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WOLF CREEK TABLE OF CONTENTS

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.

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£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£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 §!£ -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°F);(2)hotshutdown(reactorsub-critical,350°F>T-avg>200°F);and(3)coldshutdown(reactorsub-critical,T-avg<200°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°F and 360 psig. Continued residual heat removal. The RHR system is utilized to achieve cold shutdown temperature (Tavg <200°F ).

Cooldown is accomplished by increasing the steam dump from the ARVs to attain a

primary side cooldown rate of approximately 50°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°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°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°Ftolessthan200°F.ItisthenusedtomaintainthetemperatureoftheRCSlessthan200°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°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 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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°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£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£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£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

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

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°F to 2300°F (for operating 3 3 Verification plants - 200°F to 1650F)

C.1.1 Core Exit Temperature 1 200°F to 2300°F (for operating 1 3 Detection of potential for breach plants - 200°F to 1650°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°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

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

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°F subcooling 2 Verification of analysis and plant

to 35°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°F subcooled to RP081A Y 022 Y - - NPIS 2,000°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

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°F to 400°F 3 Verification

Temperature 1

B.2.2 RCS Cold Leg Water 50°F to 750°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°F TE-413B Y 021 Y 022 N NPIS

B.2.2 Wide Range T Cold 0-700°F TE-423B Y 021 Y 022 N NPIS 0-700°F TE-433B Y - - 022 N NPIS

0-700°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°F to 700°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°F to 750°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°F TE-413A Y 021 Y 022 N NPIS

Wide Range T hot 0-700°F TE-423A Y 021 Y 022 N NPIS 0-700°F TE-433A Y - - 022 N NPIS

0-700°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°F to 700°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

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°F to 750°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°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°F* to 350°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°F to 350°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°F to 400°F is adequate to monitor any expected conditions leaving the heat exchanger. The minimum temperature of the RHR system is 60°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°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°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°F to 40°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°F to 400°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°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°F to 250°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°F is 30 gpm. The design

flow below 400°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

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°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°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°F to 200°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

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

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 2µCi/ccofparticulateradioiodinesandparticulatesotherthanradioiodines,andanaveragegammaphotonenergyof0.5MeVperdisintegration.Rev.13 WOLFCREEKNOTESTOTABLE7A-3(Sheet3) 14Forestimatingreleaseratesofradioactivematerialsreleasedduringanaccident.

15Tomonitorradiationandairborneradioactivityconcentrationsinmanyareasthroughoutthefacilityandthesiteenvironswhereitisimpracticaltoinstallstationarymonitorscapableof coveringbothnormalandaccidentlevels.

16GuidanceonmeteorologicalmeasurementswasdevelopedinaProposedRevision1toRegulatoryGuide1.23,"MeteorologicalProgramsinSupportofNuclearPowerPlants." 17Thetimefortakingandanalyzingsamplesshouldbe3hoursorlessfromthetimethedecisionismadetosample,exceptforchloridewhichshouldbewithin24hours.

18Aninstalledcapabilityshouldbeprovidedforobtainingcontainmentsump,ECCSpumproomsumps,andothersimilarauxiliarybuildingsumpliquidsamples.

19Appliesonlytoprimarycoolant,nottosump.Rev.0

ML18093A895

Contents

Text

7.1 INTRODUCTION

1.3 REFERENCES

2.1 DESCRIPTION

2.4 REFERENCES

3.9 REFERENCES

4.1 INTRODUCTION

7.1 DESCRIPTION

7.3 REFERENCES

7.1 INTRODUCTION

1.3 REFERENCES

2.1 DESCRIPTION

2.4 REFERENCES

3.9 REFERENCES

4.1INTRODUCTION

7.1 DESCRIPTION

7.3 REFERENCES

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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ML18093A895

Text

7.1 INTRODUCTION

1.3 REFERENCES

2.1 DESCRIPTION

2.4 REFERENCES

3.9 REFERENCES

4.1 INTRODUCTION

7.1 DESCRIPTION

7.3 REFERENCES

7.1 INTRODUCTION

1.3 REFERENCES

2.1 DESCRIPTION

2.4 REFERENCES

3.9 REFERENCES

4.1INTRODUCTION

7.1 DESCRIPTION

7.3 REFERENCES

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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