L-2016-198, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 7, Instrument and Control
Text
TABLE OF CONTENTS
Section Title Page 7 INSTRUMENT AND CONTROL 7.1-1
7.1 General Design Criteria 7.1-1 7.1.1 Instrumentation and Control Systems Criteria 7.1-1 Instrumentation and Control Systems 7.1-1 NUREG-0700 "Guidelines for Control Room Design Review" 7.1-1 Regulatory Guide 1.97, Revision 3, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" 7.1-1a 7.1.2 Related Criteria 7.1-2 7.1.3 References 7.1-3
7.2 Protective Systems 7.2-1 7.2.1 Design Bases 7.2-1 Core Protection Systems 7.2-1 Engineered Safety Features Protection Systems 7.2-2 Protection Systems Reliability 7.2-3 Protection Systems Redundancy and Independence 7.2-4 Protection Against Multiple Disability for Protection Systems 7.2-5 Demonstration of Functional Operability of Protection Systems 7.2-6 Protection Systems Failure Analysis Design 7.2-6 Redundancy of Reactivity Control 7.2-7 Reactivity Control Systems Malfunctions 7.2-7 Principles of Design 7.2-7 Redundancy and Independence 7.2-7 Manual Actuation 7.2-8 Channel Bypass or Removal from Operation 7.2-8 Capability for Test and Calibration 7.2-8 Information Readout and Indication of Bypass or Removal from Operation 7.2-9 Vital Protective Functions and Functional Requirements 7.2-9 Completion of Protection Action 7.2-10 Multiple Trip Settings 7.2-10 Interlocks 7.2-10 Protective Actions 7.2-10 Indication 7.2-11 Annunciators 7.2-11 Digital Data Processing System (DDPS) 7.2-11 Distributed Control System (DCS) Safety Parameter Display System (SPDS)/ Emergency Response Data Acquisition & Display System (ERDADS) 7.2-12
7-i Revised 09/20/2016 C28 TABLE OF CONTENTS (Continued)
Section Title Page
7.2.2 System Design 7.2-13 Reactor Protection System Description 7.2-13 System Safety Features 7.2-13 Separation of Redundant Protection Channels 7.2-13 Loss of Power 7.2-14 Reactor Trip Signal Testing 7.2-15 Process Channel Testing 7.2-15 Logic Channel Testing 7.2-16 Primary Power Source 7.2-19 Protective Actions 7.2-19 Reactor Trip Description 7.2-19 Manual Trip 7.2-20 High Nuclear Flux (Power Range) Trip 7.2-20 High Nuclear Flux (Intermediate Range) Trip 7.2-20 High Nuclear Flux (Source Range) Trip 7.2-21 Overtemperature T Trip 7.2-21 Overpower T Trip 7.2-21 Low Pressurizer Pressure Trip 7.2-22 High Pressurizer Pressure Trip 7.2-22 High Pressurizer Water Level Trip 7.2-22 Low Reactor Coolant Flow Trip 7.2-23 Safety Injection System (SIS) Actuation Trip 7.2-23 Turbine Generator Trip 7.2-24 Steam/Feedwater Flow Mismatch Trip 7.2-24 Low-Low Steam Generator Water Level Trip 7.2-24 Rod Stops 7.2-24 Rod Drop Detection 7.2-25 Control Group Rod Insertion Monitor 7.2-26 Setpoint Methodology 7.2-26 7.2.3 System Evaluation 7.2-28 Reactor Protection System and DNB 7.2-28 Specific Control and Protection Interactions 7.2-29 Coolant Temperature 7.2-29 Pressurizer Pressure 7.2-29 Pressurizer Level 7.2-31 Steam Generator Water Level; Feedwater Flow 7.2-31 Steam Line Pressure (Hi Steam Line Flow) 7.2-34 Normal Operation Environment 7.2-34 7.2.4 ATWS Mitigating System Actuation Circuitry (AMSAC) 7.2-35 7.2.5 Steam Generator Overfill Protection 7.2-37 7.2.6 Eagle 21 Protection System 7.2.38 7.2.7 References 7.2-40
7-ii Revised 08/17/2016 C28 TABLE OF CONTENTS (Continued)
Section Title Page 7.3 Regulating Systems 7.3-1 7.3.1 Design Basis 7.3-1 7.3.2 System Design 7.3-4 RCCA Arrangements 7.3-4 Con trol Group Rod Control 7.3-5 Shutdown Groups Control 7.3-6 Interlocks 7.3-7 Rod Drive Performance 7.3-8 Full Length RCCA Position Indication 7.3-8 Individual RCCA Position Indication 7.3-10 Demand Position Indication 7.3-10 Rod Deviation 7.3-10 Turbine Bypass 7.3-11 Feedwater Control 7.3-11 Pressure Control 7.3-12 7.3.3 System Design Evaluation 7.3-13 Unit Stability 7.3-13 Step Load Changes Without Steam Dump 7.3-13 Loading and Unloading 7.3-14 Loss of Load With Turbine Bypass 7.3-15 Turbine-Generator Trip With Reactor Trip 7.3-15
7.4 Nuclear Instrumentation 7.4-1 7.4.1 Design Bases 7.4-1 Fission Process Monitors and Controls 7.4-1 Primary Nuclear Instrumentation 7.4-1 Backup Nuclear Instrumentation 7.4-2 7.4.2 System Design 7.4-2 Protection Philosophy 7.4-3 Source Range Instrumentation 7.4-4 Intermediate Range Instrumentation 7.4-5 Power Range Instrumentation 7.4-6 Equipment Design Basis 7.4-7 7.4.3 Detailed Description 7.4-7 Detectors 7.4-7 Source Range 7.4-8a Source Range Auxiliary Equipment 7.4-12 Visual - Audio Count Rate 7.4-12 Remote Count Rate Meter 7.4-13 Remote Recorder 7.4-13 Start-up Rate Circuitry 7.4-13 Intermediate Range 7.4-14 Intermediate Range Auxiliary Equipment 7.4-16 Power Range 7.4-17 Power Range Auxiliary Equipment 7.4-21 Comparator 7.4-21 Remote Recorder 7.4-21
7-iii Revised 04/17/2013
TABLE OF CONTENTS (Continued)
Section Title Page
Remote Meter 7.4-22 Overpower Recover 7.4-22 Remote Meter (Delta Flux) 7.4-22a Axial Flux Comparator 7.4-22a Flux Deviation and Miscellaneous Control and Indication Drawer 7.4-23 7.4.4 System Evaluation 7.4-23 Philosophy and Set Points 7.4-23 Reactor Trip Protection 7.4-24 Rod-Drop Protection 7.4-25 Control and Alarm Functions 7.4-26 Source Range 7.4-26 Intermediate Range 7.4-26 Power Range 7.4-27 Loss of Power 7.4-28 Safety Factors 7.4-28 7.4.5 Regulatory Guide 1.97, Revision 3 7.4-28
7.5 Engineered Safety Features Instrumentation 7.5-1 7.5.1 Design Basis 7.5-1 Engineered Safety Features Protection Systems 7.5-1 7.5.2 System Design 7.5-2 Engineered Safety Feature Actuation Instrumentation Description 7.5-2 Feedwater 7.5-3 Indication 7.5-3 Engineered Safety Features Instrumentation 7.5-3 Containment Pressure 7.5-3 Refueling Water Storage Tank Level 7.5-4 Safety Injection Pumps Discharge Pressure/Flow 7.5-4 Safety Injection Pump Energization 7.5-4 Radioactivity 7.5-4 Valve Position 7.5-4 Emergency Containment Coolers 7.5-5 Containment Level Instrumentation 7.5-5 Miscellaneous Instrumentation 7.5-5 Alarms 7.5-6 Instrumentation Used During LOCA 7.5-6
7-iv Revised 04/17/2013
TABLE OF CONTENTS (Continued)
Section Title Page 7.5.3 System Evaluation 7.5-7 Pressurizer Pressure 7.5-7 Steam Generator Level Control During Unit Cooldown 7.5-8 Environmental Capability 7.5-8 7.5.4 Regulatory Guide 1.97, Revision 3 7.5-8 7.5.4.1 Regulatory Guide 1.97 (Revision 3) Requirements 7.5-8 7.5.4.2 Evaluation Criteria 7.5-10 7.5.4.2.1 Environmental Qualification Criteria 7.5-10 7.5.4.2.2 Seismic Qualification Criteria 7.5-11 7.5.4.2.3 Redundance 7.5-12 7.5.4.2.4 Power Sources 7.5-13 7.5.4.2.5 Display and Recording 7.5-14 7.5.4.2.6 Range 7.5-15 7.5.4.3 Type A Variables 7.5-15 7.5.4.4 References 7.5-16 7.6 In-Core Instrumentation 7.6-1 7.6.1 Design Basis 7.6-1 7.6.2 System Design 7.6-1 Thermocouples 7.6-2 Movable Miniature Neutron Flux Detectors 7.6-2 Mechanical Configuration 7.6-2 Control and Readout Description 7.6-3 7.6.3 System Evaluation 7.6-5 7.6.4 Regulatory Guide 1.97, Revision 3 7.6-5
7.7 Operating Control Stations 7.7-1 7.7.1 Design Basis 7.7-1 7.7.2 System Design 7.7-2 7.7.2.1 Control Room 7.7-2 7.7.2.2 Remote (Alternate) Shutdown Capabilities 7.7-3
7.7.3 System Evaluation - Human Factors Engineering 7.7-7 7.7.3.1 HFE Program 7.7-7 7.7.3.2 Detailed Control Room Design Review Implementation 7.7-7 Technical Approach 7.7-8 Assessment 7.7-9 Implementation 7.7-9 7.7.3.3 DCRDR Implementation Evaluation 7.7-10 7.7.4 References 7.7-12
7-v Revised 04/17/2013 TABLE OF CONTENTS (Continued)
Section Title Page 7.8 Miscellaneous Alarms 7.8-1 7.8.1 Design Basis 7.8-1 Loose Parts Detection System 7.8-1 7.8.2 System Design 7.8-1 7.8.3 Alarm Indication 7.8.1
7.9 Leading Edge Flow Meter (LEFM) 7.9-1 7.9.1 Design and Operation 7.9-2 7.9.2 Operational Restrictions 7.9-2 7.9.3 References 7.9-3
7-vi Revised 04/17/2013 C26 APPENDICES
Appendix 7A Distributed Control System (DCS) /Safety Assessment System (SAS) / Emergency Response Data Acquisition and Display System (ERDADS)
7-vii Revised 04/17/2013
C26 LIST OF TABLES
Table Title 7.2-1 Reactor Trip List
7.2-2 Permissive Circuits
7.2-3 Rod Stops
7.4-1 Source Range
7.4-2 Intermediate Range
7.4-3 Power Range
7.5-1 Parameter Listing Summary Sheets Unit 3
7.5-2 Parameter Listing Summary Sheets Unit 4
7.9-1 LEFM Calorimetric Instrumentation 7.9-2 Reduced Power Limits Applicable to Inoperable LEFM Calorimetric Instrumentation 7A-1 DELETED 7A-2 DELETED
7-viii Revised 04/17/2013
C26 LIST OF FIGURES
Figure` Title 7.2-1 Typical Illustration of High T ( T vs T avg) 7.2-2 Reactor Protection Systems
7.2-3A Reactor Protection System - Redundant Channel Separation Design Configuration
7.2-3B ESF Actuation System - Redundant Channel Separation Design Configuration
7.2-4 Reactor Protection System - Typical Process Channel Testing Configuration
7.2-5 Reactor Trip Signals
7.2-6a Reactor Protection System - Typical Logic Relay Testing Configuration
7.2-6b ESF Actuation System - Typical Logic Relay Testing Configuration
7.2-7 RPS Logic Channel Test Panels
7.2-8a Pressurizer Caused Reactor Trip and Safety Injection Logic Diagram
7.2-8b Steam Generator Caused Reactor Trip and Safety Injection Logic Diagram
7.2-8c Primary Coolant System - Reactor Trips and T avg Interlock Logic Diagram
7.2-8d Nuclear Instrumentation Trip Signals Logic Diagram
7.2-8e Safeguards Actuation and Steam Line Actuation Logic Diagram
7.2-8f Nuclear Instrumentation Permissives and Block Logic Diagram
7.2-8g Setpoint Relationships
7.2-9a Rod Control System - Control System Diagram
7.2-9b T avg Control and Insertion Limit Alarms - Control System Diagram
7.2-10 Index and Symbols for Logic Diagrams
7.2-11a Pressurizer Pressure Protection and Overpressure Mitigation System - Control System Diagram
7.2-11b Pressurizer Pressure Control - Control System Diagram
7.2-12 Pressurizer Level Control and Protection and Charging Pump Control - Control System Diagram
7.2-13 Steam Generator Level Control and Protection - Control System Diagram
7-ix Revised 04/17/2013 C26C26 LIST OF FIGURES (Continued)
Figure Title
7.2-14a ATWS Mitigation System - Actuation Circuitry (AMSAC) - Logic Diagram (Unit 3) 7.2.14b ATWS Mitigation System - Annunciation Circuitry (AMSAC) - Logic Diagram (Unit 3)
7.3-1 Steam Dump to Condenser Logic Diagram
7.3-1a Steam Dump to Condenser Logic Diagram
7.4-1 Neutron Detectors and Range of Operation
7.4-2a Nuclear Instrumentation Trip Signals - Logic Diagram
7.4-2b Nuclear Instrumentation Permissives and Blocks Logic Diagrams
7.4-3 Plan View Indicating Detector Location Relative to Core
7.6-1 In-core Instrumentation Guide Tube Pressure Seals - Typical Configurations
7.7-1 Control Room Equipment Locations
7.7-2a Control Console Equipment Layout Sections 3C01
7.7-2b Control Console Front View Section 3C02
7.7-3 Vertical Panel "A" Front View Section 3C04
7.7-4 Vertical Panel "A" Front View Section 3C03
7.7-5 Vertical Panel "B" and "C" Front View Section 3C05
7.7-6 Vertical Panel "B" Front View Section 3C06
7.7-7 Control Console Front View Section 4C01
7.7-8 Control Console Front View Section 4C02
7.7-9 Vertical Panels "A" and "C" Front View Section 4CO4
7.7-10 Vertical Panel "A" Front View Section 4CO3
7.7-11 Vertical Panel "B" Front View Section 4CO5
7.7-12 Vertical Panel "B" Front View Section 4CO6
7.8-1 Loose Parts Monitoring System Units 3 & 4
7A-1 DCS(ERDADS) Cable Block Diagram
7-x Revised 04/17/2013 C26
- 7. INSTRUMENTATION AND CONTROL Supervision of the operation of the nuclear and turbine-generator portions of
each unit is accomplished by the instrumentation and control systems which
provide the control room operator with required information to operate the units
in a safe and efficient manner. The systems are designed to permit periodic on
line tests to demonstrate the operability of the reactor protection system.
7.1 GENERAL DESIGN CRITERIA
Criteria applying in common to all instrumentation and control systems are given
in the following listing. Thereafter, criteria which are specified to any one
of the instrumentation and control systems are discussed in that section in
which the system is described.
7.1.1 INSTRUMENTATION AND CONTROL SYSTEMS CRITERIA
Instrumentation and Control Systems Criterion: Instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential
reactor facility operating variables. (GDC 12)
Instrumentation and controls are provided to monitor and maintain all operationally important reactor parameters within prescribed operating ranges as
required by the stated criterion. Process variables which are required on a
continuous basis for the startup, power operation and shutdown are indicated, recorded, and controlled from the control room which is a controlled access
area. The quantity and types of instrumentation provided is adequate for safe
and orderly operation of all systems and processes over the full operating
range.
NUREG-0700 "Guidelines for Control Room Design Review" The control room design shall consider the control room workspace,
instrumentation, controls, and other equipment from a Human Factors Engineering
point of view that takes into account both system demands and operator
capabilities.
7.1-1 Rev 6 7/88 Regulatory Guide 1.97, Revision 3, "Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" Regulatory Guide 1.97, Revision 3, divides all instrumentation used to
monitor Post Accident variables into five functional types as defined in
Subsection 7.5.4.1. The requirements for this instrumentation are listed in
Table 1 of Regulatory Guide 1.97, Revision 3. The criteria are separated
into three separate groups or categories that provide a graded approach to
requirements depending on the importance to safety of the measurement of a
specific variable. Category 1 provides the most stringent requirements and
is intended for key variables. Category 2 provides less stringent
requirements, and generally applies to instrumentation designated for
indicating system operating status. Category 3 is intended to provide
requirements that will ensure that high-quality off-the-shelf instrumentation
is obtained and applies to backup and diagnostic instrumentation. It is also
used where the state of the art will not support requirements for higher
qualified instruments. Subsection 7.5.4 provides an in depth description of
Regulatory Guide 1.97, Revision 3.
7.1-1a Rev 6 7/88 7.1.2 RELATED CRITERIA Several criteria are related to all instrumentation and control systems but are more specific to other features or systems. These are therefore discussed in other chapters or references, as listed.
Criterion Discussion Suppression of Power Oscillations (GDC-7) Section 3.1 Reactor Core Design (GDC-6) Section 3.1 Quality Standards (GDC-1) Section 4.1 Performance Standards (GDC-5) Section 4.1 Fire Protection (GDC-3) Reference 1 Missile Protection (GDC-4O) Section 5.1 Emergency Power (GDC-39) Section 8.1
7.
1.3 REFERENCES
- 1. STD-M-006, Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4.
7.1-2 Revised 09/20/2016 C28C28C28 7.2 PROTECTIVE SYSTEMS The protection systems consists of the control and instrumentation associated
with the Engineered Safety Features and the Reactor Protection System. The
Engineered Safety Features Instrumentation is discussed further in Section
7.5.
This section contains Figures 7.2-5, 7.2-8a, 7.2-8b, 7.2-8c, 7.2-8d, 7.2-8e, 7.2-8f and 7.2-8g which illustrate logic along with the nominal trip
setpoints. The trip setpoint values are also contained in the Technical
Specifications and Table 7.2-1.
7.2.1 DESIGN BASES
Core Protection Systems
Criterion: Core protection systems, together with associated equipment, shall be designed to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits. (1967 Proposed GDC
- 14) If the reactor protection system receives signals which are indicative of an
approach to unsafe operating conditions, the system actuates alarms, prevents
control rod withdrawal, initiates load cutback, and/or opens the reactor trip
breakers.
The basic reactor operating philosophy is to define an allowable region of
power and coolant temperature conditions. This allowable range is defined by
the primary tripping functions, the overpower T trip, the over-temperature T trip and the nuclear overpower trip. The operating region below these trip settings is designed so that no combination of power, temperatures and
pressure could result in DNBR less than the safety analysis limit value with
all reactor coolant pumps in operation. A complete list of tripping
functions may be found in Table 7.2-1.
7.2-1 Revised 04/17/2013 C26 RCCA (rod cluster control assemblies) withdrawal is prevented by a dropped RCCA signal to provide additional core protection. The dropped RCCA is
indicated from individual RCCA position indicators, rod bottom bistables or
by a rapid flux decrease on any of the power range nuclear channels.
Rod stops from nuclear overpower, overpower T and overtemperature T deviation are provided to prevent abnormal power conditions which could
result from excessive control rod withdrawal initiated by operator violation
of administrative procedures. Automatic rod withdrawal by the reactor
control system has been permanently disabled. The overpower T and overtemperature T rod stop setpoints are the same as the reactor trip setpoints, effectively negating these functions.
Engineered Safety Features Protection Systems
Criterion: Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features. (1967 Proposed GDC 15)
Instrumentation and controls provided for the protective systems are designed
to trip the reactor, when necessary, to prevent or limit fission product
release from the core and to limit energy release; to signal containment
isolation; and to control the operation of engineered safety features
equipment.
The engineered safety features systems are actuated by the engineered safety
features actuation channels. Each coincidence network energizes an
engineered safety features actuation device that operates the associated
engineered safety features equipment, motor starters and valve operators.
The channels are designed to combine redundant sensors, and independent
channel circuitry, coincident trip logic and different parameter measurements
so that a safe and reliable system is provided in which a single failure will
not defeat the channel function. The action initiating sensors, comparators
and logic is shown in the figures included in the detailed
7.2-2 Revised 04/17/2013 Engineered Safety Features instrumentation description given in Section 7.5.2. The Engineered Safety Features instrumentation system actuates (depending on the severity of the condition) the Safety injection System, the Containment Isolation System, containment Emergency Containment Cooling System and Containment Spray System.
The passive accumulators of the Safety Injection System do not require signal or power sources to perform their function. A description of the actuation of the active portion of the Safety Injection System may be found in Table 7.2-1. Containment isolation is as tabulated in Table 7.2-1.
Protection Systems Reliability Criterion: Protection system shall be designed for high functional reliability and in-service testability necessary to avoid undue risk to the health and safety of the public.(1967 Proposed GDC 19) Protection channels are designed with sufficient redundancy for individual channel calibration and test to be made during power operation without degrading the reactor protection. In general, removal of the channel for calibration/surveillance is accomplished by placing the channel in a partial-trip mode. For example, a two-out-of-three channel becomes a one-out-of-two channel. Testing will not cause a trip unless a trip condition exists in a concurrent channel. Channel bypass capability exists for Eagle 21 (overpower T, overtemperature T and Hi Pressurizer Level) and the Nuclear Instrumentation System (Source Range and Intermediate Range) utilizes channel bypass for calibration/surveillance.
Protection and operational reliability is achieved in part by providing redundant instrumentation channels for each protective function. These redundant channels are electrically independent and physically separated.
The channel design incorporates separate sensors, separate power supplies, separate rack and panel mounted equipment and separate relays for the actuation of the protective function. For protective functions where two-out-of-three or two-out-of-four redundant-coincident actuation is provided, a single channel failure will not impair the protective function nor will it cause an unnecessary unit shutdown.
7.2-3 Revised 08/17/2016 Two of the three high-high containment pressure channels are powered from the same source. However, loss of either power source will not impair the
protective function nor will it cause an unnecessary actuation of containment
spray.
Protection Systems Redundancy and Independence
Criterion: Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure on removal
from service of any component or channel of such a system will
result in loss of the protection function. The redundancy
provided shall include, as a minimum, two channels of protection
for each protection function to be served. (1967 Proposed GDC
- 20)
The reactor protection system (for which credit is taken in the accident
analyses), is designed so that the most probable modes of failure in each
channel result in a signal calling for the protective trip. The protection
system design combines redundant sensors and channel independence with
coincident trip philosophy so that a safe and reliable system is provided, in
which a single failure will not defeat the channel function, cause a spurious
trip, or violate reactor protection criteria.
Channel independence is carried throughout the system, extending from the
sensor to the relay actuating the protective function. The protective and
control functions when combined are combined only at the sensor. The
protective and control functions are fully isolated in the remaining part of
the channel, control being derived from the primary protection signal path
through an isolation device. A failure in the control circuit, therefore, does not affect the protection channel.
A discussion of Engineered Safety Features (ESF) instrumentation may be found
under Section 7.5.1.
In the Reactor Protection System, two reactor trip breakers are provided to
interrupt power to the RCCA drive mechanisms. The breaker main contacts are
connected in series with each other and with the power supply so that opening
either breaker interrupts power to all full length RCCA drive mechanisms
permitting the RCCAs to free fall into the core.
7.2-4 Revised 04/17/2013 Further detail on redundancy is provided through the descriptions of the respective systems covered by the various subsections in this section.
Required continuous power supply for the protection systems is discussed in Section 8.
In summary, reactor protection is designed to meet all presently defined reactor protection criteria and is in accordance with the proposed IEEE 279 "Standard for Nuclear Plant Protection Systems" August 1968. The Eagle 21 instrumentation system is compliant with IEEE 279-1971 (see Section 7.2.6)
Protection Against Multiple Disability for Protection Systems
Criterion: The effects of adverse conditions to which redundant channels or protection systems might be exposed in common, either under normal conditions or those of an accident, shall not result in loss of the protection function or shall be tolerable on some other basis. (1967 Proposed GDC 23)
The components of the protection system are designed and laid out so that adverse environment accompanying an emergency situation, in which the components are required to function, does not interfere with that function.
Separation of redundant process protection channels originates at the process sensors and continues through the field wiring and containment penetrations to the process protection racks. Physical separation is used to the maximum practical extent 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 channel. Redundant process equipment is separated by locating components in different protection racks. Each channel is energized from a separate instrument bus.
Two of the three high-high containment pressure channels are powered from the same source. However, loss of either power source will not result in loss of the protective function nor will it cause unnecessary actuation of containment spray.
Wiring between vital elements of the system outside of equipment housing is routed and protected so as to maintain the true redundancy of the systems with respect to physical hazards.
7.2-5 Revised 08/17/2016 C28 Demonstration Of Functional Operability Of Protection Systems Criterion: Means shall be included for suitable testing of the active components of protection systems while the reactor is in
operation to determine if failure or loss of redundancy has
occurred. (1967 Proposed GDC 25)
The signal conditioning equipment of each protection channel in service at
power is capable of being calibrated and tested independently by simulated
analog input signals to verify its operation without tripping the reactor.
For the RPS, the logic testing scheme includes checking through the trip
logic relays to the trip breakers. For the ESF, the logic testing scheme
includes checking through the trip logic relays, but does not include the
master and slave relays. The master and slave relays are tested during
Engineered Safeguards Integrated Testing. Thus, the operability of each trip
channel can be determined conveniently and without ambiguity.
Protection System Failure Analysis Design
Criterion: The protection systems shall be designed to fail into a safe state or into a state established as tolerable on a defined basis if conditions such as disconnection of the system, loss of
energy (e.g., electrical power, instrument air), or adverse
environments (e.g., extreme heat or cold, fire, steam, or water)
are experienced. (1967 Proposed GDC 26)
Each reactor protection channel (for which credit is taken in the accident analyses), is designed on the "de-energize to operate" principle; an open
channel or a loss of power causes that channel to go into its trip mode. The
Turbine Emergency Trip Header Pressure (Low) is designed to energize the
associated logic relays. A loss of DC control power to this relay matrix
will still result in a reactor trip.
Reactor trip is implemented by simultaneously interrupting power to the
magnetic latch mechanisms on each drive allowing the full length rod clusters
to insert by free fall. The entire reactor protection system is thus
inherently safe in the event of a loss of power.
Each engineered safety feature channel (Instrumentation and logic relay) is
designed on the "de-energized to operate" principle; an open channel or a
loss-of-power causes that channel to go into its actuate mode. To achieve
ESF actuation, the master and slave relays for each ESF feature (e.g., SI, MSIS,AFW initiation, etc.) must energize to actuate that feature.
7.2-6 Revised 04/17/2013 C26 The components of the protection system are designed and laid out so that adverse environment accompanying an emergency situation, in which the
components are required to function, does not interfere with that function.
Refer to Appendix 8A for additional information pertaining to Environmental
Qualification.
Redundancy of Reactivity Control
Criterion: Two independent control systems, preferably of different principles, shall be provided. (1967 Proposed GDC 27)
One of the two reactivity control systems employs rod cluster control
assemblies to regulate the position of the neutron absorbers within the
reactor core. The other reactivity control system employs the Chemical and
Volume Control System to regulate the concentration of boric acid solution
neutron absorber in the Reactor Coolant System. These systems are described
in Sections 3.2 and 9.2, respectively.
Reactivity Control Systems Malfunction
Criterion: The reactor protection system shall be capable of protecting against any single malfunction of the reactivity control system, such as unplanned continuous withdrawal (not ejection or
dropout) of a control rod, by limiting reactivity transients to
avoid exceeding acceptable fuel damage limits. (1967 Proposed
Reactor shutdown with RCCA is completely independent of the normal control
functions since the trip breakers interrupt the power to the full length rod
mechanisms regardless of existing control signals. Effects of continuous
withdrawal of a RCCA and of deboration are described in Sections 7.3, 9.2,
and 14.1.
Principles of Design
Redundancy and Independence
The protective systems are redundant and independent for all vital inputs and
functions. Each channel is functionally independent of every other channel
and receives power from an independent source. Each train is functionally
independent of the other train and receives power from an independent source.
Separation of redundant protection channels is described in further detail in
Section 7.2.2.
7.2-7 Revised 04/17/2013 Manual Actuation Means are provided for manual initiation of protective system action.
Failure in the automatic system does not prevent the manual actuation of
protective functions. Manual actuation is designed to require the operation
of a minimum of equipment.
Channel Bypass or Removal from Operation
The system is designed to permit any one channel to be maintained, tested or
calibrated during power operation without system trip. During such operation
the active parts of the system continue to meet the single failure criterion, since the channel under test is either tripped or makes use of superimposed
test signals which do not negate the process signal.
Channel bypass capability exists for Eagle 21 (overpower T, overtemperature T and Hi Pressurizer level) and the Nuclear Instrumentation System (Source Range and Intermediate Range).
The systems with bypass capability are permitted to violate the single
failure criterion during channel bypass, since acceptable reliability has
been demonstrated and bypass time interval is short.
Capability for Test and Calibration
The rack portions of the protective system (e.g., relays, comparators, etc.)
provide trip or actuation signals only after signals from process portions of
the system reach preset values. Capability is provided for calibrating and
testing the performance of the rack portion of protective channels and
various combinations of the logic networks during reactor operation.
7.2-8 Revised 04/17/2013 The operational availability of each system input sensor, during reactor operation, is accomplished by cross checking between redundant channels or
between channels which bear a known relationship to each other and which have
readouts available. Provisions have been made for transmitter calibrations
during normal power operation, if deemed necessary.
The design provides for administrative control for the purpose of removing
the channels from service for test and calibration purposes and for
adjustment. The design provides for administrative control of access to all
trip settings, module calibration adjustments, test points, and signal
injection points.
Information Readout and Indication of By-Pass or Removal from Operation
The protective system provides the operator with complete information
pertinent to system status and safety.
Indication is provided by the annunciation system if some part of the system
has been administratively bypassed or taken out of service.
Trips are indicated and identified down to the channel level.
Vital Protective Functions and Functional Requirements
The Reactor Protection System monitors all parameters related to safe
operation of the reactor. The system is designed to trip the reactor so as
to protect the core against fuel rod cladding damage caused by departure from
nucleate boiling (DNB), and to protect the Reactor Coolant System against
damage caused by over-pressure. The Engineered Safety Features
Instrumentation System monitors parameters to detect failure of the Reactor
Coolant System, and initiates Engineered Safety Features operation. The
Engineered Safety Features Instrumentation System is described in 7.5.1.
7.2-9 Revised 04/17/2013 Completion of Protective Action 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 part of the protective system and are designed in accordance with the applicable criteria of this section.
The protective systems are so designed that, once initiated, a protective action goes to completion. Return to normal operation requires administrative action by the operator.
Multiple Trip Settings
For monitoring nuclear flux, multiple trip settings are used. When it is necessary to change to a more restrictive trip setting to provide adequate protection for a particular mode of operation or set of operating conditions, the design provides positive means of assuring that the more restrictive trip setting is used. The devices used to prevent improper use of less restrictive trip settings are considered a part of the protective system and are designed in accordance with the applicable criteria of this section.
Interlocks
Interlocks required to limit the consequences of fault conditions other than those specified as limits for the protective function comply with the applicable protective system criteria.
Protective Actions
The Reactor Protection System automatically trips the reactor when the applicable conditions listed in Table 7.2-1 exist. Interlocking functions of the Reactor Protection System prevent control rod withdrawal when certain specified parameters reach values less than the values at which reactor trip is initiated.
7.2-10 Revised 08/17/2016 For anticipated abnormal conditions, protective systems in conjunction with inherent characteristics and engineered safety features are designed to
assure that limits for energy release to the containment and for radiation
exposure (as in 10 CFR 50.67) are not exceeded.
Indication
All transmitted signals (flow, pressure, temperature, etc) which can lead to
a reactor trip are indicated and/or recorded for every channel.
All nuclear flux power range currents (top detector, bottom detector and
algebraic difference and average of bottom and top detector currents) are
indicated and/or recorded.
Annunciators are also used to alert the operator of deviation from normal
operating conditions so that he may take corrective action to avoid a reactor
trip. Further, actuation of any rod stop or trip of any reactor trip channel
will actuate an annunciator.
Digital Data Processing System (DDPS)
Various plant signals are connected to the Digital Data Processing System (DDPS), which is integrated in plant Distributed Control System (DCS).
Information is displayed at consoles provided for the reactor control
operators in the control room.
The DDPS provides the following information:
- 1. Sequence of events.
- 2. Data collection and limited processing for:
- a. Heat rate determination.
- b. Calorimetric reactor output measurement.
- c. Reactor core analysis.
- d. Primary Coolant System Loose Parts Vibration.
- e. Auxiliary Feedwater Pump Parameters 3. Data collection and storage for post trip review.
Information for sequence of events is printed on a printer, located in the
control room.
7.2-11 Revised 04/17/2013 C26C26C26 Distributed Control System / Safety Parameter Display System / Emergency Response Data Acquisition and Display System
The Safety Parameter Display System (SPDS) / Emergency Response Data Acquisition and Display System (ERDADS), Which is integrated in the plant
Distributed Control System (DCS), consists of plant process and environmental
signals that provide an electronic display of plant parameters, from which
the safety status of plant operation may be determined in the control room, Technical Support Center (TSC) and Emergency Operations Facility (EOF). The
primary function of the Safety Parameter Display System (SPDS) is to aid
operating personnel in the control room in making rapid assessments of the
status of plant safety. Duplication of the SPDS / DCS displays in the
Technical Support Center and Emergency Operating Facility improves the
communication between these facilities and the control room and assists
corporate and plant management in the recovery decision-making process.
The Emergency Response Data Acquisition and Display System (ERDADS), which
includes the Safety Parameter Display System, is a real time computer based
data acquisition and display system designed to assist control room personnel
in evaluating the safety status of the plant. The ERDADS aids in the
coordinated control of the reactor during upset conditions, while
concurrently providing information of concern to the public. The SPDS
includes a set of predetermined displays designed to yield relevant, timely, accurate, and unambiguous information to the control room operators, the
technical support advisors, and the offsite public safety officials. The
SPDS / DCS displays a small but critical subset of the parameters available
in the control room, thus reducing the problems associated with information
overload and parameter selection. At the same time, by preselecting and
grouping critical parameters for each display, the SPDS / DCS facilitates
comprehension of the prevailing plant and public safety conditions. This is
achieved by presenting high-level displays which summarize plant safety
function status, plant system performance, and radiological and
meteorological data. Printers and plotters are available for hard copy
reports. For details on ERDADS refer to Section 7.5.4.
7.2-12 Revised 04/17/2013 C26C26C26C26 7.2.2 SYSTEM DESIGN
Reactor Protection System Description
Figure 7.2-1 illustrates typical core limits and shows the maximum trip
points which are used for the protection system. The solid lines indicate a
typical locus of DNBR equal to the safety analysis limit value (in this
example, 1.30) at four pressures, and the dashed lines indicate maximum
permissible trip points for the overtemperature T reactor trip. Actual setpoints (the final setpoints will be given in the Technical Specifications)
are lower to allow for measurement and instrumentation errors. The overpower T reactor trip limits the maximum core power independent of the DNBR.
Adequate margins exist between the maximum nominal steady state operating
point (which includes allowances for temperature, calorimetric, and pressure
errors) and required trip points to preclude a spurious trip during design
A block diagram of the Reactor Protection System showing various reactor trip
functions and interlocks is shown in Figure 7.2-2.
System Safety Features
Separation of Redundant Protection Channels
The Reactor Protection System is designed to achieve separation between
redundant protection channels. The channel design is applied to the process
and the logic portions of the protection system, and is shown in Figure 7.2-
3A. Also shown in Figure 7.2-3B is the configuration for the Engineered
Safety Features Actuation Logic. The reactor trip on loss of 4160V Bus
voltage and underfrequency (Trip of RCP breaker) differs from the typical RPS
scheme shown in Figure 7.2-3A. They are illustrated by Figure 7.2-8c.
Separation of redundant process channels originates at the process sensors
and continues along the field wiring and through containment penetrations to
the process protection racks. Isolation of field wiring is achieved using
separate wireways, cable trays, conduit runs and containment penetrations for
each redundant channel.
7.2-13 Revised 04/17/2013 Process equipment is isolated by locating redundant components in different protection racks. Each channel is energized from a separate AC power feed.
Logic equipment separation is achieved by providing separate racks, each
associated with individual trip breakers. Physical separation is provided
between these racks.
The reactor trip comparators are mounted in the process protection racks and
are the final operational component in a process protection channel. Each
comparator drives two logic relays ("X-A" & "X-B"). The contacts from the "X-A" relays are interconnected to form the required actuation logic for Trip
Breaker No. A. The transition from channel identity to logic identity is
made at the logic relay coil/relay contact interface. As such, there is both
electrical and physical separation between the process and the logic portions
of the protection system. The above logic network is duplicated for Trip
Breaker No. B using the contacts from the "X-B" relays. Therefore, the two
redundant reactor trip logic channels will be physically separated and
electrically isolated from one another. The Reactor Protection System is
comprised of identifiable channels which are physically, electrically and
functionally separated from one another.
Loss of Power
With the exception of Emergency Trip Header Pressure (low), a loss of AC
power to any RPS logic relay (Reactor Trip Comparator Output) causes the
affected channel to trip. Emergency Trip Header Pressure (low) is designed
to energize the associated logic relays (to trip). A loss of DC control
power to the RPS logic matrix results in a reactor trip.
A loss of AC power to any ESF logic relay causes the affected channel to
trip. Availability of DC control power to the logic matrix is required for
train operability. Availability of DC control power to the ESF logic matrix
is continuously monitored and annunciated in the control room.
Containment pressure (High-High coincident with High) differs from any other
ESF functions in that the channel and train relays may utilize common DC
power sources. No single failure of the DC power sources will result in an
inadvertent actuation or render the system inoperable. Availability of DC
control to the CIS channels and logic matrix is continuously monitored and
annunciated in the control room.
7.2-14 Revised 04/17/2013 C26 Reactor Trip Signal Testing Provisions are made for process variables to manually place the output of the
comparators in a tripped condition for "at power" testing of all portions of
each trip circuit including the reactor trip breakers. Administrative
procedure requires that the final element in a trip channel (required during
power operation) is placed in the trip mode before that channel is taken out
of service for repair or testing, so that the single failure criterion is met
by the remaining channels. In the source and intermediate ranges where the
trip logic is one-out-of-two for each range, bypasses are provided for this
testing procedure.
Nuclear instrument power range channels are tested by superimposing a test
signal on the sensor signal so that the reactor trip protection is not
bypassed. Based upon coincident logic (2/4) this will not trip the reactor;
however, a trip will occur if a reactor trip is required.
Channel bypass capability exists for Eagle 21 (overpower T, overtemperature T and Hi Pressurizer level).
Provision is made for the insertion of test signals in each process loop.
Verification of the rack component response is made by portable instruments
at test points specifically provided for this purpose. This enables testing
and calibration of meters and comparators. Redundant sensor readouts are
checked against each other during normal power operation to monitor
transmitter performance. Provisions have been made for transmitter
calibrations using precision read-out equipment during normal power operation
if deemed necessary.
Process Channel Testing
The basic elements of a process protection channel are shown in Figure 7.2-4.
Rack door alarms are arranged on a protection channel basis to annunciate
entry to more than one redundant protection channel at any time. Each
process protection rack includes a test panel containing those switches, test
jacks and related equipment needed to test the channels contained in the
rack. A hinged cover encloses the test panel. Opening the cover or placing
the test-operate switch in the "TEST" position will initiate an alarm. The
test panel cover is designed such that it cannot be closed (and the alarm
cleared) unless the test signal plugs (described below) are removed.
7.2-15 Revised 04/17/2013 Closing the test panel cover will mechanically return the test switches to the "OPERATE" position. Each digital protection rack includes a test panel, which is used to interface with a portable Man Machine Interface (MMI) test
cart.
Administrative procedures will require that the bistable trip switch, in the
channel under test, be placed in the tripped mode prior to test. This places
a proving lamp across the comparator output so that the comparator trip point
can be checked during channel surveillances and calibration. The comparator
trip switches must be manually reset after completion of a test. Closing the
test panel cover will not restore these switches to the untripped mode.
However, the annunciator on the RTG board cannot be reset until these
comparators are returned to the untripped mode.
Administrative procedures allow the nuclear instrumentation source range and
intermediate range protection channels to be placed in bypass during periodic
testing. Annunciation is provided whenever the NIS (Source Range and
Intermediate Range) bypass selector switch is placed in bypass. Power range
overpower protection is not disabled since this function is not affected by
the testing of circuits. Channel bypass capability exists for Eagle 21 (overpower T, overtemperature T and Hi Pressurizer level). Annunciation is provided whenever any of the Eagle 21 trips are placed in bypass.
Administrative procedures also allow the power range dropped-rod annunciation
to be placed in bypass during testing. Annunciation is provided whenever the
power range dropped rod and rod stop protection bypass selector switch is
placed in bypass. In addition, the rod position system would provide
indication and associated corrective actions for a dropped rod condition.
Channel calibration consists of inserting a test signal from an external
source into the test signal injection point. Where applicable, the channel
power supply will serve as a power source for the calibration, which permits
verifying the output load capacity of the power supply. Test points, located
in the process channel, provide independent means of measuring the output of
the calibration components.
Logic Channel Testing
The general design features of the logic system are described below. The
trip logic channels for typical two-out-of-three RPS and ESFAS trip functions
are shown in Figures 7.2-3A and 7.2-3B. The typical RPS and ESFAS logic
relay testing configurations are shown in Figure 7.2-6A and 7.3-6B. Each
comparator drives two relays.
7.2-16 Revised 04/17/2013 Contacts from the Train "A" relays are arranged in a 2/3 or 2/4 trip matrix for Trip Breaker A. The above configuration is duplicated for Trip Breaker B
using contacts from the Train "B" relays. A series configuration is used for
the trip breakers since they are actuated (opened) by undervoltage coils and
shunt trip relays. This approach is consistent with a de-energize-to-trip
preferred failure mode. The logic system testing includes exercising the
reactor trip breakers to demonstrate system integrity. By-pass breakers are
provided for this purpose. During normal operation, these by-pass breakers
are open. Administrative procedures will be used to minimize the amount of
time these breakers are closed. An interlock is provided to preclude the
closing of both bypass breakers (Train A and B). Indication of a closed
condition of either by-pass breaker is provided locally on the test panel and
is annunciated in the control room.
As shown in Figure 7.2-5, the trip signal from the logic network is
simultaneously applied to the main trip breaker associated with the specific
logic chain as well as the by-pass breaker associated with the alternate trip
breaker. Should a valid trip signal occur while by-pass breaker (BYA)is
by-passing reactor trip breaker (RTA), RTB will be opened through its
associated logic train. The trip signal applied to RTB is simultaneously
applied to BYA thereby opening the by-pass around BYA. BYA would either have
been opened manually as part of the test or would be opened through its
associated logic train which would be operational or tripped during a test.
An auxiliary relay is located in parallel with the undervoltage coils of the
trip breakers. This relay is connected to ERDADS which can provide a
sequence of events printout which is used to indicate transmission of a trip
signal through the logic network during testing. Lights are also provided to
indicate the status of the logic relays.
Two shunt trip relays are connected in parallel with the undervoltage coil.
These relays provide additional assurance for opening the trip breakers on an
automatic trip signal by energizing the breaker trip coil (i.e., shunt trip
attachment).
The following procedure illustrates the method used for testing Reactor Trip
Breaker A and its associated logic network.
- a. With the BYA in the test position, close and trip BYA to verify operation.
7.2-17 Revised 04/17/2013
- b. Rack-in and close BYA. Test the undervoltage and shunt trips (independently) for RTA.
- c. Sequentially de-energize the trip relays for each logic combination (1-2, 1-3,2-3). Verify that the logic network de-energizes the
undervoltage coil on RTA for each logic combination. When the
appropriate logic is actuated, the signal applied to the undervoltage
coil is verified by use of the test panel test lights. (Note:
operation of the shunt trip attachment is tested independently of the
undervoltage coil).
- d. Repeat "C" for every logic combination in each matrix.
- e. Reset RTA. Trip and rack-out BYA.
In order to minimize the possibility of operational errors (such as tripping
the reactor inadvertently or only partially checking all logic combinations)
each logic network includes a logic channel test panel. This panel includes
those switches, test lights and recorders needed to perform the logic system
test. This arrangement is illustrated in Figure 7.2-7. The test switches
used to de-energize the trip comparator relays operate through inter-posing
relays as shown in Figure 7.2-4 and 7.2-6. This approach avoids violating
the separation philosophy used in the process channel design. Thus, although
test switches for redundant channels are conveniently grouped on a single
panel to facilitate testing, physical and electrical separation of redundant
protection channels are maintained by the inclusion of the interposing relay
which is actuated by the logic test switches.
7.2-18 Revised 04/17/2013 Primary Power Source The primary power sources for the Reactor Protection System are the
instrument buses described in Section 8. The source of electrical power for
the sensors and the actuation of circuits in the engineered safety features
instrumentation is also from these buses.
Protective Actions
Reactor Trip Description
Rapid reactivity shutdown is provided by the insertion of full length RCC
assemblies by free fall. Duplicate series-connected reactor trip breakers
maintain all power to the full length control rod drive mechanisms. The full
length RCCA must be energized to remain withdrawn from the core. Automatic
reactor trip occurs upon the loss of power to the full length control rods.
The reactor trip breakers are opened by the undervoltage coils on both
breakers. The undervoltage coils (which are normally energized) become
de-energized by any one of the several trip signals. In order to provide
additional assurance of tripping the reactor trip breakers per NRC, Generic
Letter 83-28 Item 4.3 (Reference 4), the reliability of the reactor
protection system is enhanced by a design change to also use the shunt trip
attachments to open the reactor trip breakers automatically. The automatic
shunt trip function is considered safety related. The breaker closing
circuit is electrically separated from the tripping circuit and is considered
non-safety related.
The design of the devices providing signals to the reactor trip breaker
undervoltage trip coils is such as to cause this coil to trip the breaker on
a reactor trip signal or power loss.
Certain reactor trip channels are automatically bypassed at low power where
they are not required for safety. Nuclear source range and intermediate
range trips are specifically provided for protection at low power or
subcritical operation, and at higher power operations they are bypassed by
manual action.
7.2-19 Revised 04/17/2013 During power operation, a sufficient amount of rapid shutdown capability in
the form of control rods is administratively maintained by means of the
control rod insertion limit monitors. Administrative control requires that
all shutdown group rods be in the fully withdrawn position during power
operation.
A list of reactor trips, means of actuation, required setpoints, and the
coincident circuit requirements is given in Table 7.2-1. The interlock
circuits, referred to in Table 7.2-1, are listed in Table 7.2-2.
Manual Trip
The manual actuating devices are independent of the automatic trip circuitry, and are not subject to failures which make the automatic circuitry
inoperable. Either of two manual trip devices located in the control room can
initiate a reactor trip.
High Nuclear Flux (Power Range) Trip
This circuit trips the reactor when two of the four power range channels read
above the trip set-point. There are two independent trip settings, a high
and a low setting. The high trip setting provides protection during normal
power operation. The low setting, which provides protection during startup, can be manually bypassed when two out of the four power range channels read
above approximately 10% power (P10). Three out of the four channels below
10% automatically reinstates the trip function. The high setting is always
active.
High Nuclear Flux (Intermediate Range) Trip
This circuit trips the reactor when one out of the two intermediate range
channels reads above the trip set-point. This trip, which provides
protection during reactor startup, can be manually bypassed if two out of
four power range channels are above approximately 10% (P10): Three out of
four channels below this value automatically reinstates the trip function.
The intermediate channels (including detectors) are separate from the power
range channels.
7.2-20 Revised 04/17/2013 High Nuclear Flux (Source Range) Trip This circuit trips the reactor when one of the two source range channels
reads above the trip set-point. This trip, which provides protection during
reactor startup can be manually bypassed when one of two intermediate range
channels reads above the P6 setpoint value and is automatically reinstated
when both intermediate range channels decrease below this value (P6). This
trip is also bypassed by two out of four high power range signals (P10). The
trip function can also be reinstated below P10 by an administrative action
requiring coincident manual actuation. The trip point is set between the
administrative source range cutoff power level setpoint and the maximum
source range power level.
Overtemperature T Trip The purpose of this trip is to protect the core against DNB. This trips the
reactor on coincidence of two out of the three signals, with one set of
temperature measurements per loop. The set point for this reactor trip is
continuously calculated for each loop by solving the equation provided in
Section 2.2 of the Technical Specifications.
Three of the four power range detectors provide input (one per channel) to
the overtemperature T trip function. Thus, a single failure neither defeats the function nor causes a spurious trip. Changes in f (I) can only lead to a decrease in trip setpoint.
A rod stop is initiated when
T > T rod stop where
T rod stop = T setpoint - B p B P = a set point bias The setpoint bias is set to zero, effectively negating this rod stop.
Overpower T Trip The purpose of this trip is to protect against excessive power level (fuel
rod rating protection). This trips the reactor on coincidence of two out of
the three signals, with one set of temperature measurements per loop.
7.2-21 Revised 04/17/2013 The set point for this reactor trip is continuously calculated for each channel by solving equations provided in Section 2.2 of the Technical
Specifications.
A similar rod stop function is provided for overpower protection. The
setpoint bias is also set to zero, effectively negating this rod stop.
Low Pressurizer Pressure Trip
The purpose of this trip is to protect against excessive core steam voids
which could lead to DNB. This trips the reactor on coincidence of two out of
the three low pressurizer pressure signals. This trip is blocked when three
of the four power range channels and two of two turbine inlet pressure
channels read below approximately 10% power (P7).
High Pressurizer Pressure Trip
The purpose of this trip is to limit the range of required protection from
the overtemperature T trip and to protect against Reactor Coolant System overpressure. The reactor is tripped on coincidence of two out of the three
high pressurizer pressure signals.
High Pressurizer Water Level Trip
This trip is provided as a backup to the high pressurizer pressure trip. The
coincidence of two out of the three high pressurizer water level signals
trips the reactor. This trip is blocked when three of the four power range
channels and two of two turbine inlet pressure channels read below
approximately 10% power (P7).
7.2-22 Revised 04/17/2013 C26C26 Low Reactor Coolant Flow Trip This trip protects the core from DNB following a low flow or loss of coolant
flow. The means of sensing low flow and a loss of coolant flow accident are
as follows:
- a. Measured low flow in the reactor coolant piping.
The low reactor flow trip is actuated by the coincidence of 2/3 signals for any reactor coolant loop. The loss of flow in any two loops causes
a reactor trip in the power range above approximately 10% (P7). Above
approximately 45% power (P8), the loss of flow in any loop causes a
reactor trip. The flow measurement utilizes an elbow tap which is
discussed in Section 4.2.
