ML25290A107
| ML25290A107 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/13/2025 |
| From: | Talen Energy, Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25290A004 | List:
|
| References | |
| PLA-8177 | |
| Download: ML25290A107 (1) | |
Text
SSES-FSAR Text Rev. 56 FSAR Rev. 72 15C-1 APPENDIX 15C SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT -
CYCLE SPECIFIC DATA 15C.1 Appendix C Contents 15C.1.1 Content Discussion This Section presents results that are typical of cycle-specific analyses. Actual cycle-specific results may be found in, or calculated from, Reference 15C.1.2-1.
15C.1.2 REFERENCES 15C.1.2-1 ANP-4079P, Revision 0, Susquehanna Unit 1 Cycle 24 Reload Safety Analysis, Framatome Inc., February 2024 (General Reference per NEI 98-03).
SSES-FSAR Table Rev. 66 FSAR Rev. 71 Page 1 of 4 TABLE 15C.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux
% of Rated Maximum Dome Pressure psig Maximum Vessel Pressure psig Maximum Steam line Pressure psig Maximum Core Average Surface Heat Flux,%
CPR Frequency Category Number of Valves -
1st Blowdown Duration of Blowdown 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 15.1.1 Loss of Feedwater Heater NOTE 5 NOTE 5 NOTE 5 NOTE 5 NOTE 5 0.198 Moderate 0
0 sec 15.1.2 15C.1.2-1 through 15C.1.2-4 Feedwater Controller Failure (100% Power, 108 Mlbm/hr, Max Allowable Scram Time) EOC RPT Operable 314 1259 1286 1260
~135 0.27 -
A10 0.34 -
A11 Moderate 15.1.3 15C.1.3-1 Pressure Regulator Failure - Open 100 1051 1094 1043 100 0.01 -
A10 0.01 -
A11 Moderate 2
See Text 15.1.4 Inadvertent Opening of Safety or Relief Valves See Text Moderate 15.1.6 RHR Shutdown Cooling Malfunction See Text Moderate 15.2 INCREASE IN REACTOR PRESSURE 15.2.1 Pressure Regulator Failure - Closed See Text Moderate 15.2.2 Generator Load Reject - Bypass Operable See Text and Appendix 15E Moderate 15.2.2 15C.2.2-1 through 15C.2.2-4 Generator Load Reject-Without Bypass (100% Power, 108 Mlbm/hr, max allowable Scram Time) EOC RPT Operable 389 1300 1332 1321
~135 0.42 -
A10 0.44 -
A11 Moderate 14 10 sec estimate 15.2.3 Turbine Trip - Bypass Operable See Text and Appendix 15E Moderate
SSES-FSAR Table Rev. 66 FSAR Rev. 71 Page 2 of 4 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux
% of Rated Maximum Dome Pressure psig Maximum Vessel Pressure psig Maximum Steam line Pressure psig Maximum Core Average Surface Heat Flux,%
CPR Frequency Category Number of Valves -
1st Blowdown Duration of Blowdown 15.2.3 15C.2.2-1 through 15C.2.2-4 Turbine Trip - Without Bypass (100% Power, 108 Mlbm/hr, Max Allowable Scram Time) EOC RPT Operable 389 1300 1332 1321
~135 0.42 -
A10 0.44 -
A11 Moderate 14 10 sec estimate 15.2.4 Inadvertent MSIV Closure See Text and Appendix 15E Moderate 15.2.5 Loss of Condenser Vacuum See Text and Appendix 15E Moderate 15.2.6 Loss of Auxiliary Power Transformer See Text and Appendix 15E Moderate 15.2.6 Loss of All Grid Connections See Text and Appendix 15E Moderate 15.2.7 Loss of All Feedwater Flow See Text and Appendix 15E Moderate 15.2.8 Feedwater Piping Break See Section 15.6.6 15.2.9 Failure of RHR Shutdown Cooling See Text 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 15.3.1 Trip of One Recirculation Pump Motor See Text and Appendix 15E Moderate 15.3.2 Trip of Both Recirculation Pump Motors See Text and Appendix 15E Moderate
