STP 3 & 4

Rev. 11
STP 3 & 4
Final Safety Analysis Report
15A Plant Nuclear Safety Operational Analysis (NSOA)
The information in this appendix of the reference ABWR DCD, including all
subsections, tables, and figures, is incorporated by reference with the following
departures.
STD DEP Admin (Figures 15A-13, 17, 19, 21, 25, 27, 29, 37, 38, 39, 40, 48, 51, 52,
53, 63, 64, 67, 68, 69, and 70)
STD DEP T1 2.14-1 (Figure 15A-7)
15A.6.2.3.11 Control Rod Worth Control
STD DEP Admin
Any time the reactor is not shut down and is generating less than 20% power (State
D), a limit is imposed on the control rod pattern to assure that control rod worth is
maintained within the envelope of conditions considered by the analysis of the control
rod drop accident rod withdrawal error (1-4).
15A.6.3.1 General
STD DEP Admin
The safety requirements and protection sequences for moderate frequency incidents
(anticipated operational transients) are described in the following subsections for
Events 7 through 22 23, 26, 27, 38-40, 44, 45, 48, and 49. The protection sequence
block diagrams show the sequence of frontline safety systems (Figures 15A-12
through 15A-27). The auxiliaries for the frontline safety systems are presented in the
auxiliary diagrams (Figures 15A-6 and 15A-7) and the commonality of auxiliary
diagrams (Figures 15A-65 through 15A-70).
Plant Nuclear Safety Operational Analysis (NSOA)
15A-1
STP 3 & 4
15A-2
Rev. 11
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Figure 15A-7 Safety System Auxiliaries—Group 2
Rev. 11
STP 3 & 4
Final Safety Analysis Report
EVENT 8
LOSS OF PLANT INSTRUMENT
OR NITROGEN SUPPLY
SYSTEM
STATES A, B, C, D
STATES A, C
STATES B, D
STATES C, D
PLANNED
OPERATION
SCRAM SIGNAL
WHEN SCRAM
SOLENOIDS
LOSE AIR
CONTINUED ON
FIGURE 15A-64
REACTOR
PROTECTION
SYSTEM
S F
INSERT
CONTROL
RODS
STATES A, B
HIGH PRESSURE
LIFTS VALVE
TRANSFERRING
HEAT TO
SUPPRESSION
POOL
CONTROL ROD
DRIVE SYSTEM
PRESSURE
RELIEF
SYSTEM
RHR
SUPPRESSION
POOL COOLING
S F
S F
PRESSURE
RELIEF
EXTENDED
CORE COOLING
S F
SCRAM
Figure 15A-13 Protection Sequence for Loss of Plant
Instrument or Service Air System
Plant Nuclear Safety Operational Analysis (NSOA)
15A-3
Rev. 11
STP 3 & 4
Final Safety Analysis Report
EVENT 12
ISOLATION OF ALL MAIN
STEAM LINES
STATES C AND D
STATE D
STATES C, D
STATE C
PLANNED
OPERATION
SHUTDOWN
COOLING
CONTINUED ON
FIGURE 15A-64
REACTOR
PROTECTION
SYSTEM
SCRAM SIGNAL
