDAR' REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c REACTOR IUII.DING CONTAINMENT PENETRATION NO. P-63 DESCRIPTION*
SERVICE AIR COMPRESSED AIR SYSTEM REFERENCE SECTIONCSl 9.3.1 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONT AJNMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 45 OF 74 I L---------------------*--------------------------------------------------------------------------------------------------------------------'-
UPDATED SAFETY ANALYSIS REPORT WOLF CREEK CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 VALVE POSITION TIME (SEC.)
CLOSURE MAXIMUM SIGNAL ACTUATION SECONDARY SIGNAL ACTUATION PRIMARY SOURCE POWER OPERATOR VALVE TYPE VALVE DIRECTION FLOW NORMAL CONT.OUTSIDE INSIDE/SIZE, IN.VALVE LINE/NO.VALVE 1/1 1/1 1/1 1/1 INSIDE OUTSIDE OUTSIDE OUTSIDE OUT OUT OUT N/A GLOBE GLOBE GLOBE GLOBE
1 4 1 N/A CIS-A CIS-A CIS-A N/A REM/MAN REM/MAN REM/MAN N/A 5 5 5 N/A CLSOED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED N/A CLOSED CLOSED CLOSED CLOSED OPEN OPEN CLOSED N/A NORMAL SHUTDOWN FAIL PRIMARY SECONDARY FEATURES SYS. YES NO ASSOCIATED WITH A SAFETY ISOLATION VALVE: 8.6 LENGTH OF PIPING TO OUTERMOST GDC NO. 55 APPLICABLE X SJHV-128 SJHV-129 SJHV-130 SJV-106
REQUIREMENT APPENDIX J C C C N/A
GENERAL COMMENT
S:
PAGE 45a OF 74 SOLENOID SOLENOID SOLENOID FLUID CONTAINED: REACTOR COOLANT MANUAL BCB BCD TC 29 29 REV. 29 REFERENCE SECTION(S) 18.2.3/9.3.2 RC LOOP AND PRESSURIZER LIQUID SAMPLE DESCRIPTION:
CONTAINMENT PENETRATION NO. P-64 VltVE Lll£/ INSIDE/ VM.VE OUTSDE NO. SIZE, IN. CONT. GSHV-20 6/6 INSIDE GSHV-21 6/6 OUTSIDE GSV-041 /1 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
CONT. AIR LENGTH OF PIPING TO OUTERMOST ISOLATION V/'LVE: 5.5ft APPLICABLE CDC NO. 56
GENERAL COMMENT
S:
NONE NORMAL VM..VE VltVE POWER PRIMARY SECONDARY MAXIMUM V I'L VE POSITION APPENDIX J FLOW ACTUATION ACTUATION CLOSURE DIRECTION TYPE OPERATOR SOURCE SIGNAL SIGNAL TIME <SEC.l NORMAL SHUTDOWN FAIL PRIMARY REQUIREMENT OUT BUTTERFLY MOTOR 1 CIS-A REM/MAN 5 CLOSED CLOSED AS IS CLOSED OPEN c OUT BUTTERFLY MOTOR 4 CIS-A REM/MAN 5 CLOSED CLOSED AS IS CLOSED OPEN c N/A GLOBE MANUAL N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A HV HV II 21 L CONTAINMENT AIR !1------1 1------11--
) r i wmT<M HBD TC CONTAINMENT PENETRATION NO. 65 DESCRIPTION:
CONT.H2 PURGE CONTAINMENT HYDROGEN CONTROL SYSTEM REFERENCE SECTION<Sl 6.2.5 REV.13 WOLF CREEK UPDATED SAFBTY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 46 OF 74 VN..VE LINE/ INSIDE/ Vf>J..VE OUTSIDE NO. SIZE, IN. CONT. ENHV-12 10/10 OUTSIDE ENV-080 1/1 OUTSIDE ENV-017 0/10 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!) NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OU'TERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. 56 GENERfoJ..
COMMENTS:
THII PINITIIATION II -IA11!0 WITH THE COIIITAINMINT .RAY IYITIM, WHICH II R!OUIRID TO MITIGoiiTI THI C:Oal!* QU!NCI!I OP A LOCA. A I:HICIC VALVIII PROVIDED INIIDI THE CONTAINMENT, AND A REMO'noMANUAI.
IIOI.ATION VALVE PIIOYIDI!D CXITIIOE TH1 CONTAINMENT, A liNGLE ACTIVE OR PAIIIVI PAIWRI ClAN II& loGOOMMODATJD IlNCE THE t\'IITEM II CLCIEO 0U'111DI THI COIIITAIN
.. IIIT AND II DIIIGNIO AND COWTII\ICTJO IURATI WITH TH! 111110111 AND TION OF THE OOIIITAINM!NT.
LIAUCIE OITICnON FROM LINE 0111'1101 THE CONTAINM!NT
VN..VE POWER FLOW Vf>J..VE DIRECTION TYPE OPERATOR SOURCE IN GATE MOTOR 4 N/A GLOBE MANUAL N/A IN CHECK N/A N/A A PRNARY SECONDAAY MAXIMU.! ACTUATION ACTUATION CLOSURE SIGNI>J..
SIGNI>J..
TIME !SEC.) CSAS REM/MAN N/A N/A N/A N/A N/A N/A N/A HV 1:1 _J HCI ,-, Y*017 HCB I > t TC HCD VALVE POSITION APPENDIX J NORMAL SHUTDOWN CLOSED CLOSED CLOSED CLOSED CLOSED CLOSED TC<O HCD iXi 0 I > HCB FAIL AS IS N/A N/A PRIMARY REQUIREMENT OPEN CLOSED A CLOSED N/A N/A OPEN CLOSED A NOZZLES CONTAINMENT PENETRATION NO. P-66 DESCRIPTION:
CONTAINMENT SPRAY CONTAINMENT SPRAY SYSTEM REFERENCE SECTION
6.2.2 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 47 OF 74 r-------------------------.--------
.. ----.. ----*---.-.---
- ----------. -------------------------
.. --------------.----------.-----------.-----.---------.----------.---------.-----------.----------.---.----------------------.--------
l l . j l ) ) ' 1 l I i 1 1 l l I . j h i ) l ' j , ! i 1 j 1 I. l I I
- l
- i ) l ' ) i 1 l 1 I VN...VE LNE/ INSIDE/ NO. VN...VE OUTSIDE SIZE, IN. CONT. KCHV-253 4/4 OUTSIDE KCV-478 4/4 INSIDE KCV-431 1. INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO I!] FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: APPLICABLE GDC NO.
GENERAL COMMENT
S=
NONE 15.4 56 NORMAL VN...VE VN...VE FLOW TYPE OPERATOR DIRECTION IN GATE MOTOR IN CHECK N/A N/A GLOBE MANUAL POWER PRIMARY SOURCE ACTUATION SIGNN... 1 CIS-A N/A N/A N/A N/A SECONDARY t.IAl<IMUt.l VALVE POSITION APPENDIX J ACTUATION CLOSURE SIGNAL Til.£ [SEC.l NORt.lAL SHUTDOWN NONE 30 CLOSED CLOSED N/A N/A CLOSED CLOSED N/A N/A CLOSED CLOSED HV _j V*471 HBB --, HBB:tKFB s KFB TCIIV FAIL AS IS N/A N/A PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A c CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-67 DESCRIPTION:
FIRE PROTECTION FIRE PROTECTION SYSTEM REFERENCE SECTION!Sl 9.5.1 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 48 OF 74 ' ' I ' ' L.---*-*------*-------------------*-*---*---*---*---------*---*-*---*---*---*-*--*----*---*----*---*---*----*----*---*-*-----------*---*
I I I I I I I i I I i I I I I I r-------.--------------------------.-
- ---------*--------------.-----------------------------------------------------------.-----.----------------.--------------.-----------------.---------.----------------.-----.---------------., I i it V/ILVE l.MI INSIDE/ NORI.W.. V/ILVE V/ILVE POWER NO. VPJ..VE OUTSIDE FLOW TYPE OPERATOR SOURCE SIZE, IN. CONT. DIRECTION SJHV-12 111 INSIDE OUT GATE SOLENOID 4 SJHV-13 r11 OUTSIDE OUT GATE SOLENOID , SJV-071 111 OUTSIDE N/A GLOBE MANUAL N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!] FLUID CONTAINED:
STEAM LENGTH Of PIPING TO OUTERMOST ISOLATION VALVE: 7.7ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
NONE PRIMAAY SECONDARY MAXIt.lJM ACTUATION ACTUATION CLOSURE SIGNPJ.. SIGNAL TIME CSECJ NORMAL CIS-A NONE 5 CLOSED CIS-A NONE 5 CLOSED N/A N/A N/A CLOSED HV 12 BCD TC VALVE POSITION APPENDIX J SHUTDOWN FAIL CLOSED CLOSED CLOSED CLOSED CLOSED N/A HV 13 PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A c CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-69 DESCRIPTION:
PRESSURIZER VAPOR SAMPLE LINE NUCLEAR SAMPLING SYSTEM REFERENCE SECTIONCSl 9.3.2 REV. 13 WOLF CREEK il !I ,, il ., I i ! UPDATED SAFETY ANALYSIS REPORT i! i ! II CONTAINMENT PENETRATIONS
!t FIGURE 6.2.4-1 ! I I PAGE 49 OF 74 t* I i 1 t 1 L -"
- =
.. -.-=-
-" =*
=*
.. -: *=-
..
___ J.
,--;--.,_------
- r VN.VE LINE/ INSIIJE/ VN.V£ OUTSIDE NO. SIZE, IN. CONT. EFHY-31 4/14 OUTSIDE EFY-276 11 INSIDE EFHY-33 4/14 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!] NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION Y/>LYE* 15.5 ft APPLICABLE GDC NO. 56 GENER/>L COMMENTS*
'I'H. PIIIIITIIATION Ill -IATID IIITH THI laiiiiTIAL .IIYICE WAT!II -TEM, WHICH II AICIUIII!D TO MITIGATE THE CQNgQUINCD OP II LDCII. A MANUAL I'OIVIII-GPIIIATIO YIILVI II LCCATIID INaiDI, AND A llliMOTI-MANUIIL POVIIII-OPIIIATID YALYI II !.DCA TID OUT* 1101 TH1 CONTAINMINT.
THQE YALYI!I IIIII I'OWI!III!O PROM THE SAM! PDIIIIII IDUIICI! fOil DIII!IITIII l'f8-T!M IIELIAIIILITY.
II IINDLI ACTIYI OR P_,YI PAILUIII CAN IIIIIIICOMMODAtliD IlNCE THE IIYITIM II A a I.OIID IVITIM I .. IDE T111 CONTAINMI!NT, WHICH ra DE* IIONI!D AND CDIIIri'IIUCTIIO IN ANCI WITH A1M1 IEC110N Ill, CLAa S REDUIREM6111'&
THE -IITIAL IEAYIGI WAftR LINEI ARE NOT YINTI!D 011 IIiD DUlliNG A TYPE A 'JUT IlNCE THE IIIII COOLIIIS MAY Ill IIIQUIIIED TO COOL THI CONTAINMiiNT.
A TYPII 0 naT II PIIIPDIIM*
EO. NORMN. VALVE VN.VE POWER FLOW DIRECTION TYPE OPERATOR SOURCE IN BUTTERFLY MOTOR 1 N/A GATE MANUAL N/A IN BUTTERFLY MOTOR 1 HV 31 W"TEI! IIUPI'L Y HBC HCB HBB PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNN. TIME CSEC.l NORM />d. SIGNN.. SIS REM/MAN N/A OPEN N/A N/A N/A CLOSED SIS REM/MAN N/A OPEN HV 33' 10 HBB HCB HBC > D Y/>LVE POSITION APPENDIX J SHUTDOWN OPEN CLOSED OPEN FAIL PRIMARY REQUIREMENT AS IS OPEN CLOSED c N/A CLOSED N/A N/A AS IS OPEN CLOSED c CONTAINIIINT All! COOLII!S CONTAINMENT PENETRATION NO. P-71 DESCRIPTION; ESW TO CONTAINMENT AIR COOLER ESSENTI/>L SERVICE WATER SYSTEM REFERENCE SECTION(S) 6.2.2 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 50 OF 74 =-=---=-=-=---=
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Vf>lVE LINE/ INSIDE/ NO. Vf>lVE OUTSIDE SIZE, IN. CONT. EFHV-45 14/14 INSIDE EFV-277 111 INSIDE EFHV-49 14/14 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!] NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VAI...VE:
N/A APPLICABLE GDC NO. 56 GENERM.. COMMENTS:
'I'll. -TIIAnGN 18 AIKICIATID WITH ?Nil -NTIAL IIIIVICI WATIII W!III;tl Ill IIICIUIIIID TO IIITIIIATI Till .-auiNCIII GP A UICA. A IIIMOft. IMIIftW. POIIIa.aNIIATID YALVII II LOCATIO u*IDI MD A lltiiiiJTI.MNIUAL VALVI II LCICATIG IIUT* trill TMI ld!TAIIIIIIIIT.
Til-VALVIII Alii NWIMD I'IICIII'Itll IMIIt I'CIIIIII .UIICI 1'011 tiiiiAIIII W. on. IIR.-uTT, A -LI MTIVI 011 PAll lVI PAILUIII GAM
- ACCOIIIIIDQITID 111111:11 Till lftnlll II A CLCIIID lftTIIII -DI Till IIIDIIITA_,., IIMIM II IIIINID llle CDII!IriiUCTID 1111 ANCI llnM .aTION Ill, a... I IIIQUIR-rmL Till -'IITW. IIIIVICI W...TIII LINII Alii NOT YINftlt 1111 DIIAIIIU DUll-A TYN A Tln'IIIICI TM1 Alii COOUIIII MAY
- IIIIIUIIIID
'nil coaL 1MIIXIII'TAINMIIft, ,.,.,._ DTal .UIIFOIIIIID.
NORMf>l Vf>lVE Vf>lVE FLOW TYPE OPERATOR DIRECTION OUT BUTTERFLY MOTOR N/A GATE MANUM.. OUT BUTTERFLY MOTOR HV CONTAINioiENT > ) AIR COOLERS POWER PRIMARY SECONDARY MAXIMUM SOURCE ACTUATION ACTUATION CLOSURE SIGNf>l SICNf>l TIME CSEC.) NORMAL 1 SIS REM/MAN N/A OPEN N/A N/A N/A N/A CLOSED 1 SIS REM/MAN N/A OPEN D VM..VE POSITION N'PENOIX J SHUTDOWN OPEN CLOSED OPEN FAIL AS IS N/A AS IS PRIMARY REQl.!REMENT OPEN CLOSED c CLOSED N/A N/A OPEN CLOSED c CONTAINMENT PENETRATION NO. P-73 DESCRIPTION*
ESW FROM CONTAINMENT AIR COOLER ESSENTIAL SERVICE WATER SYSTEM REFERENCE SECTION(S) 6.2.2 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 51 OF 74 ------------------.-----
- ---*-------------------------
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_-_ :-:------=--j I VN...VE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. EGHV-58 2/12 OUTSIDE EGV-090 111 OUTSIDE EGV-204 2/12 INSIDE EGHV-127 12/12 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!) FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 11.9 ft APPLICABLE GOC NO.
GENERAL COMMENT
S:
liD LAnDI MITCH PROVIDED IN TN! .OF VALVE HV*I27 56 NORMN... VN...VE FLOW DIRECTION TYPE IN GATE N/A GLOBE IN CHECK IN GATE COMPONENT COOLING WATER VPJ..VE OPERATOR MOTOR MANUAL N/A MOTOR POWER SOL!1CE HV 71 1 N/A N/A 4 0> CCI 0 I > TC PRIMARY SECONDARY ACTUATION ACTUATION SIGNAL SIGNN... CIS-B NONE N/A N/A N/A N/A REM/MAN NONE HV .. MAXIMUM VPJ..VE POSITION APPENDIX J CLOSURE TIME !SEC.l NORMAL SHUTDOWN 30 OPEN OPEN N/A CLOSED CLOSED N/A OPEN OPEN N/A CLOSED CLOSED _J V*204 -,-0 0> 0 ' > FAIL AS IS N/A N/A AS IS PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CLOSED N/A c CONTAINMENT PENETRATION NO. P-74 DESCRIPTION:
CCW TO REACTOR COOLANT PUMPS COMPONENT COOLING WATER SYSTEM REFERENCE SECTION!Sl 9.2.2 REV. 13 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 52 OF 74 l ! I ! I ' I i . ! . . I -------------------------------------------------------------------------------------------------------------------------------J
=----.----.-----.---------
- ----------.---.----
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- --*-----------------.--------.-----------------.----------
- ---. ----.------------
---*-----------.-
- -:* -----.---------------.-----.----.-----------.---------.-:I Vlt.VE LINE/ INSIDE/ NO. Vlt.VE OUTSIDE SIZE, IN. CONT. EGHV-60 12/12 INSIDE EG-V372 11 OUTSIDE EGHV-59 12/12 OUTSIDE EGHV-131 12/12 OUTSIDE EGHV-130 12/12 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 11.1 ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S:
11UU11111-.TI:H I'IIGWIIIED
.. lHE CD1111111L 11111111 FOIJIIIIIEII LICIUIUTQF VALVES HV-1311, Ill NOR Mit. Vlt.VE Vlt.VE POWER FLOW TYPE OPERATOR SOURCE DIRECTION OUT GATE MOTOR 4 N/A GLOBE MANUAL N/A OUT GATE MOTOR 1 OUT GATE MOTOR 4 OUT GATE MOTOR 1 v V-126 PRit.IARY SECONDARY t.IAXIMUM ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME CSEC.l NORMAL CIS-B NONE JD OPEN N/A N/A N/A CLOSED CIS-8 NONE JO OPEN REM/MAN NONE N/A CLOSED REM/MAN NONE N/A CLOSED HV HV liD TC 58 !** L r VALVE POSITION N'PENDIX J SHUTDOWN OPEN CLOSED OPEN CLOSED CLOSED FAIL AS IS N/A AS IS AS IS AS IS PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CLOSED N/A c CLOSED N/A c CONTAINMENT PENETRATION NO. P-75 DESCRIPTION:
CCW RETURN COMPONENT COOLING WATER SYSTEM REFERENCE SECTION
9.2.2 REV.13 WOLF CREEK UPDATED SAFETY ANALYSIS RBPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 53 OF 74 !* ,, :I !, !I i* il ,, !I ., I :I '* -----.------------------
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, ! I ' l . I I I I I I . I ! I I VN..VE LINE/ INSIDE/ NORMN.. VN..VE VN..VE POWER PRIMARY SECONDARY MAXIMUM V N.. VE POSITION APPENDIX J VN..VE OUTSIDE FLOW ACTUATION ACTUATION CLOSURE I ! i I i NO. SIZE, IN. CONT. EGHV-62 !4/4 INSIDE EGV-371 1/1 OUTSIDE EGHV-61 !4/4 OUTSIDE EGHV-132 4/4 INSIDE EGHV-133 4/4 OUTSIDE ASSOCIATED WITH A SN"ETY FEATURES SYS. YESO NO I!) FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VN..VE: 7.9 ft. APPLICABLE GDC NO. 56 GENERN.. COMMENTS*
- IIOLATIDN IWTICH PROVIDED IN THI j CDITROL 110011 FOIIPOW!R LOCKOUT DF . VALYU HV112 AND H\1111 DIRECTION TYPE OPERATOR our GATE MOTOR 4 N/A GLOBE MANUAL N/A our GATE MOTOR 1 OUT GATE MOTOR 1 our GATE MOTOR 4 CBD v .... CBC V-369 otu In 1D u u s SIGNN.. SIGNN.. TIME CSECJ NORMN.. CIS-B NONE 30 OPEN N/A N/A N/A CLOSED CIS-B NONE 30 OPEN REM/MAN NONE N/A CLOSED REM/MAN NONE N/A CLOSED HV HV 12 11 L r V-371 SHUTDOWN OPEN CLOSED OPEN CLOSED CLOSED FAIL AS IS N/A AS IS AS IS AS IS PRIMAAY SECONDAAY REQUIREMENT CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CLOSED N/A c CLOSED N/A c CONTAINMENT PENETRATION NO. P-76 DESCRIPTION:
CCW FROM RCP THERMAL BARRIER COMPONENT COOLING WATER SYSTEM REFERENCE SECTION<Sl 9.2.2 REV.13 WOLF CREEX UPDATED SAFBTY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 54 OF 74 i
VALVE LINE/ INSCE/ VALVE OUTSIDE NO. SIZE, IN. CONT. BMV-045 f313 INSIDE BMV-302 111 OUTSIDE BMV*046 OUTSIDE ASSOCIATED WITH A SAF'ETY rEATURES SYS. YESO NO[!] FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: A('>PLICABLE GDC NO.
GENERAL COMMENT
S*
NONE 6.1 ft. 56 NORMAL VALVE VALVE POWER PRIWARY FLOW ACTUATI<m DIRECTION TYPE OPERATOR SOURCE SICNAI. OUT GATE MANUAL N/A N/A N/A GLOBE MANUAL N/A N/A OUT GATE MANUAL N/A N/A V-045 SECONDARY YAXIIIUII VALVE POSITION N'PENDIX J ACTUATION a.OSURE SICNAI. TIME CSEC.I NORMAL SHUTDOWN rAIL PRIMARY REOUREIIENT N/A N/A CLOSED CLOSED N/A CLOSED N/A c N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A c HBD TC ., ,::: Ul HBB <*l L r CONTAINMENT PENETRATION.
NO. P-78 . DESCRIPTION:
STEAM GENERATOR DRAIN LINE STEAM GENERATOR SLOWDOWN SYSTEM RErERENCE SECTIONCSl 10.4.8 WOLF CREEK REV. 28 UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE PAGE 55 OF 74 VALVE LINE/ INSIJE/ NO. VALVE OUTSIDE SIZE,IN. CONT. EJHV-87D1A 12/12 INSIDE EJB70BA 3/3 INSIDE EJV-154 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!g NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A />PPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THI III!IIOUAL HEAT IIEIIOYAL IYIII"EII sucnDN LINE FROM THI REAC'IOR COOLANT IIYIII"EM CONTAINS "IWO NORWW LY CI.QIE.D, POWER.OPii:RATED RIMD'n MANUAL YALYD IN IERIEI 1 .. 1111! THE CONTAINMENT.
rH1! VALVES ARE ALIO INTERLOCKED TD PllEWNT TH*M 'ROM III!INO INAIWERtENTL Y OPIIIIID. MENT 180LATION II Mllll!ED IV IYiftM IIOLATION
¥ALVEI Cl.lla!IIT TO THii 00111-TAINMENT AIIID TM& CL!la!D aTSTEM OUT* IIDB THB CONTAINII!NT, WHICH II DE8101fo ID AND CONITIIUCT1ED DOMMENIIURATI WITH THE DQION ANP -RUCTION OF "niE CONTAIIIIM&IIT.
LEAKAGE P&TEC. TION ,ROM THII LilliE DUTIIDE THI CONo TAIIIMINT II PIIOVIDID, M D-111110 IIIIEcTIONI_..&
LDCAL TI!ITIND OF THE VAL¥& OR THE CLDIED SYliTEIII OIIT8IDI THE MINT I& NOT 111!0UIRIOIINCI"n4EIIYIITEII IS OPERATED AND INIIPEeTED IIUIIIKO NORMAL PLANT QPSRATION TO AIIIIIRR THAT THE INT1EORITV II BEING It>. L ________________________
_ NORMAL VALVE VALVE FLOW TYPE OPERATOR DIRECTION OUT GATE MOTOR N/A RELIEF NIA N/A GLOBE MANUAL BCB D BCD POWER PRIMARY SO\JlCE ACTUATION SIGNAL 1 REM/MAN NIA IVA N/A N/A 4 ,&t SECONDARY IAAXIMUM V/U..YE POSITION riPPENDIX J ACTUATION CLOSURE REQUREMENT SIGNAL TIME (SECJ NORMIU.. SHUTDOWN FAIL PRIMARY NONE N/A CLOSED OPEN NIA N/A CLOSED CLOSED N/A N/A CLOSED CLOSED RCS PRESSUR IZI 1**'-.... u ECO .,j-t'd 'f! L ,. ECB r ECB "' v 0 > ECD TCtV AS IS NIA N/A CLOSED N/A A CLOSED NIA ( N/A3_ CLOSED N/A N/A AREA OF CHANGE--SEE PENETRATION P*1& CONTAINMENT RHIIPUMP SUCTION CONTAINMENT PENETRATION NO. P-79 DESCRIPTION:
RHR SHUTDOWN LINES RESIDUAL HEAT REMOV /U.. SYSTEM REFERENCE SECTIONCSl 5.4.7 and 6.3 REV. 15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 56 OF 74 I"-----------------------------------------------*-----------------------
- i ! l i l ! ! ! ! I ! LINE/ INSIDE/ NORMN.. POWER PRIMARY SECONDAA'I' MAXIMUM V f>L VE POSITION APPENDIX J VN..VE V/li..VE Vlli..VE ! ! NO. VN..VE OUTSIDE FLOW TYPE OPERATOR SO\IlCE ACTUATION ACTUATION CLOSURE REQUIREt.1ENT SIZE, IN. CONT. DIRECTION SIGNN.. S!GNN.. TIME CSEC.l NORMAL SHUTDOWN FAIL PRIMARY I ! i I i I I I I I I i ! i ! ! I I I ! i BGHV-8105 fs/3 OUTSIDE BGV-342 INSIDE BG8381 fs/3 INSIDE ASSOCIATED WITH A S/>FETY FEATURES SYS. YESO FLUID CONT AINED* WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 18.1 ft APPLICABLE GDC NO. 55
GENERAL COMMENT
S*
NONE IN GATE MOTOR 4 N/A GLOBE MANUAL N/A IN CHECK N/A N/A .. , SIS NONE 10 OPEN N/A N/A N/A CLOSED N/A N/A N/A OPEN TCBcV HV v ,. BCD > "' --h BCB L OPEN CLOSED OPEN BCD BCD AS IS N/A NIA HV 1108 CLOSED N/A c CLOSED N/A N/A CLOSED N/A c CONTAINMENT PENETRATION NO. P-80 DESCRIPTION*
CHARGING LINE CHEMICAL AND VOLUME CONTROL SYSTEM REFERENCE SECTIONCSl 9.3.4 REV.13 WOLF CREEK I i ' I I I ( UPDATED SAFBTY ANALYSIS REPORT ! I CONTAINMENT PENETRATIONS I FIGURE 6.2.4-1 i 1 PAGE 57 OF 74 1 I I i I I -*------*---------------------------------------------------------------------------------------------------------------------------J A Ylt.VE Lit£/ INSI:lE/ VH.VE. OUTSilE NO. SIZE, IN. CONT. EJHV*8809A 0/10 1 OUTSIDE EJV*054 /1 OUTSIDE EP8818A INSIDE EP8818B 616 INSIDE EJHC.V*8890A f.lf4 INSIDE EJV-134,V*136 f.tf.t INSIDE EJV-132 l/1 INSIDE *'f4 INSIDE EJV*I73, EJV-174 I<SSOCIATED WiTH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A
- APPLICABLE GDC NO. 55
GENERAL COMMENT
S:
THII. Kllli."'JIATIOfll II .USOCIATilO WITH' THI! HIGH PIIUIUIIS c:oot.AliiT IN.'ICTJOti I'YI'I'IM>
WHICK MI'TJOATI THE COIIIIUIIIDICIS Of A L-A Clfl¢rt VALW II NoVIMD FOR liaciCI llll.u!Ctt LifE 1-TKll COift'AINIIIEIIII'.
oUID' A llliMC)TUoiANIIAt .elATION VAIO.VE II PIIOVIO.O
-TKI!' CO!ffA-ltT.
A tlfiQLI AC:TIVEOIIP-"I!VE PAILUIII!
eM Ill!
me......,.., a Ci.OGIJ*exiJHJE' THE COift'A-MD IS DISIGlllD MDCOIII'niUCnD
_... IUilATlt l'nTit Tltll* OQIOIIMD cotenJIUI()
1'1011 01' ntl OOIIT-. Ll!ldUIGI Dtrl!lmOil fiiOM ftf. IJHE OUftllle"I'Hl!
c:oHTANilliiT lll'llOVID£D, At Diiil:iiUD lfiCcm<<<NIJIJL LOCAL turlfO OP THE VALVO Oil THl CLOIIID ft'ITUII IWniiDI! 'fiil COHTAIJII.
_, 111101' IIIOIIIIIID-nte nS'inl IS ClfliftATliO'
-t.a:nD OURING* -L I'UifT Ol'liMTJON TO _... fHAT THe l!ITI!OIIl'rY IIIII!IIM MAINTA** 10. NOTE' ALL VENTS, DRAINS AND FLOW POIN.TS loS INDICATED BELOW. TP BCD NORIIIt. Ylt.VE Ylt.VE POWER FLOW DfiECTION TYPE OPERATOR SOORCE ; ' IN GATE MOTOR , N/A GLOBE MANUAL N/A IN CHECK N/A. N/A IN CHECK N/A N/A out GLOBE: , AIR 1 I N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A N/A GLOBE MANUAL N/A PRIIARY ACTUATION SIGHit. NONE N/A N/A N/A CIS*A N/A N/A N/A 8809 A SECONDARY
' MAXMJM V AL.V£ POSITION Jf'PEiOX J ACTUATION
- Q.()Sl.fi( REOUAEr.tENT SIGNAL TNE ISECJ NORMAL SHUTDOWN FAIL PRIMARY SECONDARY REM/MAN N/A OPEN OPEN loS IS OPEN CLOSED A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A N/A N/A CLOSED CLOSED N/A OPEN CLOSED A N/A N/A CLOSED CLOSED N/A OPEN CLOSED A NONE 13 CLOSED CLOSED CLOSED CLOSED HIA A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A SIS 1'040 BCD eca . IICb TEST LINE 0 :u '\Ill/ ( a.t* .,. RCSCOI..O LEG >
- ._,_ .. ___ ...JI::, __ ) .. ) ' '
1..001'2 $:! CONTAINMENT PENETRATION NO. P-82 ). DESCRIPTION:
FP F'P COLO LEC INJECTION RESIDUAL HEAT REMOVAL SYSTEM REFERENCE SECTIONCSI 5.4.7 ond 6.3 REV.27 I; WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 58 OF 74 V/ILVE UN£/ INSilE/ VALVE OUTSIDE NO. SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES(!g NOD FLUID CONTAINED' WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE' N/A APPLICABLE GDC NO. NONE
GENERAL COMMENT
S' THI CONTAI-NT PENIITRATIONI
-0. CIATID W1T11 1HII ITEAM O!NERATORS ARB NOT GI&I!CT TD GDCe>, aiN.,_ TH& COHTA .. M!Nf BARRIER INTEGRITY IS NOT IR!ACHID, TH!
OR 8ARRIIA AGAiicr FilliON PR-..c:r LRAI<M! THii ENVIRONU&NT 18 THI IIOIDE OP THI aHAM liiiNIIIATDII TUIU AND THE OUT* IIDE OF TH! LINEI EMANATING FADM THEF!!AM THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORM/IL VN...VE POWER FLOW V/ILVE TYPE OPERATOR SOUlCE DIRECTION DBB TUBING s PRIMARY ACTUATION SIGNAL l SECONDARY MAXIMUM VALVE POSITION N'lmJ ACTUATION CLOSURE REQ t.tENT SIGNAl. TIME CSEC.l NORMAL SHUTDOWN FAIL PRIMARY AREA OF CHANGE r i'( PENETRATION TESTED PER TYPE A APPENDEX J REQUIREMENTS.
L r V*ll42 OBB DBD s H:L DBB DBD CONTAINMENT PENETRATION NO. P-
8.3 DESCRIPTION
'
STEAM GENERATOR 0 SAMPLE LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTION<Sl 10.4.8 REV.15 WOLF CREEK. UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS I FIGURE 6.2.4-1 I *-----------*------------------------
PAGE 59 OF 74 _j
VALVE LM/ INSilE/ VALVE OUTSIDE NO. SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!g NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: N/A APPLICABLE GDC NO. NONE
GENERAL COMMENT
S:
rH11 CONTAINMDT PEN1!TRATJONI All(). CIATiiP WITH THI! aTI!AM IIINIAATDAI
&Ill! NOT IUiollCT TD GPC.J, IlNCE THE Cotn'AIWMINT IIARAII!A IWTEIIRITY IS NOT TNI _,UNDAIIY D11 IIAAIII!II AGAINST PIBDN PRODUCT LIAKAIIE TO TNI INVIAONMENT IS THE INIIDI Df THE &TEAM GENE!IAmA TUllES AND THI Dill'* IIDE 01' THI LINM I!MMATING PIIOM TH! CI'BAM GENERATOR IHELU. THIS PENETRATION IS INCLUDED F'OR F'IGURE COMPLETENESS.
NONE OF' THE VALVES SHOWIII ARE CONSOEREO CONTAINMENT ISOLATION VAIL VES. NORMAL VALVE FLOW TYPE DIRECTION POWER PRIMARY SECONDARY MAXIMUM VALVE POSITION APPENDIX J VALVE ACTUATION ACTUATION CLOSURE REOUmNT SOUlCE OPERATOR SIGNAL SIGNAL TIME ISEC.l NORMAL SHUTDOWN FAIL PRIMARY SECONDAR AREA OF CHANGE f )E PENETRATION TESTED PER TYPE A APPEf\DEX J REQUIREMENTS.
-* s 088 TUBING s L r H:l DBB <:8 DBO V-008 +HV-18 TUBING DBB "'-F" 0 SEE P-10 DBB DBD CONTAINMENT PENETRATION NO. P-84 DESCRIPTION:
STEAM GENERATOR A SAMPLE LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTION<Sl 10.4.8 REV. 15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 60 OF 74 L.._ .. _________ , ___________________________________________________________________
_
V/llVE LNE/ INSID£1 VALVE OUTSIO£ NO. SIZE, IN. CONl. ASSOCIATED WITH A SAFETY FEATURES SYS, YES(!] NOD FLUID CONTAINED' WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* N/A APPLICABLE GDC NO. NONE
GENERAL COMMENT
S*
TH! CONTAINMEIIIT P!NETIATIONI CIATED WITH THB IITIAM GINIRATOIII Altl IIGT IUaECT TO GDOGJ'. stNCa THI CONTAI-T IIARRIIR IN"miAITY IIIIOT .REACHID.
TH* .OUHDARV DR IIARRIIR AOAINIT F .... N I'IIDDUCT LIAKMII TO Tt111 I!NYI-ENT 1* THI INIIDI DP TKE ITEAM OI!NEIIATDR TUBER AND TH1 OUT* IlK OF THI LIND I!MANATINO FADM THE STEAM GINIRATOR DIELLI. THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSIDERED CONTAINMENT ISOLATION VALVES. NORMAL Y/llVE FLOW D!RECMN TYPE POWER PRt.IARY SECONOARY t.IAXO.IJM VALVE POSITION APPENDIX J VlllVE ACTUAnON ACTUATION CLOSURE OP£RATOR SOURCE SIGNAL SIGNAl TlME ISECJ NORMAL SHUTDOWN FAIL PRIMARY SECONDAR AREA OF CHANGEr H PENETRATION TESTED PER TYPE A APPENDEX J REQUIREMENTS._
L r V-019 HV*20 TUBING+DBB . : IIL----1 DBD SEEP.11 s NUCLEAR H:L DBB DBD SAMPLING CONTAINMENT PENETRATION NO. P-85 DESCRIPTION*
STEAM GENERATOR B SAMPLE LINE STEAM GENERATOR BLOWOOWN SYSTEM REFERENCE SECTIONISl 10.4,8 REV. 15 WOLF CREEJt UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PE NET RAT IONS FIGURE 6.2.4-1 PAGE 61 OF 74 I _=-_j ViLVE LINE/ INSilE/ NO. VALVE OUTSIDE SIZE, IN. CONT. ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!] NOD FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* N/A APPLICABLE GDC NO. NONE
GENERAL COMMENT
S:
THI CONTAINIIII!NT PIINRTRATIONI A$10-CIA TliO WITH THI 8TIIAM GIN!IIATORI ARE NM TO CON,. IINOI THI CONTAINMBNT IIAAII ... INTaGIIITY
.. NOT BREACHED, 'IMI BDUNDARY DA U.AAIIR AGAINST F-DN PIUXJUI:T LEAKAGI! TO THE ENYIIIO,..EMT If THE IIIIIIOE OF THE sntAIII GENERATOR 1\1-AJID THii OUT* liD! OF THE LINEI EMANAnNG FIIOM THE ITEAM GENERATOR THIS PENETRATION IS INCLUDED FOR FIGURE COMPLETENESS.
