STP 3 & 4

STP 3 & 4
Rev. 0
15 Sept 2007
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
STD DEP T1 2.14-1 (Figure 15A-7)
The hydrogen recombiner requirements elimination was provided in ABWR Licensing Topical
Report NEDE-33330P “Hydrogen Recombiner Requirements Elimination,” dated May 18,
2007. Figure 15A-7 is incorporated by reference from the LTR.
15A.6.2.3.11 Control Rod Worth Control
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
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
Rev. 0
15 Sept 2007
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
CONTROL ROD
DRIVE SYSTEM
HIGH PRESSURE
LIFTS VALVE
TRANSFERRING
HEAT TO
SUPPRESSION
POOL
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
15A-2
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 0
15 Sept 2007
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
Plant Nuclear Safety Operational Analysis (NSOA)
15A-3
Rev. 0
15 Sept 2007
STP 3 & 4
Final Safety Analysis Report
EVENT 14
LOSS OF A FEEDWATER
FLOW
STATES C AND D
STATE D
REACTOR
PROTECTION
SYSTEM
SCRAM
ON LOW
LEVEL
CONTINUED ON
FIGURE 15A-64
S F
CONTROL
ROD DRIVE
SYSTEM
S F
SCRAM
Figure 15A-19 Protection Sequence for Loss of All Feedwater Flow
15A-4
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 0
15 Sept 2007
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
Plant Nuclear Safety Operational Analysis (NSOA)
15A-5
15A-6
F
F
F
SCRAM
S
CONTROL ROD
DRIVE SYSTEM
S
REACTOR
PROTECTION
SYSTEM
S
NEUTRON
MONITORING
SYSTEM
BELOW
40% POWER
F
S
F
REACTIVITY
CONTROL
S
TRIP OF
FOUR RIPs
F
PRESSURE
RELIEF
≤ 85% OPEN
S
PRESSURE
RELIEF
SYSTEM
CLOSE ON LOW
CONDENSOR
VACUUM
OPEN ON
TURBINE
TRIP
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
F
VESSEL
ISOLATION
STEAM
BYPASS
SYSTEM
S
STEAM
BYPASS
SYSTEM
S
F
MAIN STEAM LINE
ISOLATION
VALVES
Figure 15A-25 Protection Sequences for Loss of Main Condenser Vacuum
INSERT
CONTROL
RODS
F
TURBINE STOP
VALVE CLOSURE
ABOVE
40% POWER
CONTINUED ON
FIGURE 15A-64
SCRAM SIGNAL ON
NEUTRON MONITOR
SYSTEM TRIP OR
TURBINE STOP
VALVE CLOSURE
HIGH
NEUTRON
FLUX
S
MAIN
TURBINE
TRIP
STATE D ONLY
EVENT 20
LOSS OF MAIN
CONDENSER VACUUM
STATES C AND D
CLOSE ON LOW
CONDENSER
VACUUM
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
F
S
F
F
REACTOR
VESSEL
ISOLATION
S
F
STATES C, D
P > 9.5 kg/cm2g
RESTORE
AC POWER
FAST BUS
TRANSFER
PLANNED OPERATION
SHUTDOWN COOLING
CONTINUED ON
FIGURE 15A-64
ELECTRICAL
SYSTEM
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
F
PRESSURE
RELIEF
S
PRESSURE RELIEF
SYSTEM
INITIATE MAIN STEAM
LINE ISOLATION ON
LOW CONDENSER
VACUUM
F
MAIN STEAM LINE
ISOLATION VALVES
S
LEAK DETECTION
AND ISOLATION
SYSTEM
S
SBPC
STATES C, D
P < 9.5 kg/cm2g
