Knowledge and Abilities Catalog for Nuclear Power Plant Operators

NUREG-2104
Knowledge and Abilities
Catalog for Nuclear Power
Plant Operators
Advanced Boiling Water Reactors
Draft Report for Comment
Office of New Reactors
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NUREG-2104
Knowledge and Abilities
Catalog for Nuclear Power
Plant Operators
Advanced Boiling Water Reactors
Draft Report for Comment
Manuscript Completed: August 2011
Date Published: December 2011
Prepared by
R. Pelton
J. Kellum
Office of New Reactors
COMMENTS ON DRAFT REPORT
Any interested party may submit comments on this report for consideration by the NRC staff.
Comments may be accompanied by additional relevant information or supporting data. Please
specify the report number NUREG-2104, draft, in your comments, and send them by December
31, 2018 to the following address:
Cindy Bladey, Chief
Rules, Announcements, and Directives Branch
Division of Administrative Services
Office of Administration
Mail Stop: TWB-05-B01M
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
Electronic comments may be submitted to the NRC by e-mail at [email protected] or
[email protected].
For any questions about the material in this report, please contact:
Richard Pelton
TWFN Mail Stop 07-D24
U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Phone: 301-415-1028
E-mail: [email protected]
James Kellum
TWFN Mail Stop 07-D24
U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Phone: 301-415-5305
E-mail: [email protected]
Please be aware that any comments that you submit to the NRC will be considered a public
record and entered into the Agencywide Documents Access and Management System
(ADAMS). Do not provide information you would not want to be publically available.
NUREG-2104
ii
ABSTRACT
The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling
Water Reactors (NUREG-2104, Revision 0) provides the basis for the development of contentvalid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs).
The examinations developed using the ABWR Catalog along with the Operator Licensing
Examination Standards for Power Reactors (NUREG-1021) will sample the topics listed under
Title 10, Code of Federal Regulations, Part 55 (10 CFR 55).
The catalog is organized into six major sections: Organization of the Catalog, Generic
Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency
and Abnormal Plant Evolutions, Components and Theory.
This is a new Knowledge and Abilities catalog developed specifically to address the General
Electric Advanced Boiling Water Reactor.
iii
NUREG-2104
TABLE OF CONTENTS
page
ABSTRACT
iii
1.0
ORGANIZATION OF THE CATALOG
1-1
1.1
1.2
1.3
1.4
1.5
1.6
1.7
1.8
1.9
1.10
1.11
1.12
1.13
1.14
1.15
Introduction
Part 55 of Title 10 of the Code of Federal Regulations
RO Written Examination
SRO Written Examination
RO and SRO Operating Test Items
Senior Reactor Operator Limited to Fuel Handling
Organization of the ABWR Catalog
Generic Knowledge and Abilities
Plant Systems
Emergency and Abnormal Plant Evolutions
Components
Theory
Importance Ratings
Rules of Use
General Guidance
1-1
1-1
1-1
1-1
1-1
1-1
1-2
1-2
1-3
1-7
1-10
1-10
1-11
1-11
1-12
2.0
GENERIC KNOWLEDGES AND ABILITIES
2-1
2.1
2.2
2.3
2.4
Conduct of Operations
Equipment Control
Radiation Control
Emergency Procedures / Emergency Plan
2-3
2-7
2-11
2-13
3.0
PLANT SYSTEMS
3.1-1
3.1
Safety Function 1: Reactivity Control
3.1-1
SF1CRD
SF1FMCRD
SF1RFC
SF1RRS
SF1RCIS
SF1SLC
Control Rod Drive System
Fine Motion Control Rod Drive Mechanism
Recirculation Flow Control System
Reactor Recirculation System
Rod Control and Information System
Standby Liquid Control System
3.1-1
3.1-5
3.1-8
3.1-12
3.1-16
3.1-20
3.2
Safety Function 2: Reactor Water Inventory Control
3.2-1
SF2HPCF
SF2CD
SF2RCIC
SF2FW
SF2RWCU
SF2FWC
High Pressure Core Flooder System
Condensate System
Reactor Core Isolation Cooling System
Feedwater System
Reactor Water Cleanup System
Feedwater Control System
3.2-1
3.2-5
3.2-9
3.2-13
3.2-17
3.2-21
v
NUREG-2104
SF2RHRLPFL Residual Heat Removal System: Low Pressure Flooder
Injection Mode
SF2AFI
Alternate Feedwater Injection System
3.2-24
3.2-28
3.3
Safety Function 3: Reactor Pressure Control
3.3-1
SF3ADS
SF3NBS
SF3EHC
3.3-1
3.3-4
SF3SRV
Automatic Depressurization System
Main and Reheat Steam System
Turbine Pressure Control/Steam Bypass and Pressure Control
System
Safety/Relief Valves
3.3-8
3.3-13
3.4
Safety Function 4: Heat Removal from Reactor Core
3.4-1
SF4NBS
SF4MT
SF4RCIC
SF4RRS
SF4RHRSDC
Main and Reheat Steam System
Main Turbine Generator and Auxiliary Systems
Reactor Core Isolation Cooling System
Reactor Recirculation System
Residual Heat Removal System: Shutdown Cooling Mode
3.4-1
3.4-5
3.4-9
3.4-13
3.4-17
3.5
Safety Function 5: Containment Integrity
3.5-1
SF5PCS
SF5LDIS
SF5RPV
SF5HRSPC
SF5RHRSPR
SF5SEC
Primary Containment System and Auxiliaries
Leak Detection and Isolation System
Reactor Vessel Internals
Residual Heat Removal System: Suppression Pool Cooling Mode
Residual Heat Removal System: Drywell and Wetwell Spray Mode
Secondary Containment
3.5-1
3.5-5
3.5-9
3.5-12
3.5-16
3.5-20
3.6
Safety Function 6: Electrical
3.6-1
SF6EPDS
SF6DC
SF6EDGCTG
SF6VAC
SF6I&C
AC Electrical Distribution System
Direct Current Power Supply System
Emergency Generators (Diesel/Combustion Turbine Generators)
Vital AC Power Supply System
Instrumentation and Control Power Supply System
3.6-1
3.6-4
3.6-7
3.6-11
3.6-14
3.7
Safety Function 7: Instrumentation
3.7-1
SF7APR
SF7ATLM
SF7APRM
Automatic Power Regulator System
Automated Thermal Limit Monitor
Average Power Range Monitor/Local Power Range
Monitor System
Nuclear Boiler Instrumentation
Radiation Monitoring System
Reactor Trip and Isolation System
Multi-Channel Rod Block Monitor System
Rod Worth Minimizer System
Startup Range Neutron Monitoring System
3.7-1
3.7-4
SF7NBI
SF7RAD
SF7RTIS
SF7MRBM
SF7RWM
SF7SRNM
NUREG-2104
vi
3.7-7
3.7-11
3.7-14
3.7-17
3.7-21
3.7-24
3.7-27
SF7ATIP
SF7ELCS
SF7PICS
SF7SPTM
SF7RSS
Automated Traversing In-Core Probe System
Engineered Safety Function Logic and Control System
Plant Information and Control System
Suppression Pool Temperature Monitoring System
Remote Shutdown System
3.7-30
3.7-33
3.7-36
3.7-41
3.7-44
3.8
Safety Function 8: Plant Service Systems
3.8-1
SF8FPS
SF8FHS
SF8IAS
SF8BCW
SF8RSW
Fire Protection System
Fuel Handling Equipment
Instrument Air System
Reactor Building Cooling Water System
Reactor Service Water System
3.8-1
3.8-4
3.8-7
3.8-10
3.8-14
3.9
Safety Function 9: Radioactivity Release
3.9-1
SF9OG
SF9HVAC
SF9RAD
SF9RW
SF9RPV
SF9FPC
SF9SGTS
SF9CRHVAC
Offgas System
Plant Ventilation Systems
Radiation Monitoring System
Radwaste System
Reactor Vessel Internals
Fuel Pool Cooling and Clean-up System
Standby Gas Treatment System
Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System
3.9-1
3.9-5
3.9-8
3.9-11
3.9-14
3.9-17
3.9-21
4.0
EMERGENCY PLANT AND ABNORMAL PLANT EVOLUTIONS
4.1-1
4.1
Emergency Plant Evolutions
4.1-1
EPE1001
EPE1002
EPE1003
EPE1004
EPE1005
EPE1006
EPE1007
EPE1008
EPE1009
EPE1010
4.1-1
4.1-3
4.1-5
4.1-6
4.1-7
4.1-8
4.1-9
4.1-12
4.1-13
EPE1011
EPE1012
EPE1013
EPE1014
High Drywell Pressure
High Reactor Pressure
Suppression Pool High Water Temperature
High Drywell Temperature
High Suppression Pool Water Level
Low Suppression Pool Water Level
Reactor Low Water Level
High Secondary Containment Area Temperature
High Secondary Containment Area Radiation Levels
Reactor Building Heating, Ventilation, and Air Conditioning
Exhaust High Radiation
Secondary Containment High Differential Pressure
Secondary Containment High Floor Drain Sump/Area Water Level
Scram Condition and Reactor Power >5% or Unknown
High Off-Site Release Rate
4.2
Abnormal Plant Evolutions
4.2-1
APE2001
Partial or Complete Loss of Forced Core Flow Circulation
4.2-1
vii
3.9-24
4.1-15
4.1-16
4.1-17
4.1-19
4.1-21
NUREG-2104
APE2002
APE2003
APE2004
APE2005
APE2006
APE 2007
APE2008
APE2009
APE2010
APE2011
APE2012
APE2013
APE2014
APE2015
APE2016
APE2017
APE2018
APE2019
APE2020
APE2021
APE2022
APE2023
APE2024
Loss of Main Condenser Vacuum
Partial or Complete Loss of AC Power
Partial or Complete Loss of DC Power
Main Turbine Generator Trip
Reactor Scram
High Reactor Pressure
High Reactor Water Level
Low Reactor Water Level
High Drywell Pressure
High Drywell Temperature
High Suppression Pool Temperature
Inadvertent Reactivity Addition
Incomplete Scram
Control Room Evacuation
High Off-Site Release Rate
Partial or Complete Loss of Reactor Building Cooling Water
Partial or Complete Loss of Instrument Air
Inadvertent Containment Isolation
Loss of Shutdown Cooling
Loss of Control Rod Drive Pumps
Refueling Accidents
Plant Fire On-Site
Generator Voltage and Electric Grid Disturbances
4.2-3
4.2-5
4.2-7
4.2-9
4.2-11
4.2-13
4.2-14
4.2-16
4.2-17
4.2-19
4.2-20
4.2-21
4.2-23
4.2-25
4.2-27
4.2-29
4.2-30
4.2-32
4.2-34
4.2-36
4.2-37
4.2-38
4.2-40
5.0
COMPONENTS
5-1
291001
291002
291003
291004
291005
291006
291007
291008
Valves
Sensors and Detectors
Controllers and Positioners
Pumps
Motors and Generators
Heat Exchangers and Condensers
Demineralizers and Ion Exchangers
Breakers, Relays and Disconnects
5-1
5-2
5-4
5-5
5-6
5-7
5-8
5-9
6.0
THEORY
6.1-1
6.1
Reactor Theory
6.1-1
292001
292002
292003
292004
292005
292006
292007
292008
Neutrons
Neutron Life Cycle
Reactor Kinetics and Neutron Sources
Reactivity Coefficients
Control Rods
Fission Product Poisons
Fuel Depletion and Burnable Poisons
Reactor Operational Physics
6.1-1
6.1-2
6.1-3
6.1-4
6.1-5
6.1-6
6.1-7
6.1-8
NUREG-2104
viii
6.2
Thermodynamics
6.2-1
293001
293002
293003
293004
293005
293006
293007
293008
293009
293010
Thermodynamic Units and Properties
Basic Energy Concepts
Steam
Thermodynamic Process
Thermodynamic Cycles
Fluid Statics
Heat Transfer and Heat Exchangers
Thermal Hydraulics
Core Thermal Limits
Brittle Fracture and Vessel Thermal Stress
6.2-1
6.2-2
6.2-3
6.2-4
6.2-5
6.2-6
6.2-8
6.2-9
6.2-12
6.2-14
ix
NUREG-2104
1.0
ORGANIZATION OF THE CATALOG
1.1
Introduction
The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling
Water Reactors, NUREG-2104, provides the basis for development of content-valid written and
operating licensing examinations for reactor operators (ROs) and senior reactor operators
(SROs). The Catalog is designed to ensure equitable and consistent examinations.
1.2
Part 55 of Title 10 of the Code of Federal Regulations
The catalog is used in conjunction with NUREG-1021 "Operator Licensing Examination
Standards for Power Reactors." NUREG-1021 provides policy and guidance and establishes
the procedures and practices for examining licensees and applicants for RO and SRO licenses
pursuant to 10 CFR 55. All knowledge and abilities (K/As) in this catalog are directly linked by
item number to 10 CFR 55. Throughout the catalog, 10 CFR 55 section references are shown
in parentheses following the appropriate K/A statement, such as (CFR: 41.x / 43.x / 45.x).
1.3
RO Written Examination
The guidance for preparation of RO written examination is presented in NUREG-1021. The
specific items for RO written examinations are specified in 10 CFR 55.41(b).
1.4
SRO Written Examination
The guidance for preparation of the SRO written examination is presented in NUREG-1021. In
addition to the RO items specified in 10 CFR 55.41(b), additional items for SRO written
examinations are presented in 10 CFR 55.43(b).
1.5
RO and SRO Operating Test Items
The items for operating tests for ROs and SROs are presented in 10 CFR 55.45(a). The
guidance for preparation of the operating tests is presented in NUREG-1021. The operating
test should include a representative selection of K/As derived from those items listed in 10 CFR
55.45(a).
1.6
Senior Reactor Operator Limited to Fuel Handling
The specifications for examinations for Senior Operators Limited to Fuel Handling (LSRO) are
provided in NUREG 1021. The LSRO examination process includes both a written examination
and an operating test. This examination and test include, but are not limited to, items
associated with 10 CFR 55.43(b) items 5 through 7, and 10 CFR 55.45(a) items 5 and 6.
1-1
NUREG-2104
1.7
Organization of the ABWR Catalog
This catalog is organized into 6 major sections. K/As are grouped according to the major
section to which they pertain. This organization is outlined below.
1.0
ORGANIZATION OF THE ABWR CATALOG
2.0
GENERIC KNOWLEDGE AND ABILITIES
Conduct of Operations
Equipment Control
Radiation Control
Emergency Procedures / Emergency Plan
3.0
PLANT SYSTEMS
Knowledge Categories (K1 - K6)
Ability Categories (A1 - A4)
4.0
EMERGENCY PLANT AND ABNORMAL PLANT EVOLUTIONS
Knowledge Categories (EK/AK 1 - EK/AK 3)
Ability Categories (EA/AA 1 – EA/AA 2)
5.0
COMPONENTS
Component Knowledge Categories
6.0
THEORY
Reactor Theory Knowledge Categories
Thermodynamics Knowledge Categories
1.8
Generic Knowledge and Abilities
Generic knowledge and abilities are generally administrative knowledges and abilities with
broad application across systems and operations. They are listed in Section 2 of the catalog.
The four (4) categories of generic K/As are:
2.1
2.2
2.3
2.4
Conduct of Operations K/As
Equipment Control K/As
Radiation Control K/As
Emergency Operating Procedures/Plan K/As
The K/As for "Conduct of Operations," are used to evaluate the applicant's knowledge of daily
operation of the facility. The types of information evaluated in this category include shift
turnover, operator responsibilities, and procedure usage.
The K/As for "Equipment Control" are used to evaluate the applicant’s knowledge of the
administrative issues associated with the management and control of plant systems and
equipment. The types of information evaluated in this category include maintenance and
temporary modifications of systems. Fuel handling and refueling K/As are organized into this
topic area due to the equipment control aspect of fuel handling.
NUREG-2104
1-2
The generic K/As for "Radiation Control" are used to evaluate the applicant's knowledge of
radiation protection and radiation (personnel and public). The types of information under in this
category include radiation hazards, radiation work permits, and radiation monitoring systems.
The K/As for "Emergency Operating Procedures/Plan" are used to evaluate the applicant's
general knowledge of emergency operations. The K/As are designed to evaluate knowledge of
the emergency operating procedures use. The emergency plan K/As are used to evaluate the
applicant's knowledge of the plan, including, as appropriate, the RO's or SRO's responsibility to
decide whether it should be executed and the duties assigned under the plan.
1.9
Plant Systems
1.9.1
Plant System Organization by Safety Function
Nine (9) major safety functions must be maintained to ensure safe nuclear power plant
operation. The safety functions groups are:
Safety Function 1
Safety Function 2
Safety Function 3
Safety Function 4
Safety Function 5
Safety Function 6
Safety Function 7
Safety Function 8
Safety Function 9
Reactivity Control
Reactor Water Inventory Control
Reactor Pressure Control
Heat Removal From Reactor Core
Containment Integrity
Electrical
Instrumentation
Plant Service Systems
Radioactivity Release
Plant systems have been included in this catalog based on their relationship and importance to
nine safety functions. Table 1 contains a list of these plant systems, arranged within each
safety function. It should be noted that some plant systems contribute to more than one safety
function.
Each plant system has been assigned a unique alphanumeric identifier. Plant systems K/As are
in Section 3 of this catalog.
Table 1
Plant Systems by Safety Function
3.1
Safety Function 1: Reactivity Control
SF1CRD
SF1FMCRD
SF1RFC
SF1RRS
SF1RCIS
SF1SLC
Control Rod Drive System
Fine Motion Control Rod and Drive Mechanism
Recirculation Flow Control System
Reactor Recirculation System
Rod Control and Information System
Standby Liquid Control System
1-3
NUREG-2104
3.2
Safety Function 2: Reactor Water Inventory Control
SF2HPCF
High Pressure Core Flooder System
SF2CD
Condensate System
SF2RCIC
Reactor Core Isolation Cooling System
SF2FW
Feedwater System
SF2RWCU
Reactor Water Cleanup System
SF2FWC
Feedwater Control System
SF2RHRLPFL Residual Heat Removal: Low Pressure Flooder Injection Mode
SF2AFI
Alternate Feedwater Injection System
3.3
Safety Function 3: Reactor Pressure Control
SF3ADS
SF3NBS
SF3EHC
SF3SRV
Automatic Depressurization System
Main and Reheat Steam System
Turbine Pressure Control/Steam Bypass and Pressure Control System
Safety/Relief Valves
3.4
Safety Function 4: Heat Removal From Reactor Core
SF4NBS
SF4MT
SF4RCIC
SF4RRS
F3RHRSDC
Main and Reheat Steam System
Main Turbine Generator and Auxiliary Systems
Reactor Core Isolation Cooling System
Reactor Recirculation System
Residual Heat Removal System: Shutdown Cooling Mode
3.5
Safety Function 5: Containment Integrity
SF5PCS
SF5LDIS
SF5RPV
SF5RHRSPC
SF5RHRSPR
SF5SEC
Primary Containment System and Auxiliaries
Leak Detection and Isolation System
Reactor Vessel Internals
Residual Heat Removal System: Suppression Pool Cooling Mode
Residual Heat Removal System Drywell/Wetwell Spray Mode
Secondary Containment
3.6
Safety Function 6: Electrical
SF6EPDS
SF6DC
SF6DGCTG
SF6VAC
SF6I&C
AC Electrical Distribution System
Direct Current Power Supply System
Emergency Generators (Diesel/Combustion Turbine Generators)
Vital AC Power Supply System
Instrumentation and Control Power Supply System
3.7
Safety Function 7: Instrumentation
SF7APR
SF7ATLM
SF7APRM
SF7NBI
SF7RMS
Automatic Power Regulator System
Automated Thermal Limit Monitoring System
Average Power Range Monitor/Local Power Range Monitor System
Nuclear Boiler Instrumentation
Radiation Monitoring System
NUREG-2104
1-4
SF7RTIS
SF7MRBM
SF7RWM
SF7SRNM
SF7ATIP
SF7ELCS
SF7PICS
SF7SPTM
SF7RSS
Reactor Trip and Isolation System
Multi-Channel Rod Block Monitor System
Rod Worth Minimizer System
Startup Range Neutron Monitor System
Automated Traversing In-Core Probe System
Engineered Safety Function Logic and Control System
Plant Information and Control System
Suppression Pool Temperature Monitoring System
Remote Shutdown System
3.8
Safety Function 8: Plant Service Systems
SF8FPS
SF8FH
SF8IAS
SF8RBCW
SF8RSW
Fire Protection System
Fuel Handling Equipment
Instrument Air System
Reactor Building Cooling Water System
Reactor Service Water System
3.9
Safety Function 9: Radioactivity Release
SF9OG
SF9HVAC
SF9RMS
SF9RMS
SF9RPV
SF9FPC
SF9SGTS
SF9CRHVAC
Offgas System
Plant Ventilation Systems
Radiation Monitoring System
Radwaste System
Reactor Vessel Internals
Fuel Pool Cooling and Clean-up System
Standby Gas Treatment System
Control Room Habitability Area Heating, Ventilation, and Air Conditioning
System
1.9.2
Knowledge and Ability Stem Statements for Plant Systems
The knowledge and abilities for each plant system are organized into six types of knowledge
and four types of ability. If there are no knowledge or ability statements following a stem
statement, then there are no applicable knowledge or ability statements.
The applicable 10 CFR 55.41 / 43 / and 45 references are included with each stem statement.
In most cases the K/As associated with the stem statements can be used for both the written
examination and the operating test. See Table 2 lists the Plant System Stem Statements:
1-5
NUREG-2104
Table 2
Knowledge and Ability Stem Statements for Plant Systems
Knowledge Stem Statements
K1
Knowledge of the physical or control/protection logic relationships between the
(SYSTEM) and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1
Contains the systems that have a connection to system XXS. The selected systems
listed have either a plant protection/control logic relationship or physical piping
relationship to system XXS. The electrical systems were not included in K 1
because they are addressed in K 2.
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2
Lists the power supplies to system components for which knowledge of power
supplies is testable.
K3
Knowledge of the effect that a loss or malfunction of the (system) will have on the
following:
(CFR: 41.7 / 45.4)
K3
Lists the systems included in K1 that are directly affected by a loss of system XXS.
K4
Knowledge of (SYSTEM) design feature(s) and or interlock(s) which provide for the
following:
(CFR: 41.7)
K4
Contains the plant protection/control design features and interlocks.
K5
Knowledge of the operational implications or cause-effect relationships as they apply
to the (SYSTEM):
(CFR: 41.5 / 45.3)
K5
Contains theoretical concepts related to the operation of the system.
K6
Knowledge of the effect of the following plant conditions, system malfunctions or
component malfunctions will have on the (SYSTEM):
(CFR: 41.7 / 45.7)
K6
Lists the systems included in K1 that will have an effect on system XXS if the listed
system is lost. It also lists the components of system XXS whose failure can affect
the operation of the system XXS.
NUREG-2104
1-6
Ability Stem Statements
A1
Ability to predict and/or monitor changes in parameters associated with operating the
(system) controls including:
(CFR: 41.5 / 45.5)
A1
Lists the parameters monitored to verify proper operation of system XXS.
A2
Ability to (a) predict the impacts of the following on the (system) and (b) based on
those predictions, use procedures to correct, control, or mitigate the consequences
of those abnormal operations:
(CFR:41.5/45.6)
A2
Lists the ability to predict and mitigate the consequences of selected items from K 6.
A3
Ability to monitor automatic operations of the (system) including:
(CFR: 41.7 / 45.7)
A3
Contains the automatic features of system XXS identified in K 4 that can be
monitored from the control room
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4
Contains the features of system XXS listed in A 3 that can also be manually
performed as well as the features of system XXS system can only be manually
performed and monitoring parameters. A4 includes system monitoring associated
with the listed manual actions.
1.10
Emergency and Abnormal Plant Evolutions
Section 4 of this catalog contains emergency plant evolutions and abnormal plant evolutions.
An emergency plant evolution is any condition, event, or symptom which leads to entry into the
plant-specific emergency operating procedures. An abnormal plant evolution is any degraded
condition, event, or symptom not directly leading to an emergency operating procedure entry
condition, but, nonetheless, adversely affecting a safety function. The listing of emergency plant
evolutions and abnormal plant evolutions was developed to include those integrative situations
crossing several plant systems and/or safety functions.
The emergency plant evolution strategies described in the Boiling Water Reactor Owners Group
Emergency Procedures Guidelines, cover five broad areas:
1. Reactor Pressure Vessel Control
2. Reactor Pressure Vessel Control with SCRAM Condition Present and Reactor Power
>5% or Unknown.
3. Primary Containment Control
4. Secondary Containment Control
5. Radioactivity Release Control.
1-7
NUREG-2104
If the operator controls the five broad areas of emergency plant evolutions listed above, the
plant safety functions will be safely maintained. Table 3 contains a list of the emergency plant
evolutions and abnormal plant evolutions covered by this catalog. The emergency plant and
abnormal plant evolutions each have a unique evolution designator.
Table 3
Emergency Plant and Abnormal Plant Evolutions
Emergency Plant Evolutions
EPE1001
EPE1002
EPE1003
EPE1004
EPE1005
EPE1006
EPE1007
EPE1008
EPE1009
EPE1010
EPE1011
EPE1012
EPE1013
EPE1014
High Drywell Pressure
High Reactor Pressure
Suppression Pool High Water Temperature
High Drywell Temperature
High Suppression Pool Water Level
Low Suppression Pool Water Level
Reactor Low Water Level
High Secondary Containment Area Temperature
High Secondary Containment Area Radiation Levels
Reactor Building Heating, Ventilation, and Air Conditioning Exhaust High Radiation
Secondary Containment High Differential Pressure
Secondary Containment High Floor Drain Sump / Area Water Level
Scram Condition and Reactor Power >5% or Unknown
High Off-Site Release Rate
Abnormal Plant Evolutions
APE2001
APE2002
APE2003
APE2004
APE2005
APE2006
APE2007
APE2008
APE2009
APE2010
APE2011
APE2012
APE2013
APE2014
APE2015
APE2016
APE2017
APE2018
APE2019
APE2019
APE2021
APE2022
APE2023
APE2024
Partial or Complete Loss of Forced Core Flow Circulation
Loss of Main Condenser Vacuum
Partial or Complete Loss of AC Power
Partial or Complete Loss of DC Power
Main Turbine Trip
Reactor Scram
High Reactor Pressure
High Reactor Water Level
Low Reactor Water Level
High Drywell Pressure
High Drywell Temperature
High Suppression Pool Water Temperature
Inadvertent Reactivity Addition
Incomplete Scram
Control Room Evacuation
High Off-Site Release Rate
Partial or Complete Loss of Reactor Building Cooling Water
Partial or Complete Loss of Instrument Air
Inadvertent Containment Isolation
Loss of Shutdown Cooling
Loss of Control Rod Drive Pumps
Refueling Accidents
Plant Fire on Site
Generator Voltage and Electric Grid Disturbances
NUREG-2104
1-8
1.10.1 Knowledge and Ability Stem Statements for Emergency and Abnormal Plant Evolutions
The information delineated within each emergency or abnormal evolution is organized into three
(3) different types of knowledge and two (2) different types of ability. If there are no knowledge
or ability statements following a stem statement there is no applicable K/A.
The applicable 10 CFR 55.41, 43, and 45 item numbers are included with each stem statement.
In most cases the K/As associated with the stem statements can be used for both the written
and operating examinations. See Table 4, below:
Table 4
Knowledge and Ability Stem Statements for
Emergency Plant and Abnormal Plant Evolutions
Knowledge Stem Statements
E/AK 1 Knowledge of the operational implications of the following concepts as they apply
to the (Emergency Plant or Abnormal Plant Evolution):
(CFR: 41.8 to 41.10)
E/AK 1 Lists the operational implications applicable to the procedure. These items can
come from the procedure bases, PRA, OE, procedure notes and cautions.
E/AK2
Knowledge of the interrelations between (Emergency Plant or Abnormal Plant
Evolution) and the following:
(CFR: 41.7 / 45.8)
E/AK 2 Lists the systems required to be monitored and/or operated by the procedure.
E/AK 3 Knowledge of the reasons for the following responses as they apply to (Emergency
Plant or Abnormal Plant Evolution):
(CFR: 41.5 / 45.6)
E/AK 3 Lists the actions and bases taken in the procedure.
Ability Stem Statements
E/AA 1 Ability to operate and / or monitor the following as they apply to (Emergency Plant
or Abnormal Plant Evolution):
(CFR: 41.7 / 45.6)
EA 1
Lists the system and/or components required to be monitored and/or operated by
the procedure.
E/AA 2 Ability to determine and / or interpret the following as they apply to (Emergency
Plant or Abnormal Plant Evolution):
(CFR: 41.10 / 43.5 / 45.13)
EA 2
Lists the parameters and/or conditions that are monitored to verify successful
implementation of the procedure.
1-9
NUREG-2104
1.11 Components
Basic components such as valves and pumps are found in many systems. NUREG-1021 lists 8
categories of components. The eight categories of components for which additional knowledge
statements are necessary are listed below and delineated in Section 5 of the ABWR catalog.
The component knowledge statements are more detailed than those provided in the system
listing, yet at the same time they are generic to the component types. Each component group
has a numeric identifier and a 10 CFR 55.41(b) reference. See Table 5, below.
Table 5
Components
291001
291002
291003
291004
291005
291006
291007
291008
1.12
Valves (CFR: 41.3)
Sensors and Detectors (CFR: 41.7)
Controllers and Positioners (CFR: 41.7)
Pumps (CFR: 41.3)
Motors and Generators (CFR: 41.7)
Heat Exchangers and Condensers (CFR: 41.4)
Demineralizers and Ion Exchangers (CFR: 41.3)
Breakers, Relays and Disconnects (CFR: 41.7)
Theory
NUREG-1021 lists theory items. General fundamental knowledge which underlies safe
performance on the job is delineated in Section 6 of the ABWR Catalog. These theory
topics represent general fundamental concepts related to plant operation. Each theory topic
has a numeric identifier. The applicable 10 CFR 55.41(b) reference is provided for Reactor
Theory and Thermodynamics Theory fundamental knowledge areas.
Reactor Theory (CFR: 41.1)
292001
292002
292003
292004
292005
292006
292007
292008
Neutrons
Neutron Life Cycle
Reactor Kinetics and Neutron Sources
Reactivity Coefficients
Control Rods
Fission Product Poisons
Fuel Depletion and Burnable Poisons
Reactor Operational Physics
Thermodynamics Theory (CFR: 41.14)
293001
293002
293003
293004
293005
293006
NUREG-2104
Thermodynamic Units and Properties
Basic Energy Concepts
Steam
Thermodynamic Process
Thermodynamic Cycles
Fluid Statics
1-10
293007
293008
293009
293010
1.13
Heat Transfer and Heat Exchangers
Thermal Hydraulics
Core Thermal Limits
Brittle Fracture and Vessel Thermal Stress
Importance Ratings
Importance, in this context, considers direct and indirect impact of the K/A on safe plant
operation in a manner ensuring personnel and public health and safety. Importance Ratings of
the K/As are given for Reactor Operators, Senior Reactor Operators, and, as appropriate, for a
combined RO/SRO (licensed operator) knowledge next to each knowledge or ability statement
in the catalog. These ratings reflect ratings that were derived by concensus of a panel of utility
experts. The rating scale is presented in Table 6, below.
Table 6
RO and SRO Importance Ratings
Rating
5
4
3
2
1
Importance for safe operation
Essential
Very important
Fairly important
Of limited importance
Insignificant importance
A rating of 2.0 or below represents a statement of limited or insignificant importance for the safe
operation of a plant. Such statements are generally not considered as appropriate content for
NRC licensing examinations. The use of statements having importance ratings less than 2 can
be used on an NRC licensing examination if justified based on plant-specific priorities.
1.14
Rules of Use
To ensure consistency in applying this catalog, the following terms are defined as:
•
“Parameter” any characteristic of a system and/or component that is measured.
•
“Actuation” includes actuation logic, signals, blocks, bypasses, permissives, interlocks,
and resets.
1-11
NUREG-2104
1.15
General Guidance
The following strategies and principles are utilized in this catalog:
•
The use of set points is minimized. Values included are specific to titles or
procedures. If a value included in the catalog changes, the statement is still testable
if it meets the intent of the statement.
•
K/A statement overlap in multiple sections is minimized. K/As are assigned to the
most appropriate section.
•
All importance ratings are single column format except A 2 and Generic K/As and
fuel handling. Fuel handling is not a RO license activity and will have N/A marked in
the RO column.
NUREG-2104
1-12
2.0
GENERIC KNOWLEDGES AND ABILITIES
2.0.1
Knowledge / Ability statements that reference Technical Specifications includes the
Short Term Availability Controls, Core Operating Limits Report, and Offsite Dose
Calculation Manual.
2.0.2
For Knowledge / Ability statements that reference Technical Specifications, the term
“apply” for an RO means to perform the Technical Specifications actions.
2.0.3
Knowledge / Ability statements including the words “such as” list suggested topical
areas as examples and are not intended to be all inclusive.
2.1
Conduct of Operations
2.1.1
Knowledge of conduct of operations requirements.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.2
Knowledge of operator responsibilities during all modes of plant operation.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.3
Knowledge of shift or short-term relief turnover practices.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.4
Knowledge of individual licensed operator responsibilities related to shift
staffing, such as medical requirements, “no-solo” operation, maintenance of
active license status, 10CFR55, etc.
(CFR: 41.10 / 43.2)
IMPORTANCE
RO 3
SRO 4
2.1.5
Ability to use procedures related to shift staffing, such as minimum crew
complement, overtime limitations, etc.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 3
SRO 4
2.1.6
Ability to manage the control room crew during plant transients.
(CFR: 41.10 / 43.5 / 45.12 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.1.7
Ability to evaluate plant performance and make operational judgments based
on operating characteristics, reactor behavior, and instrument interpretation.
(CFR: 41.5 / 43.5 / 45.12 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.1.8
Ability to coordinate personnel activities outside the control room.
(CFR: 41.10 / 45.5 / 45.12 / 45.13)
IMPORTANCE
RO 3
SRO 4
2-1
NUREG-2104
2.1
Conduct of Operations (continued)
2.1.9
Ability to direct personnel activities inside the control room.
(CFR: 41.10 / 45.5 / 45.12 / 45.13)
IMPORTANCE
RO 3
SRO 5
2.1.10
Ability to direct non-licensed personnel activities inside the control room.
(CFR: 41.10 / 43.5 / 45.9 / 45.10)
IMPORTANCE
RO 3
SRO 3
2.1.11
Knowledge of facility requirements for controlling vital / controlled access.
(CFR: 41.10 / 43.5 / 45.9 / 45.10)
IMPORTANCE
RO 3
SRO 3
2.1.12
Knowledge of criteria or conditions that require plant-wide announcements,
such as pump starts, reactor trips, and mode changes, etc.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 3
SRO 3
2.1.13
Knowledge of administrative requirements for temporary management
directives, such as standing orders, night orders, operations memos, etc.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 3
SRO 3
2.1.14
Ability to make accurate, clear, and concise verbal reports.
(CFR: 41.10 / 45.12 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.15
Ability to make accurate, clear, and concise logs, records, status boards, and
reports.
(CFR: 41.10 / 45.12 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.16
Ability to use Visual Display Units and other plant displays to evaluate system
or component status.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.1.17
Ability to use integrated control systems to operate plant systems or
components
(CFR: 41.10 /45.8 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.1.18
Ability to interpret and execute procedure steps.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 5
SRO 5
2.1.19
Ability to verify that a copy of a controlled procedure is the proper revision.
(CFR: 41.10 / 45.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-2
2.1
Conduct of Operations (continued)
2.1.20
Ability to perform specific system and integrated plant procedures during all
modes of plant operation.
(CFR: 41.10 / 43.5 / 45.2 / 45.6)
IMPORTANCE
RO 4
SRO 4
2.1.21
Ability to interpret reference materials, such as graphs, curves, tables, etc.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 4
Knowledge of industrial safety procedures such as rotating equipment,
electrical, high temperature, high pressure, caustic, chlorine, oxygen and
hydrogen, etc.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 3
SRO 4
2.1.22
2.1.23
Knowledge of system purpose and/or function.
(CFR: 41.7)
IMPORTANCE
RO 4
SRO 4
2.1.24
Knowledge of the purpose and function of major system components and
controls.
(CFR: 41.7)
IMPORTANCE
RO 4
SRO 4
2.1.25
Knowledge of how to conduct system lineups, such as valves, breakers,
switches, etc.
(CFR: 41.10 / 45.1 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.1.26
Ability to locate and operate components, including local controls.
(CFR: 41.7 / 45.7)
IMPORTANCE
RO 4
SRO 4
2.1.27
Ability to locate control room switches, controls, and indications, and to
determine that they correctly reflect the desired plant lineup.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 5
SRO 4
2.1.28
Ability to explain and apply system warnings, cautions, and precautions and
limitations.
(CFR: 41.10 / 43.2 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.1.29
Knowledge of reactor coolant system and balance of plant chemistry controls
including parameters measured and reasons for the control.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 3
SRO 4
2-3
NUREG-2104
2.1
Conduct of Operations (continued)
2.1.30
Knowledge of the fuel-handling responsibilities of Senior Reactor Operators
such as assessment of fuel handling equipment surveillance requirement
acceptance criteria, prerequisites for vessel disassembly and reassembly,
decay heat assessment, assessment of surveillance requirement for the
refueling mode, etc.
(CFR: 41.10 / 43.7)
IMPORTANCE
RO N/A
SRO 4
2.1.31
Knowledge of procedures and limitations involved in core alterations.
(CFR: 41.10 / 43.6 / 45.7)
IMPORTANCE
RO 3
SRO 4
2.1.32
Knowledge of procedures, guidelines, or limitations associated with reactivity
management.
(CFR: 41.1 / 43.6 / 45.6)
IMPORTANCE
RO 4
SRO 5
2.1.33
Knowledge of the station's requirements for verbal communications when
implementing procedures.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.1.34
Knowledge of conservative decision making practices.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.1.35
Knowledge of refueling administrative requirements such as approvals
required to amend core loading sheets, etc.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.1.36
Knowledge of the refueling process.
(CFR: 41.2 / 41.10 / 43.6 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.1.37
Knowledge of new and spent fuel movement procedures.
(CFR: 41.10 / 43.7 / 45.13)
IMPORTANCE
RO 3
SRO 3
2.1.38
Ability to use procedures to determine the effects on reactivity of plant
changes, such as reactor coolant system temperature, balance of plant, fuel
depletion, etc.
(CFR: 41.10 / 43.6 / 45.6)
IMPORTANCE
RO 4
SRO 4
2.1.39
Knowledge of Reactor Operator duties in the control room during fuel handling
such as responding to alarms from the fuel handling area, communications
with the refueling floor, systems operated from the control room in support of
fueling operations, and supporting instrumentation.
(CFR: 41.10 / 43.7 / 45.12)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-4
2.1
Conduct of Operations (continued)
2.1.40
Ability to identify and interpret diverse indications to validate the response of
another indication.
(CFR: 41.7 / 43.5 / 45.4)
IMPORTANCE
RO 4
SRO 4
2-5
NUREG-2104
2.2
Equipment Control
2.2.1
Ability to perform pre-startup procedures for the facility, including operating
those controls associated with plant equipment that could affect reactivity.
(CFR: 41.5 / 41.10 / 43.5 / 43.6 / 45.1)
IMPORTANCE
RO 4
SRO 4
2.2.2
Ability to manipulate the controls as required to operate the facility between
shutdown and designated power levels.
(CFR: 41.6 / 41.7 / 45.2)
IMPORTANCE
RO 5
SRO 4
2.2.3
(multi-unit license) Knowledge of the design, procedural, and operational
differences between units.
(CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.2.4
(multi-unit license) Ability to explain the variations in control station/control
room layouts, systems, instrumentation, and procedural actions between units
at a facility.
(CFR: 41.6 / 41.7 / 41.10 / 45.1 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.2.5
Knowledge of the process for making design or operating changes to the
facility such as 10 CFR 50.59 screening and evaluation processes,
administrative process for temporary modifications, administrative processes
for disabling annunciators, administrative processes for the installation of
temporary instrumentation, etc.
(CFR: 41.10 / 43.3 / 45.13)
IMPORTANCE
RO 2
SRO 3
2.2.6
Knowledge of the process for making changes to procedures.
(CFR: 41.10 / 43.3 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.7
Knowledge of the process for conducting special or infrequently performed
tests or evolutions.
(CFR: 41.10 / 43.3 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.8
Knowledge of surveillance procedures.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.2.9
Knowledge of tagging and clearance procedures.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.2.10
Knowledge of the process for controlling equipment configuration or status.
(CFR: 41.10 / 43.3 / 45.13)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-6
2.2
Equipment Control (continued)
2.2.11
Ability to determine the expected plant configuration using design and
configuration control documentation, such as drawings, line-ups, tag-outs, etc.
(CFR: 41.10 / 43.3 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.2.12
Knowledge of the process for managing maintenance activities during power
operations, such as risk assessments, work prioritization, and coordination
with the transmission system operator, etc.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.13
Knowledge of the process for managing maintenance activities during
shutdown operations, such as risk assessments, work prioritization, etc.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.14
Knowledge of maintenance work order requirements.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 2
SRO 3
2.2.15
Knowledge of the process for managing troubleshooting activities.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.16
Knowledge of pre- and post-maintenance operability requirements.
(CFR: 41.10 / 43.2)
IMPORTANCE
RO 3
SRO 4
2.2.17
Knowledge of limiting conditions for operations and safety limits.
(CFR: 41.5 / 43.2 / 45.2)
IMPORTANCE
RO 4
SRO 5
2.2.18
Ability to track Technical Specification limiting conditions for operations.
(CFR: 41.10 / 43.2 / 45.13)
IMPORTANCE
RO 3
SRO 5
2.2.19
Knowledge of the bases in Technical Specifications for limiting conditions for
operations and safety limits.
(CFR: 41.5 / 41.7 / 43.2)
IMPORTANCE
RO N/A
SRO 5
2.2.20
Ability to determine Technical Specification Mode of Operation.
(CFR: 41.7 / 41.10 / 43.2 / 45.13)
IMPORTANCE
RO 4
SRO 5
2-7
NUREG-2104
2.2
Equipment Control (continued)
2.2.21
Ability to analyze the effect of maintenance activities, such as degraded power
sources, on the status of limiting conditions for operations.
(CFR: 41.10 / 43.2 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.2.22
Ability to determine operability and/or availability of safety related equipment.
(CFR: 41.7 / 43.5 / 45.12)
IMPORTANCE
RO N/A
SRO 5
2.2.23
Knowledge of conditions and limitations in the facility license such as
reporting requirements when the maximum licensed thermal power output is
exceeded, administration of fire protection program requirements such as
compensatory actions associated with inoperable sprinkler systems or fire
doors, processes for Technical Specification or FSAR changes, the required
actions associated for not meeting administrative controls listed in Technical
Specification Section 5, etc
(CFR: 41.7 / 41.10 / 43.1 / 45.13)
IMPORTANCE
RO 2
SRO 5
2.2.24
Knowledge of less than or equal to one hour Technical Specification action
statements. (This Knowledge / Ability statement does not include Action
Statements of one hour or less that follow the expiration of a completion time for a
Technical Specification condition for which an Action Statement has already been
entered.).
(CFR: 41.7 / 41.10 / 43.2 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.2.25
Ability to apply Technical Specifications with action statements of less than or
equal to one hour.
(CFR: 41.10 / 43.2 / 43.5 / 45.3)
IMPORTANCE
RO 3.4
SRO 4.7
2.2.26
Ability to determine and/or interpret Technical Specifications with action
statements of greater than one hour.
(CFR: 43.2 / 43.5 / 45.3)
IMPORTANCE
RO N/A
SRO 4.7
2.2.27
Ability to apply Technical Specifications for a system, such as application of
Required Actions and Surveillance Requirements in accordance with rules of
application requirements, application of generic Limiting Condition for
Operation (LCO) requirements (LCO 3.01 thru 3.0.7 and SR 3.01 thru 3.04), etc..