- b. Monitored electrical supply to the reactor coolant pumps
The voltage and frequency of the buses which supply power to the reactor coolant pumps is monitored. Under voltage on both buses on either Train
A or B logic will cause a reactor trip above approximately 10% power (P-7). Under frequency will cause a pump breaker trip which then will
cause a reactor trip as follows:
- 1) Above approximately 10% power a loss of 2 of the 3 pumps will cause a trip (P-7).
- 2) Above approximately 45% power a loss of 1 of the 3 pumps will cause a trip (P-8).
Safety Injection System (SIS) Actuation Trip
A reactor trip occurs when the safety injection system is actuated by signals
as listed in Table 7.2-1.
7.2-23 Revised 04/17/2013 Turbine Generator Trip A turbine trip is sensed by two out of three signals from Emergency Trip Header pressure or 2/2 stop valves closed. A turbine trip causes a direct reactor trip above approximately 10% power (P7) and a controlled short term release of steam to the condenser which removes sensible heat from the reactor coolant system and thereby avoids steam generator safety valve actuation.
The turbine control system automatically trips the turbine generator under any of the following conditions:
- a. Turbine overspeed
- b. Generator lock-out
- c. Low condenser vacuum
- d. High thrust bearing wear e. Low bearing oil pressure
- f. Reactor trip
- g. Manual trip
- h. AMSAC signal
- i. High-High steam generator level j. Safeguards actuation
Steam/Feedwater Flow Mismatch Trip
This trip protects the reactor from a sudden loss of its heat sink. The trip is actuated by a steam/feedwater flow mismatch (1/2) in coincidence with low water level (1/2) in any steam generator.
Low-Low Steam Generator Water Level Trip
The purpose of this trip is to protect the steam generators in the case of a sustained steam/feedwater flow mismatch of insufficient magnitude to cause a flow mismatch reactor trip. The trip is actuated on two out of the three (2/3) low-low water level signals in any steam generator.
Rods Stops
Rod stops are added to prevent a reactor trip or prevent an abnormal condition from increasing in magnitude.
7.2-24 Revised 08/17/2016 C28 A list of rod stops is given in Table 7.2-3. Some of these have been previously noted under permissive circuits, but are listed again for
completeness.
Rod Drop Detection
Two independent systems are provided to sense a dropped rod, (1) rod position
system rod bottom bistables and (2) nuclear instrumentation power range
circuits which sense sudden reduction in out-of-core neutron flux. These
systems are not reactor protection systems.
A dropped RCCA would be detected by the rod bottom signal derived for each
rod from its individual position indication system. With the position
indication system, initiation of action is not dependent on location, reactivity worth or power distribution changes.
Backup is provided by use of the out-of-core power range nuclear detectors
and is particularly effective for larger nuclear flux reductions occurring in
the region of the core adjacent to the detectors.
The rod drop detection circuit from nuclear flux consists basically of a
comparison of each of the four ion chamber signals with the same signal taken
through a first order lag network. Since a dropped RCC assembly will rapidly
depress the local neutron flux, the decrease in flux will be detected by one
or more of these circuits.
Such a sudden decrease in ion chamber current will be seen as a changed
channel level. A negative signal output greater than a preset value (approximately 5 percent) from any one of the four power range channels will
initiate the rod drop annunciation. Automatic rod withdrawal by the reactor
control system has been permanently disabled. Manual rod withdrawal is not
blocked by nuclear instrumentation system power range rod drop detection.
Figure 7.4-2b indicates schematically the Nuclear Instrumentation System,
including the dropped RCCA alarm.
7.2-25 Revised 04/17/2013 Control Group Rod Insertion Monitor The control group rod insertion limits, Z LL , are calculated as a linear function of power and reactor coolant average temperature. The equation is:
Z LL = A (T)avg + B (T avg) + C where A, B are preset manually adjustable gains and C is a preset manually
adjustable bias. The (T)avg and (T avg) are the average of the individual temperature differences and the coolant average temperatures respectively
measured from the reactor coolant hot leg and the cold leg.
An insertion limit monitor with two alarm set points is provided for the
control banks. A description of control and shutdown rod groups is provided
in Section 7.3. The "Low" alarm alerts the operator of an approach to a
reduced shutdown reactivity situation requiring boron addition by following
procedures with the Chemical and Volume Control System. If the actuation of
the "Low-Low" alarm occurs, the operator should take immediate action to add
boron to the system.
Setpoint Methodology The nominal trip setpoints (NTS) for the reactor trip system and engineering safety features are provided in Table 7.2-1. The NTS values are the Limiting Safety System Setting (LSSS) values that are calculated based on limits derived from the safety analyses and process instrumentation and adjusted to account for the specific instrument uncertainties. The instrument uncertainties for the trip setpoints affected by the EPU are based on the methodology described in WCAP-17070P, Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4 (Power Uprate to 2644 MWt -
Core Power) (Reference 5). The guidance of Technical Specification Task Force (TSTF) No. 493, Rev. 4, Option A, "Clarify Application of Setpoint Methodology for LSSS Functions," (Reference 6) is applied to the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) setpoints and surveillance requirements impacted by EPU.
EPU impacted RTS functions include power range high neutron flux, Overtemperature AT, Overpower AT, reactor coolant low flow, steam generator low-low water level, steam/feedwater flow mismatch coincident with steam generator low water level, and turbine trip on emergency trip header pressure (Table 7.2-1 items 2, 3, 4, 8, 13, 12, & 11, respectively).
7.2-26 Revised 04/17/2013 C26 EPU impacted ESFAS functions include safety injection on high steam line flow coincident with low steam generator pressure, steam line isolation on high steam line flow coincident with low steam generator pressure, feedwater isolation on high-high steam generator water level, and auxiliary feedwater actuation on low-low steam generator water level (Table 7.2-1 items 19e, 22, 26a, & 25a, respectively). The setpoint methodology establishes the NTS and Allowable Value (AV) for each of the affected functions. The AVs at Turkey Point are "performance based" and are determined by adding (or subtracting) the rack calibration accuracy (RCA) of the device tested during the Channel Operational Test (COT) to the NTS in the non-conservative direction, i.e., toward or closer to the Safety Analysis Limit (SAL) for the application. See Figure 7.2-8g for an illustration of setpoint relationships between SAL, Channel Statistical Allowance (GSA), RCA, As-Found Tolerance (AFT), and As-Left Tolerance (ALT) are shown and AFT=ALT=RCA where the RCA uncertainty term is based on equipment manufacturer's performance specifications.
Surveillance limits are established to verify that RTS and ESFAS instrumentation operates within the boundaries of applicable instrument uncertainty calculations. These limits are implemented in plant procedures in accordance with TS Notes (a) and (b) below which are consistent with the wording provided in TSTF-493 Rev 4. These notes specify operability criteria and require that out-of-tolerance conditions detected during surveillances be evaluated before returning the channel to service. The notes have been inserted into TS Table 4.3-1, RTS Instrumentation Surveillance Requirements and TS Table 4.3-2, ESFAS Instrumentation Surveillance Requirements. The methods used to determine the NTS and AV values and summaries of the associated calculations are described in WCAP-17070-P (Reference 5).
Note (a) states: "If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service." Note (b) states: "The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in UFSAR Section 7.2." 7.2-27 Revised 04/17/2013 C26 7.2.3 SYSTEM EVALUATION
Reactor Protection System and DNB
The following is a description of how the reactor protection system prevents
DNB.
The variables affecting the DNB ratio are:
Thermal Power Coolant flow Coolant temperature Coolant pressure Core power distribution (hot channel factors)
Figure 7.2-1 illustrates the typical core safety limits for which DNBR for
the hottest fuel rod is equal to the safety analysis limit value (in this
example, 1.30) and shows the overpower and overtemperature T reactor trips locus as a function of T avg and pressure. This illustration is derived from the inlet temperature versus power relationships. Figure 7.2-9b illustrates
T avg control and insertion limit alarms and is typical for one reactor coolant loop. Figure 7.2-9a illustrates the rod control system.
Reactor trips for a fixed high pressurizer pressure and for a fixed low
pressurizer pressure are provided to limit the pressure range over which core
protection depends on the overpower and overtemperature T trips.
Reactor trips on nuclear overpower and low reactor coolant flow are provided
for direct, immediate protection against rapid changes in these parameters.
However, for all cases in which the calculated DNBR approaches the safety
analysis limit value, a reactor trip on overpower and/or overtemperature T would also be actuated.
The Reactor Protection System actuates a reactor trip for a set of conditions
for which the calculated DNBR for the worst fuel rod approaches the safety
analysis limit value. Because of the statistical nature of the DNB
correlation used and the statistical makeup of a portion of the hot channel
factors, there exists a finite probability of a few rods being in DNB for a
calculated ratio equal to the safety analysis limit value for the worst fuel
rod (Section 3.2.2).
7.2-28 Revised 04/17/2013 For the anticipated abnormal conditions, it is highly unlikely that the exact
combination of conditions (reactor coolant pressure, temperature and core
power, instrumentation inaccuracies, etc.) that cause a DNBR equal to the
safety analysis limit value will be approached before a reactor trip. The
simultaneous loss of power to all of the reactor coolant pumps is the
accident condition most likely to approach the DNBR limit value for the
calculated worst fuel rod. In any event the DNBR at the worst fuel rod is
near the limit value for only a few seconds.
Typically, the hottest fuel rods are not adjacent to one another. They are
located near the RCCA thimbles. Fuel rods located in the immediate vicinity
of the hottest fuel rod have a DNBR higher than that rod.
In the event of a difference between the upper and lower power range detector
signals that exceeds the desired range, automatic feedback signals are
provided to reduce the overtemperature trip setpoint and block rod
withdrawal.
Specific Control and Protection Interactions
Four power-range nuclear flux channels are provided for overpower protection.
Isolated output from one of these channels is used for automatic control rod
regulation of power. If any channel fails in such a way as to produce a low
output, that channel is incapable of proper overpower protection.
Two-out-of-four overpower trip logic will ensure an overpower trip, if
needed, even with an independent failure in another channel.
A rapid decrease of any nuclear flux signal will annunciate rod drop. An
overpower signal from any nuclear channel will block manual rod withdrawal.
The set point for this rod stop is below the reactor trip set point.
Coolant Temperature
Each overtemperature-overpower protection channel calculates T avg and T based on the temperature measurements from the associated RCS loop. The median T avg signal (of the three separate channels) is used for automatic control rod
regulation of power and temperature. Two out of three (2/3) trip logic is
used to ensure that a trip occurs, if needed, even with an independent
failure in another channel.
7.2-29 Revised 04/17/2013 Manual rod withdrawal blocks will occur if any one of four nuclear channels indicates an overpower condition or if any two of three overtemperature or
overpower channels exceed the trip setpoint.
Finally, as shown in Section 14.1, the combination of trips on nuclear
overpower, high pressurizer water level, and high pressurizer pressure also
serve to limit an excursion for any rate of reactivity insertion.
Pressurizer Pressure
Three pressure channels are used for high and low pressure protection and as
part of overtemperature protection (See Figures 7.2-11a and 7.2-11b).
Pressure control is accomplished by spray, power-operated relief valves, and
heaters which are controlled by output signals from two separate pressure
control channels. The pressurizer safety valves are adequately sized to
prevent system overpressure.
a) Low Pressure
A spurious high pressure signal from the control channel can cause low RCS pressure by spurious actuation of spray and/or a relief
valve. Additional redundancy is provided in the protection system
to ensure underpressure protection, i.e., two-out-of-three low
pressurizer pressure reactor trip logic and two-out-of-three low
pressurizer pressure safety injection logic.
b) High Pressure
The pressurizer heaters are incapable of overpressurizing the reactor coolant system. Maximum steam generation rate with
heaters is about 13,000 lbs/hr., compared with a total capacity of
941,478 lbs/hr. for the three safety valves and a total capacity
of 420,000 lbs/hr. for the two power-operated relief valves.
Therefore, additional redundancy for overpressure protection is
not required for a pressure control failure. Two-out-of-three high
pressurizer pressure trip logic is therefore used.
In addition, either of the two relief valves can easily maintain pressure below the high pressure trip point. The two relief
valves are controlled by independent pressure channels, one of
which is independent of the pressure channel used for heater
control. Finally, the rate of pressure rise achievable with
heaters is slow, and ample time and pressure alarms are available
for operator action.
7.2-30 Revised 04/17/2013 C26 Pressurizer Level
Three pressurizer level channels in a two-out-of-three logic (2/3) for high
pressurizer level are used for reactor trip. This function is not relied
upon as a primary trip function in the plant safety analysis. It may perform
as a backup trip for any significant heatup transient which results in a
large specific volume change for the RCS primary coolant. Isolated output
signals from these channels are used for volume control, increasing or
decreasing water level.
A level control failure could fill or empty the pressurizer at a slow rate (on the order of half an hour or more). Therefore, ample time and alarms
exist for operator action in the event of increasing or decreasing water
level in the pressurizer. (See Figure 7.2-12).
(a) High Level
A reactor trip on pressurizer high level is provided to prevent rapid thermal
expansions of reactor coolant fluid from filling the pressurizer: the rapid
change from high rates of steam relief to water relief can be damaging to the
safety valves and the relief piping and pressure relief tank. However, a
level control failure cannot actuate the safety valves because the high
pressure reactor trip is set below the safety valve set pressure. With the
slow rate of charging available, overshoot in pressure before the trip is
effective is much less than the difference between reactor trip and safety
valve set pressures. Therefore, a control failure does not require
protection system action.
In addition, ample time and alarms are available for operator action.
(b) Low Level
Ample time and alarms exist for operator action in the event of a decreasing water level in the pressurizer.
Steam Generator Water Level; Feedwater Flow
Before describing control and protection interaction for these channels, it
is beneficial to review the protection system basis for this instrumentation.
(See Figure 7.2-13).
7.2-31 Revised 04/17/2013 The basic function of the reactor protection circuits associated with low steam generator water level and low feedwater flow is to preserve the steam
generator heat sink for removal of long term residual heat. Should a
complete loss of feedwater occur with no protective action, the steam
generators would boil dry. and cause an overtemperature-overpressure
excursion in the reactor coolant. Reactor trips on temperature and pressure
will trip the unit before there is any damage to the core or reactor coolant
system. Residual heat would cause thermal expansion after trip and discharge
of the reactor coolant to the pressurizer relief tank through the pressurizer
relief valves.
Redundant auxiliary feedwater pumps are provided to prevent this. Reactor
trips act before the steam generators are dry to reduce the required capacity
and starting time requirements of these pumps and to minimize the thermal
transient on the reactor coolant system and steam generators. Independent
trip circuits are provided for each steam generator for the following
reasons:
- 1. Should severe mechanical damage occur to the feedwater line to one steam generator, it is difficult to ensure the functional integrity of level
and flow instrumentation for that unit. For instance, a major pipe
break between the feedwater flow element and the steam generator would
cause high flow through the flow element. The rapid depressurization of
the steam generator would drastically affect the relation between
downcomer water level and steam generator water inventory.
- 2. It is desirable to minimize thermal transient on a steam generator for credible loss of feedwater accidents.
It should be noted that controller malfunctions caused by a protection system failure affect only one steam generator. Also, they do not
impair the capability of the main feedwater system under either manual
control or automatic control. Hence, these failures are far from being
the worst case with respect to decay heat removal with the steam
generators.
(1) Feedwater Flow
The feedwater flow signal is monitored by the control system for sudden changes such that a spurious high signal from the feedwater flow
channel being used for control would not cause a significant reduction
in feedwater flow. The feedwater controller will reject to MANUAL when
the spurious signal is detected.
7.2-32 Revised 04/17/2013 C26 This condition is alarmed such that the failure is promptly detected. The spurious high signal will prevent that channel from tripping from steam/feedwater flow mismatch coincident with low
steam generator level. A reactor trip on steam generator low-low
water level, independent of indicated feedwater flow, will ensure a
reactor trip if needed.
In addition, the three-element feedwater controller incorporates reset on level, such that with expected controller settings a rapid
increase in the flow signal would cause only a small decrease in
level before the controller re-opened the feedwater valve. A slow
increase in the feedwater signal would have no effect at all.
(2) Steam Flow
A spurious low steam flow signal would have the same effect as a high feedwater signal, discussed above.
(3) Level
The level signal is monitored by the control system for sudden changes such that a spurious high water level signal from the
protection channel used for control will not close the feedwater
control valve; instead, the feedwater controller will reject to
MANUAL when the spurious signal is detected. This condition is
alarmed such that failure is promptly detected. This level channel
is independent of the level channels used for reactor trip on
steam/feedwater flow mismatch coincident with low steam generator
level.
a) A rapid increase in the level signal will reject the feedwater controller to MANUAL and generate an alarm. If the alarm is not
properly responded to this will lead to an actuation of a reactor
trip on steam/feedwater flow mismatch coincident with low level.
b) A slow increase in the level signal may not actuate a low feedwater signal. Since the resulting level decrease is slow, the operator has time to respond to low level alarms. Since only
one steam generator is affected, automatic protection is not
mandatory and reactor trip on two-out-of-three low-low level is
acceptable.
7.2-33 Revised 04/17/2013 C26C26C26C26 (4) Median T avg is used as an index to select gain and reset tuning parameters for feedwater control at low power levels. A spurious change to T avg from a protection channel will cause the T avg median signal selector to select the median channel and a small change in T avg. Small changes in T avg will have minimal impact on tuning parameters and no adverse effects on feedwater control.
Steam Line Pressure (Hi Steam Line Flow)
High steam flow in 2 out of 3 steam generators coincident with low steam line
pressure in 2 out of 3 steam lines or Lo-T avg in 2 out of 3 loops will actuate safety injection and close the main steam isolation valves (steam break
protection).
Normal Operating Environment
The control room is maintained at the personnel comfort level of (70 +
10)o F. Protective equipment inside the room is designed to operate within design tolerance over this temperature range and will perform its protective
function in an ambient of 120 o F and 95% relative humidity (i.e., there will be no loss-of-function in an ambient temperature of 120 o F).
The operating environment for equipment within the containment will normally
be controlled to less than 120 o F. Operation with elevated normal bulk containment temperatures up to 125 o F for short periods of time during the summer months has been evaluated and is acceptable; refer to Section 14.0.
The Reactor Protective System instrumentation within the containment is
designed for continuous operation. The temperature of the out-of-core
neutron detectors is maintained at or below 135 o F by the normal containment air cooling system. The detectors are designed for continuous operation at
135 o F and will withstand operation at 175 o F for short durations.
Typical test data (or reasonable engineering extrapolation based on test
data) will be used to verify that protection systems equipment will meet, on
a continuing basis, the functional requirements under the anticipated normal
ambient conditions.
7.2-34 Revised 04/17/2013 C26 7.2.4 ATWS MITIGATING SYSTEM ACTUATION CIRCUITRY (AMSAC)
An Anticipated Transient Without Scram (ATWS) event is an operational
transient (e.g., loss of load, loss of feedwater, loss of off-site power)
followed by a failure of the Reactor Protection System (RPS) to shutdown the
reactor. Title 10 CFR 50.62 requires that all pressurized water reactors
have backup equipment, diverse from the RPS, to automatically
initiate the Auxiliary Feedwater System and turbine trip under conditions
indicative of an ATWS event.
This requirement has been satisfied by the addition of ATWS Mitigating System
Actuating Circuitry (AMSAC), which in addition to the requirements of 10 CFR
50.62 to automatically initiate the Auxiliary Feedwater System and trip the
turbine, will trip the control rod MG set output breakers which will trip the
reactor. AMSAC serves as a non-safety related backup protective system to
the RPS by preventing overpressurization of the Reactor Coolant System, conservation of steam generator inventory, and insertion of the reactor
control rods following an ATWS event. AMSAC actuation logic is shown in
Figure 7.2-14a.
AMSAC is initiated when low steam generator level is sensed and the RPS fails
to respond with an automatic reactor trip. The AMSAC nominal trip setpoint
is based on the low steam generator level RPS safety analysis limit (4%) plus
an allowance for the total loop uncertainty of the AMSAC steam generator
level input signals. A low level on two of three steam generators for both
Channels I and II with turbine power greater than 40% (minus an allowance for
the total loop uncertainty), as indicated by turbine inlet pressure, is
required for AMSAC to initiate. The 1-5 volt input signals to AMSAC are
obtained from the voltage drop across the existing 250 ohm test point
resistors in the 4-20 milliampere secondary loops for Channels I and II steam
generator level and Channels III and IV turbine inlet pressure. Qualified
isolators are used in addition to the existing secondary loop isolators to
provide for electrical isolation between AMSAC and RPS circuitry in
accordance with the requirements for the Safety Evaluation Report for the
Westinghouse Owners Group Topical Report WCAP-10858 "AMSAC Generic Design
Package".
The logic for AMSAC is developed using two microprocessors (A & B) with
Channel I steam generator level input signals aligned to Microprocessor A and
Channel II steam generator level input signals aligned to Microprocessor B.
7.2-35 Revised 04/17/2013 C26C26C26 The inlet turbine pressure input signals are aligned to both microprocessors.
Normally, both microprocessors must be in service for AMSAC to be
operational; however, a "processor selector switch" is provided on the AMSAC
panel that allows for a single microprocessor mode of operation to facilitate
microprocessor maintenance without loss of AMSAC. In addition, AMSAC can be
completely bypassed by placing the "normal/bypass switch", located on the
AMSAC panel, into the bypass position. The microprocessors perform periodic, self-diagnostic testing to enhance the overall reliability of the system and
are designed in a fault-tolerant configuration that reduces the possibility
of inadvertent actuation.
The input isolators are powered from a vital uninterruptible instrumentation
power source, either 3P08 (Unit 3) or 4P08 (Unit 4). The microprocessors are
powered from a non-vital uninterruptible instrumentation power source, either
3P31 (Unit 3) or 4P31 (Unit 4). A loss of power to the isolators, the
microprocessors, or the input signal loops will disable AMSAC.
The output signals from AMSAC generate turbine trip, reactor trip and
auxiliary feedwater initiation. The AMSAC signal energizes the auxiliary
feedwater auto-start relays, which open the steam supply motor operated
valves to admit steam to the auxiliary feedwater pump steam turbines, open
the auxiliary feedwater trip and throttle valves (if electrically closed),
and close the steam generator blowdown and sampling isolation valves.
Qualified relays are used as isolation devices between the non-safety related
AMSAC output modules and the safety related auxiliary feedwater auto-start
relays. The AMSAC signal energizes the turbine trip solenoids to generate a
turbine trip. In addition, the AMSAC signal energizes the control rod MG set
output breakers trip coil. Tripping the breakers causes a loss of power to
the control rod drive mechanisms causing insertion of the control rods.
Since the turbine trip solenoid circuits and control rod MG set output
breaker circuits are non-safety related, electrical isolation from AMSAC is
not required.
There is no manual initiation capability available for AMSAC, since manual initiation of turbine trip, auxiliary feedwater and reactor trip is currently
available. The AMSAC signal can be reset from both the main control room at
panels, 3C04 (Unit 3) and 4C04 (Unit 4), and the AMSAC panels, 3C391 (Unit 3)
and 4C391 (Unit 4). Main control room indication is provided for AMSAC
actuation, AMSAC signal and dual microprocessor mode of operation, and AMSAC bypass.
7.2-36 Revised 04/17/2013 C26 A single annunciator window is provided for alarm of any of the following signals: (1) Low of Voltage (input signal loops, isolators, or processor);
(2) AMSAC Actuated; (3) AMSAC Bypass; or (4) Processor A/B Trouble. Input to
the plant computer (DDPS) for Microprocessor A and/or B actuation is also
provided. Local indication and digital readout is provided to give specific
AMSAC status. The Units 3 and 4 AMSAC panels are located in the Cable
Spreading Room and are seismically qualified and mounted to preclude adverse
affects on safety related components and circuits due to a postulated seismic
event. AMSAC annunciation logic is shown in Figure 7.2-14b.
7.2.5 STEAM GENERATOR OVERFILL PROTECTION
As a result of the technical resolution of the Unresolved Safety Issue (USI)
A-47,"Safety Implication of Control Systems in LWR Nuclear Power Plants," the
NRC concluded that protection should be provided for certain control system
failures and that selected emergency procedures should be modified to assure
that plant transients resulting from control system failures do not
compromise plant safety. The NRC concluded that all PWR plants should
provide automatic steam generator overfill protection, and that plant
procedures and Technical Specifications should include provisions to
periodically verify the operability of the overfill protection and to assure
that automatic overfill protection is available to mitigate main feedwater
overfill events during reactor power operation.
In response to these conclusions, the NRC issued Generic Letter 89-19 (Reference 1),"Request for Action Related to Resolution of Unresolved Safety
Issue A-47...," which requested that licensees incorporate features of the
steam generator overfill protection into plant procedures and plant Technical
Specifications.
In response to NRC Generic Letter 89-19, FPL submitted a proposed license
amendment to the NRC (Reference 2), which addressed the recommendations from
the Generic Letter and revised the Technical Specifications to include
appropriate limiting condition of operation (LCO) and surveillance
requirements for steam generator overfill protection. The NRC approved
Technical Specification changes (Reference 3) implemented the requested
improvements and included the addition of SG high-high level feedwater
isolation signals to Technical Specification Tables 3.3-2, 3.3-3 and 4.3-2
under the heading of "SG Water Level - High-High" along with a corresponding
discussion for Section 4.3 of Technical Specification Bases.
7.2-37 Revised 04/17/2013 Steam generator overfill protection is achieved by utilizing the existing steam generator level high-high signal. The high-high signal is actuated when the level in any steam generator exceeds the high-high setpoint and isolates feedwater by closing the feedwater valves and initiates other associated actions. The instrumentation, setpoints and surveillances already exist, however, they were used for equipment protection.
The steam generator level Protection Channels I, II, and III are designed to combine redundant sensors, independent channel circuitry, coincident trip logic of 2 out of 3, and varied parameter measurement to ensure that a safe and reliable system is provided.
The steam generator overfill protection function is not part of the Engineered Safety Features Actuation System (ESFAS), but was added to the ESFAS Technical Specification tables without modification of the existing design. This function was specifically developed to meet commitments to the NRC criteria contained in Generic Letter 89-19. Although the steam generator overfill protection feature uses much of the same instrumentation as the steam generator low-low trip (reactor trip circuitry), portions of the circuitry for steam generator high-high level overfill protection may not meet all the criteria which apply to ESFAS functions. This is because the steam generator high-high level function was not originally designed to be part of the ESFAS system.
7.2.6 EAGLE 21 PROTECTION SYSTEM Prior to a modification (References 7 and 8) performed on each unit in the early 1990s, reactor coolant temperature measurements used for reactor protection and control were made by Resistance Temperature Detectors (RTD) located in reactor coolant loop bypass manifolds. Due to maintenance and radiation exposure problems associated with the bypass manifolds, a temperature measurement modification was implemented that eliminated the manifold piping and valves and that uses three dual element RTDs mounted in thermowells in each coolant loop.
The modification also included the removal of the analog protection modules and circuits used in the T avg, Delta T, and Pressurizer Level protection functions and replaced them with a digital system (Eagle 21). The Eagle 21 Protection System provides the reactor trip functions of Overpower T, Overtemperature T, and Pressurizer Water Level - High, and the same redundancy as it replaced the analog protection channels on a one-for-one basis. The Pressurizer Water Level - High instrumentation was included in the Eagle 21 modification because two channels were located in the same instrument racks associated with the RTD bypass elimination modification.
7.2-38 Revised 08/17/2016 C28 The Eagle 21 Protection System meets the requirements of IEEE-279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations," and IEEE-323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations," for normal and accident environments. The design verification and validation process is in accordance with Westinghouse Design Standard 408A47, Replacement Hardware Design, Verification and Validation Plan, Revision 3, which is modeled after the guidance in Regulatory Guide 1.152, "Criteria for Programmable Digital Computer System Software in Safety-Related Systems in Nuclear Plants," November 1985 and IEEE/ANSI 7-4.3.2-1982, "Application Criteria for Programmable Digital Computer Systems in Safety Systems of Nuclear Power Generating Stations". WCAP-12632, Revision 1 (Reference 9), describes the application of the Eagle 21 Protection System to the Turkey Point units. WCAP-12374, Revision 1 (Reference 10), is the generic topical report for the Eagle 21 Protection System which provides a more detailed discussion of system design including applicable codes and standards.
7.2-39 Revised 08/17/2016 C28 7.
2.7 REFERENCES
- 1. NRC Generic Letter 89-19,"Request for Action Related to Resolution of Unresolved Safety Issue A-47,`Safety Implication of Control Systems in LWR Nuclear Power Plants' Pursuant to 10 CFR 50.54(f)," dated September 20, 1989.
- 2. FPL letter to the NRC L-93-276,"Proposed License Amendment - Steam Generator Overfill Protection (Generic Letter 89-19)," dated December 28, 1993.
- 3. NRC letter to FPL,"Issuance of Amendments RE: Steam Generator Overfill Protection (TAC NO.s M88560 and M88561)," dated April 28, 1994.
- 4. NRC Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events," July 8, 1983.
- 5. WCAP-17070-P, Rev.1, "Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4 (Power Uprate to 2644 Mwt - Core Power)" January 2011.
- 6. Technical Specification Task Force (TSTF) No. 493, Rev. 4, "Clarify Application of Setpoint Methodology for LSSS Functions," July 2009.
- 7. Unit 3 PC/M No.90-220, RTD Bypass Elimination Modification and Eagle 21 Installation, (EC 244881).
- 8. Unit 4 PC/M No.90-221, RTD Bypass Elimination Modification and Eagle 21 Installation, (EC 244882).
- 9. WCAP-12632, RTD Bypass Elimination Licensing Report for Turkey Point Units 3 and 4, Revision 1, November 1990.
- 10. WCAP-12374, Topical Report Eagle-21 Microprocessor-based Process Protection System, Revision 1, December 1991.
7.2-40 Revised 08/17/2016 C28 TABLE 7.2-1 Sheet 1 of 8 LIST OF REACTOR TRIPS & CAUSES OF ACTUATION OF: ENGINEERED SAFETY FEATURES, CONTAINMENT ISOLATION AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER ACTUATION
REACTOR TRIP TRIP SETPOINT COINCIDENCE CIRCUITRY AND INTERLOCKS COMMENTS
- 1. Manual NA 1/2, no interlocks
- 2. Power Range 108% RTP* 2/4, no interlocks High and low settings;
High Neutron Flux 25% RTP* 2/4, manual block manual block and automatic permitted by permissive P-10 reset of low setting by P-10, Table 7.2-2.
- 3. Overtemperature T Note 1 2/3, no interlocks Note 2
- 4. Overpower T Note 3 2/3, no interlocks Note 4
- 5. Low Pressurizer Pressure >
1835 psig 2/3, interlocked with P-7 (fixed set point)
- 6. High Pressurizer Pressure <
2385 psig 2/3, no interlocks (fixed set point)
- 7. High Pressurizer Water Level <
92% of instrument 2/3, interlocked with P-7 span
- 8. Low Reactor Coolant Flow 90% of loop thermal 2/3, per loop, interlocked with Low flow in 2 loops
design flow** P-7, and P-8 permitted below P-7. Low flow in 1 loop permitted below P-8.
- 9. Monitored Electrical Supply to
Reactor Coolant Pumps:
9a. Undervoltage - 4.16 KV >70% bus voltage 1/2, on both buses, interlocked with P-7 Buses A and B
9b. Underfrequency - Trip of Reactor >56.1 HZ Under frequency on 1 out of 2 on Under frequency on any Coolant Pump Breaker(s) Open either bus bus will trip minimum of one reactor coolant pump and consequently cause a reactor trip; reactor trip interlocked with P-7 and P-8
- RTP = Rated Thermal Power
- Loop thermal design flow = 86,900 gpm
Revised 04/17/2013 C26 C26 C26 TABLE 7.2-1 (Continued)
Sheet 2 of 8
REACTOR TRIP TRIP SETPOINT COINCIDENCE CIRCUITRY AND INTERLOCKS COMMENTS
9c. Reactor Coolant Pump Breakers NA interlocked with P-7 and P-8
- 10. Safety Injection Signal NA (Actuation) See Item 19
- 11. Turbine-Generator Trip 2/3, low Emergency Trip Header Pressure interlocked with P-7, or 2/2 stop Emergency Trip Header Pressure 1000 psig valve closure indication (interlocked with P-7)
Turbine Stop Valve Fully closed***
- 12. Steam/Feedwater Flow Mismatch, Feed flow <20% 1/2, steam/feedwater flow mismatch coincident with : below steam flow in coincidence with 1/2 low steam generator water level per loop Low Steam Generator Water Level 16% of narrow range
instrument span
- 13. Low-Low Steam Generator Water 16% of narrow range 2/3, per loop Level instrument span
- 14. Intermediate Range Neutron Flux <25% of RTP* 1/2, manual block permitted by P-10 Manual block and automatic reset
- 15. Source Range Neutron Flux <
10 5 CPS 1/2, manual block permitted by P-6, Manual block and automatic interlocked with P-10 reset by P-6, automatic block by P-10
- 16. Phase A - Safety Injection Signal NA See Item 19 (except manual isolation); Actuates all non-essential service containment isolation trip valves.
Manual Initiation 2 momentary push buttons, pressing of either push button (1/2)
will actuate.
- RTP = Rated Thermal Power
- Limit switch is set when turbine stop valves are fully closed.
Revised 04/17/2013 C26 C26 C26 TABLE 7.2-1 (Continued) Sheet 3 of 8 CONTAINMENT ISOLATION ACTUATION TRIP SETPOINT COINCIDENCE CIRCUITRY AND INTERLOCKS COMMENTS
- 17. Phase B - Containment Pressure 2/3 high containment in coincidence Actuates all essential High High Coincident with 20 psig with 2/3 high-high pressure service containment Containment Pressure High 4 psig isolation trip valves
Manual Initiation NA 2/2, No Interlocks
CONTAINMENT VENTILATION ISOLATION
18a. High Containment Activity Note 5 High activity signal, from air This additional signal particulate detector or radiogas closes containment purge detector. (1/2) supply and exhaust valves.
18b. Phase A Containment Isolation Manual 18c. Phase B Containment Isolation Manual 18d. Safety Injection See Item 19
ENGINEERED SAFETY FEATURES ACTUATION
- 19. Safety Injection Signal (A) See Item 10
- a. Manual Initiation NA 1/2, no interlocks
- b. Containment Pressure - High <4 psig 2/3, no interlocks c. Pressurizer Pressure - Low >1730 psig 2000 psig (Pzr Press) d. High Differential Pressure <100 psid 2/3, manual block permitted below 2000 psig Between the Steam Line (pressurizer pressure)
Header and any Steam Line
- e. Steam Line Flow - High A function defined as 1/2 in 2/3 steam generators, manual
follows: A p corres- block permitted below 543
ûF(Tavg Temp) ponding to 40% steam
flow at 0% load increasing
linearly from 20% load to a value corresponding to 114%
steam flow at full load.
coincident with:
Steam Generator 614 psig 2/3, manual block permitted below 543 o F (Tavg Temp)
Pressure - Low, or
T avg - Low >543 oF 2/3, manual block permitted below 543 o F (Tavg Temp) 20. Containment Spray Signal (P)
Containment Pressure - <20.0 psig 2/3 high containment pressure in High - High coincident with: coincidence with 2/3 High-High Containment Containment Pressure - High <4.0 psig pressure
- 21. Emergency Containment Cooling NA Safety injection signal initiates the start of two of three ECCs in accordance with the Safety Injection Starting Sequence. The third swing ECC will automatically start upon failure of
either of the other two ECCs to start.
Revised 04/17/2015 C27 TABLE 7.2-1 (Continued)
Sheet 4 of 8
STEAM LINES ISOLATION ACTUATION TRIP SETPOINT COINCIDENCE CIRCUITRY AND INTERLOCKS COMMENTS
- 22. Steam Flow 1/2 High steam line flow in 2 out of 3 loops coincident with either low T avg Steam Line Flow - High A function defined as in 2 out of 3 loops or low steam line follows: A p corres- pressure in 2 out of 3 loops ponding to 40% steam Manual block is permitted below 534
ûF flow at 0% load increasing (T avg Temp) linearly from 20% load to a value corresponding to 114%
steam flow at full load.
coincident with:
Steam Line 614 psig
Pressure - Low or
T avg - Low >
543 o F
- 23. Containment Pressure 2/3 high containment pressure signal in coincidence with 2/3 high-high High <4.0 psig containment pressure High - High <
20.0 psig
- 24. Manual per Steam Loop NA 1/1 per steam line
AUXILIARY FEEDWATER ACTUATION
25a. Low-Low Steam Generator Level 16% NRS 2/3 per loop, no interlocks
- b. Safety Injection Signal N/A See Item 19
- c. Feedwater Pump Trip N/A Trip of all operating feed pumps
- d. Bus stripping N/A
- e. AMSAC 8.65% 2/3 Low Steam Generator Level, for Both channels, with turbine power greater than 40% (minus an allowance for instrument
uncertainty)
MAIN FEEDWATER ISOLATION
26a. Close Main Feedwater Control Actuated by: This function is related to Valves (fast closure) 1. Safety injection (see #19) the Steam Generator Overfill
- 2. 2/3 high-high level Protection function; (80%) in steam generator
- 3. Reactor trip coincident
with low T avg (slow closure)
26b. Close Bypass Feedwater 1. Safety injection (see item 19)
Control Valves 2. 2/3 high-high level(80%) in any steam generator
26c. Close Backup Feedwater Isolation Valves Safety injection signal (See Item 19)
- 27. a) Trip Steam Generator Feed Pumps Safety injection signal (See Item 19)
b) Turbine Trip 2/3 high-high level (80%)
in any steam generator
Revised 04/17/2013 C26 C26 C26 C26 TABLE 7.2-1 (Continued)
Sheet 5 of 8 TABLE NOTATIONS
NOTE 1: OVERTEMPERATURE T Those values denoted with [*] are specified in the COLR (Chapter 14, Appendix A)
Where: T = Measured T by RTD Instrumentation 1 + 1 S = Lead/Lag compensator on measured T; 1 = [*]s, 2 = [*]s 1 + 2 S 1
= Lag compensator on measured T; 3 = [*]s 1 + 3 S T o = Indicated T at RATED THERMAL POWER
K 1 = [*] K 2 = [*]/o F; 1 + 4 S = The function generated by the lead-lag compensator for T avg 1 + 5 S dynamic compensation; 4 , 5 = Time constants utilized in the lead-lag compensator for T avg , 4 = [*]s, 5 = [*]s; T = Average temperature, o F; 1
= Lag compensator on measured T avg; 6 = [*]s 1 + 6 S T'
< [*] o F (Indicated Loop T avg at RATED THERMAL POWER);
K 3 = [*]/psi;
Revised 04/17/2013 I f - P - P K ]T - S 1 1 [T S 1S 1 K - K T S 1 1 S)(1S)(1T 1 3 6 T 5 T 4 T21o 3 T 2 1C26 C26 C26 C26 C26 C26 TABLE 7.2-1 (Continued)
Sheet 6 of 8 TABLE NOTATIONS (Continued)
NOTE 1: (Continued)
P = Pressurizer pressure, psig;
P' > [*] psig (Nominal RCS operating pressure);
S = Laplace transform operator, s-1;
and f 1 (I) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(1) For q t - q b between -[*]% and [*]%, f 1 (I) = 0, where q t and q b are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q t + q b is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q t - q b exceeds -[*]%, the T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q t - q b exceeds [*]%, the T Trip Setpoint shall be automatically reduced by [*]% of its value at RATED THERMAL POWER.
NOTE 2: The Overtemperature T function Allowable Value shall not exceed the nominrl trip setpoint by more than 0.5% T span for the T channel; 0.2% T span for the Pressurizer Pressure Channel; and 0.4% T span for the F(I) channel. No separate Allowable Value is provided for T avg because this function is part of the T value.
Revised 04/17/2013
C26 C26 TABLE 7.2-1 (Continued)
Sheet 7 of 8 TABLE NOTATIONS (Continued)
NOTE 1: OVERTEMPERATURE T Those values denoted with [*] are specified in the COLR (Chapter 14, Appendix A)
Where: T = As defined in Note 1; 1 + 1 S = As defined in Note 1; 1 + 2 S 1
= As defined in Note 1; 1 + 3 S T o = As defined in Note 1;
K 4 = [*];
K 5 > [*]/o F for increasing average temperature and
[*]/o F for decreasing average temperature;
7 S = The function generated by the lead-lag compensator for T avg dynamic 1 + 7 S compensation; 7 = Time constants utilized in the lead-lag compensator for T avg , 7 > [*]s, 1
= As defined in Note 1; 1 + 6 S
Revised 04/17/2013 I f - ]T - S 1 1 - [T K - T S 1 1 S 1 S K - K T S 1 1 S)(1S)(1T 2 T 6 6 T 7 T 7 T54o 3 T 2 1 6C26 C26 C26 TABLE 7.2-1 (Continued)
Sheet 8 of 8 TABLE NOTATIONS (Continued)
NOTE 3: (Continued)
K 6 = [*]/o F for T > T" and K 6 = [*] for T < T" ;
T = As defined in Note 1;
T" <
[*] o F (Indicated Loop T avg at RATED THERMAL POWER);
S = As defined in Note 1, and
f 2 (I) = [*]
NOTE 4: The Overpower T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the channel.
No separate Allowable Value is provided for T avg because this function is part of the T value.
NOTE 5: Particulate (R-11)
<6.1 x 10 5 CPM Gaseous (R-12)
Containment Gaseous Monitor Setpoint =
(3.2 x 10 4) CPM, ( F )
Containment Gaseous Monitor Allowable Value =
(3.5 x 10 4) CPM, ( F )
where F =
Actual Purge Flow Design Purge Flow (35,000 CFM)
Setpoint may vary according to current plant conditions provided that the release rate does not exceed allowable limits provided
in the Offsite Dose Calculation Manual.
Revised 04/17/2013 C26 C26 C26 C26 TABLE 7.2-2 PERMISSIVE CIRCUITS
Number Function Required input
1 Prevent rod withdrawal 1/4 high nuclear flux (power range) or on overpower 1/2 high nuclear flux (intermediate range) or 2/3 overtemperature T or 2/3 overpower T. 2*
3*
4*
5 Steam dump to condenser Rapid decrease of MWe load signal permissive, fast load drop (turbine inlet pressure) arms system
6 Manual block of source 1/2 high intermediate range flux range trip allows manual block, 2/2 low intermediate range defeats block.
7 Permissive power (block 2/4 high nuclear flux (power range) various trips). Required or 1/2 high turbine power (inlet only at power.
pressure) enables trips.
3/4 low nuclear flux (power range) and 2/2 low turbine power (inlet pressure) blocks trips.
8 Block single primary loop 2/4 high nuclear flux (power range) loss of flow trip blocks trip.
9*
10 Manual block of low power 2/4 high nuclear flux (power range) range trip and high allows manual block, intermediate range trip 3/4 low nuclear flux (power
range) defeats manual block.
Manual block of safety 2/3 low pressurizer pressure, injection 2/3 low T avg temperature
- Not applicable to this plant.
Revised 04/17/2013 C26C26C26 TABLE 7.2-3 ROD STOPS
Rod Motion to
Rod Stop Actuation Signal be blocked
- 1. Nuclear 1/4 high power range nuclear Manual Overpower flux or 1/2 high intermediate Withdrawal range nuclear flux
- 2. High T 2/3 overpower T or 2/3 over- Manual temperature T Withdrawal
Notes: 1. The overpower T and overtemperature T rod stop setpoints are the same as the reactor trip setpoints, effectively negating these functions.
Revised 10/23/2006 C22
Revised 04/17/2013)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 TYPICAL ILLUSTRATION OF HIGH T TRIP (T vs. Tavg) FIGURE 7.2-1 C26 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT REACTOR PROTECTION SYSTEMS FIGURE 7.2-2
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 REACTOR PROTECTION SYSTEM REDUNDANT CHANNEL SEPARATION DESIGN CONFIGURATION FIGURE 7.2-3A
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 EFS ACTUATION SYSTEM REDUNDANT CHANNEL SEPARATION DESIGN CONFIGURATION FIGURE 7.2-3B
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 REACTOR PROTECTION SYSTEM TYPICAL PROCESS CHANNEL TESTING CONFIGURATION FIGURE 7.2-4
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-5 REFER TO ENGINEERING DRAWING 5613-T-L1 , SHEET 2 5614-T-L1 , SHEET 2
Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 REACTOR TRIP SIGNALS FIGURE 7.2-5 C26C26
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 REACTOR PROTECTION SYSTEM TYPICALLOGIC RELAY TESTING CONFIGURATION FIGURE 7.2-6A
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 EFS ACTUATION SYSTEM TYPICAL LOGIC RELAY TESTING CONFIGURATION FIGURE 7.2-6B
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 RPS LOGIC CHANNEL TEST PANELS FIGURE 7.2-7
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8a REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 18
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PRESSURIZER CAUSED REACTOR TRIP & SAFETY INJECTION LOGIC DIAGRAM FIGURE 7.2-8a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8b REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 19
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR CAUSED REACTOR TRIP & SAFETY INJECTION LOGIC DIAGRAM FIGURE 7.2-8b
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8c REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 20
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PRIMARY COOLANT SYSTEM REACTOR TRIP & TAVG INTERLOCK LOGIC DIAGRAM FIGURE 7.2-8c
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8d REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 16
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 NUCLEAR INSTRUMENTATION TRIP SIGNALS LOGIC DIAGRAM FIGURE 7.2-8d
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8e REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 11
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SAFEGUARDS ACTUATION AND STEAM LINE ISOLATION LOGIC DIAGRAM FIGURE 7.2-8e
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-8f REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 17
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 NUCLEAR INSTRUMENTATION PERMISSIVES AND BLOCKS LOGIC DIAGRAM FIGURE 7.2-8f
Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SETPOINT RELATIONSHIPS FIGURE 7.2-8g C26
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-9a REFER TO ENGINEERING DRAWING 5610-T-D-12A , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 ROD CONTROL SYSTEM CONTROL SYSTEM DIAGRAM FIGURE 7.2-9a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-9b REFER TO ENGINEERING DRAWING 5610-T-D-12B , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 TAVG CONTROL AND INSERTION LIMIT ALARMS CONTROL SYSTEM DIAGRAM FIGURE 7.2-9b
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-10 REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 INDEX AND SYMBOLS FOR LOGIC DIAGRAMS FIGURE 7.2-10
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-11a REFER TO ENGINEERING DRAWING 5610-T-D-16A , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PZR PRESSURE PROTECTION &
OVERPRESSURE MITIGATION SYSTEM CONTROL SYSTEM DIAGRAM FIGURE 7.2-11a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-11b REFER TO ENGINEERING DRAWING 5610-T-D-16B , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PRESSURIZER PRESSURE CONTROL CONTROL SYSTEM DIAGRAM FIGURE 7.2-11b
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-12 REFER TO ENGINEERING DRAWING 5610-T-D-15 , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PZR LEVEL CONTROL & PROTECTION AND CHARGING PUMP CONTROL CONTROL SYSTEM DIAGRAM FIGURE 7.2-12
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-13 REFER TO ENGINEERING DRAWING 5613-T-D-17 , SHEET 1 5614-T-D-17 , SHEET 1
Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 STEAM GENERATOR LEVEL CONTROL AND PROTECTION FIGURE 7.2-13 C26C26
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-14a REFER TO ENGINEERING DRAWING 5613-T-L1, SHEET 33A
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 ATWS MITIGATING SYSTEM ACTUATION CIRCUITRY (AMSAC)
LOGIC DIAGRAM FIGURE 7.2-14a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.2-14b REFER TO ENGINEERING DRAWING 5613-T-L1, SHEET 33B
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 ATWS MITIGATING SYSTEM ANNUNCIATION CIRCUITRY (AMSAC)
LOGIC DIAGRAM FIGURE 7.2-14b
7.3 REGULATING SYSTEM 7.3.1 DESIGN BASIS
The reactor automatic control system is designed to respond to a rapid change in indicated nuclear flux versus steam demand (0/N-QT) through automatic rod
insertion only. This system does not have the capability for automatic rod
withdrawal. Overall reactivity control is achieved by the combination of
chemical shim and Rod Cluster Control Assemblies (RCCA). Long-term
regulation of core reactivity is accomplished by adjusting the concentration
of boric acid in the reactor coolant. Short-term reactivity control for
power changes or reactor trip is accomplished by moving RCCAs.
The function of the Reactor Control System is to provide automatic control of
the RCCAs (rod insertion only) during power operation of the reactor. The
system uses input signals including coolant temperature and turbine load.
The Chemical and Volume Control System (Section 9) supplements the reactor
control system by the addition and removal of varying amounts of boric acid
solution.
There is no provision for a direct continuous visual display of primary
coolant boron concentration. When the reactor is critical, the best
indication of reactivity status in the core is the position of the control
group in relation to power and average coolant temperature. There is a
direct relationship between control rod position and power and it is this
relationship which establishes the lower insertion limit calculated by the
rod insertion limit monitor. There are two alarm setpoints to alert the
operator to take corrective action in the event a control group approaches or
reaches its lower limit.
7.3-1 Rev. 16 10/99 Any unexpected change in the position (insertion) of the control group 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, periodic samples are taken for
determination of the coolant boron concentration. The variation in
concentration during core life provides a further check on the reactivity
status of the reactor including core depletion.
The Reactor Control System is designed to enable the reactor to follow load
reductions automatically when the output is above 15% of nominal power.
Control rod positioning (insertion) may be performed automatically when
output is above this value. Control rod positioning may be performed
manually at any time.
The operator is able to select any single bank of rods for manual operation.
This is accomplished with a multiposition switch so that he may not select
more than one bank. He may also select automatic or manual reactor control, in either case, however, the control banks can be moved only in their normal
sequence with some overlap as one bank reaches its full withdrawal position
and the next bank begins to withdraw. Relay interlocks, designed to meet the
single failure criterion, are provided to preclude simultaneous withdrawal of
more than one bank of control and shutdown rods except in overlap regions.
The system enables the nuclear units to accept a step load reduction of 10%
and a ramp reduction of 5% per minute within the load range of 100% to 15%
without reactor trip subject to possible xenon limitations. With automatic
rod withdrawal disabled, ramp load increase to 5% per minute is performed
manually. Manual rod withdrawal will be needed to bring the reactor coolant
average temperature to the programmed value following a 10% load increase
The control system is capable of restoring coolant average temperature
following a scheduled or transient reduction in load.
7.3-2 Revised 04/17/2013 C26 The pressurizer water level is programmed to be a function of the average coolant temperature. This is to minimize the requirements on the Chemical
and Volume Control and Waste Disposal System resulting from coolant density
changes during loading and unloading from full power to zero power.
Following a reactor and turbine trip, sensible heat stored in the reactor
coolant is removed without actuating the steam generator safety valves by
means of controlled steam bypass to the condenser and by injection of
feedwater to the steam generators. Reactor coolant system temperature is
reduced to the no load condition. This no load coolant temperature is
maintained by steam bypass to the condensers which removes residual heat.
