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| number = ML18093A895
| number = ML18093A895
| issue date = 03/08/2018
| issue date = 03/08/2018
| title = Revision 31 to Updated Safety Analysis Report, Chapter 7, Instrumentation and Controls
| title = 1 to Updated Safety Analysis Report, Chapter 7, Instrumentation and Controls
| author name =  
| author name =  
| author affiliation = Wolf Creek Nuclear Operating Corp
| author affiliation = Wolf Creek Nuclear Operating Corp
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:WOLF CREEK TABLE OF CONTENTS
{{#Wiki_filter:}}
 
CHAPTER 7.0
 
INSTRUMENTATION AND CONTROLS
 
Section                                                  Page
 
==7.1 INTRODUCTION==
7.1-1
 
7.1.1 IDENTIFICATION OF SAFETY-RELATED SYSTEMS 7.1-1
 
7.1.1.1 Reactor Trip System 7.1-2 7.1.1.2 Engineered Safety Feature Actuation Systems 7.1-2 7.1.1.3 Systems Required for Safe Shutdown 7.1-3 7.1.1.4 Safety-Related Display Instrumentation 7.1-3 7.1.1.5 All Other Instrumentation Systems Required for Safety 7.1-3 7.1.1.6 Control Systems Not Required for Safety 7.1-4
 
7.1.2 IDENTIFICATION OF SAFETY CRITERIA 7.1-4
 
7.1.2.1 Design Bases 7.1-5 7.1.2.2 Independence of Redundant Safety-Related Systems 7.1-5 7.1.2.3 Physical Identification of Safety-Related Equipment 7.1-9 7.1.2.4 Conformance to Criteria 7.1-11 7.1.2.5 Conformance to NRC Regulatory Guides 7.1-11 7.1.2.6 Conformance to IEEE Standards 7.1-15
 
7.
 
==1.3 REFERENCES==
7.1-16
 
7.2 REACTOR TRIP SYSTEM 7.2-1
 
7.
 
==2.1 DESCRIPTION==
7.2-1
 
7.2.1.1 System Description 7.2-1 7.2.1.2 Design Bases Information 7.2-17 7.2.1.3 Final Systems Drawings 7.2-19
 
7.2.2 ANALYSES 7.2-20
 
7.2.2.1 Failure Mode and Effects Analyses 7.2-20 7.2.2.2 Evaluation of Design Limits 7.2-20 7.2.2.3 Specific Control and Protection Interactions 7.2-35 7.2.2.4 Additional Postulated Accidents 7.2-39
 
7.0-i                        Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
Section                                                  Page
 
7.2.3 TESTS AND INSPECTIONS 7.2-39 7.
 
==2.4 REFERENCES==
7.2-40
 
7.3 ENGINEERED SAFETY FEATURE SYSTEMS 7.3-1
 
7.3.1 CONTAINMENT COMBUSTIBLE GAS CONTROL SYSTEM 7.3-2
 
7.3.1.1 Description 7.3-2 7.3.1.2 Analysis 7.3-5
 
7.3.2 CONTAINMENT PURGE ISOLATION SYSTEM 7.3-9
 
7.3.2.1 Description 7.3-9 7.3.2.2 Analysis 7.3-12
 
7.3.3 FUEL BUILDING VENTILATION ISOLATION 7.3-13
 
7.3.3.1 Description 7.3-13 7.3.3.2 Analysis 7.3-15
 
7.3.4 CONTROL ROOM VENTILATION ISOLATION 7.3-16
 
7.3.4.1 Description 7.3-16 7.3.4.2 Analysis 7.3-19
 
7.3.5 DEVICE LEVEL MANUAL OVERRIDE 7.3-19
 
7.3.5.1 Description 7.3-19 7.3.5.2 Analysis 7.3-20
 
7.3.6 AUXILIARY FEEDWATER SUPPLY 7.3-20
 
7.3.6.1 Description 7.3-20 7.3.6.2 Analysis 7.3-24
 
7.3.7 MAIN STEAM AND FEEDWATER ISOLATION 7.3-25
 
7.3.7.1 Description 7.3-25 7.3.7.2 Analysis 7.3-27 
 
7.3.8 NSSS ENGINEERED SAFETY FEATURE ACTUATION SYSTEM 7.3-29
 
7.3.8.1 Description 7.3-28 7.3.8.2 Analysis 7.3-41 7.3.8.3 Summary 7.3-56 
 
7.0-ii                      Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
Section Page
 
7.
 
==3.9 REFERENCES==
7.3-58
 
7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.4-1 
 
7.
 
==4.1 INTRODUCTION==
7.4-1
 
7.4.2 SAFE SHUTDOWN OVERVIEW 7.4-1
 
7.4.3 SAFE SHUTDOWN SCENARIO 7.4-2
 
7.4.3.1 Hot Standby Systems 7.4-3
 
7.4.3.1.1 Reactor Coolant System 7.4-3
 
7.4.3.1.1.1. Pressurizer 7.4-4
 
7.4.3.1.2 Main Steam (Steam Generators) 7.4-4
 
7.4.3.1.2.1 Water level for each steam generator 7.4-4
 
7.4.3.1.2.2 Pressure for each steam generator 7.4-4
 
7.4.3.1.3 Auxiliary Feedwater 7.4-4
 
7.4.3.1.4 Chemical and Volume Control (CVCS) 7.4-5
 
7.4.3.1.5 Essential Service Water (ESWS) 7.4-5
 
7.4.3.1.6 Component Cooling Water (CCW) 7.4-5
 
7.4.3.2 Hot Standby Discussion 7.4-5
 
7.4.3.3 Cold Shutdown Discussion 7.4-7
 
7.4.4 PLANT SAFE SHUTDOWN (PSSD) 7.4-9
 
7.4.5 POST-ACCIDENT SAFE SHUTDOWN 7.4-9
 
7.4.6 SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM 7.4-10
 
7.4.6.1 Description 7.4-10
 
7.4.6.1.1. Auxiliary Shutdown Panel 7.4-11
 
7.4.6.1.2 Controls at Switchgear Motor Control Centers, and Other Locations 7.4-11
 
7.4.7 CONTROLS FOR EXTENDED HOT STANDBY 7.4-11
 
7.0-iii Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
7.4.8 DESIGN BASIS 7.4-12
 
7.4.8.1 Initiating circuits 7.4-12
 
7.4.8.2 Logics 7.4-12
 
7.4.8.3 Bypass 7.4-13
 
7.4.8.4 Interlock 7.4-13
 
7.4.8.5 Redundancy 7.4-13
 
7.4.8.6 Diversity 7.4-13
 
7.4.8.7 Actuated devices 7.4-13
 
7.4.8.8 Supporting systems 7.4-13
 
7.4.8.9 Consequences Analysis 7.4-13
 
7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 7.5-1
 
7.5.1 REACTOR TRIP SYSTEM 7.5-2 7.5.2 ENGINEERED SAFETY FEATURE SYSTEM 7.5-2
 
7.5.2.1 System Actuation Parameters 7.5-2 7.5.2.2 System Bypasses 7.5-4 7.5.2.3 System Status 7.5-7 7.5.2.4 System Performance 7.5-9
 
7.5.3 SAFE SHUTDOWN 7.5-10
 
7.5.3.1 Hot Standby Control 7.5-10 7.5.3.2 Cold Shutdown Control 7.5-11 7.5.3.3 System Bypasses 7.5-11 7.5.3.4 System Status 7.5-12 7.5.3.5 System Performance 7.5-12
 
7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 7.6-1 7.6.1 INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM 7.6-1 7.6.2 RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES 7.6-1
 
7.0-iv                      Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
Section Page
 
7.6.2.1 Description 7.6-1 7.6.2.2 Analysis 7.6-2
 
7.6.3 REFUELING INTERLOCKS 7.6-3 7.6.4 ACCUMULATOR MOTOR-OPERATED VALVES 7.6-3 7.6.5 SWITCHOVER FROM INJECTION TO RECIRCULATION 7.6-5 7.6.6 INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION 7.6-5 7.6.6.1 Analysis of Interlocks 7.6-7 7.6.7 ISOLATION OF ESSENTIAL SERVICE WATER (ESW)
 
TO THE AIR COMPRESSORS 7.6-8 7.6.7.1 Description 7.6-8 7.6.7.2 Analysis 7.6-9 7.6.8 ISOLATION OF THE NONSAFETY-RELATED PORTION OF THE COMPONENT COOLING WATER (CCW) SYSTEM 7.6-10
 
7.6.8.1 Description 7.6-10 7.6.8.2 Analysis 7.6-12 7.6.9 FIRE PROTECTION AND DETECTION 7.6-13 7.6.10 INTERLOCKS FOR PRESSURIZER PRESSURE RELIEF SYSTEM 7.6-13 7.6.10.1 Description of Pressurizer Pressure Relief System 7.6-13 7.6.10.2 Description of Pressurizer Pressure Relief System Interlocks 7.6-13 7.6.11 SWITCHOVER OF CHARGING PUMP SUCTION TO RWST ON LOW-LOW VCT LEVEL 7.6-14 7.6.11.1 Description 7.6-14 7.6.11.2 Evaluation of Switchover of Charging Pump Suction 7.6-15 7.6.12 INSTRUMENTATION FOR MITIGATING CON- SEQUENCES OF INADVERTENT BORON DILUTION 7.6-15
 
7.6.12.1 Description 7.6-15 7.6.12.2 Analysis 7.6-15 7.6.12.3 Qualification 7.6-15 7.6.13 MONITORING OF RCS LEVEL DURING REDUCED INVENTORY (MID-LOOP) OPERATIONS 7.6-16 7.6.14 INCORE THERMOCOUPLES 7.6-16 
 
7.0-v                      Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
Section Page
 
7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 7.7-1
 
7.
 
==7.1 DESCRIPTION==
7.7-1
 
7.7.1.1 Reactor Control System 7.7-3 7.7.1.2 Rod Control System 7.7-4 7.7.1.3 Plant Control Signals for Monitoring and Indicating 7.7-10 7.7.1.4 Plant Control System Interlocks 7.7-16 7.7.1.5 Pressurizer Pressure Control 7.7-17 7.7.1.6 Pressurizer Water Level Control 7.7-18 7.7.1.7 Steam Generator Water Level Control 7.7-19 7.7.1.8 Steam Dump Control 7.7-19 7.7.1.9 Neutron Flux Detectors 7.7-21 7.7.1.10 Boron Concentration Monitoring System 7.7-23 7.7.1.11 ATWS Mitigation System Actuation Circuitry 7.7-25 
 
7.7.2 ANALYSIS 7.7-27
 
7.7.2.1 Separation of Protection and Control System 7.7-33 7.7.2.2 Response Considerations of Reactivity 7.7-33 7.7.2.3 Step Load Changes Without Steam Dump 7.7-36 7.7.2.4 Loading and Unloading 7.7-36 7.7.2.5 Load Rejection Furnished By Steam Dump System 7.7-37 7.7.2.6 Turbine-Generator Trip With Reactor Trip 7.7-38
 
7.
 
==7.3 REFERENCES==
7.7-39
 
App. 7A COMPARISON TO REGULATORY GUIDE 1.97 7A-1 
 
7A.1 INTRODUCTION 7A-1
 
7A.2 ORGANIZATION 7A-1
 
7A.3 WCGS DESIGN BASIS COMPARISON TO REGULATORY GUIDE 1.97 7A-2 7A.3.1 TYPE A VARIABLES 7A-2
 
7A.3.2 REDUNDANCY AND DIVERSITY FOR CATEGORY 1 VARIABLES 7A-3 
 
7A.3.3 RECORDERS 7A-4
 
7A.3.4 INSTRUMENT RANGES 7A-4
 
7.0-vi                        Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
 
Section                                                  Page
 
7A.3.5  UNNECESSARY VARIABLES                          7A-4
 
7A.3.6  QUALIFICATION FOR CATEGORY 1 PARAMETERS        7A-4
 
7A.3.7  QUALIFICATION FOR CATEGORY 2 PARAMETERS        7A-5
 
7A.3.8  QUALIFICATION FOR CATEGORY 3 ITEMS              7A-5
 
7.0-vii                      Rev. 14 WOLF CREEK TABLE OF CONTENTS (Continued)
 
LIST OF TABLES
 
Number                          Title
 
7.1-1                Instrumentation Systems Identification 
 
7.1-2                Identification of Safety Criteria
 
7.1-3                Conformance to Regulatory Guide 1.22 
 
7.1-4                Conformance to Regulatory Guide 1.53
 
7.1-5                Conformance to Regulatory Guide 1.62
 
7.1-6                Conformance to Regulatory Guide 1.105 
 
7.1-7                Conformance to Regulatory Guide 1.118, Rev. 2
 
7.2-1                List of Reactor Trips
 
7.2-2                Protection System Interlocks
 
7.2-3                Reactor Trip System Instrumentation 
 
7.2-4                Reactor Trip Correlation 
 
7.3-1                Containment Combustible Gas Control
 
System Actuated Equipment List
 
7.3-2                Containment Combustible Gas Control
 
System Failure Modes and Effects
 
Analysis
 
7.3-3                Containment Purge Isolation Actuation
 
System Actuated Equipment List
 
7.3-4                Containment Purge Isolation Actuation
 
System Failure Modes and Effects Analysis
 
7.3-5                Fuel Building Ventilation Isolation
 
Actuation System Actuated Equipment List
 
7.3-6                Fuel Building Ventilation Isolation 
 
Actuation System Failure Modes and 
 
Effects Analysis 
 
7.3-7                Deleted
 
7.0-viii                      Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
LIST OF TABLES Number Title
 
7.3-8 Control Room Ventilation Isolation Control System Actuated Equipment List
 
7.3-9 Control Room Ventilation Isolation Actuation System Failure Modes and Effects Analysis
 
7.3-10 Device Level Manual Override Failure Modes and Effects Analysis
 
7.3-11 Auxiliary Feedwater Actuation System Failure Modes and Effects Analysis
 
7.3-12 Auxiliary Supporting Engineered Safety Feature Systems
 
7.3-13 NSSS Instrumentation Operating Condition for Engineered Safety Features
 
7.3-14 NSSS Instrument Operating Conditions for Isolation Functions
 
7.3-15 NSSS Interlocks for Engineered Safety Feature Actuation System
 
7.4-1 AUXILIARY SHUTDOWN PANEL EQUIPMENT LIST 
 
7.4-1.1 Auxiliary Shutdown Panel Controls and Monitoring Indicators
 
7.4-1.2 Controls at Switchgear Motor Control Centers, and Other Locations
 
7.4-2 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.139 REV. 1, DRAFT 2 DATED FEBRUARY 25, 1980 TITLED 
 
"GUIDANCE FOR RESIDUAL HEAT REMOVAL TO ACHIEVE AND 
 
MAINTAIN COLD SHUTDOWN" 
 
7.4-3 DESIGN COMPARISON OF TABLE 1 OF BTP RSB 5-1 FOR POSSIBLE SOLUTIONS FOR FULL COMPLIANCE
 
7.4-4 RESIDUAL HEAT REMOVAL - SAFETY RELATED COLD SHUTDOWN OPERATIONS - FAILURE MODES AND EFFECTS ANALYSIS (FMEA) 
 
7.4-5 Systems Required to Achieve and Maintain Post-Accident Safe Shutdown
 
7.4-6 Post-Accident Safe Shutdown components 
 
7.5-1 Engineered Safety Features - Displays
 
7.5-2 Post Accident Safe Shutdown Display Information
 
7.5-3 WCGS Plant Design Comparison with Regulatory Guide 1.47 Dated May 1973, Titled "Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems"
 
7.0-ix Rev. 29 WOLF CREEK TABLE OF CONTENTS (Continued)
LIST OF TABLES Number Title
 
7.5-4 Safety-Related Display Instrumentation Located on the Control Board - (NSSS Scope of Supply)
 
7.5-5 Safety-Related Display Instrumentation Located on the Control Board - (BOP Scope of Supply)
 
7.7-1 Plant Control System Interlocks
 
7.7-2 Boron Concentration Measurement System
 
Specifications
 
7.7-3 Loss of Any Single Instrument
 
7.7-4 Loss of Power to a Protection Separation
 
Group
 
7.7-5 Loss of Power to a Control Separation
 
Group
 
7.7-6 Break of Common Instrument Lines
 
7A-1 Regulatory Guide 1.97 Variable List 
 
7A-2 Summary Comparison to Regulatory Guide 1.97 
 
7A-3 Data Sheets 
 
7.0-x                      Rev. 29
 
WOLF CREEK CHAPTER 7 - 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 #*
7.1-1 0 Protection System Block Diagram 7.2-1 1Functional Diagrams (Index and Symbols) M-744-00018 7.2-1 2Functional Diagrams (Reactor Trip Signals) M-744-00019 7.2-1 3Functional Diagrams (Nuclear Instrumentation and Manual Trip Signals) M-744-00020 7.2-1 4Functional Diagrams (Nuclear Instrumentation Permissives and Blocks) M-744-00021 7.2-1 5Functional Diagrams (Primary Coolant System Trip SignalsM-744-00022 7.2-1 6Functional Diagrams (Pressurizer Trip Signals) M-744-00023 7.2-1 7Functional Diagrams (Steam Generator Trip Signals) M-744-00024 7.2-1 8Functional Diagrams (Safeguards Actuation Signals) M-744-00025 7.2-1 9Functional Diagrams (Rod Controls and Rod Blocks) M-744-00026 7.2-1 10Functional Diagrams (Steam Dump Control) M-744-00027 7.2-1 11Functional Diagrams (Pressurizer Pressure and Level Control) M-744-00028 7.2-1 12Functional Diagrams (Pressurizer Heater Control) M-744-00029 7.2-1 13Functional Diagrams (Feedwater Control and Isolation)M-744-00030 7.2-1 14Functional Diagrams (Feedwater Control and Isolation)M-744-00031 7.2-1 15Functional Diagrams (Auxiliary Feedwater Pumps Start-up)M-744-00032 7.2-1 16Functional Diagrams (Turbine Trips, Runbacks and Other Signals) M-744-00033 7.2-1 17Functional Diagram (Pressurizer Pressure Relief System Train A) M-744-00039 7.2-1 18Functional Diagram (Pressurizer Pressure Relief System Train B) M-744-00040 7.2-2 0Setpoint Reduction Function for Overpower and Over-temperature  T Trips 7.2-3 0 Reactor Trip/Engineered Safety Features Actuation Mechanical Linkage 7.3-1 1 Engineered Safety Features Actuation System (BOP)7.3-1 2Logic Diagram Engineered Safety Features Actuation System (BOP) J-104-00390 7.3-1 3Logic Diagram Engineered Safety Features Actuation System (BOP) 7.3-2 0 Typical Engineered Safety Feature Test Circuits 7.3-3 0Engineered Safeguards Test Cabinet (Index, Notes, and Legend)      7.0-xi    Rev. 17 WOLF CREEK CHAPTER 7 - 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 #*
7.6-1 1Logic Diagram for Outer RHRS Isolation Valve 7.6-1 2Logic Diagram for Inner RHRS Isolation Valve 7.6-2 0Function Block Diagram of Accumulator Isolation Valve 7.6-3 1Safety Injection System Recirculation Sump and RHR Suction Isolation Valves 1, 2, & 3) 7.6-3 2Safety Injection System Recirculation Sump and RHR Suction Isolation Valves 1, 2, & 3) 7.6-3 3Safety Injection System Recirculation Sump and RHR Suction Isolation Valves 1, 2, & 3) 7.6-4 1Train B Functional Diagram Showing Logic Requirements for Pressurizer Pressure Relief SystemM-744-00039 7.6-4 2Train A Functional Diagram Showing Logic Requirements for Pressurizer Pressure Relief SystemM-744-00040 7.6-4 3Functional Diagram of Logic Requirements for Pressurizer Pressure Relief System 7.6-5 1Logic Diagram for VCT Outlet Isolation Valve Interlocks on Switchover to RWST 7.6-5 2Logic Diagram for RWST Valves Interlocks on Switchover to RWST 7.6-6 0 (deleted)7.7-1 0Simplified Block Diagram of Reactor Control System 7.7-2 0Control Bank Rod Insertion Monitor 7.7-3 0Rod Deviation Comparator 7.7-4 0Block Diagram of Pressurizer Pressure Control System 7.7-5 0Block Diagram of Pressurizer Level Control System 7.7-6 0Block Diagram of Steam Generator Water Level Control System 7.7-7 0Block Diagram of Main Feedwater Pump Speed Control System 7.7-8 0Block Diagram of Steam Dump Control System 7.7-9 0 Basic Flux-Mapping System 7.7-10 0Sampler Assembly 7.7-11 0Sampler Subassembly 7.7-12 0 Process Assembly Block Diagram 7.7-13 0 Boron Concentration Monitoring System Linearity Curve Over Normal Plant Operating Range of Boron Concentrations 7.7-14 0Simplified Block Diagram of Rod Control System 7.7-15 0 Control Bank D Partial Simplified Schematic Diagram of Power Cabinets 1BD and 2BD      7.0-xii    Rev. 17 WOLF CREEK CHAPTER 7.0 INSTRUMENTATION AND CONTROLS
 
==7.1  INTRODUCTION==
 
This section describes the various plant instrumentation and control systems
 
and the functional performance requirements, design bases, system descriptions, design evaluations, and tests and inspections for each. The information
 
provided in this chapter emphasizes those instruments and associated equipment
 
which constitute the protection system, as defined in IEEE Standard 279-1971, "IEEE Standard:  Criteria for Protection Systems for Nuclear Power Generating
 
Stations."
 
The instrumentation and control systems provide automatic protection and exercise proper control against unsafe and improper reactor operation during steady state and transient power operations (Conditions I, II and III) and to
 
provide initiating signals to mitigate the consequences of emergency and
 
faulted conditions (Condition III and IV). ANS conditions are discussed in
 
Chapter 15.0.
 
Applicable criteria and codes are listed in Table 7.1-2.
 
7.1.1  IDENTIFICATION OF SAFETY-RELATED SYSTEMS
 
Safety-related instrumentation and control systems and their supporting systems
 
are those systems required to ensure:
: a. The integrity of the reactor coolant pressure boundary.
: b. The capability to shut down the reactor and maintain it  in a safe shutdown condition following any design basis accident.
: c. The capability to prevent or mitigate the consequences of
 
accidents which could result in potential offsite
 
exposures comparable to the guideline exposures of 10 CFR
 
100.
 
The definitions provided below are used to classify the instrumentation systems
 
into the categories defined for Chapter 7.0 by Regulatory Guide 1.70.
 
A listing of these systems, by categories, that are comparable to those of
 
nuclear power plants of similar design is given in Table 7.1-1. Table 7.1-1
 
also identifies the systems that are different with references to discussions
 
of those differences.
 
7.1-1                          Rev. 19 WOLF CREEK The plant's control and instrumentation systems are grouped into the following
 
categories:
: a. Reactor trip system (RTS)
: b. Engineered safety feature actuation systems (ESFAS)
: c. Systems required for safe shutdown
: d. Safety-related display instrumentation
: e. Other instrumentation systems required for safety
: f. Systems not required for safety.
Descriptions of the above are given in Sections 7.1.1.1 through 7.1.1.6. Table
 
7.1-2 identifies which instrumentation systems are safety-related.
 
7.1.1.1  Reactor Trip System The RTS is described in Section 7.2  Figure 7.1-1 is a single line diagram of
 
this system.
 
7.1.1.2  Engineered Safety Feature Actuation System The ESFAS are those instrumentation systems which are needed to actuate the
 
equipment and systems required to mitigate the consequences of postulated
 
design basis accidents. The engineered safety features requiring actuation are:
: a. Main steam line and feedwater isolation (Sections 6.2.4
 
and 7.3.7)
: b. Containment combustible gas control (Sections 6.2.5 and
 
7.3.1)
: c. Containment purge isolation (Sections 6.2.4 and 7.3.2)
: d. Fuel building exhaust isolation (Sections 9.4.2 and 7.3.3)
: e. Control room ventilation isolation (Sections 9.4.1 and
 
7.3.4).
: f. Auxiliary feedwater supply (Sections 10.4.9 and 7.3.6)
: g. NSSS ESFAS (Section 7.3.8)
 
7.1-2                          Rev. 1 WOLF CREEK
: h. Control room ventilation isolation (Section 9.4.1)
: i. Auxiliary feedwater supply (Section 10.4.9)
 
The equipment which provides the engineered safety feature actuation functions
 
for the systems listed above is identified and discussed in Section 7.3. 
 
Design bases for these engineered safety feature actuation systems are also
 
given in Section 7.3. For auxiliary supporting systems, see Section 7.3.8.1.1i
 
and Table 7.3-12.
 
7.1.1.3  Systems Required for Post Accident Safe Shutdown Systems required for post accident safe shutdown are defined as those essential for pressure and reactivity control, coolant inventory makeup, and removal of
 
residual heat once the reactor has been brought to a subcritical condition. 
 
These functions are categorized according to the shutdown modes defined in the
 
Technical Specifications.
 
Identification of the equipment and systems required for post-accident safe
 
shutdown is provided in Section 7.4. Additional information regarding hot
 
standby provisions for shutdown from outside the control room is also provided
 
in Section 7.4.
 
7.1.1.4  Safety-Related Display Instrumentation Safety-related display instrumentation is instrumentation which provides
 
information for the operator to manually perform reactor trip, engineered
 
safety feature actuation, post-accident monitoring or post-accident safe shutdown functions.
 
Identification of the equipment and systems for safety-related display
 
instrumentation is provided in Section 7.5. Description of other indicating
 
systems which provide information for monitoring equipment and processes is
 
also provided in Section 7.5 and Table 7.5-1.
 
7.1.1.5  All Other Instrumentation Systems Required for Safety The other instrumentation systems required for safety - other than the RTS, the
 
ESFAS, safety-related display and the post accident safe shutdown systems - are discussed in Section 7.6. They are those systems and components which have a
 
preventive role in reducing the effects of accidents. Single failures in these
 
systems will not inhibit reactor trip, engineered safety feature actuation, or
 
functions required for post accident safe shutdown. The other instrumentation systems required for safety consist of the following:
 
7.1-3                          Rev. 19 WOLF CREEK
: a. Instrumentation and control power supply system
: b. Residual heat removal system isolation valve interlocks
: c. Refueling interlocks
: d. Accumulator motor-operated isolation valve interlocks
: e. Emergency core cooling system switchover from injection
 
mode to recirculation mode
: f. Interlocks for RCS pressure control during low
 
temperature operation
: g. Isolation of nonseismic Category I piping from seismic Category I cooling systems
: h. Interlocks for pressurizer pressure relief system.
: i. Switchover of charging pump suction to RWST on low-low
 
VCT level
: j. Instrumentation for mitigating the consequences of
 
inadvertent boron dilution
: k. Charging pump miniflow interlock
: l. Neutron flux monitoring
 
Item a above is described in Chapter 8.0. Item c is described in Section 9.1.4. The remaining items are described in Section 7.6.
 
7.1.1.6  Control Systems Not Required for Safety Control systems not required for safety are those automatic and manual systems
 
designed for the primary purpose of which is normal load control, startup, and
 
shutdown of the main power generating system. As shown in Section 7.7, malfunctions in these systems do not result in unsafe conditions.
 
7.1.2  IDENTIFICATION OF SAFETY CRITERIA
 
Considerations for instrument errors are included in the accident analyses
 
presented in Chapter 15.0. Functional requirements, developed on the basis of
 
the results of the accident analyses, that  have  utilized  conservative 
 
assumptions and parameters are
 
7.1-4                          Rev. 0 WOLF CREEK used in designing these systems. A preoperational testing program verified the
 
adequacy of the design. Accuracies are given in Sections 7.2, 7.3, and 7.5.
 
The criteria listed in Table 7.1-2 were considered in the design of the systems given in Section 7.1.1. A discussion of compliance with each criterion for
 
systems in its scope is provided in the referenced sections given in Table 7.1-
: 2. Because some criteria were established after design and testing had been
 
completed, the equipment documentation may not meet the format requirements of
 
some standards. Justification for any exceptions taken to each document for
 
systems in its scope is provided in the referenced sections.
 
7.1.2.1  Design Bases The design bases for the safety-related systems are provided in the respective
 
sections of Chapter 7.0.
 
7.1.2.2  Independence of Redundant Safety-Related Systems The safety-related systems are designed to meet the independence and separation
 
requirements of GDC-22 and Section 4.6 of IEEE Standard 279-1971.
 
The electrical power supply, instrumentation, and control conductors for
 
redundant circuits of the WCGS have physical separation to preserve the
 
redundancy and to ensure that no single credible event will prevent operation
 
of the associated function. Critical circuits and functions include power, control, and analog instrumentation associated with the operation of the
 
safety-related systems. Events considered credible and considered in the
 
design include the effects of short circuits, pipe rupture effects, missiles, fire and earthquakes.
 
7.1.2.2.1  General The physical separation criteria for redundant safety-related system sensors, sensing lines, wireways, cables, and components on racks meet the
 
recommendations contained in Regulatory Guide 1.75 with the following comments:
: a. The protection systems use redundant instrumentation
 
channels and actuation trains and incorporate physical
 
and electrical separation to prevent faults in one
 
channel from degrading any other protection channel.
: b. Where no redundant circuits share a single compartment of
 
a safety-related instrumentation rack and these redundant
 
7.1-5                          Rev. 1 WOLF CREEK safety-related instrumentation racks are physically separated, the
 
recommendations of Position C.16 of Regulatory Guide 1.75 do not apply.
: c. Redundant isolated control signal cables leaving the protection racks are brought into close proximity
 
elsewhere in the plant, such as the control board. It
 
could be postulated that electrical faults or
 
interference at these locations might be propagated into
 
all redundant racks and degrade protection circuits
 
because of the close proximity of protection and control
 
wiring within each rack. Regulatory Guide 1.75
 
(Regulatory Position C.4) and IEEE Standard 384-1974
 
(Section 4.5(3)) provide the option to demonstrate by
 
tests that the absence of physical separation could not significantly reduce the availability of Class 1E circuits.
 
Westinghouse test programs have demonstrated that Class
 
1E protection systems (the nuclear instrumentation
 
system, the solid state protection system, and the 7300
 
process protection system) are not degraded by non-Class
 
1E circuits sharing the same enclosure. Conformance to
 
the requirements of IEEE Standard 279-1971 and Regulatory
 
Guide 1.75 has been established and accepted by the NRC, based on the following which is applicable to WCGS.
 
Tests conducted on the as-built designs of the nuclear
 
instrumentation system and solid state protection system
 
were reported and accepted by the NRC in support of the Diablo Canyon application (Docket Nos. 50-275 and 50-323). Westinghouse considers these programs as
 
applicable to all plants, including Wolf Creek Generating
 
Station. Westinghouse tests on the 7300 process control
 
system were covered in a report entitled, "7300 Series
 
Process Control System Noise Tests," subsequently
 
reissued as Reference 2. In a letter dated April 20, 1977 (Ref. 3), the NRC accepted the report in which the
 
applicability to Wolf Creek Generating Station is
 
established.
 
In the Westinghouse 7300 process control system containment spray circuitry, one exception is made regarding isolation of class 1E and non-class 1E circuits. Since containment spray is an energized to actuate function, annunciation in the control room was provided to alert the operators when one of the channels is in test. This is accomplished by using contacts from the channel test cards as inputs to the annunciator. The channel test cards which provide the signals are safety related and the annunciator is non-safety related,          however, the channel test cards have not been qualified as class 1E isolation devices. An analysis of the
 
7.1-6                          Rev. 6 
 
WOLF CREEK circuits shows that a fault in the non-class 1E portion of the annunciator circuit will not degrade the safety related containment spray circuitry below an acceptable level or cause any common mode failure.
: d. The physical separation criteria for instrument cabinets
 
within the NSSS scope meet the recommendations contained
 
in Section 5.7 of IEEE Standard 384-1974. Compliance
 
with specific positions of Regulatory Guide 1.75 is given in Sections 8.1.4.3 and 8.3.1.4.
 
7.1.2.2.2  Specific Systems
 
Independence is maintained throughout each system, extending from the  sensor 
 
through  to  the  devices  actuating  the  protective
 
function. Physical separation is used to achieve separation of redundant
 
transmitters. Separation of field wiring is achieved using separate wireways, cable trays, conduit runs, and containment penetrations for each redundant protection channel set. Redundant analog equipment is separated by locating modules in different protection rack sets. Each redundant channel set is
 
energized from a separate ac power feed.
 
There are four separate protection sets. Each protection set contains several
 
channels, each channel sensing a different variable. Separation of redundant
 
analog channels begins at the process sensors and is maintained in the field
 
wiring, containment penetrations, and process protection cabinets. Protection
 
sets are formed at the process protection cabinets and SSPS transmit the
 
required signals to the redundant trains in the logic racks (Figure 7.1-1). 
 
Redundant analog channels are separated by locating modules in different
 
cabinets. Since all equipment within any cabinet is associated with a single
 
protection set, there is no requirement for separation of wiring and components
 
within the cabinet. See Section 7.1.2.3 for additional information.
 
In the nuclear instrumentation system and the solid state protection system cabinets where redundant channel instrumentation is physically adjacent, there
 
are no wireways or cable penetrations which would permit a fire resulting from
 
electrical failure in one channel to propagate into redundant channels.
 
Two reactor trip breakers are actuated by two separate logic matrices to
 
interrupt power to the control rod drive mechanisms. The breaker main contacts
 
are connected in series with the power supply so that opening either breaker
 
interrupts power to all control rod drive mechanisms, permitting the rods to
 
free fall into the core.
 
7.1-7                          Rev. 6 
 
WOLF CREEK
: a. Reactor trip system
: 1. Separate routing is maintained for the four basic
 
reactor trip system channel sets, analog sensor
 
signals, bistable output signals, and power supplies for these systems. The separation of these four
 
channel sets is maintained from sensors to instrument
 
cabinets to logic system input cabinets.
: 2. Separate routing of the redundant reactor trip
 
signals from the redundant logic system cabinets is
 
maintained, and, in addition, the cables carrying
 
these signals are separated (by spatial separation or
 
by provision of barriers or by separate cable trays
 
or wireways) from the four analog channel sets.
: b. Engineered safety feature actuation system
: 1. Separate routing is maintained for the four basic
 
sets of engineered safety feature actuation system
 
analog sensing signals, bistable output signals, and
 
power supplies for these systems. The separation of
 
these four channel sets is maintained from sensors to
 
instrument cabinets to logic system input cabinets.
: 2. Separate routing of the engineered safety feature
 
actuation signals from the redundant logic system
 
cabinets is maintained. In addition, they are
 
separated by spatial separation or by provisions of
 
barriers or by separate cable trays or wireways from the four analog channel sets.
: 3. Separate routing of control and power circuits
 
associated with the operation of engineered safety
 
feature equipment is required to retain redundancies
 
provided in the system design and power supplies.
: c. Instrumentation and control power supply system
 
The separation criteria presented also apply to the power
 
supplies for the load centers and busses distributing
 
power to redundant components and to the control of these
 
power supplies.
 
Reactor trip system, engineered safety feature actuation system, and other
 
safety-related system analog circuits may be routed in the same wireways
 
provided the circuits have the same power supply and channel set identified (I, II, III, or IV).
 
7.1.2.2.3  Fire Protection
 
For electrical equipment, noncombustible or fire retardant material is
 
specified.
 
Braided sheathed material used in the cables is noncombustible. For in-field
 
wiring, cables in the power trays are sized using derating factors listed in
 
IPCEA Publication P-46-426.
 
7.1-8                          Rev. 6 WOLF CREEK For early warning protection against propagation of electrical fires, high
 
sensitivity detectors are provided for fire detection and alarm in remote
 
wireways or other unattended areas where large concentrations of cables are
 
installed.
 
Details of the plant's fire protection system are provided in Section 9.5.1.
 
The electrical power supply, instrumentation, and control wiring for redundant
 
circuits have physical separation to preserve redundancy and ensure that no
 
single credible event will prevent operation of the associated function. 
 
Critical circuits include power, control, and analog instrumentation associated
 
with the operation of the reactor trip system or engineered safety feature
 
actuation systems. Credible events include the effects of short circuits, pipe
 
rupture, pipe whip, high-pressure jets, missiles, fire, and earthquake. These events are considered in the basic plant design.
 
Physical space or barriers are provided between separation groups performing
 
the same protective function.
 
In locations where a specific hazard exists (missile, jet, etc.) which could
 
produce damage to safety-related controls and instrumentation required as an
 
active functional part of a nuclear safety-related system, the physical
 
separation, structural protection, or armor provided is adequate to ensure that
 
no multiple failures can result from a single event.
 
The minimum protection or spacing maintained between redundant safety-related
 
control and instrumentation components is:
: a. In open space See the discussion of compliance with Regulatory Guide
 
1.75 (Appendix 3A).
: b. Inside control panels or cabinets, except as noted in
 
Section 7.1.2.2.lb, the minimum separation criteria are:
: 1. Six inches of free space, or
: 2. If a barrier is present, one inch plus the barrier.
 
See also Section 8.3.1.4.1.
 
The criteria and bases for the independence of electrical cable, including
 
routing, marking, and cable derating, are covered in Section 8.3. Fire
 
detection and protection in the areas where wiring is installed is covered in
 
Section 9.5.1.
 
7.1.2.3  Physical Identification of Safety-Related Equipment All components required as part of the safety-related control and
 
instrumentation systems are identified as safety-related components requiring
 
formal quality assurance and supporting documentation. Specific requirements for each type of component are covered in its procurement specification. The
 
Operating Quality program is described in the Quality Program Manual.
7.1-9                          Rev. 21 WOLF CREEK All panels and cabinets which contain one or more safety-related devices are subject to the requirements for safety-related systems.
 
Instrument racks and trays containing tubing or wiring connected to safety-
 
related instrumentation devices are subject to the requirements for safety-related systems.
 
Safety-related systems and their component devices are identified as to their
 
separation group. Each protection set described in Section 7.1.2.2.2 is
 
included in its respective separation group.
 
There are four separation groups identifiable with process equipment associated
 
with the RTS and ESFAS. A separation group may be comprised of more than a
 
single process equipment cabinet. The color coding of each process equipment
 
rack nameplate coincides with the color code established for the separation group of which it is a part. Redundant BOP channels are separated by locating them in different equipment cabinets. Separation of redundant channels begins
 
at the process sensors and is maintained in the field wiring, containment
 
penetrations, and equipment cabinets to the redundant trains in the logic
 
racks. The NSSS solid state protection system input cabinets and the NSSS
 
engineered safety feature actuation systems are divided into isolated
 
compartments, each serving one of the redundant input channels. Horizontal
 
1/8-inch-thick solid steel barriers, coated with fire retardant paint, separate
 
the compartments. One-eighth-inch-thick solid steel wireways coated with fire
 
retardant paint enter the input cabinets vertically. The wireway for a
 
particular compartment is open only into that compartment so that flame could
 
not propagate to affect other channels. At the logic racks, the separation
 
group color coding for redundant channels is clearly maintained until the
 
channel loses its identity in the redundant logic trains. The color coded
 
nameplates described below provide identification of equipment associated with protective functions and their channel group association:
 
Protection Set I Separation Group 1: red with white lettering Protection Set II Separation Group 2: white with black lettering Protection Set III Separation Group 3: blue with white lettering Protection Set IV Separation Group 4: yellow with black lettering Nonsafety-related: black with white lettering
 
Within the control panels, where more than one separation group is present, wiring is identified by separation group or if the wiring is enclosed by
 
conduit the separation group identification is located on the conduit.
 
7.1-10    Rev. 0 WOLF CREEK Within a cabinet or panel associated and identified with a single safety-
 
related separation group, no identification of the safety-related wiring is
 
required. The separation group of the panel or cabinet, however, is clearly
 
identified.
 
Within a panel or cabinet otherwise associated and identified with a single
 
safety-related separation group, nonsafety-related wiring is clearly
 
identified. However, provided such nonsafety-related wiring is maintained at a
 
small quantity, identification of the safety-related wiring is not required.
 
All noncabinet-mounted protective equipment and components are provided with an
 
identification tag or nameplate. Small electrical components, such as relays, have nameplates on the enclosure which houses them. All cables are numbered
 
with identification tags. In congested areas, such as under or over the control boards, instrument racks, etc., cable trays and conduits containing redundant circuits shall be identified, using permanent markings. The purpose
 
of such markings is to facilitate cable routing identification for future
 
modifications or additions. Positive permanent identification of cables and/or
 
conductors are made at all terminal points. There are also identification
 
nameplates on the input panels of the solid state protection system.
 
Fire-resistive cables, with stainless steel jacketing, are routed as separate conduits, and numbered with permanent identification.
7.1.2.4  Conformance to Criteria
 
A listing of applicable criteria and the sections where conformance is
 
discussed is given in Table 7.1-2.
 
7.1.2.5  Conformance to NRC Regulatory Guides 7.1.2.5.1  General
 
Conformance of BOP equipment to Regulatory Guides 1.22, 1.53, 1.62, 1.105, and 1.118 is addressed in Tables 7.1-3, 4, 5, 6, and 7, respectively.
 
Other regulatory guides pertinent to this section are:  1.7, 1.11, 1.21, 1.26, 1.29, 1.30, 1.40, 1.45, 1.47, 1.63, 1.68, 1.73, 1.75, 1.80, 1.89, 1.97, 1.100, 1.106 and 1.139. References to discussions of these regulatory guides are
 
provided in Appendix 3A.
 
An additional discussion of the NSSS conformance to Regulatory Guide 1.22 and
 
IEEE-338 and -379 is given in the following sections.
 
7.1-11    Rev. 24 WOLF CREEK 7.1.2.5.2  Conformance to Regulatory Guide 1.22
 
Periodic testing of the reactor trip and engineered safety feature actuation
 
systems, as described in Sections 7.2.2 and 7.3, complies with Regulatory Guide 1.22, "Periodic Testing of Protection System Actuation Functions."
 
Where the ability of a system to respond to a bona fide accident signal is
 
intentionally bypassed for the purpose of performing a test during reactor
 
operation, each bypass condition is automatically indicated to the reactor
 
operator in the main control room by a separate annunciator for the channel in
 
test. Test circuitry does not allow two channels to be tested at the same time
 
so that extension of the bypass condition to the redundant system is prevented.
 
The actuation logic for the RTS and ESFAS is tested as described in Sections 7.2 and 7.3. As recommended by Regulatory Guide 1.22, where actuated equipment is not tested during reactor operation it has been determined that:
: a. There is no practicable system design that would permit
 
operation of the actuated equipment without adversely
 
affecting the safety or operability of the plant.
: b. The probability that the protection system would fail to
 
initiate the operation of the actuated equipment is, and
 
can be maintained, acceptably low without testing the
 
actuated equipment during reactor operation.
: c. The actuated equipment can routinely be tested when the
 
reactor is shut down.
 
The list of equipment that is not tested at full power so as not to damage equipment or upset plant operation is:
: a. Manual actuation switches (RTS and ESFAS)
: b. Main turbine trip system (actual trip)
: c. Main steam isolation valves (actual full closure)
: d. Main feedwater isolation valves (actual full closure)
: e. Feedwater control valves (actual full closure)
: f. Main feedwater pump trip solenoids
: g. Reactor coolant pump seal water return valves (actual
 
full closure)
 
7.1-12    Rev. 0 WOLF CREEK
: h. Five selected slave relays
: i. Pressurizer power operated relief valves
 
The justifications for not testing the above items at full power are discussed
 
below.
: a. Manual actuation switches for RTS and ESFAS
 
These would cause initiation of their protection system
 
function at power, causing plant upset and/or reactor
 
trip. It should be noted that the reactor trip function
 
that is derived from the automatic safety injection
 
signal is tested at power in the same manner as the other
 
analog signals and as described in Section 7.2.2.2.3.
 
The processing of these signals in the solid state
 
protection system wherein their channel orientation
 
converts to a logic train orientation is tested at power
 
by the built-in semiautomatic test provisions of the
 
solid state protection system. The reactor trip breakers
 
are tested at power, as discussed in Section 7.2.2.2.3.
: b. Main turbine trip system
 
Testing of the main turbine trip function under power
 
operation is discussed in Section 10.2.3.6. Since the actual Turbine Trip cannot be actuated during power operation, the ESFAS Turbine Trip function is tested in a series of overlapping tests through the Ovation Turbine Control System (TCS) Testable Dump Manifold (TDM) solenoid valves as discussed in Section 7.3.8.
: c. Closing the main steam isolation valves
 
See Table 7.1-3.
: d. Closing the main feedwater isolation valves
 
See Table 7.1-3.
: e. Closing the feedwater control valves
 
These valves are routinely tested during refueling
 
outages. To close them at power would adversely affect
 
the operability of the plant. The verification of
 
operability of feedwater control valves at power is
 
ensured by confirmation of proper operation of the steam
 
generator water level control system. The actuation
 
function of the solenoids, which provides the closing
 
function, is periodically tested at power, as discussed
 
in Section 7.3. The operability of the slave relay which
 
actuates the solenoid, which is the actuating device, is
 
verified during this test. Although the closing of these
 
control valves is blocked when the slave relay is tested, all functions are tested to ensure that no electrical
 
malfunctions have occurred which could defeat the
 
protective
 
7.1-13    Rev. 27 WOLF CREEK function. It is noted that the solenoids work on the
 
deenergize-to-actuate principle, so that the feedwater
 
control valves will fail close upon either the loss of
 
electrical power to the solenoids or loss of air
 
pressure.
 
Based on the above, the testing of the isolating function
 
of feedwater control valves meets the guidelines of
 
Regulatory Position D.4 of Regulatory Guide 1.22.
: f. Main feedwater pump trip solenoids
 
Automatic tripping of the feedwater pumps is not part of the primary success path for accident mitigation, and, therefore, this function does not require periodic testing online. This function can be tested during the refueling outage.
: g. Reactor coolant pump seal water return valves (close)
 
Seal water return line isolation valves are routinely
 
tested during refueling outages. Closure of these valves
 
during operation would cause the seal water system relief
 
valve to lift with the possibility of valve chatter.
 
Valve chatter could damage this relief valve. Testing of
 
these valves at power could cause equipment damage.
 
Therefore, these valves are tested during scheduled
 
refueling outages. Thus, the guidelines of Regulatory
 
Position D.4 of Regulatory Guide 1.22 are met.
: h. Five selected slave relays
 
Slave relays K602, K622, K630, K740, and K741 and their
 
actuated equipment are tested at least once per 18 months
 
during refueling and during each cold shutdown exceeding
 
24 hours unless they have been tested within the previous
 
90 days. Justification for the extended test interval is
 
based on plant operational concerns, and was presented in
 
detail in Reference 4.
: i. Pressurizer power operated relief valves
 
The pressurizer power operated relief valves require
 
system pressure to enter special actuator ports in order
 
to drive the valves open. The pressurizer relief
 
isolation valves must be open to ensure the necessary
 
system pressure is available for pressurizer power
 
operated relief valve actuation. Since the pressurizer
 
relief isolation valves must also be open, opening of the
 
pressurizer power operated relief valves at power would
 
initiate a RCS depressurization transient which would
 
cause a plant upset. All functions are tested to ensure
 
that no electrical malfunctions have occurred which could
 
defeat the protective function.
 
7.1-14    Rev. 27 WOLF CREEK 7.1.2.6  Conformance to IEEE Standards
 
7.1.2.6.1  Conformance to IEEE Standard 379-1972
 
The principles described in IEEE Standard 379-1972 were used in the design of
 
the Westinghouse protection system. The system complies with the intent of
 
this standard and the additional guidance of Regulatory Guide 1.53, although
 
the formal analyses have not been documented exactly as outlined. Westinghouse
 
has gone beyond the required analyses and has performed a fault tree analysis (Ref. 1).
 
The referenced report provides details of the analyses of the protection
 
systems previously made to show conformance with the single failure criterion
 
set forth in Section 4.2 of IEEE Standard 279-1971. The interpretation of the single failure criterion provided by IEEE Standard 379-1972 does not indicate substantial differences with the Westinghouse interpretation of the criterion, except in the methods used to confirm design reliability. The RTS and ESFAS
 
are each redundant safety systems. The required periodic testing of these
 
systems discloses any failures or loss of redundancy which could have occurred
 
in the interval between tests, thus ensuring the availability of these systems.
 
7.1.2.6.2  Conformance to IEEE Standard 338-1971
 
The periodic testing of the RTS and ESFAS conforms to the requirements of IEEE
 
Standard 338-1971 with the following comments:
: a. The surveillance requirements in the Technical
 
Specifications for the protection system ensure that the
 
functional operability is maintained comparable to the original design standards. Periodic tests at the established intervals demonstrate this capability for the
 
system.
 
For sensors, the WCGS design permits periodic response
 
time testing. The methods of testing fall into two
 
categories as follows:
 
PRIMARY  -  For resistance temperature detectors (RTDs),
a loop current step response methodology is
 
used as endorsed in NUREG-0809 and described
 
in detail in EPRI report NP-834 (Vol. 1).
 
                  -  For pressure sensors, the EPRI developed
 
method described in report NP-267 is used.
 
This pressure ramp testing is also discussed
 
in ISA dS-67.06.
 
7.1-15    Rev. 5 WOLF CREEK AUXILIARY -  RTDs and pressure sensors may be tested using the noise analysis method which
 
functions  on the principle that, in the
 
protection system, sensors are sensitive to
 
process noise created by natural perturbations in variables, including
 
temperature, pressure, and flow. The noise
 
analysis method testing system is designed
 
to measure sensor response time and/or
 
assess degradation by measurement of the
 
sensors' efficiency to detect high-frequency
 
noise.
 
Nuclear instrumentation detectors are excluded since
 
delays attributable to them are negligible in the overall channel response time required for safety.
 
The measurement of response time at the specified time
 
intervals provides assurance that the protective and
 
engineered safety feature function associated with each
 
channel is completed within the time limit assumed in the
 
accident analyses.
: b. The reliability goals specified in Section 4.2 of IEEE
 
Standard 338-1971 are consistent with the test frequency
 
in the Technical Specifications .
: c. The periodic time interval discussed in Section 4.3 of
 
IEEE Standard 338-1971, and specified in the Technical
 
Specifications, is selected to ensure that equipment associated with protection functions has not drifted beyond its minimum performance requirements. The
 
adequacy of the interval will be verified by results of
 
testing or the interval will be reevaluated on the basis
 
of actual experience.
: d. The test interval discussed in Section 5.2 of IEEE
 
Standard 338-1971 is developed primarily on past
 
operating experience and modified, if necessary, to
 
ensure that system and subsystem protection is reliably
 
provided.
 
7.
 
==1.3  REFERENCES==
: 1. Gangloff, W.C. and Loftus, W.D., "An Evaluation of Solid
 
State Logic Reactor Protection in Anticipated Transients,"
WCAP-7706-L (Proprietary) and WCAP-7706 (Non-Proprietary),
July, 1971.
: 2. Marasco, F.W. and Siroky, R.M., "Westinghouse 7300 Series
 
Process Control System Noise Tests," WCAP-8892-A, June, 1977.
: 3. Letter dated April 20, 1977, R.L. Tedesco (NRC) to
 
C. Eicheldinger (Westinghouse).
: 4. Letter dated February 27, 1984, N. A. Petrick (SNUPPS) to
 
Mr. Harold R. Denton (NRC), SLNRC 84-0038.
 
7.1-16    Rev. 5 WOLF CREEK TABLE  7.1-1 INSTRUMENTATION SYSTEMS IDENTIFICATION
 
Designer  Similar To Plant    Safety-Related Systems or Categories Westinghouse Bechtel  Comanche Peak  W. B. McGuire and Watts Bar Other  Remarks              1. Reactor trip  X    X  X
: 2. Engineered safety feature actuation system
: a. Emergency core cooling  X    X  X
: b. Main steam and feedwater isolation  X    X  X    MSFIS actuators and controls replaced in RF16. See 7.3.7  c. Containment isolation  X    X  X
: d. Containment heat removal  X    X  X
: e. Containment combustible gas control    X      Millstone Unit 2    f. Containment purge isolation    X        New (see 7.3.2)  g. Fuel building exhaust isolation    X      ---  New (see 7.3.3)  h. Control room ventilation isolation    X      ---  New (see 7.3.4)  i. Auxiliary feedwater  X  X  X  X  ---  New supply configuration (see 7.3.6)
: 3. Systems required for post accident safe shutdown
: a. Hot standby  X    X  X
: b. Cold shutdown  X    X  X
: c. Shutdown from outside control room  X  X  X    ---  New (see 7.4.3)
: 4. Safety-related display instrumentation
: a. Reactor trip  X    X  X
: b. Engineering safety feature actuation system  X  X  X  X      c. Systems required for post accident safe shutdown  X  X  X  X Rev. 24 WOLF CREEK WOLF CREEK TABLE  7.1-1 (sheet 2)
INSTRUMENTATION SYSTEMS IDENTIFICATION Designer  Similar To Plant    Safety-Related Systems or Categories Westinghouse Bechtel  Comanche Peak  W. B. McGuire and Watts Bar Other  Remarks              5. Other instrumentation systems required for safety
: a. Vital instrument ac power supply    X      Trojan    b. Residual heat removal isolation valves  X    X        c. Refueling interlocks  X    X
: d. Monitoring combustible gas in containment    X      Millstone Unit 2 
 
Rev. 14 WOLF CREEK IDENTIFICATION OF SAFETY CRITERIA TABLE 7.1-2 SHEET 1                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)                                CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVES  CENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER  COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLING SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9)1QUALITY STANDARDS AND RECORDS*******************2DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA*******************
I. OVERALL 3FIRE PROTECTION*******************
REQUIREMENTS4ENVIRONMENTAL AND MISSILE DESIGN BASES*******************5SHARING OF STRUCTURES, SYSTEMS AND COMPONENTS*******************
II. PROTECTION 12SUPPRESSION OF REACTOR POWER OSCILLATIONS
*__________________BY MULTIPLE13INSTRUMENTATION AND CONTROL*******************
FISSION14REACTOR COOLANT PRESSURE BOUNDARY
___________________PRODUCT15REACTOR COOLANT SYSTEM DESIGN
*__________________BARRIERS19CONTROL ROOM*******************20PROTECTION SYSTEMS FUNCTIONS
*_*****************
III. PROTECTION21PROTECTION SYSTEM RELIABILITY AND TESTABILITY*******************
AND REACTIVITY22PROTECTION SYSTEM INDEPENDENCE*******************
CONTROL23PROTECTION SYSTEM FAILURE MODES
*_*****************
SYSTEMS24SEPARATION OF PROTECTION AND CONTROL SYSTEMS*******************25PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS
*______*___________SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC GENERAL DESIGN CRITERIA (NOTE 1 AND 5)ENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 WOLF CREEK IDENTIFICATION OF SAFETY CRITERIA TABLE 7.1-2 SHEET 2                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVESCENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLINGSAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9)30QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY
___________________34RESIDUAL HEAT REMOVAL
_____*__*****_*____35EMERGENCY CORE COOLING
_____********_*____37TESTING OF EMERGENCY CORE COOLING SYSTEM
_____********_*____IV. FLUID38CONTAINMENT HEAT REMOVAL
_______*___*_**____ SYSTEMS 40TESTING OF CONTAINMENT HEAT REMOVAL SYSTEM
______**___*_**____41CONTAINMENT ATMOSPHERE CLEANUP
_**____*______*____43TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEM
_**____*______*____44COOLING WATER
_____*__*__****__*_46TESTING OF COOLING WATER SYSTEM
_____*__*__****__*_54PIPING SYSTEMS PENETRATING CONTAINMENT
_**__**____**______V. REACTOR 55REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT
___________________CONTAINMENT56PRIMARY CONTAINMENT ISOLATION
_**________________57CLOSED SYSTEMS ISOLATION VALVES
_____**____**______VI. FUEL AND60CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT
_***_____________*_RADIOACTIVITY63MONITORING FUEL AND WASTE STORAGE
___*_______________CONTROL64MONITORING RADIOACTIVITY RELEASE
__**_____________*_SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC GENERAL DESIGN CRITERIA (NOTE 1 AND 5)(NOTE 2)ENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 WOLF CREEKIDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET 3 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)
BYPASSES (SECTION 7.5..2.2)
STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)
COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)
BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)SECTION 7.71QUALITY STANDARDS AND RECORDS
__*___**___**_**_*_I. OVERALL 2DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA
__*___**___**_**_*_REQUIREMENTS3FIRE PROTECTION
__*___**___**_**_*_4ENVIRONMENTAL AND MISSILE DESIGN BASES
__*___**___**_**_*_5SHARING OF STRUCTURES, SYSTEMS AND COMPONENTS
__*___**___**_**_*_II. PROTECTION 12SUPPRESSION OF REACTOR POWER OSCILLATIONS
__________________*BY MULTIPLE13INSTRUMENTATION AND CONTROL
***_****_****_****_FISSION14REACTOR COOLANT PRESSURE BOUNDARY
___________________PRODUCT15REACTOR COOLANT SYSTEM DESIGN
____________*______BARRIERS19CONTROL ROOM
__*__***__*_*_**_*_20PROTECTION SYSTEMS FUNCTIONS
____________*_*__*_III. PROTECTION21PROTECTION SYSTEM RELIABILITY AND TESTABILITY
___________**_**_*_AND REACTIVITY22PROTECTION SYSTEM INDEPENDENCE
___________**_**_*_CONTROL23PROTECTION SYSTEM FAILURE MODES
_______________*_**SYSTEMS24SEPARATION OF PROTECTION AND CONTROL SYSTEMS
___________________25PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS
___________________NRC GENERAL DESIGN CRITERIA (NOTE 1 AND 5)SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)RTSESFS(SECTION7.5.2)SAFESHUTDOWN(SECTION7.5.3)CONTROLSYSTEMS NOTREQUIREDFORSAFETY Rev. 11 WOLF CREEKIDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET 4 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)
BYPASSES (SECTION 7.5..2.2)
STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)
COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)
BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)SECTION 7.730QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY
____________*_*____34RESIDUAL HEAT REMOVAL
____________*__*___35EMERGENCY CORE COOLING
____________*_**_*_37TESTING OF EMERGENCY CORE COOLING SYSTEM
__*__**___*_*_**_*_IV. FLUID38CONTAINMENT HEAT REMOVAL
_________________*_ SYSTEMS 40TESTING OF CONTAINMENT HEAT REMOVAL SYSTEM
__*__**__________*_41CONTAINMENT ATMOSPHERE CLEANUP
___________________43TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEM
__*__*_____________44COOLING WATER
_______________*_*_46TESTING OF COOLING WATER SYSTEM
__*__*___________*_54PIPING SYSTEMS PENETRATING CONTAINMENT
____________*______V. REACTOR 55REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT
____________*______CONTAINMENT56PRIMARY CONTAINMENT ISOLATION
___________________57CLOSED SYSTEMS ISOLATION VALVES
____________*______VI. FUEL AND60CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT
___________________RADIOACTIVITY63MONITORING FUEL AND WASTE STORAGE
___________________CONTROL64MONITORING RADIOACTIVITY RELEASE
__*________________NRC GENERAL DESIGN CRITERIA (NOTE 1 AND 5)SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)CONTROLSYSTEMS NOTREQUIREDFORSAFETYRTSESFS(SECTION7.5.2)SAFESHUTDOWN(SECTION7.5.3)(NOTE 2)Rev. 13 WOLF CREEK IDENTIFICAITON OF SAFETY CRITERIA TABLE 7.1-2 SHEET 5                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)                                CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVES CENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLING SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9) 1.7 CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS OF COOLANT ACCIDENT
_**________________1.11INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT
_**____*___________1.21MEASURING AND REPORTING OF EFFLUENTS FROM NUCLEAR POWER PLANTS
__**____________*__1.22PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS (NOTE 3A)
********_____***___1.26QUALITY GROUP CLASSIFICATIONS AND STANDARDS
_******_******_*_*_1.29SEISMIC DESIGN CLASSIFICATION
*******************
1.30 QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF
 
INSTRUMENTATION AND ELECTRIC EQUIPMENT (NOTE 3B)
*******************
1.40 QUALIFICATION TESTS OF CONTINUOUS -DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF
 
WATER-COOLED NUCLEAR POWER PLANTS (NOTE 3C)
_*___________
*_____1.45REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEM (NOTE 3A)
_____________*_____1.47 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS (NOTE 3A)******************
_1.53 APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION
 
SYSTEMS (NOTES 3A AND 3D)
*******************1.62MANUAL INITIATION OF PROTECTIVE ACTIONS
*_******___________SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC REGULATORY GUIDES (NOTE 4)ENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 WOLF CREEK IDENTIFICAITON OF SAFETY CRITERIA TABLE 7.1-2 SHEET 6                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVES CENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLING SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9) 1.63 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (NOTE 3E)
**_____*___***_____1.68INITIAL TEST PROGRAM FOR WATER-COOLED REACTOR POWER PLANTS
*******************
1.73 QUALIFICATION TESTS OF ELECTRIC VLAV3E OPERATORS INSTALLED INSIDE THE CONTAINMENT
 
OF NUCLEAR POWER PLANTS (NOTE 3F)
_**________***_____1.75PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS (NOTES 3A AND 3G)
*******************1.80PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS
_____**_**_*__*____1.89QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS (NOTES 3H AND 3I)
*******************
1.97 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT AND
 
ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT (ICSB BTP 10)
_*____________
*____1.100 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS (NOTES 3H AND
 
3I)  (ICSB BTP 10)
*******************1.105INSTRUMENT SETPOINTS
*_*******____***__1.106 THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR OPERATED VALVES (ICSB
 
BTP 27)_*****__*_****_***_1.118PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS (NOTES 3A, 3J AND 3K)
*******************1.139GUIDANCE FOR RESIDUAL HEAT REMOVAL (SEE SECTION 5.0 APPENDIX 5.4A)
___________________SAFETY-RELATED SAFETY-RELATED AS DEFINED IN SECTION 7.1.1
*******************
IEEE STANDARD279CRITERIA FOR PROTECTION SYSTEMS FOR NUCLEAR POWER GENERATING STATIONS
*******************SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC REGULATORY GUIDES (NOTE 4)ENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 WOLF CREEKIDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET 7 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)BYPASSES (SECTION 7.5..2.2)STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)FIRE PROTECTION AND DETECTION (SECTION 7.6.9)SECTION 7.7 1.7 CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS OF COOLANT ACCIDENT
__*__*______________1.11INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT
___________________*1.21MEASURING AND REPORTING OF EFFLUENTS FROM NUCLEAR POWER PLANTS
___________________*1.22PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS (NOTE 3A)
__*__***__**________1.26QUALITY GROUP CLASSIFICATIONS AND STANDARDS
____________*_*__*_*1.29SEISMIC DESIGN CLASSIFICATION
__*___**___**_**_*__1.30 QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING
 
OF INSTRUMENTATION AND ELECTRIC EQUIPMENT (NOTE 3B)
__*___**___**_**_*__1.40 QUALIFICATION TESTS OF CONTINUOUS -DUTY MOTORS INSTALLED INSIDE THE
 
CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS (NOTE 3C)____________________1.45REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS (NOTE 3A)
____________________1.47 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY
 
SYSTEMS (NOTE 3A)
_*_**___**_**_**_*__1.53 APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT
 
PROTECTION SYSTEMS (NOTES 3A AND 3D)
__*___**___**_**_*__1.62MANUAL INITIATION OF PROTECTIVE ACTIONS
__*_________________SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)
RTSESFS(SECTION7.5.2)SAFESHUTDOWN(SECTION7.5.3)CONTROLSYSTEMS NOTREQUIREDFORSAFETYNRC REGULATORY GUIDES (NOTE 4)
Rev. 11 WOLF CREEKIDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET 8 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)BYPASSES (SECTION 7.5..2.2)STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)FIRE PROTECTION AND DETECTION (SECTION 7.6.9)SECTION 7.7 1.63 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (NOTE 3E)
__*********_*_*__**_1.68INITIAL TEST PROGRAM FOR WATER-COOLED REACTOR POWER PLANTS*******************_1.73 QUALIFICATION TESTS OF ELECTRIC VLAV3E OPERATORS INSTALLED INSIDE THE
 
CONTAINMENT OF NUCLEAR POWER PLANTS (NOTE 3F)____________
*_*__**_1.75PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS (NOTES 3A AND 3G)
__*___**____*_**_*__1.80PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS
_________________*__1.89 QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS (NOTES 3H AND 3I)__*___**___**_**____1.97 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS
 
PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT (ICSB BPT 10)
__*___*____*___*____1.100 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER POINTS (NOTES
 
3H AND 3I)  (ICSB BPT 10)
__*___**___**_**_*__1.105INSTRUMENT SETPOINTS
_______________*_*__1.106 THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR OPERATED
 
VALVES (ICSB BPT 10)____________
*_**_*__1.118 PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS (NOTES 3A , 3J AND 3K)__*___**___**_**_*__1.139GUIDANCE FOR RESIDUAL HEAT REMOVAL (SEE SECTION 5.0 APPENDIX 5.4A)
____________________SAFETY-RELATED SAFETY-RELATED AS DEFINED IN SECTION 7.1.1
__*___**___**_**_*__IEEE STANDARD279CRITERIA FOR PROTECTION SYSTEMS FOR NUCLEAR POWER GENERATING STATIONS
__*___**___**_**_*__SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)CONTROLSYSTEMS NOTREQUIREDFORSAFETY RTSESFS(SECTION7.5.2)NRC REGULATORY GUIDES (NOTE 4)SAFESHUTDOWN(SECTION 7.5.3 Rev. 11 WOLF CREEK IDENTIFICATION OF SAFETY CRITERIA TABLE 7.1-2 SHEET 9                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)                                CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVES CENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLING SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9)1BACKFITTING OF THE PROTECTION AND EMERGENCY POWER SYSTEMS OF NUCLEAR REACTORS
___________________3ISOLATION OF LOW PRESSURE SYSTEMS FROM THE HIGH PRESSURE REACTOR COOLANT SYSTEM
___________________4REQUIREMENTS  ON MOTOR-OPERATED VALVES IN THE ECCS ACCUMULATOR LINES
___________________5SCRAM BREAKER TEST REQUIREMENTS TECHNICAL SPECIFICATIONS
*__________________9DEFINITION AND USE OF "CHANNEL CALIBRATION"-TECHNICAL SPECIFICATIONS
*__________________12 PROTECTION SYSTEM TRIP POINT CHANGES FOR OPERATION WITH REACTOR COOLANT PUMPS OUT OF SERVICE
*__________________13DESIGN CRITERIA FOR AUXILIARY FEEDWATER SYSTEMS
_____*_*_*_________14SPURIOUS WITHDRAWALS OF SINGLE CONTROL RODS IN PRESSURIZED WATER REACTORS
___________________15REACTOR COOLANT PUMP BREAKER QUALIFICATION
___________________16CONTROL ELEMENT ASSEMBLY (CEA) INTERLOCKS IN COMBUSTION ENGINEERING REACTORS
___________________18 APPLICATION OF THE SINGLE FAILURE CRITERION TO MANUALLY-CONTROLLED ELECTRICALLY-
 
OPERATED VALVES
*******************SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC ICBS TECHNICAL POSITIONSENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 WOLF CREEK IDENTIFICATION OF SAFETY CRITERIA TABLE 7.1-2 SHEET 10                                                  SYSTEM SAFETY CRITERIAREACTOR TRIP SYSTEM (RTS) (SECTION 7.2)CONTAINMENT COMBUSTIBLE GAS CONTROL  (SECTION 7.3.1)CONTAINMENT PURGE ISOLATION  (SECTION 7.3.2)
FUEL BUILDING VENTILATION ISOLATION  (SECTION 7.3.3)
CONTROL ROOM VENTILATION ISOLATION (SECTION 7.3.4)
AUXILIARY FEEDWATER SUPPLY  (SECTION 7.3.6)
MAIN STEAM AND FEEDWATER ISOLATION  (SECTION 7.3.7)
NSSS ESFAS  (SECTION 7.3.8)
AUXILIARY FEEDWATER CONTROL ATMOSPHERIC RELIEF VALVES CENTRIFUGAL CHARGING SYSTEM CONTROLS ESSENTIAL SERVICE WATER COMPONENT COOLING WATER CONTAINMENT COOLERS EMERGENCY DIESEL GENERATORS CONTROL ROOM VENTILATION EMERGENCY VENTILATION SPENT FUEL POOL COOLING SAFE SHUTDOWN FROM OUTSIDE THE CONTROL ROOM  (SEE NOTE 9) 19 ACCEPTABILITY OF DESIGN CRITERIA FOR HYDROGEN MIXING AND DRYWELL VACUUM RELIEF SYSTEMS___________________
20 DESIGN OF INSTRUMENTATION AND CONTROLS PROVIDED TO ACCOMPLISH CHANGEOVER FROM
 
INJECTION TO RECIRCULATION MODE
_______*___________
21 GUIDANCE FOR APPLICATION OF REGULATORY GUIDE 1.47, "BYPASSED AND INOPERABLE STATUS
 
INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS"___________________
22 GUIDANCE FOR APPLICATION OF REGULATORY GUIDE 1.22, "PERIODIC TESTING OF PROTECTION
 
SYSTEM ACTUATION FUNCTIONS"******************
_25 GUIDANCE FOR THE INTERPRETATION OF GENERAL DESIGN CRITERION 37 FOR TESTING THE
 
OPERABILITY OF THE EMERGENCY CORE COOLING SYSTEM AS A WHOLE
_______*___________26REQUIREMENT OF REACTOR PROTECTION SYSTEM ANTICIPATORY TRIPS
*__________________SYSTEMSREQUIRED FORSAFE SHUTDOWN(SECTION 7.4)NRC ICBS TECHNICAL POSITIONSENGINEERED SAFETYFEATURE SYSTEMS(ESFS)(SECTION 7.3)
Rev. 14 IDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET 11 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)BYPASSES (SECTION 7.5..2.2)STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)FIRE PROTECTION AND DETECTION (SECTION 7.6.9)SECTION 7.7 1 BACKFITTING OF THE PROTECTION AND EMERGENCY POWER SYSTEMS OF NUCLEAR REACTO4S____________________3 ISOLATION OF LOW PRESSURE SYSTEMS FROM THE HIGH PRESSURE REACTOR COOLANT SYSTEM____________*_______4REQUIREMENTS  ON MOTOR-OPERATED VALVES IN THE ECCS ACCUMULATOR LINES
______________*_____5SRAM BREAKER TEST REQUIREMENTS TECHNICAL SPECIFICATIONS
____________________9DEFINITION AND USE OF "CHANNEL CALIBRATION"-TECHNICAL SPECIFICATIONS
*_*__***____________12 PROTECTION SYSTEM TRIP POINT CHANGES FOR OPERATION WITH REACTOR COOLANT PUMPSOUTOFSERVICE
____________________13DESIGN CRITERIA FOR AUXILIARY FEEDWATER SYSTEMS____________________14 SPURIOUS WITGHDRAWALS OF SINGLE CONTROL RODS IN PRESSURIZED WATER
 
REACTORS___________________*15REACTOR COOLANT PUMP BREAKER QUALIFICATION
____________________16 CONTROL ELEMENT ASSEMBLY (CEA) INTERLOCKS IN COMBUSTION ENGINEERING
 
REACTORS____________________18 APPLICATION OF THE SINGLE FAILURE CRITERION TO MANUALLY-CONTROLLED
 
ELECTRICALLY-OPERATED VALVES____________
*_*__*__SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)
RTSESFS(SECTION7.5.2)SAFESHUTDOWN(SECTION7.5.3)CONTROLSYSTEMS NOTREQUIREDFORSAFETY Rev. 11 WOLF CREEKIDENTIFICATION OF SAFETY CRITERIATABLE 7.1-2SHEET12 SYSTEM SAFETY CRITERIATRIP PARAMETERS (SECTION 7.5.1)TRIP STATUS (SECTION 7.5.1)
ACTUATION PARAMETERS (SECTION 7.5.2.1)BYPASSES (SECTION 7.5..2.2)STATUS (SECTION 7.5.2.3)
PERFORMANCE (SECTION 7.5..2.4)
HOT  SHUTDOWN CONTROL (SECTION 7.5.3.1)COLD SHUTDOWN CONTROL (SECTION 7.5.3.2)BYPASSES (SECTION 7.5.3.3)
STATUS (SECTION 7.5.3.4)
PERFORMANCE (SECTION 7.5.3.5)INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM (SECTION 7.6.1)RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES (SECTION 7.6.2)REFUELING INTERLOCKS (SECTION 7.6.3)
ACCUMULATOR MOTOR-OPERATED ISOLATION VALVES INTERLOCKS (SECTION 7.6.4)SWITCHOVER FROM INJECTION TO RECIRCULATION (SECTION 7.6.5)INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION  (SECTION 7.6.6)ISOLATION OF NON-SEISMIC-CATEGORY I COOLING SYSTEMS (SECTIONS 7.6.7 AND 7.6.8)FIRE PROTECTION AND DETECTION (SECTION 7.6.9)SECTION 7.7 19 ACCEPTABILITY OF DESIGN CRITERIA FOR HYDROGEN MIXING AND DRYWELL VACUUM RELIEF SYSTEMS____________________20 DESIGN OF INSTRUMENTATION AND CONTROLS PROVIDED TO ACCOMPLISH CHANGEOVER
 
FROM INJECTION TO RECIRCULATION MODE
__*____________
*____21 GUIDANCE FOR APPLICATION OF REGULATORY GUIDE 1.47, "BYPASSED AND INOPERABLE
 
STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS"___*____*___________
22 GUIDANCE FOR APPLICATION OF REGULATORY GUIDE 1.22, "PERIODIC TESTING OF
 
PROTECTION SYSTEM ACTUATION FUNCTIONS"___________
**_**_*__25 GUIDANCE FOR THE INTERPRETATION OF GENERAL DESIGN CRITERION 37 FOR TESTING
 
THE OPERABILITY OF THE EMERGENCY CORE COOLING SYSTEMS AS A WHOLE______________
**____26REQUIREMENT OF REACTOR PROTECTION SYSTEM ANTICIPATORY TRIPS____________________SAFETY-RELATEDDISPLAYINSTRUMENTATION(SECTION 7.5)ALL OTHERINSTRUMENTATIONSYSTEMS REQUIREDFOR SAFETY(SECTION 7.6)CONTROLSYSTEMS NOTREQUIREDFORSAFETY RTS ESFS(SECTION7.5.2)NRC ICBS TECHNICAL POSITIONSSAFESHUTDOWN(SECTION7.5.3)Rev. 11 WOLF CREEK INDENTIFICATION OF SAFETY CRITERIA FOR THIS TABLE: * = APPLICABLE WITH QUALIFICATIONS IDENTIFIED IN TEXT  _ = NOT APPLICABLE NOTES: 1 CONFORMANCE TO THE GENERAL DESIGN CRITERIA AND REGULATORY GUIDES IS ADDRESSED IN CHAPTER 3.0 2 CONFORMANCE TO GDCs 60, 63, AND 64 ARE ADDRESSED IN SECTIONS 11.5 AND 12.3.4.
3 INCLUDES DISCUSSION OF COMPLIANCE WTH IEEE:
A. 279,    B. 336,    C. 334,    D. 379, E,  317,    F. 382,    G. 384,    H. 323,   
 
I. 344,    J. 338,    K. 308 4 THE FOLLOWING REGULATORY GUIDES LISTED IN STANDARD REVIEW PLAN TABLE 7-1 ARE MORE
 
APPLICABLE TO OTHER SECTIONS OF THE USAR AND ARE
 
DISCESSED ELSEWHERE. SEE SECTION 3A FOR AN INDEX
 
OF THE CROSS-REFERENCES TO THESE OTHER SECTIONS. REGULATORY GUIDES:  1.6 INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR
 
DISTRIBUTION SYSTEMS. 1.12 INSTRUMENTATION FOR EARTHQUAKES  1.32 USE OF IEEE STANDARD 308 "CRITERIA FOR CLASS IE ELECTRIC SYSTEMS FOR NUCLEAR POWER
 
GENERATING STATIONS"  1.67 INSTALLATION OF OVERPRESSURE PROTECTION DEVICES  1.70 STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS  1.78 ASSUMPTIONS FOR EVALUATING THE HABITABILITY OF A NUICLEAR POWER PLANT CONTROL ROOM
 
DURING A POSTULATED HAZARDOUS CHEMICAL
 
RELEASE  1.95 PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST ACCIDENTAL CHLORINE
 
RELEASES  1.120 FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS
 
TABLE 7.1-2 SHEET 13
 
5THE FOLLOWING GDCs LISTED IN STANDARD REVIEW PLAN TABLE 7-1 ARE MORE APPLICABLE TO OTHER SECTIONS OF THE USAR AND ARE DISCUSSED ELSEWHERE. SEE SECTION 3.1 FOR AN
 
INDEX OF THE CROSS-REFERENCES TO THESE OTHER SECTIONS.
GDCs:  10 REACTOR DESIGN  20 REACTIVITY CONTROL SYSTEMS REDUNDANCY AND CAPABILITY 27 COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITY  28 REACTIVITY LIMITS  29 PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCES  33 REACTOR COOLANT MAKEUP  50 CONTAINMENT DESIGN BASES 6THE IEEE STANDARDS LISTED IN STANDARD REVIEW PLAN TABLE 7-1 ARE ALL TREATED UNDER THE DISCUSSION OF APPLICABLE
 
REGULATORY GUIDES. SEE NOTE 4. 710CFR50 PARAGRAPHS 34 AND 55a ARE VERY BROAD IN SCOPE AND COVER THE ENTIRE USAR AND ARE NOT SPECIFIC TO
 
CHAPTER 7.0. PARAGRAPH 36 ON TECHNICAL SPECIFICATIONS
 
IS COVERED IN THE WCGS TECHNICAL SPECIFICATIONS. 8THE FOLLOWING BRANCH TECHNICAL POSITIONS LISTED IN STANDARD REVIEW PLAN TABLE 7-1 HAVE BEEN REPLACED BY
 
REGULATORY GUIDES WHICH ARE DISCUSSED AT LENGTH
 
ELSEWHERE IN THE USAR. SEE SECTION 3A FOR AN INDEX OF
 
THE CROSS-REFERENCES TO THESE OTHER USAR SECTIONS.
T HESE REGULATORY GUIDES ARE LISTED ON SHEETS 5 THROUGH 8 OF TABLE 7.1-2.
BRANCH TECHNICAL POSITIONS REPLACEMENT REGULATORY GUIDE ICSB 10 1.100  ICSB 23 1.97  ICSB 24 1.118  ICSB 27 1.106 9APPLIES ONLY TO THE ESSENTIAL SHORT-T ERM INDICATIONS AND CONTROLS  Rev. 11 WOLF CREEK TABLE 7.1-3 CONFORMANCE TO REGULATORY GUIDE 1.22, REV 0, 2/72, "PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS"  This table demonstrates the conformance of the design of BOP equipment to
 
Regulatory Guide 1.22.
 
Regulatory Guide
 
1.22 Position WCGS Position
: 1. The protection system should    1. The protection system is designed
 
be designed to permit periodic      to permit periodic testing to extend
 
testing to extend to and include    to and include the actuation devices
 
the actuation devices and actu-      and actuated equipment.
 
ated equipment.  (The actuated
 
equipment is included in the
 
periodic tests to provide assur-
 
ance that the protection system
 
can initiate its operation, as
 
required by General Design
 
Criterion 21. This safety guide
 
does not address the functional
 
performance testing of actuated
 
equipment required by other
 
General Design Criteria; neither
 
does it preclude a design that
 
fulfills more than one testing
 
requirement with a single test.)
 
a)  The periodic tests should        a) and b) The periodic tests
 
duplicate, as closely as prac-      do duplicate, as closely as practi-
 
ticable, the performance that        cable, the performance that is
 
is required of the actuation        required of the actuation devices
 
devices in the event of an          in the event of an accident. The
 
accident.                            only actuation devices for which
 
the tests do not completely dupli-
 
b)  The protection system and    cate the performance that is
 
the systems whose operation it      required in the event of an acci-
 
initiates should be designed to      dent are:
 
permit testing of the actuation
 
devices during reactor operation.        (i)    The main steam and feed-
 
water isolation valve actuators--
 
full performance testing of these
 
actuators would result in full
 
closure of the main steam and
 
feedwater isolation valves. The
 
transients that would result
 
under power-generating conditions
 
in the plant would include steam
 
generator water level oscilla-
 
tions, or low-low steam generator
 
Rev. 0 WOLF CREEK TABLE 7.1-3 (Sheet 2)
 
Regulatory Guide
 
1.22 Position WCGS Position
 
water level, and would probably result
 
in reactor trip. The valve actuators can be fully tested, including full closure at high speed, whenever the plant is not in operation.
(ii)    The main turbine trip
 
function--a trip of the main turbine
 
under power-generating conditions would
 
result in a trip of the reactor. The
 
turbine trip function can be fully
 
tested whenever the turbine is not in
 
operation. Testing of the main turbine
 
trip function is further discussed in
 
Section 10.2.3.6.
: 2. Acceptable methods of in-        2.a. through d. In general, the
 
cluding the actuation devices in    protection systems can be tested in
 
the periodic tests of the pro-      accordance with method a. The only
 
tection system are:                  protection systems that cannot be
 
tested in accordance with method a
: a. Testing simultaneously      are the main steam and feedwater
 
all actuation devices and actua-    isolation systems and the auxiliary
 
ted equipment associated with        feedwater system. The systems not
 
each redundant protection system    tested in accordance with method a
 
output signal;                      can all be tested in accordance with
 
methods b and c. Method d need not
: b. Testing all actuation        be used. See Section 10.2.3.6 re-
 
devices and actuated equipment      garding the main turbine trip system.
 
individually or in judiciously
 
selected groups;
 
Rev. 24 WOLF CREEK TABLE 7.1-3 (Sheet 3)
 
Regulatory Guide
 
1.22 Position WCGS Position
: c. Preventing the operation
 
of certain actuated equipment
 
during a test of their actuation
 
devices;
: d. Providing the actuated
 
equipment with more than one
 
actuation device and testing
 
individually each actuation
 
device.
 
Method a. set forth above is the
 
preferable method of including the
 
actuation devices in the periodic
 
tests of the protection system.
 
It shall be noted that the accept-
 
ability of each of the four above
 
methods is conditioned by the
 
provisions of Regulatory Positions
: 3. and 4. below.
: 3. Where the ability of a system    3.a. and b. System bypasses are
 
to respond to a bona fide accident  generally not required for testing;
 
signal is intentionally bypassed    in most cases, the actuated equipment
 
for the purpose of performing a      actually responds to the test signals
 
test during reactor operation:      The only exceptions to these criteria
 
are:
: a. Positive means should be
 
provided to prevent expansion of          i. Bistables--test signals
 
the bypass condition to redundant    are substituted for the actual plant
 
or diverse systems, and              inputs during bistable tests, and
 
provisions are included for bypass-
: b. Each bypass condition        ing bistable outputs. The bistables
 
should be individually and          not under test, all digital inputs, automatically indicated to the      and all other portions of the pro-
 
reactor operator in the main        tection system are not affected.
 
control room.
 
ii. Main steam and feedwater
 
isolation valves--the signals to these
 
valves are held in a condition that
 
prevents valve motion during a portion
 
of the test.
 
iii. Auxiliary feedwater sys-
 
tem--the auxiliary feedwater system
 
configuration is altered during test
 
to prevent accidental injection of
 
Rev. 0 WOLF CREEK TABLE 7.1-3 (Sheet 4)
 
Regulatory Guide
 
1.22 Position WCGS Position
 
auxiliary feedwater into the steam
 
generators and to prevent the
 
introduction of essential service
 
water, which is not chemically
 
controlled, into the chemically
 
controlled portions of the system.
 
Test signal injection into a bistable
 
is effected by means of a momentary
 
test switch so that the normal input
 
signal cannot continue to be overridden
 
after the operator releases the
 
switch. Bistable bypass can be
 
effected only by means of keylock
 
switches. The keying and access to the
 
keys and to the equipment cabinets is
 
controlled to avoid the possibility of
 
testing or bypassing more than one
 
bistable at any one time. Bistable
 
bypass is indicated by a light and by
 
key position at the location of the
 
bistables and by means of the plant
 
annunciation system in the main control
 
room.
 
Bypass of any portion of the aux-
 
iliary feedwater system or of the main
 
steam and feedwater isolation valves is
 
indicated in the main control room.
: 4. Where actuated equipment is      4. Actuated equipment is tested
 
not tested during reactor opera-    during reactor operation, except
 
tion, it should be shown that,      for the equipment addressed in Section
 
7.1.2.5.2.
 
a)  There is no practicable
 
system design that would permit
 
operation of the actuated equip-
 
ment without adversely affecting
 
the safety or operability of the
 
plant;
 
b)  The probability that the
 
protection system will fail to
 
initiate the operation of the
 
actuated equipment is, and can be
 
maintained, acceptably low
 
Rev. 0 WOLF CREEK TABLE 7.1-4 CONFORMANCE TO REGULATORY GUIDE 1.53, REV 0, 6/73, "APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS" This table demonstrates the conformance of the design of BOP equipment to Regulatory Guide 1.53.
Regulatory Guide 1.53 Position WCG S Position    The guidance in trial-use
 
IEEE S td 379-1972 for applying the single-failure criterion to
 
the design and analysis of nuclear
 
power plant protection systems is
 
generally acceptable and provides
 
an adequate interim basis for com-
 
plying with S ection 4.2 of IEEE S td 279-1971, subject to the fol-lowing: 1. Because of the trial-use        1. Complies with IEEE 379-1972 status of IEEE S td 379-1972, it      in its entirety.
may be necessary in specific
 
instances to depart from one
 
or more of its provisions.
: 2. S ection 5.2 of IEEE S td          2. Complies. The testability of 379-1972 should be supplemented      the systems is designed to posi-as follows:                          tively identify failures "The detectability of a single failure is predicated on the
 
assumption that the test results
 
in the presence of a failure
 
are different from the results
 
that would be obtained if no
 
failure is present. Thus, incon-
 
clusive testing procedures such
 
as "continuity checks" of relay
 
circuit coils in lieu of relay
 
operations should not be con-
 
sidered as adequate bases to
 
classify as detectable all po-
 
tential failures which could
 
negate the functional capability
 
of the tested device."            Rev. 1 WOLF CREEK TABLE 7.1-4 (S heet 2)    Regulatory Guide 1.53 Position WCG S Position 3. S ection 6.2 of IEEE S td          3. Complies.
S witches are either 379-1972 should be supplemented      for single trains, or there are
 
as follows:                          two switches, either of which can
 
actuate both trains. For the latter "Where a single mode switch          type switch, proper separation is
 
supplies signals to redundant        included in the design.
 
channels, it should be con-
 
sidered that the single-failure
 
criterion will not be satisfied
 
if either (a) individual switch
 
sections supply signals to
 
redundant channels, or (b)
 
redundant circuits controlled
 
by the switch are separated by
 
less than six inches without
 
suitable barriers." 4. S ection 6.3 and 6.4 of IEEE      4. Complies. The FMEA is S td 379-1972 should be inter-        performed on the basis of a preted as not permitting            system defined as starting separate failure mode analyses      with the sensors and con-
 
for the protection system logic      tinuing through the actuated
 
and the actuator system. The        devices.
 
collective protection system
 
logic-actuator system.should
 
be analyzed for single-failure
 
modes which, though not negating
 
the functional capability of
 
either portion, act to disable
 
the complete protective function.
 
[An example of such a potential
 
failure mode is a misapplication
 
of Regulatory Guide 1.6 (S afety Guide 6) wherein a single d-c
 
source supplies control power
 
for one channel of protection
 
system logic and for the
 
redundant actuator circuit.]            Rev. 1 WOLF CREEK TABLE 7.1-5 CONFORMANCE TO REGULATORY GUIDE 1.62, REV 0, 10/73, "MANUAL INITIATION OF PROTECTIVE ACTIONS" This table demonstrates the conformance of the design of BOP equipment to Regulatory Guide 1.62.
Regulatory Guide 1.62 Position WCG S Position 1. Means should be provided for    1. Complies. Manual switches are
 
manual initiation of each pro-      provided for system actuation.
 
tective action (e.g., reactor
 
trip, containment isolation) at
 
the system level, regardless of
 
whether means are also provided
 
to initiate the protective action
 
at the component or channel level (e.g., individual control rod, individual isolation valve).
: 2. Manual initiation of a pro-      2. Complies. Manual actuation of tective action at the system        the protective systems has the same
 
level should perform all actions    result as automatic actuation.
 
performed by automatic initiation
 
such as starting auxiliary or
 
supporting systems, sending sig-
 
nals to appropriate valve-actuating
 
mechanisms to assure correct
 
valve position, and providing
 
the required action-sequencing
 
functions and interlocks.
: 3. The switches for manual          3. Complies. Manual switches for initiation of protective actions    protective systems are provided
 
at the system level should be        in the control room.
 
located in the control room
 
and be easily accessible to the
 
operator so that action can be
 
taken in an expeditious manner.
: 4. The amount of equipment com-    4. Complies. The manual and auto-mon to both manual and automatic    matic initiation of protective
 
initiation should be kept to a      functions are separate.
 
minimum. It is preferable to
 
limit such common equipment to
 
the final actuation devices and
 
the actuated equipment. However, action-sequencing functions and
 
interlocks (of Position 2) asso-
 
ciated with the final actuation            Rev. 1 WOLF CREEK TABLE 7.1-5 (S heet 2)    Regulatory Guide 1.62 Position WCG S Position devices and actuated equipment may
 
be common if individual manual
 
initiation at the component or
 
channel level is provided in the
 
control room. No single failure
 
within the manual, automatic, or
 
common portions of the protection
 
system should prevent initiation
 
of protective action by manual or
 
automatic means.
: 5. Manual initiation of pro-        5. Complies. In some cases, one tective actions should depend        switch will actuate both trains.
 
on the operation of a minimum        In all other cases, one switch
 
equipment, consistent with          actuates one train.
 
1, 2, 3, and 4 above.
: 6. Manual initiation of pro-        6. Complies. Once manual initiation tective action at the system        occurs, the protective action goes
 
level should be so designed          to completion.
 
that once initiated, it will
 
go to completion as required
 
in S ection 4.16 of IEEE S td 279-1971.            Rev. 1 WOLF CREEK TABLE 7.1-6 CONFORMANCE TO REGULATORY GUIDE 1.105, REV 1, 11/76, "INSTRUMENT SETPOINTS"  This table demonstrates the conformance of the design of BOP equipment to
 
Regulatory Guide 1.105. The NSSS response to this Regulatory Guide is given
 
in Appendix 3A.
 
Note that the implementation date for this Regulatory Guide (plants with
 
construction permits docketed after December 15, 1976), is after the
 
construction permit docketing date for WCGS (1974).
 
Regulatory Guide 1.105 Position  WCGS Position
 
The following are applicable
 
to instruments in systems impor-
 
tant to safety.
: 1. The setpoints should be 1. Complies. The setpoints have established with sufficient been established with sufficient margin between the technical margin to allow for instrument specification limits for the inaccuracies, calibration uncertain-process variable and the ties, and potential instrument nominal trip setpoints to drift between calibration checks.
 
allow for (a) the inaccuracy
 
of the instrument, (b) uncer-
 
tainties in the calibration, and (c) the instrument drift
 
that could occur during the
 
interval between calibrations.
: 2. All setpoints should be 2. Complies. The instrument established in that portion spans have been established to of the instrument span which ensure that the accuracy at ensures that the accuracy, as setpoint is sufficient.
 
required by regulatory posi-
 
tion 4 below, is maintained.
 
Instruments should be cali-
 
brated so as to ensure the
 
required accuracy at the
 
setpoint.
: 3. The range selected for the 3. Complies. The instrument instrumentation should encompass ranges have been established to the expected operating range of ensure that saturation does the process variable being not negate the required in-monitored to the extent that strument operation.
 
saturation does not negate the
 
required action of the instru-
 
ment.
 
Rev. 1 WOLF CREEK TABLE 7.1-6 (Sheet 2)
Regulatory Guide 1.105 Position                        WCGS Position
: 4. The accuracy of all              4. Complies. The instrument setpoints should be equal to        accuracies are adequate to ensure or better than the accuracy          actuation within the limits assumed assumed in the safety analysis,      in the safety analyses, and are not which considers the ambient          unacceptably degraded by annealing, temperature changes, vibration,      stress relieving or work hardening
 
and other environmental condi-      under design conditions. Compliance
 
tions. The instruments should      with Regulatory Guide 1.89 is discussed
 
not anneal, stress relieve, or      in Sections 3.11(B), 8.1 and work harden under design condi-      Appendix 3A.
 
tions to the extent that they
 
will not maintain the required      Note that the accident analyses
 
accuracy. Design verification      generally assume absolute values
 
of these instruments should be      for the various parameters, demonstrated as part of the          rather than assuming nominal
 
instrument qualification program    values with specified accuracies.
 
recommended in Regulatory Guide
 
1.89, "Qualification of Class lE Equipment for Nuclear Power Plants." 5. Instruments should have a        5. Complies. The bistable set-securing device on the setpoint      point adjustments are not accessible
 
adjustment mechanism unless it      when the cabinet doors are closed.
 
can be demonstrated by analysis      Locks are provided on the cabinet
 
or test that such devices will      doors, and access to the cabinet
 
not aid in maintaining the re-      area is under administrative con-quired setpoint accuracy and        trol. There is sufficient fric-minimizing setpoint changes.        tion in the setpoint adjustment The securing device should be        mechanism to ensure that the designed so that it can be          adjustment does not slip during secured or released without          normal operation or seismic altering the setpoint and            excitation.
should be under administrative control.6. The assumptions used in          6. The derivation of the set-selecting the setpoint values        point values from the limiting
 
in regulatory position 1 and the    safety system settings has been
 
minimum margin with respect to      thoroughly documented.
 
the limiting safety system set-
 
tings, setpoint rate of devia-
 
tion (drift rate), and the
 
relationship of drift rate to
 
testing interval (if any) should
 
be documented.
Rev. 13 WOLF CREEK TABLE 7.1-7 CONFORMANCE TO REGULATORY GUIDE 1.118, REV 2, 6/78 "PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS" This table demonstrates the conformance of the design of BOP equipment to Regulatory Guide 1.118. The N SSS response to this Regulatory Guide is given in Appendix 3A.
Regulatory Guide 1.118 Position WCG S Position    The requirements and recom-
 
mendations contained in IEEE S td 338-1977 are considered acceptable
 
methods for the periodic testing
 
of electric power and protection
 
systems, subject to the following:
: 1. The term "safety system" is      1. Complies. The systems listed in used in IEEE S td 338-1977 in        position 1 are considered and many places. For the purposes      designed as safety-related systems.
 
of this guide, "safety system" should be understood to mean, collectively, the electric, instrumentation, and controls
 
portions of the protection
 
system; the protective action
 
system; and auxiliary or sup-
 
porting features that must be
 
operable for the protection sys-
 
tem and protective action system
 
to perform their safety-related
 
functions.
: 2. Item (6) of S ection 5 of IEEE    2. Complies. Protection systems S td 338-1977 lists alternative      are tested during operation under means of including the actuated      conditions specified in Item equipment in the periodic            (6)(a) or (b). Full tests that
 
testing of protection system        would interfere with operation
 
equipment. The method in which      are performed with the plant
 
actuated equipment is simulta-      shutdown. Partial closure of
 
neously tested with the as-          main steam and feedwater isolation
 
sociated protection system          valves is performed during reactor
 
equipment is preferred by            operation, as described in
 
the NRC staff; however,              Table 7.1-3 on conformance
 
overlap testing is accept-          to Regulatory Guide 1.22.
S ee able. In addition to the            additional discussions in S ection requirements of item (2) in          7.1.2.5.2 and Table 7.1-3.
 
S ection 6.1, complete systems tests should be performed at
 
suitable intervals.            Rev. 1 WOLF CREEK TABLE 7.1-7 (S heet 2)    Regulatory Guide 1.118 Position WCG S Position 3. Item (11) of S ection 5 of      3. Complies. The testing is performed IEEE S td 338-1977 should be          by perturbing the monitored variable supplemented by the following:      wherever practical. Where perturbing
 
the monitored variable is not prac-
 
    "Where perturbing the            tical, substitute inputs are intro-
 
monitored variable is not            duced into the sensor.
 
practical, the proposed
 
substitute tests shall be
 
shown to be adequate." 4. S ection 5 of IEEE S td            4. Complies. Bypass of a system 338-1977 should be supplemented      does not bypass any other system by the following:                    on the same train or on redundant
 
trains. Redundant components are
 
    "(13) Means shall be in-        tested independently.
 
cluded in the design to prevent
 
the expansion of any bypass
 
condition to redundant channels
 
or load groups during testing
 
operations. Where simulated
 
signals are used to test pro-
 
tective channels or load groups
 
or in other cases where such
 
equipment can be effectively
 
bypassed during a test, care
 
shall be exercised to ensure
 
that more channels are not
 
bypassed than are necessary
 
to perform the test. The
 
remaining channels (those
 
not bypassed) shall provide
 
that safety function con-
 
sistent with the provisions
 
of item (4) in S ection 5 of IEEE S td 338-1977."    "(14) Where redundant
 
components are used within
 
a single channel or load
 
group, the design shall
 
permit each to be tested
 
independently." 5. S ection 6.3.4 of IEEE            5. Not applicable to BOP S td 338-1977 should be              equipment.
supplemented by the following:            Rev. 1 WOLF CREEK TABLE 7.1-7 (S heet 3)    Regulatory Guide 1.118 Position WCG S Position    "For neutron detectors (1)
 
tests of detector-cable assemblies
 
for increased capacitance, (2)
 
monitoring of noise characteristics
 
of neutron detector signals, or (3)
 
some other test that does not re-
 
quire removal of detectors from their
 
installed location should be used to
 
confirm neutron detector response
 
time characteristics to avoid undue
 
radiation exposure of plant personnel
 
unless such tests are not capable of
 
detecting response time changes beyond
 
acceptable limits." 6. S ection 6.4(5) of IEEE S td 338-1977 should be supplemented by the following:
    "... makeshift test setups except as follows:
    "a. Temporary jumper wires      6.a. Complies. Facilities may be used with portable test      for connection of test equip-
 
equipment where the safety          ment include screw terminal
 
system equipment to be tested        blocks at the back of the
 
is provided with facilities          cabinet.
 
specifically designed for con-
 
nection of this test equipment.
 
These facilities shall be
 
considered part of the safety
 
system and shall meet all the
 
requirements of this standard, whether the portable test
 
equipment is disconnected or
 
remains connected to these
 
facilities.
    "b. Removal of fuses or        6.b. Complies. Removal of opening a breaker is permitted      fuses or opening of input cir-
 
only if such action causes (1)      cuit breakers is done only if
 
the trip of the associated pro-      it causes the trip of the asso-
 
tection system channel, or (2)      ciated channel or actuation of
 
the actuation (startup and          the associated load group.
 
operation) of the associated
 
Class 1E load group."            Rev. 1 WOLF CREEK TABLE 7.1-7 (S heet 4)    Regulatory Guide 1.118 Position WCG S Position 7. In addition to items (1)        7. Complies. Changes in failure
 
through (7) of S ection 6.5.1        rates are considered in testing of IEEE S td 338-1977, the            intervals.
ability to detect signifi-
 
cant changes in failure rates
 
should be considered in the
 
selection of initial test
 
intervals.
: 8. The following provisions of IEEE S td 338-1977 have been added in the 1977 version of this stan-
 
dard. These provisions will be
 
considered by the NRC staff and
 
endorsed or supplemented in a
 
future revision of this regulatory
 
guide. a. S ection 4, eighth para-      8.a. Complies with IEEE 338-1977.
graph, now excludes the process      Process to sensor coupling and to sensor coupling and the          actuated equipment to process
 
actuated equipment to process        coupling are not considered in
 
coupling from response time          response times.
 
testing required by the
 
standard. b. S ection 5, first para-      8.b. Complies with IEEE 338-1977.
graph, items (2) and (3) now        Tripping or bypass of channels allow tripping of the channel        being tested is done only for
 
being tested, or bypass of          the short period of the test.
 
the equipment consistent with
 
availability requirements, during test of redundant
 
channels or load groups.
: c. S ection 6.6.2, item (8)      8.c. Complies with IEEE 338-1977.
now only requires listing of        The written procedures do provide anticipated responses in test        the anticipated response, when
 
procedures "when required as        required, as a precautionary
 
a precautionary measure."            measure immediately before the step
 
which will produce the response. The
 
means by which the response is to be
 
observed is included in the acceptance
 
criteria.            Rev. 1 ANALOG PROTECTION SYSTEM NUCLEAR IMSTRUM[NlATIOM SYSTEM OR FIELD CONTACTS CONTROL BOARD SWITCHES TRAIN B CONTROL BOARD SWITCHES TRAIN l PROTECTION SYSTEM TRA IM B WOLF CREEK MASTER AND SLAVE RELAYS ACTUATE TRAIN 8 SAFEGUARDS C!J.IPUTER DEMU> ISOLATION COMPUTER MONITOR I KG "OR" CABLE ISOLATION CONTROL BOARD MONITORING CONTROL BOARD DEMUX CABINET 1----------+--
ACTUllE TRAIN A SAfEGUARDS iO ROD DRIVE ROD CON IROL SYST(M BYPASS BkR B TRIP BKR B "' ._----liuv ( ROD CONTROL M*G SiTS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.1-1 PROTECTION SYSTEM BLOCK DIAGRAM WOLF CREEK 7.2  REACTOR TRIP SYSTEM 7.
 
==2.1  DESCRIPTION==
 
7.2.1.1  System Description The reactor trip system (RTS) automatically keeps the reactor operating within a safe region by shutting down the reactor whenever the limits of the region are approached. The safe operating region is defined by several considerations, such as mechanical/hydraulic limitations on equipment and heat transfer phenomena. Therefore, the reactor trip system keeps surveillance on process variables which are directly related to equipment mechanical
 
limitations, such as pressure and pressurizer water level (to prevent water discharge through safety valves and uncovering heaters), and also on variables
 
which directly affect the heat transfer capability of the reactor (e.g., flow and reactor coolant temperatures). Still other parameters utilized in the
 
reactor trip system are calculated from various process variables. Whenever a direct process or calculated variable exceeds a setpoint, the reactor is shut
 
down in order to protect against either damage to fuel cladding or loss of
 
system integrity, which could lead to the release of radioactive fission
 
products into the containment.
The following systems make up the reactor trip system (see Ref. 1, 2, and 3 for additional background information).
: a. Process instrumentation and control system
: b. Nuclear instrumentation system
: c. Solid state logic protection system
: d. Reactor trip switchgear
: e. Manual actuation circuit
 
The reactor trip system consists of sensors that monitor various plant parameters and are connected with analog circuitry, consisting of two to four
 
redundant channels, and digital circuitry, consisting of two redundant logic
 
trains, that receives inputs from the analog channels to complete the logic necessary to automatically open the reactor trip breakers.
Each of two logic trains, A and B, is capable of opening a separate and independent reactor trip breaker, RTA and RTB, respectively. The two trip
 
breakers in series connect three-phase ac power from the rod drive motor
 
generator sets to the rod drive 7.2-1                        Rev. 0 WOLF CREEK power cabinets, as shown in Figure 7.2-1 (Sheet 2). During plant power operation, a dc undervoltage coil on each reactor trip breaker holds a trip
 
plunger out against its spring, allowing the power to be available at the rod control power supply cabinets. For reactor trip, a loss of dc voltage to the undervoltage coil as well as energization of the shunt trip coil releases the
 
trip plunger and trips open the breaker. When either of the trip breakers opens, power is interrupted to the rod drive power supply, and the control rods fall, by gravity, into the core. The rods cannot be withdrawn until the trip
 
breakers are manually reset. The trip breakers cannot be reset until the abnormal condition which initiated the trip is corrected. Bypass breakers BYA
 
and BYB are provided to permit testing of the trip breakers, as discussed in Section 7.2.2.2.3.
An Auto Shunt Trip design modification has been implemented which monitors the Reactor Protection System outputs to the reactor trip breaker's undervoltage coils and provides trip signals to the shunt trip coils upon receipt of an
 
automatic trip signal to the undervoltage coils. This was accomplished by
 
providing a rotary type interposing relay between the trip breaker undervoltage
 
coil circuit and the shunt trip coil circuit. This Auto Shunt Trip Relay is
 
energized from the Reactor Protection System voltage which is provided to the
 
undervoltage coil. When the voltage is removed by an automatic reactor trip
 
signal, the Auto Shunt Trip Relay de-energizes, closing a contact to energize
 
the shunt trip coil. Thus, the breaker trip shaft is actuated by both the
 
undervoltage and shunt trip attachments. This design modification applies only
 
to the reactor trip breakers; the bypass breaker shunt trip coils will not
 
receive an automatic trip signal.
The added hardware consists of qualified shunt trip coils and panels which include the relays and test hardware. The shunt trip attachments and Auto
 
Shunt Trip Panels are qualified in accordance with IEEE standards 323-1974 and 344-1975. The panels are mounted at the reactor trip switchgear.
The Auto Shunt Trip Panels are provided with two pushbutton switches for use during periodic on-line testing to independently confirm the operability of the undervoltage and shunt trip attachments. The Auto Shunt Trip Block pushbutton
 
switch is used to prevent the shunt trip coil from energizing when the
 
undervoltage trip is being tested. The Auto Shunt Trip Test pushbutton switch
 
is  used to de-energize the Auto Shunt Trip Relay, energizing the shunt trip
 
coil while the undervoltage coil remains energized.
The Auto Shunt Trip Panels are also equipped with test jacks to facilitate breaker response time testing. These jacks are wired directly to an auxiliary
 
switch contact (closed when the breaker 7.2-2                          Rev. 0 WOLF CREEK is closed) to provide indication that the breaker has tripped. Another set of test jacks are connected across the Auto Shunt Trip Relay coil through
 
resistors to provide indication of initiation of a trip. The resistors are provided to assure that accidental shorts or grounds applied through the test points do not result in an inadvertent reactor trip or an overload on the
 
Reactor Protection System output.
7.2.1.1.1  Functional Performance Requirements
 
The reactor trip system automatically initiates reactor trip:
: a. Whenever necessary to prevent fuel damage for an anticipated operational transient (Condition II)
: b. To limit core damage for infrequent faults (Condition III)
: c. So that the energy generated in the core is compatible with the design provisions to protect the reactor coolant
 
pressure boundary for limiting fault conditions
 
(Condition IV).
The reactor trip system initiates a turbine trip signal whenever reactor trip is initiated to prevent the reactivity insertion that would otherwise result from excessive reactor system cooldown. This eliminates unnecessary actuation of the engineered safety feature actuation system.
The reactor trip system provides for manual initiation of reactor trip by operator action.
7.2.1.1.2  Reactor Trips The reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the reactor trip system reaches a preset level. To ensure a reliable system, high quality design, components, manufacturing, quality control, and testing are used. In addition to redundant channels and
 
trains, the design approach provides a reactor trip system that monitors
 
numerous system variables, therefore providing protection system functional
 
diversity. The extent of this diversity has been evaluated for a wide variety of postulated accidents.
Table 7.2-1 provides a list of reactor trips that are described below. Table 7.2-2 provides a listing of all protection system interlocks which are
 
designated P-(number).
: a. Nuclear overpower trips 7.2-3                          Rev. 0 WOLF CREEK The specific trip functions generated are as follows:
: 1. Power range high neutron flux trip
 
The power range high neutron flux trip circuit trips the reactor when two out of the four power range channels exceed the trip setpoint.
There are two bistables, each with its own trip setting used for a high- and a low-range trip
 
setting. The high trip setting provides protection
 
during normal power operation and is always active.
 
The low trip setting, which provides protection
 
during startup, can be manually bypassed when two out
 
of the four power range channels read above
 
approximately 10-percent power (P-10). Three out of
 
the four channels below 10 percent automatically
 
reinstate the trip function.
: 2. Intermediate range high neutron flux trip The intermediate range high neutron flux trip circuit trips the reactor when one out of the two
 
intermediate range channels exceeds the trip setpoint. This trip, which provides protection during reactor startup, can be manually blocked if
 
two out of the four power range channels are above P-
: 10. Three out of the four power range channels below this value automatically reinstate the intermediate range high neutron flux trip. The intermediate range
 
channels (including detectors) are separate from the
 
power range channels. The intermediate range channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing
 
during plant shutdown or prior to startup. This bypass action is annunciated on the control board.
: 3. Source range high neutron flux trip
 
The source range high neutron flux trip circuit trips the reactor when one out of the two source range
 
channels exceeds the trip setpoint. This trip, which
 
provides protection during reactor  startup and plant shutdown, can be manually bypassed when one out of the two intermediate range channels reads above the
 
P-6 setpoint value and is automatically reinstated
 
when both intermediate range channels decrease below
 
the P-6 setpoint value. This trip is also
 
automatically  bypassed by  two-out-of-four  logic
 
from the 7.2-4                          Rev. 1 WOLF CREEK power range protection interlock (P-10). This trip
 
function can also be reinstated below P-10 by an
 
administrative action requiring manual actuation of
 
two control board mounted switches. Each switch will reinstate the trip function in one out of the two
 
protection logic trains. The source range trip point
 
is set between the P-6 setpoint (source range cutoff
 
power level) and the maximum source range power
 
level. The channels can be individually bypassed at
 
the nuclear instrumentation racks to permit channel
 
testing during plant shutdown or prior to startup. 
 
This bypass action is annunciated on the control
 
board.
: 4. Power range high positive neutron flux rate trip
 
This circuit trips the reactor when a sudden abnormal
 
increase in nuclear power occurs in two out of the
 
four power range channels. This trip  provides DNB protection against certain rod ejection and rod withdrawal accidents (see Chapter 15.0).
: 5. Power range high negative neutron flux rate trip
 
This circuit trips the reactor when a sudden abnormal
 
decrease in nuclear power occurs in two out of the four power range channels. This trip provides
 
protection against two or more dropped rods and is
 
always active. Protection against one dropped rod is
 
not required to prevent occurrence of DNB per Section
 
15.4.3.
 