SSES-FSAR Table Rev. 66 FSAR Rev. 71 Page 3 of 4 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux
% of Rated Maximum Dome Pressure psig Maximum Vessel Pressure psig Maximum Steam line Pressure psig Maximum Core Average Surface Heat Flux,%
CPR Frequency Category Number of Valves -
1st Blowdown Duration of Blowdown 15.3.3 15C.3.3-6 Seizure of One Recirculation Pump (Single Loop Operation) 67 1052 1065 1068 67 0.33 - A10 0.77 - A11 Limiting Fault 15.3.4 Recirculation Pump Shaft Break See Text Limiting Fault 15.4 REACTIVITY AND POWER ANOMALIES 15.4.1.1 RWE - Refueling See Text Infrequent 15.4.1.2 RWE - Startup See Text Infrequent 15.4.2 RWE - At Power, 108 Mlbs/hr, Bypass Operable See Text Note 5 Note 5 Note 5 Note 5 0.22 Moderate 15.4.3 Control Rod Maloperation See Subsections 15.4.1 and 15.4.2 15.4.4 Startup of Idle Recirculation Loop See Text and Appendix 15E Moderate 15.4.5 Recirculation Flow Controller Failure(3)
See Text NOTE 5 NOTE 5 NOTE 5 NOTE 5 0.34 Moderate 15.4.7 Misplaced Bundle Accident See Text Note 5 Note 5 Note 5 Note 5 See Text Infrequent 15.4.7 Rotated Bundle Accident See Text Note 5 Note 5 Note 5 Note 5 See Text Infrequent
SSES-FSAR Table Rev. 66 FSAR Rev. 71 Page 4 of 4 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux
% of Rated Maximum Dome Pressure psig Maximum Vessel Pressure psig Maximum Steam line Pressure psig Maximum Core Average Surface Heat Flux,%
CPR Frequency Category Number of Valves -
1st Blowdown Duration of Blowdown 15.5 INCREASE IN REACTOR INVENTORY 15.5.1 Inadvertent HPCI Pump Start (at 60% power)
See Text and Appendix 15E 0.39 Moderate 15.5.3 BWR Transients That Increase Reactor Coolant Inventory See Sections 15.1 and 15.2 Notes
- 1.
Unless otherwise stated, the plant initial condition listed in this table for transients is: 100% Power, 108 Mlbs/hr Flow, EOC-Reactor Pump Trip Operable, Bypass Operable, Realistic Scram Time.
- 2.
Minimum MCPR operating limit for Single Loop Operation, see Text.
- 3.
Recirculation Flow Controller Failure transients are initiated from low power/low flow conditions. This one started at 62 Mlbs/hr flow with main steam bypass operable.
- 4.
Steam line pressure is at the turbine stop valve for events in which the turbine trips. For other transients the steam line pressure is assumed to be no higher than the reactor vessel dome pressure.
- 5.
These Anticipated Operational Occurrences are analyzed as steady-state events.
SSES-FSAR Table Rev. 65 FSAR Rev. 71 Page 1 of 3 TABLE 15C.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 1
- 1.
Thermal Power Level, MWT Rated Value Analysis Value 3952 (100%)
4031(102%)
- 2.
Steam Flow, Mlbs/hr (At 100% Power and 100 Mlbs/hr) 16.624
- 3.
Maximum Core Flow, Mlbs/hr 108.0(3)
- 4.
Feedwater Flow Rate, Mlbs/hr (At 100% Power and 100 Mlbs/hr) 16.592
- 5.
Feedwater Temperature, F (At 100% Power and 100 Mlbs/hr) 403.3
- 6.
Vessel Dome Pressure, psig (At 100% Power and 100 Mlbs/hr) 1035.7
- 7.
Vessel Core Pressure, psig at Channel exit (At 100% Power and 100 Mlbs/hr) 1047.4
- 8.
Turbine bypass Capacity, % Rated 21.5%
- 9.
Core Coolant Inlet Enthalpy, BTU/lb (At 100% Power and 100 Mlbs/hr) 523.6(2)
- 10.