The control system is designed to operate the system over the full range of
automatic control throughout core life.
7.3-3 Rev. 12 5/95 7.3.2 SYSTEM DESIGN
The Power Regulating System can be broken down into two subsystems as
follows:
- 1. Rod Control System
- a. Rod Drive Programmer
- b. Rod Position Indication
(1) Individual (2) Group
A control diagram of the Rod Control System is shown in Figure 7.2-9a.
- 2. Steam Dump Control Control logic for steam dump to condenser is shown in Figure 7.3-1 and Figure 7.3-1a.
RCCA Arrangements
There are 45 total RCCAs. The rods are divided into (1) a shutdown group
comprising two shutdown banks of 8 rod clusters each, (2) a control group
comprising 4 control banks containing 8, 8, 8, and 5 rod clusters. Figure
3.2.1-1 shows the location of RCCAs within the core. The four banks of the
control group are the only rods that can be manipulated under
7.3-4 Rev. 16 10/99 automatic control. The banks are divided into subgroups to obtain smaller incremental reactivity changes. All RCCAs in a subgroup are electrically
paralleled to move simultaneously. There is individual position indication
for each RCCA. The drive mechanism for the RCCAs is described in Section
3.2.3.
Control Group Rod Control
The reactor control system is capable of restoring programmed average
temperature following a reduction in load. The coolant average temperature
increases linearly from zero to full power.
Reactivity changes caused by fuel depletion and/or xenon transients are
initially compensated through manual rod control. Final compensation for
these two effects is made by adjusting the boron concentration. The control
system may then readjust (insert) the control group rods to respond to
changes in coolant average temperature resulting from changes in boron
concentration.
The coolant temperatures are measured by the hot leg and the cold leg
resistance temperature detectors. There is one average temperature per loop.
The median of three loop average temperatures is the main control signal.
This signal is sent to the control group rod programmer through a lead/lag
compensation unit. The control group rod programmer determines the direction
and speed of control group rod motion.
The RCCAs are divided into six main banks, and each bank into two
subgroups, to follow load changes over the full range of power operation.
Each subgroup in a bank is driven by the same variable speed rod drive
control unit which moves the subgroups sequentially one step at a time. The
sequence of motion is reversible; that is, a withdrawal sequence is the
reverse of the insertion sequence. The variable speed sequential rod control
affords the ability to insert a small amount of reactivity at low speed to
accomplish fine control of reactor coolant average temperature about a small
temperature deadband.
7.3-5 Rev. 16 10/99 Manual control is provided to move a control bank in or out at a preselected
fixed speed.
Proper sequencing of the RCCA is assured first, by fixed programming
equipment in the Rod Control System, and second, through administrative
control of the
reactor operator. Startup is accomplished by first manually withdrawing the
shutdown rods to the full out position. This action requires that the
operator select the SHUTDOWN BANK position on a control board mounted
selector switch and then position the IN-HOLD-OUT lever (which is spring
return to the HOLD position) to the out position.
RCCA are then withdrawn under manual control of the operator by positioning
the IN-HOLD-OUT lever to the OUT position. In the MANUAL selector switch
position, the rods are withdrawn (or inserted) in a predetermined programmed
sequence by the automatic programming equipment.
Programming is set so that as the first bank out reaches a preset position
near the top of the core, the second bank out begins to move out
simultaneously with the first bank. This staggered withdrawal sequence
continues until the unit reaches the desired power level. The programmed
insertion sequence is the opposite of the withdrawal sequence, i.e., the last
control bank out is the first control bank in.
With the simplicity of the rod program, the minimal amount of operator
selection, and two separate direct position indications available to the
operator, there is very little possibility that rearrangement of the control
rod sequencing could be made.
Shutdown Groups Control
The shutdown groups of control rods together with the control groups are
capable of shutting the reactor down. They are used in conjunction with the
adjustment of chemical shim and the control groups to provide shutdown
margin of at least one per cent following reactor trip with the most
reactive control rod in the fully withdrawn position for all normal operating
conditions.
7.3-6 Rev. 4 7/86 The shutdown groups are manually controlled during normal operation and are moved at a constant speed. Any reactor trip signal causes them to fall into
the core. They are fully withdrawn during power operation and are withdrawn
first during startup. Criticality is always approached with the control
groups after withdrawal of the shutdown groups.
Interlocks
The manual controls are interlocked with measurements of T and rod position system rod bottom bistables to prevent approach to an overpower condition.
7.3-7 Revised 04/27/2010 C24 Rod Drive Performance The control is driven by a sequencing, variable speed rod drive programmer.
In the control group of RCC assemblies, control subgroups (each containing a
small number of RCC assemblies) are moved sequentially in a cycle such that
all subgroups within a group are maintained within one step of each other.
The sequence of motion is reversible, that is, withdrawal sequence is the
reverse of the insertion sequence. The sequencing speed for rod insertion is
proportional to the control signal from the Reactor Control System. This
provides control group speed control proportional to the demand signal from
the control system.
A rod drive mechanism control center is provided to receive sequenced signals
from the programmer and to actuate SCRs in series with the coils of the rod
drive mechanisms. Two reactor trip breakers are placed in series with the
supply for these coils. To permit on-line testing, a bypass breaker is
provided across each of the two trip breakers.
Full Length RCCA Position Indication
Two separate systems are provided to sense and display control rod position
as described below:
a) Analog System - An analog signal is produced for each RCCA by a
linear position transmitter.
7.3-8 Rev 16 10/99 An electrical coil stack is placed above the stepping mechanisms of the control rod magnetic jacks external to the pressure housing. When the
associated control rod is at the bottom of the core, the magnetic
coupling between a primary and secondary is small and there is a small
voltage induced in the secondary. As the control rod is raised by the
magnetic jacks, the relatively high permeability of the lift rod causes
an increase in magnetic coupling. Thus, an analog signal proportional
to rod position is derived.
Direct, continuous readout of every RCC assembly position is presented
to the operator by individual meter indications, without need for
operator selection or switching to determine rod position.
Lights are provided for rod bottom positions for each rod. The lights
are operated by bistable devices in the analog system.
b) Digital System - The digital system counts pulses generated in the rod drive control system. One counter is associated with each group (or
subgroups) of RCCAs. Readout of the digital system is in the form of
add-subtract counters reading the number of steps of rod withdrawal
with one display for each group or subgroup. These readouts are
mounted on the control panel.
The digital and analog systems are separate systems; each serves as backup
for the other. Operating procedures require the reactor operator to compare
the digital and analog readings upon recognition of any apparent malfunction.
Therefore, a single failure in rod position indication does not in itself
lead the operator to take erroneous action in the operation of the reactor.
A detailed description of the solid state rod control power supply will be
available in a WCAP report.
7.3-9 Rev 16 10/99 Individual RCCA Position Indication This system derives the position signal directly from measurements of the
drive rod position utilizing a linear variable differential transformer (LVDT) as a detector. The drive shaft varies the amount of coupling between
the primary and secondary windings of the coils which generates an analog
signal proportional to rod position. The LVDT signal is conditioned and
displayed on individual meters mounted on the operating console.
Demand Position Indication
The bank demand position signal is derived from the programmer and is
displayed on an add-subtract pulse counter mounted in the control console.
Rod Deviation
Control rods - The actual rod position signals are monitored by rod deviation
monitoring equipment which provides an alarm whenever an individual rod
position signal deviates from any other rod in a bank by a preset limit.
Shutdown rods - An alarm is generated whenever any shutdown rod is inserted a
preset amount from the fully withdrawn position.
7.3-10 Turbine By-Pass A turbine by-pass system is provided to accommodate a reactor trip with
turbine trip, or 50% loss of load without reactor and turbine trip. The
turbine by-pass system removes steam to reduce the transient imposed upon the
reactor coolant system. The control rod system can then reduce the reactor
power to a new equilibrium value without causing overtemperature and/or
overpressure conditions.
The turbine by-pass is actuated when the median average coolant temperature
exceeds the programmed value by a given value and the turbine inlet pressure
decrease is greater than a given value. All the turbine by-pass valves
stroke to full open immediately upon receiving the maximum by-pass signal.
The by-pass valves are modulated after they are full open by the median
coolant average temperature signal. The turbine bypass flow reduces
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 turbine by-pass steam capacity varies from approximately 27.2 to 34.4 percent of full load steam flow based on the full power average temperature and steam pressure operating window.
Feedwater Control
Each steam generator is equipped with a three-element feedwater controller (see Figure 7.2-13) which maintains a programmed water level as a function of
load on the secondary side of the steam generator. The three-element
feedwater controller regulates the feedwater valve by continuously comparing
the feedwater flow signal, the water level signal and the steam flow signal
which is compensated by a steam pressure signal. The feedwater controller
gain and reset tuning parameters are adjusted as a function of steam flow, feed flow, or T avg to provide optimal controller performance over the entire operating range. The steam generators are operated in parallel, both on the
feedwater and on the steam side.
7.3-11 Revised 04/17/2013 C26C26C26 Continued delivery of feedwater to the steam generators is required as a sink for the heat stored and generated in the reactor coolant following a reactor
trip and turbine trip. An override signal closes the feedwater valves when
the average coolant temperature is below a given temperature or when the
respective steam generator level rises to a given value. Manual override of
the feedwater control systems is also provided.
Pressure Control
The reactor coolant system pressure is maintained at constant value by using
either the heaters (in the water region) or the spray (in the steam region of
the pressurizer). The electrical immersion heaters are located near the
bottom of the pressurizer. A portion of the heater groups are proportional
heaters which are used to control small pressure variations. These
variations are due to heat losses, including heat losses due to a small
continuous spray. The remaining (backup) heaters are turned on when the
pressurizer pressure controller signal is below a given value.
The spray nozzle is located at the top of the pressurizer. Spray is
initiated when the pressure controller signal is above a given set point.
The spray rate increases proportionally with increasing pressure until it
reaches a maximum value. 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.
Two power operated relief valves limit system pressure to 2350 psia for large
load reduction transients.
Three spring-loaded safety valves limit system pressure to 2750 psia
following a complete loss of load without direct reactor trip or turbine
by-pass.
7.3-12 Revised 04/17/2013 C26 7.3.3 SYSTEM DESIGN EVALUATION
Unit Stability
The Rod Control System is designed to maintain coolant average temperature about the control system set point within acceptable values. Because stability is more difficult to maintain, at low power under automatic control, no provision is made to provide automatic control below 15 percent of full power.
Step Load Changes Without Steam Dump
A typical power control requirement is to restore equilibrium conditions, without a trip, following a minus 10 percent step change in load demand, over the 15 to 100 percent power range for automatic control. The design must necessarily be based on conservative conditions and a greater transient capability is expected for actual operating conditions. A load demand greater than full power is prohibited by the Turbine Control System (TCS) load limiting software.
The function of the control system is to minimize the reactor average coolant temperature deviation during the transient within a given value. Excessive pressurizer pressure variations are prevented by using spray and heaters in the pressurizer.
7.3-13 Revised 08/17/2016 C28 The margin between over-temperature T set-point and the measured T is of primary concern for the step load changes. This margin is influenced by nuclear flux, pressurizer pressure, average reactor coolant temperature, and
temperature rise across the core.
Loading and Unloading
Ramp loading and unloading is performed under manual control. The function
of the control system is to respond to a rapid change in indicated nuclear
flux versus steam demand (0
/N-QT). The minimum control rod speed provides a sufficient reactivity rate to compensate the reactivity changes resulting
from the moderator and fuel temperature changes.
The average coolant 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
heaters limit the resulting system pressure decrease. The pressurizer level
is programmed such that the water level is above the setpoint at which the
heaters cut out during the loading and unloading transients.
The primary concern for the loading is to limit the overshoot in average
coolant temperature so that a margin is provided for the over-temperature T
set point.
The automatic load controls are designed to safely adjust the unit generation
to match load requirements within the limits of the unit capability and
licensed rating.
7.3-14 Rev 4 7/86 Loss of Load With Turbine By-Pass The reactor control system is designed to accept 50% loss of electrical load.
No reactor trip or turbine trip should be actuated. The automatic turbine
by-pass steam capacity varies from approximately 27.2 to 34.4 percent of full
load steam flow based on the full power average temperature and steam
pressure operating window. The turbine by-pass system is actuated during a
load rejection transient to reduce the effects that the transient imposes
upon the reactor coolant system. The reactor power is reduced at a rate
consistent with the capability of the rod control system. Reduction of the
reactor power is automatic down to 15 percent of full power, at which point
the operator places the rod motion control selector switch to MANUAL. The
by-pass flow reduction is as fast as RCCAs are capable of inserting negative
reactivity.
The pressurizer relief valves might be actuated for the most adverse
conditions, e.g., the most negative Doppler coefficient, and the minimum
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 conditions.
Turbine - Generator Trip With Reactor Trip
Whenever the turbine-generator unit trips at an operating level above 10%
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 saturation temperature
corresponding to the steam generator pressure at the safety valve set point.
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 steam temperature, a heat sink is required to
remove heat stored in the reactor coolant to prevent actuation of steam
generator safety valves for this trip from full power. This heat sink is
provided by the combination of controlled release of steam to the condenser
and by makeup of cold feedwater to the steam generators.
7.3-15 Revised 04/17/2013 C26 The turbine by-pass system is controlled from the reactor average coolant temperature signal whose set point values are reset upon trip to the no load
value. Actuation of the turbine by-pass must be rapid to prevent actuation
of the steam generator safety valves. With the by-pass valves open the
average
coolant temperature starts to reduce quickly to the no load set point. A
direct feedback of temperature acts to proportionally close the valves to
minimize the total amount of steam which is by-passed.
Following the turbine trip, the steam voids in the steam generators will
collapse and the fully opened feedwater valves will provide sufficient
feedwater flow to restore water level in the downcomer. The feedwater flow
is cut off when the average coolant temperature decreases below a given
temperature value or when the steam generator water level reaches a given
high level.
Additional feedwater makeup is then controlled manually to restore and
maintain steam generator level while assuring that the reactor coolant
temperature is at the desired value. Residual heat removal is maintained by
the steam generator pressure controller (manually selected) which controls
the amount of steam flow to the condensers. This controller operates the
same bypass 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. If 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 turbine by-pass and feedwater control systems are designed to prevent the
average coolant temperature falling below the programmed no load temperature
following the trip to ensure adequate reactivity shutdown margin.
7.3-16 Rev 16 10/99
FINAL SAFETY ANALYSIS REPORT FIGURE 7.3-1 REFER TO ENGINEERING DRAWING 5610-T-L1, SHEET 22A
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 STEAM DUMP TO CONDENSER LOGIC DIAGRAM FIGURE 7.3-1
FINAL SAFETY ANALYSIS REPORT FIGURE 7.3-1a REFER TO ENGINEERING DRAWING 5610-T-L1, SHEET 22B
REV. 16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 STEAM DUMP TO CONDENSER LOGIC DIAGRAM FIGURE 7.3-1a
7.4 NUCLEAR INSTRUMENTATION 7.4.1 DESIGN BASES
Fission Process Monitors and Controls
Criterion: Means shall be provided for monitoring or otherwise measuring and maintaining control over the fission process throughout core life
under all conditions that can reasonably be anticipated to cause
variations in reactivity of the core. (1967 Proposed GDC 13)
Primary Nuclear Instrumentation
The Primary Nuclear Instrumentation is utilized primarily for reactor
protection by permitting monitoring of neutron flux and by generating
appropriate trip and alarm functions for various phases of reactor operating
and shutdown conditions (including accidental criticality monitoring). It
also provides a secondary control function and indicates reactor status
during startup and power operation. The Primary Nuclear Instrumentation
System utilizes information from three separate types of instrumentation
channels to provide three discrete protection levels. Each range of Primary
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 from
source to intermediate and low power ranges. As the reactor power increases, the overpower protection level is increased administratively 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 a completely
shutdown condition to 120 percent of full power. The power range channels
are capable of recording overpower excursions up to 200 percent of full
power.
7.4-1 Rev. 16 10/99 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 residual
neutron source in the reloaded fuel and the primary and/or secondary neutron
source(s) (if installed) in the core and the core multiplication associated
with the shutdown reactivity. This is generally greater than one count 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 slightly
more than two decades of the total instrumentation range. This is a linear
range that overlaps with the higher portion of the intermediate range. The
overlap for all ranges is shown in Figure 7.4-1 in terms of leakage neutron
flux for a typical PWR plant. Start-up-rate indication for the source and
intermediate range channels is provided at the control console and nuclear
instrumentation panel.
7.4-1a Rev. 16 10/99 The system described above provides control room indication and recording of reactor neutron flux during core-loading, shutdown, start-up and power operation as well as during subsequent refueling. Reactor trip and rod-stop control and alarm signals are provided by this system for safe operation.
Control and permissive signals are transmitted to the Reactor Control and Protection System for automatic control. Equipment failures and test status information are annunciated in the control room.
Backup Nuclear Instrumentation
The Backup Nuclear Instrumentation is utilized for providing additional independent neutron flux indication in the control room, on the alternate shutdown panel and on the DCS / SPDS. This instrumentation does not interface with the reactor trip protection circuitry, and does not perform any control functions. It meets the requirements of Regulatory Guide 1.97, Rev. 3, and 10 CFR 50.48(c), NFPA 805.
The Backup Nuclear Instrumentation also provides alarms in the control room for system trouble and high flux at shutdown, and interfaces with the containment evacuation alarm system.
This system utilizes fission chambers to monitor the leakage neutron flux from a completely shutdown condition to 200 percent of full power which exceeds the range requirements of Regulatory Guide 1.97, Rev. 3.
7.4.2 SYSTEM DESIGN
The nuclear instrumentation system (Figures 7.4-2a and 7.4-2b) consists of ten independent channels: two of these being source range, two the intermediate range, four the power range channels and two wide range channels.
In addition, there are three auxiliary channels, the visual-audio count rate channel, the comparator channel, and the startup rate channel. The various detectors associated with the ten channels are shown in relative position with respect to the core configuration on Figure 7.4-3.
7.4-2 Revised 09/20/2016 C28 Protection Philosophy Nuclear protection assurance is obtained from the three ranges of out-of-core
nuclear instrumentation. Separation of redundant protective channels is
maintained from the neutron sensor with its associated cables to the signal
conditioning equipment in the control room with its associated output wiring, indicating or recording devices and protective devices. Where redundant
protective channels are combined to provide non-protective functions, the
required signals are derived through isolation amplifiers. These devices are
designed so that open or short circuit conditions as well as the application
of 120 VAC or 140 VDC to the isolated side of the circuit will have no effect
on the input or protection side of the circuit. As such, failures on the
non-protective side of the system will not affect the individual protection
channels. Redundant channels are powered from independent power sources, each channel being provided with the necessary power supplies for its
detectors, signal conditioning equipment, trip bistables and associated trip
relays. The nuclear instrumentation channels are mounted in four separate
racks to provide the necessary physical separation between redundant
channels.
The overpower protection provided by the out-of-core nuclear instrumentation
consists of three discrete levels. Continuation of start-up 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-of-two intermediate range permissive signal (P6) is required prior to
source range level trip blocking and detector high voltage cutoff. Source
range level trips are automatically reactivated and high voltage restored
when both intermediate range channels are below the permissive (P6) level.
There are provisions for administratively reactivating the source range level
trip and detector high voltage if required. Source
7.4-3
range level trip block and high voltage cutoff are automatically maintained
by the power range permissive (P10).
The intermediate range level trip and low-range, power-range level trip can
only be blocked after satisfactory operation and permissive information are
obtained from two-of-four power range channels. Individual blocking switches
are provided so that the low-range, power-range trip and intermediate range
trip can be independently blocked. These trips are automatically reactivated
when any three of the four power range channels are below the permissive (P10) level, thus ensuring automatic activation of more restrictive trip
protection.
Blocking of any reactor trip function is indicated by the control board
status lights. Channels which provide reactor protection through one-of-two
or one-of-four logic matrices are equipped with positive detent type trip-bypass
switches to enable channel testing. The trip-bypass condition for individual
channels is indicated at the control board and at the nuclear instrumentation
racks. The reactor protection afforded by the high-range, power-range trip
is never blocked or bypassed.
Source Range Instrumentation
Two independent source range channels are provided. Each receives pulse-type
signals from a proportional counter. The preamplified detector signal is
received by the source range instrumentation conditioning equipment located
in the control room racks. The detector signal, which is a random count rate
proportional to leakage neutron flux, is conditioned for conversion to an
analog signal proportional to the logarithm of the neutron flux count rate.
The isolated analog signals from each channel are sent to various recording
and indicating devices to provide the operator with necessary startup
information. Bistable units also located in the racks, are used to
7.4-4 generate alarms and reactor trip signals. Trip signals from the bistables are transmitted to relays in the protection relay racks where the necessary
logic involved in generating reactor trip signals is performed.
An isolated count rate signal derived from either channel is connected to a
scaler-timer. This same signal also feeds the audio count rate channel which
provides an audible count rate signal, proportional to the neutron flux.
Speakers are provided both in the containment and in the control room.
Start-up rate indication is also provided for each source range channel.
These signals are generated from the isolation amplifier output since there
is no protection function involved.
Two additional wide range channels are also provided. Each receives signals
from dual fission chambers. The signal is received by the backup NIS
instrumentation conditioning equipment located in the rod drive rooms. The
signals are conditioned for conversion to an analog signal proportional to
the logarithm of the neutron flux count rate.
The isolated analog signals from each channel are sent to remote meters to
provide the operator with additional start-up information. Bistable units
are also provided to generate alarms only.
Intermediate Range Instrumentation
Two independent compensated ionization chambers provide extended flux
coverage from the upper end of the source range to approximately rated power.
The equipment for each channel, including the high voltage and compensating
voltage power supplies are located in separate drawers. To maintain
separation between these redundant channels, the drawers are mounted in
separate racks. The signal conditioning equipment furnishes an analog output
voltage proportional to the logarithm of the neutron flux spectrum. Each
channel covers approximately 8 decades of leakage flux. Isolation amplifiers (for start-up-rate circuits, remote recording, remote indication, etc) and
bistable devices (for permissives, rod-stop and reactor trip) use this analog
voltage to indicate status and provide the necessary protection functions.
All relays associated with control or protection are located in the reactor
protection or auxiliary relay racks.
7.4-5 Rev. 16 10/99 Power Range Instrumentation Four dual section, uncompensated ionization chambers are used for power range
flux detection. Each chamber provides two current signal outputs (one from
each section) to signal conditioning equipment in the control room nuclear
instrumentation racks. Each power range channel has an independent high
voltage power supply. The individual current signals obtained from each
section of the detector are proportional to upper core and lower core neutron
flux respectively. These provide core flux status information at the
instrument racks and through isolation amplifiers the same information at the
control console. A separate output furnishes bias signals used in the overpower and overtemperature T reactor trip functions. The individual current signals are combined to provide an average signal proportional to average core flux in the associated core quadrant. This average signal is
conditioned to provide an analog voltage signal for use in permissive and
protection bistable amplifiers.
Isolation amplifiers, which provide remote control signals and core power
status information to the operator, also utilize the average power analog
signal. The four power range channels are operated from separate AC sources
and are housed in separate racks so that a single failure will not cause loss
of protection functions. Redundant relays for the protection functions are
located in the logic portion of the protection system.
Isolated analog outputs from each power range channel are compared in a
separate auxiliary channel drawer. This comparator provides the operator
with annunciation of deviations in average power between the four power range
channels. Switches are provided to defeat this comparison for a failed
channel so that subsequent deviations or failures among the three remaining
channels are annunciated.
Two additional dual fission chamber detector assemblies are used for wide
range neutron flux detection. Each assembly provides signal outputs to the
backup NIS signal conditioning equipment in the rod drive rooms. Each
channel has an independent power supply.
Isolated wide range outputs provide remote indication of percent full power
in the control room, the Alternate Shutdown Panel (channel B only), and the
SPDS.
7.4-6 Rev 16 10/99 Equipment Design Basis The out-of-core nuclear instrumentation system consists of various plug-in
type modules which perform the functions indicated on Figure 7.4-2 for the
source, intermediate and power ranges. Components designed to military
specifications are used, where possible, in conjunction with a conservative
design stressing reliability, derating of components and circuits, and the
use of field-proven circuits. On-line testing and calibration features are
provided for each channel. The source and power range test signals can be
superimposed on the normal sensor signal during operation.
The backup NIS components are qualified to IEEE standards. The backup NIS
provides indication and alarm functions only and is completely independent of
the primary NIS subsystem.
7.4.3 DETAILED DESCRIPTION
Detectors The primary nuclear instrumentation system employs six detector radial
locations containing a total of eight detectors (two proportional counters,
two compensated ionization chambers and four, dual section uncompensated
ionization chamber assemblies) installed around the reactor in the primary
shield. Windows in the primary shield minimize leakage flux attenuation and
distortion.
BF 3 proportional counters having a nominal thermal neutron sensitivity of ten counts per neutron per square centimeter per second (cps/nv), provide pulse
signals to the source range channels. These detectors are installed on
opposite "flat" portions of the core at an elevation approximating the
quarter core height.
7.4-7 Rev 16 10/99 Compensated ionization chambers serve as neutron sensors for the intermediate range channels and are located in the same instrument wells and detector
assemblies as the source range detectors. These detectors have a nominal
thermal neutron sensitivity of 7.6 x 10
-14 amperes per neutron per square centimeter per second. Gamma sensitivity is less than 3 x 10
-11 amperes per Roentgen per hour when operated uncompensated, and is reduced to
approximately 3 x 10
-13 amperes/R/hr in compensated operation. The detectors are positioned at an elevation corresponding to the center mid core height.
The detector assemblies containing one each of the above mentioned detectors
use aluminum enclosures. High density polyethylene, used as a moderator-
insulator within the detector assemblies, will be confined at temperatures
associated with the loss-of-coolant accident. The detectors are connected to
the junction box at the top of the detector well by special high temperature, radiation resistant cables.
The remaining four detector assemblies contain the power range ionization
chambers. Each provides two current signals corresponding to the neutron
flux in the upper and lower sections of a core quadrant. These detectors
have a total neutron sensitive length of ten feet and a nominal thermal neutron
sensitivity for each section of 3.1 x 10
-13 amperes per neutron per square centimeter per second. Gamma sensitivity of each section is approximately
10-10 amperes per Roentgen per hour.
The detector assemblies for power range operation are installed vertically
and located equidistant from the reactor vessel at all points, and, to
minimize neutron flux pattern distortions, within one foot of the reactor
vessel. Cabling from individual detector wells to the containment
penetrations and to the instrument racks in the control room are routed in
individual conduits, with physical separation between the penetrations and
conduits associated with redundant protective channels.
7.4-8 Rev 16 10/99 The Backup Nuclear Instrumentation System employs two detector radial
locations, each containing one detector assembly (2 fission chambers)
installed around the reactor within the primary shield.
These detector assemblies have a nominal thermal neutron sensitivity of 25
counts per second per neutron per square centimeter per second (cps/nv).
Each of these detectors provides neutron flux indication over the range of
10-8 to 200 percent full power. These detector assemblies are installed on opposite `flat' portions of the core with the sensitive center line of the
detector
assembly aligned with the center line of the reactor.
These detectors and associated components are qualified to IEEE Standards
323-1974 and 344-1975 and are designed to function during and after a design
basis accident.
Cabling from these detector assemblies to the containment penetrations to the
signal processing equipment in the rod drive rooms is routed in dedicated
conduit and safety related qualified cable trays.
Source Range
The source range output information is tabulated in Table 7.4-1. The
detector for each source range channel is a BF 3 proportional counter, except for the backup detectors. The signal received from the counter has a range
of 1 to 10 6 cps.
7.4-8a Rev 16 10/99 pulses per second randomly generated and is received through a variable gain pulse preamplifier located outside the containment. The preamplifier
optimizes the signal-to-noise ratio and also furnishes high voltage coupling
to the detector.
The preamp has internal provisions for generating self-test frequencies of
10counts per second (CPS) and 10.24 x 10 3 CPS. These test oscillator circuits are energized by a switch located on the associated source range drawer. The
source range channel power supplies furnish low voltage for preamp operation
as well as low voltage for the drawer-mounted modules. The preamp is solid
state in design with discrete components and includes an impedance matching
network between the preamp output and the 75-ohm triaxial cable.
The preamp output is received at the amplifier located on the source range
drawer. This module provides amplification and discrimination, both of which
are adjustable. Discrimination is provided between neutron flux pulses and
combined noise and gamma-generated pulses. The discriminator supplies two
outputs: one output (isolated) to a scaler-timer unit on the visual-audio
channel drawer (see source range auxiliary equipment); and the other to a
pulse shaper (transistorized flip-flop circuit) which supplies a constant
amplitude pulse to the log integrator module within the source range drawer.
Logarithmic integration of the pulse signal is performed in another modular
unit to obtain an analog DC signal. The log signal is then amplified for
local indication on the front panel of the source range drawer, and is also
delivered through a parallel run to the source range level bistables and
isolation amplifier. The analog output signal is proportional to the
logarithm of the count rate being received from the sensor and is displayed
by the front panel meter on a scale calibrated logarithmically from 10 0 to 10 6 counts per second. The solid state isolation amplifier provides analog
outputs, all of which are adjustable through attenuator controls. The
outputs are used as follows: as remote indication (0-1 ma); as
7.4-9 Rev 16 10/99 remote recording (0-37.5 mv DC). A 0-10 VDC output is used by the start-up-rate amplifier to produce a start-up rate indication at the main
control board. A spare output (0-5 DC) is available.
All bistables will employ a basic plug-in module with the external wiring
determining the mode of operation (latching or non-latching and direction of
output change with rising power). Bistables will have two adjustments "Trip
Level" and "Differential". The first adjustment determines the trip point of
the bistable, while the second determines the "dead zone" difference between
the trip and release points of the bistable. The bistable module card will
include a relay driver circuit made up of a silicon controlled rectifier (SCR) and full-wave bridge configuration. The bistable output will control
the SCR gate which, in turn, controls conduction of the full-wave bridge
supplying the power to drive up to four 115 VAC Westinghouse BF relays. All
relays are located remote from the NIS racks.
Of the three bistables monitoring the source range level amplifier signal, one is a spare, one is used to monitor shutdown flux level only, and the
third monitors source range operation during shutdown and start-up operation and
provides a reactor trip on high flux level. The reactivity of the core
during shutdown is monitored by a bistable to ensure protection of plant
personnel
working in the containment. Bistable tripping will initiate local visual and
audible annunciation and remote audible annunciation of any abnormal increase
in core activity. Visual annunciation occurs at the NIS rack and on the main
control board. Audible annunciation is handled by the annunciator located in
the control room, and the evacuation horn located in the containment.
These annunciators ensure that plant personnel will be alerted to any
potentially unsafe condition. This bistable action will be manually blocked
by deliberate operator action during start up. Blocking
7.4-10 Rev 16 10/99 is continuously annunciated at the control board during source range operation and is automatically blocked above permissive P10. The bistable
trip point is approximately one-half decade above the flux level recorded
during full shutdown.
The source range level bistable monitors the core activity during the full
span of source range operation until such time as the intermediate range
channels assume control of that portion of the reactor protection which is
being supplied by nuclear instrumentation. At that time, when the
intermediate range permissive P6 is available, the source range reactor trip
bistable may be manually blocked and high voltage removed from the BF 3 detector by the operator actuating two momentary-contact switches located on
the main control board.
A fourth bistable-relay driver unit is used as a high voltage failure
monitor. Loss of this voltage actuates the bistable, the relay driver and
then the
associated relay. The relay provides control board annunciation through a
one of two matrix formed with a similar relay controlled by the other source
range. Failure of either source range high voltage actuates this common
annunciator on the main control board. During normal operation the source
range high voltage will be cut off (mentioned above) when manual block of the
source range trips is initiated. In this instance, loss of high voltage
annunciation will be intentionally defeated to prevent the alarming of a
condition which is not abnormal.
A test-calibrate module is also included in each source range drawer for self
check of that particular channel. A multi-position switch on the source
range front panel controls this module and also the operation of the built-in
oscillator circuits in the preamp. The module is capable of injecting test
signals of either 60, 10 3 , 10 5 and 10 6 counts per second at the input to the pulse amplifier, 10 or 10.24 x 10 3 counts per second to the preamplifier, or a variable d.c. voltage corresponding to 1 to 10 6 counts per second at the input to the level amplifier. An interlock
7.4-11 Rev 16 10/99 between the trip bypass switch and the test-calibrate switch will prevent inadvertent actuation of the reactor trip circuits, (i.e., the channel cannot
be put in the test mode unless the trip is defeated). Trip bypass will be
annunciated on the source range drawer and on the control board. Operation
of the test-calibrate module will be annunciated on the control board as "NIS
Channel Test." This common annunciator for all NIS channels will be alarmed
when any channel is placed in the test position and will alert the operator
that a test is being performed at the NIS racks.
Source Range Auxiliary Equipment
- a. Visual-Audio Count Rate
The visual-audio count rate receives a signal from each of the source range channels. This isolated signal originates at the discriminator
output in each source range. A switch on the audio count rate drawer
selects either source range channel for monitoring. The selected signal
is fed to a scaler-timer unit which permits count accumulation in the
preset time or preset count mode. A visual display to five decimal
places is presented through counting strips located on the front of the
audio count rate drawer.
A "Scale Factor" switch permits division of the scaler output signal by 10, 100, 1000, or 10000. This signal, derived from the binary coded
decimal output of the scaler, is conditioned and sent to two of the
audio amplifiers which power two speakers: one speaker located in the
control room, and the other in the containment. These speakers give
plant personnel an audible indication of the count rate. Since the
audio amp signal is taken from the coded scaler output, adjustment of
the scale factor switch will alter only the audible count rate. This
enables the operator to maintain the audible count rate at a
distinguishable level.
7.4-12 Rev 16 10/99
- b. Remote Count Rate Meter
The remote meter indication is an analog signal proportional to the count rate being received, and is obtained from the 0-1 ma isolation
amplifier output.
The meter is mounted on the control board and calibrated logarithmically from 10 0 to 10 6 cps. This meter gives the same indication at the control board as is displayed by the local meter on the corresponding
source range drawer.
The Backup NIS remote meter indication is an analog signal proportional to the count rate being received, and is obtained from a 4-20ma DC isolated output from the signal processor. This meter is mounted on the
control console and calibrated logarithmically from 0.1 to 10 5 counts per second.
- c. Remote Recorder
These multi-channel recorders are capable of recording all NIS channels.
Each NIS signal is directly connected to both recorders. The operator selects the signals to be displayed. In the case of the source range channels, 0-50 mVDC signals that are proportional to the count rate range of 10 0 to 10 6 CPS are supplied from isolation amplifiers for recording during source range operation.
- d. Start-up Rate Circuitry
The start-up rate drawer receives four input signals (0-10VDC) one from each of the primary source and intermediate range channels. Four rate
amplifier modules condition these signals and output four rate signals
to the respective control room S.U.R. meters (-.5 to 5 decades/minute).
A test module is provided which can inject a test signal into any one of
the rate circuits and can be monitored on a test meter mounted on the
front panel of this drawer. Two power supplies are provided to assure
rate indication from at least one Primary Source and intermediate Range
channel pair.
7.4-13 Revised 06/06/2005
Intermediate Range Intermediate Range output information is tabulated in Table 7.4-2. Each
intermediate range channel receives a direct current signal from a
compensated ion chamber and supplies positive high voltage and compensating (negative) high voltage to its respective detector. The compensating high
voltage is used to cancel the effects of gamma radiation on the signal
current being delivered to the intermediate range channel. Both high voltage
supplies will be adjustable through controls located inside the channel
drawer. The detector signal is received by the intermediate range
logarithmic amplifier. The modular unit, comprised of several operational
amplifiers and associated discrete solid state components, produces an analog
voltage output signal which is proportional to the logarithm of the input
current. This signal is used for local indication and it is monitored by the
isolation amplifier and the various bistable relay-driver modules within the
intermediate range drawer. A 10
-11 ampere signal is continuously inserted and serves as a reference during gamma compensation. Local indication is
provided by a meter mounted on the front panel of the drawer which has a
logarithmic scale calibration of 10
-11 to 10 3 amperes. The isolation amplifier is the same solid state module that is used in the source range; it supplies
the same outputs and for the same usage. Six bistable relay-driver units are
used in the intermediate range drawer to provide the following functions:
One monitors the positive high voltage
One monitors the compensating high voltage
One provides the permissive P6
One provides rod-stop (blocks rod withdrawal)
One provides reactor trip
One serves as a spare
7.4-14 Rev 16 10/99 The intermediate range permissive P-6 bistable drives two Westinghouse BF relays (for redundancy) and the relays from each channel are combined in 1 of
2 matrices to provide the permissive function and control board annunciation
of permissive availability. Permissive P-6 permits simultaneous manual
blocking of the source range trips and removal of the source range detector
high voltage. Once source range blocking has been performed, the operator
may, through administrative action, defeat permissive P-6 and reactivate the
source range high voltage and trip functions if required. This defeat is
accomplished by the coincident operation of two control board mounted, momentary-contact switches. This provision, however, is only operational
below permissive P-10 which is supplied by the power range channels. Above
P-10, the defeat circuit is automatically bypassed and permissive P-6 is
maintained which, in effect, maintains source range cutoff. The level
bistable relay-driver unit which provides the intermediate range rod-stop
function also drives two Westinghouse BF relays. Again, 1 of 2 matrices
formed by the relays from the two intermediate range channels supply the
rod-stop function and control board annunciation. Blocking of the outputs of
these matrices is administratively performed when nuclear power is above
permissive P-10 and can only be accomplished by deliberate operator action on
two control board mounted switches.
The intermediate range reactor trip function is provided by a similar circuit
arrangement, the only difference being the trip point of the bistable units.
The same control board switches which control blocking of the rod stop
matrices also provide blocking action for the reactor trip matrices. These
blocks are manually inserted when the power range of instrumentation
indicates proper operation through activation of the P-10 permissive
function. On decreasing power, however, the more restrictive intermediate
range trip functions are automatically reinserted in the protective system.
While these trips are blocked, there will be continuous illumination on the
main control board of "Intermediate Range Trip Blocked." The high voltage
failure monitors provide both local and remote annunciation upon failure of
the respective high voltage supplies. A common "Intermediate Range Loss of
Detector Voltage" and separate "Intermediate Range Loss of Compensate
Voltage" are provided as control board annunciators for the intermediate
ranges.
7.4-15 Rev 16 10/99 Administrative testing of each intermediate range channel is provided by a built-in test-calibrate module which injects a test signal at the input to
the log amplifier. The signal is controlled by a multi-position switch on
the front of each intermediate range drawer. A fixed 10
-11 through 10
-3 ampere signal is available along with a variable 10
-10 through 10
-3 signal, selectable in decade increments.
As in source range testing, the test switch on the intermediate range must be
operated in coincidence with a trip bypass on the drawer. An interlock
between these switches prevents injection of a test signal, until the trip
bypass is in operation. Removal of the trip bypass also removes the test
signal.
Intermediate Range Auxiliary Equipment
- a. Remote Meter
The remote meter indication is in the form of an analog signal (0-1 ma) proportional to the ion chamber current. The isolation amplifier in
each channel supplies this output to a separate meter. Meter
calibration is 10
-11 to 10-3 amperes.
- b. Remote Recorder
These multi-channel recorders are capable of recording all NIS channels.
Each NIS signal is directly connected to both recorders. The operator selects the signals to be displayed. In the case of the intermediate range channels, 0
-50 mVDC signals that are proportional to the ion chamber current range of 10
-11 to 10-3 amperes are supplied from isolation amplifiers for recording during intermediate range operation.
7.4-16 Revised 06/06/2005
- c. Start-Up-Rate Circuitry
The start-up rate drawer receives four input signals (0-10VDC) one from each of the source and intermediate range channels. Four rate amplifier
modules condition these signals and output four rate signals to the
respective control room S.U.R. meters (-.5 to 5 decades/minute). A test
module is provided which can inject a test signal into any one of the
rate circuits and can be monitored on a test meter mounted on the front
panel of this drawer. Two power supplies are provided to assure rate
indication from at least one Source and Intermediate Range channel pair.
Power Range
The power range output information is tabulated in Table 7.4-3. The power
range detector is a long uncompensated ion chamber assembly which is
comprised of two separate neutron sensitive sections. Each section supplies
a current signal to the associated power range. There is one high voltage
power supply per channel and it supplies voltage to both sections of the
associated detector. The two signals are received at the channel input and
handled through separate ammeter-shunt assemblies. Four full-scale ranges
can be selected for each ammeter through switches located on the front panel
of the power range drawer, 100 ua, 500 ua, 1 ma, and 5 ma D. C. The switch
selects shunt resistors for the meter but never interrupts the ion chamber
signal to the power range channel. The circuit is so designed that a failure
of the meter or switch will not interrupt the signal to the average power
circuitry.
The individual currents are displayed on the two front panel ion chamber
current meters and are then sent to separate isolation amplifiers. There are
two isolation amplifiers monitoring each of the two individual current signals. The unit feeding the T protection function is being used for its impedance matching characteristics rather than isolation. All of the isolation amplifiers are capable of providing the same output ranges as the
isolation amplifiers previously described in relation to the source
7.4-17 and intermediate ranges. Two of the isolation amplifiers, one monitoring each of the currents, supply signals to the T setpoint program. The other two isolation amplifiers provide output for the remote meter, axial deviation comparators, and upper and lower section comparators. The individual current
signals are summed and then sent to a summing amplifier module which outputs
a linear 0-10V D. C. signal proportional to reactor power. The summing and
level amplifier has two controls: one is a "Zero" adjust located on the
module itself, while the other is a "Gain" adjust with a calibrated dial
located on the drawer's front panel. The output signal from this unit
corresponds to 0 to 120 percent of full power and is displayed on a percent
full power meter on the front panel of the power range drawer. This same
signal is delivered directly to three isolation amplifiers, a dropped rod
sensing assembly, and six bistable relay-driver modules. These isolation
amplifiers are identical to those previously described and the outputs are
the same in number and range but are used in different functions. (Specific
outputs from the amplifiers are discussed in the auxiliary equipment section
which follows.)
The dropped-rod sensor assembly is an operational amplifier unit which
incorporates an adjustable lag network at one input and a non-delayed signal
on the other. The unit compares the actual power signal with the delayed
power signal received through the lag network and amplifies the difference.
This amplified differential signal is delivered to a bistable relay-driver
unit which trips when the level of this signal exceeds a preset amount.
Tripping of this unit indicates a power level change over the lag period
which would be indicative of a dropped rod. This bistable unit is a latching
type, ensuring that the necessary action will be initiated and carried to
completion. Specifically, the unit controls dual Westinghouse NBF relays
which, in 1 of 4 logic matrices provide a control board annunciation signal
and a spare input signal. Automatic rod withdrawal by the reactor control
system has been permanently disabled. Manual rod withdrawal is not blocked
by nuclear instrumentation system power range rod drop detection. A reset
switch on the associated power range drawer must be operated manually to
remove the trip functions and reset the bistable.
7.4-18 Revised 04/27/2010 C24 The bistable units which sense the power level signal as derived by the linear amplifier are non-latching and perform the following functions: 1)
overpower rod-stop (blocks manual rod withdrawal); 2) permissive functions (provisions for three are incorporated in the design but are not required on
all plants); 3) low-range reactor trip; and 4) high-range reactor trip.
The overpower rod-stop and permissive bistables are units which trip on high
power level and control Westinghouse NBF relays in the remote relay racks.
The rod-stop relay matrices (1 of 4) provide a rod-stop function to the rod
control system and a main control board annunciation. Two-of-four logic, developed by relays controlled through the respective power range bistables, provide the signals required for the permissive functions. One set of relays
provide permissive P-10, as was previously discussed with regard to its use
in the source range and intermediate range. Two other groups of relays are
available to provide inputs to two additional permissive functions when
required. These bistable functions, when used, provide permissive P-8.
Permissive P-8 and P-10 are supplied solely by nuclear instrumentation.
For this reason, the nuclear instrumentation design provides for status light
indication of P-8 and P-10 availability. Permissive P-10 is used in all
three ranges of nuclear instrumentation while P-8 is provided by nuclear
instrumentation for use in the reactor protection system.
The low range trip bistable actuates two Westinghouse NBF relays in the logic
system. The two relays provide redundancy within the logic portion of the
protection system. Each relay is used in a separate matrix with the relays
from the other power range channels to continue the redundancy. The logic
circuitry formed by the contacts on these relays provide for 1 of 4 and 2 of
4 logic outputs. The low range trip relays provide the following functions:
- 1) spare input ;2) low range trip annunciation (2 of 4 coincidence); 3)
reactor-trip signal to reactor protection system (2 of 4 conicidence); and 4)
annunciation of "Single Channel Low Range Trip" (1 of 4).
7.4-19 Revised 04/27/2010 C24 Provisions for manually blocking these functions become available when 2 of 4 power ranges exceed the permissive P-10 level. Operator action on two
control board mounted momentary-contact switches then initiates the blocking
action. A control board permissive status light, "Power Range Low Range Trip
Blocked", will be illuminated continuously when the trip function is blocked.
On decreasing power, 3 of 4 power ranges below the P-10 power level will
automatically reactivate the low range trip.
The high range reactor trip logic circuitry is developed identical to the low
range reactor trip circuitry, but no provision for blocking is included.
The high range trip remains active at all times to prevent any continuation
of an overpower condition.
An additional bistable unit monitors the high voltage power supply in the
power range. Operation of this unit is identical to that for the source and
intermediate ranges. The bistable provides relay actuation in the remote
relay racks on failure of power range high voltage. While there is a
separate relay for each power range, they control a common "Power Range Loss
of Detector Voltage" annunciator on the main control board. Separate local
indication of high voltage failure is provided on the power range drawers.
The test-calibrate module which is provided on each power range is capable of
injecting test signals at several points in the channel. In all cases, the
test signals are superimposed on the normal signal. An interlock between the
bypass switch and channel test switch is provided as was done in the source
and intermediate ranges. The bypass switch from each power range will
activate a common annunciator,
7.4-20 Rev. 13 10/96
"NIS Trip Bypass," but individual bypass status lights will identify the particular channel. The remaining bistables which will be affected during
channel test do not require bypasses since they operate in 2 of 4 logic.
Test signals can be injected independently or simultaneously at the input of
either ammeter-shunt assembly to appear as the individual ion chamber
currents. Operation of the test-calibrate switch on any power range will
cause the "Channel Test" annunciator to be alarmed on the control board.
Power Range Auxiliary Equipment
- a. Comparator
The comparator received an isolated signal from each of the four power ranges. These signals are conditioned in separate operational
amplifier circuits and then compared with one another to determine if a
preset amount of deviation of power levels has occurred between any two
power ranges. Should such a deviation occur, the comparator output
will operate a remote relay to actuate the control board annunciator, "Power Range Channel Deviation". This alarm will alert the operator to
either a power unbalance being monitored by the power ranges or to a
channel failure. Through other indicators, the operator can then
determine the deviating channel(s) and take corrective action. Should
correction of the situation not be immediately possible (e.g., a
channel failure, rather than reactor condition), provisions are
available to eliminate the failed channel from the comparison function.
The comparator can then continue to monitor the active channels.
- b. Remote Recorder
These multi-channel recorders are capable of recording all NIS channels. Each NIS signal is directly connected to both recorders.
The operator selects the signals to be displayed. In the case of the power range channels, 0-50 mVDC signals that are proportional to 0-120%
full power are supplied from isolation amplifiers to allow continuous monitoring during power range operation.
7.4-21 Revised 06/06/2005 A signal input is also provided to the Safety Parameter Display System (DCS / SPDS) for display and recording the power range flux.
- c. Remote Meter
The Primary NIS remote meters receive the 0-1 ma isolated output that is available from each power range. This indication corresponds to
that shown on the power range drawer. The signal is displayed on a
meter scale calibrated from 0 to 120 percent of full power. The backup
NIS remote meters receive a 4-20 ma DC isolated output that is
available from each channel. The meter scales are calibrated from 10
-8 to 200 percent of full power.
- d. Overpower Recorder
Inputs routed to vertical panel recorders monitor the individual average power indications from the four primary power ranges, capable
of displaying overpower excursions up to 200 percent of full power. A
power range isolated output of 0-50 mVDC will correspond to the range
of zero percent to 200 percent of full power.
7.4-22 Revised 01/31/2013 C26
- e. Remote Meter (Delta Flux)
Four control board mounted meters display the flux difference between the upper and lower ion chambers directly for each of the primary power
range detectors.
- f. Axial Flux Comparator
The axial flux monitoring system contains four comparators, each of which receives an upper and lower section signal from one primary power
range. If the axial flux difference for any primary channel exceeds
setpoint, an alarm is actuated.
7.4-22a Rev 16 10/99 Flux Deviation and Miscellaneous Control and Indication Drawer Indicating lights (one per power range channel) are provided on this drawer
to be used during test of the dropped rod annunciation. Illumination of one
of the lights indicates completion of the relay tripping function for the
channel under test.
Switches are also provided on this drawer to permit a failed power range
channel's overpower-rod stop function to be bypassed. Upper and lower
section comparators are provided. The upper section comparator actuates an
alarm when any upper section signal deviates from the average of the upper
section signals by a preset amount. The lower section comparator performs in
a similar manner. Switches are provided to bypass a failed channel.
7.4.4 SYSTEM EVALUATION
Philosophy and Set Points
During shutdown and operation, three discrete independent levels of nuclear
protection are provided from the three ranges of out-of-core nuclear
instrumentation. The basic protection philosophy is that the level
protection is present in all three ranges to provide a reliable, rapid and
restrictive protection system which is not dependent upon operation of higher
range instrumentation.
Reliability is obtained by providing redundant channels which are physically
and electrically separated. Fast trip response is an inherent advantage of
using level trip protection in lieu of start-up rate protection (with a long
time constant) during start-up. More restrictive operation is an inherent
feature since an increase in power cannot be performed
7.4-23 Revised 04/27/2010 C24 until satisfactory operation is obtained from higher range instrumentation which permits administrative bypass of the lower range instrumentation. On
decreasing power level, protection is automatically made more restrictive.
Startup accidents while in the source range are rapidly terminated without
significant increases in nuclear flux and with essentially no power
generation or reactor coolant temperature increase.
The indications and administrative actions required by this protection system
are readily available to the operator and should result in a safe, uncomplicated increase of power.
Reactor Trip Protection
During reactor start-up the operator will be made aware of satisfactory
operation of one or more intermediate range channels by annunciation (audible
and visual) at the control board. The source and intermediate range flux
level information is also readily available on recorders and indicators at
the control console. At this time, if both intermediate range channels are
functioning properly, the operator would depress the two manual block
switches associated with the source range logic circuitry, thus causing
cutoff of source range detector voltages and blocking the trip logic outputs.