NONE OF THE VALVES SHOWN ARE CONSOERED CONTAINMENT ISOLATION VALVES. NORMiL ViLVE ViLVE POWER FLOW TYPE OPERATOR SOUlCE DIRECTION DDB TUBING s *----., PRIMNlY SECONDNlY MAXIMUM V&VE POSITION J ACTUATION ACTUATION CLOSURE REQ EMENT SIGNAL SIGNAL TIME ISEC.l NORM& SHUTDOWN FAIL PRIMARY AREA OF CHANGE f
- PENETRATION TESTED PER TYPE A APPENDEX J REQUIREMENTS.
-L r DBB DBD D SEE P.12 s ":1-DBB DBD CONTAINMENT PENETRATION NO. P-86 DESCRIPTION:
STEAM GENERATOR C SAMPLE LINE STEAM GENERATOR SLOWDOWN SYSTEM REFERENCE SECTIONISl 10.4.8 REV. 15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 62 OF 74 -------------------------------------*----
I
1--------------------------------------------------
-*-------------------------------------------------------------------------------------------------------------------------
! { i ! l I I ! ! I V/>lVE LINE/ INSIDE/ NORMAL VALVE VALVE POWER PRIMARY SECONI>ARY MAXIMUM VALVE POSITION APPENDIX J VALVE OUTSIDE FLOW ACTUATION ACTUATION CLOSURE REQUIREMENT NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SMCE SIGNAL SIGNAL TIME !SECJ NORMAL SHUTDOWN FAIL PRIMARY I ' ! ! EMHV-8802A 4/4 OUTSIDE IN GATE MOTOR 1 NONE REM/t.1AN N/A CLOSED CLOSED AS IS CLOSED OPEN A i i I I EMHV-8881 r.,r. INSIDE our GLOBE AIR 1 CIS-A NONE 10 CLOSED CLOSED CLOSED CLOSED N/A A EMV-001 212 INSIDE IN CHECK N/A N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED OPEN A i I I I EMV-002 1212 INSIDE IN CHECK N/A N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED OPEN A I ! I EMV-052, V-053, lr4t¥4 N/A V-055, V-056, INSIDE N/A GLOBE MANUAL N/A N/A N/A N/A CLOSED CLOSED N/A CLOSED N/A V-184, V-185 i I I i ! i I ! I ! I ! i EMV-051 111 OUTSIDE EMV-186, V-187 n11 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES[!] NOD FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* N/A AIPPLICABLE .GDC NO. 55
GENERAL COMMENT
S*
THII PENETRATION
- o\IIOCIATED WITM THI HIGH PR-URI COOLANT INJECTION IY&T&M, WHICH II IIEQUIIIED TQ MITIIATI THE CIIIIIEOU&NUil OP A LOCA. A CHICK YALVI II PROIIIDID FDII EACH IIIANOH LINI INIIDE THE CONTAINMENT, AND A IIIMOTE-MANUAL ISOLATION VALVI II PIIOVIDED DUniDE THI CONTAirAIINT.
A liND Ll ACTIVI OR P-VII PAILUIII.
DAN II ACCOIIIIIDIJATI.D .NCB THE IVITIM
- OLDIIIII OU'IIIIDE THI CONTAINMENT AND II DIIIQNID AND CONITIIUCTID IIIRATII WITH THI DillON AND CONI.TJIIUC>
TION OP THI CONTAINIIIIINT.
UAICAGI DETECTION PIIDM THII LINI DUniDI '1111 CONTAINMSRT II PROVIDED, AI DII!ICIIIRO IN DOTIDN 1.3.1. LQC'AL TU'T1Niil Dl' THE VALVII OR THE CLDIED IYITIM 0\JmDE THII MENT II NOT I'I!IWIIIIID IINCI '1111IYITIM II OPIRATID IWD IN.ICTID DIIIIINll NOIIMAL PLANT DPIIIATION TO AIIUIII TH ... T '1111 INTIGRITV II BEING MAINTAIN*
- m. NOTE: ALL VENTS.DRAINS AIND FLOW POINTS AS INDICATED BELOW. ' N/A GLOBE MANUAL N.A N/A N/A N/A GLOBE MANUAL N.A N/A N/A JIIW8T < t--8802 \Q :* <H I D TCIID N/A N/A PI 1: CLOSED CLOSED N/A CLOSED N/A N/A CLOSED CLOSED N/A CLOSED N/A N/A s RCSHOTLEG 2 y ** ) > ! eca BCA ICB *I*ICA VoOIIO IICI HOT LEO
- 3 Y-002 CONTAINMENT PENETRATION NO. P-87 DESCRIPTION:
HOT LEG INJECTION HIGH PRESSURE COOLANT INJECTION SYSTEM REFERENCE SECTION!Sl 6:3 REV.13 WOLF Cl\EEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS
,.p FIGURE 6.2.4-1 ! PAGE 63 OF 74 , 1! BCD i ------------------------------------------------------------------------------------------------------------*----*--------------------J
---.. ----.-----*----.-----------"--------------------.----------------------.----------.-------------
--*-----* -----------
- ----------.--------------
.. -----.----
*-. --------------------------.--------------.------
I t ' I . I I I I I ! ' , I I I VILVE LINE/ INSIDE/ NORMIL VILVE VN...VE POWER PRIMARY SECONDAA'!'
MAXIMUM VN..VE POSITION APPENDIX J VN...VE OUTSIDE FLOW ACTUATION ACTUATION CLOSURE I i I ' NO. SIZE, IN. CONT. DIRECTION TYPE OPERATOR SOURCE SIGNIL SIGNIL TIME <SEC.) NORMAL SHUTDOWN FAIL PRIMARY SECONDAR'I' REQUIREMENT I I EMHV-8801A 4/4 OUTSIDE IN GATE MOTOR 1 SIS NONE N/A CLOSED CLOSED AS IS OPEN N/A A EMHV-8801B 4/4 OUTSIDE IN GATE MOTOR 4 SIS NONE N/A CLOSED CLOSED AS IS OPEN N/A A 1 I i l-' i I I i ! i I ! I I I I 1 EMV-077 111 OUTSIDE EMHV-8843 f4/f4 INSIDE EMV-8815 INSIDE EMV-151 11 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YES I!] NOD FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE* N/A APPLICABLE GDC NO. 55
GENERAL COMMENT
S*
THII PINITIIATIDN II -CIATID WITH THI HIGH CCICIUINT INJICTIOII IYITI,._ WltiCH II RIIWIIIID THE CO-QUIIICII DF A LDCA, A OHf.CK VALVII II RlR IACH BRANCH LIN& INIIDI THI CIDNTAINMIIIT, AIID A . IIEIIDTI.foiAIIUAL IIULATIOII VALVE II PIICIVIDID Rlll IACH IIIANCH LINI DUTIIDE THI CDNTAIIIMI!NT.
A IINDLI AIITIV! 011 PAIIIVI FAII.URI CAN IE .AQDDIWDDATID IIIICII THI IYITIM B CLCIIED CUTIIDI THI CONTAIIIIM&NT AND . II DQIIINID AND CDNITIIUCTIID CCMMEN* IUAATE WITH 1HII:IIIICIN AND CDNITAIJO.
TION QP Till CICIITAIMINT.
LIAICAGI DITf.CTIQN PIIDM THII Lilli OUTIIDS TH& CONTAINMENT II PIIOYIDEII, M DIICRIIEII IN .IITIDN U.J, LOCAL TilTING QP TME YALVII DR THE CLOIII!D IYITir. CUTIIDE THI CDNTAI ... MINT 18 NOT REGUIIIIIIIIIICE THI SYaTIM II CII'IIRATED AND 1_.,-10 DUIIIIIO NDIIIWAL PLANT DPIIIATION TD AIIUIIII THAT THI IN'IIiiCIAITV B DIIICI MAINTAIN*
ID, N/A GLOBE MANUAL IN GLOBE AIR IN CHECK N/A OUT GLOBE MANUAL IOIIOfll INJECTION TANK N/A N/A N/A N/A CLOSED CLOSED 4 CIS-A NONE 10 CLOSED CLOSED N/A N/A N/A N/A CLOSED CLOSED N/A N/A N/A N/A CLOSED CLOSED BCD .. --BC8 .
BCD TelcO BCB N/A CLOSED N/A N/A CLOSED N/A N/A CLOSED N/A A OPEN N/A A CLOSED N/A N/A RCS L ) ..,. CONTAINMENT PENETRATION NO. P-88 DESCRIPTION*
BORON INJECTION TO COLO LEGS HIGH PRESSURE COOLANT INJECTION SYSTEM REFERENCE SECTION(Sl 6.3 REV. 1J WOLF CREEK I i I UPDATED SAFETY ANALYSIS REPORT ! ' CONTAINMENT PENETRATIONS
! FIGURE 6.2.4-1 j 1 PAGE 64 OF 74 1 I ! i I I ---------------------------------------------------------------------------------------------------------------------------------------J VN..VE LINE/ INSIDE/ VALVE OUTSIDE NO. SIZE, IN. CONT. ENHV-06 10/10 OUTSIDE ENV-076 111 OUTSIDE ENV-01.3 10/10 INSIDE ASSOCIATED WITH A SAFETY FEATURES SYS.
NOD FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VI>J...VE:
N/A APPLICABLE GDC NO. 56
GENERAL COMMENT
S*
TH. PliNiiTRATICN II AIIOCIATED WITM THE CONTAINMENT IPRAY IYITEM, WHICH II RIOUII!.ED TO MITIGATE THE CONIE* QUENCH 011' A LOCA. A CHECK VALVE II PROVIDID INIID! THI! CONTAINMiiNT, AND A REMOTE-MANUAL IIOLATION VALYii II PROVIDED OliTIIDE THE CONTAINMENT.
A IINOLii ACTIVE OR PAIIIYI! fAILURE CAN IE ACCOMMODAnD
... CE THI! IVITEM II CLDIID OU'RIDE THE CONTAINMENT AND II DEIIGNI!D AND CONSTRUCTED MENIURATI WI'TH THI DillON AND CON* STRUCTIDN Of THE CONTAINMENT, LIAI(. AG& DITiiCTIDN PROM THII LINI DUTIID& THI CONTAINMENT II PROVIDED, AI Oil* CRIIIiD IN II!CTIDN U.l. LOCAL TilTING OF THE VALVES DR THI CLCIIID IY8T1SM OUTSIDE THE CONTAIN* MINT II NOT REQUIRiiO IlNCE THE IVITI!M II DPIRATI!D AND IN.I!CTED DURING NORMAL PLANT DFIIIATIDN TO AllURE THAT THI! INTEGRITY II lElNO MAINTAIN*
ED. NORMAL VM..VE FLOW VN....VE DIRECTION TYPE OPERATOR IN GATE MOTOR N/A GLOBE MANUAL IN CHECK N/A CONT .... AV PUMP A POWER PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SOURCE SIGNAL SIGNAL TIME CSECJ NORM.AL 1 CSAS REM/MAN N/A CLOSED N/A N/A N/A N/A CLOSED N/A N/A N/A N/A CLOSED ** _j V-013 HCB
"' TC HCD VN..VE POSITION NJPENDIX J SHUTDOWN CLOSED CLOSED CLOSED FAJL PRIMARY ECONDARY REQUIREMENT AS IS OPEN CLOSED A N/A CLOSED N/A N/A N/A OPEN CLOSED A CONT ** RAY NOZZLES CONTAINMENT PENETRATION NO. P-89 DESCRIPTION:
CONTAINMENT SPRAY CONTAINMENT SPRAY SYSTEM REFERENCE SECTIONCSl 6.2.2 REV. 1.3 WOLF CR.EEK UPDATED SAFETY ANALYSIS R.EPOR.T CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 65 OF 74 -, ' I
--
---
--___ j
1 VH.VE LINE/ INSilE/ NORMH. NO. VH.VE OUTSIDE FLOW SIZE, IN. CONT. DIRECTION EMHV-8964 YciY.. OUTSIDE OUT EMV*15J ¥ctY.. II\ISIDE N/A EMHV-8871 YctYc INSIDE OUT EMV-038 Y4tY.. OUTSIDE N/A EMPI-929 Yc OUTSIDE N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO(!] FLUID CONTAINED*
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 16.7 ft APPLICABLE CDC NO. 56
GENERAL COMMENT
S:
NONE V&VE TYPE GLOBE GLOBE GLOBE GLOBE N/A VH.VE OPERATOR AIR MANUAL AIR MANU/>J...
N/A v POWER SOlllCE 1 N/A 4 N/A N/A 1'1 -"' .., 0 > PRIMARY SECONDARY ACTUATION ACTUATION SIGN& SIGNH. CIS-A NONE N/A N/A CIS-A NONE N/A N/A N/A N/A B+--I ro BCD MAXIMUM CLOSURE TIME CSECJ NORM/>J...
10 CLOSED N/A CLOSED 10 CLOSED N/A CLOSED N/A N/A BCB BCO V/>J...VE POSITION APPENDIX J SHUTDOWN CLOSED CLOSED CLOSED CLOSED N/A FAIL PRIMARY REQUREMENT CLOSED CLOSED N/A c N/A CLOSED N/A N/A CLOSED CLOSED N/A c N/A CLOSED N/A N/A N/A N/A N/A 7 ' I ) L AREA OF CHANGE !CCI TESI'LINU CONTAINMENT PENETRATION NO. P-92 DESCRIPTION:
ECCS TEST LINE RETURN HIGH PRESSURE COOLANT INJECTION SYSTEM REFERENCE SECTIONCSl 6.3 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT UPDATED SAFETY ANALYSIS REPORT WOLF CREEK CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 VALVE POSITION TIME (SEC.)
CLOSURE MAXIMUM SIGNAL ACTUATION SECONDARY SIGNAL ACTUATION PRIMARY SOURCE POWER OPERATOR VALVE TYPE VALVE DIRECTION FLOW NORMAL CONT.OUTSIDE INSIDE/SIZE, IN.VALVE LINE/NO.VALVE 1/1 1/1 1/1 1/1
INSIDE OUTSIDE OUTSIDE OUTSIDE
OUT OUT N/A OUT
GLOBE GLOBE GLOBE GLOBE
MANUAL
4 1 N/A 4
CIS-A CIS-A N/A CIS-A
NONE NONE N/A NONE
5 5 N/A 5
OPEN OPEN CLOSED CLOSED
CLOSED CLOSED CLOSED CLOSED
CLOSED CLOSED N/A CLOSED
CLOSED CLOSED CLOSED CLOSED
N/A N/A N/A N/A
NORMAL SHUTDOWN FAIL PRIMARY SECONDARY FEATURES SYS. YES NO ASSOCIATED WITH A SAFETY ISOLATION VALVE: 10.3 ft LENGTH OF PIPING TO OUTERMOST GDC NO. 55 APPLICABLE FLUID CONTAINED: WATER X SJHV-5 SJHV-6 SJV-069 SJHV-127
REQUIREMENT APPENDIX J C C N/A C
GENERAL COMMENT
S:
PAGE 67 OF 74 V-069 SOLENOID SOLENOID SOLENOID BCB BCD HOT LEG SAMPLE LOOP 1 29 SAMPLE SYSTEM 29 REV. 29 REFERENCE SECTION(S) 9.3.2/18.2.3
RC LOOP LIQUID SAMPLES DESCRIPTION:
CONTAINMENT PENETRATION NO. P-93 i I I i ! i I ! i i I !I I II ,, II '* ii I' ! I ! i !I I c I I I ; I I i i I I i I ! I ! VN..VE LINE/ INSIDE/ V/llVE OUTSIDE NO. SIZE, IN. CONT. SJHV-18 111 INSIDE SJHV-19 /1 OUTSIDE SJV-D66 111 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO [!J FLUID CONTAINED:
WATER LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: APPLICABLE GDC NO. GENER/>L COMMENTS*
NDNE 10.3 ft 55 NORMN.. V/llVE V/llVE POWER FLOW DIRECTION TYPE OPERATOR OUT GATE SOLENOID 4 OUT GATE SOLENOID 1 NIA GLOBE MANU AIL NIA PRIMARY SECONDARY MAXIMUM ACTUATION ACTUATION CLOSURE SIGNN.. SIGN/ll TIME <SEC.) NORMAL CIS-A NONE 5 CLOSED CIS-A NONE 5 CLOSED N/A N/A N/A CLOSED HV 18 L ECB ECD TC V I'L VE POSITION APPENDIX J SHUTDOWN CLOSED CLOSED CLOSED HV 18 FAIL CLOSED CLOSED N/A PRIMARY REQUIREMENT CLOSED N/A c CLOSED N/A c CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-95 DESCRIPTION:
ACCUMULATOR SAMPLING NUCLEAR SAMPLING SYSTEM REFERENCE SECTION<Sl 9.3.2 REV.13 WOLF CREEK. UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 68 OF 74 I i ! i
- ---*-*------*---*---------*----*---*---*---*-*--*--_J UPDATED SAFETY ANALYSIS REPORT WOLF CREEK CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 VALVE POSITION TIME (SEC.)
CLOSURE MAXIMUM SIGNAL ACTUATION SECONDARY SIGNAL ACTUATION PRIMARY SOURCE POWER OPERATOR VALVE TYPE VALVE DIRECTION FLOW NORMAL CONT.OUTSIDE INSIDE/SIZE, IN.VALVE LINE/NO.VALVE 1/1 1/1 1/1
INSIDE OUTSIDE OUTSIDE
IN IN N/A
GATE GATE GLOBE
MANUAL
1 1 N/A
CIS-A CIS-A N/A
REM/MAN REM/MAN N/A
5 5 N/A
CLOSED CLOSED CLOSED
CLOSED CLOSED CLOSED
CLOSED CLOSED N/A
CLOSED CLOSED CLOSED
OPEN OPEN N/A
NORMAL SHUTDOWN FAIL PRIMARY SECONDARY FEATURES SYS. YES NO ASSOCIATED WITH A SAFETY ISOLATION VALVE: N/A LENGTH OF PIPING TO OUTERMOST GDC NO. 56 APPLICABLE FLUID CONTAINED: CONT. AIR X GSHV-18 GSHV-17 GSV-036
REQUIREMENT APPENDIX J A,C A,C N/A
GENERAL COMMENT
S:
PAGE 69 OF 74 SOLENOID SOLENOID 2 THE HYDRO-GEN ANALYZER HCB HCD HCB HCD TC V-035 29 REV. 29 REFERENCE SECTION(S) 6.2.5
HYDROGEN CONTROL SYSTEM H SAMPLE RETURN DESCRIPTION:
CONTAINMENT PENETRATION NO. P-97 29 VM.VE LINE/ INSilE/ NORh.IM. VM.VE VM.VE POWER NO. VALVE OUTSIDE FLOW TYPE OPERATOR SOI.IlCE SIZE, IN. CONT. DIRECTION GSHV-JJ 1/1 OUTSIDE IN GATE SOLENOID 4 GSHV*J4 1/1 INSIDE IN GATE SOLENOID , GSV-052 111 OUTSIDE N/A GLOBE MANUAL N/A ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO(!) FLUID CONTAINED*
CONT. ATM LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 7.B n. APPLICABLE GDC NO. 56 ....,
GENERAL COMMENT
S-NONE TO CONTAINMENT ATMOSPH.ERE PRIMARY SECONDARY h.IAXIMUh.l ACTUATION ACTUATION CLOSURE SIGNAL SIGNAL TIME ISECJ NORMAL CIS-A REM/MAN 5 OPEN CIS-A REM/MAN 5 OPEN N/A N/A N/A CLOSED L 3 ...... r HCB t A > HCD TC AREA OF CHANGE-VALVE POSITION APPENDIX J SHUTDOWN FAIL OPEN CLOSED OPEN CLOSED CLOSED N/A CONTAINMENT A1111108PHERE MONITOR GT-RE-32 / I PRIMARY SECONDAR' REQUilEMENT CLOSED OPEN ) A,C t CLOSED OPEN l A,C_) CLOSED N/A N/A CONTAINMENT PENETRATION NO. P-97 DESCRIPTION:
SAMPLE RETURN CONTAINMENT ATMOSPHERE MONITOR REFERENCE SECTION<Sl 9.4.6 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 69a OF 74 VALVE LINE/ INSIDE/ V.ALVE OUTS ICE NO. SIZE, IN. CONT. KBV-001 212 INSIDE KBV-002 2/2 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO[!) FLUID CONTAINED:
AIR LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 7.1 ft. APPLICABLE CDC NO. 56
GENERAL COMMENT
S:
NONI: NORf.l.AL VALVE FLOW DIRECTION TYPE IN GLOBE IN GLOBE BREATHING V.ALVE POWER OPERATOR SOURCE MANUAL N/A MANUAL N/A HCD IICI L.:.a TOI.D tp PRIMARY SECOOARY MAXIMUM ACTUATION ACTUATION CLOSI.fiE SIGNAL SIGNAL TIME CSEC.l NORMAL N/A N/A N/A CLOSED N/A N/A N/A CLOSED MO. MGD TC loY TIOtiAL POLLOWING A LOCA OR ITEAM BIIIAIC, THEil LINQ IENU THI PAEIIURI OF CONTAINMINT ATMO.MIRI 011 THE INIIDE AND Alii CDNNICTID TO PIIEBUIIE TRAN ... I1'TIIIII ON TNI Cll.ITSIDI.
IIGNALI FROM THEil! TRANIMITTERI CAN INITIATE IAFIT't IN.IECTION AND COifTAINMIIfT Jill). LATION ON HlCIH OONTAINM!NT Pll!auRI.
'IHIY ALIO, UPON CONTAINMINT PIIQIUIII, PRODUCE THE ONLY IIGNAL TO INITIATE CONTAINMENT
.. RAY, IN YIIW OP THB PUNOTION, IT II lllftNTIAL THAT THI! LINI IIIMAIN CII'IN AI'D NOT II IIOLATID FOLLOWING All AI:CIDENT.
IAIID ON THJI RIOUIIIEM&NT, A IIALED I!NIING Lilli. AI DUCIIIIBD IELOW. II UUO. PT.-AND "-Alii THI MDII RANQI CONTAINMENT PIIRIURI llEGUIREO BY NURI,_D717 AND TORY GUIDa 1.17. &AOH OF THE FOUR QMANNILI HAl A IIPAIIATI PliNITRATION, AND !ACH PRUIURI TAANIMm&ll II I.OCATI!D IMMBDIATI!LY Ao.IACiiNT TO THI OU .... IIII OP T1t11 OONTAINMENT If II CON* NiOTID TO A IEAUO B!I.LOWI.
LOCATIO IMMBIIATILY AD.IACINT TO THI INIIDI CONTAINMENT III'LL. IY MI!ANI OF A IIALI!D PLUIO FILLID TUII. THII TUBING, ALONO WITH TNI TR,_ITTIR AND B OONIEIIYAnYELY DIIIQNBO I'ND IUamOT TO ITIIICT QUI'LITY CON* TROL AND TO IIEGULAR IN .. IIIYICI IN.I.CTIONI TO AIIUII! 1111 INftOIIIT'I'.
THII AIIIIANGIEMENT PROYIOU A DOUBLI BARRIIII tONI INIIDE AND ONE OUTIIDEI IITWIIN TNI! CONTAINMINT AND TMI OUTBDI &TMOa'NiiRI.
IHOULO A I.EAI< OCCUR OUTIIDE THE CONTAINMENT, THE IIALI!D BEL.LOWII INIIIDB TN& OONTAIN* MINT, WHICH II DEIIGNIO TO WI11GTAND FULL CONTAINMIIfT DUION PIIIBIIII, WILL PIII!YINT TN! IICAPE OP THI CION* T&INMENT ATMO.HIIII.
IIIOULD A LIAIC OCCUR 111110& THE CONTAINMINT, THI DIAPifiiAGM IN THI TRANIMITTIII, WHICH II DIIIGNED TO WITHSTAND FULL TAINMENT OUIGIN PRIIIIIRI., IIILL PRE* YEIIT ANY i.:API PilON THI MINT. THB AIIIIANGINENT PIIOYIDEI AUTOMATIC IIOUILI..,...IIIIII IIOI.ATION WITHOUT DPERATOII ACmON AND WITHOUT IACRIFICINO ANY RILIAIILITY.
lOTH THII IlL LOlli AND TUBING I NIIDI THE TAINMENT ARE ENCLOIID IY PROTECTIVE IHII!LDINQ, THII IHIIL.DINGI (BOX, NIL 011 GUARD PIPit, I.TC.) PA!VENT MICHANICAL DAMMII TO THI ENTI PROM MIIIILII, WATIR IITII, DIIOP. PIO TCOLI,I!TO.
1!-2&8 L P-1 PT*934 .. r CONTAINMENT PENETRATION NO. P-103 8c 104 DESCRIPTION:
E-256 CONTAINMENT PRESSURE TRANSMITTERS REFERENCE SECTIONCS) 6.3, 9.4 REV.11 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONTAINMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 72 OF 74 BICAUIE OP TH. BEALED FLUID ,.LL!D IVITEM, A I'CIBTUI.ATiiD IEYIRANCI OP THI LINE DURING IITHIII NOIIMAL III'ER* !
- L _____________________________
VALVE LINE/ INSilE/ NO. VALVE OUTSIDE SIZE,IN. CONT. GTHZ-9 36/30 OUTSIDE GTHZ-8 36/36 INSIDE GTHZ-12 18/18 OUTSIDE CTHZ-11 18/18 INSIDE GTV0223 1/1 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESD NO[!) FLUID CONTAINED:
CONT. ATM. LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 12.0 ft APPLICABLE GDC NO. 56
GENERAL COMMENT
S*
NOTE1 THI&YALVIIBINTERMITT&NTI.Y OPENI:D TO PROVIDII POR CONTAINMENTIIIINI.PYRQB DURINQ POWER OPERATION I .&..-*--------*-*-*---------------
NORMAL VALVE VALVE POWER PRIMARY SECONDARY MAXI MUlA FLOW TYPE OPERATOR SMCE ACTUATION ACTUATION CLOSURE DIRECTION SIGNAL SIGNAL TIME ISECJ NORMAL OUT BUTTERFLY AIR/SPR 1 CPIS NONE 10 CLOSED OUT BUTTERFLY AIR/SPR 4 CPIS NONE 10 CLOSED OUT BUTTERFLY AIR/SPR 1 CPIS NONE J NOTE 1 OUT BUTTERFLY AIR/SPR 4 CPIS NONE J NOTE 1 N/A GATE MANUAL N/A N/A N/A N/A CLOSED AREA OF TC '"'
"' pp ... .....,. MZI Vl "' HZI pp ... ... 0< '4F ' I C0HTA1NMINT J-ot ,; --, NBO I'URIU 4-SNP ...... ; Jt_ Z11 ) _, .. "'T ;:!;,,
0 pp ...........
pp --------------*--*--*-*--------------., VM..VE POSITION N'PENDIX J SHUTDOWN FAIL PRIMARY REQUIREMENT OPEN CLOSED CLOSED N/A c OPEN CLOSED CLOSED N/A c CLOSED CLOSED CLOSED N/A c CLOSED CLOSED CLOSED N/A c CLOSED N/A CLOSED N/A N/A f. J I CONTAINMENT ATMOSPHERE CONTAINMENT PENETRATION NO. V-160 DESCRIPTION:
CONTAINMENT PURGE SYSTEM REFERENCE SECTIONCSI 9.4 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONT AJNMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 73 OF 74 VILVE UNE/ INSilE/ NO. VILVE OUTSIDE SIZE, IN. CONT. GTHZ-7 36/36 INSIDE GTHZ-5 18/18 INSIDE GTHZ-4-18/18 OUTSIDE GTHZ-6 36136 OUTSIDE GTV0222 1/1 OUTSIDE ASSOCIATED WITH A SAFETY FEATURES SYS. YESO NO [!I FLUID CONTAINED:
AIR. LENGTH OF PIPING TO OUTERMOST ISOLATION VALVE: 12.5 ft AIPPLICABLE GDC NO. 56
GENERAL COMMENT
S:
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pp t.4AXIMUM ClOSURE TIME ISEC.l NORMAL 10 CLOSED J NOTE 1 J NOTE 1 10 CLOSED N/A CLOSED HZ8 pp .t-...._1 I VI>J...VE POSITION N'PENOIX J SHUTDOWN OPEN CLOSED CLOSED OPEN CLOSED l FAIL PRIMARY REQUREMENT CLOSED CLOSED N/A c CLOSED CLOSED N/A c CLOSED CLOSED N/A c CLOSED CLOSED N/A c N/A CLOSED N/A N/A CONTAINMENT MINI-PURGE SUPPLY CONTAINMENT SHUTDOWN PURGE SUPPLY CONTAINMENT PENETRATION NO. V-161 DESCRIPTION:
CONTAINMENT PURGE SYSTEM REFERENCE SECTIONISl 9.4 REV.15 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT CONT AJNMENT PENETRATIONS FIGURE 6.2.4-1 PAGE 74 OF 74 WOLF CREEK I I I WATER LEVEL I I I I I I* I ----BARRIER I I I I I I I I I I I I l Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.2.4-2 STEAM GENERATOR AND ASSOCIATED SYSTEMS AS A BARRIER TO THE REL-EASE OF RADIOACTIVITY POST LOCA
- l 2.5 :z: 0 r.l () :z: 2 8 r.l :z:
- r.l 0 1..5 * -
-**. ---------, WOLFCREEK TIME POST LOCA (DAYS) Rev. 8 WOLF UPDATED SAFETY ANALYSIS REPORT Figure 6.2.5-2 HYDROGEN VOLUME CONCENTRATION IN CONTAINMENT WITH ONE RECOMBINER OPERATING AT ONE DAY
- WOLFCREEK
-100 I I I I I ::::> 0 ::c ...... rn 0 10 -::s z 0 I> rz:l z 1 -rz:l 0 z 0 -0::
Aluminum.
xmc Rev. 8 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT Figure 6.2.5-4 HYDROGEN ACCUMULATION IN CONTAINMENT
""' X --z Q 1-< a:: 1-z w u z 0 (.) w :::E ::> --' 0 > z w (.) 0 a:: 0 r I WOLF CHEEK 8 6 4 2 0 0 19 20 30 40 so 60 TIME POST LOCA .<DAYS) Rev. 8 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 6.2.5-5 HYDROGEN VOLUME CONCENTRATION IN CONTAINMENT ASSUMING NO PREVENTIVE ACTION TAKEN
...... X z Q 1-<(. a:: 1-z w 2.5 u z 0 u w ::2 :::> _J 2 0 > z w *e,:) 0 .a:: 0 :t.S :>-J: WOLF CREEK tiME POST LOCA {DAYS> Rev. 8 WOLF CREEK UPDATED $Ai'"ETY ANAL YSI$ REPORT Figure 6.2.5-6 HYDROGEN VOLUME CONCENTRATION IN CONTAINMENT WITH PURGING AFTER 4 DAYS
.0.011'35 a:: :r: I 1-1 "10 -2 !-i-a (/) ....... (/) w -' 0 I 1"10-3 CD -' w 1-< a:: z 0 . -4 Vi 0 1"10 a:: a:: 8 WOLF CREEK TIME POST LOCA (SECONDS>
WOlf CREEK 1 1'10 Rev .. 8 UPDATED SAFElY ANALYSIS.
RtPORT Figure 6.2.5-7 ALUMINUM CORROSION RATES USED IN THE HYDROGEN GENERATION CALCULATION
..... a:: .::c I t-lL. 0 (/) ...... (/) w ...J 0 I CD ...J ..... t..a..i t-<( a:: z 0 .(/) 0 a:: a:: .0 u: WOLF CREEK 1"10-s 1"10 -6 TIME POST LOCA <SECONDS>
Rev. 8 WOLF CREEK UPDATED SAFETY ANAL Y$15 REPORT Figure 6.2.5-8 ZINC CORROSION RATES USED IN THE HYDROGEN GENERATION CALCULATION 2SO g 6::1 200 ff ::e 150 WOLFCREEK TIME POST LOCA (SECONDS)
Rev. 8 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT-Figure 6.2.5-9 TEMPERATURE PROFILE USED TO ADJUST CORROSION RATES FOR THE HYDROGEN GENERATION CALCULATION I I ** * *-
WOLF CREEK 6.3 EMERGENCY CORE COOLING SYSTEM The emergency core cooling system (ECCS) is designed to cool the reactor core
and provide shutdown capability following initiation of the following accident
conditions:
- a. Loss-of-coolant accident (LOCA), including a pipe break or a spurious relief or safety valve opening in the reactor coolant system (RCS) which would result in a discharge larger than that which could be made up by the normal makeup system.
- b. Rupture of a control rod drive mechanism, causing a rod cluster control assembly ejection accident.
- d. A steam generator tube failure.
The primary function of the ECCS is to provide emergency core cooling (ECC) in
the event of a LOCA resulting from a break in the primary reactor coolant
system (RCS) or to provide emergency boration in the event of a steam/or
feedwater break accident.
6.3.1 DESIGN BASES
6.3.1.1 Safety Design Basis The ECCS is safety-related and is required to function following a DBA to
achieve and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - Except for the refueling water storage tank (RWST),
the ECCS is protected from the effects of natural phenomena, such as
earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). The
RWST was designed to seismic Category I requirements only.
SAFETY DESIGN BASIS TWO - The ECCS was designed to remain functional after an
SSE and to perform its intended function following the postulated hazards of
fire, internal missiles, or pipe break (GDC-3 and 4).
6.3-1 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a
single active component failure coincident with the loss of offsite power (GDC-
35).
SAFETY DESIGN BASIS FOUR - The active components are capable of being tested
during plant operation. Provisions are made to allow for inservice inspection
of components at appropriate times specified in the ASME Boiler and Pressure
Vessel Code, Section XI (GDC-36 and 37).
SAFETY DESIGN BASIS FIVE - The ECCS was designed and fabricated to codes
consistent with the quality group classification assigned by Regulatory Guide
1.26 and the seismic category assigned by Regulatory Guide 1.29. The power
supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The capability to isolate components or piping was provided so that the ECCS safety function is not compromised. This includes
isolation of components to deal with leakage or malfunctions and to isolate
safety-related portions of the system (GDC-35).
SAFETY DESIGN BASIS SEVEN - The containment isolation valves in the system were
selected, tested, and located in accordance with the requirements of GDC-54 and
55 and 10 CFR 50, Appendix J, Type A testing.
SAFETY DESIGN BASIS EIGHT - ECCS equipment design qualifications ensures
acceptable performance for all environments anticipated under normal, testing, and design basis accident conditions.
SAFETY DESIGN BASIS NINE - The functional requirements of the ECCS are derived
from Appendix K limits for fuel cladding temperature, etc., following any of the above accidents, as delineated in 10 CFR 50.46. The subsystem functional parameters are integrated so that the Appendix K requirements are met over the
range of anticipated accidents and single failure assumptions.
6.3.1.2 Power Generation Design Basis There are no power generation design bases for the ECCS function. Portions of
the ECCS are also portions of the residual heat removal system (RHRS) and
chemical and volume control system (CVCS) and are used during normal power operation. Power generation design bases for these portions of the ECCS are
discussed in Sections 5.4.7 and 9.3.4, respectively.
6.3-2 Rev. 0 WOLF CREEK 6.3.2 SYSTEM DESCRIPTION
6.3.2.1 General Description
The ECCS components are designed so that a minimum of three accumulators, one
centrifugal charging pump, one safety injection pump, and one residual heat removal pump, together with their associated valves and piping, ensure adequate
core cooling in the event of a design basis LOCA or provide boration in the
event of a steam/or feedwater break accident. The redundant onsite emergency
diesels assure adequate emergency power to at least one train of electrically
operated components in the event that a loss of offsite power occurs simultaneously with a LOCA.