Figure 7.2-1 (Sheet 3) shows the logic for all of the
 
nuclear overpower and rate trips.
: b. Core thermal overpower trips The specific trip functions generated are as follows:
: 1. Overtemperature  T trip              This trip protects the core against low DNB and trips
 
the reactor on coincidence, as listed in Table 7.2-1, with one set of temperature measurements per loop.
 
The setpoint for this trip is continuously calculated 
 
by analog circuitry for each loop by solving the 
 
following equation: 
 
7.2-5                          Rev. 25 WOLF CREEK DT Setpoint = DT o K 1 - K 2
 
1 + t 1 s 1 + t 2 sT avg 1 + t 4 s T o avg + K 3(P-2235-f(Df))
Where: T o      =  programmed  T at rated thermal power T
avg      =  average reactor coolant temperature, F T
o        =  programmed Tavg at rated thermal avg        power P        =  pressurizer pressure, psig
 
K1        =  preset bias K2        =  preset gain which compensates for the effects of temperature on the DNB limits K3        =  preset gain which compensates for the effect of pressure on the DNB limits  1, 2    =  preset constants which compensate for piping and instrument time delay,                          seconds  4        =  preset time constant for signal conditioning s        =  laplace transform operator, seconds-1 f()    =  a function of the neutron flux difference between the upper and lower long ion chambers (refer to Figure 7.2-
: 2)
A separate long ion chamber unit supplies the flux signal for each overtemperature T trip channel.
Increases in  beyond a predefined deadband result in a decrease in trip  setpoint  (refer to  Figure 7.2-2).
The required one pressurizer pressure parameter per loop is obtained from separate sensors connected to
 
three pressure taps at the top of the pressurizer.
 
Four pressurizer  pressure  signals are obtained from 7.2-6                          Rev. 0 WOLF CREEK the three taps by connecting one of the taps to two pressure transmitters. Refer to Section 7.2.2.3.3 for an analysis of this
 
arrangement.
Figure 7.2-1 (Sheet 5) shows the logic for over-temperature T trip function.
: 2. Overpower T trip              This trip protects against excessive power (fuel rod rating protection) and trips the reactor on coincidence, as listed in Table 7.2-1, with one set of temperature measurements per loop. The setpoint for each channel is continuously calculated, using the following equation: T SETPOINT =T o [K 4 - K 5 (t 3 s 1 + t 3 s)
 
T avg 1 + t 4 s                          -K 6
 
T avg 1 + t4s - T o avg - f(Df)]
Where: T o    =  programmed T at rated thermal power f()  =  a function of the neutron flux difference between upper and lower long ion chamber
 
section K
4      =  a preset bias K
5    =  a constant which compensates for piping and instrument time delay
 
K 6      =  a constant which compensates for the change in density flow and heat capacity of the water with temperature T
o      =  programmed Tavg at rated thermal avg        power T
avg    =  average reactor coolant temperature, F  3      =  preset time constant, seconds  4      =  preset time constant for signal conditioning s      =  laplace transform operator, seconds-1 7.2-7                          Rev. 0 WOLF CREEK The source of temperature and flux information is identical to that of the overtemperature T trip,              and the resultant  T setpoint is compared to the sameT. Figure 7.2-1 (Sheet 5) shows the logic for this trip function.
: c. Reactor coolant system pressurizer pressure and water level trips The specific trip functions generated are as follows:
: 1. Pressurizer low pressure trip
 
The purpose of this trip is to protect against low pressure which could lead to DNB. The parameter being sensed is reactor coolant pressure as measured in the pressurizer. Above P-7, the reactor is
 
tripped when the pressurizer pressure measurements
 
(compensated for rate of change) fall below preset limits. This trip is blocked below P-7 to permit
 
startup. The trip logic and interlocks are given in
 
Table 7.2-1.
The trip logic is shown on Figure 7.2-1 (Sheet 6).
: 2. Pressurizer high pressure trip
 
The purpose of this trip is to protect the reactor coolant system against system overpressure.
The same sensors and transmitters used for the pressurizer low pressure trip are used for the high
 
pressure trip, except that separate bistables are
 
used for trip. These bistables trip when
 
uncompensated pressurizer pressure signals exceed preset limits on coincidence as listed in Table  7.2-
: 1. There are no interlocks or permissives associated with this trip function.
The logic for this trip is shown on Figure 7.2-1 (Sheet 6).
: 3. Pressurizer high water level trip
 
This trip is provided as a backup to the high pressurizer pressure trip and serves to prevent water relief through the pressurizer safety valves. This
 
trip is blocked below P-7 to permit startup. The coincidence logic and interlocks of pressurizer high water level signals are given in Table 7.2-1.
7.2-8                          Rev. 1 WOLF CREEK The trip logic for this function is shown on Figure
 
7.2-1 (Sheet 6).
: d. Reactor coolant system low flow trips
 
These trips protect the core from DNB in the event of a
 
loss-of-coolant flow situation. Figure 7.2-1 (Sheet 5)
 
shows the logic for these trips. The means of sensing
 
the loss-of-coolant flow are as follows:
: 1. Low reactor coolant flow
 
The parameter sensed is reactor coolant flow. Four
 
elbow taps in each coolant loop are used as a flow device that indicates the status of reactor coolant flow. The basic function of this device is to
 
provide information as to whether or not a reduction
 
in flow has occurred. An output signal from two out
 
of the three bistables in a loop would indicate a low
 
flow in that loop.
 
The coincidence logic and interlocks are given in
 
Table 7.2-1.
: 2. Reactor coolant pump undervoltage trip
 
This trip protects against low flow which can result
 
from loss of voltage to the reactor coolant pump
 
motors (e.g., from plant blackout or reactor coolant pump breakers opening).
 
There is one undervoltage sensing relay connected to
 
each pump at the motor side of each reactor coolant
 
pump breaker. These relays provide an output signal
 
when the pump voltage goes below approximately 10,578V.
Signals from these relays are time delayed to prevent spurious trips caused by short-term voltage perturbations.
: 3. Reactor coolant pump underfrequency trip
 
This trip protects against low flow resulting from
 
pump underfrequency, for example a major power grid
 
frequency disturbance. The function of this trip is to trip the reactor for an underfrequency condition
 
greater than approximately 2.4 Hz/second. The 
 
setpoint of the underfrequency relays is adjustable 
 
between 40 and 70 Hz. 
 
7.2-9                          Rev. 25 WOLF CREEK There is one underfrequency sensing relay for each
 
reactor coolant pump motor. Signals from one or both
 
relays from both busses of the pump motors (time
 
delayed up to approximately 0.2 seconds to prevent spurious trips caused by short-term frequency
 
perturbations) will trip the reactor if the power
 
level is above P-7. The coincidence logic and
 
interlocks are given in Table 7.2-1.
: e. Steam generator low-low water level trip
 
The specific trip function generated is low-low steam
 
generator water level trip.
 
This trip protects the reactor from loss of heat sink.
 
This trip is actuated on two out of four low-low water
 
level signals occurring in any steam generator.
 
The logic is shown on Figure 7.2-1 (Sheet 7).
: f. Reactor trip on a turbine trip (anticipatory)
 
The reactor trip on a turbine trip is actuated by two-
 
out-of-three logic from emergency trip fluid pressure
 
signals or by all closed signals from the turbine steam
 
stop valves. A turbine trip causes a direct reactor trip
 
above P-9. The reactor trip on turbine trip provides
 
additional protection and conservatism beyond that
 
required for the health and safety of the public. This
 
trip is included as part of good engineering practice and
 
prudent design.
 
The turbine provides anticipatory trips to the reactor
 
protection system from contacts which change position when the turbine stop valves close or when the turbine emergency trip fluid pressure goes below its setpoint.
 
Components specified for use as sensors for input signals
 
to the reactor protection system for "emergency trip oil
 
pressure low" and "turbine stop valves close" conform to
 
the requirements of IEEE 279-1971 and as environmentally
 
qualified. However, seismic criteria are not included in
 
qualification regarding mounting and location for that
 
portion of the trip system located within nonseismic
 
Category I structures.
 
Evaluations indicate that the functional performance of
 
the protection system would not be degraded by credible
 
electrical faults such as opens and shorts in the
 
circuits associated with reactor trip or the generation
 
of
 
7.2-10    Rev. 25 WOLF CREEK the P-7 interlock. The solid state protection system cabinets are provided with the fuse protection for the turbine stop valve reactor trip cabling in the Turbine Building to preclude degradation of required solid state protection system functions. Faults on the turbine building cables going to the control valve hydraulic oil pressure transmitters will not degrade the protection system as they are isolation from the protection system by the process instrumentation (Foxboro) cabinets in the main control room.
Loss of signal caused by open circuits would produce either a partial or full reactor trip. Faults on the first stage
 
turbine pressure circuits would result in upscale, conservative output for open circuits and a sustained
 
current, limited by circuit resistance, for short circuits.
 
Multiple failures imposed on redundant circuits could
 
potentially disable the P-13 interlock. In this event, the
 
nuclear instrumentation power range signals would provide the
 
P-7 safety interlock. Refer to functional diagram, Sheet 4 of Figure 7.2-1. The sensors for the  P-13 interlock are
 
seismically qualified.
Evaluations provided in Section 7.6.1 for the trip fluid pressure transmitter loops indicate that credible electrical
 
faults would not degrade the functional performance of the
 
safety-related BOP instrumentation.
In addition, the following measures were taken to ensure the integrity of the cabling to the reactor protection system (RPS):        1. Inputs from the turbine steam stop valves originate
 
from four separate limit switches (one per valve),
each of which is dedicated to providing an input to one channel of the RPS. Cables carrying these
 
signals are routed in individual conduits. The four
 
circuits are separated from one another, from non-
 
Class 1E circuits, and identified according to the
 
criteria imposed on Class 1E circuits from their
 
source up to their terminations with the RPS
 
cabinets.
: 2. Inputs from the emergency trip oil pressure and P-13 interlock instrumentation are routed in a similar
 
manner as are the turbine stop valve inputs.
The logic for this trip is shown on Figure 7.2-1 (Sheet 16).
: g. Safety injection signal actuation trip
 
A reactor trip occurs when the safety injection system is actuated. The means of actuating the safety injection
 
system are described in Section 7.3. This trip protects
 
the core following a loss of reactor coolant or a steam
 
line rupture.      7.2-11    Rev. 9 WOLF CREEK Figure 7.2-1 (Sheet 8) shows the logic for this trip.
: h. Manual trip
 
The manual trip consists of two switches with two outputs on each switch. One output is used to actuate the train A reactor trip breaker; the other output actuates the train B reactor trip breaker. Operating a manual trip
 
switch removes the voltage from the undervoltage trip coil and energizes the shunt trip coil of each breaker.
There are no interlocks which can block this trip.
Figure 7.2-1 (Sheet 3) shows the manual trip logic. The design conforms to Regulatory Guide 1.62, as shown in Figure 7.2-3.
7.2.1.1.3  Reactor Trip System Interlocks See Table 7.2-2 for the list of protection system interlocks.
: a. Power escalation permissives The overpower protection provided by the out-of-core nuclear instrumentation consists of three discrete but overlapping, ranges. Continuation of startup operation or power increase requires a permissive signal from the
 
higher range instrumentation channels before the lower
 
range level trips can be manually blocked by the
 
operator.
A one out of two intermediate range permissive signal (P-6) is required prior to source range trip blocking and
 
detector high voltage cutoff. Source range trips are
 
automatically reactivated and high voltage restored when
 
both intermediate range channels are below the permissive
 
(P-6) setpoint. There are two manual reset switches for
 
administratively reactivating the source range level trip
 
and detector high voltage when between the permissive P-6
 
and P-10 setpoints, if required. Source range level trip block and high voltage cutoff are always maintained when above the permissive P-10 setpoint.
The intermediate range level trip and power range (low setpoint) trip can only be blocked after satisfactory operation and permissive information are obtained from two out of four power range channels. Four individual blocking switches are provided so that the low range
 
power range trip and intermediate  range trip can be      7.2-12    Rev. 0 WOLF CREEK independently blocked (one switch for each train). These trips are automatically reactivated when any three out of
 
the four power range channels are below the permissive (P-10) setpoint, thus ensuring automatic activation to more restrictive trip protection. The development of
 
permissives P-6 and P-10 is shown on Figure 7.2-1 (Sheet 4). All of the permissives are digital; they are derived from analog signals in the nuclear power range and inter
 
mediate range channels.
: b. Blocks of reactor trips at low power
 
Interlock P-7 blocks a reactor trip at low power (below approximately 10 percent of full power) on a low reactor
 
coolant flow in more than one loop, reactor coolant pump
 
undervoltage, reactor coolant pump underfrequency, pressurizer low pressure or pressurizer high water
 
level. See Figure 7.2-1 (Sheets 5 and 6) for permissive
 
applications. The low power signal is derived from three
 
out of four power range neutron flux signals below the
 
setpoint in coincidence with two out of two turbine
 
impulse chamber pressure signals below the setpoint (low
 
plant load). See Figure 7.2-1 (Sheets 4 and 16) for the
 
derivation of P-7.
The P-8 interlock blocks a reactor trip when the plant is below approximately 50 percent of full power, on a low
 
reactor coolant flow in any one loop. The block action
 
(absence of the P-8 interlock signal) occurs when three
 
out of four neutron flux power range signals are below the setpoint. Thus, below the P-8 setpoint, the reactor has the capability to operate with one inactive loop and trip will not occur until two loops are indicating low
 
flow. See Figure 7.2-1 (Sheet 4) for derivation of P-8
 
and Sheet 5 for applicable logic.
Interlock P-9 blocks a reactor trip following a turbine trip below 50 percent power. See Figure 7.2-1 (Sheet 16)
 
for the implementation of the P-9 interlock and Sheet 4
 
for the derivation of P-9.
7.2.1.1.4  Coolant Temperature Sensor Arrangement
 
One hot leg and one cold leg temperature reading are provided from each coolant loop to use for protection. Narrow range thermowell-mounted Resistance Temperature Detectors (RTDs) are provided for each coolant loop. In the hot legs, sampling scoops are used because the flow is stratified. That is, the fluid temperature is not uniform over a cross section of the hot leg. One dual element RTD is mounted in a thermowell in each of the three sampling scoops associated with each hot leg. The scoops extend into the flow stream at locations 120° apart in the cross sectional plane.      7.2-13    Rev. 6 WOLF CREEK Each scoop has five orifices which sample the hot leg flow along the leading edge of the scoop. Outlet ports are provided in the scoops to direct the sampled fluid past the sensing element of the RTDs. One of each of the RTD's dual elements is used while the other is an installed spare. Three readings from each hot leg are averaged to provide a hot leg reading for that loop.
One dual element RTD is mounted in a thermowell associated with each cold leg.
No flow sampling is needed because coolant flow is well mixed by the reactor coolant pumps. As is the case with the hot leg, one element is used while the other is an installed spare.
Certain control signals are derived from individual protection channels through isolation cards. The isolation cards are classified as a part of the protection system. The rod control system uses the auctioneered (high) value of four isolated T-AVG signals.
The RTDs are a fast response design which conform to the applicable IEEE standards and 10 CFR 50.49 requirements.
7.2.1.1.5  Pressurizer Water Level Reference Leg Arrangement The design of the pressurizer water level instrumentation employs the usual tank level arrangement, using differential pressure between an upper and a
 
lower tap on a column of water. A reference leg connected to the upper tap is
 
kept full of water by condensation of steam at the top of the leg.
7.2.1.1.6  Analog System The analog system consists of two instrumentation systems - the process instrumentation system and the nuclear instrumentation system.      7.2-14    Rev. 6 WOLF CREEK Process instrumentation includes those devices (and their interconnection into systems) which measure temperature, pressure, fluid flow, fluid level as in
 
tanks or vessels, and occasionally physiochemical parameters, such as fluid conductivity or chemical concentration. Process instrumentation specifically excludes nuclear and radiation measurements. The process instrumentation
 
includes the process measuring devices, power supplies, indicators, recorders, alarm actuating devices, controllers, signal conditioning devices, etc., which are necessary for day-to-day operation of the NSSS, as well as for monitoring
 
the plant and providing initiation of plant protective functions.
The primary function of nuclear instrumentation is to protect the reactor by monitoring the neutron flux and generating appropriate trips and alarms for various phases of reactor operating and shutdown conditions. The instrumentation also provides a secondary control function and indicates
 
reactor status during startup and power operation. The nuclear instrumentation
 
system uses information from three separate types of instrumentation channels
 
to provide three discrete protection levels. Each range of instrumentation (source, intermediate, and power) provides the necessary overpower reactor trip
 
protection required during operation in that range. The overlap of instrument
 
ranges provides reliable continuous protection beginning with source level
 
through the intermediate and low power level. As the reactor power increases, the overpower protection level is increased by administrative procedures after
 
satisfactory higher range instrumentation operation is obtained. Automatic
 
reset to more restrictive trip protection is provided when reducing power.
Various types of neutron detectors, with appropriate solid state electronic circuitry, are used to monitor the leakage neutron flux from subcritical
 
conditions to 120 percent of full power. The power range channels are capable
 
of recording overpower excursions up to 200 percent of full power. The neutron
 
flux covers a wide range between these extremes. Therefore, monitoring with several ranges of instrumentation is necessary.
The lowest range ("source" range) covers six decades of leakage neutron flux.
The lowest observed count rate depends on the strength of the neutron sources in the core and the core multiplication associated with the shutdown
 
reactivity. This is generally greater than two counts per second. The next
 
range ("intermediate" range) covers eight decades. Detectors and
 
instrumentation are chosen to provide overlap between the higher portion of the
 
source range and the lower portion of the intermediate range. The highest range of instrumentation ("power" range) covers approximately two decades of the total instrumentation range. This is a linear range that overlaps with the higher portion of the intermediate range.      7.2-15    Rev. 0 WOLF CREEK The system described above provides control room indication and recording of signals proportional to reactor neutron flux during core loading, shutdown, startup, and power operation, as well as during subsequent refueling. Startup rate indication for the source and intermediate range channels is provided at the control board. Reactor trip, control rod stop, and control and alarm
 
signals are transmitted to the reactor control and protection system for automatic plant control. Equipment failures and test status information are annunciated in the control room.
See References 1 and 2 for additional background information on the process and nuclear instrumentation.
7.2.1.1.7  Solid State Logic Protection System
 
The solid state logic protection system takes binary inputs (voltage/no voltage) from the process and nuclear instrument channels corresponding to
 
conditions (normal/abnormal) of plant parameters. The system combines these
 
signals in the required logic combination and generates a trip signal (no
 
voltage) to the undervoltage coils of the reactor trip circuit breakers when
 
the necessary combination of signals occur. This trip signal also deenergizes
 
the auto shunt trip relay which, in turn, closes a contact that energizes the
 
shunt trip coil. The system also provides annunciator, status light, and
 
computer input signals which indicate the condition of bistable input signals, partial trip and full trip functions, and the status of the various blocking, permissive, and actuation functions. In addition, the system includes means
 
for semiautomatic testing of the logic circuits. See Reference 3 for
 
additional background information.
7.2.1.1.8  Isolation Amplifiers
 
In certain applications, control signals are derived from individual protection channels through isolation amplifiers contained in the protection channel, as permitted by IEEE Standard 279-1971.
In all of these cases, analog signals derived from protection channels for nonprotective functions are obtained through isolation amplifiers located in
 
the analog protection racks. By definition, nonprotective functions include
 
those signals used for control, remote process indication, and computer
 
monitoring. Refer to Section 7.1.2.2.1 for a discussion of electrical
 
separation of control and protection functions.
7.2.1.1.9  Energy Supply and Environmental Variations
 
The energy supply for the reactor trip system, including the voltage and frequency variations, is described in Chapter 8.0. The environmental variations, throughout which the system will perform, are given in Section 3.11(N) and Chapter 8.0.      7.2-16    Rev. 13 WOLF CREEK 7.2.1.1.10  Setpoints The setpoints that require trip action are given in the Technical Specifications. A detailed discussion on setpoints is found in Section
 
7.3.8.1.2.7.
7.2.1.1.11  Seismic Design The seismic design considerations for the reactor trip system are given in Section 3.10(N). This design meets the requirements of GDC-2 (refer to Section
 
3.1).7.2.1.2  Design Bases Information The information given below presents the design bases information requested by Section 3 of IEEE Standard 279-1971. Functional diagrams are presented in
 
Figure 7.2-1.
7.2.1.2.1  Generating Station Conditions
 
The following are the generating station conditions requiring reactor trip.
: a. DNBR approaching thermal design limit DNBR (see Section 4.4.1.1).
: b. Linear power density (kilowatts per foot) approaching rated value for Condition II events (see Chapter 4.0 for fuel design limits).
: c. Reactor coolant system overpressure creating stresses approaching the limits specified in Chapter 5.0.
7.2.1.2.2  Generating Station Variables The following are the variables required to be monitored in order to provide reactor trips (see Table 7.2-1).
: a. Neutron flux
: b. Reactor coolant temperature
: c. Reactor coolant system pressure (pressurizer pressure)
: d. Pressurizer water level
: e. Reactor coolant flow      7.2-17    Rev. 13 WOLF CREEK
: f. Reactor coolant pump operational status (voltage and frequency)
: g. Steam generator water level
: h. Turbine-generator operational status (trip fluid pressure and stop valve position) 7.2.1.2.3  Spatially Dependent Variables
 
The only spatially dependent variable is the reactor coolant temperature. See Section 7.3.8.1.2 for a discussion of this spatial dependence.
7.2.1.2.4  Limits, Margins, and Setpoints The parameter values that will require reactor trip are given in Chapters 15.0 and the WCGS Technical Specifications. The accident analyses in Chapter 15.0
 
demonstrate that the setpoints used in the Technical Specifications are
 
conservative.
The setpoints for the various functions in the reactor trip system have been analytically determined so that the operational limits so prescribed will
 
prevent fuel rod clad damage and loss of integrity of the reactor coolant
 
system as a result of any Condition II event (anticipated malfunction). As
 
such, during any Condition II event, the reactor trip system limits the following parameters to:
: a. Minimum DNBR = thermal design limit DNBR (see Section 4.4.1.1)
: b. Maximum system pressure = 2,750 psia
: c. Fuel rod maximum linear power for determination of protection setpoints = 18.0 kW/ft The accident analyses described in Section 15.4 demonstrate that the functional requirements specified for the reactor trip system are adequate to meet the
 
above considerations, even assuming the conservative, adverse combinations of
 
instrument errors (refer to Table 15.0-4). A discussion of the safety limits associated with the reactor core and reactor coolant system, plus the limiting
 
safety system setpoints, are presented in the Technical Specifications.
7.2.1.2.5  Abnormal Events
 
The malfunctions, accidents, or other unusual events which could physically damage reactor trip system components or could cause environmental changes are
 
as follows:      7.2-18    Rev. 7 WOLF CREEK
: a. Earthquakes (see Chapters 2.0 and 3.0)
: b. Fire (see Section 9.5.1)
: c. Explosion - hydrogen buildup inside containment (see Section 6.2)
: d. Missiles (see Section 3.5)
: e. Flood (see Chapters 2.0 and 3.0)
: f. Wind and tornadoes (see Section 3.3)
 
The reactor trip system fulfills the requirements of IEEE Standard 279-1971 to provide automatic protection and to provide initiating signals to mitigate the
 
consequences of faulted conditions. The reactor trip system is protected from
 
fires, explosions, floods, winds, and tornadoes (see each item above).
7.2.1.2.6  Minimum Performance Requirements (See WCGS Technical Specifications for additional information on minimum
 
performance requirements.)
: a. Reactor trip system response times
 
Typical time delays in generating the reactor trip signal are tabulated in Table 7.2-3. See Section 7.1.2.6.2 for a discussion of periodic response time verification
 
capabilities.
: b. Reactor trip accuracies Typical reactor trip accuracies are tabulated in Table 7.2-3. An additional discussion on accuracy is found in Section 7.3.8.1.2.7.
: c. Protection system ranges Typical protection system ranges are tabulated in Table 7.2-3. Range selection for the instrumentation covers the expected range of the process variable being
 
monitored during power operation. Limiting setpoints are
 
at least 5 percent from the end of the instrument span.
7.2.1.3  Final Systems Drawings Functional block diagrams, electrical elementaries, and other drawings required to assure electrical separation and perform a safety review are provided in the
 
Safety-Related Drawing Package (refer to Section 1.7).      7.2-19    Rev. 17 WOLF CREEK 7.2.2  ANALYSES 7.2.2.1  Failure Mode and Effects Analyses An analysis of the reactor trip system has been performed. Results of this study and a fault tree analysis are presented in Reference 4.
7.2.2.2  Evaluation of Design Limits While most setpoints used in the reactor protection system are fixed, there are variable setpoints, most notably the overtemperature T and overpower T setpoints. All setpoints in the reactor trip system have been selected on the basis of engineering design or safety studies. The capability of the reactor trip system to prevent loss of integrity of the fuel cladding and/or reactor coolant system pressure boundary during Condition II and III transients is
 
demonstrated in Chapter 15.0. Accident analyses are carried out using those
 
setpoints determined from results of the engineering design studies. Setpoint
 
limits are presented in the Technical Specifications. A discussion of the intent for each of the various reactor trips and the accident analyses (where
 
appropriate) which utilize this trip are presented below. It should be noted
 
that the selected trip setpoints provide for a margin to allow for
 
uncertainties and instrument errors. The design meets the requirements of GDC-
 
10 and 20 (refer to Section 3.1).
7.2.2.2.1  Trip Setpoint Discussion The DNBR existing at any point in the core for a given core design can be determined as a function of the core inlet temperature, power output, operating
 
pressure, and flow. Core safety limits in terms of a DNBR equal to the thermal
 
design limit DNBR (Refer to Section 4.4.1.1) for the hot channel can be developed as a function of core T, T avg and pressure for a specified flow, as illustrated by the solid lines in Figure 15.0-1. Also shown as solid lines in Figure 15.0-1 are the loci of conditions equivalent to 118 percent of power as a function of T and Tavg representing the overpower (kW/ft) limit on the fuel.
The dashed lines indicated the maximum permissible setpoint (T) as a function of Tavg and pressure for the overtemperature and overpower reactor trip.
Actual setpoint constants in the equation representing the dashed lines are as
 
given in the Technical Specifications. These values are conservative to allow
 
for instrument errors. The design meets the requirements of GDC-10, 15, 20, and 29 (refer to Section 3.1).
DNBR is not a directly measurable quantity; however, the process variables that determine DNBR are sensed and evaluated. Small isolated changes in various
 
process variables may not individually result in violation of a core safety limit, whereas the combined variations, over sufficient time, may cause the
 
overpower or      7.2-20    Rev. 12 WOLF CREEK overtemperature safety limit to be exceeded. The reactor trip system provides reactor trips associated with individual process variables in addition to the
 
overpower/overtemperature safety limit trips. Process variable trips prevent reactor operation whenever a change in the monitored value is such that a core or system safety limit is in danger of being exceeded should operation
 
continue. Basically, the high pressure, low pressure, and overpower/overtemperatureT trips provide sufficient protection for slow transients as opposed to such trips as low flow or high flux which will trip the reactor for rapid changes in flow or neutron flux, respectively, that would result in fuel damage before actuation of the slower responding T trips could be effected.
Therefore, the reactor trip system has been designed to provide protection for fuel cladding and reactor coolant system pressure boundary integrity where:  1) a rapid change in a single variable or factor will quickly result in exceeding
 
a core or a system safety limit and 2) a slow change in one or more variables
 
will have an integrated effect which will cause safety limits to be exceeded.
Overall, the reactor trip system offers diverse and comprehensive protection against fuel cladding failure and/or loss of reactor coolant system integrity
 
for Condition II and III accidents. This is demonstrated by Table 7.2-4, which
 
lists the various trips of the reactor trip system, the corresponding Technical
 
Specification sections on safety limits and safety system settings, and the appropriate accident discussed in the safety analyses in which the trip could
 
be utilized.
The reactor trip system automatically provides core protection during nonstandard operating configuration, i.e., operation with a loop out of
 
service. Although operating with a loop out of service over an extended time
 
is considered to be an unlikely event and is prohibited by Technical
 
Specifications, no protection system setpoints would need to be reset. This is
 
because the nominal value of the power (P-8) interlock setpoint restricts the
 
power so that DNBRs less than the thermal design limit DNBR are not realized
 
during any Condition II transients occurring during this mode of operation.
 
This restricted power is considerably below the boundary of permissible values as defined by the core safety limits for operation with a loop out of service.
Thus the P-8 interlock acts essentially as a high nuclear power reactor trip
 
when operating with one loop not in service. By first resetting the coefficient setpoints in the overtemperature T function to more restrictive values, as listed in the Technical Specifications, the P-8 setpoint can then be increased to the maximum value consistent with maintaining DNBR above 1.30 for
 
Condition II transients in the one loop shutdown mode. The resetting of the
 
overtemperature DT trip      7.2-21    Rev. 12 WOLF CREEK and P-8 is carried out under prescribed administrative procedures, under the direction of authorized supervision, and with the plant conditions prescribed
 
in the Technical Specifications.
The design meets the requirements of GDC-21 (refer to Section 3.1).
Preoperational testing is performed on reactor trip system components and systems to determine equipment readiness for startup. This testing serves as a
 
further evaluation of the system design.
Analyses of the results of Condition I, II, III, and IV events, including considerations of instrumentation installed to mitigate their consequences, are presented in Chapter 15.0. The instrumentation installed to mitigate the consequences of load rejection and turbine trip is given in Section 7.4.
7.2.2.2.2  Reactor Coolant Flow Measurement
 
The elbow taps used on each loop in the primary coolant system are instrument devices that indicate the status of the reactor coolant flow. The basic
 
function of this device is to provide information as to whether or not a
 
reduction in flow has occurred. The correlation between flow and elbow tap
 
signal is given by the following equation:
DP DP o  =
W W o 2 whereP o is the pressure differential at the reference flow W o and P is the pressure differential at the corresponding flow, W. The full flow reference point is established during initial plant startup. The low flow trip point is
 
then established by extrapolating along the correlation curve. The expected absolute accuracy of the channel is within  10 percent of full flow, and field results have shown the repeatability of the trip point to be within  1 percent.7.2.2.2.3  Evaluation of Compliance to Applicable Codes and Standards The reactor trip system meets the criteria of the GDC, as indicated. The reactor trip system meets the requirements of Section 4 of IEEE Standard 279-
 
1971, as indicated below.      7.2-22    Rev. 0 WOLF CREEK
: a. General functional requirement The protection system automatically initiates appropriate protective action  whenever a condition  monitored by the system reaches a preset level. Functional performance requirements are given in Section 7.2.1.1.1. Section 7.2.1.2.4 presents a discussion of limits, margins, and levels; Section 7.2.1.2.5 discusses unusual (abnormal) events; and Section 7.2.1.2.6 presents minimum performance requirements.
: b. Single failure criterion
 
The protection system is designed to provide two, three,          or four instrumentation channels for each protective
 
function and two logic train circuits. These redundant channels and trains are electrically isolated and physically separated. Thus, any single failure within a channel or train would not prevent protective action at the system level when required. Loss of input power to a channel or logic train, the most likely mode of failure,          will result in a signal calling for a trip. This design meets the requirements of GDC-23 (refer to Section 3.1).
To prevent the occurrence of common mode failures, such additional measures as functional diversity, physical separation, and testing, as well as administrative
 
control during design, production, installation, and
 
operation, are employed, as discussed in Reference 4.
 
The design meets the requirements of GDC-21 and 22 (refer to Section 3.1).
: c. Quality of components and modules
 
For a discussion on the quality of the components and modules used in the reactor trip system, refer to Chapter 17.0. The quality assurance applied conforms to GDC-1 (refer to Section 3.1).
: d. Equipment qualification For a discussion of the type tests made to verify the performance requirements, refer to Section 3.11(N). The
 
test results demonstrate that the design meets the
 
requirements of GDC-4 (refer to Section 3.1).      7.2-23    Rev. 12 WOLF CREEK
: e. Channel integrity Protection system channels required to operate in accident conditions maintain necessary functional capability under extremes of conditions relating to environment, energy supply, malfunctions, and accidents.
The energy supply for the reactor trip system is described in Section 7.6 and Chapter 8.0. The
 
environmental variations throughout which the system will perform are given in Section 3.11(N).
: f. Independence Channel independence is carried throughout the system,          extending from the sensor through to the devices actuating the protective function. Physical separation
 
is used to achieve separation of redundant transmitters.
 
Separation of wiring is achieved using separate wireways, cable trays, conduit runs, and containment penetrations
 
for each redundant channel. Redundant analog equipment
 
is separated by locating modules in different protection
 
cabinets. Each redundant protection channel set is
 
energized from a separate ac power feed. This design
 
meets the requirements of GDC-21 (refer to Section 3.1).
Two reactor trip breakers, which are actuated by two separate logic matrices, interrupt power to the control
 
rod drive mechanisms. The breaker main contacts are
 
connected in series with the power supply so that opening
 
either breaker interrupts power to all control rod drive mechanisms, permitting the rods to free fall into the core (see Figure 7.1-1).
The design philosophy is to make maximum use of a wide variety of measurements. The protection system continuously monitors numerous diverse system variables.
Generally, two or more diverse protection functions would terminate an accident before limits are exceeded. This design meets the requirement of GDC-22 (refer to Section 3.1).
: g. Control and protection system interaction
 
The protection system is designed to be independent of the control system. In certain applications, the control signals and other nonprotective functions are derived from individual protection channels through isolation      7.2-24    Rev. 1 WOLF CREEK amplifiers, as described in Section 7.2.1.1.8. The isolation amplifiers are classified as part of the
 
protection system and are located in the analog protection racks. Nonprotective functions include those signals used for control, remote process indication, and
 
computer monitoring. The isolation amplifiers are designed such that a short circuit, open circuit, or the application of credible fault voltages from within the
 
cabinets on the isolated output portion of the circuit (i.e., the nonprotective side of the circuit) will not
 
affect the input (protective) side of the circuit. The signals obtained through the isolation amplifiers are never returned to the protective racks. This design meets the requirements of GDC-24 and Section 4.7 of IEEE Standard 279-1971 (refer to Section 3.1).
The results of applying various malfunction conditions on the output portion of the isolation amplifiers show that
 
no significant disturbance to the isolation amplifier
 
input signal occurred.
: h. Derivation of system inputs To the extent feasible and practical, protection system inputs are derived from signals which are direct measures of the desired variables. Variables monitored for the various reactor trips are listed in Section 7.2.1.2.2.
: i. Capability for sensor checks
 
The operational availability of each system input sensor during reactor operation is accomplished by cross checking between channels that bear a known relationship
 
to each other and that have readouts available. Channel
 
checks are discussed in the Technical Specifications.
: j. Capability for testing
 
The reactor trip system is capable of being tested during power operation. Where only parts of the system are tested at any one time, the testing sequence provides the necessary overlap between the parts to ensure complete
 
system operation. The testing capabilities are in
 
conformance with Regulatory Guide 1.22, as discussed in
 
Section 7.1.2.5.2.
The protection system is designed to permit periodic testing of the analog channel portion of the reactor trip
 
system during reactor power operation without initiating      7.2-25    Rev. 0 WOLF CREEK a protective action, unless a trip condition actually exists. This is because of the coincidence logic
 
required for reactor trip. These tests may be performed at any plant power from cold shutdown to full power.
Before starting any of these tests with the plant at
 
power, all redundant reactor trip channels associated with the function to be tested must be in the normal (untripped) mode in order to avoid spurious trips.
 
Setpoints are referenced in the precautions, limitations,          and setpoints portion of the plant technical manual.
Analog Channel Tests Analog channel testing is performed at the analog instrumentation rack set by individually introducing
 
dummy input signals into the instrumentation channels and
 
observing the tripping of the appropriate output tables.
Process analog output to the logic circuitry is interrupted during individual channel test by a test switch which, when thrown, deenergizes the associated logic input and inserts a proving lamp in the bistable output. Interruption of the bistable output to the logic
 
circuitry for any cause (test, maintenance purposes, or
 
removed from service) will cause that portion of the
 
logic to be actuated (partial trip), accompanied by a partial trip alarm and channel status light actuation in the control room. Each channel contains those switches,          test points, etc. necessary to test the channel. See References 1 and 2 for additional background information.
The following periodic tests of the analog channels of the protection circuits are performed:
: 1. T avg and T protection channel testing
: 2. Pressurizer pressure protection channel testing
: 3. Pressurizer water level protection channel testing
: 4. Steam generator water level protection channel testing
: 5. Reactor coolant low flow, underfrequency, and under-voltage protection channel testing
: 6. Steam pressure protection channel testing
: 7. Containment pressure channel testing      7.2-26    Rev. 0 WOLF CREEK Nuclear Instrumentation Channel Tests The power range channels of the nuclear instrumentation system are tested by superimposing a test signal on the actual detector signal being received by the channel at the time of testing. The output of the bistable is not placed in a tripped condition prior to testing. Also,          since the power range channel logic is two out of four,          bypass of this reactor trip function is not required.
To test a power range channel, a "TEST-OPERATE" switch is provided to require deliberate operator action.
 
Operation of this switch will initiate the "CHANNEL TEST" annunciator in the control room. Bistable operation is
 
tested by increasing the test signal to its trip setpoint
 
and verifying bistable relay operation by control board
 
annunciator and trip status lights. It should be noted
 
that a valid trip signal would cause the channel under
 
test to trip at a lower actual reactor power level.
A reactor trip would occur when a second bistable trips.
No provisions have been made in the channel test circuit
 
for reducing the channel signal level below that signal
 
being received from the nuclear instrumentation system
 
detector.
A nuclear instrumentation system channel which can cause a reactor trip through one of two protection logic
 
(source or intermediate range) is provided with a bypass
 
function which prevents the initiation of a reactor trip
 
from that particular channel during the short period that
 
it is undergoing test. These bypasses are annunciated in
 
the control room.
The following periodic tests of the nuclear instrumentation system are performed:
: 1. Testing at plant shutdown:  a) source range testing,              b) intermediate range testing, and c) power range
 
testing
: 2. Testing between P-6 and P-10 permissive power levels:  a) Source range, and b) power range testing
: 3. Testing above P-10 permissive power level: power range testing      7.2-27    Rev. 0 WOLF CREEK Any deviations noted during the performance of these tests are investigated and corrected in accordance with
 
the established calibration and trouble shooting procedures provided in the plant technical manual for the nuclear instrumentation system. Control and protection
 
trip settings are indicated in the plant technical manual under precautions, limitations, and setpoints.
For additional background information on the nuclear instrumentation system, see Reference 2.
Solid State Logic Testing The reactor logic trains of the reactor trip system are designed to be capable of complete testing at power.
 
After the individual channel analog testing is complete, the logic matrices are tested from the train A and train B logic rack test panels. This step provides overlap between the analog and logic portions of the test program. During this test, all of the logic inputs are actuated automatically in all combinations of trip and nontrip logic. Trip logic is not maintained sufficiently
 
long enough to permit opening of the reactor trip
 
breakers. The reactor trip undervoltage coils and auto
 
shunt trip relays are "pulsed" in order to check continuity. During logic testing of one train, the other train can initiate any required protective functions.
Annunciation is provided in the control room to indicate when a train is in test (train output bypassed) and when a reactor trip breaker is bypassed. Logic testing can be
 
performed in less than 30 minutes.
A direct reactor trip resulting from undervoltage or underfrequency on the reactor coolant pump busses is
 
provided, as discussed in Section 7.2.1 and shown on
 
Figure 7.2-1. The logic for these trips is capable of
 
being tested during power operation. When parts of the trip are being tested, the sequence is such that an
 
overlap is provided between parts so that a complete
 
logic test is provided. Thus complete testing of the RTS is possible.
This design complies with the testing requirements of IEEE Standard 279-1971 and IEEE Standard 338-1971 discussed in Section 7.1.2.6.2.
The permissive and block interlocks associated with the reactor trip system and engineered safety feature
 
actuation system are given on Tables 7.2-2 and 7.3-15 and      7.2-28    Rev. 0 WOLF CREEK designated protection or "P" interlocks. As a part of the protection system, these interlocks are designed to
 
meet the testing requirements of IEEE Standard 279-1971 and IEEE Standard 338-1971.
Testing of all protection system interlocks is provided by the logic testing and semiautomatic testing capabilities of the solid state protection system. In
 
the solid state protection system, the undervoltage coils and auto shunt trip relays (reactor trip) and master
 
relays (engineered safeguards actuation) are pulsed for all combinations of trip or actuation logic with and without the interlock signals. For example, reactor trip on low flow (two out of four loops showing two out of three low flow) is tested to verify operability of the
 
trip above P-7 and nontrip below P-7 (see Figure 7.2-1, Sheet 5). Interlock testing may be performed at power.
Testing of the logic trains of the reactor trip system includes a check of the input relays and a logic matrix
 
check. The following sequence is used to test the
 
system:
: 1. Check of input relays During testing of the process instrumentation system and nuclear instrumentation system channels, each channel bistable is placed in a trip mode, causing one input relay in train A and one in train B to deenergize. A contact of each relay is connected to a universal logic printed circuit card. This card performs both the reactor trip and monitoring functions. Each reactor trip input relay contact causes a status lamp and an annunciator on the control board to operate. Either the train A or train B input relay operation will light the status lamp and annunciator.
Each train  contains a multiplexing test switch. At the start of a process or nuclear instrumentation system test, this switch (in either train) is placed in the A + B position. The A + B position alternately allows information to be transmitted from the two trains to the control board. A steady status lamp and annunciator indicates that input relays in both trains have been deenergized. A flashing lamp means that the input relays in the two trains did not both deenergize. Contact inputs to the logic protection system, such as reactor coolant pump bus underfrequency relays, operate input      7.2-29    Rev. 1 WOLF CREEK relays which are tested by operating the remote contacts as described above and using the same type of indications as those provided for bistable input relays.
Actuation of the input relays provides the overlap between the testing of the logic protection system and the testing of those systems supplying the inputs to the logic protection system. Test indications are status lamps and annunciators on the control board.
Inputs to the logic protection system are checked one channel at a time, leaving  the other channels in service. For example, a function that trips the reactor when two out of four channels trip becomes a one out of three trip when one channel is placed in the trip mode. Both trains of the logic protection system remain in service during this portion of the test.
: 2. Check of logic matrices Logic matrices are checked one train at a time.
Input relays are not operated during this portion of
 
the test. Reactor trips from the train being tested
 
are inhibited with the use of the input error inhibit
 
switch on the semiautomatic test panel in the train.
 
At the completion of the logic matrix tests, one
 
bistable each channel of process instrumentation or nuclear instrumentation is tripped to check closure
 
of the input error inhibit switch contacts.
The logic test scheme uses pulse techniques to check the coincidence logic. All possible trip and nontrip combinations are checked. Pulses from the tester are
 
applied to the inputs of the universal logic card at
 
the same terminals that connect to the input relay
 
contacts. Thus there is an overlap between the input relay check and the logic matrix check. Pulses are fed back from the reactor trip breaker undervoltage coil and auto shunt trip relay to the tester. The
 
pulses are of such short duration that the reactor trip breaker undervoltage trip attachment (UVTA) trip lever and shunt trip attachment (STA) armature cannot respond mechanically.
Test indications that are provided are an annunciator in the control room indicating that reactor trips
 
from the train have been blocked and that the train
 
is being tested and green and red lamps on the semi-
 
automatic tester to indicate a good or bad logic      7.2-30    Rev. 1 WOLF CREEK matrix test. Protection capability provided during this portion of the test is from the train not being
 
tested.
: 3. General warning alarm reactor trip Each of the two trains of the solid state protection system is  continuously monitored by the general
 
warning alarm reactor trip subsystem. The warning circuits are actuated if undesirable train conditions
 
are set up by improper alignment of testing systems,              circuit malfunction or failure, etc., as listed below. A trouble condition in a logic train is indicated in the control room.
However, if any of the conditions exist in both trains at the same time, the general warning alarm
 
circuits will automatically trip the reactor. a. Loss of either 48 volt dc power supply  b. Loss of either 15 volt dc power supply  c. Printed circuit card improperly inserted  d. Input error inhibit switch in the INHIBIT position  e. Slave relay tester mode selector in TEST position  f. Multiplexing selector switch in INHIBIT position  g. Loss of ac power in relay cabinets  h. Bypass breaker racked in and closed  i. Permissive test switch not in OFF position  j. Memory test switch not in OFF position  k. Logic A test switch not in OFF position  l. Master relay selector switch not in OFF position The testing capability meets the requirements of GDC-21 (refer to Section 3.1).
Testing of Reactor Trip Breakers Normally, reactor trip breakers 52/RTA and 52/RTB are in service, and bypass breakers 52/BYA and 52/BYB are with-drawn (out of service). In testing the protection logic,      7.2-31    Rev. 11 WOLF CREEK pulse techniques are used to avoid tripping the reactor trip breakers, thereby eliminating the need to bypass
 
them during this testing. The following procedure describes the method used for testing the trip breakers:
: 1. With bypass breaker 52/BYA racked out, manually close and trip it to verify its operation.
: 2. Rack in and close 52/BYA.
: 3. Manually trip 52/RTA through a protection system logic matrix while at the same time depressing the Auto Shunt Trip Block pushbutton switch on the Auto Shunt Trip Panel. This verifies the operation of the
 
UVTA when the breaker trips.
: 4. Release the Auto Shunt Trip block pushbutton switch.
After reclosing 52/RTA, trip it again by depressing the Auto Shunt Trip test pushbutton switch on the Auto Shunt Trip Panel. This verifies the operation
 
of the STA when the breaker trips.
: 5. Reclose 52/RTA.
: 6. Open and rack out 52/BYA.
: 7. Repeat above steps to test trip breaker 52/RTB, using bypass breaker 52/BYB.
Auxiliary contacts of the bypass breakers are connected into the alarm system of their respective trains so that if either train is placed in test while the bypass
 
breaker of the other train is closed both reactor trip
 
breakers and both bypass breakers will automatically
 
trip.
Auxiliary contacts of the bypass breakers are also connected in such a way that if an attempt is made to close the bypass breaker in one train while the bypass breaker of the other train is already closed both bypass
 
breakers will automatically trip.
The train A and train B alarm systems operate separate annunciators in the control room. The two bypass breakers also operate an annunciator in the control
 
room. Bypassing of a protection train with either the
 
bypass breaker or with the test switches would result in
 
audible and visual indicators.      7.2-32    Rev. 12 WOLF CREEK  Auxiliary switch contacts (P-4) of the reactor trip  breakers which initiate protective functions can be tested on-line to verify proper operation. Testing is  accomplished using selector switches and voltmeters  mounted on the front panels of the reactor trip switchgear cabinets. The complete reactor trip system is normally required to  be in service. However, to permit on-line testing of the various protection channels or to permit continued operation in the event of a subsystem instrumentation channel failure, the Technical Specifications define the required number of operable channels. The Technical  Specifications also define the required restriction  to operation in the event that the channel operability requirements cannot be met. k. Channel bypass or removal from operation The protection system is designed to permit periodic  testing of the analog channel portion of the reactor trip  system during reactor power operation without initiating  a protective action, unless a trip condition actually  exists. This is because of the coincidence logic  required for reactor trip. Additional information is  given in Section 7.2.2.2.2. 1. Operating bypasses Where operating requirements necessitate automatic or  manual bypass of a protective function, the design is such that the bypass is removed automatically whenever permissive conditions are not met. Devices used to achieve automatic removal of the bypass of a protective function are considered part of the protective system and are designed in accordance with the criteria of this section. Indication is provided in the control room if some part of the system has been administratively bypassed or taken out of service. m. Indication of bypasses Bypass indication is addressed in Table 7.5-3.
: n. Access to means for bypassing The design provides for administrative control of access  to the means for manually bypassing channels or  protective functions.      7.2-33    Rev. 13 WOLF CREEK
: o. Multiple setpoints For monitoring neutron flux, multiple setpoints are used. When a more restrictive trip setting becomes necessary to provide adequate protection for a particular mode of operation or set of operating conditions, the protective system circuits are designed to provide positive means or administrative control to ensure that the more restrictive trip setpoint is used. The devices used to prevent improper use of less restrictive trip settings are considered part of the protective system and are designed in accordance with the criteria of this section.
: p. Completion of protective action The protection system is so designed that, once initiated, a protective action goes to completion.
Return to normal operation requires action by the operator.
: q. Manual initiation Switches are provided on the control board for manual initiation of protective action. Failure in the automatic system does not prevent the manual actuation of the protective functions. Manual actuation relies on the operation of a minimum of equipment.
: r. Access The design provides for administrative control of access to all setpoint adjustments, module calibration adjustments, and test points.
: s. Identification of protective actions Protective channel identification is discussed in Section 7.1.2.3. Indication is discussed in item t below.
: t. Information readout The protective system provides the operator with complete information pertinent to system status and safety. All transmitted signals (flow, pressure, temperature, etc.)
which can cause a reactor trip are either indicated or recorded for every channel,      7.2-34    Rev. 1 WOLF CREEK including all neutron flux power range currents (top detector, bottom detector, algebraic difference, and
 
average of bottom and top detector currents).
Any reactor trip actuates an alarm and an annunciator.
Such protective actions are indicated and identified down to the channel level.
Alarms and annunciators are also used to alert the operator of deviations from normal operating conditions
 
so that he may take appropriate corrective action to
 
avoid a reactor trip. Actuation of any rod stop or trip
 
of any reactor trip channel actuates an alarm.
: u. System repair
 
The system is designed to facilitate the recognition,          location, replacement, and repair of malfunctioning
 
components or modules. Refer to the discussion in item j
 
above.
7.2.2.3  Specific Control and Protection Interactions 7.2.2.3.1  Neutron Flux Four power range neutron flux channels are provided for overpower protection.
An isolated auctioneered high signal is derived by auctioneering the four
 
channels for automatic rod control. If any channel fails in such a way as to
 
produce a low output, that channel is incapable of proper overpower protection
 
but will not cause control rod movement because of the auctioneer. Two-out-of-
 
four overpower trip logic ensures an overpower trip if needed, even with an
 
independent failure in another channel.
In addition, channel deviation signals in the control system give an alarm if any neutron flux channel deviates significantly from the average of the flux
 
signals. Also, the control system responds only to rapid changes in indicated
 
neutron flux; slow changes or drifts are compensated by the temperature control signals. Finally, an overpower signal from any nuclear power range channel blocks manual and automatic rod withdrawal. The setpoint for this rod stop is below the reactor trip setpoint.
7.2.2.3.2  Coolant Temperature
 
The accuracy of the narrow range resistance temperature detector (RTD) temperature measurements is demonstrated during plant startup tests by comparing temperature measurements from these      7.2-35    Rev. 6 WOLF CREEK RTDs with one another as well as with the temperature measurements obtained from the wide range RTD located in the hot leg and cold leg piping of each loop. The comparisons are done with the reactor coolant system in an isothermal condition. The linearity of the T measurements obtained from the hot leg and cold leg RTDs as a function of plant power is also checked during plant startup tests. The absolute value of T versus plant power is not important, per se, as far as reactor protection is concerned. Reactor trip system setpoints are based upon percentages of the indicated T at nominal full power rather than on absolute values of T. This is done to account for loop differences which are inherent. The percent T scheme is relative, not absolute, and therefore provides better protective action without the requirement of absolute accuracy. For this reason, the linearity of theT signals as a function of power is of importance rather than the absolute values of the T.Reactor control is based upon signals derived from protection system channels after isolation by isolation amplifiers such that no feedback effect can perturb the protection channels. Since control is based on the average
 
temperature of the loop with the highest temperature, the control rods are
 
always moved based upon the most pessimistic temperature measurement with
 
respect to margins to DNB. A spurious low average temperature measurement from
 
any loop temperature control channel causes no control action. A spurious high average temperature measurement causes rod insertion (safe direction).
Channel deviation signals in the control system give an alarm if any temperature channel deviates significantly from the auctioneered (highest) value. Automatic rod withdrawal blocks and turbine runback (power demand reduction) will also occur if any two out of the four overtemperature or overpowerT channels indicate an adverse condition.
7.2.2.3.3  Pressurizer Pressure The pressurizer pressure protection channel signals are used for high and low pressure protection and as inputs to the overtemperature T trip protection function. Isolated output signals from      7.2-36    Rev. 6 WOLF CREEK these channels are used for pressure control. These are used to control pressurizer spray and heaters and power-operated relief valves. Pressurizer
 
pressure is sensed by fast response pressure transmitters.
A spurious high pressure signal from one channel can cause decreasing pressure by actuation of either spray or relief valves. Additional redundancy is provided in the low pressurizer pressure reactor trip and in the logic for safety injection to ensure low pressure protection.
Overpressure protection is based upon the positive surge of the reactor coolant produced as a result of turbine trip under full load, assuming that the core
 
continues to produce full power. The self-actuated safety valves are sized on the basis of steam flow from the pressurizer to accommodate this surge at a setpoint of 2,500 psia and an accumulation of 3 percent. No credit is taken for the relief capability provided by the power-operated relief valves during
 
this surge.
In addition, operation of any one of the power-operated relief valves can maintain pressure below the high pressure trip point for most transients. The
 
rate of pressure rise achievable with heaters is slow, and ample time and
 
pressure alarms are available to alert the operator of the need for appropriate
 
action.Redundancy is not compromised by having a shared tap (see Section 7.2.1.1.2) since the logic for this trip is two out of three. If the shared tap is plugged,the affected channels will remain static. If the impulse line bursts, the indicated pressure will drop to zero. In either case, the fault is easily
 
detectable, and the protective function remains operable.
7.2.2.3.4  Pressurizer Water Level Three pressurizer water level channels are used for reactor trip. Isolated signals from these channels are used for pressurizer water level control. A
 
failure in the level control system could fill or empty the pressurizer at a
 
slow rate (on the order of half an hour or more).
The high water level trip setpoint provides sufficient margin so that the undesirable condition of discharging liquid coolant through the safety valves
 
is avoided. Even at full power conditions, which would produce the worst
 
thermal expansion rates, a failure of the water level control would not lead to any liquid discharge through the safety valves. This is due to the automatic high pressurizer pressure reactor trip actuating at a pressure sufficiently
 
below the safety valve setpoint.      7.2-37    Rev. 0 WOLF CREEK For control failures which tend to empty the pressurizer, two-out-of-four logic for safety injection action on low pressure ensures that the protection system
 
can withstand an independent failure in another channel. In addition, ample time and alarms exist to alert the operator of the need for appropriate action.
7.2.2.3.5  Steam Generator Water Level
 
The basic function of the reactor protection circuits associated with low-low steam generator water level is to preserve the steam generator heat sink for
 
removal of long term residual heat. Should a complete loss of feedwater occur, the reactor would be tripped on low-low steam generator water level. In
 
addition, redundant auxiliary feedwater pumps are provided to supply feedwater
 
to maintain residual heat removal capability after trip. This reactor trip
 
acts before the steam generators are dry. This reduces the required capacity, increases the time interval before auxiliary feedwater pumps are required, and
 
minimizes the thermal transient on the reactor coolant system and steam
 
generators.
Therefore, a low-low steam generator water level reactor trip circuit is provided for each steam generator to ensure that sufficient initial thermal
 
capacity is available in the steam generator at the start of the transient.
 
Two-out-of-four low-low steam generator water level trip logic ensures a
 
reactor trip, if needed, even with an independent failure in another channel
 
used for control and when degraded by an additional second postulated random
 
failure.A spurious low signal for the feedwater flow channel being used for control would cause an increase in feedwater flow. The mismatch between steam flow and
 
feedwater flow produced by the spurious signal would actuate alarms to alert
 
the operator of the situation in time for manual correction (see Figure 7.2-1, sheets 13, 14). If the condition continues, a two-out-of-four high-high steam generator water level signal in any loop, independent of the indicated feedwater flow, will cause feedwater isolation and trip the turbine. The
 
turbine trip will result in a subsequent reactor trip if power is above the P-9
 
setpoint. The high-high steam generator water level trip is an equipment
 
protective trip preventing excessive moisture carryover which could damage the
 
turbine blading.
In addition, the three-element feedwater controller incorporates reset action on the level error signal, such that with expected controller settings a rapid increase or decrease in the flow signal would cause only a small change in level before the controller would compensate for the level error. A slow
 
change in the feedwater signal would have no effect at all. A spurious low      7.2-38    Rev. 0 WOLF CREEK or high steam flow signal would have the same effect as high or low feedwater signal, discussed above. A spurious high steam generator water level signal
 
from the protection channel used for control will tend to close the feedwater valve. A spurious low steam generator water level signal will tend to open the feedwater valve. Before a reactor trip would occur, two out of four channels
 
in a loop would have to indicate a low-low water level. Any slow drift in the water level signal will permit the operator to respond to the level alarms and take corrective action.
Automatic protection is provided in case the spurious high level reduces feedwater flow sufficiently to cause low-low level in the steam generator.
 
Automatic protection is also provided in case the spurious low level signal increases feedwater flow sufficiently to cause high level in the steam generator. A turbine trip and feedwater isolation would occur on two-out-of-four high-high steam generator water level in any loop.
7.2.2.4  Additional Postulated Accidents Loss of plant instrument air or loss of component cooling water is discussed in Section 7.3.8.2. Load rejection and turbine trip are discussed in further detail in Section 7.7.
The control interlocks, called rod stops, that are provided to prevent abnormal power conditions which could result from excessive control rod withdrawal are discussed in Section 7.7.1.4 and listed in Table 7.7-1. Excessively high power
 
operation (which is prevented by blocking of automatic rod withdrawal), if allowed to continue, might lead to a safety limit (as given in the Technical Specifications) being reached. Before such a limit is reached, protection is available from the reactor trip system. At the power levels of the rod block setpoints, safety limits have not been reached; and, therefore, these rod
 
withdrawal stops do not come under the scope of safety-related systems, and are considered as control systems.
7.2.3  TESTS AND INSPECTIONS The reactor trip system meets the testing requirements of IEEE Standard 338-1971, as discussed in Section 7.1.2.6.2. The testability of the system is
 
discussed in Section 7.2.2.2.3. The initial test intervals are specified in
 
the Technical Specifications. Written test procedures and documentation, conforming to the requirement of IEEE Standard 338-1971, are available for audit by responsible personnel. Periodic testing complies with Regulatory Guide 1.22, as discussed in Sections 7.1.2.5.2 and 7.2.2.2.3.      7.2-39    Rev. 0 WOLF CREEK 7.
 
==2.4  REFERENCES==
: 1. Reid, J. B., "Process Instrumentation for Westinghouse  Nuclear Steam Supply Systems (4 Loop Plant Using WCID 7300  Series Process Instrumentation)," WCAP-7913, January, 1973.  (Additional background information only.) 2. Lipchak, J. B., "Nuclear Instrumentation System," WCAP-8255,  January, 1974.  (Additional background information only.) 3. Katz, D. N., "Solid State Logic Protection System  Description," WCAP-7488-L (Proprietary), January, 1971 and  WCAP-7672 (Non-Proprietary), June, 1971.  (Additional  background information only.) 4. Gangloff, W. C. and Loftus, W. D., "An Evaluation of Solid  State Logic Reactor Protection in Anticipated Transients,"
WCAP-7706-L (Proprietary) and WCAP-7706 (Non-Proprietary),
July, 1971.      7.2-40    Rev. 0 WOLF CREEK TABLE 7.2-1 LIST OF REACTOR TRIPS Coincidence      Protection Reactor Trip Logic Interlocks Comments
: 1. Power range high neutron        2/4          Manual block of low    High and low setting; manual
 
flux                                        setting permitted      block and automatic reset
 
by P-10                of low setting by P-10
: 2. Intermediate range              1/2          Manual block per-      Manual block and automatic high neutron flux                            mitted by P-10        reset
: 3. S ource range high neutron      1/2          Manual block per-      Manual block and automatic flux                                        mitted by P-6;        reset; automatic block above interlocked with      P-10
 
P-10
: 4. Power range high positive      2/4          No interlocks                      --
neutron flux rate
: 5. Power range high negative      2/4          No interlocks                      --
neutron flux rate
: 6. OvertemperatureT            2/4          No interlocks                      --
: 7. OverpowerT                  2/4          No interlocks                      --
: 8. Pressurizer low pressure        2/4          Interlocked with      Blocked below P-7 P-7 Rev. 1 WOLF CREEK TABLE 7.2-1 (S heet 2)                                  Coincidence    Protection Reactor Trip Logic Interlocks Comments
: 9. Pressurizer high pressure      2/4        No interlocks                    --
: 10. Pressurizer high water        2/3        Interlocked with    Blocked below P-7 level                                      P-7
: 11. Low reactor coolant        2/3 in        Interlocked with    Low flow in one loop will cause flow                        any loop      P-7 and P-8        a reactor trip when above P-8, and a low flow in two loops
 
causes a reactor trip when above
 
P-7; blocked below P-7 1/4        Interlocked        Blocked below P-8 with P-8
: 12. Reactor coolant pump        1/2 in        Interlocked        Low voltage on all busses undervoltage                both          with P-7            permitted below P-7
 
busses
: 13. Reactor coolant pump        1/2 in        Interlocked        Underfrequency on one underfrequency              both          with P-7            motor in both busses trips
 
busses                            all reactor coolant pump
 
breakers and cause reactor
 
trip; reactor trip blocked
 
below P-7
: 14. Low-low steam              2/4 in        No interlocks                    --
generator water            any loop
 
level Rev. 0 WOLF CREEK TABLE 7.2-1 (S heet 3)                                  Coincidence      Protection Reactor Trip Logic Interlocks Comments
: 15. S afety injection            Coincident      Interlocked with S ee S ection 7.3 for with actua-    P-11 (If reactor      engineered safety features
 
tion of        coolant pressure      actuation conditions safety          is less than P-11 injection      Tech S pec valve,                                                  P-11 allows manual block)
: 16. Turbine (anticipatory
 
trip)
 
a)  Low trip fluid            2/3          Interlocked with P-9  Blocked below P-9
 
pressure b)  Turbine stop valve        4/4          Interlocked with P-9  Blocked below P-9 close
: 17. Manual                        1/2          No interlocks                  --
Rev. 8 WOLF CREEK TABLE 7.2-2 PROTECTION SYSTEM INTERLOCKS Desig-nation          Derivation Function I. Power Escalation Permissives P-6      Presence of P-6:  1/2          Allows manual block of neutron flux (intermediate      source range reactor trip
 
range) above setpoint Absence of P-6:  2/2            Defeats the block of neutron flux (intermediate      source range reactor trip
 
range) below setpoint P-10      Presence of P-10:  2/4          Allows manual block of neutron flux (power range)      power range (low setpoint)
 
above setpoint                  reactor trip Allows manual block of intermediate range reactor
 
trip and intermediate range
 
rod stops (C-1)
Blocks source range reactor trip (back-up for P-6)
Absence of P-10:  3/4          Defeats the block of power neutron flux (power range)      range (low setpoint)
 
below setpoint                  reactor trip Defeats the block of intermediate range reactor
 
trip and intermediate range
 
rod stops (C-1)
Input to P-7
 
P-11      2/3 pressurizer pressure        Allows manual block of below setpoint                  safety injection actuation on low pressurizer pressure
 
signal 2/3 pressurizer pressure        Defeats manual block of above setpoint                  safety injection actuation Opens all accumulator isolation valves Rev. 0 WOLF CREEK TABLE 7.2-2 (Sheet 2)
Desig-nation            Derivation Function II. Blocks of Reactor Trips P-7      Absence of P-7:  3/4            Blocks reactor trip on:
neutron flux (power            low reactor coolant flow range) below setpoint          in more than one loop, (from P-10)                    undervoltage, under-
 
and                    frequency, pressurizer
 
2/2 turbine impulse            low pressure, and chamber pressure below          pressurizer high level setpoint (from P-13)
P-8      Absence of P-8:  3/4            Blocks reactor trip on neutron flux (power            low reactor coolant flow
 
range) below setpoint          in a single loop P-9      Absence of P-9:  3/4            Blocks reactor trip on neutron flux (power            turbine trip
 
range) below setpoint P-13      2/2 turbine impulse            Input to P-7 chamber pressure below
 
setpoint Rev. 0 WOLF CREEK TABLE 7.2-3 REACTOR TRIP SYSTEM INSTRUMENTATION (Typical for Westinghouse Four Loop PWR)
 
Typical Trip                Typical Time
 
Reactor Trip Signal                  Typical Range                Accuracy                Response (sec)*
: 1. Power range high neutron        1 to 120% of full power        + 5.3% of full scale              0.2 flux
: 2. Intermediate range high        8 decades of neutron          + 12.3% of full scale;            0.2 neutron flux                    flux overlapping source       
 
range by 2 decades
: 3. Source range high neutron      6 decades of neutron          + 11.9% of full scale              0.2 flux                            flux (1 to 10 6 counts/sec)
: 4. Power range high positive      +15% of full power            + 2.3% of full scale              0.2 neutron flux rate
: 5. Power range high negative      -15% of full power            + 2.3% of full scale              0.2 neutron flux rate
: 6. Overtemperature  T            T H 530 to 650°F                + 6.8 F                          5.0 
 
T C 510 to 630°F                                                 
 
T AV 530 to 630°F 
 
P PRZR 1,700 to 2,500 psig F() -50 to +50
 
T setpoint 0 to 150% power
: 7. Overpower  T                  TH 530 to 650°F                + 3.9°F                          5.0 
 
TC 510 to 630°F                                                 
 
TAV 530 to 630° F T setpoint 0 to 150% power
: 8. Pressurizer low pressure        1,700 to 2,500 psig            18 psi (compensated            0.6 signal)
: 9. Pressurizer high pressure      1,700 to 2,500 psig            18 psi (noncompensated          0.6 signal)
 
Rev. 30 WOLF CREEK TABLE 7.2-3 (Sheet 2)
 
Typical Trip                Typical Time
 
Reactor Trip Signal Typical Range Accuracy Response (sec)*
: 10. Pressurizer high water          Entire cylindrical            +
3.5% of full range              1.2 level                          portion of pressurizer        p between taps at design (distance between taps)        temperature and pressure
: 11. Low reactor coolant flow        0 to 120% of rated flow        +
2.5% of full flow within        0.3 range of 70 to 100% of full
 
flow
: 12. Reactor coolant pump            8400 - 12000 Volts            +
7.4% of Span                    1.2 undervoltage
: 13. Reactor coolant pump            40 to 70 Hz                    +
2.0% of Span                    0.3 underfrequency
: 14. Low-low steam generator        +
6 feet approximately        + 22.8% of full scale            1.2 water level                    from nominal full load        over pressure range of 
 
water level                    700 to 1,200 psig
: 15. Turbine trip                            -                                  -                    0.3
 
*The overall allowable response time for each reactor trip channel is given in the Technical
 
Specifications Bases Table B.3.3.1-2. The channel response time value is the elapsed time from when the parameter being sensed by the channel reaches the safety set point until the undervoltage trip coil in the reactor trip breaker is de-
 
energized. The additional time until rods are free to fall into the core is 0.3 second, or less, for the breaker
 
mechanism.
 
Rev. 25 WOLF CREEK TABLE 7.2-4 REACTOR TRIP CORRELATION (a)                    (b)                                Technical  (c)
Trip              Accident Specification
: 1. Power range        Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,    high neutron      Assembly Bank Withdrawal From a          Function 2.b flux trip          Subcritical or Low Power Startup (low setpoint)    Condition (15.4.1)
Feedwater System Malfunction that Results in a Decrease in Feedwater
 
Temperature (15.1.1)
Spectrum of Rod Cluster Control Assembly Ejection Accidents (15.4.8)
: 2. Power range        Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,    high neutron      Assembly Bank Withdrawal From a          Function 2.a flux trip          Subcritical or Low Power Startup (high setpoint)    Condition (15.4.1)
Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power
 
(15.4.2)
Startup of an Inactive Reactor Coolant Pump at an Incorrect
 
Temperature (15.4.4)
Feedwater System Malfunctions that Result in a Decrease in
 
Feedwater Temperature (15.1.1)
Rev. 13 WOLF CREEK TABLE 7.2-4 (Sheet 2)
 
(a)                    (b)                                Technical  (c)
 
Trip              Accident Specification
 
Excessive Increase in Secondary
 
Steam Flow (15.1.3)
 
Inadvertent Opening of a Steam
 
Generator Atmospheric Relief or
 
Safety Valve (15.1.4)
 
Spectrum of Steam System Piping
 
Failures Inside and Outside of
 
Containment in a PWR (15.1.5)
 
Spectrum of Rod Cluster Control
 
Assembly Ejection Accidents
 
(15.4.8)
See Note d, 3. Intermediate      Uncontrolled Rod Cluster Control 3.3.1, Table 3.3.1-1, range high        Assembly Bank Withdrawal From a Function 4 neutron flux      Subcritical or Low Power Startup 
 
trip              Condition (15.4.1)
See Note d, 4. Source range      Uncontrolled Rod Cluster Control 3.3.1, Table 3.3.1-1, high neutron      Assembly Bank Withdrawal From a  Function 5 flux trip          Subcritical or Low Power Startup 
 
Condition (15.4.1)
: 5. Power range Spectrum of Rod Cluster Control 3.3.1, Table 3.3.1-1, high positive Assembly Ejection Accidents  Function 3.a neutron flux (15.4.8) and Rod Withdrawal at      rate trip Power Accidents (15.4.2)
: 6. Power range        Rod Cluster Control Assembly 3.3.1, Table 3.3.1-1, high negative      Misalignment (15.4.3)  Function 3.b
 
flux rate trip Rev. 25 WOLF CREEK TABLE 7.2-4 (Sheet 3)
(a)                    (b)                                Technical  (c)
Trip              Accident Specification
: 7. Overtemperature    Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,T trip          Assembly Bank Withdrawal at Power Function 6 (15.4.2)
Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the
 
Reactor Coolant (15.4.6)
Loss of External Electrical Load (15.2.2)
Turbine Trip (15.2.3)
 
Feedwater System Malfunctions that Result in a Decrease in Feedwater
 
Temperature (15.1.1)
Ecessive Increase in Secondary Steam Flow (15.1.3)
Inadvertent Opening of a Pressurizer Safety or Relief Valve (15.6.1)
Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve (15.1.4)
Loss-of-Coolant Accidents Resulting from the Spectrum
 
of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary (15.6.5)
Rev. 13 WOLF CREEK TABLE 7.2-4 (Sheet 4)
(a)                    (b)                                Technical  (c)
Trip              Accident Specification
: 8. Overpower          Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,T Trip          Assembly Bank Withdrawal at              Function 7 Power (15.4.2)
Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (15.1.1)
Excessive Increase in Secondary Steam Flow (15.1.3)
Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve (15.1.4)
Spectrum of Steam System Piping Failures Inside and Outside of
 
Containment in a PWR (15.1.5)
: 9. Pressurizer        Inadvertent Opening of a Pres-          3.3.1, Table 3.3.1-1,    low pressure      surizer Safety or Relief Valve          Function 8.a trip              (15.6.1)
 
Loss-of-Coolant Accidents Resulting from the Spectrum of Postulated Piping Breaks within the Reactor
 
Coolant Pressure Boundary (15.6.5)
Steam Generator Tube Failure (15.6.3)
: 10. Pressurizer        Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,    high pressure      Assembly Bank Withdrawal of Power        Function 8.6 trip              (15.4.2)
Rev. 13 WOLF CREEK TABLE 7.2-4 (Sheet 5)
(a)                    (b)                                Technical  (c)
Trip              Accident Specification Loss of External Electrical Load (15.2.2)
Turbine Trip (15.2.3)
: 11. Pressurizer        Uncontrolled Rod Cluster Control        3.3.1, Table 3.3.1-1,    high water        Assembly Bank Withdrawal at              Function 9 level trip        Power (15.4.2)
Loss of External Electrical Load (15.2.2)
Turbine Trip (15.2.3)
: 12. Low reactor        Partial Loss of Forced Reactor          3.3.1, Table 3.3.1-1,    coolant flow      Coolant Flow (15.3.1) Function 10 Loss of Non-Emergency AC Power to the Station Auxiliaries (15.2.6)
Complete Loss of Forced Reactor Coolant Flow (15.3.2)
: 13. Reactor coolant    Complete Loss of Forced Reactor          3.3.1, Table 3.3.1-1,    pump under-        Coolant Flow (15.3.2) Function 12 voltage trip
: 14. Reactor coolant    Complete Loss of Forced                  3.3.1, Table 3.3.1-1,    pump under-        Reactor Coolant Flow (15.3.2)Function 13 frequency trip Rev. 13 WOLF CREEK TABLE 7.2-4 (Sheet 6)
(a)                    (b)                                Technical  (c)
Trip              Accident Specification15. Low-low steam      Loss of Normal Feedwater Flow 3.3.1, Table 3.3.1-1,      generator water    (15.2.7) Function 14
 
level trip Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (15.1.2)
Loss of Non-Emergency AC Power (15.2.6)
Feedwater System Pipe Break (15.2.8) 16. Reactor trip      Loss of External Electrical See Note d      on turbine        Load (15.2.2) 3.3.1, Table 3.3.1-1, trip Function 16 Turbine Trip (15.2.3)                        Loss of Non-Emergency AC Power                          (15.2.6)  17. Safety injec-      Inadvertent Opening of a  Steam See Note e      tion signal        Generator Atmospheric Relief or 3.3.1, Table 3.3.1-1, actuation trip    Safety Valve (15.1.4) Function 17 18. Manual trip        Available for all Accidents See Note d                        (Chapter 15.0) 3.3.1, Table 3.3.1-1, Function 1    Rev 15 WOLF CREEK Table 7.2-4 (Sheet 7)
NOTES: (a)  Trips are listed in order of discussion in Section 7.2. (b)  References refer to accident analysis presented in Chapter 15.0.
(c)  References refer to Technical Specifications as approved by the NRC. (d)  This trip is not assumed to function in the accident.(e)  Accident assumes that the reactor is tripped at end-of-life, which is the worst initial condition for this case.                Rev. 15 WOLF CREEK -NEUTRON FLUX DIFFERENCE BETWEEN UPPER AND LOWER LONG ION CHAMBERS A 1 , A 2 -LIMIT OF F DEADBAND B 1 , B 2-SLOPE OF,RAMP; DETERMINES RATE AT WHICH FUNCTION REACHES IT'S MAXIMUM VALUE ONCE DEADBANO IS EXCEEDED C -MAGNITUDE OF MAXIMUM VALUE THE FUNCTION MAY ATTAIN Rev. 0 WOLP CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.2-2 SETPOINT REDUCTION FUNCTION FQR OVERPOWER AND OVER-TEMPERATURE L\T TRIPS MAIN CONTROL BOARD MECHAM I CAL LINK AND BARRIER WOLF CREEK MECHAM I CAL Ll NK AND BARRIER MAIN CONTROL BOARD
! --..,._!--+-T-RIP (B) MOMENTARY if-MOMENTARY
....__ _
I RESET (A) TRIP (B) JRIP (A) !/ RESET (B) L..M-OM_E
.... NT_A_RY_,_
__ ::=MO=M:ENr-TA_R__,Y
-r-'-1-M_O_ME-rN-TA_R..,jY RESET REACTOR TRIP (A) () REACTOR TRIP (A) AND SHUNT COIL TO (A) REACTOR TRIP SWGR I I I RESET REACTOR TRIP (B) I _1 (). REACTOR TRIP (A) AND UNDERVOLTAGE COIL TO {A) LOGIC CABINET, SSPS () REACTOR TRIP (B) ANO SHUNT COIL TO (B) REACTOR TRIP SWGR I ' L _I () REACTOR TRIP {B) AND UNDERVOLTAGE COIL TO (B) LOGIC CABINET, SSPS WOLF CREEK Rev. 0 UPDATED SAFETY ANALYSIS REPORT FIGURE 7.2-3 REACTOR TRIP/ENGINEERED SAFETY FEATURES ACTUATION MECHANICAL LINKAGE WOLF CREEK 7.3  ENGINEERED SAFETY FEATURE SYSTEMS The engineered safety feature actuation system (ESFAS) is comprised of
 
the instrumentation and controls to sense accident situations and
 
initiate the operation of necessary engineered safety features. The
 
occurrence of a limiting fault, such as a loss-of-coolant accident (LOCA) or a steam line break, requires a reactor trip plus actuation
 
of one or more of the engineered safety features in order to prevent
 
or mitigate damage to the core and reactor coolant system components
 
and ensure containment integrity.
 
In order to accomplish these design objectives, the engineered safety
 
feature systems (ESFS) have proper and timely initiating signals which
 
are supplied by the sensors, transmitters, and logic components making
 
up the various instrumentation channels of the ESFAS.
 
A power interruption to the ESFS, in conjunction with a LOCA or other
 
postulated accident, is believed to be a highly improbable event. 
 
However, the accident analyses for WCGS assume a loss of offsite power
 
coincident with certain postulated events, such as a LOCA. In
 
addition, it is assumed that a single failure occurs which causes the
 
loss of one of the two onsite emergency diesel generators.
 
In response to IE Bulletin 80-06 a review was conducted of the
 
drawings for all systems serving safety-related functions at the
 
schematic level to determine whether or not, upon reset of an ESF
 
actuation signal, all associated safety-related equipment remains in
 
its emergency mode. The review revealed that certain equipment would, in particular circumstances, change state upon ESF reset. The
 
affected equipment included the control room and electrical equipment
 
room air-conditioning units, the containment air coolers, the hydrogen
 
mixing fans, and the component cooling water heat exchanger
 
temperature control valves. The control circuits for this equipment
 
were revised to provide seal-in features so that an ESF reset would
 
not change the safeguards state of the equipment.
 
If a loss of offsite power (LOOP) occurs following the onset of a LOCA, the load shedder emergency load sequencer (LSELS) will load the
 
ESF busses with loads required for a LOCA in the proper sequence if
 
the safety injection signal (SIS) is still present.
 
If one assumes that SIS has been reset prior to the LOOP, the LSELS will function to load the ESF busses with only those loads required
 
for LOOP. Operator action would be required to actuate the loads required for LOCA. Guidance for the Operator Actions is provided in
 
the Emergency Operating Procedures
 
7.3-1                        Rev. 31 WOLF CREEK The piping and instrumentation diagrams for the ESFS are included as figures in those sections of this USAR where the mechanical systems are described. The
 
location and layout drawings are referenced in Section 1.2. The electrical
 
schematic diagrams and the control logic diagrams are referenced in Section
 
1.7. The engineered safety feature actuation logic diagrams are included as figures in this section, and are referenced in the appropriate ESF discussions
 
below.
 
The auxiliary supporting ESFS function as described in Chapters 8.0, 9.0, and
 
10.0. Their controls function to support the primary ESF system, as described
 
in the support section. For each primary ESF system, a list of these auxiliary
 
supporting engineered safety feature systems is provided in Table 7.3-12.
 
7.3.1  CONTAINMENT COMBUSTIBLE GAS CONTROL SYSTEM
 
7.3.1.1  Description The concentration of hydrogen in the containment atmosphere is monitored by the
 
system described in Section 6.2.5. The containment combustible gas control
 
equipment (described briefly below and more completely in Section 6.2.5) maintains this hydrogen concentration below the minimum concentration capable
 
of combustion. The emergency exhaust fans are described in Section 9.4.2.
 
7.3.1.1.1  System Description
: a. Initiating circuits
 
The containment combustible gas control equipment is
 
operated manually from control switches located in the
 
main control room. It is not necessary for either recombiner or purge equipment to be initiated automatically because it would take approximately 4 days for the H 2 concentration to reach the control limit of 3 percent H 2 by volume with no H 2 reduction system in operation. The hydrogen mixing fans automatically run at slow speed upon receipt of a safety injection signal.
The operation of the hydrogen mixing fans following an accident is not required. Refer to Section 6.2.5 for additional information.
 
7.3-2                          Rev. 8 WOLF CREEK
: b. Logic
 
The combustible gas control system is manually
 
controlled, as shown by the drawings referenced in Section 1.7.
: c. Bypass
 
Indication of system bypass is provided as described in
 
Section 7.5. The CIS isolates the H2 sampling and purge
 
lines which can manually be reopened when necessary.
: d. Interlocks
 
There are no interlocks on these controls.
: e. Sequencing
 
On loss of offsite power coincident with SIS, the fans
 
(which are MCC loads) are picked up as soon as the diesel
 
generator output breaker is closed onto the bus.
: f. Redundancy
 
Controls are provided on a one-to-one basis with the
 
mechanical equipment so that the controls preserve the
 
redundancy of the mechanical equipment.
: g. Diversity Diversity of control is provided in that the combustible
 
gas control equipment may be controlled from local
 
controls at motor control centers, as well as from the
 
main control room panels.
: h. Actuated devices
 
Table 7.3-1 lists the actuated devices.
: i. Supporting systems
 
The supporting systems required for these controls are
 
the Class IE ac system (described in Section 8.3) and the
 
containment atmosphere monitoring system (described in
 
Section 6.2.5).
 
7.3.1.1.2  Design Bases
 
Design bases for the containment combustible gas control system are that
 
operation is controlled manually from the main control
 
7.3-3                          Rev. 0 WOLF CREEK room and that no single failure shall prevent the containment combustible gas
 
control system from functioning. In addition, the following conditions are
 
considered for the control system components:
: a. Range of transient and steady state conditions and
 
circumstances
 
The electrical power supply characteristics for the
 
controls on this system are as described in Section 8.3.
 
The range of possible environmental conditions for these
 
controls is as described in Section 3.11(N).
: b. Malfunctions, accidents, or other unusual events Fire            Fire protection is discussed in Section
 
9.5.1.
 
Missile        Missile protection is discussed in
 
Section 3.5.
 
Earthquake      Earthquake protection is discussed in
 
Sections 3.7(B) and 3.7(N).
 
7.3.1.1.3  Drawings
 
There is no automatic actuation signal for this system, although the equipment
 
controls include interfaces with sensors and with other devices. However, at
 
the device level, the H2 mixing fans automatically start, and the H2 sampling system isolation valves automatically close, on receipt of CIS. References to the drawings associated with this system are provided as described in the
 
introductory material for this section.
 
NOTE:  The hydrogen mixing fans are not required to function following an accident. Refer to Section 6.2.5 for additional information.
The final control logic diagrams for the individual devices are referenced in
 
Section 1.7. These compare with the PSAR as follows:
: a. Recombiner and emergency exhaust fan controls
: 1. Recombiners:  no functional change, added fault
 
protection.
: 2. Emergency exhaust fans.  (See Section 7.3.3.1.3.)
: b. Mixing fan controls
 
Functionally the hydrogen mixing fans operate as shown in
 
the diagrams referenced in Section 1.7. Details of motor 7.3-4                          Rev. 8 WOLF CREEK overload protection have been added since the PSAR. The control switch maintains contact in slow, fast and normal and stop. The hydrogen mixing fans are loaded onto the diesel generators as soon as the diesels are able to accept loads. The diesel generator load sequencing 
 
signal shown in the PSAR is, therefore, not shown on the 
 
control logic diagram for the hydrogen 
 
mixing fans.
 
NOTE:  The hydrogen mixing fans are not required to function following an accident. Refer to Section 6.2.5 for additional information.
The electrical schematic diagrams in Chapter 8.0 are in accordance with the
 
control logic diagrams.
 
7.3.1.2  Analysis
: a. Conformance to NRC general design criteria
 
The applicable criteria are listed in Table 7.1-2. No
 
deviations or exceptions to those criteria are taken (see Section 3.1).
: b. Conformance to Regulatory Guide 1.7 is described in
 
Section 6.2.5.
: c. Conformance to IEEE Standard 279-1971
 
The design of the control system is based on the
 
applicable requirements of IEEE Standard 279-1971, as
 
follows:
: 1. General Functional Requirement - Paragraph 4.1  The hydrogen mixing fans are able to function automatically and reliably over the full range of transients for all plant conditions for which credit was originally taken in the analyses. The hydrogen mixing fans have been determined to be unnecessary to assure post accident containment atmosphere mixing. They are no longer relied upon by any analyses. The rest of the system functions for all of these plant conditions when manually initiated. The system response time and
 
accuracy are as required in the accident analyses. The H 2 sampling line is manually actuated.
: 2. Single Failure Criterion - Paragraph 4.2 Through use of redundant, independent systems, as previously described, any single failure or multiple
 
failures resulting from a single credible event will not prevent the system
 
from performing its intended function, when required.
: 3. Quality of Components and Modules - Paragraph 4.3 Components and
 
modules used in the construction of the system exhibit a quality consistent
 
with the                              7.3-5                          Rev. 8 WOLF CREEK nuclear power plant design life objective, require
 
minimum maintenance, and have low failure rates.
 
The program for quality assurance is described in
 
Chapter 17.0.
: 4. Equipment Qualification - Paragraph 4.4
 
The system is qualified to perform its intended
 
functions under the environmental conditions
 
specified in Sections 3.10(B) and (N) and 3.11(B) and
 
(N).
: 5. Channel Integrity - Paragraph 4.5
 
All channels will maintain functional capability under all conditions described in Section 7.3.1.1.2.
: 6. Channel Independence - Paragraph 4.6
 
Discussions of the means used to ensure channel
 
independence are given in Sections 7.1.2.2 and
 
8.3.1.4.
: 7. Control  and Protection System Interaction -
 
Paragraph 4.7.
 
No credible failure at the output of an isolation
 
device will prevent the associated channel from
 
performing its intended function. No single random failure in one channel will prevent the other channel from performing the intended function.
: 8. Derivation of System Outputs - Paragraph 4.8
 
To the extent feasible, the system inputs are from
 
direct measurement of the desired variable.
: 9. Capability for Sensor Checks - Paragraph 4.9
 
Sufficient means have been provided to check the
 
operational availability of the system.
: 10. Testing and Calibration - Paragraph 4.10
 
The control system has the capability of testing the
 
devices used to derive the final system output.
 
7.3-6                          Rev. 0 WOLF CREEK
: 11. Channel Bypass or Removal from Operation - Paragraph
 
4.11
 
Testing of one channel can be accomplished during reactor operation without initiating a protective
 
action at the system level.
: 12. Operating Bypasses - Paragraph 4.12
 
There are no permissive conditions on bypasses.
 
Bypass of one channel will not bypass the other
 
channel. Bypass of one system will not bypass any
 
other system.
: 13. Indication of Bypass - Paragraph 4.13
 
If the protective action of any part of the system
 
has been bypassed or deliberately rendered
 
inoperative, the fact is continuously indicated in
 
the control room, as described in Section 7.5.
: 14. Access to Means for Bypassing - Paragraph 4.14
 
Appropriate administrative controls are applied to
 
ensure that access to the means for manually
 
bypassing the system is adequately protected.
: 15. Multiple Set Points - Paragraph 4.15
 
The system is designed so that there are no multiple setpoints.
: 16. Completion of  Protective Action Once It is Initiated
 
            - Paragraph 4.16
 
The system is designed so that once protective action
 
is initiated, it is carried through to completion.
: 17. Manual Initiation - Paragraph 4.17
 
Manual initiation of each function is provided in the
 
control system with a minimum of equipment, by direct
 
control of motor control centers and solenoid valves
 
from panel-mounted control switches. System level
 
actuation of the safety function is not provided
 
since the time required for operation of these
 
functions allows the station operator to take
 
individual action for each controlled device.
 
7.3-7                          Rev. 0 WOLF CREEK
: 18. Access to Set Point Adjustments, Calibration and Test
 
Points - Paragraph 4.18
 
Appropriate administrative controls are applied to ensure that access to the means for adjusting, calibrating, and testing the system is adequately
 
protected.
: 19. Identification of Protective Actions - Paragraph 4.19
 
System protective actions are described and
 
identified down to the channel level.
: 20. Information Readout - Paragraph 4.20
 
Sufficient information is provided to allow the
 
station operator to make a prompt decision regarding
 
the system operating requirements. The indications
 
required for these decisions are provided by supporting systems, as listed in the system description discussed in Section 7.3.1.1.li.
: 21. System Repair - Paragraph 4.21
 
The system is designed to facilitate the recognition, location, replacement, repair, and adjustment of
 
malfunctioning components or modules.
: 22. Identification - Paragraph 4.22
 
Protection system components are identified, as
 
described in Section 7.1.2.3.
: d. Conformance to NRC regulatory guides The applicability of regulatory guides is as shown in
 
Table 7.1-2. References to the discussions of these
 
regulatory guides are presented in Section 7.1.2.5.1.
: e. Failure modes and effects analysis
 
See Table 7.3-2.
: f. Periodic testing
 
Periodic testing of the mechanical equipment associated
 
with this system is discussed in Section 6.2.5. There is
 
no automatic  actuation  equipment for the entire system,
 
7.3-8                          Rev. 1 WOLF CREEK but there is automatic device actuation, as described in
 
Section 7.3.1.1.3. Provisions for periodic testing of
 
the actuation system are discussed in the Technical
 
Specifications and USAR Section 6.2.5.4.
 
7.3.2  CONTAINMENT PURGE ISOLATION SYSTEM
 
7.3.2.1  Description
 
The containment purge isolation system detects any abnormal amount of
 
radioactivity in the containment atmosphere or in the containment purge
 
effluent and initiates appropriate action to ensure that any release of radioactivity to the environs is controlled. The containment purge systems are
 
also isolated by CIS.
 
7.3.2.1.1  System Description
: a. Initiating circuits
 
Redundant and independent gaseous radiation monitors
 
measure the radioactivity levels of the containment
 
atmosphere and of the containment purge effluent. These monitors provide analog radioactivity signals to bistable units in the ESF actuation system. The bistables
 
generate redundant trip signals, and transmit them to the
 
automatic actuation logic. Since the dampers also close
 
on CIS, the initiating logic for CIS shown in Figure
 
7.2-1 (Sheet 8) is also applicable.
: b. Logic
 
A logic diagram for the ESF actuation system is provided
 
as Figure 7.3-1. This diagram shows only the actuation
 
systems; it does not detail the bypass, bypass interlock, or test provisions. The logic for the containment purge
 
isolation actuation subsystem is included in this figure.
 
The ESFAS hardware consists of solid-state bistables and logic elements, with electromechanical relays as the
 
final output devices. The output relays are all
 
energize-to-actuate, with contact operation as required
 
for each actuated device.
 
The ESFAS is divided into three input-logic-output
 
channels. These channels all meet the independence and
 
separation criteria, as described elsewhere in this
 
chapter. The logic  channels are uniquely  associated
 
with the output channels. The input signals from all
 
7.3-9 Rev. 13 WOLF CREEK three input channels are isolated as necessary, and the 
 
isolated signals are transmitted to the logic channels as
 
shown in Figure 7.3-1.
 
Interconnection of differing separation groups within the
 
BOP ESFAS is by means of digital signal isolation
 
modules. Analog signal isolation modules are included to
 
provide isolated analog signals to the plant computer.
Adequate physical separation or barriers are provided between differing separation groups, and wiring is routed
 
in separated wireways, where appropriate. The wiring is
 
color-coded with regard to separation group.
 
The digital signal isolation modules utilize optical
 
isolators with appropriate signal and power conditioning
 
circuits. The output circuits are powered by the devices
 
receiving signals from the isolation modules, so no power
 
isolation is required. There are no connections between
 
the input and output circuits, except for the optical
 
coupling in the isolation devices.
 
The analog signal isolation modules utilize transformers
 
as the isolation devices. The analog input signals and the input power are converted to pulse trains and applied to the primary windings, and then they are reconstructed
 
by circuits connected to the transformer secondaries.
 
There are no connections between the input and output
 
circuits  except for the magnetic coupling in the
 
transformers.
 
Both the analog and the digital signal isolation modules
 
are tested to ensure a minimum isolation potential of
 
1,500 Vac rms between the input terminals and the output
 
terminals (all input terminals shorted together and all
 
output terminals shorted together), and between the
 
terminals and ground (all terminals shorted together).
 
The 1,500 Vac rms test voltage was applied for at least
 
60 seconds for each test.
Once generated, any actuation signal remains present
 
until it is manually reset. Each bistable automatically
 
resets when its input signal returns to the "safe" side
 
of the setpoint-deadband region.
 
An automatic test system is provided. The system
 
periodically checks the operability of the channel  and
 
alerts the plant operator, via the annunciator and plant computer
 
7.3-10    Rev. 21 WOLF CREEK alarms, if a fault is detected. Provision is also
 
included for manual testing. These test provisions do
 
not compromise the integrity of any channel. They are
 
isolated and will not propagate any fault, and the automatic test function is overridden by any actuation
 
input. Bistable  bypass switches  are provided to permit
 
the testing of bistables. The switches are key-locked, and the key cannot be removed from the lock unless the
 
switch is in the "OPERATE" position. Visual indication
 
of any bypass of any bistables is provided at the ESFAS
 
cabinets; channel-level bypass indication is provided on
 
the main control board.
: c. Bypass There is no device level override on this system.
: d. Interlocks
 
There are no interlocks on these controls.
: e. Sequencing
 
There is no automatic sequencing of operation. The
 
system is permanently connected to the diesel bus and is
 
energized as soon as the diesel output breaker closes.
: f. Redundancy
 
Controls are provided on a one-to-one basis with the mechanical equipment so that the controls preserve the
 
redundancy of the mechanical equipment.
: g. Diversity
 
Diversity of sensing is provided in that containment
 
purge isolation can be actuated by the containment
 
atmosphere gaseous radioactivity monitors, by the
 
containment purge gaseous radioactivity monitors, and by
 
the CIS.
: h. Actuated devices
 
Table 7.3-3 lists the actuated devices.
: i. Supporting systems
 
Supporting systems for the containment purge isolation
 
are the four 125-V dc power supplies discussed in Section
 
8.3 and the instrument air system described in Section 
 
7.3-11    Rev. 0 WOLF CREEK 9.3.1. The isolation function is fail-safe with respect
 
to all of these support systems, that is to say, loss of
 
these support systems will not prevent isolation.
 
7.3.2.1.2  Design Bases
 
The design bases for the containment purge isolation system are described in
 
Section 6.2.4.1.1 (Safety Design Bases 3 and 6) and Section 7.3.1.1.2.
 
7.3.2.1.3  Drawings
 
The logic for the containment purge isolation system is shown on the engineered
 
safety feature actuation system logic diagram, Figure 7.3-1. The differences
 
between this logic and that provided in the PSAR are as follows:
: a. Logic memories are provided at the final actuation
 
outputs, rather than on each digital input.
: b. The indications and alarms for this system have been
 
revised.
: c. Purge supply fans (shutdown and mini):  Additional
 
details on overload protection, stop on containment purge
 
isolation signal (CPIS) (isolated) from Westinghouse-supplied ESFAS, and stop on supply air low temperature.
: d. Purge exhaust fans (shutdown and mini):  Additional
 
details on overload protection, stop on CPIS, and stop on
 
high charcoal temperature in the exhaust filter-adsorber
 
unit.
: e. The purge system containment isolation dampers operate as
 
shown in Figure 7.3-1. The system differs from the PSAR
 
in that the CIS is replaced by the CPIS.
: f. The containment minipurge fan discharge damper opens when
 
the fan is running and closes when the fan is stopped.
 
7.3.2.2  Analysis
: a. Conformance to NRC general design criteria
 
The applicable criteria are listed in Table 7.1-2. No deviations or exceptions to those criteria are taken.
: b. Conformance to IEEE Standard 279-1971
 
The design of the control system conforms to the
 
applicable requirements of IEEE Standard 279-1971, as
 
listed
 
7.3-12    Rev. 1 WOLF CREEK and discussed in Section 7.3.1.2, except that the system
 
actuation is automatic. The ranges and setpoints are in
 
the Technical Specifications.
: c. Conformance to NRC regulatory guides
 
The applicability of the regulatory guides is as shown in
 
Table 7.1-2. References to the discussions of these
 
regulatory guides are presented in Section 7.1.2.5.1.
: d. Failure modes and effects analysis
 
See Table 7.3-4.
: e. Periodic testing
 
Periodic testing of the mechanical equipment associated
 
with this system is discussed in Section 9.4. Periodic
 
testing of the actuation system is discussed in the
 
Technical Specifications.
 
7.3.3  FUEL BUILDING VENTILATION ISOLATION
 
7.3.3.1  Description Upon detection of high radioactivity by the fuel building exhaust gaseous
 
radioactivity monitors, the fuel building ventilation system is automatically
 
realigned through the ESFAS to meet the following requirements:
: a. Isolate normal ventilation.
: b. Initiate operation of the emergency exhaust system to
 
maintain the fuel building atmosphere at a negative
 
pressure.
: c. Reduce the flow of fuel building air to the outside
 
atmosphere to a minimum consistent with maintaining the
 
required building negative pressure.
: d. Filter the exhaust air through HEPA and charcoal filters.
 
A description of the entire fuel building ventilation
 
system is given in Section 9.4.
 
7.3-13    Rev. 0 WOLF CREEK 7.3.3.1.1  System Description
: a. Initiating circuits
 
Two independent gaseous radioactivity monitors measure
 
the radioactivity level in the fuel building exhaust line
 
and provide analog radioactivity signals to bistable
 
units in the ESF actuation system. The bistable units
 
generate two redundant trip signals  and transmit them to
 
the automatic actuation logic.
 
The emergency exhaust system is on standby for an
 
automatic start following receipt of a fuel building
 
isolation signal or an SIS. The initiation of the LOCA mode of operation (SIS signal) takes precedence if both signals are received so that the emergency ventilation is
 
directed to the auxiliary building (see Section 9.4).
: b. Logic
 
The logic for the fuel building ventilation isolation
 
actuation system is included in Figure 7.3-1. The
 
actuation signal is transmitted to each actuated device, and causes each device to assume its "safe" state.
: c. Bypass
 
There is no device level override on this system.
: d. Interlocks
 
There are no interlocks on these controls.
: e. Sequencing
 
There is no automatic sequencing of operation. The
 
system is permanently connected to the diesel bus and is
 
energized as soon as the diesel output breaker closes.
: f. Redundancy
 
Controls are provided on a one-to-one basis with the
 
mechanical equipment so that the controls preserve the
 
redundancy of the mechanical equipment. There are two
 
channels of actuation initiated by redundant
 
radioactivity monitors or redundant manual initiation
 
switches.
 
7.3-14    Rev. 0 WOLF CREEK
: g. Diversity
 
Diversity of control is provided in that the fuel
 
building ventilation isolation system can be actuated by either automatic signals or manual control.
: h. Actuated devices
 
Table 7.3-5 lists the actuated devices.
: i. Supporting systems
 
Supporting systems for the fuel building ventilation
 
isolation system actuation are the four 125-V dc power supplies discussed in Section 8.3 and the instrument air system described in Section 9.3.1. The isolation
 
function is fail-safe with respect to all of these
 
support systems; that is to say, loss of these support
 
systems will not prevent isolation.
 
7.3.3.1.2  Design Bases
 
The design bases for the fuel building ventilation isolation system are
 
discussed in Section 9.4.2.1.1  (Safety Design Bases 1, 3, 4, and 6).
 
Additionally, the design bases described in Section 7.3.1.1.2 are applicable
 
for the control system components.
 
7.3.3.1.3  Drawings The logic diagram for the fuel building ventilation isolation actuation system
 
is included in Figure 7.3-1. The differences between this logic and that
 
provided in the PSAR are the same as those for the containment purge isolation
 
system (see Section 7.3.2.1.3). In addition, actuation system reset is not
 
provided in the fuel building.
 
The control logic diagrams, the electrical schematic diagrams, the piping and
 
instrument diagrams, and the physical location drawings for this system are
 
included in the references in the introductory material for this section.
 
7.3.3.2  Analysis
: a. Conformance to NRC general design criteria
 
The applicable criteria are listed in Table 7.1-2. No deviations or exceptions to those criteria are taken.
 
7.3-15 Rev. 0 WOLF CREEK
: b. Conformance to IEEE Standard 279-1971
 
The design of the control system conforms to the
 
applicable requirements of IEEE Standard 279-1971, as listed and discussed in Section 7.3.1.2c, except that the
 
system functions automatically. The setpoints are
 
provided in the Technical Specifications.
: c. Conformance to NRC regulatory guides
 
The applicability of the regulatory guides is as shown in
 
Table 7.1-2. References to the discussions of
 
conformance to these regulatory guides are presented in
 
Section 7.1.2.5.1.
: d. Failure modes and effects analysis
 
See Table 7.3-6.
: e. Periodic testing
 
Periodic testing of the mechanical equipment associated
 
with this system is discussed in Section 9.4.2.
 
Provisions for the periodic testing of the actuation
 
system are discussed in the Technical Specifications.
 
7.3.4  CONTROL ROOM VENTILATION ISOLATION
 
7.3.4.1  Description Upon detection of high gaseous radioactivity levels the normal supply of outside air to the control room is terminated, as described in Section 6.4. In
 
this event, the control room air is recycled and filtered, and a small supply
 
of fresh makeup air is provided. The control room is maintained at a set
 
positive pressure to prevent the ingress of the local ambient atmosphere. 
 
Normal ventilation is restored only by manual operation by the plant operator, and is maintained only if the local ambient atmosphere poses none of the
 
monitored hazards.
 
7.3.4.1.1  System Description
: a. Initiating circuits
 
The gaseous radioactivity level of the air provided to the main control room from the local ambient atmosphere is monitored by two separate and independent monitoring systems.
 
7.3-16    Rev. 8 WOLF CREEK The analog signals from these monitors are transmitted to bistables in the ESFAS. If acceptable levels are exceeded, the control room is isolated, as described above. Monitors are also provided to measure the  particulate and iodine radioactivity levels in the normal supply air.
 
In addition to the above, control room isolation is initiated upon:
: 1. Fuel building ventilation isolation
: 2. Containment isolation Phase A
: 3. Manual initiation
: 4. High containment atmosphere radioactivity level (GTRE0031 and GTRE0032)
: 5. High containment purge radioactivity level    (GTRE0022 and GTRE0033)
: b. Logic
 
The control room ventilation isolation actuation system logic is included in Figure 7.3-1. The actuation signal  is transmitted to each actuated device, and, subject to the provisions of override, causes each device to assume its "safe" state.
: c. Override
 
Manual override is available by means of pull-to-lock switches on the fans.
: d. Interlocks There are no interlocks on these controls.
 
7.3-17    Rev. 17 WOLF CREEK
: e. Sequencing
 
CRVIS is sequenced to Class IE and control room HVAC
 
units.
: f. Redundancy
 
Controls are provided on a one-to-one basis with the
 
mechanical equipment so that the controls preserve the
 
redundancy of the mechanical equipment. Redundancy is
 
provided in the chlorine and gaseous radioactivity
 
monitors, the actuation signals, and manual actuation
 
switches.
: g. Diversity
 
Diversity of actuation is provided in that the control
 
room ventilation system may be isolated by either an
 
automatic system or by operator manual actuation.
 
Diversity is provided actuation from the gaseous
 
radioactivity and manual switches.
: h. Actuated devices
 
Table 7.3-8 lists the actuated devices.
: i. Supporting system
 
The supporting system required for the controls is the
 
vital Class IE ac system described in Section 8.3.
 
7.3.4.1.2  Design Bases
 
The design bases for the control room ventilation isolation system are that no
 
single failure shall prevent the isolation of the control room ventilation
 
system. The trip points are provided in the Technical Specifications.
Additionally, the design bases described in Section 7.3.1.1.2 are applicable to
 
the control system components.
 
7.3.4.1.3  Drawings
 
The logic diagram for the control room ventilation isolation actuation is
 
included in Figure 7.3-1. Other drawings pertaining to this system are
 
included i the references in Section 1.7.
 
7.3-18    Rev. 8 WOLF CREEK 7.3.4.2  Analysis
: a. Conformance to NRC general design criteria
 
The applicable criteria are listed in Table 7.1-2. No deviations or exceptions to those criteria are taken.
: b. Conformance to IEEE Standard 279-1971
 
The design of the control system conforms to the
 
applicable requirements of IEEE Standard 279-1971, as
 
listed and discussed in Section 7.3.1.2c, except that the
 
system is automatically actuated. The setpoints are
 
provided in the Technical Specifications.
: c. Conformance to NRC regulatory guides
 
The applicability of regulatory guides is as shown in
 
Table 7.1-2. References to the discussions of these
 
regulatory guides are presented in Section 7.1.2.5.1.
: d. Failure mode  and effects analysis
 
This analysis is given in Table 7.3-9.
: e. Periodic testing
 
Periodic testing of the mechanical equipment associated
 
with this system is discussed in Section 9.4.1.
 
Provisions for the periodic testing of the actuation system are discussed in the Technical Specifications.
 
7.3.5  DEVICE LEVEL MANUAL OVERRIDE
 
7.3.5.1  Description The purpose of device level manual override is to provide the capability for
 
manually overriding the actuation signal command when there is an operational
 
need to do so in the post-event situation. This equipment is only included in the designs of the post-event monitoring and sampling systems to allow manual
 
override of the containment isolation signal. When the override function has
 
been achieved, an amber light on the main control board indicates that the
 
device has been removed from the state initiated by  the  actuation  signal. 
 
Operation of the  control 
 
7.3-19    Rev. 5 WOLF CREEK switch to the position corresponding to the actuation signal command is
 
indicated by extinguishing the amber light indication. Logic diagrams are
 
referenced in Section 1.7.
 
7.3.5.2  Analysis The design of the override feature is in conformance with the criteria, guides, and standards applicable to the control circuits to which it is applied. The
 
failure modes and effects analysis is provided in Table 7.3-10.
 
7.3.6  AUXILIARY FEEDWATER SUPPLY
 
7.3.6.1  Description The auxiliary feedwater system (AFS) consists of two motor-driven pumps, one
 
steam turbine-driven pump, and piping, valves, instruments, and controls, as
 
shown in Figure 10.4-9. The pumps are started automatically on receipt of signals from the actuation logic, as shown in Figure 7.3-1. All three pumps
 
can also be started manually from control switches in the control room or at
 
the auxiliary shutdown control panel.
 
The preferred source of water for the AFS is the condensate storage tank (CST). 
 
However, this tank is not seismic Category I.
 
An automatic subsystem is provided, therefore, to monitor the water supply
 
pressure from this tank and initiate switchover to the essential service water
 
system should the supply from the condensate storage tank be interrupted.
Each motor-driven pump feeds two steam generators through individual motor-
 
operated flow control valves. AFS flow can be regulated manually from the
 
control room or from the auxiliary shutdown control panel.
 
The turbine-driven pump feeds all four steam generators through individual air-
 
operated flow control valves. AFS flow can be regulated manually from the
 
control room or from the auxiliary shutdown control panel.
 
AFS flow indication is provided for each steam generator in the control room
 
and at the auxiliary shutdown control panel.
 
The AFS pump turbine is supplied with motive power from two main steam lines
 
through two normally closed, air-operated steam supply valves. A normally
 
closed motor-operated trip and throttle valve is also provided at the inlet to the pump driver. Control of the steam supply and trip and throttle valve, as well as manual speed 
 
7.3-20    Rev. 0 WOLF CREEK control for the turbine-driven pump, is provided in the control room and at the
 
auxiliary shutdown control panel.
 
The status of the motor-driven pumps, the turbine-driven pump, the turbine steam supply valves, and the trip and throttle valve is indicated in the
 
control room and at the auxiliary shutdown control panel. The AFS flow to each
 
steam generator is indicated on both the main control board and at the
 
auxiliary shutdown control panel.
 
The AFS equipment is described in Section 10.4.9.
 
In addition to initiating functions described above, the auxiliary feedwater
 
actuation signal (AFAS) closes the steam generator blowdown and sample
 
isolation valves, when auxiliary feedwater is required by plant conditions.
All remote manually operated valves in the normal suction from the CST and in the discharge to the steam generators are normally open.
 
7.3.6.1.1  System Description
: a. Initiating circuits
 
The motor-driven pumps are started on the occurrence of
 
any one of the following signals:
: 1. Manual start
: 2. Safeguards sequence signal (initiated by safety
 
injection signal or loss-of-offsite-power)
: 3. Auxiliary feedwater actuation (AFAS-M)
 
AFAS-M is generated on the occurrence of any one of the
 
following events:
: 1. Trip of both main feedwater pumps (Manual block of
 
the main feed pump trip signals is provided at the
 
main control board, and is indicated on the ESFAS
 
status panel. This block permits startup and
 
shutdown of the plant without automatic start of the
 
AFPs, while allowing the AFPs to remain available to
 
respond to a demand from any other source.)
: 2. 2 out of 4 low-low level signals in any one steam
 
generator
: 3. ATWS Mitigation System Activation Circuitry (AMSAC)
AMSAC is not required for safety.  (For discussion see Section 7.7.1.11.) 
 
7.3-21    Rev. 4 WOLF CREEK
: 4. Manual AFAS-M initiation 
 
The turbine-driven pump is started on the occurrence of
 
either of the following signals:
: 1. Manual start
: 2. Auxiliary feedwater actuation (AFAS-T)
 
AFAS-T is generated on the occurrence of any one of the
 
following events:
: 1. Loss-of-offsite-power
: 2. Low-low level in any two steam generators
: 3. ATWS Mitigation System Activation Circuitry (AMSAC).
AMSAC is not required for safety.  (For discussion see Section 7.7.1.11.)
: 4. Manual AFAS-T initiation The steam generator sample line containment isolation
 
valves and the steam generator blowdown isolation valves
 
are all automatically closed on the occurrence of a
 
safety-injection signal, a loss-of-offsite-power signal,          or an AFAS. The signal which causes this closure is
 
reset automatically upon reset of the AFAS.
: b. Logic
 
See Figure 7.3-1.
: c. Bypass
 
There is no device level override on this system.
: d. Interlocks
 
The auxiliary feedwater supply valves from the condensate
 
storage tank and from the ESW system are interlocked with
 
the CST supply pressure  sensors,  so that receipt  of
 
2-out-of-3 low pressure signals initiates switchover to
 
essential service water.
: e. Redundancy
 
Sufficient actuation and control channels are provided
 
throughout the auxiliary feedwater system to ensure the
 
required flow to at least two steam generators in the
 
event of a single failure.
 
