ML16203A424
| ML16203A424 | |
| 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: ML16203A424 (998) | |
Text
{{#Wiki_filter:WOLF CREEK TABLE OF CONTENTS CHAPTER 9.0 AUXILIARY SYSTEMS Section Title Page 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.1.1 NEW FUEL STORAGE 9.1-1 9.1.1.1 Design Bases 9.1-1 9.1.1.2 Facility Description 9.1-3 9.1.1.3 Safety Evaluation 9.1-4 9.1.1.4 Tests and Inspections 9.1-5 9.1.1.5 Instrumentation Application 9.1-6
9.1.2 SPENT
FUEL STORAGE AND TRANSFER 9.1-6 9.1.2.1 Design Bases 9.1-6 9.1.2.2 Facilities Description 9.1-8 9.1.2.3 Safety Evaluations 9.1-13 9.1.2.4 Tests and Inspections 9.1-15 9.1.2.5 Instrumentation Application 9.1-15 9.1.3 FUEL POOL COOLING AND CLEANUP SYSTEM 9.1 15 9.1.3.1 Design Bases 9.1-16 9.1.3.2 System Description 9.1-18 9.1.3.3 Safety Evaluation 9.1-25 9.1.4 FUEL HANDLING SYSTEM 9.1-28 9.1.4.1 Design Bases 9.1-28 9.1.4.2 System Description 9.1-29 9.1.4.3 Safety Evaluation 9 1-50 9.
1.5 REFERENCES
9.1-71 App. 9.1A FUEL STORAGE RACK ANALYSIS 9.1A-1 9.1A.1 THE HIGH DENSITY RACK (HDR) DESIGN CONCEPT 9.1A-1 9.1A.1.1 Introduction 9.1A-1 9.1A.1.2 Design Bases 9.1A-2 9.1A.1.3 Design Description 9.1A-3 9.1A.2 CRITICALITY ANALYSES FOR THE FUEL
STORAGE POOL 9.1A-4 9.1A.2.1 Description of Fuel Storage Pool
Conditions 9.1A-4 9.1A.2.1.1 Normal Operating Conditions 9.1A-4 9.1A.2.1.2 Abnormal and Accident Conditions 9.1A-6 9.1A.2.2 Analytical Methodology 9.1A-6 9.1A.2.2.1 Reference Fuel Assembly 9.1A-7 9.1A.2.2.2 High Density Reference Fuel
Storage Cell 9.1A-7 9.1A.2.2.3 Analytical Technique 9.1A-7
9.0-i Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Section Title Page
9.1A.2.2.3.1 Fuel Burnup Calculations and Uncertainties 9.1A-8 9.1A.2.2.3.2 Effect of Axial Burnup Distribution 9.1A-9 9.1A.2.2.4 Cnticality Analyses Uncertainties and Tolerances 9.1A-10 9.1A.2.2.4.1 Nominal Design 9.1A-10 9.1A.2.2.4.2 Uncertainties Due to Manufacturing Tolerances 9.1A-10 9.1A.2.2.4.2.1 Boron Loading Tolerances 9.1A-10 9.1A.2.2.4.2.2 Boral Width Tolerance 9.1A-11 9.1A.2.2.4.2.3 Tolerances in Cell Lattice Spacing 9.1A-11 9.1A.2.2.4.2.4 Stainless Steel Thickness Tolerances 9.1A-11 9.1A.2.2.4.2.5 Fuel Enrichment and Density Tolerances 9.1A-11 9.1A.2.2.4.3 Water-Gap Spacing Between Modules 9.1A-11 9.1A.2.2.4.4 Eccentric Fuel Positioning 9.1A-11 9.1A.2.2.5 Abnormal and Accident Conditions 9.1A-12 9.1A.2.2.5.1 Temperature and Water Density Effects 9.1A-12 9.lA.2.2.5.2 Lateral Rack Movement 9.1A-12 9.1A.2.2.5.3 Rack-Gap Changes 9.1A-13 9.1A.2.2.5.4 Abnormal Location of a Fuel Assembly 9.1A-13 9.1A.2.2.5.5 Dropped Fuel Assembly 9.1A-14 9.1A.2.2.6 Benchmark Calculations 9.1A-14 9.1A.2.2.6.1 Summary 9.1A-14 9.1A.2.2.6.2 Effect of Enrichment 9.1A-16 9.1A.2.2.6.3 Effect of 10 B Loading 9.1A-17 9.1A.2.2.6.4 Miscellaneous and Minor Parameters 9.1A-17 9.1A.2.2.6.4.1 Reflector Material and Spacings 9.1A-17 9.1A.2.2.6.4.2 Fuel Pellet Diameter and Lattice Pitch 9.1A-17 9.1A.2.2.6.4.3 Soluble Boron Concentration Effects 9.1A-18 9.1A.2.2.6.5 MOX Fuel 9.1A-18 9.1A.3 THERMAL AND HYDRAULIC ANALYSES 9.1A-18 9.1A.3.1 Decay Heat Load Limit 9.1A-20 9.1A.3.1.1 Decay Heat Load Calculations and conservatisms 9.1A-20 9.1A.3.2 Margin Against Boiling 9.1A-22 9.1A.3.2.1 Heat-up Calculations and Conservatism's 9.1A-22 9.1A.3.2.2 Time-to-Boil 9.1A-23 9.1A.3.3 Local Pool Water Temperature 9.1A-23 9.1A.3.3.1 Local Temperature Evaluation Methodology 9.1A-24 9.1A.3.3.2 Local Water and Fuel Cladding Temperatures 9.1A-27 9.1A.3.4 Fuel Rod Cladding Temperature 9.1A-27 9.0-ii Rev. 14 WOLF CREEK TABLE OF CONTENTS (Continued) Section Title Page
9.1A.3.5 Decay Heat Load Limits 9.1A-29 9.1A.4 STRUCTURAL AND SEISMIC CONSIDERATIONS 9.1A-29 9.1A.4.1 Analysis Methodology 9.1A-29 9.1A.4.1.1 Fuel Weights 9.1A-31 9.1A.4.1.2 Synthetic Time-Histories 9.1A-31 9.1A.4.2 WPMR Methodology 9.1A-31 9.1A.4.2.1 Model Assumptions 9.1A-32 9.1A.4.2.2 Stiffness Elements 9.1A-33 9.1A.4.2.3 Coefficients of Friction 9.1A-33 9.1A.4.2.4 Governing Equations of Motion 9.1A-34 9.1A.4.3 Structural Evaluation of the Fuel Rack Design 9.1A-34 9.1A.4.3.1 Stress Limit Evaluations 9.1A-35 9.1A.4.3.2 Loads and Loading Combinations for Fuel Storage Racks 9.1A-35 9.1A.4.3.3 Parametric Simulations 9.1A-36 9.1A.4.3.4 Time History Simulation Results 9.1A-37 9.1A.4.3.4.1 Rack Displacements 9.1A-37 9.1A.4.3.4.2 Pedestal Vertical Forces 9.1A-39 9.1A.4.3.4.3 Pedestal Friction Forces 9.1A-40 9.1A.4.3.4.4 Rack Impact Loads 9.1A-41 9.1A.4.3.4.5 Rack to Wall Impacts 9.1A-41 9.1A.4.3.4.6 Fuel to Cell Wall Impact Loads 9.1A-42 9.1A.4.3.5 Rack Structural Evaluation 9.1A-43 9.1A.4.3.5.1 Rack Stress Factors 9.1A-43 9.1A.4.3.5.2 Pedestal Thread Shear Stress 9.1A-44 9.1A.4.3.5.3 Local Stresses Due to Impacts 9.1A-45 9.1A.4.3.5.4 Assessment of Rack Fatigue Margin 9.1A-45 9.1A.4.3.5.5 Weld Stresses 9.1A-46 9.1A.4.3.5.6 Bearing Pad Analysis 9.1A-47 9.1A.4.3.5.7 Level A Evaluation 9.1A-47 9.1A.4.3.5.8 Hydrodynamic Loads on Pool Walls 9.1A-48 9.1A.4.4 Fuel Pool Structure Integrity 9.1A-49 9.1A.4.4.1 Description of Fuel Storage Pool Structure 9.1A-49 9.1A.4.4.2 Definition of Loads 9.1A-49 9.1A.4.4.2.1 Static Loading (Dead Loads and Live Loads) 9.1A-49 9.1A.4.4.2.2 Seismic Induced Loads 9.1A-50 9.1A.4.4.2.3 Thermal Loading 9.1A-50 9.1A.4.4.2.4 Pool Water Loading 9.1A-50 9.1A.4.4.3 Analysis Methodology 9.1A-50 9.1A.4.4.3.1 Finite Element Analysis Model 9.1A-50 9.1A.4.4.3.2 Analysis Technique 9.1A-51 9.14.4.4.3.3 Load Combinations 9.1A-52 9.1A.4.4.3.4 Results of Analyses 9.1A-53 9.1A.4.4.4 Pool Liner 9.1A-53 9.1A.4.4.5 Gamma Heating Considerations 9.1A-54 9.1A.4.4.6 Conclusions 9.1A-54 9.1A.5 ADMINISTRATIVE CONTROL OF FUEL MOVEMENT AND STORAGE IN REGIONS 2 AND 3 9.1A-54 9.1A.6 REFERENCES 9.1A-55 9.0-iii Rev.14 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.2 WATER
SYSTEMS 9.2-1
9.2.1 STATION
SERVICE WATER SYSTEM 9.2-1 9.2.1.1 Service Water System 9.2-1 9.2.1.2 Essential Service Water System 9.2-3
9.2.2 COOLING
SYSTEM FOR REACTOR AUXILIARIES 9.2-13
9.2.2.1 Design Bases 9.2-13 9.2.2.2 System Description 9.2-14 9.2.2.3 Safety Evaluation 9.2-19
9.2.2.4 Tests and Inspections 9.2-21
9.2.2.5 Instrumentation Applications 9.2-21
9.2.3 DEMINERALIZED
WATER MAKEUP SYSTEM 9.2-22 9.2.3.1 Design Bases 9.2-22
9.2.3.2 System Description 9.2-23
9.2.3.3 Safety Evaluation 9.2-26
9.2.3.4 Tests and Inspections 9.2-26 9.2.3.5 Instrumentation Applications 9.2-26
9.2.4 POTABLE
AND SANITARY WATER SYSTEM 9.2-27
9.2.4.1 Design Bases 9.2-27
9.2.4.2 System Description 9.2-28 9.2.4.3 Safety Evaluation 9.2-30 9.2.4.4 Tests and Inspections 9.2-30
9.2.4.5 Instrumentation Application 9.2-30
9.2.5 ULTIMATE
HEAT SINK 9.2-31 9.2.5.1 Design Bases 9.2-31 9.2.5.2 System Description 9.2-31
9.2.5.3 Safety Evaluation 9.2-36
9.2.5.4 Tests and Inspections 9.2-40
9.2.5.5 Instrumentation Application 9.2-40
9.2.6 CONDENSATE
STORAGE AND TRANSFER SYSTEM 9.2-41
9.2.6.1 Design Bases 9.2-41
9.2.6.2 System Description 9.2-41
9.2.6.3 Safety Evaluation 9.2-43 9.2.6.4 Tests and Inspections 9.2-43 9.2.6.5 Instrumentation Applications 9.2-43
9.2.7 REACTOR
MAKEUP WATER SYSTEM 9.2-43
9.2.7.1 Design Bases 9.2-43
9.2.7.2 System Description 9.2-44 9.2.7.3 Safety Evaluation 9.2-47 9.2.7.4 Tests and Inspections 9.2-47
9.2.7.5 Instrumentation Applications 9.2-48
9.0-iv Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.2.8 CLOSED
COOLING WATER SYSTEM 9.2-48
9.2.8.1 Design Bases 9.2-48
9.2.8.2 System Description 9.2-49 9.2.8.3 Safety Evaluation 9.2-50 9.2.8.4 Tests and Inspections 9.2-50
9.2.8.5 Instrument Applications 9.3-51
9.
2.9 REFERENCES
9.2-51
9.3 PROCESS
AUXILIARIES 9.3-1
9.3.1 COMPRESSED
AIR SYSTEM 9.3-1
9.3.1.1 Design Bases 9.3-1 9.3.1.2 System Description 9.3-2 9.3.1.3 Safety Evaluation 9.3-5
9.3.1.4 Tests and Inspections 9.3-6
9.3.1.5 Instrumentation Applications 9.3-7
9.3.2 PLANT
SAMPLING SYSTEMS 9.3-7 9.3.2.1 Design Bases 9 3-8
9.3.2.2 System Description 9.3-9
9.3.2.3 Safety Evaluation 9.3-12
9.3.2.4 Tests and Inspections 9.3-12 9.3.2.5 Instrumentation Applications 9.3-12
9.3.3 EQUIPMENT
AND FLOOR DRAINAGE SYSTEMS 9.3-13
9.3.3.1 Design Bases 9.3-13
9.3.3.2 System Description 9.3-15 9.3.3.3 Safety Evaluation 9.3-26 9.3.3.4 Tests and Inspections 9.3-27
9.3.3.5 Instrumentation Application 9.3-28
9.3.4 CHEMICAL
AND VOLUME CONTROL SYSTEM 9.3-29 9.3.4.1 Design Bases 9.3-30 9.3.4.2 System Description 9.3-33
9.3.4.3 Safety Evaluation 9.3-65
9.3.4.4 Tests and Inspections 9.3-68
9.3.4.5 Instrumentation Application 9.3-68
9.3.5 SERVICE
GAS SYSTEM 9.3-69
9.3.5.1 Design Bases 9.3-69
9.3.5.2 System Description 9.3-70
9.3.5.3 Safety Evaluation 9.3-72
9.3.5.4 Tests and Inspections 9.3-72 9.3.5.5 Instrumentation Application 9.3-73
9.0-v Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.3.6 BORON
RECYCLE SYSTEM 9.3-73
9.3.6.1 Design Bases 9.3-73
9.3.6.2 System Description 9.3-74 9.3.6.3 Safety Evaluation 9.3-78 9.3.6.4 Tests and Inspections 9.3-78
9.3.6.5 Instrumentation Application 9.3-78
9.4 AIR CONDITIONING, HEATING, COOLING, AND 9.4-1 VENTILATION
9.4.1 CONTROL
BUILDING HVAC 9.4-1
9.4.1.1 Design Bases 9.4-1
9.4.1.2 System Description 9.4-4 9.4.1.3 Safety Evaluation 9.4-15 9.4.1.4 Tests and Inspections 9.4-17
9.4.1.5 Instrumentation Applications 9.4-18
9.4.2 FUEL BUILDING HVAC 9.4-18 9.4.2.1 Design Bases 9.4-19 9.4.2.2 System Description 9.4-21
9.4.2.3 Safety Evaluation 9.4-27
9.4.2.4 Tests and Inspections 9.4-28
9.4.2.5 Instrumentation Applications 9.4-29
9.4.3 AUXILIARY
BUILDING 9.4-30
9.4.3.1 Design Bases 9.4-31
9.4.3.2 System Description 9.4-33
9.4.3.3 Safety Evaluations 9.4-44 9.4.3.4 Tests and Inspections 9.4-45 9.4.3.5 Instrumentation Applications 9.4-46
9.4.4 TURBINE
BUILDING HVAC 9.4-47
9.4.4.1 Design Bases 9.4-47 9.4.4.2 System Description 9.4-49 9.4.4.3 Safety Evaluation 9.4-54
9.4.4.4 Tests and Inspections 9.4-55
9.4.4.5 Instrumentation Applications 9.4-55
9.4.5 RADWASTE
BUILDING HVAC 9.4-56 9.4.5.1 Design Bases 9.4-56
9.4.5.2 System Description 9.4-57
9.4.5.3 Safety Evaluation 9.4-60
9.4.5.4 Tests and Inspections 9.4-60
9.4.5.5 Instrumentation Applications 9.4-61
9.0-vi Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.4.6 CONTAINMENT
HVAC 9.4-61
9.4.6.1 Design Bases 9.4-62
9.4.6.2 System Description 9.4-65 9.4.6.3 Safety Evaluation 9.4-75 9.4.6.4 Tests and Inspections 9.4-76
9.4.6.5 Instrumentation Applications 9.4-77
9.4.7 DIESEL
GENERATOR BUILDING VENTILATION 9.4-78 9.4.7.1 Design Bases 9.4-78 9.4.7.2 System Description 9.4-79
9.4.7.3 Safety Evaluation 9.4-82
9.4.7.4 Tests and Inspections 9.4-83
9.4.7.5 Instrumentation Applications 9.4-83
9.4.8 ESSENTIAL
SERVICE WATER PUMPHOUSE 9.4-84
VENTILATION
9.4.8.1 Design Bases 9.4-84
9.4.8.2 System Description 9.4-85 9.4.8.3 Safety Evaluation 9.4-88 9.4.8.4 Tests and Inspections 9.4-89
9.4.8.5 Instrumentation Applications 9.4-89
9.4.9 PLANT
HEATING SYSTEM 9.4-89 9.4.9.1 Design Bases 9.4-89 9.4.9.2 System Description 9.4-89
9.4.9.3 Safety Evaluation 9.4-91
9.4.9.4 Tests and Inspections 9.4-91
9.4.9.5 Instrumentation Applications 9.4-91 9.4.10 CENTRAL CHILLED WATER SYSTEM 9.4-91
9.4.10.1 Design Bases 9.4-92
9.4.10.2 System Description 9.4-92
9.4.10.3 Safety Evaluation 9.4-94 9.4.10.4 Tests and Inspections 9.4-94 9.4.10.5 Instrumentation Applications 9.4-94 9.4.11 Essential Service Water Vertical Loop Chase Ventilation
9.5 OTHER
AUXILIARY SYSTEMS 9.5-1 9.5.1 FIRE PROTECTION SYSTEM 9.5-1
9.5.1.1 Design Bases 9.5-1
9.5.1.2 System Description 9.5-2
9.5.1.3 Safety Evaluation 9.5-27 9.5.1.4 Tests and Inspection 9.5-28 9.5.1.5 Instrumentation Applications 9.5-28
9.5.1.6 Personnel Qualification and Training 9.5-28
9.5.1.7 Equipment Operability 9.5-28
9.0-vii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.5.2 COMMUNICATION
SYSTEMS 9.5-37
9.5.2.1 Design Bases 9.5-38
9.5.2.2 System Description 9.5-39 9.5.2.3 Safety Evaluation 9.5-41 9.5.2.4 Tests and Inspections 9.5-42
9.5.3 LIGHTING
SYSTEM 9.5-42
9.5.3.1 Design Bases 9.5-42 9.5.3.2 System Description 9.5-42 9.5.3.3 Failure Analysis 9.5-44
9.5.3.4 Tests and Inspections 9.5-45
9.5.4 EMERGENCY
DIESEL ENGINE FUEL OIL STORAGE 9.5-45 AND TRANSFER SYSTEM 9.5.4.1 Design Bases 9.5-45
9.5.4.2 System Description 9.5-46
9.5.4.3 Safety Evaluation 9.5-51
9.5.4.4 Tests and Inspections 9.5-52 9.5.4.5 Instrumentation Applications 9.5-52
9.5.5 EMERGENCY
DIESEL ENGINE COOLING WATER 9.5-53
SYSTEM 9.5.5.1 Design Bases 9.5-53 9.5.5.2 System Description 9.5-54 9.5.5.3 Safety Evaluation 9.5-58
9.5.5.4 Tests and Inspections 9.5-59
9.5.5.5 Instrumentation Applications 9.5-59
9.5.6 EMERGENCY
DIESEL ENGINE STARTING SYSTEM 9.5-60 9.5.6.1 Design Bases 9.5-60
9.5.6.2 System Description 9.5-61
9.5.6.3 Safety Evaluation 9.5-64
9.5.6.4 Tests and Inspections 9.5-65 9.5.6.5 Instrumentation Applications 9.5-65
9.5.7 EMERGENCY
DIESEL ENGINE LUBRICATION SYSTEM 9.5-66
9.5.7.1 Design Bases 9.5-66
9.5.7.2 System Description 9.5-67 9.5.7.3 Safety Evaluation 9.5-73 9.5.7.4 Tests and Inspections 9.5-74
9.5.7.5 Instrumentation Applications 9.5-74
9.5.8 EMERGENCY
DIESEL ENGINE COMBUSTION AIR 9.5-75 INTAKE AND EXHAUST SYSTEM 9.5.8.1 Design Bases 9.5-75 9.5.8.2 System Description 9.5-76
9.5.8.3 Safety Evaluation 9.5-80
9.5.8.4 Test and Inspections 9.5-80
9.5.8.5 Instrumentation Applications 9.5-81
9.0-viii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
Section Title Page
9.5.9 AUXILIARY
STEAM SYSTEM 9.5-81
9.5.9.1 Design Bases 9.5-81
9.5.9.2 System Description 9.5-82
9.5.9.3 Safety Evaluation 9.5-84 9.5.9.4 Test and Inspections 9.5-84 9.5.9.5 Instrumentation Applications 9.5-84
9.5.10 BREATHING AIR SYSTEM 9.5-84
9.5.10.1 Design Basis 9.5-84
9.5.10.2 System Description 9.5-85 9.5.10.3 Safety Evaluation 9.5-86 9.5.10.4 Tests and Inspections 9.5-86
9.5.10.5 Instrumentation Applications 9.5-86 9.5.11 Station Blackout Diesel Generator 9.5-86 Support Systems 9.5.11.1 General Description 9.5-86 9.5.11.2 Interfacing systems 9.5-87 9.5.11.3 System Testing 9.5-87 9.5.11.4 Component Descriptions 9.5-87
App. 9.5A DESIGN COMPARISON TO REGULATORY POSITIONS 9.5A-0 OF REGULATORY GUIDE 1.120, REVISION 1, DATED NOVEMBER 1977, TITLED "FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS" App. 9.5B FIRE HAZARDS ANALYSES 9.5B-0
App. 9.5C RESPONSES TO QUESTIONS CONTAINED IN THE 9.5C-0 NRC'S LETTERS DATED NOVEMBER 3, 1977 FROM OLAN D. PARR TO THE SNUPPS UTILITY APPLICANTS
App. 9.5D RESPONSE TO QUESTIONS CONTAINED IN THE 9.5D-0 NRC'S LETTERS DATED OCTOBER 18, 1979 FROM OLAN D. PARR TO THE SNUPPS UTILITY APPLICANTS
App. 9.5E 10 CFR PART 50 APPENDIX R COMPARISON 9.5E-0
9.0-ix Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title 9.1-1 New Fuel Storage Design Data 9.1-2 Spent Fuel Storage Design Data 9.1-3 Design Comparison to Regulatory Positions of Regulatory Guide 1.13 Revision 1, Dated December
1975 Titled "Spent Fuel Storage Facility Design
Basis" 9.1-4 Fuel Pool Cooling and Cleanup System Design Parameters 9.1-5 Fuel Pool Cooling and Cleanup System Component Design Parameters 9.1-6 Spent Fuel Pool Cooling and Cleanup System Single
Active Failure 9.1-7 Fuel Handling Crane Data 9.1-8 Deleted 9.1-9 Weight of Reactor Vessel Head 9.1-10 Final head Weight with SHA Modification Components for Drop Analysis 9.1-11 Vertical Deformation of Reactor Vessel and Supports
Due to a Postulated SHA Drop Accident 9.1A-1 Summary of the Criticality Safety Analysis for the MZTR Storage Configuration 9.1A-2 Summary of the Criticality Safety Analysis for the Interim Checkerboard Storage Configuration 9.1A-3 Reactivity Effects of Abnormal and Accident Conditions 9.1A-4 Design Basis Fuel Assembly Specifications 9.1A-5 Reactivity Effects of Manufacturing Tolerances 9.1A-6 Reactivity Effects of Temperature and Void 9.1A-7 Summary of Criticality Benchmark Calculations 9.1A-8 Comparison of MCNP4a and KENO5a Calculated Reactivities for Various
Enrichments
- 9. 1A-9 MCNP4a Calculated Reactivities for Critical Experiments with Neutron Absorbers 9.1A-10 Comparison of MCNP4a and KENO5a Calculated Reactivities for Various B10 Loadings 9.1A-11 Calculations for Critical Experiments with Thick Lead and Steel Reflectors 9.0-x Rev.
29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title
9.1A-12 Calculations for Critical Experiments with Various Soluble Boron Concentrations
9.1A-13 Calculations for Critical Experiments with MOX Fuel 9.1A-14 Data for Decay Heat Load Limit Evaluation
9.1A-15 Data for Time-To-Boil Evaluation
9.1A-16 Data for Local Temperature Evaluation
9.1A-17 Results of Decay Heat Load Limit Evaluation
9.1A-18 Rack Material Data (200°F) 9.1A-19 Time-History Statistical Correlation Results
9.1A-20 Comparison of Bounding Calculated Load/Stresses vs. Code Allowables At Impact Locations and At Welds 9.1A-21 Bending Strength Evaluation
9.1A-22 Shear Strength Evaluation
9.2-1 Service Water System Flow Requirements Normal Power Generation Operation
9.2-2 Essential Service Water System Flow Requirements
Normal Power Generation Operation 9.2-3 Essential Service Water System Flow Requirements
Post-LOCA Operation
9.2-4 Essential Service Water System Flow Requirements Normal Shutdown Operation
9.2-5 Essential Service Water System Component Data
9.2-6 Essential Service Water System Single Active Failure Analysis
9.2-7 Essential Service Water System, Indicating and
Alarm Devices
9.2-8 Essential Service Water system, Indicating and Alarm Devices
9.2-9 Component Cooling Water System Requirements Normal
Operation
9.2-10 Component Cooling Water System Requirements Shutdown (@ 4 Hours) Operations
9.2-11 Component Cooling Water System Requirements Post-
LOCA 9.2-12 Component Cooling Water System Component Data 9.0-xi Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title 9.2-13 Component Cooling Water System Single Active
Failure Analysis
9.2-14 Component Cooling Water system, Indicating and Alarm Devices
9.2-15 Major Components Supplied with Water from
Demineralized Water Storage and Transfer System 9.2-16 Plant Water Chemistry Specifications
9.2-17 ESW/UHS Cooling Water Chemistry Analysis
9.2-18 Deleted
9.2-19 Heat Loads from Station Auxiliaries Post-LOCA
9.2-20 Heat Loads from Station Auxiliaries Normal Shutdown 9.2-21 Design Comparison to Regulatory Positions of
Regulatory Guide 1.27, Revision 2, Dated January
1976, Titled "Ultimate Heat Sink for Nuclear Power
Plants" 9.2-22 Components and Systems Served by Condensate Storage
and Transfer System
9.2-23 Summary of Reactor Makeup Water Requirements 9.2-24 Components Cooled by the Closed Cooling Water
System
9.2-25 Minimum ESW Total System Temperature Rise 9.2-26 Condensate Storage and Transfer System Component Data
9.3-1 Component Description Compressed Air System
9.3-2 Safety-Related Pneumatically Operated Valves
9.3-3 Primary Sampling System Sample Point Design Data
9.3-4 Radwaste Sampling System Sample Point Design Data
9.3-5 Process Sampling System Sample Point Design Data
9.3-6 List of Grab Sample Points for Primary and Radwaste
Sampling Systems
9.0-xii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title
9.3-7 Floor and Equipment Drainage System Single Active
Failure Analysis 9.3-8 Chemical and Volume Control System Design
Parameters
9.3-9 Chemical and Volume Control System Principal Component Data Summary
9.3-10 Failure Mode and Effects Analysis - Chemical and
Volume Control System Active Components - Normal
Plant Operation and Safe Shutdown 9.3-11 Service Gas Requirements
9.3-12 Laboratory Gas Requirements
9.3-13 Boron Recycle System Principal Component Data Summary
9.4-1 Outside Environment Design Conditions
9.4-2 Design Comparison to Regulatory Positions of Regulatory Guide 1.52, Revision 2, Dated March
1978, Titled "Design, Testing, and Maintenance
Criteria for Post-Accident Engineered-Safety-
Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants"
9.4-3 Design Comparison to Regulatory Positions of
Regulatory Guide 1.140, Revision 0, Dated March 1978, Titled "Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air
Filtration and Adsorption Units of Light-Water-
Cooled Nuclear Power Plants" 9.4-4 Design Data for Control Building HVAC System Components
9.4-5 Single-Failure Analyses Control Building HVAC
Systems
9.4-6 Design Data for Fuel Building HVAC System Components
9.0-xiii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title
9.4-7 Single-Failure Analysis - Emergency Exhaust System, Spent Fuel Pool Pump Room Coolers, and Fuel Building HVAC Isolation
9.4-8 Design Data for Auxiliary Building HVAC System
Components
9.4-9 Single-Failure Analyses - Pump Room Coolers, Penetration Room Coolers and Emergency Exhaust
System
9.4-10 Design Data for Turbine Building Ventilation System Components
9.4-11 Design Data for Radwaste Building HVAC System
Components
9.4-12 Design Data for the Containment HVAC System Components
9.4-13 Comparison of Containment Minipurge Containment
Isolation Valves with BTP CSB 6-4 9.4-14 Design Data for the Diesel Generators Building
Ventilation System Components
9.4-15 Single-Failure Analyses - Diesel Generator Building Ventilation System
9.4-16 Essential Service Water Pumphouse and UHS
Electrical Equipment Room Ventilation System
Components 9.4-17 Single-Failure Analyses - Essential Service Water
Pumphouse Ventilation System
9.4-18 Design Data for Plant Heating System Components 9.4-19 Design Data for Central Chilled Water System
Components
9.5.1-1 Fire Protection System Design Codes and Standards
9.5.1-2 Fire Protection System Fire Suppression Systems
9.5.1-3 Deleted
9.0-xiv Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued) Table No. Title
9.5.2-1 Communication Systems in Plant Areas Required to be Manned for Post Fire Safe Shutdown Following Control Room Evacuation 9.5.4-1 Emergency Diesel Engine Fuel Oil Storage and
Transfer System Component Data
9.5.4-2 Emergency Diesel Engine Fuel Oil Storage and Transfer System Indicating and Alarm Devices
9.5.4-3 Comparison of the Design to Regulatory Positions of
Regulatory Guide 1.137, Revision 0, dated January 1978, "Fuel-Oil Systems for Standby Diesel Generators"
9.5.5-1 Emergency Diesel Engine Cooling Water System
Component Data (per Diesel Engine) 9.5.6-1 Emergency Diesel Engine Starting System Component
Data
9.5.7-1 Emergency Diesel Engine Lubrication System Component Data
9.5.7-2 Emergency Diesel Engine Lubrication System
Indicating Devices
9.5.8-1 Emergency Diesel Engine Combustion Air Intake and Exhaust System Component Data
9.5.9-1 Components Supplied by Auxiliary Steam
9.5A-1 WCGS Fire Protection Comparison to APCSB 9.5-1, Appendix A
9.5E-1 WCGS Fire Protection Comparison to 10CFR50,
Appendix R
9.0-xv Rev. 29 WOLF CREEK CHAPTER 9 - 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 Title Drawing #* 9.1-1 0 New Fuel Storage Rack 9.1-2 1 Elevation View of Rack Layout 9.1-2 2 Rack Structure & Misc. Details 9.1-2 3 Pool Layout for Wolf Creek 9.1-3 1 Fuel Pool Cooling and Cleanup System M-12EC01 9.1-3 2 Fuel Pool Cooling and Cleanup System M-12EC02 9.1-4 0 Arrangement Drawing Containment Building Polar Crane 9.1-5 0 Hook Limits for Containment Building Polar Crane 9.1-6 0 Arrangement Drawing Fuel Building Cask Handling Crane 9.1-7 0 Hook Limits for Fuel Building Cask Handling Crane 9.1-8 0 Arrangement Drawing Spent Fuel Bridge Crane 9.1-9 0 Hook Limits for Spent Fuel Pool Bridge Crane 9.1-10 0 Refueling Machine 9.1-11 0 New Fuel Elevator 9.1-12 0 Fuel Transfer System 9.1-13 0 Rod Cluster Control Charging Fixture 9.1-14 0 Spent Fuel Handling Tool 9.1-15 0 New Fuel Handling Tool 9.1-16 0 Upper Core Barrel Handling Fixture 9.1-17 0 Quick-Acting Stud Tensioner 9.1-18 0 Model of Critical Buckling Load for the Drive Rod 9.1-19 0 Buckling Load for Section 2 9.1-20 0 Buckling Load for Section 1 9.1-21 0 Spring Mass System of the Head Upper Package and Reactor Vessel 9.1-22 0 Upper Portion of Vessel 9.1-23 0 Outlet Nozzle 9.1-24 0 Details of Seismic Restraint Assembly-Plan View 9.1-25 0 Details of Snubber Earthquake Restraint Assembly 9.1-26 0 Polar Crane in the "Parked Position" at AZ. 315 during Plant Operation 9.1A-1 0 Representation of the KENO5a Reference MZTR Calculational Model 9.1A-2 0 Representation of the KENO5a Reference Checkerboard Calculational Model 9.1A-3 0 Minimum Required Fuel Assembly Burnup as a Function of Nominal Enrichment to Permit Storage in Regions 2 and 3 (Fuel assemblies with enrichments less than 2.0wt% 235U will conservatively be required to meet the burnup requirements of 2.0 wt% 235 U assemblies).
9.0-xvi Rev. 29 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.1A-4 0 MCNP Calculated k-eff Values for Various Values of the Spectral Index 9.1A-5 0 KENO5a Calculated k-eff Values for Various Values of the Spectral Index 9.1A-6 0 MCNP Calculated k-eff Values at Various U-235 Enrichments 9.1A-7 0 KENO Calculated k-eff Values at Various U-235 Enrichments 9.1A-8 0 Comparison of MCNP and KEN05a Calculations for Various Fuel Enrichments 9.1A-9 0 Comparison of MCNP and KENO5a Calculations for Various Boron-10 Areal Densities 9.1A-10 0 Fuel Storage Pool Cooling Model 9.1A-11 0 Isometric View of Spent Fuel Pool CFD Model 9.1A-12 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram X Direction of Bounding OBE
Spectra (2% Damping) 9.1A-13 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram Y Direction of Bounding OBE Spectra (2% Damping) 9.1A-14 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram Z Direction of Bounding OBE
Spectra (2% Damping) 9.1A-15 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram X Direction of Bounding DBE
Spectra (4% Damping) 9.1A-15a 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram Y Direction of Bounding DBE
Spectra (4% Damping) 9.1A-16 0 Wolf Creek Elevation 2007 ft. Spent Fuel Pool Time History Accelerogram Z Direction of Bounding DBE Spectra (4% Damping) 9.1A-17 0 Wolf Creek Fuel Storage Pool Rack Impact Spring Numbering Scheme (Bottom) 9.1A-18 0 Wolf Creek Fuel Storage Pool Rack Impact Spring Numbering Scheme (Top) 9.1A-19 0 Wolf Creek Half-Full Fuel Storage Pool Bottom Springs at Baseplate 9.1A-20 0 Wolf Creek Half-Full Fuel Storage Pool Top Springs 9.1A-21 0 Wolf Creek Cask Loading Pool Bottom Springs 9.1A-22 0 Wolf Creek Cask Loading Pool Top Springs
9.0-xvii Rev. 19 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.1A-23 0 Isometric View of the Fuel Storage Pool Area 9.1A-24 0 Plan View and Dimensions of the Fuel Storage Pool Area 9.1A-25 0 Plot of Gap Between Racks 13 and 14 at Spring No. 496 in Full SFP Model 9.1A-26 0 Plot of Gap Between Racks 14 and 15 at Spring No. 504 in Full SFP Model 9.1A-27 0 Plot of Gap Between Racks 14 and the wall at Spring No. 501 in Full SFP Model 9.2-1 1 Service Water System M-12EA01 9.2-1 2 Service Water System M-12EA02 9.2-1 3 Service Water System M-0022 Sheet 1 9.2-2 1 Essential Service Water System M-12EF01 9.2-2 2 Essential Service Water System M-12EF02 9.2-2 3 Essential Service Water System M-K2EF01 9.2-2 4 Essential Service Water System M-K2EF03 9.2-3 0 ESW Pumphouse Equipment Location - Plan M-KG080 9.2-4 0 ESWS Pumphouse Equipment Location - Sections M-KG081 9.2-5 1 Makeup Demineralizer System M-0025 Sheet 1 9.2-5 2 Makeup Demineralizer System M-0025 Sheet 2 9.2-5 3 Makeup Demineralizer System M-0025 Sheet 3 9.2-5 4 Makeup Demineralizer System M-0025 Sheet 4 9.2-5 4A Makeup Demineralizer System M-0025 Sheet 4A 9.2-5a 0 Potable Water System A-0503 Sheet 1 9.2-6 0 Deleted 9.2-6A 0 Heat Rejection Rate to Ultimate Heat Sink - LOCA 9.2-6B 0 Double Ended Pump Suction Guillotine Break, Maximum Safety Injection, 4 Air Coolers, Total Air Cooler Heat Removal Rate vs Time 9.2-6C 0 Double Ended Pump Suction Guillotine Break, Maximum Safety Injection, 4 Air Coolers, RHR Heat Exchanger Heat Removal Rate Vs Time 9.2-7 0 Ultimate Heat Sink Inlet and Outlet Temperatures - Worst Temperature Period 9.2-8 0 Deleted 9.2-9 0 Deleted 9.2-10 0 Ultimate Heat Sink - Elevation - Worst Evaporation Period 9.2-11 0 Transient Velocity at Wolf Creek UHS Location - Case I 9.2-12 0 Transient Velocity at Wolf Creek UHS Location - Case II
9.0-xviii Rev. 19 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.2-13 0 Reactor Make-up Water System M-12BL01 9.2-14 0 Closed Cooling Water System M-12EB01 9.2-15 1 Component Cooling Water System M-12EG01 9.2-15 2 Component Cooling Water System M-12EG02 9.2-15 3 Component Cooling Water System M-12EG03 9.2-16 0 Demineralized Water Storage and Transfer System M-12AN01 9.2-17 1 Domestic Water System M-12KD01 9.2-17 2 Domestic Water System M-12KD02 9.2-18 0 Integrated Heat Rejection to Ultimate Heat Sink - LOCA 9.2-19 0 Heat Rejection Rate to Ultimate Heat Sink - Shutdown 9.2-20 0 Integrated Heat Rejected to Ultimate Heat Sink - Shutdown 9.2-21 0 Emergency Makeup Water Requirement - LOCA 9.2-22 0 Integrated Emergency Makeup Water Requirement 9.2-23 0 Condensate Storage and Transfer System M-12AP01 9.2-24 1 Waste Water Treatment Facility M-12WT01 9.2-25 1 Waste Water Treatment Facility M-12WT03 9.3-1 1 Compressed Air System M-12KA01 9.3-1 2 Compressed Air System (Service Air) M-12KA02 9.3-1 3 Instrument Air System M-12KA03 9.3-1 4 Instrument Air System M-12KA04 9.3-1 5 Compressed Air System M-12KA05 9.3-1 6 Compressed Air System M-12KA06 9.3-1 7 Compressed Air System M-12KA07 9.3-2 1 Nuclear Sampling System M-12SJ01 9.3-2 2 Nuclear Sampling System M-12SJ03 9.3-3 0 Nuclear Sampling System M-12SJ02 9.3-4 1 Process Sampling System M-12RM01 9.3-4 2 Process Sampling System M-12RM02 9.3-4 3 Process Sampling System M-12RM03 9.3-5 1 Sanitary Lift Station & Turb. Bldg. Sanitary Drainage System M-12LA01 9.3-5 2 Comm. Corridor & Control Bldg. Sanitary Drainage System M-12LA02 9.3-5 3 Chemical and Detergent Waste M-12LD01 9.3-5 4 Turbine Bldg. and Aux. Feedwater Pump Rooms Oily Waste System M-12LE01
9.0-xix Rev. 19 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.3-5 5 Control and Diesel Generator Bldg. Oily Waste System M-12LE02 9.3-5 6 Turbine Bldg. and Aux. Boiler Room Oily Waste System M-12LE03 9.3-5 7 Tendon Access Gallery and Turbine Bldg. Oily Waste System M-12LE04 9.3-5 8 Auxiliary Building Floor and Equipment Drain (FED) System M-12LF01 9.3-5 9 Auxiliary Building Floor and Equipment Drain System M-12LF02 9.3-5 10 Auxiliary Building Floor and Equipment Drain System M-12LF03 9.3-5 11 Auxiliary Building Floor and Equipment Drain System M-12LF04 9.3-5 12 Auxiliary Building Floor and Equipment Drain System M-12LF05 9.3-5 13 Radwaste and Fuel Bldgs. FED System M-12LF06 9.3-5 14 Radwaste Bldg. FED System M-12LF07 9.3-5 15 Control and Fuel Bldgs. FED System M-12LF08 9.3-5 16 Reactor Bldg. and Hot Machine Shop FED System M-12LF09 9.3-5 17 Radwaste Bldg. and Tunnel FED System M-12LF10 9.3-6 1 Major Drainage Areas 9.3-6 2 Floor Drain for Safety-Related Rooms Aux. Building Basement 9.3-7 1 Reactor Building, Stainless Steel Liner Plate, Reactor Refueling Canal C-0L2931 9.3-7 2 Fuel Building-Area 1, Stainless Steel Liner Plate Plan, Spent Fuel Pool C-1L6111 9.3-8 1 Chemical and Volume Control System M-12BG01 9.3-8 2 Chemical and Volume Control System M-12BG02 9.3-8 3 Chemical and Volume Control System M-12BG03 9.3-8 4 Chemical and Volume Control System M-12BG04 9.3-8 5 Chemical and Volume Control System M-12BG05 9.3-9 1 Service Gas System M-12KH01 9.3-9 2 Service Gas System M-12KH02 9.3-10 0 Gas Supply Location/Interface 9.3-11 1 Boron Recycle System M-12HE01 9.3-11 2 Boron Recycle System M-12HE02 9.3-11 3 Boron Recycle System M-12HE03 9.4-1 1 Control Building HVAC M-12GK01 9.4-1 2 Control Building HVAC M-12GK02 9.4-1 3 Control Building HVAC M-12GK03 9.4-1 4 Control Building HVAC M-12GK04 9.4-2 1 Fuel Building HVAC M-12GG01 9.4-2 2 Fuel Building HVAC M-12GG02
9.0-xx Rev. 19 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.4-3 1 Miscellaneous Buildings HVAC M-12GF01 9.4-3 2 Miscellaneous Buildings HVAC M-12GF02 9.4-3 3 Auxiliary Building HVAC M-12GL03 9.4-3 4 Auxiliary Building HVAC M-12GL02 9.4-3 5 Auxiliary Building HVAC M-12GL01 9.4-4 1 Turbine Building HVAC M-12GE01 9.4-4 2 Turbine Building HVAC M-12GE02 9.4-4 3 Turbine Building HVAC M-12GE03 9.4-4 4 Turbine Building HVAC M-12GE04 9.4-5 1 Radwaste Building HVAC M-12GH01 9.4-5 2 Radwaste Building HVAC M-12GH02 9.4-6 1 Containment Cooling System M-12GN01 9.4-6 2 Containment Cooling System M-12GN02 9.4-6 3 Containment Atmospheric Control System M-12GR01 9.4-6 4 Containment Purge Systems HVAC M-12GT01 9.4-7 0 Diesel Generators Building HVAC M-12GM01 9.4-8 0 Essential Service Water Pump House HVAC M-K2GD01 9.4-9 1 Plant Heating System M-12GA01 9.4-9 2 Plant Heating System M-12GA02 9.4-10 0 Central Chilled Water System M-12GB01 9.4-11 0 Waste Water Treatment Facility HVAC M-12VW01 9.5-1 1 Fire Protection System (site) M-0023 Sheet 1 9.5-1 2 Fire Protection System (site) M-0023 Sheet 2 9.5-1 3 Fire Protection System (site) M-0023 Sheet 3 9.5-1 4 Fire Protection System (site) M-0023 Sheet 4 9.5-2 0 Outdoor Piping, Key Plan and General Notes M-0051 9.5-3 0 ESW Pump House Fire Protection 9.5.1-1 1 Fire Protection Turbine Building M-12KC01 9.5.1-1 2 Fire Protection System (power block) M-12KC02 9.5.1-1 3 Fire Protection System (power block) M-12KC03 9.5.1-1 4 Fire Protection (Halon) System M-12KC04 9.5.1-1 5 Fire Protection System (power block) M-12KC05 9.5.1-1 6 Fire Protection (Halon) System M-12KC06 9.5.1-1 7 Fire Protection (Halon) System M-12KC07 9.5.1-2 1 Fire Area Delineation el. 1974' 10466-A-1801 9.5.1-2 2 Fire Area Delineation el. 2000' 10466-A-1802 9.5.1-2 3 Fire Area Delineation el. 2026' 10466-A-1803 9.5.1-2 4 Fire Area Delineation el. 2047'-6" 10466-A-1804 9.5.1-3 1 Reactor Coolant Pump Lube Oil Collection System 9.5.1-3 2 Reactor Coolant Pump Lube Oil Collection System
9.0-xxi Rev. 19 WOLF CREEK CHAPTER 9 - 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 TitleDrawing #*9.5.2-1 0 Telephone System Riser Diagram E-14QE01 9.5.2-2 0 Public Address System Riser Diagram E-1L9903 9.5.3-1 0 Lighting Distribution Riser Diagram E-1L9901 9.5.4-1 0 Emergency Fuel Oil System M-12JE01 9.5.5-1 1 Standby Diesel Generator "A" Cooling Water System M-12KJ01 9.5.5-1 2 Standby Diesel Generator "B" Cooling Water System M-12KJ04 9.5.6-1 1 Standby Diesel Generator "A" Intake, Exh., F.0. and Starting Air System M-12KJ02 9.5.6-1 2 Standby Diesel Generator "B" Intake, Exh., F.0. and Starting Air System M-12KJ05 9.5.7-1 1 Standby Diesel Generator "A" Lube Oil System M-12KJ03 9.5.7-1 2 Standby Diesel Generator "B" Lube Oil System M-12KJ06 9.5.9-1 1 Auxiliary Boiler System M-12FA01 9.5.9-1 2 Auxiliary Steam System M-12FB01 9.5.9-1 3 Auxiliary Steam System M-12FB02 9.5.9-1 4 Auxiliary Steam Chemical Addition System M-12FE01 9.5.10-1 1 Breathing Air System M-12KB01 9.5.10-1 2 Breathing Air System M-12KB02 9.5.10-1 3 Breathing Air System M-12KB03
9.0-xxii Rev. 19 WOLF CREEK CHAPTER 9.0 AUXILIARY SYSTEMS This chapter provides information concerning the auxiliary systems included in the WCGS powerblock. Those systems that are essential for the safe shutdown of the plant or the protection of the health and safety of the public are identified. The description of each system, the design bases for the system and for critical components, a safety evaluation demonstrating how the system satisfies the design bases, the testing and inspection to be performed to
verify system capability and reliability, and the required instrumentation and controls are provided. Those aspects of the auxiliary systems that have little or no relationship to protection of the public against exposure to radiation
are described in enough detail to allow understanding of the auxiliary system design and function. Emphasis is placed on those aspects of design and
operation that might affect the reactor and its safety features or contribute to the control of radioactivity. The capability of the system to function without compromising the safe operation of the plant under both normal operating or transient situations is clearly shown by the information provided, i.e., a failure analysis. 9.1 FUEL STORAGE AND HANDLING The power block has its own fuel storage and handling facility. The onsite fuel storage and handling facilities are designed to accommodate both new and spent fuel assemblies. 9.1.1 NEW FUEL STORAGE A new fuel storage facility (NFSF) is located within the fuel building, and provides onsite dry storage for 66 new fuel elements (approximately one-third core).9.1.1.1 Design Bases The NFSF maintains the new fuel elements in a subcritical array during all postulated design basis events. 9.1-1 Rev. 0 WOLF CREEK 9.1.1.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The NFSF is protected from the effects of natural phenomena, including earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The NFSF will perform its intended function and maintain structural integrity after an SSE or following a postulated hazard, such as fire, internal missiles, or pipe break (GDC-3 and 4). SAFETY DESIGN BASIS THREE - Components of the NFSF are not shared with other units (GDC-5). SAFETY DESIGN BASIS FOUR - The NFSF is designed to store reactor core fuel assemblies in a subcritical array (GDC-62). SAFETY DESIGN BASIS FIVE - The NFSF meets the requirements of 10 CFR 73.40, 10 CFR 73.55, and 10 CFR 73.60, which require physical protection of special nuclear material while in storage. SAFETY DESIGN BASIS SIX - The NFSF, including the new fuel storage racks, precludes insertion of new fuel assemblies in other than prescribed locations within the NFSF. SAFETY DESIGN BASIS SEVEN - The new fuel storage racks are designed for the following loads and combinations thereof:
- a. Dead loads
- b. Live loads (fuel assemblies)
- c. Crane uplift load (maximum of 5,000 pounds)
- d. Safe shutdown earthquake loads
- e. Operating basis earthquake loads
SAFETY DESIGN BASIS EIGHT - The NRC issued an exemption to the requirements of 10CFR70.24 to WCNOC on June 24, 1997. On November 12, 1998 the NRC issued 10CFR50.68, which provides eight criteria that may be followed in lieu of criticality monitoring per 10CFR70.24. One of these criteria require that radiation monitors are provided in storage areas when fuel is present to detect excessive radiation levels. Monitors meeting the provisions of GDC-63 are provided in the NFSF area to provide prompt warning of high radiation. SAFETY DESIGN BASIS NINE - The capability to inspect the NFSF is provided (GDC-61).9.1.1.1.2 Power Generation Design Basis There are no power generation design bases associated with the NFSF. 9.1-2 Rev. 13 WOLF CREEK 9.1.1.2 Facility Description The NFSF is a separate and protected area containing fuel storage racks, and is enclosed by a reinforced concrete structure with an associated steel plate top containing hinged openings covering each fuel assembly. The concrete vault is described in Section 3.8. Drainage is provided to prevent accumulation of water within the vault. The new fuel storage racks are carbon steel support structure with stainless steel guides where the rack comes into contact with the fuel assembly. New fuel assemblies are received, inspected, and stored in the new fuel storage racks in the NFSF or in the Fuel Storage Pool. A total of 66 new fuel assemblies can be stored in the NFSF racks in a lattice array having a minimum center-to-center distance of 21 inches in both horizontal
directions. The NFSF is shown in Figures 1.2-20 and 1.2-21. Figure 9.1-1 shows a typical new fuel storage rack module. The new fuel storage rack modules are designed and fabricated as four vertical continuous cells for the storage of fuel assemblies. The cells are continuous
stainless steel tubes to ensure good vertical alignment and stability for the fuel assemblies in storage position. Design, fabrication, and installation of the new fuel storage racks are based on the ASME Code specifications. Stresses
in a fully loaded rack are below the design stress level defined in the ASME Code, Section III, Appendix XVII. The new fuel storage racks are designed to seismic Category I criteria, and are anchored to the seismic Category I floor and walls of the NFSF. The criticality analysis (Reference 5) shows that the spacing between fuel assemblies in the storage racks is sufficient to maintain the array in a subcritical condition, even when fully loaded. New fuel is stored in 21-inch, center-to-center racks in the NFSF, with no water present, but which are designed to prevent accidental criticality even if unborated water is present.
For the flooded condition, assuming new fuel of the highest anticipated
enrichment in place, the effective multiplication factor does not exceed 0.95.
The maximum enrichment that may be stored in the NFSF is 5.0 w/o U-235]. The effective multiplication factor does not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming possible sources of moderation, such
as aqueous foam or mist. In the analysis for the storage facilities, the fuel assemblies are assumed to be in their most reactive condition, namely fresh or undepleted, and with no control rods or removable neutron absorbers present. Credit is taken for the inherent neutron-absorbing effect of materials of construction for the racks. 9.1-3 Rev. 19 WOLF CREEK Assemblies cannot be closer together than the design separation provided by the storage facility, except in special cases such as in fuel shipping containers where analyses are carried out to establish the acceptability of the design. The mechanical integrity of the fuel assembly is assumed. Section 9.1.4 provides an evaluation to demonstrate that the new fuel storage racks can withstand a maximum crane uplift force of 5,000 pounds. A dropped fuel assembly cannot impact the racks, since a steel cover is provided over the new fuel storage area. To further ensure that no fuel can be damaged, each storage cell is designed to prevent any portion of a fuel assembly or core component (e.g., control rods)
from extending above support or guiding surfaces of the storage cell. See Table 9.1-1 for the design data for the NFSF. 9.1.1.3 Safety Evaluation The safety evaluations given below correspond to the safety design bases in Section 9.1.1.1.1. SAFETY EVALUATION ONE - The NFSF is located in the fuel building. The fuel 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 the building. SAFETY EVALUATION TWO - The NFSF is 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 the facility is properly protected. SAFETY EVALUATION THREE - The power block has an NFSF capable of storing one-third of a core. No sharing is necessary. SAFETY EVALUATION FOUR - The criticality analysis (Reference 5) demonstrates that a 21-inch (square pitch) center-to-center storage spacing of fuel assemblies in both horizontal directions ensures the subcriticality of new fuel assemblies within the NFSF. SAFETY EVALUATION FIVE - The new fuel may be stored in a totally enclosed vault with reinforced concrete walls and a steel plate top. The new fuel storage vault is located within the fuel building. The security measures taken for the
protection of the new fuel against industrial sabotage and theft are discussed in the Physical Security Plan. 9.1-4 Rev. 11 WOLF CREEK SAFETY EVALUATION SIX - A steel checker plate cover is provided over the entire new fuel storage concrete vault. Hinged covers are provided directly over each fuel storage position. The covers and fuel racks are sized to prevent insertion of a fuel assembly in other than its prescribed location. SAFETY EVALUATION SEVEN - The new fuel storage racks, loaded with fuel, are designed to minimize the distortion or buckling of rack arrangements. Stresses in the fully loaded racks do not exceed stresses specified by the ASME Code, Section III, Appendix XVII. This condition ensures a k eff 0.98. The new fuel storage equipment is designed to meet seismic Category I requirements. The crane hookup to the new fuel assemblies is done manually and under administrative control. The new fuel storage racks are designed to withstand a maximum uplift force of 5,000 pounds. The impact load of a dropped fuel assembly is taken by the checker plate covering the new fuel assemblies. The checker plate has been analyzed and determined capable of sustaining the maximum fuel assembly drop. The probability of a dropped mass accident occurring is remote since:
- a. New fuel storage racks in the new fuel storage vault are protected from dropped objects by a steel protective
cover.
- b. Safe handling features, as described in Section 9.1.4, are incorporated into the new fuel assembly handling tools.
SAFETY EVALUATION EIGHT - As described in Section 9.1.1.5, a monitoring system is provided to initiate an audible alarm if high radiation occurs. SAFETY EVALUATION NINE - As described in Section 9.1.1.4, the NFSF is accessible for periodic inspection. 9.1.1.4 Tests and Inspections The NFSF requires no shielding and is completely accessible to plant personnel. Prior to initial use, the new fuel storage racks and modules were inspected to ensure the absence of any binding using a dummy assembly. For each cell, the dummy assembly was inserted and removed. Thereafter, the cells are
periodically inspected. 9.1-5 Rev. 11 WOLF CREEK 9.1.1.5 Instrumentation Application As described in Section 12.3.4, two area radiation monitors are provided near the NFSF which will provide a distinct audible and visual alarm to alert personnel in the vicinity. The monitors provide a hi-hi radiation alarm at 15 mrem/hr which will give prompt warning of high radiation. These monitors are provided in accordance with GDC-63. Criticality is precluded from occurring, however, by design and proper operation of the fuel handling system, as described in Section 9.1.4.
9.1.2 SPENT
FUEL STORAGE AND TRANSFER A fuel storage facility (FSF) is located within the fuel building and provides onsite storage for spent fuel elements. Fuel storage racks are located in the fuel storage pool, which is constructed of reinforced concrete with a stainless steel lining and is an integral part of the fuel building. The fuel storage pool consists of the spent fuel pool and the cask loading pool (with fuel storage racks installed). The fuel storage pool provides a cooling and shielding medium for the spent fuel. The facility provides protection for spent fuel assemblies under conditions such as tornadoes, hurricanes, earthquakes, and flooding and provides an efficient method for safe and
reliable fuel handling operations within the fuel storage pool. 9.1.2.1 Design Bases The FSF is safety related, and is required to ensure a subcritical array during all normal, abnormal, and accident conditions. It also provides a shielding and cooling medium for the spent fuel. 9.1.2.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The FSF is capable of withstanding the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The FSF is designed to maintain structural integrity after an SSE to perform its intended function following a postulated hazard, such as fire, internal missiles, or pipe break. The FSF uses the design and fabrication codes commensurate with Category I structures and the seismic category assigned by Regulatory Guide 1.29 (GDC-3 and 4). 9.1-6 Rev. 16 WOLF CREEK SAFETY DESIGN BASIS THREE - Components of this system are not shared with other units (GDC-5). SAFETY DESIGN BASIS FOUR - The fuel storage pool is designed to maintain fuel assemblies in a subcritical array with k eff 0.95 when fuel assemblies are inserted into prescribed locations (GDC-62). SAFETY DESIGN BASIS FIVE - The fuel handling area and equipment are designed to prevent a drop of an unacceptable object into the fuel storage pool. The FSF
is designed to prevent the loss of cooling water within the pool that could
uncover the stored fuel or prevent cooling capability. A redundant seismic
Category I emergency makeup water supply is provided. The fuel building is a controlled air leakage facility. SAFETY DESIGN BASIS SIX - The fuel storage racks are designed for the following loads and combinations thereof:
- a. Dead loads
- b. Live loads (fuel assemblies)
- c. Crane uplift load (the spent fuel pool bridge crane - 2 tons)
- d. Safe shutdown earthquake loads
- e. Operational basis earthquake loads
- f. Thermal loads
- g. Fuel assembly drop load h. Spent fuel pool transfer gate drop load SAFETY DESIGN BASIS SEVEN - The FSF is designed to meet the requirements of 10 CFR 73.55 and 10 CFR 73.60, which require physical protection of special
nuclear material while in storage. SAFETY DESIGN BASIS EIGHT - The fuel storage racks are constructed so as to preclude insertion of spent fuel assemblies into other than prescribed storage locations. If a fuel assembly is accidentally lowered or dropped onto the top
of the racks or into the annular space between the spent fuel racks and the pool wall, subcriticality is maintained in all cases with a shutdown margin of at least 0.05 (k eff <0.95).SAFETY DESIGN BASIS NINE - The NRC issued an exemption to the requirements of 10CFR70.24 to WCNOC on June 24, 1997. On November 12, 1998 the NRC issued 10CFR50.68, which provides eight criteria that may be followed in lieu of criticality monitoring per 10CFR70.24. One of these criteria require that
radiation monitors are provided in storage areas when fuel is present to detect excessive radiation levels. Monitors meeting the provisions of GDC-63 are
provided in the FSF area to provide prompt warning of high radiation. 9.1-7 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS TEN - The capability to inspect the SFSF is provided (GDC-61).9.1.2.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - Shielding for the FSF is sufficient to prevent exposure of the plant personnel to radiation levels greater than 2.5 mrem/hr during normal operations and 10 mrem/hr during fuel handling operations, except where appropriate access controls are implemented. Gaseous radioactivity above the spent fuel pool is maintained below the limits, as defined in Table l, Column l of Appendix B to 10 CFR 20. POWER GENERATION DESIGN BASIS TWO - A leak chase and collection system is provided for the detection of leaks in the spent fuel pool liner plate. POWER GENERATION DESIGN BASIS THREE - Borated demineralized reactor makeup water may be used to fill and to supplement water inventory in the spent fuel
pool. Boration is not essential for maintaining the subcriticality of the stored fuel assemblies. An alternate source of makeup water is supplied from the refueling water storage tank. POWER GENERATION DESIGN BASIS FOUR - Fuel handling devices have provisions to avoid dropping or jamming of fuel assemblies and to avoid applying or carrying improper loads during the transfer operation. POWER GENERATION DESIGN BASIS FIVE - Cranes and hoists which are used to lift spent fuel have a maximum lift height so that the minimum required depth of water is maintained for shielding. In addition to crane and hoist limitations, a long-handled tool is utilized when handling spent fuel. 9.1.2.2 Facilities Description The fuel storage pool provides storage space for irradiated spent fuel. The fuel storage pool is a reinforced concrete structure with a stainless steel liner having a normal water volume of approximately 55,260 cubic feet (413,400 gallons). The fuel storage pool water is borated to a concentration of no less
than 2165 ppm. Figures 1.2-20 through 1.2-22 depict the storage facility.
Figure 9.1-2 is a possible fuel storage rack arrangement. See Table 9.1-2 for design data for the fuel storage pool. When Wolf Creek received its low power operating license in September 1985, the spent fuel pool was authorized to store no more than 1344 fuel assemblies to be
located in 12 spent fuel storage racks in the spent fuel pool. With the NRC approval of the Wolf Creek storage pool rerack amendment in 1999, and with the completion of the rerack modification to the spent fuel pool, Wolf Creeks expanded storage space is increased to a capability to store 2363 fuel assemblies. The modification replaced the original 12 fuel storage racks with 15 high density storage racks in the spent fuel pool. Additional capability to add three high density storage racks within the cask loading pool could be made available with supporting evaluations, during a future campaign. The three high density racks within the cask loading pool would be capable of storing 279 fuel assemblies. 9.1-8 Rev. 14 WOLF CREEK Under the high density storage design, the fuel storage pool can be defined as a Mixed Zone Three Region (MZTR) storage configuration. Fuel Storage configuration patterns are setup using administrative controls to establish storage areas specifically designated for low burnup fuel, including fresh (unburned) fuel. Selected configurations ensure that a full core discharge can be accommodated with some allowance for other fuel assemblies that could also require Region 1 storage. Cells reserved for storage of fresh fuel, and spent fuel without any burnup limitations is designated as Region 1. Region 2 and 3 cells have associated minimum burnup requirements for unrestricted fuel storage. The MZTR storage configurations are described in Appendix 9.1A. As an alternative to MZTR storage, Region 1 fuel storage may be accomplished in a
checkerboard pattern without any enrichment/burnup restrictions. The new racks have a closer assembly to assembly spacing to allow for more fuel storage capability. The rack modules are designed as cellular structures such
that each fuel assembly has a square opening with conforming lateral support and a flat horizontal bearing surface. The design maximizes structural integrity while minimizing inertial mass. Each rack assembly is supported by four legs which are remotely adjustable. Therefore, the racks can be made
vertical and the tops of the racks can easily be made co-planer with each other. The rack module support legs are engineered to accommodate undulations in the pool floor flatness. A bearing pad is interposed between the rack pedestals and the pool liner. It serves to diffuse the dead load of the loaded racks into the reinforced concrete structure of the pool slab. The composite
box subassembly, baseplate, and support legs constitute the principle components of the rack module. The rack modules are free-standing and self supporting. They are primarily made from Type 304L austenitic stainless steel in a prismatic array
interconnected through longitudinal welds. They are separated by a gap of
approximately 1.5 inches from one another. Along the pool walls, a nominal gap
is provided which varies for each wall. The minimum allowable cell to wall dimension is 3/4 inches and the maximum nominal dimension is 7.57 inches. The racks contain Boral as an active neutron absorber. The Boral provides fixed neutron absorbtion for primary reactivity control in the high density
racks. The Boral absorbers in the racks have been sized to sufficiently shadow the active fuel height of all fuel assembly designs stored in the pool. The criticality analysis (including the associated assumptions and input parameters) given in Appendix 9.1A shows that the spacing between fuel
assemblies in the storage racks is sufficient to maintain the array, when fully loaded and flooded with nonborated water, in a subcritical condition, i.e., k eff of less than 0.95. This is based upon fuel with an original enrichment of 5.00 weight percent U-235 with at least 16 IFBA rods or 4.6 percent U-235 without IFBA. Fresh unirradiated fuel assemblies are either stored in the NFSF or in Region 1 of an MZTR or checkerboard configuration in the fuel storage pool (or both). Appendix 9.1A provides a discussion of the criticality analysis for the fresh unirradiated fuel stored wet in the fuel storage pool. Burnable poison rod assemblies, unirradiated rod control clusters, and thimble plug devices are normally stored in the fuel assemblies in the spent fuel pool. Items such as damaged fuel inserts, burnable poison rods, and other debris from fuel reconstitution may be stored in a container located in the fuel storage
pool. 9.1-9 Rev. 18 WOLF CREEK The fuel storage rack configuration does not prevent accidental lowering or dropping of a fuel assembly across the top of the racks or into the space between the racks and the pool wall. Criticality under these conditions is addressed in Safety Evaluation Eight. To further assure that no fuel can be damaged, each storage cell is designed to prevent any portion of a fuel
assembly or core component from extending above the top of the rack. The fuel storage racks are also designed to withstand the impact resulting from a
falling fuel assembly under normal loading and unloading conditions and are designed to meet seismic Category I requirements. Design, fabrication, and installation of the fuelstorage pool racks are based on applicable ASME Codes. Allowable stresses are expressed as percentages of yield stresses obtained from
Section III of the ASME Code. The structural, seismic, criticality, and thermal hydraulic analyses (including the associated assumptions and input parameters) given in Appendix 9.1A show that the racks are designed so that subcriticality is maintained during all normal, abnormal, or accident conditions. The fuel storage racks installed in the fuel storage pool were designed and manufactured by Holtec International. The rack modules are freestanding on the floor liner plate of the spent fuel pool. Time-history seismic analyses have been performed and demonstrate that no lateral supports from the pool walls or fastenings to the pool floor are required. The supports for the racks are sufficiently large in area to prevent damage to the spent fuel pool liner and floor leakchase system from concentrated loads. The rack modules are constructed from stainless steel square tubes arranged in an alternating pattern such that the connection of the tube corners from storage cells. A Boral (aluminum and boron carbide) panel centered on each side is attached to the walls of the stainless steel tubes by a stainless steel sheathing. Peripheral cells use a stainless steel sheathing on the outside wall to attach the Boral panel. The fuel assemblies are nominally located in the center of each storage cell on a nominal lattice spacing of approximately 8.99 inches. Each storage cell has a hole in or near the bottom and a rectangular opening on the top of the cell to allow cooling water to flow through the storage cell. The size of the openings precludes blockage by any crud accumulations. Adjacent to the spent fuel pool are two small pools and a washdown pit. One pool is the fuel transfer canal which has a normal water volume of approximately 13,990 cubic feet (104,659 gallons) and is connected to the refueling pool (inside the containment) by the fuel transfer tube. A leaktight
gate is provided to separate the fuel storage pool and the fuel transfer canal. This allows the fuel transfer canal to be drained for maintenance of the fuel transfer system mechanisms. 9.1-10 Rev. 14 WOLF CREEK The second pool is the spent fuel shipping cask loading pool, which has a normal water volume of approximately 12,200 cubic feet (91,268 gallons). It is designed for loading spent fuel assemblies into the spent fuel shipping cask. A leaktight gate is provided to separate the spent fuel pool from the cask loading pool in the event that the cask loading pit is drained. When fuel storage racks are installed in the cask loading pool, and the cask loading pool leaktight gate, which separates the cask loading pool from the spent fuel pool is permanently opened, the cask loading pool becomes part of the fuel storage pool. Since the cask loading pool is deeper than the spent fuel pool. Platforms will be installed beneath the cask loading pool racks to allow installation at the same elevation as the spent fuel pool racks. The concrete structures for the refueling pool, spent fuel pool, cask loading pool, and fuel transfer canal are designed in accordance with the criteria for seismic Category I structures contained in Sections 3.7(B) and 3.8. As such, they are designed to maintain leaktight integrity to prevent the loss of cooling water from the pools. In the event of a loss of integrity of the watertight gate, while one of the small pools is drained, a minimum of 10 feet of water is maintained above the top of the fuel. In addition, all piping penetrations into the pool are designed to preclude draining the pool down to an unacceptable limit, as described in Section 9.1.3. For the purpose of providing an easily decontaminable surface and to provide a construction form for the concrete pour, a liner plate surface which serves no safety function is provided for these pools. The liner plate is fabricated from 1/4-inch 304L stainless steel, which has been hot rolled, annealed, pickled, and then cold rolled to provide a smooth finish. The joint welds are provided with a leakchase system for initial testing and subsequent monitoring of weld integrity. Following installation and testing, a breach of the liner plate (which could result in any significant
loss of water through the leakchase system) is not considered credible. A monitoring system is provided for the leakchase system, as described in Section 9.3.3. Any water collected is directed to the floor and equipment drain system and transferred to the liquid radwaste system for processing. The liner plate is anchored to the concrete walls by welding to steel angles which are embedded in the concrete. An analysis has been performed which demonstrates that the liner plate will not, as a result of an SSE, break away
from the walls and fall on top of the fuel storage racks. Consequently, the liner plate is prevented from either inflicting mechanical damage to the spent fuel or from blocking the flow of cooling water around the fuel. The watertight gates are also seismically designed to preclude their failure during an SSE and falling onto the fuel storage racks. 9.1-11 Rev. 14 WOLF CREEK If weld repair of the liner plate is made in the future, the repair will be in accordance with the following:
- a. Materials used, including weld rod, will be verified in accordance with ASTM specifications or equivalent
- b. Repair procedures will be in accordance with the original fabrication specifications or equivalent
- c. Welders will be qualified in accordance with ASME Section IX or equivalent
- d. Non-destructive examination of the weld repairs will be in accordance with the original fabrication specification or equivalent Should repairs be necessary with water in the fuel pool, special procedures may be required and modifications to the above criteria may be required due to the particular circumstances.
The fuel pool cooling and cleanup system functions to limit the fuel storage pool temperature to 140 F during normal plant conditions; remove impurities from the water to improve visual clarity; and limit the radiation dose to the operating personnel to 2.5 mrem/hr during normal operations and 10 mrem/hr during refueling operations. The above dose rates consider the contribution
from spent fuel and the fuel storage pool water. No other radioactive equipment that would significantly contribute to the dose rate is stored in the pool. A description of the fuel storage pool cooling and cleanup system is provided in Section 9.1.3. The fuel transfer tube is completely shielded with permanent shielding to within radiation zone limits. No special access control, radiation monitoring, or posting is required. The expansion bellows for the fuel transfer tube are under water in the fuel transfer canal in the fuel building (see revised Figure 3.8-48). There is no bellows inspection room or opening. The fuel transfer tube is completely surrounded by concrete or water, with the exception of the seismic gaps, so that no personnel access is possible. The fuel transfer tube in the seismic gap between the containment wall and the internal containment structure and in the seismic gap between the containment wall and the fuel building is shielded, using permanently installed lead loaded silicone foam rubber, to meet the radiation zone limits (Figure 12.3-2). Therefore, there 9.1-12 Rev. 14 WOLF CREEK is no unshielded portion of the fuel transfer tube. See Section 12.2.1.3.1 for additional information. Section 9.1.4 discusses the load-bearing capability of all of the cranes serving the FSF. Section 9.1.4 also provides an evaluation which demonstrates that the maximum uplift force is due to the spent fuel pool bridge crane and the maximum impact load that is due to a dropped fuel assembly. The racks are designed to withstand these loads with no increase in k eff.9.1.2.3 Safety Evaluations The safety evaluations given below correspond to the safety design bases in Section 9.1.2.1.1. SAFETY EVALUATION ONE - The FSF is located in the fuel building. The fuel 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 FSF is designed to remain functional after an SSE. Appendix 9.1A provides the design loading conditions that were considered.
Sections 3.5, 3.6, and 9.5.1 provide the hazards analyses to assure that the facility is properly protected. SAFETY EVALUATION THREE - The plant has a FSF capable of storage of 2363 fuel assemblies, with the capability of ultimately expanding to 2642 fuel assemblies pending addition of storage racks in the Cask Loading Pool and the supporting evaluations, during a future campaign. SAFETY EVALUATION FOUR - The criticality analyses described in Appendix 9.1A demonstrate that the MZTR loading configuration or alternate Region 1
checkerboard pattern configuration satisfies the subcriticality condition, assuming fresh fuel of up to 5.0 weight percent enrichment with sixteen IFBAs (4.6 weight percent without IFBAs) and unborated water in the pool. 9.1-13 Rev. 20 WOLF CREEK Procedures and precautions described in Appendix 9.1A are employed to ensure that each fuel assembly has achieved the burnup VS initial enrichment that is specified in Appendix 9.1A before it is stored in Region 2 or Region 3 in the fuel storage pool. SAFETY EVALUATION FIVE - As described in Section 9.1.2.2, the spent fuel is stored within a concrete pool which has no penetrations which can result in an unacceptable loss of water. As described in Section 9.1.3, a system provides cooling and emergency makeup water for the spent fuel pool. Section 9.4.6 describes the ventilation system provided for the fuel building. Table 9.1-3 indicates compliance with Regulatory Guide 1.13 positions. SAFETY EVALUATION SIX - The structural, seismic, criticality, and thermal-hydraulic analyses provided in Appendix 9.1A demonstrate that the fuel storage racks are designed to withstand normal, abnormal, and accident conditions without causing a decrease in the degree of subcriticality. Section 9.1.4 evaluates the bases for external loads on the fuel storage racks. The probability of a dropped mass accident occurring is remote because of the safe handling features described in Section 9.1.4. SAFETY EVALUATION SEVEN - The spent fuel is stored within a reinforced concrete wall pool in the fuel building. The security measures taken for the protection of the new and spent fuel against industrial sabotage and theft are discussed in the Physical Security Plan. SAFETY EVALUATION EIGHT - Criticality analyses, described in Appendix 9.1A, show that if a fuel assembly is dropped on top of the racks or into the gap between the racks and the pool wall, the subcriticality criteria are maintained. The worst geometric configuration is for a fresh fuel assembly to be inadvertently loaded into an empty cell in the checkerboard configuration with the remainder of the rack fully loaded with fuel from the highest permissible reactivity. If it is assumed that all fuel assemblies are new fuel and the pool water is unborated, k eff <0.95.SAFETY EVALUATION NINE - As described in Section 9.1.2.5, a monitoring system is provided to initiate an audible alarm if high radiation occurs. 9.1-14 Rev. 14 WOLF CREEK SAFETY EVALUATION TEN - Access to the FSF is provided for periodic inspection as shown in Figures 1.2-20 through 1.2-22. 9.1.2.4 Tests and Inspections The fuel storage racks were shop tested by insertion of a dummy assembly which is 0.17 inch minimum wider than an actual fuel assembly to ensure there is no significant resistance. 9.1.2.5 Instrumentation Application As described in Section 12.3.4, two area radiation monitors are provided near the FSF which will provide a distinct audible and visual alarm to alert personnel in the vicinity. The monitors provide a hi-hi radiation alarm at 15
mrem/hr which will give prompt warning if high radiation occurs. These
monitors are provided in accordance with GDC-63. Criticality is precluded from occurring, however, by design and proper operation of the fuel handling system, as described in Section 9.1.4. 9.1.3 FUEL POOL COOLING AND CLEANUP SYSTEM The fuel pool cooling and cleanup system (FPCCS) is designed to maintain the fuel storage pool water temperature below prescribed limits by removing decay heat generated by stored spent fuel assemblies and to remove impurities from the refueling pool water, the spent fuel pool water, the transfer canal water, and the water in the cask loading pool in order to ensure optical clarity and
to limit the concentration of specific activity in the water. This section describes the FPCCS. The FPCCS consists of three subsystems:
- a. Fuel pool cooling system
- b. Fuel pool cleanup system
- c. Fuel pool surface skimmer system Each of these subsystems has specific functions and design bases. 9.1-15 Rev. 14 WOLF CREEK 9.1.3.1 Design Bases 9.1.3.1.1 Safety Design Bases The portion of the FPCCS associated with the cooling of spent fuel is safety-related.SAFETY DESIGN BASIS ONE - The safety-related portion of the FPCCS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The FPCCS is 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). SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power.
Components of this system are not shared with other units (GDC-5 and 44). 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-45 and 46). SAFETY DESIGN BASIS FIVE - The safety-related portions of the FPCCS use the design and fabrication 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 is provided so that the FPCCS's safety function is not compromised. This includes isolation of components to deal with leakage or malfunctions and to isolate nonsafety-related portions of the FPCCS (GDC-44). 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 C testing. SAFETY DESIGN BASIS EIGHT - The fuel pool cooling system maintains the fuel storage pool water temperature below 170°F, considering the 9.1-16 Rev. 14 WOLF CREEK maximum decay heat generation rate resulting from the maximum anticipated spent fuel inventory with the maximum anticipated fuel burnup (GDC-44 and 61 and Regulatory Guide 1.13). SAFETY DESIGN BASIS NINE - System piping is arranged so that loss of piping integrity or operator error does not result in draining of the fuel storage pool below a minimum depth above the stored fuel to ensure sufficient cooling media for cooling the stored spent fuel (Regulatory Guide 1.13). SAFETY DESIGN BASIS TEN - Redundant seismic Category I makeup water supplies from the essential service water system are provided to ensure adequate makeup capability. SAFETY DESIGN BASIS ELEVEN - A monitoring system is provided for the FPCCS to detect conditions that could result in the loss of decay heat removal capabilities (GDC-63). 9.1.3.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The fuel pool cooling system is comprised of two cooling trains, each of which limits the fuel storage pool water temperature to a maximum of 140 F during a partial core offload and 170 F during a full core offload. Assumptions and heat loads for both design conditions are given in Table 9.1-4. POWER GENERATION DESIGN BASIS TWO - The fuel pool cleanup and surface skimmer systems maintain the optical clarity of the pool water so that fuel handling operations are not hampered by limited visibility. POWER GENERATION DESIGN BASIS THREE - The fuel pool cleanup system limits the fission and corrosion product concentrations in the refueling pool water, the transfer canal water, and the fuel storage pool water to permit operator access to the fuel storage area and for fuel handling operations. POWER GENERATION DESIGN BASIS FOUR - The fuel pool cleanup system contains two pumps and two filters to allow for continuous system operation at a reduced
capacity during filter cartridge changing and pump maintenance. POWER GENERATION DESIGN BASIS FIVE - The fuel pool cleanup system provides the means for filtering and demineralizing the contents in the refueling water storage tank (RWST). 9.1-17 Rev. 14 WOLF CREEK 9.1.3.2 System Description 9.1.3.2.1 General Description The FPCCS shown in Figure 9.1-3 consists of two cooling trains, a cleanup loop, and a surface skimmer loop. The system design parameters are given in Table 9.1-4.9.1.3.2.1.1 Fuel Pool Cooling System The fuel pool cooling system consists of two 100-percent-capacity cooling trains for the removal of decay heat generated by irradiated fuel stored in the fuel storage pool. Each train consists of a horizontal centrifugal pump, a shell and U-tube heat exchanger, a strainer, manual valves, and the instrumentation required for system operation. The fuel pool cooling heat exchangers are serviced by the component cooling water system on the shell side with motor-operated isolation valves provided. Spent fuel is placed in the fuel storage during a refueling sequence, and is stored there until it is shipped offsite. During a refueling, approximately 2/5 of a core may be transferred to the fuel storage pool or the entire core may be transferred to the fuel storage pool during a core-offload. The decay heat generated is transferred from the fuel pool cooling system through the fuel pool cooling heat exchangers to the component cooling water system. During normal system operation, one fuel pool cooling pump takes suction from the fuel storage pool and transfers the pool water through a fuel pool cooling heat exchanger back to the fuel storage pool. The fuel pool cooling pump suction is protected by a permanent strainer located at the terminal end of the suction piping within the fuel storage pool. The pump suction line penetrates the spent fuel pool wall, near the normal fuel storage pool water level. The return line terminates at nominal elevation 2038 feet, approximately 17 feet above the top of the storage racks in the fuel storage pool. In order to prevent the draining of the fuel storage pool by siphoning action, an antisiphon hole is located in each return line, near the surface of the pool
water.Normal makeup water to the fuel storage pool is supplied by the reactor makeup water system. An alternate source of makeup water is the RWST via the fuel pool cleanup pumps. Boron addition to the fuel storage pool is normally accomplished by supplying borated water from the boric acid tanks via the boric
acid blender. Boron may also be added by using the RWST as the source of
makeup water to the fuel storage pool. All makeup and boron addition operations require manual action. Isolation of nonsafety-related portions of the FPCCS is
a manual action. Whenever irradiated fuel assemblies are in the fuel storage pool, at least 23 feet of water is maintained over the top of irradiated fuel assemblies seated in the storage racks. The water level in the fuel storage pool is determined to be above the minimum required depth once per 7 days. With water level less than 23 feet, suspend crane operations with loads in the fuel storage areas. Technical Specification 3.7.15 specifies the required actions taken during
movement of irradiated fuel assemblies with water level less than 23 feet. 9.1-18 Rev. 14 WOLF CREEK An FPCCS leak is detected when an abnormally high amount of makeup water is required for the fuel storage pool. Leakage is also detected by the floor drain system, as described in Section 9.3.3. Once a significant leak is found, the affected item is isolated and repaired. 9.1.3.2.1.2 Fuel Pool Cleanup System
The fuel pool cleanup system contains two inline centrifugal pumps and two filters in parallel, a mixed bed demineralizer, and a wye-type strainer. The pumps and filters are designed for 50 percent of the system capacity, and the
demineralizer and strainer are designed for 100-percent system capacity. The
demineralizer removes ionic corrosion impurities and fission products. The
filters are provided to remove particulate matter which would have otherwise entered the demineralizer, and the wye strainer downstream of the demineralizer removes resin fines which may be released from the resin bed. The fuel pool cleanup system provides the capability for purification of the water in the fuel storage pool, the transfer canal, the refueling pool, and the RWST. The water chemistry specifications are given in Table 9.2-16. The fuel pool cleanup pump design is based upon both fuel pool cleanup pumps
running. One fuel storage pool volume is processed per day. The design flow rate allows one volume change of the RWST contents in less than 25 hours, or the contents of the refueling pool in less than 22 hours. 9.1.3.2.1.3 Fuel Pool Surface Skimmer System
Surface debris is removed from the fuel storge pool, the fuel transfer canal, and refueling pool by a surface skimmer system. This system is comprised of surface intakes containing float-type strainers positioned just below the water surface. Lines from both pools and the fuel transfer canal are tied into a common header containing a pump and filter which discharges back into the fuel storage pool or refueling pool. 9.1.3.2.2 Component Description FPCCS component design parameters are given in Table 9.1-5. Codes and standards applicable to the FPCCS are listed in Tables 3.2-1 9.1-19 Rev. 14 WOLF CREEK and 9.1-5. The FPCCS is designed and constructed in accordance with the following quality group requirements: containment penetrations are quality group B, the separate and redundant cooling loops are quality group C, and the cleanup and skimmer loops are quality group D. The quality group B and C portions are designed to seismic Category I criteria. Fuel Pool Cooling Pumps - The pumps are 100-percent-capacity, horizontal/centrifugal units. All wetted surfaces are austenitic stainless steel. Each pump takes suction from the fuel storage pool via separate suction lines. Each pump is sized to include an additional 5-percent margin on flow at the design head to accommodate normal degradation of performance due to
impeller wear. Fuel Pool Skimmer Pump - The inline centrifugal pump takes suction from movable surface skimmers, circulates the water through a pool surface strainer and a high efficiency filter, and returns it to the spent fuel pool. All wetted surfaces of the pump are austenitic stainless steel. Fuel Pool Cleanup Pumps - These inline centrifugal pumps are used to circulate fuel storage pool and refueling pool water through the fuel pool cleanup filters, demineralizer, and wye strainer for removal of particulate and ionic impurities. Each pump is designed to provide 50 percent of the design flow in the loop. All wetted parts of the pumps are austenitic stainless steel. The contents of the RWST may also be circulated through the cleanup loop, using these pumps. Fuel Pool Cooling Heat Exchangers - The heat exchangers are the shell and U-tube type. Fuel pool water circulates through the tubes while component cooling water circulates through the shell. Each of the two heat exchangers is sized for 100 percent of the design heat load. Fuel Pool Cleanup Demineralizer - A flushable, mixed bed demineralizer is used to provide adequate fuel pool water purity for Zone B access of plant personnel to the pool working areas. This demineralizer is also used to purify the contents of the RWST. Fuel Pool Cleanup Filters - Two filters in parallel, each sized at 50 percent of design flow in the cleanup loop, are located in the purification train, upstream of the demineralizer, to prevent possible particulates from being passed to the demineralizer. 9.1-20 Rev. 14 WOLF CREEK Fuel Pool Skimmer Filter - The fuel pool skimmer filter is used to remove particles which are swept from the fuel pool water surface and not removed by the basket strainer in the floating skimmer. Fuel Pool Strainers - A strainer is located in each fuel pool cooling pump suction line to prevent the introduction of relatively large particles which might otherwise foul the fuel pool cooling heat exchangers or damage the fuel pool cooling pumps. Fuel Pool Skimmer Strainer - A strainer is located in the floating skimmer inlet to remove relatively large debris from the skimmer process flow. Fuel Pool Cleanup Strainer - A wye-type strainer is provided downstream of the fuel pool cleanup demineralizer to prevent the entry of resin fines into the fuel pool and to trap any resin beads released in the event of retention
element failure. Refueling Pool Drain Inlet Strainer - a strainer is located in the inlet of refueling pool drain line to the suction of the reactor coolant drain tank pumps to prevent the introduction of relatively large particles which might
otherwise restrict the refueling pool drain flow and/or damage the reactor coolant drain tank pumps. Valves - Manual stop valves are used to isolate equipment and manual throttle valves to provide flow control. Valves in contact with fuel pool water are
austenitic stainless steel. Motor-operated isolation valves are provided in the CCW line from each fuel pool cooling heat exchanger. Piping - All piping in contact with the fuel pool water is austenitic stainless steel. The piping is welded, except where flanged connections are used to
facilitate maintenance. 9.1.3.2.3 System Operation 9.1.3.2.3.1 Fuel Pool Cooling System Normal operations of the fuel pool cooling system are manual and intermittent. The system is started, operated, and secured locally as required to maintain the water temperature below the established temperature limit for the fuel storage pool and to minimize the starting and stopping of a fuel pool cooling pump. During refueling, the refueling pool and the reactor core are cooled by the RHR system, as described in Section 5.4.7. The fuel pool cooling system is
used only for removal of the decay heat generated by the irradiated fuel in the fuel storage pool. The fuel storage pool water is borated to a concentration of no less than 2165 ppm. 9.1-21 Rev. 14 WOLF CREEK Boron addition to the fuel storage pool is normally supplied from the boric acid tanks via the boric acid transfer pumps and the boric acid mixing tee, using a feed-and-bleed process. Boron may also be added to the pool water by supplying borated water from the RWST via the fuel pool cleanup pumps. These operations require manual action by the operator. Makeup water to the fuel storage pool is normally provided by the reactor makeup water system via a manually operated valve. Makeup water may also be supplied from the RWST via the fuel pool cleanup pumps if the reactor makeup water system is unavailable. These makeup supplies compensate for normal evaporative losses from the fuel storage pool. The flow rate to the fuel storage pool is locally controlled by a manually operated valve. When a complete irradiated core is unloaded from the reactor and stored in the fuel storage pool, the fuel pool cooling system has the capability to maintain the fuel storage pool water temperature below 170 F with only one cooling train operating in the normal mode (see Table 9.1-4). Following a loss of normal power, without a loss-of-coolant accident (LOCA), the fuel pool cooling pumps can be switched manually to the standby power
system to maintain cooling of the fuel storage pool. Following a LOCA with loss of offsite power, the fuel pool cooling pumps trip, and cooling of the fuel storage pool is interrupted. The fuel storage pool water temperature is increased from the initial temperature, indicated in Table 9.1-4, Item 5c. Post-LOCA, at the start of the recirculation phase, component cooling water (CCW) flow to the fuel pool heat exchangers will be automatically isolated and CCW flow to the RHR heat exchangers will be automatically started. Flow is manually reestablished to the fuel pool heat exchangers after approximately 4 hours, when sufficient excess capacity exists in the CCWS.
When the FPCCS is reestablished, the fuel pool cooling pumps are manually
loaded to the Class 1E power source, and CCW flow is established to at least one of the fuel pool cooling heat exchangers. The heat load contribution from the fuel storage pool for determination of the maximum total heat load to the UHS, as described in Section 9.2.5, is based on the decay heat rate shown in Table 9.1-4, item 5C. The fuel storage pool is assumed to contain fuel from the recent full core discharge (193 fuel assemblies) with the balance of the rack cells (2449 cells) occupied by fuel from old discharges. The heat loads, shown in Table 9.1-4, item 5c, occur 15 days after shutdown, which corresponds to the earliest time for reactor
startup. 9.1-22 Rev. 14 WOLF CREEK The bulk fuel storage pool and in-cell thermal hydraulic analysis (including the associated assumptions and input parameters) given in Appendix 9.1A supports the data provided in Table 9.1-4. Redundant, safety-related sources of makeup water are supplied to the fuel
storage pool by the ESW system, via manually operated valves. This source of
makeup is to be used only when nonsafety-related makeup sources are not
available. During normal shutdown of the reactor, the RHR system is utilized to remove
core decay heat, as described in Section 5.4.7. At approximately 4 hours after
shutdown, the RHR heat exchanger represents sufficient available CCW duty to
require reducing or terminating CCW flow to the FPCCS heat exchangers. Under this mode of operation, the fuel storage pool temperature is allowed to rise to a maximum of 170 F, at which point flow is reestablished to the FPCCS heat exchangers and sufficient excess CCWS capacity exists to handle the fuel storage pool duty.
The fuel storage pool lowest temperature, as evaluated in Appendix 9.1A reactivity analysis, is 39 F. Intermittent use of the fuel pool cooling system to maintain the pool temperature at or above 39 F is acceptable. 9.1.3.2.3.2 Fuel Pool Cleanup System
Normal fuel pool cleanup system operation is manual and intermittent. The system is started, operated, and secured locally, as required, to maintain
optical clarity and to limit ionic corrosion and fission product concentration
in the fuel storage pool and the refueling pool. During normal system
operation, both fuel pool cleanup pumps can be run to obtain maximum system
capability. Samples are periodically taken from the cleanup loop to determine the quality of the water.
During a refueling, after the refueling pool is filled with borated water from
the RWST, the fuel pool cleanup pump(s) take suction from the refueling pool
and transfer the water through the fuel pool cleanup filter(s) and the fuel pool cleanup demineralizer and back to the refueling pool. The cleanup of the refueling pool by the fuel pool cleanup system is augmented by the CVCS via the
RHR system to expedite the cleanup process. These operations are continued
during the entire refueling process to maintain water clarity for refueling and
to minimize the radiation dose to operators. Following transfer of the irradiated fuel to
9.1-23 Rev. 27 WOLF CREEK the fuel storage pool, the cleanup lines to the refueling pool are manually isolated and drained, and fuel storage pool cleanup is initiated as required. After the refueling pool is drained to the RWST, the fuel pool cleanup system is isolated from the fuel storage pool, and the RWST is manually aligned for cleanup by the fuel pool cleanup filter(s) and demineralizer, via the fuel pool cleanup pump(s). Upon high differential pressure or indication by manual sampling that the demineralizer resins are spent, the demineralizer resins are transferred to the solid radwaste system, as described in Section 11.4. Upon high differential pressure across the fuel pool cleanup or skimmer filters, that filter is
isolated, and the cartridge is replaced by the filter handling system of the solid radwaste system, as described in Section 11.4. However, if the system is unable to maintain sufficient clarity of the pool water and radiation levels adjacent to the pool when operated continuously, the
filter and/or resin is replaced. No set radiation sampling frequency has been established for the pool water. In general, sampling is more frequent during and immediately after a refueling or if pool water radiation levels are higher
than at other times. Design parameters for the fuel pool cleanup system are as follows: Filter Demineralizer
- 1. Decontamination factor Iodine 1 100 Cesium and rubidium 1 10 Other nuclides 1 100
- 2. Radiation level NA NA (See Section 12.2.1.3.2)
The fuel pool cleanup system is manually secured when the fuel pool water temperature exceeds 140 F to prevent damage to the fuel pool cleanup system resin beds. 9.1-24 Rev. 19 WOLF CREEK 9.1.3.2.3.3 Fuel Pool Surface Skimmer System The fuel pool surface skimmer system is aligned and operated, as required, to clean the refueling pool water surface and/or the fuel storage pool water surface. All operations require manual operator action. 9.1.3.3 Safety Evaluation The safety evaluations given below correspond to the safety design bases in Section 9.1.3.1.1. SAFETY EVALUATION ONE - The safety-related portions of the FPCCS are located in the reactor, auxiliary, and fuel 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 FPCCS are designed to remain functional after a safe shutdown earthquake. Sections 3.7(B).2 and
3.9(B) provide the design loading conditions that were considered. Sections
3.5 and 3.6 provide hazards analyses to assure that safe shutdown, as outlined in Section 7.4, can be achieved and maintained. SAFETY EVALUATION THREE - Complete redundancy is provided and, as indicated by Table 9.1-6, 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 FPCCS was initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.1.3.4. Section 6.6 provides the ASME Boiler and Pressure Vessel Code Section XI requirements that are appropriate for the FPCCS. 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. Section 9.1.3.2.2 shows that the components meet the design and fabrication codes given in Section 3.2. All the power
supplies and the control functions necessary for safe function of the FPCCS are
Class 1E, as described in Chapters 7.0 and 8.0. 9.1-25 Rev. 19 WOLF CREEK SAFETY EVALUATION SIX - Section 9.1.3.2.1.1 describes provisions made to identify and isolate leakage or malfunction and to isolate the affected portion 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 - The maximum decay heat generation rate that can be removed is governed by the capabilities of the fuel pool cooling system. The maximum decay heat loading was determined to be 63.41 MBtu/hr as discussed in appendix 9.1A.5.5 (see also Table 9.1-4). Under peak heat load, one fuel pool cooling train can maintain the fuel storage pool water temperature at or below 170 F. The fuel storage pool temperature may exceed 150 F for a maximum duration of 9 days following a full core offload (reference 9). Cycle-specific decay heat analysis and administrative controls ensure that the maximum pool thermal loading limit is not exceeded. The maximum anticipated heat load to be removed by the fuel pool cooling system is based on the decay heat generated by a full core (193 assemblies) removed
from the reactor and stored in the fuel storage pool 106 hours following a reactor shutdown, while the spent fuel assemblies from all of the previous (i.e., 2642-193=2449 assemblies) refuelings remain in the fuel storage pool. The fuel pool cooling system is controlled manually. Assuming that one fuel pool cooling train fails, the fuel storage pool is large enough that an extended period of time is required for the water to heat up significantly, if cooling were interrupted. Therefore, there is sufficient time for the operator to manually switch to the backup cooling train. Table 9.1-4 contains the heatup rates for the design basis conditions. Table 9.1-3 indicates compliance
with the Regulatory Guide 1.13 position. SAFETY EVALUATION NINE - The fuel pool cooling piping and other piping penetrating the fuel storage pool enter and terminate near the normal pool water level to preclude the possibility of draining the pool. The cooling
water return lines, which terminate at nominal elevation 2038 feet of the fuel storage pool, each contain a vent hole near the normal spent pool storage water level, so that the pool cannot be siphoned. SAFETY EVALUATION TEN - The redundant seismic Category I essential service water system intertie with the fuel storage pool ensures adequate fuel storage
pool makeup water, considering the maximum anticipated evaporation rates of the fuel storage pool water, as given in Table 9.1-4. SAFETY EVALUATION ELEVEN - As described in Section 9.1.3.5, a monitoring capability is provided to verify fuel storage pool level and bulk temperature. 9.1-26 Rev. 20 WOLF CREEK 9.1.3.4 Tests and Inspections Preoperational testing is discussed in Chapter 14.0. Provisions are incorporated in the design to allow for periodic starting of the nonoperating pump for verification of the required cooling flowpath. These operations demonstrate the operability, performance, and structural and leaktight integrity of all FPCCS components. The safety-related components of the system, i.e., pumps, valves, heat exchangers, and piping (to the extent practicable), are designed and located to permit preservice and inservice inspections. 9.1.3.5 Instrumentation Applications The instrumentation provided for the fuel pool cooling and cleanup system is discussed below. Alarms and indications are provided as noted. a. Temperature Instrumentation is provided to measure the temperature of the water in the spent fuel pool and give main control room indication as well as annunciation when normal temperatures are exceeded. Instrumentation is also provided to indicate the temperature of the fuel pool water as it leaves each heat exchanger.
- b. Pressure Instrumentation is provided to measure and give local indication of the pressures in the suction and discharge lines of the fuel pool cooling pumps. Local pressure indication is provided on the discharge of the fuel pool cleanup pumps and fuel pool skimmer pump.
Differential pressure instrumentation is also provided at
the fuel pool cleanup demineralizer and filters and the fuel pool skimmer filter so that the pressure differential across these components can be determined. 9.1-27 Rev. 14 WOLF CREEK
- c. Flow
Instrumentation is provided to measure and give local and main control room indication of the flow in the outlet line of the fuel pool cleanup pumps. A low-flow alarm is
located in the main control room, in addition to a local
low-flow alarm.
Instrumentation is also provided to measure and give local and main control room indication of the flow in the
discharge lines of the fuel pool cooling pumps.
- d. Level A Class 1E level switch is provided to protect each fuel pool cooling pump from loss of suction on low water level in the fuel storage pool. Instrumentation is also provided to measure the water level of the fuel storage pool and give local and main control room indication and annunciation of high or low pool levels.
A non-Class 1E wire-guided wave radar level instrument is provided to reliably monitor the spent pool water level under adverse environmental conditions resulting from a beyond design basis external event (BDBEE) as described in Appendix 3D. Redundant (designated Primary and Backup) level indications are located in the Auxiliary Building HVAC area, room 1512 for the Primary indication and room 1501 for Backup indication. The level indications provide a diverse, wide-range indication of spent fuel pool levels. 9.1.4 FUEL HANDLING SYSTEM
The fuel handling system (FHS) provides a safe means for handling fuel
assemblies and control components from the time of receipt of new fuel
assemblies to shipment of spent fuel. This includes equipment necessary for reactor vessel servicing.
Design considerations include maintaining occupational radiation exposures
ALARA during transportation and handling.
The fuel handling system is composed of cranes, equipment, special fuel handling devices, and a fuel transfer system that are designed to meet the
seismic and safety classifications shown in Section 3.2.
9.1.4.1 Design Bases 9.1.4.1.1 Safety Design Bases
The portions of the FHS that are safety related are the containment isolation
features of the fuel transfer tube and the crane structural components which
prevent the falling of major crane components onto fuel assemblies or safe shutdown equipment.
9.1-28 Rev. 29 WOLF CREEK SAFETY DESIGN BASIS ONE - The FHS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The FHS is designed to remain intact after an SSE or following the postulated hazards of fire, internal missiles, or pipe breaks (GDC-3 and 4). SAFETY DESIGN BASIS THREE - The FHS components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection and testing of components at appropriate times. SAFETY DESIGN BASIS FOUR - The FHS is designed and fabricated to codes consistent with the seismic category assigned by Regulatory Guide 1.29 and industry standard specifications. SAFETY DESIGN BASIS FIVE - The containment isolation provisions for the system are selected, tested, and located in accordance with the requirements of GDC-54 and 10 CFR 50, Appendix J, Type B testing. SAFETY DESIGN BASIS SIX - The FHS is designed and arranged so that there are no loads which, if dropped, could result in damage, leading to the release of radioactivity in excess of 10 CFR 100 guidelines, or impair the capability to safely shut down the plant. This meets the requirements of Regulatory Guide 1.13 and excludes the system from the requirements of Regulatory Guide 1.104. 9.1.4.2 System Description 9.1.4.2.1 General Description The fuel handling system consists of the equipment needed to refuel the reactor core. Basically, this equipment is composed of cranes, handling equipment, and a fuel transfer system. The associated fuel handling structures are divided into seven areas. In general, these areas are:
- a. The refueling pool
- b. The fuel transfer canal 9.1-29 Rev. 0 WOLF CREEK
- c. The spent fuel pool
- d. The shipping cask loading pool
- e. The cask washdown pit
- f. The new fuel storage vault
- g. The new fuel receiving and inspection area Figures 9.1-4 through 9.1-17 show equipment configurations and the areas of movement of the spent fuel and cask handling cranes.
The new fuel assemblies are removed one at a time from the shipping container utilizing a new fuel handling tool suspended from the monorail on the cask
handling crane, inspected in the new fuel inspection area, and stored in the new fuel storage racks within the new fuel storage facility or moved to the new fuel elevator where it is lowered and transferred to a fuel storage pool storage location. The new fuel is moved from its storage rack or inspection area, utilizing the new fuel handling tool suspended on the monorail hoist on the cask handling
crane, and transferred to the new fuel elevator. The new fuel elevator is used to lower the new fuel assemblies into the cask loading pool. The new fuel is moved from the new fuel elevator, utilizing the spent fuel handling tool with
the spent fuel bridge crane, and either placed in the fuel storage pool or transferred to the upended fuel containers located in the transfer canal, and then moved through the fuel transfer tube to the refueling pool where it is handled by the refueling machine. The fuel transfer system includes a rod cluster control (RCC) storage rack and refueling machine. These facilitate the exchange of control rods between spent
fuel and new fuel. The RCC storage rack may be used for temporary storage of new fuel during the refueling operations. Spent fuel is removed from the reactor with the refueling machine, transferred to the fuel storage pool by the fuel transfer system and the spent fuel pool bridge crane, and deposited in a fuel storage pool rack. In the fuel storage pool, fuel assemblies are moved by the spent fuel bridge crane. When lifting spent fuel assemblies, a long-handled tool is used to
ensure that sufficient radiation shielding is maintained. 9.1-30 Rev. 14 WOLF CREEK After a sufficient decay period in the fuel storage pool (90-150 days), the spent fuel may be removed from the storage racks and transferred to the spent fuel shipping cask within the cask loading pool. The cask is sealed in the loading pool and then is decontaminated in the washdown pit to meet applicable transportation regulations and shipped off site. In order to meet Department of Transportation regulations, dose rates below the maximum of 200 mrem/hr at the surface of the transporting vehicle and 10 mrem/hr at 6 feet from the surface must be attained prior to shipping. Reactor servicing consists of those operations necessary to support refueling, maintenance, and inservice inspection. 9.1.4.2.2 Component Description Principal codes and standards applicable to the FHS are listed in Tables 3.2-1 and 9.1-7 (Sheet 2). REFUELING MACHINE - The refueling machine is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling pool. The bridge spans the refueling pool and runs on rails set into the pool edges. The bridge
and trolley motions are used to position the vertical mast over a fuel assembly in the core. The bridge, trolley, main hoist, and hoist controls are interlocked through the use of the same control panel. A long tube with a pneumatic gripper on the end is lowered from the mast to grip the fuel assembly. The gripper tube is long enough so that the upper end is contained
in the mast when the gripper end contacts the fuel assembly. A winch mounted
on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel is transported to its new position while inside the mast tube. All controls for the refueling machine are mounted on a removable console on the trolley. The bridge and trolley are positioned in relation to a grid pattern referenced to the core. Bridge and trolley positions are indicated by a pointer-ruler system, and absolute encoders. The outer mast is mounted on the trolley structure on a support bearing that allows rotation of the mast to allow a fuel assembly that is not properly
oriented with the core position to be picked up and rotated into proper alignment. In the event a fuel assembly must be turned, the stops can be disconnected and the mast turned manually. With the mast rotated from normal
operating position, the hoist is run at slow speed. 9.1-31 Rev. 14 WOLF CREEK Fuel assemblies can be placed in the core in only one way relative to the core centerlines. Orientation of the fuel is maintained by the gripper which can
engage the fuel only when the relative orientation is correct. Indications are observed by the operator at the console. The drives for the
bridge and trolley are variable speed. The maximum speed for the bridge is
approximately 60 feet per minute. The trolley and hoist maximum speeds are
both approximately 40 feet per minute. The auxiliary monorail hoist on the refueling machine has a variable speed controller to give hoisting speeds of 0 to 20 feet per minute.
Electrical interlocks, Programmable Logic Controller, and resolvers on the
bridge and trolley drives prevent damage to the fuel assemblies. The winch is also provided with limit switches plus a mechanical stop to prevent a fuel assembly from being raised above a safe shielding depth should the limit switch
fail. In an emergency, the bridge, trolley, and winch can be operated
manually, using a handwheel, or equivalent, on the motor shaft. Suitable restraints are provided between the bridge and trolley structures and their respective rails to prevent derailing.
A conservative design approach is used for all load-bearing parts. The static
design load for the crane structure and all lifting components is normal dead
and live loads plus three times the fuel weight with a RCC assembly inserted. The design load on the wire rope hoisting cables does not exceed 0.20 times the average breaking strength. Where two cables are used, each is assumed to carry
one-half the load.
A single finger on the fuel gripper can support the weight of a fuel assembly and RCC assembly without exceeding the requirements given in Table 9.1-7.
All components critical to the operation of the equipment or located so that
parts can fall into the reactor are assembled with the fasteners positively
restrained from loosening under vibration. The refueling machine design includes the following provisions to ensure safe
handling of fuel assemblies:
9.1-32 Rev. 27 WOLF CREEK
- a. Safety interlocks Operations which could endanger the operator or damage the fuel are prohibited by mechanical or fail-safe electrical interlocks, or by redundant electrical interlocks. All other interlocks are intended to provide equipment protection and may be implemented either mechanically or by electrical interlock, not necessarily fail-safe. The following interlocks are provided on the refueling machine: 1. When the gripper is engaged, the machine cannot move outside the core region unless the guide tube is in its full up position.
- 2. When the gripper is disengaged, the machine cannot traverse unless the gripper is withdrawn into the
mast.
- 3. Vertical motion of the guide tube is permitted only in a controlled area over the reactor (avoiding the
vessel guide studs), fuel transfer system, rod
cluster control change fixture, gripper disengage plate or load testing fixture.
- 4. Traverse of the trolley and bridge is limited to the areas of item 3 and a clear path connecting those
areas.
- 5. A key-operated interlock bypass switch is provided to defeat interlocks 1 through 4. That switch also operates a flashing red light to indicate that the interlocks are bypassed.
- 6. The gripper is monitored by limit switches to confirm operation to the fully engaged or fully
disengaged position. An audible and a visual alarm are actuated if both engaged and disengaged switches are actuated at the same time or if neither is actuated. A time delay may be used to allow for recycle time of normal operation.
- 7. The loaded fuel gripper will not release unless it is in its down position in the core, or in the fuel transfer system or rod cluster control change fixture, and the weight of the fuel is off the mast. 9.1-33 Rev. 13 WOLF CREEK
- 8. Raising of the guide tube is not permitted if the gripper is disengaged and the load monitor indicates that it is still attached to the fuel assembly.
- 9. Raising of the guide tube is not allowed by two overload interlocks if the hoist loading exceeds specified values. The setpoints for the two inter-locks are specified as the force in lbs. above the total weight of guide tube plus fuel assembly plus rod cluster control assembly. For wet conditions the setpoint for the overload interlock is less than or equal to 250 lbs. and the setpoint for the master
overload is less than or equal to 150 lbs. above the overload setting. At 2 inches above the full down position in the core, upender, RCCA change fixture, overload interlock will be bypassed.
- 10. Lowering of the guide tube is not allowed by an underload interlock if the hoist loading falls below
a specified value. The setpoint for the interlock
is specified as the force in lbs. below the total
weight of the guide tube plus fuel assembly plus lightest component. The setpoint for the underload is less than or equal to 250 lbs. At 2 inches above the full down position in the core, upender, and RCCA change fixture, the underload interlock will be bypassed.
- 11. The guide tube is prevented from rising to a height where there is less than 10 feet of nominal water
coverage over the fuel.
- 12. The guide tube is prevented from lowering completely out of the mast.
- 13. The guide tube travels only at a controlled speed of about 2.5 fpm when: a) the bottom of the fuel begins to enter the core, and b) the gripper approaches the top of the core. In addition, just
above those points, the guide tube automatically stops lowering, and requires acknowledgment from the operator before proceeding. If the guide tube is in open water (greater than two inches in both X and Y directions, or greater than the width of a fuel assembly
plus two inches in one direction from any fuel assembly), the guide tube may operate at maximum speed.
- 14. The fuel transfer system lifting arm is prevented from moving unless the loaded gripper is in the full up position or the unloaded gripper is withdrawn into the mast, or unless the refueling machine is out of the upender zone. An interlock is provided from the refueling machine to the fuel transfer
system to accomplish this. 9.1-34 Rev. 14 WOLF CREEK
- b. Bridge and trolley holddown devices
The refueling machine bridge and trolley are both horizontally restrained on the rails by two pairs of guide rollers, one pair at each wheel location on one
truck only. The rollers are attached to the bridge truck
and contact the vertical faces on either side of the rail
to prevent horizontal movement. Vertical restraint is accomplished by antirotation bars located at each of the four wheels for both the bridge and trolley. The
antirotation bars are bolted to the trucks and extend
under the rail flange. Horizontal and vertical
restraints are both adequately designed to withstand the forces and overturning moments resulting from the SSE.
- c. Main hoist braking system
The main hoists are equipped with two independent braking systems. A solenoid-release, spring-set electric brake is mounted on the motor shaft. This brake operates in
the normal manner to release upon application of current
to the motor and set when current is interrupted. The
second brake is a mechanically actuated load brake internal to the hoist gear box that engages if the load starts to overload the hoist. It is necessary to apply
torque from the motor to raise or lower the load. In
raising, this motor cams the brake open; in lowering, the motor slips the brake, allowing the load to lower. This brake actuates upon loss of torque from the motor for any reason and is not dependent on any electrical
circuits. Both brakes are rated at 125 percent of the
hoist design load.
- d. Fuel assembly support system
The main hoist system is supplied with redundant paths of
load support so that failure of any component will not
result in free-fall of the fuel assembly. Two wire ropes are anchored to the winch drum and carried to a load-equalizing mechanism on the top of the gripper tube. In
addition, supports for the equalizing mechanism are
backed up by passive restraints to pick up the load in
the event of the failure of this primary support. During each refueling outage and prior to removing fuel, the gripper and hoist
system are load tested to 125 percent of the maximum setting on the secondary
hoist load limit.
The in-Mast sipping system is a set of hardware that provides the means for performing on line, qualitative leak testing of irradiated fuel assemblies during normal core off load operations. The hardware includes adapters for the refueling machine mast to inject air at the bottom of the mast. A mean is provided to sample the fission gasses at the top of the mast and direct the sample to the portable analysis equipment. The air injection manifold is a two piece ring bolted to the bottom of the stationary mast to support the four air injection nozzles. The purpose of this manifold is to inject a stream of bubbles into the mast once the fuel assembly is in the up position. The stream of bubbles will strip off and fission gasses migrating out of the assembly and carry them to the top of the mast for collection. 9.1-35 Rev. 27 WOLF CREEK The air collection manifold is a single piece ring mounted on the top of the stationary mast to provide a means to collect and sample the fission gasses from the fuel assembly. The final mast components are the roller covers to close off openings around the guide rollers in the stationary mast and air supply tubing mounted on the outside of the stationary mast to connect the air nozzle manifold to the air supply. CASK HANDLING CRANE - The cask handling crane is a Crane Manufacturers Association of America (CMAA) No. 70, Class A, indoor electrical overhead
traveling bridge crane with a single trolley and all the necessary motors, controls, and brakes, and a festooned pendant control station. The crane hoist
is rated at 150 tons. The crane and accessories are used to handle spent fuel shipping casks between the railroad cars or trucks, the loading pool, and the washdown pit.
The main hoist and the main bridge trolley have an inching feature for
positioning of the crane at desired locations. The cask handling crane is equipped with a monorail and hoist which is used to
transfer new fuel from the new fuel storage vault to the new fuel elevator.
The monorail is also used for moving new fuel shipping containers. The
monorail hoist is rated at 5 tons. The festooned pendant control station or radio control unit is utilized for controlling the cask handling crane and the monorail hoist.
The handling tool of the cask handling crane is designed to prevent a shipping
cask from dropping into the spent fuel pool. Under normal use, limit switches and mechanical stops are located to prevent
any crane (other than the spent fuel pool bridge crane) from traveling over the
spent fuel pool. During scheduled maintenance periods, the cask handling crane
is used to provide access over the spent fuel pool, for example, for servicing of light bulbs and fire detectors. During these periods, the rail stops are removed to allow crane travel. These rail stops, which are not heavy loads, are hinged such that they can be rotated out of the path of the cask handling
crane. The hinged connections are outside the crane rails and the stops rotate
away from the center of the fuel building to allow crane travel. These stops do not require lifting to clear the cask handling crane, but are permanently attached to the crane rail support girder, thus precluding a drop.
Administrative procedures are used to control removal and replacement of the
interlock and stops and to position the hoist and hook so as not to travel
above the pool during use of the cask handling crane above the pool. Geared-type upper and lower limit switches are used in the control circuit of
each hoist system of the fuel building cask handling crane. In addition to the
geared-type limit switches, a weight-operated hoist upper limit switch is used
in each hoist system of the cask handling crane. The two types of hoist upper
limit switches are redundant and independent. If the geared-type limit switch were to fail, the weight-operated limit switch would cut off power to the hoist, thus preventing vertical motion of the lifting block and the occurrence
of a two-blocking event.
Specific data for the cask handling crane travel speeds and lifting capacities are shown on Table 9.1-7.
9.1-36 Rev. 27 WOLF CREEK SPENT FUEL POOL BRIDGE CRANE - The spent fuel bridge crane is a CMAA No. 70, Class B type. The crane is designed to maintain its integrity during an SSE. The crane consists of a 5-ton-capacity wheeled bridge structure with steel deck walkway, a 2-ton motorized monorail trolley, and a 5-ton manual push-type trolley. The crane has interlocking capabilities with the new fuel elevator, fuel storage pool transfer gate, and cask loading gate. The crane also has a 1/4-inch bridge and trolley positioning capability. The spent fuel bridge crane is used to transport new and spent fuel to and from various locations inside the fuel building. These locations include the new
fuel elevator, fuel storage racks, spent fuel shipping cask, upending device of the fuel transfer car, and fuel storage pool transfer gates. The handling tools for the new and spent fuel are different to prevent interchanging of the
same. The hoist travel and tool length are designed to limit the maximum lift
of a fuel assembly to a safe shielding depth. The 2-ton electric hoist of the crane is primarily used to transfer spent fuel and new fuel assemblies. Control is from a pendant station supported from the trolley. The 5-ton manual chain hoist and trolley are used to move the fuel storage pool transfer gates to and from their normal storage positions. The hoists share the same monorail. While moving the transfer gates, the gates are secured by a redundant support to preclude the dropping of a gate on the fuel storage racks. The spent fuel pool bridge crane has a limited maximum lift height so that the
minimum required depth of the water shielding is maintained when the spent fuel
is handled. This is accomplished by the use of limit switches. Geared-type upper and lower limit switches are used in the control circuit of the electric hoist system of the spent fuel pool bridge crane. In addition to the geared-type limit switches, a weight-operated hoist upper limit switch is used in the electric hoist system of the spent fuel pool bridge crane. The two types of hoist upper limit switches are redundant and independent. If the geared-type limit switch were to fail, the weight-operated limit switch would cut off power to the hoist, thus preventing vertical motion of the lifting
block and the occurrence of a two-blocking event. Specific data pertaining to the travel speeds are shown on Table 9.1-7. 9.1-37 Rev. 14 WOLF CREEK CONTAINMENT BUILDING POLAR CRANE - The polar crane is a CMAA No. 70, Class C type.The containment has a 260/25-ton polar crane which is used, in conjunction with the various lifting rigs, to remove the Simplfied Head Assembly (SHA), the reactor vessel upper internals, and the lower internals. The 25-ton auxiliary hook on the polar crane, in conjunction with strategically located 3-ton-capacity jib cranes, is used for routine maintenance and inservice inspection. The crane is controlled from its bridge-mounted cab or a portable radio control unit. The polar crane is used during plant outages for maintenance activities, in plant modes 3 to 6. The heaviest load lifted is the SHA. The polar crane is designed to maintain its structural integrity under operating basis
earthquake (OBE) and safe shutdown earthquake (SSE) conditions. An analysis has been performed to show that the polar crane will not derail under an OBE during plant operation. It is also designed to carry 260 Ton under an SSE, generally during outages. The polar crane bridge is equipped with seismic restraints (snubbers), one in each corner of the crane girders G1 and G2 (corner #1 is snubber #1,corner #2 is snubber #2, etc), refer to Figures 9.1-24 to 9.1-26. Girder G1 is between corner #1 and corner #2 and Girder G2 is between corner #3 and corner #4. Each snubber consists of two wheels, each wheel contained in a frame. The two frames are pinned into a holding frame and thus are able to move with respect to each other. The wheels are pushed toward each other by a spring loaded hydraulic snubber. Thus, the wheels rest against the face of the girder flange
on which the crane rests. In case of a seismic event (OBE or SSE), the wheels will stay in contact with the girder flange face, while the shock absorbers
prevent the crane from moving more than 1/4 of an inch in the horizontal plane. The snubber wheels will be in contact with the face of the girder flange during plant modes 1 and 2. The snubbers in corners #1 to #4 may be retracted 3/4 in all plant modes. Structural integrity during an SSE with the snubbers retracted to 3/4 has been analyzed in all plant modes. Vertical motion of the polar crane is restrained through the use of upkick lugs on the snubber frame, which project under the girder flange face, during seismic events. Positive means are also provided to limit motion of the polar crane trolley during a seismic event. Trolley earthquake restraints are provided to limit
vertical motion of the trolley. These restraints are attached to both sides of the trolley girders (G1 and G2) and project under the flanges supporting the rails on which the trolley runs. To help limit horizontal motion of the trolley during a seismic event, rail capture bars are provided. The main hoist of the polar crane has a micro-drive, which enables the operator to move the main hoist hook at a speed of 3 inches per minute. The auxiliary hoist of the polar crane has an "inching" feature, which allows the operator to raise or lower the load at approximately 1/16 of an inch increments. Geared-type upper and lower limit switches are used in each hoist system of the containment building polar crane. The geared-type limit switch is driven off the hoist drum shaft through an eccentric pin and crank arrangement. As the switch drive shaft rotates, it rotates two cam gears. Cam screws lock the cam
wheels to their respective cam gears. Snap switches are actuated when the
lobes on their associated cam wheels contact the switch 9.1-38 Rev. 23 WOLF CREEK pushers. Snap switches open or close the contacts, thereby breaking or completing the electrical circuit to the hoist motor and holding brake. In addition to the geared-type limit switches, a weight-operated hoist upper limit switch is used in each hoist system of the containment building polar crane. This assembly is used as a safety-type limit switch or final upper stop to prevent over-travel of the hook block as it approaches its upper limit. Thus, the block and load are prevented from coming into contact with any portion of the trolley, and an unsafe condition is avoided (two blocking event).The two types of hoist upper limit switches, geared and weight-operated, are redundant and independent. This is to say that if the geared-type limit switch were to fail, the weight-operated limit switch would stop the load block from rising higher and would prevent the occurrence of a two-blocking event. The polar crane main and auxiliary hooks are administratively controlled by procedure to prevent travel over the reactor vessel in all modes except cold shutdown and refueling. Once the upper internals have been removed and fuel is in the reactor vessel, crane hook travel will be prohibited over the open vessel except for the occasional need for reversing the orientation of the
main/auxiliary hoists and for required vessel servicing activities such as
irradiation sample removal. When there is fuel in the vessel, administrative procedures will not allow raising or lowering the hook while traveling over the open vessel to reverse the hoist orientation and the only item attached to
either hook may be the load cell linkage attacked to the main hook. Note that during irradiation sample removal, the loads on the hoist hook, which are
carried over the vessel are light (less than 600 pounds). Specific data pertaining to the crane travel speeds and lifting capacity are shown on Table 9.1-7. FUEL TRANSFER TUBE AND ASSOCIATED COMPONENTS - The fuel transfer system permits the safe underwater transfer of new and spent fuel assemblies between the fuel transfer canal in the fuel building and the refueling pool in the reactor
building. Connecting these two areas is the fuel transfer tube which is a steel pipe 20 inches outside diameter and approximately 20 feet long. The pipe is inserted in a sleeve which is embedded in the concrete walls separating the two areas. Angle rails forming a track and extending from the refueling canal through the transfer tube and into the transfer canal permit the controlled travel of the
fuel car. During the fuel transfer operations, the fuel assemblies are supported by the fuel car. Attached to the car is the transfer car container which holds the fuel assembly. This container is a tube and is equipped with a
centrally located pivot which allows the fuel assembly to be rotated from a
vertical to a horizontal orientation for easier transfer. The fuel transfer
car and container assembly travel through the transfer tube as one unit. 9.1-39 Rev. 22 WOLF CREEK Positioned at the fuel bldg side of the transfer tube are mechanical stops. Water-activated hydraulic lifting arms which are the mechanisms that allow the fuel assembly to be pivoted are provided at both ends of the transfer tube. Each hydraulic drive is operated by a hydraulic pump. The travel of the fuel assembly, transfer car, and container is achieved by the use of a pusher arm. This arm is connected to two stainless steel cables near
the floor of the fuel transfer canal. These cables are driven by a motor-winch assembly and directed by a series of sheaves so that the cables will push the transfer car in one direction and pull in the other direction to move the container from the fuel building to the reactor building and vice versa. The motor-winch assembly is located near the operating floor of the fuel building. The fuel transfer car is equipped with an emergency pullout cable to withdraw the car from the transfer tube should a system breakdown occur. During reactor operation, the transfer car is stored in the fuel storage area. A blind flange is bolted on the reactor building cavity end of the transfer tube to seal the reactor containment. The terminus of the tube outside the containment is closed by a gate valve in the fuel building. The following safety features are provided in the fuel transfer system during operation with NO bypass funcitons in effect:
- a. Transfer car permissive switch The transfer car controls are located in the fuel storage area, and conditions in the containment are, therefore, not visible to the operator. The transfer car permissive switch allows a second operator in the containment to stop the car movement if conditions visible to him
warrant such control. Transfer car operation is possible only when both lifting arms are in the down position, as indicated by the limit
switches. The permissive switch is a backup for the
transfer car lifting arm interlock. Assuming that the
fuel container is in the upright position in the containment and the lifting arm interlock circuit fails in the permissive condition, the operator in the fuel
storage area still cannot operate the car because of the
permissive switch interlock. The interlock, therefore, can withstand a single failure. 9.1-40 Rev. 12 WOLF CREEK b. Lifting arm (transfer car position) Two redundant interlocks allow a lifting arm operation only when the transfer car is at the respective end of its travel and, therefore, can withstand a single failure. Of the two redundant interlocks which allow a lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device on the transfer car that is opened by the car moving into position. c. Transfer car (valve open) An interlock on the transfer tube valve permits transfer car operation only when the transfer tube valve position switch indicates that the valve is fully open. d. Transfer car (lifting arm) The transfer car lifting arm is primarily designed to protect the equipment from overload and possible damage if an attempt is made to move the car when the fuel container is in the vertical position. This interlock is redundant and can withstand a single failure. The basic interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device that is opened by the weight of the fuel container when it is in the horizontal position. e. Lifting arm (refueling machine) The refueling canal lifting arm is interlocked with the refueling machine. The fuel transfer system lifting arm is prevented from moving unless the loaded gripper is in the full up position or the unloaded gripper is withdrawn into the mast, or unless the refueling machine is out of the upender zone. f. Lifting arm (spent fuel pool bridge crane (SFPBC)) The lifting arm is interlocked with the SFPBC. The lifting arm cannot be operated unless the spent fuel pool bridge crane 2 Ton electric hoist is not over the lifting arm area. 9.1-41 Rev. 14 WOLF CREEK ROD CLUSTER CONTROL CHANGING FIXTURE - The RCC changing fixture is used for periodic RCC element inspections and for the transfer of RCC elements from one fuel assembly to another. The major subassemblies which comprise the changing fixture are the frame and track structure, the carriage, the guide tube, the gripper, and the drive mechanism. The carriage is a movable container supported by the frame and track structure. The tracks provide a guide for the four flanged carriage wheels and allow horizontal movement of the carriage during the changing operation. The positioning stops on the carriage and frame locate each of the three carriage compartments directly below the guide tube. Two of these compartments are designed to hold individual fuel assemblies while the third is made to support a single RCC element. The guide tube is situated above the carriage and is mounted on the refueling canal wall. The guide tube provides for the guidance and proper orientation of the gripper and RCC element as they are being raised and lowered. The gripper is a pneumatically actuated mechanism responsible for engaging the RCC element.
It has two flexure fingers which can be inserted into the top of the RCC
element when air pressure is applied to the gripper piston. Normally, the fingers are locked in the radially extended position. Mounted on the operating deck is the drive mechanism assembly which is composed of the manual carriage
drive mechanism, the revolving stop operating handle, the pneumatic selector valve for actuating the gripper piston, and the electric hoist for elevation control of the gripper. NEW FUEL ELEVATOR - The new fuel elevator consists of a box-shaped assembly with its top end open. The elevator is sized to house only one fuel assembly. It is located on the wall of the cask loading pool and is used primarily to lower a new fuel assembly to the pool bottom. It is also used to support other activities and may contain an irradiated fuel assembly, fuel rodlet container, or other equipment or components weighing less than the design value for the elevator. When it is at the bottom of the pool, the fuel assembly is transported to either the fuel storage pool storage racks or to the container of the fuel transfer car by the use of the spent fuel pool bridge crane. SPENT FUEL ASSEMBLY HANDLING TOOL - The spent fuel assembly handling tool, also referred to as the long-handling tool, is used to manually handle the new and
spent fuel in the fuel storage pool. An operator on the spent fuel pool bridge crane guides and operates the tool. The tool is designed to maintain its integrity during an SSE. The tool employs four cam-actuated latching fingers which grip the underside of the fuel assembly top nozzle. When the fingers are 9.1-42 Rev. 14 WOLF CREEK latched, a lock pin is inserted into the operating handle to prevent the fingers from being accidentally unlatched during fuel handling operations. The tool weighs approximately 376 pounds and is preoperationally tested at 125 percent the weight of one fuel assembly with control assembly inserted (1,600 pounds).NEW FUEL ASSEMBLY HANDLING TOOL - The new fuel assembly handling tool is a short-handled device located on the cask handling crane monorail. It is used to handle new fuel on the operating deck of the fuel building, to remove the new fuel from the shipping container, and to facilitate inspection and storage of the new fuel and loading of fuel into the new fuel storage racks and the new
fuel elevator. The new fuel assembly handling fixture employs four cam-actuated latching fingers which grip the underside of the fuel assembly top nozzle. When the
fingers are latched, the safety mechanism on the side of the tool is turned in
to prevent accidental unlatching of the fingers. The tool weighs approximately 80 pounds and is preoperationally tested at 125 percent the weight of one fuel assembly with control rod inserted (1,600
pounds).REACTOR CAVITY SEAL RING - A permanent, watertight reactor cavity seal ring is mounted between the reactor vessel flange and the cavity liner at the bottom of the refueling pool. The permanent cavity seal ring (PCSR) covers the annulus around the vessel permitting the cavity to be flooded for refueling. The PCSR is designed to remain in place during all plant operations as well as during refueling. Access covers are provided to allow a ventilation flow path during reactor operation. The PCSR is not an ASME Code class item and it is classified as a non-nuclear safety class item in accordance with ANSI N18.2 (1973).9.1.4.2.3 System Operation The fuel handling equipment is designed to handle the spent fuel assembly under water from the time it leaves the reactor vessel until it is placed in a
container for shipment from the site. Underwater transfer of spent fuel assemblies provides an effective, economic, and transparent radiation shield, as well as a reliable cooling medium for the removal of decay heat. 9.1-43 Rev. 6 WOLF CREEK Fuel is moved between the reactor vessel and the refueling canal by the refueling machine. A RCC changing fixture is located in the refueling canal for transferring control elements from one fuel assembly to another. The fuel transfer system is used to move fuel assemblies between the containment building and the fuel storage building. After a fuel assembly is placed in the fuel container, the lifting arm pivots the fuel assembly to the horizontal position for passage through the fuel transfer tube. The fuel transfer tube is fitted with a blind flange on the refueling pool end and a gate valve on the fuel transfer canal end. After the transfer car transports the fuel assembly through the transfer tube, the lifting arm at that end of the tube pivots the assembly to a vertical position so that the assembly can be lifted out of the fuel container. During nonrefueling operations, a blind flange seals the containment side of the transfer tube in order to ensure the leaktight integrity of the
containment. Two 0-ring seals are located around the periphery of the blind flange with leak-check provisions between them in order to perform a Type B test per 10 CFR 50, Appendix J. In the fuel storage building, fuel assemblies are moved about by the fuel handling machine. When lifting fuel assemblies, the hoist uses a long-handled tool to assure that sufficient radiation shielding is maintained. A shorter tool is used to handle new fuel assemblies initially, but the new fuel elevator
must be used to lower the assembly to a depth at which the fuel handling machine, using the long-handled tool, can place the new fuel assemblies into or out of the fuel storage racks. In MODES 5, 6, and defueled, during spent fuel pool bridge crane operations, one offsite AC circuit and one EDG are required to be operable. When either
the required offsite AC circuit or the required EDG is determined to be
inoperable, operation of the spent fuel pool bridge crane is suspended. The suspension of movement of the spent fuel pool bridge crane shall not preclude completion of movement of the component to a safe position. Decay heat, generated by the spent fuel assemblies in the fuel storage area, is removed by the fuel pool cooling and cleanup system. After a sufficient decay period, the spent fuel assemblies may be removed from the fuel racks and loaded into shipping containers for removal from the site. 9.1.4.2.3.1 Fuel Handling System Operations NEW FUEL RECEIVING AND STORAGE - New fuel assemblies are delivered to the site by truck or rail in approved containers. New fuel containers are removed from
the truck or railcar in the fuel shipping and unloading area of the fuel
building and then, with the use of the new fuel handling tool suspended on the
monorail hoist on the cask handling crane, are moved to the new fuel inspection
area (a strongback is used initially to 9.1-44 Rev. 18 WOLF CREEK upend and prevent the bowing of the new fuel assembly). While the new fuel assembly is in this area, any shipping spacers present are removed, the cleanliness is verified, and the assembly is visually inspected for any damage. Following inspection, the new fuel is transferred by means of the new fuel handling tool suspended from the monorail hoist on the cask handling crane to the new fuel storage racks in the new fuel storage vault, or moved to the new fuel elevator where it is lowered and transferred to a fuel storage pool location.REFUELING PROCEDURE - The refueling operation follows a detailed procedure
which provides a safe, efficient, refueling operation. The following
significant points are assured by the refueling procedures: a. The refueling water and the reactor coolant are maintained at no less than the boron concentration specified in the COLR. This concentration, together with the negative reactivity of the control rods, is sufficient to keep the core approximately 5-percent /k subcritical during the refueling operations. It is also sufficient to maintain the core subcritical in the unlikely event that all of the RCC assemblies were removed from the core.
- b. The water level in the refueling pool is high enough to keep the radiation levels within acceptable limits when
the fuel assemblies are being removed from the core. The refueling operation is divided into five major phases: Phase I - Preparation Phase II - Reactor disassembly Phase III - Fuel handling
Phase IV - Reactor reassembly Phase V - Preoperational checks and startup Phase I - Preparation The reactor is shut down, borated, and cooled to cold shutdown conditions (< 140 F) with a final k eff <0.95 and all rods inserted. Following a radiation survey, the containment is cleared for entry. The reactor coolant system level is lowered to a point slightly below the reactor vessel flange. The fuel transfer equipment, refueling machine and polar crane are checked for proper operation. 9.1-45 Rev. 18 WOLF CREEK Phase II - Reactor Disassembly Prior to reactor vessel head disassembly, several items must be disconnected and removed. The tie rods which anchor the control rod drive mechanism (CRDM) seismic support platform to the refueling pool walls are disconnected. All cables connected to the reactor vessel head (rod position indication, CRDM power cables, upper instrumentation thermocouple leads, upper head loose parts monitoring leads, and head vent valves) are disconnected. The power cables to the CRDM fans are disconnected. RVLIS is disconnected. The reactor vessel head shield doors are closed for shielding. The insulation is removed from the vessel head and studs detensioned and removed, guide studs installed and stud holes plugged. The refueling cavity is prepared for flooding by installing blind flanges on the refueling pool drain holes, checking underwater lights, tools and the fuel transfer system with the fuel transfer system blind flange removed. The shield plugs and cover plates on the permanent cavity seal ring are installed. With the refueling cavity prepared for flooding, the reactor vessel head is lifted slightly and lift rig inspections completed. The reactor vessel head is gradually raised to clear CRDM drive shafts and core exit thermocouple bullet noses, cavity flood up is commenced and the reactor vessel head is moved to storage area. The remainder of the refueling cavity is flooded to the normal refueling level specified in Technical Specifications. The control rod drive shafts are disconnected from the RCC assemblies, and the upper internals are removed from the vessel and stored in the refueling cavity. The fuel assemblies and RCC assemblies are now free of obstructions and the core is ready for refueling. 9.1-46 Rev. 18 WOLF CREEK Phase III - Fuel Handling The reactor shall be determined to have been subcritical for at least 76 hours by verification of the date and time of subcriticality prior to movement of
irradiated fuel in the reactor vessel. With the reactor subcritical for less than 76 hours, suspend all operations involving movement of irradiated fuel in
the reactor vessel. This requirement is consistent with the assumptions of Section 15.7.4.5.1.2. Prior to initiation of the refueling sequence, the refueling pool water level is raised to the same level as the fuel storage pool, and the gate valve, which
normally isolates the fuel building side of the transfer tube is opened. In this condition, there is communication between the fuel building pools and the refueling pool; therefore, level monitoring, including a low level alarm, is
provided by the fuel pool cooling and cleanup system. The refueling sequence is started with the refueling machine. Spent fuel assemblies are removed from the core in the sequence prepared by plant personnel before each refueling. During those activities that are defined as core alterations in the Technical Specifications, direct communications between the Control Room and refueling station are demonstrated. Communications is verified within 1 hour prior to the start of core alterations and at least once per 12 hours during core alterations. Without direct communications between the Control Room and the
refueling station, core alterations are suspended. The general fuel handling sequence is:
- a. The refueling machine is positioned over a fuel assembly in the core.
- b. The fuel assembly is lifted by the refueling machine to a predetermined height sufficient to clear the reactor vessel and still leave sufficient water covering to
eliminate any radiation hazard to the operating
personnel. 9.1-47 Rev. 21 WOLF CREEK
- c. The fuel transfer car is moved into the refueling pool from the fuel transfer canal.
- d. The spent fuel assembly is moved by the refueling machine to the fuel transfer car.
- e. The fuel assembly container on the transfer system is pivoted to the vertical position by the upender.
- f. The refueling machine is moved to line up the spent fuel assembly with the fuel assembly container, and the spent fuel assembly is loaded into the assembly container on the transfer system.
- g. The container is pivoted to the horizontal position by the upender.
- h. The fuel and container are moved through the fuel transfer tube to the fuel transfer canal by the transfer
car.
- i. The fuel assembly and container are pivoted to the vertical position. The spent fuel assembly is unloaded
by the spent fuel handling tool attached to the spent fuel pool bridge crane. j. The spent fuel assembly is transferred through the spent fuel storage pool transfer gate, and placed in a designated location in the spent fuel storage racks. k. This sequence continues until the core is off-loaded. The core may be either partially or fully off-loaded. l. Fuel is reloaded into the reactor vessel by reversing the steps necessary for fuel removal. 9.1-48 Rev. 21 WOLF CREEK The reactor is now ready for the reassembly phase. Phase IV - Reactor Reassembly The reactor reassembly, following refueling, is essentially achieved by reversing the operations given in Phase II - Reactor Disassembly. During those activities that are defined as core alterations in Technical Specifications, direct communications between the Control Room and refueling station are demonstrated. Communications is verified within 1 hour prior to the start of core alterations and at least once per 12 hours during core
alterations. Without direct communications between the Control Room and the refueling station, core alterations are suspended. Phase V - Preoperational Checks and Startup
After the refueling pool has been drained, the reactor assembled, and the fuel transfer tube has been isolated, cleanup of the fuel handling areas within the containment building is performed in accordance with the established station
housekeeping procedures. 9.1-49 Rev. 21 WOLF CREEK The blind flanges covering the refueling pool drain holes are removed and stored in designated locations, and the refueling pool strainers are replaced. Any maintenance which is required on fuel handling equipment inside the containment is done during this general cleanup phase of refueling. SPENT FUEL SHIPMENT - The spent fuel assemblies are stored on site, in the fuel storage pool, until fission product inventory and the decay heat is low enough to permit shipment. A spent fuel shipping cask (either by rail or truck) is brought into the shipping/receiving area. The 150-ton cask handling crane upends the cask and places it in the cask washdown pit where it is thoroughly cleaned. The cask head is then disengaged from the top of the cask and stored in its designated storage area. The cask is then lowered into the cask loading pool. The cask loading pool is normally flooded and considered part of the fuel storage pool. The removable
gate is used to isolate the spent fuel pool from the cask loading pool, if a leak should occur in the cask loading pool. The spent fuel handling tool on the spent fuel bridge crane transfers the spent fuel from the fuel storage racks to the cask loading pool and places it into
the shipping cask. After the cask is loaded, it is capped in the pool and then carried to the cask washdown pit for decontamination. The cask is then placed on the railcar or truck bed. The cask is monitored for radioactivity and verification made to
ensure that the provisions of state and federal regulations are met. Spent Fuel Casks (used to remove fuel from the plant) shall not be immersed in, or carried over, the Cask Loading Pool while spent fuel is stored in storage racks within this pool, unless one of the following is performed prior to the
load lift:
- 1. The Cask Handling Crane is upgraded to single failure proof.
- 2. An evaluation is performed to demonstrate the acceptability of damage to the stored fuel, rack modules, pool liner, and structure. 3. Effective means (such as crane stops, limits, barriers or impact limiters, etc.) are implemented to preclude damage to the stored fuel, rack modules, and structure.
9.1.4.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.1.4.1.1. SAFETY EVALUATION ONE - The safety-related portions of the FHS are located in the reactor and fuel 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 FHS are designed to remain intact after an SSE. Section 3.7(B) provides the design loading
conditions that were considered. Sections 3.5, 3.6, and 9.5.1 provide the required hazards analysis. 9.1-50 Rev. 21 WOLF CREEK SAFETY EVALUATION THREE - The FHS is initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.1.4.4. The fuel transfer tube is inspected in accordance with the technical requirements of ASME Section XI. SAFETY EVALUATION FOUR - Section 3.2 delineates the seismic category applicable to the safety-related portions of this system. Table 9.1-7 shows that the components meet the design and fabrication codes given in Section 3.2. SAFETY EVALUATION FIVE - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability. SAFETY EVALUATION SIX - In the event of a fuel handling accident in the fuel building, the radiological consequences analyzed in Chapter 15.0 demonstrate that the 10 CFR Part 100 guideline values are not exceeded. The circumstances
resulting in a handling accident are limited to the following conditions.
- a. Fuel drop from a lifting device
- b. Improper operation of the transfer equipment and cranes
- c. Drop of the fuel shipping cask
- d. Drop of the RV head The fuel handling equipment is designed to prevent a fuel assembly drop by providing special gripping devices which are locked in a manner which will not allow the release of the fuel assembly during transfer. The special features are described in Section 9.1.4.2.2.
Improper operation of the fuel transfer system is prevented by the location of special limit switches and interlocks which will not allow the movement of fuel
assemblies unless they are properly oriented, horizontal, thus avoiding a fuel handling accident. Further description of these devices is given in Section 9.1.4.2.2. Limit switches and interlocks located on the fuel handling cranes, in conjunction with administrative limits, prevent any improper operations which may result in a fuel handling accident. The limiting devices on the refueling machine and spent fuel pool bridge crane do not allow fuel to be moved unless it is in the proper orientation and handled correctly in the gripping tool of
the crane. 9.1-51 Rev. 21 WOLF CREEK Concerning the handling of loads over fuel in the fuel storage pool, administrative controls are employed to prevent the handling of loads that have a greater potential energy than those which have been analyzed. The cask handling crane is restricted from moving the spent fuel cask over the spent fuel pool, the new fuel vault, the fuel pool cooling system, or engineered safety features systems which could be damaged by dropping the cask, and is limited to moving in such a manner as to avoid the possibility of falling or tipping into the spent fuel pool, in accordance with the regulatory position of Regulatory Guide 1.13 and General Design Criterion 61 of Appendix A to 10 CFR 50. The following cask drops, with respect to the cask falling from a vertical or tipped position, have been evaluated:
- a. Dropping the cask into the decontamination (decon) pit with cask center-of-gravity (cg) over the decon pit and
part of the cask over the concrete slab between the cask decon pit and cask loading pool.
- b. Dropping the cask over the concrete slab between the cask decon pit and the cask loading pool with cask cg over the concrete slab.
- c. Dropping the cask into the cask loading pool with cask cg over the cask loading pool and part of the cask over the concrete slab between the decon pit and the cask loading pool.
- d. Dropping the cask directly into the decon pit with cask cg centered on the decon pit.
If the cask is dropped with its cg over the decon pit and with some part over the concrete slab between the decon pit and loading pool, it hits the top east edge of the decon pit. Because the cg of the cask is over the decon pit, it falls in a westerly direction and hits the west edge of the decon pit. The
cask shifts its position until its bottom edge passes into the pit. If the cask is dropped with its cg directly over the edge of the decon pit east wall, it falls in a westerly direction or remains standing. If the cask falls, it falls in the same manner as before. After hitting the west wall of the decon pit, the cask shifts its position until its bottom edge slides past the
concrete slab that is supporting it. After sliding past the concrete slab, the
cask falls into the decon pit. If the cask's pivot point is 9.1-52 Rev. 21 WOLF CREEK not constant as it falls over, its bottom edge slides past the east edge of the pit as its top hits the west edge of the pit. The cask subsequently falls into the pit. If the cask is dropped directly into the decon pit and tilts southward toward the railroad bay areas, a restraining wall prevents a second cask drop to ground level. The wall is designed to withstand the impact of the cask falling southward from a leaning position after it strikes the bottom of the decon pit. If the cask is dropped with its cg over the concrete slab between the decon pit and the loading pool, it remains in a standing position. From a height of 3 inches above the concrete slab (maximum possible distance due to redundant
limit switches), the maximum tipping angle obtainable is approximately 1-1/2 degrees.If the cask is dropped in such a way as to produce the maximum tip angle, the righting moment about the edge touching the concrete slab will right the cask.
The cask remains standing. If the cask is swinging from the crane when dropped, the maximum swing obtainable in the direction of the spent fuel pool before the bottom of the
cask strikes the concrete slab is approximately 4 degrees. If the cask is dropped from this position, the cask rights itself due to the cg of the cask being approximately 4 feet 3 inches to the north of the pivot
point. The cask comes to rest with its cg more than 10 feet from the edge of the spent fuel pool. When the cask strikes the concrete slab, it tends to right and stop swinging. In a SHA (Simplified Head Assembly) removal or reassembly, it is postulated that the polar crane cable fails. If this unlikely event should occur, various consequences would prevail depending upon the position of the SHA in relation to the reactor vessel at the time of the polar crane cable failure. Nine accident cases have been defined that envelope the effects of dropping the SHA at critical points along the path from the reactor vessel to the head storage stand.The accident cases that have been defined are as follows:
- 1. (W) The Old Head Assembly falls approximately 14 through the air with the total load (318,673 lbs), with polar crane rope failure, while engaged on guide studs and impacts the reactor vessel flange (RV). This case has been explicitly analyzed by Westinghouse in WCAP-9198, revisions 0 and 1.
- 2. (W) The Old Head Assembly falls 4 through the air and 24.50 through water and impacts the RV flange with the total load (318,673 lbs) when the polar crane rope fails. The water medium provides additional damping compared with air. Westinghouse in WCAP-9198, revisions 0 and 1, determined that the impact velocity will be lower than case 1.
- 3. (W) The Old Head Assembly falls 4 through the air and 6.50 through water and strikes the guide studs, falls 18 through water and impacts the RV flange at an angle of 2.83 , with the total load (318,673 lbs) when the polar crane rope fails. This case is analyzed in Westinghouse WCAP 9198, revisions 0 and 1. 9.1-53 Rev. 23 WOLF CREEK
- 4. (W) The Old Head Assembly falls 4 through the air and 24.50 through water and lands partially on the RV flange and partially on concrete. This case considers the total load (318,673 lbs) when the polar crane rope fails.
This case is analyzed in Westinghouse WCAP-9198, revisions 0 and 1.
- 5. (W) The Old Head Assembly falls 4 through the air and 24.50 through water and strikes the cavity floor. This case considers the total load (318,673 lbs) when the polar crane rope fails. This case in analyzed in Westinghouse WCAP-9198, revisions 0 and 1.
- 6. (W) The Old Head Assembly falls 24.50 through water, rotates into the refueling pool and strikes the concrete. This case considers the total load (318,673 lbs) when the polar crane rope fails. This case in analyzed in Westinghouse WCAP-9198, revision 0 and 1.
- 7. (A) The Simplified Head Assembly falls 28.50 through air and impacts the RV flange. This case considers the total load (357,000 lbs) when the polar rope crane fails. This case envelops Cases 1 through 6 analyzed in Westinghouse WCAP-9198, revisions 0 and 1.
- 8. (A) The Simplified Head Assembly falls 32 through air and impacts the RV flange. This case considers the total load (357,000 lbs) when the polar rope crane fails.
- 9. (A) The Simplified Head Assembly falls 35 through air and impacts the RV flange. This case considers the total load (357,000 lbs) when the polar rope crane fails. This is a one time evaluation only
.Notes: (W) = Westinghouse. (A) = ARES Corporation. The first six cases, Cases 1 to 6, were evaluated generically in Reference 1. It was found that Case 2 is the limiting accident case in terms of maximum impact velocity and medium participating mass at impact. In Part I of the WCGS head drop accident analysis, the reactor vessel nozzles are analyzed for a Case 2 situation using the methods and assumptions of Reference 1. However, the necessary changes from Reference 1 in weights, stiffnesses, and drop heights (assume there is no water in the refueling pool, the vessel head is positioned 28.50 feet above the vessel flange and the polar crane cable is postulated to fail) were made in order to make this analysis more specific to WCGS while remaining conservative in terms of reactor vessel nozzle evaluation. Cases 7 to 9 were analyzed by ARES corporation in Reference 11. Cases 8 and 9 were performed to justify post plant events. Case 7 envelopes all Westinghouse cases where the SHA load drop occurs over the RV. The load considered is 357,000 lbs. This load is the limiting load that can be carried. The studs, nuts and bolts, and the tie rods are not included in this weight. The absolute maximum lift height over the reactor vessel flange is 32-0. In order to more fully address the question of maintenance of core cooling capability, a reactor vessel support evaluation for the postulated head drop accident was also performed and is reported as Part II of the WCGS head drop accident analysis. 9.1-54 Rev. 23 WOLF CREEK Part I - Reactor Vessel Nozzle Evaluation (28.50 and 32-0) Head assembly falls 28.50 or 32 feet feet through air and impacts the vessel flange.During SHA disassembly or reassembly, the vessel head is positioned 28.50 feet or 32 feet above the vessel flange. When the SHA is directly above the reactor vessel and at the maximum lift height, the polar crane cable is postulated to fail. The SHA falls, engages on the guide studs, and lands directly on the reactor vessel flange. a. Head assembly impact velocity calculation Assumptions 1. Final velocity is assumed to be equivalent to that of a 28.50 and 32-0 drop through air. One time evaluation for 35 drop. (Ref. 11) 2. The head does not bind with the guide studs. Analysis The head will impact on the reactor vessel flange after dropping. The head velocity just before the impact is given by the following equation Vgh2 where: V = impact velocity g = acceleration of gravity (32.2 ft/sec
- 2) h = height = 28.50; 32 therefore V=
2 (32.2 ft/sec
- 2) (28.5 ft) = 42.84 ft/sec or 514.08 in./sec For h = 28.50 feet, collision occurs after 1.503 seconds. At that time, the velocity is 514 inches/second. For h=32, collision occurs after 419 seconds. therefore V=
sec/545 sec/40.45)32)(sec/'2.32 (2 2 in or ft b. Consideration of fuel assemblies The fuel assemblies, and specifically the fuel cladding, must retain their integrity in order to ensure no release of fission-product gases. During this accident, the head assembly itself does not come in contact with the fuel assemblies. The drive rods, which extend above the reactor vessel flange, are carefully inserted into the head during normal refueling operations. However, during the accident, it cannot be assumed that all the drive rods 9.1-55 Rev. 23 WOLF CREEK enter the head penetrations. The drive rods will buckle under the weight of the falling head, and the buckling load of each buckling drive rod must be able to be withstood by the corresponding fuel assembly. The drive rod buckling load is the only major force experienced by the fuel assemblies and is calculated hereafter. Model of Critical Buckling Load for the Drive Rod Refer to Figure 9.1-18.
- 1. Buckling Load of Section l Before the buckling load of Section 1 can be calculated, the end condition at its base must be defined. The end condition for Section 1 is
determined by the reaction and buckling load of Section 2. Section 2 is considered to have 2 pinned ends because of the small radial clearance (0.325 inch).
- 2. Buckling Load for Section 2 (refer to Figure 9.1-19)
P cr = 2 2 L EI P cr = critical buckling load E = modulus of elasticity I = moment of inertia of an area L = length K = effective length factor ~ 0.82 D = average of major and minor drive rod thread diameters d = inside diameter of drive rod Calculate I I = 64)(4 4 d D 9.1-56 Rev. 23 WOLF CREEK D = 1.75 + 1.475 2 = 1.613 I = 4 4 4.3031.0 64)875.0 613.1 (in E = 28.3 x 10 6 lb/in.2 L = 12.50 P cr =2 6 2)50.12 ()3031.0)(10 3.28 (x P cr = 541,816 lbs The buckling load for Section 2 is 541,816 pounds. A P cr load for Section 1 of anything less than 541,816 pounds will indicate that the two act independently of each other and Section 2 will not buckle, (refer to Figure 9.1-20). P cr =2 2 L EI E = 28.3 x lO 6 lb/in.2 I = 0.3031 in. 4 L = 138 K = 0.82 (Effective Length Factor) P = 2 6)138 82.0 ()3031.0)(10 3.28 (2 x x P cr = 6611 lbs The maximum vertical force on the fuel assembly is
the buckling load of Section 1. An impact force of this value will not impart enough damage to the fuel assembly, and fuel cladding integrity is maintained. 9.1-57 Rev. 23 WOLF CREEK
- c. Consideration of reactor vessel nozzles Description The impact load of the head assembly on the vessel is transmitted through the vessel to the four supported
vessel nozzles. The nozzles must be able to support this
load without exceeding the allowable stress limits. The effects on the nozzles were evaluated by conservatively assuming the head drop through 28 feet of air. Assumptions
- 1. The head assembly is assumed to drop 28.50 feet and 32 feet through air.
- 2. If it is assumed that the stresses due to the impact load are distributed throughout any elastic body exactly as in the case of static loading, then it can be shown that the vertical deformation, i and the stresses i produced in any such body by the vertical impact of a body falling from a height (h) are greater than the deformation and stress produced by the weight of the same body applied as a static load in the ratio (Reference 4): h 2+1+1==i i If h=0, we have the case of sudden loading and i = = 2 i as assumed.
The above approximate formula is derived on the assumption that the impact load strains the elastic body in the same way (though not in the same degree) as static loading and that all the kinetic energy of the moving body is expended in producing this strain. 9.1-58 Rev. 23 WOLF CREEK Actually in the impact some kinetic energy is dissipated and this loss, which can be found by equating the momentum of the entire system before and after impact, is more conveniently taken into account by multiplying the available energy by a factor K, the value of which is as follows (Reference 4): Energy Dissipation K = 1 + 1 3 M 1 M 1 + 1 2 M 1 M 2 Factor where: M = mass of the moving body = W g M 1 = mass of the body struck by the moving body
=
W 1 g From the above equations, the impact load Wi can be derived as follows: WW1 + 1 + 2Kh1 = 3. The rigidity of the vessel flange causes the impact loads to be distributed evenly to the four
supporting nozzles.
- 4. The reactor vessel is supported by two inlet and two outlet nozzles.
- 5. Load deformation of the head at impact is neglected.
- 6. The area and moment of inertia for the inlet nozzle are larger than for the outlet nozzle at the nozzle to shell juncture region. Similar difference exists
also for the cross section at the integral pad
location. Hence, the outlet nozzle was evaluated
for impact stresses. 9.1-59 Rev. 21 WOLF CREEK Analysis Determination of the Impact Load W i The head upper package and reactor vessel can be idealized as a simple spring mass system as shown in Figure 9.1-21. Determination of Spring Constant k v The upper portion of the reactor vessel was idealized as spring k
- v. To simplify the analysis, the upper portion of the vessel was conservatively assumed to be a cylindrical member with the cross section and parameters as follows (see Figure 9.1-22):
PL=A E L AE= P k v where: R = outside radius r = inside radius
t = thickness A = area L = length = 91 (Ref. 1) E = modulus of elasticity for carbon molysteel at 70 F A = (R 2-r 2) = (97.50 2 - 85.0 2) = 7166.80 in. 2 9 6 10 2.36=91)10 9.29)(80.7166 (x x k vlb/in 9.1-60 Rev. 23 WOLF CREEK Spring constants for the inlet (k in) and outlet (k on) regions were determined from a 3-D finite element analysis of the reactor vessel. Determination of equivalent spring constant k e of the system shown in Figure 9.1-21: k in = 79.8 x 10 6 lbs/in. k on = 71.7 x 10 6 lbs/in. k s = 58.0 x 10 6 lbs/in. (Ref. 1) k v = 2.36 x 10 9 lbs/in. For the nozzles and supports in series: k inlet-support
=ins s in k k k/1/1 1 k inlet-support = 3.36 x 10 7 lbs/in. = k ins k outlet-support =s on k 1+k 1 1 k outlet-support = 3.21 x 10 7 lbs/in. = k ons For springs in parallel:
k p = 2 k ons + 2 k ins k p = 13.14 x 10 7 lbs/in. 9.1-61 Rev. 23 WOLF CREEK For the springs in series: 7 10 45.12 1 1 1 x k k k p v e The weight of the upper package, head assembly, and crane block (W) is 357,000 pounds (Table 9.1-10). The weight of the vessel flange and nozzle shell (W
- 1) is 290,000 pounds.
0.643=357,000 290,000 2 1+1 357,000 290,000 3 1+1=W W 2 1+1 W W 3 1+1 2 2 1 1K Energy dissipation factor (K) is 0.643. e k W Static deflection () is 0.0029 in. Impact Load
- For 28.50 and 32 drops A]28.50 drop (W1):
Equation 2Kh+1+1 W Wi becomes: 138 , 140 0029.0)12)(50.28)(643.0)(2 (1 1 000 , 357Wi Impact load = 140.138 x 10 6 lbs Impact Displacement = inches 1.126 12.45x10 140.138x10 =(Ke)Stiffness Load Impact 7 6 Assuming a perfect drop and four supported nozzles equally share the impact load: Impact force/nozzle = lbs.35.034x10=4 140.138x10 6 6 Determine the stress developed in the outlet nozzle due to impact load (refer to Figure 9.1-23). 9.1-62 Rev. 23 WOLF CREEK Impact Load (W
- 1) for 32 feet drop Equation W 1= W 2 1 1 Kh becomes W 1 = 357,000 0029.0 12 32 643.0 2 1 1x x x W 1 = 148,472 x 10 6 lbs Impact load = 148,472 x 10 6 lbs Impact Displacement = Impact Load / Stiffness = (148.472 x 10 6 lbs)/(12.45 x 10 7 lbs/in) = 1.193 inch.
Consistent with the assumption made in References 1 and 2, four supported nozzles equally share the impact load. Impact force/nozzle = 148.472 x 10 6 lbs/4 = 37.118 x 10 6 lbs. 9.1-63 Rev. 23 WOLF CREEK Nozzle Evaluation for 28.50 Drop: R = 23.03 in. r = 14.705 in. Moment of Inertia (I) = 4 5 4 4 10 82.1 4 in x r R Maximum Bending Stress in the outlet nozzle 075 , 63I MR B psi where: M=(Impact Force/Nozzle)*(a)=(35.034 x 10 6)*(14.40)=504.49 x 10 6 in-lb Shear Stress Maximum load at the nozzle cross section = impact force/nozzle = 35.034 x 10 6 lbs. 499 , 35 705.14 03.23 10 034.35 R)le Force/Nozz Impact (6 2 2 x r avg psi Maximum Principal Stress 019 , 79 4 1 2 1 2 max 2 2avg B B psi Therefore, max is less than the allowable limit of 84,000 psi (lesser of 3.6S m or 1.05S u).Nozzle Evaluation for 32 feet Drop HeightBending Moment (M) = Impact force/nozzle * (a) = (37.118x10 6 lbs)x(14.4 in.)=534,500 x 10 6 in-lb Bending Stress () = I MR = 66,823 psi Shear Stress in Outlet Nozzle Maximum load at the nozzle cross section = Impact force/nozzle = 37.118x10 6 lbs avg =psi 37,610 14.705-23.03 37.118x10=r-R le Force/Nozz Impact 2 2 6 2 2 9.1-64 Rev. 23 WOLF CREEK Maximum Principal Stress in Outlet Nozzle 719 , 83 4 1 2 1 2 max 2avg B B psi Therefore, max is less than the allowable limit of 84,000 psi (lesser of 3.6 S m or 1.05 S u).The results of the preceding analysis show that the reactor vessel nozzles are not stressed above allowable limits. In order to more completely address the question of maintenance of core cooling capability, an evaluation of the reactor vessel supports was performed, and the results are reported in Part II.
- d. Consideration of core barrel In a normal reassembly of the reactor vessel, the SHA first contacts the upper internals flange applying pressure against the core barrel holddown spring. The upper internals depress the holddown spring until the SHA contacts the vessel flange.
During this accident case the above reassembly description occurs compressing the hold-down spring. Any amplified effects could cause some yielding on the outer portion of the core barrel and upper internals flanges. The bottom of the core barrel is designed with supports for a hypothetical accident in which the core barrel support, the flange, might fail and allow the core barrel to fall. These supports will limit its travel to approximately 1-1/4 inches in a cold condition without any failure to the fuel. Therefore, in an unlikely event of this accident case causing failure of the core barrel the lower internals supports would limit the core barrel travel as in the hypothetical accident to approximately 1-1/4 inches and still maintain the integrity of the core. Part II - Reactor Vessel Support Evaluation Due to the high impact loads on the reactor vessel nozzles reported in Part I, a reactor vessel support evaluation was performed for the effects caused by a postulated vessel head drop accident. In an effort to reduce the probability of vessel support damage due to high impact loads, the accident evaluation input parameters were revised to more closely reflect actual WCGS plant conditions during refueling. These changes are described as follows.
- a. The mass assumed to resist the impact of the falling SHA is given in Part I as 290 kips. This is the weight of the vessel nozzles, vessel flange, and the upper portion of the vessel barrel. Since, in fact, the reactor vessel shell, internals, fuel, and water in the vessel are not isolated from one another in terms of the response to a vertical load on the vessel flange, the entire weight of the vessel shell, vessel internals, 9.1-65 Rev. 23 WOLF CREEK fuel, and water was assumed for Part II to resist the impact of the falling head assembly. This weight was found to be 1650 kips.
The WCGS reactor vessel supports were evaluated to determine the maximum vertical displacement of the reactor vessel due to the postulated head drop accident. This was done using energy balance techniques, taking into consideration energy losses at impact as determined by the change in velocity of the total system based on conservation of momentum principles. The vessel supports are made up of a cooling box structure designed by Westinghouse that, in turn, is supported by a Bechtel design of structural
steel framing partially embedded in the primary shield wall. The vertical stiffness of the Westinghouse-designed cooling box was found to be 4.07 x 10 5 kips/inch. This value is the resultant of several stiffnesses acting in series that were calculated separately along the height of the support. The separate
calculations were necessary due to geometric changes in the cooling box which
affected the bearing area of the vertical plates of which the box is composed. The minimum moment of inertia along the height of the support was found to be 9,355 in.4. This value was assumed constant throughout the unbraced length of 21.5 inches in order to find the minimum Euler buckling load of the cooling
box. The buckling load ranges between 5.59 x 10 6 kips for pinned-pinned end conditions and 22.37 x 10 6 kips for fixed-fixed end conditions. The minimum yield load of the cooling box was found by multiplying the cross-sectional area at the critical section (the same area used to determine the minimum moment of inertia) by the nominal yield strength of the steel. These values are 242.25 square inches and 50 ksi, respectively, and give a resulting yield load for each cooling box of 12.11 x 10 3 kips. A comparison of the yield load and minimum buckling load shows that the cooling box will yield before it buckles.The yield displacement of the cooling box was found by dividing the yield load of 12.11 x 10 3 kips by the vertical stiffness, 4.07 x 10 5 kips/inch. The resulting displacement is 0.0298 inch. This defines the limit of elastic vertical deformation. Any further deformation was assumed to be perfectly plastic. 9.1-66 Rev. 23 WOLF CREEK The Bechtel-designed portion of the reactor vessel supports has the following properties, as provided by Bechtel: Stiffness: 34,880 kips/inch Elastic Capacity: 7,560 kips per support Ductility Ratio: 10 From the above data, the yield displacement was found to be 0.22 inch with an ultimate deformation of 2.2 inches. As in the cooling box portion of the support, any deformation beyond the yield displacement was assumed to be
perfectly plastic. The stiffness of the reactor vessel upper barrel used in the support evaluation was revised to 1.550 x 10 6 kips/inch. This was the result of changes in the following parameters affecting the stiffness calculation. These values were taken from the vessel drawing. E = 28,000 ksi R = 96.35 inches r = 85.60 inches A = 6,145 square inches
L = 111 inches The parameter labels are defined in the reactor vessel nozzle evaluation of Part I. The total system stiffness was found by adding the following stiffnesses in series: a. Reactor vessel inlet and outlet nozzle stiffnesses (two each, from Part I)
- b. Cooling box and embedded structural steel stiffnesses (four each)
- c. Reactor vessel upper barrel stiffness 9.1-67 Rev. 23 WOLF CREEK The resulting total system stiffness is 85,270 kips/inch.
A bilinear load-displacement curve describing the system response was next developed. A comparison of the minimum yield loads of the cooling box and embedded structural steel showed that the embedded structural steel will yield first at 7,560 kips per support or a total of 30,240 kips for all four supports. Dividing this yield load by the system stiffness gives a yield displacement of the total system of 0.36 inch. The following load-displacement curve can then be developed: 0.36 Total 30,240 load (kips) Displacement (inches) The area under the load-displacement curve is the internal resistance of the system to the impact load and system deformation caused by the postulated head
drop accident. The area under the load-displacement curve, with X being the total displacement, is found as follows: I = 30,240 (X) - 0.5 (30,240) 0.36 Where I = internal resistance 30,240 = minimum yield load of support system (kips) 0.36 = yield displacement of support system (inches) The external work on the system is done by the falling head assembly and is the sum of the kinetic energy remaining after impact and the work required to deform the system vertically a distance X. The total kinetic energy at impact is modified to account for that which is lost by dissipation. The modification
is by a factor determined by conservation of momentum and is given as follows: 9.1-68 Rev. 23 WOLF CREEK For 28.5 feet Drop Height The external work is given as follows: E=357 x (342) x 0.178 + 357 x (X) Where, 342 inches = height of drop Assuming the internal resistance equals the external work less the dissipated energy, the total system deformation was found by equating the expressions for I and E. I = E 30,240 (X) - 0.5 x (30,240) x 0.36 = 357 x (342) x 0.178 + 357 x (X) Solving for X, X = 0.91 inch For 32 feet Drop Height The external work is given as follows: E = 357 x (384) x 0.178 + 384 x (X) Where, 384 inches = height of drop (inches) Assuming the internal resistance equals the external work less the dissipated energy, the total system deformation was found by equating the expressions for I and E. I = E 30,240 (X) - 0.5 x (30,240) x 0.36 = 357 x (384) x 0.178 + 357 x (X) Solving for X, X = 1.00 inch The distribution of the elastic deformation of the various components involved in supporting the reactor vessel is shown in Table 9.1-11. These values were determined by multiplying the total system elastic deformation of 0.36 inch by the ratio of the stiffness of the specific component to the total system stiffness. The plastic deformation of 0.55 (=0.91-0.36) inch for 28.5 feet drop and 0.64(=1.00-0.36) inch for 32 feet drop occurred entirely in the embedded structural steel portion of the vessel supports, since the minimum yield load assumed for the total system was that of the embedded steel. The total deformation of the embedded steel was found to be 0.77 (=0.22+0.55) inch for 28.5 feet drop height and 0.86 (=0.22+0.64) inch for 32 feet drop height. These deformations are less than the limit of 2.20 inches given previously. 9.1-69 Rev. 23 WOLF CREEK After the maximum vessel displacement was determined, the routing of the essential auxilairy lines attached to each loop was examined to locate any possible interferences with civil structures, equipment, or pipe whip restraints. No interferences were found for the total vessel displacement indicated in Table 3-2. Table 3-3 shows the design margins based on maximum nozzle stresses for 28.5 and 32 feet drop heights. The results of the support evaluation for the postulated head drop accident show that the reactor vessel supports and, hence, the reactor vessel and nozzles will displace vertically a maximum of 0.91 inch for 28.5 feet drop height and 1.00 inch for 32 feet drop height. Permanent deformation of the support system is 0.55 inch for 28.5 feet drop height and 0.64 inch for 32 feet drop height. The maximum displacement of the reactor vessel will not have an effect on the ability of the reactor coolant loop piping and essential auxilairy piping to circulate borated water to the core and remove residual heat. Therefore, core cooling capability and fuel cladding integrity are maintained. The cranes and fuel transfer system are provided with limit switches to guard against improper travel and operations and to ensure correct and safe handling of the fuel assemblies. More specific details and descriptions of the limit switch applications are given in Section 9.1.4.2.2 for each major component where they may apply. 9.1-70 Rev. 23 WOLF CREEK 9.
1.5 REFERENCES
- 1. Alexander, D. W., Shakely, R., and Dudek, D. F., "Reactor Vessel Head Drop Analysis," WCAP-9198, Westinghouse Letter SAP-07-45 dated June 27, 2007.
- 2. Hunsaker, J. C. and Rightnire, B. G., Engineering Applications of Fluid Mechanics , page 183.
- 3. Hoemer, J. F., Fluid-Dynamic Drag , page 317.
- 4. Roark, R. J., Formulas For Stress and Strain - Fourth Edition , pages 340, 370, and 371. 5. Holtec Documentation No. 981968, Rev. 2, Criticality Analysis of the Wolf Creek New Fuel Vault with 5% Enriched Fuel (calculation) DCN C-175A-00088-W036. Letter from J. C. Stone, USNRC, to O. L. Maynard, WCNOC dated June 24, 1997, Request For Exemption From 10 CFR 70.24 Criticality Monitoring Requirements - Wolf Creek Generating Station (TAC NO. M89161).7. Holtec Documentation No. HI-971768, Rev. 5, Thermal-Hydraulic Evaluation of the Reracked Callaway and Wolf Creek Spent Fuel Pool DCN c-175A-00052-W06.8. Calculation No. AN-99-032, Rev. 0, The Heat Rejection Rate from the Spent Fuel Pool to the Ultimate Heat Sink following a LOCA 9. Letter ET 98-0103, dated December 8, 1998, from R. A. Muench, WCNOC, to USNRC, Docket No. 50-482: Follow-up To December 3, 1998, NRC/Wolf Creek Nuclear Operating Corporation/AmerenUE Meeting on Spent Fuel Pool Rerack10. Holtec Documentation No. HI-971754, Rev. 3, Analysis of the Mechanical Accidents for Callaway and Wolf Creek Nuclear Plants (C-175A-00053) 11. Calculation 0720517.01-C-001: WCNOC Reactor Vessel Head Drop Analysis 12. Westinghouse calculation CN-CSE-02-40: Simplified Head Assembly (SHA) loads. 9.1-71 Rev. 23 WOLF CREEK TABLE 9.1-1 NEW FUEL STORAGE DESIGN DATA Component Requirements Design Data New fuel storage vault capacity 66 new fuel assemblies (approximately 34% of a core)
New fuel storage vault size 21 feet - 0 inches wide, 24 feet - 10 inches long, 15 feet - 0 inches deep (clear dimensions) Module array Center-to-center cell lattice array is 21 x 21 inches with 33 dual cell
modules which are arranged in three rows of 11 modules each. Rev. 0 WOLF CREEK TABLE 9.1-2 SPENT FUEL STORAGE DESIGN DATA Component Requirements Design Data Type High density Spent fuel pool storage capacity With the Wolf Creek fuel storage pool rerack modification (SFP) in the year 2000, Wolf Creek total fuel storage space was increased from 1340 to 2363 fuel assemblies in the Spent Fuel Pool (SFP). The modification made in 2000 replaced the 12 original fuel storage racks with 15 high density storage racks. The license amendment in support of the rerack modification also included an additional capability to add 3 more high density storage racks, which would make a total of 18 high density storage racks. If the 3 additional high density storage racks are installed in the cask loading pool in the future, this would allow the capability to hold an additional 279 fuel assemblies, which would allow the total storage space of 2642 fuel assemblies in the SFP. Fuel Storage pool size 28 feet - 6 inches wide (does not include Cask loading pool) 50 feet - 0 inches long 41 feet - 0 inches deep Module array Refer to Figure 9.1-2. Fuel Storage pool design temperature, F 170 Rev. 20 WOLF CREEK TABLE 9.1-3 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.13 REVISION 1, DATED DECEMBER 1975, TITLED "SPENT FUEL STORAGE FACILITY DESIGN BASIS" Regulatory Guide 1.13 Position WCGS
- 1. The spent fuel storage facility (in- 1. Complies as described cluding its structures and equipment, except in Section 9.1.2.1.1.
as noted in Paragraph 6 below) should be de-signed to Category I seismic requirements.
- 2. The facility should be designed (a) 2. Complies as described to keep tornadic winds and missiles generated in Section 3.5, and
by these winds from the fuel storage pool and 3.8.(b) to keep missiles generated by tornadic winds from contacting fuel within the pool.
- 3. Interlocks should be provided to prevent 3. Complies as described cranes from passing over stored fuel (or near in Section 9.1.4.
stored fuel in a manner such that if a crane failed the load could tip over on stored fuel) when fuel handling is not in progress. During
fuel handling operations, the interlocks may be bypassed and administrative control used to prevent the crane from carrying loads that are
not necessary for fuel handling over the stored fuel or other prohibited areas. The facility should be designed to minimize the need for by-passing such interlocks.
- 4. A controlled leakage building should en- 4. Complies as described close the fuel pool. The building should be in Section 9.4.2 and equipped with an appropriate ventilation and 15.7.4.
filtration system to limit the potential release
of radioactive iodine and other radioactive ma-terials. The building need not be designed to withstand extremely high winds, but leakage
should be suitably controlled during refueling operations. The design of the ventilation and filtration system should be based on the assump-
tion that the cladding of all of the fuel rods in one fuel bundle might be breached. The in-ventory of radioactive materials available for leakage from the building should be based on the assumptions given in Regulatory Guide 1.25,"Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors" (Safety Guide 25). Rev. 0 WOLF CREEK TABLE 9.1-3 (Sheet 2) Regulatory Guide 1.13 Position WCGS
- 5. The spent fuel storage facility should 5. Complies as described have at least one of the following provisions in Section 9.1.4.
with respect to the handling of heavy loads, including the refueling cask:
- a. Cranes capable of carrying heavy loads should be prevented, preferably by design rather than by interlocks, from moving into the vicin-ity of the pool; or
- b. Cranes should be designed to provide single-failure-proof handling of heavy loads, so that a single failure will not result in loss of capability of the crane-handling system to perform its safety function; or
- c. The fuel pool should be designed to withstand, without leakage that could uncover
the fuel, the impact of the heaviest load to be carried by the crane from the maximum height to which it can be lifted. If this
approach is used, design provisions should be made to prevent the crane, when carrying heavy loads, from moving in the vicinity of stored fuel.
- 6. Drains, permanently connected mechanical 6. Complies as described or hydraulic systems, and other features that in Section 9.1.2 and by maloperation or failure could cause loss 9.1.3.
of coolant that would uncover fuel should not
be installed or included in the design. Systems for maintaining water quality and quantity should be designed so that any
maloperation or failure of such systems (including failures resulting from the safe shutdown earthquake) will not cause fuel to
be uncovered. These systems need not other-wise meet Category I seismic requirements.