The P&IDs for the ECCS are shown in Figures 5.4-7, 6.3-1 and 9.3-8. ECCS flow
diagrams are shown in Figure 6.3-2. Pertinent design and operating parameters
for the components of the ECCS are given in Table 6.3-1. The design parameters
shown represent the values specified in procurement specifications. Operating
parameters are typical for WCGS, however, minor variations in performance
characteristics exist between individual components. The accident analyses
contain adequate margin to account for these individual component variations.
The component interlocks used in the different modes of system operation are listed below.
- a. The SIS initiates the following actions:
- 1. Emergency diesel generators start
- 2. Centrifugal charging pumps start
- 3. RWST suction valves to charging pumps open
- 4. Boron injection tank suction and discharge parallel isolation valves open
- 5. Normal charging path valves close
- 6. Safety injection pumps start
6.3-3 Rev. 15 WOLF CREEK
- 7. Residual heat removal pumps start
- 8. Volume control tank outlet isolation valves close
- 9. RWST discharge isolation valves to the fuel storage pool cooling and cleanup system close
- b. Switchover from injection mode to recirculation involves the following interlocks:
- 2. The safety injection pump and charging pump recirculation suction isolation valves, EJ-HV-8804A and B, can be opened provided that either the safety injection system miniflow isolation valve, BN-HV-8813, or both safety injection pump miniflow isolation valves, EM-HV-8814A and B, are closed.
Additionally, one of the two RHR hot leg suction valves on Loop 1, BB-PV-8702A and EJ-HV-8701A, and on Loop 4, BB-PV-8702B and EJ-HV-8701B, must be closed.
6.3.2.2 Equipment and Component Descriptions Codes and standards applicable to the ECCS are listed in Tables 3.2-1 and 6.3-
- 1.
The component design and operating conditions are specified as the most severe
conditions to which each respective component is exposed, during either normal
plant operation or operation of the ECCS. For each component, these conditions
are considered in relation to the code to which it is designed. By designing
the components in accordance with applicable codes, and with due consideration
for the design and operating conditions, the fundamental assurance of
structural integrity and operability of the ECCS components was maintained.
Components of the ECCS are designed to withstand the appropriate seismic
loadings, in accordance with their safety class as given in Table 3.2-1. It is important that the ECCS is sufficiently filled with water to ensure that the subsystems can reliably perform their intended function under all LOCA and non-LOCA conditions that require makeup to the RCS. Maintaining the piping from the ECCS pumps to the RCS sufficiently full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand and water hammers in the injection line are precluded.
6.3-4 Rev. 22 WOLF CREEK The elevated temperature of the sump solution during recirculation is well
within the design temperature of all ECCS components. In addition, consideration has been given to the potential for corrosion of various types of
metals exposed to the fluid conditions prevalent immediately after the accident or during long-term recirculation operations.
The following is a discussion of the major components of the ECCS:
Accumulators The accumulators are pressure vessels partially filled with borated water and
pressurized with nitrogen gas. During normal operation, each accumulator is
isolated from the RCS by two check valves in series. Should the RCS pressure fall below the accumulator pressure, the check valves open and borated water is
forced into the RCS. One accumulator is attached to each of the cold legs of
the RCS. Mechanical operation of the swing-disc check valves is the only
action required to open the injection path from the accumulators to the core
via the cold leg.
Connections are provided for remotely adjusting the level and boron
concentration of the borated water in each accumulator during normal plant
operation, as required. Accumulator water level may be adjusted either by
draining to the recycle holdup tank or by pumping borated water from the RWST to the accumulator. Samples of the solution in the accumulators are taken periodically for checks of boron concentration.
Accumulator pressure is provided by a supply of nitrogen gas, and can be
adjusted, as required, during normal plant operation. However, the
accumulators are normally isolated from this nitrogen supply. Gas relief
valves on the accumulators protect them from pressures in excess of design
pressure. Accumulator gas pressure is monitored by indicators and alarms.
Solenoid-operated vent valves are provided to depressurize the accumulators
during emergencies, if needed.
The accumulators are located within the containment but outside of the
secondary shield wall which protects the tanks from missiles generated from a
postulated LOCA.
Refueling Water Storage Tank The borated refueling water storage facility consists of a large outside
storage tank (i.e., RWST) with connections for borated demineralized water
delivery to and receipt from the fuel pool cooling and cleanup system, the chemical and volume control system, the containment spray system, and the ECCS.
6.3-5 Rev. 0 WOLF CREEK The RWST is a passive seismic Category I component and is required only during
the short term following a LOCA, MSLB, or any other accident requiring ECCS.
Therefore, neither redundancy nor tornado missile protection is required. The
safety-related level instrumentation and the temperature monitoring instrumentation associated with the RWST are designed with redundancy.
The RWST is vented directly to the atmosphere. Tank overflow is directed to
the waste holdup tank in the liquid radwaste system via the floor and equipment
drain system. Sample connections are also provided to allow periodic analysis
of the RWST contents.
Prior to accepting makeup water from the CVCS, the current RWST sample results
must be used to ensure the proper final boron concentration in the tank.
Samples are taken periodically from the RWST for analysis to assure that the quality of the contents meets the water chemistry specifications given in Table
9.2-16. If the tank contents require purification, they are circulated through
the fuel pool cooling and cleanup system. To maintain the boron concentration
within specification, a strong boric acid solution (4 wt percent) or reactor
makeup water can be added via the chemical and volume control system.
An automatic heater system is provided to prevent the contents of the RWST from
freezing. The heater system consists of steam coils wrapped around the outside
of the RWST, insulation on the RWST, electrical heat tracing on the exposed
nonessential piping, and a heated enclosure for the essential piping, valves, and instrumentation. These steam coils are serviced by the auxiliary steam
system. For freeze protection during colder periods of the year, the RWST is automatically maintained above a nominal 50 F temperature. A temperature control valve is provided to control steam flow to the steam coil heaters. A continuous steam flow is maintained to the heating coils during plant
winterization via the temperature control valve bypass line. This ensures that
the condensate return lines will not freeze. Provisions are in place to put
the RWST in continuous recirculation through its return line when necessary to
assure minimum flow capability of the SI pump recirculation line. Redundant
temperature instrumentation is provided to inform the operator of any
degradation of the heating capability for the RWST.
Since the RWST is not normally used as a source of water during power
operation, the tank level is administratively maintained. The water level is
maintained above the minimum level (94%) consistent with the requirements for
injection, transfer allowances, and instrument error allowances. A tank level
above 94% assures that the RWST contains a volume greater than 394,000 gallons.
The RWST levels and volumes shown on Figure 6.3-7 are based on using the most
conservative instrument uncertainty associated with each RWST level setpoint.
For example, if the use of indication error results in a more conservative
calculated volume, it is assumed that the operator will use indication instead
of alarm for establishing the RWST level.
Injection Mode Allowance The injection mode of ECCS operation consists of the ECCS pumps (centrifugal
charging pumps, safety injection pumps, and residual heat removal pumps) and
the containment spray pumps taking suction from the RWST and delivering to the reactor coolant system (RCS) and containment, respectively. The minimum RWST
volume available for ECCS pump injection mode operation is 236,993 gallons.
The maximum RWST volume available for ECCS pump injection is 261,302 gallons.
This is the volume between 94% level and Lo-Lo-1 setpoint or full tank level and Lo-Lo-1 setpoint. See Figure 6.3-7.
6.3-6 Rev. 23 WOLF CREEK Containment and RCS pressures are conservatively assumed to be 0 psig to maximize flow out of the RWST.
Flow out of the RWST during the injection mode includes conservative allowances
for two pumps of each type operating at the following flow rates:
Safety injection pump 450 gpm per pump Centrifugal charging 450 gpm per pump pump RHR pump 4,500 gpm per pump Containment spray pump 3,725 gpm per pump
Total RWST outflow rate during injection mode operation is 18,250 gpm.
Based on the above minimum available RWST volume for injection mode operation and maximum total flow rate out of the RWST, the shortest injection mode
operation time is approximately 13.0 minutes.
ECCS Pumps Transfer Allowance - RHR, Charging, SI This is the volume of water between Lo-Lo-1 and Lo-Lo-2 level setpoints.
Upon receipt of the RHR auto switchover alarm (Lo-Lo-1), the operator initiates the manual operations required to complete switchover as described in Table 6.3-11 in a timely manner.
The ECCS switchover from injection to cold leg recirculation is initiated
automatically upon receipt of the RHR auto switchover signal and is completed via timely operator action at the main control board. Switchover is initiated
via automatic opening of the containment recirculation sump isolation valves
(8811 A/B). This automatic action aligns the suction of the RHR pumps to the
containment recirculation sump to ensure continued availability of a suction
source. Manual actions as described in Table 6.3-11 must be performed following switchover initiation prior to loss of the ECCS transfer allowance to ensure that all ECCS pumps are protected with suction flow available from the
containment sump. The ECCS switchover procedure is structured to facilitate the operator in accomplishing the switchover of both trains of the ECCS from injection to recirculation without the loss of suction head to any pumps.
The time available for switchover is dependent on the flow rate out of the RWST
as the switchover manual actions are performed. As ECCS valves are
repositioned, the flow rate out of the RWST is reduced in magnitude. In order
to analyze the time available for switchover, the following conservative bases are established:
6.3-7 Rev. 13 WOLF CREEK
Based on the above criteria, the minimum time available for the operator to
accomplish the switchover of the ECCS pumps is 9.46 minutes. The large break with single failure constitutes the condition where RWST outflow is the greatest. The worst single failure is for RWST/RHR isolation valve (8812A or
8812B) not to close. This failure increases the backflow from RWST to the
containment. The operators must take additional manual actions as described in
Table 6.3-12 to secure the affected RHR pump and the associated containment recirculation sump/RHR (8811A or 8811B) isolation valve. The minimum time
available for the operator to accomplish the switchover of the ECCS pumps for a
large break and the single failure is 8.15 min.
Containment Spray Pumps Transfer Allowance
The RWST volume between the Lo-Lo-2 setpoint and the empty setpoint is required
for containment spray pump switchover from the RWST to the sump. The minimum
available volume is 11,930 gallons. With both spray pumps operating, this volume provides a minimum switchover time of 2.18 minutes. The maximum volume available for containment spray pumps transfer is 31,756 gallons.
Combined Transfer Allowance for ECCS and CSS Pumps
The water volumes shown on Figure 6.3-7 for ECCS and CSS pump transfer
allowance do not include the water within the instrument uncertainty band for
Lo-Lo-2 setpoint. This water is depleted and provides additional allowance.
The minimum combined transfer allowance is 113,293 gallons and the maximum combined allowance is 128,573 gallons.
Total RWST Water Available for Containment
The minimum water volume available for transferring to the containment based on
a tank level of 94% and EMPTY level setpoint is 356,481 gallons. The maximum water volume between a full tank level and EMPTY level setpoint is 384,519 gallons.
6.3-8 Rev. 23 WOLF CREEK Setpoints and Instrument Error
The level measurement system for the RWST includes four level transmitters, each of which have five setpoints, High, Low, Lo-Lo-1, Lo-Lo-2, and Empty. Two
out of four level transmitters sensing a Lo-Lo-1 condition will initiate the automatic action and the associated alarm. One out of four level transmitters
sensing a High, Low, Lo-Lo-2, or Empty condition initiates the appropriate
alarm for operator action. The operators maintain water in the RWST during
normal plant operation above the Low Alarm. The minimum required volume is
assured by taking periodic readings on the level indication in the control
room. A level indication greater than 94% assures that the RWST contains the
minimum required volume of 394,000 gallons.
The maximum instrument uncertainty for the control room indication is +3.6/-
3.1%. The maximum uncertainty for the Lo-Lo-1 at which the automatic switchover of the RHR pumps occurs is assumed to be +
3.3%. Since the emergency procedures used for ECCS and CSS pumps switchover from injection mode to recirculation allow the operators to use alarms as well as indication, the
Injection, ECCS and CSS pumps' transfer volumes are based on the most limiting
instrument errors, assuming a starting RWST level of 94% or full tank level.
Boron Injection Tank During the preliminary design phase, provisions of heat tracing and fluid
recirculation were made in the BIT system to accommodate highly concentrated
boric acid. It has been determined that an RWST concentration of only 2,400-2,500 ppm boron is required for plant safety considerations. The heat tracing
and fluid recirculation provisions will not be operated and have been
permanently disabled. This change does not affect the favorable conclusions of
the safety analysis. During normal operation, the boron concentration in the
BIT will likely vary between the RCS and RWST concentration.
The boron injection tank (BIT) is connected to the discharge of the centrifugal
charging pumps. Upon actuation by an SIS, the isolation valves associated with
the BIT open automatically, and the centrifugal charging pumps inject the boric
acid solution from the RWST into the RCS.
6.3-9 Rev. 23 WOLF CREEK Boron Injection Surge Tank
The boron injection surge tank (BIST) has no safety significance now that the
BIT boron concentration is analyzed for 0 ppm. The recirculation lines to and
from the BIST have been permanently disconnected from high head injection lines. This system has been abandoned in place and is not planned to be used.
Residual Heat Removal Pumps Two residual heat removal (RHR) pumps are provided. Each pump is a single-
stage, vertical, centrifugal pump. In the event of a LOCA, the RHR pumps are
started automatically on receipt of an SIS. The RHR pumps take suction from the RWST during the injection phase and from the containment sump during the
recirculation phase.
A minimum flow bypass line is provided for each pump to recirculate and return
the pump discharge fluid to the pump suction should these pumps be started with
the RCS pressure above their shutoff head. Once flow is established to the
RCS, the bypass line is automatically closed. This line prevents deadheading
of the pumps and permits pump testing during normal operation.
The RHR pumps are discussed further in Section 5.4.7. A typical pump performance curve is given in Figure 6.3-3.
Centrifugal Charging Pumps Two centrifugal charging pumps are provided. Each pump is a multistage
diffuser design, barrel-type casing with vertical suction and discharge
nozzles. In the event of an accident, the centrifugal charging pumps are started automatically on receipt of an SIS and are automatically aligned to
take suction from the RWST during the injection phase. These high head pumps
deliver flow through the BIT to the RCS at the prevailing RCS pressure. During
the recirculation phase, suction is provided from the RHR pump discharge.
A minimum flow bypass line is provided on each pump discharge to recirculate
flow to the pump suction after cooling, via the seal water heat exchanger, during normal plant operation. Each minimum flow bypass line contains a motor operated valve that is normally open during operations.
Upon receipt of an SIS signal, each valve will close at a calculated flow assuming minimum safeguards actuation. This ensures that peak clad temperatures will not exceed those assumed in the analysis. Depending upon the number of pumps injecting via the Boron Injection Tank (BIT) the flow switch will function to ensure that the minimum recirculation valve opens to maintain at least 60 gallons per minute for pump protection during RCS repressurization.
The SIS also aligns the parallel suction valves from the RWST and closes the series valves from the volume control tank.
6.3-10 Rev. 17 WOLF CREEK The centrifugal charging pumps may be tested during power operation via the minimum flow bypass line.
A typical pump performance curve for the centrifugal charging pumps is
presented in Figure 6.3-4. The required pump performance curve, based upon the accident analysis, lies below this characteristic curve.
Safety Injection Pumps Two safety injection pumps are provided. Each pump is a multi-stage, diffuser
design, split-case centrifugal pump with side suction and side discharge.
In the event of an accident, the safety injection pumps are started
automatically on receipt of an SIS; take suction from the RWST via normally
open, motor-operated valves and deliver water to the RCS during the injection
phase; and take suction from the containment sump via the RHR pumps during the
recirculation phase.
A minimum flow bypass line is provided on each pump discharge to recirculate
flow to the RWST in the event that the pumps are started with the RCS pressure
above pump shutoff head. This line also permits pump testing during normal
plant operation. Two parallel valves in series, with a third valve located in a downstream common header, are provided in this line. These valves are manually closed from the control room as part of the ECCS realignment from the
injection to the recirculation mode. A typical pump performance curve for the
safety injection pumps is presented in Figure 6.3-5. The required pump
performance curve based upon the accident analysis lies below this
characteristic curve.
Boron Injection Recirculation Pumps These pumps provide the capability to recirculate boric acid solution
continuously around a closed loop consisting of the BIT, the BIST, and
associated piping. The pumps were initially provided when the BIT concentration was to be 12 weight percent. The pumps are no longer required
and have no safety significance for a BIT boron concentration of nominal RCS
concentration. The pumps have been permanently disabled from operation.
6.3-11 Rev. 13 WOLF CREEK RHR Heat Exchangers
The RHR heat exchangers are conventional shell and U-tube type units. During
normal cooldown operation, the RHR pumps recirculate reactor coolant through
the tube side while component cooling water flows through the shell side.
During the ECCS operation, water from the containment sump flows through the
tube side. The tubes are seal welded to the tube sheet.
A further discussion of the RHR heat exchangers is found in Section 5.4.7.
Valves Design features employed to minimize valve leakage include:
- a. Valves which are normally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.
- b. Normally, closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water.
- c. Relief valves are enclosed, i.e., they are provided with a closed bonnet.
Motor-Operated Valves The seating design of the motor-operated valves is of the Crane flexible wedge
design. This design releases the mechanical holding force during the first
increment of travel so that the motor operator works only against the frictional component of the hydraulic unbalance on the disc and the packing
box friction. The disc is guided throughout the full disc travel to prevent
chattering and to provide ease of gate movement. The seating surfaces are hard
faced to prevent galling and to reduce wear.
6.3-12 Rev. 13 WOLF CREEK Where a gasket is employed for the body-to-bonnet joint, it is either a fully
trapped, controlled compression, spiral wound asbestos gasket with provisions
for seal welding, or it is of the pressure seal design with provisions for seal
welding. The valve stuffing boxes are packed with a full set of graphite packing. A full set of packing is defined as a depth of packing equal to 1-1/2
times the stem diameter. Figure 6.3-6 illustrates a typical motor-operated
valve.
Maximum opening and closing times for the motor-operated valves used in the
ECCS operations are given in Table 6.3-1.
The motor operator incorporates a "hammer blow" feature that allows the motor
to impact the discs away from the backseat upon opening or closing. This "hammer blow" feature not only impacts the disc but allows the motor to attain its operational speed prior to impact. Valves which must function against system pressure are designed so that they function with a pressure drop equal
to full system pressure across the valve disc.
Manual Globes, Gates, and Check Valves Gate valves employ a wedge design and are straight through. The wedge is
either split or solid. All gate valves have backseat and outside screw and
yoke.
Globe valves, "T" and "Y" style, are full ported with outside screw and yoke
construction.
Check valves are spring loaded, lift piston types for sizes 2 inches and
smaller and swing type for sizes 2-1/2 inches and larger. Stainless steel
check valves have no penetration welds other than the inlet, outlet, and
bonnet. The check hinge is serviced through the bonnet.
The stem packing and gasket of the stainless steel manual globe and gate valves are similar to those described above for motor-operated valves. Carbon steel manual valves are employed to pass nonradioactive fluids only and, therefore, do not contain the seal weld provisions.
6.3-13 Rev. 13 WOLF CREEK Accumulator Check Valves (Swing-Disc)
The accumulator check valve is designed with a low pressure drop configuration
with all operating parts contained within the body.
Design considerations and analyses which assure that leakage across the check
valves located in each accumulator injection line will not impair accumulator
availability are as follows:
- a. During normal operation, the check valves are in the closed position with a nominal differential pressure across the disc of approximately 1,650 psi. Since the valves remain in this position except for testing or when called upon to open following an accident and are, therefore, not subject to the abuse of flowing operation or impact loads caused by sudden flow reversal and seating, they do not experience significant wear of the moving parts, and are expected to function with minimal backleakage. This backleakage can be checked via the test connection, as described in Section 6.3.4.
- b. Testing is performed on the check valves in accordance with the Technical Specifications. This testing confirms the seating of the disc and whether or not there has been an increase in the leakage since the last test.
- c. The experience derived from the check valves employed in the emergency injection systems indicates that the system is reliable and workable; check valve leakage has not been a problem. This is substantiated by the satisfactory experience obtained from operation of the Robert Emmett Ginna plant and subsequent plants where the usage of check valves is identical to WCGS.
- d. The accumulators can accept some in-leakage from the RCS without affecting availability. Continuous in-leakage would require, however, that the accumulator water volume and boron concentration be adjusted periodically to meet the Technical Specification requirements.
6.3-14 Rev. 0 WOLF CREEK Relief Valves
Relief valves are installed in various sections of the ECCS to protect lines
which have a lower design pressure than the RCS. The valve stem and spring
adjustment assembly are isolated from the system fluids by a bellows seal between the valve disc and spindle. The closed bonnet provides an additional
barrier for enclosure of the relief valves. Table 6.3-2 lists the system's
relief valves with their capacities and setpoints.
Butterfly Valves Each main residual heat removal line has an air-operated butterfly valve which
is normally open and is designed to fail in the open position. The actuator is
arranged so that air pressure on the diaphragm overcomes the spring force, causing the linkage to move the butterfly to the closed position. Upon loss of
air pressure, the spring returns the butterfly to the open position. These
valves are left in the full-open position during normal operation to maximize
flow from this system to the RCS during the injection mode of the ECCS
operation. These valves are used during normal RHR system operation to control
cooldown flowrate.
Each RHR heat exchanger bypass line has an air-operated butterfly valve, which
is normally closed and is designed to fail closed. Those valves are used
during normal cooldown to avoid thermal shock to the residual heat removal heat exchanger.
Net Positive Suction Head Available and required net positive suction head (NPSH) for ECCS pumps are
shown in Table 6.3-1. Table 6.2.2-7 provides the assumptions and results of
the NPSH analyses for the RHR pumps. The safety intent of Regulatory Guide 1.1 is met by the design of the ECCS so that adequate NPSH is provided to system
pumps. In addition to considering the static head and suction line pressure
drop, the calculation of available NPSH in the recirculation mode assumes that
the vapor pressure of the liquid in the sump is equal to the containment
ambient pressure. This ensures that the actual available NPSH is always
greater than the calculated NPSH. To ensure that the required NPSH is
available during the recirculation phase of ECCS operation, restriction
orifices are provided in the four discharge lines into the RCS cold legs and in
the two discharge lines into the RCS hot legs. The orifices are sized to
provide the RHR flow rates specified in the notes to Figure 6.3-2.
6.3-15 Rev. 0 WOLF CREEK Accumulator Motor-Operated Valve
As part of the plant shutdown administrative procedures, the operator is
required to close these valves. This prevents a loss of accumulator water
inventory to the RCS and is done after the RCS has been depressurized below the safety injection unblock setpoint. The redundant pressure and level alarms on
each accumulator would remind the operator to close these valves, if any were
inadvertently left open. Power is disconnected at the motor control center
after the valves are closed. In the event that the operator is unable to close
any of these valves, the accumulator vent valve is opened to depressurize the accumulator and avoid the addition of excess water inventory into the RCS.
During plant startup, the operator is instructed, via procedures, to energize
and open these valves before the RCS pressure reaches the safety injection unblock setpoint. Monitor lights in conjunction with an audible alarm will alert the operator should any of these valves be left inadvertently closed once
the RCS pressure increases beyond the safety injection unblock setpoint. After
these valves have been opened, power to these valves is disconnected at the
motor control center.
The accumulator isolation valves are not required to move during power
operation or in a post-accident situation, except for valve testing. For a
discussion of limiting conditions for operation and surveillance requirements
of these valves, refer to Technical Specifications.
For further discussions of the instrumentation associated with these valves, refer to Sections 6.3.5 and 7.6.4.
Motor-Operated Valves and Controls Remotely operated valves for the injection mode which are under manual control (i.e., valves which normally are in their ready position and do not require an
SIS) have their positions indicated on a common portion of the control board.
If a component is out of its proper position, its monitor light will indicate
this on the control panel. At any time during operation when one of these
valves is not in the ready position for injection, this condition is shown
visually on the board, and an audible alarm is sounded in the control room.
The ECCS delivery lag times are given in Chapter 15.0. The accumulator
injection time varies as the size of the assumed break varies, since the RCS
pressure drop will vary proportionately to the break size.
6.3-16 Rev. 13 WOLF CREEK Spurious movement of a motor-operated valve due to an electrical fault in the
motor actuation circuitry, coincident with a LOCA, has been analyzed (Ref. 1)
and found to be an acceptably low probability event. In addition, power
lockout in accordance with BTP ICSB-18 is provided for those valves whose spurious movement could result in degraded ECCS performance. Power lockout is
provided by providing a control power isolation switch for each of these valves
on the main control board. Table 6.3-3 provides a listing of the motor-
operated isolation valves in the ECCS, showing interlocks, automatic features, position indication, and which valves are provided with the power lockout
isolation switch.
The supporting auxiliaries which are required to function and support the ECCS
are the Class 1E emergency busses, the essential service water system, the component cooling water system, and the engineered safety features ventilation systems. The safeguards electrical busses are required to provide electrical
power to the ECCS pumps and motor-operated valves. The essential service water
system and the component cooling water system are required to provide cooling
for the ECCS pumps and the RHR heat exchanger (during recirculation only). The engineered safety features ventilation system is required to provide cooling
for the ECCS pump rooms to maintain the ambient environment within the design
of the pump motors.
Periodic visual inspection and operability testing of the motor-operated valves
in the ECCS ensures that there is no potential for impairment of valve
operability due to boric acid crystallization which could result from valve
stem leakage.
In addition, the location of all motor-operated valves within the containment have been examined to identify any motor operators which may be submerged following a postulated LOCA. Based on a maximum post-LOCA flood level at
El.2004'-6", none of the valves require qualification for submerged operation.
The submerged valves are either not required for accident mitigation, not
closed prior to being flooded, or not required to change position after a LOCA.
Failure modes after flooding have been evaluated for potential effects on valve
position and operator information. Therefore, the flooding of these motor
operators and any resultant postulated failure do not present any problems for
either the short- or long-term ECCS operations, containment isolation, or any
other safety-related function.
6.3.2.3 Applicable Codes and Construction Standards The applicable codes and construction standards for the ECCS are identified in
Tables 3.2-1 and 6.3-1 and discussed in Section 3.2.
6.3-17 Rev. 1 WOLF CREEK 6.3.2.4 Material Specifications and Compatibility
Materials employed for components of the ECCS are given in Table 6.3-4.
Materials are selected to meet the applicable material requirements of the
codes in Table 3.2-1 and the following additional requirements:
- a. All the parts of the components in contact with borated water are fabricated of or clad with austenitic stain-less steel or equivalent corrosion-resistant material.
- b. All the parts of the components in contact (internal) with the sump solution during recirculation are fabricated of austenitic stainless steel or equivalent corrosion-resistant material.
wear.
- d. Valve stem materials are selected for their corrosion resistance, high tensile properties, and resistance to surface scoring by the packing.
6.3.2.5 System Reliability Reliability of the ECCS is considered in all aspects of the system, from
initial design to periodic testing of the components during plant operation.
The ECCS is a two train, fully redundant, standby emergency safety feature.
The system has been designed and proven by analysis to withstand any single
credible active failure during injection or active or passive failure during
recirculation and maintain the performance objectives desired in Section 6.3.1.
Two trains of pumps, heat exchangers, and flow paths are provided for
redundancy as only one train is required to satisfy the performance
requirements. The initiating signals for the ECCS, as described in Section
7.3, are derived from independent sources as measured from process (e.g., low
pressurizer pressure) or environmental variables (e.g., containment pressure).
Redundant, as well as functionally independent variables, are measured to initiate the safety injection signals. Each train is physically separated and protected, where necessary, so that a single event cannot initiate a common
failure. Power sources for the ECCS are divided into two independent trains
supplied from the Class 1E emergency busses from offsite power. Sufficient diesel generating capacity is maintained onsite to provide required power to each train. The diesel generators and their auxiliary systems are completely
independent and each supplies power to one of the two ECCS trains.
6.3-18 Rev. 1 WOLF CREEK The reliability program extended to the procurement of the ECCS components so that only designs which were proven by past use in similar applications were
acceptable for use. For example, the ECCS pumps (safety injection, centrifugal
charging, and residual heat removal pumps) are the same type of pumps that have been used extensively in other operating plants. Their function during
recurrent normal power and cooldown operations in such plants as Zion, D.C.
Cook, Trojan, and Farley has successfully demonstrated their performance
capability. Reliability tests and inspections (see Section 6.3.4.2) further
confirmed their long-term operability. Nevertheless, design provisions were included that would allow maintenance on ECCS pumps if necessary during long-
term operation.
The preoperational testing program assured that the systems, as designed and
constructed, met the functional requirements calculated in the design.
The ECCS is designed with the ability for on-line testing of most components so
the availability and operational status can be readily determined.
In addition to the above, the integrity of the ECCS is assured through examination of critical components during the routine inservice inspection.
A failure modes and effects analysis is provided in Table 6.3-5. Consideration of an active failure of any Westinghouse nuclear steam supply system (NSSS) check valve is excluded from Tables 6.3-5 and 6.3-6 since the NSSS check valves are not considered to be active (powered) components per the Westinghouse ECCS design, particularly with respect to ECCS failure modes and effects and single active failure analyses. As discussed in Section 3.9(N).3.2.1, NSSS check valves are characteristically simple in design and their operation is not affected by seismic accelerations or the maximum applied nozzle loads. Their design is compact and there are no extended structures or masses whose motion could cause distortions that could restrict operation of the valve. The nozzle loads due to maximum seismic excitation do not affect the functional ability of the valve since the valve disc is typically designed to be isolated from the body wall. The clearance supplied by the design around the disc prevents the disc from becoming bound or restricted due to any body distortions caused by nozzle loads. Therefore, the design of the these valves is such that once the structural integrity of the valve is ensured using standard methods, the ability of the valve to operate is ensured by the design features.
Although the design of the NSSS check valves provides assurance of their ability to operate, these NSSS check valves undergo in-shop hydrostatic and seat leakage testing (prior to installation) as well as periodic in-situ valve exercising and inspection to ensure their functional capability. (As discussed in Section 3.1.1.1, the definition of an active component for the purpose of supporting the pump and valve operability program includes NSSS check valves.
These check valves, although not powered components, meet the definition of having mechanical motion and are therefore included in Table 3.9(N)-11.)
- a. Active Failure Criteria
The ECCS is designed to accept a single failure following the incident without loss of its protective function. The system design will tolerate the failure of any single active component in the ECCS itself or in the necessary associated service systems at any time during the period of required system operations following the incident.
A single active failure analysis is presented in Table 6.3-6, and demonstrates that the ECCS can sustain the failure of any single active component in either the short or long term and still meet the level of performance for core cooling.
6.3-19 Rev. 26 WOLF CREEK Since the operation of the active components of the ECCS following a steam line rupture is identical to that following a LOCA, the same analysis is applicable, and the ECCS can sustain the failure of any single active component and still meet the level of performance for the addition of shutdown reactivity.
- b. Passive Failure Criteria
The following philosophy provides for necessary redundancy in the component and system arrangement to meet the intent of the GDC on single failure, as it specifically applies to failure of passive components in the ECCS. Thus, for the long term, the system design is based on accepting either a passive or an active failure.
A single passive failure analysis is presented in Table 6.3-7. It demonstrates that the ECCS can sustain a single passive failure during the long-term phase and still retain an intact flow path to the core to supply sufficient flow to keep the core covered and effect the removal of decay heat. The procedure followed to establish the alternate flow path also isolates the component that failed.
Redundancy of Flow Paths and Components for Long-Term Emergency Core Cooling The following criteria were utilized in the design of the
ECCS:
- 1. During the long-term cooling period following a postulated loss-of-coolant accident, the emergency core cooling flow paths shall be separable into two subsystems, either of which can provide minimum core cooling functions and return spilled water from the floor of the containment back to the RCS.
- 2. Either of the two subsystems can be isolated and removed from service in the event of a leak outside the containment.
- 3. Should one of these two subsystems be isolated in this long-term period, the other subsystem remains
operable.
- 4. Adequate redundancy of the check valves is provided to tolerate failure of a check valve during the long term as a passive component.
6.3-20 Rev. 0 WOLF CREEK
- 5. Provisions are made in the design to detect leakage from components outside the containment, collect this leakage, and provide for maintenance of the affected equipment. For further discussion, see Section 9.3.3 concerning the equipment and floor drainage system.
Thus, for the long-term emergency core cooling function, adequate core cooling capacity exists with one flow path removed from service.
Subsequent Leakage from Components in the ECCS Leakage from mechanical equipment outside the containment will be detected before it propagates to major proportions by a program for periodic visual inspection and leak detection. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump shaft seal.
Evaluation of leak rate, assuming only the presence of a seal retention ring around the pump shaft, showed flows less than 7.5 gpm would result. Piping leaks, valve packing leaks, or flange gasket leaks have been of a nature to build up slowly with time and are considered less severe than the pump seal failure. The auxiliary building floor and equipment drain system leakage detection capability is discussed in Section 9.3.3.
Larger leaks in the ECCS are prevented by the following:
- 1. The piping is classified in accordance with ANS Safety Class 2 and receives a quality assurance program in accordance with 10 CFR 50, Appendix B (refer to Section 3.2).
3.2).
- 3. The system piping is located within a controlled area of the plant.
- 4. The piping system receives periodic pressure tests, and is accessible for periodic visual inspection.
- 5. The piping is austenitic stainless steel which, due to its ductility, can withstand severe distortion without failure.
6.3-21 Rev. 0 WOLF CREEK Process Flow Diagram
Figure 6.3-2 is a simplified illustration of the ECCS. The notes provided with
Figure 6.3-2 contain information relative to the operation of the ECCS in its
various modes. The modes of operation illustrated are full operation of all
ECCS components, cold leg recirculation with RHR pump B operating, and hot leg
recirculation with RHR pump A operating. These are representative of the
operation of the ECCS during accident conditions.
Lag Times
Lag times for initiation and operation of the ECCS are limited by pump startup
time and consequential loading sequence of these motors onto the Class 1E
busses. Most valves are normally in the required position for the ECCS to
fulfill its safety function.
Therefore, valve opening time is not considered for these valves. Power to the
valve operators is available anytime the Class 1E busses are energized. If
there is no loss of offsite power, all pump motors are still sequenced on the
Class 1E busses upon receipt of an SIS. In the case of a loss of offsite
power, a 12-second delay is assumed for diesel startup, then pumps are loaded according to the sequencer. For sequencer times, see Figure 8.3-2.
Potential Boron Precipitation
Boron precipitation in the reactor vessel after a postulated LOCA is precluded
by a backflush of cooling water through the core to reduce boil-off and
resulting concentration of boric acid in the water remaining in the reactor
vessel. This is accomplished by switching from cold leg to hot leg
recirculation approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following an accident.
Three flow paths are available for the hot leg recirculation of sump water.
Each safety injection pump can discharge to two hot legs with suction taken
from RHR pump discharge either directly or indirectly via the charging pump
cross connect. One RHR pump is also aligned to deliver flow to the hot leg
injection header.
Loss of one pump or one flow path does not prevent hot leg recirculation since
redundant methods are available for use. (e.g., If the RHR hot leg
recirculation valve, EJHV8840, does not open due to it being pressure locked, adequate hot leg recirculation flow is provided by the operating safety
injection pumps through the safety injection hot leg isolation valves
EMHV8802A & B.
6.3-22 Rev. 29 WOLF CREEK 6.3.2.6 Protection Provisions
The provisions taken to protect the system from damage that might result from
dynamic effects are discussed in Section 3.6. The provisions taken to protect
the system from missiles are discussed in Section 3.5. The provisions to protect the system from seismic damage are discussed in Sections 3.7(B) and (N), 3.9(B) and (N), and 3.10(B) and (N). Thermal stresses on the RCS are
discussed in Section 5.2.
6.3.2.7 Provisions for Performance Testing Test lines are provided for performance testing of the ECCS, as well as
individual components. These test lines and instrumentation are shown in
Figure 6.3-1. All pumps have miniflow lines for use in testing operability.
Additional information on testing can be found in Section 6.3.4.2.