WHEN 3 MAIN
STEAM LINES
CLOSED ≥ 15%
S F
CONTROL ROD
DRIVE SYSTEM
INSERT
CONTROL
RODS
PRESSURE
RELIEF
SYSTEM
RECIRCULATION
PUMP TRIP
S F
S F
PRESSURE
RELIEF
REACTIVITY
CONTROL
S F
SCRAM
HIGH PRESSURE
LIFTS VALVE
TRANSFERRING
HEAT TO
SUPPRESSION
POOL
HIGH
PRESSURE
TRIP
Figure 15A-17 Protection Sequences for Isolation of All Main Steamlines
15A-4
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 11
STP 3 & 4
Final Safety Analysis Report
Figure 15A-19 Protection Sequence for Loss of All Feedwater Flow
Plant Nuclear Safety Operational Analysis (NSOA)
15A-5
Rev. 11
STP 3 & 4
Final Safety Analysis Report
EVENT 16
FEEDWATER CONTROLLER
FAILURE – RUNOUT OF ONE
FEEDWATER PUMP
STATES C AND D
NO
MAIN STEAM
LEVEL 8 TRIP
SYSTEM
S F
TURBINE
TRIP
CONTINUED ON
FIGURE 15A-24
OTHER
FEEDWATER
PUMP IN
OPERATION
YES
FEEDWATER
CONTROL
SYSTEM
REDUCES
OTHER
FEEDPUMP
FLOW
S F
OPERATOR
CONTROLS FEED
FLOW
P
ACCEPTABLE
STEADY STATE
OPERATION
Figure 15A-21 Protection Sequence for Feedwater Controller Failure—Runout of One
Feedwater Pump
15A-6
Plant Nuclear Safety Operational Analysis (NSOA)
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
EVENT 20
LOSS OF MAIN
CONDENSER VACUUM
STATES C AND D
STATE D ONLY
MAIN
TURBINE
TRIP
S
S
F
CONTROL ROD
DRIVE SYSTEM
S
F
HIGH
NEUTRON
FLUX
TURBINE STOP
VALVE CLOSURE
S
CLOSE ON LOW
CONDENSER
VACUUM
F
OPEN ON
TURBINE
TRIP
STEAM
BYPASS
SYSTEM
S
CLOSE ON LOW
CONDENSOR
VACUUM
TRIP OF
FOUR RIPs
S
F
≤ 85% OPEN
SCRAM SIGNAL ON
NEUTRON MONITOR
SYSTEM TRIP OR
TURBINE STOP
VALVE CLOSURE
INSERT
CONTROL
RODS
S
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
F
STEAM
BYPASS
SYSTEM
F
REACTIVITY
CONTROL
PRESSURE
RELIEF
VESSEL
ISOLATION
Figure 15A-25 Protection Sequences for Loss of Main Condenser Vacuum
15A-7
Final Safety Analysis Report
SCRAM
F
MAIN STEAM LINE
ISOLATION
VALVES
Rev. 11
S
S
ABOVE
40% POWER
F
REACTOR
PROTECTION
SYSTEM
PRESSURE
RELIEF
SYSTEM
F
BELOW
40% POWER
NEUTRON
MONITORING
SYSTEM
CONTINUED ON
FIGURE 15A-64
STP 3 & 4
15A-8
EVENT 22
LOSS OF UNIT AUXILIARY
TRANSFORMER STATES
C AND D
STATE D
STATES C, D
P < 9.5 kg/cm2g
SBPC
S
REACTOR
PROTECTION
SYSTEM
S
MAIN
TURBINE
TRIP
F
S
F
LEAK DETECTION
AND ISOLATION
SYSTEM
S
TRIP OF
FOUR RIPs
F
F
S
S
F
MAIN STEAM LINE
ISOLATION VALVES
S
F
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
PRESSURE
RELIEF
FAST BUS
TRANSFER
RESTORE
AC POWER
F
INSERT CONTROL
RODS
SCRAM
REACTIVITY
CONTROL
Rev. 11
S
PRESSURE RELIEF