Turbine Inlet Pressure, psia 976.3
- 11.
Fuel Types ATRIUM-10 ATRIUM-11
- 12.
Core Average Gap Conductance, BTU/hr-ft2-°F 500 to 1600(1)
- 13.
Core Leakage Flow, %
10%(2)
- 14.
Required MCPR Operating Limit See Unit 1 COLR (FSAR section 16.3 - TRMs)
- 15.
MCPR Safety Limit See Table 15C.0-3
- 16.
Doppler Coefficient See Note 4
SSES-FSAR Table Rev. 65 FSAR Rev. 71 Page 2 of 3 TABLE 15C.0-2 (continued)
INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 1
- 17.
Void Coefficient See Note 4
- 18.
Core Average Rated Void Fraction See Note 4
- 19.
Scram Reactivity Analysis Data See Note 4
- 20.
Control Rod Scram Times Table 15C.0-5
- 21.
Jet Pump Ratio 2.1
- 22.
Safety Relief Valve Capacity (16 Valves)
Percent of Rated Steam Flow 87%
- 23.
Relief Function Delay, sec 0.1
- 24.
Relief Function Response, sec 0.15 25a.
Relief Mode Set Points for Safety/Relief Valves, psig 2 @ 1106 4 @ 1116 4 @ 1126 3 @ 1136 3 @ 1146 26b.
Safety mode Set Points for Safety/Relief valves, psig 2 @ 1175 6 @ 1195 8 @ 1205
- 26.
Number of Valve Groups Simulated 3
- 27.
High Flux Trip, % Rated Analysis set point 122
- 28.
High Pressure Trip, Analysis Set Point, psig 1105
- 29.
Vessel Level Trips, Nominal Setpoints Inches Above (+), Below (-) Dryer Skirt Bottom, (See Note 5)
High Level Low Level Low Low Level Low Low Low Level (L8) 54 (L4) 30 (L3) 13 (L2) -38 (L1)-129
- 30.
APRM Thermal Trip, Analytical Set Point, % Rated 118
SSES-FSAR Table Rev. 65 FSAR Rev. 71 Page 3 of 3 TABLE 15C.0-2 (continued)
INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 1
- 31.
Recirculation Pump Trip Delay, sec 0.175
- 32.
Recirculation Pump Trip Inertia for Analysis, lbm-ft2 16,800 NOTES
- 1. Gap conductance for reactor system behavior is determined for the fuel types within the core as a function of power and exposure. The hot bundle gap conductance is based on the fuel type that is expected to be limiting. It is also determined based on the initial hot bundle power and exposure.
- 2. Inlet enthalpy and leakage flow are determined for each initial condition analyzed.
- 3. Core flow shown is the maximum. It is varied depending on the initial conditions being analyzed.
- 4. The physics characteristics are based on initial conditions determined from a 3-D simulation of the core over a range of power, flow, and pressure conditions. For certain transient analyses this data is transferred and collapsed for use in a 1-D reactor core/system transient simulation model of SSES unit 1.
- 5. Analytical limits for level setpoints include drift and uncertainty allowances.
SSES-FSAR Table Rev. 63 FSAR Rev. 69 Page 1 of 1 TABLE 15C.0-3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT (ALL FUEL)
UNIT 1 MCPRSL for Two Loop Operation MCPRSL for Single Loop Operation Refer to TS 2.1.1.2
SSES-FSAR Table Rev. 59 FSAR Rev. 71 Page 1 of 1 TABLE 15C.0-4 UNIT 1 MINIMUM MCPR REQUIREMENT FOR SINGLE LOOP OPERATION (Based on Analysis of Pump Seizure Accident in Single Loop Operation)
MINIMUM MCPR REQUIREMENT FOR TWO LOOP OPERATION (Based on Analysis of Pump Seizure Accident in Two Loop Operation)
- ATRIUM 11 results are typical and based on cycle-specific delta-CPR results assuming a single-loop MCPR safety limit of 1.12.