6.3.2.8 Manual Actions No manual actions are required of the operator for proper operation of the ECCS
during the injection mode of operation. Only limited manual actions are
required by the operator to realign the system for the cold leg recirculation mode of operation, and, after approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, for the hot leg
recirculation mode of operation. The actions required to switch to cold leg recirculation are delineated in Table 6.3-8. Based on the containment pressure-temperature analyses provided in Section 6.2.1, which assume runout flows of all pumps, including the containment spray pumps, which draw from the
RWST, switchover of the RHR pumps occurs approximately 13.7 minutes after the
accident.
The changeover from the injection mode to recirculation mode is initiated
automatically and completed manually by operator action from the main control
room. Protection logic is provided to automatically open the two safety
injection system recirculation sump isolation valves when two out of four RWST
level channels indicate an RWST level less than a low-low-1 level setpoint in
conjunction with the initiation of the engineered safeguards actuation signal (SIS). When the containment sump recirculation valves are fully opened, RHR
pump suction from the RWST is automatically isolated. This automatic action
aligns the two RHR pumps to take suction from the containment sump and to
deliver water directly to the RCS. The RHR pumps continue to operate during this changeover from injection mode to recirculation mode.
Motor operated valves, such as EJHV8840, are normally electrically operated from the control room but may be manually operated locally based on acceptable radiation levels as delineated in applicable emergency operating procedures (EMG's)
6.3-23 Rev. 10 WOLF CREEK The two centrifugal charging pumps and the two safety injection pumps continue to take suction from the RWST, following the above automatic action, until
manual operator action is taken to align these pumps in series with the RHR
pumps.
The RWST level protection logic consists of four level channels with each level channel assigned to a separate process control protection set. Four RWST
transmitters provide level signals to corresponding normally de-energized level
channel bistables. Each level channel bistable would be energized on receipt
of an RWST level signal less than the low-low-1 level setpoint.
A two-out-of-four coincident logic is utilized in both protection cabinets, A
and B, to ensure a trip signal in the event that two-out-of-four level channel
bistables are energized. This trip signal, in conjunction with the SIS, provides the actuation signal to automatically open the corresponding
containment sump isolation valves.
The low-low-1 RWST level signal is also alarmed to inform the operator to
initiate the manual action required to realign the centrifugal charging and safety injection pumps for the recirculation mode.
The manual switchover sequence that must be performed by the operator is
delineated in Table 6.3-8. Following the automatic and manual switchover
sequence, the two RHR pumps take suction from the containment sump and deliver borated water directly to the RCS cold legs. A portion of the A RHR pump
discharge flow is used to supply the two centrifugal charging pumps, which also
deliver water directly to the RCS cold legs. A portion of the discharge flow
from the B RHR pump is used to provide suction to the two safety injection
pumps, which also deliver directly to the RCS cold legs. As part of the manual
switchover procedure (see Table 6.3-8, Step 4), the suctions of the safety
injection and centrifugal charging pumps are cross connected so that one RHR
pump can deliver flow to the RCS and both safety injection and centrifugal
charging pumps, in the event of the failure of the second RHR pump.
See Section 7.5 for process information available to the operator in the control room following an accident.
The consequences of the operator failing to act altogether will be loss of the
high head safety injection pumps and centrifugal charging pumps.
6.3-24 Rev. 15 WOLF CREEK 6.3.3 SAFETY EVALUATION
Safety evaluations are numbered to correspond to the safety design bases in
Section 6.3.1.1.
SAFETY EVALUATION ONE - Except for the RWST, the ECCS is located in the reactor
and auxiliary buildings. These buildings are designed to withstand the effects
of earthquakes, tornadoes, hurricanes, floods, external missiles, and other
appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide
the bases for the adequacy of the structural design of these buildings.
The events which could result in the loss of function of the RWST (i.e.,
tornado missile) will not also cause a DBA. For these events, the boric acid
transfer system is available to provide a borated source of water to achieve
and maintain the plant in a safe shutdown. Therefore, no protection of the RWST is required.
SAFETY EVALUATION TWO - The ECCS is designed to remain functional after an SSE.
Sections 3.7(B).2, 3.9(B), and 3.9(N) provide the design loading conditions
that were considered. Sections 3.5, 3.6, and Appendix 3B provide the hazards analyses to assure that a post accident safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
SAFETY EVALUATION THREE - The ECCS is completely redundant and, as indicated by
Tables 6.3-6 and 6.3-7, no single failure compromises the system's safety
functions. All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0.
SAFETY EVALUATION FOUR - The ECCS is initially tested with the program given in
Chapter 14.0. Periodic inservice functional testing is done in accordance with
Section 6.3.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the ECCS.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portion of this system and supporting systems. Table 6.3-1 shows that the components meet
the design and fabrication codes given in Section 3.2. All the power supplies
and control functions necessary for safe function of the ECCS are Class 1E, as
described in Chapters 7.0 and 8.0.
6.3-25 Rev. 19 WOLF CREEK SAFETY EVALUATION SIX - Section 6.3.2.5 describes provisions made to identify
and isolate leakage or malfunction and to isolate the nonsafety-related
portions of the system.
SAFETY EVALUATION SEVEN - Sections 6.2.4 and 6.2.6 provide the safety
evaluation for the system containment isolation arrangement and testability.
SAFETY EVALUATION EIGHT - As described in Sections 3.11(B) and 3.11(N), all
components of the ECCS required to perform a safety function are designed to
and environmentally qualified to all environments anticipated under normal, testing, and design basis accident conditions.
SAFETY EVALUATION NINE - Chapter 15.0 accidents that result in ECCS operation.
- 1. Increase in heat removed by the secondary system
- a. Inadvertent opening of a steam generator atmospheric relief or safety valve.
- b. Steam system piping failure.
- 2. Decrease in heat removed by the secondary system.
- a. Feedwater system pipe break.
- 3. Decrease in reactor coolant system inventory.
- a. Steam generator tube failure
- b. Loss of coolant accident from a spectrum of postulated piping breaks within the system.
- c. Spectrum of rod cluster control assembly (RCCA) ejection accidents.
- 4. Increase in reactor coolant system inventory
- a. Inadvertent operation of the ECCS during power
operation.
Safety injection system actuation results from any of the following:
6.3-26 Rev. 13 WOLF CREEK
- a. Low pressurizer pressure
- b. Low steam line pressure
- c. High-1 containment pressure
- d. Manual actuation
A safety injection signal rapidly trips the main turbine, closes all feedwater
control valves, trips the main feedwater pumps, and closes the feedwater
isolation valves. However, no direct credit is taken for the turbine trip function, and trip of the main feedwater pumps is not part of the primary success path for accident mitigation.
Following the actuation signal, the suction of the centrifugal charging pumps
is diverted from the volume control tank to the RWST. Simultaneously, the
valves isolating the BIT from the centrifugal charging pumps and the valves
isolating the BIT from the injection header automatically open. The
centrifugal charging pumps then force the boron solution from the RWST into the
cold legs of each loop. The safety injection pumps also start automatically
but operate at shut off head when the RCS is at normal pressure. The passive
accumulator system and the low head system also provide no flow at normal RCS
pressure.
INCREASE IN HEAT REMOVED BY THE SECONDARY SYSTEM
Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve
The most severe core conditions resulting from an accidental depressurization
of the main steam system are associated with an inadvertent opening of a single
steam dump, atmospheric relief, or safety valve.
The assumed steam release is typical of the capacity of any single steam dump (Figure 10.3-1-03), atmospheric relief (Figure 10.3-1-01), or safety valve (Figure 10.3-1-02). The ECCS injection of the boron solution provides
sufficient negative reactivity to meet the DNB design basis. The cooldown for
this case is more rapid than the actual case of steam release from all steam
generators through one steam dump, atmospheric relief, or safety valve. The
transient is quite conservative with respect to cooldown, since no credit is
taken for the energy stored in the system metal other than that of the fuel
elements or the energy stored in the steam generators. Since the transient
occurs over a period of about 5 minutes, the neglected stored energy is likely
to have a significant effect in slowing the cooldown. The analysis provided in
Section 15.1.4 demonstrates that there will be no consequential damage to the
core or reactor coolant system after reactor trip, assuming a stuck rod cluster
control assembly, with offsite power available, and assuming a single failure
in the engineered safety features. It also concludes that the DNB design
limits are not exceeded.
6.3-27 Rev. 27 WOLF CREEK Steam System Pipe Failure
The steam release arising from a rupture of a main steam pipe would result in
energy removal from the RCS, causing a reduction of coolant temperature and
pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. There is an increased
possibility that the core will become critical and return to power.
The core is ultimately shut down by the boric acid injection delivered by the
safety injection system. Capability for injection of the boric acid solution
is maintained, assuming any single failure in the safety injection system.
For cases where offsite power is assumed to be available, the sequencing of
events in the safety injection system is the following. After the generation
of the SIS (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the centrifugal charging pumps start. In 12 seconds, the valves are assumed to be in their final
position, and the pumps are assumed to be at full speed. This delay, described
above, is included in the calculations.
In cases where offsite power is not available, an additional 12-second delay is
assumed to start the diesels and to load the necessary safety injection
equipment onto them.
The analysis has shown that even assuming a stuck RCCA with or without offsite
power, and assuming a single failure in the engineered safeguards, the core
remains in place and intact. Radiation doses will not exceed 10 CFR 100
guidelines.
DECREASE IN HEAT REMOVED BY THE SECONDARY SYSTEM Feedwater System Pipe Break
A major feedwater line rupture is defined as a break in a feedwater line large
enough to prevent the addition of sufficient feedwater to the steam generators
to maintain shell side fluid inventory in the steam generators. If the break is postulated in a feedwater line between the check valve and the steam
generator, fluid from the steam generator may also be discharged through the
break. Further, a break in this location could preclude the subsequent
addition of auxiliary feedwater to the affected steam generator. (A break
upstream of the feedwater line check valve would affect the NSSS only as a loss
of feedwater. This case is covered by the evaluation in Sections 15.2.6 and
15.2.7).
6.3-28 Rev. 0 WOLF CREEK Depending upon the size of the break and the plant operating conditions at the
time of the break, the break could cause either an RCS cooldown (by excessive
energy discharge through the break) or an RCS heatup. Potential RCS cooldown
resulting from a secondary pipe rupture is evaluated in Section 15.1.5.
Therefore, only the RCS heatup effects are evaluated for a feedwater line
rupture.
A feedwater line rupture reduces the ability to remove heat generated by the
core from the RCS for the following reasons:
- a. Feedwater flow to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase prior to reactor trip.
- b. Fluid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip.
- c. The break may be large enough to prevent the addition of any main feedwater after trip.
An auxiliary feedwater system functions to ensure the availability of adequate
feedwater so that:
- a. No substantial overpressurization of the RCS occurs (less than 110 percent of design pressures); and
- b. Sufficient liquid in the RCS is maintained so that the core remains in place and geometrically intact with no loss of core cooling capability.
The engineered safety systems assumed to function are the auxiliary feedwater
system and the safety injection system. For the auxiliary feedwater system, the worst case configuration has been used, i.e., only three nonfaulted steam
generators receive auxiliary feedwater following the break. The flow from the
motor-driven auxiliary feedwater pump feeding the faulted steam generator was
assumed to spill completely through the break. The second motor-driven
auxiliary feedwater pump has been assumed to fail. The turbine-driven
auxiliary feedwater pump delivers 470 gpm to the three nonfaulted steam
generators. This assumption is conservative because it maximizes the purge
time in the feedwater lines before auxiliary feedwater enters the unaffected
6.3-29 Rev. 0 WOLF CREEK A safety injection signal from either low steamline pressure or high
containment pressure initiates flow of cold borated water into the RCS. The
amount of safety injection flow is a function of RCS pressure.
Results of the analyses show that for the postulated feedwater line rupture, the assumed auxiliary feedwater system capacity is adequate to remove decay
heat, to prevent overpressurizing the RCS, and to prevent uncovering the
reactor core. Radioactivity doses from the postulated feedwater line rupture
are less than those previously presented for the postulated steamline break.
All applicable acceptance criteria are therefore met.
DECREASE IN REACTOR COOLANT SYSTEM INVENTORY Steam Generator Tube Failure
The accident postulated and analyzed is the complete severance of a single
steam generator tube, and is assumed to occur at power.
Assuming normal operation of the various plant control systems, the following
sequence of events is initiated by a tube failure:
- a. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side, there is a steam flow/feedwater flow mismatch before the trip as feedwater flow to the affected steam generator is reduced due to the additional break flow which is not being supplied to that unit.
- b. The steam generator blowdown liquid monitor and the condenser offgas radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system, and will automatically terminate steam generator
blowdown.
- c. Continued loss of reactor coolant inventory leads to a reactor trip on low pressurizer pressure or over-temperature T. The resultant plant cooldown leads to a continued reduction in pressurizer level and SIS initiation. The SIS automatically terminates normal feedwater supply and initiates auxiliary feedwater addition. After reactor trip, the break flow reaches equilibrium at the point where incoming safety injection flow is balanced by outgoing break flow. The resultant break flow persists from plant trip until operator action is taken to bring the primary system and
6.3-30 Rev. 15 WOLF CREEK affected steam generator secondary system pressures into equilibrium.
- d. The reactor trip automatically trips the turbine, and if offsite power is available the steam dump valves open, permitting steam dump to the condenser. In the event of a coincident station loss of offsite power, the steam dump valves would automatically close to protect the condenser. The steam generator pressure would rapidly increase, resulting in steam discharge to the atmosphere through the steam generator safety and/or atmospheric relief valves.
- e. Following reactor trip, the continued action of the auxiliary feedwater supply and borated safety injection flow (supplied from the RWST) provide a heat sink which absorbs some of the decay heat. Thus, steam bypass to the condenser or, in the case of loss of offsite power, steam relief to the atmosphere is attenuated during the transient until the recovery procedure leading to isolation is being carried out.
A steam generator tube rupture, as demonstrated in the analyses provided in
Section 15.6.3, causes no subsequent damage to the RCS or the reactor core. An
orderly recovery from the accident can be completed, even assuming simultaneous
loss of offsite power.
LOCA From a Spectrum of Postulated Piping Breaks Within the System Small Break LOCA - Small ruptured pipes, cracks in large pipes, or ejection of a control rod.
A LOCA is defined as a rupture of the RCS piping or of any line connected to
the system from which the break flow exceeds the flow capability of the normal makeup/charging system. Ruptures of small cross-sections will cause expulsion
of the reactor coolant at a rate which can be accommodated by the charging pumps maintaining an operational water level in the pressurizer, permitting the operator to execute an orderly shutdown.
The maximum break size for which the normal makeup system can maintain the
pressurizer level is obtained by comparing the calculated flow from the RCS through the postulated break against the centrifugal charging pump makeup flow
at normal RCS pressure, i.e., 2,250 psia. A makeup flow rate from one
centrifugal
6.3-31 Rev. 15 WOLF CREEK charging pump is adequate to sustain pressurizer level at 2,250 psia for a
0.375-inch-diameter hole. This break results in a loss of approximately 17.5
lb/sec (127 gpm at 130°F and 2,250 psia).
The SIS stops normal feedwater flow by closing the main feedwater isolation
valves and initiates emergency feedwater flow by starting the auxiliary
feedwater pumps.
The small break analyses deal with breaks of up to 1.0 ft 2 in area, where the safety injection pumps play an important role in the initial core recovery because of the slower depressurization of the RCS.
The analysis of this break, as provided in Section 15.6, demonstrates that the
high head portion of the ECCS, together with accumulators, provides sufficient core flooding to keep the calculated peak clad temperature below the required
limits of 10 CFR 50.46. Hence, adequate protection is afforded by the ECCS in
the event of a small break LOCA.
Large Break LOCA A major LOCA is defined as a 1.0 ft 2 or larger rupture of the RCS piping, including the double-ended rupture of the largest pipe in the RCS or of any line connected to that system. The boundary considered for LOCA, as related to
connecting piping, is defined in Section 3.6.
Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. Reactor trip occurs and the safety injection
system is actuated when the pressurizer low pressure trip setpoint is reached.
Reactor trip and safety injection system actuation may be provided by a high
containment pressure signal, depending on the actual break size. These
countermeasures will limit the consequences of the accident in two ways:
- a. Reactor trip and borated water injection provide additional negative reactivity insertion to supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
- b. Injection of borated water ensures sufficient flooding of the core to prevent excessive clad temperatures.
6.3-32 Rev. 11 WOLF CREEK When the pressure falls below approximately 600 psi, the accumulators begin to
inject borated water. The conservative assumption is made that accumulator
water injected bypasses the core and goes out through the break until the
expulsion or entrainment mechanisms for bypassing are calculated not to be effective. This conservatism is consistent with the acceptable features of
ECCS Evaluation Models, as defined by Appendix K, 10 CFR 50.
The pressure transient in the reactor containment during a LOCA affects ECCS
performance in the following ways. The time at which end of blowdown occurs is
determined by a zero break flow which is a result of achieving pressure
equilibrium between the RCS and the containment. In this way, the amount of
accumulator water bypass is also affected by the containment pressure, since
the amount of accumulator water discharged during blowdown is dependent on the
length of the blowdown phase and RCS pressure at end of blowdown. During the reflood phase of the transient, the density of the steam generated in the core is dependent on the existing containment pressure. The density of this steam
affects the amount of steam which can be vented from the core to the break for
a given downcomer head, the core reflooding process, and, thus, the ECCS
performance. It is through these effects that containment pressure affects
ECCS performance.
For breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS limits the clad temperature to below 2200 F and ensure that the core remains in place and substantially intact with its essential heat transfer geometry preserved. See Section 15.6.5 for ECCS sequence of events.
For these breaks, Section 15.6 demonstrates that the ECCS meets the Acceptance
Criteria presented in 10 CFR 50.46. That is:
- a. The calculated peak fuel element clad temperature is less than 2,200 F. b. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
- c. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching.
- d. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.
6.3-33 Rev. 1 WOLF CREEK INCREASE IN REACTOR COOLANT SYSTEM INVENTORY
Inadvertent Operation Of The Emergency Core Cooling System During Power Operation Spurious emergency core cooling system (ECCS) operation at power could be
caused by operator error or a false electrical actuation signal. A spurious
signal may originate from any of the safety injection actuation channels, as described in Section 7.3.
A safety injection signal (SIS) normally results in a reactor trip followed by
a turbine trip. However, it cannot be assumed that any single fault that
actuates the ECCS will also produce a reactor trip. If a reactor trip is
generated by the spurious SIS, the operator should determine if the spurious
signal was transient or steady state in nature. The operator must also
determine if the SIS should be blocked. For a spurious occurrence, the
operator would terminate ECCS and maintain the plant in the hot standby
condition.
If the reactor protection system does not produce an immediate trip as a result
of the spurious SIS, the reactor experiences a negative reactivity excursion
due to the injected boron, causing a decrease in reactor power. The power
mismatch causes a drop in T AVG and consequent coolant shrinkage. The pressurizer pressure and water level decrease. Load will decrease due to the
effect of reduced steam pressure on load after the turbine throttle valve is
fully open. If automatic rod control is used, these effects will be lessened
until the rods have moved out of the core. The transient is eventually
terminated by the reactor protection system low pressurizer pressure trip or by
Results of the analysis show that spurious ECCS operation without immediate
reactor trip presents no hazard to the integrity of the RCS.
If the reactor does not trip immediately, the low pressurizer pressure reactor
trip is actuated. This trips the turbine and prevents excess cooldown, thereby
expediting recovery from the incident.
Criteria Used to Judge the Adequacy of the ECCS
(
Reference:
- a. The peak clad temperature calculated shall not exceed 2,200°F.
6.3-34 Rev. 1 WOLF CREEK
- b. The calculated total oxidation of the clad shall nowhere exceed 0.17 times the total clad thickness before
oxidation.
- c. The calculated total amount of hydrogen generated from the chemical reaction of the clad with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the clad cylinders surrounding the fuel, excluding the clad around the plenum volume, were to react.
- d. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
- e. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by long lived radioactivity remaining in the core.
In addition to and as an extension of the Final Acceptance Criteria, two
accidents have more specific criteria, as shown below.
In the case of the inadvertent opening of a steam generator atmospheric relief or safety valve, an additional criteria for adequacy of the ECCS is: Assuming a stuck RCCA, offsite power available, and a single failure in the engineered
safety features, there will be no return to criticality after reactor trip for
a steam release equivalent to the spurious opening with failure to close, of
the larger of a single steam dump, relief, or safety valve.
For a steam system piping failure, the added criteria is: Assuming a stuck RCCA
with or without offsite power, and assuming a single failure in the engineered
safety features, the core remains in place and intact.
Use of Dual Function Components The ECCS contains components which have no other operating function, as well as
components which are shared with other systems. Components in each category
are as follows:
- a. Components of the ECCS which perform no other function
are:
- 1. One accumulator for each loop which discharges borated water into its respective cold leg of the reactor coolant loop piping.
6.3-35 Rev. 13 WOLF CREEK
- 2. Two safety injection pumps, which supply borated water for core cooling to the RCS. (May be used during check valve testing also.)
- 3. One BIT
- 4. One BIST
- 5. ssociated piping, valves, and instrumentation
- b. Components which also have a normal operating function are as follows:
- 1. RHR pumps and the RHR heat exchangers These components are normally used during the latter stages of normal reactor cooldown and when the reactor is held at cold shutdown for core decay heat removal or for flooding the refueling cavity.
However, during all other plant operating periods they are aligned to perform the low head injection
function.
- 2. Centrifugal charging pumps
These pumps are normally aligned for charging service. As a part of the chemical and volume control system, the normal operation of these pumps is discussed in Section 9.3.4.
- 3. RWST
This tank is used to fill the refueling canal for refueling operations and to provide makeup to the fuel storage pool.
However, during all other plant operating periods it is aligned to the suction of the safety injection pumps and the RHR pumps. The charging pumps are automatically aligned to the suction of the RWST upon receipt of an SIS or a VCT low level alarm. During normal operation, they take suction from the volume control tank.
An evaluation of components required for operation of the ECCS demonstrates
that either:
6.3-36 Rev. 15 WOLF CREEK
- a. The component is not shared with other systems, or
- b. If the component is shared with other systems, it is either aligned during normal plant operation to perform its accident function or, if not aligned to its accident function, two valves in parallel are provided to align the system for injection, and two valves in series are provided to isolate portions of the system not utilized for injection. These valves are automatically actuated by the SIS.
Table 6.3-9 indicates the alignment of components during normal operation and
the realignment required to perform the accident function.
In all cases of component operation, safety injection has the priority usage such that an SIS will override all other signals and start or align systems for
injection.
Limits on System Parameters The analyses show that the design basis performance characteristic of the ECCS
is adequate to meet the requirements for core cooling following a LOCA with the
minimum engineered safety features equipment operating. In order to ensure this capability in the event of the simultaneous failure to operate any single
active component, reactor operating limits are established (see Technical
Specifications).
Normal operating status of the ECCS components is given in Table 6.3-10.
6.3.4 TESTS AND INSPECTIONS
6.3.4.1 ECCS Performance Tests 6.3.4.1.1 Preoperational Test Program at Ambient Conditions
Preliminary operational testing of the ECCS was conducted with the system cold and aligned for normal power operation with the exception that the BIT was
filled with refueling water instead of concentrated boric acid. An SIS was
initiated, and the breakers on the lines supplying offsite power were tripped
manually so that operation of the emergency diesels was tested in conjunction
with the safety injection system. System testing provided the following
verifications of system performance:
6.3-37 Rev. 0 WOLF CREEK
- a. Satisfactory SIS generation and transmission
- b. Proper operation of the emergency diesel generators, including sequential load pickup
- c. Valve operating times
- d. Pump starting times
- e. Pump delivery rates at runout conditions (one point on the operating curve)
Further details of each preoperational test performed are discussed in Chapter
14.0. 6.3.4.1.2 Components
Pumps Separate flow tests of the pumps in the ECCS were conducted during the
preoperational testing (with the reactor vessel head off) to check capability
for sustained operation. The centrifugal charging, safety injection, and RHR pumps discharge into the reactor vessel through the injection lines, the
overflow from the reactor vessel passes into the refueling pool. Each pump was
tested separately with water drawn from the RWST. Data is taken to determine
pump head and flow at this time. Pumps are then run on miniflow circuits and
data taken to determine a second point on the head flow characteristic curve.
Section 6.2.2.1.4 discusses the hydraulic model testing used to verify that the
available net positive suction head is adequate when the RHR pumps and
containment spray pumps take suction from the containment recirculation sumps.
Accumulators Each accumulator is filled with water from the RWST and pressurized with the
motor-operated valve on the discharge line closed. Then the valve is opened
and the accumulator allowed to discharge into the reactor vessel as part of the preoperational testing with the reactor cold and the vessel head off.
6.3.4.2 Reliability Tests and Inspections 6.3.4.2.1 Description of Tests Planned
Routine periodic testing of the ECCS components and all necessary support systems at power is planned. Valves which operate after a LOCA are operated
through a complete cycle, and pumps are operated
6.3-38 Rev. 0 WOLF CREEK individually in this test on their miniflow lines, except the centrifugal charging pumps, if they have been tested by their normal charging function. If
such testing indicates a need for corrective maintenance, the redundancy of
equipment in these systems permits such maintenance to be performed without
shutting down or reducing load under certain conditions. These conditions
include considerations, such as the period within which the component should be
restored to service and the capability of the remaining equipment to provide
the minimum required level of performance during such a period.
The operation of the remote stop valve is tested per the required in-service
testing (ASME code, Section XI). The operation of the check valve in each
accumulator tank discharge line is tested per the required in-service testing (ASME Code, Section XI).
Where series pairs of check valves form the high pressure to low pressure
isolation barrier between the RCS and safety injection system piping outside
the reactor containment, periodic testing of these check valves is performed to
provide assurance that certain postulated failure modes do not result in a
loss-of-coolant from the low pressure system outside the containment with a
simultaneous loss of safety injection pumping capacity.
The safety injection system test line subsystem provides the capability for
determining the integrity of the pressure boundary formed by series check
valves. The tests performed verify that each of the series check valves can
independently sustain differential pressure across its disc and also verify
that the valve is in its closed position. The required periodic tests are to
be performed after each refueling just prior to plant startup, after the RCS
has been pressurized. Temporary Modification Order (TMO) 15-015-EM-00 installed a line crimp on a portion of the safety injection system test line subsystem, which blocks a portion of the test line piping. The line crimp is installed to stop leakage of nitrogenated water out of the 'A' Accumulator through valve EPHV8879A. TMO 15-015-EM-00 must be removed prior to the next required performance of the testing of the affected check valves, no later than plant startup at the end of Refuel 21 (Fall 2016).
Lines in which the series check valves are to be tested are the safety
injection pump cold and hot leg injection lines and the RHR pump cold and hot
leg injection lines.
The Technical Specifications state the periodic component testing requirements.
During periodic system testing, a visual inspection of pump seals, valve
packings, flanged connections, and relief valves is made to detect leakage.
Inservice inspection provides further confirmation that no significant
deterioration is occurring in the ECCS fluid boundary.
Each ECCS subsystem is demonstrated Operable by performance of a flow test, during shutdown, following completion of modifications to the ECCS subsystems
that alter the subsystem flow characteristics.
- 1. For the centrifugal charging pump lines, with a single pump running, the sum of the injection line flow rates, excluding the highest flow
rate, is verified to be greater than or equal to 330 gpm and total
pump flow rate is less than or equal to 556 gpm.
- 2. For the safety injection pump lines, with a single pump running, the sum of the injection line flow rates excluding the highest flow rate, is verified to be greater than or equal to 450 gpm and total pump flow
rate is less than or equal to 670 gpm.
- 3. For the residual heat removal pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to
3800 gpm and the total pump flow rate is less than or equal to 5500
gpm. 6.3-39 Rev. 29 WOLF CREEK Design measures have been taken to assure that the following testing can be
performed:
- a. Active components may be tested periodically for operability (e.g., pumps on miniflow, certain valves, etc.).
- b. An integrated system actuation test
- c. An initial flow test of the full operational sequences can be performed.
The design features which assure this test capability are specifically:
- a. Power sources are provided to permit individual actuation of each active component of the ECCS.
- b. The safety injection pumps can be tested periodically during plant operation, using the minimum flow recirculation lines provided.
- c. The RHR pumps are used every time the RHRS is put into operation. They can also be tested periodically when the plant is at power, using the miniflow recirculation
lines.
- d. The centrifugal charging pumps are either normally in use for charging service or can be tested periodically on
miniflow.
- e. Remote-operated valves can be exercised during routine plant maintenance.
- f. Level and pressure instrumentation is provided for each accumulator tank, for continuous monitoring of these parameters during plant operation.
- Details of the testing of the sensors and logic circuits associated with the generation of an SIS, together with the application of this signal to the operation of each active component, are given in Section 7.2.
6.3-40 Rev. 0 WOLF CREEK
- g. Flow from each accumulator tank can be directed through a test line in order to determine valve operability. The test line can be used, when the RCS is pressurized, to ascertain backleakage through the accumulator check valves.
Pressure instrumentation is also provided in these lines.
- i. An integrated system test can be performed when the plant is cooled down and the RHRS is in operation. This test does not introduce flow into the RCS but does demonstrate the operation of the valves, pump circuit breakers, and automatic circuitry, including diesel starting and the automatic loading of ECCS components on the diesels (by simultaneously simulating a loss of offsite power to the vital electrical busses).
See Technical Specifications for the selection of test frequency, acceptability
of testing, and measured parameters. A description of the inservice inspection
program is included in Section 6.6. ECCS components and systems are designed
to meet the intent of the ASME Code, Section XI for inservice inspection.
6.3.5 INSTRUMENTATION REQUIREMENTS
Instrumentation and associated analog and logic channels employed for
initiation of ECCS operation are discussed in Section 7.3.
This section describes the instrumentation employed for monitoring ECCS components during normal plant operation and also ECCS postaccident operation.
Alarms are annunciated in the control room.
6.3.5.1 Temperature Indication BIT Temperature
Two temperature indicators provide local indication.
RHR Heat Exchanger Temperature
The fluid temperature at both the inlet and the outlet of each RHR heat
exchanger is recorded in the control room.
6.3-41 Rev. 10 WOLF CREEK
6.3.5.2 Pressure Indication
BIT Pressure
BIT pressure is indicated in the control room.
Centrifugal Charging Pump Inlet, Discharge Pressure
There is local pressure indication at the suction and discharge of each
centrifugal charging pump.
Safety Injection Pump Suction Pressure There is a locally mounted pressure indicator at the suction of each safety
injection pump.
Safety Injection Header Pressure Safety injection pump discharge header pressure is indicated in the control
room.
Accumulator Pressure Duplicate pressure channels are installed on each accumulator. Pressure
indication in the control room and high and low pressure alarms are provided by
each channel.
Test Line Pressure A local pressure indicator used to check for proper seating of the accumulator
check valves between the injection lines and the RCS is installed on the
leakage test line.
RHR Pump Suction Pressure Local pressure indication is provided at the inlet to each RHR pump.
RHR Pump Discharge Pressure
RHR discharge pressure for each pump is indicated in the control room. A high
pressure alarm is actuated by each channel.
6.3-42 Rev. 10 WOLF CREEK 6.3.5.3 Flow Indication
Centrifugal Charging Pump Injection Flow
Injection flow to the reactor cold legs is indicated in the control room.
Safety Injection Pump Header Flow
Flow through the safety injection pump header is indicated in the control room.
Safety Injection Pump Minimum Flow
A flow indicator is installed in the safety injection pump minimum flow line.
Test Line Flow
Local indication of the leakage test line flow is provided to check for proper
seating of the accumulator check valves between the injection lines and the
RCS, and for testing other check valves in the ECCS.
RHR Pump Cold Leg Injection Flow The flow from each residual heat removal subsystem to the RCS cold legs is
recorded in the control room. These instruments also control the RHR bypass
valves, maintaining constant return flow to the RCS during normal cooldown.
RHR Pump Minimum Flow A flowmeter installed in each RHR pump discharge header provides control for
the valve located in the pump minimum flow line.
6.3.5.4 Level Indication RWST Level
Water level indicator channels, which indicate in the control room, are
provided for the RWST. Each channel is provided with a high, low, low-low-1, low-low-2, and empty level alarm. The high level alarm is provided to protect against possible overflow of
6.3-43 Rev. 11 WOLF CREEK the RWST. The low level alarm is provided to assure that a sufficient volume
of water is always available in the RWST. The low-low-1 level alarm, as well
as the level indication, alerts the operator to realign the ECCS from the
injection to the recirculation mode following an accident and automatically opens the sump isolation valves. The low-low-2 level alarm, as well as the
level indication, alerts the operator to realign the containment spray pumps
for recirculation. The empty alarm indicates that the usable volume of the
RWST has been exhausted.
Accumulator Water Level Duplicate water level channels are provided for each accumulator. Both
channels provide indication in the control room and actuate high and low water
level alarms.
6.3.5.5 Valve Position Indication Motor/Air-Operated Valves
Valve positions are indicated on the control boards by red and green position
indication lights associated with the control switch for the valve. In
addition, a status monitoring panel is provided which indicates that a valve is in its proper position for safety features system operation by a white light.
A potential bypass of automatic operation is indicated by an amber light. See
Section 7.5.2.2.1 for additional discussion.
Manual Valves Control room position indication and alarms are provided for the following ECCS
manual valves to ensure correct system alignment.
RWST discharge (VOll on Figure 6.3-1, Sheet 1)
RHR recirculation (8717 on Figure 6.3-1, Sheet 1)
Accumulator Isolation Valve Position Indication The accumulator motor-operated valves are provided with red (open) and green (closed) position indicating lights located at the control switch for each
valve. These lights are powered by valve control power and actuated by valve motor operator limit switches.
6.3-44 Rev. 13 WOLF CREEK A monitor light that is on when the valve is not fully open is provided in an
array of monitor lights that are all off when their respective valves are in
proper position. This light is energized from a separate monitor light supply
and actuated by a valve motor-operator limit switch. Additionally, an ESF status panel bypass indication is provided whenever any of these valves leaves
the fully open position.
An alarm annunciator point is activated by both a valve motor operator limit
switch and by a valve position limit switch activated by stem travel whenever
an accumulator valve is not fully open for any reason with the system at
pressure (the pressure at which the safety injection block is unblocked is
approximately 1,970 psig). A separate annunciator point is used for each
accumulator valve.
6.3.6 REFERENCE
6.3-45 Rev. 26 WOLF CREEK TABLE 6.3-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS Accumulators Number 4 Design pressure, psig 700
Design temperature, F 300
Operating temperature, F 45* to 120*
Normal operating pressure, psig 602 to 648
Minimum pressure, psig 585 Total volume, ft 3 (each) 1350 Normal operating water volume, ft 3 (each) 850 Volume N 2 gas, ft 3 (each) 500 Boric acid concentration, ppm boron
(range) 2,300 to 2,500
Relief valve setpoint, psig 700
Seismic Category I
Design code ASME III, Class 2
Material Stainless steel Centrifugal Charging Pumps Number 2 Design pressure, psig 2,800
Design temperature, F 300
Design flow (A), gpm 150 Design head, ft 5,800
Maximum flow, gpm 550
Head at maximum flow, ft 1,400
Discharge head at shutoff, ft 6,200
Required NPSH at maximum flow, ft 28
Available NPSH, ft 44
Design code ASME III, Class 2
Seismic design Category I
Driver:
Type Electric motor
Horsepower, hp 600
Rpm 1,800
Power 4,160 V, 60 Hz, 3-phase, Class IE
Start time <5 sec
Design code NEMA (A) Includes miniflow
- The accumulator operating conditions, as stated in the ASME Section III design specification, are 60F to 150 F. The accumulator tanks can operate at temperatures as low as 45 F based on an ASME Section XI evaluation. Operation is limited to 120 F by the initial containment temperature assumed in the containment integrity accident analyses.