SYSTEM
PLANNED OPERATION
SHUTDOWN COOLING
CONTINUED ON
FIGURE 15A-64
ELECTRICAL
SYSTEM
REACTOR
VESSEL
ISOLATION
Figure 15A-27 Protection Sequence for Loss of Normal AC Power—Auxiliary Transformer Failure
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
CONTROL ROD
DRIVE SYSTEM
F
INITIATE MAIN STEAM
LINE ISOLATION ON
LOW CONDENSER
VACUUM
INITIATE SCRAM ON
GENERATOR TRIP OR
TURBINE TRIP
STATES C, D
P > 9.5 kg/cm2g
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 11
Final Safety Analysis Report
15A-9
Figure 15A-29 Protection Sequences for Inadvertent Opening of a Safety Relief Valve
CONTINUED ON
FIGURE 15A-38
STATE D
CONTROL ROD DRIVE
HOUSING RUPTURE
REACTOR
PROTECTION
SYSTEM
S
LEAK DETECTION
AND ISOLATION
SYSTEM
S
SCRAM SIGNAL ON
LOW WATER LEVEL OR
HIGH CONTAINMENT
PRESSURE
F
F
LEAK DETECTION
AND ISOLATION
SYSTEM
S
F
RADIATION
MONITORING
INTAKE AIR
REACTOR
BUILDING VENT
(PASSIVE)
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
STOP ROD
EJECTION
CONTROL
ROOM
ENVIRONMENTAL
CONTROL
PRESSURE
RELIEF
STANDBY GAS
TREATMENT
SYSTEM
S
VESSEL AND
PRIMARY
CONTAINMENT
ISOLATION
F
F
F
PRIMARY
CONTAINMENT
(PASSIVE)
S
F
F
OFF GAS VENT
SYSTEM (PASSIVE)
SECONDARY
CONTAINMENT
ISOLATION
Figure 15A-37 Protection Sequences for Loss of Coolant Piping Breaks in RCPB—Inside Containment
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
SCRAM
S
PRESSURE RELIEF
SYSTEM
Rev. 11
S
MAIN STEAM LINE
ISOLATION VALVES
SMALL BREAKS
ONLY
CONTROL ROOM
HEATING
VENTILATING AND
AIR CONDITIONING
SYSTEM
S
F
REACTOR BUILDING
ISOLATION
CONTROL SYSTEMS
S
CONTROL ROD
DRIVE SYSTEM
CONTROL ROD
HOUSING SUPPORT
(PASSIVE)
STP 3 & 4
15A-10
EVENT 32
LOCA-PIPE BREAK INSIDE
CONTAINMENT STATES
C AND D
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
EVENT 32
LOCA-PIP BREAK INSIDE
PRIMARY CONTAINMENT
STATES C AND D
CONTINUED
FROM
FIGURE
15A-37
EVENT 32
MAIN STEAM
INSTRUMENTATION
SYSTEM
INITIATE ECCS ON LOW
WATER LEVEL OR HIGH
DRYWELL PRESSURE
LARGE BREAKS
OTHER SMALL BREAKS
< 6.5 cm2
RHRA
RHRB
RHRC
P
P
P
HPCFC
INITIAL
CORE
COOLING
L
MANUAL
RELIEF
VALVE
OPERATION
P
15A-11
PRIMARY
CONTAINMENT
COOLING
RCIC
HPCFB
HPCFC
ADS8
MANUAL
RELIEF
VALVE
OPERATION
P
RHR
LPFLC
EXTENDED
CORE
COOLING
MANUAL
RELIEF
VALVE
OPERATION
P
RHR
LPFLA
RCIC
REMAINING
HPCF
B OR C
MANUAL
RELIEF
VALVE
OPERATION
P
ADS8
INITIAL
CORE
COOLING
MANUAL
RELIEF
VALVE
OPERATION
P
ADS3
RHR
LPFLB
EXTENDED
CORE
COOLING
RHR
LPFLC
ADS3
RHR