- 7. Reliable and frequently tested monitoring 7. Complies as described equipment should be provided to alarm both in Section 9.1.3.
locally and in a continuously manned location if the water level in the fuel storage pool fails below a predetermined level or if high local-radiation levels are experienced. The
high-radiation-level instrumentation should also actuate the filtration system. Rev. 0 WOLF CREEK TABLE 9.1-3 (Sheet 3) Regulatory Guide 1.13 Position WCGS
- 8. A seismic Category I makeup system should 8. Complies as described be provided to add coolant to the pool. Ap- in Section 9.1.3.
propriate redundancy or a backup system for filling the pool from a reliable source, such as a lake, river, or onsite seismic Category I water-storage facility, should be provided.
If a backup system is used, it need not be a permanently installed system. The capacity of the makeup systems should be such that water can be supplied at a rate determined by con-sideration of the leakage rate that would be expected as the result of damage to the fuel
storage pool from the dropping of loads, from earthquakes, or from missiles originating in high winds. Rev. 0 WOLFCREEKTABLE9.1-4FUELPOOLCOOLINGANDCLEANUPSYSTEMDESIGNPARAMETERS1.Fuelpoolstoragecapacity,assembliesSpentfuelpool2363Ultimate26422.Spentfuelpoolwatervolume(1),gal(waterlevel1'-6"fromtopofpool)Nominal400,0003.Boronconcentrationofthespentfuelpoolwater,ppm(minimum)2,1654.Evaporationrates,gpm @140°F1.80@170°F5.575a.HeatLoadsandBulkPoolTemperatureforaPartialCoreOffload(2),(5),(8),(9)HeatRateSFPBulkTemperature(MBtu/hr)(°F)________ 27.151405b.HeatLoadsandBulkPoolTemperatureforFullCoreOffload(3),(6),(8),(9)HeatRateSFPBulkTemperature(MBtu/hr)(°F)________ 63.411705c.HeatLoadsandBulkPoolTemperature15DaysAfterShutdownforRefueling(4),(7),(8),(9)HeatRateSFPBulkTemperature(MBtu/hr)(°F)________16.94140Rev.14 WOLFCREEKTABLE9.1-4(Sheet2) NOTES:(1)Forcomputationofthermalparameters,thespentfuelpoolisconsideredisolatedfromthefueltransfercanalandthecaskloading pool.(2)Theinitialpoolheatuprate[uponadditionofspentfuelfromrefuel26(whenthefuelstoragepoolisconsideredfull)]is8.0 °F/hr,assumingthatFSPcoolingisnotoperational.(3)Poolheatuprateforcoreoffloadis18.7 °F/hr,assumingthatFSPcoolingisnotoperational.(4)Poolheatuprate15daysaftershutdownis4.91 °F/hr,assumingthatFSPcoolingisnotoperational.(5)Themaximumheatloadisbasedondecayheatgenerationfromtheprojectednumberoffuelassembliesthataredischargedfromapartialcoreoffload,plus26previousrefuelingbatches.(6)Themaximumanticipatedheatloadisbasedonthedecayheatgeneratedbyafullcore(193assemblies)removedfromthereactorandstoredinthespentfuelpool,whilethespentfuelassembliesfrom26previousrefuelingsremaininthespentfuelpool.(7)Themaximumheatloadisbasedondecayheatgenerationfromthefuelassembliesthatareoffloadedfromthecoreandstoredinthepoolafterreactorshutdown,plus26previousrefuelingbatches.Thecoolingtimeisbasedontheassumptionthatthereactorcouldbebroughtbackonline15days(360)hoursfollowingshutdownforrefueling.(8)Useofonecoolingwatertrainwith105°FCCWisassumed.(9)SeeUSARSection9.1A.3forexplanationofhowtheMaximumHeatRateis determined.Rev.14 WOLF CREEK TABLE 9.1-5 FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARAMETERS Fuel Pool Cooling PumpQuantity2TypeHorizontal centrifugalDesign pressure, psig150 Design temperature, °F225Design flow, gpm3,250Design head, ft124 MaterialAustenitic stainless steelDesign codeASME Section III, Class 3Motor data150 hp/460 V/3 phase/60 HzSeismic categoryI Fuel Pool Skimmer PumpQuantity1TypeInline centrifugalDesign pressure, psig300 Design temperature, °F160Design flow, gpm100Design head, ft156.7 MaterialAustenitic stainless steelDesign codeMS Rev. 0 WOLF CREEK TABLE 9.1-5 (Sheet 2) Fuel Pool Cleanup PumpQuantity2TypeInline centrifugal Design pressure, psig300 Design temperature, °F160Design flow, gpm150 Design head, ft161.4 MaterialAustenitic stainless steelDesign codeMS Fuel Pool Cooling Heat ExchangerQuantity2 Design heat transfer, Btu/hr 15.09 x 10 6*Heat transfer area: gross, ft25270Design codesASME Section III Class 3 and TEMA "R"Seismic categoryI Shell TubeDesign pressure, psig150150 Design temperature, °F200250 Design flow, gpm3,0003,250 Inlet temperature, °F105125 Outlet temperature, °F115.1115.7 Fluid circulatedComponentFuel poolcooling watercooling waterMaterialCarbon steelAustenitic stainless steel*This nominal heat load was used to determine the heat exchanger surface requirement. The spent fuel pool heat exchanger has been verified to be capable of maintaining the pool temperature within
acceptable limit for full core offload conditions. Rev. 9 WOLF CREEK TABLE 9.1-5 (Sheet 3) Fuel Pool Cleanup Demineralizer Quantity 1 Design pressure, psig 150 Design temperature, °F 250 Design flow, gpm 300 Resin volume, ft3145 Design pressure drop (fouled), psi 12-15 Material Austenitic stainless steel Design code ASME Section VIII, Div. 1 Fuel Pool Cleanup Filters Quantity 2 Design pressure, psig 150 Design temperature, °F 250 Design flow, gpm 150 Design pressure drop (clean/fouled), psi 2*/30 Filtration requirement Absolute Filtration program target size is .1 micron at > 99.98% efficiency Material, vessel Austenitic stainless steel Design code ASME Section VIII, Div. 1 Fuel Pool Skimmer Filter Quantity 1 Design pressure, psig 150 Rev. 16 WOLF CREEK TABLE 9.1-5 (Sheet 4) Design temperature, °F 250 Design flow, gpm 100 Design pressure drop (clean/fouled), psi 1*/30 Filtration requirement Absolute Filtration program target size is .1 micron at > 99.98% efficiency
- The clean filter differential pressure indicated here reflects the original filter housing design pressure drop. The filter cartridges
used as a part of the absolute filtration program may have an initial clean pressure drop much higher than this due to the smaller pore size. Rev. 16 WOLFCREEKTABLE9.1-6SPENTFUELPOOLCOOLINGANDCLEANUPSYSTEMSINGLEACTIVEFAILUREComponentFailure Comments1.FPCpumpFailtostartTwopumpsarewhenmanuallyprovided.Onestarted.pumpissuffi-cientforre-sidualheatre-
moval.2.FuelStorageFailstostopTwolevelswitchesPoollevelpumpuponlowareprovided-oneswitchlevelinpool.dedicatedtoeachFPCpump.Oneswitchissuffi-cientforpro-tectionofone pump3.Motor-operatedFailstocloseTwoseparatecom-isolationvalveuponautomaticponentcooling forCCWoutletactuationatstartwaterloopsare toFPCheatex-ofpost-LOCAprovided.One changerrecirculation.CCWloopprovides100percentof post-LOCAcooling
capacity.FailstoopenTwoseparateuponmanualcoolingloopsare actuationwhenprovided.One fuelpoolcoolingloopprovides heatloadcanbe100percentof acceptedafteraresidualheatre-LOCA.moval.Rev.14 WOLF CREEK TABLE 9.1-7 FUEL HANDLING CRANE DATA (1) Name of Crane Parameters Polar Crane Cask Handling Crane Spent Fuel Pool Bridge Crane Refueling Machine Capacity of main hoist 260 tons (5) 150 tons2 tons2.4 tons Capacity of auxiliary
monorail hoist (const) 25 tons5 tons 5 ton manual Capacity of auxiliary
monorail hoist (normal) 25 tons 5 tons & 2 tons (2) 1.5 tons Capacity of main trolley 260 tons (5) 130 tons 5 tons 2.4 tons Capacity of lift bea m 500 tons Maximum main hoist speed (normal) 6.4 fp m3.75 fpm 21 fpm 40 fpm Minimum main hoist speed (normal) 3 ip m2.5 ipm 7 fpm Maximum auxiliary monorail hoist speed (normal) 51 fp m Minimum auxiliary monorail hoist speed (normal) 3 fp m Maximum trolley speed (normal) 51.5 fp m20 fpm 30 fpm 40 fpm Minimum trolley speed (normal) 6 fp m18 ipm 10 fpm Maximum bridge speed (normal) 51.5 fp m20 fpm 30 fpm 60 fpm Minimum bridge speed (normal) 6 fp m6 ipm 10 fpm Maximum load during
plant Operation 167.5 tons125 tons 1,966 lbs (Dry) 3,166 lbs Normal expected load
Range 0-167.5 tons 0 - 125 tons 0 - 1,671
lbs (wet) Maximum construction Load 475 tons Maximum main hoist speed (constr) 5 fp m Minimum main hoist
speed (constr) 3 ip m Maximum trolley speed (constr) 51.5 fp m Minimum trolley speed (constr) 6 fpm Maximum bridge speed (constr) 51.5 fpm Minimum bridge speed (constr) 6 fp m Rev. 27 WOLF CREEK TABLE 9.1-7 (Sheet 2) Name of Crane Parameters Polar Crane Cask Handling Crane Spent Fuel Pool Bridge Crane Refueling Machine Normal load range (constr)0 - 475 tons Maximum monorail hoist
Speed 20 fpm 20 fpm Minimum monorail hoist Speed 10 fpm 0 fpm Maximum monorail trolley
Speed 32 fpm Minimum monorail trolley Speed 10.7 fpm Lifting limitation 28.5/32 ft (above vessel flange)See Note 7 Cask bottom 3 inches above fl El. 2047-625-4(Hook limit is 2066 11-15/16)See Note 6 Seismic class (3) (3)(3) (4)Design standards General CMAA No. 70 CMAA No. 70 CMAA No. 70 CMAA No. 70 (1975) (1975) (1975)(1975) Structural Covered by CMAA Covered by CMAA Covered by CMAA ASME Sect. III, App. XVII, Subarticle XVII-2200 Electrical NFPA Vol. 5 Art. 610 1974-1975 NFPA Vol. 5 Art. 610 1974-1975 NFPA Vol. 5 Art. 610 1974-1975 NFPA Vol. 5 Art. 610 1974-1975 Materials ASTM Std'sASTM Std'sASTM Std's ASTM Std's Others OSHA 29 CFR 1910 & 1926 OSHA 29 CFR 1910 & 1926 OSHA 29 CFR 1910 & 1926 OSHA 29 CFR 1910 & 1926 NOTES:(1)Rated speeds given are within 10 percent of the actual speeds. (2)Refer to Figure 9.1-7: a 2-ton limit to the monorail hoist exists only over area B on Figure 9.1-7. (3)Seismic Category I (4)Component is non-Seismic Category I. Component is seismically designed and constructed if Position C.2 of Regulatory Guide 1.29 applies per Table 3.2-3. (5)Capacity of main hoist & trolley = 260 tons concurrent with a SSE event (6)When lifting a loaded RCC Handling Tool, the Spent Fuel Pool Bridge Crane hook height shall not exceed 2057' 10". When lifting a loaded BPRA Handling Tool the
maximum hook height shall not exceed 2060' 2". (7)28.50' SHA drop through the air, 32' maximum drop through the air. Rev. 23 WOLF CREEK [TABLE 9.1-8 HAS BEEN DELETED] Rev. 0 WOLF CREEK TABLE 9.1-9 Original Head Weight without SHA Modification Components for Drop Analysis Items Weight (lbs) Head 165,150 CRDM (full length) 74,100 Rod position indicator coil stack 14,895 Seismic platform 11,100 Stud tensioner hoist 900 Dummy cans 848 Sling block platform 570 Head insulation 500 Lifting rig and vent shroud (15,125 + 5,250 lbs) 20,375 Load block 16,000 Studs, Nuts and Washers 37,150 Head shield and support structure 20,000 Contingency 10,000 Total (lbs) 371,588 Use (lbs) 375,000 Items Weight (lbs) Plenum (Including fans) 20,390 Cable support structure 2,937 Shroud extension 2,861 Radiation shield 20,373 CRDM cable bridges 730 DRPI cable bridge 1,530 Cables on CSS 1,460 Cables on CRDM bridge 557 Cables on DRPI bridge 891 Transition Assembly 554 Total 52,283
Rev. 25 WOLF CREEK TABLE 9.1-10 Final Head Weight with SHA Modification Components for Drop Analysis Item No.Item Description Weight (lbs)Reference 1 Head 161,150 Reference 12 2 CRDM (full length) 74,100 Reference 12 3 Rod position indicator coil stack 14,895 Reference 12 4 Seismic platform 12,550 Reference 12 5 Stud tensioner hoist 900 Reference 12 6 Dummy Cans 848 Reference 12 7 Sling block platform 597 Reference 12 8 Head insulation 1,700 Reference 12 9 Shroud support ring 3,619 Reference 12 10 Upper cooling shroud 1,508 Reference 12 11 Lifting rig 15,936 Reference 12 12 Tie rods 2,259 Reference 12 13 Ladder 149 Reference 12 14 Head vent pipe 657 Reference 12 15 Studs, nuts, and washers (not to be considered, see Attachment B) 37,150 Reference 12 16 Added SHA Weight 52,283 From Table 9.1-9 17 Total Head Assembly Weight without Items 12 and 15 340,892-18 Bottom Block Assembly 11,750 Dwg. M-063-00087 19 Weight of rope for a conservative length of 60 feet = 2 lbs/ft x 16 ropes x 60 ft 2,880 20 Weight of Lead Blankets 300 21 Total Head Assembly Weight (Add Items 17 to 20) 355,822 22 Critical Life Load -P ir for Drop Analysis 357,000 Reference 11 Rev. 23 WOLF CREEK Table 9.1-11 Vertical Deformation of Reactor Vessel and Supports due to a Postulated SHA Drop Accident Supporting Component Elastic Deformation (inch) Plastic Deformation for 28.5 feet Drop Height (inch) Plastic Deformation for 32 feed Drop* Height (inch Vessel Plus Nozzles 0.12 None None Westinghouse Cooling Box 0.02 None None Bechtel Embedded Framing 0.22 0.55 0.64 Subtotals 0.36 0.55 0.64 Total Displacement (Elastic + Plastic) 0.91 inch for 28.5 feet Drop Height 1.00 inch for 32 feet Drop Height Allowable Displacement 2.20 inch Reactor Vessel Nozzle Stresses due to Postulated SHA Drop Accident.Supporting Component Allowable Nozzle Stress (psi) Principal Nozzle Stress for 28.5 feet Drop Height (psi)Principal Nozzle Stress for 32 feet Drop Height (psi)*Vessel Nozzles 84,000 79,019 83,719 Margin Factor (Nozzle Stress/Allowable Stress)-0.94 1.00 Rev. 23 WOLF CREEK 12999 Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-1 NEW FUEL STORAGE RACK ... ) 0 ... 0 \ CASK LOADING POOL -------,,------, ,, ,, ,, ,, FUTURE RACKS & SUPPORT PLATFORM ,, ,, ,, ,, ,, ----,' n ----" II " " " .. ... 0 " " " .. " " u ... 0 II " .. ... 0 ... a FUEL STORAGE POOL ... 0 loL "' "' u "' u .loj ... ... 0 0 CASK UJAOING P 0 0 L FLDDR ELEV. 200 I' -o* OPERA liNG DECK ... ELEV. 2047' -fi
- 0 ... 0 ** 0 w_ -:10 FUEL STORAGE P OOL ... FU:JlJ? ELEV. 2006' -I 0
- WOLF CREEK UPDATED SAIFETY ANALYSIS REPORT ELEVATION VIEW OF RACK LAYOUT FIGURE 9.1-2 CSHEET 1 OF 3) REV. 14 PICI'OR!AL Vlit OF TYPICAL RACI STRUCTURE PN"EL (ACTUAL RACK SIZE VARIES TOTAL OF STORAGE CELLSJ FINIJ CELL TYPICAL ARRAY OF STORAGE CELLS INON-FLUX TRAP CONSTRUCTION I FUEL ASSEMBLY , '
'
- 1: ===1m PN"EL-FL!JW HOLE I TYPI BASEPLATE
1 I--
L_,_J ' ' PWR CELL IN ELEVATION VIEJ WOLF CREEK UPDATED SAFETY ANALYSIS REPORT RACK STRUCTURE & MISC. DETAILS FIGURE 9.1-2 CSHEET 2 OF 31 REV. 14 FUTURE RACKS o_ f--z = -1.5" (TYP.l I ':I 1111111111111111111111111111111111111111111111111111111111111111111111 +1111111111111111111111111111111111111111111111111111111111111111111111 WOLIF CREEK UPDATED SAIFETY ANALYSIS REPORT POOL LAYOUT FOR WOLF CREEK FIGURE 9.1-2 (SHEET 3 OF 3l REV. 14 r ----WOLF CREEK NEAR SIDE FAR SIDE RIGHT END LEFT END: TOP OF RAIL ... u. ;:j ___ _,__RUNWAY RAIL CAB 181 YES ONO DYES 181NO J81YES 0NO DYES liZ! NO CAPACITY-MAIN 260 TONS CAPACITY-TONS LIFT-MAIN l.Ql_FT__!_l_!N ll FT-AUX. __g,L FT _2_! N A 134 FT_O_IN 1 .. I P WHEELBASE SECTION "A-A" MAX. LOAD ON EACH TROLLEY WHEEL ____ LBS CAB T OPERATING FLOOR EL. 2047'-6" H__Q_FT_O_IN U---LFT __Q_IN J _l8_FT_6 _IN K N/AFTN/AtN W--.ll.!.FT _Q__I N L __ 9_FT_9 _IN M_6_FT_6_tN Y -.-LFT_6 __ IN B 72 FT_!Q_IN(HIGH HOOK) N_3_FT_O_IN LENGTH OF MAIN LINE N/A FTN/AIN c N/A FT N/A1N o_li_FT_Q__IN RUNWAY D 3 FT_O_IN MAX. LOAD ON EACH E 0 FT_O_IN Q_lQ_FT_O_JN WHEEL LBS F 0 FT _ _Q_tN R__l_FT_L_IN RUNWAY RAIL G 0 FT_O_tr\1 S __ FT __ IN ],.75 LBS T _1!1:_ FT __LIN NOTE: "NEAR SIDE" & "LEFT/RIGHT"* FACING CRANE DRIVE SIDE. Top of Rail Elevation 2122'-6" "N" is 1'-8" to Obstruction Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9. 1 -4 ARRANGEMENT DRAWING CONTAINMENT BUILDING POLAR CRANE NORMAl.. 2'0/25 TON POL._R "q., CitA.NE COUPLED 10 '2.r.,() TON* ct*; NORMAL POLAR I I I \..ow MA.lN Llt.\\T I El 'ZOO I 1-0" I i I I LOW 1\Ul(. LIMIT e.L. ioo I'* ott '" a SPlllA'( P\PlWG - 2135'-0" TOP OF RAIL *-tt I MAIN HOOKH** TO REACH . THIS POINT 67!.o" TO Ci. RAIL EL 10'*0" R LOW EVE EL. O" ACCESS TO CRANE IS AT ELEVATION 2113'-8" NOTES: (1) Top/Reactor Flange Elevation= 2021'-6" (2) Total height Of Simplified Head Assembly (SHA) = 574.40"(M-797-00147-W05) WOLF CREEK REV.22 UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-5 HOOK LIMITS FOR CONTAINMENT BUILDING POLAR CRANE F CRANE ORIENTATION !_ "'( )II_!_ 1 J H -OBSTRUCTION -.-,;.!.--f--,-------- ___ --------- -:.. ," Ill
- I "I [,.,.. . .J.:...,._), _I G,.. ',, I -A .. TOPOP RAIL ., , .. , MAX. LOAD ON EACH TROLLEY WHEEL lw;.l PENOANT* ,. RUNWAY OI---L-RAIL I 0 .I I l-.l L-*.>:, J VIMINI
- Looking North CAPACITY
- MAIN 150 TONS Aux 5 T01111** UFT
- MAIN_l!LFT_jLIN Lift-Aux.
71 Ft ._jL_In. A 7S FT_LIN i *j 1.. p.w:ELBASE f.-/ AREA OF I 1 CHANGE SECTION A-A OPeRATING FLOOR EL. 2047'-6" H__!LFT_O_IN J ....!!..FT. 8-1/2 IN (MAX.AU.allABLE) K-fT __ IN L...iQ_FT-lQ..IN M.....12._FT....Q...IN V---LFT--.LIN W....J!llFT X---LFT -l...JN N.....L_n_3_IN LENGTH OF MAIN 'LINE o..JLFT_6_IN RUNWAY 100 FT..J!....IN B 35 FT ...J!.,..IN (KlGH HOOX) C 11 FT 11-i/2 Ill P _ll_n 0 IN \ MAX. LOAD ON EACH O-FT-IN AREA OF WHEEL ______ LBS R-FT-IN CHANGE RUNWAY RAIL S-FT-IN Size 1ft 1 FT 3 IN1. E u FT ..!!.. IN F 0 FT_Q..IN G 0 FT__!!_IN -175 LBS Z _8_ FT .....!L.IN
- NOTE: *NEAR SIDF." & "LEl'T/RIGHT"'
- FACING CRANE DRIVE SIDE.
- HIGH PENDANT Top of Bail -2083'-6 1/2" ** 5 tons, 2 I:ODII over nev fuel storage ama Aux. boist is bridge mounted monorail syatem REV. 14 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 9.1-6, REV. ARRANGEMENT DRAWING FUEL BUILDING CASK HANDLING CRANE I F._i; Of' -R SlOP & (. COI.II"'N LlfoiE F-7 PLAN .._DeE Of' 8LI'II'(R STill' AREA OF CHANGE N _ Ri'llL 'I:. NEW FUEL SHIPPINC.l CONTAINER S....._ AllEA OECON PIT t I SECTION A-A AREA "MM. ONLY AREA OF CHANGE WOLF CREEK REV. 14 UPDATED SAFETY ANALYSIS REPORT Figure 9.1-7, REV . HOOK LIMITS FOR FUEL BUILDING CASK HANDLING CRANE
. SI'DIT FUEL P<>OL SECTION A*A HOOKt
- TOll> OF PLATFORM-SfOP (TYR) -.t. REQ'D SEC.TION 6*1!. 4'!..91"'
ri" t -WHEE.L -{ 20Go*'<III'MoAM .. L'EW '2041 D-1."-;.1 -,. <l-Z5'*0u MAX . '! OIIEA14E"0 l &I! AM _/ I .1; 1!.! HillS II RAIL AIL 1.[ SPAN PLAN NOTES: FOR INFORMATION REGARDING SPENT FUEL BRIDGE CRANE ACCESS TO THE FAR NORTH EDGE OF FUEL POOL,SEE CCP 09356. CAPM:ITY: 1111.1-E Q.K* IIOI'ioT I'IAIN"'L KOI!>T MA'L 1.1FT SI"EEO M I Ill LIFT SP EE.O t'\All, IIRIDGII .P£1.0 "'1N.IIIP.IOGE 5PEI!.D "'--AREA OF . CHANGE -S TON -'l "!!K -21FPM I 107*
- 7FPM 1: ID'7o *IOFPM t *10 FPM 110-K WOLF CREEK Rev. 15 30FPM"' lOll. 10 l"l'li\t .,o,.. UPDATED SAFETY ANALYSIS REPORT flo\llli!MN<.f:
PLIITfoaJ!o\
- NOJ 5"0""1'1 1DPGI&IUIIII.
- ...e.VIIIII'IGN 2047'-7-l
- FIGURE 9.1-B ARRANGEMENT DRAWING SPENT FUEL BRIDGE CRANE I 8--&---\ 'b.\ ii. l'<<>LF CREEK I"'D" CUIII.I
- l ... _, "F l ' ***DG< CRA!tl RML ... TOP T yp I ** t I ' ' I H--I * ./*h I I c * ----.... *.s I I I I I I I I 'I _.; : -----:-t ---:---------+ ----* ..
- I FUEL STORAGE RACKS I I **
- CASK LOADING I I TOTAL STORAGE CAPACITY -2363 ASSEMBLIES I POOL ,:, (-B I I CNOTE 3) I I I I 1 -l---:-------------"J I I I I ri=§. .. i NOT[5 '* SPAt,NCt FUlL rLJMllt1 TO BE 8.99" a .. Ct:r-..1"1'(.
- lttt)t:A.
Po.,.,.,OtUMC. Or" C.'RAN1:. TO e.'t TO \-lllt.NOl.'E
- I
- FUTURE ' 1 I ; I ASSEMBLIES II I; I I. .. -I i ,_ . I I !I ,., ...... (HJ.NI .... I 'hq-oJ
- t 2. SHOWS NORMAL CRANE TRAVEL LIMITS. THESE LIMITS ARE NOT APPLICABLE WHEN CONCERNING CELLS LOCATED NEAR THE PERIPHERY OF THE FUEL STORAGE POOL WALLS. ;I 1-sv-y 1-IO"*Q-f* _.I..,0 1-Q 1' PLAN ----.. N c J CAREA; CHANGE WOLF CREEK UPDATED SAFETY ANALYSIS REPORT HOOK LIMITS FOR SPENT FUEL POOL BRIDGE CRANE FIGURE 9.1-9 REV. 19 L. r CWII£SSf1l_j_
I I I""'" ii.IUST II R ill ID - """"" -*, I 11111!1 l _,. 2" etwn L llJIIHt _j <$ 11--llll:: 151 \__,.,"'""" ATUCHEO TO OF'a'Afi"'G T.V. CAM£U FOF IPI l'GE I Tll)'lf WOLF CREEK (lf H)J IIJCZLE fUEL FUtL COIISltlE TOI'M tniVE Rev. 0 ! ' J *e-t.:j ""' ! 2999 -.......__F\IH WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-10 REFUELING MACHINE WOLF CREEK A_j ELEVATOR Rev. 14 I I I I I I I .l VIEW A-A WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-11 NEW FUEL ELEVATOR
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,. J __ :: 7::ii:t! ti-' / QL Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-13 ROD CLUSTER CONTROL CHANGING FIXTURE 368.00" Rev. 0 , .I WOLF CREEK
- 2.94" t WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-14 SPENT FUEL HANDLING TOOL
, I WOLF CREgK 25.25" 2.94" t Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-15 NEW FUEL HANDLING TOOL WOLF CREEK "8 II SEE "A" ll FT I NG HOLES PLAN VIEW OF UPPER CORE SUPPORT STRUCTURE Ll FTING RIG UPPER CORE BARREL INTERNALS REACTOR VESSEL flANGE Rev. 0 "A" SEE "B" WOLF CREEK L UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-16 UPPER CORE BARREL HANDLING FIXTURE I , LIFTING LUG WOLF CREEK Rev. 0 HYDRAULIC PISTONS LATCHING MECHANISM WOLF CREEK 17942-1 UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-17 QUICK-ACTING STUD TENSIONER \ I WOLF CREEK GUIDE TUBE CAAD'Y SECTION 1 SECTION 2 OCCURS A TOTAL OF 6 TIMES OOWN THE GUIDE TUBE RADIAl CLEARANCE BETWEEN DFIIVE ROD AND CARD"' 0.325" LARGEST DIAMETER THAT CAN PASS THROUGH GUIDE TUBE CARD-2:.4" REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 9.1-18 MODEL OF CRITICAL BUCKLING LOAD FOR THE DRIVE ROD WOLF CREEK Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-19 BUCKLING LOAD FOR SECTION 2 \ l WOLF CREEK A I. 138" I ; = -p REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 9.1-20 BUCKLING LOAD FOR SECTION 1 WOLF CREEK w W "' WEIGHT OF THE UPPER PACKAGE, HEAD POLAR CRANE HOOKS, AND CABLE w 1 = WEIGHT OF THE VESSEL FLANGE, NOZZLES, AND REGION IN BETWEEN kin ,. SPRING CONSTANT OF INLET NOZZLE REGION k 0 n "" SPRING CONSTANT OF OUTLET NOZZLE REGION ks "" SPRING CONSTANT OF SUPPORTS kv = SPRING CONSTANT OF VESSEL AND FLANGE USING EQUIVALENT CYLINDER ANALYSIS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-21 SPRING MASS SYSTEM OF THE HEAD UPPER PACKAGE AND REACTOR VESSEL \ l A Rev. 0 I WOLF CREEK A WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1-22 UPPER PORTION OF VESSEL ' I 1 WOLF CREEK Table 9.1A-1 Summary of the Criticality Safety Analysis for the MZTR Storage Configuration Design Basis Burnups at 5.0 0.05 wt% 235 Uinitial enrichment 0 in Region 1 50 in Region 2 40.75 in Region 3 Temperature for Analysis 20 C Uncertainties Manufacturing tolerances (Table 4.5.1) 0.0059 Water-gap (horizontal) 0.0014 Water-gap (vertical) 0.0003 Burnup (Region 2) 0.0056 Burnup (Region 3) 0.0001 Eccentricity in position negativ e KENO5a statistics (95%/95%) 0.0003 Bias statistics (95%/95%) 0.0012 Statistical combination of uncertainties
- 0.0084 Region 1 Fuel Description 5.0 wt%235 Uwith 16 IFBA rods 4.6 wt%235 Uwith no IFBA rods Reference k eff (KENO5a) 0.9266 0.9294 Total Uncertainty (above) 0.0084 0.0084 Calculational Bias (see Appendix A) 0.0030 0.0030 Axial Burnup Effect negative negative Temperature Correction to 4 C (39 F)0.0020 0.0020 Maximum k eff 0.9400 0.9428 Limiting k eff 0.9500 0.9500* Square root of the sum of the squares.
Rev. 14 SEISMIC RESTRAINT ROLLER <22" DIA.) SNUBBER Ref. 10466-M-063-00070 VERTICAL FACE of RUNWAY CRANE GIRDER REV. 23 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 9.1-24 PLAN VIEW DETAILS OF SEISMIC RESTRAINT ASSEMBLy POLARCRANE R AIL SEISMIC RESTRAINT VERTICAL FACE of the POLAR CRANE SUPPORT POLAR CRANE KAIL UPKICK LUG POLAR CRANE GIRDER G1 & G2 Ref. 10466-M-063-00070 REV. 23 WOLF CREEK UPDATED SAFETY AN A LYSIS REPORT Figure 9.1-25 DETAILS of SNUBBER EARTHQUAKE ASSEMBLY Polar Crane Parked Position REV. 23 WOLF CREEK UPDATED SAFETY AN A LYSIS REPORT Figure 9.1-26 POLAR CRANE in the "PARKED POSITION" at AZ. 315° during PLANT OPERATION WOLF CREEK APPENDIX 9.1A FUEL STORAGE RACK ANALYSIS 9.1A.1 THE HIGH DENSITY RACK (HDR) DESIGN CONCEPT 9.1A.1.1 Introduction Historically, spent fuel rack designs have been based on conservative assumptions that could easily be accommodated since it was not planned to store large numbers of high exposure spent fuel assemblies on-site. Previously it was anticipated that only small amounts of high exposure fuel assemblies (1/4 to 1/2 of a full core load) would normally be stored in the spent fuel pool at any one time. Additionally it was anticipated that occasionally (e.g., for
inservice inspection of the reactor vessel internals) the entire core would be unloaded and temporarily stored in the initial spent fuel pool. Therefore, the spent fuel storage rack design was based on the conservative assumption that all fuel rack storage positions would be occupied by fresh unirradiated fuel assemblies of the highest initial enrichment that was foreseen as being usable in that facility. The penalty in achievable fuel storage density associated with this conservative design assumption was relatively small under the circumstances anticipated and easily accommodated by a conservative fuel rack design. The potential penalty associated with this conservative design basis is no longer small when long-term on-site storage of spent fuel is a necessity. There is no situation where more than one full core load of fresh unirradiated fuel assemblies is to be stored in the fuel storage pool. Therefore, it is unnecessary and wasteful to base the entire fuel storage rack design on the assumption of fresh unirradiated fuel of the highest initial enrichment. In the previous maximum density rack (MDR) design concept utilized by Wolf Creek, the spent fuel pool was divided into two separate and distinct regions which, for the purpose of criticality considerations may be considered as separate pools. Suitability of this design assumption regarding pool separability was assured through appropriate design restrictions at the boundaries between Region 1 and Region 2. The smaller region, Region 1, of the pool was designed on the basis of accepted conservative criteria which allowed for the safe storage of a number of fresh unirradiated fuel assemblies (including a full core loading if that should prove necessary). The larger region of the pool, Region 2, was designed to store irradiated fuel assemblies. The change in criteria was the recognition of actual fuel and fission product inventory, accompanied by a system for checking fuel prior to moving any fuel assembly from Region 1 to Region 2. In the HDR design concept, currently utilized by Wolf Creek, the rack modules for the fuel storage pool are designed for storage of both new fuel and spent fuel. Spent fuel storage is designated into Regions based upon initial enrichment and accumulated burnup. Region 1 is designed to accommodate new fuel with a maximum nominal enrichment of 5.0 weight percent of U-235, with a minimum of 16 IFBA or 4.6 weight percent U-235, no IFBA. Region 2 and Region 3 are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within an acceptable domain as detailed in Section 3.7 of the Technical Specifications. 9.1A-1 Rev. 14 WOLF CREEK 9.1A.1.2 Design Bases The high density fuel storage racks are designed to assure that the effective neutron multiplication factor (keff) in the fuel storage pool is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at the temperature within the operating range corresponding to the highest reactivity. The fuel storage racks are designed to accommodate any and all of the following Westinghouse fuel assembly types: 17x17 OFA, 17x17 Standard, and 17x17 Vantage 5H (V5H), with a maximum nominal initial enrichment of 5.0 wt% 235U and a minimum of 16 Integral Fuel Burnable Absorber (IFBA) rods. The OFA designation is used generically throughout this discussion and includes V-5 and V+ fuel. Additional restrictions are specified to allow the storage of these fuel assembly types without IFBA rods. USAR Section 9.1A.6 lists the applicable codes, standards, and regulations or pertinent sections thereof relied on for the criticality safety analysis. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances. All uncertainties are statistically combined, such that the final keff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. Enrichments less than 5.0 wt% 235U are also evaluated, and soluble boron concentrations necessary to protect against postulated accidents are determined. USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, keff, including bias, uncertainties, and calculational statistics, shall be less than or equal to 0.95, with 95% probability at the 95% confidence level. To assure that reactivity in the fuel storage pool is always less than the calculated maximum reactivity, the following conservative assumptions were made in performing the criticality safety analysis: Moderator is unborated water at a temperature that results in the highest reactivity (4C, corresponding to the maximum possible moderator density). No soluble poison or control rods are assumed to be present for normal operations, although the additional margin due to the presence of soluble boron is identified. The effective multiplication factor of an infinite radial array of fuel assemblies was used except for the assessment of peripheral effects and certain abnormal/accident conditions where neutron leakage is inherent. Neutron absorption in minor structural members is conservatively neglected, i.e., spacer grids are replaced by water. Depletion calculations assume conservative operating conditions; highest fuel and moderator temperature and an allowance for the soluble boron concentrations during incore operation. The assemblies with IFBA rods are assumed to contain the minimum possible number of IFBA rods (i.e., 16), in a conservative loading pattern, with a conservative length of 120 inches. Further, the IFBA loading used in the analyses is reduced by an uncertainty of 5%. 9.1A-2 Rev. 14 WOLF CREEK 9.1A.1.3 Design Description Because the fuel storage racks are designed to accommodate any and all of the following Westinghouse fuel assembly types: 17x17 OFA, 17x17 Standard, and 17x17 Vantage 5H (V5H), with a maximum initial enrichment of 5.0 wt% 235 U, the most reactive assembly type was identified via independent criticality calculations. To assure the acceptability of the racks for storage of any and all of these assembly types, the most reactive assembly is used in the criticality analyses. At zero burnup, the 17x 17 OFA assembly has the greatest reactivity in the storage racks, and is therefore used as the design basis fuel assembly.The Mixed-Zone Three-Region (MZTR) configuration uses fuel assemblies with high discharge burnup as barrier fuel to isolate fresh fuel assemblies in order to achieve an acceptable keff in the fuel storage pool. Three separate storage regions are provided, with independent criteria defining the highest potential reactivity in each of the three regions: Region 1 is designed to accommodate new un-irradiated (fresh) fuel with a maximum nominal enrichment of 5.0 wt% 235 U and a minimum of 16 IFBA rods, or fuel of equivalent reactivity (e.g., 4.6 wt% 235U maximum enrichment without IFBA rods). Further, Region 1 cells on the periphery of the pool, that are adjacent to a concrete wall, may accommodate fresh fuel assemblies with maximum nominal enrichment of 5.0 wt% 235 U and no IFBA rods. Region 2 is designed to accommodate fuel with a maximum nominal initial enrichment of 5.0 wt% 235 U and high ( 50 MWd/kgU) discharge fuel burnup, or fuel of initial enrichment and burnup combinations yielding an equivalent reactivity. Region 2 locations are used to isolate Region 1 fuel assemblies from other Region 1 and Region 3 fuel assemblies. Region 3 is designed to accommodate fuel with a maximum nominal initial enrichment of 5.0 wt% 235U and typical (40.75 burnup 50 MWd/kgU) discharge fuel burnup, but can accommodate any spent fuel with discharge fuel burnup greater than or equal to 40.75 MWd/kgU. Additionally, fuel of initial enrichment and burnup combinations yielding an equivalent reactivity are acceptable for storage in Region 3. The water in the fuel storage pool normally contains soluble boron. The presence of this soluble boron results in a large sub-criticality margin under actual operating conditions. However, NRC guidelines specify that the criticality limit, keff 0.95 for normal storage, remain valid under accident conditions that also assume the loss of all soluble boron in the fuel storage pool. Under the double contingency principle given in ANSI N-16.1-1975 (Reference 6) and in the April 1978 NRC letter (Reference 3), credit for soluble boron under abnormal or accident conditions, however, is allowed, because only a single independent accident need be considered at one time. The consequences of abnormal and accident conditions are evaluated for the fuel storage pool. "Abnormal" refers to conditions which may reasonably be expected to occur during the lifetime of the plant, and "accident" refers to conditions which are not expected to occur, but nevertheless must be protected against. 9.1A-3 Rev. 14 WOLF CREEK 9.1A.2 CRITICALITY ANALYSES FOR THE FUEL STORAGE POOL 9.1A.2.1 Description of Fuel Storage Pool Conditions 9.1A.2.1.1 Normal Operating Conditions In the MZTR configuration, the fresh fuel cells (Region 1 ) are located alternately along the periphery of the fuel storage pool (where neutron leakage reduces reactivity) or along the boundary between two storage modules (where the water gap provides a flux-trap which reduces reactivity). High burnup fuel in Region 2 affords a low-reactivity barrier between fresh fuel assemblies and Region 3 fuel of intermediate burnup. Numerous configurations of the various assemblies within the fuel storage pool are possible. The criteria for determining an acceptable loading arrangement in the MZTR configuration for fuel of different burnups are as follows: Region 1 cells are only located along the outside periphery of the storage modules and must be separated by one or more Region 2 (burnup > or = 50 MWd/kgU for 5.0 wt% 235U, or equivalent burnup/enrichment combinations) cells.Region 1 cells may be located directly across from one another when separated by a water gap. Along the interface between storage modules the water gap is 1.5" +/1/8" (excluding sheathing). The outer rows of alternating Region 1 and Region 2 cells must be further separated (isolated) from the internal Region 3 cells by one or more Region 2 cells. Fresh fuel assemblies without IFBA rods and a maximum enrichment of 5.0 wt% 235U may be stored in any periphery Region 1 cell location that is next to a concrete wall. Prior to approaching the reactor end-of-life, not all storage cells are needed for spent fuel. Therefore, an alternative (interim) configuration may be used in which the cells of selected modules may be loaded in a checkerboard pattern of fresh fuel (or spent fuel of any burnup) with empty cells. A checkerboard configuration is intended primarily to develop a simple configuration of Region 1 cells and facilitate storage of fresh (unburned) and low burnup fuel. The principles involved in the design and specification of an acceptable loading arrangement in the interim checkerboard configuration are as follows: Fuel with maximum nominal enrichment of 5 wt% 235U and a minimum of 16 IFBA rods, or fuel of equivalent reactivity (e.g., 4.6 wt% 235 U maximum enrichment without IFBA rods), is placed in an alternating checkerboard style pattern with empty cells (i.e., fuel assemblies are surrounded on all four sides by empty cells). Fuel assemblies may not be located directly across from one another, even when separated by a water gap. 9.1A-4 Rev. 14 WOLF CREEK So long as the checkerboard pattern is maintained in a linear array greater than or equal to 2x2, the arrangement may be used anywhere in the pool. More than one checkerboard pattern may be used, as long as the limitations discussed herein are adhered to. A checkerboard region may be bounded by either a water gap, empty rack cells, Region 2 fuel assemblies, or Region 3 fuel assemblies. MZTR and checkerboard storage shall not be developed within the same rack.Non-fueled items such as trash baskets and dummy fuel assemblies may be stored anywhere in the fuel storage pool. Damaged fuel storage baskets must be stored in any cell that allows fuel assembly storage. Figure 9.1A-3 defines the acceptable burnup domains for spent fuel and illustrates the limiting burnup for fuel of various initial enrichments for both Region 2 (upper curve) and Region 3 (lower curve). Both curves assume that the fresh fuel (Region 1) has a maximum nominal enrichment of 5.0 wt% 235 U.Criticality analyses demonstrate that the most reactive configuration occurs along the boundary between modules where the water gap affords a neutron flux trap. Along the periphery of the modules facing the concrete wall of the pool, the reactivity is substantially lower due to neutron leakage. The bounding criticality analyses are summarized in Table 9.1A-1 for the design basis MZTR storage configuration and in Table 9.1A-2 for the interim checkerboard storage configuration. In both cases, the single accident condition of the loss of all soluble boron is assumed. The calculated maximum reactivity of 0.943 (corresponding to the design basis MZTR storage configuration) is within the regulatory limit of 0.95. This maximum reactivity includes calculational uncertainties and uncertainties in reactivity due to manufacturing tolerances (95% probability at the 95% confidence level), an allowance for uncertainty in depletion calculations, and the evaluated effect of the axial distribution in burnup.The value of keff in Table 9.1A-1 assumes no soluble boron to be present. For normal operations, a minimum soluble boron concentration of 2165 ppm is maintained in the Wolf Creek fuel storage pool. This concentration of soluble boron provides a large safety margin for sub-criticality. As cooling time increases in long-term storage, decay of 241Pu (and growth of 241Am) results in a continuous decrease in reactivity, which provides an increasing sub-criticality margin with time. No credit is taken for this decrease in reactivity other than to indicate conservatism in the calculations. The burnup criteria identified in Figure 9.1A-3, for acceptable storage in Region 2 and Region 3, are used in appropriate administrative procedures to assure verified burnup as specified in the proposed Regulatory Guide 1.13, Revision 2 (Reference 14). Soluble poison is present in the pool water during fuel handling operations, and this serves as a further margin of safety and as a precaution in the event of fuel misplacement during fuel handling operations. 9.1A-5 Rev. 14 WOLF CREEK 9.1A.2.1.2 Abnormal and Accident Conditions Although credit for the soluble poison normally present in the fuel storage pool water is permitted under abnormal or accident conditions, most abnormal or accident conditions will not result in exceeding the limiting reactivity (keff of 0.95) even in the absence of soluble poison. The effects on reactivity of credible abnormal and accident conditions are discussed in Section 9.1A.2.2.5 and summarized in Table 9.1A-3. Of these abnormal or accident conditions, only two have the potential for a more than negligible positive reactivity effect. These include: (1) the inadvertent misplacement of a fresh fuel assembly and (2) the mis-location of a fresh fuel assembly into a position external and adjacent to a storage rack. The inadvertent misplacement of a fresh fuel assembly has the potential for exceeding the limiting reactivity, should there be a concurrent and independent accident condition resulting in the loss of all soluble poison. Assuring the presence of soluble poison during fuel handling operations will preclude the possibility of the simultaneous occurrence of the two independent accident conditions. The largest reactivity increase would occur if a fresh fuel assembly of 5.0 wt% 235U enrichment were to be inadvertently loaded into an empty cell in the checkerboard configuration with the remainder of the rack fully loaded with fuel of the highest permissible reactivity. For the MZTR configuration, when a fresh fuel assembly of 5.0 wt% 235U enrichment is inadvertently loaded into a Region 2 location (with the remainder of the rack fully loaded with fuel of the highest permissible reactivity), the overall reactivity is slightly less reactive. However, it still exceeds the limiting value without the presence of soluble boron. Under these accident conditions, credit for the presence of soluble poison is permitted by the NRC guidelines. Calculations indicate that 500 ppm soluble boron would be adequate to reduce the keff to below the reference keff value (Table 9.1A-1). This soluble boron concentration bounds all other accidents and is well below the 2165 ppm soluble boron concentration that is maintained in the Wolf Creek fuel storage pool. It is possible for a fuel assembly to be dropped or mis-located in the fuel storage pool such that it may be situated outside and adjacent to a storage rack. The calculated keff value for the worst case situation exceeds the limit on reactivity in the absence of soluble boron. Because this case is less severe than the misplaced fresh fuel assembly accident, it requires less than 500 ppm soluble boron to reduce the keff to the reference value (Table 9.lA-l). Multiple misplaced fuel assemblies were also considered. This accident is bounded by assuming every cell is filled with 5.0% wt% 235U, 16 IFBA fuel, and requires 2165 ppm boron to control keff.9.1A.2.2 Analytical Methodology To assure the acceptability of the racks for storage of all fuel assembly design types, the most reactive assembly type was identified by independent criticality calculations. This most reactive assembly is the reference assembly used in the criticality calculations. In addition, a nominal fuel storage cell is also used in the criticality calculations. This nominal fuel storage cell represents the fuel pool storage cells. 9.1A-6 Rev. 14 WOLF CREEK 9.1A.2.2.1 Reference Fuel Assembly The fuel storage pool racks are designed to accommodate any and all of the following Westinghouse fuel assembly types: 17x17 OFA, 17x17 Standard, and 17x17 Vantage 5H (V5H), with a maximum nominal initial enrichment of 5.0 wt% 235U. Additional restrictions are specified to allow the storage of any of the aforementioned fuel assembly types without IFBA rods. Independent criticality calculations were performed to identify the most reactive assembly type. The results of these calculations show that at zero burnup the 17xl 7 OFA assembly has the greatest reactivity in the storage racks, and thus, is the design basis fuel assembly. The Westinghouse OFA is a 17 x 17 array of fuel rods with 25 rods replaced by 24 control rod guide tubes and l instrument thimble. Table 9.1A-4 summarizes the fuel assembly design specifications. At burnups beyond approximately 25 MWd/kgU, the 17x17 Standard and 17x17 Vantage 5H become the most reactive assembly types. These two assembly types are essentially identical. Therefore, for the determination of the equivalent enrichments associated with Regions 2 and 3, the reactivity of the V5H assembly was related to an initial enrichment for the 17x17 OFA assembly. The fresh fuel assemblies were assumed to contain the minimum possible number of IFBA rods (i.e., 16) in a conservative loading pattern with a conservative length of 120 inches. The IFBA rods are characterized by a thin ZrB 2 coating on the outside of the fuel pellets. Because B-10 in ZrB 2 is a strong neutron absorber, it reduces the assembly reactivity, and thus, enables the storage of fuel with high initial enrichment. The IFBA loading was assumed to be 2.25 mg B-10/inch with an uncertainty of 5%. The IFBA loading was assumed to be reduced by the 5% uncertainty in this analysis. With 16 IFBA rods present, the reactivity of the assembly does not exhibit a peak with burnup, and thus the calculated reactivity of the fresh assembly is bounding. The IFBA rods are modeled in the fresh fuel assemblies only; no credit is taken for residual IFBA in the Region 2 and Region 3 fuel assemblies. 9.1A.2.2.2 High Density Reference Fuel Storage Cell A nominal fuel storage pool cell was used for the criticality calculations for the fuel storage pool cells. Stainless steel boxes are arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes. The walls of the stainless steel boxes contain a Boral panel (attached by a stainless steel sheathing) centered on each side. Peripheral cells use stainless steel sheathing on the outside wall to attach the Boral panel. The fuel assemblies are normally located in the center of each storage cell on a nominal lattice spacing of 8.99 inches.9.1A.2.2.3 Analytical Technique The principal method for criticality analysis of the high density storage racks is the three-dimensional Monte Carlo KENO5a (Reference l) code, as developed by the Oak Ridge National Laboratory as part of the SCALE 4.3 package. Independent verification calculations were performed with the MCNP (version 4a) code (Reference 2), a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. The KENO5a calculations used the 238-group SCALE cross-section library and NITAWL (Reference3) for 238 U resonance shielding effects (Nordheim integral treatment). Benchmark calculations, presented in section 9.1A.2.2.6, indicate a bias of 0.0030 with an uncertainty of 0.0012 for KENO5a and 0.0009 0.0011 for MCNP4a, both evaluated at the 95% probability, 95% confidence level (Reference 4). 9.1A-7 Rev. 14 WOLF CREEK Fuel depletion analyses during core operation were performed with CASMO-3, a two-dimensional multigroup transport theory code based on capture probabilities (References 5-7). Restarting the CASMO-3 calculations in the storage rack geometry at 4C yields the two-dimensional infinite multiplication factor (k)for the storage rack. Parallel calculations with CASMO-3 for the storage rack at various enrichments enable a reactivity equivalent enrichment (fresh fuel) to be determined that provides the same reactivity in the rack as the depleted fuel. CASMO-3 was also used to determine the small reactivity uncertainties (differential calculations) of manufacturing tolerances. In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells. KENO5a and MCNP4a Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the KENO5a-calculated reactivity and to assure convergence, a minimum of 5 million neutron histories in 1,000 generations of 5,000 neutrons per generation were accumulated in each single assembly infinite array calculation. A minimum of 20 million neutron histories in 2,000 generations of 10,000 neutrons per generation were accumulated in each multiple assembly (MZTR and checkerboard) configuration. Figure 9.1A-1 represents the reference MZTR geometric model used in the KENO5a calculations. This figure is intended to show the arrangement of fuel assemblies modeled, and not the specific details of the model. With reflecting boundary conditions, this model effectively describes the entire pool in the MZTR configuration, including the water gap between storage modules. In the axial direction, the full length 144-inch fuel assembly was described assuming 30-cm water reflector, top and bottom. In addition, the axial variation in burnup was explicitly modeled and resulted in a slightly lower reactivity than the reference design calculation (which assumes uniform axial burnup). Figure 9.1A-2 represents the reference checkerboard geometric model used in the KENO5a calculations. With reflecting boundary conditions, this model effectively describes the entire pool in the checkerboard configuration, including the water gap between storage modules. These large models were also used to investigate uncertainties in the configurations and the consequences of potential accident conditions, including a misplaced fresh fuel assembly. Because NITAWL-KENO5a does not have burnup capability, burned fuel was represented by fuel of equivalent enrichment as determined by CASMO-3 calculations in the storage cell (i.e. an enrichment which yields the same reactivity in the storage cell as the burned fuel). In tracking long-term (30-year) reactivity effects of spent fuel, previous CASMO-3 calculations have demonstrated a continuous reduction in reactivity with time (after Xe decay) (Reference 8) due primarily to 241Pu decay and 241 Am growth. 9.1A.2.2.3.1 Fuel Burnup Calculations and Uncertainties CASMO-3 was used for burnup calculations in the hot operating condition. CASMO-3 has been extensively benchmarked (References 7 and 9) against cold, clean, critical experiments (including plutonium-bearing fuel), Monte Carlo calculations, reactor operations, and heavy element concentrations in irradiated fuel. In addition to burnup calculations, CASMO-3 was used for evaluating the small reactivity increments (by differential calculations) associated with manufacturing tolerances and for determining temperature effects. 9.1A-8 Rev. 14 WOLF CREEK In the CASMO-3 geometric model, each fuel rod and its cladding were described explicitly and reflective boundary conditions were used at the centerline of the Boral and steel plates between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells in the X-Y plane and provide a conservative estimate of the uncertainties in reactivity attributed to manufacturing tolerances. Conservative assumptions of moderator and fuel temperatures and the average operating soluble boron concentrations were used to assure the highest plutonium production and hence conservatively high values of reactivity during burnup. Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations, an allowance for uncertainty in reactivity was assigned based upon the assumption of 5% uncertainty in burnup. At the design basis burnups of 40.75 and 50 MWd/kgU, the uncertainties in burnup are 2.04 and 2.5 MWd/kgU respectively. These uncertainties correspond to approximately 0.013 k and 0.016 k in the fuel infinite multiplication factor. (The majority of the uncertainty in depletion calculations derives from uncertainties in fuel and moderator temperatures and the effect of reactivity control methods (e.g., soluble boron). For depletion calculations, bounding values of these operating parameters were assumed to assure conservative results in the analyses). To evaluate the reactivity consequences of the uncertainties in burnup, independent MZTR calculations were made with fuel of 38.5 and 47.5 MWd/kgU burnup in Regions 2 and 3, and the incremental change from the reference burnups assumed to represent the net uncertainties in reactivity attributable to uncertainty in depletion calculations. These calculations resulted in an incremental reactivity uncertainty in keff of 0.0056k for Region 2 and 0.0001k for Region 3. These effects would be lower for lower initial enrichments and burnups. The fresh unburned fuel in Region 1 strongly dominates the reactivity which tends to minimize the reactivity consequences of uncertainties in depletion calculations. The allowance for uncertainty in the burnup calculations is believed to be conservative, particularly in view of the substantial reactivity decrease with time as the spent fuel ages. 9.1A.2.2.3.2 Effect of Axial Burnup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution. As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower regions. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of high neutron leakage. Consequently, it is expected that over most of the burnup history, fuel assemblies with distributed burnups will exhibit a slightly lower reactivity than that calculated for the uniform average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup. Among others, Turner (Reference 10) has provided generic analytic results of the axial burnup effect based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup. 9.1A-9 Rev. 14 WOLF CREEK Based on axial burnup distributions of spent fuel (axial burnup data for assemblies from the Wolf Creek plant with average burnups of 50.10 and 36.84 MWd/kgU were normalized to the Region 2 and 3 burnups, 50 and 40.75 MWd/kgU, respectively), three-dimensional KENO5a calculations were performed. In these calculations, the axial height of the Region 2 and 3 fuel was divided into 5 axial zones, each with an average enrichment equivalent to the burnup of that zone. The selection of the five axial zones was based on the shapes of the axial burnup distributions. The resulting keff was 0.007 k less than the reference keff (which assumes uniform axial burnup). Fuel of lower initial enrichments (and lower burnup) would have a more negative reactivity effect as a result of the axial variation in burnup. These estimates are believed to be conservative since smaller axial increments in the calculations have been shown to result in lower incremental reactivities (Reference 10). 9.1A.2.2.4 Criticality Analyses Uncertainties and Tolerances A number of tolerances result in reactivity uncertainties which must be considered in the criticality analyses. 9.1A.2.2.4.l Nominal Design For the nominal MZTR storage configuration, the bounding criticality analyses are summarized in Table 9.1A-1. The NITAWL-KENO5a calculated keff value is combined with all the known uncertainties and corrected for bias and temperature (see Section 9.lA.2.2.5.l for temperature correction), to determine the maximum keff value with a 95% probability at the 95% confidence level (Reference 4). For the interim loading pattern of fresh fuel checkerboarded with empty cells, the bounding criticality analyses are summarized in Table 9.1A-2. An alternate calculation with a 2X2 checkerboard pattern bordered on all sides with Region 3 fuel assemblies resulted in a maximum keff of 0.903 with a 95% probability at the 95% confidence level. Therefore, the checkerboard loading pattern may be used anywhere in any module provided that the checkerboard pattern is a linear array greater than or equal to 2X2 and is bordered by any of the following: the water gap between rack modules, the water gap between a rack module and the pool wall, empty rack cells, Region 2 fuel assemblies, and/or Region 3 fuel assemblies.9.1A.2.2.4.2 Uncertainties Due to Manufacturing Tolerances The uncertainties due to manufacturing tolerances are summarized in Table 9.1A-5 and discussed below. 9.1A.2.2.4.2.1 Boron Loading Tolerances The Boral absorber panels are manufactured with a tolerance limit in B-10 content which assures that at any point, the minimum B-10 areal density will not be less than 0.030 g/cm
- 2. Differential CASMO-3 calculations for an infinite array of fresh assemblies with the minimum tolerance B-10 loading results in an incremental reactivity uncertainty of 0.0044k. This value was conservatively assumed to be the B-10 loading uncertainty. 9.1A-10 Rev. 14 WOLF CREEK 9.1A.2.2.4.2.2 Boral Width Tolerance The differential CASMO-3 calculated reactivity uncertainty is 0.0010k, when the reference storage cell design has the minimum tolerance for Boral panel thickness.9.1A.2.2.4.2.3 Tolerances in Cell Lattice Spacing The differential CASMO-3 calculations determine an uncertainty of 0.0016k in the calculated reactivity when the minimum manufacturing tolerance on the inner box dimension is used. The minimum manufacturing tolerance on the inner box dimension directly affects the storage cell lattice spacing between fuel assemblies.9.1A.2.2.4.2.4 Stainless Steel Thickness Tolerances The nominal stainless steel thickness for the stainless steel box also has an impact on the calculation of reactivity. The reactivity uncertainty of the expected stainless steel thickness tolerances was calculated with CASMO-3 and was determined to be 0.0002k.9.1A.2.2.4.2.5 Fuel Enrichment and Density Tolerances The design maximum enrichment is 5.0 0.05 wt% 235U. Separate CASMO-3 burnup calculations were made for fuel of the maximum enrichment (5.05 wt% 235 U) and for the maximum UO 2 density (10.61 g/cm 3). Reactivities in the storage cell were then calculated using the restart capability in CASMO-3. For fresh fuel, the incremental reactivity uncertainties were 0.0023k for the enrichment tolerance and 0.0026k for the tolerance in fuel density. 9.1A.2.2.4.3 Water-Gap Spacing Between Modules The water-gap between modules, which is 1.5 inches (excluding sheathing), constitutes a neutron flux-trap for the storage cells of facing racks. KENO5a calculations were made with the reference MZTR model to determine the uncertainty associated with a water-gap tolerance. Due to the asymmetries in the MZTR pool configuration, the effect of the horizontal and vertical water gaps (see Figure 9.1A-1) were calculated separately. From these calculations, it was determined that the incremental reactivity consequence (uncertainty) for a water-gap tolerance of 1/8 inches is 0.0014 k (horizontal gap) and 0.0003k (vertical gap). The racks are constructed with the base plate extending beyond the edge of the cells which assures that the minimum spacing between storage modules is maintained under all credible conditions. 9.1A.2.2.4.4 Eccentric Fuel Positioning The fuel assembly is assumed to be centered in the storage rack cell.
Calculations were made using KEN05a assuming the fuel assemblies were located in the corners of the storage rack cells (four-assembly clusters at the closest possible approach). These calculations indicated that the reactivity effect is small and negative. Therefore, the reference case in which the fuel assemblies are centered is controlling and no uncertainty for eccentricity is necessary. 9.1A-11 Rev. 14 WOLF CREEK 9.1A.2.2.5 Abnormal and Accident Conditions The reactivity effects of abnormal and accident conditions are summarized in Table 9.1A-3. 9.1A.2.2.5.1 Temperature and Water Density Effects The moderator temperature coefficient of reactivity is negative; a moderator temperature of 4C (39F) was assumed for the reference calculations, which assures that the true reactivity will always be lower over the expected range of water temperatures. Temperature effects on reactivity have been calculated (CASMO-3) and the results are shown in Table 9.1A-6. In addition, the introduction of voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown in Table 9.1A-6. With soluble boron present, the temperature coefficients of reactivity would differ from those listed in Table 9.1A-6. However, the reactivities would also be substantially lower at all temperatures with soluble boron present. The data in Table 9.1A-6 is pertinent to the higher-reactivity unborated case. For the dominant Region 1 fuel, the value of k between calculations at 20 C and 4 C is 0.0020 k. Since the KENO5a code cannot properly handle temperature dependence, all KENO5a calculations were performed at 20C and a temperature correction factor (+0.0020 k) was applied to the results. 9.lA.2.2.5.2 Lateral Rack Movement The possibility of reductions in the rack-to-rack gaps and the resulting criticality consequences have also been reviewed. Criticality evaluations are sensitive to these gap dimensions, since the inter-rack gaps provide a flux trap which reduces the reactivity. Rack-to-rack gap reductions are a concern subsequent to dynamic events which are severe enough to displace the racks laterally or produce fuel-to-rack cell wall impacts of sufficient magnitude to exceed cell wall material yield strength (i.e., produce plastic deformation). The criticality analyses are based on the minimum nominal rack to rack gap of
1.5 inches
(excluding sheathing). Thus, the outer sheathing wall-to-outer sheathing wall gap is 1.35 inches. This gap dimension is maintained during initial installation and subsequent to dynamic loadings, and is ensured by fabrication of the 3/4 inch base plate extensions on each rack. Momentary reductions in these gaps may be caused by the swaying of the tops of the racks during seismic events, during which the tops of the cells may actually come into contact. Even under these circumstances, the bottoms of the cells in adjacent racks are still maintained at the 1.5 inch dimension due to the base-plate extensions. Transient reduction in the inter-rack gap dimension below 1.5 inches is acceptable because of the presence of soluble boron which may be credited during seismic events. Additionally, a time-history plot of the inter-rack gaps (see Figure 9.1A-25 through Figure 9.1A-27) indicates that the gaps are reduced for a very short period of time before being restored to the minimum of 1.5 inches. 9.1A-12 Rev. 14 WOLF CREEK 9.1A.2.2.5.3 Rack-Gap Changes Another consideration which could potentially reduce the inter-rack gap is the impact of the fuel assembly on the inside of the cell wall during seismic events. If these impacts are of sufficient magnitude to allow plastic deformation of the cell wall membrane, then permanent displacement of the cell would take place, thus reducing the inter-rack gap. The largest fuel assembly-to-cell wall impact load is determined to be 840 pounds (see USAR Section 9.1A.4.3.4.6). Evaluations on the local cell wall integrity (see USAR Section 9.1A.4.3.5.3) have determined that the load required to produce permanent deformation (i.e., exceed the cell membrane material yield strength) exceeds the calculated load of 840 pounds by a factor of approximately 4. Therefore, there are no criticality concerns related to the reductions in inter-rack gaps from plastic deformation of the cell wall. 9.1A.2.2.5.4 Abnormal Location of a Fuel Assembly In the MZTR configuration, the abnormal location of a fresh unirradiated fuel assembly of 5.0 wt% 235U enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation (keff of 0.95). This would occur if a fresh fuel assembly of the highest permissible enrichment were to be inadvertently loaded into either a Region 2 or Region 3 storage cell. Calculations (KENO5a) confirmed that the highest reactivity, including uncertainties, for the worstcase postulated accident condition (fresh fuel assembly in Region 2) would exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations indicate that a soluble poison concentration of 440 ppm boron would be required to limit the maximum reactivity to the reference keff value (Table 9.lA-1), including all uncertainties and biases, under this maximum postulated accident condition. In the checkerboard configuration, the worst case postulated accident condition (fresh fuel assembly inadvertently loaded into an empty cell) would also exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the fuel storage pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations indicate that a soluble poison concentration of 500 ppm boron would be required to limit the maximum reactivity to the reference keff value (Table 9.1A-1), including all uncertainties and biases, under this maximum postulated accident condition. The accident scenario of all rack cells fully loaded with fresh fuel assemblies, each with a minimum of 16 IFBA (1.5x) rods, determined the concentration of soluble boron required to maintain the keff in the spent fuel pool less than or equal to the regulatory limit (keff 0.95). The most reactive assembly type at zero burnup was identified to be the Westinghouse OFA assembly and the design basis fuel assembly was defined to be the Westinghouse OFA assembly with a maximum nominal enrichment of 5.0% wt% 235 U and a minimum of 16 IFBA (1.5x) rods, (in a conservative loading pattern). Consistent with the original analyses, the KENO5a computer code was used. Also, the manufacturing uncertainties as determined in the original analyses are used in this analysis. Because this analysis involves an infinite array of fresh fuel assemblies, the uncertainties associated with the rack-to-rack gap thickness and burnup are not applicable. 9.1A-13 Rev. 14 WOLF CREEK 9.1A.2.2.5.5 Dropped Fuel Assembly For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel in the rack of more than 19 inches, including the potential deformation under seismic or accident conditions. At this separation distance, the effect on reactivity is insignificant. Furthermore, the soluble boron in the pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident. It is possible for a fuel assembly to be mis-located adjacent to a storage rack in the northwest (area near the opening to the fuel transfer canal) and southeast (area near the opening to the cask loading pool) corners of the fuel storage pool. The worst case postulated accidents are: (1) in the southeast corner of the MZTR configuration, a fresh fuel assembly could be dropped and come to rest in the corner made up by a fresh assembly to the north and a Region 2 assembly to the west and (2) in the northwest comer of the checkerboard configuration, a fresh fuel assembly could be mis-located in a corner with fresh assemblies on two sides. The keff values for these two cases are very similar, and exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. These cases are less severe than the misplaced fresh fuel assembly accidents, and thus, are bounded by them. 9.1A.2.2.6 Benchmark Calculations The methodologies for determining criticality safety have been verified by comparison with critical experiment data for configurations that impose a stringent test of the capability of the analytical methodologies. These benchmark calculations have been made on selected critical experiment, chosen, in so far as possible to bound the range of variables in fuel storage rack designs, including the Wolf Creek high density racks. 9.1A.2.2.6.1 Summary Two independent methods of analysis were used in performing the Wolf Creek fuel storage rack criticality safety analyses. These two methods differ in cross section libraries and in the treatment of the cross sections. MCNP4a (Reference 17) is a continuous energy Monte Carlo code and KENO5a (Reference
- 18) uses group-dependent cross sections. For the KENO5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II (Reference 18) program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238-group library was chosen to avoid or minimize the errors (trends) that have been reported (e.g., References 19-21 ) for calculations with collapsed cross section sets. Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices. 9.1A-14 Rev. 14 WOLF CREEK In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the 10B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used).
Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses. Table 9.1A-7 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality. One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KEN05a computes and prints the "energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO5a, the number of fissions in each group may be collected and the EALF determined (post-processing). Figures 9.lA-7 and 9.1A-8 show the calculated keff for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (UO 2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental error in performing the critical experiments within each laboratory, as well as between the various testing laboratories. A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals. Linear regression analysis of the data in Figures 9.1A-7 and 9.1A-8 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from a keff of exactly 1.000) for the two methods of analysis are shown in the table below. Calculational Bias of MCNP4a and KENO5a MCNP4a 0.0009 0.0011 KENO5a 0.0030 0.0012 The bias and standard error of the bias were derived directly from the calculated keff values in Table 9.1A-7 using the following equations, with the standard error multiplied by the one-sided K-factor for 95% probability at the 95% confidence level from NBS Handbook 91 (Reference 34) (for the number of cases analyzed, the K-factor is ~2.05 or slightly more than 2). These equations may be found in any standard text on statistics, for example, Reference 22 (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO5a. 9.1A-15 Rev. 14 WOLF CREEK k n k i i n1 (1) k i i n i n kk n nn 2 2 1 1 2 1 1/() (2) Biask K k1. (3)where k i are the calculated reactivities of n critical experiments; k is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95% probability at the 95% confidence level (NBS Handbook 91 (Reference 34)). Formula (3) is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented in the Table above. The first portion of the equation, ( 1-k ), is the actual bias which is added to the MCNP4a and KENO5a results. The second term, k K is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively. The bias values are used to evaluate the maximum keff values for the rack designs. KENO5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data (References 19-21 ) would indicate for collapsed cross section sets in KENO5a (SCALE) calculations. 9.1A.2.2.6.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 9.1A-8 show the calculated keff values (Table 9.1A-7) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments. As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO5a for various enrichments. The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 9.1A-8 and Figure 9.1A-8, confirm no significant difference in the calculated values of keff for the two independent codes as evidenced by the 45 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias. 9.1A-16 Rev. 14 WOLF CREEK 9.1A.2.2.6.3 Effect of 10 B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but with one exception (a flux-trap experiment), the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed. Table 9.1A-9 lists the subset of experiments using thin neutron absorbers (from Table 9.1A-7) and shows the reactivity worth (k) of the absorber. The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (k) change in reactivity due to the absorber. No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 9.1A-9, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors. To further confirm the absence of a significant trend with 10 B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (as suggested in Reg. Guide 3.41). Results are shown in Figure 9.1A-9 and Table 9.1A-10 for a typical geometry. These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 45 line, within an expected 95% probability limit). 9.1A.2.2.6.4 Miscellaneous and Minor Parameters9.1A.2.2.6.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors. Analysis of these critical experiments are listed in Table 9.1A-11 (subset of data in Table 9.1A-7). There appears to be a small tendency toward over prediction of keff at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward over prediction at close spacing means that therack calculations may be slightly more conservative than otherwise. 9.1A.2.2.6.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 9.1A-7, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs. 9.1A-17 Rev. 14 WOLF CREEK 9.1A.2.2.6.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 9.1A-12. Analyses of the very high boron concentration experiments (>1300 ppm) show a tendency to slightly over predict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative. 9.1A.2.2.6.5 MOX Fuel The number of critical experiments with PuO 2 bearing fuel (MOX) is more limited than for UO 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 9.1A-13. Results of these analyses are generally above a keff of 1.00, indicating that when Pu is present, both MCNP4a and KENO5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KEN05a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in keff for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated keff over a wide range of the spectral index (energy of the average lethargy causing fission). 9.1A.3 THERMAL AND HYDRAULIC ANALYSES The Wolf Creek reracked fuel storage pool (spent fuel pool and cask loading pool with fuel storageracks installed) and the Fuel Pool Cooling and Cleanup System (FPCCS) comply with the provisions of Section III of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978). The methods, models, analyses, and numerical results are summarized below. The thermal-hydraulic qualification analyses for the rack arrays fall into the following categories:
- 1. Evaluation of the maximum decay heat load limit as a function of the bulk temperature limit for the postulated discharge scenario.
- 2. Evaluation of the postulated loss-of-forced cooling scenarios to establish that pool boiling will not occur.
- 3. Determination of the maximum temperature difference between the pool local temperature and the bulk pool temperature at the instant when the bulk temperature reaches its maximum value. 4. Evaluation of the maximum temperature difference between the fuel rod cladding temperature and the local pool water temperature to establish that nucleate boiling at any location around the fuel is not possible with forced cooling available. 9.1A-18 Rev. 14 WOLF CREEK Because the thermal and hydraulic analyses bound both the Callaway and Wolf Creek fuel storage pools, the pool water volume is conservatively based on the minimum east-west and north-south dimensions of the two pools. This conservatism results in a lower bound thermal inertia and outer periphery downcomer dimension in the thermal-hydraulic calculations.
FPCCS at Wolf Creek is described in Section 9.1.3. The fuel pool cooling system consists of two 100% capacity cooling trains for the removal of decay heat generated by irradiated fuel stored in the fuel storage pool. The decay heat generated by the stored fuel in the pool is transferred from the fuel pool cooling system through the fuel pool cooling heat exchangers. Normal makeup water to the fuel storage pool is supplied by the reactor makeup water system. An alternate source of makeup water is the RWST via the fuel pool cleanup pumps. Emergency makeup water is supplied from the Essential Service Water system. Boron addition to the fuel storage pool is normally accomplished by supplying borated water from the boric acid tanks via the boric acid blending tee. Boron may also be added by using the RWST as the source of makeup water to the fuel storage pool. Isolation of non-safety related portions of the FPCCS is a manual action. The fuel pool cleanup system provides the capability for purification of the water in the spent fuel pool, the cask loading pool, the transfer canal, the refueling pool, and the RWST. The cleanup system is an essential adjunct to the FPCCS system to maintain clarity and water chemistry control in the fuel storage pool. Consistent with the current plant practice, two discharge scenarios are postulated when considering fuel storage pool cooling: i. partial core offload ii. full-core offload In lieu of prescribing a batch size and cooling period for the partial core offload, the maximum pool heat load is determined for a scenario of only one cooling train operating and a limit on the steady state bulk pool temperature of 140 F.Similarly, the full core offload scenario is required to be executed so that the maximum pool heat load will not allow for bulk pool boiling at the end of a postulated 2 hour loss of forced cooling transient which occurs immediately after the full core offload. More specifically, the bulk water temperature is sought to be limited to 207F (which includes 5F of margin) after two hours of pool heat-up in the absence of all forced cooling paths. Evaluation of these two scenarios provides maximum flexibility in batch sizes and cooling periods prior to offload into the pool. In both scenarios, the component cooling water (CCW), used to remove heat from the fuel pool cooler, is assumed to be at its maximum design temperature. During the partial core offload scenario CCW flow is assumed to be at its nominal rate. During full core offload conditions, CCW flow is assumed to be at its design basis flow rate. With the thermal effectiveness of the fuel pool cooler thus fixed, the requirement of the ceiling on the bulk pool temperature essentially translates into a limit on the total heat generation rate in the pool. Finally an evaluation is performed for a loss of cooling accident occurring some time after restart. This evaluation considers a four hour long loss of forced cooling in the FPCCS followed by a twenty hour long period with cooling provided at one-half the normal coolant flow rate. Under this scenario, the fuel pool does not reach the bulk boiling temperature during the 24hour period. For this evaluation, the component cooling water to the heat exchanger is assumed to be at an elevated temperature and reduced flow rate. 9.1A-19 Rev. 14 WOLF CREEK 9.1A.3.1 Decay Heat Load Limit The heat load imposed on the pool is from the decay heat generated by fuel assemblies discharged into the pool. The primary safety function of the FPCCS is to adequately transport this heat load to the CCW system and thereby maintain the bulk pool temperature within specified limits. Compliance with the limiting heat load will be ensured through adjustments to the cooling system performance and/or adjustments to the fuel offload rate. Commonly used decay heat calculation methods based upon ASB 9-2, ANS 5. 1, or ORIGEN2 are used to provide conservative estimates of decay heat values for specific fuel pool inventories. 9.1A.3.1.1 Decay Heat Load Calculations and Conservatisms The following conservatisms are applied in the decay heat load limit calculations.FPCCS heat exchanger thermal performance is based on the design maximum fouling and plugging level. This will conservatively minimize the heat rejection capability of the FPCCS. Thermal inertia induced transient effects resulting in a lag in bulk pool temperature response are neglected. This conservatively lowers the calculated decay heat load limit by forcing the peak decay heat load to coincide with the peak pool temperature. In calculating the fuel storage pool evaporation heat losses, the building housing the fuel storage pool is assumed to have a conservative ambient air temperature of 110°F and 100% relative humidity. This minimizes the evaporative heat loss component, maximizing the heat duty burden on the pool cooling system. The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 9.1A-10. Referring to the fuel pool cooling system, the governing differential equation can be written by utilizing conservation of energy as: C dT d Q Q T Q THXEV()where: C = Pool thermal capacity, Btu/ F T = Pool bulk temperature, °F = Time after reactor shutdown, hr Q() = Time varying decay heat generation rate, Btu/hr Q HX(T) = Temperature dependent FPCCS heat rejection rate, Btu/hr Q EV (T) = Temperature dependent evaporative heat loss, Btu/hr Subject to the second of the conservatisms listed above, this differential relationship can be reduced to the following algebraic relationship: 0 = Qlimit - Q HX (Tlimit) - Q EV (Tlimit)where: Tlimit, is the maximum bulk pool temperature limit, °F Qlimit is the decay heat load limit, Btu/hr 9.1A-20 Rev. 14 WOLF CREEK Q HX(T) is a function of the bulk pool temperature and the coolant water flow rate and temperature, and can be written in terms of the temperature effectiveness (p) as follows: HXtti Q T WC pT twhere: W t = Coolant water flow rate, lb/hr C t = Coolant water specific heat capacity, Btu/(lb x °F) p = FPCCS heat exchanger temperature effectiveness T = Bulk pool water temperature, °F t i= Coolant water inlet temperature, °F The temperature effectiveness, a measure of the heat transfer efficiency of the FPCCS heat exchangers, is defined as: p tt T t i i0 where 0 t is the coolant outlet temperature (°F) and all other terms are as defined above. Q EV(T) is a nonlinear function of the pool temperature and ambient temperature. Q EV contains the heat evaporation losses from the pool surface, natural convection and thermal radiation from the pool surface, and heat conduction through the pool walls and slab. Experiments show that the heat conduction takes only about 4% of the total heat loss (Reference 35). The evaporation heat loss and natural convection heat loss can be expressed as Reference 36: EVaawa QThAT t A Tt A PP44 where: h = Natural convection heat transfer coefficient, Btu/(hr x ft 2 x °F) A = Pool surface area, ft 2 t a = Ambient pool building temperature, °F = Emissivity of pool water = Stephan-Boltzmann constant = Evaporation rate constant, Btu/(hr x ft 2 x psi) P w = Vapor pressure of water at pool temperature, psi P a = Vapor pressure of water at ambient temperature, psi The algebraic heat balance equation is solved for the decay heat load limit by rearranging the equation given above and substituting the maximum temperature limit for pool water temperature (T). The major input values for this analysis are summarized in Table 9.1A-14. 9.1A-21 Rev. 14 WOLF CREEK 9.1A.3.2 Margin Against Boiling To ensure that the pool bulk temperature will remain less than 207 F (i.e., adequate margin against boiling) compliance is required under the following conditions: (1) all forced cooling paths are lost following a full core offload and cooling is not restored for two hours, and (2) a loss of coolant accident occurs after restart and partial cooling is restored after 4 hours. The FPCCS system has two independent trains, both of which are seismically qualified and safety-related, so a complete loss of forced cooling is not possible under single failure criteria. Regardless of this fact, these evaluations are performed for postulated non-mechanistic loss of forced cooling accidents. 9.1A.3.2.1 Heat-up Calculations and Conservatisms The following conservatisms are applied in the heat-up calculations. The decay heat load and bulk pool temperature are assumed to be the calculated decay heat load limit and corresponding maximum pool temperature limit. Maximizing the initial temperature and the decay heat load conservatively minimizes the time-to-boil. The LOCA scenario, with its four hour loss of cooling to the FPCCS, is based on the decay heat load limit and corresponding peak temperature limit of the previously evaluated partial core discharge. These conditions would occur during an offload, and would therefore bound a post-restart condition. The transient reduction in decay heat over time is conservatively neglected. This maximizes the decay heat load at all points in time and minimizes the time-to-boil. Calculations verify that sufficient makeup water exists to prevent the pool water level from dropping, but no credit is taken for the reduced temperature of the makeup water.This assumes that makeup water is provided at the bulk pool temperature, conservatively minimizing the time-to-boil.In calculating the fuel storage pool evaporation heat losses, the building housing the fuel storage pool is assumed to have a conservative ambient air temperature of 110°F and 100% relative humidity. This conservatively minimizes the credit for evaporative heat loss. The temperature rise of the water in the pool over any period of time is a direct function of the average net decay heat load during that period. Therefore, maximizing the decay heat load will maximize the pool temperature increase rate and minimize the corresponding time-to-boil. As a transient decay heat load would necessitate a reduced average net heat load, the steady-state assumptions are conservative. 9.1A-22 Rev. 14 WOLF CREEK The governing enthalpy balance equation for this condition, subject to these conservative assumptions, is written as: C dT d QQ TitEVlim where is the time after cooling is lost (hr) and all other terms are the same as defined in Section 9.1A.3.1. This differential equation is solved using a numerical solution technique to obtain the bulk pool temperature as a function of time. The major input values for the analysis are summarized in Table 9.1A-15. 9.1A.3.2.2 Time-to-Boil When the FPCCS forced pool cooling becomes unavailable, the pool water will begin to rise in temperature and eventually will reach the normal bulk boiling temperature of 212F. In order to maintain some margin to this boiling condition, the analyses are performed with the acceptance criterion of a bulk pool temperature that is 207 F. The time to reach the boiling point is the shortest when the loss of forced cooling occurs at the point in time when the pool bulk temperature is at its maximum calculated value. Although the probability of the loss-of-cooling event coinciding at the instant when the pool water has reached its peak value is extremely remote, the calculations were performed under this extremely unlikely scenario. Analysis shows that, for postulated full-core discharge, and a maximum bulk temperature of 207F after two hours without cooling, the maximum allowable decay heat load is 63.41 MBtu/hr. The steady-state FSP temperature at this heat load would be 169.68 F.For the loss of coolant accident scenario, the bulk temperature after four hours without cooling would be 172.1 °F. Once partial cooling is reestablished, the steady-state temperature would be less than 175 F, thereby precluding the possibility of boiling even with continued reduced cooling capacity.9.1A.3.3 Local Pool Water Temperature A single conservative evaluation for a bounding amalgam of conditions was performed to evaluate the local pool water temperature. The result of the single evaluation is a bounding temperature difference between the maximum local water temperature and the bulk pool temperature. In order to determine an upper bound on the maximum local water temperature, a series of conservative assumptions are made. The most important of these assumptions are: 9.1A-23 Rev. 14 WOLF CREEK With a full core discharged into the racks farthest from the coolant water inlet, the remaining cells in the spent fuel pool are postulated to be occupied with previously discharged fuel. The hottest assemblies, located together in the pool, are assumed to be located in "pedestal" cells of the racks. These cells have a reduced water entrance area, caused by the pedestal blocking the baseplate hole, and a correspondingly increased hydraulic resistance. The coolant water inlet temperature, and therefore the bulk pool temperature, is minimized to conservatively maximize the fluid viscosity. This assumption will maximize the head losses for water flowing through the fuel racks and fuel assemblies. No downcomer flow is assumed to exist between the rack modules. All rack cells are conservatively assumed to be 50% blocked at the cell outlet to account for drop accidents resulting in damage to the upper end of the cells. This blocked cell portion is conservative, since structural evaluations have shown that only about 20% of the cell is blocked subsequent to the impact of dropped objects. Westinghouse 17x17 STD assembly, which is most resistive to axial fluid flow, is assumed to populate the entire storage region. Thus, the hydraulic resistance to heat transfer is maximized. 9.1 A.3.3. 1 Local Temperature Evaluation Methodology The inlet piping which returns cooled pool water from the FPCCS terminates above the level ofthe fuel racks. To demonstrate adequate cooling of hot fuel in the pool, it is necessary to rigorously quantify the velocity field in the pool created by the interaction of buoyancy driven flows and water injection/egress. A Computational Fluid Dynamics (CFD) analysis for this demonstration is required. The objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for all postulated fuel discharge/cooling alignment scenarios. The local thermal-hydraulic analysis is performed such that partial cell blockage and slight fuel assembly variations are bounded. An outline of the CFD approach is described in the following. 9.1A-24 Rev. 14 WOLF CREEK There are several significant geometric and thermal-hydraulic features of the fuel storage pool which must be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the spent fuel pool/cask loading pool region where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the fuel storage racks located near the bottom of the fuel storage pool. In this region, water flow is directed vertically upwards due to buoyancy forces through relatively small flow channels formed by the Westinghouse 17x17 fuel assembly rod arrays in each rack cell. This situation shall be modeled as a porous solid region in which fluid flow is governed by the classical Darcy's Law: P X Ki V CV V i i i2 where P X i is the pressure gradient, Ki, i V and C are the corresponding permeability, velocity and inertial resistance parameters and is the fluid viscosity. The permeability and inertial resistance parameters for the rack cells loaded with Westinghouse 17x 17 fuel were determined based on the friction factor correlations for the laminar flow conditions typically encountered due to the low buoyancy induced velocities and the small size of the flow channels. The fuel storage pool geometry required an adequate portrayal of large scale and small scale features, spatially distributed heat sources in the fuel storage racks and water inlet/outlet configuration. Relatively cooler bulk pool water normally flows down between the fuel rack outline and pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow turns from a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water.An adequate modeling of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to capture the bulk and local features of the flow field. The distributed heat sources in the fuel storage pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, bounding peak effects, and presence of background decay heat from old discharges. Three heat generating zones were modeled. The first consists of background fuel from previous discharges, the remaining two zones consist of fuel from a bounding full-core-discharge scenario. The two full core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat generation. The background decay heat load is determined such that the total decay heat load in the pool is equal to the calculated decay heat load limit. This is a conservative model, since all of the fuel with higher than average decay heat is placed in a contiguous area. A uniformly distributed heat generation rate was applied throughout each distinct zone. 9.1A-25 Rev. 14 WOLF CREEK The CFD analysis was performed on the industry standard FLUENT (Reference 40) fluid flow and heat transfer modeling program. The FLUENT code enabled buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds's Stresses" to the mean bulk flow quantities with the following turbulence modeling options: (i) k- Model (ii) RNG k- Model (iii) Reynolds Stress Model The k- Model is considered most appropriate for the twin site CFD analysis. The k- turbulence model is a time-tested, general purpose turbulence model. This model has been demonstrated to give good results for the majority of turbulent fluid flow phenomena. The Renormalization Group (RNG) and Reynolds Stress models are more advanced models that were developed for situations where the k- Model does not provide acceptable results, such as high viscosity flow and supersonic shock. The flow regime in the bulk fluid region is such that the k- Model will provide acceptable results. Rigorous modeling of fluid flow problems requires a solution to the classical Navier-Stokes equations of fluid motion (Reference 37). The governing equations (in modified form for turbulent flows with buoyancy effects included) are written as: 0 0 0 0 0 i ij ij i j j ii i ij j u t uu xx u x u x p x TT g uu x''where i u are the three time-averaged velocity components. < u i' u j'> are time-averaged Reynolds stresses derived from the turbulence induced fluctuating velocity components, i u', o is the fluid density at temperature, T o , is the coefficient of thermal expansion, is the fluid viscosity, i g are the components of gravitational acceleration and j x are the Cartesian coordinate directions. The Reynolds stress tensor is expressed in terms of the mean flow quantities by defining a turbulent viscosity t and a turbulent velocity scale k1/2 as shown below (Reference 38): ij ij t i j j iuuku u x u x/23 The procedure to obtain the turbulent viscosity and velocity length scales involves a solution of two additional transport equations for kinetic energy (k) and rate of energy dissipation (). This methodology is known as the k- model for turbulent flows as described by Launder and Spalding (Reference 39). Some of the major input values for this analysis are summarized in Table 9.1A-
- 16. An isometric view of the assembled CFD model is presented in Figure 9.1 A-
- 11. 9.1A-26 Rev. 14 WOLF CREEK 9.1A.3.3.2 Local Water and Fuel Cladding Temperatures Consistent with the approach to make conservative assessments of temperature, the local water temperature calculations are performed for a pool with decay heat generation equal to the maximum calculated decay heat load limit. Thus, the local water temperature evaluation is a calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of a highly heat emissive fuel bundle).
The CFD study has analyzed a single bounding local thermal-hydraulic scenario. In this scenario, a bounding full-core discharge is considered in which the 193 assemblies are located in the pool, farthest from the cooled water inlet, while the balance of the rack cells are postulated to be occupied by fuel from old discharges. In this analysis, the difference between the peak local temperature and the coincident bulk pool temperature was conservatively calculated to be 64.6 F.The peak fuel cladding superheat is determined for the hottest cell location in the pool as obtained from the CFD model for the twin site pools. The maximum temperature difference between the fuel cladding and the local water (T c) is calculated to be less than 67.4F. Applying this calculated cladding T c , along with the maximum temperature difference between the local water temperature and the bulk pool temperature, to the bulk maximum normal operating pool temperature of 170F yields a conservatively bounding 234.6 F maximum local water temperature and a conservatively bounding 302F peak cladding temperature. The maximum local water temperature is lower than the 239 F local boiling temperature on top of the racks, thereby precluding nucleate boiling in the subchannel. The heat fluxes are too low to support a departure from nucleate boiling (DNB) condition. Thus, nucleate and departure from nucleate boiling do not occur anywhere within the Wolf Creek fuel storage pool. 9.1A.3.4 Fuel Rod Cladding Temperature The temperature of the fuel rod cladding is performed for a single, bounding scenario. The maximum fuel cladding superheat above the local water temperature is calculated. The maximum specific power of a fuel array A q can be given by: A q q xy Fwhere: xy F = Radial peaking factor q= Average fuel assembly specific power, Btu/hr 9.1A-27 Rev. 14 WOLF CREEK The peaking factors are given in Table 9.1A-16. The maximum temperature rise of pool water in the most disadvantageously placed fuel assembly, defined as the one which is subject to the highest local pool water temperature, was computed for all loading cases. Having determined the maximum local water temperature in the pool, it is now possible to determine the maximum fuel cladding temperature. A fuel rod can produce z F times the average heat emission rate over a small length, where z F is the axial rod peaking factor. The axial heat distribution in a rod is generally a maximum in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux over an infinitesimal area is given by the equation: cxyz c q q FF Awhere c A is the total cladding external heat transfer area in the active fuel length region. Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett (Reference 41) report a Nusselt-number based heat transfer correlation for laminar flow in a heated channel. The film temperature driving force (T f) at the peak cladding flux location is calculated as follows: f h h D Kw Nuf c f T q hwhere, f h is the water side film heat transfer coefficient, h D is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is Nusselt number for laminar flow heat transfer. In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit resistance R c (equal to 1.67x10 -4 ft 2-hr-F/Btu), which covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (T c) is given by: cfc cTTR q 9.1A-28 Rev. 14 WOLF CREEK 9.1A.3.5 Decay Heat Load Limits The calculated decay heat load limit is summarized in Table 9.1A-17. Because all transient effects were excluded from the evaluations, this decay heat load corresponds to the invariant heat load which results in a steady-state bulk pool temperature which will not exceed the temperature limit for either the partial core or full core offload scenario. This calculated decay heat load limit is not based on any specific discharge conditions, but is a mathematically derived quantity. Any conservative decay heat calculation used to determine the operational limits (i.e. in-core hold time requirement) necessary to avoid exceeding this decay heat load provides conservative operational limits. The operational limits are determined based on the decay heat load limit in Table 9.1A-17. Based on this limit, the fuel storage pool cooling system will remain in compliance. 9.1A.4 STRUCTURAL AND SEISMIC CONSIDERATIONS The structural adequacy of the high density spent fuel racks are considered under all loadings postulated for normal, seismic, and accident conditions. The fuel storage racks must remain fully functional during and after a seismic disturbance. The seismic adequacy is demonstrated in response to both a Safe Shutdown Earthquake (SSE) and the Operational Design Basis Earthquake (OBE). The analyses undertaken to confirm the structural integrity of the racks are performed in compliance with the USNRC Standard Review Plan (Reference 42) and the OT Position Paper (Reference 43). The response of a free-standing rack module to seismic inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in impacts and friction effects. Some unique attributes of rack dynamic behavior include a large fraction of the total structural mass in a confined rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced adjacent structures. Whole Pool Multi-Rack (WPMR) analysis simulates the dynamic behavior of the storage rack structures. The walls separating the Spent Fuel Pool and the Cask Loading Pool allow rack configurations to be dynamically analyzed as two separate WPMR models. 9.1A.4. 1 Analysis Methodology An accurate simulation is obtained by direct integration of the nonlinear equations of motion with three pool slab acceleration time-histories applied as the forcing functions acting simultaneously. Reliable assessment of the stress field and kinematic behavior of the rack modules incorporates key attributes of the actual structure in a conservative dynamic model. The model must have the capability to execute the concurrent motion forms compatible with the free-standing installation of the modules. Calculations must incorporate momentum transfers due to the rattling of fuel assemblies inside storage cells; the lift-off and subsequent impact of support pedestals with the pool liner (or bearing pad); and quantification of fluid coupling due to water mass in the interstitial spaces around rack modules. In short, there are a large number of parameters with potential influence on the rack kinematics. The comprehensive structural evaluation must deal with all of these without sacrificing conservatism. 9.1A-29 Rev. 14 WOLF CREEK The model must be capable of effecting momentum transfers which occur due to rattling of fuel assemblies inside storage cells and the capability to simulate lift-off and subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be modeled in an accurate manner. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool. The 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the following array of variables: Interface Coefficient of Friction Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data which have been found to be bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and O.5, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piece-wise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner. Rack Beam Behavior Rack elasticity, relative to the rack base, is included in the model by introducing linear springs to represent the elastic bending action, twisting, and extensions. Impact Phenomena Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements. The term "nonlinear spring" is a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement. Fuel Loading Scenarios The fuel assemblies are conservatively assumed to rattle in unison which obviously exaggerates the contribution of impact against the cell wall. Fluid Coupling The computer code DYNARACK (Reference 47) handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and in improving predictions of rack kinematics. 9.1A-30 Rev. 14 WOLF CREEK For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. In WPMR analysis, all rack modules are modeled simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis methodology was used. 9.1A.4.1.1 Fuel Weights For the dynamic rack simulations, the dry fuel weight is conservatively taken to be 1647 lb. This is a higher fuel weight value to account for rod control cluster assemblies (RCCAs) being stored along with fuel assemblies. Therefore, the analyses conservatively consider an RCCA to be stored along with an assembly at every location. 9.1A.4.1.2 Synthetic Time-Histories The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP 3.7.1 (Reference 48). A preferred criterion for the synthetic time-histories in SRP
3.7.1 calls
for both the response spectrum and the power spectral density corresponding to the generated acceleration time-history to envelope their target (design basis) counterparts with only finite enveloping infractions. The time-histories for the pools have been generated to satisfy this preferred (and more rigorous) criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3.7.1. Figures 9.1A-12 through 9.1A-16 provide plots of the time-history accelerograms which were generated over a 25 second duration for OBE and SSE events, respectively. These artificial time-histories are used in all non-linear dynamic simulations of the racks. Results of the correlation function of the three time-histories are given in Table 9.1A-19. Absolute values of the correlation coefficients are shown to be
less than 0.15, indicating the statistical independence of the three data sets. 9.1A.4.2 WPMR Methodology The WPMR methodology incorporates both stress and displacement criteria. The following summarizes the sequence steps undertaken for model development:
- a. Suitable 3-D dynamic models for a time-history analysis of the new maximum density racks are prepared. These models include the assemblage of all rack modules in each pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Multi-Rack model. 9.1A-31 Rev. 19 WOLF CREEK
- b. 3-D dynamic analyses are performed on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies). Appropriate displacement and load outputs from the dynamic model for post-processing are archived.
- c. A stress analysis of high stress areas for the limiting case of all the rack dynamic analysis is performed to demonstrate compliance with ASME Code Section III, Subsection NF limits on stress and displacement.9.1A.4.2.1 Model Assumptions The dynamic modeling of the rack structure considers all nonlinearities and parametric variations. The following assumptions are used in the Whole Pool Multi-Rack analysis of racks:
- a. The fuel rack structure motion is captured by modeling the rack as a 12 degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. In this manner, the response of the module, relative to the baseplate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.
- b. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H,0.75H,0.5H,0.25H, and at the rack base (H is the rack height measured above the baseplate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.
- c. Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and, therefore, yields conservative results.
- d. Fluid coupling between rack and fuel assemblies, and between rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy. These effects uses the methods (References 51 and 52) for rack/assembly coupling and for rack-to-rack coupling. The fluid coupling effect in its simplest form considers the proximate motion of two bodies under water, where one body vibrates adjacent to a second body, and both bodies are submerged in frictionless fluid. During a seismic event all racks in the pool are subject to the input excitation simultaneously. The WPMR model simulates 3-D motion of all rack modules simultaneously and encompasses interaction between every set of racks in the pool, (i.e., the motion of one rack produces fluid forces on all other racks and on the pool walls). 9.1A-32 Rev. 14 WOLF CREEK
- e. Fluid damping and form drag are conservatively neglected.
- f. Sloshing is found to be negligible at the top of the rack and is, therefore, neglected in the analysis of the rack.
- g. Potential impacts between the cell walls of the new racks and the contained fuel assemblies are accounted for by appropriate compression only gap elements between masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two horizontal directions. Bottom gap elements are located at the baseplate elevation. The initial gaps reflect the presence of baseplate extensions, and the rack stiffnesses are chosen to simulate local structural detail.
- h. Pedestals are modeled by gap elements in the vertical direction and as "rigid links" for transferring horizontal stress. Each pedestal support is linked to the pool liner (or bearing pad) by two friction springs. The spring rate for the friction springs includes any lateral elasticity of the stub pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack elasticity just above the pedestal.
- i. Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap in order to provide a conservative measure of fluid resistance to gap closure.
- j. The model for the rack is considered supported, at the base level, on four pedestals modeled as nonlinear compression only gap spring elements and eight piece-wise linear friction spring elements; these elements are properly located with respect to the centerline of the rack beam, and allow for arbitrary rocking and sliding motions. 9.1A.4.2.2 Stiffness Elements Three element types are used in the rack models. Type 1 are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linear friction springs used to develop the appropriate forces between the rack pedestals and the supporting bearing pads. Type 3 elements are non-linear gap elements which model gap closures and subsequent impact loadings (i.e., between fuel assemblies and the storage cell inner walls, and rack outer periphery spaces. 9.1A.4.2.3 Coefficients of Friction Multiple simulations were performed to adjust the friction coefficient ascribed to the support pedestal/pool bearing pad interface. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data (Reference 50). Simulations are also performed by imposing intermediate value friction coefficients developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the entire WPMR simulation in order to obtain reproducible results, closer to realistic structural conditions. 9.1A-33 Rev. 14 WOLF CREEK 9.1A.4.2.4 Governing Equations of Motion Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation (Reference 53). The system kinetic energy includes contributions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix form: M d q dt Q G 2 2where:[M] = total mass matrix (including structural and fluid mass contributions).
The size of this matrix will be 22n x22n for a WPMR analysis (n = number of racks in the model). q = the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)[G]= a vector dependent on the given ground acceleration [Q] = a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom 2 2 11 d q dtMQMGThis equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the computer program DYNARACK (Reference 47). 9.1A.4.3 Structural Evaluation of the Fuel Rack Design There are two sets of criteria to be satisfied by the rack modules: a. Kinematic Criteria Per Reference (Reference 42), in order to be qualified as a physically stable structure it is necessary to demonstrate that an isolated rack in water would not overturn when an event of magnitude:1.5 times the upset seismic loading condition is applied. 1.1 times the faulted seismic loading condition is applied. b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations provided herein. 9.1A-34 Rev. 14 WOLF CREEK 9.1A.4.3.1 Stress Limit EvaluationsThe stress limits that apply to the rack structure are derived from the ASME Code, Section III, Subsection NF (Reference 55). Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code Appendices (Reference 56), and are listed in Table 9.1A-18. For convenience, the stress results are presented in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting value. The limiting value of each stress factor is 1.0, based on the allowable strengths for each level, for Levels A, B, and D. Stress factors reported are: R 1 = Ratio of direct tensile or compressive stress on a net section to its allowable value(note pedestals only resist compression) R 2 = Ratio of gross shear on a net section in the x-direction to its allowable value R 3 = Ratio of maximum x-axis bending stress to its allowable value for the section R 4 = Ratio of maximum y-axis bending stress to its allowable value for the section R 5 = Combined flexure and compressive factor (as defined in the foregoing) R 6 = Combined flexure and tension (or compression) factor (as defined in the foregoing) R 7 = Ratio of gross shear on a net section in the y-direction to its allowable value 9.1A.4.3.2 Loads and Loading Combinations for Fuel Storage Racks The applicable loads and their combinations which must be considered in the seismic analysis of rack modules is excerpted from Refs. (Reference 43) and (Reference 57). The load combinationsconsidered are identified below: Loading Combination Service Level D + L D + L + T o D + L + T o + E Level A D + L + T a+ E D + L + T o+ P f Level B D + L + T a + E' D + L + T o +Level D The functional capability of the fuel racks must be demonstrated. 9.1A-35 Rev. 14 WOLF CREEK Where: D = Dead weight-induced loads (including fuel assembly weight) L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path) P f = Upward force on the racks caused by postulated stuck fuel assembly F d = Impact force from accidental drop of the heaviest load from the maximum possible height. E = 0perating Basis Earthquake (OBE) E' = Safe Shutdown Earthquake (SSE) T o = Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition) T a = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions) T a and T o produce local thermal stresses. The worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.9.1A.4.3.3 Parametric Simulations The table below presents a complete listing of the parametric simulations performed. Consideration of the parameters described above resulted in the following 19 runs. Run Pool COF Event 1 SFP 0.8 SSE 2 SFP 0.2 SSE 3 SFP Random SSE 4 SFP 0.8 OBE 5 SFP 0.2 OBE 6 SFP Random OBE 7 Cask Loading Pool 0.8 SSE 8 Cask Loading Pool 0.2 SSE 9 Cask Loading Pool Random SSE 10 Cask Loading Pool 0.8 OBE 11 Cask Loading Pool 0.2 OBE 12 Cask Loading Pool Random OBE 13 SFP (half full) 0.8 SSE 14 SFP (half full) 0.2 SSE 15 SFP (half full) Random SSE 16 Single Rack - Overturning Check 0.8 OBE x 1.5 17 Single Rack - Overturning Check 0.2 SSE x 1.1 9.1A-36 Rev. 14 WOLF CREEK 9.1A.4.3.4 Time History Simulation Results The results from the DYNARACK runs are presented by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stress factors. 9.1A.4.3.4.1 Rack Displacements Selected rack to wall and rack to rack gaps were evaluated over the entire duration of the 0.8 COF SSE simulation (run 1) for three selected locations around the perimeter of SFP rack no. 14. This simulation produced the largest displacement (0.677") of any rack in the SFP. Rack 2 in the Cask Loading Pool (run 9) experiences a larger displacement of (1.274"). However, in the Cask Loading Pool the rack to wall gaps are larger. Therefore, SFP Rack 14 is chosen for displacement plotting because the displacements are of greater significance for the SFP simulations. Rack to rack impacts may be identified when rack gaps are momentarily reduced to a value of zero or less. Rack to wall impacts did not occur under any of the simulations. A tabulated summary of the maximum displacement for each simulation is provided below with the location/direction terms defined as follows: uxt, uyt = displacement of top corner of rack, relative to the slab, in the North-South and East-West directions, respectively. The maximum displacements for every simulation, including the single rack tipover analyses, occurred at the top of the racks shown in the last table column.Simulations 16 and 17 were performed to evaluate the potential for overturning of a rack to account for the unlikely possibility of a seismic event occurring during the installation process. The heaviest racks with the narrowest pedestal stance are chosen for these simulations, since these racks are expected to produce the greatest displacements during seismic events. All of these simulations were performed with half loaded racks to further increase displacements. The largest displacement is less than 0.5 inches and is not a tipover concern. 9.1A-37 Rev. 14 WOLF CREEK The following maximum rack displacements (in inches) are obtained for each of the runs: Pool Event Ru n COF Maximum Displacement (inches)Location/Direction Rac k Spent Fuel Pool SSE 1 0.8 0.677 uxt 14 SSE 2 0.2 0.428 uyt 7 SSE 3 Random 0.642 uxt 15 OBE 4 0.8 0.341 uyt 1 OBE 5 0.2 0.280 uyt 1 OBE 6 Random 0.343 uyt 1 Cask Loading Pool SSE 7 0.8 1.274 uyt 2 SSE 8 0.2 0.720 uxt 1 SSE 9 Random 0.965 uyt 3 OBE 10 0.8 0.275 uxt 1 OBE 11 0.2 0.275 uxt 1 OBE 12 Random 0.275 uxt 1 Half full SFP SSE 13 0.8 0.3921 uxt 6 SSE 14 0.2 0.562 uyt 8 SSE 15 Random 0.578 uyt 6 Tipover OBE 16 0.8 0.288--Tipover SSE 17 0.8 0.338--Note: All of the maximum displacements occurred at the tops of the storage racks, as expected from swaying, bending, and tipping behavior. 9.1A-38 Rev. 14 WOLF CREEK 9.1A.4.3.4.2 Pedestal Vertical Forces Pedestal number 1 for each rack is located in the northeast corner of the rack. Numbering increases counterclockwise around the periphery of each rack. The following bounding vertical pedestal forces (in kips) are obtained for each run: Pool Event Run COF Maximum Pedestal Load (kips)Rac k Ped.Spent Fuel Pool SSE 1 0.8 291 12 3 SSE 2 0.2 235 3 2 SSE 3 Random 267 1 4 OBE 4 0.8 203 1 4 OBE 5 0.2 188 1 2 OBE 6 Random 204 1 4 Cask Loading Pool SSE 7 0.8 197 3 3 SSE 8 0.2 162 1 4 SSE 9 Random 159 3 4 OBE 10 0.8 103 1 1 OBE 11 0.2 103 1 2 OBE 12 Random 103 1 2 Half full SFP SSE 13 0.8 211 6 4 SSE 14 0.2 220 4 3 SSE 15 Random 255 3 2 The highest pedestal load of 291,000 lbs occurs in run 1. 9.1A-39 Rev. 14 WOLF CREEK 9.1A.4.3.4.3 Pedestal Friction Forces The maximum (x or y direction) shear load (in kips) bounding all pedestals in the simulation are reported below and are obtained by inspection of the complete tabular data. Pool Event Run COF Maximum Friction Load (kips)Spent Fuel Pool SSE 1 0.8 103.0 SSE 2 0.2 46.8 SSE 3 Random 99.3 OBE 4 0.8 39.7 OBE 5 0.2 31.4 OBE 6 Random 41.4 Cask Loading Pool SSE 7 0.8 58.9 SSE 8 0.2 30.7 SSE 9 Random 57.5 OBE 10 0.8 15.1 OBE 11 0.2 15.7 OBE 12 Random 15.4 Half full SFP SSE 13 0.8 61.0 SSE 14 0.2 39.6 SSE 15 Random 84.7 9.1A-40 Rev. 14 WOLF CREEK 9.1A.4.3.4.4 Rack Impact Loads A freestanding rack, by definition, is a structure subject to potential impacts during a seismic event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and, in some instances, from localized impacts between the racks. The following instantaneous maximum impact forces and locations are identified for each of the simulations performed. Listings are only given for those simulations within which impact occurred. It may be noted that all impact loads occurred at the bottom of the racks where the gap was modeled as only 1/8 inch. No impacts occurred for the 0.8 COF condition, since under higher friction the relative rack displacement at the base plate level was reduced to less than the 1/8" gap. The element numbering is identified in Figures 9.1A-17 through 9.1A-22. Pool Event Run COF Maximum Impact Load (kips)Element Spent Fuel Pool SSE 2 0.2 25.20 396 SSE 3 Random 41.03 408 Half full SFP SSE 14 0.2 47.9 2139.1A.4.3.4.5 Rack to Wall Impacts Storage racks do not impact the pool walls under any simulation. 9.1A-41 Rev. 14 WOLF CREEK 9.1A.4.3.4.6 Fuel to Cell Wall Impact Loads A review of all simulations performed allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site. The maximum fuel/cell wall impact load values are reported in the following table: Pool Event Run COF Maximum Impact Load (lb.)Rack Spent Fuel Pool SSE 1 0.8 641 12 SSE 2 0.2 625 10 SSE 3 Random 641 13 OBE 4 0.8 370 6 OBE 5 0.2 371 6 OBE 6 Random 371 6 Cask Loading Pool SSE 7 0.8 710 1 SSE 8 0.2 590 2 SSE 9 Random 659 2 OBE 10 0.8 403 1 OBE 11 0.2 378 3 OBE 12 Random 360 1 Half full SFP SSE 13 0.8 565 6 SSE 14 0.2 781 4 SSE 15 Random 840 7 Based on fuel manufacturer's data, loads of this magnitude will not damage the fuel assembly. 9.1A-42 Rev. 14 WOLF CREEK 9.1A.4.3.5 Rack Structural Evaluation9.1A.4.3.5.1 Rack Stress Factors With time history results available for pedestal normal and lateral interface forces, the limiting bending moment and shear force at the bottom casting-pedestal interface is available as a function of time. In particular, maximum values for the previously defined stress factors can be determined for every pedestal in the array of racks. With this information available, the structural integrity of the pedestal can be assessed and reported. The net section maximum (in time) bending moments and shear forces can also be determined at the bottom casting-rack cellular structure interface for each spent fuel rack in the pool. With this information in hand, the maximum stress in the limiting rack cell (box) can be evaluated. An evaluation of the stress factors for all of the simulations performed, leads to the conclusion that all stress factors, as defined in Section 9.1A.4.3. 1, are less than the mandated limit of 1.0 for the load cases examined. From all of the simulations reported in the tables, the bounding stress factors are summarized below. The maximum stress factor is always R 6 , defined in section 9.1A.4.3.1. Pool Event Run COF Maximum Stress Factor Rack Spent Fuel Pool SSE 1 0.8 0.389 12 SSE 2 0.2 0.264 3 SSE 3 Random 0.338 12 OBE 4 0.8 0.441 1 OBE 5 0.2 0.423 1 OBE 6 Random 0.442 1 Cask Loading Pool SSE 7 0.8 0.330 2 SSE 8 0.2 0.289 1 SSE 9 Random 0.309 2 OBE 10 0.8 0.378 1 OBE 11 0.2 0.377 1 OBE 12 Random 0.378 1 Half full SFP SSE 13 0.8 0.241 6 SSE 14 0.2 0.261 8 SSE 15 Random 0.289 8 The requirements of Section 9.1A.4.3 are satisfied for the load levels considered for every limiting location in every rack in the array. Stress factors for SSE are calculated based on SSE allowable strengths, while stress factors for OBE simulations are based on OBE allowable strengths. 9.1A-43 Rev. 14 WOLF CREEK 9.1A.4.3.5.2 Pedestal Thread Shear Stress The complete post-processor results give thread stresses under faulted conditions for every pedestal for every rack in the pool. The average shear stress in the engagement region is given below for the limiting pedestal in each simulation. Pool Event Run COF Maximum Thread Shear Stress (psi)Rack Spent Fuel Pool SSE 1 0.8 9268 12 SSE 2 0.2 7484 3 SSE 3 Random 8503 1 OBE 4 0.8 6465 1 OBE 5 0.2 5987 1 OBE 6 Random 6497 1 Cask Loading Pool SSE 7 0.8 6274 3 SSE 8 0.2 5159 1 SSE 9 Random 5064 3 OBE 10 0.8 3280 1 OBE 11 0.2 3280 1 OBE 12 Random 3280 1 Half full SFP SSE 13 0.8 6720 6 SSE 14 0.2 7006 4 SSE 15 Random 8121 3 The ultimate strength of the female part of the pedestal is 66,200 psi. The yield stress for this material is 21,300 psi. The allowable shear stress for Level B conditions is 0.4 times the yield stress which gives 8,520 psi and is much larger than the maximum calculated stress value of 6,497 psi for the OBE simulations. The allowable shear stress for Level D conditions is the lesser of: 0.72 S y = 15,336 psi or 0.42 S u = 27,804 psi. Therefore, the former criteria controls and the allowable is much larger than the maximum calculated stress value of 9,268 psi for the SSE condition. Therefore, thread shear stresses are acceptable under all conditions. 9.1A-44 Rev. 14 WOLF CREEK 9.1A.4.3.5.3 Local Stresses Due to Impacts Impact loads at the pedestal base produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. However, impact loads on the cellular region of the racks produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses. Even though limits on secondary stresses are not prescribed in the Code for class 3 NF structures, evaluations must be made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the subcriticality of the stored fuel array. a. Impact Loading Between Fuel Assembly and Cell Wall Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analysis is used to obtain the limiting impact load which would lead to gross permanent deformation. Fuel impacts are demonstrated not to represent a significant concern with respect to fuel rack cell deformation. b. Impacts Between Adjacent Racks The bottom of the storage racks will impact each other at a few locations during seismic events. Since the loading is presented edge-on to the 3/4" baseplate membrane, the distributed stresses after local deformation will be negligible. The impact loading will be distributed over a significant portion of the entire baseplate length. The resulting compressive stress from the highest impact load is negligible. This is a conservative computation, since the simulation assumes a local impact site. Therefore, any deformation will not effect the configuration of the stored fuel. 9.1A.4.3.5.4 Assessment of Rack Fatigue Margin Deeply submerged high density spent fuel storage racks arrayed in close proximity to each other in a free-standing configuration behave primarily as a nonlinear cantilevered structure when subjected to 3-D seismic excitations. In addition to the pulsations in the vertical load at each pedestal, lateral friction forces at the pedestal/bearing pad-liner interface, which help prevent or mitigate lateral sliding of the rack, also exert a time-varying moment in the baseplate region of the rack. The friction-induced lateral forces act simultaneously in x and y directions with the requirement that their vectorial sum does not exceed V, where is the limiting interface coefficient of friction and V is the concomitant vertical thrust on the liner (at the given time instant). As the vertical thrust at a pedestal location changes, so does the maximum friction force, F, that the interface can exert. In summary, the horizontal friction force at the pedestal/liner interface is a function of time; its magnitude and direction of action varies during the earthquake event. 9.1A-45 Rev. 14 WOLF CREEK The time-varying lateral (horizontal) and vertical forces on the extremities of the support pedestals produce stresses at the root of the pedestals in the manner of an end-loaded cantilever. The stress field in the cellular region of the rack is quite complex, with its maximum values located in the region closest to the pedestal. The maximum magnitude of the stresses depends on the severity of the pedestal end loads and on the geometry of the pedestal/rack baseplate region. Alternating stresses in metals produce metal fatigue if the amplitude of the stress cycles is sufficiently large. In high density racks designed for sites with moderate to high postulated seismic action, the stress intensity amplitudes frequently reach values above the material endurance limit, leading to expenditure of the fatigue "usage" reserve in the material. Because the locations of maximum stress (viz., the pedestal/rack baseplate junction) and the close placement of racks, a post-earthquake inspection of the high stressed regions in the racks is not feasible. Therefore, the racks are engineered to withstand multiple earthquakes without reliance of nondestructive inspections for post-earthquake integrity assessment. The fatigue life evaluation of racks is an integral aspect of a sound design. A time-history analysis was performed to provide the means to obtain a complete cycle history of the stress intensities in the highly stressed regions of the rack.To evaluate the cumulative damage factor, a finite element model of a portion of the spent fuel rack in the vicinity of a support pedestal is constructed in sufficient detail to provide an accurate assessment of stress intensities. The finite element solutions for unit pedestal loads in three orthogonal directions are combined to establish the maximum value of stress intensity as a function of the three unit pedestal loads. Using the archived results of the spent fuel rack dynamic analyses (pedestal load histories versus time), enables a time-history of stress intensity to be established at the most limiting location. This permits establishing a set of alternating stress intensity ranges versus cycles for an SSE and an OBE event. Following ASME Code guidelines for computing U, it is found that U = 0.404 due to the combined effect of one SSE and twenty OBE events. This is well below the ASME Code limit of 1.0. 9.1A.4.3.5.5 Weld Stresses Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at cell-to-cell connections. Bounding values of resultant loads are used to qualify the connections. a. Baseplate-to-Rack Cell Welds For Level A or B conditions, Reference 55 permits an allowable weld stress of = 0.3 S u. The allowable value may be increased for Level D by the ratio 1.8. The highest predicted weld stress for OBE is calculated from the highest R 6 value. The highest predicted weld stress is less than the allowable weld stress value. The highest predicted weld stress for SSE is less than the allowable weld stress value as shown in Table 9.1A-20. Therefore, all weld stresses between the baseplate and cell wall base are acceptable. 9.1A-46 Rev. 14 WOLF CREEK b. Baseplate-to-Pedestal Welds The weld between the baseplate and the support pedestal is checked using finite element analysis to determine the maximum stress under a Level B or Level D event. The calculated stress values are below the allowable values.c. Cell-to-Cell Welds Cell-to-cell connections are by a series of connecting welds along the cell height. Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall. These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld. Table 9.1A-20 gives results for the maximum allowable load that can be transferred by these welds based on the available weld area. An upper bound on the load required to be transferred is also given in Table 9.1A-20 and is much lower than the allowable load. This upper bound value is very conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, and multiplying the result by 2 to account for the simultaneous impact of two assemblies. An equilibrium analysis at the connection then yields the upper bound load to be transferred. It is seen from the results in Table 9.1A-20 that the calculated load is well below the allowable. 9.1A.4.3.5.6 Bearing Pad Analysis To protect the pool slab from high localized dynamic loadings, bearing pads are placed between the pedestal base and the slab. Fuel rack pedestals impact on these bearing pads during a seismic event and pedestal loading is transferred to the liner. Bearing pad dimensions are set to ensure that the average pressure on the slab surface due to a static load plus a dynamic impact load does not exceed the American Concrete Institute, ACI-349 (Reference 58) limit on bearing pressures. The maximum vertical pedestal load is 291,000 Lb. (SSE event). The maximum allowable concrete bearing pressure is 2,380 psi. Calculations show that the average pressure at the slab/liner interface is well below the allowable value of 2,380 psi. The stress distribution in the bearing pad is also evaluated. The maximum bending stress in the bearing pad under the peak vertical load is acceptable. Therefore, the bearing pad design devised for the Wolf Creek is appropriate for the prescribed loadings. 9.1A.4.3.5.7 Level A Evaluation The Level A condition is not a governing condition for spent fuel racks since the general level of loading is far less than Level B loading. To illustrate this, the heaviest spent fuel rack is considered under the dead weight load. It is shown below that the maximum pedestal load is low and that further stress evaluations are unnecessary. 9.1A-47 Rev. 14 WOLF CREEK LEVEL A MAXlMUM PEDESTAL LOAD Dry Weight of Largest Holtec Rack (Bl is 13x13 cells) = 25,970 Lbf. Dry Weight of 169 Fuel Assemblies = 278,343 Lbf. Total Dry Weight = 304,313 Lbf. Total Buoyant Weight (0.87 x Total Dry Weight) = 264,752 Lbf. Load per Pedestal = 66,188 Lbf. The stress allowables for the normal condition is the same as for the upset condition, which resulted in a maximum pedestal load of 204,000 Lb. Since this load (and the corresponding stress throughout the rack members) is much greater than the 66,188 lb. load calculated above, the upset (OBE) condition controls over normal (Gravity) condition. Therefore, no further evaluation is performed.9.1A.4.3.5.8 Hydrodynamic Loads on Pool Walls The maximum hydrodynamic pressures (in psi) that develop between the fuel racks and the spent fuel pool walls due to fluid coupling are listed below. The runs are selected to represent the worst case conditions. Pool Run Maximum Pressure (psi)Minimum Pressure (psi)Spent Fuel Pool 1 10.3-9.3 2 11.1-13.4 3 9.9-10.2 4 4.6-5.3 5 4.6-5.3 6 4.7-5.4 Cask Loading Pool 7 3.7-4.5 8 4.6-4.9 9 4.3-3.6 10 2.2-1.9 11 2.3-1.9 12 2.2-2.0 These hydrodynamic pressures were considered in the evaluation of the Spent Fuel Building and Pool Structure. 9.1A-48 Rev. 14 WOLF CREEK 9.1A.4.4 Fuel Pool Structure Integrity The Wolf Creek fuel storage pool is a safety related, seismic category 1, reinforced concrete structure. Spent fuel is to be placed within storage racks located in the fuel storage pool. The fuel storage pool includes the spent fuel pool and the cask loading pool with fuel storage racks installed. The area is collectively referred to in this section as the fuel pool structure. An analysis was performed to demonstrate the structural adequacy of the pool structure, as required by Section IV of the USNRC OT Position Paper (Reference 63). The fuel storage pool regions are analyzed using the finite element method. Results for individual load components are combined using factored load combinations mandated by SRP 3.8.4 (Reference 64) based on the "ultimate strength" design method. It is demonstrated that for the critical bounding factored load combinations, structural integrity is maintained when the pools are assumed to be fully loaded with spent fuel racks, as shown in Figure 9.1-2 with all storage locations occupied by fuel assemblies. The regions examined in for the fuel storage pool include the floor slabs and the highly loaded wall sections adjoining the slabs. Both moment and shear capabilities are checked for concrete structural integrity. Local punching and bearing integrity of the slab in the vicinity of a rack module support pedestal pad is evaluated. All structural capacity calculations are made using design formulas meeting the requirements of the American Concrete Institute (ACI). 9.1A.4.4.1 Description of Fuel Storage Pool Structure The FSP is located inside the Fuel Building and is supported on a two way, reinforced concrete base mat which is founded six feet below grade. The minimum thickness of the mat is 6.5 feet and the mat beneath the Spent Fuel Pool is thirteen feet thick. The Cask Loading Pool is located to the South of the SFP and is supported by the base mat which is 7.5 feet thick in this vicinity. The SFP and Cask Loading Pool are separated by a three foot thick reinforced concrete wall. Figure 9.1A-23 shows an isometric view of the SFP, Cask Loading Pool and surrounding major structural features of interest (i.e., Fuel Transfer Canal and Cask Washdown Pit). Figure 9.lA-24 shows the major structural dimensions of the pool. The floor liner plate of the SFP is located at elevation 2006.5 and the floor liner plate of the Cask Loading Pool is 5.5 feet lower at elevation 2001. The spent fuel area operating floor is at elevation 2047.5. 9.1A.4.4.2 Definition of Loads Pool structural loading involves the following discrete components: 9.1A.4.4.2.1 Static Loading (Dead Loads and Live Loads)
- 1. Dead weight of pool structure includes the weight of the Fuel Building concrete upper structure.
- 2. Maximum dead weight of rack modules and fuel assemblies stored in the modules based on 2363 storage locations in the Spent Fuel Pool and 279 storage locations in the Cask Loading Pool, as shown in Figure 9.1-2.
- 3. Dead weight of a shipping cask including yoke of 250 kips. 9.1A-49 Rev. 14 WOLF CREEK
- 4. The Cask Handling Crane and Spent Fuel Handling Machine (Refueling Platform) are designed to move along the N-S direction. The dead weight and the rated lift weight of these cranes are considered as dead load and live load, respectively.
- 5. The hydrostatic water pressure. 9.1A.4.4.2.2 Seismic Induced Loads
- 1. Vertical loads transmitted by the rack support pedestals to the slab during a SSE or OBE seismic event.
- 2. Hydrodynamic inertia loads due to the contained water mass and sloshing loads (considered in accordance with (Reference 66)) which arise during a seismic event.
- 3. Hydrodynamic pressures between racks and pool walls caused by rack motion in the pool during a seismic event.
- 4. Seismic inertia force of the walls and slab. 9.1A.4.4.2.3 Thermal Loading Thermal loading is defined by the temperature existing at the faces of the pool concrete walls and slabs. Two thermal loading conditions are evaluated: The normal operating temperature and theaccident temperature. The effect of gamma heating on the concrete was also considered and requires the implementation of administrative controls to maintain concrete temperatures within acceptable ranges, as discussed in section 9.1A.4.4.5. 9.1A.4.4.2.4 Pool Water Loading The loadings described above were considered for two possible scenarios: one considers the Cask Loading Pool full of water and the other considers the Cask Loading Pool empty. 9.1A.4.4.3 Analysis Methodology9.1A.4.4.3.1 Finite Element Analysis Model The finite element model encompasses the entire Spent Fuel Pool and three other reinforced concrete structures located immediately adjacent to the Spent Fuel Pool (the Cask Loading Pool
, the Transfer Canal and the Cask Washdown Pit). The interaction with the rest of the Fuel Building reinforced concrete, which is not included in the finite-element model, is simulated by imposing appropriate boundary conditions. The structural area of interest for the fuel storage pool includes only two, the spent fuel pool and the cask loading pool. However, by augmenting these areas of interest with the addition of the Transfer Canal and the Cask Washdown Pit, the constructed finite-element model and numerical investigation are enhanced because the perturbation induced by the boundary conditions on the stress field distribution for the area of interest is minimized. 9.1A-50 Rev. 14 WOLF CREEK The preprocessing capabilities of the STARDYNE computer code (Reference 67) are used to develop the 3-D finite-element model. The STARDYNE finite-element model contains 9866 nodes, 3696 solid type finite-elements, 4411 plate type finite-elements and 16 hydro-dynamic masses. The dynamic behavior of the water mass contained in the Spent Fuel Pool and Cask Loading Pool during a seismic event is modeled according to the guidelines set in TID-7024 (Reference 66). To simulate the interaction between the modeled region and the rest of the Fuel Building a number of boundary restraints were imposed upon the described finite-element model. The behavior of the reinforced concrete existing in the structural elements (walls, slab and mat) is considered elastic and isotropic. The elastic characteristics of the concrete are independent of the reinforcement contained in each structural element for the case when the un-cracked cross-section is assumed. This assumption is valid for all load cases with the exception of the thermal loads, where for a more realistic description of the reinforced concrete cross-section including the assumption of cracked concrete is used. To simulate the variation and the degree of cracking patterns, the original elastic modulus of the concrete is modified in accordance with Reference (Reference 65). 9.1A.4.4.3.2 Analysis Technique The structural region isolated from the Fuel Building and comprised of four pools (the Spent Fuel Pool, the Cask Loading Pool, the Transfer Canal and the Cask Washdown Pit) is numerically investigated using the finite element method. The pool walls and their supporting reinforced concrete mat are represented by a 3-D finite-element model. The individual loads considered in the analysis are grouped in five categories: dead load (weight of the pool structure, dead weight of the rack modules and stored fuel, dead weight of the reinforced concrete Fuel Building upper structure, the dead weight of the Cask Handling Crane (CHC) and the Spent Fuel Handling Machine (SFHM), and the hydro-static pressure of the contained water), live loads (CHC and SFHM suspended loads), thermal loads (the thermal gradient through the pool walls and slab for normal operating and accident conditions) and the seismic induced forces (structural seismic forces, interaction forces between the rack modules and the pool slab, seismic loads due to self-excitation of the pool structural elements and contained water, and seismic hydro-dynamic interaction forces between the rack modules and the pool walls for both OBE and SSE conditions). The dead and thermal loads are considered static acting loads, while the seismic induced loads are time-dependent. The material behavior under all type of loading conditions is described as elastic and isotropic representing the un-cracked characteristics of the structural elements cross-section, with the exception of the thermal load cases where the material elasticity modulus is reduced in order to simulate the variation and the degree of the crack patterns. This approach (Reference 65) acknowledges the self-relieving nature of the thermal loads. The degree of reduction of the elastic modulus is calculated based on the average ultimate capacity of the particular structural element. 9.1A-51 Rev. 14 WOLF CREEK The numerical solution (displacements and stresses) for the cases when the structure was subjected to dead and thermal loads is a classical static solution. For the time-dependent seismic induced loads the displacement and stress field are calculated employing the spectra (shock) method. This method requires a prior modal eigenvector and eigenvalues extraction. Natural frequencies of the 3-D finite-element model are calculated up to the rigid range, considered as greater than 34 Hz. Three independent orthogonal acceleration spectra are applied to the model. The acceleration spectra are considered to act simultaneously in three-directions. The SRSS method is used to sum the similar quantities calculated for each direction. Results for individual load cases are combined using the factored load combinations discussed below considering two scenarios: first, when the Spent Fuel Pool and the Cask Storage Pool are full of water (SC1); and second, when only the Spent Fuel Pool (SC2) is full of water. The combined stress resultants are compared with the ultimate moments and shear capacities of all structural elements pertinent to the Spent Fuel Pool and Cask Storage Pool, which are calculated in accordance with the ACI 318-(Reference 59) to develop the safety factors. 9.1A.4.4.3.3 Load Combinations The various individual load cases are combined in accordance with the NUREG-0800 Standard Review Plan (Reference 64) requirements with the intent to obtain the most critical stress fields for the investigated reinforced concrete structural elements. For "Service Load Conditions" the following load combinations are: Load Combination No. 1 = 1.4*D + 1.7*L Load Combination No. 2 = 1.4*D + 1.7*L + I.9*E Load Combination No. 3 = 1.4*D + 1.7*L - l.9*E Load Combination No. 4 = 0.75*(1.4* D + 1.7*L + I.9*E +1.7*T o)Load Combination No. 5 = 0.75*(1.4* D + 1.7*L - l.9*E +1.7*T o)Load Combination No. 6 = 1.2*D + 1.9*E Load Combination No. 7 = 1.2*D - 1.9*E For "Factored Load Conditions" the following load combinations are: Load Combination No. 8 = D + L + T o + E' Load Combination No. 9 = D + L + T o - E' Load Combination No. 10 = D + L + T a + 1.25*E Load Combination No. 11 = D + L + T a - 1.25*E Load Combination No. 12 = D + L + T a + E' Load Combination No. 13 = D + L + T a - E' 9.1A-52 Rev. 14 WOLF CREEK where: D = dead loads; L = live loads; T o = thermal load during normal operation; T a = thermal load under accident condition; E = 0BE earthquake induced loads; E' = SSE earthquake induced loads. 9.1A.4.4.3.4 Results of Analyses The STARDYNE computer code was used to obtain the stress and displacement fields for 18 individual load cases covering the two scenarios: SC1 (spent fuel pool and cask loading pool full of water) and SC2 (spent fuel pool fuel pool full of water and the cask loading pool empty). The STARDYNE postprocessing capability was employed to form the appropriate load combinations and to establish the limiting bending moments and shear forces in various sections of the pool structure. A total of 26 load combinations were computed. Section limit strength formulas for bending loading were computed using appropriate concrete and reinforcement strengths. For Wolf Creek the concrete and reinforcement allowable strengths are: concrete f c' = 4,000 psi reinforcement f y = 60,000 psi Table 9.1A-21 shows results from potentially limiting load combinations for the bending strength of the slab and walls. For each section, we define the limiting safety margins as the limited strength bending moment or shear force defined by ACI for that structural section divided by the calculated bending moment or shear force (from the finite element analyses). The major regions of the pool structure consist of six concrete walls delimiting the SFP and Cask Storage Pool. Each area is searched independently for the maximum bending moments in different bending directions and for the maximum shear forces. Safety margins are determined from the calculated maximum bending moments and shear forces based on the local strengths. The procedures are repeated for all the potential limiting load combinations. Therefore, limiting safety margins are determined. Table 9.1A-21 demonstrates that the limiting safety margins for all sections are above 1.0 as required. Table 9.1A-22 shows results of shear capacity calculations for the slab and walls. Calculated margins are again to be compared with an allowable margin of 1.0.9.1A.4.4.4 Pool Liner The pool liner is subject to in-plate strains due to movement of the rack support feet during the seismic event. Analyses are performed to establish that the liner will not tear or rupture under limiting loading conditions in the pool, and that there is no fatigue problem under the condition of 1 SSE event plus 20 OBE events. These analyses are based on loadings imparted from the most highly loaded pedestal in the pool assumed to be positioned in the most unfavorable position. Bearing strength requirements are shown to be satisfied by conservatively analyzing the most highly loaded pedestal located in the worst configuration with respect to underlying leak chases. 9.1A-53 Rev. 14 WOLF CREEK 9.1A.4.4.5 Gamma Heating Considerations The effect of gamma heating was evaluated along with the temperature differentials across the wall from normal and accident conditions. The concrete and rebar stresses were shown to be acceptable for all conditions. However, gamma heating produces concrete temperatures above 150 F under some short term conditions for the 36 inch thick wall along the south side of the Spent Fuel Pool. However, the excessive temperature will be of short duration due to the rapid reduction in gamma bombardment over the cooling period of the fuel.Although temperatures in excess of the 150°F range are allowed by the ACI code, the effect from gamma heating can be remedied by storage of fuel with longer cooling time along the pool periphery. Therefore, in lieu of performing additional evaluations to determine the acceptability of the gamma heating on the 36" thick wall, administrative controls are provided to ensure spent fuel is cooled at least one year prior to storage along the peripheral rack cells on the South end of the Spent Fuel Pool, or the peripheral rack cells on the North end of the Cask Loading Pool (i.e., in storage module cells adjacent to the 36" wall separating the two pools).9.1A.4.4.6 Conclusions Regions affected by loading the fuel pool completely with high density racks are examined for structural integrity under bending and shearing action. It has been determined that adequate safety margins exist assuming that all racks are fully loaded with a bounding fuel weight and that the factored load combinations are checked against the appropriate structural design strengths. It is also shown that local loading on the liner does not compromise liner integrity under a postulated fatigue condition and that concrete bearing strength limits are not exceeded. 9.1A.5 ADMINISTRATIVE CONTROL OF FUEL MOVEMENT AND STORAGE IN REGION 2 AND 3 Control of fuel movement in the plant and the placement of fuel in Region 2 and 3 of the Fuel Storage Pool is under strict administrative control. This control precludes the possibility of erroneous placement of fuel which has not attained the required burnup (see Figure 9.1A-3) in Region 2 or 3 of the Fuel Storage Pool. Movement of spent fuel or fuel handling in the fuel storage pool with fuel storage present is carried out under supervision of a licensed operator. Under other conditions, fuel movement on site is carried out by a trained operator. Detailed approved procedures are used which give step-by-step action for each fuel movement. When new fuel is received on site, a fuel status record is initiated for each fuel assembly which records the assembly movement throughout the plant including new fuel storage, spent fuel storage and reactor core locations. Material transfer reports are also completed which record, sequentially, each fuel assembly movement. Fuel assembly identification numbers are positively identified when placed in the new fuel storage racks or fuel storage pool. During refueling, after core loading is complete, an inventory of fuel in the core is completed to verify that the core has been loaded in accordance with design documents. Additionally, when the fuel in the fuel storge pool is to be transferred into Region 2 or 3, SNM records are used to identify candidates for Region 2 or 3 storage. 9.1A-54 Rev. 14 WOLF CREEK Prior to the start of transfer of fuel to Region 2 or 3, the history of each fuel assembly to be transferred is reviewed and calculations are performed to determine the amount of burnup each assembly has received. Once it has been determined that a fuel assembly has attained the required burnup, it is added to the list of assemblies designated for movement to Region 2 or 3 of the pool. Following this verification, fuel from Region 1 may be transferred to Region 2 or 3. Fuel is moved through the use of the spent fuel bridge crane described in Section 9.1.4.2.2 and shown in Figures 9.1-8 and 9.1-9. To minimize the possibility of error, a specific fuel assembly from the approved listing is identified for movement along with its current storage location and planned storage location in Region 2 or 3. When the spent fuel bridge crane has been positioned above the fuel assembly, its number is verified. The assembly is then moved to its designated location in Region 2 or 3 of the pool. This process is repeated until all fuel designated for relocation has been
properly transferred to Region 2 or 3 of the pool. In addition to these precautions and controls, physical inventories by
comparing fuel assembly identification numbers with SNM records are implemented prior to each refueling outage which verifies fuel location in the new fuel storage racks and fuel storage pool. These controls, procedures, checks and verifications ensure that the fuel
stored in each location in Region 2 or 3 is the fuel that was designated for storage in that location and that the fuel has attained the required burnup. 9.1A.6 REFERENCES 1. Petrie and N. F. Landers, "KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping," Volume 2, Section F11 from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990. 2. J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).3. N. M. Greene, L.M. Petrie and R.M. Westfall, 3. "NITAWLII: Scale System Module for Performing Shielding and Working Library Production," Volume 1, Section F1 from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR0200, Rev. 4, January 1990.4. M.G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, August 1963. 5. A. Ahlin, M. Edenius, and H. Haggblom, "CASMO -.A Fuel Assembly Burnup Program", AERF-76-4158, Studsvikreport. 6. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, Vol. 26, p. 604, 1977. 7. M. Edenius, A. Ahlin and B.H. Forssen, "CASMO-3 A Fuel Assembly Burnup Program User's Manual," Studsvik/NFA-86/7, Studsvik Energitechnik AB, November 1986. 8. S. E. Turner, "Waterford Criticality Analysis," HI-961562, 1996. 9. M. Edenius and A. Ahlin, "CASMO-3: New Features, Benchmarking, and Advanced Applications," Nucl. Sci. Eng, 100 (1988). 10. S. E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988. 11. General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling.12. USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3 - July 1981. 9.1A-55 Rev. 14 WOLF CREEK 13. USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979. 14. USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981. 15. ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors. 16. ANSI/ANS 8.1 (N16.1) - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors. 17. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993). 18. SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995. 19. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460) Oak Ridge National Laboratory, September 1994. 20. W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, December 1986. 21. O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated. 22. R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications , Prentice-Hall, 1986. 23. M. N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW- 1484-7, Babcock and Wilcox Company, July 1979. 24. G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW - 16454, Babcock & Wilcox Company, November 1991. 25. L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984. 26. J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods," Trans. Am. Nucl. Soc. 33: 362-364 (1979). 27. S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235U Enriched UO2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981. 28. S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235u Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981. 29. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977. 30. R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990. 31. B.M. Durst et al., Critical Experiments with 4.31 wt % 235U Enriched UO 2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982. 32. S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981. 33. E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965. 9.1A-56 Rev. 14 WOLF CREEK 34. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963. 35. Heat Loss to the Ambient from Spent Fuel Pools: Correlation of Theory with Experiment", Holtec Report HI-90477, Rev. 0, April 3, 1990. 36. "An Improved Correlation for Evaporation from Spent Fuel Pools", Holtec Report HI- 971664, Rev. 0. 37. Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press, 1967. 38. Hinze, J.O., "Turbulence", McGraw Hill Publishing Co., New York, NY, 1975.39. Launder, B.E., and Spalding, D.B., "Lectures in Mathematical Models of Turbulence", Academic Press, London, 1972. 40. "QA Documentation and Validation of the FLUENT Version 4.32 CFD Analysis Program", Holtec Report HI-961444. 41. Rohsenow, N.M., and Hartnett, J.P., "Handbook of Heat Transfer", McGraw Hill Book Company, New York, 1973. 42. USNRC NUREG-0800, Standard Review Plan, June 1987. 43. (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, and January 18, 1979 amendment thereto. 44. Soler, A.I. and Singh, K.P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp. 315-329 (1984). 45. Soler, A.I. and Singh, K.P., "Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool", INNM Spent Fuel Management Seminar X, January, 1993. 46. Singh, K.P. and Soler, A.I., "Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991). 47. Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre&Post Processors & Solver, August, 1997. 48. USNRC Standard Review Plan, NUREG-0800 (Section 3.7.1, Rev. 2, 1989). 49. Holtec Proprietary Report HI-89364 - Verification and User's Manual for Computer Code GENEQ, January, 1990. 50. Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976. 51. Singh, K.P. and Soler, A.I., "Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.52. Fritz, R.J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172. 53. Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering," McGraw Hill, 1976.54. Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment", (Proprietary), NUSCO/Holtec Report HI-88243. 55. ASME Boiler & Pressure Vessel Code, Section III, Subsection NF, 1989 Edition.56. ASME Boiler & Pressure Vessel Code, Section III, Appendices, 1989 Edition.57. USNRC Standard Review Plan, NUREG-0800 (Section 3.8.4, Rev. 2, 1989). 58. ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985. 59. ACI 318-71, Building Code requirements for Structural Concrete. 60. R. W. Westfall and J. H. Knight, "SCALE System Cross-Section Validation with Shipping Cask Critical Experiments", ANS Transactions , Vol. 33, p. 368, November 1979. 9.1A-57 Rev. 14 WOLF CREEK 61. Scavuzzo, R. J., Stokey, W. F., and Radke, E. F., "Dynamic Fluid-Structure Coupling of Rectangular Modules in Rectangular Pools", ASME Special Publication, Dynamics at Fluid-Structure Systems in the Energy Industry, PVP-39 June, 1979, pp. 77-86. 62. Specification C-175, "Technical Specification for Purchase of Spent Fuel Storage Racks for the Standardized Nuclear Unit Power Plant System (SNUPPS)", Appendix B. 63. OT Position for Review and Acceptance of Spent Fuel Handling Applications, by B.K. Grimes, USNRC, Washington, D.C., April 14, 1978. 64. NUREG-0800, SRP-3.8.4, Rev. 1., July 1981. 65. ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit Michigan. 66. "Nuclear Reactors and Earthquakes, U.S. Department of Commerce, National Bureau of Standards, National Technical Information Service, Springfield, Virginia (TID 7024). 67. STARDYNE User's Manual, Research Engineers, Inc., Rev. 4.4. 9.1A-58 Rev. 14 WOLF CREEK Table 9.1A-1 Summary of the Criticality Safety Analysis for the MZTR Storage Configuration Design Basis Burnups at 5.0 0.05 wt% 235 Uinitial enrichment 0 in Region 1 50 in Region 2 40.75 in Region 3 Temperature for Analysis 20 C Uncertainties Manufacturing tolerances (Table 4.5.1) 0.0059 Water-gap (horizontal) 0.0014 Water-gap (vertical) 0.0003 Burnup (Region 2) 0.0056 Burnup (Region 3) 0.0001 Eccentricity in position negativ e KENO5a statistics (95%/95%) 0.0003 Bias statistics (95%/95%) 0.0012 Statistical combination of uncertainties
- 0.0084 Region 1 Fuel Description 5.0 wt%235 Uwith 16 IFBA rods 4.6 wt%235 Uwith no IFBA rods Reference k eff (KENO5a) 0.9266 0.9294 Total Uncertainty (above) 0.0084 0.0084 Calculational Bias (see Appendix A) 0.0030 0.0030 Axial Burnup Effect negative negative Temperature Correction to 4 C (39 F)0.0020 0.0020 Maximum k eff 0.9400 0.9428 Limiting k eff 0.9500 0.9500* Square root of the sum of the squares.
Rev. 14 WOLF CREEK Table 9.1A-2 Summary of the Criticality Safety Analysis for the Interim Checkerboard Storage Configuration Temperature for Analysis 20 C Uncertainties Manufacturing tolerances (Table 4.5.1) 0.0059 Water-gap (horizontal) 0.0014 Water-gap (vertical) 0.0003 Burnup (Region 2) N/A Burnup (Region 3) N/A Eccentricity in position negative KENO5a statistics (95%/95%) 0.0004 Bias statistics (95%/95%) 0.0012 Statistical combination of uncertainties
- 0.0062 Fuel Description 5.0 wt%235 Uwith 16 IFBA rods 4.6 wt%235 Uwith no IFBA rods Reference k eff (KENO5a) 0.8439 0.8490 Total Uncertainty (above) 0.0062 0.0062 Calculational Bias (see Appendix A)0.0030 0.0030 Axial Burnup Effect negative negative Temperature Correction to 4 C (39 F)0.0020 0.0020 Maximum k eff 0.8551 0.8602 Limiting k eff 0.9500 0.9500* Square root of the sum of the squares.
Rev. 14 WOLF CREEK Table 9.1A-3 Reactivity Effects of Abnormal and Accident Conditions Abnormal/Accident Conditions Reactivity Effect Temperature Increase (above 4 C)Negative (Table 9.1A-6) Void (boiling) Negative (Table 9.1A-6) Assembly Drop (on top of rack) Negligible Assembly Drop (adjacent to rack)Positive - controlled by < 500 ppm soluble boron Lateral Rack Movement Included in Tolerances Misplacement of a fresh fuel assembly Positive - controlled by 500 ppm soluble boron Multiple misplaced fuel assemblies Positive - controlled by 2165 ppm soluble boron Rev. 14 WOLF CREEK Table 9.1A-4 Design Basis Fuel Assembly Specifications Fuel Rod Data Assembly type OFA Standard Vantage-5H Fuel pellet outside diameter, in. 0.3088 0.3225 0.3225 Cladding thickness, in. 0.0225 0.0225 0.0225 Cladding outside diameter, in. 0.360 0.374 0.374 Cladding material Zr Zr Zr Pellet density, % T.D. 95.0 95.0 95.0 Maximum nominal enrichment, wt% 235 U 5.0 5.0 5.0 Fuel Assembly Data Fuel rod array 17 x 17 17 x 17 17 x 17 Number of fuel rods 264 264 264 Fuel rod pitch, in. 0.496 0.496 0.496 Number of control rod guide and instrument thimbles 25 25 25 Thimble outside diameter, in.0.474 0.482 0.474 Thimble thickness, in. 0.016 0.016 0.016 Number of IFBA rods 16 16 16 Active fuel Length, in. 144 144 144 Rev. 14 WOLF CREEK Table 9.1A-5 Reactivity Effects of Manufacturing Tolerances Tolerance Reactivity Effect, k Minimum boral loading (0.03 g/cm 2 , 0.0324 g/cm 2 nominal) 0.0044 Minimum boral width (7.4375, 7.5 nominal) 0.0010 Minimum box I.D. (8.73, 8.77 nominal) 0.0016 Maximum SS thickness (0.079, 0.075 nominal) 0.0002 Density tolerance (10.61 g/cm 3 , 10.41 g/cm 3 nominal) 0.0019 Enrichment tolerance (5.05%, 5.0% nominal) 0.0023 Total (statistical sum)
- 0.0059
- Square root of the sum of the squares Rev. 14 WOLF CREEK Table 9.1A-6 Reactivity Effects of Temperature and Void Reactivity Effect, k Case Region 1 (Fresh fuel)Region 2 (50 MWd/kgU)
Region 3 (40.75 MWd/kgU) 4 C (39F) reference reference reference 20 C (68F) -0.002-0.0036-0.0034 60 C (140 F)-0.0095-0.0137-0.0134 120 C (248 F)-0.0252-0.0314-0.0313 120 C w/ 10% void -0.0496-0.0484-0.0501 Rev. 14 WOLF CREEK Table 9.1A-7 Summary of Criticality Benchmark Calculations Calculated k eff 1 EALF (eV)Reference Identification Enri ch.MCNP4a KENO5a MCNP4a KENO5a 1 B&W-1484 (Ref. 23) Core I 2.46 0.9964 0.0010 0.9898 0.0006 0.17589 0.1753 2 B&W-1484(Ref. 23) Core II 2.46 1.0008 0.0011 1.0015 0.0005 0.2553 0.2446 3 B&W-1484 (Ref. 23) Core III 2.46 1.0010 0.0012 1.0005 0.0005 0.1999 0.1939 4 B&W-1484 (Ref. 23) Core IX 2.46 0.9956 0.0012 0.9901 0.0006 0.1422 0.1426 5 B&W-1484 (Ref. 23) Core X 2.46 0.9980 0.0014 0.9922 0.0006 0.1513 0.1499 6 B&W-1484 (Ref. 23) Core XI 2.46 0.9978 0.0012 1.0005 0.0005 0.2031 0.1947 7 B&W-1484 (Ref. 23) Core XII 2.46 0.9988 0.0011 0.9978 0.0006 0.1718 0.1662 8 B&W-1484 (Ref. 23) Core XIII 2.46 1.0020 0.0010 0.9952 0.0006 0.1988 0.1965 9 B&W-1484 (Ref. 23) Core XIV 2.46 0.9953 0.0011 0.9928 0.0006 0.2022 0.1986 10 B&W-1484 (Ref. 23) Core XV 2.46 0.9910 0.0011 0.9909 0.0006 0.2092 0.2014 11 B&W-1484 (Ref. 23) Core XVI 2.46 0.9935 0.0010 0.9889 0.0006 0.1757 0.1713 12 B&W-1484 (Ref. 23) Core XVII 2.46 0.9962 0.0012 0.9942 0.0005 0.2083 0.2021 13 B&W-1484 (Ref. 23) Core XVIII 2.46 10.36 0.0012 0.99310 0.0006 0.1705 0.1708 14 B&W-1484 (Ref. 23) Core XIX 2.46 0.9961 0.0012 0.9971 0.0005 0.2103 0.2011 15 B&W-1484 (Ref. 23) Core XX 2.46 1.0008 0.0011 0.9932 0.0006 0.1724 0.1701 16 B&W-1484 (Ref. 23) Core XXI 2.46 0.9994 0.0010 0.9918 0.0006 0.1544 0.1536 17 B&W-1645 (Ref. 24) S-Type Fuel, w/886 ppm B 2.46 0.9970 0.0010 0.9924 0.0006 1.4475 1.4680 18 B&W-1645 (Ref. 24) S-Type Fuel, w/746 ppm B 2.46 0.9990 0.0010 0.9913 0.0006 1.5463 1.5660 Rev. 14 WOLF CREEK Table 9.1A-7 (sheet 2) 19 B&W-1645 (Ref. 24) SO-Type Fuel, w/1156 ppm B 2.46 0.9972 0.0009 0.9949 0.0005 0.4241 0.4331 20 B&W-1810 (Ref. 25) Case 1 1337 ppm B 2.46 1.0023 0.0010 NC 0.1531 NC 21 B&W-1810 (Ref. 25) Case 12 1899 ppm B 2.46/4.02 1.0060 0.0009 NC 0.4493 NC 22 French (Ref. 26) Water Moderator 0 gap 4.75 0.9966 0.0013 NC 0.2172 NC 23 French (Ref. 26) Water Moderator 2.5 cm gap 4.75 0.9952 0.0012 NC 0.1778 NC 24 French (Ref. 26) Water Moderator 5 cm gap 4.75 0.9943 0.0010 NC 0.1677 NC 25 French (Ref. 26) Water Moderator 10 cm gap 4.75 0.9979 0.0010 NC 0.1736 NC 26 PNL-3602 (Ref. 27) Steel Reflector, 0 separation 2.35 NC 1.0004 0.0006 NC 0.1018 27 PNL-3602 (Ref. 27) Steel Reflector, 1.321 cm sepn. 2.35 0.9980 0.0009 0.9992 0.0006 0.1000 0.0909 28 PNL-3602 (Ref. 27) Steel Reflector, 2.616 cm sepn. 2.35 0.9968 0.0009 0.99640 0.0006 0.0981 0.0975 29 PNL-3602 (Ref. 27) Steel Reflector, 3.912 cm sepn. 2.35 0.9974 0.0010 0.9980 0.0006 0.0976 0.0970 30 PNL-3602 (Ref. 27) Steel Reflector, infinite sepn. 2.35 0.9962 0.0008 0.9939 0.0006 0.0973 0.0968 31 PNL-3602 (Ref. 27) Steel Reflector, 0 sepn. 4.306 NC 1.0003 0.0007 NC 0.3282 32 PNL-3602 (Ref. 27) Steel Reflector, 0.312 cm sepn. 4.306 0.9997 0.0010 1.0012 0.0007 0.3016 0.3039 33 PNL-3602 (Ref. 27) Steel Reflector, 2.616 cm sepn. 4.306 0.9994 0.0012 0.9974 0.0007 0.2911 0.2927 34 PNL-3602 (Ref. 27) Steel Reflector, 5.405 cm sepn. 4.306 0.9969 0.0011 0.9951 0.0007 0.2825 0.2860 35 PNL-3602 (Ref. 27) Steel Reflector, infinite sepn. 4.306 0.9910 0.0020 0.9947 0.0007 0.2851 0.2864 36 PNL-3602 (Ref. 27) Steel Reflector, with Boral Sheets 4.306 0.9941 0.0011 0.9970 0.0007 0.3135 0.3150 37 PNL-3926 (Ref. 28) Lead Reflector, 0 cm sepn. 4.306 NC 1.0003 0.0007 NC 0.3159 38 PNL-3926 (Ref. 28) Lead Reflector, 0.55 cm sepn. 4.306 1.0025 0.0011 0.9997 0.0007 0.3030 0.3044 39 PNL-3926 (Ref. 28) Lead Reflector, 1.956 cm sepn. 4.306 1.0000 0.0012 0.9985 0.0007 0.2883 0.2930 40 PNL-3926 (Ref. 28) Lead Reflector, 5.405 cm sepn. 4.306 0.9971 0.0012 0.9946 0.0007 0.2831 0.2854 Rev. 14 WOLF CREEK Table 9.1A-7 (sheet 3) 41 PNL-2615 (Ref. 29) Experiment 004/0.3 - no absorber 4.306 0.9925 0.0012 0.9950 0.0007 0.1155 0.1159 42 PNL-2615 (Ref. 29) Experiment 030 - Zr plates 4.306 NC 0.9971 0.0007 NC 0.1154 43 PNL-2615 (Ref. 29) Experiment 013 - Steel plates 4.306 NC 0.9965 0.0007 NC 0.1164 44 PNL-2615 (Ref. 29) Experiment 014 - Steel plates 4.306 NC 0.9972 0.0007 NC 0.1164 45 PNL-2615 (Ref. 29) Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 0.0010 0.9981 0.0007 0.1172 0.1162 46 PNL-2615 (Ref. 29) Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 0.0012 0.9982 0.0007 0.1161 0.1173 47 PNL-2615 (Ref. 29) Exp. 031 - Boral plates 4.306 0.9994 0.0012 0.9969 0.0007 0.1165 0.1171 48 PNL-7167 (Ref. 30) Experiment 214R - with flux trap 4.306 0.9991 0.0011 0.9956 0.0007 0.3722 0.3812 49 PNL-7167 (Ref. 30) Experiment 214V3 - with flux trap 4.306 0.9969 0.0011 0.9963 0.0007 0.3742 0.3826 50 PNL-4267 (Ref. 31) Case 173 - 0 ppm B 4.306 0.9974 0.0012 NC 0.2893 NC 51 PNL-4267 (Ref. 31) Case 177 - 2550 ppm B 4.306 1.0057 0.0010 NC 0.5509 NC 52 PNL-5803 (Ref. 32) MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 0.0011 1.0046 0.0006 0.9171 0.8868 53 PNL-5803 (Ref. 32) MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 0.0012 1.0036 0.0006 0.2968 0.2944 54 PNL-5803 (Ref. 32) MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 0.0011 0.9989 0.0006 0.1665 0.1706 55 PNL-5803 (Ref. 32) MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 0.0011 0.9966 0.0006 0.1139 0.1165 56 WCAP -3385 (Ref. 33) Saxton Case 52 PuO2 0.52 pitch 6.6% Pu 0.9996 0.0011 1.0005 0.0006 0.8665 0.8417 57 WCAP -3385 (Ref. 33) Saxton Case 52 U 0.52 pitch 5.74 1.0000 0.0010 0.9956 0.0007 0.4476 0.4580 58 WCAP -3385 (Ref. 33) Saxton Case 56 PuO2 0.56 pitch 6.6% Pu 1.0036 0.0011 1.0047 0.0006 0.5289 0.5197 59 WCAP -3385 (Ref. 33) Saxton Case 56 borated PuO2 6.6% Pu 1.0008 0.0010 NC 0.6389 NC Rev. 14 WOLF CREEK Table 9.1A-7 (sheet 4) 60 WCAP -3385 (Ref. 33) Saxton Case 56 U 0.56 pitch 5.74 0.9994 0.0011 0.9967 0.0007 0.2923 0.2954 61 WCAP -3385 (Ref. 33) Saxton Case 79 PuO2 0.79 pitch 6.6% Pu 1.0063 0.0011 1.0133 0.0006 0.1520 0.1555 62 WCAP -3385 (Ref. 33) Saxton Case 79 U 0.79 pitch 5.74 1.0039 0.0011 1.0008 0.0006 0.1036 0.1047Notes: NC stands for not calculated. EALF is the energy of the average lethargy causing fission. These experimental results appear to be statistical outliers (3) suggesting the possibility of unusually large experimental error. Although they could justifiably be excluded, for conservatism, they were retained in determining the calculational basis. Rev. 14 WOLF CREEKTable9.1A-8ComparisonofMCNP4aandKENO5aCalculatedReactivitiesforVariousEnrichments Calculated k eff+/-+/-+/-+/- 1Enrichment MCNP4a KENO5a 3.0 0.8465+/- 0.0011 0.8478+/- 0.0004 3.5 0.8820+/- 0.0011 0.8841+/- 0.0004 3.75 0.9019+/- 0.0011 0.8987+/- 0.0004 4.0 0.9132+/- 0.0010 0.9140+/- 0.0004 4.2 0.9276+/- 0.0011 0.9237+/- 0.0004 4.5 0.9400+/- 0.0011 0.9388+/- 0.0004 Based on the GE 8x8R fuel assemblyRev.14 WOLFCREEKTable9.1A-9MCNP4aCALCULATEDREACTIVITIESFORCRITICALEXPERIMENTSWITHNEUTRONABSORBERS Ref.ExperimentkWorth of Absorber MCNP4a Calculated k eff+/-+/-+/-+/-1EALF(eV)29 PNL-2615BoralSheet 0.0139 0.9994+/-0.0012 0.1165 23 B&W-1484CoreXX 0.0165 1.0008+/-0.0011 0.1724 29 PNL-26151.62%Boron-steel 0.0165 0.9996+/-0.0012 0.1161 23 B&W-1484CoreXIX 0.0202 0.9961+/-0.0012 0.2103 23 B&W-1484CoreXXI 0.0243 0.9994+/-0.0010 0.1544 23 B&W-1484CoreXVII 0.0519 0.9962+/-0.0012 0.2083 27 PNL-3602BoralSheet 0.0708 0.9941+/-0.0011 0.3135 23 B&W-1484CoreXV 0.0786 0.9910+/-0.0011 0.02092 23 B&W-1484CoreXVI 0.0845 0.9935+/-0.0010 0.1751 23 B&W-1484CoreXIV 0.1575 0.9953+/-0.0011 0.2022 23 B&W-1484CoreXIII 0.1738 1.0020+/-0.0011 0.1988 30 PNL-7167Expt214Rflux trap 0.1931 0.9991+/-0.0011 0.3722EALFistheenergyoftheaveragelethargycausingfission.Rev.14 WOLF CREEK Table 9.1A-10 COMPARISON OF MCNP4a and KENO5a CALCULATED REACTIVITIES for VARIOUS 10 B LOADINGS Calculated k eff 1 10 B, g/cm 2 MCNP4a KENO5a 0.005 1.0381 0.0012 1.0340 0.0004 0.010 0.9960 0.0010 0.9941 0.0004 0.015 0.9727 0.0009 0.9713 0.0004 0.020 0.9541 0.0012 0.9560 0.0004 0.025 0.9433 0.0011 0.9428 0.0004 0.03 0.9325 0.0011 0.9338 0.0004 0.035 0.9234 0.0011 0.9251 0.0004 0.04 0.9173 0.0011 0.9179 0.0004 Based on a 4.5% enriched GE 8x8R fuel assembly Rev. 14 WOLFCREEKTable9.1A-11CALCULATIONSFORCRITICALEXPERIMENTSWITHTHICKLEADANDSTEELREFLECTORSRef.Case E, wt%Separation,cm MCNP4a k eff+/-+/-+/-+/-1KENO5a k eff+/-+/-+/-+/-127 Steel Reflector 2.35 2.35 2.35 2.35 1.321 2.616 3.9120.9980+/-0.0009 0.9968+/-0.0009 0.9974+/-0.0010 0.9962+/-0.0008 0.9992+/-0.0006 0.9964+/-0.0006 0.9980+/-0.0006 0.9939+/-0.0006 27 Steel Reflector 4.306 4.306 4.306 4.306 1.321 2.616 3.4050.9997+/-0.0010 0.9994+/-0.0012 0.9969+/-0.0011 0.9910+/-0.0020 1.0012+/-0.007 0.9974+/-0.0007 0.9951+/-0.0007 0.9947+/-0.0007 28 Lead Reflector 4.306 4.306 4.306 0.55 1.956 5.405 1.0025+/-0.0011 1.0000+/-0.0012 0.9971+/-0.0012 0.9997+/-0.0007 0.9985+/-0.0007 0.9946+/-0.0007Arrangedinorderofincreasingreflector-fuelspacing.Rev.14 WOLFCREEKTable9.1A-12CALCULATIONSFORCRITICALEXPERIMENTSWITHVARIOUSSOLUBLEBORONCONCENTRATIONSCalculatedk eff+/-+/-+/-+/-1Reference Experiment Boron Concentration, ppm MCNP4a KWNO4a 31 PNL-4267 0 0.9974+/-0.0012-24 B&W-1645 886 0.9970+/-0.0010 0.9924+/-0.0006 25 B&W-1810 1337 1.0023+/-0.0010-25 B&W-1810 1899 1.0060+/-0.0009-31 PNL-4267 2550 1.0057+/-0.0010-Rev.14 WOLF CREEK Table 9.1A-13 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO5a Reference Case+k eff 1 EALF++k eff 1 EALF++PNL-5803 (Ref. 32) MOX Fuel - Exp. No. 21 MOX Fuel - Exp. No. 43 MOX Fuel - Exp. No. 13 MOX Fuel - Exp. No. 32 1.0041 0.0011 1.0058 0.0012 1.0083 0.0011 1.0079 0.0011 0.9171 0.2968 0.1665 0.1139 1.0046 0.0006 1.0036 0.0006 0.9989 0.0006 0.9966 0.0006 0.8868 0.2944 0.1706 0.1165 WCAP-3385-54 (Ref. 33) Saxton @ 0.52 pitch Saxton @ 0.56 pitch Saxton @ 0.56 pitch borated Saxton @ 0.79 pitch 0.9996 0.0011 1.0036 0.0011 1.0008 0.0010 1.0063 0.0011 0.8865 0.5289 0.6389 0.1520 1.0005 0.0006 1.0047 0.0006 NC 1.0133 0.0006 0.8417 0.5197 NC 0.1555 Note: NC stands for not calculated + = Arranged in order of increasing lattice spacing ++ = EALF is the energy of the average lethargy causing fission. Rev. 14 WOLF CREEK Table 9.1A-14 DATA FOR DECAY HEAT LOAD LIMIT EVALUATION Length of Spent Fuel Pool (min.) 597.56 inch Width of Spent Fuel Pool (min.) 339 inch Pool Building Ambient Temperature 110 F Emissivity of Water 0.96 Specific Heat of Water 0.998 Btu/lb. x F HX Temperature Effectiveness 0.48981 (partial core) 0.3116 (full core) Coolant Water Inlet Temperature 105 F 130 F (post LOCA) Coolant Water Flow Rate 1.50x10 6 lb/hr (partial core) 0.75x10 6 lb/hr (post LOCA) 3.00x10 6 lb/hr (full core)*
- For practical purpose, it is acceptable to use one train of the CCW cooling with a normal CCW flow rate of 1.50x10 6 lb/hr (i.e., 3000 gpm) as long as the calculated maximum bulk pool temperature of 170 F would not be exceeded during the full core offload process. A fuel storage pool cooling calculation, with consideration of the planned outage schedule including the timing for commencement and completion of the core offload, will be performed on a cycle specific basis to confirm the acceptability of the normal CCW flow rate.
Rev. 14 WOLFCREEKTable9.1A-15DATAFORTIME-TO-BOILEVALUATION Length of Spent Fuel Pool 597.56 inch Width of Spent Fuel Pool 339 inch Depth of Spent Fuel Pool 37.25 ft.Total Fuel Rack Weight 411,320 lb. Number of Fuel Assemblies 2,642 assemblies Bounding Assembly Weight 1,467 lb.Pool Building Ambient Temperature 110°F Emissivity of Water 0.96 Pool Thermal Capacity 3.144x10 6 Btu/°F Specific Heat of Water 0.998 Btu/(lb. x °F)Rev.14 WOLF CREEKTable9.1A-16DATAFORLOCALTEMPERATUREEVALUATION Bounding Assembly Weight 1467 lb Maximum Fuel Assembly Heat Flux 1870 Btu/hr-ft 2 Radial Peaking Factor 1.65 Total Peaking Factor
2.5 Number
of Fuel Assemblies 2642 SFPCCS Water Flow Rate 1.625x10 6 lb/hr Type of fuel assembly Westinghouse 17 x17 Std.Fuel Rod Outer Diameter 0.374 in Rack Cell Inner Dimension 8.77 in Active Fuel Length 144 Number of Fuel Rods per Assembly* 289 rods Rack Cell Length 169 inMinimum Bottom Plenum Height 5 in* Note: Fuel assembly is modeled as a square array with all locations containing fuel rods.Rev.14 WOLF CREEKTABLE9.1A-17RESULTSOFDECAYHEATLOADLIMITEVALUATION Scenario Number of SFPCCS Trains Maximum Bulk Temperature Maximum Decay Heat Load Limit (Btu/hr x 10 6)Required Makeup Water Volume (gpm)Partial- core offload 1 140°F (limit)27.15 1.80 Full-core offload 1 170°F (calculated) 63.41 5.57Rev.14 WOLF CREEKTable9.1A-18RACKMATERIALDATA(200 °F)(ASME-SectionII,PartD) RACK MATERIAL DATA (200 °°°°F)(ASME - Section II, Part D) Material Young's Modulus E (psi)Yield Strength S y (psi)Ultimate Strength Su (psi)SA240; 304L S.S. 27.6 x 10 6 21,300 66,200 SUPPORT MATERIAL DATA (200 °°°°F)SA240, Type 304L (upper part of support feet) 27.6 x 10 6 21,300 66,200 SA-564-630 (lower part of support feet; age hardened at 1100°F)28.5 x 10 6 106,300 140,000Rev.14 WOLF CREEKTable9.1A-19TIMEHISTORYSTATISTICALCORRELATIONRESULTS TIME-HISTORY STATISTICAL CORRELATION RESULTS OBE Data1 to Data2 0.0793 Data1 to Data3 0.0174 Data2 to Data3 0.0464 DBE Data1 to Data2 0.0061 Data1 to Data3 0.0127 Data2 to Data3 0.0874 Data1 corresponds to the time-history acceleration values along the X axis (North) Data2 corresponds to the time-history acceleration values along the Y axis (West) Data3 corresponds to the time-history acceleration values along the Z axis (Vertical)Rev.14 WOLF CREEKRev.14Table9.1A-20COMPARISONOFBOUNDINGCALCULATEDLOADS/STRESSES VS.CODEALLOWABLESATIMPACTLOCATIONSANDATWELDS OBE SSE Item/Location Calculated Allowable Calculated Allowable Fuel assembly/cell wall impact, lbf. 403 3,404840 3,404Rack/baseplate weld, psi 12,111 19,860 22,514 35,748 Femalepedestal/baseplate weld, psi 8,099 19,860 21,617 35,748Cell/cell welds, lbf. 1,1403,195 2,546 5,751 fBasedonthelimitloadforacellwall.Theallowableloadonthefuelassemblyitselfmaybelessthanthisvaluebutisgreaterthan840lbs ffBasedonthefuelassemblytocellwallimpactloadsimultaneouslyappliedintwoorthogonal directions. WOLF CREEKTable9.1A-21SHEARSTRENGTHEVALUATION Location Limiting Safety Margin Critical Load Combinations (see Section 9.1A.4.3.4.6) CP + SFP East Wall 1.07 SC1 (12)SFP West Wall 3.61 SC2 (13)SFP North Wall 3.23 SC1 (12)CP North + SFP South Wall 3.43 SC2 (13)Cask Loading Pit West Wall 1.49 SC1 (13)Cask Loading Pit South Wall 2.40 SC1 (13)Note:SC1 corresponds to the condition SFP and CP full of water and SC2 corresponds to the condition SFP full and CP empty.Rev.14 WOLF CREEKTable9.1A-22BendingStrengthEvaluation Location Limiting Safety Margin Critical Load Combinations (see Section 9.1A.4.3.4.6) CP + SFP East Wall 1.07 SC1 (12)SFP West Wall 3.61 SC2 (13)SFP North Wall 3.23 SC1 (12)CP North + SFP South Wall 3.43 SC2 (13)Cask Loading Pit West Wall 1.49 SC1 (13)Cask Loading Pit South Wall 2.40 SC1 (13)Note:SC1correspondstotheconditionSFPandCPfullofwaterandSC2correspondstotheconditionSFPfullandCPempty.Rev.14 Region 2 Region 3 Region 3 Reg. 3 Region 2 Region 3 Region 3 Reg. 3 Region 2 Region 3 Region 3 Reg. 3 Region 2 Region 2 Region 2 Region 2 Region 2 Reg. 2 Region 2 Region 3 Region 3 Reg. 3 Region 2 Region 3 Region 3 Reg. 3 Reg. Region 2 Region 2 Region 3 Region 3 3 Rev.14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-1 REPRESENTATION OF THE KEN05a REFERENCE MZTR CALCULATIONAL MODEL Rev.14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-2 REPRESENTATION OF THE KEN05a REFERENCE CHECKERBOARD CALCULATIONAL MODEL ' w It: :::) 111 0 D... X w w > i= :s :::) ::::E :::) u ID w Vl Vl "" ....J w :::) ..... -: - -AC EPTAB E D)MAIN v -FOR REGIOI 2 AND 3 S. ORAGE *
- uv o.Al ..