The manual block should not be initiated, however, until at least one decade
of satisfactory intermediate range operation is obtained. If one
intermediate range channel is not functioning, normal power increase could be
performed if desired. The permissive P6 annunciation is continuously
displayed by the control board status lights.
Continuation of the start-up procedure in the intermediate range would result
in a normal power increase and the receipt of a permissive signal from the
power range channels when two-of-four channels exceed 10 percent of full
power. The operator would be alerted to this condition by a control board
permissive status light. Indicators (one per channel) and a recorder also
indicate unit status in terms of percent full power. If the operator does
not block the I.R. trip and continues the power increase,
7.4-24 a rod stop will automatically occur from either of the intermediate range channels. The operator should then depress the momentary "Manual Block" push
buttons associated with the intermediate range rod stop and reactor trip
logic. This would transfer protection to the low-range trips for the four
power range channels. The permissive P-10 status light would be continuously
displayed as was P-6. The low-range manual block switches (two) must be
depressed to initiate blocking prior to continuation of the power increase.
The permissive functions associated with administrative trip blocking and
automatic reactivation are provided with the same separation and redundancy
as the trip functions.
When decreasing power operation to lower levels, more restrictive trip
protection is automatically afforded when 3 of 4 power range channels are
below P-10 permissive and when 2 of 2 intermediate range channels are below
the permissive P6.
Rod-Drop Protection
Rod drop annunciation is provided by the power range instrumentation. Rod
position system rod bottom bistables provide rod drop annunciation and the
protective function of manual rod withdrawal block. The nuclear
instrumentation rod-drop annunciation is provided by comparison of the
average nuclear power signal with the same signal which is conditioned by an
adjustable lag network. This method provides a response to dynamic signal
changes associated with a dropped rod condition, but does not respond to the
slower signal changes associated with normal operation. Annunciation from at
least one of the four power range channels will occur for any dropped rod
condition.
7.4-25 Revised 04/27/2010 C24C24 Control and Alarm Functions Various control and alarm functions are obtained from the three ranges of
out-of-core primary nuclear instrumentation during shutdown, startup and
power operation. These functions are used to alert the operator of
conditions which require administrative action and alert personnel of unsafe
reactor conditions. The power and intermediate ranges provide manual rod
withdrawal block signals to the rod control system to avoid unnecessary
reactor trips; auto rod withdrawal signals are permanently disabled.
- a. Source Range
No control functions are obtained from the source range channels.
Alarm functions are provided, however, to alert the operator of any
inadvertent changes in shutdown reactivity. Visual annunciation of
this condition is at the control board, with audible annunciation
performed in the containment and control room. This alarm can either
be blocked prior to startup or can serve as the startup alarm in conjunction with administrative procedures.
The backup nuclear instrumentation system provides visual and audible annunciation in the control room and audible annunciation in the containment for hi-flux at shutdown.
- b. Intermediate Range
Both alarm and control functions are supplied by the primary NIS intermediate range channels. Blocking of rod withdrawal is initiated by either intermediate range on high flux level. This condition is
alarmed at the control board to alert the operator that rod-stop has
been initiated. In addition, the primary NIS intermediate ranges
provide status light indication when either channel exceeds the P-6
permissive level. This alerts the operator to the fact that he must
take administrative action to manually block the source range trips to
prevent an inadvertent trip during normal power increase.
The backup nuclear instrumentation system does not provide any control or alarm functions for the intermediate range.
7.4-26 Revised 04/27/2010 C24
- c. Power Range
The primary NIS power ranges provide alarm and control functions similar to those in the primary NIS intermediate ranges. An overpower
rod-stop function from any of the four power range channels inhibits
manual rod withdrawal and is alarmed at the control board. The power
ranges also provide status light indication when 2 of 4 channels exceed
permissive P-10 level. As in the case of P-6 in the intermediate
range, this alerts the operating personnel that administrative action (namely, blocking of intermediate and low range trips) is required
before any further power increase may take place.
The primary NIS power ranges also have provision for an additional permissive function P-8. A permissive status light is provided for
P-8, "Single Loop Flow Trip Blocked". The extinguishing of the P-8
permissive status light alerts the operator that the low flow trips and "pump breaker open" trips are now active. These trips are blocked
while the status light is alarmed. Additional functions are provided
in the power range of operation. A dropped control rod will be sensed
by one or more of the power range channels, and this condition will
annunciate.
Another function of the primary NIS is a power range channel deviation alarm. This alarm is furnished by the comparator channel through a
comparison of the average power level signals being supplied by the
power ranges. Actuation of this alarm alerts the operator to a power
unbalance between the channels so that corrective action can be taken.
Finally, two signals, one signal from each ion chamber isolation
amplifier, are supplied by power ranges 1, 2, and 3 to the reactor
protection system.
The backup nuclear instrumentation system does not provide any control or alarm functions for the power range.
7.4-27 Revised 04/27/2010 C24C24 Loss of Power The nuclear instrumentation draws its primary power from the vital instrument
buses whose reliability is discussed in Section 8. Redundant NIS channels
are powered from separate buses. Loss of a single vital instrument bus would
result in the initiation of all reactor trips associated with the primary NIS
channels deriving power from that source. During power operation, the loss
of a single bus would not result in a reactor trip since the power range
reactor trip function operates from a 2-of-4 logic. If the bus failure
occurred during source or intermediate range operation (1-of-2 logic) a
reactor trip condition would result.
The backup nuclear instrumentation system does not perform any protective or
control functions.
Safety Factors
The relation of the power range channels to the Reactor Protective System has
been described in Section 7.2. To maintain the desired accuracy in trip
action, the total error from drift in the primary NIS Power range channels will be held to 1.0 percent at full power (0.5% for Power Range Neutron Flux - High Setpoint). Routine tests and recalibration will ensure that this degree of deviation is not exceeded. Bistable trip set points of the primary NIS power range channels will also be held to an accuracy of 0.5 percent of full power.
7.4.5 REGULATORY GUIDE 1.97, REVISION 3
A review of Turkey Point Units 3 and 4 accident monitoring instrumentation
and control systems was conducted against the requirements of Regulatory
Guide 1.97, Revision 3. Section 7.5.4 presents the requirements of Regulatory
Guide 1.97, and the results of the conducted review.
7.4-28 Revised 04/17/2013 C26 TABLE 7.4-1
SOURCE RANGE
Signal and Source Destination and/or Function
- 1. Isolation Amplifier
- a. 0-10VDC Auxiliary Channel (S.U.R.)/DDPS
- b. O-5VDC Spare
- c. O-5VDC SPDS/SAS/DCS
- d. 0-1 mADC Remote Meter (CPS)
- e. 0-50 mVDC Remote Recorder
- 2. Bistable Amplifiers
- a. 115VAC Misc. Proc. Relay Rack (Spare)
- b. 115VAC Misc. Proc. Relay Rack
(Hi Flux Level @ Shutdown)
- c. 115VAC Reac. Prot. Relay Rack
(Source Range Reactor.Trip)
- d. 115VAC Misc. Proc. Relay Rack
(Annunciate "Source Range Loss
of Detector Voltage")
- 3. Manual Block (115 VAC) Misc. Proc. Relay Rack
(Block Hi Flux Level @ Shutdown)
- 4. Trip Bypass (115VAC) Reac. Prot. Relay Rack
(Block of S. R. Reactor Trip)
- 5. Test-Calibrate (115VAC) Misc. Proc. Relay Rack
("NIS Channel Test" - CB)
- 6. Discriminator (1-10E 6 Cps) Source Range Auxiliary Channel
(Visual-Audio)
BACKUP NIS SOURCE RANGE
- 1. Isolator
- a. 4-20 mADC Control Room Meter (CPS)
- b. 4-20 mADC Alternate Shutdown Panel Meter (CPS) Ch B Only
- c. 4-20 mADC SPDS
- 2. Bistable
- a. N.O. Contact Vertical Panel Annunciator
(Hi Flux Level @ Shutdown)
Communications Box
(CTMT Evacuation Alarm)
Revised 01/31/2013 C26 TABLE 7.4-2
INTERMEDIATE RANGE
Signal and Source Destination and/or Function
- 1. Isolation Amplifier
- a. 0-10 VDC Auxiliary Channel (S.U.R.)/DDPS
- b. 0-1 mADC Remote Meter (Ampere)
C. O-5O mVDC Remote Recorder
- d. 0-5 VDC SPDS/SAS/DCS
- e. 0-5 VDC Spare
- 2. Bistable Amplifiers
- a. 115 VAC Relay Rack (Spare)
- b. 115 VAC Reac. Prot. Relay Rack
(Intermediate Range Permissive P-6)
C. 115 VAC Misc. Proc. Relay Rack
(Intermediate Range Rod-Stop)
- d. 115 VAC Reac. Prot. Relay Rack
(Intermediate Range Reactor Trip)
- e. 115 VAC Misc. Proc. Relay Rack
(Annunciate "I.R. Loss of Detector
Voltage")
- f. 115 VAC Misc. Proc. Relay Rack
(Annunciate "I.R. Loss of Compensating
voltage")
- 3. Trip Bypass (115 VAC) Reac. Prot. Relay Rack
(Block of Rod-Stop and Reactor
Trip)
- 4. Test-Calibrate (115 VAC) Misc. Proc. Relay Rack
("NIS Channel Test" - CB)
Revised 01/31/2013 C26 TABLE 7.4-3 SHEET 1 of 3 POWER RANGE Signal and Source Destination and/or Function
- 1. Isolation Amplifier (Ion
Chamber A)
- a. 0-10 VDC Upper Section Comparator/DDPS
- b. 0-5 VDC Axial Flux Deviation Panel
- c. 0-1 mADC Remote Meter (Delta Flux)
- d. 0-5 VDC SPDS/SAS/DCS
- e. 0-50 mVDC Spare
- 2. Isolation Amplifier
(Ion Chamber A)
- a. 0-10 VDC Overpower-Overtemperature T (Power ranges Compensation 1, 2 & 3 only)
- b. 0-5 VDC Spare
- c. 0-1 mADC Spare
- d. 0-5 VDC Spare
- e. 0-50 mVDC Spare
- 3. Isolation Amplifier (Ion
Chamber B)
- a. 0-10 VDC Lower Section Comparator/DDPS
- b. 0-5 VDC Axial Flux Deviation Panel
C. 0-1 mADC Remote Meter (Delta Flux)
- d. 0-5 VDC SPDS/SAS/DCS
- e. 0-50 mVDC Console Recorder (Delta Flux)
- 4. Isolation Amplifier (Ion Chamber B)
- a. 0-10 VDC Overpower-Overtemperature T (Power ranges Compensation 1,2 & 3 only)
- b. 0-5 VDC Spare
- c. 0-1 mADC Spare
- d. 0-5 VDC Spare
- e. 0-50 mVDC Spare
Revised 01/31/2013 C26C26 TABLE 7.4-3 (cont'd)
SHEET 2 of 3
Signal and Source Destination and/or Function
- 5. Isolation Amplifier
(Average Power)
- a. 0-10 VDC DDPS
- b. 0-5 VDC Spare
- c. 0-1 mADC Remote Meter (Percent Full Power)
- d. 0-50 mVDC Remote Recorder
- e. O-5 VDC SPDS/SAS/DCS
- 6. Isolation Amplifier (Average Power)
- a. 0-10 VDC Power Mismatch (Power Range 4 only)
- b. 0-5 VDC Spare
- c. 0-1 mADC Spare
- d. 0-50 mADC Spare
- e. 0-5 VDC Spare
- 7. Isolation Amplifier (Average Power)
- a. 0-10 VDC Comparator
- b. 0-5 VDC Spare
- c. 0-1 mADC Spare
- d. 0-50 mVDC Overpower Recorder
- e. 0-5 VDC Spare
- 8. Bistable Amplifiers
- a. 115 VAC Reac. Prot. Relay Rack
Annunciation ERDADS
- b. 115 VAC Misc. Proc. Relay Rack
(Overpower Rod Stop)
- c. 115 VAC Reac. Prot. Relay Rack
(Permissive P-8)
Revised 01/31/2013 C26 TABLE 7.4-3 (cont'd)
SHEET 3 of 3
Signal and Source Destination and/or Function
- d. 115 VAC Reac. Prot. Relay Rack
(Permissive P-10)
- e. 115 VAC Reac. Prot Relay Rack
(Spare Permissive)
- f. 115 VAC Reac. Prot. Relay Rack
(Low Range Reactor Trip)
- g. 115 VAC Reac. Prot. Relay Rack
(High Range Reactor Trip)
- h. 115 VAC Misc. Proc. Relay Rack
(Annunciate "Power Range Loss of
Detector Voltage")
- 9. Test-Calibrate (115 VAC) Misc. Proc. Relay Rack
(NIS Channel Test-CB)
- 10. Block Rod Drop (115 VAC) Reac. Prot. Relay Rack
Detection (Block of Rod-Drop Circuit)
BACKUP NIS POWER RANGE
- 1. Isolator
- a. 4-20 mADC Control Room Meter (Percent Full Power)
- b. 4-20 mADC Alternate Shutdown Panel Meter (Percent Full Power) CH-B only
- c. 4-20 mADC SPDS/DCS
- 2. Bistable
- a. N.O. Contact Vertical Panel Annuciator (System Trouble)
Revised 01/31/2013 C26
REV. 3-7/85 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT NEUTRON DETECTORS AND RANGE OF OPERATION FIGURE 7.4-1
FINAL SAFETY ANALYSIS REPORT FIGURE 7.4-2a REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 16
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 NUCLEAR INSTRUMENTATION TRIP SIGNALS LOGIC DIAGRAM FIGURE 7.4-2a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.4-2b REFER TO ENGINEERING DRAWING 5610-T-L1 , SHEET 17
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 NUCLEAR INSTRUMENTATION PERMISSIVES AND BLOCKS LOGIC DIAGRAM FIGURE 7.4-2b
REV.16 (10/99)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 PLAN VIEW INDICATING DETECTOR LOCATION RELSTIVE TO CORE FIGURE 7.4-3
7.5 ENGINEERED SAFETY FEATURES INSTRUMENTATION 7.5.1 DESIGN BASIS
The engineered safety features instrumentation measures temperatures, pressures, flows, and levels in the reactor coolant system, steam system, reactor containment and auxiliary systems, actuates the engineered safety
features, and monitors their operation. Process variables required on a
continuous basis for the startup, operation, and shutdown of the unit are
indicated, recorded and controlled from the control room. The quantity and
types of process instrumentation provided ensures safe and orderly operation of
all systems and processes over the full operating range of the units.
Certain controls and indicators which require a minimum of operator attention, or are only in use intermittently, are located on local control panels near the
equipment to be controlled. Monitoring of the alarms of such control systems
is provided in the control room.
Engineered Safety Features Protection Systems
Criterion: Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered
safety features. (1967 Proposed GDC 15)
Instrumentation and controls provided for the protective systems are designed
to trip the reactor, when necessary, to prevent or limit fission product
release from the core and to limit energy release; and to control the operation
of Engineered Safety Features equipment.
The engineered safety features systems are actuated by the engineered safety
features actuation channels. Each coincidence network energizes an engineered
safety features actuation device that operates the associated engineered safety features equipment, motor starters and valve operators. The channels are
designed to combine redundant sensors, independent channel circuitry, coincident trip logic and different parameter measurements so that a safe and
reliable system is provided in which a single failure will not defeat the
protective function. The action initiating sensors, bistables and logic are
shown in the figures included in the detailed Engineered Safety Features
Instrumentation Description given in the System Design section. The
Engineered Safety Features instrumentation system actuates (depending on the
severity of the condition) the Safety Injection System, containment
isolation, the Emergency Containment Cooling System and the Containment Spray
System.
7.5-1 Revised 04/17/2013 C26 Availability of DC control power to the logic matrix is required for train
operability. Availability of DC control power is continuously monitored and
annunciated in the control room. The loss of instrument power to an
engineered safety features instrument channel (comparator and logic relay),
places that channel in the trip mode.
The passive accumulators of the Safety Injection System do not require signal
or power sources to perform their function. The actuation of the active
portion of the Safety Injection System is obtained from low pressurizer
pressure, high containment pressure, high differential steam line pressure or
high steam line flow in coincidence with low steam generator pressure or low
Tavg.
The Containment Emergency Cooling System is in the automatic sequence which
actuates the Engineered Safety Features upon receiving the necessary signals
indicative of an accident condition.
Containment spray is actuated by 2/3 High coincident with 2/3 High-High
containment pressure signals as noted in Table 7.2-1.
The containment isolation signals provide the means of isolating the various
pipes passing through the containment walls as required to prevent the
release of radioactivity to the outside environment in the event of a loss-
of-coolant accident. The actuation of the containment isolation may be found
in Table 7.2-1 or in Figure 7.2-8e.
7.5.2 SYSTEM DESIGN
Engineered Safety Features Actuation Instrumentation Description
Figures 7.2-8a, 7.2-8b, 7.2-8c and 7.2-8e show the actuation logic for the
engineered safety features.
The same channel isolation and separation criteria as described for the
reactor protection circuits are applied to the engineered safety features
actuation circuits.
The Engineered Safety Features actuation instrumentation automatically
commences the protective actions as noted on Table 7.2-1.
7.5-2 Revised 04/17/2013 C26 Feedwater Any safety injection signal will isolate the main feedwater lines by closing
all control valves (main and bypass valves), tripping the main feedwater
pumps and thereby closing the pump discharge valves, and all backup feedwater
isolation valves. The auxiliary feedwater system is actuated by the safety
injection signal.
Indication
All transmitted signals (flow, pressure, temperature, etc.) which can cause
actuation of the engineered safety features are either indicated or recorded
for every channel.
The d-c control supply associated with the engineered safety features is
designed to meet the single failure criterion such that one failure will not
prevent actuation of sufficient engineered safety features, to meet the core
and containment cooling criterion.
Engineered Safety Features Instrumentation
The following instrumentation ensures monitoring of the effective operation
of the Engineered Safety Features.
Containment Pressure
Two containment pressure channels derived from pressure taps reflect the
effectiveness of the containment and cooling systems and other Engineered
Safety Features. Redundant pressure transmitters are provided for the narrow
range (-6 to +18 psig), and additional redundant pressure transmitters cover
the wide range of 0-180 psig. High pressure indicates high temperatures and
reduced pressure indicates reduced temperatures. Indicators and alarms are
provided in the control room to inform the operator of system status and to
guide actions taken during recovery operations. Containment pressure
indication will be used to distinguish between various incidents.
2/3 High coincident with 2/3 High-High containment pressure signals are
required to completely isolate the containment. Each train of containment
pressure switches (located in the cable penetration room) have their own
connection to the containment. Each switch provides input to annunciation in
the control room.
7.5-3 Revised 04/17/2013 C26 Refueling Water Storage Tank Level
Level instrumentation for the refueling water storage tank consists of two
independent channels. Each channel provides remote indication (on the main
control board). Each channel provides annunciation for the technical
specification minimum level, in addition to high, low, and low-low level
alarms.
Safety Injection System Pumps Discharge Pressure
A Discharge pressure channel clearly shows that the Safety Injection System
pumps are operating. A post-accident flow channel is provided for safety
injection flow indication. These transmitters are outside the containment.
Safety Injection Pump Energization
Safety Injection pump motor power feed breakers indicate that they have
closed by energizing indicating lights on the control board.
Radioactivity
Means are provided to measure the radioactivity in the containment atmosphere
after the incident, since this information will be required for any
subsequent entry into the containment following a LOCA. The containment
system particulate and gaseous monitoring equipment could provide information
useful in post-accident recovery operations, providing containment pressure
is below 5 psig.
Valve Position
All Engineered Safety Features remote-operated valves have position
indication on the control board to show proper positioning of the valves.
Air-operated and solenoid-operated valves move in a preferred direction with
the loss of air or power. After a loss of power to the motors, motor-operated valves remain in the same position as they were prior to the
loss of power.
7.5-4 Revised 04/17/2013 Emergency Containment Coolers
The total cooling water discharge flow is indicated, and the exit temperature
of each of the coolers is recorded in the control room. In addition, each
CCW return header is monitored for radiation and alarmed in the control room
if high radiation should occur. These monitors are common to each CCW return
header and the faulty cooler can be located by remote valving.
Containment Level Instrumentation
The containment level instrumentation consists of two sub-systems: (1)
containment sump (narrow range), and (2) containment level (wide range).
Both the containment sump and containment level instrumentation consist of
redundant level transmitters designed to operate in a post-accident
environment. The signal receivers are located outside containment, remote
from the sensing elements. Indication and recording are provided in the
control room.
The containment sump level transmitter is a multi-element system with a
36-inch unit at the bottom of the sump and four 90-inch units, which extend
up to just below the containment 14'-0" elevation slab. Adjacent units
overlap for continuous level indication.
The containment level transmitter is a single-element device (90-inch unit),
which provides a range from just above the 14'-0" elevation to approximately
21'- 6" elevation.
Miscellaneous Instrumentation
In addition to the above, the following local instrumentation is available.
- a. Residual heat removal pumps discharge pressure b. Residual heat exchanger exit temperatures
- c. Containment spray test lines total flow
- d. Safety injection test line pressure and flow
7.5-5 Revised 04/17/2013 Alarms Visual and audible alarms are provided to call attention to abnormal
conditions. The alarms are of the individual acknowledgement type. That is, the operator must recognize and acknowledge the alarm for each alarm point.
Operators have the means to silence the alarms with a timed auditory silence
for use during transient conditions coincident with high alarm conditions.
Instrumentation Used During Loss-of-Coolant Accident
Instruments to be provided and designed to function following a major loss-
of-coolant accident are those which initiate or otherwise govern the
operation of engineered safety features. Pressurizer pressure and level, and steam generator level and main steam flow are typical examples of sensors
that are located inside the containment because an equivalent signal cannot
be obtained from a sensor location outside containment.
It should be emphasized, however, that for a large loss-of-coolant accident
the initial suppression of the transient is independent of any detection or
actuation signal. That is, the passive accumulators begin the rapid
reflooding of the core. Complete reflood of the core is dependent upon the
pumped water ejected from the ECCS pumps initiated via the SI actuation
signal.
All pumps used for safety injection, emergency containment cooling and
containment spray and associated instrumentation are located outside the
containment. The emergency containment cooler fans and associated
instrumentation are located inside containment and have been environmentally
qualified accordingly. The operation of the equipment can be verified by
instrumentation that reads in the control room.
Depending upon the magnitude of the loss-of-coolant incident, information
relative to the pressure of the Reactor Coolant System will be required to
determine which pumps will be used for recirculation. Wide range RCS
pressure instrumentation will be used to decide when the charging pumps can
be used, if available, for make-up water, such as for a relatively small
loss-of-coolant accident. Otherwise, the discharge pressure of the charging
pumps as read on instrumentation outside the containment, will be sufficient.
In conjunction with the available accumulator instrumentation, a full range
of system pressure can be determined.
7.5-6 Revised 04/17/2013 C26 Core recirculation and containment spray recirculation (if necessary) will be
manually accomplished when the containment level provides sufficient NPSH and
the refueling water storage tank reaches the low-low alarm setpoint.
Considerations have been given to all the instrumentation and information
that will be necessary for the recovery time following a loss-of-coolant
accident. Instrumentation external to the containment; such as radioactivity
monitoring
equipment, will not be affected by this postulated incident and will be
available to the operator.
7.5.3 SYSTEM EVALUATION
Redundant instrumentation has been provided for all inputs to the protective
systems and vital control circuits.
Where wide process variable ranges and precise control are required, both
wide range and narrow range instrumentation is provided.
Instrumentation components were originally selected from standard
commercially available products with proven operating reliability.
Replacements are upgraded (whenever practical) with nuclear grade components
when required.
All electrical and electronic instrumentation required for safe and reliable
operation is supplied from the vital instrumentation buses.
Pressurizer Pressure
Low pressurizer pressure provides primary input for the actuation of
emergency core cooling. Two-out-of-three logic will prevent false actuation
of the SIS in the event of a spurious pressure signal. Figure 7.2-8e
provides additional details concerning the initiation of safety injection.
A safety injection block switch is provided to permit the Reactor Coolant
System to be depressurized for maintenance and refueling operations without
actuation of the Safety Injection System.
7.5-7 Revised 04/17/2013 This manual block switch will be interlocked with pressurizer pressure in such a way that the blocking action will automatically be removed above a
preset pressure as operating pressure is approached. If two-out-of-three
pressure signals are above this preset pressure, blocking action cannot be
initiated. The block condition will be indicated by status lights on
vertical panel A.
Steam Generator Level Control During Unit Cooldown
The successful operation of the engineered safety features involves only
actuation functions, with one exception. This exception is the steam
generator level control function associated with cooldown using the auxiliary
feedwater pumps. This level control system involves remote manual
positioning of feedwater flow and auxiliary feedwater control valves in order
to maintain proper steam generator water level. Steam generator water level
indication and controls are located in the control room and locally.
Environmental Capability
The components of the ESFAS are designed and laid out so that adverse
environments accompanying an emergency situation, in which components are
required to function, do not interfere with that function. Refer to Appendix
8A for additional information pertaining to Environmental Qualification.
7.5.4 REGULATORY GUIDE 1.97, REVISION 3
A review of Turkey Point Units 3 and 4 instrumentation was conducted against
Regulatory Guide 1.97, Revision 3, "Instrumentation for Light-Water Cooled
Nuclear Power Plants to Assess Plant and Environs Conditions During and
Following an Accident." The references listed in Section 7.5.4.4 identify
related correspondence between FPL and NRC.
7.5.4.1 REGULATORY GUIDE 1.97 (REVISION 3) REQUIREMENTS
Regulatory Guide 1.97, Revision 3, divides all instrumentation used for Post Accident Monitoring into five functional types as defined below:
Type A Variables:
Those variables to be monitored that provide the primary information required to permit the control room operator to take specific
manually controlled actions for which no automatic control is provided and
that are required for safety systems to accomplish their safety function for
design basis accident events.
7.5-8 Revised 04/17/2013 Primary information is information that is essential for the direct accomplishment of the specified safety functions; it does not include those
variables that are associated with contingency actions that may also be
identified in written procedures.
Type B Variables:
Those variables that provide information to indicate whether plant safety functions are being accomplished. Plant safety
functions are (1) reactivity control, (2) core cooling, (3) maintaining
reactor coolant system integrity, and (4) maintaining containment integrity (including radioactive effluent control).
Type C Variables:
Those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission
product releases. The barriers are (1) fuel cladding, (2) primary coolant
pressure boundary, and (3) containment.
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.
Type E Variables
- Those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and
continually assessing such releases.
Table 1 of Regulatory Guide 1.97, Revision 3 provides design and
qualification criteria for Post Accident Monitoring Instrumentation used to
measure the various variables identified in Table 3 (for PWRs). The criteria
are divided into three categories depending on the importance to safety of
the specific variable.
In general, Category 1 provides for full qualification, redundancy, and
continuous realtime display and requires onsite (standby) power. Category 2
provides for qualification but is less stringent in that it does not (of
itself) include seismic qualification, redundancy, or continuous display and
requires only a high-reliability power source (not necessarily standby
power). Category 3 is the least stringent. It provides for high-quality
commercial grade equipment that requires only offsite power.
7.5-9 Revised 04/17/2013 7.5.4.2 EVALUATION CRITERIA
The Regulatory Guide 1.97, Revision 3, requirements cover the requirements of
10 CFR 50.49, NUREG-0737, and Generic Letter 82-33.
The following is the evaluation criteria used to develop the parameter
listing summary sheets presented in Tables 7.5-1 and 7.5-2. The information
provided in these tables was developed in response to Regulatory Guide 1.97, Revision 3, and Generic Letter 82-33.
7.5.4.2.1 ENVIRONMENTAL QUALIFICATION CRITERIA
- 1. Category 1 Instrumentation
- a. Instrumentation located in harsh environments should comply with the requirements of 10 CFR 50.49. An entry of "Comply" in the
column headed "Environ Qual" of the parameter listing summary
sheets indicates that the instrumentation is required to meet the
requirements of 10 CFR 50.49 and is included in the Turkey Point
Environmental Qualification Program.
- b. Instrumentation located in mild environments are not required to be environmentally qualified. This is denoted in the parameter
listing summary sheets by an entry of "N/A"in the column headed "Environ Qual."
- 2. Category 2 Instrumentation
For Category 2 instrumentation the same criteria are used as for Category 1 instrumentation.
- 3. Category 3 Instrumentation
Environmental qualification of Category 3 instrumentation is not required. This is denoted in the parameter listing summary sheets by
an entry of "N/A" in the column headed "Environ Qual."
7.5-10 Revised 04/17/2013 7.5.4.2.2 SEISMIC QUALIFICATION CRITERIA
- 1. Category 1 Instrumentation
The original plant licensing basis for Turkey Point did not include any commitment to seismically qualify plant equipment to the requirements
of IEEE Standard 344-1975 (Regulatory Guide 1.100). "Original
equipment" complies with the seismic qualification approach which was
the basis for plant licensing. Original mechanical and electrical
equipment was purchased under specifications that included a
description of the seismic design criteria for the plant. However, no
seismic specifications were employed in the original instrumentation
purchase orders. Type testing documented in Westinghouse's WCAP 7397-1
provides verification of the seismic design objective for
instrumentation.
For new/replacement instrumentation, qualification to IEEE Standard 344-1975 is implemented whenever practicable. As a minimum, such
equipment must meet the seismic criteria for original equipment.
Based on the above, the entry of "Comply" in the column headed "Seismic Qual" associated with the parameter listing summary sheets is only
intended to denote that seismic qualification is required.
- 2. Category 2 and 3 Instrumentation Regulatory Guide 1.97, Revision 3, does not identify any specific provisions for Category 2 and 3 instrumentation. Therefore, an entry of "N/A" in the column headed "Seismic Qual" of the parameter listing
summary sheets indicates that there is no requirement for seismic
qualification.
7.5-11 Revised 04/17/2013 7.5.4.2.3 REDUNDANCE
- 1. Category 1 Instrumentation
Table 1, Design and Qualification Criteria for Instrumentation of Regulatory Guide 1.97, Revision 3 identifies specific provisions for
redundancy including physical independence of instrument channels in
accordance with Regulatory Guide 1.75. Turkey Point's licensing basis
does not include any commitment to the requirements of Regulatory Guide
1.75. However, electrical and physical separation of Post-Accident
Monitoring circuits is provided.
For category 1 variables, separation of redundant channels is used, to the maximum practical extent, beginning at the process sensors.
Separation of redundant channels of field wiring continues through
containment penetrations to the analog protection racks. Physical
separation of field wiring for category 1 variables is achieved using
separate raceway and containment penetrations for each redundant
channel. Such separation ensures that physical damage affecting one
channel will not affect its redundant channel. Isolation devices are
provided for SPDS/ERDADS/DCS computer interfaces in accordance with
The entry in the column headed "Redundance" of the parameter listing summary sheets identifies the redundant component(s) credited for
compliance to the regulatory guide. In most instances, where
components under the column headed "Tag No." are credited for recording
function, an entry of "N/A" has been used to indicate that redundance
is not required by the regulatory guide. "N/A" has also been entered
in the "Redundance" column associated with Containment Isolation Valve
Position Indication (i.e., item no. B15). This denotes that redundant
valve position indication is not required. However, an evaluation has
been performed to demonstrate the capability of the Control Room
operator to verify isolation of containment penetrations.
- 2. Category 2 and 3 Instrumentation
For category 2 and 3 instrumentation, Regulatory Guide 1.97, Rev. 3 identifies no specific provision for redundancy. An entry of "N/A" in
the column headed "Redundance" associated with the parameter listing
summary sheets denotes that redundance is not required.
7.5-12 Revised 04/17/2013 C26 7.5.4.2.4 POWER SOURCES
- 1. Category 1 Instrumentation
Category 1 instrumentation should be powered from one of the following Class 1E power sources:
- a. 120 VAC uninterruptable power supply (inverters)
- b. 120 VAC power backed up by the Emergency Diesel Generators
- c. 125 VDC safety-related batteries
The column headed "Power Supply" associated with the parameter listing summary sheets contains a reference to a note identifying which of the
above power sources is used to power the main instrument loop providing
Post-Accident Monitoring.
- 2. Category 2 Instrumentation
Category 2 instrumentation should be supplied from a high reliability power source which can be either from:
- a. 120 VAC safety-related or nonsafety-related uninterruptable power supply, or b. 120 VAC power backed up by an Emergency Diesel Generator, or
- c. 125 VDC safety-related battery, or
- d. 125 VDC nonsafety-related battery
The column headed "Power Supply" associated with the parameter listing summary sheets contains a reference to a note identifying which of the
above power sources is used to supply the main instrument loop
providing Post-Accident Monitoring.
- 3. Category 3 Instrumentation
Regulatory Guide 1.97, Revision 3, does not identify any specific provisions for Category 3 instrumentation. Therefore, an entry of "N/A" in the column headed "Power Supply" of the parameter listing
summary sheets indicates that there are no power supply requirements.
7.5-13 Revised 04/17/2013 7.5.4.2.5 DISPLAY AND RECORDING
- 1. Category 1 Instrumentation
Category 1 instrumentation should be displayed on a real-time display.
The indicator may be on a dial, digital display, electronic display or
strip chart recorder.
Recording of instrumentation readout should be provided for at least one redundant channel. Where dedicated strip chart recorder is not
provided, recording should be updated and sorted in computer memory and
displayed on demand.
- 2. Category 2 Instrumentation
Category 2 instrumentation should be displayed on an individual instrument or it may be processed for display on demand. Signal from
effluent radioactivity and area monitors should be recorded.
- 3. Category 3 Instrumentation
For Category 3 Instrumentation the same criteria are used as for Category 2 instrumentation except that signals from effluent
radioactivity, area and meteorology monitors should be recorded.
In general, instrumentation located on Main Control Room Boards and panels
are credited for Post-Accident Monitoring and are identified on the Parameter
Listing Summary Sheets in the column headed "Tag No." However, for various
Regulatory Guide 1.97 variables, the SPDS/ERDADS/DCS has been credited for
indication and/or recording capability. In these instances, SPDS or ERDADS
is identified on the parameter listing summary sheets in the column headed "CR (Control Room) Display Location." SPDS or ERDADS is indicated only when
computer capability is required in order to comply with the minimum
requirements of Regulatory Guide 1.97.
7.5-14 Revised 04/17/2013 C26C26 7.5.4.2.6 RANGE
Control Room instrumentation should meet the range specified for the variable
in Regulatory Guide 1.97. In general, if two or more instruments are
required to cover a particular range, overlapping of instrument span shall be
provided. The parameter listing summary sheets in the column headed "Existing
Instrument Range." identifies the instrument range associated with a variable
as provided by Control Room indication. The column headed "Required
Instrument Range" identifies the range required by Table 3 "PWR Variables" of
Regulatory Guide 1.97 Revision 3.
7.5.4.3 TYPE A VARIABLES
Regulatory Guide 1.97, Revision 3, states that Type A variables are plant
specific. In order to identify the Type A variables which are specific to a
plant, a review must be performed of those EOP's which are pertinent to a
design basis accident event (i.e., anticipated operational occurrences or
serious events outside the design basis are not considered). A review was
conducted against the EOPs and the following parameters were designated as
Type A variables:
- 1. RCS Pressure
- 2. RCS Hot Leg Temperature
- 3. RCS Cold Leg Temperature
- 4. Steam Generator Level Narrow Range
- 5. Refueling Water Storage Tank Level
- 6. Pressurizer Level
- 7. Core Exit Temperature
- 8. Steam Generator Pressure
- 9. Containment Sump Water Level Wide Range
- 10. Safety Injection Pump Status
- 11. EDG Output (KW)
- 12. 4KV Bus Voltage
The assumptions, methodology and results of the EOP review are documented in
Reference 4.
7.5-15 Revised 04/17/2013 7.5.
4.4 REFERENCES
- 2. Letter L-85-176A, J W Williams, Jr (FPL) to S A Varga (NRC), dated May 10, 1985.
- 3. Letter, "Instrumentation to Follow the Course of an Accident -
Conformance to Regulatory Guide 1.97, Revision 3," D G McDonald (NRC) to
C O Woody (FPL), dated March 20, 1986.
- 4. Regulatory Guide 1.97 Emergency Operating Procedure dated September 1990 prepared by Engineering Planning and Management, Inc. (EPM) (Attachment 5 of PC/M 90-391).
- 5. EPM Letter No. EL06090-158, dated October 1, 1990, "Supporting Documentation Associated with the FSAR Revision of Regulatory Guide 1.97
Commitments."
- 7. NRC Letter to FPL, dated April 13, 1992, Docket Nos. 50-250 and 50-251.
This letter provides justification to downgrade the SI accumulator
instrumentation from RG 1.97 Category 2 to Category 3.
- 8. Letter L-2011-046, "License Amendment Request No. 214, Accident Monitoring Instrumentation Technical Specification Changes Regarding High Range - Noble Gas Effluent Monitors - Main steam Lines Accident Monitoring Instrumentation," August 17, 2011.
7.5-16 Revised 04/17/2013
C26 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 1 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION A1 RCS PRESSURE PT-404 RCS PRESSUREA10-3000 PSIG PLANT SPECIFIC COMPL YCOMPLPT-406NOTE 1*YESYESNOTES 1A,1B PT-406 RCS PRESSUREA10-3000 PSIG PLANT SPECIFIC COMPL YCOMPLPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY AA10-3000 PSIG PLANT SPECIFICN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-3000 PSIG PLANT SPECIFICN/ACOMPLQSPDS ANOTE 1YES------
A2 RCS HOT LEG WATER TEMPERATURE TE-413A RCS HOT LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-413BNOTE 1---YESYES TE-413B RCS HOT LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-413ANOTE 1---YESYES TE-423A RCS HOT LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-423BNOTE 1---YESYES TE-423B RCS HOT LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-423ANOTE 1---YESYES TE-433A RCS HOT LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-433BNOTE 1---YESYES TE-433B RCS HOT LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-433ANOTE 1---YESYES TR-413RCS HOT LEG TEMP. RECORD LOOP A,B,C FOR TR. AA10-750 F PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------QSPDS ADISPLAY AA10-750 F PLANT SPECIFICN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-750 F PLANT SPECIFICN/ACOMPLQSPDS ANOTE 1YES------
A3 RCS COLD LEG WATER TEMPERATURE TE-410A RCS COLD LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-410BNOTE 1---YESYES TE-410B RCS COLD LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-410ANOTE 1---YESYES TE-420A RCS COLD LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-420BNOTE 1---YESYES TE-420B RCS COLD LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-420ANOTE 1---YESYES TE-430A RCS COLD LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-430BNOTE 1---YESYES TE-430B RCS COLD LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFIC COMPL YCOMPLTE-430ANOTE 1---YESYES TR-410RCS COLD LEG TEMP. RECORD LOOP A,B,C FOR TR. AA10-750 F PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------QSPDS ADISPLAY AA10-750 F PLANT SPECIFICN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-750 F PLANT SPECIFICN/ACOMPLQSPDS ANOTE 1YES------
A4 RWST LEVELLT-6583ARWST CH. A LEVELA10-335,000 GALPLANT SPECIFICN/ACOMPLLT-6583BNOTE 1**YESYESNOTES 1A,1BLI-6583ARWST CH. A LEVEL IND.A10-335,000 GALPLANT SPECIFICN/ACOMPLLI-6583BNOTE 1YES------
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 2 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLT-6583BRWST CH. B LEVELA10-335,000 GALPLANT SPECIFICN/ACOMPLLT-6583ANOTE 1**YESYESNOTES 1A,1BLI-6583BRWST CH. B LEVEL IND.A10-335,000 GALPLANT SPECIFICN/ACOMPLLI-6583ANOTE 1YES------
A5 S. G. LEVEL NARROW RANGE LT-474 S.G. `A' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-475;LT-476NOTE 1---YESYESNOTE 1LLI-474S.G. `A' LVL. CH. I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-475;LI-476NOTE 1YES------NOTE 1LLT-475S.G. `A' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-474;LT-476NOTE 1---YESYESNOTE 1LLI-475S.G. `A' LVL. CH. II NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-474;LI-476NOTE 1YES------NOTE 1L LT-476S.G. `A' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-474;LT-475NOTE 1---YESYESNOTE 1LLI-476S.G. `A' LVL. CH. III NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-474;LI-475NOTE 1YES------NOTE 1L FR-478S.G. `A' LVL. CH. I, II, III NARROW RANGE RECORDERA10-100%PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------NOTE 1L LT-484S.G. `B' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-485;LT-486NOTE 1---YESYESNOTE 1LLI-484S.G. `B' LVL. CH.I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-485;LI-486NOTE 1YES------NOTE 1L LT-485S.G. `B' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-484;LT-486NOTE 1---YESYESNOTE 1LLI-485S.G. `B' LVL. CH. II NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-484;LI-486NOTE 1YES------NOTE 1L LT-486S.G. `B' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-484;LT-485NOTE 1---YESYESNOTE 1LLI-486S.G. `B' LVL. CH. III NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-484;LI-485NOTE 1YES------NOTE 1L FR-488S.G. `B' LVL. CH. I, II, III NARROW RANGE RECORDERA10-100%PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------NOTE 1L LT-494S.G. `C' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-495;LT-496NOTE 1---YESYESNOTE 1LLI-494S.G. `C' LVL. CH. I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-495;LI-496NOTE 1YES------NOTE 1L LT-495S.G. `C' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-494;LT-496NOTE 1---YESYESNOTE 1LLI-495S.G. `C' LVL. CH. II NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-494;LI-496NOTE 1YES------NOTE 1L LT-496S.G. `C' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.2 2 PLANT SPECIFIC COMPL YCOMPLLT-494;LT-495NOTE 1---YESYESNOTE 1LLI-496S.G. `C' LVL. CH. III NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLLI-494;LI-495NOTE 1YES------NOTE 1LFR-498S.G. `C' LVL. CH. I, II, III NARROW RANGE RECORDERA10-100%PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------NOTE 1L A6 REACTIVITY CONTROL - NEUTRON FLUXND-6649ANEUTRON FLUX DETECTORA11E-8 TO 200% FULL P O PLANT SPECIFIC COMPL YCOMPLND-6649BNOTE 1**YESYESNOTES 1A,1B C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 3 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONNI-6649A-2NEUTRON FLUX INDICATORA11E-8 TO 200% FULL P O PLANT SPECIFICN/ACOMPLNI-6649B-2NOTE 1YES------ND-6649BNEUTRON FLUX DETECTORA11E-8 TO 200% FULL P O PLANT SPECIFIC COMPL YCOMPLND-6649ANOTE 1**YESYESNOTES 1A,1BNI-6649B-2NEUTRON FLUX INDICATORA11E-8 TO 200% FULL P O PLANT SPECIFICN/ACOMPLNI-6649A-2NOTE 1YES------
A7 CORE EXIT TEMPERATURETE-1E THRU TE-51CORE EXIT TEMPERATUREA132-2300 F PLANT SPECIFICN/ACOMPL2 CHANNEL PER QNOTE 1*YESYESNOTES 1A,1B,QSPDS ADISPLAY 'A'A132-2300 FPLANT SPECIFICN/ACOMPLQSPDS BNOTE 1YES------& 9QSPDS BDISPLAY 'B'A132-2300 FPLANT SPECIFICN/ACOMPLQSPDS ANOTE 1YES------
A8 CONTAINMENT SUMP WATER LEVELLT-6309ACTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECCOMPL YCOMPLLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WATER LEVEL IND.A1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLLI-6309BNOTE 1YES------NOTES 1C,1K LR-6308ACTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLN/ANOTE 1YES------NOTE 1C LT-6309BCTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECCOMPL YCOMPLLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WATER LEVEL IND.A1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLLI-6309ANOTE 1YES------NOTES 1C,1K LR-6308BCTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLN/ANOTE 1YES------NOTE 1C A9 PRESSURIZER WATER LEVEL LT-459 PRZR LEVEL CH. IA10-100%(150" TO 334")PLANT SPECIFIC COMPL YCOMPLLT-460;LT-461NOTE 1---YESYESLI-459APRZR LEVEL CH. I IND.A10-100%PLANT SPECIFICN/ACOMPLLI-460;LI-461NOTE 1YES------
LT-460PRZR LEVEL CH. IIA10-100%(150" TO 334")PLANT SPECIFICCOMPL YCOMPLLT-459;LT-461NOTE 1---YESYESLI-460PRZR LEVEL CH. II IND.A10-100%PLANT SPECIFICN/ACOMPLLI-459A;LI-461NOTE 1YES------
LT-461PRZR LEVEL CH. IIIA10-100%(150" TO 334")PLANT SPECIFICCOMPL YCOMPLLT-459;LT-460NOTE 1---YESYESLI-461PRZR LEVEL CH. III IND.A10-100%PLANT SPECIFICN/ACOMPLLI-459A;LI-460NOTE 1YES------
LR-459PRZR LEVEL RECORDER FOR LT-459, 460, 461A10-100%PLANT SPECIFICN/ACOMPLN/ANOTE 1YES------
A10 STEAM GENERATOR PRESSURE PT-474 S.G. 'A' STEAM PRESSURE CH. IIA10-1400 PSIG PLANT SPECIFIC COMPL YCOMPLPT-475;PT-476NOTE 1*YESYESNOTES 1A,1BPI-474S.G. 'A' STEAM PRESSURE CH. II IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-475;PI-476NOTE 1YES------
PT-475S.G. 'A' STEAM PRESSURE CH. IIIA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-474;PT-476NOTE 1*YESYESNOTES 1A,1BPI-475S.G. 'A' STEAM PRESSURE CH. III IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-474;PI-476NOTE 1YES------
PT-476S.G. 'A' STEAM PRESSURE CH. IVA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-474;PT-475NOTE 1*YESYESNOTES 1A,1B C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 4 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION PI-476 S.G. 'A' STEAM PRESSURE CH. IV IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLPI-474;PI-475NOTE 1YES------
PT-484 S.G. 'B' STEAM PRESSURE CH. IIA10-1400 PSIG PLANT SPECIFIC COMPL YCOMPLPT-485;PT-486NOTE 1*YESYESNOTES 1A,1BPI-484S.G. 'B' STEAM PRESSURE CH. II IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-485;PI-486NOTE 1YES------PT-485S.G. 'B' STEAM PRESSURE CH. IIIA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-484;PT-486NOTE 1*YESYESNOTES 1A,1BPI-485S.G. 'B' STEAM PRESSURE CH. III IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-484;PI-486NOTE 1YES------
PT-486S.G. 'B' STEAM PRESSURE CH. IVA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-484;PT-485NOTE 1*YESYESNOTES 1A,1BPI-486S.G. 'B' STEAM PRESSURE CH. IV IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-484;PI-485NOTE 1YES------
PT-494S.G. 'C' STEAM PRESSURE CH. IIA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-495;PT-496NOTE 1*YESYESNOTES 1A,1BPI-494S.G. 'C' STEAM PRESSURE CH. II IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-495;PI-496NOTE 1YES------
PT-495S.G. 'C' STEAM PRESSURE CH. IIIA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-494;PT-496NOTE 1*YESYESNOTES 1A,1BPI-495S.G. 'C' STEAM PRESSURE CH. III IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-494;PI-496NOTE 1YES------
PT-496S.G. 'C' STEAM PRESSURE CH. IVA10-1400 PSIGPLANT SPECIFICCOMPL YCOMPLPT-494;PT-495NOTE 1*YESYESNOTES 1A,1BPI-496S.G. 'C' STEAM PRESSURE CH. IV IND.A10-1400 PSIGPLANT SPECIFICN/ACOMPLPI-494;PI-495NOTE 1YES------
A11 EDG OUTPUT 3K4A EMERGENCY DIESEL GENERATOR `3A' OUTPUTA10-4 MEGAWATTSPLANT SPECIFICN/ACOMPLEDG `3B' OUTPUTNOTE 5NOTE 1YESYESNOTES 1A,1B 3K4B EMERGENCY DIESEL GENERATOR `3B' OUTPUTA10-4 MEGAWATTSPLANT SPECIFICN/ACOMPLEDG `3A' OUTPUTNOTE 5NOTE 1YESYESNOTES 1A,1B A12 4KV BUS VOLTAGE 3AA`3A' 4KV BUS VOLTAGEA10-5000 VOLTSPLANT SPECIFICN/ACOMPL3B 4KV BUS VOLT ANOTE 5YES------NOTE 63AB`3B' 4KV BUS VOLTAGEA10-5000 VOLTSPLANT SPECIFICN/ACOMPL3A 4KV BUS VOLT ANOTE 5YES------NOTE 6 A13 SAFETY INJECTION PUMP STATUS 3AA13 SI PUMP '3A' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPL3AB12;4AA13;4AB1NOTE 3**YESYES SI PP. 3A IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLPP3B,4A/B IND LG HNOTE 3YES------3AB12SI PUMP '3B' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPL3AA13;4AA13;4AB1NOTE 3**YESYESSI PP. 3B IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLPP3A,4A/B IND LG HNOTE 3YES------4AA13SI PUMP '4A' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPL3AA13;3AB12;4AB1NOTE 3**YESYESSI PP. 4A IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLPP3A/B,4B IND LG HNOTE 3YES------4AB12SI PUMP '4B' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPL3AA13;3AB12;4AA1NOTE 3**YESYESSI PP. 4B IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLPP3A/B,4A IND LG HNOTE 3YES------
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 5 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION B1 REACTIVITY CONTROL - NEUTON FLUXND-6649ANEUTRON FLUX DETECTORB11E-8 TO 200% FULL P O 1E-6 TO 100% FULL POWE R COMPL YCOMPLND-6649BNOTE 1**YESYESNOTES 1A,1BNI-6649A-2NEUTRON FLUX INDICATORB11E-8 TO 200% FULL P O 1E-6 TO 100% FULL POWE RN/ACOMPLNI-6649B-2NOTE 1YES------ND-6649BNEUTRON FLUX DETECTORB11E-8 TO 200% FULL P O 1E-6 TO 100% FULL POWE R COMPL YCOMPLND-6649ANOTE 1**YESYESNOTES 1A,1BNI-6649B-2NEUTRON FLUX INDICATORB11E-8 TO 200% FULL P O 1E-6 TO 100% FULL POWE RN/ACOMPLNI-6649A-2NOTE 1YES------
B2 REACTIVITY CONTROL - CONTROL ROD POSITION G5 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E9 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J11 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L7 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J5 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E7 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES G11 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L9 CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F2 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES B10 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K14 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES P6 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K2 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES B6 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F14 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES P10 CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F4 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES D10 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 6 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION K12 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES M6 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K4 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES D6 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F12 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES M10 CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES D8 CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES M8 CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H4 CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H8 CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H12 CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES G3 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C9 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J13 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES N7 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J3 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C7 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES G13 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES N9 SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E5 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L11 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L5 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E11 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H6 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H10 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C26 01/27/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 7 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION F8 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K8 SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL FULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES B3 REACTIVITY CONROL - RCS BORON CONCENTRATION AE-6424 BORON ANALYZER RCS SOLUBLE BORON CONCENTRATIO NB30-6000 PPM 0-6000 PPMN/AN/AN/AN/A**YESYESNOTE 1B B4 REACTIVITY CONTROL - RCS COLD LEG WTR. TEMP.