LPFLA
RHR
LPFLB
RHR
LPFLC
EXTENDED
CORE
COOLING
Figure 15A-38 Protection Sequence for Loss of Coolant Piping Breaks in RCPB – Inside Primary Containment
Final Safety Analysis Report
ADS
SUPPRESSION
POOL TEMP
LIMIT TO
START VESSEL
DEPRESSURIZATION
RHR
LPFLB
RHR
LPFLC
INITIAL
CORE
COOLING
S F
RHR
LPFLA
RHR
LPFLB
Rev. 11
SUPPRESSION
POOL COOLING,
WETWELL AND
DRYWELL
SPRAYS
HPCFB
RHR
LPFLA
SMALL BREAKS (< 279 cm2)
HPCF LINE BREAK
< 6.5 cm2
CONTINUED ON
FIGURE 15A-40
STATE D ONLY
LARGE BREAKS
PRESSURE RELIEF
SYSTEM
S
FLOW RESTRICTORS
(PASSIVE)
F
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
REACTOR
PROTECTION
SYSTEM
S
RADIATION
MONITORING
INTAKE AIR
CONTROL ROOM
ENVIRONMENTAL
(1) LOCA PIPE BREAKS CONSIDERED:
1. REACTOR CLEANUP SYSTEM
2. RHR/SHUTDOWN COOLING
3. MAIN STEAM LINE
4. FEEDWATER LINE
5. RCIC STEAM LINE
6. HPCF LINE
7. BOTTOM HEAD DRAIN
(2) VARIOUS INDICATIONS:
1. FEED SIGNALS TO PUMPS
2. FEED TEMPERATURE
3. SPACE TEMPERATURE
4. FLOW INDICATIONS
5. REACTOR VESSEL WATER FLOW
6. FEEDWATER FLOW
7. HOT WELL LEVEL
8. VISUAL INSPECTION
9. LEAKAGE INDICATIONS
LEAK DETECTION
AND ISOLATION
SYSTEM
S
SCRAM SIGNAL ON
LOW WATER LEVEL OR
MAIN STEAM LINE
ISOLATION
CONTROL ROD
DRIVE SYSTEM
LEAK DETECTION
AND ISOLATION
SYSTEM
F
S
ISOLATE ON LOW
WATER LEVEL HIGH
FLOW OR HIGH AREA
TEMPERATURE
ISOLATE ON VARIOUS
INDICATIONS (2)
MAIN STEAM LINE
ISOLATION VALVES
S
SCRAM
F
F
REACTOR
VESSEL
ISOLATION
Figure 15A-39 Protection Sequences for Liquid and Steam, Large and Small Piping Breaks Outside Containment
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
RESTRICT LOSS
OF REACTOR
COOLANT
(PASSIVE)
F
SMALL BREAKS
Rev. 11
PRESSURE
RELIEF
CONTROL ROOM
HEATING,
VENTILATING, AND
AIR CONDITIONING
SYSTEM
S F
STP 3 & 4
15A-12
EVENT 33
LOCA (1) OUTSIDE
CONTAINMENT STATES
C AND D
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
EVENT 33
LOCA-PIPE BREAK OUTSIDE
PRIMARY CONTAINMENT
STATES C AND D
CONTINUED FROM
FIGURE 15A-39
EVENT 33
MAIN STEAM
INSTRUMENTATION
SYSTEM
RHRA
RHRB
P
RHRC
P
HPCFB
HPCFC
INITIATE
ECCS ON
LOW WATER
LEVEL
ADS
RCIC
P
MANUAL
RELIEF VALVE
P OPERATION
S F
L
ADS
SUPPRESSION POOL
TEMPERATURE LIMIT
TO START VESSEL
DEPRESSURIZATION
PRIMARY
CONTAINMENT
COOLING
MANUAL
RELIEF VALVE
P OPERATION
RHR
LPFLA
RHR
LPFLB
EXTENDED
CORE
COOLING
RHR
LPFLC
PLANNED OPERATION
WITH RHR –
SHUTDOWN COOLING
Figure 15A-40 Protection Sequence for Liquid and Steam, Large and Small Piping Breaks Outside Primary Containment