(CFR: 41.10 / 43.2 / 43.5 / 45.3)
IMPORTANCE
RO N/A
SRO 5
2.2.28
Ability to obtain and interpret station electrical and mechanical drawings.
(CFR: 41.10 / 45.12 / 45.13)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-8
2.2
Equipment Control (continued)
2.2.29
Ability to recognize system parameters that are Technical Specifications entrylevel conditions.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3)
IMPORTANCE
RO 4
SRO 5
2.2.30
Knowledge of the process used to track inoperable alarms.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 3
2.2.31
Ability to interpret control room indications to verify the status and operation
of a system, and understand how operator actions and directives affect plant
and system conditions.
(CFR: 41.5 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 4
2-9
NUREG-2104
2.3
Radiation Control
2.3.1
Knowledge of radiation exposure limits under normal or emergency
conditions.
(CFR: 41.12 / 43.4 / 45.10)
IMPORTANCE
RO 3
SRO 4
2.3.2
Ability to use radiation monitoring systems, such as fixed radiation monitors
and alarms, portable survey instruments, personnel monitoring equipment,
etc.
(CFR: 41.11 / 41.12 / 43.4 / 45.9)
IMPORTANCE
RO 3
SRO 3
2.3.3
Ability to approve liquid release permits.
(CFR: 41.13 / 43.4 / 45.10)
IMPORTANCE
RO N/A
SRO 4
2.3.4
Ability to comply with radiation work permit requirements during normal or
abnormal conditions.
(CFR: 41.12 / 45.10)
IMPORTANCE
RO 4
SRO 4
2.3.5
Knowledge of plant operational thresholds which requires radiation protection
personnel to be informed such as plant mode changes, large power changes,
radiological alarms, dosimeter alarms, etc.
(CFR 41.12/43.4/45.10
IMPORTANCE
RO 3
SRO 4
2.3.6
Ability to control radiation releases.
(CFR: 41.11 / 43.4 / 45.10)
IMPORTANCE
RO 4
SRO 4
2.3.7
Knowledge of radiological safety principles pertaining to licensed operator
duties, such as containment entry requirements, fuel handling responsibilities,
access to locked high-radiation areas, aligning filters, etc.
(CFR: 41.12 / 45.9 / 45.10)
IMPORTANCE
RO 3
SRO 4
2.3.8
Knowledge of radiological safety procedures pertaining to licensed operator
duties, such as response to radiation monitor alarms, containment entry
requirements, fuel handling responsibilities, access to locked high-radiation
areas, aligning filters, etc.
(CFR: 41.12 / 43.4 / 45.9 / 45.10)
IMPORTANCE
RO 3
SRO 4
NUREG-2104
2-10
2.3
Radiation Control (continued)
2.3.9
Knowledge of radiation or contamination hazards that may arise during
normal, abnormal, or emergency conditions or activities such as analysis and
interpretation of radiation and activity readings as they pertain to selection of
administrative, normal, abnormal, and emergency procedures, analysis and
interpretation of coolant activity, including comparison to emergency plan/or
regulatory limits.
(CFR: 41.12 / 43.4 / 45.10)
IMPORTANCE
RO 3
SRO 4
2.3.10
Knowledge of radiation monitoring systems, such as fixed radiation monitors
and alarms, portable survey instruments, personnel monitoring equipment,
etc.
(CFR: 41.12 / 43.4 / 45.9)
IMPORTANCE
RO 3
SRO 3
2-11
NUREG-2104
2.4
Emergency Procedures / Emergency Plan
2.4.1
Knowledge of Emergency / Abnormal Operating Procedure entry conditions.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 5
SRO 5
2.4.2
Knowledge of system set points associated with Emergency / Abnormal
Operating Procedure entry conditions.
(CFR: 41.7 / 45.7 / 45.8)
IMPORTANCE
RO 5
SRO 5
2.4.3
Ability to identify post-accident instrumentation.
(CFR: 41.6 / 45.4)
IMPORTANCE
RO 4
SRO 4
2.4.4
Ability to recognize abnormal indications for system operating parameters that
are entry-level conditions for Emergency / Abnormal Operating Procedures.
(CFR: 41.10 / 43.2 / 45.6)
IMPORTANCE
RO 5
SRO 5
2.4.5
Knowledge of the organization of the operating procedures network for
normal, abnormal, and emergency evolutions.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.6
Knowledge of Emergency / Abnormal Operating Procedure mitigation
strategies.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.4.7
Knowledge of how abnormal operating procedures are used in conjunction
with Emergency Operating Procedures.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.4.8
Knowledge of low power/shutdown implications in accident (e.g., loss of
coolant accident or loss of residual heat removal) mitigation strategies.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.9
Knowledge of crew responsibilities during emergency / abnormal operations.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.4.10
Knowledge of crew roles and responsibilities during Emergency / Abnormal
Operating Procedure usage.
(CFR: 41.10 / 45.12)
IMPORTANCE
RO 4
SRO 5
NUREG-2104
2-12
2.4
Emergency Procedures / Emergency Plan (continued)
2.4.11
Knowledge of general guidelines for Emergency / Abnormal Operating
Procedure usage.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 5
2.4.12
Knowledge of Emergency / Abnormal Operating Procedure implementation
hierarchy and coordination with other support procedures or guidelines such
as, operating procedures, abnormal operating procedures, and severe
accident management guidelines, etc.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.13
Knowledge of Emergency / Abnormal Operating Procedure terms and
definitions.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.14
Knowledge of the specific bases for Emergency / Abnormal Operating
Procedures.
(CFR: 41.10 / 43.1 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.15
Knowledge of Emergency / Abnormal Operating Procedure layout, symbols,
and icons.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.16
Knowledge of the operational implications of Emergency / Abnormal Operating
Procedure warnings, cautions, and notes.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.17
Knowledge of the parameters and logic used to assess the status of
Emergency / Abnormal Operating Procedure key parameters to ensure
reactivity control, core cooling and heat removal, reactor coolant system
integrity, containment conditions, radioactivity release control, etc.
(CFR: 41.7 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 5
2.4.18
Knowledge of the bases for prioritizing actions during emergency / abnormal
operations.
(CFR: 41.7 / 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.4.19
Knowledge of the bases for prioritizing emergency operating procedure
implementation.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2-13
NUREG-2104
2.4
Emergency Procedures / Emergency Plan (continued)
2.4.20
Knowledge of emergency operating procedure exit conditions such as an
emergency condition no longer exists or severe accident guideline entry is
required, etc.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.21
Knowledge of fire protection procedures.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.22
Knowledge of facility protection requirements, including fire brigade and
portable firefighting equipment usage.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 3
SRO 4
2.4.23
Knowledge of procedures relating to a security event (non-safeguards
information).
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.24
Knowledge of the emergency plan including emergency plan implementing
procedures.
(CFR: 41.10 / 43.5 / 45.11)
IMPORTANCE
RO 3
SRO 4
2.4.25
Knowledge of events related to system operation/status that must be reported
to internal organizations or external agencies, such as the State, the NRC, or
the transmission system operator.
(CFR: 41.10 / 43.5 / 45.11)
IMPORTANCE
RO 3
SRO 4
2.4.26
Knowledge of annunciator alarms, indications, or response procedures.
(CFR: 41.10 / 45.3)
IMPORTANCE
RO 4
SRO 4
2.4.27
Knowledge of operator response to a loss of all annunciators.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.28
Knowledge of Reactor Operator tasks performed outside the main control
room during an emergency and the resultant operational effects.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
2.4.29
Knowledge of non-license operator tasks during an emergency and the
resultant operational effects.
(CFR: 41.10 / 43.5 / 45.13)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-14
2.4
Emergency Procedures / Emergency Plan (continued)
2.4.30
Knowledge of the lines of authority during implementation of the emergency
plan, emergency plan implementing procedures, emergency operating
procedures, or severe accident guidelines.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.31
Ability to take actions called for in the facility emergency plan and emergency
plan implementing procedures, including supporting or acting as emergency
coordinator if required.
(CFR: 41.10 / 43.5 / 45.11)
IMPORTANCE
RO N/A
SRO 4
2.4.32
Knowledge of Reactor Operator responsibilities in emergency plan
implementation.
(CFR: 41.10 / 45.11)
IMPORTANCE
RO 4
SRO 4
2.4.33
Knowledge of Senior Reactor Operator responsibilities in emergency plan
implementation.
(CFR: 41.10 / 43.5 / 45.11)
IMPORTANCE
RO N/A
SRO 5
2.4.34
Knowledge of the emergency action level thresholds and classifications.
(CFR: 41.10 / 43.5 / 45.11)
IMPORTANCE
RO N/A
SRO 5
2.4.35
Knowledge of emergency response facilities.
(CFR: 41.10 / 45.11)
IMPORTANCE
RO 3
SRO 4
2.4.36
Knowledge of emergency communications systems and techniques.
(CFR: 41.10 / 45.13)
IMPORTANCE
RO 3
SRO 4
2.4.37
Knowledge of emergency plan protective action recommendations.
(CFR: 41.10 / 41.12 / 43.5 / 45.11)
IMPORTANCE
RO N/A
SRO 4
2.4.38
Ability to prioritize and interpret the significance of each annunciator or alarm.
(CFR: 41.10 / 43.5 / 45.3 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.4.39
Ability to verify that the alarms are consistent with the plant conditions.
(CFR: 41.10 / 43.5 / 45.3 / 45.12)
IMPORTANCE
RO 4
SRO 4
2-15
NUREG-2104
2.4
Emergency Procedures / Emergency Plan (continued)
2.4.40
Ability to diagnose and recognize trends in an accurate and timely manner
utilizing the appropriate control room reference material.
(CFR: 41.10 / 43.5 / 45.12)
IMPORTANCE
RO 4
SRO 4
2.4.41
Ability to perform, without reference to procedures, those actions that require
immediate operation of system components and controls.
(CFR: 41.10 / 43.2 / 45.6)
IMPORTANCE
RO 5
SRO 4
2.4.42
Ability to verify system alarm setpoints and operate controls identified in the
alarm response manual.
(CFR: 41.10 / 43.5 / 45.3)
IMPORTANCE
RO 4
SRO 4
NUREG-2104
2-16
3.0
PLANT SYSTEMS
3.1
Safety Function 1: Reactivity Control
System:
SF1CRD Control Rod Drive System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between Control Rod Drive System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Condensate, feedwater and condensate air extraction system
Makeup water condensate system
Feedwater system - control rod drive system return to vessel
Reactor building cooling water system
Reactor trip and isolation system
Instrument air system
Rod control and information system
Plant information and control system
Reactor recirculation system
AC electrical power distribution system
Drywell cooling system
Heating, ventilation, and air conditioning system
Reactor pressure vessel system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
Control rod drive pumps
Scram valve solenoids
Backup scram valve solenoids
Alternate rod insertion valve solenoids
K3
Knowledge of the effect that a loss or malfunction of the Control Rod Drive
System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor internal pumps
Reactor water level
Fine motion control rod drive mechanisms
Reactor water cleanup pumps
IMPORTANCE
3.1-1
3
3
3
3
3
3
3
2
3
3
2
2
3
3
2
4
4
3
3
3
3
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1CRD Control Rod Drive System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Control Rod Drive System design feature(s) and/or interlocks
which provide for the following:
(CFR 41.7)
K4.01
Protection against pump runout during scram conditions
(location of the control rod drive system flow element and a
restricting orifice in the accumulator charging water line)
Fine motion control rod drive mechanisms purge water flow
Scramming control rods with inoperative scram solenoid
valves (back-up scram valves)
Control rod scram
Controlling purge water flow during fine motion control
rod drive mechanisms insertion
Controlling control rod drive system flow
Motor cooling
Auto start of standby control rod drive pump due to low pump
discharge pressure
Auto stop of running control rod drive pump due to low pump
suction pressure
Auto stop of running control rod drive pump due to low lube
oil pressure
Auto start of lube oil pump during control rod drive pump
running due to low lube oil pressure
Controlling control rod drive pump discharge water temperature
(minimum flow bypass valve)
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
K4.12
IMPORTANCE
3
3
4
4
3
3
2
3
3
3
3
2
K5
Knowledge of the operational implications of the following concepts or
cause and effect relationships as they apply to Control Rod Drive System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
Reactor internal pumps purge water
Fine motion control rod drive mechanisms
Reactor water cleanup pump purge water
Fine motion control rod drive purge water header
NUREG-2104
3.1-2
3
3
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1CRD Control Rod Drive System (continued)
K/A NO.
ABILITY
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the
Control
Rod Drive System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
Condensate system
Condensate storage tanks
Plant information and control system
Instrument air system
Reactor trip and isolation system
AC electrical power distribution system
Reactor building cooling water system
Rod control and information system
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Control Rod Drive System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
Control rod drive system flow
Hydraulic control unit pressure/level
Reactor water level
Pump amps
Fine motion control rod drive mechanisms purge water flow
Fine motion control rod drive mechanisms purge water to reactor
differential pressure
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Control Rod Drive System; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.5)
RO SRO
Pumps trips
3
3
Valve closures
3
3
Power supply failures
3
3
Scram conditions
4
4
Discharge filter becoming plugged
3
3
Suction filter becoming plugged
3
3
Flow control valve failure
3
3
Inadequate system flow
3
3
Loss of applicable plant air systems
3
3
Low hydraulic control unit accumulator pressure/high level
4
4
Valve openings
3
3
Low fine motion control rod drive mechanisms purge water flow
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
A2 06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
IMPORTANCE
3.1-3
3
3
3
3
4
3
3
3
3
3
3
2
3
3
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1CRD Control Rod Drive System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Control Rod Drive System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
A3.10
Valve operation
Pump start
System pressure
System flow
Reactor water level
Reactor power
HCU accumulator pressure/level
Fine motion control rod drive mechanisms purge water flow
Indications and alarms
Fine motion control rod drive mechanisms purge water to reactor
differential pressure
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Control rod drive pumps
Control rod drive system flow control valve
NUREG-2104
IMPORTANCE
3.1-4
3
3
3
3
3
3
3
3
3
3
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1FMCRD Fine Motion Control Rod Drive Mechanism
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Fine Motion Control Rod Drive Mechanism and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
Control rod drive system
Rod control and information system
Reactor pressure vessel system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Fine motion control rod drive mechanism motor
K3
Knowledge of the effect that a loss or malfunction of the Fine Motion
Control Rod Drive Mechanism will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Reactor power
Flux shaping
Shutdown margin
K4
Knowledge of Fine Motion Control Rod Drive Mechanism design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
Rod drop prevention function (fine motion control rod drive latches
and rod block signal)
Rod ejection accident prevention (fine motion control rod drive
brake)
Detection of an uncoupled rod
Slowing the drive mechanism near the end of its travel following
a scram
The use of accumulator to scram the control rod
Rod position indication
Uncoupling the control rod from the drive mechanism
Maintaining the control rod at a given location
The use of fine motion control rod drive motor to auto insert control
rod upon failure to insert hydraulically on a scram signal
Detection of control rod drift
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
IMPORTANCE
3
3
3
3
3
3
4
3.1-5
4
3
4
2
4
3
3
3
4
4
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1FMCRD Fine Motion Control Rod Drive Mechanism (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Fine Motion Control Rod Drive Mechanism:
(CFR: 41.5 / 45.3)
K5.01
Reactor pressure vessel water level
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Fine Motion
Control Rod Drive Mechanism:
(CFR: 41.7 / 45.7)
K6.01
K6.02
Control rod drive system
Loss of power to fine motion control rod drive motor
IMPORTANCE
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Fine Motion Control Rod Drive Mechanism controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
Reactor power
Control rod position
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Fine Motion Control Rod Drive
Mechanism; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
Stuck rod
Uncoupled rod
Drifting rod
Control rod pair scram
Reactor scram
Loss of fine motion control rod drive charging water flow
Loss of fine motion control rod drive purge water flow
Low hydraulic control unit accumulator pressure/high water level
Excessive scram time for a given drive mechanism
NUREG-2104
4
3
3.1-6
RO SRO
3
4
4
4
3
4
4
4
4
4
3
3
3
3
4
4
3
4
3.1
Safety Function 1: Reactivity Control
System:
SF1FMCRD Fine Motion Control Rod Drive Mechanism (continued)
K/A NO.
KNOWLEDGE
A3
Ability to monitor automatic operations of the Fine Motion Control Rod
Drive Mechanism including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Fine motion control rod position
Control rod position following a scram
A4
Ability to manually operate and/or monitor the Fine Motion Control Rod
Drive Mechanism in the control room:
(CFR: 41.7 / 45.5 to 45.8)
IMPORTANCE
4
4
None
3.1-7
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RFC Recirculation Flow Control System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Recirculation Flow Control System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Reactor recirculation system
Steam bypass and pressure control system
AC electrical power distribution system
Neutron monitoring system
Rod control and information system
Automatic power regulator system
Feedwater control system
Reactor trip and isolation system
Control rod drive system
Plant information and control system
Reactor water cleanup system
Engineered Safety Function logic and control system
Reactor pressure vessel instrumentation system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Reactor internal pump adjustable speed drives
Reactor internal pump motor generator sets
K3
Knowledge of the effect that a loss or malfunction of the will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Core flow
Reactor power
Reactor water level
Reactor internal pump speed
NUREG-2104
IMPORTANCE
3.1-8
4
3
3
3
3
3
3
3
4
3
3
4
3
3
3
4
4
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1RFC Recirculation Flow Control System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Recirculation Flow Control System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
Reactor internal pump speed control
Signal failure detection
Automatic load following
Minimum and maximum pump speed setpoints
Normal reactor internal pump speed runback
Fast reactor internal pump speed runback
Selected control rod run in
Recirculation pump trip
Reactor internal pump response to a momentary reactor internal
pump -adjustable speed drive voltage drop
Prevention of reactor internal pump speed increase for various
core flow and/or reactor power conditions
Prevention of reactor internal pump speed increase due to automated
thermal limit monitor trip
K4.10
K4.11
IMPORTANCE
3
3
2
3
3
3
3
3
2
3
3
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to the Recirculation Flow Control System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
Reactor power
Reactor core flow
Reactor internal pump MG set
Feedwater flow
Reactor water level
Rod pattern
Reactor internal pump-adjustable speed drives
Technician interface unit
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Recirculation
Flow Control System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
Electrical power distribution system
Recirculation system
Feedwater control system
Low reactor water level
Neutron monitoring signal input
Automatic power regulator system
3.1-9
4
4
4
3
3
3
3
2
3
3
4
3
4
4
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RFC Recirculation Flow Control System (continued)
K/A NO.
KNOWLEDGE
K6.07
Steam bypass and pressure control system
3
K6.08
K6.09
Reactor trip and isolation system
Reactor pressure vessel instrumentation
3
3
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Recirculation Flow Control System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
Reactor internal pump speed
MG set drive motor amps
MG set generator current, power, voltage
Reactor water level
Reactor power
Reactor core flow
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Recirculation Flow Control System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO SRO
Reactor internal pump-adjustable speed drives trip
3
3
Reactor internal pump-adjustable speed drives normal or fast
speed runback
3
3
Loss of AC electrical power distribution system
3
3
Low reactor water level
3
3
Loss of feedwater signal inputs
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
IMPORTANCE
3
3
2
3
4
3
A3
Ability to monitor automatic operations of the Recirculation Flow Control
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Indications and alarms
Reactor internal pump speed
NUREG-2104
3.1-10
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1RFC Recirculation Flow Control System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Recirculation Flow Control
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
MG sets
Adjustable speed drives
Indications and alarms
Reactor internal pumps speed
IMPORTANCE
3.1-11
3
3
3
3
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RRS Reactor Recirculation System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Recirculation System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
Reactor building cooling water system
AC electrical power distribution system
Control rod drive system
Recirculation flow control system
Plant information and control system
Makeup water purified system
Reactor pressure vessel system
Radioactive drain transfer system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Reactor internal pumps
Recirculation system MG sets
Motor generator set oil pumps
K3
Knowledge of the effect that a loss or malfunction of the Reactor
Recirculation System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Core flow
Load following capabilities
Reactor power
Reactor water level
Reactor recirculation system motor generator sets
Vessel bottom head drain temperature
Primary containment integrity
NUREG-2104
IMPORTANCE
3.1-12
3
3
3
4
3
2
3
2
3
3
2
4
3
4
4
3
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1RRS Reactor Recirculation System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Recirculation System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
K4.12
K4.13
K4.14
K4.15
Adequate reactor internal pump NPSH
Reactor internal pump motor cooling
Controlled purge flow
Automatic voltage/frequency regulation
Motor generator set trips
Pump minimum flow limit
Pump start permissives
Minimization of reactor vessel bottom head temperature gradients
End of cycle recirculation pump trip
Anticipated Transient without scram - Recirc pump trip
Selected control rods run in circuitry
Reactor internal pump runback
Reactor internal pump startup
Automatic MG set start sequencing
Core flow rapid reduction logic
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to the Reactor Recirculation System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
Reactor internal pump vibration characteristics
Restart of reactor internal pumps while operating at power
Core flow
Reactor power
Reactor moderator temperature
Reactor pressure
Recirculation flow control system motor-generator sets
Nuclear boiler instrumentation (reactor water level/pressure/core
plate d/p)
Vessel bottom head drain temperature
Residual heat removal shutdown cooling mode
Reactor water level
Anticipated transient without scram circuitry
End-of-cycle recirculation pump trip circuitry
Selected control rods run in circuitry
Recirculation motor inflatable shaft seal subsystem
Recirculation motor cooling subsystem
Reactor internal pump adjustable speed drives
K5.09
K5.10
K5.11
K5.12
K5.13
K5.14
K5.15
K5.16
K5.17
IMPORTANCE
3.1-13
3
3
3
3
3
3
3
3
4
4
4
4
3
3
3
2
3
4
4
3
3
4
3
3
3
4
4
4
4
2
3
3
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RRS Reactor Recirculation System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor
Recirculation System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
Reactor building cooling water systems
AC electrical power distribution system
Control rod drive system
Recirculation system motor-generator sets
Low reactor water level
Recirculation motor inflatable shaft seal subsystem
Makeup water purified system
Reactor internal pump motor cooling subsystem
Turbine trip/load rejection
IMPORTANCE
3
3
3
3
3
3
2
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Recirculation System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
A1.12
A1.13
A1.14
A1.15
A1.16
Reactor internal pump flow
Core flow
Reactor water level
Reactor power
Reactor internal pump motor amps
Reactor internal pump speed
Recirculation cooling water flow
Vessel bottom head drain temperature
Reactor internal pump differential pressure
Reactor internal pump motor temperature
Reactor internal pump MG set temperatures
Reactor internal pump MG drive motor amps
Reactor internal pump MG set generator current, power, voltage
Reactor internal pump motor purge flow
Reactor internal pump vibration
Core differential pressure
NUREG-2104
3.1-14
4
4
3
4
3
3
3
3
3
2
2
2
2
3
2
3
3.1
Safety Function 1: Reactivity Control
System:
SF1RRS Reactor Recirculation System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Recirculation System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO SRO
Recirculation system leak
4
4
Single reactor internal pump trip
4
4
Multiple reactor internal pump trip
4
4
Inadvertent recirculation flow increase
4
4
Inadvertent recirculation flow decrease
4
4
Reactor internal pump speed mismatch
3
3
Reactor internal pump flow mismatch
3
3
Low reactor water level
4
4
Loss of reactor feedwater
4
4
High reactor pressure (ATWS circuitry initiation)
4
4
End of cycle recirculation pump trip circuitry initiation
4
4
Selected control rods run in circuitry actuation
4
4
Loss of motor cooling
3
3
Loss of AC power
3
3
Loss of reactor building cooling water
3
3
Incomplete start sequence
3
3
Loss of reactor internal pump purge flow
3
3
Reactor internal pump speed runback
3
3
Increase in reactor internal pump vibration
2
2
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
A2.18
A2.19
IMPORTANCE
A3
Ability to monitor automatic operations of the Reactor Recirculation
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
Pump/MG set start sequence
System flow
Indications and alarms
Pump speed
Reactor internal pump trips
Reactor internal pump runbacks
Recirculation system motor generator set trip
A4
Ability to manually operate and/or monitor the Reactor Recirculation
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Reactor internal pumps
System flow
Core flow
3
3
3
3
3
3
3
4
4
4
3.1-15
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RCIS Rod Control and Information System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Rod Control and Information System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
Neutron monitoring system
Control rod drive system
Recirculation flow control system
Reactor trip and isolation system
Plant information and control system
Automatic power regulator system
Refueling equipment
AC electrical power distribution system
Vital AC power supply system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
RCIS redundant controllers
Stepping motor driver modules
Rod brake controller cabinets
K3
Knowledge of the effect that a loss or malfunction of the Rod Control and
Information System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
Reactor startup
Reactor shutdown
Flux shaping
Recirculation flow control system
Refueling equipment
Automatic power regulator system
Control rod drive system
Reactor trip and isolation system
K4
Knowledge of Rod Control and Information System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Limiting the effects of a control rod accident
Rod withdrawal block signals
Rod insertion block signals
Automatic control rod run-in following a scram
NUREG-2104
IMPORTANCE
3.1-16
3
4
3
3
3
3
3
2
3
2
3
3
4
3
3
3
3
3
4
3
3
4
4
3
3.1
Safety Function 1: Reactivity Control
System:
SF1RCIS Rod Control and Information System (continued)
K/A NO.
KNOWLEDGE
K4.05
Insertion of selected control rods for core thermal-hydraulic stability
control or loss of feedwater heating event
Alternate control rod insertion on a failure to scram (control rod
run-in on ARI)
K4.06
IMPORTANCE
3
3
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Rod Control and Information System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
Ganged rod withdrawal sequence
Rod gangs
Fine motion control rod drive mechanism
Target rod pattern
Low power setpoint
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Rod Control and
Information System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
Rod position signal
AC electrical power distribution system
Reactor trip and isolation system
Plant information and control system
Neutron monitoring system
Automatic power regulator system
Vital AC power supply system
3
2
4
3
4
3
3
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Rod Control and Information System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
First stage shell pressure/turbine load
Reactor power
Reactor water temperature
3.1-17
3
3
2
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1RCIS Rod Control and Information System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Rod Control and Information System;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO SRO
High flux (neutron monitoring)
4
4
Position indication failure
3
3
Insert block
3
3
Withdraw block
3
3
Local Power Range Monitor upscale/down scale
3
3
Accumulator fault
3
4
Rod uncoupled
4
4
Rod drift
4
4
Emergency rod insertion (SCRRI, ARI, or scram-follow)
4
4
Rod misalignment
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
IMPORTANCE
A3
Ability to monitor automatic operations of the Rod Control and Information
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
Dedicated operator interface indications
Rod display module indications
Verification of proper functioning/operability
Annunciator and alarm signals
Emergency rod insertion (ARI, SCCRI, scram follow function)
Rod movement in automatic mode (step/notch/continuous)
Automatic self-bypass of RCIS protective features (Rod Worth
Minimizer System and Automated Thermal Limit Monitoring System)
A4
Ability to manually operate and/or monitor the Rod Control and Information
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
Dedicated operator interface (switches or screen manipulations)
Back panel switches or screens
Initiation of SCRRI
Control rods in semi-automatic or manual mode
step/notch/continuous)
Bypassing an inoperable control rod
Bypassing a single channel of RCIS
A4.05
A4.06
NUREG-2104
3.1-18
4
4
3
3
4
3
3
4
3
3
3
3
3
3.1
Safety Function 1: Reactivity Control
System:
SF1SLC Standby Liquid Control System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Standby Liquid Control System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Service air system
Makeup water purified system
Reactor pressure vessel system
High pressure core flooder system
Engineered Safety Function logic and control system
Leak detection and isolation system
Sampling system
Neutron monitoring system
AC electrical power distribution system
Post accident monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
Standby liquid control storage tank outlet valve
Standby liquid control pumps
Standby liquid control injection motor operated valves
SLC tank heater power
Standby liquid control lubricating pump
K3
Knowledge of the effect that a loss or malfunction of the Standby Liquid
Control System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Ability to shutdown the reactor in certain conditions
K4
Knowledge of Standby Liquid Control System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
Zero leakage to the reactor (demineralized water)
Component and system testing
Keeping sodium pentaborate in solution
Dispersal of boron upon injection into the vessel
Reactor water cleanup system isolation
System initiation upon operation of SLC controls
Over pressure protection
Automatic standby liquid control system initiation
IMPORTANCE
3.1-19
3
3
4
4
3
3
2
3
2
2
3
3
3
2
2
4
3
3
4
3
4
4
3
4
NUREG-2104
3.1
Safety Function 1: Reactivity Control
System:
SF1SLC Standby Liquid Control System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications and/or cause-effect
relationships as they apply to Standby Liquid Control System:
(CFR: 41.5 / 45.3)
K5.01
Effects of the moderator temperature coefficient of reactivity
on the boron
Chugging (as it pertains to boron mixing)
Shutdown margin
Tank heater operation
K5.02
K5.03
K5.04
IMPORTANCE
3
3
3
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Standby Liquid
Control System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
Service air system
Makeup water purified system
AC electrical power distribution system
High pressure core flooder system
Engineered Safety Function logic and control system
Neutron monitoring system
Leak detection and isolation system
2
3
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Standby Liquid Control System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Tank level
Pump discharge pressure
Valve operations
Pump amps
Reactor power
Reactor water cleanup system lineup
Standby liquid control system lineup
Indications and alarms
NUREG-2104
3.1-20
4
4
4
3
4
4
4
4
3.1
Safety Function 1: Reactivity Control
System:
SF1SLC Standby Liquid Control System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Standby Liquid Control System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
Standby liquid control pump trip
AC electrical power failures
Inadequate system flow
Loss of standby liquid control tank heaters
Failure to scram
A3
Ability to monitor automatic operations of the Standby Liquid Control
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
Pump discharge pressure
Tank level
Injection valve indications
Reactor power
Reactor water cleanup system isolation
Indications and alarms
System initiation
A4
Ability to manually operate and/or monitor Standby Liquid Control System
in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Standby liquid control system controls
Reactor water cleanup system isolation
System initiation
IMPORTANCE
3.1-21
RO SRO
4
4
3
3
3
3
3
3
4
4
4
4
4
4
4
4
4
4
4
4
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2HPCF High Pressure Core Flooder System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the High Pressure Core Flooder System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
Makeup water condensate system
Standby liquid control system
Suppression pool cleanup system
Reactor building cooling water system
Reactor pressure vessel system
High pressure nitrogen gas supply system
Instrument air system
Automatic depressurization system
Remote shutdown system
Engineered Safety Function logic and control system
Residual heat removal system
Reactor core isolation cooling system
AC electrical power distribution system
Plant information and control system
Post accident monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
High pressure core flooder pumps
High pressure core flooder test return to suppression pool valves
High pressure core flooder injection valves
High pressure core flooder CST suction valves
High pressure core flooder suppression pool suction valves
Initiation logic
K3
Knowledge of the effect that a loss or malfunction of the High Pressure Core
Flooder System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor water level
Standby liquid control system
Adequate core cooling
Automatic depressurization system
IMPORTANCE
3.2-1
3
3
2
2
4
2
2
3
3
3
3
2
3
2
2
3
3
3
3
3
3
4
3
4
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2HPCF High Pressure Core Flooder System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of High Pressure Core Flooder System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
Prevents water hammer
Prevents overfilling reactor vessel
Prevents pump over heating
Testable check valve operation
Motor operated valve operation
High pressure core flooder pump operation
Override of reactor water level interlock
K5
Knowledge of the operational implications and/or cause and effect
relationships as they apply to High Pressure Core Flooder System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
Suppression Pool
Emergency core cooling system room coolers
Adequate core cooling
Suppression pool suction strainers
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the High Pressure
Core Flooder System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
AC electrical power distribution system
Makeup water condensate system
Reactor building cooling water system
Suppression pool suction strainer
Suppression pool water level
Engineered Safety Function logic and control system
Remote shutdown system
A1
Ability to predict and/or monitor changes in parameters associated with
operating the High Pressure Core Flooder System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
High pressure core flooder flow
High pressure core flooder pressure
Reactor water level
Reactor pressure
Suppression pool water level
NUREG-2104
IMPORTANCE
3.2-2
3
3
2
2
2
4
3
4
2
4
2
4
3
3
3
3
3
3
4
3
4
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2HPCF High Pressure Core Flooder System (continued)
K/A NO.
KNOWLEDGE
A1.06
A1.07
A1.08
Motor amps
System lineup
Condensate storage tank level
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the High Pressure Core Flooder System ; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO SRO
System initiation
4
4
High pressure core flooder pump trip
4
4
AC electrical failure
3
3
High pressure core flooder line break
3
4
Pump seal failure
3
3
Inadequate system flow
3
3
Loss of room cooling
2
3
Low suppression pool level
3
4
High suppression pool level
3
4
Low condensate storage tank level
3
4
High suppression pool temperature
3
3
Clogged suppression pool suction strainers
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
IMPORTANCE
3
3
3
A3
Ability to monitor automatic operations of the High Pressure Core Flooder
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
Valve operation
Pump start
System pressure
System flow
Reactor water level
Indications and alarms
A4
Ability to manually operate and/or monitor High Pressure Core Flooder
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
A4.06
High pressure core flooder pump
High pressure core flooder suction valves
High pressure core flooder injection valve
High pressure core flooder minimum flow valve
Manual initiation controls (PRA)
High pressure core flooder testable check valve
3
4
3
4
4
4
3.2-3
4
4
4
3
3
2
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2HPCF High Pressure Core Flooder System (continued)
K/A NO.
KNOWLEDGE
A4.07
A4.08
A4.09
A4.10
Line fill
System flow
High pressure core flooder test return valve
High pressure core flooder initiation reset controls
NUREG-2104
IMPORTANCE
3.2-4
3
4
3
4
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2CD Condensate System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Condensate System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
Main turbine
Feedwater system
Condensate demineralizer system
Condensate filter system
Control rod drive system
Extraction steam system
Exhaust hood spray system
Instrument air system
Circulating water system
Turbine building cooling water system
Makeup water condensate system
Radwaste system
Oxygen injection system
Zinc injection system
Hydrogen water chemistry system
AC electrical power distribution system
Heater drains and vent system
Noble metal injection system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
Condensate pumps
Condensate booster pumps
Condensate pump discharge valves
Condensate booster pump discharge valves
K3
Knowledge of the effect that a loss or malfunction of the Condensate System
will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
Main turbine/main generator
Control rod drive system
Feedwater system
Steam jet air ejectors
Gland seal steam system
Reactor water level
IMPORTANCE
3
3
3
3
3
3
2
2
3
3
3
2
3
2
3
3
3
2
3
3
2
2
3
3
4
3
3
4
3.2-5
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2CD Condensate System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Condensate System design feature(s) and/or interlocks which
provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
Condensate or condensate booster pump auto start
CRD pump suction
Condensate or condensate booster pump protection
Maintenance of water quality
Maintenance of 100% system flow if a feedwater string isolates
Cascading heater drains
Initial main condenser vacuum
Non-condensable gas removal
Auto condensate pump trip due to feedwater line break detection
K5
Knowledge of the operational implications and/or cause and effect
relationships as they apply to Condensate System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Steam jet air ejector condenser
Gland seal steam condenser
Reactor water level
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Condensate
System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
Instrument air system
Circulating water system
Extraction steam system
AC electrical power distribution system
Turbine building cooling water systems
Feedwater system
Makeup water condensate system
Main turbine
Main steam system
Condensate demineralizer system
Condensate filter system
NUREG-2104
IMPORTANCE
3.2-6
3
3
3
3
3
2
3
3
4
3
3
4
3
3
3
3
3
3
2
3
3
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2CD Condensate System (continued)
K/A NO.
KNOWLEDGE
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Condensate System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
System flow
Pump amps
System pressure
Hotwell level
Condensate storage tank level
Reactor water level
System lineup
System water quality
Feedwater temperature
Condenser vacuum
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Condensate System; and (b) based on
those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
A2.18
A2.19
A2.20
IMPORTANCE
3
2
3
3
3
4
3
3
3
3
Pump trips
Condensate pump discharge valve closures
Condensate booster pump discharge valve closure
Valve openings
AC power failures
Inadequate system flow
Low hotwell level
High hotwell level
High feedwater heater level
Low feedwater heater level
Main turbine trip
Loss of circulating water system
Loss of turbine building cooling water systems
Loss of instrument air system
Low Condensate storage tank level
Abnormal water quality
High demineralizer differential pressure
Feedwater heater string isolation
Loss of SJAE
Condensate filter high differential pressure
3.2-7
RO SRO
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
2
2
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2CD Condensate System (continued)
K/A NO.
KNOWLEDGE
A3
Ability to monitor automatic operations of the Condensate System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
Valve operation
Pump starts
System pressure
System flow
Indications and alarms
Hotwell level
Feedwater heater level
Feedwater temperature
Feedwater heater drain tank level
A4
Ability to manually operate and/or monitor the Condensate System in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Condensate/ condensate booster pumps
System motor operated valves
Hotwell level controls
NUREG-2104
IMPORTANCE
3.2-8
3
3
3
3
3
3
3
3
3
3
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RCIC Reactor Core Isolation Cooling System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Reactor Core Isolation Cooling System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Makeup water condensate system
Main steam system
Residual heat removal system
Instrument air system
Leak detection and isolation system
Engineered Safety Function logic and control system
Feedwater system
Suppression pool suction strainer
AC electrical power distribution system
DC power supply system
High pressure core flooder system
Radioactive drain transfer system
Heating ventilation and air conditioning system
Post accident monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
RCIC steam admission valves
Condensate storage tank suction valve
Suppression pool suction valve
RCIC test return to suppression pool valves
RCIC initiation signals (logic)
Reactor core isolation cooling drain pump
K3
Knowledge of the effect that a loss or malfunction of the Reactor Core
Isolation Cooling System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor water level
Reactor vessel pressure
Decay heat removal
Adequate core cooling
IMPORTANCE
4
4
3
2
3
3
4
2
2
3
2
2
2
2
3
3
3
2
3
2
4
4
4
4
3.2-9
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Core Isolation Cooling System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
Prevent water hammer
Prevent overfilling reactor vessel
Override of high reactor water level interlock
Prevents pump over heating
Prevents turbine damage
Manual initiation
Alternate supplies of water
K5
Knowledge of the operational implications or cause and effect relationships
as they apply to Reactor Core Isolation Cooling System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Assist core cooling
Suppression pool
Main condenser
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Core
Isolation Cooling System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
AC electrical power distribution system
Instrument air system
Suppression pool water supply
Makeup water condensate system
Main steam system
Suppression pool suction strainer
DC power supply system
Engineered Safety Function logic and control system
High pressure core flooder system
Radioactive drain transfer system
Heating ventilation and air conditioning system
Feedwater system
NUREG-2104
IMPORTANCE
3
3
3
3
3
4
4
4
4
3
3.2-10
3
2
4
4
4
3
3
3
2
2
2
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Core Isolation Cooling System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Reactor core isolation cooling flow
Reactor core isolation cooling pressure
Reactor water level
Reactor pressure
Reactor core isolation cooling turbine speed
Condensate storage tank level
Suppression pool level
Suppression pool temperature
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Core Isolation Cooling System;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO
SRO
System initiation signal
4
4
Reactor core isolation cooling turbine trips
4
4
Injection valve closure
3
3
AC power loss
2
2
DC power loss
3
3
Loss of instrument air system
2
2
Loss of reactor core isolation cooling drain pump
2
2
Turbine control system failures
3
3
Inadequate system flow
3
3
Loss of room cooling
3
3
Steam line break
4
4
Low condensate storage tank level
4
3
High suppression pool level
3
3
Low suppression pool level
3
3
High suppression pool temperature
4
4
High drain tank level
2
2
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
IMPORTANCE
3.2-11
4
3
4
4
4
3
3
4
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Reactor Core Isolation Cooling
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
Valve operation
Turbine startup
System pressure
System flow
Reactor water level
Indications and alarms
A4
Ability to manually operate and/or monitor the Reactor Core Isolation
Cooling System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Reactor core isolation cooling turbine speed
Turbine trip throttle valve reset
System valves
NUREG-2104
IMPORTANCE
4
4
4
4
4
4
3.2-12
4
4
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FW Feedwater System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Feedwater System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
Reactor pressure vessel system
Reactor water cleanup system
Extraction steam system
Condensate system
Instrument air system
AC electrical power distribution system
Feedwater control system
Turbine building cooling water system
Feedwater pump lube oil subsystem
Reactor core isolation cooling system
Residual heat removal system
Heater drains and vent system
Sampling system
Control rod drive system
Plant information and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Feedwater pump(s)
Feedwater pump discharge valves
Reactor feedwater pump auxiliary oil pumps
K3
Knowledge of the effect that a loss or malfunction of the Feedwater System
will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Reactor water level
Feedwater control system
Reactor water cleanup system
Reactor internal pump NPSH
Core inlet subcooling
Condensate system
Reactor core isolation cooling system
Extraction steam system
Residual heat removal system
Reactor power
IMPORTANCE
3.2-13
4
3
3
3
3
3
4
3
3
4
4
2
2
3
3
3
2
2
4
4
3
3
3
3
3
3
3
4
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FW Feedwater System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Feedwater System design feature(s) and/or interlocks which
provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
Auto start of the feedwater pumps
Feedwater heating
Feedwater pump minimum flow
Dispersal of feedwater in the reactor vessel
Feedwater pump protection
Feedwater pump lubrication
Feedwater pump motor cooling
System isolation from the reactor vessel (check valves, double valve
isolation inside/ outside containment)
Feedwater pump runbacks due to ATWS
Reactor internal pump speed runbacks
Condensate pumps trip due to feedwater line break detection
Selected control rod run-in (SCCRI) initiation
Up-rate feedwater flow measurement
K4.09
K4.10
K4.11
K4.12
K4.13
IMPORTANCE
4
3
3
3
3
3
3
3
3
4
3
3
2
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Feedwater System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
Reactor water level
Thermal power calculation
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Feedwater System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
Instrument air system
Condensate system
AC electrical power
Extraction steam
Turbine building cooling water systems
Feedwater control system
Feedwater pump motor ventilation
Feedwater pump lube oil system
DC electrical power
Engineered Safety Function logic and control system
Heater drains
Plant information and control system
NUREG-2104
3.2-14
4
2
3
3
3
3
3
4
2
3
2
3
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FW Feedwater System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Feedwater System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
K1.05
K1.06
Feedwater flow/pressure
Feedwater inlet temperature
FP motor amps
Feedwater heater level
Feedwater pump speed
Feedwater control valve operation
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Feedwater System; and (b) based on those
predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
Feedwater heater isolation
Loss of condensate or condensate booster pump(s)
Loss of extraction steam
Loss of instrument air system
Loss of AC electrical power
Feedwater control system malfunctions
Loss of DC electrical power
Loss of turbine building cooling water system
Loss of feedwater pump
Loss of high pressure drain pump
A3
Ability to monitor automatic operations of the Feedwater System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
A3.10
A3.11
Feedwater pump auto start
Feedwater pump motor amps:
System flow
Reactor water level
Feedwater inlet temperature
Pump discharge pressure
Indications and alarms
Pump trips
Feedwater pump runbacks due to ATWS
Feedwater pump speed
Feedwater pump minimum flow valve position
IMPORTANCE
3.2-15
3
3
3
3
3
2
RO SRO
3
3
4
4
3
3
3
3
3
3
4
4
3
3
3
3
3
3
3
3
3
2
3
4
3
3
3
3
3
3
2
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FW Feedwater System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Feedwater System in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Manually start/control a feedwater pump
Feedwater heater/drain controls
System valves
NUREG-2104
IMPORTANCE
3.2-16
4
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RWCU Reactor Water Cleanup System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Reactor Water Cleanup System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.21
Reactor pressure vessel system
Feedwater system
Reactor building cooling water system
Instrument air system
Service air system
Low conductivity waste system
Leak detection and isolation system
Makeup water condensate system
Residual heat removal system
Sampling system
Control rod drive system
Feedwater control system
Automatic power regulator
Feedwater system
Engineered Safety Function logic and control system
High pressure nitrogen gas supply system
Electrical power distribution system
Plant information and control system
Fuel pool cooling and cleanup system
Post accident monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Reactor water cleanup pumps
Containment isolation valves
K3
Knowledge of the effect that a loss or malfunction of the Reactor Water
Cleanup System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
Reactor water quality
Reactor water level
Reactor building cooling water systems
Reactor water temperature
Area temperature
Area radiation levels
Drywell temperature
Drywell pressure
Automatic power regulator system
IMPORTANCE
3
3
3
3
2
3
4
3
3
3
3
3
3
3
3
2
3
3
2
2
2
2
3.2-17
3
3
2
3
2
3
2
2
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RWCU Reactor Water Cleanup System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Water Cleanup System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
Pump protection
Piping over-pressurization protection
Over temperature protection for system components
System isolation upon-receipt of isolation signals
Double valve isolation from the reactor
Maximize plant efficiency (use of regenerative heat exchanger)
Draining of reactor water to various locations
Head spray to RPV vessel head for head removal prior to refueling
operation
Control of RPV water level during startup and shutdown
Decay heat removal
Control of reactor water quality
K4.09
K4.10
K4.11
IMPORTANCE
3
3
3
4
3
3
3
2
3
2
3
K5
Knowledge of the operational implications or cause and effect relationships
as they apply to Reactor Water Cleanup System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
Main condenser
Reactor water level
Reactor water quality
Initiation of standby liquid control system
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Water
Cleanup System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
K6.14
Reactor building cooling water systems
Main condenser
Radwaste
Instrument air system
Service air system
High pressure nitrogen gas supply system
AC Electrical power distribution system
Feedwater system
Standby liquid control system logic
Leak detection and isolation system
Control rod drive system
Feedwater control system
Engineered Safety Function logic and control system
Plant information and control system
NUREG-2104
3.2-18
3
3
3
4
3
2
2
3
2
2
3
2
3
4
3
3
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RWCU Reactor Water Cleanup System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Water Cleanup System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
Reactor water level
Component cooling water temperature
Reactor water temperature
System flow
System pressure
System temperature
RWCU drain flow
Main condenser hotwell level
Reactor water conductivity
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Water Cleanup System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
Loss of reactor building cooling water
Flow control valve failure
RWCU pump trips
AC failure
Loss of instrument air system
Loss of service air system
Loss of room coolers
Inadequate system flow
Excessive drain flow rates
Signal received which results in a system isolation
System high temperature
Cleanup demineralizer high differential pressure
Loss of purge flow to reactor water cleanup pump
A3
Ability to monitor automatic operations of the Reactor Water Cleanup System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
Reactor water quality
Response to system isolations
Response to interlocks and trips designed to protect system
components
Reactor water temperature
A3.04
IMPORTANCE
3.2-19
3
3
3
3
3
3
3
2
3
RO SRO
3
3
3
3
3
3
3
3
3
3
2
2
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
4
3
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RWCU Reactor Water Cleanup System (continued)
K/A NO.