7.3-22    Rev. 4 WOLF CREEK
: f. Diversity
 
The auxiliary feedwater system is diversified by
 
utilizing a turbine-driven pump with air and dc motor-operated valves and two ac motor-driven pumps with ac
 
motor-operated valves. Diversity in initiating signals
 
can be seen on Figure 7.3-1.
: g. Actuated devices
: 1. Auxiliary feedwater pump turbine steam supply valves
 
(2)
: 2. Auxiliary feedwater pump trip and throttle valve (1)
: 3. Auxiliary feedwater flow control valves (8) (manual
 
only)
: 4. Auxiliary feedwater pump electric motors (2)
: 5. Essential service water supply valves (4)
: 6. Condensate storage tank supply valves (3)
: 7. Steam generator blowdown isolation valves (4)
: 8. Steam generator blowdown sample isolation valves (8)
: h. Supporting systems The Class IE electric system is required for auxiliary
 
feedwater control. The pressurized gas supply required
 
for motive force is normally supplied from the instrument
 
air header, which is not safety related. In addition, each valve has a seismic Category I auxiliary gas supply
 
(see Section 9.3.1).
: i. Portion of system not required for safety
 
Instrumentation provided for monitoring system
 
performance (refer to Section 7.5.3.5) is not required
 
for safety.
 
7.3.6.1.2  Design Bases
 
Auxiliary feedwater is required, as described in Section 10.4.9. No single
 
failure shall prevent this system from operating.
 
7.3-23    Rev. 0 WOLF CREEK Additionally, Section 7.3.1.1.2 is applicable to the control system components.
 
The auxiliary feedwater pumps must achieve full operating speed within 60 seconds of the detection of any condition requiring auxiliary feedwater. This will provide the required flow based on the pumps performance curves.
 
7.3.6.1.3  Drawings
 
The logic diagram for the auxiliary feedwater supply actuation system is
 
included in Figure 7.3-1. The differences between this logic and that provided in the PSAR are the same as those discussed for the containment purge isolation
 
system.
 
Other drawings pertaining to this system are included in Section 7.3. The
 
logic associated with automatic switchover to the ESW has been added.
 
7.3.6.2  Analysis
: a. Conformance to NRC general design criteria
: 1. General Design Criterion 13
 
Instrumentation necessary to monitor station
 
variables associated with hot shutdown is provided in
 
the main control room and on the auxiliary shutdown
 
control panel. Controls for the auxiliary feedwater
 
system are provided at each location. A description
 
of the surveillance instrumentation is provided in
 
Section 7.5.
: 2. General Design Criterion 19 All controls and indications required for safe
 
shutdown of the reactor are provided in the main
 
control room. In the event that the main control
 
room must be evacuated, adequate controls and
 
indications are located outside the main control room
 
to (1) bring to and maintain the reactor in a hot
 
standby condition and (2) provide capability to
 
achieve cold shutdown.
 
The auxiliary shutdown control panel, located outside
 
the main control room, is described in Section 7.4.3.
: 3. General Design Criterion 34
 
The auxiliary feedwater system provides an adequate supply of feedwater to the steam generators to remove 
 
7.3-24    Rev. 11 WOLF CREEK reactor decay heat following reactor trip. Two steam
 
generators with auxiliary feedwater supply are
 
sufficient to remove reactor decay heat without
 
exceeding design conditions of the reactor coolant system.
: 4. Other general design criteria
 
The remaining applicable general design criteria are
 
listed in Table 7.1-2 and Section 10.4.9. No
 
exceptions are taken to those criteria.
: b. Conformance to IEEE Standard 279-1971
 
The design of the control system conforms to the applicable requirements of IEEE Standard 279-1971, as
 
listed and discussed in Section 7.3.1.2c, except that
 
this system is automatically actuated. The setpoints are
 
provided in the Technical Specifications.
: c. Conformance to NRC regulatory guides
 
The applicability of regulatory guides is shown in Table
 
7.1-2. References to the discussions of these regulatory
 
guides are presented in Section 7.1.2.5.1.
: d. Failure modes and effects analysis
 
See Table 7.3-11.
: e. Periodic testing
 
Periodic testing of the mechanical equipment associated
 
with this system is discussed in Section 10.4.9.4.
 
Provisions for the periodic testing of the actuation
 
system are discussed in the Technical Specifications.
 
7.3.7  MAIN STEAM AND FEEDWATER ISOLATION
 
7.3.7.1  Description The signals that initiate automatic closure of the main steam and feedwater
 
isolation valves are generated in the ESFAS described in Section 7.3.8. The
 
logic diagrams for the generation of these signals are shown in Figure 7.2-1 (Sheets 8 and 13). The remainder of this section concentrates on the non-
 
Westinghouse portion of the main steam and feedwater isolation system.
 
7.3-25    Rev. 0 WOLF CREEK The main steam and feedwater isolation valves are operated by system medium actuators. These actuators are controlled by a series of six electric solenoid pilot valves, which direct the system fluid to either the upper piston chamber (UPC) or the lower piston chamber (LPC), or a combination thereof. The six solenoid pilot valves are divided into two trains that are independently powered and controlled. Either train can independently perform the safety function to fast close the valve.
 
7.3.7.1.1  System Description
: a. Initiating circuits
 
The main steam and feedwater isolation valves close
 
automatically upon receipt of an automatic close signal
 
from the Westinghouse solid state protection system
 
(SSPS). Manual operation is also provided.
: b. Logic
 
See drawings referenced in Section 1.7. In addition to
 
the manual and automatic trip modes of operation, manual
 
controls are provided for the slow opening or closing of each valve.
: c. Bypass
 
See Section 7.3.8.
: d. Interlocks
 
See Section 7.3.8.
: e. Redundancy
 
Two complete actuation systems are provided for each
 
valve operator. Each system is capable of closing the
 
valve as required.
: f. Diversity
 
See Section 7.3.8 for a discussion of diversity with regard to the automatic actuation signal. The valve controls provide sufficient built-in diversity within the programmable portion of the controls such that common cause failures of that programming is adequately addressed.
: g. Actuated devices
 
The actuated devices are the main steam and feedwater
 
isolation valves.
 
7.3-26    Rev. 23 WOLF CREEK
: h. Supporting systems
 
The system makes use of the Class IE dc power system.
: i. Each of the two actuation systems provide a means for opening the valve. Both actuation systems are required to open the valve. None of these provisions are required for safety.
 
7.3.7.1.2  Design Bases
 
The design bases for the main steam and feedwater isolation actuation system
 
are provided in Section 7.3.8. The design bases for the remainder of the main steam and feedwater isolation system are that the system isolates the main
 
steam and feedwater when required, and that no single failure can prevent any
 
valve from performing its required function. See Section 7.3.8 for additional
 
discussion.
 
7.3.7.1.3  Drawings
 
See Figures 7.2-1 (Sheet 8), 7.3-2, and 7.3-3. Other drawings pertaining to
 
this system are included in the introductory material for this section.
 
7.3.7.2  Analysis
: a. Conformance to NRC general design criteria
 
See Section 7.3.8.
: b. Conformance to IEEE Standard 603-1991 The design of the valve control system conforms to the
 
applicable requirements of IEEE Standard 603-1991, as listed below. The setpoints are provided in the Technical Specifications.
: 1. Single-Failure Criteria - Clause 5.1 Through use of redundant, independent systems, as previously described, any single failure or multiple failures resulting from a single credible event will not prevent the system from performing its intended function, when required.
: 2. Completion of Protective Action - Clause 5.2 The valve control system is designed so that once protective action is initiated, it is carried through to completion.
: 3. Quality - Clause 5.3 Components and modules used in the construction of the system exhibit a quality consistent with the nuclear power plant design life objective, require minimum maintenance, and have low failure rates. The program for quality assurance is described in Chapter 17.0.
: 4. Equipment Qualification - Clause 5.4 The system is qualified to perform its intended functions under the environmental conditions specified in Sections 3.10(B) and (N) and 3.11(B) and (N).
: 5. System Integrity - Clause 5.5 The control system will maintain functional capability under all conditions described in Section 7.3.1.1.2.
7.3-27 Rev. 23 WOLF CREEK
: 6. Independence - Clause 5.6 The electrical power supply, instrumentation, and control conductors for redundant circuits of the WCGS have physical separation to preserve the redundancy and to ensure that no single credible event will prevent operation of the associated function. Critical circuits and functions include power, control, and analog instrumentation associated with the operation of the safety-related systems. Events considered credible and considered in the design include the effects of short circuits, pipe rupture effects, missiles, fire and earthquakes.
Further discussions of the means to ensure channel independence are given in Section 8.3.1.4.
: 7. Capability for Testing and Calibration - Clause 5.7 The valve control system has the capability of testing the devices used to derive the final system output.
: 8. Information Displays - Clause 5.8 Sufficient information is provided to allow the station operator to make a prompt decision regarding the system operating requirements.
: 9. Control of Access - Clause 5.9 The valve control system is located in an area of the plant which is secured by the plant security system in a manner that allows only authorized personnel access.
: 10. Repair - Clause 5.10 The valve control system is designed to facilitate the recognition, location, replacement, repair, and adjustment of malfunctioning components or modules.
: 11. Identification - Clause 5.11 The valve controls system protective actions are described and identified down to the channel level.
: 12. Auxiliary Features - Clause 5.12 There are no auxiliary features within the valve control system that perform a  function not required for the valve control system to accomplish the intended safety function. Therefore, Clause 5.12 of IEEE 603-1991 is not applicable.
: 13. Multi-Unit Stations - Clause 5.13 WCGS is not a multi-unit station, therefore Clause 5.13 of IEEE 603-1991 is not applicable.
: 14. Human Factors Considerations - Clause 5.14 Human factors for the valve control system were considered at the initial stages and throughout the design process.
: 15. Reliability - Clause 5.15 The valve controls system's calculated mean time between failure is 3.28 years, which exceeds the reliability goal of two years.
: c. Conformance to NRC regulatory guides.
 
See Section 7.3.8.
 
7.3-28 Rev. 23 WOLF CREEK
: d. Failure modes and effects analysis
 
Failure mode and effects analyses have been performed on the valve control system equipment, and the results are provided in reference 5.
: e. Periodic testing
 
The valve control system includes provisions for
 
verifying the proper operation of the electronic logic circuits. The frequency of actuation system testing is provided in the Technical Specifications. The mechanical system testing provisions are given in the Technical Specifications and Section 10.3.4.
 
Note that each valve can be closed within the appropriate
 
time limit by either actuator side. Testing is
 
administratively controlled to ensure that both sides of
 
a given actuator will not be set to "TEST" mode simultaneously.
 
7.3.8  NSSS ENGINEERED SAFETY FEATURE ACTUATION SYSTEM
 
7.3.8.1  Description The Westinghouse solid state protection system (SSPS) consists of two parts:
 
the reactor trip system (RTS), which is described in Section 7.2, and the
 
engineered safety feature actuation system (ESFAS), which is described here.
The ESFAS monitors selected plant parameters and, if predetermined safety
 
limits are exceeded, transmits signals to logic matrices sensitive to
 
combinations indicative of primary or secondary system boundary ruptures (Condition III or IV events). When certain logic combinations occur, the
 
system sends actuation signals to the appropriate engineered safety feature
 
components. The ESFAS meets the requirements of GDCs 13, 20, 21, 22, 23, 24, 25, 27, 28, 34, 35, 37, 38, 40, 41, 43, 44, 46, 54, 55, and 56.
 
7.3.8.1.1  System Description
 
The equipment which provides the actuation functions is listed below and discussed in this section.  (For additional background information, see
 
References 1, 2, and 3.)
: a. Process instrumentation and control system (Ref. 1)
: b. Solid state logic protection system (Ref. 2)
: c. Engineered safety feature test cabinet (Ref. 3)
: d. Manual actuation circuits The ESFAS consists of two discrete portions of circuitry:  1) an analog portion consisting of three or four redundant channels per parameter or variable to monitor various plant parameters, such as the reactor coolant system and steam system pressures, temperatures and flows, and containment pressures; and 2) a digital portion consisting of two redundant logic trains which receive inputs from the analog protection channels and perform the logic needed to actuate the engineered safety features. Each digital train is capable of actuating the engineered safety feature equipment required. Any single failure within the engineered safety feature actuation system does not prevent system action, when required.
 
7.3-29 Rev. 23 WOLF CREEK The redundant concept is applied to both the analog and logic portions of the system. Separation of redundant analog channels begins at the process sensors
 
and is maintained in the field wiring, containment vessel penetrations, and
 
analog protection racks terminating at the redundant safeguards logic racks. 
 
The design meets the requirements of GDCs 20, 21, 22, 23, and 24.
 
The variables are sensed by the analog circuitry, as discussed in Reference 1
 
and in Section 7.2. The outputs from the analog channels are combined into
 
actuation logic, as shown in Figure 7.2-1 (Sheets 5, 6, 7, and 8). Tables 7.3-
 
13 and 7.3-14 give additional information pertaining to logic and function.
 
Analog Circuitry The process analog sensors and racks for the engineered safety feature
 
actuation system are described in Reference 1. This reference discusses the
 
parameters to be measured, including pressures, flows, tank and vessel water levels, and temperatures, as well as the measurement and signal transmission
 
considerations. These latter considerations include the transmitters, orifices
 
and flow elements, resistance temperature detectors, as well as automatic
 
calculations, signal conditioning, and location and mounting of the devices.
 
The sensors  monitoring the primary system are shown on Figure 5.1-1. The
 
secondary system sensor locations are shown on the steam system flow diagrams
 
given in Chapter 10.0.
 
Containment pressure is sensed by four physically separated differential pressure transmitters located outside of the containment (which are connected to the containment atmosphere by a filled and sealed hydraulic transmission system). The distance from penetration to transmitter is kept to a minimum, and separation is maintained. This arrangement, together with the pressure sensors external to the containment, forms a double barrier and conforms to GDC-56 and Regulatory Guide 1.11.
Digital Circuitry The engineered safety feature logic racks are discussed in detail in Reference
: 2. The description includes the considerations and provisions for physical and electrical separation, as well as details of the circuitry. Reference 2 also covers certain aspects of on-line test provision, provisions for test points, consideration for the instrument power source, and considerations for accomplishing physical separation. The outputs from the analog channels are combined into actuation logic, as shown on Sheets 5, 6, 7, 8, and 14 of Figure 7.2-1.      a. Initiating circuits
: 1. Containment pressure (see Table 7.3-14)
: 2. Steam line pressure (see Table 7.3-14)
: 3. Steam line pressure rate (see Table 7.3-14)
: 4. Manual (see Tables 7.3-13 and 14)
Manual actuation switches are provided on the main control board for the safety injection signal (SIS),              the containment isolation signal phase-A (CIS-A), and containment isolation signal phase-B/containment spray actuation signal (CIS-B/CSAS). The switches are momentary-contact and are arranged and operate as follows:
 
7.3-30    Rev. 23 WOLF CREEK (a)  SIS:  Two switches, each with two sets of contacts connected mechanically but electrically
 
isolated. One set of contacts in each switch is
 
wired to separation group 1, the other to
 
separation group 4. Operation of either switch actuates both trains of the SIS. The switch
 
wiring is in accordance with the separation
 
requirements of IEEE 279-1971.
 
(b)  CIS-A:  Two switches arranged and wired as
 
described for SIS. Operation of either switch
 
actuates both trains of the CIS-A.
 
(c)  CIS-B/CSAS:  Two sets of two switches each, each
 
switch arranged and wired as described for SIS.
Operation of both switches in either set activates both trains of both CIS-B and CSAS. 
 
Operation of any one switch, or of any two
 
switches not in the same set does not actuate
 
CIS-B/CSAS.
 
Manual controls in the control room are also provided
 
to switch from the injection to the recirculation
 
phase after a LOCA.
: b. Logic
 
The actuation logic is shown in Figure 7.2-1 (Sheets 5, 6, 7, and 8). Tables 7.3-13 and 7.3-14 give additional
 
information pertaining to the logic.
: c. Bypass
 
Bypasses are designed to meet the requirements of IEEE
 
Standard 279-1971, Sections 4.11, 4.12, 4.13, and 4.14.
 
Bypasses are provided to permit testing of the fast-close
 
logic circuitry. However, access to the bypass switches
 
is administratively controlled to prevent simultaneous
 
bypass of both actuation channels for any one valve. The
 
bypass condition is indicated in the main control room.
: d. Interlocks
 
Interlocks are also discussed in Sections 7.2, 7.6, and
 
7.7. The protection (P) interlocks are given on Tables
 
7.2-2 and 7.3-15. The safety analyses demonstrate that
 
the protective systems ensure that the NSSS would be put
 
into and maintained in a safe state following a Condition
 
II, III or IV accident commensurate with pertinent
 
criteria in the Technical Specifications. The protective
 
systems have been designed to meet IEEE Standard 279-1971
 
and are entirely redundant and separate, including all
 
permissives and blocks. All blocks of a protective
 
function are automatically cleared whenever the
 
protective function is required to function in accordance
 
with GDC-20, GDC-21, and GDC-22 and Sections 4.11, 4.12, and 4.13 of IEEE Standard 279-1971. Control interlocks
 
(C) are identified in Table 7.7-1. Because control
 
interlocks are not safety-related, they have not been
 
specifically designed to meet the requirements of IEEE Protection System Standards.
 
7.3-31    Rev. 23 WOLF CREEK The interlocks associated with NSSS engineered safety
 
feature actuation system are outlined in Table 7.3-15.
: e. Sequencing
 
The containment spray pumps start 15 seconds after a CSAS
 
with no undervoltage condition present. With an
 
undervoltage condition, 12 seconds must be added for
 
diesel startup.
: f. Redundancy
 
Redundancy for the system is provided by redundant
 
process channels which are physically and electrically separated. Redundant train logic is also provided in the SSPS, which is physically and electrically separated.
 
The process signals are combined from the process control
 
systems into the SSPS according to the prescribed logic
 
defined in Sections 7.2 and 7.3 to produce actuation
 
signals for RTS and ESFAS operations.
: g. Diversity
 
Functional diversity, as described in Reference 4, has
 
been designed into the system. The extent of diverse
 
system variables has been evaluated for postulated
 
accidents. Generally, two or more diverse protection
 
functions would automatically terminate an accident
 
before unacceptable consequences could occur.
: 1. Regarding the engineered safety feature actuation
 
system for a LOCA, a safety injection signal can be
 
obtained manually or by automatic initiation from
 
either of two diverse parameter measurements.
 
(a)  Low pressurizer pressure.
 
(b)  High containment pressure (Hi-1).
: 2. For a steam line break accident, safety injection
 
signal actuation is provided by:
 
(a)  Lead-lag compensated low steam line pressure.
 
(b)  For a steam line break inside containment, high
 
containment pressure (Hi-1) provides an additional parameter for generation of the signal.
 
7.3-32    Rev. 1 WOLF CREEK (c)  Low pressurizer pressure.
 
All of the above sets of signals are redundant and
 
physically separated and meet the requirements of IEEE Standard 279-1971.
: h. Actuated devices
 
Function Initiation The specific functions which rely on the ESFAS for
 
initiation are:
: 1. A reactor trip, provided one has not already been
 
generated by the reactor trip system.
: 2. Cold leg injection isolation valves which are opened
 
for injection of borated water by safety injection
 
pumps into the cold legs of the reactor coolant
 
system.
: 3. Charging pumps, safety injection pumps, residual heat
 
removal pumps, and associated valving which provide emergency makeup water to the cold legs of the reactor coolant system following a LOCA.
: 4. Containment air recirculation fans and cooling system
 
which serve to cool the containment and limit the
 
potential for release of fission products from the
 
containment by reducing the pressure following an
 
accident.
: 5. Those pumps which serve as part of the heat sink for
 
containment cooling (e.g., essential service water
 
and component cooling water pumps).
: 6. Motor-driven auxiliary feedwater pumps.
: 7. Phase A containment isolation, whose function is to prevent fission product release (isolation of all
 
lines not essential to reactor protection).
: 8. Steam line isolation to prevent the continuous, uncontrolled blowdown of more than one steam
 
generator and thereby uncontrolled reactor coolant
 
system cooldown (see Section 7.3.7).
: 9. Main feedwater line isolation as required to prevent
 
or mitigate the effect of excessive cooldown.
 
7.3-33    Rev. 0 WOLF CREEK
: 10. Start the emergency diesels to ensure a back-up
 
supply of power to the emergency and supporting
 
systems components.
: 11. Isolate the control room intake ducts to meet
 
control room occupancy requirements following a LOCA
 
(see Section 7.3.4).
: 12. Containment spray actuation which performs the
 
following functions:
 
(a)  Initiates containment spray to reduce
 
containment pressure and temperature following a
 
loss-of-coolant or steam line break accident inside of the containment. Initiation of containment spray causes sodium hydroxide to be
 
introduced to the spray to remove airborne
 
iodine.
 
(b)  Initiates Phase B containment isolation which
 
isolates the containment following a LOCA, or a
 
steam or feedwater line break within the
 
containment to limit radioactive releases.
Final Actuation Circuitry
 
The outputs of the solid state logic protection system
 
(the slave relays) are energized to actuate, as are most
 
final actuators and actuated devices. These devices are listed as follows:
: 1. Safety injection system pump and valve actuators.
 
See Chapter 6.0 for flow diagrams and additional
 
information.
: 2. CIS-A isolates all nonessential process lines on
 
receipt of safety injection signal. CIS-B isolates
 
the remaining process lines on receipt of a 2/4 hi-3 containment pressure signal. For further information, see Section 6.2.4.
: 3. Emergency fan coolers (see Section 6.2)
: 4. Essential service water pumps and valve actuators
 
(see Chapter 9.0)
 
7.3-34    Rev. 14 WOLF CREEK
: 5. Auxiliary feedwater pumps start (see Chapter 10.0)
: 6. Diesel start (see Chapter 8.0)
: 7. Feedwater isolation (see Chapter 10.0)
: 8. Ventilation isolation valves and damper actuator (see
 
Chapter 6.0)
: 9. Steam line isolation valve actuators (see Section
 
7.3.7 and Chapter 10.0)
: 10. Containment spray pump and valve actuators (see
 
Chapter 6.0)
If an accident is assumed to occur coincident with a loss
 
of offsite power, the engineered safety feature loads
 
must be sequenced onto the diesel generators to prevent
 
overloading them. This sequence is discussed in Chapter
 
8.0. The design meets the requirements of GDC-35.
: i. Support systems
 
The following systems are required for support of the
 
engineered safety features:
: 1. Essential service water system - heat removal (see
 
Chapter 9.0)
: 2. Component cooling water system - heat removal (see Chapter 9.0)
: 3. Electrical power distribution systems (see Chapter
 
8.0)
: 4. Essential HVAC systems (see Section 9.4)
 
Table 7.3-12 provides a list of the auxiliary support ESF
 
systems.
: j. Portion of system not required for safety
 
The system produces annunciator, status light, and
 
computer input signals to indicate individual channel
 
status. The system provides signals to the reactor trip
 
annunciators for sequence of events indication, and
 
indicates the condition of blocks and permissives. Semi-
 
automatic testing features are provided for on-line 
 
7.3-35    Rev. 0 WOLF CREEK testing. All monitoring for the testing is at the
 
protection system cabinets. Equipment used to accomplish
 
these functions is isolated from the protection functions
 
and is not required for the safety of the plant.
 
7.3.8.1.2  Design Bases
 
The functional diagrams presented in Figure 7.2-1 (Sheets 5, 6, 7, and 8)
 
provide a graphic outline of the functional logic associated with requirements
 
for the ESFAS. Requirements of the ESFS are given in Chapter 6.0. The design
 
bases information required in IEEE Standard 279-1971 is given in Sections
 
7.3.1.2c and 7.3.8.2b.
: a. Automatic actuation requirements The ESFAS receives input signals (information) from the
 
reactor plant and containment and automatically provides
 
timely and effective signals to actuate the components
 
and subsystems comprising the ESFAS.
: b. Manual actuation requirements
 
The ESFAS has provisions in the control room for manually
 
initiating the functions of the engineered safety feature
 
system
: c. Equipment protection
 
Equipment related to safe operation of the plant is designed, constructed, and installed to protect it from damage. This is accomplished by conformance to accepted
 
standards, criteria, and consideration of potential
 
environmental conditions. The criteria for equipment
 
protection are given in Chapter 3.0. As an example, certain equipment is seismically qualified in accordance
 
with IEEE Standard 344-1975. During construction, independence and separation was achieved, as required by IEEE Standard 279-1971, IEEE Standard 384-1974, and Regulatory Guide 1.75, either by barriers, physical separation, or demonstration test. This serves to protect against complete destruction of a system by
 
fires, missiles, or other hazards.
 
7.3.8.1.2.1  Generating Station Conditions
 
The following is a summary of those generating station conditions requiring
 
protective action:
 
7.3-36    Rev. 1 WOLF CREEK
: a. Primary system
: 1. Rupture in small pipes or cracks in large pipes.
: 2. Rupture of a reactor coolant pipe (LOCA).
: 3. Steam generator tube rupture.
: b. Secondary system
: 1. Minor secondary system pipe breaks resulting in steam
 
release rates equivalent to a single dump, atmospheric
 
relief, or safety valve.
: 2. Rupture of a major steam pipe.
 
7.3.8.1.2.2  Generating Station Variables
 
The following list summarizes the generating station variables required to be
 
monitored for the automatic initiation of safety injection during each accident
 
identified in the preceding section. Post-accident monitoring requirements are
 
given in Table 7.5-1.
: a. Primary system accidents
: 1. Pressurizer pressure
: 2. Containment pressure (not required for steam generator tube rupture)
: b. Secondary system accidents
: 1. Pressurizer pressure
: 2. Steam line pressures and pressure rate
: 3. Containment pressure
 
7.3.8.1.2.3  Spatially Dependent Variables
 
The only variable sensed by the ESFAS which has spatial dependence is reactor
 
coolant temperature. The effect on the measurement is negated by taking
 
multiple samples from the reactor coolant hot leg and electrically averaging these samples at the process cabinets.
 
7.3-37 Rev. 13 WOLF CREEK 7.3.8.1.2.4  Limits, Margins, and Levels
 
Operational limits, available margins, and setpoints are discussed in Chapters
 
15.0 and the WCGS Technical Specifications.
 
7.3.8.1.2.5  Abnormal Events
 
The malfunctions, accidents, or other unusual events which could physically
 
damage protection system components or could cause environmental changes are as
 
follows:
: a. LOCA (see Chapter 15.0)
: b. Steam and feedwater breaks (see Chapter 15.0)
: c. Earthquakes (see Chapters 2.0 and 3.0)
: d. Fire (see Section 9.5.1)
: e. Missiles (see Section 3.5)
: f. Flood (see Chapters 2.0 and 3.0)
 
7.3.8.1.2.6  Minimum Performance Requirements
 
Minimum performance requirements are as follows:
: a. System response times
 
The ESFAS response time is defined as the interval required for the ESF sequence to be initiated subsequent to the time that the appropriate variable(s) exceed this setpoint(s). The ESF sequence is
 
initiated by the output of the ESFAS, which is by the operation of the
 
dry contacts of the slave relays (600 and 700 series relays) in the
 
output cabinets of the solid state protection system. The response
 
times listed below include the interval of time which will elapse
 
between the time the parameter as sensed by the sensor exceeds the
 
safety setpoint and the time the solid state protection system slave
 
relay dry contacts are operated. These values (as listed below) are
 
maximum allowable values consistent with the safety analyses and the
 
Technical Specification Bases and were systematically verified during plant preoperational startup tests. For the overall ESF response time, refer to the Technical Specification Bases. In a similar manner for the overall reactor trip system instrumentation response time, refer to the Technical Specification Bases. These maximum delay times include all compensation and, therefore, require that any such network
 
be aligned and operating during verification testing.
 
7.3-38 Rev. 13 WOLF CREEK
 
The ESFAS is always capable of having response time tests
 
performed, using the same method as those tests performed
 
during the preoperational test program or following significant component changes.
 
Maximum allowable time delays in generating the actuation
 
signal for loss-of-coolant protection are:
: 1. Pressurizer pressure              2.0 seconds
 
Maximum allowable time delays in generating the actuation
 
signal for steam line break protection are:
: 1. Steam line pressure              2.0 seconds
: 2. Steam line pressure rate          2.0 seconds
: 3. High containment pressure
 
for closing main steam
 
line stop valves                  2.0 seconds
: 4. Actuation signals for auxiliary
 
feedwater pumps                  2.0 seconds
: b. System Accuracies
 
Accuracies required for generating the required actuation
 
signals for loss-of-coolant protection are:
: 1. Pressurizer pressure (uncompensated)                    18 psi          Accuracies required in generating the required actuation
 
signals for steam line break protection are:
: 1. Steam line pressure                4 percent of span
: 2. Containment pressure signal        1.8 percent of full scale
: c. Ranges of sensed variables to be accommodated until
 
conclusion of protective action is ensured
 
7.3-39 Rev. 13 WOLF CREEK Ranges required in generating the required actuation
 
signals for loss-of-coolant protection are:
: 1. Pressurizer pressure              1,700 to 2,500 psig
: 2. Containment pressure              0 to 60 psig
 
Ranges required in generating the required actuation
 
signals for steam line break protection are:
: 1. T avg                              530 to 630 F
: 2. Steam line pressure              0 to 1,300 psig
: 3. Containment pressure              0 to 60 psig
 
7.3.8.1.2.7  Bistable Trip Setpoints
 
There are three values applicable to engineered safety feature actuation:
: a. Safety analysis limit
: b. Allowable value
: c. Nominal trip setpoint
 
The safety analysis limit is the value assumed in the accident analysis.
 
The allowable value is in the Technical Specifications and is obtained by adding or subtracting a calculated allowance from the nominal trip setpoint.
This calculated allowance accounts for instrument error, process uncertainties such as flow stratification and transport factor effects, etc.
 
The nominal trip setpoint is the value set into the equipment and is obtained by adding or subtracting allowances for instrument drift, rack calibration accuracy, and rack comparator setting accuracy from the safety anaylis limit.
The nominal trip setpoint allows for the normal expected instrument safety anaylis limit drift, such that the Technical Specification limits are not exceeded under normal operation.
 
The setpoints that require trip action are given in the Technical
 
Specifications. A further discussion on setpoints is found in Section
 
7.2.2.2.1.
 
As described above, allowance is then made for process uncertainties, instrument error, instrument drift, and calibration uncertainty to obtain the
 
nominal trip setpoint which is actually set into the equipment. The only requirement on the instrument's accuracy value is that over the instrument span the error must always be 
 
7.3-40 Rev. 13 WOLF CREEK less than or equal to the error value allowed in the accident analysis. The
 
instrument does not need to be the most accurate at the setpoint value as long
 
as it meets the minimum accuracy requirement. The accident analysis accounts
 
for the expected errors at the actual setpoint.
 
Range selection for the instrumentation covers the expected range of the
 
process variable being monitored, consistent with its application. The design
 
of the reactor protection and engineered safety features systems is such that the bistable trip setpoints do not require process transmitters to operate within 5 percent of the high and low end of their calibrated span or range. 
 
Functional requirements established for every channel in the reactor protection
 
and engineered safety feature systems stipulate the maximum allowable errors on
 
accuracy, linearity, and reproducibility. The protection channels have the capability for and are tested to ascertain that the characteristics throughout
 
the entire span, in all aspects, are acceptable and meet functional requirement
 
specifications. As a result, no protection channel operates normally within 5
 
percent of the limits of its specified span.
 
The specific functional requirements for response time, setpoint, and operating
 
span are based on the results and evaluation of safety studies to be carried
 
out using data pertinent to the plant. This establishes adequate performance requirements under both normal and faulted conditions, including consideration of process transmitters margins such that even under a highly improbable
 
situation of full power operation at the limits of the operating map [as
 
defined by the high and low pressure reactor trip, DT overpower and
 
overtemperature trip lines (DNB protection), and the steam generator safety valve pressure setpoint] adequate instrument response is available to ensure
 
plant safety.
 
7.3.8.1.3  Final System Drawings
 
The schematic diagrams for the systems discussed in this section are listed in
 
Section 1.7.
 
7.3.8.2  Analysis
: a. Conformance to GDCs
 
Conformance to GDCs is described in Section 7.1
: b. Conformance to IEEE 279-1971
: 1. Single Failure Criteria
 
7.3-41    Rev. 1 WOLF CREEK The discussion presented in Section 7.2.2.2.3 is
 
applicable to the engineered safety feature actuation
 
system, with the following exception.
 
In the engineered safety feature, a loss of
 
instrument power calls for actuation of engineered
 
safety feature equipment controlled by the specific
 
bistable that lost power (containment spray
 
exempted). The actuated equipment must have power to
 
comply. The power supply for the protection system
 
is discussed in Section 7.6 and Chapter 8.0. For
 
containment spray, the final bistables are energized
 
to trip to avoid spurious actuation. In addition, manual containment spray requires a simultaneous actuation of two manual controls. This is considered acceptable because spray actuation on hi-hi
 
containment pressure signal provides automatic initiation of the system via protection channels meeting the criteria in Reference 3. Moreover, two sets (two switches per set) of containment spray
 
manual initiation switches are provided to meet the
 
requirements of IEEE Standard 279-1971. Also it is
 
possible for all engineered safety feature equipment (valves, pumps, etc.) to be individually manually
 
actuated from the control board. Hence, a third mode
 
of containment spray initiation is available. The
 
design meets the requirements of GDCs 21 and 23.
: 2. Equipment Qualification
 
Equipment qualifications are discussed in Sections
 
3.10(N) and 3.11(N).
: 3. Channel Independence
 
The discussion presented in Section 7.2.2.2.3 is
 
applicable. The engineered safety feature slave
 
relay outputs from the solid state logic protection
 
cabinets are redundant, and the actuation signals
 
associated with each train are energized up to and
 
including the final actuators by the separate ac
 
power supplied which powers the logic trains.
: 4. Control and Protection System Interaction
 
The discussions presented in Section 7.2.2.2.3 are
 
applicable.
 
7.3-42    Rev. 1 WOLF CREEK
: 5. Capability for Sensor Checks and Equipment Test and
 
Calibration
 
The discussions of system testability in Section 7.2.2.2.3 are applicable to the sensors, analog
 
circuitry, and logic trains of the ESFAS.
 
The following discussions cover those areas in which
 
the testing provisions differ from those for the
 
reactor trip system.
 
Testing of ESFAS To facilitate engineered safety feature actuation
 
testing, four cabinets (two per train) are provided
 
which enable operation, to the maximum practical extent, of safety feature loads on a group-by-group
 
basis until actuation of all devices has been
 
checked.
 
The testing program meets the requirements of GDCs
 
21, 37, 40, and 43 and Regulatory Guide 1.22, as
 
discussed in Section 7.1.2.5.2. The tests described
 
in item 3 above and further discussed in Section
 
6.3.4 meet the requirements on testing of the
 
emergency core cooling system, as stated in GDC-37,              except for the operation of those components that will cause an actual safety injection. The test, as
 
described, demonstrates the performance of the full
 
operational sequence that brings the system into
 
operation, the transfer between normal and emergency
 
power sources, and the operation of associated
 
cooling water systems. The safety injection and
 
residual heat removal pumps are started and operated
 
and their performance verified in a separate test
 
described in Section 6.3.4. When the pump tests are
 
considered in conjunction with the emergency core
 
cooling system test, the requirements of GDC-37 on
 
testing of the emergency core cooling system are met
 
as closely as possible without causing an actual
 
safety injection.
The system design, as described in Sections 6.3 and
 
7.2.2.3 item 3 above, provides completed periodic
 
testability during reactor operation of all logic and
 
components associated with the emergency core cooling
 
system. This design meets the requirements of
 
Regulatory Guide 1.22, as discussed in the above
 
sections. The testing capability is as follows:
 
7.3-43    Rev. 0 WOLF CREEK (a)  Prior to initial plant operations, ESFAS tests
 
were conducted.
 
(b)  Subsequent to initial startup, ESFAS tests are
 
conducted during each regularly scheduled
 
refueling outage.
 
(c)  During on-line operation of the reactor, all of
 
the engineered safety feature analog and logic
 
circuitry can be fully tested. In addition, essentially all of the engineered safety feature
 
final actuators can be fully tested. The
 
remaining few final actuators whose operation is
 
not compatible with continued on-line plant
 
operation can be checked by means of continuity
 
testing or a series of overlapping tests as discussed below.
 
(d)  During normal operation, the operability of
 
testable final actuation devices of the ESFS can
 
be tested by manual initiation from the control
 
room.
 
Performance Test Acceptability Standard for the SIS and for the Automatic Demand Signal for CSAS Generation
 
During reactor operation, the basis for ESFAS
 
acceptability is the successful completion of the
 
overlapping tests performed on the initiating system
 
and the engineered safety feature actuation system
 
(see Figure 7.3-2). Checks of process indications
 
verify operability of the sensors. Analog checks and
 
tests verify the operability of the analog circuitry
 
from the input of these circuits through to and
 
including the logic input relays, except for the
 
input relays associated with the containment spray
 
function which are tested during the solid state
 
logic testing. Solid state logic testing also checks
 
the digital signal path from and including logic
 
input relay contacts through the logic matrices and
 
master relays and performs continuity tests on the
 
coils of the output slave relays. Final actuator
 
testing operates the output slave relays and verifies
 
the operability of those devices which require
 
safeguards actuation and which can be tested without
 
causing plant upset. A continuity check is performed
 
on the actuators of the untestable devices.
 
Operation of the final devices is confirmed by 
 
control board indication and visual observation that 
 
the appropriate pump breakers close and automatic 
 
valves have completed their travel.
 
7.3-44  Rev. 27 WOLF CREEK The basis for acceptability for the engineered safety
 
feature interlocks is control board indication of
 
proper receipt of the signal upon introducing the
 
required input at the appropriate setpoint.
 
Maintenance checks (performed during regularly
 
scheduled refueling outages), such as resistance to
 
ground of signal cables in radiation environments, are based on qualification test data which identifies
 
what constitutes acceptable radiation, thermal, etc.,
degradation.
 
Frequency of Performance of Engineered Safety Feature Actuation Tests During reactor operation, complete system testing
 
(excluding sensors or those devices whose operation
 
would cause plant upset) is performed periodically,              as specified in the Technical Specifications.
 
Testing, including the sensors, is also performed
 
during scheduled plant shutdown for refueling. See
 
the Technical Specifications for frequency of
 
testing.
 
Engineered Safety Feature Actuation Test Description The following sections describe the testing circuitry
 
and procedures for the on-line portion of the testing
 
program. The guidelines used in developing the circuitry and procedures are:
 
(a)  The test procedures must not involve the
 
potential for damage to any plant equipment.
 
(b)  The test procedures must minimize the potential
 
for accidental tripping.
 
c)  The provisions for on-line testing must minimize
 
complication of engineered safety feature actuation circuits so that their reliability is not degraded.
 
Description of Initiation Circuitry Several systems comprise the total engineered safety
 
feature system, the majority of which may be
 
initiated by different process conditions and be reset independently of each other.
 
7.3-45    Rev. 0 WOLF CREEK The remaining functions (listed in item h of Section
 
7.3.8.1.1) are initiated by a common signal (safety
 
injection) which in turn may be generated by
 
different process conditions.
 
In addition, operation of all other vital auxiliary
 
support systems, such as auxiliary feedwater, component cooling, and essential service water, is
 
initiated by the safety injection signal.
 
Each function is actuated by a logic circuit which is
 
separated between each of the two redundant trains of
 
the engineered safety feature initiation circuits.
 
The output of each of the initiation circuits consists of a master relay which drives slave relays
 
for contact multiplication as required. The logic, master, and slave relays are mounted in the solid
 
state logic protection cabinets designated train A
 
and train B, respectively, for the redundant
 
counterparts. The master and slave relay circuits
 
operate various pump and fan circuit breakers or
 
starters, motor-operated valve contactors, solenoid-
 
operated valves, emergency generator starting, etc.
 
Analog Testing Analog testing is identical to that used for reactor
 
trip circuitry and is described in Section 7.2.2.2.3.
 
An exception to this is containment spray, which is energized to actuate 2/4 and reverts to 2/3 when one
 
channel is in test.
 
Solid State Logic Testing Except for containment spray channels, solid state
 
logic testing is the same as that discussed in
 
Section 7.2.2.2.3. During logic testing of one train, the other train can initiate the required
 
engineered safety feature function. For additional
 
details, see Reference 2.
 
Actuator Testing At this point, testing of the initiation circuits
 
through operation of the master relay and its
 
contacts to the coils of the slave relays has been accomplished. The engineered safety feature logic slave relays  in the solid  state  protection system output cabinets are subjected to coil continuity 
 
7.3-46    Rev. 1 WOLF CREEK tests by the output relay tester in the solid state
 
protection system cabinets. Slave relays (K601, K602, etc.) do not operate because of reduced voltage
 
applied to their coils by the mode selector switch (TEST/OPERATE). A multiple position master relay
 
selector switch selects the master relays and
 
corresponding slave relays to which the coil
 
continuity test voltage is applied. The master relay
 
selector switch is returned to OFF before the mode
 
selector switch is placed back in the OPERATE mode.
 
However, failure to do so will not result in defeat
 
of the protective function. The engineered safety
 
feature actuation system slave relays are activated
 
during the testing by the on-line test cabinet, so that overlap testing is maintained.
 
The engineered safety feature actuation system final
 
actuation device or actuated equipment testing is
 
performed from the solid state protection test
 
cabinets. These cabinets are located near the solid
 
state logic protection system equipment. There is
 
one set of test cabinets provided for each of the two
 
protection trains, A and B. Each set of cabinets
 
contains individual test switches necessary to
 
actuate the slave relays. To prevent accidental
 
actuation, test switches are of the type that must be
 
rotated and then depressed to operate the slave
 
relays. Assignments of contacts of the slave relays
 
for actuation of various final devices or actuators have been made such that groups of devices or actuated equipment can be operated individually
 
during plant operation without causing plant upset or
 
equipment damage. In the unlikely event that a
 
safety injection signal is initiated during the test
 
of the final device that is actuated by this test, the device is already in its proper position to
 
perform its safety function.
 
During this last procedure, close communication is
 
maintained between the main control room operator and
 
the tester at the test cabinet. Prior to the
 
energizing of a slave relay, the operator in the main
 
control room assures that plant conditions will permit operation of the equipment that is actuated by the relay. After the tester has energized the slave relay, the main control room operator observes that
 
all equipment has operated, as indicated by 
 
appropriate indicating lamps, monitor lamps, and 
 
annunciators of the control board, and records all 
 
7.3-47    Rev. 1 WOLF CREEK operations. He then resets all devices and prepares
 
for operation of the next slave relay actuated
 
equipment.
 
By means of the procedure outlined above, all
 
engineered safety feature devices actuated by
 
engineered safety feature actuation systems
 
initiation circuits, with the exceptions noted in
 
Section 7.1.2.5.2 under a discussion of Regulatory
 
Guide 1.22, are operated by the automatic circuitry.
 
The ESFAS slave relays (Train A and B) which initiate Turbine Trip are tested in a series of overlapping tests through the Ovation Turbine Control System (TCS) Testable Dump Manifold (TDM) solenoid valves. Slave relay contacts are wired to separate digital input modules in the TCS. Redundant two-out-of-three trip paths through the ETS and OA/OPC controllers ensure that a single failure will not cause a loss of trip function or result in a spurious trip. Each controller receives inputs from a Train A and B slave relay. In the ETS controller the third input for the two-out-of-three trip logic is provided by an additional Train A contact. An additional Train B contact is provided to the OA/OPC controller.
The slave relay test is initiated from a test switch provided in the Solid State Protection System (SSPS) Test Cabinet. Each train is tested separately to prevent an actual turbine trip. Prior to initiating the slave relay test, a single input channel from the train under test is blocked using Ovation Maintenance/OOS logic. Only one Ovation input channel can be blocked at a time. During the Train A slave relay test, a single ETS controller Train A input channel is blocked. During the Train B slave relay test, a single OA/OPC controller Train B input is blocked. Indication of the channel bypass status is provided in the main control room via Ovation alarm and display graphics indication. Slave relay contact operation is verified by the SOE input module as indicated on Ovation display graphics.
Turbine trip actuation devices are tested using Ovation TDM test logic. Each Ovation controller has an associated solenoid-operated valve TDM. Each manifold operates on a two-out-of-three coincidence voting logic so that a single failure will not cause a loss of trip function or result in a spurious trip. This configuration permits online testing of individual output relays and TDM solenoid valves without an actual turbine trip. Testing of the turbine trip function under power operation is discussed in Section 10.2.3.6.
 
Actuator Blocking and Continuity Test Circuits
 
Those few final actuation devices that cannot be
 
designed to be actuated during plant operation
 
(discussed in Section 7.1.2.5.2) have been assigned
 
to slave relays for which additional test circuitry
 
has been provided to individually block actuation of
 
a final device upon operation of the associated slave
 
relay during testing. Operation of these slave
 
relays, including contact operations, and continuity
 
7.3-48  Rev. 27 WOLF CREEK of the electrical circuits associated with the final devices control are checked in lieu of actual operation. The circuits provide for monitoring of the slave relay contacts, the devices' control circuit cabling, control voltage, and the devices' actuation solenoids. Interlocking prevents blocking the output from more than one output relay in a protection train at a time. Interlocking between trains is also provided to prevent continuity testing in both trains simultaneously. Therefore, the redundant device associated with the protection train not under test is available in the event protection action is required. If an accident occurs during testing, the automatic actuation circuitry will override testing, as noted above. One exception to this is that if the accident occurs while testing a slave relay whose output must be blocked, those few final actuation devices associated with this slave relay are not actuated; however, the redundant devices in the other train would be operational and would perform the required safety function.
Actuation devices to be blocked are identified in Section 7.1.2.5.2.
The continuity test circuits for these components that cannot be actuated on-line are verified by proving lights on the safeguards test racks.
The typical schemes for blocking operation of
 
selected protection function actuator circuits are
 
shown in Figure 7.3-3 as details A and B. The
 
schemes operate as explained below and are duplicated
 
for each safeguards train.
 
Detail A shows the circuit for contact closure for
 
protection function actuation. Under normal plant
 
operation and equipment not under test, the test
 
lamps "DS*" for various circuits are energized.
 
Typical circuit path is through the normally closed
 
test relay contact "K8*" and through test lamp
 
connections 1 to 3. Coils "Xl" and "X2" are capable
 
of being energized for protection function actuation
 
upon closure of solid state logic output relay
 
contacts "K*."  Coil "Xl" is typical for a motor
 
control center starter coil.  "X2" is typical for a
 
breaker closing auxiliary coil, motor starter master
 
coil, coil of a solenoid valve, auxiliary relay, etc. When the contacts "K8*" are opened to block
 
energizing of coil "Xl" or "X2," the white lamp is
 
deenergized, and the slave relay "K*" may be 
 
energized to perform continuity testing. The 
 
operability of the blocking relay in both blocking
 
and restoring normal service can be verified by
 
opening the blocking relay contact in series with
 
lamp terminal 1, which deenergizes the test lamp, and
 
by closing the  blocking relay contact in series with
 
lamp terminal 1, which energizes the test lamp and
 
verifies that the circuit is now in its normal, i.e.,
operable condition.
 
7.3-49  Rev. 27 WOLF CREEK Detail B shows the circuit for contact opening for protection function actuation. Under normal plant operation, and equipment not under test for 125-volt dc actuation devices, the white test lamps "DS*" for the various circuits are energized, and the green test lamp "DS*" is deenergized. Typical circuit path for white lamp "DS*" is through the normally closed solid state logic output relay contact "K*" and through test lamp connections 3 to 1. Coil "Y2" is capable of being deenergized for protection function actuation upon opening of solid state logic output relay contact "K*."  Coil "Y2" is typical for a solenoid valve coil, auxiliary relay, etc. When the contact "K8*" is closed to block deenergizing of coil "Y2," the green test lamp is energized, and the slave relay "K*" may be energized to verify operation (opening of its contacts). To verify operability of the blocking relay in both blocking and restoring normal service, close the blocking relay contact to
 
the green lamp - the green test lamp should now be
 
energized also; open this blocking relay contact -
 
the green test lamp should be deenergized, which
 
verifies that the circuit is now in its normal, i.e.,
operable position.
 
Time Required for Testing
 
It is estimated that analog testing can be performed
 
at a rate of several channels per hour. Logic
 
testing of both trains A and B can be performed in
 
less than 30 minutes. Testing of actuated components
 
(including those which can only be partially tested)
 
is a function of control room operator availability.
 
It is expected to require several shifts to
 
accomplish these tests. During this procedure, automatic actuation circuitry will override testing, except for those few devices associated with a single
 
slave relay whose outputs must be blocked and then
 
only while blocked. It is anticipated that
 
continuity testing associated with a blocked slave
 
relay could take several minutes. During this time, the redundant devices in the other train would be
 
functional.
 
Summary of On-Line Testing Capabilities
 
The procedures described provide capability for
 
checking completely from the process signal to the
 
logic cabinets and from there to the individual pump
 
and fan circuit breakers or starters, valve
 
contactors, pilot solenoid valves, etc., including
 
all field cabling actually used in the circuitry
 
called upon to operate for an accident condition.
 
For those few devices whose operation could adversely
 
affect plant or equipment operation, the same
 
procedure provides for checking from the process
 
signal to the logic rack. To check the final
 
actuation device, a continuity test of the individual
 
control circuits is performed.
 
7.3-50  Rev. 27 WOLF CREEK The procedures require testing at various locations.
(a)  Analog testing and verification of bistable setpoint are accomplished at process analog racks. Verification of bistable relay operation is done at the main control room status lights.
(b)  Logic testing through operation of the master
 
relays and low voltage application to slave
 
relays is done at the solid state protection
 
system logic rack test panel.
 
(c)  Testing of pumps, fans, and valves is done at
 
the test panel located in the vicinity of the
 
solid state protection system logic racks in
 
combination with the control room operator.
 
(d)  Continuity testing for those circuits that
 
cannot be operated is done at the same test
 
panel mentioned in item c above. The ESFAS Turbine Trip function is tested in a series of overlapping tests through the Ovation Turbine Control System (TCS) Testable Dump Manifold (TDM) solenoid valves as discussed above.
 
The reactor coolant pump essential service isolation
 
valves consist of the isolation valves for the
 
component cooling water return and the seal water
 
return header.
 
The main reason for not testing these valves
 
periodically is that the reactor coolant pumps may be
 
damaged. Although pump damage from this type of test
 
would not result in a situation which endangers the
 
health and safety of the public, it could result in
 
unnecessary shutdown of the reactor for an extended
 
period of time while the reactor coolant pump or
 
certain of its parts are replaced.
 
Testing During Shutdown
 
Emergency core cooling system tests are performed
 
periodically as stated in the Technical
 
Specifications, with the reactor coolant system
 
isolated from the emergency core cooling system by
 
closing the appropriate valves. A test safety
 
injection signal is then applied to initiate
 
operation of active components (pumps and valves) of
 
the emergency core cooling system. This is in
 
compliance with GDC-37.
 
Containment spray system tests are performed at each
 
major fuel reloading. The tests are performed with
 
the isolation valves in the spray supply lines at the
 
containment and spray additive tank blocked closed
 
and are initiated by tripping the normal actuation
 
instrumentation.
 
Periodic Maintenance Inspections
 
Preventive maintenance procedures have been developed 
 
and are performed based on service conditions and 
 
experience with comparable equipment. The frequency 
 
of performance depends on the operating conditions 
 
7.3-51  Rev. 27 WOLF CREEK and requirements of the reactor power plant. The scope and frequency of preventive maintenance procedures will be revised and updated as experience is gained with the equipment. 
 
The balance  of the  requirements listed  in IEEE
 
279-1971 (Sections 4.11 through 4.22) is discussed in
 
Section 7.2.2.2.3. Section 4.20 receives special
 
attention in Section 7.5.
: 6. Manual Resets and Blocking Features
 
The manual reset feature associated with containment
 
spray actuation is provided in the standard design of
 
the Westinghouse solid state protection system design
 
for two basic purposes:  first, the feature permits
 
the operator to start an interruption procedure of
 
automatic containment spray in the event of false
 
initiation of an actuate signal; second, although spray system performance is automatic, the reset feature enables the operator to start a manual
 
takeover of the system to handle unexpected events
 
which can be better dealt with by operator appraisal
 
of changing conditions following an accident.
 
Manual control of the spray system does not occur
 
once actuation has begun by just resetting the
 
associated logic devices alone. Components will seal
 
in (latch) so that removal of the actuate signal, in
 
itself, will neither cancel or prevent completion of
 
protective action, nor provide the operator with
 
manual override of the automatic system by this
 
single action. In order to take complete control of
 
the system to interrupt its automatic performance,              the operator must deliberately unlatch relays which have "sealed in" the initial actuate signals in the
 
associated motor control center, in addition to
 
tripping the pump motor circuit breakers, if stopping
 
the pumps is desirable or necessary.
 
The manual reset feature associated with containment
 
spray, therefore, does not perform a bypass. It is
 
merely the first of several manual operations
 
required to take control from the automatic system or
 
interrupt its completion should such an action be
 
considered necessary.
 
7.3-52    Rev. 1 WOLF CREEK In the event that the operator anticipates system
 
actuation and erroneously concludes that it is
 
undesirable or unnecessary and imposes a standing
 
reset condition in one train (by operating and holding the corresponding reset switch at the time
 
the initiate signal is transmitted) the other train
 
will automatically carry the protective action to
 
completion. In the event that the reset condition is
 
imposed simultaneously in both trains at the time the
 
initiate signals are generated, the automatic
 
sequential completion of system action interrupted, and control has been taken by the operator. Manual
 
takeover is maintained, even though the reset
 
switches are released, if the original initiate signal exists. Should the initiate signal then clear and return again, automatic system actuation will
 
repeat.
 
Note also that any time delays imposed on the system
 
action are to be applied after the initiating signals
 
are latched. Delay of actuate signals for fluid systems lineup, load sequencing, etc., does not provide the operator time to interrupt automatic completion, with manual reset alone, as would be the
 
case if time delay were imposed prior to sealing of
 
the initial actuate signal.
 
The manual block features associated with pressurizer
 
and steam line safety injection signals provide the
 
operator with the means to block initiation of safety
 
injection during plant startup. These block features meet the requirements of Section 4.12 of IEEE Standard 279-1971 in that automatic removal of the block occurs when plant conditions require the
 
protection system to be functional.
: 7. Manual Initiation of Protective Actions
 
There are four individual main steam isolation valve
 
momentary control switches (one per steam line)
 
mounted on the control board. Each switch, when
 
actuated, will isolate one of the main steam lines.
 
In addition, there are two system level switches.
 
Each switch actuates  all four main steam line
 
isolation and bypass valves. Automatic initiation of
 
switchover to recirculation with manual completion is
 
in compliance with Section 4.17 of IEEE Standard 279-1971, with the following comment.
 
7.3-53    Rev. 1 WOLF CREEK Manual initiation of either one of two redundant
 
safety injection actuation main control board mounted
 
switches provides for actuation of the components required for reactor protection and mitigation of
 
adverse consequences of the postulated accident, including delayed actuation of sequenced started emergency electrical loads as well as components providing switchover from the safety injection mode
 
to the cold leg recirculation mode following a loss
 
of primary coolant accident. Therefore, once safety
 
injection is initiated, those components of the
 
emergency core cooling system (see Section 6.3) which are realigned as part of the semiautomatic switchover
 
go to completion on low refueling storage tank water
 
level without any manual action.
 
Manual operation of other components or manual verification of proper position as part of the
 
emergency procedures are not precluded nor otherwise
 
in conflict with the above-described compliance to
 
Section 4.17 of IEEE Standard 279-1971 of the semiautomatic switchover circuits.
No exception to the requirements of IEEE Standard
 
279-1971 has been taken in the manual initiation
 
circuit of safety injection. Although Section 4.17
 
of  IEEE Standard  279-1971  requires  that a single failure within common portions of the protective
 
system shall not defeat the protective action by
 
manual or automatic means, the standard does not
 
specifically preclude the sharing of initiated
 
circuitry logic between automatic and manual
 
functions. It is true that the manual safety
 
injection initiation functions associated with one
 
actuation train (e.g., train A) share portions of the
 
automatic initiation circuitry logic of the same
 
logic train; however, a single failure in shared functions does not defeat the protective action of the redundant actuation train (e.g., train B). A
 
single failure in shared functions does not defeat
 
the protective action of the safety function. It is
 
further noted that the sharing of the logic by manual
 
and  automatic initiation is consistent with the
 
system level action requirements of the IEEE Standard
 
279-1971, Section 4.17 and consistent with the
 
minimization of complexity.
 
7.3-54    Rev. 1 WOLF CREEK
: c. Conformance to regulatory guides and associated IEEE
 
standards
 
Conformance to regulatory guides and associated IEEE standards is provided in Sections 7.1.2.5 and 7.1.2.6.
: d. Failure mode and effects analyses
 
Failure mode and effects analyses have been performed on
 
the engineered safety feature systems' equipment, and the results are provided in Reference 3. The interface
 
criteria provided in Appendices B and C of Reference 3
 
have been met in the WCGS design.
 
In addition to the consideration given in this reference
 
a loss of instrument air or loss of component cooling
 
water to vital equipment has been considered. Neither
 
the loss of instrument air nor the loss of cooling water
 
(assuming no other accident conditions) can cause safety
 
limits, as given in the Technical Specifications, to be exceeded. Likewise, loss of either of the two do not adversely affect the core or the reactor coolant system
 
nor prevent an orderly shutdown if this is necessary.
 
Furthermore, all pneumatically operated valves and
 
controls assume a preferred operating position upon loss
 
of instrument air. It is also noted that, for
 
conservatism during the accident analysis (Chapter 15.0),
credit is not taken for the instrument air systems nor
 
for any control system benefit.
 
The design does not provide any circuitry which directly
 
trips the reactor coolant pumps on a loss of component
 
cooling water. Normally, indication in the control room
 
is provided whenever component cooling water is lost.
 
The reactor coolant pumps can run 10 minutes after a loss of component cooling water. This provides adequate time for the operator to correct the problem or trip the
 
plant, if necessary.
 
The initiation and operation of the auxiliary feedwater
 
system are described in Section 7.3.6.
: e. Periodic testing
 
Periodic testing is described in Section 7.3.8.2b.
 
Testing frequency is provided in the Technical
 
Specifications.
 
7.3-55    Rev. 1 WOLF CREEK 7.3.8.3  Summary
 
The effectiveness of the engineered safety feature actuation system is
 
evaluated in Chapter 15.0, based on the ability of the system to contain the effects of Condition III and IV events, including loss-of-coolant and steam
 
line break accidents. The engineered safety feature actuation system
 
parameters are based upon the component performance specifications which are
 
given by the manufacturer or verified by test for each component. Appropriate
 
factors to account for uncertainties in the data are factored into the
 
constants characterizing the system.
 
The ESFAS must detect Condition III and IV events and generate signals which
 
actuate the engineered safety features. The system must sense the accident
 
condition and generate the signal actuating the protection function reliably and within a time determined by and consistent with the accident analyses in Chapter 15.0.
 
Much longer times are associated with the actuation of the mechanical and fluid
 
system equipment associated with engineered safety features. This includes the
 
time required for switching, bringing pumps and other equipment to speed, and
 
the time required for them to take load.
 
Operating procedures require that the complete engineered safety feature
 
actuation system normally be operable. However, redundancy of system
 
components is such that the system operability assumed for the safety analyses
 
can still be met with certain instrumentation channels out of service. 
 
Channels that are out of service are to be placed in the tripped mode or bypass
 
mode in the case of containment spray.
 
7.3.8.3.1  Loss-of-Coolant Protection
 
By analysis of LOCA and in system tests, it has been verified that, except for
 
very small coolant system breaks which can be protected against by the charging
 
pumps followed by an orderly shutdown, the effects of various LOCAs are
 
reliably detected by the low pressurizer pressure signal; the emergency core cooling system is actuated in time to prevent or limit core damage.
 
For large coolant system breaks, the passive accumulators inject first, because
 
of the rapid pressure drop. This protects the reactor during the unavoidable
 
delay associated with actuating the active emergency core cooling system phase.
 
High containment pressure also actuates the emergency core cooling system. 
 
Therefore, emergency core cooling actuation can be brought about by sensing
 
this other direct consequence of a primary system break; that is, the
 
engineered safety feature actuation system detects the leakage of the coolant
 
into the 
 
7.3-56    Rev. 1 WOLF CREEK containment. The generation time of the actuation signal of about 1.5 seconds, after detection of the consequences of the accident, is adequate.
 
Containment spray provides additional emergency cooling of containment and also limits fission product release upon sensing elevated containment pressure (Hi-
: 3) to mitigate the effects of a LOCA.
 
The delay time between detection of the accident condition and the generation
 
of the actuation signal for these systems is assumed to be about 1.0 second, well within the capability of the protection system equipment. However, this
 
time is short compared to that required for startup of the fluid systems.
 
The analyses in Chapter 15.0 show that the diverse methods of detecting the
 
accident condition and the time for generation of the signals by the protection systems are adequate to provide reliable and timely protection against the effects of loss of coolant.
 
7.3.8.3.2  Steam Line Break Protection
 
The emergency core cooling system is also actuated in order to protect against
 
a steam line break. About 2.0 seconds elapse between sensing low steam line
 
pressure (as well as high steam pressure rate) and generation of the actuation
 
signal. Analysis of steam line break accidents, assuming this delay for signal
 
generation, shows that the emergency core cooling system is actuated for a
 
steam line break in time to limit or prevent further core damage for steam line
 
break cases.
 
Additional protection against the effects of steam line break is provided by
 
feedwater isolation which occurs upon actuation of the emergency core cooling system. Feedwater line isolation is initiated in order to prevent excessive cooldown of the reactor vessel, protect the reactor coolant system boundary, and limit the containment pressure.
 
Additional protection against a steam line break accident is provided by
 
closure of all steam line isolation valves in order to prevent uncontrolled
 
blowdown of all steam generators. The generation of the protection system
 
signal (about 2.0 seconds) is again short, compared to the time to trip the
 
fast-acting steam line isolation valves which are designed to close in less
 
than 5 seconds against the flows associated with line breaks on either side of the valve, assuming the most limiting normal operating conditions prior to the occurrence of the break.
 
7.3-57 Rev. 24 WOLF CREEK 7.3-58  In addition to actuation of the engineered safety features, the effect of a steam line break accident also generates a signal resulting in a reactor trip
 
on overpower or following emergency core cooling system actuation. However, the core reactivity is further reduced by the highly borated water injected by the emergency core cooling system.
 
The analyses in Chapter 15.0 of the steam line break accidents and an
 
evaluation of the protection system instrumentation and channel design show
 
that the engineered safety feature actuation systems are effective in
 
preventing or mitigating the effects of a steam line break accident.
 
7.
 
==3.9  REFERENCES==
: 1. Reid, J. B., "Process Instrumentation for Westinghouse Nuclear Steam Supply System (4 Loop Plant Using WCID 7300 Series Process Instrumentation)," WCAP-7913, January 1973.
 
(Additional background information only)
: 2. Katz, D. N., "Solid State Logic Protection System
 
Description," WCAP-7488-L (Proprietary), January 1971, and
 
WCAP-7672 (Non-Proprietary), June 1971.  (Additional
 
background information only)
: 3. Mesmeringer, J. C., "Failure Mode and Effects Analysis (FMEA)
 
of the Engineered Safety Features Actuation System," WCAP-
 
8584, Revision 1 (Proprietary) and WCAP-8760, Revision 1
 
(Non-Proprietary), February 1980.
: 4. Gangloff, W. C. and Loftus, W. D., "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"      WCAP-7706-L (Proprietary) and WCAP-7706 (Non-Proprietary),
July 1971.
: 5. J-105A-00031, System Reliability Analysis for Advanced Logic System.
 
7.3-58    Rev. 23 WOLF CREEK TABLE 7.3-1 CONTAINMENT COMBUSTIBLE GAS CONTROL SYSTEM ACTUATED EQUIPMENT LIST Actuating Channel
 
Description                                    1      4 Ctmt H 2 Purge Inside Valve                          X Ctmt H 2 Purge Outside Valve                                X Ctmt H 2 Sample 1 Delivery Inside Valves            X Ctmt H 2 Sample 2 Delivery Inside Valves                    X Ctmt H 2 Sample 1 Delivery Outside Valve            X Ctmt H 2 Sample 2 Delivery Outside Valve                    X Ctmt H 2 Sample 1 Return Valve                      X Ctmt H 2 Sample 2 Return Valve                              X Ctmt H 2 Mixing Fan 1*                              X Ctmt H 2 Mixing Fan 2*                                      X Ctmt H 2 Mixing Fan 3*                              X Ctmt H 2 Mixing Fan 4*                                      X Emergency Exhaust Fans                              X      X Ctmt H 2 Thermal Recombiner 1                        X Ctmt H 2 Thermal Recombiner 2                                X Key:  Ctmt = Containment Additional details are provided on the electrical schematic diagrams and the control logic diagrams referenced in Section 1.7.
* The hydrogen mixing fans are not required to operate following an accident. Refer to Section 6.2.5 for additional information.
Rev. 8 WOLF CREEK TABLE 7.3-2
 
CONTAINMENT COMBUSTIBLE GAS CONTROL SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure              Effect on System Detection Remarks Loss of one ac power    Loss of redundancy    Immediate - indicator    Remaining channel fully channel control                                lights                    operable Loss of one dc power    Spurious valve        Immediate - indicator    Remaining channel fully channel control          closure                lights                    operable Control switch  OPEN    Loss of redundancy    Periodic testing or      Loss of control from or                                    spurious operation        main control room wiring failure  SHORT    Spurious operation
 
may occur Loss of instrument      No effect                                        There are no air-operated air system                                                                components in this system Rev. 0 WOLF CREEK TABLE 7.3-3 CONTAINMENT PURGE ISOLATION ACTUATION SYSTEM ACTUATED EQUIPMENT LIST Actuating Channel Description                                      1      4 Ctmt Shutdown Purge Supply Valve Inside                X Ctmt Shutdown Purge Supply Valve Outside                      X
 
Ctmt Shutdown Purge Exhaust Valve Inside                      X Ctmt Shutdown Purge Exhaust Valve Outside              X Ctmt Shutdown Purge Supply                            Nonsafety Fan and Damper Ctmt Shutdown Purge Exhaust                            Nonsafety Fan and Damper Ctmt Mini-purge Supply Valve Inside                    X Ctmt Mini-purge Supply Valve Outside                          X Ctmt Mini-purge Exhaust Valve Inside                          X
 
Ctmt Mini-purge Exhaust Valve Outside                  X
 
Ctmt Mini-purge Supply                                Nonsafety Fan and Damper Ctmt Mini-purge Exhaust                                Nonsafety Fan and Damper Key:  Ctmt = Containment Additional details are provided on the electrical schematic diagrams and the control logic diagrams referenced in Section 1.7.
Rev. 13 WOLF CREEK TABLE 7.3-4 CONTAINMENT PURGE ISOLATION ACTUATION SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Mode                  Effect on System Detection Remarks Loss of one ac power channel    No effect                  Immediate-annunciator      Air-operated valves are controlled by dc solenoids Loss of one dc power channel    System isolates            Immediate-annunciation      Trip - isolates on loss of bus. Periodic
 
test on individual device
 
level Loss of instrument air system    Purge valves fail closed  Immediate-indicator        Valves fail in safe lights and annunciation    position Analog sensor wiring (a) HI      (a) System isolates        (a) Immediate-annunciator  (a) Trip - isolates (b) LO      (b) System becomes 1      (b) Immediate-computer      (b) Channel failure
 
out of 3                                              alarm operates Analog sensor wiring (a) OPEN    (a) System becomes 1      (a) Immediate-computer      (a) Channel failure out of 3                                              alarm operates
 
(b) SHORT  (b) System isolates        (b) Immediate-annunciator  (b) Trip - isolates Bistable fails                  System becomes either      Immediate-annunciator      Either trip or detected 1 out of 3 or trips                                    by periodic testing CIS input open                  Loss of one sensing        Periodic testing            Diverse inputs (radiation,                                parameter in one channel                              manual) fully operable CIS input shorted                Spurious trip              Immediate-annunciator      Spurious closure; however,                                                                                        valves are normally closed Manual input open                Loss of system level      Periodic testing            Automatic actuation and manual initiation in                                  device level control
 
one train                                              fully operable Manual input shorted            Spurious trip              Immediate-annunciator      Spurious closure; however,                                                                                        valves are normally closed Output relay coil open or        Spurious trip              Immediate-annunciator      Spurious closure; however, shorted                                                                                valves are normally closed Output relay mechanically        No automatic actuation    Periodic testing            Manual control not impaired, jammed                          of associated devices                                  other train will isolate
 
in one channel Rev. 13 WOLFCREEKTABLE7.3-5FUELBUILDINGVENTILATIONISOLATIONACTUATIONSYSTEMACTUATEDEQUIPMENTLIST Actuating ChannelDescription14FuelBuildingExhaustFanAXFuelBuildingExhaustFanBX OutsideAirSupplyIsolationDamper*XOutsideAirSupplyIsolationDamper*XFuelBuildingSupplyFanA**Nonsafety FuelBuildingSupplyFanB**Nonsafety EmergencyFilterSupplyIsolation(A)X EmergencyFilterSupplyIsolation(B)XFuelStoragePoolExhaustIsolation(A)XFuelStoragePoolExhaustIsolation(B)X*Thesetwodampersareinseries.**Normallyoperating;triponFBVIS.
AdditionaldetailsareprovidedontheelectricalschematicdiagramsandthecontrollogicdiagramsreferencedinSection1.7.Rev.14 WOLF CREEK TABLE 7.3-6 FUEL BUILDING VENTILATION ISOLATION ACTUATION SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Mode                  Effect on System          Detection                    Remarks Loss of one ac power channel    Loss of redundancy        Immediate-annunciator        Partial trip
 
Loss of one dc power channel    Disables the associated    Immediate-annunciation      Reduces system to minimum actuation channel          of loss of bus. Periodic    sufficiency
 
test on individual device
 
level Loss of instrument air system    None - not applicable      Not applicable              Vent dampers are electrically operated Radiation    HI                System isolated            Immediate-annunciator        Safe condition Sensor LO                System becomes 1 out of 1  Immediate-computer          Reduces system to minimum
 
sufficiency Analog        OPEN              System becomes 1 out of 1  Immediate-computer          Channel failure alarm sensor                                                                                  operates
 
wiring fails
 
SHORT              System isolates            Immediate-annunciator        Spurious actuation Bistable fails                  System either isolates or  Immediate-annunciator        Either trip or detected becomes 1 out of 1                                      by periodic testing Manual input open                Loss of system level      Periodic testing            Automatic actuation and manual initiation for                                  device level control
 
one channel                                            fully operable Manual input shorted            Spurious trip              Immediate-annunciator        Spurious operation does not impair power genera-
 
tion over short term Output relay coil open          Actuation of associated    Immediate-annunciator        Control devices are or shorted                      devices                                                actuated
 
Output relay mechanically        No automatic actuation    Periodic testing            Manual control and
 
jammed                          of associate devices                                    redundant equipment are
 
in one train only                                      not impaired Output        OPEN              Loss of redundancy        Periodic testing            Loss of device control wiring                                                                                  from main control room SHORT              May produce spurious      Spurious isolation          Spurious isolation isolation Rev. 13 WOLF CREEK TABLE 7.3-7 Table Has Been Deleted Rev. 11 WOLF CREEK TABLE 7.3-8 CONTROL ROOM VENTILATION ISOLATION CONTROL SYSTEM I. ACTUATED EQUIPMENT LI S T                                                        Actuation Channel
 
Description                                  Number Control Room Filtration S ystem A Dampers                  1 Control Room Filtration S ystem B Dampers                  4 Upper Cable S preading Room Ventilation Isolation          4 (2 dampers)
 
Control Room A/C Unit A                                    1
 
Control Room A/C Unit B                                    4
 
Control Room Ventilation Isolation (2 dampers)            4
 
Lower Cable S preading Room Ventilation Isolation          1 (2 dampers)
 
Control Building Outside Air S upply Unit
* Control Building Exhaust Fan A
* Control Building Exhaust Fan B
* Access Control Exhaust Fan A
* Access Control Exhaust Fan B
* Control Building Outside Air S ystem Isolation              1 & 4 (5 dampers)
 
S wgr and Battery Room Ventilation Isolation                4 (2 dampers)
 
E S F S WGR Rm Ventilation Isolation (2 dampers)              4 Access Control Area Ventilation Isolation                  4
 
(3 dampers)
 
Control Room Pressurization S ystem A                      1 Control Room Pressurization S ystem B                      4 Class 1E A/C S ystem B                                      1 Class 1E A/C S ystem B                                      4 Control Building Exhaust S ystem Isolation                  1 & 4 (5 dampers)
 
Control Building Access Control Exhaust S ystem            4 & 1 Isolation (3 dampers)
 
Hot Laboratory Ventilation Isolation (2 dampers)          1
 
Hot Laboratory Ventilation Isolation (2 dampers)          4
 
Chase and Tank Area Ventilation Isolation (2 dampers)      1
 
Chase and Tank Area Ventilation Isolation (2 dampers)      4
 
Companion Power Unit E S FA S , CRVI S (where applicable)      1 Companion Power Unit E S FA S , CRVI S (where applicable)      4
*Nonsafety related Additional details are provided on the electrical schematic diagrams and the control logic diagrams referenced in S ection 1.7. The method for achieving isolation damper redundancy is
 
shown in Table 7.3-8, S heet 2. Rev. 1 WOLF CREEK TABLE 7.3-8 (S heet 2)              II. CONTROL BUILDING I S OLATION DAMPER S                                            Corresponding Channel 1 Dampers                          Channel 4 Dampers Control Building Exhaust S ystem D311 (HZ 13G)                              D312 (HZ 184D)
D014 (HZ 59B)                              D301 (HZ 184B)
 
DO18 (HZ 13B)                              D016 (HZ 55B)
 
DO18 (HZ 13B)                              D013 (HZ 98B)
 
D018 (HZ 13B)                              D012 (HZ 122B)
 
D018 (HZ 13B)                              D219 (HZ 59B)
 
D283 (HZ 13E)                              D015 (HZ 57B)
Control Building S upply S ystem D309 (HZ 13F)                              D310 (HZ 184C)
D279 (HZ 13D)                              D005 (HZ 57A)
D006 (HZ 59A)                              D300 (HZ 184A)
 
D002 (HZ 13A)                              D004 (HZ 55A)
 
D002 (HZ 13A)                              D007 (HZ 98A)
 
D002 (HZ 13A)                              D008 (HZ 122A)
 
D002 (HZ 13A)                              D009 (HZ 123A)
 
D198 (HZ 172A)                            D199 (HZ 173A)
Access Control Exhaust S ystem D025 (HZ 13C)                              D220 (HZ 123C)
D313 (HZ 13H)                              D314 (HZ 184E)
D203 (HZ 172B)                            D202 (HZ 173B)
Rev. 0 WOLF CREEK TABLE 7.3-9 CONTROL ROOM VENTILATION ISOLATION ACTUATION SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Mode                  Effect on System Detection Remarks Loss of one ac power channel    Loss of redundancy        Immediate-annunciator      Other channel is still operable Analog sensor  HI                Spurious trip              Immediate-annunciator      System placed in safe fails                                                                                  condition LO                Loss of redundancy        Periodic testing            Other sensor is still operable Analog        OPEN              Loss of redundancy        Periodic testing            Other sensor is sensor                                                                                  still operable
 
wiring fails
 
SHORT            Spurious trip              Immediate-annunciator      System placed in
 
safe condition Bistable fails                  Loss of one channel        Periodic testing            Other channel is of automatic actuation                                still operable Manual input open                Loss of system level      Periodic testing            Redundant train and manual initiation for                                  automatic actuation
 
one channel                                            and device level control
 
are fully operable on
 
affected train Manual input shorted            Spurious trip              Immediate-annunciator      System placed in safe condition
 
Output relay open or shorted    Actuation of associated    Immediate by automatic      Controlled devices are
 
devices                    testing system              actuated Output relay mechanically        Loss of redundancy        Periodic testing            Redundant train is jammed                                                                                  still operable Output        OPEN              Loss of redundancy        Periodic testing            Redundant train is wiring                                                                                  still operable
 
fails          SHORT            May produce spurious      Periodic testing or        Spurious operation
 
isolation                  spurious isolation Rev. 13 WOLF CREEK TABLE 7.3-10 DEVICE LEVEL MANUAL OVERRIDE FAILURE MODES AND EFFECTS ANALYSIS Failure Mode                Effect on System Detection Remarks Failure of bypass switch:
: a. Fails open                a. Blocks automatic        a. Instantaneous (bypass    a. Redundant train actuation in one            lamp illuminated) sys-      can operate train                        tem level annunciator
: b. Fails closed              b. Loss of bypass          b. Periodic testing        b. No loss of automatic function                                                  or manual actuation Output wiring/opens          Loss of redundancy or may    a. Periodic testing or      Redundant channel will or shorts                    produce spurious operation      immediate testing if    operate
 
spurious operation
 
is indicated Failure of indicating        No loss of control          Periodic testing            Function will be achieved light                                                                                  without indication; system level bypass annunciation and indica-
 
tion are provided Rev. 0 WOLF CREEK TABLE 7.3-11 AUXILIARY FEEDWATER ACTUATION SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Mode Effect on S ystem                    Detection Remarks Loss of one Class 1E      Loss of one motor-driven auxi-      Immediate-annunciator    The redundant motor-driven aux-ac power supply          liary feed pump, associated feed                              liary feed pump and steam-driven
 
control valves, and one essential                              auxiliary feed pump are still service water suction valve in                                available. The redundant suction affected train and one in the                                  valve in the turbine-driven pump
 
turbine-driven pump suction                                    supply is not affected.
Loss of one Class 1E dc power supply:
Loss of S eparation        Loss of control power to one        Immediate-annunciator    The redundant motor-driven auxi-Group 1                  motor-driven auxiliary feed pump.                              liary feed pump and steam-driven Two of the feed regulating valves                              auxiliary feed pump are still
 
for the turbine-driven pump fail                              available. The other two feed
 
open.                                                          regulating valves for the
 
turbine-driven pump function normally.
Loss of S eparation        Loss of turbine-driven pump          Immediate-annunciator    The two motor-driven auxiliary Group 2                  due to dc-controlled steam                                    feed pumps and associated valves supply valves                                                  remain completely functional.
Loss of S eparation        No effect                            --                        No auxiliary feedwater components Group 3                                                                                  are controlled from this group.
 
Loss of S eparation S ame as for S eparation Group 1,      Immediate-annunciator S ame as for S eparation Group 1 Group 4                  except that it occurs to the other train Loss of one Class 1E      Loss of one indication train        Immediate-annunciator    Redundant train(s) still instrument power supply  Loss or partial trip of one                                    available.
 
protection train Loss of instrument air    Does not affect system              Immediate-annunciator    Air reservoirs are utilized supply                    function                                                      as a backup air supply.
S afety injection signal  Loss of "safety injection            Periodic testing          Does not affect manual initiation open                      signal" auto initiation in                                    or other auto initiations.
one channel only S afety injection signal S tarts motor-driven auxiliary        Immediate-annunciator    Operator override to terminate shorted                  feed pump and closes steam                                    auxiliary feedwater supply is generator blowdown and sample                                  possible after assessment of
 
valves.                                                        situation.
Rev. 13 WOLF CREEK TABLE 7.3-11 (S heet 2)    Failure Mode Effect on S ystem                    Detection Remarks Loss of power signal      Loss of "loss of power" auto        Periodic testing          Does not affect manual initiation open                      initiation in one channel                                    or other auto initiations.
Loss of power signal S tarts the steam-driven            Immediate-annunciator    Operator override to terminate shorted                    auxiliary feed pump and closes                                auxiliary feedwater supply is steam generator blowdown and                                  possible after assessment of
 
sample valves                                                situation.
Blackout sequence signal  Loss of blackout auto initia-      Periodic testing          Redundant train and turbine-open                      tion in one motor-driven pump                                driven pump operate Blackout sequence signal S tarts one motor-driven auxi-      Immediate-annunciator    Operator override to terminate shorted                    liary feed pump and closes                                    supply of auxiliary feedwater steam generator blowdown and                                  is possible after assessment
 
sample valves                                                of situation.
Blackout signal or safety  Loss of one motor-driven pump      Periodic testing          Other motor-driven pump and injection signal shorted                                                                turbine-driven pump will
 
operate manual controls still operable to start affected pump.
Main feed pump trip        Main feed pump trip will not        Periodic testing          Does not affect manual initiation signal open                initiate auxiliary feedwater                                  or other auto initiations.
Main feed pump trip S tarts motor-driven auxiliary      Immediate-annunciator    Operator override to terminate signal shorted            feed pump and closes steam                                    supply of auxiliary feedwater generator blowdown and sample                                is possible after assessment
 
valves                                                        of situation.
Manual control switch      Loss of manual initiation of        Periodic testing          Does not affect auto initiations failure open              the associated function                                      or manual initiation of
 
other equipment.
Manual control switch S tarts associated auxiliary        Immediate-annunciator    Operator regulates to shorted                    feed pump and closes steam                                    proper level.
generator blowdown and sample
 
valves Rev. 1 WOLF CREEK TABLE 7.3-12 AUXILIARY SUPPORTING ENGINEERED SAFETY FEATURE SYSTEMS System                                          Section Component cooling water                          9.2.2 Essential service water (ESW)                    9.2.1.2
 
Containment spray                                6.0, 7.3 Emergency exhaust                                9.4.2 Diesel generator building ventilation            9.4.7
 
ESW pump house ventilation                      9.4.8
 
Main steam                                      10.3
 
Main feedwater                                  10.4.7 Rev. 0 WOLF CREEK TABLE 7.3-13 NSSS INSTRUMENTATION OPERATING CONDITION FOR ENGINEERED SAFETY FEATURES No. of No. of            Channels
 
No. Functional Unit Channels To Trip
: 1. Safety Injection
: a. Manual                2                  1
: b. Containment pressure (Hi-1)        3                  2
: c. Low steam line        12 (3/steam line)  2 in any one pressure lead-lag                          steam line
 
compensated
: d. Pressurizer low        4                  2 pressure  (a)
: 2. Containment Spray
: a. Manual  (b)            4                  2
: b. Containment            4                  2 pressure (Hi-3)
NOTES (a)  Permissible bypass if reactor coolant pressure is less than 2,000 psig.(b)  Manual actuation of the containment spray system requires the simultaneous operation of two separate switches, as
 
described in Section 7.3.8.1.1. Note that this also
 
initiates phase B containment isolation. The requirement for
 
the simultaneous operation of two switches is desirable to prevent the inadvertent actuation of this system.
Rev. 0 WOLF CREEK TABLE 7.3-14 NSSS INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS No. of No. of              Channels
 
No. Functional Unit Channels to Trip
: 1. Containment Isolation
: a. Automatic safety      See item 1 (b) injection (Phase A)  through (d) of Table 7.3-13
: b. Containment          See item 2 (b) pressure              of Table 7.3-13
 
(Phase B)
: c. Manual Phase A              2                    1
 
Phase B              See item 2 (a) of
 
Table 7.3-13
: 2. Steam Line Isolation
: a. High steam line      12 (3/steam line)    2/steam line negative pressure                          in any steam
 
rate                                      line
: b. Containment          3                    2 pressure (Hi-2)
: c. Safety injection    See item 1 (c) of Table 7.3-13
: d. Manual              2*                    1*
Rev. 0 WOLF CREEK TABLE 7.3-14 (Sheet 2)
No. of No. of              Channels
 
No. Functional Unit Channels to Trip
: 3. Feedwater Line Isolation
: a. Safety injection      See item 1 of Table 7.3-13
: b. Steam generator      4/loop                2/loop high-high level
 
2/4 on any steam
 
generator
: c. Steam generator      4/loop                2/loop low-low level
 
2/4 on any
 
steam generator
: d. Reactor coolant      1/loop                2 (Note 1) low average
 
temperature 2/4
: e. Manual                2*                    1*
 
*Manual actuation of either train closes all main feedwater isolation valves or all main steamline isolation and bypass
 
valves. It is also possible to operate these valves individually.
 
However, those controls are provided for convenience only, and
 
they do not meet the requirements of safety-related controls.
NOTE 1:  The feedwater line will isolate on low T AVG only in          conjunction with reactor trip.
Rev. 0 WOLF CREEK TABLE 7.3-15 NSSS INTERLOCKS FOR ENGINEERED SAFETY FEATURE ACTUATION SYSTEM Function Designation        Input Performed P-4      Reactor trip              Actuates turbine trip Closes main and bypass feedwater valves on TAVG below setpoint Prevents opening of main and bypass
 
feedwater valves which
 
were closed by safety
 
injection or high-high
 
steam generator water
 
level Allows manual block of the automatic
 
reactuation of safety injection Reactor not tripped        Defeats the block preventing automatic
 
reactuation of safety
 
injection P-11    2/3 pressurizer pressure    Allows manual block of below setpoint              safety injection
 
actuation on low
 
pressurizer pressure
 
signal Allows manual block of safety injection actuation and steam
 
compensated steam line
 
pressure signal and
 
allows steam line
 
isolation on high steam
 
line negative pressure
 
rate Rev. 13 Definition Hand Switch input to logic ProCIISS Switch input to logic Output exisu only when all inputs are pre.ent Output exists only when one or more inputs are Output exists only when input is not present Output exists only when input has been continuously present for a preset *time and remains prasent Output exists only when input is present and for a preset time after the input is not present Set output exists when set input is present & continues until the reset input is prasent. Reset outi>ut exists only when set output is not present. Output exnts only when at least A out of B inputs are present Output exists under special conditions not otherwise noted. Digital output exists only when input ;. lower than setpoint Digital output exists only when input is higher than setpoint Output is electrically isolated from input Test signal can be inserted manually in place of normal signal RED (RI -Operating I GREEN IGI -Not operating
\ 1 J.0601 AMBER IAI -Warning, take note ( re
* WHITE IWI -Advisory mformatron l Input to annuncrator.
Input to computer Resultant action intiated by loyi.: Logic continuation DRAWING NUMBERING Numbering coniOYms tn Bechtel Engineeting Procedure 6-f. Sheet numbeu correspond to instrument loop numbers. Function MANUAL INPUT PROCESS INPUT AND OR NOT ON DELAY OFF DELAY (TIMED MEMORY! MEMORY COINCIDENCE MATRIX SPECIAL LOW BISTABLE HIGH BISTABLE ISOLATION TEST DEVICE LIGHT ANNUNCIATOR COMPUTER OUTPUT ACTION CONTINUA liON ! . WOLF CREEK Symbol
* s.p 0---r---= TE!.T . 0 =o 0 -L> 'd Goner al Notes 1. logtc symbcls represent system functions and do not necessarily duplicate cit* cuit arrangement or devices. logic diagrams do not inherently imply -.gized, de-energized, or other circuit operation states. 2. Process equi1m>ent will change state when a change rs initiated, and will remain in thos srate until a change to another state is initiated.
: 3. Process equipment will remain in, or return to, the original sta1e alta. a loss and restoration of power, unless otherwise noted. 4. Inherent equipment interlocks such as circuit breaker trip. free and reyllfling starting cross tnterloclcs are not shown. 5. Some protection actions are shown also as start premissives.
Trip-free design prevents equipment operation when a protection action exists, IIYeR if a start perntissive is not provided.
: 6. Final instrument setpoints are shown elsewhere.
Setpoints shown on control logic diagran1s are approximate.
: 7. See electrical drawings for details of equipment elettrical overcurrent short circuit, and differential protection and space heaters. ' 8. The memory, reset, and start premmive logic associated with the operation of electrical protection devrces is not shown. Electrical auxiliary system breakers are reset by operation of the control. room switch to trip. Mechanical auxiliary system circuits are reset by operation of a switch at the switchgear or motor control center. 9. The test control switches at the switchgear which funciion only when a circuit breaker is in --he te1t position are not shown. 10. All circuit cor trois, except interlocks with othe. equipment, function when . 1 circuit breaker is in the test position to allow circuit testing. I 1. The logic to shovo* that valve and damper position tights are both on wfMn the equipment is in a intermediate position is not st.own. 12. Limit aJd torque switches to stop valve and damper motor actuators at the end of_ travel are not shown on the logic. The valve typa and required actions wtll be noted on tha diagram when available.
: 13. Solenoid pilot operated valves are held in position by limit switc:hes (or relays} unless otherwise noted. "' z 0 Z<t -> 1-W <ta: t.lCD OlD ...lot 000 001-(1<19 100 101-8!1!1 900-999 LC MCC SWGR r **Local \n main conttol Joom -Main control room panel irel. dwg. 10466-J*OJ3621l -Local in field -Field control panel (ref. dwg. 10466-J-0650)
**Plant COflliJllter
-480 V load center -* Motor control center --Switchgear WOLP CREEK Rev. 0 UPDA'l'ED SA!'E'l'Y ANALYSIS REPOR"r . .. . . . .*.
FIGURE 7.3-1 .... J-. ENGTN&#xa3;rRED SAFETY FEATURES ACTUATION SYSTEM (8QP) (SHEET 1) i ALPT3"1* ( 5EPA.F\ATION CJROUP I ) FROM MMIUALJ AC.TUATIOtJ i\NO RESET BUTTON <!:!ROUP I) (APP. A. 51-ff. 2.) (B-7)
* TRANSFER A.U'JC. FEEDWF\TER PUMP SUCTION TO E6W ( 6EPP..RP..TION C:IROUP I) WOLF CREEK (SEPARATION C!ROUP 2) >--.. TEST FROM. ,e..\JX. FE.EDWI'-"TER MOT"OR. DRIVEN AC..TUA.TION (GROUP 1) (,._PP. 5HT.2.) ( ,._-7)
,._UX.FEEDWA.TER TURBINE DRIVEN AcrU,<<>.TION (C:!ROUP {A.PP. A. SHT. 2)(A.-Co)
'
(5EPA.RA.TION GROUP 4) >---eTEST
/ NOT&#xa3;5: I. FOR LE.ClE.ND OF SYMBOLS '!lEE SHEET I. Z.. 516NAL 150\...ATION 15 TO BE BY THE. eE.LLER FOP. ,._LL 1"-NNUNCIATOR AND COMPUTER INPUT5. 3.
* BUYER 5UPPLIED FROM A.UX. FEEOWIUER MorOR DRIVEN N:.TUATION (GROUP4) ("'PP. A SHT. 2.)(A-7) I 4 ALPT37i ALPT36,, AL.PT39 F'RE56URE FOR FEEDWATE.R PUI"\P HEADER U)W 5UC.TION PRESSURE.
S. BiSTABLE "TRip" LIGHTS MAY BE LOCA"TED ON THE FROM AUX. FEEDWATER TUR61f.JE DRIVEN Ac:TUAT\01-J (GROUP 2.) (NJP. A 5)-IT.e.) (A-<o) FROM MANUAL AC..TUATIOtJ AND RE5ET 6llTTOIJ ( SEPARATIOI-J C:.ROUP 4) (APP. A.
z.) TRANSFER A.UY.. FEEDWATER. ( 6-7) . PUMP SUC. TION TO E5W ( 5EPF\R,..TION 6ROUP 4) / R_e_v_.* 0 WOLI!' CREE!t UPDATED SAFETY ANALYSIS REPORT FIGURE 7.3-1 LOGIC DIAGRAM ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (BOP) (J104 APP. A, SHEET 3) 
, .. 1-ANALOG TESTING LOGIC TESTING BISTABLE IN PUT f.----LOGIC CIRCUIT WOLF CREEK MASTER RELAY TESTING .. , *I -MASTER RELAY f--r-SLAVE RELAY 1 SLAVE RELAY f-SLAVE RELAY SLAVE RELAY w I I L.j _____ .,.l I_J SLAVE RELAY 11---_____, FINAL DEVICE OR ACTUATOR TESTING SOLENOID I VALVES MOTOR STARTERS MOTOR OPER. VALVES SOLENOID VALVES MOTOR STARTERS MOTOR OPER. VALVES BREAKER PUMP MOTORS ACTUATORS l --IL. __ A_c_Tu_A_To_R_s
_ _, WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.3-2 TYPICAL ENGINEERED SAFETY FEATURE TEST CIRCUITS Rev. 0 X. .ill. (I) WOLF CREEK TEST LIGHT DS* DEY 3 'ILLUMINATED PUSHBurTDM SWITCH WITH 28V lAMP NO 327 (EXCEPT AS NOTED) REAR OF PANEL !. /" TYP. TERMINAL NUMBERS ill (5) -h2 ' .Jt,* CONTACT LOCATION SCHEME GENERAL MOTES.:
* I. CIRCUITRY AND HARDWARE FOR REDUNDANT PROTECTION TRAINS "A" AND "8" TEST CABINETS ARE DUPLICATE EXCEPT AS NOTED A -TRAIN "A" ONLY a
* TRAIN *a* ONLY 2. IN DETAILS A & a THE SYMBOL
* REPRESENTS THE SUFFIX NUMBERS OF THE DEVICE REFERENCED.
EXAMPLE! K* -SPS RELAY, K601, K602, ETC. K(O) -OPERATING COIL K(R)
* RESET COIL s* -SIC TEST SWITCH, S802, S83ij .ETC. K8* -SIC RELAY, KSII, K817, ETC. OS* -SIC .LIGHT, DS8009, DS8077, ETC. 3. "DETAIL A" & *a* TYPE CIRCUITS ARE DETAILED ON THE SCHEMATICS. "DETAIL B" CIRCUITS WILL BE SUBSTITUTED FOR "DETAIL A" CIRCUITS WHERE REOUIRED.
LOCATION LEGEND SPs -SOLID STATE PROTECTION SYSTEM &sect;!&#xa3; -SAFEGUARDS TEST CABINET ! -SWGR, NCC, AUXILIARY RELAY RACK, ETC. A.S.k -AUXILIARY SAFEGUARDS CAB I NET DETAIL A TYPICAL PROTECTION ACTUATION CIRCUIT BLOCKING SCHEMES (CONTACT CLOSURE FOR ACTUATION)
DETAIL B TYPICAL PROTECTION ACTUATION CIRCUIT BLOCKING SCHEMES DETAILS A AND B OF THIS FIGURE ARE NOT TO BE CONFUSED WITH ALPHA DESIGNATION OF LOGIC TRAINS A AND B. (CONTACT OPENING FOR ACTUATlON) .
Rev
* 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.3-3 ENGINEERED SAFEGUARDS TEST CABINET (INDEX, NOTES, AND LEGEND)
WOLFCREEK7.4SYSTEMSREQUIREDFORSAFESHUTDOWN7.
 