Figure 15A-27 Protection Sequence for Loss of Normal AC Power—Auxiliary Transformer Failure
REACTIVITY
CONTROL
INSERT CONTROL
RODS
SCRAM
S
F
TRIP OF
FOUR RIPs
S
MAIN
TURBINE
TRIP
INITIATE SCRAM ON
GENERATOR TRIP OR
TURBINE TRIP
F
CONTROL ROD
DRIVE SYSTEM
S
REACTOR
PROTECTION
SYSTEM
STATE D
EVENT 22
LOSS OF UNIT AUXILIARY
TRANSFORMER STATES
C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-7
15A-8
SCRAM
SCRAM
CRD
S F
RPS
S F
INSERT
CONTROL
RODS
INITIATE SCRAM
ON HIGH
SUPPRESSION
POOL
TEMPERATURE
S F
CRD
S F
RPS
AUTOMATIC
OR
MANUAL
SCRAM
STATE D
FEEDWATER
OPERATING
STATES C, D
CONTAINMENT
(SP, WW DW)
COOLING
S F
RHR
SUPPRESSION
POOL COOLING
AUTOMATIC
ON 35°C SP
TEMPERATURE
OR MANUAL
Figure 15A-29 Protection Sequences for Inadvertent Opening of a Safety Relief Valve
INSERT
CONTROL
RODS
INITIATE SCRAM
ON LOW WATER
LEVEL
S F
SYSTEM
NUCEAR BOILER
CONTINUED ON
FIGURE 15A-64
NO FEEDWATER
EVENT 24
INADVERTENT OPENING OF
A SAFETY RELIEF VALVE
STATES C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
F
F
VESSEL AND
PRIMARY
CONTAINMENT
ISOLATION
PRIMARY
CONTAINMENT
(PASSIVE)
S
F
F
F
SECONDARY
CONTAINMENT
ISOLATION
OFF GAS VENT
SYSTEM (PASSIVE)
S
STANDBY GAS
TREATMENT
SYSTEM
REACTOR
BUILDING VENT
(PASSIVE)
S
REACTOR BUILDING
ISOLATION
CONTROL SYSTEMS
S
LEAK DETECTION
AND ISOLATION
SYSTEM
STOP ROD
EJECTION
CONTROL ROD
HOUSING SUPPORT
(PASSIVE)
F
SMALL BREAKS
ONLY
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
F
PRESSURE
RELIEF
S
PRESSURE RELIEF
SYSTEM
RADIATION
MONITORING
INTAKE AIR
CONTROL
ROOM
ENVIRONMENTAL
CONTROL
S
CONTROL ROOM
HEATING
VENTILATING AND
AIR CONDITIONING
SYSTEM
CONTROL ROD DRIVE
HOUSING RUPTURE
CONTINUED ON
FIGURE 15A-38
Figure 15A-37 Protection Sequences for Loss of Coolant Piping Breaks in RCPB—Inside Containment
SCRAM
S
MAIN STEAM LINE
ISOLATION VALVES
F
LEAK DETECTION
AND ISOLATION
SYSTEM
S
SCRAM SIGNAL ON
LOW WATER LEVEL OR
HIGH CONTAINMENT
PRESSURE
F
CONTROL ROD
DRIVE SYSTEM
S
REACTOR
PROTECTION
SYSTEM
STATE D
EVENT 32
LOCA-PIPE BREAK INSIDE
CONTAINMENT STATES
C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-9
15A-10
P
P
P
RHRC
MANUAL
RELIEF
VALVE
OPERATION
P
HPCFC
RHR
LPFLB
EXTENDED
CORE
COOLING
RHR
LPFLC
RHR
LPFLA
RHR
LPFLB
RHR
LPFLC
HPCFC
RHR
LPFLA
EXTENDED
CORE
COOLING
MANUAL
RELIEF
VALVE
OPERATION
P
RHR
LPFLC
ADS8
RHR
LPFLB
ADS3
INITIAL
CORE
COOLING
HPCFB
MANUAL
RELIEF
VALVE
OPERATION
P
RCIC
< 6.5 cm2
OTHER SMALL BREAKS
LARGE BREAKS
MANUAL
RELIEF
VALVE
OPERATION
P
RHR
LPFLA
RHR
LPFLB
EXTENDED
CORE
COOLING
ADS3
INITIAL
CORE
COOLING
RCIC
REMAINING
HPCF
B OR C
RHR
LPFLC
MANUAL
RELIEF
VALVE
OPERATION
P
INITIATE ECCS ON LOW
WATER LEVEL OR HIGH
DRYWELL PRESSURE
SMALL BREAKS (< 279 cm2)