15A-13
Final Safety Analysis Report
MANUAL
RELIEF VALVE
P OPERATION
ADS
Rev. 11
INITIAL
CORE
COOLING
SUPPRESSION POOL
COOLING, WETWELL
AND DRYWELL
SPRAYS
Rev. 11
STP 3 & 4
Final Safety Analysis Report
EVENT 41
TRIP OF ALL RIPs
STATES C AND D
STATE D
H IGH WATER
LEVEL L-8 NOT
REACHED
PLANNED
OPERATION
S C R A M ON
LOW CORE
FLOW
H IGH WATER
LEVEL L-8
REACHED
RECIRCULATION
FLOW CONTROL
SYSTEM
OPERATION
IN STATE C
S F
SCRAM
SIGNAL
REACTOR
PROTECTION
SYSTEM (RPS)
S F
L-8
TRIP OF
FEEDWATER
PUMPS
FEEDWATER
CONTROL
SYSTEM
CONTROL ROD
DRIVES (CRD)
S F
S F
CONTINUED ON
FIGURE 15A-64
SCRAM
TURBINE
TRIP
S F
PRESSURE
RELIEF
SYSTEM
S F
PRESSURE
RELIEF
Figure 15A-48 Protection Sequence for Trip of All Reactor Internal Pumps (RIPs)
15A-14
Plant Nuclear Safety Operational Analysis (NSOA)
STP 3 & 4
STATES A, B
STATE D
STATES C, D
~
Plant Nuclear Safety Operational Analysis (NSOA)
EVENT 44
FEEDWATER CONTROLLER
FAILURE – RUNOUT OF TWO
FEEDWATER PUMPS
STATES A, B, C AND D
PLANNED
OPERATION
NUCLEAR BOILER
INSTRUMENTATION
S
P < *40%
S
S
F
P ≥ *40%
HIGH FLUX
SCRAM
SIGNAL
TURBINE STOP
VALVE CLOSURE
F
STOP VALVES
POSITION
SCRAM
STEAM BYPASS
AND PRESSURE
CONTROL
S F
SCRAM SIGNAL FROM
TURBINE TRIP (RUN
MODE) OR NEUTRON
MONITORING SYSTEM
PRESSURE
RELIEF
SYSTEM
BYPASS
VALVES
S F
INSERT
CONTROL
RODS
TRIP OF
FOUR RIPs
S
PRESSURE
RELIEF
F
REACTIVITY
CONTROL
15A-15
Figure 15A-51 Protection Sequences for Feedwater Controller Failure—Runout of Two Feedwater Pumps
Final Safety Analysis Report
CONTROL ROD
DRIVE SYSTEM
S
F
F
REACTOR
PROTECTION
SYSTEM
CONTINUED ON
FIGURE 15A-64
Rev. 11
NEUTRON
MONITORING
SYSTEM
HIGH LEVEL TURBINE
TRIP AND FEEDWATER
TRIP
STP 3 & 4
15A-16
EVENT 45
PRESSURE REGULATOR
FAILURE – OPENING OF ALL BYPASS
AND CONTROL VALVES
STATES C AND D
STATE D
TURBINE
BYPASS
SYSTEM
S
STATE D
F
MAIN STEAM
INSTRUMENTATION
SYSTEM
PRESSURE RELIEF
SYSTEM
S
TURBINE TRIP
F
PRESSURE
RELIEF
REACTOR
PROTECTION
SYSTEM
S
F
CONTROL ROD
DRIVE SYSTEM
S
F
SCRAM
SCRAM SIGNAL FROM
• MAIN STEAM LINE
ISOLATION
• TURBINE TRIP (RUN
MODE: POWER > 40%)
• HIGH PRESSURE
• LOW WATER LEVEL
INSERT
CONTROL
RODS
LEAK DETECTION
AND ISOLATION
SYSTEM
S
F
MAIN STEAM LINE
ISOLATION VALVES
S
INITIATE ISOLATION
ON:
1. DEPRESSURIZATION
TO 5.2 MPaG
(RUN MODE:
POWER 0 – 100%)
SCRAM
SIGNAL
F
Rev. 11
TRIP OF
FOUR RIPs
HIGH
WATER
LEVEL
L-8
CONTINUED ON
FIGURE 15A-64
REACTOR
VESSEL
ISOLATION