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<b v (J<,;-1/,-: -/¥ UNACt:EPTAELE BU NUP COMAIN . FOR 2 OR 3 STO . -0 I I I I I I I I I I I I I I I I I I I I 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 NOMINAL INITIAL ENRICHMENT {w/o U-235) Minimum Required Fuel Assembly Bunmp as a FUDCtion of NomiJJal Initial Enrichment to Permit Srorage in Regions 2 ind 3 (Fuel assemblies with enrichmenrs less than 2.0 wt% 135 U will conservatively be required to meet the burnup requirements of2.0 wt% mu assemblies). WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-3 REV. 14
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..., *
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- Energy of Average Lethargy Causing Fission (Log Scale) 1 KEN05a CALCULATED k-eff VALUES FOR VARIOUS VALUES OF THE SPECTRAL INDEX I *
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lXI ..., 0 c: Gl u --Gl 0 u c: 0 :;: 0 D '-0 u ..c: -3: c 0 .. .. Gl '-01 0 Q: '-0 0 c: ::::i 0 0 Clllll I i1 I ..l. .1. I r -= 11 , r -r 'I' I I I 1/) I 0 --I m I I I I I I I I I -' Ul L..L ...l II I I I 'I nll'T "fill I I I I I I I I I 1 r 1_,1 .: ... IT I 0 j I I I I I I I I I 10 0 0 I I I 'I I 111111111 0 Ul 0 01 0 01 -0 9A!J:l9JJ9-l! p&JDin:liD:> 0 01 en 0 0 1--:"' ,..:-,;.; f.. ..... _Ul ,... f.. f.. 1: ..... 1: -:"' 1-1-..... f: 'Ill I f.. 111111111 -. Ul co 01 0 10 ..., ('I I :::::1 0 ........ 3 -c
- E .s:: 0 -.::: c w WOLF CREEK Vl 1-VlZ WL&J ::::l:::E
...JI <Cu >-a:::: -z 'QjL&J 110 oN wl 1-::::l <( ....lV) ::::l:::J uo ...J-<(0::: u<C > a ...... z<C w UPDATED SAFETY ANALYSIS FIGURE 9.1A-7 REV. 14 .. c: 0 -a :I () a u --.. I ..Y. 0 I() 0 z w :l.:: --0.94 . . . --0.92 -. . -0.90 ---0.88 . . -0.86 ---: -0.84 0.84 ( 4.5l / f-A E v 3. 511 E E v v v f 3.0. . . . . ' .... .... ... ... ' .. ....... '"' 0.86 0.88 0.90 0.92 0.94 MCNP k-eff Calculations COMPARISON OF MCNP AND KENOSA CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS WOLF CREEK E UPDATED SAFETY ANALYSIS REPORT ---J -------------- FIGURE 9.1A-8 '----------------------- REV. 14
0.. z 1 .04 1.03 1.02 1.01 : : --: -: 1.00 ----:!:: 0. 99 "C Q) +-0 :::1 0 a u -> -0 a Q) a::: : : 0.98 -.. , -0.97 -0.96 -: -0.95 : -0.94 -0.93 : : 0.92 --I I I 0.91 0.900 nno; ' -r I / r ........ L I 015 g.l'en.q / j_ -B.02B g/co q ¥,.1125 /crr.q :i .030 g/cme :,ia.wa g/cmllq 1 0.04 ...... ' ' ' ' ' ' ' ' ' ' ' I I I I I I 0.920 0.940 0.960 0.980 1.000 1.020 Reactivity Calculated with KEN05a COMPARISON OF MCNP AND KEN05a CALCULATIONS FOR VARIOUS BORON-1 C AREAL DENSITIES WOLIF CREEK I ' UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-9 REV. 14 EVAPORATION HEAT LOSS FUEL STORAGE POOL HEAT EXCHANGER r, P L_ ________ T COOLANT WOLIF CREEK L ---------------------------- tJPJDA"f'ED-SAFE'JFY-1\.NA:I.;YSIS -REPORT FUEL STORAGE POOL COOLING MODEL FIGURE 9.1A-10 REV. 14 WOLF CREEK 3-D CFD MODEL WOILF CREEK I UPDATE][) SAFETY ANALYSIS REPOR1' ISOMETRIC VIEW OF FUEL STORAGE POOL CFD MODEL ' FIGURE 9.1A-11 I REV.14 ,.... c: ..J E a. o E ' c.. Cl I' 010 (S)o (S)c..lfl NCDN c CD 0 0 0 ..J 0 c.. ...><(....> 0 0 > >. CD CDC.. C. -o(l) W-' coW _:x_..JCD QJIO Cl 01 u .5 .s 4--1--u -c: 0-::J s 0 0 oCD 0... c: (]) 0 Ol..J o_, L 0 0 +-' <D (/) c.. ..J (])Q :::J LLx . . (S) (S) ( 6 26 ) (S) (S) . (S) . (S) U0)1Q..J9i988\;j (S) (S) LD N (S) (S) ,....... (S) (S) .....--(S) (S)x *
- 0 CD (S) (IJ (S) '--J CD t: --' 1-(S) (S) . (S) (S) LD . (S) -----------------------------
WOLF -cREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-12 REV. 14 01 c: -' E a. o E ' L 0 f'. 010 <Slo NmN -....... c: ()) 0 0 0 -' 0 L ....><(....> 0 0 > >. CD (J)LQ. -o(l) W...> CIIW ..Y-'ill Q)IO Q) L CD 01 U E c:: '+----'-' ol---u 5 c:: -::J 0 0 am 0... c Q) 0 01-' 0...> L 0 2 CD Ul '--' Q)Q ::J lL>-(S) N (S) . (S) (S) (S) . . (S) . <Sl I u o 1 1 g ... re 1 e o o 'if (S) (S) (S) LO N (S) (S) ,....... 0 0 ..---(S)
- 0 CD (S) (J) (S) '-J CD E -' I-(S) (S) . (S) (S) LO . <Sl I I ----------------------------
.-----------------'----, WOLF CREEK UPDATED SAFETY ANALYSIS REIPORT FIGURE 9.1A-13 REV. 14 ('S) """' ('S) 0) c ..J E a. c E ' L 0 " rnO (S} 0 (S} LIR (S} N mN ('S) -'-" c m 0 0 0 ,...... -' 0 L 0 ..J<(..., 0 0 0 > >-m CILO. ......... -oOO W-' IS) wW "t:-'ffi <l!IO ([) L Cl 0)
- U E c 0 '+--' ..J ot--u (])
IS) (/) 0 0 om IS) '-../ (L c ([) 0 (]) O"l-' E O..J L 0 ....J 0 CD 1-+-' UJL ..J <l!O IS) ::J IS) LLN (S} (S} IS) IS) (S)(S) N (S) N (S} . . . (S} (S) IS) (S)" (5 I I I ) u 01 1 g ..J.e 1 eo ov
- L ---------------------------
WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-14 REV. 14 ----------------------------------------------------------------------------------------------------
(S) ,....,. (S) 01 c E a. In o E N ' L 0 r-.. 010 (S) 0 (S) LI:R (S) N CD"f" (S) c (I) 0 0 0 ,...._, 0 l. (S) -'<C .... 0 0 (S) > >. (I) Q)l.Q. '""" ...:.;o(l) w .... (S) wW (S)x _:,[_
- GJIO Q) '--(1)01
- u E c ...... 0 '+----1---u 0 ()) :s_; (S) (/) 0 0 om (S) '-J Q_
- c ()) Q) 0 E o_, ...... '--0 0 +-' (I) I-(f) L GJQ (S) ::::J (S) LLx * (S) (S) L.O (S) (S) (S) .. (S!. (S) (S) (S)(S) M N ...... (S) ...... N M . . . . . . (S) (S) (S) (S) (S) (S) (S) (6 I I % ) u 0 ") 1 g ... i"e 1 e 8 8\:j
- L ___________________________
UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-15 REV. 14
- o fTl < _,. '1 G"l c ;
- o fTl <D )> ' (Jl )> d, ;gl il>, 0, "'I l<liO ":lt"'
il-l(") z'::>:l >1&1 t"',P'I ;;JI >-< " !:Ill --------------------------------------------------------W o If Creek E 1 e v e1 l Lon 2007' FueiStoroge Pool Tlme Hlslory Y Dlrecllon 8oundlng D8E Speolro Domplng) 0.40 .---.. 0) I. II .II J Ill II II II. H e. 20 4 .,_, c 0 * ...J ..._) o -0 . 00 IAJ'\ m 1 (,. Q) Q) 0 -0 . 20 l I I p . 1111 II II I I I 1111 I I*IIUilllll '1111 'I I I ' I I I -0 . 40 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 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 0 . ee s00. ee 1 00e. ee 1 see. e0 20ee . 00 2se0 . 00 T l me ( sec
- X *1 00) QL_, _____ _ ::>:l "' -l
r-------. Ol,-------,-------,;-------.------:----riS) E a. D E ' L 0 I' 010 <Slo N em 1-o 0 0 0 L ...><( .... D 0 > >. CD me.. a. ..:.o(/) W...> mW (])IO (]) '-m OJ U E c 4-1--"C 0 c 5-::l 0 0 om 0... c (]) 0 0...> L 0 o m _,_, Ul L (])0 ::J LLN . <S) (S) N (S) (S) (S) IS) I (S) N . (S) I U011Q..J9!988\j IS) lSI N WOLF
- I-UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-16 REV. 14
... NORTII 351 362 ]77 J78 J9Q -.401 402 41] 414 m CD 0 (] CD 366 )71 383 395 OJ l HI J6& 310 lSI 382 JQ] 194 405 406 417 418 <>> <<D (j) @ m 11 *m n ill m m m ffi m <k J @ u @ u @ u @ Ll @ L v I "3 474 481 482 489 *90 497 41111 50S SD& WOLF CREEK FUEL STORAGE POOL MODEl (IJLJ!OINATE RACK IHPACT SPRING NUMBERING SCHEME (BOTTOM! AXES WOLF CREEK UPDATED SAJFETY ANALYSIS RtPORT -------------------------------------------- FIGURE 9 .1A-17 REV. 14 .... NORTH 361 ))4 m 380 1111 m .&Ol -404 415 418 3B7r--------,376 3118 400
- mr--------,,.,.
CD 0 G) 0 31111 liS lB7 l99 II ll 37J 174 lB5 3BB "R1 l99 409 410 <11:?1 4i!l m *oo 0 @ Q 43 51 *n m ill
- 471 480 4BB 496 5()4 512 @ @ @ @ <72 79 87 951 Js.,; 511 y m m ffi soo Lx HIDEL COORDINATE AXES WOLF CREEK FUEL STORAGE POOL RACK IHPACT SPRING NUYBERING SCHEHE !TOP I WOLF' CREEK _j_ UPDATED SAF'ETY ANALYSIS REPORT L ________________________________________
_ ' FIGURE 9.1A-18 REV. 14 .... NORTII 19l 194 2119 210 221 222 2ll 234 245 246 J91'r--------2!M 216 11B 2-lo*r--------151 CD 0 G) 0 198 Dl 115 m m 151 101 101 Ill 114 115 216 ll7 llB 249 150 251 264 172 0 ISB ltil <71 261 262 1&11 270 277 194 y 178 IBl IBI 181 MODEL COO I NATE AXES WOLF CREEK HALF-FULL FUEL STORAGE POOL BOTTOM SPRINGS AT BASEPLATE WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-19 REV. 14 T .... MORTI! 195 1!111 211 212 m m z:r; 23& 2<7 2<a 199**------- 21111 220 2l2 2<4 256 CD cv G) 0 :roo zo> 19l lnd l>o1 z.;, 205 lOll 2ll 21B 22B 210 2'1 z.42 253 254 259 2138 216 (j) m 285 268 27] 274 2m 3m 290 297 y L 2BS 2ll6 Lx HIJOEL ([)[]liD! NATE AXES WOLF CREEK HALF-FULL FUEL STORAGE POOL TOP SPRINGS WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.1A-20 REV. 14 ' L1 KJJEL CIIJ2DINA1E AXES ---------------- ..... NORTH 73 74 77 B4 CD 78 BJ Bl B2 101 102 89 96 .---------...,108 0 G) oo 1m 9] 94 105 106 WOLF CREEK CASK LOADING POOL BOTTOM SPRINGS WOLF' CREEK UPDATED SAF'ETY ANALYSIS REPORT --------------------------------------------
______ ] FIGURE 9.1A-21 REV. 14 y Lz 14DEL CIIRJINATE AXES 75 76 N CD so 87 85 86 103 104 91 IQQr---------,112 0 G) 92 I lgg Ill 97 98 lOll 110 WOLF CREEK CASK LOADING POOL TOP SPRINGS WOLIF CREEK UPDATED SAFETY ANALYSIS REPORT!
FIGURE 9.1A-22 REV. 14 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT --------------------------- _j--,:-=--::-:=-=--=--=-==-=-=-==-=-=-==-=--=-=-=,....,...,=-=--=-:l ISOMETRIC VIEW OF THE FUEL STORAGE POOL AREA FIGURE 9.1A-23 REV. 14 -a...._ NORTH . .. 22' *8" ....,..... 20' 0" 20 -o 20' *0" ---r---: . I rn= -I--T T .O.C. IL. 20 * *0" FUlL TUIISFE C!HAL I -,-t L 3'-D" -3'-0"'-' I I I FLOOI LIIBI PLAT IL. 2012' *2", I '1 I 87'-8" i;> , CASK IASHDDIN P T ., I FLOOR EL. 2020' D"' I
- I-I I I I :II _ 2* -0" 1101 I 3* -0" -l..-.1 I I IALL I I I I, l k l I;.,. I I I I I bl!UEL STORAGE POOL I I Fs ----'-
L.
- rr ._.... j. CASK LOADING T,oL _j I I i' I . ;. 1 lp C FL. LINEI PL TE 3' *0" io *' :. EL. 2001' *0 , --, I I N I " ;.I 16' -o" ID' -D" s* -tl L i::i I I I I __, f-. I'-D" 1. 1-1-3'-D" I I 1i
- j j j I ------1'*8" TIP. 3'*0" TYP. I !:L, .. , WOLF CREEK -------------------------------------------_/_.UPDA-TED -ANALYSIS-REPOR-T-PLAN VIEW AND DIMENSIONS OF THE FUEL STORAGE POOL AREA FIGURE 9.1A-24 REV. 14 2.1 j 1.9 I ,1.75 ::;. i 1.55 11.35 1.15 0.95. 0 fl (\ II 1\ *VV'v v v v \f\ v -* 5 : 10 15 20 Simulation Time Step (Seconds)
Plot of Gap Between Racks 13 and 14 at Sprtng No. 4961n Full SFP Model WOLF CREEK v 25 UPDATED SAFETY ANAILYSIS REPORT -------------------------------------------- '-------fiGt::JRE---g-.tfl.-2'5' -------I-* REV. 14 1.9 1.8 1.7 11.6 1.5 1.3 f'l f\v. -"'I\ -v\ IN v v
1.2 1.1 1 0 5 10 15 20 25 SimUlation Time Step (Seconds)
Plot of Gap 14 and.15 at Spring No. 504 in FuJI SFP Madel WOLF CREEK I UPDATED SAFETY ANALYSIS REPORT --------------------------------------------*------------------- FIGURE 9.1A-26 REV. 14 -------------------
,-------------------
1.1 I 1.05 0.95 -"' 0.85 :;:. I 0.75 lo.e5 i5 0.55 0.45 0.35. 0 .-. fi V' u I' {\ I . -v J --------5 1 o* 15 20 25 Simulation Time Step (Seconds) Plot of Gap Between Rack 14 and the wall atSpr1ng No. 5011n Full SFP Modal WOLF CREEK UIPDATED SAFETY ANALYSIS REIPqRT REV. 14 WOLF CREEK
9.2 WATER
SYSTEMS
9.2.1 STATION
SERVICE WATER SYSTEM The station service water system consists of the service water system (SWS) and the essential service water system (ESWS). The SWS is used during normal operating and normal shutdown conditions. The ESWS is used during normal shutdown conditions and abnormal conditions, such as loss of offsite power and a LOCA. In addition to the following descriptions of the SWS and the ESWS, refer to Sections 9.2.1.1 and 9.2.1.2, respectively. 9.2.1.1 Service Water System The SWS is a nonsafety-related system which provides a source of heat rejection for plant auxiliaries which require cooling during normal plant operation and
normal plant shutdown. The system also supplies cooling water to the safety-related ESWS during normal operation. The heated service water is discharged to the circulating water system and to the Ultimate Heat Sink. 9.2.1.1.1 Design Bases 9.2.1.1.1.1 Safety Design Bases The SWS serves no safety-related function. 9.2.1.1.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The SWS provides sufficient cooling water for the heat removal from nonessential auxiliary plant equipment and from the ESWS over the full range of the normal reactor operation and normal shutdown. 9.2.1.1.2 System Description 9.2.1.1.2.1 General Description The SWS consists of piping, valves, pumps, strainers, instrumentation, and traveling water screens, as shown in Figures 9.2-1 and 10.4-1. The pumps, strainers and traveling water screens are located at the Circulating Water Screenhouse (CWSH). The SWS consists of three one-half capacity service water pumps and one low flow and startup pump, traveling screens and automatic
backwash strainers, all located in the screenhouse. During normal plant
operation, the SWS supplies cooling water to the turbine plant auxiliary
equipment, steam generator blowdown nonregenerative heat exchanger, and CVCS chiller, as well as components served by the ESWS. The service water system is the normal water supply for the Demineralized Water Makeup System. The components cooled by the SWS and their respective SWS flow rates and heat loads are given in Table 9.2-1. 9.2-1 Rev. 19 WOLF CREEK 9.2.1.1.2.2 Component Description The majority of the SWS piping and valves are carbon steel. Some portions may be stainless steel. All piping and valves are designed to meet the requirements of ANSI B31.1. The design ratings of the SWS supply lines in the
power block are 200 psig and 150 F, and discharge lines to the circulating
water system are 85 psig and 150 F. Trash is removed from the circulating water influent by traveling water screens operated as per system operating procedures. The traveling water screens can
be rotated and backwashed, manually or automatically, due to differential
pressure across the screens. Debris is automatically deposited in a basket for
periodic removal by plant personnel. The plant service water return discharges into the circulating water discharge. This discharge is directed to the station cooling lake. Each service water pump is sized to deliver 25,000 gpm of service water at a discharge pressure of approximately 185 feet. To prevent organic fouling of the system, chlorination or liquid or solid, oxidizing or non-oxidizing biocide
addition of the service water is performed on an intermittent basis at the
CWSH. Cathodic protection of the underground SWS piping is provided to
minimize long term corrosion problems from the soil. Connections are provided for localized chemical treatment of normally low flow and stagnant lines. The CWSH and the SWS pumps are designed to accommodate the expected range of lake elevations. Freeze protection of the screenhouse inlet from the cooling
lake is accomplished by a warming line from the circulating water discharge. 9.2.1.1.2.3 System Operation Upon loss of offsite power or the receipt of a SIS, the system is isolated from the ESWS, as described in Section 9.2.1.2. The SWS is designed to operate satisfactorily for all normal plant operating conditions throughout the range of cooling lake levels from 1075.5 feet m.s.l. minimum to 1095 feet m.s.l. maximum. The SWS piping is filled with water and pressurized, with at least one pump operating at all times. The system is controlled by manually energizing or de-
energizing any of the service water pumps and/or the low flow pump. The number of pumps in operation depends on the SWS header pressure, which is displayed on
the Main Control Board (MCB), CWSH standby control panel and locally. The SWS header pressure downstream of the strainers is maintained at a minimum of 80 psig. This provides the required minimum pressure to the power block interface and prevents pump run-out (excessive flow). If pressure falls below this
value, the event is alarmed on the MCB and CWSH standby control panels, and
recorded by the plant computer. 9.2-2 Rev. 18 WOLF CREEK 9.2.1.1.3 Safety Evaluation The SWS is not required for the safe shutdown of the plant. Equipment essential to the safe shutdown of the plant can be cooled by the ESWS and the component cooling water system. 9.2.1.1.4 Test and Inspection The SWS is designed to permit periodic inspection and testing of system components to ensure that the design bases are met. The system is designed to
permit periodic functional testing of components that are not in normal use.
The service water pumps and strainers are proven operable by their use during
normal station operation. The electropotential between the subsurface SWS piping and the ground is measured periodically to verify the effectiveness of the cathodic protection system. Preoperational testing is described in Chapter 14.0. The performance and structural and leaktight integrity of all cooling water system components is demonstrated by continuous operation. 9.2.1.1.5 Instrumentation Applications
The SWS instrumentation is designed to facilitate automatic operation, remote control, and continuous indication of system parameters. Local pressure test points and temperature indicators are provided at various components which are served by the SWS to allow checking of system performance.
Control valves are provided to control water flow where necessary. The pump discharge header pressure and header pressure downstream of the strainers are indicated in the control room. Pressure downstream of the strainers is alarmed
on low pressure. 9.2.1.2 Essential Service Water System The ESWS removes heat from plant components which require cooling for post fire or post accident safe shutdown of the reactor or following a DBA. The ESWS also provides emergency makeup to the fuel storage pool and component cooling water systems, and is the backup water supply to the auxiliary feedwater system. The ESWS consists of two redundant cooling water trains. The ESWS does not directly interface with radioactive systems. 9.2-3 Rev. 19 WOLF CREEK 9.2.1.2.1 Design Bases 9.2.1.2.1.1 Safety Design Basis The ESWS is safety-related, is required to function following a DBA, and is required to achieve and maintain the plant in a post accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The ESWS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2). SAFETY DESIGN BASIS TWO - The ESWS is designed to remain functional after an SSE and to perform its intended function following the postulated hazards of fire, internal missile, or pipe break. Failure of any adjacent non-seismic
Category I structure will not constitute a hazard to the ESWS (GDC-3 or 4). SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC-44). Components of this system are not shared with other units (GDC-5). 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-45 and 46). SAFETY DESIGN BASIS FIVE - The ESWS 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 to isolate components or piping is provided so that the ESWS's safety function are not compromised. This includes isolation of components to deal with leakage or malfunctions and to isolate nonsafety-related portions of the ESWS (GDC-44). 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 C testing. SAFETY DESIGN BASIS EIGHT - The ESWS is designed to remove heat from components important to mitigating the consequences of a LOCA or MSLB and to transfer the heat to the ultimate heat sink (GDC-44). 9.2-4 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS NINE - The ESWS operates in conjunction with the component cooling water and other reactor auxiliary components and the ultimate heat sink
to provide a means to cool the reactor core and RCS to achieve and maintain a safe shutdown.
SAFETY DESIGN BASIS TEN - The ESWS provides emergency makeup to the fuel
storage pool and component cooling water systems, and is the backup water
supply to the auxiliary feedwater system. SAFETY DESIGN BASIS ELEVEN - The ESWS is protected from long term organic
fouling and corrosion problems.
9.2.1.2.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The ESWS provides sufficient cooling water
for removing heat from essential plant equipment over the full range of the
normal reactor operation.
9.2.1.2.2 System Description
9.2.1.2.2.1 General Description
The ESWS is shown in Figure 9.2-2 and consists of two separate 100-percent capacity trains of traveling screen pumps, pump prelube storage tanks, self cleaning strainers, piping, valves, and instrumentation. One pump supplies
cooling water to each flow path. Each flow path is fed from the ultimate heat
sink(refer to Sections 9.2.5 and 3.8.4.1.9.). The essential service water
pumps draw water from the ultimate heat sink at a maximum temperature of 95øF and a minimum temperature of 32øF. Each train of the ESWS serves through the associated train of safety-related components. Each train of the ESWS is
interconnected with the SWS. Two motor-operated isolation valves are provided
in each crosstie header where it connects to the SWS. In addition, cooling
water flow is maintained following a DBA to a nonsafety-related air compressor and associated after-cooler. The air compressor is automatically isolated on high flow (indicative of leakage) or it can be remote manually isolated. The
ESWS pumphouse equipment locations are shown in Figures 9.2-3 and 9.2-4. To
prevent organic fouling of the ESW, chemical addition of liquid microbial
control agent is performed during system operation at the ESW pump house. Connections are provided for periodic localized chemical treatment in lines which are normally stagnant or receive low flows.
The water chemistry of the ESWS fluid is given in Table 9.2-17. The carbon steel piping in the ESWS is designed with a corrosion tolerance and the use of piping inspection gauging (PIGing) hardware to assure that there is no long-term degradation of the system. The sections of the ESWS piping that cannot be inspected with PIGing hardware are stainless steel to prevent corrosion through the piping walls and assure that there is no long-term degradation of the system. The components cooled by or supplied with makeup water from the ESWS and their respective heat loads and flow rates are given in Tables 9.2-2 through 9.2-4. The basis for the heat loads and flow rates is given in Tables 9.2-2 through 9.2-4.
9.2-5 Rev. 28 WOLF CREEK The ESWS normally supplies water at a higher pressure than the cooled safety-related component. Therefore, if leakage occurs it is into the system being
cooled or, in the case of ESW piping and valves, in the floor drain system described in Section 9.3.3. Once a significant leak is found, an affected item is isolated and repaired.
The essential service water pumps are located in a seismic Category I pumphouse (refer to Section 3.8.4.1.8). Each flow path is protected by interior walls from internally generated missiles, jet impingement, and flooding that may result from failures in adjacent flow path piping. Equipment protection from
high winds and floods is discussed in Sections 3.3 and 3.4, respectively.
Tornado missile protection is discussed in Section 3.5.2.
To prevent the potentially damaging effects of column closure water hammer to the ESWS, vertical loop piping with high point vacuum breakers and check valves in the service water cross tie are used to reduce the potential for drain down of the ESWS. The vertical loop piping is part of the ESW return piping. 9.2.1.2.2.2 Component Description
Codes and standards applicable to the ESWS are listed in Table 3.2-1. The ESWS
is designed and constructed in accordance with the following quality group
requirements: Containment penetrations are quality group B, the separate and redundant cooling loops for safety-related equipment are quality group C, and lines to other nonessential equipment are quality group D. The quality group B
and C portions are seismic Category I. Design data for the ESWS pumps, prelube
storage tanks, self-cleaning strainers, piping and valves is provided in Table
9.2-5. ESSENTIAL SERVICE WATER PUMPS - The two essential service water pumps each have
a capacity of 100 percent of the flowrate required during normal operation.
These designs exceed the required accident condition flowrate. Pumps are sized
to include an additional wear margin on the flow at the design head to accommodate normal degradation of performance due to impeller wear. The pumps are of the vertical centrifugal type. The pump manufacturer performed a full
range performance test at the minimum submergence of 8 feet and demonstrated no
vortexing or cavitation during the test. These pumps take their suction from
the ultimate heat sink. ESSENTIAL SERVICE WATER PUMP PRELUBE STORAGE TANKS - Each pump is provided with
a prelube storage tank. The tank supplies the pump lineshaft bearings with
water to prevent the bearings above the pit water level from running dry during
startup. Tank size is based on supplying a minimum of a five-minute supply of water, at six gpm, with no makeup from the pump discharge line. When the pump is operating, the bearings are lubricated by the pumped fluid.
9.2-6 Rev. 29 WOLF CREEK ESSENTIAL SERVICE WATER SELF CLEANING STRAINERS - One self cleaning strainer is provided for each essential service water flow path. One hundred percent of
the essential service water flow is filtered through 1/16-inch slotted openings in the strainer element. On high differential pressure, the strainer element is automatically backwashed to eject the accumulated debris.
ESSENTIAL SERVICE WATER TRAVELING WATER SCREENS - Two traveling water screens
are provided, one per train of the ESWS. Screens are designed for continuous operation to protect the essential service water pump suction from large debris. Screen spray water is provided from the essential service water supply
header downstream of the self-cleaning strainer.
AUXILIARY HEAT EXCHANGERS - Tables 9.2-2 through 9.2-4 list the various components in the ESWS and their heat loads and flow requirements. In general, essential service water flows through the tube side, and the cooled fluid flows
through the shell side. Further description of these items is included in the
referenced sections.
PIPING AND VALVES - Piping within the standard power block is carbon steel except for portions replaced by stainless steel due to erosion or corrosion.
Piping to and from the ultimate heat sink and yard piping is carbon steel.
Portions of the ESWS that cannot be inspected with PIGing hardware are
stainless steel. The design condition for the supply water is 200 psi and 100 °F. The design condition for the return line downstream of the last isolation valve and upstream of flow orifice EF-FO-00021/22 is 130 psi and 200 °F. The
design condition between orifices EF-FO-00021/22 and EF-FO-00023/24 is 95 psi
and 200 °F. Downstream of flow orifices EF-FO-00023/24, the design condition
is 65 psi and 200 °F. Two entirely separate redundant lines are provided and designed to ASME Section III, Class 3, except for containment penetrations which are designed to ASME Section III, Class 2. Nonsafety-related portions of
the system are designed to ANSI B31.1.
For the components located inside the containment, supply and return lines are provided with containment isolation valves, as described in Section 6.2.4.
Power-operated valves are provided to permit isolation of nonsafety-related or
nonessential service following a DBA.
To mitigate the adverse effects of column closure water hammer following a LOOP, check valves are installed in the service water cross tie connection.
9.2.1.2.2.3 System Operation
POWER GENERATION OPERATION - During normal plant operations, the ESW within the standard power block receives water from the SWS and supplies water to the
safety-related components and air compressors. After cooling the equipment, the heated water is returned to the SWS and/or to the Ultimate Heat Sink.
Manual bypass valves are provided around the main outlet isolation valve to the component cooling water heat exchangers. During normal operation, these valves are adjusted for proper flow for safety functions and locked into position, and
the main outlet isolation valves are normally closed but can be throttled as
needed.
9.2-7 Rev. 29 WOLF CREEK The normal makeup water to the fuel storage pool and component cooling water
system is from other plant sources, and the ESWS is only used if the other systems are unable to supply water.
PLANT COOLDOWN AND SHUTDOWN - No changes to the valving arrangement are
required from the normal operation to initiate cooldown of the plant. During
the cold shutdown condition, various components may be isolated if no heat loads are generated. The source of water is normally from the SWS; however, if offsite power is not available the Class IE ESW pumps will provide the water
source.
EMERGENCY OPERATION - Following a DBA, auxiliary feedwater low suction pressure, or loss of offsite power, or safety injection signal, the ESWS is isolated from the SWS by closing the associated motor-operated isolation valves Following a LOOP, the ESW/SW cross tie check valves reduce the potential for drain down of the ESWS. Both essential service water pumps are automatically started by the emergency diesel load sequencer, and receive power from the preferred power supply or the emergency diesel generators. Pump A starts 20 seconds and pump B 25 seconds after receipt of the SIS. The pumps start immediately upon receipt of auxiliary feedwater low suction pressure. Pump A
starts 32 seconds and pump B starts 37 seconds after receipt of loss of offsite
power signal (Loop). The pumps supply cooling water from the ultimate heat sink to the safety-related components and air compressors. The main isolation valves on the component cooling water heat exchanger outlet are automatically
closed, to decrease the cooling water flow rate as dictated by the service
requirements. After cooling the equipment, the heated water is returned to the
ultimate heat sink. As described in Section 10.4.9, the ESWS will automatically supply water to the
auxiliary feedwater system in the event condensate storage tank water is
unavailable. In addition, the ESWS provides emergency suction supply to the
auxiliary feedwater system (AFS) upon receipt of an engineered safety features actuation signal (ESFAS) initiated by an AFS low suction pressure signal. The low AFS suction pressure signal also provides the open signal to the AFS pump
suction supply valves and closes the ESWS/SWS supply isolation valves located
at the power block inlet. This assures essential service water supply to the
AFS following a SSE without an accompanying accident or loss of offsite power. During this event, operator action is required to realign the ESWS to the ultimate heat sink. Operational procedures ensure realignment of the ESWS to
the ultimate heat sink. An allowance of 5 acre-feet loss of UHS volume, equivalent to about 1 hour at maximum ESWS flow, is provided for operator
alignment to be performed.
9.2-8 Rev. 29 WOLF CREEK The traveling water screens automatically start with the same signals that start the essential service water pumps. The screens protect the essential
service water pumps from large debris. Water is sprayed on the screens to clean debris that may collect on the screens.
The prelube storage tank is continuously supplied with water by a connection on
the essential service water pump discharge, downstream of the check valve and
self-cleaning strainer. Tank level is automatically maintained by action of the supply line to the pump lineshaft bearings, stuffing box, and the open tank overflow, and by the manual setting of the globe valve in the tank supply line.
The tank provides water to the lineshaft bearings and stuffing box continuously
during periods when the essential service water pumps are idle. The discharge
lines are normally pressurized by the service water pumps. When the essential service water pump is running, flow in the supply line from the tank reverses and discharges through the overflow. If an undetected failure of the SW pumps
is assumed, this could result in loss of prelube supply prior to operator
action to start the ESW pump. However, the pump will start and continue to run
satisfactorily in an emergency situation with dry bearings. Bronze lineshaft bearings are provided in the ESW pumps because of this possibility. The alternate bearing material (cutless rubber) would have a greater tendency to
seize during this transient.
The self-cleaning strainers filter the supply water to the power block. High differential pressure caused by accumulated debris on the strainer element is corrected automatically by backwashing the element to the ultimate heat sink.
Backwash of the filters may also be initiated manually.
Trash is removed from the essential service water influent by traveling water screens operated as per system operating procedures. The essential service water screens can also be rotated and backwashed manually or automatically. In
automatic, the essential service water screens are cleaned whenever the
essential service water pumps are running. Any debris on the screens is washed
back into the Wolf Creek Lake. Freeze protection for the ESWS intake structure is provided by a warming line
which branches from each essential service water return line. Each warming
line can divert a minimum of 4,200 gpm from 13,755 gpm of returning flow, to
warm a pump suction flow rate of 15,000 gpm. These flow parameters, in conjunction with a minimum ESW system temperature rise of 0.72F, provided by operating safety related equipment, ensures that frazil ice will not block the
trash racks. Following an automatic ESW pump start due to an SIS or loss of
off site power, sufficient safety related equipment is automatically started to
ensure that the minimum required heat load is available. Following an automatic ESW pump start due to auxiliary feedwater low suction pressure, sufficient heat load is ensured by operational procedures that require
containment coolers to be operating when in Modes 3 and above. Operational
procedures ensure that the minimum heat load is present prior to manual ESW
pump starts. Table 9.2-25 summarizes the minimum ESW heat loads available in
terms of equivalent temperature rises at design ESW flow rates for different modes of plant operation. The warming line flow rates will vary as the ESWS returning flow rates vary, and the warming lines are active during ESWS and
normal service water system operation. The warming lines will not be in
service when the lake is warm to avoid exceeding the ESWS supply design
temperature. The adverse effects of column closure water hammer, following a loss of offsite power, are mitigated in the ESWS by the installation of the vertical pipe loops on both trains of the return piping. In the event of a loss of offsite power vacuum breakers at the high point of the vertical loops mitigate the formation 9.2-9 Rev. 29 WOLF CREEK of column separation in the system by allowing atmospheric air to enter the return piping. Additionally check valves in the service water cross-tie piping close when service water flow is stopped. 9.2.1.2.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases. SAFETY EVALUATION ONE - Except for the buried piping between the ESWS pumphouse and the power block, the safety-related portions of the ESWS are located in the
reactor, auxiliary, control, diesel, and essential service water pump
pumphouse, and ESW vertical loop chase 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 buried piping is also designed to withstand
these natural phenomena, as described in Section 9.2.1.2. Each ESWS return
line to the UHS has a warming line branch. The warming lines are used to prevent ice from blocking the pump house intake structure as described in section 9.2.1.2. Each ESWS return line in the power block has a vertical loop. The vertical loops are used to mitigate the adverse effects of column closure water hammer as described in 9.2.1.2.2.3. SAFETY EVALUATION TWO - The safety-related portions of the ESWS 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 post-accident safe shutdown, as outlined in
Section 7.4, can be achieved and maintained. SAFETY EVALUATION THREE - The ESWS services two identical trains of engineered
safety features equipment which are required for safe shutdown of the plant.
Only one train of the redundant plant components is required for the safe
shutdown of the plant after any postulated accident condition. Water is supplied to each train of components by a separate pump and header. No single failure will compromise the system's safety functions.
Both essential service water trains are capable of supplying the required
cooling water flows to meet the single failure criterion. All vital power can be supplied from either onsite or offsite power systems as described in Chapter
- 8.
The single active failure analysis is presented in Table 9.2-6. This provides
the basis for the technical specifications with regard to limiting conditions for operation and surveillance.
SAFETY EVALUATION FOUR - The ESWS is initially tested in accordance with the
program given in Chapter 14.0. Periodic inservice functional testing is done
in accordance with Section 9.2.1.2.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for the ESWS.
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 9.2.1.2.2.2 shows that the components meet the design and fabrication codes given in Section 3.2. All the
power supplies and the control function necessary for safe function of the ESWS
are Class IE, as described in Chapters 7.0 and 8.0.
9.2-10 Rev. 29 WOLF CREEK All structures and components of the system are located so that the failure of any non-seismic Category I structure will not constitute a hazard to the ESWS. The location of the ESWS structures and components is such that:
- a. Adequate separation from all non-seismic
Category I structures is provided.
- b. The essential service water lines and Category I electrical duct banks are placed below non-Category I lines at points of
intersection except where they cross the non-
seismic Category I circulating and warming
water lines. In order to prevent possible erosion of the bedding for the duct banks, the non-seismic lines below are surrounded by
seismic Category I reinforced concrete
encasements. In order to support the ESWS piping from the possible erosion of the bedding if the non-seismic Category I circulating water line ruptured, caissons are
used to support the ESWS piping on each side
of the circulating water piping. These
encasements and caissons are described in Section 3.8.4.1.5.
- c. Any hazards to the system from the failure
of man-made structures, such as the failure
of slopes or the postulated rupture of storage tanks are precluded.
The seismic Category I essential service water pumphouse is located 1,895 feet
east and 380 feet south of the centerline of the reactor. The seismic Category
I essential service water piping at the discharge point is located approximately 2,505 feet east and 1,260 feet south of the pumphouse (Refer to Figure 9.2-2).
The ESWS pumphouse is designed to withstand the effects of high winds and wave
forces and is protected from flooding due to high water levels associated with the probable maximum flood in the cooling lake as described in Sections 3.3 and 3.4.
The essential service water pumps are designed to operate throughout the range
of expected water levels in the UHS from 1068 to 1095' - 3" m.s.l. SAFETY EVALUATION SIX - Section 9.2.1.2.2.1 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 - The minimum flow rates required to remove heat from
the containment and necessary safety-related components from a postulated LOCA
or MSLB and dissipate it to the ultimate heat sink are listed in Table 9.2-3. Determination of these minimum flows is based on a maximum ultimate heat sink temperature of 95 F, as determined in Section 9.2.5. The ESWS design assures
that the flow requirements are met by operation of an ESWS pump and proper
realignment of the valves to the accident configuration.
9.2-11 Rev. 29 WOLF CREEK Each train of the ESWS and each train of the safety-related systems served by the ESWS are 100 percent redundant. This arrangement ensures that the full-heat dissipating capacity is available following an accident and an assumed single failure. SAFETY EVALUATION NINE - The minimum ESWS flow rate required to remove decay heat from the RCS and other necessary components to achieve and maintain post accident safe shutdown is listed in Table 9.2-4. The flow rates are based on an ultimate heat sink temperature of 95 F. The ESWS design assures that the flow requirements are met by operation of an ESWS pump and proper realignment of the valves to the accident configuration. SAFETY EVALUATION TEN - The ESWS is capable of supplying emergency makeup to the fuel storage pool and component cooling water systems and the backup water
to the auxiliary feedwater system. The values are listed in Table 9.2-3. The ESWS design assures that the flow requirements are met by operation of an ESWS pump in each train and proper realignment of the associated valves. SAFETY EVALUATION ELEVEN - The portions of the ESWS which are sensitive to organic fouling, such as heat exchangers, are served the majority of the time by the SWS which is chemically treated. The chemical treatment of the service water mitigates the detrimental effects of organic fouling on the ESWS. The ESWS chemical addition system can be operated during ESW pump tests to provide long term protection to the ESW piping which has very low flow during normal
plant operation. During periods of long duration of operation, daily chemical
injection rate will be used to treat the overall ESWS. Cathodic protection of the underground ESWS piping is provided to minimize long term corrosion problems with the soil. 9.2.1.2.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. The performance and structural and leaktight integrity of all cooling water system components is
demonstrated by continuous operation. The ESWS is testable through the full operational sequence that brings the system into operation for reactor shutdown and for LOCAs, including operation of applicable portions of the protection system and the transfer between normal and standby power sources. An analog channel operational test of the differential pressure instrumentation for automatic isolation of ESW to the air compressors is performed every 184 days. A channel calibration of the differential pressure instrumentation for automatic isolation of ESW to the air compressors is performed every 18 months during shutdown. The safety-related components of the ESWS, i.e., pumps, valves, heat exchangers, and piping (to the extent practicable), are designed and located to permit preservice and inservice inspections. The electropotential between the subsurface ESWS piping and the ground is measured periodically to verify the effectiveness of the cathodic protection system. 9.2-12 Rev. 20 WOLF CREEK 9.2.1.2.5 Instrumentation Applications The ESWS instrumentation, as described on Tables 9.2-7 and 9.2-8, is designed to facilitate automatic operation and remote control of the system and to
provide continuous indication of system parameters. Redundant controls are
provided to initiate the start of the ESWS and to isolate it from the SWS upon
receipt of an SIS and/or loss of offsite power. Redundant and independent power supplies for pump controls and instrumentation are provided from Class IE busses. Refer to Chapter 8.0. Indicating and alarm devices for the system are provided in Tables 9.2-7 and 9.2-8. Thermowells and pressure indicator connections are provided where required for testing and balancing the system. Portable ultrasonic flow indicators are utilized for initial balancing of the flows in the system and for verifying
flows during plant operation.
9.2.2 COOLING
SYSTEM FOR REACTOR AUXILIARIES The cooling system for the reactor auxiliaries is the component cooling water system (CCWS). The CCWS provides cooling water to selected essential and non-
essential components during normal plant operation, including shutdown, and also provides cooling water to several engineered safety feature systems (ESFS) during a LOCA or MSLB. During an emergency cold shutdown CCWS provides cooling to essential components in containment. This system is a closed loop system
which serves as an intermediate barrier between the SWS or ESWS and potentially
radioactive systems in order to eliminate the possibility of an uncontrolled
release of radioactivity. 9.2.2.1 Design Bases 9.2.2.1.1 Safety Design Basis Portions of the CCWS are safety related and are required to function following a DBA and to achieve and maintain the plant in a post-accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The CCWS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The CCWS is designed to remain functional after a SSE and to perform its intended function following the postulated hazards of 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-44).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-45 and 46). 9.2-13 Rev. 14 WOLF CREEK SAFETY DESIGN BASIS FIVE - The CCWS 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 to isolate components or piping is
provided so that the CCWS's safety function is not compromised. This includes
isolation of components to deal with leakage or malfunctions and to isolate nonsafety-related portions of the system (GDC-44).
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 C testing. SAFETY DESIGN BASIS EIGHT - The CCWS is designed to remove heat from components
important to mitigating the consequences of a LOCA or MSLB and to transfer the
heat to the essential service water system (GDC-44).
SAFETY DESIGN BASIS NINE - The CCWS, operating in conjunction with the RHR, chemical and volume control systems, and the water systems, provides a means to
cool the reactor core and primary systems to achieve and maintain post accident
safe shutdown.
SAFETY DESIGN BASIS TEN - The CCWS, in conjunction with the Essential Service Water System (ESWS), provides sufficient heat energy to maintain the ESWS inlet
trash racks from being blocked with frazil ice.
9.2.2.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The CCWS provides a continuous supply of
cooling water during plant power generation operation to those auxiliary
components in which the potential for radioactive leakage exists.
POWER GENERATION DESIGN BASIS TWO - The system is designed to allow for pump maintenance without interruption of the cooling function.
9.2.2.2 System Description 9.2.2.2.1 General Description
The CCWS is shown in Figure 9.2-15. The system consists essentially of two
separate 100-percent-capacity trains which serve engineered safety features, and includes a loop, common to both trains. The common loop consist of the non-essential radwaste and essential containment loads. The radwaste isolation valves automatically close upon low level in the surge tank, SIS, or high flow
and the containment loads are isolated after a CISB. The safety related containment loop services essential loads such as the excess letdown heat
exchanger (Section 7.4) and the RCP thermal barrier cooling coils (section
5.4.1).
9.2-14 Rev. 29 WOLF CREEK The major components of the system are two component cooling water heat exchangers, four CCWS pumps, two surge tanks, a chemical addition tank, piping, valves, controls, and instruments. The components cooled by the CCWS and their respective CCWS heat loads and flow rates are given in Tables 9.2-9 through 9.2-11. The basis for the heat loads and flow rates is given in the appropriate section referenced in Tables 9.2-9 through 9.2-11.
9.2.2.2.2 Component Description Codes and standards applicable to the CCWS are listed in Tables 3.2-1 and 9.2-
- 12. The component cooling water system is designed and constructed in
accordance with the following quality group requirements: Containment
penetrations are quality group B, the separate and redundant cooling loops for engineered safety features equipment are quality group C, and lines to other equipment are quality groups C and D. The quality group B and C portions are
seismic Category I.
COMPONENT COOLING WATER HEAT EXCHANGERS - Component cooling water heat exchangers are of the horizontal shell and straight tube type. The tube side is supplied with water from the service water or the ESWS, and the shell side
is supplied with water from the discharge of the component cooling water pump.
The overall heat transfer coefficient for the CCW HX during post-LOCA conditions is 192 Btu/hr-ft 2_ F approximately. The post-accident peak containment pressure and temperature was determined by assuming a constant CCW inlet temperature of 130 F to the RHR HX, which envelops the maximum CCW temperature expected during the accident conditions. COMPONENT COOLING WATER PUMPS - Each of the four component cooling water pumps
has a capacity of 100 percent of the flowrate required during normal operation.
This exceeds the required accident condition flowrate. The pumps are of the
horizontal, centrifugal type. Pumps are sized to include an additional 5-percent margin on the flow at the design head to accommodate normal degradation of performance due to impeller wear.
Flooded suction is ensured by surge tanks. Each safety-related component
cooling water train has two 100-percent capacity pumps. The two pumps in each component cooling water train are powered by the same Class IE bus; however, only one pump per train is automatically started upon receipt of an SIS or loss of offsite power signal.
The installation of two 100-percent capacity pumps per train is provided only to avoid a shutdown which would otherwise be required by the technical specifications during prolonged maintenance or repair of component cooling
water pump.
SURGE TANKS - One safety-related component cooling water surge tank is provided in each of the two safety-related CCW trains to accommodate volumetric changes in the system due to thermal transients or leakage. Provisions are included
for automatic makeup to the system from the demineralized water storage and
transfer system.
CHEMICAL ADDITION TANK - One nonsafety-related CCWS chemical addition tank is provided with connections to both of the surge tanks. Provisions for
demineralized makeup water and addition of the chemicals are included.
9.2-15 Rev. 25 WOLF CREEK AUXILIARY HEAT EXCHANGERS - Tables 9.2-9 through 9.2-11 list the various components in the CCWS and their heat loads and flow requirements. In general, component cooling water flows through the shell side, and the potentially radioactive liquid flows through the tube side. Further description of these items is included in the referenced sections.
PIPING AND VALVES - Piping to and from the CCW heat exchangers is of carbon
steel. Portions of the piping have been replaced with stainless steel to prohibit erosion/corrosion. The valves, which when open permit component cooling water to flow into or return from the nonsafety-related portion of the
CCWS, are designed to fail closed. These valves automatically close upon low
level in the surge tank, SIS, or high flow (indicative of a break in the
nonsafety-related portion of the system). Additionally, a supply header flow restricting orifice and two return header check valves prevent a break in the non-safety related portion of the system from compromising the safety related portion. Component cooling water inlet valves to the RHR heat exchanger are motor operated and fail as is. These valves are automatically opened at the start of the post-LOCA recirculation phase.
For the components located inside the containment, supply and return lines are
provided with containment isolation valves, as described in Section 6.2.4. For the reactor coolant pump thermal barriers, a separate return header arrangement is utilized to isolate primary coolant in-leakage to the component cooling
water system in the event of a rupture in the thermal barrier. This header has
its own containment penetration and automatic isolation valves. A CIS-B
automatically closes the normally open motor-operated containment isolation valves. The normally closed (with power lockout) parallel sets of containment isolation valves may be opened in an emergency by the operator to establish
cooling water to the Reactor Coolant Pumps (RCPs) and the excess letdown heat
exchanger when the normal valve can not be recovered from its closed (single
failure) position. These valves are not credited for operation under LOCA or MSLB conditions as reflected in USAR Section 3.11(B), Table 3.11(B)-3 because; (1) no single failure can disable both means of cooling the RCP seals (CCW
cooling to the thermal barrier or seal injection), (2) the excess letdown flow
path is not required post LOCA or post MSLB and (3) the valves are closed with
power lockout preserving the integrity of their containment penetration. A separate source of emergency makeup is provided from the ESWS to each train
of the CCWS.
9.2.2.2.3 System Operation GENERAL - The entire CCWS is a closed-cycle system; cooling water is
continuously recirculated through the system by the CCWS pumps. Heat is
dissipated from the system by the flow of service water or essential service
water through the tube side of the component cooling water heat exchangers. A
component cooling water heat exchanger shell side bypass arrangement in conjunction with throttling of ESW Valves maintains the minimum CCW temperature at 35°F or above, during cold service water conditions.
Following a Safety Injection Signal, the bypass valve closes, thereby forcing
all the CCW flow through the CCW heat exchanger. These additional flows will cause the CCW temperature to fall below 35°F depending upon the lake water temperature, but will not impact the safety function of the CCW system or
components.
The essential service water and service water systems are described in Section
9.2.1. Supply
of normal makeup water to the system is provided from the demineralized water storage tank. 9.2-16 Rev. 29 WOLF CREEK The water in the CCWS is analyzed periodically for pH, conductivity, and corrosion-inhibitor concentration. The specifications which apply to CCWS
chemistry are given in the WCGS Chemistry Specification Manual. Abnormal chemistry is corrected by eliminating the source and by draining water from the system, as required, and adding makeup water to the surge tank. Periodic
addition of water to the CCWS is required to make up for losses due to system
leakage and sampling. For control of long-term corrosion, corrosion inhibitors
are added, as required, via the chemical addition tank. The level indication on the surge tank and the rate of water addition will alert the operator of any abnormal leakage from the system. In addition, the floor drain system
described in Section 9.3.3 provides additional detection capabilities. Once a
significant leak is found, the affected item will be isolated and repaired.
Radiation monitors are installed to monitor the water in each CCWS train to indicate radioactivity leaking into the system. Alerted to radioactive
inleakage, the operator may identify the leaking component by selective
isolation of heat exchangers and determination of the rate of increase of CCW
radioactivity at the component while the suspect component remains isolated. Once the source is determined, the component may remain isolated until repaired. A high radiation signal will automatically close the surge tank vent
and the makeup water valve.
The excess letdown heat exchanger, Reactor Coolant Drain Tank Heat Exchanger, RCP seal thermal barrier and RCP motor bearing cooler are located inside the containment. Cooling is provided to the excess letdown heat exchanger for
emergency cold shutdown operations or to maximize the normal cooldown rate.
The chemical and volume control system injection path to the RCP seals is a
totally diverse cooling means to the CCW supply. The reactor coolant pump motor bearings are qualified for 10 minutes of operation without cooling water. Pump operation can be terminated or the cooling water can be established within
10 minutes. The parallel-series valve arrangements for the containment
penetrations ensure that CCW can be supplied to the containment, considering
any single active failure. Redundant safety-related indication of CCW flow to components located inside the containment is provided on the MCB. A low flow alarm is also provided in the control room.
POWER GENERATION OPERATION - The system is normally operated with one pump on
one of the safety-related CCWS trains supplying the associated safety-related train and essential containment loads in the common loop. The redundant safety-related CCWS train is isolated from the remainder of the system, and the CCWS pumps in that train are not operated. Should an operating pump fail, a
low-pressure switch in the pump discharge header will start the standby pump in
9.2-17 Rev. 29 WOLF CREEK the same train automatically after a 4-second delay. Also, an interlock is provided to start a CCWS pump when the corresponding centrifugal charging pump
of the same train in the chemical and volume control system is started. In this mode, a minimum CCWS flow of 10,400 gpm exists, with a maximum CCWS supply water temperature to plant components not exceeding 105 F. Cooling water flow
through the fuel pool cooling heat exchangers normally exists as needed during periods when spent fuel is stored in the pool except for a short time (approx. 4 hours) following ECCS switchover to recirculation post-LOCA. Additionally, cooling water flow to the fuel pool cooling heat exchangers is the preferred method during normal operation of assuring that the CCWS pump minimum flow of
3000 gpm is met for each train of the CCWS. The above flowrate and temperature conditions apply during spent fuel storage. In this mode of operation, heat
absorbed by the CCWS is dissipated to the SWS. PLANT COOLDOWN AND SHUTDOWN - During the plant cooldown phase following initiation of normal plant shutdown, if a 20-hour cooldown capability is desired, both CCWS trains are placed in operation since flow through both RHR heat exchangers is required. An additional CCWS pump is started in the train which is supplying the other auxiliary loads, such as those associated with the chemical and volume control system, to assist in providing the necessary cooling flow. The plant may be brought safely to the cold shutdown condition with one RHR heat exchanger in operation, but this would require more than 20 hours.The evaluated conditions for a 20-hour cooldown of the plant use 90F cooling lake water. Under these conditions CCWS equilibrium temperature at the CCWS heat exchanger outlet may rise to 120F, after initiation of the CCW flow to the RHR heat exchanger (approximately 4 hours after shutdown). Under these evaluated conditions, CCW flow to the Fuel Storage Pool Heat Exchanger is
reduced or terminated. Flow is resumed prior to the FSP temperature reaching 170F or approximately 8 hours after shutdown when CCWS duty is low enough to accommodate the fuel storage pool load. If the heat loads are less, colder cooling water is used, or the cooldown occurs later in time than the evaluated conditions, then a reduction in CCW flow to a FSP Heat Exchanger may not be necessary. Cycle-specific decay heat analysis and administrative controls
ensure that the maximum pool thermal loading limit is not exceeded. During periods of plant cold shutdown for maintenance or refueling operations following the plant cooldown phase, one of the trains of the CCWS may be shut
down, and one of the RHR heat exchangers may be taken out of service. At least one CCWS pump and one RHR heat exchanger are required to be in operation during
cold shutdown to remove decay heat from the reactor coolant system. Cooling water flow through the fuel pool cooling heat exchangers is also required to cool spent fuel. In this mode of operation, heat absorbed by the CCWS is
dissipated to the service water system. EMERGENCY OPERATION - Upon receipt of a safety injection signal (SIS), the isolation valves for the nonseismic Category I CCWS loop are closed automatically, and one of the pumps in the non-operating, safety-related CCWS
train is started. Component cooling water flows through normally open valves to
the seal coolers for 9.2-18 Rev. 15 WOLF CREEK the residual heat removal pumps and to the oil coolers for the centrifugal charging pumps and safety injection pumps. Upon a loss of offsite power, any
operating CCWS pump stops, and two pumps, one in each train, start in accordance with the standby diesel generator loading sequence (see Chapter 8.0).At the initiation of the post-LOCA recirculation phase, Component Cooling water flow is automatically initiated through the residual heat removal heat exchangers by the opening of normally closed, motor-operated valves at the inlets to the heat exchangers and closing of outlet valves associated with the
fuel storage pool heat exchanger. This establishes a cooling water flow to remove heat from the containment sump water which is flowing through the
residual heat removal system. After 4 hours, the flow may be resumed to the fuel storage pool heat exchanger. The complete switchover sequence is described in Section 6.3. In the emergency mode of operation, the two cooling trains are normally operated in a parallel
configuration. Remotely operated isolation valves are provided to permit complete separation of the two trains. In the event that an emergency cold shutdown (CSD) must be achieved using only safety-related systems and components, the CCW system can withstand a single
active failure and still remove heat from components important to achieving and maintaining post accident safe shutdown. These components include the reactor coolant pump thermal barrier and motor bearings, seal water return heat exchanger, excess letdown heat exchanger, RHR heat exchanger, and fuel storage
pool cooling heat exchanger. 9.2.2.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.2.2.1. SAFETY EVALUATION ONE - The safety-related portions of the CCWS are located in the reactor, auxiliary, and fuel 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 CCW 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained. 9.2-19 Rev. 19 WOLF CREEK SAFETY EVALUATION THREE - The CCWS is completely redundant and, as indicated by Table 9.2-13, 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 CCWS is initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.2.2.4. Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for the CCW 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 systems. Table 9.2-10 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 CCW system
are Class IE, as described in Chapters 7.0 and 8.0. SAFETY EVALUATION SIX - Section 9.2.2.2 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 - The minimum component cooling water flow rates required to remove heat from the reactor coolant system during and after a postulated LOCA are listed in Table 9.2-11. Determination of these minimum flows includes the assumption of the highest anticipated component cooling water temperature. The CCWS design assures that at least these minimum flows are achieved by the system in its accident configuration, i.e., one pump per train in operation and
the nonseismic Category I CCW loop isolated. The system is shifted into the
accident configuration upon receipt of an SIS signal which starts a pump in
each train and isolates the nonessential flow paths. The CCWS flow to the RHR heat exchanger is automatically initiated as the RHR heat exchanger (primary side) is placed in service at the start of the recirculation mode. The CCWS heat is dissipated to the ESWS, as discussed in Section 9.2.1. The CCWS and each of the safety-related systems served by the CCWS are 100-percent
redundant. This arrangement ensures that full-heat dissipating capacity is available following a LOCA and an assumed single failure. 9.2-20 Rev. 13 WOLF CREEK SAFETY EVALUATION NINE - The minimum CCWS flow rates required to remove decay heat from the RCS during and following achieving post accident safe shutdown are listed in Table 9.2-10. Flow instrumentation indications and alarms are identified in Table 9.2-14. Determination of these minimum flows includes the assumption of the highest anticipated component cooling water temperature. The
CCWS heat is dissipated to the ESWS or SWS, as discussed in Section 9.2.1. The CCWS design assures that the flow requirements are met by operation of a CCWS
pump and -proper realignment of the associated valves. SAFETY EVALUATION TEN - As described in Section 9.2.1.2.2.3 and Table 9.2-25, the CCWS supports the ESWS by providing part of the heat energy needed to maintain the ESWS inlet trash racks from being blocked with frazil ice. 9.2.2.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. The performance and structural and leaktight integrity of all cooling water system components is
demonstrated by continuous operation. The CCWS is testable through the full operational sequence that brings the system into operation for reactor shutdown and for LOCAs, including operation of applicable portions of the protection system and the transfer between normal and standby power sources. -The safety-related components of the CCWS, i.e., pumps, valves, heat exchangers, and piping (to the extent practicable) are designed and located to permit preservice and inservice inspections. 9.2.2.5 Instrumentation Applications The CCWS instrumentation was designed to facilitate automatic operation and remote control of the system and to provide continuous indication of system
parameters. High flow switches in the return lines from each reactor coolant pump thermal barrier cooling coil will initiate the rapid closure of isolation valves to isolate the applicable reactor coolant pump in the event of a leak in the thermal barrier. Control room indication of surge tank levels keeps the operator informed of any leakage into or out of the CCWS. Actuation of the surge tank makeup valves is a-utomatically initiated by the low level switch. The low-low level switch will automatically isolate the nonseismic Category I part of the nonsafety-related train from the system. A radiation detection system is provided in each CCWS train to alarm abnormally high radioactivity which would be indicative of 9.2-21 Rev. 19 WOLF CREEK inleakage from one of the components. High radiation will isolate the CCWS surge tank vent in the affected flow train to prevent escape of radioactivity
prior to isolation of the leak. Thermowells and pressure indicator connections are provided where required for testing and balancing the system. Flow indicator taps are provided at strategic points in the system for initial balancing of the flows in the system and for verifying flows during plant operation. Table 9.2-14 summarizes CCWS alarms and indications of status, temperature, flow, etc.
9.2.3 DEMINERALIZED
WATER MAKEUP SYSTEM The demineralized water storage and transfer system (DWSTS) stores water for use upon demand for makeup within the plant. The DWSTS receives filtered and
demineralized water from the demineralized water makeup system (DWMS). For
reactor makeup water, a degasifier removes oxygen from the demineralized water as it is transferred. The effluent from several systems which process waste, which can be recycled within the plant, are passed through the DWSTS degasifier
before being transferred to the reactor makeup water storage tank (RMWST). 9.2.3.1 Design Bases 9.2.3.1.1 Safety Design Bases The DWSTS serves no safety function and has no safety design basis. 9.2.3.1.2 Power Generation Design Bases The Make Up Demineralized System (WM) is supplied by either the John Redmond Reservoir or the Service Water System. POWER GENERATION DESIGN BASIS ONE - The DWSTS and DWMS maintain chemistry specifications required by the plant components. Actual water chemistry specifications are located in the WCGS Chemistry Specification Manual. The demineralized water makeup system shall produce a final effluent that meets the following water quality requirements as a minimum: 9.2-22 Rev. 18 WOLF CREEK pH at 25 C. 6.0 to 8.0 Specific Cation Conductivity less than 1.0 micro-siemens/cm
Sodium less than 0.010 ppm Silica less than or equal 0.1 ppm Suspended Solids less than or equal 0.1 ppm
Total Solids less than 0.5 ppm
Chloride and Floride less than or equal 0.1 ppm
Potassium less than 0.01 ppm Aluminum less than 0.02 ppm Calcium less than 0.02 ppm
Magnesium less than 0.02 ppm POWER GENERATION DESIGN BASIS TWO - The DWSTS provides demineralized water to the equipment and systems shown in Table 9.2-15. The demineralized water makeup system shall provide reactor quality water for preoperational tests, hydrostatic tests, startup, restarts and normal operation. POWER GENERATION DESIGN BASIS THREE - The DWSTS's capacity is sufficient to supply the anticipated normal makeup demand in any 24-hour period. POWER GENERATION DESIGN BASIS FOUR - The demineralized water storage tank (DWST) has sufficient storage capacity to augment the condensate and reactor
makeup water storage facilities so that a 3-day supply of normal anticipated makeup demand to both the secondary and the primary systems is maintained. POWER GENERATION DESIGN BASIS FIVE - The DWSTS degasifies selected processed in-plant liquid waste and supplies it to the reactor makeup system with less
than 0.1 ppm dissolved oxygen. 9.2.3.2 System Description 9.2.3.2.1 General Description The DWSTS is shown in Figure 9.2-16. The system consists of one demineralized water storage tank, two 100-percent system capacity transfer pumps (connected
in parallel), one degasifier, and associated piping, valves, controls, and
instrumentation. Check valves are provided to preclude backflow from the
demineralized water transfer system to the demineralized water storage tank, assuring that contamination of the source is precluded. Water normally is supplied to the Demineralized Water Makeup System from the Service Water System (Figure 9.2-1) via a cross-tie with the raw water supply from the John Redmond Reservoir (Figure 9.2-5 sheet 2). The raw water system provides an alternate supply source if the Service Water System is not available. 9.2-23 Rev. 19 WOLF CREEK Demineralized water makeup equipment located inside the Shop Building has been retired-in-place. Demineralized water makeup to the DWST is supplied by a vendor owned and operated water treatment system located in the Water Treatment Building North. The system is capable of providing up to 400 gpm of deaerated, demineralized water continuously. The wastewater stream from the vendor water treatment system normally is discharged to the Wolf Creek cooling lake from a sump in the Water Treatment Building North. A wastewater line from the building sump is routed to the Wastewater Treatment Recirculation Pump discharge line which discharges to the lake via the Circulating Water Outfall. A back-up wastewater discharge line is routed to the Wastewater Treatment Facility via the TDS pump discharge. The piping, pumps, vessels, and other equipment associated with the demineralized water makeup system are constructed of corrosion-resistant
materials that prevent contamination of the demineralized water. Section 3.6 provides an evaluation demonstrating that the pipe routing of the DWSTS is physically separated from essential systems to the maximum extent
practicable. Protection mechanisms that may be required are also discussed in
Section 3.6. Samples may be taken from the discharge of the demineralized water transfer pumps. Chemical specifications for the demineralized water are given in Table
9.2-16 for the services indicated in Table 9.2-15. 9.2.3.2.2 Component Description The DWSTS is designed and constructed in accordance with quality group D specifications. DEMINERALIZED WATER STORAGE TANK - The demineralized water storage tank is a covered, vented, and insulated tank constructed of stainless steel and located outdoors. The tank has both a fixed roof and an internal floating roof. The stainless steel floating roof reduces the absorption of atmospheric oxygen into the water. Freeze protection is provided by external steam heating coils and tank insulation. The capacity of the tank is 50,000 gallons. 9.2-24 Rev. 19 WOLF CREEK DEMINERALIZED WATER TRANSFER PUMPS - The pumps are constant speed, electric motor-driven, vertical, centrifugal pumps. All parts in contact with the
pumped fluid are stainless steel. The two pumps are connected in parallel with common suction and discharge lines. DEGASIFIER - The degasifier consists of a level control valve, degasifier tank with integral 300-gallon storage section, two 100-percent system capacity, degasified water transfer pumps, and two 100-percent design capacity vacuum pumps. All parts in contact with the demineralized water are stainless steel, with the exception of the piping which is saran-lined and the degasifier column
packing which is polypropylene. 9.2.3.2.3 System Operation The water treatment vendor has the responsibility for operation and maintenance of equipment provided under contract. A control signal is provided from the DWST level switches for automatic control of the demineralized water supply to
the tank. A low-low level switch trips the demineralized water transfer pumps to prevent loss of suction. High and low tank level alarms are provided in the main
control room. Tank level indication is provided locally and in the main
control room. The tank can be bypassed, if necessary. The level indication read in the control room on the DWST and an imbalance of water addition versus controlled water discharge will indicate any abnormal leakage from the system. In addition, the floor drain system described in Section 9.3.3 provides additional detection capabilities. 9.2-25 Rev. 21 WOLF CREEK The reactor makeup water system (described in Section 9.2.7) must receive deaerated water to meet the chemistry specifications. Aerated plant waste streams which are recycled must pass through the degasifier. The DWST can be used as a source of makeup to the primary side water inventory once the other sources are exhausted. The flowrate of water to the degasifier is automatically controlled by the throttling of the inlet control valve in proportion to the degasifier column level. Degasification is accomplished by the spraying of the degasifier influent over the column packing while the column is maintained under vacuum. Vacuum is maintained by one of two centrifugal, oilsealed vacuum pumps. An
alternate source of vacuum is via the Condenser Vacuum Pumps. The vacuum pump
exhaust is combined with the exhaust of the condenser air removal system so that it is discharged directly to the unit vent. Two 100-percent system capacity degasified water transfer pumps supply the reactor makeup water system
on a demand basis. 9.2.3.3 Safety Evaluation The DWSTS serves no safety-related function. 9.2.3.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
The operability, performance, structural and leaktight integrity of all system components are demonstrated by continuous operation. The demineralized water makeup system was subjected to preoperational tests in order to ensure operability, reliability, and integrity of the system.
Portions of the systems can be isolated during normal station operation to
permit testing and maintenance. 9.2.3.5 Instrumentation Applications All instrumentation required for operation of the water treatment system is included with the vendor equipment installed in the Water Treatment Building North. A remote common trouble alarm is provided from the vendor system to the main control room to indicate a system malfunction. 9.2-26 Rev. 19 WOLF CREEK Local and control room indication of DWST level is provided. A local pressure gauge and control room low pressure alarm are provided for the demineralized water transfer pump header. The low pressure switch does not start the standby pump. The degasifier tank has a level gauge and local and control room high
and low level alarms.
9.2.4 POTABLE
AND SANITARY WATER SYSTEM The domestic water system (DoWS) provides chlorinated potable water for drinking and cooking, and for showers, laundry, and toilet facilities within
the standardized power block. Refer to Section 9.3.3 for the sanitary water
system.9.2.4.1 Design Bases 9.2.4.1.1 Safety Design Bases The potable and sanitary water systems serve no safety function and have no safety design basis. 9.2.4.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The potable and sanitary water systems provide potable water supplies and sewage treatment necessary for normal
station operation, shutdown periods, and the construction period. A
chlorinated, pressurized source is provided to adequately meet the requirements
of all outlets. POWER GENERATION DESIGN BASIS TWO - The potable and sanitary water system is designed so that there are no interconnections with systems which might contain radioactivity. In addition, the branches with outlets in areas where a radioactive hazard exists are designed with backflow-prevention capability to ensure that radioactive contamination cannot enter the system. 9.2-27 Rev. 19 WOLF CREEK POWER GENERATION DESIGN BASIS THREE - The sanitary waste water treatment system effluent shall meet the quality standards required by the Kansas State
Department of Health and Environment. 9.2.4.2 System Description 9.2.4.2.1 General Description
The potable water system is supplied from RWD #3 by the Melvern lake processing plant. A 70,000 gallon storage tank is filled directly from the incoming RWD
- 3 line. A pump skid takes water from this tank and supplies the main water line for the plant site. The potable water system is shown on Figure 9.2-5a.
The DoWS is shown in Figure 9.2-17. The system consists of hot water storage heaters and the necessary interconnecting valves and piping. The treated water
supply is provided from outside the power block, as described in Section 9.2.4. Potable water is delivered to all points in the plant by way of the domestic
water distribution piping system. Hot water at 140 F is supplied from a branch of the system, using thermostatically controlled heaters. This system serves shower and other
domestic fixtures. The system is designed to provide quantities of water adequate to enable proper functioning of the plumbing fixtures in all parts of the plant. No cross connections exist between the DoWS and any radioactive or potentially radioactive system. All cross connections to a non potable source or a potable source that has a different supply will be governed by the KDHE minimum standards. In addition, the main header serving these areas is provided with a
reduced pressure backflow preventer device. Hot water supply and recirculation
lines connected to the main hot water storage heater do not serve such areas. Protection against pollution from any equipment which takes water from the DoWS but uses this water for purposes other than drinking, cooking, or washing is
provided by passing the flow supplying such equipment through air gaps or a backflow prevention device of the reduced-pressure zone type. 9.2-28 Rev. 10 WOLF CREEK Section 3.6 provides an evaluation demonstrating that pipe routing of the DoWS is physically separate from essential systems, to the maximum extent
practicable. In addition, the floor drain system described in Section 9.3.3 provides leakage detection capabilities to assure that any abnormal leakage is detected and repaired. The sewage lagoon is of the non discharging type and is designed to evaporate all the liquid and store the solid residue for the remainder of plant design life.9.2.4.2.2 Component Description
The DoWS was designed in compliance with the Wisconsin Administrative Code, Section H62.13 and H62.14, the Uniform Plumbing Code, and the Occupational Safety and Health Standards (OSHA), Sections 1910.141 and 1910.151, dated
October 18, 1972. Future designs shall be in compliance with the Kansas State
Department of Health and Environment, the Uniform Plumbing Code, (UPC) and the
Occupational Safety and Health Standards (OSHA), Sections 1910.141 and 1910.151.HOT WATER HEATERS - Two 200 gallon electric water heaters are provided as the main source of heated water for the DoWS. These heaters supply hot water to
the main shower and toilet areas and to other plumbing fixtures, outlets, and equipment requiring domestic hot water service in the auxiliary, and control buildings. The design hot water supply temperature is 140°F. The hot water
heaters are designed to prevent the hot water supply temperature from exceeding 145°F for all flow rates and operating conditions. The remote toilet locations in the north and south ends of the turbine building which require hot potable water are not supplied by the main hot water branch but are provided with local electric water heaters. VALVES - Pressure relief valves are provided on the hot water heaters. Isolation valves are provided for each fixture to permit local shutoff. Pressure-reducing valves are provided, as necessary, to maintain fixture supply
pressure at or below 65 psig. An expansion tank of approximately 50 gallon is
provided on the outlet header of the heaters to accomodate thermal expansion of
the water. 9.2-29 Rev. 10 WOLF CREEK PIPING - All DoWS piping is copper with copper or bronze fittings. This will prevent the introduction of objectionable tastes, odors, discoloration, and
toxic conditions into the system. The piping system conforms to the provisions of the applicable local, state, and national plumbing codes. 9.2.4.2.3 System Operation The potable and sanitary water system receives potable water from RWD #3. A 70,000 gallon tank is filled directly from the incoming RWD #3 line. A pump skid takes water from this tank and supplies the main water line for site buildings. The DoWS uses this water and automatically supplies water when an intermittent demand is created at an outlet. Hot water at 140 F is maintained automatically by hot water heaters for distribution on a demand basis. Sanitary wastes collected by the sewage system are discharged into a Non Discharging Sewage Lagoon. The Non Discharging Sewage Lagoon is designed to operate at normal rate of 10,000 gallons of sewage per day and a maximum rate of 20,000 gallons of sewage per day. Sewage sludge is stored in the bottom of the lagoon. 9.2.4.3 Safety Evaluation Since no safety-related functions are performed by these systems, a safety evaluation is not required. 9.2.4.4 Tests and Inspections
Preoperational testing is discussed in Chapter 14.0. Proper operation of the various components is verified by satisfactory use. 9.2.4.5 Instrumentation Application Instrumentation is not required for the potable and sanitary water systems. However, thermostats, high temperature limit switches, temperature gauges, pressure relief devices, and pressure gauges are installed on the hot water storage heaters in the DoWS. 9.2-30 Rev. 10 WOLF CREEK
9.2.5 ULTIMATE
HEAT SINK
9.2.5.1 Design Bases
The ultimate heat sink (UHS) meets the following design bases:
9.2.5.1.1 Safety Design Basis SAFETY DESIGN BASIS ONE - The UHS provides a reliable source of cooling water to dissipate the heat of an accident safely and to achieve and maintain safe
shutdown of one nominal 3579 MW (th) unit following a DBA (GDC-44). SAFETY DESIGN BASIS TWO - The UHS supplies emergency makeup water to the fuel storage pool and component cooling water systems, and is the backup water supply for the auxiliary feedwater system. SAFETY DESIGN BASIS THREE - The UHS provides sufficient water volume and
cooling capability to shut down and cool down one nominal 3579 MW (th) unit and to maintain it in a post-accident safe shutdown condition.
SAFETY DESIGN BASIS FOUR - The UHS was designed to withstand, without loss of
function, the most severe natural phenomena, including seismic events such as
the safe shutdown earthquake (SSE). SAFETY DESIGN BASIS FIVE - The UHS was designed to withstand postulated site-
related events, such as loss of the main cooling lake.
9.2.5.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The UHS at Wolf Creek serves to dissipate
heat generated during plant operation.
9.2.5.2 System Description 9.2.5.2.1 General Description
The UHS for the Wolf Creek Generating Station consists of a normally submerged Seismic Category I cooling pond. The UHS is formed by providing a volume of 455 acre-feet with no sedimentation behind a Seismic Category I dam built in one finger of the main cooling lake. The water surface elevation of the UHS (the UHS dam crest) is 1070 msl feet. The normal cooling lake elevation is
1087 msl feet and
9.2-31 Rev. 25 WOLF CREEK cooling lake low water elevation is 1076 msl feet. The main cooling lake, main dam, saddle dams, and control structures are designed for stability under
drawdown, OBE, and probable maximum flood conditions. Availability of water at the intakes to the essential service water system
pumps is assured by location of the UHS adjacent to the Seismic Category I
essential service water system pumphouse. The essential service water intake structure and discharge point are described in Section 3.8.
UHS cooling water chemistry analysis is provided for information in Table 9.2-
- 17.
9.2.5.2.2 Heat Loads Analysis of the UHS is based on the total heat dissipation requirements
assuming a design-basis LOCA or shutdown using UHS. In the event Wolf Creek
undergoes a loss of offsite power (LOOP), the heat rejection rates are assumed
essentially equal to those of a LOCA. Heat is rejected to the UHS directly or indirectly via the ESW system following
a LOCA or shutdown from the four sources, namely, (1) containment air coolers, (2) residual heat removal system, (3) station auxiliary systems, and (4) fuel
storage pool. Tabulated below are the maximum heat rejection rates for the specified time
intervals, taken from Figures 9.2-6A and 9.2-19.
Normal Shutdown LOCA (10 6 Btu/hr) (10 6 Btu/hr) 0 to 1 hour 73 495
1 hour to 10 hours 299 357
10 hours to 1 day 184 169 1 day to 3 days 116 136 3 days to 15 days 103 113
15 days to 30 days 43 82
The above figures represent heat from all sources, including heat from sensible heat, fission product and heavy element decay, heat from the station auxiliary systems, and heat due to pump work.
9.2.5.2.2.1 Heat Load Following a LOCA
Heat rejected to the UHS following a LOCA is based on the assumption the two ESW trains are in operation for the entire 30 days. Following a postulated
LOCA, the two ESW trains are used to remove heat from the containment via four
containment air coolers and two residual heat removal heat exchangers. Heat
rejected to the UHS via the containment air coolers and the residual heat
removal system is based upon the results of the containment pressure/temperature analysis using full capacity of the containment heat removal system as described in Table 6.2.1-3 of Section 6.2.1.
9.2-32 Rev. 28 WOLF CREEK Figures 9.2-6A and 9.2-18 provide the heat load to the UHS as a function of time for 30-day period following a LOCA. In Figure 9.2-6A, the curved line
represents the analyzed heat curve which is the summation of the heat rejection to the UHS from the following sources:
- a. Containment air coolers (Figure 9.2-6B)
- b. Residual heat removal system (Figure 9.2-6C)
- c. Station auxiliary systems, and
- d. Fuel storage pool.
The heat rejection rates actually used in the analysis of the UHS are also shown in Figure 9.2-6A by the dashed line approximation to the analyzed curve.
Heat rejection rates from the containment air coolers and the residual heat
removal system as shown in Figure 9.2-6B and 9.2-6C are obtained from the results of the containment pressure/temperature analysis using full capacity of the containment heat removal system as described in Section 6.2.1. The average
heat rejection rate of the station auxiliary systems listed in Table 9.2-19
following a LOCA is 43.8 x 10 6 Btu/hr and is assumed to remain constant over a 30-day period. The heat load from the fuel storage pool cooling system is described in Section 9.1.3. Cooling to the fuel storage pool is assumed to be lost for the first 4 hours following a postulated LOCA.
The average heat load during the 30-day period following a LOCA is 86.8 x 10 6 Btu/hr. The total integrated heat released for the 30-day period is 62.5 x 10 9 Btu from the following:
- a. Total integrated heat released from the containment air coolers and the residual heat removal system is 22.2 x 10 9 Btu. b. For the station auxiliary systems, the heat rejection rate after a LOCA is 43.8 x 10 6 Btu/hr and is taken as constant over the 30-day period. Thus, total integrated heat released by the station auxiliaries is 31.5 x 10 9 Btu.
9.2-33 Rev. 25 WOLF CREEK
- c. The average decay heat load from the fuel storage pool cooling system is 12.2 x 10 6 Btu/hr. For a 30-day period, the total integrated heat due to fuel storage pool cooling is 8.8 x 10 9 Btu. This includes loads from the auxiliary equipment in the fuel pool heat removal system.
- d. Therefore, total heat rejection to the UHS is 62.5 x 10 9 Btu. 9.2.5.2.2.2 Heat Loads Following Shutdown Using UHS
Figures 9.2-19 and 9.2-20 provide the heat load as a function of time for a normal plant cooldown. The average heat load per unit during the 30 days following a plant cooldown is 57.5 x 10 6 Btu/hr. The total integrated heat load for the 30-day period is 41.4 x 10 9 Btu. Heat rejected to the UHS following a shutdown is based on the following
assumptions.
- a. Two-train RHR operation for the initial 20 hours at which time the unit is in cold shutdown condition.
- b. Diesel generators are shut down after 7 days when off-site power becomes available.
- c. Two ESW trains in operation for the entire period.
The heat rejection rate of the station auxiliary sytems is listed in Table 9.2-20 is 29.8 x 10 6 Btu/hr/train and is assumed to be constant for the first 23 hours. After 23 hours, the heat loads are assumed to be 20.4 x 10 6 Btu/hr/train and after 7 days when the diesel generator load is discontinued (because off-site power becomes available again), the heat load is assumed to
be 3.6 x 10 6 Btu/hr/train. The total integrated heat released by the station auxiliaries for 30 days is 11.6 x 10 9 Btu. The heat load from the fuel storage pool cooling system is assumed to be
constant at a maximum rate of 2.9 x 10 6 Btu/hr/train. The total integrated heat released by the fuel storage pool for 30 days is 4.18 x 10 9 Btu. This is the load expected at the end of the operating cycle which corresponds to the largest total heat loads from all sources.
9.2-34 Rev. 25 WOLF CREEK 9.2.5.2.3 Emergency Makeup Water Requirement Power block emergency makeup water requirements are given in Figures 9.2-21 and 9.2-22. It includes total makeup for the systems described below. The makeup water required to supply the auxiliary feedwater system when the condensate storage tank is unavailable or exhausted, is based on system operation as described in Section 10.4.9. Makeup water required to replace evaporative losses from the fuel storage pool is based on system operation, as described in Section 9.1.3 and Table 9.1-4. Makeup water may also be required to replace evaporative losses or minor leakage from the component cooling water system. 9.2.5.2.4 Component Description A design comparison of the UHS to the positions of Regulatory Guide 1.27 is provided in Table 9.2-21. 9.2.5.2.5 System Operation The capacity of the UHS is sufficient to provide cooling for the required period of 30 days with no makeup water under both normal and accident-mode operating conditions. The UHS is assumed to supply cooling water to the essential service water system (ESWS) at a rate of 30,000 gpm for the entire 30-day period for this analysis. The UHS has sufficient capacity to supply emergency makeup water to the fuel storage pool and component cooling water systems and to serve as the backup water supply for the auxiliary feedwater system. The UHS also has sufficient capacity to allow up to 140 gpm of continuous losses throughout the 30 day period from leakage from the ESWS conponents and other system components which would be supplied makeup from the
ESWS during accident conditions. At the start of the analysis, the UHS was assumed to have lost a volume of 155 acre-feet due to sediment. The UHS design assures that the design-basis temperatures of safety-related equipment are not exceeded at the flow rates outlined above. The design-basis temperature of water supplied to the plant is 95°F. 9.2-35 Rev. 14 WOLF CREEK 9.2.5.3 Safety Evaluation SAFETY EVALUATION ONE - The UHS is capable of providing enough cooling water for a post-accident safe shutdown and for continued cooling of the reactor for 30 days following an accident. The UHS water supply is sufficient to meet the design-basis plant heat rejection rates based on UO2 fuel as described in 9.2.5.2.2. The design basis heat load for the UHS is tabulated in Section 9.2.5.2.2. Section 9.2.5.3.1 provides a safety evaluation which demonstrates that the UHS
capacity is sufficient for a 30-day supply of cooling water at a maximum
temperature of 95 F, assuming maximum engineered safety feature operation with minimum heat transfer coefficient, as described in Section 6.2.1. This is based on maximum heat load and the most severe-meteorological conditions. SAFETY EVALUATION TWO - The minimum UHS reserve requirement to provide emergency makeup water to the fuel storage pool and component cooling water systems and backup water to the auxiliary feedwater system is provided in Figures 9.2-21 and 9.2-22. Section 9.2.5.3.1 provides the safety evaluation
which demonstrates that the UHS has an adequate capacity to meet these needs. SAFETY EVALUATION THREE - All Seismic Category I requirements are satisfied by the UHS. Cut slopes of the UHS and intake channel and the UHS dam embankment are designed for stability under end-of-construction, steady-state, drawdown, and SSE conditions. In addition, the UHS will withstand the unlikely event of
a breach of the main dam and subsequent rapid drawdown. The UHS dam embankment
structure will withstand overflow conditions that would result if the main cooling lake were to be drawn down below the UHS dam crest elevation. 9.2.5.3.1 UHS Analysis
FLOWAVE, a computer program developed by the Tennessee Valley Authority (TVA) and modified by S&L, was used for unsteady flood routing. This unsteady flow model is discussed in Subsection 2.4.4.2.2. The theoretical discharge and depth at the cross section of the Main Dam were computed as outlined by Stoker (Ref. 2) and used in Case I. The discharge computed was for the instantaneous
complete failure of the cooling lake Main Dam and hence conservative. In Case II, a similar approach was used for the instantaneous failure of the Baffle Dike 'A' in front of the UHS. The flood wave was routed through the cooling lake in both cases. Figures 9.2-11 and 9.2-12 show the transient average 9.2-36 Rev. 14 WOLF CREEK velocities through the cross section at the UHS Dam location for Cases I and II, respectively. From these figures, the maximum average velocity for Cases I
and II are 7.6 and 9.5 feet per second, respectively. During the unlikely postulated total loss of the main cooling lake dam and baffle dike 'A', the slopes and crest of the UHS dam will be subjected to a flow of water over the crest. Adequate erosion protection has been provided for the upstream and downstream slopes as well as for the crest of the dam. The techniques for the design of rock sections for overtopping were presented by Olivier (Ref. 1). A series of laboratory tests were made with various sizes
of stones to develop parameters for different flow rates. The test results were applied to the design of the UHS dam slope protection. Following the
criteria that the filter and riprap materials satisfy the quality requirements of concrete aggregates as given in ASTM C-33 and in accordance with the guidelines established in the Corps of Engineers publication entitled "Stability of Riprap and Discharge Characteristics, Overflow Embankments, Arkansas River, Arkansas" (Ref. 3) the riprap is a well-graded material with
the following gradations: Maximum Size Weight: 3200 pounds 85% Size Weight: 1500 to 2200 pounds
50% Size Weight: 190 to 400 pounds
15% Size Weight: 25 to 50 pounds Minimum Size Weight: 5 pounds The basic criteria or conditions for the UHS dam are quite similar to those experienced and investigated in the Corps of Engineers publication (Ref. 3).
The side slopes used in their study, 4 horizontal to 1 vertical, are the same as those for the UHS dam. The duration of the overtopping is approximately equal in both cases. The gradation for the riprap for the UHS dam was made to compare to the A-gradation used by the Corps. The UHS dam riprap is twice as
thick as that used by the Corps (4 feet as opposed to 2 feet), and is
complimented by two 18-inch filters consisting of a fine filter and a coarse filter. The maximum average water velocity expected over the UHS dam is less than 10 fps, while the Corps had experienced velocities as high as 13 fps. By examining various flow conditions over the UHS dam which take the tailwater elevation downstream and the headwater elevation upstream 250 feet from the crest of the dam (in contrast to the 100-foot distance used by the Corps), the riprap was found to be in the stable region for nonaccesstype embankments with
a gradation of A-1 as shown in Army Corps of Engineers Plate 48 (Ref. 3). This
A-1 gradation performed similarly to the A-gradation, as described in the Corps
of Engineers publication (Ref. 3). 9.2-37 Rev. 8 WOLF CREEK The riprap material is 4 feet thick, measured perpendicular to the slopes of the embankment. The filter material (coarse and fine beddings) to be placed
under the riprap was designed according to the criteria established in Subsection 2.5.6.4.1.4.2. Based on these criteria, the following gradation sizes were required for the filter material: Coarse Filter Fine Filter Sieve
%Passing Sieve
%Passing 4 inch 100 3/4 inch 100 3 inch 85-100 1/2 inch 90-100
1 1/2 inch 55-85 3/8 inch 70-100
3/4 inch 30-65 No. 10 20-65
3/8 inch 10-30 No. 30 8-35 No. 4 0-15 No. 50 3-15 No. 10 0-3 No. 200 0-5 Each of the coarse and fine bedding layers is 18 inches thick, measured perpendicular to the side slopes. Details of the riprap and filter are shown on Figures 2.5-116 and 2.5-117. The design water level of the UHS and crest of the cohesive embankment of the UHS Dam is at elevation 1070 and the elevation of the top of riprap is at elevation 1077. As shown in Figure 2.5-116, the riprap extends into the abutment to the point where natural grade is at elevation 1077. Therefore, any flow below elevation 1077 will be through areas protected by filter bedding and
riprap.The abutments are protected with adequate riprap and filter as described above and hence will not be eroded during the postulated overflow conditions. The UHS design was based upon adverse hydrological and meteorological conditions. The maximum temperature and maximum evaporation periods for recorded weather conditions were considered in sizing of the UHS. Selection of the critical weather periods was based upon a computer analysis (UHSAVG) of meteorological data for a 16-year period which included a severe drought, estimated to have a recurrence interval of 50 years. A weather tape scan of
surface weather data for Chanute, Kansas, and of precipitation data for Iola, Kansas, for the period of 1949-1964 was performed. These data included the historic drought years of 1952-1957. The 16-year weather data were used to evaluate water surface temperatures and evaporation rates for a prescribed rate of heat rejection from the surface of
the UHS. The worst evaporation period was obtained by selecting the weather conditions corresponding to the 30 consecutive days for which evaporation 9.2-38 Rev. 8 WOLF CREEK loss was maximum. The worst temperature period was obtained by saving the conditions for the 5 consecutive days, 1 day, and 30 consecutive days resulting
in highest average water temperature after which these three periods were combined in the indicated order to produce a synthetic 36-day worst-weather period. The maximum evaporative and temperature periods were determined to have the
following dates: Maximum Evaporation Period Worst 30 days: June 24, 1954 to July 23, 1954
Maximum Temperature Period Worst 5 days: June 30, 1949 to July 5, 1949
Worst 1 day: July 2, 1949 (Noon) to July 3, 1949 (Noon)
Worst 30 days: July 16, 1951 to August 15, 1951 For the above listed weather periods, ultimate heat sink draw-down and plant
inlet temperatures were evaluated as a function of time using a computer model (LAKET-5) which predicts the transient response of the heat sink to external
conditions. Heat rejection rates were taken as those corresponding to a LOCA or Normal
shutdown with UHS in one unit employing UO 2 fuel. In this mode of operation, a total UHS flow of 30,000 gpm is assumed. In addition, it was assumed that all
of the water in the UHS was at 90°F at the start of this analysis. This temperature is conservative since prior transient analysis of the main cooling lake, with one unit operating and an annual average load factor of 100% showed
that the main cooling lake surface temperatures at the locations of the UHS
were not in excess of 87.7°F. Because the UHS is submerged, actual UHS
temperatures will be less than 87.7°F. From evaluation of the UHS computer analysis, the maximum plant inlet (UHS
outlet) temperature occurring during the worst temperature period is predicted
to be 95°F. The plant inlet temperature was usually well below 95°F, as
evidenced by the average temperature over the entire period being slightly below 90°F and 95% of the time below 94°F.
The maximum drawdown under worst evaporative conditions, including water loss
due to lake seepage, was found to be approximately 1.65 feet from the initial
elevation of 1070 feet. This corresponds to a decrease in UHS volume of about 39% of the volume existing at the start of the accident. At this point the UHS water level is at 1068.35 feet, and the UHS has already provided the required
390,000 gallons of emergency makeup water to the fuel storage pool and component cooling water systems, back-up to the auxiliary feedwater system, and 140 gpm of ESWS losses throughout the 30 days. The UHS thus has the capability
for the safe plant shutdown in the event of postulated LOCA or shutdown using the UHS, assuming extreme evaporative conditions. Results obtained for the heat sink inlet and outlet temperatures are shown in Figure 9.2-7 for the worst temperature period drawdown and in Figure 9.2-10 for worst evaporation
conditions.
9.2-39 Rev. 14 WOLF CREEK Technical Specifications allow the UHS water temperature to exceed the limiting 90 F design-basis temperature provided that the level of the main cooling lake is equal to or greater than 1075 feet msl (i.e. the main cooling lake dam is intact) and the UHS temperature does not exceed 94 F. An additional evaluation, reference 4, was performed that assumes the main dam and water control structures for the lake will remain intact during a LOCA, the initial lake temperature could be as high as 94 F, and credit is taken for the volume of water directly above the UHS, up to 1075 feet msl. The maximum lake temperature following the most severe DBA, will remain below the accident analysis temperature of 95 F. Conservative projections of previous studies, assuming the main dam is not intact (i.e., the lake water elevation is at the UHS dam elevation of 1070 feet msl) and the initial water temperature is 94 F results in UHS temperatures rising above 95 F for short periods of time. Administrative tube plugging limits have been established for the emergency diesel generator (EDG) heat exchangers that are based on an intake cooling water temperature of 96 F. Consequently, there are no adverse affects on EDG operation. Other equipment operation abnormalities are not expected from a slight room temperature excursion if it were to occur. Availability of water in UHS is not affected by failure of makeup lines to, or blowdown lines from, the main cooling lake during the required 30-day period, since a substantial quantity of water has been shown to remain in the UHS even under extreme evaporative conditions. Dredging of the UHS will be performed whenever necessary to maintain a minimum capacity and adequate flow to the ESWS pumps. Sedimentation is discussed in more detail in Section 2.4.11. 9.2.5.4 Tests and Inspections The UHS is inspected periodically to determine degree of siltation. See Section 2.4.11.6 for a further discussion of siltation. The UHS dam will be monitored periodically as discussed in Section 2.5.5, and an annual verification that the crest of the UHS dam below the rip rap cover is at or above elevation 1070 Mean Sea Level, USGS datum is performed. 9.2.5.5 Instrument Application See Sections 2.4.11.6 and 2.5.5. 9.2-40 Rev. 17 WOLF CREEK
9.2.6 CONDENSATE
STORAGE AND TRANSFER SYSTEM
The condensate storage and transfer system (CSTS) consists of one 450,000-gallon condensate storage tank (CST), one non-safety-related auxiliary feedwater pump (NSAFP), and associated valves and piping. The CST serves as a reservoir to supply or receive condensate, as required by the condenser hotwell level control system. The tank is also a nonseismically designed source of
water to the auxiliary feedwater system. The NSAFP provides a diverse backup to the motor driven auxiliary feedwater pumps (MDAFPS) and turbine driven auxiliary feedwater pump (TDAFP), but is not credited for accident mitigation.
9.2.6.1 Design Bases 9.2.6.1.1 Safety Design Bases
The CSTS serves no safety function and has no safety design basis.
9.2.6.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The minimum usable volume of the CST
provides sufficient water to the suction of the auxiliary feedwater pumps to:
a) remove decay heat to maintain hot standby for 4 hours and b) to remove decay
heat and sensible heat to cool down the reactor to 350 F. POWER GENERATION DESIGN BASIS TWO - CSTS permits periodic testing of the
auxiliary feedwater pumps.
POWER GENERATION DESIGN BASIS THREE - The gross capacity of the CST is sufficient to fill the condensate system, feedwater system, and the steam generators.
POWER GENERATION DESIGN BASIS FOUR - The CSTS is designed to limit the
dissolved oxygen in the CST to less than 0.1 ppm. 9.2.6.2 System Description
9.2.6.2.1 General Description
The CSTS is shown in Figure 9.2-23. The system consists of one CST, one NSAFP and associated piping, valves, controls, and instrumentation. Section 3.6 provides an evaluation that demonstrates that the pipe routing of the CSTS is physically separated from the essential systems to the maximum extent
practicable. Protection mechanisms that may be required are also discussed in Section 3.6.
A sample is periodically taken from the makeup line to the condenser hotwell
for analysis to assure that the quality of the water stored in the CST meets
the chemical specifications given in Table 9.2-16 for the services indicated in
Table 9.2-22.
9.2-41 Rev. 27 WOLF CREEK 9.2.6.2.2 Component Description
Codes and standards applicable to the CSTS are listed in Table 3.2-1. The system is designed and constructed in accordance with quality group D specifications.
CONDENSATE STORAGE TANK - The CST is a covered, insulated tank constructed of
stainless steel. The tank has both a fixed roof and an internal floating roof. The floating roof prevents the absorption of atmospheric oxygen into the water and is constructed of stainless steel. The tank is located outdoors. Freeze
protection is provided by thermal insulation and external steam heating coils.
The capacity of the tank is 450,000 gallons.
NON-SAFETY AUXILIARY FEEDWATER PUMP - The NSAFP is driven by an AC-powered electric motor supplied with power from independent, dedicated diesel generator installed outdoors, immediately east of the northeast corner of the Turbine Building. This nonsafety-related 2 MW generator provides sufficient power to the NSAFP motor for starting and running the NSAFP at full system load. The NSAFP is a 10-stage horizontal centrifugal pump that takes suction from the non-safety-related condensate storage tank (CST) and discharges to the Auxiliary Feedwater System (AFS) downstream of the turbine-driven auxiliary feedwater pump (TDAFP). Pump design capacity includes manually controlled minimum flow recirculation back to the CST.
9.2.6.2.3 System Operation
The supply of demineralized water makeup to the CST is manually or
automatically controlled by the tank level. In automatic operation, low and high tank level signals cycle a control valve in the line from the demineralized water makeup system. Level HILO and Level LOLO 1 alarms in the
main control room are provided by CST level instrumentation. Also, a Level
LOLO 2 alarm is provided by the Auxiliary Feedwater System pump suction
pressure instrumentation. Tank level indication is provided locally and in the main control room.
The level indication of the CST and an imbalance of water addition versus
controlled water discharge will alert the operator of any abnormal leakage from
the system. In addition the floor drain system described in Section 9.3.3 provides additional detection capabilities. Section 3.6 demonstrates that a storage tank failure would have no detrimental effect on safety-related
structures or equipment.
Deaeration of the CST during initial startup operations can be accomplished via the main condenser. The tank contents circulate to the deaerating hotwell of the condenser and are pumped back to the CST by a condensate pump. At low
plant loads and at startup, the condenser spargers aid in the deaeration
process. Nitrogen sparging of the CST can also be performed to reduce
dissolved oxygen levels both during power operations and during refueling
outages. Overflow of the CST is directed to the secondary liquid waste system, as
described in Section 9.3.3. Isolation valves are provided for all lines which
penetrate the tank, with the exception of the overflow. The minimum volume
required for auxiliary feedwater is 281,000 gallons. During normal operation, the CSTS contains no radioactive contaminants. In the
event of primary-to-secondary system leakage due to a steam generator tube
leak, it is possible for the CST contents to become radioactively contaminated.
A discussion of the 9.2-42 Rev. 27 WOLF CREEK radiological aspects of primary-to-secondary leakage is included in Chapter 11.0. When the auxiliary feedwater system is started, the condensate storage tank provides a clean source to feed to the steam generators. If during emergency
operation the condensate storage tank is unavailable or exhausted, emergency
backup water is automatically supplied from the essential service water system, as described in Section 10.4.9. The function of the NSAFP is to provide an alternate source of cooling water to the steam generators through the Auxiliary Feedwater System. The NSAFP and dedicated diesel generator will be manually aligned upon the following events occurring simultaneously: Loss of off-site power, loss of on-site power, and failure of the TDAFP. 9.2.6.3 Safety Evaluation
The CSTS serves no safety-related function. 9.2.6.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0. The performance, structural, and leaktight integrity of all system components is demonstrated by continuous operation. Technical Specification requires monitoring of volume.
9.2.6.5 Instrumentation Applications
Local and control room indication of CST level is provided. A nominal minimum tank temperature of 50 F is maintained automatically by steam heating coils. A temperature control valve is provided to supply steam in response to actual
tank temperature. 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.
9.2.7 REACTOR
MAKEUP WATER SYSTEM
The reactor makeup water system (RMWS) stores deaerated water to be used upon
demand for primary makeup within the plant. The RMWS receives filtered, deaerated, demineralized water from the demineralized water storage and transfer system.
9.2.7.1 Design Bases 9.2.7.1.1 Safety Design Bases
Except for an associated containment penetration, the RMWS is not a safety-
related system.
SAFETY DESIGN BASIS ONE - 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.
9.2-43 Rev. 27 WOLF CREEK 9.2.7.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The reactor makeup water storage tank (RMWST) is designed to meet peak demands from the RCS in conjunction with the design makeup requirements for the fuel storage pool and the refueling pool. The peak demand from the RCS is experienced when the plant is going to cold shutdown at approximately 80-percent core life, with 200 ppm boron in the
reactor coolant system, with no control rods, and with equilibrium xenon, followed by a startup. The design makeup requirement for the fuel storage pool and the refueling pool is based on a pool temperature of 125 F. POWER GENERATION DESIGN BASIS TWO - The reactor makeup water transfer pumps are designed to deliver 120 gpm to the boric acid blending tee, which is equivalent to the maximum letdown flow from the reactor coolant system. These pumps are also designed to deliver 150 gpm, as an alternate source, for cooling the contents of the pressurizer relief tank from 200 F to 120 F in one hour
following a pressurizer safety valve discharge. POWER GENERATION DESIGN BASIS THREE - The RMWS is designed to supply high quality degasified water to minimize corrosion in the systems supplied and
which is compatible with the RCS water chemistry. 9.2.7.2 System Description 9.2.7.2.1 General Description The RMWS, shown in Figure 9.2-13, consists of one storage tank, two transfer pumps and a tank steam coil heater, and the associated piping, valves, and instrumentation. The RMWST receives water from the demineralized water system. The RMWS tank is provided with a diaphragm which is continuously in contact with the tank water surface. This prevents absorption of gases which would lower the water quality below that necessary 9.2-44 Rev. 14 WOLF CREEK for use as reactor makeup water. The reactor makeup water chemistry specifications are given in Table 9.2-16. Overpressure/overflow protection for the RMWST is provided by a loop seal which drains to the waste holdup tank in the liquid radwaste system. The RMWST level is maintained above a specified minimum by manual replenishment from the demineralized water system. The water temperature in the tank is maintained above freezing by an automatically controlled heater system. The heater system consists of steam
coils wrapped around the outside of the reactor makeup water storage tank.
Steam to these coils is provided by the auxiliary steam system and is
controlled automatically to maintain the tank contents above a nominal 50 F temperature. 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. The reactor makeup water transfer pumps, taking suction from the RMWST, are employed for various makeup and flushing operations throughout the nuclear steam supply system auxiliaries, the radwaste systems, and the fuel pool
cooling and cleanup system. Table 9.2-23 gives a summary of the reactor makeup
water requirements. A sample connection is provided to allow periodic analysis of the RMWST contents.The pipe routing of the RMWS is physically separated from essential systems, to the maximum extent practicable. 9.2.7.2.2 Component Description The containment penetration associated with the RMWS is designed and constructed to quality group B and seismic Category I requirements. The balance of the system is designed and constructed in accordance with quality group D specifications. REACTOR MAKEUP WATER STORAGE TANK - The tank is covered, vented, and insulated and is constructed of stainless steel. The tank is located outdoors. The tank contains a diaphragm and loop seal to maintain airtight integrity. Freeze protection is provided by external steam heating coils. The usable capacity of
the tank is 126,000 gallons. 9.2-45 Rev. 14 WOLF CREEK REACTOR MAKEUP WATER TRANSFER PUMPS - The pumps are 100-percent system capacity, inline centrifugal type and are driven by constant speed electric
motors. All parts in contact with the pumped fluid are stainless steel. The two pumps are connected in parallel with common suction and discharge lines. 9.2.7.2.3 System Operation In order to maintain the RMWST at a specified minimum water level, water from the demineralized water system is manually supplied to the RMWST by opening the air-operated supply valve. If the RMWST is out of service, this demineralized water supply can be manually transferred to the suction of the reactor makeup water transfer pumps. The RMWS is normally kept at pressure by operating one of the two pumps in the run mode. Thus, reactor makeup water is available upon demand. The second
pump is kept in the auto mode and will start upon low pressure in the discharge
header or on demand from the reactor makeup water control system (RMWCS). Once
automatically started, the additional pump is manually stopped when the surge demand has passed. The RMWCS is used to maintain proper boron concentration in the reactor coolant system and the preset level in the volume control tank by
supplying makeup water to the boric acid blending tee. The reactor makeup water transfer pumps are also operated locally and from the main control room to supply, as required, makeup and flush water to the various systems given in Table 9.2-23. These operations require manual actuation of
the valves normally isolating the various connections to the given systems. 9.2-46 Rev. 14 WOLF CREEK A recirculation line from the reactor makeup water transfer pump discharge to the RMWST is provided to protect the pumps during periods of system operation
when there is little or no demand from the systems normally being supplied. This maintains the system in a ready state for any of the automatic demands. Grab samples may be taken from the RMWST for analysis to assure that the quality of this makeup meets the chemical specifications for service in the RCS, as given in Table 9.2-16. If the reactor makeup requires purification, it can be recirculated through the demineralized water until the water chemistry is within specifications. The level indication on the RMWST and an imbalance of water addition versus controlled water discharge will alert the operator of any abnormal leakage from the system. In addition, the floor drain system described in Section 9.3.3 provides additional detection capabilities. In the event of a LOCA, the containment must be isolated as described in Section 6.2.4. The containment penetration associated with reactor makeup water has a check valve, inside the containment, and a power-operated valve which automatically closes on a CIS-A, outside the containment. 9.2.7.3 Safety Evaluation Except for an associated containment penetration, the RMWS is not a safety-related system. SAFETY EVALUATION ONE - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability. 9.2.7.4 Tests and Inspections Preoperational testing is discussed in Chapter 14.0.
The operability, performance, and structural and leaktight integrity of all system components is demonstrated by continuous operation. 9.2-47 Rev. 14 WOLF CREEK 9.2.7.5 Instrumentation Applications Instrumentation is provided to measure the water level in the reactor makeup water storage tank and to give the main control room indication and annunciation of high and low levels. Level-control instrumentation is provided
to stop the pumps on tank low-low level. Instrumentation is provided to measure the water temperature in the reactor makeup water storage tank and to give the main control room indication as well as annunciation of high and low temperatures. Temperature-control
instrumentation is provided to initiate and terminate the auxiliary steam supply to the heater coil. 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. Local pressure indicators are provided for the suction and discharge of the reactor makeup water transfer pumps. Instrumentation is provided to measure the pressure in the common discharge line of the reactor makeup water transfer pumps, to give the main control room indication and annunciation of low system pressure, and to start the backup pump to maintain system pressure.
9.2.8 CLOSED
COOLING WATER SYSTEM The closed cooling water system (ClCWS) receives heat from the turbine building miscellaneous plant equipment and rejects it to the service water system. 9.2.8.1 Design Bases 9.2.8.1.1 Safety Design Bases The ClCWS serves no safety function and has no safety design basis. 9.2.8.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The ClCWS provides corrosion-inhibited, demineralized cooling water to the equipment shown in Table 9.2-24. POWER GENERATION DESIGN BASIS TWO - During power operation, the ClCWS operates to provide a continuous supply of cooling water, at a maximum temperature of 105 F, to turbine plant equipment, with a service water inlet temperature of 90 F. 9.2-48 Rev. 11 WOLF CREEK POWER GENERATION DESIGN BASIS THREE - The system is designed to permit the maintenance of any single active component without interruption of the cooling
function.POWER GENERATION DESIGN BASIS FOUR - Makeup to the system can be provided at a rate of 5 percent of the closed cooling water flow. POWER GENERATION DESIGN BASIS FIVE - The surge tank is sized to provide at least 30 seconds of active storage. 9.2.8.2 System Description 9.2.8.2.1 General Description The ClCWS is shown on Figure 9.2-14. The system consists of one surge tank, one chemical addition tank, two pumps, two heat exchangers (connected in parallel), and associated piping, valves, controls, and instrumentation. Heat
is removed from the ClCWS, via the closed cooling water heat exchanger, by the service water system, which is described in Section 9.2.1.1. A sample is periodically taken for analysis to assure that the water quality meets the chemical specifications given in the WCGS Chemistry Specification Manual for the services indicated in Table 9.2-24. 9.2.8.2.2 Component Description The ClCWS is designed and constructed in accordance with quality group D specifications. CLOSED COOLING WATER SURGE TANK - The closed cooling water surge tank is covered, vented, and constructed of carbon steel. The tank is located in the turbine building. Demineralized water make-up is provided by a mechanical
level control valve. The capacity of the tank is 866 gallons. CLOSED COOLING WATER CHEMICAL ADDITION TANK - The closed cooling water chemical addition tank is constructed of carbon steel. Provisions for make-up water and addition of the chemicals are included. The tank is located in the turbine building. The capacity of the tank is 75 gallons. CLOSED COOLING WATER PUMPS - The pumps are constant speed, electric motor-driven, horizontal centrifugal pumps. The two pumps are connected in parallel with common suction and discharge lines. The pumps operate at approximately
1062 gpm. 9.2-49 Rev. 12 WOLF CREEK CLOSED COOLING WATER HEAT EXCHANGERS - The closed cooling water heat exchangers are of horizontal shell and straight tube design. The tube side is supplied
with service water, and the shell side is supplied with closed cooling water. The surface area is based on normal heat load.
9.2.8.2.3 System Operation
During normal power operation, one of the two 100-percent-capacity closed cooling water pumps circulates demineralized water through the shell of one of the two 100-percent-capacity closed cooling water heat exchangers. In the
closed cooling water heat exchanger, heat is rejected to the service water
passing through the tubes.
Cooling water flow rate to the electrohydraulic control (EHC) coolers, steam generator feed pump turbine lube oil coolers, and generator exciter air cooler
is regulated by automatic control valves. Control valves in the cooling water
outlet from these units are throttled in response to temperature signals from
the fluid being cooled. Cooling water flow rate to the steam generator feed pump hydraulic power unit coolers is regulated by automatic control valves in the cooling water inlet.
The flow rate of cooling water to all of the other coolers is manually
regulated, by individual throttling valves located on the cooling water outlet from each unit.
The closed cooling water surge tank is connected to the pumps' suction and is
located at a relatively high elevation in the system to provide ample suction
head for the pumps. The surge tank also provides a reservoir for small amounts of leakage from the system and for the expansion and contraction of the cooling fluid with changes in the system temperature.
Demineralized water makeup to the ClCWS is controlled automatically by a level
control valve which is actuated by sensing surge tank level. A corrosion inhibitor is manually added to the system.
9.2.8.3 Safety Evaluation
The ClCWS does not serve a safety-related system. 9.2.8.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0. The performance, structural, and leaktight integrity of all system components is demonstrated by continuous operation.
9.2-50 Rev. 27 WOLF CREEK 9.2.8.5 Instrument Applications Local indication of closed cooling water surge tank level is provided. Surge tank low and high level alarms are provided in the control room via the plant computer. Each pump discharge contains a pressure gauge. Pressure indicator connections are provided where required for testing and balancing the system. Flow indicator taps are provided at strategic points in the system for initial balancing of the flows and for verifying flows during plant operation. 9.
2.9 REFERENCES
- 1. Olivier, H., "Through and Overflow Rockfill Dams - New Design Techniques," Proceedings of the Institute of Civil Engineers
, Paper No. 7012, Vol. 36, March 1967.
- 2. Stoker, J.J., 1957, Water Waves: Interscience Publishers, New York, p. 333-513.
- 3. U.S. Army Corps of Engineers, "Stability of Riprap and Discharge Characteristics, Overflow Embankments, Arkansas River, Arkansas," Publication No. 2-650, June 1964.4. SLWC-0588, Sargent & Lundy Letter, Transmittal of Final Results of UHS Thermal Analysis Study, June 15, 2000. 9.2-51 Rev. 17 WOLF CREEK TABLE 9.2-1 SERVICE WATER SYSTEM FLOW REQUIREMENTS NORMAL POWER GENERATION OPERATION Flow Operating/ Duty Each Total Standby(5) Total (7) No./ (x 10 6 Duty (x 10 6 Component Each (gpm) Flow (gpm) In Use(5) Btu/hr) Btu/hr)
- a. Closed cooling 776/776 1,552 2/1 4.22 4.22 water heat exchangers
- b. Central chiller 1,350/556 1,906 2/1 9.56 9.56
condensing units
- c. Central chiller ~0.5/~0.5 ~1 2/0 0 0
pumpout units
- d. Steam packing 1,153/741 1,153 1/1 12.5 12.5
exhauster
- e. Deleted
- f. Generator hydrogen 1,450 2,900 2/2 11.99 23.98 coolers
- g. Generator stator 2,388/1,038 3,426 2/1 17.9 17.9 liquid coolers
- h. Turbine-generator 1,608/1,608 3,216 2/1 10.8 10.8 lube oil coolers
- i. CVCS chiller 300/153 300 1/1 2.3 2.3
- j. Steam generator 1,526/1,526 1,526 1/1 18.5 18.5 blowdown non-regenerative heat
exchanger (max.)
- k. Condenser vacuum 630/630 1,890 3/2 .87 1.74
pump seal water coolers (8)
- l. Water box venting Unit C: 55/55 229 3/2 .17 .34 pump seal water Units A&B: 87/87 coolers (8)
- m. Motor-driven steam 24 24 1/0 .48 0 generator feedwater
pump and coolers (2)
- n. Make Up Demineralized N/A 700 (9) N/A N/A N/A System Supply Rev. 27 WOLF CREEK TABLE 9.2-1 (Sheet 2)
Total Duty (x 10 6 Total Flow (gpm) Btu/hr)
Service water system total 18,823 101.84(3)
Total essential service water system (normal operation)(1) 24,246 65.496
Total power block service 43,069 167.336 water requirement (6)
(1) From Table 9.2-2 (2) Flow is continuous - duty applies during start-up only (3) Does not include item m. or c.
(4) Deleted (5) Units on standby will receive the standby flow to control microbiologically induced corrosion (MIC). MIC flow to the standby turbine generator lube oil coolers and central chiller condensers is optional, if chemical treatment or
periodic inspections and cleaning is provided. (6) Does not include 2000 gpm for site auxiliaries. (7) These flow rates are nominal design flow rates. Actual operating flow rates may be lower because when the lake temperature is below the design temperature of 90°F, less flow is required to remove the heat load of the heat exchangers. Also, in some components the flow is throttled to control the temperature of the process fluid. Additionally, if the desired MIC flow is not maintained, periodic MIC inspection are performed. (8) Number of running pumps will be as required to maintain system vacuum. (9) Flow will vary from 0 to 700 gpm depending on Make Up Demineralizer System demand.
Rev. 27 WOLF CREEK TABLE 9.2-2 ESSENTIAL SERVICE WATER SYSTEM FLOW REQUIREMENTS NORMAL POWER GENERATION OPERATION
'A' Train 'B' Train Total Flow (gpm)/ Flow (gpm)/ Duty Equipment Section Number/ Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr)(2) Btu/hr)(2) Btu/hr)
Component 9.2.2 2/1 8,800/48.42 7,350/0 48.42
cooling water heat exchanger
Containment 6.2.2 4/4 1,850/6.76 1,850/6.76 13.52 air cooler
Diesel 9.5.5 2/0 1,200/0 1,200/0 0.0 generator
cooler Component 9.4.3 2/1 128/0.153 128/0 0.153
cooling water pump room cooler
Centrifugal 9.4.3 2/1 128/0.162 128/0 0.162 charging pump room cooler Auxiliary 9.4.3 2/0 128/0 128/0 0.0 feedwater
pump room
cooler Safety 9.4.3 2/0 88/0 88/0 0.0
injection pump room cooler RHR pump 9.4.3 2/0 88/0 88/0 0.0 room cooler
Rev. 29 WOLF CREEK TABLE 9.2-2 (Sheet 2)
'A' Train 'B' Train Total Flow (gpm)/ Flow (gpm)/ Duty Equipment Section Number/ Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr)(2) Btu/hr)(2) Btu/hr)
Containment 9.4.3 2/0 88/0 88/0 0.0
spray pump room cooler Penetration 9.4.3 2/2 100/0.161 100/0.170 0.331 room cooler (electrical) Fuel pool 9.4.2 2/1 32/0.077 32/0 0.077 cooling pump room cooler Control 9.4.1 2/1 140/0.663 140/0 0.663 room a/c unit condenser Class IE 9.4.1 2/2 66/0.485 66/0.485 0.97 switchgear a/c condenser Air compressor 9.3.1 2/2 40/0.60 40/0.60 1.20 and after
cooler (1)
Total "A" train --- --- 12,848/57.481 --- 57.481 Total "B" train --- --- --- 11,398/8.015 8.015 Flow to 10.4.9 --- O/NA O/NA ---
auxiliary feedwater system Makeup to 9.1.3 --- O/NA O/NA --- fuel storage pool cooling & cleanup
system Makeup to 9.2.2 --- O/NA O/NA ---
component cooling
water system _________________ ____________ ____________ ______ Total(3) --- --- 12,848/57.481 11,398/8.015 65.496 (1) Values may vary with plant conditions; both are assumed to be operating. (2) These flow rates are nominal design flow rates. Actual operating flow rates may be lower because when the lake
temperature is below the design temperature of 90°F, less flow is required to remove the heat load of the heat exchangers. Also, in some components the flow is throttled to control the temperature of the process fluid. Additionally, if the desired MIC flow is not maintained, periodic MIC inspection are performed.
(3) Pre-Lube Storage Tank flow is not included in this total.
Rev. 26 WOLF CREEK TABLE 9.2-3 ESSENTIAL SERVICE WATER SYSTEM FLOW REQUIREMENTS POST-LOCA OPERATION
'A' Train 'B' Train Total Flow (gpm)/ Flow (gpm)/ Duty
Equipment Section Number/ Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr)(3) Btu/hr)(3) Btu/hr)(2)
Component 9.2.2 2/2 7,350/164.14 7,350/164.14 (2) cooling water heat exchanger (1) and (9) Containment 6.2.2 4/4 2,000/141.4 2,000/141.4 (2)
air cooler (10) Diesel 9.5.5 2/2 1,200/16.8 1,200/16.8 33.6
generator cooler Component 9.4.3 2/2 128/0.280 128/0.280 0.560 cooling water pump
room cooler (13) Centrifugal 9.4.3 2/2 128/0.171 128/0.171 0.342
charging pump room cooler (13)
Auxiliary 9.4.3 2/2 128/0.320 128/0.320 0.640 feedwater pump room cooler (7) (13)
Safety 9.4.3 2/2 88/0.165 88/0.165 0.330 injection pump room
cooler (13) RHR pump 9.4.3 2/2 88/0.17 88/0.17 0.34
room cooler (13) Containment 9.4.3 2/2 88/0.174 88/0.174 0.348
spray pump room cooler (13)
Penetration 9.4.3 2/2 100/0.108 100/0.114 0.222 room cooler (electrical) (13)
Rev. 29 WOLF CREEK TABLE 9.2-3 (Sheet 2)
'A' Train 'B' Train Total
Flow (gpm)/ Flow (gpm)/ Duty Equipment Section Number/ Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr)(3) Btu/hr)(3) Btu/hr)(2)
Fuel pool 9.4.2 2/2 29/0.075 29/0.075 0.150 cooling pump room cooler (13) Control 9.4.1 2/2 85/0.663 85/0.663 1.326 room a/c unit condenser (13)
Class IE 9.4.1 2/2 73.3/0.485 73.3/0.485 0.970 switchgear
a/c condenser (13)
Air 9.3.1 2/2 40/0.60 40/0.60 1.20 compressor and
after cooler (4) ______________ _____________ _____________ ______ Total for - - 11,548/(12) 11,548/(12) (12) operating
ESW train Maximum 10.4.9 - 1,120/NA 1,120/NA -
flow to auxiliary feedwater
system (5&6) Makeup to 9.1.3 - 25/NA 25/NA -
fuel storage pool cooling & cleanup
systems (6) Maximum 9.2.2 - 100/NA 100/NA -
makeup to component cooling water
system (6) ______ ______
Total flow(11) - - 12,793 12,793 -
Rev. 29 WOLF CREEK TABLE 9.2-3 (Sheet 2a) NOTE: (1) Load does not occur until post-LOCA recirculation mode is initiated.
(2) The CCW heat exchanger and CAC duties given are limiting design values corresponding to a single train cooldown. These number are not additive since duties per train are significantly less for a two train cooldown. CCW heat exchanger and CAC peak duties for a two train LOCA cooldown are 240 MBTU/hr (120 MBTU/hr per train) and 272 MBTU/hr (136 MBTU/hr per train), respectively (see Figures 9.2-6C and 9.2-6B.
RHR and CCW peak duties are essentially the same).
(3) Peak duty is shown for each component. Total duty is actually less and will reduce long term, as described
in Section 9.2.5.
(4) Values may vary with plant conditions.
(5) Auxiliary feedwater system may be used to maintain steam generator water level post-LOCA.
(6) Flow shown would be maximum intermittent value expected.
(7) Heat load shown would be maximum intermittent value expected.
(8) Deleted.
(9) The essential service water flow to the CCW heat exchangers may be reduced to as low as 7150 gpm in order to ensure that the design bases cooling water flows are provided to the remaining ESW system components.
(10) At least 42.7 psig must be maintained at the CAC return ESW header (measured at the top of the cooler housing) to prevent boiling in the tubes.
(11) Pre-Lube Storage Tank, Traveling water screen wash and Strainer Backwash Flows are not included in this total.
(12) These duties are not additive since peak CCW and CAC duties do not occur at the same time. Peak analyzed ESW duty for two train LOCA cooldown is 494.6 MBTU/hr (see Figure 9.2-6A). (13) For these units the flow and duty listed are nominal vendor design values and reflect a target value that is used when flow balancing the system. The minimum flow requirements for removing Post LOCA heat loads are less than the values listed and are documented in controlled Engineering Calculations.
Rev. 16 WOLF CREEK TABLE 9.2-4 ESSENTIAL SERVICE WATER SYSTEM FLOW REQUIREMENTS NORMAL SHUTDOWN OPERATION
'A' Train 'B' Train Total
Flow (gpm)/ Flow (gpm)/ Duty(2) Equipment Section Number/ Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr) Btu/hr) Btu/hr)
Component 9.2.2 2/2 8,800/135.78 8,800/117.11 261.70 cooling water
heat exchanger (1)
Containment 6.2.2 4/4 1,850/6.76(5) 1,850/6.76(5) 13.52(5) air cooler
Diesel 9.5.5 2/0 1,200/0 1,200/0 0.0 generator cooler Component 9.4.3 2/2 128/0.263 128/0.280 0.543 cooling water pump room cooler(8)
Centrifugal 9.4.3 2/1 128/0.162 128/0 0.162 charging pump room
cooler (8) Auxiliary 9.4.3 2/0 128/0 128/0 0 feedwater pump room cooler (8)
Safety 9.4.3 2/0 88/0 88/0 0.0 injection
pump room cooler (8)
RHR pump 9.4.3 2/2 88/0.17 88/0.17 0.34 room cooler (8)
Containment 9.4.3 2/0 88/0 88/0 0.0 spray pump room cooler (8)
Penetration 9.4.3 2/2 100/0.161 100/0.170 0.331 room cooler (8) Rev. 29 WOLF CREEK TABLE 9.2-4 (Sheet 2)
'A' Train 'B' Train Total Flow (gpm)/ Flow (gpm)/ Duty(2)
Equipment Section Number Duty (x 10 6 Duty (x 10 6 (x 10 6 Description Number In Use Btu/hr) Btu/hr) Btu/hr) Fuel pool cooling 9.4.2 2/2 32/0.056 32/0.056 0.112 pump room cooler (8) Control room a/c 9.4.1 2/1 140/0.663 140/0 0.663 unit condenser (8) Class IE switch- 9.4.1 2/2 66/0.485 66/0.485 0.97 gear a/c condenser (8) Air compressor 9.3.1 2/2 40/0.600 40/0.600 1.20
and after cooler (3) _____________ _____________ ______ Total - - 12,848/153.91 12,848/125.631 279.541 Flow to 10.4.9 - - 0 - auxiliary feed-water system Makeup to spent 9.1.3 - - 0 - fuel pool cooling & cleanup system Makeup to 9.2.2 - - 0 - component cooling water system ______ ______ Total flow(6) 12,848 12,848 Rev. 16 WOLF CREEK TABLE 9.2-4 (Sheet 2a) NOTE: (1) Maximum duty from CCW occurs 4 hours after initiation of shutdown when the RHR system is brought into service, as described in Section 9.2.2. (2) Peak duty is shown for each component. Total duty to UHS is actually less and will reduce long term. (3) Values may vary with plant conditions; both are assumed to be operating. (4) Estimated temperatures based on a maximum service water supply temperature of 90F. These temperatures apply 4 hours after shutdown. (5) The heat load per train for Normal Shutdown with loss of off-site power is 3.96 x 10 6 Btu/hr. The total heat load is 7.92 x 10 6 Btu/hr. (6) Pre-Lube Storage Tank flow is not included in this total. (7) Deleted. (8) For these units the flow and duty listed are nominal vendor design values and reflect a target value that is used when flow balancing the system. The minimum flow requirements for removing Normal Shutdown heat loads are less than the values listed and are documented in controlled Engineering Calculations. Rev. 16 WOLF CREEK TABLE 9.2-5 ESSENTIAL SERVICE WATER SYSTEM COMPONENT DATA Essential Service Water Pump (all data is per pump) Quantity 2 (100% each) Type Vert centrifugal - 2 stg. with packed stuffing boxes Capacity, gpm 15,000 TDH, ft 361
Submergence required, ft 9 Material Case Carbon steel Impeller Aluminum - Bronze
Shaft Stainless Steel
Design Codes ASME Section III, C1. 3 Driver Type Electric motor
Horsepower 1,750
RPM 885
Power Supply 4,000 V 60 Hz, 3-phase, Class 1E Design code NEMA Seismic design Category I Essential Service Water Pump Prelube Storage Tanks (all data is per tank) Quantity 2 Type Vertical Capacity, gallons 43 Design pressure Atm. Design temperature, F 122 Shell material Carbon steel
Corrosion Allowance 1/16 inch Design code ASME Section III, C1. 3 Seismic design Category I Essential Service Water Self Cleaning Strainers (all data is per strainer) Quantity per unit 2 Capacity, gpm 15,000
Pressure drop, clean 3.5 psi (typical)
Pressure drop, dirty* 5.56 psi
Strainer openings 1/16 inch
Design pressure psig 200 Design temperature, F 100
- At start of backwash Rev. 11 WOLF CREEK TABLE 9.2-5 (Sheet 2)
Design Code ASME Section III, C1. 3 Driver Type Electric Motor
Horsepower 1
Rpm 1750
Power Supply Class 1E Design Code NEMA Seismic design Category I
Essential Service Water Traveling Water Screens
(all data is per screen)
Quantity 2
Capacity, gpm 15,000
Size, height x width 43'-0" x 8'-0" Screen openings 3/8" Screen material 304 SS Pressure drop, clean 0.05 inches of water
Pressure drop, dirty (20% clean) 0.691 inches of water
Water velocity @ design capacity 0.74 ft/sec
Design code AGMA* Seismic design Category I Driver Type Electric motor
Horsepower 3.0/0.75
Rpm 1,740/430 Power Supply Class 1E Design Code NEMA
Seismic design Category I
Piping, Fittings, and Valves Design pressure, psig 200
Design temperature, F 200
Material Carbon steel, Stainless Steel Design Code ASME Section III, C1. 3
- AGMA - American Gear Manufacturers Association
Rev. 28 WOLF CREEK TABLE 9.2-6 ESSENTIAL SERVICE WATER SYSTEM SINGLE ACTIVE FAILURE ANALYSIS Component Failure Comments
- 1. ESW pump and associated Fails to start on Two pumps are provided.
supporting items automatic signal. One is sufficient for post-LOCA heat removal.
- 2. Supply isolation valve Fails to close on Second valve in series between SW and ESW automatic signal. provides isolation.
system
- 3. Supply valve to air Fails to close Continued use of the compressor upon small break. system results in minimal loss of water.
100 percent of the heat load is removed by the redundant train.
- 4. CCW heat exchanger Fails to open on Two CCS heat exchangers inlet valve automatic signal. and two paths are pro-vided. One loop provides 100% cooling capacity.
- 5. CCW heat exchanger Fails to close on Results in lower flows main outlet valve automatic signal. to other components (i.e.
Note bypass valve containment air cooler), provides desired hence reducing their flow. efficiency. 100 percent of the heat load is
removed by the redundant train.
- 6. Return isolation Fails to close on Second valve in series valve between automatic signal. provides isolation.
SW and ESW system
- 7. Return isolation Fails to open on 100 percent of the heat valve to ultimate automatic signal. load is removed by the heat sink. redundant train. Rev. 6 WOLF CREEK TABLE 9.2-7 ESSENTIAL SERVICE WATER SYSTEM, INDICATING AND ALARM DEVICES Indication Control Room Local Control Room Alarm ESW header flow rate Yes No No ESW header temperature Yes No No High pressure drop to air Yes No No compressor and after cooler Power-operated valve Yes Yes No position (all valves)
Rev. 8 WOLF CREEK TABLE 9.2-8 ESSENTIAL SERVICE WATER SYSTEM INDICATING AND ALARM DEVICES Indication Control Room Local Essential service water pump Yes Yes discharge pressure Essential service water pump No Yes discharge strainer high
pressure differential Essential service water pump No Yes intake water level Alarms Essential service water pump Yes No low discharge pressure Essential service water pump Yes No
discharge strainer high
pressure differential Rev. 5 WOLF CREEK TABLE 9.2-9 COMPONENT COOLING WATER SYSTEM REQUIREMENTS NORMAL OPERATION 'A' Train 'B' Train Flow (gpm)/ Flow (gpm)/ Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr) Btu/hr) (x 10 6 Btu/hr)Essential ComponentsResidual heat removal heat 5.4.7 & 6.3 2/0 0/0 0/0 0.0 exchangers RHR pump seal coolers 5.4.7 & 6.3 2/1 4/0 0/0 0.0 Centrifugal charging pump 9.3.4 2/1 55/0.0755 0/0 0.0755 bearing oil coolers Safety injection pump 6.3 2/1 25/0 0/0 0.0 bearing oil coolers Fuel pool cooling heat 9.1.3 2/1 3,000/15.46 0/0 15.46 exchangers Excess letdown heat exchanger 9.3.4 1/1 183 (9)/0 NA 0.0 Reactor coolant pumps(4) 5.0 4/4 2,064/8.38 NA 8.38 Motor air coolers Upper bearing coolers Lower bearing coolers Thermal barrier cooling coils Nonessential Components (1)(4)Reactor coolant drain 11.2 1/1 225/2.23 NA 2.23 Tank heat exchanger Letdown heat exchanger 9.3.4 1/1 700/10.5 (10) NA 10.50 (2)Seal water heat exchanger 9.3.4 1/1 375/1.88 NA 1.88 Rev. 15 WOLF CREEK TABLE 9.2-9 (Sheet 2)
'A' Train 'B' Train Flow (gpm)/ Flow (gpm)/ Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr) Btu/hr) (x 10 6 Btu/hr) Recycle evaporator 9.3.6 1/0 0/0 (12) NA 0 (12) Package (11) (12) (13) Aux. steam radiation 1/1 46/0.23 NA 0.23 monitor RE-50 Deleted (5)
Waste gas compressor (2) 11.3 2/1 100/0.135 NA 0.135 Catalytic hydrogen 11.3 2/1 20/0.075 NA 0.075 recombiner (2) Nuclear sample system 9.3.2 1/1 82.4/0.63 (6) NA 0.63 sample cooler Waste evaporator package 11.2 1/0 0/0 (12) NA 0 (12) (12) (13)
Secondary waste evaporator 10.4.10 1/0 0/0 (12) NA 0 (12) package (12) (13)
Reverse osmosis unit (3) 11.2 1/1 0/0 (3) NA 0.0 (3)
Total 6880/39.61 0/0 39.61 Notes: (1) Nonessential components feed from either train. (2) Each of the two components is supplied with cooling water at a rate of one-half of the flow rate shown; however, only one of the components is accepting a heat load. (3) Reverse Osmosis Unit has been removed, however there is a flow loop via throttling EGV0423. EGTV0423 is locked shut. (4) The flows and related duty for these components are nominal values used for calculational purposes. Actual flows and duties may vary with actual plant operating conditions. (5) PASS is no longer used due to Amendment 137. (6) Flow as low as 60 gpm is acceptable for use. (7) Deleted. (8) Deleted. (9) 170 gpm for -15°F T hot Reduction (10) 9.9 x 10 6 Btu/Hr for -15°F T hot Reduction (11) Deleted. (12) Flow to these components is permanently isolated. (13) Equipment is abandoned in place. Rev. 29 WOLF CREEK TABLE 9.2-10 COMPONENT COOLING WATER SYSTEM REQUIREMENTS SHUTDOWN (@ 4 HOURS) OPERATIONS
'A' Train 'B' Train Flow (gpm)/ Flow (gpm)/
Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr)(1)(2) Btu/hr)(2)
(x 10 6 Btu/hr)Essential Components Residual heat removal heat 5.4.7 & 6.3 2/2 7,600/117.0 7,600/117.0 234.0 exchanger RHR pump seal cooler 5.4.7 & 6.3 2/2 4/0.03 4/0.03 0.06 Centrifugal charging pump 9.3.4 2/2 55/0.0755 55/0.0755 0.151 bearing oil cooler Safety injection pump 6.3 2/2 25/0 25/0 0.0 bearing oil cooler
Fuel pool cooling heat 9.1.3 2/0 0/0 0/0 0.0 exchanger (3)
Excess letdown heat exchanger 9.3.4 1/1 183 (13)/0 NA 0.0 Reactor coolant pumps(9) 5.0 4/1 2,064/2.1 NA 2.1 Motor air coolers Upper bearing coolers Lower bearing coolers Thermal barrier cooling coils Nonessential Components(9) Reactor coolant drain 11.2 1/0 225/0 NA 0.0 Tank heat exchanger Letdown heat exchanger 9.3.4 1/1 1,000/4.8 NA 4.8 Seal water heat exchanger 9.3.4 1/1 375/1.88 NA 1.88 Rev. 15 WOLF CREEK TABLE 9.2-10 (Sheet 2)
'A' Train 'B' Train Flow (gpm)/ Flow (gpm)/
Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr)(1)(2) Btu/hr)(2) (x 10 6 Btu/hr) Recycle evaporator 9.3.6 1/0 0/0 NA 0 Package (15) (16) Waste gas compressor (5) 11.3 2/1 100/0.135 NA 0.135 Aux. steam radiation 1/1 46/0.23 NA 0.23
monitor RE-50 Deleted (7) Catalytic hydrogen 11.3 2/1 20/0.075 NA 0.075
recombiner (5) Nuclear sample system 9.3.2 1/1 82.4/0.63 (8) NA 0.63
sample cooler Waste evaporator package 11.2 1/0 0/0 NA 0 (15) (16) Secondary waste evaporator 10.4.10 1/0 0/0 NA 0.0
Package (15) (16) Reverse osmosis unit (6) 11.2 1/0 0/0.0 (6) NA 0.0 (6) Total 11,780/126.97 7,684/117.11 244.08
NOTE: (1) Two CCW pumps are operating in Train A, for cooling nonessential loads. Nonessential loads can be fed from either train.
(2) Peak duty is shown. Heat load is lower as shutdown progresses. (3) Peak duty from fuel storage pool cooling system is 13.3 x 10 6 Btu/hr during refueling. This load can be shed for up to 4 hours after initiation of RHR load. As the RHR heat load is reduced, or selected nonessential loads are dropped, the fuel pool cooling may be placed back into service as indicated in Section 9.2.5. (4) During emergency shutdowns, prior to placing the RHR system into service at 4 hours after shutdown, the seal water heat exchanger is required to cool a maximum of 120 gpm (Btu/hr) of water recirculated from the discharge of the charging pumps; and the excess letdown heat exchanger is required to cool a maximum of 60 gpm (Btu/hr) letdown flow to the PRT. (5) Each of the two components is supplied with cooling water at a rate of one-half of the flow rate shown; however, only one of the components is accepting a heat load. (6) Reverse Osmosis Unit has been removed, however there is a flow loop via throttling EGV0423. EGV0423 is locked shut. (7) PASS is no longer used due to Amendment 137. (8) Flow as low as 60 gpm is acceptable for use. (9) The flows and related duty for these components are nominal values used for calculational purposes. Actual flows and duty may vary with actual plant operating conditions. (10) Deleted. (11) Deleted. (12) Deleted.
(13) 170 gpm for -15°F T hot Reduction (14) Deleted. (15) Flow to these components is permanently isolated. (16) Equipment is abandoned in place. Rev. 29 WOLF CREEK TABLE 9.2-11 COMPONENT COOLING WATER SYSTEM REQUIREMENTS POST LOCA
'A' Train 'B' Train Flow (gpm)/ Flow (gpm)/
Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr)(1) Btu/hr)(1)
(x 10 6 Btu/hr)Essential Components Residual heat removal heat 5.4.7 & 6.3 2/2 7,600/164 7,600/164 (5) exchanger RHR pump seal cooler 5.4.7 & 6.3 2/2 4/0.03 4/0.03 0.06 Centrifugal charging pump 9.3.4 2/2 55/0.08 55/0.08 0.16 bearing oil cooler Safety injection pump 6.3 2/2 25/0.024 25/0.024 0.05 bearing oil cooler Fuel pool cooling heat 9.1.3 2/1 0/0 0/0 0.0 exchanger (2)
Excess letdown heat exchanger 9.3.4 1/0 0/0 NA 0.0 Reactor coolant pumps 5.0 4/0 0/0 NA 0.0 Motor air coolers Upper bearing coolers
Lower bearing coolers Thermal barrier cooling coils Nonessential Components Reactor coolant drain 11.2 1/0 0/0 NA 0.0 Tank heat exchanger Letdown heat exchanger 9.3.4 1/0 0/0 NA 0.0 Seal water heat exchanger 9.3.4 1/0 0/0 NA 0.0 Rev. 15 WOLF CREEK TABLE 9.2-11 (Sheet 2)
'A' Train 'B' Train
Flow (gpm)/ Flow (gpm)/ Section Number/ Duty (x 10 6 Duty (x 10 6 Total Duty Equipment Description Number In Use Btu/hr)(1) Btu/hr)(1) (x 10 6 Btu/hr) Recycle evaporator package(6, 7) 9.3.6 1/0 0/0 NA 0.0 Waste gas compressor 11.3 2/0 0/0 NA 0.0
Aux. steam radiation 1/0 0/0 NA 0.0 monitor RE-50
Deleted (3)
Catalytic hydrogen 11.3 2/0 0/0 NA 0.0
recombiner Nuclear sample system 9.3.2 1/0 0/0 NA 0.0
sample cooler Waste evaporator package (6, 7) 11.2 1/0 0/0 NA 0.0 Secondary waste evaporator 10.4.10 1/0 0/0 NA 0.0 Package (6, 7) Reverse osmosis unit (8) 11.2 1/0 0/0 NA 0.0 Total 7,684.9/164.14 7,684.0/164.13 (5)
NOTE: (1) Peak duty is shown representing start of recirculation mode. Total duty is reduced long term.
(2) Flow rate and heat load for fuel pool cooling heat exchangers are not included in totals, since they are not added until 4 hours after start of the recirculation mode when heat from other sources has been significantly reduced, as shown in Section 9.2.5. (3) PASS is no longer used due to Amendment 137. (4) Deleted. (5) The RHR heat exchanger duties given are limiting design values corresponding to a single train cooldown.
Total RHR heat exchanger peak duty for a two train LOCA cooldown is 240 MBTU/hr (120 MBTU/hr per train, see Figure 9.2-6C). Since the other CCW loads are relatively small, CCW system peak duties are essentially the same as for the RHR heat exchangers themselves. (6) Equipment is Abandoned in Place. (7) Flow to these components is permanently isolated. (8) Reverse Osmosis Unit has been removed, however there is a flow loop via trottling EGV0423. EGV0423 is locked shut. Rev. 29 WOLF CREEK TABLE 9.2-12 COMPONENT COOLING WATER SYSTEM COMPONENT DATA Component Cooling Water Pump (all data is per pump) Quantity 4 (100% each) Type Horizontal centrifugal, split case dual volute with mechanical
seals Capacity, gpm (each) 11,025 TDH, ft 195
NPSH required, ft 20 NPSH available, ft (min.) 40 Material
Case Carbon steel Impeller Bronze Shaft Alloy steel
Design codes ASME Section III, Class 3 Driver Type Electric motor
Horsepower, hp 700 with a 1.0 service factor RPM 1,180 Power supply 4,160 V, 60 Hz, 3-phase, Class IE Design code NEMA Seismic design Category I Component Cooling Water Heat Exchangers (all data is per exchanger) Quantity 2 (100% each) Type Horizontal shell and straight tube
Design duty normal operation 77.18 x 10 6 (each), Btu/hr U-Factor, Btu/hr-ft 2 , F Clean 580 Dirty 221 Area, ft 2 31,900 Tube Side: Fluid Service water/essential service water
Number of passes 2 Temperature, in/out F 95/106.4 Flow rate, gpm 13,500
Design pressure, psig 200 Design temperature, F 200 Material
Tubes Copper-nickel Tube sheet Carbon steel Codes and standards ASME Section III, Class 3, TEMA R Seismic design Category I Rev. 19 WOLF CREEK TABLE 9.2-12 (Sheet 2) Shell Side: Fluid CCW Number of passes 2 Temperature, in/out F 119.7/105
Flow rate, gpm 10,500 Design pressure, psig 150 Design temperature, F 200 Material Carbon steel Codes and standards ASME Section III, Class 3, TEMA R
Seismic design Category I Component Cooling Water Surge Tank Quantity 2 Type Vertical
Capacity (each), gallon 5,000 Operating pressure/temp- Atm/120 erature, psig/F
Design pressure/temperature, 150/200 psig/F Material Carbon steel
Code ASME Section III, Class 3 Seismic design Category I Component Cooling Water Chemical Addition Tank Quantity 1 Type Vertical Capacity, gallons 500 Operating pressure/temp- 90/ambient erature, psig/F Design pressure/temperature, 150/200 psig/F
Material Carbon steel Code ASME Section VIII Piping, Fitting, and Valves Design pressure, psig 150 Design temperature 200 Material Carbon steel Design code
Containment penetrations ASME Section III, Class 2 Safety-related portion ASME Section III, Class 3 Nonsafety-related portion ANSI B31.1 Rev. 0 WOLFCREEKTABLE9.2-13COMPONENTCOOLINGWATERSYSTEMSINGLEACTIVEFAILUREANALYSISComponentFailure Comments1.CCWpumpFailstostartonautomaticsignal.Fourpumpsareprovided.Onepumpissufficientforpost-LOCAheatremoval.2.CCWheatexchangerbypassControlloopdrivesvalvefullopen.Twoseparatedcoolingloopsareprovided. valveOneloopprovides100%coolingcapacity3.Motor-operatedisolationvalveFailstoopenonautomaticsignalTworesidualheatexchangersandtwoonresidualheatexchangerpostLOCA.flowpathsareprovided.Flowisrequired inlettoonlyoneRHRheatexchangerpostLOCA.4.CCWflowpath,includingheatFailureofpressureboundaryre-Twoseparatecoolingloopsareprovided.exchangershellsultinginabnormalleakageandOneloopprovides100%coolingcapacity.lossofsystemfluid.5.PowersupplyFailureofbothnormalandpreferredAllClassIEcomponentsautomaticallypowersupplies.switchtooperationfrompowersuppliedfromemergencydieselgenerators.6.PowersupplyFailureofpowersupplybustooneTheothertrainissuppliedfromantrain.independentandphysicallyseparatedbus.Eachtrainprovides100%cooling
capacity.7.EssentialservicemakeupwaterFailuretoopenvalveifmakeupisTwoseparatecoolingloopsareprovided.supplyvalvesrequired.Makeuptoeitherloopissufficient.8.Motor-operatedisolationOnevalvefailstocloseonSISorTwovalvesareprovidedinseries.Onevalvesforsupplytonon-lowsurgetanklevel.valveclosingprovidesisolation. essentialcomponents9.Motor-operatedisolationOnevalvefailstocloseonSISorTwovalvesareprovidedinseries.Onevalvesforreturnfromnon-lowsurgetanklevel.valveclosingprovidesisolation. essentialcomponents10.Motor-operatedisolationOnevalvefailstocloseonCIS-B.Twovalvesareprovidedinseries.Onevalvesforsupplytovalveclosingprovidesisolation.essentialcomponentsinsideEithervalveclosesuponreceiptofcontainmentspurioussignal.Valvesareprovidedinparallel.Openingvalvewithin10minutesissufficientforRCPmotorbearingheatremoval. CVCSsealinjectionprovidesdiverse coolingforRCPseals.11.ContainmentisolationvalvesOnevalvefailstoclose.Twovalves(onecheckandonemotorforsupplytoessentialoperatedclosedbyCIS-B)areprovidedcomponentsinsidecontainmentinseries.Onevalveclosingprovides isolation.12.Motor-operatedvalveonFailstocloseonautomaticTwoRHRheatexchangersandtwoflowoutletoffuelstoragesignalpostLOCApathsprovided.FlowisrequiredtopoolcoolingheatonlyoneRHRheatexchangerpostLOCA. exchanger.Rev.14 WOLF CREEK TABLE 9.2-13 (Sheet 2) Component Failure Comments
- 13. Containment isolation valves One valve fails to close on CIS-B. Two valves are provided in series. One for return line for reactor valve closing provides isolation.
coolant pump thermal barrier Either valve closes upon receipt
of spurious signal. CVCS seal injection provides diverse cooling for RCP seals. Long-term requirements are met by opening parallel
valves.
- 14. Containment isolation valves One valve fails to close on CIS-B. Two valves are provided in series. One
for return line for other valve closing provides isolation. essential components in- Either valve closes upon receipt side containment of spurious signal. Valves are provided in parallel.
Opening valve within 10 minutes sufficient for RCP motor bearing heat removal.
- 15. CCW to Radwaste Return One valve fails to close following a Two valves are provided in series. One Check Valves (EGV0448 break in the non safety related valve closing provides isolation. And EGV0449) radwaste service loop portion of CCW piping.
Rev. 29 WOLF CREEK TABLE 9.2-14 COMPONENT COOLING WATER SYSTEM, INDICATING AND ALARM DEVICES Control Room Indication Control Room Local Alarm CCW heat exchanger flow Yes Yes No CCW heat exchanger inlet temperature Yes Yes No CCW heat exchanger outlet temperature Yes Yes Yes CCW pump suction pressure No Yes No
CCW pump discharge pressure Yes Yes Yes
CCW motor running lights Yes No No
CCW heat exchanger inlet pressure No Yes No CCW heat exchanger outlet pressure No Yes No CCW flow to redundant safety-related equipment trains Yes Yes Yes CCW flow to incontainment service Yes Yes Yes CCW temperature out of safety-related equipment No Yes No CCW surge tank level Yes Yes Yes
Radiation level of fluid Yes No Yes CCW flow to RCPs No Yes Yes Rev. 0 WOLF CREEK TABLE 9.2-15 MAJOR COMPONENTS SUPPLIED WITH WATER FROM DEMINERALIZED WATER STORAGE AND TRANSFER SYSTEM A. Condensate storage tank B. Reactor makeup water storage tank
C. Component cooling water system
D. Closed cooling water system
E. Auxiliary steam system
F. D-G cooling water expansion tank
G. Chilled water system
H. Hot water system
I. Miscellaneous laboratory and sampling requirements
J. Miscellaneous flushing requirements
K. Miscellaneous makeup requirements
L. Condensate pump seals
M. Condensate and chemical addition Rev. 0 WOLF CREEK TABLE 9.2-16 PLANT WATER CHEMISTRY SPECIFICATIONS(1)
(typical)
Demineralized Condensate Reactor Makeup Refueling Water _ Item/Service Water Storage Tk Water Tank Storage Tank (3) Fuel Pool (3)pH @ 25C 6.0 - 8.0 6.0 - 8.0 6.0 - 8.0 4.0 - 4.7 4.0 - 8.0 Cation conductivity <1.0 <1.0 <1.0 - -(6) @25C ( mho)/cm Specific conductivity - - <2.0 - - @ 25C ( mho)/cm Sodium (ppm) <0.01 <0.01 - - -
Silica (ppm) <0.1 <0.2 <0.1 <0.3 -
Chlorides (ppm) (4) - (4) <0.15 <0.15 Fluorides (ppm) (4) - (4) <0.15 <0.15 Boric acid (ppm B) - - <1.0 2000 +/- 502000 Potassium (ppm) <0.01 - <0.01 - -
Aluminum (ppm) <0.02 - <0.02 <0.08 -
Calcium (ppm) <0.02 - <0.02 <0.08 <1.0 Magnesium (ppm) <0.02 - <0.02 <0.08 <1.0 Dissolved oxygen (ppm) - <0.1 <0.10 - -
Suspended solids (ppm)(5) <0.1 <0.1 <0.1 <2.0 -
Total solids (ppm) <0.5 - <0.5(2) - NOTES: (1) Actual plant water chemistry specifications can be found in the WCGS Chemistry Specification Manual. (2) Excluding boric acid.
(3) Makeup water must meet the specification for reactor makeup water, except for dissolved
oxygen.
(4) Total chlorides and fluorides must be <0.10.__ (5) Solids concentration determined by filtration through filter having an 0.45 micron pore
size.
(6) No specification requirement for items marked by a dash (-). Rev. 12 WOLF CREEK TABLE 9.2-17 ESW/UHS COOLING WATER CHEMISTRY ANALYSIS Cations, PPM CaCO3 Calcium, Ca 948 Magnesium,Mg 385
Sodium, Na 220
Potassium, K 12
Total cations 1,565 Anions, PPM CaCO 3 Bicarbonate, HCO 3 66 Sulfate, SO 4 1,322 Chloride, CI 170 Nitrate, NO 3 7 Phosphate, PO (Negl.)
Total anions 1,565 TDS, PPM 1,700 pH 7.4
Silica, SiO 2 PPM 1.5 Fe + Mn, PPM 1.4
Ammonia nitrogen, PPM (Negl.)
Suspended solids, PPM 15-50 Makeup source Redmond Res'v'r.
Concentration cycles 3
Drought Rev. 0 WOLF CREEK TABLE 9.2-18 DELETED Rev. 7 WOLF CREEK TABLE 9.2-19 HEAT LOADS FROM STATION AUXILIARIES POST LOCA Actual Duty Analyzed Duty Section (Per Operating Train) (Per Operating Train) Component Number (x 10 6 Btu/hr) (x 10 6 Btu/hr)RHR pump seal cooler 5.4.7 & 6.3 0.03 0.07 Centrifugal charging pump bearing oil cooler 9.3.4 0.08 0.10 Safety injection pump bearing oil cooler 6.3 0.024 0.10 Diesel generator cooler 9.5.5 16.8 18.39 Component cooling water pump room cooler 9.4.3 0.280 0.32 Centrifugal charging pump
room cooler 9.4.3 0.171 0.32 Auxiliary feedwater pump
room cooler 9.4.3 0.28 0.32 Safety injection pump
room cooler 9.4.3 0.165 0.22 RHR pump room cooler 9.4.3 0.17 0.22 Containment spray pump room cooler 9.4.3 0.174 0.22 Penetration room cooler (Electrical) 9.4.3 0.12 0.10 Fuel pool cooling pump room cooler 9.4.2 0.075 0.07 Control room ac unit condenser 9.4.1 0.66 0.66 Class IE switchgear ac condenser 9.4.1 0.49 0.49 Air compressor and after cooler 9.3.1 0.60 0.30 Total 20.119 21.90 Rev. 20 WOLF CREEK TABLE 9.2-20 HEAT LOADS FROM STATION AUXILIARIES NORMAL SHUTDOWN Actual Duty Section (Per Operating Train) Analyzed Duty Component Number (x 10 6 Btu/hr) (x 10 Btu/hr) Reactor coolant pumps 5.0 2.1 2.375 Motor air coolers Upper bearing coolers Lower bearing coolers Thermal barrier cooling coils RHR pump seal cooler 5.4.7, 6.3 0.03 0.03
Centrifugal charging pump 9.3.4 0.0755 0.0755 bearing oil cooler Letdown heat exchanger 9.3.4 4.80 4.80 Seal water heat exchanger 9.3.4 1.88 1.88 Nuclear sample system sample cooler 9.3.2 0.63 0.63 Diesel generator cooler 9.5.5 16.8 16.8
Component cooling water pump room cooler 9.4.3 0.28 0.32 Centrifugal charging pump room cooler 9.4.3 0.162 0.32 Auxiliary feedwater pump
room cooler 9.4.3 0.25 0.32 RHR pump room cooler 9.4.3 0.17 0.22 Penetration room cooler 9.4.3 0.170 0.170 Fuel pool cooling pump room cooler 9.4.2 0.06 0.07 Control room ac unit
condenser 9.4.1 0.66 0.66 Class IE switchgear ac condenser 9.4.1 0.49 0.49 Air compressor and after-cooler 9.3.1 0.60 0.60 Total 29.16 29.76 Rev. 15 WOLFCREEKTABLE9.2-21(Sheet1of7)DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.27,REVISION2DATEDJANUARY1976,TITLED"ULTIMATEHEATSINKFORNUCLEARPOWERPLANTS"RegulatoryGuide1.27PositionWCGSPositionI.1.TheultimateheatsinkshouldbeI.1.Compliescapableofprovidingsufficientcool-RefertoSection9.2.5.1ingforatleast30days(a)topermitsimultaneoussafeshutdownandcooldownofallnuclearreactorunitsthatit servesandtomaintaintheminasafeshutdowncondition,and(b)intheevent ofanaccidentinoneunit,tolimitthe effectsofthataccidentsafely,topermitsimultaneousandsafeshutdownof theremainingunits,andtomaintaintheminasafeshutdowncondition. Proceduresforensuringacontinued capabilityafter30daysshouldbe available.Sufficientconservatismshouldbeprovidedtoensurethata30-daycoolingsupplyisavailableandthatdesignbasistemperaturesofsafety-related equipmentarenotexceeded.Forheat sinkswherethesupplymaybelimited and/orthetemperatureofplantintake waterfromthesinkmayeventually becomecritical(e.g.,ponds,lakes, coolingtowers,orothersinkswhere recirculationbetweenplantcooling waterdischargeandintakecanoccur), transientanalysesofsupplyand/or temperatureshouldbeperformed.2.Themeteorologicalconditionsresultin2.CompliesmaximumevaporationanddriftlossshouldRefertoSection9.2.5.3Rev.0 WOLFCREEKTABLE9.2-21(Sheet2of7)RegulatoryGuide1.27PositionWCGSPositionbetheworst30-dayaveragecombinationofcontrollingparameters(e.g.,dewpointdepression,windspeed,solarradiation.Themeteorologicalconditionsresultinginminimumwatercoolingshouldbetheworstcombinationofcontrolling parameters,includingdiurnalvariations whereappropriate,forthecriticaltime period(s)uniquetothespecificdesign ofthesink.Thefollowingareacceptablemethodsforselectingtheseconditions:a.Basedonregionalclimatologicalinformation,selectthemostsevereobservationforthecriticaltime period(s)foreachcontrolling parameterorparametercombination, withsubstantiationconservatismof thesevaluesforsiteuse.The individualconditionsmaybecombinedwithoutregardtohistorical occurrence.b.Selectthemostseverecombinationofcontrollingparameters,including diurnalvariationswhereappropriate, forthetotalofthecriticaltime period(s),basedonexaminationof regionalclimatologicalmeasurementsthataredemonstratedtoberepresentativeofthesite.If significantlylessthan30yearsofRev.0 WOLFCREEKTABLE9.2-21(Sheet3of7)RegulatoryGuide1.27PositionWCGSPositionrepresentativedataareavailable,otherhistoricalregionaldatashouldbeexaminedtodetermine controllingmeteorologicalconditionsforthecriticaltime period(s).Iftheexaminationof otherhistoricalregionaldata indicatesthatthecontrolling meteorologicalconditionsdidnot occurwithintheperiodofrecordfortheavailablerepresentativedata,thentheseconditionsshould becorrelatedwiththeavailable representativedataandappropriate adjustmentsshouldbemadeforsite conditions.c.Lessseveremeteorologicalcondi-tionsmaybeassumedwhenitcanbe demonstratedthattheconsequences ofexceedinglesserdesignbasis conditionsforshorttimeperiods areacceptable.Informationon magnitude,persistence,and frequencyofoccurrenceof controllingmeteorological parametersthatexceedthedesign basisconditions,basedon acceptabledataasdiscussedabove, shouldbepresented.Theaboveanalysisrelatedtothe30-daycoolingsupplyandtheexcesstemperatureshouldinclude sufficientinformationto substantiatetheassumptionsandRev.0 WOLFCREEKTABLE9.2-21(Sheet4of7)RegulatoryGuide1.27PositionWCGSPositionanalyticalmethodsused.Thisinformationshouldincludeactualperformancedataforasimilarcoolingmethodoperating underloadnearthespecifieddesignconditionsorjustificationthat conservativeevaporationanddriftloss andheattransfervalueshavebeenused.3.Acoolingcapacityoflessthan30days3.Notapplicablemaybeacceptableifitcanbedemon-stratedthatreplenishmentoruseofanalternatewatersupplycanbeeffected toassurethecontinuouscapabilityof thesinktoperformitssafety functions,takingintoaccounttheavailabilityofreplenishmentequipment andlimitationsthatmaybeimposedon "freedomofmovement"followingan accidentortheoccurrenceofsevere naturalphenomena.II.1.Theultimateheatsinkcomplex,whetherII.1.Compliescomposedofsingleormultiplewatersources, shouldbecapableofwithstanding,withoutlossofthesinksafetyfunctionsspecifiedinRegulatoryPositionI,followingevents:a.Themostseverenaturalphenomenaexpectedatthesite,withappropriateambientconditions,butwithnotwoor moresuchphenomenaoccurring
simultaneously,Rev.0 WOLFCREEKTABLE9.2-21(Sheet5of7)RegulatoryGuide1.27PositionWCGSPositionb.Thesite-relatedevents(e.g.,transportationaccident,riverdiversion)thathistorically haveoccurredorthatmay occurduringtheplant lifetime,c.Reasonablyprobablecombina-tionsoflessseverenaturalphenomenaand/orsite-related events,d.Asinglefailureofmanmadestructuralfeatures,2.Ultimateheatsinkfeatures,which2.Notapplicableareconstructedspecificallyforthenuclearpowerplantandwhicharenot requiredtobedesignedtowithstandthe SafeShutdownEarthquakeortheProbableMaximumFlood,shouldatleastbedesignedandconstructedtowithstandtheeffectsoftheOperatingBasisEarthquake(asdefinedin10CFRPart100,AppendixA) andwaterflowbasedonseverehistoricaleventsintheregion.III.1.TheUltimateHeatSink(UHS)shouldIII.1.TheUltimateHeatSink(UHS)isaconsistofatleasttwosourcesofwater,singlewatersourcefortheunit. includingtheirretainingstructures,TheUHSisseismicCategoryIand eachwiththecapabilitytoperformtheitsuppersurfaceislocated17 safetyfunctionsspecifiedinRegulatoryfeetbelowthenormalpooleleva-PositionI,unlessitcanbedemonstratedtionoftheWCGSCoolingLake.thatthereisanextremelylowprobabilityHence,thereisanextremelylowoflosingthecapabilityofasinglesource.probabilityoflosingitscapa-bility.Itisconsideredsingle failureproof.Rev.0 WOLFCREEKTABLE9.2-21(Sheet6of7)RegulatoryGuide1.27PositionWCGSPosition2.Forclose-loopcoolingsystemsthere2.Compliesshouldbeatleasttwoaqueductsconnectingthesource(s)withtheintakestructuresofthenuclearpowerunitsandatleasttwoaqueductstoreturnthecoolingwatertothesource,unlessit canbedemonstratedthatthereisextremelylowprobabilitythatasingle aqueductcanfunctionallyfailentirely asaresultofnaturalorsite-related
phenomena.3.Foronce-throughcoolingsystems,there3.Notapplicableshouldbeatleasttwoaqueductscon-nectingthesource(s)withtheintakestructuresofthenuclearpowerunitsand atleasttwoaqueductstodischargethecoolingwaterwellawayfromthenuclear powerplanttoensurethatthereisno potentialforplantfloodingbythedischargedcoolingwater,unlessitcan bedemonstratedthatthereisextremelylowprobabilitythatasingleaqueductcanfunctionallyfailasaresultof naturalorsite-relatedphenomena.4.Allwatersourcesandtheirassociated4.Compliesaqueductsshouldbehighlyreliableand shouldbeseparatedandprotectedsuch thatfailureofanyonewillnotinducefailureofanyother.Rev.0 WOLFCREEKTABLE9.2-21(Sheet7of7)RegulatoryGuide1.27PositionWCGSPositionIV.ThetechnicalspecificationsfortheplantIV.NoplantTechnicalSpecificationsareshouldincludeprovisionsforactionstorequiredbecause:(1)noconditionsbetakenintheeventthatcapabilityofthethreatenpartiallossofUHScapability UltimateHeatSinkortheplanttemporarilyand(2)theplantsatisfiesRegulatorydoesnotsatisfyRegulatoryPositionsIandPositionsIandIIIduringoperation. IIIduringoperation.Rev.0 WOLF CREEK TABLE 9.2-22 COMPONENTS AND SYSTEMS SERVED BY CONDENSATE STORAGE AND TRANSFER SYSTEM USAR COMPONENT/SYSTEM SECTION Condenser air removal system 10.4.2 Condenser hotwells 10.4.1
Condensate demineralizer 10.4.6
Auxiliary steam condensate recovery and storage tank 9.5.9 Auxiliary feedwater pumps 10.4.9 Rev. 0 WOLF CREEK TABLE 9.2-23
SUMMARY
OF REACTOR MAKEUP WATER REQUIREMENTS Minimum(1) Required Connection to System Flow (gpm) Purposes Boric acid blending tee(2) 120 To dilute the con-centrated boric acid as required. Boric acid blending tee 120 To supply makeup water to the refueling water
storage tank (RWST). Chemical mixing tank 1 For chemical addition. Boric acid batch tank 80 Used in the production of the boric acid
solution. Boron thermal regeneration 60 Alternate bed re-demineralizers generation. Emergency boration fill 5 To flush the line. line Recycle evaporator package 11 To flush the package. (3) Recycle evaporator reagent 5 To flush the tank (3) tank and to provide makeup water. Recycle evaporator 55 Water cleanup of (3) condensate demineralizer RMWST. Pressurizer relief tank 150 @ 90 For alternate cooling. psig Reactor coolant pump 10 Provide periodic de-standpipes gassed purge water to the RCP #3 seal on demand. Chemical drain tank 5 To flush waste from drumming line back into tank. Rev. 14 WOLF CREEK TABLE 9.2-23 (Sheet 2) Minimum(1) Required Connection to the System Flow (gpm) PurposesWaste evaporator package 11 To flush package. (3) Waste evaporator reagent 5 To flush tank and to tank provide makeup water. Catalytic hydrogen 5 To force gases out of recombiner equipment prior to maintenance. Waste gas compressor 5 Compressor seal usage. Gas decay tanks 30 @ 100 Displace gas in decay psig tanks prior to maintenance. Liquid radwaste demineralizer 20 Flush skid Evaporator bottoms tank & 20 To flush tank pump and pump (primary & secondary) associated piping for maintenance. Spent resin storage tank 20 To flush waste from (secondary) drumming line back into
tank and to provide demineralizer sluicing water to the tank. Resin charging tanks 20 each To provide sluicing (radwaste & CVCS) water Spent resin storage tank 20 To flush waste from (primary) drumming line back into
tank and to provide demineralizer sluicing water to tank. Sample sinks 5 General laboratory requirements. Fuel storage pool 20 Fuel storage pool water makeup. Cask washdown pit 40 Decontamination of the spent fuel shipping
cask. Rev. 19 WOLF CREEK TABLE 9.2-23 (Sheet 3) Minimum(1) Required Connection to the System Flow (gpm) Purposes Spray booth 40 For preliminary decontamination prior to use in the chemical
tanks. Decontamination exhaust 0.5 Water supply to scrubbers scrubber unit. Reactor vessel head 40 Decontamination storage area of the reactor vessel head. Secondary liquid waste 22/55 To provide makeup (3) evaporator and periodic flushing.
Demineralized water 120 To remove dissolved degasifier oxygen in reactor makeup water. Electrical control and 3 To provide makeup hydraulic power unit to ECH power unit reservoir. Notes: (1) Intermittent services. Atmospheric pressure at the connection unless otherwise specified. (2) Maximum letdown rate. (3) Equipment permanently out of service. Rev. 14 WOLF CREEK TABLE 9.2-24 COMPONENTS COOLED BY THE CLOSED COOLING WATER SYSTEM Flow Total Duty Total Equipment Number/ Each Flow Each Duty Description in Use (gpm) (gpm) (Btu/hr) (Btu/hr) Generator isophase 2/1 149 149 1.5 x 10 6 1.5 x l0 6 bus duct coolers
Steam generator 4/2 240 480 6.0 x 10 5 1.2 x l0 6 feed pump turbine lube oil coolers
Generator exciter 1/1 205 205 3.6 x l0 5 3.6 x l0 5 air cooler
EHC coolers 2/1 30 60 4.3 x 10 4 4.3 x 10 4 (1) Condensate pump 3/3 8 24 1.5 x 10 4 4.6 x 10 4 motor bearing oil coolers Secondary system 12/12 7 (4) 80 Varies 8.5 x 10 5 sample coolers Heater drain pump 2/2 6 12 3.4 x 10 4 6.8 x 10 4 motor bearing oil coolers Auxiliary boiler and 2/2 7 14 4.9 x 10 3 9.7 x 10 3 (5) auxiliary steam reboiler sample coolers Steam generator wet 4/4 7 28 4.9 x 10 3 1.9 x 10 4 (2) layup recirculation sample coolers Feedwater corrosion 1/1 4 4 5.2 x 10 4 5.2 x 10 4product sample cooler Heater drain Corrosion product Sample cooler 1/1 4 4 5.2 x 10 4 5.2 x 10 4 Degasifier Vacuum 2/1 3 3 1.9 x 10 4 1.9 x 10 4 (2) Pumps SGFP HPU coolers 4/2 10 20 2.5 X 10 4 5.0 x 10 4 Total 1,045 (3) 4.22 x 10 6 NOTES: (1) Flow is to two coolers but heat load is only from one cooler.
(2) Used only during plant shutdown, not included in total duty or total flow. (3) Total system flow may differ slightly from pump design flow.
(4) Sample Cooler ERM13 has a flow rate of 3 GPM.
(5) The Aux Boiler Sample Cooler is normally not in operation and is not included in total duty or total flow.
Rev. 27 WOLF CREEK Table 9.2-25 MINIMUM ESW TOTAL SYSTEM TEMPERATURE RISE Case 1 The minimum possible ESW temperature rise is 0.55F. This could occur on one train of ESW after a cold shutdown when containment coolers are no longer needed or required and the corresponding train of CCW is not operating. 0.55F is due to the energy input from the ESW pump. Case 2 The minimum design ESW temperature rise is 0.72F. This could occur after a cold shutdown when containment coolers are no longer needed or required and the corresponding train of CCW is operating without any significant load. 0.72F is due to the energy input from the ESW pump (0.55F) and one CCW pump (0.17F). Case 3 The minimum ESW temperature rise in Modes 1, 2, 3 or 4 would be 0.87F for the train not carrying CCW. 0.87F is due to the energy input from the ESW pump (0.55F) and one containment air cooler (0.32F). In Modes 1, 2, 3 or 4 all four containment coolers are normally in operation, but by design only three containment coolers are needed for normal operation in the winter. Likewise, only one train of CCW is required to be in operation. Case 4 The minimum ESW temperature rise in Modes 1, 2, 3 or 4 would be 2.98F for the train carrying CCW. 2.98F is due to the energy input from the ESW pump (0.55F), one CCW pump (0.17F), normal CCW system loads (1.62F), and two containment coolers (0.32F each). Case 5 The minimum ESW temperature rise following loss of off site power is 1.17F. 1.17 is due to the energy input from the ESW pump (0.55F) and one EDG (0.62F). Note Other accidents or events may differ, but the minimum ESW temperature will be bounded by Case 3 when in Modes 4 or above, and by case 2 in Modes 5 or below. Rev. 10 WOLF CREEK TABLE 9.2-26 CONDENSATE STORAGE AND TRANSFER SYSTEM COMPONENT DATA Non-Safety Auxiliary Feedwater Pump Quantity 1 Type Horizontal centrifugal, multistage, split case Nominal capacity, gpm 500 TDH, ft. 3,460 NPSH required, ft. 32 NPSH available, ft. 45 (nominal) Material Case Alloy steel Impellers Stainless steel Shaft Stainless steel Design code Not applicable Seismic design Non-seismic Driver Type Electric motor Horsepower, hp 700 RPM 3,575 Power supply 4,160 V, 60 Hz, 3-phase Design code NEMA Seismic design Non-seismic Rev. 27
- 1,000.0 :::1: -::::>> tD -w 100.0 i 8 :::1: ---HEAT REJECTION
ANALYSIS INPUT 10.0 100 1,000 10,000
- 100,000 TIME (Secondst WOlf CfiEEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-6A HEAT REJECTION RATE TO ULTIMATE HEAT SINK 30 DAYS 1,000,000
- Rev. 7 10,000,000
-a: :I: ---:::> 1-al :E -w 1-<( a: ...J <( > 0 :E w a: 1-<( w :I: a: w ...J 0 0 u a: <(
- 300 250 200 150 100 50 0 1E+00 1E+01 1E+02 WOLFCREEK UPOA TED SAFETY ANALYSIS REPORT FIGURE 9.2-68 DOUBLE ENDED PUMP SUCTION GUILLOTINE BREAK, MAXIMUM SAFETv INJECTION, 4 AIR COOLERS, TOTAL 1E+03 1E+04 TIME (SECONDS)
- AIR COOLER HEAT REMOVAL RATE VS. TIME 1E+05 1E+06 Rev. 7 1E+07
- 300 .------------------------------------------------------------------------
Rev. 7 -0: 250 J: WOLFCREEK UPDATED SAFETY ANALYSIS IEPORT ....... => .,_ m -> t-:::> c 0: w (!) z <( J: u X w .,_ <( w J: 0: J: 0: 200 150 100 50 FIGURE 9.2-6C DOUBLE ENDED PUMP SUCTION GUILLOTINE BREAK, MAXIMUM SAFETY INJECTION, 4 AIR COOLERS, RHR HEAT EXCHANGER HEAT REMOVAL RATE VS. TIME 0 1E+01 1E+02 1"E+03 1E+04 1E+05 1E+06 1E+07 TIME (SECONDS)
8 4 si .... -0 0 1991 5 7 9 11 13 15 17 19 21 23 25 27 29 SEP 3 5 7 PLANT DISCHARGE AND INTAKE TEMPERATURES DURING LOCA WITH WORST TEMPERATURE WEATHER PERIOD WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-7 ULTIMATE HEAT SINK Re"-I N LET AN D 0 U T LET T E M P E RAT U RES WORST TEMPERATURES PERIOD " U H S IN LET T EM P E RAT U R E x U H S 0 U T LET T E M P E RAT U R E I-w w LL 0 0 0 ..... 2 0 <<> o) U) S2 0 <<> oi <<> 2 0 ot o) <0 S2 0 C"' oi <<> 2 0 0 oi <0 2 0 <<> o:) <<> S2 0 U) o:) U) 2 8 o:) U) "-...