HPCF LINE BREAK
< 6.5 cm2
MAIN STEAM
INSTRUMENTATION
SYSTEM
ADS8
Figure 15A-38 Protection Sequence for Loss of Coolant Piping Breaks in RCPB – Inside Primary Containment
RHR
LPFLA
INITIAL
CORE
COOLING
HPCFB
SUPPRESSION
POOL TEMP
LIMIT TO
START VESSEL
DEPRESSURIZATION
PRIMARY
CONTAINMENT
COOLING
ADS
L
S F
SUPPRESSION
POOL COOLING,
WETWELL AND
DRYWELL
SPRAYS
RHRB
RHRA
CONTINUED
FROM
FIGURE
15A-37
EVENT 32
EVENT 32
LOCA-PIP BREAK INSIDE
PRIMARY CONTAINMENT
STATES C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
FLOW RESTRICTORS
(PASSIVE)
6. FEEDWATER FLOW
7. HOT WELL LEVEL
8. VISUAL INSPECTION
9. LEAKAGE INDICATIONS
(2) VARIOUS INDICATIONS:
1. FEED SIGNALS TO PUMPS
2. FEED TEMPERATURE
3. SPACE TEMPERATURE
4. FLOW INDICATIONS
5. REACTOR VESSEL WATER FLOW
S
SCRAM
CONTROL ROD
DRIVE SYSTEM
SCRAM SIGNAL ON
LOW WATER LEVEL OR
MAIN STEAM LINE
ISOLATION
F
S
F
REACTOR
VESSEL
ISOLATION
S
MAIN STEAM LINE
ISOLATION VALVES
ISOLATE ON VARIOUS
INDICATIONS (2)
F
LEAK DETECTION
AND ISOLATION
SYSTEM
SMALL BREAKS
ISOLATE ON LOW
WATER LEVEL HIGH
FLOW OR HIGH AREA
TEMPERATURE
F
LEAK DETECTION
AND ISOLATION
SYSTEM
LARGE BREAKS
CONTINUED ON
FIGURE 15A-40
Figure 15A-39 Protection Sequences for Liquid and Steam, Large and Small Piping Breaks Outside Containment
5. RCIC STEAM LINE
6. HPCF LINE
7. BOTTOM HEAD DRAIN
CONTROL ROOM
ENVIRONMENTAL
S
REACTOR
PROTECTION
SYSTEM
RADIATION
MONITORING
INTAKE AIR
CONTROL ROOM
HEATING,
VENTILATING, AND
AIR CONDITIONING
SYSTEM
S F
(1) LOCA PIPE BREAKS CONSIDERED:
1. REACTOR CLEANUP SYSTEM
2. RHR/SHUTDOWN COOLING
3. MAIN STEAM LINE
4. FEEDWATER LINE
RESTRICT LOSS
OF REACTOR
COOLANT
(PASSIVE)
TRANSFER DECAY
HEAT TO
SUPPRESSION POOL
F
PRESSURE
RELIEF
S
PRESSURE RELIEF
SYSTEM
STATE D ONLY
EVENT 33
LOCA (1) OUTSIDE
CONTAINMENT STATES
C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-11
15A-12
S F
P
ADS
SUPPRESSION POOL
TEMPERATURE LIMIT
TO START VESSEL
DEPRESSURIZATION
RHRC
PRIMARY
CONTAINMENT
COOLING
MANUAL
RELIEF VALVE
P OPERATION
L
SUPPRESSION POOL
COOLING, WETWELL
AND DRYWELL
SPRAYS
P
RHRB
HPCFB
ADS
RHR
LPFLA
RHR
LPFLB
EXTENDED
CORE
COOLING
RCIC
MANUAL
RELIEF VALVE
P OPERATION
INITIAL
CORE
COOLING
HPCFC
MAIN STEAM
INSTRUMENTATION
SYSTEM
MANUAL
RELIEF VALVE
P OPERATION
PLANNED OPERATION
WITH RHR –
SHUTDOWN COOLING
RHR
LPFLC
ADS
INITIATE
ECCS ON
LOW WATER
LEVEL
Figure 15A-40
Protection Sequence for Liquid and Steam, Large and Small Piping Breaks Outside Primary Containment
P
RHRA
CONTINUED FROM
FIGURE 15A-39
EVENT 33
EVENT 33
LOCA-PIPE BREAK OUTSIDE
PRIMARY CONTAINMENT
STATES C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 0
15 Sept 2007
STP 3 & 4
Final Safety Analysis Report
EVENT 41
TRIP OF ALL RIPs