Rev. 13 WOLF CREEK TABLE 6.3-1 (Sheet 2)
Safety Injection Pumps Number 2 Design pressure, psig 1,750 Design temperature, F 300 Design flow rate, gpm 440 Design head, ft 2,780 Maximum flow rate, gpm 660 Head at maximum flow rate, ft 1,760 Discharge head at shutoff, ft 3,645 Required NPSH 25 Available NPSH 44 Design code ASME III, Class 2 Seismic design Category I
Driver:
Type Electric motor Horsepower, hp 450 Rpm 3,600 Power 4,160 V, 60 Hz, 3-phase, Class IE Start time <5 sec Design code NEMA Seismic design Category I
Residual Heat Removal Pumps Number 2 Design pressure, psig 600 Design temperature, F 400 Design flow, gpm 3,800 Design head, ft 350 NPSH required at 4,760 gpm, ft 21.01 Available NPSH at 4,760 gpm, ft 23.79* Design code ASME III, Class 2 Seismic design Category I
Driver:
Type Electric motor Horsepower, hp 500 Rpm 1,800 Power 4,160 V, 60 Hz, 3-phase, Class IE Start time <5 sec Design code NEMA Seismic design Category I
Residual Heat Exchangers (See Section 5.4.7 for design parameters)
Rev. 25 WOLF CREEK TABLE 6.3-1 (Sheet 3)
Refueling Water Storage Tank Quantity 1 Maximum volume (to overflow), gal 419,000
Minimum Water Volume Required, gal 394,000
Boric acid concentration, ppm
boron (range) 2,400 to 2,500
Type Vertical, field
erected
Diameter, ft-in 40-0
Side height, ft-in 46-0
Design pressure, psig Atmospheric
Design temperature, F 120/-60
Material Austenitic stainless
steel
Design code ASME III, Class 2
Seismic design Category I Boron Injection Tank Number 1 Total volume, gal 900
Usable volume at operating conditions, solution, gal 900
Boron concentration, ppm (nominal)
- Design pressure, psig 2,735
Operating pressure Atmospheric
Design temperature, F 300
Operating temperature, F Ambient
Heaters **
Type Strip Design code ASME III, Class 2
Seismic design Category I Boron Injection Surge Tank**
Number 1 Total volume, gal 75
Boron concentration, ppm
(nominal) 0
Design pressure Atmospheric
Operating pressure Atmospheric
Design temperature, F 200
Operating temperature, F Ambient
Heaters **
Type Immersion Design code ASME III, Class 3
Seismic design Category I
- Between 2400 ppm and RCS concentration
- Heaters, pumps, and Boron Injection Surge Tank are no longer required due to
lower system boron concentration. They remain installed but have been
permanently disabled.
Rev. 14 WOLF CREEK TABLE 6.3-1 (Sheet 4)
Boron Injection Tank Recirculation Pumps**
Number 2 Design pressure, psig 150
Design temperature, F 250 Design flow rate, gpm 20 Design head, ft 100
Design code ASME III, Class 3
Seismic design Category I Maximum Opening Or Motor-Operated Valves Closing Time Up to and including 8 inches, time, sec 15 Over 8 inches, time, sec* Valve size (inches) inches 1 min 49 min X 60 sec
- Excluding valves EJ-HV-8809A,B and EJ-HV-8840 and EJ-HV-8716A,B, which have 15-second maximum opening/closing
times. Other exceptions are: EMHV8801A/B 20 seconds EMHV8803A/B 20 seconds BNHV8812A/B 25 seconds EJHV8804A/B 30 seconds
have been permanently disabled.
Rev. 16 WOLF CREEK TABLE 6.3-2 EMERGENCY CORE COOLING SYSTEM RELIEF VALVE DATA Description Fluid Discharged Fluid Inlet Temperature Normal (F)
Set Pressure (psig)Backpressure Constant (psig)Maximum Total Backpressure(psig)Capacity N 2 supply to accumulators N 2 120 700 0 0 1,500 scfm Safety injection pump discharge Water 120 1,825 0 to 15 50 20 gpm Residual heat
removal pump
safety injection lineWater 120 600 0 to 15 50 20 gpm Safety injection
pumps suction
headerWater 100 220 0 to 15 50 25 gpm Accumulator to containment N 2 Gas 120 700 0 0 1,500 scfm Rev. 18 WOLF CREEKTABLE6.3-3MOTOR-OPERATEDISOLATIONVALVESINTHEEMERGENCYCORECOOLINGSYSTEM Location Valve Identification Interlocks Automatic Features Position Indication AlarmsAccumulatorisolation valves8808A,B,C,DPowerlockoutprovidedOpensonSISifpoweronvalveandRCSpressureunblockMCBYes-outof positionSafetyinjectionpump suctionfromRWST8806A&B8923A&BNoneNoneMCBYes-outof positionRHRsuctionfromRWST8812A&BCannotbeopenedunlesssumpvalveclosedClosesonSIScoincident withRWSTlow-low-1level andsumpvalvefully
openMCBYes-outof positionRHRdischarged tosafety injection/chargingpumpsuction8804A&BCannotbeopenedunlesssafetyin-jectionpumpmini-flowisolatedandRHRsuctionvalvefrom RCSclosedNoneMCBYes-outof positionSafetyinjection hotleginjection8802A&BPowerlockout providedNoneMCBYes-outof positionRHRhotleg injection8840Powerlockout providedNoneMCBYes-outof positionContainmentsumpisolationvalve8811A&BCannotbeopenedinnormaloper-ationunlessRHR suctionvalves fromRWST&from RCSclosedOpensonRWST low-low-1with
SISMCBYes-outof positionRev.0 WOLF CREEKTABLE6.3-3(Sheet2)MOTOR-OPERATEDISOLATIONVALVESINTHEEMERGENCYCORECOOLINGSYSTEM Location Valve Identification Interlocks Automatic Features Position Indication AlarmsCVCSsuctionfrom RWSTLCV-112D&ESISOpenonSISMCBYes-outof positionCVCSnormal
suctionLCV-112B&CSISClosesonSISifCVCSsuctionvalvesfromRWSTopenMCBYes-outof positionSafetyinjectionpumptocoldleg8835Powerlockout providedNoneMCBYes-outof positionCVCSnormal
discharge 8105 8106SISClosesonSISMCBNoneBoroninjectiontanksuction8803A&BSISOpensonSISMCBYes-outof positionBoroninjectiontankdischarge8801A&BSISOpensonSISMCBYes-outof position Charging pump/safety injectionpump
crossover8801A&B 8924NoneNoneMCB(8807A&B Only))Yes-outof positionRHRtoRCScold legs8809A&BPowerlockout providedNoneMCBYes-outof positionSafetyInjection pumpminiflow 88138814A&BCannotbeopened unlessRHRdischarge tosafetyinjection&
tochargingpumps closed.Power lockouton8813onlyNoneMCBYes-outof positionRev.9 WOLF CREEKTABLE6.3-3(Sheet3)MOTOR-OPERATEDISOLATIONVALVESINTHEEMERGENCYCORECOOLINGSYSTEM Location Valve Identification Interlocks Automatic Features Position Indication AlarmsRHRcrossconnect8716A&BNoneNoneMCBYes-outof positionSafetyinjectionpumpcross connect8821A&BNoneNoneMCBYes-outof positionChargingpump miniflow 8110 8111SISClosedonSISMCBYes-outof positionMCB-maincontrolboardRev.0 WOLFCREEKTABLE6.3-4MATERIALSEMPLOYEDFOREMERGENCYCORECOOLINGSYSTEMCOMPONENTSComponentMaterialAccumulatorsCarbonsteelcladwithausteniticstainlesssteelBoroninjectiontankAusteniticstainlesssteelBoroninjectionsurgetankAusteniticstainlesssteel PumpsCentrifugalchargingAusteniticstainlesssteelSafetyinjectionAusteniticstainlesssteel ResidualheatremovalAusteniticstainlesssteelRHRheatexchangersShellCarbonsteelShellendcapCarbonsteelTubesAusteniticstainlesssteelChannelAusteniticstainlesssteel ChannelcoverAusteniticstainlesssteel TubesheetAusteniticstainlesssteel ValvesMotor-operatedvalves containingradioactive
fluidsPressurecontainingAusteniticstainlesssteel partsorequivalentBody-to-bonnetLowalloysteelboltingandnutsSeatingsurfacesStelliteNo.6orequivalent StemsAusteniticstainlesssteelor17-4PHstainlessRev.10 WOLFCREEKTABLE6.3-4(Sheet2)ComponentMaterialDiaphragmvalvesAusteniticstainlesssteelAccumulatorcheckvalvesPartscontactingAusteniticstainlesssteelboratedwaterClapperarmshaft17-4PHstainlessReliefvalvesStainlesssteelbodiesStainlesssteelCarbonsteelbodiesCarbonsteel Allnozzles,discs,spindles,andguidesAusteniticstainlesssteelBonnetsforstainlessStainlesssteelorplatedsteelvalveswithoutacarbonsteel balancingbellowsAllotherbonnetsCarbonsteel PipingAllpipingincontactAusteniticstainlesssteelwithboratedwaterRev.0 WOLF CREEK TABLE 6.3-5 FAILURE MODE AND EFFECTS ANALYSIS - EMERGENCY CORE COOLING SYSTEM - ACTIVE COMPONENTS
Effect Failure
Component*** Failure Mode ECCS Operation Phase On System Operation* Detection Methods** Remarks
- 1. Motor-operated Fails to close Injection - cold legs Failure reduces redun- Valve position indi- Valve is elec-gate valve LCV- on demand of RC loops dancy of providing VCT cation (open to trically inter-112B (LCV-112C discharge isolation. closed position locked with iso-
analogous) No effect on safety change) at MCB. lation valve for system operation; Valve close position LCV-112D. Valve isolation valves LCV- monitor light and closes on actua-
112C and 8440 provide alarm for group tion by an SIS back-up tank discharge monitoring of provided isolation
isolation. components at MCB. valve LCV-112D is at a full open
position.
- 2. Motor-operated Fails to open Injection - cold legs Failure reduces redun- Same methods of Valve is elec-gate valve LCV- on demand of RC loops dancy of providing detection as those trically inter-112D (LCV-112E fluid flow from RWST to stated for item 1, locked with the analogous) suction of HHSI/CH except open position instrumentation
pumps. No effect on monitor light and that monitors
safety for system opera- alarm for group moni- fluid level of tion. Alternate iso- toring of components, the VCT. Valve lation valve LCV-112E and closed to open opens upon actu-opens to provide back- position change indi- ation by a "low-
up flow path to suction cation at MCB. low level" VCT of both HHSI/CH pumps. signal.
- 3. Centrifugal Fails to Injection and recir- Failure reduces redun- HHSI/CH pump dis- One HHSI/CH pump charging pump deliver working culation - cold legs dancy of providing charge header flow is used for nor-1 (pump 2 fluid of RC loops emergency coolant to (FI-917A) at MCB. mal charging of analogous) the RCS via the BIT at BIT discharge pres- RCS during plant prevailing incident RCS sure (PI-947) at operation. Pump pressure. Fluid flow MCB. Open pump circuit breaker from HHSI/CH pump 1 switchgear circuit aligned to close will be lost. Minimum breaker indication on actuation by
flow requirements at on MCB. Circuit an SIS.
prevailing high RCS breaker close posi-
pressures will be met tion monitor light
by HHSI/CH pump 2 for group monitoring
delivery via BIT. of components at MCB.
Common breaker trip
alarm at MCB.
- See list at end of table for definition of acronyms and abbreviations used.
- As part of plant operation, periodic tests, surveillance inspections, and instrument calibrations are made to monitor equipment and performance. Failures may be detected during such monitoring of equipment in addition to detection methods noted. *** NSSS check valves are not considered to be active (powered) components in the Westinghouse design with respect to the active components considered in this Emergency Core Cooling System (ECCS) Failure Modes and Effects Analysis (FMEA). Rev. 27
WOLF CREEK
TABLE 6.3-5 (Sheet 2)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Methos**
Remarks
- 4. Motor-operated Fails to close Injection - cold legs Failure prevents iso- Same methods of de- Valve aligned to globe valve on demand of RC loops lation of HHSI/CH tection as those close upon act ua- 8110 (8111 pump 1 (pump 2) mini- stated for item 1. tion by a coin- analogous) flow line. No effect cident SIS and on safety for system charging pump flow operation. Alternate > 258.9 gpm.
isolation valve 8111 =
in HHSI/CH pump 2
(pump 1) provides
miniflow isolation
and assures adequate
HHSI/CH pump flow.
Fails to open Injection - cold legs Failure prevents open- Same methods of de- Valve aligned to on demand of RC loops ing of HHSI/CH pump 1 tection as those open a coinici- (pump 2) miniflow line. stated for item 1. dent SIS and
No effect on safety charging pump flow for system operation. > 173.5 gpm.
Alternate valve 8111 =
(8110) in HHSI/CH
pump 2 (pump 1) pro-
vides adequate miniflow.
- 5. Motor-operated Fails to close Injection cold legs Failure reduces redun- Same methods of de- Valve aligned
gate valve on demand of RC loops dancy of providing iso- tection as those to close upon
8105 (8106 lation of HHSI/CH pump stated for item 1. actuation by
analogous) discharge to normal an SIS.
charging line of CVCS.
No effect on safety for
system operation. Alter-
nate isolation valve
8106 provides back-up
normal CVCS charging
line isolation.
- 6. Motor-operated Fails to open Injection - cold legs Failure reduces redun- Same methods of de- Valve aligne d to gate valve on demand of RC loops dancy of fluid flow tection as those open upon ac tua 8803A (8803B paths from HHSI/CH stated for item 2. tion by an S IS. analogous) pumps to the RCS via
BIT. No effect on
safety for system oper-
ation. Alternate isola-
tion valve 8803B opens
to provide back-up flow
path from HHSI/CH pumps
to BIT.
Rev. 26
WOLF CREEK
TABLE 6.3-5 (Sheet 3)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 7. Motor-operated Fails to open Injection - cold legs Failure reduces redun- Same methods of de- Valve aligned to gate valve on demand of RC loops dancy of fluid flow tection as those open upon actu a- 8801A (8801B paths from HHSI/CH stated for item 2. tion by an SIS. analogous) pumps to the RCS via
BIT. No effect on
safety for system opera-
tion. Alernate isola-
tion valve 8801B opens
to provide back-up flow
path from HHSI/CH pumps
to BIT.
- 8. Motor-operated a. Fails to Injection - cold legs Failure reduces work- Valve position indi- Valve is regu-gate valve close on of RC loops ing fluid delivered cation (open to lated by signa l FCV-610 (FCV- demand to RCS from RHR pump closed position from flow tran s- 611 analogous) 1. Minimum flow change) at MCB. mitter located requirements for LHSI RHR pump return in pump discha rge will be met by LHSI/ line to cold header. The
RHR pump 2 delivering legs flow indication control valve
working fluid to RCS. (FI-618) at MCB. opens when the RHR pump dis-
charge flow is less than ~816 gpm and closes whe n the flow excee ds ~1,650 gpm.
Rev. 26
WOLF CREEK
TABLE 6.3-5 (Sheet 4)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- b. Fails Injection - cold legs Failure results in an Same methods of
closed of RC loops insufficient fluid flow detection as those
through LHSI/RHR pump stated for item 8.a, 1 for a small LOCA or except closed to
steam line break result- open position change
ing in possible pump indication at MCB.
damage. If pump becomes
inoperatrive minimum flow
requirements for LHSI
will be met by LHSI/RHR
pump 2 delivering working
fluid to RCS.
- 9. RHR pump 1 Fails to Injection - cold legs Failure reduces redun- RHR pump return line The RHR pump i s (pump 2 deliver working of RC loops dancy of providing emer- to coldlegs flow sized to deliv er analogous) fluid gency collant to the indication (FI-618) reactor coolan t RCS from the RWST at and low flow alarm through the RH R low RCS pressure (195 at MCB. RHR pump heat exchanger psig). Fluid flow discharge pressure to meet plant
from LHSI/RHR pump 1 (PI-614) at MCB. cooldown
will be lost. Minimum Open pump switch- requirements a nd flow requirements for gear circuit breaker is used during LHSI will be met by indication at MCB plant cooldonw LHSI/RHR pump 2 Circuit breaker close and startup
delivering working positioni monitor operations. T he fluid. lighjt and alarm for pump circuit
group monitoring of breaker is
components at MCB. aligned to clo se Common breaker trip on actuation b y alarm at MCB. on SIS.
(pump 2 deliver working of RC loops ancy of providing pressure (PI-919) at breaker aligne d analogous) fluid emergency coolant to MCB. SI pump dis- to close on ac- the RCS from the RWST charge flow (FI-918) tuation by an SIS. at high RCS pressure at MCB. Open pump
(1,520 psi). Fluid switchggear circuit
flow from HHSI/SI pump breaker indication
1 will be lost. Mini- at MCB. Circuit
mum flow requirements breaker close posi-
for HHSI will be met tion monitor light
by HHSI/SI pump 2 and alamr for group
delivering working monitoring of compo-
fluid. nents at MCB. Commmon
breaker trip alarm at
MCB.
Rev. 26
WOLF CREEK
TABLE 6.3-5 (Sheet 5)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 11. Motor-operated Fails to open Recirculation - cold Failure reduces redun- Same methods of Valve is actu-gate valve on demand legs of RC loops dancy of providing detection as those ated to open b y 8811A (8811B fluid from the con- stated for item 2. an SIS in
analogous) tainment sump to the In addition, failure coincidence
RCS during recirculat- may be detected with two out o f tion. LHSI/RHR pump through monitoring four "low-low-1 1 will not provide of RHR pump return level" RWST si g- recirculation flow. line to cold legs nals. Valve i s Minimum LHSI flow flow indication (FI- electrically
requirements will be 618) and RHR pump interlocked fr om met through opening of discharge pressure remotely being isolation valve 8811B (PI-614) at MCB. opened from MC B and recirculation of by isolation
fluid by LHSI/RHR valves 8812A, pump 2. 8701A, and 870 2A.
- 12. Motor-operated Fails to close Recirculation - cold Failure reduces redun- Same methods of Valve is elec-gate valve on demand legs of RC loops dancy of providing flow detection as those trically inter- 8812A (8812B isolation of contain- stated for item 1. locked with is o- analogous) ment sump from RWST. lation valve
No effect on safety 8811A and may
for system operation. not be opened
Alternate check isola- unless valve
tion valve 8958A pro- 8811A is close
- d. vides back-up isolation.
- 13. Motor operated Fails to close Recirculation - cold Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing detection as those
8716A (8716B LHSI/RHR pump train stated for item 1.
analogous) separation for recircu-
lation of fluid to cold
legs of RCS. No effect
on safety for system
operation. Alternate
isolation valve 8716B
provides back-up iso-
lation for LHSI/RHR pump
train separation.
Rev. 26
WOLF CREEK
TABLE 6.3-5 (Sheet 6)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 14. Motor-operated Fails to close Recirculation - cold Failure reduces redun- Same methods of Valve is elec-globe valve on demand legs of RC loops dancy of providing iso- detection as those trically inter- 8813 lation of HHSI/SI pumps stated for item 1. locked with is o- miniflow line isolation lation valves
from RWST. No effect 8804A and 8804 B on safety for system and may not be operation. Alternate opened unless
isolation valve 8814A these valves a re and 8814B in each pumps closed.
miniflow line provide
back-up isolation.
- 15. Motor-operated Fails to close Recirculation - cold Failure reduces redun- Same methods of Same remark as globe valve on demand legs of RC loops dancy of providing detection as those that stated fo r 8814A (8814B isolation of HHSI/SI stated for item 1. item 16.
analogous) pump 1 miniflow isolation
from RWST. No effect
on safety for system
operation. Alternate
isolation valve 8813 in
main miniflow line pro-
vides back-up isolation.
- 16. Motor-operated Fails to open Recirculation - cold Failure reduces redun- Same methods of Valve is elec-gate valve on demand legs of RC loops dancy of providing detection as those trically inter- 8804A NPSH to suction of stated for item 2. locked with is o- HHSI/CH pumps from lation valves
LHSI/RHR pumps. No 8814A, 8814B, effect on safety for 8813, 8701A an d system operation. Mini- 8702A. Valve
mum NPSH to HHSI/CH cannot be open ed pump suction will be unless valve 8 813 met by flow from LHSI/ or valves 8814 A RHR pump 2 via cross- and 8814B are
tie line and opening closed and val ve of isolation valve 8701A or 8702A is 8807A or 8807B and closed.
isolation valve 8804B.
Rev. 26 WOLF CREEK TABLE 6.3-5 (Sheet 7)
Effect Failure Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 17. Motor-operated Fails to open Recirculation - cold Failure reduces redun- Same methods of Valve is elec-gate valve on demand legs of RC loops dancy of providing detection as those trically inter- 8804B NPSH to suction of stated for item 2. locked with is o- HHSI/SI pumps from lation valves
LHSI/RHR pumps. No 8814A, 8814B, effect on safety for 8813, 8701B, system operation. Mini- and 8702B.
mum NPSH to HHSI/SI pump Valve cannot b e suction will be met by opened unless
flow from LHSI/RHR pump valve 8813 or
1 via cross-tie line and valves 8814A a nd opening of isolation valve 8814B are clos ed 8807A or 8807B and isola- and valve 8701 B tion valve 8804A. or 8702B is
closed.
- 18. Motor-operated Fails to open Recirculation - cold Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing detection as those
8807A (8807B fluid flow through stated for item 2.
analogous) cross-tie between
suction of HHSI/CH
pumps and HHSI/SI pumps.
No effect on safety for
system operation. Alter-
nate isolation valve
8807B open to provide
back-up flow path through
cross-tie line.
- 19. Motor-operated Fails to close Recirculation - cold Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing detection as those
8806A (8806B flow isolation of stated for item 1.
analogous) HHSI/SI pump suction
from RWST. No effect
on safety for system
operation. Alternate
check isolation valve
8926A provides back-up
isolation.
- 20. Motor-operated Fails to close Recirculation - cold Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing detection as those
LCV-112D (LCV- flow isolation of suc- stated for item 2.
112E analogous) tion of HHSI/CH pumps
from RWST. No effect
on safety for system
operation. Alternate
check isolation valve
8546 provides back-up
isolation.
Rev. 26 WOLF CREEK
TABLE 6.3-5 (Sheet 8)
Effect Failure
Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 21. RHR pump 1 Fails to Recirculation - cold Failure reduces redun- Same methods of
(pump 2 deliver working legs of RC loops dancy of providing re- detection as those
analogous) fluid circulation of coolant stated for item 11.
to the RCS from the
containment sump.
Fluid flow from LHSI/RHR
pump 1 will be lost.
Minimum recirculation
flow requirements for
LHSI flow will be met
by LHSI/RHR pump 2 de-
livering working fluid.
- 22. SI pump 1 Fails to Recirculation - cold Failure reduces redun- Same methods of
(pump 2 deliver working or hot legs of RC dancy of providing re- detection as those
analogous) fluid loops circulation of coolant stated for item 12.
to the RCS from the
containment sump to
cold legs of RC loops
Fluid flow from HHSI/SI
pump 1 will be lost.
Minimum recirculation
flow requirements for
HHSI flow will be met by
HHSI/SI pump 2 deliver-
ing working fluid.
- 23. Motor-operated Fails to close Recirculation - hot Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing re- detection as those
8809A circulation of coolant stated for item 1.
to the RCS from the
containment sump to
hot legs of RC loops.
Fluid flow from LHSI/
RHR pump 1 will con-
tinue to flow to cold
legs of RC loops.
Minimum recirculation
flow requirements to hot
legs of RC loops will be
met by LHSI/RHR pump 2
recirculation fluid to
RC hot legs via HHSI/SI
pumps.
Rev. 26 WOLF CREEK TABLE 6.3-5 (Sheet 9)
Effect Failure Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 24. Motor-operated Fails to open Recirculation - hot Failure reduces redun- Valve position
gate valve on demand legs of RC loops dancy of providing re- indication (closed
8716A (8716B circulation of coolant to open position
analogous) to the RCS from the change) at MCB.
containment sump to Valve close posi-
the hot legs of RC tion monitor light
loops. Fluid flow and alarm at MCB.
from LHSI/RHR pump 1 In addition, RHR
will be lost. Mini- pump discharge
mum recirculation flow pressure (PI-614)
requirements to hot at MCB.
legs of RC loops will
be met by LHSI/RHR pump
2 recirculating fluid
to RC hot legs via
HHSI/SI pumps.
- 25. Motor-operated Fails to open Recirculation - hot Same effect on system Same methods of
gate valve on demand legs of RC loops operation as that detection as those
8840 stated for item 26. stated for item 2.
In addition, RHR pump
discharge pressure
(PI-614) at MCB.
- 26. Motor-operated Fails to close Recirculation - hot Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing re- detection as those
8809B circulation of coolant stated for item 1.
to the RCS from the
containment sump to
the hot legs of RC
loops. Fluid flow
from LHSI/RHR pump 2
will continue to flow
to cold legs of RC loops.
Minimum recirculation
flow requirements to
hot legs of RC loops
will be met by LHSI/
RHR pump 1 recirculating
fluid to RC hot legs.
- 27. Motor-operated Fails to close Recirculation - hot Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing detection as those
8821A (8821B flow isolation of stated for item 1.
analogous) HHSI/SI pump flow to
cold legs of RC loops.
No effect on safety for
system operation. Alter-
nate isolation valve
8835 provides back-up
isolation against flow to cold legs of RC loops. Rev. 26 WOLF CREEK TABLE 6.3-5 (Sheet 10)
Effect Failure Component
- Failure Mode ECCS Operation Phase On System Operation*
Detection Method**
Remarks
- 28. Motor-operated Fails to open Recirculation - hot Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing re- detection as those
8802A (8802B circulation of coolant stated for item 2.
analogous) to the hot legs of RCS In addition, SI pump
from the containment discharge pressure
sump via HHSI/SI pumps. (PI-919) and flow
Minimum recirculation (FI-918) at MCB.
flow requirements to
hot legs of RC loops
will be met by LHSI/
RHR pump 1 recirculating
fluid from containment
sump to hot legs of RC
loops and HHSI pump 2
recirculating fluid to
hot legs 1 and 4 of RC
loops through the open-
ing of isolation valve
8802B.
- 29. Motor-operated Fails to close Recirculation - hot Failure reduces redun- Same methods of
gate valve on demand legs of RC loops dancy of providing flow detection as those
8835 isolation of HHSI/SI stated for item 1.
pump flow to cold legs
of RC loops. No effect
on safety for system
operation. Alternate
isolation valves 8821A
and 8821B in cross-tie
line between HHSI/SI
pumps provide back-up
isolation against flow
to cold legs of RC loops.
- 30. RHR pump 1 Fails to de- Recirculation - hot Failure reduces redun- Same methods of
(pump 2 liver working legs of RC loops dancy of providing re- detection as those
analogous) fluid circulation of coolant stated for item 11.
to the RCS from the
containment sump to
the hot legs of RC
loops. Fluid flow
from LHSI/RHR pump 1
will be lost. Mini-
mum flow requirements
to hot legs of RC loop
will ve met by LHSI/RHR
pump 2 recirculating
fluid to RC hot legs
via HHSI/SI pumps.
Rev. 26 WOLF CREEK
TABLE 6.3-5 (Sheet 11)
List of acronyms and abbreviations
RC - Reactor coolant
BIT - Boron injection tank RCD - Reactor coolant system
BIST - Boron injection surge tank RHR - Residual heat removal
CH - Charging RWST - Refueling water storage tank
HHSI - High head safety injection SI - Safety Injection
LHSI - Low head safety injection VCT - Volume control tank
MCB - Main control board
NPSH - Net positive suction head
Rev. 10 WOLFCREEKTABLE6.3-6SINGLEACTIVEFAILUREANALYSISFOREMERGENCYCORECOOLINGSYSTEMCOMPONENTSComponentMalfunction CommentsInjectionPhase1.Pumpsa.CentrifugalchargingFailstostartTwoprovided;evaluationbasedonoperationofone.b.SafetyinjectionFailstostartTwoprovided;evaluationbasedonoperationofone.c.ResidualheatremovalFailstostartTwoprovided;evaluationbasedonoperationofone.2.Automaticallyoperatedvalvesa.Boroninjectiontankisolation(1)InletFailstoopenTwoparallelpaths;eachpathconsistingoftwoisolationvalvesinparallellines;onevalveineitherlinerequiredtoopen.(2)OutletFailstoopenTwoparallellines;onevalveineitherlinerequiredtoopen.Rev.10 WOLFCREEKTABLE6.3-6(Sheet2)ComponentMalfunction Commentsb.Centrifugalchargingpumps(1)Suctionlinefromre-FailstoopenTwoparallelvalves;onlyonefuelingwaterstoragevalverequiredtoopen.
tank(2)DischargelinetotheFailstocloseTwovalvesinseries;onlyonenormalchargingpathvalverequiredtoclose.(3)MiniflowbypasslineFailstocloseTwoparallelvalves;onlyonevalverequiredtoclose.(4)SuctionfromvolumeFailstocloseTwovalvesinseries;onlyonecontroltankvalverequiredtoclose.RecirculationPhase1.Valvesoperatedautomaticallyduringswitchovertorecirculationa.Residualheatremovalpumps(1)SuctionlinefromFailstoopenTwoparallellines;onlyonecontainmentsumpvalveineitherlinerequiredtoopen.(2)Suctionlinefromre-FailstocloseCheckvalveinserieswithafuelingwaterstoragegatevalveineachparallelline;tankoperationofonlyonevalveineachlinerequired.Rev.10 WOLFCREEKTABLE6.3-6(Sheet3)ComponentMalfunction Comments2.Valvesoperatedmanuallyfromthecontrolrooma.SafetyinjectionpumpFailstocloseCheckvalveinserieswithtwosuctionlinefromrefuelinggatevalvesineachparallelwaterstoragetankline;operationofonlyonevalveineachlinerequired.b.CentrifugalchargingpumpFailstocloseCheckvalveinserieswithagatesuctionlinefromrefuelingvalveineachparallelline;oper-waterstoragetankationofonlyonevalveineachlinerequired.c.HighheadpumpsuctionFailstoopenSeparateandindependentpathstolineatdischargeofsafetyinjectionpumpsandchargingresidualheatexchangerpumpstakesuctionfromdischargeofresidualheatexchangers;oper-ationofonlyonevalverequired.d.ResidualheatremovalFailstocloseTwovalvesinseries;operationcross-connectlineofonerequired.e.SafetyinjectionpumpFailstocloseTwoparallelvalvesprovidedminiflowlinesinserieswithathird;operationofeitherbothparallelvalvesor thesingleseriesvalverequired.f.Safetyinjection/chargingFailstoopenTwoparallelvalvesprovided;cross-connectlineinsuctionoperationofonerequired.
headerg.Safetyinjection/residualFailstoopenThreeflowpathsavailable;heatremovalhotlegisola-adequateflowtocoreisassured tionvalvesbyanytwo.h.Safetyinjection/residualFailstocloseRedundantvalvesprovidedwithheatremovalcoldlegiso-suitablearrangements.lationvalvesRev.0 WOLFCREEKTABLE6.3-7EMERGENCYCORECOOLINGSYSTEMRECIRCULATIONPIPINGPASSIVEFAILUREANALYSISLONG-TERMPHASEFlowPathIndicationofLossofFlowPathAlternateFlowPathLowHeadRecirculationFromcontainmentsumptolowAccumulationofwaterinaresidualViatheindependent,headinjectionheaderviatheheatremovalpumpcompartmentoridenticallowheadresidualheatremovalpumpsandauxiliarybuildingsumpflowpath,utilizingtheresidualheatexchangersthesecondresidualheatexchangerandresidualheatremoval pumpHighHeadRecirculationFromcontainmentsumptotheAccumulationofwaterinaFromcontainmenthighheadinjectionheaderresidualheatremovalpumpsumptothehighviaresidualheatremovalpump,compartmentortheauxiliaryheadinjectionresidualheatexchanger,andbuildingsumpheadersviaalter-thehighheadinjectionpumpsnateresidualheatremovalpump,residualheatex-injection,orchargingpumpRev.0 WOLFCREEKTABLE6.3-8SEQUENCEOFCHANGEOVEROPERATIONFROMINJECTIONTORECIRCULATIONWithoutbeingstopped,theRHRpumpsarerealignedfortherecirculationmodebytheautomaticopeningofthesumpisolation valves,whichoccursuponreceiptoftheRWSTlow-low-1level signalandanSIS.TheisolationvalveineachRHRsuctionline fromtheRWSTisthenautomaticallyclosed.Atthesametime,the ComponentCoolingWaterSystemisautomaticallyalignedtoprovidecoolingtotheRHRheatexchanger,andtoterminateflowtothe fuelpoolcoolingheatexchangers.Thefollowingremotemanual operatoractionsfromthecontrolroomarerequiredtocomplete thechangeoveroperationfromtheinjectionmodetothe recirculationmode.1.Closethetworemotemotor-operatedvalvesinthecrossoverlinedownstreamoftheresidualheatremovalheatexchangers(8716AandB).2.Closethethreemotor-operatedisolationvalvesinthesafetyinjectionpumpminiflowlines(8814AandB;8813).3.Openthemotor-operatedvalveinthelinefromtheARHRpumpdischargetothechargingpumpsuctionandthemotor-operatedvalveinthelinefromtheBRHRpumpdischargetothesafetyinjectionpumpsuction(8804Aand
B).4.Openthetwoparallelmotor-operatedvalvesinthecommonsuctionlinebetweenthechargingpumpsuctionandthesafetyinjectionpumpsuction(8807AandB).5.Closethetwoparallelmotor-operatedvalvesinthelinefromtheRWSTtothechargingpumpsuctionandthevalvesinthe linefromtheRWSTtothesafetyinjectionpumpsuction(LCV 112DandE;8806AandB).NOTE:Theseoperatoractionsdonotincludeallthestepslistedintheemergencyoperatingprocedures.Theoperatorsare trainedtoaccomplishtheECCSpumpsswitchoverintimely mannerwithintheavailabletime.Rev.14 WOLFCREEKTABLE6.3-9EMERGENCYCORECOOLINGSYSTEMSHAREDFUNCTIONSEVALUATIONComponentNormalOperatingArrangementAccidentArrangementRefuelingwaterLineduptosuctionofsafetyLineduptosuctionofstoragetankinjectionandresidualheatre-centrifugalcharging,movalpumpssafetyinjectionandre-sidualheatremovalpumpsCentrifugalLinedupforchargingserviceSuctionfromrefueling chargingpumpssuctionfromvolumecontrolwaterstoragetank,dis-tank,dischargevianormalchargelineduptoinlet charginglineofboroninjectiontank.Valvesforrealignmentmeet singlefailurecriteriaResidualheatLineduptocoldlegsofreactorLineduptocoldlegsof removalpumpscoolantpipingreactorcoolantpipingResidualheatLineduptocoldlegsofreactorLineduptocoldlegsofexchangerscoolantpipingreactorcoolantpipingRev.0 WOLFCREEKTABLE6.3-10NORMALOPERATINGSTATUSOFEMERGENCYCORECOOLINGSYSTEMCOMPONENTSFORCORECOOLINGNumberofsafetyinjectionpumpsoperable2Numberofcentrifugalchargingpumpsoperable2 NumberofRHRpumpsoperable2 NumberofRHRheatexchangersoperable2 RWSTvolume,gallons,minmaintained394,000BoronconcentrationinRWST,ppm(range)2,400to 2,500Boronconcentrationinaccumulatortank,ppm2,300to(range)2,500Numberofaccumulatortanks4 Minimumaccumulatorpressure,psig585Nominalaccumulatorwatervolume,ft 3 850Systemvalves,interlocks,andpipingrequiredfortheabovecomponentswhichareoperableAllRev.13 WOLF CREEK
TABLE 6.3-11 RWST OUTFLOW (LARGE BREAK) - NO FAILURES Auto/Manual Actions Min Time Available (Min)
RWST Outflow (gals)
RWST Volume Avail Min/Max (gals)
Transfer suction of RHR pumps to containment recirculation sumps (Automatic); Reset SIS; Transfer
suction of CCPs and SIPs to the discharge of RHRHX; Reset CIS Phases A and B and CSS actuation.