- ELEVATION
'-....... Rev. 5 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-10 ULTIMATE HEAT SINK -ELEVATION WORST EVAPORATION PERIOD AUG 1901 3 5 7 9 FIG. 10 11 13 15 17 19 21 2.3 25 ELEVATION OF UHS DURING LOCA WITH WORST EVAPORATION WEATHER PERIOD 27 29
- 10 9 u 8 lL. 7 6 L-9 5 w > 4 w (.!) 3 I ! I l )"' 7 l v '__.11 i ! I I I i I I Wolf Creek I
- T i I I l I I I I I l ! I "' I '"---I ! I !-;--*-l I ' I i j_ I I * -+----+-----+-----+ , ! T j , I ' I I I I '/ I I I I I I I * ., : 'I 'i J" ....... ? I I l I I l I I j I i f i -I 1 lr I ! ! I I I 1 I 1 I I I I ! I : , ! ; I I JzJ I I t I t I I t -r r*---t r*j 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 TIME IN MINUTES (AFTER COOUNG LAKE DAM FAILURE) WOLF CREEK Rev .. 0 UPDATED SAFETY ANALYSIS REPORT I FIGURE I TRANSIENT VELOCITY AT WOLF CREEK . UHS LOCATION -CASE I I 10 9 . 8 7 t f 6 9 5 . 4 UJ 3 2 0 Wolf Creek I T l I I I 1 I l 1-1 I I I I --T I I --I /1 r I I I I I I J . I I _. J I L I T I\./\J61 /-\, -I -, i --I l I I I l I ... ,, I I I I l I I -, ..... I I l u I I I I 1 T I I ! I T I I I l I I I I I I I I I -i 0 2 4 I --I 6 8 10 12 14 16 18 20 22 24 26 28 30 TIME IN MINUTES (AFTER THE DIKE FAILLJ(E)
Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-12 TRANSIENT VELOCITY AT WOLF CREEK UHS LOCATION -CASE II
- 100,000.0 10,000.0 1,000.0 100.0 10.0 100 1,000 10,000 TIME (Seconds) 100,000 30 DAYS WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-18 INTEGRATED HEAT REJECTION TO ULTIMATE HEAT SINK LOCA
- Rev. 7 1,000,000 10,000,000 1000.0 iC :::1: -::;) Iii 100.0 s :::1: 10.0 100
- 1,000 ,---1 I I I I I I I I I I I I I I ---HEAT REJECTION
ANALYSIS INPUT 10,000 TIME (Seconds)
- 100,000 WOLF ClEEK UPDATED SAR1Y ANALYSIS REPORT FIGURE 9.2-19 HEAT REJECTION RATE TO ULTIMATE HEAT SINK SHUTDOWN 30 DAYS 1,000,000 10,000,000
- Rev. 7
- 100,000.0 10,000.0 1,000.0 100.0 10.0 100 1,000 10,000 TIME (Seconds) 100,000 30 DAYS WOU:CREEX UPDAtED SAFETY ANALYSIS REPORT FIGURE 9.2-20 INTEGRATED HEAT REJECTION TO ULTIMATE HEAT SINK SHUTDOWN Rev. 7 1,000,000 10,000,000 Q.. < a: ..... :; u.. 1000 9 8 7 6 5 2 100 9 8 7 6 5 Wolf Creek Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS FIGURE 9.2-21 EMERGENCY MAKEUP WATER REQUIREMENT
-LOCA ______ j __ . I I I I If/ i I I 1,-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 I I I I
- I I I I I I I I I I I 1 oo 2 3 4 5 s 7 e 9 1 ooo 2 3 4 5 s 7 a 9 1 o 4 2 3 4 5 s 1 a 9 1 o5 2 3 4 5 s 1 a 9 1 o6 2 3 4 5 s 1 a 9 TIME {SECONDS)
Wolf Creek 9 8 7 6 5 4 3 v; 2 0 z 0 u w w 9 8 ..... 7 v; 6 5 5 j 4 < (,!) M 0 2 w :..: < ::2 10 ..... 9 < 8 ffi 7 6 5 4 3 2 TIME (SECONDS) 30 DAYS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.2-22 INTEGRATED EMERGENCY MAKEUP WATER REQUIREMENT 2 3 4 567891 WOLF CREEK
9.3 PROCESS
AUXILIARIES
9.3.1 COMPRESSED
AIR SYSTEM The compressed air system (CAS) provides a reliable, continuous supply of filtered, dry, and oil-free air for instrument and control operations. The system also provides station air at service outlets throughout the plant for operation of pneumatic tools and other service requirements. The CAS provides a reliable backup supply of compressed gas for the main feedwater control valves. The system also provides a safety-related backup compressed gas supply for the auxiliary feedwater control valves and the main steam atmospheric relief valves. 9.3.1.1 Design Bases 9.3.1.1.1 Safety Design Bases The following safety design bases are applicable to the safety-related functions of containment isolation, the 8-hour backup air supply for the
auxiliary feedwater control valves and the 8-hour backup air supply for the
main steam atmospheric relief valves. SAFETY DESIGN BASIS ONE - Portions of the CAS are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - Portions of the CAS remain functional after an SSE and perform their intended function following postulated hazards of fire, internal missiles, or pipe break (GDC-3 and 4). SAFETY DESIGN BASIS THREE - Component redundancy is provided so that safety functions are performed, assuming a single active component failure coincident with the loss of offsite power (GDC-34). SAFETY DESIGN BASIS FOUR - Active components of the CAS 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. 9.3-1 Rev. 0 WOLF CREEK SAFETY DESIGN BASIS FIVE - The CAS is in accordance 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 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.9.3.1.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The CAS provides compressed air for service outlets located throughout the plant and a continuous supply of filtered, dried, and essentially oil-free air for pneumatic instruments and valves. POWER GENERATION DESIGN BASIS TWO - The combined air receiver storage capacity is adequate to supply instrument air requirements for a 28-second period. This is enough time for the standby compressor to come up to pressure in the event of an operating compressor failure. POWER GENERATION DESIGN BASIS THREE - The CAS has the capability of providing instrument air during loss of offsite power. POWER GENERATION DESIGN BASIS FOUR - The CAS serving the main feedwater control valves provides a nonseismically qualified, nonsafety-related backup compressed gas system that will provide 4 hours of reliable compressed gas in the event of
the loss of the normal air supply. 9.3.1.2 System Description 9.3.1.2.1 General Description The CAS includes four skidmounted, at least 100-percent-capacity air compressing trains. Three main units consist of an air inlet filter/silencer, a compressor unit, an aftercooler, an air receiver, and interconnecting piping
and valving. The three air receivers are connected in parallel by a common
header which branches into the instrument air and service air subsystems. Service air goes directly to distribution, while instrument air first passes through a drying/filtering train. A piping connection is provided to allow the
construction air connection source, with administrative restrictions, to be
used as an instrument air supply. All of the above components are located in
the turbine building. Localized compressed gas cylinders and control valves are provided for the backup gas supply systems. Safety-related pneumatically operated valves are listed in Table 9.3-2. The CAS is shown in Figure 9.3-1. In addition, the fourth air compressing train, located outside the turbine building provides additional air for the service air side of the Compressed Air System. This train consists of an air inlet filter/silencer, a compressor unit, an aftercooler, a refrigerated air dryer, and interconnecting piping and valving. This air compressor is located outside on the west side of the
Turbine Building. The refrigeration unit is located inside the Turbine
Building in the west corridor. 9.3-2 Rev. 16 WOLF CREEK 9.3.1.2.2 Component Description COMPRESSORS - The three 932 scfm air compressors are rotary, non-lubricated, water cooled, two-stage, motor driven units. Each of the compressors is rated to deliver 932 scfm at 125-psig discharge pressure. The compressors are sized
so that each can supply an adequate supply of air for average instrument air
requirements. Two of the compressors are non-Class IE devices which are
powered from different Class IE busses, and the third is powered from a non-Class IE bus. The two compressors powered by the Class IE busses are cooled by service water during normal plant operation and essential service water for all
loss of offsite power conditions. Both compressors are shed from the Class IE busses on a safety injection signal but may be realigned to the busses
manually. The third air compressor, powered from a non-Class 1E bus, is cooled by central chilled water system. A fourth air compressor is an air cooled motor driven, lubricated, rotary screw air compressor. This unit is rated to deliver 1500 scfm at system pressure, and is powered off the non-Class IE bus. This train is provided with air filtration and refrigerated air drying capability sufficient to meet Service Air Component user requirements. AFTERCOOLERS - Each of the three main compressors is provided with an aftercooler to cool the flow of air from the associated air compressor to 110 F. The aftercoolers are TEMA Class C air/water heat exchangers cooled from the same water system as the air compressors. The fourth air compressor has an
air/air after-cooler train backed up with a refrigerated air dryer. This air-cooler also cools the compressors lubricating oil. AIR RECEIVERS - Compressed air from the outlet of each aftercooler on the main air compressors flows through one of the three air receivers. The air
receivers serve as storage volume to supply a limited amount of compressed air
following a compressor failure. The combined volume of the three air receivers
is sized to provide greater than a 28-second supply of instrument air at rated flow while allowing air pressure to drop from 122 psig to no less than 80 psig and to allow time for a standby compressor to come up to pressure. The air receiver design pressure is 125 psig. DRYER/FILTER TRAIN - The instrument air dryer/filter train consists of a series arrangement of two parallel prefilters, two parallel dual tower dryer units, and two parallel afterfilters. One or both parallel trains may be in use as
required. Parallel filters allow cleaning or changing of filters during one
train system operation by diverting air flow through the parallel filter. Each
air dryer section consists of an interconnected set of two desiccant chambers. Air flow is automatically alternated through each chamber and the drying of desiccant in the other chamber. Drying of the desiccant is accomplished by
depressurizing the dessicant chamber and purging dry air through it. Permanent
valving is provided such that temporary connections can be given so that the
instrument air may be processed via a portable filter/dryer unit. 9.3-3 Rev. 18 WOLF CREEK ACCUMULATORS - The backup gas systems provided for the auxiliary feedwater control valves, main feed control valves, and main steam atmospheric relief
valves as part of the compressed air system utilize carbon steel accumulators to store nitrogen for use in the event of loss of operation of the CAS. The supply of nitrogen is from the service gas system (see Section 9.3.5). The main feedwater control valve accumulator and the auxiliary feedwater control valve/main steam atmospheric relief valves accumulators are designed to ASME
Section III, Class 3. Accumulators have design and operating pressures as noted in Table 9.3-1. The accumulator system is shown in Figure 9.3-1. The main feed pump mini flow valves receive backup gas pressure directly from the low
pressure nitrogen system. 9.3.1.2.3 System Operation The CAS provides a reliable, continuous supply of filtered, dry, and essentially oil-free instrument air for pneumatic instrument operation and the control of pneumatic valves. The CAS also supplies service air to service outlets throughout the plant for the operation of pneumatic tools and other service requirements. One of the three main air compressors is normally available for instrument air service at all times, and the other compressors are on standby. In the event
of loss of the operating compressor or heavy loads, the resulting low pressure initiates an automatic start of a standby compressor. If the pressure continues to drop, a second and if necessary, a third standby compressor comes on line. Automatic starts occur only on low pressure. The three instrument
air side compressors automatically load and unload in response to small system
pressure variations to minimize the amount of compressor starts and stops required. The compressor automatically shuts down after running unloaded for 15 minutes. The sequence of compressor starting can be varied to permit equal operating time for all three air compressors. System functions and abnormal conditions are annunciated in the control room. The discharge line from each air receiver is connected in parallel to a common header. The service air subsystem takes its supply from this common header and
from a fourth air compressor to a separate service air header for direct
distribution to the service air outlets located throughout the plant. The
instrument air subsystem is supplied from this common header and from the fourth air compressor, if needed, to the dryer/filter train, where the air is processed to the required cleanliness and dew point. The train provides air
meeting industry standards and design specifications. 9.3-4 Rev. 11 WOLF CREEK The service air line is provided with an isolation valve that will automatically isolate the service air subsystem from the compressed air supply
when the service air header pressure drops below 110 psig. This arrangement is provided to direct all of the compressed air to the instrument air subsystem to maintain instrument air pressure in the event of excessive demand. The accumulator backup gas system is designed to supply compressed gas to designated valves in the event of the loss of the normal instrument air supply. A check valve in the nitrogen supply line feeding each accumulator and a check valve in the instrument air supply line feeding the valve actuators prevent
stored N2 from being vented from the accumulator during any loss of pressure in the nitrogen feed and/or normal instrument air supply line. The pressure-reducing valve downstream of each accumulator is manually adjusted to supply accumulator air below the nominal instrument air line pressure. A relief valve is provided for each accumulator and in the pipe downstream of each pressure-reducing valve. 9.3.1.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases of Section 9.3.1.1. SAFETY EVALUATION ONE - The safety-related portions of the CAS 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 safety-related portions of the CAS 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 and 3.6 provide the hazards analyses to assure that a safe shutdown, as outlined in Section 7.4, can be achieved and maintained. 9.3-5 Rev. 19 WOLF CREEK SAFETY EVALUATION THREE - The safety-related portions of the CAS are completely redundant. Therefore, 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 CAS is initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.3.1.4. Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for the 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 systems. Table 9.3-1 shows that the components meet the design and fabrication codes given in Section 3.2. All the power supplies and controls necessary for safety-related functions of the CAS are Class 1E, as described in Chapters 7.0 and 8.0. SAFETY EVALUATION SIX - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability. 9.3.1.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. Leaktight integrity of the CAS is demonstrated by a hydrostatic test performed per the requirements of applicable codes. Air compressors and associated components on standby can be checked and operated periodically. Air filters are inspected for cleanliness, and the desiccant is changed when it no longer performs according to the manufacturer's specifications. During the initial plant testing prior to reactor startup, all engineered safety features systems utilizing compressed air were tested to ensure fail-safe operation upon loss of compressed air pressure. The compressed gas accumulator systems can be isolated from the regular compressed air system and tested to ensure proper operation characteristics. Inservice inspections are performed for the safety-related portions of the system per the technical requirements of ASME Section XI, as described in
Section 6.6. 9.3-6 Rev. 12 WOLF CREEK 9.3.1.5 Instrumentation Applications The compressors and associated equipment are provided with local control panels. Each panel consists of temperature and pressure switches, indicators, and automatic protection devices. Indicating lights are located in the control
room to indicate equipment status. Control room alarm is provided for air compressor header pressure. Service air/instrument air isolation valve status
is also provided in the control room. The instrument air dryer assembly consists of two dryer units in parallel. Each dryer is equipped with local pressure indicators. Local control panel and control room alarms are provided for high differential pressure across the dryer package, pre-filters, afterfilters, and high package discharge humidity and low package discharge pressure. Local hand switches are provided to permit the operators to open the standby dryer train isolation valves. Control room
indication is provided for instrument dryer header pressure. Control room indication of the pressure and low pressure alarms for each safety-related accumulator are provided along with alarms for low accumulator pressure. Local pressure indicators are provided downstream of the accumulator
pressure-reducing valves to permit local monitoring of the system supply
pressure.Local pressure indicators are provided for the air lines feeding the fuel storage pool transfer gate seals to permit local monitoring of the seal pressure.9.3.2 PLANT SAMPLING SYSTEMS The plant sampling systems consist of the following subsystems: 1) the nuclear sampling system, which is further divided into the primary sampling system (PrSS), and a radwaste sampling system (RWSS), 2) a process sampling system (PSS) for secondary side sampling, and 3) local grab sample provisions. These subsystems include equipment to collect representative samples of the various process fluids in a safe and convenient manner. The RWSS is located in the
radwaste building, the PrSS in the auxiliary building sample room, and the PSS
in the turbine building. These systems include sample lines, valves, coolers, and automatic analysis equipment. A description of the equipment comprising these systems and their features relating to safety is presented in this section. Certain process sampling components are discussed in other sections.
A safety-related containment hydrogen analyzer provided to monitor the containment atmosphere following a postulated LOCA is described in Section
6.2.5.5.4. A discussion of process radiation monitoring is provided in Section 11.5. A discussion of gas analysis associated with the gaseous radwaste hydrogen recombiner is provided in Section 11.3. 9.3-7 Rev. 14 WOLF CREEK 9.3.2.1 Design Bases 9.3.2.1.1 Safety Design Basis The plant sampling system serves no safety function and has no other safety design basis, except for a containment isolation provision. SAFETY DESIGN BASIS ONE - The containment isolation valves in the system are selected, tested, and located in accordance with 10 CFR 50, Appendix A, General Design Criteria 54, 55, and 56, and 10 CFR 50, Appendix J, Type C Testing. 9.3.2.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The PrSS is designed to collect representative samples of fluids in the reactor coolant system and auxiliary
system process streams, as listed in Table 9.3-3, for analysis by the plant
operating staff. Chemical and radiochemical analyses are performed on these
samples to determine:
- 1. Boron concentration
- 2. Fission and corrosion product activity levels
- 3. Dissolved gas concentration
- 4. Halide concentration
- 5. pH and conductivity levels
- 6. Fission gas content
- 7. Gas compositions in various vessels The results are used to:
- 1. Monitor core reactivity
- 2. Monitor fuel rod integrity
- 3. Evaluate ion exchanger and filter performance
- 4. Specify chemical additions to the various systems
- 5. Maintain acceptable hydrogen levels in the reactor coolant system.
9.3-8 Rev. 0 WOLF CREEK
- 6. Detect radioactive material leakage POWER GENERATION DESIGN BASIS TWO - The RWSS is designed to collect samples of the fluids in the radwaste systems, as listed in Table 9.3-4, for analysis by the plant operating staff. Chemical and radiochemical analyses are performed on
these samples to determine treatment or disposition of the collected batches. POWER GENERATION DESIGN BASIS THREE - The PSS is designed to continuously monitor water samples from the turbine cycle systems, as listed in Table 9.3-5. Water quality analyses are performed on these samples to determine:
- 1. pH and conductivity levels
- 2. Dissolved oxygen
- 3. Residual hydrazine
- 4. Sodium concentration The above measurements are used to control water chemistry and to permit appropriate corrective action by the operating staff. In addition, grab sample
capabilities are provided at each of these monitoring points to monitor other
chemical species. POWER GENERATION DESIGN BASIS FOUR - Local grab sampling stations, as listed in Table 9.3-6, are provided for process points which require heat tracing or
sampling at a frequency of not more than once a week. POWER GENERATION DESIGN BASIS FIVE - The PrSS, RWSS, and PSS are designed and built to the codes listed in Table 3.2-1. 9.3.2.2 System Description 9.3.2.2.1 Primary Sampling System
The PrSS collects samples from the reactor coolant system and the auxiliary systems, as listed in Table 9.3-3, and brings them to a common location in a sample room in the auxiliary building. The PrSS consists of a primary sampling rack and a sampling panel. To minimize the source volume exposed at the primary sampling panel, the sampling station components that retain potentially radioactive fluids, such as sample coolers, isolation valves, throttle valves, rod-in-tube flow control valves, and associated piping and tubing, are mounted
on the primary sampling rack. 9.3-9 Rev. 14 WOLF CREEK The rack is located behind a 2-foot-thick concrete wall which provides radiation shielding. The primary sampling panel, located in front of the
radiation shield wall, contains the grab sampling facilities. The PrSS is shown in Figure 9.3-2. The PrSS rack contains sample coolers which reduce the temperature of the samples to below 110 F (to permit the safe handling of
samples). The PrSS sample coolers are cooled by the component cooling water
system. Relief valves protect the system from overpressurization. After temperature and pressure reduction, the PrSS samples are routed to a manual sample facility within an exhaust-ventilated, hooded enclosure to
confine any leakage or spillage of radioactive fluids. Temperature and pressure indicators are provided to verify the sample conditions. Within the
vented sampling hood are grab sample points for each stream and the sample pressure vessels. Any liquid leakage is collected in the sink and drained to the floor drain tank or the holdup tank for processing through the liquid radwaste system. The PrSS is manually operated to provide samples for laboratory analysis, except for steam generator blowdown samples which are continuously monitored for radioactivity by one process radiation monitor (described in Section 11.5) common to the four samples. Sample lines are purged before each sample is drawn to ensure that representative samples are obtained. Continuous monitoring of the water quality of the steam generator blowdown sample is provided on the PSS. The steam generator blowdown sample lines are provided with solenoid valves which are closed automatically if radioactivity approaching the limits
discussed in Section 11.5 is detected in the steam generator sample, or if a containment isolation signal occurs. If the steam generator blowdown samples are needed after an automatic closure of the blowdown sample valves due to high radiation, the valves can be opened manually at the nuclear sampling panel. Continuous monitoring of the CVCS letdown line (failed fuel monitor) is discussed in Section 11.5. The operating conditions of the PrSS are given in Table 9.3-3. The high-pressure reactor coolant system samples are collected at full process pressure and reduced temperature in sample pressure vessels. Samples can also be taken
at reduced pressure through the rod-in-tube flow control valves. These vessels are designed for 3,000 psig at 600 F, and are equipped with quick-disconnect couplings to facilitate removal to the radiochemical laboratory for analysis. The RCS hot leg sample lines include a delay coil (sufficiently long tubing run) to permit the decay of N-16 before the sample leaves the containment. The
reactor coolant system, chemical and volume control system, and accumulator samples require sufficient purge to ensure representative samples. System 9.3-10 Rev. 8 WOLF CREEK pressure provides the motive force for the purging flows. Purge time is determined for each sample by the flow rate and the individual sample line
volume. Primary coolant purge flow is discharged to the volume control tank in the chemical and volume control system. Other purge flows are returned to the auxiliary building floor drain tank and elsewhere, as shown in Figure 9.3-2.
The sample sink drain, which may be contaminated with particulates or cleaning
solutions, is also routed to the auxiliary building floor drain tank. 9.3.2.2.2 Radwaste Sampling System The RWSS collects samples from the radwaste systems, as listed in Table 9.3-4, and brings them to the sample room in the radwaste building. The RWSS is
manually operated on an intermittent basis to provide samples for laboratory analysis. The RWSS is shown in Figure 9.3-3. The RWSS samples are routed to a manual sample facility within an exhaust-ventilated, hooded enclosure. Within the vented sampling hood are grab sample points for each stream. Sample lines
are purged before each sample is drawn to ensure that representative samples
are obtained. The design conditions of the RWSS are given in Table 9.3-4. 9.3.2.2.3 Process Sampling System The purpose of the PSS is to provide the data necessary to implement procedures for controlling the water quality of the secondary plant systems listed in
Table 9.3-5. The PSS, which is located in the turbine building, is shown in
Figure 9.3-4. The operating conditions of the PSS samples are given in Table 9.3-5. Roughing coolers are provided for the samples whose temperatures exceed 140 F. All samples to analyzers, except the corrosion product sample panels, are conditioned by a chilled water, constant-temperature bath. The corrosion product sample panels are located near the points of sample origin. Samples are analyzed, and the results are used for automatic or manual control of the process fluids. All analyzers are continuously monitoring
representative samples. The sample line and sample sink drains in the PSS are collected in the secondary liquid waste system where they are processed for reuse.9.3.2.2.4 Manual Grab Sample Stations Manual grab sample stations are provided for the liquid and gaseous sample points which require sampling at a frequency of less than once a week or on a
nonscheduled basis. All gas sampling stations are of the inline type which returns purge 9.3-11 Rev. 16 WOLF CREEK gases to the process lines. Quick-disconnect type couplings are used for sampling bottle connections to provide a convenient and expeditious way of
sampling for the nuclear sampling system. Grab sample points for primary and radwaste liquid and gases are identified in Table 9.3-6. No sample point is provided on the chemical mixing tank of the chemical and volume control system since chemical additives are preanalyzed before they are added to the mixing tank. 9.3.2.3 Safety Evaluation Except for the associated containment penetrations, the plant sampling system is not a safety-related system. SAFETY EVALUATION ONE - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability. All plant sampling system lines penetrating the containment can be isolated at the containment boundary by solenoid valves that close either upon receipt of a containment isolation signal or by manual actuation. (See Section 6.2.4 for a
discussion of containment isolation.) 9.3.2.4 Tests and Inspections Proper operation of the PrSS, RWSS, and PSS is initially demonstrated during preoperational testing. The proper operation and availability of the PrSS and RWSS are proved in service by their use during normal plant operation. Samples from the PrSS and RWSS are drawn manually for laboratory analysis. The results of this analysis
are checked by calibrating the laboratory instruments against known
compositions or check sources. The PSS draws continuous samples from the turbine cycle systems for automatic or manual water quality analysis. The operation of the PSS is verified by observing that continuous sample flow is maintained through the analyzers. The calibration of the analyzers is checked periodically by comparing it with laboratory analysis of a grab sample from the same process. The output of the continuous analyzers is recorded, and abnormal values are alarmed. 9.3.2.5 Instrumentation Applications The plant sampling systems use local pressure, temperature, and flow indicators to facilitate manual operation and to verify sample conditions before samples are drawn. 9.3-12 Rev. 14 WOLF CREEK A radiation element continuously monitors the steam generator blowdown sample for primary-to-secondary tube leaks. In the event the steam generator blowdown
samples exhibit high radioactivity, approaching the limits given in Section 11.5, the sample line isolation valves are automatically closed. Facilities for obtaining these samples are also provided at the nuclear sampling panel. The PSS is equipped with continuous analyzers to monitor specific water quality conditions. Certain measurements, as indicated in Figure 9.3-4, are used to automatically control the chemical addition for pH and corrosion control. Indicators and manual controls are provided on the sampling panel to maintain
the proper sample conditions of the water entering the analyzers. Grab sample points are also provided for laboratory analysis verification of analyzer
calibration. Chilled water is provided to condition samples to the standard condition required by instrumentation.
9.3.3 FLOOR
AND EQUIPMENT DRAINAGE SYSTEMS The floor and equipment drainage system (FEDS) collects, monitors, and directs liquid waste generated within the plant to the proper area for processing or disposal.9.3.3.1 Design Bases 9.3.3.1.1 Safety Design Bases
The following safety design bases are applicable to those portions of the FEDS which have safety-related functions of containment isolation, leak detection in
safety-related pump rooms following a LOCA, isolation of auxiliary building drainage system discharge paths following a LOCA, leak detection in the diesel generator rooms, leak detection in the basement of the control building, and backflow prevention rooms housing redundant trains of safety-related equipment. SAFETY DESIGN BASIS ONE - The FEDS is protected from the effects of all appropriate natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The FEDS is designed to remain functional after a SSE or to perform its intended function following postulated hazards of fire, internal missile, or pipe break (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-34). 9.3-13 Rev. 10 WOLF CREEK SAFETY DESIGN BASIS FOUR - The FEDS is designed so that 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. SAFETY DESIGN BASIS FIVE - The FEDS uses design and fabrication 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 is provided so that the FEDS' safety function will not be compromised. This
includes isolation of components to deal with leakage or malfunctions. Drainage from safety-related equipment rooms is designed to prevent flooding via drainage piping backflow. SAFETY DESIGN BASIS SEVEN - Instrumentation is provided which permits the detection of leakage from safety-related systems following a LOCA. SAFETY DESIGN BASIS EIGHT - 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 Criterion 56, and 10 CFR 50, Appendix J, Type C
testing.SAFETY DESIGN BASIS NINE - Instrumentation is provided which permits the detection of water accumulation that could affect the operation of safety-
related equipment. 9.3.3.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - All FEDS subsystems are designed to prevent the uncontrolled discharge of radioactive effluent from the power block. POWER GENERATION DESIGN BASIS TWO - All nonradioactive sub-systems are designed to minimize the introduction of potentially radioactive contaminated materials. POWER GENERATION DESIGN BASIS THREE - The design and arrangement of the sanitary drainage subsystem ensures that the introduction of potentially radioactive contaminated materials will not occur. 9.3-14 Rev. 0 WOLF CREEK POWER GENERATION DESIGN BASIS FOUR - The FEDS is designed to adequately handle and process normal anticipated drainage without sump overflow. Radioactive and
nonradioactive wastes are handled by separate subsystems. POWER GENERATION DESIGN BASIS FIVE - The FEDS contains provisions for normal plant operation leakage detection. The FEDS contains provisions for the detection of leakage from the reactor coolant system pressure boundary, containment cooler coil section, the spent fuel pool, transfer canal, cask loading pool, and refueling pool. The FEDS serves to identify leakage that may occur in the event of a pipe rupture within the plant. The collection piping within the FEDS is normally empty and is not a source of leakage. The discharge lines from the sump pump are normally full of water.
Section 3.6 provides an evaluation which demonstrates that the pipe routing of
the FEDS is physically separated from the essential systems to the maximum extent practical. Protection mechanisms, as required, are also discussed in Section 3.6. 9.3.3.2 System Description 9.3.3.2.1 General Description The FEDS is shown in Figure 9.3-5. Major drainage areas are shown in Figure 9.3-6. The FEDS consists of several subsystems, as described below. Areas of
the plant are served by the appropriate FEDS, based on the potential source of leakage into the subject area. This allows segregation of radioactive and nonradioactive sources. In addition, provisions are made in the appropriate subsystems for leak detection and isolation of portions of the subsystem to
preclude degradation of safety-related functions. Equipment and floor drainage for site structures (fuel oil pump house, administration building, shop building and circulating water screenhouse) are
provided separate from the power block structures. 9.3.3.2.1.1 Radioactive Drainage Areas Radioactive FED subsystems include:
- a. Potentially radioactive low tritium level waste (DRW)
- b. Tritiated waste (CRW) 9.3-15 Rev. 14 WOLF CREEK
- c. Chemical waste (ARW)
- d. Detergent waste (SRW)
- e. Potentially radioactive secondary liquid waste (LRW)
These subsystems are directed to and processed or disposed of by appropriate systems, as indicated in Figure 11.1A-1. The discharge of the liquid radioactive effluents from the power block is discussed in Section 11.2.2. A check valve is provided in the site portion of
the discharge line in order to preclude backflow of water to the powerblock. DRW SUBSYSTEM - The DRW subsystem consists of a network of floor and equipment drains arranged to collect potentially radioactive wastes with relatively low
tritium levels from mechanical components, valve stem leakoffs, and maintenance drainage in the auxiliary, fuel, containment, radwaste, and control buildings.
Each building is provided with a separate sump or group of sumps from which the collected waste is pumped to the floor drain tank for processing. The system also collects potentially radioactive tritiated waste from continuous equipment
drains within the engineered safety features pump rooms and liquid collected by
the leak detection subsubsystem of the DRW subsystem. The leak detection subsubsystem consists of a network of leak chases, collection piping, and flow measuring standpipes for the refueling pool and
fuel storage pool; standpipes for the containment coolers; level indicators for the containment normal sumps, RHR pump rooms sump, control building sumps, and the auxiliary building sump pit; and high level alarms for all sumps. The fuel storage pool standpipe measures combined leakage from the spent fuel pool, fuel transfer canal, and cask loading pool. In addition, all sump pump start and stop times are monitored and recorded by the plant computer. The refueling pool and fuel storage pool leak detection systems utilize gravity flow leak collection chases positioned behind the liner plate welds. Vertical liner plate welds have structural steel channels seal welded behind the weld
lines forming the collection chases. Horizontal liner plate welds have structural steel channels positioned under the weld lines. These horizontal channels are not seal welded at the top of the channel to the bottom of the pool liner. The leak chases are segregated into isolatable zones to facilitate leak location with only the vertical chases having leak tight capability. Refer to Figures 9.3-5 sheets 16 and 17 and 9.3-7 for the leak chase zone and
standpipe configuration. Each standpipe is capable of detecting a 1-gallon per
minute leak within 60 minutes after leak initiation. 9.3-16 Rev. 14 WOLF CREEK A condensate measuring standpipe is provided for each of the four containment coolers. The standpipes are similar in design to the pool standpipes, except
that they measure containment cooler condensate in lieu of pool leakage. The standpipes are designed to preclude condensate backup into the containment coolers in the event of high condensate flow rates. The standpipes are capable
of measuring a 1.0 gpm flow rate within 60 minutes after leak initiation. The
condensate flow rate during normal plant operation is used as the base rate
when evaluating condensate flow rates to determine if abnormal flows are occurring. The containment cooler standpipes measure unidentifiable leakage rates, as defined in Section 5.2.5. Sump level instruments in the containment, RHR pump rooms, auxiliary building, and instrument tunnel are used in conjunction with the plant computer to determine leak rates in the various buildings. The safety-related level instrumentation in the sumps in the auxiliary building, RHR pump rooms, control building, and containment is located in a
protected corner of the respective sump. The RHR pump room instrumentation provides the earliest possible indication of a potential flooding condition in the safety-related pump rooms and, therefore, serves to protect the safety-related pumps. The auxiliary building sump
instrumentation provides the earliest possible indication of a potential flooding condition in the auxiliary building at El. 1,974 corridors and areas open to the corridors and, therefore, serves to protect all equipment in the
auxiliary building. The control building instrumentation provides early
indication of a potential flood condition in the basement of the control
building and, therefore, serves to protect the safety-related essential service water system components in that area. The containment instrumentation provides early indication of a potential flood condition in the containment and
therefore serves to protect all safety-related equipment in the containment. Each of the safety-related level indication units is designed to provide the control room with an analog indication of the water level within the instrument measurement range. Refer to Appendix 3B for the design basis flood level. The containment sumps and incore instrumentation sump indicators serve to measure unidentifiable leakage, as defined in Section 5.2.5. 9.3-17 Rev. 6 WOLF CREEK In addition to providing the plant operators with a safety-related, Class IE indication of water levels in the RHR pump room and the auxiliary building
sump, the level indicators provide input to the plant computer. High level alarms with control room annunciation are provided for all sumps. The alarm points are set above the highest normal sump pump actuation level. All sumps within the turbine building are provided with a common annunciator as are all the CRW and DRW sumps serving nonsafety-related systems. All other sumps are provided with unique and separate annunciators. Safety-related components which are located in the lowest elevation of the auxiliary building are housed within watertight compartments. The drainage
arrangement for that area is such that external drain or flood water is prevented from back flow into these areas, and flooding within rooms of one train of the safety-related components cannot communicate with the areas associated with the redundant train. Figure 9.3-6 shows the drainage
arrangement associated with this area. Redundant check valves on the DRW
subsystem discharge line from the control building are also provided to assure that there is no backflow. The DRW subsystem is designed with a segregated collection system for each of the safety-related pump trains so that crossflooding between trains will not
occur. One sump for each safety-related train is provided and located in the RHR pump room. Sump pump discharge lines for these sumps are routed above the minimum watertight level and are provided with check valves internal and
external to each room to preclude the sump pump discharge of one room from
backflowing into the redundant room. The CRW subsystem equipment drains for
the safety-related rooms are routed to a common sump external to the equipment rooms watertight boundary. The CRW equipment drains are capped during normal plant operation to prevent equipment room flooding from an external source. The
CRW equipment drain caps may be removed during controlled maintenance
operations to facilitate equipment drainage. In the event of a LOCA, it is necessary to assure that any leakage from the ECCS be retained within the auxiliary building since airborne releases can be
controlled and filtered, as discussed in Section 6.5. Redundant safety-related
sump pump discharge isolation valves are provided which isolate on any SIS
signal and prevent the discharge of the auxiliary building and RHR pumproom DRW sump pumps from leaving the auxiliary building. Seal failure and the resultant maximum seal leakage of 7.5 gpm from the ECCS and containment spray pumps is the only major 9.3-18 Rev. 0 WOLF CREEK credible source of leakage outside the containment following a LOCA. This leak rate is based on gross seal failure, as discussed in Section 6.3. Containment isolation provisions on the DRW subsystem line which penetrates the containment include a normally open motor-operated valve inside and a normally closed air-operated valve outside. Both valves automatically close upon receipt
of a CIS-A signal. An additional nonsafety-related solenoid is provided for the
air-operated valve which, when energized, will open the valve. This occurs upon receipt of an indication of a running containment sump pump, except when a CIS-A signal is present. A high water level in a sump activates the associated
pump.An oil/water separator is provided for the DRW subsystem drains of the hot machine shop and decontamination room. The separator segregates machine shop oil from collected effluent prior to discharge to the floor drain tank. Sumps collecting liquids for processing through oil/water separators (hot machine shop and all LRW subsystem sump pumps) are furnished with low shear double diaphragm pumps to preclude oil emulsification prior to oil/water separation. CRW SUBSYSTEM - The CRW subsystem collects liquid waste which may contain relatively high tritium levels from equipment and valves within the auxiliary, radwaste, and fuel buildings. Separate sumps are provided in each building for effluent collection. Equipment drains only are provided for this system. The
subsystem sump pumps discharge all collected effluent to the waste hold-up tank
for processing. ARW SUBSYSTEM - The ARW subsystem collects waste from selected laboratory sample sinks, maintenance drains from evaporator reagent tanks, and washdown
wastes from the hot machine shop decontamination room decontamination tank.
Waste collected by the hot laboratory sample sinks flows by gravity to a
collection sump and is pumped to the chemical drain tank for processing and solidification. Waste from the reagent tanks, a radwaste building sample station, and the decontamination tank flow directly by gravity to the chemical drain tank. SRW SUBSYSTEM - The subsystem collects waste from laboratory dishwashers, washing machines, deep sinks, and the laundry area. The collected waste flows by gravity to a stainless steel collection tank. Two pumps take suction from
the tank and operate 9.3-19 Rev. 14 WOLF CREEK alternatively or in parallel to pump the effluent to the laundry water storage tank for processing and recycle. A basket strainer is provided in each pump
suction line to filter out lint and debris. LRW SUBSYSTEM - The LRW subsystem collects normally nonradioactive but potentially radioactive turbine building drains and portions of the auxiliary building drains which do not normally house radioactive components. Two sumps are provided in the turbine building and one in the auxiliary building. The system normally discharges collected effluent to the secondary liquid waste processing system for recycle within the plant or discharge. The drainage can also be directed to the oil waste (OW) subsystem. Drain routing is an operator
judgment based on prior knowledge of secondary system chemistry and radioactive
contamination, in conjunction with ODCM limitations. The subsystem discharge can also be aligned to pump to the OW system when nonsecondary side liquid is
collected, such as may be the case following a fire protection system sprinkler
actuation. The LRW system also includes a condensate collection tank and pump designed to hold and transfer recyclable condensate back to the condenser. The OW and LRW subsystems both serve the turbine building. Six-inch curbs are provided between the subsystem drainage areas to assure that proper segregation of equipment leakages is maintained. The sump pump discharge lines for the two
subsystems have independent discharge line isolation valves and a valved crossconnection. These pneumatically operated valves can be remotely operated so that the LRW subsystem can be aligned to discharge to the OW subsystem header or the OW subsystem aligned to discharge to the LRW subsystem header. Two 20-inch LRW drain lines are provided for the main steam/main feedwater isolation valve room in the auxiliary building to preclude flooding in the event of a postulated pipe break. The drain lines discharge into El. 2,000 of the turbine building. 9.3.3.2.1.2 Nonradioactive Drainage Areas Nonradioactive FED subsystems include:
- a. Sanitary waste (SAN)
- b. Roof drains (RD)
- c. Potentially oily waste (OW) 9.3-20 Rev. 13 WOLF CREEK SAN SUBSYSTEM - The SAN subsystem collects sanitary waste from service facilities, pantry facilities, electric water coolers, clean showers, plumbing
fixtures, and toilet floor drains within nonradioactive areas. The system is completely trapped and vented. The sanitary waste from all the buildings in the plant is collected in gastight and vented concrete sumps and pumped by a
duplex arrangement of sump pumps to the sanitary sewer system to a lift station
which sends the sanitary waste to a Non Discharging Sewage Lagoon located north
east of the plant.
Equipment drains in the shop building are routed to the site water treatment
system, which is located in the shop building (separate from the power block) and routed to a lime sludge pond (located north of the power block).
RD SUBSYSTEM - The roof drain subsystem collects water resulting from
precipitation on building roofs. The roof drain subsystem is sized at a design
rainfall rate, as shown in Chapter 2.0. The collected rainwater is conveyed by gravity to the site storm drainage system.
Building roof drainage for the power block area is routed to the storm drain
system, which eventually drains into the cooling lake southwest of the plant
site. Any potentially oily waste is directed to the oil separator located south of the plant building. The effluent from the oil separator joins the storm drain system and empties into the cooling lake.
OW SUBSYSTEM - The OW subsystem collects nonradioactive liquid waste from the
turbine building, diesel generator building, communications corridor, control building, and selected areas of the auxiliary building. Equipment and floor drainage for site structures (fuel oil pump houses, administration building, shop building and circulating water screenhouse) are provided separate from the
powerblock. Nonradioactive wastes are collected in sumps and pumped to an
oil/water separator outside of the powerblock for processing and disposal. Radiation monitoring and automatic system isolation is provided in the discharge line to preclude the discharge of radioactive fluid from the
powerblock. Potentially oily wastes are routed to the cooling lake after
passing through the oil separator.
A portion of the OW system includes the Essential Service Water Vertical Loop Chase (ESWVLC) where a pump, discharge piping, and isolation valves are installed to remove water from the Chase. This portion of the OW system provides no safety-related function and does not adversely affect any design functions of safety-related SSCs. However, there is a potential that water collected in the Chase may contain traces of contamination from other sources outside the Chase. Additionally, the sump effluent may contain water from the ESW system. Therefore, the sump contents are manually sampled and tested for contamination constituents prior to free release to the storm sewer. The pump is manually operated and is located in the sump, which is designed to be a low point in the foundation for placement of the sump pump. The sump effluent is manually sampled and tested for contaminated constituents prior to free release, after which the effluent is pumped to the site oil/water separator via the storm sewer. If the effluent is unacceptable for free release, the effluent will be pumped through an alternate flow path, via hose connection and alignment of the isolation valves in the discharge piping, for collection, treatment, and disposal or release.
9.3-21 Rev. 29 WOLF CREEK The diesel generator building sumps are provided with safety-related level instruments located in a protected corner of the respective sump. They provide
the control room with analog level indication for earliest possible indication of a potential flooding condition in the diesel rooms and therefore serve to protect the diesels and associated switchgear. The OW drainage system serving the control room is provided with a loop seal to facilitate control room pressurization. Means for checking and maintaining the loop seal level are provided. Trapped and vented drains are provided for powerblock battery rooms to assure that potentially noxious and corrosive
vapors are retained within the battery rooms in the event of a gross battery failure. Acid neutralization tanks are also provided for the battery room
drain headers to assure that the potentially corrosive effluents are neutralized with respect to pH prior to discharge from the powerblock. The Containment Mini Purge Air Supply Unit provided with a loop seal that prevents the flow of air through the oily waste drain line le256xnd-4 during the negative pressure condition when auxiliary building supply fan (SGL01) is in operation. 9.3.3.2.2 Component Description Codes and standards applicable to the FEDS are listed in Table 3.2-1. Except as discussed below, the design and construction of the FEDS is non-seismic Category I and quality group D. The containment penetration associated with
the FEDS is designed and constructed to quality group B and seismic Category I requirements. Sump pump discharge isolation valves and level instrumentation for ECCS and containment spray pump areas which are required following LOCA are designed and constructed to quality group C and seismic Category I. COLLECTION PIPING - In areas of potential radioactivity, the collection piping is stainless steel. Stainless steel is also provided for nonradioactive
battery room drains in the control building and drains in the turbine building for the collection of secondary side leakages and drainage. In non-radioactive areas where the collected effluent is discharged from the powerblock (OW, SAN, and RD subsystems), all embedded piping is cast iron. Suspended piping is galvanized steel or cast iron. The fabrication and installation of piping
provides for a minimum uniform slope of 1/8 inch per foot to induce waste to flow in the piping. The piping is embedded where necessary for radiation shielding. Equipment drainage piping is terminated not less than 3 inches
above the finished floor. 9.3-22 Rev. 7 WOLF CREEK EQUIPMENT DRAINS - Piped-up equipment drains are either routed directly to an embedded stub-up and seal welded in place or routed to an embedded stub-up and
terminated in the open end. The connection to the stub-up varies as required for a particular application. In general, CRW subsystem drains carrying liquids with the potential for relatively high tritium levels are terminated
with seal-welded connections while all others are terminated in the open end of
drain hubs. Several equipment drains may terminate in one stub-up. FLOOR DRAINS - All floor drains are installed with rims flush with the low point elevation of the finished floor. Floor drains in areas of potential radioactivity are welded directly to the collection piping and are provided with threaded plugs of the same material. The plugs are used to seal the floor drains during hydrostatic testing of the drainage systems and during all required leak rate test procedures. They are also installed, as required, to preserve the integrity of the drainage systems. Floor drains in areas not restricted due to potential radioactivity are
provided with caulked or threaded connections. TRAPS - Inlets to the sanitary drainage system are provided with a water seal in the form of a vented P-trap to minimize entry of vermin and foul odors into
the building. Air pressure vent lines to the outside atmosphere are provided
downstream of the P-traps to prevent excessive backpressures which could cause blowout or siphonage of the water seal. A trapped header is provided to facilitate control room pressurization during control room isolation, as
indicated in Section 6.4. Means for testing and filling the control room trap is provided. Trapped and vented drain lines are also provided for battery rooms in the control and turbine buildings. Traps are not installed at inlets in oily, detergent, and chemical drainage subsystems or in areas of potential radioactivity to preclude the accumulation of radioactive liquids, oil, or detergents. CLEANOUTS - The DRW, OW, LRW, and SRW subsystems are provided with cleanouts when practical at the base of each vertical riser and at intervals of not more than 50 feet. Floor and equipment drains without traps are considered to be
cleanouts for design purposes. The CRW, ARW subsystems, and leak detection
subsystems are not provided with cleanouts because the effluents collected have a very low percentage of suspended solids. The sanitary drainage subsystem and roof drain are provided with cleanouts. 9.3-23 Rev. 1 WOLF CREEK COLLECTION SUMPS - Sumps collecting potentially radioactive liquid (except those inside the containment) are lined with stainless steel. The sumps are
provided with a 1/2-inch-thick carbon steel cover used to support the sump controls and sump pumps. The sumps (except LRW sumps) are vented to a filtered building exhaust system. The sumps inside the containment are lined with
carbon steel. Radwaste building sumps and the ARW sumps are designed to accept
a 12-inch-thick concrete cover in addition to the steel cover for additional
shielding. Sumps collecting nonradioactive liquid consist of concrete pits covered with 1/2-inch carbon steel cover plates. Sumps collecting nonradioactive fluids are
locally vented, except for the sanitary drainage sump which is vented outside
the powerblock through an independent vent. All sump capacities are equal to or greater than the amount pumped from it in 5 minutes with one pump running. All sumps have removable covers and/or inspection openings to facilitate sump cleaning and pump and controls inspection. COLLECTION TANKS - Horizontal stainless steel tanks are used to collect SRW subsystem wastes and DRW reactor coolant pump lubricating oil leakage. The SRW
subsystem tank is located in the control building and vented to a filtered exhaust system whereas the DRW subsystem tank is located in the containment and vented locally. A horizontal carbon steel tank in the LRW subsystem is used in
the turbine building to collect main steam condensate. This tank is vented to a
turbine building exhaust system. All tanks are provided with overflow
connections. ACID NEUTRALIZATION TANKS - Each battery room floor drain network is provided with an acid neutralization tank designed to neutralize the amount of acid
contained within approximately 25 percent of the battery cells in the event of
a break in the batteries. The tanks are stainless steel and filled with limestone as the neutralization agent. Liquid flows by gravity through the tanks and into the OW subsystem. The tanks are vented outside for removal of
C0 2 generated during the neutralization process. PUMPS - Vertical centrifugal sump pumps or double diaphragm sump pumps are provided for all sumps. Sumps lined with stainless steel have duplex stainless steel pumps while pumps in the concrete sumps are cast iron. Double diaphragm
pumps are used to pump water which is to be processed through an oil-water
separator and subsequently recycled. Duplex arrangements of sump pumps are
provided in every case, except for pumps in the tendon access 9.3-24 Rev. 0 WOLF CREEK gallery, which are simplex. Submersible pumps are used in the incore instrumentation tunnel and the auxiliary building - radwaste building pipe
tunnel. All sump pumps, except the sanitary lift station pumps, are provided with suction strainers designed to preclude the pumping of particles greater than 1/2-inch diameter. The sanitary lift station sump pumps are capable of
pumping a spherical mass less than or equal to 2 1/2-inches in diameter. Pump
discharge rates are equal to or greater than the maximum anticipated drainage
rates to the sumps during normal plant and/or maintenance operations. 9.3.3.2.3 System Operation All of the FEDS subsystems utilize gravity drainage for the collection of the various effluents. All subsystems except the OW, roof drain, and leak detection subsystems utilize duplex arrangements of pumps at the collection point. The OW subsystem utilizes single pumps for the tendon access gallery and miscellaneous condensate drain tank and duplex pumps for all other applications. The roof drain and leak detection subsystems do not require
pumps. The pumps (lead pump in a duplex configuration) are automatically activated when a predetermined high water level in the sump or tank is reached. Two pumps are actuated (duplex assemblies) when the water level rises to a
predetermined high-high level. One pump will stop automatically when the
liquid level falls below the high level set point. The lag pump will continue
to operate until the water level is pumped down to a predetermined low level. The alternator automatically changes the actuation sequence for the lead and lag pump. High level alarms with computer annunciation are provided for all
sumps and tanks. The pneumatically operated containment isolation valve is normally closed and opens only in the event of a containment sump pump start in the absence of a containment isolation signal. The motor-operated containment isolation valve is normally open and closes in the event of a containment isolation signal. The auxiliary building isolation valves are motor operated and normally open and close only in the event of a safety injection signal. The leak detection system in the containment determines leak rates by calculating fill rates in sumps and standpipes. Standpipes utilize base-mounted
pressure transmitters, to monitor standpipe water levels. The plant computer utilizes the pressure information and calculates incoming flow rates based on level changes resulting from the filling standpipes. The standpipes for the
containment cooler, the refueling pool, and the fuel storage pool automatically drain following a stand-pipe high level and 9.3-25 Rev. 14 WOLF CREEK reset for continued operation. The level transmitters for the containment sumps provide an analog level signal to the plant computer. The plant computer
is programmed to periodically calculate incoming flow rate and produce an alarm message if the flow rate increases by a predetermined amount. 9.3.3.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases of Section 9.3.3.1.1. SAFETY EVALUATION ONE - The safety-related portions of the FEDS 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 safety-related portions of the FEDS are designed to remain functional after a safe shutdown earthquake. 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained. SAFETY EVALUATION THREE - The safety-related portions of the FEDS are completely redundant and, as indicated by Table 9.3-7, 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 safety-related portions of the FEDS were initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.3.3.4. Section 6.6 provides the
ASME Boiler and Pressure Vessel Code, Section XI requirements that are
appropriate. 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. Figure 9.3-5 shows that the components meet the design and fabrication codes given in Section 3.2. All the power supplies and controls necessary for safety-related functions of the FEDS are Class IE, as described in Chapters 7.0 and 8.0. 9.3-26 Rev. 19 WOLF CREEK SAFETY EVALUATION SIX - Section 9.3.3.2 describes the isolation provisions incorporated in the drainage system to ensure that any leakage that could occur
following a LOCA is retained within an area which is served by a safety-related filtration exhaust system. Section 9.3.3.2 also describes the segregated drainage system for each watertight safety-related component area and the barriers which prevent backflow.SAFETY EVALUATION SEVEN - Safety-related level indicators are provided in each of the watertight areas which house the ECCS and containment spray pumps. Seal
leakage from these pumps is the only major credible source of leakage following
a LOCA. Redundant level indication is provided in the auxiliary building sump located in the sump pit of the basement of the auxiliary building to detect any long-term accumulation of fluid leaking from safety-related systems operating after a LOCA. Level instrumentation is discussed in Section 9.3.3.5. The drain configuration is indicated in Figure 9.3-6. SAFETY EVALUATION EIGHT - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability. SAFETY EVALUATION NINE - Safety-related level indicators are provided in the basement of the control building and in each diesel generator room to provide indication of a potential flooding condition in those areas. Refer to Section 9.5.1 for a description of flood damage protection during fire fighting
operations. 9.3.3.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. The performance and structural and leaktight integrity of system components is demonstrated by continuous operation. The FEDS is testable through the full operational sequence that provides isolation following a LOCA, including operation of applicable portions of the protection system and transfer between normal and standby power. The safety-related components are located to permit preservice and inservice inspections. 9.3-27 Rev. 0 WOLF CREEK 9.3.3.5 Instrumentation Application The FEDS instrumentation is designed to facilitate automatic operation and remote control of the system and to provide continuous indication of system parameters. Safety-related float-type level indicators are provided in the RHR pump room sumps, the auxiliary building sump, the control building sump, the diesel generators building sumps, and the containment normal sumps. In the event of a LOCA, these level devices monitor the performance of the safety-related systems by detecting leakage significant enough to result in detectable accumulation. The RHR pump room sump and auxiliary building sump level devices can detect the accumulated leakage resulting from a leak as small as 5 gpm within 30 minutes of the leak initiation. The level devices in the sumps will indicate the water level up to 6 feet from the bottom of the sump for the control building sump (high level alarm at 2 feet 2 inches); up to 6 feet 6 inches above the top of the sump for the RHR pump rooms (high level alarm at 2 feet 10 inches from the bottom of the sump); between 3 feet 6 inches from the bottom of the sump to 6 feet from the bottom of the diesel generator building sumps (high level alarm at 1 foot 9 inches); and up to 7 feet 6 inches from the bottom of the sump for the containment normal sumps (high level alarm at 3 feet 3 inches from the bottom of the sump). Each FEDS normal sump operating level line which can discharge outside the standard power block is provided with a radiation monitor which will isolate the discharge path upon a high level indication. Section 11.5 discusses the process radiation monitors. High level alarms with control room annunciation are provided for all sumps and tanks. All sumps within the turbine building are provided with a common
annunciator as are all floor and equipment subsystem sumps serving nonsafety-related equipment. The sanitary drainage subsystem sump, chemical drain subsystem sump, detergent drain subsystem tank, sump serving the auxiliary feedwater pumps, sumps within the diesel generator building, sump serving the auxiliary boiler room, sumps within the containment, and sumps inside of the safety features equipment rooms are provided with unique and separate
annunciators. Each high level alarm is set to annunciate at a level above that required to actuate both pumps of a duplex sump pump arrangement or one pump of a simplex sump arrangement. 9.3-28 Rev. 4 WOLF CREEK Hand switches with indicator lights are provided in the control room for sumps inside the containment and the sump pumps within the safety features pump rooms
to permit remote sump pump actuation. All other pumps are provided with local hand switches. Instrumentation is provided for the oily waste discharge line to measure and indicate the radiation level of the pumped effluent. Digital readout is provided in the control room. Instrumentation is provided for the fuel storage pool and refueling pool to measure pool leak rates and each containment cooler to measure condensate flow rates. Standpipes with automatic drain controls are used and can detect a one gallon per minute leak within 60 minutes of leak initiation. A periodic update of the leak rate is provided by the plant computer. Also, instrumentation is provided for the instrument tunnel sump and the containment normal sumps, which
provide data to the plant computer for leak rate calculations. The detergent waste system basket strainers are provided with instrumentation to determine strainer pressure drop. A high pressure drop condition is alarmed in the control room. Controls and instrumentation are provided for pneumatic and motor-actuated valves to permit remote operation and provide indication of valve position. This includes containment isolation valves, fuel storage pool standpipe valves (leak detection subsystem), refueling pool standpipe valves (leak detection subsystem), oily waste discharge isolation valves, containment cooler standpipe
valves (leak detection subsystem), auxiliary building sump discharge isolation
valves, and the secondary liquid waste to oily waste system isolation valves.
9.3.4 CHEMICAL
AND VOLUME CONTROL SYSTEM The chemical and volume control system (CVCS) performs the following functions: a. The CVCS maintains the required water inventory in the reactor coolant system (RCS) during normal operation, power changes, startup, and shutdown, including pressurizer
auxiliary spray for depressurization. The CVCS also provides
reactor grade water to the reactor coolant pump seals for cooling and sealing purposes. b. The CVCS varies the RCS soluble neutron absorber (boron) concentration to compensate for core burn-up. The CVCS provides sufficient boron, in the form of boric acid, to maintain the required shutdown margin during refueling. 9.3-29 Rev. 14 WOLF CREEK
- c. The CVCS and boron thermal regeneration subsystem (BTRS) vary the RCS boron concentration to compensate for xenon
transients and other reactivity changes which occur when the reactor power changes during load following, shutdowns, and startups.
- d. The CVCS functions to maintain the desired RCS water chemistry conditions and reduce the radioactivity level.
- e. Portions of the CVCS (i.e., charging pump subsystem) provide an injection flow to the RCS upon receiving a safety
injection signal.
- f. The CVCS provides normal makeup to the RWST and fuel storage pool.
- g. For safety grade cold shutdown, part of the CVCS functions in conjunction with other systems of the cold shutdown design.
The boron recycle system is discussed in Section 9.3.6. 9.3.4.1 Design Bases 9.3.4.1.1 Safety Design Basis
Portions of the CVCS associated with emergency boration (via BAT or RWST), charging for ECCS, reactor coolant pressure boundary isolation and containment
isolation are safety related. These portions are required to function following a DBA and to achieve and maintain the plant in a safe shutdown condition. SAFETY DESIGN BASIS ONE - The CVCS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The pressure boundary of the CVCS is designed to remain intact after an SSE, some of the system components are designed to
remain functional after an SSE, and the system is designed to perform its intended function following postulated hazards, internal missiles, or pipe breaks (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-26 and 35). 9.3-30 Rev. 14 WOLF CREEK 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-32, 36, and 37). SAFETY DESIGN BASIS FIVE - The CVCS 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 to isolate components or piping is provided so that the CVCS's safety function will not be 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 SEVEN - The containment isolation valves in the system are selected, tested, and located in accordance with the requirements of GDC-55 and
10 CFR 50, Appendix J, Type C Testing. SAFETY DESIGN BASIS EIGHT - The CVCS provides diverse means of borating the RCS to a concentration that exceeds the requirement for a safe shutdown of the
reactor from any operating condition, assuming that the control rod cluster
with the highest reactivity worth is stuck in its fully withdrawn position and in the unlikely event that safe shutdown is initiated from peak xenon conditions. This amount of boric acid also exceeds the amount required to bring
the reactor to a hot shutdown condition and to compensate for the subsequent
reactivity transient resulting from xenon decay (GDC-27 and 29). SAFETY DESIGN BASIS NINE - The CVCS has sufficient makeup capacity to maintain the required RCS water inventory in the event of a reactor coolant system leak resulting from an equivalent pipe break opening of 3/8-inch (liquid service), 3/4-inch (steam service), diameter or less (GDC-33). SAFETY DESIGN BASIS TEN - The centrifugal charging pump subsystem of the CVCS in conjunction with other systems, provides a borated injection flow to the RCS
upon receipt of a safety injection signal. The charging pump subsystem of the
CVCS is an integral part of the ECCS. SAFETY DESIGN BASIS ELEVEN - Should only safety-related equipment be available, the centrifugal charging pump subsystem of the CVCS functions in conjunction
with other systems of the cold shutdown design to borate the RCS to a cold
shutdown concentration. 9.3-31 Rev. 0 WOLF CREEK 9.3.4.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The CVCS regulates the concentration of chemical neutron absorber (boron) in the reactor coolant to control reactivity changes resulting from the change in reactor coolant temperature between cold
shutdown and hot full-power operation, burnup of fuel and burnable poisons, buildup of fission products in the fuel, and xenon transients. The CVCS is
capable of borating the RCS through either one of two flow paths and from either one of two boric acid sources. POWER GENERATION DESIGN BASIS TWO - The CVCS is capable of controlling the changes in the reactor coolant boron concentration to compensate for the xenon
transients during load-follow operations, without adding makeup for either boration or dilution. This is accomplished by the boron thermal regeneration process, which is designed to allow load-follow operations as required by the design load cycle. POWER GENERATION DESIGN BASIS THREE - The CVCS maintains the coolant inventory in the RCS within the allowable pressurizer level range for all normal modes of operation, including startup from cold shutdown, full power operation, and
plant cooldown. This system also has sufficient makeup capacity to maintain the
minimum required inventory in the event of minor RCS leaks. POWER GENERATION DESIGN BASIS FOUR - The CVCS is capable of removing fission and activation products, in ionic form, gaseous form, or as particulates, from
the reactor coolant in order to provide access to those process lines carrying
reactor coolant during operation and to reduce activity releases due to leaks. POWER GENERATION DESIGN BASIS FIVE - The CVCS provides a means for adding chemicals to the RCS to control the pH of the coolant during initial startup and subsequent operation, scavenge oxygen from the coolant during startup, and
counteract the production of oxygen in the reactor coolant due to radiolysis of
water in the core region. Oxygen control is also provided by maintaining dissolved hydrogen in the reactor coolant to scavenge oxygen. The CVCS is capable of maintaining the oxygen content and pH of the reactor coolant within the limits specified in Table 5.2-5. POWER GENERATION DESIGN BASIS SIX - The CVCS is able to continuously supply filtered water to each reactor coolant pump seal, as required by the reactor
coolant pump design and as specified in Table 9.3-8. 9.3-32 Rev. 0 WOLF CREEK POWER GENERATION DESIGN BASIS SEVEN - (deleted) POWER GENERATION DESIGN BASIS EIGHT - The letdown and excess letdown lines between the points where they connect to the reactor coolant system and the points where they penetrate the secondary shield wall contain sufficient volume
to delay the flow for 60 seconds during maximum letdown to allow the N-16 activity to decay. POWER GENERATION DESIGN BASIS NINE - The purification and BTRS portions of the CVCS use design and fabrication codes consistent with quality group D (augmented), as assigned by Regulatory Guide 1.143 for radioactive waste
management systems. The codes and standards to which individual components of
the CVCS are designed are listed in Section 3.2. 9.3.4.2 System Description 9.3.4.2.1 General Description The CVCS is shown in Figure 9.3-8, with system design parameters listed in Table 9.3-8. The CVCS consists of several subsystems: the charging, letdown, and seal water system; the reactor coolant purification and chemistry control
system; the reactor makeup control system; and the boron thermal regeneration
system. CVCS operation during accident mitigation is discussed in Section 6.3. During refuel 6, non-safety related temperature monitoring instrumentation has been temporarily installed on the CVCS Auxiliary Spray piping to the
pressurizer. The instrumentation is strapped around the piping in three
locations. Two of the locations fall on portion of this piping in the Reactor Coolant System (RCS). The instrumentation consists of two RTDs at each of thethree locations (Total 6 RTDs) and associated cabling to a data logger. The
RTDs provide input to an existing data logger located in the containment, which ultimately feeds the data via an existing datalink to a Personal Computer
located in the I&C Hot Shop. Section 3.6 provides an evaluation demonstrating that pipe routing of the CVCS is physically separated from essential systems to the maximum extent
practicable. Protection mechanisms that are required are also discussed in
Section 3.6. 9.3.4.2.1.1 Charging, Letdown, and Seal Water System The charging and letdown functions of the CVCS are employed to maintain a programmed water level in the RCS pressurizer, thus maintaining a proper reactor coolant inventory during all phases of plant operation. This is achieved by means of a continuous 9.3-33 Rev. 10 WOLF CREEK feed-and-bleed process during which the feed rate is automatically controlled, based on the pressurizer water level. The bleed rate can be chosen to suit various plant operational requirements by selecting the proper combination of letdown orifices in the letdown flow path. Reactor coolant is let down to the CVCS from a reactor coolant loop cross-over leg. It then flows through the shell side of the regenerative heat exchanger where its temperature is reduced by heat transfer to the charging flow passing through the tubes. The coolant then experiences a large pressure reduction as it passes through the letdown orifice(s) and flows through the tube side of the letdown heat exchanger where its temperature is further reduced. Downstream of
the letdown heat exchanger, a second pressure reduction occurs. This second pressure reduction is performed by the low pressure letdown valve, which maintains upstream pressure and thus prevents flashing downstream of the
letdown orifices. The coolant then flows through one of the mixed bed demineralizers. The flow may then pass through the cation bed demineralizer, which is used intermittently when additional purification of the reactor coolant is required. From a point upstream of the BTRS or from a point upstream of the reactor coolant filters, a small sample flow may be diverted from the letdown stream to the boron concentration measurement system (see Section 7.7). The read-out on the boron concentration is given in the main control room. The boron concentration measurement system has been abandoned-in-place. During reactor coolant boration and dilution operations, especially during load follow, the letdown flow leaving the demineralizers may be directed to the BTRS. The coolant then flows through the reactor coolant filter and into the volume control tank (VCT) through a spray nozzle in the top of the tank. Hydrogen is continuously supplied to the VCT where it mixes with fission gasses
which are stripped from the reactor coolant into the tank gas space. The contaminated hydrogen is vented to the gaseous waste processing system. The partial pressure of the hydrogen gas mixture in the VCT determines the
concentration of hydrogen dissolved in the reactor coolant for control of the oxygen produced by radiolysis of the water in the core. Three charging pumps (one normal pump and two standby pumps) are provided to take suction from the volume control tank and return the purified reactor coolant to the RCS. Normal charging flow is handled by the normal charging
pump. This charging flow splits into two paths. The bulk of the charging flow
is pumped back to the RCS cold leg through the tube side of the regenerative heat exchanger. The letdown flow in the shell side of the regenerative heat exchanger raises the charging flow to a temperature approaching the reactor coolant temperature. Two charging paths are provided from a point downstream of the regenerative heat exchanger. A flow path is also provided from the regenerative heat exchanger outlet to the 9.3-34 Rev. 15 WOLF CREEK pressurizer spray line. An air-operated valve in the spray line is employed to provide auxiliary spray to the vapor space of the pressurizer during plant
cooldown. This provides a means of cooling and mixing the pressurizer contents near the end of plant cooldown, when the reactor coolant pumps, which normally provide the driving head for the pressurizer spray, are not operating. Should
only safety grade equipment be available, depressurization could be performed
by the cold shutdown design, also described in Section 7.4. A portion of the charging flow is directed to the reactor coolant pumps (RCP) (nominally 8 gpm per pump) through a seal water injection filter. The flow is
directed to a point above the pump shaft bearing. Here the flow splits, and a
portion (nominally 5 gpm per pump) enters the RCS through the labyrinth seals
and thermal barrier. The remainder of the flow is directed upward along the pump shaft to the number 1 seal leakoff. The number 1 seal leakoff flow from the four RCPs discharges to a common manifold, exits from the containment, and then passes through the seal water return filter and the seal water heat exchanger to the suction side of the charging pumps, or by alternate path to
the volume control tank. Measurement of the leakage from the RCP seals when the RCS pressure is 2235 20 psig is performed every 31 days. When RCP number 1 seal leakoff is found to be greater than 8 gpm per RCP, action is taken in accordance with the off normal procedure for RCP malfunction. A very small
portion of the seal flow leaks through to the number 2 seal. A stand-pipe provides a head for the number 3 seal which provides a final barrier to leakage of reactor coolant to the containment atmosphere. The number 2 seal leakoff
flow is discharged to the reactor coolant drain tank in the liquid waste processing system. The number 3 seal overflow is discharged to the containment
sump (this leakoff flow consists of a portion of the reactor makeup water which is supplied by the RCP seal standpipe). As discussed in Section 5.4.1.2.2, the RCP shaft seal system is designed for continued operation with either seal
water injection or component cooling water to the RCP thermal barrier. The excess letdown path is provided as an alternate letdown path from the RCS in the event that the normal letdown path is inoperable or provides insufficient capacity. Reactor coolant can be discharged from a crossover leg
to flow through the tube side of the excess letdown heat exchanger where it is
cooled by component cooling water. Under emergency shutdown conditions, the
letdown flow can be diverted downstream of the excess letdown heat exchanger to the pressurizer relief tank. Under normal conditions, downstream of the heat exchanger, a remote-manual control valve is used to control the letdown flow.
The flow normally joins the RCP number 1 seal discharge manifold and passes through the seal water return filter and heat exchanger to the suction side of the charging pumps. The excess letdown flow can also be directed to the VCT or the reactor coolant drain tank. When the normal letdown 9.3-35 Rev. 14 WOLF CREEK line is not available, the purification path is also not in operation. Therefore, this alternate condition would allow continued power operation for a
limited period of time, dependent on RCS chemistry and activity. The excess letdown flow path is also used to provide additional letdown capability during the final stages of plant heatup. This path removes some of the excess reactor
coolant due to coolant expansion as a result of the RCS temperature increase.
Should RCS inventory letdown be required, a safety grade letdown path via the
excess letdown heat exchanger to the pressurizer relief tank (PRT) is provided. This assures the capability to provide an RCS inventory letdown path should normal letdown paths become unavailable. This path may be used in conjunction
with other features of the safety grade cold shutdown system which is discussed in Section 7.4. A normally open cross-tie line is provided between the normal letdown and charging systems. The purpose of the cross-tie is to provide makeup water from charging to letdown to collapse any steam bubble(s) formed in the letdown system due to letdown system cooldown, post-isolation. The cross-tie includes
a mechanical pressure gauge, installed outside the primary shield wall, to enable Operations to ensure that the pressure across letdown system isolation valves BG LCV459/460 is equalized before the valves are reopened. Surges in the RCS inventory due to load changes are accommodated for the most part in the pressurizer. The volume control tank provides additional surge capacity for reactor coolant expansion not accommodated by the pressurizer. If the water level in the volume control tank exceeds the normal operating range of 30 - 60%, a proportional controller modulates a three-way valve downstream of the reactor coolant filter to divert a portion of the letdown to the boron recycle system. If the high level limit in the volume control tank is reached, an alarm is actuated in the control room and the letdown flow is completely diverted to the boron recycle system, which is described in Section 9.3.6. Low level in the volume control tank initiates makeup from the reactor makeup control system. If the reactor makeup control system does not supply sufficient makeup to keep the volume control tank level from falling to a lower level, a low alarm is actuated. Manual action may correct the situation or, if the level continues to decrease, a low-low level signal from both level
channels causes the suction of the charging pumps to be transferred from the
volume control tank to the refueling water storage tank and closes the volume control tank outlet isolation valves. 9.3-36 Rev. 19 WOLF CREEK 9.3.4.2.1.2 Reactor Coolant Purification and Chemistry Control System Reactor coolant water chemistry specifications are given in Table 5.2-5. pH CONTROL - The pH control chemical employed is lithium hydroxide. This chemical is compatible with the materials and water chemistry of borated water/stainless steel/zirconium/inconel systems. In addition, lithium-7 is produced in the core region due to the irradiation of the dissolved boron in the coolant. The concentration of lithium-7 in the RCS is maintained in the range specified for pH control (see Table 5.2-5). If the concentration exceeds this range the cation bed demineralizer is employed in the letdown line in series operation with a mixed bed demineralizer. Since the amount of lithium to be removed is small and its buildup can be readily calculated, the flow through the cation bed demineralizer is not required to be full letdown flow. The cation
demineralizer is in use approximately 10 percent of the time. If the concentration of lithium-7 is below the specified limits, lithium hydroxide can be introduced into the RCS via the charging flow. The solution is prepared in
the laboratory and poured into the chemical mixing tank. Reactor makeup water
is then used to flush the solution to the suction manifold of the charging
pumps.OXYGEN CONTROL - During reactor startup from the cold condition, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution is introduced in accordance with plant operating procedures. Hydrazine is not employed at any time other than startup from the cold shutdown state. Dissolved hydrogen is employed to control and scavenge oxygen produced due to radiolysis of water in the core region. A sufficient partial pressure of
hydrogen is maintained in the VCT so that the specified concentration of
hydrogen is maintained in the reactor coolant. A pressure control valve maintains a minimum pressure in the vapor space of the volume control tank. This valve can be adjusted to provide the correct equilibrium hydrogen concentration (25 to 50 cc hydrogen at STP per kilogram of water). Hydrogen is
supplied from the hydrogen manifold in the service gas system. 9.3-37 Rev. 16 WOLF CREEK Mixed bed demineralizers are provided in the letdown line to provide cleanup for the letdown flow. The demineralizers remove ionic corrosion products and
certain fission products. One demineralizer is in continuous service and can be supplemented intermittently by the cation bed demineralizer, if necessary, for additional purification. The cation resin removes principally cesium and
lithium isotopes from the purification flow. The second mixed bed
demineralizer serves as a standby unit for use if the operating demineralizer
becomes exhausted during operation. A further cleanup feature is available for use during cold shutdown and operation of the residual heat removal system (RHRS). A remote-operated valve
admits a bypass flow from the RHRS into the letdown line upstream of the
letdown heat exchanger. The flow passes through the heat exchanger, a mixed bed demineralizer, and the reactor coolant filter to the VCT. The fluid is then returned to the RCS via the normal charging route. Filters are provided at various locations to ensure filtration of particulate and resin fines and to protect the seals on the reactor coolant pumps. Fission gases are removed from the reactor coolant by purging of the VCT to the gaseous waste processing system. 9.3.4.2.1.3 Reactor Makeup Control System The soluble neutron absorber (boric acid) concentration is controlled by the BTRS and by the reactor makeup control system which controls the makeup water
concentration at a pre-set value between 0 and 4.0 weight percent nominal boric
acid solution. The reactor makeup control system is also used to maintain proper reactor coolant inventory. In addition, for emergency boration and makeup, the redundant capability exists to supply borated water, at 1.4 weight percent nominal boric acid, directly from the refueling water storage tank to
the suction of the charging pumps. When this source is used for boration, letdown from the RCS is required. Emergency boration utilizing only safety grade equipment is discussed in Section 7.4. The reactor makeup control system provides a manually pre-selected makeup concentration of boric acid to the charging pump suction header or to the
volume control tank. The makeup control functions are those of maintaining desired operating level in the VCT and adjusting reactor coolant boron concentration for reactivity control. Reactor makeup water and boric acid
solution (4 weight percent nominal) are blended together to achieve the desired boron concentration for use as makeup to maintain volume control tank level or to change the reactor coolant boron concentration. 9.3-38 Rev. 14 WOLF CREEK A boron concentration measurement system (see Section 7.7) is provided to monitor the boron content of the reactor coolant in the letdown line. The
boron concentration is indicated in the main control room. The boron concentration measurement system has been abandoned-in-place. Nominal four weight percent boric acid is stored in two boric acid tanks. Two
boric acid transfer pumps are provided which are capable of supplying boric
acid, at nominal 4 weight percent, directly to the charging pumps' suction header upon remote manual demand from the main control room. The boric acid transfer pumps are normally aligned to recirculate the boric acid tank contents
via the minimum flow lines and will supply boric acid to the boric acid
blending tee upon demand of the reactor makeup control system. This boric acid
is blended with reactor makeup water and delivered to the VCT inlet or outlet for injection into the reactor coolant system. The boric acid transfer pumps are Class IE devices which are normally supplied by the Class IE power source, but have non-Class IE controls and are shed from the busses upon accident initiation. They can be manually loaded on the standby diesel generator as
needed, if offsite power is lost. All portions of the CVCS which normally contain concentrated boric acid
solution (nominally 4.0 to 4.5 weight percent boric acid), except for the normal and emergency boration lines in RM 1113, are located within a heated area in order to maintain the solution temperature at 65 F, as discussed in Section 9.4. Heat trace is installed on the normal and emergency boration lines in Room 1113 since the temperature in the room can drop below 65F, which is the solubility limit for boric acid at 7700 ppm.
The reactor makeup water pumps, taking suction from the reactor makeup water storage tank, are employed for various makeup and flushing operations throughout the systems. One of these pumps starts on demand from the reactor
makeup controller and provides flow to the boric acid blending tee or chemical
mixing tank. During reactor operation, changes are made in the reactor coolant boron
concentration for the following conditions:
- a. Reactor startup - Boron concentration must be decreased from shutdown concentration to achieve criticality.
- b. Load follow - Boron concentration must be either increased or decreased to compensate for the xenon
transient following a change in load.
- c. Fuel burnup - Boron concentration must be decreased to
compensate for fuel burnup and the buildup of fission
products in the fuel.
- d. Cold shutdown - Boron concentration must be increased to the cold shutdown concentrations.
(When in modes 3, 4 and 5, LCV112A is in the full VCT position unless boron
concentration or CVCS inventory changes are required.)
9.3-39 Rev. 25 WOLF CREEK The BTRS is normally used to control boron concentration to compensate for xenon transients. Boron thermal regeneration can also be used in conjunction
with dilution operations of the reactor makeup control system to reduce the amount of effluent to be processed by the boron recycle system. The reactor makeup control system (RMCS) can be set up for the following modes of operation:
- a. Automatic Makeup The "automatic makeup" mode of operation of the reactor makeup control system provides blended boric acid
solution to the RCS at the desired concentration. Automatic makeup compensates for minor leakage of reactor coolant without causing significant changes in the reactor coolant boron concentration. Under normal plant operating conditions, the mode selector switch is set in the "automatic makeup" position. This switch position establishes a pre-set
control signal to the total makeup flow controller and
establishes positions for the makeup stop valves for
automatic makeup. The boric acid flow controller is set to blend to the desired concentration of boric acid. A preset low level signal from the VCT level controller causes the automatic makeup control action to start a reactor makeup water pump, start a boric acid transfer pump, open the makeup stop valve to the charging pump suction, and position the boric acid flow control valve and the reactor makeup water flow control valve. The flow controllers then blend the makeup stream according to the
preset concentration. Makeup addition to the charging pump suction header causes the water level in the VCT to rise. At a preset high level point, the makeup is stopped. This operation may be terminated manually at
any time. If the automatic makeup fails or is not aligned for operation and the VCT level continues to decrease, a low level alarm is actuated. Manual action may correct the situation or, if the level continues to decrease, a low-
low level signal opens the isolation valves in the
refueling water supply line to the charging pumps and closes the isolation valves in the VCT outlet line. 9.3-40 Rev. 16 WOLF CREEK
- b. Dilution The "dilute" mode of operation permits the addition of a preselected quantity of reactor makeup water at a preselected flow rate to the RCS. The operator sets the
mode selector switch to "dilute," the total makeup flow
controller setpoint to the desired flow rate, and the
total makeup batch integrator to the desired quantity and initiates system start. This opens the reactor makeup water flow control valve, opens the makeup
isolation valve to the VCT inlet, and starts a reactor makeup water pump. Excessive rise of the VCT water
level is prevented by automatic actuation (by the tank level controller) of a three-way diversion valve which routes the reactor coolant letdown flow to the boron recycle system. When the preset quantity of water has
been added, the batch integrator causes the makeup to
stop. The operation may be terminated to stop and the operation may be terminated manually at any time. Dilution can also be accomplished by operating the BTRS in the boron storage mode, as described in Section
9.3.4.2.1.4.
- c. Alternate Dilution The "alternate dilute" mode of operation is similar to the dilute mode, except that a portion of the dilution water flows directly to the charging pump suction and a portion flows into the VCT via the spray nozzle and then
flows to the charging pump suction. This decreases the
delay in diluting the RCS caused by directing dilution
water to the VCT inlet.
- d. Boration The "borate" mode of operation permits the addition of a preselected quantity of concentrated boric acid solution at a preselected flow rate to the RCS. The operator sets the mode selection switch to "borate," the
concentrated boric acid flow controller setpoint to the
desired flow rate, and the concentrated boric acid batch
integrator to the desired quantity and initiates system start. This opens the makeup isolation valve to the charging pumps suction, positions the boric acid flow
control valve, and starts the selected boric acid transfer pump, which delivers a nominal 4 weight percent 9.3-41 Rev. 4 WOLF CREEK boric acid solution to the charging pumps suction header. The total quantity added in most cases is so
small that it has only a minor effect on the VCT level. When the preset quantity of concentrated boric acid solution is added, the batch integrator causes the
makeup to stop. Also, the operation may be terminated
manually at any time. Boration can also be accomplished by operating BTRS in the boron release mode, as described in Section
9.3.4.2.1.4.
- e. Manual The "manual" mode of operation permits the addition of a preselected quantity and blend of the boric acid
solution to the refueling water storage tank, to the
recycle holdup tanks in the boron recycle system, to the boron thermal regeneration system, and to the fuel storage pool, or to some other location via a temporary connection. While in the manual mode of operation, automatic makeup to the RCS is precluded. The
discharge flow path must be prepared by opening the manual valves in the desired path. The operator sets the mode selector switch to "manual," the boric acid and total makeup flow controllers to the
desired flow rates, and the boric acid and total makeup batch integrators to the desired quantities and actuates the makeup start switch. Reactor makeup water can be used to de-borate the boron thermal regeneration demineralizers using the manual switch settings. The start switch actuates the boric acid flow control valve and the reactor makeup water flow control valve
and starts the reactor makeup water pump and boric acid transfer pump. When the preset quantities of boric acid and reactor makeup water have been added, the batch integrators
cause the makeup to stop. This operation may be stopped manually by actuating the makeup stop switch. If either batch integrator is satisfied before the other has recorded its required total, the pump and valve associated with the integrator which has been satisfied
will terminate the flow. The flow controlled by the other integrator will continue until that integrator is satisfied. In the manual mode, the boric acid flow is terminated first to prevent the piping systems from remaining filled with nominal 4 weight percent boric
acid solution. 9.3-42 Rev. 14 WOLF CREEK The quantities of boric acid and reactor makeup water injected are totalized by the batch counters, and the
flow rates are recorded on strip recorders. Deviation alarms sound for both boric acid and reactor makeup water if the flow rates deviate from the setpoints. 9.3.4.2.1.4 Boron Thermal Regeneration System Downstream of the mixed bed demineralizers, the letdown flow can be diverted to the BTRS when boron concentration changes are desired. After processing by the
BTRS, the flow is returned to the letdown flow path at a point upstream of the
reactor coolant filter. The boron concentration measurement system (see Section 7.7) is used to monitor the boron concentration in the letdown stream before it is diverted to the BTRS for processing or to monitor the adjusted boron concentration of the letdown stream after it has been treated by the thermal regeneration process. The boron concentration measurement system has been abandoned-in-place. Storage and release of boron is determined by the temperature of the fluid entering the thermal regeneration demineralizers. A chiller unit and a group of heat exchangers are employed to provide the desired fluid temperatures at the demineralizer inlets for either storage or release operation of the system. The flow path through the boron thermal regeneration system is different for the dilution and the boration operations. During dilution, the letdown stream enters the moderating heat exchanger, and from there it passes through the letdown chiller heat exchanger. These two heat exchangers cool the letdown stream prior to its entering the demineralizers. The letdown reheat heat exchanger is valved out on the tube side and performs no function during boron storage operations. The temperature of the letdown stream at the point of entry to the demineralizers is controlled
automatically by the temperature control valve which controls the shell side flow to the letdown chiller heat exchanger. After passing through the demineralizers, the letdown enters the moderating heat exchanger shell side
where it is heated by the incoming letdown stream before returning to letdown line.For dilution, a decrease in the boric acid concentration in the reactor coolant is accomplished by sending the letdown flow at relatively low temperatures to the thermal regeneration demineralizers. The resin, which was depleted of
boron at high temperature during a prior boron release operation, is now capable of 9.3-43 Rev. 14 WOLF CREEK storing boron from the low temperature letdown stream. Reactor coolant with a decreased concentration of boric acid leaves the demineralizers and is returned
to the letdown line. During the boration operation, the letdown stream enters the moderating heat exchanger tube side, bypasses the letdown chiller heat exchanger, and passes through the shell side of the letdown reheat heat exchanger. The moderating and letdown reheat heat exchangers heat the letdown stream prior to its entering the resin beds. The temperature of the letdown at the point of entry to the demineralizers is controlled automatically by the temperature control
valve which controls the flow rate on the tube side of the letdown reheat heat exchanger. After passing through the demineralizers, the letdown stream enters
the shell side of the moderating heat exchanger, passes through the tube side of the letdown chiller heat exchanger, and then goes to the VCT via the reactor coolant filter and letdown line. The temperature of the letdown stream entering the VCT is controlled automatically by adjusting the shell side flow
rate on the letdown chiller heat exchanger. Thus, for boration, an increase in
the boric acid concentration in the reactor coolant is accomplished by sending the letdown flow at relatively high temperatures to the thermal regeneration demineralizers. The water flowing through the demineralizers now results in
boron being released which was stored by the resin at low temperature during a
previous boron storage operation. The boron enriched reactor coolant is
returned to the RCS via the charging system portion of the CVCS. Although the boron thermal regeneration system is primarily designed to compensate for xenon transients occurring during load follow, it can also be
used to handle boron changes during other modes of plant operation. During
startup dilution, for example, the resin beds are first saturated, then washed off. This operation continues until the desired dilution in the RCS is obtained. This method of startup serves to reduce the effluents diverted to
the boron recycle system. As an additional function, a thermal regeneration demineralizer can be used as a deborating demineralizer, which can be used to dilute the RCS down to very low boron concentrations toward the end of a core cycle. To make such a bed
effective, the effluent concentration from the bed must be kept very low, close
to zero ppm boron. This low effluent concentration can be achieved by using
fresh resin. Use of fresh resin can be coupled with the normal replacement cycle of the resin, one resin bed being replaced during each core cycle. 9.3-44 Rev. 0 WOLF CREEK 9.3.4.2.2 Component Description Codes and standards applicable to the CVCS are listed in Tables 3.2-1 and 9.3-
- 9. The CVCS is designed and constructed in accordance with the following quality group requirements:
Reactor coolant system boundary valves and piping are quality group A; the letdown, charging, and seal water system and associated containment penetrations are quality group B; the boric acid transfer system is quality group C; and the coolant purification and BTRS are quality group D (augmented)
in accordance with Regulatory Guide 1.143 for radioactive waste management
systems. The quality group A, B, and C portions are seismic Category I. The
entire CVCS is located within seismic Category I structures. CHARGING PUMPS - Three charging pumps are supplied to inject coolant into the RCS. The pumps are of the single speed, horizontal, centrifugal type. The
100-percent redundant centrifugal charging pumps are powered from separate
Class IE sources, while the normal charging pump is powered from a non-Class IE source. In the USAR, the safety related pumps are always referred to as centrifugal charging pumps and non-safety related pumps are referred to as normal charging pumps. Where the term charging pump is used, it implies any of three pumps. All parts in contact with the reactor coolant are fabricated of austenitic stainless steel or other corrosion-resistant material. To prevent leakage to the atmosphere, the charging pump seals are provided with leakoffs to collect the leakage. There are minimum flow recirculation lines to protect the charging pumps from a closed discharge valve condition. The charging flow rate is determined from a pressurizer level signal. Charging
flow control is accomplished by a modulating valve on the discharge side of the charging pumps. The standby centrifugal charging pumps also serve as high-head safety injection pumps in the emergency core cooling system. A description of the centrifugal charging pump function upon receipt of a safety injection signal is given in Section 6.3.2.2. BORIC ACID TRANSFER PUMPS - Two 100-percent redundant canned motor pumps are supplied per unit. The pumps are Class IE devices powered through a qualified
isolation device from Class IE sources with non-Class IE controls and are shed
on a safety injection 9.3-45 Rev. 15 WOLF CREEK signal. In the event of loss-of-offsite power, the pumps can be manually loaded on separate Class IE (diesel backed) sources. A complete description of
this capability is provided in Chapters 7.0 and 8.0. The boric acid transfer pumps are normally aligned to supply boric acid to the suction header of the charging pumps. Manual or automatic initiation of the reactor coolant makeup
system will start one pump to provide normal makeup of boric acid solution to
the suction header of the charging pumps. Mini-flow from this pump flows back
to the associated boric acid tank and helps maintain thermal equilibrium. The standby pump can be used intermittently to circulate the boric acid solution through the other tank to maintain thermal equilibrium in this part of the
system. The transfer pumps also function to transfer boric acid solution from the batching tank to the boric acid tanks. Emergency boration, in which nominal 4 weight percent boric acid solution is supplied directly to the suction of the charging pumps, can be accomplished by manually starting either or both pumps. This is the preferred emergency boration mode if all components are available, rather than using the refueling
water storage tank. The pumps are located in a heated area to prevent crystalization of the boric acid solution. All parts in contact with the solution are of austenitic stainless steel. An alternate discussion on
boration is provided in Section 7.4 in conjunction with a discussion of the features of safety related cold shutdown designs. CHILLER PUMPS - Two centrifugal pumps circulate the water through the chilled water loop in the BTRS. One pump is normally operated, with the second serving
as a standby. REGENERATIVE HEAT EXCHANGER - The regenerative heat exchanger is designed to recover heat from the letdown flow by reheating the charging flow, which reduces the thermal effects on the charging connections to the reactor coolant
loop piping. The letdown stream flows through the shell of the regenerative heat exchanger while the charging stream flows through the tubes. The unit is constructed of austenitic stainless steel, and is of all-welded construction. The temperatures of both outlet streams from the heat exchanger are monitored with indication given in the control room. A high temperature alarm is actuated on the main control board if the temperature of the letdown stream exceeds the desired limits. LETDOWN HEAT EXCHANGER - The letdown heat exchanger cools the letdown stream to the operating temperature of the mixed bed demineralizers. Reactor coolant flows through the tube side of 9.3-46 Rev. 14 WOLF CREEK the exchanger while component cooling water flows through the shell side. All surfaces in contact with the reactor coolant are austenitic stainless steel, and the shell is carbon steel. The low pressure letdown valve, located downstream of the heat exchanger, maintains the pressure upstream of the heat exchanger in a range sufficiently high to prevent flashing downstream of the letdown flow orifices. Pressure indication and high pressure alarm are provided on the main control board. The letdown temperature control indicates and controls the temperature of the letdown flow exiting from the letdown heat exchanger. The exit temperature of
the letdown stream is controlled by regulating the component cooling water flow
through the letdown heat exchanger. Temperature indication is provided on the main control board. If the outlet temperature from the heat exchanger is excessive, a high temperature alarm is actuated, and a temperature controlled three-way valve diverts the letdown directly to the reactor coolant filter bypassing the CVCS demineralizers. The outlet temperature from the shell side of the heat exchanger is allowed to vary over an acceptable range compatible with the equipment design parameters
and required performance of the heat exchanger in reducing letdown stream
temperature. EXCESS LETDOWN HEAT EXCHANGER - The excess letdown heat exchanger cools reactor coolant excess letdown flow. The flow rate is equivalent to the portion of the
nominal seal injection flow which flows into the RCS through the reactor
coolant pump labyrinth seals. The excess letdown heat exchanger can be employed either when normal letdown is temporarily out of service to maintain the reactor in operation, to supplement
maximum letdown during the final stages of heatup, or to provide a letdown path
from the RCS to the pressurizer relief tank. The letdown flows through the
tube side of the unit, and component cooling water is circulated through the shell. All surfaces in contact with reactor coolant are austenitic stainless steel, and the shell is carbon steel. All tube joints are welded. A temperature detector measures the temperature of the excess letdown flow downstream of the excess letdown heat exchanger. Temperature indication and high temperature alarm are provided on the main control board. 9.3-47 Rev. 0 WOLF CREEK A pressure sensor indicates the pressure of the excess letdown flow downstream of the excess letdown heat exchanger and excess letdown control valve.
Pressure indication is provided on the main control board. Redundant temperature detectors measure the temperature of the letdown flow from the excess letdown heat exchanger to the pressurizer relief tank. Redundant flow detectors measure the flow rate of the letdown flow from the excess letdown heat exchanger to the pressurizer relief tank. SEAL WATER HEAT EXCHANGER - The seal water heat exchanger is designed to cool fluid from three sources: reactor coolant pump number 1 seal leakage, reactor
coolant discharged from the excess letdown heat exchanger, and miniflow from the charging pumps. Reactor coolant flows through the tube side of the heat exchanger, and component cooling water is circulated through the shell. The design flow rate through the tube side is equal to the sum of the nominal excess letdown flow, maximum design reactor coolant pump seal leakage, and miniflow from one charging pump. The unit is designed to cool the above flow to the temperature normally maintained in the VCT. All surfaces in contact
with reactor coolant are austenitic stainless steel, and the shell is carbon
steel.MODERATING HEAT EXCHANGER - The moderating heat exchanger operates as a regenerative heat exchanger between incoming and outgoing streams to and from the boron thermal regeneration demineralizers. The incoming letdown flow enters the tube side of the moderating heat exchanger. The shell side fluid, which comes directly from the thermal regeneration demineralizers, enters at low temperature during boron storage and high temperature during boron release. LETDOWN CHILLER HEAT EXCHANGER - During the boron storage operation, the process stream enters the tube side of the letdown chiller heat exchanger after leaving the tube side of the moderating heat exchanger. The letdown chiller heat exchanger cools the process stream to allow the thermal regeneration demineralizers to remove boron from the coolant. The desired cooling capacity
is adjusted by controlling the chilled water flow rate passed through the shell
side of the heat exchanger. The letdown chiller heat exchanger is also used during the boron release operation to further cool the liquid leaving the moderating heat exchanger shell side to ensure that its temperature does not exceed that of normal
letdown to the VCT. 9.3-48 Rev. 15 WOLF CREEK LETDOWN REHEAT HEAT EXCHANGER - The letdown reheat heat exchanger is used only during boration operations to heat the process stream. Water used for heating
is diverted from the letdown line upstream of the letdown heat exchanger, passed through the tube side of the letdown reheat heat exchanger, and then returned to the letdown stream upstream of the letdown heat exchanger. VOLUME CONTROL TANK - The VCT provides surge capacity for part of the reactor coolant expansion volume not accommodated by the pressurizer. When the level in the tank reaches the high level setpoint, the remainder of the expansion volume is accommodated by diversion of the letdown stream, upstream of the VCT, to the boron recycle system. To prevent large hydrogen pressure swings in the VCT, and undesirable hydrogen concentrations in the RCS, it is desirable to maintain the level at the upper limit of the normal range. For optimal hydrogen pressure control, the level should be maintained between 50 and 60% whenever possible. The tank also provides a means for introducing hydrogen to the coolant to maintain the required equilibrium concentration of 25 to 50 cc hydrogen (at STP) per kilogram of water and is used for degassing the reactor coolant. It also serves as a head tank for the charging pumps. A spray nozzle located inside the tank on the letdown line provides liquid-to-gas contact between the incoming fluid and the hydrogen atmosphere in the tank. Hydrogen (from the service gas system) is supplied to the VCT. If it is desired to remove gaseous fission products, which are stripped from the reactor coolant and collected in this tank, a remotely operated vent valve, discharging to the gaseous waste processing system, can be placed in service. Relief protection, gas space sampling, and nitrogen purge connections are also provided. The tank can also accept the seal water return flow from the reactor coolant pumps, although this flow normally goes directly to the suction of the charging pumps. VCT pressure is monitored with indication given in the control room. An alarm is actuated in the control room for high and low pressure conditions. The VCT pressure control valve is automatically closed by the low pressure signal. Three level channels govern the water inventory in the VCT. Redundant level indication is provided on the main control board from two level channels.
Local level indication with a high-low alarm on the main control board is provided from the third channel. If the VCT level rises above the normal operating range of 30 - 60%, one level channel provides an analog signal to the proportional controller which modulates the three-way valve downstream of the reactor coolant filter to maintain the VCT level within the normal operating band. The three-way valve can split letdown flow so that a 9.3-49 Rev. 19 WOLF CREEK portion goes to the boron recycle system and a portion to the VCT. The controller would operate in this fashion during a dilution operation when
reactor makeup water is being fed to the VCT from the reactor makeup control system.If the modulating function of the channel fails and the VCT level continues to rise, the high level alarm will alert the operator to the malfunction, and the full letdown flow is diverted to the recycle hold-up tank. During normal power operation, a low level in the VCT initiates automatic makeup which injects a preselected blend of boric acid solution and reactor
makeup water into the charging pump suction header. When the volume control
tank level is restored to normal, automatic makeup stops. If the automatic makeup fails or is not aligned for operation and the tank level continues to decrease, a low level alarm is actuated. Manual action may
correct the situation, or if the level continues to decrease, a low-low signal
from either of the two redundant level channels opens its associated stop valve in the RWST supply line, and closes its associated stop valve in the VCT outlet line. For a description of the VCT level controls, refer to Section 7.6.11. BORIC ACID TANKS - The combined BAT capacity is sized to store sufficient boric acid solution for refueling plus enough for a cold shutdown from full-power operation immediately following refueling with the most reactive control rod not inserted. The concentration of boric acid solution in storage is maintained between 4 and
4.5 percent
nominal by weight. Periodic manual sampling and corrective action, if necessary, assure that these limits are maintained. Therefore, measured amounts of boric acid solution can be delivered to the reactor coolant to
control the prevailing boron concentration. A temperature sensor provides the temperature measurement of the contents of each tank. Temperature indication, as well as high and low temperature alarms, are provided on the main control board. Two level detectors indicate the level in each boric acid tank. Level indication with high, low, low-low, and empty level alarms is provided on the main control board. The high alarm indicates that the BAT may soon overflow. The low alarm warns the operator 9.3-50 Rev. 4 WOLF CREEK to start makeup to the BAT. The low-low alarm is set to indicate the minimum level of boric acid in the BAT to ensure that sufficient boric acid is
available for a cold shutdown with one stuck rod. The empty level alarm is set to give warning of loss of pump suction. BATCHING TANK - The batching tank is used for mixing a makeup supply of boric acid solution for transfer to the boric acid tanks. A local sampling point is provided for verifying the solution concentration prior to transferring it out of the tank. The tank is provided with an
agitator to improve mixing during batching operations and a steam jacket for
heating the boric acid solution. CHEMICAL MIXING TANK - The primary use of the chemical mixing tank is in the preparation of caustic solutions for pH control, hydrazine solution for oxygen scavenging, and chemicals for corrosion products oxidation during a refueling shutdown.CHILLER SURGE TANK - The chiller surge tank handles the thermal expansion and contraction of the water in the chiller loop. The surge volume in the tank also
acts as a thermal buffer for the chiller. The fluid level in the tank is
monitored with level indication, and high and low level alarms are provided on
the main control board. MIXED BED DEMINERALIZERS - Two flushable mixed bed demineralizers assist in maintaining reactor coolant purity. A lithium form cation resin and hydroxyl
form anion resin are charged into the demineralizers for normal operation while
a hydrogen form cation resin and hydroxyl form anion resin can be used during refueling outage cleanup. The anion resin is converted to the borate form in operation. Both types of resin remove fission and corrosion products. The
resin bed is designed to reduce the concentration of ionic isotopes in the
purification stream, except for cesium, yttrium, and molybdenum, by a minimum
factor of 10. Each demineralizer has more than sufficient capacity for one core cycle with 1 percent of the rated core thermal power being generated by defective fuel rods. One demineralizer is normally in service with the other in standby. A temperature sensor monitors the temperature of the letdown flow downstream of the letdown heat exchanger. If the letdown temperature exceeds the maximum
allowable resin operating temperature (approximately 140 F), a three-way valve
is automatically actuated so that the flow bypasses the demineralizers.
Temperature indication and high alarm are provided on the main control board. The air-operated, three-way valve failure mode directs flow to the VCT via the reactor coolant filter. 9.3-51 Rev. 9 WOLF CREEK CATION BED DEMINERALIZER - A flushable cation resin bed in the hydrogen form is located downstream of the mixed-bed demineralizers and is used intermittently
to control the concentration of Li-7 (pH control) in the reactor coolant system. Its size is based upon the estimated production of Li-7 in the reactor core region due to the B 10 (n) Li 7 reaction during base load operation. The demineralizer also has sufficient capacity to maintain the cesium-137 concentration in the coolant below 1.0Ci/cc with 1 percent defective fuel. The resin bed is designed to reduce the concentration of ionic isotopes, particularly cesium, yttrium, and molybdenum, by a minimum factor of 10. The demineralizer has more than sufficient capacity for one core cycle with 1 percent of the rated core thermal power being generated by defective fuel rods. THERMAL REGENERATION DEMINERALIZERS - The function of the thermal regeneration demineralizers is to store the total amount of boron that must be removed from the RCS to accomplish the required dilution during a load cycle in order to compensate for xenon buildup resulting from a decreased power level. Furthermore, the demineralizers must be able to release the previously stored
boron to accomplish the required boration of the reactor coolant during the load cycle in order to compensate for a decrease in xenon concentration resulting from an increased power level. The thermally reversible ion storage capacity of the resin applies only to borate ions. The capacity of the resin to store other ions is not thermally
reversible. Thus, during boration, when borate ions are released by the resin, there is no corresponding release of the ionic fission and corrosion products stored on the resin. The thermal regeneration demineralizer resin capacity is directly proportional to the solution boron concentration and inversely proportional to the
temperature. Further, the differences in capacity as a function of both boron concentration and temperature are reversible. For the 50 F to 140 F temperature cycle, this reversible capacity varies from the beginning of a core cycle to the end of core life by a factor of about 2. The demineralizers are of the type that can accept flow in either direction. The flow direction during boration is therefore always opposite to that during release. This provides faster response when the beds are switched from storage to release and vice versa than would be the case if the demineralizers could
accept flow in only one direction. 9.3-52 Rev. 1 WOLF CREEK Temperature instrumentation is provided upstream of the thermal regeneration demineralizers to control the temperature of the process flow. During dilution operations, it controls the flow through the shell side of the letdown chiller heat exchanger to maintain the process flow at 50 F as it enters the demineralizers. During boration operations, it controls the flow through the tube side of the letdown reheat heat exchanger to maintain the process flow at
140 F as it enters the demineralizers. Temperature indication and a high
temperature alarm are provided on the main control board. An additional temperature instrument is provided to protect the demineralizer resins from a high temperature condition. On reaching the high temperature set
point, an alarm is sounded on the main control board, and the letdown flow is
diverted to the VCT from a point upstream of the mixed bed demineralizers. Failure of the temperature controls resulting in hot water flow to the demineralizers would result in a release of boron stored on the resin with a
resulting increase in reactor coolant boron concentration and increased margin
for shutdown. If the temperature of the resin rises significantly above 140 F, the number of ion storage sites on the resin will gradually decrease, thus reducing the capability of the resin to remove boron from the process stream.
Degradation of ion-removal capability will occur for temperatures of approximately 160 F and above. The extent of the degradation and rate at which
it will occur depend upon the temperature experienced by the resin and the length of time that the resin experiences this elevated temperature. Failure of the temperature control system resulting in cold water flow to the demineralizers would result in storage of boron on the resin and reduction of
the reactor coolant boron concentration. The amount of reduction in the reactor coolant boron concentration is limited by the capacity of the resin to remove boron from the water. As the boron concentration is reduced, the control rods
would be driven into the core to maintain the power level. If the rods were to
reach the shutdown limit set point, an alarm would be actuated informing the
operator that emergency boration of the RCS is necessary in order to maintain the capability of shutting the reactor down with control rods alone. REACTOR COOLANT FILTER - The reactor coolant filter is located in the letdown line upstream of the VCT. The filter collects resin fines and particulates
from the letdown stream. The nominal flow capacity of the filter is greater than the maximum letdown flow rate. A differential pressure indicator is provided to show the differential pressure drop across the reactor coolant
filter. 9.3-53 Rev. 13 WOLF CREEK SEAL WATER INJECTION FILTERS - Two seal water injection filters are located in parallel in a common line to the reactor coolant pump seals; they collect
particulate matter that could be harmful to the seal faces. Each filter is sized to accept flow in excess of the normal seal water flow requirements. A differential pressure indicator monitors the pressure drop across each seal water injection filter and gives local indication with high differential pressure alarm on the main control board. Redundant safety related flow monitoring is provided downstream of seal water injection filters. SEAL WATER RETURN FILTER - This filter collects particulates from the reactor coolant pump seal water return and from the excess letdown flow. The filter is designed to pass the sum of the excess letdown flow and the maximum design leakage from all reactor coolant pumps. A differential pressure indicator is provided to show the differential pressure across the seal water return filter. BORIC ACID FILTER - The boric acid filter collects particulates from the boric acid solution being pumped from the boric acid tanks by the boric acid transfer
pumps. The filter is designed to pass the design flow of two boric acid transfer pumps operating simultaneously. A differential pressure indicator is provided to show the differential pressure across the boric acid filter. LETDOWN ORIFICES - Three letdown orifices are provided to reduce the letdown pressure from reactor conditions and to control the flow of reactor coolant leaving the RCS. The orifices are placed into or out of service by remote
operation of their respective isolation valves. Two of the orifices are
designed for a normal letdown flow of 75 gpm, and the third orifice is designed for 45 gpm. During normal power operation 45 gpm orifice may be used to attain the desired letdown flowrate. Any combination of the three orifices may be
used for flow control at low RCS pressures, such as plant startup, when maximum
letdown is desirable. Each orifice consists of an assembly which provides for
permanent pressure loss without recovery, and is made of austenitic stainless steel or other adequate corrosion resistant material. A flow monitor provides indication in the control room of the letdown flow rate and an alarm to indicate unusually high flow. 9.3-54 Rev. 0 WOLF CREEK A low pressure letdown controller located downstream of the letdown heat exchanger controls the pressure upstream of the letdown heat exchanger to
prevent flashing of the letdown liquid. Pressure indication and high pressure alarm are provided on the main control board. CHILLER - The chiller is located in a chilled water loop containing a surge tank, chiller pumps, the letdown chiller heat exchanger, piping, valves, and controls.The purpose of the chiller is twofold:
- a. To cool down the process stream during storage of boron on the resin.
- b. To maintain an outlet temperature from the BTRS at or below 115 F during release of the boron.
VALVES - Where functional requirements permit, elastomere diaphragm-type valves or packless globe valves are used to essentially eliminate leakage to the atmosphere. All packed valves which are larger than 2 inches and which are
designated for radioactive services are provided with graphite packing. All control (modulating) and three-way valves are either provided with stuffing box and leakoff connections or are totally enclosed. Leakage to the atmosphere is essentially zero for these valves. Basic material of construction is stainless steel for all valves which handle radioactive liquid or boric acid solutions. All active, power-operated valves which are required to realign the CVCS for emergency core cooling, to isolate the containment, or are utilized as part of
the safety-related cold shutdown design are energized from Class IE sources. Normal letdown, purification, reactor makeup control, and BTRS power-operated valves, which are not required for emergency core cooling or containment isolation, fail to the safe position and are powered from non-Class IE sources.
However, in the event of a loss of offsite power, selected valves in the boric
acid transfer system can be manually loaded on a Class IE (diesel-backed) bus. Relief valves are provided for lines and components that might be pressurized above design pressure by improper operation or component malfunction. 9.3-55 Rev. 5 WOLF CREEK PIPING - All CVCS piping that handles radioactive liquid is austenitic stainless steel. 9.3.4.2.3 System Operation Operation of the CVCS is described for the various phases of reactor plant operation presented below. 9.3.4.2.3.1 Plant Startup Plant startup is defined as the operations which bring the reactor plant from the cold shutdown condition to normal, no-load operating temperature and pressure, and subsequently to full-power operation. During cold shutdown, the CVCS is employed periodically to provide cleanup of a portion of the refueling water being cooled by the residual heat removal system (RHRS).The charging and RHR pumps may be used to initially fill and pressurize the RCS (RCS temperature must be above 70 F). During filling, makeup water is drawn from the reactor makeup water storage tank and blended, using the reactor makeup control system, with boric acid, to provide makeup water at the administratively controlled RCS boron concentration. The RWST and RHUT are also available as borated water sources. A vacuum may be applied to the RCS to enhance the filling process via the reactor vessel head and the pressurizer. The reactor coolant system is vented via the reactor vessel head. The pressurizer and RCS are vented, as a minimum to the containment purge system, containment atmosphere, or pressurizer relief tank. Following the filling operations, the RCS is cold and water solid at low pressure. Special
precautions are exercised to assure that an overpressurization transient does
not occur. Overpressure protection is discussed in Section 5.2.2. The charging pumps are employed to increase the RCS pressure while letdown continues through the RHRS to the CVCS. (Note that throttling the RHRS or stopping of the RHR pump can result in a RCS pressure increase of 100-150 psi.)
The operator assures that the reactor vessel's allowable pressure/temperature
relationship is not exceeded. The RCS pressure and letdown flowrate is monitored as the flow control valve is manually opened. The manual throttle valves in each of the RCP seal water supply lines are set to provide a flow of 8 to 13 gpm per RCP. Seal water is supplied to the idle reactor coolant pumps by throttling BGHCV182. When the reactor coolant temperature exceeds 150 F, seal water and component cooling water are supplied to the RCPs. 9.3-56 Rev. 14 WOLF CREEK The reactor coolant pressure is initially maintained by the letdown pressure control valve PCV-131. All three letdown orifices are normally in use. The initial low pressure in the RCS will result in insufficient flow through the orifices, with the requirement that a bypass loop must be established. This is accomplished by diverting a portion of the RHRS flow to either the letdown heat exchanger (downstream of the orifices) normally by fully opening either EJ-V001, or EJ-V002, and throttling HCV-128. A nitrogen cover gas of 15 psig minimum should be maintained in the VCT. The rate of increase of system pressure is controlled by a manual operation of
the letdown pressure control valve PCV-131, the charging flow control valve FCV-121, and the RHR cleanup flow control valve HCV-128. Gradually, the letdown pressure control valve is closed to increase the RCS pressure to approximately 400 psig. If desired, the letdown pressure control valve may be reset in AUTO to maintain a pressure of about 350 psig in the RCS. Pressurization is then controlled by the charging system. When the reactor coolant system pressure has reached an indicated pressure of 325-425 psig, and the P across the RCP's No. 1 seals are satisfactory, the RCPs may be started. The reactor coolant system heats up due to the reactor coolant pump heat input, pressurizer heaters, and residual heat addition; hence, excess coolant resulting from fluid expansion will accumulate in the VCT. The VCT level rises and the nitrogen cover gas is expelled to the gaseous radwaste system. As soon as high level is reached in the VCT, the nitrogen supply is secured, and the hydrogen makeup valve is brought into operation. During this operation, the VCT pressure is maintained at 15 to 50 psig by the pressure control valve PCV-115 in the gaseous vent line. The VCT is allowed to decrease
to normal volume by manually diverting the letdown to the recycle holdup tanks. This operation establishes the hydrogen over-pressure in the volume control tank.Chemical treatment such as hydrazine addition is accomplished as required, prior to the RCS reaching 250 F. The mixed bed demineralizers are bypassed
during chemical treatment to avoid driving lithium off the bed and replacing it with the ammonia which is formed during startup as a result of the hydrazine/oxygen reaction. After oxygen scavenging is complete, the pressurizer spray valves are closed to allow the pressurizer to heat up independent of the main reactor coolant loops. 9.3-57 Rev. 14 WOLF CREEK When the pressurizer is raised to the saturation temperature corresponding to the RCS pressure, a steam bubble is formed in the pressurizer. The RHR loop is
isolated. The low-pressure letdown control valve is now set to AUTO to maintain a pressure of approximately 300-350 psig downstream of the letdown orifices. The charging flow control valve (if centrifugal pump is used) is
controlled manually to obtain normal water level (no-load) in the pressurizer. The normal charging pump flow control valve BGFCV462 can be placed in AUTO following attainment of normal water level in the pressurizer. As heatup proceeds, it is necessary to provide extra letdown flow capability in order to
maintain maximum heatup rate. The amount of letdown flow attainable is limited by the regenerative heat exchanger (a maximum of 380 F is allowed at the outlet
from the heat exchanger upstream of the letdown orifices), and the rate of expansion of the coolant due to heatup as reflected by the pressurizer level. The excess letdown heat exchanger may be employed as the reactor coolant
temperature approaches no load T-avg to allow continued maximum RCS heatup
rate.Following chemical analysis to confirm that water quality, boron concentration, and hydrogen concentration are within specification, criticality is achieved by appropriate rod withdrawal; prior reduction of boron concentration by dilution
will be required. The RMCS dilution mode, alternate dilution mode may be used. Exception to this is initial startup following refueling, during low power physics testing, where criticality is achieved through boron dilution. The dilution mode of operation permits the addition of a preselected quantity of reactor makeup water at a preselected flow rate to the reactor coolant
system.Although the BTRS is primarily designed to compensate for xenon transients occurring during load follow, it can also be used to reduce the RCS concentration during a startup. During RCS dilution, for example, the letdown flow passes through the resin beds until they become saturated. The beds can then be washed off by reactor makeup water into the recycle holdup tanks. This operation continues until the desired dilution in the RCS is obtained. As compared with a feed-and-bleed operation, the use of the BTRS demineralizers
for dilution during startup reduce the amount of liquid diverted to the recycle holdup tanks. Further adjustments in boron concentration by operation of the RMCS to establish preferred control-group rod positions and to compensate for xenon
buildup are also necessary. 9.3-58 Rev. 15 WOLF CREEK Following attainment of full RCS pressure and temperature, the letdown orifices are set for normal letdown, and the excess letdown flow may be terminated. During the heatup phase it should not be necessary to adjust the seal water injection valves; however, some adjustment of the charging line flow control valve may be required to maintain the required seal injection flow rate. 9.3.4.2.3.2 Normal Operation Normal operation includes operation at steady power (base load) level, load follow operation, and hot standby. BASE LOAD - At a constant power level, the rates of charging and letdown are dictated by the requirements for seal water to the reactor coolant pumps and the normal purification of the reactor coolant system. One charging pump is employed, and the flow is controlled automatically from pressurizer level. The
only adjustments in boron concentration are those necessary to compensate for
core burnup. These adjustments are made to maintain the maneuvering band of the rod control groups within their allowable limit. Rapid variations in power demand will be
accommodated automatically by control rod movement. If variations in power
level occur, and the new power level is sustained for long periods, some adjustment in boron concentration may be necessary to ensure the preservation of the shutdown margin. During normal operation, the letdown flow is 75 or 120 gpm, and one mixed bed demineralizer is in service. Reactor coolant samples are taken at frequent intervals to check boron concentration, water quality, pH, and activity level. The normal charging-pump flow control valve (FCV-462) is modulated by the
pressurizer water level at the set point programmed for a prevailing reactor
coolant average temperature. During normal operation with maximum
purification, the letdown flow is 120 gpm. If a standby centrifugal charging pump is employed, the charging flow control valve (FCV-121) is modulated by pressurizer water level. Operation of the BTRS is automatic. A master switch is provided which puts the BTRS in the right mode of operation for release or storage of boron on the resin beds. This switch performs the following functions for storage (RCS
dilution): 9.3-59 Rev. 15 WOLF CREEK
- a. Aligns the proper flow path for dilution.
Valves open Valves closed Valves modulating HV-7054 HV-8245 TCV-386 UV-7002A UV-7040 UV-7002B UV-7041
UV-7056 TCV-381A Deleted UV-7057
UV-7045 UV-7046 UV-7022
- b. Shuts off the letdown reheat heat exchanger tube side flow which puts this heat exchanger out of operation
(closes TCV-381A).
- c. Transfers control of TCV-386, the control valve at the letdown chiller heat exchanger shell side outlet, to
TCY-381B which is located between the letdown reheat heat exchanger and the BTRS demineralizers. The temperature set point is 50 F.
- d. Starts chiller and chiller pump.
For release (RCS boration), the master switch performs the following functions:
- a. Aligns the proper flow path for boration.
Valves open Valves closed Valves modulating HV-7054 HV-8245 TCV-386 UV-7041 UV-7002A TCV-381A UV-7057 UV-7002B Deleted Deleted UV-7056 UV-7046 UV-7045 UV-7040 UV-7022
- b. Energizes the control of TCV-381A for the tube side flow rate to the letdown reheat heat exchanger by a signal from TCY-381A located between this heat exchanger and
the BTRS demineralizers. The temperature set point is 140 F.
- c. Transfers control of the control valve TCV-386 at letdown chiller heat exchanger shell side outlet to TCY-386 located in the line leading from the moderating heat exchanger to the reactor coolant filter. The temperature set point is 115 F.
- d. Starts chiller and chiller pump. 9.3-60 Rev. 0 WOLF CREEK The BTRS is put into operation as follows:
For dilution of the RCS (storage):
- a. Cool down the chiller loop to about 40 F. This is not a requirement, but it will provide a faster cooldown transient of the BTRS.
- b. Put the master switch in the dilute position.
- c. Control the rate of dilution by positioning 3-way Valve HCV-387. The flow rate through the BTRS is dictated by the desired dilution rate of the RCS.
For boration of the RCS (releases):
- a. Put the master switch in the boration position.
- b. Control the rate of boration by positioning 3-way valve HCV-387. The flow rate through the BTRS is dictated by
the desired boration rate of the RCS. The BTRS is shut down by placing the master switch in the off position. Several resin beds in the BTRS can be used as deborating demineralizers, which toward the end of the core life are used to dilute the RCS down to very low
boron concentrations. To make such beds effective, the effluent concentration
from the beds must be kept very low, close to zero ppm. This can be achieved by using fresh resin. This should be coupled with the normal replacement cycle of the resin beds. HOT SHUTDOWN - If required for periods of maintenance, or following spurious reactor trips, the reactor can be held subcritical, but with the capability to return to full power within the period of time it takes to withdraw the control rods. During this hot shutdown, the average temperature is maintained at no-
load T-avg by initiating steam dumping to provide residual heat removal or at
later stages by running the reactor coolant pumps to maintain the system
temperature. Following shutdown, xenon buildup occurs and increases the degree of shutdown; i.e., initially, all control rods are inserted, and the core is maintained at a minimum of 1.3 percent k/k subcritical. The effect of the xenon buildup is to increase this value to a maximum of about 4 percent k/k at about 8 hours following shutdown from equilibrium full power conditions. 9.3-61 Rev. 0 WOLF CREEK If a return to power is anticipated, the reactor is taken critical by withdrawing the control banks. The xenon transient is followed by rod movement
and boration, as necessary, to maintain the control banks above the low insertion limit. If a prolonged shutdown is required, the reactor coolant is borated to the hot standby, xenon-free value and the control rods are inserted. 9.3.4.2.3.3 Reactor Cooldown Reactor cooldown is the operation which takes the reactor from hot standby to cold shutdown conditions (reactor is subcritical by at least 1.3 percent k/k and Tavg 200 F). Normal Cold Shutdown While initiating a cold shutdown, the RCS hydrogen concentration is lowered by reducing the volume control tank overpressure, by replacing the volume control
tank hydrogen atmosphere with nitrogen, and by continuous purging to the
gaseous radwaste system. Boration is one of the methods that can be used for reactivity control during a controlled plant cooldown. The boron concentration (in conjunction with Control Rods if available) is adjusted to maintain adequate shutdown margin as required due to the reactivity changes from the Reactor System cooldown. The RCS cold shutdown concentration is ensured by process control, i.e., knowledge of initial RCS boron concentrations and knowledge of amounts and concentrations of injected fluid ensures that the cold shutdown concentration is obtained. If desired the reactor coolant boron concentration may be increased to the cold shutdown value before cooldown and depressurization of the reactor plant is initiated. After the boration is completed and reactor coolant samples verify that the concentration is correct, the operator resets the Reactor Makeup Control System for leakage and system contraction makeup at the shutdown
reactor coolant boron concentration. Contraction of the coolant during cooldown of the RCS results in actuation of the pressurizer level control to maintain normal pressurizer water level. The
charging flow is increased, relative to letdown flow, and results in a
decreasing volume control tank level. The volume control tank level controller
automatically initiates makeup to maintain the inventory. Depressurizaton is performed by cooling the vapor space of the pressurizer with spray flow from an RCS loop with an operating reactor coolant pump. After the RHRS is placed in service and the reactor coolant pumps are shut down, further cooling of the pressurizer liquid is accomplished by charging through the auxiliary spray line from the outlet of the CVCS regenerative heat exchanger. Coincident with plant cooldown, a portion of the reactor coolant
flow is diverted from the RHRS to the CVCS for cleanup. 9.3-62 Rev. 16 WOLF CREEK If required, operation of the mixed-bed demineralizers and gas stripping are started in advance of a planned shutdown; demineralization of ionic radioactive
impurities and stripping fission gases reduce the reactor coolant activity level to permit personnel access for maintenance or refueling operations. Safety-related Cold Shutdown It is expected that the portions of the CVCS that are relied upon to perform reactor coolant system (RCS) purification, boration, letdown and depressurization operations, following an event that requires eventual cooldown
and long term cooling, will function in the normal manner. Additional safety-
related features have been designed and incorporated into the CVCS design to
ensure that certain functions relied upon to take the reactor from the hot standby mode to the cold shutdown mode will be available; in other words, the safety grade features have been provided to augment normal shutdown features should equipment availability become a concern. The following discussion describes the functioning of the CVCS using only safety-related equipment. Before cooldown and depressurization of the RCS is initiated, the RCS boron concentration is increased to the cold shutdown value.
Borated water from the RWST is delivered to the RCS through the Emergency Core
Cooling System (ECCS) cold leg injection lines via the boron injection tank (BIT) path and to the RCS through the reactor coolant pump (RCP) seals via seal injection lines. Charging flow is provided by the standby centrifugal charging pumps. Should RCS inventory letdown be required, this function can be
accomplished by releasing RCS fluid to the pressurizer relief tank (PRT) via
the excess letdown heat exchanger. Following the initial RCS boration/letdown operation, the RCS is depressurized by venting the pressurizer to the PRT through the pressurizer power-operated relief valves (PORVs). RCS pressure control will be maintained by using the
CVCS centrifugal charging pumps to provide RCS inventory control/makeup in
conjunction with the use of the PORVs. In the event normal charging and letdown paths are not available, RCS boration and inventory control functions will be maintained by utilizing redundant safety grade paths with the necessary throttling capability. Section 7.4 provides a systems integrated discussion on safe shutdown/cold shutdown. 9.3-63 Rev. 14 WOLF CREEK 9.3.4.2.3.4 Emergency Boration If emergency boration is required to achieve and maintain a safe shutdown, as described in Section 7.4, then the redundant centrifugal charging pumps can take suction from either the RWST or from the BAT via the boric acid transfer
pumps and discharge either through the RCP seals, the charging line, or the
boron injection tank path. The preferred mode, if offsite power is available, is the normal shutdown operation described in Section 9.3.4.2.3.3. This mode utilizes the boric acid
transfer system and normal seal injection and charging. If offsite power is
unavailable, then the RWST and the boron injection tank path are used to borate
the core. Further, if the RWST is rendered unavailable, then the boric acid transfer pumps can be loaded on the diesel and appropriate valves opened to provide suction to the centrifugal charging pumps. Sufficient flow can be delivered, either through the BIT injection path or through the reactor coolant pump seals, to borate the reactor coolant system to a cold shutdown concentration. In either case, the flow is throttled to permit orderly matching of the letdown and RCS shrinkage to the charging flow. The
centrifugal charging pumps are protected with open miniflow recirculation lines
during low flow operations. The emergency letdown path to the PRT can also be
throttled. 9.3.4.2.3.5 Emergency Core Cooling The charging portion of the CVCS plays an integral part in the emergency core cooling requirements for accidents involving small breaks or inadvertent valve lifting in the main steam or feedwater systems. The centrifugal charging pumps deliver borated water at the prevailing RCS pressure to the cold legs of the RCS. During the injection mode, the
centrifugal charging pumps take suction from the refueling water storage tank. The delivery of the boric acid provides negative reactivity to counteract the positive reactivity caused by the system cooldown. The safety injection function of the CVCS is automatically actuated by a safety injection signal (SIS). For a RCS equivalent pipe break opening of 3/8-inch (liquid service) diameter or less, the charging system can maintain the pressurizer level at the normal operating level and pressure. Therefore, the emergency core cooling system would not be automatically actuated, and is not
required. Details of the response by the CVCS are presented in Section 6.3. 9.3-64 Rev. 19 WOLF CREEK 9.3.4.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.3.4.1.1. SAFETY EVALUATION ONE - The safety-related portions of the CVCS 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 safety-related portions of the CVCS 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 and 3.6 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 safety-related portions of the CVCS are completely redundant and, as indicated by Table 9.3-10, 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 CVCS is initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with
Section 9.3.4.4. Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for the CVCS. 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 9.3-9 shows that the components meet the design and fabrication codes given in Section 3.2. Except for the control functions for the BAT system, all the power supplies and control functions
necessary for safe function of the CVCS are Class IE, as described in Chapters
7.0 and 8.0. The controls for the BAT system are only required when the RWST
is rendered inoperable, and the design of the control system is adequate for this situation. SAFETY EVALUATION SIX - Section 9.3.4.2.1 describes provisions made to identify and isolate leakage or malfunction and to isolate the nonsafety-related
portions of the system. 9.3-65 Rev. 19 WOLF CREEK 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 - Any time that the plant is critical at power, the quantity of boric acid retained and ready for injection is always equal to or greater than that quantity required for normal cold shutdown, assuming that the
control assembly of greatest worth is in its fully withdrawn position. This
quantity always exceeds the quantity of boric acid required to bring the reactor to hot shutdown and to compensate for subsequent xenon decay. An adequate quantity of boric acid is available in either the refueling water
storage tank or the boric acid tanks to achieve cold shutdown. When the reactor is subcritical (i.e., during cold shutdown, hot shutdown, hot standby) LCV112A is placed in the VCT position and Volume Control Tank level provides indication of an inadvertent boron dilution Transient. Upon the detection of a Hi level during any of the aforementioned modes of operation, an
alarm is sounded to alert the operator, and valve movement to terminate the
dilution and start boration is manually initiated. These corrective actions prevent the core from becoming critical. (See also Section 15.4.6 for discussion of boron dilution accident.) As a design capability, the rate of boration, with a single boric acid transfer pump operating, is sufficient to take the reactor from full power operation to one percent shutdown in the hot condition, with no rods inserted, in less than 90 minutes. In less than 90 additional minutes, enough boric acid can be injected to compensate for xenon decay, although xenon decay below the equilibrium operating level will not begin until approximately 25 hours after
shutdown. Additional boric acid is employed if it is desired to bring the reactor to cold shutdown conditions. Three separate and independent flow paths are available for reactor coolant boration; i.e., the charging line, the reactor coolant pump seal injection lines, and the boron injection tank path. A single active failure does not result in the inability to borate the RCS. If the normal charging line is not available, charging to the RCS may be continued via reactor coolant pump seal injection at the rate of approximately
5 gpm per pump. At the charging rate of 20 gpm (5 gpm per reactor coolant pump), approximately 5 hours are required to add enough nominal 4 weight percent boric acid solution to counteract xenon decay, although xenon decay below the full power equilibrium operating level will not begin until
approximately 25 hours after the reactor is shut down. 9.3-66 Rev. 13 WOLF CREEK As backup to the normal boric acid supply, the operator can align the refueling water storage tank outlet to the suction of the charging pumps. Should the normal charging and RCP seal injection paths not be available for boration or makeup, redundant safety related flow paths with necessary throttling capability are provided by the ECCS cold leg injection headers via
the BIT path using either the BIT inlet MOVs or the 1" bypass lines using
manual valves EMV151, EMV246, and EMV247. The CVCS is capable of borating the RCS to cold shutdown concentration at a rate that is compatible with meeting the objectives of the cold shutdown
design, described in Section 7.4. (Letdown to accommodate boration is also discussed in Section 7.4.) The CVCS is also capable of providing sufficient borated water from the refueling water storage tank to make up for primary shrinkage due to cooling or RCS inventory discharged during cooldown. Since inoperability of a single component does not impair the ability to meet boron injection requirements, plant operating procedures allow the components to be temporarily out of service for repairs. However, with an inoperable component, the ability to tolerate additional component failure is limited. Therefore, operating procedures require immediate action to effect repairs of
an inoperable component, restrict permissible repair time, and require demonstration of the operability of the redundant component. SAFETY EVALUATION NINE - As discussed in Section 9.3.4.2, the CVCS is capable of making up for a small RCS leakoff up to approximately 120 gpm, using one centrifugal charging pump, and still maintaining seal injection flow to the
reactor coolant pumps. This also allows for a minimum RCS cooldown contraction. This is accomplished with the letdown isolated. SAFETY EVALUATION TEN - Section 6.3 provides the safety evaluation for the emergency core cooling operation of the CVCS. Portions of the CVCS are relied upon for safe shutdown and accident mitigation. The failure mode and effects analysis summarized in Table 9.3-10 demonstrates that single active component failures do not compromise the CVCS safe shutdown
functions of boration and makeup. This analysis also shows that single
failures occurring during CVCS operation do not compromise the ability to prevent or mitigate accidents. The capabilities are accomplished by a combination of suitable redundancy, instrumentation for indication and/or alarm
of abnormal conditions, and relief valves to protect piping and components
against malfunctions. 9.3-67 Rev. 14 WOLF CREEK Portions of the CVCS are also relied upon to provide safety-related boration and makeup. The capability of the CVCS to perform in conjunction with other
systems of the cold shutdown design is presented in the Table 5.4A-3. The CVCS shares components with the ECCS and containment isolation functions. These safeguard functions of the CVCS are addressed in Chapter 6.0. SAFETY EVALUATION ELEVEN - Section 7.4 demonstrates how cold shutdown, including the function of boration, is achieved with the use of only safety-related equipment. 9.3.4.4 Tests and Inspections As part of plant operation, periodic tests, surveillance inspections, and instrument calibrations are made to monitor equipment condition and
performance. Most components are in use regularly; therefore, assurance of the availability and performance of the systems and equipment is provided by control room and/or local indication. Further information concerning preoperational and startup testing is described in Chapter 14.0. Technical Specifications have been established concerning calibration, checking, and sampling of the CVCS. 9.3.4.5 Instrumentation Application Process control instrumentation is provided to acquire data concerning key parameters about the CVCS. The location of the instrumentation is shown on Figure 9.3-8. The instrumentation furnishes input signals for monitoring and/or alarming purposes. Indications and/or alarms are provided for the following parameters:
- a. Temperature
- b. Pressure
- c. Flow
- e. Water level The instrumentation also supplies input signals for control purposes. Some specific control functions are:
- a. Letdown flow is diverted to the volume control tank upon high temperature indication upstream of the mixed bed demineralizers. 9.3-68 Rev. 14 WOLF CREEK
- b. Pressure upstream of the letdown heat exchanger is controlled to prevent flashing of the letdown liquid
downstream of the letdown orifices.
- c. Charging flow rate is controlled during charging pump operation.
- d. Water level is controlled in the volume control tank.
- e. Temperature of the boric acid solution in the batching tank is maintained.
- f. Reactor makeup is controlled.
- g. Temperature of letdown flow to the boron thermal regeneration system is controlled.
- h. Temperature of the chilled water flow to the letdown chiller heat exchanger is controlled.
- i. Temperature of letdown flow return from the boron thermal regeneration demineralizers is controlled.
- j. Letdown flow rate to PRT is controlled.
- k. Letdown temperature to PRT is controlled.
- 1. Seal injection flow is controlled
- m. BIT flow is controlled n. Normal charging pump recirculation isolation valve is automatically controlled by the pump discharge flow switch.
9.3.5 SERVICE
GAS SYSTEM The service gas system (SGS) provides nitrogen, hydrogen, carbon dioxide, oxygen, and laboratory gases to plant systems, as required. Bulk storage of service gases is described in Section 2.2. The compressed air system is described in Section 9.3.1, and the diesel generator starting air system is described in Section 9.5.6. 9.3.5.1 Design Bases 9.3.5.1.1 Safety Design Bases The nitrogen, hydrogen, carbon dioxide, and oxygen systems serve no safety function, and there are no system safety design bases. Since the service gas storage vessels are maintained at high pressure, they are a potential hazard, and the location and design of the tanks and adjacent structures and/or barriers are consistent with the following safety design bases. 9.3-69 Rev. 10 WOLF CREEK SAFETY DESIGN BASIS ONE - Rupture of a compressed gas storage vessel or piping will not cause unacceptable impairment of a safety-related system, structure, or component from blast forces, missile impacts, or pipe whipping. SAFETY DESIGN BASIS TWO - Rupture of a compressed gas storage vessel or piping will not cause a deficiency of oxygen for breathing purposes in the control rooms.SAFETY DESIGN BASIS THREE - Rupture of a hydrogen or oxygen storage vessel or piping will not cause the failure of safety-related components, systems, or
structures as a result of delayed ignition or explosion. 9.3.5.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The SGS transports low pressure nitrogen from the site storage facility for use as a cover gas, purge gas for corrosion
prevention, and carrier gas. The SGS also transports high pressure nitrogen to
be used as a source of potential energy. POWER GENERATION DESIGN BASIS TWO - The SGS transports hydrogen from the site storage facility and stores hydrogen, in limited quantities within the power
block, for use, in recombination with oxygen, as a cover gas, as a cooling
medium, and as a stripping agent. POWER GENERATION DESIGN BASIS THREE - The SGS transports carbon dioxide from the site storage facility to be used as a purge gas. POWER GENERATION DESIGN BASIS FOUR - The SGS transports oxygen from the site storage facility to be used in recombination with hydrogen. 9.3.5.2 System Description 9.3.5.2.1 General Description The SGS, which is shown in Figure 9.3-9, consists of a network of piping conveying nitrogen, hydrogen, carbon dioxide, and oxygen from site storage
facilities to the standard power block for various uses. For each gas entering
the power block, a master shutoff valve and a pressure regulator are provided. In addition, for hydrogen lines entering the power block, an excess flow check valve was provided. Section 2.2 describes the gas storage facilities. 9.3-70 Rev. 0 WOLF CREEK High and low pressure nitrogen enter the power block through separate headers. A separate low volume source of hydrogen or nitrogen is provided to supply a cover gas for the reactor coolant drain tank which is located inside the containment. This source is located in the hot machine shop which is outside any safety-related building. Each of the other gases enters the power block at a single location. Figure 9.3-10 shows where each of the gases enters the power block. The service gas main headers are all 2-inch lines, with the exception of the high pressure nitrogen supply line which is 1 inch to reduce the potential for failure. From the headers, 1-inch service gas lines are routed to their associated service location. Table 9.3-11 lists the various components supplied by the SGS. In addition to the major gas distribution headers, gas bottles are located within the plant in nonsafety-related areas to provide small quantities of specialty gases for laboratory analysis or localized testing. An exception to this are small cylinders of oxygen and hydrogen/nitrogen mix located on the
hydrogen analyzer skid on elevation 2047 near the equipment hatch and in the containment purge supply air handling room. Their location, which is shown in Figure 9.3-10, is in accordance with safety codes, including the Wisconsin Administrative Code. A list of laboratory gases is provided in Table 9.3-12. 9.3.5.2.2 Component Description The gas storage facilities outside the power block are described in Section 2.2.Piping and valves are designed and fabricated to meet the requirements of the Power Piping Code, ANSI B31.1. Packless valves are used to minimize gaseous leakage. All headers are carbon steel, except the oxygen piping which is constructed of welded stainless steel. Storage facilities for laboratory gases are provided in the nonsafety-related building of the power block as indicated in Figure 9.3-10. 9.3.5.2.3 System Operation During normal operation, service gas received from the site storage facility is maintained at the required pressure through pressure regulators. Service gas flow is controlled by those systems being served. 9.3-71 Rev. 15 WOLF CREEK During plant startup and shutdown, service gas for filling and purging is manually controlled. 9.3.5.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design basis. SAFETY EVALUATION ONE - Section 2.2 provides an evaluation which demonstrates that any hazards due to the gas storage facilities will not adversely affect safety-related structures, systems, or components due to blast force, missile
impact, or pipe whipping. SAFETY EVALUATION TWO - The routing of service gases within the power block will not allow escaping gas to enter the control building air intakes. The effects of a high pressure nitrogen pipe rupture are discussed in Section 3.6.
Section 2.2 provides an evaluation of gas storage facilities outside the power block.SAFETY EVALUATION THREE - All lines associated with the distribution of hydrogen within the safety-related structure are less than l inch in diameter and carry moderate energy fluid, hence no break needs to be assumed per NRC Branch Technical Position MEB 3-1. In addition, if a rupture were to occur
there is insufficient volume associated with the bottle storage for the reactor coolant drain tank to create an explosive mixture. For the hydrogen bulk storage supply, an excess flow check valve is provided to keep the maximum
blowdown below an explosive mixture. The maximum rate of blowdown for hydrogen from a ruptured pipe is eight scfm. The minimum ventilation rate in areas where hydrogen gas lines are routed is 300 scfm, which results in a maximum hydrogen gas concentration of less than 3
volume percent. Thus, an explosive mixture cannot form. Oxygen is not routed
within safety-related structures. 9.3.5.4 Tests and Inspections Preoperational testing is performed, as outlined in Chapter 14.0. The system is inspected to verify that the applicable plans, drawings, and specification are
met. Applicable code-required testing is performed. The service gas system operates continuously throughout the life of the plant, thus demonstrating the structural and leaktight integrity of all the components. 9.3-72 Rev. 0 WOLF CREEK 9.3.5.5 Instrumentation Application Pressure indication is provided in most systems served by the service gas system.9.3.6 BORON RECYCLE SYSTEM The boron recycle system (BRS) receives reactor coolant effluent for the purpose of storage until it can either be reused or disposed of by processing it through the Liquid Radwaste System. 9.3.6.1 Design Bases 9.3.6.1.1 Safety Design Basis The BRS serves no safety function.
9.3.6.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The BRS collects and processes plant effluents which can be potentially reused. For the most part, this effluent is the deaerated, tritiated, borated, and radioactive water from the letdown and process drains. The BRS is designed to collect the excess reactor coolant that results from certain plant operations, as described in Section 9.3.6.2.1. POWER GENERATION DESIGN BASIS TWO - The BRS is designed to process the total volume of water collected during a core cycle as well as shortterm surges. The design surge is that produced by a cold shutdown and subsequent startup during
the latter part of a core cycle or by a refueling shutdown and startup. POWER GENERATION DESIGN BASIS THREE - Deleted. POWER GENERATION DESIGN BASIS FOUR - The BRS uses design and fabrication codes consistent with quality group D (augmented), as assigned by Regulatory Guide 1.143 for radioactive waste management systems. 9.3-73 Rev. 14 WOLF CREEK 9.3.6.2 System Description 9.3.6.2.1 General Description The BRS is shown in Figure 9.3-11. The BRS is designed to collect, via the letdown line in the chemical and volume control system (CVCS), the excess reactor coolant that results from the following plant operations during one core cycle:
- a. Dilution for core burnup from approximately 1,600 ppm boron at the beginning of an eighteen (18) month core cycle to approximately 10 ppm near the end of the core cycle.
- b. Hot shutdowns and startups. Four hot shutdowns are assumed to take place during an annual core cycle.
- c. Cold shutdowns and startups. Three cold shutdowns are assumed to take place during an annual core cycle.
- d. Refueling shutdown and startup.
The BRS also collects water from the following sources:
- a. Reactor coolant drain tank (liquid waste processing system) - collects leakoff type drains from equipment
inside the containment.
- b. Volume control tank and charging pump suction pressure reliefs (CVCS), safety injection pump pressure reliefs, and RHR pump pressure reliefs.
- c. Boric acid blending tee (CVCS) - provides for the storage of boric acid if a boric acid tank must be emptied for maintenance. The boric acid solution is
stored in a recycle holdup tank after first being
diluted with reactor makeup water by the blending tee, if necessary, to ensure against precipitation of the boric acid in the unheated recycle holdup tank.
- d. Accumulators (safety injection system) - collect effluent resulting from leak testing of accumulator check valves.
- e. Deleted. 9.3-74 Rev. 14 WOLF CREEK
- f. Fuel transfer canal (via the fuel storage pool cooling and cleanup system) - provides a means of storing the fuel transfer canal water in the event maintenance is required on the transfer equipment.
- g. Safety injection system - accepts flush water when the boron injection tank valves are being tested or the
system flushed.
- h. Deleted.
When water is directed to the BRS, the flow first passes through the recycle evaporator feed demineralizers and filters and then into the recycle holdup tanks. The recycle evaporator feed pumps can then be used to transfer liquid from the recycle holdup tanks to the Liquid Radwaste System or used to recirculate water through the recycle evaporator feed demineralizers for additional cleanup. Water can also be transferred to the fuel transfer canal or to the suction of the charging pumps (CVCS) for refilling the RCS. Water can also be transferred from one recycle holdup tank to the other, if desired. 9.3-75 Rev. 14 WOLF CREEK 9.3.6.2.2 Component Descriptions Codes and standards applicable to the BRS are listed in Table 3.2-1 and 9.3-13. The BRS is designed and constructed in accordance with quality group D (augmented), as assigned by Regulatory Guide 1.143 for radioactive waste
management systems. The BRS is housed within a seismically designed building, as described in Section 3.8.6. The performance parameters to which the
individual components of the BRS are designed are listed in Table 9.3-13. RECYCLE EVAPORATOR FEED PUMPS - Two centrifugal, canned pumps supply feed to the Liquid Radwaste System from the recycle holdup tanks. The pumps can also be used to recirculate water from the recycle holdup tanks through the recycle evaporator feed demineralizers for cleanup, if desired. An auxiliary discharge connection is provided to return water to the fuel transfer canal from the recycle holdup tanks, if those tanks were used for storage of fuel transfer canal water during refueling equipment maintenance. Another auxiliary discharge connection is provided to supply water to the suction of the charging pumps (CVCS) for refilling the RCS after loop or system draindown. RECYCLE HOLDUP TANKS - Two recycle holdup tanks provide storage for radioactive fluid which is discharged from the RCS during startup, shutdown, load changes, and boron dilution. The sizing criteria is based on the design surge that is
produced by a cold shutdown and subsequent startup during the latter part of core cycle or by refueling shutdown and startup. Each tank has a diaphragm which prevents air from dissolving in the water and prevents the hydrogen and fission gases in the water from mixing with the air.
The volume in the tank above the diaphragm is continuously ventilated with building supply air, and any gas which accumulates below the diaphragm is intermittently vented to the gaseous waste processing system via the recycle
holdup tank vent eductor. In addition to the collection of effluents, the recycle holdup tanks provide the following functions:
- a. Serve as a head tank for the recycle evaporator feed pumps. 9.3-76 Rev. 14 WOLF CREEK
- b. Provide holdup for a RCS drain to the centerline of the reactor vessel nozzles, including the pressurizer and
steam generators.
- c. Provide storage for fuel transfer canal water during refueling equipment maintenance.
- d. Collect discharges from the various relief valves.
RECYCLE EVAPORATOR FEED DEMINERALIZERS - Two flushable, mixed bed demineralizers remove fission products from the fluid directed to the recycle
holdup tanks. The demineralizers also provide a means of cleaning the recycle
holdup tank contents via recirculation. RECYCLE EVAPORATOR FEED FILTER - This filter collects resin fines and particulates from the fluid entering the recycle holdup tanks. RECYCLE HOLDUP TANK VENT EDUCTOR - The eductor is designed to pull gases from under the diaphragm in a recycle holdup tank and deliver them to the gaseous waste processing system. Nitrogen, provided by the standby waste gas
compressor, provides the motive force. 9.3.6.2.3 System Operation The BRS is manually operated, with the exception of a few automatic protection functions. These automatic functions protect the recycle evaporator feed demineralizers from high inlet temperature and high differential pressure, prevent high vacuum from being drawn on the recycle holdup tank diaphragm, protect the recycle evaporator feed pumps from low net positive suction head. The BRS has sufficient instrumentation readouts and alarms to provide the operator information to assure proper system operation. RECYCLE HOLDUP TANK VENTING - Because hydrogen is dissolved in the reactor coolant at a concentration of 25 - 50cc hydrogen per kilogram of reactor coolant, a portion of the hydrogen along with fission gases will come out of solution in the recycle holdup tank under the diaphragm. The hydrogen and
fission gases are vented to the gaseous radwaste system, as required. The total integrated flow from the letdown line and the reactor coolant drain tank to the recycle holdup tanks is monitored. An alarm indicates when a sufficient amount of water has passed to the recycle holdup tanks to require venting of the accumulated gases. When venting of either recycle holdup tank is required, the following steps are observed: a. The standby gas compressor is started up, and the vent from the holdup tank is opened. The vent flow is throttled to approximately 1 scfm. At this time, a sample of the vent gases can be taken to check the composition. 9.3-77 Rev. 14 WOLF CREEK b. When the gases have been vented from the recycle holdup tank, the pressure in the vent line decreases, which automatically trips the recycle holdup tank vent isolation valve closed. The recycle holdup tank vent isolation valve may also be closed manually. c. After the vent isolation valve closes, the manual vent valve is closed, and the gas compressor is shut down. 9.3.6.3 Safety Evaluation The BRS has no safety-related functions. 9.3.6.4 Tests and Inspections The BRS is in intermittent use throughout normal reactor operation. Periodic visual inspection and preventive maintenance are conducted using normal
industry practice. Refer to Chapter 14.0 for further information concerning preoperational and startup testing. 9.3.6.5 Instrumentation Application The instrumentation available for the BRS is discussed below. Alarms are provided as noted. There is also a common alarm on the main control board which indicates any alarms on the BRS panel. TEMPERATURE - Instrumentation is provided to measure the temperature of the
inlet flow to the recycle evaporator feed demineralizers and to control a three-way bypass valve. If the inlet temperature becomes too high, the instrumentation aligns the valve to bypass the demineralizers. Local
temperature indication and a high temperature alarm on the BRS panel are provided by this instrumentation. PRESSURE - Instrumentation is provided to measure the pressure differential across the recycle evaporator feed demineralizers and to control the same three-way valve as discussed above (but independently of the temperature
control). If the pressure drop through the demineralizers is too high, this instrumentation aligns the valve to divert flow directly to the recycle evaporator feed filter. Local pressure differential indication and a high alarm on the BRS panel are provided by this instrumentation. 9.3-78 Rev. 14 WOLF CREEK TABLE 9.3-1 COMPONENT DESCRIPTION COMPRESSED AIR SYSTEM Component Air Compressors Type Nonlubricated, rotary Capacity, scfm 932 Quantity 3 Motor horsepower 250 Operating pressure, psig 122 Design pressure, psig 150 Type (Sullair) Lubricated, rotary Capacity, scfm 1500 Quantity 1 Motor Horsepower 350 Operating Pressure, psig 100-110 Design Pressure, psig 150 Air Receivers Type Vertical Quantity 3 Capacity, ft3, each 52 Operating pressure, psig 122 Design pressure, psig 125 Supply gas pressure, psig 122 Stored energy, ft-lb, each 3.67 x 10 6 Design code ASME, Section VIII Prefilters and Afterfilters Type Cartridge, disposable Quantity 2 Capacity, scfm, each 1200 and 1070 Operating pressure, psig 122
Design pressure, psig 150 Design code ASME, Section VIII Air Dryers Type Heatless, desiccant Quantity 2 Capacity, scfm, each 1200 Operating pressure, psig 122
Design pressure, psig 150 Design code ASME, Section VIII Type Refrigerated Quantity 1 Capacity, scfm 1500 Operating Pressure, psig 100 Design Pressure, psig 150 Rev. 16 WOLF CREEK TABLE 9.3-1 (Sheet 2) Component Aux. Feedwater Control Valve and Main Steam Atmospheric Relief Valve
Accumulator Type Horizontal, carbon steel Quantity 4 Capacity, ft3, each 25 Design pressure, psig 850 Operating Range, psig 650-750 Nominal Supply pressure, psig 750 Stored energy, ft-lb, each 9.63 x 10 6 Code requirements ASME III, Class 3 Valve operating time provided, 8 (One valve cycle hrs/valve every 20 min./aux. F.W. control valve)
(One valve cycle every 10 min./M.S. atm.
relief valve) Main Feedwater Control Valve Accumulator Type Vertical, carbon steel Quantity l Capacity, ft3, each 30 Design pressure, psig 825 Operating Range, psig 650-750 NominalSupply pressure, psig 750 Stored energy, ft-lb, each 1.55 x 10 7 Code requirements ASME III, Class 3 Valve operating time provided, 4 (One valve cycle hrs/valve every 30 min.) Rev. 11 WOLF CREEK TABLE 9.3-2 SAFETY-RELATED PNEUMATICALLY OPERATED VALVES Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes AB-HV-05 Loop 2 Steam Supply to AFW Pump Turbine Open Open Admit steam to AFW pump turbine and secondary side pressure boundary isolation AB-HV-06 Loop 3 Steam Supply to AFW Pump Turbine Open Open Admit steam to AFW pump turbine and secondary side pressure boundary isolation AB-HV-12 Main Steam Iso. Bypass Valve Loop 4 Closed Closed Secondary side pressure boundary isolation and steam line warmup AB-HV-15 Main Steam Iso. Bypass Valve Loop 1 Closed Closed Secondary side pressure boundary isolation and steam line warmup AB-HV-18 Main Steam Iso. Bypass Valve Loop 2 Closed Closed Secondary side pressure boundary isolation and steam line warmup AB-HV-21 Main Steam Iso. Bypass Valve Loop 3 Closed Closed Secondary side pressure boundary isolation and steam line warmup AB-HV-48 Loop 2 Steam Supply to AFW Turbine Closed Closed Secondary side pressure boundary Bypass isolation and steam line keep warm AB-HV-49 Loop 3 Steam Supply to AFW Turbine Closed Closed Secondary side pressure boundary Bypass isolation and steam line keep warm AB-LV-07 Main Steam Line Drain Valve Loop 3 Closed Closed Secondary side pressure boundary isolation and condensate drain AB-LV-08 Main Steam Line Drain Valve Loop 2 Closed Closed Secondary side pressure boundary isolation and condensate drain AB-LV-09 Main Steam Line Drain Valve Loop 1 Closed Closed Secondary side pressure boundary isolation and condensate drain Rev. 24 WOLF CREEK TABLE 9.3-2 (Sheet 2) Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes AB-LV-10 Main Steam Line Drain Valve Loop 4 Closed Closed Secondary side pressure boundary isolation and condensate drain AB-PV-01 Steam Gen. A Atm. Relief Valve Closed Closed Secondary side pressure boundary 2 isolation, secondary side heat removal, and pressure relief AB-PV-02 Steam Gen. B Atm. Relief Valve Closed Closed Secondary side pressure boundary 2 isolation, secondary side heat removal, and pressure relief AB-PV-03 Steam Gen. C Atm. Relief Valve Closed Closed Secondary side pressure boundary 2 isolation, secondary side heat removal, and pressure relief AB-PV-04 Steam Gen. D Atm. Relief Valve Closed Closed Secondary side pressure boundary 2 isolation, secondary side heat removal, and pressure relief AE-FV-43 Steam Gen. A Chemical Control Closed Closed Secondary side pressure boundary isolation and chemistry control AE-FV-44 Steam Gen. B Chemical Control Closed Closed Secondary side pressure boundary isolation and chemistry control AE-FV-45 Steam Gen. C Chemical Control Closed Closed Secondary side pressure boundary isolation and chemistry control AE-FV-46 Steam Gen. D Chemical Control Closed Closed Secondary side pressure boundary isolation and chemistry control AE-FCV-510 Feedwater Control Valve Loop 1 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-520 Feedwater Control Valve Loop 2 Closed Closed Backup valve for secondary side pressure boundary isolation Rev. 24 WOLF CREEK TABLE 9.3-2 (Sheet 3) Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes AE-FCV-530 Feedwater Control Valve Loop 3 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-540 Feedwater Control Valve Loop 4 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-550 Feedwater Control Bypass Valve Loop 1 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-560 Feedwater Control Bypass Valve Loop 2 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-570 Feedwater Control Bypass Valve Loop 3 Closed Closed Backup valve for secondary side pressure boundary isolation AE-FCV-580 Feedwater Control Bypass Valve Loop 4 Closed Closed Backup valve for secondary side pressure boundary isolation AL-HV-06 Turbine AFP Disch. to Steam Gen. D Open Open Control AFW flow to steam generators; 1 isolation of AFW to broken loop AL-HV-08 Turbine AFP Disch. to Steam Gen. A Open Open Control AFW flow to steam generators; 1 isolation of AFW to broken loop AL-HV-10 Turbine AFP Disch. to Steam Gen. B Open Open Control AFW flow to steam generators; 1 isolation of AFW to broken loop AL-HV-12 Turbine AFP Disch. to Steam Gen. C Open Open Control AFW flow to steam generators; 1 isolation of AFW to broken loop BB-HV-8026 Ctmt. Iso. Valve - Nitrogen to PRT Closed Closed Containment isolation BB-HV-8027 Ctmt. Iso. Valve - Nitrogen to PRT Closed Closed Containment isolation
BG-HV-8152 Ctmt. Iso. Valve - Letdown Line Closed Closed Containment isolation BG-HV-8160 Ctmt. Iso. Valve - Letdown Line Closed Closed Containment isolation BL-HV-8047 Ctmt. Iso. Valve - Reactor Makeup Water Closed Closed Containment isolation
BM-HV-01 Steam Gen. A to SGBD Flash Tank Valve Closed Closed Secondary side pressure boundary isolation BM-HV-02 Steam Gen. B to SGBD Flash Tank Valve Closed Closed Secondary side pressure boundary isolation BM-HV-03 Steam Gen. C to SGBD Flash Tank Valve Closed Closed Secondary side pressure boundary isolation Rev. 0 WOLF CREEK TABLE 9.3-2 (Sheet 4) Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes BM-HV-04 Steam Gen. D to SGBD Flash Tank Valve Closed Closed Secondary side pressure boundary isolation BN-HCV-8800A RWST Iso. Valve to SFP Cleanup Closed Closed System pressure boundary isolation BN-HCV-8800B RWST Iso. Valve to SFP Cleanup Closed Closed System pressure boundary isolation EF-HV-43 ESW to Air Compressor Iso. Valve Closed Closed System pressure boundary isolation EF-HV-44 ESW to Air Compressor Iso. Valve Closed Closed System pressure boundary isolation EG-HV-69A CCW Supply Waste Header Iso. Valve Closed Closed System pressure boundary isolation EG-HV-69B CCW Return Waste Header Iso. Valve Closed Closed System pressure boundary isolation EG-HV-70A CCW Supply Waste Header Iso. Valve Closed Closed System pressure boundary isolation
EGLV-0001 Demin Water to CCW Surge Tank A Closed Closed Level control and CCW pressure boundary EGLV-0002 Demin Water to CCW Surge Tank B Closed Closed Level control and CCW pressure boundary EG-HV-70B CCW Return Waste Header Iso. Valve Closed Closed System pressure boundary isolation EG-TV-29 CCW Heat Exchanger A Bypass Iso. Valve Closed Closed Maintain CCW heat exchanger discharge temperature and isolate bypass flow EGRV-0009 CCW Surge Tank A Vent Closed Closed Close on high radiation and CCW pressure boundary EGRV-0010 CCW Surge Tank B Vent Closed Closed Close on high radiation and CCW pressure boundary EG-TV-30 CCW Heat Exchanger B Bypass Iso. Valve Closed Closed Maintain CCW heat exchanger discharge temperature and isolate bypass flow EJ-FCV-0618 RHR HX A Bypass Closed Closed Isolate bypass flow EJ-FCV-0619 RHR HX B Bypass Closed Closed Isolate bypass flow EJ-HCV-0606 RHR HX A Discharge Open Open Remain open to ensure flow path EJ-HCV-0607 RHR HX B Discharge Open Open Remain open to ensure flow path EJ-HV-8825 Test Line Iso. Valve - Hot Leg Injection Closed Closed Containment Isolation EJ-HV-8890A Test Line Iso. Valve - Cold Leg Injection Closed Closed Containment Isolation EJ-HV-8890B Test Line Iso. Valve - Cold Leg Injection Closed Closed Containment Isolation
EM-HV-8823 Test Line Iso. Valve - SI to RCS Cold Leg Closed Closed Containment Isolation EM-HV-8824 Test Line Iso. Valve - Hot Legs 1 and 4 Closed Closed Containment Isolation EM-HV-8843 Test Line Iso. Valve - BIT Line Closed Closed Containment Isolation Rev. 5 WOLF CREEK TABLE 9.3-2 (Sheet 5) Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes EM-HV-8871 Ctmt. Iso. Valve - SI Test Line Closed Closed Containment isolation EM-HV-8881 Test Line Iso. Valve - Hot Legs 2 and 3 Closed Closed Containment isolation
EM-HV-8882 BIT Test Line Iso. Valve Closed Closed System pressure boundary isolation EM-HV-8888 Ctmt. Iso. Valve - Accumulator Fill Closed Closed Containment isolation
EM-HV-8889A HL 1 SI Test Line Closed Closed System pressure boundary isolation
EM-HV-8889B HL 2 SI Test Line Closed Closed System pressure boundary isolation EM-HV-8889C HL 3 SI Test Line Closed Closed System pressure boundary isolation EM-HV-8889D HL 4 SI Test Line Closed Closed System pressure boundary isolation EM-HV-8964 Ctmt. Iso. Valve - SI Test Line Closed Closed Containment isolation EP-HV-8875A N2 Supply Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8875B N2 Supply Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8875C N2 Supply Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8875D N2 Supply Iso. Valve Closed Closed System pressure boundary isolation
EP-HV-8877A Acc. Tank A to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8877B Acc. Tank B to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8877C Acc. Tank C to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation
EP-HV-8877D Acc. Tank D to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8878A Acc. Tank Fill from SI Pump Closed Closed System pressure boundary isolation EP-HV-8878B Acc. Tank Fill from SI Pump Closed Closed System pressure boundary isolation
EP-HV-8878C Acc. Tank Fill from SI Pump Closed Closed System pressure boundary isolation EP-HV-8878D Acc. Tank Fill from SI Pump Closed Closed System pressure boundary isolation Rev. 10 WOLF CREEK TABLE 9.3-2 (Sheet 6) Failure Mode Valve Safe on Loss of Number Description Position Air Supply Safety Function Notes EP-HV-8879A Acc. Tank A to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8879B Acc. Tank B to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8879C Acc. Tank C to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8879D Acc. Tank D to SIS Test Line Iso. Valve Closed Closed System pressure boundary isolation EP-HV-8880 Ctmt. Iso. Valve - N2 Supply to Accum. Closed Closed Containment isolation FC-FV-310 AFP Steam Trap Isolation Valve Closed Closed Condensate removal FCLV-0010 Aux Feedwater pump turbine bypass trap Closed Closed Level control and Aux Feedwater to Cond Valve piping pressure boundary HB-HV-7126 Ctmt. Iso. Valve - RCDT to Waste Gas Closed Closed Containment isolation Comp. HB-HV-7136 Ctmt. Iso. Valve - RCDT to Recy. Closed Closed Containment isolation Holdup Tank HB-HV-7150 Ctmt. Iso. Valve - RCDT to Recy. Closed Closed Containment isolation Holdup Tank HB-HV-7176 Ctmt. Iso. Valve - RCDT to Waste Closed Closed Containment isolation Gas Comp. KA-FV-29 Ctmt. Iso. Valve - Inst. Air Line Closed Closed Containment isolation LF-FV-96 Ctmt. Iso. Valve - Sump to Floor Drain Closed Closed Containment isolation Tank NOTES: (1) Provided with backup compressed gas supply to open for safety functions.
(2) Provided with backup compressed gas supply to modulate valve as required during cooldown from hot shutdown condition to cold shutdown.
Rev. 24 WOLF CREEK TABLE 9.3-3 PRIMARY SAMPLING SYSTEM SAMPLE POINT DESIGN DATA Sample Conditions (Operating) SamplePoint Sample Name Pressure Temp No. Primary Sampling System psig** F** Typical Analysis* 1 Steam generator blowdown A 1091 557 Gross activity by liquid monitor 2 Steam generator blowdown B 1091 557 Gross activity by liquid monitor 3 Steam generator blowdown C 1091 557 Gross activity by liquid monitor 4 Steam generator blowdown D 1091 557 Gross activity by liquid monitor 5 RCS hot legs sample (loop 1 or 3) 2235 618 Gross activity, tritium, hydrogen, oxygen, lithium, radioiodine, pH, conductivity, boron, chloride/fluoride, silica, Ca, Mg, sulfate, ammonia, gross alpha, total suspended solid, dose equivalent iodine, fuel reliability indicator 6 Pressurizer liquid space 2235 653 Boron, oxygen 7 Pressurizer vapor space 2235 653 Hydrogen, oxygen, nitrogen, fission gases, helium 8 CVCS letdown upstream of demineralizer 300 115 Gross activity, activity by liquid monitor Note: other data same as RCS 9 CVCS letdown downstream of demineralizer 300 115 Gross activity, chloride, fluoride, ph, silica, decontamination factor 10 Reactor makeup water storage tank 100 80 Chloride/fluoride, total solid, total suspended solid, oxygen, pH, conductivity, boron, Sodium, tritium, silica, Al, Ca, Mg 11 Accumulator tanks A, B, C, and D 650 150 Boron, chloride/fluoride 12 Boric acid tank A and B Atmospheric 120 Boron, sulfate, silica, Al, Ca, Mg, gross activity, Chloride/fluoride, iron 13 Reactor coolant drain tank 134 140 Hydrogen, oxygen, nitrogen, helium, fission gases 14 *Residual heat removal heat exchanger A and B 540 140 Boron, chloride/fluoride, conductivity, lithium, gross activity, sulfate, tritium, Mg, pH, silica, Ca, total suspended solid 15 Boron thermal regeneration demin- eralizer effluent 180 115 Boron
- Sampling frequencies are specified within applicable chemistry procedures.
- The pressures and temperatures listed in this table represent nominal values at the time of plant licensing and should not be viewed as actual values observabrele in the field.
Rev. 17 WOLF CREEK TABLE 9.3-4 RADWASTE SAMPLING SYSTEM SAMPLE POINT DESIGN DATA Sample Conditions (Operating)Sample Point Sample Name Pressure Temp No Radwaste Sampling System psig** F** Typical Analysis* 1 Waste holdup tank 108 100 Gross activity, pH, silica, conductivity 2 Waste evaporator condensate tank 108 120 Gross activity, general surveillance 3 Waste evaporator distillate 65 120 Gross activity, general surveillance 4 Chemical drain tank 108 100 Conductivity, pH 5 Laundry and hot shower tank 108 100 Gross activity, general surveillance 6 Laundry water storage tank 108 100 Gross activity, general surveillance 7 Floor drain tanks A and B 108 100 Gross activity, silica, pH, conductivity, oil and grease 8 Waste monitor tank A 108 100 Gross activity, tritium, dissolved and entrained gas, total suspended solid, oil and grease 9 Waste monitor tank B 108 100 Gross activity, tritium, dissolved and entrained gas, total suspended solid, oil and grease 10 Steam generator blowdown surge tank 200 150 Gross activity, tritium, dissolved and entrained gas, total suspended solid, oil and grease 11 Recycle holdup tanks A and B 110 115 Gross activity, chloride, fluoride, boron, conductivity 12 Recycle evaporator condensate 65 210 Gross activity, general surveillance demineralizer 13 Fuel pool cleanup demineralizer inlet 134 140 Gross activity, chloride, fluoride, boron, pH, total suspended solid, silica, sulfate, tritium, gross alpha Rev. 17 WOLF CREEK TABLE 9.3-4 (sheet 2) RADWASTE SAMPLING SYSTEM SAMPLE POINT DESIGN DATA Sample Conditions (Operating) SamplePoint Sample Name Pressure Temp No Radwaste Sample System psig** F** Typical Analysis* 14 Fuel pool cleanup demineralizer outlet 134 140 Gross activity, chloride, fluoride 15 Secondary liquid waste evaporator distillate 150 120 Gross activity, general surveillance 16 Secondary liquid waste monitor tanks 150 115 Gross activity, tritium, dissolved and entrained gas, total suspended solid, oil and grease
- Sampling frequencies are specified within applicable chemistry procedures.
- The pressures and temperatures listed in this table represent nominal values at the time of plant licensing and should not be viewed as actual values observable in the field.
Rev. 17 WOLF CREEK TABLE 9.3-5 PROCESS SAMPLING SYSTEM SAMPLE POINT DESIGN DATA Operating Conditions Alarm Purpose for Measurement and
Sampling Point Pressure Temp. F Analysis Ranges Hi/Lo General Comments 1 Condenser hotwell 29.5" Hg 126 Cation Cond 0.055-1.0 S/cm 0.3/- Monitor condenser leakage and vac Na+ 0.01-10 ppb 0.1/- identify leak point Support conductivity data related to cond tube leak 2 Condensate pump 536 psig 127 Spec Cond 1-100 S/cm 60.0/- Aids in condensate and feedwater pH discharge header control - Oxygen Control Chemical control Cation Cond 0.055-1.0 S/cm 0.3/- Monitor condenser tube leak pH 6-11 pH 10.5/8.9 Support conductivity data related to pH control Na+ 0.01-10 ppb 0.1/- Monitor condenser tube leak Dissolved O 2 0.1-20 ppb 7/- Monitor condenser air leak and control oxygen control chemical feed Corrosion - - Monitor condensate corrosion product Product transport Concentration 3 Condensate demin 545 131 Cation Cond 0.055-1.0 S/cm 0.3/- Detect early breakthrough of outlet Na+ 0.01-10 ppb 0.3/- demineralizers Detect unexpected pickup or breakthrough 4 LP feedwater 375 336 Spec Cond 1-100 S/cm 60.0/- Control amine feed rate, pH heater outlet control pH 6-11 pH 10.5/9.0 Check effectiveness of pH control by spec cond 5 Steam generator 1104 445 Cation Cond 0.055-1.0 S/cm 0.3/- Monitor quality of feedwater feedwater pH 6-11 pH 10.5/8.9 Monitor alkalinity of feedwater Dissolved O 2 0.1-20 ppb 5/- Monitor effectiveness of oxygen control chemical feed Hydrazine 0-200 ppb -/60 Monitor residual oxygen control
chemical in feedwater Corrosion - - Monitor feedwater corrosion product Product transport
Concentration 6 Main steam A, B, 985 560 Spec Cond 1-100 S/cm 60.0/- Monitor steam purity and carry C, and D over tendency Rev. 25 WOLF CREEK TABLE 9.3-5 (Sheet 2) Operating Conditions Alarm Purpose for Measurement and Sampling Point Pressure Temp. F Analysis Ranges Hi/Lo General Comments 7 Stm gen blowdown 1095 560 Cation Cond 0.055-10 S/cm 1.0/- Determine TDS in S.G. and A, B, C, and D determine blowdown rate pH 6-11 10.5/9.0 Minimize corrosion and deposit
Na+ 0.1-100 ppb 5.0/- Monitor condenser leak and its concentration Determine blowdown rate 8 Heater drain pump 392 372 Cation Cond 0.055-10 S/cm 0.3/- disch/MSR drains A, B, C, and D Corrosion - - Monitor heater drain corrosion product Product transport
Concentration 9 Demin wtr degasi- F125 F150 Dissolved O 2 0.1-100 ppb 75/- Monitor effectiveness of fier in/out degasifier 10 MSR Drains 163 372 Spec Cond 1-100 S/cm 60/- Monitor water quality A,B,C and D pH 6-11 10.5/9.3
Rev. 25 WOLF CREEK TABLE 9.3-6 LIST OF GRAB SAMPLE POINTS FOR PRIMARY AND RADWASTE SAMPLING SYSTEMS Typical Sample Point Type of Sampling No. Sample Name Sample Frequency 1 Pressurizer relief tank, vapor space Noble gas As required 2 Volume control tank, vapor space Noble gas As required 3 Volume control tank Liquid As required 4 CVCS letdown chiller heat exchanger Liquid * (chill water) 5 Boric acid batch tank Liquid As required 6 Reactor makeup water storage tank Liquid
- 7 Steam generator blown demineralizer inlet Liquid
- and outlet 8 Refueling water storage tank Liquid
- 9 Fuel pool cooling heat exchangers A and B Liquid
- 10 Component cooling water heat exchanger Liquid
- 11 Residual heat removal pumps A and B Liquid
- 12 Refueling water storage tank return from Liquid As required RHR system 13 Safety injection pump test return to refueling Liquid As required water storage tank 14 Boron injection surge tank Liquid As required 15 Boron injection tank Liquid As required 16 Refueling water storage tank to spray Liquid As required additive eductor 17 Containment spray additive tank Liquid
- 18 Gas decay tanks A, B, C, D, E, F, and G Noble gas
- 19 Waste evaporator condensate Liquid As required 20 Liquid waste charcoal absorber Liquid As required
- Sampling frequencies are specified within applicable chemistry procedures Rev. 17 WOLF CREEK TABLE 9.3-6 (Sheet 2)
Sample Point Type Typical of Sampling No. Sample Name Sample Frequency 21 Waste monitor tank demineralizer inlet and Liquid As required outlet 22 Waste evaporator condensate demineralizer Liquid As required 23 Reactor coolant drain tank vapor space Noble gas As required 24 Waste evaporator concentrate Liquid As required 25 Evaporator bottom tank (primary) Liquid As required 26 Evaporator bottom tank (secondary) Liquid As required 27 Spent resin sluice filter (primary) Liquid As required 28 Spent resin sluice filter (secondary) Liquid As required 29 Recycle evaporator feed demin A and B Liquid As required 30 Recycle holdup tank vapor space A and B Noble gas As required 31 Recycle evaporator concentrate Liquid As required 32 Sec. liquid waste drain collection tank Liquid As required 33 Sec. liquid waste charcoal absorber Liquid As required 34 Sec. liquid waste demineralizer Liquid As required 35 High TDS collector tank Liquid As required 36 Low TDS collector tank Liquid As required 37 Sec. liquid waste evaporator entrainment Liquid As required separator rack spray water 38 Reactor containment sump Liquid As required 39 Laundry and hot shower charcoal absorber Liquid As required 40 SLW evaporator concentrates Liquid As required Rev. 17 WOLF CREEK TABLE 9.3-7 FLOOR AND EQUIPMENT DRAINAGE SYSTEM SINGLE ACTIVE FAILURE ANALYSIS Separation Component Failure Position Group Comments and Analysis FV95 As is 1 One of two containment isolation valves.
If valve fails, the other valve in sepa-ration group 4 will operate. FV96 Closed 4 One of two containment isolation valves. If valve fails, the other valve in separa-tion group 1 will operate. HV105 As is 4 One of two motor-operated auxiliary building sump pump discharge isola-tion valves. If valve fails, the other valve in separation group 1 will operate. HV106 As is 1 One of two motor-operated auxiliary building sump pump discharge iso-
lation valves. If valve fails, the other valve in separation group 4 will operate. LE102, LIT102 Anywhere in range 1 One level transmitter is provided for each RHR pump room with power supplied by the same separation group as the safety-related equipment in the associ-ated room. If one fails, the indication system for the other pump room train will operate. LE101, LIT101 Anywhere in range 4 One level transmitter is provided for each RHR pump room with power supplied Rev. 0 WOLF CREEK TABLE 9.3-7 (Sheet 2) Separation Component Failure Position Group Comments and Analysis by the same separation group as the
safety-related equipment in the associated room. If one fails, the indication system for the other pump room train will operate. LE103, LIT103 Anywhere in range 1 One of two auxiliary building sump level transmitter and indication systems. If one fails, the other train will operate. LE104, LIT104 Anywhere in range 4 One of two auxiliary building sump level transmitter and indication systems. If
one fails, the other train will operate. LE105, LIT105 Anywhere in range 1 One level transmitter is provided for each diesel generator buidling sump with power supplied by the same separation group as the safety-related equipment in the
associated room. If one fails, the other train will be protected by the other indication system. LE106, LIT106 Anywhere in range 4 One level transmitter is provided for each diesel generator building sump with power
supplied by the same separation group as the safety-related equipment in the associated room. If one fails, the other
train will be protected by the other indication system. Rev. 0 WOLF CREEK TABLE 9.3-7 (Sheet 3) Separation Component Failure Position Group Comments and Analysis LE124, LIT124 Anywhere in range 1 One of two control building basement level
transmitter and indication systems. If one fails, the other train will operate. LE125, LIT125 Anywhere in range 4 One of two control building basement level transmitter and indication systems. If one fails, the other train will operate. LE09, LIT09 Anywhere in range 1 One of two containment normal sump level transmitter and indication systems. If
one fails, the other train will operate. LE10, LIT10 Anywhere in range 4 One of two containment normal sump level transmitter and indication systems. If one fails, the other train will operate. Rev. 0 WOLF CREEK TABLE 9.3-8 CHEMICAL AND VOLUME CONTROL SYSTEM DESIGN PARAMETERS General Seal water supply flow rate, for four reactor coolant pumps, nominal, gpm 32 Seal water return flow rate, for four reactor coolant pumps, nominal, gpm 12 Letdown flow Normal, gpm 75
Maximum, gpm 120 Charging flow (excludes seal water) Normal, gpm 55
Maximum, gpm 100 Temperature of letdown reactor coolant entering system, F <560 Temperature of charging flow directed to reactor coolant system, F 518 Temperature of effluent directed to boron recycle system, F 115 Centrifugal charging pump miniflow, each, gpm 60 Normal charging pump miniflow, gpm 45
Amount of 4 weight percent boric acid solution required to meet cold shutdown requirements
shortly after full power operation, gal *18,500*Design nominal combined boric acid tank (BAT) boric solution volume. Rev. 13 WOLF CREEK TABLE 9.3-9 CHEMICAL AND VOLUME CONTROL SYSTEM PRINCIPAL COMPONENT DATA
SUMMARY
Normal charging Pump Number 1 Design pressure, psig 3,100 Design temperature, F 300 Design flow, gpm 130 Design head, ft 5,900 Material Austenitic stainless steel Design code ASME III-Class 2 Driver Type Electric motor RPM 3,600 Power supply 600 hp, 4000V, 3 Non-Class IE Seismic design Motor Non-Category I Pump Category I (pressure boundary) Centrifugal Charging Pumps Number 2 Design pressure, psig 2,800 Design temperature, F 300 Design flow, gpm 150 Design head, ft 5,800 Material Austenitic stainless steel Cooling water, gpm 55 Design code ASME III-Class 2 Driver Type Electric motor RPM 1,800 Power supply 600 hp, 4000V, 3 Class IE Seismic design Category I Boric Acid Transfer Pump Number 2 Design pressure, psig 150 Design temperature, F 250 Design flow, gpm 75 Rev. 10 WOLF CREEK TABLE 9.3-9 (Sheet 2) Design head, ft 235 Material Austenitic stainless steel Design code ASME III Class 3 Driver Type Electric motor RPM 3,450 Power supply 20.8hp, 460V, 3 Seismic design Motor Diesel backed/ Pump Non-Class IE Category I Boron Injection Makeup Pump Number 1 Design pressure, psig 150 Design temperature, F 250 Design flow, gpm 40 Design head, ft 233 Material Austenitic stainless steel Design code MS Seismic design Non-Category I Chiller Pumps Number 2 Design pressure, psig 150 Design temperature, F 200 Design flow, gpm 400 Design head, ft 150 Material Carbon steel
Design code MS Seismic design Non-Category I Regenerative Heat Exchanger Number 1 Heat transfer rate at design conditions, Btu/hr 11 0 x 10 6 Shell Side Tube Side Design pressure, psig 2,485 3,100 Design temperature, F 650 650 Fluid Borated reac- Borated reactor tor coolant coolant Material Austenitic Austenitic stain-stainless steel less steel Design code ASME III, Class 2 ASME III Class 2 Seismic design Category I Category I Rev. 15 WOLF CREEK TABLE 9.3-9 (Sheet 3) Shell Side Tube Side Flow, lb/hr 37,300 27,300 Inlet temperature, F 560 130 Outlet temperature, F 290 518 Letdown Heat Exchanger Number 1 Heat transfer rate at design conditions, Btu/hr 16.1 x 10 6 Shell Side Tube Side Design pressure, psig 150 600 Design temperature, F 250 400 Design flow, lbm/hr 498,000 59,600 Fluid Component cool- Borated reactor ing water coolant Material Carbon steel Austenitic stain-
less steel Design code ASME III, ASME III, Class 2 Class 3 Seismic design Category I Category I Excess Letdown Heat Exchanger Number 1 Heat transfer rate at design conditions, 5.2 x l0 6 Btu/hr Shell Side Tube Side Design pressure, psig 150 2,485 Design temperature, F 250 650 Design flow, lb/hr 129,000 12,410 Inlet temperature, F 105 560 Outlet temperature, F 145 165 Fluid Component cool- Borated reactor ing water coolant Material Carbon steel Austenitic stainless steel Design code ASME III, Class 3 ASME III, Class 2 Seismic design Category I Category I Seal Water Heat Exchanger Number 1 Heat transfer rate at design conditions, Btu/hr 2.0 x 10 6 Rev. 0 WOLF CREEK TABLE 9.3-9 (Sheet 4) Shell Side Tube Side Design pressure, psig 150 220 Design temperature, F 250 250 Design flow, lb/hr 186,000 51,900 Inlet temperature, F 105 156 Outlet temperature, F 121 115 Fluid Component Borated reactor cooling water coolant Material Carbon steel Austenitic stainless steel Design code ASME III, Class 3 ASME III, Class 2 Seismic design Category I Category I Moderating Heat Exchanger Number 1 Heat transfer rate at design conditions, Btu/hr 2.53 x 10 6Design pressure, psig 300 300 Shell Side Tube Side Design temperature, F 200 200 Design flow, lb/hr 59,600 59,600 Design inlet temperature, boron storage mode, F 50 115 Design outlet temperature, boron storage mode, F 92.4 72.6 Inlet temperature, boron release mode, F 140 115 Outlet temperature, boron release mode, F 123.2 131.8 Material Austenitic Austenitic stainless steel stainless steel Design code (1) ASME VIII ASME VIII Seismic design Non-Category I Non-Category I Letdown Chiller Heat Exchanger Number 1 Heat transfer rate at design conditions, boron storage mode, Btu/hr 1.65 x 10 6 Rev. 19 WOLF CREEK TABLE 9.3-9 (Sheet 5) Shell Side Tube Side Design pressure, psig 150 300 Design temperature, F 200 200 Design flow, boron storage mode, lb/hr 175,000 59,600 Design inlet temperature, boron storage mode, F 39 72.6 Design outlet temperature, boron storage mode, F 48.4 45 Shell Side Tube Side Flow, boron release mode, 175,000 59,600 lb/hr Inlet temperature, boron release mode, F 90 123.7 Outlet temperature, boron release mode, F 99.8 94.9 Material Carbon steel Austenitic stainless steel Design code (1) ASME VIII ASME VIII Seismic design Non-Category I Non-Category I Letdown Reheat Heat Exchanger Number 1 Heat transfer rate at design conditions, Btu/hr 1.49 x 10 6 Shell Side Tube Side Design pressure, psig 300 600 Design temperature, F 200 400 Design flow, lb/hr 59,600 44,700 Inlet temperature, F 115 280 Outlet temperature, F 140 246.7 Material Austenitic Austenitic Stainless steel Stainless steel Design code (1) ASME VIII ASME VIII, Class 2 Seismic design Non-Category I Non-Category I Volume Control Tank Number 1 Volume, ft3 400 Design pressure, psig 75 Design temperature, F 250 Material Austenitic stainless steel
Design code ASME III, Class 2 Seismic design Category I Rev. 0 WOLF CREEK TABLE 9.3-9 (Sheet 6) Boric Acid Tanks Number 2 Capacity, usable, gal 24,000 Design pressure, psig 10 Design temperature, F 200 Material Austenitic stainless steel
Design code ASME III, Class 3 Seismic design Category I Batching Tank Number 1 Capacity, gal 800 Design pressure Atmospheric vessel steam jacket, psig Design temperature, F 150 (steam jacket) 400 Material Austenitic stainless steel Design code ASME VIII Seismic design Non-Category I Chemical Mixing Tank Number 1 Capacity, gal 5 Design pressure, psig 150 Design temperature, F 200 Material Austenitic stainless steel Design code ASME VIII Seismic design Non-Category I Chiller Surge Tank Number 1 Volume, gal 500 Design pressure Atmospheric Design temperature, F 200 Material Carbon steel Design code ASME VIII Seismic design Non-Category I Mixed Bed Demineralizers Number 2 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 120 Resin volume, each, ft 3 39 Rev. 0 WOLF CREEK TABLE 9.3-9 (Sheet 7) Material Austenitic stainless steel Design code (1) ASME VIII Seismic design Non-Category I Cation Bed Demineralizers Number 1 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 120 Resin volume, ft3 39 Material Austenitic stainless steel Design code (1) ASME VIII Seismic design Non-Category I Thermal Regeneration Demineralizers Number 5 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 250 Resin volume, ft3 74.3 Material Austenitic stainless steel Design code (1) ASME VIII Seismic design Non-Category I Reactor Coolant Filter Number 1 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 250 Particle retention Absolute filtration program target size is .1 micron at > 99.98% efficiency Material, vessel Austenitic stainless steel Design code ASME III Class 2 Seismic design Category I Seal Water Injection Filters Number 2 Design pressure, psig 3,100 Design temperature, F 250 Design flow, gpm 80 Particle retention Absolute filtration program target size is .1 micron at > 99.98% efficiency Material, vessel Austenitic stainless steel Design code ASME III Class 2 Seismic design Category I Rev. 10 WOLF CREEK TABLE 9.3-9 (Sheet 8) Seal Water Return Filter Number 1 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 250 Particle retention Absolute filtration program target size is .1 micron at > 99.98% efficiency size Material, vessel Austenitic stainless steel Design code ASME III, Class 2 Seismic design Category I Boric Acid Filter Number 1 Design pressure, psig 300 Design temperature, F 250 Design flow, gpm 250 Particle retention Absolute filtration program target size is .1 micron at > 99.98% efficiency size Material, vessel Austenitic stain-less steel Design code ASME III, Class 3 Seismic design Category I Letdown Orifice 45 gpm 75 gpm Number 1 2 Design flow, lb/hr 22,200 37,300 Differential pressure at design flow, psig 1,525 1,525 Design pressure, psig 2,485 2,485 Design temperature, F 650 650 Material Austenitic Austenitic stainless stainless steel steel Design code ASME III, ASME III, Class 2 Seismic design Class 2 Category I Category I Chiller Unit Number 1 Capacity, Btu/hr 1.66 x 10 6 (ice tons) 138 Design code MS Seismic design Non-Category I Rev. 10 WOLF CREEK TABLE 9.3-9 (Sheet 9) Note 1 - Table indicates the required code based on its safety-related importance as dictated by service and functional requirements and by the consequences of their failure. Note that the actual equipment may be supplied to a higher principal construction code than required. Rev. 0 WOLF CREEK TABLE 9.3-10 FAILURE MODE AND EFFECTS ANALYSIS-CHEMICAL AND VOLUME CONTROL SYSTEM ACTIVE COMPONENTS - NORMAL PLANT OPERATION AND SAFE SHUTDOWN CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- 1. Air diaphragm- a. Fails open Charging and volume Failure reduces redundancy Valve position indi- Valve is desig ned operated globe control - letdown of providing letdown flow cation (open to to fail "close d" valve LCV-459 flow isolation to protect PRZ closed position and is electri
- (LCV-460 heaters from uncovering change) at CB. cally wired so
analogous) at low water level in the that the elect ri- PRZ. No effect on system cal solenoid o f operation. Alternate the air diaphr agm isolation valve (LCV-460) operator is en er- provides back-up letdown gized to open the flow isolation. valve. Soleno id is de-energize d to close the valv e upon the gener a- tion of a low
level PRZ cont rol signal. The v alve is electricall y interlocked wi th three letdown ori- fice isolation
valves and may not be opened manu ally from the CB if any of these valve s is at an open pos ition. b. Fails Charging and volume Failure blocks normal Valve position indi-
closed control - letdown letdown flow to VCT. cation (closed to
flow Minimum letdown flow open position change)
requirements for boration at CB; letdown flow
of RCS to safe shutdown temperature indica-
concentration level may tion (TI-127) at CB;
be met by establishing letdown flow-pressure
letdown flow through indication (PI-131)
alternate excess letdown at CB; letdown flow
flow path. If the alter- indication (FI-132)
nate excess letdown flow at CB; and VCT level
path to VCT is not avail- indication (LI-185) able due to a single fail- and low water level ure (loss of instrument alarm at CB.