==4.1INTRODUCTION==
WolfCreekGeneratingStation(WCGS)hasbeendesignedtoenabletheplanttobeplacedinasafeshutdowncondition,followinganyaccidentornaturalhazard,usingonlysafety-relatedsystems.TheintentofRegulatoryGuide1.139andBTPRSB5-1aremet.ClarificationsandspecificexceptionstotheseguidesarediscussedinTables7.4-2and7.4-3,respectively.Appendices3Band9.5BprovidetheresultsofintegratedhazardsanalyseswhichdemonstratethatWCGShasbeendesignedtowithstandpostulatedevents.Itemsconsideredincludetornadoes,floods,missiles,pipebreaks,fires,andseismicevents.Theabilitytoachieveasafeshutdownconditionafterafireoccurs(post-firesafeshutdown)hasbeenanalyzedandisdocumentedinAppendix9.5B.ThesinglefailurecriteriautilizedinthedesignarediscussedinSection 3.1.Thesafeshutdownfunctionsdescribedinthissectionarecontrolledandmonitoredfromthecontrolroomortheauxiliaryshutdownpanel(ASP).Foradiscussionofsafeshutdownusingcontrolsandindicationsentirelyoutsideofthecontrolroom,seeSection7.4.6.7.4.2SAFESHUTDOWNOVERVIEWThedesignofWolfCreekenablestheplanttoachieveasafeshutdownundernormalplantconditions,post-accident(DBA)conditions,andpost-fireconditions.ThreeTechnicalSpecificationoperatingconditionsareconsideredassafeshutdown:(1)hotstandby(reactorsub-critical,T-avg>350&deg;F);(2)hotshutdown(reactorsub-critical,350&deg;F>T-avg>200&deg;F);and(3)coldshutdown(reactorsub-critical,T-avg<200&deg;F).10CFR50AppendixR,SectionIII.L.1.c,requiresthatPWR'sachieveandmaintainHotStandby,andeventually,ifnecessary,ColdShutdown.HotShutdownistraversedwhilegoingfromHotStandbytoColdShutdown,andisalsoconsideredasafeshutdowncondition,butitisnotnormallyaconditionthatismaintainedatWCGS.Therefore,WCGS'spost-accidentandpost-fireanalysesutilizetheterms"HotStandby"and"ColdShutdown".Whensafelyshuttingdowntheplantthereactormustbebroughttoasub-criticalconditionwithdecayandsensibleheatbeingremoved.Thisisaccomplishedbyinsertingcontrolrods,boratingthereactorcoolantsystem(RCS),andmaintainingtheRCStemperatureutilizingthemainsteam,essentialservicewater,andsupportsystems.Inordertomaintainasafeshutdownconditioncertainparametersmustbecontrolled:(1)reactivity;(2)decayandsensibleheat;and(3)reactorcoolantinventory.Theseparametersarecontrolledandmanipulatedinsafeshutdownconditionsandrequireuseofcertainprocessmonitoringinstrumentationandsupportsystemfunctions.Thedesignationofsystemsthatcanbeusedforpost-accidentsafeshutdowndependsonidentifyingthosesystemswhichprovidethefollowingcapabilities:a.Circulationofreactorcoolantb.Reactivitycontrol/borationc.Residualheatremovald.Depressurization7.4-1Rev.14 WOLFCREEKTable7.4-5providesalistingofsystemsthatarerequiredtoachieveandmaintainapost-accidentsafeshutdown.Thesystemshaveredundancy/diversity,andnosinglefailurewillcompromisesafetyfunctions.AllpowersuppliesandcontrolfunctionsforrequiredportionsofthesesystemsareClassIE,asdescribedinChapters7.0and8.0,exceptasdescribedfortheboricacidtransfersystem(Section9.3.4).AsdiscussedinSection3.2,allcomponentsmeettherequirementsofRegulatoryGuides1.26and1.29EachoftheWCGSSystemDescriptionsforthesystemsidentifiedinTable7.4-5identifiestheintegralrolethatthesystemplaysinachievingandmaintainingasafeshutdown.7.4.3SAFESHUTDOWNSCENARIOTheplantisdesignedwithanumberofsystemswhichareused,ifavailableundernormaloremergencyconditions,tosafelyshutdowntheplant.Thefollowingshutdownscenariodemonstratesthattheplantcanbetakentobothhotstandbyandcoldshutdownconditionsusingonlysafety-relatedequipment.Althoughtheuseofcertainnonsafety-relateditemswouldbepreferableinmostsituations,thisscenariodoesnottakecreditfornonsafety-relateditemsbecauseoftheassumptionsstatedinSection3.1forthesinglefailure criteria.AtWCGStherearenosystemsdedicatedassafeshutdownsystems,perse.However,proceduresforsecuringandmaintainingtheplantinasafeconditionprovidealignmentofselectedsystemsinthenuclearsteamsupplysystemandBOP.Thediscussionofthesesystems,togetherwiththeapplicablecodes,criteria,andguidelines,isfoundinothersectionsofthissafetyanalysisreport.Inaddition,thealignmentofshutdownfunctionsassociatedwiththeengineeredsafetyfeatures,whichareinvokedunderpostulatedlimitingfaultsituations,isdiscussedinChapter6.0andSection7.3.WCGSsafeshutdowndesignbasisishotstandby.HotstandbyisasafeandstableplantconditionforareactorplantthatincorporatesaWestinghouseNSSS.ExaminationofConditionII,III,andIVeventsfortheWestinghouseNSSShasrevealednonethatrequirecooldowntocoldshutdownconditionsforsafetyreasons.Whiletheplantisinthehotstandbycondition,theauxiliaryfeedwater(AFW)systemandthesteamgeneratorsafetyvalvesoratmosphericreliefvalves(ARVs)canbeusedtoremoveresidualheattomeetallsafetyrequirements.Thelong-termsafetygradesupplyofAFWallowsextendedoperationathotstandbyconditions.BorationduringthehotstandbyconditionisdiscussedinSection7.4.3.2.Eventualachievementofcoldshutdownconditionsmayberequiredforlong-termrecovery.However,thereisnosafetyreasonwhythismustbeaccomplishedinsomelimitedperiodoftime.Nothingintheplantdesignprecludestheeventualachievementofcoldshutdown,evenassumingasafeshutdownearthquake(SSE),alossofoffsitepower,andthemostlimitingsinglefailure.7.4-2Rev.14 WOLFCREEK7.4.3.1HotStandbySystemsHotstandbyisdefinedastheconditioninwhichthereactorissubcriticalandthereactorcoolantsystemtemperatureandpressureareinthenormaloperatingrange.Theminimumsystemsandfunctionsrequiredtomaintainhotstandbyunderanaccidentconditionarediscussedbelow.
* ReactorCoolantSystem(RCS)
* Reactorcoolantcirculation
* Pressurizer
* Waterlevel/RCSinventory
* Pressure(RCSpressureorpressurizerpressure)
* MainSteam(SteamGenerators)
* Atmosphericreliefvalvescontrol
* Steamgeneratorwaterlevelindication
* Steamgeneratorpressureindication
* AuxiliaryFeedwaterSystem
* Auxiliaryfeedwatersupply
* Suctionpressureforeachauxiliaryfeedwaterpump
* ChemicalandVolumeControlSystem(CVCS)
* Boricacidtanklevel
* Emergencyletdownflow
* ReactorCoolantPump(RCP)sealwaterflow
* Boroninjectionflow
* EssentialServiceWater(ESW)
* Provideslastresortwatersupplyforauxiliaryfeedwater
* Providescoolingwaterforcomponentcoolingwatersystem(CCWS) components
* ComponentCoolingWaterSystem(CCWS)
* Flowindicationtocomponentsinsidecontainment 7.4.3.1.1ReactorCoolantSystemCirculationismaintainedbyoperatingtheReactorCoolantPumps(RCP's).Post-accidentsafeshutdownutilizes"naturalcirculation"whenoffsitepowerisnotavailable,Section7.4.5.7.4-3Rev.14 WOLFCREEK7.4.3.1.1.1PressurizerRCSinventoryismaintainedbyaddingwaterviatheCVCS.RCSpressureismaintainedbycontrollingcooldownrateusingthesteamgeneratoratmosphericreliefvalves(ARVs)andrelievingpressure,asnecessary,utilizingthepressurizerpoweroperatedreliefvalves(PORV's).7.4.3.1.2MainSteam(SteamGenerators)MainsteamsystemcomponentsareutilizedastheheatsinkfortheRCSandtocontroltheRCScooldownrate.Steamgeneratorsandatmosphericsteamreliefvalvesaretheprimarycomponentsutilized.7.4.3.1.2.1WaterlevelforeachsteamgeneratorWaterlevelforthesteamgeneratorsismaintainedbyuseoftheauxiliaryfeedwatersystem.Controlsandindicationareinthecontrolroomandontheauxiliaryshutdownpanel(ASP).7.4.3.1.2.2PressureforeachsteamgeneratorSteamgenerator(SG)pressureismaintainedbycyclingtheARVs.SGpressurecontrolisimportanttothecontrolofRCScooldownrate.TheARVsareutilizedtocontrolpost-accidentRCScooldownrate.Theinstrumentationandcontrolsfortheatmosphericsteamreliefsystemconsistofcontrols,transmitters,andindicatorstoprovideautomaticormanualactuationoftheARVstoremovereactorheatfromthereactorcoolant system.BoththesafetyvalvesandtheARVsarelocatedupstreamofthemainsteamisolationvalves,outsideofthecontainment,andbothprovideameansofremovingreactorheatinahotstandbycondition.Thesafetyvalvesarefull-capacity,spring-loadedvalveswhichareactuatedbyhighmainsteamlinepressure.TheyaredescribedmorefullyinChapter10.0.TheARVs,however,arethepreferredmodeofsteamrelieftoavoidprolongedoperationofthesafetyvalves.ApressuretransmitterandpressurecontrollerareprovidedforeachofthesteamgeneratorstoactuatetheARVandcontrolthesteampressureatapredeterminedsetting.Manualcontrolcapabilityisprovidedbothinthecontrolroomandontheauxiliaryshutdownpanelforatmosphericreliefvalveregulation.ThestatusoftheARVsisindicatedbyopenandclosedindicatinglightsandbythecontrolleroutputindication.7.4.3.1.3AuxiliaryFeedwaterAuxiliaryfeedwaterprovidesthecoolingmediumforthereactorwhilemitigatingmostdesignbasisaccidents(DBAs).Itreceivesitswatersupplyfromthecondensatestoragetank(CST)oressentialservicewater(ESW)system.Feedwaterflowiscontrolledtothesteamgeneratorswhichserveastheheatsinkforthereactorcoolant.ReactorcoolanttemperatureiscontrolledbycyclingARV's.Theauxiliaryfeedwaterpumpsstartautomatically,asdescribedinSection7.3.6,orcanbestartedmanually.Start/stoppumpcontrolslocatedontheauxiliaryshutdownpanel(aswellasinsidethecontrolroom)areprovided,aswellascontrolfortheflowcontrolvalves.Foracompletediscussionoftheinitiatingcircuits,logic,bypasses,interlocks,redundancy,diversityandsupportingsystems,seeSection7.3.6.1.1.7.4-4Rev.14 WOLFCREEK7.4.3.1.4ChemicalandVolumeControlSystem(CVCS)CVCSprovidesseveralfunctionsduringshutdown:RCSboration;RCPsealwater;andRCSinventorycontrol.Instrumentationandcontrolsrequiredforthesefunctionsareassociatedwith:Boricacidtanklevel,Boroninjectionflow,RCPsealwaterflow,andEmergencyletdownflow.ForadetaileddiscussionoftheCVCS,seeSection9.3.4.7.4.3.1.5EssentialServiceWaterSystem(ESWS)TheESWSprovidesthesafetyrelatedandlongtermwatersupplytothesuctionoftheauxiliaryfeedwaterpumps.TheCSTisthepreferredsourceofwatertotheAFW,butitisnotprotectedagainsttheeffectsofnaturalhazards.Italsocontainsalimitedvolumeofwater.TheinstrumentsimportanttosafeshutdownarethepressureinstrumentswhichmonitorthewatersupplypressurefromthistankandinitiateswitchovertotheESWSshouldthesupplyfromtheCSTbeinterrupted.ESWSalsoprovidescoolingtotheComponentCoolingWaterSystem.ForadetaileddiscussionoftheESWS,seeSection9.2.1.2.7.4.3.1.6ComponentCoolingWaterSystem(CCWS)CCWprovidescoolingwaterforsafeshutdowncomponents.ForadetaileddiscussionoftheCCWS,seeSection9.2.2.7.4.3.2HotStandbyDiscussionReactorcoolantiscirculatedbyRCP'swhenachievinganormalplantsafeshutdown.Ifnormaloffsitepowerisnotavailable,thereactorcoolantiscirculatedusingnaturalcirculation.SeeSection7.4.5foradiscussionofnaturalcirculation.Post-accidentsafeshutdowncanbeachievedwithoutoffsitepowerbeingavailable.Theauxiliaryfeedwatersystem,inconjunctionwiththesafety-relatedportionofthemainsteamsystem,isinitiallyreliedupontotransferresidualcoreheatfromtheRCS,viathesteamgenerators,totheatmosphere.Thisisaccomplishedbyreleasingsteamfromthesecondarysideofthesteamgenerators,whilemaintainingsteamgeneratorpressure.SteamisreleasedviatheARV's.Theauxiliaryfeedwatersystemisusedtomaintainalevelinthesteamgeneratorsduringthisperiodoftime.Waterisnormallyprovidedtotheauxiliaryfeedwaterpumps(AFP)fromthecondensatestoragetank(CST);however,inthecaseofanSSEortornadohazard,theunprotectedCSTmaybeunavailable.Inthiscase,redundantpressuretransmittersinthesuctionlinesoftheauxiliaryfeedwaterpumps(AFP)detectlossofAFPsuctionpressureandisolatetheAFPsuctionheaderfromtheCST.Concurrently,withCSTisolation,theessentialservicewater(ESW)pumpsarestarted,andthevalvesintheESWSheadersareopenedtoadmitESWtotheAFP.TheCSTisolationvalvesareinserieswithcheckvalvestofurtherprecludetheshort-circuitingofESWflowtotheCST.TheESWauxiliaryfeedwatersupplyvalvesaresegregatedbytrainrelationshipstothemotor-drivenAFPwiththeturbine-drivenAFPbeingfedbybothtrainAandBESWheaders.Therefore,evenwithasinglefailure,ESWisalignedtoaminimumofonemotor-drivenAFPandtheturbine-drivenAFP.7.4-5Rev.14 WOLFCREEKThemotor-operatedAFPdischargevalvesaresegregatedbytrainrelationshiptothemotor-drivenAFPs,eachpumpfeedingtwosteamgenerators.Theturbine-drivenAFPhastwoair-operateddischargevalvesofonetrainandtwooftheother,soastohaveredundantandoppositetrainsegregationtothemotor-operatedvalvesassociatedwiththemotor-drivenAFPs.Asafety-relatedgassupplyisprovidedforthesevalvesasbackuptotheairsupply.Inallcases,adequateauxiliaryfeedwaterissuppliedtothesteamgeneratorsforRCSheat removal.ThesteamgeneratorARVsareair-operatedvalvessegregatedbytrainrelationshipwiththesteamgenerators,suchthatadequatereliefcapabilityexistsatalltimestoaccomplishRCSheatremoval.TheARVsareremotelycontrolledvalves,whichcaneitherautomaticallymaintainapresetpressureinthemainsteampipingorcanbemanuallycontrolledfromeitherthemain-controlboardortheauxiliaryshutdownpanel.Asafety-relatedbackupgassupplyisprovidedforthesevalves.Inordertomaintainanextendedhotstandby(greaterthan24hours),additionalnegativereactivitymustbeaddedtotheRCS.ThisisaccomplishedbyboratingtheRCSwhilerelyingonreactorcoolantpumpsornaturalcirculationintheRCStoensureadequatemixingoftheinjectedboricacidwithinthereactorcoolant.Thedesignborationconditionisbasedonaddingsufficientboricacidtobringthereactortoaxenon-freecoldshutdownconditionfromthehotfull-powerpeakxenoncondition.TheCVCSprovidesdiversemeansofboratingtheRCStoaconcentrationthatexceedstherequirementforasafeshutdownofthereactorfromanyoperatingcondition,assumingthatthecontrolrodclusterwiththehighestreactivityworthisstuckinitsfullywithdrawnpositionandintheunlikelyeventthatsafeshutdownisinitiatedfrompeakxenonconditions.Theadditionof3600gallonsof4-weight-percentboricacidisrequiredwithin25hoursafterreactorshutdowntomaintainthereactorinahotstandbycondition.Thisisequivalentto10,500gallonsofwaterfromtherefuelingwaterstoragetank(RWST)(at2400ppmboron).IntermsofboronconcentrationintheRCS,thiscorrespondstoapproximately300ppmboron,assumingzeroboronconcentrationintheRCSinitially.Atotalof13,450gallonsofmakeupisrequiredtomaintaintheRCSinahotstandbycondition.Borationmaybeaccomplishedbyusingtheboricacidtransferpumps(BATPs)andboricacidtanks(BAT)orbyusingtheRWSTandthecentrifugalchargingpumps.AtleastoneBATPisavailableundermostplantconditions.Eachpumpispoweredfromaredundantseparationgroupoftheonsiteemergencypowerdistributionsystem.Thesupplycircuitbreakersareshunttrippedonlyupontheoccurrenceofasafetyinjectionsignal(SIS).However,operationoftheBATPscannotbeassuredfollowingaseismiceventoruponoccurrenceofanSIS.Whenthisisthecase,theRWSTisusedasthesourceofboratedmakeuptotheRCS.TheboricacidtransfersystemisavailableforalleventsfollowingwhichtheRWSTisassumedtobeunavailable.RedundantlevelindicationfortheBATandRWSTisprovidedonthemaincontrolboard(MCB).Theselevelindicationsareusedtodeterminethatsufficientboronconcentrationhasbeenattainedforsafeshutdown.Ifthenormalchargingpathisunavailable,boronisaddedthroughoneoftwodiverseflowpathsinthechargingsystem(reactorcoolantpumpsealsortheboroninjectionpath).EachpathiscapableofdeliveringacontrolledflowofboratedwaterfromtheRWSTorboricacidtransfersystem,whichcanbematchedtotheletdownrateinordertomaintainpressurizerlevel.Theemergencysafety-relatedletdownpathdivertsletdowncooledbytheexcessletdownheatexchangertothepressurizerrelieftank(PRT).Inaddition,letdownfromtheRCSmayalsobeaccomplished,utilizingthepressurizerPORVs.Thesevalvesarepoweredbyredundantpowertrains.7.4-6Rev.14 WOLF CREEK The PRT has a total volume capacity of 13,500 gallons. Although not normal operating procedure, prior to initiating letdown through the excess letdown heat exchanger, the 10,000 gallons of relatively clean water in the PRT can be discharged to the containment normal sump at a controlled rate. This makes the PRT available to contain the cooled letdown from the excess letdown heat exchanger and, thereby, minimize release of airborne radioactivity to the
 
containment.
During the hot standby condition, the reactor coolant pump seals require cooling by either seal injection or component cooling water. Normally, a
 
continuous source of component cooling water is provided. Seal injection is
 
via the charging pump. RCS leakage past the seal, with no seal return, goes to
 
the PRT via the seal return line relief valve if the containment isolation valves are isolated. The loss is considered in the RCS inventory.
7.4.3.3  Cold Shutdown Discussion
 
Should an event occur which would place the plant under a Limiting Condition for Operation, or if recovery from the event will cause the plant to be shutdown for an extended period of time, the plant may be taken from a hot
 
standby condition to a cold shutdown condition.
With the RCS in a hot standby condition, cold shutdown procedures may be initiated. The essential functions which must be continued or initiated to
 
achieve cold shutdown are: Continued residual heat removal via the steam generators, utilizing auxiliary feedwater and the atmospheric relief valves. Continued circulation of the coolant in the RCS. Letdown and boration to cold shutdown boric acid concentrations. RCS depressurization. Initiation of the residual heat removal (RHR) system when the RCS reaches
 
350&deg;F and 360 psig. Continued residual heat removal. The RHR system is utilized to achieve cold shutdown temperature (Tavg <200&deg;F ).
Cooldown is accomplished by increasing the steam dump from the ARVs to attain a
 
primary side cooldown rate of approximately 50&deg;F /hr. Boration during cooldown
 
is required by procedure to maintain adequate shutdown margin. In conjunction
 
with this portion of the cooldown, the charging pumps are used to deliver
 
refueling water to makeup for primary coolant contraction due to cooling.
Letdown and boration to achieve cold shutdown boric acid concentrations are identical to procedures described above for the hot standby condition. The
 
completion of this step requires that the RCS boron concentration be increased
 
to approximately 1700 ppm boron at the beginning of an operating fuel cycle and
 
to approximately 690 ppm boron at the end of the cycle. These concentrations range from approximately 300 to 690 ppm higher than the boron concentrations at hot full power equilibrium xenon. The specific required cold shutdown
 
concentration for any time during any fuel cycle and for the actual xenon condition may be calculated using a written procedure. 7.4-7 Rev. 16 WOLFCREEKBorationisoneofthemeansusedforreactivitycontrolwhencoolingdowntheRCStocoldshutdown.Boronconcentrationisadjustedtomaintainadequateshutdownmarginasrequiredduetoreactivitychangesfromthecooldown.Inordertomaintainpressurizerlevelwithinthedefinedoperatingband,borationisacombinedchargingandletdownprocess.Atypicalvolumeofwatertobechargedandletdownfromhotfullpoweratthebeginningofafuelcycleatpeakxenonconditionsis33,500gallonsandtheendofafuelcyclewatervolumerequirementfromfullpoweris83,754gallons,wheretheRWSTisthesourceoftheboratedwater.TheRCScoldshutdownconcentrationisensuredbyprocesscontrol,i.e.,knowledgeofinitialRCSboronconcentrationsandknowledgeofamountsandconcentrationsofinjectedfluid(eitherRWSTorBATfluids)ensuresthatthecoldshutdownconcentrationisobtained.ContinuedcirculationofRCSisaccomplishedbyreactorcoolantpumpsresultingfromheatremovalviathesteamgenerators.Naturalcirculationisutilizedwhenoffsitepowerisnotavailable.DepressurizationoftheRCSisachievedthroughtheuseofthepressurizerPORV's.Aspreviouslystated,eachvalvehasanindependentsafety-relatedpoweractuationtrain.AfterreducingRCSpressurebelow1000psig,itisnecessarytoensurethatallaccumulatortankisolationvalvesareintheclosedpositiontoavoidtheirdischargetotheRCS.Fortwoofthefourvalves(thosepoweredbythedieselgenerator),thisisaccomplishedfromtheassociatedmotorcontrolcenter(MCC).Ifpowerisnotavailablefortheremainingtwovalves,theoperatorcanventcovergasfromtheaffectedaccumulator.WhentheRCShasbeencooledanddepressurizedtoapproximately350&deg;Fand360psig,theRHRsystemisputinservice.ThisisdonebyestablishingcomponentcoolingwaterflowthroughtheRHRheatexchangerbyopeningtheassociatedmotor-operatedvalveandbyclosingthemotor-operatedisolationvalvestotheRCScoldlegsandtotheRHRpumpsuctionfromtheRWST.ThenextoperationrequiresthattheRCS/RHRisolationvalvesbeopened.Withalossofoffsitepowerinconjunctionwiththefailureofonedieselgenerator,oneofthetwoisolationvalvesineachRHRsuctionlinecannotbeopenedfromthemaincontrolboard.Inordertoinitiatesystemoperation,electricalpowertoopentheclosedisolationvalvecanbesuppliedtoitsMCCbyinstallingatemporaryjumperfromtheoppositeelectricalpenetrationroompoweredbytheoperationaldieselgenerator
.AfteropeningtheRCS/RHRisolationvalves,theRHRpumpismanuallystartedtocirculateflowthroughtheminiflowline.Theminiflowbypassvalvewillautomaticallyopentomaintainminimumflowbasedonthesignalreceivedfromtheflowindicatingswitchintheoutletpipingofthepump.AtthispointitispossibletoobtainamanualsampleoftheRHRloopfluid.RCSboronconcentrationwillbemaintainedatgreaterthanorequaltotherequiredcoldshutdownRCSconcentration.Thissamplecanbeobtainedatanyofseveraldrainandventconnections,thelocalsampleconnection,orviathedirectconnectiontothenuclearsamplingsystem,ifavailable.Duringthisperiodoftime,theoperatorhasdeterminedthestatusoftheRHRsystemcontrolsandispreparedtoputtheRHRsystemintooperation.ThenextstepistoestablishflowfromtheRCShotlegtotheRCScoldlegsviatheRHRpumpandheatexchanger.7.4-8Rev.14 WOLFCREEKDuringthefinalcooldownphase,sincetheair-operatedflowcontrolvalvesmaynotbefunctional,variousmeasurescanbetakentoavoidexcessiveheatloads(andresultingexcessiveduty)onthecomponentcoolingwatersystem.Thesemeasuresincludethefollowing:1)onlyoneRHRpumpmaybeoperatedortwoRHRpumpscanbestarted/stoppedoveranextendedperiodoftimetolimitthetotalheatloadontheRHRheatexchangers,2)throttlingoftheCCWtotheRHRheatexchangercanbeaccomplished.Thiswillresultinlessflow,thoughatahighertemperature,backtotheCCWheatexchangers,or,3)nitrogenbottlescanbetemporarilyinstalledandconnectedintotheairlinestooperatetheair-operatedflowcontrolvalves.ContinuedoperationinthismodedecreasestheRCStemperaturetocoldshutdownconditions(<200&deg;F).ThecapabilityoftheRHRsystemtoaccommodateasinglecomponentfailureandstillperformasafetygradecooldownisdemonstratedinthefailuremodeandeffectsanalysis(FMEA)oftheRHRsystemforsafety-relatedcoldshutdownoperationsprovidedasTable7.4-4.7.4.4PLANTSAFESHUTDOWN(PSSD)Fornormalplantsafeshutdowntheplantistakenfromfullpoweroperationtoacoldshutdownconditionusingsafety-relatedandnon-safety-relatedsystemsfollowingthescenariopresentedinSection7.4.3.Plantsystemsandnormaloperatingproceduresareutilizedtobringtheplanttosafeshutdownconditionsforanoutageorrepairofequipment.Noaccidents,malfunctionsorhazardsareassumedtooccur.Intheeventofaturbineorreactortripduringnormaloperationstheplantisplacedinandmaintainedatahotstandbyconditionasdescribedin7.4.3.2.IfTechnicalSpecificationsorrecoveryfromaneventcausestheplanttobeshutdownforanextendedperiodoftime,theplantistakentoacoldshutdownconditionasdescribedin7.4.3.3.Ineithercondition,anadequateheatsinkisprovidedtoremovereactorcoreresidualheat.Borationcapabilityisprovidedtocompensateforxenondecayandtomaintaintherequiredcoreshutdownmargin.7.4.5POST-ACCIDENTSAFESHUTDOWNToeffectapost-accident(orposthazard)safeshutdown,theunitisbroughttoandmaintainedat,asafeshutdownconditionundercontrolfromthemaincontrolroomortheauxiliaryshutdownpanel.Post-accidentsafeshutdownisachievedintwophases:(1)HotStandbyand(2)ColdShutdown.Hotstandbyisachievedbyinsertingthecontrolrods,initiatingcooldownviathemainsteamsystem,andinitiatingborationoftheRCSasdescribedin7.4.3.2withtheprimaryexceptionsbeingreactortripandtheuseof"naturalcirculation".Naturalcirculationisdescribedbelow:Inthephysicallayoutofthereactorcoolantsystem,thereactorcoreisatalowerelevationwithrespecttothesteamgenerators;consequently,thehighertemperatureheatsourceisbelowtheheatsink.Thisconfigurationensuresheatwillbetransportedfromthereactorcoretothesteamgeneratorsviathefreeconvectionflowphenomena(natural circulation).7.4-9Rev.14 WOLFCREEKImbalanceofforcesisneededtoinitiateaconvectiveflow.Athinlayeroffluidneartheheattransfersurfacesinthereactorcoreisheated,generatingagradientintemperatureanddensity.Whenaparticleofheatedfluidisdisplacedfromneartheheattransfersurface,itentersaregionofgreateraveragedensityandis,therefore,subjecttoabuoyantforce.Thebuoyancyforceisopposedbyviscousdragandbyheatdiffusion.Convectionbeginswhenbuoyancyovercomesthedissipativeeffectofviscousdragandheatdiffusion.Asthetemperatureofthefluidinthereactorcoreisincreasedrelativetothetemperatureofthefluidinthesteamgenerators,aconvectiveflowwillbemaintainedthroughoutthereactorcoolantsystem.TheportionsofthereactortripsystemrequiredtoachievetheshutdownconditionaredescribedinSections7.2and7.5.Post-accidentcoldshutdownisachievedutilizingthesystems,components,indication,andcontrolsasdescribedinSection7.4.3.3.withtheadditionofthesafety-relatedRHRSystemTheRHRsystemhasalowerdesignpressurethantheRCS.Therefore,cooldownfromhotstandbytocoldshutdownrequiresatwo-stepprocess.Duringthefirststep,transferofdecayandsensibleheat,afterreactorshutdown,willbeviathesteamgenerators.Duringthesecondstep,theRHRsystemwillbeutilizedasameansofheattransfer.TheRHRsystemisusedtocooltheRCSdownfromapproximately350&deg;Ftolessthan200&deg;F.ItisthenusedtomaintainthetemperatureoftheRCSlessthan200&deg;Ftoassurecoldshutdownismaintained.Limitationsplacedoncoldshutdownarethoserelatedtoequipmentdesignandfailurecausedbytheaccident.TheRHRsystemisdescribedindetailinSection5.4.7.Theinstrumentationandcontrolsforcoldshutdownsystemsmayrequiresometemporarymodificationsinorderthattheirfunctionsmaybeperformedfromoutsidethecontrolroom.Notethatthereactorplantdesignincludesattainingthecoldshutdownconditionfromoutsidethecontrolroom.Thesystemandcomponentcontrolsandmonitoringindicatorsprovidedontheauxiliaryshutdownpanel(ASP)arediscussedinsection7.6.andlistedinTables7.4-1,7.4-1.1and7.4-1.2.SomeASPcomponentscanbeisolatedfromthecontrolroom.SuchcomponentsareidentifiedinTable7.4-1.7.4.6SAFESHUTDOWNFROMOUTSIDETHECONTROLROOM7.4.6.1DescriptionIftemporaryevacuationofthecontrolroomisrequiredbecauseofsomeabnormalplantcondition,theoperatorscanestablishandmaintaintheplantinahotstandbyconditionfromoutsidethecontrolroomthroughtheuseofcontrolslocatedattheauxiliaryshutdownpanel,attheswitchgear,oratmotorcontrolcenters,andotherlocalstations.Hotstandbyisastableplantcondition,automaticallyreachedfollowingplantshutdown.Thehotstandbyconditioncanbemaintainedsafelyforanextendedperiodoftime.Intheunlikelyeventthataccesstothecontrolroomisrestricted,theplantcanbesafelykeptatahotstandby,untilthecontrolroomcanbereentered,bytheuseoftheessentialmonitoringindicatorsandthecontrolslistedinTables7.4-1.1and7.4-1.2.TheauxiliaryshutdownpanelroomislocatedinthenortheastcorneroftheauxiliarybuildingonelevelbelowthecontrolroomatElevation2026.Therearetwodistinctauxiliaryshutdownpanelsatthislocation;onepanelisassociatedwithinstrumentationandcontrolcircuits7.4-10Rev.14 WOLFCREEKusedforcontrollingsafeshutdownequipmentintrainA,andtheotherpanelisassociatedwithinstrumentationandcontrolcircuitsusedforcontrollingsafeshutdownequipmentintrainB.Bothpanelsareelectricallyseparatedandareassociatedwiththesamesafety-gradecircuitsthatservetheirrespectivetrains.TheauxiliaryshutdownpaneldesignalsoprovideselectricalisolationofinstrumentationandcontrolcircuitsfortheequipmentcontrolledbetweentrainBauxiliaryshutdownpanelandthecontrolroom.SwitchesareprovidedonBauxiliaryshutdownpaneltoisolateandremovecontrolfromthecontrolroomforthetrainBsafeshutdownequipmentnecessarytotaketheplanttoandmaintaintheplantinasafehotstandbyconditionindependentofthecontrolroom.Thiscapabilityisassuredintheeventapostulatedfirecausesdamageinthecontrolroomandsubsequentevacuationoftheoperators.TrainBinstrumentationandcontrolswereselectedtobeisolatedbecausetheinstrumentationandcontrolfortheturbine-drivenauxiliaryfeedwaterpumparelocatedonthispanel.ThecontrolroomfireisanalyzedinAppendix9.RefertoTable7.4-1forthelistofinstrumentationandcontrolsonBauxiliaryshutdownpanelthathaveanisolationfeature.Althoughtheprimaryintentoftheauxiliaryshutdownpanelisthemaintainingofhotstandbyfromoutsidethecontrolroom,thispanelcanalsobeusedforcertainfunctionswhenimplementingcoldshutdownfromoutsidethecontrol room.7.4.6.1.1AuxiliaryShutdownPanel(ASP)Theauxiliaryshutdownpanelisutilizedtoachieveandmaintainsafeshutdownconditionswhenitisnecessarytoevacuatethecontrolroom.TheASPcontrolsandindicatorsarelistedinTable7.4-1.1.7.4.6.1.2ControlsatSwitchgearMotorControlCenters,andOtherLocationsInadditiontothecontrolsandmonitoringindicatorslistedinTable7.4-1.1otheressentialcontrolsareprovidedoutsideofthecontrolroomwithacommunicationnetworkbetweenthesecontrollocationsandtheauxiliaryshutdownpanel:SeeTable7.4-1.2.7.4.7ControlsforExtendedHotStandbyInordertomaintainanextendedhotstandby(greaterthan24hours),additionalnegativereactivitymustbeaddedtotheRCStoaccommodatethepositivereactivityaddedthroughxenondecay.Thiscanbeaccomplishedbymanualcontrolofthenormalchargingandletdownsystemsviacontrolsattheauxiliaryshutdownpanel,motorcontrolcenters,switchgear,andcontrolofindividualequipmentatthedevicelocation.However,extendedhotstandbyconditionscanbemaintainedfromoutsidethecontrolroomthroughtheuseofredundant,safety-gradesystemsonly.ThisisaccomplishedbymeansofthecontrolsandindicationsontheASPandtheadditionalcontrolslistedinTables7.4-1.1and7.4-1.2.Priortoapproximately25hoursafterreactorshutdown,sufficientboronwouldbeaddedtotheRCStocanceltheeffectsofxenondecay.Borationcanbeaccomplishedfromoutsidethecontrolroomusingonlyredundantsafety-gradeequipmentbyoperatingoneoftwocentrifugalchargingpumps,takingsuctionfromtheRWST,andchargingintotheRCSthrougheitherthenormalchargingpathortheboroninjectiontankflowpath.7.4-11Rev.14 WOLFCREEKIntheabsenceofanSIS,onecentrifugalchargingpumpwouldbestartedfromitsswitchgearandisolationofnormalletdownfromtheASPwouldcauseautomaticrealignmentofpumpsuctionfromthevolumecontroltank(VCT)totheRWSTwhenVCTlowleveloccursviaaVCTlowlevelsignal.ChargingintotheRCScouldbethroughthenormalchargingline,inwhichallair-operatedvalvesarefail-open.Analternativechargingpathistheboroninjectiontankflowpath.Thenormallyclosedvalvesinthatpathcanbeopenedusinglocalswitchesatmotorcontrolcenters.ToprovidesufficientvolumefortheinjectionofadditionalboratedwatertotheRCS,areductionofRCSaveragetemperaturecanbeaccomplishedbymanuallycontrollingsteamreleasetotheatmospherefromtheredundantsecondary-sideARV's.NecessaryinstrumentationandcontrolsareontheASP.UndertheconditionsofRCSmakeupfromtheRWST,noletdown,andpressurizerlevelmaintainedwithinthenormalrange,sufficientboroncanbeaddedtotheRCStomaintainK eff0.99atalltemperaturesbetweennormaloperatingtemperatureand80&deg;Fatanytimeincorelife,assumingthatthexenonconcentrationinthecoreatthetimeofshutdownwastheequilibriumvalueorless.Inaddition,sufficientboroncanbeaddedinthismannertomaintainextendedhotstandbyconditions.Therefore,theWCGSdesignpermitsachievementofextendedhotstandbyconditionsfromoutsidethecontrolroomusingonlyredundant,safety-gradesystemsandequipment.Inadditiontothenormalchargingandletdownsystems,thesystemsdiscussedmaybeusedtomaintainanextendedhotstandbybylocalactionsoutsidethecontrolroom.BorationoftheRCStothecoldshutdownconcentrationisaccomplishedasdescribedin7.4.3.3.7.4.8DESIGNBASISSystemsandcomponentsdesignbasesarediscussedinapplicablesectionsoftheUSARreferencedthroughoutSection7.4,Table7.4-5andother7.4tables.Generaldescriptionsareprovidedbelowfor:initiatingcircuits;logic;bypasses;interlocks;redundancy;diversity;andactuateddevices.FordetailsrefertotheapplicableUSARsectionsinTable7.4-5.Applicabledesigncriterion,regulatoryguidance,etc.areidentifiedintheapplicableUSARsectionand/orSystemDescriptions.7.4.8.1InitiatingCircuitsInitiatingcircuitsforpost-accidentsafeshutdownareinitiatedautomaticallybyreactorprotectionsystems,engineeredsafetyfeaturesactuationsystem(ESFAS),and/orsafety-relatedplantprocessinginstrumentation.Manualinitiation/actuationisaccomplishedvialocalhandswitches(HS)orremotehandindicatingswitches(HIS).7.4.8.2LogicsLogicsemployedforpost-accidentsafeshutdownarethereactorprotection logics.7.4-12Rev.14 WOLFCREEK7.4.8.3BypassesIsolationofcertainsignalsoccursbydesignormanualimplementationviaprocedureasplantconditionsdictateforaspecifictypeofaccident.7.4.8.4InterlocksInterlocksbetweensystemsandcomponentshavebeendesignedandinstalledforplantoperationandprotection.Theyaredependeduponasoperatoraidsincontrollingplantconditions.7.4.8.5RedundancyTrainsAandBarethetwoSSDequipmenttrainsatWCGS.Electricallythereare6separationgroups:1through4aresafety-relatedwhileGroups5and6arenonsafetyrelated.Electricalgroups1and3areassociatedwithTrainAandgroups2and4areassociatedwithTrainB.Ingeneral,groups1and4areredundanttoeachother.Thefewexceptionsareidentifiedinthearea analyses.Therearesystemsthathavemorethanoneredundantcapabilitysuchassourcerangemonitoring.Thisredundancymaybeanalyzedandreliedupontoachieveandmaintainsafeshutdownconditions.7.4.8.6DiversityWCGShasbeendesignedwithbackupcapabilitiesforsafeshutdown(SSD)functions.Forexample:
* ReactivityControlControlRodInsertionandBoration
* MakeupWaterRWSTandBAT7.4.8.7ActuatedDevicesSSDisaccomplishedbytheoperationofpumps,motors,motoroperatedvalves,pressureinstruments,temperatureinstruments,andplantprocessindication.7.4.8.8SupportingSystemsSystemsorcomponentsrequiredtosupportpost-accidentsafeshutdownaresafety-relatedorhavebeenanalyzedandalternativecapabilitiesidentifiedandimplementedthroughstationprocedures.7.4.8.9ConsequencesAnalysisMaintenanceofasafeshutdownwiththesystemsandassociatedinstrumentationandcontrolsidentifiedinSection7.4hasincludedconsiderationoftheaccidentconsequencesthatmightjeopardizepost-accidentsafeshutdownconditions.Theaccidentconsequencesthatarerelevantarethosethatwouldtendtodegradethecapabilitiesforboration,adequatesupplyofauxiliaryfeedwater,andresidualheatremoval.TheresultsofdesignbasisaccidentanalysesarepresentedinChapter15.0.Ofthese,thefollowingproducetheconsequencesthataremostrelevant:7.4-13Rev.14 WOLFCREEK* Chemicalandvolumecontrolsystemmalfunctionthatresultsinadecreaseintheboronconcentrationinthereactorcoolant(uncontrolledborondilution)
(15.4.6)
* Lossofnormalfeedwaterflow(15.2.7)
* Lossofexternalelectricalloadand/orturbinetrip(15.2.2and15.2.3)
* Lossofnon-emergencyacpowertothestationauxiliaries(15.2.6)Theseanalysesshowthatsafetyisnotadverselyaffectedbytheseincidents,withtheassociatedassumptionsbeingthattheinstrumentationandcontrolsdiscussedinsection7.4.3areavailabletocontroland/ormonitorshutdown.Redundancyofsystemsandcomponentsisprovidedtoenablecontinuedmaintenanceofthehotstandbycondition.Ifrequired,itisassumedthatpermanentortemporaryrepairscanbemadetocorrectorcircumventanyfailureswhichmightotherwiseimpedeeventuallytakingtheplanttothecoldshutdowncondition.RHRDesignTheresultsoftheanalysiswhichdeterminedtheapplicabilityofthenuclearsteamsupplysystemcoldshutdownsystems(RHRS)totheregulatorypositionsofRegulatoryGuides1.139arepresentedinTable7.1-2.7.4-14Rev.14 WOLFCREEK TABLE 7.4-1 AUXILIARY SHUTDOWN PANEL EQUIPMENT LISTInstrumentUnitSep. No.                                No.                            Service                                      Group BB-PI-455B All Pressurizer Pressure NV BB-LI-459B All Pressurizer Level 1*BB-LI-460B All Pressurizer Level 4*BB-PI-406X All RCS Pressure (wide range) 4 BB-PI-405X All RCS Pressure (wide range) 1 BB-HIS-51B AllPzr Htrs Backup Gp A NV*BB-HIS-52B AllPzr Htrs Backup Gp B NV AB-PI-516X All SG A Pressure 4 AB-PI-524B All SG B Pressure 1 AB-PI-535X All SG C Pressure 4 AB-PI-544B All SG D Pressure 1 AE-LI-501A All SG A Level (wide range) 1*AE-LI-502A All SG B Level (wide range) 4 AE-LI-503A All SG C Level (wide range) 1*AE-LI-504A All SG D Level (wide range) 4 AB-PIC-lB AllSG A Stm Dump to Atmos Ctrl 1*AB-PIC-2B AllSG B Stm Dump to Atmos Ctrl 2 AB-PIC-3B AllSG C Stm Dump to Atmos Ctrl 3*AB-PIC-4B AllSG D Stm Dump to Atmos Ctrl 4*AB-HIS-6B All AB-HV-6 Solenoid Valve 2 AB-HS-1 AllSG A Stm Dump Ctrl Xfr Sw 1 AB-HS-2 AllSG B Stm Dump Ctrl Xfr Sw 2 AB-HS-3 AllSG C Stm Dump Ctrl Xfr Sw 3 AB-HS-4 AllSG D Stm Dump Ctrl Xfr Sw 4 AB-ZL-lB AllSG A Stm Dump to Atmos Vlv Posn 1*AB-ZL-2B AllSG B Stm Dump to Atmos Vlv Posn 2 AB-ZL-3B AllSG C Stm Dump to Atmos Vlv Posn 3*AB-ZL-4B AllSG D Stm Dump to Atmos Vlv Posn 4 BG-HIS-8149AB All Letdown Orifice A Isol Vlv NV BG-HIS-8149BB All Letdown Orifice B Isol Vlv NV BG-HIS-8149CB All Letdown Orifice C Isol Vlv NV*BG-HIS-8152A AllLetdown Ctmt Isol Vlv 4 BG-HIS-8160A AllLetdown Ctmt Isol Vlv 1*AL-HK-5B AllSG D Aux Fw Ctrl Vlv MD Pmp B 4 AL-HS-5 AllSG D Aux Fw Ctrl Vlv Xfr Sw 4*AL-ZL-5B AllSG D Aux Fw Ctrl Vlv Posn 4 AL-HK-6B AllSG D Aux Fw Ctrl Vlv to Pmp 1 AL-HS-6 AllSG D Aux Fw Ctrl Vlv Xfr Sw 1 AL-ZL-6B AllSG D Aux Fw Ctrl Vlv Posn 1 AL-HK-7B AllSG A Aux Fw Ctrl Vlv MD Pmp B 4 AL-HS-7 AllSG A Aux Fw Ctrl Vlv Xfr Sw 4 AL-ZL-7B AllSG A Aux Fw Ctrl Vlv Posn 4 AL-HK-8B AllSG A Aux Fw Ctrl Vlv to Pmp 1Rev.14 WOLF CREEK TABLE 7.4-1  (Sheet 2)
AUXILIARY SHUTDOWN PANEL EQUIPMENT LIST Instrument Unit Sep.
No. No. Service Group AL-HS-8 All SG A Aux Fw Ctrl Vlv Xfr Sw 1 AL-ZL-8B All SG A Aux Fw Ctrl Vlv Posn 1 AL-HK-9B All SG B Aux Fw Ctrl Vlv MD Pmp A 1 AL-HS-9 All SG B Aux Fw Ctrl Vlv Xfr Sw l AL-ZL-9B All SG B Aux Fw Ctrl Vlv Posn 1 *AL-HK-10B All SG B Aux Fw Ctrl Vlv to Pmp 4 AL-HS-10 All SG B Aux Fw Ctrl Vlv Xfr Sw 4 *AL-ZL-10B All SG B Aux Fw Ctrl Vlv Posn 4
 
AL-HK-llB All SG C Aux Fw Ctrl Vlv MD Pmp A 1 AL-HS-11 All SG C Aux Fw Ctrl Vlv Xfr Sw 1 AL-ZL-llB All SG C Aux Fw Ctrl Vlv Posn 1 AL-HK-12B All SG C Aux Fw Ctrl Vlv to Pmp 4 AL-HS-12 All SG C Aux Fw Ctrl Vlv Xfr Sw 4 AL-ZL-12B All SG C Aux Fw Ctrl Vlv Posn 4 *AL-FI-lB All SG D Aux Fw Flow 4 AL-FI-2B All SG A Aux Fw Flow 1 *AL-FI-3B All SG B Aux Fw Flow 2 AL-FI-4B All SG C Aux Fw Flow 3 AL-PI-15B All MD Aux Fw Pmp B Disch Press NV AL-PI-18B All MD Aux Fw Pmp A Disch Press NV AL-PI-21B All Turb Driven Aux Fw Pmp Disch PressNV AL-PI-25B All MD Aux Fw Pmp A Suct Press 1 *AL-PI-24B All MD Aux Fw Pmp B Suct Press 4
*AL-PI-26B All Turb Drive Aux Fw Pmp Suct Press 2
*AL-HIS-22B All MD Aux Fw Pmp B 4
 
AL-HIS-23B All MD Aux Fw Pmp A 1 *FC-ZL-312AD, AE, A FAll AFPT Trip & Throt Vlv Posn 2 *FC-HIS-312B All Turb Driven Aux Fw Pmp Trip A Throt Vlv 2 *FC-HIS-313B All Aux Fp Turb Speed Gov Ctrl 2 *AB-HIS-5B All Turb Drvn Aux Fw Pmp Stm Isol Vlv 2
*AB-HIS-6B All Turb Drvn Aux Fw Pmp Stm Isol Vlv 2
 
AP-LI-4B All Cond Stor Tank Level NV *AL-HIS-30B All ESW to MD Aux Fw Pmp B 4 AL-HIS-31B All ESW to MD Aux Fw Pmp A 1 AL-HIS-32B All ESW to Turb Driven Aux Fw Pmp 1 *AL-HIS-33B All ESW to Turb Driven Aux Fw Pmp 4
*AL-HIS-34B All CST to MD Aux Fw Pmp B 4
 
AL-HIS-35B All CST to MD Aux Fw Pmp A 1 AL-HIS-36B All CST to Turb Driven Aux Fw Pmp 1 BB-TI-413X All W.R. RCS Cold Leg Temp Loop 1 NV *BB-TI-423X All W.R. RCS Cold Leg Temp Loop 2 4
 
BB-TI-433X All W.R. RCS Cold Leg Temp Loop 3 NV BB-TI-443X All W.R. RCS Cold Leg Temp Loop 4 NV
 
Rev. 28 WOLFCREEK TABLE 7.4-1  (Sheet 3)
AUXILIARY SHUTDOWN PANEL EQUIPMENT LISTInstrumentUnitSep. No.                                No.                            Service                                      Group SE-NI-31C All Source Range Nuclear Inst NV*SE-NI-61X All Source Range Neutron Flux 4 AE-LI-517X All SG A Level (narrow range) 4 AE-LI-528X All SG B Level (narrow range) l AE-LI-537X All SG C Level (narrow range) 4 AE-LI-548X All SG D Level (narrow range) 1 BG HIS-459A AllRCS Letdown to Regen Hx NV BG HIS-460A AllRCS Letdown to Regen Hx NV FC-HS-313 AllAFPT Gov Ctrl Sel Sw 2 SE-NI-35C AllIntermediate Range Nuclear Inst NV*SE-NI-61Y All Power Range Neutron Flux 4*FC-ZL-315B, 317B All AFPT Gov Vlv Position 2*FC-ZL-312DB AllAFPT Throttle Vlv Trip Mech Pos 2*BB-TI-443A All W.R. RCS Hot Leg Temp Loop 4 4 BB-TI-413Y All W.R. RCS Hot Leg Temp Loop l NV*RP-HIS-1 AllCtrl Rm Instr Xfr Sw 2*RP-HIS-2 AllCtrl Rm Instr Xfr Sw 4*RP-HIS-3 AllCtrl Rm Instr Xfr Sw NV NV - NON-VITAL* -INSTRUMENTATION AND CONTROLS ON RP118B THAT CAN BE ISOLATED FROM CONTROL ROOM CIRCUITSRev.14 WOLFCREEKTable7.4-1.1AuxiliaryShutdownPanelControlsandMonitoringIndicators1.Controlsa)START/STOPcontrolforeachmotor-drivenauxiliaryfeedwaterpump(1)(5)(6)b)START/STOPcontrolsfortheturbine-drivenauxiliaryfeedwaterpump(steamsupplyandtripandthrottlevalvecontrols)(5)
(6)c)MANUALcontrolforallauxiliaryfeedwaterflowcontrolvalves(2)(5)d)OPEN/CLOSEcontrolforessentialservicewatertotheauxiliaryfeedwaterpumpsuctionvalvesandcondensatestoragetanktotheauxiliaryfeedwaterpumpsuctionvalves(1)(5)(6)e)Auxiliaryfeedwaterpumpturbinespeedcontrol(2)(5)f)AUTOMATIC/MANUALcontrolforeachatmosphericreliefvalve(2)
(5)g)ON/OFF/AUTOcontrolfortwopressurizerbackupheatergroups(3)(6)h)OPEN/CLOSEcontrolforthecontainmentisolationvalvesintheletdownline(1)(5)(6)i)OPEN/CLOSEcontrolfortheshutoffvalvesintheletdownlineupstreamoftheregenerativeheatexchangerandfortheletdownorificeisolationvalves2.Monitoringindicators(4)a)Waterlevelforeachsteamgenerator(bothwiderangeandnarrowrange)(5)b)Pressureforeachsteamgenerator(5)c)Reactorcoolantsystempressure(widerange)(5)d)Pressurizerpressuree)Pressurizerlevel(5)f)Suctionpressureforeachauxiliaryfeedwaterpump(5)g)Auxiliaryfeedwaterpumpturbinespeed(rpm)h)Dischargepressureforeachauxiliaryfeedwaterpumpi)Auxiliaryfeedwaterflowtoeachsteamgeneratorj)Condensatestoragetanklevelk)Reactorcoolant(coldleg)widerangetemperatureRev.14 WOLFCREEKTable7.4-1.1(Sheet2)a)Sourcerangenuclearpowerindicatorsb)Intermediaterangenuclearpowerindicatorsc)Indicatinglights(on-off/open-closed)forallpower-operatedequipmentlistedina.above.d)Reactorcoolant(hotleg)wide-rangetemperature(twoloops)e)AnequipmentlistfortheauxiliaryshutdownpaneliscontainedinTable7.4-1.
NOTES:1.TrainAparalleledwiththecontrolswitchinthecontrolroom(controlcanbeaccomplishedfromeitherlocationwithoutuseofatransferswitch;theequipmentrespondstothelastcommandfromeitherlocation).2.Transferofthecontrolcircuitwithswitchattheauxiliaryshutdownpanelisprovidedfortheanaloginstrumentcontrol loop.3."AUTO"modeisnotoperableaftertransfer.4.Alistofmonitoringinstrumentation,includingnumberofchannels,isprovidedinTable7.5-2.5.Essentialmonitoringindicatororcontrol.TrainBcontrolsinthemaincontrolroomcanbeisolatedfromtheauxiliaryshutdownpanelcontrols.Controlistransferredthroughatransferswitchlocatedattheauxiliaryshutdownpanel.Rev.14 WOLF CREEK Table 7.4-1.2 Controls at Switchgear Motor Control Centers, and Other Locations
: 1. Reactor trip capability at the reactor trip switchgear.
: 2. START/STOP controls for both centrifugal charging pumps. Location:
Centrifugal Charging pump switchgear.
: 3. START/STOP controls for the component cooling water pumps. Location:
Component cooling water pumps switchgear.
: 4. START/STOP controls for the containment fan cooler units. Location:
Cooler fan motor control centers.
: 5. START/STOP controls for the control room air-conditioning units.
Location:  At the equipment.
: 6. START/STOP controls for the diesel generators. Location:  Each diesel generator local control panel.
: 7. START/STOP controls for the essential service water pumps. Location:
Essential service pump switchgear.
Rev. 15 WOLF CREEK TABLE 7.4-2 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.139 REV. 1, DRAFT 2 DATED FEBRUARY 25, 1980 TITLED "GUIDANCE FOR RESIDUAL HEAT REMOVAL TO ACHIEVE AND MAINTAIN COLD SHUTDOWN" REGULATORY POSITION WCGS POSITION
: 1. FUNCTIONAL
 
The method utilized to take the reactor from normal operating conditions to cold shutdown should satisfy the functional guidance presented below. a. The design should be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-related systems that satisfy General Design Criteria 1 through 5 (design in compliance with GDC 1).
: a. The reactor coolant system, in conjunction with several supporting systems, can be brought to a cold shutdown condition following any given hazard (GDCs 2, 3, and 4) using safety-related systems (design in compliance with GDC 1) except as noted
 
in Appendix 9.5(B). b. These safety-related systems should have suitable redundancy in components and features and
 
suitable interconnection, leak detection and
 
containment, and isolation capabilities to ensure
 
that, for onsite electric power system operation (assuming offsite power is not available) and for
 
offsite electric power system operation (assuming
 
onsite power is not available), the system safety
 
function can be accomplished assuming a single
 
failure.In demonstrating that the method can be utilized to perform its function assuming a single failure, limited operator action outside the control room
 
would be acceptable if suitably justified.
 
Necessary operator actions to maintain hot
 
shutdown or proceed from that plant condition to
 
cold shutdown should be planned no sooner than one
 
hour from the time when shutdown is commenced.
 
This limited operator action should not result in
 
an exposure beyond the allowed limits assuming
 
high radioactivity in the reactor coolant or containment building environment.
: b. Complies. Section 3.1.2 provides the single failure criteria that is used, including the bases for operator action outside the control room.
 
Table 7.4-3 provides a safety-related cold shutdown (CSD) FMEA.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 2)
: c. The method should be capable of bringing the reactor to a hot shutdown condition, where RHR cooling may be initiated, within approximately 36 hours following shutdown with only offsite power
 
or onsite power available, assuming the most
 
limiting single failure.
: c. Complies
: d. Instrumentation and controls including protective measures and interlocks associated with the safety-related systems required to achieve or
 
maintain cold shutdown should meet the
 
requirements of IEEE Standards 279-1971, 323, 384, and 344 and the guidance provided in Regulatory
 
Guides 1.89, 1.75, and 1.100
: d. Except for the boric acid transfer system controls and the pressurizer heaters, the instrumentation and controls are designed in accordance with applicable Regulatory Guides and IEEE standards.
The highly reliable design of the pressurizer
 
heaters and the boric acid transfer system (both of
 
which are capable of being manually loaded on the
 
diesels) are described in Sections 7.4 and 8.0.
: e. The safety-related systems should be classified as Seismic Category I and meet the guidance provided in Regulatory Guide 1.29.
: e. Except as discussed in 1d, all components and systems comply.2. REACTIVITY CONTROL A safety-related system should meet GDC 1-5, 26, and 27 and be capable of controlling and monitoring boron
 
concentration in order to ensure reactor subcriticality
 
from operating conditions through cold shutdown. Complies.
: 3. HEAT REMOVAL TO REDUCE THE RCS FROM PLANT OPERATING CONDITIONS TO RHR SYSTEM OPERATING CONDITIONS
: a. Auxiliary Feedwater System A safety-related auxiliary feedwater system
 
should be designed and constructed to provide a
 
reliable source of cooling water at PWR plants
 
in accordance with GDC 1-5, 44, 45, and 46.
The auxiliary feedwater system complies with these requirements, as discussed in 10.4.9. The
 
essential service water system (Section 9.2.1.2)
 
which provides the ultimate water supply has
 
adequate inventory to supply short-term and long-
 
term requirements.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 3)
The safety-related water supply for the auxiliary feedwater system for a PWR should have sufficient inventory to permit operation at hot-standby conditions for at least 4 hours
 
followed by cooldown to the conditions
 
permitting operation of the RHR system. The
 
inventory needed for cooldown should be based
 
on the longest cooldown time needed with either
 
only onsite or only offsite power available with an assumed single failure.
The capability should exist for providing cooling water from the ultimate heat sink prior to exhaustion of the safety-related water
 
supply. Automatic initiation should be
 
provided for the auxiliary feedwater system.
 
The automatic initiation signals and circuits
 
should be safety-related and be designed so
 
that a single failure will not result in the
 
loss of AFWS function. Testability of the
 
initiating signals and circuits should be a
 
feature of the design. Manual initiation
 
capability from the control room should be
 
safety-related and be designed so that a single
 
failure will not result in the loss of system
 
function. The a-c motor-driven pumps and
 
valves in the AFWS should be included in the
 
automatic actuation (simultaneous and/or
 
sequential) of the loads to the emergency
 
buses. The automatic initiating signals and
 
circuits should be designed so that their failure will not result in the loss of manual
 
capability to initiate the AFWS from the control room. A safety-related redundant system should be provided for indication in the control room of auxiliary feedwater flow to each steam generator.
Safety-related (Class IE) indication of the AFW flow to each generator is provided in the control room.Safety-related steam generator level indication provides a backup means of determining the AFW flow.b. Steam Relief  A safety-related redundant atmospheric secondary side steam relief system should be designed to provide for reduction of the RCS
 
temperature to RHR system operating conditions. Complies, as discussed in Section 10.3.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 4)
: c. Steam Generator Inventory  Each steam generator should be equipped with a safety-related redundant water level indication and alarm system. Complies, as discussed in Section 10.4.7.
: 4. RESIDUAL HEAT REMOVAL The RHR system should meet GDC 1-5 and 34 with at least two redundant trains of pumps and heat exchangers. Beginning 4 hours after reactor
 
shutdown, each train should have sufficient heat removal capability (a) for maintaining the RCS at hot
 
shutdown (RHR system initial operating conditions) at
 
that time in core life when the greatest amount of
 
decay and residual heat is present, and (b) to provide for cooldown of the RCS from hot shutdown to
 
cold shutdown conditions.
The RHR system meets the applicable GDCs, as described in Section 5.4.7
: a. RHR System Isolation  1) Isolation of the suction side of each RHR system train from direct RCS pressure should be provided by at least two power-operated
 
valves in series, with valve position
 
indicated in the control room. Alarms in
 
the control room should be provided to alert
 
the operator if either valve is open when
 
the RCS pressure exceeds the RHR system
 
design pressure. The isolation valve system
 
should have two or more independent
 
interlocks to prevent the valves from being opened unless the RCS pressure is below the
 
RHR system design pressure. Upon loss of
 
actuating power, isolation valves should not
 
change position unless movement is to a position that provides greater safety. The isolation valve system should have two or more independent protective measures to close any open valve in the event of an increase in the RCS pressure above the RHR
 
system design pressure. 1) Complies, except that the automatic closure of these valves upon high RCS pressure is not considered necessary function based on the
 
analysis performed. See Section 5.4.7.
Rev. 20 WOLF CREEK TABLE 7.4-2 (sheet 5)
: 2) One of the following should be provided the discharge side of the RHR system to isolate it from the RCS:  2) Complies. Meets Paragraph C.    (a) The valves, position indicators, alarms, and interlocks described in item (1). (b) One or more check valves in series with a normally closed power-operated valve. The position of the power-
 
operated valve should be indicated in
 
the control room. If the RHR system
 
discharge line is used for an ECCS
 
function, the power-operated valve
 
should be opened upon receipt of a
 
safety-injection signal once the reactor coolant pressure has decreased
 
below the ECCS design pressure.  (c) Two check valves in series.
: b. RHR System Pressure Relief To protect the RHR system against accidental over pressurization when it is in operation (not isolated from the RCS), pressure relief in
 
the RHR system should be provided with
 
relieving capacity in accordance with the ASME
 
Boiler and Pressure Vessel Code. The most
 
limiting pressure transient during the plant
 
operating condition when the RHR system is not
 
isolated from the RCS should be considered when selecting the pressure relieving capacity of
 
the RHR system. For example, during shutdown
 
cooling in a PWR with no steam bubble in the
 
pressurizer, inadvertent operation of an
 
additional charging pump in the normal charging
 
mode or a high head ECCS pump (for those plants at which the high head pumps serve a dual function) should be consideredin selecting the design bases.
Complies, as described in Sections 5.2.2.10 and 5.4.7.2.5 Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 6)
Fluid discharge through the RHR system pressure relief valves should be collected and contained so that a relief valve that is stuck in the open position will not: (1) Result in flooding of any safety-related equipment.(2) Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA. (3) Result in a non-isolation situation in which the water provided to the RCS to maintain the core in a safe condition is
 
discharged outside the containment.
If interlocks are provided to automatically
 
close the isolation valves when the RCS pressure exceeds the RHR design pressure, relief capacity should be provided during the
 
time that the valves are closing such as to prevent the RHR design pressure from being
 
exceeded.c. RHR System Pump Protection The design and operating procedures of the RHR
 
system and plant operating procedures should be such that no single failure or single operator error can result in loss of the RHR function due to damage of the RHR system pumps including
 
overheating, cavitation, or loss of adequate
 
pump suction head.
Complies. See Section 5.4.7.
: d. RHR System Testing For the RHR system, the isolation valve
 
operability and interlock circuits should be designed to permit on-line testing when operating in the RHR mode. System testing should meet the requirements of IEEE Standard 338 and the guidance of Regulatory Guide 1.118.
Complies. See Chapters 7.0 and 8.0 for IEEE testing.
See the responses to Regulatory Guides1.22 and 1.68 for testing.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 7)
The pre-operational and initial startup test program should be in conformance with the Regulatory Guide 1.68. In addition, the programs for pressurized water reactors should include tests with supporting analysis to confirm (a) that
 
adequate mixing of borated water added to the
 
reactor coolant system prior to or during cooldown
 
can be achieved under natural circulation
 
conditions and permit estimation of the times required to achieve such mixing and (b) that the
 
cooldown under natural circulation conditions can be achieved with the guidelines specified in the
 
emergency operating procedures.
The RHR system should be designed to permit on-line pressure and functional testing to assure (1) the structural and leak tight integrity of its components, (2) the operability and performance of the active components of the system, (3) and the operability of the system as a whole and, under conditions as close to design as practical, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. e. RHR System Operation Indication e. Complies.
Indication of isolation valve position, system pressure and flow, and pump operating status should be available in the control room. f. RHR System Integrity f. Complies.
The RHR system should be designed and constructed to have the capability to remove
 
heat from the reactor coolant during normal and
 
following accident conditions. Since the
 
reactor coolant may be highly radioactive
 
following accident conditions, the RHR system
 
integrity should be such that radioactivity is
 
not released to the environment beyond accepted limits. The Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 8) design should include features to prevent unacceptable degradation of long-term heat removal capability and leakage resulting from a degraded core condition or the containment post-accident environment. In addition, the
 
system should be designed so that the operator
 
can assess the status, isolate, maintain and
 
repair the RHR system, as needed.
 
Specifically, the RHR system integrity should meet the following criteria: (1) Leakage from the system such as from valves and pump seals should be monitored and controlled. The leakage limits at which an RHR train is to be declared inoperable and isolated should be stated
 
in the Plant Technical Specifications.
Indication of the amount of leakag e, such as sump level indication, radiation levels and system isolation should be available locally and in the control
 
room. Valve lineup and isolation capability should be such as to preclude the possibility that highly radioactive
 
sump water can automatically transferred to the radwaste processing system. (1) Complies, except that RHR leakage is addressed in the Reactor Coolant Sources Outside Containment program as discussed in the technical specifications and USAR Section
 
18.3.4. Leakage detection is discussed in Section 9.3.3. (2) Shielding should be provided to maintain personnel exposure as low as is reasonably achievable (ALARA). Shielding
 
protection should also be provided for instruments, components, or other items which might be adversely af fected by high radiation fields. Provisions should be made for access to, and minor repair of, equipment outside containment which may
 
fail during a post-incident recovery
 
period. Provisions should be made for
 
tie-in of additional equipment or systems
 
in the event that major repair is necessary. Area temperature monitoring and control should be provided for the RHR system environment with indication
 
and control in the control room. (2) Complies, except that area temperature monitoring is not provided for WCGS. High temperature alarm is provided in the MCR for the RHR pump rooms. Compliance with ALARA requirements are discussed in Section 12.3.1.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 9) (3) The RHR system including the leakage collection sump should be located in a closed area which is equipped with an engineered safety feature filtration system (as given in Regulatory Guide
 
1.52) and radiation monitors. These
 
areas should be maintained at a
 
sufficient negative pressure (typically, at least - 1/8 inch, water gauge) with respect to the ambient atmosphere to
 
prevent exfiltration of activity which could bypass the ESF filter system. (3) The emergency exhaust filtration system which serves this function following a LOCA is discussed in Sections 9.4.2 and 9.4.3. g. RHR Cooling Water Supply System g. Complies, except that the radiation monitors are not located on the outlet of the RHR heat The safety-related system should be designed
 
and constructed with at least two independent sub-systems or trains such that each has the capacity to adequately remove heat from the reactor coolant in accordance with GDC 1, 2, 3, 4, 5, 44, 45 and 46. Cooling water
 
radioactivity should be monitored at the output
 
of the RHR heat exchangers with indication and
 
an alarm in the control room.
exchanger. Instead, each train of component cooling water is provided with radiation monitors within the system. See Section 9.2.2.
: 5. NATURAL CIRCULATION COOLING FOR PWR PLANTS To ensure the capability to achieve and maintain natural circulation within the primary system, redundant emergency power, which meets General Design
 
Criteria 17 and 18, should be provided to each of the
 
following Complies. See response to Regulatory Guides 1.22 and 1.68. A natural circulation test was performed on Diablo Canyon, which is similar to WCGS.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 10)
: a. The minimum number of pressurizer heaters required to maintain natural circulation conditions.
: a. WCGS is provided with two groups of backup pressurizer heaters. The heater groups and their associated controls are powered from a diesel-backed bus through qualified isolation devices that shed their load only upon an SIS or emergency bus undervoltage signal. If desired, these devices can be manually reclosed from the control room, following reset of the initiating trip signals.
The emergency diesel generators are sized in excess of that required to carry all connected pressurizer heaters concurrent with the loads required for a LOCA. They are provided with a full complement of status indication in the control room.
: b. The control and motive power systems for the power operated relief valves and associated block valves, and b.The pressurizer is provided with two Class IE power-operated relief valves (PORV) and two Class IE power-operated relief valve isolation valves (PORVIV).These valves are powered from the onsite emergency power supply, with redundant Class IE power
 
supplying the two valves associated with each flow
 
path.c. The pressurizer level indication instrument channels.c. Three loops of the pressurizer level instrumentation are powered from Class IE power supplies. In addition, a fourth non-safety grade instrumentation loop is provided.
: 6. REACTOR COOLANT SYSTEM INVENTORY A safety-related system should be designed and
 
constructed to meet GDC 1-5 and 33 and capable of providing reactor coolant makeup and letdown control with a sufficient water supply to account for
 
cooldown shrinkage, required letdown for boration, and technical specification allowed leakage from operating conditions to cold shutdown.
Complies. The chemical and volume control system is described in Section 9.3.4.
Rev. 14 WOLF CREEK TABLE 7.4-2 (sheet 11)
: 7. OPERATIONAL PROCEDURES The operational procedures for bringing the plant from normal operating power to cold shutdown should be in conformance with Regulatory Guide 1.33. For pressurized water reactors, the operational procedures should include specific procedures and information required for cooldown under natural circulation conditions. In addition, plant procedures for all activities should provide instruction in such a manner that will not lead to a loss of the RHR system.
Complies. Backup heaters are supplied by a non-Class IE MCC with a Class IE diesel backed power supply.
Emergency procedures should address cooldown during
 
or after an accident, including natural circulation
 
cooldown in the case of PWR plants. These emergency
 
procedures should include guidance on safe shutdown to cold conditions in the event of failure of non-safety related equipment and single failures of safety-related equipment. Other cases which the
 
emergency procedures should address are RHR heat
 
exchanger tube leak, high radioactivity in the reactor coolant, and high airborne radioactivity in the RHR system room.
Emergency procedures should be prepared to address the transfer of the pressurizer heaters to the
 
emergency power source in the event that this action is necessary. The method and time required to
 
accomplish the transfer of the pre-selected
 
pressurizer heaters to the emergency buses should be
 
described in written approved procedures and be
 
consistent with the timely initiation and maintenance of natural circulation.
Rev. 14 WOLFCREEKTABLE7.4-3DESIGNCOMPARISONOFTABLE1OFBTPRSB5-1FORPOSSIBLESOLUTIONSFORFULLCOMPLIANCEDesignRequirementsofBTPRSB5-1Processand(SystemorComponent)PossibleSolutionforFullCompliance WCGSI.FunctionalRequirementforTakingtoColdShutdowna.Capabilityusingonlysafetygradesystemsb.Capabilitywitheitheronlyonsiteoronlyoff-sitepowerandwithsinglefailure(limitedactionoutsideCRtomeet SFc.Reasonabletimeforcooldown,assumingmostlimitingSFandonlyoffsiteoronlyonsite powerLong-termcooling(RHRdropline)Providedoubledropline(orvalvesinparallel)topreventsinglevalvefailurefromstoppingRHRcoolingfunction(Note:Thisrequirementinconjunctionwithmeetingeffectsofsinglefailureforlong-termcoolingandisolationrequire-mentsinvolvein-creasednumberofindependentpowersuppliesandpossiblymorethanfourvalves.)Seriespower-operatedvalvesareprovidedinbothRHR/RCSshutdownlines.Designcanwith-standasinglefailure,asdis-cussedinSection 5.4.7.Rev. 14 WOLFCREEKTABLE7.4-3(sheet2)DesignRequirementsofBTPRSB5-1Processand(SystemorComponent)PossibleSolutionforFullCompliance WCGSHeatremovalandRCScirculationduringcooldowntocoldshutdown.(NoteNeedSGcoolingtomaintainRCScir-culationevenafterRHRSinoperationwhenundernaturalcir-culation)(atmo-sphericrelief valves.)Providesafety-gradeatmosphericrelief valves,operators,andpower supply,etc.sothatmanual actionshouldnotberequiredafterSSE,excepttomeetsingle failure.Complies.Depressurization (Pressurizerauxiliarysprayorpower-operatedreliefvalves)Provideupgradingandadditionalvalvestoensureoperationofauxiliarypressurizerspray,usingonlysafety-gradesub-systemmeetingsinglefailure.Possiblealternativemayin-volveusingpressurizerpower-operatedreliefvalveswhichhavebeenupgraded.MeetSSEandsinglefailurewith-outmanualoperationinsidecontainment.Complies.FullyqualifiedClassIEpressurizerpower-operatedreliefvalvesareprovided.Rev.14 WOLFCREEKTABLE7.4-3(sheet3)DesignRequirementsofBTPRSB5-1Processand(SystemorComponent)PossibleSolutionforFullCompliance WCGSBorationforcoldshutdown(CVCSandboronsampling)Provideprocedureandupgradingwhereneces-sary,suchthatborationtocoldshutdowncon-centrationmeetsthere-quirementsofI.Solu-tioncouldrangefrom(1)upgradingandadd-ingvalvestohavebothletdownandchargingpathssafetygradeandmeetsinglefailureto(2)useofbackupproceduresinvolvinglesscost.Forexample,borationwithoutletdownmaybeacceptableandeliminateneedforupgradingletdownpath.UseofECCSforinjectionofboratedwatermayalsobeacceptable.Needsurveillanceofboronconcentration(boronometerand/orsamp-ling).Limitedoperatoractioninsideorout-sideofcontainmentifjustified.Theexcoredetectoralertstheoperatorofanycriticalitypotential.ChargingtoandletdownfromtheRCSarecontrolledquantities.Noboronsamplingisrequired.
Rev. 14 WOLFCREEKTABLE7.4-3(sheet4)DesignRequirementsofBTPRSB5-1Processand(SystemorComponent)PossibleSolutionforFullCompliance WCGSII.RHRIsolationRHRSystemComplywithoneofallowablearrange-mentsgiven.II.Complies.SeeSection5.4.7.III.RHRPressureReliefb.Collectandcon-tainrelief discharge.RHRSystemDeterminepiping,etc.,neededtomeetrequire-mentandprovidein design.III.Complies.SeeSection5.4.7.V.TestRequirementb.MeetR.G.1.68ForPWRs,testplusanalysisforcooldownundernaturalcirculationtoconfirmadequatemixingandcooldownwithinlimitsspecifiedin EOP.Runtestsandconfirminganalysistomeetrequire-ment.V.Complies.SeeR.G.1.68re-sponse.Anat-uralcirculationtestwasperform-edonDiabloCanyon,whichissimilartoWCGS.VI.OperationalProcedurea.MeetR.G.1.33.ForPWRs,includespecificproceduresandinformationforcooldownundernaturalcirculation.Developproceduresandinformationfromtestandanalysis.VI.Complies.
Rev. 14 WOLFCREEKTABLE7.4-3(sheet5)DesignRequirementsofBTPRSB5-1Processand(SystemorComponent)PossibleSolutionforFullCompliance WCGSVII.AuxiliaryFeedwater SupplyEmergencyFeedwater Supplya.SeismicCategoryIsupplyforauxiliaryFWforatleast4hoursathotshut-downpluscool-downtoRHRcut-inbasedonlongesttimeforonlyonsiteoronlyoffsitepowerandassumedsinglefailure.FromtestsandanalysisobtainconservativeestimateofauxiliaryFWsupplytomeetrequirementandprovideseismicCategoryIsupplyVIII.Preoperationalandstartuptestestablishedtheamountofmakeupwaterrequired.Essentialservicewateristheseis-micCategoryI supply.Rev. 14 WOLF CREEK TABLE 7.4-4 RESIDUAL HEAT REMOVAL - SAFETY RELATED COLD SHUTDOWN OPERATIONS - FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks 1. Motor-operated gate valve
 
8701A (8701B analogous) a. Fails to open on demand.
Provides isola-tion of fluid
 
flow from the RCS to RHR pump 1 (pump 2).
Failure blocks RC flow from hot
 
leg of RC loop 1 through train "A" of RHRS. 
 
Fault reduces
 
redundancy of
 
RHR coolant
 
trains provide
: d. No effect on safety for
 
system operation.
Plant cooldown
 
requirements
 
will be met by
 
RC flow from hot
 
leg of RC loop 4
 
flowing through
 
train "B" of
 
RHRS. However, time required to
 
reduce RCS temperature will be extended.
 
Valve position
 
indication (Closed to open
 
position change)
 
at CB; RCS loop wide range
 
pressure indication (PI-405) at CB;
 
RHR train "A" discharge flow
 
indication (FI-
 
618) and low flow alarm at
 
CB; and RHR pump
 
discharge
 
pressure indication (PI-
 
614) at CB.
: 1. Valve is
 
electrically
 
interlocked
 
with the containment sump isolation v alve 8811A and the RWST i solation valve 8 812A, with RHR to charging
 
pump suction
 
line isolation v alve 8804A and with a "prevent-open" pressure interlock (PB-405A) of seal
 
to RC loop 1
 
hot leg. The
 
valve canno t be o pened remotely from the CB if one of the
 
indicated
 
isolation valves is open
 
or if RC loop
 
pressure exceeds 360
 
psig. The
 
valve can be
 
manually opened.
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 2)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks      2. If both trains of RHRS are u navailable for plant cooldown d ue to multiple component
 
failures, the
 
auxiliary
 
feedwater system and SG
 
atmospheric
 
relief valves
 
can be used to
 
perform the s afety function of removing residual heat.
: 2. Motor-operated gate valve 8702A (8702B
 
analogous)
Same failure modes as those
 
stated for item
 
1.a. Same function as that stated
 
for item 1.a.
Same effect on
 
system operation
 
as that stated
 
for item 1.a.
 
Same methods of
 
detection as
 
those stated for
 
item 1.a.
 
S ame remarks as those stated for item 1.a, except for
 
pressure interlock (PB-
 
403A) control.
 
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 3)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks 3. RHR pump 1, APRH (RHR pump  2 analogous)
 
Fails to deliver working fluid.
Provides fluid flow of RC through RHR heat exchanger
 
1 (heat exchanger 2) to
 
reduce RCS
 
temperature
 
during cooldown
 
operation.
Failure results in loss of RC
 
flow from hot leg of RC loop 1 through train "A" of RHRS. 
 
Fault reduces
 
redundancy or
 
RHR coolant t rains provided.
No effect on safety for
 
system operation.
Plant cooldown
 
requirements
 
will be met by
 
RC flow from hot
 
leg or RC loop 4
 
flowing through
 
train "B" of
 
RHRS. How-ever, time required to
 
reduce RCS temperature will be extended.
 
Open pump
 
switchgear
 
circuit breaker
 
indication at
 
CB; circuit
 
breaker close position monitor
 
light for group
 
monitoring of
 
components at
 
CB; common
 
breaker trip
 
alarm at CB; RCS
 
loop wide range pressure indication (PI-
 
405) at CB; RHR
 
train "A" discharge flow
 
indication (FI-
 
618) and low
 
flow alarm at
 
CB; and pump
 
discharge
 
pressure indication (PI-
 
614) at CB.
 
The RHRS shares
 
components with
 
the ECCS. Pumps
 
are tested as
 
part of the ECCS
 
testing program (see Section
 
6.3.4). Pump
 
failure may also
 
be detected
 
during ECCS
 
testing.   
 
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 4)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks 4. Motor-operated gate valve FCV-610 (FCV-
 
611 analogous)
: a. Fails to open  on demand Provides regulation of
 
fluid flow through miniflow bypass
 
line to suction
 
of RHR pump 1 (pump 2) to
 
protect against
 
overheating of
 
the pump and
 
loss of discharge flow from the pump.
Failure blocks mini-flow line
 
to suction of RHR pump "A" during cooldown
 
operation of
 
checking boron
 
concentration
 
level of coolant
 
in train "A" of
 
RHRS.
Circulation
 
through miniflow line is not available. If
 
the Operator
 
does not secure
 
RHR pump 'A'
 
before cavitation
 
occurs, failure
 
will reduce the
 
redundancy of
 
RHR coolant
 
trains. No
 
effect on safety
 
for system
 
operation.
Valve position indication (closed to open position change) at CB.
Valve is automatically
 
controlled to open when pump
 
discharge is
 
less than ~816
 
gpm and close
 
when the discharge
 
exceeds ~1,650
 
gpm.               
 
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 5)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks  b. Fails to close  on demand.
Same function as that stated for item 4.a.
Failure allows for a portion of RHR heat exchanger "A" discharge flow
 
to be bypassed
 
to suction of
 
RHR pump "A." 
 
RHRS train "A" is degraded for
 
the regulation
 
of coolant
 
temperature by RHR heat exchanger "A." 
 
No effect on
 
safety for
 
system operation. 
 
Cooldown of RCS
 
within established
 
specification
 
cooldown rate
 
may be accomplished
 
through operator
 
action of
 
throttling flow
 
control valve
 
HCV-606 and
 
controlling
 
cooldown with
 
redundant RHRS
 
train "B."
Valve position indication (open
 
to closed position change) and RHRS train "A" discharge
 
flow indication (FI-618) at CB.
 
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 6)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks 5. Air diaphragm-operated butter-fly
 
valve FCV-618 (FCV-619 analogous)
: a. Fails to open  on demand.
Controls rate of fluid flow
 
by-passed around RHR heat exchanger 1 (heat exchanger
: 2) during
 
cooldown operation.
Failure prevents coolant discharged for RHR pump "A" from bypassing
 
RHR heat exchanger "A" resulting in
 
mixed mean
 
temperature of
 
coolant flow to
 
RCS being low. 
 
RHRS train "A" is degraded for the regulation
 
of controlling
 
temperature of
 
coolant. No
 
effect on safety
 
for system
 
operation. 
 
Cooldown of RCS
 
within established
 
specification
 
rate may be
 
accomplished
 
through operator
 
action of
 
throttling flow
 
control valve
 
HCV-606 and
 
controlling
 
cooldown with
 
redundant RHRS
 
train "B."
RHR pump "A" discharge flow
 
temperature and RHRS train "A" discharge to RCS
 
cold leg flow
 
temperature
 
recording (TR-
 
612) at CB; and
 
RHRS train "A" discharge to RCS
 
cold leg flow
 
indication (FI-618) at CB.
: 1. Valve is designed to fail "closed" and is electrically
 
wired so that
 
electrical
 
solenoid of the
 
air diaphragm
 
operator is
 
energized to o pen the valve.
Valve is normally "closed" to
 
align RHRS for ECCS operation
 
during plant
 
power operation
 
and load follow.
: 2. Valve operation is not required
 
for safety grade cold shutdown operations.
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 7)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks  b. Fails to close on demand.
Same function as that stated for item 5.a.
Failure allows coolant discharged from RHR pump "A" to
 
by-pass RHR heat
 
exchanger "A",
resulting in
 
mixed mean
 
temperature of
 
coolant flow to
 
RCS being high. 
 
RHRS train "A" is degraded for the regulation of controlling
 
temperature of
 
coolant. No
 
effect on safety
 
for system
 
operation. 
 
Cooldown of RCS
 
within established
 
specification
 
rate may be
 
accomplished
 
through operator
 
action of throttling flow control valve
 
HCV-606 and
 
controlling
 
cooldown with
 
redundant RHRS
 
train "B."
: However, cooldown time
 
will be extended.
Same methods of detection as
 
those stated for item 5.a.
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 8)
Component  Failure Mode Function Effect on System Operation*** Failure Detection Method**  Remarks 6. Air diaphragm-operated butter-fly
 
valve HCV-606 (HCV-607 analogous)
: a. Fails to close on demand for flow reduction.
Controls rate of fluid flow through RHR heat exchanger 1 (heat exchanger 2)
 
during cool-
 
down operation.
Failure prevents control of
 
coolant discharge flow from RHR heat
 
exchanger "A,"
resulting in
 
loss of mixed
 
mean temperature
 
coolant flow
 
adjustment to
 
RCS. No effect
 
on safety for system operation. 
 
Cooldown of RCS
 
within established
 
specification
 
rate may be
 
accomplished by
 
operator action
 
of controlling
 
cool-down with
 
redundant RHRS
 
train "B."
Same methods of detection as
 
those stated for Item 5.a. In
: addition, monitor light
 
and alarm (valve
 
closed) for
 
group monitoring
 
of components at
 
CB.
: 1. Valve is designed to fail "open".
Valve is normally "open" to
 
align RHRS
 
for ECCS operation
 
during plant
 
power operation and
 
load follow.
: 2. Valve operation is
 
not required
 
for safety
 
grade cold
 
shutdown operations. b. Fails to open on demand for
 
increased
 
flow. Same function as that stated
 
for item 6.a.
Same effect on system operation as that stated
 
for item 6.a.
 
Same methods of
 
detection as
 
those stated for
 
item 6.a.
 
Rev. 26 WOLF CREEK TABLE 7.4-4 (sheet 9)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 7. Motor-operated gate valve 8812A (8812B analogous)
Fails to close on demand.
RWST to RHR discharge isolation.
Failure prevents isolation of RWST from RHR pump 1 (pump 2).
Negligible effect on safety for system operation.
Alternate RHR train is available by isolating RWST to RHR pump 2 (pump 1) via isolation valve 8812B (8812A).
Only effect is an increase in time required to reduce RCS temperature.
Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components.
: 1. Valve is normally open during plant operation (for alignment of ECCS). Valve interlocked so it must be closed before valves 8701A and 8702A (8701B and 8702B) can be opened. 2. See item 3 "Effect on System Operation."  8. Solenoid-operated globe valve 8154A (8154B analogous)
: a. Fails to open on demand.
Provides isolation of fluid flow from the RCS to the PRT via the excess letdown heat exchanger.
Failure reduces redundancy of providing flow from the RCS to the PRT.
Negligible effect on safety for system operation.
Letdown flow provided by parallel letdown path through alternate isolation valve 8154B (8154A).
Valve open/close position indication at CB; and letdown high temperature indication and alarm at CB.
The letdown path to the PRT provides fluid flow out of the RCS to accommodate boration makeup flow into the RCS. Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 10)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks  b. Fails to close  on demand.
Same function as that stated for item 8.a.
Failure reduces redundancy of iso-lation flow from the RCS to the PRT.
Negligible effect on safety for system operation. RCS letdown flow isolation provided by alternate series isolation valve 8153A (8153B).
Same methods of detection as those stated for item 8.a.
: 9. Solenoid-operated globe valve 8153A (8153B analogous)
: a. Fails to open on demand.
Same function as that stated for item 8.a.
Same effect on system operation as that stated for item 8.a, except for alterante isolation valve 8153B (8153A).
Same methods of detection as those stated for item 8.a.
Same remarks as those stated for item 8.a.
: b. Fails to close on demand.
Same function as that stated for item 8.a.
Same effect on system operation as that stated for item 8.a, except for alternate isolation valve 8153A (8153B).
Same methods of detection as those stated for item 8.a.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 11)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 10. Motor-operated globe valve 8157A (8157B analogous)
Fails to open on demand.
Provides safety grade letdown flow path.
Same effect on system operation as that stated for item 8.a, except for alternate parallel isolation valve 8157B (8157A). Same methods of detection as those stated for item 8.a.
Same remarks as those stated for item 8.a.
: 11. Solenoid-operated power-operated relief valve PCV-456A (PCV-455A analogous)
: a. Fails to open on demand. Provides isolation of fluid flow from pres- surizer to PRT. Failure reduces redundancy of providing flow from pressurizer to PRT.
Negligible effect on safety for system operation.
Pressurizer vent flow provided by a parallel pressurizer vent path through alternate isolation valve PCV-455A.
Valve open/closed position indication at CB; pressurizer poweroperated relief valve outlet temperature indication at CB. Pressurizer vent path to the PRT provides fluid flow out of the RCS to permit RCS depressurizatio n to RHRS initiation conditions.
: b. Fails to close on demand. Same function as that stated for item 11.a.
Failure reduces redundancy of isolating flow from the pressurizer to the PRT.
Negligible effect on safety for system operation.
Pressurizer vent flow isolation provided by alternate series isolation valve 8000B (8000A).
Same methods of detection as those stated for item 11.a.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 12)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 12. Motor-operated gate valve 8000A (8000B analogous)
Fails to close on demand.
Same function as that stated for item 11.a.
Same effect on system operation as that stated for item 11.a, except pressurizer vent flow isolation provided by alternate series isolation valve PCV-455A (PCV-456A). Same methods of detection as those stated for item 11.a.
Same remarks as those stated for item 11.a.
: 13. Motor-operated gate valve 8808A 8808B, 8808C, 8808D analogous)
Fails to close on demand.
Provides isolation of fluid flow from accumulator 1 (accumulator 2, accumulator 3, accumulator 4) to the RCS.
Failure prevents isolation of accumulator 1 (accumulator 2, accumulator 3, accumulator 4) from the RCS.
Negligible effect on safety for system operation.
Accumulator 1 (accumulator 2, accumulator 3, accumulator  4) is depressurized by opening vent isolation 8950A (8950B or C, 8950D or E, 8950F). Valve open/closed position indication at CB; valve (closed) monitor light and alarm at CB; and accumulator pressure indication and low alarm at CB.
Accumulators are isolated or vented during plant cooldown to not affect RCS depressurization to RHRS initiation conditions.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 13)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 14. Solenoid operated globe valve 8950A (8950F analogous)
Fails to open on demand.
Provides venting of nitrogen gas from accumulator 1 (accumulator 4) to containment.
Failure prevents venting of accumulator 1 (accumulator 4) to containment.
Negligible effect on safety for system operation.
Accumulator 1 (accumulator 4) is isolated from RCS by closing isolation valve 8808A (8808D).
Valve open/closed position indication at CB and accumulator pressure indication and low alarm at CB.
Same remarks as those stated for item 13. 15. Solenoid-operated globe valve 8950B/8950D (8950C/8950E analogous)
Fails to open on demand.
Provides venting of nitrogen gas from accumulator 2/accumulator
: 3. Failure reduces redundancy in venting accumulator 2/accumulator 3.
Negligible effect on safety for system operation.
Accumulator 2/accumulator 3 venting capability provided by valves 8950C/8950E if accumulator isolation valves 8808B/8808C cannot be closed. Same methods of detection as those stated for item 14. Same remarks as those stated for item 13. Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 14)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 16. Centrifugal charging pump 1 (pump 2 analogous)
Fails to deliver working fluid. Provides fluid flow of borated water from the RWST to the RCS. Failure reduces redundancy of providing borated water to the RCS at high RCS pressures.
Fluid flow from charging pump 1 (pump 2) will be lost. Minimum flow requirements for boration and makeup will be met by charging pump 2 (pump 1).
Charging pump discharge header pressure and flow indication at CB. Open/closed pump switch gear circuit breaker indication on CB. Circuit breaker closed position monitor light for group monitoring of c omponent at CB.
Common breaker trip alarm at CB. The charging pumps provide boration, seal injection, and makeup flow to the RCS during safety grade cold shut-down operations.
: 17. Motor-operated gate valve LCV 112C (LCV-112B analogous)
Fails to close on demand.
Provides isolation of fluid discharge from the VCT to the suction of charging pumps Failure reduces redundancy of providing VCT discharge isolation.
Negligible effect on safety for system operation.
Alternate isolation valve LCV-112B (LCV-112C) provides back-up tank discharge isolation.
Valve open/closed position indication at CB and valve (closed) monitor light and alarm at CB. The charging pumps' suction is isolated from the VCT and aligned to the RWST (for boration/make-up) during safety grade cold shutdown operations.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 15)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 18. Motor-operated gate valve 8105 (8106 analogous)
Fails to close on demand.
Provides isolation of fluid flow from the charging pump discharge header to the CVCS normal charging line to the RCS.
Failure reduces redundancy of providing isolation of charging pump discharge to normal charging line of CVCS.
Negligible effect on safety for system operation.
Alternate isolation valve 8106 (8105) provides backup normal CVCS charging line isolation.
Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components at CB. Normal charging line is isolated during safety grade cold shutdown operations.
Boration and makeup flow provided to RCS through redundant ECCS headers to the RCS cold legs.
: 19. Motor-operated  gate valve LCV-112E (LCV-112D analogous)
Fails to open on demand.
Provides isolation of fluid discharge from the RWST to the suction of charging pumps. Failure reduces redundancy of providing fluid flow from RWST to suction of charging pump 2.
Negligible effect on safety for system operation.
Alternate isolation valve LCV-112D (LCV-112E) opens to provide backup flow path to suction of charging pump 1.
Valve open/closed position indication at CB and valve (open) monitor light and alarm at CB.
The charging pumps' suction is aligned to the RWST for make-up/boration to the RCS during safety grade cold shutdown operations.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 16)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 20. Motor-operated globe valve 8110 (8111 analogous)
Fails to open on demand.
Provides isolation of charging pump mini-flow line.
Failure reduces redundancy of providing boration/make-up flow from the RWST to the RCS under low flow throttled conditions where charging pump minimum flow requirements cannot be met with-out mini-flow. Negligible effect on safety for system operation.
Alternate charging-pump 2 (charging-pump
: 1) minimum flow requirements will be met utilizing mini-flow isolation valve 8111 (8110). Boration/makeup flow requirements are satisfied by the redundant alternate train.
Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components at CB. 1. Valve aligned to close upon receipt of a safe g uards "S" signal. 2. Normally open valve.          Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 17)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 21. Motor-operated globe valve 8357A (8357B analogous)
Fails to open on demand.
Provides safety grade seal injection flow path. Failure reduces redundancy of providing seal injection flow to the RCP seals. Negligible effect on safety for system operation.
Alternate valve 8357B (8357A) opens to provide a seal injection flow path to the RCPs. Seal injection flow requirements are satisfied by the redundant alternate path.
Valve open/closed position indication at CB. 22. Motor-operated gate valve 8716A (8716B analogous)
Fails to close on demand.
Provides separation between the two RHR trains during cooldown operation.
Failure reduces retion RHR trains during cooldown.
Negligible effect on system operation.
Isolation valve 8716B (8716A) provides backup isolation between the two RHR trains.
Valve open/closed position indication at CB and valve (closed) monitor light and alarm at CB.      Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 18)
Component  Failure Mode Function Effect on System Operation*  Failure Detection Method**  Remarks 23. Motor-operated gate valve 8037A (8037B analogous)
Fails to open on demand.
PRT to containment sump isolation.
Failure reduces redundancy of providing flow from the PRT to containment sump. Negligible effect on safety for system operation.
Letdown flow provided by parallel path through alternate isolation valve 8037B (8037A).
Valve position indication (closed to open position change) at CB. Valve open position monitor light and alarm for group monitoring of components.
Letdown path to containment sump provides flow out of PRT to accommodate flow out of RCS during shutdown operations.
: 24. Motor-operated gate valve 8801A (8801B analogous)
Fails to open on demand.
BIT discharge to RCS. Failure reduces redundancy of providing flow via the BIT to RCS. Negligible effect on safety for system operation. Flow path provided by parallel isolation valve 8801B (8801A).
Valve position indication (closed to open position change) at CB. Valve open position monitor light and alarm for group monitoring of components.
Path utilized for boration/makeup flow to RCS for safety grade cold shutdown operation.
Rev. 14 WOLF CREEK TABLE 7.4-4 (sheet 19)
List of acronyms and abbreviations
 
CB  - Control board
 
CVCS  -Chemical and volume control system
 
ECCS  -Emergency core cooling system RC  - Reactor coolant RCS  - Reactor coolant system RHR  - Residual heat removal
 
RHRS  -Residual heat removal system
 
RWST  -Refueling water storage tank SG  - Steam generator
 
BIST  -Boron injection surge tank RCP  - Reactor coolant pump VCT  - Volume control tank PRT  - Pressurizer relief tank BIT  - Boron injection tank
* 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.
*** If failure occurs on the single RHR SDC train when RCS temperature is between 350 F - 225 F the RHR ECCS standby train will be placed in the SDC mode.
 
Rev. 26 WOLFCREEKTABLE7.4-5SYSTEMSREQUIREDTOACHIEVEANDMAINTAINPOST-ACCIDENTSAFESHUTDOWNReactorCoolantSystem(SeeChapter5.0)MainSteamSystem(SeeSection10.3)AuxiliaryFeedwaterSystem(SeeSection10.4.9)ChemicalandVolumeControlSystem(SeeSection9.3.4)BoratedRefuelingWaterSystem(SeeSection6.3)ResidualHeatRemovalSystem(SeeSection5.4.7)ComponentCoolingWaterSystem(SeeSection9.2.2)EssentialServiceWaterSystem(SeeSection9.2.1.2)SupportiveHVACSystems(SeeSection9.4)EmergencyDieselGenerators(SeeSections9.5.4through9.5.8)FuelPoolCoolingSystem(SeeSection9.1.3)SupportivePortionsofInstrumentAirSystem(SeeSection9.3.1)SupportivePortionsofElectricalDistributionSystem(SeeChapter8.0)Rev.14 WOLF CREEK TABLE 7.4-6 POST-ACCIDENT SAFE SHUTDOWN COMPONENTS (EXTRACTED FROM TABLE 3.11(B)-3)
Component ID Component name Room NoSpec No Hot SD Cold SDAB007 AUX. RELAY RACK 3413 E-093 X X AB008 AUX. RELAY RACK 1320 E-093 X X AB009 AUX. RELAY RACK 1408 E-093 X X ABFHC0002 ABPV0002 Local Controller 1509 J-601B X X ABFHC0003 ABPV0003 Local Controller 1509 J-601B X X ABHV0005 VALVE TERMINAL BOX 1412 E-028 X X ABHV0006 VALVE TERMINAL BOX 1412 E-028 X X ABHV0011 MAIN STEAM ISO VALVE LOOP 4 1508 M-628 X X ABHV0012  MAIN STEAM ISOL BYPASS VALVE LOOP 4 1508 J-601A X X ABHV0014 MAIN STEAM ISO VALVE LOOP 1 1508 M-628 X X ABHV0015  MAIN STEAM ISOL BYPASS VALVE LOOP 1 1508 J-601A X X ABHV0017 MAIN STEAM ISO VALVE LOOP 2 1509 M-628 X X ABHV0018  MAIN STEAM ISOL BYPASS VALVE LOOP 2 1509 J-601 X X ABHV0020 MAIN STEAM ISO VALVE LOOP 3 1509 M-628 X X ABHV0021  MAIN STEAM ISOL BYPASS VALVE LOOP 3 1509 J-601A X X ABHV0048 VALVE TERMINAL BOX 1412 E-028 X X ABHV0049 VALVE TERMINAL BOX 1412 E-028 X X ABHY0005 ABHV0005 SOLENOID VALVE 1412 J-601A X X ABHY0006 ABHV0006 SOLENOID VALVE 1412 J-601A X X ABHY0012A ABHV0012 SOLENOID VALVE 1508 J-601A X X ABHY0012A VALVE TERMINAL BOX (ABZS0012A) 1508 E-028 X X ABHY0012B ABHV0012 SOLENOID VALVE 1508 J-601A X X ABHY0012B VALVE TERMINAL BOX (ABZS0012B) 1508 E-028 X X ABHY0015A ABHV0015 SOLENOID VALVE 1508 J-601A X X ABHY0015A VALVE TERMINAL BOX (ABZS0015A) 1508 E-028 X X ABHY0015B ABHV0015 SOLENOID VALVE 1508 J-601A X X ABHY0015B VALVE TERMINAL BOX (ABZS0015B) 1508 E-028 X X ABHY0018A ABHV0018 SOLENOID VALVE 1509 J-601A X X ABHY0018A VALVE TERMINAL BOX (ABZS0018A) 1509 E-028 X X ABHY0018B ABHV0018 SOLENOID VALVE 1509 J-601A X X ABHY0018B VALVE TERMINAL BOX (ABZS0018B) 1509 E-028 X X ABHY0021A ABHV0021 SOLENOID VALVE 1509 J-601A X X ABHY0021A VALVE TERMINAL BOX (ABZS0021A) 1509 E-028 X X ABHY0021B ABHV0021 SOLENOID VALVE 1509 J-601A X X ABHY0021B VALVE TERMINAL BOX (ABZS0021B) 1509 E-028 X X ABHY0048A ABHV0048 SOLENOID VALVE 1412 J-601A X X ABHY0049A ABHV0049 SOLENOID VALVE 1412 J-601A X X ABLT0007  MAIN STM LINE DRN VLV LP 3 LV 1331 J-301 X X ABLT0008  MAIN STM LINE DRN VLV LP 2 LV 1331 J-301 X X ABLT0009  MAIN STM LINE DRN VLV LP 1 LV 1326 J-301 X X ABLT0010  MAIN STM LINE DRN VLV LP 4 LV 1325 J-301 X X ABLV0007 MAIN STEAM LINE DRAIN VALVE LOOP 3 LEVEL 1412 J-601A X X ABLV0008 MAIN STEAM LINE DRAIN VALVE LOOP 2 LEVEL 1412 J-601A X X ABLV0009 MAIN STEAM LINE DRAIN VALVE LOOP 1 LEVEL 1411 J-601A X X ABLV0010 MAIN STEAM LINE DRAIN VALVE LOOP 4 LEVEL 1411 J-601A X X ABLY0007A ABLV0007 SOLENOID VALVE 1412 J-601A X X ABLY0007A VALVE TERMINAL BOX (ABLY0008B) 1412 E-028 X X
 
Rev. 21 WOLF CREEK TABLE 7.4-6 (Sheet 2)
Component ID Component name Room NoSpec No Hot SD Cold SDABLY0007B ABLV0007 SOLENOID VALVE 1412 J-601A X X ABLY0007B VALVE TERMINAL BOX (ABLY0008A) 1412 E-028 X X ABLY0008A ABLV0008 SOLENOID VALVE 1412 J-601A X X ABLY0008B ABLV0008 SOLENOID VALVE 1412 J-601A X X ABLY0009A ABLV0009 SOLENOID VALVE 1411 J-601A X X ABLY0009A VALVE TERMINAL BOX (ABLY0010B) 1411 E-028 X X ABLY0009B ABLV0009 SOLENOID VALVE 1411 J-601A X X ABLY0009B VALVE TERMINAL BOX (ABLY0010A) 1411 E-028 X X ABLY0010A ABLV0010 SOLENOID VALVE 1411 J-601A X X ABLY0010B ABLV0010 SOLENOID VALVE 1411 J-601A X X ABPI0514A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0515A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0516A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0516X STEAMLINE PRESSURE 1413 ESE-14 X X ABPI0524A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0524B STEAMLINE PRESSURE 1413 ESE-14 X X ABPI0525A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0526A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0534A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0535A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0535X STEAMLINE PRESSURE 1413 ESE-14 X X ABPI0536A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0544A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0544B STEAMLINE PRESSURE 1413 ESE-14 X X ABPI0545A STEAMLINE PRESSURE 3601 ESE-14 X X ABPI0546A STEAMLINE PRESSURE 3601 ESE-14 X X ABPIC0001A STM GEN A ATM STEAM DUMP 3601 J-110 X X ABPIC0001B STM GEN A ATM STEAM DUMP 1413 J-110 X X ABPIC0002A STM GEN B ATM STEAM DUMP 3601 J-110 X X ABPIC0002B STM GEN B ATM STEAM DUMP 1413 J-110 X X ABPIC0003A STM GEN C ATM STEAM DUMP 3601 J-110 X X ABPIC0003B STM GEN C ATM STEAM DUMP 1413 J-110 X X ABPIC0004A STM GEN D ATM STEAM DUMP 3601 J-110 X X ABPIC0004B STM GEN D ATM STEAM DUMP 1413 J-110 X X ABPSL0011A ABHV0011 ACCUM PRESS SWITCH 1508 M-628 X X ABPSL0011B ABHV0011 ACCUM PRESS SWITCH 1508 M-628 X X ABPSL0014A ABHV0014 ACCUM PRESS SWITCH 1508 M-628 X X ABPSL0014B ABHV0014 ACCUM PRESS SWITCH 1508 M-628 X X ABPSL0017A ABHV0017 ACCUM PRESS SWITCH 1509 M-628 X X ABPSL0017B ABHV0017 ACCUM PRESS SWITCH 1509 M-628 X X ABPSL0020A ABHV0020 ACCUM PRESS SWITCH 1509 M-628 X X ABPSL0020B ABHV0020 ACCUM PRESS SWITCH 1509 M-628 X X ABPT0001 STM GEN A STEAMLINE PRESSURE 1304 J-301 X X ABPT0002 STM GEN B STEAMLINE PRESSURE 1305 J-301 X X ABPT0003 STM GEN C STEAMLINE PRESSURE 1305 J-301 X X ABPT0004 STM GEN D STEAMLINE PRESSURE 1304 J-301 X X ABPT0011A  ABHV0011 ACCUM PRESS TRANSMITTER 1508 M-628 X X ABPT0011B  ABHV0011 ACCUM PRESS TRANSMITTER 1508 M-628 X X ABPT0014A  ABHV0014 ACCUM PRESS TRANSMITTER 1508 M-628 X X ABPT0014B  ABHV0014 ACCUM PRESS TRANSMITTER 1508 M-628 X X ABPT0017A  ABHV0017 ACCUM PRESS TRANSMITTER 1509 M-628 X X ABPT0017B  ABHV0017 ACCUM PRESS TRANSMITTER 1509 M-628 X X ABPT0020A  ABHV0020 ACCUM PRESS TRANSMITTER 1509 M-628 X X ABPT0020B  ABHV0020 ACCUM PRESS TRANSMITTER 1509 M-628 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 3)
Component ID Component name Room NoSpec No Hot SD Cold SDABPT0514 MAIN STEAM PRESSURE LOOP 1 1411 ESE-1A X X ABPT0515 MAIN STEAM PRESSURE LOOP 1 1411 ESE-1A X X ABPT0516 MAIN STEAM PRESSURE LOOP 1 1411 ESE-1A X X ABPT0524 MAIN STEAM PRESSURE LOOP 2 1412 ESE-1A X X ABPT0525 MAIN STEAM PRESSURE LOOP 2 1412 ESE-1A X X ABPT0526 MAIN STEAM PRESSURE LOOP 2 1412 ESE-1A X X ABPT0534 MAIN STEAM PRESSURE LOOP 3 1412 ESE-1A X X ABPT0535 MAIN STEAM PRESSURE LOOP 3 1412 ESE-1A X X ABPT0536 MAIN STEAM PRESSURE LOOP 3 1412 ESE-1A X X ABPT0544 MAIN STEAM PRESSURE LOOP 4 1411 ESE-1A X X ABPT0545 MAIN STEAM PRESSURE LOOP 4 1411 ESE-1A X X ABPT0546 MAIN STEAM PRESSURE LOOP 4 1411 ESE-1A X X ABPV0001 VALVE TERMINAL BOX 1508 E-028 X X ABPV0001 STM GEN A ATM RELIEF VLV 1508 J-601B X X ABPV0002 VALVE TERMINAL BOX 1509 E-028 X X ABPV0002 STM GEN B ATM RELIEF VLV 1509 J-601B X X ABPV0003 VALVE TERMINAL BOX 1509 E-028 X X ABPV0003 STM GEN C ATM RELIEF VLV 1509 J-601B X X ABPV0004 VALVE TERMINAL BOX 1508 E-028 X X ABPV0004 STM GEN D ATM RELIEF VLV 1508 J-601B X X ABPY0001 ABPV0001 I/P CONVERTER 1508 J-601B X X ABPY0002 ABPV0002 I/P CONVERTER 1509 J-601B X X ABPY0003 ABPV0003 I/P CONVERTER 1509 J-601B X X ABPY0004 ABPV0004 I/P CONVERTER 1508 J-601B X X ABV0045  SAFETY VALVES LOOP 4 1508 M-140 X X ABV0046  SAFETY VALVES LOOP 4 1508 M-140 X X ABV0047  SAFETY VALVES LOOP 4 1508 M-140 X X ABV0048  SAFETY VALVES LOOP 4 1508 M-140 X X ABV0049  SAFETY VALVES LOOP 4 1508 M-140 X X ABV0055  SAFETY VALVES LOOP 1 1508 M-140 X X ABV0056  SAFETY VALVES LOOP 1 1508 M-140 X X ABV0057  SAFETY VALVES LOOP 1 1508 M-140 X X ABV0058  SAFETY VALVES LOOP 1 1508 M-140 X X ABV0059  SAFETY VALVES LOOP 1 1508 M-140 X X ABV0065  SAFETY VALVES LOOP 2 1509 M-140 X X ABV0066  SAFETY VALVES LOOP 2 1509 M-140 X X ABV0067  SAFETY VALVES LOOP 2 1509 M-140 X X ABV0068  SAFETY VALVES LOOP 2 1509 M-140 X X ABV0069  SAFETY VALVES LOOP 2 1509 M-140 X X ABV0075  SAFETY VALVES LOOP 3 1509 M-140 X X ABV0076  SAFETY VALVES LOOP 3 1509 M-140 X X ABV0077  SAFETY VALVES LOOP 3 1509 M-140 X X ABV0078  SAFETY VALVES LOOP 3 1509 M-140 X X ABV0079  SAFETY VALVES LOOP 3 1509 M-140 X X ABZC0001  ABPV0001 POSITIONER 1508 J-601B X X ABZC0002  ABPV0002 POSITIONER 1509 J-601B X X ABZC0003  ABPV0003 POSITIONER 1509 J-601B X X ABZC0004  ABPV0004 POSITIONER 1508 J-601B X X ABZS0001 ABPV0001 LIMIT SWITCH 1508 J-601B X X ABZS0002 ABPV0002 LIMIT SWITCH 1509 J-601B X X ABZS0003 ABPV0003 LIMIT SWITCH 1509 J-601B X X ABZS0004 ABPV0004 LIMIT SWITCH 1508 J-601B X X ABZS0005 ABHV0005 LIMIT SWITCH 1412 J-601A X X ABZS0006 ABHV0006 LIMIT SWITCH 1412 J-601A X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 4)
Component ID Component name Room NoSpec No Hot SD Cold SDABZS0007A ABLV0007 LIMIT SWITCH 1412 J-601A X X ABZS0007B ABLV0007 LIMIT SWITCH 1412 J-601A X X ABZS0008A ABLV0008 LIMIT SWITCH 1412 J-601A X X ABZS0008B ABLV0008 LIMIT SWITCH 1412 J-601A X X ABZS0009A ABLV0009 LIMIT SWITCH 1411 J-601A X X ABZS0009B ABLV0009 LIMIT SWITCH 1411 J-601A X X ABZS0010A ABLV0010 LIMIT SWITCH 1411 J-601A X X ABZS0010B ABLV0010 LIMIT SWITCH 1411 J-601A X X ABZS0011A ABHV0011 LIMIT SWITCH 1508 M-628 X X ABZS0011A MSIV LIMIT SWITCH CONNECTOR LOOP 4 1508 HE-8 X X ABZS0011B ABHV0011 LIMIT SWITCH 1508 M-628 X X ABZS0011B MSIV LIMIT SWITCH CONNECTOR LOOP 4 1508 HE-8 X X ABZS0011C ABHV0011 LIMIT SWITCH 1508 M-628 X X ABZS0011C MSIV LIMIT SWITCH CONNECTOR LOOP 4 1508 HE-8 X X ABZS0011D ABHV0011 LIMIT SWITCH 1508 M-628 X X ABZS0011D MSIV LIMIT SWITCH CONNECTOR LOOP 4 1508 HE-8 X X ABZS0012A ABHV0012 LIMIT SWITCH 1508 J-601A X X ABZS0012B ABHV0012 LIMIT SWITCH 1508 J-601A X X ABZS0014A ABHV0014 LIMIT SWITCH 1508 M-628 X X ABZS0014A MSIV LIMIT SWITCH CONNECTOR LOOP 1 1508 HE-8 X X ABZS0014B ABHV0014 LIMIT SWITCH 1508 M-628 X X ABZS0014B MSIV LIMIT SWITCH CONNECTOR LOOP 1 1508 HE-8 X X ABZS0014C ABHV0014 LIMIT SWITCH 1508 M-628 X X ABZS0014C MSIV LIMIT SWITCH CONNECTOR LOOP 1 1508 HE-8 X X ABZS0014D ABHV0014 LIMIT SWITCH 1508 M-628 X X ABZS0014D MSIV LIMIT SWITCH CONNECTOR LOOP 1 1508 HE-8 X X ABZS0015A ABHV0015 LIMIT SWITCH 1508 J-601A X X ABZS0015B ABHV0015 LIMIT SWITCH 1508 J-601A X X ABZS0017A ABHV0017 LIMIT SWITCH 1509 M-628 X X ABZS0017A MSIV LIMIT SWITCH CONNECTOR LOOP 2 1509 HE-8 X X ABZS0017B ABHV0017 LIMIT SWITCH 1509 M-628 X X ABZS0017B MSIV LIMIT SWITCH CONNECTOR LOOP 2 1509 HE-8 X X ABZS0017C ABHV0017 LIMIT SWITCH 1509 M-628 X X ABZS0017C MSIV LIMIT SWITCH CONNECTOR LOOP 2 1509 HE-8 X X ABZS0017D ABHV0017 LIMIT SWITCH 1509 M-628 X X ABZS0017D MSIV LIMIT SWITCH CONNECTOR LOOP 2 1509 HE-8 X X ABZS0018A ABHV0018 LIMIT SWITCH 1509 J-601A X X ABZS0018B ABHV0018 LIMIT SWITCH 1509 J-601A X X ABZS0020A ABHV0020 LIMIT SWITCH 1509 M-628 X X ABZS0020A MSIV LIMIT SWITCH CONNECTOR LOOP 3 1509 HE-8 X X ABZS0020B ABHV0020 LIMIT SWITCH 1509 M-628 X X ABZS0020B MSIV LIMIT SWITCH CONNECTOR LOOP 3 1509 HE-8 X X ABZS0020C ABHV0020 LIMIT SWITCH 1509 M-628 X X ABZS0020C MSIV LIMIT SWITCH CONNECTOR LOOP 3 1509 HE-8 X X ABZS0020D ABHV0020 LIMIT SWITCH 1509 M-628 X X ABZS0020D MSIV LIMIT SWITCH CONNECTOR LOOP 3 1509 HE-8 X X ABZS0021A ABHV0021 LIMIT SWITCH 1509 J-601A X X ABZS0021B ABHV0021 LIMIT SWITCH 1509 J-601A X X ABZS0048 ABHV0048 LIMIT SWITCH 1412 J-601A X X ABZS0049 ABHV0049 LIMIT SWITCH 1412 J-601A X X AEFV0039 FEEDWATER ISOLATION VALVE LOOP 1 1411 M-630 X X AEFV0040 FEEDWATER ISOLATION VALVE LOOP 2 1412 M-630 X X AEFV0041 FEEDWATER ISOLATION VALVE LOOP 3 1412 M-630 X X AEFV0042 FEEDWATER ISOLATION VALVE LOOP 4 1411 M-630 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 5)
Component ID Component name Room NoSpec No Hot SD Cold SD AEFV0043  STM GEN A CHEM CONTROL 1411 J-601A X X AEFV0044  STM GEN B CHEM CONTROL 1412 J-601A X X AEFV0045  STM GEN C CHEM CONTROL 1412 J-601A X X AEFV0046  STM GEN D CHEM CONTROL 1411 J-601A X X AEFY0043 AEFV0043 SOLENOID VALVE 1411 J-601A X X AEFY0044 AEFV0044 SOLENOID VALVE 1412 J-601A X X AEFY0045 AEFV0045 SOLENOID VALVE 1412 J-601A X X AEFY0046 AEFV0046 SOLENOID VALVE 1411 J-601A X X AELI0501 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0501A STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0502 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0502A STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0503 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0503A STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0504 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0504A STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0517 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0517X STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0518 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0519 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0527 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0528 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0528X STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0529 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0537 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0537X STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0538 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0539 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0547 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0548 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELI0548X STEAM GENERATOR WATER LEVEL 1413 ESE-14 X X AELI0549 STEAM GENERATOR WATER LEVEL 3601 ESE-14 X X AELT0501 SG LEVEL WIDE RANGE LOOP 1 2201 ESE-3C X X AELT0501 SG LEVEL WIDE RANGE LOOP 1 2201 ESE-3A X X AELT0502 SG LEVEL WIDE RANGE LOOP 2 2201 ESE-3C X X AELT0502 SG LEVEL WIDE RANGE LOOP 2 2201 ESE-3A X X AELT0503 SG LEVEL WIDE RANGE LOOP 3 2201 ESE-3C X X AELT0503 SG LEVEL WIDE RANGE LOOP 3 2201 ESE-3A X X AELT0504 SG LEVEL WIDE RANGE LOOP 4 2201 ESE-3C X X AELT0504 SG LEVEL WIDE RANGE LOOP 4 2201 ESE-3A X X AELT0517 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3A X X AELT0517 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3C X X AELT0518 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3A X X AELT0518 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3C X X AELT0519 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3A X X AELT0519 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3C X X AELT0527 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3C X X AELT0527 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3A X X AELT0528 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3A X X AELT0528 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3C X X AELT0529 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3A X X AELT0529 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3C X X AELT0537 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3A X X AELT0537 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3C X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 6)
Component ID Component name Room NoSpec No Hot SD Cold SDAELT0538 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3C X X AELT0538 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3A X X AELT0539 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3A X X AELT0539 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3C X X AELT0547 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3C X X AELT0547 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3A X X AELT0548 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3A X X AELT0548 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3C X X AELT0549 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3A X X AELT0549 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3C X X AELT0551 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3C X X AELT0551 SG LEVEL NARROW RANGE LOOP 1 2201 ESE-3A X X AELT0552 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3C X X AELT0552 SG LEVEL NARROW RANGE LOOP 2 2201 ESE-3A X X AELT0553 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3C X X AELT0553 SG LEVEL NARROW RANGE LOOP 3 2201 ESE-3A X X AELT0554 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3C X X AELT0554 SG LEVEL NARROW RANGE LOOP 4 2201 ESE-3A X X AEPSL0039A AEFV0039 ACCUM PRESS SWITCH 1411 M-630 X X AEPSL0039B AEFV0039 ACCUM PRESS SWITCH 1411 M-630 X X AEPSL0040A AEFV0040 ACCUM PRESS SWITCH 1412 M-630 X X AEPSL0040B AEFV0040 ACCUM PRESS SWITCH 1412 M-630 X X AEPSL0041A AEFV0041 ACCUM PRESS SWITCH 1412 M-630 X X AEPSL0041B AEFV0041 ACCUM PRESS SWITCH 1412 M-630 X X AEPSL0042A AEFV0042 ACCUM PRESS SWITCH 1411 M-630 X X AEPSL0042B AEFV0042 ACCUM PRESS SWITCH 1411 M-630 X X AEPT0039A  AEFV0039 ACCUM PRESS TRANSMITTER 1411 M-630 X X AEPT0039B  AEFV0039 ACCUM PRESS TRANSMITTER 1411 M-630 X X AEPT0040A  AEFV0040 ACCUM PRESS TRANSMITTER 1412 M-630 X X AEPT0040B  AEFV0040 ACCUM PRESS TRANSMITTER 1412 M-630 X X AEPT0041A  AEFV0041 ACCUM PRESS TRANSMITTER 1412 M-630 X X AEPT0041B  AEFV0041 ACCUM PRESS TRANSMITTER 1412 M-630 X X AEPT0042A  AEFV0042 ACCUM PRESS TRANSMITTER 1411 M-630 X X AEPT0042B  AEFV0042 ACCUM PRESS TRANSMITTER 1411 M-630 X X AEV0120  FEEDWATER ISOL VALVE LOOP 2 (CHECK) 2201 M-224A X X AEV0121  FEEDWATER ISOL VALVE LOOP 1 (CHECK) 2201 M-224A X X AEV0122  FWTR ISO VLV LOOP 4 (CHECK) 2201 M-224A X X AEV0123  FEEDWATER ISO VALVE LOOP 3 (CHECK) 2201 M-224A X X AEV0124  AUX FDWTR ISO VLV LOOP 2 (CHECK) 1305 M-224B X X AEV0125  AUX FEEDWATER ISOL VALVE LOOP 1 (CHECK)1304 M-224B X X AEV0126  AUX FWTR ISO VLV LOOP 4 (CHECK) 1304 M-224B X X AEV0127  AUX FWTR ISO VALVE LOOP 3 (CHECK) 1305 M-224B X X AEV0132  AMMONIA & HYDRAZINE ISO VLV LOOP 2 (CHECK) 1412 M-231C X X AEV0133  AMMONIA & HYDRAZINE ISO VLV LOOP 1 (CHECK) 1411 M-231C X X AEV0134  AMMONIA & HYDRAZINE ISO VLV LOOP 4 (CHECK) 1411 M-231C X X AEV0135  AMMONIA & HYDRAZINE ISO VALVE LOOP 3 (CHECK) 1412 M-231C X X AEV0702  S.G. A LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0704  S.G. A LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0706  S.G. B LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0708  S.G. B LT VENT ISO VALVE (MAN) 2201 J-705 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 7)
Component ID Component name Room NoSpec No Hot SD Cold SD AEV0710  S.G. C LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0712  S.G. C LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0714  S.G. D LT VENT ISO VALVE (MAN) 2201 J-705 X X AEV0716  S.G. D LT VENT ISO VALVE (MAN) 2201 J-705 X X AEZS0039A AEFV0039 LIMIT SWITCH 1411 M-630 X X AEZS0039A MFIV LIMIT SWITCH CONNECTOR LOOP 1 1411 HE-8 X X AEZS0039B AEFV0039 LIMIT SWITCH 1411 M-630 X X AEZS0039B MFIV LIMIT SWITCH CONNECTOR LOOP 1 1411 HE-8 X X AEZS0039C AEFV0039 LIMIT SWITCH 1411 M-630 X X AEZS0039C MFIV LIMIT SWITCH CONNECTOR LOOP 1 1411 HE-8 X X AEZS0039D AEFV0039 LIMIT SWITCH 1411 M-630 X X AEZS0039D MFIV LIMIT SWITCH CONNECTOR LOOP 1 1411 HE-8 X X AEZS0040A AEFV0040 LIMIT SWITCH 1412 M-630 X X AEZS0040A MFIV LIMIT SWITCH CONNECTOR LOOP 2 1412 HE-8 X X AEZS0040B AEFV0040 LIMIT SWITCH 1412 M-630 X X AEZS0040B MFIV LIMIT SWITCH CONNECTOR LOOP 2 1412 HE-8 X X AEZS0040C AEFV0040 LIMIT SWITCH 1412 M-630 X X AEZS0040C MFIV LIMIT SWITCH CONNECTOR LOOP 2 1412 HE-8 X X AEZS0040D AEFV0040 LIMIT SWITCH 1412 M-630 X X AEZS0040D MFIV LIMIT SWITCH CONNECTOR LOOP 2 1412 HE-8 X X AEZS0041A AEFV0041 LIMIT SWITCH 1412 M-630 X X AEZS0041A MFIV LIMIT SWITCH CONNECTOR LOOP 3 1412 HE-8 X X AEZS0041B AEFV0041 LIMIT SWITCH 1412 M-630 X X AEZS0041B MFIV LIMIT SWITCH CONNECTOR LOOP 3 1412 HE-8 X X AEZS0041C AEFV0041 LIMIT SWITCH 1412 M-630 X X AEZS0041C MFIV LIMIT SWITCH CONNECTOR LOOP 3 1412 HE-8 X X AEZS0041D AEFV0041 LIMIT SWITCH 1412 M-630 X X AEZS0041D MFIV LIMIT SWITCH CONNECTOR LOOP 3 1412 HE-8 X X AEZS0042A AEFV0042 LIMIT SWITCH 1411 M-630 X X AEZS0042A MFIV LIMIT SWITCH CONNECTOR LOOP 4 1411 HE-8 X X AEZS0042B AEFV0042 LIMIT SWITCH 1411 M-630 X X AEZS0042B MFIV LIMIT SWITCH CONNECTOR LOOP 4 1411 HE-8 X X AEZS0042C AEFV0042 LIMIT SWITCH 1411 M-630 X X AEZS0042C MFIV LIMIT SWITCH CONNECTOR LOOP 4 1411 HE-8 X X AEZS0042D AEFV0042 LIMIT SWITCH 1411 M-630 X X AEZS0042D MFIV LIMIT SWITCH CONNECTOR LOOP 4 1411 HE-8 X X AEZS0043 AEFV0043 LIMIT SWITCH 1411 J-601A X X AEZS0044 AEFV0044 LIMIT SWITCH 1412 J-601A X X AEZS0045 AEFV0045 LIMIT SWITCH 1412 J-601A X X AEZS0046 AEFV0046 LIMIT SWITCH 1411 J-601A X X ALFI0001A AUX FDWTR PMP TO STEAM GEN D 3601 J-110 X  ALFI0001B AUX FDWTR PMP TO STEAM GEN D 1413 J-110 X  ALFI0002A AUX FDWTR PMP TO STEAM GEN A 3601 J-110 X  ALFI0002B AUX FDWTR PMP TO STEAM GEN A 1413 J-110 X  ALFI0003A AUX FDWTR PMP TO STEAM GEN B 3601 J-110 X  ALFI0003B AUX FDWTR PMP TO STEAM GEN B 1413 J-110 X  ALFI0004A AUX FDWTR PMP TO STEAM GEN C 3601 J-110 X  ALFI0004B AUX FDWTR PMP TO STEAM GEN C 1413 J-110 X ALFO0009  MAFP B TO CST FLOW ORIFICE 1325 M-021 X ALFO0010  MAFP A TO CST FLOW ORIFICE 1326 M-021 X ALFO0011  TAFP TO CST FLOW ORIFICE 1331 M-021 X  ALFT0001 AUX FDWTR PMP TO STEAM GEN D 1304 J-301 X  ALFT0002 AUX FDWTR PMP TO STEAM GEN A 1304 J-301 X  ALFT0003 AUX FDWTR PMP TO STEAM GEN B 1305 J-301 X 
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 8)
Component ID Component name Room NoSpec No Hot SD Cold SDALFT0004 AUX FDWTR PMP TO STEAM GEN C 1305 J-301 X  ALFT0007 AUX FDW FLO TO SG A 1304 J-301 X  ALFT0009 AUX FDW FLO TO SG B 1305 J-301 X  ALFT0011 AUX FDW FLO TO SG C 1305 J-301 X  ALHK0005A MOT AUX FW PMP B DISCH TO STM GEN D 3601 J-110 X  ALHK0005B MOT AUX FW PMP B DISCH TO STM GEN D 1413 J-110 X  ALHK0006A TURB AFP DISCH TO STEAM GEN D 3601 J-110 X  ALHK0006B TURB AFP DISCH TO STEAM GEN D 1413 J-110 X  ALHK0007A MOT AUX FW PMP B DISCH TO STM GEN A 3601 J-110 X  ALHK0007B MOT AUX FW PMP B DISCH TO STM GEN A 1413 J-110 X  ALHK0008A TURB AFP DISCH TO STEAM GEN A 3601 J-110 X  ALHK0008B TURB AFP DISCH TO STEAM GEN A 1413 J-110 X  ALHK0009A MOT AUX FW PMP A DISCH TO STM GEN B 3601 J-110 X  ALHK0009B MOT AUX FW PMP A DISCH TO STM GEN B 1413 J-110 X  ALHK0010A TURB AFP DISCH TO STEAM GEN B 3601 J-110 X  ALHK0010B TURB AFP DISCH TO STEAM GEN B 1413 J-110 X  ALHK0011A MOT AUX FW PMP A DISCH TO STM GEN C 3601 J-110 X  ALHK0011B MOT AUX FW PMP A DISCH TO STM GEN C 1413 J-110 X  ALHK0012A TURB AFP DISCH TO STEAM GEN C 3601 J-110 X  ALHK0012B TURB AFP DISCH TO STEAM GEN C 1413 J-110 X  ALHV0005 MOT AUX FDWTR PMP B DISCH ISO 1324 J-601A X  ALHV0006 VALVE TERMINAL BOX 1327 E-028 X  ALHV0007 MOT AUX FDWTR PMP B DISCH ISO 1324 J-601A X  ALHV0008 VALVE TERMINAL BOX 1327 E-028 X  ALHV0009 MOT AUX FDWTR PMP A DISCH ISO 1328 J-601A X  ALHV0010 VALVE TERMINAL BOX 1330 E-028 X  ALHV0011 MOT AUX FDWTR PMP A DISCH ISO 1328 J-601A X  ALHV0012 VALVE TERMINAL BOX 1330 E-028 X  ALHV0030 ESW TO AUX. FEED PUMP B 1206 M-236A X  ALHV0031 ESW TO AUX. FEED PUMP A 1206 M-236A X  ALHV0032 TURBINE AF PUMP SUCTION FROM ESW A 1207 M-236A X  ALHV0033 TURBINE AF PUMP SUCTION FROM ESW B 1207 M-236A X  ALHV0034 CST TO AUXILIARY FEED PUMP B 1206 LIMITORQUE X  ALHV0035 CST TO AUXILIARY FEED PUMP A 1206 LIMITORQUE X  ALHV0036 TURBINE AF PUMP SUCTION FROM CST 1207 LIMITORQUE X  ALHY0006 ALHV0006 I/P CONVERTER 1327 J-601A X  ALHY0008 ALHV0008 I/P CONVERTER 1327 J-601A X  ALHY0010 ALHV0010 I/P CONVERTER 1330 J-601A X  ALHY0012 ALHV0012 I/P CONVERTER 1330 J-601A X  ALPI0024A MOT AUX FDWTR PMP B SUCT PRESS 3601 J-110 X  ALPI0024B MOT AUX FDWTR PMP B SUCT PRESS 1413 J-110 X  ALPI0025A MOT AUX FDW PMP A SUCT PRESS 3601 J-110 X  ALPI0025B MOT AUX FDW PMP A SUCT PRESS 1413 J-110 X  ALPI0026A TURB AUX FDWTR PMP SUCT PRESS 3601 J-110 X  ALPI0026B TURB AUX FDWTR PMP SUCT PRESS 1413 J-110 X  ALPI0037 AFW SPLY PRESS FROM COND STR TK 3601 J-110 X  ALPI0038 AFW SPLY PRESS FROM COND STR TK 3601 J-110 X  ALPI0039 AFW SPLY PRESS FROM COND STR TK 3601 J-110 X  ALPT0024 MOT AUX FEEDWATER PMP B SUCT PRESS 1325 J-301 X  ALPT0025 CST TO MOT AUX FDW PMP A SUCT PRESS 1326 J-301 X  ALPT0026 TURB AUX FEEDWATER PMP SUCT PRESS 1331 J-301 X  ALPT0037 ESFAS LOW SUCTION PRESS 1207 J-301 X  ALPT0038 ESFAS LOW SUCTION PRESS 1207 J-301 X  ALPT0039 ESFAS LOW SUCTION PRESS 1207 J-301 X 
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 9)
Component ID Component name Room NoSpec No Hot SD Cold SD ALV0001  TAF PMP SUCT - CST (CHECK) 1331 M-223A X ALV0002  CST TO AUX FWTR PUMP A (CHECK) 1326 M-223A X ALV0003  CST TO AF PUMP B (CHECK) 1206 M-223A X ALV0006  ESW TO AUX FEED PUMP B (CHECK) 1325 M-223A X ALV0009  ESW TO AUX FEED PUMP A (CHECK) 1326 M-223A X ALV0012  TAF PMP SUCT - ESW A (CHECK) 1331 M-223A X ALV0015  TAF PMP SUCT - ESW B (CHECK) 1331 M-223A X ALV0029  AF PUMP B RECIRC LINE (CHECK) 1325 M-231C X ALV0030  AF PMP B DISCHARGE (CHECK) 1325 M-224B X ALV0033  AF PUMP B DISCHARGE TO SG A (CK) 1324 M-224B X X ALV0036  AF PUMP B TO SG D (CK) 1324 M-224B X X ALV0041  AUX FEED PUMP A RECIRC LINE (CK) 1326 M-231C X ALV0042  AF PUMP A DISCHARGE (CK) 1326 M-224B X ALV0045  AF PUMP A TO SG C (CK) 1328 M-224B X X ALV0048  AF PUMP A DISCHARGE TO SG B (CK) 1328 M-224B X X ALV0053  TAF PMP RECIRC LINE (CHECK) 1331 M-224B X ALV0054  TAF PMP DISCHARGE (CHECK) 1331 M-224A X ALV0057  TAF PMP DISCHARGE TO SG A (CK) 1327 M-224B X X ALV0062  TAF PMP TO SG D (CK) 1327 M-224B X X ALV0067  TAF PMP DISCHARGE TO SG B (CK) 1330 M-224B X X ALV0072  TAF PMP TO SG C (CK) 1330 M-224B X X ALZS0006 ALHV0006 LIMIT SWITCH 1327 J-601A X  ALZS0008 ALHV0008 LIMIT SWITCH 1327 J-601A X  ALZS0010 ALHV0010 LIMIT SWITCH 1330 J-601A X  ALZS0012 ALHV0012 LIMIT SWITCH 1330 J-601A X  BB007 D.C. CONTACTOR FOR PORV 1403 E-018A X X BB008 D.C. CONTACTOR FOR PORV 1408 E-018A X X BB8010A  PRESS SAFETY RELIEF VALVE 2602 M-724-3-2 X X BB8010B  PRESS SAFETY RELIEF VALVE 2602 M-724-3-2 X X BB8010C  PRESS SAFETY RELIEF VALVE 2602 M-724-3-2 X X BB8038A  ISO FOR PRT INLET LINE (CHECK) 2000 M-724-8 X X BB8038B  ISO FOR PRT INLET LINE (CHECK) 2000 M-724-8 X X BB8378A  CVCS CHARGING ISO TO RCS LOOP 1 (CHECK)2000 M-724-8 X X BB8378B  CVCS CHARGING ISO TO RCS LOOP 1 (CHECK)2000 M-724-8 X X BB8379A  CVCS ALTERNATE CHARGING VLV (CHECK) 2000 M-724-8 X X BB8379B  CVCS ALTERNATE CHARGING VLV (CHECK) 2000 M-724-8 X X BB8948A  SAFETY INJ ACCUM ISO (CHECK) 2000 M-724-8 X X BB8948B  SAFETY INJ ACCUM ISO (CHECK) 2000 M-724-8 X X BB8948C  SAFETY INJ ACCUM ISO (CHECK) 2000 M-724-8 X X BB8948D  SAFETY INJ ACCUM ISO (CHECK) 2000 M-724-8 X X BB8949A  SAFETY INJ ISO VLV LOOP 1 (CHECK) 2000 M-724-8 X X BB8949B  SI & RHR ISO VLV LOOP 2 (CHECK) 2000 M-724-8 X X BB8949C  SI & RHR ISO VLV LOOP 3 (CHECK) 2000 M-724-8 X X BB8949D  SAFETY INJ ISO VLV LOOP 4 (CHECK) 2000 M-724-8 X X BBFI0017  RCP 1A THRM BARR CLG FLOW 2000 J-517A X X BBFI0018  RCP 1B THRM BARR CLG FLOW 2000 J-517A X X BBFI0019  RCP 1C THRM BARR CLG FLOW 2000 J-517A X X BBFI0020  RCP 1D THRM BARR CLG FLOW 2000 J-517A X X BBFT0017 RCP 1A THRM BARR CLG FLOW 2000 J-301 X X BBFT0018 RCP 1B THRM BARR CLG FLOW 2000 J-301 X X BBFT0019 RCP 1C THRM BARR CLG FLOW 2000 J-301 X X BBFT0020 RCP 1D THRM BARR CLG FLOW 2000 J-301 X X BBFT0414 RCS FLOW TRANSMITTER LOOP 1 2201 ESE-4 X X BBFT0415 RCS FLOW TRANSMITTER LOOP 1 2201 ESE-4 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 10)
Component ID Component name Room NoSpec No Hot SD Cold SDBBFT0416 RCS FLOW TRANSMITTER LOOP 1 2201 ESE-4 X X BBFT0424 RCS FLOW TRANSMITTER LOOP 2 2201 ESE-4 X X BBFT0425 RCS FLOW TRANSMITTER LOOP 2 2201 ESE-4 X X BBFT0426 RCS FLOW TRANSMITTER LOOP 2 2201 ESE-4 X X BBFT0434 RCS FLOW TRANSMITTER LOOP 3 2201 ESE-4 X X BBFT0435 RCS FLOW TRANSMITTER LOOP 3 2201 ESE-4 X X BBFT0436 RCS FLOW TRANSMITTER LOOP 3 2201 ESE-4 X X BBFT0444 RCS FLOW TRANSMITTER LOOP 4 2201 ESE-4 X X BBFT0445 RCS FLOW TRANSMITTER LOOP 4 2201 ESE-4 X X BBFT0446 RCS FLOW TRANSMITTER LOOP 4 2201 ESE-4 X X BBHV0013 RCP THERMAL BARRIER COOLER ISOLATION VALVE 2000 LIMITORQUE X X BBHV0014 RCP THERMAL BARRIER COOLER ISOLATION VALVE 2000 LIMITORQUE X X BBHV0015 RCP THERMAL BARRIER COOLER ISOLATION VALVE 2000 LIMITORQUE X X BBHV0016 RCP THERMAL BARRIER COOLER ISOLATION VALVE 2000 LIMITORQUE X X BBHV8000A PRESSURIZER PORV ISOLATION VALVE 2601 LIMITORQUE X X BBHV8000B PRESSURIZER PORV ISOLATION VALVE 2601 LIMITORQUE X X BBHV8001A VALVE TERMINAL BOX 2000 E-028  X BBHV8001A REACTOR VESSEL HEAD VENT VALVE 2000 HE-10A  X BBHV8001A REACTOR VESSEL HEAD VENT VALVE CONNECTOR 2000 HE-8  X BBHV8001B VALVE TERMINAL BOX 2000 E-028  X BBHV8001B REACTOR VESSEL HEAD VENT VALVE CONNECTOR 2000 HE-8  X BBHV8001B REACTOR VESSEL HEAD VENT VALVE 2000 HE-10A  X BBHV8002A VALVE TERMINAL BOX 2000 E-028  X BBHV8002A REACTOR VESSEL HEAD VENT VALVE 2000 HE-10A  X BBHV8002A REACTOR VESSEL HEAD VENT VALVE CONNECTOR 2000 HE-8  X BBHV8002B VALVE TERMINAL BOX 2000 E-028  X BBHV8002B REACTOR VESSEL HEAD VENT VALVE 2000 HE-10A  X BBHV8002B REACTOR VESSEL HEAD VENT VALVE CONNECTOR 2000 HE-8  X BBHV8010A VALVE TERMINAL BOX 2000 E-028 X X BBHV8010B VALVE TERMINAL BOX 2000 E-028 X X BBHV8010C VALVE TERMINAL BOX 2000 E-028 X X BBHV8037A PRT EMERGENCY DRAIN LINE 2000 LIMITORQUE X X BBHV8037B PRT EMERGENCY DRAIN LINE 2000 LIMITORQUE X X BBHV8157A EXCESS LETDOWN PATH TO PRT ISOLATION 2000 M-231E X X BBHV8157B EXCESS LETDOWN PATH TO PRT ISOLATION 2000 M-231E X X BBHV8351A RCP SEAL INJECTION CONTAINMENT ISOLATION VALVE 1322 LIMITORQUE X X BBHV8351B RCP SEAL INJECTION CONTAINMENT ISOLATION VALVE 1322 LIMITORQUE X X BBHV8351C RCP SEAL INJECTION CONTAINMENT ISOLATION VALVE 1322 LIMITORQUE X X BBHV8351D RCP SEAL INJECTION CONTAINMENT ISOLATION VALVE 1322 LIMITORQUE X X BBLI0459A PRESSURIZER LEVEL 3601 ESE-14 X X BBLI0459B PRESSURIZER LEVEL 1413 ESE-14 X X BBLI0460A PRESSURIZER LEVEL 3601 ESE-14 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 11)
Component ID Component name Room NoSpec No Hot SD Cold SDBBLI0460B PRESSURIZER LEVEL 1413 ESE-14 X X BBLI0461 PRESSURIZER LEVEL 3601 ESE-14 X X BBLT0459 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3A X X BBLT0459 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3C X X BBLT0460 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3C X X BBLT0460 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3A X X BBLT0461 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3C X X BBLT0461 PRESSURIZER LEVEL TRANSMITTER 2201 ESE-3A X X BBPCV0455A PRESSURIZER POWER OPERATED RELIEF VLV CONNECTOR 2601 HE-8 X X BBPCV0455A PRESSURIZER POWER-OPERATED RELIEF VALVE2601 HE-9 X X BBPCV0455A VALVE TERMINAL BOX 2601 E-028 X X BBPCV0456A PRESSURIZER POWER OPERATED RELIEF VLV CONNECTOR 2601 HE-8 X X BBPCV0456A PRESSURIZER POWER-OPERATED RELIEF VALVE2601 HE-9 X X BBPCV0456A VALVE TERMINAL BOX 2601 E-028 X X BBPI0403 REACTOR COOLANT PRESSURE INDICATOR 3601 ESE-14 X X BBPI0405 REACTOR COOLANT PRESSURE INDICATOR 3601 ESE-14 X X BBPI0405X REACTOR COOLANT PRESSURE INDICATOR 1413 ESE-14 X X BBPI0406 REACTOR COOLANT PRESSURE INDICATOR 3601 ESE-14 X X BBPI0406X REACTOR COOLANT PRESSURE INDICATOR 1413 ESE-14 X X BBPT0403 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1202 ESE-2D (13) X X BBPT0403 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1202 ESE-1A X X BBPT0405 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1320 ESE-1B X X BBPT0405 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1320 ESE-1C X X BBPT0406 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1202 ESE-1A X X BBPT0406 REACTOR COOLANT SYSTEM PRESSURE WIDE RANGE 1202 ESE-2D (13) X X BBPT0455 PRESSURIZER PRESSURE TRANSMITTER 2201 J-301 X X BBPT0456 PRESSURIZER PRESSURE TRANSMITTER 2201 J-301 X X BBPT0457 PRESSURIZER PRESSURE TRANSMITTER 2201 J-301 X X BBPT0458 PRESSURIZER PRESSURE TRANSMITTER 2201 J-301 X X BBPV8702A RHR PUMP SUCTION ISOLATION VALVE 2000 LIMITORQUE X X BBPV8702B RHR PUMP SUCTION ISOLATION VALVE 2000 LIMITORQUE X X BBTE0411A1 RCS HOT LEG RTD TEMP ELEMENT LOOP 1 2201 J-564 X X BBTE0411A2 RCS HOT LEG RTD TEMP ELEMENT LOOP 1 2201 J-564 X X BBTE0411A3 RCS HOT LEG RTD TEMP ELEMENT LOOP 1 2201 J-564 X X BBTE0411B RCS COLD LEG RTD TEMP ELEMENT LOOP 1 2201 J-564 X X BBTE0413A RCS HOT LEG RTD CONNECTOR (WR) LOOP 1 2201 HE-8 X X BBTE0413A RCS HOT LEG TEMPERATURE ELEMENT (WR)
LOOP 1 2201 ESE-6 X X BBTE0413B RCS COLD LEG RTD CONNECTOR (WR) LOOP 1 2201 HE-8 X X BBTE0413B RCS COLD LEG TEMP ELEMENT (WR) LOOP 1 2201 ESE-6 X X BBTE0421A1 RCS HOT LEG RTD TEMP ELEMENT LOOP 2 2201 J-564 X X BBTE0421A2 RCS HOT LEG RTD TEMP ELEMENT LOOP 2 2201 J-564 X X BBTE0421A3 RCS HOT LEG RTD TEMP ELEMENT LOOP 2 2201 J-564 X X BBTE0421B RCS COLD LEG RTD TEMP ELEMENT LOOP 2 2201 J-564 X X BBTE0423A RCS HOT LEG RTD CONNECTOR (WR) LOOP 2 2201 HE-8 X X BBTE0423A RCS HOT LEG TEMP ELEMENT (WR) LOOP 2 2201 ESE-6 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 12)
Component ID Component name Room NoSpec No Hot SD Cold SDBBTE0423B RCS COLD LEG RTD CONNECTOR (WR) LOOP 2 2201 HE-8 X X BBTE0423B RCS COLD LEG TEMP ELEMENT (WR) LOOP 2 2201 ESE-6 X X BBTE0431A1 RCS HOT LEG RTD TEMP ELEMENT LOOP 3 2201 J-564 X X BBTE0431A2 RCS HOT LEG RTD TEMP ELEMENT LOOP 3 2201 J-564 X X BBTE0431A3 RCS HOT LEG RTD TEMP ELEMENT LOOP 3 2201 J-564 X X BBTE0431B RCS COLD LEG RTD TEMP ELEMENT LOOP 3 2201 J-564 X X BBTE0433A RCS HOT LEG RTD CONNECTOR (WR) LOOP 3 2201 HE-8 X X BBTE0433A RCS HOT LEG TEMPERATURE ELEMENT (WR)
LOOP 3 2201 ESE-6 X X BBTE0433B RCS COLD LEG RTD CONNECTOR (WR) LOOP 3 2201 HE-8 X X BBTE0433B RCS COLD LEG TEMPERATURE ELEMENT (WR)
LOOP 3 2201 ESE-6 X X BBTE0441A1 RCS HOT LEG RTD TEMP ELEMENT LOOP 4 2201 J-564 X X BBTE0441A2 RCS HOT LEG RTD TEMP ELEMENT LOOP 4 2201 J-564 X X BBTE0441A3 RCS HOT LEG RTD TEMP ELEMENT LOOP 4 2201 J-564 X X BBTE0441B RCS COLD LEG RTD TEMP ELEMENT LOOP 4 2201 J-564 X X BBTE0443A RCS HOT LEG RTD CONNECTOR (WR) LOOP 4 2201 HE-8 X X BBTE0443A RCS HOT LEG TEMPERATURE ELEMENT (WR)
LOOP 4 2201 ESE-6 X X BBTE0443B RCS COLD LEG RTD CONNECTOR (WR) LOOP 4 2201 HE-8 X X BBTE0443B RCS COLD LEG TEMPERATURE ELEMENT (WR)
LOOP 4 2201 ESE-6 X X BBTI0413A TH WIDE RANGE LOOP 1 HOT LEG TEMP 3601 ESE-14 X X BBTI0413B TH WIDE RANGE LOOP 1 COLD LEG TEMP 3601 ESE-14 X X BBTI0423A TH WIDE RANGE LOOP 2 HOT LEG TEMP 3601 ESE-14 X X BBTI0423B TH WIDE RANGE LOOP 2 COLD LEG TEMP 3601 ESE-14 X X BBTI0423X TH WIDE RANGE LOOP 2 COLD LEG TEMP 1413 ESE-14 X X BBTI0443A TH WIDE RANGE LOOP 4 HOT LEG TEMP 1413 ESE-14 X X BBTW0413A  RCS HL LOOP 1 WR TH. WELL 2201 M-714 X X BBTW0413B  RCS CL LOOP 1 WR TH. WELL 2201 M-714 X X BBTW0423A  RCS HL LOOP 2 WR TH. WELL 2201 M-714 X X BBTW0423B  RCS CL LOOP 2 WR TH. WELL 2201 M-714 X X BBTW0433A  RCS HL LOOP 3 WR TH. WELL 2201 M-714 X X BBTW0433B  RCS CL LOOP 3 WR TH. WELL 2201 M-714 X X BBTW0443A  RCS HL LOOP 4 WR TH. WELL 2201 M-714 X X BBTW0443B  RCS CL LOOP 4 WR TH. WELL 2201 M-714 X X BBV0001  INJ TANK INJ LINE ISO (CHECK) 2000 M-231C X X BBV0022  BORON INJ TANK INJ LINE ISO (CHECK) 2000 M-231C X X BBV0040  BORON INJ TANK INJ LINE ISO (CHECK) 2000 M-231C X X BBV0059  SIS BORON INJ TANK ISO (CHECK) 2000 M-231C X X BBV0118  RCP SEAL INJ CTMT ISOL VLV (CHECK) 2000 M-231C X X BBV0120  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0121  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0124  RCP A THERM BARRIER COOLING COIL PRESS
 