ABILITY
A3.05
A3.06
Indications and alarms
Reactor water level control by feedwater control system
A4
Ability to manually operate and/or monitor the Reactor Water Cleanup
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
System pumps
Valve controls
Heat exchanger operation
NUREG-2104
IMPORTANCE
3
3
3
3
3
3.2-20
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FWC Feedwater Control System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Feedwater Control System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
Reactor trip and isolation system
Feedwater system
Steam bypass and pressure control system
Recirculation flow control system
Main steam system
Reactor water cleanup
Plant information and control system
Engineered Safety Function logic and control system
Neutron monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Feedwater control system circuitry
Feedwater pump adjustable speed drive
K3
Knowledge of the effect that a loss or malfunction of the Feedwater Control
System will have on the following:
(CFR: 41.7 / 45.4 to 45.8)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Reactor water level
Reactor feedwater system
Recirculation flow control system
Main turbine
Reactor water level indication
Recirculation flow control system
Reactor water cleanup
K4
Knowledge of Feedwater Control System design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4. 01
K4.02
K4.03
K4.04
K4.05
K4.06
Feedwater pump runout protection
Reactor water level setpoint setdown following a reactor scram
Control signal failure
Feedwater pump speed control
Single element control (reactor water level provides the only input)
Three element control (main steam flow, reactor feedwater flow and
reactor water level provide input)
Manual and automatic control of the system
K4.07
IMPORTANCE
3.2-21
4
4
3
3
3
3
3
3
3
2
3
4
4
3
3
3
3
3
3
3
3
4
3
3
4
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FWC Feedwater Control System (continued)
K/A NO.
KNOWLEDGE
K4.08
K4.09
K4.10
Feedwater pump speed lockup
Selection of various instruments to provide reactor water level input
Simultaneous Manual and Auto operation of the system
(i.e. 1 FP in Auto, 1 FP in Manual)
Feedwater runback due to ATWS
K4.11
IMPORTANCE
4
3
3
4
K5
Knowledge of the operational implications or cause and effect relationships
as they apply to Feedwater Control System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
Moisture carryover
Moisture carryunder
Main steam flow
Reactor water level
Reactor feedwater flow
Technician interface unit
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Feedwater Control
System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
AC power
Main steam flow input
Reactor feedwater flow input
Reactor water level input
2
2
3
4
4
2
3
3
3
4
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Feedwater Control System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
Reactor water level
Reactor feedwater flow
Reactor power
Reactor water level control controller indications
Low flow control position.
Motor driven reactor feedwater pump speed
NUREG-2104
3.2-22
4
4
4
4
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2FWC Feedwater Control System (continued)
K/A NO.
KNOWLEDGE
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Feedwater Control System; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
Loss of any number of main steam flow inputs
Loss of any number of reactor feedwater flow inputs
Loss of reactor water level input
Loss of controller signal output
Feedwater runback due to ATWS
A3
Ability to monitor automatic operations of the Feedwater Control System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
Runout flow control
Changes in reactor water level
Changes in main steam flow
Changes in reactor feedwater flow
Changes in reactor power
Reactor water level setpoint setdown following a reactor scram
FWRV lockup
Motor driven feedwater pump speed lockup
A4
Ability to manually operate and/or monitor the Feedwater Control System in
the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
All individual component controllers in the manual mode
All individual component controllers when transferring
from manual to automatic modes
Motor driven feedwater pump lockup reset controls
All individual component controllers when transferring
from automatic to manual mode
Setpoint setdown reset controls
A4.03
A4.04
A4.05
IMPORTANCE
3.2-23
RO SRO
3
3
3
3
4
4
3
3
4
4
3
3
3
3
3
4
4
4
4
4
3
4
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RHRLPFL Residual Heat Removal System: Low Pressure Flooder
Injection Mode
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Residual Heat Removal System: Low Pressure Flooder Injection Mode
and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
K1.25
K1.26
Makeup water condensate system
Automatic depressurization system
AC electrical power distribution system
Nuclear boiler instrumentation system
Instrument air
Radioactive drain transfer system
Reactor building cooling water system
Reactor pressure vessel system
Fire protection system (PRA)
Remote shutdown system
Reactor core isolation cooling
High pressure core flooder system
Fuel pool cooling and cleanup system
Sampling system
High pressure nitrogen gas supply system
Suppression pool temperature monitoring system
Engineered Safety Function logic and control system
Plant information and control system
Leak detection and isolation system
Reactor water cleanup system
Low conductivity waste system
Atmosphere control system
Heating, ventilation and air conditioning system
Suppression pool water drainage system
Post accident monitoring system
Feedwater system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
K2.07
K2.08
Residual heat removal pumps
Residual heat removal injection valves
Residual heat removal pump minimum flow valves
Residual heat removal test return valves
Residual heat removal heat exchanger outlet valves
Residual heat removal heat exchanger bypass valves
Residual heat removal suppression pool suction valves
Initiation logic
NUREG-2104
IMPORTANCE
3.2-24
3
4
4
4
3
2
3
4
3
3
3
3
3
2
2
3
4
3
3
3
2
2
2
2
2
4
4
3
2
2
2
2
2
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RHRLPFL Residual Heat Removal System: Low Pressure Flooder
Injection Mode (continued)
K/A NO.
KNOWLEDGE
K3
Knowledge of the effect that a loss or malfunction of the Residual Heat
Removal System: Low Pressure Flooder Injection Mode will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor water level
Suppression pool level
Automatic depressurization logic
Adequate core cooling
K4
Knowledge of Residual Heat Removal System: Low Pressure Flooder
Injection Mode design feature(s) and/or interlocks which provide for the
following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
Automatic system initiation/ injection
Prevention of piping overpressurization
Pump minimum flow protection
Pump motor cooling
Prevention of water hammer
Adequate pump net positive suction head (interlock valve open)
Emergency diesel generator load sequencing
Pump operability testing
Surveillance for all operable components
Dedicated injection system during automatic system
initiation (injection valve interlocks)
System redundancy
The prevention of leakage to the environment through
Residual heat removal heat exchanger
Operation from remote shutdown panel
Pump runout protection
K4.11
K4.12
K4.13
K4.14
IMPORTANCE
4
4
4
4
4
3
3
3
3
4
4
3
3
4
4
3
4
3
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Residual Heat Removal System: Low Pressure Flooder
Injection Mode:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
Suppression pool
Keep fill
Reactor pressure
Shutdown cooling
Drywell or wetwell spray
Drywell pressure
Reactor water level
4
3
4
4
4
4
4
3.2-25
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RHRLPFL Residual Heat Removal System: Low Pressure Flooder
Injection Mode (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Residual Heat
Removal System: Low Pressure Flooder Injection Mode:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
AC electrical power
Keep fill
Makeup water condensate system
Instrument air system
ECCS room cooling
Nuclear boiler instrumentation
Reactor building cooling water system
Automatic depressurization system
ECCS room integrity
Nuclear boiler system
High pressure nitrogen gas supply system
Engineered Safety Function logic and control system
Remote shutdown system
IMPORTANCE
4
3
3
3
3
3
3
4
3
3
2
4
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Residual Heat Removal System: Low Pressure Flooder
Injection Mode controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Reactor water level
Reactor pressure
System flow
System pressure
Suppression pool level
Motor amps
Emergency diesel generator loading
Reactor building cooling water systems
NUREG-2104
3.2-26
4
4
4
4
4
2
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2RHRLPFL Residual Heat Removal System: Low Pressure Flooder
Injection Mode (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Residual Heat Removal System: Low
Pressure Flooder Injection Mode; and (b) based on those predictions, use
procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
Residual heat removal pump trips
AC failures
Pump seal failure
Inadequate room cooling
Inadequate system flow
Nuclear boiler instrument failures
Motor operated injection valve failures
Pump runout
Initiating logic failure
Loss of coolant accident
Keep fill failure
A3
Ability to monitor auto Residual Heat Removal System: Low Pressure
Flooder Injection Mode automatic operations including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
Valve operation
Pump start
Pump discharge pressure
System flow
Reactor water level
Indications and alarms
System initiation sequence
Emergency diesel generator load sequencing
A4
Ability to manually operate and/or monitor the Residual Heat Removal
System: Low Pressure Flooder Injection Mode in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
A4.06
A4.07
Pumps
System valves
Keep fill
Heat exchanger cooling flow
Manual initiation controls
System reset following automatic initiation
Testable check valves
IMPORTANCE
3.2-27
RO SRO
4
4
4
4
3
3
3
3
3
3
3
4
3
4
3
3
4
4
4
5
3
4
4
4
4
4
4
4
4
4
4
4
3
4
4
4
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2AFI Alternate Feedwater Injection System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Alternate Feedwater Injection System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.06
K1.07
K1.08
K1.09
AC electrical power distribution system
Reactor feedwater system (via RWCU return line)
Radioactive drain transfer system
Makeup water preparation system
Plant information and control system
Main steam system
Atmosphere control system
Nuclear boiler instrumentation system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Alternate feedwater injection pump
3
Alternate feedwater injection system motor operated injection valves 2
K3
Knowledge of the effect that a loss or malfunction of the Alternate Feedwater
Injection System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
Reactor water level
Reactor pressure
K4
Knowledge of Alternate Feedwater Injection System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
Sufficient water capacity to maintain core covered for 24 hours
following the beyond design basis event
Ability for the AFI pumps to take a suction from numerous water
sources
K4.02
IMPORTANCE
3
3
2
3
3
2
3
2
3
3
3
3
K5
Knowledge of the operational implications or cause and effect relationships
as they apply to Alternate Feedwater Injection System):
(CFR: 41.5 / 45.3)
K5.01
K5.02
Reactor pressure vessel water level
Reactor pressure vessel pressure
NUREG-2104
3.2-28
3
3
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2AFI Alternate Feedwater Injection System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect that following plant conditions, system malfunctions
or component malfunctions will have on the Alternate Feedwater Injection
System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
AC electrical power distribution system
Reactor feedwater system
Radioactive drain transfer system
Makeup water preparation system
Main steam system
Atmosphere control system
Nuclear boiler instrumentation system
IMPORTANCE
3
3
2
3
2
3
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Alternate Feedwater Injection System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
Reactor vessel water level
Reactor vessel pressure
Suppression chamber pressure
Suppression chamber level
Demineralized water storage tank level
System flow
Alternate feedwater pump discharge pressure
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Alternate Feedwater Injection System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
Trip of a running alternate feedwater injection pump
Low demineralized storage tank level
3.2-29
4
4
3
3
3
3
2
RO SRO
4
4
3
3
NUREG-2104
3.2
Safety function 2: Reactor Water Inventory Control
System:
SF2AFI Alternate Feedwater Injection System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Alternate Feedwater Injection
System including:
(CFR: 41.7 / 45.7)
A3.01
Automatic closing of the solenoid operated vent valves when
the respective motor operated injection valve opens
A4
Ability to manually operate and/or monitor the Alternate Feedwater Injection
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
IMPORTANCE
None
NUREG-2104
3.2-30
3
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3ADS Automatic Depressurization System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
Automatic Depressurization System and the following systems:
(CFR: 41.2 to 41.9 /45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
Residual heat removal system
High pressure core flooder system
Nuclear boiler instrumentation system
High pressure nitrogen gas supply system
Neutron monitoring system
Engineered Safety Function logic and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
ADS logic
ADS valve solenoids
K3
Knowledge of the effect that a loss or malfunction of the Automatic
Depressurization System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Restoration of reactor water level after a break that
does not depressurize the reactor when required
Ability to rapidly depressurize the reactor
K3.02
IMPORTANCE
4
4
4
3
3
3
3
3
4
4
K4
Knowledge of Automatic Depressurization System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Allows manual initiation of ADS logic
4
ADS logic initiation
4
Insures adequate nitrogen supply to ADS valves
4
Auto inhibit of ADS with power above Average Power Range Monitor downscale
setpoint
3
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Automatic Depressurization System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
ADS logic operation
Drywell/containment pressure
Safety/relief valves
Suppression pool
Reactor pressure vessel
4
4
4
3
3
3.3-1
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3ADS Automatic Depressurization System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Automatic
Depressurization System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
RHR pump discharge pressure
High pressure core flooder pump discharge pressure
Nuclear boiler instrumentation system (level indication)
Nitrogen supply to ADS valves
Direct current power supply system
Primary containment instrumentation
Average Power Range Monitor downscale signal to ADS logic
Engineered Safety Function logic and control system
IMPORTANCE
4
4
4
4
4
3
4
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Automatic Depressurization System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
ADS valve tail pipe temperatures
ADS valve nitrogen supply pressure
Reactor pressure
Reactor water level
Suppression pool temperature
Average Power Range Monitor indication
Suppression pool level
ADS valve position indication
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Automatic Depressurization System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
Small steam line break LOCA
Large break LOCA
Loss of nitrogen supply to ADS valves
ADS failure to initiate
Loss of DC power to ADS valves
ADS initiation signals present
Failure to auto inhibit
NUREG 2104
3.3-2
3
3
4
4
4
3
3
3
RO SRO
4
4
4
4
3
4
4
4
3
4
4
4
3
3
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3ADS Automatic Depressurization System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Automatic Depressurization
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
ADS valve operation
ADS valve tail pipe temperatures
Drywell/wetwell pressure
Suppression pool level
Suppression pool temperature
Indication and alarms
Reactor pressure
Reactor vessel water level
A4
Ability to manually operate and/or monitor the Automatic Depressurization
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
ADS valves (PRA)
ADS logic initiation controls (PRA)
ADS logic reset controls
ADS inhibit controls
IMPORTANCE
3.3-3
4
4
4
4
4
4
4
4
4
4
4
4
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3NBS Main and Reheat Steam System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Main and Reheat Steam System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
Reactor pressure vessel system
Main turbine
Moisture separator reheaters
Steam bypass and pressure control system
Offgas system
Condenser air removal system
Steam seal/gland seal system
Extraction steam system
High pressure heater drains and vents
Instrument air system
High pressure nitrogen gas supply system
Plant information and control system
Process radiation monitoring system
Containment system
Reactor core isolation cooling system
Feedwater control system
Reactor trip and isolation system
Leak detection and isolation system
Main condenser
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Main steam isolation valve solenoids
K3
Knowledge of the effect that a loss or malfunction of the Main and Reheat
Steam System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Turbine generator
Condenser
Containment
Decay heat removal
Reactor core isolation cooling system
Moisture separator reheaters
Feedwater control system
Safety/relief valves
Reactor vessel and internals
Steam jet air ejectors
NUREG 2104
IMPORTANCE
3.3-4
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
4
4
3
3
3
3
3
3
3
2
4
4
3
3
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3NBS Main and Reheat Steam System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Main and Reheat Steam System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Automatic isolation of steam lines
Automatic isolation and opening of drain valves
Insures that steam released from a steam line break
will not bypass suppression pool
Limits steam flow during a steam line rupture to 200%
Steam flow measurement
Over pressure control
Removal of non condensable gases from reactor head area
Equalization of pressure across the MSIV's before opening
Moisture removal from steam lines prior to admitting steam
Turbine bypass valve inhibit when MSIVs are closed
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
IMPORTANCE
4
3
3
3
3
4
3
3
3
3
K5
Knowledge of the operational implications or cause and effect relationships
as they apply to Main and Reheat Steam System:
(CFR: 41.5 / 45.3)
K5.01
Definition and reason for steam blanketing of moisture separator
reheater
Air/nitrogen operated MSIV's
Decay heat removal
Safety/Relief valves
Head vent
K5.02
K5.03
K5.04
K5.05
2
3
3
4
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Main and Reheat
Steam System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
AC electrical power distribution system
Instrument air system
High pressure nitrogen gas supply system
Safety/relief valve safety function operability
Relief valve operability
Steam line integrity
MSIV isolation signal
Main condenser vacuum
Leak detection and isolation system
Plant information and control system
Steam bypass and pressure control system
3.3-5
3
3
4
3
3
4
4
3
4
3
3
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3NBS Main and Reheat Steam System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Main and Reheat Steam System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
Main steam pressure
Main steam temperature
Reheat steam pressure
Reheater temperature
Main steam line radiation monitors
Air ejector process radiation monitor
Reactor water level
Reactor pressure
Main steam flow
Reactor power
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Main and Reheat Steam System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
IMPORTANCE
Change in steam demand and its effect on reactor pressure and
power
MSIV closure
Main steam line low pressure
Main steam line high radiation
Main steam tunnel high temperature or differential high temperature
Closure of one or more MSIV's at power
Steam line break
Leak detection and isolation system actuation
High reactor water level
Inadvertent initiation of HPCF (steam quality and steam flow)
4
3
2
2
4
3
4
4
3
4
RO SRO
4
4
4
4
4
4
4
4
4
3
A3
Ability to monitor automatic operations of the Main and Reheat Steam
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
Isolation of main steam system
Opening and closing of drain valves as turbine load changes
Moisture separator reheat steam supply
Isolation of moisture separator reheater
NUREG 2104
3.3-6
4
3
3
3
4
4
4
4
4
4
4
4
4
4
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3NBS Main and Reheat Steam System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Main and Reheat Steam
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Main steam isolation valves
Main steam line drain valves
IMPORTANCE
4
3
3.3-7
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3EHC Turbine Pressure Control/Steam Bypass and Pressure Control
System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Turbine Pressure Control/Steam Bypass and Pressure Control System
and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
Reactor trip and isolation system
AC electrical power distribution system
DC power supply system
Turbine building cooling water systems
Recirculation flow control system
Automatic power regulator system
Power generation control system
Plant information and control system
Feedwater control system
Nuclear boiler instrumentation system
Main turbine
Main generator
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Hydraulic power unit pumps
Turbine pressure control logic
Steam bypass and pressure control logic
K3
Knowledge of the effect that a loss or malfunction of the Turbine Pressure
Control/Steam Bypass and Pressure Control System will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
Reactor power
Reactor pressure vessel pressure
Reactor pressure vessel water level
Reactor steam flow
Main turbine steam flow
Main turbine bypass valves
Main turbine stop valves
Main turbine control valves
Combined intermediate valves
Reactor trip and isolation system
Turbine chest warming
Main turbine acceleration
Main turbine speed
NUREG 2104
IMPORTANCE
3.3-8
4
3
3
3
3
4
3
3
3
3
4
3
2
2
2
4
4
4
4
4
4
3
4
3
4
3
3
3
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3EHC Turbine Pressure Control/Steam Bypass and Pressure Control
System (continued)
K/A NO.
KNOWLEDGE
K3.14
K3.15
K3.16
K3.17
K3.18
K3.19
K3.20
K3.21
K3.23
K3.24
K3.25
Main turbine inlet pressure
Main turbine trip
Recirculation flow control system
Main generator
Turbine trip testing
Reactor heatup
Reactor cooldown
Turbine protection
Turbine monitoring
Automatic power regulator system
Plant information and control system
K4
Knowledge of Turbine Pressure Control/Steam Bypass and Pressure Control
System design feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
Reactor pressure control
Turbine speed control
Turbine acceleration control
Reactor scram
Turbine trip
Turbine chest warming
Turbine trip testing
Automatic hydraulic power unit pump start
Reactor cooldown
Turbine protection
Main turbine bypass valve control
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Turbine Pressure Control/Steam Bypass and Pressure Control
System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
K5.09
K5.10
Reactor power vs. reactor pressure
Turbine inlet pressure vs. reactor pressure
Turbine speed measurement
Reactor power
Reactor pressure vessel pressure
Reactor pressure vessel water level
Reactor pressure vessel steam flow
Main turbine steam flow
Main turbine bypass valves
Main turbine stop valves
IMPORTANCE
3.3-9
3
3
3
3
3
3
3
3
2
4
3
4
3
3
4
4
2
3
3
3
3
4
4
3
2
4
4
4
4
4
4
4
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3EHC Turbine Pressure Control/Steam Bypass and Pressure Control
System (continued)
K/A NO.
KNOWLEDGE
K5.11
K5.12
K5.13
K5.14
K5.15
K5.16
K5.17
K5.18
K5.19
K5.20
K5.21
K5.22
K5.23
Main turbine control valves
Combined intermediate valves
Bearing oil
Turbine chest warming
Main turbine acceleration
Main turbine speed
Main turbine inlet pressure
Main turbine trip
Main condenser vacuum
Reactor startup
Reactor heatup
Reactor cooldown
Main turbine protection
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Turbine Pressure
Control/Steam Bypass and Pressure Control System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
K6.14
K6.15
K6.16
K6.17
AC electrical power distribution system
Turbine building cooling water system
Condenser vacuum
Reactor pressure vessel pressure
Reactor power
Main turbine steam flow
Main turbine bypass valves
Main turbine stop valves
Main turbine control valves
Combined intermediate valves
Main turbine speed signal
Main generator
Reactor trip and isolation system
Power generation control system
Nuclear boiler instrumentation system
Feedwater control system
Automatic power regulator system
NUREG 2104
IMPORTANCE
3.3-10
4
3
3
2
3
3
3
3
3
3
3
3
3
3
3
3
4
4
3
4
3
3
3
2
3
3
3
3
3
3
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3EHC Turbine Pressure Control/Steam Bypass and Pressure Control
System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Turbine Pressure Control/Steam Bypass and Pressure Control
System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
A1.12
A1.13
A1.14
A1.15
A1.16
A1.17
A1.18
A1.19
A1.20
A1.21
Reactor pressure vessel pressure
Reactor power
Reactor vessel pressure water level
Main turbine inlet pressure
Reactor steam flow
Ma in turbine steam flow
Main turbine bypass valve position
Main turbine control valve position
Main turbine stop valve position
Combined intermediate valve position.
Reactor/turbine pressure regulating system oil pressure
Reactor/turbine pressure regulating system load set/reference
Main turbine speed
Pressure setpoint/pressure demand
Maximum combined flow limit
Load limit set
Main condenser vacuum
Reactor cooldown
Main turbine vibration
Main turbine eccentricity
Main turbine expansion
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Turbine Pressure Control/Steam Bypass
and Pressure Control System; and (b) based on those predictions, use
procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
High reactor pressure
Failed open/closed bypass valve(s)
Failed open/closed main turbine control valve(s)
Failed open/closed main turbine stop valve(s)
Loss of main condenser vacuum
Main turbine overspeed
Loss of generator load
Loss of AC electrical power
IMPORTANCE
3.3-11
4
4
4
3
4
3
4
3
3
3
3
3
3
3
3
3
3
3
3
3
3
RO SRO
4
4
4
4
4
4
4
4
4
4
3
3
3
4
3
3
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3EHC Turbine Pressure Control/Steam Bypass and Pressure Control
System (continued)
K/A NO.
ABILITY
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
Loss of turbine building cooling water systems
Loss of main turbine speed feedback
Loss of reactor pressure signal)
Main turbine trip
Main generator trip
Reactor scram
Main turbine high vibration
Main turbine high eccentricity
Main turbine high differential expansion
A3
Ability to monitor automatic operations of the Turbine Pressure
Control/Steam Bypass and Pressure Control System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
A3.10
Main turbine speed control
Main turbine acceleration control
Reactor pressure vessel pressure control
Main turbine bypass valve operation
Main turbine control valve operation
Main turbine stop valve operation
Combined intermediate valve operation
Turbine trip testing
Indications and alarms
Main turbine startup
A4
Ability to manually operate and/or monitor the Turbine Pressure
Control/Steam Bypass and Pressure Control System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Main turbine bypass valves
Turbine trip
Turbine panel controls
NUREG 2104
IMPORTANCE
3.3-12
RO SRO
3
3
2
2
3
3
4
4
4
4
4
4
3
3
3
3
2
3
3
3
3
4
3
3
3
3
3
3
4
4
4
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3SRV Safety/Relief Valves
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships between
the Safety/Relief Valves and the following systems
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Main steam system
SPDS
Nuclear boiler instrumentation system
High pressure nitrogen gas supply system
Automatic depressurization system
Remote shutdown system
Direct current power supply system
Post accident monitoring system
Primary containment system
Alternate feedwater system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
SRV solenoids
K3
Knowledge of the effect that a loss or malfunction of the Safety/Relief Valves
will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Reactor pressure control
Reactor over pressurization
Ability to rapidly depressurize the reactor
K4
Knowledge of Safety/Relief Valves design feature(s) and/or interlocks which
provide for the following:
(CFR: 41.7)
K4.01
Prevents siphoning of water into SRV discharge piping and limits
loads on subsequent actuation of SRV's
Ensures even distribution of heat load to suppression pool, and
adequate steam condensing
Allows for SRV operation from more than one location
Detection of valve leakage
Minimum steam pressure required to keep SRV open or to open SRV
Opening of the SRV from either an electrical or mechanical signal
Manual opening of the SRV
Methods for determining position of SRV
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
IMPORTANCE
4
3
4
3
4
3
3
2
3
3
3
3.3-13
4
4
4
3
3
4
4
3
4
4
3
NUREG 2104
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3SRV Safety/Relief Valves (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Safety/Relief Valves:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
Relief function of SRV operation
Safety function of SRV operation
Tail pipe temperature monitoring
Discharge line quencher operation
Vacuum breaker operation
Suppression pool
Reactor pressure vessel overpressurization event
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Safety/Relief
Valves:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
Nuclear boiler instrumentation system (pressure indication)
High pressure nitrogen gas supply system
Direct current power supply system
Discharge line vacuum breaker
Remote shutdown system
IMPORTANCE
3
4
3
3
3
4
4
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Safety/Relief Valves controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
Tail pipe temperature
Nitrogen supply
Reactor pressure
Reactor water level
Reactor power
Turbine load
Suppression pool water temperature
NUREG 2104
3.3-14
3
3
4
4
4
3
4
3.3
Safety Function 3: Reactor Pressure Control
System:
SF3SRV Safety/Relief Valves (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Safety/Relief Valves; and b) based on
those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
Stuck open vacuum breakers
Leaky SRV
Stuck open SRV
ADS actuation
Low reactor pressure
Reactor high pressure
A3
Ability to monitor automatic operations of the Safety/Relief Valves including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
SRV operation after ADS actuation
SRV operation on high reactor pressure
Tail pipe temperatures
Suppression pool temperature
Reactor pressure
Reactor water level
Indications and alarms
A4
Ability to manually operate and/or monitor the Safety/Relief Valves in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
SRV's
IMPORTANCE
RO SRO
3
3
3
3
4
4
4
4
3
3
4
4
4
4
4
4
4
4
4
4
3.3-15
NUREG 2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4NBS Main and Reheat Steam System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Main and Reheat Steam System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
Reactor pressure vessel system
Main turbine
Moisture separator reheaters
Steam bypass and pressure control system
Offgas system
Condenser air removal system
Steam seal/gland seal system
Extraction steam system
High pressure heater drains and vents
Instrument air system
High pressure nitrogen gas supply system
Plant information and control system
Process radiation monitoring system
Containment system
Reactor core isolation cooling system
Feedwater control system
Reactor trip and isolation system
Leak detection and isolation system
Main condenser
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Main steam isolation valve solenoids
K3
Knowledge of the effect that a loss or malfunction of the Main and Reheat
Steam System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Turbine generator
Condenser
Containment
Decay heat removal
Reactor core isolation cooling system
Moisture separator reheaters
Feedwater control system
Safety/relief valves
Reactor vessel and internals
Steam jet air ejectors
IMPORTANCE
3.4-1
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
4
4
3
3
3
3
3
3
3
2
4
4
3
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4NBS Main and Reheat Steam System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Main and Reheat Steam System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Automatic isolation of steam lines
Automatic isolation and opening of drain valves
Insures that steam released from a steam line break
will not bypass suppression pool
Limits steam flow during a steam line rupture to 200%
Steam flow measurement
Over pressure control
Removal of non condensable gases from reactor head area
Equalization of pressure across the MSIV's before opening
Moisture removal from steam lines prior to admitting steam
Turbine bypass valve inhibit when MSIVs are closed
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
IMPORTANCE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Main and Reheat Steam System:
(CFR: 41.5 / 45.3)
K5.01
Definition and reason for steam blanketing of moisture separator
reheater
Air/nitrogen operated MSIVs
Decay heat removal
Safety/Relief valves
Head vent
K5.02
K5.03
K5.04
K5.05
4
3
3
3
3
4
3
3
3
3
2
3
3
4
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Main and Reheat
Steam System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
AC electrical power distribution system
Instrument air system
High pressure nitrogen gas supply system
Safety/relief valve safety function operability
Relief valve operability
Steam line integrity
MSIV isolation signal
Main condenser vacuum
Leak detection and isolation system
Plant information and control system
Steam bypass and pressure control system
NUREG 2104
3.4-2
3
3
4
3
3
4
4
3
4
3
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4NBS Main and Reheat Steam System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Main and Reheat Steam System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
Main steam pressure
Main steam temperature
Reheat steam pressure
Reheater temperature
Main steam line radiation monitors
Air ejector process radiation monitor
Reactor water level
Reactor pressure
Main steam flow
Reactor power
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Main and Reheat Steam System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
IMPORTANCE
4
3
2
2
4
3
4
4
3
4
Change in steam demand and its effect on reactor pressure
and power
MSIV closure
Main steam line low pressure
Main steam line high radiation
Main steam tunnel high temperature or differential high temperature
Closure of one or more MSIV's at power
Steam line break
Leak detection and isolation system actuation
High reactor water level
Inadvertent initiation of HPCF/RCIC (steam quality and steam flow)
RO SRO
4
4
4
4
4
4
4
4
4
3
A3
Ability to monitor automatic operations of the Main and Reheat Steam
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
Isolation of main steam system
Opening and closing of drain valves as turbine load changes
Moisture separator reheat steam supply
Isolation of moisture separator reheater
3.4-3
4
4
4
4
4
4
4
4
4
4
4
3
3
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4NBS Main and Reheat Steam System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Main and Reheat Steam
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Main steam isolation valves
Main steam line drain valves
NUREG 2104
IMPORTANCE
3.4-4
4
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4MT Main Turbine Generator and Auxiliary Systems
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Main Turbine Generator and Auxiliary Systems and the
following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
AC electrical power distribution system
Condensate system
Main steam system
Reactor trip and isolation system
Extraction steam system
Turbine building cooling water system
Instrument air system
Steam bypass and pressure control system
DC power supply system
Generator cooling system
Hydrogen gas cooling system
Generator sealing oil system
Turbine lubricating oil system
Plant information and control system
Turbine control system
Turbine gland steam system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
Generator cooling pumps
Turbine lube oil pumps
Generator sealing oil pumps
Turbine supervisory instrumentation
K3
Knowledge of the effect that a loss or malfunction of the Main Turbine
Generator and Auxiliary Systems will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
AC electrical power distribution system
Reactor pressure
Reactor power
Feedwater system (feedwater heaters)
Condensate system
Reactor trip and isolation system
Steam bypass and pressure control system
Turbine control system
IMPORTANCE
3.4-5
3
3
3
4
3
3
3
3
3
3
3
3
3
3
3
3
2
2
2
2
4
4
4
3
3
4
4
4
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4MT Main Turbine Generator and Auxiliary Systems (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Main Turbine Generator and Auxiliary Systems design
feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
Bearing lubrication
Generator cooling
Sealing to prevent hydrogen leakage
Hydrogen cooling
Turbine protection
Generator protection
Generator voltage regulation
Moisture removal from turbine steam
Turbine control
Extraction steam
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Main Turbine Generator and Auxiliary
Systems:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
Turbine operation and limitations
Turbine speed measurement
Relief valve operation
Turbine shaft sealing
Generator operations and limitations
Generator cooling
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Main Turbine
Generator and Auxiliary Systems:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
Gland seal
Steam bypass and pressure control system
Hydrogen seal oil
Hydrogen gas cooling system
Generator cooling water system
Electrical distribution
Extraction steam
Main steam system
Voltage regulation
Turbine lube oil system
NUREG 2104
IMPORTANCE
3.4-6
2
2
3
2
3
3
3
2
3
3
3
2
2
3
3
2
3
4
3
3
3
3
2
3
2
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4MT Main Turbine Generator and Auxiliary Systems (continued)
K/A NO.
ABILITY
K6.11
K6.12
K6.13
K6.14
Plant information and control system
Turbine control system
Moisture separator reheater
Condenser vacuum
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Main Turbine Generator and Auxiliary Systems controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Generator megawatts
Turbine speed
Turbine valve position
Steam flow
Reactor pressure
Condenser vacuum
First stage turbine pressure
Generator output voltage/reactive load
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Main Turbine Generator and Auxiliary
Systems; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
IMPORTANCE
3
3
2
3
3
3
3
3
4
3
3
3
RO SRO
4
4
3
4
4
4
4
4
4
4
3
3
4
4
2
3
3
3
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
Turbine trip
Loss of lube oil
Loss of condenser vacuum
Reactor scram
Generator trip
Loss of extraction steam
Steam bypass and pressure control system
Turbine rotor bow
Turbine vibration
Turbine pressure control system
A3
Ability to monitor automatic operations of the Main Turbine Generator and
Auxiliary Systems including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Turbine trip
Turbine roll to rated speed
3.4-7
4
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4MT Main Turbine Generator and Auxiliary Systems (continued)
K/A NO.
ABILITY
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
A3.10
A3.11
A3.12
Generator megawatt output
Turbine speed
Control valve operation
Turbine lube oil pressure
Hydrogen seal oil pressure
Hydrogen gas pressure
Generator output voltage
Stator cooling
Generator lockout
Turbine lube oil temperature
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
Turbine lube oil pumps
Generator controls
Stator water cooling pumps
Hydrogen seal oil pumps
Turbine controls
NUREG 2104
IMPORTANCE
3.4-8
3
3
3
3
3
3
3
3
2
2
3
3
3
3
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RCIC Reactor Core Isolation Cooling System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Core Isolation Cooling System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Makeup water condensate system
Main steam system
Residual heat removal system
Instrument air system
Leak detection and isolation system
Engineered Safety Function logic and control system
Feedwater system
Suppression pool suction strainer
AC electrical power distribution system
DC power supply system
High pressure core flooder system
Radioactive drain transfer system
Heating ventilation and air conditioning system
Post accident monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
RCIC steam admission valves
Condensate storage tank suction valve
Suppression pool suction valve
RCIC test return to suppression pool valves
RCIC initiation signals (logic)
Reactor core isolation cooling drain pump
K3
Knowledge of the effect that a loss or malfunction of the Reactor Core
Isolation Cooling System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor water level
Reactor vessel pressure
Decay heat removal
Adequate core cooling
IMPORTANCE
4
4
3
2
3
3
4
2
2
3
2
2
2
2
3
3
3
2
3
2
4
4
4
4
3.4-9
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Core Isolation Cooling System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
Prevent water hammer
Prevent overfilling reactor vessel
Override of high reactor water level interlock
Prevents pump over heating
Prevents turbine damage
Manual initiation
Alternate supplies of water
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Reactor Core Isolation Cooling System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Assist core cooling
Suppression pool
Main condenser
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Core
Isolation Cooling System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
AC electrical power distribution system
Instrument air system
Suppression pool water supply
Makeup water condensate system
Main steam system
Suppression pool suction strainer
DC power supply system
Engineered Safety Function logic and control system
High pressure core flooder system
Radioactive drain transfer system
Heating ventilation and air conditioning system
Feedwater system
NUREG 2104
IMPORTANCE
3
3
3
3
3
4
4
4
4
3
3.4-10
3
2
4
4
4
3
3
3
2
2
2
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Core Isolation Cooling System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Reactor core isolation cooling flow
Reactor core isolation cooling pressure
Reactor water level
Reactor pressure
Reactor core isolation cooling turbine speed
Condensate storage tank level
Suppression pool level
Suppression pool temperature
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Core Isolation Cooling System ;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
System initiation signal
RCIC turbine trips
Injection valve closure
AC power loss
DC power loss
Loss of instrument air system
Loss of RCIC drain pump
Turbine control system failures
Inadequate system flow
Loss of room cooling
Steam line break
Low condensate storage tank level
High suppression pool level
Low suppression pool level
High suppression pool temperature
High drain tank level
A3
Ability to monitor automatic operations of the Reactor Core Isolation
Cooling System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Valve operation
Turbine startup
IMPORTANCE
4
3
4
4
4
3
3
4
RO SRO
4
4
4
4
3
3
2
2
3
3
2
2
2
2
3
3
3
3
3
3
4
4
4
3
3
3
3
3
4
4
2
2
4
4
3.4-11
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RCIC Reactor Core Isolation Cooling System (continued)
K/A NO.
ABILITY
A3.03
A3.04
A3.05
A3.06
System pressure
System flow
Reactor water level
Indications and alarms
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Reactor core isolation cooling turbine speed
Turbine trip throttle valve reset
System valves
NUREG 2104
IMPORTANCE
4
4
4
4
3.4-12
4
4
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RRS Reactor Recirculation System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Recirculation System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
Reactor building cooling water system
AC electrical power distribution system
Control rod drive system
Recirculation flow control system
Plant information and control system
Makeup water purified system
Reactor pressure vessel system
Radioactive drain transfer system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Reactor internal pumps
Recirculation system MG sets
Motor generator set oil pumps
K3
Knowledge of the effect that a loss or malfunction of the Reactor
Recirculation System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Core flow
Load following capabilities
Reactor power
Reactor water level
Reactor recirculation system motor generator sets
Vessel bottom head drain temperature
Primary containment integrity
IMPORTANCE
3.4-13
3
3
3
4
3
2
3
2
3
3
2
4
3
4
4
3
3
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RRS Reactor Recirculation System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Recirculation System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
K4.12
K4.13
K4.14
K4.15
Adequate reactor internal pump NPSH
Reactor internal pump motor cooling
Controlled purge flow
Automatic voltage/frequency regulation
Motor generator set trips
Pump minimum flow limit
Pump start permissives
Minimization of reactor vessel bottom head temperature gradients
End of cycle recirculation pump trip
Anticipated Transient without scram - Recirc pump trip
Selected control rods run in circuitry
Reactor internal pump runback
Reactor internal pump startup
Automatic MG set start sequencing
Core flow rapid reduction logic
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to the Reactor Recirculation System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
Reactor internal pump vibration characteristics
Restart of reactor internal pumps while operating at power
Core flow
Reactor power
Reactor moderator temperature
Reactor pressure
Recirculation flow control system motor-generator sets
Nuclear boiler instrumentation (reactor water level/pressure/core
plate d/p)
Vessel bottom head drain temperature
Residual heat removal shutdown cooling mode
Reactor water level
Anticipated transient without scram circuitry
End-of-cycle recirculation pump trip circuitry
Selected control rods run in circuitry
Recirculation motor inflatable shaft seal subsystem
Recirculation motor cooling subsystem
Reactor internal pump adjustable speed drives
K5.09
K5.10
K5.11
K5.12
K5.13
K5.14
K5.15
K5.16
K5.17
NUREG 2104
IMPORTANCE
3.4-14
3
3
3
3
3
3
3
3
4
4
4
4
3
3
3
2
3
4
4
3
3
4
3
3
3
4
4
4
4
2
3
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RRS Reactor Recirculation System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor
Recirculation System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
Reactor building cooling water systems
AC electrical power distribution system
Control rod drive system
Recirculation system motor-generator sets
Low reactor water level
Recirculation motor inflatable shaft seal subsystem
Makeup water purified system
Reactor internal pump motor cooling subsystem
Turbine trip/load rejection
IMPORTANCE
3
3
3
3
3
3
2
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Recirculation System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
A1.12
A1.13
A1.14
A1.15
A1.16
Reactor internal pump flow
Core flow
Reactor water level
Reactor power
Reactor internal pump motor amps
Reactor internal pump speed
Recirculation cooling water flow
Vessel bottom head drain temperature
Reactor internal pump differential pressure
Reactor internal pump motor temperature
Reactor internal pump MG set temperatures
Reactor internal pump MG drive motor amps
Reactor internal pump MG set generator current, power, voltage
Reactor internal pump motor purge flow
Reactor internal pump vibration
Core differential pressure
3.4-15
4
4
3
4
3
3
3
3
3
2
2
2
2
3
2
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RRS Reactor Recirculation System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Recirculation System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
A2.18
A2.19
Recirculation system leak
Single reactor internal pump trip
Multiple reactor internal pump trip
Inadvertent recirculation flow increase
Inadvertent recirculation flow decrease
Reactor internal pump speed mismatch
Reactor internal pump flow mismatch
Low reactor water level
Loss of reactor feedwater
High reactor pressure (ATWS circuitry initiation)
End of cycle recirculation pump trip circuitry initiation
Selected control rods run in circuitry actuation
Loss of motor cooling
Loss of AC power
Loss of reactor building cooling water
Incomplete start sequence
Loss of reactor internal pump purge flow
Reactor internal pump speed runback
Increase in reactor internal pump vibration
A3
Ability to monitor automatic operations of the Reactor Recirculation
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
Pump/MG set start sequence
System flow
Indications and alarms
Pump speed
Reactor internal pump trips
Reactor internal pump runbacks
Recirculation system motor generator set trip
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
Reactor internal pumps
System flow
Core flow
NUREG 2104
IMPORTANCE
RO SRO
4
4
4
4
4
4
4
4
4
4
3
3
3
3
4
4
4
4
4
4
4
4
4
4
3
3
3
3
3
3
3
3
3
3
3
3
2
2
3
3
3
3
3
3
3
4
4
4
3.4-16
3.4
Safety Function 4: Heat Removal from Reactor Core
System
SF4RHRSDC Residual Heat Removal System: Shutdown Cooling Mode
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Residual Heat Removal System: Shutdown Cooling Mode and
the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
Fuel pool cooling assist
Reactor building cooling water system
AC electrical power distribution system
Reactor water cleanup system
Radioactive drain transfer system
Reactor pressure vessel system
Remote shutdown system
Sampling system
Feedwater system
High pressure nitrogen gas supply system
Instrument air system
Plant information and control system
Leak detection and isolation system
Engineered Safety Function logic and control system
Makeup water condensate system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Residual heat removal pumps
RHR shutdown cooling suction valves
K3
Knowledge of the effect that a loss or malfunction of the Residual Heat
Removal System: Shutdown Cooling Mode will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Reactor pressure
Reactor pressure vessel water level
Reactor pressure vessel temperatures (moderator, vessel, flange)
K4
Knowledge of Residual Heat Removal System: Shutdown Cooling Mode
design feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
High RHR equipment area temperature isolation
High reactor pressure vessel isolation
Low reactor water level isolation
Adequate pump NPSH
IMPORTANCE
3.4-17
3
3
3
2
1
3
3
2
3
2
2
3
3
3
2
3
3
3
3
4
3
4
4
3
NUREG-2104
3.4
Safety Function 4: Heat Removal from Reactor Core
System
SF4RHRSDC Residual Heat Removal System: Shutdown Cooling Mode
(continued)
K/A NO.