- 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. Rev. 0 WOLF CREEK WOLF CREEK TABLE 9.3-10 (Sheet 2)
CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks air supply) affecting
the opening operation of
valves in each flow path, the plant operator can
borate the RCS to a safe shutdown concentration level without letdown flow by
utilizing the steam space
available in the PRZ.
- 2. Air diaphram- a. Fails open Charging and volume Failure prevents isolation Valve position indi- Valve is of si mi- operated globe control - letdown of normal letdown flow cation (open to lar design as that valve LCV-8149B flow through regenerative heat closed position stated for ite m 1. (LCV-8149C and exchanger when bringing change) at CB. Solenoid is de
- 8149A analogous) the reactor to a cold shut- energized to c lose down condition after the the valve upon the RHRS is placed into opera- generation of a tion. No effect on safe low water leve l shutdown operation. Con- PRZ signal or tainment isolation valve closing of let down 8152 or 8160 may be re- isolation valv es motely closed from the CB LVC-459 and LC V- to isolate letdown flow 460 upstream o f through the heat exchanger. the regenerati ve heat exchanger . b. Fails Charging and volume Failure blocks normal let- Same methods of
closed control - letdown down flow to VCT. Normal detection as those
flow letdown flow to VCT may be stated for item 1.b.
maintained by opening alter-
nate letdown orifice isola-
tion valve 8149C. Minimum
letdown flow requirements
for boration of RCS to safe
shutdown concentration level
may be met by opening let-
down orifice isolation valve LCV-8149A or LCV-8149C. If a single failure (loss of
instrument air) prevents
opening of these valves
the plant operator can borate
the RCS to a safe shutdown
concentration level without
letdown flow by utilizing the
steam space available in PRZ. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 3) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- 3. Air diaphragm- a. Fails Charging and volume Same effect on system Same methods of Valve is of si m- operated globe closed control - letdown operation as that stated detection as those lar design as that valve 8152 flow for item 1.b. stated for item 1.b. stated for ite m 1. (8160 analogous) In addition, close Solenoid is de
- position group moni- energized to c lose toring light at CB. the valve upon the generation of an ESF "T" signal . b. Fails open Charging and volume Failure has no effect on Valve position indi-control - letdown CVCS operation during cation (open to
flow normal plant operation closed position
and load follow. However, change) at CB.
under accident conditions
requiring containment iso-
lation, failure reduces the
redundancy of providing isolation of normal let-down line.
- 4. Air diaphragm- a. Fails open Boron concentration Failure inhibits use of Letdown heat ex- 1. Valve is operated globe control - boron ther- BTRS for load follow changer tube dis- designed t o valve TCV-381B mal regeneration operation (boration) due charge flow (FI-132) fail "open
" (boration) to low temperature of and pressure (PI-131) and is ele c- letdown flow entering indications at CB trically w ired BTRS demineralizers. and BTR deminer- so that th e Alternate boration of alizer inlet flow electrical reactor coolant for temperature indica- solenoid o f the load follow is possible, tion (TI-381) at CB air diaphr agm using RMCS of CVCS. No if BTRS is in operator i s effect of operation to operation. energized to bring reactor to safe close the valve. shutdown condition.
- 2. BTRS opera tion is not req uired in operati ons of the CVC S used to br ing the reacto r to hot standb y condition.
- b. Fails Boron concentration Failure inhibits use of Same methods of closed control - boron BTRS for load follow detection as those
thermal regeneration operation (boration) due stated for item 1.b, (boration) to loss of temperature except no "closed to Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 4) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks control of letdown flow open position change" entering BTRS deminer- indication at CB. If
alizers. Failure also BTRS is not operating, blocks normal letdown BTRS status indication flow to VCT when BTRS is (off) light at CB.
not being used for load
follow. Minimum letdown flow requirements for boration of RCS to hot
standby concentration
level may be met as stated
for effect on system opera-
tion for item 1.b.
- 5. Air diaphragm- a. Fails open Charging and volume Failure prevents control Letdown heat ex- 1. Same remar k as operated globe control - letdown of pressure to prevent changer tube dis- stated for item valve PCV-131 flow flashing of letdown flow charge flow indica- 4, in rega rd in letdown heat exchanger tion (FI-132) and to valve d esign. and also allows high high flow alarm at pressure fluid to mixed CB; temperature 2. As a desig n bed demineralizers. indication (TI-130) transient the Relief valve 8119 opens and high temperature letdown he at ex- in demineralizer line to alarm at CB; and changer is
release pressure to VCT pressure indication designed f or and valve TCV-129 changes (PI-131) at CB. complete l oss of position to divert flow charging f low. to VCT. Boration of RCS to safe shutdown concentra-tion level is possible
with valve failing open.
- b. Fails Charging and volume Same effect on system Letdown heat ex-closed control - letdown operations as that stated changer discharge
flow for item 1.b. flow indication
(FI-132), pressure
indication (PI-131)
and high pressure
alarm at CB.
- 6. Air diaphragm- a. Fails open Charging and volume Letdown flow bypassed Valve position indi- 1. Electrical sole- operated three for flow control - letdown from flowing to mixed cation (VCT) at CB noid of ai r way valve TCV- only to flow bed demineralizers and and RCS activity diaphragm opera- 129 VCT BTRS. Failure prevents level when sampling tor is ele ctri- ionic purification of letdown flow. cally wire d so letdown flow and inhi- that solen oid is bits operation of BTRS. energized to Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 5)
CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks Boration of RCS to safe open valve for shutdown concentration flow to th e level is possible with mixed bed demin- valve failing open for eralizers. flow only to VCT. Valve open s for flow to VC T on "high letd own temperatur e" or on "high l etdown reheat hea t ex- changer ou tlet temperatur e." 2. Technical
Specificat ions provide a limit on RCS act ivity. b. Fails open Charging and volume Continuous letdown to Valve position indi-
for flow control - letdown mixed bed demineralizers cation (demineralizer)
only to flow and BTRS. Failure pre- at CB. If BTRS is
mixed bed vents automatic isola- in operation, BTR
deminer- tion of mixed bed demin- demineralizer return
alizer eralizers and BTRS under flow indication (FI-
fault condition of high 385) indicating flow
letdown flow temperatures. during an alarm con-
These systems may be dition of high let-manually isolated, using down reheat heat ex-local valves 8524A and changer outlet tem-
8524B at mixed bed demin- perature or high let-
eralizers. Boration of down temperature.
RCS to safe shutdown con-centration level is possible with valve failing
open for flow only to demin-
eralizer.
- 7. Air diaphragm- Fails closed Boron concentration Failure inhibits use Valve position indi- 1. Valve is operated dia- control - boron of BTRS for load fol- cation (closed to designed t o phragm valve thermal regeneration low operation (boration open position change) fail "clos ed" 7054 or storage or dilution) due to flow at CB; BTRS operation and is ele ctri- isolation of the BTRS. indication (borate or cally wire d so Alternate boration or dilute) at CB and BTR that the e lec- dilution of reactor demineralizer return trical sol enoid coolant for load follow flow indication of air dia
- may be accomplished using (FI-385) and inlet phragm ope rator RMCS of CVCS. No effect flow temperature is energiz ed to Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 6) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks on operation to bring indication (TI-381) open the v alve. reactor to safe shutdown at CB. condition. 2. BTRS not r e- quired to bring reactor to safe shutdown c ondi- tion.
- 8. Air diaphragm- a. Fails Boron concentration Failure inhibits use of BTRS operation indi- Same remarks a s operated dia- closed control - boron BTRS for load follow cation (dilute) at those stated f or phragm valve storage operation (dilution) due CB; letdown reheat item 7.
7002A to flow isolation of heat exchanger out-
letdown chiller heat let temperature
exchanger. Alternate (TI-381) at CB;
dilution of reactor and RCS boron level
coolant for load follow when sampling letdown
may be accomplished using flow. RMCS of CVCS. No effect on operations to bring
reactor to safe shutdown
condition.
- b. Fails open Boron concentration Failure inhibits use of BTRS operation indi-control - boron BTRS for load follow cation (boration) at
thermal regeneration operation (boration) due CB; BTRS return flow
to flow through letdown temperature indica-
chiller heat exchanger. tion (TI-386) at CB; Alternate boration of BTR return flow indi-reactor coolant for load cation (FI-385) at
follow may be accomplished CB; and RCS boron
using RMCS of CVCs. No level when sampling
effect on operation to letdown flow. bring reactor to safe shutdown condition.
- 9. Air diaphragm- a. Fails Boron concentration Same effect on system Same methods of Same remarks a s operated dia- closed control - boron operation as that stated detection as those those stated f or phragm valve storage for item 8.a. stated for item 8.a. item 7.
7002B
- b. Fails open Boron concentration Failure inhibits use of Same methods of control - boron BTRS for load follow detection as those
thermal regeneration operation (boration) due stated for item 8.b.
to bypass of letdown flow
from letdown reheat heat
exchanger. Alternate bora-
tion of reactor coolant may Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 7) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks be accomplished, using RMCS of CVCS. No effect on
operation to bring reactor to safe shutdown condition.
- 10. Relief valve Fails open Charging and volume Letdown flow is relieved High temperature Radioactive fl uid 8117 control - letdown to pressurizer relief tank. relief line indica- contained.
flow Failure inhibits use of tion (TI-125) and demineralizers for reactor alarm at CB and VCT
coolant purification and level indication
use of BTRS. Normal let- (LI-185) and low
down line can be isolated level alarm at CB.
and minimum letdown flow
requirements for hot
standby may be met by
establishing letdown flow through alternate excess letdown flow path.
- 11. Relief valve Fails open Charging and volume Letdown flow is relieved RCS activity level Radioactive fl uid 8119 control - letdown to VCT. Failure inhibits when sampling let- contained.
flow use of demineralizers for down flow. When
reactor coolant purifica- BTRS is operating, tion and use of BTRS. low BTR demineral-
Normal letdown line can izer return flow
be isolated and minimum indication (FI-385)
letdown flow requirement at CB. for hot standby may be met by establishing flow
through alternate excess
letdown flow path.
- 12. Air diaphragm- a. Fails Boron concentration Normal purification of BTRS operation indi- Same remarks a s operated dia- closed control - boron reactor coolant using only cation (off) at CB those stated f or phragm valve thermal regeneration mixed bed demineralizers and RCS activity item 4.
8245 or storage cannot be performed. level when sampling
Failure also blocks normal letdown flow. Valve
letdown flow. Boration position indication
of RCS to safe shutdown (closed to open concentration level may be position change) at met as stated for effect CB.
on system operation for
item 1.b. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 8) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- b. Fails open Boron concentration Failure inhibits use of RCS boron level when control - boron BTRS for load follow opera- sampling letdown flow.
thermal regeneration tion (boration or dilution) If BTRS is operating, or storage due to bypass of letdown BTRS operating indica-flow from BTRS. Alternate tion (borate or dilute)
boration or dilution of at CB and low BTR demin-
reactor coolant for load eralizer return flow follow may be accomplished indication (FI-385) at using RMCS of CVCS. No CB. Valve position
effect on operations to indication (open to
bring reactor to hot stand- closed position change)
by condition. at CB.
- 13. Air diaphragm- a. Fails Boron concentration Failure inhibits use of RCS boron level when See remarks as operated dia- closed control - boron BTRS for load follow oper- sampling letdown those stated f or phragm valve storage ation (dilution) due to flow. If BTRS is item 7.
7056 (7045 flow isolation of BTR operating, BTRS analogous) demineralizers. Alternate operation indication dilution of reactor cool- (dilute) at CB and low
ant for load follow may be BTR demineralizer
accomplished, using RMCS of return flow indication
CVCS. No effect on oper- (FI-385) at CB.
ations to bring reactor to
safe shutdown condition.
- b. Fails open Boron concentration Failure inhibits use of RCS boron level when control - boron BTRS for load follow opera- sampling letdown flow.
thermal regeneration tion (boration) due to flow If BTRS is operating, bypass of BTR demineral- BTRS operation indica-
izers. Alternate boration tion (borate) at CB.
of reactor coolant for load
follow may be accomplished
using RMCS of CVCS. No
effect on operations to
bring reactor to safe
shutdown condition.
- 14. Air diaphragm- a. Fails open Boron concentration Failure inhibits use of RCS boron level when Same remarks a s operated dia- control - boron BTRS for load follow oper- sampling letdown those stated f or phragm valve storage ation (dilution) due to flow. If BTRS is item 4.
7057 (7046 flow bypass of BTR demin- operating, BTRS
analogous) eralizers. Alternate dilu- operation indication
tion of reactor coolant (dilute) at CB.
for load follow may be
accomplished using RMCS of
CVCS. No effect on opera-
tions to bring reactor to
safe shutdown condition. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 9) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- b. Fails Boron concentration Failure inhibits use of RCS boron level when closed control - boron BTRS for load follow oper- sampling letdown flow.
thermal regeneration ation (boration) due to If BTRS is operating, flow isolation of BTR BTRS operation indica-demineralizers. Alternate tion (borate) at CB and
boration of reactor cool- low BTR demineralizer
ant for load follow may be return flow indication accomplished, using RMCS of (FI-385) at CB. CVCS. No effect on opera-
tions to bring reactor to
safe shutdown condition.
- 15. Air diaphragm- a. Fails open Boron concentration Same effect on system Same methods of Same remarks a s operated dia- control - boron operation as that stated detection as those those stated f or phragm valve storage for item 14.a. stated for item item 4.
7040 14.a.
- b. Fails Boron concentration Failure inhibits use of Same metods of detec-closed control - boron BTRS for load follow oper- tion as those stated
thermal regeneration ation (boration) due to for item 14.b.
blockage of return letdown
flow from letdown chiller
heat exchanger. Alternate
boration of reactor coolant
for load follow may be
accomplished, using RMCS of
CVCS. No effect on opera-tions to bring reactor to hot standby condition.
- 16. Air diaphragm- a. Fails open Boron concentration Failure inhibits use of RCS boron level when Same remarks a s operated dia- control - boron BTRS for load follow opera- sampling letdown those stated f or phragm valve storage tion (dilution) due to flow. If BTRS is item 4.
7041 flow bypass of letdown operating, BTRS
chiller heat exchanger. operation indication
Alternate dilution of (dilute) at CB and
reactor coolant for load letdown reheat heat
follow may be accomplished, exchanger outlet tem-
using RMCS of CVCS. No perature indication effect on operation to (TI-381) at CB. bring reactor to safe
shutdown condition. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 10) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- b. Fails Boron concentration Failure inhibits use of Same methods of detec-closed control - boron BTRS for load follow oper- tion as those stated
thermal regeneration ation (boration) due to for item 14.b. flow isolation of letdown reheat heat exchanger and
BTR demineralizers. Alternate
boration of reactor coolant for load follow may be accom-plished using RMCS of CVCS.
No effect on operation to
bring reactor to safe shut-
down condition.
- 17. Air diaphragm- a. Fails Boron concentration Failure inhibits use of RCS boron level when Same remarks a s operated dia- closed control - boron BTRS for load follow oper- sampling letdown those stated f or phragm valve storage ation (dilution) due to flow. If BTRS is item 7.
7022 flow blockage of return operating, BTRS letdown flow from modera- operation indication ting heat exchanger. (dilute) at CB and
Alternate dilution of low BTR demineralizer
reactor coolant for load return flow indication
follow may be accomplished, (FI-385) at CB.
using RMCS of CVCS. No
effect on operation to bring
reactor to safe shutdown
condition.
- b. Fails open Boron concentration Failure inhibits use of RCS boron level when control - boron BTRS for load follow sampling letdown flow.
thermal regeneration operation (boration) due If BTRS is operating, to bypass of flow from BTRS operate indication
letdown chiller heat ex- (borate) at CB and BTRS changer of return letdown return flow temperature flow. Alternate boration indication (TI-386) and
of reactor coolant for high temperature alarm
load follow may be accom- at CB.
plished, using RMCS of
CVCS. No effect on oper-
ation to bring reactor to safe shutdown condition.
- 18. Air diaphragm- Fails closed Boron concentration Failure inhibits use of RCS boron level when 1. Valve is operated butter- control - boron BTRS for load follow oper- sampling letdown designed t o fly valve TCV-386 thermal regeneration ation (boration and dilu- flow. If BTRS is fail "clos ed." and storage tion) due to flow blockage operating, BTRS
of chiller flow through return flow tempera- 2. BTRS not u sed letdown chiller heat ture indication (TI- to bring t he exchanger. Alternate 386) and high temper- reactor to safe boration and dilution of ature alarm at CB; shutdown c ondi- reactor coolant for load and chiller surge tion. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 11) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks follow may be accomplished, tank temperature indi-using RMCS of CVCS. No cation (TI-379) at CB.
effect on operations to bring reactor to safe shutdown condition.
- 19. Air diaphragm- Fails open Boron concentration Failure inhibits use of RCS boron level when 1. Valve is operated control - boron BTRS for load follow sampling letdown designed t o butterfly valve thermal regeneration operation (boration and flow. If BTRS is fail "open
." FCV-375 and storage dilution) due to flow operating, BTRS bypass of chiller flow return flow tempera- 2. BTRS not u sed from letdown chiller heat ture indication (TI- to bring t he exchanger. Alternate bora- 386) and high temper- reactor to safe tion and dilution of ature alarm at CB and shutdown c ondi- reactor coolant for load chiller surge tank tion.
follow may be accomplished, temperature indication using RMCS of CVCS. No (TI-379) at CB. effect on operations to bring reactor to safe shut-
down condition.
- 20. Chiller unit, Fails to cool Boron concentration Failure inhibits use of Same methods of BTRS not used to AHCU liquid control - boron BTRS for load follow detection as those bring the reac tor thermal regeneration operation (boration and stated for item 19. to safe shutdo wn and storage dilution) due to loss of In addition, BTRS condition.
cooling capability of let- operation indication
down chiller heat ex- (borate or dilute) changer. Alternate bora- at CB. tion and dilution of reac-
tor coolant for load follow
may be accomplished, using
RMCS of CVCS. No effect
on operations to bring
reactor to hot standby
condition.
- 21. Chiller pump 1, Fails to Boron concentration No effect on BTRS opera- Local pump discharge Both chiller p umps APCI (pump 2 deliver work- control - boron tion. Redundant chiller flow pressure indica- operate simult an- analogous) ing fluid thermal regeneration pump 2 provides necessary tion (PI-377A) and eously during BTRS and storage delivery of working fluid MCC contactor posi- operation.
for chiller unit operation. tion indication
BTRS not required in (open) at CB.
operations to bring reactor
to hot standby condition.
- 22. Air diaphragm a. Fails Boron concentration Failure inhibits use of RCS boron level when Same remarks a s operated closed control - boron BTRS for load follow sampling letdown those stated f or diaphragm valve thermal regeneration operation (boration) due flow. If BTRS is item 7.
7002A (7002B and storage
analogous) Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 12) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks to flow isolation of shell operating, letdown side of letdown reheat reheat heat exchanger
heat exchanger. Alternate outlet temperature boration of reactor cool- indication (TI-381) ant for load follow may be at CB.
accomplished using RMCS of
CVCS. No effect on opera-tions to bring reactor to safe shutdown condition.
- b. Fails open Boron concentration Failure inhibits use of RCS boron level when control - boron BTRS for load follow sampling letdown flow.
storage operation (dilution) due If BTRS is operating, to passage of CVCS letdown letdown reheat heat
flow through tube side of exchanger outlet tem-
letdown reheat heat ex- perature indication changer. Alternate dilu- (TI-381) at C.B. tion of reactor coolant may be accomplished using
RMCS of CVCS. No effect
on operations to bring
reactor to safe shutdown
condition.
- 23. Solenoid- a. Fails Charging and volume Failure reduces redundancy Valve open/closed 1. If normal let- operated closed control - letdown of the excess letdown fluid position indica- down and e xcess globe valve flow system of the CVCS as an tion at CB; letdown flow is no t 8153A alternate system that may high temperature available for (8154A be used for letdown flow indication and alarm safe shutd own analogous; control during normal at CB. operations ,plant 8153B and plant operation and operator c an 8154B reduces redundancy of the borate RCS to similar) excess letdown system to safe shutd own control water level in the concentrat ion, pressurizer of the RCS using stea m during final stage of space avai lable plant startup due to in PRZ.
flow blockage.
- b. Fails open Charging and volume Failure reduces redun- Valve position indi-control - letdown dancy of providing excess cation (open to
flow Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 13) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks letdown flow isolation dur- closed position change) ing normal plant operation at CB.
and for plant startup. No effect on system operation. Alternate isolation valves
closed to provide back-up
flow isolation of excess letdown line.
- 24. Air diaphragm- a. Fails Charging and volume Same effect on system Same methods of Same remarks a s operated globe closed control - letdown operation as that stated detection as those those stated valve HCV-123 flow for item 23.a. Redundant stated for item 23.a, for item 23
valves 8157A or 8157B except for val ve may be opened to provide position indic ation a path to the PRT. at CB.
- b. Fails open Charging and volume Failure prevents manual Excess letdown heat control - letdown adjustment at CB of RCS exchanger outlet flow pressure downstream of pressure indication
excess letdown heat (PI-124) at CB, and
exchanger to a low pres- seal water return flow
sure consistent with recording (FR-156) and
number 1 seal leakoff back- low flow alarm at CB.
pressure requirements.
When using excess letdown
system, failure leads to a
decrease in seal water pump shaft flow for cooling pump bearings.
- 25. Air disphragm- a. Fails Charging and volume No automatic makeup of Valve position indi- 1. Same remar k as operated dia- closed control - seal water seal water to seal stand- cation (closed to that state d for phragm valve flow pipe that services number open position change) item 7 in regard LCV-181 3 seal of RC pump 1. No and low standpipe to valve d esign. (LCV-178, LCV- effect on operations to level alarm at CB.
179 and LCV- bring the reactor to safe 2. Low level stand- 180 analogous) shutdown condition. pipe alarm con- servativel y set to allow a ddi- tional tim e for RC pump op era- tion witho ut a complete l oss of seal water from being inje cted to number 3 seal after soun ding of alarm. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 14) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- b. Fails open Charging and volume Overfill of seal water Valve position indi-
control - seal water standpipe and dumping of cation (open to closed
flow reactor makeup water to position change) and
containment sump during high standpipe level
automatic makeup of alarm at CB.
water for number 3 seal
of RC pump 1. No effect on operations to bring reactor to safe shutdown
condition.
- 26. Relief valve Fails open Charging and volume RC pump seal water return Decrease in VCT 1. The capaci ty of 8121 control - seal water flow and normal excess level, causing RMCS the relief valve flow letdown flow bypassed to of CVCS to operate. equals max imum PRZ relief tank of RCS. flow from four Failure inhibits use of the RC pump se als excess letdown fluid sys- plus norma l ex- tem of the CVCS as an cess letdo wn alternate system that may flow.
be used for letdown flow
control during normal plant 2. Radioactiv e operation and inhibits use fluid con-
of normal excess letdown sys- tained.
tem to control water level in
the PRZ of the RCS during final 3. Same remar k as stage of a plant startup. that state d for item 23 (#2).27. Motor-operated a. Fails open Charging and volume Failure has no effect on Valve position indi- 1. Valve is n or- globe valve control - seal water CVCS operation during cation (open to mally at a full 8112 (8100 flow and excess let- normal plant operation closed position open posit ion, analogous) down flow and load follow. However, change) at CB. and motor oper- under accident conditions ator is en er- requiring containment isola- gized to c lose tion, failure reduces redun- the valve upon dancy of providing isolation the genera tion of seal water flow and normal of an ESF "T" excess letdown flow. signal.
- 2. If normal let- down and n ormal excess let down flow is no t available for safe shutd own operation, plant operator c an borate RCS to safe shutd own concentrat ion, using stea m space avai lable in PRZ and excess Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 15)
CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks letdown path t o PRT.
- b. Fails Charging and volume RC pump seal water return Valve position indi-closed control - seal water flow and normal excess let- cation (closed to open
flow and excess let- down flow blocked. Failure position change) at
down flow inhibits use of the normal CB; group monitoring excess letdown fluid system light at CB; and seal of the CVCS as an alternate water return flow
system that may be used for recording (FR-157) and
letdown flow control during low seal water return
normal plant operation. flow alarm at CB.
However, excess letdown path
to PRT will be available along
with increased steam space in
PRZ. Also degrades cooling
capability of seal water in cooling RC pump bearings. CCW should be established to
the seals and seal injection
terminated. This minimizes
water loss to PRT via relief
valve 8121. Valve 8121 will
continue to pass seal leakage
to PRT (5 gpm per seal) until
the RCS pressure is reduced.
- 28. Motor-operated a. Fails open Charging and volume Failure has no effect on Valve position indi- Valve is nor-gate valve control - charging CVCS operation during cation (open to mally at a ful l 8105 (8106 flow normal plant operation closed position open position, and analogous) and load follow. However, change) at CB. motor operator is under accident conditions energized to c lose requiring isolation of the valve upon the charging line, failure generation of a reduces redundancy of pro- safety injecti on viding isolation of normal "S" signal.
charging flow.
- b. Fails Charging and volume Failure inhibits use of Valve position indi-closed control - charging normal charging line to cation (closed to open
flow RCS for boration, dilution, position change) and
and coolant makeup opera- group monitoring light
tions. Seal water injec- (valve closed) at CB;
tion and BIT paths remain letdown temperature
available for boration of indication (TI-127) and
RCS to a safe shutdown con- high temperature alarm
centration level and makeup at CB; charging flow
of coolant during operations temperature indication
to bring the reactor to (TI-126) at CB; seal
safe shutdown condition. water flow pressure indication (PI-120A) at Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 16) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks CB; VCT level indica-
tion (LI-185) and high
level alarm at CB.
- 29. Air diaphragm- a. Fails open Charging and volume Failure prevents manual Seal water flow pres- Same remark as that operated globe control - charging adjustment at CB of seal sure indication (PI- stated for ite m 4 valve HCV-182 flow and seal water water flow through the 120A) at CB; seal in regard to flow control of backpressure water return record- design of valv
- e. in charging header, result- ing (FR-157); and
ing in a reduction of flow low seal water
to RC pump seals leading to return flow alarm
a reduction in flow to RCS at CB.
via labyrinth seals and
pump shaft flow for cool-
ing pump bearings. Bora-tion of RCS to a safe shut-down concentration level
and makeup of coolant during
operations to bring reac-
tor to safe shutdown condi-
tion is still possible
through normal charging
flow path or BIT path.
- b. Fails Charging and volume Same effect on system Same methods of closed control - charging operation as that stated detection as those flow for item 28.b. stated for item 28.b.
- 30. Motor-operated a. Fails open Charging and volume Failure has no effect on Valve position indi- 1. Same remar ks as globe valve control - charging CVCS operation during cation (open to those stat ed for 8110 (8111) flow and seal water normal plant operation closed position item 28.
analogous) flow and load follow. How- change) at CB. ever, under accident con-ditions requiring isola-
tion of centrifugal
charging pump miniflow line to suction of pumps via seal water heat ex-
changer, failure results
in reduction of delivered
flow for one 100 percent
train only. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 17) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- b. Fails Charging and volume Failure blocks miniflow Valve position indi-closed control - charging to suction of centrifu- cation (closed to
flow and seal water gal charging pumps via open position change) flow seal water heat exchanger. at CB; group monitoring Normal charging flow and light (valve closed)
seal water flow prevents and alarm at CB; and
deadheading of pumps when charging and seal water used. Boration of RCS flow indication (FI-121A) to a safe shutdown concen- and high flow alarm at CB.
tration level and makeup
of coolant during opera-
tions to bring reactor to
safe shutdown condition is
accomplished by the opposite
train which provides 100
percent of the flow require-ments. 31. Air diaphragm- a. Fails open Charging and volume Failure has no effect on Valve position indi- Same remark as tha t operated globe control - charging CVCS operation during normal cation (open to stated for item 4 valve 8146 flow plant operation, load follow closed position in regard to and safe shutdown operation. change) at CB. design of valve . BG HV-8147 can be removed from service, if required, since only one return path is required. In the event that auxiliary spray is being used to cooldown the PZR, the charging return flow path is isolated to maximize auxiliary spray flow. Cold shutdown of the reactor is still possible; however, time for cooling down the PZR will be extended. b. Fails Charging and volume Isolates one of the two Valve position indi- closed control - charging available charging return cation (closed to flow flow paths to the RCS. open position change) BG HV-8146 and BG HV-8147 at CB; charging flow are both charging returns indication (TI-126) to the RCS. BG HV-8147 can at CB; regenerative be placed in-service, if heat exchanger shell required. No effect on CVCS side exit temperature operations during normal indication (TI-127) plant operation, load and high temperature following or safe shutdown alarm at CB; and char- operation. ging and seal water flow indication (FI-121A) and low flow alarm at CB. Rev. 17 WOLF CREEK TABLE 9.3-10 (Sheet 18) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks32. Air diaphragm- a. Fails Charging and volume Isolates one of the two Valve position indi- Same remark as that operated globe closed control - charging available charging return cation (closed to stated for item 4 valve 8147 flow flow paths to the RCS. open position in regard to BG HV-8146 and BG HV-8147 change) at CB. design of valve. are both charging returns to the RCS. BG HV-8146 can be placed in-service, if required. No effect on CVCS operations during normal plant operation, load following or safe shutdown operation. b. Fails Charging and volume Failure has no effect on Valve position indi- open control - charging CVCS operation during normal cation (open to closed flow plant operation, load follow position change) at and safe shutdown operation. CB. BG HV-8146 can be removed from service, if required, since only one return path is required. In the event that auxiliary spray is being used to cooldown the PZR, the charging return flow path is isolated to maximize auxiliary spray flow. Cold shutdown of the reactor is still possible; however, time for cooling down the PZR will be extended.
- 33. Air diaphragm- a. Fails open Charging and volume Failure results in inadver- Valve position indi- Same remark as that operated globe control - charging tent operation of auxili- cation (open to stated for ite m 7 valve 8145 flow ary spray that results in closed position in regard to
a reduction of PRZ pres- change) at CB and design of valv
- e. sure during normal plant PRZ pressure record-
operation and load follow. ing (PR-455) and
PRZ heaters operate to low pressure alarm
maintain required PRZ pres- at CB.
sure. Boration of RCS to
a safe shutdown concen-
tration level and makeup of coolant during opera-tion to bring reactor to a
safe shutdown condi-
tion is still possible.
- b. Fails Charging and volume Failure has no effect Valve position indi-closed control - charging on CVCS operation during cation (closed to
flow normal plant operation, open position change)
load follow, and safe at CB.
shutdown operation. Valve
is used during cold shut-down operation to active auxiliary spray for cool-
ing down the PRZ after
operation of RHRS.
- 34. Relief valve Fails open Charging and volume Failure results in a por- Local pressure indi- Radioactive fl uid 8123 control - charging tion of seal water return cation (PI-118 and contained.
flow flow and centrifugal PI-119) in discharge
charging pump miniflow line of centrifugal
being bypassed to VCT. charging pumps.
Boration of RCS to a safe shutdown concentration level and makeup of cool-
ant during operations to
bring reactor to safe
shutdown condition is
still possible.
Rev. 17 WOLF CREEK TABLE 9.3-10 (Sheet 19) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks35. Motor operated Fails open Charging & volume Failure results in portion Local charging Normal pump globe valve control - charging of charging and seal water and seal water may be isolated HV-8109 flow and seal water flow from normal charging flow indication by closing of m anual isolation flow pump being bypassed to (FI-121A) and valves in pump the seal water heat pressure discharge and exchanger. No effect on indication suction lines. normal plant operation (PI-186 & PI-463) load follow, or bringing reactor to safe shutdown condition. Normal charging pump may be taken out of service, and an alternate centrifugal charging pump used for delivery of charging and seal water flow. Fails Closed Failure results in loss of Same as that Same as that minimum recirculation flow for fail open for fail open for normal charging pump, case case loss of pump protection for abnormal operating conditions. No effect on normal plant operation, load follow, or bringing reactor to safe shutdown condition. Normal charging pump may be taken out of service, and an alternate centrifugal charging pump used for delivery of charging and seal water flow.
- 36. Air diaphragm- a. Fails open Charging and volume Failure reduces redundancy Charging and seal 1. Same remar k as operated globe control - charging of providing charging and water flow indica- that state d for valve FCV-121 flow and seal water seal water flow to RCS. tion (FI-121A) and item 4 in re- (FCV-462 flow No effect on normal plant high flow alarm at gard to de sign analogous) operation, load follow, CB, and PRZ level of valve.
or bringing reactor to recording (LR-459) safe shutdown condition. and high level 2. Methods of Normal charging alarm at CB. detection apply pump normally used for when a sta ndby delivery of charging and centrifuga l seal water flow to RCS. charging p ump Check valves 8481A and is in oper ation 8481B provide isolation
of normal charging
pump flow to discharge of
centrifugal charging pump if valve fails "open" dur-
ing operation of normal charging pump.
- b. Fails Charging and volume Failure reduces redundancy Charging and seal
closed control - charging of providing charging and water flow indication
flow and seal water seal water flow to RCS. (FI-121A) and low
flow flow alarm at CB, and PRZ level recording (LR-459) Rev. 10 WOLF CREEK TABLE 9.3-10 (Sheet 20) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks No effect on and low level alarm system operation at CB.
during normal plant
operation load follow, or
bringing reactor to safe
shutdown condition.
Normal charging pump
normally used for delivery of charging and seal water flow to RCS. 36a. Motor- Fails Charging and volume Failure reduces redun- Valve position in-operated closed control - alternate dancy of providing seal dication (ZI-8357A, globe valve seal water flow water flow during accident B) at CB; seal water
8357A (8357B conditions. No effect flow indication
analogous) on safety for system (FK-215A, B) at CB;
operation. Seal water seal water return
flow under an accident recording (FR-157); condition is provided and low seal water by alternate flow path return flow alarm
through valve 8357B (FAL-154).
(8357A).
- 37. Check valve Fails open Charging and volume Failure reduces redun- Charging and seal 1. Normal 8497 control - charging dancy of providing water flow indica- charging flow and seal water charging and seal water tion (FI-121A) and may be iso lated flow to RCS. Discharge of low flow alarm at by the clo sing normal charging CB, and PRZ level of manual valves pump remains open to recording (LR-459) in pump's suc- "back-flow" when a and low level tion and d is- centrifugal charging alarm at CB. charge lin es. pump is placed into operation. No effect 2. Methods of
on normal plant operation, detection apply load follow, or bringing when centr ifu- reactor to safe shutdown gal chargi ng condition; normal pump 1 is in charging pump normally operation. used for delivery of charging and seal water flow.
- 38. Check valve Fails open Charging and volume Failure reduces redun- Same methods of 1. Centrifuga l 8481A control - charging dancy of providing detection as those charging p ump 1 (8481B flow and seal water charging and seal water stated for item may be iso lated analogous) flow flow to RCS. Discharge 37. by the clo sing of centrifugal charging of manual valves pump 1 is open to "back- in pump's suc- flow" when centrifugal tion and d is- charging pump 2 is charge lin es. placed into operation after failure of centri- 2. Method s of fugal charging pump 1 detection apply to deliver charging and when centr ifu- seal water flow. No gal chargi ng effect on normal plant pump 2 is in operation, load follow, operation
. or bringing reactor to
safe shutdown condi-
tion; normal charging pump nor-mally used for delivery
of charging and seal
water flow.
Rev. 1 0 WOLF CREEK TABLE 9.3-10 (Sheet 21) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- 39. Normal Fails to Charging and volume Failure reduces redun- Pump circuit breaker Flow rate is c ontrolled charging pump deliver control - charging dancy of providing position indication by a modulatin g valve working flow and seal water charging and seal water (open) at CB; common (FCV-462) in d ischarge fluid flow flow to RCS. No effect pump breaker trip header for the normal on normal plant operation, alarm at CB; charg- charging pump.
load follow, or bringing ing and seal water reactor to safe shutdown flow indication (FI-condition. Centrifugal 121A) and low flow
charging pumps (1 and 2) alarm at CB; and are placed into PRZ level recording operation for delivery (LR-459) and low of charging and seal water level alarm at CB.
flow.
- 40. Centrifugal Fails to Charging and volume Failure reduces redun- Same methods of Flow rate for a charging pump, deliver control - charging dancy of providing detection as those centrifugal ch arg- 1, APCH (pump working flow and seal water charging and seal water stated for item 39 ing pump is co n- 2 analogous) fluid flow flow to RCS. Alternate when centrifugal trolled by a delivery of charging and charging pump 1 modulating val ve seal water flow by redun- is in operation. (FCV-121) in d is- dant certrifugal charging charge header for pump 2 (pump 1) is avail- the centrifuga l able. No effect on normal charging pumps
. plant operation, load fol-
low or bringing reactor to
safe shutdown condition.
Normal charging pump normally used for delivery
of charging and seal water
flow.
- 41. Air diaphragm- Fails Chemical control, Failure blocks hydrogen VCT pressure indi- 1. Valve is operated globe closed purification, and flow to VCT and leads to cation (PI-115) and designed t o valve 8156 makeup - oxygen loss of venting of VCT low pressure alarm fail "clos ed." control (vent line PCV-115 closes at CB. Periodic
on low VCT pressure), sampling of gas 2. Plant Tech nical resulting in loss of gas mixture in VCT. Specificat ions stripping of fission pro- set limits on ducts from RCS coolant. RCS activi ty No effect on operation to level.
bring the reactor to safe
shutdown condition.
- 42. Relief valve Fails open Charging and volume Failure allows VCT liquid Decrease in VCT Radioactive fl uid 8120 control - charging to be released to BRS level, causing RMCS contained.
flow and seal water recycle holdup tank, re- to operate; VCT
flow sulting in a loss of VCT level indication Rev. 10 WOLF CREEK TABLE 9.3-10 (Sheet 22) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks liquid and makeup coolant (LI-185) and low
available for charging and level alarm at CB;
seal water flow during and BRS recycle
normal plant operation, holdup tank level
load follow, and brining increase.
the reactor to a safe
shutdown condition. VCT isolation valves 112B and LCV-112C close on low-low
tank level signal, causing
the suction of charging
pumps to be transferred
to the RWST for an alternate
supply of borated coolant.
- 43. Motor-operated a. Fails open Charging and volume Failure has no effect on Valve position indi- During normal plant gate valve control - charging CVCS operation during nor- cation (open to operation and load LCV-112B flow and seal water mal plant operation, load closed position follow, valve is at (LCV-112C flow follow, and bringing re- change) at CB. a full open po si- analogous) actor to a safe shutdown tion and motor oper- condition. However, ator is energi zed under accident conditions to close the v alve requiring isolation of VCT, upon the gener ation failure reduces redundancy of a VCT low-l ow of providing isolation for level signal o r upon discharge line of VCT. the generation of safety injecti on "S" signal.
- b. Fails Charging and volume Failure blocks fluid flow Valve position indi-closed control - charging from VCT during normal cation (closed to
flow and seal water plant operation, load open position change)
flow follow, and when bringing at CB; group moni-
the reactor to a safe toring light and
shutdown condition. Alter- alarm (valve closed)
nate supply of borated at CB; charging and
coolant from the RWST to seal water flow indi-suction of charging pumps cation (FI-121A) and
can be established from low flow alarm at CB; the CB by the operator and PRZ level record-through the opening of ing (LR-459) and low
RWST isolation valves level alarm at CB.
LCV-112D and LCV-112E. Rev. 8 WOLF CREEK TABLE 9.3-10 (Sheet 23) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- 44. Air diaphragm- Fails closed Chemical control, Failure blocks venting of Valve position indi- 1. Same remar k as operated dia- purification, and VCT gas mixture to gas cation (closed to that state d for phragm valve makeup - oxygen waste processing sytem open position item 7 in re- PCV-115 control (waste gas compressors) change) at CB and gard to va lve for stripping of fission VCT pressure indica- design.
products from RCS coolant tion (PI-115) at CB.
during normal plant opera- Periodic sampling of 2. Same remar k as tion and load follow. gas mixture in VCT. that state d for No effect on operations to item 41 in re- bring the reactor to safe gard to RC S shutdown condition. activity.
- 45. Air diaphragm- Fails closed Boron concentration Failure blocks fluid flow Valve position indi- Same remark as that operated dia- control - reactor from reactor makeup con- cation (closed to stated for ite m 7 phragm valve makeup control, bora- trol system for automatic open position in regard to v alve FCV-110B tion, automatic make- boric acid addition and change) at CB; total design.
up, and alternate reactor water makeup dur- makeup flow devia-dilution. ing normal plant operation tion alarm at CB; and load follow. Failure and VCT level indi-
also reduces redundancy of cation (LI-185) and
fluid flow paths for dilu- low level alarm at
tion of RC by reactor CB.
makeup water and blocks
fluid flow for boration of
the RC when bringing the
reactor to a safe shutdown
condition. Boration (at BA tank boron concentration level) of RCS coolant to
bring the reactor to safe
shutdown condition may be
possible by opening alter-
nate BA tank isolation
valve 8104 at CB.
- b. Fails open Boron concentration Failure allows for alter- Valve position indi-control - reactor nate dilute mode type cation (open to
makeup control, operation for system op- closed position
boration, automatic eration of normal dilution change) at CB. makeup, and alter- of RCS coolant. No effect nate dilution on CVCS operation during
normal plant operation and
load follow and when
bringing the reactor to a
safe shutdown condition. Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 24) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks
- 46. Air diaphragm- a. Fails Boron concentration Failure blocks fluid flow Same methods of Same remark as that operated dia- closed control - reactor from RMCS for dilution detection as those stated for ite m 7 phragm valve makeup control, of RCS coolant during stated for item in regard to v alve FCV-111B dilution, and alter- normal plant operation 45.a. design.
nate dilution and load follow. No
effect on CVCS operation.
Operator can dilute RCS coolant by establishing "alternate dilute" mode
of system operation.
Dilution of RCS coolant
not required when bringing
the reactor to a safe shut-
down condition.
- b. Fails open Boron concentration Failure allows for alter- Valve position indi-control - reactor nate dilute mode type cation (open to makeup control, operation for system op- closed position dilution, and alter- eration of boration and change) at CB.
nate dilution automatic makeup of RCS
coolant. No effect on
CVCS operation during
normal plant operation
and load follow and when
bringing the reactor to
a safe shutdown operation.
- 47. Relief valve Fails open Charging and volume Failure allows for a por- Decrease in VCT Radioactive fl uid 8124 control - charging tion of flow to suction level, causing RMCS contained.
and seal water flow header of charging pumps to operate; VCT
to be relieved to BRS level indication
recycle holdup tank. (LI-185) and low Boration of RCS coolant level alarm at CB; to bring reactor to safe and BRS recycle
shutdown condition is holdup tank level
still possible. increase.
- 48. Air diaphragm- a. Fails open Boron concentration Failure prevents the addi- Valve position indi- Same remark as that operated globe control - reactor tion of a preselected cation (open to stated for ite m 4 valve FCV-110A makeup control, quantity of concentrated closed position in regard to v alve boration, and auto boric acid solution at change) at CB; and design.
matic makeup a preselected flow rate BA flow recording Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 25) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks to the RCS coolant during (FR-110) and flow normal plant operation, deviation alarm at
load follow, and when CB. bringing the reactor to a safe shutdown condition.
Boration to bring the re-
actor to a safe shutdown condition is possible; however, flow rate of
solution from BA tanks
cannot be automatically
controlled.
- b. Fails Boron concentration Failure blocks fluid flow Valve position indi-closed control - reactor of BA solution from BA cation (closed to
makeup control, tanks during normal plant open position change) boration, and auto- operation, load follow, at CB; and BA flow matic makeup and when bringing the recording (FR-110) reactor to a safe shutdown and flow deviation
condition. Boration (at alarm at CB.
BA tank boron concentra-
tion level) of RCS cool-
ant to bring the reactor
to safe shutdown condition
may be possible by opening
of alternate BA tank isola-
tion valve 8104 at CB.
- 49. Air diaphragm- a. Fails Boron concentration Failure blocks fluid flow Valve position indi- Same remark as that operated globe closed control - reactor of water from RMCS during cation (closed to stated for ite m 7 valve FCV-111A makeup control, normal plant operation and open position in regard to v alve dilute, alternate load follow. No effect change) at CB; VCT design.
dilute, and auto- on system operation when level indication matic makeup bringing the reactor to (LI-185) and low
a safe shutdown condition. level alarm at CB;
and makeup water
flow recording (FR-
110) and flow devia-
tion alarm at CB.
- b. Fails open Boron concentration Failure prevents the addi- Valve position indi-control - reactor tion of a preselected cation (open to
makeup control, quantity of water makeup closed position
dilute, alternate at a preselected flow change) at CB and
dilute, and auto- rate to the RCS coolant makeup water flow
matic makeup during normal plant opera- recording (FR-110) Rev. 0 WOLF CREEK TABLE 9.3-10 (Sheet 26) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks tion and load follow. No and flow deviation effect on system operation alarm at CB.
when bringing the reactor to a safe shutdown condition.
- 50. Motor-operated a. Fails Boron concentration Failure reduces redundancy Valve position indi- 1. Valve is a t a globe valve closed control - reactor of flow paths for supply- cation (closed to closed pos ition 8104 makeup control, bora- ing BA solution from BA open position during nor mal tion, and automatic tanks to RCS via charging change) at CB and RMCS opera tion. makeup pumps. No effect on CVCS flow indication operation during normal (FI-183A) at CB. 2. If both fl ow plant operation, load paths from BA follow, or safe shutdown tanks are
operation. Normal flow blocked du e to path via RMCS may be failure of iso- available for boration lation val ves of RCS coolant. FCV-110A a nd 8104, bora ted water from RWST is availab le by opening is o- lation val ve LCV-112D o r LCV-112E.
- b. Fails open Boron concentration Failure prevents the addi- Valve position indi-control - reactor tion of a preselected cation (open to
makeup control, bora- quantity of concentrated closed position tion, and automatic BA solution at a pre- change) at CB and makeup selected flow rate to the flow indication (FI-
RCS coolant during normal 183A) at CB.
plant operation, load
follow, and when bringing
the reactor to a safe
shutdown condition. Bora-
tion to bring the reactor
to a safe shutdown condition
is possible; however, flow
rate of solution from BA tanks cannot be automati-cally controlled.
- 51. BA transfer Fails to Boron concentration No effect on CVCS opera- Pump motor start Both BA transf er pump 1, APBA deliver work- control - reactor tion during normal plant relay position indi- pumps operate (pump 2 ing fluid makeup control, operation, load follow, cation (open) at CB simultaneously for analogous) boration, and auto- or bringing reactor to safe and local pump dis- RMCS boration
matic makeup shutdown condition. Redun- charge pressure operation. Rev. 8 WOLF CREEK TABLE 9.3-10 (Sheet 27) CVCS Operation Effect on System Operation Failure Component Failure Mode Function and Shutdown* Detection Method** Remarks dant BA transfer pump 2 indication (PI-113).
provides necessary
delivery of working fluid
for CVCS operation.
- 52. Air diaphragm- Fails open Charging and volume Failure bypasses normal Valve position indi- Valve is desig ned operated three for flow control - letdown letdown flow to BRS re- cation (holdup tank) to fail open f or way valve only to BRS flow cycle holdup tank, result- at CB; VCT water flow to VCT an d is LCV-112A recycle hold- ing in excessive use of level indication electrically w ired up tank RMCS. No effect on opera- (LI-185) and low so that electr ical tion to bring reactor to level alarm at CB; control soleno ids safe shutdown condition. and increase water for valve are level in BRS recycle energized for flow holdup tank. to BRS recycle
holdup tank. Valve opens to flow to BRS
recycle holdup
tank on high V CT water level si gnal.List of acronyms and abbreviations BA - Boric acid BRS - Boron recycle system BTR - Boron thermal regeneration
BTRS - Boron thermal regeneration system
CB - Control board
CVCS - Chemical and volume control system
MCC - Motor control center
PRZ - Pressurizer RC - Reactor coolant RCS - Reactor coolant system
RHRS - Residual heat removal system
RWST - Refueling water storage tank
RMCS - Reactor makeup control system VCT - Volume control tank NOTE: Portions of the CVCS are relied upon to perform as part of the safety-grade cold shutdown designs; therefore, see Section 5.4.7 and Section 7.4 for further discussions. Rev. 14 WOLF CREEK TABLE 9.3-11 SERVICE GAS REQUIREMENTS Component Serviced Service Gas with Nitrogen FunctionSafety injection Cover gas, source of accumulator tanks potential energy Pressurizer relief Cover gas tankVolume control tank Purge gas (during shutdown) Spent resin tanks Sluice spent resins to solid radwaste system Gas decay tanks Maintenance during shutdown Spray additive tank Cover gas Feedwater heaters Purge and cover gas during layup Steam generator Purge and cover gas (shell side) during layup Auxiliary steam generator Purge and cover during and reboiler layup Chilled water Cover gas expansion tank Chemical addition tanks Cover gas Electrical penetration Testing assembliesSteam generator blowdown Chemical mixing in steam System generators Hydrogen recombiners Purge gas Hydrogen recombiners' gas Instrument calibration analyzer racks (GAR) and purge Back up compressed gas Source of potential energy system accumulators Condensate Storage Tank Purge gas to reduced dissolved oxygen concentration Rev. 24 WOLF CREEK TABLE 9.3-11 (Sheet 2) Component Serviced Service Gas with Hydrogen FunctionMain generator Cooling medium for generator field Volume control tank Recombine free oxygen and stripping agent Reactor coolant drain Cover gas (from two 194 SCF Tank local cylinders outside containment) Gaseous radwaste system Testing (from a portable hydrogen recombiners 20 SCF storage cylinder) Component Serviced Service Gas with Carbon Dioxide Function Main generator gas Atmospheric and hydrogen System purge Component Serviced Service Gas with Oxygen Function Gaseous radwaste system Recombination with free H2 from hydrogen recombiners volume control tank and other miscellaneous sources Gaseous radwaste system Instrument supply hydrogen recombiners' gas analyzer racks (GAR) Rev. 3 WOLF CREEK TABLE 9.3-12 LABORATORY GAS REQUIREMENTS Count Room Radwaste Cold Type and Hot Lab Lab Lab Argon Yes No No
Propane Yes No Yes
Oxygen No Yes No
Hydrogen Yes No No
Nitrous oxide Yes No No
P-10 Yes No No
Acetylene Yes No No
Nitrogen Yes No No
Helium Yes No No Bottle size - less than 300 pounds of gas. Small cylinders of oxygen and hydrogen/nitrogen mix are located on the hydrogen analyzer skid on elevation 2047' in the auxiliary building (rooms 1505 and 1506). Rev. 13 WOLFCREEKTABLE9.3-13BORONRECYCLESYSTEMPRINCIPALCOMPONENTDATA
SUMMARY
RecycleEvaporatorFeedPumpsNumber2Designpressure,psig150 Designtemperature,F250 Designflow,gpm30/100 Designhead,ft250/200 MaterialAusteniticstainless steelDesigncode(1)MSRecycleHoldupTanksNumber2Capacity,usable,gal60,700 DesignpressureAtmospheric Designtemperature,F200 MaterialAusteniticstainless steelDesigncode(1)API650RecycleEvaporatorReagentTankNote3Number1Capacity,gal5 Designpressure,psig150 Designtemperature,F200 MaterialAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1RecycleEvaporatorFeedDemineralizersNumber2 Designpressure,psig300Designtemperature,F250Designflow,gpm120Resinvolume,ft 339max.MaterialAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1RecycleEvaporatorCondensateDemineralizerNote3Number1Designpressure,psig300Rev.14 WOLFCREEKTABLE9.3-13(Sheet2)Designtemperature,F250Designflow,gpm120 Resinvolume,ft 339max.MaterialAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1RecycleEvaporatorFeedFilter(FHE05)Number1Designpressure,psig300Designtemperature,F250 Designflow,gpm250Particleretention(SeeNote2Below)Material,vesselAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1RecycleEvaporatorCondensateFilter(FHE06)*Note3Number1Designpressure,psig200 Designtemperature,F250Designflow,gpm35Particleretention(SeeNote2Below)micronsize(max)Material,vesselAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1RecycleEvaporatorConcentratesFilter(FHE04)*Note3Number1Designpressure,psig200 Designtemperature,F250Designflow,gpm35Particleretention(SeeNote2Below)micronsize(max)Material,vesselAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1
- Standardfiltercartridgesareavailablewithvariableparticleretentioncharacteristics,andtheselectionofthefilter cartridgeisbasedonoperatingdataandindustryexperience.Rev.14 WOLFCREEKTABLE9.3-13(Sheet3)RecycleEvaporatorPackageNote3Number1Designflow,gpm15 Concentrationofconcentrate,boricacid,wtpercent4Concentrationofcondensate<10ppmboronasH3BO3MaterialAusteniticstainless steelDesigncode(1)ASMEVIII,Div.1;MS;TEMA-R;ANSIB31.1RecyleHoldupTankVentEductorNumber1 Designpressure,psig150Designtemperature,F200Suctionflow,scfm1+0.2Motiveflow,scfm40MaterialCarbonsteel Designcode(1)MSNote1-Tableindicatestherequiredcodebasedonitssafety-relatedimportanceasdictatedbyserviceandfunctionalrequirementsandbytheconsequencesof theirfailure.Notethattheequipmentmaybesupplied toahigherprincipalconstructioncodethanrequired.Note2-Theselectionoffiltercartridgeandparticleretentioncharacteristicsisbasedontheflowrate,filterfunction,operatingdataandindustryexperience.Note3-Therecycleevaporatorandrelatedequipmentarepermanentlyoutof service.Rev.14 RM>W/>..SlE.
DRW C:.RW uJ ;z §o. 0:: u .&:0 > c:£ b ;:I: FUE.L e,uiLDINq DRW C.RW a WOLF CREEK DIE.SE.L CONTROL &UILDING ow DRW ow DRW CRW LRW ow I I DRW go! u-_a; 2':c£ :JO :t<J :I: 3 A.UXILIJ\R"t' BOILE.R ROOM ow LRW WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Rev. 0 FIGURE 9.3-6 MAJOR DRAINAGE AREAS {SHEET 1) 0 RHR PUMP WOLF CREEK 0 RI-IR I'UHP Rev. 0 tRW '\\_&oRON INJECTION PUHP& WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 9.3-6 FLOOR DRAIN FOR SAFETY-RELATED ROOMS AUX. BUILDING BASEMENT (SHEET 2) 02 0 L.P. N2 --r--*--ItA' ...__ t<<X.E' CREEK 0 0 @ ...... ,..... C)/ -., 0 H.P. Building Cantral end CommuniAtlon Turbine DleMI Gener1tor full Rld-tl @ for H 2 G* Supply to R.C. f'=' B L1b, G11 c Hydrogen Analyzer Calibration Gas CEI. 2047'-0"))
* * "2'co2 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 9.3-10, REV. 13 GAS SUPPLY LOCATION/INTERFACE
__ WOLF CREEK 9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION The following sections provide the design bases, descriptions and evaluations of the HVAC systems for each building. Section 3.11(B) provides a summary of the environmental conditions that result from the systems described herein. Table 9.4-1 provides the design outside ambient conditions.
9.4.1 CONTROL
BUILDING HVAC The control building HVAC systems consist of the control building supply and
exhaust systems, the control room, Class IE electrical equipment and access
control air-conditioning systems, the access control exhaust system, and the
counting room recirculation system. A local air handling unit serves the Secondary Alarm Station (SAS) room.
The control building supply system provides conditioned outside air for ventilation and cooling to each level of the control building. The control
building exhaust system provides a means of normal exhaust and of purging smoke following a postulated fire from the clean areas (radiation Zone A areas) of the control building.
The control room air-conditioning system, including the control room filtration
system, the control room pressurization system, and the SAS room air handling unit provide a suitable atmosphere for personnel and equipment within the control room.
The Class IE electrical equipment air-conditioning system provides a suitable
environment for the Class IE electrical equipment. The access control air-conditioning system provides a suitable environment for personnel comfort. The access control exhaust system exhausts the potentially
contaminated areas of the access control area and provides a means of purging
smoke following a postulated fire. The counting room recirculation system and the SAS room air conditioning system
provide a suitable environment for personnel and equipment located in the
counting room and the SAS room respectively.
9.4.1.1 Design Bases 9.4.1.1.1 Safety Design Bases
The control room air-conditioning system, the Class IE air-conditioning system, and portions of the control building supply, control building exhaust, and the access control exhaust systems
9.4-1 Rev. 0 WOLF CREEK are safety related and are required to function following a DBA and to achieve and maintain the plant in a post accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The control room air-conditioning system, the Class IE electrical equipment air-conditioning system, and the control building
isolation provisions are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). SAFETY DESIGN BASIS TWO - The control room air-conditioning system, the Class IE electrical equipment air-conditioning system, and the provisions for control
building isolation are designed to remain functional after an SSE and to
perform their intended functions following a postulated hazard, such as internal missiles, or pipe break (GDC-4). SAFETY DESIGN BASIS THREE - Safety functions of the control building HVAC
systems can be performed, assuming a single active component failure coincident
with the loss of offsite power.
SAFETY DESIGN BASIS FOUR - Active components of the control building HVAC systems are capable of being tested during plant operation. Provisions are
made to allow for inservice inspection of the ASME Section III components of
the safety-related air-conditioning units at appropriate times specified in the
ASME Boiler and Pressure Vessel Code, Section XI. SAFETY DESIGN BASIS FIVE - The control room air-conditioning system, the Class
IE electrical equipment air-conditioning system, and the safety-related control
building isolation provisions are 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 all nonsafety-related HVAC
system penetrations of the control building boundary is provided so that the occupation and habitability of the control room, as discussed in Section 6.4, is not compromised (GDC-2 and 19).
SAFETY DESIGN BASIS SEVEN - The control room air-conditioning system provides
the control room with a conditioned atmosphere during all modes of plant operation, including post-accident operation (GDC-19). The control room filtration system and the control room pressurization system charcoal adsorbers
comply with Regulatory Guide 1.52 to the extent discussed in Table 9.4-2.
9.4-2 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS EIGHT - The Class IE electrical equipment air-conditioning system provides a suitable atmosphere for the Class IE electrical switchgear
during all modes of plant operation, including loss of preferred power and post-accident operation. 9.4.1.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The control building supply system provides the necessary outside air needed for the required cooling and ventilating of the cable spreading rooms. The control building supply system also provides
ventilation and supplemental cooling for each of the other levels of the
control building. The control building ventilation system is designed to
provide fresh air ventilation at a minimum rate of 0.1 cfm per square foot of floor area. POWER GENERATION DESIGN BASIS TWO - The control building exhaust system serves
to remove from the control building the hydrogen generated by the batteries
during normal operation. POWER GENERATION DESIGN BASIS THREE - The access control air-conditioning
system provides the first aid room, the HP office, the sign-in/out areas of the
access control area, and the nonvital electric equipment areas of the
electrical and mechanical equipment level with an environment suitable for personnel comfort and electrical equipment operation.
POWER GENERATION DESIGN BASIS FOUR - The access control exhaust system collects
and processes the effluents from the potentially contaminated regions of the
access control area. The exhaust system is designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the as-low-as-reasonably achievable dose objective of 10 CFR 50, Appendix I. The access control exhaust system charcoal adsorption train complies with Regulatory Guide 1.140, to the
extent discussed in Table 9.4-3.
POWER GENERATION DESIGN BASIS FIVE - The counting room recirculation system provides adequate cooling, humidity control, and filtering of the counting room
environment for personnel and equipment.
POWER GENERATION DESIGN BASIS SIX - The SAS Room air handling unit provides the SAS room with an environment suitable for personnel comfort and electrical equipment operation.
9.4-3 Rev. 19 WOLF CREEK 9.4.1.2 System Description
9.4.1.2.1 General Description The control building HVAC systems are shown in Figure 9.4-1. The systems consist of the control building supply system, control room air-conditioning
system with supplemental filtration and pressurization systems and SAS Room air
handling unit, Class IE electrical equipment air-conditioning system, access control air conditioning system, counting room recirculation system, control building exhaust system, and the access control exhaust system. The design
conditions for these systems are presented in Table 3.11(B)-1. Potential
radiation doses in the control room are discussed in Chapter 15.0.
The control building is serviced by an outside-air-supply system which provides fresh cooled or heated air to each of the various levels of the building.
Self-contained air-conditioning units serve the control room elevation and the Class IE electrical equipment floors. Local fan-coil units serve the access
control floor and nonvital areas of the electrical and mechanical equipment level and the counting room and the SAS room. All outside air intakes, both essential and nonessential, are provided with
labyrinth missile barriers. The barriers are designed to withstand and absorb
missile impacts and to prevent the propagation of a missile trajectory in line with essential equipment.
Two exhaust systems also service the building. The control building exhaust
system takes suction from the clean areas of the building, and the access
control exhaust system takes suction from the potentially contaminated areas of the access control floor. The control building exhaust system discharges directly to the atmosphere, while the access control exhaust system processes
the exhaust air through charcoal adsorbers prior to discharging through the
unit vent.
Based on the source terms provided in Section 11.1 and the dose evaluation provided in Section 11.3, the access control exhaust system meets the objective of 10 CFR 50, Appendix I, and the limits of 10 CFR 20.
9.4.1.2.2 Component Description Codes and standards applicable to the control building HVAC systems are listed
in Tables 3.2-1 and 9.4-4. The control room air-conditioning system, including
the control room filtration and pressurization systems, the Class IE air-
conditioning system, and
9.4-4 Rev. 19 WOLF CREEK safety-related HVAC penetrations of the control building boundaries are designed and constructed in accordance with codes and standards comparable with
quality group C. The control room ac system coils and condenser and the Class IE electrical equipment ac system coils and condensers are designed and constructed in accordance with quality group C.
NONESSENTIAL AIR HANDLING UNITS - Those nonessential air handling units which
make up a part of the control building HVAC system are the control building supply air unit, access control fan coil unit, and the counting room fan coil unit and the SAS Room air handling unit.
The control building supply air unit consists of a particulate filter, hot-
water heating coil, chilled-water cooling coil, centrifugal fan, and electric motor driver.
The access control fan coil unit consists of a particulate filter, chilled-water cooling coil, centrifugal fan, and electric motor driver.
The counting room fan-coil unit consists of an inlet filter module, a chilled-water cooling coil module, a fan module, a diffuser module, and an HEPA filter module. The SAS Room air handling unit consists of a moderate efficiency prefilter, direct expansion cooling coil, centrifugal fan and motor driver. The SAS Room condensing unit consists of a compressor, condensing coil and condenser fan. SAFETY-RELATED AIR HANDLING UNITS - The control building HVAC system contains
two safety-related air handling units, the control room air-conditioning unit, and the Class IE electrical equipment air-conditioning unit.
Both the control room air-conditioning unit and the Class IE electrical
equipment air-conditioning unit consist of 85% efficiency filters, a self-
contained refrigeration system utilizing essential service water as the heat sink, centrifugal fans, and electric motor drivers.
NONESSENTIAL FILTER UNITS - The control building HVAC system contains two
nonessential filter units, the access control exhaust filter adsorber unit, and the counting room filter unit. The access control exhaust filter adsorber unit consists of moderate efficiency
prefilters, HEPA filters, and charcoal adsorption beds.
The counting room filter unit consists of moderate efficiency prefilters and HEPA filters.
9.4-5 Rev. 20 WOLF CREEK SAFETY-RELATED FILTER UNITS - Those safety-related filter units which are a part of the control building HVAC system are the control room filtration system
filter adsorber units and the control room pressurization system filter adsorber units. Each control room filtration system filter adsorber unit consists of moderate
efficiency prefilters, HEPA filters, and charcoal adsorption beds.
Each control room pressurization system filter adsorber unit consists of a demister, electric heater, HEPA filters, and charcoal adsorption beds.
NONESSENTIAL FANS - There are two pairs of nonessential fans in the control
building HVAC system -- the access control exhaust fans and the control building exhaust fans.
The access control exhaust fans are centrifugal fans with an electric motor driver. The control building exhaust fans are vaneaxial fans with an electric motor driver.
SAFETY-RELATED FANS - The control building HVAC system contains two pairs of
safety-related fans, the control room filtration fans, and the control room pressurization fans. Both the control room filtration system fans and the control room
pressurization system fans are centrifugal fans with electric motor drivers.
SUPPLEMENTAL HEATER - Supplemental heating is provided by nonessential electric duct heaters and electric unit heaters.
Electric duct heaters supplement the heating of the control room, access
control area, the HVAC equipment room (El. 2,016), and the nonvital areas of the dc battery and switchgear area.
Electric unit heaters supplement the heating of the upper and lower cable
spreading rooms, the ESF switchgear rooms, the pipe chase/tank area, and the
control room air-conditioning equipment room. Each unit heater consists of a coil and a fan with an electric motor driver. FIRE DAMPERS - Fire dampers are located between fire barriers, as necessary, to
maintain the fire ratings of the barriers. Dampers are the 3-hour-rated
curtain type.
9.4-6 Rev. 13 WOLF CREEK ISOLATION DAMPERS - Where a means of system isolation is required, parallel-blade-type dampers are utilized. The type of operator employed is dependent
upon the specific design and/or usage requirements. FLOW CONTROL DAMPERS - Single-blade-type or opposed-blade-type dampers are
utilized, as necessary, to provide a means of system balancing. In general, these are manually operated. However, some utilize power operators to allow
compensation for changes occurring during system operation. BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to maintain
the proper direction of flow.
TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects of extreme wind or tornado conditions is required. These dampers close with the flow produced by the differential pressure associated with tornadoes or
high winds.
9.4.1.2.3 System Operation GENERAL - The control building is serviced by an outside air supply system
which provides fresh cooled or heated air to each of the various levels of the
building. Self-contained air-conditioning units serve the control room
elevation and the Class IE electrical equipment floors. Local fan-coil units serve the access control floor and the nonvital areas of the electrical and mechanical equipment level, the counting room and the SAS Room.
Two exhaust systems also service the building. The control building exhaust
system takes suction from the clean areas of the building, and the access control exhaust system takes suction from the potentially contaminated areas of the access control floor and the basement beneath. The control building
exhaust system discharges directly to the atmosphere while the access control
exhaust system processes the exhaust air through a charcoal adsorber train
prior to discharging through the unit vent. Cooling water for the nonessential units is supplied by the central chilled
water system (Section 9.4.10), and cooling water for the safety-related units
is supplied by the essential service water system (Section 9.2.1). Hot water
for the control building supply air unit is supplied by the plant heating system (Section 9.4.9).
The control room has two sources of outside air. The normal source of outside
air is provided by the control building supply air unit which draws from the
intake plenum located on top of the auxiliary building. This intake plenum is identified as an HVAC
9.4-7 Rev. 27 WOLF CREEK penthouse located between building column lines A-3 and A-1, and A-J and A-H on USAR Figure 1.2-14. The emergency source of outside air is provided by the
control room pressurization system which draws air from intake louvers located on the west wall of the control building. These louvers are shown on USAR Figures 1.2-24 (grid H-4) and 1.2-28 (grid E-3).
The relative locations of all power block buildings and the location of the
radiation release points are shown on USAR Figure 1.2-1. USAR Figure 11.3-2 identifies the release points of potentially radioactive gaseous effluents.
The calculated total control room leakage with a 1/4-inch w.g. differential
pressure is 80 cfm, of which approximately 95 percent is attributable to the
doors. The remaining leakage paths are both individually and collectively insignificant in terms of total control room leakage.
Nonetheless, the following specific criteria were included in the control room isolation damper procurement specification to ensure that the required leak-
tightness is provided:
- a. For dampers with a surface area equal to or greater than
2 ft 2 , the maximum allowable leakage at a pressure differential of 6 inches w.g. is 20 cfm/ft
- 2. b. For dampers with a surface area of less than 2 ft 2 , but greater than 1 ft 2 , the maximum allowable leakage at a differential pressure of 6 inches w.g. is 30 cfm/ft
- 2. c. For dampers with a surface area of less than 1 ft 2 , the maximum allowable leakage at a differential pressure of 6
inches w.g. is 30 cfm.
Discussed below are the power generation operations, fire operation, and
emergency operations of the control building HVAC systems. Shutdown operations are identical to the power generation operations.
POWER GENERATION OPERATION - The control building supply air system draws in
outside air, filters it through particulate prefilters, either cooling it with a chilled-water coil or heating it with a hot-water coil, and distributes the conditioned air to separate floors of the control building.
The control building supply air system intake is in a penthouse atop the auxiliary building, which is located approximately 15 feet below and 135 feet horizontally from the diesel exhaust discharge point. This separation is sufficient to provide
9.4-8 Rev. 14 WOLF CREEK significant dilution of the diesel exhaust gases; therefore, operation of the diesel during normal plant operations poses no danger to the occupants of the
control room or other areas of the building. The heating or cooling mode of operation of the outside air supply unit is a function of the outside air temperature only. When the outside air temperature
exceeds 65 F, conditioned outside air is supplied to the building. When the
outside air temperature is between 65 and 50 F, unconditioned outside air is supplied to the building. When the outside air temperature is below 50 F, the heating system is operational. These operations are controlled by temperature
switches, located in the ductwork upstream of the coils, which sense the
outside air temperature and function accordingly.
When the outside air temperature rises above 65 F, the temperature switch associated with the cooling system activates the supply unit cooling control
system. This control system then functions to maintain a constant supply air temperature of 60 F by modulating the flow of chilled water to the coil.
While the outside air temperature is between 65 and 50 F, the supply unit continues to operate, supplying unconditioned air to the building.
When the outside air temperature falls below 50 F, a temperature switch
activates the supply unit heating control system. This control system then functions to maintain the temperature of the air leaving the coil at 65 F. The supply unit heating coil is supplied from a secondary hot-water loop to prevent
the possible freezeup of the coil when the outside air temperature falls below
32 F. A temperature switch is provided in the outside air unit, downstream of
the coils. This temperature switch will trip the supply unit, should the supply temperature drop below 40 F, to protect the coils from freezing. Air from the control building supply system is supplied to the space above the access control area to remove the heat generated by electric cables. This cooling is provided to minimize the amount of cooling required for the spaces below. During periods of control building isolation, cooling is not required since the ambient temperature in the area will not exceed the ambient design rating (50 C) of the Class IE power cables.
Supplemental heating for the access control area is provided by electric duct heaters located in the supply air mains serving that area. The heaters are interlocked with the supply fan, and operation of the heaters is controlled by
room temperature switches which function to maintain space temperatures.
9.4-9 Rev. 19 WOLF CREEK Supplemental heating is also provided by electric unit heaters strategically located in the upper cable spreading room, the lower cable spreading room, the
ESF switchgear rooms, the basement areas, and the control room air-conditioning equipment rooms. Each heater is sized for its specific location and is thermostatically controlled to maintain the space design temperature requirements of 60 F or above.
Air from the clean areas of the control building is exhausted by the control building exhaust system. Air from the potentially contaminated areas of the control building is exhausted by the access control exhaust system. Exhaust
air from the access control exhaust system is processed through a charcoal
filtration train for cleanup prior to discharge through the unit vent. Exhaust
hoods are provided in the laundry room, over the sorting table, and in the hot lab over the rinse sink and over the sample test area. The hoods in the hot lab contain an integral exhaust air bypass arrangement for periods when flow
through a hood is not required. The hoods are used as part of the normal exhaust from the spaces and, therefore, contain no isolation provisions.
One of each of the two control building exhaust fans and access control exhaust fans runs continuously during normal plant operations. The motor-operated
discharge isolation dampers (one associated with each control building exhaust
fan) operate in conjunction with their corresponding fans. Automatic back-draft
dampers (one associated with each access control exhaust fan) operate in conjunction with their corresponding fans.
The control building exhaust system serves to remove the hydrogen generated by
the batteries during normal plant operation. The quantity of air exhausted
from each of the battery rooms is well in excess of that which was calculated as necessary to maintain the concentration of hydrogen in the rooms, under the worst conditions, below the flammability limit.
A differential pressure indicator controller, located across the access control
filter adsorber unit, modulates a damper downstream of the filter train to maintain a constant system resistance as the particulate filters load up. This control arrangement will assure a constant system flow.
Each charcoal adsorber is monitored for charcoal bed temperature. Should the
bed temperature approach 200 F, an alarm is received in the control room to alert the operators of excessive bed heating. Subsequently, should the bed temperature continue to rise, a 300 F alarm is received in the control room. Each particulate filter bank is provided with differential pressure
transmitters wired to the plant computer which alarms at excessive pressure
drops.
9.4-10 Rev. 14 WOLF CREEK The access control air-conditioning system operates in a continuous recirculation mode to provide supplemental cooling or heating of the nonvital
equipment areas of the electrical and mechanical equipment room and the first aid room, sign-in/out area, and the HP office and record storage area of the access control area.
The system cooling mode of operation is controlled by a temperature controller
which senses return air (space) temperature. If the temperature falls below 74 F, no cooling is provided. If the temperature falls below 65 F, the heating mode is initiated.
The system heating mode is controlled by a temperature controller located in the unit return air ductwork. This controller energizes the electric duct heater, as necessary, to maintain the return air (space) temperature at 65 F. Additional heating of the two mechanical equipment rooms is provided by an electric duct heater in the branches serving those spaces. These heaters are each sized for the specific room served and are thermostatically controlled to maintain the space design temperature requirements of 60 F or above. The control room air-conditioning system operates in a continuous recirculation
mode to maintain the control room temperature. The amount of cooling provided
by the self-contained refrigeration system is self-regulating and, therefore, automatically compensates for changes in the control room heat load, including latent load due to presence of moisture. The control room air-conditioning unit limits humidity to 70% RH in the room by dehumidifying supply air.
Heating, if required, is provided by an electric duct heater. This heater is
thermostatically controlled to maintain temperatures. The heater serves no safety function. The SAS room air handling unit normally operates to supply conditioned air to
the SAS room to maintain the room temperature. The amount of cooling provided
is regulated by regulating the amount of refrigerant which bypasses the
condensing coil. The system draws in air from the control room, cools it and discharges it to the SAS room. The amount of bypass is controlled by a thermostat located in the SAS room.
During operation of the SAS room air conditioning system, the supply air branch
from the control room air conditioning system to the SAS room is isolated. Should the SAS room air conditioning system fail, the supply air branch from the control room air conditioning system would be manually opened to provide
minimal cooling for the SAS room.
9.4-11 Rev. 19 WOLF CREEK The Class IE electrical equipment air-conditioning system is operated in a continuous recirculation mode to maintain the ESF switchgear room, the battery
rooms, and the dc switchgear rooms at or below a temperature of 90 F. The amount of cooling provided by the self-contained refrigeration system is self-regulating and, therefore, automatically compensates itself for changes in the
room heat loads.
The counting room fan-coil unit operates in a continuous recirculation mode to provide the necessary cooling and filtration of the counting room atmosphere to maintain a suitable ambience for the electronic equipment and personnel in the
room. The amount of cooling is controlled by a temperature controller located
in the counting room. The controller functions to maintain the space air
temperature. Prefilter and an HEPA filter are provided to minimize the airborne particulates in the counting room. A smoke detector installed in the ductwork down stream of the fan-coil unit will shut the unit down if smoke is
detected.
The control building supply air unit intake, the control building exhaust system, control room pressurization, and the access control exhaust system contain dampers capable of withstanding the effects of extreme wind or tornado
conditions (3 psi total at a rate of 2 psi/second per Regulatory Guide 1.76).
These dampers close with a tornado or high winds. The dampers located in the exhaust systems are spring loaded to prevent closure during normal system operations.
Based on the outside air design conditions, design space heat loads and
operation of the control building HVAC systems, as described above, no area of
the control building (except for the laundry and locker areas of the access control area) exceed a relative humidity of 70 percent.
EMERGENCY OPERATION - Located in the control building supply system ductwork, downstream of the control room filter adsorbers, are redundant radiation
monitors.
9.4-12 Rev. 27 WOLF CREEK A smoke detector is located upstream of the control room filter adsorbers. These monitors sense contaminants in the influent and alarm in the control room when limits are exceeded. The high radiation monitors initiate isolation of the control building normal supply and exhaust systems. The nonsafety-related systems which penetrate the boundary of the control building are provided with automatic isolation capabilities. This isolation
consists of two dampers, aligned in a series arrangement and powered from separate IE sources. The ductwork located between the two isolation dampers is designed to meet seismic Category I requirements. Upon receipt of the control
building isolation signal, these dampers close, thus isolating the control
building from all other adjacent buildings and outside air.
The control building isolation signal also automatically bypasses portions of the control room air-conditioning system flow through the associated
particulate filter charcoal adsorber train for cleanup and initiates operation
of the control room pressurization system. The control room pressurization system draws in outside air, processing it through a particulate filter charcoal adsorber train for cleanup. This outside air is diluted with air drawn from the cable spreading rooms and the electrical equipment floor levels
and distributed back into those spaces for further dilution. The control room
filtration system takes a portion of air from the supply side of this system, for dilution with portions of the return air from the control room air-conditioning system and processes it through the control room filtration system adsorption train for additional cleanup. This air is then further diluted with
the remaining control room air-conditioning system return air, cooled, and
supplied to the control room. This process maintains the control room under a
positive pressure of 1/4 inch w.g. (min.). This assures exfiltration from the control room, thus preventing any unprocessed contaminants from entering the control room. (The control room is classified as Type B, per the requirements of Regulatory Guide 1.78, with an air exchange rate exceeding 0.06 volume per
hour.) If the control room were isolated but unpressurized, the amount of inleakage resulting from a differential pressure of 1/4 inch w.g., caused by temperature, barometric, or wind variations, would be less than 80 cfm. Leakage rates are
calculated in accordance with "Conventional Buildings for Reactor Containment," NAA-SR-101000. The primary paths which contribute to this leakage are (1) the gap
9.4-13 Rev. 10 WOLF CREEK between the floor and the ceiling and building walls, (2) the joints between the stairway walls and the chase walls and building walls, (3) the doors to the
communication corridor, the electrical chases, and the stairway, (4) the door frames, (5) the ductwork, piping, and electrical penetrations, (6) penetration inserts, and (7) the ductwork isolation dampers.
The major contributors to the leakage are the doors, which account for
approximately 95 percent of the total leakage. The remaining paths are both individually and collectively insignificant in terms of the total control room leakage.
The control room pressurization system intake is in the west wall of the
control building and is located approximately 80 feet below and 80 feet horizontally from the diesel exhaust discharge point. This separation is sufficient to provide significant dilution of the diesel exhaust gases;
therefore, operation of the diesel during periods of control room isolation poses no danger to the occupants of the control room.
Indication of a loss of preferred ac power, a LOCA, or a fuel handling accident will automatically initiate the Class IE electrical equipment air-conditioning
systems if they are not in operation.
During normal plant operations, the battery rooms are purged with fresh air by the control building supply system and the control building exhaust system. This purging maintains the local concentration of hydrogen well below 0.2 volume percent.
During periods of control building isolation associated with a tornado, dilution air is not provided. This isolation can be maintained for approximately 3 days before purging is required to prevent local hydrogen concentration from approaching 2.0 volume percent (the lower flammability
limit). This is based on all batteries at full charge throughout the time
period. During periods of control building isolation, following an accident condition, the hydrogen concentration is maintained well below 0.5 volume percent by
dilution with air provided by the control room pressurization system.
The ambient temperature in the battery rooms, under any mode of operation, is between 60 F and 90 F.
During normal plant operations the SAS room is served by the non-vital SAS room
air handling unit. During periods of control room operation following loss of offsite power, ventilation air for the SAS room can be provided from the control room air-conditioning system by manual alignment of dampers.
9.4-14 Rev. 26 WOLF CREEK FIRE OPERATIONS - The operation of the HVAC systems during a fire and the interface between the ventilation systems and the fire protection system (Section 9.5) vary, depending on the type of fire protection and detection systems employed. In those areas where smoke detectors and automatic sprinklers are employed (upper and lower cable spreading rooms and access control area) or manual fire-
fighting is used (control room), no interface with or automatic isolation of the HVAC system(s) is provided. If it is determined, following receipt of a fire alarm signal in the control room, that it is necessary to isolate the HVAC
system(s) serving the alarmed area, then the operator can initiate isolation
from the control room.
In those areas where a halon extinguishing system is employed (dc switchgear and nonvital ac switchgear, El. 2,016 and IE ac switchgear, El. 2,000), the
HVAC system(s) serving those areas are interlocked to provide the necessary isolation upon receipt of a halon actuation signal. A halon release in either
of the IE ac switchgear rooms automatically isolates the portion of the control building supply air system and the control building exhaust serving that area and stop the associated Class IE air-conditioning unit.
A halon release in either of the nonvital ac switchgear rooms isolates only
that portion of the HVAC system serving the affected room. A halon release in any one of the dc switchgear rooms automatically initiates
isolation of that portion of the control building supply air system and control
building exhaust system serving that level and stops the respective Class IE
air-conditioning unit. 9.4.1.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in
Section 9.4.1.1. SAFETY EVALUATION ONE - The safety-related portions of the control building
HVAC systems are located in the control 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 safety-related portions of the control building
HVAC systems are designed to remain functional after a
9.4-15 Rev. 0 WOLF CREEK safe shutdown earthquake. 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
SAFETY EVALUATION THREE - The system description for the control building HVAC systems shows that complete redundancy is provided and, as indicated by Table 9.4-5, 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 control room system, the Class IE air-conditioning
system, and the control building isolation provisions are initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.4.1.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the applicable portions of the control room air-conditioning system and the Class IE electrical equipment system.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group
classification and seismic category applicable to the safety-related portions
of the control building HVAC systems. All the power supplies and control functions necessary for safe function of the control room air-conditioning system, the Class IE electrical equipment air-conditioning system, and the
control building isolation provisions are Class IE, and 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. Section 6.4 evaluates the isolation
requirements of Regulatory Guides 1.78 and 1.95.
SAFETY EVALUATION SEVEN - Completely redundant control room air-conditioning systems are provided for the control room. Each system is powered from independent Class IE power sources, and headered on separate essential service
water systems. Operation of these systems, as discussed in Section 9.4.1.2.3, maintains the design conditions specified in Section 3.11(B).
SAFETY EVALUATION EIGHT - Completely redundant Class IE electrical equipment air-conditioning systems are provided for the Class IE Switchgear and Battery
areas. Each system is powered from independent Class IE power sources and
headered on separate essential
9.4-16 Rev. 19 WOLF CREEK service water systems. Operation of this system, as discussed in Section 9.4.1.2.3, maintains the design conditions specified in Section 3.11(B).
9.4.1.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
Filters and adsorbers for the access control exhaust system, the control room pressurization system, and the control room filtration system are tested in the manufacturer's shop, after initial installation, and subsequent to each filter
or adsorber change. Following installation of the filters and adsorbers for
the safety-related filtration units (control room pressurization and control
room filtration units), interim tests and inspections are performed in accordance with the requirements of Regulatory Guide 1.52, as discussed in Table 9.4-2. Following installation of the filters and adsorbers for the
nonsafety-related filtration units (access control filtration unit), interim
tests and inspections are performed in accordance with the requirements of
Regulatory Guide 1.140, as discussed in Table 9.4-3. During technical specification surveillance testing (TS 5.5.11a., b., f.) of the Pressurization System filter absorber unit, a Pressurization System flow rate of 2200 cfm + 10% is verified. All charcoal adsorbs are factory tested in accordance with RDT M-16-lT to exhibit a decontamination efficiency of no less than 99.9 percent for elemental iodine and 98 percent for methyl iodide. Sample charcoal canisters are tested
for impregnant efficiency in an independent laboratory using radio-methyl
iodide tracers. Inplace testing is performed with a suitable refrigerant, in accordance with the procedures set forth in ANSI N510, to check for bed bypass leakages.
Prefilters do not undergo factory or inplace testing since no credit is taken
for removal of particulates. HEPA filters are factory tested with DOP aerosol to demonstrate a minimum
particulate removal efficiency of no less than 99.97 percent for 0.3 micron
particulates. Inplace leak testing is carried out with cold polydispersed DOP.
Testing is in accordance with the procedures set forth in ANSI N510. One of each type of safety-related fan (control room air-conditioning system, control room filtration system, control room pressurization system, and Class
IE electrical equipment air-conditioning system) are tested in accordance with
AMCA standards. All other fans are AMCA rated. One control room air-conditioning unit and one Class IE electrical equipment
air-conditioning unit are performance tested by the manufacturer to assure
design heat removal capabilities.
9.4-17 Rev. 13 WOLF CREEK Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.1.5 Instrumentation Applications Indication of fan operational status is provided in the control room.
All fans, except the counting room fan coil unit fan, are operable from the control room.
An indication of the position of all isolation dampers is provided in the
control room.
Thermostats, located in the various levels and the ductwork, control space temperatures.
A humidistat located in the control room controls the minimum relative humidity
in the room. A duct-mounted airflow switch interlocks the operation of the electric steam humidifier with the control building supply air unit. In addition, duct-mounted moisture limit switches de-energize the humidifier when
supply air moisture relative humidity exceeds their high limit.
The indication of the amount of filter loading for all filters associated with the essential and nonessential air handlers is provided at each of the air handlers.
Alarms are provided in the control room to indicate high charcoal bed
temperatures in the control room filtration, control room pressurization and access control filtration units and high room temperature in the ESF switchgear and dc switchgear rooms.
An alarm is provided in the control room to indicate high hydrogen concentrations in a battery room. Alarms are provided in the control room to indicate high radiation and smoke in
the control building intake.
All instrumentation provided with the filtration units is as required by Regulatory Guide 1.52 or 1.140, as applicable.
9.4.2 FUEL BUILDING HVAC
The fuel building ventilation system consists of the fuel building supply system which includes the fuel building heating coil, the fuel building supply air unit, and the fuel handling area cooling coil; the emergency exhaust
system, including the emergency exhaust heating coil; the auxiliary/fuel
building normal exhaust system; the fuel storage pool cooling pump room coolers; and the unit heaters. Since both the emergency exhaust system and the auxiliary/fuel building normal exhaust system also serve the auxiliary building, their operation in the auxiliary building is discussed in Section 9.4.3.
9.4-18 Rev. 14 WOLF CREEK The fuel building supply system provides conditioned outside air for ventilation and cooling or heating, as required, to all areas of the fuel
building. The auxiliary/fuel building normal exhaust system exhausts air from the area above the fuel storage pool during normal operation and provides a means of purging smoke following a postulated fire.
In the event of a fuel handling accident, the emergency exhaust system collects
and processes the airborne particulates in the fuel building. In the event of a LOCA, the emergency exhaust system processes the atmosphere of the auxiliary building.
The fuel storage pool cooling pump room coolers provide a suitable ambient
temperature for the electric motor drives of the safety-related pumps. The fuel building unit heaters provide supplemental heating for the fuel
building, when required.
9.4.2.1 Design Bases 9.4.2.1.1 Safety Design Bases
The emergency exhaust system, the fuel storage pool cooling pump room coolers, and those portions of the fuel building supply system and the auxiliary/fuel building normal exhaust system which are required to provide isolation of the fuel building are safety related and are required to function following a DBA
and to achieve and maintain the plant in a post accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The emergency exhaust system, the fuel storage pool cooling pump room coolers, and the HVAC penetrations of the fuel building
boundaries are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The emergency exhaust system, the fuel storage pool cooling pump room coolers, and the HVAC penetrations of the fuel building
boundary remain functional after a SSE and perform their intended function
following a postulated hazard, such as internal missiles, or pipe break (GDC-4). SAFETY DESIGN BASIS THREE - The safety functions of the fuel building HVAC
systems can be performed, assuming a single active component failure coincident
with the loss of offsite power.
SAFETY DESIGN BASIS FOUR - Active components of the fuel storage pool cooling pump room coolers, the emergency exhaust system and the fuel building HVAC
boundary penetration isolation provisions
9.4-19 Rev. 19 WOLF CREEK are capable of being tested during plant operation. Provisions are made to allow for inservice inspection of components at appropriate times, as specified
in the ASME Boiler and Pressure Vessel Code, Section XI. SAFETY DESIGN BASIS FIVE - The emergency exhaust system, the fuel storage pool cooling pump room coolers, and the HVAC penetrations of the fuel building boundaries are 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 ability to isolate the HVAC system penetrations
of the fuel building boundaries is provided, when required, so that the emergency exhaust system functions are not compromised.
SAFETY DESIGN BASIS SEVEN - Means are provided to assure both the control and
monitoring of radioactive releases following a fuel handling accident (GDC-60
and GDC-64). Radiological consequences of a fuel handling accident are evaluated in Chapter 15.0.
SAFETY DESIGN BASIS EIGHT - The fuel storage pool cooling pump rooms' ambient temperature is limited to assure operability of the fuel storage pool cooling pump. 9.4.2.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The fuel building ventilation system maintains the space temperature between 60 and 110 F during normal and fuel handling operations.
POWER GENERATION DESIGN BASIS TWO - The auxiliary/fuel building normal exhaust
system is sized to exhaust slightly more air than is being supplied to inhibit unprocessed exfiltration from the building.
POWER GENERATION DESIGN BASIS THREE - The fuel building ventilation system is
designed to maintain the airborne radioactivity levels within the fuel building
below the maximum permissible concentrations (MPC), as defined by 10 CFR 20. The exhaust system is designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the ALARA dose objective of 10 CFR 50, Appendix I.
9.4-20 Rev. 14 WOLF CREEK 9.4.2.2 System Description 9.4.2.2.1 General Description The fuel building ventilation system is designed to provide fresh air, heated or cooled, as required, for the fuel building. The fuel building and auxiliary
building share common ventilation exhaust systems for normal and emergency
operation. The auxiliary/fuel building normal exhaust system is described in Section 9.4.3. The fuel building HVAC systems are shown in Figure 9.4-2.
The emergency exhaust system will collect and process the fuel building
atmosphere in the event of a fuel handling accident. During operation of the emergency exhaust system, the nonessential fuel building HVAC air paths are isolated and the building exhausted to assure that fission products and particulate matter are collected and processed. The fuel building intake air
system is provided with two motor-operated dampers in a series arrangement. Indication of high radiation levels in the fuel building will initiate automatic transfer to the emergency exhaust system. Each fuel storage pool pump room is provided with a local independent room
cooler.
During a tornado or extreme wind conditions, the fuel building is vented to equalize pressures.
Based on the source terms provided in Section 11.1 and the dose evaluation
provided in Section 11.3, the exhaust system meets the objective of 10 CFR 50, Appendix I, and the limits of 10 CFR 20. 9.4.2.2.2 Component Description
Codes and standards applicable to the fuel building HVAC systems are listed in
Table 3.2-1. Design data for major components of the Fuel Building HVAC systems are presented in Table 9.4-6. The emergency exhaust system, fuel storage pool cooling system pump room coolers, and the safety-related HVAC
penetrations of the fuel building boundaries are designed and constructed in
accordance with codes and standards comparable with quality group C. The fuel
storage pool cooling system pump room cooling coils are designed and constructed in accordance with quality group C.
9.4-21 Rev. 19 WOLF CREEK NONESSENTIAL AIR HANDLING UNITS - The fuel building supply air units are the nonessential air handling units in the fuel building. Each unit consists of a
chilled-water cooling coil and centrifugal fan with electric motor driver. SAFETY-RELATED ROOM COOLERS - The only safety-related room coolers located in the fuel building are the fuel storage pool cooling pump room coolers. Each unit consists of an essential service water cooling coil and centrifugal fan with electric motor drive. SAFETY-RELATED FILTRATION UNITS - The emergency exhaust filter/adsorber units are located in the fuel building. Each filter train consists of moderate efficiency prefilters, HEPA filters, and charcoal adsorption beds.
SAFETY-RELATED FANS - The emergency exhaust system fans are located in the fuel building. These fans are centrifugal fans with electric motor drivers.
HEATING EQUIPMENT - Heating of the fuel building is provided by a hot-water
heating coil located in the fuel building supply air ductwork and by unit heaters located in various areas of the building. Heating of the air to each of the emergency exhaust system filter/adsorber units is provided by a safety-
related electric duct heater. Unit heaters are either hot-water type, consisting of a coil and fan with an electric motor, or electric heating coil type. SUPPLEMENTAL COOLING - Additional cooling is provided, when required, by a
chilled-water cooling coil located in the supply air system ductwork to the
fuel storage pool area. ISOLATION DAMPERS - Where a means of system isolation is required, parallel-blade-type dampers are utilized. The type of operator employed is dependent upon the specific design and/or usage requirements.
FLOW CONTROL DAMPERS - Opposed-blade-type dampers are utilized, as necessary, to provide a means of system balancing. BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to maintain
the proper direction of flow.
TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects of extreme wind or tornado conditions is required. These dampers close with
the flow produced by the
9.4-22 Rev. 14 WOLF CREEK differential pressure associated with the tornadoes or high winds. FIRE DAMPERS - Fire dampers are located in fire barriers, as necessary, to maintain the fire ratings of the barriers. Dampers are the 3-hour-rated curtain type.
9.4.2.2.3 System Operation
The fuel building is served by an outside air supply system which provides fresh outside air, either heated or cooled as required, to all areas of the
fuel building. The supply air unit has provisions for operating in a
recirculation mode. Additional cooling for the fuel handling area is provided
by a cooling coil located in the duct supplying that area. Within the fuel building, the auxiliary/fuel building normal exhaust system
takes suction from the area above the fuel storage pool and mixes that air with
the air from the auxiliary building prior to processing it through the
auxiliary/fuel building filter adsorber train and discharging it to the unit vent. The emergency exhaust system collects and processes the fuel building
atmosphere in the event of a fuel handling accident. During operation of the
emergency exhaust system, the fuel building nonessential HVAC air paths are isolated and the building exhausted to assure that fission products and particulate matter are collected and processed. The fuel building intake air
system is provided with two motor-operated dampers in a series arrangement.
Each damper is powered from a separate Class IE source to assure closure. This
transfer occurs automatically upon receipt of a fuel building isolation signal. The emergency exhaust system maintains a minimum negative pressure of greater than or equal to 1/4 in. w.g. to assure that all leakage is into the building.
The emergency exhaust system is on standby for an automatic start following
receipt of a fuel building isolation signal or a safety injection signal (SIS). The initiation of the LOCA mode of operation (SIS) takes precedence over any other mode of operation.
Each fuel storage pool cooling pump room is provided with a local independent
cooling unit. These cooling units utilize essential service water as the heat sink (service water during normal plant
9.4-23 Rev. 27 WOLF CREEK operation) and are powered by the same Class IE power supply as the associated pump to be cooled. Each unit has the capacity to provide 100 percent of the
cooling required. During a tornado or extreme wind conditions, the fuel building is vented to equalize pressures. Missile protection is provided to prevent a tornado
missile from damaging HVAC equipment required during safe shutdown.
Discussed below are the power generation operations and the emergency operations of the fuel building HVAC systems.
The differences between shutdown operations and power operations are few and
are, therefore, covered under the power generation operations. POWER GENERATION OPERATION - Outside air is drawn in by one of the two supply
air units, filtered through the particulate filter, conditioned as required, and distributed to the various areas of the fuel building. Depending on the
space temperature requirements, the outside air is either heated by the hot-water coil located in the outside air intake or cooled by the supply air unit's chilled-water coil. Each fuel storage pool cooling pump room is provided with
a room cooler to maintain the ambient temperature within limits. Space heating
is provided by the outside air intake heating coil and supplemented by hot-
water and electric unit heaters. The heating or cooling mode of operation of the outside air intake unit is
controlled by the outside air temperature. When the outside air temperature
exceeds 78 F, the chilled-water-cooled outside air is supplied to the building.
When the outside air temperature is between 78 F and 50 F, outside air is supplied directly into the building. When the outside air temperature is below 50 F, the heating system is operational. These operations are controlled by temperature switches located in the inlet ductwork, upstream of the coils.
When the outside air temperature rises above 78 F, the temperature switch associated with the cooling system activates the supply unit cooling control system. This control system then functions to maintain a constant supply air
temperature, by modulating the flow of chilled water to the coil.
During fuel handling operations, the supplementary chilled-water coil located in the supply air duct may be manually actuated. Once actuated, the chilled-water flow to the coil is automatically modulated to limit the ambient
temperature in the fuel handling area to 110 F.
9.4-24 Rev. 23 WOLF CREEK When the outside air temperature falls below 50 F, a temperature switch activates the heating coil control system. The heating coil is supplied from a
secondary hot-water loop which is, in turn, supplied from the plant heating system. This arrangement is provided to circulate water through the coil to prevent a possible freezeup of the coil.
The heating control system consists of temperature transmitters located in
various spaces, which sense the space temperature and transmit a corresponding signal to a single temperature controller. When any of these signals indicates that a space temperature is below 60 F, the temperature controller then
modulates the amount of heating accordingly. The temperature controller
controls the secondary loop temperature by regulating the amount of hot water
which enters the secondary loop of the heating coil. A temperature switch is provided in the supply air duct downstream of the
heating coil. This temperature switch will isolate the supply air intake and trip the supply fan, should the supply air temperature drop below 40 F, to
protect the coils from freezing. The fuel building supply air system intake is in the side of the west wall of
the fuel building, and is located approximately 70 feet below and 165 feet
horizontally from the diesel exhaust discharge point. This separation is
sufficient to provide significant dilution of the diesel exhaust gases; therefore, operation of the diesel during normal plant operations will result in no significant ingestion of exhaust gases into the fuel building.
Supplemental heating is provided by unit heaters located throughout the
building. Each unit heater is sized for its location, and each is thermostatically controlled to maintain the space design requirements of 60 F or above.
The auxiliary/fuel building normal exhaust system components are located in the
auxiliary building and are described in Section 9.4.3. All normal exhaust from the fuel building is through the auxiliary/fuel building normal exhaust system.
The auxiliary/fuel building normal exhaust system and/or the emergency exhaust
system provide a means of purging smoke following a postulated fire. The operation of the supply air units is interlocked with the operation of the
auxiliary/fuel building normal exhaust fans. A supply air unit will operate in
the supply mode only if an auxiliary/fuel building normal exhaust fan is
operating at fast
9.4-25 Rev. 0 WOLF CREEK speed. When a normal exhaust fan is operating at slow speed, the supply air unit operates only if in the recirculation mode of operation.
The fuel building intake air isolation system consists of two dampers in a series arrangement, each powered by a separate Class IE source. The dampers are designed to close automatically upon a high radiation indication within the
fuel building.
EMERGENCY OPERATIONS - Actuation of the emergency mode of operation is initiated either manually by operator action or automatically upon detection of
high radiation levels in the fuel building. All air from the fuel building is processed through the emergency exhaust filter adsorber train, prior to release from the unit vent, and all makeup air is by infiltration only. Actuation of the FBVIS isolates the outside air intake system, trips the supply air handling units and closes the corresponding damper in the normal exhaust ductwork to the auxiliary building in order to isolate the fuel building.
In the event of a LOCA, the SIS trips off the fuel building supply fan and closes the dampers in the normal exhaust system to isolate the fuel building
exhaust ductwork from the auxiliary building. The SIS concurrently energizes the emergency exhaust fan, opens the corresponding damper in the emergency
ductwork from the auxiliary building and ensures that the appropriate damper in the emergency ductwork from the fuel building is in its throttled position to provide a negative pressure of greater than or equal to 1/4 in. w.g. in the
auxiliary building. All exhaust is processed through the fuel building
emergency exhaust filter train. Under this mode, all nonessential fuel building
HVAC is out of service. Each charcoal adsorber train is monitored for charcoal bed temperature. Should
the bed temperature approach 200 F, an alarm is received in the control room to
alert the operators of excessive bed heating. Subsequently, if bed temperature continues to rise, a 300 F alarm is received in the control room.
9.4-26 Rev. 19 WOLF CREEK To prevent backflow through the system, upstream isolation is provided by a backdraft damper located at the inlet to the filter train. Each particulate
filter bank is provided with differential pressure transmitters wired to the plant computer which will alarm excessive pressure drops. The emergency exhaust system is provided with electric heating coils to
maintain the relative humidity of the air entering the charcoal filters below
70 percent. The fuel storage pool cooling pump room coolers are activated when their
associated pump starts. Each pump room cooler is full capacity, utilizes
service water (normal operations) or essential service water (accident
operations) as the cooling medium, and is powered from the same Class IE source as the associated fuel storage pool cooling pump. Each pump room cooler is located in its respective pump room and operates in a complete recirculation
mode. Each pump room is monitored for space temperature; should the space temperature exceed 122 F, the condition will be alarmed in the control room via
the plant computer. Fresh air for ventilation is provided during normal plant operation by the fuel building supply system.
9.4.2.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.4.2.1. SAFETY EVALUATION ONE - The safety-related portions of the fuel building HVAC
systems are located in the fuel building and the auxiliary building. 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 fuel building HVAC systems are designed to remain functional after a safe shutdown earthquake. 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained. SAFETY EVALUATION THREE - Complete redundancy is provided and, as indicated by Table 9.4-7, 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.
9.4-27 Rev. 19 WOLF CREEK SAFETY EVALUATION FOUR - The fuel storage pool cooling pump room coolers, the emergency exhaust system, and the fuel building HVAC boundary penetration
isolation provisions are initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.4.2.4. Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the fuel storage pool cooling pump room coolers. 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. All the power supplies and control functions necessary for safe function of the fuel storage pool cooling pump room coolers, emergency exhaust system, and the fuel building HVAC boundary
penetration isolation provisions are Class IE, as described in Chapters 7.0 and
8.0. SAFETY
EVALUATION SIX - Section 9.4.2.2.3 describes the provisions made to assure the isolation of the auxiliary building following a DBA.
SAFETY EVALUATION SEVEN - The emergency exhaust system maintains a negative
pressure of no less than 1/4 in. w.g. in the fuel building to prevent unprocessed exfiltration following a fuel handling accident which releases radioactivity. The emergency exhaust system is monitored for radioactivity
both upstream and downstream of the filter adsorber unit prior to release to
the site. The filter adsorber unit limits the radiological consequences of a
fuel handling accident to less than 10 CFR 100 limits. SAFETY EVALUATION EIGHT - Room coolers are installed in each fuel storage pool cooling pump room and are designed to limit pump room ambient temperature to assure operability of the fuel storage pool cooling pump. 9.4.2.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
Filters and adsorbers for the emergency exhaust system are tested in the manufacturer's shop, after initial installation and subsequent to each filter or adsorber change. After installation, interim tests and inspections are
performed after every 720 hours of operation and once per 18 months in
accordance with the requirements of Regulatory Guide 1.52 and the Technical
Specifications, to detect any deterioration of components that may develop under service or standby conditions.
9.4-28 Rev. 20 WOLF CREEK Prefilters do not undergo factory or inplace testing since no credit is taken for removal of particulates in meeting permissible dose rates. However, unloaded prefilters will exhibit a 55-percent efficiency (min.) for the removal of coarse particulates when tested in accordance with ASHRAE-52. HEPA filters are factory tested with monodispersed DOP aerosol to demonstrate a
minimum particulate removal efficiency of no less than 99.97 percent for 0.3
micron particulates. Inplace leak testing is carried out with cold polydispersed DOP. Testing is in accordance with procedures set forth in ANSI N510.
Charcoal adsorbers are qualified per Regulatory Guide 1.52 and are factory
tested in accordance with RDT M-16-IT to exhibit a decontamination efficiency of no less than 99.9 for elemental iodine and 98 percent for methyl iodide. Sample charcoal canisters are tested for impregnant efficiency in an
independent laboratory, using radiomethyl iodide tracers. Inplace testing is performed with a suitable refrigerant, in accordance with the procedures set
forth in ANSI N510, to check for bed bypass leakages. The emergency exhaust system, the fuel building HVAC boundary penetration
isolation provisions, and fuel storage pool cooling pump room coolers will undergo preoperational testing prior to plant startup. The remaining system undergoes acceptance testing. One fan from each of the emergency exhaust fans and fuel storage pool cooling pump room cooler fans are tested in accordance with AMCA standards. All other
fans are AMCA rated.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.2.5 Instrumentation Applications Indication of the operational status of all fuel building HVAC fans is provided in the control room. All fans, except the fuel storage pool cooling pump room cooler fans, are operable from the control room. An indication of the position of all isolation dampers is provided in the
control room. Thermostats, located in various areas of the fuel building and in the HVAC ductwork, control space temperatures.
9.4-29 Rev. 14 WOLF CREEK The amount of filter loading for the supply air unit intake filter is indicated at the supply unit.
Indication of the levels of gaseous particulate and iodine radioactivity being exhausted from the fuel building during all modes of operation is available in the control room.
All instrumentation provided with the emergency exhaust filter/adsorber unit is as required by Regulatory Guide 1.52.
A high temperature computer alarm for each of the fuel storage pool cooling
pump rooms is provided in the control room.
9.4.3 AUXILIARY
BUILDING
The auxiliary building ventilation system consists of the auxiliary building supply system, the auxiliary/fuel building normal exhaust system (including the
decontamination tanks exhaust scrubbers), the emergency exhaust system, the main steam tunnel supply system, the main steam tunnel exhaust system, and the access tunnel transfer fan. Local fan coil units serve the electrical
equipment room, the component cooling water pump room, the ground floor
corridor, the hot machine shop/hot instrument shop, the normal charging pump
room, and the basement corridor. Local room coolers serve the safety injection pump rooms, the component cooling water pump rooms, the RHR pump rooms, the centrifugal charging pump rooms, the containment spray pump rooms, the auxiliary feedwater (motor-driven) pump rooms, and the electrical penetration rooms. A room air-conditioner serves the I&C Hot Shop.
Since both the auxiliary/fuel building normal exhaust system and the emergency exhaust system also serve the fuel building, their operation in the fuel building is described in Section 9.4.2. All modes of operation discussed in
this section are applicable to the auxiliary building only.
The auxiliary building supply system and the main steam tunnel supply system function to provide conditioned outside air for ventilation and cooling or
heating, as required, to each level of the auxiliary building. During normal
operations, the auxiliary/fuel building normal exhaust system and the main
steam tunnel exhaust system operate to provide the required exhaust from the building. These systems also provide a means of purging smoke following a postulated fire.
9.4-30 Rev. 15 WOLF CREEK During periods when the decontamination system tank(s) are in use, the decontamination tanks exhaust scrubber(s) operate to exhaust any fumes evolving
from the tanks. Following a LOCA, the emergency exhaust system serves to collect and process airborne particulates in the auxiliary building and exhausts the air purged
from the containment via the containment hydrogen control system.
The fan coil units serve to provide supplemental cooling of the auxiliary building, as required.
The pump room coolers provide a suitable ambient environment for the electric
motor drivers for the safety-related pumps. The penetration room coolers provide a suitable atmosphere for the safety-
related electrical equipment located in the electrical penetration rooms.
The I&C Hot Shop air-conditioning unit provides a suitable environment for equipment and personnel working in the shop.
The access tunnel transfer fan functions to supply air from the auxiliary
building basement corridor to the radwaste tunnel.
The auxiliary building unit heaters provide supplemental heating to the auxiliary building, when required.
9.4.3.1 Design Bases 9.4.3.1.1 Safety Design Bases
The pump room coolers, the penetration room coolers, the emergency exhaust system, and those portions of the auxiliary building and the main steam tunnel
supply systems and the auxiliary/fuel building normal exhaust and main steam enclosure building exhaust systems which are required to provide isolation of the auxiliary building are safety-related and are required to function
following a DBA and to achieve and maintain the plant in a post accident safe shutdown condition. The I&C Hot Shop air-conditioning unit serves no safety function, and a failure of this unit will not affect the safe shutdown of the plant. SAFETY DESIGN BASIS ONE - The pump room coolers, the penetration room coolers, the emergency exhaust system, and the auxiliary building isolation provisions
are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The pump room coolers, the penetration room coolers, the emergency exhaust system, and the isolation provisions for the auxiliary
building remain functional after a safe shutdown earthquake and perform their
intended function following a postulated hazard, such as internal missiles, or pipe break (GDC-4).
9.4-31 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS THREE - The safety functions of the auxiliary building HVAC systems can be performed, assuming a single active component failure coincident
with the loss of offsite power. SAFETY DESIGN BASIS FOUR - Active components of the auxiliary building safety-related HVAC systems 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.
SAFETY DESIGN BASIS FIVE - The pump room coolers, the penetration room coolers, the emergency exhaust system, and the safety-related auxiliary building
isolation provisions are consistent with the quality group classification assigned by Regulatory Guide 1.26 and with the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions must be in
accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The capability to isolate both the safety and nonsafety-related HVAC system penetrations of the auxiliary building boundary is provided so that the safety-related HVAC systems' functions are not
compromised.
SAFETY DESIGN BASIS SEVEN - Means are provided to assure both the control and monitoring of gaseous radioactive releases following a LOCA. The radiological consequences are evaluated in Section 15.0.
SAFETY DESIGN BASIS EIGHT - The ESF pump room coolers limit the ESF pump room
ambient temperatures to assure operability of the ESF pumps. 9.4.3.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The auxiliary building supply system
provides conditioned outside air to maintain the ground floor level of the auxiliary building at or below 104 F. The auxiliary building supply system also provides supplemental cooling for each of the other floor levels of the
auxiliary building. The auxiliary building supply system provides fresh air
ventilation at a rate of 0.1 cfm/ft 2 of floor area or greater. POWER GENERATION DESIGN BASIS TWO - The main steam tunnel supply and exhaust systems limit the temperature to a maximum of 120 F and a minimum of 50 F and
provide fresh air ventilation.
POWER GENERATION DESIGN BASIS THREE - The auxiliary/fuel building normal exhaust system exhausts slightly more air than is being
9.4-32 Rev. 0 WOLF CREEK supplied, to inhibit exfiltration of the air from the auxiliary building. The main steam tunnel exhaust system exhausts an amount of air equal to that being supplied. POWER GENERATION DESIGN BASIS FOUR - The auxiliary building ventilation system maintains the auxiliary building sample room between 60 F and 104 F, the hot machine shop and the hot instrument shop between 60°F and 85°F. The electrical equipment room ventilation system maintains the room at or below 80°F. All other areas of the auxiliary building are maintained between 60°F and 104°F except as noted in Table 3.11(B)-1. The boric acid storage tank areas, the pipe chase containing the boric acid piping which runs between the tanks and the boric acid filters, and the boric acid filter valve gallery are maintained at a minimum of 75°F to prevent crystallization of the boron in the lines.
POWER GENERATION DESIGN BASIS FIVE - The auxiliary building air flow patterns are from levels of lower contamination potential to levels of higher contamination potential.
POWER GENERATION DESIGN BASIS SIX - The design exhaust flow of a single decontamination tank's exhaust scrubber is sufficient to prevent the
escape of fumes from the largest tank to the room. Each scrubber removes sufficient caustic and acidic fumes from the exhaust air so that the exhaust air can be discharged to the auxiliary/fuel building
normal exhaust system without adverse effects on the system components. POWER GENERATION DESIGN BASIS EIGHT - The ventilation exhaust system is
designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the ALARA dose objective of 10 CFR 50, Appendix I. The auxiliary/fuel building normal exhaust system filter adsorber unit
complies with Regulatory Guide 1.140, to the extent discussed in Table 9.4-3. POWER GENERATION DESIGN BASIS NINE - The I&C Hot Shop air-conditioning unit limits the I&C Hot Shop temperature to a maximum of 78°F.
9.4.3.2 System Description 9.4.3.2.1 General Description
The auxiliary building HVAC system is shown in Figure 9.4-3. The auxiliary building is served by an outside air supply system which
provides fresh outside air, either heated or cooled as required, to all levels of the auxiliary building. Local fan coil units serve the basement corridor area, the normal charging pump room, the ground floor
corridor, the hot machine shop/hot instrument shop, the component cooling water pump room, and the electrical equipment room areas of the auxiliary building. Local cooling units are provided to minimize
ductwork requirements.
9.4-33 Rev. 16 WOLF CREEK The main steam tunnel is served by a unit which provides fresh outside air, either heated or cooled as required. The auxiliary/fuel building normal exhaust system takes suction from the areas of greater contamination potential of the auxiliary building, mixes this exhaust air with the exhaust air from the fuel building, and processes the exhaust air through a charcoal adsorber train prior to
discharge through the unit vent. Based on the source terms provided in Section 11.1 and the dose evaluation provided in Section 11.3, the exhaust system meets the objective of 10 CFR 50, Appendix I, and the
limits of 10 CFR 20.
The main tunnel exhaust system takes suction from all levels of the main steam tunnel and discharges directly to the unit vent.
Both the auxiliary/fuel building normal exhaust system and the main steam tunnel exhaust system are monitored for activity, in accordance
with the requirements of Regulatory Guide 1.21. The emergency exhaust system serves the auxiliary building only
following a LOCA to assure that all ECCS leakage to the auxiliary building atmosphere and the containment air purged via the hydrogen purge system are processed. All ductwork which is not required for
operation of the emergency exhaust system and penetrates the auxiliary building boundary is automatically isolated. These nonessential systems are provided with two motor-operated dampers in a series
arrangement at the boundary penetrations. These will close automatically following receipt of an SIS. The emergency exhaust system maintains a negative pressure of greater than or equal to 1/4
in. w.g. to assure that all leakage is into the building. Each area containing safety-related equipment that is heat sensitive is
provided with a local independent cooling unit. These cooling units
utilize essential service water as the heat sink and are powered by the same Class IE supply as the associated equipment to be cooled.
The decontamination tanks exhaust scrubbers collect and remove the caustic vapors which evolve from the decontamination tanks. After processing the exhaust air, the scrubbers discharge to the
auxiliary/fuel building normal exhaust system. Outside air is induced into the decontamination room for makeup when the scrubber(s) are in use. Supplemental heating is provided by unit heaters located throughout the building.
The access tunnel transfer fan transfers air from the auxiliary building to the radwaste tunnel. The I&C Hot Shop utilizes a room air-conditioner for cooling. The I&C Hot Shop room air-conditioner operates in a complete recirculation mode, however can be manually switched to a 100% exhaust mode as required.
9.4-34 Rev. 13 WOLF CREEK An evaluation of the effects of the postulated inability to maintain preferred air flow patterns in the auxiliary building is summarized below: a. Loss of Auxiliary Building Supply System The auxiliary/fuel building normal exhaust system has the
capability of operating at a reduced flow following the postulated loss of the supply system. Depending on physical resistance to building infiltration and fan
characteristics, both exhaust fans may be operated in a
parallel arrangement to maintain approximate design flow rates.
The ductwork distribution system is designed to supply directly to the clean areas, such as corridors, and exhaust from the potentially contaminated areas, such as
equipment compartments. With the postulated loss of supply air, the exhaust pattern from the potentially contaminated areas is maintained. The source of makeup
air is building infiltration which flows toward the potentially contaminated areas. Therefore, the effect of this event is negligible.
- b. Loss of Auxiliary/Fuel Building Normal Exhaust System
The auxiliary/fuel building normal exhaust system is provided with redundant, full-capacity fans. However, assuming a loss of the exhaust air flow, the supply
system automatically shuts down to prevent building pressurization. The supply fan is interlocked with the exhaust system so that the exhaust system must be
operating before the supply system can be started or
operated. Therefore, a postulated loss of the exhaust system
results in a complete loss of direct outside air movement within the auxiliary building. Natural air flow patterns may be established, depending on thermal gradients and
the flow paths existing within and across the auxiliary building. Assuming uniform mixing of the auxiliary building atmosphere as the most conservative case, there
would be negligible effect in relation to operator exposure if the ventilation system is returned to service within several hours. Actions are taken to remove
unnecessary equipment from service if it contributes to personnel exposure in order to maintain exposures ALARA.
9.4-35 Rev. 0 WOLF CREEK The loss of normal ventilation will have no impact on those areas with safety-related equipment. Other areas of the building are periodically monitored, depending upon operating loads and duration of the loss of ventilation, to determine the impact on equipment. The equipment room housing the auxiliary building exhaust
components is located on the operating floor which houses radioactively clean components. In addition, the exhaust fans are belt-driven centrifugal fans. The parts which
are normally susceptible to wear include belts, motor
drive, and bearings. These parts are readily available for replacement and can be easily installed within a few hours.
9.4.3.2.2 Component Description
Design data for major components of the auxiliary building HVAC systems are presented in Table 9.4-8. Codes and standards applicable to the auxiliary building HVAC systems are listed in Table 3.2-1. The pump
room coolers, penetration room coolers, emergency exhaust system, and the safety-related penetrations of the auxiliary building boundaries are designed and constructed in accordance with codes and standards
comparable with quality group C. The pump room cooler cooling coils and the penetration room cooler cooling coils are designed and constructed in accordance with quality group C.
NONESSENTIAL AIR HANDLING UNITS - Listed and described below are those nonessential air handling units which make up a part of the auxiliary
building HVAC system. The auxiliary building supply air unit consists of particulate filters, hot-water heating coil, chilled-water cooling coil, centrifugal fan, and electric motor driver. The electrical equipment room fan coil units, the component cooling
water pump room fan coil units, the ground floor fan coil unit, the hot machine shop/hot instrument shop fan coil unit, the normal charging pump room fan coil unit, and the basement corridor fan coil unit each
consist of particulate filters, chilled-water cooling coil, centrifugal fan, and electric motor. The I&C Hot Shop room air-conditioner consists of a particulate filter, a centrifugal blower, an electric motor driver, an evaporator, a compressor, and a condenser.
9.4-36 Rev. 13 WOLF CREEK The decontamination tanks exhaust scrubbers consist of a stainless steel housing containing a water spray, filter media, and centrifugal fan with electric motor driver. SAFETY-RELATED ROOM COOLERS - Those room coolers which provide safety-related cooling are described below.
The SI pump room coolers, the RHR pump room coolers, the component cooling water pump room coolers, the centrifugal charging pump room coolers, the containment spray pump room coolers, the auxiliary
feedwater pump room coolers, and the penetration room coolers each
consist of coils utilizing essential service water as the cooling medium, centrifugal fans, and electric motor drivers. Units which normally operate are provided with particulate filters.
NONESSENTIAL FILTER UNITS - The auxiliary building HVAC systems contain only one nonessential filter unit--the auxiliary/fuel building normal
exhaust filter adsorber unit. This unit consists of moderate efficiency prefilters, HEPA filters, and charcoal adsorption beds.
SAFETY-RELATED FILTER UNITS - The auxiliary building HVAC systems contain no safety-related filter units. The emergency exhaust filter adsorber units are described in Section 9.4.2.2.2.
NONESSENTIAL FANS - There are four nonessential fans in the auxiliary building HVAC system. The auxiliary/fuel building normal exhaust fans, and the main steam tunnel exhaust fans are centrifugal fans with electric motor drivers. The access tunnel transfer fan is a propeller fan with an electric motor driver. The auxiliary building fume hood booster fan is a vane axial fan with an electric motor driver. SAFETY-RELATED FANS - The auxiliary building HVAC systems contain no safety-related fans. The emergency exhaust system fans are described in Section 9.4.2.2.2. SUPPLEMENTAL HEATING - Supplemental heating is supplied by electric
duct heaters and electric and hot-water unit heaters. Electric duct heaters are used in the scrubber makeup air ductwork in
the decontamination room to ensure that minimum temperatures are maintained during winter operation.
The hot-water and electric unit heaters are located throughout the auxiliary building to provide supplemental heating. Each unit heater consists of a coil and a fan with an electric motor driver.
FIRE DAMPERS - Fire dampers are located between fire barriers, as necessary, to maintain the fire ratings of the barriers. Dampers are
the 3-hour-rated curtain type.
9.4-37 Rev. 13 WOLF CREEK ISOLATION DAMPERS - Where a means of system isolation is required, parallel-blade-type dampers are utilized. The type of operator employed is dependent upon the specific design and/or usage requirements.
FLOW CONTROL DAMPERS - Opposed-blade-type or single-blade-type dampers are utilized, as necessary, to provide a means of system balancing. In
general, these are manually operated. However, some utilize power operators to allow compensation for changes occurring during system operation.
BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to maintain the proper direction of flow.
TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects of extreme wind or tornado conditions is required. These dampers close with the flow produced by the differential pressure
associated with tornadoes or high winds.
9.4.3.2.3 System Operation
GENERAL - The auxiliary building is served by two outside air supply units, one which serves all areas of the auxiliary building, except the
main steam tunnel, and one which serves only the main steam tunnel. Recirculation units (both essential and nonessential) are utilized
throughout the building to supplement the outside air units' cooling (nonessential), provide cooling for the safety-related equipment, and minimize ductwork requirements.
Three exhaust systems serve the auxiliary building. The main steam tunnel exhaust system takes suction from the main steam tunnel and
discharges to the atmosphere through the unit vent. The auxiliary/fuel
building normal exhaust system takes suction from the potentially contaminated areas of the auxiliary building and processes it through a charcoal adsorber train prior to release through the unit vent. The
emergency exhaust system exhausts from the auxiliary building following a LOCA and processes the air through a charcoal adsorber train prior to releasing it through the unit vent. The emergency exhaust system also
exhausts a limited amount of air from the fuel building following a LOCA to prevent excessive negative pressure in the Auxiliary building.
The scrubbers exhaust through hoods over the decontamination tanks to remove the vapors which evolve from the tanks when in use. This air is cleaned and discharged to the auxiliary/fuel building normal exhaust
system.
9.4-38 Rev. 27 WOLF CREEK The access tunnel transfer fan transfers air from the auxiliary building to the clean side of the radwaste tunnel. This air is then exhausted through the hot side of the tunnel by the auxiliary/fuel building normal exhaust. Cooling water for the nonessential air handlers is supplied by the central chilled water system (Section 9.4.10), and cooling water for the safety-related room coolers is supplied by the essential service water system (Section 9.2.1).
Hot water for the supply air unit and the unit heaters is supplied by the plant heating system (Section 9.4.9).
Discussed below are the power generation operations and emergency operations of
the auxiliary building HVAC systems. Shutdown operations are identical to the
power generation operations. POWER GENERATION OPERATION - Operation of the auxiliary building supply system, the auxiliary/fuel building normal exhaust system, the main steam tunnel supply
system, and the main steam tunnel exhaust system is initiated manually from the control room. These systems operate continuously during normal plant operations.
The auxiliary building supply air unit draws in outside air, filters it through
low efficiency particulate filters, either cooling with a chilled-water coil or
heating with a hot-water coil, and distributes the conditioned air to the separate floors of the auxiliary building. In addition to the outside air cooling, local cooling is provided by supplemental fan-coil units which utilize
chilled water coils for cooling. Space heating is provided by the outside air
unit and unit heaters.
The heating or cooling mode of operation of the outside air supply unit is a function of the outside air temperature only. When the outside air temperature exceeds 65 F, conditioned outside air is supplied to the building. When the
outside air temperature is between 65 and 50 F, unconditioned outside air is
supplied to the building, and when the outside air temperature is below 50 F, the
9.4-39 Rev. 13 WOLF CREEK heating system is operational. These operations are controlled by temperature switches, one associated with each coil, located in the ductwork upstream of
the coils which sense the outside air temperature and function accordingly. When the outside air temperature rises above 65°F, the temperature switch associated with the cooling system activates the supply unit cooling control
system. This control system then functions to maintain a constant supply air
temperature of 60°F by modulating the flow of chilled water to the coil. While the outside air temperature is between 65 and 50°F, the supply unit
continues to supply unconditioned air to the building.
When the outside air temperature falls below 50°F, the temperature switch associated with the heating coil activates the supply unit heating control system. The supply unit heating coil is supplied from a secondary hot-water
loop. This arrangement is provided to prevent the possible freezeup of the coil when the outside air temperature falls below 32°F.
The supply unit heating control system consists of temperature transmitters, located on each level, which sense the corridor temperature and transmit a
corresponding signal to a single temperature controller. When any one of these
signals indicates that a corridor temperature is below 60°F, the temperature
controller then increases the amount of heating from the supply unit heating coil to maintain a minimum of 60°F on all levels. The temperature controller controls the secondary loop temperature by regulating the amount of hot water
which enters the secondary loop of the heating coil.
A temperature switch is provided in the supply air duct downstream of the outside air unit. This temperature switch will trip the supply unit, should the supply temperature drop below 40°F, to protect the coils from freezing.
The previous description of the operation of an auxiliary building air unit is, in general, applicable to the main steam tunnel supply air unit. However, cooling is provided as a function of exhaust temperature only and the setpoint for heating may be higher than 50°F to minimize the variation of the area temperature. The auxiliary building supply air system and the main steam enclosure supply air system intakes are in a penthouse atop the auxiliary building, which is located approximately 15 feet below and 135 feet horizontally from the diesel exhaust discharge point. This separation is sufficient to provide significant dilution of the diesel exhaust gases; therefore, operation of the diesel during
normal plant operations will result in no ingestion of exhaust gases into the auxiliary building.
9.4-40 Rev. 8 WOLF CREEK The basement corridor fan coil unit, the normal charging pump room fan coil unit, the hot instrument and hot machine shop fan coil unit, the ground floor
fan coil unit, the component cooling water pump room's fan coil units, and the electrical equipment room fan coils units operate to provide supplemental cooling of the auxiliary building. The operation of these units is controlled by a temperature switch located in the respective room and/or area served. This
switch activates the unit fan when the room or area temperature exceeds the
design limits. The basement corridor fan coil unit, the normal charging pump fan coil unit, the ground floor corridor fan coil unit, the electrical equipment room fan coil unit (SGL02), and the component cooling water pump room
fan coil unit temperature switches are set to initiate operation of the unit
when the room temperature increases and to stop the unit when the room
temperature falls below setpoint with the exception of the temperature switch for the electrical equipment room fan coil unit (SGL02). The second electrical equipment room fan coil unit (SGL20) operates continuously with the chilled
water flow regulated to control the cooling of the unit. The two electrical equipment room fan coil units (SGL02 and SGL20) operate in parallel to maintain the room temperature at or below 80 F. The hot machine/hot instrument shop fan coil unit switches are set to initiate operation of the cooler when the space temperature exceeds 80°F and to stop the cooler when the temperature falls
below 75°F. The NCP room cooler is oversized for normal plant operating conditions (normally <104 F ambient). To prevent high cycling of the cooling fan and premature tripping of the motor thermal overloads, cooling flow is reduced to the low end of the allowable flow band and set the room thermostat low enough to ensure extended fan run times. This ensures the room temperature does not exceed the high and low limits during pump operation and when the pump is secured.
Supplemental heating is provided by unit heaters located in the basement corridor, the hot machine shop, the decontamination room, the hot instrument
shop, the intermediate floor corridor, the auxiliary building operating floor HVAC equipment room, the containment personnel access area, the boric acid storage tanks area, pipe chase and filter valve gallery. Each heater is sized
for its specific location and is thermostatically controlled to maintain the
space design temperature requirements of 60°F or above. The boric acid storage
tank area, pipe chase, and the boric acid filter valve gallery unit heaters maintain a space temperature of 75°F or above to prevent crystallization of the 4-weight-percent boric acid in the tanks and/or lines.
The auxiliary building supply air unit intake, the auxiliary/fuel building normal exhaust system discharge and the auxiliary/fuel building normal exhaust system, the emergency exhaust system, the access tunnel transfer fan, the hot machine shop/hot instrument shop fan coil unit, and the decontamination tanks
exhaust scrubbers exhaust line penetrations of the auxiliary building
boundaries contain tornado dampers capable of withstanding the effects of
extreme wind or tornado conditions (3 psi total at a rate of 2 psi per second per Regulatory Guide 1.76). These dampers close with the flow produced by the differential pressure associated with tornadoes or high winds. The dampers
located in those systems whose normal flow is in the same direction as would be
the flow produced by the differential pressure are spring loaded to prevent
closure during normal system operations. Missile barriers are provided externally to the isolation system to prevent propagation of a tornado missile.
9.4-41 Rev. 24 WOLF CREEK Each pump room is provided with an individual cooler. Each penetration room cooler serves one electrical penetration room. The pump room coolers and the
penetration room coolers are completely contained within their respective spaces and form closed loop systems. The pump room coolers may be operated manually during occupation of the room
but normally operate only in conjunction with the pump motors they serve. The
pump room coolers start automatically upon initiation of the respective pumps. Operation of the penetration room coolers is controlled by a handswitch or SIS.
One of the two auxiliary/fuel building normal exhaust fans runs continuously
during normal plant operations. The standby fan is designed to start either automatically on failure of the operating fan or on manual initiation.
A differential pressure indicator controller, located across the charcoal adsorber, modulates a damper downstream of the filter train to maintain a
constant system resistance as the filters load up. This control arrangement will assure a constant system flow. Exhaust hoods are provided over the decontamination spray booth located in the decontamination room and the sample
sink located in the auxiliary building sample room. Exhaust flow through the
sample room hood is constant whether the sink is in use or not.
An individual exhaust hood is provided over each of the decontamination tanks. During normal operation, exhaust is limited to one larger hood (3200 cfm) or
two small hoods (1600 cfm each). Use is limited due to the quantities of air
required for fume control over the tanks. Operation of the scrubbers is
provided manually by means of handswitches located in the decontamination room. When each scrubber is in use, a corresponding damper, located in ductwork downstream of an outside air louver, opens to provide makeup air for the scrubber. Located in the makeup air ductwork is a heating coil which, when the
outside air temperature is below 60 F, tempers the entering air to maintain it
at a minimum of 60 F. An audible alarm is provided in the decontamination room and a pressure switch is provided in the water supply line for each scrubber to prevent operation of the scrubber and to alarm the operator if there is no
water flow to the scrubbers. This feature is provided to prevent damage to
those auxiliary/fuel building normal exhaust system components located
downstream of the scrubbers.
9.4-42 Rev. 0 WOLF CREEK The main steam tunnel exhaust system operation is initiated manually. One of the two main steam enclosure building exhaust fans runs continuously during
normal plant operation. The standby fan is designed to start either automatically on failure of the operating fan or on manual initiation. The motor-operated discharge isolation dampers (one associated with each fan) operate in conjunction with their corresponding fans.
The auxiliary building ductwork infiltration air isolation system consists of two dampers in a series arrangement in each system which penetrates the auxiliary building and must be isolated following a LOCA. Each damper of a
pair is powered from a separate Class IE source to assure closure and will
close automatically upon receipt of an SIS.
In the event of a radioactive release from a fuel handling accident in the fuel building, the portion of the auxiliary/fuel building normal exhaust system
serving the fuel building is automatically isolated, and the operating auxiliary/fuel building normal exhaust fan is manually switched to the low
speed to maintain the exhaust flow from the auxiliary building. The charcoal adsorber train is monitored for charcoal bed temperature. Should
the bed temperature exceed 200 F, an alarm is received in the control room to alert the operators of excessive bed heating. Should the bed temperature continue to rise and exceed 300 F, a second alarm is received in the control room. To prevent backflow through the system, upstream isolation is provided
by a backdraft damper located at the inlet to the filter train. All
particulate and HEPA filter banks are provided with local differential pressure
indication and differential pressure switches which will alarm excessive
pressure drops via the plant computer. In the event of a fire, the auxiliary/fuel building normal exhaust system can
function to purge the auxiliary building of smoke.
Based on the outside air design conditions, as described in Table 9.4-1, design space heat loads, and operation of the auxiliary building HVAC systems as described above, the relative humidity in the auxiliary building will not
normally exceed 70 percent.
Following a loss of offsite power, certain areas of the auxiliary building will experience temperatures higher than their normal ambient design temperatures. These areas and their resultant temperatures are given in Table 3.11(B)-1. In
none of the affected areas does the temperature increase affect either the safe
operation or the ability to achieve and maintain the safe shutdown of the
plant. I&C HOT SHOP AIR-CONDITIONING UNIT - The I&C Hot Shop room air-conditioner is
manually initiated from the unit's control panel. The room air-conditioner is
controlled by the unit thermostat, which cycles the compressor. The room
temperature is limited to a maximum of 78 F. Room occupants can manually
adjust the unit's fan speed and may switch from complete recirculation to 100% exhaust as necessary.
9.4-43 Rev. 14 WOLF CREEK EMERGENCY OPERATION - In the event of a LOCA, those systems which penetrate the auxiliary building boundaries (excluding the emergency exhaust system) are
automatically isolated and the emergency exhaust system is automatically started. The emergency exhaust system takes suction on all levels of the auxiliary building and processes the exhaust air through the emergency exhaust charcoal adsorption train (Section 9.4.2) for cleanup prior to monitoring and
discharge through the unit vent. The emergency exhaust system also exhausts a
limited amount of air from the fuel building following a LOCA to prevent excessive negative pressure in the auxiliary building.
Following a LOCA, if the hydrogen purge system is used, the air purged from the
containment is ducted to the emergency exhaust system for processing and
release through the unit vent. To protect the ductwork system from over-pressurization and to provide a means of maintaining the hydrogen purge system flow within design limits, a globe valve is located downstream of the outboard
hydrogen purge containment isolation valve. The portion of the system between the outboard isolation valve and the globe valve is piping, and the portion
after the globe valve is ductwork. The valve is located in the south electrical penetration room and is locked throttled to provide a purge flow of 100 scfm at a containment pressure of 4 psig. Provided immediately downstream of the globe
valve is a pressure indicator for monitoring system pressure.
9.4.3.3 Safety Evaluations Safety evaluations are numbered to correspond to the safety design bases in
Section 9.4.3.1. SAFETY EVALUATION ONE - The safety-related portions of the auxiliary building HVAC systems are located in the auxiliary and fuel 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 auxiliary building
HVAC systems 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
9.4-44 Rev. 19 WOLF CREEK SAFETY EVALUATION THREE - The design of the auxiliary building HVAC systems provides complete redundancy and, as indicated by Table 9.4-9, 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. SAFETY
EVALUATION FOUR - The pump room coolers, the penetration room coolers, the emergency exhaust system, and the auxiliary building isolation provisions are initially tested with the program given in Chapter 14.0. Periodic in-service functional testing is done in accordance with Section 9.4.3.4. Section
6.6 provides
the ASME Boiler and Pressure Vessel Code, Section XI requirements
that are appropriate for the pump room and the penetration room coolers.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portions
of the auxiliary building HVAC systems. All the power supplies and control functions necessary for safe function of the pump room coolers, penetration
room coolers, emergency exhaust system, and auxiliary building isolation provisions are Class IE, as described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - Section 9.4.3.2.3 describes the provisions made to
assure the isolation of the auxiliary building following a DBA.
SAFETY EVALUATION SEVEN - The emergency exhaust system maintains a negative pressure in the auxiliary building of not less than 1/4 inch w.g., following a
LOCA. The system collects and processes potential ECCS leakages and the
effluent purged from the containment via the hydrogen purge system. The system
is monitored for radioactivity upstream of the filter adsorber unit prior to release through the unit vent. SAFETY EVALUATION EIGHT - The ESF pump rooms coolers have sufficient cooling
capacity to maintain the ESF pump rooms at 122 F or below when the ESF pumps
are operating at rated load. 9.4.3.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. Filters and adsorbers
located within the auxiliary/fuel building normal exhaust filter adsorber unit are tested in the manufacturer's shop, after initial installation, and subsequent to each filter or adsorber change. Interim tests and inspections
are performed annually in accordance with the requirements of Regulatory Guide
1.140.
9.4-45 Rev. 0 WOLF CREEK All charcoal adsorbers are factory tested in accordance with RDT-M-16-IT to exhibit a decontamination efficiency of no less than 99.9 percent for elemental
iodine and 98 percent for methyl iodide. Sample charcoal canisters are tested for impregnant efficiency in an independent laboratory, using radiomethyl iodide tracers. Inplace testing is performed with a suitable refrigerant in accordance with the procedures set forth in ANSI N510.
Prefilters do not undergo factory or inplace testing since no credit is taken for removal of particulates.
HEPA filters are factory tested with monodispersed DOP aerosol to demonstrate a
minimum particulate removal efficiency of no less than 99.97 percent for 0.3
micron particulates. Inplace leak testing is carried out with cold polydisperse DOP. Testing is in accordance with the procedures set forth in ANSI N510.
All safety-related systems and boundary isolation provisions undergo preoperational testing prior to plant startup. All nonsafety-related systems
underwent acceptance testing prior to plant startup. One fan in each group of identical safety-related fans (pump room coolers and
penetration room coolers) is tested in accordance with AMCA Standards. All
other fans are AMCA rated.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.3.5 Instrumentation Applications The auxiliary/fuel building normal exhaust fans, the emergency exhaust fans, the main steam enclosure building exhaust fans, the auxiliary building supply air unit fan, and the main steam enclosure building supply air unit fan are operable from the control room and have indication of operation status in the control room. All other fans are locally operable by manual and/or automatic means by association with a pump start or thermostat.
An indication of the position of all isolation dampers is provided in the
control room.
Thermostats, located in both the various levels of the building and the HVAC ductwork, control space temperatures.
The amount of filter loading for all filters associated with both the air
handlers and the filter adsorbers is available at the unit.
9.4-46 Rev. 13 WOLF CREEK All instrumentation provided with the auxiliary/fuel building normal exhaust filter adsorber unit is as required by Regulatory Guide 1.140.
High temperature alarms for each of the ESF pump rooms and the electrical penetration room are provided in the control room.
Low water flow to the scrubbers is alarmed in the decontamination room. The
scrubbers are operable from the decontamination room.
9.4.4 TURBINE
BUILDING HVAC
The turbine building HVAC systems consist of the main building heating and
ventilation systems, the lube oil room ventilation and heating system, the computer room HVAC system, the instrument shop HVAC system, the condenser air removal filtration system, the battery room ventilation and cooling system, and
the EHC cabinet room air-conditioning system, the Oxygen control and pH control chemical storage room air-conditioning system and the Turbine Deck Office Mezzanine room air-conditioning system. The main building ventilation system provides outside air for ventilation and
cooling for each level of the turbine building. The main building ventilation
system serves the turbine building, the communication corridor, and the battery
rooms. The lube oil room ventilation and heating system provides outside air for
ventilation and cooling or heating, as required, for the equipment within the
lube oil room.
The computer room HVAC system provides a suitable environment for the equipment and personnel comfort.
The instrument shop HVAC system provides a suitable environment for personnel
comfort. The condenser air removal filtration system collects and processes the
noncondensables from the condenser.
The battery room cooling and ventilation system serves to dilute the hydrogen emitted from the batteries.
The EHC cabinet room air-conditioning system provides a suitable environment
for the equipment.
The Turbine Deck Office Mezzanine room air-conditioning system provides a suitable environment for personnel comfort. 9.4.4.1 Design Bases
9.4.4.1.1 Safety Design Bases The turbine building HVAC systems serve no safety function; however, those dampers and ductwork in the condenser air removal filtration system which are
required to provide isolation of the
9.4-47 Rev. 16 WOLF CREEK auxiliary building are safety-related and are required to function following a DBA and to achieve and maintain the plant in a post accident safe shutdown
condition. SAFETY DESIGN BASIS ONE - The auxiliary building isolation provisions are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2). The isolation
provisions for the auxiliary building remain functional after a safe shutdown earthquake and perform their intended function following a postulated hazard, such as internal missiles, or pipe break (GDC- 4).
SAFETY DESIGN BASIS TWO - The safety functions of the condenser air removal
filtration system can be performed, assuming a single active component failure coincident with the loss of offsite power.
SAFETY DESIGN BASIS THREE - Active components of the condenser air removal filtration system are capable of being tested during plant operation.
SAFETY DESIGN BASIS FOUR - The safety-related auxiliary building isolation provisions are consistent with the quality group classification assigned by
Regulatory Guide 1.26 and with the seismic category assigned by Regulatory
Guide 1.29. The power supply and control functions must be in accordance with
Regulatory Guide 1.32. 9.4.4.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The main building ventilation system
supplies outside air for ventilation and cooling to maintain the turbine building average ambient temperature below 110 F. The main building heating system is designed to maintain the average ambient temperature above 60 F.
POWER GENERATION DESIGN BASIS TWO - The lube oil room ventilation and heating
system supplies outside air for cooling or heating, as required, to maintain the lube oil room average ambient temperature between 60 F and 110 F.
POWER GENERATION DESIGN BASIS THREE - The Operations Relief Area HVAC system maintains the average room temperature between 60 F and 72 F and a relative humidity of 50 10 percent. Full-capacity, redundant air-conditioning units are provided. The majority of plant computer equipment has been removed from the computer room. The area has been converted to office space. The HVAC system has been modified to accommodate the reduced heat load now in the area. The HVAC system for the room is controlled by a thermostat and is set by personnel occupying the room.
POWER GENERATION DESIGN BASIS FOUR - The instrument shop HVAC system maintains the average room temperature between 60 F and 78 F.
9.4-48 Rev. 21 WOLF CREEK POWER GENERATION DESIGN BASIS FIVE - The ventilation exhaust systems are designed to meet the requirements of the discharge concentration limits of 10
CFR 20 and the as-low-as-reasonably-achievable dose objective of 10 CFR 50, Appendix I. No filtration of the main turbine building exhaust is required. The condenser air removal filtration system charcoal adsorption train complies with Regulatory Guide 1.140 to the extent discussed in Table 9.4-3.
POWER GENERATION DESIGN BASIS SIX - The condenser air removal filtration system monitors radioactivity in accordance with Regulatory Guide 1.21.
POWER GENERATION DESIGN BASIS SEVEN - The battery room cooling and ventilation
system maintains the average ambient temperature between 60°F and 90°F.
POWER GENERATION DESIGN BASIS EIGHT - The EHC cabinet room air-conditioning and ventilation system limits the temperature to a maximum of 80°F.
POWER GENERATION DESIGN BASIS NINE - The Oxygen control and pH control chemical Storage Room Air-Conditioning System provides adequate cooling to maintain the room temperature below 75°F. 9.4.4.2 System Description
9.4.4.2.1 General Description Figure 9.4-4 shows the flow diagram of the turbine building HVAC systems.
The main building ventilation system utilizes outside air as a cooling medium.
Air is distributed throughout the turbine building and communication corridor by supply units located on the periphery of the building. Outside air, supplied directly, provides cooling for the summer months. During the winter
months of plant operation, a reduced quantity of outside air is required for
building cooling. The outside air is mixed with turbine building (recirculated) air for tempering during this mode of operation. During plant shutdown in the winter, heating is provided by strategically located electric and hot water unit heaters.
The turbine building air is exhausted to the atmosphere by exhaust fans located
within louvered penthouses on the roof. Smoke removal in the turbine building in the event of a fire is discussed in
Section 9.5.1.
The battery rooms are ventilated by a branch duct from the outside air supply units. A chilled-water coil is provided within the supply air duct to maintain the temperature conditions within these rooms. The rooms are pressurized
slightly by the supply air
9.4-49 Rev. 14 WOLF CREEK and relieved through a transfer grille into the turbine building. The rate of supply air into the battery rooms is sufficient to dilute the hydrogen emitted
from the batteries to a value well below the flammability, and, hence, the explosive limits.
The lube oil room ventilation system utilizes outside air as a cooling medium.
A heating coil is provided for tempering the outside air during winter plant
operation and plant shutdown. The lube oil room is exhausted to the atmosphere.
The computer room HVAC system utilizes chilled water for cooling and
dehumidification, a hot-water coil for heating, and a humidifier. The computer
room air-conditioning unit operates in a complete recirculation mode. Fresh air is provided by a branch duct from an outside air supply unit (servicing the communication corridor) and is relieved through a transfer grille.
The condenser air removal filtration system collects and processes the
noncondensables from the condenser (through the mechanical vacuum pumps) and other potential sources of radioactivity. The effluents from these components are diluted with turbine building air, approximately 10 to 1, upstream of the
filtration unit to dilute the concentration of noble gases, and moisture
content. The condenser air removal filtration system is monitored for
radioactivity upstream of the adsorber train. Redundant fans are provided to assure system reliability. The condenser air removal filtration system discharges through the unit vent after processing through the adsorber train.
Based on the source terms provided in Section 11.1 and the dose evaluation
provided in Section 11.3 the exhaust systems meet the objective of 10 CFR 50, Appendix I, and the limits of 10 CFR 20.
The instrument shop air-conditioning system utilizes chilled water for cooling.
The instrument shop air-conditioning unit operates in a complete recirculation
mode. Fresh air is provided by a branch duct from an outside air supply unit (servicing the communication corridor) and is relieved through a transfer grille.
The EHC cabinet room air-conditioning system utilizes two direct - expansion (DX) rooftop air conditioners for cooling. Each EHC cabinet room air conditioner has a manual damper for operation in full recirculation mode or with a mix of up to 50% outside air.
The Oxygen control and pH control chemical storage room air-conditioning system
utilizes a room air-conditioner for cooling. The Turbine Deck Office Mezzanine room air-conditioning system utilizes three
room air-conditioners for cooling.
9.4.4.2.2 Component Description
Codes and standards applicable to turbine building ventilation systems are listed in Tables 3.2-1 and 9.4-10.
9.4-50 Rev. 27 WOLF CREEK The turbine building ventilation systems are designed and constructed in accordance with codes and standards comparable with quality group D.
AIR HANDLING UNITS - Each main building supply unit consists of a fan section, a medium-capacity filter box (with provisions for future filters), and an
electric motor driver. Some units are also equipped with mixing boxes.
The communication corridor supply unit consists of a fan section, a medium-capacity filter box, a mixing box, and an electric motor driver.
The lube oil room supply air unit consists of particulate filters, a hot-water
heating coil, a centrifugal fan, and an electric motor driver.
Each computer room fan-coil unit consists of particulate filters, a chilled-water cooling coil, a hot-water heating coil, a humidifier, a centrifugal fan, and an electric motor driver.
The instrument shop fan-coil unit consists of particulate filters, a chilled-water cooling coil, a centrifugal fan, and an electric motor driver. An electric duct heater provides heating.
Each EHC cabinet room air conditioner consists of particulate filters, a refrigerant coil, a direct drive plenum fan, and an electric motor driver.
ROOM AIR CONDITIONERS - The Oxygen control and pH control chemical storage room air-conditioner and the Turbine Deck Office Mezzanine room air-conditioner
consists of a particulate filter, a centrifugal blower, an electric motor
driver, an evaporator, a compressor, and a condenser. COOLING COILS - Each battery room cooling coil is chilled-water type.
FILTER UNIT - The condenser air removal filtration unit consists of moderate
efficiency prefilters, HEPA filters, and charcoal adsorption beds. FANS - The main building exhaust fans are vaneaxial fans with electric motor
drivers.
The lube oil room and condenser air removal system exhaust fans are centrifugal fans with electric motor drivers.
The toilet areas and the elevator machine room exhaust fans are centrifugal
fans with electric motor drivers.
UNIT HEATERS - Hot-water unit heaters are used to provide heating in the main building. Electric unit heaters provide heating in the stairwells. Each unit
heater consists of a coil and a fan with an electric motor driver.
9.4-51 Rev. 27 WOLF CREEK FIRE DAMPERS - Fire dampers are located in fire barriers as necessary to maintain the fire rating of the barriers. Dampers are the 3-hour-rated curtain
type. ISOLATION DAMPERS - Where a means of system isolation is required, parallel-
blade-type dampers are utilized. The type of operator used is dependent upon
the specific design and/or usage requirements.
FLOW CONTROL DAMPERS - Opposed-blade-type dampers are utilized, as necessary, to provide a means of system balancing.
BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to maintain
the proper direction of air flow. TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects
of extreme wind or tornado conditions is required. These dampers close with
the flow produced by the differential pressure associated with the tornado or
high winds. Further information regarding the turbine building ventilation system
components is provided in Table 9.4-10.
9.4.4.2.3 System Operation MAIN BUILDING HEATING AND VENTILATION SYSTEM - During the summer mode of
operation, the system utilizes 100 percent outside air for cooling, with all
supply units operating. During the winter mode of system operation with the
plant operating, only selected outside air supply units are operating, and these are in a partial recirculation mode. Likewise, the number of exhaust fans operating is reduced to correspond to the outside air requirements during
this mode of system operation. During the winter mode of operation, six supply
units, two per floor, operate in a partial recirculation mode in conjunction
with the two small roof exhaust fans. Unit heater operation may be initiated to provide supplemental heating in low equipment heat load areas. The turbine building is maintained between 60 and 110 F.
Unit heaters are provided for building heating during plant shutdown. Hot
water is utilized as the heating medium. The unit heaters are controlled by local thermostats which energize the heater whenever the building temperature reaches 60 F.
BATTERY ROOM COOLING AND VENTILATION SYSTEM - The branch ducts from the main
building ventilation system to the battery rooms are provided with chilled-water coils for cooling. The battery rooms are maintained between 60 and 90 F.
9.4-52 Rev. 27 WOLF CREEK LUBE OIL ROOM VENTILATION AND HEATING SYSTEM - The lube oil room is served by independent supply and exhaust ventilation systems. The lube oil supply system
takes suction from the outside and supplies it directly to the space. Cooling is accomplished by the outside air. A hot-water heating coil, located in the supply system, provides the required heating. The lube oil room exhaust system takes suction from the space and discharges it directly to the atmosphere. The
lube oil room is maintained between 110 and 60 F.
While the outside air temperature is above 60 F, the lube oil supply air unit continues to operate, supplying outside air to the building.
When the outside air temperature falls below 60 F, the temperature switch
associated with the heating coil activates the supply unit heating control system. The supply unit heating coil is supplied from a secondary hot-water loop which is, in turn, supplied from the plant heating system.
The lube oil room supply unit heating control system consists of a temperature
controller. When the temperature is below 60 F, the temperature controller modulates the amount of heating accordingly. The temperature controller controls the secondary loop temperature by regulating the amount of hot water
which enters the secondary loop of the heating coil.
A temperature switch located downstream of the coil is provided to trip the supply air unit, should the supply air temperature drop below 40 F, to protect the coil from freezing.
Operations Relief Area HVAC SYSTEM - The majority of plant computer equipment has been removed from the computer room and this area has been turned into an office space. The HVAC for this room is controlled by a thermostat. The thermostat controls cooling and heating by modulating the respective three-way
mixing valves. The room temperature will be controlled by the personnel
occupying this room.
INSTRUMENT SHOP HVAC SYSTEM - The instrument shop air conditioning unit is controlled by a room thermostat which cycles the supply fan. Cooling is
accomplished by circulating chilled water through the cooling coil. The
required heating is provided by an electric duct heater located in the outside
air branch duct serving the room. The room temperature is maintained between 60 and 78 F.
9.4-53 Rev. 21 WOLF CREEK CONDENSER AIR REMOVAL FILTRATION SYSTEM - The condenser air removal filtration system is manually initiated by a local handswitch. One of the two exhaust
fans runs continuously during normal operations. The standby fan is designed to start either automatically on failure of the operating fan or on manual initiation.
A differential pressure-indicating controller, located across the charcoal
adsorber, modulates a damper downstream of the filter train to maintain a constant system resistance as the particulate filters load up. This control arrangement will assure a constant system flow.
EHC CABINET ROOM AIR-CONDITIONING AND VENTILATION SYSTEM - The EHC cabinet room
air conditioners are manually initiated from a single control panel. The room air conditioners are controlled by the temperature switches, which cycle the compressors. The room temperature is limited to a maximum of 80 F. OXYGEN CONTROL AND PH CONTROL CHEMICAL STORAGE ROOM AIR-CONDITIONING SYSTEM - The Oxygen control and pH control chemical storage room air-conditioning is manually initiated from the unit control panel. The room air-conditioner is
controlled by the unit thermostat, which cycles the compressor. The room
temperature is limited to a maximum of 75°F.
TURBINE DECK OFFICE MEZZANINE ROOM AIR-CONDITIONING SYSTEM - The Turbine Deck Office Mezzanine room air-conditioners are manually initiated from the unit
control panel. The room air-conditioners are controlled by the unit
thermostats, which cycle the compressors. The room temperature is controlled
to ensure personnel comfort. 9.4.4.3 Safety Evaluation
Safety evaluations are numbered to correspond to the safety design bases in
Section 9.4.4.1. SAFETY EVALUATION ONE - The safety-related portions of the condenser air
removal filtration system are located in the auxiliary 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 this building. The safety-related portions of the
condenser air removal filtration 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 safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
SAFETY EVALUATION TWO - The design of the safety-related portions of the
condenser air removal filtration system provides complete redundancy, and no single failure compromises the system's safety functions. All vital power can be supplied from either onsite or offsite power system as described in Chapter
8.0. SAFETY
EVALUATION THREE - The auxiliary building isolation provisions are initially tested with the program given in Chapter 14.0. Periodic in-service functional testing is done in accordance with Section 9.4.4.4.
9.4-54 Rev. 27 WOLF CREEK SAFETY EVALUATION FOUR - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portions
of the condenser air removal filtration system. All the power supplies and control functions necessary for safe function of the auxiliary building isolation provisions are Class 1E, as described in Chapters 7.0 and 8.0.
9.4.4.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. Filters and adsorbers for the condenser air removal filtration system are tested in the shop, after initial installation, and subsequent to each filter
or adsorber change. Following installation of the filters and adsorbers, interim tests and inspections are performed in accordance with the requirements of Regulatory Guide 1.140.
Charcoal adsorbers are qualified per Regulatory Guide 1.140. Charcoal batch
samples are factory tested with radiomethyl iodide tracers at 25 C and 70 percent relative humidity to exhibit a decontamination efficiency of no less than 99.9 percent for elemental iodine and 98 percent for methyl iodide.
Sample charcoal is tested for impregnant efficiency in an independent
laboratory using radiomethyl iodide tracers. Inplace testing is performed with a suitable refrigerant, in accordance with the procedures set forth in ANSI N510, to check for bed bypass leakages.
Prefilters do not undergo factory or inplace testing since no credit is taken
for removal of particulates.
HEPA filters are factory tested with monodispersed DOP aerosol to demonstrate a minimum particulate removal efficiency of no less than 99.97 percent for 0.3
micron particulates. Inplace leak testing is carried out with cold
polydispersed DOP. Testing is in accordance with the procedures set forth in
ANSI N510. Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
All safety-related boundary isolation provisions underwent preoperational testing prior to plant start-up.
9.4.4.5 Instrumentation Applications Indication of condenser air removal system exhaust fan operational status is provided in the control room.
9.4-55 Rev. 0 WOLF CREEK An indication of the position of all building isolation dampers associated with the condenser air removal filtration system is provided in the control room.
A temperature-sensor element is provided for the charcoal adsorber to indicate excessive bed heating.
An alarm is provided in the control room to indicate high radiation in the
condenser air removal filtration system. All instrumentation provided with the filter/adsorber unit is as required by
Regulatory Guide 1.140.
Thermostats, located in the various levels, control space temperature. Local differential pressure indication is provided across each particulate
filter bank.
9.4.5 RADWASTE
BUILDING HVAC The radwaste building HVAC system functions to provide a suitable atmosphere
for equipment and personnel occupation.
9.4.5.1 Design Bases 9.4.5.1.1 Safety Design Bases
This system serves no safety function. Failure of the system does not affect safe shutdown of the plant. 9.4.5.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The radwaste building HVAC system provides
fresh air ventilation at a rate of at least 0.1 cfm per square foot of floor area. POWER GENERATION DESIGN BASIS TWO - The radwaste building HVAC system maintains
the temperature in the control rooms, the sample laboratory, and drumming area
between 60 and 85 F. All other areas of the building are maintained between 60 and 104 F.
POWER GENERATION DESIGN BASIS THREE - The exhaust system inhibits exfiltration
by exhausting approximately 10 percent more air than is being supplied to the
building, the difference being made up by infiltration into the building.
9.4-56 Rev. 0 WOLF CREEK POWER GENERATION DESIGN BASIS FOUR - The ventilation exhaust system is designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and
the ALARA dose objective of 10 CFR 50, Appendix I. The charcoal adsorption train complies with Regulatory Guide 1.140 to the extent discussed in Table 9.4-3. 9.4.5.2 System Description 9.4.5.2.1 General Description The radwaste building heating, ventilating, and air-conditioning system shown in Figure 9.4-5 consists of fans, heating coils, cooling coils, a filter train, and its associated ductwork, dampers, and controls. Local unit heaters are used to provide supplemental heating. Fan-coil units are used for cooling the evaporator rooms, control rooms, and the sample laboratory. Local fan-coil
units are used to provide additional supplemental cooling.
The radwaste building is served by an outside air supply system which provides fresh cooled or heated air to each of the various levels of the building.
The radwaste building exhaust system takes suction from all levels of the
radwaste building, processes the exhaust through the filter adsorber train, and discharges it through the building vent. The building vent extends 10 feet above the roof of the radwaste building. Radiation monitors are provided to sample effluents. All exhaust air from the radwaste building is through the
radwaste building exhaust system.
The radwaste building exhaust system is designed to inhibit exfiltration. Air flow patterns are from areas of potentially lesser contamination to areas of greater contamination.
Based on the source terms provided in Section 11.1 and the dose evaluation
provided in Section 11.3, the exhaust system meets the objective of 10 CFR 50, Appendix I, and the limits of 10 CFR 20.
9.4.5.2.2 Component Description
Codes and standards applicable to the radwaste building HVAC system are listed in Tables 3.2-1 and 9.4-11.
The radwaste building HVAC systems are designed and constructed in accordance
with codes and standards comparable with quality group D.
9.4-57 Rev. 0 WOLF CREEK SUPPLY AIR UNIT - The radwaste building supply air unit consists of particulate filters, a hot-water heating coil, a chilled-water cooling coil, a centrifugal
fan, and an electric motor driver. RECIRCULATION UNITS - Fan-coil units are used to provide the cooling for the evaporator rooms, control rooms, and sample laboratory. Local fan-coil units
are used to provide supplemental cooling of the basement and ground floors.
Each unit consists of particulate filters, a chilled-water cooling coil, a centrifugal fan, and an electric motor driver. The main control room fan-coil unit is also provided with a hot-water heating coil to provide heating.
FILTER UNIT - The radwaste building filtration unit consists of moderate
efficiency prefilters, HEPA filters, and charcoal adsorption beds. FANS - The radwaste building exhaust fans are centrifugal fans with electric
motor drivers.
The access tunnel transfer fan is a propeller type with an electric motor driver. UNIT HEATERS - Hot-water unit heaters are used to provide supplemental heating.
Each unit heater consists of a coil and a fan with an electric motor driver.
FIRE DAMPERS - Fire dampers are located in fire barriers, as necessary, to maintain the fire rating of the barriers. Dampers are the 3-hour-rated curtain
type.
ISOLATION DAMPERS - Where a means of system isolation is required, parallel-blade-type dampers are utilized. The type of operator used is dependent upon the specific design and/or usage requirements.
FLOW CONTROL DAMPERS - Opposed-blade-type dampers are utilized, as necessary, to provide a means of system balancing. BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to maintain
the proper direction of air flow.
9.4.5.2.3 System Operation The radwaste building supply air unit is started manually. The supply air unit
draws in outside air, filters it through low efficiency particulate filters, either cooling with a chilled-water coil or heating with a hot-water coil, and
distributes the conditioned air to the various floors of the radwaste building.
9.4-58 Rev. 0 WOLF CREEK Fan-coil units are used for cooling the evaporator rooms, control rooms, and sample laboratory. Local fan-coil units are used to provide supplemental
cooling in the basement and ground floors. Space heating is provided by the outside air unit and local unit heaters. The heating or cooling mode of operation of the outside air supply unit is
controlled by the outside air temperature. When the outside air temperature
exceeds 80 F, chilled-water cooled outside air is supplied to the building. When the outside air temperature is between 48 and 80 F, unconditioned outside air is supplied to the building. When the outside air temperature is below 48
F, the heating system is operational. These operations are controlled by
temperature switches, one associated with each coil, located in the ductwork
upstream of the coils, which sense the outside air temperature and function accordingly.
When the outside air temperature rises above 80 F, the temperature switch associated with the cooling system activates the supply unit cooling control
system. This control system then functions to maintain a constant supply air temperature of 75 F by modulating the flow of chilled water to the coil.
While the outside air temperature is between 48 and 75 F, the supply unit
continues to supply unconditioned air to the building.
When the outside air temperature falls below 48 F, a temperature switch activates the supply unit heating control system. The supply unit heating coil
is supplied from a secondary hot-water loop which is, in turn, supplied from
the plant heating system.
The supply unit heating control system consists of temperature transmitters, located on each level, which sense the corridor temperature and transmit a corresponding signal to a single temperature controller. When any one of these
signals indicates that a corridor temperature is below 60 F, the temperature
controller then modulates the amount of heating accordingly. The temperature controller controls the secondary loop temperature by regulating the amount of hot water which enters the secondary loop of the heating coil.
A temperature switch is provided downstream of the coils. This temperature
switch trips the supply unit, should the supply air temperature drop below 40 F, to protect the coils from freezing.
Supplemental heating is provided by unit heaters located throughout the
building. Each unit heater is controlled thermostatically to maintain the
space design requirements of 60 F or above.
9.4-59 Rev. 0 WOLF CREEK The radwaste building exhaust system is started manually. One of the two radwaste building exhaust fans runs continuously during normal operations. The
standby fan is designed to start either automatically on failure of the operating fan or on manual initiation. The radwaste building supply air unit fan and the exhaust fans are interlocked
to prevent the supply fan from being started before the exhaust fan.
A differential pressure-indicating controller, located across the filter train, modulates a damper downstream of the filter train to maintain a constant system
resistance as the particulate filters load up. This control arrangement will
assure a constant system flow and thus assures that the system will always
inhibit exfiltration. The recirculation fan-coil units are initiated manually and are controlled by
their respective thermostats which cycle the supply fans.
9.4.5.3 Safety Evaluation The operation of the radwaste ventilation system is not required for the safe shutdown of the plant or for mitigating the consequences of a design basis
accident. 9.4.5.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
Filters and adsorbers are tested in the manufacturer's shop, after initial installation and subsequent to each filter or adsorber change. Following installation of the filters and the adsorbers, interim tests and inspections
are performed in accordance with the requirements of Regulatory Guide 1.140.
Charcoal adsorbers are qualified per Regulatory Guide 1.140. Charcoal batch samples are factory tested with radiomethyl iodide tracers at 25 C and 70 percent relative humidity to exhibit a decontamination efficiency of no less
than 99.9 percent for elemental iodine and 98 percent for methyl iodide.
Sample charcoal is tested for impregnant efficiency in an independent
laboratory using radiomethyl iodide tracers. Inplace testing is performed with a suitable refrigerant, in accordance with the procedures set forth in ANSI N510, to check for bed bypass leakage.
9.4-60 Rev. 0 WOLF CREEK HEPA filters are factory tested with monodispersed DOP aerosol to demonstrate a minimum particulate removal efficiency of no less than 99.97 percent for 0.3
micron particulates. Inplace leak testing is carried out with cold polydisperse DOP. Testing is in accordance with the procedures set forth in ANSI N510.
Prefilters do not undergo factory or inplace testing, since no credit is taken
for the removal of particulates. Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.5.5 Instrumentation Applications Fan-running lights and fan-trip alarms are provided in the control room.
All exhaust fans are operable from the control room. Thermostats, located at the various levels, control space temperatures.
Temperature-sensor elements are provided downstream of the charcoal adsorbers
to indicate excessive bed heating. Local differential pressure indication is
provided across each particulate filter bank. Alarms are provided in the control room to indicate high radiation in the
radwaste building and high temperature in the charcoal adsorber beds.
All instrumentation provided with the filtration units is as required by Regulatory Guide 1.140.
9.4.6 CONTAINMENT
HVAC
The containment HVAC system consists of the containment shutdown purge, containment minipurge, containment atmosphere control, control rod drive mechanism (CRDM) cooling, cavity cooling, pressurizer skirt cooling, elevator
machine room exhaust, the hydrogen mixing fans, and containment cooling system.
The containment shutdown purge system operates during reactor outages (mode 6 and Defueled) to supply outside air into the containment for ventilation and cooling or heating and may also be used, when the reactor is in the cold shutdown mode (mode 5), to reduce the concentration of noble gases within the containment prior to and during personnel access. The containment shutdown purge system is the preferred system for operation in modes 5 and 6 and Defueled to maintain a more suitable containment environment for personnel access due to the larger volume of air movement and heating and cooling capability.
9.4-61 Rev. 19 WOLF CREEK The containment minipurge system may be used during reactor power operations to reduce the concentration of noble gases within the containment prior to and
during personnel access or to equalize internal and external pressures. The containment minipurge system may also be used in modes 5 and 6 and Defueled, but must not be run in parallel with the containment shutdown purge system in order to prevent damage to fans and ductwork.
During RCS draindown activities, hydrogen from the pressurizer can be directed to the containment shutdown purge system or the containment minipurge system via temporary hoses. The containment atmospheric control system functions, as required, to reduce
the concentrations of radioiodine and particulate activity within the containment prior to and during personnel access or purging of the containment.
The CRDM cooling system maintains a suitable atmosphere during normal operation
within the CRDM shroud to protect and prolong the life of the CRDM coils.
The cavity cooling system maintains a suitable atmosphere within the reactor cavity during normal operation to protect the concrete, the ex-core neutron
detectors, and the neutron streaming shield.
The pressurizer cooling fan provides the necessary cooling of the lower portion of the pressurizer (skirt and heater connections) when the containment cooler serving that compartment is out of service.
The machine room exhaust fan provides the required ventilation of the
containment elevator machine room during normal plant operations. The functional bases for the containment cooling system and the hydrogen mixing
system are described in Sections 6.2.2 and 6.2.5, respectively.
9.4.6.1 Design Bases 9.4.6.1.1 Safety Design Bases
The containment HVAC systems are not safety related, with the exception of associated containment penetrations, the hydrogen mixing fans, and the containment cooling system. A complete description of the design of the containment cooling system and containment hydrogen mixing system is provided
in Section 6.2.2.2.
SAFETY DESIGN BASIS ONE - 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.
9.4-62 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS TWO - The containment purge system containment isolation valves are capable of rapid closure, following their respective DBA (FHA for
the shutdown purge valves and LOCA for the minipurge valves), to limit the escape of fission products from the containment. 9.4.6.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The containment shutdown purge system is designed to maintain a containment ambient air temperature between 50 and 90 F, when the reactor is shut down. The shutdown purge system supplies fresh air
into the containment at a rate of approximately one containment volume air
change per every 2 hours for fresh air ventilation.
POWER GENERATION DESIGN BASIS TWO - The Containment minipurge system has a design basis flow of 4,000 CFM which is based on continuous system operation with an assumed weekly occupancy of 5 hours for any one individual.
The assumptions used in determining the flow rate and resultant airborne activities are consistent with NUREG-0017, Reference 7, Section 12.2. Table 12.2-11 provides the assumed RCS specific activities, failed fuel percentages, RCS leakage rates, and partition factors. Table 12.2-12 provides the airborne
concentrations within the containment, assuming a continuous 4,000 cfm purge.
Any individual is allowed to be exposed to the concentrations of Table I Column I of 10 CFR 20, Appendix B, for 40 hours per week or to greater concentrations
for a corresponding lesser amount of time. The design bases for the minipurge
results in the most limiting factor being approximately 7 times those listed in
Table I, Column I and therefore occupancy for an individual would be allowed for nearly 6 hours. In addition, the philosophy of Regulatory Guide 8.15 is to minimize the requirement for wearing respirators through improved ventilation.