TE-410A RCS COLD LEG WTR. TEMP. LOOP `A'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TE-410B RCS COLD LEG WTR. TEMP. LOOP `A'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TE-420A RCS COLD LEG WTR. TEMP. LOOP `B'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TE-420B RCS COLD LEG WTR. TEMP. LOOP `B'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TE-430A RCS COLD LEG WTR. TEMP. LOOP `C'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TE-430B RCS COLD LEG WTR. TEMP. LOOP `C'B30-750 F50-400 FN/AN/AN/AN/A---YESYES TR-410RCS COLD LEG TEMP. RECORD LOOP A, B, C FOR TR. AB30-750 F50-400 FN/AN/AN/AN/AYES------QSPDS ADISPLAY `A'B30-750 F50-400 FN/AN/AN/AN/AYES------QSPDS BDISPLAY `B'B30-750 F50-400 FN/AN/AN/AN/AYES------
B5 CORE COOLING - RCS HOT LEG WTR. TEMP.
TE-413A RCS HOT LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 F COMPL YCOMPLTE-413BNOTE 1---YESYES TE-413B RCS HOT LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 F COMPL YCOMPLTE-413ANOTE 1---YESYES TE-423A RCS HOT LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 F COMPL YCOMPLTE-423BNOTE 1---YESYES TE-423B RCS HOT LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 F COMPL YCOMPLTE-423ANOTE 1---YESYES TE-433A RCS HOT LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 F COMPL YCOMPLTE-433BNOTE 1---YESYES TE-433B RCS HOT LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 F COMPL YCOMPLTE-433ANOTE 1---YESYES TR-413RCS HOT LEG TEMP. RECORD LOOP A, B, C FOR TR. AB10-750 F 50-700 FN/ACOMPLN/ANOTE 1YES------QSPDS ADISPLAY `A'B10-750 F 50-700 FN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-750 F 50-700 FN/ACOMPLQSPDS ANOTE 1YES------
B6 CORE COOLING - RCS COLD LEG WTR. TEMP.
TE-410A RCS COLD LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 F COMPL YCOMPLTE-410BNOTE 1---YESYES TE-410B RCS COLD LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 F COMPL YCOMPLTE-410ANOTE 1---YESYES TE-420A RCS COLD LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 F COMPL YCOMPLTE-420BNOTE 1---YESYES C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 8 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION TE-420B RCS COLD LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 F COMPL YCOMPLTE-420ANOTE 1---YESYES TE-430A RCS COLD LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 F COMPL YCOMPLTE-430BNOTE 1---YESYES TE-430B RCS COLD LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 F COMPL YCOMPLTE-430ANOTE 1---YESYES TR-410RCS COLD LEG TEMP. RECORD LOOP A, B, C FOR TR. AB10-750 F 50-700 FN/ACOMPLN/ANOTE 1YES------QSPDS ADISPLAY `A'B10-750 F 50-700 FN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-750 F 50-700 FN/ACOMPLQSPDS ANOTE 1YES------
B7 CORE COOLING - RCS PRESSURE PT-404 RCS PRESSUREB10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-406NOTE 1*YESYESNOTES 1A,1B PT-406 RCS PRESSUREB10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS ANOTE 1YES------
B8 CORE COOLING - CORE EXIT TEMPERATURETE-1E THRU TE-51CORE EXIT TEMPERATUREB332-2300 F 200-2300 FN/AN/AN/AN/A---YESYESNOTE 9QSPDS ADISPLAY `A'B332-2300 F 200-2300 FN/AN/AN/AN/AYES------QSPDS BDISPLAY `B'B332-2300 F 200-2300 FN/AN/AN/AN/AYES------
B9 CORE COOLING - COOLANT INVENTORYICCS RVL-A (HJTCREACTOR VESSEL WTR. LVL. CH. `A'B10-100% (CORE TOP/V S BTM HOT LEG TO TOP OF VN/ACOMPLRVL-BNOTE 1*YESYESNOTES 1A,1BICCS RVL-B (HJTCREACTOR VESSEL WTR. LVL. CH. `B'B10-100% (CORE TOP/V S BTM HOT LEG TO TOP OF VN/ACOMPLRVL-ANOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-100% (CORE TOP/V S BTM HOT LEG TO TOP OF VN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-100% (CORE TOP/V S BTM HOT LEG TO TOP OF VN/ACOMPLQSPDS ANOTE 1YES------
B10 CORE COOLING - DEGREES OF SUBCOOLINGQSPDS ARCS TEMP. SATURATION MARGIN CH. `A'B2700 TO -2100 F200 SUBCLNG TO 35 SUPR HN/AN/AN/ANOTE 1YESYESYESQSPDS BRCS TEMP. SATURATION MARGIN CH. `B'B2700 TO -2100 F200 SUBCLNG TO 35 SUPR HN/AN/AN/ANOTE 1YESYESYES B11 MAINTAINING RCS INTEGRITY - RCS PRESSURE PT-404 RCS PRESSUREB10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-406NOTE 1*YESYESNOTES 1A,1B PT-406 RCS PRESSUREB10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS BNOTE 1YES------
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 9 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONQSPDS BDISPLAY `B'B10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS ANOTE 1YES------
B12 MAINTAINING RCS INTEGRITY - CTMT. SUMP WTR. LEVELLT-6308ACTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)COMPL YN/AN/ANOTE 1---YESYESNOTE 1CLI-6308ACTMT. SUMP WTR. LVL. IND.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308ACTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1JLT-6308BCTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)COMPL YN/AN/ANOTE 1---YESYESNOTE 1CLI-6308BCTMT. SUMP WTR. LVL. IND.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308BCTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1J B13 MAINTAINING RCS INTEGRITY - CTMT. SUMP WTR. LEVELLT-6309ACTMT. WTR. LVL.B1397"-487" WIDE RANGE (PLANT SPE C COMPL YCOMPLLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WTR. LVL. IND.B1397"-487"WIDE RANGE (PLANT SPE CN/ACOMPLLI-6309BNOTE 1YES------NOTES 1C,1KLR-6308ACTMT. WTR. LVL.B1397"-487" WIDE RANGE (PLANT SPE CN/ACOMPLN/ANOTE 1YES------NOTE 1CLT-6309BCTMT. WTR. LVL.B1397"-487" WIDE RANGE (PLANT SPE C COMPL YCOMPLLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WTR. LVL. IND.B1397"-487"WIDE RANGE (PLANT SPE CN/ACOMPLLI-6309ANOTE 1YES------NOTES 1C,1KLR-6308BCTMT. WTR. LVLB1397"-487" WIDE RANGE (PLANT SPE CN/ACOMPLN/ANOTE 1YES------NOTE 1C B14 MAINTAINING RCS INTEGRITY - CTMT. PRESSUREPT-6306ACTMT. PRESS. WIDE RANGEB10-180 PSIG 0 PSIG TO DESIGN PRESS UN/ACOMPLPT-6306BNOTE 1---YESYESPI-6306ACTMT. PRESS. WIDE RANGE IND.B10-180 PSIG0 PSIG TO DESIGN PRESS UN/ACOMPLPI-6306BNOTE 1YES------PR-6306ACTMT. PRESS. WIDE RANGEB10-180 PSIG O PSIG TO DESIGN PRESS UN/ACOMPLN/ANOTE 1YES------PT-6306BCTMT. PRESS. WIDE RANGEB10-180 PSIG 0 PSIG TO DESIGN PRESS UN/ACOMPLPT-6306ANOTE 1---YESYESPI-6306BCTMT. PRESS. WIDE RANGE IND.B10-180 PSIGO PSIG TO DESIGN PRESS UN/ACOMPLPI-6306ANOTE 1YES------PR-6306BCTMT. PRESS. WIDE RANGEB10-180 PSIG 0 PSIG TO DESIGN PRESS UN/ACOMPLN/ANOTE 1YES------
B15 MAINTAINING CTMT. INTEGRITY - CTMT. ISO VAVLE POS.MOV-744A (LS)RHR TO COLD LEG I.C.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2**YESYESNOTE1A,B,H,2 MOV-744A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-744B (LS)RHR TO COLD LEG I.C.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2**YESYESNOTE1A,B,H,2 MOV-744B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 10 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONMOV-716B (LS)RCP THERMAL BARRIER CCWB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-716B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-626 (LS)RCP A, B, C THERMAL BARRIER COOLING WTR.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-626 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-730 (LS)CCW FROM RCP A, B, C COOLER BEARINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-730 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-739 (LS)EXCESS LETDOWN HEAT EXCHANGERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-739 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-1417 (LS)CCW TO NORMAL CTMT. COOLINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1417 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-1418 (LS)CCW FROM NORMAL CTMT. COOLINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1418 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-200A (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-200A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-200B (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-200B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-200C (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-200C IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-204 (LS)LETDOWN LINE LOW PRESS.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-204 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-381 (LS)RCP SEAL WTR. RETURN VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-381 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-4658A (LS)RCDT VENT VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 11 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION CV-4658A STATUS LIGHT ON PHASE A ISOL. BOARDB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-4658B (LS)RCDT VENT VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4658B STATUS LIGHT ON PHASE A ISOL. BOARDB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-4668A (LS)RCDT DISCH. TO HOLD-UP TANKB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4668A STATUS LIGHT ON PHASE A ISOL. BOARDB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-4668B (LS)RCDT DISCH. TO HOLD-UP TANKB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4668B STATUS LIGHT ON PHASE A ISOL. BOARDB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-4659A (LS)RCDT LINE TO HYDROGEN ANAL.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-4659A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-4659B (LS)RCDT LINE TO HYDROGEN ANAL.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-4659B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-880A (LS)CTMT. SPRAY PUMP A DISCH. VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-880A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-880B (LS)CTMT. SPRAY PUMP B DISCH. VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-880B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-956A (LS)PRZR STM. SPACE SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-956A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-956B (LS)PRZR LIQUID SPACE SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-956B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------SV-6428 (LS)HOT LEG RCS SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 SV-6428 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------SV-2912 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2,11 SV-2912 IND. LIGHTSB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------SV-2911 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2,11 SV-2911 IND. LIGHTSB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------
9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 12 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONSV-2913 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2,11 SV-2913 IND. LIGHTSB1CLOSED/NOT CLOSE DCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------CV-519A (LS)DEMIN. WTR. TO PRZR RELIEF TNK.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2CV-519A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2600 (LS)CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11POV-2600 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2601 (LS)CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2601 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2602 (LS)CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2602 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2603 (LS)CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2603 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2604 (LS)MAIN STM. MSIV S.G. `A'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2604 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-1403 (LS)MAIN STM. LINE `A'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 MOV-1403 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------POV-2605 (LS)MAIN STM. MSIV S.G. `B'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 POV-2605 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-1404 (LS)MAIN STM. LINE `B'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-1404 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------POV-2606 (LS)MAIN STM. MSIV S.G. `C'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 POV-2606 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-1405 (LS)MAIN STM. LINE `C'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 MOV-1405 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------FCV-478 (LS)S.G. `A' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 13 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION FCV-478 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------FCV-488 (LS)S.G. `B' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2 FCV-488 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------FCV-498 (LS)S.G. `C' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1---YESYESNOTES 1H,2 FCV-498 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------
CV-2816 AUX. FEEDWATER TO S.G. `A'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1401AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1E CV-2831 AUX. FEEDWATER TO S.G. `A'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1401BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1E CV-2817 AUX. FEEDWATER TO S.G. `B'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1457AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1E CV-2832 AUX. FEEDWATER TO S.G. `B'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1457BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1E CV-2818 AUX. FEEDWATER TO S.G. `C'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1458AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1E CV-2833 AUX. FEEDWATER TO S.G. `C'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2HIC-1458BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 1YES------NOTE 1ECV-6275A (LS)S.G. `A' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-6275A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-6275B (LS)S.G. `B' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-6275B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-6275C (LS)S.G. `C' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2 CV-6275C IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-1427 (LS)S.G. `A' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1427 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------
9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 14 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONMOV-1426 (LS)S.G. 'B' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1426 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-1425 (LS)S.G. `C' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1425 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2903 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COOLER BB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2903 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2904 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COOLER CB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2904 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2905 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COOLER AB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2905 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2810 (LS)CCW FROM EMERGENCY CTMT. COOLER B BYPASSB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2810 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2906 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. COOLER BB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2906 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2812 (LS)CCW FROM EMERGENCY CTMT. COOLER C BYPASSB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2812 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2907 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. COOLER CB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2907 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2814 (LS)CCW FROM EMERGENCY CTMT. COOLER A BYPASSB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2814 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2908 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. COOLER AB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 CV-2908 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-872 (LS)LOW HEAD SAFETY INJECT.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-872 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-855 (LS)NITROGEN SUPPLY TO ACCUMULATORSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 15 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION CV-855 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-956D (LS)ACCUMULATOR SAMPLE LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-956D IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-843A (LS)BORON INJ. TANK OUT STOP VALVEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-843A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------MOV-843B (LS)BORON INJ. TANK OUT STOP VALVEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-843B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-2821 (LS)CTMT. SUMP DISCH.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2821 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-2822 (LS)CTMT. SUMP DISCH.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2822 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-2819 (LS)INST. AIR BLEEDB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2819 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------CV-2826 (LS)INST. AIR BLEEDB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2826 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------MOV-6386 (LS)RCP SEALB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YCOMPLN/ANOTE 2---YESYESNOTES 1H,2 MOV-6386 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 2YES------CV-516 (LS)GAS ANALYZER SAMPLE VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2,11 CV-516 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------SV-6385 (LS)GAS ANALYZER SAMPLE VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3---YESYESNOTES 1H,2 SV-6385 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLN/ANOTE 3YES------
9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 16 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION B16 MAINTAINING CTMT. INTEGRITY - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESSUREB10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.B10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.B10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------PT-6306BCTMT. WIDE RANGE PRESS.B10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6306ANOTE 1---YESYESPI-6306BCTMT. WIDE RANGE PRESS. IND.B10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6306ANOTE 1YES------PR-6306BCTMT. WIDE RANGE PRESSB10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------PT-6425ACTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6425BNOTE 1---YESYESPI-6425ACTMT. NARROW RANGE PRESS. IND.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6425BNOTE 1YES------PR-6306ACTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------PT-6425BCTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6425ANOTE 1---YESYESPI-6425BCTMT. NARROW RANGE PRESS. IND.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6425ANOTE 1YES------PR-6306BCTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------
C1 FUEL CLADDING - CORE EXIT TEMPERATURETE-1E THRU TE-51CORE EXIT TEMPERATUREC132-2300 F 200-2300 FN/ACOMPL2 CHANNEL PER QNOTE 1*YESYESNOTES 1A,1B,QSPDS ADISPLAY `A'C132-2300 F200-2300 FN/ACOMPLQSPDS BNOTE 1YES------& 9QSPDS BDISPLAY `B'C132-2300 F200-2300 FN/ACOMPLQSPDS ANOTE 1YES------
C2 FUEL CLADDING - RADIOACTIVITY CONCENTRATION OR RADIATION LEVEL IN CIRCULATING PRIMARY COOLANT NONERADIOACTIVITY CONCENTRATION OR RADIATION LEVELC1GRAB SAMPLE1/2 TO 100 X T.S. LIMIT---------------------NOTE 3 C3 FUEL CLADDING - ANALYSIS OF PRIMARY COOLANT NONE Rx COOL WATER RADIOACTIVITY ANALYSISC3GRAB SAMPLE10 micro Ci/ml to 10 Ci/ml---------------------NOTE 12 C4 Rx COOLANT PRESSURE BOUNDARY - RCS PRESSURE PT-404 RCS PRESS.C10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-406NOTE 1*YESYESNOTES 1A,1B PT-406 RCS PRESS.C10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'C10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'C10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS ANOTE 1YES------
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 17 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION C5 Rx COOLANT PRESSURE BOUNDARY - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------
PT-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6306ANOTE 1---YESYES PI-6306BCTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6306ANOTE 1YES------
PR-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------
C6 Rx COOLANT PRESSURE BOUNDARY - CTMT. PRESSUREPT-6425ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6425BNOTE 1---YESYESPI-6425ACTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6425BNOTE 1YES------
PR-6306ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------
PT-6425BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPT-6425ANOTE 1---YESYES PI-6425BCTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLPI-6425ANOTE 1YES------
PR-6306BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSN/ACOMPLN/ANOTE 1YES------
C7 Rx COOLANT PRESSURE BOUNDARY - CTMT. SUMP WTR. LEVELLT-6308ACTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)COMPL YN/AN/ANOTE 1---YESYESNOTE 1CLI-6308ACTMT. SUMP WATER LEVEL IND.C25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1I LR-6308ACTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1J LT-6308BCTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)COMPL YN/AN/ANOTE 1---YESYESNOTE 1CLI-6308BCTMT. SUMP WATER LEVEL IND.C25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1I LR-6308BCTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1J C8 Rx COOLANT PRESSURE BOUNDARY - CTMT. SUMP WTR. LEVELLT-6309ACTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECCOMPL YCOMPLLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WATER LEVEL IND.C1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLLI-6309BNOTE 1YES------NOTES 1C,1K LR-6308ACTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLN/ANOTE 1YES------NOTE 1C LT-6309BCTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECCOMPL YCOMPLLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WATER LEVEL IND.C1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLLI-6309ANOTE 1YES------NOTES 1C,1K 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 18 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLR-6308BCTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECN/ACOMPLN/ANOTE 1YES------NOTE 1C C9 Rx COOLANT PRESSURE BOUNDARY - CTMT. AREA RADIATIONRAD-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A'C31 TO 1E8 R/Hr1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESRAI-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A' IND.C31 TO 1E8 R/Hr1 TO 1E4 R/HrN/AN/AN/AN/AYES------RAD-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B'C31 TO 1E8 R/Hr1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESRAI-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B' IND.C31 TO 1E8 R/Hr1 TO 1E4 R/HrN/AN/AN/AN/AYES------
C10 Rx COOLANT PRESS. BOUNDARY EFFLUENT RADIOACTIVITY - NOBLE GAS EFFLUENT FROM COND. AIR REMOVAL SYS. EXH.RAD-6417AIR EJECTOR CONDENSER EXHAUSTC31E-7 TO 1E5 micro Ci/
C1E-6 TO 1E-2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1B C11 CONTAINMENT - RCS PRESSURE PT-404 RCS PRESSUREC10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-406NOTE 1*YESYESNOTE 1B PT-406 RCS PRESSUREC10-3000 PSIG 0-3000 PSIG COMPL YCOMPLPT-404NOTE 1*YESYESNOTE 1BQSPDS ADISPLAY `A'C10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'C10-3000 PSIG 0-3000 PSIGN/ACOMPLQSPDS ANOTE 1YES------
C12 CONTAINMENT - CTMT. HYDROGEN CONCENTRATIONAE-6307ACTMT. HYDROGEN MONITORC30-20%0-10 VOL.%N/AN/AN/AN/A---YESYESAI-6307ACTMT. HYDROGEN INDICATORC30-20%0-10 VOL.%N/AN/AN/AN/AYES------RAR-6311ACTMT. HYDROGEN RECORDERC30-10%0-10 VOL.%N/AN/AN/AN/AYES------AE-6307BCTMT. HYDROGEN MONITORC30-20%0-10 VOL.%N/AN/AN/AN/A---YESYESAI-6307BCTMT. HYDROGEN INDICATORC30-20%0-10 VOL.%N/AN/AN/AN/AYES------RAR-6311BCTMT. HYDROGEN RECORDERC30-10%0-10 VOL.%N/AN/AN/AN/AYES------
C13 CONTAINMENT - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG
-5 PSIG TO 3X DESIGN PR EN/ACOMPLPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG
-5 PSIG TO 3X DESIGN PR EN/ACOMPLN/ANOTE 1YES------PT-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG
-5 PSIG TO 3X DESIGN PR EN/ACOMPLPT-6306ANOTE 1---YESYESPI-6306BCTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPI-6306ANOTE 1YES------PR-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG
-5 PSIG TO 3X DESIGN PR EN/ACOMPLN/ANOTE 1YES------PT-6425ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPT-6425BNOTE 1---YESYES C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 19 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONPI-6425ACTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPI-6425BNOTE 1YES------PR-6306ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLN/ANOTE 1YES------PT-6425BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPT-6425ANOTE 1---YESYESPI-6425BCTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLPI-6425ANOTE 1YES------PR-6306BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PR EN/ACOMPLN/ANOTE 1YES------
C14 CONTAINMENT EFFLUENT RADIOACTIVITY NOBLE GAS FROM IDENTIFIED RELEASE POINTSRAD-6304VENT STACK WIDE RANGE MONITORC21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E-2 micro Ci/CCN/AN/AN/A**YESYESNOTE 1BRAD-6417AIR EJECTOR CONDENSER EXH.C21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E-2 micro Ci/CCN/AN/AN/A**YESYESNOTE 1BRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONLY)C21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E-2 micro Ci/CCN/AN/AN/A**YESYESNOTE 1B C15 CONTAINMENT EFFLUENT RADIOACTIVITY NOBLE GAS (FROM BUILDINGS OR AREAS, ETC.)RAD-6304VENT STACK WIDE RANGE MONITORC21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E3 micro Ci/CCN/AN/AN/A**YESYESNOTE 1BRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONLY)C21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E3 micro Ci/CCN/AN/AN/A**YESYESNOTE 1B D1 RHR SYSTEM - RHR SYSTEM FLOW FT-605 RHR SYSTEM FLOWD20-8500 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYES FI-605 RHR SYSTEM FLOW INDICATORD20-8500 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D2 RHR SYSTEM - RHR Hx OUTLET TEMPERATURE TE-606 RHR Hx OUTLET TEMPERATURED250-400 F 40-350 F COMPL YN/AN/ANOTE 1---YESYESNOTE 3 TR-604 RHR Hx OUTLET TEMPERATURE RECORDERD250-400 F 40-350 FN/AN/AN/ANOTE 1YES------NOTE 3 D3 S.I.S. ACCUMULATOR TANK LEVEL LT-920 ACCUMULATOR TANK LEVEL `A'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 LI-920 ACCUMULATOR TANK LEVEL`A' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 LT-922 ACCUMULATOR TANK LEVEL `A'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 LI-922 ACCUMULATOR TANK LEVEL `A' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 LT-924 ACCUMULATOR TANK LEVEL `B'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 LI-924 ACCUMULATOR TANK LEVEL `B' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 LT-926 ACCUMULATOR TANK LEVEL `B'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 LI-926 ACCUMULATOR TANK LEVEL `B' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 LT-928 ACCUMULATOR TANK LEVEL `C'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 20 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION LI-928 ACCUMULATOR TANK LEVEL `C' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 LT-930 ACCUMULATOR TANK LEVEL `C'D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 LI-930 ACCUMULATOR TANK LEVEL `C' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-921 ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-921 ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-923 ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-923 ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-925 ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-925 ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-927 ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-927 ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-929 ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-929 ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-931 ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-931 ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 D4 S.I.S ACCUMULATOR TANK PRESSUREN/AN/AN/APT-921ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1---YESYESPI-921ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------
PT-923ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1---YESYES PI-923ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------
PT-925ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1---YESYES PI-925ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------
PT-927ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1---YESYES PI-927ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------
PT-929ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1---YESYES PI-929ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIGO-750 PSIGN/AN/AN/ANOTE 1YES------
9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 21 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION PT-931 ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1---YESYES PI-931 ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------
D5 S.I.S ACCUMULATOR ISOLATION VALVE POSITIONMOV-865A (LS)ACCUMULATOR TANK ISOLATION VALVE `A'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYES MOV-865A IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------MOV-865B (LS)ACCUMULATOR TANK ISOLATION VALVE 'B'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYES MOV-865B IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------MOV-865C (LS)ACCUMULATOR TANK ISOLATION VALVE `C'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYESMOV-865C IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------
D6 S.I.S. BORIC ACID CHARGING FLOW FT-943 BORIC ACID CHARGING FLOWD20-1000 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYESFI-943BORIC ACID CHARGING FLOW IND.D20-1000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D7 S.I.S. FLOW IN HPI SYSTEM FT-940 HPI SYSTEM FLOWD20-1000 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYESFI-940HPI SYSTEM FLOW IND.D20-1000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D8 S.I.S. FLOW IN LPI SYSTEM FT-605 LPI SYSTEM FLOWD20-8500 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYESFI-605LPI SYSTEM FLOW IND.D20-8500 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D9 S.I.S. REFUELING WATER STORAGE TANKLT-6583ARWST LEVELD20-335,000 GALTOP TO BOTTOMN/AN/AN/ANOTE 1---YESYESLI-6583ARWST LEVEL INDICATORD20-335,000 GALTOP TO BOTTOMN/AN/AN/ANOTE 1YES------LT-6583BRWST LEVELD20-335,000 GALTOP TO BOTTOMN/AN/AN/ANOTE 1---YESYES LI-6583BRWST LEVEL INDICATORD20-335,000 GALTOP TO BOTTOMN/AN/AN/ANOTE 1YES------
D10 PRIMARY COOLANT SYSTEM - RCP MOTOR STATUS 3P200A RCP `A' MTR. CURRENT INDICATORD30-1200 AMP MTR. CURRENTN/AN/AN/AN/AYESYESYES3P200BRCP `B' MTR. CURRENT INDICATORD30-1200 AMPMTR. CURRENTN/AN/AN/AN/AYESYESYES 3P200CRCP `C' MTR. CURRENT INDICATORD30-1200 AMPMTR. CURRENTN/AN/AN/AN/AYESYESYES D11 PRIMARY COOLANT SYSTEM - PRIMARY SYSTEM SAFETY RELIEF VALVE POSITIONPCV-455C (LS)PRZR PORV POSITIOND2OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 3---YESYES 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 22 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONPCV-456 (LS)PRZR PORV POSITIOND2OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 3---YESYES POSITION INDICATION LIGHTS FOR PCV 455C & 456D2OPEN/CLOSEDCLOSED/NOT CLOSEDN/AN/AN/ANOTE 3YES------ZS-6303APRIMARY SYSTEM SAFETY R.V. CODE SAFETY VALVED20-100%CLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESZS-6303BPRIMARY SYSTEM SAFETY R.V. CODE SAFETY VALVED20-100%CLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESZS-6303CPRIMARY SYSTEM SAFETY R.V. CODE SAFETY VALVED20-100%CLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYES LIGHT IND. FOR ZS-6306A,B,CD20-100%CLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------
D12 PRIMARY COOLANT SYSTEM - PRESSURIZER LEVEL LT-459 PRZR LEVEL CH. ID10-100%(150" TO 334")TOP TO BOTTOMCOMPL YCOMPLLT-460;LT-461NOTE 1---YESYES LI-459A PRZR LEVEL CH.I IND.D10-100%TOP TO BOTTOMN/ACOMPLLI-460;LI-461NOTE 1YES------
LT-460 PRZR LEVEL CH. IID10-100%(150" TO 334")TOP TO BOTTOMCOMPL YCOMPLLT-459;LT-461NOTE 1---YESYES LI-460 PRZR LEVEL CH. II IND.D10-100%TOP TO BOTTOMN/ACOMPLLI-459A;LI-461NOTE 1YES------
LT-461 PRZR LEVEL CH. IIID10-100%(150" TO 334")TOP TO BOTTOMCOMPL YCOMPLLT-459;LT-460NOTE 1---YESYES LI-461 PRZR LEVEL CH. III IND.D10-100%TOP TO BOTTOMN/ACOMPLLI-459A;LI-460NOTE 1YES------
LR-459 PRZR LEVEL RECORDER FOR LT-459, 460, 461D10-100%TOP TO BOTTOMN/ACOMPLN/ANOTE 1YES------
D13 PRIMARY COOLANT SYSTEM - PRESSURIZER HEATER STATUS 3B11 PRZR HEATER STATUS CONTROL GROUPD20-500 KW CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B 3B12 PRZR HEATER STATUS BACKUP GROUP 3AD20-600 AMP CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B 3B13 PRZR HEATER STATUS BACKUP GROUP 3BD20-600 AMP CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B D14 PRIMARY COOLANT SYSTEM - QUENCH TANK LEVEL LT-470 PRZR RELIEF TANK LEVEL (QUENCH TANK)D30-100%(0-100")TOP TO BOTTOMN/AN/AN/AN/A---YESYES LI-470PRZR RELIEF TANK LEVEL INDICATOR (QUENCH TANK)D30-100%TOP TO BOTTOMN/AN/AN/AN/AYES------
D15 PRIMARY COOLANT SYSTEM - QUENCH TANK TEMPERATURE TE-471 PRZR RELIEF TANK TEMPERATURED350-350 F 50-750 FN/AN/AN/AN/A---YESYESNOTE 3 TI-471 PRZR RELIEF TANK TEMPERATURE INDICATORD350-350 F 50-750 FN/AN/AN/AN/AYES------NOTE 3 D16 PRIMARY COOLANT SYSTEM - QUENCH TANK PRESSURE PT-472 PRZR RELIEF TANK PRESSURED30-120 PSIG0-DESIGN PRESSUREN/AN/AN/AN/A---YESYES PI-472 PRZR RELIEF TANK PRESSURE INDICATORD30-120 PSIG0-DESIGN PRESSUREN/AN/AN/AN/AYES------
D17 SECONDARY SYSTEM (STEAM GEN.) - S. G. LEVEL LT-474 S.G. `A' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-475;LT-476NOTE 1---YESYESNOTE 1L C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 23 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION LI-474 S.G. `A' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-475;LI-476NOTE 1YES------NOTE 1LLT-475S.G. `A' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-474;LT-476NOTE 1---YESYESNOTE 1LLI-475S.G. `A' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-474;LI-476NOTE 1YES------NOTE 1LLT-476S.G. `A' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-474;LT-475NOTE 1---YESYESNOTE 1LLI-476S.G. `A' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-474;LI-475NOTE 1YES------NOTE 1LFR-478S.G. `A' LVL CH. I, II, III NARROW RANGE RECORDD10-100%FROM TUBE SH TO SEPAR AN/ACOMPLN/ANOTE 1YES------NOTE 1L LT-484 S.G. `B' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-485;LT-486NOTE 1---YESYESNOTE 1LLI-484S.G. `B' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-485;LI-486NOTE 1YES------NOTE 1LLT-485S.G. `B' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-484;LT-486NOTE 1---YESYESNOTE 1LLI-485S.G. `B' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-484;LI-486NOTE 1YES------NOTE 1LLT-486S.G. `B' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-484;LT-485NOTE 1---YESYESNOTE 1LLI-486S.G. `B' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-484;LI-485NOTE 1YES------NOTE 1LFR-488S.G. `B' LVL CH. I, II, III NARROW RANGE RECORDD10-100%FROM TUBE SH TO SEPAR AN/ACOMPLN/ANOTE 1YES------NOTE 1L LT-494 S.G. `C' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-495;LT-496NOTE 1---YESYESNOTE 1LLI-494S.G. `C' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-495;LI-496NOTE 1YES------NOTE 1LLT-495S.G. `C' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-494;LT-496NOTE 1---YESYESNOTE 1LLI-495S.G. `C' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-494;LI-496NOTE 1YES------NOTE 1LLT-496S.G. `C' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.2 2 FROM TUBE SH TO SEPAR A COMPL YCOMPLLT-494;LT-495NOTE 1---YESYESNOTE 1LLI-496S.G. `C' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPAR AN/ACOMPLLI-494;LI-495NOTE 1YES------NOTE 1LFR-498S.G. `C' LVL CH. I, II, III NARROW RANGE RECORDD10-100%FROM TUBE SH TO SEPAR AN/ACOMPLN/ANOTE 1YES------NOTE 1L D18 SECONDARY SYSTEM (STEAM GEN.) - S. G. PRESSURE PT-474 S.G. `A' STEAM PRESSURE CH. IID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYESPI-474S.G. `A' STEAM PRESSURE CH. II IND.D20-1400 PSIGPa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-475 S.G. `A' STEAM PRESSURE CH. IIID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYESPI-475S.G. `A' STEAM PRESSURE CH. III IND.D20-1400 PSIGPa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-476 S.G. `A' STEAM PRESSURE CH. IVD20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES 9/24/04 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 24 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION PI-476 S.G. `A' STEAM PRESSURE CH. IV IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-484 S.G. `B' STEAM PRESSURE CH. IID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-484 S.G. `B' STEAM PRESSURE CH. II IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-485 S.G. `B' STEAM PRESSURE CH. IIID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-485 S.G. `B' STEAM PRESSURE CH. III IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-486 S.G. `B' STEAM PRESSURE CH. IVD20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-486 S.G. `B' STEAM PRESSURE CH. IV IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-494 S.G. `C' STEAM PRESSURE CH. IID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-494 S.G. `C' STEAM PRESSURE CH. II IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-495 S.G. `C' STEAM PRESSURE CH. IIID20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-495 S.G. `C' STEAM PRESSURE CH. III IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
PT-496 S.G. `C' STEAM PRESSURE CH. IVD20-1400 PSIG Pa TO 20% ABV MIN SV SE T COMPL YN/AN/ANOTE 1---YESYES PI-496 S.G. `C' STEAM PRESSURE CH. IV IND.D20-1400 PSIG Pa TO 20% ABV MIN SV SE TN/AN/AN/ANOTE 1YES------
D19 SECONDARY SYSTEM (STEAM GEN.) - SAFETY/RELIEF VALVE POSITIONS OR MAIN STEAM FLOWD19FT-474 S.G. 'A' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-474 S.G. 'A' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------D19FT-475 S.G. 'A' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-475 S.G. 'A' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------D19FT-484 S.G. 'B' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-484 S.G. 'B' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------D19FT-485 S.G. 'B' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-485 S.G. 'B' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------D19FT-494 S.G. 'C' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-494 S.G. 'C' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------D19FT-495 S.G. 'C' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPL YN/AN/ANOTE 1---YESYESD19FI-495 S.G. 'C' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------
Revised 04/17/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 25 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION D20 SECONDARY SYSTEM (STEAM GEN.) - SAFETY/RELIEF VALVE POSITIONS OR MAIN STEAM FLOW FT-476 S.G. `A' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-476 S.G. `A' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
FT-477 S.G. `A' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-477 S.G. `A' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
FT-486 S.G. `B' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-486 S.G. `B' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
FT-487 S.G. `B' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-487 S.G. `B' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
FT-496 S.G. `C' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-496 S.G. `C' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
FT-497 S.G. `C' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES FI-497 S.G. `C' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
D21 AUXILIARY FEEDWATER - AUXILIARY FEEDWATER FLOWFT-1401AAUX. F.W. FLOW TO S.G. `A'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1401AAUX. F.W. FLOW TO S.G. `A' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1401BAUX. F.W. FLOW TO S.G. `A'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1401BAUX. F.W. FLOW TO S.G. `A' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1457AAUX. F.W. FLOW TO S.G. `B'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1457AAUX. F.W. FLOW TO S.G. `B' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1457BAUX. F.W. FLOW TO S.G. `B'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1457BAUX. F.W. FLOW TO S.G. `B' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1458AAUX. F.W. FLOW TO S.G. `C'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1458AAUX. F.W. FLOW TO S.G. `C' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1458BAUX. F.W. FLOW TO S.G. `C'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1458BAUX. F.W. FLOW TO S.G. `C' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
Revised 04/17/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 26 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION D22 AUXILIARY FEEDWATER - CONDENSATE STORAGE TANK WATER LEVELLT-6384ACONDENSATE STORAGE TANKD10-100% (19" TO 583")PLANT SPECIFICN/ACOMPLLT-6384BNOTE 1*YESYESNOTES 1A,1BLI-6384ACONDENSATE STORAGE TANK IND.D10-100% (0 - 250K GAL)PLANT SPECIFICN/ACOMPLLI-6384BNOTE 1YES------LT-6384BCONDENSATE STORAGE TANKD10-100% (19" TO 583")PLANT SPECIFICN/ACOMPLLT-6384ANOTE 1*YESYESNOTES 1A,1BLI-6384BCONDENSATE STORAGE TANK IND.D10-100% (0 - 250K GAL)PLANT SPECIFICN/ACOMPLLI-6384ANOTE 1YES------
D23 CONTAINMENT COOLING SYSTEM CONTAINMENT SPRAY FLOW CONTAINMENT SPRAY PP. A IND. LIGHTSD2START-STOP0-110% DESIGN FLOWN/AN/AN/ANOTE 3YESYESYESNOTES 1F,3MOV-880A (LS)CONTAINMENT SPRAY PP. A OUTLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-880A IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3MOV-864A (LS)CONTAINMENT SPRAY PP. A INLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-864A IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3 CONTAINMENT SPRAY PP. B IND. LIGHTSD2START-STOP0-110% DESIGN FLOWN/AN/AN/ANOTE 3YESYESYESNOTES 1F,3MOV-880B (LS)CONTAINMENT SPRAY PP. B OUTLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-880B IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3MOV-864B (LS)CONTAINMENT SPRAY PP. B INLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-864B IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3 D24 CONTAINMENT COOLING SYSTEM - HEAT REMOVAL BY THE CTMT. FAN HEAT REMOVAL SYSTEMCV-2905 (LS)EMERGENCY CONTAINMENT COOLER A INLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2905 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2904 (LS)EMERGENCY CONTAINMENT COOLER C INLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2904 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2903 (LS)EMERGENCY CONTAINMENT COOLER B INLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2903 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2908 (LS)EMERGENCY CONTAINMENT COOLER A OUTLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2908 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2907 (LS)EMERGENCY CONTAINMENT COOLER C OUTLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2907 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1G C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 27 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONCV-2906 (LS)EMERGENCY CONTAINMENT COOLER B OUTLET VALVED2OPEN/CLOSEDPLANT SPECIFIC COMPL YN/AN/ANOTE 2---YESYESNOTE 1G CV-2906 IND. LIGHTSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GEMERGENCY CONTAINMENT COOLER FAN A IND. LIGHTSD2START-STOPPLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1GEMERGENCY CONTAINMENT COOLER FAN B IND. LIGHTSD2START-STOPPLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1GEMERGENCY CONTAINMENT COOLER FAN C IND. LIGHTSD2START-STOPPLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1G FT-613A CCW HEADER FLOWD20-14000 GPMPLANT SPECIFICN/AN/AN/ANOTE 1---YESYESNOTE 1G FI-613A CCW HEADER FLOW IND.D20-14000 GPMPLANT SPECIFICN/AN/AN/ANOTE 1YES------NOTE 1G FT-613B CCW HEADER FLOWD20-14000 GPMPLANT SPECIFICN/AN/AN/ANOTE 1---YESYESNOTE 1G FI-613B CCW HEADER FLOW IND.D20-14000 GPMPLANT SPECIFICN/AN/AN/ANOTE 1YES------NOTE 1G D25 CONTAINMENT COOLING SYSTEM - CTMT. ATMOS. TEMPERATURE TE-6700 CTMT. ATMOS. TEMPERATURED20-300 F 40-400 F COMPL YN/AN/ANOTE 1*YESYESNOTES 1B,3 TE-6701 CTMT. ATMOS. TEMPERATURED20-300 F 40-400 F COMPL YN/AN/ANOTE 1*YESYESNOTES 1B,3 TE-6702 CTMT. ATMOS. TEMPERATURED20-300 F 40-400 F COMPL YN/AN/ANOTE 1*YESYESNOTES 1B,3 D26 CONTAINMENT COOLING SYSTEM - CTMT. SUMP WTR. TEMPERATURE TE-604A RHR HX A INLET TEMPERATURED250-400 F 50-250 F COMPL YN/AN/ANOTE 1---YESYESNOTE 3 TE-604B RHR HX B INLET TEMPERATURED250-400 F 50-250 F COMPL YN/AN/ANOTE 1---YESYESNOTE 3 TR-604 RHR HX INLET TEMPERATURE RECORDERD250-400 F 50-250 FN/AN/AN/ANOTE 1YES------NOTE 3 D27 CHEMICAL & VOLUME CONTROL SYSTEM - MAKEUP FLOW FT-122 CHARGING FLOWD20-150 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYES FI-122A CHARGING FLOW IND.D20-150 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D28 CHEMICAL & VOLUME CONTROL SYSTEM - LETDOWN FLOW FT-150 LO PRESSURE LETDOWN FLOWD20-150 GPM0-110% DESIGN FLOWCOMPL YN/AN/ANOTE 1---YESYES FI-150 LO PRESSURE LETDOWN FLOW IND.D20-150 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D29 CHEMICAL & VOLUME CONTROL SYSTEM - VOLUME CONTROL TANK LEVEL LT-115 VOL. CONTROL TANK LEVELD20-100%(0-59")TOP TO BOTTOMN/AN/AN/ANOTE 1---YESYES LI-115 VOL. CONTROL TANK LEVEL IND.D20-100%TOP TO BOTTOMN/AN/AN/ANOTE 1YES------
D30 COOLING WATER SYSTEM - COMPONENT COOLING WATER TEMP. TO ESF SYSTEM TE-607A COMPONENT COOLING Hx OUTLET TEMPERATURED250-200 F 40-200 FN/AN/AN/ANOTE 1---YESYESNOTE 3 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 28 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION TI-607ACOMPONENT COOLING Hx OUTLET TEMPERATURE IND.D250-200 F 40-200 FN/AN/AN/ANOTE 1YES------NOTE 3 TE-607B COMPONENT COOLING Hx OUTLET TEMPERATURED250-200 F 40-200 FN/AN/AN/ANOTE 1---YESYESNOTE 3 TI-607BCOMPONENT COOLING Hx OUTLET TEMPERATURE IND.D250-200 F 40-200 FN/AN/AN/ANOTE 1YES------NOTE 3 D31 COOLING WATER SYSTEM - COMPONENT COOLING WATER FLOW TO ESF SYSTEM FT-613A CCW HEADER FLOWD20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYES FI-613A CCW HEADER FLOW IND.D20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
FT-613B CCW HEADER FLOWD20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYES FI-613B CCW HEADER FLOW IND.D20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D32 RADWASTE SYSTEMS - HIGH LEVEL RADIOACTIVITY LIQUID TANK LEVEL LIT-1001 WASTE HOLDUP TANK LEVELD30-24,000 GALTOP TO BOTTOMN/AN/AN/AN/A**YESYESNOTE 1B D33 RADWASTE SYSTEMS - RADIOACTIVE GAS HOLDUP TANK PRESSURE PT-1036 GAS DECAY TANK `A' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B PT-1037 GAS DECAY TANK `B' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B PT-1038 GAS DECAY TANK `C' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B PT-1039 GAS DECAY TANK `D' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B PT-1052 GAS DECAY TANK `E' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B PT-1053 GAS DECAY TANK `F' (HOLDUP) PRESSURED30-160 PSIG 0-150% DESIGN PRESSUR EN/AN/AN/AN/A**YESYESNOTE 1B D34 VENTILATION SYSTEM - EMERGENCY VENTILATION DAMPER POSITIOND1-A (LS)C.R. NORMAL INTAKE DAMPER POSITIOND2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-1A IND. LIGHTS ASSOCIATED WITH HIS-6552AD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-1B (LS)C.R. NORMAL INTAKE DAMPER POSITIOND2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-1B IND LIGHTS ASSOCIATED WITH HIS-6552BD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------
D-2 (LS)C.R. EMERGENCY AIR INTAKE EASTD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-2 IND. LIGHTS ASSOCIATED WITH HIS-6541D2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 1YES------
D-3 (LS)C.R. EMERGENCY AIR INTAKE WESTD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-3 IND. LIGHTS ASSOCIATED WITH HIS-6542D2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 1YES------D-11A (LS)C.R. FILTER RECIRCULATION AIRD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-11A IND. LIGHTS ASSOCIATED WITH HIS-6543AD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-11B (LS)C.R. FILTER RECIRCULATION AIRD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 29 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION D-11B IND. LIGHTS ASSOCIATED WITH HIS-6543BD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-14 (LS)TOILET EXHAUSTD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-14 IND. LIGHTS ASSOCIATED WITH HIS-6550D2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-22 (LS)KITCHEN EXHAUSTD2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-22 IND. LIGHTS ASSOCIATED WITH HIS-6549D2OPEN/CLOSEDOPEN/CLOSEDN/AN/AN/ANOTE 2YES------
D35 POWER SUPPLIES - STATUS OF STANDBY POWER & OTHER ENERGY SOURCES IMPORTANT TO SAFETY 3AA`3A' 4KV BUS VOLTAGED20-5000 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3AA02 3A 4KV BUS-AUX XFRMER CURRENTD20-4000 AMPSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3AA05 3A 4KV BUS-S/U XFRMER CURRENTD20-4000 AMPSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3AB`3B' 4KV BUS VOLTAGED20-5000 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3AB02 3B 4KV BUS-AUX XFRMER CURRENTD20-4000 AMPSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3AB05 3B 4KV BUS-S/U XFRMER CURRENTD20-4000 AMPSPLANT SPECIFICN/AN/AN/ANOTE 5YES------
3K4A EMERGENCY DIESEL GENERATOR `3A' VOLTAGED20-5000 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3K4A EMERGENCY DIESEL GENERATOR `3A' CURRENTD20-600 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3K4B EMERGENCY DIESEL GENERATOR `3B' VOLTAGED20-5000 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3K4B EMERGENCY DIESEL GENERATOR `3B' CURRENTD20-600 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3B01 LOAD CENTER 3A STATUSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3BO2 LOAD CENTER 3B STATUSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3B03 LOAD CENTER 3C STATUSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3B04 LOAD CENTER 3D STATUSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES 3B50 LOAD CENTER 3H STATUSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B05 MCC-3A VITAL BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 30 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION 3B06 MCC-3B VITAL BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B07 MCC-3C VITAL BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B08 MCC-3D VITAL BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B52 MCC 3K BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B53 MCC 3L BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3B54 MCC 3M BUS VOLTSD20-600 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3D01 3A BATTERY AND CHARGER BUS VOLTSD20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 4D01 4B BATTERY AND CHARGER BUS VOLTSD20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3D23 3B BATTERY AND CHARGER BUS VOLTSD20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 4D23 4A BATTERY AND CHARGER BUS VOLTSD20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3Y01 INVERTER 3A CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y01 INVERTER 3A VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y02 INVERTER 3B CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y02 INVERTER 3B VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y05 INVERTER 3C CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y05 INVERTER 3C VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y07 INVERTER 3D CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y07 INVERTER 3D VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y04 INVERTER AS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y04 INVERTER AS VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 4Y04 INVERTER BS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 4Y04 INVERTER BS VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y06 INVERTER CS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3Y06 INVERTER CS VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 31 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION 4Y06 INVERTER DS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 4Y06 INVERTER DS VOLTAGED20-150 VOLTSPLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B 3AA02NO.3 AUX XFRMER TO 3A 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AA05NO.3 S/U XFRMER TO 3A 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AB02NO.3 AUX XFRMER TO 3B 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AB05NO.3 S/U XFRMER TO 3B 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AA22NO.4 S/U XFRMER TO 3A 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AA20 EDG `3A' TO 3A 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AB20 EDG `3B' TO 3B 4160 VOLT BUS BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AA09 3B 4KV BUS TIE TO 3AC13 AND 3AB22 BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YES------
3AB22 3B 4KV BUS TIE TO 3AA09 AND 3AC13 BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YES------
3AA08 3A 4KV BUS TO LOAD CENTER 3A BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AB09 3B 4KV BUS TO LOAD CENTER 3B BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AA14 3A 4KV BUS TO LOAD CENTER 3C BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3AB14 3B 4KV BUS TO LOAD CENTER 3D BKR STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES 3P06 120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3P07 120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3P08 120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B 3P09 120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSEDPLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B PS-2323AFW NITRO. BACKUP SUPPLY STA. 1 TR. 1 LOW PRESS.D2NORMAL/LOW PRESS U PLANT SPECIFICN/AN/AN/ANOTE 1---YESYES AFW NITRO. STA. 1 TR. 1 LOW PRESS. IND. LIGHTD2NORMAL/LOW PRESS U PLANT SPECIFICN/AN/AN/ANOTE 1YES------
PS-2322AFW NITRO. BACKUP SUPPLY STA. 1 TR. 2 LOW PRESS.D2NORMAL/LOW PRESS U PLANT SPECIFICN/AN/AN/ANOTE 1---YESYES AFW NITRO. STA. 1 TR. 2 LOW PRESS. IND. LIGHTD2NORMAL/LOW PRESS U PLANT SPECIFICN/AN/AN/ANOTE 1YES------
Revised 04/17/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 32 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION E1 CONTAIMENT RADIATION - CONTAIMENT AREA RADIATION HI RANGERAD-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A'E11 TO 1E8 R/Hr1 TO 1E7 R/Hr COMPL YCOMPLRAD-6311BNOTE 1---YESYESRAI-6311ACTMT. HIGH RANGE RAD. MONITOR CH. 'A' IND.E11 TO 1E8 R/Hr1 TO 1E7 R/HrN/ACOMPLRAI-6311BNOTE 1YES------RAR-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A' RECORDERE11 TO 1E8 R/Hr1 TO 1E7 R/HrN/ACOMPLRAR-6311BNOTE 1YES------
RAD-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B'E11 TO 1E8 R/Hr1 TO 1E7 R/HrCOMPL YCOMPLRAD-6311ANOTE 1---YESYESRAI-6311BCTMT. HIGH RANGE RAD. MONITOR CH. 'B' IND.E11 TO 1E8 R/Hr1 TO 1E7 R/HrN/ACOMPLRAI-6311ANOTE 1YES------
RAR-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B' RECORDERE11 TO 1E8 R/Hr1TO 1E7 R/HrN/ACOMPLRAR-6311ANOTE 1YES------
E2 AREA RADIATION - RADIATION EXPOSURE RATE RD-1417 EAST END OF E/W CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1418 WEST END OF E/W CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1419 SPENT FUEL PIT EXHAUSTE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1420 CONTROL ROOME31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1415 NORTH END OF N/S CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1416 SOUTH END OF N/S CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1413 OUTSIDE SAMPLE RM.-UNIT 3E31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 RD-1414 OUTSIDE SAMPLE RM.-UNIT 4E31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4 R-1405 RECORDERE31E-4 TO 1E4 R/HR1E-1 TO 1E4 R/HRN/AN/AN/AN/AYES------NOTE 4 E3 AIRBORNE RADIOACTIVE MATERIALS RELEASED FROM PLANT NOBLE GAS & VENT FLOW RATE NONE CTMT. OR PURGE EFFLUENTE2THIS DESIGN NOT US E1E-6 TO 1E5 micro Ci/CC---------------------NOTE 5 NONE CTMT. OR PURGE EFFLUENT (FLOW)E2THIS DESIGN NOT US E0-110% DESIGN FLOW---------------------NOTE 5 C26* DCS(SPDS)
Revised 04/17/2013 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 33 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION NONE REACTOR SHIELD BLDG ANNULUSE2THIS DESIGN NOT US E1E-6 TO 1E4 micro Ci/CC---------------------NOTE 5 NONE AUXILIARY BLDGE2THIS DESIGN NOT US E1E-6 TO 1E3 micro Ci/CC---------------------NOTE 5 NONE AUXILIARY BLDG.(FLOW)E2THIS DESIGN NOT US E0-110% DESIGN FLOW---------------------NOTE 5 E4 CONDENSER AIR REMOVAL SYSTEMRAD-6417AIR EJECTOR CONDENSER EXH.E21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E5 micro Ci/CCN/AN/AN/A**YESYESNOTE 1B NONE AIR EJECTOR CONDENSER FLOWE2- - -- - ----------------------NO INST.