9.46 90,469 90,469/107,711 Transfer suction of CSPs to containment recirculation sumps 2.18 11,930 11,930/31,756 NOTES: (1) See Table 6.3-8 for additional details for the transfer of ECCS pumps for cold leg recirculation. (2) See Table 6.2.2-3 for a description of CSPs switchover.
(3) The operators are trained to accomplish the switchover of the ECCS and CSS pumps within the available time per emergency operating procedures. (4) The RWST volume available is the minimum and maximum volumes of water between Lo-Lo-1 and Lo-Lo-2 or Lo-Lo-2 and Empty level setpoints. (5) The minimum time available is based on the RWST outflow and maximum RWST depletion rate.
Rev. 23 WOLF CREEK
TABLE 6.3-12 RWST OUTFLOW (LARGE BREAK) - WORST SINGLE FAILURE Auto/Manual Actions Min Time Available (Min)
RWST Outflow (gals)
RWST Volume Avail Min/Max (gals)
Transfer suction of RHR pumps to containment recirculation sumps (Automatic); Reset SIS; Secure the affected RHR pump and sump valve of the
train with the failed valve; Transfer
suction of CCPs and SIPs to the discharge of RHR pumps; Reset CIS Phases A and B and CSS actuation.
8.15 90,469 90,469/107,711 Transfer suction of CSPs to containment
recirculation sumps 2.18 11,930 11,930/31,756
NOTES:
(1) The worst single active failure is the failure of one of the RWST supply valves to RHR pumps to close following the opening of containment recirculation sump valves. The opertor mitigates the consequences of this failure by securing the affected RHR pump and the associated containment recirculation sump valve. (2) If a single active failure of Valves BNHV0003 or BNHV0004 were to occur (valve fails to close), the operator has 2.25 minutes for accomplishing the switchover of the CSS pumps and secure the pump on the faulted train. The risk associated with this failure is minimal as only one train is required for protection. (3) The operators are trained to accomplish the switchover of the ECCS and CSS pumps within the available time per emergency operating procedures. (4) The RWST volume available is the minimum and maximum volumes of water between Lo-Lo-1 and Lo-Lo-2 or Lo-Lo-2 and Empty level setpoints. (5) The minimum time available is based on the RWST outflow and maximum RWST depletion rate.
Rev. 23
- * * *-----*---*--*-*--*--------------*-----------------------------------------------------------------*---*-------------------*-----*-----------
REV. 10 WOLP CRBH UPDATBD IAI'BTY ANALYSII RBPOR.T FIGURE 6.3-2 CSHEET U Et.IERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM . .
NOTE:THISDIAGRAMISASIMPLIFICATIONOFTHESYSTEMINTENDEDTOFACILITATETHEUNDERSTANDINGOFTHEPROCESS.FORDETAILSOFTHEPIPING,VALVES,INSTRUMENTATION,ETC.REFERTOTHEENGINEERINGFLOWDIAGRAM.REFERTO PROCESSFLOWDIAGRAMTABLESFORTHECONDITIONATEACHNUMBEREDPOINT.REV.0WOLFCREEKUPDATEDSAFETYANALYSISREPORT FIGURE 6.3-2EMERGENCYCORECOOLINGSYSTEMPRPOCESSFLOW DIAGRAM,SHT2 WOLF CREEK NOTES TO FIGURE 6.3-2 MODES OF OPERATION Mode A -Injection This mode presents the process conditions for the case of maximum safeguards, i.e., all pumps operating, following accumulator delivery.
Two residual heat removal (RHR) pumps, two safety injection (SI) pumps, and two centrifugal charging (CC) pumps operate, taking suction from the RWST and delivering to the reactor through the cold leg tions. Note that the flow from each pump is less than its maximum runout since the pump discharge piping is shared by the two pumps of each subsystem.
Note also that the SI pump branch connections to the residual lines are assumed very close to their discharge into the accumulator lines, thereby eliminating any increase in RHR branch line head loss due to the combined flows of the RHR and SI pumps. The RHR line resistance was assumed to be the minimum of the allowable bank presented in the limiting pressure drop and elevation head design requirements, allowing maximum RHR injection flow. Mode B -Cold Leg Recirculation This mode presents the process conditions for the case of cold leg recirculation, assuming RHR pump number 2 ating, SI pumps numbers 1 and 2 operating, and cc pumps numbers 1 and 2 operating.
In this mode, the ECCS pumps operate in series, with only the RHR pump capable of taking suction from the containment sump. The recirculation coolant is then delivered by the RHR pump to both of the SI pumps which deliver to the reactor through their cold leg connections and to both of the CC pumps, which deliver to the reactor through their cold leg connections.
The RHR pump also delivers flow directly to the reactor through two cold legs since the RHR discharge cross-connect valves are closed when making the transfer from injection to recirculation.
Mode c -Hot Leg Recirculation This mode I)Pesents the process conditions for the case of hot leg recirculation, assuming RHR pump number 1 operating, cc pumps numbers 1 and 2 operating, and SI pumps numbers 1 and 2 operating.
Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 2) In this mode, the ECCS pumps again operate in series only the RHR pump taking suction from the containmen The recirculated coolant is then delivered by the to both of the cc pumps which continue to deliver to reactor through their cold leg connections and to bo the SI pumps which deliver to the reactor through th leg connections.
The RHR pump also delivers directl reactor through two hot leg connections.
with sump. pump the of ir hot to the Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 3)
VALVE ALIGNEMNT CHART Operational ModesValve No.A B C1OCC2OCC 3OCC4OCC5OCC6OOC7OOC 8CCO9CCO10CCC11CCC12COO13COO14CCC15CCC 16CCC17CCC 18OOO 19OOO20COO 21COO22OCC 23OCO 24OOC 25CCO 26OOC 27CCC 28OCC 29COO 30CCC 31CCC 32OOO35OOO36CCC37CCC38CCC 39CCC 40CCC O - open C - closed Rev. 10 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 4)
MODE A - INJECTION PHASE (RUNOUT CONDITIONS FOLLOWING ACCUMULATOR DELIVER)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)1Refueling waterAtm tank100--370,0002(a)10016,9052,333-313 psia10016,0252,211-4-1009,6951,338-5-1008,8561,222-611 psia100880121-7-1006,330873-8>10 psia100839116-9>10 psia10044061.5-1010 psia10044061.5-11116510044061.5-12<25100395-1310 psia10044061.5-14116510044061.5-15<25100395-16-1007811-171050100802111-1873100200.528-19-1002,414.5333-20-1002,414.5333-21Borated Water010000-22Borated Water010000850 (ft 3)(b)23Nitrogen010000500 (ft 3)24Reactor coolant-10000-Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 5)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)25Refueling water01004,428611-261381004,428611-27-1004,428611-28471004,428611-298610000-30-10000-31-1004,428611-32861004,428611-338610000-34Reactor coolant-10000-35Refueling water01004,428611-361381004,428611-37-1004,428611-38471004,428611-398610000-40-10000-41-1004,428611-42861004,428611-43Recirc. CoolantContainment pressure12000-4412000-4512000-46Refueling waterLow Pressure10000-4710000-4810000-4910000-5010000-Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 6)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)51Refueling waterLow pressure10000-5210000-53>10 psia100839116-54-10000-551,51910041958-56-10000-5710 psia10041958-5810 psia10041958-591,51910041958-601,51610012417-61~010012417-621,45610071499-632000 ppm boron-10071499900 64Refueling water1,39610071499-651,008100178.524.6-66388100178.524.6-NOTES:
(a)At reference conditions, 100 F and 0 psig (b)Minimum allowable volume at normal operating conditions Rev. 10 l WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 7) MODE B -COLD LEG RECIRCULATION (PUMP NUMBER 2 OPERATING)
Flow ' Pressure Temperature (gpm) (a) Volume Location Flyid (psi g) (F) (lb/sec) (gal) 1 Refueling
' Atm tank 100 <5000 water 2 II 100 0 0 3 If 100 0 0 4 If 100 0 0 5 If 100 0 0 6 Recirc. 186 0 0 coolant 7 100 0 0 water 8 II 100 0 0 9 Recirc. -35 186 1,278 170 water 10 If 186 440 59 11 If -1,165 186 -440 59 12 Refueling 100 0 0 water 13 Recirc. -35 186 440 59 coolant 14 II -1,165 186 -440 59 15 Refueling 100 0 0 water 16 II 100 0 0 17 Recirc. 1,050 186 880 117 coolant 18 II 73 186 220 29 19 II 186 1,761 235 20 If 186 220 29 21 Nitrogen 0 Ambient 0 0 950Cft 3> 22 Nitrogen 0 Ambient 0 0 23 If 0 Ambient 0 0 400( ft 3) Rev. 0
l l<<>LF CREEK NOTES TO FIGURE 6.3-2 (Sheet 8) Flow Pressure Temperature
{qpm} (a) Volume Location Fluid {EsigJ {F} {lbLsec} (gal} 1 24 Recir<J* 212 0 0 coolant 25 II -12 212 4,800 640 26 It 113 212 4,800 640 27 ** 212 4,800 640 28 It 29 186 3,082 411 29 u 56 186 0 0 30 .. 60 186 1,718 229 31 It 65 186 4,800 640 32 II 55 186 3,082 411 33 " 0 186 0 0 34 II 212 0 0 35 Refueling 100 0 0 water 36 II 100 0 0 37 " 100 0 0 38 " 100 0 0 39 u 100 0 0 40 II 100 0 0 41 II 100 0 0 42 II 100 0 0 43 .Recirc. containment 212 ""350,000 coolant pressure 44 II ** 212 4,800 640 45 II II 212 0 0 46 -Re£ueli:ncT--
LOW 0 -****-..
water pressure 47 II u 100 0 0 48 II " 100 0 0 49 II II 100 0 0 so " II 100 0 0 Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 9)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)51Refueling waterLow pressure10000-52Recirc. Coolant-18600-53-18600-54-18600-55~1519186~1,41956-56730186838111-57~3018641956-58~3018641956-59~1,51918641956-601,51618612416-61018612416-621,45618671495-63-18671495-64Recirc. Coolant1,39618671495-651,008186178.524-66388186178.524-NOTES:(a)At reference conditions, 212 F and 0 psig(b)Minimum water volume at operating conditions.
Rev.10 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 10)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)1Refueling waterAtm tank100--<50002-10000-3-10000-4-10000-5-10000-6Recirc. Coolant-18200-7Refueling water-10000-8-10000-9Recirc. Coolant~25<18666088-10~25<18666088-11~715<18666088-12Refueling water-10000-13Recirc. Coolant~25<18666088-14~715<18666088-15Refueling water-10000-16-10000-17Recirc. Coolant0<18600-18-<18600-19-18600-20-18600-Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 11)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)21Nitrogen-Ambient00-22Nitrogen0Ambient00950 (ft 3)(b)230Ambient00400 (ft 3)24Recirc. Coolant-21200-25-<21200-26-<21200-27-<21200-28-<18600-29-<18600-30-<18600-31-<18600-32-<18600-3350<1862,641352-34-21200-35122124,800640-361132124,800640-37-2124,800640-38-<18600-3955<18600-4060<1862,158288-4165<1864,800640-4255<1862,641352-43Recirc. CoolantContainment pressure212---4421200-452124,800640-467<1862,642352-475<1861,321176-48645<18666088-49-<1861,651220-50645<18666088-Rev. 0 WOLF CREEK NOTES TO FIGURE 6.3-2 (Sheet 12)
Location Fluid Pressure (psig)Temperature (F)Flow (gpm)(a) (lb/sec)Volume (gal)51-<18633044-52-<18600-53-<18600-54-<1862,158288-55~1,519<18641956-56<35<1861,320180-57~35<18641856-58~35<18641956-59~1,519<18641956-601,516<18612416-61~0<18612416-621,456<18671495-63--<18671495-64Recirc. Coolant1,896<18671495-651,008<186178.524-66388<186178.524-NOTES:(a)At reference conditions, 212F and 0 psig.(b)Minimum water volume at operating conditions.
Rev. 10 500 350 -300 1-LIJ Li.J l.L 250 0 <[ Li.J 200 ::c 150 100 so 0. 0 1000 WOLF CREEK 2000 3000 1+000 5000 FLOW (GPM) WOLP CREEK 20 1-Li.J Li.J l.L 10 ---0 6000 ::c C/) :z: Rev. 0 UPDATED SAFETY ANALYSIS REPORT FIGURE 6.3-3 TYPICAL RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVE 7000 6000 5000 000 1-U.I I.U u... 0 < UJ :z: 3000 2000 1000 0 0 100 WOLF CREEK 200 300 FLOW ( GPM) 4-00 500 WOLF CREEK 600 20 1-I..LJ UJ u... 10 _ _.. 0 :r: en Q.. z: Rev. 0 UPDATED SAFETY ANALYSIS REPORT FIGURE 6.3-4 CENTRIFUGAL CHARGING PUMP PERFORMANCE CURVE Lj ::::i < c.J a: >-!=. 0 8 ('I) WOLF CREEK 8*8 tD N N N 8 00 {133:1} OV3H (133.:J} HSdN 0 0 (s;) 0 g 8 ::::t--::E 0... (,!:) 8 u.. ('I) 8 N 8 0 8 8 tD N Rev. 0 WOLF CREEK SAFETY ANALYSIS REPORT FIGURE 6.3-5 TYPICAL SAFETY INJECTION PUMP PERFORMANCE CURVE OPERATOR ---STEM --YOKE ---LEAKOFF
< PLUO<IED on CIIPPE I NG BOX Rev. 5 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 6.3-6 GATE VALVE ASSEMBLY f!WST LEVEL HI ALAf!M NORMAL LEVEL RW ST LEVEL LO ALARM (506.2 IN)
INJECTION VOLUME (208 IN I ALARM ECCS PUMP TRANSFER ALLOWANCE
[81 IN) (53 IN)
(0 IN) EL 2000 FT 61N H EA 0 ER _/ ______ __, L___ 24" DIAMETEA PIPE NOTES: 1. The minimum volume (394,000 gals) reguired by Tech. Spec. is assured by verifying that the RWST level is greater than 94 %. The maximum volume (419, 000 gals) is up to the overflow nozzle. 2. The minimum/maximum volumes are dependent on in,strument unceL*tainties.
The injection volumes represent water volume between 94%level or overflow nozzle & LOL0-1, ECCS transfer volumes represent water volumes between LOL0-1 & LOL0-2 and CSS transfer volumes represent water volumes between LOL0-2 and EMPTY setpoints_
- 3. The RHR pumps auto switchover occurs LOL0-1 and the operators initiate css pumps switchover@
LOL0-2. 4. Tank volume (gal./ft) is as follows: 0 in_ to 184 in_ -9361 gal.(ft_ 184 in. to 276 in. -9371 gal./ft. 276 in. to 368 in. -9376 gal.(ft_ 368 in. to 537 in. -9385 gal./ft_ CREEK REV. 23 UPDATED WOLF SAFETY AN A LYSIS REPORT Figure 6.3-7 RWST LEVELS AND VOLUMES WOLF CREEK 6.4 HABITABILITY SYSTEMS The control room habitability systems include missile protection, radiation shielding, radiation monitoring, and smoke detection capability, control room filtration, pressurization and air conditioning, lighting, personnel support, and manual fire protection. These habitability systems are provided to permit
access to and occupancy of the control room during normal plant operations, as
well as during and following emergency conditions.
There is no equipment located within the control room boundary, such as batteries, which emits noxious gases.
The only potential sources for the release of any gases into the control room are the discharge of the fire extinguishers, the discharge of the Halon system
into the cable trenches and chases, and leakage of the control room air-
conditioning unit refrigerant. The release of any one of these gases would not result in a toxicity level which would be hazardous to the control room operators.
For a discussion of the control room ventilation, refer to USAR Section 9.4.1.
The ventilation and air-conditioning equipment discussed in this section is the same control room and control building equipment discussed in Section 9.4.1, Control Building HVAC. This section only addresses emergency service
requirements and responses, including operation of control room ventilation and
air-conditioning equipment under emergency conditions. Lighting systems are
discussed fully in Section 9.5.3, and are not discussed herein. Other
equipment and systems are described only as necessary to define their
connection with control room habitability and, accordingly, reference is made
to other appropriate sections.
6.4.1 DESIGN BASES
6.4.1.1 Safety Design Bases
The control room filtration, pressurization and air-conditioning systems, and the radiation monitoring system, the emergency lighting system, the isolation
dampers in the control building supply air, exhaust, and access control exhaust
ducting are treated as safety-related items and are required to function 6.4-1 Rev. 10 WOLF CREEK under emergency conditions. These habitability systems are required to function following a DBA to enable the plant operators to achieve and/or
maintain the plant in a safe shutdown condition. The following safety design
bases are met:
SAFETY DESIGN BASIS ONE - The habitability systems are housed within a structure capable of withstanding the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The habitability systems are designed to remain functional after an SSE and to perform their intended function following a
postulated hazard, such as a fire, internal missiles, or pipe break (GDC-3 and
4).SAFETY DESIGN BASIS THREE - Habitability system redundancy is provided so that safety functions can be performed, assuming a single active component failure coincident with a loss of offsite power.
SAFETY DESIGN BASIS FOUR - The habitability systems are designed so that the active components are capable of being tested during plant operation.
Provisions are made to allow for inservice inspection of appropriate components of the control room air-conditioning system.
SAFETY DESIGN BASIS FIVE - The habitability systems are designed and fabricated according to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The capability to isolate all nonsafety-related HVAC system penetrations of the control building boundary is provided, if required, so that the occupation and habitability of the control room will not be compromised.
SAFETY DESIGN BASIS SEVEN - The radiation exposure of control room personnel throughout the duration of any one of the postulated DBAs discussed in Chapter 15.0 does not exceed the guideline values of GDC-19.
SAFETY DESIGN BASIS EIGHT - Throughout the duration of any one of the postulated hazardous chemical releases discussed in Section 2.2 or DBAs
discussed in Chapter 15.0, the habitability systems maintain the control room atmosphere at environmental conditions suitable for occupancy per GDC-19. The habitability systems comply with Regulatory Guides 1.78 and 1.95. 6.4-2 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS NINE - The control room ventilation system is capable of automatic transfer from its normal operational mode to its emergency mode upon
detection of conditions which could result in accidental radiation exposure of control room personnel in excess of GDC-19 limits.
6.4.1.2 Power Generation Design Bases The control room ventilation and air-conditioning system power generation design bases are discussed in Section 9.4.1.1.2.
6.4.2 SYSTEM DESIGN 6.4.2.1 Definition of Control Room Envelope The control room envelope includes the control room and all areas in or adjacent to the control room containing plant information and equipment that may be needed during an emergency, including pantry, sanitary facilities, and Class 1E air-conditioning equipment rooms (the control room air-conditioning units are located in the Auxiliary Building).
6.4.2.2 Ventilation System Design The control building (including the control room) HVAC systems are described in Section 9.4.1 and shown in Figure 9.4-1. Codes and standards applicable to the
control building HVAC systems are listed in Table 3.2-1. Elevation and plan
views are shown in Figures 1.2-25, 1.2-27, and 1.2-28.
The control room ventilation and air-conditioning system is a recirculation system. The system is designed to control the level of airborne contamination in the control room atmosphere and to control the temperature and humidity for personnel safety and comfort.
Upon actuation of the system to the emergency mode of operation, as outlined in Section 9.4.1, the control building exhaust isolation dampers and the control
building supply air isolation dampers close; the air-conditioning system
switches to emergency recirculation.
Redundant control room emergency pressurization systems are used to pressurize the control room envelope during emergency recirculation. Supply air and
exhaust system isolation dampers close in less than 10 seconds, in accordance
with Regulatory Guide 1.95.
6.4-3 Rev. 8 WOLF CREEK Redundant radiation monitors are provided to control ventilation system operation. The radiation monitors are located in the control building supply
air system ductwork, downstream of the supply unit.
6.4.2.3 Leaktightness During the emergency mode of operation, the control room is maintained at an overpressure of 1/4 inch w.g. (minimum) by the control room pressurization system to prevent infiltration from surrounding areas of unfiltered air.
Potential leak paths are listed in Section 9.4.1.2.3.
Detailed design calculations for control room leakage rates and required pressurization system design conditions have been performed.
For an analysis of the radiological consequences to the control room occupants in the unlikely event of a LOCA, see Section 15.6.5.
6.4.2.4 Interaction With Other Zones and Pressure-Containing Equipment The control room envelope is isolated and pressurized during the accident involving the release of radioactive gases in the surrounding zones. The control room air-conditioning system is operated in the emergency recirculation
mode, with outside filtered air used to maintain control room pressurization.
The control room pressurization system maintains the control room at a slight positive pressure during emergency operation. If smoke is detected in the
control building supply air system, it is alarmed in the control room.
Those doors which form part of the control room pressure boundary are designed to maintain their specified leaktightness at a positive control room pressure of 1/4 inch w.g.
The use of fire extinguishers located in the control room envelope does not yield a hazardous concentration of toxic gas. Piping not connected or related to control room equipment is routed outside the pressurized boundary. Portable
self-contained breathing apparatus that satisfy Regulatory Guide 1.95 guidelines are readily available for use by the control room operators.
6.4-4 Rev.19 WOLF CREEK 6.4.2.5 Shielding Design A description of the radiation sources and shielding required to maintain the habitability of the control room during normal operations and during the course of postulated accidents is provided in Section 12.3. The shielding design is based on the requirements specified in GDC-19. A plan view drawing of the
control room and associated structures identifying distances and shield thicknesses is shown in Figure 12.3-3.
6.4.3 SYSTEM OPERATIONAL PROCEDURES NORMAL MODE - Control room ventilation system operation in the normal mode is described in Section 9.4.1.2.3. Normal operation of the fire protection system
is described in Section 9.5.1.2.3.
EMERGENCY MODE - Control room ventilation system operation in the emergency mode is described in Section 9.4.1.2.3.
6.4.4 DESIGN EVALUATIONS Safety evaluations are numbered to correspond with the safety design bases.
SAFETY EVALUATION ONE - The safety-related portions of the habitability systems are located in the auxiliary and control buildings. These buildings are
designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of the
structural design of these buildings.
SAFETY EVALUATION TWO - The safety-related portions of the habitability systems are designed to remain functional after an SSE. Sections 3.7(B).2 and 3.9(B)
provide the design loading conditions that were considered. Sections 3.5, 3.6, and 9.5.1 provide the hazards analyses to assure that a safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
SAFETY EVALUATION THREE - The system design for the safety-related portions of the habitability systems provides for complete redundancy, and, as indicated by
Table 9.4-5, no single failure will compromise the systems' safety functions.
All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0.
SAFETY EVALUATION FOUR - The habitability systems were initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is
done in accordance with Section 6.4.5. 6.4-5 Rev. 13 WOLF CREEK Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for portions of the control room air-
conditioning system.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portions of these systems and supporting systems. The power supplies and control
functions necessary for safe functioning of the safety-related portions of the
habitability systems are Class 1E, as described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - Section 9.4.1.2.3 describes the provisions made to assure the isolation of the control room.
SAFETY EVALUATION SEVEN - The direct radiation exposure of a control room occupant throughout the duration of any one of the postulated DBAs discussed in Chapter 15.0 does not exceed GDC-19 requirements. A detailed discussion of the dose calculation model for control room operators is discussed in Appendix 15A.
Control room shielding design, based on the most limiting design basis LOCA
fission product release, is discussed in Section 12.3.
SAFETY EVALUATION EIGHT - Throughout the duration of any of the postulated hazardous chemical releases discussed in Section 2.2 or DBAs discussed in
Chapter 15.0, the habitability system maintains the control room environmental conditions below those established by Regulatory Guides 1.78 and 1.95 and GDC-
- 19. Compliance with Regulatory Guides 1.78 and 1.95 is provided in Tables 6.4-1 and 6.4-2, respectively.
SAFETY EVALUATION NINE - Upon detection of high radiation in the induction trunk, the control room ventilation system is capable of automatic transfer from normal to emergency mode so as to minimize the exposure of control room
personnel.
6.4.5 TESTS AND INSPECTIONS Testing and inspection of control room HVAC systems are described in Section 9.4.1.4.The emergency mode of the control room HVAC system has been subjected to an acceptance test to verify that the system will maintain a 1/4-inch w.g.
positive pressure in the emergency zone. Testing complies with Regulatory Guide 1.95, as described in Table 6.4-2.
6.4-6 Rev. 13 WOLF CREEK The control room is classified as Type B per Regulatory Guide 1.78. Since the air exchange rate exceeds 0.06 air exchanges per hour for the control room, periodic testing of the control room pressurization system is not required per
the exclusion provisions of regulatory guides 1.78 and 1.95. The periodic
testing is not required based on the adequacy of a 400 cfm pressurization flow
rate (Ref. 1).
The control room pressurization system flow rate is an optimized value, based on detailed design calculations, which provides acceptable protection for the
control room operators during both radiological and toxic gas accidents. The
pressurization makeup rate provides for approximately 0.24 volume changes per
hour. Periodic verification testing of the presurization system is, therefore, not justified in this case.
During the PSAR review stage, the issue of periodic testing of the pressurization system flow was discussed in detail as open item B-8, "Outstanding Issues and Positions for the Callaway Plant," for Supplement 1 to the Callaway SER. During a meeting with the NRC staff on June 26, 1975, the
NRC staff agreed that periodic testing was not required for the SNUPPS (Wolf Creek and Callaway) units based on the adequacy of a 400 cfm pressurization flow rate. Refer to the NRC meeting summary dated September 8, 1975.
6.4.6 INSTRUMENTATION REQUIREMENT
Safety-related instrumentation and isolation signals are discussed in Sections 9.4.1.2.3 and 7.3.
Indication of all fan operational status is provided in the control room.
An indication of the position of all isolation dampers is provided in the control room.
All instrumentation associated with filtration units complies with Regulatory Guide 1.52, as described in Table 9.4-2.
Alarms indicate induction trunk airborne gaseous radioactivity per the nominal values listed in Table 11.5-3. A smoke detector is also provided in the control building supply air intake with an alarm in the control room.
6.4-7 Rev. 13 WOLF CREEK Redundant chlorine detectors are installed in the control building supply air
intake duct. With the removal of one-ton storage containers from the WCGS
site, the chlorine detectors will not be maintained operable and will not
initiate control room ventilation isolation. These chlorine detectors are
spared in place. If single chlorine containers greater than 150 lbs are
brought back on site, the chlorine detectors shall be made operable and provide
alarm indication at the control room and automatically isolate the control room
in seven seconds for chlorine concentrations of 5 ppm or grater in control
building supply air intake duct. Chlorine accident analysis will be performed
and the USAR and Technical Specifications revised to reflect the use of greater
than 150 lb. containers of chlorine.
A discussion of the range, alarm points, isolation setpoint, and minimum
sensitivity for the redundant radiation monitors installed in the control
building supply air induction trunk is presented in Section 11.5.
6.4.7 REFERENCE
- 1. NRC Staff meeting summary for June 26, 1975, dated September
8, 1975.
6.4-8 Rev. 29 WOLF CREEK TABLE 6.4-1
COMPARISON OF THE DESIGN TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.78, DATED JUNE 1974 TITLED "ASSUMPTIONS FOR EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM
DURING A POSTULATED HAZARDOUS CHEMICAL RELEASE" Regulatory Guide WCGS Position 1.78 Position In evaluating the habitability of a nuclear power plant control room during a postulated
hazardous chemical release, the following assumptions should be made:
- 1. If major depots or storage tanks of 1. See Section 2.2.
hazardous chemical such as the chemical
listed in Table C-1 of the guide are known
or projected to be present within a five-
mile radius of the reactor facility, these
chemicals should be considered in the
evaluation of control room habitability.
Whether a major depot or storage area
constitutes a hazard is determined on the
basis of the quantity of stored chemicals, the distance from the nuclear plant, the inleakage characteristics of the control
room, and the applicable toxicity limits (see Regulatory Position 4 for
definition). Table C-2 gives the criteria
to be used in evaluating the hazards of
chemicals to control rooms. A procedure
for adjusting the quantities given to
Table C-2 to appropriately account for the
toxicity limit of a specific chemical, meteorology conditions of a particular
site, and air exchange rate of a control
room is present in Appendix A of this
guide. Rev. 0 WOLF CREEK TABLE 6.4-1 (Sheet 2)
Regulatory Guide WCGS Position 1.78 Position
Chemical stored or situated at distances
greater than five miles from the facility
need not be considered because, if a release occurs at such a distance, atmospheric dispersion will dilute and
disperse the incoming plume to such a
degree that there should be sufficient
time for the control room operators to
take appropriate action. In addition, the
probability of a plume remaining within a
given sector for a long period of time is
quite small.
- 2. If hazardous chemicals such as 2. See Section 2.2.
those indicated in Table 1 are known or projected to be frequently shipped by rail, water, or road routes within a five-
mile radius of a nuclear power plant, estimates of these shipments should be
considered in the evaluation of control room habitability. The weight limits of
Table C-2 (adjusted for the appropriate
toxicity limit, meteorology, and control
room air exchange rate) apply also to
frequently shipped quantities of hazardous
chemicals. Shipments are defined as being
frequent if there are 10 per year for
truck traffic, 30 per year for rail
traffic, or 50 per year for barge
traffic. If the quantity, per shipment, of hazardous chemicals frequently shipped past a site is less than the adjusted
quantity shown on Appendix A for the control room type being evaluated, the shipments need not be considered in the
analysis.
Rev. 23
WOLF CREEK TABLE 6.4-1 (Sheet 3)
Regulatory Guide WCGS Position 1.78 Position
- 3. In the evaluation of control room 3. See Section 2.2.
habitability during normal operation, the
release of any hazardous chemical to be stored on the nuclear plant site in a
quantity greater than 100 pounds should be
considered. Any hazardous chemical stored
onsite should be accompanied by instrumen-
tation that will detect its escape, set
off an alarm, and provide a readout in the
control room.
- 4. The toxicity limits should be taken 4. See Section 2.2.
from appropriate authoritative sources, such as those listed in the References section. For each chemical considered, the values of importance are the human
detection threshold and the maximum
concentration that can be tolerated for
two minutes without physical incapaci-
tation of an average human (i.e., severe
coughing, eye burn, or severe skin
irritation). The latter concentration is
considered the "toxicity limit." Table 1 gives the toxicity limits (in ppm by volume and mg/m
- 3) for the chemicals listed. Where these data are not available, a determination of the values
to be used is made on a case-by-case
basis.
- 5. Two types of industrial accidents 5. See Section 2.2.
should be considered for each source of
hazardous chemicals; maximum concentration
chemical accidents and maximum
concentration-duration chemical accidents.
Rev. 23
WOLF CREEK TABLE 6.4-1 (Sheet 4)
Regulatory Guide WCGS Position 1.78 Position
- a. For a maximum concentration accident, the quantity of the hazardous chemical to be
considered is the instantaneous release of the total contents of one of the following:
(1) The largest storage container falling
within the guidelines of Appendix A and located at a nearby stationary facility, (2) the largest shipping container (or for
multiple containers of equal size, the
failure of only one container unless the
failure of that container could lead to successive failures) falling within the
guidelines of Appendix A and frequently transported near the site, or (3) the largest container stored onsite (normally the total
release from this container unless the
containers are interconnected in such a
manner that a single failure could cause a release from several containers.)
For chemicals that are not gases at 100 F and
normal atmospheric pressure but are liquids
with vapor pressures in excess of 10 torr, consideration should be given to the rate of
flashing and boiloff to determine the rate of
release to the atmosphere and the appropriate
time duration of the release.
The atmospheric diffusion model to be used in the evaluation should be the same as or
similar to the model presented in Appendix B
of the guide.
- b. For a maximum concentration-duration
accident, the continuous release of hazardous
chemicals from the largest
Rev. 23
WOLF CREEK TABLE 6.4-1 (Sheet 5)
Regulatory Guide WCGS Position 1.78 Position
safety relief valve on a stationary, mobile, or onsite source falling within
the guidelines of Appendix A should be considered. Guidance on the atmospheric
diffusion model is presented in Regulatory
Guide 1.3, "Assumptions used for
Evaluating the Potential Radiological
Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors," and
Regulatory Guide 1.4, "Assumptions Used
for Evaluating the Potential Radiological
Consequences of a Loss-of-Coolant Accident
for Pressurized Water Reactors."
- 6. The value of the atmospheric dilution 6. See Section 2.2.
factor between the release point and the
control room that is used in the analysis
should be that value that is exceeded only 5 percent of the time.
When boiloff or a slow leak is analyzed, the effects of density on vertical
diffusion may be considered if adequately
substantiated by reference to data from
experiments. Density effect to heavier-
than-air gases should not be considered
for releases of a violent nature or for
material that becomes entrained in the
turbulent air near buildings.
- 7. For both types of accidents described 7. See Section 2.2 for those hazardous
in Regulatory Position 5 above, the capa- chemicals stored onsite.
bility of closing the air ducts of the control room with dampers and thus isolating the control room should be considered in the
evaluation of control room habitability.
Rev. 23
WOLF CREEK TABLE 6.4-1 (Sheet 6)
Regulatory Guide WCGS Position 1.78 Position In particular, the time required to shut off or redirect the intake flow should be
justified. The detection mechanism for each hazardous chemical should be considered. Human detection may be
appropriate if the buildup of the
hazardous chemical in the control room is
at a slow rate due to slow air turnover.
The air flows for infiltration, makeup, and recirculation should be considered for
both normal and accident conditions. The
volume of the control room and all other
rooms that share the same ventilating air, during both normal conditions and accident
conditions, should be considered. The
time required for buildup of a hazardous chemical from the detection concentration
to the toxicity limit should be
considered. Table C-3 of the guide
contains a sample list of the chemical and
control room data needed for the
evaluation of control room habitability.
- 8. In the calculation of the rate of air 8. See below.
infiltration (air leaking into the control
room from ducts, doors, or other openings)
with the control room isolated and not
pressurized, use of the following
assumptions is suggested:
- a. A pressure differential of 1/8-inch a. Complies.
water gauge across all leak paths.
- b. The maximum design pressure differential b. Not applicable. Control room isolation for fresh air dampers on the suction side of isolates all systems and stops all fans which
recirculation fans. penetrate the control room boundary.
Rev. 0 WOLF CREEK TABLE 6.4-1 (Sheet 7)
Regulatory Guide WCGS Position 1.78 Position
- 9. When the makeup air flow rate required 9. Complies.
to pressurize the control room is calcu-
lated, a positive pressure differential of 1/4-inch water gauge should be assumed in the control room relative to the space
surrounding the control room.
- 10. To account for the possible increase 10. Complies.
in air exchange due to ingress or egress, an additional 10 cfm of unfiltered air should
be assumed for those control rooms without
airlocks. This additional leakage should be
assumed whether or not the control room is
pressurized.
- 11. If credit is taken in the evaluation 11. Complies. No credit is taken for for the removal of hazardous chemicals by removal.
filtration or other means, the experimental
basis for the dynamic removal capability
of the removal system for the particular
chemical being considered should be
established.
- 12. Concurrent chemical release of container 12. See Section 2.2.
contents during an earthquake, tornado, or
flood should be considered for chemical
container facilities that are not designed
to withstand these natural events. It may
also be appropriate to consider release
from a single onsite container or pipe
coincident with the radiological
consequences of a design basis loss-of-
coolant accident, if the container facilities are not designed to withstand
an earthquake.
Rev. 0 WOLF CREEK TABLE 6.4-1 (Sheet 8)
Regulatory Guide WCGS Position 1.78 Position
- 13. If consideration of possible accident 13. Complies per the requirements of for any hazardous chemical indicates that Regulatory Guide 1.95. See Emergency Plans.
the applicable toxicity limits may be ex-ceeded, self-contained breathing apparatus There are six self-contained breathing appara-of at least one-half hour capacity or a tus stored in the Control Room. An eight-
tank source of air with manifold outlets hour supply of air, which includes spare
and protective clothing, if required, bottles for changeout, is available. Standard
should be provided for each operator in annual respiratory protection training is the control room. Additional air capacity required and the equipment is inventoried and with appropriate equipment should be cleaned on a monthly basis. It is estimated
provided if a chemical hazard can persist that the length of time for deploying and don-
longer than one-half hour. For accidents ning the equipment is two minutes.
of long duration, sufficient air for six hours (coupled with provisions for obtaining
additional air within this time period) is adequate. Each operator should be taught to distinguish the smells of hazardous
chemicals peculiar to the air. Instruction
should include a periodic refresher course.