- Pump Seizure Accident in Two Loop Operation is a non-limiting event for Atrium-11. Therefore, these results assume a two-loop MCPR safety limit of 1.09 and were not reevaluated.
MCPR Safety Limit 1.12 ATRIUM-10 Minimum MCPR Requirement 1.45 Typical ATRIUM-11 Minimum MCPR Requirement*
1.89 MCPR Safety Limit 1.09 ATRIUM-10 Minimum MCPR Requirement 1.31 Typical ATRIUM-11 Minimum MCPR Requirement**
1.39
SSES-FSAR Table Rev. 64 FSAR Rev. 69 Page 1 of 1 TABLE 15C.0-5 AVERAGE SCRAM INSERTION TIMES UNIT 1 Refer to COLR Section 5.0
SSES-FSAR Table Rev. 56 TABLE 15C.1.1-1 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING UNIT 1 TIME, SECONDS 0
2
~40 (estimate)
~60 (estimate) 600 (estimate)
FSAR Rev. 60 EVENT Initiate a 100°F temperature reduction into the feedwater system.
Initial effect of unheated feedwater starts to raise core power level and steam flow, (Transport delay in feedwater piping is neglected).
APRM high neutron flux alarm sounds.
Reactor variables settle into new steady state, (below Scram trip point).
Operator begins to reduce core flow.
The above times are estimates. This event is a relatively slow transient and the analysis was performed as a series of steady-state calculations.
Page 1 of 1
SSES-FSAR Table Rev. 65 FSAR Rev. 71 Page 1 of 1 TABLE 15C.1.2-1 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND UNIT 1 (TYPICAL)
TIME, SECONDS EVENT 0
Initiate simulated failure of 127% upper limit on feedwater flow.
18.285 L8 vessel level setpoint trips main turbine and feedwater pumps.
18.350 Reactor scram trip actuated from main turbine stop valve position switch.
18.380 Bypass Valves actuated 18.460 Recirculation pump trip (RPT) actuated by stop valve position switch.
Initial Conditions:
Power
= 100%
Flow
= 108 Mlbs/hr Bypass
= Maximum Allowable Exposure
= EOC
SSES-FSAR Table Rev. 57 FSAR Rev. 71 Page 1 of 1 TABLE 15C.1.3-1 SEQUENCE OF EVENTS FOR PRESSURE REGULATOR FAILURE - OPEN UNIT 1 TIME, SECONDS EVENT 0
Initial conditions, maximum limit on steam flow to turbine.
0.2 Main turbine bypass valves full open
~7.5 Main steamline isolation trip occurs.
~8.0 Initiation of scram trip signal, 0.06 seconds after the Main steam isolation valves reach 85% open position.
~12.0 Pressure in reactor vessel reaches a minimum and starts to increase.
~12.5 MSIVs are fully closed.
48 (est)
Relief valves at lowest setting start to cycle to remove decay heat.
SSES-FSAR Table Rev. 65 FSAR Rev. 71 Page 1 of 1 TABLE 15C.2.2-1 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS UNIT 1 (TYPICAL)
TIME, SECONDS EVENTS 0
Turbine-generator detection of loss electrical load.
0 Generator lockout relays act to initiate turbine control valve fast closure.
0 Turbine bypass valves fail to operate.
0.000 Turbine control valves closure on GLR (Generator Load Reject) 0.075 Reactor scram trip actuated from main turbine stop valve position switch.
0.185 EOC-Reactor Pump Trip initiated.
~2 Actuation of safety/relief valves.
Initial Conditions Power: 100%
Flow: 108 Mlbs/hr Bypass: Inoperable Scram: Maximum Allowable RPT: Operable
SSES-FSAR Table Rev. 59 FSAR Rev. 64 Page 1 of 1 TABLE 15C.3.3-1 PUMP SEIZURE ACCIDENT FROM TWO LOOP OPERATION SEQUENCE OF EVENTS UNIT 1 TIME, SEC EVENT 0.0 Single Pump Seizure was Initiated 0.8 Jet Pump Diffuser Flow Reverses in Seized Loop 1.31 Minimum CPR Note: Figures include a 0.5 second null transient.