Therefore, not using the minipurge would be contrary to the philosophy of both 10 CFR 20 and Regulatory Guide 8.15. In order to pass the required flow of
4,000 cfm and use only one set of valves in accordance with the recommendations of BTP CSB-4, an 18-inch isolation valve was utilized in lieu of the recommended 8-inch valve.
Good engineering practice limits the flow velocities and pressure drops through
system valves. With an 18-inch line (velocity = 2,264 fpm), the design flow can be maintained by the supply and exhaust fans designed for a differential pressure of 4.25 and 5.0 inches w.g., respectively. If the system lines
remained as designed and the isolation valves were replaced with 8-inch valves
with reducers on either side, the supply and exhaust system pressure drops
would increase to 9.02 and 10.5 inches w.g. at the design flow. Since the fans cannot create these high differential pressures, the design flow would not be realized and the system would not perform its design function.
9.4-63 Rev. 13 WOLF CREEK The charcoal adsorbers in the discharge of the system comply with Regulatory Guide 1.140, to the extent discussed in Table 9.4-3.
POWER GENERATION DESIGN BASIS THREE - The CRDM cooling system limits the normal ambient temperature within the CRDM shroud to approximately 160 F by inducing 120 F containment air for cooling. The cooling of this air is provided by the containment coolers.
POWER GENERATION DESIGN BASIS FOUR - The cavity-cooling system is designed to limit the normal ambient temperature to acceptable limits around the out-of-core neutron detectors. The cavity concrete temperature is limited to 150 F, except for the area directly below the seal ring support, by air cooling of the
reactor vessel supports and reactor coolant pipe whip restraints. The cooling of the cavity air is provided by the containment coolers.
POWER GENERATION DESIGN BASIS FIVE - The pressurizer cooling fan limits the
temperature in the area below the pressurizer skirt to approximately 120 F by
inducing air from the containment for cooling. POWER GENERATION DESIGN BASIS SIX - The elevator machine room exhaust fan is
designed to provide sufficient air changes in the machine room to maintain a suitable environment for the equipment located there.
POWER GENERATION DESIGN BASIS SEVEN - The containment atmospheric control system has two functions--to reduce the containment airborne concentrations of
radioiodine and particulates to acceptable levels during occupancy of the
containment and to reduce the amount of airborne radioiodine and particulates
released to the environment to meet the as-low-as-reasonably achievable dose objective of 10 CFR 50, Appendix I during containment purges. The system operates, as needed, prior to and during purging to provide internal clean-up
of the containment atmosphere by recirculation through charcoal adsorbers and
HEPA filters. The charcoal adsorbers are responsive to Regulatory Guide 1.140, to the extent discussed in Table 9.4-3.
9.4-64 Rev. 16 WOLF CREEK 9.4.6.2 System Description 9.4.6.2.1 General Description Piping and Instrumentation Diagrams for the containment shutdown purge system, the containment minipurge system, the containment atmospheric control system, and the CRDM and cavity-cooling system described below are shown in Figure 9.4-
- 6. The containment shutdown purge supply system supplies fresh outside air, tempered or cooled, as required, to the containment. During operation of the
containment shutdown purge supply system, the containment shutdown purge
exhaust fan takes suction from the containment through the containment purge exhaust system and containment purge filtration unit and discharges it through the unit vent.
Prior to entrance into the containment during reactor power operations, the
containment minipurge system and the containment atmospheric control system may be used to reduce the concentration of noble gases, particulates, and halogens in the containment atmosphere. The containment minipurge supply system
supplies air to the containment. The containment minipurge exhaust fan
exhausts from the containment through the containment purge exhaust system.
The exhaust air is processed through the containment purge filter-adsorber unit and discharged through the unit vent. When required, the containment atmospheric control system collects and processes airborne and particulate
fission products through charcoal adsorbers. The system operates completely
within the containment. The containment minipurge system and the containment
atmospheric control system maintains containment occupant exposure from airborne activity to less than that specified in 10 CFR 20. Based on the source terms provided in Section 11.1 and the dose evaluation provided in Section 11.3, the exhaust system meets the objective of 10 CFR 50, Appendix I.
The containment atmospheric control systems are located on the laydown area above the operating floor, on the same side of the containment. The systems take suction from this region and discharge it upwards to supplement the normal
containment air pattern.
The CRDM cooling system induces containment air into the CRDM shroud and exhausts it through the fans. Normally, two of the three fans are in operation to provide the required flow of cooling air.
9.4-65 Rev. 19 WOLF CREEK The cavity-cooling system induces air supplied to the incore instrument tunnel by the containment coolers into the cavity for cooling. The rate of airflow
for the cavity cooling fan is based on dissipating the heat from the vessel, nozzle support system, insulation losses, vessel piping, and gamma heating. The cavity-cooling system fans exhaust from the cavity to the containment atmosphere. Normally, one fan is in operation to provide the necessary
airflow. As described in Section 6.2.2.2 and shown in Figure 6.2.2-7, the containment coolers supply air to the lower portions of the steam generator compartments.
The air is exhausted from these compartments by means of the hydrogen mixing
fans, which have a high discharge velocity, directing the air-stream upward.
This action in conjunction with the operation of the CRDM cooling system and the cavity cooling systems, which take suction from the lower area of the containment and discharge it upwards, produces a normal containment air flow
circulation path from the bottom to the top of the containment.
9.4.6.2.2 Component Description Codes and standards applicable to the containment HVAC systems are listed in
Tables 3.2-1 and 9.4-12. The containment penetrations and containment
isolation valves are designed and constructed in accordance with the
requirements of quality group B and are seismic Category I. The cavity cooling system, CRDM cooling system, containment atmospheric control system, pressurizer skirt cooling, elevator machine room exhaust, and the containment
purge systems (excluding the containment isolation provisions) are designed and
constructed in accordance with codes and standards comparable with quality
group D. NONESSENTIAL AIR HANDLERS - The only nonessential air handlers in the containment HVAC systems are the containment shutdown purge supply air unit and
the containment minipurge supply air unit.
The containment shutdown purge supply air unit consists of particulate filters, heating coil, cooling coil, centrifugal fan, and electric motor driver.
The containment minipurge supply air unit consists of particulate filters, heating coil, and centrifugal fan with electric motor driver.
9.4-66 Rev. 0 WOLF CREEK NONESSENTIAL FANS - Those nonessential fans in the containment HVAC systems are the shutdown purge exhaust fan, minipurge exhaust fan, CRDM cooling fans, cavity cooling fans, pressurizer cooling fan, elevator machine room exhaust fan, and the containment atmospheric control system fans. The shutdown purge exhaust fan, minipurge exhaust fan, and the elevator machine
room exhaust fan are centrifugal fans with electric motor drivers.
The CRDM cooling fans, cavity cooling fans, pressurizer cooling fan, and containment atmospheric control system fans are vaneaxial fans with directly
coupled electric motor drivers.
NONESSENTIAL FILTER UNITS - There are two nonessential filter units in the containment HVAC systems--the containment purge exhaust filter/adsorber unit and the containment atmospheric control system filter adsorber units. Each
unit consists of moderate efficiency prefilters, HEPA filters, and charcoal adsorption beds.
CONTAINMENT ISOLATION VALVES - The containment purge system is the only containment HVAC system which penetrates the containment. The supply and
exhaust system both contain four isolation valves. These valves are air-
operated butterfly valves.
CONTAINMENT ISOLATION PENETRATION TEST VALVES - Each containment purge system penetration is provided with a manually operated gate valve for test connection which is normally locked closed and capped. DEBRIS SCREENS - As shown on Figure 9.4-6, Sheet 4, debris screens are provided on the containment side of the minipurge supply and exhaust isolation valves to prevent the entry of lightweight debris which could preclude tight valve
closure. The piping which contains the screens is ANSI B31.1 (150 pound design pressure) piping which is seismically analyzed in accordance with Position C.3
of Regulatory Guide 1.29. The screens are located approximately two pipe diameters away from the isolation valves and are inherently designed to withstand post-LOCA differential pressures due to their rugged design and the
negligible pressure drop through the screen material (No. 2 mesh, .063 inch
wire with a 76.4 percent free area). The screen material is welded over the
17-inch-diameter opening in a 1/4-inch-thick flange which is bolted into place. The purge isolation valves and debris screens are located adjacent to the
containment wall, outside of the secondary shield walls, and are protected from
missiles which could be postulated following a LOCA. Also, motor-operated dampers are located one pipe diameter away from the screens on the containment side. These dampers and the connecting piping provide additional protection for the wire mesh screens.
9.4-67 Rev. 15 WOLF CREEK CONTAINMENT COOLERS - See Section 6.2.2.2.
HYDROGEN MIXING FANS - See Sections 6.2.2.2 and 6.2.5. ISOLATION DAMPERS - Where a means of system isolation is required, parallel-
blade-type dampers are utilized. The type of operator employed is dependent
upon the specific design and/or usage requirements.
FLOW CONTROL DAMPERS - Opposed-blade-type dampers are utilized, as necessary, to provide a means of system balancing. In general, these are manually
operated. However, some utilize power operators to allow compensation for
changes in system resistance occurring during system operation.
BACKDRAFT DAMPERS - Backdraft dampers are employed, where required, to prevent system backflow.
TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects
of extreme wind or tornado conditions is required. These dampers close with the flow produced by the differential pressure associated with the tornado or high winds.
FIRE DAMPERS - Fire dampers are located in fire barriers, as necessary, to
maintain the fire ratings of the barriers. Dampers are the 3-hour-rated curtain type. 9.4.6.2.3 System Operation
The containment shutdown purge supply system supplies fresh outside air, tempered or conditioned as required, into the containment. During operation of the containment shutdown purge supply system, the containment shutdown purge
exhaust fan operates to take suction from the containment through the
containment purge exhaust system. The containment exhaust is monitored and
processed through the containment purge filter/adsorber unit prior to being released through the unit vent.
The containment minipurge system and the containment shutdown purge system
intakes are in a penthouse atop the auxiliary building, which is located
approximately 15 feet below and 135 feet horizontally from the diesel exhaust discharge point. This separation is sufficient to provide significant dilution of the diesel exhaust gases; therefore, operation of the diesel during normal
plant operations results in no significant ingestion of exhaust gases into the
containment.
9.4-68 Rev. 27 WOLF CREEK The containment shutdown purge supply system ductwork runs along the inside periphery of the containment and discharges to the operating floor. The
containment coolers aid in the distribution of the air throughout the remainder of the containment. The containment shutdown purge supply system and the containment shutdown purge
exhaust fan may be operated continuously during shutdown to provide the
containment ventilation. This system also serves as the means of heating the containment during plant shutdown.
Prior to entrance into the containment during reactor power operations or in modes 3, 4, 5, 6 or Defueled, the containment minipurge system operates in conjunction with the containment atmospheric control system. Operation of the containment minipurge system may be continuous or intermittent. The minipurge system is designed to reduce the containment noble gas concentration. The containment atmospheric control system collects and processes airborne and
particulate fission products through charcoal adsorbers and operates only as
required. The CRDM cooling system induces containment air into the CRDM shroud for
cooling. Normally, two of the three fans are in operation to provide the required flow of cooling air. The ultimate cooling is provided by the containment coolers. The temperature of the cooling air in the missile shield plenum is monitored via the plant computer. Each CRDM fan is provided with a manual backdraft damper to prevent bypass flow through the idle fan.
The cavity cooling system induces air into the cavity for cooling. This air is
induced from the instrument tunnel (where it is supplied by the containment coolers), through the hot leg and cold leg restraints and around the neutron shield. The rate of airflow is based on dissipating the heat from the nozzle support system, insulation losses from the reactor vessel, reactor coolant
piping, and the hot and cold leg restraints, and gamma heating. The cavity
cooling system fans exhaust from the cavity to the containment atmosphere. Air is exhausted from the upper regions of the cavity through the reactor vessel supports and through the neutron detector wells. One operating fan has the
capability to provide the necessary airflow. The ultimate cooling is provided
by the containment coolers. The effluent air temperature from one reactor
vessel support, from one detector well, and in one upper cavity region exhaust leg is monitored by the plant computer. In addition, temperature elements are embedded in the cavity, below each reactor vessel support, to monitor concrete
temperature.
9.4-69 Rev. 19 WOLF CREEK The pressurizer cooling fan is located near the bottom of the pressurizer compartment. The fan takes suction from the lower region of the pressurizer
compartment (and therefore the coolest) and through the ductwork and discharges it in the area immediately below the pressurizer skirt. The fan will operate only when the associated containment cooler is out of service.
The machine room exhaust fan is located on the roof of the machinery equipment
room and takes suction from the room. Makeup air is induced from the containment through transfer grilles located in the walls of the room. The machine room exhaust fan operates during normal plant operations and during
shutdown. It should not be operated during ILRT, to prevent overloading of the
fan motor.
Cooling water for the shutdown purge supply unit is supplied by the central chilled water system (Section 9.4.10). Hot water for both the containment
shutdown purge supply unit and the containment minipurge supply unit is supplied by the plant heating system (Section 9.4.9).
Discussed below are the power generation operations and shutdown operations of the containment HVAC systems. Because the emergency operation consists only of
closing the containment isolation valves, it is discussed under the power
generation and shutdown operations.
POWER GENERATION OPERATION - The minipurge system is designed to minimize occupational exposures to as-low-as-reasonably-achievable (ALARA) levels.
Instead of personnel entering the containment with airborne activities much
greater than permissible DAC limits and at odds with the philosophy of Regulatory Guide 8.15, the containment will be purged to reduce airborne radioactivity concentrations and exposures in line with the philosophy of 10 CFR 20 and Regulatory Guide 8.15. The minipurge system is designed to be operated continuously to achieve these objectives. The need for continuous
operation includes consideration for planned and unplanned entries into the containment and the need to periodically vent excess air from the containment to maintain the pressure near atmospheric conditions.
- a. Preplanned Entries
During the first years of commercial operation, daily entry into the containment is assumed for planning purposes. This frequency is used by other PWRs. These
entries are assumed to be from 1/2 to 1-1/2 hours in
length, depending on the conditions found within the
containment. This type of operation would allow correction of leaks, (much smaller than the Technical
9.4-70 Rev. 13 WOLF CREEK Specification limits). Early correction, prior to the formation of large mounds of boric acid crystals and the
release of significant amounts of radioactivity, enhances the overall ALARA program at the plant. The frequency of planned entries is expected to be significantly less than assumed above.
- b. Unplanned Entries Unplanned entries include those responding to abnormal
indications from within the containment. These
indications include leaks, equipment malfunctions, and
instrumentation failures. Since these failures could have a significant impact on the continued safe operation of the plant, immediate response is most preferable.
Without the continuous operation of the minipurge, the doses received from containment entries are much higher, unless entries are delayed for significant amounts of time. For instance, if the containment had not been purged for 2 weeks, it would take 65 hours to bring the
airborne activity down to the same levels as those
maintained with its operation.
- c. Containment Pressure Reduction
Instrument air is continuously being vented to the
containment from air-operated valves. These valves also
dump the air from their accumulators upon actuation. In order to maintain the containment pressure near atmospheric conditions, the minipurge system is used to release excess air.
One operating plant has experienced over a 1 psig pressure buildup in 24 hours. If this rate were experienced at WCGS, the containment would have to be
vented at least every other day to maintain the
containment pressure within the Technical Specification
upper limit of +1 1/2 psig. The containment minipurge system is manually initiated from the control room.
Exhaust from the containment is processed through the containment purge exhaust
system charcoal adsorption train prior to being discharged through the unit
vent. The containment purge exhaust system is monitored for radioactivity, both upstream and downstream of the charcoal adsorber. The containment purge exhaust system is provided with redundant particulate and gaseous radiation
monitors in a seismic Category I section of ductwork directly downstream of the (exhaust) containment isolation valves. Downstream monitoring of the
containment purge exhaust system is provided by the radiation monitor in the
unit vent.
9.4-71 Rev. 0 WOLF CREEK A temperature controller located downstream of the containment minipurge system maintains an offcoil temperature of 50 F during the winter months' operation.
The containment minipurge supply unit has no cooling coil and, therefore, when the outside ambient temperature rises above 50 F untempered outside air is supplied to the containment. A temperature switch, located immediately downstream of the shutdown purge supply unit cooling coil, stops the supply fan
if the supply air temperature falls below 40 F, to prevent freezing of the
coils. Normally, each of the four containment coolers are operating to provide
containment cooling capabilities. Although only three coolers are required to
provide the proper cooling, four coolers operate to provide the required air
flow distribution. 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. 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 is also monitored in the area of each containment cooler intake. Control room indication is provided for both the
leaving air and inlet air temperatures. The hydrogen mixing fans are located in the hatches above the reactor coolant pump motors. Air is drawn from the steam generator compartments by the fans
and discharged toward the upper regions of containment. The discharge of the
fan has provisions for a minimum throw (distance travel by the air stream) of 100 feet to minimize stratification in the upper regions of containment.
The CRDM cooling system operates normally with two of the three fans. The
system is manually initiated from the control room.
The CRDM cooling fan operation is as follows:
- 1. When RCS temperature is >165ºF and no fans are running, the missile shield plenum temperature is monitored to ensure that the cooling air remains lower than 165ºF.
- 2. When RCS temperature is >200ºF, then at least one fan must be operating.
- 3. When RCS temperature is >350ºF, then
- a. two fans must be operating, or b. one fan must be operating, and the missile shield plenum temperature is monitored to ensure that the cooling air remains less than 165ºF.
The CRDM cooling system removes residual heat from the CRDM following a trip of
the rods. This is a nonessential operation, but, if available, will protect the CRDM.
The cavity cooling system fan induces containment air from the instrument
tunnel, through the reactor coolant piping penetrations, and into the cavity
for cooling. Portions of this air are
9.4-72 Rev. 22 WOLF CREEK exhausted directly through the reactor vessel supports and the out-of-core neutron detector wells for cooling, and the remaining air is exhausted from the
upper portions of the cavity. The cavity cooling system maintains the concrete at a temperature of no greater than 150 F, except for the area directly below the seal ring support which is limited to 220 F. Temperatures outside the reactor cavity are described in Section 6.2.2.
The pressurizer cooling fan induces air from the containment to provide cooling
of the pressurizer skirt and heater connections when the containment cooler serving the pressurizer compartment is out of service. The system is manually initiated from the control room.
The elevator machine room exhaust fan provides the required ventilation of the machinery equipment room. The fan is manually operated from the control room. The containment atmospheric control system, when required, is started manually
from the control room. The system operates in a recirculation mode whenever
cleanup of the containment atmosphere is required.
During containment atmospheric control system operations, as filter loading increases, a constant flow is maintained by utilizing a modulating damper
located downstream of the filter train. As the filter loads, a pressure
differential indicator controller monitors the change in pressure drop across
the charcoal bed (the charcoal bed will not load up and thus has a constant pressure drop for a given flow) and modulates the damper accordingly to maintain the system resistance.
Each containment atmospheric control system filtration unit is provided with thermistors in the airstream between the charcoal beds to alarm via the plant computer at 200 F and 300 F. The filtration unit is open at both ends so that any pressure buildup which may result from the excessive bed heating is vented to the containment atmosphere.
SHUTDOWN OPERATIONS - Once cold shutdown (mode 5) is achieved, the containment shutdown purge system is the desired system to be operated. The elevator machine room exhaust fan may also be operated. The minipurge system may be operated in this mode and also in mode 6 or Defueled, but not in parallel with the shutdown purge system. The containment shutdown purge system and the minipurge system are manually initiated from the control room. The containment purge exhaust system is
9.4-73 Rev. 19 WOLF CREEK monitored for radioactivity, both upstream and downstream of the charcoal adsorber. The containment purge exhaust system is provided with redundant
particulate and gaseous radiation monitors in a seismic Category I section of duct-work directly downstream of the (exhaust) containment isolation valves. Downstream monitoring of the containment purge exhaust system is provided by the radiation monitor located in the unit vent.
A differential pressure indicator controller, located across the charcoal adsorber, modulates a damper downstream of the filter train to maintain constant system resistance as the particulate filters load up. This control
arrangement assures a constant system flow.
The containment purge charcoal adsorber train is monitored for charcoal bed temperature. Should the bed temperature exceed 200 F, an alarm is received in the control room to alert the operators of excessive bed heating. Should the bed temperature continue to rise and exceed 300 F, a second alarm is received in the control room. To prevent backflow through the system, upstream isolation is provided by a backdraft damper located at the inlet to the filter
train.
Temperature controllers, located downstream of the containment shutdown purge supply unit, regulate the flow of chilled water or hot water to the respective coils to ensure that the containment is maintained between the design temperatures of 50 F and 90 F during shutdown. A temperature switch, located immediately downstream of the supply unit cooling coil, stops the supply fans
if the supply air temperature falls below 40 F to prevent freezing of the coils. The elevator machine room exhaust fan provides the required ventilation of the machinery equipment room. The fan is manually operated from the control room.
The containment coolers and the hydrogen mixing fans may be operated during
refueling operations to provide supplemental air distribution within the containment. Both the hydrogen mixing fans and the containment cooler fans are operated at low speed to reduce noise levels within the containment. The
coolers may be operated with service water to provide supplemental cooling or
without service water for supplemental heating.
The containment coolers may be operated during containment integrated leak rate testing (ILRT) to maintain uniform containment temperature. If in-service, the coolers are operated with service water to provide
9.4-74 Rev. 25 WOLF CREEK cooling and without service water to provide heating during the test procedure. If in-service the fans are operated at low speeds during this elevated pressure condition. EMERGENCY OPERATIONS - Both the containment shutdown purge and the containment
minipurge isolation valves are automatically closed upon receipt of a containment purge isolation signal (CPIS). The CPIS is initiated by receipt of an SIS, indication of high radioactivity levels in the containment atmosphere by one of the containment atmospheric radiation monitors, or by indication of high radioactivity levels in the purge exhaust system process effluents by one
of the purge exhaust radiation monitors. Sections 7.2.2 and 7.3.2 discuss these
various signals which generate a CPIS.
The containment purge isolation valves are designed for rapid closure to minimize release of containment effluents following postulated accident
conditions. The containment minipurge isolation valves are designed for tight
closure within 3 seconds after receipt of an isolation signal. Wire screens
are provided on the inboard (containment) side of these valves to preclude the entrance of debris which could prevent tight closure of the minipurge valves. The shutdown purge containment isolation valves are tested by the manufacturer
under static conditions for valve closure within 10 seconds. Both these valves
are designed to fail closed by spring action upon a loss of power or instrument
air. Spectacle flanges are provided inboard the inboard valves and outboard of the outboard valves to facilitate integrated leak rate testing of each individual valve. Also, a test valve is provided outside containment at each
penetration to facilitate integrated leak rate testing.
Sections 6.2.2 and 6.2.5 provide the description of the containment coolers and the hydrogen mixing fans following a postulated DBA.
9.4.6.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.4.6.1. SAFETY EVALUATION ONE - Sections 6.2.4 and 6.2.6 provide the safety evaluation
for the system containment isolation arrangement and testability.
SAFETY EVALUATION TWO - A testing program, implemented by the manufacturer, verifies a minipurge containment isolation valve closure time of 3 seconds or
less. The containment minipurge containment isolation valves comply with BTP
CSB 6-4 to the extent discussed in Table 9.4-13. The shutdown purge
containment isolation valve is tested by the manufacturer to verify a closure time of 10 seconds or less under static conditions.
9.4-75 Rev. 25 WOLF CREEK 9.4.6.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. Filters and adsorbers in the containment purge exhaust system and the containment atmospheric control system are tested in the shop and after initial
installation. The containment purge exhaust system filters and adsorbers are
also tested subsequent to each filter or adsorber change. Interim tests and inspection are performed annually, after installation, in accordance with the requirements of Regulatory Guide 1.140 except as noted below.
Prefilters do not undergo factory or inplace testing since no credit is taken for removal of particulates in meeting permissible dose rates. However, unloaded prefilters will exhibit a 55-percent efficiency (min.) for the removal
of coarse particulates, when tested in accordance with ASHRAE-52.
HEPA filters are factory tested with monodispersed DOP aerosol to demonstrate a minimum particulate removal efficiency of no less than 99.97 percent for 0.3 micron particulates. Inplace leak testing is carried out with cold
polydisperse DOP. Testing is in accordance with the procedures set forth in
ANSI N510. The containment atmospheric control system is exempt from this subject inplace testing. Charcoal adsorbers are qualified per Regulatory Guide 1.140 and are factory
tested in accordance with RDT-M-16-IT to exhibit a decontamination efficiency of no less than 99.5 percent for elemental iodine and 95 percent for methyl iodide. Inplace testing is performed with a suitable refrigerant in accordance with the procedures set forth in ANSI N510. The containment atmospheric control system is exempt from this subject inplace testing.
The containment shutdown purge system and the containment minipurge system, excluding the containment isolation valves, the containment atmospheric control system, CRDM cooling system, cavity cooling system, pressurizer cooling system, and elevator machine room exhaust fan undergo acceptance testing prior to plant
startup.
The containment purge valves undergo preoperational testing prior to plant startup. Each valve is leak rate tested in accordance with 10 CFR 50, Appendix J. Fans are rated in accordance with AMCA standards.
Major components located outside the containment are accessible during normal plant operation for inspection, maintenance, and periodic testing. Components
located inside the containment are accessible during plant shutdown.
9.4-76 Rev. 12 WOLF CREEK 9.4.6.5 Instrumentation Applications Indication of the operational status of the containment purge exhaust fans and all the fans in the containment is provided in the control room. All fans and air handlers are operable from the control room.
An indication of the position of all isolation dampers is provided in the control room.
Temperature controllers located in the containment purge ductwork control the
containment temperature during shutdown.
The amount of filter loading for all filters associated with both the air handlers and the filter adsorbers is available at the unit.
All instrumentation provided with the containment atmospheric control system
and the containment purge system filter adsorber units is as required by Regulatory Guide 1.140.
Indication of the levels of gaseous, particulate, and iodine radioactivity
being exhausted from the containment and being released through the unit vent
is available in the control room. The temperature of the air leaving one of the detector wells, leaving one
reactor vessel support, and leaving the upper cavity area, as well as the
concrete temperature below each reactor vessel support, is available in the
control room. The containment pressure relative to the auxiliary building, the containment temperature, and the containment relative humidity are available in the control
room. 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 is also monitored in the area of each containment cooler intake. Control room indication is provided for the inlet
air temperatures. Each containment cooler fan is operable from the control room.
Each hydrogen mixing fan is operable from the control room and is monitored for
fan vibration.
9.4-77 Rev. 22 WOLF CREEK
9.4.7 DIESEL
GENERATOR BUILDING VENTILATION
The function of the diesel generator building (DGB) ventilation system is to provide a combustion air makeup rate and an environment suitable for the operation of the diesel generators.
9.4.7.1 Design Bases 9.4.7.1.1 Safety Design Bases The DGB HVAC system, excluding unit heaters, is safety related and is required to function following a DBA and to achieve and maintain the plant in a post accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The DGB ventilation system is protected from the
effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles.
SAFETY DESIGN BASIS TWO - The DGB ventilation system is designed to remain functional after a SSE and to perform its intended function following a
postulated hazard, such internal missiles, or pipe break (GDC-4). SAFETY DESIGN BASIS THREE - The safety functions of the DGB ventilation system can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC-44). SAFETY DESIGN BASIS FOUR - The DGB ventilation system is designed so that the
active components are capable of being tested during plant operation. SAFETY DESIGN BASIS FIVE - The DGB ventilation system uses the design and
fabrication 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 must be in accordance with Regulatory Guide 1.32.
SAFETY DESIGN BASIS SIX - The DGB ventilation system maintains a suitable
atmosphere in the DGB while the diesel is operating. Cooling is accomplished by the outside air. 9.4.7.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The DGB heating system limits the minimum
room temperature to 60 F during periods when the diesel is not operating (See Note 8 of Table 3.11(B)-1 for clarification).
9.4-78 Rev. 19 WOLF CREEK 9.4.7.2 System Description 9.4.7.2.1 General Description Figure 9.4-7 is the piping and instrumentation diagram for the DGB ventilation
and heating system.
The DGB ventilation system provides cooling for the diesel generators, using outside air as the cooling medium. Air is supplied into the building, pressurizing the building slightly, and is vented from the building through
exhaust air louver openings. Each diesel generator room is provided with a
separate system. Electric unit heaters are provided in each room for heating.
The ventilation system serves as a source of makeup air which is used for combustion air by the diesel.
9.4.7.2.2 Component Description
Codes and standards applicable to the DGB ventilation system are listed in Tables 3.2-1 and 9.4-14. The DGB ventilation system is designed and
constructed in accordance with codes and standards comparable with quality
group C. SAFETY-RELATED FANS - A DGB supply fan is located in each diesel generator room. These fans are vaneaxial fans with electric motor drivers.
UNIT HEATERS - Heating of the diesel generator building is provided by electric
unit heaters. Each unit heater consists of a coil and a fan with an electric motor driver.
ISOLATION DAMPERS - Where a means of system isolation is required, parallel-
blade-type dampers are utilized. The type of operator employed is dependent
upon the specific design and/or usage requirements. FLOW CONTROL DAMPERS - Opposed-blade-type dampers are utilized, as necessary, to provide a means of system balancing. In general, these are manually
operated. However, some utilize power operators to allow compensation for
changes occurring during system operation. TORNADO DAMPERS - Tornado dampers are employed where isolation from the effects
of extreme wind or tornado conditions is required. These dampers close with
the flow produced by the differential pressure associated with the tornado or
high winds.
9.4-79 Rev. 27 WOLF CREEK 9.4.7.2.3 System Operation
The DGB ventilation system is automatically activated when the room temperature exceeds 90°F and automatically shuts down when room temperature falls below 85°F. The ventilation system can be manually activated, if necessary, to provide cooling during occupation of the building. When the ventilation system is in operation, the supply fans take suction from the outside air and supply
air directly to their respective diesel generator room for maximum cooling requirements. However, each system is provided with a recirculation mode, whereby a portion of the room air may be mixed with the outside air. This recirculation mode is primarily for winter operation to prevent freezing. The
recirculated room air is utilized for tempering the outside air. Outside air
intake and exhaust louvers are selected on the basis of adverse environmental conditions. Louver blades are fixed and, hence, cannot become inoperable due to freezing or icing. They are designed to reduce cascading and reentrainment
of water into the airstream. Design of the louvers is for air inlet velocities
below 500 fpm to prevent moisture carryover. Electrical unit heaters are
provided in each room to limit the minimum room temperature to 60 F when the diesels are not operating. These unit heaters operate automatically and independently from the ventilation system.
The fire protection system provided for the diesel generators is a preaction
sprinkler system. Carbon dioxide is not utilized as the extinguishing medium. Hence, there is no possibility of CO2 being drawn into the combustion air. The exhaust stack is located approximately 65 feet horizontally from the air intake
and discharges approximately 35 feet above the air intake. The distances
between the diesel intake and exhaust, the exit velocity of the gases from the
exhaust stack, and the buoyancy of the hot exhaust gases are sufficient to reduce the possibility of exhaust gases being drawn into the combustion air stream to insignificant levels.
The probability of inducing exhaust gases into the intake air stream, due to
the loss of the stacks, is slight since the distance between the intake and exhaust (65 feet) is sufficient to prevent a short-circuiting of the exhaust gases.
Discussed below are the emergency operations of the DGB ventilation system.
Except for operation of the unit heaters, the power generation operations and shutdown operations are identical to the emergency operations.
9.4-80 Rev. 9 WOLF CREEK EMERGENCY OPERATIONS - The DGB ventilation system is automatically activated when the room temperature exceeds 90°F and automatically shuts down when room
temperature falls below 85°F. The ventilation system can be manually activated, if necessary, to provide cooling during occupation of the building.
For maximum cooling, the supply fans take suction from the outside and supply
directly to their respective rooms. All cooling is accomplished by the outside
air. Each fan is provided with a mixing box arrangement. When maximum cooling is not required, a portion of the room air is mixed with the outside air. The proportion of outside air and room air is controlled by a room thermostat to
maintain the ambient temperature within its specified range, when outside
temperatures permit. This mixing mode is primarily for winter operation to
prevent freezing and to minimize cycling of the fans. The room air is utilized for tempering the outside air.
The supply air system, when in operation, serves as a source for combustion air
to be used by the diesel generators. The exhaust air flowpath also supplies
combustion air for the diesels regardless of the mode of operation of the supply air system. The exhaust flowpath of the ventilation system is provided with a damper which is designed to fail in the open position. This exhaust
damper opens automatically on a diesel start to ensure that the maximum
quantity (24,000 cfm) of combustion air required by the diesel is provided.
The exhaust damper is normally closed when the supply air system is not operating to prevent cold outside air from entering the building. With the diesel operating at full load, the ventilation system providing the required
outside air for combustion (17,900 cfm at -30F), and excluding the unit heaters
as a source of heat, the room temperature will remain above freezing.
The diesel room is pressurized slightly by the air supply system and relieved through the exhaust louver. The exhaust damper, located upstream of the
exhaust louver and tornado damper, provides building isolation against outside
air infiltration during system shutdown.
The diesel generators building supply air system intake and the exhaust system ductwork contain dampers capable of withstanding the effects of extreme wind or
tornado conditions (3 psi at a rate of 2 psi/sec per Regulatory Guide 1.76). These dampers close with the flow produced by the differential pressure
associated with high winds or tornadoes. The damper located in the exhaust system ductwork is spring loaded to prevent closure during normal system operation.
9.4-81 Rev. 27 WOLF CREEK Missile barriers are provided externally to the isolation system to prevent damage by a tornado missile.
Both the supply air intake and the recirculation ductwork are provided with modulating dampers operated by electrohydraulic actuators. These dampers modulate, as required, to provide the required mixing of the supply and
recirculation air to maintain the room temperature within the specified limits.
Modulation of the dampers is controlled by a Class IE control circuit which senses the room temperature and operates the dampers accordingly. This control circuit serves to start the fan should the room temperature rise above 90°F, as
well as initiate opening of the exhaust damper. This control circuit also
serves to alarm the control room, via the plant computer, of low room
temperature and high room temperature. The ventilation system with its recirculation mode of operation, and cut-off
below approximately 85°F whenever the diesels are operating, can maintain room ambient temperatures between 60 and 122 F when the outside ambient temperatures are between 7 and 97 F. When the diesels are in standby, heating is provided by strategically located unit heaters.
Electrical unit heaters, each individually controlled by its associated room thermostat, are provided in each room to limit the minimum room temperature to 60 F during periods when the diesel is not operating. These unit heaters
operate automatically and independently from the ventilation system.
9.4.7.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 9.4.7.1.
SAFETY EVALUATION ONE - The safety-related portions of the DGB ventilation system are located in the diesel building, which 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 DGB ventilation 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 safe shutdown, as outlined in Section 7.4 can be achieved and maintained.
9.4-82 Rev. 19 WOLF CREEK SAFETY EVALUATION THREE - Complete redundancy is provided for the DGB ventilation system and, as indicated by Table 9.4-15, 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. SAFETY EVALUATION FOUR - The DGB ventilation system was initially tested with
the program given in Chapter 14.0. Periodic inservice functional testing is
done in accordance with Section 9.4.7.4. 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. All the power supplies and control
functions necessary for safe function of the DGB ventilation systems are Class IE, as described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - The DGB ventilation system has sufficient cooling capability to maintain the diesel room at 122 F or below with the diesel
operating at rated load and with ambient outside air temperature of 97 F. The supply fan is stopped should the room temperature drop below 85°F. 9.4.7.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0. One of the two redundant DGB fans is tested in accordance with standards of the Air Moving and Conditioning Association (AMCA) to assure fan characteristic
performance curves.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.7.5 Instrumentation Applications Indication of the DGB fan operational status is provided in the control room. All DGB fans are operable from the control room.
An indication of the position of all exhaust dampers is provided in the control room. Thermostats control the room temperatures.
The DGB room temperature is available in the control room.
9.4-83 Rev. 9 WOLF CREEK High and low DGB room temperature is alarmed in the control room.
Exhaust dampers are operable from the control room.
9.4.8 ESSENTIAL
SERVICE WATER PUMPHOUSE VENTILATION
The function of the essential service water (ESW) pumphouse ventilation system
is to provide an environment suitable for operation of the essential service water pump motors and associated electrical equipment.
9.4.8.1 Design Bases 9.4.8.1.1 Safety Design Bases The ESW pumphouse ventilation system excluding unit heaters, is safety-related, is required to function following a DBA, and is required to achieve and
maintain the plant in a post accident safe shutdown condition. SAFETY DESIGN BASIS ONE - The ESW pumphouse ventilation 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 ESW pumphouse ventilation system remains functional after an SSE and performs its intended function following a postulated hazard, such as internal missiles, or pipe break (GDC-4). SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a
single active component failure coincident with the loss of offsite power (GDC-44). SAFETY DESIGN BASIS FOUR - Active components of the ESW pumphouse ventilation
system are capable of being tested during plant operation.
SAFETY DESIGN BASIS FIVE - The ESW pumphouse ventilation system uses the design and fabrication 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 must be in accordance with
Regulatory Guide 1.32. SAFETY DESIGN BASIS SIX - The ESW pumphouse ventilation system is designed to
limit the building to a maximum ambient temperature of 122 F (50 C). Cooling is accomplished by the outside air.
9.4-84 Rev. 19 WOLF CREEK 9.4.8.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - The essential service water pumphouse heating system will limit the minimum room temperature to 50°F during periods when the pumps are not operating (See Note 8 of Table 3.11(B)-1 for
clarification). The temperature in the yard piping access vaults is required to be monitored during cold weather to assure freezing conditions do not occur. Access vaults AV2 and AV6 do not have temperature indication. Access vaults AV2 and AV6 will not reach a freezing temperature based on normal flow conditions of the return piping in the vaults and the maximum credited stagnation time.
9.4.8.2 System Description
9.4.8.2.1 General Description
The ESW pumphouse ventilation system is shown in Figure 9.4-8.
The ESW pumphouse ventilation system provides cooling for the essential service water pump motors, using outside air as the cooling medium. Air is supplied
into the building, pressurizing the building slightly, and is vented from the
building through exhaust air louver openings. Each ESW pumproom is provided
with a separate system. Electric unit heaters are provided in each room for heating. 9.4.8.2.2 Component Description
Codes and Standards applicable to the ESW pumphouse ventilation system are listed in Tables 3.2-1 and 9.4-16. The ventilation system, excluding unit heaters, is designed and constructed in accordance with codes and standards
comparable with quality group C. The unit heaters are designed and constructed
in accordance with codes and standards comparable with quality group D.
SAFETY-RELATED FANS - An ESW pumphouse supply fan is located in each pumproom. These fans are vaneaxial fans with electric motor operators.
UNIT HEATERS - Heating of the ESW pumphouse is provided by electric unit
heaters. TORNADO DAMPERS - Tornado dampers are provided in the ventilation intake and
exhaust paths. These dampers close with the flow produced by the differential
pressure associated with high winds or tornadoes.
ISOLATION DAMPERS - The ventilation exhaust paths employ air-operated, parallel-blade-type dampers for isolation.
9.4-85 Rev. 28 WOLF CREEK FLOW CONTROL DAMPERS - The ventilation supply system employs opposed-blade-type dampers as the means for controlling the mixture of recirculation and outside
air being supplied to the pumproom. These dampers are operated by means of electro/hydraulic actuators. 9.4.8.2.3 System Operation
Each ESW pumphouse ventilation subsystem is automatically actuated by a start of the associated essential service water pump. The supply fans take suction from the outside and supply air directly to the respective pumprooms.
However, when maximum cooling is not required, the system is provided with a
recirculation mode whereby a portion of the room air is mixed with the outside air. This recirculation mode is primarily for winter operation to prevent freezing and continuous cycling of the fans. The recirculated room air is
utilized for tempering the outside air.
Each room is provided with dampers in the supply and exhaust ductwork to isolate the outside openings during a tornado.
Electric unit heaters are provided in each room to limit the minimum room
temperature to 50 F. These unit heaters operate automatically and
independently of the ventilation system. Discussed below are the emergency operations only, since there are no power
generation operations associated with the ESW pumphouse ventilation system.
EMERGENCY OPERATIONS - The ESW pumphouse ventilation system is automatically activated upon starting the associated ESW pump and automatically shut down when the pump shuts down. During periods when the pumps or fans are shut down, the ventilation system can be manually activated to provide cooling, if
necessary, during occupation of the building. The system may be started
manually, either by the local handswitch located in the room, or by the remote handswitch located in the control room.
For maximum cooling, the supply fans take suction from the outside and supply
directly to their respective rooms. All cooling is accomplished by outside
air. Each fan is provided with a mixing box arrangement. When maximum cooling is not required, a portion of the room air is mixed with the outside air. The proportion of outside air to room air is controlled by a room thermostat to
maintain ambient temperature within a specified range. This
9.4-86 Rev. 0 WOLF CREEK mixing mode is primarily for winter operation to prevent freezing and to minimize cycling of the fans. The room air is used for tempering the outside
air. The exhaust flow path is provided with a damper which is designed to fail in the open position. The exhaust damper is normally closed, when the pumps are
not operating, to prevent cold outside air from entering the building. This
exhaust damper opens automatically upon initiation of pump or fan operation to assure an exhaust air flow path, regardless of the mode of operation of the supply air system.
The ESW pumphouse is pressurized slightly by the air supply system and relieved
through the exhaust system. The exhaust damper, located upstream of the exhaust louver and tornado damper, provides building isolation against outside air infiltration during system shutdown.
The supply air system intake and the exhaust system ductwork contain dampers
capable of withstanding the effects of extreme wind or tornado conditions (a differential pressure of 3 psi and a differential pressure rate of 2 psi per second per Regulatory Guide 1.76). These dampers close with the flow produced
by the differential pressure associated with the high winds or tornado. The
damper located in the exhaust system ductwork is spring loaded to prevent
closure during normal system operation. Missile barriers are provided externally to the isolation system to prevent
damage by a tornado missile.
Both the supply air intake and the recirculation ductwork are provided with modulating dampers operated by electro-hydraulic actuators. The dampers modulate, as required, to provide the required mixing of the supply and recirculation air to maintain the room temperature within the specified limits.
Modulation of the dampers is controlled by a Class IE control circuit which
senses the room temperature and operates the dampers accordingly. This control circuit also serves to alarm the control room, via the plant computer, on low room temperature and high room temperature and to shut down the fan should the room temperature fall below 65 F. The ventilation system with its recirculation mode of operation, whenever the pumps are operating, can maintain room ambient temperatures between 50 and 122 F when the outside ambient temperatures are between the minimum and maximum site design temperatures (see Table 9.4-16).
9.4-87 Rev. 12 WOLF CREEK When the pumps, and hence ventilation system, are in standby, heating is provided by strategically located unit heaters. Electric unit heaters, each
individually controlled by its associated room thermostat, are provided in each room to limit the minimum room temperature to 50 F. These unit heaters are nonseismic Category I, and operate automatically and independently from the
ventilation system.
The essential service water yard piping access vaults are required to be monitored during cold weather to assure that freezing temperatures do not occur in the vaults. A temperature indicator (EF system designation) is provided for
each vault, to allow monitoring the vault temperatures. Access vaults AV2 and AV6 do not have temperature indication. Access vaults AV2 and AV6 will not reach a freezing temperature based on normal flow conditions of the return piping in the vaults and the maximum credited stagnation time.
9.4.8.3 Safety Evaluation
Safety evaluations are numbered to correspond to the safety design bases in Section 9.4.8.1.
SAFETY EVALUATION ONE - The safety-related portions of the ESW pumphouse
ventilation system are located in the ESW pumphouse which 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
this building.
SAFETY EVALUATION TWO - The safety-related portions of the ESW pumphouse ventilation systems are designed to remain functional after 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 safe
shutdown, as outlined in Section 7.4 can be achieved and maintained.
SAFETY EVALUATION THREE - The system description shows that complete redundancy of the ESW pumphouse ventilation system is provided and, as indicated by Table
9.4-17 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 ESW pumphouse ventilation system is initially
tested with the program given in Chapter 14.0. Periodic inservice functional
testing is done in accordance with Section 9.4.8.4.
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. All the power supplies and control
functions necessary for safe function of the ESW pumphouse ventilation system
are Class IE, as described in Chapters 7.0 and 8.0.
9.4-88 Rev. 28 WOLF CREEK SAFETY EVALUATION SIX - The ESW pumphouse ventilation system has sufficient cooling capacity to maintain the rooms at 122 F or below when the ESW pump
motors are operating at rated load and the outside air is at the maximum site design ambient. 9.4.8.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. One of each group of ESW pumphouse ventilation fans is tested in accordance with standards of the Air Moving and Conditioning Association (AMCA) to ensure
fan characteristic performance curves.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.8.5 Instrumentation Applications Indication of ESW pumphouse status is provided in the control room.
The ESW pumphouse fans are operable from the control room.
Thermostats control the room temperatures. Each room's temperature is available in the control room.
High and low room temperature is alarmed in the control room.
9.4.9 PLANT
HEATING SYSTEM
The plant heating system (PHS) serves as the heating medium for air to provide
a suitable environment for personnel and equipment. 9.4.9.1 Design Bases 9.4.9.1.1 Safety Design Basis
This system serves no safety function. Failure of the system does not affect safe shutdown of the plant.
9.4.9.1.2 Power Generation Design Basis
POWER GENERATION DESIGN BASIS ONE - The PHS provides hot water for the heating coils and the unit heaters.
9.4.9.2 System Description
9.4-89 Rev. 0 WOLF CREEK 9.4.9.2.1 General Description
The PHS is shown in Figure 9.4-9. The PHS is composed of redundant hot-water pumps, a steam-to-water heat exchanger, and a supply and return piping system. Each of the hot-water pumps
is rated at 100 percent of the total system flow to ensure system operation in
the event of the failure of one of the pumps. An expansion tank is provided on the suction side of the hot-water pumps to accommodate the volume of water expansion and maintain suction pressure for the pumps.
The steam-to-water heat exchanger is located on the discharge side of the hot-
water pumps and heats the water flowing through it. During normal plant operation, the heat exchanger utilizes steam from the reboiler as the heating medium, and during plant shutdown it utilizes steam produced by the auxiliary
boiler. In-line, secondary loop, hot-water pumps are provided with the heating coils for all outside supply air units.
9.4.9.2.2 Component Description
Codes and standards applicable to the PHS is listed in Table 9.4-18. The plant heating system is designed and constructed in accordance with codes and standards comparable with quality group D.
The PHS consists of a steam-to-water heat exchanger, two 100-percent-capacity
pumps, electric motor drivers, expansion tank, and associated piping, valves, instruments, and controls. HEAT EXCHANGER - The plant heating heat exchanger is the steam-to-water type
and consists of a shell and tubes. The steam is supplied to the shell, and the
water to be heated flows through the tubes. PUMPS - The main hot-water pumps are the centrifugal horizontal split case type
with electric motor drivers.
Design data for the plant heating system components are given in Table 9.4-18. 9.4.9.2.3 System Operation
The heating system, which utilizes hot water as the heating medium, provides
the source of heat for the ventilation system heating coils and unit heaters. The hot water is pumped by one of
9.4-90 Rev. 0 WOLF CREEK the two hot-water pumps through the supply main to the heat exchanger, where its temperature is raised to 198 F. The 198 F water then flows to the various
heating coils where the air is heated. It then leaves the coils at approximately 158 F and flows through the return main to the hot-water pumps.
Operation of the hot-water pumps is initiated either manually by operator
action or automatically upon indication of low outside air temperature.
Temperature sensors are located outside the auxiliary and control buildings near the air intakes. Either sensor will automatically initiate the hot-water pumps and energize the inlet steam temperature control valve. The pumps will
start with an outside air temperature of approximately 60 F, or less.
Overpressure protection is provided for both the shell and tube sides of the plant heating heat exchanger.
9.4.9.3 Safety Evaluation
The operation of the PHS is not required for the safe shutdown of the plant or for mitigating the consequences of a design basis accident.
9.4.9.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0. The hot-water system is hydrostatically tested in accordance with ANSI B31.1.
Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing. 9.4.9.5 Instrumentation Applications
An alarm is provided in the control room to indicate high and low water levels
in the expansion tank. Local pressure indication is provided upstream and down stream of each hot-
water pump.
9.4.10 CENTRAL CHILLED WATER SYSTEM The central chilled water system (CeCWS) serves as the cooling medium for air
to provide a suitable environment for personnel and equipment and to reduce the
outside air requirements.
9.4-91 Rev. 0 WOLF CREEK 9.4.10.1 Design Bases 9.4.10.1.1 Safety Design Basis The CeCWS serves no safety function. Failure of the system does not affect the safe shutdown of the plant.
9.4.10.1.2 Power Generation Design Basis POWER GENERATION DESIGN BASIS ONE - The CeCWS provides chilled water for the
cooling coils.
9.4.10.2 System Description 9.4.10.2.1 General Description
The CeCWS is shown on Figure 9.4-10. The CeCWS provides the cooling medium, when required, for the ventilation system cooling coils. The CeCWS is composed of redundant chilled-water pumps
and chillers and a supply and return piping system. Each of the chilled-water
pumps is rated at 100 percent of the total system load, as are the chillers, to
ensure system operation in the event of the failure of one of the components. A nitrogen blanketed expansion tank is provided on the suction side of the chilled-water pumps to accommodate the volume of water expansion and maintain
suction pressure for the pumps.
The chillers are located on the discharge side of the chilled-water pumps and cool the water flowing through them. The service water system serves as the heat sink for the chillers.
The CeCWS operates in a closed loop mode. To prevent rusting and deterioration
of the piping, chilled water is demineralized and corrosion inhibitors are introduced on the suction side of the pumps. Strainers are placed in the line at the inlet to the pumps to protect the equipment.
9.4.10.2.2 Component Description
Codes and standards applicable to the CeCWS are listed in Table 9.4-19. The central chilled water system is designed and constructed in accordance with
codes and standards comparable with quality group D.
The CeCWS consists of two 100-percent-capacity chillers, two 100-percent-capacity pumps, electric motor drivers, an expansion tank, and the associated piping, valves, instruments, and controls.
9.4-92 Rev. 0 WOLF CREEK CENTRAL CHILLERS - The central chillers are the centrifugal type, and each chiller consists of a compressor, an evaporator, and a water-cooled condenser.
PUMPS - The chilled-water pumps are the centrifugal type with electric motor drivers. Design data for components of the CeCWS are given in Table 9.4-19.
9.4.10.2.3 System Operation
The CeCWS provides the cooling medium for various ventilation cooling coils all year around, except for those cooling coils that are isolated & drained for plant winterization. Operation of the chilled-water pumps and chillers is manually initiated from a
local control panel. The chillers and the chilled-water pumps are arranged in
parallel. This permits manual alignment, so that either chilled-water pump may
service either chiller. During system operation, the chilled water is pumped by one of the two chilled-
water pumps through the supply main to the chiller, where its temperature is
lowered to approximately 44 F. The 44 F water then flows to the various
cooling zones, where it absorbs heat from the air passing over the coils. Water leaves the coils at approximately 62 F and flows through the return main to the chilled-water pumps. The heat absorbed at the coils is transferred to the
chiller, which, in turn, rejects this heat to the service water.
Fully automatic condenser water flow control is provided to regulate the water flow rate to maintain the condenser head pressure. This arrangement serves as a means of head pressure control to maintain chiller operation within the more
efficient range and to preclude tripping.
A means of adjusting chiller capacity in proportion to the variation in the design load is obtained through the use of temperature control. This is accomplished by means of variable inlet guide vanes at the suction to the
compressor. This control reduces the capacity of the chiller by varying the
angle at which the suction gas is directed into the eye of the impeller. A
chilled-water temperature sensor, located in the main header downstream of the pumps, automatically maintains the leaving chilled-water temperature at 44 F.
When the temperature changes, the temperature sensor signals the chilled-water
temperature controller to reposition the capacity-regulating vanes, which
change the capacity of the chiller to
9.4-93 Rev. 12 WOLF CREEK maintain the desired temperature. When the vanes reach the closed position and the leaving temperature of the chilled water continues to decrease to the
predetermined minimum, approximately 40 F, the low chilled-water temperature cut-out switch stops the compressor.
9.4.10.3 Safety Evaluation
The operation of the CeCWS is not required for the safe shutdown of the plant or for mitigating the consequences of a design basis accident.
9.4.10.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0. Major components are accessible during normal plant operation for inspection, maintenance, and periodic testing.
9.4.10.5 Instrumentation Applications
An alarm is provided in the control room to indicate high and low water levels
in the expansion tank.
Local differential pressure indication is provided across each strainer upstream of the pumps, and local pressure indication is provided downstream of each pump.
A temperature sensor located in the main header downstream of the pumps is
provided to control the leaving chilled-water temperature. A low chilled-water temperature cut-out switch stops the chiller.
9.4.11 ESSENTIAL SERVICE WATER VERTICAL LOOP CHASE VENTILATION Louvered ventilation openings located on opposite sides and elevations of the ESW vertical Loop Chase provide natural circulation of air through the chase to limit the internal temperature to a maximum temperature of 120°F to ensure health and safety of personnel working in the ESW Vertical Loop Chase for a limited time. This temp erature has no impact to the ESW or Ultimate Heat Sink (UHS) since this is less than the maximum temperature of the lines. The increased heat load on the Control Building HVAC systems due to he internal chase temperature being greater than the outside temperature requires an increased minimum air flow less than the design supply air flow for the coolers. The ESW Vertical Loop Chase is unheated. ESW passing through the chase will not freeze due to fiberglass insulation on the pipe and the short duration of time water will spend in the chase based on the high rate of flow through the pipe. The pipe chase has no impact on the winter heating load for te Control Building.
9.4-94 Rev. 29 WOLF CREEK TABLE 9.4-1 OUTSIDE ENVIRONMENT DESIGN CONDITIONS Summer Winter____ Dry Wet Dry Wind Bulb Bulb Bulb Velocity
(F) (F) (F) (mph)
Normal Design Conditions 97 79 7 15 (See Note 1)
Extreme Design Conditions 120 -- (-)30 --
(See Note 2)
NOTES:
- 1. The outdoor ambient temperatures are taken from the 1972 ASHRAE Handbook of Fundamentals, Weather Data and Design Conditions, Chapter 33, Table 1. Summer 97-1/2 percent values are used. The winter wind velocity was assumed for conservatism. Portions of WCGS may be designed to the original SNUPPS normal winter temperature of -25°F.
- 2. All safety-related HVAC systems and components which are exposed to the outside environment are capable of
sustaining the WCGS extreme temperature conditions without loss of function (see Section 3.11). However, no HVAC system (safety-related or nonsafety-related) is
designed to maintain space design temperatures while
operating during the WCGS extreme temperature conditions.
Portions of WCGS may be designed to the original SNUPPS extreme winter temperature of -60°F. Rev. 6 WOLFCREEKTABLE9.4-2DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.52,REVISION2,DATEDMARCH1978,TITLED"DESIGN,TESTING, ANDMAINTENANCECRITERIAFORPOST-ACCIDENTENGINEERED-SAFETY-FEATUREATMOSPHERECLEANUPSYSTEMAIRFILTRATIONANDADSORPTIONUNITSOFLIGHT-WATER-COOLEDNUCLEARPOWERPLANTS."DesignrequirementsofthisRegulatoryGuideareapplicabletothe followingexhaustsystems:a.Emergencyexhaustb.Controlroomfiltrationc.ControlroompressurizationDesignrequirementsfornonsafety-relatednormalexhaustsystems are discussedinTable9.4-3.RegulatoryGuide1.52PositionWCGS1.EnvironmentalDesign1.EnvironmentalDesignCriteriaCriteriaa.Thedesignofana.Complies. engineered-safety-featureatmospherecleanupsystem shouldbebasedonthemaxi-mumpressuredifferential,radiationdoserate,rela-tivehumidity,maximumand minimumtemperature,and otherconditionsresulting fromthepostulatedDBAand onthedurationofsuch
conditions.Rev.0 WOLFCREEKTABLE9.4-2(Sheet2)RegulatoryGuide1.52PositionWCGSb.Thedesignofeachb.Complies.ESFsystemshouldbebasedon theradiationdosetoessential servicesinthevicinityofthe adsorbersection,integrated overthe30-dayperiodfollowing thepostulatedDBA.The radiationsourcetermshould beconsistentwiththeas-sumptionsfoundinRegulatory Guides1.3(Ref.5),1.4 (Ref.6)and1.25(Ref.7). Otherengineeredsafety features,includingpertinent componentsofessentialser-vicessuchaspower,air,and controlcablesshouldbe adequatelyshieldedfromthe ESFatmospherecleanupsys-
tems.c.Thedesignofeachc.Complies.adsorbershouldbebasedon theconcentrationandrela-tiveabundanceoftheiodine species(elemental,particu-late,andorganic),which shouldbeconsistentwith theassumptionsfoundin RegulatoryGuides1.3(Ref. 5),1.4(Ref.6),and1.25 (Ref.7).d.Theoperationofanyd.Complies.ESFatmospherecleanupsystem shouldnotdeleteriously affecttheoperationofother engineeredsafetyfeatures suchasacontainmentspray system,norshouldtheopera-tionofotherengineered safetyfeaturessuchasa containmentspraysystem deleteriouslyaffectthe operationofanyESFatmo-spherecleanupsystem.Rev.0 WOLFCREEKTABLE9.4-2(Sheet3)RegulatoryGuide1.52PositionWCGSe.Componentsofsys-e.Complies.temsconnectedtocompart-mentsthatareunheated duringapostulatedaccident shouldbedesignedforpost-accidenteffectsofboththe lowestandhighestpredicted
temperatures.2.SystemDesignCriteria2.SystemDesignCriteriaa.ESFatmospherea.1.Controlroomcleanupsystemsdesignedpressurizationsystemandinstalledforthepurposecomplies. ofmitigatingaccidentdoses shouldberedundant.Thea.2.Emergencyexhaust systemsshouldconsistofsystemcomplies,except thefollowingsequentialthatdemistersarenot components:(1)demisters,provided.Waterdroplets (2)prefilters(demistersarenotentrainedin mayservethisfunction),theairstream. (3)HEPAfiltersbeforethe adsorbers,(4)iodineadsor-a.3.Controlroom bers(impregnatedactivatedfiltrationsystem carbonorequivalentadsor-complies,exceptthat bentsuchasmetalzeolites),demistersarenot (5)HEPAfiltersaftertheprovided.Water adsorbers,(6)ductsanddropletsarenoten-valves,(7)fans,and(8)trainedintheair-relatedinstrumentation.stream.Humidity Heatersorcoolingcoilscontrolisprovided usedinconjunctionwithbysafety-relatedair-heatersshouldbeusedwhenconditioningsystem thehumidityistobecon-whichhasprovisions trolledbeforefiltration.fordehumidifying.b.TheredundantESFb.Complies.atmospherecleanupsystems shouldbephysicallyseparated sothatdamagetoonesystem doesnotalsocausedamage tothesecondsystem.The generationofmissilesfrom high-pressureequipment rupture,rotatingmachinery failure,ornaturalphenomena shouldbeconsideredinthedesign forseparationandprotection.Rev.0 WOLFCREEKTABLE9.4-2(Sheet4)RegulatoryGuide1.52PositionWCGSc.Allcomponentsofc.Complies.anengineered-safety-feature atmospherecleanupsystem shouldbedesignatedas SeismicCategoryI(seeReg-ulatoryGuide1.29(Ref.8)) iffailureofacomponent wouldleadtotherelease ofsignificantquantities offissionproductstothe workingoroutdoorenviron-
ments.d.IftheESFatmo-d.Notapplicable.Thespherecleanupsystemissystemsarelocated subjecttopressuresurgesoutsideofthecon-resultingfromthepostu-tainmentandnot latedaccident,thesystemexposedtopressure shouldbeprotectedfromsurges. suchsurges.Eachcompo-nentshouldbeprotected withsuchdevicesaspressure reliefvalvessothatthe overallsystemwillperform itsintendedfunctionduring andafterthepassageofthe pressuresurge.e.Inthemechanicale.Complies.designoftheESFsystem, thehighradiationlevels thatmaybeassociatedwith buildupofradioactivemate-rialsontheESFsystemcom-ponentsshouldbegivenpar-ticularconsideration.ESF systemconstructionmate-rialsshouldeffectively performtheirintendedfunc-tionunderthepostulated radiationlevels.Theef-fectsofradiationshould beconsiderednotonlyfor thedemisters,heaters, HEPAfilters,adsorbers,Rev.0 WOLFCREEKTABLE9.4-2(Sheet5)RegulatoryGuide1.52PositionWCGSandfans,butalsoforany electricalinsulation,con-trols,joiningcompounds, dampers,gaskets,andother organic-containingmaterials thatarenecessaryforoper-ationduringapostulated
DBA.f.Thevolumetricairf.Complies.flowrateofasingleclean-uptrainshouldbelimited toapproximately30,000ft3 /min.Ifatotalsystemair flowinexcessofthisrate isrequired,multipletrains shouldbeused.Foreaseof maintenance,afilterlayout threeHEPAfiltershighand tenwideispreferred.g.TheESFatmosphereg.Complies,exceptthatcleanupsystemshouldbein-flowratesarenot strumentedtosignal,alarm,recorded.Highandlow andrecordpertinentpres-differentialpressure suredropsandflowratesalarmsinthecontrol atthecontrolroom.roomprovideindicationofanyabnormalityin flowrates.h.Thepowersupplyandh.Complies.electricaldistributionsys-temfortheESFatmosphere cleanupsystemdescribedin SectionC.2.aaboveshouldbe designedinaccordancewith RegulatoryGuide1.32(Ref.9). Allinstrumentationand equipmentcontrolsshouldbe designedtoIEEEStandard279 (Ref.10).TheESFsystem shouldbequalifiedandtested underRegulatoryGuide1.89 (Ref.11).Totheextent applicable,RegulatoryGuides 1.30(Ref.12),1.100(Ref.13), and1.118(Ref.14)andIEEE Standard334(Ref.15)shouldbe consideredinthedesign.Rev.0 WOLFCREEKTABLE9.4-2(Sheet6)RegulatoryGuide1.52PositionWCGSi.Unlesstheapplica-i.Complies.bleengineered-safety-feature atmospherecleanupsystem operatescontinuouslyduring alltimesthataDBAcanbe postulatedtooccur,the systemshouldbeautomatically activatedupontheoccurrence ofaDBAby(1)aredundant
engineered-safety-feature signal(i.e.,temperature, pressure)or(2)asignalfrom redundantSeismicCategoryI radiationmonitors.j.Tomaintainradiationj.Complies.exposurestooperatingper-sonnelaslowasisreasonably achievableduringplant maintenance,ESFatmosphere cleanupsystemsshouldbe designedtocontrolleakageand facilitatemaintenancein accordancewiththeguidelines ofRegulatoryGuide8.8(Ref. 16).TheESFatmospherecleanup trainshouldbetotally enclosed.Eachtrainshouldbe designedandinstalledina mannerthatpermitsreplacement ofthetrainasanintactunit orasaminimumnumberofsegmentedsectionswithout removalofindividual
components.k.Outdoorairintakek.Complies.openingsshouldbeequippedwithlouvers,grills,screens, orsimilarprotectivedevices tominimizetheeffectsof highwinds,rain,snow,ice, trash,andothercontaminants ontheoperationofthesys-tem.Iftheatmospheresur-roundingtheplantcouldcon-tainsignificantenvironmental contaminants,suchasRev.0 WOLFCREEKTABLE9.4-2(Sheet7)RegulatoryGuide1.52PositionWCGSdustsandresiduesfromsmoke cleanupsystemsfromadjacent coalburningpowerplantsor industry,thedesignofthe systemshouldconsider thesecontaminantsand preventthemfromaffecting theoperationofanyESF atmospherecleanupsystem.1.ESFatmosphereclean-1.Complies.upsystemhousingsandduct-workshouldbedesignedto exhibitontestamaximum totalleakagerateasdefined inSection4.12ofANSIN509-1976(Ref.1).Ductand housingleaktestsshould beperformedinaccordance withtheprovisionsof Section6ofANSIN510-1975 (Ref.2).3.ComponentDesignCri-3.ComponentDesignCri-teriaandQualificationteriaandQualificationTestingTestinga.Demistersshouldbea.1.Notapplicabledesigned,constructed,andtoemergencyexhausttestedinaccordancewiththeandcontrolroom requirementsofSection5.4filtrationsystem. ofANSIN509-1976(Ref.1).SeeresponsetoDemistersshouldmeetUnder-RegulatoryPositionwriters'Laboratories(UL)2.aabove. Class1(Ref.17)require-ments.a.2.Controlroompres-surizationsystem complies.b.Airheatersshouldbeb.1.Notapplicabletodesigned,constructed,andcontrolroomfiltra-testedinaccordancewiththetionsystem.See requirementsofSection5.5responsetoRegula-ofANSIN509-1976(Ref.1).toryPosition2.aabove.b.2.Controlroom pressurizationsystem andemergencyexhaust systemcomplies.Rev.0 WOLFCREEKTABLE9.4-2(Sheet8)RegulatoryGuide1.52PositionWCGSc.Materialsusedinc.Complies,excepttheprefiltersshouldwith-thatnoprefilters standtheradiationlevelsareusedforcon-andenvironmentalconditionstrolroompressur-prevalentduringthepostu-izationsystem. latedDBA.Prefilters shouldbedesigned,con-structed,andtestedin accordancewiththeprovi-sionsofSection5.3of ANSIN509-1976(Ref.1).d.TheHEPAfiltersd.Complies.shouldbedesigned,con-structed,andtestedin accordancewithSection5.1ofANSIN509-1976(Ref.1).EachHEPAfiltershouldbetestedforpenetrationof dioctylphthalate(DOP)in accordancewiththeprovi-sionsofMIL-F-51068(Ref. 19)andMIL-STD-282(Ref.
20).2e.Filterandadsorbere.Complies.mountingframesshouldbe constructedanddesignedin accordancewiththeprovi-sionsofSection5.6.3of ANSIN509-1976(Ref.1).f.Filterandadsorberf.Complies.banksshouldbearrangedin accordancewiththerecom-mendationsofSection4.4of ERDA76-21(Ref.3).g.Systemfilterg.Complies.housings,includingfloors anddoors,shouldbecon-structedanddesignedin accordancewiththeprovi-sionsofSection5.6of ANSIN509-1976(Ref.1).Rev.14 WOLFCREEKTABLE9.4-2(Sheet9)RegulatoryGuide1.52PositionWCGSh.Waterdrainsshouldh.Complies.bedesignedinaccordance withtherecommendationsof Section4.5.8ofERDA76-21 (Ref.3).i.Theadsorbersectioni.Complies,exceptthatoftheESFatmosphereclean-therepresentativeupsystemmaycontainanysampleswillbeadsorbentmaterialdemonstratedlaboratorytestedintoremovegaseousiodineaccordancewiththe(elementaliodineandorganicrequirementsofASTMiodides)fromairattheD3803-1989.requiredefficiency.Since impregnatedactivatedcarbonis commonlyused,onlythis adsorbentisdiscussedinthis guide.Eachoriginalor replacementbatchofimpregnated activatedcarbonusedinthe adsorbersectionshouldmeetthe qualificationandbatchtest resultssummarizedinTable5.1 ofANSIN509-1976(Ref.1).In thistable,a"qualification test"shouldbeinterpretedto meanatestthatestablishesthe suitabilityofaproductfora generalapplication,normallya one-timetestreflecting historicaltypicalperformance ofmaterial.Inthistable,a "batchtest"shouldbe interpretedtomeanatestmade onaproductionbatchofproduct toestablishsuitabilityfora specificapplication.A"batch ofactivatedcarbon"shouldbe interpretedtomeanaquantity ofmaterialofthesamegrade, type,andseriesthathasbeen homogenizedtoexhibit,within reasonabletolerance,thesame performanceandphysicalRev.14 WOLFCREEKTABLE9.4-2(Sheet10)RegulatoryGuide1.52PositionWCGScharacteristicsandforwhich themanufacturercandemon-stratebyacceptabletests andqualitycontrolpractices suchuniformity.Allmaterialinthesamebatchshouldbeactivated,impregnated,andotherwise treatedunderthesame processconditionsand proceduresinthesame processequipmentandshould beproducedunderthesame manufacturingreleaseand instructions.Material producedinthesamecharge ofbatchequipment constitutesabatch;material producedindifferentcharges ofthesamebatchequipment shouldbeincludedinthe samebatchonlyifitcanbe homogenizedasabove.The maximumbatchsizeshouldbe 350ft3ofactivatedcarbon.Ifanadsorbentotherthanimpregnatedactivated carbonisproposedorifthe meshsizedistributionis differentfromthe specificationsinTable5.1 ofANSIN509-1976(Ref.1), theproposedadsorbentshouldhavedemonstratedthe capabilitytoperformaswell asorbetterthanactivated carboninsatisfyingthe specificationsinTable5.1 ofANSIN509-1976(Ref.1).Ifimpregnatedactivatedcarbonisusedastheadsor-bent,theadsorbersystem shouldbedesignedforan averageatmosphereresidence timeof0.25secpertwo inchesofadsorbentbed.Rev.0 WOLFCREEKTABLE9.4-2(Sheet11)RegulatoryGuide1.52PositionWCGSTheadsorptionunitshould bedesignedforamaximum loadingof2.5mgoftotal iodine(radioactiveplus stable)pergramofactivated carbon.Nomorethan5%of impregnant(50mgofimpreg-nantpergramofcarbon) shouldbeused.Theradia-tionstabilityofthetype ofcarbonspecifiedshould bedemonstratedandcertified (seeSectionC.1.bofthis guideforthedesignsource
term).j.Adsorbercellsshouldj.Complies.bedesigned,constructed,and testedinaccordancewiththe requirementsofSection5.2of ANSIN509-1976(Ref.1).k.Thedesignofthek.1.Emergencyexhaustadsorbersectionshouldcon-systemcomplies.Charcoal siderpossibleiodinede-bedtemperatureismaintained sorptionandadsorbentauto-belowdesorptionrangeby ignitionthatmayresultassuringaminimumairflow fromradioactivity-inducedacrosstheloadedbed. heatintheadsorbentand concomitanttemperaturerise. Acceptabledesignsincludeak.2.Controlroomfil-low-flowairbleedsystem,trationandcontrol coolingcoils,waterspraysroompressuriza-fortheadsorbersection,ortionsystemscomply. othercoolingmechanisms.Anticipatedcharcoal Anycoolingmechanismshouldbedloadingisnot satisfythesingle-failuresufficienttoraise criterion.Alow-flowairbedtemperatureto bleedsystemshouldsatisfythedesorptionrange. thesingle-failurecriterionHowever,manually forprovidinglow-humidityactuatedwatersprays (lessthan70%relativehu-areprovided,ifrequiredto midity)coolingairflow.preventormitigateignition.Rev.0 WOLFCREEKTABLE9.4-2(Sheet12)RegulatoryGuide1.52PositionWCGSl.Thesystemfan,itsl.Complies.mounting,andtheductwork connectionsshouldbede-signed,constructed,and testedinaccordancewith therequirementsofSections 5.7and5.8ofANSIN509-1976(Ref.1).m.Thefanorblowerm.Complies.usedontheESFatmosphere cleanupsystemshouldbe capableofoperatingunder theenvironmentalcondi-tionspostulated,including
radiation.n.Ductworkshouldben.Complies.designed,constructed,and testedinaccordancewith theprovisionsofSection 5.10ofANSIN509-1976 (Ref.1).o.Ductsandhousingso.Complies.shouldbelaidoutwitha minimumofledges,protru-sions,andcrevicesthat couldcollectdustand moistureandthatcould impedepersonnelorcreate ahazardtothemintheper-formanceoftheirwork.Straighteningvanesshould beinstalledwhererequired toensurerepresentativeair flowmeasurementanduniform flowdistributionthrough cleanupcomponents.p.Dampersshouldbep.Complies.Dampersdesigned,constructed,andaredesigned,construct-testedinaccordancewiththeed,andtestedinaccor-provisionsofSection5.9ofdancewithcodesand ANSIN509-1976(Ref.1).standardscomparablewiththeprovisionsofSection5.9ofANSI
N509-1976.Rev.0 WOLF CREEK TABLE 9.4-2 (Sheet 13) Regulatory Guide 1.52 Position WCGS
- 4. Maintenance 4. Maintenance
- a. Accessibility of a. Complies.
components and maintenance should be considered in the
design of ESF atmosphere cleanup systems in accor-dance with the provisions
of Section 2.3.8 of ERDA 76-21 (Ref. 3) and Section 4.7 of ANSI N509-1976 (Ref. 1).
- b. For ease of main- b. Complies where
tenance, the system design internal removal of should provide for a minimum components is required. of three feet from mounting
frame to mounting frame between banks of components. If components are to be re-
placed, the dimension to be provided should be the maxi-mum length of the component
plus a minimum of three feet.
- c. The system design c. Complies.
should provide for permanent test probes with external connections in accordance
with the provisions of Sec-tion 4.11 of ANSI N509-1976 (Ref. 1).
- d. Each ESF atmosphere d. Each train is operated for cleanup train should be greater than or equal to 15 operated at least 10 hours continuous minutes in per month, with the heaters accordance with the technical on (if so equipped), in specifications.
order to reduce the build-
up of moisture on the adsorbers and HEPA filters.
- e. The cleanup compo- e. Complies.
nents (i.e., HEPA filters, prefilters, and adsorbers) should not be installed while active construction
is still in progress. Rev. 28 WOLFCREEKTABLE9.4-2(Sheet14)RegulatoryGuide1.52PositionWCGS5.In-PlaceTesting5.In-PlaceTestingCriteriaCriteriaa.Avisualinspectiona.Complies.oftheESFatmosphereclean-upsystemandallassociated componentsshouldbemade beforeeachin-placeairflow distributiontest,DOPtest, oractivatedcarbonadsorber sectionleaktestin accordancewiththe provisionsofSection5of ANSIN510-1975(Ref.2).b.Theairflowdistri-b.Complies,exceptbutiontotheHEPAfiltersthattestingconductedandiodineadsorbersshouldinaccordancewithANSI betestedinplaceforuni-N510utilizesthe1980 formityinitiallyandaftereditioninlieuofthe maintenanceaffectingthe1975edition.However, flowdistribution.Thedis-airflowdistribution tributionshouldbewithintestingisperformed only+20%oftheaverageflowonthedown-streamsideperunit.ThetestingshouldofthefirstHEPA filtersbeconductedinaccordanceinlieuofeachfilter,withtheprovisionsofSec-asstatedinSection8.1 tion9of"IndustrialVenti-ofANSIN510-1980. lation"(Ref.21)andSec-tion8ofANSIN510-1975 (Ref.2).c.Thein-placeDOPtestc.Complies,exceptthattheforHEPAfiltersshouldcon-TechnicalSpecification formtoSection10ofANSIacceptancecriteriaof N510-1975(Ref.2).HEPAlessthan1.0percent filtersectionsshouldbein-placepenetrationand testedinplace(1)ini-bypassleakageshallbe tially,(2)atleastonceemployed.TheTechnical per18monthsthereafter,Specificationrequire-and(3)followingpainting,ments,althoughless fire,orchemicalreleasestringentthanthe inanyventilationzonecom-Reg.Guide,stillallow municatingwiththesystemaconservativedesign, toconfirmapenetrationofastheaccidentdose lessthan0.05%atratedevaluationassumesa flow.Anengineered-safety-95percentefficiency.Rev.0 WOLFCREEKTABLE9.4-2(Sheet15)RegulatoryGuide1.52PositionWCGSfeatureairfiltrationsystemThein-placeDOPtesting satisfyingthisconditionisperformedinaccor-canbeconsideredtodancewithSection10of warranta99%removalANSIN510-1075,butthe efficiencyforparticulatesprerequisitetestingin inaccidentdoseevaluations.Sections8and9isper-HEPAfiltersthatfailtoformedinaccordance satisfythisconditionwiththe1980versionin wouldbereplacedwithlieuofthe1975version filtersqualifiedpursuantexceptthatDOPisnot toregulatorypositionC.3.dinjectedbetweenthetwo ofthisguide.IftheHEPAHEPAbanks. filterbankisentirelyor onlypartiallyreplaced,an in-placeDOPtestshouldbe conducted.Ifanywelding repairsarenecessaryon, within,oradjacenttothe ducts,housing,ormounting frames,thefiltersand adsorbersshouldberemoved fromthehousingduringsuch repairs.Therepairsshould becompletedpriorto periodictesting,filter inspection,andin-place testing.Theuseofsilicone sealantsoranyother temporarypatchingmaterial onfilters,housing,mounting frames,orductsshouldnot beallowed.d.Theactivatedcarbond.Complies,exceptthatadsorbersectionshouldbetheprerequisitetesting leaktestedwithagaseousinSections8and9is halogenatedhydrocarbonre-performedinaccordance frigerantinaccordancewithwithANSIN510-1980. Section12ofANSIN510-1975 (Ref.2)toensurethat bypassleakagethroughthe adsorbersectionisless than0.05%.Afterthetest iscompleted,airflow throughtheunitshouldbeRev.0 WOLFCREEKTABLE9.4-2(Sheet16)RegulatoryGuide1.52PositionWCGSmaintaineduntiltheresidualrefrigerantgasintheeffluent islessthan0.01ppm.Adsorberleaktestingshouldbeconducted(1)initially,(2)atleastonce per18monthsthereafter,(3) followingremovalofanadsorber sampleforlaboratorytestingiftheintegrityoftheadsorbersectionisaffected,and(4) followingpainting,fire,or chemicalreleaseinany ventilationzonecommunicating withthesystem.6.LaboratoryTestingCriteriaforActivatedCarbon6.LaboratoryTestingCriteriaforActivatedCarbona.TheactivatedcarbonadsorbersectionoftheESF atmospherecleanupsystemshould beassignedthedecontamination efficienciesgiveninTable2 forelementaliodineandorganic iodidesifthefollowing conditionsaremet:a.Complies,exceptthattherepresentativesampleswillbelaboratorytestedinaccordancewiththerequirementsofASTMD3803-1989.(1)TheadsorbersectionmeetstheconditionsgiveninregulatorypositionC.5.dofthisguide.(2)NewactivatedcarbonmeetsthephysicalpropertyspecificationsgiveninTable 5.1ofANSIN509-1976(Ref.1), and(3)Representativesamplesofusedactivatedcarbonpassthelaboratorytestsgivenin Table2.Iftheactivatedcarbonfailstomeetanyoftheabove conditions,itshouldnotbeusedinengineered-safety-featureadsorbers.Rev.14 WOLFCREEKTABLE9.4-2(Sheet17)RegulatoryGuide1.52PositionWCGSb.Theefficiencyoftheactivatedcarbonadsorber sectionshouldbedeterminedbylaboratorytestingofrepresentativesamplesofthe activatedcarbonexposedb.Complies,exceptthattherepresentativesamples willbelaboratorytestedinaccordancewiththerequirementsofASTM
D3803-1989.simultaneouslytothesameserviceconditionsasthe adsorbersection.Eachrepresentativesampleshouldbenotlessthantwoinchesinboth lengthanddiameter,andeach sampleshouldhavethesame qualificationandbatchtestcharacteristicsasthesystemadsorbent.Thereshouldbea sufficientnumberof representativesampleslocated inparallelwiththeadsorber sectiontoestimatetheamount ofpenetrationofthesystem adsorbentthroughoutitsservicelife.Thedesignofthe samplersshouldbeinaccordance withtheprovisionsofAppendix AofANSIN509-1976(Ref.1). Wherethesystemactivated carbonisgreaterthantwo inchesdeep,eachrepresentative samplingstationshouldconsist ofenoughtwo-inchsamplesin seriestoequalthethicknessof thesystemadsorbent.Once representativesamplesare removedforlaboratorytest, theirpositionsinthesampling arrayshouldbeblockedoff.Laboratorytestsofrepresentativesamplesshouldbe conducted,asindicatedinTable 2ofthisguide,withthetest gasflowinthesamedirection astheflowduringRev.14 WOLFCREEKTABLE9.4-2(Sheet18)RegulatoryGuide1.52PositionWCGSserviceconditions.Similar laboratorytestsshouldbe performedonanadsorbent samplebeforeloadingintothe adsorberstoestablishan initialpointforcomparison offuturetestresults.The activatedcarbonadsorber sectionshouldbereplaced withnewunusedactivated carbonmeetingthephysical propertyspecificationsof Table5.1ofANSIN509-1976 (Ref.1)if(1)testingin accordancewiththefrequency specifiedinFootnotecof Table2resultsinarepre-sentativesamplefailingto passtheapplicabletestin Table2or(2)norepresen-tativesampleisavailable fortesting. 1ThepertinentqualityassurancerequirementsofAppendixB,"QualityAssuranceCriteriaforNuclearPowerPlantsandFuelReprocessingPlants,"to10CFRPart50applytoallactivities affectingthesafety-relatedfunctionsofHEPAfilters. 2TheUSDepartmentofEnergy(USDOE)operatesanumberoffiltertestfacilitiesqualifiedtoperformHEPAfilterefficiencytests.ThesefacilitiesarelistedinthecurrentUSDOEEnvironmental SafetyandHealthInformationBulletinforFilterUnitInspectionandTestingService(Ref.18).*Allstatementsapplytoallthreeexhaustsystemslistedabove,unlessotherwiseindicated.Rev.0 WOLFCREEKTABLE9.4-3DESIGNCOMPARISONTOREGULATORYPOSITIONSOFREGULATORYGUIDE1.140,REVISION0,DATEDMARCH1978,TITLED"DESIGN,TESTING,ANDMAINTENANCECRITERIAFOR NORMALVENTILATIONEXHAUSTSYSTEMAIRFILTRATIONANDADSORPTIONUNITSOFLIGHT-WATER-COOLEDNUCLEARPOWERPLANTS."DesignrequirementsofthisRegulatoryGuideareapplicabletothe followingstandardizedpowerplantexhaustsystems:a.Condenserairremovalfiltrationb.Radwastebuildingexhaustc.Accesscontrolexhaust d.Containmentatmosphericcontrol e.Containmentpurge f.Auxiliary/fuelbuildingnormalexhaustDesignrequirementsforsafety-relatedexhaustsystemsare discussed inTable9.4-2.RegulatoryGuide1.140PositionWCGS1.EnvironmentalDesign1.EnvironmentalDesignCriteriaCriteriaa.Thedesignofeacha.Complies.atmospherecleanupsystemin-stalledinanormalventila-tionexhaustsystemshouldbe basedonthemaximumantici-patedoperatingparametersoftemperature,pressure, relativehumidity,and radiationlevels.The cleanupsystemshouldbe designedbasedoncontinuous operationfortheexpected lifeoftheplantorthemaximumanticipatedservice lifeofthecleanupsystem.Rev.0 WOLFCREEKTABLE9.4-3(Sheet2)RegulatoryGuide1.140PositionWCGSb.Iftheatmosphereb.Complies.cleanupsystemislocated inanareaofhighradi-ationduringnormalplant operation,adequate shieldingofcomponents fromtheradiationsource shouldbeprovided.c.Theoperationofanyc.Complies.atmospherecleanupsystemin anormalventilationexhaust systemshouldnotdeleteri-ouslyaffecttheexpected operationofanyengineered-safety-featuresystemthat mustoperateafteradesign basisaccident.d.Thedesignofthed.Complies.atmospherecleanupsystem shouldconsideranysignifi-cantcontaminantssuchas dusts,chemicals,orother particulatematterthat coulddeleteriouslyaffect thecleanupsystem'sopera-
tion.2.SystemDesignCriteria2.SystemDesignCriteriaa.Atmospherecleanupa.Complies.Heaterssystemsinstalledinnormalorcoolingcoilsareventilationexhaustsystemsnotrequired. neednotberedundantnor designedtoseismicCategoryI classification,butshould consistofthefollowingse-quentialcomponents:(1) HEPAfiltersbeforethead-sorbers,(2)iodineadsor-bers(impregnatedacti-vatedcarbonorequivalent adsorbentsuchasmetal zeolites),(3)ductsand dampers,(4)fans,andRev.0 WOLFCREEKTABLE9.4-3(Sheet3)RegulatoryGuide1.140PositionWCGS(5)relatedinstrumentation. Ifitisdesiredtoreduce theparticulateloadon theHEPAfiltersandextend theirservicelife,thein-stallationofprefilters upstreamoftheinitial HEPAbankissuggested. Considerationshouldalso begiventotheinstalla-tionofaHEPAfilterbank downstreamofcarbonadsor-berstoretaincarbonfines. Heatersorcoolingcoilsused inconjunctionwithheaters shouldbeusedwhenthehu-midityistobecontrolled beforefiltration.b.Thevolumetricairb.Complies,exceptthatflowrateofasingleclean-airflowratefor uptrainshouldbelimitedauxiliary/fuelbuilding toapproximately30,000normalexhaustis ft 3/min.Ifatotalsystem32,000cfm(multipleairflowinexcessofthistrainsarenotusedrateisrequired,multipleduetospacelimi-trainsshouldbeused.Fortation). easeofmaintenance,a filterlayoutthatisthree HEPAfiltershighandten wideispreferred.c.Eachatmospherec.Complies.Adif-cleanupsystemshouldbeferentialpressure locallyinstrumentedtoindicatorcontroller monitorandalarmpertinentmodulatesadamper pressuredropsandflowdownstreamofthefilterratesinaccordancewithtraintomaintainatherecommendationsofSec-constantsystemresis-tion5.6ofERDA76-21tanceasthefilters(Ref.3).loadup.Thisarrangementassuresaconstantsystemflow.Rev.14 WOLFCREEKTABLE9.4-3(Sheet4)RegulatoryGuide1.140PositionWCGSHighandlow differentialpressurealarmsprovide indicationofanyabnor-malityinflowrates.d.Tomaintainthed.Complies.radiationexposuretooper-atingpersonnelaslowasis reasonablyachievableduring plantmaintenance,atmosphere cleanupsystemsshouldbede-signedtocontrolleakageand facilitatemaintenanceinac-cordancewiththeguidelines ofRegulatoryGuide8.8, "InformationRelevanttoEn-suringthatOccupational RadiationExposuresatNu-clearPowerStationsWill BeAsLowAsIsReasonably Achievable"(Ref.5).e.Outdoorairintakee.Complies.openingsshouldbeequipped withlouvers,grills,screens, orsimilarprotectivedevices tominimizetheeffectsof highwinds,rain,snow,ice, trash,andothercontaminants ontheoperationofthesys-tem.Iftheatmospheresur-roundingtheplantcouldcontainsignificantenviron-mentalcontaminants,suchas dustsandresiduesfrom smokecleanupsystemsfrom adjacentcoalburningpower plantsorindustry,the designofthesystemshould considerthesecontaminants andpreventthemfromaf-fectingtheoperationof anyatmospherecleanup
system.Rev.0 WOLFCREEKTABLE9.4-3(Sheet5)RegulatoryGuide1.140PositionWCGSf.Atmospherecleanupf.Complies.systemhousingsandductwork shouldbedesignedtoex-hibitontestamaximum totalleakagerateasde-finedinSection4.12of ANSIN509-1976(Ref.1). Ductandhousingleaktests shouldbeperformedinac-cordancewiththeprovisions ofSection6ofANSIN510-1975(Ref.2).3.ComponentDesignCriteria3.ComponentDesignCriteriaandQualificationTestingandQualificationTesting a.Adsorptionunitsa.Notapplicabletofunctionefficientlyatathesesystems.Seerelativehumidityof70%orresponsetoRegulatory less.IftherelativePosition2.a.above. humidityoftheincoming atmosphereisexpectedto begreaterthan70%during normalreactoroperation, heatersorcoolingcoils usedinconjunctionwith heatersshouldbedesigned toreducetherelative humidityoftheincoming atmosphereto70%.Heaters shouldbedesigned,con-structed,andtestedinac-cordancewiththerequire-mentsofSection5.5of ANSIN509-1976(Ref.1) exclusiveofsizingcriteria.b.TheHEPAfiltersb.Complies.shouldbedesigned,con-structed,andtestedinac-cordancewiththerequire-mentsofSection5.1of ANSIN509-1976(Ref.1). EachHEPAfiltershouldbe testedforpenetrationof dioctylphthalate(DOP)in accordancewiththeprovi-sionsofMIL-F-51068(Ref. 6)andMIL-STD-282(Ref.7).Rev.0 WOLFCREEKTABLE9.4-3(Sheet6)RegulatoryGuide1.140PositionWCGSc.Filterandadsorberc.Complies.mountingframesshouldbe designedandconstructedin accordancewiththeprovi-sionsofSection5.6.3of ANSIN509-1976(Ref.1).d.Filterandadsorberd.Complies.banksshouldbearrangedin accordancewiththerecom-mendationsofSection4.4 ofERDA76-21(Ref.3).e.Systemfilterhous-e.Complies.ings,includingfloorsand doors,andelectricalcon-duits,drains,andpiping installedinsidefilter housingsshouldbedesigned andconstructedinaccor-dancewiththeprovisions ofSection5.6ofANSI N509-1976(Ref.1).f.Ductworkassociatedf.Complies.withtheatmospherecleanup systemshouldbedesigned, constructed,andtestedin accordancewiththeprovi-sionsofSection5.10of ANSIN509-1976(Ref.1).g.Theadsorbersec-g.Complies.tionoftheatmosphereclean-upsystemmaycontainanyad-sorbentmaterialdemonstrated toremovegaseousiodine (elementaliodineandorganic iodides)fromairatthe requiredefficiency.Since impregnatedactivatedcarbon iscommonlyused,onlythis adsorbentisdiscussedin thisguide.Eachoriginal orreplacementbatchofim-pregnatedactivatedcarbonRev.0 WOLFCREEKTABLE9.4-3(Sheet7)RegulatoryGuide1.140PositionWCGSusedintheadsorbersec-tionshouldmeetthequali-ficationandbatchtestre-sultssummarizedinTable1 ofthisguide.Ifanadsorbentotherthanimpregnatedactivatedcarbonisproposedorifthe meshsizedistributionis differentfromthe specificationsinTable1, theproposedadsorbent shouldhavedemonstratedthe capabilitytoperformas wellasorbetterthan activatedcarboninsatis-fyingthespecificationsin Table1.Ifimpregnated activatedcarbonisusedas theadsorbent,theadsorber systemshouldbedesigned foranaverageatmosphere residencetimeof0.25sec pertwoinchesofadsorbent
bed.h.Adsorbercellsh.Complies.shouldbedesigned,con-structed,andtestedinac-cordancewiththerequire-mentsofSection5.2of ANSIN509-1976(Ref.1).i.Thesystemfanandi.Complies.motors,mounting,andduct-workconnectionsshouldbe designed,constructed,and testedinaccordancewith therequirementsofSec-tions5.7and5.8ofANSI N509-1976(Ref.1).j.Thefanorblowerj.Complies.usedintheatmosphere cleanupsystemshouldbe capableofoperatingunder theenvironmentalcondi-tionspostulated.Rev.0 WOLFCREEKTABLE9.4-3(Sheet8)RegulatoryGuide1.140PositionWCGSk.Ductsandhousingsk.Complies.shouldbelaidoutwitha minimumofledges,protru-sions,andcrevicesthat couldcollectdustand moistureandthatcould impedepersonnelorcreatea hazardtotheminthe performanceoftheirwork. Straighteningvanesshould beinstalledwhererequired toensurerepresentativeair flowmeasurementanduniform flowdistributionthrough cleanupcomponents.l.Dampersshouldbel.Complies.Dampersaredesigned,constructed,anddesigned,constructed, andtestedinaccordancewithtestedinaccordance withtheprovisionsofSectioncodesandstandardscom-5.9ofANSIN509-1976parablewiththeprovi-(Ref.1).sionsofSection5.9ofANSIN509-1976.4.Maintenance4.Maintenancea.Accessibilityofa.Complies.componentsandmaintenance shouldbeconsideredinthe designofatmosphereclean-upsystemsinaccordancewiththeprovisionsofSec-tion2.3.8ofERDA76-21 (Ref.3)andSection4.7of ANSIN509-1976(Ref.1).b.Foreaseofmain-b.Complieswhereinternaltenance,thesystemdesignremovalofcomponentis shouldprovideforamini-required. mumofthreefeetfrom mountingframetomounting framebetweenbanksofcom-ponents.Ifcomponentsare tobereplaced,thedimen-sionstobeprovidedshould bethemaximumlengthof thecomponentplusamini-mumofthreefeet.Rev.0 WOLFCREEKTABLE9.4-3(Sheet9)RegulatoryGuide1.140PositionWCGSc.Thesystemdesignc.Complies.shouldprovideforpermanent testprobeswithexternal connectionsinaccordance withtheprovisionsofSec-tion4.11ofANSIN509-1976(Ref.1).d.Thecleanupcompo-d.Complies.nents(e.g.,HEPAfilters andadsorbers)shouldbein-stalledafterconstruction iscompleted.5.In-PlaceTesting5.In-PlaceTestingCriteria Criteriaa.Avisualinspectiona.Complies.oftheatmospherecleanupsystemandallassociated componentsshouldbemade beforeeachin-placeair-flowdistributiontest, DOPtest,oractivatedcar-bonadsorbersectionleak testinaccordancewiththe provisionsofSection5of ANSIN510-1975(Ref.2).b.Theairflowdistri-b.Complies,exceptthatbutiontotheHEPAfilterstestingconductedin andiodineadsorbersshouldaccordancewithANSIbetestedinplaceforuni-N510utilizesthe1980formityinitiallyandaftereditioninlieuofthe maintenanceaffectingthe1975edition.However, flowdistribution.Theairflowdistribution distributionshouldbetestingisperformed onlywithin+20%oftheaverageonthedownstreamsideflowperunitwhentestedofthefirstHEPA inaccordancewiththepro-filtersinlieuofeach visionsofSection9offilterasstatedin"IndustrialVentilation"Section8.1ofANSI(Ref.8)andSection8ofN510-1980.ANSIN510-1975(Ref.2).Rev.0 WOLFCREEKTABLE9.4-3(Sheet10)RegulatoryGuide1.140PositionWCGSc.Thein-placeDOPc.Complies,exceptthattestforHEPAfiltersshouldanacceptancecri-conformtoSection10ofteriaoflessthan ANSIN510-1975(Ref.2).1.0percentin-place HEPAfiltersectionsshouldpenetrationandbypass betestedinplaceinitiallyleakageisemployed andatafrequencynottotobeconsistentwith exceed18monthsthereaftertheacceptancecriteria (duringascheduledreactorutilizedforthe shutdownisacceptable).engineeredsafety-TheHEPAfilterbankfeaturefiltration upstreamoftheadsorberunits.Thisrequire-sectionshouldalsobement,althoughless testedfollowingpainting,thantherequired.05 fire,orchemicalreleaseinpercent,stillallows anyventilationzoneaconservative communicatingwiththedesign,asthefilter systeminsuchamannerthatefficiencyisassumed theHEPAfilterscouldtobe95percentin becomecontaminatedfromtheaccidentanalyses. fumes,chemicals,orforeignThein-placeDOP materials.DOPpenetrationtestingisperformed testsofallHEPAfilterinaccordancewith banksshouldconfirmaSection10ofANSI penetrationoflessthanN510-1975,butpre-0.05%atratedflow.Arequisitetestingon filtrationsystemsatisfyingSections8and9is thisconditioncanbeconsid-performedinaccor-eredtowarranta99%removaldancewithANSIN510-efficiencyforparticulates.1980,exceptthatDOP HEPAfiltersthatfailtoisnotinjectedbetween satisfythein-placetestthetwoHEPAbanks. criteriashouldbereplacedThisrequirementdoeswithfiltersqualifiedpur-notapplytothesuanttoregulatoryposi-ContainmentAtmospherictionC.3.bofthisguide.ControlSystem.IftheHEPAfilterbankisentirelyofonlypartiallyreplaced,anin-placeDOPtestshouldbeconducted.Ifanyweldingrepairsarenecessaryon,within,oradjacenttotheducts, housing,ormountingframes,thefiltersandadsorbersshouldberemovedfromthe housingduringsuchrepairs.Rev.4 WOLFCREEKTABLE9.4-3(Sheet11)RegulatoryGuide1.140PositionWCGSTheserepairsshouldbecom-pletedpriortoperiodic testing,filterinspection, andin-placetesting.The useofsiliconesealantsor anyothertemporarypatching materialonfilters,housing, mountingframes,orducts shouldnotbeallowed.d.Theactivatedcarbond.Complies,exceptthatadsorbersectionshouldbetheprerequisitetestingleak-testedwithagaseousinSections8and9ishalogenatedhydrocarbonperformedinaccordancerefrigerantinaccordancewithANSIN510-1980,andwithSection12ofANSIthatthisrequirement doesN510-1975(Ref.2)toensurenotapplytothe Contain-thatbypassleakagethroughmentAtmosphericControltheadsorbersectionislessSystems.than0.05%.Afterthetestiscompleted,airflowthroughtheunitshouldbemaintaineduntiltheresidual refrigerantgasinthe effluentislessthan0.01 ppm.Adsorberleaktesting shouldbeconducted(1) initially,(2)atafrequency nottoexceed18months thereafter(duringa scheduledreactorshutdownis acceptable),(3)following removalofanadsorbersample forlaboratorytestingifthe integrityoftheadsorbersectionisaffected,and(4)followingpainting,fire,or chemicalreleaseinany ventilationzonecommuni-catingwiththesystemin suchamannerthatthe charcoaladsorberscould becomecontaminatedfromthe fumes,chemicals,orforeign
materials.Rev.4 WOLFCREEKTABLE9.4-3(Sheet12)RegulatoryGuide1.140PositionWCGS6.LaboratoryTesting6.LaboratoryTestingCriteriaforActivatedCriteriaforActivated CarbonCarbona.Theactivatedcarbona.Complies,exceptthattheadsorbersectionoftherepresentativesamplesatmospherecleanupsystemwillbelaboratorytested shouldbeassignedtheinaccordancewiththe decontaminationefficienciesrequirementsofASTMD-giveninTable2forradio-3803asrecommendedbyiodineifthefollowingtheassociatedFinalconditionsaremet:TechnicalEvaluation(1)TheadsorberReportfortheNRC/INELsectionmeetstheconditionsActivatedCarbonTestinggiveninregulatorypositionProgram,ReportNo.EGG-C.5.dofthisguide,CS-7643,byC.D.Scarpellino (2)NewactivatedandC.W.SillofEG&GIdaho, carbonmeetsthephysicalInc. propertyspecifications giveninTable1,and(3)Representativesamplesofusedactivatedcarbonpassthelaboratory testsgiveninTable2.Iftheactivatedcarbonfailstomeetanyoftheaboveconditions,it shouldnotbeusedin adsorptionunits.b.Theefficiencyofb.Complies,exceptthatthetheactivatedcarbonadsorberrepresentativesampleswillsectionshouldbedeterminedbelaboratorytestedin bylaboratorytestingofaccordancewiththerequire-representativesamplesofthementsofASTM-D-3803asactivatedcarbonexposedrecommendedbythe associatedsimultaneouslytothesameFinalTechnicalEvaluationserviceconditionsastheReportfortheNRC/INEL adsorbersection.Eachrep-ActivatedCarbonTesting resentativesampleshouldnotProgram,ReportNo.EGG-CS-belessthantwoinchesinboth7643,byC.D.Scarpellino lengthanddiameter,andeachandC.W.SillofEG&GIdaho, sampleshouldhavethesameInc. qualificationandbatchtest characteristicsasthesystem
adsorbent.Rev.3 WOLFCREEKTABLE9.4-3(Sheet13)RegulatoryGuide1.140PositionWCGSThereshouldbeasufficient numberofrepresentative sampleslocatedinparallel withtheadsorbersectionto estimatetheamountof penetrationofthesystem adsorbentthroughoutits servicelife.Thedesignof thesamplersshouldbein accordancewiththepro-visionsofAppendixAofANSI N509-1976(Ref.1).Where thesystemactivatedcarbon isgreaterthantwoinches deep,eachrepresentative samplingstationshould consistofenoughtwo-inch samplesinseriestoequal thethicknessofthesystem adsorbent.Oncerepresen-tativesamplesareremoved forlaboratorytest,their positionsinthesampling arrayshouldbeclockedoff.Laboratorytestsofrepresentativesamplesshould beconducted,asindicatedin Table2ofthisguide,with thetestgasflowinthesame directionastheflowduring serviceconditions.Similar laboratorytestsshouldbe performedonanadsorbent samplebeforeloadinginto theadsorberstoestablishan initialpointforcomparison offuturetestresults.The activatedcarbonadsorber sectionshouldbereplaced withnewunusedactivated carbonmeetingthephysical propertyspecificationsof Table1if(1)testingin accordancewiththefrequencyRev.0 WOLFCREEKTABLE9.4-3(Sheet14)RegulatoryGuide1.140PositionWCGSspecifiedinFootnotec ofTable2resultsina representativesample failingtopasstheappli-cabletestinTable2or (2)norepresentative sampleisavailablefor
testing.*Allstatementsapplytoall exhaustsystemslistedabove, unlessotherwiseindicated.Rev.0 WOLF CREEK TABLE 9.4-4 DESIGN DATA FOR CONTROL BUILDING HVAC SYSTEM COMPONENTS I. Control Building Supply System A. Supply Air Unit Quantity 1 Air flow, cfm 15,000 Static pressure, in. w.g. 5.39
Motor horsepower, hp 25 Total cooling capacity, Btu/hr 1,114,000 Total heating capacity, Btu/hr 1,479,000 Chilled water flow, gpm 131 Hot water flow, gpm 73 Design codes and standards
Unit MS Motor NEMA Coil MS
Seismic design Non-Category I B. Control Room Electric Duct Heater
Quantity 1 Heater rating, kW 35
Design standards MS Seismic design Non-Category I C. Access Control Supply System Booster Coil Quantity 1 Heater rating, kW 7 Design standards MS Seismic design Non-Category I D. Access Control Supply System Booster Coil Quantity 1 Heater rating, kW 8 Design standards MS
Seismic design Non-Category I II. Control Building Exhaust System Quantity 2 Type Vaneaxial
Air flow, cfm each 8,900 Total pressure, in. w.g. 4.09 Rev. 19 WOLF CREEK TABLE 9.4-4 (Sheet 2) Motor horsepower, hp 10 Design codes and standards Fan MS Motor NEMA Seismic design Non-Category I III. Access Control Exhaust System A. Adsorber Train Quantity 1 Particulate filters 4
HEPA filters 8 Charcoal, lbs 1,070 (approx.) Design criterion (unit) Reg. Guide 1.140
Seismic design Non-Category I B. Fans Quantity 2 Type Centrifugal
Air flow, cfm each 6,000 Static pressure, in. w.g. 9.40 Motor horsepower, hp 15
Design codes and standards Fan MS Motor NEMA
Seismic design Non-Category I IV. Control Room Air-Conditioning System A. Control Room Air-Conditioning Unit Quantity 2 Flow, cfm each 20,400 Static pressure, in. w.g. 4.00
Motor horsepower, hp 40 Total cooling capacity, Btu/hr each 549,720 Compressor power input, kW 48.6
Condenser water flow, gpm each 85 Fouling factor (service water) .002
Design codes and standards
Unit IEEE-323 and 344 Condenser ASME Section III, Class 3 (water side), ASME Section VIII, Div. 1, refrigerant side) Seismic design Category I B. Control Room Filtration System Adsorber Train Quantity 2 Particulate filters, each 2 HEPA filters, each 4 Rev. 18 WOLF CREEK TABLE 9.4-4 (Sheet 3) Charcoal, lbs. each 270 (approx.) Design criterion Reg. Guide 1.52
Seismic design Category I C. Control Room Filtration Fan
Quantity 2 Air flow, cfm each 2,000
Static pressure, in. w.g. 6.80
Motor horsepower, hp 5.0
Design codes and standards
Fan MS
Motor IEEE-323
Seismic design Category I D. Control Room Pressurization System Adsorber Train
Quantity 2 Demisters, each 1
HEPA filters, each 2
Charcoal, lbs. each 135 (approx.)
Electric heater, quantity each 1
Heater rating, kW each 5 Design codes and standards
Unit Reg. Guide 1.52
Heater IEEE-323
Seismic design Category I E. Control Room Pressurization System Fan
Quantity 2 Air flow, cfm. each 2,200
Static pressure, in. w.g. 8.5
Motor horsepower, hp 7.5
Design codes and standards
Fan MS
Motor IEEE-323
Seismic design Category I V. Class IE Electrical Equipment Air-Conditioning System
Quantity 2 Flow, cfm each 11,500
Static pressure, in. w.g. 3.5
Motor horsepower, hp 15.0
Total cooling capacity, Btu/hr each 398,400
Compressor power input, kW 33.7 Rev. 11 WOLF CREEK TABLE 9.4-4 (Sheet 4) Condenser water flow, gpm each 73 Fouling factor (service water) .002 Design codes and standards
Unit IEEE-323 and 344
Condenser ASME Section III, Class 3 (water side), ASME Section VIII, Div. 1, (refrigerant side) Seismic design Category I VI. Access Control Air-Conditioning System A. Access Control Fan Coil Unit
Quantity 1 Air flow, cfm 5,000
Static pressure, in. w.g. 4.93
Motor horsepower, hp 7.50
Total cooling capacity, Btu/hr 150,000
Fouling factor (chilled water) 0.0005
Chilled water flow, gpm 18
Design codes and standards
Unit MS
Motor NEMA
Coil MS
Seismic design Non-Category I B. Access Control Air-Conditioning System Booster Coil Quantity 1 Heating rating, kW 16
Design standards MS
Seismic design Non-Category I C. Mechanical Equipment Room Booster Coil Quantity 2 Heater Rating, kW each 6
Design standards MS
Seismic design Non-Category I VII. Unit Heaters
A. Upper Cable Spreading Room Unit Heater Quantity 2 Type Electric
Heater rating, kW each 40
Design standards MS
Seismic design Non-Category I Rev. 18 WOLF CREEK TABLE 9.4-4 (Sheet 5) B. Lower Cable Spreading Room Unit Heater Quantity 2 Type Electric Heater rating, kW each 10
Design standards MS Seismic design Non-Category I C. ESF Switchgear Room Unit Heater Quantity 2 Type Electric
Heater rating, kW each 7.5 Design standards MS Seismic design Non-Category I D. Pipe Chase and Tank Area Unit Heater Quantity 3 Type Electric Heater rating, kW each 5
Design standards MS Seismic design Non-Category I E. Control Room Air-Conditioning Equipment Room Unit Heater Quantity 1 Type Electric Heater rating, kW 25 Design standards MS
Seismic design Non-Category I F. Control Room Air-Conditioning Equipment Room Unit Heater Quantity 1 Type Electric Heater rating, kW 15
Design standards MS Seismic design Non-Category I VIII. Counting Room Recirculation System A. Counting Room Fan Coil Unit
Quantity 1 Air flow, cfm 3,500 Static pressure, in. w.g. 5.52 Rev. 20 WOLF CREEK TABLE 9.4-4 (Sheet 6) Motor horsepower, hp 7.5 Total cooling capacity, Btu/hr 90,900 Chilled water flow, gpm 12.12 Design codes and standards Unit MS / UL Motor MS / NEMA Coil MS / ARI Seismic Design Non-Category I B. Counting Room Filter Unit Quantity 1 Particulate filters 1 HEPA filters 1
Design codes and standards (unit) MS / Ul Seismic design Non-Category I IX. SAS Room Cooling System A. SAS Room Air Handling Unit Quantity 1 Air Flow, cfm 2,200
Static pressure, in.w.g 1.5 (ESP) Motor horsepower, hp 2.0 Total cooling capacity, Btu/hr 59,400
Design Codes and Standards Unit MS Motor NEMA
Coil MS Seismic Design Non-Category I B. SAS Room Condensing Unit Quantity 1 Condenser fan airflow, cfm 5,000 Condenser fan motor, hp 1/2
Compressor motor, hp 11-1/2 Condenser capacity, Btu/hr 67,440 Design codes and standards
Condenser MS Compressor MS Rev. 20 WOLF CREEK TABLE 9.4-5 SINGLE FAILURE ANALYSES CONTROL BUILDING HVACSYSTEMS Component Malfunction ConsequencesI. Control Room Air-ConditioningSystemParticulate filtersExcessive dust load-Systemissizeding,reduced airflow for full flow with fully loaded filters. Air-conditioning Casing failure, Partial loss of unit casing airbypasses coil cooling;redundant (evaporator) unitis available for cooling. Fan Fails to start Loss of one unit; redundant fan is available. Compressor Fails to operate Loss of one (refrigerant)system;redundantsystemis available for cooling.Condenser Tuberupture Loss of condenser;redundantsystemis available. Piping (refrigerant) Rupture, loss of Loss of one system;refrigerantredundantsystemis available.
System Isolation Damper fails to open; Loss of one system; Dampers flow path not avail-redundantsystem able.is available.II. Control Room FiltrationSystemParticulate filtersExcess dust loading; Reduced cleanup airflowisreduced capabilities;redundantsystemis available for cleanup. Fan Fails to start One unitis out of service;redundant unit is available for cleanup. Rev. 0 WOLF CREEK TABLE 9.4-5 (Sheet 2) Component Malfunction Consequences System isolation Damper fails to open; Loss of one system; damper flow path not avail- redundant system is
able. available. III. Control Room Pressurization System
Particulate filters Excessive dust load- System is sized for ing, reduced airflow full flow with
fully loaded filters Fan Fails to start One unit is out of service; redundant unit is available for operation. System isolation Damper fails to open; Loss of one system; dampers flow path not avail- redundant system is
able. available. IV. Class IE Electrical Equipment Air-Conditioning System Particulate filters Excessive dust load- System is sized for ing, reduced airflow full flow with fully
loaded filters. Air-conditioning Casing failure, air Partial loss of unit casing bypasses coil cooling; redundant (evaporator) unit is available for cooling. Fan Fails to start Loss of one unit; redundant fan is available. Compressor Fails to operate Loss of one system; (refrigerant) redundant system is available. Piping (refrigerant) Rupture, loss of Loss of one system; refrigerant redundant system is available. Rev. 0 WOLF CREEK TABLE 9.4-5 (Sheet 3) Component Malfunction Consequences V. Control building supply system, control building exhaust system, and access control exhaust system penetrations of the common
auxiliary/control building boundary Building isolation Damper fails to close Loss of isolation dampers one one side of penetration; redun-dant damper closes. Rev. 0 WOLF CREEK TABLE 9.4-6 DESIGN DATA FOR FUEL BUILDING HVAC SYSTEM COMPONENTS I. Fuel Building Supply Air System A. Supply Air Heating Coil Quantity 1 Total heating capacity, Btu/hr 2,196,800
Hot water flow, gpm 110 Design standards MS Seismic design Non-Category I B. Supply Air Unit Quantity 2 Air flow, cfm each 18,000 Static pressure, in. w.g. each 6.4 Motor, hp each 25
Total cooling capacity, Btu/hr each 477,900
Chilled water flow rate, gpm 64 Design codes and standards Unit MS
Coil MS Motor NEMA Seismic design Non-Category I C. Fuel Handling Area Chilled Water Coil Quantity 1 Total cooling capacity Btu/hr 175,500 Chilled water flow, gpm 28 Design standards MS
Seismic design Non-Category I II. Fuel Storage Pool Pump Room Coolers Quantity 2 Air flow, cfm each 3,500 Static pressure, in. w.g. each 0.92
Motor horsepower, bhp each 2 Total cooling capacity Btu/hr each 72,000 Water flow rate, gpm each 29 Fouling factor 0.002 Tube material 90/10 Cu-Ni &/or ASME SB- 676/UNS NO8367 Design codes and standards Unit MS
Coil ASME Section III, Class 3 Motor IEEE-323
Seismic design Category I Rev. 15 WOLF CREEK TABLE 9.4-6 (Sheet 2) III. Unit Heaters A. Lower Level Unit Heater (East Wall) Quantity 1 Heating capacity, Btu/hr 168,900 Hot water flow rate, gpm 9 Design standards
Motor NEMA Coil MS Seismic design Non-Category I B. Lower Level Unit Heater (West Wall) Quantity 1 Heating capacity, Btu/hr 168,900 Hot water flow rate, gpm 9 Design standards
Motor NEMA
Coil MS Seismic design Non-Category I C. Lower Level Unit Heater (South Wall) Quantity 1 Heating capacity, Btu/hr each 126,700
Hot water flow rate, gpm each 7 Design standards Motor NEMA
Coil MS
Seismic design Non-Category I D. Upper Level Unit Heater D.1. Quantity 2 Heating capacity, Btu/hr each 260,000 Hot water flow rate, gpm each 13
Design standards Motor NEMA Coil MS
Seismic design Non-Category I D.2. Quantity 4 Rating, kw 30
Design Standard MS Seismic Design NON-Category I E. Stairway Unit Heater (Electric) Quantity 1 Rating, kW 15 Seismic Category Non-Category I Rev. 6 WOLF CREEK TABLE 9.4-6 (Sheet 3) IV. Emergency Exhaust System A. Adsorber Train Quantity 2 Particulate filters, each 6 HEPA filters, each 12 Charcoal, lbs each 1,937 (approx.)
Design criterion (unit) Reg. Guide 1.52
Seismic design Category I B. Electric Heaters Quantity 2 Rating, kW each 37
Design standard IEEE-323 Seismic design Category I C. Fans Quantity 2 Type Centrifugal Air flow, cfm each 9,000* Static pressure, in. w.g. each 11.75
Motor brake horsepower, bhp each 25 Design codes and standards Fan MS Motor IEEE-323
- Technical Specifications currently require an emergency exhaust flowrate of 6500 cfm + 650. At 6500 cfm and a fan speed of 2250 rpm, the static pressure of each fan is 10.0 in w.g. Rev. 6 WOLFCREEKTABLE9.4-7SINGLE-FAILUREANALYSIS-EMERGENCYEXHAUSTSYSTEM,FUELSTORAGEPOOLPUMPROOMCOOLERS,ANDFUELBUILDINGHVACISOLATIONComponentMalfunction ConsequencesExhaustsystemfilter/Excessdustloading;Systemissizedforfull-adsorberunitparticu-airflowisreducedflowwithfullyloadedlatefiltersfilters.ExhaustfanFailstostartOneunitisoutofservice;redundantunitiscapableofproviding
cleanup.Exhaustsystemdis-Failsclosed;flowOneunitisoutofser-chargedamperpathisnotavail-vice;redundantunitisable.capableofproviding cleanup.Auxiliary/fuelbuild-DamperfailstoOnedamperisoutof ingnormalexhaustclose;flowpathisservice;redundantdam-systemisolationnotisolatedperiscapableofpro-dampersvidingrequiredclosure.FuelbuildingsupplyDamperfailstoOnedamperisoutofsystemductworkiso-close;flowpathisservice;redundant lationdampersnotisolateddamperiscapableofprovidingrequiredclosure.PumproomcoolerFailstostartOneunitisoutofser-fanvice;redundantunitiscapableofprovidingcoolingrequirements.FuelBuildingDamperfailstoOnedamperisoutofemergencyexhaustopenfollowingservice;redundant systemisolationFBVISsignaldamperiscapabledampersofprovidingrequired opening.DamperfailstoOnedamperisoutofachieveitsclosedservice;redundant positionfollowingdamperiscapable anSISsignalofprovidingrequired(i.e.damperisopening.openedtoomuch)Rev.14 WOLF CREEK TABLE 9.4-8 DESIGN DATA FOR AUXILIARY BUILDING HVAC SYSTEM COMPONENTS I. Auxiliary Building Supply Air Unit Quantity 1
Air flow, cfm 10,800
Static pressure, in. w.g. 4.84 Motor horsepower, hp 15 Total cooling capacity, Btu/hr 802,000
Total heating capacity, Btu/hr 1,167,000
Chilled water flow, gpm 95
Hot water flow, gpm 60 Design codes and standards Unit MS
Motor NEMA Codes MS
Seismic design Non-Category I II. Main Steam Tunnel Supply Air Unit
Quantity 1
Air flow, cfm 20,000 Static pressure, in. w.g. 5.20 Motor horsepower, hp 40 Heating capacity, Btu/hr 2,186,000
Hot water flow rate, gpm 109
Cooling capacity, Btu/hr Sensible 807,300 Total 1,440,300 Chilled water flow, gpm 106 Design codes and standards Unit MS
Motor NEMA Codes MS Seismic design Non-Category I
III. Auxiliary/Fuel Building Normal Exhaust System
A. Adsorber Train Quantity 1
Particulate filters, each 25
HEPA filters, each 50
Charcoal, lbs each 6,300 Design criterion Reg. Guide 1.140 Seismic design Non-Category I
Rev. 25
WOLF CREEK TABLE 9.4-8 (Sheet 2) B. Fans Quantity 2 Type Centrifugal Air flow, cfm each 32,000 Static pressure, in. w. g. 13.72 Motor horsepower, hp each 100 Design codes and standards Fan MS Motor NEMA
Seismic design Non-Category I C. Decontamination Tanks Exhaust Scrubber Quantity 2 Flow, cfm each 3,200 Static pressure, in. w.g. 7.00 Motor horsepower, hp 7.5
Water flow rate gpm each 0.50 Contaminant removal efficiency, % 90 Housing material Stainless steel Design codes and standards
Fan MS Motor NEMA Seismic design Non-Category I D. Scrubber Makeup Air Duct Heater Quantity 2 Heater rating, kW each 90 Design standards MS Seismic design Non-Category I E. Auxiliary Building Fume Hood Booster Fan Quantity 1 Type Vane axial Air flow, CFM 1180 Static pressure, in. w.g. .25
Motor horsepower, hp .33
Seismic design Non-Category I Rev. 8 WOLF CREEK TABLE 9.4-8 (Sheet 3) IV. Main Steam Tunnel Exhaust System Quantity 2 Air flow, cfm each 23,000 Static pressure, in. w.g. 4 Motor horsepower, hp each 25 Design codes and standards Fan MS Motor NEMA Seismic design Non-Category I V. Nonessential Recirculation Units A. Electrical Equipment Room Fan Coil Units
- 1. Unit SGL02 Quantity 1 Air flow, cfm 3,100
Static pressure in. w.g. 3.52
Motor horsepower, hp 5 Total cooling capacity, Btu/hr 164,000 Chilled water flow rate, gpm 20
Design codes and standards Unit MS Motor NEMA Coil MS Seismic design Non-Category I
- 2. Unit SGL20 Quantity 1 Air flow, cfm 13020 Total Static pressure, in.w.g. 3.35
Motor horsepower, hp 15 Total cooling capacity, BTU/HR 381,000 Chilled water flow rate, gpm 40 Design codes and standards Unit MS
Motor NEMA
Coil MS Seismic design Non-Category I Rev. 8 WOLF CREEK TABLE 9.4-8 (Sheet 4) B. Component Cooling Water Pump Room Fan Coil Unit Quantity 2 Air flow, cfm 4,000 Static pressure, in. w.g. 3.22 Motor horsepower, hp 5 Total cooling capacity, Btu/hr 300,000 Chilled water flow rate, gpm 35 Design codes and standards Unit MS
Motor NEMA Coil MS Seismic design Non-Category I C. Ground Floor Corridor Fan Coil Unit Quantity 1 Flow, cfm 2,600
Static pressure, in. w.g. 3.27 Motor horsepower, hp 3 Total cooling capacity, Btu/hr 137,400 Chilled water flow rate, gpm 16 Design codes and standards Unit MS Motor NEMA Coil MS Seismic design Non-Category I D. Hot Machine Shop and Hot Instrument Shop Fan Coil Unit Quantity 1 Flow, cfm 1,100
Static pressure, in. w.g. 2.30 Motor horsepower, hp 1.5 Total cooling capacity, Btu/hr 58,000 Chilled water flow rate, gpm 7 Design codes and standards
Unit MS
Motor NEMA Coil MS Seismic design Non-Category I Rev. 8 WOLF CREEK TABLE 9.4-8 (Sheet 5) E. Normal Charging Pump Room Fan Coil Unit Quantity l Air flow, cfm 5,500 Static pressure, in. w.g. 2.5 Motor horsepower, hp 5 Total cooling capacity, Btu/hr 250,000 Chilled water flow rate, gpm 24 Design codes and standards Unit MS Motor NEMA Coil MS
Seismic design Non-Category I F. Basement Corridor Fan Coil Unit Quantity 1 Flow, cfm 900
Static pressure, in. w.g. 2.30 Motor horsepower, hp 1.5 Total cooling capacity, Btu/hr 49,000
Chilled water flow rate, gpm 6
Design codes and standards Unit MS Motor NEMA Coil MS
Seismic design Non-Category I VI. Safety-Related Recirculation Units A. Safety-Injection Pump Room Cooler Quantity 2 Air Flow, cfm each 10,100 Static pressure, in. w.g. 1.30 Motor horsepower, hp each 10
Total cooling capacity, Btu/hr each 220,000
Water flow rate, gpm each 88 Fouling factor 0.002 Tube material 90/10 copper-nickel
&/or ASME SB-676/UNS NO8367 Rev. 24 WOLF CREEK TABLE 9.4-8 (Sheet 6)
Design codes and standards Unit MS Motor IEEE-323
Coil ASME Section III, Class 3
Seismic design Category I B. RHR Pump Room Cooler
Quantity 2
Air Flow, cfm each 10,100 Static pressure, in. w.g. 1.30 Motor horsepower, hp each 10
Total cooling capacity, Btu/hr each 220,000
Water flow rate, gpm each 88 Fouling factor 0.002 Tube material 90/10 copper-nickel
&/or ASME SB-676/UNS
NO8367 Design codes and standards Unit MS Motor IEEE-323
Coil ASME Section III, Class 3
Seismic Design Category I C. Component Cooling Water Pump Room Cooler
Quantity 2
Air Flow, cfm each 14,679 Static pressure, in. w.g. 1.70
Motor horsepower, hp each 15
Total cooling capacity, Btu/hr each 320,000
Water flow rate, gpm each 128 Fouling factor 0.002 Tube material AL-6XN SB-676 N-438-2
Rev. 29
WOLF CREEK TABLE 9.4-8 (Sheet 7) Design codes and standards Unit MS
Motor IEEE-323
Coil ASME Section III, Class 3 Seismic design Category I
D. Containment Spray Pump Room Cooler
Quantity 2 Air Flow, cfm each 10,000 Static pressure, in. w.g. 1.30
Motor horsepower, hp each 10
Total cooling capacity, Btu/hr each 220,000 Water flow rate, gpm each 88 Fouling factor 0.002
Tube material 90/10 copper-nickel
&/or ASME SB-676/UNS
NO8367 Design codes and standards
Unit MS
Motor IEEE-323
Coil ASME Section III, Class 3 Seismic design Category I
E. Auxiliary Feedwater Pump Room Cooler
Quantity 2 Air Flow, cfm each 14,679
Static pressure, in. w.g. 1.30
Motor brake horsepower, hp each 10
Total cooling capacity, Btu/hr each 320,000 Water flow rate, gpm each 128 Fouling factor 0.002
Tube Material AL-6XN SB-676 N-438-2 Design codes and standards Unit MS Motor IEEE-323
Coil ASME Section III, Class 3
Seismic design Category I
Rev. 29
WOLF CREEK TABLE 9.4-8 (Sheet 8) F. Penetration Room Cooler Quantity 2
Air Flow, cfm each 18,300
Static pressure, in. w.g. 2.16
Motor capability, hp each 30 Total cooling capacity, Btu/hr each 170,000
Water flow rate, gpm each 100
Fouling factor 0.002
Tube material 90/10 copper-nickel &/or ASME SB-676/UNS NO8367 Design codes and standards
Unit MS
Motor IEEE-323 Coil ASME Section III, Class 3
Seismic design Category I
G. Charging Pump Room Cooler Quantity 2
Air Flow, cfm each 14,679 Static pressure, in. w.g. 1.30
Motor horsepower, hp each 10.0 Total cooling capacity, Btu/hr each 320,000
Water flow rate, gpm each 128
Fouling factor 0.002 Tube material AL-6XN SB-676 N-438-2 Design codes and standards Unit MS
Motor IEEE-323
Coil ASME Section III, Class 3 Seismic design Category I
VII. Unit Heaters
A. Auxiliary Building Basement Corridor Unit Heater
Rev. 29
WOLF CREEK TABLE 9.4-8 (Sheet 9) Quantity 2 Heating capacity, Btu/hr each 60,000 Hot water flow, gpm each 3 Design standards Motor NEMA Coil MS Seismic design Non-Category I B. Auxiliary Building Hot Instrument Shop Unit Heater Quantity 1 Heating capacity, Btu/hr 80,000
Hot water flow, gpm 4 Design standards
Motor NEMA Coil MS Seismic design Non-Category I C. Auxiliary Building Decontamination Room Unit Heater Quantity 1 Heating capacity, Btu/hr 100,000 Hot water flow, gpm 5 Design standards Motor NEMA Coil MS
Seismic design Non-Category I D. Auxiliary Building Hot Machine Shop Unit Heater Quantity 2 Heating capacity, Btu/hr each 80,000 Hot water flow, gpm each 4 Design standards Motor NEMA
Coil MS
Seismic design Non-Category I Rev. 8 WOLF CREEK TABLE 9.4-8 (Sheet 10) E. Auxiliary Building Corridor Unit Heater Quantity 2 Heating capacity, Btu/hr each 80,000 Hot water flow, gpm each 4 Design standards Motor NEMA Coil MS Seismic design Non-Category I F. Auxiliary Building HVAC Equipment Room Unit Heater Quantity 3 Heating capacity, Btu/hr each 100,000
Hot water flow, gpm each 5 Design standards Motor NEMA
Coil MS Seismic design Non-Category I G. Auxiliary Building Containment Personnel Access Area Unit Heater Quantity 2 Heating capacity, Btu/hr each 60,000 Hot water flow, gpm each 3 Design standards Motor NEMA Coil MS
Seismic design Non-Category I H. Auxiliary Building Boric Acid Storage Tank Area Unit Heater Quantity 2 Type Electric Heating rating, kW each 15 Design standards MS Seismic design Non-Category I Rev. 8 WOLF CREEK TABLE 9.4-8 (Sheet 11) I. Pipe Chase Unit Heater Quantity 2 Type Electric Heater rating, kW each 5 Design standards MS Seismic design Non-Category I J. Pipe Chase Unit Heater Quantity 4 Type Electric Heater rating, kW each 10
Design standards MS Seismic design Non-Category I VIII. Access Tunnel Transfer Fan Quantity 1 Air flow, cfm 250 Static pressure, in w.g. 0.82 Motor horsepower, hp 0.25 Design codes and standards Fan MS
Motor NEMA
Seismic design Non-Category I IX. I&C Hot Shop Air-Conditioning Unit Quantity 1 Type Room Air-Conditioner Air Flow, cfm 555-675 Total Cooling Capcity, Btu/hr 30,500 Power Input, KW 4.3-4.7
Design Standards ARI AMCA MS Seismic Design Non-Category I Rev. 13 WOLF CREEK TABLE 9.4-9 SINGLE-FAILURE ANALYSES - PUMP ROOM COOLERS, PENETRATION ROOM COOLERS AND EMERGENCY EXHAUST SYSTEM* Component Malfunction Consequences Fan Fails to start One unit is out of service; redundant unit is
capable of providing cooling requirements. Discharge damper** Fails closed, One unit is out of service; airflow is lost redundant unit is capable of providing cooling
requirements. Emergency exhaust Damper fails to One system is out of system isolation open; exhaust service; redundant damper airflow is lost system is capable of providing required exhaust flow. Isolation dampers Damper fails to Loss of isolation of one for those non- close side of penetration; safety-related HVAC redundant damper closes. systems which pene-trate the auxiliary building boundary
- Applies only to that portion of the system located within the auxiliary building.
- Applies only to penetration room coolers and one component cooling water pump room cooler.
Rev. 0 WOLF CREEK TABLE 9.4-10 DESIGN DATA FOR TURBINE BUILDING VENTILATION SYSTEM COMPONENTS I. Main Building Supply System A. Operating Floor
- 1. Supply Units
Quantity 2 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.4
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA
- 2. Supply Units
Quantity 2 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.53
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA
- 3. Supply Unit
Quantity 1 Air flow, cfm 24,000
Static pressure, in. w.g. 1.96
Motor horsepower, hp each 20
Design standards
Fan MS
Motor NEMA B. Mezzanine Floor
- 1. Supply Units
Quantity 5 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.4
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA Rev. 0 WOLF CREEK TABLE 9.4-10 (Sheet 2)
- 2. Supply Units Quantity 2 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.53
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA
- 3. Supply Unit
Quantity 1 Air flow, cfm 24,000
Static pressure, in. w.g. 1.96
Motor horsepower, hp each 20
Design standards
Fan MS
Motor NEMA C. Ground Floor
- 1. Supply Units
Quantity 3 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.4
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA
- 2. Supply Units
Quantity 2 Air flow, cfm each 45,000
Static pressure, in. w.g. 1.53
Motor horsepower, hp each 40
Design standards
Fan MS
Motor NEMA
- 3. Supply Units
Quantity 1 Air flow, cfm 24,000 Rev. 0 WOLF CREEK TABLE 9.4-10 (Sheet 3) Static pressure, in. w.g. 1.4 Motor horsepower, hp 20
Design standards
Fan MS
Motor NEMA D. Communication Corridor
Quantity 1 Air flow, cfm 24,000
Static pressure, in. w.g. 1.86
Motor horsepower, hp 15 Design standards Fan MS
Motor NEMA II. Main Building Exhaust System A. Exhaust Fans (large)
Quantity 8 Air flow, cfm each 90,000
Static pressure, in. w.g. 1.74
Motor horsepower, hp each 50 B. Exhaust Fans (small) Quantity 2 Air flow, cfm each 40,000
Static pressure, in. w.g. 1.08 Motor horsepower, hp each 10 Design standards
Fan MS
Motor NEMA C. Toilet Exhaust Fans Quantity 2 Air flow, cfm each 500
Static pressure, in. w.g. 0.73
Motor horsepower, hp each 0.25
Design standards
Fan MS
Motor NEMA D. Elevator Machine Room Exhaust Fan
Quantity 1 Air flow, cfm 500 Rev. 0 WOLF CREEK TABLE 9.4-10 (Sheet 4) Static pressure, in. w.g. 0.27 Motor horsepower, hp 1/3
Design standards
Fan MS
Motor NEMA III. Turbine Building Unit Heaters
A. Operating Floor Quantity 4 Heating capacity, Btu/hr each 330,000 Hot water flow rate, gpm each 17 Design standards
Coil MS
Motor NEMA B. Mezzanine Floor
Quantity 2 Heating capacity, Btu/hr each 220,000
Hot water flow rate, gpm each 11
Design standards
Coil MS
Motor NEMA C. Ground Floor Quantity 4 Heating capacity, Btu/hr each 220,000 Hot water flow rate, gpm each 11 Design standards
Coil MS
Motor NEMA D. Stairwells Quantity 8 Type Electric
Heating capacity, kW each 7.5
Design standards
Coil MS
Motor Manufacturer's
Standard E. Communication Corridor
- 1. Elevation 2073'-6" Quantity 2 Heating capacity, Btu/hr each 120,000 Rev. 0 WOLF CREEK TABLE 9.4-10 (Sheet 5)
Hot water flow rate, gpm each 6 Design standards
Coil MS
Motor NEMA
- 2. Elevation 2,032
Quantity 2 Heating capacity, Btu/hr each 20,000
Hot water flow rate, gpm each 1
Design standards
Coil MS
Motor NEMA
- 3. Elevation 2,000
Quantity 1 Heating capacity, Btu/hr each 60,000
Hot water flow rate, gpm 3
Design standards
Coil MS
Motor NEMA IV. Lube Oil Room Ventilation and Heating System
A. Supply System
Quantity 1 Air flow, cfm 2,000
Static pressure, in. w.g. 1.69
Motor horsepower, hp 1.5
Heating capacity, Btu/hr 220,000
Hot water flow, gpm 11
Design standards
Fan MS
Motor NEMA B. Exhaust System
Quantity 1 Air flow, cfm 2,000
Static pressure, in. w.g. 1.01
Motor horsepower, hp 1
Design standards
Fan MS
Motor NEMA Rev. 0 WOLF CREEK TABLE 9.4-10 (Sheet 6) V. Computer Room HVAC System Quantity 2 Air flow, cfm each 6,000 Static pressure, in. w.g. 4.08 Motor horsepower, hp each 20 Total cooling capacity, Btu/hr each (total) 350,500 Chilled water flow rate, gpm each 15 Heating capacity, Btu/hr each 285,000 Hot water flow rate, gpm each 15 Design standards Fan MS Motor NEMA Coil MS VI. Instrument Shop HVAC System A. Recirculation Unit Quantity 1 Air flow, cfm 1,250 Static pressure, in. w.g. 1.90 Motor horsepower, hp each 1.5 Total cooling capacity, Btu/hr (total) 41,000 Chilled water flow rate, gpm 5 Design standards Fan MS Motor NEMA Coil MS B. Duct Heater Quantity 1 Type Electric Air flow, cfm 100 Heating capacity, kW 2 Design standard MS VII. Condenser Air Removal Filtration System Quantity 2 Air flow, cfm each 1,000 Static pressure, in. w.g. 10.0 Motor horsepower, hp each 3.0 Prefilters, No. 1 HEPA filters, No. 2 Rev. 18 WOLF CREEK TABLE 9.4-10 (Sheet 7)
Charcoal, lbs 150 Design criterion Reg. Guide 1.140
Seismic design Non-Category I
VIII. Battery Room Chiller Water Coils Quantity 2
Type Chilled water
Air flow, cfm each 720
Total cooling capacity Btu/hr each 58,000 Water flow, gpm each 7
Design standard MS
IX. EHC Cabinet Room Air-Conditioning System Quantity 2
Type Direct Expansion (DX) Air flow, cfm each 3,200 Total cooling capacity, Btu/hr each 106,430 Static pressure, in w.g. 0.540 Motor horsepower, hp each
3.6 Design
standards Fan MS Motor NEMA X. Oxygen control and pH control chemical Storage Room
Air-Conditioning System
Quantity 1 Type Room air conditioners
Air flow, cfm 675
Total cooling capacity, Btu/hr 32,500 (min)
Power input, kW 4.3 (max) Design standards MS
XI. Turbine Deck Office Mezzanine Room Air-Conditioning System
Quantity 3 Type Room air Conditioners
Air flow, cfm each 165
Total cooling capacity, Btu/hr each 8,250
Power input, kW 0.83
Design standards MS
Rev. 27 WOLF CREEK TABLE 9.4-11 DESIGN DATA FOR RADWASTE BUILDING HVAC SYSTEM COMPONENTS I. The supply air system consists of a particulate filter, hot water heating coil, chilled water cooling coil, centrifugal fan, and electric motor driver. Quantity 1 Air flow, cfm 10,800
Static pressure, in. w.g. 2.47
Motor horsepower 7.5
Cooling capacity, Btu/hr (total) 345,100
Heating capacity, Btu/hr 1,108,00
Chilled water flow rate, gpm 41
Hot water flow rate, gpm 55
Design standards
Fan MS
Motor NEMA
Coil MS II. Unit heaters are used or provide supplemental heating.
A. Basement Floor
Quantity 1 Heating capacity, Btu/hr 140,000
Hot water flow rate, gpm 7
Type Hot water
Design standards
Coil MS
Motor NEMA B. Ground Floor
Quantity 2 Heating capacity, Btu/hr each 78,600
Hot water flow rate, gpm each 4
Type Hot water
Design standards
Coil MS
Motor NEMA C. El. 2,021
Quantity 2 Heating capacity, Btu/hr each 140,000
Hot water flow rate, gpm each 7
Type Hot water Rev. 0 WOLF CREEK TABLE 9.4-11 (Sheet 2) Design standards Coil MS
Motor NEMA D. Second Floor
Quantity 4 Heating capacity, Btu/hr each 195,000
Hot water flow rate, gpm each 10
Type Hot water
Quantity 1
Heating capacity, Btu/hr 100,000
Hot water flow rate, gpm 5
Type Hot water
Design standards
Coil MS
Motor NEMA E. Drumming Area Unit Heaters
Quantity 3 Type Electric
Heater rating, kW each 40
Design standards MS III. Local fan-coil units are used to provide supplemental cooling on the basement and ground floors. Each unit consists of a chilled water cooling coil, centrifugal
fan, and electric motor driver. Quantity 2 Air flow, cfm each 1,700
Static pressure, in. w.g. 3.0
Motor horsepower, each 2
Total cooling cap. Btu/hr each 90,000
Chilled water flow rate, gpm each 11
Design standards
Fan MS
Motor NEMA
Coil MS IV. Fan-coil units are provided to supplement the cooling of the evaporator rooms. Each unit consists of a chilled
water cooling coil, centrifugal fan, and electric motor
driver. A. Recycle Evaporator Room
Quantity 1 Air flow, cfm 3,200
Static pressure, in. w.g. 2.2 Rev. 0 WOLF CREEK TABLE 9.4-11 (Sheet 3) Motor horsepower 3 Total cooling cap., Btu/hr 169,000
Chilled water flow rate, gpm 20
Design standards
Fan MS
Motor NEMA
Coil MS B. Waste Evaporator Room
Quantity 1 Air flow, cfm 3,200
Static pressure, in. w.g. 2.2
Motor horsepower 3
Total cooling cap., Btu/hr 169,000
Chilled water flow rate, gpm 20
Design standards
Fan MS
Motor NEMA
Coil MS C. SLWS Evaporator Room
Quantity 1 Air flow, cfm 3,200
Static pressure, in. w.g. 2.2
Motor horsepower 3
Total cooling cap., Btu/hr 169,000
Chilled water flow rate, gpm 20
Design standards
Fan MS
Motor NEMA
Coil MS V. Fan-coil units are used to provide the cooling of the control rooms and the sample laboratory. Each unit
consists of a chilled water cooling coil, centrifugal
fan, and electric motor driver. In addition, the main
control room fan-coil unit is provided with a heating
coil. A. Control Room (Solidification)
Quantity 1 Air flow, cfm 500
Static pressure, in. w.g. 1.3
Motor horsepower 0.75
Total cooling, cap., Btu/hr 15,100
Chilled water flow rate, gpm 2.0 Rev. 0 WOLF CREEK TABLE 9.4-11 (Sheet 4) Design standards Fan MS
Motor NEMA
Coil MS B. Sample Laboratory
Quantity 1 Air flow, cfm 500
Static pressure, in. w.g. 1.3
Motor horsepower 0.75
Total cooling cap., Btu/hr 15,100
Chilled water flow, gpm 2.0
Design standards
Fan MS
Motor NEMA
Coil MS C. Control Room
Quantity 1 Air flow, cfm 1,700
Static pressure, in. w.g. 2.5
Motor horsepower 1.5
Total cooling cap., Btu/hr
(total) 42,100
Heating capacity, Btu/hr 40,400 Chilled water flow rate, gpm 5
Hot water flow rate, gpm 2.0
Design standards
Fan MS
Motor NEMA
Coil MS VI. Transfer Fan
A fan is provided to transfer air from the radwaste building to the personnel access tunnel. The unit
consists of a propeller fan and electric motor driver. Quantity 1 Air flow, cfm 250
Static pressure, in. w.g. 0.81
Motor horsepower 0.25
Design standards
Fan MS
Motor NEMA VII. The radwaste building exhaust system consists of partic-ulate filters and charcoal adsorption trains, centri-
fugal fans, and electric motor drivers. Rev. 0 WOLF CREEK TABLE 9.4-11 (Sheet 5) A. Adsorber Train Quantity 1 Particulate filters 9
HEPA filters 18
Charcoal, lbs 1800
Bed depth, in. 2
Type Gasketless
Design criterion Reg. Guide 1.140
Seismic design Non-Category I B. Fans
Quantity 2 Type Centrifugal
Air flow, cfm each 12,000
Static pressure, in. w.g. 9.16
Motor horsepower, each 25
Design standards
Fan MS
Motor NEMA Rev. 0 WOLF CREEK TABLE 9.4-12 DESIGN DATA FOR THE CONTAINMENT HVAC SYSTEM COMPONENTS I. Containment Shutdown Purge Supply System A. Supply Air Unit Quantity 1 Air flow, cfm 20,000 Static pressure, in. w.g. 4.94 Motor horsepower, hp 25
Total cooling capacity, Btu/hr 2,680,000 Hot water flow, gpm 106 Design codes and standards
Unit MS Motor NEMA Coils MS
Seismic Design Non-Category I B. Containment Isolation Valves
Quantity 4 Type Butterfly (wafer)
Material Carbon steel Actuation Air cylinder Failure mode Closed
Size, in. 36 Design codes ASME Section III, Class 2 Seismic design Category I C. Containment Penetration Size, in. 36 Material Carbon steel Design codes ASME Section III, Class E
Seismic design Category I II. Containment Minipurge Supply System
A. Supply Air Unit Quantity 1 Air flow, cfm 4,000 Static pressure, in. w.g. 4.25
Motor horsepower, hp 5 Total heating capacity, Btu/hr 324,000 Hot water flow, gpm 17
Design codes and standards Unit MS Motor NEMA
Coils MS Seismic design Non-Category I Rev. 0 WOLF CREEK TABLE 9.4-12 (Sheet 2) B. Containment Isolation Valves Quantity Isolation Valves Quantity 4 Type Butterfly (wafer) Material Carbon steel
Actuation Air cylinder Failure mode Closed Size, in. 18
Design codes ASME Section III, Class 2 Seismic design Category I III. Containment Purge Exhaust System A. Exhaust Fans
- 1. Containment Shutdown Purge Exhaust Fan Quantity 1 Air flow, cfm 20,000 Static pressure, in w.g. 13.3
Motor horsepower, hp 60 Design codes and standards Fan MS
Motor NEMA
- 2. Containment Minipurge Exhaust Fan
Quantity l Air flow, cfm 4,000
Static pressure, in. w.g. 5.0 Motor horsepower, hp 5 Design codes and standards
Fan MS Motor NEMA B. Filter Adsorber Unit Quantity 1 Particulate filters 20 HEPA filters 40 Charcoal, lbs 3,500
Bed depth, in. 2 Type Gasketless Design criterion Reg. Guide 1.140
Seismic design Non-Category I Rev. 0 WOLF CREEK TABLE 9.4-12 (Sheet 3) C. Containment Isolation Valves
- 1. Containment Shutdown Purge Quantity 2 Type Butterfly (wafer)
Material Carbon steel
Actuation Air cylinder Failure mode Closed Size in. 36
Design codes ASME Section III, Class 2 Seismic design Category I
- 2. Containment Minipurge Quantity 2 Type Butterfly (wafer)
Material Carbon steel Actuation Air cylinder
Failure mode Closed Size in. 18 Design codes ASME Section III, Class 2
Seismic design Category I D. Containment Penetration
Size, in. 36 Material Carbon steel
Design codes ASME Section III, Class 2 Seismic design Category I IV. Containment Atmospheric Control System Fans Quantity 2 Fan type Vaneaxial Arrangement 4 Air flow, cfm each 10,000
Static pressure, in w.g. 7.27 Motor horsepower, hp 20 Design codes and standards
Fan MS Motor NEMA V. Containment Atmospheric Control System Filter Units Particulate filters, quantity each 8 HEPA filters, quantity each 16 Adsorbers, lbs of charcoal each 1,900 Rev. 0 WOLF CREEK TABLE 9.4-12 (Sheet 4) Bed depth, in. 2 Type Gasketless Design criterion Reg. Guide 1.140 Seismic design Non-Category I VI. CRDM Cooling Fans Quantity 3 Type Vaneaxial Air flow, cfm each 25,000 Static pressure, in. w.g. 3.9 Motor horsepower, hp each 3 @ 30 hp Design codes and standards Fan MS Motor NEMA VII. Cavity Cooling Fans Quantity 2 Type Vaneaxial Air flow, cfm each 16,000 Static pressure, in. w.g. 9.0 Motor horsepower, hp each 50 Design codes and standards Fan MS Motor NEMA VIII. Machine Room Exhaust Fan Quantity 1 Type Centrifugal Air flow, cfm each 1,000 Static pressure, in. w.g. 0.1 Motor horsepower, hp each 0.5 Design codes and standards Fan MS Motor NEMA IX. Pressurizer Cooling Fan Quantity 1 Type Vaneaxial Air flow, cfm each 2,400 Static pressure, in. w.g. 0.92 Motor horsepower, hp each 0.75 Design codes and standards Fan MS Motor NEMA Rev. 19 WOLF CREEK TABLE 9.4-12 (Sheet 5) X. Ductwork Material - inside secondary shield wall Stainless steel Material - outside secondary shield wall Galvanized steel Rev. 0 WOLF CREEK TABLE 9.4-13 COMPARISON OF CONTAINMENT MINIPURGE CONTAINMENT ISOLATION VALVES WITH BTP CSB 6-4 BTP Item No. Minipurge isolation valve B.1.a Complies. B.1.b System employs two lines, one for the supply system and one for the exhaust system. B.1.c Minipurge containment isolation valve size is 18-inch-diameter butterfly valve. A minipurge flow rate of 4,000
cfm is required to maintain inplant containment doses, based on the assumptions and source terms of
Regulatory Guide 1.112, at 7 MPC during occupation (see Section 11.3). At a flow rate of 4,000 cfm, 8-inch-diameter valves and system result in prohibitive velocities and pressure drops. B.1.d Complies. B.1.e Complies.
B.1.f Complies. B.1.g Complies. B.2 Complies.
B.3 Complies. B.4 Complies. B.5.a Complies. (see Chapter 15.0)
B.5.b The purge system has no safety-related fans, filters or ductwork beyond the
isolation valves. B.5.c Complies. (see Section 6.2.1)
B.5.d. Complies. Rev. 0 WOLF CREEK TABLE 9.4-14 DESIGN DATA FOR THE DIESEL GENERATORS BUILDING VENTILATION SYSTEM
COMPONENTS I. Diesel Generators Building Ventilation Supply Fans Quantity 2 Type Vaneaxial
Airflow, cfm each 120,000
Static pressure, in. w.g. 2.25
Motor horsepower, hp 100
Design codes and standards
Fan AMCA
Motor IEEE-323
Seismic design Category I II. Diesel Generators Building Unit Heaters
A. Quantity 4 Heating capacity, kW each 30
Type Electric
Design standards NEC, UL
Seismic design Non-Category I B. Quantity 4 Heating capacity, kW each 40
Type Electric
Design standards NEC, UL
Seismic design Non-Category I Rev. 0 WOLF CREEK TABLE 9.4-15 SINGLE-FAILURE ANALYSES - DIESEL GENERATOR BUILDING VENTILATION SYSTEM Component Malfunction Consequences Inlet damper Fails to open; Loss of one system; loss of air redundant system is cooling available for cooling the redundant diesel. Recirculation Fails to open Low-temperature damper during winter cutouts on ventilation conditions; sub- system isolate the freezing air being fans. supplied into diesel
generator room Fans Fails to start Loss of one system; redundant system is available for cooling the redundant diesel. Rev. 0 WOLF CREEK TABLE 9.4-16 ESSENTIAL SERVICE WATER PUMPHOUSE VENTILATION SYSTEM COMPONENTS I. Site Ambient Design Temperature Summer 97 F
Winter 7 F II. Supply Fans Type Vaneaxial
Quantity, per room 1
Air flow, cfm each 32,000
Static pressure, in. w.g. each 1.85
Motor horsepower, hp each 20.0
Design Codes and Standards
Fan MS
Motor IEEE-323
Seismic Design Category I III. Unit Heaters Quantity, per room 1
Heating capacity, kW each 25
Type Electric
Quantity, per room 3
Heating capacity, kW each 15
Type Electric
Design Codes and Standards MS
Seismic Design Non-Category I Rev. 0 WOLF CREEK TABLE 9.4-17 SINGLE-FAILURE ANALYSES - ESSENTIAL SERVICE WATER PUMPHOUSE VENTILATION SYSTEM Component Malfunction Consequences Inlet damper Fails to open; loss Loss of one system; of air cooling redundant system is
available for cooling
the redundant pump. Discharge Fails to open; loss Loss of one system; damper of discharge air redundant system is
available for cooling
the redundant pump. Fans Fails to start Loss of one system; redundant system is
available for cooling
the redundant pump. Thermostats Fails to operate Loss of one system; redundant system is
available for cooling
the redundant pump. Rev. 0 WOLF CREEK TABLE 9.4-18 DESIGN DATA FOR PLANT HEATING SYSTEM COMPONENTS I. Main Hot Water Pumps Quantity 2 Type Centrifugal
Capacity, gpm each 918
Head, ft of water, each 150
Motor horsepower, each 60
Casing material Cast iron
Impeller material Bronze
Design standards
Pump MS
Motor NEMA II. Secondary Loop Hot Water Pumps
Quantity 8 Type In line
Hot water flow rate, gpm 11-130*
Head, ft of water 12-15*
Motor horsepower 0.25-1*
Casing material Cast iron
Impeller material Bronze
Design standard MS III. Heat Exchanger Hot water temperature in, F 158 Hot water temperature out, F 198
Inlet steam pressure, psig 25
Steam temperature in, F 267
Condensate temperature out, F 212
Steam flow rate, lb/hr 19,660
Design pressure, psig
Tube side 150
Shell side 150
Tube material Carbon steel, SA-285
SB-111
Shell material Carbon steel, SA-285
Fouling factor 0.0014
Design code ASME Section VIII
- Indicates range for eight pumps.
Rev. 0 WOLF CREEK TABLE 9.4-18 (Sheet 2) IV. Hot Water Expansion Tank Quantity 1 Capacity, gal. 400
Operating pressure, psig 30
Design pressure, psig 125
Material Carbon steel
Shell: SA-414G
Heads: SA-414F
Design Code ASME Section VIII Rev. 0 WOLF CREEK TABLE 9.4-19 DESIGN DATA FOR CENTRAL CHILLED WATER SYSTEM COMPONENTS I. Chilled Water Pumps Quantity 2 Type Centrifugal
Motor horsepower, each 75
Capacity, gpm, each 720
Head, ft each 227
Casing material Cast iron
Impeller material Bronze
Design standards
Pump MS
Motor NEMA II. Central Chillers Quantity 2 Type Centrifugal
Capacity, tons, each 540
Refrigerant R12
Design code ASME Section VIII A. Condenser
Design pressure, psig Tube side 200
Shell side 225
Entering water
Temperature, F 90 Condenser water flow, gpm 1,350 Fouling factor
(service water) 0.002
Tube material 90/10 copper-nickel
Shell material Carbon steel B. Compressor kW, input 691 Type Centrifugal, hermetic C. Evaporator Design pressure, psig Tube side 150
Shell side 185
Chilled water entering
temperature, F 62 Rev. 5 WOLF CREEK TABLE 9.4-19 (Sheet 2) Chilled water leaving temperature, F 44
Chilled water flow, gpm 720
Fouling factor
(chilled water) 0.0005
Tube material Copper
Shell material Carbon steel III. Chilled Water Expansion Tank
Quantity 1 Design pressure, psig 50 Operating pressure, psig 35 Design capacity, gal 315
Material Carbon steel
Design code ASME Section VIII Rev. 4 WOLF CREEK
9.5 OTHER
AUXILIARY SYSTEMS 9.5.1 FIRE PROTECTION SYSTEM
Fire protection is provided in accordance with the requirements of 10 CFR 50, Appendix A, GDC-3.
The fire protection system (FPS) is designed to detect fires, protect the plant
against damage from fire, minimize hazards to personnel, and reduce property
loss due to fire.
Appendix 9.5B provides an evaluation of the effects of postulated fires within
the plant, to ensure the integrity of the reactor coolant system boundary, enable the plant to be placed in a safe condition, and minimize the release of
radioactivity. 9.5.1.1 Design Bases 9.5.1.1.1 Safety Design Bases
Safety-related structures, systems, and components are designed and located to minimize the fire hazard consistent with other safety requirements.
Noncombustible and heat resistant materials are used wherever practical
throughout the unit to minimize the fire intensity in any fire area. This
requirement is in compliance with 10 CFR 50, Appendix A, General Design Criterion 3, Fire Protection.
The basic fire protection for safety-related items is achieved by fire inception
avoidance and through remote separation of systems serving the same safety
function or by fire barriers between such installations.
Therefore, except for an associated containment penetration, the FPS is not a
safety-related system.
SAFETY DESIGN BASIS ONE - The containment isolation valves in the FPS 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.
9.5.1.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The fire protection system is designed to
minimize the effects of fires. It is designed to provide the capability to
extinguish fires encountered in all plant areas. Areas which are protected by
means of manual fire protection are accessible with respect to heat, smoke, toxic combustion products, and radiation.
9.5-1 Rev. 25 WOLF CREEK POWER GENERATION DESIGN BASIS TWO - The plant fire protection water supply system is capable of supplying rated flow with the largest pump out of service.
The fire protection water supply system is rated to supply simultaneously the
maximum design flow for any sprinkler or water spray system and 500 gpm for fire hoses, assuming the shortest fire main flow path is valved out of service.
POWER GENERATION DESIGN BASIS THREE - The fire protection water supply yard main
is arranged so that each branch line from the main may be supplied with water
from the pumps by alternate flow paths. Two-way hydrants with hose houses are installed at about 250-foot intervals along the main.
POWER GENERATION DESIGN BASIS FOUR - Fixed water suppression systems are
installed as required in areas with a high fire or loss potential. Criteria for
determining the need for these systems is in substantial compliance with the
American Nuclear Insurers (ANI) "Basic Fire Protection for Nuclear Power Plants" (March 1976).
POWER GENERATION DESIGN BASIS FIVE - Halon 1301 flooding systems are provided in
the control room cable trenches and chases, switchgear rooms, ESF switch gear rooms, MG sets room, and cable penetrations rooms.
POWER GENERATION DESIGN BASIS SIX - Standpipe connections and hose stations are
provided in areas adjacent to and within stair towers and other points not
greater than 100 feet apart in all normally accessible areas.
POWER GENERATION DESIGN BASIS SEVEN - Portable fire extinguishers are provided
throughout normally accessible areas of the plant in accordance with applicable
NFPA, OSHA, and ANI regulations and recommendations.
POWER GENERATION DESIGN BASIS EIGHT - Alarms are provided in the control room
and signal upon activation of the automatic fire protection systems. Fire and
smoke monitoring and detection systems are installed as required where a
potential for fire exists. These systems alarm in the control room and, if
personnel can be in the vicinity, locally. 9.5.1.2 System Description 9.5.1.2.1 General Description
The powerblock fire suppression systems are shown schematically in Figure 9.5.1-
- 1. The fire area boundaries are indicated in Figure 9.5.1-2.
9.5-2 Rev. 12 WOLF CREEK A comparison of the powerblock design with Regulatory Guide 1.120 is presented in Appendix 9.5A. The powerblock fire protection system is comprised of
diversified monitoring, detection, alarm, and suppression facilities
particularly selected to protect the area or equipment from damage by fire and
includes the following major features:
- a. Suppression systems/capabilities
- b. Fire and smoke detection and alarm system
- c. Fire barriers
- d. Smoke and heat isolation and ventilation
- e. Other fire protection features The capability of the fire detection and extinguishing systems provided for each
area associated with safe shutdown structures, systems, and components is
evaluated in Appendix 9.5B.
The essential service water pumphouse and the yard area around the refueling water storage tank are the only site-specific, safety-related areas which contain combustible materials or are exposed to other fire hazards in their
immediate vicinity. Appendix 9.5B-8 gives a detailed analysis of fire control
in these areas.
The site Fire Protection System (FPS) may be viewed as two subsystems, the
suppression system and the detection/alarm system. Each subsystem has numerous
components; the following describes major components of each system.
Where required, portions of the fire protection system that pass through areas
containing safety-related equipment are seismically analyzed and supported to
prevent damage to this equipment. The system is designed to preclude adverse
flooding of safety-related equipment under seismic conditions, as discussed in
Section 3.6.
Adequate drainage is provided to prevent accumulation from suppression system discharge from causing water damage to components required for safe shutdown of the plant.
Accessibility is provided for the safety-related equipment areas which rely on
manual fire protection.
Site-related buildings have relatively low heights thus eliminating the
potential of fire and smoke spreading by the stack
9.5-3 Rev. 25 WOLF CREEK effect. Air locks are provided at building entrance doors, wherever possible, to minimize the effect of winds on air flow patterns inside each building.
Egress is provided from normally manned areas by clearly marked fire exits and
passageways.
Egress is provided through the power block in accordance with 29 CFR 1910, Subpart E, "Means of Egress."
Appendix 9.5B presents an area-by-area analysis of the powerblock fire loading
and the associated fire detection and suppression systems. In the unlikely
event that the fire cannot be extinguished, fire barriers or physical separation
of redundant components will prevent the fire from causing the failure of
redundant components and systems required for safe shutdown. For redundant
systems which cannot physically be located in separate fire areas, protection is
provided by separation distance and a combination of fire resistive wraps and fire detection and suppression systems.
Noncombustible construction is employed throughout all the buildings to minimize
fire potential. Employment of heat- and flame-resistant materials of
construction and fire resistant coatings reduces the potential for fire, particularly in areas which contain or may interact with safety-related
equipment or rely solely on manual fire protection. Figure 9.5.1-2 shows the
applicable areas of the plant.
Heat and smoke are vented by the normal ventilation exhaust systems in the
auxiliary, radwaste, fuel, diesel, and control buildings, and the auxiliary boiler room. Heat and smoke are vented by the automatic actuation of heat and smoke vents in the turbine building. All flammable gases used at WCGS are stored outside safety-related areas so that a fire involving these gases cannot cause the failure of any safety-related equipment, as discussed in Section
9.3.5. An exception to this is hydrogen/nitrogen test gas cylinders associated
with the hydrogen analyzers. This test gas is less than 10% hydrogen and any
leakage would quickly dissipate below the 4% lower explosive limit (LEL) for hydrogen in air.
Thermal antisweat insulation with appropriate Underwriters' Laboratories ratings
of 25 or less for flame spread is provided for powerblock piping which is
located in safety-related areas.
The powerblock fire suppression systems are designed so that in the event of the
failure of the primary suppression system in a given area, a backup suppression
system is available. In areas so protected, the primary suppression system is a
fixed water or gas-type system with secondary protection from portable extinguishers and hose stations. In areas of low fire loading, the primary method of fire extinguishment is by portable extinguishers with secondary
protection from hose stations. All hose stations are
9.5-4 Rev. 14 WOLF CREEK located so that if a fire occurs at a hose station, preventing access to the hose, the fire may be extinguished by the hose stream from an adjacent hose
station.
The powerblock FPS components in safety-related equipment areas utilize proven
components and have been selected to minimize the risks of inadvertent
operation. Drip-proof safety-related pump motors and electrical equipment are
used, when feasible, to minimize the possibility of damage should fire fighting
operations be required. Wet-pipe sprinkler systems are not used in electric
motor-driven safety-related pump rooms and electrical equipment rooms.
Extinguishing materials used in the FPS are compatible with the equipment in the
areas served to avoid damage to the equipment in the event of a break in the
system. Adequate drainage is provided in the areas where sprinkler or
waterspray systems are used.
The basic fire protection for a safety-related area is achieved through
separation or by fire barriers. The fire protection system, designed to detect, control, and extinguish any fire rapidly and effectively, is not a safety-
related system and is, therefore, a nonseismic Category I system. The design
requirements of the FPS are based on the American Nuclear Insurers (ANI) Guide "Basic Fire Protection for Nuclear Power Plants". Based on the recommendations
of this guide, the FPS includes two 100 percent capacity fire pumps (one
electric motor driven, one diesel engine driven) and two motor driven jockey pumps. One jockey pump is maintained in a standby condition, serving as a backup for the operating jockey pump.
The codes and standards considered in the design of the fire protection system
are listed in Table 9.5.1-1.
The fire protection system is designed so that an inadvertent actuation will not
prevent a safe shutdown of the plant. In most cases, where fixed water
suppression systems are required for safety-related equipment, preaction systems
are provided. The preaction system is provided with closed-head nozzles and is
pressurized with air. Inadvertent opening of a sprinkler head will result in a
loss of air pressure, which is alarmed in the control room. Opening of the
deluge valve is alarmed in the control room. Inadvertent opening of the deluge
valve does not affect the integrity of the safety-related equipment, since the
fire suppression system remains intact with the closed heads.
Standpipes which service safety-related equipment are located outside the
boundary of the equipment room, where possible, so
9.5-5 Rev. 28 WOLF CREEK that an inadvertent pipe failure does not create a flooding condition in the vicinity of the safety-related equipment. Manual valves are provided to isolate
the failed standpipe. The safety-related equipment located in the basement of
the auxiliary building is enclosed by watertight doors and walls to prevent a
flooding condition within the equipment room. Standpipes in the control building
are routed in the stairwells, where possible, to preclude pipe failures, creating a flooding condition in the vicinity of the safety-related equipment.
Floor drains have been provided throughout the control building to preclude
flooding at any elevation due to a failure or if water is required to extinguish
a fire.
Fire detection circuits are continuously supervised for circuit continuity, and
open circuit failure is annunciated by the circuits' supervisory alarm.
The FPS supply piping to the containment was designed to protect the system from
a single active failure. The inside containment isolation valve is a check
valve which is highly reliable by design and considered to be exempt from active
failure due to the absence of any external electrical or control signals which
may be disabled from a fire inside the containment. Refer to Section 3.1.1 for
a discussion of the single failure criteria. A fire inside the containment does
not disable the operability or impair the access to the outside isolation
valves, which may be operated manually, either remotely or locally.
9.5.1.2.2 Component Description
Codes and standards applicable to the fire protection system are listed in Table
3.2-1 and Table 9.5.1-1. The powerblock fire protection system is designed in
accordance with applicable sections of Title 29, Chapter XVIII, Part 1910 (Occupational Safety and Health Standards) of the Code of Federal Regulations.
It is designed in substantial compliance with the requirements of the American
Nuclear Insurers (ANI) and the National Fire Codes of the National Fire
Protection Association (NFPA).
9.5.1.2.2.1 Suppression Systems/Capabilities
FIRE PROTECTION WATER SUPPLIES, YARD MAINS, AND HYDRANTS - Water supply for the
permanent fire protection installation is based on the maximum automatic
sprinkler or fixed water spray system demand with the simultaneous flow of 500
gpm.
9.5-6 Rev. 25 WOLF CREEK The WCGS cooling lake provides water for the FPS. The FPS does not take water from the ultimate heat sink except for the hose stations in the Essential
Service Water (ESW) pumphouse when the ESW system is operating (see Section
9.5B.8). The circulating water screenhouse serves as the intake point and
provides physical protection for the fire pumps and controllers. Figure 10.4-1, sheet 5, shows the location of the fire protection equipment within the
circulating water screenhouse.
The major components of the water supplied fire protection system are: two fire
pumps (one electric motor driven, the other diesel engine driven), two jockey pumps (one primary and one backup)for system pressure maintenance, distribution mains, fire hydrants with hose houses, standpipes and sprinklers. Figure 9.5-1
provides details concerning system components and operations. Figure 9.5-2
shows a site-layout plan for the FPS.
Each pump is equipped with a control panel for manual or automatic operation.
The jockey pump maintains pressure in the system at all times when the fire
pumps are not in operation. The pumping capacity of the FPS is maintained under
conditions of failure of one fire pump. Additional water may be added to the
FPS by way of fire department connections at the circulating water screenhouse
or at any yard hydrant. This additional water could be supplied by fire
department pumpers by taking draft from the WCGS cooling lake by temporary
means. Fire pump flow testing equipment and associated piping are also
contained within the circulating water screenhouse.
To minimize water hammer/pressure surge concerns from fire pump starts, a surge
tank has been installed on the fire pumps common discharge header.
The fire distribution main is looped around the power block area and provides
water to all water supplied fire protection systems with the exception of the
hose stations at the ESW screenhouse. (Refer to Figure 9.5-3 and Section 9.5B.8
for ESW Fire Protection details).
Isolation valves are provided throughout the system in order to maintain maximum
coverage even when portions of the system must be shut down for emergency repair
or routine maintenance. All system control valves designated ZA, i.e., position alarmed, in Figure 9.5-1 are equipped with limit switches and
indicating lights in the main control room to indicate if the valve is fully
open.
Fire hydrants are located along the distribution main. In the power block loop
the maximum distance between hydrants is 250 feet. Each hydrant is located
within a hose house equipped per NFPA 24. Hydrants are also equipped with curb
box shut off valves.
9.5-7 Rev. 28 WOLF CREEK AUTOMATIC WET-PIPE SPRINKLER SYSTEMS - Automatic wet-pipe sprinklers are provided based on NFPA Nos. 13-1975 or 1976 or 1991 to protect the area and
equipment shown on Table 9.5.1-2. Each system consists of a network of piping
which distributes water to closed head sprinklers or spray nozzles. The wet-
pipe systems for the cable chases and the cable area above the access control area have pendant type spray nozzles with fusible link closure.
The powerblock wet-pipe system for the vertical cable chases in the auxiliary
and control buildings is equipped with closed head spray nozzles. Systems are
provided with an alarm check valve equipped to alarm locally and in the control
room when a leakage or flow greater than 10 gpm occurs. The alarm check valve
trim includes a spring-loaded auxiliary valve, a restriction orifice, and a
self-draining retarding chamber in the alarm line to prevent false alarms.
The auxiliary feedwater pipe chase area automatic wet-pipe system is equipped with upright style fusible link sprinklers. This system is provided with a flow alarm device in combination with a gate valve in the main supply header. The
flow through the piping is annunciated locally and in the control room. This
valve is normally locked open and can be used to isolate the piping network for
maintenance purposes.
The sprinkler system in the access control area covers all associated rooms with
the exception of the toilet and shower areas, based on their low fire loadings. Wet standpipe hose stations are spaced at approximately 100 foot intervals
throughout the administration building, shop building, warehouse, Radwaste
Storage Building, Essential Service Water pumphouse, Security Building, and
Technical Support Center. Each hose station is equipped with a shut off valve, a minimum of 50 feet of hose and a variable fog nozzle.
Wet pipe sprinkler systems are provided in the diesel fire pump room in the
circulating water screenhouse, throughout the main warehouse, and in designated
fire hazard areas of the shop building, Security Building, Security diesel-
generator building, Radwaste Storage Building and Technical Support Center, and
Support Building West. Each system is provided with an isolation valve and flow
alarm check valve. Each of the sprinkler systems is automatically initiated by
fusible sprinkler heads located in the hazard area. Operation of any wet pipe
sprinkler system simultaneously sounds an alarm near the fire hazard area and in the main control room with the exception of the Security Building which has a local sprinkler alarm only.
AUTOMATIC WATER SPRAY SYSTEM - Automatic water spray systems for the powerblock, hydraulically designed based on NFPA No. 15-1973, are provided to protect the
equipment as shown on Table 9.5.1-2. Each system utilizes directional solid-cone
nozzles and provides a spray density of 0.25-0.30 gpm
9.5-8 Rev. 13 WOLF CREEK per square foot. Automatic systems are provided with diaphragm-type deluge valves. The valve trim consists of ball check valve, main and auxiliary drains, ball drip valve, drip funnel and support, pressure gauges to indicate pressure
below diaphragm, and a manual pull control station. Piping downstream of the
deluge valves is galvanized.
The deluge valves are tripped by solenoid valves. Pressure switches are
provided on the alarm lines from the deluge valves to alarm in the control room
on the establishment of a water flow. It is possible to periodically test the
alarm line circuit without tripping the deluge valve. Each automatic spray
system has a local control panel that performs the following functions:
- a. On detection, transmits a fire alarm to the control room
and to local bell.
- b. Initiates operation of the deluge valve.
- c. Transmits water flow alarm to the control room.
- d. Supervises deluge valve actuation device circuits and
transmits a trouble alarm to the control room in the
event of a malfunction or power failure.
Failure of the detection system does not trip the deluge valve but does register
a trouble alarm in the control room.
MANUAL WATER SPRAY SYSTEM - Manual water spray systems for the powerblock, hydraulically designed based on NFPA No. 15-1973, are provided to protect the
equipment, as shown on Table 9.5.1-2. Each system utilizes directional solid-
cone nozzles and provides a spray density of 0.25-0.30 gpm per square foot. Manually actuated systems are provided with normally closed outside screw and yoke (OS&Y) isolation (gate) valves with limit switches to indicate the position
of the valves in the control room. Refer to Fire and Smoke Detection Alarm
Systems of this section (9.5.1.2.2) for further description of valve
supervision. The charcoal adsorbers are equipped with water spray piping with a
quick connect/disconnect coupling. Hose from the nearest hose station can be connected through the quick connect/disconnect coupling to supply water to the spray piping.
AUTOMATIC PREACTION SPRINKLER SYSTEM - Automatic preaction sprinklers for the
powerblock are provided based on NFPA Nos. 13-1975 or 1976 to protect the areas and equipment as shown on Table 9.5.1-2. Each preaction system includes a
deluge valve, a check valve, and a network of distribution piping with closed
head sprinklers or spray nozzles. All areas served by the preaction
9.5-9 Rev. 12 WOLF CREEK systems have upright or pendant heads with a fusible link closure. The piping downstream of the deluge valve is galvanized and is normally dry and pressurized
to approximately 20 psig with air. Service air is used, and necessary pressure
regulators are provided. A pressure switch is installed to alarm in the control
room on loss of air pressure (indicating either a breakage or fusing of sprinkler heads) or high pressure (indicating a malfunction of air supply
control).
Refer to the automatic water spray system described above for a description of
the deluge valve and related trim and the local control panel.
MANUAL PREACTION SPRINKLER SYSTEM - Manual preaction sprinklers for the
powerblock are provided based on NFPA No. 13-1976 to protect the areas and
equipment, as shown on Table 9.5.1-2. Manual preaction sprinkler systems are
similar to the automatic preaction sprinklers, with the following exceptions:
- a. The system is not pressurized. Therefore, a check valve, air pressure regulators, and air pressure switch are not
provided.
- b. Since the actuation is manual, there are no automatic release devices. Therefore, local water flow alarms are not provided.
- c. The Containment Fire Protection is normally maintained in a "DRY" condition. The entire system is charged by local and remote manual actuation of the isolation valves. Under normal operating conditions, the deluge valves inside of the containment are maintained in a tripped (or open) condition by placing the manual release valve for each system into the OPEN position. This action prevents the deluge valve from being held closed.
The powerblock system is designed to provide 0.30 gpm per square foot of floor
area for the most remote 1,000 square feet. The system utilizes pendant-type
spray nozzles with fusible link closures.
9.5-10 Rev. 18 WOLF CREEK The fuel oil storage tank is provided with a dry-pipe mechanical type foam
extinguishing system. The foam extinguishing system is manually activated by
opening of the water supply isolation valve located in the fuel oil pumphouse.
A manual hose station with foam generating capability is located on the outside wall of the fuel oil pumphouse.
STANDPIPES AND HOSE RACKS - Wet standpipes for fire hoses are designed based on
the requirements for Class II service of NFPA No. 14-1976.
Except in the containment, hose racks are supplied water from wet standpipes
located throughout the plant. Hose racks are provided for use by the plant
personnel and are located adjacent to stairways and at interior columns so that
no more 100 feet separates adjacent hose racks. Since the fire hazard analysis
must consider fires by transient combustibles, any hose station in the plant may be blocked by fire. However, a fire at any hose station may be reached and extinguished by water from an adjacent hose station. In containment, an
additional length of hose can be added, if required.
The standpipes inside the containment are normally dry. Hand pull stations are
provided adjacent to each hose station. Actuation of any station registers an
alarm in the control room. Plant operations and firebrigade personnel locally and remotely open the isolation valves and charge the standpipes. Hose racks located inside the containment are spaced no more than 100 feet from an adjacent hose rack. Additional hose racks are provided in the truck bay of the radwaste
building to obviate the need for sprinklers or fire barriers in this area.
Coverage by interior hose racks is provided for every accessible area of the
power block and the ESW pumphouse.
Four-inch standpipes are provided for multiple hose stations, and 2-1/2-inch
standpipes are provided for single hose stations. Each standpipe hose station
is equipped with a pressure-reducing 1-1/2-inch angle hose valve with the hose
rack assembly for use by the plant personnel and, except for inside the
containment, a 2-1/2-inch hose valve suitable for connection to a fire
department hose. The fire department hose connections are spaced throughout the
plant so that no more than 130 feet separates two adjacent hose valves.
9.5-11 Rev. 18 WOLF CREEK Each hose rack is provided with 75 feet of 1-1/2-inch hose and adjustable nozzle, with the exception of the diesel generator rooms, cable spreading rooms, and the north end of Corridor No. 2, room 1408, of the Aux Building, elevation
2026, which are protected by hose racks having 100 feet of hose.