E5 COMMON VENT - NOBLE GASESRAD-6304VENT STACK W.R. RAD. MONITORE21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E3 micro Ci/CCN/AN/AN/A**YESYESNOTE 1B FT-6584 VENT STACK-FLOWE20-150,000 CFM0-110% DESIGN FLOWN/AN/AN/A**YESYESNOTE 1B E6 VENT FROM STEAM GENERATOR SAFETY RELIEF VALVERAD-6426STEAM LINE RAD. MONITORE21E-1 TO 1E3 micro Ci/
C1E-1 TO 1E3 micro Ci/CCN/AN/AN/A**YESYESNOTES 1B E6A ALL OTHER IDENTIFIED RELEASE POINTSRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONLY)E21E-7 TO 1E5 micro Ci/
C1E-6 TO 1E2 micro Ci/CCN/AN/AN/A**YESYESNOTE 1B E7 PARTICULATES & HALOGENS - ALL IDENTIFIED PLANT RELEASE POINTSRAD-6304VENT STACKE3NOTE 31E-3 TO 1E2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1BRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONLY)E3NOTE 31E-3 TO 1E2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1B E8 ENVIRONS RADIATION AND RADIOACTIVITY GRAB SAMPLE &
M AIRBORNE RADIOHALOGENS AND PARTICULATESE3NOTE 11E-9 TO 1E-3 micro Ci/CCN/AN/AN/A------------RO-2,RO-2A & RO-PLANT AND ENVIRONS RADIATIONE3NOTE 2 NOTE 1N/AN/AN/A------------PORTABLE MCAPLANT AND ENVIRONS RADIOACTIVITYE3ISOTOPIC ANALYSISISOTOPIC ANALYSISN/AN/AN/A------------
E9 METEOROLOGY - WIND DIRECTION & SPEED ESTIMATE OF ATMOSPHERIC STABILITY 10 M. W.D. SO. DA D METEOROLOGY 10 METER WIND DIRECTIONE30-540 DEGREE0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.D. TURKE Y METEOROLOGY 10 METER WIND DIRECTIONE30-540 DEGREE0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 60 M. W.D. SO. DA D METEOROLOGY 60 METER WIND DIRECTIONE30-540 DEGREE0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.S. SO. DA D METEOROLOGY 10 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.S. TURKE Y METEOROLOGY 10 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1B 60 M. W.S. SO. DA D METEOROLOGY 60 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1BDELTA T "A" S.D.ESTIMATE OF ATMOSPHERIC STABILITYE3-5 TO +15 F-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,3DELTA T "B" S.D.ESTIMATE OF ATMOSPHERIC STABILITYE3-5 TO +15 F-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,3 C27* DCS(SPDS)
Revised 01/08/2014 TABLE 7.5-1 PARAMETER LISTING
SUMMARY
SHEETS UNIT 3 TURKEY POINTSHEET 34 OF 34 VARIABLE INSTRUMENT RANGE DISPLA YITEMTAG NO.
ENVIRO N S EISMI CREDUNDANCEPOWER LOCATIONSCHEDULE
/DESCRIPTION T YP E CA T EXISTINGREQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION 10M SIGMA THET A ESTIMATE OF ATMOSPHERIC STABILITYE30-100 DEGREES-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,3 E10 ACCIDENT SAMPLING CAPABILITY - PRIMARY COOLANT AND SUMP NONE RCS ACTIVITY GROSS CPS NOTE 410 micro Ci/ml to 10 Ci/mlN/AN/AN/AN/AN/AN/AN/A NONE CTMT. AIR-ISOTOPIC ANALYSIS GAMMA SPECTRUM NOTE 4ISOTOPIC ANALYSISN/AN/AN/AN/AN/AN/AN/A NONEBORON ANALYZER RCS SOLUBLE BORON CONCENTRATIONNOTE 4 0-6000 PPMN/AN/AN/AN/AN/AN/AN/A NONE RCS CHLORIDE ANALYSIS OF PRIMARY COOLANT NOTE 4 0-20 PPMN/AN/AN/AN/AN/AN/AN/A NONE DISSOLVED HYDROGEN ANALYSIS OF PRIMARY COOLANT NOTE 4 0 - 2000 CC/KGN/AN/AN/AN/AN/AN/AN/A NONE DISSOLVED OXYGEN ANALYSIS OF PRIMARY COOLANT NOTE 4 0-20 PPMN/AN/AN/AN/AN/AN/AN/A NONE RCS pH ANALYSIS OF PRIMARY COOLANT NOTE 4 1-13 pHN/AN/AN/AN/AN/AN/AN/AAE-6307ACTMT. HYDROGEN CONCENTRATION CH. AE30-10% AND 0-20%0-10 VOL. %N/AN/AN/AN/A*YESYESNOTE 1BAE-6307BCTMT. HYDROGEN CONCENTRATION CH. BE30-10% AND 0-20%0-10 VOL. %N/AN/AN/AN/A*YESYESNOTE 1B NONE CONTAINMENT OXYGEN NOTE 4 0-30 VOL. %N/AN/AN/AN/AN/AN/AN/A NONE CTMT. AIR GAMMA SPECTRUM NOTE 4ISOTOPIC ANALYSISN/AN/AN/AN/AN/AN/AN/A C26 C26* DCS(SPDS)
Revised 01/31/2013 NOTES FOR TABLE 7.5-1
TURKEY POINT UNIT 3 Sheet 1 OF 9 For Tag No. Column (LS) = Limit Switch Associated with Valve For Existing Instrument Range Column1.Portable sampling with onsite analysis capability is capable of providing a range from less than 1E-9 micro Ci/CC to greater than 1E-3 micro Ci/CC.2.Portable instrumentation provides a range of:A.1E-3 R/HR to values greater than 1E4 R/HR photons; and B.1E-3 R/HR to values greater than 1E4 R/HR beta and low-energy photons
- 3. Existing range monitors up to 7.4E-2 micro Ci/CC. Plant specific analysis justifies smaller range. Particulates and halogens collected on
filter cartridge and monitored in lab after sample collection period (30
minutes design for accident situations).
- 4. No instrument is provided for this variable. Elimination of the need to provide on-site analysis capability for this variable has been accepted by the NRC in their safety analysis report related to technical specification amendments 211/205, dated 1/31/2001.
For Required Instrument Range Column1.RG 1.97 requires the following ranges:A.1E-3 R/HR to 1E4 R/HR photons; andB.1E-3 R/HR to 1E4 R/HR beta and low-energy photons For Environmental Qualification Column1.The Safety Injection Accumulator Discharge Valves MOV-865A, B and C are administratively controlled and are required to be in the open position during normal operation. These valves are not required to change
position under accident conditions. Administrative control is
accomplished by locking open the associated motor control center circuit
breakers. Since administrative control via electrical de-energization of
the valves ensures that the valves will be in their safe position during 06/18/2001 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 2 OF 9 an accident, environmental qualification of the limit switches providing position indication is not required.
For Power Supply Column
Power source is identified as:
- 1. Class 1E, 120 VAC uninterruptable power supply (inverters)
- 2. Class 1E, 120 VAC power backed up by the Emergency Diesel Generator
- 3. Class 1E, 125 VDC safety-related battery
- 4. Non-Class 1E, 120 VAC uninterruptable power supply
- 6. Transducers internal to the inverter providing computer display signals for inverter current and voltage are powered by the inverter internals.
- 7. The SPING monitors communicate with both primary and backup control terminals which are powered from plant inverters and backed up by the
safety-related batteries. SPING Monitors RAD-3(4)-6417 are powered from
non-vital lighting panels capable of being powered from the emergency
diesel generators. SPING Monitors RAD-3-6418, RAD-6426, and RAD-6304
are powered from a vital AC power panel which is automatically backed up
by an emergency diesel generator.
For Display Location
- 1. Control Room metering is credited for primary indication of Emergency Diesel Generator Output (MW). Recording capability for this variable is
also available via DCS/ERDADS.
For Schedule/Justification Column
- 1. The following notes referenced under the "Schedule/Just" column of the
Revised 04/17/2013 C26C26 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 3 OF 9 Parameter Listing Summary Sheets correspond to the technical justifications identified below:
A. This justification demonstrates the acceptability of the existing uninterruptable power source (UPS) associated with the DCS (SPDS/ERDADS) for the monitoring of Category 1 variables. This
acceptability is based upon the existing UPS allowing the DCS (SPDS/ERDADS) to perform its credited RG 1.97 functions:
(1) Recording of Category 1 Variables -
Control Room indication is normally used to provide trending while DCS (SPDS/ERDADS) is used only as a backup to those
instruments. In those cases where DCS (SPDS/ERDADS) is
being used to trend Category 1 variables, either the
trending is not necessary to the Control Room operator's
decisions or the operator can obtain the real time
information via the monitoring of Control Room indication.
(2) Indication of Category 1 Variables -
DCS (SPDS/ERDADS) is only used as a backup means of indication for certain containment isolation valves but is
not credited for RG 1.97 indication for any other Category 1
variable.
(3) Containment Isolation Valve Indication -
In the few instances where DCS (SPDS/ERDADS) is credited for backup indication associated with containment isolation
valves, computer power will be available from the UPS
battery for at least the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the accident.
This period of operability is sufficient to allow the
completion of containment isolation.
B. This justification demonstrates that the DCS (SPDS/ERDADS), although classified as non-nuclear safety-related, is capable of
Revised 01/31/2013 C26C26C26C26C26 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 4 OF 9 providing the necessary Regulatory Guide functions for which it is credited.
(1) The DCS (SPDS/ERDADS) is not essential to the monitoring of Category 1 variables. The computer is credited only for
backup indication of a few containment isolation valves (i.e., valve position indication) but is not credited for
either primary or backup indication for any other Category 1
variables.
(2) The DCS (SPDS/ERDADS) is not essential in providing the Control Room operators with vital trending or recording
information. Control Room indication is normally used to
provide trending while DCS (SPDS/ERDADS) is used only as a
backup to those instruments. In those cases where DCS (SPDS/ERDADS) is being used to trend Category 1 variables, either the trending is not necessary to the operator's
decisions or the operator can obtain the real time
information via the monitoring of Control Room indication.
(3) The DCS (SPDS/ERDADS) provides primary indication of certain Category 2 and 3 variables. In general, the DCS (SPDS/ERDADS) complies with the Category 2 and 3 design and
qualification criteria identified in Table 1 of RG 1.97.
(4) The DCS (SPDS/ERDADS) does not diminish the capability of the Control Room operators to obtain the necessary post-
accident monitoring information or in achieving the safe
shutdown of the plant. Based upon conclusions (1) and (2)
above, it can be further concluded that the DCS (SPDS/ERDADS) does not perform an essential function with
respect to Category 1 post-accident monitoring.
C. This justification demonstrates that the lack of overlap between the ranges of Containment Sump Water Level narrow and wide range
instrumentation does not jeopardize the capability of providing the
Control Room operators the critical information required during
Revised 01/31/2013 C26C26C26C26C26C26 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 5 OF 9 plant accident conditions. This is based on an analysis which provides the following:(1)The deadband in Containment Sump Water Level indication between 369" and 397" causes less than 6% error in indication.(2)The resulting error in indication is introduced in a non-critical range of the required indication. Thus the deadband does not prevent the operator from obtaining the required information:(a)Low level (narrow range) indication of the initial ingress of water into the sump to allow the assessment of water source and rate.(b)High level (wide range) indication for operator response to containment flooding.(c)Determination of the ability to transfer to cold leg recirculation in the event of loss of reactor or secondary
coolant based upon having achieved minimum pump NPSH.D.This justification clarifies the inconsistency between the Accumulator Tank Level ranges identified in the previous FPL RG 1.97 submittals of
January 26, 1984 and May 10, 1985, and the existing Control Room
instrumentation range of 6,500 to 6,750 gals. The existing Control Room
range of 6,500 to 6,750 gals. uses the same basis for justification as
identified and approved by NRC in its Safety Evaluation dated March 20, 1986. Accumulator tank pressure is also credited for determining
accumulator tank level. As pressure drops in the accumulators, application of the Ideal-Gas state equation provides indication of how
much water remains in the accumulator following actuation. As an
operator aid, a curve has been made available to the operator which correlates accumulator pressure to accumulator level. The accumulator instrumentation have been down graded, per NRC letter dated April 13, 1992 Docket Nos. 50-250 and 50-251, from R.G. 1.97 Category 2 to Category
3.Rev 11 11/93 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 6 OF 9E.This justification clarifies the use of flow meters integral to hand indicating controllers as a means of providing valve position indication. The integral flow meters provide "closed" position
indication by indicating zero flow and "not closed" position
indication by indicating higher than zero flow.F.This justification identifies alternative instrumentation being credited for the monitoring of Containment Spray Flow. An
alternative method of monitoring this variable was identified in to FPL RG 1.97 submittal dated May 10, 1985. The
alternative instrumentation provides monitoring of the operation of
the Containment Spray System, as intended by RG 1.97. This is
accomplished by monitoring the proper alignment of Containment Spray
valves and operation of the Containment Spray pumps. In addition, the monitoring of containment temperature and pressure assures that
containment cooling systems are performing their required function.
Monitoring of RWST level provides indirect indication of the
Containment Spray flow function.G.This justification identifies alternative instrumentation being credited for the monitoring of Containment Fan Heat Removal. An
alternative method of monitoring this variable was identified in to FPL RG 1.97 submittal dated May 10, 1985. The
method used to address this variable monitors the operation of the
Emergency Containment Cooling (ECC) fans and verifies that Component
Cooling Water (CCW) flow has been established to the ECC coolers.
In addition, the monitoring of containment pressure and temperature
provides indirect indication of the Containment Fan Heat Removal
function.H.This justification provides the rationale for not recording containment isolation valve position (Category 1 variable).
Recording of containment isolation valve position is not essential
for operator action. Containment valve position is available to the
operators via Control Room indicating lights. The operators depend
on the real time information provided by indicating lights to verify
containment isolation. Thus the operators do not need trending of
valve position to verify isolation.
Rev. 10 7/92 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 7 OF 9I.This justification provides the basis for the acceptability of the existing range for Containment Sump Water Level narrow range indication. The existing range of LI-6308A&B includes a 0-5 inch deadband (i.e., no specific reading can be obtained). However, since the 0-5 inch deadband is outside of the loop measurement range and insignificant compared to the span of 364 inches, the lower limit of the indicator scale of 0-5 inches is acceptable. J.This justification provides the basis for the acceptability of the existing range for Containment Sump Water Level narrow range recording. The existing range of LR-6308A&B includes a 0-5 inch deadband (i.e., no specific reading can be obtained). However, since the 0-5 inch deadband is insignificant compared to the span of 364 inches, the lower limit of the recorder scale of 0-5 inches is acceptable.K.This justification demonstrates that the lack of units of measure (i.e., inches) associated with the Control Room indication for Containment Sump Water Level wide range, LI-6309A&B, will not mislead the Control Room operators. This is based on the operators being familiar with the applicable units of measure via training.L.Wide range monitoring for Steam Generator Level is provided via a single non-Class 1E wide range level loop. This justification demonstrates that, although wide range monitoring may not be available during an accident scenario, the Control Room operator will have sufficient information to identify and mitigate an accident and to determine the availability of the steam generators as heat sinks. This is based upon the following:(1)Steam generator level will either remain within narrow range level indication or, if steam generator level has fallen below narrow range indication, that Auxiliary Feedwater has been initiated and will result in the recovery of steam generator level to within narrow range limits. This is accomplished via the associated emergency operating procedures.
Rev. 10 7/92 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 8 OF 9(2)RCS temperature (i.e., hot and cold leg water temperature) and pressure are available to determine the effectiveness of the steam generators as heat sinks.2.Since the original containment isolation design for Turkey Point was not required to provide redundant valve position indication, the redundancy criteria of RG 1.97 are not applicable to the existing plant design. As a result, in order to address the RG 1.97 concern for ensuring Control Room capability to verify isolation status, an RG 1.97 Containment Isolation Valve Evaluation was performed. The evaluation considers the effects of single failure of valve indication and demonstrates the capability for the Control Room operator to verify isolation of Containment penetrations.3.An exception to this variable has been accepted by NRC in its Safety Evaluation Report dated March 20, 1986.4.All 24 channels of the Area Radiation Monitoring System (ARMS) have been replaced by PC/M 89-462 to comply with commitments made to the NRC in FPL letter L-88-290 (Reference 6). L-88-290 commitments require the use of instrumentation with a range of 10
-3 R/hr to 10 2 R/hr. Instrumentation installed under PC/M 89-462 has a range of 10
-4 R/hr to 10 4 R/hr, which exceeeds both Regulatory Guide 1.97 recommendations and L-88-290 commitments.5.No instrumentation has been provided since effluent discharge is through a common plant vent.6.No recording capability exists for 4KV Bus Voltage (Category 1 variable).
The emergency operating procedures presently credit the monitoring of 4KV Bus Voltage to allow the Control Room operator to determine the loss of power to a 4KV bus. Control Room meter indication of 4KV bus voltage is available and is adequate to allow the operator to identify the loss of bus voltage on a realtime basis. Trending of bus voltage is not necessary. Therefore, recording of the variable is not essential.
Rev. 10 7/92 NOTES FOR TABLE 7.5-1 (Continued)
Sheet 9 OF 9
- 7. NOT USED
- 8. NOT USED
- 9. The original plant design included 51 core exit thermocouples. Due to the potential for individual sensor failures, the actual number of
operable thermocouples may be reduced below this value.
- 10. NOT USED
- 11. Existing instrument range of "OPEN/CLOSED" is derived from a single
limit switch contact. The contact provides CLOSED/NOT CLOSED position
to ERDADS which then defines and displays the position as OPEN or CLOSED
at the CR, TSC and EOF consoles.
- 12. On-site analysis capability for this variable has been eliminated in
favor of grab samples and offsite analysis. This change is consistent
with commitments documented in NRC safety evaluation for technical
specification amendments 211/205, dated 1/31/2001.
Revised 04/17/2013 C26C26C26 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 1 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION A1 RCS PRESSUREPT-404RCS PRESSUREA10-3000 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-406NOTE 1*YESYESNOTES 1A,1BPT-406RCS PRESSUREA10-3000 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY AA10-3000 PSIG PLANT SPECIFICN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-3000 PSIG PLANT SPECIFICN/ACOMPLYQSPDS ANOTE 1YES------
A2 RCS HOT LEG WATER TEMPERATURETE-413ARCS HOT LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-413BNOTE 1---YESYESTE-413BRCS HOT LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-413ANOTE 1---YESYESTE-423ARCS HOT LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-423BNOTE 1---YESYESTE-423BRCS HOT LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-423ANOTE 1---YESYESTE-433ARCS HOT LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-433BNOTE 1---YESYESTE-433BRCS HOT LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-433ANOTE 1---YESYESTR-413RCS HOT LEG TEMP. RECORD LOOP A,B,C FOR TA10-750 F PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------QSPDS ADISPLAY AA10-750 F PLANT SPECIFICN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-750 F PLANT SPECIFICN/ACOMPLYQSPDS ANOTE 1YES------
A3 RCS COLD LEG WATER TEMPERATURETE-410ARCS COLD LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-410BNOTE 1---YESYESTE-410BRCS COLD LEG WTR. TEMP. LOOP AA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-410ANOTE 1---YESYESTE-420ARCS COLD LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-420BNOTE 1---YESYESTE-420BRCS COLD LEG WTR. TEMP. LOOP BA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-420ANOTE 1---YESYESTE-430ARCS COLD LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-430BNOTE 1---YESYESTE-430BRCS COLD LEG WTR. TEMP. LOOP CA10-750 F PLANT SPECIFICCOMPLYCOMPLYTE-430ANOTE 1---YESYESTR-410RCS COLD LEG TEMP. RECORD LOOP A,B,C FO RA10-750 F PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------QSPDS ADISPLAY AA10-750 F PLANT SPECIFICN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY BA10-750 F PLANT SPECIFICN/ACOMPLYQSPDS ANOTE 1YES------
A4 RWST LEVELLT-6583ARWST CH. A LEVELA10-335,000 GAL PLANT SPECIFICN/ACOMPLYLT-6583BNOTE 1**YESYESNOTES 1A,1B C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 2 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLI-6583ARWST CH. A LEVEL IND.A10-335,000 GAL PLANT SPECIFICN/ACOMPLYLI-6583BNOTE 1YES------LT-6583BRWST CH. B LEVELA10-335,000 GAL PLANT SPECIFICN/ACOMPLYLT-6583ANOTE 1**YESYESNOTES 1A,1BLI-6583BRWST CH. B LEVEL IND.A10-335,000 GAL PLANT SPECIFICN/ACOMPLYLI-6583ANOTE 1YES------
A5 S. G. LEVEL NARROW RANGELT-474S.G. `A' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-475;LT-476NOTE 1---YESYESNOTE 1LLI-474S.G. `A' LVL. CH. I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-475;LI-476NOTE 1YES------NOTE 1LLT-475S.G. `A' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-474;LT-476NOTE 1---YESYESNOTE 1LLI-475S.G. `A' LVL. CH. II NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-474;LI-476NOTE 1YES------NOTE 1LLT-476S.G. `A' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-474;LT-475NOTE 1---YESYESNOTE 1LLI-476S.G. `A' LVL. CH. III NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-474;LI-475NOTE 1YES------NOTE 1LFR-478S.G. `A' LVL. CH. I, II, III NARROW RANGE RECOR DA10-100%PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1LLT-484S.G. `B' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-485;LT-486NOTE 1---YESYESNOTE 1LLI-484S.G. `B' LVL. CH.I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-485;LI-486NOTE 1YES------NOTE 1LLT-485S.G. `B' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-484;LT-486NOTE 1---YESYESNOTE 1LLI-485S.G. `B' LVL. CH. II NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-484;LI-486NOTE 1YES------NOTE 1LLT-486S.G. `B' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-484;LT-485NOTE 1---YESYESNOTE 1LLI-486S.G. `B' LVL. CH. III NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-484;LI-485NOTE 1YES------NOTE 1LFR-488S.G. `B' LVL. CH. I, II, III NARROW RANGE RECOR DA10-100%PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1LLT-494S.G. `C' LVL. CH. I NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-495;LT-496NOTE 1---YESYESNOTE 1LLI-494S.G. `C' LVL. CH. I NARROW RANGE IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-495;LI-496NOTE 1YES------NOTE 1LLT-495S.G. `C' LVL. CH. II NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-494;LT-496NOTE 1---YESYESNOTE 1LLI-495S.G. `C' LVL. CH. II NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-494;LI-496NOTE 1YES------NOTE 1LLT-496S.G. `C' LVL. CH. III NARROW RANGEA10-100%(30.1" TO 138.22")PLANT SPECIFICCOMPLYCOMPLYLT-494;LT-495NOTE 1---YESYESNOTE 1LLI-496S.G. `C' LVL. CH. III NARROW RANGE IND.A10-100%
PLANT SPECIFICN/ACOMPLYLI-494;LI-495NOTE 1YES------NOTE 1LFR-498S.G. `C' LVL. CH. I, II, III NARROW RANGE RECORA10-100%
PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1L A6 REACTIVITY CONTROL - NEUTRON FLUXND-6649ANEUTRON FLUX DETECTORA11E-8 TO 200% FULL POW E PLANT SPECIFICCOMPLYCOMPLYND-6649BNOTE 1**YESYESNOTES 1A,1B C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 3 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONNI-6649A-2NEUTRON FLUX INDICATORA11E-8 TO 200% FULL POW E PLANT SPECIFICN/ACOMPLYNI-6649B-2NOTE 1YES------ND-6649BNEUTRON FLUX DETECTORA11E-8 TO 200% FULL POW E PLANT SPECIFICCOMPLYCOMPLYND-6649ANOTE 1**YESYESNOTES 1A,1BNI-6649B-2NEUTRON FLUX INDICATORA11E-8 TO 200% FULL POW E PLANT SPECIFICN/ACOMPLYNI-6649A-2NOTE 1YES------
A7 CORE EXIT TEMPERATURE TE-1E THRU T CORE EXIT TEMPERATUREA132-2300 F PLANT SPECIFICN/ACOMPLY2 CHANNEL PE RNOTE 1*YESYESNOTES 1A,1B,QSPDS ADISPLAY 'A'A132-2300 FPLANT SPECIFICN/ACOMPLYQSPDS BNOTE 1YES------& 9QSPDS BDISPLAY 'B'A132-2300 FPLANT SPECIFICN/ACOMPLYQSPDS ANOTE 1YES------
A8 CONTAINMENT SUMP WATER LEVELLT-6309ACTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECIFICCOMPLYCOMPLYLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WATER LEVEL IND.A1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYLI-6309BNOTE 1YES------NOTES 1C,1KLR-6308ACTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1CLT-6309BCTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECIFICCOMPLYCOMPLYLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WATER LEVEL IND.A1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYLI-6309ANOTE 1YES------NOTES 1C,1KLR-6308BCTMT. WATER LEVELA1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1C A9 PRESSURIZER WATER LEVELLT-459PRZR LEVEL CH. IA10-100%(150" TO 334")PLANT SPECIFICCOMPLYCOMPLYLT-460;LT-461NOTE 1---YESYESLI-459APRZR LEVEL CH. I IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-460;LI-461NOTE 1YES------LT-460PRZR LEVEL CH. IIA10-100%(150" TO 334")PLANT SPECIFICCOMPLYCOMPLYLT-459;LT-461NOTE 1---YESYESLI-460PRZR LEVEL CH. II IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-459A;LI-461NOTE 1YES------LT-461PRZR LEVEL CH. IIIA10-100%(150" TO 334")PLANT SPECIFICCOMPLYCOMPLYLT-459;LT-460NOTE 1---YESYESLI-461PRZR LEVEL CH. III IND.A10-100%PLANT SPECIFICN/ACOMPLYLI-459A;LI-460NOTE 1YES------LR-459PRZR LEVEL RECORDER FOR LT-459, 460, 461A10-100%
PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------
A10 STEAM GENERATOR PRESSUREPT-474S.G. 'A' STEAM PRESSURE CH. IIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-475;PT-476NOTE 1*YESYESNOTES 1A,1BPI-474S.G. 'A' STEAM PRESSURE CH. II IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-475;PI-476NOTE 1YES------PT-475S.G. 'A' STEAM PRESSURE CH. IIIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-474;PT-476NOTE 1*YESYESNOTES 1A,1BPI-475S.G. 'A' STEAM PRESSURE CH. III IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-474;PI-476NOTE 1YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 4 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONPT-476S.G. 'A' STEAM PRESSURE CH. IVA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-474;PT-475NOTE 1*YESYESNOTES 1A,1BPI-476S.G. 'A' STEAM PRESSURE CH. IV IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-474;PI-475NOTE 1YES------PT-484S.G. 'B' STEAM PRESSURE CH. IIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-485;PT-486NOTE 1*YESYESNOTES 1A,1BPI-484S.G. 'B' STEAM PRESSURE CH. II IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-485;PI-486NOTE 1YES------PT-485S.G. 'B' STEAM PRESSURE CH. IIIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-484;PT-486NOTE 1*YESYESNOTES 1A,1BPI-485S.G. 'B' STEAM PRESSURE CH. III IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-484;PI-486NOTE 1YES------PT-486S.G. 'B' STEAM PRESSURE CH. IVA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-484;PT-485NOTE 1*YESYESNOTES 1A,1BPI-486S.G. 'B' STEAM PRESSURE CH. IV IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-484;PI-485NOTE 1YES------PT-494S.G. 'C' STEAM PRESSURE CH. IIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-495;PT-496NOTE 1*YESYESNOTES 1A,1BPI-494S.G. 'C' STEAM PRESSURE CH. II IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-495;PI-496NOTE 1YES------PT-495S.G. 'C' STEAM PRESSURE CH. IIIA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-494;PT-496NOTE 1*YESYESNOTES 1A,1BPI-495S.G. 'C' STEAM PRESSURE CH. III IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-494;PI-496NOTE 1YES------PT-496S.G. 'C' STEAM PRESSURE CH. IVA10-1400 PSIG PLANT SPECIFICCOMPLYCOMPLYPT-494;PT-495NOTE 1*YESYESNOTES 1A,1BPI-496S.G. 'C' STEAM PRESSURE CH. IV IND.A10-1400 PSIG PLANT SPECIFICN/ACOMPLYPI-494;PI-495NOTE 1YES------
A11 EDG OUTPUT4K4AEMERGENCY DIESEL GENERATOR `4A' OUTPUTA10-4 MEGAWATTSPLANT SPECIFICN/ACOMPLYEDG `4B' OUTP UNOTE 5NOTE 1YESYESNOTES 1A,1B4K4BEMERGENCY DIESEL GENERATOR `4B' OUTPUTA10-4 MEGAWATTSPLANT SPECIFICN/ACOMPLYEDG `4A' OUTP UNOTE 5NOTE 1YESYESNOTES 1A,1B A12 4KV BUS VOLTAGE4AA`4A' 4KV BUS VOLTAGEA10-5000 VOLTS PLANT SPECIFICN/ACOMPLY4B 4KV BUS VO LNOTE 5YES------NOTE 64AB`4B' 4KV BUS VOLTAGEA10-5000 VOLTSPLANT SPECIFICN/ACOMPLY4A 4KV BUS VO LNOTE 5YES------NOTE 6 A13 SAFETY INJECTION PUMP STATUS3AA13SI PUMP '3A' MTR BREAKERA1START-STOP PLANT SPECIFICN/ACOMPLY3AB12;4AA13;4 ANOTE 3**YESYESSI PP. 3A IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLYPP3B,4A/B IND LNOTE 3YES------3AB12SI PUMP '3B' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPLY3AA13;4AA13;4 ANOTE 3**YESYESSI PP. 3B IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLYPP3A,4A/B IND LNOTE 3YES------4AA13SI PUMP '4A' MTR BREAKERA1START-STOPPLANT SPECIFICN/ACOMPLY3AA13;3AB12;4 ANOTE 3**YESYESSI PP. 4A IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLYPP3A/B,4B IND LNOTE 3YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 5 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION4AB12SI PUMP '4B' MTR BREAKERA1START-STOP PLANT SPECIFICN/ACOMPLY3AA13;3AB12;4 ANOTE 3**YESYESSI PP. 4B IND. LIGHTSA1START-STOPPLANT SPECIFICN/ACOMPLYPP3A/B,4A IND LNOTE 3YES------
B1 REACTIVITY CONTROL - NEUTON FLUXND-6649ANEUTRON FLUX DETECTORB11E-8 TO 200% FULL POW E1E-6 TO 100% FULL POWERCOMPLYCOMPLYND-6649BNOTE 1**YESYESNOTES 1A,1BNI-6649A-2NEUTRON FLUX INDICATORB11E-8 TO 200% FULL POW E1E-6 TO 100% FULL POWERN/ACOMPLYNI-6649B-2NOTE 1YES------ND-6649BNEUTRON FLUX DETECTORB11E-8 TO 200% FULL POW E1E-6 TO 100% FULL POWERCOMPLYCOMPLYND-6649ANOTE 1**YESYESNOTES 1A,1BNI-6649B-2NEUTRON FLUX INDICATORB11E-8 TO 200% FULL POW E1E-6 TO 100% FULL POWERN/ACOMPLYNI-6649A-2NOTE 1YES------
B2 REACTIVITY CONTROL - CONTROL ROD POSITION G5CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E9CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J11CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L7CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J5CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E7CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESG11CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L9CONTROL ROD BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F2CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESB10CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESK14CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES P6CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K2CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES B6CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESF14CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESP10CONTROL ROD BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES F4CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESD10CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C26 C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 6 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONK12CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES M6CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K4CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES D6CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESF12CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESM10CONTROL ROD BANK `C' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES D8CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES M8CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H4CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H8CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESH12CONTROL ROD BANK `D' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES G3SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C9SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J13SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES N7SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES J3SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C7SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESG13SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES N9SHUTDOWN BANK `A' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES E5SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESL11SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES L5SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESE11SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES H6SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYESH10SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES C26 01/27/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 7 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION F8SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES K8SHUTDOWN BANK `B' ROD BOTTOM LIGHTSB3FULL IN OR NOT FULL INFULL IN OR NOT FULL INN/AN/AN/AN/AYESYESYES B3 REACTIVITY CONROL - RCS BORON CONCENTRATIONAE-6424BORON ANALYZER RCS SOLUBLE BORON CON CB30-6000 PPM 0-6000 PPMN/AN/AN/AN/A**YESYESNOTE 1B B4 REACTIVITY CONTROL - RCS COLD LEG WTR. TEMP.TE-410ARCS COLD LEG WTR. TEMP. LOOP `A'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTE-410BRCS COLD LEG WTR. TEMP. LOOP `A'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTE-420ARCS COLD LEG WTR. TEMP. LOOP `B'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTE-420BRCS COLD LEG WTR. TEMP. LOOP `B'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTE-430ARCS COLD LEG WTR. TEMP. LOOP `C'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTE-430BRCS COLD LEG WTR. TEMP. LOOP `C'B3 0-750 F 50-400 FN/AN/AN/AN/A---YESYESTR-410RCS COLD LEG TEMP. RECORD LOOP A, B, C FOB3 0-750 F 50-400 FN/AN/AN/AN/AYES------QSPDS ADISPLAY `A'B3 0-750 F 50-400 FN/AN/AN/AN/AYES------QSPDS BDISPLAY `B'B3 0-750 F 50-400 FN/AN/AN/AN/AYES------
B5 CORE COOLING - RCS HOT LEG WTR. TEMP.TE-413ARCS HOT LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 FCOMPLYCOMPLYTE-413BNOTE 1---YESYESTE-413BRCS HOT LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 FCOMPLYCOMPLYTE-413ANOTE 1---YESYESTE-423ARCS HOT LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 FCOMPLYCOMPLYTE-423BNOTE 1---YESYESTE-423BRCS HOT LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 FCOMPLYCOMPLYTE-423ANOTE 1---YESYESTE-433ARCS HOT LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 FCOMPLYCOMPLYTE-433BNOTE 1---YESYESTE-433BRCS HOT LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 FCOMPLYCOMPLYTE-433ANOTE 1---YESYESTR-413RCS HOT LEG TEMP. RECORD LOOP A, B, C FO RB10-750 F 50-700 FN/ACOMPLYN/ANOTE 1YES------QSPDS ADISPLAY `A'B10-750 F 50-700 FN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-750 F 50-700 FN/ACOMPLYQSPDS ANOTE 1YES------
B6 CORE COOLING - RCS COLD LEG WTR. TEMP.TE-410ARCS COLD LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 FCOMPLYCOMPLYTE-410BNOTE 1---YESYESTE-410BRCS COLD LEG WTR. TEMP. LOOP `A'B10-750 F 50-700 FCOMPLYCOMPLYTE-410ANOTE 1---YESYES C26 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 8 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONTE-420ARCS COLD LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 FCOMPLYCOMPLYTE-420BNOTE 1---YESYESTE-420BRCS COLD LEG WTR. TEMP. LOOP `B'B10-750 F 50-700 FCOMPLYCOMPLYTE-420ANOTE 1---YESYESTE-430ARCS COLD LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 FCOMPLYCOMPLYTE-430BNOTE 1---YESYESTE-430BRCS COLD LEG WTR. TEMP. LOOP `C'B10-750 F 50-700 FCOMPLYCOMPLYTE-430ANOTE 1---YESYESTR-410RCS COLD LEG TEMP. RECORD LOOP A, B, C FOB10-750 F 50-700 FN/ACOMPLYN/ANOTE 1YES------QSPDS ADISPLAY `A'B10-750 F 50-700 FN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-750 F 50-700 FN/ACOMPLYQSPDS ANOTE 1YES------
B7 CORE COOLING - RCS PRESSUREPT-404RCS PRESSUREB10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-406NOTE 1*YESYESNOTES 1A,1BPT-406RCS PRESSUREB10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS ANOTE 1YES------
B8 CORE COOLING - CORE EXIT TEMPERATURE TE-1E THRU T CORE EXIT TEMPERATUREB332-2300 F 200-2300 FN/AN/AN/AN/A---YESYESNOTE 9QSPDS ADISPLAY `A'B332-2300 F 200-2300 FN/AN/AN/AN/AYES------QSPDS BDISPLAY `B'B332-2300 F 200-2300 FN/AN/AN/AN/AYES------
B9 CORE COOLING - COOLANT INVENTORY ICCS RVL-A (H REACTOR VESSEL WTR. LVL. CH. `A'B10-100% (CORE TOP/VSL T BTM HOT LEG TO TOP OF VES LN/ACOMPLYRVL-BNOTE 1*YESYESNOTES 1A,1B ICCS RVL-B (H REACTOR VESSEL WTR. LVL. CH. `B'B10-100% (CORE TOP/VSL T BTM HOT LEG TO TOP OF VES LN/ACOMPLYRVL-ANOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-100% (CORE TOP/VSL T BTM HOT LEG TO TOP OF VES LN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'B10-100% (CORE TOP/VSL T BTM HOT LEG TO TOP OF VES LN/ACOMPLYQSPDS ANOTE 1YES------
B10 CORE COOLING - DEGREES OF SUBCOOLINGQSPDS ARCS TEMP. SATURATION MARGIN CH. `A'B2700 TO -2100 F200 SUBCLNG TO 35 SUPRHTN/AN/AN/ANOTE 1YESYESYESQSPDS BRCS TEMP. SATURATION MARGIN CH. `B'B2700 TO -2100 F200 SUBCLNG TO 35 SUPRHTN/AN/AN/ANOTE 1YESYESYES B11 MAINTAINING RCS INTEGRITY - RCS PRESSUREPT-404RCS PRESSUREB10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-406NOTE 1*YESYESNOTES 1A,1BPT-406RCS PRESSUREB10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'B10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS BNOTE 1YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 9 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONQSPDS BDISPLAY `B'B10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS ANOTE 1YES------
B12 MAINTAINING RCS INTEGRITY - CTMT. SUMP WTR. LEVELLT-6308ACTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)COMPLYN/AN/ANOTE 1---YESYESNOTE 1CLI-6308ACTMT. SUMP WTR. LVL. IND.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308ACTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1JLT-6308BCTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)COMPLYN/AN/ANOTE 1---YESYESNOTE 1CLI-6308BCTMT. SUMP WTR. LVL. IND.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308BCTMT. SUMP WTR. LVL.B25"-369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1J B13 MAINTAINING RCS INTEGRITY - CTMT. SUMP WTR. LEVELLT-6309ACTMT. WTR. LVL.B1397"-487" WIDE RANGE (PLANT SPECIFI CCOMPLYCOMPLYLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WTR. LVL. IND.B1397"-487"WIDE RANGE (PLANT SPECIFI CN/ACOMPLYLI-6309BNOTE 1YES------NOTES 1C,1KLR-6308ACTMT. WTR. LVL.B1397"-487"WIDE RANGE (PLANT SPECIFI CN/ACOMPLYN/ANOTE 1YES------NOTE 1CLT-6309BCTMT. WTR. LVL.B1397"-487" WIDE RANGE (PLANT SPECIFI CCOMPLYCOMPLYLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WTR. LVL. IND.B1397"-487"WIDE RANGE (PLANT SPECIFI CN/ACOMPLYLI-6309ANOTE 1YES------NOTES 1C,1KLR-6308BCTMT. WTR. LVLB1397"-487"WIDE RANGE (PLANT SPECIFI CN/ACOMPLYN/ANOTE 1YES------NOTE 1C B14 MAINTAINING RCS INTEGRITY - CTMT. PRESSUREPT-6306ACTMT. PRESS. WIDE RANGEB10-180 PSIG0 PSIG TO DESIGN PRESSUREN/ACOMPLYPT-6306BNOTE 1---YESYESPI-6306ACTMT. PRESS. WIDE RANGE IND.B10-180 PSIG0 PSIG TO DESIGN PRESSUREN/ACOMPLYPI-6306BNOTE 1YES------PR-6306ACTMT. PRESS. WIDE RANGEB10-180 PSIG O PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6306BCTMT. PRESS. WIDE RANGEB10-180 PSIG0 PSIG TO DESIGN PRESSUREN/ACOMPLYPT-6306ANOTE 1---YESYESPI-6306BCTMT. PRESS. WIDE RANGE IND.B10-180 PSIG O PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6306ANOTE 1YES------PR-6306BCTMT. PRESS. WIDE RANGEB10-180 PSIG0 PSIG TO DESIGN PRESSUREN/ACOMPLYN/ANOTE 1YES------
B15 MAINTAINING CTMT. INTEGRITY - CTMT. ISO VAVLE POS.