STATES C AND D
STATE D
HIGH WATER
LEVEL L-8 NOT
REACHED
PLANNED
OPERATION
S CRA M ON
LOW CORE
FLOW
HIGH 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)
Plant Nuclear Safety Operational Analysis (NSOA)
15A-13
15A-14
INSERT
CONTROL
RODS
F
REACTIVITY
CONTROL
S
F
CONTROL ROD
DRIVE SYSTEM
S
S
F
TRIP OF
FOUR RIPs
TURBINE STOP
VALVE CLOSURE
HIGH LEVEL TURBINE
TRIP AND FEEDWATER
TRIP
P ≥ *40%
SCRAM SIGNAL FROM
TURBINE TRIP (RUN
MODE) OR NEUTRON
MONITORING SYSTEM
F
HIGH FLUX
SCRAM
SIGNAL
F
REACTOR
PROTECTION
SYSTEM
S
NEUTRON
MONITORING
SYSTEM
P < *40%
S
NUCLEAR BOILER
INSTRUMENTATION
BYPASS
VALVES
S F
STEAM BYPASS
AND PRESSURE
CONTROL
STATES C, D
CONTINUED ON
FIGURE 15A-64
STOP VALVES
POSITION
SCRAM
STATE D
PRESSURE
RELIEF
S F
PRESSURE
RELIEF
SYSTEM
Figure 15A-51 Protection Sequences for Feedwater Controller Failure—Runout of Two Feedwater Pumps
PLANNED
OPERATION
STATES A, B
~
EVENT 44
FEEDWATER CONTROLLER
FAILURE – RUNOUT OF TWO
FEEDWATER PUMPS
STATES A, B, C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
F
HIGH
WATER
LEVEL
L-8
S
F
PRESSURE
RELIEF
S
PRESSURE RELIEF
SYSTEM
F
F
SCRAM
S
CONTROL ROD
DRIVE SYSTEM
S
REACTOR
PROTECTION
SYSTEM
STATE D
INSERT
CONTROL
RODS
SCRAM SIGNAL FROM
• MAIN STEAM LINE
ISOLATION
• TURBINE TRIP (RUN
MODE: POWER > 40%)
• HIGH PRESSURE
• LOW WATER LEVEL
CONTINUED ON
FIGURE 15A-64
F
F
REACTOR
VESSEL
ISOLATION
S
MAIN STEAM LINE
ISOLATION VALVES
S
LEAK DETECTION
AND ISOLATION
SYSTEM
SCRAM
SIGNAL
INITIATE ISOLATION
ON:
1. DEPRESSURIZATION
TO 5.2 MPaG
(RUN MODE:
POWER 0 – 100%)
Figure 15A-52 Protection Sequences for Pressure Regulator Failure—Opening of All Bypass and Control Valves
REACTIVITY
CONTROL
TRIP OF
FOUR RIPs
TURBINE TRIP
MAIN STEAM
INSTRUMENTATION
SYSTEM
STATE D
TURBINE
BYPASS
SYSTEM
EVENT 45
PRESSURE REGULATOR
FAILURE – OPENING OF ALL BYPASS
AND CONTROL VALVES
STATES C AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-15
Rev. 0
15 Sept 2007
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
F
PRESSURE
RELIEF
MAIN STEAM
INSTRUMENTATION
SYSTEM
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
15A-16
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
CORE
REACTIVITY
CONTROL
S F
SLCS
RCIC
RHRS
SUPPRESSION
POOL COOLING
P
MODE
HPCFB
EXTENDED
CORE
COOLING
RHRS
SHUTDOWN
COOLING
P MODE
REMOVE DECAY
HEAT FROM
SUPPRESSION
POOL
HPCFC
START HPCF,
RCIC SYSTEM
ON LOW
WATER LEVEL
WHEN PRESSURE
< SDC INTERLOCK
TRANSFER
HEAT TO
SUPPRESSION
POOL
< SDC INTERLOCK
STATES B, D
> SDC INTERLOCK
REACTOR ISOLATED
FROM MAIN CONDENSER
NUCLEAR BOILER
INSTRUMENTATION
SYSTEM
MAINTAIN
WATER
LEVEL
PLANNED OPERATION
CONTROL COOLDOWN
USING NORMAL
EQUIPMENT
REACTOR NOT ISOLATED
FROM MAIN CONDENSER
PRESSURE