Isolation valves for main supply headers and water suppression systems have
supervisory switches. Isolation valves in the standpipes are locked open.
Pressure indicators are provided at the top of the most remotely located
standpipes in the plant.
The hose stations and standpipes provided for the WCGS are in accordance with
the requirements of BTP 9.5-1, Appendix A for plants which received a
construction permit before July 1, 1976 which does not require a Seismic
Category I water system. It should be noted that portions of the fire water
supply piping have been seismically designed to the requirements of Regulatory Position C.2 of Regulatory Guide 1.29 where their failure could affect safety-related equipment.
Vacuum breakers have been installed on standpipe risers to mitigate the
potential vacuum effect of standpipe voiding as a result of a system flow
demand.
PORTABLE FIRE EXTINGUISHERS - Portable fire extinguishers for manual
extinguishment of fires are provided throughout normally accessible areas of the
plant, based on NFPA No. 10-1975 and OSHA regulations.
Where possible, the hand extinguishers are located conveniently and ready for
immediate use. Carbon dioxide, pressurized water, water mist, and potassium bicarbonate-(dry chemical) type portable extinguishers are provided as appropriate for the class of combustible and the type of equipment located in
the hazard area. Figure 9.5.1-2 shows the approximate location and type of fire
extinguishers installed in the power block buildings.
Carbon dioxide extinguishers are 20-pound capacity with a minimum UL rating of
10 B:C. Dry chemical extinguishers (Purple K) are at least 20-pound capacity
with a minimum UL rating of 80 B:C. The capacity of the pressurized water
extinguishers is 2-1/2 gallons each, with a UL rating of 2-A. The water mist extinguishers have a capacity of 2 or 2 1/2 gallons and a UL rating of 2-A:C. In addition, a wheeled dry chemical extinguisher(s) (Purple K) is provided for the turbine building.
Self-contained breathing apparatus and protective clothing is available in the
plant to permit access to hazard areas during and after a fire.
HALON 1301 SYSTEM - Halon 1301 fire suppression systems for the powerblock are
designed based on NFPA No. 12A-1973 and are provided locally to protect the
areas shown in Table 9.5.1-2. The systems are total flooding type and are
designed to maintain a minimum 5-percent concentration at the height of the
highest
9.5-12 Rev. 26 WOLF CREEK combustible in the room for 10 minutes. The 5-percent concentration time may be less than 10 minutes for the upper section of the vertical control room chase.
The Halon 1301 storage cylinders are mounted on racks located outside the hazard
areas. A 100-percent reserve bank is provided for each bank of cylinders. The
one piece extruded storage cylinders are charged to 600 psig and are designed to meet the requirements of the U.S. Department of Transportation. Each cylinder
has a control head (normally closed) which is opened by applying pilot pressure
from the pilot cylinder and a pressure indicator. All manifold and distribution
piping is galvanized. A local control panel is provided with each system to
perform the following functions:
- a. Sound a local alarm horn and initiate Halon discharge on
second level of detection.
- b. Transmit discharge alarm to the control room.
- c. Close selected ventilation dampers and shut off
associated ventilation and/or air conditioning fan
motors.
- d. Supervise and transmit a trouble alarm to the control
room on power failure.
Check valves are provided to prevent the loss of Halon if any cylinder is
disconnected. Bleeder valves are provided to prevent accidental reserve bank
discharge after the main bank has operated. Where one bank serves more than one
area, solenoid-operated selector valves are installed.
Halon 1301 total flooding protection is provided for the QA records vault in the
administration building, the computer room and underfloor area in the Technical Support Center. The Halon system is actuated automatically by cross zoned ionization detectors or by a manual pull station located in or near the hazard area. Actuation of the Halon system will energize an audible and visual alarm
near the hazard and in the main control room and automatically seals off HVAC
ducts into and out of the respective Halon protected area to contain the Halon.
Upon actuation of the system, the Halon discharge substantially completes within ten seconds, filling the protected space with a five to seven percent
concentration of Halon 1301. Manual HVAC controls are provided to purge the QA
records vault in the administration building of Halon after a discharge. Other
Halon protected areas utilize the normal HVAC system and/or portable blowers to
purge the protected space of Halon.
9.5-13 Rev. 9 WOLF CREEK Electronic supervision of the Halon equipment is maintained both locally and in the control room for annunciation of normal, trouble and alarm status.
9.5.1.2.2.2 Fire and Smoke Detection and Alarm System
Automatic fire and smoke detection systems are provided, as indicated in
Appendix 9.5B, based on NFPA Numbers 72D-1975 (Class A) and 72E-1978.
The fire detection and alarm system is divided into convenient zones. There are
provisions for at least 259 zones made up as follows:
No. of zones
- a. Power block and the ESW 210
pumphouse and spares
- b. Site related 49 Site-related fire detection devices are provided in the administration building, shop building, electrical room of the circulating water screenhouse, ESW
pumphouse, Security Building, Technical Support Center, Support Building West
and in the fuel oil pumphouse with the fire detection devices within each
building grouped into zones. Activation of a fire detection device is alarmed
in the main control room and locally. Each building except the ESW pumphouse
contains an annunicator panel to indicate the zone within that building which
initiated the alarm.
Fire detection for the main warehouse and radwaste storage building is
accomplished by indirect means through the monitoring of flow and/or pressure
indicating switches for the associated fire protection water sprinkler systems.
There are four alarm control units (ACU), one of which is site related associated with the fire detection and alarm system. The ACU is of a solid state modular construction. Each ACU has: one Central Processor Unit, wired
initiating device circuits as required to monitor remote dry contact type
devices, (i.e., "Yard Loop Isolation Valves", "Fuel Building Shutoff Valves", etc.), addressable initiating device circuits as needed to monitor smoke and/or
heat sensors and hand pull stations, notification appliance circuits to operate
existing area alarm horns, bells, etc., auxiliary relay control modules for
operation of miscellaneous functions, a network interface board which interfaces
each ACU together as a node within the network, and several power supplies.
Each ACU is supplied 120 V AC, and reduces/converts the voltage internally to a
nominal 24 V DC.
Due to the contact arrangement of the ACU Inverter, the "Primary" power for each ACU is the non-Class 1E 125 V DC system, and the "Secondary" power for each ACU is the non-Class 1E instrument ac system. A switch on the ACU Inverter labels
either the AC Input as "Primary" or the DC Input as "Primary". The DC Input has
been labeled as "Primary " in order to receive a notification of switch-over
from primary power to secondary power. The Inverter will continue to supply power to the load if the AC power is lost, and if the DC power is lost the
inverters static transfer switch disconnects the Inverter and connects the AC
line directly to the load.
The non-Class 1E 125 V DC system is supplied through batteries and battery
chargers. The battery chargers are sized to carry the total connected load
indefinitely. The battery chargers are normally fed from the Class 1E emergency
power system. Upon failure of a battery charger, each separation group battery
can carry the total connected load for 2 hours. Additional load carrying time
can be obtained by selective load shedding and/or closing the bus tie switches between the separation group buses.
9.5-14 Rev. 25 WOLF CREEK The non-Class 1E instrument AC system is continuously supplied by the Class 1E ac emergency power system. The preferred and normal source of the Class 1E power system is the offsite power system. Two physically independent sources of offsite power are fed to the onsite power system. Secondary power is provided by the station emergency diesel generator to each 4.16kV bus. The arrangement, fuel supply, etc., of the WCGS station diesel exceed the minimum requirements of NFPA 72D. The primary power for the remote fire protection panels is provided by the non-
Class 1E 125 V dc system. Each ACU is utilized as power distribution panels for
the remote panels. The non-Class 1E 125 V dc system is continuously supplied by
the 480 V 1E bus via the battery chargers. Two physically independent offsite power sources provide the normal and preferred source to this system.
The standby power source for the secondary supply to the local panels is
provided by the station emergency diesel generator. The arrangement, fuel
supply, etc., of the station diesel exceeds the minimum requirements of NFPA
72D.
The non-Class 1E 125 V dc system is supplied through batteries and battery
chargers. The battery chargers are sized to carry the total connected load
indefinitely. The battery chargers are fed from the Class 1E emergency power system. In the event of a battery charger failure, each battery can carry the dc loads
for approximately 6 hours. This assumes that ac sources are still available for
other non-1E loads. This exceeds the 4-hour requirement of NFPA 72D.
Cables to remote fire protection panels are routed in conduit and supervised for
integrity. Loss of power to these panels is immediately alarmed in the control
room on the fire protection annunciator.
The ACU powers and supervises all detectors, except those for extinguishing
system actuation in the following areas:
- a. Turbine building, including the transformers
- b. Fuel building railroad bay
- c. Diesel generator rooms
The detection systems in these areas actuate the automatic suppression systems
installed in the above areas directly, and are powered by local control panels.
The ACU supplies the power to these panels.
9.5-15 Rev. 10 WOLF CREEK For area detection and alarm systems and for detection for actuation of extinguishing systems in areas not listed above, the signal is received by the
ACU. Alarm and control functions are then initiated by the ACU with the
appropriate annunciation in the control room. For valve supervision, extinguishing system discharge, and all other alarm and trouble signals
generated by the local extinguishing system control panels, the signals feed
through the ACU multiplexer to the control room panel.
The four ACU units are connected to the Network Display Unit and the Color
Graphics Control Unit located in the fire protection control panel in the
control room. The wiring between each node of the system consists of two data
loops. The primary node constantly monitors the system for alarm and circuit
information, and displays this information on the fire protection control panel.
The fire protection control panel houses the network display unit, color
graphics touch screen, terminal board, primary CPU, site-related fire pump
controls, and an Inverter for power. The touch screen color graphic unit, mounted in KC008, will provide pictorial layouts of monitored areas of the
facility with identifying icons for each type initiating device. These icons
will be highlighted when in an alarm or trouble condition and the operator can
acknowledge and reset the alarms from this central control unit. The central
control Network Display Unit (NDU), also mounted within KC008, will log events
in a historic log and will display the current network status on an LCD readout.
There is another unit similar to the control room fire protection panel, located
in the Walter P. Chrysler Building Instrumentation and Controls Shop. This unit is capable of monitoring the fire detection system as the KC008 Panel is. This
unit is used for testing and troubleshooting of the fire alarm circuits. No
acknowledging or resetting of alarms will occur at this unit except during test
mode.
The fire alarm system meets the requirements for Class A systems per NFPA 72D-
1975 paragraph 1321 as detailed below and the requirements for Class 1 circuits
as stated in the National Electrical Code - 1978, Article 725-11. Each
signaling line circuit between the multiplexers and the fire protection control
panel in the main control room is capable of operating for its intended
signaling services during a single break or a single ground fault condition in
the circuit. All initiating device circuits are continuously supervised and
provide a trouble alarm in the event of a break in the circuit. Additionally, initiating device circuits (detectors) serving pre-action suppression systems
for safety-related areas are capable of operating during a single break or a
single ground fault condition. The initiating device circuits for remote alarm
pressure switches are designed in accordance with NFPA 72D-1975, paragraph 1322
or 1323.
Supervision of the fire protection panel is not the primary function of the
plant operator assigned to monitoring the panel. Since there will be few and
infrequent signals to this panel, a full-time supervisor is not justified. Upon
receipt of an alarm or trouble signal at the panel, an audible alarm alerts all
the operators in the control room of this condition.
9.5-16 Rev. 28 WOLF CREEK The fire detection and alarm system includes the following, which is supplemental to the requirements of NFPA 72D, Class A:
- a. For Halon extinguishing systems which are actuated by
two zones of detection in the same hazard area, each zone is not designed to maintain detection capabilities
during a single ground fault or break. Upon generation
of a trouble signal in one of the fire detection zones, a trouble alarm is sent to the control room. In this
condition, the system automatically discharges the Halon
on receipt of an alarm signal from the second zone of
detection.
- b. Upon receipt of a trouble signal on the fire protection
annunciation panel in the control room, an operator is dispatched immediately to the respective zone to investigate the cause of the trouble signal.
The alarm system is of the limited power type as defined in Article 760 of the
N.E.C. and meets the requirements of Class I circuits given in the N.E.C., 1978, Article 725-11.
Suppression system and main header isolation valves are provided with position
switches which are grouped into zones for annunciation of out-of-normal
condition in the control room. Valves which are not electrically supervised will
be subject to administrative supervision which will consist of locking valves
open and periodic inspection. This, however, does not include the drain, vent, and hose valves which are normally maintained closed and instrument root valves
which are normally maintained open since reversing the valve position will not
inhibit operation of the FPS. The general area alarm is by electric horns located throughout the power block.
The sound pressure level is either 90 db or 110 db, both measured 10 feet from
the source, depending upon background noise levels. In addition, integral
flashing lights are installed with the horn within each diesel generator room.
Local manual pull stations are installed throughout the power block buildings.
The detector bases are equipped with lights to quickly identify the detector
that has actuated.
Line-type thermal detectors are provided for the reactor coolant pumps as well
as the cable trays inside the containment. As a backup to these detectors, ionization-type duct detectors are provided in the containment cooling system.
As noted in Appendix 9.5B, all detectors alarm locally as well as in the control
room.
9.5-17 Rev. 0 WOLF CREEK 9.5.1.2.2.3 Fire Barriers
Fire barrier walls, floors, and ceilings are provided as indicated necessary by
the results of the fire hazards analysis, Appendix 9.5B. The fire barrier
ratings and locations are indicated in Figure 9.5.1-2 (drawing series A-1800). Fire barrier configurations that are unique or not directly bounded by fire testing are evaluated in M-663-00017A.
The design, construction, test method, and acceptance criteria for the fire
barriers and related items are as follows:
- a. Fire Rated Barriers - Per applicable sections (determined by construction type) of ASTM E-119, UL
standards, and state building codes.
- b. Fire Barrier Penetration Seals - All fire rated cable
tray penetration seals were tested by an independent
testing laboratory utilizing the following for test
guidance:
- 1. ASTM E 119
- 2. IEEE 634
- 3. ANI/MAERP standard test method
The test program, procedures, and results were approved
by ANI. The tests consisted of exposing all typical
penetration seals (installed in a test slab) to an ASTM
E 119 standard controlled test fire. All powerblock
penetrations passed the test including a hose stream
test.
Fire stops are provided for powerblock cable trays at
each penetration of a fire-rated wall, floor, or
ceiling. The cable rating is compatible with the fire
stop construction. Vertical tray runs that are not
protected by the automatic fire detection and
suppression system generally do not exceed 20 feet in length. In isolated cases where this is not the case, auxiliary protection is provided in the form of fire
retardant coatings, automatic sprinklers, or other means
deemed necessary by the Fire Hazards Analysis.
Horizontal tray runs are generally protected by
separating the corresponding redundant circuits. This
is accomplished by 3-hour fire barriers between the
redundant circuits. Where this is not feasible, a
combination of alternate fire protection means is used
such as fire resistance materials, automatic fire
detection, and suppression system or other provisions
deemed necessary by the Fire Hazards Analysis. Sprayed
on fireproofing is not utilized on cables in safety-
related systems since this would affect cable friction
and thus damping ratios.
9.5-18 Rev. 19 WOLF CREEK The results of the tests, which confirm the integrity of the fire penetration seals, are documented in various fire test reports.
- c. The notes contained on drawing A-1801 define the structural steel fire proofing requirements. The structural steel of all of the ceilings of the fire areas is fire proofed for 3-hour protection with the exception that no fireproofing is provided on the underside of the fuel building roof.
- d. Fire Dampers/Fire Doors - Test methods and acceptance criteria for dampers requiring a fire rating are based on the guidelines established in Underwriter Laboratories standard UL-10B. The devices carry the UL label and are installed in sleeves of 10 gauge (minimum) steel which are attached to the ductwork and supported by the wall. The devices are positioned between the two wall surfaces, with the exception of two fire dampers in the auxiliary building.
Test methods and acceptance criteria for standard type doors requiring a fire rating are based on the guidelines established in ASTM-E-152 or they are U.L.
labeled.
Labeled fire doors are provided, where feasible, with an equivalent rating to the fire barrier in which they are installed. Where required, security devices, thresholds, door sweeps, and weatherstripping are installed on door assemblies. The installation of these devices is judged not to adversely affect the performance of the door and frame
assembly.
- e. Metal Deck Roof - Construction of the metal deck roof assemblies conforms to FM category 1 or UL Class A. The test method and acceptance criteria are as specified in UL-790 or the FM-approved guide.
9.5-19 Rev. 19 WOLF CREEK 9.5.1.2.2.4 Smoke and Heat Isolation and Ventilation
Smoke and heat vents are provided in the turbine building, and auxiliary boiler
room. The turbine building and auxiliary boiler room vents are fusible link
operated, whereas the diesel rooms utilize the exhaust air flow path. The turbine building vents are sized on the basis of at least 1 square foot of
venting area for each 100 square feet of floor area. The auxiliary boiler room
and diesel generator room vents are sized on a basis of at least l square foot
of venting area for each 200 square feet of floor area. The ventilation systems
in other areas of the plant are designed to isolate and confine the smoke and
heat from a fire in the affected areas. Portable equipment is used to vent the
smoke, as required.
Smoke and heat transfer from one area to another during a fire are restricted, and the normal plant ventilation system and portable equipment are used after the fire is extinguished.
- a. The ventilation system equipment used in the fire and smoke isolation of areas in the auxiliary, fuel, radwaste, and control buildings consists of fusible-link actuated fire dampers, power-operated isolation dampers, and centrifugal and vaneaxial fans.
9.5-20 Rev. 19 WOLF CREEK
- b. Ventilation system operations are controlled from the control room. Fire and smoke are automatically isolated in all areas of the auxiliary, radwaste, fuel, and control buildings by the fusible-link actuated fire dampers located
in all rated fire barrier walls or by remote manual operation of area
ventilation systems. Individual areas of the control building can be isolated from the normal supply and exhaust HVAC systems by control switches
in the control room.
- c. Following a fire, the fire dampers are accessible only from that area where they are located. All other systems and components would be remotely
operable from the control room.
- d. All exhaust fans, with the exception of the control building fans, are centrifugal with the motor located outside of the airstream, making them less
susceptible to high gas temperatures. The fans are capable of processing air of temperatures at least as high as the fusible link melting temperature (160 F) of the fire dampers. The control building exhaust fans are vaneaxial with
the motors located in the process airstream. The fan motor is designed for a
minimum 150°F temperature rise.
Since the exhaust fans are all downstream of
the system filter units, they are not subject
to damage from high temperature particles.
Since the HVAC systems of site structures are independent from that of the
powerblock, the ventilation air paths and duct penetrations of the site HVAC
systems will not cause the spreading of fire or smoke from site structures to
any powerblock safety-related structures, systems or components. Fire dampers
are provided at each penetration through any fire rated walls or partitions.
9.5.1.2.2.5 Other Fire Protection Features
EMERGENCY LIGHTING - Section 9.5.3 describes the emergency lighting system and
the design features provided for post fire access and egress.
RADIO COMMUNICATION - A portable radio annunciation system is provided which may
be used by the fire brigade and other operations personnel involved in safe
plant shutdown.
9.5-21 Rev. 19 WOLF CREEK FIRE RESISTANT AND NONCOMBUSTIBLE MATERIALS - Construction materials are noncombustible or Class A rated (regarding flame spread and smoke development)
to the maximum extent practical. Insulation over metal roof decking and the
vapor barrier is securely attached by approved noncombustible adhesive and
perimeter fastening.
Suspended acoustical ceilings and their supports are of UL listed, noncombustible construction. Insulation for pipes and ducts and their adhesives
are noncombustible and UL listed, where practical. Concealed spaces are devoid
of combustibles, as practical. Materials which give off toxic fumes when exposed
to fire are prohibited, where practical. Excluding charcoal adsorbers, all
ventilation prefilters for filtration units are UL Class 1.
COMBUSTIBLE OIL - Areas in which combustible oil-filled equipment is located are
prepared to eliminate the spread of the combustible oil from the immediate area of the equipment. An enclosed gravel filled pit is located beneath the yard transformers. The pit is sized to contain oil from the largest transformer
served by the pit and the water from two transformer water spray systems
operating for 10 minutes (unless the pit serves only one transformer). All
transformers inside the building are the dry type. A fire barrier of at least
2-hour rating is provided between all oil-filled transformers which are
separated by less than 50 feet (except between the station service and start-up
transformers which are within 40 feet of each other without requiring a wall).
The underground diesel fuel storage tanks are set on a firm foundation, backfilled with noncorrosive sand surrounding the tank (6 inches minimum) and
provided with a covering of 2 feet (minimum) of earth.
Each diesel fuel oil day tank is provided with protection features to preclude
the uncontrolled leakage of diesel fuel. The design features provided for the day tank were reviewed and accepted by the NRC at the Wolf Creek Fire Protection Audit of February 6 to 9, 1984.
An oil collection system to collect and contain the lubricating oil for each
reactor coolant pump is installed. See section 9.5B RB.4 Fire Protection for
further discussion of the lube oil collection tanks.
ION EXCHANGE RESINS - To ignite and to sustain combustion is relatively
difficult in ion exchange resins that are in a hydrated form (as opposed to
those in dehydrated form).
9.5-22 Rev. 19 WOLF CREEK The ion exchange resins are received and stored in a hydrated form in nonsafety-related areas of the plant (radwaste and turbine buildings) which are separated
from the safety-related areas by 3-hour fire barriers. The only safety-related
area of the plant which normally contains resins is Fire Area A-8 at elevation
2000 in the auxiliary building. All of the resin, however, is contained in ASME pressure vessels filled with water. Spent resin is sluiced in a spent resin
storage tank in the radwaste building. Fresh resins, still the hydrated form, are introduced into the ion exchange vessels from elevation 2026 of the
auxiliary building Fire Area A-26. Administrative controls ensure that resin
only in quantities required for immediate use in recharging the vessels is
brought into this area, and the containers are hauled away in a timely manner.
Fire Area A-26 is separated from adjacent areas by a 3-hour-rated fire barrier, and automatic smoke detection is provided at the ceiling of the room. This Fire
area is within 75 feet of the hose stations in the corridor (Fire Area A
see Figure 9.5.1-2, Sheet 3). Portable extinguishers are installed outside the room in the corridor.
Since there is no safe shutdown equipment in this fire area and since a fire in
this fire area is confined by the fire barriers until extinguished manually, an
automatic suppression system is not installed.
9.5.1.2.3 System Operation
In addition to the standard fire suppression uses identified below, the system
is also used as a source of water for fire brigade training and as a backup
source of raw water for plant safe shut down for design basis accidents other
than fire.
Automatic wet-pipe sprinkler system operation is initiated on a rise in the
ambient temperature to the melting point of fusible links on sealed sprinkler heads, thus permitting the heads to open. Flow of water through mechanical alarm valves or devices energizes local alarms and registers an alarm condition
on the audio-visual fire protection control panel in the control room. Once
initiated, wet-pipe sprinkler system operation is terminated manually by
shutting an isolation gate valve. The status of these valves is
administratively controlled or electrically supervised and annunciated in the
control room unless specifically stated otherwise.
Water spray system actuation is either manual or automatic, depending upon the
hazard. Automatic operation is initiated by rate compensated thermal detectors.
These sensors detect attainment of a high fixed temperature or rapid rate of
temperature rise and release a tripping device which opens the deluge valve and
thus supplies water under pressure to the spray nozzles. Actuation of the heat
responsive device also initiates a local alarm and registers an alarm condition
on the audio-visual fire protection control panel in the control room. A
pressure switch in the alarm lines from the deluge valve transmits an alarm to
the fire protection control panel to indicate the deluge valve trip and water
flow. System operation is terminated by shutting an isolation gate valve
manually. Closure of this valve registers an
9.5-23 Rev. 19 WOLF CREEK alarm in the fire protection control panel in the control room. Local hand pull stations are provided for alarm and/or actuation function. The local hand pull
stations for the main and startup transformers provide alarm functions only; a
key operated switch on their respective fire panels provides the means for
electric/ manual actuation of the suppression system. The local hand pull stations on all other automatic water spray systems are provided to trip the
deluge valves manually. Manual tripping is annunciated in the control room.
The main and startup transformers are automatically deenergized upon actuation
of the water spray system to prevent damage to the transformer bushings. To
avoid inadvertent spray system actuation spurious transformer trips, dual zone
detection is provided for these transformers. The first zone of detection
alarms locally and in the control room. Detection by both zones trips the
deluge valve and deenergizes the transformer.
Manual water spray system is actuated by opening a normally locked closed isolation valve. Manual opening of the valve is alarmed in the control room.
Preaction sprinkler system operation is initiated by an automatic smoke or
thermal detector, as appropriate for the hazard. These sensors detect either
particles of combustion (ionization and photoelectric detectors) or attainment
of a fixed high temperature (thermal detectors) and release a tripping device to
open the deluge valve, thus supplying water under pressure to fill and
pressurize the system. Actuation of a detection device also initiates a local
alarm, and registers the alarm condition on the audio-visual fire protection
control panel in the control room. In addition, water flow is annunciated in
the control room. Preaction sprinkler system operation is continued on rise in
ambient temperature to the melting point of fusible links on sealed sprinkler
heads, thus permitting the heads to open. Once initiated, system operation is
terminated by manually shutting an isolation gate valve. Closure of this valve is annunciated in the control room. The piping downstream of the deluge valve is pressurized to approximately 20 psig with air. Low air pressure in this
piping gives an audible and visual alarm in the control room, thus loss of air
pressure, which is indicative of open sprinkler heads, will be alarmed.
Local hand pull stations are provided to trip the deluge valve manually. Manual
tripping is annunciated in the control room and locally.
9.5-24 Rev. 19 WOLF CREEK The containment manual preaction sprinkler systems operation is initiated by the operation of the system isolation valve. Thermal detectors, sensing high temperature, initiates a local alarm and registers the alarm condition on the audio-visual fire protection control panel in the control room. The preaction valve release device is maintained in the tripped position. Upon manually opening the system isolation valve, water under pressure is released to fill the system piping. In addition, water flow is annunciated in the control room. The sprinkler system operation is continued on rise in ambient temperature to the melting point of the fusible link on the sealed sprinkler heads, thus permitting the heads to open. Once activated, system operation is terminated by manually shutting an isolation gate valve. Local hand pull stations are provided to trip the deluge valve manually.
Manually tripping is annunciated in the control room and locally.
Hose racks are operated manually by plant personnel. Each rack is controlled by a normally closed hose valve which may be opened without release of water until
the last fold of hose is removed from the rack. Hose nozzles are fully
adjustable from complete shutoff to a straight stream, except in areas where
high voltage electrical equipment presents a shock hazard. In such areas, hose
nozzles without the straight stream capability are provided.
Halon 1301 system operation is initiated by a cross-zoned ionization smoke
detection system. The detectors are mounted in the ceiling of the area
protected, sense particles of combustion, and provide early warning of fire to
activate the suppression system, thus preventing the development of a deep seated cable tray fire. The first zone of detection initiates an alarm locally and in the control room. Detection by both zones sounds a local horn to warn
against impending discharge and starts a discharge time delay device to permit
personnel to leave the area.
Halon is discharged after a preset time delay by actuation of a solenoid valve
on the pilot Halon cylinder, applying pilot pressure to the control heads on
other cylinders in the bank, as required. A minimum 5-percent Halon 1301
concentration is achieved in the enclosure to be protected. In addition, the
system is designed to provide a 5-percent concentration for 10 minutes at the
elevation of highest combustible material in each area protected. The 5-percent
concentration time may be less than 10 minutes for the upper section of the
vertical control room chase. Halon that is expelled is manifolded and piped to
the hazard area and discharged through nozzles. Halon 1301 system piping does
not contain Halon until the system is activated.
9.5-25 Rev. 26 WOLF CREEK A pressure switch is provided to alarm in the control room, indicating Halon discharge. Prior to Halon discharge, selected ventilation dampers close and
selected ventilating and/or air conditioning fan motors associated with the
hazard area shut down. Except for the Control Room Trenches Halon System, a
transfer valve or switch is provided to manually transfer the actuation to the reserve bank after a main bank discharge. For the control room trenches and
chase, automatic discharge of a second cylinder in the main bank is timed after
the discharge of the primary main bank cylinder. The Control Room Trenches
system uses a replacement reserve; the main cylinders, if discharged, must be
removed and replaced by the reserve cylinders which are stored adjacently.
Where one bank of cylinders serves more than one area, solenoid-operated
selector valves are installed to direct the discharge to the affected area.
Local manual actuation is possible by pulling a lever in the pilot cylinder. A
pull pin and seal prevents accidental operation of the manual lever. Pushbuttons
are provided for areas served by Halon 1301 system for remote manual actuation. Manual actuation is similar to detection by both zones of detection. The time delay device is adjustable. Each Halon system is provided with a momentary
contact abort switch to delay the discharge for evacuation purposes. A keylock
switch is also provided to disable the system during maintenance operations.
A pressure gauge on each Halon 1301 storage cylinder is provided, and the
pressure reading is periodically monitored to ensure that the required pressure
is maintained. With the local application of Halon 1301, a leaking cylinder or
leakage to adjacent areas poses no immediate danger, since the resulting
concentrations are less than 10-volume percent. The activation of the Halon
system in an adjacent area, including the cable trenches in the control room, will not endanger the inhabitants within the control room, since the amount of
Halon 1301 is sized only for the areas served and the control room is normally
pressurized. The control building ventilation system and control room
pressurization are discussed in Section 9.4.1. The fire and smoke detection and alarm devices are activated by the several
stages of fire. Ionization detectors alarm at the presence of invisible
combustion particles during the incipient stage of a fire. Flame detectors
respond directly to the infrared radiation emanating from a flickering flame
sustained for at least 5 seconds. These are located in the areas where fire
develops rapidly with a minimum or no incipient stage. Photoelectric smoke
detectors respond directly to visible smoke concentrations of not
9.5-26 Rev. 19 WOLF CREEK less than 1.2 percent per foot of light obscuration caused by smoke. These devices are located in those areas where the use of ionization detectors is
precluded by high air flow in the area or due to the particles of combustion, such as truck exhaust in the radwaste building truck bay, normally found in the
area. Thermal detectors react to the attainment of a high fixed temperature and provide release service for certain automatic systems as discussed above. Air
duct detectors sample the air moving through ducts and alarm at the presence of
particles of combustion, or the presence of smoke (concentrations of not less
than 1.2 percent per foot of light obscuration caused by smoke as above). All
air monitoring, detection, and alarm devices are supervised for reliability in
accordance with NFPA 72D, 1975.
The detection and alarm system is powered by four control units. The alarm
control units also initiate the actuation of automatic suppression systems and
perform other fire-related functions, such as driving area alarm horns and tripping miscellaneous equipment. The multiplexer units installed within the alarm control units constantly interrogate in a preprogrammed sequence the
status of contacts powered by the alarm control units and the open dry contacts (such as valve supervision, water flow alarms, etc.) connected directly to them.
The multiplexer units transmit this status information to the fire protection
control panel located in the control room. The fire protection control panel
also provides an audio-visual display of specific trouble areas, based on the
status input from the multiplexer units. The area local alarm control unit can
be silenced and reset from the main fire protection control panel.
9.5.1.3 Safety Evaluation A comparison of the FPS design with NRC Branch Technical Position, APCSB 9.5-1, Appendix A is presented in Appendix 9.5A. An evaluation of the fires that could
indirectly or directly affect Category I safety-related structures, systems, and components, and other post fire safe shutdown equipment, is included in Appendix 9.5B. The powerblock has been designed to provide protection for safety-related
equipment from hazards and events which could reasonably be expected to occur.
This protection is provided to ensure that recovery from the event is possible, to ensure the integrity of the reactor coolant pressure boundary, to minimize
the release of radioactivity, and to enable the plant to be placed in a safe
condition.
Appendix 9.5B provides the basis for and the results of integrated fire hazards
analyses for the plant to demonstrate that the plant can be safely shutdown following a fire in any fire area of the plant. Even though each area of the plant and each system was designed individually to properly consider the above events, an integrated analysis of rooms, systems, and events is performed to ensure that the above objectives are realized for
each postulated event.
9.5-27 Rev. 19 WOLF CREEK Except for an associated containment penetration, the FPS is not a safety-related system.
SAFETY EVALUATION ONE - Sections 6.2.4 and 6.2.6 provide the safety evaluation
for the system containment isolation arrangement and testability.
9.5.1.4 Tests and Inspection The equipment and systems are inspected and tested after installation in
accordance with applicable codes and requirements including the requirements of
local and state authorities and fire insurance underwriters who have recognized fire protection association standards.
The fire protection system reliability is ensured by periodic test and
inspection.
The pumps are shop tested to ensure that pump characteristics are as specified.
Operational checks, inspection and testing required to maintain fire protection, detection and alarm systems integrity are discussed in the WCGS Fire Protection
Program. Preoperational testing is described in Chapter 14.0. The performance and structural and leaktight integrity of the FPS is routinely demonstrated during
plant operation.
9.5.1.5 Instrumentation Applications Section 9.5.1.2 provides a description of the instruments used to monitor and
actuate the various functions of the FPS.
9.5.1.6 Personnel Qualification and Training The WCGS Fire Protection Program describes the personnel qualifications and
training for the station personnel responsible for firefighting, including those
responsible for maintaining and inspecting the FPS equipment.
9.5.1.7 Equipment Operability The technical requirements of the Fire Protection System (FPS) are described in
Fire Protection Administrative Procedures. The Fire Protection Administrative
Procedures also describe the operability requirements and actions required, if not operable, of the FPS. A list of all the surveillance procedures required to
prove operability of all FPS equipment is maintained in the Fire Protection
Administrative Procedures. This list also describes the operability requirement
that each procedure covers.
9.5-28 Rev. 19 WOLF CREEK 9.5.1.7.1 Fire Detection Instrumentation
Operability of the fire detection instrumentation ensures that both adequate
warning capability is available for the prompt detection of fires and that Fire
Suppression Systems, that are actuated by fire detectors, will discharge extinguishing agents in a timely manner. Prompt detection and suppression of
fires will reduce the potential for damage to safety-related equipment and is an
integral element in the overall facility fire protection program.
Fire detectors that are used to actuate fire suppression systems represent a
more critically important component of a plant's fire protection program than
detectors that are installed solely for early fire warning and notification.
Consequently, the compensatory action for inoperable fire detection
instrumentation is more stringent.
The loss of area wide detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire
protection for any area due to the subsequent loss of automatic suppression
capability. The establishment of frequent fire patrols in the affected areas is
required to provide detection capability until the inoperable instrumentation is
restored to operability.
9.5.1.7.2 Fire Suppression Systems
The operability of the fire Suppression Systems ensures that adequate fire
suppression capability is available to confine and extinguish fires occurring in
any portion of the facility where safety-related equipment is located. The Fire
Suppression Systems consists of the water system, spray, and/or sprinklers, Halon, and fire hose stations. The collective capability of the Fire
Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility Fire Protection Program.
In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the
affected areas until the inoperable equipment is restored to service. When the
inoperable fire-fighting equipment is intended for use as a backup means of fire
suppression, a longer period of time is allowed to provide an alternate means of
fire fighting than if the inoperable equipment is the primary means of fire
suppression.
9.5-29 Rev. 19 WOLF CREEK The Surveillance Requirements provide assurance that the minimum operability requirements of the Fire suppression Systems are met. An allowance is made for
ensuring a sufficient volume of Halon in the Halon storage tanks by verifying
either the weight or the level of the tanks. Level measurements are made by
either a U.L., or F. M. approved method, or by ultrasonic measurement corrected for temperature using equipment calibrated to standards traceable to NIST. The
term, "simulated fire" test signal, is interpreted to mean actuation of an
automatic Fire Protection System by any of the release mechanisms provided, e.g., fire detectors, hand pull stations, fusible line/mechanical, manual, hydro/mechanical, etc.
In the event the Fire Suppression Water System becomes inoperable, immediate
corrective measures must be taken since this system provides the major fire
suppression capability of the plant.
9.5.1.7.3 Fire Barrier Penetrations
The functional integrity of the fire barrier penetrations ensures that fires
will be confined or adequately retarded from spreading to adjacent portions of
the facility. This design feature minimizes the possibility of a single fire
rapidly involving several areas of the facility prior to detection of and
extinguishing of the fire. The fire barrier penetrations are a passive element
in the facility fire protection program and are subject to periodic inspections.
Fire barrier penetrations, including cable penetration barriers, and fire doors
are considered functional when the visually observed condition is the same as
the as-designed condition. Fire dampers accessible for drop testing are
considered functional if they fully close, as designed, when subjected to
periodic drop testing outlined in plant procedures. Fire dampers inaccessible
for drop testing may be visually inspected to assess functionality, provided that accessible damper drop testing results do not indicate an adverse trend. Fire damper test sampling and frequency are discussed in procedure AP 10-100. For those fire barrier penetrations that are not in the as-designed condition, an evaluation shall be performed to show that the modification has not degraded
the fire rating of the fire barrier penetration.
During periods of time when a barrier is not functional (Outside of Containment), either: (1) a continuous fire watch is required to be maintained
in the vicinity of the affected barrier, or (2) the fire detectors on at least
one side of the affected barrier must be verified operable and an hourly fire
watch patrol established, until the barrier is restored to functional status.
During periods of time when a barrier is not functional (Inside of Containment), either: (1) Verify the operability of the Containment Cooler Fire Detection Zone
219, or (2) Provide compensatory measures described in procedure AP 10-103, Attachment B. 9.5.1.7.4 Fire Protection Plan
AP 10-100, Fire Protection, explains the requirements for the entire fire
protection plan. It is an administrative control procedure which describes the Fire Protection Program in section 9.5.1.7.5.
9.5.1.7.5 Fire Protection Program
The WCGS Fire Protection Program uses a defense-in-depth approach to fire
protection to ensure that the plant can be safely shut down and public health
and safety maintained in the event of fire. This protection is accomplished
through automatic detection and suppression systems, trained personnel, fire
emergency procedures, and administrative controls.
9.5-30 Rev. 25 WOLF CREEK 9.5.1.7.5.1 Fire Protection Program Organization and Individual Responsibilities 9.5.1.7.5.1.1 Administrative Organization
The administrative organization of WCGS as it relates to the Fire Protection
Program is described in Section 13.1.
9.5.1.7.5.1.2 Operational Organization
The operational organization as it relates to the Fire Protection Program is
described in Section 13.1.
9.5.1.7.5.1.2.1 Fire Brigade Leader
The Fire Brigade Leader is a member of the Fire Brigade designated by the Shift Manager to direct the efforts of
the Brigade. In the event of a fire, the Fire Brigade
Leader selects a command location and directs the
other Brigade members in the attack on the fire. He
establishes communications with the Control Room and keeps
the Shift Manager informed of the progress of the fire
fighting effort. He is responsible for the safety of the
other Brigade members. He evaluates the hazards that the
particular circumstances present and ensures that Brigade
members are properly equipped to deal with those hazards. In
the event that off-site fire fighting personnel are present, he uses them as he deems appropriate.
9.5.1.7.5.1.2.2 Fire Brigade The WCGS Fire Brigade is composed of a minimum of five
persons from the on duty work force. The Brigade does not
include any of the plant physical security personnel
required to be available to fulfill the response
requirements of paragraph 73.55(h)(2) of 10 CFR Part 73, "Physical Protection of Plants and Materials." The Fire
Brigade is responsible to the Shift Manager for those
actions necessary to assist in extinguishing and containing
the spread of fire. Individual and collective efforts of
the members of this Brigade are directed by the Fire Brigade
Leader.
9.5.1.7.5.2 Training and Qualifications
9.5.1.7.5.2.1 Fire Brigade
9.5.1.7.5.2.1.1 Physical Qualification
Each person assigned to the Fire Brigade receives an annual
physical examination that evaluates those conditions which
are not acceptable for strenuous fire fighting activities.
9.5-31 Rev. 19 WOLF CREEK 9.5.1.7.5.2.1.2 Initial Training for Qualification of Fire Brigade Members
Before being assigned to the Fire Brigade, a person must successfully complete a formal training program established by the Supervisor Fire Protection. This program may consist of off-site training conducted by outside organizations, on-site training conducted by WCGS personnel, or a combination of both. Training should include, but is not limited to the following:
- a. Indoctrination to the plant fire fighting plan with specific coverage of each individual's responsibilities.
- b. Identification of the fire hazards and associated types of fires that could occur in the plant and an identification of the location of such hazards.
- c. The toxic characteristics of expected products of combustion.
- d. Identification of the location of fire fighting equipment for each fire area and familiarization with the layout of the plant, including access and egress routes to each area.
- e. The proper use of available fire fighting equipment and the correct method of fighting each type of fire. The types of fires in cables and cable trays, hydrogen fires, fires involving flammable and combustible liquids or hazardous process chemicals, construction fires, and record file fires.
- f. The proper use of communication, lighting, ventilation, and emergency breathing equipment.
- g. The proper method for fighting fires inside buildings and confined spaces.
- h. The direction and coordination of the fire fighting activities (Fire Brigade Leaders only).
NOTE: Fire Brigade Leaders and all other operations personnel, assigned to Fire Brigade duty, also receive training in the following areas:
- i. Detailed review of fire fighting strategies and procedures.
9.5-32 Rev. 19 WOLF CREEK
- j. Review of the latest plant modifications and corresponding changes in fire fighting plans.
9.5.1.7.5.2.1.3 Practice Sessions
All Fire Brigade members participate in training exercises
designed to provide experience in actual fire extinguishment
and the use of emergency breathing apparatus under
strenuous conditions, as those which may be encountered in
fire fighting. These exercises are included in each Fire
Brigade member's initial training. In order to maintain
qualifications, each Fire Brigade member must repeat
participation in a practice session at least once every 365 days
after completion of the initial practice session. For purposes of
scheduling this activity, a grace period of one calendar quarter may be applied at the discretion of Management to support unscheduled plant outages or regularly scheduled refueling outages. Individual Fire
Brigade members will be allowed a 31 day grace period in order to make up
missed practice sessions.
9.5.1.7.5.2.1.4 Assignment to Fire Brigade Duty
Upon the completion of the training outlined in Sections
9.5.1.7.5.2.1.2 and 9.5.1.7.5.2.1.3 the individual's name is
placed on the Fire Brigade roster. He may then be assigned
to Fire Brigade duty as needs require.
9.5.1.7.5.2.1.5 Complete Training for Fire Brigade Members
Over each two year period following initial qualification, Brigade members receive periodic refresher training such that all areas of Section 9.5.1.7.5.2.1.2 are covered within
the two year period. A two year period is defined as any eight
consecutive calendar quarters. For purposes of scheduling this activity, a grace period of one calendar quarter may be applied at the discretion
of Management to support unscheduled plant outages and regularly
scheduled refueling outages. Individual Fire Brigade members will be
allowed a 31 day grace period to make up missed periodic refresher
training. The training may consist of a combination of classroom work
and drills as described in Section 9.5.1.7.5.2.6. Training sessions are
conducted at least once per calendar quarter. To maintain active
status on the Fire Brigade Roster, each member must complete all
of the refresher training, practice sessions and must participate
in at least two fire drills per year. For the purpose of drills, a year
is defined as a period from January 1 to December 31 of each subsequent
year of service.
9.5.1.7.5.2.2 Training for Personnel Authorized Unescorted Access to
WCGS
All WCGS personnel, vendor personnel, etc., requiring unescorted access to the
WCGS receive basic fire safety training during initial indoctrination.
Refresher training is given as part of orientation training. Training
includes, but is not limited to, the following:
- a. Recognition and response to alarms and
announcements
- b. WCGS fire protection policies
9.5-33 Rev. 19 WOLF CREEK 9.5.1.7.5.2.3 Training for Off-Site Fire Department Personnel
The off-site fire department personnel receive yearly training in the following
areas:
- a. Radiological precautions, principles, and personal
exposure monitoring
- b. Site layout, access routes, and major plant fire
hazards
- c. Contamination monitoring and decontamination
techniques
- d. Site firefighting equipment and locations
- e. Coordination with WCGS Fire Brigade
Agreements with off-site firefighting organizations are on file as part of the
WCGS Emergency Planning basis.
9.5.1.7.5.2.4 Training for Fire Watches
Personnel performing duties as fire watches receive training in the following
areas:
- a. Purpose and duties of fire watches
- b. WCGS fire prevention policies
- c. Plant layout and locations of portable and installed
firefighting equipment
- d. Operation of assigned portable equipment for fire
watches
9.5.1.7.5.2.5 Training for Maintenance and Inspection Personnel
Personnel performing maintenance and inspection functions on Fire Protection
equipment are trained in the maintenance or inspection activity. Periodic
retraining of these personnel is conducted.
9.5-34 Rev. 19 WOLF CREEK 9.5.1.7.5.2.6 Drills
9.5.1.7.5.2.6.1 Frequency
9.5.1.7.5.2.6.1.1 Each shift Fire Brigade participates in a planned drill at least once per calendar quarter. Each shift
Fire Brigade participates in both an unannounced and a backshift drill yearly. Yearly is defined as a period from January 1 to December 31 of a given year.
9.5.1.7.5.2.6.1.2 The off-site fire department participates in at least one WCGS fire drill yearly. Yearly is defined as
a period from January 1 to December 31 of a given year.
This may be in combination with other drills.
9.5.1.7.5.2.6.1.3 At 3 year intervals, a randomly selected unannounced drill is critiqued by qualified individuals independent of the licensee's staff. A copy of the written report from such individuals shall be available for NRC review. A 3 year interval is defined as occurring at least once in a 3 calendar year period.
9.5.1.7.5.2.6.2 Drill Monitoring
Drill monitoring is performed by the Fire Protection Staff with
the assistance of other plant personnel as designated by the
Supervisor Fire Protection. 9.5.1.7.5.2.6.3 Drill Critiques Drill critiques are conducted for drill participants by the
Fire Protection Specialist and the drill monitors. The
Fire Protection Specialist determines what corrective
actions, if any, are required as a result of the drill and
ensures that these actions are taken in a timely manner.
9.5.1.7.5.3 Administrative Controls for Fire Prevention
Implementing procedures have been developed in the following fire prevention
areas:
9.5.1.7.5.3.1 Transient Ignition Sources
A permit system controls the use of transient ignition sources. The Supervisor
Fire Protection designates those individuals who are qualified to conduct an
evaluation of the potential fire hazards of each work activity which requires
the use of a transient ignition source. Personnel performing the evaluation of
such activities are responsible for specifying any necessary precautions or
limitations necessary for the safe completion of the prescribed work activity.
9.5-35 Rev. 19 WOLF CREEK 9.5.1.7.5.3.2 Combustible Materials
A permit system controls the storage and handling of combustible materials.
Welding, cutting and acetylene-oxygen gas systems inside or adjacent to safety-related areas of the WCGS are controlled by the Transient Ignition Source
program.
9.5.1.7.5.3.3 Inoperative Detection, Alarm, or Suppression Equipment
A permit system controls those activities that render fire detection, alarm, or
suppression equipment inoperative. This system specifies actions and any
precautionary measures required prior to and during deactivation of any such
equipment.
9.5.1.7.5.3.4 Testing and Maintenance of Fire Protection Equipment
An inspection plan for fire protection equipment has been developed. Personnel
responsible for maintenance and inspections of the fire protection system are
listed in Sections 13.1.2.2.2 and 13.1.2.2.7. The inspection plan includes the types, frequency and detailed procedures for inspection. Test procedures specify the steps necessary to verify conformance with design and system performance
requirements following modification, repair or replacement of portions of the
fire protection system.
9.5.1.7.5.3.5 Leak Testing
Leak testing is accomplished by approved methods only. The use of open flame or
combustion smoke for this purpose is specifically prohibited.
9.5.1.7.5.3.6 Smoking
Smoking is prohibited in the Protected Area Boundary except where specifically
designated.
9.5.1.7.5.4.0 Fire Emergency Procedure and Pre-Fire Plans
9.5.1.7.5.4.1 A Fire Emergency Procedure directs the response to fire, explosion, and unusual hazardous conditions. This procedure details:
- a. Actions to be taken by the person discovering a
fire
9.5-36 Rev. 19 WOLF CREEK
- b. Actions to be taken by the Control Room operators, including announcement of alarm, notification and activation of the Fire
Brigade, initiation of suppression systems, and
notification and coordination of off-site fire department.
- c. General instructions to Fire Brigade including
assembly points, protective equipment
requirements, communications, and location of
special fire protection equipment
- d. Salvage actions to be taken to minimize post-
fire damage and to implement restoration
9.5.1.7.5.4.2 Pre-Fire Plans have been developed for potential fire hazards which could affect safety-related systems. Plan content is based upon the USAR comparison to 10 CFR 50, Appendix R. Reference Table 9.5E-1, Section III.K.12.
9.5.1.7.5.5 Quality Assurance
The Quality Assurance Program for fire protection is specified in Section C of
Table 9.5A-1.
9.5.2 COMMUNICATION
SYSTEMS
The communication systems include internal (in-plant) and external
communications designed to provide convenient and effective communications among various plant locations, and between the plant and locations external to the plant.
9.5-37 Rev. 19 WOLF CREEK The public address, and maintenance jack systems each have their own dedicated conduit systems. The public address system conduit can be used as a raceway, where documented on engineering drawings, for non-metallic fiber optic cable used for data transmission in systems that are not used for plant process equipment control. To the extent practicable these conduits are embedded to minimize the systems exposure to hazards. The Telephone Conduit system is also
dedicated except that it is also used as a raceway for non-metallic fiber optic
cable which supports the Local Area Network (LAN) Computer Communications.
A malfunction of a given system component will disable that particular component and hence, communications would have to be resumed using one of the remaining
systems from the station. An accident, such as a fire, that disables a PA system
loop would disable that particular communications loop, and thus communications
would have to revert to one of the remaining systems for that entire loop. The
maintenance jack system, if disabled by fire, would require repair before it
could be restored to service.
9.5.2.1 Design Bases 9.5.2.1.1 Safety Design Bases
There is no safety design basis for the communication systems.
9.5.2.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - Intraplant voice communication is provided
by a public address (PA) and telephone system.
POWER GENERATION DESIGN BASIS TWO - A maintenance jack system, utilizing plug-in
telephone type handsets and handsets with 5channel jack stations, is provided to
supplement the public address system.
POWER GENERATION DESIGN BASIS THREE - An evacuation alarm system is provided to serve the entire plant.
POWER GENERATION DESIGN BASIS FOUR - A dial telephone system is provided for
plant-to-offsite communication on a continuous basis.
POWER GENERATION DESIGN BASIS FIVE - Telephone communication is provided between
the control room and various plant locations.
POWER GENERATION DESIGN BASIS SIX - A control-room-to-offsite communication
system is also provided for emergency purposes.
POWER GENERATION DESIGN BASIS SEVEN - An offsite communications system is
provided for reliable plant-to-offsite communications and will consist of:
- a. Touchtone Telephone System
- b. Touchtone Telephones (System Independent)
- c. Wide Area Fiber Optic System (Supplied by Western Resources)
- d. Security Radio System
- e. Plant Radio System
POWER GENERATION DESIGN BASIS EIGHT - Non-process Computer communications are
provided between various plant locations for normal operations support.
9.5-38 Rev. 25 WOLF CREEK 9.5.2.2 System Description The plant communication systems are illustrated schematically in Figures 9.5.2-1
and 9.5.2-2.
9.5.2.2.1 Intraplant Communications
Communications within the plant are provided as follows:
- a. For operating purposes, a public address system is provided, consisting of handset stations and loud-speaker assemblies, each having its own plug-in
amplifier.
The system provides six separate independent communication channels - one general page and five party lines. In the
Control Room, the system can operate in one of two
selectable modes: "outage" and "normal." In the outage
mode, communication between parties in the plant and the
Control Room can be easily and quickly established using the
Control Room page channel (channel 1). Communication
between parties within the plant can be easily and quickly
established by using the general page channel. In the
normal mode all paging is performed on the general page
channel. The party line channel is normally used after the page call is completed. As many as five party lines may communicate simultaneously. The portion of the PA system connecting the fuel transfer area in the containment, the fuel storage area and new fuel handling area in the fuel building, and the control room can be isolated from the remainder of the PA system from the control room. This permits extended use of the fuel handling communications system without disruption to the remainder of the system.
The PA system is supplied power from two separate 208/120-V instrument busses through a transfer switch. In case of failure of the normal power source, an automatic transfer is made to the alternate source. Each instrument bus is fed through an isolation transformer and can be supplied power by one of the emergency diesel generators.
Each PA amplifier unit is equipped with an adjustable volume control which may be turned up in high noise
areas.
Handset stations are designed with a noise-canceling mouthpiece for use in high noise areas.
A wall-mounted handset station is provided for communication between the auxiliary control panel and other areas of the plant.
9.5-39 Rev. 19 WOLF CREEK b. For communications between the control room and equipment being maintained, calibrated, or tested, a five-channel maintenance jack system consisting of a permanently interconnected series of jack stations is provided. The system provides two-way communication between multiple stations on a preselected channel by means of plug-in headsets. Power is provided through the same sources as the PA system described above.
- c. An audible evacuation alarm system is provided by means of a multi-tone generator whose output is broadcast throughout the plant via the public address system. The evacuation alarm tone is discernible from that of the fire alarm. A volume control bypass relay provides maximum sound. The audible alarm system is supplemented by visual alarms in high noise areas. Manual activation of the system is from the main control room.
- d. For administrative purposes, an outside automatic dial type telephone system is provided with extensions for intra use.
9.5.2.2.2 Plant-to-Offsite Communications
9.5.2.2.2.1 Touchtone Telephone System
The Touchtone Telephone System uses VoIP (Voice Over Internet Protocol) technology, which transmits calls using a digital signal over an IP network (i.e. corporate computer network). The phones system supports telephone communications throughout the power block, in the main Control Room, security building, administration building and various other buildings around the site. The new system has diverse routing consisting of a minimum of four Primary Rate Interfaces (PRI). Each PRI has 24 trunks supporting inward/outward calls to the local public telephone system.
The VoIP system is powered through a battery backup system which can provide about 8 hours of service after loss of offsite power.
9.5.2.2.2.2 Touchtone Telephone (System Independent)
In the event of a total system failure there is a minimum of 4 touchtone telephones which remain operational for access to the local Burlington exchange.
9.5-40 Rev. 28 WOLF CREEK 9.5.2.2.2.3 Wide Area Fiber Optic System
Provide communications connectivity to extensions at WCGS for communication with
the Operating Agent Home Office.
9.5.2.2.2.4 Security Radio System
Refer to the WCGS Physical Security Plan.
9.5.2.2.2.5 Plant Radio System
The Plant Radio System consists of two repeater sites: the Turbine Building and
the Meteorological Tower. The sites have multiple repeaters to support
communications inside the Plant structures as well as in a 5 mile radius of the
Plant. The system provides two-way portable and mobile communications for normal and emergency Plant operation. Portable and mobile users have communications capability with fixed operator positions in the Control Room, TSC
and EOF, and with security radio consoles, if desired.
9.5.2.3 Safety Evaluation Diverse systems are provided to assure a means of communication. For additional
reliability, the PA system is supplied from either of two 208/120-V
instrumentation busses, which can be supplied by redundant diesel generators. Power to each of the instrumentation busses is fed through isolation
transformers connected to a Class IE motor control center.
The PA and maintenance jack systems are provided with power from the same
sources, but are completely independent.
Those plant areas which must be manned for hot shutdown have been evaluated to
ensure that the expected noise levels will not make the provided communication
systems ineffective. Each of these areas, and the communication systems
available in each area, are listed in Table 9.5.2-l. Of all the areas listed in Table 9.5.2-1, the only area where high noise levels
are expected is the diesel generator room. However, all PA amplifiers may be
turned up by means of the volume control.
A11 communication systems circuits are enclosed in conduit or site designated
raceway to provide protection for cables.
Should the PA and maintenance jack systems become inoperable, two-way radios and
dial telephones are available.
There are no safety functions associated with the communication systems.
9.5-41 Rev. 19 WOLF CREEK 9.5.2.4 Tests and Inspections Systems of the types described above are conventional and have a history of
successful operation at existing plants.
All communication systems are inspected and tested at the completion of
installation to ensure proper coverage and audibility. During plant operations, the routine use of the normal communication systems ensures their reliability.
Periodic inspection and testing is performed on the backup systems. Where
applicable, the radio equipment is checked and calibrated in accordance with
Federal Communication Commission guidance.
9.5.3 LIGHTING
SYSTEM
The plant lighting systems include normal, standby, and emergency lighting designed to provide adequate lighting during normal operation, accident conditions, a loss of offsite power, and postulated fires including a fire in
the control room which requires evacuation of the control room.
9 5.3.1 Design Bases 9.5.3.1.1 Safety Design Bases
The lighting system has no safety design bases.
9.5.3.1.2 Power Generation Design Bases
POWER GENERATION DESIGN BASIS ONE - Adequate lighting systems are provided in
areas used during shutdown or emergency, including the appropriate access or
exit routes.
POWER GENERATION DESIGN BASIS TWO - Lighting intensities are designed for those
levels recommended by the Illuminating Engineering Society.
POWER GENERATION DESIGN BASIS THREE - Mercury-vapor fixtures are not used inside the containment or directly above the fuel storage pool .
POWER GENERATION DESIGN BASIS FOUR - The main control room is given special
attention to reduce glare and shadows at the control boards.
9.5.3.2 System Description The plant lighting distribution systems are illustrated schematically in Figure
9.5.3-1.
9.5-42 Rev. 19 WOLF CREEK 9.5.3.2.1 Normal Lighting System
The normal lighting system consists of a complete distribution network of
cables, raceways, transformers, lighting panels, fixtures, receptacles, and
switches.
This system is fed from the non-Class IE auxiliary power system and is designed
to provide adequate illumination levels for normal plant operating and service
conditions. A selected number of normal lighting fixtures are chosen to be used
in the standby lighting system.
9.5.3.2.2 Standby Lighting System
The standby lighting system consists of selected fixtures of the normal lighting
system in the auxiliary, control, reactor, and turbine buildings. These fixtures
are supplied from the emergency diesel generators during the loss of offsite
power and are isolated from the Class IE power source on the occurrence of an SI
signal. These circuits are treated as non-Class IE, non-associated.
9.5.3.2.3 Emergency Lighting Systems
The emergency lighting system consists of individual sealed-beam, self-
contained, battery units to provide silhouette lighting, that is, to provide
shadows and to highlight obstructions to personnel for access and egress. Eight-
hour battery units are located throughout the plant, including areas requiring
operator actions for safe shutdown following a fire.
The locations of emergency lighting fixtures have been selected to provide for
access and egress to/from the auxiliary, control, fuel, diesel generator, reactor, and radwaste buildings; the communication corridor; and the ESW pump
house. Lighting to and from the radwaste building and the ESW pump house is
provided by the site. The emergency lighting provided ensures egress from these
areas in the event of a loss of off-site power, a design basis event, or a fire, should the normal lighting and standby lighting (powered by the emergency
diesels) be unavailable. All emergency lights credited to meet 10 CFR 50, Appendix R requirements are provided with 8-hour batteries.
As described in Section 7.4, the WCGS design ensures that there are no
preplanned manual operations outside of the control room for maintenance of hot
shutdown following a design basis event, a severe natural phenomenon, or a loss
of offsite power. The WCGS design also includes provisions to achieve and
maintain hot and cold shutdown using safety-related equipment.
9.5-43 Rev. 27 WOLF CREEK For cold shutdown, operator actions may be required in the electrical penetration rooms (1409 and 1410) to isolate the accumulator tanks and to open
the RHR suction valves from the hot legs. These actions may be taken as late as
72 hours following an event. The safe shutdown scenario does not require access
to the containment for hot shutdown but could require access to containment for cold shutdown.
The 8-hour battery-backed emergency lighting fixtures provide lighting for fire
fighting activities in the areas listed previously. For fires in the general
plant areas, operator actions may be required outside of the control room to maintain hot standby. In the event of a fire, the standby lighting system should remain operable; however, the emergency lighting provides access to the areas
and lighting within the areas where operator actions are required. These
provisions meet the requirements of 10 CFR 50, Appendix R, Paragraph III.J.
In areas required to be manned for safe shutdown, sufficient lighting is
directed at the control panels to enable operation of controls. This includes
the following:
- a. Main control board
- b. Auxiliary shutdown panel(s)
- c. Diesel generator control panel(s)
In the area above the main control board and operator's console, the emergency
lighting system consists of emergency lights with 8 hour battery packs and
fixtures supplied from a Class IE battery through a normally deenergized
contactor. The contactor control circuit monitors the normal ac lighting feed
and automatically energizes the fixtures from one Class IE battery upon loss of
ac power. The contactor, switch, wiring, raceways, and fixture mounting for this
system are equivalent to Class IE with regard to separation, color coding, and
seismic supports.
One and one-half hour battery units are used in the turbine building and the hot
machine shop. Each unit is connected to the normal lighting ac source for
maintaining the charge and is automatically transferred to its internal
batteries upon loss of ac power.
9.5.3.3 Failure Analysis
The emergency lighting system is designed to provide lighting in areas used
during safe-shutdown, design basis events or fire fighting activities. In the
event of loss of offsite power, the emergency lighting is maintained by
batteries, as outlined in Section 9.5.3.2.3. The standby lighting system in
these areas is powered from the emergency diesel generators in the event of the
loss of offsite power. Refer to Section 9.5.3.2.2.
9.5-44 Rev. 27 WOLF CREEK 9.5.3.4 Tests and Inspections AC lighting circuits are normally energized and require no periodic testing. The
dc emergency lighting is inspected and tested periodically to ensure the
operability of the automatic switches and other components in the system.
9.5.4 EMERGENCY
DIESEL ENGINE FUEL OIL STORAGE AND TRANSFER SYSTEM
The emergency diesel engine fuel oil storage and transfer system (EDEFSTS)
provides onsite storage and transfer of fuel oil to the diesel engines.
9.5.4.1 Design Bases 9.5.4.1.1 Safety Design Bases
The EDEFSTS is safety related and is required to function following a loss of offsite power to achieve and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The EDEFSTS is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The EDEFSTS remains functional after a SSE and
performs its intended function following the postulated hazards of fire, internal missile, or pipe break.
SAFETY DESIGN BASIS THREE - 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 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-45 and 46).
SAFETY DESIGN BASIS FIVE - The EDEFSTS 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 - Following a loss of offsite power, the system provides
onsite storage and delivery of fuel oil for at least 7 days of operation of the diesel generators at their continuous rating.
9.5-45 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS SEVEN - Following a loss of offsite power, the EDEFSTS is designed to supply fuel oil at all times under the most severe environmental
conditions probable at the plant site. The EDEFSTS complies with Regulatory
Guide 1.137 to the extent discussed in Table 9.5.4-3.
9.5.4.1.2 Power Generation Design Basis
The EDEFSTS serves no power generation function.
9.5.4.2 System Description 9.5.4.2.1 General Description
The EDEFSTS is shown schematically in Figure 9.5.4-1. Each diesel engine has its own individual fuel oil storage and transfer system. The EDEFSTS for each diesel
engine has an underground storage tank with a transfer pump, day tank, strainers
and filters, piping, valves, instruments, and controls. The oil fill connection
to the underground storage tank is located above grade and includes a strainer.
A truck connection, normally isolated by a locked closed valve, is provided on
the transfer pump discharge piping to empty the fuel oil storage tank, if
necessary, using the transfer pump.
Two strainers are installed in parallel on the transfer pump discharge piping to
the day tank with an isolation capability so that the flow can be diverted to either strainer without disrupting the system operation.
An interconnecting pipe with normally locked closed valves is installed between
the two transfer systems to enable the supply of fuel oil from either storage
tank to be transferred to either day tank. Figure 9.5.4-1 indicates that the
cross-connection piping between the two fuel oil tanks is Seismically Supported.
This capability to supply fuel oil to either engine from either tank is not
utilized.
The following measures prevent the accumulation of moisture and residual
sediment in the bottom of the fuel storage tanks, and insure a supply of quality
fuel if an event should occur that would require replenishment of fuel oil
without the interruption of diesel generator operation.
Physical arrangement of the fuel oil system is such that the fuel oil fill point and the transfer pump outlet points are more than 40-feet apart. The
minimum required oil level in the tank dissipates the turbulence effect of
the incoming fuel stream. There are strainers, filters and some gradient in
the piping and tanks to ensure a supply of clean fuel to the diesel
generators. Prior to adding new fuel oil to the supply tanks, onsite samples of the new fuel are taken for testing of specific gravity, viscosity, water and
sediment. Fuel oil stored in the tanks meets the requirements of the Diesel Fuel Oil Testing Program. Fuel tanks are physically checked for water monthly.
9.5-46 Rev. 19 WOLF CREEK Accumulated condensate is removed from the bottom of the buried storage tanks when its presence is detected or suspected. Fuel tanks are emptied and cleaned of accumulated sediment at 10-year intervals. Protection against corrosion of the buried fuel tanks is provided by: o External - protective coating and an impressed current type cathodic protection system. o Internal - Frequent removal of water eliminates this source of corrosion Design of tank provides a slope to direct water to sump System operation provides flow to motivate water toward sump Tank replenishment provides similar motive force Treatment of tank contents with biocide to eliminate biologic growth which promotes waste production and acidifies contents of tank The immersed surface provides a passive coating of fuel oil Visual examinations are conducted to check for leakage, structural distress or corrosion. Records of inspections and tests are maintained to assist in evaluating the extent of degradation of the corrosion protection systems. The day tank supplies fuel oil to the diesel engine by gravity. Duplex basket
strainers and duplex oil filters are installed in series on the fuel oil lines
from the day tank to the engine.
The fuel oil day tank is located more than 20 feet horizontally from the diesel
engine and well below the insulated diesel exhaust piping and, therefore, is not
exposed to any high temperature surfaces.
There is no elevated fuel oil piping adjacent to the engine. The fuel oil piping
between the engine and the day tank drops down from the tank and runs along the
floor until it reaches the engine. The diesel engine itself sets on a 6-inch
skid and therefore is elevated above the floor.
There are no open flames in the diesel generator room.
Open flames in the diesel generator area as well as in other plant areas are
controlled by plant administrative procedures.
The excess fuel from the engine is returned to the day tank. Leakage from the
injection nozzles is drained by gravity to the fuel oil storage tank.
If any growth of algae should occur in the tank detection would be accomplished
by either periodic sampling of the fuel oil or visual inspection of the tank
interior. Should any algae be found a decision would be made at that time as to what methods of treatment would be employed to prevent future occurrences. Cleanup would be a manual operation. Should any algae occur and get into the
fuel oil system, the system strainers and filters would remove it before it
entered the diesel engine.
9.5-47 Rev. 19 WOLF CREEK 9.5.4.2.2 Component Description
Codes and standards applicable to the EDEFSTS are listed in Tables 3.2-1 and
9.5.4-1. The EDEFSTS is designed and constructed in accordance with quality group C and Seismic Category I.
- a. Emergency Fuel Oil Storage Tanks
Two cylindrical emergency fuel oil storage tanks, one for each diesel engine, are provided. The tanks are horizontal and have elliptical heads.
The tanks are buried underground near the diesel generator building. The capacity of each tank is based on the fuel consumption by one diesel engine for operation at continuous rating for 7 days. The tank is vented, via a flame arrester, to the atmosphere outside the diesel building at a location above all the tank
connections.
The fuel oil storage tank fill and vent lines terminate outside of the diesel generator building; however, they are routed underground from the tank to the building and from the building to the outside. The vent line has a flame arrestor, which is goosenecked downward. The bottom of the flame arrestor is approximately 15 feet above grade. The fill connection is capped and penetrates the building wall at approximately 3 feet above grade. The maximum probable flood level does not exceed grade and, therefore, the vent and fill connections are not subject to flood conditions.
As noted, the fill connection is capped and the vent goosenecked down and, therefore, neither allow the entrance of water into the system during adverse environmental conditions.
A concrete vault is provided on top of each tank to permit access to the manhole, the pump, the pump discharge piping and conduits, level transmitters, and sample line.
The storage tanks have integral sumps. Each tank is sloped to the sump. Sample lines extend from the sumps to the vaults for periodic bottom sampling and water draw-off. The sample lines can be used to empty the storage tanks when the fuel oil level falls below the transfer pump suction.
An additional 6 inch sample line is provided to enable multiple level samples to be taken and to verify the fuel oil level manually.
9.5-48 Rev. 19 WOLF CREEK The storage tanks are buried below grade at a sufficient depth to prevent floating when the tanks are emptied. Fill lines are installed above the probable maximum ` flood level to prevent any entry of water into the tank.
The exterior surfaces of the tanks are coated with Bitumastic. An impressed-current-type cathodic protection is provided for the tanks.
- b. Emergency Fuel Oil Transfer Pumps
Two transfer pumps are provided, one for each diesel generator. The pumps are the horizontal centrifugal type and are submerged in their respective fuel oil storage tanks. Each pump motor is powered from the same Class IE bus its associated diesel generator serves. The capacity of each transfer pump is approximately twice the consumption rate of the diesel engine at its continuous
rating.
- c. Emergency Fuel Oil Day Tanks
Two cylindrical day tanks are provided, one for each diesel engine. The day tanks are horizontal and have ASME (torispherical) heads. Each day tank is installed in the room of the engine it serves, and the tank elevation ensures adequate net positive suction head on the diesel engine-driven fuel oil pump at all times. Each day tank has a capacity equal to approximately 80 minutes of operation of the diesel engine at its continuous rating. The tanks are vented, via a flame arrester, to outside the diesel generator building. The overflow and drain connections on the day tank are piped to the emergency fuel oil storage tank. A sampling connection is provided to the bottom of the tank for periodic sampling of the fuel oil for quality and for drawing off any accumulated condensation and sediment.
The interiors of the day tanks are waterproofed with a coating of bitumastic.
Instrumentation is provided, as described in Section 9.5.4.5. The level settings ensure that there is at least a l-hour supply of oil in each day tank for the diesel engine (based on fuel consumption at a load of 100% of the engine continuous rating plus a minimum margin of 10%) at the level where the oil is automatically added to the day tanks by the transfer pumps in the storage tanks.
Fuel oil storage tank low level and low-low level, fuel oil system strainer high pressure differential, and day tank low level, high level, and low stand pipe level are alarmed directly in the control room.
9.5-49 Rev. 19 WOLF CREEK The level of fuel oil in the day tank is indicated in the control room.
The diesel engine basket strainer high pressure differential alarm, fuel filter high differential pressure alarm, and low fuel oil pressure alarm all result in a control room "diesel trouble" light and alarm. An operator would go to the alarm panel in the diesel generator room to determine the specific alarm.
None of the above malfunctions result in harmful effects to the diesel engine, and none result in the tripping of the diesel engine. Station operating procedures give the operators guidance for responding to these alarms.
- d. Piping and Valves
All piping in the EDEFSTS is carbon steel. The exterior surfaces of the underground piping are coated with coal tar and wrapped. Cathodic protection is provided for underground piping.
9.5.4.2.3 System Operation
Each diesel engine has its own independent fuel oil pumping train from the fuel
oil storage tank to the day tank, with tie lines normally isolated between the
two flow paths. Level transmitters installed on the day tanks initiate the
signals to start the transfer pumps on low level and stop the pumps on a high
level. If the diesel generators are running, the transfer pumps will run continuously.
A fire detection signal from the diesel building stops the fuel oil transfer
pump. However, the fuel oil pump will not be stopped if the diesel generator is
running to preclude any spurious trips from the fire protection system under
accident conditions.
Fuel oil is supplied by gravity to the diesel engine-driven fuel oil pumps.
The storage tanks are replenished by delivery trucks through the oil fill
connections located above grade. Accumulated moisture may be withdrawn prior to
adding new fuel oil through the sample nozzle to minimize the possibility of
degrading the overall quality of the new fuel. The strainers and filters in the
system will ensure that any sediment stirred up during replenishment does not
reach the injection nozzles. Refer to Section 9.5.4.2.1 for additional
discussion of methods used to prevent degrading overall fuel quality during tank
fill operations.
Based on availability, there are numerous suppliers that could deliver to WCGS
within a few hours of contact.
9.5-50 Rev. 19 WOLF CREEK Approximate
Company Location Distance (Miles) Mobil Oil Kansas City, MO 105 Security Oil Wichita, KS 130
There are multiple access routes to WCGS. WCGS is sited above PMF levels and it
is anticipated that not all access routes to WCGS would be blocked for an
extended period of time and fuel oil can be delivered as needed.
9.5.4.3 Safety Evaluation Safety evaluations are numbered to correspond to safety design bases in Section
9.5.4.1. SAFETY EVALUATION ONE - With the exception of the fill and vent connections, the
above-ground portions of the EDEFSTS are located inside the diesel generator
building. This building is designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other 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 this building. The underground portions of the EDEFSTS
have adequate earth coverage for missile protection. The access vaults for the
storage tanks are missile protected. The missile covers and vaults form
watertight barriers to prevent water entry into the tanks from ground water and
flooding. SAFETY EVALUATION TWO - The safety-related portions of the EDEFSTS are designed
to remain functional after a safe shutdown earthquake. 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 design of the EDEFSTS provides complete
redundancy; therefore 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.
In the event that one of the oil storage tanks must be replenished without
interruption of the associated diesel generator operation, accumulated moisture
may be withdrawn prior to adding new fuel oil by the normal procedure (i.e. through the sample nozzle) to minimize the possibility of degrading the overall quality of the new fuel. Refer to Section 9.5.4.2.1 for additional discussion
of methods used to prevent degrading overall fuel quality during tank fill
operations.
SAFETY EVALUATION FOUR - The EDEFSTS is initially tested with the program given
in Chapter 14.0. Periodic inservice functional testing is done in accordance
with Section 9.5.4.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the EDEFSTS.
9.5-51 Rev. 19 WOLF CREEK SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification
and seismic category applicable to the safety-related portion of this system.
Table 9.5.4-1 shows that the components meet the design and fabrication codes
given in Section 3.2. All the power supplies and control functions necessary for safe function of the EDEFSTS are Class IE, as described in Chapters 7.0 and
8.0. SAFETY
EVALUATION SIX - The capacity of each emergency fuel oil storage tank is
sufficient for 7 days of operation of one diesel generator at its continuous
rating. Within this period, additional fuel can be delivered to the plant site
by truck or rail.
SAFETY EVALUATION SEVEN - Maintenance of the fuel oil temperature is achieved by
enclosing the equipment in heated buildings for portions of the system above ground or by burial below the frostline of underground portions of the system. Plant procedures assure that the "cloud point" of the fuel oil is lower than the
lowest expected temperature in the vaults over the storage tanks.
9.5.4.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
The EDEFSTS is tested periodically, along with the complete diesel generator system. This test demonstrates system performance and structural and leaktight
integrity of accessible system components.
With the exception of underground portions of the system, all equipment and
components are readily available for inspection and maintenance. Provisions are
made to pressure test the underground portions of the system. The fuel oil
transfer pumps can be tested independently of the diesel generator by draining
the day tanks (manually to the storage tanks) to the levels that automatically
start the pumps. The pump flowrate can be verified by monitoring the day tank
level indicators and/or observing flow indicators. Level annunciators in the storage tanks can be used to verify the leaktightness of the tanks.
The fuel oil in the storage tank and day tanks is periodically sampled to verify
quality. Degenerated fuel oil can be pumped out of the storage tanks by truck
connections provided on the discharge of each fuel oil transfer pump.
Accumulated moisture may be removed periodically, via the sample line, to
minimize degradation of the fuel oil. Refer to Section 9.5.4.2.1 for a
discussion of methods used to prevent the accumulation of moisture and sediment
in the storage tanks.
9.5.4.5 Instrumentation Applications The EDEFSTS instrumentation is designed to provide indication of system
parameters and automatic operation of the transfer pumps.
9.5-52 Rev. 19 WOLF CREEK The emergency fuel oil storage tanks have level transmitters to alarm in the
control room on low level (corresponding to 7-day capacity) and a low-low level
to indicate low suction head for the transfer pumps. A local outdoor level
indicator is also provided.
The day tanks have level transmitters to automatically start and stop the
transfer pumps. In addition, control room annunciation of high level and low
levels is provided to indicate system malfunction. The low level alarm is
provided to allow sufficient time for plant staff to accomplish minor repairs if
required, before all fuel in the day tank is consumed. Day tank level
indicators are provided in the local diesel engine control panels and in the
main control room.
The strainers and filters installed in the system have pressure differential switches and pressure differential indicators. High differential pressure across the strainer in the transfer pump discharge is alarmed in the control
room, whereas a high differential pressure across the strainers and filters on
the diesel engine skid is annunciated on the local control panel. Low fuel oil
pressure downstream of the diesel engine-driven pump is annunciated in the local
control panel. A common alarm is provided in the control room for any local
annunciator.
Test connections are provided on the fuel oil transfer pump discharge lines to
monitor the pressure or temperature, if desired.
Table 9.5.4-2 summarizes the EDEFSTS alarms and indicators of various system
parameters.
9.5.5 EMERGENCY
DIESEL ENGINE COOLING WATER SYSTEM The emergency diesel engine cooling water system (EDECWS) provides cooling water
to the emergency diesel engines. This is a closed cycle system, and serves as an
intermediate system between the diesel engines and the essential service water
system.
9.5.5.1 Design Bases 9.5.5.1.1 Safety Design Bases
The EDECWS is safety related and is required to function following a DBA and to achieve and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The EDECWS is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The EDECWS remains functional after a SSE and performs
its intended function following the postulated hazards of fire, internal
missiles, or pipe break (GDC-3 and 4).
9.5-53 Rev. 19 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-44).
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-45 and 46).
SAFETY DESIGN BASIS FIVE - The EDECWS is designed and fabricated to codes
consistent with the 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 EDECWS is designed to remove heat from the diesel
engines to permit their operation at continuous nameplate rating. SAFETY DESIGN BASIS SEVEN - The EDECWS is designed to maintain the diesel engine
in a hot standby condition to ensure quick starting of the diesel engine.
9.5.5.1.2 Power Generation Design Bases
The EDECWS has no power generation design bases.
9.5.5.2 System Description 9.5.5.2.1 General Description
The EDECWS is shown schematically in Figure 9.5.5-1. Each diesel engine has its own cooling water system. Each cooling water system consists of a jacket
cooling water system and an intercooler cooling system.
The EDECWS rejects heat to the essential service water system.
Each jacket cooling water system consists of an engine-driven pump, a jacket
water heat exchanger, an electric motor-driven keep-warm pump, an electric keep-
warm heater, piping, valves, controls, and instrumentation. The engine-driven
pump circulates water through the cylinder jackets and the jacket water heat
exchanger, where the extracted heat is transferred to the essential service water system. When on standby status, the electric motor-driven pump circulates water through the electric heater and the engine cylinder jackets to keep the
engine warm.
9.5-54 Rev. 19 WOLF CREEK Each intercooler cooling system consists of an engine-driven intercooler pump, intercooler heat exchanger, piping, valves, controls, and instrumentation. The
engine-driven intercooler pump circulates water through the intercooler heat
exchanger and the engine-mounted intercoolers. Turbocharged air is cooled by
the intercoolers prior to its entry into the combustion air manifold, and the extracted heat is transferred to the essential service water system at the
intercooler heat exchanger.
An expansion tank is provided to accommodate any volumetric changes in the
EDECWS due to thermal transients or leakage and to absorb pump pulsations. The
expansion tank maintains adequate suction at the engine-driven pumps and
provides a release point for undissolved gases in the system.
The jacket water and intercooler cooling systems have high point vents which are
piped to the jacket water expansion tank. This ensures that the systems are filled with water at all times.
9.5.5.2.2 Component Description
Codes and standards applicable to the EDECWS are listed in Tables 3.2-1 and
9.5.5-1. The safety-related portions of the EDECWS are designed and constructed
in accordance the quality groups listed in Table 3.2-1 and seismic Category I.
ENGINE-DRIVEN COOLING WATER PUMPS - The jacket cooling water pump and the
intercooler pump are driven by the engine. The pumps are the horizontal
centrifugal type. Adequate suction is provided by the jacket water expansion
tank. A failure of either of these pumps constitutes an engine failure.
HEAT EXCHANGERS - The heat exchangers in the EDECWS are the horizontal shell and
tube type. Essential service water is supplied to the tube side. The heat exchangers are arranged in series so that the essential service water first flows through the intercooler heat exchanger and then the jacket water heat
exchanger. Heat exchanger design by the engine manufacturer is based on maximum
heat rejection requirements and a specified 0.002 fouling factor for the lube
oil cooler and jacket water heat exchangers, a specified fouling factor of
0.0019 for the intercooler heat exchanger, and 95 F entering water temperature.
Because these design conditions are inherently conservative, the diesel engine
cooling system contains a suitable margin for operation under all design
conditions.
EXPANSION TANK - One expansion tank is provided in the EDECWS to accommodate
volumetric changes in the jacket cooling water and intercooler cooling water
systems due to thermal transients or leakage. The expansion tank serves to
absorb any pump
9.5-55 Rev. 19 WOLF CREEK pulsations. The tank is a horizontal cylindrical type and is located at a suitable elevation to provide adequate suction head to the engine-driven pumps.
The makeup to the expansion tank is from the demineralized water storage and
transfer system which is nonseismic Category I. The makeup quantities are controlled automatically by level switches. The capacity of the expansion tank
is based on providing sufficient reserve capacity for operation of the diesel at
continuous rating for at least 7 days. Provisions are included for the addition
of chemicals, as required.
JACKET COOLING WATER KEEPWARM PUMP AND HEATER - An electric motor-driven
keepwarm pump is provided to circulate water to the cylinder liners through an
electric heater to keep the engine warm on standby. The pump and heater
combination on each diesel skid is powered from the same Class IE bus served by
their associated diesel generator. PIPING AND VALVES - All piping in the EDECWS is carbon steel. The inlets to the
heat exchangers are controlled by self-contained thermostatic valves. Section
9.2.1 describes
the piping and valves associated with the essential service
water system. Due to the manufacturer's service and design experience, the flex
connections used within the EDECWS are of the nonmetallic type, are designed and
constructed to manufacturer's standards, and have proven reliable for the
intended service.
9.5.5.2.3 System Operation
GENERAL - The jacket cooling water system and the intercooler cooling water
system are closed-cycle systems. High points in these systems are vented to the
expansion tank. This assures that all spaces are filled with water when a water
inventory is maintained in the expansion tank. The EDECWS uses demineralized water with a suitable corrosion inhibitor. The demineralized water chemistry meets EPRI Closed Cooling Water Guidelines, which exceeds the manufacturer's recommendations. When the engine is on standby, the jacket water keepwarm pump circulates water
through the electric heater and the cylinder jacket. This keeps the engine warm
to facilitate starting. The heater is controlled by a temperature switch. A
failure of the keepwarm system will lower the jacket cooling water temperature. As described in Section 8.3, low jacket cooling water temperature will be
alarmed locally and annunciated in the control room as a common trouble alarm.
9.5-56 Rev. 19 WOLF CREEK During diesel engine operation, the keepwarm system is automatically shut off. The engine-driven jacket cooling water pump and the intercooler cooling water
pump circulate cooling water. The heat extracted by the cooling water is
transferred to the essential service water system at the jacket water and
intercooler heat exchangers. The cooling water to each heat exchanger is modulated by a self-contained thermostatic valve, so that the temperature
differentials across the engine in the jacket cooling water and the intercooler
cooling systems remain at the design minimum. The thermostatic valves bypass the
heat exchangers when the engine is started so that the cooling water is rapidly
brought to normal operating temperatures. The thermostatic valves are designed
to permit the full volume of cooling water to flow through the engine. The
expansion tank makes up for any volumetric changes due to thermal transients and
minor leakage.
The WCGS engine cooling water characteristics have been reviewed and accepted by the diesel manufacturer. A corrosion inhibitor is used in the diesel generator cooling water system. The manufacturer includes in his instruction manual
cooling water treatment guidelines and recommends the use of a corrosion
inhibitor.
Treatment of the essential service water system that serves the cooling water
heat exchanger is described in Section 9.2.
During normal plant operation, the service water system through the essential
service water system piping provides a heat sink to the EDECWS. When the diesel
is started during an emergency on a safety injection signal, the essential
service water system absorbs the heat from the EDECWS. However, the essential
service water pumps do not activate immediately because they are connected to
power from the diesel generators. The normal period of diesel engine operation
prior to the start of the essential service water flow is less than a minute. This includes the diesel generator start and acceleration to its rated speed, energizing the sequencer for starting the essential service water pump motor, and starting and accelerating the essential service water pump motor to rated
conditions. The diesel engines are designed to operate at a continuous
nameplate rating without a cooling water supply for 3 minutes.
Loss of water from the EDECWS is detected by monitoring both the operation of
the D-G room sump pump and the number of times makeup water is introduced into
the system by monitoring the operation of the makeup water solenoid valve.
There are no mechanical limitations within the EDECWS that would restrict the
operation of the diesel generator when less than full electrical power
generation is required.
9.5-57 Rev. 19 WOLF CREEK 9.5.5.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases.
SAFETY EVALUATION ONE - The EDECWS is located in the diesel generators 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 this building.
SAFETY EVALUATION TWO - The safety-related portions of the EDECWS are designed
to remain functional after a safe shutdown earthquake. 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 design of the emergency diesel generators provides
for complete redundancy; therefore no single failure of the EDECWS portion
compromises the diesel generators safety functions. Vital power can be supplied
from either onsite or offsite power systems, as described in Chapter 8.0. There
are no interconnections between the two engine cooling systems. Therefore, no
failure of or between any of the engine cooling subsystems would result in any
degradation of the other diesel engine.
SAFETY EVALUATION FOUR - The EDECWS is initially tested with the program given
in Chapter 14.0. Periodic inservice functional testing is carried out in
accordance with Section 9.5.5.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the EDECWS. 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. Table 9.5.5-1 shows that the components meet the design and
fabrication codes given in Section 3.2. All the power supplies and control
functions necessary for the safe function of the EDECWS are Class IE, as
described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - The EDECWS components are sized to remove heat from the
engine when operating at continuous nameplate rating and transfer this heat to
the essential service water system.
9.5-58 Rev. 19 WOLF CREEK SAFETY EVALUATION SEVEN - The EDECWS has a jacket water keepwarm system designed to keep the engine in hot standby. This allows the quick start and loadings
required for emergencies.
9.5.5.4 Tests and Inspections
Preoperational testing is described in Chapter 14.0.
The EDECWS is tested periodically, along with the complete diesel generator system. This test demonstrates the performance, structural, and leaktight
integrity of all the system components.
The safety-related portions of the EDECWS are designed and arranged (to the
extent practical) to permit preservice and inservice inspection.
9.5.5.5 Instrumentation Applications The EDECWS instrumentation is designed to permit automatic operation and remote
control of the system and to provide continuous indication of system parameters.
Refer to Section 8.3 for details of instrumentation. The local control panel has indicators for coolant pressure and coolant inlet and outlet temperatures.
The frequency of makeup to the expansion tank is monitored by the data logger
printout of the opening of the makeup water valve.
All applicable instrument controls, sensors, and alarms for the diesel cooling
water system are shown on Figure 9.5.5-1, Sheets 1 and 2.
Those temperatures and pressures which are alarmed in the diesel generator room
but result only in the general control room "diesel trouble" alarm are high
jacket water temperature from the engine, low jacket water temperature from the engine, low jacket water pump discharge pressure, low jacket water expansion tank level, high intercooler water temperature from the engine, low intercooler
water temperature from the engine, and low intercooler pump discharge pressure.
An operator would go to the alarm panel in the diesel generator room to
determine the specific alarm.
There are no cooling water system alarms which alarm directly in the control
room.
Local indication in the diesel generator room is provided for jacket water
temperature to and from the engine, intercooler water temperature to and from
the engine, water temperature from the generator outboard bearing, jacket water
pump discharge pressure, and intercooler pump discharge pressure.
9.5-59 Rev. 19 WOLF CREEK None of the above malfunctions which alarm in the control room result in harmful effects to the diesel or shutdown, except for the high jacket water temperature
from the engine. High jacket water temperature is sensed by four separately
mounted temperature switches which have three discrete temperature setpoints, two of the four sensors are at the highest of the three temperature setpoints. Operation of any one switch sounds an alarm, and operation of any two (one of
which must be at the highest temperature setpoint) results in engine shutdown.
Station operating procedures give the operators guidance for responding to these
alarms.
9.5.6 EMERGENCY
DIESEL ENGINE STARTING SYSTEM
The emergency diesel engine starting system (EDESS) provides a reliable method
for starting the emergency diesel engines for all modes of operation. The EDESS is divided into two parts -- a safety-related portion which is that
portion downstream of and including the air start tank check valve and the
remainder of the system which is nonsafety-related.
9.5.6.1 Design Bases 9.5.6.1.1 Safety Design Bases
The safety-related portion of the EDESS is required to function following a DBA and to achieve and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The EDESS is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The safety-related portion of the EDESS remains
functional after a SSE and performs its intended function following the
postulated hazards of fire, internal missiles, or pipe break.
SAFETY DESIGN BASIS THREE - 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 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-45 and 46).
9.5-60 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS FIVE - The safety-related portion of EDESS 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 to isolate components, systems, or
piping is provided, when required, so that the system's safety function is not
compromised. This includes isolation of components to deal with leakage or
malfunctions and to isolate nonsafety-related portions of the system.
SAFETY DESIGN BASIS SEVEN - The safety-related portion of the EDESS is capable
of storing sufficient air to allow for at least five consecutive crank cycles of
approximately 3 seconds or 2 to 3 revolutions of the diesel engine without
external support or assistance.
9.5.6.1.2 Power Generation Design Basis
The EDESS serves no power generation function.
9.5.6.2 System Description
9.5.6.2.1 General Description
The EDESS is shown schematically in Figure 9.5.6-1. Each diesel engine has its
own starting system. The starting system for each diesel engine has two
redundant, independent starting air trains, one for each bank of cylinders.
Each starting air train consists of a compressor, dryer, starting air tank, filters, strainers, piping, valves, controls, and instruments. Each bank of
engine cylinders has its own engine-driven air start distributor with a pilot
air connection to each cylinder for operation of the cylinder air start valves.
The engine starts on either or both banks of cylinders.
The WCGS emergency diesel generator is started by using air admitted directly to
the cylinder through an air start distributor. The air is stored in two
separate reservoirs, which are charged from two separate compressors. Dryness
of the air is assured by the use of an air dryer on each compressor.
The air dryers are of the automatic recharging type, using purge flow to effect
re-charge of the membrane. They are designed to provide air dried to a dew point of -40 degrees F at the design flow rate of 50scfm, which is well below the lowest design room temperature of 60 degrees F (See Note 8 of Table 3.11(B)-
1 for clarification).
9.5-61 Rev. 27 WOLF CREEK Oil carryover from the compressor is controlled by use of a prefilter upstream of the dryer.
Starting air pressure is also used to operate the governor servorack booster
which opens the fuel injection pump racks to ensure adequate fuel at startup.
For emergency shutdown, starting air pressure is used to operate a fuel rack
shutdown cylinder to close the fuel racks.
The pressure transmitters associated with the pressure indicators on the local
control panel are supplied with air from the starting air system.
9.5.6.2.2 Component Description
Codes and standards applicable to the EDESS are listed in Tables 3.2-1 and
9.5.6-1. The safety-related portion of the EDESS is designed and constructed in
accordance with the quality groups listed in Table 3.2-1 and seismic Category I.
COMPRESSORS - Each train of the diesel starting system has an electric motor-
driven compressor. The compressor is a 3-stage air-cooled design with
sufficient capacity to charge its associated air tank from minimum to maximum
starting air pressure in less than 30 minutes. The compressor start/stop
functions are automatically controlled by a pressure switch monitoring the
starting air tank pressure. The compressor is nonsafety related, and the
compressor motor is powered from a non-IE source.
Each compressor has air cooled intercoolers and an aftercooler on the compressor
itself to cool the compressed air.
DRYERS - A membrane air dryer is provided in each starting air train. The dryer is the automatic regeneration type and includes a prefilter. The dryer provides moisture-free air with a dew point temperature (at rated pressure) of minus 40
F. The dryer package is nonsafety-related.
STARTING AIR TANKS - Compressed air is stored in the starting air tanks. Two
starting air tanks are provided for each emergency diesel engine - one for each
redundant starting air train. Each tank has sufficient capacity for five
cranking cycles without recharging. The air tanks are equipped with safety
valves and normally closed drains to blow down any accumulated moisture and
sediments periodically. The starting air tank is safety-related.
9.5-62 Rev. 27 WOLF CREEK STRAINERS AND FILTERS - Each compressor inlet is fitted with a filter. A prefilter is installed in each dryer package. This assures that the dryer efficiency is maintained and that moisture-free clean air is supplied to the
starting air tanks. A wye strainer with a 740-micron particle retention
capability is provided downstream of each starting air tank. In addition, a wye
strainer with a 149-micron particle retention capability is provided upstream of
each air start solenoid valve in each starting air train. The strainers minimize
the possibility of a malfunction of the components in the starting air system by
particle entrapment.
PIPING AND VALVES - Carbon steel piping and valves are installed in the EDESS.
The dryers provide moisture-free air. Manual drains are provided in the starting
air tanks to blow down periodically any accumulated moisture. Therefore, rust
formation in the carbon steel piping and valves is minimized. The strainers
provide an additional safeguard against carryover of any rust particles to the
starting air system components.
9.5.6.2.3 System Operation
Upon initiation of the diesel engine start sequence, the air start solenoid
valves in the redundant starting air trains open to release sufficient air from
the starting air tanks to the engine-mounted air start valves and the engine-
driven air start distributors located on both banks of the engine cylinders.
The air pressure operates the governor servorack booster to open the fuel racks
to ensure adequate fuel during starting. The engine can be started from either
or both banks of cylinders. The engine is maintained in a hot standby condition
to facilitate quick start. The engine start and acceleration to synchronous
speed at rated voltage and frequency is accomplished within 12 seconds. An
engine start failure is annunciated in the control room (see Section 8.3).
The starting air tanks are automatically charged by the compressors. Pressure
switches are installed in each starting air train, at the starting air tanks.
These switches start and stop both compressors to maintain the required pressure
in the tanks. An interconnecting pipe with a normally closed valve is provided
between the two starting air trains, upstream of the starting air tanks, administratively controlled so that either of the two compressors can charge
both starting air tanks. Low starting air pressure is annunciated in the local
panel and in the control room (see Section 8.3). Safety relief valves are
installed on each tank and compressor for overpressurization protection.
9.5-63 Rev. 27 WOLF CREEK A barring gear interlock is provided in each starting air train to prevent the starting of the diesel engine when the barring gear is engaged. Engagement of
the barring gear is annunciated in the local panel and in the control room.
Adequate isolation capabilities are provided in the EDESS to isolate the nonsafety-related portions of the system from the safety-related portion. The
inlet piping to the starting air tanks has nonreturn valves and manual valves to
isolate the tanks from the compressor circuit. Excess flow valves are installed
in the air supply line to the various pressure transmitters on the engine skid.
9.5.6.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases.
SAFETY EVALUATION ONE - The safety-related portion of the EDESS is located in the diesel generators 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 portion of the EDESS is designed to
remain functional after a safe shutdown earthquake. 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 design of the emergency diesel generators provides
for complete redundancy; therefore single failure of the EDESS portion does not
compromise the diesel generators safety function. All vital power can be
supplied from either onsite or offsite power systems, as described in Chapter
8.0. SAFETY
EVALUATION FOUR - The EDESS was initially tested with the program given
in Chapter 14.0. Periodic inservice functional testing is carried out in
accordance with Section 9.5.6.4.
9.5-64 Rev. 19 WOLF CREEK Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI requirements that are appropriate for the EDESS.
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 9.5.6-1 shows that the components meet the design and
fabrication codes given in Section 8.2. All the power supplies and control
functions necessary for safe function of the EDESS are Class IE, as described in
Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - Section 9.5.6.2 describes provisions made to identify
and isolate leakage or malfunction and to assure isolation of the nonsafety-
related portions of the system.
SAFETY EVALUATION SEVEN - The redundant starting air trains in the engine starting system have independent starting air tanks. Each tank has a sufficient capacity to provide at least five diesel engine crank cycles without external
support or assistance. The duration of each crank cycle is 3 seconds or a
minimum of 2 shaft revolutions.
9.5.6.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
The EDESS is tested periodically, along with the complete diesel generator system. This test demonstrates the performance, structural, and leaktight
integrity of all the system components. When the engine is on standby, the
starting air system is normally pressurized up to the air start solenoid valves.
Instrumentation is provided to indicate and alarm loss of air pressure. This
provides an additional means of verification of the structural and leaktight
integrity of the system when the engine is on standby.
The safety-related portions of the EDESS are designed and located (to the extent
practicable) to permit preservice and inservice inspections.
9.5.6.5 Instrumentation Applications The EDESS instrumentation is designed to facilitate automatic operation of the
system and to provide continuous indication of system parameters. Refer to
Section 8.3 for details of instrumentation. Local pressure indicators are provided on the starting air tanks. Pressure indicators are also installed in
the local control panel for each starting air train.
9.5-65 Rev. 19 WOLF CREEK All applicable instruments, controls, sensors, and alarms for the diesel starting air system are shown on Figure 9.5.6-1, Sheets 1 and 2.
The only system function which is alarmed in the diesel generator room is low
air system pressure. This alarm also generates a general control room "diesel trouble" alarm. This malfuction does not result in any harmful effects to the
diesel engine.
9.5.7 EMERGENCY
DIESEL ENGINE LUBRICATION SYSTEM
The emergency diesel engine lubrication system (EDELS) provides essential
lubrication and cooling for the components of the emergency diesel engine.
9.5.7.1 Design Bases 9.5.7.1.1 Safety Design Bases
The EDELS, excluding operation of the keepwarm components, is safety related and is required to function following a DBA and to achieve and maintain the plant in
a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The EDELS is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The EDELS remains functional after a SSE and performs
its intended function following the postulated hazards of fire, internal
missiles, or pipe break (GDC-3 & 4). SAFETY DESIGN BASIS THREE - 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 active components are capable of being tested
during plant operation. Provisions are made to allow for the inservice
inspection of components at appropriate times specified in the ASME Boiler and
Pressure Vessel Code, Section XI (GDC-45 & 46).
SAFETY DESIGN BASIS FIVE - To the extent practicable, the EDELS 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 to isolate components or piping is provided to deal with leakage or malfunctions (GDC-44).
9.5-66 Rev. 19 WOLF CREEK SAFETY DESIGN BASIS SEVEN - The EDELS is designed to provide adequate lubrication and cooling for the various moving parts of the engine to permit it
to be operated at continuous nameplate rating for at least 7 days without
replenishing the system.
SAFETY DESIGN BASIS EIGHT - The EDELS is designed to maintain the lubricating
oil in a warm condition when the engine is on standby to facilitate quick
starting, when required.
9.5.7.1.2 Power Generation Design Basis
The EDELS has no power generation design basis.
9.5.7.2 System Description 9.5.7.2.1 General Description
The EDELS is shown schematically in Figure 9.5.7-1. Each diesel engine is furnished with an independent lubrication system. The EDELS consists of two
separate systems - the main oil system and the rocker oil system.
The main oil system supplies lubricating oil to the main bearings, pistons, camshaft bearings, cam followers, fuel injection pumps, camshaft, and accessory
drive gears. The system consists of an engine-driven main oil pump, oil cooler, electric motor-driven prelube/keepwarm pump, keep-warm electric heater, auxiliary lubricating oil makeup tank, bypass filter, duplex full-flow strainer, piping, valves, controls, and instrumentation.
During engine operation, the engine-driven pump draws oil from the engine sump and delivers it through the oil cooler and strainer to the main engine oil
header. The header supplies oil under pressure to lubricate and cool various
components. After lubrication, the oil flows back to the sump through a return
header.
The capacity of the engine oil sump provides a sufficient volume of oil such that the emergency diesel engine and generator can operate at nameplate continuous rating for at least 7 days without replenishing the oil in the sump. The design specification lube oil consumption at continuous load rating is approximately 60 gallons per day. The engine sump is considered empty at 300 gallons. At the dipstick full mark the sump contains 1200 gallons. This provides 900 gallons of consumable oil, which calculates to 15 days of operation at rated load consuming 60 gallons per day. The low level alarm has a set point corresponding to 963 gallon in the sump (663 consumable, 11 days consuming 60 gallons per day). At the dipstick add oil mark there is 948 gallons in the sump (648 consumable, 10.8 days consuming 60 gallons per day). The engine is also equipped with an automatic oil makeup system. It has set points to add lube oil when the sump reaches 1063 gallons and stops adding when the sump oil reaches 1143 gallons. The lube oil makeup system has an auxiliary tank holding up to 260 gallons of lube oil and can add up to an additional 4 days (at 60 gallons per day consumption) to that provided by the sump alone. On an engine standby status, the electric prelube/keepwarm pump draws oil from
the engine sump and delivers it through an electric heater, filter, and strainer
to the main engine lubricating oil header. The keepwarm system thus serves the
following functions:
9.5-67 Rev. 23 WOLF CREEK
- a. Maintains the oil in the sump in a warm condition to facilitate a quick start.
- b. Prelubricates the essential engine components to
minimize the possibility of oil starvation.
- c. Maintains oil purity by continuous filtration and
straining.
The prelube/keepwarm pump operates when the engine is running in order to
provide a path for the bypass filtration of the oil in the sump.
To protect the crankcase oil from contamination by cooling water and fuel leaks
at the cylinder head upper deck level, the valve rockers are lubricated and
drained by a separate rocker lubricating oil system. The system consists of an engine-driven pump, strainer, reservoir electric motor-driven prelube pump, piping, valves controls and instrumentation. The engine driven pump draws oil
from an engine-mounted reservoir and discharges it under pressure through a
strainer to a header. The header feeds lubricating oil to each cylinder head
rocker assembly. A drain header returns the oil to the reservoir. The electric
motor-driven prelube pump serves as a backup to the engine-driven pump. When the
engine is on standby, the prelube pump is used to lubricate the rocker arm
assembly periodically in accordance with the engine manufacturer's
recommendations.
An auxiliary lubricating oil makeup tank, external to the engine skid, is
provided for a convenient reservoir of makeup oil to the engine sump. The
makeup to the sump is controlled automatically by level switches in an external
level control tank which monitors level in the sump. The volume of oil in the
auxiliary lubricating oil makeup tank is not necessary to meet the requirement for the engine to operate at nameplate continuous rating for at least seven days without replenishing the oil in the sump.
A crankcase vacuum system is provided to maintain a slight negative pressure in
the crankcase. The negative pressure in the crankcase reduces oil leakage out
of the engine. The system consists of an ejector driven by combustion air, an
oil separator, piping, valves, and instrumentation. The ejector discharge is
piped outside the diesel building.
Administrative procedures control the use of lubricating oils and their
containers to prevent inadvertent addition of the wrong type of oils.
Maintenance procedures ensure acceptable lube oil quality through a program of
analysis and actions based on these results.
9.5.7.2.2 Component Description
Codes and standards applicable to the EDELS are listed in Tables 3.2-1 and
9.5.7-1. The safety-related portions of the EDELS are designed and constructed
in accordance with the quality groups listed in Table 3.2-1 and seismic Category
I.
9.5-68 Rev. 19 WOLF CREEK MAIN OIL PUMP - The main oil pump is driven by the engine. It is a positive
displacement, rotary pump. The pump draws oil from the engine sump and delivers
it under pressure to the lubricating oil system.
A suction strainer is provided in the engine sump. A relief valve is built into
the pump for overpressure protection. A failure of this pump constitutes an
engine failure. The pump failure is detected by low lubricating oil pressure or
by a rise in the bearing temperature. See Section 8.3 for the details of
instrumentation.
LUBRICATING OIL COOLER - The lubricating oil cooler is a horizontal shell and
tube-type heat exchanger. The essential service water leaving the jacket water
heat exchanger is circulated through the tubeside of the cooler.
AUXILIARY LUBRICATING OIL MAKEUP TANK - One lubricating oil makeup tank is
provided per engine. The tank is external to the engine skid and is located in
the same room it serves. This tank provides a convenient reservoir of makeup
oil to the engine sump. The volume of oil in the auxiliary lubricating oil
makeup tank is not necessary to meet the requirement for the engine to operate
at nameplate continuous rating for at least seven days without replenishing the
oil in the sump. The tank is horizontal cylindrical type. Connections are
provided on the tank for manual fill, vent, overflow, drain, and level
instrumentation.
KEEPWARM PUMP AND HEATER - The keepwarm pump is a positive displacement pump, driven by an electric motor. A strainer is installed in the suction piping to the pump. A relief valve is provided on the pump to prevent overpressurization.
The keepwarm pump circulates oil through an electric heater, which is
thermostatically controlled when the diesel engine is in stand-by. The heater
is de-energized when the engine runs. The pump and the heater are powered from
an IE source.
ROCKER LUBRICATING OIL PUMPS - The main rocker lubricating oil pump is engine
driven. A backup electric motor-driven pump is also provided. The pumps are of
the positive displacement type. In addition to the relief valves built into the
pumps, a pressure regulator is provided in the system to prevent
overpressurization. The backup pump motor is powered from an IE source.
STRAINERS AND FILTERS - Strainers and filters are provided in the EDELS to
maintain the oil purity at a level required for satisfactory operation of the
diesel and to protect the positive displacement pumps in the system. All the
oil to the engine lubricating oil header is delivered through a duplex basket
type strainer. The keepwarm pump circulates a portion of the oil in the engine
sump continuously through a filter and delivers it to the main lubricating oil
system upstream of the basket strainer. The rocker lubricating oil system has a
strainer between the pump and the engine oil header.
The main lubricating oil and rocker arm lube oil strainers are a full flow, removable, basket type, duplex strainers. The strainers have a 30-micron
nominal particle retention capability.
9.5-69 Rev. 27 WOLF CREEK The filters are a cartridge type made of cellulose with 5-micron nominal particle retention capability. Instrumentation is provided to alarm on a high
pressure differential across the filters and strainers (see Sections 8.3 and
9.5.7.5 for details).
PIPING AND VALVES - All piping in the EDELS is carbon steel. Due to the
manufacturer's service and design experience the flex connections used in the
EDELS are a nonmetallic type, are designed and constructed to manufacturer's
standards, and have proven reliable for the intended service. The oil flow to
the lube oil cooler is controlled by a self-contained thermostatic valve. The
lubricating oil makeup to the engine sump is controlled automatically by a
solenoid-operated valve that is actuated by the level switches in the sump.
Section 9.2.1 describes the piping and valves associated with the essential
service water system. 9.5.7.2.3 System Operation
On engine standby, the keepwarm pump draws oil from the engine sump and delivers
it to the engine lubricating oil header through an electric heater, a bypass
filter, and the main lube oil strainer. The heater is controlled by a
temperature switch. A failure in the keepwarm system will result in a lowering
of the sump oil temperature. As described in Section 8.3, this condition is
monitored and alarmed locally and in the control room.
During diesel engine operation, the engine-driven main oil pump draws oil from
the engine sump and delivers the oil to the engine lubricating oil header
through the lube oil cooler and the main lube oil strainer. The oil header
supplies oil under pressure to the engine components requiring lubrication and
cooling. The oil then drains back to the sump. Essential service water is used to cool the lubricating oil at the cooler. The
oil flow to the shell side of the cooler is modulated by a self-contained
thermostatic valve, so that the temperature differential between the oil inlet
and outlet to the engine remains at the design minimum. The thermostatic valve
bypasses the cooler when the engine is started so that the lubricating oil is
brought to normal operating temperature rapidly. The valve is designed to
permit the full volume of oil flow through the engine.
During normal plant operation, the service water system through the essential
service water system piping provides the cooling water to the cooler. When the
diesel is started during an emergency, the essential service water system
supplies the cooling
9.5-70 Rev. 19 WOLF CREEK water to the cooler. No cooling is available for the time required to bring the essential service water pumps into service with the power from the diesel
generators. As explained in Section 9.5.5.2.3, this time lag is less than a
minute. The diesel engines are designed to operate at nameplate continuous
rating without cooling water for 3 minutes.
The oil level in the engine sump is maintained automatically by the auxiliary
makeup tank. Instrumentation that senses the level in the engine sump controls
a solenoid valve installed in the inlet piping from the makeup tank. A manual
bypass around the solenoid valve is also provided.
The WCGS emergency diesel generator includes an electric motor-driven
prelube/keepwarm pump as an integral part of the lube oil system. This pump
circulates lube oil from the engine crankcase through a keepwarm heater and a
filter, then into the main lube oil system, through a strainer, and into the engine header. During engine standby, this system provides continuous prelubrication and filtering of the oil charge at keepwarm temperature. During
engine operation, this system is used for continuous filtration of the oil
charge. Additionally, the engine includes a separate rocker arm lubrication
system.
This system includes an electric motor-driven prelubrication pump, which is
manually operated and is intended to be used prior to test starts. The rocker
arm prelube pump is manually started from the Engine Gauge Panel. The pump is
operated once every week for a period of 5 to 30 minutes. After operating for
the preset time period the pump automatically shuts off. It is not considered
detrimental by the engine manufacturer for the rocker arms to operate with
reduced oil pressure for the short period of time during which the engine is
coming up to speed in an emergency start situation.
The rocker lube oil system is employed in all Colt-Pielstick diesel engines. This is true whether the engine is in maritime, commercial, or nuclear service.
The vendor (Colt) has stated that, based on both his extensive shop testing and
operational service of the Colt-Pielstick diesel engine, no cooling of the
rocker lube oil is required. Additionally, since the system is not considered
vital to emergency startup of the engine, a keepwarm feature is not provided.
The diesel generator building is maintained at a minimum of 60 F, which is
sufficient to prevent excessive cooling of the lube oil. Temporary temperature
measuring instrumentation was provided on the rocker lube oil system during
startup testing at Callaway Plant to confirm proper operation. Refer to SLNRC
84-0022 dated February 2, 1984.
9.5-71 Rev. 19 WOLF CREEK Colt-Pielstick diesel engines are installed at two operational nuclear power plants, Virgil Summer Station 1 and Farley Unit 1.
The components discussed above are shown in USAR Figure 9.5.7-1, Sheets 1 and 2.
During engine operation, the rocker assembly is lubricated by the engine-driven
rocker lubricating oil pump. The pump draws oil from a reservoir mounted on the
engine and delivers this oil through a strainer to an oil header. The header
distributes the oil to the rocker assembly. A return header is provided to drain
the oil to the reservoir. An electric motor-driven pump is provided as a backup
to the engine-driven pump. During engine standby, the backup pump is used to
prelubricate the rocker assembly at regular intervals in accordance with the
engine manufacturer's recommendations. The reservoir is filled with oil from
the engine main oil header. A float valve is installed to control the oil level
in the reservoir. The full flow strainer and the bypass filter in the keepwarm system maintain the
required oil purity. Fill connections are installed in the makeup tank and the
engine sump. The system is filled, using an offskid portable pump. A temporary
strainer is included in the portable pump package during filling operations. The
sump can also be filled, using the fill connection provided in the keepwarm
system. When this connection is used, the oil is circulated through the bypass
filter and the main oil strainer by an offskid portable pump before reaching the
sump. The quality of the oil in the sump and the makeup tank can be checked, if
required, by withdrawing samples from various points in the system.
During engine operation, a portion of the combustion air is used to drive an
ejector. The ejector is designed to maintain a negative pressure in the
crankcase. An oil separator is provided to ensure that the ejector discharge is
oil free. Instrumentation is provided to alarm on increasing crankcase pressure and to shut down the engine automatically when the pressure exceeds a design maximum. See Section 8.3 for the instrumentation details.
In addition, explosion relief doors are provided to safeguard against any sudden
pressure surges within the crankcase. The explosion relief doors are designed
to relieve the vapors from the crankcase and prevent the entry of outside air
into the crankcase.
9.5-72 Rev. 19 WOLF CREEK Excessive leakage in the main oil system decreases the system pressure and, as described in Section 8.3, the engine automatically shuts down. The keepwarm
system can be isolated from the main oil system. Valves are provided to isolate
one section of the main oil basket strainer from the other for maintenance
purposes. The sump can be isolated from the makeup tank.
There are no mechanical limitations within the EDELS that would restrict the
operation of the diesel engine generator when less than full electrical
generation is required.
9.5.7.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in
Section 9.5.7.1.
SAFETY EVALUATION ONE - The safety-related portions of the EDELS are located in
the diesel generators 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 this building.
SAFETY EVALUATION TWO - The safety-related portions of the EDELS remain
functional after a safe shutdown earthquake. 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 design of the emergency diesel generators provides
for complete redundancy; therefore a single failure of the EDELS portion does
not compromise the diesel generators' safety function. All vital power can be
supplied from either the onsite or offsite power systems, as described in
Chapter 8.0.
SAFETY EVALUATION FOUR - The EDELS is initially tested with the program given in
Chapter 14.0. Periodic inservice functional testing is done in accordance with
Section 9.5.7.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the EDELS.
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 9.5.7-1 shows that the component meets the design and
fabrication
9.5-73 Rev. 19 WOLF CREEK codes given in Section 3.2. All the power supplies and control functions necessary for safe function of the EDELS are Class IE, as described in Chapters
7.0 and 8.0.
SAFETY EVALUATION SIX - Section 9.5.7.2 describes provisions made to isolate leakage or malfunction of the system components.
SAFETY EVALUATION SEVEN - The EDELS components provide adequate lubrication and
cooling for the various moving parts of the emergency diesel engine to permit
its operation at nameplate continuous rating for at least 7 days without oil
replenishment from external sources.
SAFETY EVALUATION EIGHT - A keepwarm system is provided in the EDELS to maintain
the lubricating oil temperature in a warm condition when the engine is on
standby. This facilitates a quick engine start. 9.5.7.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
The EDELS is tested periodically along with the complete diesel generator system. This test demonstrates the performance and structural and leaktight
integrity of all the system components.
The safety-related portions of the EDELS are designed and located (to the extent
practicable) to permit preservice and inservice inspection.
9.5.7.5 Instrumentation Applications The EDELS instrumentation is designed to permit automatic operation and to
provide continuous indication of the system parameters. Refer to Section 8.3
for a list of annunciators and engine trip functions associated with the EDELS.
All appropriate instruments, controls, sensors, and alarms for the diesel engine
lube oil system are shown on USAR Figure 9.5.7-1, Sheets 1 and 2.
Those lube oil temperatures, pressures, and levels which alarm locally and
result in a control room "diesel trouble" light and alarm are high lube oil
temperature from engine, high lube oil strainer differential pressure, low lube
oil pressure to engine, low lube oil sump temperature, high lube oil filter
differential pressure, lube oil level control tank high level and low level, low
lube oil pressure to rocker arms, rocker lube oil strainer high differential pressure, and rocker lube oil reservoir high level.
9.5-74 Rev. 19 WOLF CREEK In addition to the local and control room alarms for low lube oil pressure to the engine, operation of low lube oil pressure switches in a modified two of
four logic (as discussed in Section 8.3 and depicted on Figure 8.3-5) initiates
automatic shutdown of the engine. None of the other malfunctions shutdown the
engine or result in any effects which require immediate operator action.
Station operating procedures give the operators guidance for responding to these
alarms.
Local indication is provided for lube oil temperature to and from the lube oil
cooler and to and from the engine, lube oil strainer differential pressure, lube
oil pressure to engine, auxiliary lube oil tank level, lube oil filter differential pressure, lube oil level control tank level, rocker lube oil
strainer differential pressure, and lube oil pressure to rocker arms.
Table 9.5.7-2 lists the indicators provided for the various system parameters.
9.5.8 EMERGENCY
DIESEL ENGINE COMBUSTION AIR INTAKE AND
EXHAUST SYSTEM
The emergency diesel engine combustion air intake and exhaust system (EDECAIES)
supplies combustion air of suitable quality to the diesel engines and exhausts
the combustion products from the diesel engine to the atmosphere.
9.5.8.1 Design Bases
9.5.8.1.1 Safety Design Bases
The EDECAIES is safety related and is required to function following a DBA and
to achieve and maintain the plant in a safe shutdown condition.
SAFETY DESIGN BASIS ONE - The EDECAIES is protected from the effects of natural
phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external
missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The EDECAIES remains functional after a SSE and
performs its intended function following the postulated hazards of fire, internal missiles, or pipe break (GDC-3 and 4).
SAFETY DESIGN BASIS THREE - Provisions are made to allow for inservice
inspection of components at appropriate times specified in the ASME Boiler and
Pressure Vessel Code, Section XI.
SAFETY DESIGN BASIS FOUR - To the extent practicable, the EDECAIES 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.
9.5-75 Rev. 27 WOLF CREEK SAFETY DESIGN BASIS FIVE - The EDECAIES is designed to supply combustion air to the diesel engines and to exhaust to the atmosphere the products of combustion
so that the diesel generator can be operated continuously at nameplate rating.
9.5.8.1.2 Power Generation Design Bases
The EDECAIES has no power generation design basis.
9.5.8.2 System Description 9.5.8.2.1 General Description
The EDECAIES is shown in Figure 9.5.6-1. Each emergency diesel engine has its own combustion air intake and exhaust system. The combustion air intake system
for each engine consists of intake filters, intake silencers, and piping.
Separate combustion air intake manifolds are provided for the right and left
banks of the cylinders. Combustion air is supplied to each manifold through an
intake filter and silencer. The intake system uses the air in the diesel
generator room for combustion. The air intake system is located within the
diesel generator building and, as such, is not subject to adverse weather
conditions which could potentially block the air intake system. The diesel
generators building ventilation system serves as the source of makeup air which
is used for combustion air by the diesel engine. See Section 9.4.7 for a description of the ventilation system.
A portion of the combustion air from one of the engine combustion air manifolds
is used to drive an ejector to maintain a negative pressure in the engine
crankcase (refer to Section 9.5.7 for details).
The exhaust system for each engine consists of an exhaust silencer and piping.
The products of combustion gases exhausted by the engine and piped through the
silencer are discharged outside the diesel generators building approximately 50
feet above the roof.
9.5.8.2.2 Component Description
Codes and standards applicable to the EDECAIES are listed in Tables 3.2-1 and
9.5.8-1. The system is designed and constructed in accordance with the following quality group requirements: All piping within the diesel generator rooms is quality group C. The intake filter, silencers, and flexible connector
in the intake piping are not commercially available to quality group C. They
are, therefore, designed and constructed to the manufacturer's standards. The
piping outside the diesel generators building is quality group D. Those
portions of EDECAIES inside the building are seismic Category I, and those
portions located outside the building are nonseismic Category I.
9.5-76 Rev. 19 WOLF CREEK INTAKE FILTER - Oil-bath-type air filters are used in the combustion air intake system. The filters are installed in the same room as the engine they serve.
Mist eliminator pads are installed within the filters to remove any oil mist
from the filtered air. A rain shield is provided over the air inlet to each
filter to minimize water carryover in the event the preaction sprinkler system installed for diesel building fire protection is activated. Water carryover
into the filter does not reduce the filter efficiency. The entrapped water
tends to settle and can be drawn off.
INTAKE AND EXHAUST SILENCERS - Silencers are installed in the intake system to
minimize the noise level within the diesel generator room. A silencer is
installed in the exhaust system to reduce the noise level outside the diesel
generator room. The silencers are the inline type, constructed of carbon steel, and utilize internal baffle arrangements to reduce the level of noise emitted
from the EDECAIES. PIPING - The piping in the EDECAIES is carbon steel. Expansion joints are
strategically located to accommodate the thermal growth of the exhaust piping.
The piping is sized adequately so that the total pressure drop when the engine
is operating at nameplate continuous rating is within the diesel engine
manufacturer's recommendations.
ELECTRICAL EQUIPMENT - All electrical equipment mounted outside of the control
panels for the diesel generator unit are provided with either NEMA 4 or NEMA 12
enclosures. The control panels themselves are of NEMA 12 construction with
filtered ventilation openings.
VENTILATION SYSTEM - The D/G building ventilation system employs no filters and
thus supplies outside air directly to the building. However, the system
operates primarily only when the D/G is operating and, therefore, minimizes the time during which it operates. The ventilation system intake is located on top of the D/G building and, therefore, intakes only those particulates which are
airborne (no ground dust).
9.5.8.2.3 System Operation
During engine standby, normally the minimum temperature in the diesel generator
room is maintained at 60 F. The diesel generator room ventilation system
provides the required combustion air when the engine is operating. As explained
in Section 9.4.7, the ventilation system is designed to provide combustion air
under adverse weather conditions and to perform its safety function, assuming a
failure of an active component.
9.5-77 Rev. 19 WOLF CREEK The products of combustion are exhausted to outside the diesel building. Each engine has an independent and separate exhaust stack. The stacks discharge the
exhaust gases approximately 50 feet above the diesel building roof. The exhaust
gases are released approximately 35 feet above the air intake. The intake
louvers are located 65 feet horizontally from the diesel stacks. The distances between the combustion air intake and exhaust release, the high exhaust
discharge velocity, and the buoyancy of the heated exhaust gases are sufficient
to minimize the possibilities of diluting the combustion air with exhaust.
Refer to Sections 6.4, 9.4.1, 9.4.2, and 9.4.3 for a discussion on the ingestion
of exhaust gases into the ventilating system of other buildings. Refer to
Figures 1.2-24 through 1.2-28 for the design features and relative locations of
the intake and exhaust structures.
As shown on Figure 1.2-27, the diesel ventilation intake is located in the
diesel building penthouse, which is approximately 20 feet below the top of the control building. As shown on Figure 1.2-1, the ESF transformers are located to the north of the diesel intakes. The control building intervenes between the
subject ESF transformers (which are at approximately grade elevation) and the
diesel intake. The building wake effect of the control building and the
buoyancy of the smoke and gases would tend to prevent smoke from a potential ESF
transformer fire from entering the intakes. In addition, the intake louvers are
located on the downstream side of the penthouse (from the fire) make smoke
injection even less likely.
The stacks outside the diesel generators building are non-seismic Category I
because the pressure boundary integrity of the stacks is not required for proper
operation of the diesel. However, to preclude blockage of exhaust flow from the
diesel engines due to a seismic event, the design of the stack meets seismic
Category I criteria. The design of the supports for the stacks prevents the
stacks from damaging Category I structures and/or components during a seismic event. The stacks are designed to withstand a pressure differential associated with a tornado and are separated horizontally by approximately 35 feet. With
this separation, it is improbable that a tornado missile can damage both stacks.
The diesel generator exhausts directly to the outside through the diesel exhaust
stacks.
The intake and exhaust louvers are selected on the basis of adverse
environmental conditions. Louver blades are fixed and, hence, cannot become
inoperable due to freezing or icing. They are designed to reduce cascading and
reentrainment of water into the airstream. Design of the louvers is for air
inlet velocities below 500 fpm to prevent moisture carryover.
9.5-78 Rev. 19 WOLF CREEK The size of the intake and exhaust louvers is based on the maximum required quantity of cooling air, 120,000 cfm. The maximum amount of combustion air
required by the diesel engine is approximately 24,000 cfm. Therefore, the
louvers could be as much as 80 percent restricted and still have sufficient open
area to allow operation of the diesel without affecting its performance.
Operation of the diesel building HVAC is not interlocked with the diesel engine
controls. A failure of, or spurious signal from, the HVAC does not prevent
starting of the diesel engine or shutdown of the engine once it is operating.
Refer to Section 2.2 for a discussion of the location of any gases stored on
site. There are no gases stored sufficiently close to the diesel building, such
that a release would impair operation of the diesel engine through ingestion of
the gases into the engine. Water (rain or melted snow) which has entered the
diesel exhaust systems through the exhaust stack accumulates in the bottom of the exhaust silencer. Each silencer is provided with two drain lines, to a loop seal, which allow any accumulation of water to be drained off.
The tornado and missile protection for the diesel generators building
ventilation system is discussed in Section 9.4.7.
The diesel generator may be required to be operated at no load to low loads (less than 20 percent) and rated speed for extended periods. To reduce the
possibility of accumulation of combustion and lube oil products in the exhaust
system at low loads, the engine is operated at 50 percent load for one 1-hour
period during each subsequent 24 hours, starting with the first hour of each 24-
hour period. Above 20 percent load rating, the engine may be run continuously, as required.
This method of operation is based on the manufacturer's recommendations which are now included in the instruction manual. The recommendations are based in part on past experience and in part on a 24-hour no load test conducted on a 12
cylinder Model PC-2.0 engine. That engine successfully accepted a load after 24
hours running at no load. Based on the similarity of the PC 2.5 WCGS engine to
the PC 2.0 and the manufacturer's experience with the operating characteristics
of each engine type, the manufacturer concludes that the PC 2.5 engine responds
more favorably to no load operation than does the PC 2.0 engine. This is
confirmed by a report which shows that a PC 2.5 engine has operated in a power
plant at essentially no load for at least 24 hours.
9.5-79 Rev. 19 WOLF CREEK 9.5.8.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in
Section 9.5.8.1.
SAFETY EVALUATION ONE - Portions of the EDECAIES are located inside the diesel
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 the building. Section 9.5.8.2 describes the
protection provided for the portions of EDECAIES outside the diesel generators
building against the effects of natural phenomena.
SAFETY EVALUATION TWO - The safety-related portions of the EDECAIES are designed
to remain functional after a safe shutdown earthquake. 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 EDECAIES is initially tested with the program
given in Chapter 14.0. Periodic inservice functional testing is done in
accordance with Section 9.5.8.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section XI
requirements that are appropriate for the EDECAIES.
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 9.5.8-1 shows that the component meets the design and
fabrication codes given in Section 3.2. SAFETY EVALUATION FIVE - The EDECAIES components are designed and arranged to
provide combustion air of required quality and to exhaust the combustion
products when the diesel engine is operating continuously at nameplate rating.
9.5.8.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.
The EDECAIES is tested periodically, along with the complete diesel generator system. This test demonstrates the performance, structural, and leaktight
integrity of all the system components.
9.5-80 Rev. 19 WOLF CREEK The safety-related portions of the EDECAIES are designed and located (to the extent practicable) to permit preservice and inservice inspection.
9.5.8.5 Instrumentation Applications
The EDECAIES instrumentation is designed to provide continuous indication of the
system parameters. Refer to Section 8.0 for a list of annunciators for the
EDECAIES. Section 9.4.7.5 describes instrumentation provided for the diesel engine room HVAC. Temperature indicators are installed on the local control
panel for monitoring combustion air temperature, each cylinder exhaust
temperature, cylinder exhaust to the turbocharger, and turbocharger exhaust
temperatures.
All appropriate instruments, controls, sensors, and alarms for the diesel engine
intake air and exhaust systems are shown on USAR Figure 9.5.6-1, Sheets 1 and 2.
The only malfunction which results in an individual local alarm and a general
control room "diesel trouble" alarm is low intake filter suction pressure. Local indication is provided for combustion air temperature, manifold air
pressure, intake filter suction pressure, and exhaust air temperature from each
cylinder, for the left side before and after the turbocharger and for the right
side before and after the turbocharger.
There are no instruments, controls, or sensors in the intake and exhaust air
systems which shut down the engine or when alarmed require immediate operator
action.
9.5.9 AUXILIARY
STEAM SYSTEM
The auxiliary steam system is designed to provide the steam required for plant
heating and processing during plant startup, complete shutdown, and normal
operation. 9.5.9.1 Design Bases 9.5.9.1.1 Safety Design Bases
The auxiliary steam system has no safety function. The location of the equipment and the routing of the piping in the auxiliary steam system are based
on an evaluation of the effects of both high and moderate energy line breaks.
9.5-81 Rev. 19 WOLF CREEK 9.5.9.1.2 Power Generation Design Basis
POWER GENERATION DESIGN BASIS ONE - The auxiliary steam system is designed to
provide the steam required for plant heating.
9.5.9.2 System Description
9.5.9.2.1 General Description
The auxiliary steam system is shown in Figure 9.5.9-1. The system consists of
steam generation equipment, distribution headers, and condensate return
equipment. The auxiliary steam is distributed throughout the plant to the components listed in Table 9.5.9-1.
Normal flow is from the auxiliary steam condensate recovery tank to the
auxiliary steam condensate transfer pumps. The pumps discharge from the
radwaste building to the auxiliary steam condensate recovery and storage tank
located in the auxiliary building. The auxiliary steam deaerator feed pumps
take suction from the auxiliary steam condensate recovery and storage tank and
supply the auxiliary steam deaerator. The auxiliary steam feedwater pumps take
suction from the auxiliary steam deaerator and feed the auxiliary boiler and the
auxiliary steam reboiler, depending on which is in operation. Steam generated by the auxiliary steam system is supplied to the plant heating system and process equipment. The condensate from this equipment is then returned to the
auxiliary steam condensate recovery tank or the auxiliary steam condensate
recovery and storage tank.
Boiler water quality is maintained by blowdown to an atmospheric blowdown tank.
The water levels in the steam boiler, steam reboiler, and the auxiliary steam
deaerator are maintained by automatic controls.
Condensate makeup to the auxiliary steam condensate recovery and storage tank is
from the condensate storage tank and/or the demineralized water storage and
transfer system.
An alarm for high radioactivity levels is provided on the condensate return from the auxiliary steam condensate recovery tank. This alarm automatically cut off the steam supply to the evaporators and shut down the auxiliary steam condensate
transfer pumps.
Section 3.6 provides an evaluation that demonstrates that the pipe routing of
the auxiliary steam system is physically separated from essential systems to the
maximum extent practicable. Protection mechanisms that may be required are
discussed in Section 3.6.
9.5-82 Rev. 19 WOLF CREEK 9.5.9.2.2 Component Description
Codes and standards applicable to the auxiliary steam system are listed in Table
3.2-1. The auxiliary steam system is designed and constructed in accordance
with quality group D specifications.
AUXILIARY STEAM SYSTEM AND BOILER - The auxiliary steam boiler is an oil-fired
package boiler with a rated capacity of 100,000 lb/hr of saturated steam at 125
psig. The design pressure of the steam system is 150 psig, and the system is
protected from overpressure by relieving through the safety valve on the boiler.
AUXILIARY STEAM REBOILER - The auxiliary steam reboiler is a U-tube-type heat
exchanger, using extraction steam or main steam with a rated capacity of 100,000
lb/hr of saturated steam at 125 psig. The design pressure of the steam system
id 150 psig, and the system is protected from overpressure by relieving through the safety valve on the reboiler.
AUXILIARY STEAM DEAERATOR - The deaerator is a 100-percent-capacity tray-type
unit with a vertical deaerating column and a horizontal storage section.
Auxiliary steam is used to preheat the condensate water.
CONDENSATE TANKS - One 600-gallon capacity condensate recovery tank is provided
to handle the condensate return in the radwaste building. One 2,500-gallon-
capacity condensate recovery and storage tank is provided to handle nonradwaste
condensate return and serve as surge capacity for storage of condensate fed to
the deaerator.
PUMPS - Two 100-percent-capacity auxiliary steam feedwater pumps are provided
which can feed either the auxiliary steam boiler or the auxiliary steam
reboiler. The condensate recovery tank and the condensate recovery and storage tank each have two 100-percent-capacity transfer pumps.
9.5.9.2.3 System Operation
During normal operation of the plant, extraction steam from the main turbine is
supplied to the auxiliary steam reboiler which produces auxiliary steam.
Provision is also made for supply main steam to the auxiliary steam reboiler.
During plant shutdown, the oil-fired auxiliary steam boiler is used to generate
auxiliary steam. The switchover from use of the auxiliary steam reboiler to the
oil-fired auxiliary boiler is accomplished manually.
9.5-83 Rev. 19 WOLF CREEK Operational safety features are provided within the system for the protection of plant personnel and equipment. Radiological control is inherent by supplying
steam at higher pressure for those plant processes which have interface with
nuclear process systems.
9.5.9.3 Safety Evaluation The auxiliary steam system has no safety function.
9.5.9.4 Tests and Inspections
Testing of the auxiliary steam system is performed prior to plant operation.
Components of the system are continuously monitored to ensure satisfactory operation.
Periodic operation of all equipment is utilized for additional inspection, checkout, and maintenance.
9.5.9.5 Instrumentation Applications The auxiliary steam system is provided with the necessary controls and
indicators for local or remote monitoring of the operation of the system.
9.5.10 BREATHING AIR SYSTEM
The Breathing Air System (KB) provides clean purified air for use with
respiratory protection and personnel cooling equipment in radiologically
controlled areas of WCGS.
9.5.10.1 Design Basis 9.5.10.1.1 Safety Design Basis
SAFETY DESIGN BASIS ONE - The containment isolation valves in the system are selected, tested and located in accordance with the requirements of 10CFR50, Appendix A, General Design Criteria 54 and 56 and 10CFR50, Appendix J Type C
testing.
SAFETY DESIGN BASIS TWO - Portions of the system piping are designed in
accordance with Seismic Category II/I - Special Scope requirements.
9.5.10.1.2 Power Generation Design Basis
POWER GENERATION DESIGN BASIS ONE - The breathing air system provides clean purified air to various radiologically controlled locations during all modes of operation. However, use of Breathing Air inside containment is limited by the
requirements of containment integrity.
9.5-84 Rev. 19 WOLF CREEK POWER GENERATION DESIGN BASIS TWO - The breathing air system is designed so that there are no interconnections with systems that might contain radioactivity.
POWER GENERATION DESIGN BASIS THREE - The breathing air system shall meet the
air-supply line requirements of Part II of 30 CFR for Type C supplied-air respirators and the applicable minimum grade requirements for Type I gaseous air
set forth in The Compressed Gas Association Commodity Specification for Air, G-
7-1 (Grade D or higher quality).
9.5.10.2 System Description 9.5.10.2.1 General Description
The KB system includes two identical skid mounted compressors. Each compressor skid is complete with compressor, electric motor, service liquid line
accessories, inlet filter, and discharge separator. Air leaving the discharge
separator is directed through a water cooled heat exchanger.
Once the air leaves the compressor skids, it is directed to an air receiver and
then to an air dryer/filter train.
9.5.10.2.2 Component Description
Compressors - The two air compressors are double acting liquid ring motor driven units. Each compressor is sized to deliver approximately 300 scfm at 100 psig discharge pressure. The compressors are sized so that each can supply an
adequate supply of air to meet normal breathing air and cooling air
requirements. The air compressors are non-class IE devices, powered from non-
class IE busses. The two compressors use potable water as a service liquid
within the compressor and central chilled water in the after-cooler.
Air receivers - Compressed air from the outlet of the air compressor after-
cooler flow to a 200 cubic foot galvanized air receiver. The air receiver
provides a reservoir of air during air compressor cycling activities and
provides a limited amount of breathing air following compressor failure to allow
start-up of a standby compressor.
Dryer/Filter Train - The breathing air dryer/filter train consists of a
coalescing prefilter, refrigerated air dryer, charcoal absorber after filter, CO and CO 2 converter, and final particulate filter.
9.5-85 Rev. 19 WOLF CREEK 9.5.10.2.3 System Operation
The compressors may be individually operated and controlled automatically in
response to system demand. Alternately the compressors may be operated in
tandem under a "lead-lag" control scheme which alternates the usage of the
compressors if both are not required.
Local hand switches are provided to permit the operators to start the standby
compressor.
9.5.10.3 Safety Evaluation
Safety evaluations are numbered to correspond to the safety design bases of
9.5.10.1.
Safety Evaluation One - Section 6.2.4 and 6.2.6 provide the safety evaluation
for the system containment isolation arrangement and testability.
Safety Evaluation Two - Section 3.7(B).3.13 (BP-TOP-1) provides the safety
evaluation for the interaction of seismic Category I piping with nonseismic
Category II (II/I) piping.
9.5.10.4 Tests and Inspections
The breathing Air System is in intermittent use, predominantly during refueling
outages. Periodic visual inspections and preventive maintenance are conducted
using normal industry practice. Air quality testing is periodically performed
to ensure compliance with The Compressed Gas Association Commodity Specification
for Air, G-7.1 (Grade D or higher quality).
Inservice inspections are performed for the safety related portions of the
system per the technical requirements of ASME Section IX, as described in
section 6.6. Local leak rate testing in performed in accordance with 10 CFR 50, Appendix J, Type C requirements.
9.5.10.5 Instrumentation Applications
The compressors and associated equipment are provided with local control panels.
Each panel consists of temperature and pressure switches, indicators and
automatic protection devices.
The dryer/filter train is equipped with local pressure indicators. Local
control panel alarms are provided for high differential pressure. The system
has continuous sampling of the discharge air for CO, with local alarm and
shutdown of the compressors should a high concentration exist.
9.5.11 STATION BLACKOUT DIESEL GENERATOR SUPPORT SYSTEMS 9.5.11.1 General Description The Station Blackout Power System (KU) consists of a missile barrier located outside of the protected area (PA) that contains the necessary equipment required to provide reliable power to 4.16 kV Class 1E bus NB001 or NB002 during a station blackout event. This equipment includes three diesel generators (DGs) and one Power Equipment Center (PEC). The PEC includes nine 4.16 kV switchgear sections, four control panels, and one 125 volt DC battery system in addition to other auxiliary equipment required to operation of the system. One control panel is also located in each of the ESF switchgear rooms to allow operation of the station blackout power system without the need for plant personnel to be present in the missile barrier. 9.5-86 Rev. 27 WOLF CREEK Each diesel generator is housed within its own enclosure which contains all equipment necessary to start the DGs. The connection to the NB001 and NB002 buses is made through an underground electrical raceway system. The missile barrier and underground electrical raceway are designed to withstand a tornado with 230 mph wind velocities. The diesel generators are capable of operating over a range of ambient temperatures from -30 F to 120 F. 9.5.11.2 Interfacing Systems The KU system interfaces with the following systems: KC Fire Protection NB Lower Medium Voltage 4.16 kV Class 1E Power PB Lower Medium Voltage 4.16 kV Non-Class 1E Power (Power Block) QF Public Address system RK Plant Annunciator 9.5.11.3 System Testing The Station Blackout Generator System will be periodically tested to ensure continued reliability of the system. As stated in Section 3.3.2, several different modes exist in which the system can be tested. This system will not be connected to the NB bus for testing when the NB bus is energized by normal sources. 9.5.11.4 Component Descriptions 9.5.11.4.1 Station Blackout Diesel Generator Missile Barrier (Z117) The missile barrier is constructed to provide protection from tornado winds and tornado generated missiles.
- a. The foundation slab is designed for dead loads, live loads, tornado wind and missile loads, and seismic loads according to the International Building Code (IBC). The floor elevation is 2000', which is above the probable maximum precipitation flooding elevation for the site. The slab thickness will resist overturning and frost.
- b. The missile barrier is constructed of five steel reinforced cast-in-place concrete walls running in the north-south direction and removable heavy duty steel grating acting as walls on the north and south ends.
9.5-87 Rev. 27 WOLF CREEK The two outermost (east and west) reinforced cast-in-place concrete walls are constructed to withstand tornado winds and tornado generated missiles. Reinforced cast-in-place concrete vestibules are constructed to protect the door entrances from tornado winds and tornado generated missiles. The outer doors are not missile or fire rated. The three innermost reinforced cast-in-place concrete walls are to divide the missile barrier into four bays - three for the diesel generators and one for the PEC. These walls provide fire separation between the bays. The doors between the four missile barrier bays are 3-hour fire rated. UL 924 qualified photo-luminescent exit signs are installed near the doors within the missile barrier structure. Within each diesel generator bay, a partition wall constructed of dampers that are housed by a hollow structural section tube steel frame separates the intake and exhaust sides of the diesel generators.
- c. The north and south ends of the missile barrier are constructed of removable heavy-duty steel grating panels to withstand tornado winds and tornado generated missiles. Each end of the missile barrier heavy duty grating consists of four panels, one for each bay. Metal decking is attached to the north end grating of each DG bay to act as a wind barrier against wind from the north. The heavy duty grating and metal decking ensure proper airflow to the diesel generator enclosures for cooling.
The heavy duty grating also limits the tornado wind speeds inside the missile barrier to less than or equal to 150 mph during a 230 mph tornado event. d. The roof is constructed of removable pre-cast reinforced concrete panels except for the northernmost panel in each diesel generator bay which is constructed of heavy duty grating. 9.5.11.4.2 Station Blackout Diesel Generator Enclosure Each diesel generator bay inside the missile barrier contains a DG enclosure which houses all the equipment necessary to start and run the DGs, i.e., starting battery, jacket water system, radiator, turbochargers, intercoolers, fuel tank, engine/exciter controllers, etc. The provided fuel tanks have sufficient capacity to provide a minimum of 24 hours of run time at full load. The DG enclosures are rated to withstand a wind speed of 150 mph. 9.5-88 Rev. 27 WOLF CREEK TABLE 9.5.1-1 FIRE PROTECTION SYSTEM DESIGN CODES AND STANDARDS Issued By Number Title National Fire Protection Association 10 Installation of (NFPA) Portable Fire
Extinguishers NFPA 11 Foam Extinguishing Systems NFPA 12A Halogenated Extin-guishing Agent Sys-tems - Halon 1301 NFPA 13 Installation of Sprinkler Systems 13A Maintenance of NFPA Sprinkler Systems NFPA 13E Fire Department Operations in Pro-perties Protected
by Sprinkler, Standpipe Systems NFPA 14 Standpipe and Hose Systems NFPA 15 Water Spray Fixed Systems NFPA 20 Centrifugal Fire Pumps NFPA 24 Outside Protection NFPA 30 Flammable and Com-bustible Liquids Code NFPA 37 Combustion Engines and Gas Turbines NFPA 50A Gaseous Hydrogen Systems NFPA 72A Local Protective Signaling Systems Rev. 0 WOLF CREEK TABLE 9.5-1-1 (Sheet 2) Issued By Number Title NFPA 72B Protective Sig-naling System NFPA 72C Remote Station Protective Sig-naling Systems NFPA 72D Proprietary Pro-tective Signaling Systems NFPA 72E Automatic Fire Detectors NFPA 75 Protection of Elec-tronic Computer/
Data Process Equip-ment NFPA 90A Air Conditioning and Ventilation Systems of Other Than Residential Type NFPA 321 Classification of Flammable Liquids NFPA 801 Facilities Handling Radioactive Mater-ial NFPA 803 Fire Protection for Nuclear Power Plants Rev. 0 WOLF CREEK TABLE 9.5.1-2 Fire Protection System Fire Suppression Systems System Area Design Density/Flow RateAutomatic wet- Turbine lube oil 0.30 gpm/sq. ft. over the entire area pipe sprinkler storage tank room system Auxiliary boiler 0.20 gpm/sq. ft. over the entire area room Turbine lube oil 0.30 gpm/sq. ft. over the entire area reservoir room Condenser pit 0.30 gpm/sq. ft. for the most remote (area beneath the 5,000 sq. ft. main condensers) Dry waste compactor 0.15 gpm/sq. ft. over the entire area (radwaste building) Access control 0.15 gpm/sq. ft. for the most remote area (control 1,500 sq. ft. building) Pipe space and 0.15 gpm/sq. ft. for the most remote tank area 1,500 sq. ft. (control building) Cable area above 0.30 gpm/sq. ft. for the most remote access control area 1,000 sq. ft. Vertical cable 0.50 gpm/sq. ft. at the ceiling and chases (auxiliary 0.15 gpm/sq. ft. at the intermediate building) level for the most remote sprinklers Vertical cable 0.50 gpm/sq. ft. with all heads in chases (control the most remote level open building) Aux Feedwater Pipe 0.20 gpm/sq. ft. for the most remote Chase Area 1,500 sq. ft. (Auxiliary Building) Turbine Building 0.15 gpm/sq. ft. for the entire office Outage Office and storage areas Automatic water Hydrogen seal oil 0.30 gpm/sq. ft. spray system unit Main transformer 0.25 gpm/sq. ft. Startup transformer 0.25 gpm/sq. ft. Auxiliary 0.25 gpm/sq. ft. transformer Rev. 19 WOLF CREEK TABLE 9.5.1-2 (Sheet 2) Station service 0.25 gpm/sq. ft. transformer ESF transformer 0.25 gpm/sq. ft. Manual water Auxiliary feedwater 0.30 gpm/sq. ft. spray system pump (turbine driven) Steam generator 0.30 gpm/sq. ft. feed pump Automatic pre- Fuel building rail- 0.30 gpm/sq. ft. action sprinkler road bay system Lower cable 0.30 gpm/sq. ft. of floor spreading room area for the most remote 3,000 sq. ft. Upper cable 0.30 gpm/sq. ft. of floor spreading room area for the most remote 3,000 sq. ft. Cable trays at El. 0.30 gpm/sq. ft. of associated 1974'-0", 2000'-0", floor area for the most and 2026'-0" of the remote 3,000 sq. ft. of tray auxiliary building surface Diesel generator 0.30 gpm/sq. ft. for entire Rooms space with the hydraulically most demanding level of sprinklers operating Area below turbine 0.30 gpm/sq. ft. for the generator operating most remote 5,000 sq. ft. floor and mezzanine floor Turbine generator 0.30 gpm/sq. ft. for all sprinklers operating around the two most hydraulically remote adjacent bearings Manual preaction North cable pene- 0.30 gpm/sq. ft. of floor sprinkler system tration inside the area for the most remote containment 1,000 sq. ft. South cable pene- 0.30 gpm/sq. ft. of floor tration inside the area for the most remote containment 1,000 sq. ft. Halon 1301 ESF switchgear 5 percent minimum for System rooms 10 minutes Nonvital switch- 5 percent minimum for gear and trans- 10 minutes former rooms Rev. 19 WOLF CREEK TABLE 9.5.1-2 (Sheet 3) Switchgear rooms 5 percent minimum for 10 minutes Control cabinet, 5 percent minimum for load center, and MG 10 minutes sets room Electrical pene- 5 percent minimum for tration rooms 10 minutes Control room cable 5 percent minimum for trenches and asso- 10 minutes. (Refer to ciated wall chases Sections 9.5.1.2.2.1
and 9.5.1.2.3) Rev. 8 WOLF CREEK TABLE 9.5.1-3 FIRE PROTECTION SYSTEM TECHNICAL REQUIREMENTS Deleted Rev. 16 WOLF CREEK TABLE 9.5.2-1 COMMUNICATION SYSTEMS IN PLANT AREAS REQUIRED TO BE MANNED FOR POST FIRE SAFE SHUTDOWN FOLLOWING CONTROL ROOM EVACUATION Area Major Equipment Available Communications Reactor trip Reactor trip PA switchgear switchgear Maintenance jacks room 1403 SB102A and B Class 1E Class 1E buses PA switchgear NB01 and 2; MCCs Maintenance jacks
rooms 3301 NG01A and 2A; load Telephone (NB02 side only) and 3302 centers NG02 and 04 Penetration Class 1E motor PA rooms 1409 and control centers Maintenance jacks
1410 NG01B, 2B, 2T, Telephone and 4T Auxiliary Class 1E air con- PA (South room only) building HVAC ditioning unit Maintenance jack
rooms 1501 SGK04B; MCCs and 1512 NG03C and 4C Diesel Diesel generator PA generator NE02; MCC NG04D Maintenance jacks
room 5201 Auxiliary RP118A, B shutdown Maintenance jacks shutdown panels PA Approx. panel room Telephone 8 feet
1413 outside the room Rev. 19 WOLF CREEK Table 9.5.3-1 Deleted Rev.9 WOLF CREEK TABLE 9.5.4-1 EMERGENCY DIESEL ENGINE FUEL OIL STORAGE AND TRANSFER SYSTEM
COMPONENT DATA Underground Storage Tanks Quantity 2
Type Horizontal, cylindrical
Capacity gallons, (each) 100,000
Operating pressure/temperature, Atm/35 to 80
psig/F
Design pressure/temperature, Atm/120
psig/F
Material Carbon steel
Code ASME Section III, Class 3
Seismic design Category I Fuel Oil Transfer Pumps Quantity 2
Type Horizontal, centrifugal
submersible
Capacity, gpm (each) >15 TDH, ft 75
NPSH required/available Flooded suction
Material
Case Type 316 stainless steel
Impeller Type 316 stainless steel
Shaft Type 316 stainless steel
Design Code ASME Section III, Class 3
Driver
Type Canned electric motor
Kilowatts, kW 2.5 with 1.15 service
factor
Power supply 460 V, 60 Hz, 3-phase
Class IE
Seismic design Category I Emergency Fuel Oil Day Tanks Quantity 2
Type Horizontal, cylindrical
Capacity gallons, (each) 621 (includes volume in standpipe) Operating pressure/temperature, Atm/100
psig/F
Design pressure/temperature, 5/150
psig/F
Material Carbon steel
Code ASME Section III, Class 3
Seismic design Category I Rev. 6 WOLF CREEK TABLE 9.5.4-1 (Sheet 2) Piping, Fittings, and Valves Design pressure, psig 150
Design temperature, F 100
Material Carbon steel
Design Code
Safety-related portion ASME Section III, Class 3
Nonsafety-related portion ANSI B31.1 Rev. 0 WOLF CREEK TABLE 9.5.4-2 EMERGENCY DIESEL ENGINE FUEL OIL STORAGE AND TRANSFER SYSTEM INDICATING AND ALARM DEVICES Indication Alarm________ Indication Alarm Control Room Local Control Room Local Storage tank level Yes Yes Yes No Day tank level Yes Yes Yes No Transfer pump motor-running lights Yes Yes Yes No Fuel oil pressure No Yes Yes* Yes Strainer/filter dif-ferential pressure No Yes Yes* Yes
- Common alarm in the control room for local annunciation.
Rev. 0 WOLF CREEK TABLE 9.5.4-3 COMPARISON OF THE DESIGN TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.137, REVISION 0 DATED JANUARY 1978, "FUEL-OIL SYSTEMS FOR STANDBY DIESEL GENERATORS" Regulatory Guide 1.137 Position WCGS
- 1. The requirements for the design of fuel-oil systems for diesel generators that provide standby electrical power for a nuclear power plant that are included in ANSI N195-1976, "Fuel Oil Systems for Standby Diesel Generators,"1 provide a method acceptable to the NRC
staff for complying with the pertinent requirements of General Design Criterion 17 of
Appendix A to 10 CFR Part 50, subject to the following:
- a. Throughout ANSI N195- a. No response is 1976, other documents required required.
to be included as part of the
standard are either identified at the point of reference or described in Section 7.4, "Applicable Codes, Standards, and Regulations," or Section 11, "References," of the standard.
The specific acceptability of these listed documents has been or will be addressed separately in other regulatory guides or in Commission regulations, where appropriate.
- b. Section 1, "Scope," of b. Complies.
ANSI N195-1976 states that the standard provides the design requirements for the fuel-oil Rev. 0 WOLF CREEK TABLE 9.5.4-3 (Sheet 2) Regulatory Guide 1.137 Position WCGS system for standby diesel gen-erators and that it sets forth other specific design require-ments such as safety class, materials, physical arrangement, and applicable codes and regu-lations. The standard does not
specifically address quality assurance, and in this regard ANSI N195-1976 should be used in
conjunction with Regulatory Guide 1.28, "Quality Assurance Program Requirements (Design and
Construction)," which endorses ANSI N45.2-1971, "Quality Assur-ance Program Requirements for Nuclear Power Plants," for the design, construction, and main-tenance of the fuel-oil system.
- c. Section 5.4, "Calcu- c. Complies with (1).
lation of Fuel Oil Storage Re-
quirements," of the standard sets forth two methods for the calculation of fuel-oil storage requirements. These two methods are (1) calculations based on assuming the diesel generator
operates continuously for 7 days at its rated capacity, and (2) calculations based on the time-dependent loads of the diesel generator. For the time-dependent load method, the mini-mum required capacity should include the capacity to power the engineered safety features.
Applications that use the time-dependent load method to calcu-late fuel-oil storage require-
ments will be reviewed on a case-by-case basis along with the calculations.
- d. Section 7.3, "Physical d. Complies.
Arrangement," of ANSI N195-1976
states that "the location of the day tanks of standby diesel gen- Rev. 0 WOLF CREEK TABLE 9.5.4-3 (Sheet 3) Regulatory Guide 1.137 Position WCGS erators shall be as required by the diesel-engine manufacturer." In addition to this requirement, the day tanks should be located at an elevation to ensure ade-quate net positive suction head at the engine fuel pumps at all
times. e. Section 7.3 of ANSI e. Complies. N195-1976 states that the ar-rangement of the fuel-oil system "shall provide for
inservice inspection and testing in accordance with ASME Boiler and Pressure Vessel Code, Section XI, 'Rules for In-Service Inspection of Nuclear Power Plant
Components.'"2 Although Section XI of the ASME Boiler and Pressure Vessel Code does
not specify whether its provisions apply to fuel-oil systems, they should be applied for the inservice inspection and testing program for those portions of the fuel-oil
systems for standby diesel gen-erators that are designed to Section III, Subsection ND of the Code.
- f. Section 7.3 of ANSI f. Complies.
N195-1976 states that adequate heating shall be provided for the fuel-oil system. Assurance
should be provided that fuel oil can be supplied and ignited at all times under the most severe
environmental conditions expected at the facility. This may be accomplished by use of an oil with a "Cloud Point" lower than the 3-hour minimum soak temperature (Ref. 1) expected at
the site during the seasonal periods in which the oil is to Rev. 0 WOLF CREEK TABLE 9.5.4-3 (Sheet 4) Regulatory Guide 1.137 Position WCGS be used, and/or by maintenance of the onsite fuel oil above the "Cloud Point" temperature.
- g. Section 7.5, "Other Re- g. Complies.
quirements," of the standard states that "protection against external and internal corrosion shall be
provided" for the fuel-oil system. To amplify this requirement for buried supply tanks not located
within a vault and other buried portions of the system, a water-proof protective coating and an
impressed current-type cathodic protection system should be pro-vided in accordance with NACE Standard RP-01-69 (1972 Revision), "Recommended Practice-Control of External Corrosion on Underground
or Submerged Metallic Piping Systems."3 In addition, the impressed current-type cathodic
protection system should be designed to prevent the ignition of combustible vapors or fuel oil present in the fuel-oil systems for standby diesel generators.
- h. Section 7.5 of the h. Complies. See standard includes requirements Section 9.5.1.
for fire protection for the diesel-
generator fuel-oil system. The requirements of Section 7.5 are not considered a part of this regulatory guide since this sub-ject is addressed separately in more detail in other NRC docu-
ments. Thus a commitment to follow this regulatory guide does not imply a commitment to follow
the requirements of Section 7.5 concerning fire protection.
- 2. Appendix B to ANSI N195-1976 should be used as a basis for a Rev. 0 WOLF CREEK TABLE 9.5.4-3 (Sheet 5)
Regulatory Guide 1.137 Position WCGS program to ensure the initial and continuing quality of fuel oil as supplemented by the following:
- a. The oil stored in the fuel- a. Complies. Diesel fuel oil used oil supply tank, and the oil to be for filling and refilling the used for filling and refilling the supply tank, complies with ASTM
supply tank, complies with ASTM D975-81 per Technical Specifi-D975-81, "Standard Specification cation requirements. for Diesel Fuel Oils,"4 or the
requirements of the diesel-generator manufacturer, if they are more restrictive, as well as
the fuel-oil total insolubles level specified in Appendix B of the standard and the "Cloud Point" requirements given in Regulatory Position C.2.b. Fuel oil contained in the supply tank not
meeting these requirements should be replaced in a short period of time (about a week).
- b. Prior to adding new fuel oil b. Prior to adding new fuel to the supply tanks, tests for the to the supply tanks, it will
following properties should be con- be sampled in accordance with ducted: Technical Specification require-ments. See Technical Specifi-
(1) Specific or API gravity cations.
(2) Cloud Point
(3) Water and Sediment
(4) 90% Distillation Temperature The fuel oil complies with ASTM D975-74 for the latter two analyses.
The "Cloud Point" should be less than or equal to the 3-hour mini-
mum soak temperature, or the mini-mum temperature at which the fuel oil will be maintained during the
period of time that it will be in storage. Analysis of the other properties of the fuel oil listed
in ASTM D975-74 should be completed within 2 weeks of the transfer. Rev. 11 WOLF CREEK TABLE 9.5.4-3 (Sheet 6) Regulatory Guide 1.137 Position WCGS c. The periodic sampling pro- c. Periodic sampling of fuel oil cedure for the fuel oil should be will be in accordance with in accordance with ASTM D270-1975 Technical Specifications "Standard Method of Sampling Petroleum and Petroleum Products."5 d. Accumulated condensate should d. Complies. be removed from storage tanks on: (1) a quarterly basis; (2) a monthly basis when it is suspected or known that the ground
water table is equal to or higher than the bottom of buried storage tanks; and (3) one day after the addition of new fuel. e. Day tanks and integral tanks e. Complies. Present experience should be checked for water monthly, indicates that removal of water as a minimum, and after each opera- from the day tank every 31 days tion of the diesel where the period provides sufficient protection of operation was 1 hour or longer. to the EDGs. Any accumulated water should be removed immediately. If it is suspected that water has entered the suction piping from the day or integral tank, the entire fuel-oil system between the day or integral tank and the injectors should be
flushed. f. As a minimum, the fuel oil f. Complies by removing the fuel oil stored in the supply tanks should from the storage tanks, removing be removed, the accumulated accumulated sediment, and cleaning sediment removed, and the tanks of the tanks once per 10 years. cleaned in order to perform the (See 9.5.4.2.1) ASME Section XI, Article IWD- ASME Section XI, IWD-2000 2000, "Examination Require- examinations are not a basis for ments," at the required 10-year draining and cleaning of the tanks. intervals. To preclude the ISI examinations are conducted introduction of surfactants in in accordance with the ISI Program.
the fuel system, this cleaning Rev. 17 WOLF CREEK TABLE 9.5.4-3 (Sheet 7) Regulatory Guide 1.137 Position WCGS should be accomplished using sodium hypochlorite solutions or its equivalent rather than soap or detergents.
- g. Assuming an unlikely event g. Complies.
should occur that would require Refer to 9.5.4.2.3. replenishment of fuel oil without
the interruption of operation of the diesel generators, the method of adding additional fuel oil should be
such as to minimize the creation of turbulence of the accumulated residual sediment in the bottom of
the supply tank since stirring up this sediment during the addition of acceptable new incoming fuel has the
potential of causing the overall quality of the fuel oil in the storage tank to become unacceptable.
- h. Cathodic protection sur- h. Complies.
veillance should be conducted
according to the following procedures:
(1) At intervals not ex-ceeding 12 months, tests should be conducted on each underground cathodic protection system to de-termine whether the protection is adequate. (2) The test leads required for cathodic protection should be main-
tained in such a condition that elec-trical measurements can be obtained to ensure the system is adequately
protected.
(3) At intervals not exceeding 2 months, each of the cathodic protection rectifiers should be inspected. Rev. 0 WOLF CREEK TABLE 9.5.4-3 (Sheet 8)
Regulatory Guide 1.137 Position WCGS (4) Records of each inspection and test should be maintained over the life of the facility, to assist in evaluating the extent of degradation of the corrosion protection systems. NOTES: 1. Copies may be obtained from the American Nuclear Society. 555 North Kensington Avenue, La Grange Park, Illinois 60525.
- 2. Copies may be obtained from the American Society of Mechanical Engineers. United Engineering Center. 345 East 47th Street, New York, N.Y. 10017.
- 3. Copies may be obtained from the National Association of Corrosion Engineers, 2400 West Loop South, Houston, Texas 77027.
- 4. Also designated ANSI Z11.205-1975. Copies may be obtained from the American National Standards Institute, 1430 Broadway, New
York, N.Y. 10018.
- 5. Also designated ANSI Z11.33-1976. Copies may be obtained from the American National Standards Institute, 1430 Broadway, New York, N.Y. 10018. Rev. 0 WOLF CREEK TABLE 9.5.5-1 EMERGENCY DIESEL ENGINE COOLING WATER SYSTEM COMPONENT DATA (PER DIESEL ENGINE)
Jacket Cooling Water Pump
Quantity 1 Type Horizontal centrifugal Capacity, gpm 1,054 TDH, ft 128 Design code MS Driver Engine driven Seismic design Category I
Jacket Coolant Keepwarm Pump
Quantity 1 Type Horizontal centrifugal Capacity, gpm 50 TDH, ft 20 Design code ASME Section III, Class 3
Driver Type Electric motor Horsepower, hp 0.75 Rpm 1,800 Power supply 460 V, 60 Hz, 3-phase Class IE Design Code NEMA Seismic design Category I
Jacket Cooling Water Heat Exchanger
Quantity 1 Type Horizontal shell and tube Design duty, Btu/hr 7.24 x 10 6 Seismic design Category I Codes and standards ASME Section III, Class 3 TEMA R, ASME Section VIII Div 1 Tube side: Fluid Service water/essential service water Temperature in/out, F 103.2/115.2 Flowrate, gpm 1,200 Design pressure, psig 200 Design temperature, F 200
Rev. 25 WOLF CREEK TABLE 9.5.5-1 (Sheet 2) Material: Tubes UNS N08367 class 2 Tubesheet 5A-240 Type 304/L or 5A-182 Gr. F304/L Shell side: Fluid Jacket cooling water Temperature in/out, F 180/165.8 Flowrate, gpm 1,050 Design pressure, psig 150 Design temperature, F 200 Material Carbon steel Jacket Coolant Keepwarm Heater Quantity 1 Type Electric Design rating, kW 42 Power supply 480 V, 60 Hz, 3 phase Class IE Code (pressure boundary) ASME Section III, Class 3 Seismic design Category I Expansion Tank Quantity (per engine) 1 Type Horizontal, cylindrical Capacity, gallon 100 Operating pressure/temperature, Atm./122
psig/F Material Carbon steel Code ASME Section III, Class 3 Seismic design Category I Intercooler Cooling Water Pump Quantity (per engine) 1 Type Horizontal centrifugal Capacity, gpm 1,063 TDH, ft 126 Design code MS Driver Engine driven Seismic design Category I Intercooler Heat Exchanger Quantity (per engine) 1 Type Horizontal shell and tube Design duty, Btu/hr 4.85 x 10 6 Seismic design Category I Codes and standards ASME Section III, Class 3, TEMA R Rev. 25 WOLF CREEK TABLE 9.5.5-1 (Sheet 3)
Tubeside:
Fluid Service water/essential service water Temperature in/out, F 95/105 Flowrate, gpm 1,200 Design pressure, psig 200 Design temperature, F 200
Material: Tubes SB-676 (UNS N08367 CLASS 2) Tubesheet SA-240 Type 304L
Shell side:
Fluid Intercooler cooling water Temperature in/out, F 121/110 Flowrate, gpm 1,050 Design pressure, psig 150 Design temperature, F 200 Material Carbon steel
Piping, Fittings, and Valves
Material Carbon and stainless steel Design code Safety-related portion ASME Section III, Class 3 (except flexible con-
nectors)
Flexible connectors MS
Seismic design Category I
- As noted in the NRC Safety Evaluation for License Amendment 134.
WCNOC has stated that the tube plugging limits for the Emergency
Diesel Generator Heat Exchangers will be based on a service water inlet temperature of 96 F at the inlet to the intercooler heat exchangers and 104.2ºF at the inlet to the jacket water heat exchangers.
Rev. 25 WOLF CREEK TABLE 9.5.6-1
EMERGENCY DIESEL ENGINE STARTING SYSTEM
COMPONENT DATA
Compressors
Quantity (per engine) 2
Type Reciprocating, air cooled
Capacity, scfm 31
Discharge pressure, psig 700
Air temperature leaving cooler, F 160
No. of stages 3
Design code MS
Driver
Type Electric motor
Horsepower, hp 15
Rpm 1,800
Power supply 480 V, 60 Hz, 3 phase
non-1E Seismic design Nonseismic Category I
Dryers Quantity (per engine) 2
Type Membrane, automatic regenerative Capacity, scfm 50 scfm Design pressure, psig 1200 Air inlet temperature, F 122
Dew point of air leaving dryer, F (-)40
Design code MS
Seismic design Nonseismic Category I
Starting Air Tanks
Quantity (per engine) 2
Type Horizontal, cylindrical
Capacity, cu ft 55
Design pressure/temperature
psig/F 670/142
Operating pressure/temperature, psig/F 640/122
Material Carbon steel
Code ASME Section III, Class 3
Seismic design Category I
Rev. 27 WOLF CREEK TABLE 9.5.6-1 (Sheet 2) Piping, Fittings, and Valves (Safety Related) Material Carbon steel Design code ASME Section III, Class 3
Seismic design Category I Piping, Fittings, and Valves (Nonsafety Related)
Material Carbon steel Design code MS Rev. 0 WOLF CREEK TABLE 9.5.7-1 EMERGENCY DIESEL ENGINE LUBRICATION SYSTEM COMPONENT DATA
Main Oil Pump Quantity, (per engine) 1 Type Positive displacement, rotary Capacity, gpm 631 Relief valve set pressure, psig 110-115 Design code MS Driver Engine driven Seismic design Category I
Keepwarm Pump Quantity (per engine) 1 Type Positive displacement, Screw Capacity, gpm 75 Relief valve set pressure, psig 130 Design code (Note 1) Driver Type Electric motor Horsepower, hp 20 Rpm 1800 Power supply 460 V, 60 Hz, 3 phase Class IE Design code NEMA Seismic design Category I
Oil Cooler Quantity, (per engine) 1 Type Horizontal shell and tube Design duty, Btu/hr 2.2 x 10 6 Codes and standards ASME Section III, Class 3 TEMA R Seismic design Category I Tubeside: Fluid Service water/essential service water Temperature in/out, F 118/123 Flowrate, gpm 1,200 Design pressure, psig 200 Design temperature, F 200
Material Tubes Admiralty ASTM Blll Tubesheet Muntz Metal ASTM B171 Shellside: Fluid Lubricating oil Temperature in/out, F 160/141 Flowrate, gpm 630 Design pressure, psig 150 Design temperature, F 200 Material Carbon steel Rev. 27 WOLF CREEK TABLE 9.5.7-1 (Sheet 2) Keepwarm Heater Quantity, (per engine) 1 Type Electric Design rating, kW 24 Power supply 480 V, 60 Hz, 3 phase Class IE Code (pressure boundary) ASME Section III, Class 3 Seismic design Category I Makeup Tank Quantity, (per engine) 1 Type Horizontal, cylindrical Capacity, gallon 300 Operating pressure/ temperature, psig/F Atm./amb. Material Carbon steel Code ASME Section III, Class 3 Seismic design Category I Main Oil Strainer Quantity, (per engine) 1 Type Duplex, removable basket type Flowrate, gpm 700 Particle retention capability 30 micron (nominal) 56 micron (absolute) Design pressure/temperature, psig/F 150/200 Material Screen Stainless steel Housing Carbon steel Code (pressure boundary) ASME Section III, Class 3 Seismic design Category I Bypass Filter Quantity, (per engine) 1 Type Cartridge type, simplex Flowrate, gpm 75 Design pressure temperature, psig/F 150/200 Filtering capacity, microns 5 (nominal) Material Housing Carbon steel Filter Cellulose Code (pressure boundary) ASME Section III, Class 3 Seismic design Category I Rev. 9 WOLF CREEK TABLE 9.5.7-1 (Sheet 3) Rocker Lube Oil Pump Quantity, (per engine) 1 Type Rotary, positive dis-placement Capacity, gpm 2.2 Relief valve set pressure, psig 20 Design code MS Driver Engine driven Seismic design Category I
Rocker Prelube Pump Quantity, (per engine) 1 Type Rotary, positive dis-placement Capacity, gpm 2.2 Relief valve set pressure, psig 20 Design code MS Driver Type Electric motor Horsepower, hp 0.5 Rpm 1,200 Power supply 460 V, 60 Hz, 3 phase Class IE Design code NEMA Seismic design Category I
Rocker Lube Oil System Strainer Quantity, (per engine) 1 Type Duplex, removable basket type Flowrate, gpm 2.5 Design pressure/temperature, 25/200 psig/F Capacity, microns 30/40 microns (nominal/absolute)
Material Housing Carbon steel Screen Stainless Steel Design code MS Seismic design Category I
Piping, Fittings, and Valves Material Carbon steel Design code Safety-related portion ASME Section III, (except flexible connectors) Class 3 Flexible connectors MS Seismic design Category I
Note 1: The component design and reliability has been proven through years of previous service. The standards used in design, manufacture and inspection are the manufacturer's standards developed by manufacturing and testing experience. The design meets seismic category I
requirements and is equivalent to the originally supplied ASME Section III component. Rev. 27 WOLF CREEK TABLE 9.5.7-2
EMERGENCY DIESEL ENGINE LUBRICATION SYSTEM
INDICATING DEVICES
Local Panel Engine Skid
Indication Mounted Mounted
Oil pressure to engine header Yes No
Oil pressure from engine No No
Oil temperature to engine header Yes No
Oil cooler inlet temperature No Yes
Oil cooler outlet temperature No Yes
Main oil strainer differential
pressure No Yes
Bypass filter differential
pressure No Yes
Sump oil level Yes No
Makeup tank oil level Yes No
Rocker oil header pressure Yes No
Rocker oil strainer differential
pressure No Yes
Crankcase pressure Yes Yes
Rev. 27 WOLF CREEK TABLE 9.5.8-1 EMERGENCY DIESEL ENGINE COMBUSTION AIR INTAKE AND EXHAUST SYSTEM COMPONENT DATA Air Intake Filter Quantity (per engine) 2
Type Oil bath
Design flow, cfm
Design pressure/temperature, psig/F
Material Carbon steel
Quantity of oil, gals 59
Code MS
Seismic design Category I Intake Silencer Quantity (per engine) 2
Type Horizontal
Design flow, cfm
Design pressure/temperature, psig/F
Material Carbon steel
Code MS
Seismic design Category I Exhaust Silencer Quantity (per engine) 1
Type Horizontal
Design flow, cfm
Design pressure/temperature, psig/F
Material Carbon steel
Code MS
Seismic design Category I Piping Material Carbon steel
Design code
Inside diesel building ASME Section III, (except flexible connectors in Class 3
the intake piping)
Flexible connectors (intake) MS
Outside the building ANSI B31.1
Seismic design Category I Rev. 0 WOLF CREEK TABLE 9.5.9-1 COMPONENTS SUPPLIED BY AUXILIARY STEAM Referenced Steam Rate Component Section (x 10 4 lb/hr)a. Plant heating heat exchanger 9.4.9 1.966
- b. Recycle evaporator package 9.3.6 0 Note 1
- c. Waste evaporator package 11.2 1.06
- d. Outdoor storage tank heating 6.3, 9.2 0.066
- e. Turbine steam seal system 10.4.3 1.9-3.4
- f. Secondary liquid waste evaporator 10.4.10 0 Note 1
- g. Condenser sparging 10.4.1 2.75
- h. Boric acid batching tank 9.3.4 0.05
- i. Decontamination areas in the
auxiliary building
and containment N/A 0 Note 2 j. Moisture separator
reheater tube
blanketing 10.2 0.2 Note 1: Equipment no longer in service. Note 2: The Decontamination areas in the auxiliary building and containment are no longer in service. Rev. 19 Thispageleftintentionallyblank.Thispageleftintentionallyblank. 19Thispageintentionallyleftblank. 19Thispageintentionallyleftblank. 19Thispageleftintentionallyblank. 19Thispageleftintentionallyblank. WOLF CREEK APPENDIX 9.5A Design Comparison to Regulatory Positions of Regulatory Guide 1.120, Revision 1, Dated November 1977, Titled "Fire Protection Guidelines for Nuclear Power Plants" The basis for compliance to Regulatory Guide 1.120 is the implementation of Appendix A of NRC Branch Technical Position (BTP) APCSB 9.5-1. The following
provides a summary of the compliance with APCSB 9.5-1. 9.5A-0 Rev. 0 WOLF CREEK TABLE 9.5A-1 WCGS FIRE PROTECTION COMPARISON TO APCSB 9.5-1 APPENDIX A APCSB 9.5-1 Appendix A WCGS Positions
A. Overall Requirements of Nuclear Plant Fire Protection Program
- 1. Personnel Responsibility for The President and Chief Executive the overall fire Officer is ultimately responsible protection program for the Fire Protection Program. should be assigned The implementation is delegated to to a designated the Supervisor Fire Protection who person in the upper is responsible for the administration level of management. of the Fire Protection Program. This person should retain ultimate responsibility even though formulation and assurance of program implementa- The President and Chief Executive tion is delegated. Officer and Supervisor Fire Protection Such delegation of are supported by a Fire Protection authority should Engineer. This Fire Protection Engineer be to staff personnel shall be a Professional Grade member of prepared by training the Society of Fire Protection Engineers and experience in (SFPE) or shall have qualifications fire protection and equivalent to the professional grade nuclear plant safety membership requirements of SFPE. to provide a balanced approach in directing the fire protection programs for nuclear power plants. The basic design of the fire protection system, selection of equipment, and coordination of The qualification re- layout with fire area requirements quirements for the fire was the direct responsibility of protection engineer or a licensed senior mechanical consultant who assists engineer assigned full time to the in the design and se- SNUPPS project. lection of equipment, inspects and tests the Additional assistance has been pro-completed physical vided by licensed fire protection, aspects of the system, and graduate fire protection
Rev. 25 WOLF CREEK TABLE 9.5A-1 (Sheet 2) APCSB 9.5-1 Appendix A WCGS develops the fire protection program, and assists in the fire-fighting training for the operating plant should be stated. engineers. The entire fire protection system, including the fire hazards analyses, have been reviewed by a licensed
fire protection engineer. Subsequently, the USAR should discuss the training and the updating provisions such as fire drills provided for maintaining the competence of the station fire-fighting and operating crew, including personnel responsible for maintaining and inspecting the fire protection equipment. The WCGS Fire Protection Program discusses the training and the updating provisions such as fire drills provided for maintaining the competence of the station fire fighting and operating crew, including personnel responsible for
maintaining and inspecting fire protection
equipment. Maintenance of the fire protection system is the responsibility of the Manager Operations and Manager
Maintenance. Fire prevention and fire training is maintained under the direction of the Supervisor Fire Protection who may
delegate certain of these responsibilities. The Fire Protection Staff should be responsible for: a) coordination of building layout and system design with fire
area requirements, including
consideration of potential hazards associated with postulated design basis fires, b) design and maintenance of fire detection, suppression, and
extinguishing systems, Rev. 14 WOLF CREEK TABLE 9.5A-1 (Sheet 3) APCSB 9.5-1 Appendix A WCGS c) fire prevention activities, d) training and manual fire-fighting activi-ties of plant per-sonnel and the fire
brigade.