RLF VLV 2000 M-141-2 X X BBV0148  RCP SEAL INJ CTMT ISO VLV (CHECK) 2000 M-231C X X BBV0150  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 13)
Component ID Component name Room NoSpec No Hot SD Cold SD BBV0151  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0154  RCP B THERM BARRIER COOLING COIL PRESS
 
RLF VLV 2000 M-141-2 X X BBV0178  RCP SEAL INJ CTMT ISO VALVE (CHECK) 2000 M-231C X X BBV0180  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0181  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0184  RCP C THERM BARRIER COOLING COIL PRESS
 
RLF VLV 2000 M-141-2 X X BBV0208  RCP SEAL INJ CTMT ISO VALVE (CHECK) 2000 M-231C X X BBV0210  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0211  RCP SEAL INJ RCS ISOLATION CHECK VLV (CHECK) 2000 M-231C X X BBV0214  RCP D THERM BARRIER COOLING COIL PRESS
 
RLF VLV 2000 M-141-2 X X BBV0443  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0444  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0445  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0446  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0447  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0448  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0449  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBV0450  CCW TO RCP THERMAL BARRIER COOLER 2000 M-231B X X BBZS0455A PRESSURIZER PORV LIMIT SWITCH 2601 HE-9 X X BBZS0455A PRESSURIZER PORV LIMIT SWITCH CONNECTOR2601 HE-8 X X BBZS0456A PRESSURIZER PORV LIMIT SWITCH CONNECTOR2601 HE-8 X X BBZS0456A PRESSURIZER PORV LIMIT SWITCH 2601 HE-9 X X BBZS8010A PRESSURIZER SAFETY RELIEF VALVE LIMIT SWITCH 2602 HE-7 X X BBZS8010A PRESSURIZER SAFETY RELIEF VALVE LS CONNECTOR 2602 HE-8 X X BBZS8010B PRESSURIZER SAFETY RELIEF VALVE LIMIT SWITCH 2602 HE-7 X X BBZS8010B PRESSURIZER SAFETY RELIEF VALVE LS CONNECTOR 2602 HE-8 X X BBZS8010C PRESSURIZER SAFETY RELIEF VALVE LS CONNECTOR 2602 HE-8 X X BBZS8010C PRESSURIZER SAFETY RELIEF VALVE LIMIT SWITCH 2602 HE-7 X X BBZS8702AA VALVE TERMINAL BOX (BBZS8702AB) 2000 E-028 X X BBZS8702AA RHR PUMP SUCTION ISOLATION VALVE LS CONNECTOR 2000 HE-8 X X BBZS8702AA RHR PUMP SUCTION ISOLATION VALVE LIMIT SWITCH 2000 HE-3 X X BBZS8702AB RHR PUMP SUCTION ISOLATION VALVE LS CONNECTOR 2000 HE-8 X X BBZS8702AB RHR PUMP SUCTION ISOLATION VALVE LIMIT SWITCH 2000 HE-3 X X BG8124  CCP/SI CROSSTIE RELIEF 1108 M-724-3-1 X X BG8440  CHARG PMP SUCTION (CHECK) 1318 M-724-8 X X BG8481A  CCP A DISCHARGE CHECK VLV (CHECK) 1114 M-724-8 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 14)
Component ID Component name Room NoSpec No Hot SD Cold SD BG8481B  CPP B DISCHARGE CHECK VLV (CHECK) 1107 M-724-8 X X BG8486  BORIC ACID BATCHING TK ISO FROM BATP SUCT (CK) 1117 M-724 X X BG8497  PD PMP DISCH (CHECK) 1115 M-724-8 X X BG8546A  CHARG PMP SUCT FROM RWST (CHECK) 1114 M-724-8 X X BG8546B  CHARG PMP SUCT FROM RWST (CHECK) 1107 M-724-8 X X BGFI0138A EXCESS LETDOWN FLOW TO PRT 3601 ESE-14 X X BGFI0138B EXCESS LETDOWN FLOW TO PRT 3601 ESE-14 X X BGFI0215A SEAL INJECTION FLOW 3601 ESE-14 X X BGFI0215B SEAL INJECTION FLOW 3601 ESE-14 X X BGFO0004  CCP A MINI FLOW ORIFICE 1114 M-721-5 X X BGFO0005  CCP B MINI FLOW ORIFICE 1107 M-721-5 X X BGFT0138A EXCESS LETDOWN PATH TO PRT FLOW TRANSMITTER 2000 ESE-3 X X BGFT0138B EXCESS LETDOWN PATH TO PRT FLOW TRANSMITTER 2000 ESE-3 X X BGFT0215A RCP SEAL INJECTION FLOW TRANSMITTER 1202 ESE-4A X X BGFT0215A RCP SEAL INJECTION FLOW TRANSMITTER 1202 ESE-4D X X BGFT0215B RCP SEAL INJECTION FLOW TRANSMITTER 1202 ESE-4D X X BGFT0215B RCP SEAL INJECTION FLOW TRANSMITTER 1202 ESE-4A X X BGHV8104 EMERGENCY BORATION PATH ISOLATION VALVE1113 LIMITORQUE X X BGHV8105 CHARGING LINE CONTAINMENT ISOLATION VALVE 1323 LIMITORQUE X X BGHV8106 CHARGING LINE ISOLATION VALVE 1323 LIMITORQUE X X BGHV8110 CCP A MINIFLOW ISOLATION VALVE 1114 LIMITORQUE X X BGHV8111 CCP B MINIFLOW ISOLATION VALVE 1107 LIMITORQUE X X BGHV8153A EXCESS LETDOWN/RCS ISOLATION VALVE CONNECTOR 2000 HE-8 X X BGHV8153A EXCESS LETDOWN/RCS ISOLATION VALVE 2000 HE-10A X X BGHV8153B EXCESS LETDOWN/RCS ISOLATION VALVE 2000 HE-10A X X BGHV8153B EXCESS LETDOWN/RCS ISOLATION VALVE CONNECTOR 2000 HE-8 X X BGHV8154A EXCESS LETDOWN/RCS ISOLATION VALVE 2000 HE-10A X X BGHV8154A EXCESS LETDOWN/RCS ISOLATION VALVE CONNECTOR 2000 HE-8 X X BGHV8154B EXCESS LETDOWN/RCS ISOLATION VALVE CONNECTOR 2000 HE-8 X X BGHV8154B EXCESS LETDOWN/RCS ISOLATION VALVE 2000 HE-10A X X BGHV8357A CCP A ALTERNATE DISCHARGE TO RCP SEALS 1114 M-231E X X BGHV8357B CCP B ALTERNATE DISCHARGE TO RCP SEALS 1107 M-231E X X BGLCV0112B VCT ISOLATION VALVE FROM CHARGING PUMP SUCTION 1318 LIMITORQUE X X BGLCV0112C VCT ISOLATION VALVE FROM CHARGING PUMP SUCTION 1318 LIMITORQUE X X BGLCV0459  REGEN HX FROM RCS LOOP 3 XL 2000 M-724-7-1 X X BGLCV0460  REGEN HX FROM RCS LOOP 3 XL 2000 M-724-7-1 X X BGLI0038  PBG05A LUBE OIL RESERVOIR LEVEL 1114 M-721-1 X X BGLI0039  PBG05B LUBE OIL RESERVOIR LEVEL 1107 M-721-1 X X BGLI0102 BORIC ACID TANK #1 LEVEL 3601 ESE-14 X X BGLI0104 BORIC ACID TANK 1 LEVEL 3601 ESE-14 X X BGLI0105 BORIC ACID TANK #2 LEVEL 3601 ESE-14 X X BGLI0106 BORIC ACID TANK 2 LEVEL 3601 ESE-14 X X BGLI0112 VOLUME CONTROL TANK LEVEL 3601 ESE-14 X X BGLI0185 VOLUME CONTROL TANK LEVEL 3601 ESE-14 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 15)
Component ID Component name Room NoSpec No Hot SD Cold SDBGLT0102 BORIC ACID TANK A LEVEL TRANSMITTER 1117 ESE-4 X X BGLT0104 BORIC ACID TANK A LEVEL TRANSMITTER 1117 ESE-4 X X BGLT0105 BORIC ACID TANK B LEVEL TRANSMITTER 1116 ESE-4 X X BGLT0106 BORIC ACID TANK B LEVEL TRANSMITTER 1116 ESE-4 X X BGLT0112 VCT LEVEL TRANSMITTER 1318 ESE-4 X X BGLT0185 VCT LEVEL TRANSMITTER 1320 ESE-4 X X BGPI0001  PBG05A LUBE OIL PRESSURE 1114 M-721-1 X X BGPI0002  PBG05B LUBE OIL PRESSURE 1107 M-721-1 X X BGPI0019  PBG05A LUBE OIL PRESSURE 1114 M-721-1 X X BGPI0020  PBG05B LUBE OIL PRESSURE 1107 M-721-1 X X BGPS0017  PBG03A MOTOR START SWITCH 1114 M-721-1 X X BGPS0018  PBG03B MOTOR START SWITCH 1107 M-721-1 X X BGTE0137A EXCESS LETDOWN PATH TO PRT TEMPERATURE ELEMENT 2000 ESE-42A X X BGTE0137B EXCESS LETDOWN PATH TO PRT TEMPERATURE ELEMENT 2000 ESE-42A X X BGTI0036  PBG05A THRUST BEARING TEMPERATURE 1114 M-721-1 X X BGTI0037  PBG05B THRUST BEARING TEMPERATURE 1107 M-721-1 X X BGTI0040  PBG05A LUBE OIL TEMPERATURE 1114 M-721-1 X X BGTI0041  PBG05B LUBE OIL TEMPERATURE 1107 M-721-1 X X BGTI0137A EXCESS LETDOWN TO PRT TEMP 3601 ESE-14 X X BGTI0137B EXCESS LETDOWN FLOW TO PRT TEMP 3601 ESE-14 X X BGV0091  CCP A MINIFLOW ISO VALVE (CHECK) 1114 M-231C X X BGV0095  CCP B MINIFLOW ISO VALVE (CHECK) 1107 M-231C X X BGV0147  BATP A DISCHARGE VLV (CHECK) 1117 M-221 X X BGV0154  BORIC ACID SUPPLY TO CHARGING PMPS (CK)1113 M-231C X X BGV0165  BATP B DISCHARGE VLV (CK) 1116 M-221 X X BGV0172  MANUAL VALVE IN BORIC ACID FILTER BYPASS LINE 1117 M-243 X X BGV0173  MANUAL VALVE IN BORIC ACID FILTER
 
BYPASS LINE 1117 M-243 X X BGV0174  EMERG BORATION PATH ISOL VLV (CHECK) 1113 M-221 X X BGV0177  BORIC ACID SUPPLY TO CHARGING PUMPS (MANUAL) 1113 M-243 X X BGV0180  REACTOR MAKEUP WATER ISOLATION (CHECK) 1318 M-231C X X BGV0183  MANUAL ISOLATION VALVE TO EMERG
 
BORATION PATH 1113 M-243 X X BGV0184  EMERG BORATION PATH CHECK VLV ISOL (CHECK) 1113 M-231C X X BGV0203  PRESS RLF VLV EXCESS LTDN HX 2000 M-141-2 X X BGV0207  PRESS RLF VLV CCW SEAL WTR HX 1315 M-141-2 X X BGV0259  CCP A - CCW DISCHARGE (MAN) 1114 M-231A X X BGV0268  CCP B - CCW DISCHARGE (MAN) 1107 M-231A X X BGV0524  PRESS RLF VLV CCP A CCW LINE 1114 M-141-2 X X BGV0525  PRESS RLF VLV CCP B CCW LINE 1107 M-141-2 X X BGV0591  CHARGING TO RCP SEAL INJ ISO VLV (CK) 1115 M-231C X X BGV0800A  PBG05A LUBE OIL RELIEF VALVE 1114 M-721-1 X X BGV0800B  PBG05B LUBE OIL RELIEF VALVE 1107 M-721-1 X X BGV0801A  PBG05A LUBE OIL CHECK VALVE 1114 M-721-1 X X BGV0801B  PBG05B LUBE OIL CHECK VALVE 1107 M-721-1 X X BGV0802A  PBG05A LUBE OIL CHECK VALVE 1114 M-721-1 X X BGV0802B  PBG05B LUBE OIL CHECK VALVE 1107 M-721-1 X X BMHV0001 VALVE TERMINAL BOX 1411 E-028 X  BMHV0002 VALVE TERMINAL BOX (AEFV0044) 1412 E-028 X 
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 16)
Component ID Component name Room NoSpec No Hot SD Cold SDBMHV0003 VALVE TERMINAL BOX 1412 E-028 X  BMHV0004 VALVE TERMINAL BOX (AEFV0046) 1411 E-028 X  BMHV0019 SG A OUT TO NUC SAMP SYS 2000 J-603A X  BMHV0020 SG B OUT TO NUC SAMP SYS 2000 J-603A X  BMHV0021 SG C OUT TO NUC SAMP SYS 2000 J-603A X  BMHV0022 SG D OUT TO NUC SAMP SYS 2000 J-603A X  BMHV0035 SG A TUBE SHT SAMP VLV 2000 J-603A X  BMHV0036 SG B TUBE SHT SAMP VLV 2000 J-603A X  BMHV0037 SG C TUBE SHT SAMP VLV 2000 J-603A X  BMHV0038 SG D TUBE SHT SAMP VLV 2000 J-603A X  BMHV0065 SG A TO NUC SAMP SYS 1323 J-603A X  BMHV0066 SG B TO NUC SAMP SYS 1323 J-603A X  BMHV0067 SG C TO NUC SAMP SYS 1323 J-603A X  BMHV0068 SG D TO NUC SAMP SYS 1323 J-603A X  BMHY0001A BMHV0001 SOLENOID VALVE 1411 J-601A X  BMHY0001B BMHV0001 SOLENOID VALVE 1411 J-601A X  BMHY0001B VALVE TERMINAL BOX (AEFV0043 BMZS0001B) 1411 E-028 X X BMHY0002A BMHV0002 SOLENOID VALVE 1412 J-601A X  BMHY0002B BMHV0002 SOLENOID VALVE 1412 J-601A X  BMHY0002B VALVE TERMINAL BOX (BMZS0002B) 1412 E-028 X  BMHY0003A BMHV0003 SOLENOID VALVE 1412 J-601A X  BMHY0003B BMHV0003 SOLENOID VALVE 1412 J-601A X  BMHY0003B VALVE TERMINAL BOX (AEFV0045 BMZS0003B) 1412 E-028 X X BMHY0004A BMHV0004 SOLENOID VALVE 1411 J-601A X  BMHY0004B BMHV0004 SOLENOID VALVE 1411 J-601A X  BMHY0004B VALVE TERMINAL BOX (BMZS0004B) 1411 E-028 X  BMZS0001A BMHV0001 LIMIT SWITCH 1411 J-601A X  BMZS0001B BMHV0001 LIMIT SWITCH 1411 J-601A X  BMZS0002A BMHV0002 LIMIT SWITCH 1412 J-601A X  BMZS0002B BMHV0002 LIMIT SWITCH 1412 J-601A X  BMZS0003A BMHV0003 LIMIT SWITCH 1412 J-601A X  BMZS0003B BMHV0003 LIMIT SWITCH 1412 J-601A X  BMZS0004A BMHV0004 LIMIT SWITCH 1411 J-601A X  BMZS0004B BMHV0004 LIMIT SWITCH 1411 J-601A X  BNHCV8800A VALVE TERMINAL BOX 1202 E-028 X X BNHCV8800B VALVE TERMINAL BOX 1202 E-028 X X BNHS8812B ISOLATION SWITCH FOR BNHV8812B 3302 E-028B  X BNHV8812A RHR PUMP RWST SUCTION VALVE 1111 LIMITORQUE  X BNHV8812B RHR PUMP RWST SUCTION VALVE 1109 LIMITORQUE  X BNHY8800A RWST ISOLATION VALVE FOR FUEL POOL CLEANUP PUMPS 1202 HE-5 X X BNHY8800B RWST ISOLATION VALVE FOR FUEL POOL CLEANUP PUMPS 1202 HE-5 X X BNLCV0112D CHARGING PUMP RWST SUCTION VALVE 1114 LIMITORQUE X X BNLCV0112E CHARGING PUMP RWST SUCTION VALVE 1107 LIMITORQUE X X BNZS8800A RWST ISOL VLV FOR FUEL POOL CLEANUP PUMPS LMT SW 1202 HE-6  X BNZS8800B RWST ISOL VLV FOR FUEL POOL CLEANUP PUMPS LMT SW 1202 HE-6  X BNZS8812AA RHR PUMP RWST SUCTION VALVE LIMIT SWITCH 1111 HE-6  X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 17)
Component ID Component name Room NoSpec No Hot SD Cold SDBNZS8812BA RHR PUMP RWST SUCTION VALVE LIMIT SWITCH 1109 HE-6  X CGD01A  ESW PUMP ROOM SUPPLY FAN K105 M-619.2 X X CGD01B  ESW PUMP ROOM SUPPLY FAN K104 M-619.2 X X CGM01A  DIESEL GEN. VENT. SUPPLY FAN 5203 M-619.2 X X CGM01B  DIESEL GEN. VENT. SUPPLY FAN 5201 M-619.2 X X Delete      Delete      Delete      Delete      DCGD01A ESW PUMP ROOM SUPPLY FAN MOTOR K105 M-619.2 X X DCGD01B ESW PUMP ROOM SUPPLY FAN MOTOR K104 M-619.2 X X DCGM01A DIESEL GEN. VENT. SUPPLY FAN MOTOR 5203 M-619.2 X X DCGM01B DIESEL GEN. VENT. SUPPLY FAN MOTOR 5201 M-619.2 X X Delete      Delete      Delete      Delete      DPAL01A AUX FEEDWATER PUMP A MOTOR 1326 E-012.2 X  DPAL01B AUX FEEDWATER PUMP B MOTOR 1325 E-012.2 X  DPBG02A BORIC ACID TRANSFER PUMP A MOTOR 1117 AE-3 X X DPBG02B BORIC ACID TRANSFER PUMP B MOTOR 1116 M-098 X X DPBG05A CENTRIFUGAL CHARGING PUMP A MOTOR TERMINATE KIT 1114 E-029 X X DPBG05A CENTRIFUGAL CHARGING PUMP A MOTOR 1114 AE-2 X X DPBG05B CENTRIFUGAL CHARGING PUMP B MOTOR TERMINATE KIT 1107 E-029 X X DPBG05B CENTRIFUGAL CHARGING PUMP B MOTOR 1107 AE-2 X X DPEF01A ESW PUMP A MOTOR K105 E-012.2 X X DPEF01B ESW PUMP B MOTOR K104 E-012.2 X X DPEG01A CCW PUMP A MOTOR TRAIN A 1406 E-012.2 X X DPEG01B CCW PUMP B MOTOR TRAIN B 1401 E-012.2 X X DPEG01C CCW PUMP C MOTOR TRAIN A 1406 E-012.2 X X DPEG01D CCW PUMP D MOTOR TRAIN B 1401 E-012.2 X X DPEJ01A RHR PUMP A MOTOR 1111 AE-2  X DPEJ01A RHR PUMP A MOTOR TERMINATION KIT 1111 E-029  X DPEJ01B RHR PUMP B MOTOR 1109 AE-2  X DPEJ01B RHR PUMP B MOTOR TERMINATION KIT 1109 E-029  X DPJE01A  EMERGENCY FUEL OIL TRANSFER PUMP A MOTOR (50) M-087 X X DPJE01B EMERGENCY FUEL OIL TRANSFER PUMP B MOTOR (50) M-087 X X DPKJ01A JACKET WATER KEEP WARM PUMP A MOTOR 5203 M-018 X X DPKJ01B JACKET WATER KEEP WARM PUMP B MOTOR 5201 M-018 X X DPKJ02A ROCKER PRELUBE PUMP A MOTOR 5203 M-018 X X DPKJ02B ROCKER PRELUBE PUMP B MOTOR 5201 M-018 X X DPKJ03A AUXILIARY LUBE OIL KEEP WARM PUMP A MOTOR 5203 M-018 X X DPKJ03B AUXILIARY LUBE OIL KEEP WARM PUMP B MOTOR 5201 M-018 X X DSGF02A AUX FEED PUMP ROOM COOLER MOTOR 1326 M-612 X X DSGF02B AUX FEED PUMP ROOM COOLER MOTOR 1325 M-612 X X DSGL10A RHR PUMP ROOM COOLER MOTOR 1111 M-612  X
 
Rev. 25 WOLF CREEK TABLE 7.4-6 (Sheet 18)
Component ID Component name Room NoSpec No Hot SD Cold SDDSGL10B RHR PUMP ROOM COOLER MOTOR 1109 M-612  X DSGL11A CCW PUMP ROOM COOLER MOTOR 1406 M-612 X X DSGL11B CCW PUMP ROOM COOLER MOTOR 1401 M-612 X X DSGL12A CENT. CHARGING PUMP ROOM COOLER MOTOR 1114 M-612 X X DSGL12B CENT. CHARGING PUMP ROOM COOLER MOTOR 1107 M-612 X X DSGL15A PENETRATION ROOM COOLER MOTOR 1410 M-612 X X DSGL15B PENETRATION ROOM COOLER MOTOR 1409 M-612 X X DSGN01A CONTAINMENT COOLER FAN MOTOR 2000 M-620 X X DSGN01B CONTAINMENT COOLER FAN MOTOR 2000 M-620 X X DSGN01C CONTAINMENT COOLER FAN MOTOR 2000 M-620 X X DSGN01D CONTAINMENT COOLER FAN MOTOR 2000 M-620 X X EBB01A  STM GEN LOOP A 2000 M-711 X X EBB01B  STM GEN LOOP B 2000 M-711 X X EBB01C  STM GEN LOOP C 2000 M-711 X X EBB01D  STM GEN LOOP D 2000 M-711 X X EBG02  EXCESS LETDN HX 2000 M-722 X X EBG08A  PBG05A LUBE OIL COOLER 1114 M-721-1 X X EBG08B  PBG05B LUBE OIL COOLER 1107 M-721-1 X X ECHV0011 F.P.C.H. EX. CCW ISOLATION VALVE LOOP A6105 M-236 X X ECHV0012 F.P.C.H. EX. CCW ISOLATION VALVE LOOP B6104 M-236 X X ECLSL0057 FUEL POOL LEVEL SWITCH 6106 J-481 X X ECLSL0058 FUEL POOL LEVEL SWITCH 6106 J-481 X X ECV0004  S.F. POOL PUMP A DISCHARGE (CK) 6105 M-221 X X ECV0013  S.F. POOL PUMP B DISCHARGE (CK) 6104 M-221 X X ECV0996  FUEL POOL HX B RLF VLV 6104 M-071 X X ECV0997  FUEL POOL HX A RLF VLV 6105 M-071 X X ECV0998  FUEL POOL HX B RLF VLV 6104 M-071 X X ECV0999  FUEL POOL HX A RLF VLV 6105 M-071 X X EEC01A  FUEL POOL COOLING HX LOOP A 6105 M-071 X X EEC01B  FUEL POOL COOLING HX LOOP B 6104 M-071 X X EEG01A  CCW HEAT EXCH A TRAIN A 1406 M-072 X X EEG01B  CCW HEAT EXCH B TRAIN B 1401 M-072 X X EEJ01A  RHR HX A 1310 M-722  X EEJ01B  RHR HX B 1309 M-722  X EF155 ESW CONTROL PNL K104 J-201 X X EF156 ESW CONTROL PNL K105 J-201 X X EFC02  AUX FW PUMP TURBINE L.O. HX 1331 M-021 X X EFHS0026A ISOLATION SWITCH FOR EFHV0026 3302 E-028B X X EFHS0038A ISOLATION SWITCH FOR EFHV0038 3302 E-028B X X EFHV0023 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0024 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0025 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0026 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0031 CONTAINMENT COOLER ISOLATION VALVE 1323 LIMITORQUE X X EFHV0032 CONTAINMENT COOLER ISOLATION VALVE 1322 LIMITORQUE X X EFHV0033 CONTAINMENT COOLER ISOLATON VALVE 2000 LIMITORQUE X X EFHV0034 CONTAINMENT COOLER ISOLATION VALVE 2000 LIMITORQUE X X EFHV0037 ULTIMATE HEAT SINK ISOLATION VALVE 3101 M-235 X X EFHV0038 ULTIMATE HEAT SINK ISOLATION VALVE 3101 M-235 X X EFHV0039 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0040 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0041 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0042 SERVICE WATER ISOLATION VALVE 3101 M-235 X X EFHV0043 VALVE TERMINAL BOX 1320 E-028 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 19)
Component ID Component name Room NoSpec No Hot SD Cold SDEFHV0044 VALVE TERMINAL BOX 1320 E-028 X X EFHV0045 CONTAINMENT COOLER ISOLATION VALVE 2000 LIMITORQUE X X EFHV0046 CONTAINMENT COOLER ISOLATION VALVE 2000 LIMITORQUE X X EFHV0049 CONTAINMENT COOLER ISOLATION VALVE 1323 LIMITORQUE X X EFHV0050 CONTAINMENT COOLER ISOLATION VALVE 1322 LIMITORQUE X X EFHV0051 CCW ISOLATION VALVE 1406 M-236 X X EFHV0052 CCW ISOLATION VALVE 1401 M-236 X X EFHV0059 CCW ISOLATION VALVE 1406 M-236 X X EFHV0060 CCW ISOLATION VALVE 1401 M-236 X X EFHV0091 TRAVELING WATER SCREEN A SPRAY VALVE K105 LIMITORQUE X X EFHV0092 TRAVELING WATER SCREEN B SPRAY VALVE K104 LIMITORQUE X X EFHV0097 ESW PUMP A VENT VALVE K105 LIMITORQUE X X EFHV0098 ESW PUMP B VENT VALVE K104 LIMITORQUE X X EFHY0043 EFHV0043 SOLENOID VALVE 1320 J-601A X X EFHY0044 EFHV0044 SOLENOID VALVE 1320 J-601A X X EFPDI0019 ESW 2A SELF-CLN STR DIFF PRES K105 J-110 X X EFPDI0020 ESW 2B SELF-CLN STR DIFF PRES K104 J-110 X X EFPDS0019A ESW 2A Self-Cln Str. DP Switch K105 J-301 X X EFPDS0020A ESW 2B Self-Cln Str. DP Switch K104 J-301 X X EFPDT0019 ESW 2A SELF-CLN STR DIFF PRES K105 J-301 X X EFPDT0020 ESW 2B SELF-CLN STR DIFF PRES K104 J-301 X X EFPDT0043 ESW TO AIR COMPRESOR ISO 1320 J-301 X X EFPDT0044 ESW TO AIR COMPRESOR ISO 1320 J-301 X X EFPDV0019 STRAINER A TRASH VALVE K105 LIMITORQUE X X EFPDV0020 STRAINER B TRASH VALVE K104 LIMITORQUE X X EFPI0001 ESW PUMP 1A DISCH PRESS 3601 J-110 X X EFPI0002 ESW PMP 1B DISCH PRESS 3601 J-110 X X EFPT0001 ESW PMP 1A DISCH PRESS K105 J-301 X X EFPT0002 ESW PMP 1B DISCH PRESS K104 J-301 X X EFV0001  ESW PUMP A DISCHARGE ISO (CK) K105 M-223B X X EFV0004  ESW PUMP B DISCHARGE ISO (CK) K105 M-223B X X EFV0046  AIR COMPRESSOR ISO VLV (CK) 1320 M-223C X X EFV0076  AIR COMPRESSORS ISO VLV (CK) 1320 M-223C X X EFV0241  ESW A RTN TO UHS (CK) (51) K105 M-223A X X EFV0242  ESW B RTN TO UHS (CK) (51) K104 M-223A X X EFZS0043 EFHV0043 LIMIT SWITCH 1320 J-601A X X EFZS0044 EFHV0044 LIMIT SWITCH 1320 J-601A X X EGFI0062  RCP OUT FLOW 1127 J-517A X X EGFI0128 CCW HX OUT TO RCP BY-PASS 3601 J-110 X X EGFI0129 CCW HX OUT TO RCP BY-PASS 3601 J-110 X X EGFT0062 RCP THERMAL BARRIER (TOTAL) OUTLET FLOW1127 J-301 X X EGFT0107 CCW HX'S FLOW OUT TO RWB NONESSENTIAL COMP 1314 J-301 X X EGFT0108 CCW HX'S FLOW OUT TO RWB NONESSENTIAL COMP 1314 J-301 X X EGFT0128 CCW HX OUT TO RCP BY-PASS 1320 J-301 X X EGFT0129 CCW HX OUT TO RCP BY-PASS 1320 J-301 X X EGHS0070A LOCAL CONTROL STATION 1408 E-028B  X EGHV0011 EMERGENCY MAKEUP WATER TRAIN A FROM ESW1406 M-231B X X EGHV0012 EMERGENCY MAKEUP WATER TRAIN B FROM ESW1401 M-231B X X EGHV0013 EMERGENCY MAKEUP WATER TRAIN A FROM ESW1406 M-231B X X EGHV0014 EMERGENCY MAKEUP WATER TRAIN B FROM ESW1401 M-231B X X EGHV0015 CCW COMMON HEADER RETURN ISO TRAIN A 1402 M-261  X EGHV0016 CCW COMMON HEADER RETURN ISO TRAIN B 1402 M-261  X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 20)
Component ID Component name Room NoSpec No Hot SD Cold SDEGHV0053 TRAIN A CCW TO COMMON HEADER 1402 M-261  X EGHV0054 TRAIN B CCW TO COMMON HEADER 1401 M-261  X EGHV0058 RC PUMP CCW SUPPLY CONT ISO 1323 LIMITORQUE X X EGHV0059 RC PUMP CCW RETURN CONT ISO 1323 LIMITORQUE X X EGHV0060 RC PUMP CCW RETURN CONT ISO 2000 LIMITORQUE X X EGHV0061 CONT ISO VLV CCW RTN FROM RC PMP THER BARR 1323 LIMITORQUE X X EGHV0062 CONT ISO VLV CCW RTN FROM RC PMP THER BARR 2000 LIMITORQUE X X EGHV0069A CCW SUPPLY WASTE HEADER ISO 1314 J-605A  X EGHV0069A VALVE TERMINAL BOX 1314 E-028  X EGHV0069B CCW RETURN WASTE HEADER ISO 1314 J-605A  X EGHV0069B VALVE TERMINAL BOX 1314 E-028  X EGHV0070A CCW SUPPLY WASTE HEADER ISO 1314 J-605A  X EGHV0070A VALVE TERMINAL BOX 1314 E-028  X EGHV0070B CCW RETURN WASTE HEADER ISO 1314 J-605A  X EGHV0070B VALVE TERMINAL BOX 1314 E-028  X EGHV0071 ISO VALVE CCW SUPPLY TO RC PUMP 1323 LIMITORQUE X X EGHV0101 RHR A INLET CCW ISOLATION VALVE 1408 M-236  X EGHV0102 RHR B INLET CCW ISOLATION VALVE 1402 M-236  X EGHV0126 EGHV0071 BYPASS VALVE 1323 LIMITORQUE X X EGHV0127 EGHV0058 BYPASS VALVE 1323 LIMITORQUE X X EGHV0130 EGHV0060 BYPASS VALVE 2000 LIMITORQUE X X EGHV0131 EGHV0059 BYPASS VALVE 1323 LIMITORQUE X X EGHV0132 EGHV0062 BYPASS VALVE 2000 LIMITORQUE X X EGHV0133 EGHV0061 BYPASS VALVE 1323 LIMITORQUE X X EGHY0069A EGHV0069A SOLENOID VALVE 1314 J-605A  X EGHY0069B EGHV0069B SOLENOID VALVE 1314 J-605A  X EGHY0070A EGHV0070A SOLENOID VALVE 1314 J-605A  X EGHY0070B EGHV0070B SOLENOID VALVE 1314 J-605A  X EGLI0001 CCW SURGE TK A LEVEL 3601 J-110 X X EGLI0002 CCW SURGE TK B LEVEL 3601 J-110 X X EGLT0001 CCW SURGE TK A LEVEL 1503 J-301 X X EGLT0002 CCW SURGE TK B LEVEL 1502 J-301 X X EGPT0077 CCW PMPS A&C DISCH PRESS 1406 J-301 X X EGPT0078 CCW PMPS B&D DISCH PRESS 1401 J-301 X X EGTE0031 CCW HX A OUTLET TEMP 1406 J-558B X X EGTE0032 CCW HX B OUTLET TEMP 1401 J-558B X X EGTI0031 CCW HX A OUTLET TEMP 3605 J-110 X X EGTI0032 CCW HX B OUTLET TEMP 3605 J-110 X X EGTT0031 CCW HX A OUTLET TEMP 3605 J-110 X X EGTT0032 CCW HX B OUTLET TEMP 3605 J-110 X X EGTV0029 CCW HX A BYPASS ISO 1406 J-605A X X EGTV0029 VALVE TERMINAL BOX 1406 E-028 X X EGTV0030 CCW HX B BYPASS ISO 1401 J-605A X X EGTV0030 VALVE TERMINAL BOX 1401 E-028 X X EGTY0029A EGTV0029 SOLENOID VALVE 1406 J-605A X X EGTY0030A EGTV0030 SOLENOID VALVE 1401 J-605A X X EGV0003  CCW PUMP A DISCHARGE (CK) 1406 M-223A X X EGV0007  CCW PUMP C DISCHARGE (CK) 1406 M-223A X X EGV0012  CCW PUMP B DISCHARGE (CK) 1401 M-223A X X EGV0016  CCW PUMP D DISCHARGE (CK) 1401 M-223A X X EGV0024  CCW HX A RLF VLV (THERMAL RLF) 1406 M-141 X X EGV0027  CCW HX A RLF VLV (THERMAL RLF) 1406 M-141 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 21)
Component ID Component name Room NoSpec No Hot SD Cold SDEGV0036  TRAIN A CCW TO COMMON HDR (CK) 1402 M-223B  X EGV0049  CCW HX B RLF VLV (THERMAL RLF) 1401 M-141 X X EGV0052  CCW HX B RLF VLV (THERMAL RLF) 1401 M-141 X X EGV0061  TRAIN B CCW TO COMMON HDR (CK) 1401 M-223B  X EGV0124  RCP A-D TBCC RETURN (CK) 1323 M-223C X X EGV0129  RCP A-D RETURN (CK) 1323 M-223B X X EGV0130  CCW COMMON HDR RTN ISO TRAIN A (CK) 1408 M-223B  X EGV0131  CCW COMMON HDR RTN ISO TRAIN B (CK) 1408 M-223B  X EGV0159  SURGE TANK A RLF VLV (VACUUM RLF) 1503 M-141 X X EGV0170  SURGE TANK B RLF VLV (VACUUM RLF) 1502 M-141 X X EGV0204  CTMT ISO VLV CCW SUPPLY TO RC PUMP (CK)2000 M-223B X X EGV0305  SURGE TNK RLF VLV (VACUUM RLF) 1503 M-141 X X EGV0306  SURGE TNK RLF VLV (VACUUM RLF) 1502 M-141 X X EGZS0029 EGTV0029 LIMIT SWITCH 1406 J-605A X X EGZS0030 EGTV0030 LIMIT SWITCH 1401 J-605A X X EGZS0069A EGHV0069A LIMIT SWITCH 1314 J-605A  X EGZS0069B EGHV0069B LIMIT SWITCH 1314 J-605A  X EGZS0070A EGHV0070A LIMIT SWITCH 1314 J-605A  X EGZS0070B EGHV0070B LIMIT SWITCH 1314 J-605A  X EJ8708A  RCL HL 1 TO PRT (RLF) 2000 M-724-3-1  X EJ8708B  RCL HL 4 TO PRT (RLF) 2000 M-724-3-1  X EJ8730A  RHR HEAT EXCH A OUTLET (CHECK) 1310 M-724-8  X EJ8730B  RHR HEAT EXCH B OUTLET (CHECK) 1309 M-724-8  X EJ8841A  RCS BOUNDARY VALVE RCS LOOP 2 & 3 (CHECK) 2000 M-724-8 X X EJ8841B  RCS BOUNDARY VALVE RCS LOOP 2 & 3 (CHECK) 2000 M-724-8 X X EJ8842  HL INJECTION RELIEF 1322 M-724-3-1  X EJ8856A  CL INJECTION TRAIN A RELIEF 1323 M-724-3-1  X EJ8856B  CL INJECTION TRAIN B RELIEF 1322 M-724-3-1  X EJ8958A  RHR SUCT LINE FROM RWST (CHECK) 1111 M-724-8  X EJ8958B  RHR SUCT LINE FROM RWST (CHECK) 1109 M-724-8  X EJ8969A  FEED TO CHARGING/SI PUMP SUCTIONS (CHECK) 1310 M-724-8  X EJ8969B  FEED TO CHARGING/SI PUMP SUCTIONS (CHECK) 1108 M-724-8  X EJFCV0610 RHR MINI FLOW ISOLATION VALVE LOOP A 1111 LIMITORQUE  X EJFCV0611 RHR MINI FLOW ISOLATION VALVE LOOP B 1109 LIMITORQUE  X EJFIS0610 RHR PUMP A MIN. FLOW INDICATING SWITCH 1301 ESE-40A  X EJFIS0611 RHR PUMP B MIN. FLOW INDICATING SWITCH 1107 ESE-40A  X EJHCV8890A VALVE TERMINAL BOX 2000 E-028  X EJHCV8890B VALVE TERMINAL BOX 2000 E-028  X EJHV8701A RHR SHUTDOWN SUCTION LINE ISOLATION VALVE LOOP A 2000 LIMITORQUE X X EJHV8701B RHR SHUTDOWN SUCTION LINE ISOLATION VALVE LOOP B 2000 LIMITORQUE X X EJHV8716A RHR INJECTION BALANCE LINE/HL FEED LINE VALVE 1310 LIMITORQUE  X EJHV8716B RHR INJECTION BALANCE LINE/HL FEED LINE VALVE 1309 LIMITORQUE  X
 
Rev. 27 WOLF CREEK TABLE 7.4-6 (Sheet 22)
Component ID Component name Room NoSpec No Hot SD Cold SDEJHV8804A RHR TO CHARGING SI PUMP SUCTIONS 1310 LIMITORQUE  X EJHV8804B RHR TO CHARGING/SI PUMP SUCTIONS 1108 LIMITORQUE  X EJHV8809A RHR ISOL VALVE TO COLD LEG RCS LOOPS 1 AND 2 1323 LIMITORQUE  X EJHV8809B RHR ISOL VALVE TO COLD LEG RCS LOOPS 3 AND 4 1322 LIMITORQUE  X EJHV8811A CONT RECIRC SUMP ISOLATION VALVE (ENCAPSULATED) 1204 LIMITORQUE  X EJHV8811B CONT RECIRC SUMP ISOLATION VALVE (ENCAPSULATED) 1203 LIMITORQUE  X EJHV8840 RHR ISOL VALVE TO HOT LEGS RCS LOOPS 2 AND 3 1322 LIMITORQUE  X EJHY8890A TEST LINE ISOLATION VALVE COLD LEG INJ LINE SOL 2000 HE-2  X EJHY8890A TESTLINE ISO VLV COLD LEG INJ LINE SOL CONNECTOR 2000 HE-8  X EJHY8890B TEST LINE ISOLATION VALVE COLD LEG INJ LINE SOL 2000 HE-2  X EJHY8890B TESTLINE ISO VLV COLD LEG INJ LINE SOL CONNECTOR 2000 HE-8  X EJTI0608  RHR HX A DISCHARGE TEMP 1310 M-771  X EJTI0609  RHR HX B DISCHARGE TEMP 1309 M-771  X EJV0084  PRESS RELIEF VLV EEJ01A CCW LINE (THERMAL RLF) 1408 M-141  X EJV0085  PRESS RELIEF VLV EEJ01B CCW LINE (THERMAL RLF) 1402 M-141  X EJV0156  PRESS RELIEF VLV PEJ01A CCW LINE (THERMAL RLF) 1111 M-141-2  X EJV0157  PRESS RELIEF VLV PEJ01B CCW LINE (THERMAL RLF) 1109 M-141-2  X EJXJ0015  ENCAP. TANK B INLET EXPANSION JOINT 1203 M-312  X EJXJ0016  ENCAP. TANK A INLET EXPANSION JOINT 1204 M-312  X EJXJ0017  ENCAP. TANK B OUTLET EXPANSION JOINT 1203 M-312  X EJXJ0018  ENCAP. TANK A OUTLET EXPANSION JOINT 1204 M-312  X EJZS8701BA VALVE TERMINAL BOX (EJZS8701BB) 2000 E-028 X X EJZS8701BA RHR PUMP SUCT LINE ISOL VLV LOOP B LS CONNECTOR 2000 HE-8 X X EJZS8701BA RHR SHUTDOWN SUCT LINE ISOL VLV LOOP B LMT SW 2000 HE-3 X X EJZS8701BB RHR PUMP SUCT LINE ISOL VLV LOOP B LS CONNECTOR 2000 HE-8 X X EJZS8701BB RHR SHUTDOWN SUCT LINE ISOL VLV LOOP B LMT SW 2000 HE-3 X X EJZS8804A RHR TO CHARGING/SI PUMP SUCTIONS LIMIT SWITCH 1310 HE-6  X EJZS8804BA RHR TO CHARGING/SI PUMP SUCTIONS LIMIT SWITCH 1108 HE-6  X EJZS8804BB RHR TO CHARGING/SI PUMP SUCTIONS LIMIT SWITCH 1108 HE-6  X EJZS8809AA RHR ISOL VLV TO COLD LEG RCS LOOPS 1&2 LIMIT SW 1323 HE-6  X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 23)
Component ID Component name Room NoSpec No Hot SD Cold SDEJZS8809BA RHR ISOL VLV TO COLD LEG RCS LOOPS 3&4 LIMIT SW 1322 HE-6  X EJZS8811A CONT RECIRC SUMP ISOLATION VALVE LIMIT SWITCH 1204 HE-6  X EJZS8811A CONT RECIRC SUMP ISOLATION VALVE LS CONNECTOR 1204 HE-8  X EJZS8811B CONT RECIRC SUMP ISOLATION VALVE LS CONNECTOR 1203 HE-8  X EJZS8811B CONT RECIRC SUMP ISOLATION VALVE LIMIT SWITCH 1203 HE-6  X EJZS8840A RHR ISOL VLV TO HOT LEGS RCS LOOPS 2&3 LIMIT SW 1322 HE-6  X EJZS8890A TEST LINE ISO VLV COLD LEG INJ LINE LS CONNECTOR 2000 HE-8  X EJZS8890A TEST LINE ISOL VLV COLD LEG INJ LINE LIMIT SW 2000 HE-3  X EJZS8890B TEST LINE ISO VLV COLD LEG INJ LINE LS CONNECTOR 2000 HE-8  X EJZS8890B TEST LINE ISOL VLV COLD LEG INJ LINE LIMIT SW 2000 HE-3  X EKJ01A  JACKET WATER HTR A 5203 M-018 X X EKJ01B  JACKET WATER HTR B 5201 M-018 X X EKJ02A  LUBE OIL HTR A 5203 M-018 X X EKJ02B  LUBE OIL HTR B 5201 M-018 X X EKJ03A  INTERCOOLER HEAT EXCHANGER A 5203 M-018C X X EKJ03B  INTERCOOLER HEAT EXCHANGER B 5201 M-018C X X EKJ04A  LUBE OIL COOLER A 5203 M-018 X X EKJ04B  LUBE OIL COOLER B 5201 M-018 X X EKJ06A  JACKET WATER HX A 5203 M-018E X X EKJ06B  JACKET WATER HX B 5201 M-018E X X EM8815  DISCH TO RCS (CHECK) 2000 M-724-8  X EMFI0917A CHARGING PUMP DISCHARGE FLOW 3601 ESE-14 X X EMFI0917B CHARGING PUMP DISCHARGE FLOW 3601 ESE-14 X X EMFS0917C CHARGING PUMP DISCHARGE MINI FLOW SWITCH 1126 ESE-40A X X EMFS0917D CHARGING PUMP DISCHARGE MINI FLOW SWITCH 1126 ESE-40A X X EMFT0917A CHARGING PUMP TO BIT FLOW TRANSMITTER 1126 ESE-4 X X EMFT0917B CHARGING PUMP TO BIT FLOW TRANSMITTER 1126 ESE-4 X X EMHS8843 LOCAL CONTROL STATION 1409 E-028 X X EMHV8801A BIT DISCHARGE VALVE TO RCS 1323 LIMITORQUE X X EMHV8801B BIT DISCHARGE VALVE TO RCS 1323 LIMITORQUE X X EMHV8803A CHARGING PUMP DISCHARGE VALVE TO BIT 1126 LIMITORQUE X X EMHV8803B CHARGING PUMP DISCHARGE VALVE TO BIT 1126 LIMITORQUE X X EMHV8807A RHR TO CVCS AND SI PUMPS ISOL VALVE 1113 LIMITORQUE  X EMHV8807B RHR TO CVCS AND SI PUMPS ISOL VALVE 1108 LIMITORQUE  X EMHV8843 VALVE TERMINAL BOX 2000 E-028 X X EMHV8882  BIT TEST LINE ISO VLV 2000 M-724-7-1 X X EMHV8889A  HL 1 SI TEST LINE 2000 M-724-7-1 X X EMHV8889B  HL 2 SI TEST LINE 2000 M-724-7-1 X X EMHV8889C  HL 3 SI TEST LINE 2000 M-724-7-1 X X EMHV8889D  HL 4 SI TEST LINE 2000 M-724-7-1 X X
 
Rev. 25 WOLF CREEK TABLE 7.4-6 (Sheet 24)
Component ID Component name Room NoSpec No Hot SD Cold SDEMHY8843 BIT TEST LINE AND CONT ISOL VALVE CONNECTOR 2000 HE-8 X X EMHY8843 BIT TESTLINE AND CONTAINMENT ISOLATION VALVE 2000 HE-2 X X EMV0001  CTMT ISOL (CHECK) 2000 M-231C X X EMV0002  CTMT ISOL (CHECK) 2000 M-231C X X EMV0003  CTMT ISOL (CHECK) 2000 M-231C X X EMV0004  CTMT ISOL (CHECK) 2000 M-231C X X EMV0188  SI PUMP B CCW LINE PRESS RELIEF VLV (THERM RLF) 1108 M-141-2 X X EMV0189  SI PUMP A CCW LINE PRESS RELIEF VLV (THERM RLF) 1113 M-141-2 X X EMZS8843 BIT TEST LINE CONT ISOL VLV LS CONNECTOR 2000 HE-8 X X EMZS8843 BIT TESTLINE AND CONTAINMENT ISOL VLV LIMIT SW 2000 HE-3 X X EP8818A  RC BOUNDARY VLV RHR PUMP (CHECK) 2000 M-724-8 X X EP8818B  RC BOUNDARY VLV RHR PUMP (CHECK) 2000 M-724-8 X X EP8818C  RC BOUNDARY VLV RHR PUMP (CHECK) 2000 M-724-8 X X EP8818D  RC BOUNDARY VLV RHR PUMP (CHECK) 2000 M-724-8 X X EP8956A  RC BOUNDARY VLV ACC TANK A (CHECK) 2000 M-724-8 X X EP8956B  RC BOUNDARY VLV ACC TANK B (CHECK) 2000 M-724-8 X X EP8956C  RC BOUNDARY VLV ACC TANK C (CHECK) 2000 M-724-8 X X EP8956D  RC BOUNDARY VLV ACC TANK D (CHECK) 2000 M-724-8 X X EPHV8808A ACCUMULATOR TANK A ISOLATION VALVE 2000 LIMITORQUE X X EPHV8808B ACCUMULATOR TANK B ISOLATION VALVE 2000 LIMITORQUE X X EPHV8808C ACCUMULATOR TANK C ISOLATION VALVE 2000 LIMITORQUE X X EPHV8808D ACCUMULATOR TANK D ISOLATION VALVE 2000 LIMITORQUE X X EPHV8879A  ACCUM TANK A TO SIS TEST LINE ISO VLV 2000 M-724-7-1 X X EPHV8879B  ACCUM TANK B TO SIS TEST LINE ISO VLV 2000 M-724-7-1 X X EPHV8879C  ACCUM TANK C TO SIS TEST LINE ISO VLV 2000 M-724-7-1 X X EPHV8879D  ACCUM TANK D TO SIS TEST LINE ISO VLV 2000 M-724-7-1 X X EPHV8950A ACCUMULATOR TANK A VENT VALVE 2000 HE-10A  X EPHV8950A ACCUMULATOR TANK A VENT VALVE CONNECTOR2000 HE-8  X EPHV8950B ACCUMULATOR TANK B VENT VALVE 2000 HE-10A  X EPHV8950B ACCUMULATOR TANK B VENT VALVE CONNECTOR2000 HE-8  X EPHV8950C ACCUMULATOR TANK B VENT VALVE 2000 HE-10A  X EPHV8950C ACCUMULATOR TANK B VENT VALVE CONNECTOR2000 HE-8  X EPHV8950D ACCUMULATOR TANK C VENT VALVE CONNECTOR2000 HE-8  X EPHV8950D ACCUMULATOR TANK C VENT VALVE 2000 HE-10A  X EPHV8950E ACCUMULATOR TANK C VENT VALVE 2000 HE-10A  X EPHV8950E ACCUMULATOR TANK C VENT VALVE CONNECTOR2000 HE-8  X EPHV8950F ACCUMULATOR TANK D VENT VALVE 2000 HE-10A  X EPHV8950F ACCUMULATOR TANK D VENT VALVE CONNECTOR2000 HE-8  X EPV0010  RC BOUNDARY VLV SI PUMP (CHECK) 2000 M-231C X X EPV0020  RC BOUNDARY VLV SI PUMP (CHECK) 2000 M-231C X X EPV0030  RC BOUNDARY VALVE SI PUMP (CHECK) 2000 M-231C X X EPV0040  RC BOUNDARY VLV SI PUMP (CHECK) 2000 M-231C X X EPZS8808AB ACCUMULATOR TANK A ISOL VLV LS CONNECTOR 2000 HE-8  X EPZS8808AB ACCUMULATOR TANK A ISOLATION VALVE LIMIT SWITCH 2000 HE-3  X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 25)
Component ID Component name Room NoSpec No Hot SDCold SDEPZS8808AB VALVE TERMINAL BOX 2000 E-028  X EPZS8808BB ACCUMULATOR TANK B ISOL VLV LS CONNECTOR 2000 HE-8  X EPZS8808BB ACCUMULATOR TANK B ISOLATION VALVE LIMIT SWITCH 2000 HE-3  X EPZS8808BB VALVE TERMINAL BOX 2000 E-028  X EPZS8808CB ACCUMULATOR TANK C ISOL VLV LS CONNECTOR 2000 HE-8  X EPZS8808CB ACCUMULATOR TANK C ISOLATION VALVE LIMIT SWITCH 2000 HE-3  X EPZS8808CB VALVE TERMINAL BOX 2000 E-028  X EPZS8808DB ACCUMULATOR TANK D ISOL VLV LS CONNECTOR 2000 HE-8  X EPZS8808DB ACCUMULATOR TANK D ISOLATION VALVE LIMIT SWITCH 2000 HE-3  X EPZS8808DB VALVE TERMINAL BOX 2000 E-028  X FBG04A  SEAL WATER INJECTION FILTER 1302 M-723-1 X X FBG04B  SEAL WATER INJECTION FILTER 1302 M-723-1 X X FBG07  BORIC ACID FILTER 1302 M-723 X X FBG08A  PBG05A LUBE OIL FILTER 1114 M-721-1 X X FBG08B  PBG05B LUBE OIL FILTER 1107 M-721-1 X X FC219 AFP TURBINE GAUGE AND CONTROL PANEL 1331 M-021A X X FCFO0096  TAFP LUBE OIL FLOW ORIFICE 1331 M-021 X X FCFO0097  TAFP LUBE OIL FLOW ORIFICE 1331 M-021 X X FCFO0098  TAFP LUBE OIL FLOW ORIFICE 1331 M-021 X X FCFO0099  TAFP LUBE OIL FLOW ORIFICE 1331 M-021 X X FCFV0313 TURBINE SPEED-GOVERNING VALVE 1331 M-021, M-021A X X *FCHIS0313A AUX FP TURB SPEED CTRL 3601 M-021A X X *FCHIS0313B AUX FDWTR PMP SPEED GOVERNOR 1413 M-021A X X FCHV0312 TURBINE MANUAL TRIP AND THROTTLE VALVE 1331 M-021 X X FCV0001  MAIN STEAM ISO (CK) 1412 M-224B X X FCV0002  MAIN STEAM ISO (CK) 1412 M-224B X X FCV0003  AUX STEAM ISO (CK) 1331 M-224B X X FCV0024  MAIN STEAM ISO (CK) 1412 M-224B X X FCV0025  MAIN STEAM ISO (CK) 1412 M-224B X X FCV0999  TAFP LUBE OIL RELIEF (RLF) 1331 M-021 X X FEF01A TRAVELING WATER SCREEN A K105 M-020 X X FEF01B TRAVELING WATER SCREEN B K104 M-020 X X FEF02A SELF-CLEANING STRAINER A K105 M-154 X X FEF02B SELF-CLEANING STRAINER B K104 M-154 X X FEF03A  ESW PRELUBE STD TK FILTER TR. A K105 M-105B X X FEF03B  ESW PRELUBE STD TK FILTER TR. B K104 M-105B X X FFC02  TURBINE L.O. FILTER 1331 M-021 X X FKJ02A  INTAKE AIR FILTER A 5203 M-018 X X FKJ02B  INTAKE AIR FILTER B 5203 M-018 X X FKJ02C  INTAKE AIR FILTER C 5201 M-018 X X FKJ02D  INTAKE AIR FILTER D 5201 M-018 X X FKJ06A  STARTING AIR TO INSTR. FILTER A 5203 M-018 X X FKJ06B  STARTING AIR TO INSTR. FILTER B 5201 M-018 X X FKJ07A  FUEL OIL FILTER A 5203 M-018 X X FKJ07B  FUEL OIL FILTER B 5201 M-018 X X FKJ08A  MAIN LUBE OIL STRAINER A 5203 M-018 X X FKJ08B  MAIN LUBE OIL STRAINER B 5201 M-018 X X FKJ09A  DIESEL LUBE OIL FILTER A 5203 M-018 X X 
 
Rev. 27 WOLF CREEK TABLE 7.4-6 (Sheet 26)
Component ID Component name Room NoSpec No Hot SD Cold SDFKJ09B  DIESEL LUBE OIL FILTER B 5201 M-018 X X FKJ10A  DIESEL ROCKER LUBE OIL FILTER A 5203 M-018 X X FKJ10B  DIESEL ROCKER LUBE OIL FILTER B 5201 M-018 X X FKJ11A  DIESEL OIL SEPARATOR A 5203 M-018 X X FKJ11B DIESEL OIL SEPARATOR B 5201 M-018 X X FKJ12A  DIESEL LUBE OIL SUCTION STRAINER A 5203 M-018 X X FKJ12B  DIESEL LUBE OIL SUCTION STRAINER B 5201 M-018 X X GDD0001  BACKDRAFT DAMPER ESW PUMPHOUSE K105 M-627A X X GDD0002  FLOW CONTROL DAMPER ESW PUMPHOUSE K105 M-627A X X GDD0003  TORNADO DAMPER ESW PUMPHOUSE K105 M-627A X X GDD0004  FLOW CONTROL DAMPER ESW PUMPHOUSE K105 M-627A X X GDD0008  BACKDRAFT DAMPER ESW PUMPHOUSE K104 M-627A X X GDD0009  FLOW CONTROL DAMPER ESW PUMPHOUSE K104 M-627A X X GDD0010  TORNADO DAMPER ESW PUMPHOUSE K104 M-627A X X GDD0011  FLOW CONTROL DAMPER ESW PUMPHOUSE K104 M-627A X X GDD0012  FLOW CONTROL DAMPER ESW PUMPHOUSE K105 M-627A X X GDD0013  FLOW CONTROL DAMPER ESW PUMPHOUSE K104 M-627A X X GDTE0001 ESW PMP RM TEMP 1A DAMPER K105 J-558B X X GDTE0011 ESW PMP RM TEMP 1B DAMPER K104 J-558B X X GDTT0001 ESW PMP RM TEMP 1A DAMPER 3605 J-110 X X GDTT0011 ESW PMP RM TEMP 1B DAMPER 3605 J-110 X X GDTZ0001A FLOW CONTROL DAMPER ACTUATOR - GDD0002 - ELC/HYD K105 M-627A X X GDTZ0001B FLOW CONTROL DAMPER ACTUATOR - GDD0004 - ELC/HYD K105 M-627A X X GDTZ0001C FLOW CONTROL DAMPER ACTUATOR - GDD0012 - ELC/HYD K105 M-627A X X GDTZ0011A FLOW CONTROL DAMPER ACTUATOR - GDD0009 - ELC/HYD K104 M-627A X X GDTZ0011B FLOW CONTROL DAMPER ACTUATOR - GDD0011 - ELC/HYD K104 M-627A X X GDTZ0011C FLOW CONTROL DAMPER ACTUATOR - GDD0013 - ELC/HYD K104 M-627A X X GK195A SGK05A CONTROL PANEL 3416 M-622.1A X X GK195B SGK05A POWER PANEL 3416 M-622.1A X X GK195C SGK05A POWER AND CONTROL PANEL 3416 M-622.1A X X GK196A SGK05B CONTROL PANEL 3415 M-622.1A X X GK196B SGK05B POWER PANEL 3415 M-622.1A X X GK196C SGK05B POWER AND CONTROL PANEL 3415 M-622.1A X X GK198A SGK04A CONTROL PANEL 1512 M-622.1A X X GK198B SGK04A POWER PANEL 1512 M-622.1A X X GK198C SGK04A POWER AND CONTROL PANEL 1512 M-622.1A X X GK199A SGK04B CONTROL PANEL 1501 M-622.1A X X GK199B SGK04B POWER PANEL 1501 M-622.1A X X GK199C SGK04B POWER AND CONTROL PANEL 1501 M-622.1A X X GKHZ0029A A/C SYS ISO DAMPER ACTUATOR - GKD0080 -
ELEC 1512 M-627A X X GKHZ0029B A/C SYS ISO DAMPER ACTUATOR - GKD0081 -
ELEC 1512 M-627A X X GKHZ0040A A/C UNIT ISO DAMPER ACTUATOR - GKD0084 - ELEC 1501 M-627A X X GKHZ0040B CTRL ROOM ISO DAMPER ACTUATOR -GKD0085 - ELEC 1501 M-627A X X GKV0765  SGK04A WATER REGULATING VALVE 1512 M-622.1A X X GKV0766  SGK04B WATER REGULATING VALVE 1501 M-622.1A X X GKV0767  SGK05A WATER REGULATING VALVE 3416 M-622.1A X X              Rev. 19 WOLF CREEK TABLE 7.4-6 (Sheet 27)
Component ID Component name Room NoSpec No Hot SD Cold SDGKV0768  SGK05B WATER REGULATING VALVE 3415 M-622.1A X X GLHZ0080 AUX BLDG ISO DAMPER ACTUATOR - GLD0157 - ELEC 1406 M-627A X X GLHZ0081 AUX BLDG ISO DAMPER ACTUATOR - GLD0156 - ELEC 1406 M-627A X X GMD0001  D.G. BLDG AIR INTAKE BACKDRAFT DAMPER 5203 M-627A X X GMD0002  D.G. BLDG AIR INTAKE ISO DAMPER 5203 M-627A X X GMD0003  D.G. BLDG SUPPLY ISO DAMPER 5203 M-627A X X GMD0004  D.G. BLDG EXHAUST TORNADO DAMPER 5203 M-627A X X GMD0005  D.G. BLDG EXHAUST ISO DAMPER 5203 M-627A X X GMD0006  D.G. BLDG AIR INTAKE BACKDRAFT DAMPER 5201 M-627A X X GMD0007  D.G. BLDG AIR INTAKE ISO DAMPER 5201 M-627A X X GMD0008  D.G. BLDG SUPPLY ISO DAMPER 5201 M-627A X X GMD0009  D.G. BLDG EXHAUST ISO DAMPER 5201 M-627A X X GMD0010  D.G. BLDG EXHAUST ISO DAMPER 5201 M-627A X X GNPI0938 CONTAINMENT ATM PRESSURE 3601 ESE-14 X  GNPI0939 CONTAINMENT ATM PRESSURE 3601 ESE-14 X  GNPT0938 CONTAINMENT PRESSURE TRANSMITTER WIDE RANGE 1410 J-301 or ESE-4A (14) X  GNPT0939 CONTAINMENT PRESSURE TRANSMITTER WIDE RANGE 1409 J-301 or ESE-4A (14) X  GNTE0060 CTMT COOLER A TEMP 2000 J-558B X X GNTE0061 CTMT COOLER B TEMP 2000 J-558B X X GNTE0062 CTMT COOLER C TEMP 2000 J-558B X X GNTE0063 CTMT COOLER D TEMP 2000 J-558B X X GNTI0060 CTMT ATMOS TEMP 3601 J-110 X X GNTI0061 CTMT ATMOS TEMP 3601 J-110 X X GNTI0062 CTMT ATMOS TEMP 3601 J-110 X X GNTI0063 CTMT ATMOS TEMP 3601 J-110 X X GNTR0063 CTMT ATMOS TEMP 3601 J-110 X X GNTT0060 CTMT COOLER A TEMP 3605 J-110 X X GNTT0061 CTMT COOLER B TEMP 3605 J-110 X X GNTT0062 CTMT COOLER C TEMP 3605 J-110 X X GNTT0063 CTMT COOLER D TEMP 3605 J-110 X X HBV0036  RCDT HX CCW LINE PRESS RLF VLV (THERMAL RLF) 2000 M-141-2 X X JELI0012A EMER FUEL OIL DAY TK A LEV 3601 J-110 X X JELI0012B EMER FUEL OIL DAY TK A LEV 5203 J-110 X X JELI0032A EMER FUEL OIL DAY TK B LEV 3601 J-110 X X JELI0032B EMER FUEL OIL DAY TK B LEV 5201 J-110 X X JELT0001 EMER FUEL OIL DAY TK A LEV 5203 J-301 X X JELT0010  EMER FUEL OIL DAY TK A LEV 5203 J-301 X X JELT0012 EMER FUEL OIL DAY TK A LEV 5203 J-301 X X JELT0021 EMER FUEL OIL DAY TK B LEV 5201 J-301 X X JELT0030  EMER FUEL OIL DAY TK B LEV 5201 J-301 X X JELT0032 EMER FUEL OIL DAY TK B LEV 5201 J-301 X X JEYS0001  PUMP DISCHARGE WYE STRAINER 5203 M-157 X X JEYS0002  PUMP DISCHARGE WYE STRAINER 5203 M-157 X X JEYS0003  PUMP DISCHARGE WYE STRAINER 5201 M-157 X X JEYS0004  PUMP DISCHARGE WYE STRAINER 5201 M-157 X X
 
Rev. 19 WOLF CREEK TABLE 7.4-6 (Sheet 28)
Component ID Component name Room NoSpec No Hot SD Cold SD KAPCV0101  REGULATING VLV (AIR) 1305 J-601A X X KAPCV0102  REGULATING VLV (AIR) 1305 J-601A X X KAPCV0103  REGULATING VLV (AIR) 1304 J-601A X X KAPCV0200  REGULATING VLV (AIR) 1304 J-601A X X KAV0703  BACK-UP GAS SUPPLY ACCUM TANK RELIEF VALVE 1304 M-141-2 X X KAV0704  BACK-UP GAS SUPPLY ACCUM TANK RELIEF
 
VALVE 1305 M-141 X X KAV0705  BACK-UP GAS SUPPLY ACCUM TANK RELIEF
 
VALVE 1305 M-141 X X KAV0706  BACK-UP GAS SUPPLY ACCUM TANK RELIEF
 
VALVE 1304 M-141-2 X X KAV0710  BACK-UP GAS SUPPLY LINE PRESS RELIEF
 
VALVE 1304 M-141-2 X X KAV0711  BACK-UP GAS SUPPLY LINE PRESS. RELIEF
 
VALVE 1305 M-141 X X KAV0712  BACK-UP GAS SUPPLY LINE PRESS. RELIEF VALVE 1305 M-141 X X KAV0713  BACK-UP GAS SUPPLY LINE PRESS RELIEF VALVE 1304 M-141-2 X X KFC02 AUX FEED PUMP TURBINE 1331 M-021 X X KJ121 DIESEL GAUGE AND CONTROL PANEL 5203 M-018 X X KJ122 DIESEL GAUGE AND CONTROL PANEL 5201 M-018 X X KJBS0001A  D.G. FUEL OIL STRNR A 5203 M-018 X X KJBS0101A  D.G. FUEL OIL STRNR B 5201 M-018 X X KJFO0001A  D.G. A START AIR SYS FLOW ORIFICE 5203 M-018 X X KJFO0002A  D.G. A COOLING WTR FLOW ORIFICE 5203 M-018 X X KJFO0002B  D.G. B COOLING WTR FLOW ORIFICE 5201 M-018 X X KJFO0003A  D.G. A COOLING WTR FLOW ORIFICE 5203 M-018 X X KJFO0003B  D.G. B COOLING WTR FLOW ORIFICE 5201 M-018 X X KJFO0004A  D.G. A JACKET WTR HX FLOW ORIFICE 5203 M-018 X X KJFO0004B  D.G. B JACKET WTR HX FLOW ORIFICE 5201 M-018 X X KJFO0005A  D.G. A COOLING WTR FLOW ORIFICE 5203 M-018 X X KJFO0005B  D.G. B COOLING WTR FLOW ORIFICE 5201 M-018 X X KJFO0101A  D.G. B START AIR SYS FLOW ORIFICE 5201 M-018 X X KJHV0001 ESW TO D.G. AFTERCOOLERS A AND B 5203 LIMITORQUE X X KJHV0002 ESW FROM D.G. AFTERCOOLERS A AND B 5203 LIMITORQUE X X KJHV0101 ESW TO D.G. AFTERCOOLERS A AND B 5201 LIMITORQUE X X KJHV0102 ESW FROM D.G. AFTERCOOLERS A AND B 5201 LIMITORQUE X X KJLCV0019 LUBE OIL AUX TNK FROM L.O. SUCTION
 
STRAINER A 5203 M-018 X X KJLCV0027  ROCKER RESERVOIR FROM D.G. A 5203 M-018 X X KJLCV0119 LUBE OIL AUX TNK FROM L.O. SUCTION
 
STRAINER B 5201 M-018 X X KJLCV0127  ROCKER RESERVOIR FROM D.G. B 5201 M-018 X X KJLI0031 L.O. LEVEL CONTROL TANK INDICATOR 5203 M-018 X X KJLI0038 L.O. AUX. TANK LEVEL INDICATOR 5203 M-018 X X KJLI0131 L.O. LEVEL CONTROL TANK INDICATOR 5201 M-018 X X KJLI0138 L.O. AUX. TANK LEVEL INDICATOR 5201 M-018 X X KJLS0019  L.O. LEVEL CNTRL TNK LEVEL 5203 M-018 X X KJLS0119  L.O. LEVEL CNTRL TANK LEVEL 5201 M-018 X X KJLSH0027  L.O. ROCKER RES LEVEL 5203 M-018 X X KJLSH0036  L.O. LEVEL CNTRL TNK LEVEL 5203 M-018 X X KJLSH0127  L.O. ROCKER RESERVOIR LEVEL 5201 M-018 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 29)
Component ID Component name Room NoSpec No Hot SD Cold SDKJLSH0136  L.O. LEVEL CNTRL TNK LEVEL 5201 M-018 X X KJLSL0032  L.O. LEVEL CNTRL TNK LEVEL 5203 M-018 X X KJLSL0067  JACKET WTR EXP TANK LEVEL 5203 M-018 X X KJLSL0069  JACKET WTR EXP TANK LEVEL 5203 M-018 X X KJLSL0132  L.O. LEVEL CNTRL TNK LEVEL 5201 M-018 X X KJLSL0167  JACKET WTR EXP TNK LEVEL 5201 M-018 X X KJLSL0169  JACKET WTR EXP TANK LEVEL 5201 M-018 X X KJLT0031  L.O. LEVEL CNTRL TNK LEVEL 5203 M-018 X X KJLT0038  L.O. AUX TANK PRESS 5203 M-018 X X KJLT0131  L.O. LEVEL CNTRL TNK LEVEL 5201 M-018 X X KJLT0138  L.O. AUX TANK LEVEL 5201 M-018 X X KJPCV0072A  STARTING AIR SYSTEM D.G. A 5203 M-018 X X KJPCV0072B  STARTING AIR SYSTEM D.G. A 5203 M-018 X X KJPCV0172A  STARTING AIR SYSTEM D.G. B 5201 M-018 X X KJPCV0172B  STARTING AIR SYSTEM D.G. B 5201 M-018 X X KJPDI0010  FUEL OIL PRESS 5203 M-018 X X Deleted            KJPDI0022  L.O. PRESS 5203 M-018 X X KJPDI0028  L.O. PRESS 5203 M-018 X X KJPDI0037  L.O. PRESS 5203 M-018 X X KJPDI0110  FUEL OIL PRESS 5201 M-018 X X Deleted            KJPDI0122  L.O. PRESS 5201 M-018 X X KJPDI0128  L.O. PRESS 5201 M-018 X X KJPDI0137  L.O. PRESS 5201 M-018 X X KJPDSH0010  FUEL OIL PRESS 5203 M-018 X X KJPDSH0011  FUEL OIL PRESS 5203 M-018 X X KJPDSH0022  L.O. PRESS 5203 M-018 X X KJPDSH0028  L.O. PRESS 5203 M-018 X X KJPDSH0037  L.O. PRESS 5203 M-018 X X KJPDSH0110  FUEL OIL PRESS 5201 M-018 X X KJPDSH0111  FUEL OIL PRESS 5201 M-018 X X KJPDSH0122  L.O. PRESS 5201 M-018 X X KJPDSH0128  L.O. PRESS 5201 M-018 X X KJPDSH0137  L.O. PRESS 5201 M-018 X X KJPI0003A  STARTING AIR TNK PRESS 5203 M-018 X X KJPI0003B  STARTING AIR TNK PRESS 5203 M-018 X X KJPI0007  STARTING AIR PRESS 5203 M-018 X X KJPI0072A  D.G. A START AIR PRESSURE 5203 M-018 X X KJPI0072B  D.G. A START AIR PRESSURE 5203 M-018 X X KJPI0093  D.G. A MANIFOLD AIR PRESSURE 5203 M-018 X X KJPI0095  L.O. PRESS 5203 M-018 X X KJPI0098A  D.G. A TURBOCHARGER INLET PRESS 5203 M-018 X X KJPI0098B  D.G. A TURBOCHARGER INLET PRESS 5203 M-018 X X KJPI0103A  STARTING AIR TNK PRESS 5201 M-018 X X KJPI0103B  STARTING AIR TNK PRESS 5201 M-018 X X KJPI0107  STARTING AIR PRESS 5201 M-018 X X KJPI0172A  D.G. B START. AIR PRESS 5201 M-018 X X KJPI0172B  D.G. B START. AIR PRESS 5201 M-018 X X KJPI0193  D.G. B MANIFOLD AIR PRESSURE 5201 M-018 X X KJPI0195  L.O. PRESS 5201 M-018 X X KJPI0198A  D.G. B TURBOCHARGER INLET PRESS 5201 M-018 X X
 
Rev. 27 WOLF CREEK TABLE 7.4-6 (Sheet 30)
Component ID Component name Room NoSpec No Hot SD Cold SD KJPI0198B  D.G. B TURBOCHARGER INLET PRESS 5201 M-018 X X KJPS0062 JACKET WATER BACKUP TO ELECT SPEED SWITCH 5203 M-018 X X KJPS0162 JACKET WATER BACKUP TO ELECT SPEED SWITCH 5201 M-018 X X KJPSH0023A HIGH CRANKCASE PRESSURE 5203 M-018 X X KJPSH0023B HIGH CRANKCASE PRESSURE 5203 M-018 X X KJPSH0023C HIGH CRANKCASE PRESSURE 5203 M-018 X X KJPSH0023D HIGH CRANKCASE PRESSURE 5203 M-018 X X KJPSH0123A HIGH CRANKCASE PRESS 5201 M-018 X X KJPSH0123B HIGH CRANKCASE PRESS 5201 M-018 X X KJPSH0123C HIGH CRANKCASE PRESS 5201 M-018 X X KJPSH0123D HIGH CRANKCASE PRESS 5201 M-018 X X KJPSHL0002A AIR COMPR. A START/STOP PRESS. SWITCH 5203 M-018 X X KJPSHL0002B AIR COMPR. B START/STOP PRESS. SWITCH 5203 M-018 X X KJPSHL0102A AIR COMPR. A START/STOP PRESS. SWITCH 5201 M-018 X X KJPSHL0102B AIR COMPR. B START/STOP PRESS. SWITCH 5201 M-018 X X KJPSL0006A  STARTING AIR PRESS 5203 M-018 X X KJPSL0006B  STARTING AIR PRESS 5203 M-018 X X KJPSL0012  FUEL OIL PRESS 5203 M-018 X X KJPSL0026A LOW LUBE OIL PRESS 5203 M-018 X X KJPSL0026B LOW LUBE OIL PRESS 5203 M-018 X X KJPSL0026C LOW LUBE OIL PRESS 5203 M-018 X X KJPSL0026D LOW LUBE OIL PRESS 5203 M-018 X X KJPSL0029  L.O. PRESS 5203 M-018 X X KJPSL0057  ENG DR INTERCOOLER PMP DISCH 5203 M-018 X X KJPSL0064  ENG DR JACK WTR PMP DISCH 5203 M-018 X X KJPSL0098A  D.G. A TURBOCHARGER INLET PRESS SWITCH 5203 M-018 X X KJPSL0098B  D.G. A TURBOCHARGER INLET PRESS SWITCH 5203 M-018 X X KJPSL0106A  STARTING AIR PRESS 5201 M-018 X X KJPSL0106B  STARTING AIR PRESS 5201 M-018 X X KJPSL0112  FUEL OIL PRESS 5201 M-018 X X KJPSL0126A LOW LUBE OIL PRESS 5201 M-018 X X KJPSL0126B LOW LUBE OIL PRESS 5201 M-018 X X KJPSL0126C LOW LUBE OIL PRESS 5201 M-018 X X KJPSL0126D LOW LUBE OIL PRESS 5201 M-018 X X KJPSL0129  L.O. PRESS 5201 M-018 X X KJPSL0157  ENG DR INTERCOOLER PMP DISCH PRESS 5201 M-018 X X KJPSL0164  ENG DR JACK WTR PMP DISCH PRESS 5201 M-018 X X KJPSL0198A  D.G. B TURBOCHARGER INLET PRESS SWITCH 5201 M-018 X X KJPSL0198B  D.G. B TURBOCHARGER INLET PRESS SWITCH 5201 M-018 X X KJPT0013  FUEL OIL PRESS 5203 M-018 X X KJPT0014  FUEL OIL PRESS 5203 M-018 X X KJPT0024  CRANKCASE PRESS 5203 M-018 X X KJPT0026  L.O. PRESS 5203 M-018 X X KJPT0029  L.O. PRESS 5203 M-018 X X KJPT0057  ENG DR INTERCOOLER PMP DISCH PRESS 5203 M-018 X X KJPT0064  ENG DR JACK WTR PMP DISCH PRESS 5203 M-018 X X KJPT0113  FUEL OIL PRESS 5201 M-018 X X KJPT0114  FUEL OIL PRESS 5201 M-018 X X KJPT0124  CRANKCASE PRESSURE 5201 M-018 X X KJPT0126  L.O. PRESS 5201 M-018 X X KJPT0129  L.O. PRESS 5201 M-018 X X KJPT0157  ENG DR INTERCOOLER PMP DISCH PRESS 5201 M-018 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 31)
Component ID Component name Room NoSpec No Hot SD Cold SD KJPT0164  ENG DR JACK WTR PMP DISCH PRESS 5201 M-018 X X KJPV0001A AIR START SOL VLV 1 5203 M-018 X X KJPV0001B AIR START SOL VLV 2 5203 M-018 X X KJPV0008 SHUTDOWN AIR SOL VLV 5203 M-018 X X KJPV0101A AIR START SOL VLV 1 5201 M-018 X X KJPV0101B AIR START SOL VLV 2 5201 M-018 X X KJPV0108 SHUTDOWN AIR SOL VLV 5201 M-018 X X KJTCV0034 LUBE OIL COOLER TEMP CTRL VLV 5203 M-018 X X KJTCV0056 INTERCOOLER HEAT EXCHANGER TEMP CTRL VLV 5203 M-018 X X KJTCV0060 JACKET WATER HEAT EXCHANGER TEMP CTRL VLV 5203 M-018 X X KJTCV0061 COOLING WATER SYSTEM D.G. A TEMP CTRL VLV 5203 M-018 X X KJTCV0134 LUBE OIL COOLER TEMP CTRL VLV 5201 M-018 X X KJTCV0156 INTERCOOLER HEAT EXCHANGER TEMP CTRL VLV 5201 M-018 X X KJTCV0160 JACKET WATER HEAT EXCHANGER TEMP CTRL VLV 5201 M-018 X X KJTCV0161 COOLING WATER SYSTEM D.G. B TEMP CTRL
 
VLV 5201 M-018 X X KJTI0020  LUBE OIL COOLER TEMP. INDICATOR 5203 M-018 X X KJTI0021  LUBE OIL COOLER TEMP. INDICATOR 5203 M-018 X X KJTI0049  GEN. OUTBOARD BRG. WATER TEMP. 5203 M-018 X X KJTI0120  LUBE OIL COOLER TEMP. INDICATOR 5201 M-018 X X KJTI0121  LUBE OIL COOLER TEMP. INDICATOR 5201 M-018 X X KJTI0149  GEN. OUTBOARD BRG. WATER TEMP. 5201 M-018 X X KJTS0039  D.G. A L.O. HEATER TEMP SWITCH 5203 M-018 X X KJTS0050  D.G. A JACKET WTR HTR TEMP SWITCH 5203 M-018 X X KJTS0139  D.G. B L.O. HTR. TEMP. SWITCH 5201 M-018 X X KJTS0150  D.G. B JACKET WTR HTR TEMP SWITCH 5201 M-018 X X KJTSH0033  D.G. A L.O. SYST. TEMP. SWITCH 5203 M-018 X X KJTSH0055  D.G. A COOLING WTR TEMP. SWITCH 5203 M-018 X X KJTSH0059A HIGH JACKET WATER TEMP 5203 M-018 X X KJTSH0059B HIGH JACKET WATER TEMP 5203 M-018 X X KJTSH0059C HIGH JACKET WATER TEMP 5203 M-018 X X KJTSH0059D HIGH JACKET WATER TEMP 5203 M-018 X X KJTSH0133  D.G. B L.O. SYST. TEMP. SWITCH 5201 M-018 X X KJTSH0155  D.G. B COOLING WTR TEMP SWITCH 5201 M-018 X X KJTSH0159A HIGH JACKET WATER TEMP 5201 M-018 X X KJTSH0159B HIGH JACKET WATER TEMP 5201 M-018 X X KJTSH0159C HIGH JACKET WATER TEMP 5201 M-018 X X KJTSH0159D HIGH JACKET WATER TEMP 5201 M-018 X X KJTSL0030  D.G. A L.O. SUCT. STRNR TEMP SWITCH 5203 M-018 X X KJTSL0053  D.G. A COOLING WTR TEMP. SWITCH 5203 M-018 X X KJTSL0063  D.G. A COOLING WTR SYS. TEMP. SWITCH 5203 M-018 X X KJTSL0130  D.G. B L.O. STRNR TEMP SWITCH 5201 M-018 X X KJTSL0153  D.G. B COOLING WTR TEMP. SWITCH 5201 M-018 X X KJTSL0163  D.G. B COOLING WTR TEMP. SWITCH 5201 M-018 X X KJTW0035A  D.G. A L.O. SUPPLY THERMOWELL 5203 M-018 X X KJTW0035B  D.G. A L.O. SYST. THERMOWELL 5203 M-018 X X KJTW0054A  D.G. A COOLING WTR THERMOWELL 5203 M-018 X X KJTW0054B  D.G. A COOLING WTR THERMOWELL 5203 M-018 X X KJTW0060A  D.G. A COOLING WTR SYS. THERMOWELL 5203 M-018 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 32)
Component ID Component name Room NoSpec No Hot SD Cold SD KJTW0060B  D.G. A COOLING WTR SYS. THERMOWELL 5203 M-018 X X KJTW0135A  D.G. B L.O. SYST. THERMOWELL 5201 M-018 X X KJTW0135B  D.G. B L.O. SYST. THERMOWELL 5201 M-018 X X KJTW0154A  D.G. B COOLING WTR THERMOWELL 5201 M-018 X X KJTW0154B  D.G. B COOLING WTR THERMOWELL 5201 M-018 X X KJTW0160A  D.G. B COOLING WTR THERMOWELL 5201 M-018 X X KJTW0160B  D.G. B COOLING WTR THERMOWELL 5201 M-018 X X KJV0711A D.G. A STARTING AIR/INTAKE/EXHAUST (CHECK) 5203 M-018 X X KJV0711B D.G. B STARTING AIR/INTAKE/EXHAUST (CHECK) 5201 M-018 X X KJV0712A D.G. A STARTING AIR/INTAKE/EXHAUST (CHECK) 5203 M-018 X X KJV0712B D.G. B STARTING AIR/INTAKE/EXHAUST (CHECK) 5201 M-018 X X KJV0716A D.G. A STARTING AIR/INTAKE/EXHAUST (RLF) 5203 M-018 X X KJV0716B D.G. B STARTING AIR/INTAKE/EXHAUST (RLF) 5201 M-018 X X KJV0717A D.G. A STARTING AIR/INTAKE/EXHAUST (RLF) 5203 M-018 X X KJV0717B D.G. B STARTING AIR/INTAKE/EXHAUST (RLF) 5201 M-018 X X KJV0735A  D.G. A STARTING AIR/INTAKE/EXHAUST (CK)5203 M-018 X X KJV0735B  D.G. B STARTING AIR/INTAKE/EXHAUST (CK)5201 M-018 X X KJV0742A  D.G. A STARTING AIR/INTAKE/EXHAUST (CK)5203 M-018 X X KJV0742B  D.G. B STARTING AIR/INTAKE/EXHAUST (CK)5201 M-018 X X KJV0743A  D.G. A STARTING AIR/INTAKE/EXHAUST (CK)5203 M-018 X X KJV0743B  D.G. B STARTING AIR/INTAKE/EXHAUST (CK)5201 M-018 X X KJV0757A  D.G. A STARTER AIR/INTAKE/EXHAUST (CK) 5203 M-018 X X KJV0757B  D.G. B STARTING AIR/INTAKE/EXHAUST (CK)5201 M-018 X X KJV0766A D.G. A STARTING AIR/INTAKE/EXHAUST (RLF) 5203 M-018 X X KJV0766B D.G. B STARTING AIR/INTAKE/EXHAUST (RLF) 5201 M-018 X X KJV0771A  D.G. A COOLING WATER SYSTEM (RLF) 5203 M-018 X X KJV0771B  D.G. B COOLING WATER SYSTEM (RLF) 5201 M-018 X X KJV0773A  D.G. A COOLING WATER SYSTEM (CK) 5203 M-018 X X KJV0773B  D.G. B COOLING WATER SYSTEM (CK) 5201 M-018 X X KJV0813A  D.G. A LUBE OIL SYSTEM (RLF) 5203 M-018 X X KJV0813B  D.G. B LUBE OIL SYSTEM (RLF) 5201 M-018 X X KJV0814A  D.G. A LUBE OIL SYSTEM (RLF) 5203 M-018 X X KJV0814B  D.G. B LUBE OIL SYSTEM (RLF) 5201 M-018 X X KJV0815A  D.G. A LUBE OIL SYSTEM (CK) 5203 M-018 X X KJV0815B  D.G. B LUBE OIL SYSTEM (CK) 5201 M-018 X X KJV0816A  D.G. A LUBE OIL SYSTEM (CK) 5203 M-018 X X KJV0816B  D.G. B LUBE OIL SYSTEM (CK) 5201 M-018 X X KJV0818A  D.G. A LUBE OIL SYSTEM (CK) 5203 M-018 X X KJV0818B  D.G. B LUBE OIL SYSTEM (CK) 5201 M-018 X X KJV0820A  D.G. A LUBE OIL SYSTEM (RLF) 5203 M-018 X X KJV0820B  D.G. B LUBE OIL SYSTEM (RLF) 5201 M-018 X X KJV0824A  D.G. A LUBE OIL SYSTEM (RLF) 5203 M-018 X X KJV0824B  D.G. B LUBE OIL SYSTEM (RLF) 5201 M-018 X X KJV0835A  D.G. A LUBE OIL SYSTEM (RLF) 5203 M-018 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 33)
Component ID Component name Room NoSpec No Hot SD Cold SDKJV0835B  D.G. B LUBE OIL SYSTEM (RLF) 5201 M-018 X X KJV0877A  D.G. A ENG. DR. L.O. PUMP RELIEF (RLF) 5203 M-018 X X KJV0877B  D.G. B ENG. DR. L.O. PUMP RELIEF (RLF) 5201 M-018 X X KJV0890  D.G. A COOLING WATER SYSTEM (MAN) 5203 M-018 X X KJV0891  D.G. A COOLING WATER SYSTEM (MAN) 5203 M-018 X X KJV0892  D.G. B COOLING WATER SYSTEM (MAN) 5201 M-018 X X KJV0893  D.G. B COOLING WATER SYSTEM (MAN) 5201 M-018 X X KJXJ0001A  D.G. A EXHAUST EXPANSION JOINT 5203 M-312 X X KJXJ0001B  D.G. B EXHAUST EXPANSION JOINT 5201 M-312 X X KJXJ0002A  D.G. A EXHAUST EXPANSION JOINT 5203 M-312 X X KJXJ0002B  D.G. B EXHAUST EXPANSION JOINT 5201 M-312 X X KJXJ0003A  D.G. A EXHAUST EXPANSION JOINT 5203 M-312 X X KJXJ0003B  D.G. B EXHAUST EXPANSION JOINT 5201 M-312 X X Deleted      Deleted      KJXJ0005A  D.G. A EXHAUST EXPANSION JOINT 5203 M-312 X X KJXJ0005B  D.G. B EXHAUST EXPANSION JOINT 5201 M-312 X X KJYS0001A  D.G. A START AIR Y STRAINER 5203 M-018 X X KJYS0001B  D.G. A START AIR Y STRAINER 5203 M-018 X X KJYS0001C  D.G. B START AIR Y STRAINER 5201 M-018 X X KJYS0001D  D.G. B START AIR Y STRAINER 5201 M-018 X X KJYS0002A  D.G. A START. AIR Y STRAINER 5203 M-018 X X KJYS0002B  D.G. A START AIR Y STRAINER 5203 M-018 X X KJYS0002C  D.G. B START AIR Y STRAINER 5201 M-018 X X KJYS0002D  D.G. B START AIR Y STRAINER 5201 M-018 X X KJYS0003A  D.G. A START AIR Y STRAINER 5203 M-018 X X KJYS0003B  D.G. A START AIR Y STRAINER 5203 M-018 X X KJYS0003C  D.G. B START AIR Y STRAINER 5201 M-018 X X KJYS0003D  D.G. B START. AIR Y STRAINER 5201 M-018 X X KJYS0004A  D.G. A L.O. SYS Y STRAINER 5203 M-018 X X KJYS0004B  D.G. B L.O. SYS Y STRAINER 5201 M-018 X X KKJ01A  STANDBY DIESEL ENGINE A 5203 M-018 X X KKJ01B  STANDBY DIESEL ENGINE B 5201 M-018 X X LELE0105 DIESEL GEN BLDG SUMP 5203 J-481 X X LELE0106 DIESEL GEN BLDG SUMP 5201 J-481 X X LELI0105 DIESEL GEN BLDG SUMP 3601 J-200 X X LELI0106 DIESEL GEN BLDG SUMP 3601 J-200 X X LELIT0105 DIESEL GEN BLDG SUMP 3605 J-481 X X LELIT0106 DIESEL GEN BLDG SUMP 3605 J-481 X X LFLE0101A RHR PMP ROOM LEAKAGE LEVEL 1109 J-481  X LFLE0101B RHR PMP ROOM LEAKAGE LEVEL 1109 J-481  X LFLE0102A RHR PMP ROOM LEAKAGE LEVEL 1111 J-481  X LFLE0102B RHR PMP ROOM LEAKAGE LEVEL 1111 J-481  X LFLI0101 RHR PMP ROOM SUMP LEVEL 3601 J-110  X LFLI0102 RHR PMP ROOM SUMP LEVEL 3601 J-110  X LFLIT0101 RHR PMP ROOM LEAKAGE LEVEL 3605 J-481  X LFLIT0102 RHR PMP ROOM LEAKAGE LEVEL 3605 J-481  X NB001 4.16 KV SWITCHGEAR 3301 E-009 X X NB002 4.16 KV SWITCHGEAR 3302 E-009 X X NE001 DIESEL GENERATOR 5203 M-018 X X NE002 DIESEL GENERATOR 5201 M-018 X X NE106 CONTROL & RELAY PANEL 5201 M-018 X X NE107 CONTROL & RELAY PANEL 5203 M-018 X X
 