MOV-744A (L S RHR TO COLD LEG I.C.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2**YESYESNOTE1A,B,H,2 MOV-744A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-744B (L S RHR TO COLD LEG I.C.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2**YESYESNOTE1A,B,H,2 MOV-744B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 10 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION MOV-716B (L S RCP THERMAL BARRIER CCWB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-716B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-626 (LS
)RCP A, B, C THERMAL BARRIER COOLING WTR.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2MOV-626 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-730 (LS
)CCW FROM RCP A, B, C COOLER BEARINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2MOV-730 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-739 (LS)EXCESS LETDOWN HEAT EXCHANGERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2CV-739 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-1417 (L S CCW TO NORMAL CTMT. COOLINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1417 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-1418 (L S CCW FROM NORMAL CTMT. COOLINGB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1418 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-200A (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-200A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-200B (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-200B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-200C (LS)LETDOWN LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-200C IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-204 (LS)LETDOWN LINE LOW PRESS.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-204 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-381 (LS
)RCP SEAL WTR. RETURN VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-381 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
CV-4658A (L S RCDT VENT VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 11 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION CV-4658A STATUS LIGHT ON PHASE A ISOL. BO AB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-4658B (L S RCDT VENT VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4658B STATUS LIGHT ON PHASE A ISOL. BO AB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-4668A (L S RCDT DISCH. TO HOLD-UP TANKB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4668A STATUS LIGHT ON PHASE A ISOL. BO AB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-4668B (L S RCDT DISCH. TO HOLD-UP TANKB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-4668B STATUS LIGHT ON PHASE A ISOL. BO AB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-4659A (L S RCDT LINE TO HYDROGEN ANAL.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-4659A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-4659B (L S RCDT LINE TO HYDROGEN ANAL.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-4659B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-880A (L S CTMT. SPRAY PUMP A DISCH. VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2MOV-880A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-880B (L S CTMT. SPRAY PUMP B DISCH. VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2MOV-880B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-956A (LS)PRZR STM. SPACE SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11CV-956A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-956B (LS)PRZR LIQUID SPACE SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11CV-956B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------SV-6428 (LS)HOT LEG RCS SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11SV-6428 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------SV-2912 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2,11SV-2912 IND. LIGHTSB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------SV-2911 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2,11SV-2911 IND. LIGHTSB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 12 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONSV-2913 (LS)CTMT. AIR SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2,11 SV-2913 IND. LIGHTSB1CLOSED/NOT CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------CV-519A (LS)DEMIN. WTR. TO PRZR RELIEF TNK.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-519A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2600 (L S CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2600 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2601 (L S CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2601 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2602 (L S CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2602 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2603 (L S CTMT. PURGEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2603 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2604 (L S MAIN STM. MSIV S.G. `A'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 POV-2604 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-1403 (L S MAIN STM. LINE `A'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2MOV-1403 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
POV-2605 (L S MAIN STM. MSIV S.G. `B'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 POV-2605 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-1404 (L S MAIN STM. LINE `B'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2MOV-1404 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
POV-2606 (L S MAIN STM. MSIV S.G. `C'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 POV-2606 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-1405 (L S MAIN STM. LINE `C'B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 MOV-1405 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------FCV-478 (LS)S.G. `A' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2 9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 13 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION FCV-478 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------FCV-488 (LS)S.G. `B' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2 FCV-488 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------FCV-498 (LS)S.G. `C' FEEDWATERB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1---YESYESNOTES 1H,2 FCV-498 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------CV-2816AUX. FEEDWATER TO S.G. `A'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1401AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1ECV-2831AUX. FEEDWATER TO S.G. `A'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1401BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1ECV-2817AUX. FEEDWATER TO S.G. `B'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1457AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1ECV-2832AUX. FEEDWATER TO S.G. `B'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1457BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1ECV-2818AUX. FEEDWATER TO S.G. `C'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1458AHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1ECV-2833AUX. FEEDWATER TO S.G. `C'B10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2HIC-1458BHAND INDICATING CONTROLLERB10-300 GPMCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 1YES------NOTE 1E CV-6275A (L S S.G. `A' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-6275A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-6275B (L S S.G. `B' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-6275B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
CV-6275C (L S S.G. `C' BLOWDOWNB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2 CV-6275C IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-1427 (L S S.G. `A' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2,11MOV-1427 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 14 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION MOV-1426 (L S S.G. 'B' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2,11 MOV-1426 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-1425 (L S S.G. `C' BLOWDOWN SAMPLEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2---YESYESNOTES 1H,2,11MOV-1425 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2903 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COO LB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2903 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2904 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COO LB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2904 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2905 (LS)CCW INLET VALVE TO EMERGENCY CTMT. COO LB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2905 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2810 (LS)CCW FROM EMERGENCY CTMT. COOLER B BYP AB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2810 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2906 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2906 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2812 (LS)CCW FROM EMERGENCY CTMT. COOLER C BYP AB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2812 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2907 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2907 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2814 (LS)CCW FROM EMERGENCY CTMT. COOLER A BYP AB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2814 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2908 (LS)CCW OUTLET VALVE FROM EMERGENCY CTMT. B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 CV-2908 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-872 (LS
)LOW HEAD SAFETY INJECT.B1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-872 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-855 (LS)NITROGEN SUPPLY TO ACCUMULATORSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 15 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION CV-855 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-956D (LS)ACCUMULATOR SAMPLE LINEB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-956D IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-843A (L S BORON INJ. TANK OUT STOP VALVEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-843A IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------
MOV-843B (L S BORON INJ. TANK OUT STOP VALVEB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-843B IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-2821 (LS)CTMT. SUMP DISCH.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2821 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-2822 (LS)CTMT. SUMP DISCH.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2822 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-2819 (LS)INST. AIR BLEEDB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2819 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------CV-2826 (LS)INST. AIR BLEEDB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-2826 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
MOV-6386 (L S RCP SEALB1OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYCOMPLYN/ANOTE 2---YESYESNOTES 1H,2 MOV-6386 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 2YES------CV-516 (LS)GAS ANALYZER SAMPLE VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2,11 CV-516 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------SV-6385 (LS)GAS ANALYZER SAMPLE VLV.B1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3---YESYESNOTES 1H,2 SV-6385 IND. LIGHTSB1OPEN/CLOSEDCLOSED/NOT CLOSEDN/ACOMPLYN/ANOTE 3YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 16 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION B16 MAINTAINING CTMT. INTEGRITY - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESSUREB10-180 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.B10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.B10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6306BCTMT. WIDE RANGE PRESS.B10-180 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6306ANOTE 1---YESYESPI-6306BCTMT. WIDE RANGE PRESS. IND.B10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6306ANOTE 1YES------PR-6306BCTMT. WIDE RANGE PRESSB10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6425ACTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6425BNOTE 1---YESYESPI-6425ACTMT. NARROW RANGE PRESS. IND.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6425BNOTE 1YES------PR-6306ACTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6425BCTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6425ANOTE 1---YESYESPI-6425BCTMT. NARROW RANGE PRESS. IND.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6425ANOTE 1YES------PR-6306BCTMT. NARROW RANGE PRESS.B1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------
C1 FUEL CLADDING - CORE EXIT TEMPERATURE TE-1E THRU T CORE EXIT TEMPERATUREC132-2300 F 200-2300 FN/ACOMPLY2 CHANNEL PE RNOTE 1*YESYESNOTES 1A,1B,QSPDS ADISPLAY `A'C132-2300 F200-2300 FN/ACOMPLYQSPDS BNOTE 1YES------& 9QSPDS BDISPLAY `B'C132-2300 F200-2300 FN/ACOMPLYQSPDS ANOTE 1YES------
C2 FUEL CLADDING - RADIOACTIVITY CONCENTRATION OR RADIATION LEVEL IN CIRCULATING PRIMARY COOLANTNONERADIOACTIVITY CONCENTRATION OR RADIATIO NC1GRAB SAMPLE1/2 TO 100 X T.S. LIMIT---------------------NOTE 3 C3 FUEL CLADDING - ANALYSIS OF PRIMARY COOLANTNONERx COOL WATER RADIOACTIVITY ANALYSISC3GRAB SAMPLE10 micro Ci/ml to 10 Ci/ml---------------------NOTE 12 C4 Rx COOLANT PRESSURE BOUNDARY - RCS PRESSUREPT-404RCS PRESS.C10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-406NOTE 1*YESYESNOTES 1A,1BPT-406RCS PRESS.C10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-404NOTE 1*YESYESNOTES 1A,1BQSPDS ADISPLAY `A'C10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'C10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS ANOTE 1YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 17 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION C5 Rx COOLANT PRESSURE BOUNDARY - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6306ANOTE 1---YESYESPI-6306BCTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6306ANOTE 1YES------PR-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------
C6 Rx COOLANT PRESSURE BOUNDARY - CTMT. PRESSUREPT-6425ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG
-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6425BNOTE 1---YESYESPI-6425ACTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6425BNOTE 1YES------PR-6306ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------PT-6425BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPT-6425ANOTE 1---YESYESPI-6425BCTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYPI-6425ANOTE 1YES------PR-6306BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO DESIGN PRESSUR EN/ACOMPLYN/ANOTE 1YES------
C7 Rx COOLANT PRESSURE BOUNDARY - CTMT. SUMP WTR. LEVELLT-6308ACTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)COMPLYN/AN/ANOTE 1---YESYESNOTE 1CLI-6308ACTMT. SUMP WATER LEVEL IND.C25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308ACTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1JLT-6308BCTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)COMPLYN/AN/ANOTE 1---YESYESNOTE 1CLI-6308BCTMT. SUMP WATER LEVEL IND.C25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1ILR-6308BCTMT. SUMP WATER LEVELC25" TO 369"NARROW RANGE (SUMP)N/AN/AN/ANOTE 1YES------NOTES 1C,1J C8 Rx COOLANT PRESSURE BOUNDARY - CTMT. SUMP WTR. LEVELLT-6309ACTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECIFICCOMPLYCOMPLYLT-6309BNOTE 1---YESYESNOTE 1CLI-6309ACTMT. WATER LEVEL IND.C1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYLI-6309BNOTE 1YES------NOTES 1C,1KLR-6308ACTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1CLT-6309BCTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECIFICCOMPLYCOMPLYLT-6309ANOTE 1---YESYESNOTE 1CLI-6309BCTMT. WATER LEVEL IND.C1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYLI-6309ANOTE 1YES------NOTES 1C,1K 9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 18 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLR-6308BCTMT. WATER LEVELC1397" TO 487"WIDE RANGE PLANT SPECIFICN/ACOMPLYN/ANOTE 1YES------NOTE 1C C9 Rx COOLANT PRESSURE BOUNDARY - CTMT. AREA RADIATIONRAD-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A'C31 TO 1E8 R/Hr 1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESRAI-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A' IND.C31 TO 1E8 R/Hr 1 TO 1E4 R/HrN/AN/AN/AN/AYES------RAD-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B'C31 TO 1E8 R/Hr 1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESRAI-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B' IND.C31 TO 1E8 R/Hr 1 TO 1E4 R/HrN/AN/AN/AN/AYES------
C10 Rx COOLANT PRESS. BOUNDARY EFFLUENT RADIOACTIVITY - NOBLE GAS EFFLUENT FROM COND. AIR REMOVAL SYS. EXH.RAD-6417AIR EJECTOR CONDENSER EXHAUSTC31E-7 TO 1E5 micro Ci/CC1E-6 TO 1E-2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1B C11 CONTAINMENT - RCS PRESSUREPT-404RCS PRESSUREC10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-406NOTE 1*YESYESNOTE 1BPT-406RCS PRESSUREC10-3000 PSIG 0-3000 PSIGCOMPLYCOMPLYPT-404NOTE 1*YESYESNOTE 1BQSPDS ADISPLAY `A'C10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS BNOTE 1YES------QSPDS BDISPLAY `B'C10-3000 PSIG 0-3000 PSIGN/ACOMPLYQSPDS ANOTE 1YES------
C12 CONTAINMENT - CTMT. HYDROGEN CONCENTRATIONAE-6307ACTMT. HYDROGEN MONITORC30-20%0-10 VOL.%N/AN/AN/AN/A---YESYESAI-6307ACTMT. HYDROGEN INDICATORC30-20%0-10 VOL.%N/AN/AN/AN/AYES------RAR-6311ACTMT. HYDROGEN RECORDERC30-10%0-10 VOL.%N/AN/AN/AN/AYES------AE-6307BCTMT. HYDROGEN MONITORC30-20%0-10 VOL.%N/AN/AN/AN/A---YESYESAI-6307BCTMT. HYDROGEN INDICATORC30-20%0-10 VOL.%N/AN/AN/AN/AYES------RAR-6311BCTMT. HYDROGEN RECORDERC30-10%0-10 VOL.%N/AN/AN/AN/AYES------
C13 CONTAINMENT - CTMT. PRESSUREPT-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPT-6306BNOTE 1---YESYESPI-6306ACTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPI-6306BNOTE 1YES------PR-6306ACTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYN/ANOTE 1YES------PT-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPT-6306ANOTE 1---YESYESPI-6306BCTMT. WIDE RANGE PRESS. IND.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPI-6306ANOTE 1YES------PR-6306BCTMT. WIDE RANGE PRESS.C10-180 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYN/ANOTE 1YES------PT-6425ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPT-6425BNOTE 1---YESYES C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 19 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONPI-6425ACTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPI-6425BNOTE 1YES------PR-6306ACTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYN/ANOTE 1YES------PT-6425BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPT-6425ANOTE 1---YESYESPI-6425BCTMT. NARROW RANGE PRESS. IND.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYPI-6425ANOTE 1YES------PR-6306BCTMT. NARROW RANGE PRESS.C1-6 TO +18 PSIG-5 PSIG TO 3X DESIGN PRESSN/ACOMPLYN/ANOTE 1YES------
C14 CONTAINMENT EFFLUENT RADIOACTIVITY NOBLE GAS FROM IDENTIFIED RELEASE POINTSRAD-6304VENT STACK WIDE RANGE MONITORC21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E-2 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1BRAD-6417AIR EJECTOR CONDENSER EXH.C21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E-2 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1B C15 CONTAINMENT EFFLUENT RADIOACTIVITY NOBLE GAS (FROM BUILDINGS OR AREAS, ETC.)RAD-6304VENT STACK WIDE RANGE MONITORC21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E3 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1B D1 RHR SYSTEM - RHR SYSTEM FLOWFT-605RHR SYSTEM FLOWD20-8500 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-605RHR SYSTEM FLOW INDICATORD20-8500 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D2 RHR SYSTEM - RHR Hx OUTLET TEMPERATURETE-606RHR Hx OUTLET TEMPERATURED250-400 F 40-350 FCOMPLYN/AN/ANOTE 1---YESYESNOTE 3TR-604RHR Hx OUTLET TEMPERATURE RECORDERD250-400 F 40-350 FN/AN/AN/ANOTE 1YES------NOTE 3 D3 S.I.S. ACCUMULATOR TANK LEVELLT-920ACCUMULATOR TANK LEVEL `A'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3LI-920ACCUMULATOR TANK LEVEL`A' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3LT-922ACCUMULATOR TANK LEVEL `A'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3LI-922ACCUMULATOR TANK LEVEL `A' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3LT-924ACCUMULATOR TANK LEVEL `B'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3LI-924ACCUMULATOR TANK LEVEL `B' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3LT-926ACCUMULATOR TANK LEVEL `B'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3LI-926ACCUMULATOR TANK LEVEL `B' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3LT-928ACCUMULATOR TANK LEVEL `C'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3LI-928ACCUMULATOR TANK LEVEL `C' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3LT-930ACCUMULATOR TANK LEVEL `C'D36400-6870 GAL10% TO 90% VOLUMECOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 20 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLI-930ACCUMULATOR TANK LEVEL `C' IND.D36400-6870 GAL10% TO 90% VOLUMEN/AN/AN/ANOTE 1YES------NOTES 1D,3PT-921ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3PI-921ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-923ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-923ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-925ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-925ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-927ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-927ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-929ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-929ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 PT-931ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG0-750 PSIGCOMPLYN/AN/ANOTE 1---YESYESNOTES 1D,3 PI-931ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG0-750 PSIGN/AN/AN/ANOTE 1YES------NOTES 1D,3 D4 S.I.S ACCUMULATOR TANK PRESSUREPT-921ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-921ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------PT-923ACCUMULATOR TANK PRESSURE `A'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-923ACCUMULATOR TANK PRESSURE `A' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------PT-925ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-925ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------PT-927ACCUMULATOR TANK PRESSURE `B'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-927ACCUMULATOR TANK PRESSURE `B' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------PT-929ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-929ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG O-750 PSIGN/AN/AN/ANOTE 1YES------PT-931ACCUMULATOR TANK PRESSURE `C'D30-800 PSIG 0-750 PSIGCPMLYN/AN/ANOTE 1---YESYESPI-931ACCUMULATOR TANK PRESSURE `C' IND.D30-800 PSIG 0-750 PSIGN/AN/AN/ANOTE 1YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 21 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION D5 S.I.S ACCUMULATOR ISOLATION VALVE POSITION MOV-865A (L SACCUMULATOR TANK ISOLATION VALVE `A'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYES MOV-865A IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------
MOV-865B (L SACCUMULATOR TANK ISOLATION VALVE 'B'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYES MOV-865B IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------
MOV-865C (L SACCUMULATOR TANK ISOLATION VALVE `C'D2CLOSED OR OPENCLOSED OR OPENNOTE 1N/AN/ANOTE 2---YESYES MOV-865C IND. LIGHTSD2CLOSED OR OPENCLOSED OR OPENN/AN/AN/ANOTE 2YES------
D6 S.I.S. BORIC ACID CHARGING FLOWFT-943BORIC ACID CHARGING FLOWD20-1000 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-943BORIC ACID CHARGING FLOW IND.D20-1000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D7 S.I.S. FLOW IN HPI SYSTEMFT-940HPI SYSTEM FLOWD20-1000 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-940HPI SYSTEM FLOW IND.D20-1000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D8 S.I.S. FLOW IN LPI SYSTEMFT-605LPI SYSTEM FLOWD20-8500 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-605LPI SYSTEM FLOW IND.D20-8500 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D9 S.I.S. REFUELING WATER STORAGE TANKLT-6583ARWST LEVELD20-335,000 GAL TOP TO BOTTOMN/AN/AN/ANOTE 1---YESYESLI-6583ARWST LEVEL INDICATORD20-335,000 GAL TOP TO BOTTOMN/AN/AN/ANOTE 1YES------LT-6583BRWST LEVELD20-335,000 GAL TOP TO BOTTOMN/AN/AN/ANOTE 1---YESYESLI-6583BRWST LEVEL INDICATORD20-335,000 GALTOP TO BOTTOMN/AN/AN/ANOTE 1YES------
D10 PRIMARY COOLANT SYSTEM - RCP MOTOR STATUS3P200ARCP `A' MTR. CURRENT INDICATORD30-1200 AMP MTR. CURRENTN/AN/AN/AN/AYESYESYES3P200BRCP `B' MTR. CURRENT INDICATORD30-1200 AMP MTR. CURRENTN/AN/AN/AN/AYESYESYES3P200CRCP `C' MTR. CURRENT INDICATORD30-1200 AMP MTR. CURRENTN/AN/AN/AN/AYESYESYES D11 PRIMARY COOLANT SYSTEM - PRIMARY SYSTEM SAFETY RELIEF VALVE POSITION PCV-455C (L S PRZR PORV POSITIOND2OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 3---YESYESPCV-456 (LS)PRZR PORV POSITIOND2OPEN/CLOSEDCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 3---YESYES POSITION INDICATION LIGHTS FOR PCV 455C &
4D2OPEN/CLOSEDCLOSED/NOT CLOSEDN/AN/AN/ANOTE 3YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 22 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONZS-6303APRIMARY SYSTEM SAFETY R.V. CODE SAFETY VD20-100%CLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESZS-6303BPRIMARY SYSTEM SAFETY R.V. CODE SAFETY VD20-100%CLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESZS-6303CPRIMARY SYSTEM SAFETY R.V. CODE SAFETY VD20-100%CLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYES LIGHT IND. FOR ZS-6306A,B,CD20-100%CLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------
D12 PRIMARY COOLANT SYSTEM - PRESSURIZER LEVELLT-459PRZR LEVEL CH. ID10-100%(150" TO 334")TOP TO BOTTOMCOMPLYCOMPLYLT-460;LT-461NOTE 1---YESYESLI-459APRZR LEVEL CH.I IND.D10-100%TOP TO BOTTOMN/ACOMPLYLI-460;LI-461NOTE 1YES------LT-460PRZR LEVEL CH. IID10-100%(150" TO 334")TOP TO BOTTOMCOMPLYCOMPLYLT-459;LT-461NOTE 1---YESYESLI-460PRZR LEVEL CH. II IND.D10-100%TOP TO BOTTOMN/ACOMPLYLI-459A;LI-461NOTE 1YES------LT-461PRZR LEVEL CH. IIID10-100%(150" TO 334")TOP TO BOTTOMCOMPLYCOMPLYLT-459;LT-460NOTE 1---YESYESLI-461PRZR LEVEL CH. III IND.D10-100%TOP TO BOTTOMN/ACOMPLYLI-459A;LI-460NOTE 1YES------LR-459PRZR LEVEL RECORDER FOR LT-459, 460, 461D10-100%
TOP TO BOTTOMN/ACOMPLYN/ANOTE 1YES------
D13 PRIMARY COOLANT SYSTEM - PRESSURIZER HEATER STATUS3B11PRZR HEATER STATUS CONTROL GROUPD20-500 KW CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B3B12PRZR HEATER STATUS BACKUP GROUP 3AD20-600 AMP CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B3B13PRZR HEATER STATUS BACKUP GROUP 3BD20-600 AMP CURRENTN/AN/AN/ANOTE 5**YESYESNOTE 1B D14 PRIMARY COOLANT SYSTEM - QUENCH TANK LEVELLT-470PRZR RELIEF TANK LEVEL (QUENCH TANK)D30-100%(0-100")
TOP TO BOTTOMN/AN/AN/AN/A---YESYESLI-470PRZR RELIEF TANK LEVEL INDICATOR (QUENCHD30-100%
TOP TO BOTTOMN/AN/AN/AN/AYES------
D15 PRIMARY COOLANT SYSTEM - QUENCH TANK TEMPERATURETE-471PRZR RELIEF TANK TEMPERATURED350-350 F 50-750 FN/AN/AN/AN/A---YESYESNOTE 3TI-471PRZR RELIEF TANK TEMPERATURE INDICATORD350-350 F 50-750 FN/AN/AN/AN/AYES------NOTE 3 D16 PRIMARY COOLANT SYSTEM - QUENCH TANK PRESSUREPT-472PRZR RELIEF TANK PRESSURED30-120 PSIG0-DESIGN PRESSUREN/AN/AN/AN/A---YESYESPI-472PRZR RELIEF TANK PRESSURE INDICATORD30-120 PSIG0-DESIGN PRESSUREN/AN/AN/AN/AYES------
D17 SECONDARY SYSTEM (STEAM GEN.) - S. G. LEVELLT-474S.G. `A' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-475;LT-476NOTE 1---YESYESNOTE 1LLI-474S.G. `A' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-475;LI-476NOTE 1YES------NOTE 1L C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 23 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLT-475S.G. `A' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-474;LT-476NOTE 1---YESYESNOTE 1LLI-475S.G. `A' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-474;LI-476NOTE 1YES------NOTE 1LLT-476S.G. `A' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-474;LT-475NOTE 1---YESYESNOTE 1LLI-476S.G. `A' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-474;LI-475NOTE 1YES------NOTE 1LFR-478S.G. `A' LVL CH. I, II, III NARROW RANGE RECOR DD10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYN/ANOTE 1YES------NOTE 1LLT-484S.G. `B' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-485;LT-486NOTE 1---YESYESNOTE 1LLI-484S.G. `B' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-485;LI-486NOTE 1YES------NOTE 1LLT-485S.G. `B' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-484;LT-486NOTE 1---YESYESNOTE 1LLI-485S.G. `B' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-484;LI-486NOTE 1YES------NOTE 1LLT-486S.G. `B' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-484;LT-485NOTE 1---YESYESNOTE 1LLI-486S.G. `B' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-484;LI-485NOTE 1YES------NOTE 1LFR-488S.G. `B' LVL CH. I, II, III NARROW RANGE RECOR DD10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYN/ANOTE 1YES------NOTE 1LLT-494S.G. `C' LVL CH. I NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-495;LT-496NOTE 1---YESYESNOTE 1LLI-494S.G. `C' LVL CH. I NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-495;LI-496NOTE 1YES------NOTE 1LLT-495S.G. `C' LVL CH. II NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-494;LT-496NOTE 1---YESYESNOTE 1LLI-495S.G. `C' LVL CH. II NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-494;LI-496NOTE 1YES------NOTE 1LLT-496S.G. `C' LVL CH. III NARROW RANGED10-100%(30.1" TO 138.22")FROM TUBE SH TO SEPARAT OCOMPLYCOMPLYLT-494;LT-495NOTE 1---YESYESNOTE 1LLI-496S.G. `C' LVL CH. III NARROW RANGE IND.D10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYLI-494;LI-495NOTE 1YES------NOTE 1LFR-498S.G. `C' LVL CH. I, II, III NARROW RANGE RECOR DD10-100%FROM TUBE SH TO SEPARAT ON/ACOMPLYN/ANOTE 1YES------NOTE 1L D18 SECONDARY SYSTEM (STEAM GEN.) - S. G. PRESSUREPT-474S.G. `A' STEAM PRESSURE CH. IID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-474S.G. `A' STEAM PRESSURE CH. II IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-475S.G. `A' STEAM PRESSURE CH. IIID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-475S.G. `A' STEAM PRESSURE CH. III IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-476S.G. `A' STEAM PRESSURE CH. IVD20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-476S.G. `A' STEAM PRESSURE CH. IV IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------
9/24/04 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 24 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONPT-484S.G. `B' STEAM PRESSURE CH. IID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-484S.G. `B' STEAM PRESSURE CH. II IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-485S.G. `B' STEAM PRESSURE CH. IIID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-485S.G. `B' STEAM PRESSURE CH. III IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-486S.G. `B' STEAM PRESSURE CH. IVD20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-486S.G. `B' STEAM PRESSURE CH. IV IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-494S.G. `C' STEAM PRESSURE CH. IID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-494S.G. `C' STEAM PRESSURE CH. II IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-495S.G. `C' STEAM PRESSURE CH. IIID20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-495S.G. `C' STEAM PRESSURE CH. III IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------PT-496S.G. `C' STEAM PRESSURE CH. IVD20-1400 PSIGPa TO 20% ABV MIN SV SETCOMPLYN/AN/ANOTE 1---YESYESPI-496S.G. `C' STEAM PRESSURE CH. IV IND.D20-1400 PSIGPa TO 20% ABV MIN SV SETN/AN/AN/ANOTE 1YES------
D19 SECONDARY SYSTEM (STEAM GEN.) - SAFETY/RELIEF VALVE POSITIONS OR MAIN STEAM FLOWFT-474S.G. 'A' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-474S.G. 'A' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------FT-475S.G. 'A' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-475S.G. 'A' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------FT-484S.G. 'B' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-484S.G. 'B' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------FT-485S.G. 'B' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-485S.G. 'B' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------FT-494S.G. 'C' STEAM FLOW CH.IIID20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-494S.G. 'C' STEAM FLOW CH.III IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------FT-495S.G. 'C' STEAM FLOW CH.IVD20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDCOMPLYN/AN/ANOTE 1---YESYESFI-495S.G. 'C' STEAM FLOW CH.IV IND.D20 TO 5E6 LBS/HrCLOSED/NOT CLOSEDN/AN/AN/ANOTE 1YES------
D20 SECONDARY SYSTEM (STEAM GEN.) - SAFETY/RELIEF VALVE POSITIONS OR MAIN STEAM FLOWFT-476S.G. `A' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYES C26* DCS(SPDS)
Revised 04/17/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 25 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONFI-476S.G. `A' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------FT-477S.G. `A' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYESFI-477S.G. `A' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------FT-486S.G. `B' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYESFI-486S.G. `B' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------FT-487S.G. `B' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYESFI-487S.G. `B' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------FT-496S.G. `C' F.W. FLOW CH. IVD30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYESFI-496S.G. `C' F.W. FLOW CH. IV IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------FT-497S.G. `C' F.W. FLOW CH. IIID30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/A---YESYESFI-497S.G. `C' F.W. FLOW CH. III IND.D30 TO 5E6 LBS/Hr0-110% DESIGN FLOWN/AN/AN/AN/AYES------
D21 AUXILIARY FEEDWATER - AUXILIARY FEEDWATER FLOWFT-1401AAUX. F.W. FLOW TO S.G. `A'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1401AAUX. F.W. FLOW TO S.G. `A' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1401BAUX. F.W. FLOW TO S.G. `A'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1401BAUX. F.W. FLOW TO S.G. `A' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1457AAUX. F.W. FLOW TO S.G. `B'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1457AAUX. F.W. FLOW TO S.G. `B' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1457BAUX. F.W. FLOW TO S.G. `B'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1457BAUX. F.W. FLOW TO S.G. `B' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1458AAUX. F.W. FLOW TO S.G. `C'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1458AAUX. F.W. FLOW TO S.G. `C' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-1458BAUX. F.W. FLOW TO S.G. `C'D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESHIC-1458BAUX. F.W. FLOW TO S.G. `C' IND.D20-300 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D22 AUXILIARY FEEDWATER - CONDENSATE STORAGE TANK WATER LEVELLT-6384ACONDENSATE STORAGE TANKD10-100%(19" TO 583")PLANT SPECIFICN/ACOMPLYLT-6384BNOTE 1*YESYESNOTES 1A,1B C26 C26* DCS(SPDS)
Revised 04/17/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 26 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONLI-6384ACONDENSATE STORAGE TANK IND.D10-100% (O - 250K GAL)PLANT SPECIFICN/ACOMPLYLI-6384BNOTE 1YES------LT-6384BCONDENSATE STORAGE TANKD10-100% (19" TO 583")PLANT SPECIFICN/ACOMPLYLT-6384ANOTE 1*YESYESNOTES 1A,1BLI-6384BCONDENSATE STORAGE TANK IND.D10-100% (O - 250K GAL)PLANT SPECIFICN/ACOMPLYLI-6384ANOTE 1YES------
D23 CONTAINMENT COOLING SYSTEM CONTAINMENT SPRAY FLOWCONTAINMENT SPRAY PP. A IND. LIGHTSD2START-STOP0-110% DESIGN FLOWN/AN/AN/ANOTE 3YESYESYESNOTES 1F,3 MOV-880A (L SCONTAINMENT SPRAY PP. A OUTLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-880A IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3 MOV-864A (L SCONTAINMENT SPRAY PP. A INLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-864A IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3CONTAINMENT SPRAY PP. B IND. LIGHTSD2START-STOP0-110% DESIGN FLOWN/AN/AN/ANOTE 3YESYESYESNOTES 1F,3 MOV-880B (L SCONTAINMENT SPRAY PP. B OUTLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-880B IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3 MOV-864B (L SCONTAINMENT SPRAY PP. B INLET VALVED2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2---YESYESNOTES 1F,3 MOV-864B IND. LIGHTSD2OPEN/CLOSED0-110% DESIGN FLOWN/AN/AN/ANOTE 2YES------NOTES 1F,3 D24 CONTAINMENT COOLING SYSTEM - HEAT REMOVAL BY THE CTMT. FAN HEAT REMOVAL SYSTEMCV-2905 (LS)EMERGENCY CONTAINMENT COOLER A INLET VD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G CV-2905 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2904 (LS)EMERGENCY CONTAINMENT COOLER C INLET VD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G CV-2904 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2903 (LS)EMERGENCY CONTAINMENT COOLER B INLET VD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G CV-2903 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2908 (LS)EMERGENCY CONTAINMENT COOLER A OUTLE TD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G CV-2908 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2907 (LS)EMERGENCY CONTAINMENT COOLER C OUTLE TD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G CV-2907 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1GCV-2906 (LS)EMERGENCY CONTAINMENT COOLER B OUTLE TD2OPEN/CLOSED PLANT SPECIFICCOMPLYN/AN/ANOTE 2---YESYESNOTE 1G C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 27 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION CV-2906 IND. LIGHTSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 2YES------NOTE 1G EMERGENCY CONTAINMENT COOLER FAN A IN DD2START-STOP PLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1G EMERGENCY CONTAINMENT COOLER FAN B IN DD2START-STOP PLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1G EMERGENCY CONTAINMENT COOLER FAN C IN DD2START-STOP PLANT SPECIFICN/AN/AN/ANOTE 2YESYESYESNOTE 1GFT-613ACCW HEADER FLOWD20-14000 GPM PLANT SPECIFICN/AN/AN/ANOTE 1---YESYESNOTE 1GFI-613ACCW HEADER FLOW IND.D20-14000 GPM PLANT SPECIFICN/AN/AN/ANOTE 1YES------NOTE 1GFT-613BCCW HEADER FLOWD20-14000 GPM PLANT SPECIFICN/AN/AN/ANOTE 1---YESYESNOTE 1GFI-613BCCW HEADER FLOW IND.D20-14000 GPM PLANT SPECIFICN/AN/AN/ANOTE 1YES------NOTE 1G D25 CONTAINMENT COOLING SYSTEM - CTMT. ATMOS. TEMPERATURETE-6700CTMT. ATMOS. TEMPERATURED20-300 F 40-400 FCOMPLYN/AN/ANOTE 1*YESYESNOTES 1B,3TE-6701CTMT. ATMOS. TEMPERATURED20-300 F 40-400 FCOMPLYN/AN/ANOTE 1*YESYESNOTES 1B,3TE-6702CTMT. ATMOS. TEMPERATURED20-300 F 40-400 FCOMPLYN/AN/ANOTE 1*YESYESNOTES 1B,3 D26 CONTAINMENT COOLING SYSTEM - CTMT. SUMP WTR. TEMPERATURETE-604ARHR HX A INLET TEMPERATURED250-400 F 50-250 FCOMPLYN/AN/ANOTE 1---YESYESNOTE 3TE-604BRHR HX B INLET TEMPERATURED250-400 F 50-250 FCOMPLYN/AN/ANOTE 1---YESYESNOTE 3TR-604RHR HX INLET TEMPERATURE RECORDERD250-400 F 50-250 FN/AN/AN/ANOTE 1YES------NOTE 3 D27 CHEMICAL & VOLUME CONTROL SYSTEM - MAKEUP FLOWFT-122CHARGING FLOWD20-150 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-122ACHARGING FLOW IND.D20-150 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D28 CHEMICAL & VOLUME CONTROL SYSTEM - LETDOWN FLOWFT-150LO PRESSURE LETDOWN FLOWD20-150 GPM0-110% DESIGN FLOWCOMPLYN/AN/ANOTE 1---YESYESFI-150LO PRESSURE LETDOWN FLOW IND.D20-150 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D29 CHEMICAL & VOLUME CONTROL SYSTEM - VOLUME CONTROL TANK LEVELLT-115VOL. CONTROL TANK LEVELD20-100%(0-59")
TOP TO BOTTOMN/AN/AN/ANOTE 1---YESYESLI-115VOL. CONTROL TANK LEVEL IND.D20-100%TOP TO BOTTOMN/AN/AN/ANOTE 1YES------
D30 COOLING WATER SYSTEM - COMPONENT COOLING WATER TEMP. TO ESF SYSTEMTE-607ACOMPONENT COOLING Hx OUTLET TEMPERATUD250-200 F 40-200 FN/AN/AN/ANOTE 1---YESYESNOTE 3TI-607ACOMPONENT COOLING Hx OUTLET TEMPERATUD250-200 F 40-200 FN/AN/AN/ANOTE 1YES------NOTE 3 C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 28 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONTE-607BCOMPONENT COOLING Hx OUTLET TEMPERATUD250-200 F 40-200 FN/AN/AN/ANOTE 1---YESYESNOTE 3TI-607BCOMPONENT COOLING Hx OUTLET TEMPERATUD250-200 F 40-200 FN/AN/AN/ANOTE 1YES------NOTE 3 D31 COOLING WATER SYSTEM - COMPONENT COOLING WATER FLOW TO ESF SYSTEMFT-613ACCW HEADER FLOWD20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESFI-613ACCW HEADER FLOW IND.D20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------FT-613BCCW HEADER FLOWD20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1---YESYESFI-613BCCW HEADER FLOW IND.D20-14,000 GPM0-110% DESIGN FLOWN/AN/AN/ANOTE 1YES------
D32 RADWASTE SYSTEMS - HIGH LEVEL RADIOACTIVITY LIQUID TANK LEVELLT-1001WASTE HOLDUP TANK LEVELD30-24,000 GAL TOP TO BOTTOMN/AN/AN/AN/A**YESYESNOTE 1B D33 RADWASTE SYSTEMS - RADIOACTIVE GAS HOLDUP TANK PRESSUREPT-1036GAS DECAY TANK `A' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1BPT-1037GAS DECAY TANK `B' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1BPT-1038GAS DECAY TANK `C' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1BPT-1039GAS DECAY TANK `D' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1BPT-1052GAS DECAY TANK `E' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1BPT-1053GAS DECAY TANK `F' (HOLDUP) PRESSURED30-160 PSIG0-150% DESIGN PRESSUREN/AN/AN/AN/A**YESYESNOTE 1B D34 VENTILATION SYSTEM - EMERGENCY VENTILATION DAMPER POSITIOND1-A (LS)C.R. NORMAL INTAKE DAMPER POSITIOND2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-1A IND. LIGHTS ASSOCIATED WITH HIS-6552AD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-1B (LS)C.R. NORMAL INTAKE DAMPER POSITIOND2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-1B IND LIGHTS ASSOCIATED WITH HIS-6552BD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-2 (LS)C.R. EMERGENCY AIR INTAKE EASTD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-2 IND. LIGHTS ASSOCIATED WITH HIS-6541D2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 1YES------D-3 (LS)C.R. EMERGENCY AIR INTAKE WESTD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-3 IND. LIGHTS ASSOCIATED WITH HIS-6542D2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 1YES------D-11A (LS)C.R. FILTER RECIRCULATION AIRD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-11A IND. LIGHTS ASSOCIATED WITH HIS-6543 AD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 29 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIOND-11B (LS)C.R. FILTER RECIRCULATION AIRD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYES D-11B IND. LIGHTS ASSOCIATED WITH HIS-6543 BD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-14 (LS)TOILET EXHAUSTD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-14 IND. LIGHTS ASSOCIATED WITH HIS-6550D2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------D-22 (LS)KITCHEN EXHAUSTD2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 4---YESYESD-22 IND. LIGHTS ASSOCIATED WITH HIS-6549D2OPEN/CLOSED OPEN/CLOSEDN/AN/AN/ANOTE 2YES------
D35 POWER SUPPLIES - STATUS OF STANDBY POWER & OTHER ENERGY SOURCES IMPORTANT TO SAFETY4AA`4A' 4KV BUS VOLTAGED20-5000 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4AA024A 4KV BUS-AUX XFRMER CURRENTD20-4000 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4AA054A 4KV BUS-S/U XFRMER CURRENTD20-4000 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4AB`4B' 4KV BUS VOLTAGED20-5000 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4AB024B 4KV BUS-AUX XFRMER CURRENTD20-4000 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4AB054B 4KV BUS-S/U XFRMER CURRENTD20-4000 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YES------4K4AEMERGENCY DIESEL GENERATOR `4A' VOLTAG ED20-5000 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4K4AEMERGENCY DIESEL GENERATOR `4A' CURREN TD20-600 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4K4BEMERGENCY DIESEL GENERATOR `4B' VOLTAG ED20-5000 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4K4BEMERGENCY DIESEL GENERATOR `4B' CURREN TD20-600 AMPS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4B01LOAD CENTER 4A STATUSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4BO2LOAD CENTER 4B STATUSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4B03LOAD CENTER 4C STATUSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4B04LOAD CENTER 4D STATUSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 5YESYESYES4B50LOAD CENTER 4H STATUSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B05MCC-4A VITAL BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B06MCC-4B VITAL BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 30 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION4B07MCC-4C VITAL BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B08MCC-4D VITAL BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B51MCC 4J BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B52MCC 4K BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B53MCC 4L BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4B54MCC 3M BUS VOLTSD20-600 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B3D013A BATTERY AND CHARGER BUS VOLTSD20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4D014B BATTERY AND CHARGER BUS VOLTSD20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B3D233B BATTERY AND CHARGER BUS VOLTSD20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4D234A BATTERY AND CHARGER BUS VOLTSD20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4Y01INVERTER 4A CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y01INVERTER 4A VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y02INVERTER 4B CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y02INVERTER 4B VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y05INVERTER 4C CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y05INVERTER 4C VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y07INVERTER 4D CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y07INVERTER 4D VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B3Y04INVERTER AS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B3Y04INVERTER AS VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y04INVERTER BS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y04INVERTER BS VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B3Y06INVERTER CS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B3Y06INVERTER CS VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4Y06INVERTER DS CURRENTD20-100 AMPS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B C26* DCS(SPDS)
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 31 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION4Y06INVERTER DS VOLTAGED20-150 VOLTS PLANT SPECIFICN/AN/AN/ANOTE 6**YESYESNOTE 1B4AA02NO.4 AUX XFRMER TO 4A 4160 VOLT BUS BKR S TD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AA05NO.4 S/U XFRMER TO 4A 4160 VOLT BUS BKR S TD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AB02NO.4 AUX XFRMER TO 4B 4160 VOLT BUS BKR S TD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AB05NO.4 S/U XFRMER TO 4B 4160 VOLT BUS BKR S TD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AA22NO.4 S/U XFRMER TO 4A 4160 VOLT BUS BKR S TD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AA20EDG `4A' TO 4A 4160 VOLT BUS BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AB21EDG `4B' TO 4B 4160 VOLT BUS BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AA094B 4KV BUS TIE TO 4AC14 AND 4AB22 BKR STAT UD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YES------4AB224B 4KV BUS TIE TO 4AA09 AND 4AC14 BKR STAT UD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YES------4AA084A 4KV BUS TO LOAD CENTER 4A BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AB094B 4KV BUS TO LOAD CENTER 4B BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AA144A 4KV BUS TO LOAD CENTER 4C BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4AB144B 4KV BUS TO LOAD CENTER 4D BKR STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 3YESYESYES4P06120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4P07120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4P08120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1B4P09120 VAC INVERTER PANEL-BREAKER STATUSD2OPEN/CLOSED PLANT SPECIFICN/AN/AN/ANOTE 1**YESYESNOTE 1BPS-2329AFW NITRO. BACKUP SUPPLY STA. 1 TR. 1 LOW PD2NORMAL/LOW PRESSUR E PLANT SPECIFICN/AN/AN/ANOTE 1---YESYES AFW NITRO. STA. 4 TR. 1 LOW PRESS. IND. LIGH TD2NORMAL/LOW PRESSUR E PLANT SPECIFICN/AN/AN/ANOTE 1YES------PS-2328AFW NITRO. BACKUP SUPPLY STA. 1 TR. 2 LOW PD2NORMAL/LOW PRESSUR E PLANT SPECIFICN/AN/AN/ANOTE 1---YESYES AFW NITRO. STA. 4 TR. 2 LOW PRESS. IND. LIGH TD2NORMAL/LOW PRESSUR E PLANT SPECIFICN/AN/AN/ANOTE 1YES------
Revised 04/17/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 32 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION E1 CONTAIMENT RADIATION - CONTAIMENT AREA RADIATION HI RANGERAD-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A'E11 TO 1E8 R/Hr 1 TO 1E7 R/HrCOMPLYCOMPLYRAD-6311BNOTE 1---YESYESRAI-6311ACTMT. HIGH RANGE RAD. MONITOR CH. 'A' IND.E11 TO 1E8 R/Hr 1 TO 1E7 R/HrN/ACOMPLYRAI-6311BNOTE 1YES------RAR-6311ACTMT. HIGH RANGE RAD. MONITOR CH. `A' REC OE11 TO 1E8 R/Hr 1 TO 1E7 R/HrN/ACOMPLYRAR-6311BNOTE 1YES------RAD-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B'E11 TO 1E8 R/Hr 1 TO 1E7 R/HrCOMPLYCOMPLYRAD-6311ANOTE 1---YESYESRAI-6311BCTMT. HIGH RANGE RAD. MONITOR CH. 'B' IND.E11 TO 1E8 R/Hr 1 TO 1E7 R/HrN/ACOMPLYRAI-6311ANOTE 1YES------RAR-6311BCTMT. HIGH RANGE RAD. MONITOR CH. `B' REC OE11E-4 TO 1 R/Hr 1E-1 TO 1E4 R/HrN/ACOMPLYRAR-6311ANOTE 1YES------
E2 AREA RADIATION - RADIATION EXPOSURE RATERD-1417EAST END OF E/W CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1418WEST END OF E/W CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1420CONTROL ROOME31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1415NORTH END OF N/S CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1416SOUTH END OF N/S CORRIDORE31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1413OUTSIDE SAMPLE RM.-UNIT 3E31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4RD-1414OUTSIDE SAMPLE RM.-UNIT 4E31E-4 TO 1E4 R/Hr1E-1 TO 1E4 R/HrN/AN/AN/AN/A---YESYESNOTE 4R-1405RECORDERE31E-4 TO 1E4 R/HR1E-1 TO 1E4 R/HRN/AN/AN/AN/AYES------NOTE 4 E3 AIRBORNE RADIOACTIVE MATERIALS RELEASED FROM PLANT NOBLE GAS & VENT FLOW RATENONECTMT. OR PURGE EFFLUENTE2THIS DESIGN NOT USED1E-6 TO 1E5 micro Ci/CC---------------------NOTE 5NONECTMT. OR PURGE EFFLUENT (FLOW)E2THIS DESIGN NOT USED0-110% DESIGN FLOW---------------------NOTE 5NONEREACTOR SHIELD BLDG ANNULUSE2THIS DESIGN NOT USED1E-6 TO 1E4 micro Ci/CC---------------------NOTE 5NONEAUXILIARY BLDGE2THIS DESIGN NOT USED1E-6 TO 1E3 micro Ci/CC---------------------NOTE 5NONEAUXILIARY BLDG.(FLOW)E2THIS DESIGN NOT USED0-110% DESIGN FLOW---------------------NOTE 5 E4 CONDENSER AIR REMOVAL SYSTEMRAD-6417AIR EJECTOR CONDENSER EXH.E21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E5 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1BNONEAIR EJECTOR CONDENSER FLOWE2- - -- - ----------------------NO INST.
Revised 01/31/2013 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 33 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATION E5 COMMON VENT - NOBLE GASESRAD-6304VENT STACK W.R. RAD. MONITORE21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E3 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1BFT-6584VENT STACK-FLOWE20-150,000 CFM0-110% DESIGN FLOWN/AN/AN/ANOTE 7**YESYESNOTE 1B E6 VENT FROM STEAM GENERATOR SAFETY RELIEF VALVERAD-6426STEAM LINE RAD. MONITORE21E-1 TO 1E3 micro Ci/CC1E-1 TO 1E3 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTES 1B E6A ALL OTHER IDENTIFIED RELEASE POINTSRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONL YE21E-7 TO 1E5 micro Ci/CC1E-6 TO 1E2 micro Ci/CCN/AN/AN/ANOTE 7**YESYESNOTE 1B E7 PARTICULATES & HALOGENS - ALL IDENTIFIED PLANT RELEASE POINTSRAD-6304VENT STACKE3NOTE 31E-3 TO 1E2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1BRAD-6418SPENT FUEL POOL VENT MONITOR (UNIT 3 ONL YE3NOTE 31E-3 TO 1E2 micro Ci/CCN/AN/AN/AN/A**YESYESNOTE 1B E8 ENVIRONS RADIATION AND RADIOACTIVITY GRAB SAMP L AIRBORNE RADIOHALOGENS AND PARTICULAT EE3NOTE 11E-9 TO 1E-3 micro Ci/CCN/AN/AN/A------------
RO-2,RO-2A
&PLANT AND ENVIRONS RADIATIONE3NOTE 2 NOTE 1N/AN/AN/A------------
PORTABLE M PLANT AND ENVIRONS RADIOACTIVITYE3ISOTOPIC ANALYSISISOTOPIC ANALYSISN/AN/AN/A------------
E9 METEOROLOGY - WIND DIRECTION & SPEED ESTIMATE OF ATMOSPHERIC STABILITY 10 M. W.D. S OMETEOROLOGY 10 METER WIND DIRECTIONE30-540 DEGREE 0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.D. T UMETEOROLOGY 10 METER WIND DIRECTIONE30-540 DEGREE 0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 60 M. W.D. S OMETEOROLOGY 60 METER WIND DIRECTIONE30-540 DEGREE 0-360 DEGREEN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.S. S OMETEOROLOGY 10 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1B 10 M. W.S. T UMETEOROLOGY 10 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1B 60 M. W.S. S OMETEOROLOGY 60 METER WIND SPEEDE30-100 MPH 0-50 MPHN/AN/AN/AN/A**YESYESNOTE 1B DELTA T "A" SESTIMATE OF ATMOSPHERIC STABILITYE3-5 TO +15 F-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,3 DELTA T "B" SESTIMATE OF ATMOSPHERIC STABILITYE3-5 TO +15 F-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,310M SIGMA TESTIMATE OF ATMOSPHERIC STABILITYE30-100 DEGREES-5 C TO 10 C (-9 F TO 18 F)N/AN/AN/AN/A**YESYESNOTES 1B,3 E10 ACCIDENT SAMPLING CAPABILITY - PRIMARY COOLANT AND SUMPNONERCS ACTIVITY GROSS CPSE3NOTE 410 micro Ci/ml to 10 Ci/mlN/AN/AN/AN/AN/AN/AN/ANONECTMT. AIR-ISOTOPIC ANALYSIS GAMMA SPECT RE3NOTE 4 ISOTOPIC ANALYSISN/AN/AN/AN/AN/AN/AN/ANONEBORON ANALYZER RCS SOLUBLE BORON CON CE3NOTE 4 0-6000 PPMN/AN/AN/AN/AN/AN/AN/ANONERCS CHLORIDE ANALYSIS OF PRIMARY COOLA NE3NOTE 4 0-20 PPMN/AN/AN/AN/AN/AN/AN/A C27* DCS(SPDS)
Revised 01/08/2014 TABLE 7.5-2 PARAMETER LISTING
SUMMARY
SHEETS UNIT 4 TURKEY POINTSHEET 34 OF 34 VARIABLE INSTRUMENT RANGE DISPLAYITEMTAG NO.ENVIRONSEISMICREDUNDANCEPOWER LOCATIONSCHEDULE/
DESCRIPTION T YP E CA T EXISTING REQUIREDQUAL.QUAL.
SUPPL YCRTSCEOFJUSTIFICATIONNONEDISSOLVED HYDROGEN ANALYSIS OF PRIMARY E3NOTE 4 0-2000 CC/KGN/AN/AN/AN/AN/AN/AN/ANONEDISSOLVED OXYGEN ANALYSIS OF PRIMARY C OE3NOTE 4 0-20 PPMN/AN/AN/AN/AN/AN/AN/ANONERCS pH ANALYSIS OF PRIMARY COOLANTE3NOTE 4 1-13 pHN/AN/AN/AN/AN/AN/AN/AAE-6307ACTMT. HYDROGEN CONCENTRATION CH. AE30-10% AND 0-20%0-10 VOL. %N/AN/AN/AN/A*YESYESNOTE 1BAE-6307BCTMT. HYDROGEN CONCENTRATION CH. BE30-10% AND 0-20%0-10 VOL. %N/AN/AN/AN/A*YESYESNOTE 1BNONECONTAINMENT OXYGENE3NOTE 4 0-30 VOL. %N/AN/AN/AN/AN/AN/AN/ANONECTMT. AIR GAMMA SPECTRUME3NOTE 4 ISOTOPIC ANALYSISN/AN/AN/AN/AN/AN/AN/A C26* DCS(SPDS)
Revised 01/31/2013 NOTES FOR TABLE 7.5-2
TURKEY POINT UNIT 4 Sheet 1 OF 9 For Tag No. Column (LS) = Limit Switch Associated with Valve For Existing Instrument Range Column1.Portable sampling with onsite analysis capability is capable of providing a range from less than 1E-9 micro Ci/CC to greater than 1E-3 micro Ci/CC.2.Portable instrumentation provides a range of:A.1E-3 R/HR to values greater than 1E4 R/HR photons; and B.1E-3 R/HR to values greater than 1E4 R/HR beta and low-energy photons
- 3. Existing range monitors up to 7.4E-2 micro Ci/CC. Plant specific analysis justifies smaller range. Particulates and halogens collected on
filter cartridge and monitored in lab after sample collection period (30
minutes design for accident situations).
- 4. No instrument is provided for this variable. Elimination of the need to provide on-site analysis capability for this variable has been accepted by the NRC in their safety analysis report related to technical specification amendments 211/205, dated 1/31/2001.
For Required Instrument Range Column1.RG 1.97 requires the following ranges:A.1E-3 R/HR to 1E4 R/HR photons; andB.1E-3 R/HR to 1E4 R/HR beta and low-energy photons For Environmental Qualification Column1.The Safety Injection Accumulator Discharge Valves MOV-865A, B and C are administratively controlled and are required to be in the open position during normal operation. These valves are not required to change
position under accident conditions. Administrative control is
accomplished by locking open the associated motor control center circuit
breakers. Since administrative control via electrical de-energization of
the valves ensures that the valves will be in their safe position during 06/18/2001 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 2 OF 9 an accident, environmental qualification of the limit switches providing position indication is not required.
For Power Supply Column
Power source is identified as:
- 1. Class 1E, 120 VAC uninterruptable power supply (inverters)
- 2. Class 1E, 120 VAC power backed up by the Emergency Diesel Generator
- 3. Class 1E, 125 VDC safety-related battery
- 4. Non-Class 1E, 120 VAC uninterruptable power supply
- 6. Transducers internal to the inverter providing computer display signals for inverter current and voltage are powered by the inverter internals.
- 7. The SPING monitors communicate with both primary and backup control terminals which are powered from plant inverters and backed up by the
safety-related batteries. SPING Monitors RAD-3(4)-6417 are powered from
non-vital lighting panels capable of being powered from the emergency
diesel generators. SPING Monitor RAD-6304 and steam line monitor RAD-
6426 are powered from vital AC power panels which are automatically
backed up by an emergency diesel generator.
For Display Location
- 1. Control Room metering is credited for primary indication of Emergency Diesel Generator Output (MW). Recording capability for this variable is
also available via DCS/ERDADS.
For Schedule/Justification Column
- 1. The following notes referenced under the "Schedule/Just" column of the
Revised 04/17/2013 C26C26 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 3 OF 9 Parameter Listing Summary Sheets correspond to the technical justifications identified below:
A. This justification demonstrates the acceptability of the existing uninterruptable power source (UPS) associated with the
DCS/SPDS/ERDADS computer for the monitoring of Category 1
variables. This acceptability is based upon the existing UPS
allowing the DCS/SPDS/ERDADS computer to perform its credited
RG 1.97 functions:
(1) Recording of Category 1 Variables -
Control Room indication is normally used to provide trending while DCS/SPDS/ERDADS is used only as a backup to those
instruments. In those cases where DCS/SPDS/ERDADS is being
used to trend Category 1 variables, either the trending is
not necessary to the Control Room operator's decisions or
the operator can obtain the real time information via the
monitoring of Control Room indication.
(2) Indication of Category 1 Variables -
DCS/SPDS/ERDADS is only used as a backup means of indication for certain containment isolation valves but is not credited
for RG 1.97 indication for any other Category 1 variable.
(3) Containment Isolation Valve Indication -
In the few instances where DCS/SPDS/ERDADS is credited for backup indication associated with containment isolation
valves, computer power will be available from the UPS
battery for at least the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the accident.
This period of operability is sufficient to allow the
completion of containment isolation.
B. This justification demonstrates that the DCS/SPDS/ERDADS computer, although classified as non-nuclear safety-related, is capable of
Revised 04/17/2013 C26C26C26C26C26C26 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 4 OF 9 providing the necessary Regulatory Guide functions for which it is credited.
(1) The DCS/SPDS/ERDADS computer is not essential to the monitoring of Category 1 variables. The computer is
credited only for backup indication of a few containment
isolation valves (i.e., valve position indication) but is
not credited for either primary or backup indication for any
other Category 1 variables.
(2) The DCS/SPDS/ERDADS computer is not essential in providing the Control Room operators with vital trending or recording
information. Control Room indication is normally used to
provide trending while DCS/SPDS/ERDADS is used only as a
backup to those instruments. In those cases where
DCS/SPDS/ERDADS is being used to trend Category 1 variables, either the trending is not necessary to the operator's
decisions or the operator can obtain the real time
information via the monitoring of Control Room indication.
(3) The DCS/SPDS/ERDADS computer provides primary indication of certain Category 2 and 3 variables. In general, the
DCS/SPDS/ERDADS computer complies with the Category 2 and 3
design and qualification criteria identified in Table 1 of
(4) The DCS/SPDS/ERDADS computer does not diminish the capability of the Control Room operators to obtain the
necessary post-accident monitoring information or in
achieving the safe shutdown of the plant. Based upon
conclusions (1) and (2) above, it can be further concluded
that the DCS/SPDS/ERDADS computer does not perform an
essential function with respect to Category 1 post-accident
monitoring.
C. This justification demonstrates that the lack of overlap between the ranges of Containment Sump Water Level narrow and wide range
instrumentation does not jeopardize the capability of providing the
Control Room operators the critical information required during
Revised 04/17/2013 C26C26C26C26C26C26C26C26 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 5 OF 9 plant accident conditions. This is based on an analysis which provides the following:(1)The deadband in Containment Sump Water Level indication between 369" and 397" causes less than 6% error in indication.(2)The resulting error in indication is introduced in a non-critical range of the required indication. Thus the deadband
does not prevent the operator from obtaining the required
information:(a)Low level (narrow range) indication of the initial ingress of water into the sump to allow the assessment of water source and rate.(b)High level (wide range) indication for operator response to containment flooding.(c)Determination of the ability to transfer to cold leg recirculation in the event of loss of reactor or
secondary coolant based upon having achieved minimum
pump NPSH.D.This justification clarifies the inconsistency between the Accumulator Tank Level ranges identified in the previous FPL RG 1.97
submittals of January 26, 1984 and May 10, 1985, and the existing
Control Room instrumentation range of 6,500 to 6,750 gals. The
existing Control Room range of 6,500 to 6,750 gals. uses the same
basis for justification as identified and approved by NRC in its
Safety Evaluation dated March 20, 1986. Accumulator tank pressure
is also credited for determining accumulator tank level. As
pressure drops in the accumulators, application of the Ideal-Gas
state equation provides indication of how much water remains in the
accumulator following actuation. As an operator aid, a curve has
been made available to the operator which correlates accumulator
pressure to accumulator level. The accumulator instrumentation have been downgraded, per NRC letter dated April 13, 1992, Docket Nos.
50-250 and 50-251, from R.G. 1.97 Category 2 to Category 3.
Rev. 16 10/99 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 6 OF 9E.This justification clarifies the use of flow meters integral to hand indicating controllers as a means of providing valve position indication. The integral flow meters provide "closed" position
indication by indicating zero flow and "not closed" position
indication by indicating higher than zero flow.F.This justification identifies alternative instrumentation being credited for the monitoring of Containment Spray Flow. An
alternative method of monitoring this variable was identified in to FPL RG 1.97 submittal dated May 10, 1985. The
alternative instrumentation provides monitoring of the operation of
the Containment Spray System, as intended by RG 1.97. This is
accomplished by monitoring the proper alignment of Containment Spray
valves and operation of the Containment Spray pumps. In addition, the monitoring of containment temperature and pressure assures that
containment cooling systems are performing their required function.
Monitoring of RWST level provides indirect indication of the
Containment Spray flow function.G.This justification identifies alternative instrumentation being credited for the monitoring of Containment Fan Heat Removal. An
alternative method of monitoring this variable was identified in to FPL RG 1.97 submittal dated May 10, 1985. The
method used to address this variable monitors the operation of the
Emergency Containment Cooling (ECC) fans and verifies that Component
Cooling Water (CCW) flow has been established to the ECC coolers.