SSES-FSAR Table Rev. 58 FSAR Rev. 71 Page 1 of 1 TABLE 15C.3.3-2 PUMP SEIZURE ACCIDENT FROM SINGLE LOOP OPERATION SEQUENCE OF EVENTS UNIT 1 TIME, SEC EVENT 0.0 Single Pump Seizure was Initiated N/A Jet Pump Diffuser Flow Reverses in Seized Loop
~1.5 Minimum CPR
SSES-FSAR Table Rev. 57 FSAR Rev. 64 Page 1 of 1 TABLE 15C.4.2-1 SEQUENCE OF EVENTS - RWE IN POWER RANGE UNIT 1 ELAPSED TIME EVENT 0
Core is assumed to be at rated conditions.
0 Operator selects and withdraws the maximum worth control rod.
1 sec The total core power and the local power in the vicinity of the control rod increase.
5 sec The operator ignores warning and continues withdrawal.
15 sec The RBM system indicates excessive localized peaking.
15 sec The operator ignores warning and continues withdrawal.
20 sec The RBM system initiates a rod block inhibiting signal, credit is taken for this signal. Further control rod withdrawal is blocked.
40 sec Reactor core stabilizes at higher core power level.
60 sec Operator attempts to re-insert control rod to reduce core power level.
80 sec Core stabilizes at rated conditions.
SSES-FSAR Table Rev. 56 TABLE 15C.4.5-1 SEQUENCE OF EVENTS FOR RECIRCULATION FLOW CONTROLLER FAILURE TIME, SECONDS 0
~220
~220
~230 UNIT 1 EVENT Master Flow Controller fails initiating a slow run-up of both reactor recirculation pumps Reactor high flux scram (analytical setpoint, 122%).
Two relief valves open at 1120. 7 psia.
Two relief valves reseat at 1045.7 psia.
This sequence of events is for the event initiated from:
Power
=
Flow
=
Bypass
=
Exposure =
FSAR Rev. 60 69%
60 Mlbs/hr Inoperable EOC Page 1 of 1
SSES-FSAR Table Rev. 56 TABLE 15C.4.7-1 UNIT 1 SEQUENCE OF EVENTS FOR MISLOADED BUNDLE ACCIDENT
- 1.
During core loading operation, bundle is placed in the wrong position.
- 2.
Subsequently, the bundle intended for this position is placed in the position of the previous bundle.
- 3.
During core verification procedure, error is not observed.
- 4.
Plant is brought to full power operation without detecting misplaced bundle.
- 5.
Plant continues to operate.
SEQUENCE OF EVENTS FOR ROTATED BUNDLE ACCIDENT
- 1.
During core loading operation, bundle is placed in its proper location but rotated either 90° or 180° from its proper orientation.
- 2.
During core verification procedure this error is not observed.
- 3.
Plant is brought to full power operation without detecting rotated bundle.
- 4.
Plant continues to operate.
FSAR Rev. 60 Page 1 of 1
SSES-FSAR Table Rev. 58 FSAR Rev. 72 Page 1 of 1 TABLE 15C.4.9-1 SEQUENCE OF EVENTS FOR CONTROL ROD DROP ACCIDENT UNIT 1 APPROXIMATE ELAPSED TIME EVENT Reactor is operating at rod density pattern of up to 50%.
Maximum worth control rod blade becomes decoupled from the CRD.
Operator selects and withdraws the control rod drive of the decoupled rod along with the other control rods assigned to the analyzed rod position sequence.
Decoupled control rod sticks in the fully inserted or in an intermediate bank position.
0 Control rod becomes unstuck and drops to the drive position at the nominal measured velocity plus three standard deviations.
1 second Reactor goes on a positive period and initial power increase is terminated by the Doppler effect.
1 second APRM 120% power signal scrams the reactor.
5 seconds Scram terminates the accident.