Rev. 27 WOLF CREEK TABLE 7.4-6 (Sheet 34)
Component ID Component name Room NoSpec No Hot SD Cold SDNF039A LOAD SHED/SEQ PANEL 3605 E-092 X X NF039B LOAD SHED/SEQ PANEL 3605 E-092 X X NF039C LOAD SHED/SEQ PANEL 3605 E-092 X X NG001 LOAD CENTER 3301 E-017 X X NG001A MCC 3301 E-018 X X NG001B MCC 1410 E-018 (11) X X NG001T MCC 1410 E-018 (11) X X NG002 LOAD CENTER 3302 E-017 X X NG002A MCC 3302 E-018 X X NG002B MCC 1409 E-018 (11) X X NG002T MCC 1409 E-018 (11) X X NG003 LOAD CENTER 3301 E-017 X X NG003C MCC 1512 E-018 X X NG003D MCC 5203 E-018 X X NG003T MCC 1410 E-018 (11) X X NG004 LOAD CENTER 3302 E-017 X X NG004C MCC 1501 E-018 X X NG004D MCC 5201 E-018 X X NG004T MCC 1409 E-018 (11) X X NG005E MCC K105 E-018 X X NG006E MCC K104 E-018 X X NK001 125 VDC SWITCHBOARD 3408 E-020 X X NK002 125 VDC SWITCHBOARD 3410 E-020 X X NK003 125 VDC SWITCHBOARD 3414 E-020 X X NK004 125 VDC SWITCHBOARD 3404 E-020 X X NK011 BATTERY 3407 E-050A X X NK012 BATTERY 3411 E-050A X X NK013 BATTERY 3413 E-050A X X NK014 BATTERY 3405 E-050A X X NK021 CHARGER 3408 E-051 X X NK022 CHARGER 3410 E-051 X X NK023 CHARGER 3414 E-051 X X NK024 CHARGER 3404 E-051 X X NK041 125 VDC SWITCHBOARD 3408 E-020 X X NK042 125 VDC SWITCHBOARD 3410 E-020 X X NK043 125 VDC SWITCHBOARD 3414 E-020 X X NK044 125 VDC SWITCHBOARD 3404 E-020 X X NK051 125 VDC SWITCHBOARD 3408 E-020 X X NK051A LIGHTING PANEL 3408 E-020 X X NK054 125 VDC SWITCHBOARD 3404 E-020 X X NK071 Transfer Switch 3408 E-051B X X NK072 Transfer Switch 3410 E-051B X X NK073 Transfer Switch 3414 E-051B X X NK074 Transfer Switch 3404 E-051B X X NN001 VITAL AC INST DIST PANEL 3408 E-053 X X NN002 VITAL AC INST DIST PANEL 3410 E-053 X X NN003 VITAL AC INST DIST PANEL 3414 E-053 X X NN004 VITAL AC INST DIST PANEL 3404 E-053 X X NK079 SWING INVERTER DC TRANSFER SWITCH 3301 E-051C X X NK080 SWING INVERTER DC TRANSFER SWITCH 3302 E-051C X X 
 
Rev. 29 WOLF CREEK TABLE 7.4-6 (Sheet 35)
Component ID Component name Room NoSpec No Hot SD Cold SD NN011 INVERTER 3408 M-766A X X NN012 INVERTER 3410 M-766A X X NN013 INVERTER 3414 M-766A X X NN014 INVERTER 3404 M-766A X X PA003 PT CUB. FOR RCP MOTOR DPBB01A 1410 E-009 X X PA004 PT CUB. FOR RCP MOTOR DPBB01B 1410 E-009 X X PA005 PT CUB. FOR RCP MOTOR DPBB01C 1409 E-009 X X PA006 PT CUB. FOR RCP MOTOR DPBB01D 1409 E-009 X X PAL01A  MOTOR DRIVEN AUX FEED PUMP A 1326 M-021 X  PAL01B  MOTOR DRIVEN AUX FEED PUMP B 1325 M-021 X  PAL02  TURBINE DRIVEN AUX FD PUMP 1331 M-021 X  PBG02A  BORIC ACID TRANSFER PUMP A 1117 M-721 X X PBG02B  BORIC ACID TRANSFER PUMP B 1116 M-098 X X PBG05A  CENTRIFUGAL CHARGING PUMP A 1114 M-721-1 X X PBG05B  CENTRIFUGAL CHARGING PUMP B 1107 M-721-1 X X PEF01A  ESW PUMP A K105 M-089 X X PEF01B  ESW PUMP B K104 M-089 X X PEG01A  CCW PUMP A TRAIN A 1406 M-082 X X PEG01B  CCW PUMP B TRAIN B 1401 M-082 X X PEG01C  CCW PUMP C TRAIN A 1406 M-082 X X PEG01D  CCW PUMP D TRAIN B 1401 M-082 X X PEJ01A  RHR PUMP A 1111 M-721-2  X PEJ01B  RHR PUMP B 1109 M-721-2  X PFC04  TURBINE L.O. PUMP 1331 M-021 X X PJE01A  EMERGENCY FUEL OIL TRANSFER PUMP (SUB IN TANK) (50) M-087 X X PJE01B  EMERGENCY FUEL OIL TRANSFER PUMP (SUB IN TANK) (50) M-087 X X PKJ01A  MOT DR. JACKET WATER KEEP WARM PUMP A 5203 M-018 X X PKJ01B  MOT DR. JACKET WATER KEEP WARM PUMP B 5201 M-018 X X PKJ02A  MOT DR. ROCKER PRELUBE PUMP A 5203 M-018 X X PKJ02B  MOT DR. ROCKER PRELUBE PUMP B 5201 M-018 X X PKJ03A  MOT DR. AUXILIARY LUBE OIL KEEP WARM PUMP A 5203 M-018 X X PKJ03B  MOT DR. AUXILIARY LUBE OIL KEEP WARM PUMP B 5201 M-018 X X PKJ04A  ENGINE DRIVEN FUEL OIL PUMP A 5203 M-018 X X PKJ04B  ENGINE DRIVEN FUEL OIL PUMP B 5201 M-018 X X PKJ05A  ENGINE DRIVEN INTERCOOLER PUMP A 5203 M-018 X X PKJ05B  ENGINE DRIVEN INTERCOOLER PUMP B 5201 M-018 X X PKJ06A  ENGINE DRIVEN JACKET WATER PUMP A 5203 M-018 X X PKJ06B  ENGINE DRIVEN JACKET WATER PUMP B 5201 M-018 X X PKJ07A  ENGINE DRIVEN LUBE OIL PUMP A 5203 M-018 X X PKJ07B  ENGINE DRIVEN LUBE OIL PUMP B 5201 M-018 X X PKJ08A  ENGINE DRIVEN ROCKER LUBE PUMP A 5203 M-018 X X PKJ08B  ENGINE DRIVEN ROCKER LUBE PUMP B 5201 M-018 X X PKJ09A  EJECTOR A 5203 M-018 X X PKJ09B  EJECTOR B 5201 M-018 X X RBB01  REACTOR VESSEL 2000 M-706 X X RBB02  RV INTERNALS 2000 M-703 X X RBB03  CRDM ASSEMBLIES 2000 M-709 X X RL001 MAIN CTRL BOARD 3601 J-200 X X RL002 MAIN CTRL BOARD 3601 J-200 X X RL003 MAIN CTRL BOARD 3601 J-200 X X NN015 SWING INVERTER 3301 M-766A X X NN016 SWING INVERTER 3302 M-766A X X              Rev. 29 WOLF CREEK TABLE 7.4-6 (Sheet 36)
Component ID Component name Room NoSpec No Hot SD Cold SDRL004 MAIN CTRL BOARD 3601 J-200 X X RL005 MAIN CTRL BOARD 3601 J-200 X X RL006 MAIN CTRL BOARD 3601 J-200 X X RL011 MAIN CTRL BOARD 3601 J-200 X X RL012 MAIN CTRL BOARD 3601 J-200 X X RL013 MAIN CTRL BOARD 3601 J-200 X X RL014 MAIN CTRL BOARD 3601 J-200 X X RL015 MAIN CTRL BOARD 3601 J-200 X X RL016 MAIN CTRL BOARD 3601 J-200 X X RL017 MAIN CTRL BOARD 3601 J-200 X X RL018 MAIN CTRL BOARD 3601 J-200 X X RL019 MAIN CTRL BOARD 3601 J-200 X X RL020 MAIN CTRL BOARD 3601 J-200 X X RL021 MAIN CTRL BOARD 3601 J-200 X X RL022 MAIN CTRL BOARD 3601 J-200 X X RL023 MAIN CTRL BOARD 3601 J-200 X X RL024 MAIN CTRL BOARD 3601 J-200 X X RL025 MAIN CTRL BOARD 3601 J-200 X X RL026 MAIN CTRL BOARD 3601 J-200 X X RL027 MAIN CTRL BOARD 3601 J-200 X X RL028 MAIN CTRL BOARD 3601 J-200 X X RP053AA BOP INSTR RACK 3605 J-110 X X RP053AB BOP INSTR RACK 3605 J-110 X X RP053AC BOP INSTR RACK 3605 J-110 X X RP053BA BOP INSTR RACK 3605 J-110 X X RP053BB BOP INSTR RACK 3605 J-110 X X RP053BC BOP INSTR RACK 3605 J-110 X X RP053DA BOP INSTR RACK TERMN 3605 J-110 X X RP053DB BOB INSTR RACK TERMN 3605 J-110 X X RP118A AUX SHUTDOWN PNL 1413 J-201 X X RP118B AUX SHUTDOWN PNL 1413 J-201 X X RP139 AUX RELAY RACK 3301 E-093 X X RP140 AUX RELAY RACK 3302 E-093 X X RP209 AUX RELAY RACK 1320 E-093 X X RP210 AUX RELAY RACK 1402 E-093 X X RP266 AUX RELAY RACK 1408 E-093 X X RP330 REVERSE ISOL RELAY RACK 1320 E-093 X X RP331 REVERSE ISOL RELAY RACK 1408 E-093 X X RP332 AUX. RELAY RACK 1320 E-093 X X RP333 AUX. RELAY RACK 1408 E-093 X X RP334 LOCKOUT RELAY RACK 3302 E-093 X X RP335 LOCKOUT RELAY RACK 3302 E-093 X X SA036A ESFAS CABINET 3605 J-104 X X SA036B ESFAS CABINET 3605 J-104 X X SA036C ESFAS CABINET 3605 J-104 X X SA036D ESFAS CABINET 3605 J-104 X X SA036E ESFAS CABINET 3605 J-104 X X SA075A MN STM & FW ISO ACT PNL 3605 J-105 X X SA075B MN STM & FW ISO ACT PNL 3605 J-105 X X SB029A CAB W SS PROT SYS INPUT TRN A 3605 ESE-16 X X SB029B CAB W SS PROT SYS LOGIC TRN A 3605 ESE-16 X X SB029C CAB W SS PROT SYS OUT 1 TRN A 3605 ESE-16 X X SB029D CAB W SS PROT SYS OUT 2 TRN A 3605 ESE-16 X X SB030A CAB W SAFEGUARDS TEST 1 TRN A 3605 ESE-16 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 37)
Component ID Component name Room NoSpec No Hot SD Cold SDSB030B CAB W SAFEGUARDS TEST 2 TRN A 3605 ESE-16 X X SB032A CAB W SS PROT SYS INPUT TRN B 3605 ESE-16 X X SB032B CAB W SS PROT SYS LOGIC TRN B 3605 ESE-16 X X SB032C CAB W SS PROT SYS OUT 1 TRN B 3605 ESE-16 X X SB032D CAB W SS PROT SYS OUT 2 TRN B 3605 ESE-16 X X SB033A CAB W SAFEGUARDS TEST 1 TRN B 3605 ESE-16 X X SB033B CAB W SAFEGUARDS TEST 1 TRN B 3605 ESE-16 X X SB037 CAB W PROCESS PROTECTION SET 3 3605 ESE-13 X X SB038 CAB W PROCESS PROTECTION SET 1 3605 ESE-13 X X SB041 CAB W PROCESS PROTECTION SET 4 3605 ESE-13 X X SB042 CAB W PROCESS PROTECTION SET 2 3605 ESE-13 X X SB102A CAB W REACTOR TRIP SWGR TRAIN A 1403 ESE-62A X X SB102A CAB W REACTOR TRIP SWGR TRAIN A 1403 ESE-20 X X SB102B CAB W REACTOR TRIP SWGR TRAIN B 1403 ESE-62A X X SB102B CAB W REACTOR TRIP SWGR TRAIN B 1403 ESE-20 X X SB148A CAB W CONTR RM ISOL 3302 ESE-13 X X SB148B CAB W CONTR RM ISOL 3302 ESE-13 X X SBUQ0761A SSPS POWER SUPPLY IN SB038 3605 ESE-13 X X SBUQ0761B SSPS POWER SUPPLY IN SB038 3605 ESE-13 X X SBUQ0761C SSPS POWER SUPPLY IN SB038 3605 ESE-13 X X SBUQ0762A SSPS POWER SUPPLY IN SB042 3605 ESE-13 X X SBUQ0762B SSPS POWER SUPPLY IN SB042 3605 ESE-13 X X SBUQ0762C SSPS POWER SUPPLY IN SB042 3605 ESE-13 X X SBUQ0763A SSPS POWER SUPPLY IN SB037 3605 ESE-13 X X SBUQ0763B SSPS POWER SUPPLY IN SB037 3605 ESE-13 X X SBUQ0763C SSPS POWER SUPPLY IN SB037 3605 ESE-13 X X SBUQ0764A SSPS POWER SUPPLY IN SB041 3605 ESE-13 X X SBUQ0764B SSPS POWER SUPPLY IN SB041 3605 ESE-13 X X SBUQ0764C SSPS POWER SUPPLY IN SB041 3605 ESE-13 X X SE001 WELL DET. (NE41) 2000 ESE-8 X X SE003 WELL DET. (NE43) 2000 ESE-8 X X SE005 WELL DET. (NE42) 2000 ESE-8 X X SE007 WELL DET. (NE44) 2000 ESE-8 X X SE054A W NUC INST. NIS 1 3605 ESE-10 X X SE054B W NUC INST. NIS 2 3605 ESE-10 X X SE054C W NUC INST. NIS 3 3605 ESE-10 X X SE054D W NUC INST. NIS 4 3605 ESE-10 X X SGF02A  AUXILIARY FEED PUMP ROOM COOLER 1326 M-612 X X SGF02B  AUXILIARY FEED PUMP ROOM COOLER 1325 M-612 X X SGK04A CONTROL ROOM A/C UNIT 1512 M-622.1A X X SGK04B CONTROL ROOM A/C UNIT 1501 M-622.1A X X SGK05A CLASS IE ELEC. EQUIP. A/C UNIT 3416 M-622.1A X X SGK05B CLASS IE ELEC. EQUIP. A/C UNIT 3415 M-622.1A X X SGL10A  RHR PUMP ROOM COOLER 1111 M-612  X SGL10B  RHR PUMP ROOM COOLER 1109 M-612  X SGL11A  CCW PUMP ROOM COOLER 1406 M-612 X X SGL11B  CCW PUMP ROOM COOLER 1401 M-612 X X SGL12A  CENT. CHARGING PUMP ROOM COOLER 1114 M-612 X X SGL12B  CENT. CHARGING PUMP ROOM COOLER 1107 M-612 X X SGL15A  PENETRATION ROOM COOLER 1410 M-612 X X SGL15B  PENETRATION ROOM COOLER 1409 M-612 X X SGN01A  CONTAINMENT COOLER 2000 M-620 X X SGN01B  CONTAINMENT COOLER 2000 M-620 X X SGN01C  CONTAINMENT COOLER 2000 M-620 X X
 
Rev. 19 WOLF CREEK TABLE 7.4-6 (Sheet 38)
Component ID Component name Room NoSpec No Hot SD Cold SDSGN01D  CONTAINMENT COOLER 2000 M-620 X X SKJ01A  INTAKE SILENCER A 5203 M-018 X X SKJ01B  INTAKE SILENCER B 5203 M-018 X X SKJ01C  INTAKE SILENCER C 5201 M-018 X X SKJ01D  INTAKE SILENCER D 5201 M-018 X X SKJ02A  EXHAUST SILENCER A 5203 M-018 X X SKJ02B  EXHAUST SILENCER B 5201 M-018 X X TBB03  PRESSURIZER 2000 M-713 X X TBG03A  BORIC ACID TANK A 1117 M-105B X X TBG03B  BORIC ACID TANK B 1116 M-105B X X TBG05  VOLUME CONTROL TANK 1318 M-723-2 X X TBN01  RWST SITE M-109 X  TEG01A  CCW SURGE TANK A 1503 M-105A X X TEG01B  CCW SURGE TANK B 1502 M-105A X X TEJ01A  RHR ISOLATION A 1204 M-109A  X TEJ01B  RHR ISOLATION B 1203 M-109A  X TEM01  BORON INJECTION TANK (BIT) 1126 M-723-2 X X TJE01A  EMERGENCY FUEL OIL STORAGE TANK (50) M-109 X X TJE01B  EMERGENCY FUEL OIL STORAGE TANK (50) M-109 X X TJE02A  EMER FUEL OIL DAY TANK 5203 M-105A X X TJE02B  EMER FUEL OIL DAY TANK 5201 M-105A X X TKA02  ACCUMULATOR 1304 M-105B X X TKA03  ACCUMULATOR 1305 M-105B X X TKA04  ACCUMULATOR 1305 M-105B X X TKA05  ACCUMULATOR 1304 M-105B X X TKJ01A  JACKET WTR EXPANSION TK A 5203 M-018 X X TKJ01B  JACKET WTR EXPANSION TK B 5201 M-018 X X TKJ02A  STARTING AIR TANK A 5203 M-018 X X TKJ02B  STARTING AIR TANK B 5203 M-018 X X TKJ02C  STARTING AIR TANK C 5201 M-018 X X TKJ02D  STARTING AIR TANK D 5201 M-018 X X TKJ04A  LUBE OIL AUX TANK A 5203 M-018 X X TKJ04B  LUBE OIL AUX TANK B 5201 M-018 X X TKJ05A  ROCKER RESERVOIR TANK A 5203 M-018 X X TKJ05B  ROCKER RESERVOIR TANK B 5201 M-018 X X TKJ07A  D.G. A FUEL RACK SUPPLY AIR TANK 5203 M-018 X X TKJ07B  D.G. B FUEL RACK SUPPLY AIR TANK 5201 M-018 X X TKJ09A  D.G. A LUBE OIL LEVEL CONTROL TANK 5203 M-018 X X TKJ09B  D.G. B LUBE OIL LEVEL CONTROL TANK 5201 M-018 X X XNG01 LC TRANSFORMER 3301 E-017 X X XNG02 LC TRANSFORMER 3302 E-017 X X XNG03 LC TRANSFORMER 3301 E-017 X X XNG04 LC TRANSFORMER 3302 E-017 X X XNG05 ESW MCC TRANSFORMER K105 E-075 X X XNG06 ESW MCC TRANSFORMER K104 E-075 X X DELETED      DELETED    ZFE-1 FLOW ELEMENTS (2) J-435 X X ZFE-2 FLOW ELEMENTS (2) M-771 X X ZFO FLOW ORIFICES (2) M-143A X X ZHS SWITCHES 1410 E-028A (11) X X ZHS SWITCHES 1409 E-028A (11) X X 
 
Rev. 29 WOLF CREEK TABLE 7.4-6 (Sheet 39)
Component ID Component name Room NoSpec No Hot SD Cold SDZNE268 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE268 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE269 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE269 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE277 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE277 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE278 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE278 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE279 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE279 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE287 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE287 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE288 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE288 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE295 ELECTRICAL PENETRATION MODULES 1410 E-035B X X ZNE295 ELECTRICAL PENETRATION ASSY 1410 E-035 X X ZNE296 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1204 E-035B X X ZNE296 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1204 E-035 X X ZNE297 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1203 E-035B X X ZNE297 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1203 E-035 X X ZNE298 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1204 E-035B X X ZNE298 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1204 E-035 X X ZNI268 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI268 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI269 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI269 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI277 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI277 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI278 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI278 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI279 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI279 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI287 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI287 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI288 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI288 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI295 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZNI295 ELECTRICAL PENETRATION MODULES 2000 E-035B X X ZNI296 VALVE TERMINAL BOX 1204 E-028  X ZNI296 ELECTRICAL PENETRATION MODULES 1204 E-035B  X ZNI297 VALVE TERMINAL BOX 1203 E-028  X ZNI297 ELECTRICAL PENETRATION MODULES 1203 E-035B  X ZSE215 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1204 E-035 X X
 
Rev. 17 WOLF CREEK TABLE 7.4-6 (Sheet 40)
Component ID Component name Room NoSpec No Hot SD Cold SDZSE215 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1204 E-035B X X ZSE216 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1203 E-035 X X ZSE216 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1203 E-035B X X ZSE217 ELECTRICAL PENETRATION ASSY (ENCAPSULATION) 1203 E-035 X X ZSE217 ELECTRICAL PENETRATION MODULES (ENCAPSULATION) 1203 E-035B X X ZSE218 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE218 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE219 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE219 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE233 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE233 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE234 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE234 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE243 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE243 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE249 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE249 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE250 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE250 ELECTRICAL PENETRATION MODULES 1409 E-035B X X ZSE258 ELECTRICAL PENETRATION ASSY 1409 E-035 X X ZSE258 ELECTRICAL PENETRATION MODULE 1409 E-035B X X ZSI215 VALVE TERMINAL BOX 1204 E-028  X ZSI215 ELECTRICAL PENETRATION MODULE 1204 E-035B  X ZSI216 VALVE TERMINAL BOX 1203 E-028  X ZSI216 ELECTRICAL PENETRATION MODULE 1203 E-035B  X ZSI218 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI218 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI219 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI219 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI233 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI233 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI234 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI234 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI243 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI243 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI249 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI249 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI250 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI250 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZSI258 ELECTRICAL PENETRATION ASSY 2000 E-035 X X ZSI258 ELECTRICAL PENETRATION MODULE 2000 E-035B X X ZZB 5 KV POWER CABLES (5) E-029 X X ZZC1 600 VOLT COPPER CONTROL CABLE (2) E-057 X X ZZC2 600 VOLT COPPER CONTROL CABLE (2) E-057A X X ZZC3 600 VOLT COPPER CONTROL CABLE (2) E-057B X X ZZC4 600 VOLT FIRE-RESISTIVE CONTROL AND POWER CABLE (5) E-057C X X ZZG 600 VOLT POWER CABLE (2) E-058 X X ZZJ 600 VOLT SHIELDED INSTRUMENTATION CABLE(2) E-062 X X ZZJ1 600 VOLT SHIELDED INSTRUMENTATION CABLE(2) E-062A X X ZZP PREFABRICATED CABLE ASSEMBLIES (2) E-095 X X Rev. 24 WOLF CREEK TABLE 7.4-6 (Sheet 41)
Component ID Component name Room NoSpec No Hot SD Cold SDZZR CABLE BREAKOUT KIT (2) (1) X X ZZS 600 VOLT SHIELDED INSTRUMENTATION CABLE(2) E-062 X X ZZT THERMOCOUPLE EXTENSION CABLE (2) E-061 X X ZZU 5 KV CABLE SPLICE MATERIAL (5) E-029 X X ZZV CABLE END SEAL KIT (2) (1) X X ZZW NUCLEAR MOTOR CONNECTION KITS (2) (1) X X ZZX COAXIAL & TRIAXIAL CABLE (2) E-060 X X ZZY 600 V CABLE TERMINATION MATERIAL (2) (1) X X ZZY HEAT SHRINK FLD. SPLICING SYSTEM (2) (1) X X ZZZ STUB CONNECTION KIT (2) (1) X X ZZZ TERMINAL LUGS (2) (1) X X ZZZ TRANSITION SPLICE KIT (2) (1) X X
 
Rev. 17 
 
WOLF CREEK 7.5  SAFETY-RELATED DISPLAY INSTRUMENTATION The information necessary to monitor the nuclear steam supply systems, the containment systems, and the balance of plant is displayed on the operator's console and the various control boards located within the control room. These indications include the information to control and operate the unit through all operating conditions, including anticipated operational occurrences and accident and post-accident conditions. Hot standby information is also displayed on the auxiliary shutdown control panel located outside the control room (refer to Section 7.4). This section is limited to the discussion of those display instruments which provide information to enable the operator to
 
assess reactor status, the onset and severity of accident conditions, and engineered safety feature system (ESFS) status and performance, or to enable the operator to intelligently perform vital manual actions such as safe
 
shutdown and initiation of manual ESFAS actuations. Reactivity control is
 
monitored by sampling of the reactor coolant for boron.
The surveillance instrumentation, which includes indicators, annunciators, recorders, and lights, consists of specific instrumentation for the following functions:
: a. Reactor trip
: b. Engineered safety features
: c. Safe shutdown This section discusses instrumentation that is required for safety as well as instrumentation that is only indirectly related to safety. The safety-related display instrumentation provided in the control room is listed in Table 7.5-4
 
and 7.5-5.
This section also furnishes a summary of important display instrumentation provided to monitor system status and performance. The bypassed status
 
indication is treated separately to establish a clear definition of the system
 
of bypass indication. The display instrumentation defined for bypass, status, and performance monitoring is not safety related (refer to Table 7.1-2, Sheet
: 2) since failure in no way degrades the operation of safety systems and poses
 
no threat to public health and safety.
Refer to Section 1.7 for drawings associated with auxiliary shutdown panel, safety-related display instrumentation, and main control board layouts and
 
ESFAS logic diagrams.
7.5-1                        Rev. 14 WOLF CREEK 7.5.1  REACTOR TRIP SYSTEM Display instrumentation for the reactor trip system actuation is provided by the nuclear steam system supplier and is discussed in Sections 7.2 and 7.7 and Tables 7.5-1 and 7.5-2.
7.5.2  ENGINEERED SAFETY FEATURE SYSTEM
 
Display instrumentation is provided to monitor actuation parameters, bypasses, status, and performance of the ESFSs.
7.5.2.1  System Actuation Parameters 7.5.2.1.1  Description
 
The ESFS actuation parameter display instrumentation comprises those display instrument channels which will provide for informed operator action during and
 
following an accident. The displays provide the information necessary to
 
enable the operator to determine the nature and predict the course of an
 
accident occurrence. They also allow the operator to monitor the effects of an
 
accident through key variables which reflect whether the plant is responding
 
properly to safety measures (and, consequently, whether the ESFS is functioning adequately). The information provided by the displays enables the operator to
 
estimate the magnitude of an impending threat or to determine the potential for
 
radioactive release, to manually initiate the ESFS in the unlikely event of
 
ESFS actuation equipment malfunctions or unanticipated post-accident
 
conditions, and to allow early indication of necessary actions to take to
 
protect the public.
Each parameter monitored for ESFS actuation is displayed in the main control room for operator information. Parameters associated with automatic actuation as well as those required to enable the operator to initiate manual ESFS actuation are displayed. Redundant analog instrument channels, consisting of transmitters, alarm units, and indicators, provide the required information.
Automatic actuation of the ESFS is provided by the engineered safety feature actuation system (ESFAS) described in Section 7.3. The indicators provided for the actuating parameters display the same analog signals monitored by the
 
ESFAS. One indicator is provided for each channel of each parameter.
Table 7.5-1 is a tabulation of the type of readout provided, the number of channels, and the range, accuracy, and location for display instrumentation provided to monitor the ESFS actuation parameters.
7.5-2                        Rev. 0 WOLF CREEK The accuracy and ranges are sufficient to monitor the full range of accident conditions. Predicted accident transients will result in less than full-scale
 
readings on safety-related display indicators.
Display instrumentation provided for the ESFS actuation parameters are the same as those used to monitor these parameters during normal operation.
Redundant indicators displaying the same parameter are located close enough to each other to enable visual comparison. Comparisons between duplicate information channels or between functionally related channels will enable the
 
operator to readily identify a malfunction.
The ESFS actuation parameter displays are visually discernible from other displays on the panels so that they are readily located in the event of an accident. Color-coded nameplates identify all safety-related display
 
instrumentation. Wire and cable are color-coded to differentiate between
 
redundant channels and are physically separated within the plant.
7.5.2.1.2  Analysis
 
DESIGN CRITERIA - The ESFS actuation parameter instruments are designed to remain available in the event of a single failure. Redundant indicator channels are powered from redundant Class 1E 120-V vital instrument ac power supplies (Section 8.3.1.1.5). Display instrumentation is capable of operating
 
independent of offsite power. The indication channels are designed in
 
accordance with Sections 4.2, 4.4, 4.6, and 4.10 of IEEE Standard 279-1971, except that recorders are required to be operable following, but not necessarily during, an SSE. Recorders provided by the nuclear steam system supplier are designed to withstand an SSE, but verification of operability is necessary since signal cable and power supplies are not considered vital.
Temporary modifications may be required to regain operating status. Wiring
 
associated with the ESFS actuation displays is physically separated in accordance with the requirements of Regulatory Guide 1.75 (refer to Appendix 3A). A detailed comparison of the WCGS design to the recommendations of
 
Regulatory Guide 1.97 are contained in Appendix 7A.
Refer to Table 7.1-2 for applicable guides and standards for this equipment.
 
ADEQUACY - The ESFS actuation parameter displays provide sufficient information to enable the operator to assess accident conditions and to perform the
 
necessary operation of manual ESFS actuations.
7.5-3                        Rev. 1 WOLF CREEK Each of the ESFAS parameters is displayed, providing the operator with information on those parameters indicative of accident conditions.
The information supplied by the ESFS actuation parameter displays enables the operator to perform manual actuation. Containment sump level indication and refueling water storage tank level indication provide assurance that adequate net positive suction head (NPSH) exists (Chapter 6.0). Control room ventilation monitors provide the operator with the necessary information on
 
which to base his decision for operation of control room ventilation isolation and filtration.
Containment pressure and air temperature instrumentation provides information for the operator to monitor containment conditions, assess the effectiveness of safety measures in operation, and determine if manual action is necessary.
Containment post-accident radiation monitors provide information concerning the
 
radioactive content of the containment atmosphere. Containment hydrogen
 
concentration indication provides information to judge the significance of a
 
metal-water reaction and furnishes the information necessary for manual
 
hydrogen control through the use of the combustible gas control systems.
The recorders provided for the variables furnish trend information, such as the containment pressure and temperature transients, to help predict the course of
 
an accident. In addition, the recorders provide a historical record for post-
 
accident review.
7.5.2.2  System Bypasses 7.5.2.2.1  Description Bypasses within the ESFAS are indicated on the main control boards or ESFAS cabinets by lights and are alarmed by the plant computer. Bypass of
 
containment airborne gaseous radiation actuation or of containment purge
 
isolation for periodic testing and maintenance and the bypass of low reactor coolant pressure actuation of the safety injection system for startup and shutdown are examples of such bypasses. Bypass is accomplished in the ESFAS cabinets by turning a key associated with a particular actuation bistable.
This causes a light to indicate that a bistable within that actuation channel
 
is bypassed. In the latter example, backlighted switches accomplish the bypass
 
function from the main control boards. Refer to Section 7.3 for identification of the bypass functions and their use.
7.5-4                        Rev. 0 WOLF CREEK Bypass of ESFAS equipment operation can be effected a number of ways.
Handswitch in pull-to-lock position, loss of control power, breaker in test or
 
not in operating position, and closure of manual valves for system or device testing or maintenance are some of the means by which an ESFS or vital supporting system might be rendered inoperative on a system level. The
 
following describes the system of bypass indication and annunciation provided.
The number of bypass features or devices provided for operational purposes or routine testing is minimized by design, but wherever such features or devices
 
are an integral part of the design and are used more frequently than once a
 
year and the bypass results in defeating system functions, a means of
 
indication is provided on the engineered safety feature status panel (ESFSP).
Each piece of ESFS equipment (pump, valve, fan, etc., including vital support system equipment) or small group of equipment (subsystem) which must operate
 
upon automatic or manual ESFS actuation is monitored by a status light
 
indicating availability of that component or group of components.
 
Unavailability is indicated by an amber indicating light. Thus, a bypass of a
 
component by operation of a control switch or by "racking out" a breaker which
 
results in a bypass of system function is indicated by a distinctly colored
 
light. A lighted lamp indicates improper status for ESFS operation.
The status lights for actuated ESFS equipment are arranged in groups in a central location on the main control boards, in accordance with the ESFS and
 
the train in that system. In addition to the individual component indication, annunciation is provided on a system-level basis for each ESFS train. A bypass
 
of one or more components within a system train actuates a corresponding
 
audible alarm to annunciate the fact that a train of equipment is inoperable.
Automatic system level indication of bypass and inoperable status, called for by Regulatory Guide 1.47, applies only to automatically initiated systems, including those systems which directly support the automatically initiated
 
systems but which themselves may not be automatically initiated because they
 
are normally in the operating mode.
Rendering equipment inoperable through the use of features provided strictly for infrequent maintenance (once a year or less often) is not specifically and
 
automatically indicated. Such maintenance features include manual valves
 
provided for isolation of equipment for repair, electrical cable connections, or other manual disconnects. However, manual initiation of safety features
 
equipment bypass indication on a system-level basis is provided in 7.5-5                        Rev. 1 WOLF CREEK the status display panel. Under administrative control, manual bypass indication can be set up or removed. The automatic indication feature cannot
 
be removed by operator action.
7.5.2.2.2  Analysis DESIGN CRITERIA - The system of bypass indication is designed to satisfy the requirements of IEEE Standard 279-1971 (Paragraph 4.13), Branch Technical
 
Position ICSB 21, and Regulatory Guide 1.47. Refer to Table 7.5-3 for a comparison with Regulatory Guide 1.47 recommendations. The intent of IEEE
 
Standard 308-1971, Surveillance Requirements, is satisfied to the extent of indicating control circuit power availability for ESFS equipment. Other indications responsive to IEEE Standard 308-1971 are described in Chapter 8.0.
The system of indicating lights for bypasses of ESFS actuation channels or sensor channels is located in the ESFAS cabinets and is designed to the
 
requirements of IEEE Standard 279-1971. The indicating lights and associated
 
wiring are located in the cabinets corresponding to the channel indicated and
 
are powered by the power source associated with the cabinet. The ESFAS and
 
associated bypass indication system are designed as seismic Category I
 
equipment, and also are designed to withstand all postulated environmental
 
conditions, as stated in Tables 3.11(B)-2 and 3.11(B)-3.
ADEQUACY - The system of status lights for bypass indication, together with other display information available to the operator, and periodic testing provide assurance that the operator is constantly aware of the status of the ESFS. The automatic indication system described previously assures that bypass
 
of control circuits or manual process valves, which could affect system
 
performance, is immediately made obvious.
The bypass indication system is used to supplement administrative procedures by providing indications of safety system availability or status. Administrative procedures do not require operator action based solely on the bypass
 
indicators.
The design of the bypass indication system allows testing during normal plant operation. Both indicating and annunciating functions can be verified.
Process indicators are provided for ESFS actuation parameters (Section 7.5.2.1.1) so that, for parameters that vary in value during plant operation, closure of a manual valve in the transmitter sensing line results in a discrepant indication and response when compared with the corresponding
 
indicators for the 7.5-6                        Rev. 0 WOLF CREEK redundant channels of the same parameter. The process indicators thus provide indication of impulse line blockage or bypass, which obviates the need for
 
position indication for the manual instrument valves.
For ESFS actuation parameters which do not vary during operation, sufficient redundancy is provided so that more than one manual instrument valve would have to be placed in the wrong position before system level actuation could be blocked.Diversity in actuating parameters and the capability for manual system actuation make it even more improbable that ESFS function can be blocked by
 
improper instrument valve position. For the preceding reasons, instrument valves are not included in the status light displays.
On items that do not affect the ESFS function, no indication system is provided for manual valve position or circuit bypass features.
 
Operation of manual valves, use of manual disconnects, or other operations occurring once a year or less frequently, which could impair ESFS performance, are controlled by administrative procedures. Thus, the probability for system blocks or bypasses existing undisclosed between periodic functional tests is
 
minimal.7.5.2.3  System Status 7.5.2.3.1  Description
 
The information important in evaluating the readiness of the ESFS prior to operation and the status of active components during system operation is displayed for the operator in the main control room. The display information consists of process indicators, indicating lights, alarms, and recorders. The
 
display is sufficient but supplemented by the plant computer outputs.
Table 7.5-1 lists the display information provided, together with the type of readout, number of channels, and their range, accuracy, and location.
Indicators are provided for levels, pressures, and temperatures important to safety feature operation. Each of the indicators is driven by an electronic instrument loop consisting of a transmitter, power supply, and any necessary
 
signal conditioners. Where an alarm is provided, the instrument loop includes an alarm unit providing a contact output to the plant annunciator. Many of the analog signals are monitored by the plant computer to enable 7.5-7                        Rev. 1 WOLF CREEK display or logging of status or alarm information. Recorders are provided in lieu of, or in addition to, the indicators where a trend or a time history of the process variable is desired.
Indicating lights are provided to monitor equipment status. In addition to the system level availability and bypass indicating lights described in Section 7.5.2.2, indicating lights are provided at each control switch for equipment.
Each motor-driven component (pump, fan, etc.) has ON and OFF indicating lights, each remotely controlled open-closed service valve or damper has corresponding
 
OPEN-CLOSED light indication, and each breaker control switch has its
 
associated open-closed indicating light. A red light is used to indicate an
 
operating status; for example, motor running, valve fully open, or breaker closed. The green light indicates that the equipment is not in an operating state; for example, motor off, valve fully closed, or breaker open. Amber
 
lights, where provided, signify equipment bypassed, locked out, or not in automatic readiness. The indicating lights for a given control circuit are
 
operated from the control circuit power. Thus, loss of control circuit power
 
would be accompanied by a loss of indicating lights for that device.
7.5.2.3.2  Analysis DESIGN CRITERIA - Status light switches and wiring are designed to the same standards as the associated control circuits. The analog process instruments
 
for status information which are not required for safety system operation do
 
not require special design requirements and are, therefore, of standard
 
commercial quality.
ADEQUACY - Sufficient instrumentation is provided to furnish the plant operator the necessary information and the ESFS status to enable accurate assessment of
 
the readiness of the ESFS prior to operation and the status of active
 
components during operation. The ESFS instrumentation is arranged by system on
 
the main control board to provide the plant operator with a logical arrangement
 
of information to facilitate his evaluation of the ESFS status.
Each power-operated component in the ESFS is equipped with instrumentation to provide equipment status information. Auxiliary contacts from the motor
 
starters or breakers provide motor status indication, while position
 
transmitters and position switches provide valve position indication.
Process variables important for evaluating system readiness are displayed.
Pressures and levels providing information on the ESFS  7.5-8 Rev. 21 WOLF CREEK status regarding adequate tank inventories and accumulator pressures are monitored via pressure and level transmitters and indicators.
Resistance temperature detectors and thermocouples are utilized to monitor temperatures of tanks subject to a freezing environment or tanks containing boric acid solutions to preclude undisclosed freezing or crystallization and loss of availability.
7.5.2.4  System Performance 7.5.2.4.1  Description
 
Display information important in evaluating the performance of an ESFS during periodic test, continuous normal operation, or post-accident operation is provided on the main control boards. Sufficient process indicators, alarms, and recorders are provided to enable the operator to determine whether a system
 
is performing normally or if there is some unanticipated failure within a system. The plant computer monitors selected instrument channels to supplement the display information.
Table 7.5-1 lists the display information provided for the ESFS performance, together with the type of readout, number of channels, and their range, accuracy, and location.
7.5.2.4.2  Analysis
 
DESIGN CRITERIA - The instrumentation is arranged by system on the main control board to facilitate the operator's evaluation of the system performance. The
 
performance monitoring instrumentation is not required for the operation of the
 
safety systems and does not warrant special design and is, therefore, of standard commercial quality.
ADEQUACY - Sufficient instrumentation is provided to furnish the operator with the information to assess operating ESFS performance.
Sufficient process indicators, alarms, and recorders are provided to enable the operator to determine whether a system is performing normally or if there is
 
some unanticipated failure within a system.
For fluid systems, discharge pressure indication is provided for each pump, and flow indication is provided for each system. Together, the flow and pressure
 
enable the operator to verify proper pump performance and verify fluid delivery
 
performance.
7.5-9                        Rev. 0 WOLF CREEK Temperature indication is provided for each system heat exchanger inlet and outlet. The operator has the information, together with the system flow, to
 
verify proper cooling performance.
Temperature indication is also provided for each ventilation system incorporating charcoal filtration, to verify proper temperature range for expected filter performance.
Hydrogen recombiner outlet temperature provides a measure of recombiner performance.
7.5.3  SAFE SHUTDOWN The important display information provided for operator use during post accident safe shutdown operations is briefly described, analyzed, and tabulated in this section. Further discussion of the functional adequacy and use of the hot and cold shutdown control instrumentation is provided in Section 7.4.
7.5.3.1  Hot Standby Control 7.5.3.1.1  Description
 
The hot standby control display instruments are required for manual operations to safely maintain the plant in a hot standby condition.
Table 7.5-2 lists the display information provided for hot standby control, together with the type of readout, number of channels, and their range, accuracy, and location.
These instruments are provided on the main control board in the main control room and on the auxiliary shutdown control panel outside of the main control
 
room. Two or more separate and redundant channels of display information are
 
provided for each required process variable.
7.5.3.1.2  Analysis
 
DESIGN CRITERIA - Since the hot standby information display systems are designed to protection systems standards, the display parameters remain available in the event of a single failure. Redundant indication channels are powered by redundant, 120-V vital instrument ac power supplies (Section
 
8.3.1.1.5). The indication channels are designed in accordance with the portions of IEEE Standard 279-1971 applicable to indication channels.
Refer to Table 7.1-2 for applicable guides and standards for this equipment.      7.5-10    Rev. 19 WOLF CREEK ADEQUACY - Compliance with the design criteria ensures the availability of the display instruments to present the information required to maintain the plant
 
in a hot standby condition.
Two channels of level and pressure are indicated on the main control board for
 
each steam generator, which enable the operator to control auxiliary feedwater
 
to the steam generator and to regulate atmospheric relief. Two channels of primary system pressure and pressurizer level are provided which enable the operator to control the pressurizer heaters and coolant inventory.
Similar provisions are made on the auxiliary shutdown control panel where two channels of pressure and level are displayed for each steam generator and two channels of primary system pressure and level are indicated.
7.5.3.2  Cold Shutdown Control 7.5.3.2.1  Description
 
The display instruments required to bring the plant to a cold shutdown condition are provided in the main control room. For cold shutdown from outside of the control room, see Section 7.4.
Table 7.5-2 lists the display information provided for cold shutdown control, together with the type of readout, number of channels, and their range, accuracy, and location.
7.5.3.2.2  Analysis
 
DESIGN CRITERIA - Refer to Section 7.4.
 
ADEQUACY - Refer to Section 7.4.
 
7.5.3.3  System Bypasses 7.5.3.3.1  Description
 
No bypass indicating light system is provided specifically for the shutdown systems. Certain components used for shutdown have bypass/availability indicating lights provided, if these items also have an ESFS function, but no shutdown system-level indication is provided. Those shutdown components and systems having bypass/availability indicating lights are the auxiliary feedwater system, auxiliary feedwater pump suction valves (essential service water), centrifugal charging pumps, essential      7.5-11    Rev. 14 WOLF CREEK service water pumps, component cooling water pumps, reactor building fan coolers, emergency diesel generators, and the control room ventilation system.
7.5.3.3.2  Analysis
 
The bypass indications on safe shutdown equipment are included in Table 7.5-2.
The analysis provided for the design criteria and adequacy of the ESFS bypass indications in Section 7.5.2.2.2 is applicable to safe shutdown equipment
 
bypasses.7.5.3.4  System Status 7.5.3.4.1  Description
 
Information important in evaluating the readiness of the safe shutdown systems prior to operation and the status of components during system operation is
 
displayed in the main control room. The display information consists of
 
process indicators, indicating lights, alarms, and recorders. In addition to those indicating lights provided in the control room, each control switch on
 
the auxiliary shutdown control panel is provided with associated indicating
 
lights. The plant computer may also be used to supplement the other displays
 
for additional process variables or equipment status.
The description of the equipment provided for ESFS status display information (Section 7.5.2) also applies to the safe shutdown status displays.
7.5.3.4.2  Analysis The safe shutdown system status displays are listed in Table 7.5-2. The analysis provided for the design criteria and adequacy of the ESFS status
 
displays in Section 7.5.2.3.2 is applicable.
7.5.3.5  System Performance 7.5.3.5.1  Description The display information important in evaluating the performance of safe shutdown systems during system operation and periodic tests is listed in Table 7.5-2. Indicators, alarms, and recorders are provided to enable the operator to determine whether the system is performing normally or if there is some failure within the system.      7.5-12    Rev. 1 WOLF CREEK 7.5.3.5.2  Analysis The analysis provided for the design criteria and adequacy of the ESFS performance displays is applicable to the safe shutdown systems performance displays.      7.5-13    Rev. 0 WOLF CREEK TABLE 7.5-1 ENGINEERED SAFETY FEATURES - DISPLAYS LEGEND S Type of Readout/Display Readout/Display LocationI - Linear scale indicator or log scale indicator CB - Control board (main)
* S pared in place refer to R - Recorder ++
S C - System cabinets in control room  section 6.4.6. L - Indicator light LP - Local panel A - Control room annunciator or computer alarm
 
C - Display on demand via plant computer
# - S afety related Number of Channels Channel
 
Indicated Accuracy  Type of % of Full Readout/Display
 
Displayed Parameter Readout/Display Available Required Range  S cale      Locations Engineered S afety Feature S ystem Actuation Reactor coolant system pressure  I #, R, C                3          1          0-3,000 psig 4          CB, LP Containment pressure              I #, R                  4          1          0-69 psig 4          CB Containment pressure (wide        I #, R, C                2          1          180 psig 4          CB range)
S team generator pressure          I #, R, C            3 per loop  1 per loop    0-1,300 psig 14*        CB, LP (steam line)
Reactor coolant system            I #, R, C                2          1          0-700 F 4*          CB wide range temperature (hot)
Reactor coolant system            I #, R, C                2          1          0-700 F 4*          CB, LP wide range temperature (cold)
 
Refueling water storage tank      I #, R, C                4          1          0-100 % 4*          CB level Boric acid tank level            I #, R              2 per tank  1 per tank    0-100 % 4*          CB S team generator water level      I #, R, C            4 per loop  1 per loop    0-100 % 35*        CB, LP (3 narrow, 1 wide range)
Control room air intake -        I #, A, R, C            2          1          10
-7 to 10-2 25 S C    gaseous radioactivity Ci/cc Control room air intake -        I #, A                  2          1          0 to 5 ppm 25          CB chlorine content*
Containment gasesous radio-      I #, A, R, C            2          1          10
-7 to 10-2 25 S C    activity Ci/cc                                                                                                                    Rev. 21 WOLF CREEK TABLE 7.5-1 (S heet 2)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Display
 
Displayed Parameter Readout/Display Available Required Range S cale          Locations Containment hydrogen          I#, R#, A, C                2          1          0-10 percent 5 S C, CB Containment Normal sump        I#, A, C                    2          1          1995' 6" to 4            CB level                                                                        2008' 6" Containment Normal S ump        R#                          1          1          0 - 100% 4            CB level Containment purge gaseous      I#, R, A, C                2          1          10
-7 to 10-2 25 S C    radioactivity Ci/cc Containment spray additive    I#, A                      2          1          0 - 100 % 4            CB tank level Fuel building gaseous          I#, R, A, C                2          1          10
-7 to 10-2 25 S C    radioactivity Ci/cc Containment air temperature    I#, R#, C                  4          1          0 - 400 F 4            CB Containment post accident      I#, R#, C                  2          1          10 2 to 10 6 R/hr 25          CB radiation Control bldg sump level        I#, A, C                    2          1          6 - 72" 4            CB Diesel bldg sump level        I#, A, C                    2          1          42- 72" 4            CB RHR pump room sump level      I#, A, C                    2          1          6 - 138" 4            CB Auxiliary bldg sump level      I#, A, C                    2          1          6 - 138" 4            CB Engineered S afety Feature, S ystem Bypasses Trip bistable bypass          L, A                  1 per train                On for bypass    -
S C, CB Actuation system signal        L                      1 per equip train          On for bypass    -
S C    bypass Equipment bypass              L, A                  1 per equipment            On for bypass    -            CB
 
Equipment S afety Feature S ystems S tatus Containment spray additive    C                          1                    0 - 10 psig 4            CB tank pressure
 
Control valve status**        L                      1 per valve                Open - closed    -            CB RW S T temperature              I                          2                    0 - 200 F 4            CB Auxiliary gas supply pressure  A                      1 per system              Low alarm        -            LP Equipment status              L                      1 per motor                On - Off          -            CB Rev. 16 WOLF CREEK TABLE 7.5-1 (S heet 3)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Display
 
Displayed Parameter Readout/Display Available Required Range S cale          Locations S tation 4.16-kV and 480 Volt      L                    One/power channel        Current status    -            CB load center electrical dis-
 
tribution Diesel day tank level            I, A                1 per tank                0 - 100 %        +
4            CB, LP Diesel starting air              L, A                1 per diesel              Low alarm          -            CB, LP accumulator pressure Ultimate heat sink level          C                        1 S ite dependent    +
4            CB Ultimate heat sink temperature    I                        2                    0 - 200 F        +
4            CB Boron injection tank temper-      I                        2                    50 - 200 F        +
1.5          LP ature Boron injection tank pressure    I                        1                    0 - 2,800 psig    +
1.5          CB Accumulator pressure              I, A                2 each tank              0 - 700 psig      +
1.5          CB Accumulator water                I, A                2 each tank              0 - 100 %        +
1            CB level Containment pressure              I                        1                    (-)85 to (+) 85  +
4            CB inches of water
 
Engineered S afety Feature S ystem Performance Containment spray pump dis-      I                    1 per pump                0 - 300 psig      +
4            CB charge pressure Containment spray flow            I                    1 per header              0-2 x 10 6 lb/hr  +
4            CB Essential service water pump      I #                  1 per pump                0 - 300 psig      +
4            CB discharge pressure
 
Essential service water flow      I #                  1 per header              0-10.6 x 10 6      +4            CB Component cooling water          I                    1 per header              0 - 200 F        +
4            CB temperature Rev. 13 WOLF CREEK TABLE 7.5-1 (S heet 4)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Display
 
Displayed Parameter Readout/Display Available Required Range S cale          Locations Hydrogen recombiner heater        I                    1 per unit                0 - 100 Kw        +
4 S C    power Hydrogen recombiner temperature  I                    1 per unit                0 - 2,000 F      +
4 S C Control room filtration          A, C                1 per filter              150 - 400 F      +
4            CB temperature
 
Fuel building exhaust filter      A, C                1 per filter              150 - 400 F      +
4            CB temperature
 
Diesel generator performance      -                  (see Chapter 8.0)          -                -            -
 
Residual heat exchanger          C, R                1 ea. heat exchanger      50 - 400 F        +
1            CB temperature (inlet/outlet)
 
Charging pump inlet/discharge    I                    1 each pump              0 - 150 psig      +
2            LP pressure                                                                      (inlet) 0 - 3,500 psig
 
(disch)
S afety injection pump            I                    1 each pump              0 - 200 psig      +
2            LP suction pressure
 
Residual heat removal pump        I                    1 each pump              0 - 700 psig      +
2            LP suction pressure
 
S afety injection header          I                    1 each header            0 - 2,000 psig    +
1            CB pressure
 
Residual heat removal pump        I, A                1 each pump              0 - 700 psig      +
1            CB discharge pressure
 
Charging pump injection          I, A                    1                    40 - 200 gpm      +
1            CB flow S afety injection pump            I                    1 each pump              0 - 800 gpm      +
1            CB header flow
 
Residual heat removal pump        I                    1 each pump              0 - 4,500 gpm    +
2            CB hot leg injection flow
 
Residual heat removal pump        I                    1 each pump              400 - 1,500 gpm  +
1.5          LP minimum flow
 
++S afety-related recorders are not required to function during an earthquake, but must function with the required accuracy without operator action as soon as the seismic excitation is removed.
 
*Channel accuracy in % of span.
 
**S ee S ection 6.3.5.5 for accumulator isolation valve position indication.                                      Rev. 0 WOLF CREEK TABLE 7.5-2 POST ACCIDENT SAFE SHUTDOWN DISPLAY INFORMATION LEGEND S Type of Readout/Display Readout/Display Location I - Linear scale indicator or log scale indicator        CB - Control board (main)
R - Recorder                                            AP - Auxiliary shutdown control panel
 
L - Indicator light S C - S ystem cabinets in control room A - Control room annunciator or computer alarm
 
C - Display on demand via plant computer
# - S afety related Number of Channels                      Channel
 
Indicated Accuracy Type of                                                      % of Full        Readout/Displa y Displayed Parameter Readout/Display Available Required Range S cale          Locations Hot S tandby Control S team generator water level      I#, R                2 per loop  1 per loop    0-100%            +
35*          CB (narrow range)                                                                (+7 to (-)5 ft.  (hot)
 
from nominal
 
full load level)
                                ++I#                  2 per loop  1 per loop    0 - 100 %        +
35*          AP
(+7 to (-)5 ft.  (hot) from nominal
 
full load level)
S team generator pressure          I#, R                2 per loop  1 per loop    0 - 1,300 psig    +
14*          CB (steam line)                **I#                  2 per loop  1 per loop    0 - 1,300 psig &  +
14*          AP 0 - 1,500 psig    +
4 Pressurizer water level          I#, R                    2          1          0 - 100 %        +
35*          CB I#                      2          1          0 - 100 %        +
35*          AP Reactor coolant system            I#, R                2 per loop  1 per loop    0 - 3,000 psi    +
4.3*        CB wide range pressure            I#                  2 per loop  1 per loop    0 - 3,000 psi    +
4.3*        AP (pressurizer)
Rev.
21 WOLF CREEK TABLE 7.5-2 (S heet 2)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Displa y Displayed Parameter Readout/Display Available Required Range S cale          Locations Auxiliary feedwater pump          I#, A                    3          1          0 - 100 psia      +
4            CB suction pressure              I#                      3          1          0 - 100 psia      +
4            AP Condensate storage tank          I#                      3          1          0 - 50 psia      +
4            CB supply pressure Cold S hutdown Control Those listed above for hot standby and the following:
S ource range nuclear              I, R                    2                    1 to 10 6 counts/  +
7            CB, S C    instrumentation                                                              second I                        2                    1 to 10 6 counts/  +
7            AP second Intermediate range nuclear        I, R                    2                    8 decades        +
7            CB, S C    instrumentation                I                        2                    8 decades        +
7            AP Hot S tandby S ystem Bypasses S ee S ection 7.5.3.3 Hot S tandby S ystem S tatus Condensate storage tank level    I, A                    1                    0 - 100 %        +
4            CB I                        1                    0 - 100 %        +
4            AP Condensate storage tank          C                        1                    0 - 200 F        +
4            CB temperature
 
Control valve status              L                    1 per valve assoc        Open-closed        -            CB, AP with system channel Auxiliary gas supply pressure    A                    1 per system              Low alarm          -            CB
 
Equipment status                  L                    1 per motor assoc        On-Off            -            CB, AP with system channel Centrifugal charging pump        A                    1 per room                High alarm        -            CB room temperature Rev.
14 WOLF CREEK TABLE 7.5-2 (S heet 3)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Displa y Displayed Parameter Readout/Display Available Required Range S cale          Locations Component cooling water          A                    1 per room              High alarm          -            CB
 
pump room temperature Motor Driven Auxiliary feedwater  A                    1 per room              High/low alarm      -            CB pump room temperature Fuel storage pool pump            A                    1 per room              High alarm          -            CB room temperature
 
E S F switchgear                    A                    1 per room              High alarm          -            CB room temperature
 
Electrical penetration            A                    1 per room              High alarm          -            CB room temperature Emergency diesel generator        A, C                1 per room              High alarm          -            CB room temperature Essential service water pump      A, C                1 per room              High alarm          -            CB room temperature Containment temperature          I, R                2 per channel            0 - 400 F          +
4            CB Auxiliary shutdown panel          A                        1                    High alarm          -            CB room temperature Cold S tandby S ystem S tates Those listed above for hot standby and the following:
 
Residual heat removal pump        A                    1 per room              High alarm          -            CB room temperature S afety injection pump room        A                    1 per room              High alarm          -            CB temperature
 
Hot S tandby S ystem Performance Auxiliary feedwater pump          I, A, C              1 per pump              0 - 2,000 psig    +
4            CB discharge pressure            I                    1 per pump              0 - 2,000 psig    +
4            AP Auxiliary feedwater flow          I, C                1 per stm gen            0 - 2 x 10 5 lb/hr  +4            CB I                    1 per stm gen            0 - 2 x 10 5 lb/hr  +4            AP Rev. 14 WOLF CREEK TABLE 7.5-2 (S heet 4)                                                        Number of Channels                      Channel Indicated Accuracy Type of                                                      % of Full        Readout/Displa y Displayed Parameter Readout/Display Available Required Range S cale          Locations Auxiliary feedwater pump          I                        I                    0 - 6,000 rpm      +
4            CB turbine speed                  I                        I                    0 - 6,000 rpm      +
4            AP Reactor coolant temperature
 
(see Note 1)
Loop 1 cold leg                I#, R                1 (Note 1)  Note 1      0 - 700 F          +
4            CB I                    1 (Note 1)  Note 1      0 - 700 F          +
4            AP hot leg                I#, R                1 (Note 1)  Note 1      0 - 700 F          +
4            CB I                    1 (Note 1)  Note 1      0 - 700 F          +
4            AP Loop 2 cold leg                I#, R                1 (Note 1)  Note 1      0 - 700 F          +
4            CB I#                  1 (Note 1)  Note 1      0 - 700 F          +
4            AP hot leg                I#, R                1 (Note 1)  Note 1      0 - 700 F          +
4            CB Loop 3 cold leg                R                    1 (Note 1)  Note 1      0 - 700 F          +
4            CB I                    1 (Note 1)  Note 1      0 - 700 F          +
4            AP hot leg                R                    1 (Note 1)  Note 1      0 - 700 F          +
4            CB Loop 4 cold leg                R                    1 (Note 1)  Note 1      0 - 700 F          +
4            CB I                    1 (Note 1)  Note 1      0 - 700 F          +
4            AP hot leg                R                    1 (Note 1)  Note 1      0 - 700 F          +
4            CB I#                  1 (Note 1)  Note 1      0 - 700 F          +
4            AP Nuclear power source range        I, R                    2                    1 to 10 6 counts/  +
7            CB sec I                        2                    1 to 10 6 counts/  +
7            AP sec Intermediate power range          I,R                      2                    8 decades          +
7            CB I                        2                    8 decades          +
7            AP
*Channel accuracy in % of span.
 
**Each loop utilizes one vertical indicator, range:  0 - 1,300 psig and one indicator integral to the atmospheric steam dump controller which displays steam line pressure, range: 0 - 1,500 psig.
+Accuracy is sufficient to indicate that water level is above pressurizer heaters and below 100% of span.
 
++One narrow range/one wide range per loop.
Rev. 1 WOLF CREEK TABLE 7.5-2 (S heet 5) Note 1: Two of the cold leg indicators on AP are powered from different protection sets, as are the two AP hot leg indicators.
The circuitry for the redundant indicators is isolated and runs in different nonsafety-grade separation groups. No single failure can inhibit the indication at the auxiliary shutdown panel of at least one cold leg temperature associated with a steam generator h aving both an auxiliary feedwater supply and an operable atmospheric relief valve, and at least one hot leg temperature associated wi th a steam generator having both auxiliary feedwater supply and an operable atmospheric relief valve.
Rev. 13 WOLF CREEK TABLE 7.5-3
 
WCGS PLANT DESIGN COMPARISON WITH REGULATORY GUIDE 1.47 DATED MAY 1973, TITLED BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS Regulatory Guide 1.47 Position                              WCGS Design C. Regulatory Position                The WCGS design complies with Regulatory Guide 1.47. Refer to Section 7.5.2.2.1 for
 
The following comprises an acceptable        a description of the bypasssed and inoper-method for implementing the requirements of        able status indication system.
Section 4.13 of IEEE Std 279-1971 and Cri-
 
terion XIV of Appendix B to 10 CFR Part 50
 
with respect to indicating the bypass or in-
 
operable status of portions of the protection system, systems actuated or controlled by the protection system, and auxiliary or suppor-
 
ting systems that must be operable for the
 
protection system and the system it actuates
 
to perform their safety-related functions:
: 1. Administrative procedures should be supplemented by a system that automatically indicates at the system level the bypass
 
or deliberately induced inoperability of the
 
protection system and the systems actuated
 
or controlled by the protection system.
: 2. The indicating system of C.1. above should also be activated automatically by
 
the bypassing or deliberately induced in-
 
operability of any auxiliary or supporting
 
system effectively bypasses or renders in-Rev. 0 WOLF CREEK TABLE 7.5-3 (Sheet 2)
 
Regulatory Guide 1.47 Position                                WCGS Design C. Regulatory Position (Continued) operable the protection system and the systems actuated or controlled by the protection system.
: 3. Automatic indication in accordance
 
with C.1. and C.2. above should be provided
 
in the control room for each bypass or
 
deliberately induced inoperable status that meets all the following conditions:
: a. Renders inoperable any redun-dant portion of the protection system, systems actuated or controlled by the pro-
 
tection system, and auxiliary or supporting
 
systems that must be operable for the pro-
 
tection system and the systems it actuates to perform their safety-related functions;
: b. Is expected to occur more fre-
 
quently than once per year; and
: c. Is expected to occur when the
 
affected system is normally required to
 
be operable.
: 4. Manual capability should exist in the control room to activate each system-
 
level indicator provided in accordance
 
with C.1. above.
Rev. 0 WOLF CREEK TABLE 7.5-4 SAFETY-RELATED DISPLAY INSTRUMENTATION
 
LOCATED ON THE CONTROL BOARD - (NSSS SCOPE OF SUPPLY)
 
Indicator    Notes 1 and 2 Parameter                              Tag No.
PAMS Separation Group I          II WIDE RANGE RCS T HOT LEG LOOP 1        BB-TI 413A      X WIDE RANGE RCS T HOT LEG LOOP 2        BB-TI 423A      X WIDE RANGE RCS T COLD LEG LOOP 1        BB-TI 413B                  X
 
WIDE RANGE RCS T COLD LEG LOOP 2        BB-TI 423B                  X PRESSURIZER WATER LEVEL                BB-LI 459A      X
 
PRESSURIZER WATER LEVEL                BB-LI 460A                  X
 
PRESSURIZER WATER LEVEL                BB-LI 461        X
 
STEAM GEN. LOOP 3 PRESSURE              AB-PI 534A      X
 
STEAM GEN. LOOP 1 PRESSURE              AB-PI 514A      X
 
STEAM GEN. LOOP 2 PRESSURE              AB-PI 524A      X
 
STEAM GEN. LOOP 4 PRESSURE              AB-PI 544A      X STEAM GEN. LOOP 1 PRESSURE              AB-PI 515A                  X
 
STEAM GEN. LOOP 2 PRESSURE              AB-PI 525A                  X
 
STEAM GEN. LOOP 4 PRESSURE              AB-PI 545A                  X
 
STEAM GEN. LOOP 3 PRESSURE              AB-PI 535A                  X
 
STEAM GEN. LOOP 1 PRESSURE              AB-PI 516A                  X STEAM GEN. LOOP 4 PRESSURE              AB-PI 546A                  X STEAM GEN. LOOP 2 PRESSURE              AB-PI 526A      X STEAM GEN. LOOP 3 PRESSURE              AB-PI 536A      X
 
STEAM GEN. LOOP 2 WATER LEVEL N. R. AE-LI 529        X
 
STEAM GEN. LOOP 3 WATER LEVEL N. R. AE-LI 539        X
 
STEAM GEN. LOOP 1 WATER LEVEL N. R. AE-LI 519                  X
 
STEAM GEN. LOOP 4 WATER LEVEL N. R. AE-LI 549                  X STEAM GEN. LOOP 1 WATER LEVEL N. R. AE-LI 518        X STEAM GEN. LOOP 2 WATER LEVEL N. R. AE-LI 528        X STEAM GEN. LOOP 3 WATER LEVEL N. R. AE-LI 538        X STEAM GEN. LOOP 4 WATER LEVEL N. R. AE-LI 548        X STEAM GEN. LOOP 1 WATER LEVEL N. R. AE-LI 517                  X STEAM GEN. LOOP 2 WATER LEVEL N. R. AE-LI 527                  X STEAM GEN. LOOP 3 WATER LEVEL N. R. AE-LI 537                  X STEAM GEN. LOOP 4 WATER LEVEL N. R. AE-LI 547                  X CONTAINMENT PRESSURE N. R.              GN-PI 934                  X
 
CONTAINMENT PRESSURE N. R.              GN-PI 935        X
 
CONTAINMENT PRESSURE N. R.              GN-PI 936                  X CONTAINMENT PRESSURE N. R.              GN-PI 937        X STEAM GEN. LOOP 1 W. R. WATER LEVEL    AE-LI 501        X
 
STEAM GEN. LOOP 2 W. R. WATER LEVEL    AE-LI 502                  X
 
STEAM GEN. LOOP 3 W. R. WATER LEVEL    AE-LI 503        X STEAM GEN. LOOP 4 W. R. WATER LEVEL    AE-LI 504                  X Rev. 0 WOLF CREEK TABLE 7.5-4 (Sheet 2)
Indicator    Notes l and 2 Parameter                              Tag No.
PAMS Separation Group I          II R. C. S. W. R. PRESSURE                BB-PI 405        X R. C. S. W. R. PRESSURE                BB-PI 403                  X BORIC ACID WATER LEVEL                  BG-LI 102        X
 
R. W. S. T. WATER LEVEL                BN-LI 930        X
 
R. W. S. T. WATER LEVEL                BN-LI 931                  X R. W. S. T. WATER LEVEL                BN-LI 932        X R. W. S. T. WATER LEVEL                BN-LI 933                  X CENTRIFUGAL CHARGING PUMP FLOW          EM-FI 917A      X CENTRIFUGAL CHARGING PUMP FLOW          EM-FI 917B                  X CONTAINMENT PRESSURE W. R.              GN-PI 938        X CONTAINMENT PRESSURE W. R.              GN-PI 939                  X
 
R. C. S. EXCESS LETDOWN HEAT            BG-TI 137A      X
 
EXCHANGER FLOW TO PRT TEMP
 
R. C. S. EXCESS LETDOWN HEAT            BG-TI 137B                  X
 
EXCHANGER FLOW TO PRT TEMP
 
R. C. S. EXCESS LETDOWN HEAT            BG-FI 138A      X EXCHANGER FLOW TO PRT R. C. S. EXCESS LETDOWN HEAT            BG-FI 138B                  X
 
EXCHANGER FLOW TO PRT
 
BORIC ACID WATER LEVEL                  BG-LI 104                  X BORIC ACID WATER LEVEL                  BG-LI 105        X BORIC ACID WATER LEVEL                  BG-LI 106                  X
 
VOLUME CONTROL TANK WATER LEVEL        BG-LI-112        X
 
VOLUME CONTROL TANK WATER LEVEL        BG-LI-185                  X RCS W. R. PRESSURE                      BB-PI-406                  X CHG. PUMP TO RCP                BG-FI 215A      X SEAL FLOW CHG. PUMP TO RCP                BG-FI 215B                  X SEAL FLOW
 
SOURCE RANGE NEUTRON FLUX              SE-NI-60A        X
 
POWER RANGE NEUTRON FLUX                SE-NI-60B        X SOURCE & POWER RANGE NEUTRON Flux (3)  SE-NIR-61                  X REACTOR VESSEL WATER LEVEL N. R.        BB-LI-1311      X REACTOR VESSEL PRESSURE DROP W. R.      BB-LI-1312      X
 
REACTOR VESSEL WATER LEVEL N. R.        BB-LI-1321                  X REACTOR VESSEL PRESSURE DROP W. R.      BB-LI-1322                  X RCS TEMPERATURE MARGIN TO SATURATION    BB-TI-1390A      X RCS TEMPERATURE MARGIN TO SATURATION    BB-TI-1390B                X NOTES:
: 1. PAM I routed as Separation Group 1. PAM II routed as Separation Group 4.
: 2. See Westinghouse process control block diagrams for the applicable
 
protection set.
: 3. Instrument on the MCB is a dual pen indicating recorder.
Rev. 10 WOLFCREEKTABLE7.5-5SAFETY-RELATEDDISPLAYINSTRUMENTATIONLOCATEDONTHECONTROLBOARD-(BOPSCOPEOFSUPPLY)IndicatorSeperationParameterTagNo.GroupAUXILIARYFEEDWATER-FLOWTOS.G.DAL-FI-1A4AUXILIARYFEEDWATER-FLOWTOS.G.AAL-FI-2A1 AUXILIARYFEEDWATER-FLOWTOS.G.BAL-FI-3A2AUXILIARYFEEDWATER-FLOWTOS.G.CAL-FI-4A3CONDENSATESTORAGETANK-PRESSUREAL-PI-371 CONDENSATESTORAGETANK-PRESSUREAL-PI-382 CONDENSATESTORAGETANK-PRESSUREAL-PI-394 TURBINEDRIVENAUXILIARYFEEDPUMP-SUCTIONPRESS.AL-PI-26A2MOTORDRIVENAUXILIARYFEED PUMPA-SUCTIONPRESS.AL-PI-25A1 MOTORDRIVENAUXILIARYFEEDPUMP B-SUCTIONPRESS.AL-PI-24A4 CONTROLROOMAIRINTAKED-CHLORINEGK-AI-24(sparedin CONTROLROOMAIRINTAKED-CHLORINEGK-AI-31place)
CONTROLROOMAIRINTAKED-GASEOUSRADIOACTIVITYGK-RIC-4*4CONTROLROOMAIRINTAKE-GASEOUSRADIOACTIVITYGK-RIC-5*1CONTAINMENT-GASEOUSRADIOACTIVITYGT-RIC-31*4CONTAINMENT-GASEOUSRADIOACTIVITYGT-RIC-32*1CONTAINMENT-HYDROGENGS-AI-104CONTAINMENT-HYDROGENGS-AI-191 CONTAINMENTSUMPNORMALLEVELLF-LI-104CONTAINMENTSUMPNORMALLEVELLF-LI-91CONTAINMENTPURGE-GASEOUSRADIO-ACTIVITYGT-RIC-33*4CONTAINMENTPURGE-GASEOUSRADIO-ACTIVITYGT-RIC-22*1CONTAINMENTSPRAYADDITIVETANK-LEVELEN-LI-174CONTAINMENTSPRAYADDITIVETANK-LEVELEN-LI-191 FUELBUILDING-GASEOUSRADIOACTIVITYGG-RIC-28*4FUELBUILDING-GASEOUSRADIOACTIVITYGG-RIC-27*1CONTAINMENT-AIRTEMPERATUREGN-TI-614 CONTAINMENT-AIRTEMPERATUREGN-TI-601 CONTAINMENT-AIRTEMPERATUREGN-TI-634CONTAINMENT-AIRTEMPERATUREGN-TI-621CONTROLBUILDINGSUMP-LEVELLF-LI-1254CONTROLBUILDINGSUMP-LEVELLF-LI-1241 DIESELGENERATORBUILDINGSUMP-LEVELLE-LI-1064 DIESELGENERATORBUILDINGSUMP-LEVELLE-LI-1051
 
________*DigitaldisplayonradiationmonitoringpanelSP-067.Rev.13 WOLFCREEKTABLE7.5-5(Sheet2)IndicatorSeperationParameterTagNo.GroupRHRPUMPROOMSUMP-LEVELLF-LI-1014RHRPUMPROOMSUMP-LEVELLF-LI-1021 AUXILIARYBUILDINGSUMP-LEVELLF-LIO-1044AUXILIARYBUILDINGSUMP-LEVELLF-LIO-1031NK21BATCHARGERAMPSNK-II-11 NK11BATAMPSNK-II-21 NK01125VDCBUSVOLTSNKEI11 NK22BATCHARGERAMPSNK-II-32NK12BATAMPSNK-II-42NK02125VDCBUSVOLTSNK-EI-22 NK23BATCHARGERAMPSNK-II-53 NK13BATAMPSNK-II-63 NK03125VDCBUSVOLTSNK-EI-33 NK24BATCHARGERAMPSNK-II-74 NK14BATAMPSNK-II-84 NK04125VDCBUSVOLTSNK-EI-44RWSTTEMPBN-TI-21RWSTTEMPBN-TI-54CTMTRECIRCSUMPBLEVELEJ-LI-84 CTMTRECIRCSUMPALEVELEJ-LI-71CCWSURGETANKBLEVELEG-LI-24CCWHXBDISCHTEMPEG-TI-324 ESWBPMPDISCHFLOWEF-FI-544 ESWBPMPDISCHPRESS.EF-PI-24 ESWTRAINBTEMPEF-TI-624ESWTRAINATEMPEF-TI-611ESWAPMPDISCHPRESSEF-PI-11 ESWAPMPDISCHFLOWEF-FI-531 CCWHXADISCHTEMPEG-TI-311 CCWSURGETKALEVELEG-LI-11 CCWHXTORCPFLOWEG-FI-1281 CCWHXTORCPFLOWEG-FI-1294 EMERGENCYFUELOILDAYTKALVLJE-LI-12A1EMERGENCYFUELOILDAYTKBLVLJE-LI-32A44.16KVBUSNBO1VOLTSNB-EI-114.16KVBUSNBO2VOLTSNB-EI-244.16KVBUSNBO1SYNCHROSCOPENB-EI-314.16KVBUSNBO2SYNCHROSCOPENB-EI-44Rev.13 WOLF CREEK 7.6  ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 7.6.1  INSTRUMENTATION AND CONTROL POWER SUPPLY SYSTEM
 
The instrumentation and control power supply system is described in Section 8.3.1.1.5.
Safety-related BOP transmitters not powered directly from the system described in 8.3.1.1.5 are powered by input buffers in the BOP analog equipment cabinets.
Each BOP electronic analog input buffer is able to withstand an open circuit, a short circuit, or a single or multiple-point ground on the field wiring, without affecting any other instrument loop in any separation group.
An open circuit would interrupt the field current and drive the buffer output offscale "low."  The field bus power supply voltage is not high enough to cause any damage if it were suddenly unloaded. There would be no consequential
 
damage to the electronics.
A short circuit would apply the full field bus voltage across on-board current-limiting resistors designed and provided to limit such current to a safe value.
 
The buffer output would be driven to the high limit with no consequential
 
damage to the electronics.
A single ground on an input buffer field line would connect one side of the field bus power supply to system ground through an on-board, current-limiting resistor designed and provided to limit the resultant current to a safe value.
The buffer output would take on some arbitrary value, with no consequential
 
damage to the electronics.
A ground on both field lines of an input buffer would result in a condition similar to an input line short circuit. The buffer output would be driven to the high limit, but there would be no consequential damage to the electronics.
7.6.2  RESIDUAL HEAT REMOVAL SYSTEM ISOLATION VALVES 7.6.2.1  Description The residual heat removal system (RHRS) isolation valves are normally closed and are opened only for residual heat removal system operation after system
 
pressure is reduced to approximately 360 psig and system temperature has been reduced to approximately 350&deg;F.
7.6-1                        Rev. 8 WOLF CREEK There are two motor-operated valves in series in each of the two residual heat removal pump suction lines from the reactor coolant system (RCS) hot legs. The
 
two valves nearest the RCS (valves 8702A and 8702B) are designated as the inner isolation valves, while the two valves nearest the residual heat removal pumps (valves 8701A and 8701B) are designated as the outer isolation valves. The
 
interlock features provided for the outer isolation valves, shown on Figure 7.6-1 (Sheet 2), are identical to those provided for the inner isolation valves, shown on Figure 7.6-1 (Sheet 1), except that equipment diversity is employed by virtue of the fact that PT 405 is of a different manufacture than PT 403. Each valve is interlocked so that it cannot be opened unless the RCS pressure is below a preset pressure. This interlock prevents the valve from being opened when the RCS pressure plus the residual heat removal pump pressure is
 
above the RHRS design pressure. An alarm will actuate on the main control
 
board if any one of the four(4) RHRs suction-isolation valves is not fully
 
closed in conjunction with RCS high pressure.
In addition, the valves cannot be opened unless the isolation valves in the following lines are closed:
: a. Recirculation line from the residual heat exchanger outlet to the suction of the high head safety injection pumps.
: b. RHR pump suction line from the refueling water storage tank.
: c. RHR pump suction line from the containment sump.
7.6.2.2  Analysis Based on the scope definitions in IEEE Standards 279-1971 and 338-1971, these criteria do not apply to the residual heat removal isolation valve interlocks.
 
However, because of the possible severity of the consequences of loss of
 
function, the requirements of IEEE Standard 279-1971 have been applied with the following comments.
: a. For the purpose of applying IEEE Standard 279-1971 to this circuit, the following definitions are used:
(1) Protection system The two valves in series in each line and all components of their interlocking and closure circuits.
7.6-2                        Rev. 12 WOLF CREEK (2) Protective action Once RHRS is isolated from the RCS, the isolation between RCS and RHS is maintained by the open permissive interlock when the RCS pressure is above the preset value.
: b. IEEE Standard 279-1971, Section 4.10
 
The above-mentioned pressure interlock signal and logic will be tested on-line from the analog signal through to the train signal which activates the slave relay (the
 
slave relay provides the final output signal to the valve
 
control circuit). Actuation to permit opening the valve
 
is not performed because this could leave only one
 
remaining valve to isolate the low pressure RHRS from the
 
RCS, which would reduce the safety margin.
: c. IEEE Standard 279-1971, Section 4.15
 
This requirement does not apply, as the setpoints are independent of the mode of operation and are not changed.
Environmental qualification of the valves and wiring is discussed in Section 3.11(N).7.6.3  REFUELING INTERLOCKS
 
Electrical interlocks (i.e., limit switches), as discussed in Section 9.1.4, are provided for minimizing the possibility of damage to the fuel during fuel handling operations.
7.6.4  ACCUMULATOR MOTOR-OPERATED VALVES The safety injection system accumulator discharge isolation valves are motor-operated, normally open valves which are controlled from the main control
 
board.These valves are interlocked so that:
: a. They open automatically on receipt of an SIS with the main control board switch in either the "AUTO" or "CLOSE" position.
: b. They open automatically whenever the RCS pressure is above the safety injection unblock pressure (P-11)
 
specified in Technical Specifications only when the main
 
control board switch is in the "AUTO" position.
: c. They cannot be closed as long as an SIS is present.
7.6-3                        Rev. 6 WOLF CREEK The interconnection of the interlock signals to the accumulator isolation valve meets the following criteria:
: a. Automatic opening of the accumulator valves when: (1) the primary coolant system pressure exceeds a preselected value (specified in Technical Specifications) or (2) a safety injection signal has been initiated. Both signals are provided to the valves.
: b. Utilization of a safety injection signal to automatically remove (override) any bypass features that are
 
provided to allow an isolation valve to be closed for short periods of time when the RCS is at pressure (in accordance with the provisions of Technical Specifications). As a result of the confirmatory SIS, isolation of an accumulator with the reactor at pressure
 
is acceptable.
: c. Under certain plant conditions, the motor control center start drawout units are withdrawn, under administrative
 
control, as discussed in Section 6.3.2.
The control circuit for these valves is shown on Figure 7.6-2. The valves and control circuits are further discussed in Sections 6.3.2 and 6.3.5.
The four main control board position switches for these valves provide a "spring return to auto" from the open position and a "maintain position" from
 
the closed position.
The "maintain closed" position is required to provide an administratively controlled manual block of the automatic opening of the valve at pressure above the safety injection unblock pressure (P-11). The manual block or "maintain closed" position is required when performing periodic check valve leakage tests
 
when the reactor is at pressure. The maximum permissible time that an
 
accumulator valve can be closed when the reactor is at pressure is specified in
 
Technical Specifications.
Administrative control is required to ensure that any accumulator valve which has been closed at pressures above the safety injection unblock pressure is returned to the "AUTO" position. Verification that the valve automatically returns to its normal full open position is also required.
During plant shutdown, the accumulator valves are closed. To prevent an inadvertent opening of these valves during that period, the accumulator valve
 
motor circuit breakers are opened or 7.6-4                        Rev. 0 WOLF CREEK withdrawn (see Section 6.3.2). Administrative control is again required to ensure that these valve circuit breakers are closed during the prestartup
 
procedures.
These normally open, motor-operated valves have alarms, indicating a malpositioning (with regard to their emergency core cooling system function during the injection phase). The alarms sound in the main control room.
An alarm will sound for any accumulator isolation valve, under the following conditions, when the RCS pressure is above the "safety injection unblocking
 
pressure."    a. Valve motor-operated limit switch indicates valve not open.
: b. Valve stem-operated limit switch indicates valve not open. The alarm on this switch will repeat itself at
 
given intervals.
Additionally, an ESF status panel bypass indication is provided whenever any of these valves leaves the fully open position.
7.6.5  SWITCHOVER FROM INJECTION TO RECIRCULATION
 
The details of achieving cold leg recirculation following safety injection are given in Section 6.3.2.8 and on Table 6.3-8. Figure 7.6-3 shows the logic which is used to automatically open the sump valves.
7.6.6  INTERLOCKS FOR RCS PRESSURE CONTROL DURING LOW TEMPERATURE OPERATION The basic function of the RCS pressure control during low temperature operation is discussed in Section 5.2.2. As noted in Section 5.2.2, this pressure
 
control includes automatic actuation logic for two pressurizer power-operated
 
relief valves (PORVs). The function of this actuation logic is to continuously
 
monitor RCS temperature and pressure conditions, with the actuation logic
 
unblocked only when plant operation is at a temperature below the reference nil
 
ductility temperature (RNDT). The monitored system temperature signals are
 
processed to generate the reference pressure limit program which is compared to
 
the actual monitored RCS pressure. This comparison provides an actuation signal to an actuation device which causes the PORV to automatically open, if necessary, to prevent pressure conditions from exceeding allowable limits.
 
Refer to Figure 7.6-4 for the block diagram showing the interlocks for RCS
 
pressure control during low temperature operation.
7.6-5                        Rev. 0 WOLF CREEK The generating station pressure and temperature variables required for this interlock are channelized as follows:
: a. Pressure and Temperature Inputs to PCV455A
 
(1)  Four wide range RCS temperature signals derived from channels in a Train A related protection set.
(2)  One wide range RCS pressure signal derived from a channel in a Train A related protection set.
: b. Pressure and Temperature Inputs to PCV456A (1)  Four wide range RCS temperature signals derived from channels in a Train B related protection set.
(2)  One wide range RCS pressure signal derived from a channel in a Train B related protection set.
The wide range RCS temperatures in each protection set are auctioneered in an auctioneering device in each protection set to select the lowest reading.
An alarm is actuated when the auctioneered low temperature from the RCS wide range temperature channels falls within the range of cold overpressure applicability, thereby alerting the operator to arm the RCS cold overpressure mitigation system, which automatically opens the block valve when the block valve control switch is in the automatic position.
The lowest reading is selected and input to a function generator which calculates the reference pressure limit program, considering the plant's allowable pressure and temperature limits. Also available from the related protection set is the wide range RCS pressure signal. The reference pressure from the function generator is compared to the actual RCS pressure monitored by
 
the wide range pressure channel. The error signal derived from the difference
 
between the reference pressure and the actual measured pressure will first
 
annunciate a main control board alarm whenever the actual measured pressure
 
approaches, within a predetermined amount, the reference pressure. On a
 
further increase in measured pressure, the error signal will generate an
 
actuation signal.
Logic is also provided to close the block valve automatically if the relief valve fails or sticks in the open position following some plant transient when
 
the RCS temperature is above the cold over pressurization setpoint, and the RCS
 
pressure drops below the reset pressure for the relief valve.
7.6-6                        Rev. 0 WOLF CREEK The monitored generating station variables that generate the actuation signal for the redundant PORV are processed in a similar manner.
Upon receipt of the actuation signal, the actuation device will automatically cause the PORV to open. Upon sufficient RCS inventory letdown, the operating RCS pressure will decrease, clearing the actuation signal. Removal of this signal causes the PORV to close.
7.6.6.1  Analysis of Interlocks Many criteria presented in IEEE Standards 279-1971 and 338-1971 do not apply to the interlocks for RCS pressure control during low temperature operation, because the interlocks do not perform a protective function but, rather, provide automatic pressure control at low temperatures as back-up to the
 
operator. However, although IEEE Standard 279-1971 criteria do not apply, some
 
advantages of the dependability and benefits of an IEEE Standard 279-1971
 
design have accrued by including selected elements, as noted above, in the
 
protection sets and by organizing the control of the two PORVs (either of which
 
can accomplish the RCS pressure control function) into dual channels.
The design of the low temperature interlocks for RCS pressure control is such that pertinent features include:
: a. No credible failure at the output of the protection set racks, after the output leaves the racks to interface with the interlocks, would prevent the associated protection system channel from performing its protective function because of the separation of Train B interlocks
 
from Train A (see Figure 7.6-4).
: b. Testing capability for elements of the interlocks within (not external to) the protection system is consistent
 
with the testing principles and methods discussed in
 
Section 7.2.2.2.3, item J. It should be noted that there
 
is an annunciator which provides an alarm when the block
 
valve is armed coincident with a closed position of the
 
motor-operated (MOV) pressurizer relief block valve.
 
This MOV is in the same fluid path as the PORV, with a
 
separate MOV and alarm used with the second PORV.
: c. A loss of offsite power does not defeat the provisions for an electrical power source for the interlocks because
 
these provisions are through onsite power, which is
 
described in Section 8.3.
7.6-7                        Rev. 0 WOLF CREEK 7.6.7  ISOLATION OF ESSENTIAL SERVICE WATER (ESW) TO THE AIR COMPRESSORS 7.6.7.1  Description As stated in Section 9.2.1.2.2.1, ESW flow to the nonsafety-related air compressors and associated aftercoolers is maintained following a DBA.
Instrumentation and controls are provided to automatically isolate each train of the ESW to the air compressors on high flow. ESW to the air compressors can also be isolated by remote manual means.
Each control system (one per train of the ESW) utilizes a differential pressure transmitter and bistable which senses flow through the associated isolation
 
valve. On high flow (indicative of gross leakage in the nonseismic portion of the system), the control system automatically closes the isolation valve.
The isolation valve will remain in the closed position until the valve is manually reset by the operator in the control room.
A means of remote manual isolation is provided in the control room. The status of each isolation valve is indicated by open and closed indicating lights in the control room.
The isolation valves are air operated and are designed to fail closed on the loss of air and electrical power.
: a. Initiating circuits Each isolation valve is automatically actuated by flow monitoring instrumentation. The isolation valves can
 
also be closed via control switches in the control room.
: b. Logic
 
The logic diagram for the isolation of the ESW to the air compressors is provided in Section 1.7.
: c. Bypass
 
No bypass is provided.
: d. Interlock
 
No interlock is provided.
: e. Redundancy 7.6-8                        Rev. 0 WOLF CREEK Redundancy is accomplished on a system basis. Each train of the ESW is provided with an independent control system
 
and isolation valve.
: f. Actuated devices The isolation valves are the actuated devices.
: g. Supporting systems
 
The controls for ESW isolation to the air compressors are powered from the Class 1E power system (refer to Chapter 8.0).
: h. Portion of system not required for safety Isolation valve position inputs to the station computer are not required for safety.
: i. Design bases
 
The design bases for ESW isolation to the air compressors are described in Section 9.2.1.2.1 (Safety Design Bases 5 and 6).
Additionally, Section 7.3.1.1.2a. and b. are applicable to the control system components.
7.6.7.2  Analysis
: a. Conformance to NRC regulatory guides
: 1. Regulatory Guide 1.22
 
The isolation system controls can be tested periodically.
: 2. Regulatory Guide 1.29 The isolation system controls are designed to withstand the effects of an earthquake without loss
 
of function. The isolation system controls are
 
classified seismic Category I, in accordance with the guide.
: b. Conformance to IEEE Standard 279-1971 The controls for the isolation system conform to the applicable requirements of IEEE Standard 279-1971. The 7.6-9                        Rev. 1 WOLF CREEK control circuits are designed so that any single failure will not compromise the ESW system's safety function.
 
This is accomplished by redundancy provided in the ESW system. Each isolation system utilizes control power from independent Class 1E power systems. In order to
 
prevent interaction between the redundant systems, the control channels are wired independently and separated with no electrical connections between control channels.
: c. Conformance to other criteria and standards Conformance to other criteria and standards is indicated in Table 7.1-2.
7.6.8  ISOLATION OF THE NONSAFETY-RELATED PORTION OF THE COMPONENT COOLING WATER (CCW) SYSTEM 7.6.8.1  Description The nonsafety-related, nonseismic portion of the CCW system is isolated by two isolation valves in series that are provided in both the supply and return
 
lines (see Figure 9.2-15). These valves automatically close upon low-low surge tank level, SIS, or high flow. The nonseismic portion of the CCW system can
 
also be isolated by remote manual means.
Two independent flow transmitters in the supply line sense flow through the isolation valves. On high flow (indicative of gross leakage in the nonseismic portion of the system), the isolation valves are automatically closed and will remain in the closed position until the valves are manually reset by the operator in the control room. Each flow transmitter and its associated bistable provides isolation signals to one valve in the supply line and one valve in the return line.
Two independent level transmitters (one per surge tank) are provided. On low-low surge tank level, the isolation valves are automatically closed and will
 
remain in the closed position until the valves are manually reset by the
 
operator in the control room. Each level transmitter and its associated bistable provides isolation signals to one valve in the supply line and one valve in the return line.
The isolation valves are air operated and are designed to fail closed on loss of air and electrical power.
A means of remote manual isolation is provided in the control room. The status of each isolation valve is indicated by open and closed indicating lights in
 
the control room.      7.6-10    Rev. 12 WOLF CREEK The SIS to the isolation valves is discussed in Section 7.3 and will not be discussed further in this section.
: a. Initiating circuits
 
Each isolation valve is automatically actuated by flow monitoring and level monitoring instrumentation. The isolation valves can also be closed via control switches
 
in the control room.
: b. Logic
 
The logic diagram for the isolation of the nonseismic portion of the CCW system is provided in Section 1.7.
: c. Bypass
 
No bypass is provided.
: d. Interlock
 
An interlock is provided to defeat the isolation of one set of isolation valves (one in the supply line and one
 
in the return line) on low-low surge tank level. This
 
interlock will allow continued plant operation for a period of time if the corresponding train of the CCW is out of service.
: e. Redundancy
 
Redundancy is accomplished by providing two independent sets of flow instrumentation and two independent sets of level instrumentation.
: f. Diversity
 
Diversity is accomplished by isolation on high flow or low-low surge tank level.
: g. Actuated devices
 
The isolation valves are the actuated devices.
: h. Supporting systems
 
The controls for isolation of the nonseismic portion of the CCW system are powered from two independent Class 1E power systems.      7.6-11    Rev. 1 WOLF CREEK
: i. Portion of system not required for safety Isolation valve position inputs to the station computer are not required for safety.
: j. Design bases
 
The design bases for isolation of the nonsafety-related portion of the CCW system are described in Section 9.2.2.1.1 (Safety Design Bases 5 and 6).
Additionally, Section 3.11(B).2.2 and 3.11(B).2.3 are applicable to the control system components.
7.6.8.2  Analysis
: a. Conformance to NRC regulatory guides
 
(1) Regulatory Guide 1.22
 
The isolation system controls can be tested periodically.
(2) Regulatory Guide 1.29 The isolation system controls are designed to withstand the effects of an earthquake without loss
 
of function. The isolation system controls are
 
classified seismic Category I, in accordance with the
 
guide.
: b. Conformance to IEEE Standard 279-1971 The controls for the isolation system conform to the applicable requirements of IEEE Standard 279-1971. The
 
control circuits are designed so that any single failure
 
will not compromise the CCW system's safety function.
 
This is accomplished by redundant flow and surge tank
 
level instrumentation.
The CCW isolation system, flow instrumentation, and the surge tank level instrumentation utilize power from two independent Class 1E power systems. In order to prevent interaction between the redundant systems, the control channels are wired independently and separated with no electrical connections between control channels.      7.6-12    Rev. 12 WOLF CREEK
: c. Conformance to other criteria and standards Conformance to other criteria and standards is indicated in Table 7.1-2.
7.6.9  FIRE PROTECTION AND DETECTION
 
Fire protection and detection is discussed in Section 9.5.1.
7.6.10  INTERLOCKS FOR PRESSURIZER PRESSURE RELIEF SYSTEM 7.6.10.1  Description of Pressurizer Pressure Relief System The pressurizer pressure relief (PPR) system provides the following:
: a. Capability for RCS overpressure mitigation during cold shutdown, heatup, and cooldown operations to minimize the
 
potential for impairing reactor vessel integrity when
 
operating at or near the vessel ductility limits.
: b. Capability for RCS depressurization following Condition II, III, and IV events.
: c. Interlock that, with the pressurizer PORVs and PORV block valves in auto control, closes the PORV block valves and prevents signals from the pressurizer pressure control system from opening the PORVs when pressurizer pressure is low.
7.6.10.2  Description of Pressurizer Pressure Relief System Interlocks Interlocks for the PPR system control the opening and closing of the pressurizer PORVs and the PORV block valves. These interlocks provide the
 
following functions:
: a. Pressurizer pressure control (refer to Section 7.7.1.5 for a description).
: b. RCS pressure control during low temperature operation (refer to Sections 5.2.2 and 7.6.6 for a description).
: c. RCS pressure control to achieve and maintain a cold shut-down and to heatup, using equipment that is required for safety (refer to Section 7.4 for a description).
The interlock functions that provide pressurizer pressure control are  derived from process parameters as shown on Figure 7.2-1,      7.6-13    Rev. 14 WOLF CREEK Sheet 11 and the interlock logic functions as well as process parameter inputs required for low temperature operation, as shown on Figure 7.6-4. The
 
functions shown on Figure 7.6-4 include those needed for the PORV block valves as well as the pressurizer PORVs to meet both interlock logic and manual operation requirements where manual operation is at the main control board.
7.6.11  SWITCHOVER OF CHARGING PUMP SUCTION TO RWST ON LOW-LOW VCT LEVEL 7.6.11.1  Description The suction of the charging pumps is normally supplied by a line containing two normally open motor-operated valves which connects to the bottom of the volume
 
control tank (VCT). These VCT outlet isolation valves are designated as LCV-
 
112B, which is assigned to the A train, and LCV-112C, which is assigned to the
 
B train. Each VCT outlet isolation valve is controlled by its train associated level channel. Refer to Figure 7.6-5 (Sheet 1 of 2) for the logic diagram. When the
 
control switch is in the normal position, the valve receives a signal to close on a low-low level signal from its associated channel. The valves also receive
 
a signal to close on an SIS signal.
The interlock between the above signal and the emergency makeup signal from its train associated RWST valve position prevents the valve from automatically closing unless its train associated valve from the RWST to the charging pump suction header is open. This system ensures that the charging pumps will
 
always have a source of fluid and protects them against loss of NPSH and
 
cavitation damage.
Each RWST valve is controlled by its train associated level channel. Refer to Figure 7.6-5 (Sheet 2 of 2) for the logic diagram. When the control switch is
 
in the normal position, the valve receives a signal to open on a low-low level signal from its associated channel. The valves also receive a signal to open
 
on an SIS signal.
In order to avoid any interface between control grade instrumentation functions and protection grade instrumentation channels which are derived from level transmitters LT-112 and LT-185, a third VCT level instrumentation channel
 
derived from level transmitter LT-149 is provided. This channel performs all
 
the control grade functions so that LT-112 and LT-185 may be dedicated to switchover of charging pump suction to the RWST on low-low VCT level.      7.6-14    Rev. 0 WOLF CREEK 7.6.11.2  Evaluation of Switchover of Charging Pump Suction In addition to having complete electrical separation from channels LT-112 and LT-185, the upper level tap from LT-149 is on the VCT vent line at the same pressure point as pressure transmitter PT-115. This ensures adequate physical separation of the different grades of equipment. LT-185 and LT-149 share the lower level tap. A postulated rupture of this tap would result in a false "empty" indication by the affected transmitter, which would initiate switchover.
7.6.12  INSTRUMENTATION FOR MITIGATING CONSEQUENCES OF INADVERTENT BORON DILUTION 7.6.12.1  Description Instrumentation is provided to mitigate the consequences of inadvertent addition of unborated, primary grade water into the reactor coolant system.
 
The boron dilution control system is similar to that reviewed and approved by the NRC for Comanche Peak Units 1 and 2 (Docket Nos. 50-445 and 50-446).
In the event of a boron dilution transient (modes 3, 4 and 5), redundant level transmitters in the CVCS Volume Control Tank (VCT) provide a Hi level alarm on the main control board that indicates an unplanned boron dilution. The alarm is set to give operators time to terminate the transient.  (FLUX doubling equipment previously used for automatic mitigation is still installed to provide additional information. Automatic actuation is no longer provided.)
7.6.12.2  Analysis The analysis of effects and consequences of inadvertent boron dilution transients is covered in Section 15.4.6.
7.6.12.3  Qualification Qualification of the flux doubling equipment is discussed in WCAP-8587 Supplement 1, "Equipment Qualification Data Package" ESE-47. VCT Hi level alarms are from bistables in the qualified 7300 process cabinets.
7.6.13  MONITORING OF RCS LEVEL DURING REDUCED INVENTORY (MID-LOOP) OPERATIONS In preparation for Steam Generator or Reactor Coolant Pump maintenance, RCS inventory may be reduced to a point below the top of the RCS Hot-Leg piping.
 
Such operation is known as "mid-loop" operation and the resulting of RCS level
 
is referred to as a mid-loop level. Careful control of RCS inventory is needed
 
during mid-loop operation to prevent decreases in level which could result in
 
interruption of RHR system operation and to prevent level increases which could present a personnel safety hazard. Mid-loop level instrumentation provides
 
reliable indication of RCS level in the control room to allow appropriate
 
operator actions to control RCS inventory during mid-loop operations.      7.6-15    Rev. 10 WOLF CREEK Two independent level sensing loops are provided. Each sensing loop contains a Wide Range (WR) and a Narrow Range (NR) instrument loop. The WR level instruments measure RCS level from just above the bottom of the RCS Hot-Leg to just above the lower Pressurizer tap. The NR level instruments measures RCS level from just above the bottom of the RCS Hot-Leg to approximately two feet above the top of the RCS Hot-Leg piping. Both the WR and NR level are displayed in the control room and alarms are provided for High RCS Level and
 
Low RCS Level. Each instrument loop also provides input to the Plant Computer.
The design of the mid-loop RCS level instrumentation is in accordance with WCNOC commitments to the NRC in response to Generic Letter 88-17 Loss of Residual Heat Removal. The instrumentation, tubing, supports, electrical cable, cable raceway and raceway supports are functionally non-safety related.
The instrument loops are isolated from the RCS during modes one through four by
 
normally closed manual isolation valves. These valves are opened after RCS de-
 
pressurization and prior to RCS drain down in preparation for mid-loop
 
operations. Except for field routed tubing, all aspects have been designed to
 
meet II/I and/or seismic design requirements.
7.6.14  INCORE THERMOCOUPLES The incore thermocouples are chromel-alumel thermocouples which are threaded into guide tubes that penetrate the reactor vessel head through seal assemblies, and terminate at the exit flow end of the fuel assemblies. The thermocouples are provided with two primary pressure boundary seals consisting of a core exit thermocouple nozzle assembly with a lower conoseal with a quick-acting clamp and an upper Grafoil seal in a seal carrier with a split clamp, drive sleeve and drive nut. Thermocouple readings are monitored by the computer, and the information is used for the core subcooling monitor which is classified as safety-related display instrumentation.      7.6-16    Rev. 8 
* * *
* I.U Vl Vl I.U 8: l: S! l: Vl u a: ALARM <RHRS-ISO VLV OPEN> CLOSEST TO RHR SPRING RETURN TO AUTO FRON BOTH SIDES RCS HIGH PRESSURE**
RECIRCULATION LINE ISOLATION VALVE CLOSED RHR N4P/RWST ISOLATION VALVE CLOSED Stw LINE ISOLATION VALVE CLOSED
* ALARM SETPOINT ** PREVENT OPEN SETPOIHT OPEN AUTO CLOSE CLOSE VALVE NOTE: LOGIC FOR VALVES IN E!CH FLUID SYSTEM TRAIN IS IDENTICAL Rev.6 WOL!' CUBit SAFETY ARALYSIS FIGURE 7.6-1 LOGIC DIAGRAM FOR INNER RHRS ISOLATION VALVE (SHEET 1) w 111 111 w 0: a. :I: !2 :I: 111 u 0: ALARM <RHRS-ISO VLV OPEN> CLOSEST TORe$ SPRING R&#xa3;1URN TO AUTO FRQ4 BO'Tlt S I DES RCS HIGH PRESSURe**
RECIRCULATION LINE ISOLATION VALVE CLOSED RHR P\JMP/RWST ISOLATION VALVE CLOSED sti4P LINE ISOlATION VALVE CLOSED
* ALARM SETPOINT *
* PREVENT OPEN SETPO I NT OPEN AUTO. CLOSE OPEN VALVE NOTE: LOGIC FOR VALVES IN EACH FLUID SYSTEM TRAIN IS IDENTICAL CLOSE VALVE Rev.S WOLP CUBit OPDA'l'BD SAJ!'ftl' AIIALl'SIS RBPOlt'l' FIGURE 7.6-1 LOGIC DIAGRAM FOR OUTER RHRS ISOLATION VALVE (SHEET 2) * *
* WOLF CREEK Control Board Switch Maintain Close, Spring Return From Open to Auto .. OPEN AUTO CLOSE Safety Injection System Unblock Pressure Signal {From RCPS)* Safety Injection Signal r--r----;:::==----
Safety Injection Signal AND AND Close ACCUMULATOR ISOLATION VALVE *This interlock Indicates the method of applying automatic opening of the valve, whenever the RCS pressure exceeds a limit. This signal automatically occurs at RCS pressures above the S I unblock pressure used to derive P-11. Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.6-2 FUNCTION BLOCK DIAGRAM OF ACCUMULATOR
!SOLATION VALVE WOLF CREEK RWST LEVEL CHANNEL BISTABLE$
: 1) NORMALLY DE-ENERGIZED . 2} DE-ENERGIZED ON LOSS OF POWER 3) TRIP SIGNAL PROVIDED WHEN ENERGIZED ENERGIZED ON LO-L0-1 SETPOINT A TRIP SIGNAL TO AUTOMATICALLY OPEN SUMP ISOLATION VALVE 8811A (CON'T ON SHEET 2) NOTE: WHEN 8811A IS FULL OPEN, RWST VALVE (TO RHR PUMP) 8812A WILL CLOSE (SEE SHEET 3) PROCESS CONTROL CABINETS SOLID STATE PROTECTION
""--r--CABINETS B TRIP SIGNAL TO AUTOMATICALLY*OPEN SUMP I SOLATION VALVE 88118 (CON'T ON SHEET 2) NOTE: WHEN 88118 IS FULL OPEN, RWST VA'LVE (TO RHR PUMP) 88128 WILL Rev. o CLOSE (SEE SHEET 3) WOLF CREEK UPDATED SAFETY REPORT FIGURE /.6-3 SAFETY INJECTION SYSTEM RECIRCUlATION SUMP AND RHR SUCTION -ISOLATION VALVES (SHEET 1)
RHR OUTER ISO. VALVE CLOSED MANUAL RESET SPRING RETURN OPEN CLOSE MCB RWST/RHR SUCTION I SO. VALVE CLOSED 8811A 88118 TRAIN A 88128 LMT SW#I B OPEN VALVE I r r"Ri P siGNAL! FROM 2/4
* I RWST LO-L0-11 I LEVEL s IGHALI I {CON'T FROM I CLOSE VALVE APPLICABLE VALV DESCRIPTION SUMP TO RHR PUMP A 8811 A SUMP TO RHR PUMP B 88118 LIMIT SWITCH #I IS THE NORMAL POSITION SIGNAL AND IS USED FOR POSITION SIGNALS BETWEEN VALVES ASSIGNED TO THE SAME TRAIN. LIMIT SWITCH #2 IS THE STEM MOUNTED POSITION SWITCH AND IT IS USED FOR POSITION SIGNALS BETWEEN VALVES ASSIGNED TO OPPOSITE TRAINS. Rev.14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.6-3 SAFETY INJECTION SYSTEM RECIRCULATION SUMP AND RHR SUCTION ISOLATION VALVES (SHEET2)
NCB OPEN SUt.F SUCTION ISOLATic.4 VALVE CLOSED OPEN VALVE CLOSE CLOSE VALVE INTERLOCK TABLE 8812A 88128 SUMP I SOL. VAL. .. I 8 ,.1 TRAIN A 8 SlJ.tP SUCTION ISOLATION VALVE OPEN TB-TEST BUTTON APPLICABLE VALVE DESCRIPTION RWST TO RHR PUMP A 8812A RWST TO RHR PUMP B 88128 LIMIT SWITCH #I IS THE NORMAL POSITION SIGNAL AND IS USED FOR POSITION SIGNALS BETWEEN VALVES ASSIGNED TO THE SAME TRAIN. Rev. 14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.6*3 SAFETY INJECTION SYSTEM RECIRCULATION SUMP AND RHR SUCTION ISOLATION VALVES (SHEET 3) 
* *I I
* L NOTES: WOLF CREEK PRESSURIZER LOW PRESSURE LEAD/LAG COMPENSATED ll ll ll PRESSURIZER PRESSURE R'ELIEF INTERLOCK
<REFER TO FIGURE 7.6-4 SHEETS 1 AND 2> 1. FOR NOTATION AND DRAWING CONVENTION, REFER TO FIGURE 7.2-1, SHEET 1 ll REV. 12 WOLP CREEK 2. THIS LOGIC IS REDUNDANT UPDATED SAFETY ANALYSIS REPORT FIGURE 7.6-4 FUNCTIONAL DIAGRAM OF LOGIC REQUIREMENT FOR PRESSURIZER PRESSURE RELIEF SYSTEM INTERLOCK
<SHEET 3) : ' ---------*-----------------*--------*--------------*---*---------
WOLF CREEK SPRING RETURN TO NORMAL FROM BOTH SIDES MCB OPEN NORMAL CLOSE OPEN CLOSE VCT MOTOR OPERATED VALVE(**)
NORMALLY OPEN LO*LO LEVEL EMERGENCY MAKEUP FROM RWST VALVE* OPEN LEVEL RWST INSTRUMENT VALVE TRAIN (***) (*) VCT VALVE {**) A LB*112B LCV-1120 LCV*112B B LB-1858 LCV*112E LCV*112C Rev. 0 WOLF.CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.6-5 LOGIC DIAGRAM FOR VCT OUTLET ISOLATION VALVE INTERLOCKS ON SWITCHOVER TO RWST (SHEET 1)
MCB WOLF CREEK SPRING RETURN TO NORMAL FROM BOTH SIDES CLOSE NORMAL OPEN SIS '------1-() I ; CLOSE *OPEN RWST MOTOR OPERATED VALVE(*) NORMALLY CLOSED LB .. < LO*LO LEVEL LEVEL RWST INSTRU MENT VALVE TRAIN **
* A LB*1t2B LCV*1120 B LB*185B LCV*112E Rev. 1 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT Figure 7. 6-5 Logic Diagram for RWST Valves I nter.locks On Switchover Tb RWST (Sheet 2) i I ! j WOLF CREEK 7.7  CONTROL SYSTEMS NOT REQUIRED FOR SAFETY The general design objectives of the plant control systems are:
: a. To establish and maintain power equilibrium between the primary and secondary system during steady state unit operation.
: b. To constrain operational transients so as to preclude unit trip and reestablish steady state unit operation.
: c. To provide the reactor operator with monitoring instrumentation that indicates all required input and output control parameters of the systems and provides the operator with the capability of assuming manual control of the system.
7.
 