- 2. Desiqn Bases The overall fire pro- The overall fire protection tection program should program is based upon evaluation be based upon evalua- of potential fire hazards through-tion of potential out the plant and the effect of
fire hazards postulated design basis fires throughout the plant relative to maintaining ability and the effect of to perform post-fire safe shutdown postulated design functions and minimize radioactive basis fires relative releases, as described in powerblock to maintaining abili- fire hazards analysis, Appendix 9.5B. ty to perform safety shutdown functions The Fire Protection Program in-and minimize radio- corporating those aspects of fire active releases to protection outside the power block the environment. are based upon an evaluation of potential fire hazards throughout
the plant and the effect of post-ulated design basis fires relative to maintaining ability to perform
post-fire safe shutdown functions and minimize radioactive releases as
discussed in the WCGS fire hazards
analysis, Section 9.5B.8
- 3. Backup Total reliance should Where automatic extinguishing sys-not be placed on a tems are provided, appropriate
single automatic fire backup fire suppression capa-suppression system. bility is provided. Appropriate backup fire
suppression capability should be provided. Rev. 14 WOLF CREEK TABLE 9.5A-1 (Sheet 4) APCSB 9.5-1 Appendix A WCGS
- 4. Single Failure Criterion A single failure in A single failure in the fire the fire suppression suppression system does not impair system should not both the primary and backup impair both the pri- fire suppression capability as de-
mary and backup fire scribed in Section 9.5.1.3. suppression capabiity. Two full capacity fire pumps with For example, redundant independent power sources and
fire water pumps with controls are provided. independent power supplies and controls
should be provided. Postulated fires or fire protection system fail-
ures need not be consid-ered concurrent with other plant accidents or the most severe natural phenomena. The effects of light- Lightning protection for the con-ning strikes should be tainment is provided in accordance included in the overall with NFPA No. 78-1975 and the re-
plant fire protection quirements of Underwriters' program. Laboratories, Inc. UL-96A-June, 1963.
- 5. Fire Suppression Systems Failure or inadvertent Failure or inadvertent operation operation of the fire of the fire suppression system does
suppression system not incapacitate safe shutdown should not incapacitate systems or components, as described safety-related systems in Section 9.5.1.2.1. Fire suppres-
or components. Fire sion systems that are pressurized suppression systems during normal plant operation that are pressurized meet the guidelines specified in
during normal plant APCSB Branch Technical Position operation should meet 3-1, "Protection Against Post-the guidelines speci- ulated Piping Failures in Fluid
fied in APCSB Branch Systems Outside Containment." Technical Position 3-1, "Protection Against The site-related structures and Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 5) APCSB 9.5-1 Appendix A WCGS Postulated Piping systems are remotely located from Failures in Fluid the power block and do not pose Systems Outside a hazard to safety-related struc-Containment". tures and systems.
- 6. Fuel Storage Areas Schedule for implemen- The Fire Protection Program and tation of modifica- equipment for buildings storing tions, if any, are es- new and spent reactor fuel and for tablished on a case- adjacent fire areas which could by-case basis. affect the fuel storage zone were
fully operational before fuel
was received at WCGS.
- 7. Fuel Loading Schedule for imple- The Fire Protection Program and mentation of modifi- equipment for WCGS was operational
cations, if any, are prior to initial fuel loading. established on a case-by-case basis.
- 8. Multiple-Reactor Sites Not applicable to WCGS.
- 9. Simultaneous Fires Not applicable to WCGS.
B. Administrative Procedures, Controls and Fire Brigade
- 1. Administrative proced- Administrative procedures were ures consistent with developed for control of training, the need for maintain- maintenance, and testing required ing the performance of to ensure the fire protection of
the fire protection all safety-related systems in the system and personnel plant as discussed in the WCGS Fire in nuclear power Protection Program. The listed
plants should be pro- NFPA publications were utilized for vided. guidance in development of these procedures. Guidance is contained in the following pub-
lications: NFPA 4 - Organization for Fire Services Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 6) APCSB 9.5-1 Appendix A WCGS NFPA 4A - Organization for Fire Department NFPA 6 - Industrial Fire Loss Prevention NFPA 7 - Management of Fire Emer-
gencies NFPA 8 - Management Responsibility for Effects of Fire on
Operations NFPA 27 - Private Fire Brigades
- 2. Effective administra- Administrative measures are tive measures should implemented to control the be implemented to pro- storage of combustible materials
hibit bulk storage of inside or adjacent to safety- - combustible materials related buildings or systems inside or adjacent to during normal operation or
safety-related build- maintenance periods. The ings or systems during guidance provided by Regulatory operation or mainten- Guide 1.39, "Housekeeping Re-
ance periods. Regula- quirements for Water-Cooled tory Guide 1.39, Nuclear Power Plants" is util-
"Housekeeping Require- ized in the development of
ments for Water- these measures. Cooled Nuclear Power Plants", provides
guidance on house-keeping, including the disposal of combusti-
ble materials.
- 3. Normal and abnormal Work Control Procedures, which conditions or other include identification of the anticipated operations need for special action such as such as modifications a fire watch, are utilized.
(e.g., breaking, fire Such procedures are reviewed by stops, impairment of appropriate levels of management. Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 7)
APCSB 9.5-1 Appendix A WCGS fire detection and suppression systems) and refueling activi-ties should be re-
viewed by appropriate levels of management and appropriate special
actions and procedures such as fire watches or temporary fire barriers
implemented to assure adequate fire protec-tion and reactor
safety. In particular: a) Work involving Plant operating and maintenance ignition sources personnel are trained and such as welding equipped to prevent and combat
and flame cutting fires. Personnel from offsite should be done organizations are either trained under closely prior to work involving ignition
controlled condi- sources or provided with a fire tions. Proced- watch. ures governing
such work should be reviewed and approved by per-
sons trained and experienced in fire protection.
Persons performing and directly assisting in such
work should be trained and equipped to prevent
and combat fires. If this is not possible, a person
qualified in fire protection should directly monitor
the work and func-tion as a fire watch. Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 8) APCSB 9.5-1 Appendix A WCGS b) Leak testing, and Leak testing and similar procedures similar procedures such as air flow determination do such as air flow not use open flames or combustion-determination, generated smoke where safety-
should use one of related systems could be affected. the commercially available aerosol
techniques. Open flames or combus-tion generated
smoke should not be permitted. c) Use of combustible The use of combustible materials material, e.g., inside or adjacent to safety-HEPA and charcoal related areas are administratively
filters, dry ion controlled and where possible, exchange resins or fire retardant products are used. other combustible See the WCGS Fire Protection Program. supplies, in safety-related
areas should be
controlled. Use of wood inside buildings contain-
ing safety-related systems or equip-ment should be permitted only when suitable non-combustible sub-
stitutes are not available. If wood must be used, only fire retard-ant treated wood (scaffolding, lay
down blocks) should be per-mitted. Such
materials should be allowed into safety-related
areas only when they are to be used immediately.
Their possible and Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 9) APCSB 9.5-1 Appendix A WCGS probable use should be con-sidered in the fire hazard
analysis to deter-mine the adequacy of the installed
fire protection systems.
- 4. Nuclear power plants WCGS was designed to be self-are frequently located sufficient with respect to on-site
in remote areas, at fire fighting activities. Coffey some distance from County Fire District #1 is available public fire depart- for supplemental and/or backup ments. Also, first assistance as described in the Fire response fire depart- Protection Program. No credit ments are often is taken in the fire protection volunteer. Public system design for this response. fire department re-sponse should be con-
sidered in the overall Fire Protection Pro-gram. However, the
plant should be de-signed to be self-sufficient with re-
spect to fire fight-ing activities and rely on the public
response only for supplemental or backup capability.
- 5. The need for good or- The WCGS Emergency Plan has ganization, training been prepared utilizing the
and equipping of fire guidance provided by Regulatory brigades at nuclear Guide 1.101, "Emergency Planning power plant sites re- for Nuclear Power Plants." The
quires effective WCGS Fire Protection Program measures be implement- outlines the measures to be ed to assure proper implemented to ensure the proper
discharge of these organization, training and functions. The guid- equipping of fire brigades. ance in Regulatory
Guide 1.101, "Emer-gency Planning for Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 10) APCSB 9.5-1 Appendix A WCGS Nuclear Power Plants" should be followed as applicable. a) Successful fire A test plan for fire protection fighting requires was developed (See the testing and main- WCGS Fire Protection Program).
tenance of the Procedures necessary to assure fire protection adequate fire protection during equipment, emer- refueling or maintenance are
gency lighting discussed in Items B.3 and B.3.a. and communication, as well as prac-
tice as brigades for the people who must utilize the
equipment. A test plan that lists the individuals and
their responsibili-ties in connection with routine tests
and inspections of the fire detection and protection
systems should be developed. The test plan should
contain the types, frequency and detailed proce-
dures for test-ing. Procedures should also con-
tain instructions on maintaining fire protection during
those periods when the fire protection system is impaired
or during periods of plant mainten-ance, e.g., fire
watches or tempor-ary hose connec-tions to water
systems. Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 11) APCSB 9.5-1 Appendix A WCGS b) Basic training is Details concerning Fire Brigade a necessary ele- Training and Drills can be ment in effective found in the WCGS Fire Protection fire fighting Program. operation. In
order for a fire
brigade to operate effectively, it must operate as a
team. All members must know what their individual duties are. They must be familiar with the layout of
the plant and equipment location and operation in
order to permit effective fire-fighting
operationsduring times when a particular area is
filled with smoke or is insufficiently
lighted. Such training can only be accomplished by
conducting drills several times a year (at least
quarterly) so that all members of the fire brigade have
had the opportunityto train as a team, testing itself in the major areas of the plant. The
drills should include the simulated use of
equipment in each area and should be pre-planned and
post-critiqued to Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 12) APCSB 9.5-1 Appendix A WCGS establish the train-
ing objective of the drills and determine how well
these objectives have been met. These drills
should periodically (at least annually)
include local fire department partici-pation where
possible. Such drills also permit supervising per-
sonnel to evaluate the effectiveness of communications
within the fire brigade and with the on scene fire
team leader, the reactor operator in the control room, and the off-site command post. c) To have proper During all phases of operation coverage during of the plant, each shift has
all phases of op- a sufficient number of eration, members individuals trained in fire of each shift crew prevention and suppression
should be trained to provide for appropriate in fire protection. fire protection of the facility. Training of the Local fire departments assist
plant fire brigade and participate in the training should be coordin- program to provide a mutual ated with the local understanding of responsibilities
fire department and duties, the operational so that responsi- precautions necessary bilities and duties when fighting fires at nuclear
are delineated in plant sites, the need for advance. This radioactive protection of coordination personnel and the special
should be part hazards associated with a of the training nuclear power plant site. course and imple-
mented into the Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 13) APCSB 9.5-1 Appendix A WCGS training of the The training program utilizes local fire depart- courses in fire prevention and ment staff. Local suppression which are recognized fire departments and/or sponsored by the fire
should be educated protection industry. See in the operational the WCGS Fire Protection precautions when Program for additional fighting fires on fire protection training nuclear power information. plant sites. Local fire departments should be made aware of the need
for radioactive protection of per-sonnel and the
special hazards associated with a nuclear power plant
site. d) NFPA 27, "Private See the WCGS Fire Protection Program. Fire Brigade" should be followed in organization, training, and fire drills. This standard also is
applicable for the inspection and maintenance of fire fighting equipment. Among the standards
referenced in this document, the following should
be utilized: NFPA 194, "Standard for Screw Threads and
Gaskets for Fire Hose Couplings", NFPA 196, "Standard for Fire Hose", NFPA 197, "Training Standard
on Initial Fire Attacks", NFPA Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 14) APCSB 9.5-1 Appendix A WCGS 601, "Recommended Manual of Instruc-tions and Duties for the Plant Watchman
on Guard." NFPA booklets and pamphlets listed on
Page 27-11 of Volume 8, 1971-72 are also applicable for good
training references. In addition, courses in
fire prevention and fire suppression which are recognized
and/or sponsored by the fire protection industry should be
utilized. C. Quality Assurance Program Quality Assurance (QA) The Fire Protection QA Program Programs of applicants and is a graded QA program under contractors should be de- the management of the WCGS Quality veloped and implemented to Branch.
assure that the require-ments for design, procure- The Quality Program, which ment, installation, and is applied to fire protection, testing and administrative is derived from 10CFR50 Appendix B, controls for the fire pro- and from criteria 2 and 4 through tection program for safety 10 as specified in Attachment 6 of related areas as defined in D. B. Vassallo's letter of this Branch Position are August 29, 1977. The QA Program satisfied. The program for fire protection during the
should be under the manage- Design and Construction phase is ment control of the QA given in the "Quality Assurance organization. The QA Programs for Design and Construc-program criteria that tion" document. The following apply to the fire paragraphs describe the commitments protection program made in this graded QA Program
should include the during the Plant Operations phase. following: Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 15) APCSB 9.5-1 Appendix A WCGS The QA program for fire protection was revised to include the A/E QA organization review, surveillance and audit to help
verify the effectiveness of the QA program for fire protection.
- 1. Design Control and Design control and procurement Procurement Document document control include the Control following:
Measures should be a. Measures to assure that established to assure quality standards are that all design- specified in design related guidelines of documents such as approp-
the Branch Technical piate fire protection Position are included codes and standards, and in design and procure- deviations and changes ment documents and that from these quality stand-deviations there from ards are controlled. are controlled. b. Design reviews by know-
ledgeable personnel to assure inclusion of appropiate fire protection requirements such
as:
- 1. Verification of ade-quacy of wiring isol-
ation and cable separ-ation criteria.
- 2. Verification of appro-
priate requirements for room isolation (sealing penetrations, floors, and other fire barriers).
- c. Measures to assure that pro-
curement documents adequately state and are reviewed for fire protection requirements
and that the requirements are Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 16) APCSB 9.5-1 Appendix A WCGS inspectable and controllable and the acceptance-rejection criteria are stated.
- d. Measures to assure that design and procurement docu-ment changes, including field
changes, are subject to the same controls, reviews, and approvals applicable to the
original document.
- 2. Instructions, Procedures and Drawings Inspections, tests, Inspections, tests, administra-administrative con- tive controls, fire drills, and trols, fire drills training are also addressed and training that in the WCGS Fire Protection govern the Fire Pro- Program.
tection Program
should be prescribed Instructions and procedures by by documented in- which indoctrination and training structions, pro- programs are conducted for fire cedures or drawings protection and fire fighting have and should be been developed and implemented.
accomplished in
accordance with these documents. Design, installation, inspection, test, maintenance and modification
of fire protection systems are accomplished in accordance with documented instructions, procedures and drawings. These procedures which represent the administrative control are reviewed to assure appropriate fire protection requirements are
included such as: precautions; control of ignition sources and combustibles; and provisions for
backup fire protection, if the activities require disabling a fire protection system. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 17) APCSB 9.5-1 Appendix A WCGS The installation or application of penetration seals and fire retar-dant coatings are performed in accordance with manufacturer's
instructions by personnel knowl-edgeable of these instructions. Measures exist to assure that instructions, procedures and drawings, including changes
thereto, are properly reviewed, approved, issued and distributed.
- 3. Control of Purchased Material, Equipment and Services Measures should be Measures have been established established to assure to assure that purchased material, purchased material equipment and services conform equipment and services to procurement documents. These conform to the pro- measures include:
curement documents.
- 1. Supplier evaluation and selection as appropriate.
Note that suppliers of fire protection materials and equipment are selected from
vendors known to have the ability to supply commer-cially acceptable items.
- 2. Inspection or performance testing to verify that material, equipment and
services pertaining to the site-related portions of the FPS conform to procurement
documents. Inspections occur as receipt inspections, or installation inspections, or
both as appropriate.
- 4. Inspection A program for indepen- Maintenance and modifications of dent inspection of the FPS, including emergency Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 18)
APCSB 9.5-1 Appendix A WCGS activities affecting lighting and normal plant commu-fire protection should nication equipment (plant telephone be established and and PA system), are subject to executed by, or for, inspection to assure conformance
the organization to design and installation require-performing the activ- ments. Such inspections may ity to verify conform- occur as receipt inspection or
ance with documented installation inspection, or both as installation appropriate. drawings and test
procedures for The installation of the portions accomplishing the of the FPS where performance can-activities. not be verified through
preoperational tests, such as penetration seals, fire retardant coatings, cable routing and fire
barriers is subject to inspection. Inspections are performed by either quality control or engineering personnel other than those who performed the activities being
inspected and who are knowledgeable of fire protection design and installation requirements. These
inspections are performed in accor-dance with procedures or checklists and shall include, as applicable, the following:
- a. Identification of items/activities being inspected.
- b. Individuals/organization
responsible to perform inspections.
- c. Reference design documents
and acceptance criteria.
- d. Identification of inspection method.
- e. Documentation requirements.
- f. Inspection results, inspector signoff. Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 19)
APCSB 9.5-1 Appendix A WCGS The WCGS Fire Protection Program describes the program by which the elements of the FPS are inspected to assure they are in acceptable
condition. For those materials subject to degradation (such as fire stops, seals and fire retar-
dant coatings), periodic visual inspections are performed to assure they have not deteriorated or been
damaged. Periodic inspections and/or tests are performed on the FPS, emergency breathing and auxiliary equipment, emergency lighting and normal plant
communication equipment to assure acceptable condition of these items.
- 5. Test and Test Control A test program should The WCGS Fire Protection be established and Program describes the program implemented to assure developed to verify conformance that testing is with design following modifica-performed and veri- tion, repair or replacement fied by inspection of portions of the FPS. Also
and audit to demon- discussed is a program of strate conformance periodic tests to verify with design and system system readiness requirements.
readiness require-ments. The tests The WCGS Fire Protection Pro-should be performed gram discusses training of
in accordance with personnel responsible for main-written test proce- taining and inspecting the FPS. dures; test results The programs for testing
should be properly and training are subject to QA evaluated and acted audit and surveillance. on.
Tests results are documented, evaluated and their acceptability determined by qualified groups or
individuals. Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 20) APCSB 9.5-1 Appendix A WCGS
- 6. Inspection, Test and Operating Status Measures should be Measures are developed to established to provide identify those components that for the identification have satisfactorily passed re-of items that have quired preoperational tests
satisfactorily and inspections. passed required tests and inspections.
- 7. Non-Conforming Items Measures should be Fire protection system items, established to con- including emergency lighting trol items that do and normal plant communication not conform to equipment items, that do not con-specified require- form to specific requirements ments to prevent are identified during inspections inadvertent use or and/or tests. Nonconforming installation. items identified are appropriately segregated, tagged or labeled to
prevent inadvertent use. Identification, documentation, segregation, review disposition and notification to the affected organization of nonconformances are
procedurally controlled and include identification of individuals/ groups responsible for disposition
of nonconforming items. Documentation of nonconforming items consists of identification, description, disposition and approval of disposition.
- 8. Corrective Action Measures should be Procedures have been established established to assure such that failures, malfunctions, assure that conditions deficiencies, deviations, defective adverse to fire components, uncontrolled combust-protection, such as ible material and nonconformances Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 21)
APCSB 9.5-1 Appendix A WCGS failures, malfunctions, which affect fire protection are deficiencies, devia- promptly identified, reported, tions, defective evaluated and corrected. The
components, uncon- evaluation considers the cause trolled combustible of the condition and action to material and noncon preclude recurrence.
formances are promptly identified, reported and corrected.
- 9. Records Records should be Records are maintained in accor-prepared and main- dance with ANSI N45.2.9 to show tained to furnish the criteria committed to are evidence that the being met for activities affecting criteria enumerated the Fire Protection Program.
above are being met for activities affecting the Fire Protection
Program.
- 10. Audits Audits should be con- Audits are performed to verify ducted and documented compliance with the Fire Protection
to verify compliance Program, including procedures, in-with the Fire Protec- spections and testing activities. tion Program including
design and procurement Audits are conducted by QA documents; instruc- personnel in accordance with tions, procedures written procedures or checklists.
and drawings and inspection and test Audit results are documented activities. and reviewed by supervisory Rev. 0 WOLF CREEK TABLE 9.5A-1 (Sheet 22) APCSB 9.5-1 Appendix A WCGS personnel responsible for correcting deficiencies revealed by the audit.
Followup actions are then by responsible management to correct
deficiencies identified during the audit. Audits of the Fire Protection Program are performed as Required by the Quality Program Manual, Section 18. D. General Guidelines for Plant Protection D.1. Building Design (a) Plant layouts should be Where possible, plant arranged to: layouts are arranged to:
(1) Isolate safety-related (1) Isolate safety-related systems from unacceptable systems from unaccept-
fire hazards, and able fire hazards, and (2) Separate redundant safety- (2) Separate redundant safe related systems from each shutdown systems from other so that both are not each other so that both subject to damage from a are not subject to
single fire hazard. damage from a single fire hazard. Alternatives: (a) Redundant Where this is not possible, safety related systems that protection is provided by are subject to damage from a fire resistive wraps, fire
single fire hazard should be detection, fire suppression protected by a combination systems, or a combination of of fire detection and sup- these. The details of plant
pression systems, or (b) a layout features designed to separate system to perform meet these guidelines are the safety function should given in Appendix 9.5B.
be provided. The Appendix 9.5E comparison response to 10 CFR 50 Appendix R Section III.G provides additional information
regarding the methodologies utilized for fire protection of safe shutdown capability. Rev. 24 WOLF CREEK TABLE 9.5A-1 (Sheet 23) APCSB 9.5-1 Appendix A WCGS (b) In order to accomplish 1. (b) Refer to Appendix 9.5B (a) above, safety-related for the fire hazards systems and fire hazards analysis. The hazards should be identified analysis will be reviewed
throughout the plant. There- and updated, as necessary. fore a detailed fire hazard analysis should be
made. The fire hazards analysis should be reviewed and updated as necessary.
Additional fire hazards analysis should be done after any plant modification.
(c) For multiple reactor sites, (c) Each CSR is separated cable spreading rooms should from the other areas by
not be shared between reactors. walls, floor, and ceiling Each cable spreading room having a minimum fire should be separated from resistance of 3 hours.
other areas of the plant by barriers (walls and floors) Two CSRs are provided for having a minimum fire resis- each reactor unit. Cables
tance of 3 hours. Cabling for the two redundant pairs for redundant safety divisions of shutdown divisions are should be separated by walls routed separately into
having 3-hour fire barriers. their respective CSR.
(d) Interior wall and structural (d) Interior wall and structural components, thermal insulation components are steel, rein-materials and radiation forced concrete, and other shielding materials and sound- noncombustible materials.
proofing should be non- Thermal insulation and radi-combustible. Interior finishes ation shielding materials are should be noncombustible or also noncombustible. Interior
listed by a nationally recog- wall surfaces are generally nized testing laboratory, such painted CMU or concrete. as Factory Mutual or Under- Floor areas have also been coated. writers' Laboratory, Inc. for These paints and coatings have been flame spread, smoke and fuel considered in the fire hazards analysis. contribution of 25 or less in Minimum drywall construction is used in its use configuration (ASTM E-84 the control room over steel stud framing. Test), "surface Burning
Characteristics of building
Materials"). Rev. 15 WOLF CREEK TABLE 9.5A-1 (Sheet 24) APCSB 9.5-1 Appendix A WCGS Control room carpet has a Class A flame spread rating as described in appendix 9.5B. Smoke and fuel contribution is not considered to be a significant factor with respect to fire protection within the control room, since this area is manned 24 hours/day.
(e) Metal deck roof construction (e) Metal deck roof construc-should be noncombustible tion has not been utilized (see the building materials in safety-related struc-directory of the Under- tures, except as forms for writers Laboratory, Inc.) the reinforced concrete or listed as Class I by roof; therefore, this guide-Factor Mutual system line does not apply.
Approval Guide.
(f) suspended ceilings and their (f) Suspended ceilings are non-supports should be of non- combustible and exist in the combustible construction. control room and adjacent Concealed spaces should be offices, hot chem lab, devoid of combustibles counting room, and access control area. Combustibles Adequate fire detection in concealed spaces in the and suppression systems control room, hot chem lab should be provided where and counting room consist full implementation is only of insulation for not practicable. electrical cables in trays which are not safety-related. Combustibles in concealed spaces in the access area consist of safety-related electrical cables in trays. This area is sprinkled both above and below the ceiling.
Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 25) APCSB 9.5-1 Appendix A WCGS (g) High voltage - high (g) High voltage - high amperage transformers amperage transformers installed inside buildings installed inside buildings containing safety-related containing safety related systems should be of the systems are not exposed dry type or insulated and to oil-filled transformers. cooled with noncombustible liquid.
(h) Buildings containing safety (h) Buildings containing safety-related systems having related systems having opening in exterior walls openings in exterior walls closer than 50 feet to are not closer than 50 feet flammable oil-filled trans- to flammable oil-filled formers, should be transformers.
protected from the effects of fire by: ...
(i) Floor drains, sized to remove (i) Adequate drainage is expected fire fighting water provided in all safe flow, should be provided in shutdown areas; including those areas where fixed at each elevation of water fire suppression of vertical electrical systems are installed. Drains cable chases, to remove should also be provided in fire fighting water other areas where hand hose whether from fixed lines may be used if such suppression systems or fire fighting water could manual hose stations.
cause unacceptable or damage Water-sensitive equipment to equipment in the area. components are situated Equipment should be installed above the floor level to on pedestals, or curbs should prevent damage from be provided as required to extinguishing system dis-contain water and direct it charge. The drainage to floor drains. (See NFPA system has no provisions 92M, "Water-proofing and for arresting the spread Draining of Floors.") Drains of fire within the system. in areas containing preventing However, the only safe the spread of the fire shutdown area where an throughout the drain appreciable amount of system. Water drainage from combustibles could enter areas which may contain the drainage system is in radioactivity should be the diesel generator sampled and analyzed before building. Each of the two discharge to the redundant diesels has a environment. In operating separate drainage system Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 26) APCSB 9.5-1 Appendix A WCGS plants or plants under so that a fire could not construction, if spread from one diesel accumulation of water from room to the other. All the operation of new fire drainage (except sanitary) suppression systems does is monitored for radio-not create unacceptable activity before release consequences, drains need to the environment. not be installed.
(j) Floors, walls, and ceilings (j) Where fire barriers are enclosing separate fire provided to separate areas should have minimum redundant safe shutdown fire rating of 3 hours. trains, floors, walls, Penetrations in these fire and ceilings of the barriers, including conduits enclosures have a minimum and piping, should be sealed fire rating of 3 hours.
or closed to provide a fire- Refer to Section 9.5.1.2.2. resistance rating at least Penetrations in these fire equal to that of the fire barriers, including conduits barrier itself. Door and piping, are sealed or openings should be protected closed to provide a fire with equivalent rated doors, resistance rating of 3 hours. frames, and hardware that Hatchways in the auxiliary have been tested and approved building floors are protected by a nationally recognized as detailed in 9.5.1.1.2. laboratory. Such doors should be normally closed Normally, doors, frames, and and locked or alarmed with hardware have the same rating alarm and annunciation in as the barrier. Elevator the control room. and dumbwaiter doors will be Penetrations for rated at 1-1/2 hours, since ventilation system should it is an industry standard be protected by a standard and, as stated in ANSI A17.1, "fire door damper" where such doors are acceptable for required. (Refer to NFPA use on a 2 hour rated shaft. 80, "Fire Doors and For a fire to propagate from Windows.") The fire one floor elevation to another hazard in each area should it would have to penetrate two be evaluated to determine doors. barrier requirements. If barrier fire resistance cannot be made adequate, fire detection and suppression should be provided, such as: Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 27) APCSB 9.5-1 Appendix A WCGS (i) water curtain in Doors are normally closed and case of fire, locked, except where the door (ii) flame retardant is a means of egress in which coatings, case they are closed and (iii) additional fire latched. The door between the barriers. Pantry and Control Room is normally open. This door is provided with an electromagnetic closer which closes the door upon detection of fire. Ventilation openings are protected by fire dampers having a rating equal to the barrier. The fire hazard in each area has been evaluated to determine barrier requirements. This analysis is presented in Appendix 9.5B. Control building floors and ceilings are rated as 3- hour fire barriers. Three-hour-rated fire stops are provided at each floor elevation in vertical cable chases. Auxiliary building floors and ceilings are rated as 3-hour fire barriers. The roofs of the auxiliary and control buildings are not fire barriers. The area however, constructed of non-combustible concrete with built up Class A roofing. In addition the structural steel supporting the roof is protected with 3 hour rated fire proofing. Therefore, building structural integrity is maintained. 2. Control of Combustibles (a) Safety-related systems Safety-related systems should be isolated separated are isolated or separated from combustible materials. from combustible When this is not possible materials, where because of the nature of the practical. Where this safety system or the is not practical, combustible material, special protection is Rev. 23 WOLF CREEK TABLE 9.5A-1 (Sheet 28) APCSB 9.5-1 Appendix A WCGS should be provided to prevent provided to prevent a fire from defeating the failure of both safe safety system function. Such shutdown trains by a protection may involve a single fire. Refer of
combination of automatic both safe shut-down fire suppression and trains by a single fire construction capable of Refer of Both safe
with-standing and shutdown trains by a to containing a fire that the fire Hazards consumes all combustibles Analysis, Appendix 9.5B.
present. Examples of such combustible materials Administrative controls that may not be separable have been established to
from the remainder of its control the introduction system are: of transient combustibles into the safety-related
areas.
(1) Emergency diesel generator fuel oil day tanks (2) Turbine-generator oil and hydraulic control fluid systems (3) Reactor coolant pump lube oil system.
(b) Bulk gas storage (either Flammable bulk gas storage is compressed or cryogenic), not permitted inside structures should not be permitted housing safe shutdown equipment
inside safety-related as a design restriction. All equipment. Storage of bulk gas storage facilities flammable gas such as structures housing are located
hydrogen should be located outdoors outdoors or in separate detached buildings so that
a fire or explosion does not adversely affect any safety-related systems
or equipment.
(Refer to NFPA 05A, "Gaseous Hydrogen Systems".)
Care should be taken to locate high pressure gas storage containers with
the long axis parallel to building walls. This minimizes the possibility Rev. 4 WOLF CREEK TABLE 9.5A-1 (Sheet 29) APCSB 9.5-1 Appendix A WCGS of wall penetration in the event of a container failure. Use of com-pressed gases (especially flammable and fuel gases) inside buildings should be controlled. (Refer to NFPA 6, "Industrial Fire Loss Prevention") c) The use of plastic Halogenated materials are materials should be used for electrical cable minimized. in particular, insulation and jacketing. halogenated plastics such as These materials exhibit polyvinyl chloride (PVC) acceptable electrical, me-and neoprene should be used chanical, and environmental only when substitute non- characteristics. The material combustible materials are not is minimized to the extent available. All plastic practicable by limiting the materials, including flame wall thickness to the minimum and fire retardant materials, permitted by applicable burn with an intensity and industry standards. BTU production in a range similar to that of ordinary Insulating materials used for hydrocarbons. When burning, cables routed in cable trays are: they produce heavy smoke that obscures visibility, (1) 5 and 15 Kv cable and can plug air filters, insulation - Kerite HTK especially charcoal and compound jacket - Kerite HEPA. The halogenated HTNS (5 Kv) and Kerite plastics also release free FR (15 Kv) chlorine and hydrogen chloride when burning (2) 600 V control cable which are toxic to humans insulation - cross linked and corrosive to polyethylene jacket - equipment. Neoprene (3) 600 V Power and Instru-mentation Cable insulation - Ethylene-propylene rubber jacket - chlorosulfonated poly-ethylene d) Storage of flammable Storage of Flammable liquids liquids should, as a complies with the requirements minimum, comply with of NFPA 30q-1973. the requirements of NFPA 30, "Flammable and Combustible Liquids Code". Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 30) APCSB 9.5-1 Appendix A WCGS
- 3. Electrical Cable Construction, Cable Trays and Cable Penetrations a) Only non-combustible materials Galvanized steel is used should be used for cable tray for cable tray construction.
construction. b) See Section E.3 for fire See Section E.3. protection guidelines for cable spreading rooms. c) Automatic water sprinkler Automatic sprinkler systems systems should be provided are provided for vertical
for cable trays outside the cable chases, the cable area cable spreading room. Cables above the suspended ceiling in should be designed to allow the access control area which
wetting down with deluge water contain Class 1E cables, and without electrical faulting. zones in the auxiliary building Manual hose stations and with cable concentrations.
portable hand extinguishers Manually charged, closed head should be provided as backup. sprinkler systems are provided Safety-related equipment in for the two cable penetration
the vicinity of such cable areas inside the containment. trays does not itself require Cables are designed to allow water fire protection but is wetting down without electrical
subject to unacceptable damage faulting. Manual hose stations from sprinkler water discharge, and portable hand extinguishers should be protected from are provided as backup.
sprinkler system operation or Sprinkler systems are not malfunction. installed in areas where sprinkler operation would
cause damage to safe shutdown equipment. When safety-related cables Safety-related cables satisfy do not satisfy the provisions the provisions of Regulatory of Regulatory Guide 1.75, all Guide 1.75. exposed cables should be covered with an approved fire retardant Safety-related fire-resistive cables coating and a fixed automatic exceed the intent of the provisions of water fire suppression system Regulatory Guide 1.75. should be provided. Rev. 24 WOLF CREEK TABLE 9.5A-1 (Sheet 31) APCSB 9.5-1 Appendix A WCGS d) Cable and cable Cable and cable tray penetra-tray penetration tion of fire barriers (vertical of fire barriers and horizontal) are sealed to give (vertical and protection at least equivalent to horizontal) should the barrier which they pene-be sealed to give trate. Typical horizontal and protection at vertical cable tray penetra-least equivalent tions are tested to prevent to that fire the spread of fire and retain barriers for structural soundness when horizontal and exposed to a 3-hour fire as vertical cable discussed in 9.5.1.2.2. trays should, as a minimum, meet the requirements of ASTM E 119, "Fire Test of Building Construction and Materials," in cluding the hose stream test. Where installed penetration seals are deficient with respect to fire resistance, these seals may be protected by covering both sides with an approved fire retardant mater-ial. The adequacy of using such material should be demonstrated by suitable testing. e) Fire breaks should Fire breaks are provided as be provided as deemed necessary by the fire deemed necessary hazards analysis. The cable by the fire rating is compatible with the hazards analysis. construction of the fire Flame or fire re- break. Refer to Appendix 9.5B tardant coatings and Section 9.5.1.2.2. may be used as a fire break for grouped electrical Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 32) APCSB 9.5-1 Appendix A WCGS cables to limit spread of fire in cable ventings.
(Possible cable
derating owing to use of such coating materials must be
considered during design.) f) Electrical cable Safety-related electrical constructions cable passes the IEEE 383-1974 should as a mini- flame test or meet the intent mum pass the of this requirement as discussed current IEEE No. in Appendix 9.5B.
383 flame test. (This does not Fire-resistive cables are constructed from imply that cables non-flammable materials: silicon dioxide passing this test insulation, copper nickel conductors and will not require stainless steel jacketing, and are tested additional fire to exceed any flame test requirements of protection.) IEEE 383. For cable instal-lation in opera-ting plants and plants under
construction that do not meet the IEEE No. 383 flame
test requirements, all cables must be covered with an
approved flame retardant coating and properly
derated. g) To the extent See response to D.2(c) above. practical, cable construction that does not give off
corrosive gases while burning should be used.
(Applicable to new cable installa-tions.) Rev. 24 WOLF CREEK TABLE 9.5A-1 (Sheet 33)
APCSB 9.5-1 Appendix A WCGS h) Cable trays, race- Cable trays, raceways, con-ways, conduit, duit, and cable trenches are trenches, or used for the routing of cables culverts should be only.
used only for cables. Miscellane-ous storage should
not be permitted, nor should piping for flammable or
combustible liquids or gases be installed in these areas. Installed equip-ment in cable
tunnels or cul-verts need not be removed if they
present no hazard to cable runs as determined by the
fire hazards analysis. i) The design of cable Smoke venting is discussed in tunnels, culverts, Section D.4(a). and spreading
rooms should provide for auto-matic or manual
smoke venting as required to facil-itate manual fire
fighting capa-bility. j) Cables in the Cable in the control room is control room limited to that necessary for should be kept to control room operation.
the minimum neces- Cables entering the control sary for operation room terminate there. Floor of the control trenches provide for cabling
room. All cables access to the operators con-entering the sole and other panels from the control room upper CSR.
should terminate Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 34) APCSB 9.5-1 Appendix A WCGS there. Cables Cable trenches are provided should not be with fixed automatic total installed in floor flooding Halon systems. trenches or cul-
verts in the control room. Existing cabling
installed in concealed floor and ceiling spaces
should be pro-tected with an automatic total
flooding Halon system.
- 4. Ventilation a) The products of com- Fire and smoke are automatic-bustion that need ally isolated in all areas of to be removed from the auxiliary, radwaste, fuel, a specific fire and control buildings by the
area should be fusible link actuated fire evaluated to deter- dampers in all rated fire mine how they are barrier walls or by remote
controlled. Smoke manual operations of area and corrosive ventilation systems. Indivdidual gases should gen- area of the control building erally be auto- can be isolated from the normal matically dis- supply and exhaust HVAC system by charged directly control switches in the control outside to a safe room. location. Smoke and gases contain- All exhaust fans, with the
ing radioactive exception of the control build-materials should be ing fans, are centrifugal with monitored in the the motor located outside of
fire area to deter- the airstream, thus making them mine if release to less susceptible to high gas the environment is temperatures. The fans are
within the per- capable of processing air of missible limits of temperatures at least as high the plant Technical as the fusible link melting
Specifications. temperature (160 F) of the fire The products of dampers. The control building combustion which exhaust fans are vaneaxial with
need to be removed the motors located in the from a specific process airstream. The fan Rev. 8 WOLF CREEK TABLE 9.5A-1 (Sheet 35) APCSB 9.5-1 Appendix A WCGS fire area should be motor is designed for a minimum evaluated to deter- 150 F temperature rise. mine how they are controlled. Since the exhaust fans are all
downstream of the system filter units, they are not subject to damage from high temperature
particles. The auxiliary building and fuel building can be exhausted by either the normal building system or the emergency (Class IE) system. The control building and reactor building can be exhausted by
the normal system which is located remotely to the hazard. The diesel building
utilizes the exhaust air flow path, including dampers as the means of heat and smoke
venting and takes no credit for mechanical exhaust. For areas which have a potential
for radioactivity, such as the radwaste building, releases are through normal
(or emergency) process points which are monitored for radioactivity releases.
b) Any ventilation There are no fans provided system designed specifically for the function
to exhaust smoke of smoke exhaust. The normal or corrosive and/or emergency exhaust fans gases should be may be used for the purpose.
evaluated to This arrangement limits plant ensure that releases through normal pro-inadvertent cess points, thus eliminating
operation or inadvertent releases to the single failures environment. does not violate
the controlled areas of the plant design.
This requirement includes contain-ment functions
for protection of Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 36) APCSB 9.5-1 Appendix A WCGS the public and maintaining habitability for operations
personnel. c) The power supply The power cables for exhaust and controls for fans are physically separated mechanical venti- for Class IE fans. The non-lation systems IE exhaust fans for the
should be run reactor and control buildings outside the fire are located in the auxiliary area served by building. The diesel build-the system. ing utilizes the exhaust air flow path, including dampers, as the means of heat and
smoke venting and takes no credit for mechanical exhaust. The control cable
for isolation dampers may be located within the fire area. d) Fire supression The charcoal adsorbers are systems should be sized for iodine loadings of installed to 2.5 mg/gm. Where this load-
protect charcoal ing may be approached, a low-filters in flow air-bleed system is accordance with provided per Regulatory
Regulatory Guide Guide 1.52. In addition, 1.52, "Design each charcoal adsorber unit Testing and is equipped with a high
Maintenance temperature detection system Criteria for which alarms in the control Atmospheric room and a manually activated
Cleanup Air water spray system for the Filtration." charcoal bed. e) The fresh air Exhausts from safety-related supply intakes to buildings, except the diesel areas containing and control buildings, are
safety-related through the unit vent. The equipment or diesel building takes suction systems should be and discharges through a
located remotely penthouse which has the from the exhaust louvers separated by 36 feet air outlets and and located on opposite sides
smoke vents of of the structure. The con-other fire areas trol building exhaust is Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 37) APCSB 9.5-1 Appendix A WCGS to minimize the possibility of contaminating the intake
air with the products of combustion. located approximately 30 feet away from the auxiliary building
intake. The control building takes its intake air from the auxiliary building intake. Stairwells should be designed to minimize smoke infiltration during a fire.
Staircases should serve as escape routes and access routes for fire fighting.
Fire exit routes should be clearly marked. Stairwells, elevators, and chutes should
be enclosed in masonry towers with minimum fire rating of 3 hours and
automatic fire doors, at least equal to the enclosure construction, at each Stairwells which serve as escape routes and access routes for fire fighting are enclosed in
reinforced masonry towers with a fire rating of 3 hours. Stairwell doors have a 3 hour
fire rating in seismic Category I buildings. In non-seismic Category I buildings, stairwell
doors have a 1 1/2 hour rating. Fire preplans and drills are performed to provide escape and
access routes for all areas. Elevators are enclosed in reinforced masonry towers with
doors rated for 1-1/2 hours, B label.opening into the building. Elevators should not be used during fire emergencies. Where stairwells or
elevators cannot be enclosed in 3-hour fire-rated barrier The stairwell in the communications corridor is enclosed in a masonry tower with a 2 hour fire rating. with equivalent fire doors, escape and access routes should be established by pre-fire plan and practiced
in drills by operating and fire brigade personnel. The reactor building elevator and stairs are not enclosed but preestablished escape routes are used. Rev. 14 WOLF CREEK TABLE 9.5A-1 (Sheet 38) APCSB 9.5-1 Appendix A WCGS g) Smoke and heat vents Heat and smoke venting for each may be useful in diesel generator room is pro-specific areas such vided by utilizing the exhaust as cable spreading air flow path. The free area
rooms and diesel of the exhaust air flow path fuel oil storage provides at least 1.0 square areas and switch- feet of venting area for each
gear rooms. When 200 square feet of floor area. natural-convection Smoke exhaust fans per se are ventilation is not employed. Normal ventila-
used, a minimum tion exhaust systems are uti-ratio of 1 foot2 of lized throughout for smoke venting area per removal.
200 foot2 of floor area should be provided. If
forced-convection ventilation is used, 300 CFM
should be provided for every 200 foot2 of floor areas.
See NFPA No. 204 for additional guidance or smoke
control. h) Self-contained Refer to Table 9.5E-1 comparison to breathing appara- 10CFR50 appendix R, Section III.H, for tus, using full discussion of emergency breathing air face positive equipment and supply. pressure masks, approved by NIOSH (National Insti- tute for Occupa- tional Safety and Health-approval formerly given by the U.S. Bureau of Mines) should be provided for fire brigade, damage control and control room personnel. Control room
personnel may be Rev. 17 WOLF CREEK TABLE 9.5A-1 (Sheet 39) APCSB 9.5-1 Appendix A WCGS furnished breath-ing air by a manifold system piped from a
storage reservoir if practical. Service or oper-ating life should be a minimum of
one half hour for the self-contained units. At least two extra bottles should be
located onsite for each self-con-tained breathing
unit. In addition, an on-site 6-hour supply of reserve
air should be provided and arranged to per-
mit quick and complete replen-ishment of
exhausted supply air bottles as they are returned. If
compressors are used as a source of breathing air, only units approved for breathing air
should be used. Special care must be taken to
locate the com-pressor in areas free of dust and
contaminants. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 40) APCSB 9.5-1 Appendix A WCGS i) Where total flood- Ventilation systems serving ing gas extinguish- areas protected by Halon 1301 ing systems are are provided with isolation used, area intake capabilities. The area is
and exhaust venti- isolated either by positive lation dampers closure dampers or by stopping should close upon the ventilation system fan.
initiation of gas Closure is initiated automat-flow to maintain ically upon detector actuation. necessary gas
concentration.
(See NFPA 12, "Carbon Dioxide Systems", and 12A, "Halon 1301 Systems.")
- 5. Lighting and Communication Lighting and two way voice communication are vital to safe shutdown and emer-gency response in the event of fire. Suit-
able fixed and port-able emergency light-ing and communication
devices should be pro-vided to satisfy the following require-
ments: a) Fixed emergency Fixed emergency lighting lighting should consists of sealed beam units consist of sealed with individual 8-hour minimum beam units with battery power supplies.
individual 8-hour minimum battery power supplies. b) Suitable sealed Suitable sealed beam battery beam battery powered portable hand lights
powered portable are provided for emergency hand lights should use. be provided for
emergency use. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 41) APCSB 9.5-1 Appendix A WCGS c) Fixed emergency Fixed emergency communication communication is available by use of the should use voice plant wide public address system. powered head sets The PA system is supplied power at pre-selected from two separate sources (see stations. section 9.5.2.2.1). d) Fixed repeaters The radio system repeater installed to located at elevation 2061'6" permit use of in the Communication Corridor portable radio is protected from exposure communication fire damage. units should be protected from exposure fire damage. E. Fire Detection and Suppression
- 1. Fire Detection a) Fire detection sys- The fire detection system tems should, as a complies with NFPA 72D-1975 minimum, comply for Class A systems as with NFPA 72D, detailed in 9.5.1.2.2.2 with
"Standard for the the following exceptions.
Installation, Supervision of the fire pro-Maintenance, and tection panel is not the primary
Use of Proprietary function of the plant operator Protective Sig- assigned to monitoring the naling Systems." panel. Since there are few
and infrequent signals to this Deviations from panel, a fulltime supervisor is the requirements not justified.
of NFPA 72D should be identified and justified. Supplemental to the require-ments of NFPA 72D, Class A, the following provisions are made:
(1) All initiating device circuits (detection
circuits) which actuate automatic suppression systems
serving safe shutdown areas of the plant Rev. 3 WOLF CREEK TABLE 9.5A-1 (Sheet 42) APCSB 9.5-1 Appendix A WCGS are designed to perform their detection functions in the event of a single break or
single ground fault in the circuits. For systems such as Halon
extinguishing systems which are actuated by two zones of detection
in the same hazard area, each zone is not designed to maintain detection capabilities during a single ground fault or break. Upon
generation of a trouble signal in one of the fire detection zones, a
trouble alarm is sent to the control room. In this condition, the
system automatically discharges the Halon on receipt of an alarm
signal from the second zone of detection.
(2) Upon receipt of a trouble signal on the fire annun-ciation panel in the
control room, a runner is dispatched immediately to the respective zone to
investigate the cause of the trouble signal. b) Fire detection sys- The detection system gives tem should be audible and visual alarm audible and and annunciation in the
visual alarm and control room and locally. annunciation in the control
room. Local audible alarms should also sound
at the location of the fire. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 43) APCSB 9.5-1 Appendix A WCGS c) Fire alarms should Fire alarms are distinctive be distinctive and unique. Horn and bell- and unique. They type alarms are used in should not be both the control room and capable of being locally. confused with any other plant system alarms. d) Fire detection The detection and actuation and actuation systems are connected to the systems should be non-Class IE dc system which connected to the is backed by a battery charger plant emergency supplied from the plant power supply. emergency power supply as described in 9.5.1.2.2.2.
- 2. Fire Protection Water Supply Systems a) An underground An underground yard fire main yard fire main loop is provided to furnish loop should be the design fire water re- installed to fur- quirements of the power block nish anticipated and various site structures. fire water re- The underground yard loop quirements. NFPA piping, valves, hydrants, 24-Standard for etc., conforms to NFPA 24. Outside Protec- Lined ductile-iron pipe, in tion - gives accordance with ANSI A21.51 necessary guid- (AWWA C151) is used. The ance for such piping system may be flushed installation. It through normal system opera- references other tion. If required, chemical design codes and treatment of the system may standards devel- be accomplished by injection oped by such of chemicals into the pump organizations as suction area or through other the American designated points to control National Stan- corrosion and organic fouling. dards Institute Approved visually indicating (post (ANSI) and the indicator) sectional control American Water valves are provided to iso- Works Association late portions of the under- (AWWA). Lined ground yard firemain loop. steel or cast Eighteen (18) sectional iron pipe should control valves have been placed be used to reduce in underground pits. These valves are located within the Security System Isolation Zone. Rev. 11 WOLF CREEK TABLE 9.5A-1 (Sheet 44)
APCSB 9.5-1 Appendix A WCGS
internal tubercu- The underground yard fire lation. Such main loop piping is separate tuberculation from the service water and deposits in an sanitary water system pip-unlined pipe over ing. The service water a period of years system and FPS are inter- can significantly connected, however, as one reduce water flow fire protection jockey pump through the takes suction from the combination of service water system, a second increased friction submersible jockey pump takes and reduced pipe suction from the 'A' CWSH bay. diameter. Means Only one jockey pump is required for treating and to operate to maintain fire water flushing the system pressure. (See Figure 9.5-1). systems should be provided. Approved visually Indicating sectional control valves, such as Post Indicator Valves, should be provided to isolate portions of the main for maintenance or repair without shutting off the entire system.
The fire main sys- In addition to the standard fire tem piping should suppression uses, the system is be separate from used as a source of water for fire service or brigade training and as a back up sanitary water source of raw water for plant system piping. safe shut down for design basis accidents other than fire. b) Multi Unit Site Not applicable. Fire Mains c) If pumps are re- Two 100% capacity fire pumps, one quired to meet electric motor driven, and one
Rev. 28 WOLF CREEK TABLE 9.5A-1 (Sheet 45) APCSB 9.5-1 Appendix A WCGS
system pressure or diesel engine driven are provided. flow requirements, The underground yard loop is sup-
a sufficient num- plied by two 12-inch diameter pipes ber of pumps which are separated by a distance should be provided of approximately 30 feet. This
so that 100% distance is based on a recommenda-capacity is avail- tion by ANI. Each fire pump is able with one pump provided with independent power
inactive (e.g., supplies and controller. The three 50% pumps motor-driven fire pump is located or two 100% pumps). within the general equipment area
The connection of the circulating water screen-to the yard fire house and not within a separate main loop from room. The general equipment area
each fire pump of the screenhouse also contains should be widely the fire protection jockey pumps, separated, pre- service water pumps, strainers, ferably located air compressors, and other related on opposite sides equipment. (See Figure 10.4-2) of the plant.
Each pump should The electrical equipment associ- have its own ated with the above equipment is driver with inde- located within a separate room pendent power provided with three hour rated fire supplies and barriers. Thus the general equip-control. At ment area of the screenhouse does least one pump not contain combustible (if not powered materials and no separation from the emergency for the motor driven fire pump diesels) should is deemed necessary. The diesel be driven by non- engine driven fire pump is electrical means, located within a room enclosure. preferably diesel The west wall of this room, which engine. Pumps is between the two fire pumps, is and drivers a three hour rated fire barrier. should be located Each fire pump is individually in rooms separated alarmed in the Main Control Room from the remain- to indicate pump running, driver ing pumps and availability and failure to start. equipment by a In addition, the diesel driven fire minimum three- pump has a general alarm in hour fire wall. the Main Control Room to indicate Alarms indicating system failure (overspeed, water temperature, air damper switch,
Rev. 28 WOLF CREEK TABLE 9.5A-1 (Sheet 46) APCSB 9.5-1 Appendix A WCGS pump running, crank termination and oil pres-driver avail- sure), low fuel and battery ability, or fail- failure. ure to start should
be provided in the control room. Details of the The fire pump installation conforms fire pump instal- to NFPA 20. lation should as
a minimum conform to NFPA 20, "Standard for the Installation of Centrifugal Fire Pumps". d) Two separate reli- (See Item E.2.f) able water sup-
plies should be provided. If tanks are used, two 100% (minimum of 30,000 gallons each) system cap-
acity tanks should be installed. They should be so
interconnected that pumps can take suction from
either or both. However, a leak in one tank or its
piping should not cause both tanks to drain. The
main plant fire water supply ca-pacity should be
capable of refil-ling either tank in a minimum of
eight hours. Common tanks are Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 47) APCSB 9.5-1 Appendix A WCGS permitted for fire and sanitary or service water storage. When
this is done, how-ever, minimum fire water storage
requirements should be dictated by means of a ver-
tical standpipe for other water services. e) The fire water The FPS flow rate and pressure supply (total ca- requirements are determined based pacity and flow on the maximum demand of the power
rate) should be block. The flow requirements are calculated on the based on providing 500 gpm for basis of the manual hose streams. largest expected flow rate for a
period of two hours, but not less than 300,000 gallons. This flow rate should be based (conservatively) on 1,000 gpm for manual hose streams plus the
greater of:
- 1) all sprinkler heads opened and flowing in the largest
designed fire area; or
- 2. the largest open head deluge system(s)
operating. Rev. 19 WOLF CREEK TABLE 9.5A-1 (Sheet 48) APCSB 9.5-1 Appendix A WCGS f) Lakes or fresh The source of water for fire pro- water ponds of tection is the Wolf Creek sufficient size lake. The fire pumps take suction may qualify as from a common wet pit sump in the sole source of circulating water screenhouse. Two water for fire vertical traveling screens and a protection, but bar grill are located at the inlet require at least to the sump serving the fire pumps. two intakes to An open pipe connection to the the pump supply. adjacent sump in the circulating When a common water screenhouse is provided water supply is as a second source of water to the permitted for fire pumps and to preclude de-fire protection watering of the sump serving the and the ultimate fire pumps. There are four 12- heat sink, the inch diameter pipe connections following con- between the fire pump screen- ditions should house sump and the adjacent circu- also be satisfied: lating water screenhouse sump, as shown in Figure 10.4-1-05. 1) The additional In the event of blockage of water fire protec- flow into the fire pump sump tion water through the traveling screens in requirements one bay, one connection between are designed adjacent bays and one operational into the traveling screen is all that is re- total storage quired to provide an adequate capacity; and source of water from the adjacent bay to a fire pump running at 100% capacity at the cooling lakes' lowest design level. 2) Failure of the fire pro- tection sys- tem should not degrade the function of the ultimate heat sink. The 5,090 acre Wolf Creek lake is adequate to provide fire protection water storage as well as cooling water. The FPS does not take water from the ultimate heat sink. Rev. 18 WOLF CREEK TABLE 9.5A-1 (Sheet 49) APCSB 9.5-1 Appendix A WCGS g) Outside manual Hydrants with hose houses equipped hose installation with hose and combination nozzle should be suffic- and other auxiliary equipment ient to reach recommended in NFPA 24, are lo-
any location with cated approximately every 250 feet an effective hose on the yard main system. The stream. To accom- lateral from the yard main to
plish this hyd- each hydrant is provided with a rants should be curb valve. Threads on all installed approx- hydrants, hose couplings, and
imately every 250 standpipe risers are compatible feet on the yard with those used by the fire main system. The departments of Coffey County, Kansas lateral to each and Emporia (Lyon County), Kansas. hydrant from the
yard main should
be controlled by a visually indi-cating or key
operated (curb) valve. A hose house, equipped with hose and combination nozzle, and other
auxiliary equipment recommended in
NFPA 24, "Outside Protection", should be
provided as needed but at least every 1,000
feet. Threads compatiblewith those used by local fire
departments should be provided on all
hydrants, hose couplings and standpipe risers. Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 50) APCSB 9.5-1 Appendix A WCGS
- 3. Water Sprinklers and Hose Standpipe Systems a) Each automatic The sprinkler systems (both sprinkler system and manual and automatic ) are manual hose station supplied from a header which standpipe should have is fed from each end. A
an independent connec- separate header, also fed from tion to the plant both ends, is provided for all underground water main. standpipes except the reactor
Headers fed from each building. The header arrange-end are permitted in- ment is such that no single side buildings to failure can impair both the
supply multiple sprinkler systems and the sprinkler and stand- standpipe system. For reactor pipe systems. When building system arrangement
provided, such headers and single failure discussion, are considered an ex- refer to Figure 9.5.1-1 and tension of the yard Section 9.5.1.2. main system. The header arrangement should be such that no single
failure can impair both the primary and backup fire protection
systems. Each sprinkler and Each sprinkler and standpipe standpipe system system is equipped with OS&Y should be equipped with gate valves to isolate the OS&Y (outside screw and system. Individual automatic
yoke) gate valve, or sprinkler systems are other approved shutoff equipped with water flow valve, and water flow alarms. Water flow in the
alarm. Safety-related standpipe system is indicated equipment that does not by fire pump annunciation. itself require
sprinkler water fire Where sprinkler systems are re-protection, but is quired in the vicinity of subject to unacceptable water-sensitive safe shutdown
damage if wetted by equipment, preaction-type sprinkler water dis- sprinkler systems are in-charge should be stalled. In no case are water
protected by water extinguishing systems shields or baffles. installed such that both safe shutdown trains would be
damaged by system discharge or malfunction. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 51) APCSB 9.5-1 Appendix A WCGS b) All valves in the fire Shutoff valves for each power-water systems should block fixed extinguishing system be electrically super- and each main fire protection
vised. The electrical system header are electrically supervision signal supervised. Standpipe isolation should indicate in the valves are locked in the open
control room and other position. All other power block appropriate command valves (drain, vent, and hose locations in the plant valves) are not supervised.
(See NFPA 26, "Supervision of The fire protection yard loop Valves".) isolation and yard loop branch isolation valves are electrically supervised with indication on the fire
protection panel in the Main Control Room, or are administratively controlled as discussed below. When electrical For those control and sec-supervision of fire tionalizing valves in the
protection valves is FPS that are not electric-not practicable, an ally supervised, an admin-adequate management istrative control has been
supervision program developed to implement should be provided. procedures addressing con-Such a program should trol of locked or sealed
include locking valves valves and periodic sur-open with strict key veillance of valves which control; tamper proof are not locked or sealed.
seals; and periodic, visual check of all valves. c) Automatic sprink- Automatic water extinguish-ler systems ing systems are designed, should, as a constructed, and tested minimum, con- based on NFPA 13-1975 and form to re- 15-1973, as applicable.
quirements of appropriate standards such
as NFPA 13, "Standard for the Installation
of Sprinkler Systems," and NFPA 15 "Stan-
dard for Water Spray Fixed Systems." Rev. 12 WOLF CREEK TABLE 9.5A-1 (Sheet 52) APCSB 9.5-1 Appendix A WCGS d) Interior manual hose Interior hose stations are installation capable of reaching all acces- should be able to sible areas of the plant, reach any location including inside the reactor ith at least one building, with an effective effective hose hose stream. In addition, a stream. To accom- fire in the immediate vicinity plish this, stand- of a hose station can be pipes with hose extinguished, using an adja- connections equip- cent hose station. ped with a maximum of 75 feet of 1- All hose stations, except the 1/2-inch woven- hose stations protecting the jacket-lined fire diesel generator rooms and hose and suitable cable spreading rooms and the nozzles should be north end of Corridor No. 2, equipped with 75 Room 1408, of the Aux Building feet of 1-1/2 Elevation 2026, are equipped with provided in all 75 feet of 1-1/2" inch woven buildings, includ- jacket, lined fire hose and ing containment, adjustable nozzles. The hose on all floors and stations protecting the diesel should be spaced generator rooms and cable at not more than spreading rooms are equipped 100-foot inter- with 100 feet of hose to vals. Individual provide effective coverage for standpipes should all accessible areas. The hose be of at least 4- stations at the North end of inch diameter for Corridor No. 2 are also equipped multiple hose with 100 feet of hose to facilitate connections and 2- actions required by Emergency 1/2-inch diameter Procedures in the event of a loss for single hose of all AC power The hose stations connections. are spaced at not more than 100 These systems feet from an adjacent hose station. Should follow the Standpipe risers are of at requirements of least 4-inch diameter for NFPA No. 14 for multiple hose connections. Sixing, spacing, The standpipe system is based and pipe support on NFPA 14-1976. Requirements Hose stations are located Hose stations outside entrances to normally should be located occupied areas and inside outside entrances normally occupied and un-to normally un- occupied areas, where pos- occupied areas and sible. All hose stations are inside normally equipped with pressure re- occupied areas. ducing devices where required by code. Standpipe isolation valves are located outside of safe shutdown equipment areas, where possible. Rev. 19 WOLF CREEK TABLE 9.5A-1 (Sheet 53) APCSB 9.5-1 Appendix A WCGS Standpipes serving hose stations in areas housing safety-related equipment should have shutoff valves and pres-sure reducing devices (if applicable) out-side the area. e) The proper type of Either a combination spray/ hose nozzles to be straight stream nozzle or a supplied to each spray nozzle is provided for area should be interior hose stations. The based on the fire hose stations are utilized hazard analysis. by fire brigade personnel The usual combi- only, who are trained to nation spray/ apply the appropriate spray straight-stream pattern commensurate with nozzle may cause the electrical hazard and unacceptable fire severity. mechanical damage (for example, the delicate elec- tronic equipment in the control room) and be unsuitable. Electrically safe nozzles should be provided at loca- tions where elec-trical equipment or cabling is located. f) Certain fires such No foam extinguishing systems as those involving are provided in the power flammable liquids block buildings. The fuel oil respond well to storage tank is provided with foam suppression. a dry-pipe mechanical type foam extinguishing system. Considerations should be given to use of any of the available foams Rev. 23 WOLF CREEK TABLE 9.5A-1 (Sheet 54) APCSB 9.5-1 Appendix A WCGS for such specialized protection application. These include the more common chem-
ical and mechani-cal low-expansion foams, high-
expansion foam, and the relatively new aqueous film-
forming foam (AFFF).
- 4. Halon Suppression Systems The use of Halon fire Halon extinguishing systems extinguishing agents are based on NFPA 12A-1973.
should, as a minimum, Only approved agents are used. comply with the re-quirements of NFPA 12A Each Halon system is capable of and 12B, "Halogenated attaining a 5-percent minimum
Fire Extinguishing concentration. Each system is Agent Systems - Halon designed to maintain a 5-1301 and Halon 1211." percent minimum concentration
Only UL or FM approved at the highest combustible agents should be used. material in the hazard area for 10 minutes. (Refer to Sections
In addition to the 9.5.1.2.2.1 and 9.5.1.2.3) guidelines of NFPA 12A and 12B, preventative
maintenance and test-ing of the systems, including check weigh-
ing of the Halon cylinders should be done at least
quarterly. The system actuation is by a cross-zoned, ionization-type detection system. Detection by the first zone alarms
locally and in the control room. Detection by both zones will sound a local horn, close
required dampers, shut off Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 55) APCSB 9.5-1 Appendix A WCGS associated ventilation and/or air conditioning fan motors, and discharges after a time delay for personnel evacua-
tion. A momentary contact abort switch is provided in each local panel to delay the
discharge for evacuation purposes. Each local control panel has a separate key lock
switch to disable system controls during maintenance operations. At such times, the system indicates "trouble" on the annunciator panel in the control room. A 100-
percent reserve bank is provided for each Halon system. The Halon systems are main-tained and tested based on
NFPA 12A-1973. Particular consideration The Halon system design and should also be given application considered con-to: centration and soak time;
toxicity and corrosive char-a) minimum required acteristics. Halon concentra-
tion and soak time b) toxicity of Halon
c) toxicity and corrosive charac-
teristics of thermal decomposi-tion products of
Halon.
- 5. Carbon Dioxide Suppression System The use of carbon No carbon dioxide extinguishing dioxide extinguishing systems are used in the power-systems should as a block buildings.
minimum comply with the requirments of NFPA 12, "Carbon Dioxide Extinguishing Systems." Rev. 11 WOLF CREEK TABLE 9.5A-1 (Sheet 56) APCSB 9.5-1 Appendix A WCGS 6. Portable Extinguishers Fire extinguishers Fire extinguishers are provided should be provided in based on NFPA 10-1975. (10A accordance with guide- has been dropped from the NFPA lines of NFPA 10 and Codes.) All extinguishers are 10A, "Portable Fire installed with consideration Extinguishers Instal- given to cleanup problems and lation, Maintenance adverse effects to equipment and Use." Dry chem- in the hazard area. ical extinguishers should be installed with due consideration given to cleanup problems after use and possible adverse effects on equipment installed in the area. F. Guidelines for Specific Plant Areas 1. Primary and Secondary Containment a) Normal Operation The lubricating oil system for each reactor coolant pump is Fire protection re- provided with enclosures and quirements for the drip collection pans to con- primary and secon- tain and drain away from the dary containment pump any leakage from this areas should be system. provided on the basis of specific High pressure portions of the identified lube oil system are totally hazards. For enclosed with low point drain example: connections. Low pressure portions of the system are - Lubricating oil provided with drip pans with or hydraulic low point connections. Remote fluid system lube oil fill lines for the upper for the primary and lower bearing reservoirs on coolant pumps each reactor coolant pump motor are not protected by drip pans. Due to the design of the fill lines, no lube oil leakage is postulated. The RTD Conduit Boxes (3 per motor) are not provided with drip pans, however, conduit seals and leak tight fittings are used to minimize lube oil leakage. Oil leakage at the RDT Conduit Box does not represent a fire hazard. All low point connections are piped to a remote oil collection tank (greater than 300-gallon tank for each two reactor Rev. 12 WOLF CREEK TABLE 9.5A-1 (Sheet 57) APCSB 9.5-1 Appendix A WCGS
- Cable tray coolant pumps - each pump arrangements lube oil system holds 265 and cable gallons of oil) located inside penetrations the reactor building. The
tanks have level indication
- Charcoal and level alarm annunciation filters in the control room. The tank
vent is equipped with a flame Fire suppression arrestor. Refer to Figure systems should be 9.5.1-3 for the general provided based on arrangement of the oil collec-the fire hazards tion system. The location of analysis. cable trays in the vicinity of the reactor coolant pumps Fixed fan suppres- is also indicated in Figure sion capability 9.5.1-3. should be provided for hazards that A description of the Class 1E could jeopardize cable trays in the vicinity of safe plant shut- the reactor coolant pumps is down. Automatic described in Appendix 9.5B. sprinklers are preferred. The cable penetration areas in the reactor building are pro-tected by a remote, manually actuated preaction sprinkler system. A detection system is provided for the Class 1E cable trays in the reactor building. b) Refueling and Maintenance Refueling and Management procedures and controls
maintenance necessary to assure adequate fire operations in protection during refueling or containment may maintenance are discussed in Items
introduce addi- B.3 and B.3.a). tional hazards such as contamination
control materials, decontamination supplies, wood planking, temporary wiring, welding and flame cutting
(with portable com-pressed fuel gas Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 58)
APCSB 9.5-1 Appendix A WCGS supply). Possible fires would not necessarily be in the vicinity of
fixed detection and suppression systems. Management procedures and controls necessary
to assure adequate fire protection are discussed in Section 3a. In addition, manual Standpipes with hose stations fire-fighting and portable fire extinguishers
capability should are located to protect all be permanently areas with fixed combustible installed in materials.
containment. Standpipes with hose stations and
portable fire extinguishers should be in-
stalled at stra-tegic locations throughout con-
tainment for any required manual fire-fighting
operations. Adequate self- Self-contain breathing contained breath- apparatus are available for ing apparatus use inside the containment. should be provided They are stored near the near the contain- entrance of the RCA. ment entrances for
fire-fighting and
damage control personnel. These units should be
independent of any breathing appara-tus or air supply systems provided for general plant activities. Rev. 9 WOLF CREEK TABLE 9.5A-1 (Sheet 59) APCSB 9.5-1 Appendix A WCGS
- 2. Control Room The control room is The control room is separated essential to safe from other areas of the plant reactor operation. by 3-hour-rated walls, floor, It must be protected and ceiling.
against disabling fire damage and should be separated from other areas of the plant by floors, walls, and roofs having minimum fire resistance rating of 3 hours. Control room cabinets and consoles are subject to damage from two fire hazards: a) Fire originating within a cabinet or console; and b) Exposure fire involving combus-tibles in the general room area. Hose stations adjacent Hose stations for the control to the control room room are located in the foyer with portable extin- and vestibule area just outside guishers in the con- the control room. Portable trol room are accept- extinguishers are located able. within the room. Hose sta-tions are equipped with Class Nozzles that are com- "C" spray nozzles with rubber patible with the hazard bumpers. These hose stations and equipment in the permit coverage of the area control room should be above the suspended ceiling. provided for the manual hose station. The nozzels chosed should satisfy actual fire-fighting needs, satisfy electrical safety, and minimize physical damage to electrical equipment from hose stream impingement. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 60) APCSB 9.5-1 Appendix A WCGS Fire detection in the General area products of control room cabinets combustion detectors are and consoles should be provided in the ceiling of the provided by smoke and control room and at the ceil-
heat detectors in each ing in the area behind the fire area. Alarm and control room proper. Ioniza-annunciation should be tion duct detectors are pro-
provided in the con- vided in the cabinet area trol room. Fire return air duct. In addition, alarms in other parts cabinets which contain redun-
of the plant should dant safe shutdown circuits also be alarmed and have detectors installed annunciated in the inside the cabinets. All fire control room. alarms in the plant are alarmed and annunciated in the control room. Breathing apparatus for Breathing apparatus are available control room operators for control room operators. The
should be readily control room is separated from available. Control adjacent areas by 3-hour-rated room floors, ceilings, walls. The floor and ceiling, supporting structures including structural steel, are and walls, including also rated for 3 hours. All penetrations and penetration seals are relatively
doors, should be airtight. The control building designed to a minimum ventilation system is equipped fire rating of 3 with a smoke detector in the out-
hours. All penetra- side air intake. The control of tion seals should be the system for isolation or vent-airtight. The control ing is manual.
room ventilation intake should be provided with smoke
detection capability to automatically alarm locally and isolate
the control room ventilation system to protect operators by
preventing smoke from entering the control room. Manually oper-
ated venting of the control room should be available so that
operators have the Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 61) APCSB 9.5-1 Appendix A WCGS option of venting for visibility. Manually operated ventilation systems are accept-
able. Cables should not be Cable trenches in the floor of located in concealed the control room are provided floor and ceiling with fixed automatic total spaces. All cables flooding Halon systems.
that enter the control room should terminate in the control room. That is, no cabling should be simply routed through the
control room from one area to another. If such concealed spaces
are used, however, they should have fixed automatic total flood-
ing Halon protection.
- 3. Cable Spreading Room a) The preferred accept-able methods are:
(1) Automatic water The cable spreading rooms are system such as protected by an automatic preaction
closed head sprinkler system installed in the sprinklers, ceiling of each room. Location open head del- of the sprinkler heads considers
uge, or open cable tray sizing and arrangement. directional Cables are designed to allow wett-spray nozzles. ing down with deluge water with-
Deluge and out electrical faulting. The open spray sprinkler system is equipped with systems should closed heads.
have provisions for manual operation at
a remote station; how-ever, there
should also be provisions to Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 62) APCSB 9.5-1 Appendix A WCGS preclude inad-vertent opera-tion. Location of sprinkler
heads or spray nozzles should consider cable
tray sizing and arrangements to assure adequate
water coverage. Cables should be designed to allow wetting down with del-uge water with-
out electrical faulting. Open head deluge and
open direction-al spray sys-tems should be
zoned so that a single failure will not de-
prive the en-tire area of automatic fire
suppression capability. The use of foam
is a type cap-able of being delivered by a
sprinkler or deluge system, such as an
Aqueous Film Forming Foam (AFFF).
(2) Manual hoses Manual hose stations and and portable portable extinguishers are
extinguishers located in the area for backup should be protection. provided as
backup. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 63) APCSB 9.5-1 Appendix A WCGS (3) Each cable The two cable spreading rooms spreading room are separated vertically by the of each unit control room. A 3-hour fire should have barrier separates the CSR from
divisional adjacent areas. The floor and cable separa- ceiling of room 3501 is 3-hour tion, and be rated. The floor of room 3801 is separated from 3-hour rated however, the ceiling the other and is formed by the roof of the the rest of Control Building. The roof is not the plant by a a fire barrier. It is however minimum 3-hour constructed of non-combustible rated fire concrete with built up Class A
wall (Refer to roofing. In addition, the NFPA 251 or structural steel supporting the ASTM E-119 for roof is protected with 3 hour
fire test rated fire proofing. Therefore, resistance building structural integrity rating). is maintained.
(4) At least two Two remote and separate en-remote and trances are provided to the
separate room. entrances are provided to
the room for access by fire brigade
personnel; and (5) Aisle separ- Generally, aisle separation ation pro- between tray stacks is 3 feet vided between wide by 7 feet high.
tray stacks should be at least 3 feet
wide and 8 feet high. b) For cable spread- Not Applicable. ing rooms that do not provide divi-
sional cable separation of (a)(3),... Rev. 17 WOLF CREEK TABLE 9.5A-1 (Sheet 64) APCSB 9.5-1 Appendix A WCGS c) As an alternate to Not Applicable.
(a)(1) above, automatically initiated gas
systems (Halon or CO2) may be used for primary fire
suppression, provided a fixed water system is
used as a backup. d) Plants that cannot Not Applicable. meet the guide-lines of Regula-tory Guide 1.75,...
- 4. Plant Computer Room Safety-related com- Not Applicable.
puters should be separated from other areas of the plant by barriers having a minimum 3-hour fire resistant rating. Automatic fire detec-tion should be pro-vided to alarm and annunciate in the control room and alarm
locally. Manual hose stations and portable water and Halon fire
extinguishers should be provided.
- 5. Switchgear Rooms Switchgear rooms should Control building switchgear be separated from the rooms are separated from the remainder of the plant remainder of the plant and by minimum 3-hour rated from each other by a 3-hour
fire barriers to the barrier. Automatic Halon 1301 extent practicable. extinguishing systems are Automatic fire detec- provided in each switchgear
tion should alarm and room. The detection system annunciate in the alarms and annunciates in the Rev. 12 WOLF CREEK TABLE 9.5A-1 (Sheet 65) APCSB 9.5-1 Appendix A WCGS control room and alarm control room and locally. locally. Fire hose Hose stations and portable stations and portable extinguishers are available in extinguishers should the area.
be readily available. Cables which enter the switch-Cables entering the gear room(s) without terminating switchgear room(s) that there are minimized. do not terminate there should be kept at a minimum. Acceptable protection
for cables that pass through the switchgear room is automatic
water or gas agent suppression. Such automatic suppression must consider pre-venting unacceptable damage to electrical
equipment and possible necessary containment of agent following
discharge.
- 6. Remote Safety-Related Panels The general area housing The areas housing the remote remote safety-related safety-related panels are panels should be protected by a detection provided with auto- system which alarms and annun-
matic fire detectors ciates locally and in the that alarm locally and control room. Combustible alarm and annunciate materials are controlled and
in the control room. limited to those required for Combustible materials operation. Portable extin-should be controlled guishers and manual hose and limited to those stations are provided. required for opera-tion. Portable ex-
tinguishers and manual hose stations should be provided.
- 7. Station Battery Rooms Battery rooms should be Battery rooms are separated protected against from each other and from the fire explosions. rest of the plant by 3-hour Battery rooms should fire barrier walls, floors, and Rev. 8 WOLF CREEK TABLE 9.5A-1 (Sheet 66)
APCSB 9.5-1 Appendix A WCGS be separated from ceilings. Ventilation sys-each other and other tems in the battery rooms are areas of the plant by capable of maintaining the barriers having a hydrogen concentration well
minimum fire rating below two-volume percent. of 3-hours inclusive Portable extinguishers and of all penetrations hose stations are provided.
and openings. (See NFPA 69, "Standard on Portable extinguishers and Explosion Prevention hose stations are provided in
Systems.") Venti- the corridor outside of the lation systems in the rooms. battery rooms should
be capable of main-taining the hydrogen concentration well
below 2 vol. % hydro-gen concentration. Standpipe and hose
portable extin-guishers should be provided. Alternatives: a) Provide a total The battery rooms are served fire rated barrier by two systems--the control enclosure of the building supply air system and
battery room the Class IE unit. Loss of complex that either or both of these sys-exceeds the fire tems will be alarmed in the
load contained in control room via the plant the room. computer. Each battery room is also provided with a hydro-
b) Reduce the fire gen detector which will alarm load to be within the control room whenever the the fire barrier hydrogen concentration exceeds
capability of 1- 1 volume percent in any one of 1/2 hours. the battery rooms. OR The control building supply air system supplies outside air to c) Provide a remote each of the four dc switchgear
manual actuated rooms. This air is exhausted sprinkler system from the switchgear rooms by in each room and means of the control building Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 67) APCSB 9.5-1 Appendix A WCGS provide the 1-1/2 exhaust system. The supply hour fire barrier air system provide approxi-separation. mately 1 air change per hour in each battery room. Battery rooms 1, 3, 2 and 4 are each served by the Class IE ac
system. Each battery room is supplied and exhausted separ-ately. The Class IE ac sys-
tems each operate in a com-plete recirculating mode at all times. Since these sys-tems also serve their respec-tive ESF switchgear and dc switchgear rooms, it has been
conservatively calculated that with no fresh air the system can operate for approximately
3 days before the hydrogen concentration reaches 3 volume percent. All ductwork penetrations of the battery rooms are provided
with 3-hour fire dampers.
- 8. Turbine Lubrication and Control Oil Storage and Use Areas A blank fire wall hav- The turbine oil system is ing a minimum resis- located in the turbine build-tance rating of three ing which is separated by a hours should separate 3-hour barrier from buildings all areas containing housing safe shutdown safety-related sys- equipment. Automatic wet
tems and equipment sprinklers are provided in from the turbine oil the turbine oil reservoir room system. and turbine oil storage tank
room. When a blank wall is not present, open
head deluge protec-tion should be pro-vided for the turbine
oil hazards and automatic open head Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 68) APCSB 9.5-1 Appendix A WCGS water curtain pro-tection should be provided for wall openings.
- 9. Diesel Generator Areas Diesel generators The diesel generators are sep-should be separated arated from each other and
from each other and other areas of the plant by other areas of the fire barriers having a minimum plant by fire bar- fire resistance of 3 hours.
riers having a min- The ceiling is an 18-inch imum fire resistance thick concrete slab which is rating of three supported by structural steel
hours. which has been fireproofed for 3 hours. No safety-related equipment is located above the
building's ceiling. Automatic fire suppres- Automatic preaction sprinkler sion such as AFFF foam system is provided in each or sprinklers should room. Fire detectors are be installed to combat installed in the ceiling of
any diesel generator the room. The detectors alarm or lubricating oil locally and in the control fires. Automatic fire room. Drainage is provided
detection should be for firefighting water, and provided to alarm and smoke and heat venting is annunciate in the provided utilizing the normal control room and alarm ventilation exhaust air flow path. locally. Drainage for fire fighting water Each diesel fuel oil day tank
and means for local is provided with protection manual venting of features to preclude the smoke should be pro- uncontrolled leakage of diesel
vided. fuel. The design features provided for the day tank were When day tanks cannot reviewed and accepted by the be separated from the NRC at the Wolf Creek Fire diesel generator, one Protection Audit of February 6 of the following to 9, 1984.
should be provided for the diesel generator area: Rev. 11 WOLF CREEK TABLE 9.5A-1 (Sheet 69) APCSB 9.5-1 Appendix A WCGS a) Automatic open head deluge head spray nozzle system(s) b) Automatic closed head sprinklers c) Automatic AFFF that is delivered
by a sprinkler deluge or spray system d) Automatic gas system (Halon or
CO2) may be used in lieu of foam or sprinklers to
combat diesel generator and/or lubricating oil
fires.
- 10. Diesel Fuel Oil Storage Areas Diesel fuel oil tanks The diesel oil storage tanks with a capacity are buried approximately 23 greater than 1100 feet from the diesel generator gallons should not be building wall.
located inside the buildings containing safety-related equip-
ment. They should be located at least 50 feet from any building containing safety-related equipment or, if located within 50
feet, they should be housed in a separate building with con-struction having a minimum fire resist-ance rating of 3
hours. Buried tanks Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 70) APCSB 9.5-1 Appendix A WCGS are considered as meeting the 3-hour fire resistance re-quirements. See NFPA
30, "Flammable and Combustible Liquids Code", for additional
guidance. When located in a separate building, the tank should be pro-tected by an automatic fire suppression system such as AFFF or sprinklers. Tanks, unless buried, should not be located
directly above or below safety-related systems or equipment regardless
of the fire rating of separating floors or ceilings.
- 11. Safety-Related Pumps Pump houses and rooms The fire hazards analysis, housing safety-related Appendix 9.5B, indicates that a pumps should be pro- fixed suppression system is not tected by automatic required in the safety-related sprinkler protection pump rooms and houses. Early unless a fire hazards warning fire detection is
analysis can demon- installed with alarm and annun-strate that a fire ciation locally and in the does not endanger other control room. Local hose
safety-related equip- station and portable extin-ment required for safe guishers are provided. plant shutdown. Early warning fire detection should be installed with alarm and annun-
ciation locally and in the control room. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 71) APCSB 9.5-1 Appendix A WCGS Local hose stations and portable extinguishers should also be provided.
- 12. New Fuel Area Hand portable extin- Hand portable extinguishers guishers should be and hose stations are located located within this throughout the fuel building.
area. Also, local An automatic fire detection hose stations should system is installed which be located outside but alarms and annunciates in the within hose reach of control room and locally. this area. Automatic Combustibles are limited to a fire detection should minimum. The storage area is
alarm and annunciate provided with a drainage in the control room system to preclude accumula-and alarm locally. tion of water.
Combustibles should be limited to a minimum in the new fuel area.
The storage area should be provided with a drainage system
to preclude accumula-tion of water. The storage configura- New fuel storage is designed tion of new fuels for optimum moderation condi-should always be so tions. Refer to Section 9.1.
maintained as to preclude criticality for any water density
that might occur during water applica-tion.
- 13. Fuel Storage Pool Area Protection for the Local hose stations and por- fuel storage pool area table extinguishers are pro- should be provided by vided in this area. Automatic local hose stations fire detection is provided to and portable extin- alarm and annunciate in the guishers. Automatic control room and locally. Rev. 14 WOLF CREEK TABLE 9.5A-1 (Sheet 72)
APCSB 9.5-1 Appendix A WCGS fire detection should be provided to alarm and annunciate in the control room and to
alarm locally.
- 14. Radwaste Building The radwaste building With the exception of the Radwaste should be separate Tunnel, the Radwaste Building is from other areas of Physically separated from the rest the plant by fire of the power block by at least 20 feet. barriers having at The Radwaste Tunnel (Fire Area RW-1) is least 3-hour ratings separated from the connected Auxiliary Automatic sprinklers Building by a 3-hour rated fire barrier should be used in all (wall). Power block buildings with safe areas where combus- shutdown or safety-related components have tible materials are 3-hour rated fire barriers (walls) on the located. Automatic side nearest the Radwaste Building. An fire protection should automatic sprinkler system is provided be provided to annun- over the dry waste compactor. . ciate and alarm in the Automatic detection is provided control room and alarm to annunciate and alarm in the locally. During a control room and locally. The fire, the ventilation. ventilation systems in this systems in these areas building are capable of being should be capable of isolated. Firefighting water being isolated. Water drains to liquid radwaste should drain to liquid building sumps. Portable radwaste building extinguishers and hose stations sumps. are provided throughout the building Acceptable alternative fire protection is automatic fire detec-tion to alarm and annunciate in the control room, in addi-tion to manual hose stations and portable extinguishers consist-ing of hand held and large wheeled units. Rev. 11 WOLF CREEK TABLE 9.5A-1 (Sheet 73)
APCSB 9.5-1 Appendix A WCGS
- 15. Decontamination Areas The decontamination Decontamination areas (hot areas should be pro- machine shop and access control
tected by automatic area) which contain flammable sprinklers if flammable liquids are protected by detec-liquids are stored. tors which alarm locally and in
Automatic fire detec- the control room. Portable tion should be provided extinguishers and hose stations to annunciate and alarm are also provided in these
in the control room and areas. The ventilation systems alarm locally. The in these areas are capable of ventilation system being isolated. The access
should be capable of control area of the control being isolated. Local building is provided with an hose stations and hand automatic wet-pipe sprinkler
portable extinguishers system. should be provided as backup to the sprinkler The combustible loading in the
system. hot machine shop is low. The hot machine shop is not a safety-related area, and is separated
from adjacent safety-related areas by a 3-hour fire barrier. There-fore, an automatic sprinkler sys-
tem is not installed.
- 16. Safety-Related Water Tanks Storage tanks that Safety-related water tanks are supply water for safe located in areas which contain shutdown should be low quantities of protected from the combustibles.
effects of fire. Local hose stations Hydrants and hose houses and portable extin- equipped with 250 feet of hose
guishers should be are provided within 250 feet located in nearby hose of all outdoor safety-related houses. Combustible water tanks.
materials should not be stored next to outdoor tanks. A min-
imum of 50 feet of separation should be provided between out-
door tanks and combus-tible materials where feasible. Rev. 13 WOLF CREEK TABLE 9.5A-1 (Sheet 74) APCSB 9.5-1 Appendix A WCGS
- 17. Cooling Towers Not applicable.
- 18. Miscellaneous Areas Miscellaneous areas Miscellaneous structures are such as records stor- located such that a postulated age areas, shops, fire within these structures or warehouses, and aux- the effects of a fire, includ-
iliary boiler rooms ing smoke, does not pose a should be so located hazard to safety-related struc-that a fire or effects tures and systems required for
of a fire, including safe shutdown. The heating oil smoke, will not storage tank is provided with a adversely affect any berm to contain the entire fuel
safety-related systems oil tank contents. or equipment. Fuel oil tanks for auxil-
iary boilers should be buried or pro-vided with dikes to contain the entire tank contents. G. Special Protection Guide-lines 1. Welding and Cutting, Acetylene-Oxygen Fuel Gas Systems This equipment is used Welding and cutting and in various areas acetylene-oxygen fuel gas throughout the plant. systems are controlled by a Storage locations permit system. Bottled gases are should be chosen to stored only in areas protected permit fire protection by automatic sprinklers or in by automatic sprinkler isolated structures. An exception systems. Local hose is the flammable gas cylinder stations and portable storage in Room 3102, which is equipment should be necessary for analysis activities provided as backup. in the Hot Laboratory. Hand hose The requirements of lines and portable extinguishers NFPA 51 and 51B are are located at all storage areas. applicable to these hazards. A permit system should be re- quired to utilize this equipment. (Also refer to 2f herein.) Rev. 19 WOLF CREEK TABLE 9.5A-1 (Sheet 75) APCSB 9.5-1 Appendix A WCGS
- 2. Storage Areas for Dry Ion Exchange Resins Dry ion exchange resins Dry ion exchange resins are not should not be stored stored near essential safety-near essential safety- related systems. New resins related systems. Dry are received normally in a
unused resins should be hydrated form and do not consti-protected by automatic tute a fire hazard. Only the wet pipe sprinkler containers in which the resins
installations. Detec- are stored are combustible. tion by smoke and heat The spent resins are sluiced detectors should alarm to the spent resin storage tank
and annunciate in the located in the radwaste building, control room and alarm which is not a safety-related locally. Local hose building. Administrative con-
stations and portable trols ensure that resin in extinguishers should quantities required for immed-provide backup for iate use only are introduced these areas. Storage into safety-related areas, and areas of dry resin the containers are hauled away should have curbs and as soon as they are emptied.
drains. (Refer to NFPA 92M, "Waterproofing and In addition, detection, por-Draining of Floors.") table extinguishers, and hose
stations are provided in these areas. Storage areas are provided with drains.
- 3. Hazardous Chemicals Hazardous chemicals Hazardous chemicals are stored should be stored and and protected in accordance protected in accord- with the recommendations of ance with the reactor NFPA 49-1975. Storage areas commendations of NFPA are ventilated and drained.
49, "Hazardous
Chemicals Data." Chemical storage areas should be well venti-
lated and protected against flooding conditions since some
chemicals may react with water to produce ignition. Rev. 1 WOLF CREEK TABLE 9.5A-1 (Sheet 76) APCSB 9.5-1 Appendix A WCGS
- 4. Materials Containing Radioactivity Materials that collect Materials that collect and and contain radioacti- contain radioactivity are vity such as spent ion stored in closed metal tanks or exchange resins, char- containers located in areas
coal filters, and HEPA free from ignition sources. filters should be stored in closed metal
tanks or containers that are located in areas free from igni-
tion sources or com-bustibles. These materials should be
protected from exposure to fires in adjacent areas as well. Con-sideration should be given to requirements for removal of isotopic
decay heat from en-trained radioactive materials. Rev. 1 WOLF CREEK APPENDIX 9.5B FIRE HAZARDS ANALYSES The USAR FHA has been superseded by the following documents: E-1F9905, Fire Hazard Analysis. E-1F9900, Post-Fire Safe Shutdown Manual Actions. E-1F9910, Post-Fire Safe Shutdown Fire Area Analysis. E-1F9915, Design Basis Document for OFN RP-017, Control Room Evacuation. XX-E-013, Post-Fire Safe Shutdown (PFSSD) Analysis. M-663-00017A, Fire Protection Evaluations for Unique
or Unbounded Fire Barrier Configurations. The above documents are incorporated by Reference within the USAR.
9.5B-0 Rev. 29
WOLF CREEK APPENDIX 9.5C RESPONSES TO QUESTIONS CONTAINED IN THE NRC'S LETTERS DATED NOVEMBER 3, 1977 FROM OLAN D. PARR TO THE SNUPPS UTILITY APPLICANTS (Note that all section and page numbers referenced by the NRC in the questions contained in this Appendix are those contained in the original Fire Protection
Report, dated April 1, 1977) 9.5C-0 Rev. 0 WOLF CREEK Item 1. (General) (RSP) Throughout the fire hazards analysis, you state that the fire ratings of barriers protecting various safety related equipment, such as the control room, auxiliary building, battery room, and cable
spreading room, are below that specified in Appendix
A to BTP 9.5-1. It is our position that all walls, ceilings, floors, and associated penetrations which enclose separate safety related fire areas shall
have a minimum fire rating of three hours. State your intent with regard to this staff position.
(See related items 23, 24, 26, 27, 28, 29 and 31.)
Response: See Section 9.5.1.2.2.3, Figure 9.5.1-2 and Appendix 9.5B. 9.5C-1 Rev. 0 WOLF CREEK Item 2. Section 9.5.1.2.2, Page 9.5-3, and Table 9.5-5, Sheet 4) Describe the design basis and criteria for the automatic sprinkler systems used in vertical cable chases in the auxiliary building and control
building. Provide assurance that the associated
drain systems are adequately designed. Describe how manual fire fighting and smoke removal can be accomplished. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Automatic Wet Pipe Sprinkler Systems" and the appropriate fire areas in the Fire Hazards Analysis, Appendix 9.5B. 9.5C-2 Rev. 0 WOLF CREEK Item 3. (Section 9.5.1.2.2, Page 9.5-5) Substantiate the fire resistance capacity of the following items by verifying that their construction is in accordance with a particular design that has
been fire tested:
(a) rated fire barriers, (b) fire barrier penetration seals, (c) fire stops in cable trays, (d) fire dampers/fire doors, as well as how they are installed in the ventilation ducts that
penetrate fire rated barriers of safety related areas, and (e) metal deck roof. Also identify the design and the test method used and the acceptance criteria. Response: Refer to Section 9.5.1.2.2.3. 9.5C-3 Rev. 0 WOLF CREEK Item 4. (Section 9.5.1.2.2., Page 9.5-5) Confirm that the fire alarm system meets the requirements for Class A systems, as defined in NFPA 72D, and also meets the requirements for Class 1
circuits as stated in the National Electrical Code
Alarms. Response: Refer to Section 9.5.1.2.2.2. 9.5C-4 Rev. 0 WOLF CREEK Item 5. (Section 9.5.1.3, Page 9.5-7) Identify safety related areas where a hose station may be blocked by a fire in that particular area. For any such case, indicate the location of
alternate accessible hose stations. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Standpipes and Hose Racks." 9.5C-5 Rev. 0 WOLF CREEK Item 6. (Table 9.5-4, Sheet 7) You state that redundant systems necessary for safe shutdown are separated such that both trains are not subject to the same fire hazard. Identify all the
redundant systems necessary for safe shutdown which
are separated only by distance. Describe the fire
detection system and the primary and backup suppression system provided for these systems necessary for safe shutdown. Provide the results of
an analysis to demonstrate that for a postulated fire, including an exposure fire and failure of the
primary fire suppression system, that safe shutdown can still be accomplished. For any cases where safe shutdown cannot be assured, describe the additional measures which will be taken. Response: Refer to Appendix 9.5B. 9.5C-6 Rev. 0 WOLF CREEK Item 7. (Table 9.5-4, Sheet 12) If approved Class A elevator, pressure, water-tight, and missile-resistant doors are not available for this plant application, describe how the doors to be
used will be shown to be the equivalent of Class A
doors. Response: Refer to Section 9.5.1.2.2.3.d. 9.5C-7 Rev. 0 WOLF CREEK Item 8. (Table 9.5-4, Sheet 19) Describe the manner in which fire and smoke are either isolated or ventilated via the normal and/or emergency ventilation system in all parts of the
auxiliary, radwaste, fuel, and control buildings. Include in your discussions, (a) the equipment used, (b) operator actions required for this operation, (c) the control access to the equipment and (d) the
ability of the equipment to handle high temperature
gases and particles. Response: Refer to Section 9.5.1.2.2.4. 9.5C-8 Rev. 0 WOLF CREEK Item 9. (Table 9.5-4, Sheet 24) (RSP) Your response to Section D.5 of BTP 9.5-1 Appendix A is not acceptable. It is our position that fixed self-contained emergency lighting, consisting of
fluorescent or sealed beam units with individual 8-
hour minimum battery power supplies, should be
provided in areas which must be manned for safe shutdown and for access and egress routes to all fire areas. Safe shutdown areas include those
required to be manned if the control room must be evacuated. Also, a portable radio communications
system should be provided for use by the fire brigade and other operations personnel involved in safe plant shutdown. State your intent with regard to this staff position. Response: Refer to Section 9.5.1.2.2.5, Other Fire Protection Features, paragraphs entitled "Emergency Lighting
and Radio Communication." 9.5C-9 Rev. 0 WOLF CREEK Item 10. (Table 9.5-4, Sheet 28) (RSP) Your response to Section E.3.(d) of BTP 9.5-1 Appendix A is not acceptable. It is our position that provisions be made to supply water, at least to
standpipes and hose connections, for manual fire
fighting in areas within hose reach of equipment
required for safe shutdown following an SSE. The standpipe system serving such hose stations should be analyzed for SSE loading and should be provided
with supports to assure system pressure integrity. The piping and valves for the portion of hose
standpipe system affected by this functional requirement should at least satisfy ANSI Standard B 31.1, "Power Piping." The water supply for this condition should be obtained from a normal seismic
Category I water system. State your intent with
regard to this staff position. Response: Refer to Section 9.5.1.2.2.1. 9.5C-10 Rev. 0 WOLF CREEK Item 11. (Table 9.5-4, Sheet 28) You state that hose stations within the containment will be on 150 foot centers rather than 100 foot centers. Confirm that placement of hose stations
within the containment will provide adequate
coverage such that the effective hose streams will
cover all safety related areas. Consideration should be given to situations where the access to a single hose station may be blocked by a fire; in such cases
an alternate hose station should be provided. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Standpipes and Hose Racks." 9.5C-11 Rev. 0 WOLF CREEK Item 12. (Table 9.5-4, Sheet 29) Describe the design basis and criteria used to determine the Halon requirements for the various rooms containing safety related equipment. Provide
data including soak times, concentrations, and mode
of detection for all rooms protected by a gaseous
system. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Halon 1301 System." 9.5C-12 Rev. 0 WOLF CREEK Item 13. (Table 9.5-4, Sheet 33 and Table 9.5-5, Sheet 39) Where safety related control rooms and auxiliary shutdown and control cabinets contain cables from both separation groups, describe the consequences of
a fire in each cabinet. Consider the resultant
damage to wiring and instrumentation in adjoining
safety related cabinets due to heat and corrosive vapors. Confirm that smoke detectors are provided in the cabinets and consoles in the control rooms or
justify acceptable alternatives. Describe any additional measures necessary to assure safe
shutdown. Response: Refer to Sections A.28 and C.27 of Appendix 9.5B. 9.5C-13 Rev. 0 WOLF CREEK Item 14. (Table 9.5-4, Sheet 38) Confirm that loss of the ventilation system for the battery rooms will alarm in the control room. Describe the operation of the battery room
ventilation system. Response: Refer to Appendix 9.5B, Sections C.15 and C.16. 9.5C-14 Rev. 0 WOLF CREEK Item 15. (Table 9.5-4, Sheet 39) In the event of a fire in the turbine building which might cause the turbine building roof to collapse, provide information to assure that safe shutdown can
still be accomplished. Response: Refer to Appendix 9.5B, Section T.2. 9.5C-15 Rev. 0 WOLF CREEK Item 16. (Table 9.5-4, Sheet 42) Provide justification for the deviation you have taken to the requirement of Appendix A to BTP 9.5-1 that requires automatic sprinklers for areas which
contain flammable liquids. Response: Refer to Appendix 9.5B, Sections HMS.1, C.5, and C.6. 9.5C-16 Rev. 0 WOLF CREEK Item 17. (Table 9.5-4 Sheet 44) Provide either (a) supporting evidence for your statement that the dry ion exchange resins used in the SNUPPS power block will not burn, or (b) revise
your fire protection system to provide automatic
fire suppression in any safety related area where
ion exchange resins are stored. Response: Refer to Section 9.5.1.2.2.5, paragraph entitled "Ion Exchange Resins," and Appendix 9.5B, Section A.8. 9.5C-17 Rev. 0 WOLF CREEK Item 18. (Table 9.5-5, Sheet 1) Where an automatic fire suppression system is installed, the fire hazards analysis should be made to demonstrate that the plant can achieve its safe
shutdown during a fire assuming malfunction of the
fixed automatic fire suppression system in the
area. Revise your fire hazards analysis to reflect changes resulting from this assumption. Response: Refer to Appendix 9.5B. 9.5C-18 Rev. 0 WOLF CREEK Item 19. (Table 9.5-5, Sheet 2, paragraph b) For fire protection we consider safe shutdown to be the cold shutdown operational mode. Revise your fire hazards analysis and your discussion of systems
required for safe shutdown (PSAR Section 7.4.1), as
necessary, to reflect this change. Response: Refer to Appendix 9.5B. 9.5C-19 Rev. 0 WOLF CREEK Item 20. (Table 9.5-5, Sheet 2, paragraph d) You state the Class 1E conduit and instrumentation are not included in the fire hazards analysis due to incomplete design. Provide the design basis and
criteria for assuring protection of exposed Class 1E
conduit or Class 1E instrumentation. Response: Refer to Section 8.3.1.4. 9.5C-20 Rev. 0 WOLF CREEK Item 21. (Table 9.5-5, Sheet 2, paragraph f) Provide the basis for assigning a value of 12,250 Btu/pound as the total heat release of electrical cable insulation. Response: Refer to Appendix 9.5B. 9.5C-21 Rev. 0 WOLF CREEK Item 22. (Table 9.5-5, Sheet 4) You state that Separation Groups 1 and 4 cable trays in Zone 1128 are horizontally separated by 15 feet and the total combustible loading is low. However, no provisions are indicated for fire protection and
detection for this zone. Discuss in your analysis
an exposure fire involving both separation groups and the affect on shutdown of an exposure fire assuming failure of any fixed suppression system. Response: Refer to Appendix 9.5B, Section A.1. 9.5C-22 Rev. 0 WOLF CREEK Item 23. (Table 9.5-5, Sheet 8) Discuss your rationale for not using a 3-hour fire rated wall to separate the safety related equipment in the event of a fire due to transient combustibles
in this area. Also describe your proposed method
for manual suppression of fires occurring in the
cable trays at the 40 foot ceiling or provide fixed suppression. Response: Refer to Appendix 9.5B, Section A.3. 9.5C-23 Rev. 0 WOLF CREEK Item 24. (Table 9.5-5, Sheet 11) Provide either (a) 3-hour fire rate construction for isolating the volume control tank, seal water heat exchangers, and containment isolation valves from
the remainder of the plant or (b) justification for
not providing 3-hour fire rated construction. Response: Refer to Appendix 9.5B, Sections A.8, A.24, and A.25. 9.5C-24 Rev. 0 WOLF CREEK Item 25. (Table 9.5-5, Sheets 17 and 20) Provide either (a) separation of redundant safety related equipment by fire rated barriers or (b) justification for not separating redundant safety
related equipment by fire rated barriers. Redundant safe shutdown cable separation group in room 1406 are separated by 9.5 feet. A similar situation exists in Fire Area A-19. Analyze the
affects on plant shutdown due to an exposure fire
that can effect both groups. Response: Refer to Appendix 9.5B, Sections A.16 and A.19. 9.5C-25 Rev. 0 WOLF CREEK Item 26. (Table 9.5-5. Sheet 21) Zones 1502 and 1503 do not appear to provide a 3-hour fire rated barrier between the component cooling water surge tanks. Therefore, provide for a
3-hour barrier or provide justification for not
providing such a barrier for protection against
fires due to transient combustibles in the area. Response: Refer to Appendix 9.5B, Section A.20. 9.5C-26 Rev. 0 WOLF CREEK Item 27. (Table 9.5-5, Sheet 22) Confirm that the concrete block wall separating the control room HVAC equipment is 3-hour fire rated. Also, describe the fire protection provided for the
charcoal filters in these rooms. Response: Refer to Appendix 9.5B, Sections A.21, A.22, and C.27. 9.5C-27 Rev. 0 WOLF CREEK Item 28. (Table 9.5-5, Sheet 23) Confirm that safety related equipment in Zones 1411, 1412, 1508 and 1509 are separated by means of 3-hour fire rated barriers. Revise the fire hazards
analysis considering the exposure fire effects on
redundant divisions and the need for fire rated
barriers. Response: Refer to Appendix 9.5B, Section A.23. 9.5C-28 Rev. 0 WOLF CREEK Item 29. (Table 9.5-5, Sheet 25) Discuss the consequences of an exposure fire with regard to the redundant ESW motor operated valves in Zone 3101. Also provide justification for the lack
of a 3-hour fire rated ceiling. Response: Refer to Appendix 9.5B, Section C.1. 9.5C-29 Rev. 0 WOLF CREEK Item 30. (Table 9.5-5, Sheet 29) Describe how manual fire fighting of cable fires can be conducted above the suspended ceiling in the access control area. Response: Refer to Appendix 9.5B, Sections C.5 and C.6. 9.5C-30 Rev. 0 WOLF CREEK Item 31. (Table 9.5-5, Sheet 34) Provide justification for 2-hour fire-rated floors and ceilings in Fire Areas C-15 and C-16. Also confirm that hose stations can reach all the various
rooms in these areas. Response: Refer to Appendix 9.5B, Sections C.15 and C.16. 9.5C-31 Rev. 0 WOLF CREEK Item 32. (Table 9.5-5, Sheet 39) Confirm that safety related cable is not installed in concealed spaces in the control room ceiling. If such cable installation exists, discuss the means
for fire detection and suppression. Response: Refer to Appendix 9.5B, Section C.27. 9.5C-32 Rev. 0 WOLF CREEK Item 33. (Table 9.5.5, Sheet 41) Your fire hazard analysis does not cover all areas adjacent to the control room. Discuss the fire exposure to the control room from all adjacent
rooms, such as Zones 3603, 3606, and 3609. Response: Refer to Appendix 9.5B, Section C.27. 9.5C-33 Rev. 0 WOLF CREEK Item 34. (Table 9.5-5, Sheet 44) Discuss the consequences of an exposure fire involving both safety trains in Zones 6104 and 6105 of the fuel building, at elevation 2000 feet. Response: Refer to Appendix 9.5B, Sections F.2 and F.3. 9.5C-34 Rev. 0 WOLF CREEK Item 35. (Table 9.5-5, Sheet 46) It is our position that manual fire fighting capability shall be permanently installed in containment. Therefore, either state your intent to
comply with this position or provide justification
for not installing this capability in containment. Response: Refer to Appendix 9.5B, Section RB. 9.5C-35 Rev. 0 WOLF CREEK Item 36. (Table 9.5-5, Sheet 46) Confirm that a system, to collect and contain lubricating oil for each reactor coolant pump, will be provided to assure that oil will not leak or
spray from the pump or lubricating system. State
the capacity of the collection system relative to
the pump lube oil inventory, and provide isometric drawings of the system showing it as it will be installed on the reactor coolant pump. Provide
additional arrangement drawings showing the location of the reactor coolant pumps and Class 1E cable
trays, including the separation distance and fire suppression equipment. Provide justification for not providing sprinklers to extinguish a potential oil fire. Response: Refer to Appendix 9.5B, Section RB. 9.5C-36 Rev. 0 WOLF CREEK Item 37. (Table 9.5-5, Sheet 47) Separation between redundant groups does not take into consideration an exposure fire. Discuss the possibility of an exposure fire involving both
safety divisions in the reactor building, elevation
2000 feet, cable tray area. Response: Refer to Appendix 9.5B, Section RB. 9.5C-37 Rev. 0 WOLF CREEK Item 38. (Table 9.5-5, Sheets 47, 48, and 49) Provide justification for not installing sprinklers in Fire Areas RB-2 and RB-5. Also, justify the use of manual sprinklers rather than automatic
sprinklers in Fire Areas RB-3, RB-4, RB-7 and RB-8. Response: Refer to Appendix 9.5B, Section RB. 9.5C-38 Rev. 0 WOLF CREEK Item 39. (Table 9.5-5, Sheet 51) Confirm that the hose stations provided at the ends of the radwaste pipe tunnel and cable chase can effectively reach all areas of the tunnel between
the radwaste building and the auxiliary building.
Describe the provisions for venting and draining the
tunnel. Response: Refer to Appendix 9.5B, Section RW.1. 9.5C-39 Rev. 0 WOLF CREEK Item 40. (Table 9.5-5, Sheet 56) (RSP) The emergency fuel oil day tanks are not in a separate enclosure. Appendix A to BTP 9.5-1 permits the day tanks to be installed in the diesel
generator area only if they are located in a
separate enclosure and protected by an automatic
fire suppression system. Therefore, we require that you comply with Appendix A in this regard or provide justification for deviating from Appendix A. Response: Refer to Section 9.5.1.2.2.5, paragraph entitled "Combustible Oil." 9.5C-40 Rev. 0 WOLF CREEK Item 41. (Table 9.5-5, Sheet 56) Confirm that hose stations installed for the protection of the diesel generators will reach all portions of the diesel generator building. Also
confirm that the ceiling of the diesel generator
room has a 3-hour fire rating. Response: Refer to Appendix 9.5B, Sections D.1 and D.2. 9.5C-41 Rev. 0 WOLF CREEK Item 42. (Table 9.5-4, Sheet 29) (RSP) You state that the Halon systems will be tested at 6-month intervals, in conformance with NFPA recommendations. We require that testing be
performed at 3-month intervals. State your intent
with regard to this staff position. Response: Testing frequencies are defined in plant procedures. 9.5C-42 Rev. 0 WOLF CREEK Item 43. The QA program for fire protection should be under the management control of the QA organization. This
control consists of (1) either formulating a fire protection QA program that incorporates suitable requirements and is acceptable to management
responsible for fire protection, or verifying that
the program incorporates suitable requirements and
is acceptable to management responsible for fire protection, and (2) verifying the effectiveness of the QA program for fire protection through review, surveillance, and audit. Revise the response to Mr. Boyd's letter of September 30, 1976 to clarify that
the QA program for fire protection is under the management control of QA, or provide an alternative position for the staff's evaluation. Response: Refer to a separate quality document entitled "Quality Assurance Programs for Design and
Construction." 9.5C-43 Rev. 0 WOLF CREEK Item 44. Sheet 5 of Table 9.5-4 of the response to Mr. Boyd's letter of September 30, 1976 states: "Design of the
fire protection system by Bechtel has proceeded on the basis of normal checking and reviews within Bechtel Engineering as specified by existing
procedures. Future design activities will be
carried out on the same basis..." To confirm the
acceptability of the WCGS's design control and procurement document control program for fire protection, we need a clearer definition of the
related Bechtel activities. This should be provided by either (1) committing that these activities are
in accordance with the programs described in the Bechtel QA topical report BQ-TOP-1 or (2) providing a similar level of detail in Table 9.5-4 for the staff's evaluations. Response: See response to Question 43. 9.5C-44 Rev. 0 WOLF CREEK Item 45. Under the QA program criteria in part C of APCSB 9.5-1 Appendix A, it appears that Criterion 5, Test
and Test Control, is not applicable to WCGS' activities. Conversely, it appears that the other criteria are applicable. Under the WCGS column of
Table 9.5-4, it is not clear that WCGS will meet
these criteria to the extent that the organization
is involved in these activities. Provide such clarification or identify those criteria which the WCGS organization will not meet. Response: See response to Question 43. 9.5C-45 Rev. 0 WOLF CREEK Item 46. D. B. Vassallo's letter of August 29, 1977 on fire protection provides supplemental guidance on quality
assurance. Modify your response to Mr. Boyd's letter of September 30, 1976 so that your letter is also responsive to this latest supplemental guidance
on quality assurance for fire protection. Response: Refer to a separate quality document entitled "Quality Assurance Programs for Design and
Construction." 9.5C-46 Rev. 0 WOLF CREEK APPENDIX 9.5D RESPONSES TO QUESTIONS CONTAINED IN THE NRC'S LETTERS DATED OCTOBER 18, 1979 FROM OLAN D. PARR TO THE SNUPPS UTILITY APPLICANTS (Note that all section and page numbers referenced by the NRC in the Questions contained in this Appendix are those contained in the first revision of the Fire Protection Report, dated May 3, 1978.) 9.5D-0 Rev. 0 WOLF CREEK Item 1. (Page R5-1) In the response to question "Item 5" you state that inside the containment hose stations may be 150 ft. apart and that an extra length of hose must be added
to the hose station if required. This arrangement is
unacceptable. It is our position, as stated in
Section E3d of Appendix E to BTP 9.5-1, that standpipes equipped with a maximum of 75 ft. of 1-1/2 in. woven-jacket, lined fire hose with suitable
nozzles be provided for all elevations inside containment. In addition, hose stations should be
spaced at no more than 100 ft. intervals. Revise your design accordingly. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Standpipes and Hose Racks." 9.5D-1 Rev. 0 WOLF CREEK Item 2. (Page 9.5A-4) Your response to Section B of BTP 9.5-1, Appendix A, "Administrative Procedures, Controls and Fire Brigade" is adequate. Confirm that you will follow
the staff supplemental guidance contained in
"Nuclear Plant Fire Protection Functional
Responsibilities, Administrative Controls and Quality Assurance," dated June 14, 1977. Response: See Table 9.5A Item B1. 9.5D-2 Rev. 0 WOLF CREEK Item 3. (Page 9.5-1) You state that a single failure in the fire protection system will not impair both the primary and backup fire suppression capabilities. However, Figure 9.5-1 (Sheet 2) fire protection system, shows
both the north and south cable penetration areas
protected by an automatic sprinkler system fed off the hose station standpipe system for the containment. A loss of this single penetration will
leave this area without primary and backup fire suppression systems which is not consistent with
your stated design objective. It is our position, as stated in Section A4 of Appendix A to BTP 9.5-1, that a single failure in the fire suppression system should not impair both the primary and backup fire
suppression capability. Revise your design to meet
our guidelines. Response: Refer to Appendix 9.5B, Section RB.4. 9.5D-3 Rev. 0 WOLF CREEK Item 4. (Page 9.5-4a, Item C) It is our position, as stated in Section 4C.c(4) of Appendix A, that fire stops be installed every 20 ft. along horizontal cable routings in areas that
are not protected by automatic fire suppression
systems (AFSS). Vertical cable routings should have
fire stops installed at each floor/ceiling level. Between levels or in vertical cable chases, fire stops should be installed at the mid-height if the
vertical run is 20 ft. or more but less than 30 ft. or at 15 ft. intervals in vertical runs of 30 ft. or
more unless such vertical cable routings are protected by AFSS directed on the cable trays. Individual fire stop designs should prevent the propagation of a fire for a minimum period of 30
min. when tested for the largest number of cable
routings and maximum cable density. Revise your design accordingly.
Response
Refer to Section 9.5.1.2.2 3. 9.5D-4 Rev. 0 WOLF CREEK Item 5. (Page 9.5-5, Fire and Smoke Detection and Alarm System) You state that the fire and smoke detection system is powered by the non-class IE dc system which is backed by a battery charger supplied from the
emergency power supply. This arrangement is
unacceptable. It is our position, as stated in Section E.1.(a) of Appendix A, that primary and secondary power supplies should be provided for the
fire detection system and for electrically operated control valves for automatic suppression systems by:
(a) Using normal offsite power as the primary supply, with a 4-hour battery supply as secondary supply; and (b) Having capability for manual connection to the Class 1E emergency power bus within 4 hours of loss of offsite power. Such
connection should follow the applicable
guidelines in Regulatory Guides 1.6, 1.32
and 1.75. Revise your design accordingly. Response: Refer to Section 9.5.1.2.2.2. 9.5D-5 Rev. 0 WOLF CREEK Item 6. (Page 9.5-6a, Safety Evaluation Two) You state that in most areas of high fire loading, a backup system will be available in case of failure of the primary suppression system in a given area.
However, the backup system is a portable
extinguisher or a hose station. It is our position, as stated in Section E.3(d) of Appendix A, that portable extinguishers, due to their limited capacity and effectiveness, are not considered as
secondary protection. Hose stations should be provided so that all areas of the plant can be
properly protected. Revise your design accordingly. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Standpipes and Hose Racks." 9.5D-6 Rev. 0 WOLF CREEK Item 7. It is our position, as stated in Section F2 of Appendix A, that an automatic fire detection and suppression system be provided to protect the area
above the suspended ceiling in the control room and
adjacent offices, and the access control area. The
system should be actuated by a cross-zoned smoke detection system with alarm and annunciation in the control room. Revise your design accordingly. Response: Refer to Appendix 9.5B, Section C.27. 9.5D-7 Rev. 0 WOLF CREEK Item 8. (Page 9.5A-23) Verify that self-contained breathing apparatus will have at least two extra air bottles located on site for each unit. Also it is our position, as stated
in Section D4(h) of Appendix A, that you should
provide an onsite 6 hr. supply of reserve air so
arranged to permit quick and complete replenishment of exhausted supply air bottles as they are returned. State your intent with regard to this
position. Response: See Table 9.5A-1, Item 4h. 9.5D-8 Rev. 0 WOLF CREEK Item 9. (Page 9.5A-27, Item b) Your response is incomplete. It is our position, as stated in Section E3(b) of Appendix A, that all control and sectionalizing valves in the fire water
system should be electrically supervised. The
electrical supervision signal should indicate in the
control room. Otherwise, a management supervision program should be provided. Such a program should include locking valves open with strict key control, tamper-proof seals, and periodic visual check of all valves. Revise your design accordingly. Response: Refer to Section 9.5.1.2.2.2, Fire and Smoke Detection and Alarm System. 9.5D-9 Rev. 0 WOLF CREEK Item 10. (Page 9.5A-29) You state that the Halon system is designed to maintain a 5 percent minimum concentration at the ceiling for 10 minutes. This system design is
unacceptable. It is our position, as stated in
Section E4 of Appendix A, that the total flooding
Halon 1301 concentration be increased to extinguish a deep-seated cable tray fire. Revise your system design and fire hazard analysis accordingly. Response: Refer to Section 9.5.1.2.2.1, paragraph entitled "Halon 1301 System." 9.5D-10 Rev. 0 WOLF CREEK Item 11. (Page 9.5A-39a) It is our position, as stated in Section F10 of Appendix A, that the diesel generator day tanks be limited to a maximum of 1100 gal. and that if a
diked enclosure is provided, that it have sufficient
capacity to hold 110% of the contents of the day
tank and drained to a safe location. Also, hose stations should be provided for secondary protection for the diesel generator area in case of failure of
the primary system (the pre-action system). Revise your design accordingly. Response: Refer to Appendix 9.5B, Section D.1. 9.5D-11 Rev. 0 WOLF CREEK Item 12. (Page 9.5A-41) It is our position, as stated in Section F11 of Appendix A, that both early warning fire detection with alarm and annunciation and hose stations be
provided for protection of the essential service
water pumphouse in addition to the proposed fire
extinguishers. Revise your design accordingly. Response: Refer to Item F.11 of Table 9.5A-1. 9.5D-12 Rev. 0 WOLF CREEK Item 13. (Page 9.5A-44, Welding and Cutting, Acetylene Oxygen Gas System) You indicate that portable extinguishers and hose stations are provided for the storage locations of welding and cutting, acetylene oxygen gas systems.
This provision is unacceptable. It is our position, as stated in Section D2(b) of Appendix A, that gas cylinder storage locations should not be in areas that contain or expose safety-related equipment or
the fire protection systems that serve those safety-related areas. A permit system should be required
to use this equipment in safety related areas of the plant. In addition, storage locations should be chosen to permit fire protection by automatic sprinkler systems. Local hose stations and portable
equipment should be provided as backup. Revise your
design accordingly. Response: Refer to Table 9.5A-1, Section D.2 and G.1. 9.5D-13 Rev. 0 WOLF CREEK Item 14. (Page 9.5B-2, Section 2.3) In Section 2.3 of your analysis, you state that separation of the devices for nuclear safety-related controls and instrumentation will be achieved by
physical separation or barriers between separation
groups from the same protective function in
accordance with Regulatory Guide 1.75. This separation distance is unacceptable since Regulatory Guide 1.75 does not consider the consequence of an
exposure fire. It is our position, as stated in D1(a) of Appendix A, that separate fire areas for
each division of safety-related systems be provided to separate redundant systems from a common (exposure) fire. Particular design attention to the use of separate isolated fire areas for redundant
cables should be provided to avoid loss of redundant
safety-related cables. Revise your design accordingly.
Response
Refer to Appendix 9.5B. 9.5D-14 Rev. 0 WOLF CREEK Item 15. (Page 9.5B-9, Section A.1.4.2)
(a) You state that an automatic preaction-type sprinkler system is installed over cable trays in Zones 1101, 1120, 1121, and 1122 with a design
density for the system of 0.3 gpm/ft2 for the most
remote 3000 sq. ft. of top surfaces of all trays.
This arrangement is unacceptable. It is our position that sufficient sprinkler heads be added or relocated (throughout the plant) such that the
design density flow be at the floor level and not at the top of cable trays. This was noticed throughout
your fire hazards analysis. Revise your design accordingly.
(b) Throughout your fire hazards analysis, you identify the location of various safety-related
cable trays and conduits that are separated by distance only. Your proposed design is unacceptable. It is our position, as stated in
Section D1(2) of Appendix A, that 3 hr. fire rated
barriers be provided to separate the redundant
conduits and/or cable trays. Revise your design accordingly. This applies in the following areas: Fire Area A-1 (9.5B-7) Reference Figure 9.5-1a, elevation 1974
- 1. Zone 1101
- 2. Between Zones 1130 and 1101
- 3. Between Zones 1122 and 1101
- 4. Zone 1128
- 5. Zone 1206 9.5D-15 Rev. 0 WOLF CREEK Table A.1 for Fire Area A-1, Failure Modes and Analysis Section 1-b.
Section 2-b. Section 3-b. Section 4-a. Section 9-b. Fire Area A-2 (9.5B-19) Reference Figure 9.5-1a, elevation 1974'
- 1. Zone 1114 Fire Area A-3 (9.5B-23) Reference Figure 9.5-1a-1c, elevation 1974', 2000', 2026'.
- 1. Between Zones 1116 and 1117 Fire Area A-6 (9.5B-33) Reference Figure 9.4-1a-1d, elevation 1974', 2000', 2026', 2047'.
- 1. In Zone 1127, Section A.6.7.2.1 Fire Area A-7 (9.5B-3c) Reference Figure 9.5-1a, elevation 1974'
- 1. In Zone 1126 Fire Area A-16 (9.5B-67) Reference Figure 9.5-1c, elevation 2026'
- 1. Zone 1401
- 2. Zone 1406 Table A.16 for Fire Area A-16
Section 1 Failure Modes & Analysis 3 a, b 4 b 9.5D-16 Rev. 0 WOLF CREEK Section 2 1 a, b, c 2 a 3 b 4 b 10 b
Section 3 1 b 2 c, d Fire Area A-20 (9.5B-90) Reference Figure 9.5-1D
- 1. Zone 1502
- 2. Zone 1403 Fire Area A-23 (9.5B-100) Reference Figure 9.5-1c-d
- 1. Zone 1508
- 2. Zone 1509
- 3. Zone 1412
- 4. Zone 1411 Fire Area A-27 (9.5B-112) Reference Figure 9.5-1c
- 1. Zone 1403, Section A.27.7.2.2 Fire Area A-28 (9.5B-116) Reference Figure 9.5-1c
- 1. Zone 1413, Section A.28.7.1 9.5D-17 Rev. 0 WOLF CREEK Fire Area C-1 (9.5-118) Reference Figure 9.5-1a
- 1. Zone 3101 Response: (a) Refer to Appendix 9.5B, Section A.1.4 (b) Refer to Appendix 9.5B. 9.5D-18 Rev. 0 WOLF CREEK Item 16. (Page 9.5B-115, Fire Area A-28, Reference Figure 9.5-1c, Zone 1413)
You state in your Fire Hazards Analysis how various safety-related cable trays, conduit and equipment are separated by distance from their redundant
counterpart, and the criteria that were used to
establish barriers between these redundant trains. In order to provide a defense-in-depth design, so that a fire will not prevent the performance of
necessary safe plant shutdown functions, a detailed fire hazards analysis should be conducted for each
plant area. It is essential that the analysis include the effects of postulated fire involving permanent and/or transient combustibles (exposure fires) on systems, circuit cable trays or equipment
required for safe plant cold shutdown. The fire
hazards analysis should identify all the redundant mechanical and electric systems and components necessary for safe cold shutdown which are separated
only by distance (no fire barriers and with
redundant trains 20 ft. or less from each other).
Redundant trains within 20 ft. of each other, as a minimum, will be required to be protected by a half hour fire rated barrier as well as area automatic
sprinklers. This does not mean that in some
instances, such as the cable spreading room, relay
room, and 460V and 4160V switchgear rooms, redundant trains separated by more than 20 ft. will not require additional protection. The fire hazards analyses need to demonstrate that, assuming failure of the primary suppression system, a fire on installed or transient combustibles will not result in the loss of capability to achieve safe
cold shutdown. Where this cannot be demonstrated, an alternative means of assuming safe plant shutdown
(cold shutdown) should be provided.
Demonstrate that:
(a) Safe shutdown from the main control room where a fire disables any safe shutdown equipment including conduit/cable trays controlled from remote locations. 9.5D-19 Rev. 0 WOLF CREEK (b) Safe shutdown from remote locations when the main control room is uninhabitable due to a fire
or when fire disables safe shutdown equipment of the relay room or 460V switchgear room or 4160V switchgear room. Remote location need only be provided for the essential instrumentation, controls and equipment necessary to bring the plant to a hot standby condition. Fire damage to systems necessary to
achieve and maintain cold shutdown should be limited so that repairs can be made and cold shutdown
achieved within 72 hours. A detailed breakdown of staff requirements (Attachment 1) is enclosed for:
(a) Minimum safe shutdown systems when one division of all safety systems is not available.
(b) Minimum fire protection when dedicated or alternate shutdown systems are provided.
Provide the requested information.
Response
Refer to Appendix 9.5B. 9.5D-20 Rev. 0 WOLF CREEK Item 17. (Page 9.5B-24, Fire Area A-3, Figure 9.5-1a-1c, Elevations 1974, 2000, 2026) It appears from the drawings that a fire on the upper elevations of fire area A-3 can propagate through the open grating to the redundant pieces of
safety related equipment below, namely the boric
acid tank and transfer pump on elevation 1974. This arrangement is unacceptable. It is our position that a 3 hr. fire rated barrier separate these two
pieces of safety-related equipment such that a fire will not jeopardize both trains. Revise your design
accordingly. Response: Refer to Appendix 9.5B, Section A.3. 9.5D-21 Rev. 0 WOLF CREEK Item 18. (Pages 9.5B-39, 9.5B-47, 9.5B-49, 9.5B-104, 9.5B-107, 9.5B-109) Fire Area A-9 (9.5B-47) Reference Figure 9.5-1b-9.5-1c Zone 1309, Section A.9.4.1 Fire Area A-10 (9.5B-49) Reference Figure 9.5-1b-9.5-1C Zone 1310, Section A.10.4.1 Fire Area A-24 (9.5B-104) Reference Figure 9.4-1b Zone 1323, Section A.24.4.1 Fire Area A-25 (9.5B-107) Reference Figure 9.5-1b Zone 1322, Section A.25.4.1 Fire Area A-26 (9.5B-109) Reference Figure 9.5-1c Zone 1405, Section A.25.4.1 It is our position, as stated in Section E1(b) of Appendix A, that automatic smoke detectors that
alarm locally and in the control room be installed in all areas containing safety-related equipment and/or conduit/cable. Revise your design
accordingly. Response: Refer to Appendix 9.5E, Section III.F. 9.5D-22 Rev. 0 WOLF CREEK Item 19. (Page 9.5B-64) Fire Area A-15 Reference Figure 9.5-lb Zone 1331, Section A.15.4.1 Because of the slow response of the rate-compensated fire detector to a small or incipient fire, provide automatic smoke detectors for protection for this area. The detectors should alarm and annunciate in
the control room. Response: Refer to Appendix 9.5B, Section A.15. 9.5D-23 Rev. 0 WOLF CREEK Item 20. (Pages 9.5B-86, 9.5B-90, 9.5B-94, 9.5B-97, 9.5B-100, 9.5B-204, 9.5B-207) It is our position, as stated in Section d1(j) of Appendix A, that the ceiling of the following fire areas be fire rated for 3 hours for protection of
various safety-related equipment as well as
conduit/cable trays in the area. Revise your design accordingly. Fire Area A-8, Reference Figure 9.5-1b Fire Area A-9, Reference Figure 9.5-1b and 9.5-1c Fire Area A-22, Reference Figure 9.5-1d Fire Area A-23, Reference Figure 9.5-1c-d Fire Area C-33, Reference Figure 9.5-1d Fire Area C-34, Reference Figure 9.5-1d
Response: Refer to Appendix 9.5B. 9.5D-24 Rev. 0 WOLF CREEK Item 21. (Page 9.5B-118, Fire Area C-1, Reference Figure 9.4-1a, Zone 1301) You state that fire detection system is not provided in this area, where the ESW isolation valves for both trains are located, and only portable
extinguishers are provided for fire-fighting
purpose. Your proposal is unacceptable. It is our position, as stated in Section E of Appendix A, that both an automatic smoke detection system as well as
an automatic sprinkler system be provided for this area due to the restricted mobility of various areas
of this room for manual firefighting. Both systems should alarm and annunciate in the control room. Revise your design accordingly. Response: Refer to Appendix 9.5B, Section C.1. 9.5D-25 Rev. 0 WOLF CREEK Item 22. (Page 9 5B-126, Fire Area C-4, Reference Figure 9.5-1a, Section C.4.4.4) You indicate that this area is separated from the horizontal cable (safety-related cable) area above a 2-hour-rated suspended ceiling. Because the fire
loading in this area is appreciable, to protect the
safety related cable against an exposure fire, the sprinkler system should cover all rooms of fire area C-4. Revise your design accordingly. Response: Refer to Appendix 9.5B, Sections C.5 and C.6. 9.5D-26 Rev. 0 WOLF CREEK Item 23. For the following listed areas, you state that failure by fire of all circuits contained in these areas will not prevent safe shutdown of the plant
and that the identification of the circuits in these
areas, which feed redundant safe shutdown equipment, will be made and the detailed analysis to demonstrate that their failure will not prevent safe shutdown will be provided in the FSAR.
(a) Fire Area C-5 9.5B-129, Reference Figure 9.5-1a, Section C.5.7.2 (b) Fire Area C-7 9.5B-133, Reference Figure 9.5-1a, Section C.7.7.2 (c) Fire Area C-8 9.5B-136, Reference Figure 9.5-1a, Section C.8.7.2 (d) Fire Area C-9 9.5B-139, Reference Figure 9.5-1b, Section C.9.7.2 (e) Fire Area C-10 9.5B-142, Reference Figure 9.5-1b, Section C.10.7.2 (f) Fire Area C-11 9.5B-145, Reference Figure 9.5B-1b, Section C.11.7.2 (g) Fire Area C-12 9.5B-148, Reference Figure 9.5-1b, Section C.12.7.2 (h) Fire Area C-15 9.5B-155, Reference Figure 9.5-1b, Section C.15.7.2 (i) Fire Area C-16 9.5B-158, Reference Figure 9.5-1b, Section C.16.7.2 (j) Fire Area C-17 9.5B-161, Reference Figure 9.5-1b, Section C.17.7.2 (k) Fire Area C-18 9.5B-164, Reference Figure 9.5-1b, Section C.18.7.2 (1) Fire Area C-21 9.5B-173, Reference Figure 9.5-1c, Section C.21.7.2 9.5D-27 Rev. 0 WOLF CREEK (m) Fire Area C-22 9.5B-176, Reference Figure 9.5-1c, Section C.22.7.2 (n) Fire Area C-30 9.5B-198, Reference Figure 9.5-1d, Section C.30.7.2 (o) Fire Area C-31 9.5B-200, Reference Figure 9.5.1d, Section C.31.7.2 (p) Fire Area C-33 9.5B-204, Reference Figure 9.5.1d, Section C.33.7.2 Response: Refer to Appendix 9.5B 9.5D-28 Rev. 0 WOLF CREEK Item 24. (Pages 9.5B-179, 9.5B-182)
The following areas are missing a part of the fire hazards analysis which deals with the safe shutdown capability with failure by fire of all circuits and
equipment in the fire area.
(a) Fire Area C-23, Reference Figure 9.5-1C, Section C.23.7 (b) Fire Area C-24, Reference Figure 9.5-1C, Section C.24.7 Revise your fire hazard analysis and modify your fire protection program accordingly.
Response: Refer to Appendix 9.5B. 9.5D-29 Rev. 0 WOLF CREEK Item 25. Page 9.5B-190, Fire Area C-27, Reference Figure 9.5.1d, Section C.27.2.3 (a) You indicate that some electrical cables are routed above the suspended ceiling over the main control room. It is our position, as
stated in Section F2 of Appendix A, that an
automatic fire suppression be provided for protection of the cables running above the suspended ceiling in the control room. The
system should be activated by an automatic smoke detection system located above the
ceiling. Revise your design accordingly.
(b) Verify that the area above the suspended ceiling is not being used for a supply or
exhaust air plenum for the control room.
Verify the fire rating in regards to flame spread, fuel contributed and smoke developed when tested under UL E-84 fire test.
(c) Describe how the cables from separation groups 2, 4 and 6 are protected as they pass through the suspended ceiling above the control room.
(d) It is our position, as stated in Section F2 of Appendix A, that automatic smoke detectors, which alarm and annunciate in the control room be provided for all cabinet and consoles that contain redundant safety-related conduit, cable and wiring. Revise your design accordingly.
(e) It is also our position that the peripheral rooms in the control room complex (within the 3 hr. fire rated wall) should have an automatic
smoke detector installed as well as each room
be separated from the control room by 1 hr.
fire rated construction, including all door openings. Revise your design accordingly.
(f) It is also our position, as stated in Section F2 of Appendix A, that the outside air intakes
for the control room ventilation system be provided with smoke detection capability to alarm in the control room to enable manual
isolation of the control room ventilation system and thus prevent smoke from entering the control room. Revise your design accordingly. 9.5D-30 Rev. 0 WOLF CREEK Response: Refer to Appendix 9.5B, Section C.27. 9.5D-31 Rev. 0 WOLF CREEK Item 26. (Page 9 5B-209, Fire Area F-1, Reference Figure 9.5-1b-d, Section F.1.4.1) It is our position, as stated in Section F12, F13 of Appendix A, that automatic detectors be provided throughout the fuel building and not just at the
refueling floor, elevation 2046 ft., 6 in. All
detectors should alarm and annunciate in the control room. Revise your design accordingly. Response: Refer to Appendix 9.5B, Sections F.1 through F.7. 9.5D-32 Rev. 0 WOLF CREEK Item 27. (Page 9.5B-224, Fire Area RB, Reference Figure 9.5-1b, Section RB.4.1) Specify the type of automatic fire detectors that will be installed above each reactor coolant pump. It is our position, as stated in Section F1 of
Appendix A, that an automatic fire detection system
be provided throughout the reactor building and not just over the reactor coolant pumps. Detectors should alarm and annunciate in the control room.
Revise your design accordingly. Response: Refer to Appendix 9.5B, Section RB.4. 9.5D-33 Rev. 0 WOLF CREEK Item 28. Verify that the structural steel of the hot machine shop as well as the fuel handling building are not structurally tied into the fire wall separating
these areas from the auxiliary building so that a
fire in the non-safety related equipment side will
not affect the safety-related equipment and conduit/cable trays on the other side. Response: Refer to Appendix 9.5B, Sections HMS-1 and F.1. 9.5D-34 Rev. 0 WOLF CREEK Item 29. (Page 9.5B-245 Fire Area D-1, Reference Figures 9.5-1b, Section D.1.4, Fire Area D-2, Page 9.5B-248, Section D.2.4) It is our position, as stated in Section D1(i) of Appendix A, that the floor drains as well as the trench of each diesel generator room should have provisions for preventing the spread of fire throughout the drain system. Demonstrate that a flammable liquid fire will not spread to the
adjacent room; otherwise revise your design accordingly. Response: Refer to Appendix 9.5B, Section D.1. 9.5D-35 Rev. 0 WOLF CREEK Item 30. The response to item 43 indicates that the review and audit program of the A/E's QA organization does not extend to fire protection. As indicated in
Mr. Vassallo's letter of August 29, 1977 and the
"For Comment" editions of Regulatory Guide 1.120, it
is our position that the A/E's QA organization should verify the effectiveness of the A/E's QA program for fire protection through review, surveillance, and audits. This is in addition to supervisory controls by the A/E and audits by
others. Revise the response to item 43 accordingly. Response: See Item C of Table 9.5A-1. 9.5D-36 Rev. 0 WOLF CREEK Item 31. State whether the nonconformance control program and
corrective action program for fire protection, described in the A/E's Project Engineering Procedures, meet Sections 17A.1.15 and 17A.1.16 of
the WCGS PSAR. If not, list and justify the
deviations. Response: See Item C7, Table 9.5A-1. 9.5D-37 Rev. 0 WOLF CREEK APPENDIX 9.5E This Section provides a design comparison to 10 CFR 50 Appendix R,Fire Protection Program for Nuclear Power Facilities Operating Priorto January 1, 1979. 9.5E-0 Rev. 0 WOLF CREEK TABLE 9.5E-1 WCGS FIRE PROTECTION COMPARISON TO 10CFR50 Appendix R
10CFR50 Appendix R WCGS III. Specific Requirements A. Water Supplies for Fire Suppression Systems Two separate water supplies Complies. shall be provided to furnish necessary water volume and
pressure to the fire main loop. Each supply shall consist of a storage tank, pump, piping, and appropriate isolation and control valves. Two separate redundant
suctions in one or more intake structures from a large body of water (river, lake, etc.) will
satisfy the requirement for two separated water storage tanks. These supplies shall be separated
so that a failure of one supply will not result in a failure of the other supply. Each supply of the fire water distribution system shall be
capable of providing for a period of 2 hours the maximum expected water demands as determined by
the fire hazards analysis for safety-related areas or other areas that present a fire
exposure hazard to safety-related areas. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 2) 10CFR50 Appendix R WCGS When storage tanks are used for combined service-water/fire-
water uses the minimum volume for fire uses shall be ensured by means of dedicated tanks or
by some physical means such as a vertical standpipe for other water service. Administrative
controls, including locks for tank outlet valves, are unacceptable as the only means
to ensure minimum water volume. Other water systems used as one of the two fire-water supplies shall be permanently connected to the fire main system and
shall be capable of automatic alignment to the fire main system. Pumps, controls, and
power supplies in these systems shall satisfy the requirements for the main fire pumps. The In addition to the standard use of other water systems for fire protection applications, fire protection shall not be the system is also used for incompatible with their fire brigade training and as functions required for safe a backup source of raw water plant shutdown. Failure of the to support plant safe shut other system shall not degrade down for design basis the fire main system. accidents other than fire. III. B. Sectional Isolation Valves Sectional isolation valves such Complies. as post indicator valves or key operated valves shall be in-stalled in the fire main loop Rev. 12 WOLF CREEK TABLE 9.5E-1 (Sheet 3) 10CFR50 Appendix R WCGS to permit isolation of portions of the main fire main loop for
maintenance or repair without interrupting the entire water supply. III. C. Hydrant Isolation Valves Valves shall be installed to Complies. permit isolation of outside hydrants from the fire main
for maintenance or repair without interrupting the water supply to automatic or manual
fire suppression systems in any area containing or presenting a fire hazard to
safety-related or safe shutdown equipment. III. D. Manual Fire Suppression Standpipe and hose systems shall Complies. Wet standpipes be installed so that at least one for power block fire hoses effective hose stream will be able are designed in accordance to reach any location that contains with the requirements for
or presents an exposure fire hazard Class II service of NFPA to structures, systems, or compon- No. 14-1976. Hose racks ents important to safety. are located so that no
more than 100 feet separ-Access to permit effective func- ates adjacent hose racks. tioning of the fire brigade shall Access to permit func-
be provided to all areas that con- tioning of the fire tain or present an exposure fire brigade is adequately pro-hazard to structures, systems, vided and is discussed in
or components important to safety. Appendix 9.5B. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 4) 10CFR50 Appendix R WCGS Standpipe and hose stations shall The standpipe system for be inside PWR containments and the containment is sup-BWR containments that are not plied from the fire main inerted. Standpipe and hose loop through a safety- stations inside containment may grade containment pene-be connected to a high quality tration. The containment water supply of sufficient standpipes are normally quantity and pressure other than dry and may be charged by the fire main loop if plant- plant operations and fire specific features prevent brigade personnel via local extending the fire main and control room actions. supply inside containment. For BWR drywells, standpipe, Each hose rack is pro- and hose stations shall be vided with 75 feet of placed outside the dry well 1-1/2-inch hose, except with adequate lengths of hose the diesel generator, to reach any location inside cable spreading the dry well with an effective rooms, and north end of hose stream. Corridor No. 2, room 1408, of the Aux Building, Elevation 2026, which are protected with 100 feet of hose. III. E. Hydrostatic Hose Tests Fire hose shall be hydrostatically Complies. Hose in outside tested at a pressure of 150 psi hose houses are tested or 50 psi above maximum fire annually at a pressure of 150 main operating pressure, which- psi or 50 psi above maximum ever is greater. Hose stored in fire main operating pressure, outside hose houses shall be whichever is greater. tested annually. Interior Interior standpipe hose is standpipe hose shall be tested tested every 3 years at a every three years. pressure of 150 psi or 50 psi above maximum fire main pressure, whichever is greater or the hose is replaced at least every 5 years. III. F. Automatic Fire Detection Automatic fire detection system Automatic fire and smoke shall be installed in all areas detector systems are pro- of the plant that contain or vided throughout the plant present an exposure fire hazard on the basis of the fire Rev. 18 WOLF CREEK TABLE 9.5E-1 (Sheet 5) 10CFR50 Appendix R WCGS to safe shutdown or safety- hazards analysis and related systems or components. consequences of specific
These fire detection systems postulated fires. A dis-shall be capable of operating cussion of detector types with or without offsite power. and specific locations is
provided in Appendix 9.5B on an area-by-area basis. Table 9.5B-3 provides a list of those fire areas in safety-related
buildings which do not have detectors in every room. These rooms are
listed to indicate if safe shutdown circuits exist. As shown on Table 9.5B-3, a majority of the rooms have no safe shutdown circuits. Fire Areas A-9, A-10, A-29, A-30, C-35, C-36, and C-37 have no detection provided. The
notes to Table 9.5B-3 contain brief descriptions of these areas which form
the basis for not pro-viding detection. Refer to Appendix 9.5B for a
detailed description of these areas. As indicated in Section 9.5.1.2, the fire de-tection system is provided
with a backup battery power supply. The Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 6) 10CFR50 Appendix R WCGS
batteries are served by a battery charger that can be manually connected to the plant emergency ac power supply.
The ESW pumphouse also complies.
III. G. Fire Protection of Safe Shutdown Capability
- 1. Fire protection features E-1F9905, Fire Hazards Analysis, shall be provided for struc- coupled with E-1F9910, Post Fire tures, systems, and components Safe Shutdown Fire Area Analysis, important to safe shutdown. provide an area-by-area analysis These features shall be capable of the powerblock, which demon- of limiting fire danger so strates that no single fire can that: prevent safe shutdown.
- a. One train of systems neces- For a fire outside the control room, sary to achieve and maintain fire protection features are provided hot shutdown conditions from such that post-fire hot standby can either the control room or be achieved and maintained from the emergency control station(s) is control room, with limited reliance on free of fire damage; and operator manual actions outside the control room. Predominantly, redundant safe shutdown components are separated
- b. Systems necessary to by 3-hour fire rated barriers or the achieve and maintain cold shut- equivalent protection identified by down from either the control III.G.2. Fire resistive cable for room or emergency control BNHV8812B, EGHV0016, EGHV0054 and station(s) can be repaired EMHV8801A, which has been successfully within 72 hours.
tested per the requirements of NRC Generic Letter 86-10, Supplement 1, may be used in lieu of the rated fire
- 2. Except as provided for in barrier requirement in III.G.2.a.
paragraph G.3 of this section, In some instances, operator manual actions outside the control room are utilized in lieu of III.G.2 protection. Operator manual actions have been evaluated for feasibility and reliability, considering NUREG-1852 guidance. For redundant trains of systems required to achieve and maintain cold shutdown that could potentially be affected by
Rev. 27 WOLF CREEK TABLE 9.5E-1 (Sheet 7) 10CFR50 Appendix R WCGS where cables or equipment, a single fire, repairs or including associated non- local operator actions can safety circuits that could be performed within 72 prevent operation or cause hours. maloperation due to hot shorts, open circuits, or As described in E-1F9905 shorts to ground, or re- and Section 7.4, an dundant trains of systems auxiliary shutdown panel necessary to achieve and is provided as a dedicated maintain hot shutdown means of achieving and conditions are located maintaining hot standby within the same fire area in the event that the outside of primary con- control room is tainment, one of the following uninhabitable due to a means of ensuring that one fire. of the redundant trains is free of fire damage shall be provided: The ESW pumphouse also complies. a. Separation of cables and equipment and assoc-iated non-safety circuits of In fire area A-8, the volume redundant trains by a fire control tank outlet valves barrier having a 3-hour (BGLCV0112B and BGLCV0112C) and rating. Structural steel circuits are not separated in forming a part of or sup-accordance with Section III.G.2. porting such fire barriers However, the fire protection shall be protected to features provided in fire area provide fire resistance A-8 as well as the low fixed equivalent to that re-combustible loading provides quired of the barrier; reasonable assurance that at least one valve will respond to
- b. Separation of cables a control room close signal and equipment and associated following a fire in the area.
non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no in-tervening combustible or
Rev. 27 WOLF CREEK TABLE 9.5E-1 (Sheet 8) 10CFR50 Appendix R WCGS fire hazards. In addition, fire detectors and an automatic fire
suppression system shall be installed in the fire area; or
- c. Enclosure of cable and equip-ment and associated non-safety circuits of one redundant train
in a fire barrier having a 1-hour rating. In addition, fire detec-tors and an automatic fire sup-
pression system shall be instal-led in the fire area; Inside noninerted containments one of the fire protection means specified above or one of the
following fire protection means shall be provided:
- d. Separation of cables and equipment and associated non-safety circuits of redundant
trains by a horizontal distance of more than 20 feet with no in-tervening combustibles or fire
hazards;
- e. Installation of fire detec-tors and an automatic fire suppres-sion system in the fire area; or
- f. Separation of cables and equipment and associated nonsafety circuits of redundant trains by a
noncombustible radiant energy shield. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 9) 10CFR50 Appendix R WCGS
- 3. Alternative or dedicated shutdown capability and its associated circuits,6 indepen-dent of cables, systems or components in the area, room, or zone under consideration, shall be provided:
- a. Where the protection of systems whose function is
required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section;
or
- b. Where redundant trains of systems required for hot shut-down located in the same fire area may be subject to damage
from fire suppression activi-ties or from the rupture or inadvertent operation of fire
suppression systems. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under
consideration. 6Alternative shutdown capability is provided by rerouting, relocating or modificating of existing systems; dedicated shutdown capability is
provided by installing new structures and systems for the function of post-fire shutdown. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 10) 10CFR50 Appendix R WCGS III. H. Fire Brigade A site fire brigade trained Fire brigade compliment, training, and equipped for fire fighting medical qualification, and personal shall be established to ensure protective equipment complies with adequate manual fire fighting the stipulated requirements. The capability for all areas of the fire brigade leader is not expected plant containing structures, to don a self-contained breathing systems, or components important apparatus (SCBA) in a fire event, to safety. The fire brigade shall since this could significantly deter be at least five members on from his/her primary responsibility each shift. The brigade leader of directing fire brigade actions. and at least two brigade The remaining four fire brigade members shall have sufficient members are required to don a SCBA training in or knowledge of in a fire event presenting an plant safety-related systems to atmosphere immediately dangerous to understand the effects of fire life and health (IDLH). Minimum and fire suppressants on safe emergency breathing air equipment shutdown capability. The and capacity is as follows: qualification of fire brigade members shall include an Fire Brigade: annual physical examination to determine their ability to 1. At least ten (10) face piece perform strenuous fire fighting masks are provided at the Fire activities. The shift brigade locker area. supervisor shall not be a member of the fire brigade. 2. At least four (4) of the The brigade leader shall be face piece masks are within a competent to assess the a complete SCBA unit equipped potential safety consequences with a 1-hour air cylinder that of a fire and advise control is ready for immediate fire room personnel. Such brigade use. The SCBAs are competence by the brigade located at the fire brigade leader may be evidenced by locker area. possession of an operator's license or equivalent knowledge 3. At least four (4) additional of plant safety-related 1-hour cylinders are provided systems. at the fire brigade locker area such that they are readily accessible for change out. These cylinders may be stand alone or part of a complete SCBA unit. 4. An additional 6-hour reserve air supply is provided for each of the four (4) fire brigade members. This supply may be comprised of filled cylinders and/or a refill mechanism(compressor or cascade) conforming to the power and air quality requirements specified in 10 CFR 50 Appendix R III.H. The reserve air supply is located on site and readily accessible for air change out. Rev. 17 WOLF CREEK TABLE 9.5E-1 (Sheet 11) 10CFR50 Appendix R WCGS The minimum equipment provided Control Room: for the brigade shall consist of personal protective 1. At least six (6) complete SCBA equipment such as turnout units equipped with a 1-hour coats, boots, gloves, hard hats, air cylinder that is ready for emergency communications immediate use, are provided in equipment, portable lights, the control room area. portable ventilation equipment, and portable extinguishers. 2. At least six (6) additional 1- Self-contained breathing hour cylinders are provided in apparatus using full-face the control room area such that positive-pressure masks they are readily accessible for approved by NIOSH (National change out. These cylinders Institute for Occupational may be stand alone or part of a Safety and Health approval complete SCBA unit. formerly given by the U.S. Bureau of Mines) shall be 3. An additional 6-hour reserve provided for fire brigade, air supply is provided for each damage control, and control of the six (6) control room room personnel. At least 10 SCBA units. masks shall be avail-able for fire brigade personnel. Damage Control:Control room personnel may be furnished breathing air by a Emergency breathing air supply manifold system piped from a complement is satisfied by the storage reservoir if practical. equipment provided for fire brigade Service or rated operating life and control room use. shall be a minimum of one-half hour for the self-contained units. At least a 1-hour supply of breathing air in extra bottles shall be located on the plant site for each unit of self-contained breathing apparatus. In addition, an onsite 6-hour supply of reserve air shall be provided and arranged to permit quick and complete replenishment of exhausted Rev. 17 WOLF CREEK TABLE 9.5E-1 (Sheet 12) 10CFR50 Appendix R WCGS supply air bottles as they are returned. If compressors are
used as a source of breathing air, only units approved for breathing air shall be used;
compressors shall be operable assuming a loss of offsite power. Special care must be
taken to locate the compressor in areas free of dust and contaminants. III. I. Fire Brigade Training The fire brigade training pro- Complies except a grace gram shall ensure that the capa- period of 31 days is allowed bility to fight potential fires for individuals to make up is established and maintained. missed training sessions. The program shall consist of an A full calendar quarter, initial classroom instruction beyond the 2 year requirement, program followed by periodic is allowed to accommodate classroom instruction, fire scheduling. Refer to USAR fighting practice, and fire Section 9.5.1.7.5.2 for a drills. discussion of Fire Brigade training.
- 1. Instruction
- a. The initial classroom in-struction shall include:
(1) Indoctrination of the plant fire fighting plan with speci-fic identification of each in-dividual's responsibilities.
(2) Identification of the type and location of fire hazards
and associated types of fires that could occur in the plant. Rev. 15 WOLF CREEK TABLE 9.5E-1 (Sheet 13) 10CFR50 Appendix R WCGS (3) The toxic and corrosive characteristics of expected pro-
ducts of combustion.
(4) Identification of the loca-tion of fire fighting equipment for each fire area and familiar-ization with the layout of the
plant, including access and egress routes to each area.
(5) The proper use of available fire fighting equipment and the correct method of fighting each
type of fire. The types of fires covered should include fires in energized electric
equipment, fires in cables and cable trays, hydrogen fires, fires involving flammable and
combustible liquids or hazard-ous process chemicals, fires resulting from construction or
modificatiion (welding), and record file fires.
(6) The proper use of communica-tion, lighting, ventilation, and emergency breathing equipment.
(7) The proper method for fighting fires inside buildings
and confined spaces.
(8) The direction and coordina-tion of the fire fighting activ-ities (fire brigade leaders only). Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 14) 10CFR50 Appendix R WCGS (9) Detailed review of fire fighting strategies and
procedures.
- 10) Review of the latest plant modifications and corresponding changes in fire fighting plans.
Note - Items (9) and (10) may be deleted from the training of no more than two of the non-
operations personnel who may be assigned to the fire brigade.
- b. The instruction shall be provided by qualified indi-viduals who are knowledgeable, experienced and suitably trained in fighting the types of fires that could occur in
the plant and in using the types of equipment available in the nuclear power plant.
- c. Instruction shall be pro-vided to all fire brigade mem-
bers and fire brigade leaders.
- d. Regular planned meetings shall be held at least every 3 months for all brigade mem-bers to review changes in the
fire protection program and other subjects as necessary.
- e. Periodic refresher training sessions shall be held to re-peat the classroom instruction Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 15) 10CFR50 Appendix R WCGS program for all brigade members over a two-year period. These
sessions may be concurrent with the regular planned meetings.
- 2. Practice Practice sessions shall be held for each shift fire brigade on the proper method of fighting the various types of fires that
could occur in a nuclear power plant. These sessions shall provide brigade members with
experience in actual fire extin-guishment and use of emergency breathing apparatus under
strenuous conditions encountered in fire fighting. These prac-tice sessions shall be provided
at least once per year for each fire brigade member.
- 3. Drills
- a. Fire brigade drills shall be performed in the plant so that the fire brigade can practice as a team.
- b. Drills shall be performed at regular intervals not to
exceed 3 months for each shift fire brigade. Each fire brigade member should participate in
each drill, but must participate in at least two drills per year. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 16) 10CFR50 Appendix R WCGS A sufficient number of these drills, but not less than one for each shift fire brigade per year, shall be unannounced to determine the fire fighting
readiness of the plant fire brigade, brigade leader, and fire protection systems and
equipment. Persons planning and authorizing an unannounced drill shall ensure that the
responding shift fire brigade members are not aware that a drill is being planned until
it is begun. Unannounced drills shall not be scheduled closer than four weeks.
At least one drill per year shall be performed on a "back
shift" for each shift fire brigade.
- c. The drills shall be pre-planned to establish the train-ing objectives of the drill
and shall be critiqued to deter-mine how well the training objectives have been met. Un-
announced drills shall be planned and critiqued by mem-bers of the management staff
responsible for plant safety and fire protection. Perform-ance deficiencies of a fire
brigade or of individual fire brigade members shall be remedied by scheduling addi-
Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 17) 10CFR50 Appendix R WCGS tional training for the brigade or members. Unsatisfactory
drill performance shall be followed by a repeat drill within 30 days.
- d. At 3-year intervals a ran-domly selected unannounced
drill shall be critiqued by qualified individuals indepen-dent of the licensee's staff.
A copy of the written report from such individuals shall be available for NRC review.
- e. Drills shall as a minimum include the following:
(1) Assessment of fire alarm effectiveness time, time re-
quired to notify and assemble fire brigade, and selection, placement and use of equipment, and fire fighting strategies.
(2) Assessment of each brigade member's knowledge of his or her role in the fire fighting stra-tegy for the area assumed to
contain the fire. Assessment of the brigade member's con-formance with established
plant fire fighting procedures and use of fire fighting equip-ment, including self-contained
emergency breathing apparatus, communication equipment, and ventilation equipment, to
the extent practicable. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 18) 10CFR50 Appendix R WCGS (3) The simulated use of fire fighting equipment required to
cope with the situation and type of fire selected for the drill. The area and type of
fire chosen for the drill should differ from those used in the previous drill so that
brigade members are trained in fighting fires in various plant areas. The situation selected
should simulate the size and arrangement of a fire that could reasonably occur in the
area selected, allowing for fire development due to the time required to respond, to obtain
equipment, and organize for the fire, assuming loss of automatic suppression capability.
(4) Assessment of brigade leader's direction of the fire
fighting effort as to through-ness, accuracy, and effectiveness.
- 4. Records Individual records of training provided to each fire brigade member, including drill crit-iques, shall be maintained for
at least 3 years to ensure that each member receives training in all parts of the training
program. These reoords of training shall be available for NRC review. Retraining or Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 19) 10CFR50 Appendix R WCGS broadened training for fire fighting within buildings shall
be scheduled for all those brigade members whose perform-ance records show deficiencies. III. J. Emergency Lighting Emergency lighting units with The Power block Complies. at least an 8-hour battery As stated in Section power supply shall be provided 9.5.3.2.3, emergency
in all areas needed for opera- lighting units with tion of safe shutdown equipment eight-hour batteries and in access and egress routes are located in all
thereto. plant areas required for operation of safe shutdown equipment
and also those areas necessary for access and egress. The ESW pumphouse also complies. III. K. Administrative Controls Administrative controls shall Administrative procedures be established to minimize fire define limitations to hazards in areas containing minimize fire hazards in structures, systems, and com- areas containing SSCs ponents important to safety. important to safety. These controls shall establish Administrative procedures procedures to: are also provided to promote prompt, appropriate action 1. Govern the handling and upon discovery of a fire. limitation of the use of ordi- nary combustible materials, com- bustible and flammable gases Rev. 17 WOLF CREEK TABLE 9.5E-1 (Sheet 20) 10CFR50 Appendix R WCGS and liquids, high efficiency particulate air and charcoal
filters, dry ion exchange resins, or other combustible supplies in safety-related
areas.
- 2. Prohibit the storage of combustibles in safety-related areas or establish designated storage areas with appropriate
fire protection.
- 3. Govern the handling of and limit transient fire loads such as combustible and flammable liquids, wood and plastic pro-
ducts, or other combustible materials in buildings containing safety-related systems or equip-
ment during all phases of operating, and especially during maintenance, modification, or
refueling operations.
- 4. Designate the onsite staff member responsible for the in-plant fire protection review of proposed work activities to
identify potential transient fire hazards and specify re-quired additional fire protec-
tion in the work activity procedure.
- 5. Govern the use of ignition sources by use of a flame per-mit system to control welding, flame cutting, brazing, or Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 21) 10CFR50 Appendix R WCGS soldering operations. A separate permit shall be issued
for each area where work is to be done. If work continues over more than one shift, the
permit shall be valid for not more than 24 hours when the plant is operating or for the dura-
tion of a particular job during plant shutdown.
- 6. Control the removal from the areas of all waste, debris, scrap, oil spills, or other
combustibles resulting from the work activity immediately following completion of the
activity, or at the end of each work shift, whichever comes first.
- 7. Maintain the periodic housekeeping inspections to
ensure continued compliance with these administrative controls.
- 8. Control the use of specific combustibles in safety-related areas. All wood used in safety-
related areas during mainten-ance, modification, or refuel-ing operations (such as lay-down
blocks or scaffolding) shall be treated with a flame retardant. Equipment or supplies (such as
new fuel) shipped in untreated combustible packing containers may be unpacked in safety-
related areas if required Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 22) 10CFR50 Appendix R WCGS for valid operating reasons. However, all combustible
materials shall be removed from the area immediately following the unpacking. Such transient
combustible material, unless stored in approved containers, shall not be left unattended
during lunch breaks, shift changes, or other similar periods. Loose combustible
packing material such as wood or paper excelsior, or polyethylene sheeting shall be
placed in metal containers with tight-fitting self-closing metal covers.
- 9. Control actions to be taken by an individual discovering a
fire. For example, notification of control room, attempt to extinguish fire, and actuation
of local fire suppression systems.
- 10. Control actions to be taken by the control room oper-ator to determine the need for
brigade assistance upon report of a fire or receipt of alarm on control room annunciator panel, for example, announcing location of fire over PA system, sounding fire alarms, and notifying the
shift supervisor and the fire brigade leader of the type, size, and location of the fire. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 23) 10CFR50 Appendix R WCGS
- 11. Control actions to be taken by the fire brigade after
notification by the control room operator of a fire, for example, assembling in a desig-
nated location, receiving di-rections from the fire brigade leader, and discharging specific
fire fighting responsibilities including selection and trans-poration of fire fighting
equipment to fire location, selection of protective equip-ment, operating instructions
for use of fire suppression systems, and use of preplanned strategies for fighting fires
in specific areas.
- 12. Define the strategies for fighting fires in all safety-related areas and areas present-ing a hazard to safety-related
equipment. These strategies shall designate:
- a. Fire hazards in each area covered by the specific prefire plans.
- b. Fire extinguishants best suited for controlling the
fires associated with the fire hazards in that area and the nearest location of
these extinguishants.
- c. Most favorable direction from which to attack a fire in each area in view of the ven- Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 24) 10CFR50 Appendix R WCGS tilation direction, access hallways, stairs, and doors
that are most likely to be free of fire, and the best station or elevation for fight-
ing the fire. All access and egress routes that involve locked doors should be speci-
fically identified in the pro-cedure with the appropriate precautions and methods for
access specified.
- d. Plant systems that should be managed to reduce the damage potential during a local fire and the location of local and
remote controls for such management (e.g., any hydrau-lic or electrical systems in
the zone covered by the specific fire fighting proce-dure that could increase the
hazards in the area because of overpressurization or electrical hazards).
- e. Vital heat-sensitive system components that need to be kept
cool while fighting a local fire. Particularly hazardous combustibles that need
cooling should be designated.
- f. Organization of fire fighting brigades and the assignment of special duties according to job title so that
all fire fighting functions are Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 25) 10CFR50 Appendix R WCGS covered by any complete shift personnel complement. These duties include command control of the brigade, transporting fire suppression and support equipment to the fire scenes, applying the extinguishant to the fire, communication with the control room, and coordin- ation with outside fire departments. g. Potential radiological and toxic hazards in fire zones. h. Ventilation system operation that ensures desired plant air distribution when the ventilation flow is modified for fire con-tainment or smoke clearing operations. i. Operations requiring con- trol room and shift engineer coordination or authorization. j. Instructions for plant operators and general plant personnel during fire. III. L. Alternative and Dedicated Shutdown Capability 1. Alternative or dedicated An auxiliary shutdown shutdown capability provided panel, described in for a specific fire area shall Section 7.4, in conjunction be able to (a) achieve and with certain local maintain subcritical reac- controls, provides a means tivity conditions in the of achieving and maintaining reactor, (b) maintain hot standby in the event reactor coolant inventory that the main control room (c) achieve and maintain is uninhabitable. hot standby (7) conditions Rev. 23 WOLF CREEK TABLE 9.5E-1 (Sheet 26) 10CFR50 Appendix R WCGS for PWR (hot shutdown (7) for The auxiliary shutdown a BWR); (d) achieve cold panel contains the con-shutdown conditions within trols and indication 72 hours; and (e) maintain necessary to maintain cold shutdown conditions reactor coolant system thereafter. During the inventory, remove decay postfire shutdown, the heat, and provide the reactor coolant system required boration for process variables shall hot standby. Adequate be maintained within operations shift staffing those predicted for loss is provided to achieve and of normal ac power and maintain post-fire safe shut-the fission product down "Hot Standby Conditions" boundary integrity shall in the event of a fire. not be affected i.e., Cold shutdown can be there shall be no fuel achieved and maintained clad damage, rupture of from outside the control any primary coolant room by additional manual boundary, or rupture operator action at local of the containment control sites. boundary. The auxiliary shutdown
- 2. The performance goals for panel is included in the shutdown functions shall the fire hazards anal-be: ysis, Appendix 9.5B.
- a. The reactivity control The performance criteria of function shall be capable of III.L.1 are satisfied, with the achieving and maintaining exception of maintaining reactor cold shutdown reactivity process variables within those conditions.
predicted for a loss of normal ac power. This is acceptable, as long as a control room fire will not result in the plant reaching an unrecoverable condition, which could lead to core damage. The criteria for "not reaching an unrecoverable condition" are that 1) natural circulation is maintained, and
- 2) adequate core cooling is maintained.
_____________________________________________________ 7 - As defined in the Standard Technical Specifications.
Rev. 29 WOLF CREEK TABLE 9.5E-1 (Sheet 27) 10CFR50 Appendix R WCGS
- b. The reactor coolant In general, the performance goals makeup function shall be of III.L.2 are satisfied except that capable of maintaining the in some cases pressurizer water reactor coolant level above level is not maintained within level the top of the core for BWRs indication. This is acceptable as and be within the level long as an evaluation demonstrates indication in the pressur-that unrecoverable conditions are izers for PWRs.
not reached.
- c. The reactor heat removal function shall be capable of achieving and maintaining decay heat removal.
- d. The process monitoring function shall be capable of providing direct readings of the process variables necessary to perform and control the above functions.
- e. The supporting functions shall be capable of providing the process cooling, lubri-cation, etc., necessary to permit the operation of the equipment used for safe shutdown functions.
- 3. The shutdown capability for specific fire areas may be
unique for each such area or it may be one unique combination
Rev. 29 WOLF CREEK TABLE 9.5E-1 (Sheet 28) 10CFR50 Appendix R WCGS of systems for all such areas. In either case the alterna-
tive shutdown capability shall be independent of the specific fire area(s) and
shall accommodate postfire conditions where offsite power is available and where
offsite power is not available for 72 hours. Procedures shall be in effect
to implement this capability.
- 4. If the capability to achieve and maintain cold shutdown will not be available because of fire
damage, the equipment and systems comprising the means to achieve and maintain the
hot standby or hot shutdown condition shall be capable of maintaining such conditions
until cold shutdown can be achieved. If such equipment and systems will not be
capable of being powered by both onsite and offsite electric power systems
because of fire damage an independent onsite power system shall be provided.
The number of operating shift personnel, exclusive of fire brigade members, required to
operate such equipment and systems shall be on site at all times. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 29) 10CFR50 Appendix R WCGS
- 5. Equipment and systems comprising the means to
achieve and maintain cold shutdown conditions shall not be damaged by fire or the
fire damage to such equipment and systems shall be limited so that the systems can be
made operable and cold shutdown can be achieved within 72 hours. Materials
for such repairs shall be readily available on site and procedures shall be in effect
to implement such repairs. If such equipment and systems used prior to 72 hours after
the fire will not be capable of being powered by both onsite and offsite electric
power systems because of fire damage an independent onsite power system shall be
provided. Equipment and systems used after 72 hours may be powered by offsite
power only.
- 6. Shutdown systems installed to ensure postfire shutdown capabil-ity need not be designed to meet seismic Category I criteria, single failure criteria, or other design basis accident criteria, except where required for other
reasons, e.g., because of interface with or impact on existing safety systems, or
because of adverse valve actions due to fire damage. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 30) 10CFR50 Appendix R WCGS
- 7. The safe shutdown equipment and systems for each fire area
shall be known to be isolated from associated non-safety circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe
shutdown equipment. The separator and barriers between trays and conduits containing associated
circuits of one safe shutdown division and trays and conduits containing associated circuits or
safe shutdown cables from the redundant division, or the isolation of these associated
circuits from the safe shutdown equipment, shall be such that a postulated fire involving
associated circuits will not prevent safe shutdown8. __________________________ 8An acceptable method of complying with this alternative would be to meet Regulatory
Guide 1.75 position 4 related to associated circuits and IEEE Std 384-1974 (Section 4.5) where trays from redundant safety div-
isions are so protected that postulated fires affect trays from only one safety division. Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 31) 10CFR50 Appendix R WCGS III. M. Fire Barrier Cable Penetration Seal Qualification Penetration seal designs shall utilize only Complies. As stated in noncombustible materials and shall be Section 9.5.1.2.2.3, qualified by tests that are comparable to the penetration seal tests used to rate fire barriers. The designs were tested acceptance criteria for the test shall utilizing the following
include: for test guidance:
- 1. The cable fire barrier pen- o ASTM E 119 etration seal has withstood o IEEE 634 the fire endurance test with- o ANI/MAERP stan-out passage of flame or dard test method
ignition of cables on the unexposed side for a period The test included a of time equivalent to the standard hose stream
fire resistance rating re- test. quired of the barrier.
- 2. The temperature levels re-corded for the unexposed side are analyzed and demonstrate
that the maximum temperature is sufficiently below the cable in-sulation ignition temperature;
and
- 3. The fire barrier penetration seal remains intact and does not allow projection of water beyond the unexposed surface during the
hose stream test. III. N. Fire Doors Fire doors shall be self- Complies. closing or provided with Rev. 0 WOLF CREEK TABLE 9.5E-1 (Sheet 32) 10CFR50 Appendix R WCGS closing mechanisms and shall Standard fire doors are be inspected semiannually to provided with self-
verify that automatic hold- closing devices. Doors open, release, and closing are normally closed and mechanisms and latches are locked unless the door
operable. is a means of egress, in which case they are One of the following measures closed and latched.
shall be provided to ensure they will protect the opening The door between the as required in case of fire: Control Room and Pantry
is normally open. This
- 1. Fire doors shall be kept door is provided with an closed and electrically electromagnetic closer, supervised at a continuously which closes the door upon manned location; detection of fire.
- 2. Fire doors shall be locked Special doors such and inspected weekly to verify as pressure, water-that the doors are in the tight, and missile-
closed position; resistant that are also fire doors are normally
- 3. Fire doors shall be pro- closed and locked.
vided with automatic hold-open and release mechanisms and Doors for areas pro-inspected daily to verify that tected by halon systems
doorways are free of obstruc- have self-closing tions; or mechanisms or are electrically supervised to ensure they are maintained closed.
- 4. Fire doors shall be kept closed and inspected daily to verify that they are in the closed position.
The fire brigade leader shall have ready access to keys for any locked fire doors. Rev. 6 WOLF CREEK TABLE 9.5E-1 (Sheet 33) 10CFR50 Appendix R WCGS Areas protected by automatic total flooding gas suppression systems shall have electrically supervised self-closing fire doors or shall satisfy option 1 above. III. 0. Oil Collection System for Reactor Coolant Pump The reactor coolant pump shall be The reactor coolant pumps equipped with an oil collection (RCP) are provided with an system if the containment is not oil spillage protection and inerted during normal operation. control system that consists The oil collection system shall be of a package of splash so designed, engineered, and in- guards, catch basins, and stalled that failure will not lead enclosure assembled as to fire during normal or design attachments to the RCP basis accident conditions and that motors at strategic there is reasonable assurance locations to preclude the that the system will withtand possibility of oil making the Safe Shutdown Earthquake. 9 contact with hot components and piping. Such collection systems shall be capable of collecting lube oil High pressure portions of from all potential pressurized the lube oil system are and unpressurized leakage sites totally enclosed with low in the reactor coolant pump lube point drain connections. oil systems. Leakage shall be Low pressure portions of collected and drained to a the system are provided vented closed container that with drip pans with low can hold the entire lube oil point connections. Remote system inventory. A flame lube oil fill lines for arrester is required in the the upper and lower bearing vent if the flash point charac reservoirs on each reactor teristics of the oil present coolant pump motor are not the hazard of fire flashback. protected by drip pans. Leakage points to be protected Due to the design of the shall include lift pump and fill lines, no lube oil piping overflow lines, lube leakage is postulated. The oil cooler, oil fill and drain RTD Conduit Boxes (3 per lines and plugs, flanged con- motor) are not provided with nections on oil lines, and drip pans, however, conduit lube oil reservoirs where seals and leak tight such features exist on the fittings are used to reactor coolant pumps. The minimize lube oil leakage. drain line shall be large Oil leakage at the RDT enough to accommodate the Conduit Box does not largest potential oil leak. represent a fire hazard. Rev. 11 WOLF CREEK TABLE 9.5E-1 (Sheet 34) 10CFR50 Appendix R WCGS The low points of the col- lection systems are piped to two collection tanks (each tank serves two RCPs) located in the reactor building as shown on Figure 9.5.1-3. Each collection tank has a capacity of greater than 300 gallons. Each RCP motor contains approximately 265 gallons of oil; however, it is un- likely that common failure would occur that would cause the entire inventory of oil in two RCP motors to leak out. The col-lection tanks are provided with level indication and high level alarm in the control room. Therefore, the plant operators would have an early indication of a significant oil leak and could initiate cor- rective action. Should leakage exceed the collection tank capacity before corrective actions are completed, the tank would overflow onto the containment floor. Any such leakage would flow into the drainage trenches located adjacent to the tanks (see Figure 1.2.11) and be collected in the containment normal sumps. This oil would not come into contact with hot surfaces and would not pose a significant fire hazard. The tanks are constructed to the requirements of ASME Section VIII and have flame arrestors on the vents. The drain piping is ANSI B31.1. The tanks and piping are seismically Rev. 12 WOLF CREEK TABLE 9.5E-1 (Sheet 35) 10CFR50 Appendix R WCGS supported in accordance with the requirements of
Paragraph C.2 of Regula-tory Guide 1.29. The oil collection devices mounted on the RCPs have been seismically analyzed
and qualified in accord-ance with the requirements of Paragraph C.2 of
Regulatory Guide 1.29. _____________________ 9 See Regulatory Guide 1.29 - "Seismic Design Classification", Paragraph C.2. Rev. 12}}