KNOWLEDGE
K4.05
K4.06
K4.07
Reactor pressure vessel cooldown rate
RHR pump motor cooling
Pump minimum flow
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Residual Heat Removal System: Shutdown Cooling Mode:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
Reactor pressure vessel pressure
Reactor water level
Fuel pool cooling assist
Low pressure flooder injection
Reactor pressure vessel temperatures (moderator, vessel, flange)
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Residual Heat
Removal System: Shutdown Cooling Mode:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
AC electrical power distribution system
Low reactor water level
High reactor pressure
Reactor building cooling water system
High pressure nitrogen gas supply system
Reactor water cleanup system
Reactor internal pump failure
IMPORTANCE
4
2
3
4
4
3
4
4
3
4
4
3
2
3
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Residual Heat Removal System: Shutdown Cooling Mode
controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
Heat exchanger cooling flow
SDC/RHR pump flow
SDC/RHR pump suction pressure
Reactor water level
Reactor pressure vessel temperatures (moderator, vessel, flange)
RHR pump motor amps
Heat exchanger temperatures
SDC/RHR pump/system discharge pressure
Throttle valve position
NUREG 2104
3.4-18
3
3
3
3
4
2
3
3
3
3.4
Safety Function 4: Heat Removal from Reactor Core
System:
SF4RHRSDC Residual Heat Removal System: Shutdown Cooling Mode
(continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on Residual Heat Removal System: Shutdown
Cooling Mode; and (b) based on those predictions, use procedures to
correct, control, or mitigate the consequences of those abnormal
conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
Low shutdown cooling suction pressure
Electrical power distribution failure
System isolation
SDC/RHR pump trips
Loss of RHR pump motor cooling
Loss of heat exchanger cooling
Reactor low water level
Valve operation
Reactor internal pump trips
Inadequate system flow
A3
Ability to monitor automatic operations of the Residual Heat Removal
System: Shutdown Cooling Mode including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
Valve operation
RHR pump trips
Control room indications and alarms
Shutdown cooling initiation sequence
A4
Ability to manually operate and/or monitor the Residual Heat Removal
System: Shutdown Cooling Mode in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
SDC/RHR pumps
SDC/RHR suction valves
SDC/RHR discharge valves
Heat exchanger cooling water valves
Minimum flow valves
IMPORTANCE
3.4-19
RO SRO
3
3
3
3
4
4
3
4
3
3
3
4
4
4
3
3
3
3
3
3
3
3
3
3
4
4
4
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5PCS Primary Containment System and Auxiliaries
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Primary Containment System and Auxiliaries and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Atmosphere control system
Low and high conductivity waste system
Residual heat removal system
Suppression pool cleanup system
Safety/relief valves
Instrument air system
Post accident sampling system
High pressure core flooder system
Reactor core isolation cooling system
Containment and drywell atmosphere monitoring system
Reactor building heating, ventilation, and air conditioning
Standby gas treatment system
Drywell cooling system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Containment atmosphere monitoring system
Drywell cooling fans
K3
Knowledge of the effect that a loss or malfunction of the Primary
Containment System and Auxiliaries will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Secondary containment
Drywell/wetwell temperature
Drywell/wetwell pressure
Drywell/wetwell hydrogen concentration
Drywell/wetwell oxygen concentration
Differential pressure between secondary and primary containment
Differential pressure between wetwell and drywell
Pneumatically operated valves internal to drywell
Reactor pressure vessel instrumentation
Offsite radioactivity release
IMPORTANCE
3.5-1
3
3
4
2
4
3
3
4
4
3
3
3
4
2
2
4
3
3
3
3
4
3
3
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5PCS Primary Containment System and Auxiliaries (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Primary Containment System and Auxiliaries design
feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
Allows for absorption of the energy released during a LOCA
Contains fission products after a LOCA
Containment isolation
Maintains proper wetwell to drywell differential pressure
Maintains proper secondary containment to drywell differential
pressure
Prevents localized heating of suppression pool (SRV steam
quenchers)
Containment overpressure protection
Drywell vent pipes (vertical and horizontal)
Lower drywell flooder
Quenching of ex-vessel core debris
Containment inerted by nitrogen
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
IMPORTANCE
4
4
4
3
3
3
4
4
4
3
3
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Primary Containment System and Auxiliaries:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Vacuum breaker/relief operation
Hydrogen production mechanisms
Hydrogen combustibility versus hydrogen concentration and
oxygen concentration
Hydrogen concentration measurement
Oxygen concentration measurement
Differential pressure measurement
Containment isolation/integrity
K5.04
K5.05
K5.06
K5.07
3
3
3
3
3
3
4
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Primary
Containment System and Auxiliaries:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
Drywell cooling
Suppression pool makeup
Containment atmospheric control
Drywell vacuum relief system
AC electrical distribution
DC electrical distribution
NUREG 2104
3.5-2
4
4
3
4
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5PCS Primary Containment System and Auxiliaries (continued)
K/A NO.
KNOWLEDGE
K6.07
K6.08
K6.09
K6.10
Atmospheric pressure control system
Residual heat removal system
Standby gas treatment system
Instrument air system
IMPORTANCE
3
4
3
3
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Primary Containment System and Auxiliaries controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
Drywell temperature
Drywell pressure
Wetwell air space temperature
Wetwell pressure
Drywell/wetwell oxygen concentration
Drywell to wetwell differential pressure
Suppression pool level
Suppression pool temperature
Drywell leak detection system
Reactor building to wetwell differential pressure
Inerting flow to both the drywell and wetwell
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Primary Containment System and
Auxiliaries; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
RO SRO
Loss of coolant accident
4
4
Steam bypass of suppression pool
4
4
Safety/relief valve leaking or stuck open
4
4
High drywell/wetwell hydrogen concentration
4
4
High drywell/wetwell oxygen concentration
4
4
High drywell pressure
4
4
Vacuum breaker malfunction
4
4
High drywell temperature
4
4
Abnormal suppression pool level
4
4
High suppression pool temperature
4
4
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
3.5-3
4
4
4
4
3
3
4
4
4
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5PCS Primary Containment System and Auxiliaries (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Primary Containment System
and Auxiliaries including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Vacuum breaker/relief valve operation
System indications and alarms
Drywell or wetwell response during LOCA
Drywell pressure
Drywell/wetwell differential pressure
A4
Ability to manually operate and/or monitor the Primary Containment
System and Auxiliaries in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
Containment overpressure protection system
Drywell/wetwell nitrogen makeup and purge
Drywell pneumatics
Drywell coolers
NUREG 2104
IMPORTANCE
3.5-4
3
3
4
4
3
3
3
3
4
3.5
Safety Function 5: Containment Integrity
System:
SF5LDIS Leak Detection and Isolation System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Leak Detection and Isolation System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
Main steam system
Reactor water cleanup system
Reactor building heating, ventilation, and air conditioning system
Reactor core isolation cooling
Residual heat removal system
Atmosphere control system
Sampling system
Standby gas treatment system
Automated traversing in-core probe system
Radioactive drain transfer system
Safety parameter display system
Reactor building cooling water system
Standby liquid control system
Reactor trip and isolation system
Process radiation monitoring system
Fuel pool cooling and cleanup system
Suppression pool clean-up system
Heating, ventilation, and air conditioning normal cooling
water system
Feedwater system
Engineered Safety Function logic and control system
Plant information and control system
Instrumentation and control power supply system
Reactor pressure vessel instrumentation system
K1.19
K1.20
K1.21
K1.22
K1.23
IMPORTANCE
4
3
3
4
4
3
3
3
3
3
2
3
3
3
3
3
2
2
2
3
3
3
3
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Logic power supplies
K3
Knowledge of the effect that a loss or malfunction of the Leak Detection
and Isolation System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
Reactor water level
Fuel cladding temperature
Off-site radioactivity release
Reactor building radiation level
Drainage sump levels
Turbine building radiation
2
3.5-5
4
4
4
3
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5LDIS Leak Detection and Isolation System (continued)
K/A NO.
KNOWLEDGE
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
K3.14
K3.15
K3.16
K3.17
K3.18
K3.19
K3.20
K3.21
K3.22
K3.23
K3.24
Reactor pressure
Reactor vessel temperature
Main steam system
Reactor water cleanup system
Reactor building heating, ventilation, and air conditioning system
Reactor core isolation cooling
Residual heat removal system
Atmosphere control system
Containment atmosphere sampling
Standby gas treatment system
Automated traversing in-core probe system
Radioactive drain transfer system
Reactor building cooling water systems
Primary containment integrity
Secondary containment integrity
Process radiation monitoring system
Fuel pool cooling and cleanup system
Condensate system
K4
Knowledge of Leak Detection and Isolation System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
Redundancy
Testability
Manual isolation capability
Single failures will not impair the function ability of the system
Once initiated, system reset requires deliberate operator action
Physical separation of system components (to prevent localized
environmental factors, electrical faults, and physical events from
impairing system response)
Manual defeating of selected isolations during specified emergency
conditions
K4.07
IMPORTANCE
4
3
3
3
3
3
3
3
3
3
3
3
2
3
3
2
2
3
3
3
4
3
3
3
3
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Leak Detection and Isolation System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
Primary containment integrity
Secondary containment integrity
NUREG 2104
3.5-6
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5LDIS Leak Detection and Isolation System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the LEAK
DETECTION AND ISOLATION System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
Instrumentation and control power supply system
Process radiation monitoring system
Reactor pressure vessel instrumentation
Containment instrumentation
Various process instrumentation
Reactor trip and isolation system
Engineered Safety Function logic and control system
IMPORTANCE
3
3
3
3
3
4
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the LEAK DETECTION AND ISOLATION System controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
System status indications and alarms
Valve closures
Safety parameter display system
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the LEAK DETECTION AND ISOLATION
System; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
Electrical power failures
System logic failures
Process radiation monitoring system failures
Reactor pressure vessel instrumentation failures
Containment instrumentation failures
Various process instrumentation failures
Surveillance testing
System initiation
Loss of coolant accidents
Standby liquid control system initiation
3.5-7
4
4
3
RO SRO
3
4
3
3
3
3
3
4
3
3
3
3
3
3
4
4
4
4
4
4
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5LDIS Leak Detection and Isolation System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the LEAK DETECTION AND
ISOLATION System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
System status indications and alarms
Valve closures
Safety parameter display system
Standby gas treatment system initiation
Reactor building heating, ventilation, and air conditioning isolation
Condensate pump trip (feedwater line break)
Control room habitability area Heating, Ventilation, and Air
Conditioning initiation
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
Valve closures
Manually initiate isolations
Reset isolations
Safety parameter display system
NUREG 2104
IMPORTANCE
3.5-8
3
4
3
3
3
3
3
4
4
4
3
3.5
Safety Function 5: Containment Integrity
System:
SF5RPV Reactor Vessel Internals
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Vessel Internals and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Main steam system
Reactor recirculation system
Feedwater system
Residual heat removal system
High pressure core flooder system
Reactor core isolation cooling system
Control rod drive system
Fine motion control rod drive mechanism
Standby liquid control system
Reactor water cleanup system
Automatic depressurization system
Loose parts monitoring system
Automated Traversing In-core Probe system
Neutron monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
3
3
3
3
3
3
3
3
3
3
2
3
3
None
K3
Knowledge of the effect that a loss or malfunction of the Reactor Vessel
Internals will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Reactor water level
Reactor pressure
Reactor power
Plant radiation levels
Off-site radioactive release
Leak detection and isolation system
Nuclear boiler instrumentation
K4
Knowledge of Reactor Vessel Internals design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Separation of fluid flow paths within the vessel
Core orificing
Moisture removal from generated steam
Natural circulation
3.5-9
3
3
3
3
3
3
3
3
3
3
4
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5RPV Reactor Vessel Internals (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Reactor Vessel Internals:
(CFR: 41.5 / 45.3)
IMPORTANCE
None
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Vessel
Internals:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
K6.14
K6.15
K6.16
Control rod drive system
Fine motion control rod drive mechanism
Reactor recirculation system
Feedwater system
Standby Liquid Control system
Safety/relief valves
Reactor water cleanup system
Nuclear boiler instrumentation
High pressure core flooder system
Residual heat removal system
Reactor core isolation cooling system
Automatic depressurization system
Loose parts monitoring
Automated traversing in-core probe system
Neutron monitoring system
Main steam system
3
3
3
3
3
3
3
3
3
3
3
3
2
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Vessel Internals controls including:
(CFR:41.5/45.5)
None
NUREG 2104
3.5-10
3.5
Safety Function 5: Containment Integrity
System:
SF5RPV Reactor Vessel Internals (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Vessel Internals; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
Loss of coolant accident
Overpressurization transient
Control rod drop accident
Excessive heatup/cooldown rate
Exceeding thermal limits
Exceeding safety limits
A3
Ability to monitor automatic operations of the Reactor Vessel Internals
including:
(CFR: 41.7 / 45.7)
IMPORTANCE
RO SRO
4
4
4
4
4
4
4
4
4
4
4
4
None
A4
Ability to manually operate and/or monitor the Reactor Vessel Internals in
the control room:
(CFR: 41.7 / 45.5 to 45.8)
None
3.5-11
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPC Residual Heat Removal System: Suppression Pool Cooling
Mode
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Residual Heat Removal Systems: Suppression Pool Cooling
Mode and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
AC electrical power distribution system
Reactor building cooling water system
Suppression pool temperature monitoring system
Remote shutdown system
Engineered Safety Function logic and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
RHR heat exchanger inlet and outlet valves
RHR heat exchanger bypass valve
Residual heat removal system pumps
Valve control logic
K3
Knowledge of the effect that a loss or malfunction of the Residual Heat
Removal System: Suppression Pool Cooling Mode will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
Suppression pool temperature control
Residual heat removal system drain to radwaste
K4
Knowledge of Residual Heat Removal System: Suppression Pool Cooling
Mode design feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Redundancy
Unintentional reduction in vessel injection flow during accident
conditions
Pump minimum flow protection
Pump motor cooling
Prevention of water hammer
Adequate pump net positive suction head
Heat exchanger cooling
Prevention of leakage to the environment through system heat
exchanger
Automatic initiation of suppression pool cooling
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
NUREG 2104
IMPORTANCE
3.5-12
4
3
4
3
3
3
2
3
2
4
3
4
4
3
3
3
3
3
3
4
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPC Residual Heat Removal System: Suppression Pool Cooling
Mode (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Residual Heat Removal System: Suppression Pool Cooling
Mode:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
Suppression pool
Keep fill
Residual heat removal system piping
Residual heat removal system pumps
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Residual Heat
Removal System: Suppression Pool Cooling Mode:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.06
K6.07
AC electrical power distribution system
Keep fill
Suppression pool
ECCS room cooling
Reactor pressure vessel instrumentation
Reactor building cooling water system
IMPORTANCE
4
3
4
4
3
3
4
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Residual Heat Removal System: Suppression Pool Cooling
Mode controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
Suppression pool temperature
System flow
System pressure
Suppression pool level
Motor amps
Emergency generator loading
Wetwell air temperature
3.5-13
4
4
3
3
2
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPC Residual Heat Removal System: Suppression Pool Cooling
Mode (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Residual Heat Removal System:
Suppression Pool Cooling Mode; and (b) based on those predictions, use
procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
Inadequate net positive suction head
Pumps trips
Valve closures
Valve openings
AC electrical power distribution system failure
Pump seal failure
Inadequate room cooling
Reactor pressure vessel instrument failures
Motor operated valve failures
Valve logic failure
High suppression pool temperature
Loss of coolant accident
Loss of, or inadequate, heat exchanger cooling flow
High suppression pool level
A3
Ability to monitor automatic operations of the Residual Heat Removal
System: Suppression Pool Cooling Mode including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
Valve operation
Pump start
Pump discharge pressure
System flow
Suppression pool temperature
Reactor building cooling water/ Reactor building service water start
signal
Control room alarms
System initiation sequence
A3.07
A3.08
NUREG 2104
IMPORTANCE
3.5-14
RO SRO
3
3
3
3
3
3
3
3
3
4
3
3
3
3
3
3
3
3
3
3
4
4
4
4
3
3
3
3
3
3
2
3
3
3
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPC Residual Heat Removal System: Suppression Pool Cooling
Mode (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Residual Heat Removal
System: Suppression Pool Cooling Mode in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
Residual heat removal system pumps
Valves pertaining to suppression pool cooling
Keep fill
Minimum flow valves
The overrides for suppression pool cooling valve logic
IMPORTANCE
3.5-15
4
4
3
3
4
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPR Residual Heat Removal System: Drywell and Wetwell Spray
Mode
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
Residual Heat Removal System: Drywell and Wetwell Spray Mode and the
following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
AC electrical power distribution system
Reactor pressure vessel instrumentation
Reactor building cooling water systems
Fire protection system (PRA)
Engineered Safety Function logic and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Drywell and wetwell spray motor operated valves
Residual heat removal pumps
Valve control logic
K3
Knowledge of the effect that a loss or malfunction of the Residual Heat
Removal System: Drywell and Wetwell Spray Mode will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Drywell/ wetwell pressure
Drywell/ wetwell temperature
Drywell/ wetwell components, due to continued operation with
elevated pressure and/or temperature and/or water level
K4
Knowledge of Residual Heat Removal System: Drywell and Wetwell Spray
Mode design feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
Redundancy
Reduction in vessel injection flow during accident conditions
Prevention of piping overpressurization
Pump minimum flow protection
Residual heat removal pump motor cooling
Prevention of water hammer
Adequate pump net positive suction head
Spray flow cooling
Prevention of leakage to the environment through system heat
exchanger
NUREG 2104
IMPORTANCE
3.5-16
3
3
3
3
3
2
3
2
4
4
3
3
3
2
3
2
3
2
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPR Residual Heat Removal System: Drywell and Wetwell Spray
Mode (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply Residual Heat Removal System: Drywell and
Wetwell Spray Mode:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
K5.09
K5.10
Evaporative cooling
Convective cooling
Vacuum breaker operation
Suppression pool
Residual heat removal system piping
Residual heat removal system pumps
Keep fill
Drywell spray penetration
Containment instrumentation
Wetwell spray penetration
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Residual Heat
Removal System: Drywell and Wetwell Spray Mode:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
AC electrical power distribution system
Keep fill
Suppression pool (temperature level and pressure)
ECCS room cooling
Reactor pressure vessel instrumentation
Wetwell to drywell vacuum breakers
Reactor building cooling water system
Containment integrity
Suction flow path
IMPORTANCE
2
2
3
3
4
4
3
3
3
3
3
3
3
2
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Residual Heat Removal System: Drywell and Wetwell Spray
Mode controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
Drywell pressure
Drywell temperature
Wetwell pressure
Wetwell temperature
Suppression pool temperature
3.5-17
4
3
4
3
3
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPR Residual Heat Removal System: Drywell and Wetwell Spray
Mode (continued)
K/A NO.
ABILITY
A1.06
A1.07
A1.08
A1.09
A1.10
System flow
System pressure
Suppression pool level
RHR pump motor amps
Emergency generator loading
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Residual Heat Removal System: Drywell
And Wetwell Spray Mode; and (b) based on those predictions, use
procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
Inadequate net positive suction head
RHR pumps trips
AC electrical power distribution system
Pump seal failure
Inadequate room cooling
Reactor pressure vessel instrumentation failures
Motor operated valve failures
Pump runout
Valve logic failure
High suppression pool level
High drywell/wetwell pressure
Loss of, or inadequate heat exchanger cooling flow
High drywell temperature
Low (or negative) drywell pressure during system operation
Low (or negative) wetwell pressure during system operation
Loss of coolant accident
Loss of drywell cooling system
A3
Ability to monitor automatic operations of the Residual Heat Removal
System: Drywell and Wetwell Spray Mode including:
(CFR: 41.7 / 45.7)
A3.01
Drywell and wetwell spray initiation sequence
NUREG 2104
IMPORTANCE
3.5-18
3
3
3
2
3
RO SRO
2
3
3
3
3
3
2
3
3
3
3
3
3
3
3
3
3
3
3
3
4
4
3
3
3
3
4
4
4
4
4
4
3
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5RHRSPR Residual Heat Removal System: Drywell and Wetwell Spray
Mode (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Residual Heat Removal
System: Drywell and Wetwell Spray Mode in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
A4.06
Residual heat removal system pumps
Residual heat removal system suction valves
Residual heat removal system spray valves (PRA)
Keep fill
Minimum flow valves
Valve logic override
IMPORTANCE
3.5-19
4
3
4
3
3
4
NUREG-2104
3.5
Safety Function 5: Containment Integrity
System:
SF5SEC Secondary Containment
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Secondary Containment and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
Reactor building heating, ventilation, and air conditioning system
Primary containment system
Standby gas treatment system
Instrument air system
Reactor building
Radioactive drain transfer system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
4
4
3
3
2
None
K3
Knowledge of the effect that a loss or malfunction of the Secondary
Containment will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Off-site radioactivity release
Core alterations
Movement of irradiated fuel in the secondary containment
Operations with a potential for draining the reactor vessel (OPDRVs)
K4
Knowledge of Secondary Containment design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Personnel access without breaching secondary containment
Protection against over pressurization
Fluid leakage collection
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Secondary Containment:
(CFR: 41.5 / 45.3)
None
NUREG 2104
3.5-20
4
3
3
3
4
3
3
3.5
Safety Function 5: Containment Integrity
System:
SF5SEC Secondary Containment (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Secondary
Containment:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
Reactor building heating, ventilation, and air conditioning system
Standby gas treatment system
Primary containment system
Instrument air system
Reactor building
IMPORTANCE
4
4
4
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Secondary Containment controls including:
(CFR: 41.5 / 45.5)
A1.01
High area temperature
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Secondary Containment; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
Access doors or equipment hatch failure
Excessive outleakage
High area radiation
High airborne radiation
High area temperature
High area water levels
Low secondary containment differential pressure
A3
Ability to monitor automatic operations of the Secondary Containment
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Secondary containment isolation
Secondary containment differential pressure
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
4
RO SRO
3
4
4
4
3
4
3
4
3
3
3
3
3
3
4
4
None
3.5-21
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6EPDS AC Electrical Distribution System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the AC Electrical Distribution System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Emergency diesel generators
Direct current power supply system
Offsite power system
Vital AC power supply system
Main turbine/generator
Combustion turbine generator
Engineered Safety Function logic and control system
Remote shutdown system
Plant information and control system
Instrument and control power supply system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
4
3
3
3
3
4
3
3
3
3
None
K3
Knowledge of the effect that a loss or malfunction of the AC Electrical
Distribution System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Major system loads
Emergency diesel generators
Direct current power supply system
Vital AC power supply system
Offsite power system
Combustion turbine generator
Instrumentation and control power supply system
K4
Knowledge of AC Electrical Distribution System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
Bus lockouts
Circuit breaker automatic trips
Interlocks between automatic bus transfer and breakers
Protective relaying
Paralleling of AC sources (syncroscope)
Redundant power sources to vital buses
3.6-1
4
4
3
3
3
4
3
3
3
3
3
4
4
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6EPDS AC Electrical Distribution System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to the AC Electrical Distribution System:
(CFR: 41.5 / 45.3)
K5.01
Principle involved with paralleling two AC sources
K6
Knowledge of the effect of the following plant conditions, system
malfunctions, or components malfunctions will have on the AC Electrical
Distribution System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
DC power
Loss of off-site power
Main generator trip
IMPORTANCE
3
3
4
4
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the AC Electrical Distribution System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
Effect on instrumentation and controls of switching power supplies
Effects of loads when energizing a bus
Bus voltage
Load currents
Breaker lineups
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the AC Electrical Distribution System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
NUREG-2104
Turbine/generator trip
Loss of coolant accident
Loss of off-site power
Types of loads that, if deenergized, would degrade or hinder plant
operation
Bus grounds
Deenergizing a live bus
Energizing a dead bus
Exceeding voltage limitations
3.6-2
3
3
3
3
3
RO SRO
3
4
4
4
4
4
4
3
3
3
3
4
3
3
3
3
3.6
Safety Function 6: Electrical
System:
SF6EPDS AC Electrical Distribution System (continued)
K/A NO.
ABILITY
A2.09
A2.10
A2.11
A2.12
Exceeding current limitations
Degraded system voltages
Main generator load rejection
Station blackout
A3
Ability to monitor automatic operations of the AC Electrical Distribution
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Breaker tripping ................................................................................ 3
Automatic bus transfer ...................................................................... 3
Load shedding .................................................................................. 3
Load sequencing............................................................................... 3
Synchronizing and paralleling of different AC supplies ...................... 3
A4
Ability to manually operate and/or monitor the AC Electrical Distribution
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Available breakers and disconnects (including available switchyard) 3
Synchroscope, including understanding of running and incoming
voltages ............................................................................................ 3
Local operation of breakers............................................................... 3
Synchronizing and paralleling of different AC supplies ...................... 4
Voltage, current, power, and frequency on AC buses ....................... 3
Perform ground isolation ................................................................... 2
A4.03
A4.04
A4.05
A4.06
IMPORTANCE
RO SRO
3
3
3
4
3
3
4. 4
3.6-3
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6DC Direct Current Power Supply System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Direct Current Power Supply System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
AC electrical power distribution system
Control building heating, ventilation, and air conditioning system
Control rod drive system
Engineered Safety Function logic and control system
Plant information and control system
Reactor trip and isolation system
Reactor core isolation cooling system
Fuel pool cooling and cleanup system
Turbine lube oil system
Generator seal oil system
Lighting and servicing power supply system
Emergency diesel generator system
Vital AC power supply system
Suppression pool cleanup system
Turbine control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
Alternate rod insertion valves
Backup scram relay contacts to CRD air header dump valves
Reactor core isolation cooling system DC valves
Emergency lighting system
Engineered Safety Function logic and control system
Plant information and control system
K3
Knowledge of the effect that a loss or malfunction of the Direct Current
Power Supply System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
Emergency diesel generators
Engineered Safety Function logic and control system
Reactor core isolation cooling system
Reactor trip and isolation system
Control rod drive system
Suppression pool cleanup system
Fuel pool cooling and cleanup system
Turbine lube oil system
Generator seal oil system
NUREG-2104
IMPORTANCE
3.6-4
3
3
3
4
4
4
4
3
2
2
2
3
3
2
3
3
3
3
2
3
3
3
3
3
3
3
2
2
2
2
3.6
Safety Function 6: Electrical
System:
SF6DC Direct Current Power Supply System (continued)
K/A NO.
KNOWLEDGE
K3.10
K3.12
K3.13
K3.14
K3.15
Lighting and servicing power supply system
Vital AC power supply system
Plant information and control system
AC electrical power distribution system
Automatic depressurization system
K4
Knowledge of Direct Current Power Supply System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Manual/ automatic transfers of control
Breaker interlocks, permissives, bypasses and cross ties
K5
IMPORTANCE
2
3
3
3
3
3
3
Knowledge of the operational implications or cause-effect relationships as
they apply Direct Current Power Supply System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Hydrogen generation during battery charging
Battery charger and battery
Ground detection
3
3
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Direct Current
Power Supply System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
Electrical power distribution system
Battery ventilation
3
3
ABILITY
A1
A1.01
Ability to predict and/or monitor changes in parameters associated with
operating the Direct Current Power Supply System controls including:
(CFR: 41.5 / 45.5)
Battery discharging rate
3
3.6-5
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6DC Direct Current Power Supply System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Direct Current Power Supply System;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
Grounds
Loss of ventilation during charging
Station blackout
A3
Ability to monitor automatic operations of the Direct Current Power Supply
System including:
(CFR: 41.7 / 45.7)
A3.01
Control room indications and alarms
A4
Ability to manually operate and/or monitor the Direct Current Power Supply
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Major breakers and control power fuses
NUREG-2104
IMPORTANCE
3.6-6
RO SRO
3
3
3
3
3
3
3
3
3.6
Safety Function 6: Electrical
System:
SF6EDGCTG Emergency Generators (Diesel/Combustion Turbine
Generator)
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between Emergency Generators (Diesel/Combustion Turbine Generators)
and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Electrical power distribution system
Direct current power supply system
Fire protection system
Reactor building cooling water system
Remote shutdown system
Engineered Safety Function logic and control system
Makeup water purified system
Vital AC power supply system
Reactor building heating, ventilation, and air conditioning system
Plant information and control system
Residual heat removal system
High pressure core flooder system
Grounding and lightning protection system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
Air compressor
Fuel oil pumps
Lube oil pumps
Emergency diesel generator field flash
K3
Knowledge of the effect that a loss or malfunction of the Emergency
Generators (Diesel/Combustion Turbine Generators) will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Residual heat removal system
High pressure core flooder system
Electrical power distribution system
K4
Knowledge of Emergency Generators (Diesel/Combustion Turbine
Generators) design feature(s) and/or interlocks which provide for the
following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Generator trips (normal)
Generator trips (emergency/Loss of coolant accident)
Speed droop control
IMPORTANCE
3.6-7
4
3
3
3
3
4
2
3
3
3
4
4
2
2
2
2
2
4
4
4
4
4
3
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6EDGCTG Emergency Generators (Diesel/Combustion Turbine
Generator) (continued)
K/A NO.
KNOWLEDGE
K4.04
K4.05
K4.06
K4.07
K4.08
Field flashing
Load shedding and sequencing
Governor control
Local operation and control
Automatic startup
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Emergency Generators (Diesel/Combustion
Turbine Generators):
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
Definition of frequency and synchronous frequency
Reactive power control
Real power control
Fuel oil supply subsystem
Starting air subsystem
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Emergency
Generators (Diesel/Combustion Turbine Generators):
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
Starting air
Fuel oil pumps
Lube oil pumps
Ignition system (combustion turbine generator)
Reactor building cooling water system
Electrical power distribution system
DC electrical power supply system
Engineered Safety Function logic and control system
Makeup water purified system
Loss of offsite power
Loss of coolant accident
Jacket cooling water
NUREG-2104
IMPORTANCE
3.6-8
3
3
3
3
4
2
2
2
3
3
4
4
4
3
4
4
3
4
3
3
3
3
3.6
Safety Function 6: Electrical
System:
SF6EDGCTG Emergency Generators (Diesel/Combustion Turbine
Generator) (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Emergency Generators (Diesel/Combustion Turbine
Generators controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
Lube oil temperature
Fuel consumption rate
Operating voltages, currents, and temperatures
Crank case temperature and pressure
Cylinder temperature differential
Emergency generator room temperature
Maintaining minimum load on generator (to prevent reverse power)
Diesel generator load (MWe and frequency)
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Emergency Generators
(Diesel/Combustion Turbine Generators); and (b) based on those
predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
Parallel operation of generator
Unloading prior to securing generator
Operating unloaded, lightly loaded, and highly loaded.
Consequences of operating under/over excited
Synchronization of the generator with other electrical supplies
Opening normal and/or alternate power to Class 1E AC bus
Loss of off-site power during full-load testing
Initiation of generator room fire protection system
Loss of electrical power distribution system
LOCA
Station blackout
A3
Ability to monitor automatic operations of the Emergency Generators
(Diesel/Combustion Turbine Generators) including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
Automatic starting of emergency generator
Minimum time for load pick up
Control room indications and alarms
IMPORTANCE
3.6-9
3
2
3
3
2
2
3
3
RO SRO
4
4
3
3
3
3
3
3
4
4
3
3
4
4
3
4
4
4
4
4
4
4
3
3
3
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6EDGCTG Emergency Generators (Diesel/Combustion Turbine
Generator) (continued)
K/A NO.
ABILITY
A3.04
Operation of the governor control system on frequency and voltage
control
Load shedding and sequencing
Reactor building cooling water system operation
A3.05
A3.06
IMPORTANCE
3
3
3
A4
Ability to manually operate and/or monitor Emergency Generators
(Diesel/Combustion Turbine Generator) in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
Adjustment of exciter voltage
Synchroscope
Transfer of control between manual and automatic
Manual start, loading, and stopping of generator
Transfer of generator (with load) to grid
NUREG-2104
3.6-10
3
3
3
4
4
3.6
Safety Function 6: Electrical
System:
SF6VAC Vital AC Power Supply System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Vital AC Power Supply System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
Rod control and information system
Reactor trip and isolation system
Process radiation monitoring system
Neutron monitoring system
Standby liquid control system
Plant data network
Containment atmospheric monitoring system
Electrical power distribution system
DC power supply system
Control building heating, ventilation, and air conditioning system
Recirculation flow control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
3
3
3
3
2
2
3
3
2
3
None
K3
Knowledge of the effect that a loss or malfunction of the Vital AC Power
Supply System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
Reactor trip and isolation system
Reactor recirculation flow control system
Neutron monitoring system
Standby liquid control system
Plant data network
Containment atmospheric monitoring system
Rod control and information system
Control building heating, ventilation, and air conditioning system
K4
Knowledge of Vital AC Power Supply System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
Transfer from preferred power to alternate power supplies
3.6-11
3
3
3
3
2
2
3
2
3
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6VAC Vital AC Power Supply System (continued)
K/A NO.
ABILITY
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Vital AC Power Supply System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
General principles of static inverter operation
General principles of static switch operation
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Vital AC Power
Supply System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
Electrical power distribution system
DC power supply system
Static inverter
IMPORTANCE
2
2
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Vital AC Power Supply System controls including:
(CFR: 41.5 / 45.5)
None
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Vital AC Power Supply System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
Under voltage
Over voltage
Frequency changes in the system
A3
Ability to monitor automatic operations of the Vital AC Power Supply
System including:
(CFR: 41.7 / 45.7)
A3.01
Transfer from preferred to alternate source
NUREG-2104
3.6-12
RO SRO
3
3
3
3
2
3
3
3.6
Safety Function 6: Electrical
System:
SF6VAC Vital AC Power Supply System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Vital AC Power Supply
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Transfer from alternative source to preferred source
IMPORTANCE
3.6-13
3
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6I&C Instrumentation and Control Power Supply System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Instrumentation and Control Power Supply System and the
following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
Engineered Safety Function logic and control system
Plant information and control system
Makeup water condensate system
Residual heat removal system
High pressure core flooder system
Reactor building cooling water system
Reactor service water system
Service air system
Instrument air system
Drywell cooling system
Electrical power distribution system
AC electrical power distribution system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
4
4
2
2
2
2
2
2
2
3
3
4
None
K3
Knowledge of the effect that a loss or malfunction of the Instrumentation
and Control Power Supply System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Engineered Safety Function logic and control system
Plant information and control system
Makeup water condensate system
Residual heat removal system
High pressure core flooder system
Reactor water cleanup system
Reactor building cooling water system
Reactor service water system
Service air system
Instrument air system
K4
Knowledge of Instrumentation and Control Power Supply System design
feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
None
NUREG-2104
3.6-14
4
4
2
2
2
2
2
2
2
3
3.6
Safety Function 6: Electrical
System:
SF6I&C Instrumentation and Control Power Supply System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Instrumentation and Control Power Supply
System:
(CFR: 41.5 / 45.3)
IMPORTANCE
None
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Instrumentation
and Control Power Supply System:
(CFR: 41.7 / 45.7)
K6.01
AC electrical power distribution system
4
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Instrumentation and Control Power Supply System controls
including:
(CFR: 41.5 / 45.5)
None
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Instrumentation and Control Power
Supply System; and (b) based on those predictions, use procedures to
correct, control, or mitigate the consequences of those abnormal
conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
Under voltage
Over voltage
Distribution board feed breaker trip
A3
Ability to monitor automatic operations of the Instrumentation and Control
Power Supply System including:
(CFR: 41.7 / 45.7)
RO
3
3
3
SRO
3
3
3
None
3.6-15
NUREG-2104
3.6
Safety Function 6: Electrical
System:
SF6I&C Instrumentation and Control Power Supply System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Instrumentation and Control
Power Supply System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
IMPORTANCE
None
NUREG-2104
3.6-16
3.7
Safety Function 7: Instrumentation
System:
SF7APR Automatic Power Regulator System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships and
between the Automatic Power Regulator System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
Rod control and information system
Neutron monitoring system
Recirculation flow control system
Steam bypass and pressure control system
Plant information and control system
Turbine control system
Reactor water cleanup system
Main generator
Automated thermal limit monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Automatic power regulator digital controllers
K3
Knowledge of the effect that a loss or malfunction of the Automatic Power
Regulator System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
Rod control and information system
Recirculation flow control system
Steam bypass and pressure control system
Turbine control system
Plant information and control system
K4
Knowledge of Automatic Power Regulator System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
Determination of reactor criticality when Automatic Power Regulator
is in automatic mode
Trip of Automatic Power Regulator to manual when system or
component conditions are abnormal during execution of prescribed
sequences
K4.02
IMPORTANCE
3.7-1
4
3
4
4
3
4
3
3
3
2
4
4
4
4
4
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7APR Automatic Power Regulator System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Automatic Power Regulator System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
Main generator output
Reactor power
Rod pattern control
Reactivity control
Reactor water temperature during startup and shutdown
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Automatic Power
Regulator System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
Plant information and control system
Neutron monitoring system
Reactor water cleanup system
Steam bypass and pressure control system
Turbine control system
Main generator system
Rod control and information system
Recirculation flow control system
Automated thermal limit monitoring system
IMPORTANCE
3
4
3
3
3
3
3
3
4
3
4
4
4
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Automatic Power Regulator System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
Reactor power
Reactor pressure
Main generator power output
Reactor water temperature
Power/flow operating map
NUREG-2104
3.7-2
4
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7APR Automatic Power Regulator System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Automatic Power Regulator System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
Fault detection while in automatic mode of operation
A3
Ability to monitor automatic operations of the Automatic Power Regulator
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
Dedicated operator interface indications
Verification of proper functioning/operability
Annunciator and alarm signals
Reactor power
Core flow
Control rod position
Reactor pressure
Reactor water temperature
Main generator output
A4
Ability to manually operate and/or monitor the Automatic Power Regulator
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Console controls to set up various automatic power regulator
modes of operation
IMPORTANCE
3.7-3
RO SRO
3
3
3
3
3
4
4
4
3
3
3
4
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7ATLM Automated Thermal Limit Monitor
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Automated Thermal Limit Monitor and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
Rod control and information system
Automatic power regulator system
Feedwater control system
Recirculation flow control system
Neutron monitoring system
Plant information and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Automated thermal limit monitor system controllers
K3
Knowledge of the effect that a loss or malfunction of the Automated
Thermal Limit Monitor will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Rod control and information system
Recirculation flow control system
Automatic power regulator system
K4
Knowledge of Automated Thermal Limit Monitor design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Rod withdrawal blocks/errors
Bypassing an automated thermal limit monitor channel
System testing
Inhibiting an increase in recirculation flow
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Automated Thermal Limit Monitor:
(CFR: 41.5 / 45.3)
IMPORTANCE
None
NUREG-2104
3.7-4
4
4
2
4
3
3
2
4
4
4
4
3
2
4
3.7
Safety Function 7: Instrumentation
System:
SF7ATLM Automated Thermal Limit Monitor (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunction or component malfunctions will have on the Automated
Thermal Limit Monitor:
(CFR: 41.7 / 45.7)
K6.01
K6.02
Recirculation flow control system
Feedwater control system input (cleanup water flow, feedwater
temperature and total feedwater flow to RPV)
Plant information and control system
Neutron monitoring system
K6.03
K6.05
IMPORTANCE
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Automated Thermal Limit Monitor controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.03
Rod position
Status of control rod withdrawal blocks
Status of recirculation flow increase blocks
Automated thermal limit monitor channel bypass status
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Automated Thermal Limit Monitor; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
Power supply loss
Loss of neutron monitoring input
Automatic self-bypass of the ATLM protective features
A3
Ability to monitor automatic operations of the Automated Thermal Limit
Monitor including:
(CFR: 41.7 / 45.7)
A3.01
System indications and alarms on control room panels and
backpanels
Verification of proper functioning/operability
Control rod withdrawal blocks
Inhibit of recirculation flow increase
Automatic bypass below low power setpoint
A3.02
A3.03
A3.04
A3.05
3.7-5
3
3
3
3
RO SRO
2
2
3
3
2
2
3
3
3
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7ATLM Automated Thermal Limit Monitor (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Automated Thermal Limit
Monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Controls to bypass/unbypass an Automated Thermal Limit
Monitor channel
NUREG-2104
IMPORTANCE
3.7-6
3
3.7
Safety Function 7: Instrumentation
System:
SF7APRM Average Power Range Monitor/Local Power Range Monitor
System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Average Power Range Monitor/Local Power Range Monitor
System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
Reactor trip and isolation system
Multi-channel rod block monitor system
Safety parameter display system
Plant information and control system
Rod control and information system
Automated traversing incore probe system
Reactor pressure vessel
Automatic power regulator system
Oscillation power range monitor sub-system
Recirculation flow control system
Engineered Safety Function logic and control system
Vital AC power supply system
Automated thermal limit monitor system
Primary containment system
Rod worth minimizer system
Feedwater control system
Automatic depressurization system
K2
Knowledge bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Local Power Range Monitor detector channels
Average Power Range Monitor channels
Oscillation power range monitor channels
K3
Knowledge of the effect that a loss or malfunction of the Average Power
Range Monitor/Local Power Range Monitor System will have on the
following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Reactor trip and isolation system
Recirculation flow control system
Rod control and information system
Reactor power indication
Multichannel rod block monitor
Core thermal limit calculations
Automatic power regulator system
IMPORTANCE
3.7-7
4
3
3
3
3
3
3
3
4
3
3
3
3
2
3
3
4
2
3
2
4
3
3
4
3
4
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7APRM Average Power Range Monitor/Local Power Range Monitor
System (continued)
K/A NO.
KNOWLEDGE
K3.08
K3.09
K3.10
K3.11
Oscillation power range monitor system
Engineered Safety Function logic and control system
Automated thermal limit monitor system
Rod worth minimizer system
K4
Knowledge of Average Power Range Monitor/Local Power Range Monitor
System design feature(s) and/or interlocks which provide for the
following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
Rod withdrawal blocks
Reactor scram signals
Individual Local Power Range Monitor detector replacement
Alarm seal-in
Effects of detector aging on Average Power Range Monitor/Local
Power Range Monitor readings
Average Power Range Monitor flow biased trip setpoints
Multichannel rod block monitor setpoints
Sampling of overall core power in each Average Power Range
Monitor (accomplished through Local Power Range Monitor
assignments and symmetrical rod patterns)
Detection of power oscillations
Inhibit of automatic initiation of Automatic Depressurization System
K4.06
K4.07
K4.08
K4.09
K4.10
IMPORTANCE
4
3
3
3
4
4
2
2
3
4
4
3
4
4
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Average Power Range Monitor/Local Power Range Monitor
System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
Local Power Range Monitor detector operation
Effects of voids on Local Power Range Monitor indication
Control rod symmetrical patterns
Local Power Range Monitor detector location and core symmetry
Core flow effects on Average Power Range Monitor trip setpoints
Assignment of Local Power Range Monitor's to specific Average
Power Range Monitor channels
NUREG-2104
3.7-8
3
3
3
3
4
2
3.7
Safety Function 7: Instrumentation
System:
SF7APRM Average Power Range Monitor/Local Power Range Monitor
System (continued)
K/A NO.