Practice drills should be conducted to ensure that personnel can don breathing apparatus within two minutes.
- 14. Detection instrumentation, isolation 14. The single failure criterion is met, as systems, filtration equipment, air supply described in Sections 2.2, 6.4 and 9.4.1.
equipment, and protective clothing should
meet the single-failure criterion. (In the
case of self-contained breathing apparatus and protective clothing, this may be accom-
plished by supplying one extra unit for
every three units required.)
Rev. 6 WOLF CREEK TABLE 6.4-1 (Sheet 9)
Regulatory Guide WCGS Position 1.78 Position
- 15. Emergency procedures to be initiated 15. See Section 2.2.
in the event of a hazardous chemical release
within or near the station should be Emergency supplies are maintained onsite to written. These procedures should address take care of the needs of 25 individuals for both maximum concentration accidents, and one week. These supplies consist of dehy-
maximum concentration-duration accidents drated, frozen and/or canned food. Sealed
and should identify the most probably containers of potable water are also stored.
chemical releases at the station. Methods A standard 36-unit first aid kit plus an of detecting the event by station emergency stretcher is available.
personnel, both during normal workday
operation and during minimum staffing periods (late night and weekend shift staffing),
should be discussed. Special instrumentation
that has been provided for the detection of
hazardous chemical releases should be
described, including sensitivity, action initiated by detecting instrument, level at
which this action is initiated, and Technical
Specification limitations on instrument
availability. Criteria should be defined for
the isolation of the control room, for the
use of protective breathing apparatus or
other protective measures, and for orderly
shutdown or scram. Criteria and procedures
for evacuating nonessential personnel from
the station should also be defined.
Arrangement should be made with federal, state, and local agencies or other cognizant
organizations for the prompt notification of
the nuclear power plant when accidents
involving hazardous chemicals have occurred
within five miles of the plant.
Rev. 0 WOLFCREEKTABLE6.4-2COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95PositionControlroomoperatorsshouldbeprotectedagainsttheeffectsofanaccidental chlorinereleaseasdescribedbelow.1.Liquifiedchlorineshouldnotbestored1.SeeSection2.2.within100metersofacontrolroomorits freshairinlets.(Smallquantitiesfor laboratoryuse,20poundsorless,areexempt.)2.Ifachlorinecontainerhavinganinven-2.Thecapabilityforremotemanualtoryof150poundsorlessisstoredmorethanisolationisprovidedatthecontrolroom.100metersfromthecontrolroomoritsfreshairinlets,thecapabilityformanualisolationofthecontrolroomshouldbeprovided.3.Forsinglecontainerquantitiesexceeding3.Notapplicable.150pounds,themaximumallowablechlorineinventoryinasinglecontainerstoredatspecifieddistancesfromthecontrolroomor itsfreshairinletisgiveninTable1forcontrolroomTypesIthroughVI(describedbelow).Foreachcontrolroomtype,the maximumallowablechlorineinventoryina singlecontainerisgivenasafunctionof distancefromthecontrolroom.Ifthereareseveralchlorinecontainers,onlythefailureofthelargestcontainerisnormally consideredunlessthecontainersareinter-connectedinsuchamannerthatfailureofa singlecontainercouldcauseachlorine releasefromseveralcontainers.Rev.8 WOLFCREEKTABLE6.4-2(Sheet2)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positiona.TypeIcontrolroomsshouldincludethefollowingprotectivefeatures:(1)Quick-responsechlorinedetectorslocated(1)Notapplicable.inthefreshairinlets.Within10secondsafterarrivalofthechlorine,detectionshouldinitiatecompleteclosureofiso-lationdamperstothecontrolroom.(2)Anormalfreshairmakeuprateofless(2)Complies.thanoneairchangeperhour.Thefresh airinletshouldbeatleast15metersabove grade.(3)Low-leakageconstructionwithanequi-(3)Leakagecriteriacompliesfor1/8-inchvalentairexchangerateoflessthan0.06differential.However,theWCGSdesign hr-1whenallpenetrationsareexposedtoautilizes1/4-inch(minimum)differential.
1/8-inchwatergagepressuredifferential.Applicableconstructiondetailsareshown ConstructiondetailsshouldbeprovidedtoinFigure6.4-1.
showthatthislimitismet.(4)Low-leakagedampersorvalvesinstalled(4)Notapplicable.ontheupstreamsideofrecirculationfansorotherlocationswherenegativesystemspressureexistsandwhereinleakagefromchlorine-contaminatedoutsideairis possible.Rev.8 WOLFCREEKTABLE6.4-2(Sheet3)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionb.TypeIIcontrolroomsshouldincludetheb.Notapplicable.protectivefeaturesofParagrapha,except thattheisolationtimeshouldbe4secondsorlessratherthan10secondsorless.c.TypeIIIcontrolroomsshouldincludethec.Notapplicable.protectivefeaturesofParagrapha,except thatthenormalfreshairmakeuprateshould belimitedto0.3airchangeperhourorless.d.TypeIVcontrolroomsshouldincludethed.Notapplicable.protectivefeaturesofParagrapha,except thattheisolationtimeandthenormalair exchangerateshouldbeequaltoorlessthan 4secondsand0.3airchangeperhour, respectively.Inaddition,thecontrolroomisolatedairexchangerateshouldbereduced to0.015airchangeperhourorless(see descriptionofrequiredleakrate verificationtestinRegulatoryPosition5).e.TypeVcontrolroomsshouldincludethee.Notapplicable.protectivefeaturesofParagrapha,withthe additionofremotechlorinedetectorslocated atthechlorinestorageandunloadingloca-tion.Theseadditionaldetectorsshouldbe placedandthedetectortrippointsadjusted soastoensuredetectionofeitheraleakor acontainerrupture.Adetectortripsignal shouldaccomplishautomaticisolationofthe controlroombeforechlorinearrivesatthe isolationdampers.Thedetectortripsignal shouldalsosetoffanalarmandprovideareadoutinthecontrolroom.Rev.0 WOLFCREEKTABLE6.4-2(Sheet4)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95PositionAnalternativetotheinstallationofremotedetectorswouldbetoprovideanisolation systemusinglocaldetectorsbuthavinganisolationtimeofeffectivelyzero.Thiscanbeaccomplishedbyensuringthatthetime requiredforchlorinetotravelfromthe chlorinedetectortotheisolationdamper, withintheinletducting,isequaltoorgreaterthanthetimerequiredtodetectthechlorineandclosetheisolationdamper.f.TypeVIcontrolroomsshouldincludethef.Notapplicable.protectivefeaturesinParagraphe,except thatthecontrolroomisolatedairexchange rateshouldbereducedto0.015airchange perhourorless.Forisolatedexchangeratesbetween0.015hr-1and0.06hr-1, linearinterpolationoftheweightsgivenfor controlroomTypesVandVIinTable1canbe made.Verificationtestingisrequired withinthisrangeofexchangerates(see RegulatoryPosition5).4.Thefollowingshouldbeappliedtoall4.Seebelow.controlroomtypes(IthroughVI):a.Immediatelyaftercontrolroomisolation,a.Complies.Initiationsofcontrolroomtheemergencyrecirculatingcharcoalfilterisolationalsoinitiateoperationofcontrol orequivalentequipmentdesignedtoremoveorroomfiltrationsystem.
otherwiselimittheaccumulationof chlorinewithinthecontrolroomshouldbe startedupandoperated.Rev.0 WOLFCREEKTABLE6.4-2(Sheet5)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionb.Stepsshouldbetakentoensurethattheb.Seebelow.isolatedexchangerateisnotinadvertently increasedbydesignoroperatingerror.Forinstance,thefollowingshouldbeconsidered:(1)Anadministrativeprocedureshould(1)Aftertransfertotheemergencymode,requirethatalldoorsleadingtothecontroladministrativecontrolsrequireclosureofroombekeptclosedwhennotinuse.anyopendoorleadingtothecontrolroomwhennotinuse.(2)Ventilationequipmentforthecontrol(2)Complies.Automaticisolationoftheroomandfortheadjacentzonesshouldbecontrolroomalsoautomaticallystopsall reviewedtoensurethatenhancedairexchangefansandisolatesallsystemswhichpenetratebetweentheisolatedcontrolroomandthethecontrolbuildingboundary.outsidewillnotoccur(e.g.,ifthereisa chlorinerelease,exhaustfansshouldbe stoppedand/orisolatedfromthecontrol roomventilationzonebylow-leakagedampersorvalves).(3)Acontrolroomexitleadingdirectlyto(3)Notapplicable.theoutsideofthebuildingshouldhavetwo low-leakagedoorsinseries.c.Theuseoffull-faceself-containedc.Complies.Full-face,self-containedpressure-demand-typebreathingapparatus(orpressure-demand-typebreathingapparatustheequivalent)andtheuseofprotectiveareprovided.
clothingshouldbeconsideredinthedevelop-mentofachlorinereleaseemergencyplan.
Becausecalculationsindicatethatchlorine concentrationsmayincreaserapidly, emergencyplanprovisionsandrehearsalof emergencyplanprovisionsarenecessaryto ensuredonningofbreathingapparatuson detectionofhighchlorineconcentrations.Rev.6 WOLFCREEKTABLE6.4-2(Sheet6)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95PositionStorageprovisionsforbreathingapparatusandproceduresfortheiruseshouldbesuchthat operatorscanbeginusingtheapparatuswithintwominutesafteranalarm.Adequateaircapacityforthebreathingapparatus(at leastsixhours)shouldbereadilyavailable onsitetoensurethatsufficienttimeis availabletotransportadditionalbottledairfromoffsitelocations.Thisoffsitesupplyshouldbecapableofdeliveringseveral hundredhoursofbottledairtomembersof theemergencycrew.Aminimumemergencycrew shouldconsistofthosepersonnelrequiredto maintaintheplantinasafecondition, includingorderlyshutdownorscramofthe reactor.Asaguideline,aminimumoffiveunitsofbreathingapparatusshouldbe providedfortheemergencycrew.d.Theairsupplyapparatusshouldmeetthed.Complies.Self-containedbreathingsinglefailurecriterionandbedesignatedapparatusaresuppliedtomeetthe SeismicCategoryI.(Inthecaseofself-singlefailurecriterion,asspecified containedbreathingapparatus,thesingleinRegulatoryGuide1.95.
failurecriterionmaybemetbysupplying oneextraunitforeverythreeunitsrequired.)TheisolationsystemcomponentsshouldbeComplies.ofaqualitythatensureshighreliability andavailability.Onemethodtomeetthese goalsistoprovideasystemthatmeetsthe requirementsofIEEE-279,"Criteriafor ProtectionSystemsforNuclearPower GeneratingStations."Inallcases,theRev.0 WOLFCREEKTABLE6.4-2(Sheet7)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionisolationsystem,recirculatingfiltersystem,andairconditioningsystemshouldmeetIEEE-279sincetheyarerequiredtomaintainahabitableenvironmentinthecontrolroomduringdesignbasisradiologicalevents.Specificacceptancecriteriaforthechlorinedetectionsystemandalliedactuating electronicsareasfollows:(1)ChlorineConcentrationLevel.Detectors(1)Notapplicable.shouldbeabletodetectandsignalachlorineconcentrationof5ppm.(2)SystemResponseTime.Thesystem(2)Notapplicableresponsetime,whichincorporatesthedetectorresponsetime,thevalveclosuretime,andassociatedinstrumentdelays,shouldbeless thanorequaltotherequiredisolationtime.(3)SingleFailureCriteria.Thechlorine(3)Notapplicable.detectionsystemshouldberedundantandphysicallyseparatetoaccomplishdecoupling oftheeffectsofunsafeenvironmentalfactors,electrictransients,physicalaccident,andcomponentfailure.Localdetectorsshouldconsistoftwophysi-callyseparatechannelsforeachfreshair inlet.Eachchannelshouldconsistofa separatepowersupply,detector,actuating electronics,andinterconnectingcabling.
RemotedetectorsshouldalsoconsistoftwoRev.8 WOLFCREEKTABLE6.4-2(Sheet8)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionseparatechannelshavingdetectorslocatedatthechlorineunloadingfacility.(4)SeismicQualification.Thechlorine(4)Notapplicable.detectionsystemshouldbedesignatedas SeismicCategoryIandbequalifiedassuch.(5)EnvironmentalQualification.The(5)Notapplicable.detectionsystemshouldbequalifiedforall expectedenvironmentsandforsevereenvi-ronmentsthatcouldclearlyleadtoorbe aresultofchlorinerelease.Theinstal-lationshouldensurethattheyareprotected fromadversetemperatureeffects.(6)Maintenance,Testing,andCalibration.The(6)Notapplicablemanufacturer'smaintenancerecommendationsare acceptableprovidedtheyfollowsoundengineer-ingpracticeandarecompatiblewiththeproposed application.Aroutineoperationalcheckshould beconductedatone-weekintervals.Verificationtestingandcalibrationofthechlorinedetectorsandverificationtesting ofthesystemresponsetimeshouldbe conductedatsix-monthintervals.5.Thegrossleakagecharacteristicofthe5.Complies.Pressurizationflowratecontrolroomshouldbedeterminedbypres-is400cfm.Theairexchangerateis surizingthecontrolroomto1/8-inchwatergreaterthan0.06perhour.Therefore, gageanddeterminingthepressurizationperiodictestingisnotrequired.flowrate.(TheuseofahigherpressuredifferentialisacceptableprovidedtheflowRev.13 WOLFCREEKTABLE6.4-2(Sheet9)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionrateisconservativelyadjustedtocorrespondto1/8-inchwatergage).Forairexchange ratesoflessthan0.06hr-1,periodicverifi-cationtestingshouldbeperformed.Anaccept-ablemethodforperiodictestingwouldbethe useofapermanentlyinstalledcalibrated pressurizationfan.Thesystemwouldhavea knownpressure-versus-flowcharacteristicsothattheleakratecouldbedeterminedbymeasuringthecontrolroompressure
differential.Testingshouldbeconductedatleasteverysixmonthsandafteranymajoralterationthatmay affectthecontrolroomleakage.6.Emergencyprocedurestobeinitiated6.SeeChapter13.0intheeventofachlorinereleaseshouldbe provided.Methodsofdetectingtheeventby stationpersonnel,bothduringnormalworkday operationandduringminimumstaffingperiods (latenightandweekendshiftstaffing),should bediscussed.Instrumentationthathasbeen providedforthedetectionofchlorineshould bedescribedincludingsensitivity;action initiatedbydetectinginstrumentandlevelat whichthisactionisinitiated;technicalspe-cificationlimitationsoninstrumentavailabil-ity;andinstructionsformaintenance,calibra-tion,andtesting.Criteriashouldbedefined fortheisolationofthecontrolroom,forthe useofprotectivebreathingapparatusandRev.0 WOLFCREEKTABLE6.4-2(Sheet10)COMPARISONOFTHEDESIGNTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.95,REVISION1,DATEDJANUARY1977,TITLED"PROTECTIONOFNUCLEARPOWERPLANTCONTROLROOMOPERATORSAGAINSTANACCIDENTALCHLORINERELEASE"RegulatoryGuideWCGSPosition1.95Positionotherprotectivemeasures,andformaintenanceoftheplantinasafeconditionincluding thecapabilityfororderlyshutdownorscramofthereactor.Criteriaandproceduresforevacuatingnonessentialpersonnelfromthe stationshouldalsobedefined.Rev.0 Note:Othertypicalpenetrationsealswere alsoutilized(
Reference:
M-663-00017).
19 Note:Otherductworktypicaldetailswere alsoutilized(
Reference:
M-OH1904and M-OH1905).19 Note:Othertypicalpenetration sealswerealsoutilized
(
Reference:
M-663-00017).
19 WOLF CREEK 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS Several plant features serve to reduce or limit the release of fission products following a postulated LOCA or fuel handling accident. This section provides a discussion of the function of the containment, containment spray system, and emergency filter systems to mitigate the consequences of an accident. The design of each of these engineered safety features is discussed in other referenced sections. Chapter 15.0 addresses the radiological consequences of postulated accidents and demonstrates the adequacy of the fission product
removal and control systems.
Other sections provide the design bases and safety evaluations, which demonstrate that the design and construction of these systems is commensurate with acceptable practices for engineered safety features. This includes, but is not limited to, assuring redundancy, isolation from nonsafety-related portions, seismic classification, compliance with Regulatory Guide 1.52, suitability of material for the intended service, Class IE power supply from
onsite or offsite sources, qualification testing, and the capability for
inspection and testing.
6.5.1 ENGINEERED SAFETY FEATURE (ESF) FILTER SYSTEMS The ESF filter systems include the emergency exhaust system, discussed in Sections 9.4.2 and 9.4.3, and the control building HVAC systems, discussed in Sections 6.4 and 9.4.1. The emergency exhaust system would operate following a LOCA to control and reduce fission product releases from the auxiliary building. It also would operate after a fuel handling accident to control and
reduce fission product releases from the fuel building (see Section 9.4.2).
The control building HVAC systems operate to maintain control room habitability
by removing fission products from air entering the control room (see Section
6.4). This section discusses the design basis and safety evaluation of the functional requirements of the ESF filter systems.
6.5.1.1 Design Basis 6.5.1.1.1 Safety Design Basis
SAFETY DESIGN BASIS ONE - An emergency exhaust system is provided to reduce the fission product release from the plant, following a fuel handling accident in
the fuel building or a LOCA that could potentially result in radioactive
leakage into the auxiliary building. 6.5-1 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS TWO - A control building HVAC system is provided to isolate the control building and provide the control room with a filtered supply of
fresh air.
6.5.1.1.2 Power Generation Design Basis The ESF filter systems have no power generation design basis.
6.5.1.2 System Design 6.5.1.2.1 General Description
The emergency exhaust system is shown in Figure 9.4-2, and the control building HVAC system is shown in Figure 9.4-1. A detailed description of these systems is provided in Sections 9.4.1, 9.4.2, and 9.4.3.
The ESF filter systems comply with Regulatory Guide 1.52, as discussed in Table 9.4-2.Table 6.5-1 lists the system design parameters used in the radiological consequences analysis presented in Chapter 15.0.
6.5.1.2.2 Component Description
The emergency exhaust system components are described in Sections 9.4.2 and 9.4.3. The control room HVAC system components are described in Section 9.4.1.
6.5.1.2.3 System Operation
In the event of a LOCA, the emergency exhaust system functions to limit and reduce the potential release of fission products from the auxiliary building.
Specific details of system operation following a LOCA are provided in Section
9.4.3.In the event of a fuel handling accident in the fuel building, the emergency exhaust system functions to reduce the fission product release from the fuel
building. Specific details of system operation following a fuel handling
accident are provided in Section 9.4.2.
In the event of a LOCA or fuel handling accident, the control building HVAC systems function to isolate the control building and provide the control room
with a filtered supply of air. Specific details of system operation following
a LOCA are discussed in Section 9.4.1. 6.5-2 Rev. 13 WOLF CREEK 6.5.1.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases given in Section 6.5.1.1.1.
SAFETY EVALUATION ONE - Table 6.5-1 lists the ESF filtration systems' design parameters used to determine the radiological consequences for the postulated accidents analyzed in Chapter 15.0. The results of these analyses demonstrate that the emergency exhaust system reduces and controls fission products released from the fuel building following a fuel handling accident or released from the auxiliary building following a LOCA, such that the offsite radiation exposures are within the guidelines of 10 CFR 100. The safety evaluations
which demonstrate the design and construction of the ESF filtration systems are
provided in Sections 9.4.2 and 9.4.3.
SAFETY EVALUATION TWO - The results of the analyses described in Chapter 15.0 demonstrate that the control building HVAC systems reduce and control fission product release to the control room following a LOCA, such that the offsite
radiation exposures are within the guidelines of 10 CFR 100. The safety
evaluations which demonstrate the design and construction of these control
building HVAC systems are provided in Sections 9.4.1 and 6.4.
6.5.1.4 Tests and Inspections Tests and inspections for ESF filter systems are described in Section 9.4.
6.5.1.5 Instrumentation Requirements Instrumentation and controls are provided to facilitate automatic operation and remote control of the system and to provide continuous indication of system
parameters. Further descriptions are provided in Section 9.4.
6.5.1.6 Materials The materials used for ESF filtration systems were chosen considering the environmental conditions and are commensurate with acceptable construction practices. Further information is provided in Section 9.4.
6.5.2 CONTAINMENT SPRAY SYSTEM The containment spray system (CSS) is an ESF, the functions of which are to reduce pressure and temperature in the containment atmosphere following a postulated LOCA and to remove radioactive 6.5-3 Rev. 0 WOLF CREEK fission products from the containment atmosphere. These functions are performed by spraying water into the containment atmosphere through a large
number of nozzles on spray headers located in the containment dome. Reduction of pressure and temperature in the containment with the CSS is discussed in Section 6.2.2.
Radioiodine in its various forms is the fission product of primary concern in the evaluation of a LOCA. It is absorbed by the containment spray from the containment atmosphere. To enhance this iodine absorption capacity of the
spray, the spray solution is adjusted to an alkaline pH which promotes iodine
hydrolysis, in which iodine is converted to nonvolatile forms.
The physical characteristics of the CSS are discussed in Section 6.2.2.
Discussed herein are the spray additive portion of the system and the
containment spray system's fission product removal capability following a LOCA.
6.5.2.1 Design Bases 6.5.2.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The CSS is designed to provide a spray solution while the spray additive portion of the system is in operation in the pH range of 9.0 to 11.0 and a final containment recirculation sump solution with a pH of at least 8.5.
SAFETY DESIGN BASIS TWO - The CSS is capable of reducing the iodine and particulate fission product inventories in the containment atmosphere such that the offsite radiation exposures resulting from a design basis LOCA are within the plant siting dose guidelines of 10 CFR 100.
Additional safety design bases are included in Section 6.2.2, in which the capability of the spray system to remove heat from the containment atmosphere is discussed.
6.5.2.1.2 Power Generation Design Basis The CSS has no power generation design basis.
6.5.2.2 System Design 6.5.2.2.1 General Description The containment spray additive portion of the CSS provides for eduction of 30 weight percent (nominal) sodium hydroxide into the spray injection water. This
yields a spray mixture with a pH of from 9.0 to 11.0 during the initial period
of operation, when radioiodine is being removed from the containment
atmosphere. 6.5-4 Rev. 3 WOLF CREEK The spray additive subsystem of the CSS, shown schematically in Figure 6.2.2-1, consists of one spray additive tank, two eductors, valves, and connecting
piping. The system uses the containment spray pumps and spray headers, as described in Section 6.2.2.1, to deliver and distribute the spray additive solution to the containment atmosphere. Initially, water from the refueling
water storage tank (RWST) is used for containment spraying followed by water recirculated from the containment sump. Sodium hydroxide is educted from the spray additive tank into the water from the RWST and containment sump and
pumped to the spray ring headers and nozzles.
Parts of the system in contact with borated water or the sodium hydroxide spray additive, or mixtures of the two, are stainless steel or an equivalent corrosion-resistant material.
The stainless steel spray additive tank contains sufficient 30 weight percent (nominal) sodium hydroxide spray additive solution to bring the containment sump fluid to a minimum pH of 8.5 upon mixing with the borated water from the refueling water storage tank, the boron injection tank, the accumulators, and reactor coolant. This assures continued iodine retention effectiveness of the containment sump water during the recirculation phase.
The two spray additive eductors are 3-inch mixing eductors. The units draw the 30 weight percent (nominal) sodium hydroxide spray additive solution into their suction by using borated water discharged by the containment spray pumps as their motive flow.
The spray header design, including the number of nozzles per header, nozzle spacing, and nozzle orientation, is provided in Section 6.2.2.1 and shown in Figures 6.2.2-2 and 6.2.2-4. Each spray header layout is oriented to provide
more than 90-percent area coverage at the operating deck of the reactor
building.Total containment free volume, unsprayed containment free volume, specific unsprayed regions and volumes, and post-accident ventilation between sprayed
and unsprayed volumes are provided in Table 6.5-2. Operability of dampers, ductwork, etc., for which credit is taken postaccident is discussed in Section
6.2.2.2.2.3.
6.5.2.2.2 Component Description The mechanical components of the spray additive subsystem are described in this section. Other components in the containment spray system are described in
Section 6.2.2.1. Spray additive subsystem component design parameters are
given in Table 6.5-3. 6.5-5 Rev. 13 WOLF CREEK The containment spray additive tank, located at El. 2,000 in the auxiliary building, is a stainless steel tank with a nitrogen gas blanket designed to
contain 30 percent by weight (nominal) sodium hydroxide solution. The capacity of the tank is given in Table 6.5-3. A local sample connection allows periodic chemical analysis of the contents, and fill and drain connections provide for
initial fill, concentration adjustments, and maintenance. A manway is also provided for tank internal inspection. Tank level, pressure indication, and alarm instrumentation are provided.
An interlock is provided from the tank level transmitters to preclude closure of the discharge valves before sufficient NaOH has been added to the spray
solution to comply with the sump pH criterion. Heat tracing of the spray additive tank and associated piping containing 30 weight percent (nominal) NaOH is not required since the auxiliary building is heated to maintain temperatures greater than 60 F. The containment spray additive tank is provided with
overpressure protection and vacuum relief. Setpoints of the relief devices are
provided in Table 6.5-3.
Sodium hydroxide is added to the spray liquid by a liquid jet eductor, a device which uses kinetic energy of a pressurized liquid to entrain another liquid, mix the two, and discharge the mixture against a counter pressure. The
pressurized liquid in this case is the spray pump discharge which is used to
entrain the sodium hydroxide solution and discharge the mixture into the
suction of the spray pumps. The eductors are designed to assure a minimum pH
of 9.0 for the spray mixture.
Component descriptions of the nozzles are provided in Section 6.2.2.1. Special tests performed on the spray nozzle include capacity and droplet size
distribution. Figures 6.5-1, 6.5-2, and 6.5-3 provide the test results for the
spray nozzles (Ref. 1).
The spray nozzle was flow tested at a range of inlet pressures from 3 to 100 psig to determine that the actual flow at 40 psi differential across the nozzle
was in accordance with the design value of 15.2 gpm, as depicted in Figure 6.5-
1.Droplet-size distribution measurements were performed at the design pressure of 40 psi and the design flowrate of 15.2 gpm. At these conditions, the spray
distribution was obtained by measuring the spray volume distribution in two
perpendicular planes over a timed interval (Ref. 1). 6.5-6 Rev. 3 WOLF CREEK For the droplet size distribution measurement, a television camera and light source were mounted on a flat beam. A protective covering was constructed with
a slot which allowed spray droplets to fall between the camera and light source. Measurements of drop count in each micron increment were recorded at 4-inch increments from the outer edge of the spray cone to the spray axis.
At the design pressure, the droplet size distribution was recorded by high speed photographic methods. The droplet images were measured, and droplets with a diameter in the micron increment being counted were registered. Figure
6.5-2 shows the relative frequency for each droplet size. The results of
testing performed on the spray nozzle are provided in Table 6.5-2. The containment spray envelope reduction factor as a function of post-LOCA containment saturation temperature is provided in Figure 6.5-4. This envelope reduction factor was applied to the throw distance and elliptic coverage values
presented in Table 6.5-2.
6.5.2.2.3 System Operation
Summary of the design basis LOCA chronology for the CSS is presented in Table 6.2.2-3.The method of switchover of the WCGS ECCS systems and the containment spray system from injection to recirculation evolved from a totally manual design to
one of limited and reasonable operator action during the PSAR stage. The
current design provides for the automatic switchover to the recirculation mode for the RHR pumps followed by manual realignment of the containment spray pumps, the centrifugal charging pumps, and the high head safety injection
pumps. The necessary indications and the sequence of events for the switchover
of the containment spray system are described in Section 6.2.2.1.2.3. The
sequence of events for the switchover of the ECCS systems is described in USAR
Section 6.3.2.8.
Table 6.2.2-4 provides the minimum duration of containment spray flow in the injection mode for the various assumed flow conditions. These durations provide an adequate time frame for the necessary operator actions required for
ECCS and containment spray system management.
The current design was found to be acceptable by the NRC at the PSAR stage and is the basis, in that regard, for the issuance of the WCGS Construction permit.
The thermal-hydraulic analyses provided in Section 6.2.1 and the radiological consequences of the accidents analyzed in Chapter 15.0 demonstrate the adequacy
of the existing containment spray system. 6.5-7 Rev. 0 WOLF CREEK The spray system is actuated either manually from the control room or on coincidence of two-out-of-four CSAS containment pressure signals. This signal
starts the containment spray pumps, opens the discharge valves to the spray headers, and opens the valves associated with the spray additive tank.
On actuation, approximately 5 percent of each spray pump discharge flow is diverted through each spray additive eductor to draw sodium hydroxide from the spray additive tank. The sodium hydroxide solution mixes with the liquid
entering the suction line of the pumps to give a solution suitable for the
removal of iodine from the containment atmosphere.
When the refueling water storage tank has reached its specified low-low level limit, recirculation spray flow is manually initiated. The operator can remotely initiate recirculation flow by use of either or both of the spray pumps. Sections 6.2.2.1.5 and 6.5.2.5 address the instrumentation and
information displays available to the operator, in order for manual switchover
of the CSS to take place.
System flow rates and the duration of operational modes are presented in Section 6.2.2.1.2.3.
Design operation of the CSS and the containment spray additive subsystem is such that LOCA iodine removal requirements are fulfilled during the injection phase and the amount of NaOH added is sufficient to ensure long-term iodine retention. Operation of the containment spray additive subsystem can be remote-manually terminated following the eduction of the prescribed quantity of NaOH which assures a minimum long-term sump pH of at least 8.5. Automatic isolation of the containment spray additive subsystem occurs upon receipt of a low-low level signal from the spray additive tank level instruments. The containment iodine removal credit assumed in the calculations of offsite doses, following a LOCA, is provided in Chapter 15.0.
6.5.2.3 Safety Evaluation The safety evaluations are numbered to correspond to the safety design bases.
SAFETY EVALUATION ONE - The system's capability to reduce the airborne fission product inventory is based on the pH of the spray solution for removal during injection and for retention during recirculation, and on the system's capability to provide spray for essentially all regions of the containment, considering postaccident conditions. 6.5-8 Rev. 13 WOLF CREEK The design minimum spray water pH of 9.0 coupled with the dependent parameters identified in Safety Evaluation Two below, assure the minimum elemental iodine
removal coefficient of 10.0 per hour during the spray injection phase. The design minimum sump pH of 8.5 assures long-term iodine retention in the recirculated spray liquid.
The maximum pH of the spray solution in the CSS during the spray injection phase is 11.0, based on the maximum allowable eductor sodium hydroxide flow rate and minimum boric acid concentration in the RWST.
The system is designed to provide a spray solution in the CSS during the spray recirculation phase with a maximum spray pH of less than 11.0 based on a sump pH of at least 8.5 (due to prior addition of NaOH), design spray recirculation flow rate, as noted in Table 6.2.2-2, and maximum spray additive flow rate greater than 46 gpm. To preclude closure of the valve between the spray additive tank and the spray additive eductors before sufficient NaOH has been added to meet the sump pH criterion, an interlock is provided on the motor-operated valves from the spray additive tank to prohibit closure of the valves before the prescribed amount of NaOH has been added to the sump. The total volume of sodium hydroxide added to the containment following a LOCA results in a minimum pH of 8.5 in the sump, and the rate of addition maintains the spray
solution pH in the CSS between 9.0 and 11.0 for all single failures within the
system. Single failure analysis for the spray additive subsystem is given in Table 6.5-4. The sump pH, as a function of time, is provided in Figure 6.5-5.
SAFETY EVALUATION TWO - The spray iodine removal analysis is based on the assumptions that:
- a. Only one out of two spray pumps is operating
- b. The ECCS is operating at its maximum capacity 6.5-9 Rev. 13 WOLF CREEK The spray system is assumed to spray approximately 85 percent of the total containment net free volume. This volume consists of those areas directly
sprayed plus those volumes which have good communication with the directly sprayed volumes. The remaining 15 percent of the containment free volume has restricted communication with the sprayed volumes and is assumed to be
unsprayed. A description of the unsprayed volumes is presented in Table 6.5-2.
The containment spray additive subsystem is used to maintain the spray solution at a minimum pH of 9.0 during NaOH injection to ensure efficient and rapid removal of the iodine from the containment atmosphere.
The performance of the spray system was evaluated at the containment post-LOCA calculated saturation temperature corresponding to the calculated peak
pressures provided in Table 6.2.1-8 and the containment design pressure
provided in Table 6.2.1-2. The spray design flow rate of 2,995 gpm per train
was used in the calculations provided in Appendix 6.5A.
Based on Regulatory Guide 1.4, three species of airborne iodine are postulated to exist in the containment atmosphere following a LOCA. These are elemental, particulate, and organic species.
It has been assumed in these evaluations of spray removal effectiveness that organic iodine forms are not removed by the sodium hydroxide spray. A limited
credit for the removal of airborne particulates containing iodine has been taken, assuming that the spray removal rate is 0.45 hr
-1 until a DF of 100 is attained. Credit for removal of elemental iodine is based on a spray removal rate of 10 hr
-1 until a df of 100 is attained. These assumptions underestimate the actual amounts of iodine removed and thus result in calculated accident doses higher than could realistically be expected.
Utilizing the dose analysis input parameters indicated above and in Table 6.5-2, the dose analysis of Chapter 15.0 demonstrates that offsite radiation
exposures resulting from a design basis LOCA are within the plant siting dose
guidelines of 10 CFR 100.
Appendix 6.5A provides the model used to calculate the iodine removal coefficients provided in Table 6.5-2.
6.5.2.4 Tests and Inspections All active components in the spray additive subsystem are tested both by performance tests in the manufacturer's shop and by inplace testing after installation. 6.5-10 Rev. 3 WOLF CREEK Preoperational testing is described in Chapter 14.0. During the initial preoperational tests of the system, the performance of the eductor was checked
by running the containment spray pumps with the spray additive tank filled with water. Calibration curves, which correlate water flow with 30-weight-percent NaOH flow, were provided by the manufacturer, based on shop tests. In
addition, during the initial preoperational tests, calibration curves were generated for water flow, under the conditions of periodic plant tests when the spray pump will be operating at shutoff head (miniflow only).
Routine periodic testing of the spray additive system components and the necessary support systems at power is planned. Included is a periodic sampling
of the NaOH in the spray additive tank through the local sampling connection.
The spray eductors are tested singly by opening the valves in the spray pump miniflow lines to the RWST and the valve in the eductor suction line from the RWST and running the respective pump. The operator observes the eductor
suction flow and suction pressure.
The spray additive tank isolation valves can be opened periodically for testing. The contents of the tank are periodically sampled to determine that
the required solution is maintained.
The CSS tests and inspections are discussed in Section 6.2.2.1.4, including spray nozzle tests and inspections.
6.5.2.5 Instrumentation Requirements Instrumentation and associated analog and logic channels employed for the initiation of spray additive system operation are discussed in Section 7.3.
The following describes the instrumentation which is employed for monitoring the spray additive subsystem during normal plant operation and during post-
accident operation. All alarms are annunciated in the control room.
- a. Spray Additive Tank Pressure A locally mounted indicator on the spray additive tank provides means to monitor the tank pressure while adding
nitrogen and during periodic inspections. 6.5-11 Rev. 0 WOLF CREEK
- b. Spray Additive Flow A flow element is located in each discharge line from the spray additive tank to the eductors. Readout is local and on the main control board to provide flow indication during flow testing.
- c. Spray Additive Tank Level
- 1. Redundant level instruments are provided to alarm the imminent depletion of the spray additive tank and to
provide automatic closure of the spray additive tank discharge line valves.
- 2. Redundant level instruments are also provided to annunciate at the time that sufficient additive has
been educted from the tank to meet the pH criteria of
the system. These level instruments are interlocked
with the spray additive tank discharge line valves to
preclude premature closure of those valves.