SSES-FSAR Table Rev. 64 FSAR Rev. 71 Page 1 of 1 TABLE 15C.4.9-2 CONTROL ROD DROP ACCIDENT UNIT 1 (TYPICAL)
Cycle Exposure, MWD/MTU EOC Control Rod Sequence B
Rod Group 2
Dropped Rod Location 22-55 Dropped Rod Worth
<12 mk Number of Failed Fuel Rods
<2000 Peak Deposited Enthalpy, cal/gm
<230
Key Parameters - FC_NO_N_A_100P _108F 400 -i;,,,,:,=,;='e:~=~---------------------------------------------;-
i ---1 Core Power 300 1 ----e Core Heat Flux
-I -
3 Core Flow 1 - - -4 Vsl Exit Stm Flow
-5 Feed Flow
'O Q) ca 200 a: -
0 -
C:
Q) e Q) a..
100 0
-100 +------,--........ --........ --...-----,,----,---....---.-----,.--......
0.0 5.0 10.0 15.0 20.0 25.0 Time (s)
SUSQUEHANNA STEAM ELECTRIC STATION UNITl FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure lSC.1.2-1-1, Rev. 66 FSAR Rev. 71
U)
Q)
..c
<)
C
- .=,
Sensed Water Level - FC NO N A 1 OOP 108F 600 i;:=:=.~=====;-......_---'----'----'---
-'-----'---...,__----'----j I -
1 Sensed Levell 580
~ 560 Q)
....J 540 520 ~.- -----
0.0 5.0 10.0 15.0 20.0 25.0 Time (s)
SUSQUEHANNA STEAM ELECTRIC STATION UNITl FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure lSC.1.2-1-2, Rev. 66 FSAR Rev. 71
U) --E
- 9 -
QJ (tj a:
3:
0 u::::
SRV Flow-FC NO N A 100P 108F 1 -
I -
1 Total SRV Flowj 0 -
0 0
-1 +---..----,------,------,,------,------.------,-----,----,---~
0.0 5.0 10.0 15.0 20.0 25.0 Time (s)
SUSQUEHANNA STEAM ELECTRIC STATION UNITl FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure lSC.1.2-1-3, Rev. 66 FSAR Rev. 71
Pressure - FC 1300 1 Steam Dome Pressure
f! Lower Plenum Pressure 1200 "iii
.e:
Q)
(/)
(/)
Q) 0.. 1100 *
1*
1000 0.0 5.0 10.0 FSAR Rev. 71 NO N Time (s}
A 100P 108F I
I I I I I I
...1 15.0 20.0 25.0 SUSQUEHANNA STEAM ELECTRIC STATION UNITl FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure lSC.1.2-1-4, Rev. 66
a:
1ij 0:
0 c "
~ "
- 0.
I
...I is
~
0 0..
E
.S!
(/)
Q)
Key Parameters 100 50 0
-50 +---~-----~------~--~-------+
0.0 5.0 560.0 540.0 520.0 10.0 Time (s)
Downcomer Water Level
\ I 15.0 l---< CNTRLVAR-5529 20.0 500.0 +------------------------+
0.0 1.0 0.8 0.6 0.4 0.2 0.0 5.0 10.0 Times 15.0 Valve Stem Position
-1 Turbtne Conltol V31,*e
~
Turbine Stop Valve
- 3Main Steam 1,oiation V;JNes J 3 3 3 3 3) 3 20.0
-0.2 0.0 5.0 10.0 Time(s) 15.0 20.0
- s.,
5 Ill 8l a:
1400 1200 1000 800 600 400 200 0 0.0 2000.0 5.0 Pressure 10.0 Time(s)
Flow
,- * - ~ - - ~ - ~ - - --~ - ~
I 15.0 0.0
--+-- --+---+--~-- ----+--*--+--~-.+- -illj,------
20.0
-2000.0 ;---~--~-----~--~--~-------t-0.0 5.0 FSAR REV. 71 10.0 Time(s) 15.0 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PRESSURE REGULATOR FAILED OPEN TYPICAL OF UNIT 1 FIGURE 15C.1.3-1, Rev0 20.0
"O Q) ca a:
a C:
Q)
(..)
Q) a..