ABILITY
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Average Power
Range Monitor/Local Power Range Monitor System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
Reactor trip and isolation system
Automated traversing incore probe system
Detectors
Rod control and information system
Vital AC power supply system
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Average Power Range Monitor/Local Power Range Monitor
System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
Reactor power indication
Reactor trip and isolation system status
Control rod block status
Scram and rod block trip setpoints
Indications and alarms
Average Power Range Monitor (gain adjustment factor)
Injection of cold water into the reactor pressure vessel
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Average Power Range Monitor/Local
Power Range Monitor System; and (b) based on those predictions, use
procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
Power supply degraded
Upscale or downscale trips
Inoperative trip (all causes)
Scram trip signals
Rod block signals
Loss of reactor recirculation flow signal
Core flow channels upscale/downscale
Faulty or erratic operation of detectors/systems
IMPORTANCE
3.7-9
4
2
3
3
3
4
4
4
4
3
4
3
RO SRO
3
3
4
4
4
4
4
4
4
4
4
4
3
4
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7APRM Average Power Range Monitor/Local Power Range Monitor
System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Average Power Range
Monitor/Local Power Range Monitor System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Panel indications
Annunciator and alarm signals
Reactor trip and Isolation System status
Control rod block status
Inhibit of automatic initiation of Automatic Depressurization System
A4
Ability to manually operate and/or monitor the Average Power Range
Monitor/Local Power Range Monitor System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.05
A4.06
Average Power Range Monitor back panel switches
Local Power Range Monitor back panel switches
Oscillation Power Range Monitor back panel switches
Trip bypasses
Bypass an Average Power Range Monitor/Local Power Range
Monitor channel
NUREG-2104
IMPORTANCE
3.7-10
3
3
4
4
4
3
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7NBI Nuclear Boiler Instrumentation
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Nuclear Boiler Instrumentation and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Reactor trip and isolation system
Neutron monitoring system
Engineered Safety Function logic and control system
Safety/relief valves
Recirculation flow control system
Feedwater control system
Feedwater system
Plant information and control system
Reactor pressure vessel
Steam bypass and pressure control system
Remote shutdown system
Alternate feedwater injection system
Control rod drive system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
4
3
4
4
3
4
3
3
4
4
3
3
3
None
K3
Knowledge of the effect that a loss or malfunction of the Nuclear Boiler
Instrumentation will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
K3.14
Reactor trip and isolation system
Leak detection and isolation system
Reactor core isolation cooling system
High pressure core flooder system
Residual heat removal
Automatic depressurization system
Relief/safety valves
Alternate rod insertion
Recirculation flow control system
Feedwater control system
Feedwater system
Main turbine
Emergency diesel generators
Anticipated transient without scram logic
3.7-11
4
4
4
4
4
4
4
4
3
4
3
3
4
4
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7NBI Nuclear Boiler Instrumentation (continued)
K/A NO.
KNOWLEDGE
K3.15
K3.16
K3.17
K3.18
K3.19
K3.20
K3.21
K3.22
K3.23
K3.24
K3.25
Plant information and control system
Reactor pressure vessel
Vessel temperature monitoring
Vessel level monitoring
Vessel pressure monitoring
Core flow monitoring
Core differential pressure monitoring
Loose parts detection in the primary system
Reactor internal pump flow monitoring
Recirculation system
Steam bypass and pressure control system
K4
Knowledge of Nuclear Boiler Instrumentation design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
K4.10
K4.11
K4.13
Reading of nuclear boiler parameters outside the control room
Physical separation of sensors
Redundancy of sensors
Inputs to the reactor trip and isolation system
Initiation of the emergency core cooling systems
Initiation of the leak detection and isolation system
Protection for the main turbine from high moisture carryover
Protection against filling the main steam lines from the feed system
Inputs to alternate reactor shutdown system
Reactor vessel overpressure protection
Overpressure protection for various low-pressure systems
Temperature compensation for water level indication
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Nuclear Boiler Instrumentation:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Vessel vibration measurement (loose parts monitor)
Steam flow effect on reactor water level
Anticipated transient without scram logic
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Nuclear Boiler
Instrumentation:
(CFR: 41.7 / 45.7)
K6.01
K6.02
AC electrical distribution system
DC electrical distribution system
NUREG-2104
IMPORTANCE
3.7-12
3
3
3
4
4
4
4
3
2
3
3
4
3
3
4
4
4
3
3
4
4
3
3
2
3
4
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7NBI Nuclear Boiler Instrumentation (continued)
K/A NO.
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Nuclear Boiler Instrumentation controls including:
(CFR: 41.5/45.5)
A1.01
A1.02
A1.03
A1.04
Control room indications
Removing or returning a transmitter from/to service
Surveillance testing
System venting
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Nuclear Boiler Instrumentation; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
IMPORTANCE
3
3
3
3
RO SRO
3
3
3
3
A2.01
A2.02
Loss of power supply
Reactor coolant temperature
A3
Ability to monitor automatic operations of the Nuclear Boiler
Instrumentation including:
(CFR: 41.7 / 45.7)
A3.01
Relationship between display readings and actual parameter values
A4
Ability to manually operate and/or monitor the Nuclear Boiler
Instrumentation in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Channel select controls
Process computer
3
3
3
3.7-13
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RAD Radiation Monitoring System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Radiation Monitoring System and the following systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
Main steam system
Offgas system
Plant stack
Reactor building cooling water system
Radioactive waste treatment system
Reactor building heating, ventilation, and air conditioning system
Fuel building heating, ventilation, and air conditioning system
Service building heating, ventilation, and air conditioning system
Leak detection and isolation system
Turbine building heating, ventilation, and air conditioning system
Radwaste building heating, ventilation, and air conditioning system
Plant information and control system
Emergency response information system
Vital AC power supply system
Instrument air system
Engineered Safety Function logic and control system
Standby gas treatment system
Makeup water purified system
Control building heating, ventilation, and air conditioning system
Extraction system
K2
Knowledge of bus or electrical power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
K2.07
Main steamline radiation monitors
Offgas radiation monitoring system
Plant stack radiation monitoring
Radwaste liquid radiation monitoring system
Reactor building heating, ventilation, and air conditioning monitors
Area radiation monitors
Control room ventilation monitors
K3
Knowledge of the effect that a loss or malfunction of the Radiation
Monitoring System will have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
K3.04
Station liquid effluent release monitoring
Station gaseous effluent release monitoring
Station area radiation monitoring
Offgas system
NUREG-2104
IMPORTANCE
3.7-14
2
3
3
3
3
3
3
2
4
3
3
3
2
2
2
3
3
2
3
3
3
3
3
2
3
2
2
3
3
3
4
3.7
Safety Function 7: Instrumentation
System:
SF7RAD Radiation Monitoring System (continued)
K/A NO.
KNOWLEDGE
K3.05
K3.06
K3.07
K3.08
Reactor building heating, ventilation, and air conditioning
Drywell sump liquid discharge
Radwaste building ventilation
Control building heating, ventilation, and air conditioning system
K4
Knowledge of Radiation Monitoring System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Redundancy
Automatic actions to contain the radioactive release
in the event that the predetermined release rates are exceeded
Fail safe tripping of process radiation monitoring
logic during conditions of instrument failure
K4.03
IMPORTANCE
3
2
3
3
3
4
4
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Radiation Monitoring System:
(CFR: 41.7 / 45.4)
K5.01
K5.02
K5.03
K5.04
K5.05
Hydrogen injection operation's effect on process radiation indications
Drywell sump liquid discharge
Turbine gland seal condenser exhaust
Fuel handling area ventilation exhaust
Drywell
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Radiation
Monitoring System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
Vital AC power
Plant information and control system
Leak detection and isolation system
3
2
2
3
3
3
2
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Radiation Monitoring System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
Alarms and indications associated with normal operations
Alarms and indications associated with surveillance testing
3.7-15
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RAD Radiation Monitoring System (continued)
K/A NO.
ABILITY
A2
Ability to (d) predict the impacts of the following on the Radiation
Monitoring System; and (b) based on those predictions, use procedures to
correct, control, or mitigate the consequences of those abnormal
conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
Fuel element failure
AC electrical failure
Downscale trips
Offgas system failure
Low fuel pool level
Loss of coolant accident
Leakage and/or breaks from contaminated systems to
atmosphere or to other process systems
Refuel floor handling accidents/operations
Low reactor water level during refueling operations
A2.08
A2.09
IMPORTANCE
RO SRO
4
4
3
3
3
3
3
3
3
3
4
4
3
3
3
4
4
4
A3
Ability to monitor automatic operations of the Radiation Monitoring System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
Offgas system isolation indications
Liquid radwaste isolation indications
Radwaste handing interlocks
Drywell LCW or HCW sump isolation indications
Ventilation system isolation indications
Display indications
Meter indications
Containment isolation indications
Lights and alarms
A4
Ability to manually operate and/or monitor the Radiation Monitoring
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4. 01
Manually trip process radiation monitor logic
NUREG-2104
3.7-16
4
3
2
3
3
3
3
4
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7RTIS Reactor Trip and Isolation System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Trip and Isolation System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Neutron monitoring system
Nuclear boiler system
Vital AC power supply system
Control rod drive system
Plant information and control system
Turbine main steam system
Suppression pool temperature monitoring system
Rod control and information system
Recirculation flow control system
Leak detection and isolation system
Engineered Safety Function logic and control system
Direct current power supply system
Feedwater control system
Instrumentation and control power supply system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Reactor Trip and Isolation System logic channels
K3
Knowledge of the effect that a loss or malfunction of the Reactor Trip and
Isolation System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
Leak detection and isolation system
4
Reactor Protection System logic channels
4
Scram air header solenoid operated valves
4
Reactor power
4
Reactor coolant primary system integrity
4
The magnitude of heat energy that must be absorbed by the containment
during accident/transient conditions
3
The ability of the core cooling systems to provide adequate core cooling
during loss of coolant accidents
4
Recirculation flow control system
3
Secondary containment integrity
3
Plant information and control
3
Engineered Safety Function logic and control system
3
Feedwater control system
3
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
IMPORTANCE
3.7-17
4
4
3
3
3
3
3
3
3
3
3
3
3
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RTIS Reactor Trip and Isolation System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Reactor Trip and Isolation System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
System redundancy and reliability
3
The prevention of a reactor scram following a single component
failure
4
Functional testing of the system while maintaining power operation
4
Manual system activation
4
Complete control rod insertion following scram signal generation
4
Control rod insertion following RPS system electrical failure
4
Dual-rod scram testing
3
Operation with the neutron monitoring system selection
switch in non-coincident position
3
Initiation of the “scram follow” function following a reactor scram signal4
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
IMPORTANCE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Reactor Trip and Isolation System:
(CFR: 41.5 / 45.3)
K5.01
Specific logic arrangements
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Trip and
Isolation System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
Vital AC power supply system
Neutron monitoring system
Main steam system
Main condenser vacuum
Turbine main steam system
Leak detection and isolation system
Control rod drive system
Suppression pool temperature monitoring system
Direct current power supply system
Instrumentation and control power supply system
NUREG-2104
3.7-18
3
3
3
3
3
3
3
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7RTIS Reactor Trip and Isolation System (continued)
K/A NO
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Trip and Isolation System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
Reactor power
Reactor water level
Reactor pressure
4
3
3
A1.04
Control rod position
3
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Trip and Isolation System; and
(b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
A2.16
A2.17
A2.18
A2.18
A2.20
A2.21
A2.22
RTIS power supply failure
Neutron monitoring system failure
RTIS logic channel failure
High reactor power
High reactor pressure
Low reactor level
High main steam line flow
Low main steam line pressure
High drywell pressure
Turbine stop valve closure signal
Control valve fast closure signal
Main steamline isolation valve closure
Low condenser vacuum
Low control rod drive charging header pressure
Changing mode switch position
Half-scram signal
Complete scram signal
Half-MSIV isolation signal
Complete MSIV isolation signal
High main steam line tunnel temperature
High suppression pool temperature
Short reactor period
IMPORTANCE
3.7-19
RO SRO
4
4
4
4
3
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
2
2
4
4
2
2
4
4
4
4
4
4
4
4
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RTIS Reactor Trip and Isolation System (continued)
K/A NO
ABILITY
A3
Ability to monitor automatic operations of the Reactor Trip and Isolation
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
Reactor power
Control rod position
Control room indications and alarms
End-of-cycle recirculation pump trip
A4
Ability to manually operate and/or monitor the Reactor Trip and Isolation
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
A4.04
A4.05
A4.06
A4.07
A4.08
A4.09
Reactor mode switch
Manual scram pushbuttons
Manual scram reset switch
Divisional manual main steam line isolation switch
CRD charging pressure trip bypass switch
Divisional RPS/MSIV sensor bypass switch
Divisional RPS/MSIV trip logic function trip bypass switch
Divisional trip logic function auto trip test switch
Main steam line isolation bypass switch
NUREG-2104
IMPORTANCE
3.7-20
4
4
4
4
4
4
4
3
4
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7MRBM Multi-Channel Rod Block Monitor System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Multi-Channel Rod Block Monitor System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
Average power range/local power range monitor sub-system
Rod control and information system
Plant information and control system
Vital AC power supply system
Automated thermal limit monitor system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Multi-channel rod block monitor channels
K3
Knowledge of the effect that a loss or malfunction of the Multi-Channel Rod
Block Monitor System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Rod control and information system
K4
Knowledge of Multi-Channel Rod Block Monitor System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
Rod withdrawal blocks
Allows manual or auto setup of rod block lines during power increase
Automatic setdown of rod block lines during power reduction
Initiation point
Bypass an MRBM channel
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Multi-Channel Rod Block Monitor System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Trip reference selection
Local power range monitors
Control rod selection
IMPORTANCE
3
3
3
3
3
2
3
3
3
3
3
3
3
3
3
3.7-21
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7MRBM Multi-Channel Rod Block Monitor System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Multi-Channel
Rod Block Monitor System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
Vital AC power supply system
Local power range monitor detectors
Automated thermal limit monitor system
Core flow signal
IMPORTANCE
2
3
3
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Multi-Channel Rod Block Monitor System controls including:
(CFR: 41.5 / 45.5)
A1.01
Trip reference
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Multi-Channel Rod Block Monitor
System ; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
Withdrawal of control rod in high power region of core
Loss or reduction in core flow (flow comparator)
Loss of associated Average Power Range Monitor channel
Power supply loss
Multi-channel rod block monitor upscale or inoperable
A3
Ability to monitor automatic operations of the Multi-Channel Rod Block
Monitor System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Control room indications and alarms
Verification or proper functioning/operability
Back panel indications
Automatic setup when in automatic rod withdrawal mode
Automatic setdown of setpoint during power reduction
NUREG-2104
3
3.7-22
RO SRO
3
4
3
3
3
3
3
3
3
3
3
4
4
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7MRBM Multi-Channel Rod Block Monitor System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Multi-Channel Rod Block
Monitor System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
A4.03
MRBM back panel switches, indications and indicating lights
Trip bypasses
"Setup" pushbutton
IMPORTANCE
3.7-23
3
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RWM Rod Worth Minimizer System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Rod Worth Minimizer System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
Rod control and information system
Neutron monitoring system
Reactor trip and isolation system
Vital AC power supply system
Plant information and control system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Rod worth minimizer circuitry
K3
Knowledge of the effect that a loss or malfunction of the Rod Worth
Minimizer System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Rod control and information system
K4
Knowledge of Rod Worth Minimizer System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Insert blocks
Withdraw blocks
Automatic bypass above low power setpoint
System testing
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Rod Worth Minimizer System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.02
K5.03
K5.04
K5.05
Low power set point
Low power alarm point
Rod groups and steps
Withdraw block
Insert block
Possible fuel damage due to rod withdrawal error
NUREG-2104
IMPORTANCE
3.7-24
3
3
3
3
3
2
3
3
4
3
3
3
3
3
4
4
3
3.7
Safety Function 7: Instrumentation
System:
SF7RWM Rod Worth Minimizer System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Rod Worth
Minimizer System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
Vital AC power supply system
Rod control and information system
Neutron monitoring system
Reactor trip and isolation system
Inoperable control rod
Stuck control rod
IMPORTANCE
3
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Rod Worth Minimizer System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
Rod position
Status of control rod movement blocks
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Rod Worth Minimizer System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
3
3
RO SRO
3
3
3
3
3
3
3
4
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
Power supply loss
Inoperable control rod
Stuck control rod
Out of sequence rod movement
RWM hardware/software failure
A3
Ability to monitor automatic operations of the Rod Worth Minimizer System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
System indication
Verification of proper functioning/operability
Annunciator and alarm signals
Control rod movement blocks
Automatic bypass above the low power setpoint
3.7-25
3
4
3
4
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RWM Rod Worth Minimizer System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Rod Worth Minimizer in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Bypass single channel of the rod worth minimizer circuitry
Dual channel bypass for special operations
NUREG-2104
IMPORTANCE
3.7-26
3
2
3.7
Safety Function 7: Instrumentation
System:
SF7SRNM Startup Range Neutron Monitoring System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Startup Range Neutron Monitoring System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
Reactor trip and isolation system
Rod control and information system
Plant information and control system
Reactor pressure vessel
Primary containment system
Engineered Safety Function logic and control system
Automatic power regulator system
Vital AC power supply system
Alternate reactor shutdown system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Startup range neutron monitoring channels/detectors
K3
Knowledge of the effect that a loss or malfunction of the Startup Range
Neutron Monitoring System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Reactor trip and isolation system
Rod control and information system
Reactor power indications
Automatic power regulator system
Engineered Safety Function logic and control system
Plant information and control system
Alternate reactor shutdown system
K4
Knowledge of Startup Range Neutron Monitoring System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Rod withdrawal blocks
Reactor scram signals
Gamma compensation
Different power determination methods (counting and Campbelling
technique)
Bypassing a startup range neutron monitoring channel
SRNM indications when changing reactor mode switch position
K4.05
K4.06
IMPORTANCE
3.7-27
4
3
3
3
3
3
3
3
3
3
3
3
4
3
3
3
3
4
3
2
2
3
2
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7SRNM Startup Range Neutron Monitoring System (continued)
K/A NO.
KNOWLEDGE
K4.07
K4.08
K4.09
ATWS permissive for automatic SLC initiation
Non-coincidence scram signals during initial fuel loading
Alarm seal-in
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Startup Range Neutron Monitoring System:
(CFR: 41.5 / 45.3)
K5.01
Gamma discrimination
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Startup Range
Neutron Monitoring System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
Reactor trip and isolation system
Vital AC power supply system
Startup range neutron monitoring detectors
Reactor vessel
Primary containment system
IMPORTANCE
3
2
2
2
3
3
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Startup Range Neutron Monitoring System controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
Reactor power indication
Reactor trip and isolation system status
Control rod block status
Scram, rod block, and period alarm trip setpoints
Indications and alarms
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Startup Range Neutron Monitoring
System; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
Vital AC power supply loss
Startup range neutron monitoring inoperable condition
NUREG-2104
3.7-28
4
3
4
4
3
RO SRO
3
3
3
4
3.7
Safety Function 7: Instrumentation
System:
SF7SRNM Startup Range Neutron Monitoring System (continued)
K/A NO.
ABILITY
A2.03
A2.04
Up scale and downscale trips
Faulty or erratic operation of detectors/system
A3
Ability to monitor automatic operations of the Startup Range Neutron
Monitoring System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Control room indications
Annunciator and alarm signals
Reactor trip and isolation system status
Control rod block status
ATWS permissive for SLC initiation
A4
Ability to manually operate and/or monitor the Startup Range Neutron
Monitoring in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Startup range neutron monitoring channel bypass switches
IMPORTANCE
3.7-29
RO SRO
4
4
3
4
3
3
4
4
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7ATIP Automated Traversing In-Core Probe System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between Automated Traversing In-Core Probe System and the following:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
Local power range monitors
Plant information and control system
Instrument air system
Leak detection and isolation system
DC power supply system
Electrical power distribution system
Instrumentation and control power supply system
Reactor pressure vessel
Primary containment system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
Shear valves
ATIP channels/detectors
K3
Knowledge of the effect that a loss or malfunction of the Automated
Traversing In-Core Probe System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Local power range monitor's calibration
K4
Knowledge of Automated Traversing In-Core Probe System design
feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Primary containment isolation
Radiation shielding
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Automated Traversing In-Core Probe System:
(CFR: 41.5 / 45.3)
K5.01
Increasing area radiation monitor indications
NUREG-2104
IMPORTANCE
3
3
2
3
2
2
2
3
3
2
2
3.7-30
2
3
2
3.7
Safety Function 7: Instrumentation
System:
SF7ATIP Automated Traversing In-Core Probe (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Automated
Traversing In-Core Probe System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
DC power supply system
AC electrical power distribution system
Instrumentation and control power supply system
Plant information and control system
Leak detection and isolation system
Instrument air system
Nuclear boiler system
IMPORTANCE
2
2
2
3
3
2
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Automated Traversing In-Core Probe System controls
including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
Radiation levels
Detector position
Valve status
Drive speed
Detector output
Radiation alarms
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Automated Traversing In-Core Probe
System; and (b) based on those predictions, use procedures to correct,
control, or mitigate the consequences of those abnormal conditions or
operations:
(CFR: 41.5/45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
3
3
3
2
2
3
Low reactor water level
High drywell pressure
Drive mechanism failure
Electrical power distribution system
DC power supply system
Shear valve closures
Failure to retract during accident conditions
Failure to retract to shield
3.7-31
RO SRO
3
3
3
3
2
2
2
2
2
2
2
3
3
4
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7ATIP Automated Traversing In-Core Probe (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Automated Traversing InCore Probe System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
Detector position
Detector drive speed
Valve operation
Indicating lights
Detector output
A4
Ability to manually operate and/or monitor the Automated Traversing InCore Probe System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Isolation valves
NUREG-2104
IMPORTANCE
2
2
3
2
2
3
3.7-32
3.7
Safety Function 7: Instrumentation
System:
SF7ELCS Engineered Safety Function Logic and Control System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Engineered Safety Function Logic and Control System and the
following systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
Reactor trip and isolation system
Neutron monitoring system
Standby liquid control system
Process radiation monitoring system
Plant information and control system
DC power supply system
Instrumentation and control power system
Leak detection and isolation system
Suppression pool temperature monitoring system
High pressure core flooder system
Reactor core isolation cooling system
Residual heat removal system
Automatic depressurization system
Emergency diesel generators
Standby gas treatments system
Reactor building cooling water system
Reactor service water system
Heating, ventilation, and air conditioning system emergency cooling
water system
High-pressure nitrogen gas supply system
Alternate reactor shutdown system
Feedwater system (feedwater line break circuitry)
Containment atmospheric monitoring system
Reactor building heating, ventilation, and air conditioning system
Control building heating, ventilation, and air conditioning system
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
K2
IMPORTANCE
3
3
4
3
3
3
3
3
3
4
4
4
4
3
3
3
3
3
3
3
3
3
3
3
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
None
K3
Knowledge of the effect that a loss or malfunction of the Engineered Safety
Function Logic and Control System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
Reactor trip and isolation system
Neutron monitoring system
Standby liquid control system
Process radiation monitoring system
3.7-33
3
3
4
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7ELCS Engineered Safety Function Logic and Control System
(continued)
K/A NO.
KNOWLEDGE
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
K3.14
K3.15
K3.16
Plant information and control system
Leak detection and isolation system
Suppression pool temperature monitoring system
High pressure core flooder system
Reactor core isolation cooling system
Residual heat removal system
Automatic depressurization system
Emergency diesel generators
Standby gas treatment system
Reactor building cooling water system
Reactor service water system
Heating, ventilation, and air conditioning system emergency cooling
water system
High-pressure nitrogen gas supply system
Alternate reactor shutdown system
Feedwater system (feedwater line break circuitry)
Containment atmospheric monitoring system
Reactor building heating, ventilation, and air conditioning system
Control building heating, ventilation, and air conditioning system
K3.17
K3.18
K3.19
K3.20
K3.21
K3.22
IMPORTANCE
3
3
3
4
4
4
4
3
3
3
3
3
3
3
3
3
3
3
K4
Knowledge of Engineered Safety Function Logic and Control System
design feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Redundancy of instrumentation
On-line self diagnostic testing
Initiation of the emergency core cooling systems
Inputs to alternate rod insertion circuitry
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Engineered Safety Function Logic and
Control System:
(CFR: 41.5 / 45.3)
3
3
4
4
None
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Engineered
Safety Function Logic and Control System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
Instrumentation and control power supply system
DC power supply system
NUREG-2104
3.7-34
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7ELCS Engineered Safety Function Logic and Control System
(continued)
K/A NO.
ABILITY
A 1.
Ability to predict and/or monitor changes in parameters associated with
operating the Engineered Safety Function Logic and Control System
controls including:
(CFR: 41.5/45.5)
A1.01
A1.02
Removing or returning a channel to service
Surveillance testing
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Nuclear Boiler Instrumentation; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
IMPORTANCE
3
3
RO SRO
Loss of AC or DC power supply
3
3
System initiation/isolation due to Engineered Safety Function logic and control
actuation
3
3
A3
Ability to monitor automatic operations of the Engineered Safety Function
Logic and Control System including:
(CFR: 41.7 / 45.7)
A3.01
System initiation/isolation
A4
Ability to manually operate and/or monitor the Engineered Safety Function
Logic and Control System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Removing or returning a channel to service
3
3.7-35
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7PICS Plant Information and Control System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Plant Information and Control System and the following
systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
Main steam system
Reactor recirculation system
Control rod drive system
Leak detection and isolation system
Reactor water cleanup system
Fuel pool cooling and cleanup system
Suppression pool cleanup system
Turbine main steam system
Condensate, feedwater and condensate air extraction system
High vent and drain system
Main turbine
Steam bypass and pressure control system
Turbine gland steam system
Turbine lube oil system
Moisture separator reheaters
Hydrogen gas cooling system
Generator cooling system
Generator seal oil system
Main condenser
Offgas system
Circulating water system
Makeup water purified system
Turbine building cooling water system
Heating, ventilation, and air conditioning system normal cooling
water system
Turbine service water system
Atmosphere control system
Drywell cooling system
Control building heating, ventilation, and air conditioning system
Turbine building heating, ventilation, and air conditioning system
Instrument air system
AC electrical power distribution system
Makeup water condensate system
Neutron monitoring system
Reactor trip and isolation system
Engineered Safety Function logic and control system
Rod control and information system
Feedwater control system
K1.25
K1.26
K1.27
K1.28
K1.29
K1.30
K1.31
K1.32
K1.33
K1.34
K1.35
K1.36
K1.37
NUREG-2104
IMPORTANCE
3.7-36
3
3
3
3
3
2
2
3
2
2
2
3
2
2
2
2
2
2
2
3
2
2
2
2
2
3
3
3
2
2
3
2
3
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7PICS Plant Information and Control System (continued)
K/A NO.
KNOWLEDGE
K1.38
K1.39
K1.40
K1.41
K1.42
K1.43
K1.44
K1.45
K1.46
K1.47
K1.48
K1.49
K1.50
K1.51
K1.52
K1.53
K1.54
Recirculation flow control system
Standby liquid control system
Process radiation monitoring system
Area radiation monitoring system
Alternate feedwater injection system
Automated traversing incore probe system
Automatic power regulator system
Instrumentation and control power supply system
Direct current power supply system
Loose parts monitoring system
Asset monitoring system
Plant data network
Containment atmospheric monitoring system
Fuel servicing equipment
Turbine supervisory system
Combustion turbine system
Emergency diesel generator system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
3
3
2
3
2
3
3
3
2
2
2
3
3
3
3
3
None
K3
Knowledge of the effect that a loss or malfunction of the Plant Information
and Control System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
K3.14
K3.15
K3.16
K3.17
Main steam system
Reactor recirculation system
Control rod drive system
Leak detection and isolation system
Reactor water cleanup system
Fuel pool cooling and cleanup system
Suppression pool cleanup system
Turbine main steam system
Condensate, feedwater and condensate air extraction system
High vent and drain system
Main turbine
Steam bypass and pressure control system
Turbine gland steam system
Turbine lube oil system
Moisture separator reheaters
Hydrogen gas cooling system
Generator cooling system
3.7-37
3
3
3
3
3
2
2
3
2
2
2
3
2
2
2
2
2
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7PICS Plant Information and Control System (continued)
K/A NO.
KNOWLEDGE
K3.18
K3.19
K3.20
K3.21
K3.22
K3.23
K3.24
Generator seal oil system
Main condenser
Offgas system
Circulating water system
Makeup water purified system
Turbine building cooling water system
Heating, ventilation, and air conditioning system normal cooling
water system
Turbine service water system
Atmosphere control system
Drywell cooling system
Control building heating, ventilation, and air conditioning system
Turbine building heating, ventilation, and air conditioning system
Instrument air system
AC electrical power distribution system
Makeup water condensate system
Neutron monitoring system
Reactor trip and isolation system
Engineered Safety Function logic and control system
Rod control and information system
Automated thermal limit monitoring system
Feedwater control system
Recirculation flow control system
Standby liquid control system
Process radiation monitoring system
Area radiation monitoring system
Alternate feedwater injection system
Automated traversing incore probe system
Automatic power regulator system
Instrumentation and control power supply system
Direct current power supply system
Loose parts monitoring system
Asset monitoring system
Plant data network
Containment atmospheric monitoring system
Fuel servicing equipment
Turbine supervisory system
Combustion turbine system
Emergency diesel generator system
K3.25
K3.26
K3.27
K3.28
K3.29
K3.30
K3.31
K3.32
K3.33
K3.34
K3.35
K3.36
K3.37
K3.38
K3.39
K3.40
K3.41
K3.42
K3.43
K3.44
K3.45
K3.46
K3.47
K3.48
K3.49
K3.50
K3.51
K3.52
K3.53
K3.54
K3.55
NUREG-2104
IMPORTANCE
3.7-38
2
2
3
2
2
2
2
2
3
3
3
2
2
3
2
3
3
3
3
3
3
3
3
3
2
3
2
3
3
3
2
2
2
3
2
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7PICS Plant Information and Control System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Plant Information and Control System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Redundancy of instrumentation
On-line self diagnostic testing
K5
Knowledge of the operational implications or cause-effects relationships as
they apply to Plant Information and Control System:
(CFR: 41.5 / 45.3)
K5.01
Power generation control system
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Plant Information
and Control System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
Instrumentation and control power supply system
DC power supply system
IMPORTANCE
2
2
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Plant Information and Control System controls including:
(CFR: 41.5 / 45.5)
A1.01
Control room indications associated with normal operations
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Plant Information and Control System ;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
AC or DC power supply failure
3.7-39
2
RO SRO
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7PICS Plant Information and Control System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of Plant Information and Control
System including:
(CFR: 41.7 / 45.7)
A3.01
Reactor startup, power operation, and shutdown of the plant in
Power Generation Control System “automatic or semi-automatic
mode of operation
IMPORTANCE
3
A4
Ability to manually operate and/or monitor the Plant Information and
Control System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Activate appropriate breakpoint to allow continuation of Power
Generation Control System automatic operation
NUREG-2104
3.7-40
3
3.7
Safety Function 7: Instrumentation
System:
SF7SPTM Suppression Pool Temperature Monitoring System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Suppression Pool Temperature Monitoring System and the
following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
Engineered Safety Function logic and control system
4
Plant information and control system
3
Reactor trip and isolation system
4
Primary containment system (suppression pool temperature and level)3
Remote shutdown system
3
Residual heat removal system
4
Vital AC power supply system
3
Reactor building cooling water system
4
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Suppression pool temperature monitoring logic channels
K3
Knowledge of the effect that a loss or malfunction of the Suppression Pool
Temperature Monitoring System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
Primary containment system (suppression pool temperature
and level)
Reactor trip and isolation system
Residual heat removal system
Remote shutdown system
Plant information and control system
Reactor building cooling water system
K3.02
K3.03
K3.04
K3.05
K3.06
IMPORTANCE
2
3
4
4
3
3
4
K4
Knowledge of Suppression Pool Temperature Monitoring System design
feature(s) and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
System redundancy and reliability
Arrangement of temperature sensors in suppression pool
Method of determining average temperature of suppression pool
Suppression pool temperature monitoring logic arrangement
3.7-41
3
2
3
3
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7SPTM Suppression Pool Temperature Monitoring System (continued)
K/A NO
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Suppression Pool Temperature Monitoring
System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
Low suppression pool level
Safety/relief valve operation
K6
Knowledge of the effect of the following plant malfunctions, system
malfunctions or component malfunctions will have on the Suppression
Pool Temperature Monitoring System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
Primary containment system (suppression pool temperature and level)3
Vital AC power supply system
3
Engineered Safety Function logic and control system
3
IMPORTANCE
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Suppression Pool Temperature Monitoring System controls
including:
(CFR: 41.5 / 45.5)
A1.01
Surveillance testing
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Suppression Pool Temperature
Monitoring System; and (b) based on those predictions, use procedures to
correct, control, or mitigate the consequences of those abnormal
conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
NUREG-2104
2
High suppression pool temperature
Low suppression pool level
Failure or erratic operation of suppression pool temperature/level
detector
3.7-42
RO SRO
3
3
3
3
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7SPTM Suppression Pool Temperature Monitoring System (continued)
K/A NO
ABILITY
A3
Ability to monitor automatic operations of the Suppression Pool
Temperature Monitoring System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Control room indications and alarms
Initiations and actuations due high suppression pool temperature
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
SPTM backpanel switches
IMPORTANCE
3
3
3
3.7-43
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RSS Remote Shutdown System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Remote Shutdown System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Main steam system
Nuclear Boiler Instrumentation system
Residual heat removal system
High pressure core flooder system
Reactor building cooling water system
AC electrical distribution system
Reactor service water system
Atmosphere control system
Makeup water condensate system
Emergency diesel generator system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
3
4
3
3
3
3
3
2
3
None
K3
Knowledge of the effect that a loss or malfunction of the Remote Shutdown
System will have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
Main steam system
Nuclear Boiler Instrumentation system
Residual heat removal system
High pressure core flooder system
Reactor building cooling water system
AC electrical distribution system
Reactor service water system
Atmosphere control system
Makeup water condensate system
Emergency diesel generator system
K4
Knowledge of Remote Shutdown System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
Ability to transfer control of interfacing system equipment from the
main control room and override capability (manual and automatic)
of operating the interfacing equipment from the main control room
NUREG-2104
3.7-44
3
3
4
3
3
3
3
3
2
3
3
3.7
Safety Function 7: Instrumentation
System:
SF7RSS Remote Shutdown System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Remote Shutdown System:
(CFR: 41.7 / 45.4)
K5.01
Heat removal mechanisms
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Remote
Shutdown System:
(CFR: 41.7 / 45.7)
K6.01
Conditions which cause the main control room to become
inaccessible
IMPORTANCE
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Remote Shutdown System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
A1.12
A1.13
A1.14
A1.15
Reactor pressure vessel water level
Reactor pressure vessel pressure
Residual heat removal pump discharge pressure
Residual heat removal heat exchanger inlet temperature
Residual heat removal heat exchanger outlet temperature
Residual heat removal system flow
High pressure core flooder system flow
High pressure core flooder pump discharge pressure
Reactor service water system ultimate heat sink water level
Reactor service water system ultimate heat sink temperature
Reactor building cooling water system flow
Suppression pool level
Drywell temperature
Drywell pressure
Condensate storage tank level
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Remote Shutdown System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
Control room evacuation
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
RO SRO
3
4
3.7-45
NUREG-2104
3.7
Safety Function 7: Instrumentation
System:
SF7RSS Remote Shutdown System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Remote Shutdown System
including:
(CFR: 41.7 / 45.7)
IMPORTANCE
None
A4
NUREG-2104
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
None
3.7-46
3.8
Safety Function 8: Plant Service Systems
System
SF8SFS Fire Protection System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Fire Protection System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Main generator hydrogen system
Auxiliary steam system
A.C electrical power distribution system
Residual heat removal system
Fire detection system
Makeup water preparation system
Instrument air system
Turbine lube oil system
Fuel oil transfer system
Emergency diesel generators
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Fire protection supervisory circuits
Motor driven pumps
Fire detection system
K3
Knowledge of the effect that a loss or malfunction of the Fire Protection
System will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
The ability to detect fires
Personnel protection
Plant protection
Residual heat removal system
K4
Knowledge of Fire Protection System design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.5 / 41.7 / 45.3 / 45.5 to 45.8)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
Adequate supply of water for the fire protection system
3
Automatic system initiation
3
Maintenance of fire header pressure
3
Personnel safety during halon and/or carbon dioxide system actuation 4
Fire protection capability during loss of off-site power
4
Fire suppression capability that does not rely on the displacement
of oxygen (Halon)
3
Diesel engine protection
3
Ability to provide alternate source of water to the reactor pressure
vessel or primary containment
4
K4.07
K4.08
IMPORTANCE
3.8-1
3
2
3
4
2
2
2
2
2
2
2
3
2
3
3
4
4
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8SFS Fire Protection System (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Fire Protection System:
(CFR: 41.5/45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
K5.09
Effect of carbon dioxide on fires
Effect of Halon on fires
Effect of water spray on electrical components
Diesel operations
Heat detection
Smoke detection
Reactor water level
Emergency generator rooms
Main generator exciter
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Fire Protection
System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
AC electrical power distribution system
Diesel fuel transfer system
Fire detection system
Makeup water preparation system
Instrument air system
IMPORTANCE
3
3
3
3
3
3
3
3
3
3
3
2
2
2
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Fire Protection System controls including:
(CFR: 41.5/45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
System pressure
System flow
Fire doors
Fire dampers
System lineups
NUREG-2104
3
2
3
3
3
3.8-2
3.8
Safety Function 8: Plant Service Systems
System
SF8SFS Fire Protection System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Fire Protection System; and (b) based on
those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
System logic failure
AC electrical power distribution system failure
Fire protection diesel trips
Low fire header pressure
Inadvertent system initiation
Failure to actuate when required
Pump trips
Low diesel fuel supply
A3
Ability to monitor automatic operations of the Fire Protection System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
Fire water pump start
Fire main pressure
Actuation of fire detectors
System initiation
Fire doors
Fire dampers
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Electric fire pump
Diesel fire pump
IMPORTANCE
RO SRO
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3.8-3
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8FHS Fuel Handling Equipment
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
K1
Knowledge of the physical or control/protection logic relationships
between the Fuel Handling Equipment and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
Rod control and information system
Fuel pool ventilation
Fuel pool cooling and cleanup system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
3
3
2
3
3
2
None
K3
Knowledge of the effect that a loss or malfunction of the Fuel Handling
Equipment will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
Rod control and information system
Fuel handling operations
Core modifications/alterations
K4
Knowledge of Fuel Handling Equipment design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Prevention of core alterations during control rod movements
Prevention of control rod movement during core alterations
Protection against inadvertently lifting radioactive components out
of the water
Movement of the spent fuel cask only over designated areas
K4.04
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Fuel Handling Equipment:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
Crane/hoist operation
Fuel handling equipment interlocks
Water as a shield against radiation
Spent fuel pool design
Fuel orientation
Fuel
Core components
NUREG-2104
3.8-4
3
3
3
3
4
4
3
3
4
4
3
3
4
3
3
3
3
3
3
3
3
3
4
3
4
4
3
3
3.8
Safety Function 8: Plant Service Systems
System
SF8FHS Fuel Handling Equipment (continued)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
K5.08
K5.09
K5.10
K5.11
Spent fuel cask
Reactor vessel components
Fuel pool configuration
Refuel floor ventilation
3
3
3
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Fuel Handling
Equipment
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
AC electrical power distribution system
Rod control and information system
Refueling platform air system
Fuel pool water inventory
Refuel floor ventilation
Fuel pool cooling and cleanup system
3
3
3
3
3
2
3
3
3
3
3
4
4
3
3
2
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Fuel Handling Equipment controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
Spent fuel pool level
Refuel floor radiation levels/airborne levels
Core reactivity level
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Fuel Handling Equipment; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
Interlock failure
Loss of refueling platform air system
Loss of electrical power
3.8-5
3
3
3
3
3
3
3
4
4
3
4
3
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8FHS Fuel Handling Equipment (continued)
K/A NO.
ABILITY
IMPORTANCE
RO SRO
A3
Ability to monitor automatic operations of the Fuel Handling Equipment
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Crane/refuel bridge movement
Interlock operation
A4
Ability to manually operate and/or monitor Fuel Handling Equipment at the
equipment location:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Neutron monitoring system
Control rod drive system
NUREG-2104
3
3
4
3
3.8-6
4
4
4
4
3.8
Safety Function 8: Plant Service Systems
System
SF8IAS Instrument Air System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Instrument Air System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
Service air system
High pressure nitrogen gas supply system
Reactor building cooling water system
Main steam system
Engineered Safety Function logic and control system
Control rod drive system
Automated traversing in core probe system
Offgas system
AC electrical power distribution system
Extraction system
Fire protection system
Neutron monitoring system
Main turbine
Leak detection and isolation system
Instrument and control power supply system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Instrument air compressor
K3
Knowledge of the effect that a loss or malfunction of the Instrument Air
System will have on the following:
(CFR: 41.7 / 45.6)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
Service air system
Main steam system
Control rod drive system
Automated traversing in core probe system
Offgas system
Extraction system
Fire protection system
Neutron monitoring system
Main turbine
IMPORTANCE
3
3
3
3
2
3
2
2
3
2
2
3
2
2
2
3
3.8-7
3
3
3
2
2
2
2
3
2
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8IAS Instrument Air System (continued)
K/A NO.
KNOWLEDGE
K4.
Knowledge of Instrument Air System design feature(s) and or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Manual/automatic transfers of control
Cross-over to other air systems
Securing of IAS upon loss of cooling water
K5.
Knowledge of the operational implications or cause and effect
relationships as they apply to the Instrument Air System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
Air compressors
Service air cross-connect valve
Pneumatic operated devices
K6
Knowledge of the effect of the following plant conditions, plant
malfunctions or component malfunctions will have on the Instrument Air
System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
Air compressors
Service air cross-connect valve
Engineered Safety Function logic and control system
Electrical power distribution system
High pressure nitrogen gas supply system
Reactor building cooling water system
Instrument and control power supply system
IMPORTANCE
3
3
3
3
2
2
2
3
2
3
3
2
2
ABILITY
A1
Ability to predict and / or monitor changes in parameters associated with
operating the Instrument Air System controls including:
(CFR: 41.5 / 45.5)
None
NUREG-2104
3.8-8
3.8
Safety Function 8: Plant Service Systems
System
SF8IAS Instrument Air System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Instrument Air System and (b) based on
those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal operations:
(CFR: 41.5 / 45.6)
A2.01
Air dryer and filter malfunctions
A3
Ability to monitor automatic operations of the Instrument Air System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Air pressure
Air temperature
A4
Ability to manually operate and/or monitor the Instrument Air System in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
Instrument air compressors
IMPORTANCE
RO SRO
3
3
3
2
3
3.8-9
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8RBCW Reactor Building Cooling Water System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Building Cooling Water System and the following
systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
Reactor service water system
Process radiation monitoring system
Reactor recirculation system
Remote shutdown system
Heating, ventilation, and air conditioning system emergency cooling
water system
Engineered Safety Function logic and control system
Makeup water purified system
Suppression pool cleanup system
Instrument air system
Control rod drive system
Leak detection and isolation system
Recirculation flow control system
Residual heat removal system
High pressure core flooder system
Reactor water cleanup system
Fuel pool cooling and cleanup system
Sampling system
Radioactive drain transfer system
Post accident monitoring system
AC electrical power distribution system
Emergency diesel generator system
Suppression pool temperature monitoring system
Drywell cooling system
Heating, ventilation and air conditioning system
Containment atmospheric monitoring system
Service air system
Turbine building cooling water system
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
K1.25
K1.26
K1.27
IMPORTANCE
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
RBCW pumps
RBCW containment isolation valves
NUREG-2104
3.8-10
4
3
3
3
2
3
2
2
3
3
3
3
3
3
3
3
2
2
2
3
3
3
3
3
3
2
2
3
3
3.8
Safety Function 8: Plant Service Systems
System
SF8RBCW Reactor Building Cooling Water System (continued)
K/A NO.