Figure 15A-52 Protection Sequences for Pressure Regulator Failure—Opening of All Bypass and Control Valves
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
REACTIVITY
CONTROL
Rev. 11
STP 3 & 4
Final Safety Analysis Report
EVENT 46
PRESSURE REGULATOR
FAILURE – CLOSURE OF ALL BYPASS
AND CONTROL VALVES
STATES C AND D
STATE D
STATE D
CONTINUED ON
FIGURE 15A-64
PRESSURE RELIEF
SYSTEM
S
MAIN STEAM
INSTRUMENTATION
SYSTEM
F
PRESSURE
RELIEF
S
F
TRIP OF
FOUR RIPs
S
F
REACTIVITY
CONTROL
HIGH
PRESSURE
TRIP
NEUTRON
MONITORING
SYSTEM
S F
HIGH FLUX
SCRAM
SIGNAL
REACTOR
PROTECTION
SYSTEM
S
F
CONTROL ROD
DRIVE
SYSTEM
S
F
SCRAM
Figure 15A-53 Pressure Regulator Failure—Closure of All Bypass Valves and Control
Valves
Plant Nuclear Safety Operational Analysis (NSOA)
15A-17
STP 3 & 4
15A-18
EVENT 56
REACTOR SHUTDOWN
WITHOUT CONTROL RODS
STATES B AND D
STATE D
PLANNED
OPERATIONS
CONTINUE
SHUTDOWN
COOLING
STATES
B, D
REACTOR NOT ISOLATED
FROM MAIN CONDENSER
REACTOR ISOLATED
FROM MAIN CONDENSER
> SDC INTERLOCK
PLANNED OPERATION
CONTROL COOLDOWN
USING NORMAL
EQUIPMENT
SLCS
< SDC INTERLOCK
STATES B, D
NUCLEAR BOILER
INSTRUMENTATION
SYSTEM
START HPCF,
RCIC SYSTEM
ON LOW
WATER LEVEL
TRANSFER
HEAT TO
SUPPRESSION
POOL
PRESSURE
RELIEF
SYSTEM
S F
PRESSURE
RELIEF
P
S F
RCIC
HPCFB
RHRS
SUPPRESSION
POOL COOLING
P
MODE
HPCFC
REMOVE DECAY
HEAT FROM
SUPPRESSION
POOL
WHEN PRESSURE
< SDC INTERLOCK
EXTENDED
CORE
COOLING
Figure 15A-63 Protection Sequence for Reactor Shutdown—Without Control Rods
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
RHRS
SHUTDOWN
COOLING
P MODE
Rev. 11
CORE
REACTIVITY
CONTROL
MAINTAIN
WATER
LEVEL
Rev. 11
STP 3 & 4
Final Safety Analysis Report
CONTINUED PROTECTION
SEQUENCE FOR CORE AND
CONTAINMENT
NUCLEAR BOILER
INSTRUMENTATION
SYSTEM
NON ESF
MOTOR
FEED PUMP
RCIC
HPCFB
CHPCC
CONDENSATE
PUMPS
INITIAL
CORE
COOLING
RHRA SUPPRESSION
POOL COOLING
RHRB SUPPRESSION
POOL COOLING
RHRC SUPPRESSION
POOL COOLING
EXTENDED CORE
AND
CONTAINMENT
COOLING
PLANNED OPERATION WITH
RHRS SHUTDOWN COOLING
Figure 15A-64 Protection Sequence for Core and Containment Cooling for Loss of
Feedwater and Vessel Isolations
Plant Nuclear Safety Operational Analysis (NSOA)
15A-19
CONTINUED
ON FIGURE
15A-68
STATE EVENTS
A
B
*
C
D
STATE EVENTS
A
B
*
C
D
S F
S F
RCICS
HPCF
RHRS
LPFL
MODE
HPCF
STATE EVENTS
A
B
*
C
D
STATE EVENTS
A
B
*
C
D
STATE EVENTS
A
B
*
C
D
STP 3 & 4
15A-20
REACTOR
BUILDING
COOLING
WATER
SYSTEM (RCWS)