==7.1  DESCRIPTION==
 
The plant control systems described in this section perform the following functions:
Reactor Control System
: a. Enables the nuclear plant to accept a step load increase or decrease of 10 percent and a ramp increase or decrease
 
of 5 percent per minute within the load range of 15
 
percent to 100 percent without reactor trip, steam dump, or pressurizer relief actuation, subject to possible
 
xenon limitations.
: b. Maintains reactor coolant average temperature (T avg)        within prescribed limits by creating the bank demand signals for moving groups of RCCAs during normal
 
operation and operational transients. The T avg control        also supplies a signal to pressurizer water level control and steam dump control.
Rod Control System
: a. Provides for reactor power modulation by manual or automatic control of control rod banks in a preselected sequence and for manual operation of individual banks.
7.7-1                        Rev. 0 WOLF CREEK
: b. Systems for monitoring and indicating
: 1. Provide alarms to alert the operator if the required core reactivity shutdown margin is not available due to excessive control rod insertion.
: 2. Display control rod position.
: 3. Provide alarms to alert the operator in the event of control rod deviation exceeding a preset limit.
Plant Control System Interlocks
: a. Prevent further withdrawal of the control banks when signal limits are approached that indicate the approach
 
to a DNBR limit or kW/ft limit.
: b. Limit automatic turbine load increase to values for which the NSSS has been designed.
Pressurizer Pressure Control Maintains or restores the pressurizer pressure to the design operating pressure
+35 psi (which is within reactor trip and relief and safety valve actuation setpoint limits) following normal operational transients that induce pressure
 
changes by control (manual or automatic) of heaters and spray in the
 
pressurizer. Provides steam relief by controlling the power relief valves.
Pressurizer Water Level Control Establishes and maintains the pressurizer water level within specified limits as a function of average coolant temperature. Changes in level are caused by
 
coolant density changes induced by loading, operational, and unloading
 
transients. Level changes are produced by means of charging flow control (manual or automatic) as well as by manual selection of letdown orifices.
 
Maintaining coolant level in the pressurizer within prescribed limits by
 
actuating the charging and letdown system provides control of the reactor
 
coolant water inventory.
Steam Generator Water Level Control
: a. Establishes and maintains the steam generator water level within predetermined limits during normal operating transients.
7.7-2                        Rev. 12 WOLF CREEK
: b. The steam generator water level control system also maintains the steam generator water level to within
 
predetermined limits under unit trip conditions. It regulates the feedwater flow rate so that under operational transients the water level for the reactor
 
coolant system does not decrease below a minimum value.
Steam generator water inventory control is manual or automatic through the use of feedwater control valves.
Steam Dump Control (Also Called Turbine Bypass)
: a. Permits the nuclear plant to accept a sudden loss of load without incurring reactor trip. Steam is dumped to the
 
condenser, as necessary, to accommodate excess power
 
generation in the reactor during turbine load reduction
 
transients.
: b. Ensures that stored energy and residual heat are removed following a reactor trip to bring the plant to
 
equilibrium no-load conditions without actuation of the steam generator safety valves.
: c. Maintains the plant at no-load conditions and permits manually controlled cooldown of the plant.
: d. Maintain reactor at 100 percent power with reduced main turbine load while cycling main turbine stop and control valves.
Neutron Flux Detectors Provides information on the neutron flux distribution.
AMSAC The ATWS Mitigation System Actuation Circuitry (AMSAC) automatically initiates auxiliary feedwater and a turbine trip independent of the reactor trip system.
The AMSAC system is based on a series of generic studies (Ref. 6 and 7) on ATWS which show that acceptable consequences result, provided that the turbine trips and auxiliary feedwater flow is initiated in a timely manner.
7.7.1.1  Reactor Control System The reactor control system enables the nuclear plant to follow load changes automatically, including the acceptance of step load increases or decreases of 10 percent and ramp increases or decreases of 5 percent per minute within the
 
load range of 15 percent to 100 percent without reactor trip, steam dump, or
 
pressure relief (subject to possible xenon limitations). The system is also
 
capable of restoring coolant average temperature to within the programmed
 
temperature deadband following a change in load. Manual control rod operation
 
may be performed at any time within the range of defined insertion limits. 7.7-3 Rev. 13 WOLF CREEK The reactor control system controls the reactor coolant average temperature by regulation of control rod bank position. The reactor coolant loop average temperatures are determined from hot leg and cold leg measurements in each reactor coolant loop. There is an average coolant temperature (Tavg) computed for each loop, where:
T avg  =T hot + T cold 2 The error between the programmed reference temperature (based on turbine impulse chamber pressure) and the highest of the T avg measured temperatures (which is processed through a lead-lag compensation unit) from each of the reactor coolant loops constitutes the primary control signal, as shown in
 
general on Figure 7.7-1 and in more detail on the functional diagrams shown in
 
Figure 7.2-1 (Sheet 9). The system is capable of restoring coolant average
 
temperature to the programmed value following a change in load. The programmed coolant temperature increases linearly with turbine load from zero power to the full power condition. The T avg also supplies a signal to pressurizer level control and steam dump control and rod insertion limit monitoring.
The temperature channels needed to derive the temperature input signals for the reactor control system are fed from protection channels via isolation amplifiers.
An additional control input signal is derived from the reactor power versus turbine load mismatch signal. This additional control input signal improves
 
system performance by enhancing response and reducing transient peaks. The
 
core axial power distribution is controlled during load follow maneuvers by
 
changing (a manual operator action) the boron concentration in the RCS. The control board displays (see Section 7.7.1.3.1) indicate the need for an adjustment in the axial power distribution. Adding boron to the reactor coolant reduces T avg and cause the rods (through the rod control system) to move toward the top of the core. This action reduces power peaks in the bottom of the core. Likewise, removing boron from the reactor coolant moves the rods
 
further into the core to control power peaks in the top of the core.
7.7.1.2  Rod Control System 7.7.1.2.1  Description
 
The rod control system receives rod speed and direction signals from the T avg control system. The rod speed demand signal varies over the corresponding range of 3.75 to 45 inches per minute (6 to 7.7-4                        Rev. 5 WOLF CREEK 72 steps/minute), depending on the magnitude of the input signal. Manual control is provided to move a control bank in or out at a prescribed fixed
 
speed.When the turbine load reaches approximately 15 percent of rated load, the operator may select the "AUTOMATIC" mode, and rod motion is then controlled by the reactor control systems. A permissive interlock C-5 (see Table 7.7-1) derived from measurements of turbine impulse chamber pressure prevents
 
automatic control when the turbine load is below 15 percent. In the "AUTOMATIC" mode, the rods are again withdrawn (or inserted) in a predetermined
 
programmed sequence by the automatic programming with the control interlocks (see Table 7.7-1).
The shutdown banks are always in the fully withdrawn position during normal operation, and are moved to this position at a constant speed by manual control prior to criticality. A reactor trip signal causes them to fall by gravity
 
into the core. There are five shutdown banks.
The control banks are the only rods that can be manipulated under automatic control. Each control bank is divided into two groups to obtain smaller
 
incremental reactivity changes per step. All RCCAs in a group are electrically
 
paralleled to move simultaneously. There is individual position indication for
 
each RCCA.
Power to CRDMs is supplied by two motor generator sets operating from two separate 480 Volt, three phase busses. Each generator is the synchronous type and is driven by a 200-Hp induction motor. The ac power is distributed to the
 
rod control power cabinets through the two series-connected reactor trip
 
breakers.The variable speed rod drive programmer affords the ability to insert small amounts of reactivity at low speed to accomplish fine control of reactor coolant average temperature about a small temperature deadband, as well as
 
furnishing control at high speed. A summary of the RCCA sequencing
 
characteristics is given below.
: a. Two groups within the same bank are stepped so that the relative position of the groups will not differ by more
 
than one step.
: b. The control banks are programmed so that withdrawal of the banks is sequenced in the following order; control
 
bank A, control bank B, control bank C,  and control bank 7.7-5                        Rev. 0 WOLF CREEK D. The programmed insertion sequence is the opposite of the withdrawal sequence, i.e., the last control bank
 
withdrawn (bank D) is the first control bank inserted.
: c. The control bank withdrawals are programmed such that when the first bank reaches a preset position, the second bank begins to move out simultaneously with the first bank which continues to move toward its fully withdrawn
 
position. When the second bank reaches a preset position,  the third bank begins to move out,  and so
 
on. This withdrawal sequence continues until the unit reaches the desired power level. The control bank insertion sequence is the opposite.
: d. Overlap between successive control banks is adjustable between 0 to 50 percent (0 and 115 steps), with an accuracy of  1 step.      e. Rod speeds for either the shutdown banks or manual operation of the control banks are capable of being
 
controlled between a minimum of 6 steps per minute and a
 
maximum of 72 (+0, -10) steps per minute.
7.7.1.2.2  Features Credible rod control equipment malfunctions which could potentially cause inadvertent positive reactivity insertions due to inadvertent rod withdrawal, incorrect overlap, or malpositioning of the rods are the following:
: a. Failures in the manual rod controls:
: 1. Rod motion control switch (in-hold-out)
: 2. Bank selector switch
: b. Failures in the overlap and bank sequence program control:
: 1. Logic cabinet systems
: 2. Power supply systems
 
Failures in the manual rod controls
: 1. Failure of the rod motion control switch The rod motion control switch is a three-position lever switch. The three positions are "In," "Hold,"                              7.7-6                        Rev. 0 WOLF CREEK and "Out."  These positions are effective when the bank selector switch is in manual. Failure of the
 
rod motion control switch (contacts failing short or activated relay failures) would have the potential,              in the worst case, to produce positive reactivity
 
insertion by rod withdrawal when the bank selector switch is in the manual position or in a position which selects one of the banks.
When the bank selector switch is in the automatic position, the rods would obey the automatic commands
 
and  failures in the rod motion  control switch would have no effect on the rod motion regardless of whether the rod motion control switch is in "In,"              "Hold," or "Out."              In the case where the bank selector switch is selecting a bank and a failure occurs in the rod
 
motion switch that would command the bank "Out" even
 
when the rod motion control switch was in an "In" or
 
            "Hold" position the selected bank could inadvertently
 
withdraw. This failure is bounded in the safety
 
analysis (Chapter 15.0) by the uncontrolled bank
 
withdrawal subcritical and at power transients. A
 
reactivity insertion of up to 75 pcm/sec is assumed in the analysis due to rod movement. This value of reactivity insertion rate is consistent with the
 
withdrawal of two banks.
Failure that can cause more than one group of four mechanisms to be moved at one time within a power
 
cabinet is not a credible event because the circuit
 
arrangement for the movable and lift coils would cause the current available to the mechanisms to divide equally between coils in the two groups (in a power supply). The drive mechanism is designed so
 
that it does not operate on half current. A second
 
feature in this scenario would be the multiplexing
 
failure detection circuit included in each power cabinet. This circuit would stop rod withdrawal (or insertion).
The second case considered in the potential for inadvertent reactivity insertion due to possible
 
failures is when the selector switch is in the manual
 
position. Such a case could produce, with a failure
 
in the rod motion control switch, a scenario where
 
the rods could inadvertently withdraw in a programmed
 
sequence. The overlap and bank sequence are program 7.7-7                        Rev. 1 WOLF CREEK med when the selection is in either automatic or manual. This scenario is also bounded by the
 
reactivity values assumed in the accident analysis.
In this case, the operator can trip the reactor, or the protection system would trip the reactor via
 
power range neutron flux-high, or overtemperature T.        2. Failure of the bank selector switch A failure of the bank selector switch produces no consequences when the "in-hold-out" manual switch is
 
in the "Hold" position. This is due to the following
 
design feature:
The bank selector switch is series wired with the in-hold-out lever switch for manual and individual control rod bank operation. With the in-hold-out lever switch in the "Hold" position, the bank
 
selector switch can be positioned without rod
 
movement.
Failures in the overlap and bank sequence program control The rod control system design prevents the movement of the groups out of sequence as well as limiting the rate of reactivity insertion. The main
 
feature that performs the function of preventing malpositioning produced by
 
groups out of sequence is included in the block supervisory memory buffer and control. This circuitry accepts and stores the externally generated command signals. In the event of out of sequence input command to the rods while they are in movement, this circuit inhibits the buffer memory from accepting the command. If a change of signal command appears, this circuit would stop the system after allowing the slave cyclers to finish their current sequencing.
Failure of the components related to this system also produces rod deviation
 
alarm and insertion limit alarm. Failures within the system such as failures
 
of supervisory logic cards, pulser cards, etc., also causes an urgent alarm.
 
An urgent alarm is followed by the following actions:
Automatic deenergizing of the lift coil and reduced current energizing of the stationary gripper coils and movable gripper coils Activation of the alarm light (urgent failure) on the power supply cabinet front panel Activation of rod control urgent failure annunciation window on the plant annunciator 7.7-8                        Rev. 0 WOLF CREEK The urgent alarm is produced in general by:
Regulation failure detector
 
Phase failure detector
 
Logic error detector
 
Multiplexing error detector
 
Interlock failure detector
: 1. Logic cabinet failures
 
The rod control system is designed to limit the rod speed control signal output to a value that causes the pulser (logic cabinet) to drive the control rod
 
driving mechanism at 72 steps per minute. If a
 
failure should occur in the pulses or the reactor
 
control system, the highest stepping rate possible is
 
77 steps per minute, which corresponds to one step
 
every 780 milliseconds. A commanded stepping rate
 
higher than 77 steps per minute would result in "GO" pulses entering a slave cycler while it is sequencing
 
its mechanisms through a 780 millisecond step. This
 
condition stops the control bank motion
 
automatically, and alarms are activated locally and
 
in the control room. It also causes the affected
 
slave cycler to reflect further "GO" pulses until it
 
is reset.
Failures that cause the 780 millisecond step sequence time to shorten will not result in higher rod speeds, since the stepping rate is proportional to the
 
pulsing rate. Simultaneous failures in the pulser or
 
rod control system and in the clock circuits that
 
determine the 780 millisecond stepping sequence could
 
result in higher CRDM speed; however, in the unlikely
 
event of these simultaneous multiple failures the
 
maximum CRDM operation speed would be no more than approximately 100 steps per minute due to mechanical limitation. This speed has been verified by tests
 
conducted on the CRDMs. Surveillance testing of the Reactor Control System and  the Rod Control System is performed at periodic intervals  to detect failures that could lead to an increase in the  rod speed.
7.7-9                        Rev. 16 WOLF CREEK Failures causing movement of the rods out of sequence:
No single failure was discovered (Ref. 2) that would cause a rapid uncontrolled withdrawal of Control Bank D (taken as worst case) when operating in the automatic bank overlap control mode with the reactor at near full power output. The analysis revealed
 
that many of the failures postulated were in a safe direction and that rod movement is blocked by the rod
 
urgent alarm.
: 2. Power supply system failures
 
Analysis of the power cabinet disclosed no single component failures that would cause the uncontrolled
 
withdrawal of a group of rods serviced by the power
 
cabinet. The analysis substantiates that the design
 
of a power cabinet is "fail-preferred" with regard to
 
a rod withdrawal accident if a component fails. The
 
end results of the failure is either that of blocking
 
rod movement or that of dropping an individual rod or
 
rods or a group of rods. No failure, within the
 
power cabinet, which could cause erroneous drive
 
mechanism operation would remain undetected.
 
Sufficient alarm monitoring (including "urgent" alarm) is provided in the design of the power cabinet
 
for fault detection of those failures which could
 
cause erroneous operation of a group of mechanisms.
 
As noted in the foregoing, diverse monitoring systems are available for detection of failures that cause
 
the erroneous operation of an individual control rod
 
drive mechanism.
In summary, no single failure within the rod control system can cause either reactivity insertions or mal-positioning of the control rods resulting in core
 
thermal conditions not bounded by analyses contained in Chapter 15.0.
7.7.1.3  Plant Control Signals for Monitoring and Indicating 7.7.1.3.1    Monitoring Functions Provided by the Nuclear Instrumentation System The power range channels are used to measure power level, axial flux imbalance, and radial flux imbalance. These channels are capable of recording overpower
 
excursions up to 200 percent of full power. Suitable alarms are derived from
 
these signals, as described below.      7.7-10    Rev. 0 WOLF CREEK Basic power range signals are:
: a. Total current from a power range detector (four signals from separate detectors); these detectors are vertical and have a total active length of 10 feet.
: b. Current from the upper half of each power range detector (four signals).
: c. Current from the lower half of each power range detector (four signals).
The following (including standard signal processing for calibration) are derived from these basic signals:
: a. Indicated nuclear power (four signals).
: b. Indicated axial flux imbalance (), derived from upper half flux minus lower half flux (four signals).
Alarm functions derived are as follows:
: a. Deviation (maximum minus minimum of four) in indicated clear power.
: b. Upper radial tilt (maximum to average of four) on upper half currents.
: c. Lower radial tilt (maximum to average of four) on lower half currents.
The plant computer archives the 8 ion chamber signals, i.e., upper and lower currents for each detector, nuclear power and axial imbalance. Indicators are provided on the control board for nuclear power and for axial flux imbalance.
The axial flux difference imbalance deviation  alarms are derived from the plant computer which determines the 1-minute averages of the excore detector outputs to monitor  in the reactor core and alerts the operator when  alarm conditions exist. Above a preset power level, an alarm message is output immediately upon determining a delta flux exceeding a preset band. For periods
 
during which difference is inoperable, the axial flux difference is logged, as defined in the Technical Requirements Manual. No power reduction is required during this period of manual surveillance.      7.7-11    Rev. 21 WOLF CREEK Additional background information on the nuclear instrumentation system can be found in Reference 1.
7.7.1.3.2  Rod Position Monitoring
 
Two separate systems are provided to sense and display control rod position as described below:
: a. Digital rod position indication system
 
The digital rod position indication system measures the actual position of each control rod, using a detector which consists of discrete coils mounted concentrically with the rod drive pressure housing. The coils are
 
located axially along the pressure housing and
 
magnetically sense the entry and presence of the rod
 
drive shaft through its centerline. For each detector, the coils are interlaced into two data channels, and are
 
connected to the containment electronics (Data A and B)
 
by separate multiconductor cables. By employing two
 
separate channels of information, the digital rod
 
position indication system can continue to function (at
 
reduced accuracy) when one channel fails. Multiplexing
 
is used to transmit the digital position signals from the
 
containment electronics to the control board display unit.
The control board display unit contains a column of light-emitting-diodes (LEDs) for each rod. At any given
 
time, the one LED illuminated in each column shows the position for that particular rod. Since shutdown rods are always fully withdrawn with the plant at power, their position is displayed to  4 steps only from rod bottom to 18 steps and from 210 steps to 228 steps. All intermediate positions of the rod are represented by a
 
single "transition" LED. Each rod of the control banks has its position displayed to  4 steps throughout its range of travel.
Included in the system is a rod at bottom signal for each rod that operates a local alarm. Also a control room
 
annunciator is actuated when any shutdown rod or control
 
bank A rod is at bottom.      7.7-12    Rev. 0 WOLF CREEK
: b. Demand position system
 
The demand position system counts pulses generated in the
 
rod drive control system to provide a digital readout of the demanded bank position.
 
The demand position and digital rod position indication systems are separate
 
systems, but safety criteria were not involved in the separation, which was a
 
result only of operational requirements. Operating procedures require the
 
reactor operator to compare the demand and indicated (actual) readings from the
 
rod position indication system so as to verify operation of the rod control
 
system.
 
7.7.1.3.3  Control Bank Rod Insertion Monitoring When the reactor is critical, an indication of reactivity status in the core is
 
the position of the control bank in relation to reactor power (as indicated by the reactor coolant system loop  T) and coolant average temperature.
Insertion limits for the control banks are defined as a function of reactor power. Two alarms are provided for each control bank.
: a. The "low" alarm alerts the operator of an approach to the rod insertion limits requiring boron addition by following normal procedures with the chemical and volume control system.
: b. The "low-low" alarm alerts the operator to verify shutdown margin within limits of COLR or to add boron to the reactor coolant system    by any one of several alternate methods.
The purpose of the control bank rod insertion monitor is to give warning to the
 
operator of excessive rod insertion. The insertion limit maintains sufficient
 
core reactivity shutdown margin following reactor trip, provides a limit on the
 
maximum inserted rod worth in the unlikely event of a hypothetical rod ejection, and limits rod insertion so that acceptable nuclear peaking factors
 
are maintained. Since the amount of shutdown reactivity required for the
 
design shutdown margin following a reactor trip increases with increasing
 
power, the allowable rod insertion limits must be decreased (the rods must be
 
withdrawn further) with increasing power. Two parameters which are
 
proportional to power are used as inputs to the insertion monitor. These are the T between the hot leg and the cold leg, which is a direct function of reactor power, and T avg ,  which is programmed as a function of power. The rod insertion monitor uses parameters for each control rod bank as follows:
 
7.7-13    Rev. 22 WOLF CREEK Z
LL = A(T)auct +B(T avg)auct + C  where:
Z LL = maximum permissible insertion limit for affected control bank
 
(T)auct = highest  T of all loops
 
(T avg)auct = highest T avg of all loops
 
A, B, C    = constants chosen to maintain Z LL >                    actual limit based on physics    =
calculations
 
The control rod bank demand position (Z) is compared to Z LL as follows:
If Z - Z LL  D a low alarm is actuated.
 
If Z - Z LL  E a low-low alarm is actuated.
 
Since the highest values of T avg and T are chosen by auctioneering, a conservatively high representation of power is used in the insertion limit calculation.
 
Actuation of the low alarm alerts the operator of an approach to a reduced
 
shutdown reactivity situation. Administrative procedures require the operator
 
to add boron through the chemical and volume control system. Actuation of the
 
low-low alarm requires shutdown margin verification or boration. The value for "E" is chosen so that the low-low alarm would normally be actuated before the insertion limit is reached. The value for "D" is chosen to allow the operator
 
to follow normal boration procedures. Figure 7.7-2 shows a block diagram
 
representation of the control rod bank insertion monitor. The monitor is shown
 
in more detail on the functional diagrams shown in Figure 7.2-1 (Sheet 9). In addition to the rod insertion monitor for the control banks, the plant
 
computer, which monitors individual rod positions, provides an alarm that is
 
associated with the rod deviation alarm discussed in Section 7.7.1.3.4 to warn
 
the operator if any shutdown RCCA leaves the fully withdrawn position.
 
7.7-14    Rev. 22 WOLF CREEK Rod insertion limits are established by:
: a. Establishing the allowed rod reactivity insertion at full power consistent with the purposes given above.
: b. Establishing the differential reactivity worth of the control rods when moved in normal sequence.
: c. Establishing the change in reactivity with power level by relating power level to rod position.
: d. Linearizing the resultant limit curve. All key nuclear parameters in this procedure are measured as part of the initial and periodic physics testing program.
Any unexpected change in the position of the control bank under automatic control, or a change in coolant temperature under manual control, provides a
 
direct and immediate indication of a change in the reactivity status of the
 
reactor. In addition, samples are taken periodically of coolant boron
 
concentration. Variations in concentration during core life provide an
 
additional check on the reactivity status of the reactor, including core
 
depletion.
7.7.1.3.4  Rod Deviation Alarm
 
The position of any control rod is compared to the position of other rods in the bank. A rod deviation alarm is generated by the digital rod position
 
indication system if a preset rod deviation limit is exceeded. The deviation
 
alarm of a shutdown rod is based on a preset insertion limit being exceeded.
The demanded and measured rod position signals are also monitored by the plant computer which provides a visual alarm screen whenever an individual rod position signal deviates from the other rods in the bank by a preset limit.
The alarm can be set with appropriate allowance for instrument error and within sufficiently narrow limits to preclude exceeding core design hot channel factors.Figure 7.7-3 is a block diagram of the rod deviation comparator and alarm system implemented by the plant computer. Additionally, the digital rod position indication system contains rod deviation circuitry that detects and alarms the following conditions:
: a. When any two rods within the same control bank are misaligned by a preset distance ( 12 steps), and
: b. When any shutdown rod is below the full-out position by a preset distance (18 steps) except during normal shutdown bank withdrawal or insertion or on update of rod bank
 
position on the plant computer.      7.7-15    Rev. 21 WOLF CREEK 7.7.1.3.5  Rod Bottom Alarm The rod bottom signal for the control rods in the digital rod position indication system is used to operate a control relay, which generates the "ROD BOTTOM ROD DROP" alarm.
7.7.1.4  Plant Control System Interlocks The listing of the plant control system interlocks, along with the description of their derivations and functions, is presented in Table 7.7-1. The
 
designation numbers for these interlocks are preceded by "C."  The development
 
of these logic functions is shown in the functional diagrams (see Figure 7.2-1, Sheets 9 through 16).
7.7.1.4.1  Rod Stops
 
Rod stops are provided to prevent abnormal power conditions which could result from excessive control rod withdrawal initiated by either control system
 
malfunction or operator violation of administrative procedures.
Rod stops are the C-1, C-2, C-3, C-4, and C-5 control interlocks identified in Table 7.7-1. The C-3 rod stop derived from overtemperature T and the C-4 rod stop derived from overpower T are also used for turbine runback, which is discussed below.
7.7.1.4.2  Automatic Turbine Load Runback Automatic turbine load runback is initiated by an approach to an overpower or overtemperature condition. This prevents high power operation that might lead
 
to an undesirable condition, which, if reached, is protected by reactor trip.
Turbine load reference reduction is initiated by either an overtemperature or overpowerT signal. Two-out-of-four coincidence logic is used.
A rod stop and turbine runback are initiated when T > T rod stop for both the overtemperature and the overpower condition.
For either condition in general T rod stop =T setpoint- B p      7.7-16    Rev. 0 WOLF CREEK where:
 
B p = a setpoint bias where T setpoint refers to the overtemperature T reactor trip value and the overpower T reactor trip value for the two conditions.
 
The turbine runback is continued until T is equal to or less than T rod stop or 254MW.
This function serves to maintain an essentially constant margin to trip.
 
7.7.1.4.3  Turbine Loading Stop
 
An interlock (C-16) is provided to limit turbine loading during a rapid return
 
to power transient when a reduction in reactor coolant temperature is used to
 
increase reactor power (through the negative moderator coefficient). This
 
interlock limits the reduction in coolant temperature so that it does not reach
 
cooldown accident limits and preserves satisfactory steam generator operating
 
conditions. Subsequent automatic turbine loading can begin after the interlock
 
has been cleared by an increase in coolant temperature, which is accomplished
 
by reducing the boron concentration in the coolant.
 
7.7.1.5  Pressurizer Pressure Control
 
The reactor coolant system pressure is controlled by using either the heaters (in the water region) or the spray (in the steam region) of the pressurizer
 
plus steam relief for large transients.
 
The electrical immersion heaters are located near the bottom of the
 
pressurizer. A portion of the heater group is proportionally controlled to
 
correct small pressure variations. These variations are caused by heat losses, including heat losses due to a small continuous spray. The remaining (back-up)
 
heaters are turned on when the pressurizer pressure controlled signal demands
 
approximately 100-percent proportional heater power.
 
The spray nozzle is located on the top of the pressurizer. Spray is initiated
 
when the pressure controller spray demand signal is above a given setpoint. 
 
The spray rate increases proportionally with increasing spray demand signal
 
until it reaches a maximum value.
 
7.7-17    Rev. 27 WOLF CREEK Steam condensed by the spray reduces the pressurizer pressure. A small continuous spray is normally maintained to reduce thermal stresses and thermal
 
shock and to help maintain uniform water chemistry and temperature in the pressurizer.
Power relief valves limit system pressure for large positive pressure transients. In the event of a large load reduction, not exceeding the design plant load rejection capability, the pressurizer power-operated relief valves
 
might be actuated for the most adverse conditions, e.g., the most negative Doppler coefficient and the maximum incremental rod worth. The relief capacity
 
of the power-operated relief valves is sized large enough to limit the system pressure to prevent actuation of high pressure reactor trip for the above condition.
A block diagram of the pressurizer pressure control system is shown on Figure 7.7-4.7.7.1.6  Pressurizer Water Level Control The pressurizer operates by maintaining a steam cushion over the reactor coolant. As the density of the reactor coolant varies with temperature, the
 
steam water interface is adjusted to compensate for cooling density variations
 
with relatively small pressure disturbances.
The water inventory in the reactor coolant system is maintained by the chemical and volume control system. During normal plant operation, the charging flow
 
varies to produce the flow demanded by the pressurizer water level controller.
 
The pressurizer water level is programmed as a function of coolant average temperature, with the highest average temperature (auctioneered) being used.
 
The pressurizer water level decreases as the load is reduced from full load.
This is a result of coolant contraction following programmed coolant temperature reduction from full power to low power. The programmed level is designed to match as nearly as possible the level changes resulting from the coolant temperature changes.
While raising or lowering pressurizer water level during startup and shutdown operations, the charging flow is manually regulated from the main control room.
Once normal pressurizer water level is attained, charging flow can be placed in auto. The letdown line isolation valves are closed on low pressurizer level.
A block diagram of the pressurizer water level control system is shown on
 
Figure 7.7-5.      7.7-18    Rev. 12 WOLF CREEK 7.7.1.7  Steam Generator Water Level Control Each steam generator is equipped with a three-element feedwater flow controller which maintains a programmed water level. The three-element feedwater controller regulates the feedwater valve by continuously comparing the feedwater flow signal, the water level signal, the programmed level, and the pressure compensated steam flow signal. The feedwater pump speed is varied to maintain a programmed pressure differential between the steam header and the feedwater pump discharge header. The speed controller continuously  compares the actual  P with a programmed  Pref which is a linear function of steam flow. Continued delivery of feedwater to the steam generators is required as a sink for the heat stored and generated in the reactor following a reactor trip and turbine trip. An override signal closes all feedwater valves when the
 
average coolant temperature is below a given temperature and the reactor has
 
tripped. Manual override of the feedwater control system is available at all times.When the nuclear plant is operating at very low power levels (as during startup), the steam and feedwater flow signals are not usable for control.
 
Therefore, a secondary automatic control system is provided for operation at
 
low power. This system uses the steam generator water level and nuclear power signals in a feed forward control scheme to position a bypass valve which is in parallel with the main feedwater regulating valve. Switchover from the bypass feedwater control system (low power) to the main feedwater control system is
 
initiated by the operator at approximately 25-percent power.
Block diagrams of the steam generator water level control system and the main feedwater pump speed control system are shown in Figures 7.7-6 and 7.7-7.
7.7.1.8  Steam Dump Control The steam dump system, together with control rod movement, is designed to accept a 50-percent loss of net load without tripping the reactor.
The automatic steam dump system is able to accommodate this abnormal load rejection and to reduce the effects of the transient imposed upon the reactor coolant system. By bypassing main steam directly to the condenser, an
 
artificial load is thereby maintained on the primary system. The rod control
 
system can then reduce the reactor temperature to a new equilibrium value
 
without      7.7-19    Rev. 0 WOLF CREEK causing overtemperature and/or overpressure conditions. The steam dump steam flow capacity is 40 percent of full load steam flow at full load steam
 
pressure.If the difference between the reference T avg (T ref) based on turbine impulse chamber pressure and the lead-lag compensated auctioneered T avg exceeds a predetermined amount, and the interlock mentioned below is satisfied, a demand signal will actuate the steam dump to maintain the reactor coolant system
 
temperature within control range until a new equilibrium condition is reached.
To prevent actuation of steam dump on small load perturbations, an independent load rejection sensing circuit is provided. This circuit senses the rate of decrease in the turbine load, as detected by the turbine impulse chamber pressure. It is provided to unblock the dump valves when the rate of load rejection exceeds a preset value corresponding to a 10-percent step load decrease or a sustained ramp load decrease of 5 percent per minute.
A block diagram of the steam dump control system is shown on Figure 7.7-8.
 
7.7.1.8.1  Load Rejection Steam Dump Controller
 
This circuit prevents a large increase in reactor coolant temperature following a large, sudden load decrease. The error signal is a difference between the lead-lag compensated auctioneered T avg and the reference T avg based on turbine impulse chamber pressure.
The T avg signal is the same as that used in the reactor coolant system. The lead-lag compensation for the T avg signal is to compensate for lags in the plant thermal response and in valve positioning. Following a sudden load decrease, T ref is immediately decreased and T avg tends to increase, thus generating an immediate demand signal for steam dump. Since control rods are
 
available in this situation, steam dump terminates as the error comes within
 
the maneuvering capability of the control rods.      7.7-20    Rev. 1 WOLF CREEK 7.7.1.8.2  Plant Trip Steam Dump Controller Following a reactor trip, the load rejection steam dump controller is defeated, and the plant trip steam dump controller becomes active. Since control rods are not available in this situation, the demand signal is the error signal between the lead-lag compensated  auctioneered  T avg and the no-load reference T avg. When the error signal exceeds a predetermined setpoint, the dump valves are tripped open in a prescribed sequence. As the error signal reduces in
 
magnitude, indicating that the RCS T avg is being reduced toward the reference no-load value, the dump valves are modulated by the plant trip controller to regulate the rate of removal of decay heat and thus gradually establish the
 
equilibrium hot shutdown condition.
7.7.1.8.3  Steam Header Pressure Controller
 
Residual heat removal (at operating temperature) and steam header pressure (while cycling main turbine stop and control valves) are maintained by the steam generator pressure controller (manually selected) which controls the amount of steam flow to the condenser. This controller operates a portion of the same steam dump valves to the condensers which are used during the initial transient following turbine or reactor trip on load rejection.
7.7.1.9  Neutron Flux Detectors The neutron flux detectors are movable miniature neutron detectors which can be positioned at the center of selected fuel assemblies, throughout the length of the incore flux thimble to measure neutron flux along the fuel assembly vertical axis. The incore flux thimble normally extends the full length of the fuel assembly, but may be somewhat shorter due to flux thimble repositioning.
Repositioning may be required to mitigate the consequences of interaction between the flux thimble and its supports which could reduce the tube wall
 
thickness. The basic system for insertion of the neutron detectors is shown in
 
Figure 7.7-9.
7.7.1.9.1  Movable Neutron Flux Detector Drive System Miniature fission chamber detectors can be remotely positioned in retractable guide thimbles to provide flux-mapping of the core. The stainless steel detector shell is welded to the leading end of      7.7-21    Rev. 9 WOLF CREEK the helical wrap drive cable and to stainless steel sheathed coaxial cable.
The retractable thimbles, into which the miniature detectors are driven, are
 
pushed into the reactor core through conduits which extend from the bottom of the reactor vessel down through the concrete shield area and then up to a thimble seal table. Their distribution over the core is nearly uniform with
 
about the same number of thimbles located in each quadrant.
The thimbles are closed at the leading ends, are dry inside, and serve as the pressure barrier between the reactor water pressure and the atmosphere.
 
Mechanical seals between the retractable thimbles and the conduits are provided
 
at the seal table. During reactor operation, the retractable thimbles are
 
stationary. They are extracted downward from the core during refueling to
 
avoid interference within the core. A space above the seal table is provided
 
for the retraction operation.
The drive system for the insertion of the miniature detectors consists basically of drive assemblies, 6-path transfer assemblies, and 15-path transfer
 
assemblies, as shown in Figure 7.7-9. The drive system pushes hollow helical
 
wrap drive cables into the core with the miniature detectors attached to the
 
leading ends of the cables and small diameter sheathed coaxial cables threaded
 
through the hollow centers back to the ends of the drive cables. Each drive
 
assembly consists of a gear motor which pushes a helical wrap drive cable and a
 
detector through a selective thimble path by means of a special drive box and
 
includes a storage device that accommodates the total drive cable length. Each
 
detector has access to all thimble locations via the 6- and 15-path rotary
 
assemblies.
7.7.1.9.2  Control and Readout Description The control and readout system provides means for inserting the miniature neutron detectors into the reactor core and withdrawing the detectors while
 
plotting neutron flux versus detector position. The control system is located in the control room. Limit switches in each transfer device provide feedback of path selection operation. Each gear box drives a resolver for position
 
feedback. One 6-path transfer selector is provided for each drive unit to
 
insert the detector in one of six functional modes of operation. One 15-path
 
transfer is also provided for each drive unit that is then used to route a detector into any one of up to 15 selectable paths. A common path is provided
 
to permit cross calibration of the detectors.
The control room contains the necessary equipment for control, position indication, and flux recording for each detector.      7.7-22    Rev. 5 WOLF CREEK A "flux-mapping" consists, briefly, of selecting flux thimbles in given fuel assemblies at various core quadrant locations. The detectors are driven to the
 
top of the core and stopped automatically. An x-y plot (position versus flux level) is initiated with the slow withdrawal of the detectors through the core from top to a point below the bottom. In a similar manner, other core
 
locations are selected and plotted. Each detector provides axial flux distribution data along the center of a fuel assembly.
Various radial positions of detectors are then compared to obtain a flux map for a region of the core.
The number and location of these thimbles have been chosen to permit measurement of local-to-average peaking factors to an accuracy of +5 percent (95-percent confidence). Measured nuclear peaking factors are increased by 5
 
percent to allow for this accuracy. If the measured power peaking is larger
 
than acceptable, reduced power capability is indicated.
Operating plant experience has demonstrated the adequacy of the incore instrumentation in meeting the design bases stated.
7.7.1.10 Boron Concentration Monitoring System The boron concentration monitoring system has been abandoned-in-place.
The boron concentration monitoring system utilizes a sampler assembly unit which contains a neutron source and neutron detector located in a shield tank.
 
A thermal neutron absorption technique is used. Piping within the shield tank is arranged to provide coolant sample flow between the neutron source and the neutron detector. Neutrons originating at the source are thermalized in the sample and the surrounding moderator. These neutrons then pass through the sample and impinge upon the detector. The number of neutrons which survive the transit from the source to the detector is inversely proportional to the boron concentration in the sample. The boron concentration is calculated by
 
monitoring the neutron countrate in conjunction with the proper transfer function. The neutron cross-section of the boron in the sample is also a
 
function of the neutron energy and, subsequently, the sample temperature.
 
Therefore, the sample temperature is also monitored and the transfer function
 
from the neutron countrate to boron concentration modified to compensate for the variance of temperature.
The processor assembly is used to convert the neutron countrate and temperature data from the sampler assembly to parts per million (ppm) of boron, and to
 
prepare the data for local and remote display. The system characteristics are
 
listed in Table 7.7-2.      7.7-23    Rev. 14 WOLF CREEK
: a. Sampler assembly The sampler assembly consists of a polyethylene cylinder encased in a stainless steel liner (see Figure 7.7-10).
The polyethylene serves as a neutron moderator and shield. A cavity (source tube) is located in the center of the shield into which is inserted a neutron source on the end of a polyethylene rod (source plug). Immediately
 
adjacent to the source tube is a second larger cavity into which an annulus assembly and a top plug assembly
 
are inserted. Details of these two assemblies are shown in Figure 7.7-11.
The annulus assembly consists of two concentric tubes with top and bottom plates. A neutron detector is positioned inside the smaller tube. The coolant sample
 
is circulated between the concentric tubes. The sample
 
is brought into and taken out of the annular region via
 
tubes provided for connection to plant piping. The
 
entire assembly is made of stainless steel.
The top plug assembly consists of a polyethylene plug with appropriate ports for the input and output tubes and
 
the detector signal cable. A stainless steel top plate
 
is provided for mounting to the sampler assembly.
: b. Processor assembly The processor assembly controls the operation of the system. It processes the neutron countrate and temperature data from the sampler assembly, displays the calculated boron concentration, and transmits the result serially for remote display. A block diagram which
 
depicts the functional operation of the processor
 
assembly is shown in Figure 7.7-12. The neutron
 
countrate and sample temperature measurements are
 
processed to a microprocessor. The microprocessor
 
repeatedly solves an algorithm to convert the input
 
information to a boron concentration measurement. In
 
order to make the above calculation, several constants are required. These constants are determined by calibration and are entered in the microprocessor by
 
manual interactions with a keypad. The display unit
 
presents the calculated boron concentration in units of ppm in integer format.      7.7-24    Rev. 1 WOLF CREEK
: c. Remote display assembly The function of this unit is to display the boron concentration at a location (usually in the control room) remote from the processor assembly. This remote display may be located up to 1,000 feet from the processor assembly. Boron concentration data generated at the process assembly is transmitted serially over a twisted
 
shielded pair. The remote display assembly contains the circuits necessary to decode and display the data.
The boron concentration monitoring system is designed for use as an advisory system. It is not designed as a safety system or component of a safety system.
The boron concentration monitoring system is not part of a control element or control system, nor is it designed for this use. No credit is taken for this
 
system in any accident analysis. Therefore, redundancies of measurement
 
components, self checking subsystems, malfunction annunciations, and diagnostic
 
circuitry are not included in this system. As a general operating aid, it
 
provides information as to when additional check analyses are warranted, rather
 
than a basis for fundamental operating decisions. During normal plant
 
operations, the boron concentration varies between 0 ppm at end-of-cycle to values near the RWST concentration at beginning-of-cycle. The boron concentration monitoring system operates within a +10 ppm range, as shown in Figure 7.7-13.
7.7.1.11  ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY The ATWS Mitigation System Actuation Circuitry (AMSAC) automatically initiates auxiliary feedwater flow, isolates Steam Generator blowdown and sample lines, and initiates a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event.
The AMSAC equipment is located in the control room and consists of logic assemblies, isolation devices, and interconnecting cables interfacing with other plant equipment.
Four Reactor Protection System (RPS) narrow range steam generator level loops and two turbine impulse pressure loops provide inputs to AMSAC from the 7300 racks. The AMSAC actuation outputs go to the Balance of Plant (BOP) Engineered Safety Features Actuation System (ESFAS) and Turbine Generator Electrohydraulic Control (EHC) cabinets.      7.7-25    Rev. 4 WOLF CREEK An AMSAC actuation occurs when 3 out of 4 steam generator narrow range level
 
signals fall below the AMSAC setpoint (12% of span) for more than 25 seconds
 
and the AMSAC permissive has been armed. AMSAC is armed above 34.9% equivalent
 
reactor power as indicated by both the first stage turbine impulse pressure
 
signals and is disabled 360 seconds after 1 out of 2 turbine impulse pressure
 
signals fall below 34.9% equivalent reactor power. The 25 second AMSAC
 
actuation time delay allows the RPS to operate first. Where as AMSAC must be
 
armed above 40% reactor power, the actual setpoint has been established as
 
34.9% equivalent reactor power to allow for instrument loop inaccuracies.
 
An AMSAC actuation causes the BOP-ESFAS system to start the AFW pump, close the steam generator blowdown isolation valves and close the steam generator sample isolation valves. The safety-related systems (RPS and BOP-ESFAS) are isolated from AMSAC through qualified isolation devices. AMSAC also provides signals to the EHC cabinets to trip the turbine by using one set of three AMSAC trip relay contacts wired to three separate digital input modules in the Ovation Turbine Control System (TCS) Emergency Trip System (ETS) Controller. A second set of contacts is wired to separate digital input modules in the TCS Ovation Operator Auto/Overspeed Protection and Control (OA/OPC) Controller to provide additional redundancy. Redundant trip outputs are initiated using two-out-of-three logic through the ETS controller to TDM A and OA/OPC controller to TDM B.
 
AMSAC provides two main control board annunciators and several local indicating
 
lights. One control room annunciator indicates that an AMSAC actuation will
 
occur after a 25 second time delay and the other is a trouble alarm that
 
indicates several miscellaneous AMSAC trouble conditions (e.g., Master bypass
 
switch in ON position, Operating bypass switch not in OFF position, ATWS logic
 
trouble, power supply malfunction, module out/door open). The local indicating
 
lights on the AMSAC logic cabinet indicates the several miscellaneous status (e.g., AMSAC armed, operating bypass, ATWS logic trouble, ATWS panel trouble, power OK, ATWS SG level pretrip) of AMSAC operation.
 
AMSAC cannot be manually reset by the operators. AMSAC actuates automatically;
 
there is no manual AMSAC initiation. AMSAC is automatically armed above 34.9%
 
equivalent reactor power and bypassed below 34.9% equivalent reactor power.
 
AMSAC initiates an auxiliary feedwater actuation signal in the event of an ATWS
 
event and a turbine trip within 30 seconds of an ATWS event. The auxiliary
 
feedwater pumps must be operating at full operating speed within 60 seconds of
 
an AFAS signal to deliver the required flow rate based on pump performance (response times include sensor and relay delays).
 
AMSAC utilizes three identical logic assemblies, diverse from the SSPS, and a 2
 
out of 3 actuation logic to prevent inadvertent trips due to the AMSAC
 
circuitry and improve reliability. Testing of AMSAC through the final
 
actuation devices will be performed every refueling outage.
 
7.7-26 Rev. 27 WOLF CREEK References 3-5 provide additional discussions on AMSAC diversity from the RPS, logic power supplies, safety-related interfaces via Class 1E isolation devices, graded QA program, maintenance and testing bypasses via 4 permanently installed bypass switches and annunciated by the previously mentioned trouble alarm, electrical independence and physical separation from the RPS, testability at
 
power in bypass and completion of mitigative action once initiated. The steam generator level sensors used for input to AMSAC are different than those used to drive the steam generator level control system, thereby precluding adverse
 
control system interactions. The logic power supply is independent from the RPS power supplies. AMSAC is capable of performing its intended function upon
 
a loss of offsite power. Removal of the permissive signal is delayed by 360 seconds to avoid blocking AMSAC before it can perform its function in the event a turbine trip occurs. Existing protection system level transmitters, sensing lines, and sensor power supplies are used for input to AMSAC. Input isolation is attained via 7300 isolation cards; output isolation is attained via
 
isolation relays before going to the BOP-ESFAS. AMSAC output can be disabled
 
via the master bypass switch to avoid actuation during maintenance and testing.
 
Each of the three logic assemblies is also provided with an individual bypass
 
switch to permit troubleshooting and repair.
7.7.2  ANALYSIS
 
The plant control systems are designed to assure high reliability in any anticipated operational occurrences. Equipment used in these systems is designed and constructed with a high level of reliability.
Proper positioning of the control rods is monitored in the control room by bank arrangements of the individual position columns for each RCCA. A rod deviation
 
alarm alerts the operator of a deviation of one RCCA from the other rods in
 
that bank position. There are also insertion limit monitors with visual and audible annunciation. A rod bottom alarm signal is provided to the control room for each RCCA. Four excore long ion chambers also detect asymmetrical flux distribution indicative of rod misalignment.
Overall reactivity control is achieved by the combination of soluble boron and RCCAs. Long-term regulation of core reactivity is accomplished by adjusting
 
the concentration of boric acid in the reactor coolant. Short-term reactivity
 
control for  7.7-27 Rev. 4 WOLF CREEK power changes is accomplished by the plant control system, which automatically moves RCCAs. This system uses input signals including neutron flux, coolant
 
temperature, and turbine load.
The axial core power distribution is controlled by moving the control rods through changes in RCS boron concentration. Adding boron causes the rods to move out, thereby reducing the amount of power in the bottom of the core. This allows power to redistribute toward the top of the core. Reducing the boron
 
concentration causes the rods to move into the core, thereby reducing the power in the top of the core. As a result, power is redistributed toward the bottom
 
of the core.
The plant control systems will prevent an undesirable condition in the operation of the plant that, if reached, is protected by reactor trip. The description and analysis of this protection is covered in Section 7.2. Worst-
 
case failure modes of the plant control systems are postulated in the analysis
 
of off-design operational transients and accidents covered in Chapter 15.0, such as the following:
: a. Uncontrolled RCCA bank withdrawal from a subcritical or low power startup condition.
: b. Uncontrolled RCCA bank withdrawal at power.
: c. RCCA misalignment.
: d. Loss of external electrical load and/or turbine trip.
: e. Loss of all nonemergency ac power to the station auxiliaries.
: f. Feedwater system malfunctions that result in a decrease in feedwater temperature.
: g. Excessive increase in secondary steam flow.
: h. Inadvertent opening of a steam generator atmospheric relief or safety valve.
These analyses show that a reactor trip setpoint is reached in time to protect the health and safety of the public under those postulated incidents and that
 
the resulting coolant temperatures produce a DNBR well above the thermal design
 
limit DNBR. Thus, there will be no cladding damage and no release of fission
 
products to the RCS under the assumption of these postulated worst-case failure
 
modes of the plant control system. 7.7-28 Rev. 14 WOLF CREEK RESPONSE Failures have been postulated which affect the major NSSS control systems and demonstrate that for each failure the resulting event is within the bounds of existing accident analyses. The events which are considered are:
: a. Loss of any single instrument
: b. Break of any single instrument line
: c. Loss of power to all systems powered by a single power supply system (i.e., single inverter)
The analysis is conducted for five major NSSS control systems:
: a. Reactor control system
: b. Steam dump control system
: c. Pressurizer pressure control system
: d. Pressurizer water level control system
: e. Feedwater control system
 
The initial conditions are assumed to be anywhere within the full operating power range of the plant (i.e., 0-100 percent), where applicable.
The results of the analysis indicate that, for any of the postulated events considered in a. through c. above, the Condition II accident analyses given in Chapter 15.0 are bounding.
LOSS OF ANY SINGLE INSTRUMENT Table 7.7-3, Loss of Any Single Instrument, is a sensor-by-sensor evaluation of the effect on the control systems itemized above caused by a sensor failing either high or low. The particular sensor considered is given, along with the
 
number of channels which exist, the failed channel, the control systems
 
impacted by the sensor, the effects on the control systems for failures in both directions, and the bounding USAR accident. Where no control action occurs or where control action is in a safe direction, no bounding accident is given.      7.7-29    Rev. 4 WOLF CREEK The table clearly shows that for any single instrument failure, either high or
 
low, the Condition II events itemized in Chapter 15.0 are bounding.
 
LOSS OF POWER TO A PROTECTION SEPARATION GROUP
 
Table 7.7-4, Loss of Power to A Protection Separation Group, analyzes the
 
effects on the control systems caused by the loss of power to a protection
 
separation group. The WCGS power supply is composed of eight inverters. Four inverters power protection separation group 1 through 4, respectively. Two Class 1E swing (backup) inverters (with a DC transfer switch), one per train, are installed to function as a backup for the normal inverters in that train.
One swing inverter and transfer switch can be aligned to replace a separation group 1 or 3 inverter. The other swing inverter and transfer switch can be aligned to replace a separation group 2 or 4 inverter. Two other inverters supply power to control separation group 5 and control separation group 6, discussed later in this section. The control systems affected, the sensors
 
affected, the failure direction, the effect on the control systems, and the
 
bounding USAR accident are given. Where no control action occurs and/or where
 
control action is in a safe direction, no bounding accident is given.
 
Besides the loss of power to a complete control separation group or protection
 
separation group, there is the chance of having an electrical fault on one of
 
the control system circuit cards. The control systems are designed so that
 
each card is used in only one control system. A circuit card failure cannot
 
directly impact more than one control system. A failure on a control card
 
would cause the controller to generate either an "off" or a "full on" output, depending on the type of failure. This result would be similar to having a
 
fault in a sensor feeding the control system. Therefore, the failure of or
 
loss of power in any control system circuit card would be bounded by the Loss
 
of Any Single Instrument analysis described in Table 7.7-3.
 
The analysis is conservative in the sense that, in cases where switches enable
 
the operator to choose from which protection separation group a given signal is
 
desired, it is assumed that the switch is in the position of the failed
 
protection separation group.
 
The table shows that for a loss of power to any protection separation group, the Condition II events analyzed in Chapter 15.0 are bounding.
 
7.7-30    Rev. 29 WOLF CREEK LOSS OF POWER TO CONTROL SEPARATION GROUPS Table 7.7-5 Loss of Power to Control Separation Groups, examines the effects on the control systems caused by the loss of power to control separation groups.
Loss of power to control separation group 5, which powers control groups 1 and 3 (7300 cabinets 5 and 7) is considered, followed by loss of power to control separation group 6 which powers control groups 2 and 4 (7300 cabinets 6 and 8).
The control systems affected, the equipment or signals affected, the failure direction, the effects of the failure, and the bounding accident are given.
The table shows that, for either a loss of power to control separation group 5 or control separation group 6, the resulting failure is bounded by a loss of normal feedwater flow, which is a Condition II event analyzed in the USAR.
BREAK OF COMMON INSTRUMENT LINES Table 7.7-6, Break of Common Instrument Lines, considers the scenario whereby an instrument line which supplies more than one signal ruptures, causing faulty
 
sensor readings.
Three sets of sensors used for control are located in common lines:
: a. Loop steam flow (control separation groups 5 and 6, any steam generator) and narrow range steam generator level (protection separation groups 1 or 2, any steam generator)
: b. Pressurizer level (protection separation groups 1, 2, or
: 3) and pressurizer pressure (protection separation groups
 
1, 2, 3, or 4)
: c. T cold and T hot (any loop)
Not shown on the table since they are not part of the plant control system but are used just for protection are the loop flow transmitters. There are three
 
flow transmitters in each loop with each transmitter having a common high
 
pressure tap but separate and unique low pressure taps. Therefore, a break at
 
the high pressure flow transmitter tap would result in disabling all three flow
 
transmitters in one loop, resulting in a low flow reading for all three
 
transmitters. This would result in a reactor trip if the plant is above the P-
 
8 setpoint, or an annunciation if it is below P-8.      7.7-31    Rev. 14 WOLF CREEK The only malfunction mode explicitly analyzed was a break in the common instrument line at the tap. Another possibility is to have a complete blockage
 
in the sensor tap, causing the sensor to read a constant (before blockage) value. However, this failure mode is not analyzed, since it is really not a credible event. There is no anticipated agent available that would cause a tap
 
blockage. The reactor coolant system piping and fittings and the instrument impulse line tubing are all stainless steel, so no products of corrosion are expected. Also, the water chemistry is of high quality which, along with high
 
temperature operation, precludes the presence of solids in the water and ensures the maintenance of the solubility of chemicals in the water. In
 
addition, prior to startup, and during any shutdown as well, it is routine maintenance and servicing practice for instrument lines to be blown down to a canister. Since the buildup of sludge is a slow process, any buildup would be detected during response time testing done during shutdown. Therefore, the hypothesis of the presence of a complete blockage of the sensor tap is not
 
sufficiently credible to warrant its consideration as a design basis.
In the extremely unlikely event that a complete instrument line blockage were to occur, the condition is detectable because the reading would become static (no variations over time). In an unblocked channel, a reading would always
 
vary somewhat due to noise (i.e. flow induced noise in flow channels) or slight
 
controller action (i.e. cycling operation of spray and heaters in pressurizer).
 
By a comparison of the static channel to the redundant unblocked channels, the
 
operator would be informed that a blockage in one channel has occurred.
Table 7.7-6 indicates that, even in the event of an instrument line break which supplies more than one control signal, the resulting failures are bounded by
 
the Chapter 15.0 analyses.
CONCLUSIONS The preceding tables have illustrated that failures of individual sensors, losses of power to protection separation groups or to control separation
 
groups, or breaks in common instrument lines all results in events which are
 
bounded by Chapter 15.0 analyses. Therefore, the USAR adequately bounds the
 
consequences of these fundamental failures.      7.7-32    Rev. 4 WOLF CREEK 7.7.2.1  Separation of Protection and Control System In some cases, it is advantageous to employ control signals derived from individual protection channels through isolation amplifiers contained in the protection channel. As such, a failure in the control circuitry does not adversely affect the protection channel. Test results have shown that a short circuit or the application (credible fault voltage from within the cabinets) of 120 Volt ac or 140 Volt dc on the isolated output portion of the circuit (nonprotection side of the circuit) will not affect the input (protection) side of the circuit.
Where a single random failure can cause a control system action that results in a generating station condition requiring protective action and can also prevent
 
proper action of a protection system channel designed to protect against the
 
condition, the remaining redundant protection channels are capable of providing
 
the protective action even when degraded by a second random failure. This
 
meets the applicable requirements of Section 4.7 of IEEE Standard 279-1971.
The pressurizer pressure channels needed to derive the control signals are electrically isolated from control.
7.7.2.2  Response Considerations of Reactivity Reactor shutdown with control rods is completely independent of the control functions since the trip breakers interrupt power to the CRDMs, regardless of
 
existing control signals. The design is such that the system can withstand
 
accidental withdrawal of control groups or unplanned dilution of soluble boron
 
without exceeding acceptable fuel design limits. The design meets the
 
requirements of GDC-25.
No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single RCCA from the partially inserted bank at full power operation. The operator could deliberately withdraw a
 
single RCCA in the control bank; this feature is necessary in order to retrieve
 
a rod, should one be accidentally dropped. In the extremely unlikely event of
 
simultaneous electrical failures which could result in single RCCA withdrawal, rod deviation would be displayed on the plant annunciator, and the individual
 
rod position readouts would indicate the relative positions of the rods in the bank. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, would result in activation of the same alarm and the same visual indications.      7.7-33    Rev. 4 WOLF CREEK Each bank of control and shutdown rods in the system is divided into two groups (group 1 and group 2) of up to four or five mechanisms each. The rods
 
comprising a group operate in parallel through multiplexing thyristors. The two groups in a bank move sequentially so that the first group is always within one step of the second group in the bank. The group 1 and group 2 power
 
circuits are installed in different cabinets, as shown in Figure 7.7-14, which also shows that one group is always within one step (5/8 inch) of the other group. A definite schedule of actuation or deactuation of the stationary
 
gripper, moveable gripper, and lift coils of a mechanism is required to withdraw the RCCA attached to the mechanism.
Since the stationary grippers, moveable gripper, and lift coils associated with the RCCAs of a rod group are driven in parallel, any single failure which could cause rod withdrawal would affect a minimum of one group of RCCAs. Mechanical failures are in the direction of insertion, or immobility.
Figure 7.7-15 illustrates the design features that ensure that no single electrical failure could cause the accidental withdrawal of a single RCCA from
 
the partially inserted bank at full power operation.
Figure 7.7-15 shows the typical parallel connections on the lift, movable, and stationary coils for a group of rods. Since single failures in the stationary
 
or movable circuits will result in dropping or preventing rod (or rods) motion, the discussion of single failure will be addressed to the lift coil circuits:
: 1) due to the method of wiring the pulse transformers which fire the lift coil multiplex thyristors, three of the four thyristors in a rod group could remain turned off when required to fire, if for example the gate signal lead failed
 
open at point X
: 1. Upon "up" demand, one rod in group 1 and four rods in group 2 would withdraw. A second failure at point X 2 in the group 2 circuit is required to withdraw one RCCA; 2) timing circuit failures will affect the four
 
mechanisms of a group or the eight mechanisms of the bank and will not cause a single rod withdrawal; and 3) more than two simultaneous component failures are required (other than the open wire failures) to allow withdrawal of a single
 
rod.The identified multiple failure involving the least number of components consists of open circuit failure of the proper two out of 16 wires connected to
 
the gate of the lift coil thyristors. The probability of open wire (or
 
terminal) failure is 0.016 x 10
-6 per hour by MIL-HDB-217A. These wire failures would have to be accompanied by failure, or disregard, of the
 
indications mentioned above. The probability of this occurrence is, therefore, too low to have any significance.      7.7-34    Rev. 4 WOLF CREEK Concerning the human element, to erroneously withdraw a single RCCA, the
 
operator would have to improperly set the bank selector switch, the lift coil
 
disconnect switches, and the in-hold-out switch. In addition, the three
 
indications would have to be disregarded or ineffective. Such series of errors would require a complete lack of understanding and administrative control. A
 
probability cannot be assigned to a series of errors such as these.
 
The rod position indication system provides direct visual displays of each
 
control rod assembly position. The plant computer alarms for deviation of rods
 
from their banks. In addition, a rod insertion limit monitor provides an
 
audible and visual alarm to warn the operator of an approach to an abnormal
 
condition due to dilution. The low-low insertion limit alarm alerts the
 
operator to verify shutdown margin or initiate boration. The facility reactivity control systems are such that fuel damage limits are not exceeded even in the event of a single malfunction of either system.
 
An important feature of the control rod system is that insertion is provided by
 
gravity fall of the rods.
 
In all analyses involving reactor trip, the single, highest worth RCCA is
 
postulated to remain untripped in its full out position.
 
One means of detecting a stuck control rod assembly is available from the
 
actual rod position information displayed on the control board. The control
 
board position readouts, one for each rod, give the plant operator the actual
 
position of the rod in steps. The indications are grouped by banks (e.g.,
control bank A, control bank B, etc.) to indicate to the operator the deviation
 
of one rod with respect to other rods in a bank. This serves as a means to identify rod deviation.
 
The plant computer monitors the actual position of all rods. Should a rod in a
 
control bank be misaligned from the other rods in that bank by more than 12
 
steps, or when any shutdown rod is below the full-out position by 18 steps
 
except during normal shutdown bank withdrawal or insertion, or on update of rod
 
bank position on the plant computer, the rod deviation alarm is actuated.
Misaligned RCCAs are also detected and alarmed in the control room via the flux tilt monitoring system, which is independent of the plant computer.
 
Isolated signals derived from the nuclear instrumentation system are compared
 
with one another to determine if a preset amount of deviation of average power level has occurred. Should such a deviation occur, the comparator output
 
operates a bistable unit to actuate a control board annunciator. This alarm
 
alerts the operator to a power imbalance caused by a misaligned rod. By use of
 
individual rod position readouts, the operator can determine 
 
7.7-35    Rev. 22 WOLF CREEK the deviating control rod and take corrective action. The design of the plant control systems meets the requirements of GDC-23.
Refer to Section 4.3 for additional information on response considerations due to reactivity.
7.7.2.3  Step Load Changes Without Steam Dump The plant control system restores equilibrium conditions, without a trip, following a plus or minus 10-percent step change in load demand, over the 15-
 
to 100-percent power range for automatic control. Steam dump is blocked for
 
load decrease less than or equal to 10 percent. A load demand greater than
 
full power is prohibited by the turbine control load limit devices.
The plant control system minimizes the reactor coolant average temperature deviation during the transient within a given value and restores average
 
temperature to the programmed setpoint. Excessive pressurizer pressure variations are prevented by using spray and heaters and power relief valves in
 
the pressurizer.
The control system must limit nuclear power overshoot to acceptable values, following a 10-percent increase in load to 100 percent.
7.7.2.4  Loading and Unloading Ramp loading and unloading of 5 percent per minute can be accepted over the 15-to 100-percent power range under automatic control without tripping the plant.
 
The function of the control system is to maintain the coolant average
 
temperature as a function of turbine-generator load.
The coolant average temperature increases during loading and causes a continuous insurge to the pressurizer as a result of coolant expansion. The
 
sprays limit the resulting pressure increase. Conversely, as the coolant
 
average temperature is decreasing during unloading, there is a continuous outsurge from the pressurizer resulting from coolant contraction. The
 
pressurizer heaters limit the resulting system pressure decrease. The
 
pressurizer water level is programmed such that the water level is above the
 
setpoint for heater cut out during the loading and unloading transients. The
 
primary concern during loading is to limit the overshoot in nuclear power and
 
to provide sufficient margin in the overtemperature DT setpoint.      7.7-36    Rev. 4 WOLF CREEK The automatic load controls are designed to adjust the unit generation to match load requirements within the limits of the unit capability and licensed rating.
During rapid loading transients, a drop in reactor coolant temperature is sometimes used to increase core power. This mode of operation is applied when the control rods are not inserted deep enough into the core to supply all the reactivity requirements of the rapid load increase (the boron control system is relatively ineffective for rapid power changes). The reduction in temperature
 
is initiated by continued turbine loading past the point where the control rods are completely withdrawn from the core. The temperature drop is recovered and
 
nominal conditions restored by a boron dilution operation.
Excessive drops in coolant temperature are prevented by interlock C-16. This interlock circuit monitors the auctioneered low coolant temperature indications and the programmed reference temperature, which is a function of turbine
 
impulse pressure and causes a turbine loading stop when the decreased
 
temperature reaches the setpoints.
The core axial power distribution is controlled during the reduced temperature return to power by placing the control rods in the manual mode when the operating limits are approached. Placing the rods in manual stops further
 
changes in  , and it also initiates the required drop in coolant temperature. Normally, power distribution control is not required during a rapid power increase, and the rods proceed, under the automatic rod control
 
system, to the top of the core. The bite position is reestablished at the end
 
of the transient by decreasing the coolant boron concentration.
7.7.2.5  Load Rejection Furnished By Steam Dump System When a load rejection occurs, if the difference between the required temperature setpoint of the RCS and the actual average temperature exceeds a predetermined amount, a signal actuates the steam dump to maintain the RCS temperature within control range until a new equilibrium condition is reached.
The reactor power is reduced at a rate consistent with the capability of the rod control system. Reduction of the reactor power is automatic. The steam
 
dump flow reduction is as fast as RCCAs are capable of inserting negative
 
reactivity.      7.7-37    Rev. 4 WOLF CREEK The rod control system can then reduce the reactor temperature to a new equilibrium value without causing overtemperature and/or overpressure
 
conditions. The steam dump steam flow capacity to the condenser is 40 percent of full load steam flow at full load steam pressure.
The steam dump flow decreases proportionally as the control rods act to reduce the average coolant temperature. The artificial load is therefore removed as the coolant average temperature is restored to its programmed equilibrium
 
value.The dump valves are modulated by the reactor coolant average temperature signal. The required number of steam dump valves can be tripped quickly to stroke full open or modulate, depending upon the magnitude of the temperature error signal resulting from loss of load.
7.7.2.6  Turbine-Generator Trip With Reactor Trip Whenever the turbine-generator unit trips at an operating power level above 50-percent power, the reactor also trips. The unit is operated with a programmed average temperature as a function of load, with the full load average temperature significantly greater than the equivalent saturation pressure of
 
the steam generator safety valve setpoint. The thermal capacity of the reactor
 
coolant system is greater than that of the secondary system, and because the full load average temperature is greater than the no-load temperature, a heat sink is required to remove heat stored in the reactor coolant to prevent
 
actuation of steam generator safety valves for a trip from full power. This heat sink is provided by the combination of controlled release of steam to the
 
condenser and by makeup of feedwater to the steam generators.
The steam dump system is controlled from the reactor coolant average temperature signal whose setpoint values are programmed as a function of
 
turbine load. Actuation of the steam dump is rapid to prevent actuation of the
 
steam generator safety valves.
With the dump valves open, the average coolant temperature starts to reduce quickly to the no-load setpoint. A direct feedback of temperature acts to
 
proportionally close the valves to minimize the total amount of steam which is
 
bypassed.The feedwater flow is cut off following a reactor trip when the average coolant temperature decreases below a given temperature or when the steam generator
 
water level reaches a given high level.      7.7-38    Rev. 4 WOLF CREEK Additional feedwater makeup is then controlled manually to restore and maintain steam generator water level while assuring that the reactor coolant temperature
 
is at the desired value. Residual heat removal is maintained by the steam header pressure controller (manually selected), which controls the amount of steam flow to the condensers. This controller operates a portion of the same
 
steam dump valves to the condensers, which are used during the initial transient following turbine and reactor trip.
The pressurizer pressure and level fall rapidly during the transient because of coolant contraction. The pressurizer water level is programmed so that the
 
level following the turbine and reactor trip is above the heaters. However, if
 
the heaters become uncovered following the trip, the chemical and volume
 
control system will provide full charging flow to restore water level in the
 
pressurizer. Heaters are then turned on to restore pressurizer pressure to
 
normal.The steam dump and feedwater control systems are designed to prevent the average coolant temperature from falling below the programmed no-load
 
temperature following the trip to ensure adequate reactivity shutdown margin.
7.
 
==7.3  REFERENCES==
: 1. Lipchak, J. B., "Nuclear Instrumentation System," WCAP-8255,      January, 1974.  (For additional background information only.)
: 2. Shopsky, W. E., "Failure Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Control System," WCAP-8976, August 1977.
: 3. Adler, M. R., "AMSAC Generic Design Package," WCAP-10858-P-A Revision 1, July 1987.
: 4. NRC Safety Evaluation Report for WOLF CREEK Compliance with ATWS Rule 10CFR50.62, dated December 16, 1987.
: 5. KMLNRC 86-195, WM 87-0100.
: 6. Burnett, T. W. T., et al., Westinghouse Anticipated Transients  Without Trip Analysis, WCAP-8330, August 1974. 7. Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (USNRC),  ATWS Submittal, NS-TMA-2182, December 1979.      7.7-39    Rev. 13 WOLF CREEK TABLE 7.7-1 PLANT CONTROL SYSTEM INTERLOCKS Designation Derivation Function C-1          1/2 neutron flux (intermediate    Blocks automatic and manual range) above setpoint            control rod withdrawal C-2          1/4 neutron flux (power range)    Blocks automatic and manual above setpoint                    control rod withdrawal C-3          2/4 overtemperature T above      Blocks automatic and manual setpoint                          control rod withdrawal Actuates turbine runback via load reference Defeats remote load dispatching (if remote
 
load dispatching is used)
C-4          2/4 overpower T above            Blocks automatic and manual setpoint                          control rod withdrawal Actuates turbine runback via load reference Defeats remote load dispatching (if remote
 
load dispatching is used)
C-5          1/1 turbine impulse chamber      Defeats remote load pressure below setpoint          dispatching (if remote
 
load dispatching is used)
Blocks automatic control rod withdrawal C-7          1/1 time derivative (absolute    Makes steam dump valves value) of turbine impulse        available for either
 
chamber pressure (decrease        tripping or modulation
 
only) above setpoint C-9          Condenser pressure above          Blocks steam dump to con-setpoint                          denser.
S ee Fig. 7.2-1 (S heet 10)                                                                    Rev. 0 WOLF CREEK TABLE 7.7-1 (S heet 2) Designation Derivation Function C-11          1/1 bank D control rod            Blocks automatic rod position above setpoint          withdrawal C-16          Reduce limit in coolant temp-S tops automatic turbine erature above normal setpoint    loading until condition clears P-4          Reactor trip                      Makes steam dump valves available for either
 
tripping or modulation Absence of P-4                    Blocks steam dump control via plant trip T avg con-                                                troller C-20          2/2 Turbine impulse chamber      Enables AM S AC for:              pressures above 40% equivalent    - Turbine trip Reactor power (S ee S ection        - AFW actuation 7.7.1.11)                          -
S team generator blow down and sample line isolation Rev. 4 WOLFCREEKTABLE7.7-2BORONCONCENTRATIONMEASUREMENTSYSTEMSPECIFICATIONSOperatingConditionsLinevoltage:120Voltac,+10percent,60Hz+1percentPressure:15to225psig(sample)Temperature:70to130F(sample)
Sampleflowrate:0to0.4gpmAmbienttemperature:60to105FRelativehumidity:to95percent Radiationlevels:<2mr/hr@24inchesfromalltank surfacesReadingtime:Variabledependingonboronconcentration.Maximumtimefor5,000ppmisapproximately5
 
minutes.AccuracyBoronparts/millionpartsofwaterAccuracyStandardDeviation0-1,800ppm+10ppm1,800-5,000ppm+1.25percentDrift:lessthan10ppm/weekNOTE1:Theboronconcentrationmeasurementsystemhasbeen abandoned-in-place.Rev.14 WOLF CREEK TABLE 7.7-3 LOSS OF ANY SINGLE INSTRUMENT NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Feedpump        1 per                      o  Feedwater            Lo        FW pump speed increases    No event if pump speed in Discharge      plant                        Control                        if in auto mode.  (FW      manual. New steady st ate Pressure                                                                      control valves close      reached if pump speed and                                                                              due to increased flow      FCV in auto, i.e., pum p                                                                              if in auto mode.)          speed increases and FC V lift                                                                                                          decreases. If pump sp eed in                                                                                                          auto and FCV in manual
,                                                                                                        bounding event is Exce ssive                                                                                                        FW Flow (U S AR 15.1.2).
Hi        FW pump speed decreases    No event if pump speed in                                                                              if in auto mode.  (FW      manual. Other modes r esult                                                                              control valves open        in a decreased FW flow
 
due to decrease flow      over time. Hence, bou nding                                                                              if in auto mode.)          event is Loss of Norma l FW                                                                                                          Flow (U S AR 15.2.7).
S team          1 per                      o  Feedwater            Lo        FW pump speed decreases    No event if pump speed in Header          plant                        Control                        if in auto mode.  (FW      manual. Other modes r e-Pressure                                                                      control valves open        sult in a decreased FW
 
due to decreased flow if  flow over time. Hence
,                                                                              in auto mode.)            bounding event is Loss of                                                                                                          Normal FW Flow (U S AR                                                                                                          15.2.7).
Hi        FW pump speed increases    No event if pump speed in                                                                              if in auto mode.  (FW      manual. New steady st ate                                                                              control valves close due  reached if pump speed and                                                                              to increased flow if in    FCV in auto, i.e., pum p                                                                              auto mode.)                speed increases and FC V                                                                                                        speed in auto and FCV in                                                                                                        manual, bounding event is                                                                                                          Excessive FW Flow (U S AR                                                                                                        15.1.2).
S team          1 per                      o  Feedwater            Lo        FW pump speed decreases    No event if pump speed in Header          plant                        Control                        if in auto mode.  (FW      manual. Other modes r e-Pressure                                                                      control valves open due    sult in a decreased FW o
S team Dump                      to decreased flow if in    flow over time. Hence,                                              (Pressure Mode)                auto mode.)                bounding event is Loss of                                                                                                          Normal FW Flow (U S AR                                                                                                          15.2.7).
Rev. 1 WOLF CREEK TABLE 7.7-3 (S heet 2)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT S team          1 per                                              Hi        FW pump speed increases S team dump in pressure mode Header          plant                                                        if in auto mode.  (FW      at hot standby conditi ons or Pressure                                                                      control valves open due    at very low power. He nce,                                                                              to increased flow if in    dump valves would open for                                                                              auto mode.)  Dump valves  only a very short time
 
open.  (S team dump        until Lo-Lo TAVG is reached.
blocked on Lo-Lo TAVG
 
(P-12).)                  If pump speed is in ma nual,                                                                                                        or if both pump speeds and                                                                                                          FCV are in auto, then this                                                                                                        event is bounded by
 
excessive increase in
 
secondary steam flow (U S AR                                                                                                        15.1.3). If pump spee d in                                                                                                          auto and FCV in manual
,                                                                                                        bounding event is Exce ssive                                                                                                        FW Flow (U S AR 15.1.2) because this results i n                                                                                                        excessive cooling.
Loop            2 per      1 selected    o  Feedwater            Lo        FW pump speed decreases    No event if pump speed and S team          loop        for control      Control                        if in auto mode. FW      FCV in manual. Other m odes Flow                                                                          valves close if in auto    result in decreased FW flow,                                                                              mode.                      and therefore bounding event                                                                                                          is Loss of Normal FW F low                                                                                                        (U S AR 15.2.7).
Hi        FW pump speed increases    No event if pump speed and                                                                              if in auto mode. FW      FCV in manual. Other modes                                                                              valves open if in auto    result in an increased FW                                                                              mode.                      flow, and hence, bound ing                                                                                                        event is Excessive FW Flow                                                                                                        (U S AR 15.1.2).
Loop FW        2 per      1 selected    o  Feedwater            Lo        FW valve opens if in      No event if FW valve i n Flow            loop        for control      Control                        auto mode.                manual. If in auto,                                                                                                          bounding event is Exce ssive                                                                                                        FW Flow (U S AR 15.1.2).
Hi        FW valve closes if in      No event if FW valve i n                                                                              auto mode.                manual. If in auto, bounding event is Loss of                                                                                                          Normal FW Flow (U S AR                                                                                                          15.2.7).
Narrow          4 per      1 selected    o  Feedwater            Lo        FW valve opens if in      No event if FW valve i n Range S team      for control      Control                        auto mode.                manual. If in auto, Level          Generator  I of II                                                                      bounding event is Exce s-                (two avail-                                                                              sive FW Flow (U S AR 15.1.2).
able for control)
Rev. 1 WOLF CREEK TABLE 7.7-3 (S heet 3)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Hi        FW valve closes if in      No event if FW valve i n                                                                              auto mode.                manual. If in auto, bounding event is Loss of                                                                                                          Normal FW Flow (U S AR                                                                                                          15.2.7).
Pressurizer    3 per      I or III      o  Prz. Level            Lo        Charging flow increases. Bounding event is Incr eased Level          plant                        Control                        Heaters turn off (except  Reactor Coolant Invent ory (Control)                                                                    for local control).        (U S AR 15.5.2).
Letdown isolated.  (VCT empties, charging pumps
 
take suction from
 
RW S T.)                                                                    Hi        Charging flow decreases. While heaters are on, no                                                                              Backup heaters on.        net depressurization o f                                                                              (Later, letdown iso-      RC S. After heaters blocked,                                                                              lation from interlock      the decreased charging flow                                                                              channel and heaters        acts to depressurize t he                                                                              blocked from interlock    RC S. Depressurization event channel.)                  is therefore bounded b y                                                                                                        Inadvertent Opening of a                                                                                                          Prz.
S afety or Relief Valve (U
S AR 15.6.1).
Pressurizer    3 per      II or III      o  Prz. Level            Lo        Letdown isolation. Prz. Reach new steady-state with Level          plant                        Control                        heaters blocked (except    high pressurizer level. No (Interlock)                                                                  for local control).        event.
 
(Charging flow control-
 
ler reduces flow to main-
 
tain level).
Hi        No control action, get    Not applicable.
Hi level annunciation.
Pressurizer    4 per      II of IV      o  Prz. Pressure        Lo        No control action. PORV  Not applicable.
Pressure        plant                        Control                        456A blocked from open-
 
ing. PORV 455A opens if
 
required, closes when
 
presusre falls below
 
deadband.)
Hi        PORV 456A opens.  (POR    Bounding event is Inad ver-                                                                              closes when pressure      tent Opening of a Prz.
drops below deadband.)
S afety or Relief Valve (U
S AR 15.6.1).
Rev. 0 WOLF CREEK TABLE 7.7-3 (S heet 4)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Pressurizer    4 per      I or III      o  Prz. Pressure        Lo        Backup heaters on.        Heater on causes incre ase Pressure        plant                        Control S pray remains off.        in prz. pressure to PORV PORV 455A blocked from    456A actuation. No ev ent.                                                                              opening.  (PORV 456A
 
opens if required, closes when pressure
 
falls below deadband.)
Hi        PORV 455A opens.
S pray    Bounding event is Inadver-on. (PORV 455A closes      tent Opening of a Prz.
when pressure drops be-S afety or Relief Valve low deadband).            U S AR 15.6.1).
TAVG            1 per loop  Any            o S team dump            Lo S top turbine loading/      Not applicable.
Auct.          (TAVG Mode)                    Defeat remote dis-
 
patching.  (C-16-Annun-
 
Hi        o  Reactor Control                ciation occurs).
o  Prz. Level Control Auct.      o  Turbine Loading/
Lo            Dispatching Hi        Rods in (safe direc-      No event unless reacto r                                                                              tion). Charging flow      trips, then steam dump increases until full      valves open and this i s                                                                              power prz. Level is      bounded by Excessive
 
reached (if at reduced    Increase in S econdary                                                                              power). If reactor S team Flow (U S AR 15.1.3) trips, steam dump enabled
 
and dump valves open until
 
steam dump stops when
 
Lo-Lo TAVAG (P-12) is
 
reached.)
TAVG            1 per      Any            o S team dump            Lo S top turbine loading/      Not applicable.
loop                          (Pressure                      defeat remote dis-Auct.          Mode)                          patching.  (C-16-Annun-
 
Hi                                            ciation occurs).
 
o  Reactor Control o  Prz. Level Control Auct.      o  Turbine Loading/
Lo            Dispatching Rev. 1 WCG S                                                        TABLE 7.7-3 (S heet 5)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Hi        Rods in (safe direc-      Reach steady-state wit h                                                                              tion). Charging flow      pressurizer at full-
 
increases until full      power level. No event
.                                                                              power prz level is
 
reached (if at reduced
 
power).
S teamline      3 per loop  Control        o S. Gen ARV          Lo        No control action.        Not applicable.
Pressure        for protec- Channel tion, 1 per                                        Hi S. Gen. atmospheric        Result is bounded by loop for                                                      relief valve opens.        Inadvertent Opening control                                                                                  of a S. Gen. Atmospheric (different                                                                              Relief or S afety Valve from those                                                                              (U S AR 15..11.4).
used for protection)
Intermediate    2 per      I or II        o  Reactor Control      Lo        No control action.        Not applicable.
Range Flux      plant
 
Hi        Get reactor trip (during  Not applicable.
 
startup) due to C-1
 
actuation, otherwise
 
no control action.
Turbine        w per      I              o S team Dump                      Rods in (safe direc-      Not applicable.
Impulse        tubine      (Control)        (TAVG Mode)                    tion). Auto rod with-Chamber                                                                      drawal blocked (C-5).
 
Pressure                                  o  Reactor Control      Lo        (If reactor trip occurs, steam dump unblocked and
 
o  FW Control                      dump valves open until no
 
load TAVG is reached.)
No effect on FW control since constant level
 
program.
Hi        Rods out until blocked    Result is bounded by U ncon-                                                                              by Hi flux, overpower,    trolled  Rod Cluster C on-                                                                              or overtemperature rod    trol Assembly Bank Wit h-                                                                              stop, or until pro-        drawal at Power (U S AR)                                                                              grammed TREF limits is    15.4.2).
reached.  (If reactor
 
trip occurs, steam dump
 
unblocked and dump valves
 
open until no load TAVG
 
is reached.)  No effect on
 
FW control since have
 
onstant S.G. level program.
Rev. 13 WCG S                                                        TABLE 7.7-3 (S heet 6)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Turbine        2 per      II            o S team Dump S team dump unlocked.      Not applicable Impulse        turbine    (Interlock)      (TAVG Mode)                    Rods in (safe direc-
 
Chamber                                                                      tion). Auto rod with-
 
Pressure                                  o  Reactor Control      Lo        drawal blocked (C-5).
 
(If reactor trip occurs, o  FW Control                      dump valves open until
 
no load TAVG is reached.)
 
No effect on FW control, since have constant S.G.                                                                              level program.
Hi        Rods out until blocked    Result is bounded by by Hi flux, overpower,    Uncontrolled Rod Clust er                                                                              or overtemperature rod    Control Assembly Bank With-                                                                              stop, or until pro-        drawal at Power (U S AR                                                                              grammed TREF limits is    15.4.2).
 
reached.  (If reactor
 
trip occurs, steam
 
dump occurs, steam
 
dump valves open until
 
no load TAVG is reached.)
 
No effect on FW control, since have constant S.G.                                                                              level program.
Turbine        2 per      I              o S team Dump                      Auto rod withdrawal        Not applicable.
Impulse        turbine    (Control)        (Pr. Mode)                      blocked (C-5). Rods in Chamber                                                                      (safe direction). No
 
Pressure                                  o  Reactor Control      Lo        effecton FW control, since have constant S.G.                                            o  FW Control                      level program.  (If
 
reactor trip occurs, dump
 
valves open to keep
 
steam header pressure at
 
or below setpont.)
Hi        Rods out until blocked    Result is bounded by U n-                                                                              by Hi flux, overpower      controlled Rod Cluster or overtemperature rod    Control Assembly Baak
 
stop or until pro-        Withdrawal at Power
 
grammed TREF is reached    (U S AR 15.4.2).
(If reactor trip occurs, dump valves open to keep
 
steam header pressure at
 
or below setpoint.)  No
 
effect on FW control
 
since have constant S.G.                                                                              level program.
Rev. 0 WCG S                                                        TABLE 7.7-3 (S heet 7)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Turbine        2 per      II            o S team Dump                      Auto rod withdrawal        Not applicable.
Impulse        turbine    (Interlock)      (Pr. Mode)                      blocked (C-5). Rods in
 
Chamber                                                                      (safe direction). No
 
Pressure                                  o  Reactor Control      Lo        effect on FW control
 
since have constant S.G.                                            o  FW Control                      level program.  (If
 
reactor trip occurs, dump
 
valves open to keep steam
 
header pressure at or
 
below setpoint.)
Hi        Rods out until blocked    Result is bounded by U n-                                                                              by Hi flux, overpower or  controlled Rod Cluster overtemperature rod stop  Control Assembly Bank
 
or until porgrammed TREF  Withdrawal at Power
 
is reached. (If reactor    (U S AR 15.4.2).
trip occurs, dump valves
 
open to keep steam head-
 
er pressure at or below
 
setpoint.)  No effect on
 
FW control since have
 
constant S.G. level program.
Power          4 per      Any            o  Reactor Control      Lo        No control action (auc-    Not applicable.
Range          plant                                                        tioneered Hi).
 
Flux                                      o  FW Control Hi        Auto and manual rod        Increased bypass valve withdrawal blocked        opening would be bound ed                                                                              (C-2). Rods in (safe      by Excessive FW flow
 
direction). FW bypass    (U S AR 15.1.2).
valve opens if in
 
auto.  (If reactor trip
 
occurs, dump valves
 
open, until no load
 
TAVE is reached.)
 
Rising S.G. level causes valve to close
 
till steam and feed
 
flows match.
Rev. 0 WCG S                                                        TABLE 7.7-3 (S heet 8)                  NUMBER                                          A SS UMED                    OF        FAILED                              FAILURE                                      BOUNDING S EN S OR          CHANNEL S      CHANNEL S Y S TEM            DIRECTION EFFECT EVENT Condenser      2 per      Any            o S team Dump            Lo        No control action.        Not applicable.
Available      condenser S team dump unblocked,                                                                              i.e., condenser
 
available for steam
 
dump.
Hi        No control action.        Not applicable.
 
S team dump stays blocked, ie., con-
 
denser unavailable for
 
steam dump.
TAVG            1                          o S team Dump            Lo        Charging flow decreases    Result is bounded by Un-(High                                                                        until no-load level        controlled Rod Cluster Auctioneer)                                o  Reactor Control                reached. Rods out        Control Assembly Bank
 
until blocked by Hi        Withdrawal at Power
 
o  Prz. Level                      flux, overpower or        (U S AR 15.4.2).
Control                        overtemperature rod
 
stop.
S team dump blocked (TAVG mode only).
Hi        Identical to TAVG S ee above.
channel failing high,                                                                              see analysis above.
S team Flow      2 per loop  1 selected    o S team Flow            Lo        Identical to loop steam S ee above.
Pressure                    for control                                      flow channel failing low.
Compensator S ee analysis above.
Hi        Identical to loop steam S ee above.
flow channel failing
 
high.
S ee analysis above.
Rev. 0 WCGSTABLE7.7-4LOSSOFPOWERTOAPROTECTIONSEPARATIONGROUPCONTROLEQUIPMENTORSYSTEMSSENSORSFAILUREBOUNDING
 
AFFECTEDAFFECTEDDIRECTION EFFECTS EVENTSteamDumpTurbinePressureLoNocontrolaction.Steam(control)dumpunblocked(pressuremode).S.Gen.ARV(Loop1)remainsSteamlinePressureLoclosed.ReactorControlPowerRangeFluxLoRodsin(safedirection).Powerdecreases.TurbinePressureLoStopturbineloading/defeatremotedispatching.TurbinePressureLo (Interlock)TAVG(Loop1)LoFWControlNarrowRangeLevelLoAllfeedwatervalvesclose.LossofNormalFWFlow(USAR15.2.7)eventisTurbinePressureLoboundingsinceincreasedchargingflow/isolatedFeedwaterControlValvesFcletdownhaslittleeffectrelativetothedecreasedfeedFeedwaterIsolationValvesFcflow.SteamFlowPressureLo CommpensationPrz.LevelPrz.Level(control)LoChargingflowincreases.Heatersblocked.Letdownisolated.(Allactionsoccurifonchannel1).Prz.PressurePrz.Pressure(PORV455A)LoNocontrolaction.PORV455AstaysclosedPORV455AFcSteamDumpTurbinePressureLoNocontrolaction.SteamBoundingeventiseither(Interlock)dumpunblocked(Bothmodes).ExcessiveFWFlow(USARS.Gen.(ARV(Loop2)remains15.1.2)orLossofNormalSteamlinePressureLoclosed.FWFlow(USAR15.2.7),dependingonchannelsused.Thesescenariosareboundingsinceletdownisolationhas littleeffectrelativetotheFW flowevents.Rev.14 WCGSTABLE7.7-4(Sheet2)LOSSOFPOWERTOAPROTECTIONSEPARATIONGROUPCONTROLEQUIPMENTORSYSTEMSSENSORSFAILUREBOUNDING AFFECTEDAFFECTEDDIRECTION EFFECTS EVENTReactorControlPowerRangeFluxLoRodsin(safedirection).Powerdecreases.TurbinePressureLoStopturbineloading/defeatremoteTurbinePressureLodispatching.(Interlock)TAVG(Loop2)LoWControlNarrowRangeLevelLoIfaffectedlevelsignalusedforcontrol,FCVopensinTurbinePressureaffectedloop,andFWflowin-creases(overridessteamflowSteamFlowPressureLosignal).Otherwise,channelnotCompensationconnected,getdecreasedFWflowinloopswithfailedsteamflowpressurecompensationonly.Noeffectonremainingloops.Prz.LevelPrz.LevelLoLetdownisolated.(Interlock)Heatersblocked.(Ifonchannel2).Prz.PressurePrz.Pressure(PORV456A)LoNocontrolaction.PORV456Astaysclosed.PORV456AFcSteamDumpSteamlinePressureLoNocontrolaction.Stopturbineloading/defeatremotePowerRangeFluxLodispatching.S.Gen.ARV(Loop3)staysclosed.ReactorControlTAVG(Loop3)LoRev.13 WCGSTABLE7.7-4(Sheet3)LOSSOFPOWERTOAPROTECTIONSEPARATIONGROUPCONTROLEQUIPMENTORSYSTEMSSENSORSFAILUREBOUNDING AFFECTEDAFFECTEDDIRECTION EFFECTS EVENTPrz.LevelPrz.LevelLoChargingflowincreases.Combiningeffectsof(Control)Heatersblocked.pressurizerlevelandLetdownisolated.pressurecontrolsystems,or(Ifonchannel3).couldhaveeitherin-creasingchargingflowPrz.LevelLoorwithheateroffcausinga (Interlock)depressurization,orelseheaterscausepressuretoLetdownisolated.increaseuntilPORV456AisHeatersblocked.actuated.Eitherway,eventis(Ifonchannel3).boundedbyInadvertentOpeningofaPressurizerSafetyor ReliefValve(USAR15.6.1).Prz.PressurePrz.PressureIfattectedpressuresignal(PROV455A)Lousedforcontrol,PORV455Astaysclosed,backupheaterson(ifallowbylevelsignal, seeabove)andsprayoff.SteamdumpSteamlinePressureLoNocontrolaction.S.Gen.ARV(Loop4)remainsclosed.ReactorControlPowerRangeFluxLoStopturbineloading/defeatremotedispatching.TAVG(Loop4)LoPrz.PressurePrz.PressureLoNocontrolaction.(PORV456A0PORV456Astaysclosed.FWControlFWControlValvesFcFeedwatervalvesclose.BoundingeventisLossofNormalFWFlowFWIsolationValvesFc(USAR15.2.7).Rev.14 WCGSTABLE7.7-5LOSSOFPOWERTOACONTROLSEPARATIONGROUPCONTROLEQUIPMENTORSYSTEMSSENSORSFAILUREITEMIZEDBOUNDING
 
AFFECTEDAFFECTEDDIRECTION EFFECTS EVENTSteamDumpCondenserAvailableHiNocontrolaction.Steamdumpstaysblocked.S.G.HeaderPressureLoDumpValvesFailclosedFWControlSteamFlowLo(S.G.1and3)FSFlowLo(S.G.1and3)FWControlValvesFailClosedLossofFWflowLossofnormalFWflow.(S.G.1and3)USAR15.2.7.S.G.HeaderPressureLo FWDischargePressureLo FeedPumpsCoastdownPressurizerPrz.Pressure(PORV455A)LoPORV455A PressureStaysclosed.SteamDumpAuctioneeredLoNocontrolaction.
TAVGReactorControlAuctioneeredTAVGLoRodsstationary.StopturbineControlrodsloading/defeatremotedis-patching(C-16).Annunciation
 
occurs.FWControlSteamFlowLo(S.G.2and4)FWFlowLo(S.G.2and4)FWControlValvesFailclosedLossofFWflowLossofnormalFWflow.(S.G.2and4)USAR15.2.7.Rev.14 WCGSTABLE7.7-5(SHEET2)LOSSOFPOWERTOAPROTECTIONSEPARATIONGROUPCONTROLEQUIPMENTORSYSTEMSSENSORSFAILUREITEMIZEDBOUNDING
 
AFFECTEDAFFECTEDDIRECTION EFFECTS EVENTPresurizerLevelAuctioneeredTAVGLoPrz.HeatersFoffRCSinventoryremainsrelatively constant.ChargingControlValveFopenChargingPumpCoastdownLetdownIsolationValvesFailclosedPressurizerPrz.Pressure(PORV456A)LoPORV456A Pressurestaysclosed.Rev.0 WCGS TABLE 7.7-6 LOSS OF POWER TO A PROTECTION SEPARATION GROUP FAILED                            FAILURE                                            BOUNDING SENSORS                CHANNELS SYSTEM DIRECTION EFFECTS EVENT Loop Steam Flow        I or II      o  Feedwater            Lo          FW valve closes in affected    Bounding event is Los s and Narrow Range                      Control              Hi          S.G. Pump speed decreases.      of Normal FW Flow (US AR Level                                                                                                    15.1.2).
Pressurizer Level      I            o  Prz. Level            Hi          Charging flow decreases.        This is a depressuriz a-(Interlock or Control)                Control                            (Control) Backup heaters on. tion event which is (Control) (On low level,        bounded by Inadverten t  and                                                                    letdown isolated and heaters    Opening of a Prz. Saf ety                                                                          blocked from interlock          or Relief Valve
 
channel.                        (USAR 15.6.1).
Pressurizer Pressure  I            o  Prz. Pressure        Lo          PORV 455A stays closed.(Either PORV)                          Control Pressurizer Level      II          o  Prz. Level            Hi          No control action.              Not applicable.(Interlock or Control) and Pressurizer Pressure  II          o  Prz. Pressure        Lo          PORV 456A stays closed.(Either PORV)
Pressurizer Level      III          o  Prz. Level            Hi          Charging flow decreases and    Depending on switch p osi-(Interlock or Control)                Control                            backup heaters on if on        tion, this event is a t                                                                          control channel. On low        most a depressurizati on  and                                                                    level, letdown isolated and    event which is bounde d by                                                                          heaters blocked form inter-    Inadvaertent Opening of                                                                          lock channel. No control      a Prz. Safety or Reli ef                                                                          action if on interlock          Valve (USAR 15.6.1).
channel.
Pressurizer Pressure  III and IV  o  Prz. Pressure        Lo          Either PORV (or neither).(Either PORV)                          Control                            stays closed.
Tcold and/or Thot      I, II, III,  o  Steam Dump            Lo          See failure of TAVG in "Loss or IV                                              of any Single Instrument" o  Reactor Control      or          Table 1 o  Prz. Level Control    Hi Rev. 0 WOLF CREEK THOT LEI TCOlD LEG THOT LEG TCOlD LEG THOT LEG TCOLD LEG THOT LEG TCOLD LEG AVERAGE TEMPERATURE UNIT LOOP I TH+Tc TAVG =-2-AVERAGE TEMPERATURE UNIT LOOP 2 -TH+Tc TAVG --2-AVERAGE TEMPERATURE UNIT LOOP 3 -T H+Tc TAVG --2-AVERAGE TEMPERATURE UNIT
-TH-t-TC TAVG --2-I AUCTIONEER UNITI TURBINE LOAD '----------
.... 1 HIGHEST TAVG I. SIGNAL TO STEAM ( HIGHEST TAVG NUCLEAR J ._ ________ ......_ ___ __.., TO PRESSURIZER I DUMP SYSTEM LOAD SIGNAL LEVEL PROGRAMMER t f t ,____. __ LEAD-LAG POWER MISMATCH COMPENSA Tl ON UNIT AVERAGE TEMPERATURE PROGRAMMER
+ HOTES: 1. TEMPERATURES ARE MEASURED AT STEAM GENERATOR'S INLET
... COMPENSATION UNIT I ' j ROD SPEED MANUAL. ROD -CONTROL UN IT SEQUEIITI AL ROO CONTROl UNIT (AUTOMATIC CONTROL) f PERMISSIVE CIRCUIT ROO DRIVE POWER REACTOR TRIP BREAKER I REDugq:r TRIP SIGNAL (ROD INTERLOCK)
.. CONTROL ROD ACTUATOR lr CvNTROL ROD DRIVE MECHAM ISM REACTOR TRIP ...,_ _ __. BREAKER 2 ROD DRIVE POWER WOLF CREEK Rev. 0 UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-1 SIMPLIFIED BLDCK DIAGRAM OF REACTOR SYSTEM WOLF CREEK LOW ALARM LOW-LOW ALARM COMPARATOR (6T) AUCT DEMAND BANK SIGNAL z _____ ___. TYPICAL OF ONE CONTROL BANK NOTE: I. ANALOG CIRCUITRY IS USED FOR THE COMPARATOR NETWORK 2. COMPARISON IS DONE FOR ALL CONTROL BANKS COMMON FOR ALL FOUR CONTROL BANKS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-2 CONTROL BANK ROD INSERTION MONITOR WOLF CREEK DEMAND BANK SIGNAL (ROD CONTROL) ___ __, INDIVIDUAL ROD POSITION READING OF THOSE RODS CLASSIFIED AS MEMBERS OF THAT BANK ALARM COMPARATOR NOTE: I. DIGITAL OR ANALOG SIGNALS MAY BE USED FOR THE COMPARATOR COMPUTER INPUTS. 2. THE COMPARATOR WILL EHERGIZE THE ALARM IF THERE EXISTS A POSITION DIFFERENCE GREATER THAN A PRESEt LiMIT BETWEEN ANY. INDIVIDUA-L ROD AND THE DEMAND BANK SIGNAL. * ***********-*-*--
Rev. 0
: 3. COMPARISON IS INDIVIDUALLY D()NE FOR ALL CONTROL BANKS. WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-3 ROD DEVIATION COMPARATOR 
'1 WOLF CREEK PRESSURIZER PRESSURE Sl GNAL POWER RELIEF POWER VALVE #I RELIEF VALVE #2 REFERENCE PRESSURE PID CONTROLLER TO BACKUP TO VARIABLE HEATER HEATER CONTROL CONTROL SPRAY CONTROLLER I I REMOTE MANUAL POSITIONING (TYPICAL-SEPERATE CONTROLLER FOR EACH SPRAY VALVE) PID-PROPORTIONAL+
INTEGRAL+
DERIVATIVE Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPOl FIGURE 7.7-4 BLOCK DIAGRAM OF PRESSURIZE PRESSURE CONTROL SYSTEM HEATER CONTROL CHARGING FLOW SIGNAL I WOLF CREEK PRESSURIZER WATER LEVEL SIGNAL PtD CONTROLLER PI CONTROLLER REMOTE CONTROL CHARGING FLOW CONTROL VALVE POSITIOK
* AUCTIONEERED TAVG LEVEL PROGRAMMER WOLF CREEK Rev. 0 PID-SEE FIG. 7.7-4 UPDATED SAFETY REPORT I FIG U R E 7 7-5 BLOCK DIAGRAM OF LEVEL CONTROL SYSTEM I STEAM FLOW SIGNAL WOLF CREEK LEVEL PROGRAMMER SET AT 50% FEEDWATER FL()I SIGNAL STEAM GENERATOR WATER LEVEL SIGNAL FILTER PI CONTROLLER PI CONTROLL&#xa3;R REMOTE MANUAL POSIT ION I NG PI CONTROLLER MA IN FEEDWA TE R CONTROL VALVE DYNAMICS MAIN FEEOWATER CONTROL VALVE POSITION PI -PROPORTIONAL
+ INTEGRAL POWER RANGE NEUTRON FLUX GAIN + FEEDWATER BYPASS VALVE DYNAMICS FEEDWATER BYPASS VALVE POSITION WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-6 BLOCK DIAGRAM OF STEAM GENERATOR WATER LEVEL CONTROL SYSTEM Rev. 0 
'\ WOLF CREEK MAIH FEEDWATER PUMP DISCHARGE PRESSURE TOTAL PLAHT STEAM FLOW I I STEAM HEADER PRESSURE _____ .,.. REMOTE MANUAL POSITIONING REMOTE MANUAL POSITIONING PI* SEE FIG. 7.7-6 PI COM lRSLLER ! '-------ADJUSTABLE t:i' AT HO-LOAO SETPOINT PROP CONTROLLER (TYPICAL-EACH PUMP HAS ITS OWN PROP COHTROLLER)
MAIN FEEDWATEn SPEED Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-7 BLOCK DIAGRAM oF-MAIN FEEDWATER PUMP SPEED CONTROL SYSTEM WOLF CREEK STEAM DUMP CONTROL IN MANUAL (STEAM PRESSURE CONTROL) TURBINE IMPULSE STAGE PRESSURE RATE/ LAG COMPENSATION REACTOR TRIP LOAD REJECT I ON BISTABLE STEAM PRESSURE P-4-DEFEAT LOAD REJECTION STEAM DUMP CONTROL; ALLOW PLANT TRIP STEAM DUMP CONTROL SET PRESSURE PLAin TR'I P PI CONTROLLER LOAD REJECTION CONTROL OR PLANT TRIP CONTROtl.
TAVG NO-LOAD AUCTIONEERED TAVG LEAD/LAG COMPENSATION REFERENCE f AVG BISTABLE$
Bl STABLES .. LOAD REJECTION CONTROLLER LOAD REJECT I ON. CONTROL OR PLAIT TRIP CONTROL TRIP OPEN STEAM DUMP VALVES MANUAL (STEAM PRESSURE CONTROL) AUTO (TAVG CONTROL) NOTE: FOR BLOCKING, BLOCKING SIGNAL TO CONDENSER STEAM DUMP VALVES SEE FIGURE 7.2-1 SHEET IO --..-AIR SUPPLY TO DUMP VALVES MODULATE CONDENSER llUMl IJ.ALV&#xa3;5..
UPDATED BLOCK Rev. 0 WOLF CREEK SAFETY ANALYSIS REPORT FIGURE 7.7-8 DIAGRAM OF STEAM DUMP CONTROL SYSTEM WOLF CREEK LIMIT SWHCHE INTERCONNECTING TUBING FLUX THIMBLES SAFETY SWITCHES TRANSFERS
[ Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-9 BASIC FLUX-MAPPING SYSTEM HS NUT, c: "'0 c m en c )> en s::., )>:i "'0-;;lo fnCil
:;p;; )>m )>0 en..., z:::o en* m'";'l )>m :s:-r-m aJQ c;; :::0 m "'0 0 f -1/2" UNION -SWAGELOK SOURCE TUBE SOURCE PLUG 16" I lf .........
;7: .. ,. .... 7-.. 11 ::::::::::. . . :::: .. -; ; ;; ; :
SAMPLER COVER PLATE #6 SCREW, HELICOIL PLCS) ----TOP PLUG ASSY. i/2" TEE. SWAGELOK #6 SCREW (4 PLCS) SOURCE I ASSY. POLY SHIELD HOLD DOWN GUSSETTS {q PLCS) Jq" NOTE: The boron concentration measurement system has been abandoned-in-place. 0 t"4 "tJ n ::tl l.zJ l.zJ WOLF CREEK 1/2" SWAGELOK UNION T 7-1/2" 1 1/8" SST -----TOP PLUG ASSY. 1/2" SWAGELOK TEE G-2C-5 BF3 DETECTOR ANNULUS ASSY. ALL STAINLESS NOTE: The boron concentration measurement system has been abandoned-in-place.
Rev.14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7M11 SAMPLER SUBASSEMBLY "tt ;tJ 0 ("') m en en )> en"" en_ mCi') 3:c: m;;o rm -<....., m* r-:' ()N ;;s;: c > Ci') !: c: ""0 c m c en rHo
)>0 z::u )>m r-m -<" en u; ::u m ""0 0 ::u ...a. ol:lo THUMBWHEEL SWITCH TC DIGITAL THERMOMETER r
L kEYPAD I I I I I DISPLAY POWER +/- 5 SUPPLY +/- 12 L +/- 15 ,---1.--I SBC-519 DIGITAL 1/0 CMOS RAM PREAMP I *I COUNTER/TIMER HV POWER SUPPLY * .,. H 1/LO . l I c-ALARM BUS SBC-80/IOA MICROCMUTER DATEL 0/A BUFFER BUFFER l l ANALOG TO OUTPUT REMOTE DISPLAY AND/OR PRINTER NOTE: The boron concentration measurement system has been abandoned-in-place. 0 t"' tJ:j Q tzJ tzJ ::00:
l z Cl: Q Cl: Cl: W,j WOLF CREEK '" 12 tO 8 6 " 2 0 *2 -I+ 8 *10 -12 200 "00 600 800 I 000 1200 IIWO I 600 I 800. 2000 BORON CONCENTRATION IN PPM NOTE: The boron concentration monitoring system has been abandoned-in-place.
Rev.14 WOLFCREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7*13 BORON CONCENTRATION MONITORING SYSTEM LINEARITY CURVE OVER NORMAL PLANT OPERATING RANGE OF BORON CONCENTRATIONS
(/) H 3: -u I H ...,., H ('")f"T'l
.,., 00 H z G'> -leo c: ::0 I ::0 00 m r-C":l :;;>;: ... '-..! (/.) . -<o 'J C/> H I -j)> ....... m G'> ...t:" 3::0 )> 3 0 ...,., ::a 0 0 SLAVE CYCLER I BD REACTOR H H MASTER H CONTROL PULSER SYSTEM CYCLER SLAVE I CYCLER MANUAL 2 BD SWITCH BANK BANK SELECTOR OVERLAP MULTIPLEX Cl RCUITS a I" t .. , ___ LIFTING} ; I I L1 OFF GROUP I .,g '1 L t/2...., 11LIFTING}
.. "' ( I I I GROUP 2 i I I ----------OFF 22 1-4 m tU 1-i :;Q (t) < . 0 NORMAL SEQUENCING OF GROUPS WITHIN BANK POWER CABINET BANK D I BD GROUP I LIFT COIL DISCONNECT SWITCHES POWER CONTROL CABINET BANK D 2 BD GROUP 2 I . I NOTE: ONLY CABI!tETS lBO AND 2BD SHOWN. FOR MORE COMPLETE DIAGRAM CLUDING POWER CABINETS lAC, 2AC, SCDE, AND DC HOLD SCD, SEE REF. I IN SECTION 7.7.3 i"%j 0 ?<:
STATIONARY GRIPPER COILS t()VABLE GRIPPER COILS WOLF CREEK i20 VAC CONTROL BANK D GROUP I POWER CABINET IBD LIFT COIL o DISCONNECT SWITCHES CONTROL BANK D GROUP 2 POWER CAB I NET 280 LIFT COILS LIFT COILS MULTIPLEX THYRISTORS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 7.7-15 CONTROL BANK 0 PARTIAL SIMPLIFIED SCHEMATIC DIAGRAM OF POWER CABINETS 180 AND 280 WOLF CREEK APPENDIX 7A COMPARISON TO REGULATORY GUIDE 1.97 7A.1  INTRODUCTION This appendix provides an evaluation of the instrumentation to assess plant and environs conditions following an accident. The plant instrumentation and
 
features provided at WCGS have resulted from detailed design evaluations and
 
reviews. Design features that enable the plant to be taken to cold shutdown
 
while utilizing only safety-grade equipment are described in Section 7.4, "Systems Required for Safe Shutdown". Chapter 18.0 provides a comparison of the WCGS design to the requirements of NUREG-0737.
Since most of the instrumentation in the WCGS was purchased and installed prior to the issuance of Regulatory Guide 1.97, strict compliance to the many prescriptive recommendations is not provided in all cases. However, the WCGS
 
instrumentation and control room design is adequate to allow the operators to evaluate and mitigate the consequences of postulated accidents.
This appendix provides a detailed comparison of the WCGS design to the recommendations contained in the regulatory guide.
7A.2  ORGANIZATION
 
The text of this appendix provides a summary description of the bases for the WCGS instrumentation design as they relate to the recommendations of the
 
regulatory guide. The tables provide the data necessary to perform a detailed
 
comparison of the WCGS design with the recommendations of the regulatory guide.
Table 7A-1 is a cross-reference between Table 2 of the regulatory guide and the information presented in this appendix. Table 7A-1 lists the variables in the same sequence in which they appear in the regulatory guide table, assigns variable identification numbers, and identifies the data sheet upon which the
 
detailed comparison with the WCGS design has been provided.
Table 7A-2 provides a summary of the WCGS design to the recommendations of the regulatory guide. This table also serves as an index to the data sheets in Table 7A-3.
Table 7A-3 consists of individual data sheets. One data sheet is provided for each variable or group of related variables identified in Table 2 of the
 
regulatory guide. The data sheet contains 7A-1                      Rev. 19 WOLF CREEK the recommended range, category, and purpose for the variable and includes the multiple listing requirements. A discussion is provided of the WCGS plant design bases for ranges, qualification, etc., and other pertinent data which support the adequacy of the current design or describe design modifications which are being implemented. Table 7A-3 also provides an indication of the computer into which the variable is inputted and thereby made available to the plant computer network.
7A.3  WCGS DESIGN BASIS COMPARISON TO REGULATORY GUIDE 1.97 The WCGS design bases are stated throughout the USAR. The discussions provided below summarize the WCGS design bases as they pertain to the salient recommendations of the regulatory guide. Appropriate references to other USAR sections are provided in Table 7A-3 for more detailed information. The discussions below are intended to aid the review of the WCGS design bases for
 
compliance with the intent of the regulatory guide recommendations.
7A.3.1  TYPE A VARIABLES
 
Variables classified as Type A for the WCGS design are identified in Table 7A-
: 2. The reason for the classification is provided on the corresponding data sheet in Table 7A-3.
The following criteria are the bases for identification of Type A variables for the WCGS. The terminology used in the discussion is consistent with that of the generic Emergency Response Guidelines (ERGs) for Westinghouse plants, which were submitted to the NRC by Westinghouse Owners Group letter WOG-64, dated November 30, 1981.
: a. Variables used for event diagnosis are classified as Type A because these variables direct the operator to the
 
appropriate Optimal Recovery Guidelines (formerly termed
 
Emergency Operating Instructions) or to monitoring of
 
critical Safety Functions.
: b. Variables used by the operator to perform manual actions prescribed by the Optimal Recovery Guidelines, which are associated with Condition IV events (LOCA, MSLB, and
 
SGTR), are classified as Type A. Condition I, II and III
 
events are not considered in identifying Type A variables
 
(e.g., Spurious Safety Injection).
7A-2                      Rev. 11 WOLF CREEK
: c. Variables which identify the need for operator action to correct single failures are not classified as Type A.
These actions are often identified as "Notes" or "Contingency Actions" in the ERGs.
: d. Variables associated with operator actions required for events not currently in the design bases of the plant are not identified as Type A variables.
7A.3.2  REDUNDANCY AND DIVERSITY FOR CATEGORY 1 VARIABLES
 
The following discussion summarizes salient points of the WCGS design with respect to the regulatory recommendations:
: a. Adequate redundancy is considered to exist when adequate information is available to the operator to make appropriate decisions, assuming a single failure. This
 
is done on a system, loop, or component basis, as appropriate. For the steam generator heat sink function
 
and pressurizer, it was done on a component basis. For
 
the reactor and reactor coolant loops, it was done on a
 
system basis due to the abundance of diverse or
 
associated variables which are available to indicate the
 
nature of the event and identify its cause.
: b. Diverse variables are considered to be those which vary directly with or have a direct relation with the primary
 
variable. Associated variables are those which, when considered with the primary and/or diverse variables, aid in the identification and evaluation of events and the
 
status of the plant.
: c. The need for a third reading or a diverse variable is based on the control room operators' need for the identification of the proper recovery from an event.
 