KNOWLEDGE
K3
Knowledge of the effect that a loss or malfunction of the Reactor Building
Cooling Water System will have on the following:
(CFR: 41.7 / 45.6)
K3.01
K3.02
K3.03
Reactor service water system
Reactor recirculation system
Heating, ventilation, and air conditioning system emergency cooling
water system
Process radiation monitoring system
Suppression pool cleanup system
Instrument air system
Control rod drive system
Recirculation flow control system
Residual heat removal system
High pressure core flooder system
Reactor water cleanup system
Fuel pool cooling and cleanup system
Sampling system
Radioactive drain transfer system
Emergency diesel generator system
Drywell cooling system
Heating, ventilation and air conditioning system
Containment atmospheric monitoring system
Service air system
Turbine building cooling water system
K3.04
K3.05
K3.06
K3.07
K3.08
K3.09
K3.10
K3.11
K3.12
K3.13
K3.14
K3.15
K3.16
K3.17
K3.18
K3.19
K3.20
IMPORTANCE
3
3
3
3
2
3
3
3
3
3
3
2
2
2
3
3
3
2
2
2
K4.
Knowledge of Reactor Building Cooling Water System design feature(s)
and or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
Automatic start of standby pump
Emergency diesel generator load sequencing
RBCW response to LOCA signal
RBCW response to high suppression pool temperature
Operation from the remote shutdown panel
RBCE response to a high RBCW supply temperature
K5.
Knowledge of the operational implications or cause-effect relationships as
they apply to the Reactor Building Cooling Water System:
(CFR: 41.5 / 45.3)
K5.01
Chemistry control
3
3
3
3
3
3
2
3.8-11
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8RBCW Reactor Building Cooling Water System (continued)
K/A NO.
KNOWLEDGE
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Building
Cooling Water System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
AC electrical power distribution system
Heating, ventilation, and air conditioning system emergency cooling
water system
Makeup water purified system
Plant information and control system
Engineered Safety Function logic and control system
Suppression pool cleanup system
Instrument air system
Process radiation monitoring system
K6.03
K6.04
K6.05
K6.06
K6.07
K6.11
IMPORTANCE
3
2
2
3
3
2
2
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Building Cooling Water System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
RBCW flow rate
RBCW temperature
RBCW pressure
Surge tank level
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Building Cooling Water System
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
RO SRO
Loss of RBCW pump
3
4
High/low surge tank level
3
4
High/low RBCW temperature
3
3
Radiation monitoring system alarm
3
3
Loss of AC electrical power distribution system
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
NUREG-2104
3
3
3
3
3.8-12
3.8
Safety Function 8: Plant Service Systems
System
SF8RBCW Reactor Building Cooling Water System (continued)
K/A NO.
KNOWLEDGE
A3
Ability to monitor automatic operations of the Reactor Building Cooling
Water System including:
(CFR: 41.7 / 45.7)
A3.01
Setpoints on instrument signal levels for normal operations,
warnings, and trips that are applicable to the RBCW
System alignment due to high suppression pool temperature
Emergency diesel generator load sequencing
System alignment due to LOCA signal
System alignment due to loss of offsite power
A3.02
A3.03
A3.04
A3.05
IMPORTANCE
3
4
3
4
3
A4
Ability to manually operate and/or monitor the Reactor Building Cooling
Water System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
RBCW indications and controls
3.8-13
3
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8RSW Reactor Service Water System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Service Water System and the following systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
Remote shutdown system
Engineered Safety Function logic and control system
Sampling system
Radioactive drain transfer system
Reactor building cooling water system
Instrument air system
AC electrical power distribution system
Chemical storage and transfer system
Turbine service water system
Makeup water preparation system
Plant information and control system
Circulating water system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Reactor service water pumps
K3
Knowledge of the effect that a loss or malfunction of the Reactor Service
Water System will have on the following:
(CFR: 41.7 / 45.6)
K3.01
Reactor building cooling water system
K4.
Knowledge of Reactor Service Water System design feature(s) and or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
Automatic start of standby pump
3
Emergency diesel generator load sequencing
3
Response to loss of coolant accident signal
3
Response to a loss of offsite power signal
3
Response to a reactor building cooling/reactor building service water heat
exchanger room high water level signal
3
Response to high suppression pool temperature
3
Operation from the remote shutdown panel
3
K4.06
K4.07
NUREG-2104
IMPORTANCE
3.8-14
3
3
2
2
4
2
3
2
2
2
3
2
3
4
3.8
Safety Function 8: Plant Service Systems
System
SF8RSW Reactor Service Water System (continued)
K/A NO.
KNOWLEDGE
K5.
Knowledge of the operational implications or cause-effect relationships as
they apply to the Reactor Service Water System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
Chemistry control
Cold weather operations
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Service
Water System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
Remote shutdown system
Engineered Safety Function logic and control system
Sampling system
Radioactive drain transfer system
Instrument air system
AC electrical power distribution system
Chemical storage and transfer system
Turbine service water system
Makeup water preparation system
Plant information and control system
Circulating water system
IMPORTANCE
2
2
3
3
2
2
2
3
2
2
2
3
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Service Water System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
RSW flow rate
RSW temperature
RSW header supply pressure
Ultimate heat sink basin level
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Service Water System and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
Loss of RSW pump
Low ultimate heat sink water level
3
3
3
3
3.8-15
RO SRO
3
4
3
4
NUREG 2104
3.8
Safety Function 8: Plant Service Systems
System
SF8RSW Reactor Service Water System
K/A NO.
ABILITY
A2.03
A2.04
A2.05
A2.06
Loss of AC electrical power distribution system
Loss of coolant accident
Loss of offsite power
High water level in the RBCW/RSW heat exchanger room
A3
Ability to monitor automatic operations of the Reactor Service Water
System Including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
System alignment due to high suppression pool temperature
System alignment due to LOCA signal
System alignment due to loss of offsite power
Emergency diesel generator load sequencing
Setpoints on instrument signal levels for normal operations,
warnings, and trips that are applicable to the RSW
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 /45.5 to 45.8)
A4.01
RSW indications and controls
NUREG-2104
IMPORTANCE
3.8-16
RO SRO
3
3
4
4
3
4
3
3
4
4
3
3
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9OG Offgas System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Offgas System and the following systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
Feedwater and condensate air extraction system
Process radiation monitoring system
Radioactive waste treatment system
Instrument air system
Turbine building cooling water system
AC electrical distribution system
Sampling system
Turbine building heating, ventilation, and air conditioning system
Heating steam and condensate water return system
Heating, ventilation, and air conditioning system normal cooling
water system
Makeup water purified system
Plant information and control system
Hydrogen water chemistry system
K1.11
K1.12
K1.13
IMPORTANCE
3
3
2
3
3
2
3
2
2
2
2
3
2
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Offgas system vacuum pump
K3
Knowledge of the effect that a loss or malfunction of the Offgas System will
have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
Condenser vacuum
Off-site radioactivity release
Hydrogen concentration
K4
Knowledge of Offgas System design feature(s) and/or interlocks which
provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
Dilution of hydrogen gas concentration
Prevention of water entering the recombiner catalyst
Maintenance of sufficient oxygen gas inventory to allow for complete
hydrogen recombination
The prevention of hydrogen explosions and/or fires
Redundancy
K4.04
K4.05
2
4
3
3
3.9-1
3
3
2
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9OG Offgas System (continued)
K/A NO.
KNOWLEDGE
K4.06
K4.07
K4.08
K4.09
Decay of fission product gases to particulate daughters
Maximizing charcoal bed efficiency
Automatic system isolation
Filtration of radioactive particulate
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Offgas System:
(CFR: 41.7 / 45.4)
K5.01
K5.02
K5.03
K5.04
K5.05
K5.06
K5.07
K5.08
K5.09
K5.10
Heat removal mechanisms
Heat addition mechanisms
Hydrogen concentration measurement
Oxygen concentration measurement
Catalytic recombination
Radioactive decay
Charcoal adsorption of fission product gases
Hydrogen and oxygen recombination
Decontamination factors
Explain the necessity of reducing relative humidity for carbon
bed filters.
Elevated release point
Condenser vacuum
Off-site radioactivity release
K5.11
K5.12
K5.13
IMPORTANCE
3
3
3
3
2
2
3
3
3
3
3
3
2
3
3
3
3
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Offgas System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
Instrument air system
Process radiation monitoring system
Turbine building cooling water system
Hydrogen water chemistry system
AC electrical distribution system
Feedwater and condensate air extraction system
Fuel cladding integrity
Condenser vacuum
Heating steam and condensate water return system
Makeup water purified system
Turbine building heating, ventilation, and air conditioning system
Plant information and control system
NUREG-2104
3.9-2
3
3
2
3
3
3
3
3
2
2
2
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9OG Offgas System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Offgas System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
A1.11
A1.12
A1.13
A1.14
A1.15
Condenser vacuum
Station radioactive release rate
Preheater discharge temperature
Recombiner catalyst temperature
Cooler condenser discharge temperature
Filter differential pressure
Charcoal bed humidity
System flow
Charcoal bed temperature
Charcoal vault temperature
Offgas condenser temperatures
Process radiation monitoring indications
Hydrogen gas concentration
Oxygen gas concentration
Steam supply pressures
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Offgas System; and (b) based on those
predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5/45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
Low condenser vacuum
Low dilution steam flow
Offgas system high radiation
High charcoal bed humidity
Offgas system holdup volume explosion/ fire
Low oxygen injection flow
AC electrical distribution system failures
Offgas system high flow
Offgas system low flow
Recombiner high temperature
Recombiner low temperature
Offgas filter high differential pressure
Air intrusion
Loss of offgas system loop seals
Reactor power changes
IMPORTANCE
3.9-3
3
3
2
2
2
2
2
3
2
2
2
3
3
3
3
RO SRO
3
3
3
3
4
4
3
3
4
4
3
3
3
3
3
3
3
3
3
3
2
3
3
3
3
3
3
3
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9OG Offgas System (continued)
K/A NO.
ABILITY
A3
Ability to monitor automatic operations of the Offgas System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
System isolations
System flows
System temperatures
Station radioactive release rate
System indications and alarms
System differential pressures
Process radiation monitoring system indications
Startup and shutdown of the offgas system
A4
Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Reset system isolations
Offgas system controls/components
NUREG-2104
IMPORTANCE
3.9-4
3
3
3
3
3
3
3
3
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9HVAC Plant Ventilation Systems
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Plant Ventilation Systems and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
AC electrical distribution system
Secondary containment
Standby gas treatment system
Atmosphere control system
Process radiation monitoring system
Instrument air system
Heating, ventilation, and air conditioning system normal cooling
water system
Heating, ventilation, and air conditioning system emergency cooling
water system
Reactor building cooling water system
Turbine building cooling water system
Engineered Safety Function logic and control system
Leak detection and isolation system
Radioactive drain transfer system
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
IMPORTANCE
3
3
4
3
3
3
2
3
3
2
3
3
2
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
Reactor building supply and exhaust fans
Turbine building supply and exhaust fans
Radwaste building supply and exhaust fans
K3
Knowledge of the effect that a loss or malfunction of the Plant Ventilation
Systems will have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
Secondary containment temperature
Reactor building temperature
Reactor building pressure
Secondary containment differential pressure
Turbine building temperature
Turbine building differential pressure
3.9-5
2
2
2
3
3
3
3
2
2
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9HVAC Plant Ventilation Systems (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Plant Ventilation Systems design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Automatic initiation of standby gas treatment system
Secondary containment isolation
Automatic starting and stopping of fans
Smoke removal
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to PLANT VENTILATION Systems:
(CFR: 41.7 / 45.4)
K5.01
Airborne contamination control
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Plant Ventilation
Systems:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
AC electrical distribution system
Instrument air system
Heating, ventilation, and air conditioning normal cooling water
system
Heating, ventilation, and air conditioning emergency cooling water
system
Reactor building cooling water system
Turbine building cooling water system
Engineered Safety Function logic and control system
Leak detection and isolation system
K6.04
K6.05
K6.06
K6.07
K6.08
IMPORTANCE
4
4
3
3
3
3
3
3
3
3
2
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Plant Ventilation Systems controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
Filter differential pressure
Fan differential pressure
Area temperatures
Secondary containment differential pressure
NUREG-2104
3.9-6
2
2
2
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9HVAC Plant Ventilation Systems (continued)
K/A NO.
KNOWLEDGE
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Plant Ventilation Systems; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
High drywell pressure
Low reactor water level
Loss of coolant accident
High radiation
Secondary containment differential pressure
A3
Ability to monitor automatic operations of the Plant Ventilation Systems
including:
(CFR: 41.7 / 45.7)
A3.01
Isolation/initiation signals
A4
Ability to manually operate and/or monitor the Plant Ventilation Systems in
the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Start and stop fans
Open and close dampers
IMPORTANCE
RO SRO
3
3
3
4
4
4
4
4
3
4
4
3
2
3.9-7
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9RAD Radiation Monitoring System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Radiation Monitoring System and the following systems:
(CFR: 41.2 to 41.9/45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
Main steam system
Offgas system
Plant stack
Reactor building cooling water system
Radioactive waste treatment system
Reactor building heating, ventilation, and air conditioning system
Fuel building heating, ventilation, and air conditioning system
Service building heating, ventilation, and air conditioning system
Leak detection and isolation system
Turbine building heating, ventilation, and air conditioning system
Radwaste building heating, ventilation, and air conditioning system
Plant information and control system
Emergency response information system
Vital AC power supply system
Instrument air system
Engineered Safety Function logic and control system
Standby gas treatment system
Makeup water purified system
Control building heating, ventilation, and air conditioning system
K2
Knowledge of bus or electrical power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
K2.06
K2.07
Main steamline radiation monitors
Offgas radiation monitoring system
Plant stack radiation monitoring
Radwaste liquid radiation monitoring system
Reactor building heating, ventilation, and air conditioning monitors
Area radiation monitors
Control room ventilation monitors
K3
Knowledge of the effect that a loss or malfunction of the Radiation
Monitoring System will have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
K3.04
Station liquid effluent release monitoring
Station gaseous effluent release monitoring
Station area radiation monitoring
Offgas system
NUREG-2104
IMPORTANCE
3.9-8
3
3
3
3
3
3
3
2
4
3
3
3
2
2
2
3
3
2
3
3
3
3
2
3
2
2
3
3
3
4
3.9
Safety Function 9: Radioactivity Release
System:
SF9RAD Radiation Monitoring System (continued)
K/A NO.
KNOWLEDGE
K3.05
K3.06
K3.07
K3.08
Reactor building heating, ventilation, and air conditioning
Drywell sump liquid discharge
Radwaste building ventilation
Control building heating, ventilation, and air conditioning system
K4
Knowledge of Radiation Monitoring System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
Redundancy
Automatic actions to contain the radioactive release
in the event that the predetermined release rates are exceeded
Fail safe tripping of process radiation monitoring
logic during conditions of instrument failure
K4.03
IMPORTANCE
3
2
3
3
3
4
4
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Radiation Monitoring System:
(CFR: 41.7 / 45.4)
K5.01
K5.02
K5.03
K5.04
K5.05
Hydrogen injection operation's effect on process radiation indications
Drywell sump liquid discharge
Turbine gland seal condenser exhaust
Fuel handling area ventilation exhaust
Drywell
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Radiation
Monitoring System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
Vital AC power
Plant information and control system
Leak detection and isolation system
3
2
2
3
3
3
2
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Radiation Monitoring System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
Alarms and indications associated with normal operations
Alarms and indications associated with surveillance testing
3.9-9
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9RAD Radiation Monitoring System (continued)
K/A NO.
ABILITY
A2
Ability to (d) predict the impacts of the following on the Radiation
Monitoring System; and (b) based on those predictions, use procedures to
correct, control, or mitigate the consequences of those abnormal
conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
Fuel element failure
AC electrical failure
Downscale trips
Offgas system failure
Low fuel pool level
Loss of coolant accident
Leakage and/or breaks from contaminated systems to atmosphere
or to other process systems
Refuel floor handling accidents/operations
Low reactor water level during refueling operations
A2.08
A2.09
IMPORTANCE
RO SRO
4
4
3
3
3
3
3
3
3
3
4
4
3
3
3
4
4
4
A3
Ability to monitor automatic operations of the Radiation Monitoring System
including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
A3.07
A3.08
A3.09
Offgas system isolation indications
Liquid radwaste isolation indications
Radwaste handing interlocks
Drywell LCW or HCW sump isolation indications
Ventilation system isolation indications
Display indications
Meter indications
Containment isolation indications
Lights and alarms
A4
Ability to manually operate and/or monitor Radiation Monitoring System in
the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4. 01
Manually trip process radiation monitor logic
NUREG-2104
3.9-10
4
3
2
3
3
3
3
4
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9RW Radwaste System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Radwaste System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
Makeup water condensate system
Instrument air system
Service air system
AC electrical power distribution system
Instrument and control power supply system
Reactor water cleanup system
Fuel pool cooling and cleanup system
Residual heat removal system
Turbine building service water system
Process radiation monitoring system
Condensate storage and transfer system
Makeup water purified system
Radioactive drain transfer system
Solid waste system
Make-up water condensate system
Reactor building cooling water system
Radwaste building heating, ventilation, and air conditioning system
Mobile liquid radwaste processing system
Mobile solid radwaste processing system
Low conductivity waste system
High conductivity waste system
Hot shower and storm drain system
Spent sludge system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
2
2
2
3
2
3
3
2
2
3
2
2
2
2
2
2
2
2
2
3
3
2
2
None
K3
Knowledge of the effect that a loss or malfunction of the Radwaste System
will have on the following:
(CFR: 41.5 / 45.3)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
RWCU system
Makeup water condensate system
Low and high conductivity waste sumps
Fuel pool cooling and cleanup system
Residual heat removal system
Radioactive drain transfer system
3.9-11
2
2
3
2
2
2
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9RW Radwaste System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of the Radwaste System design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
IMPORTANCE
None
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to the Radwaste System:
(CFR: 41.5 / 45.3)
None
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Radwaste
System:
(CFR: 41.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
Instrument air system
Service air system
AC electrical power distribution system
Instrument and control power supply system
Radwaste building heating, ventilation, and air conditioning
Process radiation monitoring system
Makeup water purified system
2
2
3
2
2
3
2
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Radwaste System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
Radiation level
Off-site radioactivity release
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Radwaste System; and (b) based on
those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
NUREG-2104
3
3
RO SRO
3
3
2
3
System rupture
High turbidity water
3.9-12
3.9
Safety Function 9: Radioactivity Release
System:
SF9RW Radwaste System (continued)
K/A NO.
KNOWLEDGE
A3
Ability to monitor automatic operations of the Radwaste System including:
(CFR: 41.7 / 45.7)
IMPORTANCE
None
A4
Ability to manually operate and/or monitor the Radwaste System in the
control room:
(CFR: 41.7 / 45.5 to 45.8)
None
3.9-13
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9RPV Reactor Vessel Internals
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Reactor Vessel Internals and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Main steam system
Reactor recirculation system
Feedwater system
Residual heat removal system
High pressure core flooder system
Reactor core isolation cooling system
Control rod drive system
Fine motion control rod drive mechanism
Standby liquid control system
Reactor water cleanup system
Automatic depressurization system
Loose parts monitoring system
Automated Traversing In-core Probe system
Neutron monitoring system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
IMPORTANCE
3
3
3
3
3
3
3
3
3
3
3
2
3
3
None
K3
Knowledge of the effect that a loss or malfunction of the Reactor Vessel
Internals will have on the following:
(CFR: 41.7 / 45.4)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
Reactor water level
Reactor pressure
Reactor power
Plant radiation levels
Off-site radioactive release
Leak detection and isolation system
Nuclear Boiler Instrumentation
K4
Knowledge of Reactor Vessel Internals design feature(s) and/or interlocks
which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Separation of fluid flow paths within the vessel
Core orificing
Moisture removal from generated steam
Natural circulation
NUREG-2104
3.9-14
3
3
3
3
3
3
3
3
3
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9RPV Reactor Vessel Internals (continued)
K/A NO.
KNOWLEDGE
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Reactor Vessel Internals:
(CFR: 41.5 / 45.3)
IMPORTANCE
None
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Reactor Vessel
Internals:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
K6.14
K6.15
K6.16
Control rod drive system
Fine motion control rod drive mechanism
Reactor recirculation system
Feedwater system
Standby Liquid Control system
Safety/relief valves
Reactor water cleanup system
Nuclear Boiler Instrumentation
High pressure core flooder system
Residual heat removal system
Reactor core isolation cooling system
Automatic depressurization system
Loose parts monitoring
Automated traversing in-core probe system
Neutron monitoring system
Main steam system
3
3
3
3
3
3
3
3
3
3
3
3
2
3
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Reactor Vessel Internals controls including:
(CFR: 41.5 / 45.5)
None
3.9-15
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9RPV Reactor Vessel Internals (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Reactor Vessel Internals; and (b) based
on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
LOCA
Overpressurization transient
Control rod drop accident
Excessive heatup/cooldown rate
Exceeding thermal limits
Exceeding safety limits
A3
Ability to monitor automatic operations of the Reactor Vessel Internals
including:
(CFR: 41.7 / 45.7)
IMPORTANCE
RO SRO
4
4
4
4
4
4
4
4
4
4
4
4
None
A4
Ability to manually operate and/or monitor the Reactor Vessel Internals in
the control room:
(CFR: 41.7 / 45.5 to 45.8)
None
NUREG-2104
3.9-16
3.9
Safety Function 9: Radioactivity Release
System:
SF9FPC Fuel Pool Cooling and Cleanup System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Fuel Pool Cooling and Cleanup System and the following
systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
Residual heat removal system
Makeup water condensate system
Sampling system
Instrument air system
AC electrical power distribution system
Reactor building cooling water system
Low conductivity waste system
Radwaste drain transfer system
Suppression pool cleanup system
Reactor water cleanup system
Direct current power supply system
Reactor pressure vessel instrumentation system
Plant information and control system
Spent sludge system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Fuel pool cooling and cleanup pumps
K3
Knowledge of the effect that a loss or malfunction of the Fuel Pool Cooling
and Cleanup System will have on the following:
(CFR: 41.7 /45.6)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
K3.07
K3.08
Fuel pool temperature
Fuel pool water level
Fuel pool water clarity
Fuel pool water chemistry
Fuel pool water fission product concentration
Area radiation levels
Suppression pool chemistry
Refueling operations
IMPORTANCE
3.9-17
3
2
2
2
2
3
2
3
2
2
2
2
3
2
2
3
3
3
2
3
3
2
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9FPC Fuel Pool Cooling and Cleanup System (continued)
K/A NO.
KNOWLEDGE
K4
Knowledge of Fuel Pool Cooling and Cleanup System design feature(s)
and/or interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
K4.05
K4.06
K4.07
K4.08
K4.09
Redundancy
Pool clarity
Maintenance of adequate pool temperature
Overpressure protection for fuel pool cooling system filter
Net positive suction head requirements for fuel pool cooling pumps
Maintenance of adequate pool level
Supplemental heat removal capability
Pool cooling during loss of coolant accident
Maintenance of filter/demineralizer precoat during low flow conditions
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Fuel Pool Cooling and Cleanup System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
Spent fuel decay heat generation
Mechanical filtration operation
Maximum normal heat load
Maximum (abnormal) heat load
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Fuel Pool
Cooling and Cleanup System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
K6.11
K6.12
K6.13
AC electrical power distribution system
Residual heat removal system
Makeup water condensate system
Reactor building cooling water system
Instrument air system
Radwaste drain transfer system
Reactor well seal failure
Direct current power supply system
Reactor pressure vessel instrumentation system
Plant information and control system
Reactor building heating, ventilation, and air conditioning system
Low conductivity waste system
Suppression pool cleanup system
NUREG-2104
IMPORTANCE
3.9-18
2
2
3
2
2
3
3
3
2
3
2
3
3
3
2
3
3
2
2
3
2
2
3
2
2
2
3.9
Safety Function 9: Radioactivity Release
System:
SF9FPC Fuel Pool Cooling and Cleanup System (continued)
K/A NO.
ABILITY
Al.
Ability to predict and/or monitor changes in parameters associated with
operating the Fuel Pool Cooling and Cleanup System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.06
A1.07
A1.08
A1.09
A1.10
Surge tank level
Pool level
Pool temperature
Pump discharge pressure
Filter/demineralizer differential pressure
System flow
System temperature
Pool chemistry
Pool clarity
Pool activity levels
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Fuel Pool Cooling and Cleanup System;
and (b) based on those predictions, use procedures to correct, control, or
mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
IMPORTANCE
3
3
3
2
2
3
3
2
2
2
RO SRO
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
2
2
2
2
2
2
2
2
3
3
3
3
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A2.12
A2.13
A2.14
A2.15
High pool level
Low pool level
Low surge tank level/high level
Pump trip
High fuel pool temperature
Reactor building cooling water failure
AC electrical power distribution system failures
Refueling bellows seal high flow
High spent fuel pool gate drain flow
High filter/demineralizer differential pressure
Low filter/demineralizer differential pressure
Low system flow
Low pump suction pressure
Low pool clarity
Inadequate system/pool chemistry
A3
Ability to monitor automatic operations of the Fuel Pool Cooling and
Cleanup System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
Valve operation
Pump trip(s)
System indications and alarms
3.9-19
2
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9FPC Fuel Pool Cooling and Cleanup System (continued)
K/A NO.
ABILITY
A4
Ability to manually operate and/or monitor the Fuel Pool Cooling and
Cleanup System in the control room:
CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Fuel pool cooling and cleanup system pumps
Fuel pool cooling and cleanup system valves
NUREG-2104
IMPORTANCE
3.9-20
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9SGTS Standby Gas Treatment System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Standby Gas Treatment System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
Reactor building heating, ventilation, and air conditioning system
Plant ventilation stack
Process radiation monitoring system
Atmosphere control system
Leak detection and isolation system
Fire protection system
Makeup water purified system
Radioactive drain transfer system
Plumbing and drainage system
Engineered Safety Function logic and control system
AC electrical power distribution system
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
K2.02
K2.03
K2.04
K2.05
Standby gas treatment process and cooling fans
Standby gas treatment motor operated inlet/outlet flow damper
Standby gas treatment initiation logic
Standby gas treatment filter train heaters
Standby gas treatment inlet and outlet space heaters
K3
Knowledge of the effect that a loss or malfunction of the Standby Gas
Treatment System will have on the following:
(CFR: 41.7 /45.6)
K3.01
K3.02
K3.03
K3.04
K3.05
K3.06
Secondary containment differential pressure
Off-site release rate
Primary containment pressure
Secondary containment pressure
Secondary containment radiation/contamination levels
Atmosphere control system
K4
Knowledge of Standby Gas Treatment System design feature(s) and/or
interlocks which provide for the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
K4.04
Automatic system initiation
Charcoal bed decay heat removal
Moisture removal
Radioactive particulate filtration
IMPORTANCE
3.9-21
3
3
3
3
3
2
2
2
2
3
3
2
2
2
2
2
3
4
3
3
3
3
4
3
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9SGTS Standby Gas Treatment System (continued)
K/A NO.
KNOWLEDGE
K4.05
K4.06
K4.07
Fission product iodine removal
Charcoal bed retention
Control charcoal bed relative humidity
K5
Knowledge of the operational implications or cause and effect
relationships as they apply to Standby Gas Treatment System:
(CFR: 41.5 / 45.3)
K5.01
K5.02
K5.03
K5.04
K5.05
Heat removal mechanisms
Primary containment
Secondary containment
Wetwell
Primary containment pressure
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Standby Gas
Treatment System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
K6.05
K6.06
K6.07
K6.08
K6.09
K6.10
AC electrical power distribution system
Process radiation monitoring
Drywell high pressure
Leak detection and isolation system
Fire protection system
Engineered Safety Function logic and control system
Radioactive drain transfer system
Makeup water purified system
Plant ventilation stack
Reactor building heating, ventilation, and air conditioning system
IMPORTANCE
3
2
2
2
3
3
3
3
3
3
3
3
2
3
2
2
3
3
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Standby Gas Treatment System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
System flow
Drywell/wetwell pressure
Off-site radioactive release
Secondary containment differential pressure
Standby gas treatment system filter train temperature
NUREG-2104
3.9-22
3
3
3
3
3
3.9
Safety Function 9: Radioactivity Release
System:
SF9SGTS Standby Gas Treatment System (continued)
K/A NO.
ABILITY
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Standby Gas Treatment System; and (b)
based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
A2.05
A2.06
A2.07
A2.08
A2.09
A2.10
A2.11
A1.12
A2.13
A2.14
Low system flow
High system flow
High filter train temperature
High filter train moisture content
Standby gas treatment process fan trips
Motor operated inlet/outlet flow damper closures
AC electrical power distribution system failure
Low reactor water level
High drywell/wetwell pressure
High refuel floor ventilation exhaust radiation
High secondary containment ventilation exhaust radiation
Refueling floor ventilation exhaust radiation
High charcoal bed temperature
Filter train heater trips
A3
Ability to monitor automatic operations of the Standby Gas Treatment
System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
A3.03
A3.04
A3.05
A3.06
System flow
Standby gas treatment system process fan start
Standby gas treatment system motor operated damper operation
System temperature/humidity
Secondary containment differential pressure
Standby gas treatment cooling fans
A4
Ability to manually operate and/or monitor the Standby Gas Treatment
System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4. 01
A4. 02
Standby gas treatment system fans
Standby gas treatment system motor operated dampers
IMPORTANCE
3.9-23
RO SRO
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
3
4
3
4
2
2
2
2
3
3
3
3
3
3
3
3
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9CRHVAC Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System
K/A NO.
KNOWLEDGE
K1
Knowledge of the physical or control/protection logic relationships
between the Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01
K1.02
Process radiation monitoring system
Heating, ventilation, and air conditioning emergency cooling
water system
Fire protection system
Makeup water purified system
Engineered Safety Function logic and control system
Non-radioactive drain transfer system
AC electrical power distribution system
Vital AC power supply system
Direct current power supply system
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
IMPORTANCE
K2
Knowledge of bus or division power supplies to the following:
(CFR: 41.7)
K2.01
Control room habitability area heating, ventilation, and air
conditioning supply fans
Control room habitability area heating, ventilation, and air
conditioning exhaust fans
Control room habitability area heating, ventilation, and air
conditioning emergency filtration unit supply
fans
Control room habitability area heating, ventilation, and air
conditioning motor operated normal supply
dampers
Control room habitability area heating, ventilation, and air
conditioning emergency filtration unit air
supply dampers
K2.02
K2.03
K2.04
K2.05
3
3
3
2
3
2
3
3
2
2
2
2
2
2
K3
Knowledge of the effect that a loss or malfunction of the Control Room
Habitability Area Heating, Ventilation, and Air Conditioning System will
have on the following:
(CFR: 41.7 /45.6)
K3.01
K3.02
K3.03
Control room habitability
Control room temperature
Control room pressure
NUREG-2104
4
3
3
3.9-24
3.9
Safety Function 9: Radioactivity Release
System:
SF9CRHVAC Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System (continued)
K/A NO.
KNOWLEDGE
K3.04
Control room habitability area heating, ventilation, and air
conditioning humidity
Control room habitability area heating, ventilation, and air
conditioning radioactivity
K3.05
IMPORTANCE
2
3
K4
Knowledge of Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System design feature(s) and/or interlocks which provide for
the following:
(CFR: 41.7)
K4.01
K4.02
K4.03
System initiations/reconfiguration
Control room temperature
Smoke removal mode
K5
Knowledge of the operational implications or cause-effect relationships as
they apply to Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System:
(CFR: 41.5 / 45.3)
K5.01
Airborne contamination (e.g., radiological, toxic gas, smoke) control
K6
Knowledge of the effect of the following plant conditions, system
malfunctions or component malfunctions will have on the Control Room
Habitability Area Heating, Ventilation, and Air Conditioning System:
(CFR: 41.7 / 45.7)
K6.01
K6.02
K6.03
K6.04
AC electrical power distribution system
Vital AC power supply system
Direct current power supply system
Heating, ventilation, and air conditioning emergency cooling
water system
Fire protection system
Process radiation monitoring system
Makeup water purified system
Engineered Safety Function logic and control system
Non-radioactive drain transfer system
K6.05
K6.06
K6.07
K6.08
K6.09
3.9-25
3
2
3
3
3
2
2
3
3
3
2
3
2
NUREG 2104
3.9
Safety Function 9: Radioactivity Release
System:
SF9CRHVAC Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System (continued)
K/A NO.
ABILITY
A1
Ability to predict and/or monitor changes in parameters associated with
operating the Control Room Habitability Area Heating, Ventilation, and Air
Conditioning System controls including:
(CFR: 41.5 / 45.5)
A1.01
A1.02
A1.03
A1.04
A1.05
A1.05
Filter differential pressure
Fan differential pressure
Area temperatures
Control room pressure
Control room humidity
Airborne radioactivity levels
A2
Ability to (a) predict the impacts of the following system/component
malfunctions or operations on the Control Room Habitability Area Heating,
Ventilation, and Air Conditioning System; and (b) based on those
predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.01
A2.02
A2.03
A2.04
Initiation/reconfiguration
Extreme environmental conditions
Initiation/reconfiguration failure
Initiation/failure of fire protection system
A3
Ability to monitor automatic operations of the Control Room Habitability
Area Heating, Ventilation, and Air Conditioning System including:
(CFR: 41.7 / 45.7)
A3.01
A3.02
Initiation/reconfiguration
Initiation/failure of fire protection system
A4
Ability to manually operate and/or monitor the Control Room Habitability
Area Heating, Ventilation, and Air Conditioning System in the control room:
(CFR: 41.7 / 45.5 to 45.8)
A4.01
A4.02
Initiate/reset system
Control room habitability area heating, ventilation, and air
conditioning supply and exhaust fans
Control room habitability area heating, ventilation, and air
conditioning emergency filtration unit supply
fans
Reposition dampers
A4.03
A4.04
NUREG-2104
IMPORTANCE
3.9-26
2
2
3
3
2
3
RO SRO
3
3
3
3
3
4
3
3
3
3
3
3
3
3
4.0
EMERGENCY PLANT AND ABNORMAL PLANT EVOLUTIONS
4.1
Emergency Plant Evolutions
EPE:
EPE1001 High Drywell Pressure
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Drywell Pressure:
(CFR: 41.8 to 41.10)
EK1.01
Primary containment integrity
EK2.
Knowledge of the interrelations between High Drywell Pressure and the
following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
EK2.10
EK2.11
EK2.12
EK2.13
EK2.14
EK2.16
High pressure core flooder system
Reactor core isolation cooling
Residual heat removal system/Low pressure flooder mode
Reactor trip and isolation system
Emergency diesel generators
Leak detection and isolation system
Automatic depressurization system
Drywell spray
Suppression pool cooling
Wetwell spray
Safety parameter display system
Drywell cooling system
Condensate system
Containment overpressure protection system
Atmosphere control system
EK3.
Knowledge of the reasons for the following responses as they apply to
High Drywell Pressure:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
Drywell spray operation
Wetwell spray operation
Containment overpressure protection operation
Emergency depressurization
Reactor scram
Drywell venting
IMPORTANCE
4
4.1-1
4
4
4
4
4
4
4
4
4
4
3
3
4
3
3
4
4
4
4
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1001 High Drywell Pressure (continued)
K/A NO.
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Drywell Pressure:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
EA1.04
EA1.05
EA1.06
EA1.07
EA1.08
EA1.09
EA1.10
EA1.11
EA1.12
EA1.13
EA1.14
EA1.15
High pressure core flooder system
Reactor core isolation cooling system
Residual heat removal system/Low pressure flooder mode
Reactor trip and isolation system
Emergency diesel generators
Leak detection and isolation system
Automatic depressurization system
Drywell spray
Wetwell spray
Suppression pool cooling
Drywell cooling system
Containment atmospheric monitoring
Wetwell to drywell vacuum relief breakers
Atmosphere control system
Containment overpressure protection system
EA2.
Ability to determine and/or interpret the following as they apply to High
Drywell Pressure:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
EA2.05
EA2.04
EA2.05
Drywell pressure
Drywell temperature
Suppression pool level
Wetwell pressure
Wetwell air-space temperature
Suppression pool temperature
Primary containment radiation levels
NUREG-2104
IMPORTANCE
4.1-2
4
4
4
4
4
4
4
4
4
4
3
4
3
3
3
RO SRO
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1002 High Reactor Pressure
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Reactor Pressure:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
EK1.04
EK1.05
EK1.06
Pressure effects on reactor power
Reactor pressure vessel integrity
Safety/relief valve tailpipe temperature/pressure relationships
Decay heat generation
Exceeding safety limits
Pressure effects on reactor pressure vessel water level
EK2.
Knowledge of the interrelations between High Reactor Pressure and the
following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
Reactor trip and isolation system
Alternate reactor shutdown system
Fine motion control rod motor drive insertion function (FM Control
Rod Drive run-in)
Recirculation system
Safety/relief valves
Reactor core isolation cooling system
Steam bypass and pressure control system
Reactor power
Safety parameter display system
Reactor pressure vessel water level
ESF logic and control system
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
EK2.10
EK2.11
IMPORTANCE
4
4
4
4
4
4
4
4
4
4
4
4
4
4
3
4
3
EK3.
Knowledge of the reasons for the following responses as they apply to
High Reactor Pressure:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
EK3.07
Safety/relief valve opening
4
Reactor internal pump trip
4
Alternate rod insertion
4
Standby liquid control system initiation signal
3
Steam bypass and pressure control system
4
Feedwater control system operation (feedwater pump speed runback) 4
Fine motion control rod motor driven insertion function (FM Control
Rod Drive run-in)
4
Automatic reactor scram
4
EK3.08
4.1-3
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1002 High Reactor Pressure (continued)
K/A NO.
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Reactor Pressure:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
EA1.04
EA1.05
EA1.06
EA1.07
EA1.08
EA1.09
Main steam line drains
Steam bypass and pressure control system
Safety/relief valves
Reactor core isolation cooling system
Alternate rod insertion
Standby liquid control system
Reactor trip and isolation system
Feedwater control system (feedwater pump speed runback)
Fine motion control rod motor driven insertion function (FM Control
Rod Drive run-in)
Reactor internal pump trip
Reactor water cleanup system
EA1.10
EA1.11
EA2.
EA2.01
EA2.02
EA2.03
EA2.04
EA2.05
NUREG-2104
IMPORTANCE
3
4
4
4
4
3
4
4
4
4
3
Ability to determine and/or interpret the following as they apply to High
Reactor Pressure:
(CFR: 41.10 / 43.5 / 45.13)
Reactor pressure vessel pressure
Reactor power
Suppression pool temperature
Suppression pool level
Reactor pressure vessel water level
4.1-4
RO SRO
4
4
4
4
4
4
4
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1003 Suppression Pool High Water Temperature
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Suppression Pool High Water Temperature:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
Pump net positive suction head
Steam condensation
Primary containment integrity
EK2.
Knowledge of the interrelations between Suppression Pool High Water
Temperature and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
Suppression pool cooling
Wetwell spray
Wetwell pressure
Safety parameter display system
Suppression pool level
Suppression pool temperature monitoring system
EK3.
Knowledge of the reasons for the following responses as they apply to
Suppression Pool High Water Temperature:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
EK3.07
Emergency/normal reactor pressure vessel depressurization
Suppression pool cooling
Wetwell spray
Standby liquid control system injection
Automatic reactor scram
Reactor building cooling water system automatic initiation
Reactor service water system automatic initiation
IMPORTANCE
3
4
3
4
4
3
3
4
4
4
4
4
4
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to
Suppression Pool High Water Temperature:
(CFR: 41.7 / 45.6)
EA1.01
EA1.03
Suppression pool cooling
Suppression pool temperature monitoring
EA2.
Ability to determine and/or interpret the following as they apply to
Suppression Pool High Water Temperature:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
Suppression pool water temperature.
Suppression pool level
Reactor pressure vessel pressure
4.1-5
4
4
RO SRO
4
4
4
4
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1004 High Drywell Temperature
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Drywell Temperature:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
Reactor pressure vessel water level measurement
Equipment environmental qualification
Primary containment integrity
EK2.
Knowledge of the interrelations between High Drywell Temperature and the
following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
Drywell spray
Components internal to the drywell
Reactor pressure vessel water level indication
Drywell cooling system
Safety Parameter Display System
EK3.
Knowledge of the reasons for the following responses as they apply to
High Drywell Temperature:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
Emergency depressurization.
Reactor pressure vessel flooding
Drywell spray operation
Increased drywell cooling
Reactor scram
Lower drywell flooder actuation
IMPORTANCE
4
3
3
4
3
4
4
2
4
4
4
4
4
3
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Drywell Temperature:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
Drywell spray
Drywell cooling system
Drywell pressure
EA2.
Ability to determine and/or interpret the following as they apply to High
Drywell Temperature:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
EA2.05
EA2.06
NUREG-2104
4
4
4
RO
Drywell temperature
Reactor pressure vessel pressure
Reactor pressure vessel water level
Drywell pressure
Wetwell pressure...
Wetwell air space temperature
4.1-6
SRO
4
4
4
4
4
3
4
4
4
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1005 High Suppression Pool Water Level
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications as they apply to High
Suppression Pool Water Level:
(CFR: 41.8 to 41.10)
EK1.01
EK2.02
Containment integrity
Termination of injection sources external to containment
EK2.
Knowledge of the interrelations between High Suppression Pool Water
Level and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK206
Residual heat removal system
High pressure core flooder system
Wetwell to drywell vacuum breakers
Safety relief valves and respective discharge piping
Drywell water level
Reactor core isolation cooling system
EK3.
Knowledge of the reasons for the following responses as they apply to
High Suppression Pool Water Level:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
Emergency depressurization
Lowering suppression pool water level
Reactor scram
IMPORTANCE
3
3
3
3
3
3
3
3
4
4
3
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Suppression Pool Water Level:
(CFR: 41.7 / 45.6)
EK1.01
EA102
EA1.03
High pressure core flooder system
Residual heat removal system
Reactor core isolation cooling system
EA2.
Ability to determine and/or interpret the following as they apply to High
Suppression Pool Water Level:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
Suppression pool water level
Reactor pressure
Drywell water level
3
3
3
RO SRO
4
4
4
4
3
4
4.1-7
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1006 Low Suppression Pool Water Level
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Low Suppression Pool Water Level:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
Steam condensation
Pump net positive suction head
Heat capacity
EK2.
Knowledge of the interrelations between Low Suppression Pool Water
Level and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
Reactor core isolation cooling system
Residual heat removal system/low pressure flooder mode
High pressure core flooder system
Horizontal vent submergence
SRV discharge submergence
Safety Parameter Display System
Suppression pool temperature detector submergence
EK3.
Knowledge of the reasons for the following responses as they apply to Low
Suppression Pool Water Level:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
Emergency depressurization
Reactor core isolation cooling system
High pressure core flooder system
Reactor scram
net positive suction head considerations for Emergency Core
Cooling System pumps
IMPORTANCE
4
4
4
4
4
4
4
4
3
3
4
4
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Low
Suppression Pool Water Level:
(CFR: 41.7 / 45.6)
EA1.01
Emergency Core Cooling System systems (net positive suction
head considerations)
Reactor core isolation cooling system
High pressure core flooder system
Suppression pool cleanup system
Residual heat removal system
EA1.02
EA1.03
EA1.04
EA1.05
NUREG-2104
4.1-8
4
4
4
2
4
4.1
Emergency Plant Evolutions
EPE:
EPE1006 Low Suppression Pool Water Level
K/A NO.