RELIEF
S F
PRESSURE
RELIEF
SYSTEM
STATE D
Figure 15A-63 Protection Sequence for Reactor Shutdown—Without Control Rods
P
PLANNED
OPERATIONS
CONTINUE
SHUTDOWN
COOLING
STATES
B, D
EVENT 56
REACTOR SHUTDOWN
WITHOUT CONTROL RODS
STATES B AND D
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-17
Rev. 0
15 Sept 2007
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
15A-18
Plant Nuclear Safety Operational Analysis (NSOA)
Plant Nuclear Safety Operational Analysis (NSOA)
HPCF
COOL
RHRS
EQUIPMENT
AREA
RHR
SHUTDOWN
COOLING
MODE
STATE EVENTS
A
B
*
C
D
RCIC
LEAK
DETECTION
AND
ISOLATION
SYSTEM
STATE EVENTS
A
B
*
C
D
RHRSLPFL
MODE
ADS
S F
HPCF
HPCF
STATE EVENTS
A
B
*
C
D
COOL
RHRS
EQUIPMENT
AREA
RHR
SUPPRESSION
POOL COOLING
MODE
STATE EVENTS
A
B
*
C
D
RCIC
CONTINUED
ON FIGURE
15A-68
Figure 15A-67 Commonality of Auxiliary Systems—Reactor Building Cooling Water System (RCWS)
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
NOTE: SF REQUIREMENT NOT APPLICABLE IN EVENTS 54, 55 AND 56
RCICS
HPCF
S F
S F
RHRS
LPFL
MODE
STATE EVENTS
A
B
*
C
D
STATE EVENTS
A
B
*
C
D
REACTOR
BUILDING
COOLING
WATER
SYSTEM (RCWS)
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
15A-19
15A-20
EMERGENCY
DIESEL
GENERATOR
ROOM COOLERS
CONTINUED
ON FIGURE
15A-69
Figure 15A-68 Commonality of Auxiliary Systems—Reactor Building Cooling Water System (RCWS) (Continued)
MAIN CONTROL
ROOM COOLERS
HVAC EMERGENCY
CHILLED WATER
SYSTEM (HECW)
*
EVENTS
STANDBY AC
POWER SYSTEM
STATE
A
B
C
D
S F
*
EVENTS
S F
STATE
A
B
C
D
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
NOTE: SF REQUIREMENT NOT APPLICABLE IN EVENTS 54, 55 AND 56
RHRSSHUTDOWN
COOLING MODE
RHR PUMP
COOLING
*
EVENTS
RHRSSUPPRESSION POOL
COOLING MODE
STATE
A
B
C
D
S F
*
EVENTS
S F
*
EVENTS
RHRS LPFL
STATE
A
B
C
D
S F
STATE
A
B
C
D
CONTINUED
FROM
FIGURE
15A-67
REACTOR BUILDING
COOLING WATER
SYSTEM (RCWS)
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)
Rev. 0
15 Sept 2007
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)
Plant Nuclear Safety Operational Analysis (NSOA)
15A-21
15A-22
RCIC
HPCF
RCIC
STATE EVENTS
A
B
*
C
D
PRESSURE
RELIEF
SYSTEM
STATE EVENTS
A
B
*
C
D
RHR
LPFL
HPCF
RCIC
STATE EVENTS
A
B
*
C
D
RHR
LPFL
MODE
ADS
HPCF
HPCF
RCIC
STATE EVENTS
A
B
*
C
D
MANUAL
RELIEF VALVE
SYSTEM
OPERATION
STATE EVENTS
A
B
*
C
D
CONTAINMENT
(PASSIVE)
STATE EVENTS
A
B
*
C
D
Figure 15A-70 Commonality of Auxiliary Systems—Suppression Pool Storage
* APPLICABLE EVENTS (TABLES 15A-2 THROUGH 15A-5)
HPCF
MAIN STEAM
INSTRUMENTATION
SYSTEM
STATE EVENTS
A
B
*
C
D
SUPPRESSION
POOL
STORAGE
(PASSIVE)
STP 3 & 4
Rev. 0
15 Sept 2007
Final Safety Analysis Report
Plant Nuclear Safety Operational Analysis (NSOA)