Diversity is not provided solely for TSC/EOF use, accident reconstruction, or range not associated with
 
DBEs.
: d. Since the need for a diverse variable arises upon the single failure of the primary instrumentation and that failure must result in ambiguity (e.g., the instrument fails in midscale, not offscale high or low), diverse variables may be performance or commercial grade. Many diverse variables on WCGS are qualified as Class 1E for reasons other than their diversity function.
: e. Items identified as diverse variables are not considered to be part of the post-accident  monitoring data 7A-3                      Rev. 1 WOLF CREEK base and are not included in the Emergency Response Facility Data Base solely for that purpose. Many diverse variables are part of the post-accident monitoring data base because of their primary function. Since it is highly unlikely that a variable would be required for a diversity function, the EOF/TSC may contact the control room should the need arise.
7A.3.3  RECORDERS Dedicated recorders are required only where trend information is immediately required for operator use. The current value (indicated) of the PAMs variables
 
is normally used by the operator for decision-making purposes. Where Class 1E indicators are provided, recorders may be performance grade.
7A.3.4  INSTRUMENT RANGES
 
Instrument ranges have been determined, considering the function(s) of the sensed parameters. The installed instrumentation may meet the ranges recommended in the regulatory guide, meet the intent of the recommended range, or have a range appropriate for the design function. Instrumentation that has
 
an appropriate range is identified on Table 7A-2. The ranges are justified on
 
the individual data sheets of Table 7A-3.
7A.3.5  UNNECESSARY VARIABLES
 
Several variables listed in the regulatory guide are not necessary for post-accident monitoring for the WCGS. Table 7A-2 identifies which variables are
 
considered unnecessary from a post-accident monitoring standpoint, and the
 
individual data sheets provide a discussion justifying the determination.
7A.3.6  QUALIFICATION FOR CATEGORY 1 PARAMETERS
 
Tables 7A-2 and 7A-3 show that instrumentation for all variables designated as Category 1 by the NRC and those designated as Type A herein are qualified as Class 1E from the sensor to the indicator.
Qualification of these devices is described in Section 3.11(B). All Class 1E equipment is qualified in accordance with Regulatory Guide 1.89, and Regulatory Guide 1.100 as discussed in Appendix 3A.
7A-4                      Rev. 1 WOLF CREEK 7A.3.7  QUALIFICATION FOR CATEGORY 2 PARAMETERS The WCGS design utilizes Class 1E and non-Class 1E sensors, transmitters, indicators, and power sources. There is no qualification category between these two categories, as implied by the Category 2 terminology of the regulatory guide.
Table 7A-2 shows that many of the Category 2 items are in fact fully qualified to Class 1E environmental and seismic requirements. These items exceed the regulatory recommendations.
The non-Class 1E instruments are termed performance grade. These items are purchased to perform in their anticipated service environments for the plant conditions in which they must function. The regulatory guide implies that they
 
must function in the accident environment for the area in which they are
 
located without consideration of the design function. If an instrument has to
 
function following an accident, it is fully qualified to Class 1E requirements.
If the instrument is not required following an accident, it is termed non-safety-related and purchased to performance grade requirements. The equipment service conditions are provided in the purchase specification and include
 
radiation levels and integrated doses, temperature, relative humidity, and
 
other special considerations. The current qualification levels for each item
 
reflect its importance to safety. Table 7A-3 addresses the function of performance grade items in Category 2.
Non-Class 1E equipment is supplied from Separation Groups 5 and 6, which are highly reliable (refer to Section 8.3.1.3). The non-Class 1E 125 V dc buses are backed by the emergency diesel generators.
For the purpose of compliance to the regulatory requirements for seismic
 
qualification for items identified as Category 2, the sensors/transmitters
 
continued operation is not assumed to be required, since the indicators need
 
not be qualified. Assurance of pressure boundary integrity during and after
 
seismic events is ensured for safety-related systems. No seismic requirements are placed on items in non-safety-related systems.
7A.3.8  QUALIFICATION FOR CATEGORY 3 ITEMS The Category 3 qualification guidelines of the regulatory guide imply a possible need to ensure that the instrument sensor and transmitter are qualified for an accident environment. Table 7A-2 identifies those Category 3
 
instruments located inside the containment, and the appropriate data sheet of Table 7A-3 justifies the lack of post-accident qualification.
7A-5                      Rev. 1 WOLF CREEK TABLE 7A-1 REGULATORY GUIDE 1.97 VARIABLE LIST DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
B.1            Reactivity Control B.1.1          Neutron Flux                                    1.1 B.1.2          Control Rod Position                            1.2
 
B.1.3          RCS Soluble Boron Concentration                13.1 B.1.4          RCS Cold Leg Water Temperature                  2.1 B.2            Core Cooling B.2.1          RCS Hot Leg Water Temperature                  2.2 B.2.2          RCS Cold Leg Water Temperature                  2.1
 
B.2.3          RCS Pressure                                    2.3 B.2.4          Core Exit Temperature                          1.3 B.2.5          Coolant Level in Reactor                        1.4
 
B.2.6          Degrees of Subcooling                          1.5
 
B.3            Maintaining Reactor Coolant System Integrity B.3.1          RCS Pressure                                    2.3 B.3.2          Containment Sump Water Level                    6.2
 
B.3.3          Containment Pressure                            6.1 B.4            Maintaining Containment Integrity B.4.1          Containment Isolation Valve Position            6.3 (excluding check valves)
B.4.2          Containment Pressure                            6.1
 
C.1            Fuel Cladding C.1.1          Core Exit Temperature                          1.3 Rev. 0 WOLF CREEK TABLE 7A-1 (Sheet 2)
DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
C.1.2          Radioactivity Concentration or Radiation      13.3 Level in Circulating Primary Coolant C.1.3          Analysis of Primary Coolant                    13.1 (gamma spectrum)
C.2            Reactor Coolant Pressure Boundary C.2.1          RCS Pressure                                    2.3 C.2.2          Containment Pressure                            6.1
 
C.2.3          Containment Sump Water Level                    6.2 C.2.4          Containment Area Radiation                    11.1 C.2.5          Effluent Radioactivity - Noble Gas            12.2 Effluent from Condenser Air Removal System Exhaust C.3            Containment
 
C.3.1          RCS Pressure                                    2.3 C.3.2          Containment Hydrogen Concentration              6.4 C.3.3          Containment Pressure                            6.1 C.3.4          Containment Effluent Radioactivity -          12.1 Noble Gases from Identified Release Points C.3.5          Radiation Exposure Rate (inside build-        11.2 ing or areas, e.g., auxiliary building,              reactor shield building annulus, and fuel handling building, which are in
 
direct contact with primary containment
 
where penetrations and hatches are located)
C.3.6          Effluent Radioactivity - Noble Gases          12.1 (from buildings as indicated above)
D.1            Residual Heat Removal (RHR) or Decay Heat Removal System D.1.1          RHR System Flow                                3.1 D.1.2          RHR Heat Exchanger Outlet Temperature          3.1 Rev. 0 WOLF CREEK TABLE 7A-1 (Sheet 3)
 
DATA
 
VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.
VARIABLE SHEET NO.
D.2            Safety Injection Systems
 
D.2.1          Accumulator Tank Level and Pressure            3.2
 
D.2.2          Accumulator Isolation Valve Position            3.2
 
D.2.3          Boric Acid Charging Flow                        3.3
 
D.2.4          Flow in HPI System                              3.3
 
D.2.5          Flow in LPI System                              3.1
 
D.2.6          Refueling Water Storage Tank Level              3.4
 
D.3            Primary Coolant System D.3.1          Reactor Coolant Pump Status                    2.4
 
D.3.2          Primary System Safety Relief Valve              2.5 Positions (including PORV and code
 
valves) or Flow Through or Pressure
 
in Relief Valve Lines
 
D.3.3          Pressurizer Level                              2.6
 
D.3.4          Pressurizer Heater Status                      2.7
 
D.3.5          Quench Tank Level                              2.8
 
D.3.6          Quench Tank Temperature                        2.8
 
D.3.7          Quench Tank Pressure                            2.8
 
D.4            Secondary System (Steam Generator)
D.4.1          Steam Generator Level                          4.1
 
D.4.2          Steam Generator Pressure                        4.2
 
D.4.3          Safety/Relief Valve Positions or Main          4.3
 
Steam Flow
 
D.4.4          Main Feedwater Flow                            4.4
 
Rev. 0 WOLF CREEK TABLE 7A-1 (Sheet 4)
DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
D.5            Auxiliary Feedwater or Emergency Feedwater System D.5.1          Auxiliary or Emergency Feedwater Flow          5.1 D.5.2          Condensate Storage Tank Water Level            5.2
 
D.6            Containment Cooling Systems D.6.1          Containment Spray Flow                        10.1
 
D.6.2          Heat Removal by the Containment Fan Heat        8.1 Removal System D.6.3          Containment Atmosphere Temperature              6.5 D.6.4          Containment Sump Water Temperature              6.6
 
D.7            Chemical and Volume Control System D.7.1          Makeup Flow-In                                  7.1 D.7.2          Letdown Flow-Out                                7.1
 
D.7.3          Volume Control Tank Level                      7.1 D.8            Cooling Water System D.8.1          Component Cooling Water Temperature to          9.1 ESF System D.8.2          Component Cooling Water Flow to                9.1 ESF System D.9            Radwaste System D.9.1          High-Level Radioactive Liquid Tank Level      14.1 D.9.2          Radioactive Gas Holdup Tank Pressure          14.2
 
D.10          Ventilation Systems D.10.1        Emergency Ventilation Damper Position          15.1 Rev. 0 WOLF CREEK TABLE 7A-1 (Sheet 5)
DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
D.11          Power Supplies D.11.1        Status of Standby Power and Other Energy      16.1, 16.2 Sources Important to Safety (hydraulic,              pneumatic)
E.1            Containment Radiation E.1.1          Containment Area Radiation - High Range        11.1 E.2            Area Radiation E.2.1          Radiation Exposure Rate (inside build-        11.2 ings or areas where access is required to service equipment important to safety)
E.3            Airborne Radioactive Materials Released from Plant E.3.1          Noble Gases and Vent Flow Rate E.3.1.1        o  Containment or Purge Effluent              12.1
 
E.3.1.2        o  Reactor Shield Building Annulus            NA (if in design)
E.3.1.3        o  Auxiliary Building (including any          12.1 building containing primary system
 
gases, e.g., waste gas decay tank)
E.3.1.4        o  Condenser Air Removal System Exhaust        12.2
 
E.3.1.5        o  Common Plant Vent or Multipurpose          12.1 Vent Discharging Any of Above Releases (if containment purge is included)
E.3.1.6        o  Vent From Steam Generator Safety            12.3 Valves or Atmospheric Relief Valves E.3.1.7        o  All Other Identified Release Points        12.4
 
E.3.2          Particulates and Halogens Rev. 13 WOLF CREEK TABLE 7A-1 (Sheet 6)
DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
E.3.2.1        o  All Identified Plant Release Points        12.5 (except steam generator safety valves or atmospheric relief valves and condenser air removal
 
system exhaust). Sampling with
 
Onsite Analysis Capability E.4            Environs Radiation and Radioactivity E.4.1          Radiation Exposure Meters (continuous          17.1 indication at fixed locations)
E.4.2          Airborne Radiohalogens and Particulates        17.2 (portable sampling with onsite analysis capability)
E.4.3          Plant and Environs Radiation (portable        17.3 instrumentation)
E.4.4          Plant and Environs Radioactivity              17.4 (portable instrumentation)
E.5            Meteorology E.5.1          Wind Direction                                17.5 E.5.2          Wind Speed                                    17.5
 
E.5.3          Estimation of Atmospheric Stability            17.5 E.6            Accident Sampling Capability (Analysis Capability on Site)
E.6.1          Primary Coolant                                13.1 E.6.1.1        o  Gross Activity                              13.1
 
E.6.1.2        o  Gamma Spectrum                              13.1 E.6.1.3        o  Boron Content                              13.1 E.6.1.4        o  Chloride Content (1)                          13.1 E.6.1.5        o  Dissolved Hydrogen or Total Gas            13.1 E.6.1.6        o  Dissolved Oxygen                            13.1 Rev. 13 WOLF CREEK TABLE 7A-1 (Sheet 7)
DATA VARIABLE                                                   
 
==SUMMARY==
 
IDENT. NO.                  VARIABLE SHEET NO.
E.6.1.7        o  pH                                          13.1 E.6.2          Sump                                          13.2
 
E.6.2.1        o  Gross Activity                              13.2 E.6.2.2        o  Gamma Spectrum                              13.2 E.6.2.3        o  Boron Content                              13.2
 
E.6.2.4        o  Chloride Content (1)                          13.2 E.6.2.5        o  pH                                          13.2 E.6.3          Containment Air
 
E.6.3.1        o  Hydrogen Content                            6.4 E.6.3.2        o  Oxygen Content                              13.1 E.6.3.3        o  Gamma Spectrum                              13.1 (1)The analysis can be performed on site if dose rates allow, or by an off site facility contracted to provide results within four days.
Rev. 11 WOLF CREEK TABLE 7A-2
 
==SUMMARY==
COMPARISON TO REGULATORY GUIDE 1.97 SENSOR          CHANNEL RANGE COMPARISON LOCATION QUALIFICATION NRC
 
DATA                                      QUAL.      WCGS        Complies          Appro-
 
SHEET              VARIABLE                CATE-      TYPE A        with    Meets  priate  Inside  Outside  Class  Perf.
 
NUMBER            DESCRIPTION            GORY      VARIABLE      Reg. Intent  Range    Ctmt    Ctmt      1E    Grade CORE AND REACTOR VESSEL VARIABLES 1.1  Neutron Flux                          1                      X                          X                X      X 1.2  Control Rod Position                  3                      X                          X                        X
 
1.3  Core Exit Temperature                1                      X                          X                X
 
1.4  Reactor Vessel Level                  1                      X                          X                X
 
1.5  Subcooling Monitor                    2                      X                          X RCS AND RELATED VARIABLES
 
2.1  RCS T cold                            1        Yes          X**                        X                X 2.2  RCS T hot                              1        Yes          X**                        X                X 2.3  RCS Pressure                          1        Yes          X                          X                X 2.4  RCP Status (motor current)            3                      X                                  X                X
 
2.5  Primary System Safety Relief Valve    2                      X                          X                X
 
Position
 
2.6  Pressurizer Level                    1        Yes                    X                X                X
 
2.7  Pressurizer Heater Status            2                      X                                          X      X
 
2.8  PRT Level                            3                      X                          X                        X
 
2.8  PRT Temperature                      3                                        X        X                        X
 
2.8  PRT Pressure                          3                      X                          X                        X ECCS VARIABLES
 
3.1  RHR/LPI Flow Rate                    2                      X                                  X                X 3.1  RHR/Heat Exchanger T out              2                                        X                X                X 3.2  Accumulator Tank Level                2                      NA*                      X                        X
 
3.2  Accumulator Tank Pressure            2                      NA*                      X                        X
 
3.2  Accumulator Isolation Valve          2                      X                        X                X
 
Position
 
3.3  Centrifugal Charging Pump Flow        2                      X                                  X        X
 
3.3  Safety Injection Pump Flow            2                      X                                  X                X
 
3.3  RCP Seal Injection Flow              2                      X                                  X        X
 
3.4  RWST Level                            2        Yes          X                                  X        X
* Unnecessary variables - refer to Table 7A-3
** Complies with range recommended in Revision 3 of Regulatory Guide 1.97 Rev. 2 WOLF CREEK TABLE 7A-2 (Sheet 2)
SENSOR          CHANNEL RANGE COMPARISON LOCATION QUALIFICATION NRC
 
DATA                                      QUAL.      WCGS        Complies          Appro-
 
SHEET              VARIABLE                CATE-      TYPE A        with    Meets  priate  Inside  Outside  Class  Perf.
 
NUMBER            DESCRIPTION            GORY      VARIABLE      Reg. Intent  Range    Ctmt    Ctmt      1E    Grade SECONDARY SIDE VARIABLES 4.1  Steam Generator Level- Wide          1                                X                X                X Range
 
4.1  Steam Generator Level - Narrow        1        Yes          NA                        X                X
 
Range
 
4.2  Steam Line Pressure                  1        Yes                    X                        X        X
 
4.3  Secondary Side ARV Position          2                      X                                  X        X 4.3  Secondary Side Safety Valve          2                      NA                                NA
 
Position
 
4.4  Main Feedwater Flow Rate              3                      X                                  X                X AUXILIARY FEEDWATER SYSTEM VARIABLES
 
5.1  Auxiliary Feedwater Flow Rate        2                      X                                  X        X 5.2  Condensate Storage Tank Level        1                      X                                  X        X
 
(Pressure)
CONTAINMENT VARIABLES
 
6.1  Containment Pressure - Design        1        Yes          X                          X                X Pressure Range
 
6.1  Containment Pressure - Extended      1                      X                          X                X
 
Range
 
6.2  Containment Normal Sump Level        1        Yes          X                          X                X
 
6.2  Containment RHR Sump Level            1                      X                          X                X
 
6.3  Containment Isolation Valve          1                      X                          X      X        X
 
Position
 
6.4  Containment Hydrogen Concentration    1        Yes          X                                  X        X 6.5  Containment Atmosphere Temperature    2                      X                          X                X
 
6.6  Containment Sump Temperature          2                      NA*
CHARGING AND LETDOWN SYSTEM VARIABLES
 
7.1  Normal Charging Flow                  2                                X                        X                X 7.1  Normal Letdown Flow                  2                                                  X                X
 
7.1  Volume Control Tank Level            2                                X                        X        X
 
7.1  Letdown Flow - Safety Related        2                      X                          X                X
* Unnecessary Variable - Refer to Table 7A-3 Rev. 11 WOLF CREEK TABLE 7A-2 (Sheet 3)
SENSOR          CHANNEL RANGE COMPARISON LOCATION QUALIFICATION NRC
 
DATA                                      QUAL.      WCGS        Complies          Appro-
 
SHEET              VARIABLE                CATE-      TYPE A        with    Meets  priate  Inside  Outside  Class  Perf.
 
NUMBER            DESCRIPTION            GORY      VARIABLE      Reg. Intent  Range    Ctmt    Ctmt      1E    Grade CONTAINMENT COOLING SYSTEM VARIABLES 8.1  Containment Cooler Heat Removal      2                      NA*
 
COMPONENT COOLING WATER SYSTEM VARIABLES
 
9.1  Component Cooling Water              2                      X                                  X                X Temperature to ESF
 
9.1  Component Cooling Water              2                      X                                  X                X
 
Flow Rate to ESF CONTAINMENT SPRAY SYSTEM VARIABLES
 
10.1  Containment Spray Flow Rate          2                                X                        X                X
 
AREA RADIATION MONITORING 1
 
11.1  Containment Area Radiation            1        Yes          X                          X                X 11.2  Area Radiation Monitor-              2                      NA*
 
Containment Penetrations
 
Hatches and Areas Important
 
to Safety EFFLUENT MONITORS
 
12.1  Unit Vent - Noble Gas                2                      X                                  X                X 12.2  Condensate Air Removal -              3                      X                                  X                X
 
Radiation Monitor
 
12.3  Secondary Side Radiation Release      2                                X                        X                X
 
12.4  AFW Turbine Radiation Release2                                X                          X                X
 
12.5  Vent Particulates and Halogens        3                      X                                  X                X
* Unnecessary Variable - Refer to Table 7A-3
 
Rev. 1 WOLF CREEK TABLE 7A-2 (Sheet 4)
SENSOR          CHANNEL RANGE COMPARISON LOCATION QUALIFICATION NRC
 
DATA                                      QUAL.      WCGS        Complies          Appro-
 
SHEET              VARIABLE                CATE-      TYPE A        with    Meets  priate  Inside  Outside  Class  Perf.
 
NUMBER            DESCRIPTION            GORY      VARIABLE      Reg. Intent  Range    Ctmt    Ctmt      1E    Grade SAMPLING SYSTEMS 13.1  Post-Accident Sampling System        3                      NA*
13.2  Containment Recirculation            3                      X                                  X                X Sump Sample
 
13.2  ECCS Room Sump Sample                3                      NA*
 
13.2  Auxiliary Building Sump Sample        3                      NA*
 
13.3  Radiation Level in RCS                1                      NA*
RADWASTE SYSTEM VARIABLES
 
14.1  Recycle Holdup Tank Level            3                      NA*
14.2  Waste Gas Decay Tank Pressure        3                      NA*
DAMPER POSITION
 
15.1  Emergency Ventilation Damper          2                      X                          X      X        X Position POWER SUPPLY STATUS INDICATION
 
16.1  Electric Power Supply Status          2                      X                                  X        X 16.2  Gas Accumulator Tank Pressure        2                      X                                  X                X ENVIRONMENTAL MONITORING
 
17.1  Fixed Radiation Exposure Meters      3                      NA*
17.2  Portable Emergency Monitor -          3                      X                                  X                X
 
Particulates and Halogen
 
17.3  Particulates Monitor - Plant          3                      X                                  X                X
 
and Environs
 
17.4  Plant and Environs - Gamma Spectra    3                      X                                  X                X
 
17.5  Meteorological Parameters            3                      X                                  X                X
* Unnecessary Variable - Refer to Table 7A-3
 
Rev. 20 WOLF CREEK TABLE 7A-3, DATA SHEET 1.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE                CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.1.1    Neutron Flux              10
-6% to 100% full power            1        Function detection, accomplishment of mitigation
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT 
 
IDENT. NO. VARIABLE            RANGE              SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.1.1        Neutron Flux    10
-8 to 200% power      SENE60      Y        020      Y        -        -      NPIS SENE61      Y        020      Y        020        Y      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. Redundant Class 1E neutron flux monitors, independent from the protection system, have been added to the WCGS design.
These monitors meet the stated recommendations.
 
Rev. 21
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 1.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.          VARIABLE                  RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.1.2        Control Rod Position      Full in or not full in      3                  Verification
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.      VARIABLE              RANGE            SENSOR/TRANSMITTER          CONTROL ROOM                  COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.1.2    Control Rod Positon    Full in to full out    SF0074        N      022        N  NPIS 53 rods
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets the stated recommentations.
: 2. WCGS has 53 full-length control rods arranged in four banks (A through D), and each bank is divided into two groups.
 
Each group consists of several assemblies which move together.
: 3. The rod position monitoring is performed by two separate systems:  (1) the digital rod position indication system and
 
(2) a demand position system. The position of each rod is indicated on a dedicated LED. These systems are described
 
in USAR Section 7.7.1.3.2.
 
Rev. 13
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 1.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE                CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.2.4    Core Exit Temperature 1    200&deg;F to 2300&deg;F (for operating    3 3        Verification plants - 200&deg;F to 1650F)
 
C.1.1    Core Exit Temperature 1    200&deg;F to 2300&deg;F (for operating    1 3        Detection of potential for breach plants - 200&deg;F to 1650&deg;F)                    accomplishment of mitigation, long-
 
term surveillance
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT 
 
IDENT. NO. VARIABLE                RANGE          SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.2.4    Core Exit Temperature    200 - 2300&deg;F      TE-1 through  Y      RP081A, B  Y        RP081A, B  Y      NPIS
 
C.1.1                                              TE-50
 
(47 total)
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets the stated recommentations.
: 2. All 47 thermocouples are qualified to Class 1E requirements and provide inputs to the subcooling monitor described on data sheet 1.5.
: 3. All 47 thermocouples are indicated and recorded on qualified devices in the control room. Diversity is not required due to extensive redundancy provided.
 
Rev. 21
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 1.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE              CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.2.5    Coolant Level in Reactor Bottom of core to Top of vessel      1        Verification, accomplishment of mitigation
 
(direct in-
 
dicating or
 
recording de-
 
vice not re-
 
quired)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO. VARIABLE                RANGE          SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.2.5    Reactor Vessel Water  Bottom to top of    LT 1311      Y      021      Y        080        Y      NPIS
 
Level                  vessel              LT 1312      Y      021      Y        080        Y      NPIS
 
LT 1321      Y      021      Y        -        -      NPIS
 
LT 1322      Y      021      Y        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations.
: 2. The WCGS RV level indication system will provide information on the RV water level with or without the RC pumps in
 
operation. This Class 1E system will utilize two pressure taps to cover the range from the bottom of the vessel to
 
the top of the vessel.
: 3. The design includes four indicating devices which provide redundancy (two devices) for the two design conditions.
: 4. Diversity is provided by the core exit thermocouples described on data sheet 1.3.
 
Rev. 11 
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 1.5
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE                CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.2.6    Degrees of Subcooling          200&deg;F subcooling              2            Verification of analysis and plant
 
to 35&deg;F superheat        (With con-        conditions
 
firmatory
 
operator
 
procedures)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO. VARIABLE              RANGE            SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.2.6    Subcooling Monitor    200&deg;F subcooled to    RP081A        Y      022 Y          -        -      NPIS                                2,000&deg;F superheat      RP081B        Y      022 Y          -        -      NPIS
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. WCGS subcooling moniter meets all of the stated recommentations.
: 2. The subcooling monitor design provisions are described in Section 18.2.13.2. The system is Class 1E and fully
 
qualified.
: 3. Diversity is not required, since this system is considered to be Category 2 per the regulatory recommendations;
 
however, extensive redundancy in the inputs is provided to ensure system reliability.
: 4. This system could be utilizied by the plant operators following an event; however, it is not considered a Type A
 
variable, since the operator is able to perform subcooling calculations, using existing instrumentation.
 
Rev. 13
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.1.4      RCS Cold Leg Water        50&deg;F to 400&deg;F            3            Verification
 
Temperature 1
B.2.2      RCS Cold Leg Water        50&deg;F to 750&deg;F*          1        Function detection, accomplishment of mitigation, Temperature 1                                                verification, long-term surveillance
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.1.4        RCS Temperature      0-700&deg;F        TE-413B      Y        021      Y        022        N        NPIS
 
B.2.2        Wide Range T Cold      0-700&deg;F        TE-423B      Y        021      Y        022        N        NPIS 0-700&deg;F        TE-433B      Y        -        -        022        N        NPIS
 
0-700&deg;F        TE-443B      Y        -        -        022        N        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The RCS wide-range T cold instruments are Class 1E and powered from Protection Sets I and II. Protection Set I instruments are indicated separately on a qualified indicator. The T cold and T hot readings for each loop are recorded on a dual pen recorder.
: 2. The existing range meets the recommended range of Revision 3 of Regulatory Guide 1.97. Other associated variables are
 
available to help ensure that the operator is aware of primary system parameters.
: 3. Diversity is not required due to the extensive redundancy provided; however, the operator can use the steam line
 
pressure of the associated steam generator to estimate the T cold readings. T cold will trend with T sat for each steam generator. Associated variables which provide useful information include T hot and T cold and the core exit temperatures.
: 4. This parameter is a Type A variable, and it is used throughout the EOIs.
* Revision 3 of Regulatory Guide 1.97 revised the range to 50&deg;F to 700&deg;F. Thus, the existing range now meets the
 
regulatory recommendation.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.2.1      RCS Hot Leg Water        50&deg;F to 750&deg;F*          1            Function detection, accomplishment of
 
Temperature                                                    mitigation, verification, long-term sur-
 
veillance
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.2.1        RCS Temperature      0-700&deg;F        TE-413A      Y        021      Y        022        N        NPIS
 
Wide Range T hot      0-700&deg;F        TE-423A      Y        021      Y        022        N        NPIS 0-700&deg;F        TE-433A      Y        -        -        022        N        NPIS
 
0-700&deg;F        TE-443A      Y        -        -        022        N        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The RCS wide-range T hot instruments are Class 1E and powered from Protection Sets I and II. Protection Set I instruments are indicated separately on a qualified indicator. As noted on data sheet 2.1, T hot is recorded with T cold of the same loop on a dual pen recorder.
: 2. The existing range meets the recommended range of Revision 3 of Regulatory guide 1.97.
: 3. Diversity is not required due to the extensive redundancy provided; however, the operator could use the core exit
 
thermocouples as a diverse measurement. Refer to data sheet 1.3.
: 4. This parameter is a Type A variable, and it is used throughout the EOIs.
 
*Revision 3 of Regulatory Guide 1.97 revised the range to 50&deg;F to 700&deg;F. Thus, the existing range now meets the regulatory
 
recommendations.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.2.3        RCS Pressure 1      0-3,000 psig (4,000 psig      1 2          Function detection, accomplishment of for CE plants)                            mitigation, verification, long-term sur-
 
veillance
 
B.3.1        RCS Pressure 1      0-3,000 psig (4,000 psig      1 2          Function detection, accomplishment of for CE plants)                            mitigation
 
C.2.1        RCS Pressure 1      0-3,000 psig (4,000 psig      1 2          Detection of potential or actual breach,                                  for CE plants)                            accomplishment of mitigation, long-term
 
surveillance
 
C.3.1        RCS Pressure 1      0-3,000 psig (4,000 psig      1 2          Detection of potential for breach, accom-for CE plants)                            plishment of mitigation.
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.          VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.2.3    RCS Pressure          0-3,000 psig        PT-405      Y        022      Y        022        N        NPIS
 
B.3.1                            0-3,000 psig        PT-403      Y        022      Y        022        N        NPIS
 
C.2.1                            0-3,000 psig        PT-406      Y        002      Y        -        -        -
C.3.1
 
NA        Pressurizer Pressure  1,700 to2,500 psig  PT-455      Y        002      N        022        N        NPIS
 
PT-456      Y        002      N        PR 455-Select      NPIS
 
PT-457      Y        002      N        1 of 4              NPIS
 
PT-458      Y        002      N                            NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The RCS pressure instruments meet all of the stated requirements.
: 2. RCS pressure is a Type A variable, and is used throughout the EOIs.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.3.1      Reactor Coolant Pump      Motor Current          3              To monitor operation
 
Status
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.3.1      Reactor Coolant Pump    0-600A          CT-PA0107    N        021      N        -        -        NPIS
 
Motor Current          0-600A          CT-PA0108    N        021      N        -        -        NPIS
 
0-600A          CT-PA0204    N        021      N        -        -        NPIS
 
0-600A          CT-PA0205    N        021      N        -        -        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets the stated recommendations.
 
Rev. 11 
 
WOLF CREEK WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.5
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.3.2    Primary System Safety    Closed-not closed            2            Operation status, to monitor for loss
 
Relief Valve Positions                                            of coolant
 
(including PORV and code
 
valves) or Flow Through
 
or Pressure in Relief
 
Valve Lines     
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.3.2    PORV Position          Closed-not closed  ZS-455A      Y      021        Y        -          -      NPIS
 
ZS-456A      Y      021        Y        -          -      NPIS
 
D.3.2    PORV Block            Closed-not closed  HIS-8000A    Y      021        Y        -          -      NPIS
 
Valve Position                            HIS-8000B    Y      021        Y        -          -      NPIS
 
D.3.2    Safety Valve Position  Closed-not closed  ZS-8010A      Y      021        Y        -          -      NPIS
 
ZS-8010B      Y      021        Y        -          -      NPIS
 
ZS-8010C      Y      021        Y        -          -      NPIS
 
____________________________________________________________________________________________________________________________
 
_
III. REMARKS
: 1. The WCGS design meets the stated recommendations. Section 18.2.6.2 provides mor information on these items.
: 2. Since the WCGS design provides position monitoring of the subject valves, the flow through or pressure in the
 
discharge lines to the PRT is not provided.
: 3. Diversity is not required, since this is an NRC Category 2 variable. However, the PRT parameters described on
 
data sheet 2.8 are available.
 
Rev. 11 
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.6
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                      PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.3.3      Pressurizer Level        Bottom to top            1            To ensure proper operation of pressurizer
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO. VARIABLE                RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.3.3    Pressurizer Level      Bottom to top of    LT-459      Y        002      Y        002              NPIS
 
straight shell      LT-460      Y        002      Y        Select 1 of 3    NPIS
 
LT-461      Y        002      Y                          NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The range covered meets the intent of the recommended range. Approximately 85 percent of the total volume is
 
covered. Monitoring level in the hemispherical heads is not advisable, since the volume-to-level ratio is not linear.
: 2. This is a Type A variable, and is used throughout the EOIs for operator action.
: 3. Diversity is not required due to the extensive redundancy provided.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.7
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.3.4      Pressurizer Heater      Electric current          2            To determine operating status
 
Status
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.3.4      Pressurizer Heater      0-300A          CT-NB0106    Y        015      Y        -        -        NPIS
 
Current                0-300A          CT-NB0208    Y        015      Y        -        -        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets the stated recommendations.
: 2. Diversity is not required, since this is an NRC Category 2 variable.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 2.8
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.3.5      Quench Tank Level          Top to bottom              3              To monitor operation
 
D.3.6      Quench Tank Temperature    50&deg;F to 750&deg;F              3              To monitor operation
 
D.3.7      Quench Tank Pressure        0 to design pressure 4      3              To monitor operation
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT 
 
IDENT. NO. VARIABLE                RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER 
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.3.5    Pressurizer Relief Tank    Top to bottom    LT-470      N        021      N        -        -      NPIS
 
Level
 
D.3.6    Relief Tank Temperature    50 to 350        TE-468      N        021      N        -        -      NPIS
 
D.3.7    Relief Tank Pressure      0-100 psig        PT-469      N        021      N        -        -      NPIS
 
(design)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The PRT is a horizontal, cylindrical tank. The level is measured for 100 of the 114-inch tank diameter, which is
 
essentially top to bottom.
: 2. The PRT temperature range is adequate to monitor any expected conditions in the tank. The PRT design pressure is 100
 
psig (Tsat = 327.8&deg;F), and the rupture disc release pressure is 91 psig, nominal. Following breach of the disc, the
 
temperature of the tank cannot exceed the saturation temperature associated with the existing contaiment pressure.
: 3. The PRT parameters are available in the NPIS computer; therefore, it is not necessary to provide a dedicated recorder.
: 4. Although these instruments are located inside the containment, they are not qualified for post-accident conditions, e
 
they are not required following a LOCA or MSLB. Primary and secondary loop parameters, as well as containment
 
parameters, are available to allow the operator to determine the nature and course of the accident. The EOIs do not
 
indicate any use of these parameters following an event. Refer to Section 7A.3.8.
 
Rev. 21 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                        PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.1.1      RHR System Flow      0 to 110% design flow 10      2              To monitor operation
 
D.1.2      RHR Heat Exchanger  32&deg;F* to 350&deg;F                2              To monitor operation and for analysis
 
Outlet Temperature
 
D.2.5      Flow in LPI System  0 to 115% design flow 10      2              To monitor operation
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.1.1    RHR/LPI-Inj./Recirc.      0-114%        FT-618      N        017      N        018        N        NPIS
 
Cold Leg                                FT-619      N        017      N        018        N        NPIS
 
D.2.5    LPI - Hot Leg Recircu-    0-169%        FT-988      N        018      N        -        -        NPIS
 
lation Flow
 
D.1.2    RHR Heat Exchanger A      50-400F        TE-612      N        -        -        018        N        NPIS
 
Inlet/Outlet Tempera-                    TE-604      N        -        -        018        N        NPIS
 
tures
 
D.1.2    RHR Heat Exchanger B      50-400F        TE-613      N        -        -        018        N        NPIS
 
Inlet/Outlet Tempera-                    TE-605      N        -        -        018        N        NPIS
 
tures
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
 
See next page for Remarks.
* Revision 3 to Regulatory Guide 1.97 revised the range to 40&deg;F to 350&deg;F.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.1 (Continued)
 
III. REMARKS
: 1. The proper operation of the RHR system is verified by observing pump and valve status indications provided on the main
 
control board, which contains mimic diagrams of the flow paths. These indications are fully qualified to Class 1E
 
requirements.
: 2. The RHR system (Figure 5.4-7) serves the dual function of residual heat removal and low pressure
 
injection/recirculation. The flow rates are indicated for all modes of operation; however, they are provided for
 
performance monitoring only. The flow rate and temperature monitoring is not required for any safety-related function
 
and, therefore, the instruments are not Class 1E. The proper operation of the RHR system is verified by observing
 
pump and valve status indications provided on the main control board, which contains mimic diagrams of the flow
 
paths. These indications are fully qualified to Class 1E requirements.
: 3. Since the sensors/transmitters are part of the pressure boundary, they are designed to remain intact following an SSE;
 
however, functionality is not assured.
: 4. The RHR injection phase runout flow is limited to 4,428 gpm. The range of FT-618 and 619 is 0 to 5,500 gpm. The RHR
 
hot leg recirculation flow is 2,662 gpm for one RHR pump operating. The range of FT-988 is 0 to 4,500 gpm.
: 5. Train A flow (FT-618) and temperatures (TE-604 and 612) are recorded on TR-612. Train B flow (FT-619) and
 
temperatures (TE-605 and 613) are recorded on TR-613. The heat exchanger inlet temperatures are not considered to be
 
part of the Regulatory Guide 1.97 data base.
: 6. The RHR heat exchanger outlet temperature range from 50&deg;F to 400&deg;F is adequate to monitor any expected conditions leaving the heat exchanger. The minimum temperature of the RHR system is 60&deg;F in the long term following an accident
 
due to the automatic temperature control on the CCW system, which provides cooling water to the RHR heat exchanger.
 
The air-operated  temperature control valve which bypasses flow around the CCW heat exchanger is a safety-related
 
qualified valve; however, it is supplied by a nonsafety-related instrument air system. This system will most likely
 
be available during the long term following an accident, and it may be loaded onto the emergency diesel generator.
 
If this automatic control is not available, many options exist for operator action to control the CCW and/or RHR
 
temperatures and flows to maintain a minimum RHR heat exchanger outlet temperature at or above 50&deg;F; therefore, the
 
existing range of the outlet temperature indicators is adequate. With the given decay heat, it would take several
 
days for the outlet temperature to approach the low end of the currently monitored range. With operators periodically
 
monitoring RCS water temperature after an accident, it is not deemed credible for the outlet temperature to fall below
 
50&deg;F with no remedial actions being taken by the operating staff. As evidenced by the Revision 3 change to the low
 
end of the range (from 32&deg;F to 40&deg;F), it is WCGS' position that this required range is arbitrary and not based on plant-specific requirements for post-accident monitoring.
 
Rev. 1 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.2.1    Accumulator Tank          10% to 90% volume          2                To monitor operation
 
Level and Pressure        0 to 750 psig
 
D.2.2    Acccumulator Isolation    Closed or open            2                Operation status
 
Valve Position
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.2.1    Accumulator Tank Level  14+ inches    LT-950          N    018      N        -        -        NPIS (Unnecessary)                          through
 
957
 
D.2.1    Accumulator Tank        0-700 psig    PT-960          N    018      N        -        -        NPIS
 
Pressure (Unnecessary)                  through
 
967
 
D.2.2    Accumulator Isolation    Closed/Open    ZS 8808AA,AB    Y    018      Y        -        -        NPIS
 
Valves                                  through DA,DB
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The accumulator isolation valve position indication requirements are met.
: 2. Accumulator tank level and pressure indication are unnecessary variables and need not be provided for post-accident
 
monitoring. Therefore, Category 2 instruments are not required. Remark 3 provides additional justification. Remarks
 
4 and 5 discuss the available pressure  and level monitors and their ranges. These remarks also address the adequacy
 
of the existing ranges when compared to the recommended ranges of Table 2 of Regulatory Guide 1.97. Since these
 
variables are unnecessary, the comparision is provided only for information.
: 3. Table 2 of Regulatory Guide 1.97 lists accumulator pressure and level under Type D variables which are defined therein
 
as:  "Type D Variables:  Those variables that provide information to indicate the operation of individual safety
 
systems and other systems important to safety. These variables are to help the operator make appropriate decisions in
 
using the individual systems important to safety in mitigating the consequences of an accident".
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.2 (Continued)
 
III. REMARKS (Continued)
Accumulator level and pressure indication do not provide information which is relevant to the defined purpose of a
 
Type D variable. The accumulators are designed to passively inject water into the RCS when the primary pressure
 
falls below the accumulator cover gas pressure which is maintained between 585 to 665 psig. The nitrogen cover gas is not injected until much lower pressures (around 300 psig) are reached. Since the discharge of water
 
from the accumulators is beneficial for transients resulting from RCS breaks, the accumulator discharge valves are disabled so they cannot be closed from the control room. Section 15.6 provides RCS depressurization curves for various size LOCAs. The accumulators inject water for all LOCAs analyzed except for the 3-inch LOCA wherein the analysis was terminated at 2500 seconds.
 
If the operator had determined that there is no further need or potential need for accumulator water injection and he
 
desired to preclude the addition of nitrogen during the long-term LOCA recovery phase and if the RCS pressure had not
 
dropped below 300 psig, the operator may vent the accumulators and/or isolate the discharge of the accumulators by
 
directing the power breakers to be unlocked (outside the control room), provided that this action would not violate
 
any procedures.
 
For a LOCA, there is no need to determine if accumulator water has been injected. If water has been injected, it was
 
needed or at least not adverse to the core.
 
Should there be a question as to whether the accumulators actually discharged nitrogen into a depressurized but
 
relatively intact primary system, the operator could utilize the pressurizer and RV level indication to determine if
 
nitrogen was in the pressurizer or the vessel head. These areas can be vented from the control room, if it is deemed
 
appropriate.
 
Other Condition IV events (SGTR and MSLB) do not result in RCS depressurization transients which result in discharge
 
of accumulator nitrogen into the RCS. For these events, the operating staff will isolate or depressurize the
 
accumulators prior to proceeding to a cold shutdown condition. The operating staff has two variables available to
 
them to indicate the successful completion of this action:  valve position of the accumulator discharge valves and
 
valve position of the nitrogen vent valves. The operator is capable of isolating or depressurizing the accumulators
 
even with an assumed single failure. Therefore, the accumulator level and pressure indications are unncessary for
 
these events as well as a LOCA.
: 4. The range of the accumulator tank pressure transmitter is adequate to monitor any expected pressure in the
 
accumulator. The maximum pressure allowed in the accumulators is 665 psig. No fluid addition to the tank is expected  following an accident due to the check valve in the discharge line from each accumulator. Each accumulator has a relief valve set at 700 psig. Therefore, there is no need to extend the pressure indication beyond the present 700 psig range.
: 5. The recommended range of level indication from 10 to 90 percent of tank volume is unnecessary. The plant Technical Specifications require that the content of the tank be maintained within a very narrow range (6122 to 6594 gallons).
The instrumentation provided monitors the level of the tank for a span of approximately 14 inches in which the normal level is maintained. Monitoring the level above the Technical Specification value is not required because fluid addition following an accident is not postulated.
 
Monitoring the levels between the present range and the recommended range of 10 percent of tank volume is not required
 
because the addition of water contained in that volume, as noted previously, is beneficial and of no concern following
 
an accident.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                  RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.2.3      Boric Acid Charging      0-110% design flow 10        2              To monitor operation Flow
 
D.2.4      Flow in HPI System      0-110% design flow 10        2              To monitor operation
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.      VARIABLE                RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.2.3      Centrifugal Charging      0-160%          FT-917A      Y        018      Y        -        -      NPIS
 
Pump Flow (BIT)            0-160%          FT-917B      Y        018      Y        -        -      NPIS
 
D.2.4      Safety Injection Pump      0-123%          FT-918      N        017      N        -        -      NPIS
 
Flow                      0-123%          FT-922      N        017      N        -        -      NPIS
 
D.2.4      Charging to RCP Seals      0-250%          FT-215A      Y        001      Y        -        -      NPIS
 
0-250%          FT-215B      Y        001      Y        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The SI pump flow rate is 650 gpm for hot leg recirculation. The range of FT-918 and 922 (shown on Figures 6.3-1, Sheet 2) is 0 to 800 gpm. The centrifugal charging pump flow rate to the BIT path is 714 gpm (357 gpm per pump) for injection and recirculation. The range of FT-917A and 917B (shown on Figure 6.3-1, Sheet 3) is 0 to 570 gpm.
: 2. The flow to the RCP seals (shown on Figure 9.3-8) is provided by the centrifugal charging pumps, as described in Section 9.3.4. The normal flow rate is 32 gpm (8 gpm per pump). This flow path is also utilized as part of post  accident safe shutdown with only safety-related equipment. Refer to Section 7.4. The range of FT-215A and 215B is  80 gpm. 3. The safety injection flow is provided for performance monitoring only and is not required following an accident; therefore, the transmitters are not Class 1E. The centrifugal charging pump flow elements/transmitters are used during post accident safe shutdown; therefore, they are Class 1E.
 
Rev. 19 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 3.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.            VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.2.6      Refueling Water Storage        Top to bottom            2              To monitor operation
 
Tank Level
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE              RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.2.6    Refueling Water Storage    Top to bottom    LT-930      Y        018      Y        018        N      NPIS
 
Tank Level                                  LT-931      Y        018      Y        018        N      NPIS
 
LT-932      Y        018      Y        -        N      NPIS
 
LT-933      Y        018      Y        -        N      NPIS
 
___________________________________________________________________________________________________________________________
 
REMARKS
: 1. The RWST level instrumentation is shown on Figure 6.3-1, Sheet 1, and fully meets the stated requirements.
: 2. The RWST level indications and alarms are utilized during switchover from injection to recirculation in a 2-out-of-4
 
logic. RWST level is a Type A variable, per the assumptions stated in Section 7A.3.1.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 4.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE                CATEGORY                  PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.4.1    Steam Generator Level    From tube sheet to separators      1                To monitor operation
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE            RANGE          SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.4.1  Steam Generator Level -    7 inches above    LT-501          Y        025      Y        026      N    NPIS
 
Wide Range                tube sheet to      LT-502          Y        025      Y        026      N    NPIS
 
separators        LT-503          Y        025      Y        026      N    NPIS
 
LT-504          Y        025      Y        026      N    NPIS
 
NA    Steam Generator Level -    128 inches        LT-517,518,519  Y        026      Y        -        -    NPIS
 
Narrow Range                                  LT-527,528,529  Y        026      Y        -        -    NPIS
 
LT-537,538,539  Y        026      Y        -        -    NPIS
 
LT-547,548,549  Y        026      Y        -        -    NPIS
 
LT-551,2,3&4    Y        025      N        -        -    NPIS
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The steam generator wide range instrumentation provides level indication from 7 inches above the tube sheet to the
 
moisture separators (a range of 559 inches) and meets the intent of the recommended range. The steam generator is
 
essentially dry when the level drops below the lower tap (less than 300 gallons).
: 2. The four narrow range level transmitters on each loop are fully qualified and are considered to be a Type A variable
 
per the assumptions stated in Section 7A.3.1. The narrow range transmitters are used to identify a steam generator
 
tube rupture.
: 3. The narrow range instruments provide diverse indications within their range (438 to 566 inches above the tube sheet)
 
and would indicate the failure (high or low) of a wide range instrument.
: 4. Additional diverse variables for the wide range steam generator level measurement when the steam generator level is
 
below the bottom tap of the narrow range span consist of one channel of auxiliary feedwater flow per loop and three
 
steamline pressure measurements per loop.
 
Furthermore, a review of the WCGS Emergency Operating Procedures indicates that wide range steam generator level is
 
not utilized in any application that necessitates it being a Category 1 variable. As such, one channel per steam
 
generator is adequate to meet Category 2 requirements.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 4.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                      RANGE                CATEGORY                  PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.4.2    Steam Generator          From atmospheric pressure to      2                To monitor operation
 
Pressure                20 percent above the lowest
 
safety valve setpoint.
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE            RANGE          SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E
 
D.4.2    Steam Line Pressure    0-1,300 psig      PT-514, 5, 6    Y      026        Y      026 (PT-514) N    NPIS
 
(0-110% above      PT-524, 5, 6    Y      026        Y      026 (PT-524) N    NPIS
 
lowest safety      PT-534, 5, 6    Y      026        Y      026 (PT-535) N    NPIS
 
valve setpoint)    PT-544, 5, 6    Y      026        Y      026 (PT-545) N    NPIS
 
NA        Steam Line Pressure    0-1,500 psig      PT-1            Y      006        Y        -        -        -
 
for ARV Operation      126%              PT-2            Y      006        Y        -        -        -
 
PT-3            Y      006        Y        -        -        -
 
PT-4            Y      006        Y        -        -        -
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The lowest safety valve setpoint is 1,185 psig. The steam line pressure transmitters have a range of 0 to 1,300
 
psig  which is 10 percent above the lowest setpoint. Assuming a repeatability factor of +3 percent on the opening setpoint of the safety valves and a +
3 percent total channel accuracy of the steam line pressure monitoring channels, a margin of 40 psi exists between the upper range of the steam line pressure transmitters and the opening
 
setpoint of the lowest safety valve.
 
In addition, the WCGS atmospheric relief valves are fully qualified and available for controlled heat removal and
 
steam generator level control by maintaining a steam discharge rate approximately equal to the auxiliary feedwater
 
addition rate.
 
These atmospheric relief valves are set at 1125 psig and would lift prior to the safety valve with the lowest set pressure. The operation of these valves provides another 60 psi margin between the opening of a relief valve and the 1300 psig range of the steam line pressure indicators. Using this set point, the steam line pressure
 
transmitters have a range of 0 to 115 percent. The existing range of 0 to 1300 psig is adequate for the WCGS design since it provides sufficient margins above the expected secondary side pressures.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 4.2 (Continued)
 
III. REMARKS (Continued)
: 2. The steam line pressure transmitters used for ARV operation have a range of 0 to 1,500 psig, which is 126 percent of the lowest setpoint. These instruments are not considered part of the RG 1.97 data set per the assumptions stated in
 
Section 7A.3.2 and are not inputted to the Plant Computer. These instruments are fully qualified and meet the
 
requireents of Category 2 instrumentation.
: 3. The steam line pressure is a Type A variable per the assumptions stated in Section 7A.3.1, and is used to detect an
 
SGTR and secondary side break and to identify the affected steam generator.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 4.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.4.3    Safety/Relief Valve      Closed - not closed        2                To monitor operation
 
Positions or Main
 
Steam Flow
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT 
 
IDENT. NO. VARIABLE                RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER 
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.4.3  Atmospheric Relief    Closed - not closed  ZS-1            Y        006      Y        -        -      NPIS
 
Valve Position (ARV)                      ZS-2            Y        006      Y        -        -      NPIS 
 
ZS-3            Y        006      Y        -        -      NPIS
 
ZS-4            Y        006      Y        -        -      NPIS
 
D.4.3  Safety Valve Posi-    See Note 2
 
tion (20 valves)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The atmospheric relief valve (ARV) position fully meets the stated requirements.
: 2. Main Steam Safety valves are not monitored by the plant computer and have no Main Control Board indication. 
 
Rev. 21 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 4.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                  RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.4.4    Main Feedwater Flow  0-110 percent design flow 10      3              To monitor operation
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO. VARIABLE            RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.4.4  Main Feedwater Flow    0-121 percent    FT-510          N        026      N        006        N      NPIS
 
of VWO flow      FT-511          N        026      N        -        -      NPIS
 
FT-520          N        026      N        006        N      NPIS
 
FT-521          N        026      N        -        -      NPIS
 
FT-530          N        026      N        006        N      NPIS
 
FT-531          N        026      N        -        -      NPIS
 
FT-540          N        026      N        006        N      NPIS
 
FT-541          N        026      N        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations.
: 2. The flow transmitter has a range from 0 to 4.8 x 10 6 lbs/hr. The VWO flow is 3.96 x 10 6 lbs/hr for each line.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 5.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.5.1    Auxiliary or Emergency    0-110 percent design    2 (1 for            To monitor operation
 
Feedwater Flow            flow 10                  B&W plants)
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.5.1    Auxiliary Feedwater      0-160%        FT-1            Y        006      Y        -        -      NPIS
 
Flow                                    FT-2            Y        006      Y        -        -      NPIS
 
FT-3            Y        006      Y        -        -
FT-4            Y        006      Y        -        -
NA                                0-160%        FT-7            Y        -        -        -        -
FT-9            Y        -        -        -        -      NPIS
 
FT-11          Y        -        -        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The auxiliary feedwater system is described in Section 10.4.9 and shown on Figure 10.4-9.
: 2. Auxiliary feedwater flow to each steam generator is monitored by Class 1E flow loop. Each flow transmitter is powered
 
by a different separation group (1 through 4) corresponding to the power supply for the steam line ARV. Only two of the four steam generators are required to establish a heat sink for the RCS. The required flow indication to two
 
intact steam generators is assured assuming a single failure.
: 3. A comparision of the AFWS to the NUREG-0737 requirements for reliability and flow indication is provided in Section
 
18.2.7 which shows complete compliance to all recommendations.
: 4. The flow transmitters have a range of 0 to 400 gpm. The design flow to the steam generators is 250 gpm for a normal
 
shutdown. For a MSLB the design flow to two intact steam generators is 500 gpm (250 gpm each).
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 5.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.5.2    Condensate Storage        Plant Specific            1          To ensure water supply for auxiliary feedwater
 
Tank Level                                                      (Can be Category 3 if not primary source of
 
AFW. Then whatever is primary source of AFW
 
should be listed and should be Category 1.)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.5.2  Condensate Storage Tank  Top to bottom      PT-24          Y        005      Y        -        -      NPIS
 
Level (indicated by pump                    PT-25          Y        005      Y        -        -      NPIS
 
suction pressure)                            PT-26          Y        005      Y        -        -      NPIS
 
NA    Condensate Storage Tank  Appropriate for    PT-37          Y        026      Y        -        -       
 
Level (for automatic      automatic switch-  PT-38          Y        026      Y        -        -       
 
AFWS switchover)          over to ESW        PT-39          Y        026      Y        -        -       
 
NA    Condensate Storage Tank  0-100%            LT-4          N        005      N        -        -      NPIS
 
Level
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The CST is shown on Figure 9.2-23, and the pressure transmitters are shown on Figure 10.4-9. As stated in Section 10.4.9, the CST level is determined by PT-24, 25, and 26. The automatic switchover to ESW upon the depletion of CST
 
water volume is initiated by PT-37, 38, and 39. LT-4 is non-safety grade and provides a direct level reading;
 
however, this instrument is not considered part of the RG 1.97 data base.
: 2. Since there is no manual action required for switchover to the alternate source of auxiliary feedwater (ESW), the CST
 
level measurements are not Type A variables.
 
Rev. 12 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                CATEGORY                  PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.3.3    Containment Pressure 1    0 to design pressure 4          1          Funcation detection accomplishment of (psig)                                    mitigation, verification
 
B.4.2    Containment Pressure 1    10 psia to design pressure 4    1          Same
 
C.2.2    Containment Pressure 1    10 psia to design pressure 4    1          Detection of breach, accomplishment of psig (5 psia for subatmo-                mitigation, verification, long-term
 
spheric containments)                    surveillance
 
C.3.3    Containment Pressure 1    10 psia to 3 times design      1          Detection of potential for or actual breach,                                    pressure4 for concrete (4                accomplishment of mitigation, verification
 
times design pressure for
 
steel)  (5 psia for subat-
 
mospheric containments)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT
 
IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E B.3.3  Containment Pressure    0-69 psig      PT-934          Y        018      Y        018        N      NPIS
 
B.4.2  (normal design range)                    PT-935          Y        018      Y        018        N      NPIS
 
C.2.2                                            PT-936          Y        018      Y        018        N      NPIS
 
PT-937          Y        018      Y        018        N      NPIS
 
C.3.3  Containment Pressure -  -5 to 180 psig  PT-938          Y        020      Y        020        N      NPIS
 
Wide Range                              PT-939          Y        020      Y        020        N      NPIS
 
NA      Containment Pressure    (-) 85 to      PDY-40          N        020      N        -        -      NPIS (normal operating range) (+) 85 in H 2 O ___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated requirements.
: 2. The design pressure of the containment is 60 psig. The peak calculated pressure following a LOCA and MSLB are 47.3
 
and 48.9 psig, respectively. As stated in Section 7A.3.2, diversity is not required in extended ranges not associated with DBEs.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.1 (Continued)
 
III. REMARKS (Continued)
: 3. Monitoring of subatmospheric conditions recommended in items B.4.2, C.2.2, and C.3.3 is accomplished by the wide range
 
instruments.
: 4. Normal contaiment pressure is maintained near atmospheric pressure and measured by pressure transmitters located
 
inside and outside of the containment. The difference in pressures is indicated in the control room. This
 
instrumentation is not part of the Regulatory Guide 1.97 data base.
 
Rev. 1 WOLF CREEK WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                    CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.3.2    Containment Sump Water    Narrow range (sump) Wide        2              Function detection, accomplishment of
 
Level 1                    range (bottom of contain-        1              mitigation, verification
 
ment to 600,000-gallon level
 
equivalent)
 
C.2.3    Containment Sump Water    Narrow range (sump) Wide        2              Detection of breach, accomplishment of
 
Level 1                    range (bottom of contain-        1              mitigation, verification, long-term
 
ment to 600,000-gallon                          surveillance
 
level equivalent)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT 
 
IDENT. NO.        VARIABLE            RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER 
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.3.2    Normal Sump Water Level  836,000 gallons  LIT-9          Y        018      Y        -        -      NPIS
 
C.2.3                                              LIT-10          Y        018      Y        020        Y      NPIS
 
NA      RHR Recirculation Sump  626,000 gallons  LT-7          Y        018      Y        -        -      NPIS
 
Level                                      LT-8          Y        018      Y        020        Y      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. Refer to Section 18.2.12.2 for a comparison with NUREG-0737 requirements.
: 2. The WCGS design provides for Class 1E level monitoring in each of the two containment normal sumps and in each of
 
the two RHR sumps. A single range for the two containment normal sump level indicators is used to monitor both the
 
containment normal sump level and the containment water level. Both containment normal sump level indicator ranges
 
are 13 feet (156 inches) measured from 1995' 6" (6" above sump bottom) to 2008' 6" (equivalent to 836,000 gallons).
 
A single range for the two RHR sump level indicators is used to measure the RHR sump level and the containment water level. Both RHR sump level indicator ranges are 11 feet 7 inches (139 inches) measured from 1994 feet 6 inches (30
 
inches above sump bottom) to 2006' 1" (equivalent to 626,000 gallons). The LOCA analysis results in a maximum flood
 
level in containment of 2004' 8".
: 3. Both the normal and RHR sumps are provided with twin level elements which are indicated on one continuous indicator.
 
Redundancy is provided in each type of sump. Diversity is not required, since there are four independent water level
 
measurements.
: 4. The normal sump level is a Type A variable on WCGS. The normal sump level is used for event identification. The RHR
 
sump level is not a Type A variable. Although the recirculation sump level could be used for event identification, it
 
is not required and would not be flooded with water immediately following an event since there is a 6-inch of curb
 
around it. Similarly, since switchover to recirculation is initiated automatically on low RWST level, verification of
 
containment water level is not required nor part of a preplanned manual safety function. Refer to Section 7A.3.1.
 
Rev. 27 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                      PURPOSE
 
___________________________________________________________________________________________________________________________
 
B.4.1    Containment Isolation    Closed - not closed        1                Accomplishment of isolation
 
Valve Position (ex-
 
cluding check valves)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
B.4.1  Containment Isolation  Closed - not      See Figure      Y        Misc. Y        -        -      NPIS
 
Valve Position (ex-    closed            6.2.4-1
 
cluding manual and
 
check valves)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. Refer to Section 6.2.4 and 18.2.11 for discussions of containment isolation. As noted in Section 6.2.4, manual valves
 
do not have position indication in the control room. The position of the manual valves is verified on a monthly basis
 
in accordance with Technical Specifications. In addition, these valves are under administrative control and are locked or sealed closed whenever containment integrity is required. 
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
C.3.2    Containment Hydrogen    0 to 10% (capable of oper-      1            Detection of potential for breach, Concentration          ating from 10 psia to                          accomplishment of mitigation, long-term
 
maximum design pressure4)                      surveillance
 
E.6.3.1  Hydrogen Content        0 to 10%                        3            Release assessment, verification analysis
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT
 
IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
C.3.2    Containment Hydrogen      0-10%        AT-10          Y        020      Y        020        Y      NPIS
 
E.6.3.1  Concentration                          AT-19          Y        020      Y        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The hydrogen analyzers are described in Section 6.2.5 and shown on Figure 6.2.5-1.
: 2. The hydrogen analyzers meet all of the stated requirements. Refer to Section 18.2.12.2 for a comparison with NUREG-
 
0737 requirements. The analyzers will operate properly within the recommended containment pressure ranges.
: 3. The hydrogen concentration is a Type A variable and is used for initiating the Hydrogen Recombiners when hydrogen is 
 
detected.
 
Should the need arise, the recombiners could be started following load sequencing operations should the core or
 
primary systems indicate a potential for hydrogen generation rates above any current design bases.
: 4. As stated in Section 7A.3.2.d, diverse variables need only be performance grade and not Class 1E. Refer to data  sheet 13.1.
 
Rev. 20 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.5
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.6.3    Containment Atmosphere    40&deg;F to 400&deg;F            2                To indicate accomplishment of cooling
 
Temperature
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.6.3    Containment Atmosphere    0-400&deg;F      TE-60          Y        018      Y        -        -      NPIS
 
Temperature                            TE-61          Y        018      Y        -        -      NPIS
 
TE-62          Y        018      Y        -        -      NPIS
 
TE-63          Y        018      Y        020        Y      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations.
: 2. The WCGS design utilizes containment pressure to verify that containment heat removal is being accomplished. Refer to
 
data sheet 8.1 for a further discussion.
: 3. Containment temperature is not a Type A variable, since it does not meet the requirements discussed in Section 7A.3.1.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 6.6
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.6.4    Containment Sump Water    50&deg;F to 250&deg;F            2              To monitor operation
 
Temperature
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.6.4    Containment Sump Water
 
Temperature (unnecessary
 
variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. This variable is unnecessary for the WCGS plant. The recommended purpose is to "monitor operation"; however, there is
 
no system on WCGS for it to monitor. Containment cooling is monitored by the air temperature monitors described on
 
data sheet 6.5.
: 2. Sump temperature is not required for RHR operation or assurance of NPSH available, since NPSH calculations
 
conservatively assume saturated water was present. See Safety Evaluation Eleven of Section 6.2.2.1.3 and Table
 
6.2.2-7.
: 3. Primary system, PRT, and other containment parameters are all available to help determine the plant conditions. Sump
 
level indications indicate the amount of water, and the other parameters indicate its source.
: 4. Note that proper RHR functions during the recirculation mode are provided by other variables described on data sheet
 
3.1.
: 5. The Callaway SER (NUREG-0830) in Section 6.2.1.1 (page 6-4) indicates that the NRC Staff agrees that this variable is
 
not necessary for the SNUPPS plants and finds this exception to the guidelines of Regulatory Guide 1.97 acceptable.
: 6. The Callaway SER also addresses the containment heat removal systems and similarly finds them acceptable. Page 6-10
 
indicates that the RHR system serves to remove heat from the containment during the recirculation mode following a
 
LOCA by cooling the containment sump fluid in the RHR heat exchanger. During this mode of operation, the RHR inlet
 
temperature monitors described on Data Sheet 3.1 would provide indication of the containment sump water temperature.
 
As noted on Data Sheet 3.1, the RHR heat exchanger inlet temperature is not considered to be part of the Regulatory
 
Guide 1.97 data base.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 7.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.7.1    Makeup Flow - In        0 to 110% design flow 10      2                To monitor operation
 
D.7.2    Letdown Flow - Out      0 to 110% design flow 10      2                To monitor operation
 
D.7.3    Volume Control          Top to bottom                2                To monitor operation
 
Tank Level
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.7.1    Normal Charging Flow  50 to 267%        FT-121          N        002      N        -        -      NPIS
 
D.7.2    Normal Letdown Flow    0 to 267%          FT-132          N        002      N        -        -      NPIS
 
D.7.3    Volume Control Tank    Top to bottom of  LT-185          Y        002      Y        -        -        -
 
Level                  straight shell    LT-112          Y        002      Y        -        -        -
 
LT-149          N        -        -        -        -      NPIS
 
D.7.2    Safety Related Let-    0 to 167%          FT-138A        Y        001      Y        -        -      NPIS
 
down                  0 to 167%          FT-138B        Y        001      Y        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The normal charging and letdown flow rates are described on this data sheet. The DBA-related portion of the charging
 
system is described on data sheet 3.3.
: 2. The volume control tank level is Class 1E to ensure a suction source from the RWST (automatically) on low VCT level.
: 3. The level of the VCT is monitored for the straight shell portion only. The span is 75 inches. The hemispherical
 
heads are not monitored, since the volume-to-level ratio is not linear.
: 4. Section 7.4 describes the safety grade cold shutdown system provided in the WCGS design. As part of this design, a Class 1E letdown system is provided to the PRT through the excess letdown heat exchanger. FT-138A and B have a range
 
of 0 to 50 gpm. The maximum emergency letdown flow rate at RCS loop temperatures above 400&deg;F is 30 gpm. The design
 
flow below 400&deg;F has not been established; however, it is maintained below 50 gpm.
 
Rev. 14 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 8.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.6.2    Heat Removal by the        Plant specific            2                To monitor operation
 
Containment Fan Heat
 
Removal System
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E D.6.2    Containment Cooler Heat
 
Removal - (unnecessary
 
variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. Quantification of the amount of heat being removed by the containment fan coolers is an unnecessary variable and is
 
not provided on WCGS. The accomplishment of post-accident heat removal is verified by monitoring the operation of the
 
fan coolers and monitoring the containment pressure and air temperature. Containment pressure and air temperature
 
monitors are described on Data Sheets 6.1 and 6.5.
: 2. Monitoring of containment air cooler operation is provided by three sets of indications, all of which are safety-
 
related and qualified for post-accident operation. These items do indicate that the air coolers are operating;
 
however, they do not quantify the amount of heat being removed from the containment atmosphere.
 
The handswitches for each containment air cooler fan are provided with lights which indicate the mode of operation
 
(stop, slow, or fast) for each containment air cooler.
 
The ESF status panel indicates whether the fan coolers are being provided with power (control and fan power supply).
 
If the control fuse blows or if the power breaker trips, a red trouble light appears on one of the ESF status panel
 
windows "Ctmt Cooler Fan SGN01A (B, C or D)."  Also, an audio alarm is generated.
 
The containment isolation valves serving each set of two containment air coolers are also provided with Class 1E hand
 
indication switches in the control room. These position switches indicate that the isolation valves are open and that
 
the lines to each cooler are capable of passing the cooling water flow. Since the containment isolation valves are
 
normally open and receive a confirmatory open signal on the receipt of a safety injection signal, the ESF status panel
 
also contains windows for these valves. A red light will appear and an audio alarm is sounded if any valve fails to
 
take its post-accident position (open).
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 8.1 (Continued)
 
III. REMARKS (Continued)
 
On WCGS, the heat removal capability of the containment air coolers is accurately determined by sophisticated
 
mathematical and computer modeling developed by the air cooler supplier. The accuracy of the model was verified
 
during the prototype testing of three different coils at three different post-accident pressures. Topical Report AAF-
 
TR-7101 (Reference 1 of USAR Section 6.2.2.3) provides a comparison of the measured heat removal during the tests to
 
the computer analysis predictions. The comparisons show very close agreement between the predicted and actual heat
 
removal abilities. The NRC has approved the topical report for reference in construction permit and operating license
 
applications.
: 3. During the transient of an accident, heat removal by air coolers cannot be used by an operator, since too many
 
variables are changing rapidly. The amount of energy release to the containment cannot be accurately quantified .
 
Heat removal mechanisms are those identified in Section 6.2.1 and include heat transfer to passive heat sinks, containment sprays, and containment air coolers. The operator must determine what equipment is operating and watch
 
the changes in containment pressure, temperature, sump level, and radiation levels to determine the nature of the
 
accident.
: 4. The operability of the air coolers is verified periodically throughout the life of the plant in accordance with
 
Technical Specifications, which ensures the proper operation of the system. 
 
Rev. 11
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 9.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.8.1    Component Cooling Water    32 F* to 200&deg;F            2                To monitor operation
 
Temperature to ESF
 
System
 
D.8.2    Component Cooling Water    0 to 110% design          2                To monitor operation
 
Flow to ESF System        flow10
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE            RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.8.1    CCW Heat Exchanger      0-200&deg;F          TE-31          Y        019      Y        -        -      NPIS
 
Discharge Temperature                    TE-32          Y        019      Y        -        -      NPIS
 
D.8.2    CCW Pump Discharge      0-137 percent    FT-95          N        -        -        -        -      NPIS
 
Flow                                      FT-96          N        -        -        -        -      NPIS
 
FT-97          N        -        -        -        -      NPIS
 
FT-98          N        -        -        -        -      NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The component cooling water system is described in Section 9.2.2. The WCGS design meets the recommended ranges.
: 2. Section 7A.3.7 describes the qualification of NRC Category 2 variables, as provided on WCGS. The instruments
 
described herein are located outside of the containment in areas served by Class 1E room coolers. These instruments
 
are not required for the proper operation of the system; rather, they are provided for performance monitoring only.
: 3. Since these instruments are part of the system pressure boundary, they are seismically designed to ensure integrity of
 
the system boundary.
* Revision 3 of Regulatory Guide 1.97 revised the range to 40&deg;F to 200&deg;F.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 10.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.6.1    Containment Spray Flow    0-110% design flow 10      2                To monitor operation
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE            RANGE          SENSOR/TRANSMITTER              CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.6.1    Containment Spray Flow  0-126% (design    FT-5            N        017      N        -        -      NPIS
 
flow - injection)
 
0-106% (design    FT-11          N        017      N        -        -      NPIS
 
flow - recircula-
 
tion)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The containment spray system is described in Section 6.2.2. The spray system need only operate during the injection
 
phase for cooling purposes. During this phase, the flow rate monitor exceeds the recommended range.
: 2. Section 7A.3.7 describes the qualification of NRC Category 2 items, as provided on WCGS. These instruments are
 
located outside of the containment in areas served by Class 1E room coolers. These instruments are provided for
 
performance monitoring and not to allow proper system operation.
: 3. The instruments are part of the pressure boundary and are seismically designed to ensure its integrity.
: 4. Class 1E operability indications for each containment spray train are provided in the control room. All motor-
 
operated valves in the flow paths are provided with hand indication switches and receive a CSAS to open. The
 
containment spray pumps also have hand switches and start automatically on a CSAS. The ESF status indication panel
 
provides backup information on a component and system level and indicates the system's status. Should the power
 
breakers trip or the control fuses blow, an amber light appears and an audio signal is generated.
 
Also, redundant Class 1E level indication is provided on the spray additive tank. Reducing level in this tank
 
indicates that sodium hydroxide additive is being injected into an operable spray system flow path.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 11.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                  RANGE              CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
C.2.4    Containment Area        1 R/hr to 10 4 R/hr          3 6,7            Detection of breach, verification Radiation1
 
E.1.1    Containment Area        1 R/hr to 10 7 R/hr          1 6,7            Detection of significant releases, release Radiation - High                                                    assessment, long-term surveillance, Range1                                                              emergency plant actuation
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT
 
IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
C.2.4    Containment Area      1 to 10 8 R/hr    RE-59          Y        067      Y        -        -      RMS/NPIS Radiation
 
E.1.1                                            RE-60          Y        067      Y        20        Y      RMS/NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. These instruments meet all of the stated recommendations and are further described in Section 18.2.12.2.
: 2. As described in Section 7A.3.2, diverse variables are performance grade. The WCGS design includes area radiation monitors with a range to 10 R/hr located inside the containment. 3. This is a Type A variable and is used for event identification in the EOIs.
 
Rev. 20 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 11.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE              CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
C.3.5    Radiation Exposure Rate    10
-1 R/hr to 10 4 R/hr        2 7          Indication of breach (inside buildings or
 
areas, e.g., auxiliary
 
building, reactor shield
 
building annulus, fuel
 
handling, which are in
 
direct contact with
 
primary containment
 
where penetrations and
 
hatches are located)1
 
E.2.1    Radiation Exposure Rate1  10
-1 R/hr to 10 4 R/hr        2 7            Detection of significant releases, release (inside building or                                                  assessment, long-term surveillance
 
areas where access is
 
required to service
 
equipment important
 
to safety)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT 
 
IDENT. NO.        VARIABLE            RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER 
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
C.3.5    Radiation Exposure Rate  (Unnecessary Variable)
 
E.2.1
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. Area radiation monitors are shown on Figure 12.3-2 and are provided in accordance with the criteria stated in Section
 
12.3.4.1. Process and effluent monitors are provided in accordance with the criteria stated in Section 11.5. Area
 
monitors are provided in the corridors of the auxiliary building and not in the penetration areas or equipment
 
spaces. 
 
Rev. 24 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 11.2 (Continued)
 
III. REMARKS (Continued)
: 2. The process and effluent monitors will provide indication of releases and/or breaches in the systems in operation
 
following an event. Use of extended range area monitors in the areas adjacent to the containment are not appropriate
 
since the background, direct radiation levels can be expected to be quite high. The process and effluent monitors
 
provide the required public protection.
: 3. The existing area radiation monitors provide for adequate employee protection with their range to 10R/hr. Should this
 
range be exceeded, employee entry is prohibited.
: 4. Exposure rate monitors associated with variable C.3.5 were deleted in Revision 3 of Regulatory Guide 1.97.
 
Rev. 0 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
__________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE                CATEGORY                        PURPOSE
 
___________________________________________________________________________________________________________________________
C.3.4    Containment Effluent        10
-6  Ci/cc to              2 8,9          Detection of breach, accomplishment of Radioactivity - Noble        10
-2  Ci/cc                                mitigation, verification Gases from Identified Release Points1 C.3.6    Effluent Radioactivity1      10
-6  Ci/cc to              2 8            Indication of breach Noble Gases (from            10
-3  Ci/cc            buildings or areas where penetrations
 
and hatches are
 
located)
 
E.3.1.1  Containment or Purge        10
-6  Ci/cc to 10 5  Ci/cc    2 8            Detection of significant releases,            Effluent                    0 to 110% vent design                      release assessment flow 10 (Not needed if ef-fluent discharges through
 
common plant vent)
 
E.3.1.3  Auxiliary Building1          10
-6  Ci/cc to              2 8            Detection of significant releases,            (including any building      10 3  Ci/cc                                  release assessment, long-term sur-containing primary          0 to 110% vent design                      veillance system gases, e.g.,          flow 10 (Not needed if waste gas decay tank)        effluent discharges
 
through common plant
 
vent)
 
E.3.1.5  Common Plant Vent or        10
-6  Ci/cc to              2 8            Detection of significant releases,            Multipurpose Vent Dis-      10 3  Ci/cc                                  release assessment, long-term sur-charge Any of above          0 to 110% vent design flow 10                veillance Release (if contain-        10
-6  Ci/cc to            ment purge is included)      10 4  Ci/cc ___________________________________________________________________________________________________________________________
 
Rev. 0 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.1 (Continued)
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE          RANGE              SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E E.3 1.1  Containment or Purge  10
-6 to 10 5  Ci/cc            Effluent C.3.4    Plant Unit Vent Wide  10
-7 to 10 5  Ci/cc    GT-RE-21B      N        SP010    N        SP010    N      RMS/NPIS E.3.1.5  Range Gas Radwaste Building      10
-7 to 10 5  Ci/cc    GH-RE-10B      N        SP010    N        SP010    N      RMS/NPIS Wide Range Gas
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The plant unit vent receives the discharge from the containment purge, auxiliary building, control building, fuel
 
building, and the condenser air removal filtration system. The radwaste building vent receives the discharge from the
 
radwaste building exhaust fans. The radwaste building contains the waste gas decay tanks.
: 2. The unit vent flow rate is determined by fan run contacts which are inputted to the RMS computer. Each system is
 
balanced and assumed to be operating at the design flow. The high range monitor has an isokinetic flow monitor.
 
These provisions adequately meet the requirements of the item.
: 3. The radwaste building vent is a constant flow vent receiving the discharge of the radwaste building exhaust fans.
 
Flow rate monitoring is not required. The high range monitor for the radwaste building vent also has an isokinetic
 
nozzle.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
C.2.5    Effluent Radioactivity -    10
-6 to 10-2  Ci/cc      3 8              Detection of breach, verification Nobel Gas Effluent from Condenser Air Removal System Exhaust 1  E.3.1.4  Condenser Air Removal      10
-6 to 10 5  Ci/cc        2 8              Detection of significant releases,          Exhaust1                    0 to 110 percent vent                      release assessment
 
design flow 10 (not                                      needed if effluent dis-
 
charges through common
 
plant vent)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE            SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E
 
C.2.5  Condenser Air Removal  10
-7 to 10-2 Ci/cc    RE-92          N        056      N        056      N      RMS/NPIS Exhaust Radioactivity
 
E.3.1.4 Condenser Air Removal Exhaust (not required-discharge through plant
 
vent)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The condenser air removal exhaust discharges through the plant vent; therefore, the monitor for item E.3.1.4 is not
 
required. The existing condenser air removal exhaust monitor meets the requirements of item C.2.5.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
E.3.1.6    Vent from Steam Gen-    10
-1  Ci/cc to 10 3  Ci/cc    2 12            Detection of significant release erator Safety Relief    (duration of releases in                    assessment Valves or Atmospheric    seconds and mass of steam Relief Valves            per unit time)
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE          RANGE            SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E
 
E.3.1.6  Vent from Steam Gen-  1.3 x 10
-2  Ci/cc    RE-111          N        SP010    N        SP010      N    RMS/NPIS erator Safety Relief  to 1.3 x 10 3  Ci/cc  RE-112          N        SP010    N        SP010      N    RMS/NPIS Valves or Atmospheric                      RE-113          N        SP010    N        SP010      N    RMS/NPIS Relief Valves                              RE-114          N        SP010    N        SP010      N    RMS/NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design monitors the atmospheric relief valve plumes. The atmospheric relief valves are set to open at a lower pressure than the safety relief valves and are Class 1E, highly
 
reliable components. These valves are provided with position indication. It is assumed that
 
the atmospheric relief valves are open and releasing the same concentration and distribution of
 
radio-nuclides any time any of the safety valves on the same steam line are open. 2. Radiation detectors are positioned to view the plume directly from each of the four atmospheric relief valves.
 
Rev. 21 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
E.3.1.7  All other Identified    10
-6  Ci/cc to 10 2  Ci/cc      2 8            Detection of significant releases, release Release Points          0-110 percent vent design                    assessment, long-term surveillance flow 10 (not needed if ef-fluent discharges through
 
other monitored plant vents)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE            SENSOR/TRANSMITTER              CONTROL ROOM              COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E
 
E.3.1.7  Auxiliary Feedwater 1 to 1 x 10 5          RE-385        N        SP010    N        SP010    N      RMS/NPIS Pump Turbine Ex-    MR / HR haust Monitor
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. A radiation detector monitoring the plume of the auxiliary feedwater turbine exhaust is used to determine the
 
releases.
: 2. This release is from the main steam line; thus, the monitor was designed with the same capabilities as the monitors
 
for steam generator releases (Data Sheet 12.3). The range recommended is not applicable to secondary side releases, as can be seen by the different ranges recommended here and on Data Sheet 12.3.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 12.5
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE                CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.3.2    Particulates and Halogens
 
E.3.2.1  All Identified Plant        10
-3  Ci/cc to 10 2 Ci/cc    3 13          Detection of significant release,            Release Points (except      0 to 110% vent design                      release assessment, long-term sur-steam generator safety      flow10                                    veillance
 
relief valves or at-
 
mospheric relief valves and condenser air
 
removal system exhaust).
 
Sampling with Onsite
 
Analysis Capability
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE          RANGE                SENSOR/TRANSMITTER              CONTROL ROOM            COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL  CL. 1E E.3.2.1  Unit Vent Monitors  10
-3  Ci/cc to 10 2  Ci/cc              Particulates    (See data sheet 12.5,      GT-RE-21B      N        N/A      N        -      -    RMS/NPIS Iodines          III. Remarks, Note 3)
 
Radwaste Building    10
-3  Ci/cc to 10 2  Ci/cc          Vent Monitors        (See data sheet 12.5,              Particulates    III. Remarks, Note 3)    GH-RE-10B      N        N/A      N        -      -    RMS/NPIS Iodines
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations. Refer to Sections 11.5 and 18.2.12.2 for further discussions.
: 2. Refer to data sheet 12.1 for a discussion of vent flow rate monitoring and wide range gas monitors.
: 3. The wide range noble gas monitors described on data sheet 12.1 include the capability to obtain grab samples for both
 
halogens and particulates. After collection, laboratory samples are used to quantify releases.
 
Rev. 13 WOLF CREEK WOLF CREEK
 
TABLE 7A-3, DATA SHEET 13.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.6.1    Primary Coolant          Grab Sample                3 5,18 Release assessment, verification analysis E.6.1.1  Gross Activity          10  Ci/ml to 10 Ci/ml E.6.1.2  Gamma Spectrum          (Isotopic Analysis)
 
E.6.1.3  Boron Content            0 to 6,000 ppm
 
E.6.1.4  Chloride Content (2)      0 to 20 ppm
 
E.6.1.5  Dissolved Hydrogen or Total Gas (19)
E.6.1.6  Dissolved Oxygen (19)    0 to 20 ppm
 
E.6.1.7  pH                      1 to 13
 
B.1.3    RCS Soluble Boron        0 - 6,000 ppm              3  Verification
 
Concentration
 
C.1.3    Analysis of Primary      10  Ci/gm to 10 Ci/gm or  3 5  Detail analysis, accomplishment of            Coolant (Gamma Spectrum) TID-14844 source term in  mitigation, verification, long-term                                    coolant volume    surveillance
 
E.6.3    Containment Air          Grab Sample    Release assessment, verification analysis
 
E.6.3.2  Oxygen Content          0 to 30 percent    Release assessment, verification analysis
 
E.6.3.3  Gamma Spectrum          (Isotopic Analysis)  Release assessment, verification analysis
 
___________________________________________________________________________________________________________________________
 
Rev. 20 WOLF CREEK WOLF CREEK
 
TABLE 7A-3, DATA SHEET 13.1 (Continued)
 
II. WCGS DESIGN PROVISIONS No sampling and/or analysis of these variables as part of a post accident sampling system is performed at WCGS.
Reference USQD 59 98-0071, letter WO 98-0047, letter 98-01418, and letter 01-00234 (Amendment No. 137). Should sampling be required during recovery, assessments will address taking samples from the Nuclear Sample System and the Containment Hydrogen Monitoring Equipment.
 
Rev. 20 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 13.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                CATEGORY                  PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.6.2    Sump                      Grab sample                    3 5,18    Release assessment, verification analysis E.6.2.1  o  Gross Activity        10  Ci/ml to 10 Ci/ml          3 E.6.2.2  o  Gamma Spectrum        (isotopic analysis)            3
 
E.6.2.3  o  Boron Content          0-6,000 ppm                    3
 
E.6.2.4  o  Chloride Content      0-20 ppm (4)                    3
 
E.6.2.5  o  pH                    1 to 13                        3
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT
 
IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
E.6.2  Sump Grab Sample
 
Containment Recir-    See data sheet 13.1
 
culation
 
ECCS Pump Room Sumps      Not required
 
Auxiliary Building Sumps  Not required
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. No sampling and/or analysis of these variables as part of a post accident sampling system is performed at WCGS.
Reference USQD 59 98-0071, letter WO 98-0047, letter 98-01418, and letter 01-00234 (Amendment No. 137). Should sampling be required during recovery, assessments will address taking samples from the Nuclear Sample System.
: 2. The ECCS pump room and auxiliary building sumps are provided with Class 1E level indication and operate as described
 
in Section 9.3.3. Process and effluent monitors provide indication of any airborne activity in these sumps since they
 
are directly vented to the auxiliary building normal exhaust system.
: 3. Sump sampling for the ECCS pump rooms and auxiliary building is considered unnecesary. The Class 1E level indication
 
will detect any accumulated leakage, and the isolation valves will prevent its discharge from the auxiliary building.
 
Should the leakage be from a line that contains fluid from the recirculation sump, the recirculation sump sample will
 
provide the recommended analyses, since the fluid is from the same source.
: 4. The analysis can be performed on site if dose rates allow, or by an off-site facility contracted to provide results within four days.
 
Rev. 20 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 13.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                    CATEGORY                    PURPOSE
 
___________________________________________________________________________________________________________________________
 
C.1.2    Radioactivity Concen-      1/2 Technical Specification      1              Detection of breach
 
tration or Radiation      limit to 100 times technical
 
Level in Circulating      specification, limit R/hr.
 
Primary Coolant
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT
 
IDENT. NO.        VARIABLE            RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
C.1.2    Radioactivity Concen-
 
tration (unnecessary
 
variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. As noted in comments provided by the AIF, this variable is unnecessary, and there is no presently available means of
 
providing this information. Also, there is no apparent need or use for this variable which would require its
 
classification as Category I.
 
Rev. 20 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 14.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                      PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.9.1    High-Level Radioactive    Top to bottom            3                To indicate storage volume
 
Liquid Tank
 
Level
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                          PLANT IDENT. NO.        VARIABLE          RANGE          SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E D.9.1  Recycle Holdup Tank
 
Level (Unnecessary
 
Variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design precludes the need for this variable. The liquid radwaste system is not required following an event.
 
It is located in the radwaste building, and is controlled from the radwaste building control room. System parameters
 
are not provided in the main control room.
: 2. The safety grade letdown system is located within the containment, and the containment isolation system is designed to
 
preclude inadvertent discharge from the containment.
: 3. The recycle holdup tank levels (LT-261 and LT-262) have a range from the top to bottom of the tank and indications are
 
provided in the radwaste building control room. Since the system is only operated from that room, the control room
 
operators may obtain that status of the tanks from the radwaste building control room personnel. The liquid radwaste
 
system need not be operated during an accident. It may be used during recovery, if the radwaste building is
 
habitable.
: 4. As noted on Data Sheet 13.2, the auxiliary building and ECCS pump room sumps are provided with Class 1E sump level
 
indication. These sumps would collect any long-term leakage from systems which recirculate fluids from the
 
containment sump. As described in Section 9.3.3 and shown on Figure 9.3-6, Sheet 2, the discharge lines from these
 
sumps contain Class 1E isolation valves which close on a SIS to preclude inadvertent discharge of fluids to the floor
 
drain tank in the radwaste building. The LOCA analysis includes an evaluation of a 2 gpm leak from lines
 
recirculating sump fluids. Refer to Section 15.6.5.4.1.2 for a discussion of the analysis and to Table 15.6-8 for the
 
resulting radiological consequences. Failure of this tank has been analyzed in USAR Section 15.7.2.
: 5. The containment normal and instrument tunnel sumps and the reactor coolant drain tank discharge lines are isolated by
 
a CIS-A signal. This signal is generated as a result of a safety injection signal or as a result of high containment
 
pressure. These lines are isolated subsequent to any LOCA. Refer to Section 18.2.11, which addresses NUREG-0737 Item
 
II.E.4.2, Containment Isolation Dependability. Inadvertent contamination of the radwaste or auxiliary buildings due
 
to discharge of fluids from the containment is precluded by design and is not postulated.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 14.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                    RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.9.2    Radioactive Gas Holdup  0-150% design pressure 4          3          To indicate storage capacity Tank Pressure
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.9.2    Gas Decay Tank
 
Pressure (unnecessary
 
variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The radioactive gas holdup tank is referred to as the gas decay tank (GDT). Pressure is an unnecessary variable for
 
WCGS design as described in Remark 3 below; however, Remark 2 describes the adequacy of the GDT design and the range
 
of the pressure indicators.
: 2. Addition of radioactive gases to the gaseous radwaste system following an accident is precluded by design and is not
 
postulated. Containment isolation valves on gas bearing lines from the pressurizer relief tank and the reactor
 
coolant drain tank close upon receipt of a CIS-A. Refer to Remark 5 on Data Sheet 14.1 for a further discussion of
 
containment isolation. Since there are no containment gases added to the gaseous radwaste system, there is no need to
 
monitor the available storage capacity following an accident.
: 3. The design pressure of each of the eight GDTs is 150 psig. Each tank is provided with a pressure transmitter/indi-
 
cator/alarm. The indicators are located in the radwaste building control room and have a range of 0 to 150 psig. The
 
alarms for the six GDTs used during normal operation are set at 100 psig. Two of the GDTs are used for shutdown and
 
start-up. All GDTs are provided with relief valves set at or below the tank's design pressure. The relief valves for
 
x GDTs discharge at design pressure to the shutdown GDTs which are normally at low pressure. Should an extended
 
discharge to the shutdown GDT occur, a high alarm (at 90 psig) would be received prior to the lifting of the shutdown
 
GDT relief valve at 100 psig. The discharge from the radwaste building vent is monitored by the radwaste building
 
vent monitor described on Data sheet 12.1. Failure of one of these tanks has been analyzed in USAR Section 15.7.1.
 
Based upon the protection afforded by the installed tank relief valves and the potential eventual release to the
 
radwaste building vent, the span of 0 to tank design pressure is adequate to provide information to the operating
 
staff concerning the status of the GDTs.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 15.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.10.1    Emergency Ventilation      Open-closed status        2                To indicate damper status
 
Damper Position
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE          RANGE        SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.10.1  Safety Related Damper    Open-closed  HIS-XX          Y        020      Y        -        -      NPIS
 
Position                                                        068
 
019
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The safety-related dampers which receive an automatic signal to reposition are provided with Class 1E position
 
indication in the control room. The WCGS design meets all of the stated recommendations.
 
Rev. 11
 
WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 16.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.11.1    Status of Standby Power    Voltages, currents,      2 11            To indicate system status Sources Important to Safety
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E D.11.1  Status of Standby Power
 
4160 V Class 1E Incoming
 
Current
 
Current                0-2000A          CT-NB0109        Y      RL015      N        -        -        NPIS
 
Current                0-2000A          CT-NB0111        Y      RL015      N        -        -        NPIS
 
Current                0-2000A          CT-NB0212        Y      RL015      N        -        -        NPIS
 
Current                0-2000A          CT-NB0209        Y      RL015      N        -        -        NPIS
 
Current                0-1200A          CT-PA0201        N      RL016      N        -        -        NPIS
 
4160 V Class 1E Bus Voltage
 
Voltage                0 - 5250 V      PT-101/B        Y      RL015      Y        -        -        NPIS
 
Voltage                0 - 5250 V      PT-201/B        Y      RL015      Y        -        -        NPIS
 
Diesel Gen No. 1
 
Current                0 - 1500A      CT-NE107          Y      RL015      N        -        -        NPIS
 
Voltage                0 - 5250 V      PT-NE107          Y      RL015      N        -        -        -
 
KW                    0 - 8MW        CT/PT-NE107      Y      RL015      N        -        -        NPIS
 
Vars                  0 - 8Mvar      CT/PT-NE107      Y      RL015      N        -        -        NPIS
 
Frequency              55 - 65 Hertz  PT-NE107          Y      RL015      N        -        -        NPIS
 
Diesel Gen No. 2
 
Current                0 - 1500A      CT-NE106          Y      RL015      N        -        -        NPIS
 
Voltage                0 - 5250 V      PT-NE106          Y      RL015      N        -        -        NPIS
 
___________________________________________________________________________________________________________________________
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 16.1 (Continued)
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT
 
IDENT. NO.        VARIABLE          RANGE              SENSOR/TRANSMITTER            CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E  PANEL  CL. 1E    PANEL    CL. 1E
 
KW                    0 - 8MW        CT/PT-NE106      Y      RL015      N        -        -        NPIS
 
Vars                  0 - 8Mvar      CT/PT-NE106      Y      RL015      N        -        -        NPIS
 
Frequency              55 - 65 Hertz  PT-NE106          Y      RL015      N        -        -        NPIS
 
Current to Class 1E 480 V System
 
Current                  0 - 300A          CT-NB0110      Y    RL015      N        -        -        NPIS
 
Current                  0 - 300A          CT-NB0113      Y    RL015      N        -        -        NPIS
 
Current                  0 - 300A          CT-NB0210      Y    RL015      N        -        -        NPIS
 
Current                  0 - 300A          CT-NB0213      Y    RL015      N        -        -        NPIS
 
Current                  0 - 300A          CT-NB0117      Y    RL015      N        -        -        NPIS
 
Current                  0 - 300A          CT-NB0217      Y    RL015      N        -        -        NPIS
 
Current                  0 - 100A          CT-NB0116      Y    RL015      N        -        -        NPIS
 
Current                  0 - 100A          CT-NB0216      Y    RL015      N        -        -        NPIS
 
Class 1E 125 V DC System                                        All Panel 16
 
Current Battery          (-)800 to (+)800A  Shunt-NK11      Y              Y        -        -        NPIS
 
Current Battery          (-)800 to (+)800A  Shunt-NK12      Y              Y        -        -        NPIS
 
Current Battery          (-)800 to (+)800A  Shunt-NK13      Y              Y        -        -        NPIS
 
Current Battery          (-)800 to (+)800A  Shunt-NK14      Y              Y        -        -        NPIS
 
Current Battery Charger  0 - 500A          Shunt-NK21      Y              Y        -        -        NPIS
 
Current Battery Charger  0 - 500A          Shunt-NK22      Y              Y        -        -        NPIS
 
Current Battery Charger  0 - 500A          Shunt-NK23      Y              Y        -        -        NPIS
 
Current Battery Charger  0 - 500A          Shunt-NK24      Y              Y        -        -        NPIS
 
Voltage                  0 - 150V          Batt Mon-NK11  Y              Y        -        -        NPIS
 
Voltage                  0 - 150V          Batt Mon-NK12  Y              Y        -        -        NPIS
 
Voltage                  0 - 150V          Batt Mon-NK13  Y              Y        -        -        NPIS
 
Voltage                  0 - 150V          Batt Mon-NK14  Y              Y        -        -        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations. All Class 1E 4.16-kv buses and 125 VDC system are provded with voltage and current indications. The 480 volt system is provided with current indications only.
 
Rev. 13 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 16.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE              CATEGORY                PURPOSE
 
___________________________________________________________________________________________________________________________
 
D.11.1    Status of Energy          Pressures                2 11            To indicate system status Sources Important to        Safety (hydraulic, pneumatic)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
____________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
D.11.1  Air Accumulator Tank
 
Pressures
 
AFW Control Valves and  0-800 psig      PT-108          N        -        -        -        -        NPIS
 
Secondary Side          0-800 psig      PT-110          N        -        -        -        -        NPIS
 
Atmospheric Relief      0-800 psig      PT-112          N        -        -        -        -        NPIS
 
Valves                  0-800 psig      PT-114          N        -        -        -        -        NPIS
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
: 1. The safety-related air accumulators are described in Section 9.3.1 and shown on Figure 9.3-1, Sheet 5. The WCGS
 
design meets all of the stated requirements.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 17.1
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                  CATEGORY            PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.4.1    Radiation Exposure    Range, location, and qualifi-      3          Verification of significant release and
 
Meters (continuous    cation criteria to be devel-                  local magnitudes
 
indication at fixed  oped to satisfy NUREG-0654, locations)            Section II.H.5.b and 6.b for
 
emergency radiological
 
monitoring
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
(Unnecessary Variable)
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
 
This variable has been deleted from Regulatory Guide 1.97 in Revision 3.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 17.2
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                  RANGE                CATEGORY            PURPOSE
 
___________________________________________________________________________________________________________________________
E.4.2    Airborne Radiohalogens    10
-9 to 10-3  Ci/cc          3 14        Release assessment; analysis and Particulates (port-able sampling with on-site analysis capability)
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
See Remarks Section
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
 
Health physics air sampling and analysis equipment is available on site for the monitoring and assessment of airborne
 
radioactivity concentations. Airborne sampling capabilities for particulates and radioiodines are provided by low
 
flow air samplers using glass fiber filters and TEDA-impregnated activated charcoal or silver Zeolite cartridges
 
(accident conditions). Analysis of collection media are performed by germanium gamma ray spectroscopy equipment
 
(multichannel analyzer and HPGe detector). In the auxiliary warehouse laboratory, utilization of laboratory gamma
 
spectroscopy equipment ensures the capability to analyze samples within the detection limits of 10-9  Ci to 10-3  Ci
 
for principal gamma emitters.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 17.3
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                  CATEGORY            PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.4.3    Plant and Environs    10
-3 to 10 4 R/hr photons          3 15        Release assessment; analysis Radiation (portable  10
-3 to 10 4 rads/hr beta          3 15            instrumentation)      radiations and low-energy
 
photons
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
See Remarks Section
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
 
In accordance with Regulatory Guide 1.97 recommendations, portable radiation survey instrumentation with the
 
capability to detect gamma radiation over the range of 10
-3 to 104 R/hr is maintained in the site health physics instrument inventory. The capability to measure beta radiation fields over the range of 10-3 to 104 R/hr is provided
 
by portable survey instrumentation equipped with beta-sensitive detectors.
 
Rev. 11 WOLF CREEK
 
WOLF CREEK
 
TABLE 7A-3, DATA SHEET 17.4
 
I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
___________________________________________________________________________________________________________________________
 
VARIABLE IDENT. NO.      VARIABLE                RANGE                  CATEGORY            PURPOSE
 
___________________________________________________________________________________________________________________________
 
E.4.4    Plant and Environs      Multichannel gamma-ray          3          Release assessment; analysis
 
Radioactivity (portable  spectrometer
 
instrumentation)
 
___________________________________________________________________________________________________________________________
 
II. WCGS DESIGN PROVISIONS
___________________________________________________________________________________________________________________________
 
VARIABLE                                                                                                        PLANT IDENT. NO.        VARIABLE        RANGE            SENSOR/TRANSMITTER                CONTROL ROOM                COMPUTER
 
___________________________________________________________________________________________________________________________
 
INDICATOR          RECORDER
 
IDENT. NO. CL. 1E    PANEL  CL. 1E    PANEL    CL. 1E
 
See Remarks Section
 
___________________________________________________________________________________________________________________________
 
III. REMARKS
 
A portable, battery powered, 2,048-channel multichannel analyzer is used with a 2-inch x 2-inch NaI detector for
 
quantification of radioactivity in plant and environmental radiological samples. In addition, portable single-
 
channel analyzers with NaI detectors are available in emergency kits for analysis of selected radioisotopes.
 
Rev. 11 WOLF CREEK TABLE 7A-3, DATA SHEET 17.5 I. REGULATORY GUIDE 1.97 TABLE 2 RECOMMENDATIONS
__________________________________________________________________________________________________________________________
_ VARIABLE IDENT. NO.      VARIABLE                RANGE                  CATEGORY            PURPOSE
 
___________________________________________________________________________________________________________________________
E.5.1    Wind Direction        0 to 360 degrees (+
5 degress      3            Release assessment accuracy with a deflection of
 
15 degrees). Starting speed
 
0.45 mps (1.0 mph). Damping
 
ratio between 0.4 and 0.6, distance constant <2 meters
 
                                                  =
 
E.5.2    Wind Speed            0 to 30 mps (67 mph) +
0.22 mps    3            Release assessment (0.5 mph) accuracy for wind
 
speeds less than 11 mps (24
 
mph) with a starting threshold
 
of less than 0.45 mps (1.0 mph)
E.5.3    Estimation of        Base on vertical temperature      3            Release assessement
 
Atmospheric          difference from primary system, Stability            -5 C to 10 C (-9 F to 18 F) and
 
                                +
0.15 C accuracy per 50-meter intervals (+
0.3 F accuracy per 164-foot intervals) or analogous
 
range for alternative stability
 
estimates
 
___________________________________________________________________________________________________________________________
II. WCGS DESIGN PROVISIONS VARIABLE IDENT. NO.
VARIABLE  RANGE  SENSOR/TRANSMITTER CONTROL ROOM PLANT COMPUTER    I DENT. NO.
CL. 1E INDICATOR PANEL CL. 1E RECORDER PANEL CL. 1E E.5.1 Wind Direction 0-360 degrees, +3 degrees MET ONE 020C or MET ONE 50.5 (Wind Direction)
N  NPIS E.5.2 Wind Speed 0-100 mph
+1% or + 0.15 mph (whichever is greater)  (timed average values)  Threshold 0.6 mph (010c) or .2 mph (50.5) MET ONE 010C or MET ONE 50.5 (Wind Speed) N  NPIS Rev. 21 WOLF CREEK TABLE 7A-3, DATA SHEET 17.5 (Continued)
 
II. WCGS DESIGN PROVISIONS (Continued)
VARIABLE IDENT. NO.
VARIABLE  RANGE  SENSOR/TRANSMITTER CONTROL ROOM PLANT COMPUTER    IDENT. NO.
CL. 1E INDICATOR PANEL CL. 1E RECORDER PANEL CL. 1E E.5.3 Estimate of Atmospheric Stability Temperature Temperature Difference
 
Dew Point
 
Precipitation Ground Level
-50 to +50 C
+0.3 C 
 
-4 of +6 C
+0.1 C (timed average valves)  -50 to 50 C RH between 10% and 100% RH  0 - 3 inch
+1% Platinum RTD/
T-200 Met. One
 
Platinum RTD/
T-200 Met. One
 
MET. One 083D R H Sensor
 
Met One 375 Precip. Gage N 
 
N 
 
N 
 
N 
 
NPIS 
 
NPIS 
 
NPIS 
 
NPIS 
 
III. REMARKS
: 1. The WCGS design meets all of the stated recommendations. See Table 2.3-48 for Reg. Guide 1.23 Instrumentation (No Dew Point or Precip)
: 2. The meteorological information system (site related) provides inputs to the NPIS via the meteorological monitoring system at the met towers. The NPIS converts the inputs to digital form at the met tower and transmits them to the NPIS computer in the Operations Relief Area. 3. The parameters are sampled at a frequency of 1 minute or less by the NPIS.
 
Rev. 21 WOLFCREEKNOTESTOTABLE7A-3(Sheet1)FootnotestoRegulatoryGuide1.97Table2-PWRVariables 1Whereavariableislistedformorethanonepurpose,theinstrumentationrequirementsmaybeintegratedandonlyonemeasurementprovided.
2ThemaximumvaluemayberevisedupwardtosatisfyATWS requirements.
3Aminimumoffourmeasurementsperquadrantisrequiredforoperation.Sufficientnumbershouldbeinstalledtoaccountforattrition.(Replacementinstrumentationshouldmeetthe2300F rangeprovision.)
4DesignpressureisthatvaluecorrespondingtoASMEcodevaluesthatareobtainedatorbelowcode-allowablesvaluesformaterialdesignstress.
5Samplingormonitoringofradioactiveliquidsandgasesshouldbeperformedinamannerthatensuresprocurementofrepresentativesamples.Forgases,thecriteriaofANSIN13.1shouldbe applied.Forliquids,provisionsshouldbemadeforsamplingfrom well-mixedturbulentzones,andsamplinglinesshouldbedesigned tominimizeplateoutordesposition.Forsafeandconvenient sampling,theprovisionsshouldinclude:a.ShieldingtomaintainradiationdosesALARAb.Samplecontainerswithcontainer-samplingportconnector compatibilityc.Capabilityofsamplingunderprimarysystempressureandnegativepressuresd.Handlingandtransportcapabilitye.Prearrangementforanalysisandinterpretation 6Minimumoftwomonitorsatwidelyseparatedlocations.
7Detectorsshouldrespondtogammaradiationphotonswithinanyenergyrangefrom60keVto3MeVwithanenergyresponseaccuracyof+20percentatanyspecificphotonenergyfrom0.1MeVto3MeV.Overallsystemaccuracyshouldbewithinafactoroftwoovertheentirerange.Rev.0 WOLFCREEKNOTESTOTABLE7A-3(Sheet2) 8Monitorsshouldbecapableofdetectingandmeasuringradioactivegaseouseffluentconcentrationswithcompositionsrangingfromfreshequilibriumnoblegasfissionproductmixturesto10-day-old mixtures,withoverallsystemaccuracieswithinafactoroftwo.
EffluentconcentrationsmaybeexpressedintermsofXe-133 equivalentsorintermsofanynoblegasnuclide(s).Itisnot expectedthatasinglemonitoringdevicehassufficientrangeto encompasstheentirerangeprovidedinthisregulatoryguide.
Multiplecomponentsorsystemsareneeded.Existingequipmentmay beusedtomonitoranyportionofthestatedrangewithintheequipmentdesignrating.
9ProvisionsshouldbemadetomonitorallidentifiedpathwaysforreleaseofgaseousradioactivematerialstotheenvironsinconformancewithGeneralDesignCriterion64.Monitoringof individualeffluentstreamsisonlyrequiredwheresuchstreams arereleaseddirectlyintotheenvironment.Iftwoormore streamsarecombinedpriortoreleasefromacommondischarge point,monitoringofthecombinedstreamisconsideredtomeetthe intentoftheregulatoryguide,providedsuchmonitoringhasa rangeadequatetomeasureworst-casereleases.
10Designflowisthemaximumflowanticipatedinnormaloperation.
11Statusindicationofallstandbypoweracbuses,dcbuses,inverteroutputbuses,andpneumaticsupplies.
12EffluentmonitorsforPWRsteamsafetyvalvedischargesandatmosphericreliefvalvedischargesshouldbecapableofapproximatelylinearresponsetogammaradiationphotonswithenergiesfromapproximately0.5MeVto3MeV.Overallsystem accuracyshouldbewithinafactoroftwo.Calibrationsourcesshouldfallwithintherangeofapproximately0.5MeVto1.5MeV(e.g.,CS-137,Mn-54,Na-22,andCo-60).Effluentconcentrations shouldbeexpressedintermsofanygamma-emittingnoblegas nuclidewithinthespecifiedenergyrange.Calculationalmethods shouldbeprovidedforestimatingconcurrentreleasesoflow-energynoblegasesthatcannotbedetectedormeasuredbythemethodsortechniquesemployedformonitoring.
13Toprovideinformationregardingreleaseofradioactivehalogensandparticulates.Continuouscollectionofrepresentativesamplesfollowedbyonsitelaboratorymeasurementsofsamplesfor radiohalogensandparticulates.Thedesignenvelopefor shielding,handling,andanalyticalpurposesshouldassume30minutesofintegratedsamplingtimeatsamplerdesignflow,anaverageconcentrationof10 2&#xb5;Ci/ccofparticulateradioiodinesandparticulatesotherthanradioiodines,andanaveragegammaphotonenergyof0.5MeVperdisintegration.Rev.13 WOLFCREEKNOTESTOTABLE7A-3(Sheet3) 14Forestimatingreleaseratesofradioactivematerialsreleasedduringanaccident.
15Tomonitorradiationandairborneradioactivityconcentrationsinmanyareasthroughoutthefacilityandthesiteenvironswhereitisimpracticaltoinstallstationarymonitorscapableof coveringbothnormalandaccidentlevels.
16GuidanceonmeteorologicalmeasurementswasdevelopedinaProposedRevision1toRegulatoryGuide1.23,"MeteorologicalProgramsinSupportofNuclearPowerPlants." 17Thetimefortakingandanalyzingsamplesshouldbe3hoursorlessfromthetimethedecisionismadetosample,exceptforchloridewhichshouldbewithin24hours.
18Aninstalledcapabilityshouldbeprovidedforobtainingcontainmentsump,ECCSpumproomsumps,andothersimilarauxiliarybuildingsumpliquidsamples.
19Appliesonlytoprimarycoolant,nottosump.Rev.0}}

Latest revision as of 03:58, 6 January 2025

1 to Updated Safety Analysis Report, Chapter 7, Instrumentation and Controls
ML18093A895
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 03/08/2018
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ML18093A810 List: ... further results
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