In addition, the monitoring of containment pressure and temperature
provides indirect indication of the Containment Fan Heat Removal
function.H.This justification provides the rationale for not recording containment isolation valve position (Category 1 variable).
Recording of containment isolation valve position is not essential
for operator action. Containment valve position is available to the
operators via Control Room indicating lights. The operators depend
on the real time information provided by indicating lights to verify
containment isolation. Thus the operators do not need trending of
valve position to verify isolation.
Rev. 15 4/98 NOTES FOR TABLE 7.5-2 (Continued) Sheet 7 OF 9 I. This justification provides the basis for the acceptability of the existing range for Containment Sump water Level narrow range indication. The existing range of LI-6308A&B includes a 0-5 inch deadband (i.e., no specific reading can be obtained). However, since the 0-5 inch deadband is outside of the loop measurement range and insignificant compared to the span of 364 inches, the lower limit of the indicator scale of 0-5 inches is acceptable.
J. This justification provides the basis for the acceptability of the existing range for Containment Sump water Level narrow range recording. The existing range of LR-6308A&B includes a 0-5 inch deadband (i.e., no specific reading can be obtained). However, since the 0-5 inch deadband is insignificant compared to the span of 364 inches, the lower limit of the recorder scale of 0-5 inches is acceptable.
K. This justification demonstrates that the lack of units of measure (i.e., inches) associated with the Control Room indication for Containment Sump water Level wide range, LI-6309A&B, will not mislead the Control Room operators. This is based on the operators being familiar with the applicable units of measure via training.
L. Wide range monitoring for Steam Generator Level is provided via a single non-Class 1E wide range level loop. This justification demonstrates that, although wide range monitoring may not be available during an accident scenario, the Control Room operator will have sufficient information to identify and mitigate an accident and to determine the availability of the steam generators as heat sinks. This is based upon the following:
(1) Steam generator level will either remain within narrow range level indication or, if steam generator level has fallen below narrow range indication, that Auxiliary Feedwater has been initiated and will result in the recovery of steam generator level to within narrow range limits. This is accomplished via the associated emergency operating procedures.
(2) RCS temperature (i.e., hot and cold leg water temperature) and pressure are available to determine the effectiveness of the steam generators as heat sinks.
- 2. Since the original containment isolation design for Turkey Point was not required to provide redundant valve position indication, the redundancy
Rev. 16 10/99 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 8 OF 9 criteria of RG 1.97 are not applicable to the existing plant design. As a result, in order to address the RG 1.97 concern for ensuring Control Room capability to verify isolation status, an RG 1.97 Containment
Isolation Valve Evaluation was performed. The evaluation considers the
effects of single failure of valve indication and demonstrates the
capability for the Control Room operator to verify isolation of
Containment penetrations.
- 3. An exception to this variable has been accepted by NRC in its Safety Evaluation Report dated March 20, 1986.
- 4. All 24 channels of the Area Radiation Monitoring System (ARMS) have been replaced by PC/M 89-462 to comply with commitments made to the NRC in
FPL letter L-88-290 (Reference 6). L-88-290 commitments require the use
of instrumentation with a range of 10
-3 R/hr to 10 2 R/hr. Instrumentation installed under PC/M 89-462 has a range of 10
-4 R/hr to 10 4 R/hr, which exceeeds both Regulatory Guide 1.97 recommendations and L-88-290 commitments.
- 5. No instrumentation has been provided since effluent discharge is through a common plant vent.
- 6. No recording capability exists for 4KV Bus Voltage (Category 1 variable). The emergency operating procedures presently credit the
monitoring of 4KV Bus Voltage to allow the Control Room operator to
determine the loss of power to a 4KV bus. Control Room meter indication
of 4KV bus voltage is available and is adequate to allow the operator to
identify the loss of bus voltage on a realtime basis. Trending of bus
voltage is not necessary to ensure accomplishment of this manual action.
Therefore, recording of the variable is not essential.
- 7. NOT USED
- 8. NOT USED
Revised 04/17/2013 C26C26 NOTES FOR TABLE 7.5-2 (Continued)
Sheet 9 OF 9 9. The original plant design included 51 core exit thermocouples. Due to
the potential for individual sensor failures, the actual number of
operable thermocouples may be reduced below this value.
- 10. NOT USED
- 11. Existing instrument range of "OPEN/CLOSED" is derived from a single limit switch contact. The contact provides CLOSED/NOT CLOSED position
to ERDADS which then defines and displays the position as OPEN or CLOSED
at the CR, TSC and EOF consoles.
- 12. On-site analysis capability for this variable has been eliminated in favor of grab samples and offsite analysis. This change is consistent with commitments documented in NRC safety evaluation for technical
specification amendments 211/205, dated 1/31/2001.
01/27/2013 C26 7.6 IN-CORE INSTRUMENTATION 7.6.1 DESIGN BASIS
The in-core instrumentation is designed to yield information on the neutron
flux distribution and fuel assembly outlet temperatures at selected core
locations. Using the information obtained from the in-core instrumentation
system, it is possible to confirm the reactor core design power distribution
parameters and calculated hot channel factors. The system provides means for
acquiring data and performs no operational control.
7.6.2 SYSTEM DESIGN
The in-core instrumentation system consists of the Inadequate Core Cooling
System (ICCS) and flux thimbles, which run the length of selected fuel
assemblies to facilitate measurement of the neutron flux distribution within
the reactor core.
The measured data obtained from the ICCS in-core temperature thermolcouples and flux distribution instrumentation system, in conjunction with previously determined analytical information, can be used to determine the fission power distribution in the core at any time throughout core life. This method is more accurate than using calculations alone.
Once the fission power distribution has been established, the maximum power output is primarily determined by thermal power distribution and the thermal and hydraulic limitations determine the maximum core capability.
The in-core instrumentation provides information which may be used to calculate the coolant enthalpy distribution, the fuel burnup distribution, and an estimate of the coolant flow distribution.
Both radial and azimuthal symmetry of power may be evaluated by combining the detector and thermocouple information from the one quadrant with similar data obtained from the other three quadrants.
The ICCS consists of three systems:
- 1. Core Exit Thermocouples System (CET)
- 2. Heated Junction Thermocouples System (HJTC)
- 3. Subcooled Margin Monitoring System (SMM)
7.6-1 Revised 09/18/2007 C23C23 These three systems are briefly discussed below:
- 1. Core-Exit Thermocouples System This system originally included 51 thermocouples positioned to measure fuel assembly coolant outlet temperature at preselected locations; some thermocouples have been abandoned in accordance with plant procedures.
The temperature measurement signals from these thermocouples are carried through silicon-rubber insulated cables with stainless steel protective jackets routed in redundant channels. The thermocouples for the two channels have been selected in such a way that each channel indicates the temperature of the whole core. The thermocouple outputs are recorded in the computer room and indicated in the control room.
- 2. Heated Junction Thermocouple System This system includes eight pairs of heated/unheated thermocouples located axially in a probe assembly; some probes have had pairs of heated/ unheated thermocouples abandoned in accordance with plant procedures. There are two identical probe assemblies in the reactor vessel. The measurements from these thermocouples are carried through silicon-rubber insulated cables with stainless steel protective jackets routed in two redundant channels. Two pairs of thermocouples are located in the upper head region above the upper support plate and six pairs are located in the upper plenum region between core alignment and support plates. These thermocouples provide information regarding reactor coolant inventory. The outputs from these thermocouples are processed in the computer room and indicated in the control room.
- 3. Subcooled Margin Monitoring System This system includes two pressure transmitters to measure RCS pressure and one dual RTD in each hot and cold leg to measure RCS temperature. Reactor coolant system hot leg temperature (1 per loop per QSPDS channel), cold leg temperature (1 per loop per QSPDS channel) and RCS pressure (1 per QSPDS channel) are routed in two redundant channels to the computer room for saturation margin calculations.
In the computer room, the signals for these systems are processed by a computer installed in a seismically qualified cabinet for each channel. A display unit for each channel is installed in the control room for indication of processed parameters and these are connected to the computer with a fiber optic data link. Each ICCS (QSPDS) Channel is powered from a station vital power supply.
7.6-1a Revised 08/17/2016 C28 Thermocouples Chromel-alumel thermocouples 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 seals, using high pressure screwed fittings from conduit to head. The thermocouple
column to vessel seal consists of grafoil packing rings which are compressed by
a drive sleeve to seal the annulus between the thermocouple column and the head
port adapter. The head port adapter is threaded and seal welded onto the nozzle
penetrating the vessel head. (See Figure 7.6.1.) The thermocouples are
enclosed in stainless steel sheaths within the above tubes to allow replacement
if necessary. Thermocouple outputs are recorded in the computer room and
displayed in the control room. The support of the thermocouple guide tubes in
the upper core support assembly is described in Section 3.
Movable Miniature Neutron Flux Detectors Mechanical Configuration
Five fission chamber detectors (employing U 3 O 8 which is 90 percent enriched in U 235) can be remotely positioned in retractable guide thimbles to provide flux mapping of the core. Maximum chamber dimensions are 0.188-inch in diameter and 2.10 inches in length. The stainless steel detector shell is welded to the
leading end of the helical wrap drive cable and the stainless steel sheathed
coaxial cable. Each detector is designed to have a minimum thermal neutron
sensitivity of 1.0 x 10
-17 amps/nv and a maximum gamma sensitivity of 3 x 10
-14 amps/R/hr. Operating thermal neutron flux range for these detectors is 1 x 10 10 to 8.7 x 10 13 nv. Other miniature detectors, such as gamma ionization chambers and boron-lined neutron detectors, can also be used in the system. 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 zone.
7.6-2 Revised 01/14/2010 C24 The thimbles which are dry inside are closed at the leading ends, 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 five drive assemblies, five path group selector assemblies and five rotary selector assemblies. 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 trailing ends of the drive cables. Each drive assembly generally consists of a gear motor which pushes a helical-wrap drive cable and detector through a selected thimble path by means of a special drive box and includes a storage device that accommodates the total drive cable length.
Further information on mechanical design and support is described in Section 3.2.3. During the Unit 4 Cycle 27 refueling outage, the following twenty two thimble tubes were replaced/installed: C-12, E-11, G-14, H-1, M-3, J-3, L-11, N-12, F-2, L-5, J-7, G-9, F-13, F-8, F-6, H-4, N-5, C-8, L-9, J-5, L-4, AND N-7. Thimble tubes H-1 AND M-3 were capped due to not having their isolation valves, casings, fittings, and supporting frame within each respective tube. During the Unit 4 Cycle 29 refueling outage, thimble tubes H-1 and M-3 previously capped during the Cycle 27 refueling outage, were restored to operational status. Therefore, following the Cycle 29 refueling outage, all thimble locations are available in Unit 4 for flux mapping. Capped thimble tubes are periodically repositioned to minimize tube wall wear. While inspections may result in capping at additional thimble tube locations, the remaining number of detector thimbles will not decrease below the number of required thimbles available for peaking factor verification.
During Unit 3 cycle 24 refueling outage, F-13, G-7, H-3, L-4, L-9, N-5, N-8, H-13, M-3 and J-12 thimble tube core locations were replaced. During the Unit 3 Cycle 27 refueling outage, D-12, E-11, N-10, B-7, D-10, J-10 and G-9 thimble tube core locations were replaced. While future ECT inspections may result in capping thimble tube locations, the remaining number of detector thimbles will not decrease below the number of required thimbles available for peaking factor verification.
7.6-3 Revised 06/23/2016 C28 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 at a selected speed while plotting a level of induced radioactivity versus detector position. Each detector can be driven in or out at speeds of 72 feet per minute or 12 feet per minute. In normal operation, the detectors would move at a speed of 72 feet per minute outside the reactor core and 12 feet per minute when scanning the neutron flux. The average path length external to the core is 120 feet.
Up to five separate fuel assemblies can be scanned simultaneously. A full core map can be read in one hour. The control system consists of two sections, one physically mounted with the drive units, and the other contained in the control room. Limit switches in each drive conduit provide means for pre-recording detector and cable positioning in preparation for a flux mapping operation. Each gear box drives an encoder for positional data plotting. One group path selector (5 path) is provided for each drive unit to route the detector into one of the flux thimble groups. A rotary transfer assembly is a transfer device that is used to route a detector into any one of up to ten selectable paths. Fifty manually operated isolation valves allow free passage of the detector and drive cable when open, and prevents leakage of coolant in case of a thimble rupture, when closed. A path common to each group of flux thimbles is provided to permit cross calibration of the detectors.
The control room contains the necessary equipment for control, position indication, and flux recording. Panels are provided to indicate the core position of the detectors, and for plotting the flux level versus the detector position. Additional panels are provided for such features as drive motor controls, core path selector switches, plotting and gain controls. A "flux-mapping" consists, briefly, of selecting (by panel switches) flux thimbles in given fuel assemblies at various core quadrant locations. The detectors are driven or inserted to the top of the core and stopped automatically or manually.
An x-y plot (position vs 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. Data from selected fuel assemblies are then compared to obtain a flux map of the core.
7.6-4 Revised 06/23/2016 C28 7.6.3 SYSTEM EVALUATION
The thimbles are distributed nearly uniformly over the core with about the same number of thimbles in each quadrant. The number and location of thimbles have been chosen to permit measurement of local to average peaking factors to an accuracy of 10% (95% confidence). Measured nuclear peaking factors will be increased to allow for possible instrument error. The maximum measured hot channel factor will be compared to the hot channel factors in the core operating limits. If the measured hot channel factor is larger than expected, reduced power capability will be indicated.
During the Unit 4 Cycle 20 refueling outage, two of the core mapping thimble tubes at H-1 and M-3 were modified by the insertion of a thermocouple cable.
During Unit 3 Cycle 24 refueling outage, F-13, G-7, H-3, L-4, L-9, N-5, N-8, H-13, M-3 and J-12 thimble tube core locations were replaced. During the Unit 3 Cycle 27 refueling outage, D-12, E-11, N-10, B-7, D-10, J-10 and G-9 thimble tube core locations were replaced. During the Unit 4 Cycle 22 refueling outage, two core mapping thimble tubes at locations E-11 and G-14 were removed from service by capping at their respective seal table high pressure fitting. However, since the minimum complement of thimbles are expected to remain available, there is no impact on the uncertainties assumed in the surveillance of incore peaking factors. During RFO 4-25 CET, H-1 and M-3 thimble tubes have been disconnected, spared repositioned and abandoned in place. During the Unit 4 Cycle 29 refueling outage, thimble tubes H-1 and M-3 previously capped during the Cycle 27 refueling outage, were restored back to operational status. While future ECT inspections may result in capping thimble tube locations, the remaining number of detector thimbles will not decrease below the number of required thimbles available for peaking factor verification. Capped thimble tubes are periodically repositioned to minimize tube wall wear.
7.6.4 REGULATORY GUIDE 1.97, REVISION 3
A review of Turkey Point Units 3 and 4 accident monitoring instrumentation and control systems was conducted against the requirements of Regulatory Guide 1.97, Revision 3. Subsection 7.5.4 presents the requirements of Regulatory Guide 1.97, Revision 3, and the results of the conducted review.
7.6-5 Revised 06/23/2016 C28 Revised 0714/2005 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 IN-CORE INSTRUMENTATI ON GUIDE TUBE PRESSURE SEALS TYPICAL CONFIGURATIONS FIGURE 7.6-1
7.7 OPERATING CONTROL STATIONS 7.7.1 DESIGN BASIS
Criterion: The facility shall be provided with a control room from which actions to maintain safe operational
status of the plant can be controlled. Adequate
radiation protection shall be provided to permit
access even under accident conditions to equipment in
the control room or other areas as necessary to
shutdown and maintain safe control of the facility
without excessive radiation exposures of personnel (1967 proposed GDC 11).
NUREG-0737, "Clarification of TMI Action Plan Requirements", published in October 1980, provided a comprehensive and integrated plan to
improve safety at power reactors. Clarification item I.D.1, "Control
Room Design Reviews," required all licensees and applicants for
operating licenses to develop a Detailed Control Room Design Review (DCRDR) to identify and correct design deficiencies. This review
includes an assessment of the control room layout for human factors
considerations that have an impact on operating effectiveness.
Draft NUREG-0801, "Evaluation Criteria for Detailed Control Room Design Review," provides the guidelines in determining the acceptability of the DCRDR and resultant control room improvements. Subsection 7.7.3
provides a description of the DCRDR for Turkey Point Units 3 and 4.
Clarification item III.D.3.4, "Control Room Habitability Requirements," of NUREG 0737 required all licensees to assure that control room operators would be adequately protected against the effects of an accidental release of toxic or radioactive gases, such that the unit(s) could be safely operated or shutdown if required.
To satisfy these design bases, the units are equipped with a control room which contains the controls and instrumentation necessary for operation of the reactor and turbine generator under normal, abnormal, and accident conditions. The units are also equipped with remote shutdown capability to which allow safe shutdown of the plant from outside the control room if control room evacuation is required.
7.7-1 09/14/2001 Sufficient shielding, distance, and containment integrity are provided to assure that control room personnel radiation exposure under MHA conditions does not exceed 10CFR50 Appendix A, GDC 19 limits during
occupancy of, ingress to and egress from the control room. Multiple
self-contained breathing apparatus units are in and near the control
room for use by the control room personnel during accidental release of
toxic gases. The control room ventilation consists of a system having
a large percentage of recirculated air. The fresh air intake is
automatically closed to control the intake of airborne activity upon
Containment Isolation Actuation Signal.
To ensure that the control Room operators are not impaired by an ammonia storage tank spill at Turkey Point Unit 5, a layer of floating (special surface blanketing) balls has been installed in the
impoundment basin below the ammonia storage tanks. These balls will
automatically arrange themselves into a close packed formation if a
spill occurs and reduce the release of ammonia to the atmosphere.
Consequence modeling demonstrates that the concentration of ammonia in
the control room will remain below the Occupational Safety and Health
Administration Permissible Exposure Levels (OSHA-PEL) without operator
action. 7.7.2 SYSTEM DESIGN 7.7.2.1 Control Room The principal criterion of control station design and layout is that all controls, instrumentation displays and alarms required for the safe
operation and shutdown of the unit are readily available to the
operators in the Control Room.
The Control Room arrangement provides a north-south separation. The alarms for the two units are in opposite ends of the room and have
different tones to make them distinguishable to the operator.
The control room Control Board has been designed to minimize the operators surveillance area. Control stations on the boards are
grouped according to function so as to minimize the possibility of
operator error.
The Control Room control boards consist of console and adjacent vertical panels arranged as shown on Figure 7.7-1. The console
contains those switches and control stations which are most frequently
employed during normal unit operation. The vertical panels contain
those control stations used less frequently (e.g. start-up or shut-
down). 7.7-2 Revised 04/25/2007 C23C23 Indicators and trend recorders are located on both the console and vertical panels. Those located on the vertical panels are positioned to be in front of that section of the console which contains functionally related control stations and indicators.
The Unit 3 detailed control board layout drawings are as follows: Figures 7.7-2a, 7.7-2b, 7.7-3, through 7.7-6. The Unit 4 detailed control board layout drawings are as follows: Figures 7.7-7, through 7.7-12.
7.7.2.2 Remote (Alternate) Shutdown Capabilities
Provisions have been made so that the operators can maintain the units in a safe hot standby condition by means of controls located outside the control room. Refer to Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4 (Reference 12) for a description of the remote (alternate) shutdown capabilities provided for at Turkey Point.
7.7-3 Revised 09/20/2016 C28 7.7.3 SYSTEM EVALUATION - HUMAN FACTORS ENGINEERING
7.7.3.1 HFE Program
In response to the requirement of NUREG-0737, Clarificaton item I.D.1 "Control Room Design Review", and supplement 1 to NUREG-0737, FPL established and maintains a Human Factors Engineering program to review the design of the control room and remote shutdown capabilities to identify and correct design deficiencies. The design review was performed following the guidelines of NUREG-0700, "Guidelines for Control Room Design Review" and NUREG-0801, "Evaluation Criteria for Detail Control Room Design Review".
7.7.3.2 Detail Control Room Design Review Implementation
A summary report which outlined the activities performed for the implementation of the Detailed Control Room Design Review was issued on November 1, 1983. This report was prepared following the outline recommended in Section 5.2 of NUREG-0700. This report discusses:
a) The Detailed Control Room Design Review phases.
b) The technical activities.
c) Method of assessments of discrepancies.
d) Method of identification and selection of enhancement and design solutions.
e) Review results of Human Engineering Discrepancies, Human Engineering Discrepancy Assessment, and the selected enhancement and design solutions.
f) Improvements to be made.
g) Schedule of implementation.
7.7-4 Revised 09/20/2016 An overview of the major activities and methods utilized in the Detail Control Room Design Review is presented below:
Technical Approach
The technical approach utilized in the DCRDR included those activities listed below. A detailed discussion of the methodologies and a discussion of the finding, of each of the surveys is included in Section 2.0 of the DCRDR report. o Review of operating experience o Assembly of control room documentation o Review of system functions and task analysis o Surveys - noise - lighting - control room environment - design conventions - controls and displays
- computers - emergency garmets - labeling - annunciators - anthropometrics - force/torque - communications - maintainability o Verification of task performance capability o Validation of control room functions o Assessment of discrepancies.
Each survey report addresses: o Task Objectives - The type of data to be collected or human performance variables under analysis. o Review Team - The personnel required to conduct the task. o Criteria - Generally, the review guidelines appropriate to the evaluation being conducted. o Task Definition - Steps or procedures followed in the conduct of the task.
o Outputs and Results - Task results. These are Human Engineering Discrepancies which may be drawn upon by subsequent tasks (e.g.,
Task Analysis).
7.7-5 Revised 09/20/2016
Assessment The surveys identified Human Engineering Discrepancies (HEDs). These HEDs were assessed for error inducing potential and the system consequences of the potential error. The means of resolving the HEDs were also reviewed.
The basic assessment process was divided into four steps as follows:
o Assess extent of deviation from NUREG-0700 guidelines o Assess Human Engineering Discrepancy impact on error occurrence o Assess potential consequences of error occurrence o Assign Human Engineering Discrepancy scheduling priority.
Based on the assessment of the HEDs probability of inducing errors a priority for correction was assigned. The HED priority was utilized in the establishment of a backfit schedule.
Implementation
The backfit schedule program for the correction of the HEDs was established based on the following functions: o Human engineering discrepancy priority o Engineering and procurement lead time requirements and constraints o Overall plant outage schedules.
The design solutions and/or enhancements selected for the correction of the HEDs were based on the recommendations of NUREG-0700. o Analysis of correction by enhancement o Analysis of correction by design alternatives o Assess extent of correction.
7.7-6 Revised 09/20/2016 As part of the correction of HEDs several backfit activities, plant change modifications, were implemented. These activities' objectives were to reduce the potential of human errors. Examples of these activities are: the Performance Enhancement Program (PEP) which has improved training drawings and installed more-legible Fiberglass tags on valves, the Visual Instructive Plan which has installed tags throughout the plant which have improved legibility. Panels, 4KV, 480V load center, 480V motor control center, lighting panel, and field component labeling have been modified to include unit color-code information. Additionally the appropriate unit number (i.e. 3 or 4) is used as the first character in the component number.
Emergency Operating Procedures (EOPs) and Off-Normal Operating Procedures (ONOPs) have been reviewed and changed to a new format that will reduce the potential for human error (References 1 to 11). In the new format, procedures are required to be written to the entry-level person, and have less print per page, one action per step, and cautions and warnings before, rather than after the applicable steps. A review also has been made of normal operating procedures (non-emergency) ,
maintenance procedures, health physics, and chemistry procedures, etc, with the intention of making them "user-friendly".
7.7.3.3 DCRDR Implementation Evaluation
The Turkey Point Detailed Control Room Design Review (DCRDR) Program Plan was submitted to the NRC on May 20, 1983. The program plan utilized Supplement 1 to NUREG-0737, NUREG-0700, and NUREG-0801 as the bases for the program development. The Turkey Point DCRDR Summary Report was then submitted on November 1, 1983.
7.7-7 Revised 09/20/2016 The NRC reviewed these reports and provided FPL with a draft Safety Evaluation and Technical Report of the Turkey Point DCRDR on February 2, 1984. This report indicated that a pre-implementation audit would be necessary to resolve the open or confirmatory items identified in the Safety Evaluation. The NRC then conducted the pre-implementation audit of the DCRDR program at Turkey Point on April 2 through 6, 1984.
The results of the NRC audit identified the resolved items and those items requiring additional information. The NRC stated that a meeting would be appropriate to discuss FPL plans, methods, and schedules for submittal of a supplement to the Turkey Point DCRDR Summary Report.
FPL reviewed the requirements of NUREG-0737, Supplement 1 and the operating experience review problems identified and established programs to review and resolve the open or confirmatory items. The Supplemental Summary Report, issued on April 1, 1986 describes the review process. The ten items contained in the supplementary summary report are lised below:
- 1. Operating Experience Review Problems.
- 2. LER Review.
- 3. Task Analysis.
- 4. HFE Review of Post Control Room Changes.
- 5. Additional HED Justification.
- 6. Reverification of Control Room Changes.
- 7. Reverification of Control Room Changes to Ensure No New HEDs.
- 8. Future Control Room Changes.
- 9. Supplemental Summary Report.
- 10. Integration Into Other Programs.
The methodology utilized in the review and resolution of the open or confirmatory items is contained in the DCRDR Supplemental Summary Report.
7.7-8 Revised 09/20/2016 On April 1, 1986, FPL submitted the Supplemental Summary Report on Turkey Point 3 and 4 DCRDR. A preliminary evaluation of the Supplemental Summary Report by NRC resulted in the identification of concerns regarding completion schedules for proposed DCRDR modifications. FPL responded with a September 3, 1986, submittal outlining the schedule for completion of DCRDR modifications. On December 15, 1986, the NRC transmitted a letter along with the Safety Evaluation of the Supplemental DCRDR report. The NRC concluded that FPL had conducted a comprehensive DCRDR program for Plant Turkey Point which satisfied the requirements of Supplement 1 to the NUREG-0737, item I.D.1.
7.
7.4 REFERENCES
- 1. NRC Generic Letter 82-33, NUREG-0737 Supplement 1,"Requirements for Emergency Response Capability," dated December 17, 1982.
- 2. FPL letter to the NRC L-83-237,"Supplement 1 to NUREG 0737 - Generic Letter 82-33," dated, April 15, 1983.
- 3. NRC Generic Letter 83-22,"Safety Evaluation of `Emergency Response Guidelines'," dated June 3, 1983.
- 4. NRC Confirmatory Order, "Order Confirming Commitments on Emergency Response Capability," dated February 23, 1984.
- 5. FPL letter to the NRC L-84-270,"Upgrade Emergency Operating Procedures (EOPs) - Procedures Generation Package," dated October 1, 1984.
- 6. FPL letter to the NRC L-85-472,"Emergency Operating Procedures Upgrade," dated December 23, 1985.
- 7. NRC (Office of NRR) letter to FPL, "Modification of Commission Order Dated February 23, 1984," dated December 24, 1985.
- 8. NRC (Office of NRR) letter to Westinghouse Owner's Group, "Supplemental Safety Evaluation Report by the Office of NRR in the Matter of the Westinghouse Owner's Group Emergency Response Guidelines," dated December 26, 1985.
- 9. NRC (Div. of Reactor Safety) letter to FPL, "Emergency Operating Procedure (EOP) Inspection Program," dated November 3, 1989.
7.7-9 Revised 09/20/2016
- 10. NRC (Div. of Reactor Projects) letter to FPL, "Turkey Point Units 3 and 4 - Procedures Generation Package - TMI Action Plan Items I.C.1.2 and I.C.1.3," dated December 15, 1989.
- 11. NRC Inspection Report No.s 50-250/89-53 and 50-250/89-53, dated March 14, 1990.
- 12. STD-M-006, Engineering Guidelines for Fire Protection for Turkey Point Units 3 & 4.
7.7-10 Revised 09/20/2016 C28 The NRC reviewed these reports and provided FPL with a draft Safety Evaluation and Technical Report of the Turkey Point DCRDR on February 2, 1984. This report indicated that a pre-implementation audit would
be necessary to resolve the open or confirmatory items identified in
the Safety Evaluation. The NRC then conducted the pre-implementation
audit of the DCRDR program at Turkey Point on April 2 through 6, 1984.
The results of the NRC audit identified the resolved items and those
items requiring additional information. The NRC stated that a meeting
would be appropriate to discuss FPL plans, methods, and schedules for
submittal of a supplement to the Turkey Point DCRDR Summary Report.
FPL reviewed the requirements of NUREG-0737, Supplement 1 and the operating experience review problems identified and established
programs to review and resolve the open or confirmatory items. The Supplemental Summary Report, issued on April 1, 1986 describes the review process. The ten items contained in the supplementary summary
report are lised below:
- 1. Operating Experience Review Problems.
- 2. LER Review.
- 3. Task Analysis.
- 4. HFE Review of Post Control Room Changes.
- 5. Additional HED Justification.
- 6. Reverification of Control Room Changes.
- 7. Reverification of Control Room Changes to Ensure No New HEDs.
- 8. Future Control Room Changes.
- 9. Supplemental Summary Report.
- 10. Integration Into Other Programs.
The methodology utilized in the review and resolution of the open or confirmatory items is contained in the DCRDR Supplemental Summary
Report.
7.7-11 Rev 6 7/88 On April 1, 1986, FPL submitted the Supplemental Summary Report on Turkey Point 3 and 4 DCRDR. A preliminary evaluation of the Supplemental Summary Report by NRC resulted in the identification of
concerns regarding completion schedules for proposed DCRDR
modifications. FPL responded with a September 3, 1986, submittal
outlining the schedule for completion of DCRDR modifications. On
December 15, 1986, the NRC transmitted a letter along with the Safety
Evaluation of the Supplemental DCRDR report. The NRC concluded that
FPL had conducted a comprehensive DCRDR program for Plant Turkey Point
which satisfied the requirements of Supplement 1 to the NUREG-0737, item I.D.1.
7.
7.4 REFERENCES
- 1. NRC Generic Letter 82-33, NUREG-0737 Supplement 1,"Requirements for Emergency Response Capability," dated December 17, 1982.
- 2. FPL letter to the NRC L-83-237,"Supplement 1 to NUREG 0737 - Generic Letter 82-33," dated, April 15, 1983.
- 3. NRC Generic Letter 83-22,"Safety Evaluation of `Emergency Response Guidelines'," dated June 3, 1983.
- 4. NRC Confirmatory Order,"Order Confirming Commitments on Emergency Response Capability," dated February 23, 1984.
- 5. FPL letter to the NRC L-84-270,"Upgrade Emergency Operating Procedures (EOPs) - Procedures Generation Package," dated October 1, 1984.
- 6. FPL letter to the NRC L-85-472,"Emergency Operating Procedures Upgrade," dated December 23, 1985.
- 7. NRC (Office of NRR) letter to FPL,"Modification of Commission Order Dated February 23, 1984," dated December 24, 1985.
7.7-12 Rev. 13 10/96
- 8. NRC (Office of NRR) letter to Westinghouse Owner's Group,"Supplemental Safety Evaluation Report by the Office of NRR in the Matter of the Westinghouse Owner's Group Emergency Response Guidelines," dated December 26, 1985.
- 9. NRC (Div. of Reactor Safety) letter to FPL,"Emergency Operating Procedure (EOP) Inspection Program," dated November 3, 1989.
- 10. NRC (Div. of Reactor Projects) letter to FPL,"Turkey Point Units 3 and 4 - Procedures Generation Package - TMI Action Plan Items I.C.1.2 and I.C.1.3," dated December 15, 1989.
- 11. NRC Inspection Report No.s 50-250/89-53 and 50-250/89-53, dated March 14, 1990.
7.7-13 Rev. 13 10/96
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-1 REFER TO ENGINEERING DRAWING 5610-M-63
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 CONTROL ROOM EQUIPMENT LOCATIONS FIGURE 7.7-1
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-2a REFER TO ENGINEERING DRAWING 5610-M-301-28 , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 CONTROL CONSOLE EQUIPMENT LAYOUT SECTIONS 3C01 FIGURE 7.7-2a
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-2b REFER TO ENGINEERING DRAWING 5610-M-301-28 , SHEET 2
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 CONTROL CONSOLE FRONT VIEW SECTION 3C02 FIGURE 7.7-2b
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-3 REFER TO ENGINEERING DRAWING 5610-M-301-12
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL"A" FRONT VIEW SECTION 3C04 FIGURE 7.7-3
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-4 REFER TO ENGINEERING DRAWING 5610-M-301-13, SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL "A" FRONT VIEW SECTION 3C03 FIGURE 7.7-4
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-5 REFER TO ENGINEERING DRAWING 5610-M-301-36
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANELS "B" AND "C" FRONT VIEW SECTION 3C05 FIGURE 7.7-5
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-6 REFER TO ENGINEERING DRAWING 5610-M-301-37
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL "B" FRONT VIEW SECTION 3C06 FIGURE 7.7-6
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-7 REFER TO ENGINEERING DRAWING 5610-M-301-23 , SHEET 1
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 CONTROL CONSOLE FRONT VIEW SECTION 4C01 FIGURE 7.7-7
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-8 REFER TO ENGINEERING DRAWING 5610-M-301-23 , SHEET 2
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 CONTROL CONSOLE FRONT VIEW SECTION 4C02 FIGURE 7.7-8
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-9 REFER TO ENGINEERING DRAWING 5610-M-301-20
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANELS "A" AND "C" FRONT VIEW SECTION 4C04 FIGURE 7.7-9
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-10 REFER TO ENGINEERING DRAWING 5610-M-301-26
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL "A" FRONT VIEW SECTION 4C03 FIGURE 7.7-10
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-11 REFER TO ENGINEERING DRAWING 5610-M-301-40
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL "B" FRONT VIEW SECTION 4C05 FIGURE 7.7-11
FINAL SAFETY ANALYSIS REPORT FIGURE 7.7-12 REFER TO ENGINEERING DRAWING 5610-M-301-41
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 VERTICAL PANEL "B" FRONT VIEW SECTION 4C06 FIGURE 7.7-12
7.8 MISCELLANEOUS ALARMS
7.8.1 Design Basis
Loose Parts Detection System
The loose parts monitor is a non-safety system used to detect unusually high
vibration levels in the primary reactor coolant system. This system will be
utilized to give indication of a possible loose metal part which might
accumulate in one of the steam generators or in the reactor vessel. This
system is strictly for surveillance and performs no safety function. (See
figure 7.8-1)
7.8.2 System Design
The loose parts metal impact system is comprised of thirteen active and two
spare accelerometers strategically located inside containment: two on the
reactor vessel upper head, two active and two spare at the lower area of the
reactor vessel, two just above the tube sheet on each steam generator, and
one near the feedwater inlet on each steam generator (this accelerometer
monitors the secondary side of the steam generator). Each of these locations
is a natural collection region for a potential loose part. Cable routings for
each pair of sensors monitoring the same general area are physically
separated from each other from the sensor to the control room to provide
redundancy.
7.8.3 Alarm Indication
Overall system alarm indication is provided by an annunciator window in Panel "G" in 3C06 for Unit 3 and 4C06 for Unit 4. Specific sensor alarm is
indicated at the process rack. The overall system alarm is also input to the
Digital data Processing System.
7.8-1 Rev 4 7/86
FINAL SAFETY ANALYSIS REPORT FIGURE 7.8-1 REFER TO ENGINEERING DRAWING 5610-M-1388
REV. 13 (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LOOSE PARTS MONITORING SYSTEM FIGURE 7.8-1
7.9 Leading Edge Flow Meter 7.9.1 Design and Operation The Turkey Point Extended Power Uprate (EPU) raised the licensed maximum power level to 2644 MWt. The EPU change to the maximum rated thermal power (RTP) included a 1.7% Measurement Uncertainty Recapture (MUR). Modifications required for the MUR portion of the EPU included installation of the Cameron Leading Edge Flow Meter (LEFM) Check Plus system. The use of LEFM for determination of Feedwater (FW) temperature and FW mass flow, results in an overall calorimetric uncertainty of 0.30%. The MUR uprate of 1.7% results from the difference between the original 2% power determination uncertainty (required by 10CFR50 Appendix K) and the LEFM based calorimetric uncertainty of 0.30%. The MUR is based on the following Cameron Topical Reports:
- 1) ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," dated March 1997 (NRC SER dated March 8, 1999) (Reference 1).
- 2) ER-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check System," dated May 2000 (NRC SER, dated January 19, 2001) (Reference 2).
- 3) ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or Check Plus System," dated October 2001 (NRC SER, dated December 20, 2001) (Reference 3).
The LEFM feedwater flow measurement system is a ultrasonic 8-path transit time flow meter. The LEFM Check Plus system consists of one flow element (meter) installed in each of the three FW flow headers. Each meter has two transit planes which consist of four transit paths. The individual LEFM Check Plus flow elements have been calibrated in a site-specific model test at Alden Research Laboratories with traceability to National Standards. The LEFM flow elements (meters) are installed at specific locations upstream from the existing FW venture nozzles. The resulting piping configurations were explicitly modeled as part of the LEFM meter factor and accuracy assessment testing performed at Alden Research Laboratories. Test data and results for the flow elements are documented in Cameron Engineering Reports ER-748 and ER-752, "Meter Factor Calculation and Accuracy Assessment for Turkey Point 3 and 4 Nuclear Power Plant". (Reference 4).
7.9-1 Revised 04/17/2013 C26 The Calibration factor (also know as the meter factor) and the uncertainty in the calibration factor for the LEFM Check Plus system is based on these Cameron engineering reports.
The LEFM Check Plus system is used for continuous calorimetric power determination by providing FW mass flow and FW temperature input data to the distributed control system (DCS), which is the computer system used for automated performance of the calorimetric power calculations. The LEFM system communicates with the DCS via redundant digital communication links. Individual Steam Generator Heat Rates are calculated in the DCS using the LEFM flow and temperature data along side an independent calculation using the conventional instruments (Feedwater Flow Venturis and Temperatures). The LEFM based Heat Rate data is integrated into appropriate DCS calorimetric display screens to facilitate side-by-side comparison with Heat Rate data based on the conventional instruments. For each steam generator, a correction factor is established to allow operation during a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> allowed outage time during which the LEFM condition can be corrected. The plant may remain at 100% power using the correction factor times the conventional instrument heat rates (Ventrui Corrected value).
The LEFM Check Plus system incorporates self-verification features to ensure that hydraulic profile and signal processing requirements are met within the site-specific design basis uncertainty analysis contained in the Cameron Report ER-783, "Bounding Uncertainty Analysis for Thermal Power Determination at Turkey Point Units 3 & 4 Using the LEFM Check Plus System." (Reference 5). Critical performance parameters are continually monitored for every individual meter path with alarm set points established to ensure corresponding assumptions in the uncertainty analysis remain bounding. A main control room annunciator is provided for operator notification of LEFM degraded system performance or system failure.
Operability of the LEFM instrumentation is required to support an overall calorimetric uncertainty of 0.30%. Operability requirements and associated action statements are identified below.
7.9.2 Operational Restrictions Operability of the LEFM instrumentation is required to support an overall calorimetric uncertainty of 0.30%.
7.9-2 Revised 04/17/2013 C26 Limiting Condition for Operation The LEFM instrumentation shown in Table 7.9-1 shall be OPERABLE Applicability: MODE 1 Action: a) With the number of OPERABLE LEFM / Calorimetric instrument channels less than the minimum required by Table 7.9-1, restore the inoperable channels to OPERABLE status or be in compliance with the reduced power limits of Table 7.9-2 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Operation at 2644 MWT may continue within the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> window provided the Venturi Corrected value is selected.
b) If the plant experiences a power change of greater than 2% during the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, then power level will be restricted to less than or equal to 2599.0 MWt until the LEFM system is fully OPERABLE.
7.
9.3 REFERENCES
- 1. Engineering Report, ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check System," dated March 1997 (NRC SER dated March 8, 1999).
- 2. Engineering Report, ER-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check System," dated May 2000 (NRC SER, dated January 19, 2001).
- 3. Engineering Report, ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or Check Plus System," dated October 2001 (NRC SER, dated December 20, 2001).
- 4. Engineering Reports, Cameron ER-748 and ER-752, "Meter Factor Calculation and Accuracy Assessment for Turkey Point 3 & 4 Nuclear Power Plant," June 2010.
- 5. Engineering Report, Cameron Report ER-783, "Bounding Uncertainty Analysis for Thermal Power Determination at Turkey Point 3 & 4 Using the LEFM Check Plus System." June 2010.
7.9-3 Revised 04/17/2013 C26 TABLE 7.9-1 LEFM CALORIMETRIC INSTRUMANTATION Functional Unit Total No. of Channels Minimum Channels Operable LEFM CPU 2 1 LEFM Meter Section (Path 1-4, 5-8) 6 6 Calorimetric portion of DCS 1 1
Revised 04/17/2013 C26 TABLE 7.9-2 REDUCED POWER LIMITS APPLICABLE to INOPERABLE LEFM CALORIMETRIC INSTRUMANTATION Maximum MWt Maximum Power Total Power uncertaintySelected Calorimetric Mode of Operation Description of Inoperable LEFM Calorimetric Instrument 2638.7 99.8% 0.50% LEFM One meter Section (Plane) in any LEFM in "Check Mode" (Level 2 LEFM System Status) 2599.0 98.3% 2.0% Venturi Any LEFM Meter in "Fail Mode" (Level 3 LEFM System Status) or Loss of communication with both LEFM CPUs. 2599.0 98.3% 2.0% ------ Calorimetric Portion of DCS is OOS
Revised 04/17/2013 C26 APPENDIX 7A DISTRIBUTED CONTROL SYSTEM / SAFETY ASSESSMENT SYSTEM /
EMERGENCY RESPONSE DATA ACQUISITION AND DISPLAY SYSTEM The Safety Assessment System (SAS)/Emergency Response Data Acquisition and
Display System (ERDADS), which is implemented in the plant Distributed
Control System (DCS), has been designed to meet the requirements of NUREG-
0696, "Functional Criteria for Emergency Response Facilities," including the
requirements for the Safety Parameter Display System (SPDS). This system
also meets the requirements of NUREG-1394, "Emergency Response Data System (ERDS) implementation."
1.0 DESIGN BASES
Due to the requirement for obtaining many signals from various safety
systems, the isolation/termination cabinets and the processing cabinets are
seismically qualified to IEEE 344-1975.
The remaining computer and display equipment is not required to be qualified;
and, as it is not part of any safety related system, it is not safety grade.
All equipment located in the control room is mounted to sufficiently restrain
it from affecting any other equipment in the event of an earthquake.
Although not a safety-related system, redundant computer systems are utilized
to meet the reliability requirement in NUREG-0696, "Functional Criteria for
Emergency Response Facilities."
2.0 GENERAL DESCRIPTION
The Safety Assessment System (SAS)/Emergency Response Data Acquisition and
Display System (ERDADS) which includes the Safety Parameter Display System (SPDS) is a real time computer based Distributed Control System (DCS)
designed to assist control room personnel in evaluating the safety status of
the plant. The SAS/ERDADS aids in the coordinated control of the reactor
during upset conditions while concurrently providing information of concern
to the public. The SPDS includes a set of predetermined electronic displays
designed to yield relevant, timely, accurate, and unambiguous information to
the control room operators, the technical support advisors, and the offsite
public safety officials. The SPDS displays a small but critical subset of
the parameters available in the control room, thus reducing the problems
associated with
7A-1 Revised 01/31/2013 C26C26C26C26C26 information overload and parameter selection. At the same time, by preselecting and grouping critical parameters for each display, the SPDS
facilitates comprehension of the prevailing plant and public safety
conditions. This is achieved by presenting high-level displays which
summarize plant safety function status, plant system performance, and
radiological and meteorological data. However, detailed information is not
sacrificed. Each display may be examined at an intermediate or subsystem
level as well as at the individual signal level if detailed information is
desired. Finally, the DCS/SAS/ERDADS is also designed to be useful for
normal operations, allowing the operators to become familiar with the
functions during day-to-day operation. By sharing common, near real-time
displays and pre-selected information, operators, technical support staff, and members of the emergency operations facility staff may cooperate
effectively to bring the plant to a safe condition and to assess the
potential impact on public safety.
3.0 SYSTEM OPERATION
The system consists of six major elements:
- 1) Plant Process Parameters
- 2) Signal Isolation and DCS (ERDADS) Input Signal Processing 3) DCS (ERDADS) Data Processing 4) Plant Data Network 5) Data Link Processing 6) Graphic Display/Readout Equipment
See Figure 7A-1 for the Typical DCS/SAS/ERDADS System Configuration.
7A-2 Revised 01/31/2013 C26C26C26C26 4.0 PLANT PROCESS PARAMETERS The plant process parameters consist of a preselected set of analog and
digital plant signals that are required to assess the plant's overall
condition. The majority of these signals originate from the plant's existing
instrument loops; however, some dedicated instrument loops have been
provided.
Input parameter validation consists of two types of reasonability checks.
First all signals are compared with the sensor range limits. Secondly, whenever redundant input signals are available for a single parameter, cross
checking is performed to identify significant differences between redundant
signals. The presence of suspect input signals are indicated on all displays
containing parameters derived from those signals.
5.0 SIGNAL ISOLATION
The DCS ERDADS safety related (SR) signals provided for ERDADS use are supplied with SR Foxboro DCS TAs(Termination Assemblies), FBMs (Field Bus Modules), cables and baseplates (qualified for Class 1E service).
6.0 DCS (ERDADS) DATA PROCESSING The Unit 3 and Unit 4 DCS (ERDADS) are completely separate and independent from each other. Redundant pairs of DCS (ERDADS) control processors (CP) are provided for each unit. The CP pairs act in a redundant fashion to provide bumpless transfer to the backup CP upon detection of failure in the master CP. Each CP processes the input data from the associated DCS field input signal modules (analog, digital and pulse), and performs the programmed logic functions to support the graphical displays. The DCS CP pairs communicate in peer-to-peer fashion with each other and with the graphic display workstations via the Plant Data Network.
7A-3 Revised 01/31/2013 C26C26 7.0 PLANT DATA NETWORK The Plant Data Network (PDN) is a redundant and diverse Ethernet switched network. It functions as a communication backbone for the plant DCS (ERDADS), allowing peer-to-peer communication between various DCS control processors, workstations, and archiving historians. The PDN network switches are powered from diverse station inverters to avoid single failure impacts.
A SR inverter provides backup power to the Control Room PDN switches.
Isolating fuses provide the SR to NNS boundary. The Unit 3 and Unit 4 PDNs are isolated from each other.
8.0 DATA LINK PROCESSING
SAS/ERDADS has data interface links to communicate with the Eberline
Computer, the Digital Data Processing System (DDPS), Kaye generator monitors, and Inadequate Core Cooling System (QSPDS) computer. These data interface
links are considered as non-nuclear safety related, with the exception of the
qualified fiber optic link/isolation provided between SAS/ERDADS and QSPDS.
In addition to the fiber optic link, several QSPDS inputs are hard wired to
the SAS isolation cabinets. In accordance with the requirements of NUREG-
1394, unit specific data links are provided between the ERDADS computer
output communication server and the NRC Operations Center. This data link, which would be activated during a site emergency, provides a direct near-
real- time transfer of critical data. The Emergency Response Data System (ERDS) data link has been designed to provide the following data which is
necessary to assess the severity of the accident and the potential public
impact: (1) core and reactor coolant system conditions, (2) conditions
inside containment, (3) radioactivity release rates and (4) meteorological
data. The DCS platform is unit specific system architecture. The only
common system interface with the DCS platform will occur at the Technical
Support Center, Operational Support Center and the Emergency Operations
Facility. This is required since the facilities are shared between the units
for the plant emergency plan.
Various data links provide data to the DCS (ERDADS) processor to support information displays and application programs. These data links are comprised of data/information interface for the NRC for the Emergency Response Data System (ERDS), Eberline Computer (Radiation Monitoring), Kaye Generator Monitors, The Qualified Safety Parameters Display System (QSPDS), Data Link Health Status, Annuciator Alarms and the Metrological Data Link which is transmitted serially by fiber optic modems to the computer room DCS (ERDADS).
7A-4 Revised 01/31/2013 C26C26C26 9.0 GRAPHIC DISPLAY/READOUT EQUIPMENT General purpose displays located in the control room, the computer room, the Technical Support Center and the Emergency Operating Facility, provide access to a comprehensive set of system-oriented mimic, tabular and trend displays.
In addition specific user-defined display capability is available. The DCS (ERDADS) and the DCS (SPDS) graphics are displayed on workstation and flat panel display. A continuous calorimetric flat panel display is also provided in the control room. A DCS (SPDS) touch screen display is used as a SPDS graphics navigation tool. The DCS operator workstations display ERDADS information in the Control Room. QSPDS display 3B is mounted in the Operator Console (3C256).
10.0 SYSTEM POWER Power is supplied to the DCS/SAS/ERDADS through four Uninterruptible Power Systems (UPS), which provide redundant supplies to the computer. Each UPS is
comprised of a distribution panel, a static transfer switch, a static
inverter, and a regulating transformer. The UPS ensures that the
DCS/SAS/ERDADS computer installation is not operationally impaired by power
fluctuations or failures.
Single phase, 120 volt, 60 Hz power is supplied to the distribution panels
through primary or alternate feeds by means of static transfer switches. The
primary feeds are two - 20 kVA and two - 10 kVA static inverters that are
powered by the auxiliary power upgrade 125V DC buses. In the event of a
failure in the primary feed, static transfer to the alternate feed is
accomplished. Manual switching is provided for maintenance purposes. The
alternate feeds are from the Condensate Polishing System Motor Control
Center via 480/120 volt, single phase, regulating transformers (two - 10 kVA
and two - 25 kVA). These power supplies are non-safety related.
The DCS ERDADS and PDN systems are powered from sources that collectively provide a robust power supply capable of withstanding a short duration
(<2hours) Loss of Offsite Power (LOOP) coincident with loss of any single panel, inverter, battery, or AC power feed without the interruption of service. This is accomplished by employing NNS station inverters for powering DCS and PDN equipment. Critical components are provided with redundant power feeds and are designed to automatically switch to the redundant source upon loss of one of the power sources. One of the redundant feeds to each of the PDN Zone switches in the Control Room is provided from a Safety Related inverter to provide PDN functionality during an extended LOOP.
Extended LOOP is that period of time after which the NNS batteries have discharged (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). This will provide additional DCS (ERDADS) functional capability by having one safety related source powering the PDN Zone switches.
7A-5 Revised 01/31/2013 C26C26C26C26
FINAL SAFETY ANALYSIS REPORT FIGURE 7A-1 REFER TO ENGINEERING DRAWING 5613-J-95S-4000 (Unit 3) 5614-J-95S-4000 (Unit 4)
Revised 01/31/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3/4 DCS (ERDADS) CABLE BLOCK DIAGRAM REFERENCE DWG 5613/4-J-95S-4000 FIGURE 7A-1 C26