- d. Spray Additive Eductor Suction Pressure A locally mounted indicator on the eductor suction line provides eductor suction pressure during flow testing.
- e. Containment spray instrumentation is given in Section 6.2.2.1.5.
6.5.2.6 Materials The containment spray additive subsystem is constructed primarily of corrosion-resistant austenitic stainless steel. The spray additive tank, in which the
NaOH is stored, is constructed of austenitic stainless steel. Construction
materials for the spray additive subsystem are provided in Table 6.5-3.
The chemical compositions of the NaOH stored in the spray additive tank, the containment spray fluid entering the spray header during the injection phase of
containment spray, and the containment spray fluid in the system during the
recirculation phase of containment spray (containment sump solution) are provided in Table 6.5-5.
None of the materials used is subject to decomposition by the radiation or thermal environment. All specifications require that the materials be unaffected when exposed to the equipment design temperature and total
integrated radiation dose. 6.5-12 Rev. 0 WOLF CREEK The corrosion of materials in the NSSS and the containment building, resulting from the spray solution used for iodine absorption, has been tested by the
Reactor Division at ORNL (Ref. 2). The spray solutions provided in Table 6.5-5 result in negligible corrosion, based on these studies.
Sodium hydroxide does not undergo radiolytic decomposition in the post-LOCA environment. Sodium has a low neutron absorption cross section and will not undergo significant activation.
With respect to the potential for pyrolytic decomposition, NaOH is stable to at least its melting point temperature of 604 F. It may convert to sodium oxide (Na2O) upon removal of the water.
6.5.3 FISSION PRODUCT CONTROL SYSTEMS 6.5.3.1 Primary Containment The containment consists of a prestressed post-tensioned, reinforced concrete structure with cylindrical walls, hemispherical dome, and base slab lined with
welded quarter-inch carbon steel liner plate, which forms a continuous, leaktight membrane. Details of the containment structural design are discussed in Section 3.8. Layout drawings of the containment structure and the related
items are given in the general arrangement drawings of Section 1.2.
The containment walls, liner plate, penetrations, and isolation valves function to limit the release of radioactive materials, subsequent to postulated accidents, such that the resulting offsite doses are less than the guideline values of 10 CFR 100. Containment parameters affecting fission product release accident analyses are given in Appendix 15A.
Long-term containment pressure response to the design basis accident is shown in Figure 6.2.1-1. Relative to this time period, the CSS is operated to reduce
iodine concentrations and containment atmospheric temperature and pressure
commencing with system initiation, at approximately 60 seconds, as shown in
Table 6.2.2-3 and ending when containment pressure has returned to normal. For
the purpose of post-LOCA dose calculations discussed in Chapter 15.0, two dose
models have been assumed, the 0-2 hour case and the 0-30 day case, as shown in
Appendix 15A.
The containment minipurge system may be operated for personnel access to the containment when the reactor is at power, as discussed in Section 9.4.6. 6.5-13 Rev. 0 WOLF CREEK Redundant, safety-related hydrogen recombiners are provided in the containment as the primary means of controlling post-accident hydrogen concentrations. A
hydrogen purge system is provided for backup hydrogen control. Although use of the hydrogen purge system is not expected for post-accident hydrogen control, offsite dose analyses assuming the operation of the hydrogen purge system have
been performed to determine its incremental contribution on the radiological doses. This analysis is provided in Chapter 15.0.
Containment combustible gas control systems are discussed in detail in Section 6.2.5.6.5.3.2 Secondary Containment This section is not applicable to WCGS.
6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM This section is not applicable to WCGS.
6.
5.5 REFERENCES
- 1. Spraying Systems Company Topical Report No. SSCO-15215-lC-304SS-6.3-NP, April 1977, "Containment Spray Nozzles for Nuclear Power Plants" 2. "Design Considerations of Reactor Containment Spray Systems, The Corrosion of Materials in Spray Solutions," ORNL-TM-2412 Part III , December 1969 6.5-14 Rev. 0 WOLFCREEKTABLE6.5-1ESFFILTRATIONSYSTEMSINPUTPARAMETERSTOCHAPTER15.0ACCIDENTANALYSISEmergencyexhaust90filteradsorberunitefficiencies(percent)Controlroomfilter95adsorberunitefficiency(percent)Controlroomairconditioningsystemflowrate(SCFM)per trainFilteredintakefrom<550controlbuildingFilteredrecirculation>1250fromcontrolroomRev.14 WOLFCREEKTABLE6.5-2(Sheet1)INPUTPARAMETERSANDRESULTSOFSPRAYIODINEREMOVALANALYSISUltimatecorepowerrating3,565MWtTotalcontainmentfreevolume2.50x10 6 ft 3Unsprayedcontainmentfreevolume<15.0percentAreacoverageattheoperatingdeckdesign>90percentCalculated>93percentMixingratebetweensprayedandunsprayedvolumes85,000cfm***DosemodelOneregion Minimumverticaldistancetooperatingdeckfromlowestsprayheader118feet-2in.Netsprayflowratepertrain, injectionphase3,131gpmDesignNaOHflowratepereductor44.0
+/-2gpmNumberofspraypumpsoperating1SpraysolutionpH9.0to11.0 Elementaliodineabsorptioncoefficient,s,usedinaccidentcalculations10hr
-1*Expecteds25.7hr-1**Particulateiodineabsorptioncoefficient,p,usedinaccidentcalculations0.45hr
-1*Calculatedp0.73hr
-1**Spraydropsize,designSeeFigure6.5-2Rev.13 WOLFCREEKTABLE6.5-2(Sheet2)Schmidtnumber11.58Gasdiffusivity0.064Partitioncoefficient5,000*UsedDFsofupto100.**AscalculatedfromAppendix6.5A.***AdequatemixingofthecontainmentatmospherefollowingaLOCAisensuredbyeffectsoftheinitialblowdown,containmentsprays,naturalconvectionandforcedairventilationprovidedbythecontainmentcoolerswithoutrelianceonthehydrogenmixingfans.RefertoSection6.2.5foradditionalinformation.Rev.8 WOLFCREEKTABLE6.5-2(Sheet3)SPRAYNOZZLETESTRESULTSNozzledropletspectrumFigure6.5-2NozzlecapacitycurveFigure6.5-1NozzlemassmediandiameterversusFigure6.5-3pressuredropNumbermeandiameter526micron@40psi Volumemeandiameter831micron@40psiNumbermediandiameter325micron@40psiNozzleOrientationThrowDistance*EllipticCoverage*Vertical-down0ft10ft-0in.x10ft-0in.7.5°offvertical-down2.5ft10ft-0in.x10ft-0in.
15°offvertical-down3.75ft10ft-0in.x10ft-0in.30°offvertical-down5.0ft10ft-0in.x10ft-0in.40°offvertical-down7.3ft10ft-6in.x11ft-0in.
Horizontal10.6ft12ft-6in.x12ft-0in.
30°offhorizontal-up10.8ft13ft-0in.x12ft-6in.*Basedon100-footdropandpost-LOCAsaturationtemperature.Rev.0 WOLF CREEK TABLE 6.5-2 (Sheet 4)
UNSPRAYED CONTAINMENT FREE VOLUME Unsprayed Region Volume (ft
- 3) Pressurizer enclosure and overhang 26,511
Region below the four RC pump hatches 44,245
Pressurizer safety valve enclosure 14,392
Region below the four containment coolers and two filter adsorber units 63,852
Pressurizer spray valve enclosure 8,920
Region under CRDM PLENUM/SEISMIC SUPPORT PLATE 3,189
Elevator machine room and elevator shaft 16,596
Region under concrete flooring used for structural strength and shielding 182,821
Total unsprayed free volume 360,526
Percentage of free volume unsprayed ~14.4%
Rev. 22 WOLFCREEKTABLE6.5-3SPRAYADDITIVESUBSYSTEM-DESIGNPARAMETERS EductorsQuantity2Eductorinlet(motive)OperatingfluidBoratedwater OperatingtemperatureAmbientEductorSuctionFluidNaOHconcentration,wtpercent30(nominal)Specificgravity~1.3Viscosity(design),cp~10 OperatingtemperatureAmbientMaterialStainlesssteelSprayAdditiveTankNumberlTotalvolume,usablegallons4,700 NaOHconcentration,wtpercent30 Designtemperature,F200 Externaldesignpressure,psig3 Internaldesignpressure,psig10 Operatingtemperature,FAmbient Operatingpressure,psig~1*
MaterialStainlesssteelHighpressurereliefvalvesetpoint,psig10Vacuumreliefvalvessetpoint,in.Hg2SprayAdditiveSystemPipingMaterialStainlesssteel*Duringnormalconditions,thereisa1to2psignitrogengasblanket.Duringaccidentinjection,thetankpressurefallsbelowatmosphericpressure;redundantvacuumbreakersare providedinordertoassurethattankexternaldesignpressure isnotexceededrelativetothetankinternalvacuum.Rev.13 WOLFCREEKTABLE6.5-4SPRAYADDITIVESUBSYSTEM-SINGLEFAILUREANALYSISCommentsandComponentMalfunction ConsequencesAutomaticallyFailstoopenTwoprovidedinoperatedsprayparallel.Operation additivetankofonerequired.outletisolation valveFailstoclosePotentialexistsforlosingonetrain.
Operationofonlyonetrainrequired.SprayadditiveFailstoopenTwoprovided.tankvacuumOperationofone breakerrequired.Rev.0 WOLFCREEKTABLE6.5-5CONTAINMENTSPRAYSYSTEMFLUIDCHEMISTRY1.ContainmentSprayAdditiveSodiumhydroxide,weightpercent30(nominal)Temperaturerange,F60-104II.SprayedFluid-InjectionPhaseAqueoussolution,pH4.0-11.0Chloride,ppm,max0.15 Fluoride,ppm,max0.15 Boricacid,ppmboron,max/min2,500/2,400Sodiumhydroxide,ppm0-3,000Temperaturerange,F37-120III.SprayedFluid-RecirculationPhaseAqueoussolution,pH8.5-11.0 Boricacid,ppmboron,max/min2,500/2,400 Sodiumhydroxide,ppm,max3,000Temperaturerange,F120-255IV.FinalSumpFluidAqueoussolution,pH8.5-9.0 Boricacid,ppmboron,max/min2,500/2,400 Sodiumhydroxide,ppm,max3,000 Temperaturerange,F120-255Rev.13 100 90 0 80 7 80 50 30 2 1: 20 £ .. _, r w m:: A : "" m:: a. ... --7
- 5 4 3 2. 5 2 1. 5 1 -1 1.5 2 2.1 3 WOLF CREEK I I I I I I I I I f I j_ v lL J I , J I ' :7 Rev. 0 WOLP CREEK j UPDATED SAPETY ANALYSIS REPORT FIGURE 6.5-1 CAPACITY CURVE 15215-1C-304SS-6.3 WHIRLJET NOZZLE I 4 5
- 1
- 910 15.2 20 21 30 40 50 80 70 80 100 CAPACITY GPM
>-u z w ::> 0 w a: u. w > I-<{ ...J w a:
1700 1600 11 Rev. 0 WOLP CREEK UPDATED SAPETY ANALYSIS REPORT FIGURE 6.5-2 SPATIAL DROPLET SIZE DISTRIBUTION OF 15215-1C-304SS-6.3 WHIRLJET SPRAY NOZZLE 800 900 1000 11001200 1300 1400 1soo 16oo 1100 1900 2000 21 2200 23oo PARTICLE SIZE DIAMETER IN MICRONS I I
- i j I i i !i ... i WOLF CREEK . , . ...............
Rev. 0 WOL!' CRBBlt UPDATED SA!'BTY ANALYSIS REPORT FIGURE 6.5-3 PARTICLE SIZE VS. PRESSURE 15215 1C-304-S-6.3 WHIRLJET SPRAY NOZZLE WOLF CREEK --310 PO&T lOCA OCMIT AINMEMT "'TURATIOM TEMPERATURE IF) Rev. 0 WOLF CRBBK SAFETY ANALYSIS FIGURE 6.5-4 SPRAY ENVELOPE REDUCTION FACTOR R E V. 29 W I T H ON E T R A I N O F N a 0 H B O T H CON T. SP RAY T R A I N S W I T H B O T H E DU C T O R S N a 0 H B O T H CON T. SP RAY T R A I N S WOLF CREEK APPENDIX 6.5A IODINE REMOVAL MODELS FOR THE
CONTAINMENT SPRAY SYSTEM 6.5A-1 Rev. 0 WOLF CREEK 6.5A.1 PARTICULATE IODINE MODEL The spray washout model for aerosol particles is represented in equation form as follows:
l 3hEF P = 2dV (6.5A-1)
Where:
l P = spray removal constant for particles h = fall height
E = total collection efficiency for a single drop
F = spray flow rate
d = mean drop diameter
V = volume of gas space
The capture of particles by falling drops results from Brownian diffusion, diffusiophoresis, interception, and impaction. Early in the injection phase, particles are removed mainly by impaction. Following injection, when the
larger particles have already been removed, the removal rate is controlled by
diffusiophoresis, which is the collection of particulates by steam condensing
on the spray drops. The single drop collection efficiency, E, is taken as
0.0015, the minimum value observed in experimental tests (Ref. 1). The minimum
collection efficiency, 0.0015, was only attained after the major fraction of
airborne particles was removed. For early time periods, the removal rates were much higher than the minimum values ultimately reached.
The spray removal constant (l P) for particulate iodine has been calculated to be 0.73 hr
-1 , based on equation 6.5A-1.
A limited and conservative credit for spray removal of airborne particulates containing iodine has been taken, assuming the spray removal constant is 0.45 hr-1 for the 0 to 2-hour period following the postulated LOCA (see Table 6.5-2).Particle spray removal constants considerably larger and of longer duration than those conservatively chosen above have been reported from the Battelle
Northwest Containment Systems Experiment (Ref. 2) and by the Oak Ridge National
Laboratories Nuclear Safety Pilot Plant (Ref. 4). 6.5A-2 Rev. 13 WOLF CREEK 6.5A.2 ELEMENTAL IODINE MODEL The spray system, by virtue of the large surface area provided between the droplets and the containment atmosphere, will afford an excellent means of absorbing elemental radioactive iodine released as a consequence of a LOCA.
Sodium hydroxide is added to the spray fluid to increase the solubility of iodine in the spray to the point where the rate of absorption is largely dependent on the concentration of radioiodine in the air surrounding the drops.
The basic model of the containment atmosphere and spray system is given by Parsley (Ref. 4). The containment atmosphere is viewed as a "black box" having
a sprayed volume, V, and containing iodine at some uniform concentration Cg.
Liquid enters at a flow of F volumes per unit time, containing iodine at a concentration of CLl, and leaves at the same flow, at concentration CL2. A material balance for the containment vessel as a function of time is given by:
-VdCg = F(CL2 - CLl)dt (6.5A-2)
Where:
CLl = the iodine concentration in the liquid entering the dispersed phase, gm/cm 3 CL2 = the iodine concentration in the liquid leaving the
dispersed phase, gm/cm 3 V = sprayed volume of containment, cm 3 Cg = the iodine concentration in the containment atmosphere, gm/cm 3 F = the spray flow rate, cm 3/sec t = spray time, sec A drop absorption efficiency, E, which may be described as the fraction of saturation, is defined as:
E = (CL2 - CLl)/(CL* - CLl) (6.5A-3)
In addition, the equilibrium distribution of iodine between the vapor and liquid phases is given by:
H = CL*/Cg (6.5A-4) 6.5A-3 Rev. 0 WOLF CREEK Where: H = The iodine partition coefficient (gm/liter of liquids)/(gm/liter of gas)
CL* = The equilibrium concentration in the liquid, gm/cm3
Substitution of equation 6.5A-4 into equation 6.5A-3 yields E = (CL2 - CLl)/(HCg - CLl) (6.5A-5)
Solving equation 6.5A-5 for (CL2 - CLl) and inserting the result into equation 6.5A-2 gives
-(V)dCg = EF(HCg - CLl)dt (6.5A-6)
During the injection phase, CLl = 0, so that
-(V)dCg = (EFHCg)dt (6.5A-7)
Equation 6.5A-7 can be integrated to solve for Cg. The concentration of iodine in the containment atmosphere during injection as a function of time is given
by: Cg = Cg o exp [-EHFt/V] (6.5A-8)
Where: Cg o = The initial iodine concentration in the containment atmosphere, gm/cm 3 Equation 6.5A-8 is applicable up to the time the spray solution is recirculated and is based on the following assumptions:
- a. Cg is uniform throughout the containment
- b. There are no iodine sources after the initial release
- c. The concentration of iodine in the spray solution entering the containment is zero From equation 6.5A-8, the spray removal constant, l, is given by
l = EHF V (6.5A-9) 6.5A-4 Rev. 1 WOLF CREEK The above equation for is independent of the models on which the numerical evaluation of the drop absorption efficiency, E, and the iodine partition
coefficient, H, may be based.
Absorption efficiency for elemental iodine may be calculated from the time-dependent diffusion equation for a rigid sphere with the gas film mass transfer resistance as a boundary condition. This mass transfer model was suggested by L. F. Parsley (Ref. 4), who gives the solution to the diffusion
equation, with the above given boundary condition, as:
E = 1 - n=1 oo 6 Sh 2 exp
-a 2 n Q f[]a 2 n + (Sh) (Sh -1)a 2
n (6.5A-10)
Where:
Sh = the dimensionless group = kg a/HDL a = the drop radius, cm
k g = the gas film mass transfer coefficient, cm/sec
DL = the liquid diffusivity, cm2/sec
Qf = the dimensionless drop residence time
a n = the eigenvalues of the solution It should be noted that this solution, which applies to the rigid drop model, is based on the assumption that molecular diffusion is the only mechanism by which iodine is transported from the surface to the interior of the drop.
Since a high degree of mixing is expected in the drops, particularly in the presence of sizable temperature and concentration gradients, it is apparent
that this stagnant drop model presents a conservative approach to the
calculation of iodine absorption by the drops.
The gas film mass transfer coefficient required for the above calculation is computed by the equation of Ranz and Marshall (Ref. 5).
k g =D g d ()2 x 0.6 Re 0.5 Sc 0.33 (6.5A-11) 6.5A-5 Rev. 1 WOLF CREEK Where: d = drop diameter, cm
D g = diffusion coefficient in vapor, cm2/sec Re = Reynold's number
Sc = Schmidt number
A more conservative numerical value of E is obtained from equation 6.5A-12 given below, which is quoted by Postma and Pasedag (Ref. 6):
E = 1 - exp-6 k g t e d(H + k g k L) (6.5A-12)
Where: E = drop absorption efficiency k
L = liquid phase mass transfer coefficient, cm/sec t
e = drop exposure time, sec d = drop diameter, cm
H = equilibrium partition coefficient
Equation 6.5A-12 is based on a model in which it is assumed that the drop consists of an outer stagnant film and a well-mixed interior. Though this model is basically nonconservative compared with the stagnant drop model
represented by equation 6.5A-10, conservatism is introduced into equation 6.5A-12 when the following expression is used for kL:
kL =
2p 2 D L 3d (6.5A-13)
Where: D L = liquid diffusivity of iodine in water, cm 2/sec d = drop diameter, cm 6.5A-6 Rev. 1 WOLF CREEK Equation 6.5A-13 results from a truncated approximation (Ref. 6) to the rigid drop diffusion equation due to Griffith (Ref. 7). Griffith's approximation is
conservative in that it predicts lower absorption than would be predicted without such approximation for stagnant drop absorption.
The numerical value of E obtained from equation 6.5A-12 is more conservative than the one obtained from equation 6.5A-10, as shown by Postma and Pasedag (Ref. 6) by comparing them with the numerical value of E based upon another
model. The reference model chosen by Postma and Pasedag (Ref. 6) for comparison is the completely well mixed model in which the solution in the
entire drop, including the interior as well as the gas-liquid interface, is in equilibrium with the iodine concentration in the gas phase outside the drop.
The expression in this reference model is:
E = 1 -exp
-6 k g t e dh (6.5A-14)
The absorption efficiency is a function of the drop size, the gas phase mass transfer coefficient, diffusion in the liquid phase, the partition coefficient, and the drop fall time.
Eggleton's equation (Ref. 8) for the equilibrium elemental iodine decontamination factors, DF, is given by:
Where: VG = Gaseous volume of the containment
VL = Liquid volume of the containment, which may be used for calculation of the partition coefficient, H, for a given value of the DF. However, equation 6.5A-15 was not used in the present analysis; instead, a numerical
value of 5,000 for H, the minimum found from CSE tests (Refs. 9 and 10) for sodium hydroxide spray, was used in the evaluation of .Since the spray does not consist of a uniform droplet size, a spectrum of drop sizes and their corresponding volume percentage (for the specific nozzle
design) were used to determine the individual spray removal constant for each droplet size. The total spray removal constant is equal to the sum of the individual spray removal constants, i.e.:
6.5A-7 Rev. 1 WOLF CREEK l =i=1 n li =i=1 nJ=1 m iJ (6.5A-16)
Since the fall time, t e , is dependent on distance from the spray header to the operating deck, and each spray header consists of ring headers ( ) located at various levels, i was calculated for each spray ring header ( ), utilizing the appropriate drop distance for each header.
Therefore, l i J =E iJ H F iJ V (6.5A-17)
Where: E i J = collection efficiency for a single drop of micron increment i for ring header
F i J = spray flow rate for micron increment i for header J
and, F i J = (F i/nozzle) . (N J) (6 5A-18)
Where: F i/nozzle =
15.2 (N i). (V i)i=1 n N i V i N J = number of nozzles on ring header N
i = number frequency for micron increment i (Figure 6.5-2)
V i = volume of a drop in micron increment i As the spray solution enters the high-temperature containment atmosphere, steam will condense on the spray drops. The amount of condensation is easily
calculated by a mass balance of the drop:
mh + m c h g = m'h f 6.5A-8 Rev. 1 WOLF CREEK where: m and m' = the mass of the drop before and after condensa-tion, lbs m = the mass of condensate, lbs c
h = the initial enthalpy of the drop, Btu/lb
h g and h f = The saturation enthalpy of water vapor and liquid, Btu/lb The increase in each drop diameter in the distribution, therefore, is given by:
d'd 3 =v v f.h g - h h fg Where: v f = the specific volume of liquid at saturation, ft 3/lb v = the specific volume of the drop before conden-sation, ft 3/lb h fg = the latent heat of evaporation, Btu/lb h g = the enthalpy of steam at saturation, Btu/lb d and d' = the drop diameter before and after condensation, cm Postma and Pasedag (Ref. 6) conclude that condensation will tend to increase the iodine washout rate due to the increased volume of the spray. Their effect has been conservatively ignored.
The drop exposure time calculated is based on the assumption that the drops were sprayed in such a manner that the initial downward velocity of the drops
at the spray ring header elevation was zero. The drops fall under the effect
of gravity from the spray ring header to the operating deck. The minimum height is given in Table 6.5-2. As the drop size increases, the average residence time decreases from about 20 to 5 seconds. Incorporating the above parameters into equation 6.5A-16 with the sprayed containment volume, V, and assuming a single spray header flow rate, the value of the spray removal
coefficient calculated is presented in Table 6.5-2. 6.5A-9 Rev. 0 WOLF CREEK The resulting elemental iodine spray removal constant is greater than 10 hr
-1.Only this conservative removal constant of 10 hr
-1 is assumed and used in the design basis LOCA evaluations presented in Section 15.4.
6.5A.3 REFERENCES
- 1. Hilliard, R. K., Coleman L. F., "Natural Transport Effects of Fission Product Behavior in the Containment System Experiment," BNWL-1457, Battelle Pacific Northwest Laboratories, Richland, Washington, December 1970.
- 2. Hilliard, R. K., et al, "Removal of Iodine and Particulates from Containment Atmospheres by Sprays -
Containment Systems Experiment Interim Report," BNWL-1244, 1970.
- 3. Perkins, J. F., "Decay of U235 Fission Products," Physical Science Laboratory, RR-TR-63-11 , U.S. Army Missile Command Redstone Arsenal, Alabama, July 25, 1963.
- 4. Parsley, Jr., L. F., "Design Considerations of Reactor Containment Spray Systems - Part VII," ORNL TM 2412 , Part 7, 1970.
- 5. Ranz, W.E., and Marshall, Jr., W.R., "Evaporation from Drops," Chemical Engineering Progress 48, 141-46, 173-80, 1952.
- 6. Postma, A. K., and Pasedag, W. F., "A Review of Mathematical Models for Predicting Spray Removal of Fission Products in Reactor Containment Vessels," WASH-
1329, U.S. Atomic Energy Commission, June 1974.
- 7. Griffiths, V., "The Removal of Iodine from the Atmosphere by Sprays," Report No. AHSB(S)R45, United Kingdom Atomic
Energy Authority, London, 1963.
- 8. Eggleton, A. E. J., "A Theoretical Examination of Iodine-Water Partition Coefficient," AERE (R)-4887, 1967.
- 9. Postma, A. K., Coleman, L. F., and Hilliard, R. K., "Iodine Removal from Containment Atmospheres by Boric
Acid Spray," BNP-100, Battelle-Northwest, Richland, Washington, 1970.
- 10. Coleman, L. F., "Iodine Gas-Liquid Partition," Nuclear Safety Quarterly Report, February, March, April 1970, BNWL-1315-2, Battelle-Northwest, Richland, Washington, p.
2.12-2.19, 1970. 6.5A-10 Rev. 0 WOLF CREEK 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS This section addresses the preservice inspections, inservice inspections, repairs and replacements of quality group B and C (ASME Boiler and Pressure Vessel Code, Section III, Class 2 and 3) components as required by the applicable edition of Section XI of the ASME Code, including addenda, per 10 CFR 50.55a(g), with certain exceptions and alternatives whenever specific written relief is granted by the NRC per 10 CFR 50.55a, or when Section XI Code Cases are used which either have been reviewed by the NRC and found acceptable
as documented in Regulatory Guide 1.147 or approved for use by the granting of
relief requests. The conditions for use of Regulatory Guide 1.147 approved
Code Cases are discussed in Appendix 3A. The inservice testing of pumps and valves is discussed in Section 3.9(B). The limitations and modifications that the NRC places in the ASME Code in paragraph (b) of 10 CFR 50.55a are adhered to.In addition, separate preservice/inservice inspection program documents, complying with the "NRC Staff Guidance for Complying with Certain Provisions of
10 CFR 50.55a(g) - Inservice Inspection Requirements" were submitted to the NRC. Subsequent inservice inspection program documents are prepared in accordance with the 10 year update requirements in 10 CFR 50.55a and submitted to the NRC for initial approval. The inspection program documents identify the applicable Section XI Edition and Addenda and provide the details of the areas subject to examination, method of examination, extent and frequency of examination, and applicable Code Cases. "Relief Requests" seeking relief from applicable code requirements are submitted to the NRC and become part of the
inservice inspection program. The repair and replacement program identifies
the applicable Section XI Edition and Addenda, applicable Code Cases and relief
requests, and provides the administrative controls for performing repairs and
replacements.
6.6.1 COMPONENTS SUBJECT TO INSPECTION The ASME Section XI Class 2 and 3 components are classified in accordance with the definitions of the 1974 Edition of the ASME Boiler and Pressure Vessel
Code, Section III, Paragraph NA-2140. Class 2 and 3 components subject to
inspection and the extent of preservice and inservice inspections are described
below.6.6.1.1 Preservice Inspections
Class 2 components, other than those exempted by Paragraph IWC-1220, were inspected in accordance with the requirements of Subsection IWC of Section XI in the 1977 Edition of the ASME Boiler and Pressure Vessel Code up to and including Summer 1978 Addenda. However, the extent of selection of Class 2 piping welds was determined by the requirements of the 1974 Edition of Section XI with Addenda through Summer 1975, except the residual heat removal system, which contained at least a 25 percent representative sample of all pressure boundary welds at structural discontinuities distributed among the loops, and
the high pressure coolant injection system, which contained a 7-1/2 percent sample of all pressure welds at structural discontinuities on the safety
injection pump suction lines. Class 3 components were inspected in accordance
with the technical requirements of Subsection IWD of the 1977 Edition of
Section XI with Addenda through Summer 1978, insofar as practicable.
6.6-1 Rev. 20 WOLF CREEK 6.6.1.2 First 10-Year Interval Inservice Inspections All Class 2 components other than those exempted by Paragraph IWC-1220 were inspected in accordance with the requirements of Subsection IWC of Section XI in the 1980 Edition of the ASME Boiler and Pressure Vessel Code up to and including Winter 1981 Addenda. However, the extent of selection of Class 2 piping welds was determined by the requirements of the 1974 Edition of Section XI with Addenda through Summer 1975 as allowed by 10CFR50.55a(b)(2), excepting
the high pressure coolant injection system which contained a 7-1/2 percent sample of all pressure welds at structural discontinuities on the safety
injection pump suction lines. All Class 3 components were inspected in accordance with the technical requirements of Subsection IWD of the 1980 Edition of Section XI with Addenda through Winter 1981 insofar as practicable.
6.6.1.3 Subsequent 10-Year Interval Inservice Inspections All Class 2 and Class 3 components other than those exempted by Paragraph IWC-1220 and IWD-1220, respectively, will be inspected in accordance with the
requirements of the applicable Edition and Addenda of Section XI, as described
at the beginning of section 6.6 and documented in the inservice inspection
program. Beginning in ISI interval 2, the selection of piping welds for examination is determined under a risk-informed ISI program as an NRC approved alternative to the Section XI requirements. This program is implemented under the 'Relief Request' process described at the beginning of 6.6.
6.6.2 ACCESSIBILITY The physical arrangement of the components (such as piping, pumps, and valves) and supports is designed to allow personnel access to welds requiring inservice inspection to the maximum extent practical. Modifications to the initial plant
design were incorporated where practical to provide proper inspection access.
Removable insulation was provided on those piping systems initially requiring volumetric and surface inspection. In addition, the placement of pipe hangers
and supports with respect to the welds requiring inspection was reviewed and modified, where necessary, to reduce the amount of plant support required in these areas during inspection.
Working platforms have been provided in many areas required to facilitate the servicing of pumps and valves. Temporary platforms, scaffolding, and ladders will be provided to gain access to the piping welds. The surface of the welds
initially requiring ultrasonic or surface examination within the inspection boundary has been prepared to permit effective examination.
An inservice inspection design review was undertaken to evaluate access requirements of the ASME Boiler and Pressure Vessel Code with subsequent design
modifications and/or inspection technique development to ensure Code
compliance, as required, to the extent practical. The provisions for suitable
access for inservice examinations minimizes the time required for these
inspections to be performed and reduces the amount of radiation exposure to
both plant and examination personnel.
6.6-2 Rev. 20 WOLF CREEK Space is provided to handle and store insulation, structural members, shielding, and similar material related to the inspection. Suitable hoists and
other handling equipment have also been provided. Lighting and sources of power for the inspection equipment are provided at appropriate locations.
6.6.3 EXAMINATION TECHNIQUES AND PROCEDURES
Prior to commercial operation, inspection locations, inspection techniques, inspection frequencies, and evaluation of examination data for Class 2 and 3 preservice examinations were in accordance with the technical requirements of the 1977 Edition of the ASME Boiler and Pressure Vessel Code, Section XI, with
addenda through Summer 1978. The inspection locations, techniques, extent and
frequency of inspections and the evaluation of examination data for Class 2 and 3 inservice examinations are in accordance with the technical requirements of the Edition and Addenda of ASME Section XI, as described at the beginning of USAR section 6.6 and documented in the inservice inspection program.
Furthermore, the ultrasonic examination of ferritic, austenitic, and dissimilar
metal piping welds will be performed in accordance with the same Edition and
Addenda.The visual, surface, and volumetric examination techniques and procedures are written in accordance with the requirements of Section XI, Subarticle IWA-2200.
The liquid penetrant or magnetic particle methods are used for surface examinations and radiography or ultrasonic (UT) methods (manual or remote) for volumetric examinations. Manual ultrasonic examination techniques are used for most volumetric examinations of Class 2 components. Reportable indications are mapped, and records are made of maximum signal amplitude, depth below the scanning surface, and length of the reflector. The data compilation format is such as to provide for comparison of data with subsequent examinations.
Radiographic techniques may be used where ultrasonic techniques are not applicable. For areas where manual surface examinations or direct visual examinations are to be performed, reportable indications are mapped with respect to size and location in a manner to allow comparison of data to
subsequent examinations.
6.6.4 INSPECTION INTERVALS The inservice inspection schedule for Class 2 system components is developed in accordance with the requirements of Subarticles IWA-2400 and IWC-2400.
6.6-3 Rev. 12 WOLF CREEK The schedule for the inspection of Class 3 system components is developed in accordance with the requirements of Subarticles IWA-2400 and IWD-2400.
The inspection interval, as defined in Subarticle IWA-2400 of Section XI, is a 10-year interval of service. These inspection intervals represent calendar years after the reactor facility has been placed into commercial service. The interval may be extended by as much as one year to permit inspections to be concurrent with plant outages. The examinations required by Subarticles IWC-
2400 and IWD-2400 were performed completely, once, prior to initial plant startup. Inservice examinations are primarily performed during normal plant
outages, such as refueling shutdowns or maintenance shutdowns occurring during the inspection interval. However, inservice examinations may be performed while the unit is on-line, if radiological and operational conditions permit access to the components.
6.6.5 EXAMINATION CATEGORIES AND REQUIREMENTS
Inservice inspection categories and requirements for Class 2 and 3 components and piping are in agreement with Tables IWC-2500-1 and IWD-2500-1, respectively except where an alternative is approved by the NRC as previously descried in 6.6.Preservice examinations for Class 2 and 3 components following repair or replacement meet the requirements of Subarticles IWC-2200 and IWD-2200, respectively.
6.6.6 EVALUATION OF EXAMINATIONS Prior to commercial operation, evaluation of examination results of Class 2 and 3 components for Preservice Inspection were in accordance with Article IWC-
3000/IWD-3000 of the ASME Code, Section XI, 1977 Edition with Addenda through the Summer of 1978. Inservice Inspection examination results for Class 2 and 3
components are evaluated in accordance with the Edition and Addenda of ASME
Section XI, as described at the beginning of USAR section 6.6 and documented in
the inservice inspection program.
Repair and replacement of Class 2 and 3 components are performed in accordance with the requirements of ASME Section XI, as described in the opening paragraphs of USAR section 6.6 and documented in the repair and replacement program. 6.6-4 Rev. 20 WOLF CREEK 6.6.7 SYSTEM PRESSURE TEST Class 2 systems subject to pressure tests are tested in accordance with Articles IWA-5000 and IWC-5000 and Table IWC-2500-1.
Class 3 systems subject to system pressure tests are tested in accordance with the requirements of Articles IWA-5000 and IWD-5000, and Table IWD-2500-1.
Class 2 and 3 components are visually examined during the system pressure test
in accordance with the requirements of Paragraph IWA-5240, once every 1/3 of each inspection interval. For systems, or portions of systems, required to be hydrostatically tested each inspection interval, the provisions of an applicable ASME Code Case as documented in the ISI program plan may be used to perform a system leakage test in lieu of the system hydrostatic test.
6.6.8 AUGMENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURE An augmented inservice inspection program is conducted on high-energy piping between the required pipe break restraints located inside and outside the containment beyond the isolation valves. This program is conducted in accordance with the requirements set forth in Standard Review Plan 3.6.1, Branch Technical Position APCSB 3-1, B.2.d or as required per the risk-informed process for piping as outlined in EPRI report 1006937, Rev. 0.
If pipe break restraints are not provided, the area between the containment isolation valves and/or inside containment wall interface, of high-energy pipe, including valve/pipe circumferential welds, is subject to the augmented
examinations, to the maximum extent practical.
The welds are examined using volumetric techniques for butt welds and surface examination techniques for socket welds once in each inspection interval.
High-energy fluid piping systems are defined as those fluid systems that, during normal plant conditions (i.e., reactor startup, operation at power, hot
standby, and reactor cool-down to cold shutdown conditions), are in operation
or maintained pressurized under either or both of the following conditions:
- a. Maximum operating temperature exceeds 200°F.
- b. Maximum operating pressure exceeds 275 psig. 6.6-5 Rev. 20