ABILITY
EA2.
Ability to determine and/or interpret the following as they apply to Low
Suppression Pool Water Level:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
Suppression pool level
Suppression pool temperature
Reactor pressure
Drywell/ suppression chamber differential pressure
IMPORTANCE
4.1-9
RO SRO
4
4
4
4
4
4
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1007 Reactor Low Water Level
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Reactor Low Water Level:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.02
Adequate core cooling
Natural circulation
Water level effects on reactor power
EK2.
Knowledge of the interrelations between Reactor Low Water Level and the
following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
EK2.10
EK2.11
EK2.12
EK2.13
EK2.14
EK2.15
EK2.16
EK2.17
Reactor pressure vessel water level indication
Reactor pressure vessel pressure.
Reactor core isolation cooling system
Residual heat removal system: Low pressure flooder mode
High pressure core flooder system
Automatic depressurization system
Recirculation system
Reactor trip and isolation system
Leak detection and isolation system
Alternate rod insertion circuitry
AC electrical power distribution system
Feedwater control system
Residual heat removal system: shutdown cooling mode
Standby liquid control system
ESF logic and control system
Safety parameter display system
Fire protection system
EK3.
Knowledge of the reasons for the following responses as they apply to
Reactor Low Water Level:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
EK3.07
EK3.08
EK3.09
Automatic depressurization system initiation
Core coverage
Steam cooling
Emergency depressurization
Reactor scram
Reactor core isolation cooling initiation
High pressure core flooder initiation
Residual heat removal system: Low pressure flooder mode initiation
Alternate rod insertion initiation
NUREG-2104
IMPORTANCE
4.1-10
4
4
4
4
4
4
4
4
4
3
4
4
4
3
4
4
4
3
2
4
4
4
4
4
4
4
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1007 Reactor Low Water Level (continued)
K/A NO.
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Reactor
Low Water Level:
(CFR: 41.7 / 45.6)
EA1.01
EK2.02
EA1.03
EA1.04
EA1.05
EA1.06
EA1.07
EA1.08
EA1.09
EA1.10
EA1.11
Residual heat removal system: Low pressure flooder mode
High pressure core flooder systems
Reactor core isolation system
Automatic depressurization system
Safety/relief valves
Alternate injection subsystems
Control rod drive system
Condensate system
Feedwater system
Feedwater control system
Fire protection system
EA2.
Ability to determine and/or interpret the following as they apply to Reactor
Low Water Level:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
Reactor pressure vessel water level
Reactor power
Reactor pressure vessel pressure
Adequate core cooling
RO SRO
4
4
4
4
4
4
4
4
4.1-11
NUREG-2104
IMPORTANCE
4
4
4
4
4
4
4
4
4
4
$
4.1
Emergency Plant Evolutions
EPE:
EPE1008 High Secondary Containment Area Temperature
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Secondary Containment Area Temperature:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
EK1.04
EK1.05
Personnel protection
Radiation releases
Impact of operating environment on components
Maximum normal operating temperature
Maximum safe operating temperature
EK2.
Knowledge of the interrelations between High Secondary Containment
Area Temperature and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
Area/room coolers
Reactor building HVAC system
Leak detection and isolation system
Temperature sensitive instrumentation
Systems required for safe shut-down
Systems required to suppress a fire
Systems required for adequate core cooling
Systems required to protect the containment
Primary system discharging in secondary containment
EK3.
Knowledge of the reasons for the following responses as they apply to
High Secondary Containment Area Temperature:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
Emergency depressurization
Reactor scram
Isolating affected systems
IMPORTANCE
4
4
3
3
3
4
4
4
3
4
3
3
3
3
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Secondary Containment Area Temperature:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
Reactor building HVAC system
Fire protection system
Affected systems so as to isolate damaged portions
EA2.
Ability to determine and/or interpret the following as they apply to High
Secondary Containment Area Temperature:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
Area temperature
Equipment operability
Cause of high area temperature
NUREG-2104
4.1-12
4
3
4
RO SRO
4
4
3
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1009 High Secondary Containment Area Radiation Levels
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Secondary Containment Area Radiation Levels:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
EK1.04
EK1.05
Component environmental qualifications
Personnel protection
Radiation releases.
Maximum normal operating radiation limit
Maximum safe operating radiation limit
EK2.
Knowledge of the interrelations between High Secondary Containment
Area Radiation Levels and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
Area radiation monitoring system
Process radiation monitoring system
Reactor Building HVAC system
Standby gas treatment system
Primary system discharging in to secondary containment
Systems required to suppress a fire
Systems required for adequate core cooling
Systems required to shutdown the reactor
Systems required to protect the containment
EK3.
Knowledge of the reasons for the following responses as they apply to
High Secondary Containment Area Radiation Levels:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
Emergency depressurization
Reactor scram
Isolating affected systems
Personnel evacuation
Emergency plan
IMPORTANCE
2
4
4
3
3
4
4
4
4
3
3
3
3
3
3
4
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High
Secondary Containment Area Radiation Levels:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
EA1.04
EA1.05
Area radiation monitoring system.
Process radiation monitoring system
Reactor Building HVAC system
Standby gas treatment system
Affected systems so as to isolate damaged portions
4.1-13
4
4
4
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1009 High Secondary Containment Area Radiation Levels (continued)
K/A NO.
ABILITY
EA2.
Ability to determine and/or interpret the following as they apply to High
Secondary Containment Area Radiation Levels:
(CFR: 41.10/43.5/45.13)
EA2.01
EA2.02
EA2.03
Area radiation levels
Equipment operability
Cause of high area radiation
NUREG-2104
IMPORTANCE
4.1-14
RO SRO
4
4
3
3
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1010 Reactor Building Heating, Ventilation, and Air Conditioning
Exhaust High Radiation
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Reactor Building HVAC Exhaust High Radiation:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
Personnel protection
Radiation releases
EK2.
Knowledge of the interrelations between Reactor Building HVAC Exhaust
High Radiation and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
Process radiation monitoring system
Standby gas treatment system
Reactor building HVAC system
Leak detection and isolation system
EK3.
Knowledge of the reasons for the following responses as they apply to
Reactor Building HVAC Exhaust High Radiation:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
Isolating reactor building HVAC system
Starting standby gas treatment system
Personnel evacuation.
IMPORTANCE
4
4
4
4
4
4
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Reactor
Building HVAC Exhaust High Radiation:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
Process radiation monitoring system
Reactor building HVAC system
Standby gas treatment system
EA2.
Ability to determine and/or interpret the following as they apply to Reactor
Building HVAC Exhaust High Radiation:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
Ventilation radiation levels
Cause of high radiation levels
4
4
4
RO SRO
4
4
4
4
4.1-15
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1011 Secondary Containment High Differential Pressure
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Secondary Containment High Differential Pressure:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
Secondary containment integrity
Radiation release
EK2.
Knowledge of the interrelations between Secondary Containment High
Differential Pressure and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
Reactor building HVAC system
Standby gas treatment system
Off-site release rate
EK3.
Knowledge of the reasons for the following responses as they apply to
Secondary Containment High Differential Pressure:
(CFR: 41.5 / 45.6)
EK3.01
Reactor building HVAC system response
IMPORTANCE
4
4
4
4
3
3
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Secondary
Containment High Differential Pressure:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
Reactor building HVAC system
Standby gas treatment system
EA2.
Ability to determine and/or interpret the following as they apply to
Secondary Containment High Differential Pressure:
(CFR: 41.8 to 41.10)
EA2.01
EA2.02
NUREG-2104
Secondary containment pressure
Off-site release rate
4.1-16
4
4
RO SRO
4
4
3
4
4.1
Emergency Plant Evolutions
EPE:
EPE1012 Secondary Containment High Floor Drain Sump/Area Water Level
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Secondary Containment High Floor Drain Sump/Area Water
Level:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
EK1.04
Radiation releases
Electrical ground/ circuit malfunction
Maximum normal operating limit
Maximum safe operating limit
EK2.
Knowledge of the interrelations between Secondary Containment High
Floor Drain Sump/Area Water Level and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
Reactor building high conductivity and low conductivity waste drains
Radwaste system
Systems required to shutdown the reactor
Systems required to suppress a fire
Systems required for adequate core cooling
Primary system discharging to secondary containment
EK3.
Knowledge of the reasons for the following responses as they apply to
Secondary Containment High Floor Drain Sump/Area Water Level:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
Emergency depressurization
Reactor scram
Isolating affected systems (PRA)
Pumping reactor building sumps
IMPORTANCE
3
3
3
3
3
3
3
3
3
3
3
3
4
3
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Secondary
Containment High Floor Drain Sump/Area Water Level:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
Reactor building high conductivity and low conductivity waste drains
Affected systems so as to isolate damaged portions (PRA)
Radwaste systems
4.1-17
3
4
3
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1012 Secondary Containment High Floor Drain Sump/Area Water Level
(continued)
K/A NO.
ABILITY
EA2.
Ability to determine and/or interpret the following as they apply to
Secondary Containment High Floor Drain Sump/Area Water Level:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
Operability of components within the affected area
Water level in the affected area
Cause of the high water level
NUREG-2104
IMPORTANCE
4.1-18
RO SRO
3
3
3
3
3
4
4.1
Emergency Plant Evolutions
EPE:
EPE1013 Scram Condition and Reactor Power >5% or Unknown
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to Scram Condition and Reactor Power >5% or Unknown:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
EK1.04
EK1.05
EK1.06
EK1.07
Reactor pressure effects on reactor power
Reactor water level effects on reactor power
Boron effects on reactor power (Standby liquid control system)
Hot shutdown boron weight
Cold shutdown boron weight
Cooldown effects on reactor power
Shutdown margin
EK2.
Knowledge of the interrelations between Scram Condition and Reactor
Power >5% or Unknown and the following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
EK2.10
EK2.11
EK2.12
EK2.13
Reactor trip and isolation system
Alternate reactor shutdown system
Control rod drive system
Neutron monitoring system
Safety parameter display system
Reactor pressure vessel water level
Reactor pressure vessel pressure
Rod control and information system
Alternate boron injection methods
Reactor internal pump runback and trips
Feedwater spargers
ESF logic and control system
Reactor recirculation flow control system
EK3.
Knowledge of the reasons for the following responses as they apply to
Scram Condition and Reactor Power >5% or Unknown:
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
EK3.05
EK3.06
EK3.07
EK3.08
EK3.09
Reactor internal pump runback
Trip of reactor internal pumps
Standby liquid control system injection
Lowering of reactor pressure vessel water level (PRA)
Feedwater pump runback
Hot shutdown boron weight
Cold shutdown boron weight
Maintaining heat sinks external to the containment
Various alternate methods of control rod insertion
IMPORTANCE
4.1-19
4
4
4
3
3
4
3
4
4
4
4
3
4
4
4
3
4
4
3
4
4
4
4
4
4
2
3
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1013 Scram Condition and Reactor Power >5% or Unknown (continued)
K/A NO.
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to Scram
Condition and Reactor Power >5% or Unknown:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
EA1.04
EA1.05
EA1.06
EA1.07
EA1.08
EA1.09
EA1.10
EA1.11
EA1.12
Reactor trip and isolation system
Initiation of alternate rod insertion function
Standby liquid control system
Control rod drive system
Neutron monitoring system
Rod control and information system
Safety parameter display system
Alternate boron injection methods
Leak detection and isolation system
Feedwater control system (PRA)
Reactor recirculation system
Reactor water cleanup system
EA2.
Ability to determine and/or interpret the following as they apply to Scram
Condition and Reactor Power >5% or Unknown:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
EA2.05
EA2.06
EA2.07
EA2.08
EA2.09
NUREG-2104
IMPORTANCE
Reactor power
Reactor power oscillations
Reactor pressure vessel water level
Standby liquid control system tank level
Suppression pool temperature
Control rod position
Reactor pressure vessel pressure
Containment conditions/isolations
Reactor recirculation flow
4.1-20
4
4
4
4
4
4
3
4
4
4
4
4
RO SRO
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4.1
Emergency Plant Evolutions
EPE:
EPE1014 High Off-Site Release Rate
K/A NO.
KNOWLEDGE
EK1.
Knowledge of the operational implications of the following concepts as
they apply to High Off-Site Release Rate:
(CFR: 41.8 to 41.10)
EK1.01
EK1.02
EK1.03
Biological effects of radioisotope ingestion
Protection of the general public
Meteorological effects on off-site release
EK2.
Knowledge of the interrelations between High Off-Site Release Rate and the
following:
(CFR: 41.7 / 45.8)
EK2.01
EK2.02
EK2.03
EK2.04
EK2.05
EK2.06
EK2.07
EK2.08
EK2.09
EK2.10
Radwaste system
Offgas system
Plant ventilation systems
Stack-gas monitoring system
Site emergency plan
Process radiation monitoring system
Control room habitability area HVAC system
Safety parameter display system
Post accident sample system (PASS)
Condensate air extraction system
EK3.
Knowledge of the reasons for the following responses as they apply to
High Off-Site Release Rate
(CFR: 41.5 / 45.6)
EK3.01
EK3.02
EK3.03
EK3.04
Implementation of site emergency plan
System isolations
Control room habitability area HVAC system isolation
Emergency depressurization
IMPORTANCE
3
4
3
3
4
4
4
4
3
4
3
3
3
4
4
4
4
ABILITY
EA1.
Ability to operate and/or monitor the following as they apply to High OffSite Release Rate:
(CFR: 41.7 / 45.6)
EA1.01
EA1.02
EA1.03
EA1.04
EA1.05
EA1.06
EA1.07
Stack-gas monitoring system
Meteorological instrumentation
Process radiation monitoring system
Safety parameter display system
Post accident sample system
Plant ventilation
Control room habitability area HVAC system
4.1-21
4
3
4
3
3
4
4
NUREG-2104
4.1
Emergency Plant Evolutions
EPE:
EPE1014 High Off-Site Release Rate (continued)
K/A NO.
ABILITY
EA2.
Ability to determine and/or interpret the following as they apply to High OffSite Release Rate:
(CFR: 41.10 / 43.5 / 45.13)
EA2.01
EA2.02
EA2.03
EA2.04
Off-site
Total number of curies released
Radiation levels
Source of offsite release
NUREG-2104, Rev. 0
IMPORTANCE
4.1-22
RO SRO
3
4
2
3
4
4
4
4
4.2
Abnormal Plant Evolutions
APE:
APE2001 Partial or Complete Loss of Forced Core Flow Circulation
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Partial or Complete Loss of Forced Core Flow Circulation:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
AK1.05
Natural circulation
Power/flow distribution
Thermal limits
Power oscillations
Selected control rod run-in initiation
AK2.
Knowledge of the interrelations between Partial or Complete Loss of
Forced Core Flow Circulation and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.05
AK2.06
AK2.07
Reactor recirculation system
Nuclear boiler instrumentation
Reactor pressure vessel water level
Reactor power
Core flow indication
Reactor trip and isolation system
AK3.
Knowledge of the reasons for the following responses as they apply to
Partial or Complete Loss of Forced Core Flow Circulation:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
Reactor pressure vessel water level response.
Reactor power response
Reactor scram
Reduced reactor internal pump operating requirements
Core flow indication
IMPORTANCE
4.2-1
4
3
4
3
4
4
3
4
4
4
4
3
4
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2001 Partial or Complete Loss of Forced Core Flow Circulation
(continued)
K/A NO.
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Partial or
Complete Loss of Forced Core Flow Circulation:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
Reactor recirculation system
Reactor trip and isolation system
Rod control and information system
Recirculation flow control system
Neutron monitoring system
Nuclear boiler instrumentation
Selected control rod run-in
AA2.
Ability to determine and/or interpret the following as they apply to Partial or
Complete Loss of Forced Core Flow Circulation:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
Power/flow map
Neutron monitoring
Actual core flow
Reactor internal pump operability
Nuclear boiler instrumentation
NUREG-2104
IMPORTANCE
4.2-2
4
3
3
3
3
3
4
RO SRO
4
4
3
3
3
3
3
3
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2002 Loss of Main Condenser Vacuum
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Loss of Main Condenser Vacuum:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
Plant efficiency
Turbine efficiency
Loss of heat sink
Offgas flow changes
AK2.
Knowledge of the interrelations between Loss of Main Condenser Vacuum
and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
Ak2.09
AK2.10
Main turbine
Leak detection and isolation system
Steam bypass and pressure control system
Feedwater system
Condensate system
Offgas system
Circulating water system
Seal steam
Reactor trip and isolation system
Condensate air extraction system
AK3.
Knowledge of the reasons for the following responses as they apply to
Loss of Main Condenser Vacuum:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
AK3.06
Turbine trip
Turbine bypass valve closure
Main steam isolation valve closure
Air ejector flow
Decreased main generator output
Reactor power reduction
IMPORTANCE
2
2
4
3
3
4
3
3
3
3
3
3
3
2
3
3
3
3
2
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Loss of
Main Condenser Vacuum:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.08
AA1.09
Condensate system
Offgas system
Reactor trip and isolation system
Leak detection and isolation system
Main turbine
Steam bypass and pressure control system
Circulating water system
Recirculation flow control system
Rod control and information system
4.2-3
3
3
3
3
3
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2002 Loss of Main Condenser Vacuum (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Loss of
Main Condenser Vacuum:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
Condenser vacuum/absolute pressure
Generator output.
Offgas system flow
NUREG-2104
IMPORTANCE
4.2-4
RO SRO
3
3
2
2
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2003 Partial or Complete Loss of AC Power
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Partial or Complete Loss of AC Power:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
AK1.05
AK1.06
Effect of battery discharge rate on capacity
Load shedding
Under voltage/degraded voltage effects on electrical loads
Electrical bus divisional separation
Failsafe component design
Station blackout
AK2.
Knowledge of the interrelations between Partial or Complete Loss of AC
Power and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
Station batteries
Emergency diesel generators
Combustion turbine generators
AC electrical distribution system
AC electrical loads
DC electrical loads
Reactor core isolation cooling system
AK3.
Knowledge of the reasons for the following responses as they apply to
Partial or Complete Loss of AC Power:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
Manual and auto bus transfer
Load shedding and sequencing
Ground isolation
Reactor scram
Containment isolation
IMPORTANCE
3
3
3
3
3
4
3
4
4
4
3
3
4
3
4
3
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Partial or
Complete Loss of AC Power:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA2.03
AA1.04
AA1.05
AC electrical distribution system
Emergency diesel generators
Combustion turbine generators
Systems necessary to assure safe plant shutdown
DC electrical distribution system
4.2-5
4
4
4
4
4
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2003 Partial or Complete Loss of AC Power (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Partial or
Complete Loss of AC Power:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
NUREG-2104
IMPORTANCE
Cause of partial or complete loss of AC power
Reactor power, pressure, and level
Battery status
System lineups
Whether a partial or complete loss of AC power has occurred
4.2-6
RO SRO
3
4
4
4
3
4
4
4
4
4
4.2
Abnormal Plant Evolutions
APE:
APE2004 Partial or Complete Loss of DC Power
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Partial or Complete Loss of DC Power:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
Electrical bus divisional separation
Effect of battery discharge rate on capacity
Loss of breaker protection
Prevention of inadvertent system(s) actuation upon restoration of
DC power
AK2.
Knowledge of the interrelations between Partial or Complete Loss of DC
Power and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
Battery charger
Batteries
Reactor core isolation cooling system
AC electrical power distribution system
Emergency diesel generators
AK3.
Knowledge of the reasons for the following responses as they apply to
Partial or Complete Loss of DC Power:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
Manual load shedding
Ground isolation/fault determination
Reactor scram
IMPORTANCE
3
3
3
3
3
3
4
3
3
3
3
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Partial or
Complete Loss of DC Power:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
DC electrical distribution systems
Systems necessary to assure safe plant shutdown
AC electrical power distribution system
4.2-7
3
4
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2004 Partial or Complete Loss of DC Power (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Partial or
Complete Loss of DC Power:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
Cause of partial or complete loss of DC power
Extent of partial or complete loss of DC power
Battery voltage
System lineups
NUREG-2104
IMPORTANCE
4.2-8
RO SRO
3
4
4
4
3
3
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2005 Main Turbine Generator Trip
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to MAIN TURBINE GENERATOR TRIP:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Pressure effects on reactor power
Core thermal limit considerations
Pressure effects on reactor pressure vessel water level
AK2.
Knowledge of the interrelations between MAIN TURBINE GENERATOR
TRIP and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
AK2.09
Reactor trip and isolation system
Feedwater temperature
Reactor recirculation system
Main generator protection
Extraction steam system
Seal steam evaporator
Steam bypass and pressure control system
AC electrical distribution system
Turbine protection
AK3.
Knowledge of the reasons for the following responses as they apply to
MAIN TURBINE GENERATOR TRIP:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
AK3.06
Reactor scram
Reactor internal pump trip and runback
Feedwater temperature decrease
Main generator trip
Extraction steam/moisture separator isolations
Main turbine bypass valve operation
IMPORTANCE
4
3
4
4
3
3
3
3
2
4
3
3
4
3
3
3
3
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to MAIN
TURBINE GENERATOR TRIP:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
Reactor recirculation system
Reactor trip and isolation system
Rod control and information system
Main generator controls
Steam bypass and pressure control system
4.2-9
3
4
3
3
4
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2005 Main Turbine Generator Trip (continued)
K/A NO.
ABILITY
AA1.06
AA1.07
AA1.08
Condenser vacuum breaker
AC electrical distribution
Turbine control system
AA2.
Ability to determine and/or interpret the following as they apply to Main
Turbine Generator Trip:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
AA2.07
AA2.08
AA2.09
Turbine speed
Turbine vibration
Turbine valve position
Reactor pressure vessel pressure
Reactor power
Feedwater temperature
Reactor pressure vessel water level
Electrical distribution status
Number of running reactor internal pumps and their speed
NUREG-2104
IMPORTANCE
2
3
3
4.2-10
RO SRO
3
3
2
3
3
3
4
4
4
4
3
3
4
4
3
3
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2006 Reactor Scram
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Reactor Scram:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Decay heat generation and removal
Shutdown margin
Reactivity control
AK2.
Knowledge of the interrelations between Reactor Scram and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
Reactor trip and isolation system
Feedwater control system
Control rod drive system
Reactor power
Steam bypass and pressure control system
Recirculation flow control system
Main turbine trip
AK3.
Knowledge of the reasons for the following responses as they apply to
Reactor Scram:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.06
AK3.07
Reactor pressure vessel water level response
Reactor power response
Reactor pressure vessel pressure response
Reactor pressure vessel water level setpoint setdown
Reactor internal pump speed reduction
Scram follow function
IMPORTANCE
4
3
4
4
4
4
4
4
3
4
4
4
4
3
3
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Reactor
Scram:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.08
Reactor trip and isolation system
Feedwater control system (PRA)
Steam bypass and pressure control system
Reactor recirculation system
Neutron monitoring system
Control rod drive system
Rod control and information system
Recirculation flow control system
4.2-11
4
4
4
3
4
4
4
4
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2006 Reactor Scram (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Reactor
Scram:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
Reactor power
Control rod position
Reactor pressure vessel water level
Reactor pressure vessel pressure
Whether a reactor scram has occurred
Cause of reactor scram
NUREG-2104
IMPORTANCE
4.2-12
RO SRO
4
4
4
4
4
4
4
4
4
4
4
4
4.2
Abnormal Plant Evolutions
APE:
APE2007 High Reactor Pressure
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Reactor Pressure:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Pump shutoff head
Pressure effects on reactor power
Turbine load
AK2.
Knowledge of the interrelations between High Reactor Pressure and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
Steam bypass and pressure control system
Reactor power
Residual heat removal system/Low pressure flooder mode
Shutdown cooling
AK3.
Knowledge of the reasons for the following responses as they apply to
High Reactor Pressure:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
RCIC operation
Safety/relief valve operation
Steam bypass and pressure control system operation
Reactor internal pump trip
IMPORTANCE
3
4
3
4
4
3
3
3
4
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High
Reactor Pressure:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
Reactor core isolation cooling system
Safety/relief valve operation
Steam bypass and pressure control system
AA2.
Ability to determine and/or interpret the following as they apply to High
Reactor Pressure:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
3
4
4
Reactor pressure vessel pressure
Reactor power
Reactor pressure vessel water level
RO SRO
4
4
4
4
4
4
4.2-13
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2008 High Reactor Water Level
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Reactor Water Level:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Moisture carryover.
Component erosion/damage
Feed flow/steam flow mismatch
AK2.
Knowledge of the interrelations between High Reactor Water Level and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
Feedwater system
Feedwater control system
Reactor core isolation cooling system
High pressure core flooder system
Main turbine
Steam bypass and pressure control system
Reactor water cleanup system (ability to drain)
AK3.
Knowledge of the reasons for the following responses as they apply to
High Reactor Water Level:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
Main turbine trip
Feedwater pump trip
Reactor core isolation cooling system steam supply valve closure
High pressure core flooder injection valve closure
IMPORTANCE
3
3
3
4
4
3
3
3
3
3
3
3
3
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High
Reactor Water Level:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
Feedwater control system
Reactor water cleanup (ability to drain)
Reactor core isolation cooling system
High pressure core flooder system
Main turbine
Feedwater system
NUREG-2104
4.2-14
4
3
3
3
3
4
4.2
Abnormal Plant Evolutions
APE:
APE2008 High Reactor Water Level (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to High
Reactor Water Level:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
Reactor pressure vessel water level
Steam flow/feed flow mismatch
Reactor water cleanup blowdown flow
Reactor pressure vessel water swell
IMPORTANCE
4.2-15
RO SRO
4
4
3
3
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2009 Low Reactor Water Level
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Low Reactor Water Level:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
Steam carryunder
Natural circulation
AK2.
Knowledge of the interrelations between Low Reactor Water Level and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
Reactor pressure vessel water level indication
Feedwater water level control system
Reactor recirculation system
Reactor water cleanup system
Reactor trip and isolation system
AK3.
Knowledge of the reasons for the following responses as they apply to Low
Reactor Water Level:
(CFR: 41.5 / 45.6)
AK3.01
Reactor internal pump trip and runback
IMPORTANCE
3
3
4
4
3
3
4
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Low
Reactor Water Level
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
Feedwater system
Feedwater control system
Recirculation system
Reactor water cleanup system
Condensate system (PRA)
AA2.
Ability to determine and/or interpret the following as they apply to Low
Reactor Water Level:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
Reactor pressure vessel water level
Steam flow/feed flow mismatch
Reactor water cleanup blowdown flow rate
NUREG-2104
4.2-16
4
4
3
3
4
RO SRO
4
4
4
4
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2010 High Drywell Pressure
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Drywell Pressure:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
Temperature increases
Dewpoint increases
AK2.
Knowledge of the interrelations between High Drywell Pressure and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
Suppression pool level
Drywell wetwell differential pressure
Atmosphere control system
Drywell cooling and ventilation
Instrument air system
High pressure nitrogen gas supply system
AK3.
Knowledge of the reasons for the following responses as they apply to
High Drywell Pressure:
(CFR: 41.5/45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
AK3.06
Drywell/wetwell venting
Increased drywell cooling
Radiation level monitoring
Leak investigation
Temperature monitoring
Termination of drywell inerting
IMPORTANCE
3
2
3
3
3
4
3
3
4
3
3
4
4
2
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High
Drywell Pressure:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.08
Drywell ventilation/cooling
High and low conductivity waste sumps
Nitrogen makeup
Sampling system
Drywell wetwell vent and purge
Leakage detection systems
Atmosphere control system
Process radiation monitoring system
4.2-17
3
4
3
3
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2010 High Drywell Pressure (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to High
Drywell Pressure:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
Leak rates
Drywell pressure
Drywell radiation levels
Drywell temperature
Drywell dewpoint
NUREG-2104
IMPORTANCE
RO SRO
3
4
4
4
3
4
4
4
2
2
4.2-18
4.2
Abnormal Plant Evolutions
APE:
APE2011 High Drywell Temperature
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Drywell Temperature:
(CFR: 41.8 to 41.10)
AK1.01
Pressure/temperature relationship
AK2.
Knowledge of the interrelations between High Drywell Temperature and the
following:
(CFR: 41.7 / 45.8)
AK2.01
Drywell cooling
AK3.
Knowledge of the reasons for the following responses as they apply to
High Drywell Temperature:
(CFR: 41.5 / 45.6)
AK3.01
Increased drywell cooling
IMPORTANCE
3
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High
Drywell Temperature:
(CFR: 41.7 / 45.6)
AA1.01
AA2.02
Drywell cooling system
Sampling system
AA2.
Ability to determine and/or interpret the following as they apply to High
Drywell Temperature:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
Drywell temperature
Drywell pressure
Dewpoint
4
2
RO SRO
4
4
4
4
2
2
4.2-19
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2012 High Suppression Pool Temperature
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Suppression Pool Temperature:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Pool stratification
Ambient temperature effects
Localized heating
AK2.
Knowledge of the interrelations between High Suppression Pool
Temperature and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
Ak2.03
AK2.04
AK2.05
Suppression pool cooling
Safety relief valve actuation
Reactor core isolation cooling system
Suppression pool temperature monitoring system
ESF logic and control system
AK3.
Knowledge of the reasons for the following responses as they apply to
High Suppression Pool Temperature:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
Suppression pool cooling operation
Limiting heat additions
IMPORTANCE
3
2
3
4
4
4
4
3
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High
Suppression Pool Temperature:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
Suppression pool cooling
Systems that add heat to the suppression pool
AA2.
Ability to determine and/or interpret the following as they apply to High
Suppression Pool Temperature:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
Suppression pool temperature
Localized heating/stratification
NUREG-2104
4.2-20
4
4
RO SRO
4
4
3
4
4.2
Abnormal Plant Evolutions
APE:
APE2013 Inadvertent Reactivity Addition
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Inadvertent Reactivity Addition:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
AK1.05
Prompt critical
Reactivity anomaly
Shutdown margin
Fuel thermal limits
Abnormal reactivity additions
AK2.
Knowledge of the interrelations between Inadvertent Reactivity Addition
and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
AK2.09
AK2.10
AK2.11
AK2.12
Reactor trip and isolation system
Fuel thermal limits
Fuel temperature
Void concentration
Neutron monitoring system
Moderator temperature
Reactor power
Rod control and information system
Safety limits
Recirculation flow control system
Feedwater control system
Steam bypass and pressure control system
AK3.
Knowledge of the reasons for the following responses as they apply to
Inadvertent Reactivity Addition:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
Reactor scram
Control rod blocks
Selected control rod run-in
Core flow increase block
IMPORTANCE
4
3
4
4
4
4
4
3
3
4
3
4
3
4
4
3
3
4
4
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Inadvertent
Reactivity Addition:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
Reactor trip and isolation system
Recirculation flow control system
Rod control and information system
4.2-21
4
4
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2013 Inadvertent Reactivity Addition (continued)
K/A NO.
ABILITY
AA1.04
AA1.05
AA1.06
Neutron monitoring system
Steam bypass and pressure control system
Feedwater temperature
AA2.
Ability to determine and/or interpret the following as they apply to
Inadvertent Reactivity Addition:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
Reactor power
Reactor period
Cause of reactivity addition
Violation of fuel thermal limits
Violation of safety limits
NUREG-2104
IMPORTANCE
4
3
4
RO SRO
4
4
4
4
4
4
4
4
4
4
4.2-22
4.2
Abnormal Plant Evolutions
APE:
APE2014 Incomplete Scram
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Incomplete Scram:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
Shutdown margin
Cooldown effects on reactor power
Reactivity effects
Reactor pressure
AK2.
Knowledge of the interrelations between Incomplete Scram and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
Control rod drive system
Rod control and information system
Reactor trip and isolation system
Neutron monitoring system
Safety parameter display system
Instrument air
Alternate rod insertion
AK3.
Knowledge of the reasons for the following responses as they apply to
Incomplete Scram:
(CFR: 41.5 / 45.6)
AK3.01
Bypassing rod insertion blocks
IMPORTANCE
4
4
4
4
4
3
4
4
3
4
4
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Incomplete
Scram:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
Control rod drive system
Reactor trip and isolation system
Rod control and information system
Neutron monitoring system
Safety parameter display system.
Alternate rod insertion
4.2-23
4
4
3
4
3
4
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2014 Incomplete Scram (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to
Incomplete Scram:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
Reactor power
Control rod position
NUREG-2104
IMPORTANCE
RO SRO
4
4
4
4
4.2-24
4.2
Abnormal Plant Evolutions
APE:
APE2015 Control Room Evacuation
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Control Room Evacuation:
(CFR: 41.8 to 41.10)
IMPORTANCE
None
AK2.
Knowledge of the interrelations between Control Room Evacuation and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
Remote shutdown panel
Local control stations
AK3.
Knowledge of the reasons for the following responses as they apply to
Control Room Evacuation:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
Reactor scram
Turbine trip
Disabling control room controls
4
4
4
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Control
Room Evacuation:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.08
AA1.09
AA1.10
AA1.11
AA1.12
AA1.13
Reactor trip and isolation system
AC electrical power distribution system
Reactor pressure vessel water level
Control room/local control transfer mechanisms (PRA)
Reactor pressure vessel pressure
Safety/relief valves (PRA)
Residual heat removal system (PRA)
High pressure core flooder system
Reactor building cooling water system
Reactor service water system
Atmosphere control system
Makeup water condensate system
Suppression pool temperature monitoring system
4.2-25
4
3
4
4
4
3
3
3
3
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2015 Control Room Evacuation (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Control
Room Evacuation:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
AA2.07
AA2.08
Reactor power
Reactor pressure vessel water level
Reactor pressure vessel pressure
Suppression pool temperature
Drywell pressure
Cooldown rate
Wetwell pressure
Reactor pressure vessel water temperature
NUREG-2104
IMPORTANCE
4.2-26
RO SRO
4
4
4
4
4
4
4
4
4
4
3
4
3
3
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2016 High Off-Site Release Rate
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to High Off-Site Release Rate:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Biological effects of radioisotope ingestion
Protection of the general public
Meteorological effects on off-site release
AK2.
Knowledge of the interrelations between High Off-Site Release Rate and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
AK2.09
AK2.10
AK2.11
AK2.12
AK2.13
Fission product production versus reactor power
Radwaste system
Off-gas system
Plant ventilation systems
Site emergency plan
Control room habitability area ventilation system
Safety parameter display system
Condensate air extraction system
Process radiation monitoring system
Standby gas treatment system
Reactor trip and isolation system
Leak detection and isolation system
Fuel pool cooling and cleanup system
AK3.
Knowledge of the reasons for the following responses they apply to High
Off-Site Release Rate:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
System isolations
Plant ventilation
Implementation of site emergency plan
Power reduction
Control room habitability area ventilation system
IMPORTANCE
2
4
3
3
3
3
3
3
3
3
3
3
3
3
4
2
4
3
4
4
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to High OffSite Release Rate:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.10
AA1.08
AA1.09
Radwaste system
Off-gas system
Plant ventilation systems
Safety parameter display system
Condensate air extraction system
Process radiation monitoring system
Standby gas treatment system
Reactor trip and isolation system
Leak detection and isolation system
Meteorological data
4.2-27
3
4
3
3
3
3
4
4
4
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2016 High Off-Site Release Rate (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to High OffSite Release Rate:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
Off-site release rate
Total number of curies released
Radiation levels
Source of off-site release
Meteorological data
NUREG-2104
IMPORTANCE
4.2-28
RO SRO
3
4
2
4
3
4
4
4
3
4
4.2
Abnormal Plant Evolutions
APE:
APE2017 Partial or Complete Loss of Reactor Building Cooling Water
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Partial or Complete Loss of Reactor Building Cooling Water:
(CFR: 41.8 to 41.10)
AK1.01
Effects on component/system operations
AK2.
Knowledge of the interrelations between Partial or Complete Loss of
Reactor Building Cooling Water and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
Reactor water cleanup system
Reactor recirculation system
Residual heat removal system
Fuel pool cooling and cleanup system
Plant operations
AK3.
Knowledge of the reasons for the following responses as they apply to
Partial or Complete Loss of Reactor Building Cooling Water:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
AK3.06
Isolation of non-essential heat loads
Reactor power reduction
Securing individual components (prevent equipment damage)
Starting standby pump
Placing standby heat exchanger in service
Increasing cooling water flow to heat exchangers
IMPORTANCE
4
3
3
4
3
3
3
3
3
3
3
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Partial or
Complete Loss of Reactor Building Cooling Water:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
System loads
Affected systems so as to isolate damaged portions
AA2.
Ability to determine and/or interpret the following as they apply to Partial or
Complete Loss of Reactor Building Cooling Water:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
Component temperatures
Cooling water temperature
Cause for partial or complete loss
System flow
System pressure
Surge tank level
4.2-29
3
3
RO SRO
3
3
3
3
3
3
3
3
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2018 Partial or Complete Loss of Instrument Air
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Partial or Complete Loss of Instrument Air:
(CFR: 41.8 to 41.10)
IMPORTANCE
None
AK2.
Knowledge of the interrelations between Partial or Complete Loss of
Instrument Air and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
AK2.09
AK2.10
AK2.11
Control rod drive system
Reactor building cooling water system
Feedwater system
Reactor water cleanup system
Main steam system
Offgas system
Fuel pool cooling and cleanup system
Radwaste system
Service air system
Reactor core isolation cooling
Atmosphere control system
AK3.
Knowledge of the reasons for the following responses as they apply to
Partial or Complete Loss of Instrument Air:
(CFR: 41.5 / 45.6)
AK301
AK3.02
Standby air compressor operation
Auto cross-tie of service air system to instrument air system
on lower instrument air header pressure
4
3
3
3
3
3
3
3
3
3
3
4
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Partial or
Complete Loss of Instrument Air:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
Backup air supply
Instrument air compressor power supplies
NUREG-2104
4.2-30
4
3
4.2
Abnormal Plant Evolutions
APE:
APE2018 Partial or Complete Loss of Instrument Air (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Partial or
Complete Loss of Instrument Air:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
Instrument air system pressure
Status of instrument air system loads
RO SRO
4
3
3
3
4.2-31
NUREG-2104
IMPORTANCE
4.2
Abnormal Plant Evolutions
APE:
APE2019 Inadvertent Containment Isolation
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Inadvertent Containment Isolation:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
AK1.05
Loss of normal heat sink
Power/reactivity control
Water chemistry
Bottom head thermal stratification
Loss of drywell/containment cooling
AK2.
Knowledge of the interrelations between Inadvertent Containment Isolation
and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
AK2.09
AK2.10
AK2.11
Main steam system
Sampling system
Drywell cooling
Reactor water cleanup system
Reactor core isolation cooling system
Automated traversing in-core probe system
Residual heat removal system/shutdown cooling
Low/high conductivity waste system
Standby gas treatment system
Reactor building cooling water system
Atmosphere control system
AK3.
Knowledge of the reasons for the following responses as they apply to
Inadvertent Containment Isolation:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
AK3.06
AK3.07
AK3.08
AK3.09
AK3.10
Reactor scram
Drywell/containment pressure response
Drywell/containment temperature response
Reactor pressure vessel pressure response
Reactor pressure vessel water level response
Suppression pool water level response.
Suppression pool temperature response
Wetwell pressure response
Safety relief valve operation
Reactor core isolation cooling operation
NUREG-2104
IMPORTANCE
4.2-32
4
4
2
3
3
4
3
3
3
3
3
3
3
3
3
3
4
3
3
4
4
3
3
3
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2019 Inadvertent Containment Isolation (continued)
K/A NO.
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Inadvertent
Containment Isolation:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
Leak detection and isolation system
Drywell cooling system
Reactor building HVAC system
AA2.
Ability to determine and/or interpret the following as they apply to
Inadvertent Containment Isolation:
(CFR: 41.10 / 43.5 / 45.13)
RO
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
IMPORTANCE
Drywell pressure
Drywell temperature
Reactor power
Reactor pressure vessel pressure
Reactor pressure vessel water level
Cause of isolation
4.2-33
4
3
3
SRO
4
3
4
4
4
3
4
3
4
4
4
4
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2020 Loss of Shutdown Cooling
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Loss of Shutdown Cooling:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
AK1.04
Decay heat
Thermal stratification
Adequate core cooling
Natural circulation
AK2.
Knowledge of the interrelations between Loss of Shutdown Cooling and
the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
Reactor pressure vessel water temperature
Reactor water cleanup system
Reactor building cooling water system
Fuel pool cooling and cleanup system
Reactor recirculation system
AK3.
Knowledge of the reasons for the following responses as they apply to
Loss of Shutdown Cooling:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
AK3.04
AK3.05
Raising reactor pressure vessel water level
Feeding and bleeding reactor vessel
Increasing drywell cooling
Maximizing reactor water cleanup flow
Establishing alternate decay heat removal flow paths (PRA)
IMPORTANCE
4
3
4
4
4
3
3
3
3
3
3
3
3
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Loss of
Shutdown Cooling:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
Reactor water cleanup system
RHR/shutdown cooling (PRA)
Reactor building cooling water systems
Alternate decay heat removal methods
Reactor recirculation system
Containment/ drywell temperature
NUREG-2104
4.2-34
3
4
3
4
3
3
4.2
Abnormal Plant Evolutions
APE:
APE2020 Loss of Shutdown Cooling (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to Loss of
Shutdown Cooling:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
AA2.07
Reactor pressure vessel water heatup/cooldown rate
RHR/shutdown cooling system flow
Reactor pressure vessel water level
Reactor pressure vessel water temperature
Reactor pressure vessel metal temperature
Reactor pressure vessel pressure
Reactor recirculation flow
IMPORTANCE
4.2-35
RO SRO
4
4
3
3
4
4
3
4
3
4
3
3
3
3
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2021 Loss of Control Rod Drive Pumps
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Loss of Control Rod Drive Pumps:
(CFR: 41.8 to 41.10)
AK1.01
Reactivity control
AK2.
Knowledge of the interrelations between Loss of Control Rod Drive Pumps
and the following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
Reactor recirculation system
Fine motion control rod mechanism
Hydraulic control unit accumulator pressures
Reactor pressure vessel water level
Reactor water cleanup system
AK3.
Knowledge of the reasons for the following responses as they apply to
Loss of Control Rod Drive Pumps:
(CFR: 41.5 / 45.6)
AK3.01
Reactor scram
IMPORTANCE
4
3
3
3
3
3
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Loss of
Control Rod Drive Pumps:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
Control rod drive system
Reactor trip and isolation system
Reactor recirculation system
Reactor water cleanup system
AA2.
Ability to determine and/or interpret the following as they apply to Loss of
Control Rod Drive Pumps:
(CFR: 41.10 / 43.5 / 45.13)
RO
AA2.01
AA2.02
NUREG-2104
Hydraulic control unit accumulator pressure
Control rod drive system status
4.2-36
3
4
3
3
SRO
4
3
4
3
4.2
Abnormal Plant Evolutions
APE:
APE2022 Refueling Accidents
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Refueling Accidents:
(CFR: 41.8 to 41.10)
AK1.01
AK1.02
AK1.03
Radiation exposure hazards
Shutdown margin
Inadvertent criticality
AK2.
Knowledge of the interrelations between Refueling Accidents and the
following:
(CFR: 41.7 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
Fuel handling equipment
Fuel pool cooling and cleanup system
Radiation monitoring equipment
Rod control and information system.
Reactor building HVAC system
Standby gas treatment system
AK3.
Knowledge of the reasons for the following responses as they apply to
Refueling Accidents:
(CFR: 41.5 / 45.6)
AK3.01
AK3.02
AK3.03
Refueling floor evacuation
Interlocks associated with fuel handling equipment
Reactor building HVAC system isolation
IMPORTANCE
4
3
4
3
3
3
3
4
4
4
3
3
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Refueling
Accidents:
(CFR: 41.7 / 45.6)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
Reactor building HVAC system
Fuel pool cooling and cleanup system
Fuel handling equipment
Radiation monitoring equipment
Neutron monitoring system
Standby gas treatment system
AA2.