STATE EVENTS
A
B
*
C
D
S F
RHR
SHUTDOWN
COOLING
MODE
RCIC
RHR
SUPPRESSION
POOL COOLING
MODE
HPCF
COOL
RHRS
EQUIPMENT
AREA
ADS
HPCF
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
NOTE: SF REQUIREMENT NOT APPLICABLE IN EVENTS 54, 55 AND 56
Figure 15A-67 Commonality of Auxiliary Systems—Reactor Building Cooling Water System (RCWS)
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
RHRSLPFL
MODE
RCIC
Rev. 11
COOL
RHRS
EQUIPMENT
AREA
LEAK
DETECTION
AND
ISOLATION
SYSTEM
CONTINUED
FROM
FIGURE
15A-67
CONTINUED
ON FIGURE
15A-69
STATE
A
B
C
D
EVENTS
STATE
A
B
C
D
*
S F
RHRS LPFL
EVENTS
STATE
A
B
C
D
*
S F
RHRSSUPPRESSION POOL
COOLING MODE
EVENTS
*
S F
S F
STANDBY AC
POWER SYSTEM
HVAC EMERGENCY
CHILLED WATER
SYSTEM (HECW)
EVENTS
*
S F
RHR PUMP
COOLING
*
STATE
A
B
C
D
Rev. 11
STATE
A
B
C
D
EVENTS
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
REACTOR BUILDING
COOLING WATER
SYSTEM (RCWS)
MAIN CONTROL
ROOM COOLERS
EMERGENCY
DIESEL
GENERATOR
ROOM COOLERS
RHRSSHUTDOWN
COOLING MODE
Figure 15A-68 Commonality of Auxiliary Systems—Reactor Building Cooling Water System (RCWS) (Continued)
15A-21
Final Safety Analysis Report
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
NOTE: SF REQUIREMENT NOT APPLICABLE IN EVENTS 54, 55 AND 56
Rev. 11
STP 3 & 4
Final Safety Analysis Report
REACTOR BUILDING
COOLING WATER
SYSTEM (RCWS)
CONTINUED FROM
FIGURE 15A-68
STATE
A
B
C
D
EVENTS
STATE
A
B
C
D
*
*
S F
S F
RHRSSUPPRESSION POOL
COOLING MODE
EVENTS
COOL RHRS
HEAT EXCHANGERS
RHRS
SHUTDOWN
COOLING MODE
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
NOTE: SF REQUIREMENT NOT APPLICABLE IN EVENTS 54, 55 AND 56
Figure 15A-69 Commonality of Auxiliary Systems—Reactor Building Cooling Water
System (RCWS) (Continued)
15A-22
Plant Nuclear Safety Operational Analysis (NSOA)
STP 3 & 4
Plant Nuclear Safety Operational Analysis (NSOA)
SUPPRESSION
POOL
STORAGE
(PASSIVE)
STATE EVENTS
A
B
*
C
D
MAIN STEAM
INSTRUMENTATION
SYSTEM
RCIC
HPCF
STATE EVENTS
A
B
*
C
D
STATE EVENTS
A
B
*
C
D
RHR
LPFL
RCIC
PRESSURE
RELIEF
SYSTEM
HPCF
RCIC
STATE EVENTS
A
B
*
C
D
ADS
RHR
LPFL
MODE
HPCF
RCIC
STATE EVENTS
A
B
*
C
D
CONTAINMENT
(PASSIVE)
HPCF
STATE EVENTS
A
B
*
C
D
Figure 15A-70 Commonality of Auxiliary Systems—Suppression Pool Storage
15A-23/24
Final Safety Analysis Report
MANUAL
RELIEF VALVE
SYSTEM
OPERATION
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
Rev. 11
HPCF
STATE EVENTS
A
B
*
C
D