Ability to determine and/or interpret the following as they apply to
Refueling Accidents:
(CFR: 41.10 / 43.5 / 45.13)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
Area radiation levels
Fuel pool level
Airborne contamination levels
Occurrence of fuel handling accident
Entry conditions of emergency plan
RO SRO
4
4
3
4
3
4
3
4
3
4
4.2-37
NUREG-2104
3
3
3
3
3
4
4.2
Abnormal Plant Evolutions
APE:
EPE2023 Plant Fire On-Site
K/A NO.
KNOWLEDGE
AK1
Knowledge of the operation applications of the following concepts as they
apply to Plant Fire On-Site:
(CFR: 41.10 / 43.5 / 45.11)
AK1.01
AK1.02
Fire classifications by type
Fire fighting
AK2.
Knowledge of the interrelations between Plant Fire On-Site and the
following:
(CFR: 41.10 / 43.5 / 45.11)
AK2.01
AK2.02
AK2.03
AK2.04
Sensors/detectors and dampers
Controllers and positioners
Motors
Breakers/relays/disconnects/transformers
AK3
Knowledge of the reasons for the following responses as they apply to
Plant Fire On-Site:
(CFR: 41.10 / 43.5 / 45.11)
AK3.01
AK3.02
Installation of fire detectors
Steps called out in the site fire protection plant/fire protection system
manual or fire zone manual
Fire detector surveillance test
Actions contained in the abnormal procedure for plant fire on site
(CFR: 41.10 / 43.5 / 45.11)
AK3.03
AK3.04
IMPORTANCE
3
3
3
2
3
3
2
2
2
3
ABILITY
AA1
Ability to operate and/ or monitor the following as they apply to Plant Fire
On-Site:
(CFR: 41.10 / 43.5 / 45.11)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
AA1.06
AA1.07
AA1.08
AA1.09
Respirator air pack
Re-installation of fire detector
Bypass of fire zone detector
Bypass of heat detector
Plant and control room habitability area ventilation systems
Fire alarm
Fire alarm reset panel
Fire fighting equipment used on each class of fire
Plant fire zone panel (including detector location)
NUREG-2104
4.2-38
3
2
2
2
3
3
2
3
3
4.2
Abnormal Plant Evolutions
APE:
EPE2023 Plant Fire On-Site (continued)
K/A NO.
ABILITY
AA2
Ability to determine and interpret the following as they apply to Plant Fire
On-Site:
(CFR: 41.10 / 43.5 / 45.11)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
AA2.07
AA2.08
AA2.09
AA2.10
AA2.11
AA2.12
AA2.13
AA2.14
AA2.15
AA2.16
AA2.17
AA2.18
IMPORTANCE
RO SRO
Damper position
3
3
Fire alarm
3
3
The fire's extent of potential operational damage to plant equipment 3
3
Ventilation alignment necessary to secure affected area
3
3
Need for pressurizing control room (recirculating mode)
3
3
Need for placing control room habitability area HVAC in smoke removal
mode
4
4
Need for placing the reactor building HVAC system in smoke removal
mode
4
4
Need for placing the turbine building HVAC system in smoke removal
mode
4
4
Need for placing the control building safety-related equipment area
HVAC system in smoke removal mode
4
4
Whether malfunction is due to common-mode electrical failures
3
3
Time limit of long-term-breathing air system for control room
3
3
Time limit for use of respirators
3
3
Location of vital equipment within fire zone
3
4
Need for emergency plant shutdown
3
4
Equipment that will be affected by fire suppression activities in each
zone
3
4
Requirements for establishing a fire watch
2
4
Vital equipment and control systems to be maintained and operated
during a fire
3
4
Systems that may be affected by the fire
3
4
4.2-39
NUREG-2104
4.2
Abnormal Plant Evolutions
APE:
APE2024 Generator Voltage and Electric Grid Disturbances
K/A NO.
KNOWLEDGE
AK1.
Knowledge of the operational implications of the following concepts as
they apply to Generator Voltage and Electric Grid Disturbances:
(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
AK1.01
AK1.02
Over-excitation
Under-excitation
AK2.
Knowledge of the interrelations between Generator Voltage and Electric
Grid Disturbances and the following:
(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
AK2.01
AK2.02
AK2.03
AK2.04
AK2.05
AK2.06
AK2.07
AK2.08
Motors
Breakers, relays
Sensors, detectors, indicators
Controllers, positioners
Pumps
Reactor power
Turbine/generator control
Bus frequency
AK3.
Knowledge of the reasons for the following responses as they apply to
Generator Voltage and Electric Grid Disturbances:
(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
AK3.01
AK3.02
Reactor and turbine trip criteria
Actions contained in abnormal operating procedure for voltage and
grid disturbances
IMPORTANCE
3
3
3
3
3
3
3
4
4
3
4
4
ABILITY
AA1.
Ability to operate and/or monitor the following as they apply to Generator
Voltage and Electric Grid Disturbances:
(CFR: 41.5 and 41.10 / 45.5, 45.7, and 45.8)
AA1.01
AA1.02
AA1.03
AA1.04
AA1.05
Grid frequency and voltage
Turbine/generator controls
Voltage regulator controls
Reactor controls
Engineered safety features
NUREG-2104
4
4
4
4
4
4.2-40
4.2
Abnormal Plant Evolutions
APE:
APE2024 Generator Voltage and Electric Grid Disturbances (continued)
K/A NO.
ABILITY
AA2.
Ability to determine and/or interpret the following as they apply to
Generator Voltage and Electric Grid Disturbances:
(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)
AA2.01
AA2.02
AA2.03
AA2.04
AA2.05
AA2.06
AA2.07
AA2.08
AA2.09
AA2.10
Operating point on the generator capability curve
Voltage outside the generator capability curve
Generator current outside the capability curve
VARs outside capability curve
Operational status of offsite circuit
Generator frequency limitations
Operational status of engineered safety features
Criteria to trip the turbine or reactor
Operational status of emergency diesel generators
Generator overheating and the required actions
IMPORTANCE
4.2-41
RO SRO
4
4
4
4
4
4
4
4
3
4
3
4
4
4
4
4
4
4
4
4
NUREG-2104
5.0
Components
COMPONENT:
291001 Valves
(CFR: 41.3)
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
K1.04
The operation of safety valves
The operation of relief valves
The relationship of valve position to flow rate and back pressure
Valve design for a given failed valve position (open, closed,
and, as-is positions); spring loaded valves; hydraulic,
pneumatically controlled valves; electric motor driven valves)
The significance of stem position (valve status) for gate valves
Safety concerns in the use of gate valves (protect valves seals,
open slowly)
Cautions for placing a valve controller in manual mode
Emergency operation of MOV with motor inoperable
The stroke test for a valve, including the use of a stopwatch
Principles of operation and purpose of check valves
Operation of manual valves and verification of position with
indicator lights
Reason for using globe valves versus gates valves for throttling
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
IMPORTANCE
5-1
RO SRO
3.4
3.5
3.4
3.6
2.7
2.8
2.7
2.9
2.8
2.8
2.7
3.4
3.4
2.7
3.1
2.7
3.4
3.5
2.7
3.1
3.2
2.6
3.2
2.8
NUREG 2104
COMPONENT:
291002 Sensors and Detectors
(CFR: 41.7)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Flow
K1.01
K1.02
K1.03
K1.04
K1.05
Operation of venturis and orifices
Temperature compensation requirements
Effects of gas or steam on liquid flow rate indications (erroneous
reading)
Modes of failure
Operation of a flow D/P cell type flow detector
2.4
2.4
2.5
2.5
2.5
2.9
3.1
2.6
3.1
3.1
2.8
3.2
2.9
3.2
2.8
3.3
2.9
3.3
2.4
2.5
2.3
2.8
2.9
2.5
2.9
3.1
2.3
2.6
2.4
2.8
2.5
2.7
2.3
2.4
Level
K1.06
K1.07
K1.08
K1.09
Temperature/pressure compensation requirements
Operation of a differential pressure level detector
Effects of operating environment (pressure, temperature,
and radiation)
Modes of failure
Pressure
K1.10
K1.11
K1.12
K1.13
Theory of operation of bourdon tubes, diaphragms, bellows,
and pressure detectors
Effects of operating environment (pressure, temperature,
radiation)
Operation of a pressure D/P cell
Modes of failure
Temperature
K1.14
K1.15
Theory of operation of T/C, RTD, thermostats, thermometers
(expanding fluid)
Indications of failure modes of T/C, RTD, thermometers
Position Detector_
K1.16
K1.17
NUREG-2104
Failure modes of reed switches, LVDT, limit switches, and
potentiometers
Applications of reed switches, magnets, LVDT, potentiometers,
and limit switches
5-2
COMPONENT:
291002 Sensors and Detectors (continued)
(CFR: 41.7)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Electrical
K1.18
Applications of voltmeters, ammeters, frequency, and ground
detectors
2.2
2.4
3.0
3.2
2.8
3.1
3.2
2.9
3.0
3.1
Operation of ion chambers, G-M tubes and scintillation detectors 2.8
Use of portable radiation monitoring instruments
3.1
2.9
3.2
Nuclear Instrumentation
K1.19
K1.20
K1.21
K1.22
Operation of fission chambers, ion chambers
Neutron monitoring indication units
Effects of voltage changes on neutron detector performance
Failure modes of fission chambers, ion chambers, and
proportional counters
Radiation Detection
K1.23
K1.24
5-3
NUREG 2104
COMPONENT:
291003 Controllers and Positioners
(CFR: 41.7)
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
NUREG-2104
IMPORTANCE
RO SRO
Function and operation of flow controller in manual and
automatic modes
3.5
3.7
Function and operation of a speed controller .................................. 3.5
3.6
Operation of a valve controller, including seal-in features............... 3.3
3.4
Function and operation of pressure and temperature controllers,
including pressure and temperature control valves......................... 3.3
3.3
Function and characteristics of valve positioners............................ 2.8
2.8
Function and characteristics of governors and other mechanical
controllers ...................................................................................... 2.5
2.6
Safety precautions with respect to the operation of controllers
and positioners
2.8
2.8
Theory of operation of the following types of controllers:
electronic, electrical, and pneumatic
2.2
2.2
Effects on operation of controllers due to proportional,
proportional and reset, and proportional
and integral features ...................................................................... 2.0
2.2
5-4
COMPONENT:
291004 Pumps
(CFR: 41.3)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Centrifugal
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
Identification, symptoms, and consequences of cavitation
Reasons for venting a centrifugal pump
Consequences of air binding
Consequences of operating a pump dead headed or for
extended recirculation times
Discuss relationships among head, flow, speed, and power
Need for net positive suction head (NPSH); effects of loss of
suction
Starting current and operating current interpretation
Purpose of starting a pump with discharge valve closed
Pressure and flow relationship of pumps in parallel
Pressure and flow relationship of pumps in series
Definition of pump shutoff head
"Runout" of a centrifugal pump (definition, indications, causes,
effects, and corrective measures)
Principles of operation of a centrifugal pump
Relationship between flow from a pump and suction heads
Purpose of pump minimum flow requirements
3.2
2.8
2.8
3.2
2.8
2.9
3.0
2.8
3.1
2.9
3.3
2.8
2.8
2.3
2.3
2.4
3.3
2.8
2.8
2.4
2.4
2.5
2.8
2.6
2.5
2.9
2.8
2.7
2.5
2.9
2.5
2.7
2.5
2.6
3.3
2.6
3.3
2.6
3.1
3.1
3.1
3.0
Positive Displacement
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
Discuss relationship among head, flow, speed, and power
Net positive suction head (NSPH) requirements for a positive
displacement pump
Consequences of operating a positive displacement pump
against a closed flow path
Functions and characteristics of positive displacement pumps
Reason for starting a positive displacement pump with the
discharge valve open; need to clear the flow path
Safety procedures and precautions associated with positive
displacement pumps
5-5
NUREG 2104
COMPONENT:
291005 Motors and Generators
(CFR: 41.7)
K/A NO.
KNOWLEDGE
K1.01
K1.02
Locked motor rotor, recognition from motor parameters
Potential consequences of overheating motor insulation or
motor bearings
Causes of excessive current in motors, such as low voltage,
overloading, and mechanical binding
Relationship between pump motor current (ammeter reading)
and the following: pump fluid flow, head, speed, and stator
temperature
Explain the difference between starting current and operating
(running) current in a motor
Reason for limiting the number of motor starts in a given time
period
Electrical units: volts, amps, AC, DC, and hertz
Consequences of overexcited/underexcited
Interrelations of the following: VARs, Watts, Amps, Volts,
Power factor
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
NUREG-2104
IMPORTANCE
5-6
RO SRO
2.6
2.6
2.6
2.7
2.6
2.7
2.7
2.7
2.6
2.7
2.9
2.6
2.5
3.1
2.6
2.6
2.3
2.6
COMPONENT:
291006 Heat Exchangers and Condensers
(CFR: 41.4)
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
IMPORTANCE
RO SRO
Startup/shutdown of a heat exchanger ........................................... 2.7
2.7
Proper filling of a shell and tube heat exchanger ............................ 2.6
2.6
Basic heat transfer in a heat exchanger ......................................... 2.4
2.6
Effects of heat exchanger flow rates that are too high or too low .... 2.8
2.8
Flow paths for the heat exchanger (counterflow and U-types) ........ 2.2
2.3
Components of a heat exchanger (shells, tubes, plates, etc.) ........ 2.3
2.3
Control of heat exchanger temperatures ........................................ 2.7
2.8
Relationship between flow rates and temperatures ........................ 2.9
3.0
Definition of thermal shock ............................................................. 2.7
2.8
Principle of operation of condensers .............................................. 2.8
2.8
Relationship between condenser vacuum and backpressure ......... 2.8
2.8
Causes of natural circulation .......................................................... 2.9
3.0
Use of steam tables to determine saturation pressure for a given
temperature and vice versa ............................................................ 2.7
2.9
Fluid hammer and methods of prevention ...................................... 3.1
3.2
Effects of heat exchanger tube fouling ........................................... 2.6
2.8
Effects of scaling on heat exchanger operation .............................. 2.5
2.6
Consequences of heat exchanger tube failure ............................... 2.7
2.8
Reasons for non-condensible gas removal..................................... 2.8
2.9
5-7
NUREG 2104
COMPONENT:
291007 Demineralizers and Ion Exchangers
(CFR: 41.3)
K/A NO.
KNOWLEDGE
K1.01
Effect of excessive differential pressure on demineralizer
performance
Reason for sampling inlet and outlet of demineralizer
Effects of channeling in a demineralizer
Purpose of a demineralizer
Purpose of demineralizer D/P gauge
Reason for demineralizer temperature and flow limits
Principles of demineralizer operation
Demineralizer D/P to determine condition of demineralizer
resin bed
Effects of demineralizer operation on water conductivity
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
NUREG-2104
IMPORTANCE
5-8
RO SRO
2.6
2.5
2.8
2.8
2.4
2.7
2.3
2.7
2.6
2.9
2.9
2.5
2.7
2.5
2.6
2.7
2.6
2.7
COMPONENT:
291008 Breakers, Relays and Disconnects
(CFR: 41.7)
K/A NO.
KNOWLEDGE
K1.01
Purpose for racking out breakers (de-energize components and
associated control and indication circuits)
Local indication that breaker is open, closed or tripped
Meaning of power supply circuit breaker indicator lights and
capability to remotely open and close
Operation of various push buttons, switches and handles and
the resulting action on breakers
Function of thermal overload protection device
Interpreting one-line diagram of control circuitry
Safety procedures and precautions associated with breakers,
including MCC bus breakers, high, medium and low voltage
breakers, relays and disconnects
Effects of closing breakers with current out of phase, different
frequencies, high voltage differential, low current, or too much
load
Effect of racking out breakers on control and indicating circuits
and removal of control power on breaker operation
Function, control, and precautions associated with disconnects
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
IMPORTANCE
5-9
RO SRO
3.6
3.4
3.6
3.5
3.3
3.4
3.3
3.0
3.2
3.3
3.1
3.6
3.5
3.7
3.4
3.5
3.4
3.3
3.5
3.4
NUREG 2104
6.0
Theory
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292001 Neutrons
K/A NO.
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
KNOWLEDGE
IMPORTANCE
Define fast, intermediate, and slow neutrons.
Define prompt and delayed neutrons.
Define thermal neutrons.
Describe neutron moderation.
Identify characteristics of good moderators.
Define neutron lifetime.
Define neutron generation time.
Describe fast flux, thermal flux, and flux distribution.
6.1-1
RO SRO
2.0
2.1
3.0
3.1
2.7
2.7
3.2
3.2
2.4
2.6
1.9
1.9
1.9
1.9
2.2
2.4
NUREG-2104
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292002 Neutron Life Cycle
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Describe the neutron life cycle using the following terms:
K1.01
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
NUREG-2104
fast fission factor
fast non-leakage probability factor
resonance escape probability factor
thermal non-leakage probability factor
thermal utilization factor
reproduction factor
Define critical, subcritical, and supercritical with respect to a
reactor
Define effective multiplication factor and discuss its
relationship to the state of the reactor
Define K-excess.
Define shutdown margin.
Define reactivity.
State the relationship between reactivity and effective
multiplication factor
Calculate shutdown margin using procedures and given
plant parameters
Evaluate change in shutdown margin due to changes in
plant parameters
6.1-2
1.9
1.9
2.0
1.9
1.9
1.9
1.9
1.9
2.1
2.0
2.0
1.9
3.5
3.5
2.7
2.4
3.2
3.2
2.8
2.6
3.5
3.3
2.4
2.5
1.8
2.4
2.6
2.9
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292003 Reactor Kinetics and Neutron Sources
K/A NO.
KNOWLEDGE
K1.01
K1.02
Explain the concept of subcritical multiplication.
Given the simplified formula for subcritical multiplication,
perform calculations involving steady state count rate and
source count rate.
Describe the production of delayed neutrons.
Define delayed neutron fraction and effective delayed
neutron fraction; state the reasons for variation.
Define reactor period.
Explain the effect of delayed neutrons on reactor period.
Explain prompt critical, prompt jump, and prompt drop.
Given the power equation, solve problems for power changes
and period.
Define doubling time and calculate it using the power equation.
Explain the necessity for installed neutron sources in a reactor
core.
Explain why installed sources are not needed after one cycle of
core operation.
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
IMPORTANCE
6.1-3
RO SRO
2.9
3.0
2.1
2.4
2.3
2.4
2.5
3.7
3.7
3.3
2.5
3.7
3.7
3.3
2.7
2.5
2.8
2.6
2.4
2.4
2.4
2.4
NUREG-2104
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292004 Reactivity Coefficients
K/A NO.
KNOWLEDGE
K1.01
K1.02
Define the temperature coefficient of reactivity.
Describe the effect on the magnitude of the temperature
coefficient of reactivity from changes in moderator temperature
and core age.
Explain resonance absorption.
Explain doppler broadening and self-shielding.
Define the doppler coefficient of reactivity.
K1.03
K1.04
K1.05
IMPORTANCE
RO SRO
3.2
3.2
2.5
2.6
2.6
2.9
2.6
2.7
2.7
2.9
Describe the effect on the magnitude of the doppler coefficient of reactivity for
changes in the following:
K1.06
K1.07
K1.08
K1.09
Moderator temperature
Core void fraction
Fuel temperature
Core age
2.1
2.1
2.2
1.9
2.2
2.2
2.4
2.1
K1.10
Define the void coefficient of reactivity.
3.2
3.2
Describe the effect on the magnitude of void coefficient from changes in the
following:
K1.11
K1.12
K1.13
K1.14
NUREG-2104
Core void fraction
Fuel temperature
Core age
Compare the relative magnitudes of the temperature, doppler,
and void coefficients of reactivity.
6.1-4
2.5
2.2
2.1
2.6
2.3
2.2
3.3
3.3
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292005 Control Rods
K/A NO.
KNOWLEDGE
K1.01
K1.02
Relate notch and rod position.
Name the material used for thermal neutron absorption in
control rods.
Describe nuclear properties of active material in the rod.
Predict direction of change in reactor power for a change
in control rod position.
Define rod density.
Define reactor scram.
Define control rod worth, differential control rod worth, and
integral control rod worth.
Explain the shape of curves for differential and integral CRW
versus rod position.
Explain direction of change in the magnitude of CRW for a
change in moderator temperature, void fraction, and control rod
density, and Xenon.
State the purpose of flux shaping and rod sequencing.
Define deep rods, and shallow rods.
Describe effects of deep and shallow control rods on
axial and radial flux distribution.
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
IMPORTANCE
6.1-5
RO SRO
3.2
3.3
2.5
1.9
2.6
1.9
3.5
2.5
3.7
3.5
2.6
3.8
2.4
2.6
2.1
2.3
2.5
2.8
2.4
2.6
3.3
2.5
2.6
2.9
NUREG-2104
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292006 Fission Product Poisons
K/A NO.
KNOWLEDGE
K1.01
K1.02
Define fission product poison.
State the characteristics of Xenon-135 as a fission product
poison.
Describe the production of Xenon-135.
Describe the removal of Xenon-135.
K1.03
K1.04
IMPORTANCE
RO SRO
2.7
2.8
3.1
2.9
2.9
3.1
2.9
2.9
Describe the following processes and state their effect on reactor operations:
K1.05
K1.06
K1.07
Equilibrium Xenon
Maneuvering Xenon
Xenon following a scram
2.9
2.7
3.2
2.9
2.7
3.2
K1.08
Describe the effects that Xenon concentration has on flux shape
and control rod patterns.
2.8
3.2
Plot the curve and explain the reasoning for the reactivity insertion by
Xenon-135 versus time for the following:
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
Initial reactor startup and ascension to rated power
Reactor startup with Xenon-135 already present in the core
Power changes from steady-state power to another
Reactor scram.
Reactor shutdown.
Explain the process and reasons for the Reactor Operator to
compensate for the time dependent behavior of Xenon-135
concentration in the reactor
State the characteristics of Samarium-149 as a fission product
poison.
Describe the production of Samarium-149.
Describe the removal of Samarium-149.
Define equilibrium samarium.
2.5
2.9
2.6
2.8
2.6
2.5
2.9
2.7
2.3
2.6
3.1
3.2
2.1
1.8
1.9
1.8
2.1
1.9
1.9
1.8
Plot the curve and explain the reasoning for reactivity insertion by
Samarium-149 versus time for the following:
K1.19
K1.20
Initial reactor startup and ascension to rated power
Reactor shutdown.
1.7
1.6
1.8
1.7
K1.21
K1.22
Describe effects of power changes on samarium concentration.
Compare effects of Samarium-149 on reactor operation with
those of Xenon-135.
1.7
1.8
2.4
2.4
NUREG-2104
6.1-6
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292007 Fuel Depletion and Burnable Poisons
K/A NO.
KNOWLEDGE
IMPORTANCE
K1.01
K1.02
K1.03
RO SRO
Define burnable poison and state its use in the reactor
2.9
3.1
Describe and explain distribution of burnable poisons in the core. 1.8
2.0
Given a curve of K-effective versus core age, state the reasons
for maximum, minimum, and inflection points.
2.4
2.7
6.1-7
NUREG-2104
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292008 Reactor Operational Physics
K/A NO.
KNOWLEDGE
IMPORTANCE
Startup and Approach to Criticality
K1.01
K1.02
K1.03
K1.04
K1.05
List parameters which should be monitored and controlled
during the approach to criticality.
List reactivity control mechanisms which exist for plant conditions
during the approach to criticality.
Describe count rate and period response which should be
observed for rod withdrawal during the approach to criticality.
Relate the concept of subcritical multiplication to predicted count
rate and period response for control rod withdrawal during the
approach to critical.
Explain characteristics to be observed when the reactor is very
close to criticality.
RO SRO
3.8
3.9
3.8
3.8
4.1
4.0
3.3
3.4
4.3
4.3
4.2
3.9
4.2
3.9
4.1
4.1
3.9
3.9
3.6
3.6
3.7
3.8
3.6
3.8
3.7
3.9
Criticality
K1.06
K1.07
K1.08
List parameters which should be monitored and controlled upon
reaching initial criticality.
Define criticality as related to a reactor startup.
Describe reactor power and period response once criticality is
reached.
Intermediate Range Operation
K1.09
K1.10
K1.11
K1.12
K1.13
List parameters which should be monitored and controlled during
the intermediate phase of startup.
Explain procedures for adjusting reactor period during the
intermediate phase of startup.
Discuss the concept of the point of adding heat (POAH) and its
impact on reactor power
Describe reactor power and period response prior to reaching
the POAH.
Explain characteristics to look for when the POAH is reached.
Heatup Operation
K1.14
K1.15
K1.16
NUREG-2104
Describe three parameters to be monitored and controlled during
heatup.
3.5
Describe reactor power and period response after reaching the
point of adding heat.
3.7
Explain procedures for establishing and controlling heatup rate. 3.6
6.1-8
3.5
3.7
3.7
6.1
Reactor Theory
(CFR: 41.1)
Reactor Theory: 292008 Reactor Operational Physics (Continued)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Power Operation
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
Describe three parameters to be monitored and controlled during
power operation.
3.6
Describe means by which reactor power will be increased to rated
power
3.8
Explain transient and steady-state effects of a control rod
withdrawal on reactor power and void fraction content.
3.1
Explain transient and steady-state effects of an increase in core
flow on reactor power and void fraction.
3.3
Explain the relationship between steam production rate and reactor
power given specific conditions.
2.9
Explain the effect that opening steam bypass valves, during power
operation, will have on reactor power
3.5
Explain the necessity for rod pattern exchanges.
2.6
Describe the parameters to be monitored and controlled during rod
pattern exchanges.
2.8
3.6
3.8
3.2
3.4
3.0
3.6
3.1
3.2
Reactor Response on a Scram
K1.25
Explain the shape of a curve of reactor power versus time after
a scram.
2.8
2.9
3.4
3.4
3.7
3.5
3.4
3.4
3.7
3.6
3.2
3.5
Normal Reactor Shutdown
K1.26
K1.27
K1.28
K1.29
K1.30
Explain reactor power response to a decrease in core flow.
Explain reactor power response to a control rod insertion.
Explain the necessity for inserting control rods in a predetermined
sequence during normal shutdown.
Define decay heat.
Explain the relationship between decay heat generation
and: a) power level history, b) power production, and
c) time since reaction shut down.
6.1-9
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293001 Thermodynamic Units and Properties
K/A NO.
KNOWLEDGE
K1.01
Convert between absolute and relative pressure and vacuum
scales.
Recognize the difference between absolute and relative
temperature scales.
Describe how common pressure and level sensing
instruments work.
Explain relationships between work, power, and energy.
K1.02
K1.03
K1.04
IMPORTANCE
6.2-1
RO SRO
2.2
2.3
2.1
2.1
2.5
1.8
2.7
1.9
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293002 Basic Energy Concepts
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
Identify energy and work forms.
Explain the law of conservation of energy.
Explain the difference between state and phase of a working
substance.
Explain the application of enthalpy in the monitoring of plant
processes.
Identify the relationship between heat flow during a process
and a T-S diagram representation of the process.
Define specific heat.
Apply specific heat in solving heat transfer problems.
K1.04
K1.05
K1.06
K1.07
NUREG-2104
IMPORTANCE
6.2-2
RO SRO
1.6
1.7
1.9
1.9
1.6
1.7
2.1
2.4
2.0
1.8
1.5
2.2
2.1
1.6
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293003 Steam
K/A NO.
KNOWLEDGE
K1.01
Describe effects of pressure on density or specific volume
of a liquid.
Distinguish between liquids, vapors, gases, and fluids.
Define latent heat of vaporization
Define vaporization line
Define critical point
Define vapor dome
Define saturated liquid
Define wet vapor
Define saturated vapor
Define vapor pressure
Define moisture content
Define quality
Define superheated vapor
Define supersaturated vapor
Define subcooled and compressed liquids
Define subcooling
Define specific heat
K1.02
K1.03
K1.04
K1.05
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
IMPORTANCE
RO SRO
2.3
2.2
2.3
2.0
1.8
1.8
2.7
1.8
2.5
1.8
2.3
2.5
2.3
1.8
2.4
2.8
1.9
2.4
2.3
2.4
2.1
1.8
2.0
2.8
1.9
2.6
1.9
2.3
2.6
2.4
1.8
2.4
2.8
2.1
1.7
2.1
2.2
2.1
1.7
2.2
2.3
2.2
2.9
2.8
3.2
3.1
Identify the following terms on a T-s diagram:
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
Critical point
Saturated liquid line
Saturated vapor line
Solid, liquid, gas, vapor, and fluid regions
Explain the usefulness of steam tables to the Control Room
Operator
Use saturated and superheated steam tables.
6.2-3
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293004 Thermodynamic Process
K/A NO.
KNOWLEDGE
K1.01
K1.02
Explain the relationship between real and ideal processes.
Explain the shape of the T-s diagram process line for
a typical boiler
IMPORTANCE
RO SRO
1.6
1.7
1.8
1.9
2.2
2.5
2.7
2.2
2.6
2.7
1.9
2.1
2.0
2.1
2.2
2.1
1.7
1.8
1.8
1.9
1.7
1.8
2.9
2.5
2.6
3.1
2.6
2.7
2.2
2.1
2.3
2.3
Nozzles
K1.03
K1.04
K1.05
Describe the functions of nozzles in flow restrictors.
Describe the functions of nozzles in air ejectors.
Describe the principles of operation of a jet pump
Turbines
K1.06
K1.07
K1.08
K1.09
Explain the function of nozzles, fixed blading, and moving
blading in the turbine.
Explain the reason turbines are multistages.
Define turbine efficiency.
Explain the difference between actual turbine performance
and ideal thermal efficiency.
Pumps
K1.10
K1.11
Define pump efficiency.
Explain the difference between ideal and real pumping
processes.
Condensers
K1.12
K1.13
K1.14
Discuss subcooling.
Explain vacuum formation in condenser processes.
Explain the condensing process.
Throttling and the Throttling Process
K1.15
K1.16
NUREG-2104
Define throttling.
Explain the reduction of process pressure from throttling.
6.2-4
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293005 Thermodynamic Cycles
K/A NO.
KNOWLEDGE
K1.01
K1.02
Define thermodynamic cycle.
Define thermodynamic cycle efficiency in terms of net work
produced and energy applied.
Describe the moisture effects on turbine integrity and
efficiency.
Explain steam quality effects on nuclear turbine design.
State the advantages of moisture separators/reheaters
and feedwater heaters for a typical steam cycle.
Kl.03
K1.04
K1.05
IMPORTANCE
6.2-5
RO SRO
1.7
1.8
1.7
1.7
2.6
2.3
2.7
2.4
2.7
2.8
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293006 Fluid Statics
K/A NO.
KNOWLEDGE
K1.01
K1.02
Distinguish between fluids and other substances.
Distinguish between static pressure, dynamic pressure, and
total pressure.
Define head loss.
Discuss operational considerations of viscosity as related to
head loss.
Explain operational implications of fluid hammer
K1.03
K1.04
K1.05
IMPORTANCE
RO SRO
1.7
1.8
2.0
2.4
2.2
2.5
1.7
3.2
1.9
3.3
2.5
2.5
2.6
2.6
2.5
2.8
2.7
2.4
2.6
2.9
2.8
2.5
2.9
2.9
2.6
2.7
2.2
2.3
2.3
2.4
2.1
2.3
2.6
2.7
1.9
2.1
2.7
2.4
2.4
2.2
2.2
2.9
2.4
2.6
2.3
2.3
Pumps and Pump Characteristics
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
NUREG-2104
State the purpose of a pump.
Discuss pump head.
Discuss relationship between pump speed, head, flow, and
power without using formulas or calculations.
Define cavitation.
Define net positive suction head (NPSH).
Define pump shut-off head, pump runout, and axial thrust.
Explain the importance of proper system venting for pump
operations.
Explain the results of putting centrifugal pumps in
parallel or series combinations.
Given the characteristic curve for a typical centrifugal pump,
explain the reason for its shape.
Using a centrifugal pump characteristic curve and a system
Characteristic curve, illustrate how the system operating
point changes due to system changes.
Describe how a centrifugal pump characteristic curve will
change with pump speed.
Explain how operating a centrifugal pump at shutoff head
may cause overheating of the pump and describe methods
used to avoid overheating.
Discuss the characteristic curve for a typical positive
displacement pump and explain the reason for its shape.
Describe the problems that will occur in emergency core
cooling systems if the pumps are operated at lower than
design flow for extended periods of time.
Define or explain mass flow rate
Define or explain two-phase flow
Define or explain pressure spike
Define or explain gas binding
6.2-6
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293006 Fluid Statics (continued)
K/A NO.
KNOWLEDGE
K1.24
K1.25
K1.26
Define or explain recirculation ratio
Define or explain pipe whip
Explain why flow measurements must be corrected for
density changes.
Explain the relationship between pressure head and
velocity head in a fluid system.
Describe the methods of controlling system flow rates.
K1.27
K1.28
IMPORTANCE
6.2-7
RO SRO
2.1
2.3
2.1
2.2
2.3
2.4
1.8
2.6
2.0
2.7
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293007 Heat Transfer and Heat Exchangers
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Heat Transfer
K1.01
K1.02
K1.03
Describe three mechanisms of heat transfer
Describe thermal conductivity.
Explain the manner in which fluid films affects heat transfer
3.2
2.4
2.7
3.2
2.6
2.8
1.9
2.0
2.2
2.2
2.7
2.8
2.7
2.9
3.0
2.5
3.1
2.7
2.7
2.6
2.6
2.3
2.9
3.1
2.7
2.9
Heat Exchangers
K1.04
K1.05
K1.06
K1.07
Discuss parallel flow heat exchangers.
Discuss counter-flow heat exchangers.
Discuss the factors which affect heat transfer rate in a heat
exchanger
Describe how the presence of gases or steam can affect heat
transfer and fluid flow in heat exchangers.
Condenser Applications of Heat Transfer
K1.08
K1.09
List functions of the main condenser in a power plant.
Discuss operational implications of condensate depression.
Core Thermal Power
K1.10
K1.11
K1.12
K1.13
NUREG-2104
Define core thermal power
Explain methods of calculating core thermal power
Define percent reactor power
Calculate core thermal power using a simplified heat balance.
6.2-8
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293008 Thermal Hydraulics
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Boiling Heat Transfer
K1.01
K1.02
K1.03
K1.04
K1.05
Distinguish between boiling processes and other heat
transfer mechanisms.
Describe surface or cavity nucleation.
List factors affecting bubble formation in a cavity.
Describe means by which boiling improves convection heat
transfer
Describe microconvection.
2.6
2.2
1.9
2.8
2.3
2.1
2.6
1.4
2.7
1.5
2.5
2.6
2.8
2.9
3.0
2.9
2.7
2.7
2.3
3.0
3.1
3.2
3.0
2.8
2.8
2.3
2.0
2.1
2.2
2.3
2.2
2.3
2.5
2.8
2.2
2.4
2.6
2.4
2.8
2.6
Pool Boiling Curve (T vs. Q/A)
K1.06
K1.07
K1.08
K1.09
K1.10
K1.11
K1.12
K1.13
Define a natural convection heat transfer
Define nucleate boiling, subcooled nucleate boiling, and bulk
boiling.
Describe departure from nucleate boiling
Describe onset of transition boiling.
Describe critical heat flux
Describe transition (partial film) boiling.
Describe stable film boiling.
Describe burnout and burnout heat flux.
Two Phase Flow
K1.14
K1.15
K1.16
K1.17
K1.18
Classify slug flow region along a fuel channel, experiencing
two phase flow.
Describe annular flow region along a hypothetical fuel
channel, experiencing two phase flow.
Describe dryout region or mist flow region along a
hypothetical fuel channel, experiencing two phase flow.
Describe onset of transition boiling point along a hypothetical
fuel channel, experiencing two phase flow.
Describe effects of flowrate and phase change on the heat
transfer coefficient.
Core Inlet Subcooling
K1.19
K1.20
Define core inlet subcooling.
Define carryunder
6.2-9
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293008 Thermal Hydraulics (continued)
K/A NO.
KNOWLEDGE
IMPORTANCE
Voids and Void Fraction
K1.21
K1.22
K1.23
K1.24
Define void fraction.
Explain the term void as applied to core operations
Define quality
Draw the temperature profile from the centerline of a fuel
pellet o the centerline of the channel.
RO SRO
3.0
2.9
2.5
3.0
3.0
2.7
2.4
2.5
3.2
2.9
2.9
3.2
3.1
3.0
2.3
2.5
2.8
3.0
2.7
2.9
2.5
2.7
3.0
2.6
Natural Circulation
2.4
2.6
Explain the causes of natural circulation in BWR's.
Describe problems that thermal stratification can cause.
Describe means by which the operator can determine if
natural circulation flow exists.
Describe means by which the operator can enhance natural
circulation.
2.9
3.1
3.1
3.3
3.1
3.3
3.2
3.4
Recirculation System
K1.25
K1.26
K1.27
K1.28
Explain the reason for forced core recirculation.
Explain the jet pump operating principle.
Explain the necessity of determining core coolant flow.
Describe the factors affecting single- and two-phase flow
resistance.
Core Orificing
K1.29
K1.30
K1.31
K1.32
K1.33
K1.34
K1.35
K1.36
K1.37
NUREG-2104
Describe the effects of increasing bundle power on bundle
flow resistance.
Compare the flow resistance through high powered bundles
to that of low powered bundles.
Explain the necessity of core orificing.
Describe core bypass flow.
Explain the need for adequate core bypass flow.
6.2-10
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293008 Thermal Hydraulics (continued)
K/A NO.
KNOWLEDGE
IMPORTANCE
RO SRO
Sketch the axial temperature and enthalpy profiles for a typical reactor
coolant channel and describe how they are affected by the following:
K1.38
K1.39
K1.40
K1.41
K1.42
Onset of nucleate boiling
Axial core flux
Inlet temperature
Heat generation rate
Flow rate in the channel
1.8
1.8
1.8
1.8
1.8
2.1
1.9
1.9
2.0
1.9
K1.43
Sketch the temperature profile in the axial and radial directions
for a typical fuel rod and explain the reason for its shape.
2.0
2.2
6.2-11
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293009 Core Thermal Limits
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
K1.04
K1.05
Explain radial peaking factor (RPF)
Explain axial peaking factor (APF)
Explain local peaking factor (LPF)
Explain total peaking factor (TPF)
State the reason thermal limits are necessary.
IMPORTANCE
RO SRO
2.1 2.5
2.2 2.6
2.1 2.5
2.2 2.6
3.3 3.5
Linear Heat Generation Rate (LHGR)
K1.06
K1.07
K1.08
K1.09
Define LHGR
Explain the basis of the limiting condition of LHGR
Describe the mode of fuel failure for LHGR
Define FLPD and MFLPD.
3.4
2.8
3.0
3.1
3.8
3.6
3.4
3.7
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
K1.10
K1.11
K1.12
K1.13
K1.14
K1.15
K1.16
Define average planar linear heat generation rate (APLHGR)
Explain the basis of the limiting condition for APLGHR
Describe the mode of fuel failure for APLHGR
Define MAPLHGR
Explain the mechanisms most limiting for each region of the
MAPLHGR limit curves.
Describe conditions under which radiative heat transfer
becomes the significant method of heat transfer within
a fuel bundle.
Discuss how changes in the heat generation rate and thermal
conductivity of the fuel rod affect fuel centerline temperature
3.3
2.8
2.9
3.1
3.7
3.6
3.5
3.6
2.2
2.7
2.6
3.1
2.4
2.8
3.3
3.2
2.8
3.1
3.1
2.9
2.8
2.7
3.7
3.7
3.6
3.6
3.6
3.3
3.2
3.2
2.7
3.2
2.6
3.1
Minimum Critical Power Ratio (MCPR)
K1.17
K1.18
K1.19
K1.20
K1.21
K1.22
K1.23
K1.24
K1.25
K1.26
NUREG-2104
Define critical power
Define critical power ratio
Explain the basis of the limiting condition for CPR
Describe the mode of fuel failure for CPR
Define MCPR
Describe the effects of subcooling on critical power
Describe the effects of mass flow on critical power
Describe the effects of pressure on critical power
Describe the effects of local power distribution on critical
power
Describe the effects of axial power distribution on critical
power
6.2-12
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293009 Core Thermal Limits (continued)
K/A NO.
KNOWLEDGE
K1.27
K1.28
IMPORTANCE
RO SRO
Explain the purpose of the flow biasing correlation factor, (K ), as it
relates to MCPR limits.
2.7
3.3
Define FLCPR
3.0
3.5
Thermal Time Constant
K1.29
K1.30
Define fuel thermal time constant.
2.4
Relate thermal time constant to transient operating condition. 2.3
2.7
2.7
Pellet Clad Interaction
K1.31
K1.32
K1.33
K1.34
K1.35
Describe pellet clad interaction
List the causes of PCI.
Describe the purpose of the pellet to clad gap.
Identify the possible effects of fuel densification.
Describe the effects of iodine and cadmium on pellet clad
interaction.
3.0
2.9
2.4
2.3
3.4
3.3
2.8
2.6
2.2
2.6
2.8
3.4
2.6
3.3
2.7
3.1
PCIOMR
K1.36
K1.37
K1.38
Explain the purpose for PCIOMR
Identify how the PCIOMR rules minimize the adverse effects
of pellet clad interaction.
State the items measured for each of the three core thermal
limits.
For the following plant operating or accident conditions, identify which of
the three core thermal limits are most limiting:
K1.39
K1.40
K1.41
K1.42
K1.43
Full power operation
Loss of reactor coolant
Increase in core flow
Increase in reactor pressure
Cold water addition
6.2-13
2.8
2.8
2.8
2.8
2.9
3.2
3.3
3.3
3.3
3.4
NUREG-2104
6.2
Thermodynamics Theory
(CFR: 41.14)
Thermodynamics:
293010 Brittle Fracture and Vessel Thermal Stress
K/A NO.
KNOWLEDGE
K1.01
K1.02
K1.03
K1.04
State the brittle fracture mode of failure.
State the definition of Nil-Ductility Transition Temperature.
Define reference temperature.
State how the possibility of brittle fracture is minimized by
operating limitations.
State the effect of fast neutron irradiation on reactor vessel
metals.
K1.05
NUREG-2104
IMPORTANCE
6.2-14
RO SRO
2.4
2.8
2.2
2.7
2.0
2.5
2.9
3.2
2.5
2.8
NRC FORM 335
U.S. NUCLEAR REGULATORY COMMISSION
1. REPORT NUMBER
(Assigned by NRC, Add Vol., Supp., Rev.,
and Addendum Numbers, if any.)
(12-2010)
NRCMD 3.7
BIBLIOGRAPHIC DATA SHEET
(See instructions on the reverse)
2. TITLE AND SUBTITLE
Knowledge and Abilities Catalog for Nuclear Power Plant Operators:
Advanced Boiling Water Reactors
NUREG-2104
3. DATE REPORT PUBLISHED
MONTH
YEAR
December
2011
4. FIN OR GRANT NUMBER
5. AUTHOR(S)
R. Pelton
6. TYPE OF REPORT
Technical
7. PERIOD COVERED (Inclusive Dates)
8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,
provide name and mailing address.)
Division of Construction Inspection and Operational Programs
Office of New Reactors
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,
and mailing address.)
Same as above
10. SUPPLEMENTARY NOTES
11. ABSTRACT (200 words or less)
This catalog provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and
senior reactor operators (SROs). The examinations developed using this Catalog will sample the topics listed under Title 10,
Code of Federal Regulations , Part 55 (10 CFR 55). This is a new Knowledge and Abilities catalog developed specifically to
address the General Electric Advanced Boiling Water Reactor.
This catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant
Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components and Theory.
12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)
operator licensing
examinations
reactor operator
senior reactor operator
training
task analysis
advanced boiling water reactors
13. AVAILABILITY STATEMENT
unlimited
14. SECURITY CLASSIFICATION
(This Page)
unclassified
(This Report)
unclassified
15. NUMBER OF PAGES
16. PRICE
NRC FORM 335 (12-2010)
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, DC 20555-0001
-------------------OFFICIAL BUSINESS
NUREG-2104
Draft Report
Knowledge and Abilities Catalog for Nuclear Power Plant Operators
Advanced Boiling Water Reactors
December 2011