Key Parameters - TL_NO_ T _A_ 1OOP_108F 400 -i-----'---------'------'-----'-----;:==::======::;-----i
- ----1 Core Power
---e core Heat Flux 3CoreFlow
- - -4 Vsl Exit Stm Flow 300
-5 Feed Flow
~ \
I I
/ \
200 I \
I 100
\
I I
0 I
I l
I
/
I
/
-100 +-----,-------,----,-----,----..-----,-----,------!
0.0 1.0 2.0 Time (s)
FSAR REV.71 3.0 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE lSC.2.2-1-1, Rev 65 4.0
en Q)
..c u
C: -
Sensed Water Level - TL NO T A 1 OOP 108F 570 i======:----'-------'----'------'---
---j I -
1 Sensed Levell 560
~ 550 Q)
....J I,...
Q)
~
540 530 +, ----,---..----....------,-----,---..--------
0.0 1.0 2.0 3.0 Time (s)
FSAR REV.71 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE lSC.2.2-1-2, Rev 65 4.0
ui'
---E
- 9 Q) cri a:
~
0 u:::
SRV Flow - TL NO T A 100P 108F 4000 t=====~~:::;::::::;--....__ __
I -
1 Total SRV Flo"1 3000 2000 1000 r
- ---1----- ------
1-----
/
I I
{
I I i
I I
I I I 0 +-----.--1----.---------,.-------.---------,-----!
0.0 1.0 2.0 3.0 4.0 Time (s)
FSAR REV.71 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE lSC.2.2-1-3 Rev 65
c?
"cii Q. -
Q)
- )
en en Q) a..
Pressure - TL NO T A 1 OOP 108F 1400 -t===========::----'------'--
--t 1 Steam Dome Pressure
-I! lower Plenum Pressure 1300 1200 1100 1000 +, ----,,-----,-----,-----,,----,-----,-----,----
0.o 1.0 2.0 s.o Time (s)
FSAR REV.71 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE lSC.2.2-1-4, Rev 65 4.0
Key Paramet ers
'150,
100
~ ~ --=.::-:..~....
--.i._* ~
-1/4- - --.i;,_
-0
/1- - -- ---s... _
iii
4,..
a:
..... 4...
0 E
50 -
~
IL 0 -
\
\
I,,,,
- 50 '
0.0 2'.o 4.0 6'.o a'.o Time (s)
Downcomer Water Level 590.0 580.0 0~
~ 570.0
- §.
\
~
t
-' 560.0 550.0
\
i ----'ICNTRLVAA:$9 540.0 0.0 2.0 4.0 6.0 8.0 Time (s) 10.0 10.0 Pressure 1200 1150 1100
~
., 1050 5.,
a: 1000 950 Pl>rnm-
-4Turbntlisa.i!lrPns:sme
-- Slearrlirle f
-s:sue at \'esttl Outtet 900 0.0 2.0 4.0 6.0 Tune (s)
Flow 6000.0 4000.0 I
g_
.!l 2000.0 A
0:
~
IJ..
0.0
--t-§-..---111-6---~~---....~~
l I V
-2000.0 0.0 2.0 4.0 6.0 nme (s)
FSAR REV. 71 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT TWO LOOP OPERATION TYPICAL OF UNIT 1 FIGURE 15C.3.3-5, Rev 0 8.0 8.0 10.0 10.0
cc 0
C..
ll.
90 70 50 30 10
-10
-30 Key Parameters - SLPS_A 11_67.2P _52F I
-ICo,o PowO<
-e Ct11c Heat Fku 3Coreflow
- - V!IExltStmFtow
- - --5fee<fAow
-50 f------- ~----------------'
0.0 2.0 4.0 6.0 Time s Sensed Water Level - SLPS_A 11_67.2P _52F 555 550 ~----------------------~
o.o 2.0 4.0 6.0 Times Pressure - SLPS_A11_67.2P _52F
--1 Steam Dom& P1eS$Ure
--Q: LI,Wof Plonum PIOSSVfO 1050
~
.e
~
~
~ 1000 ll.
950.,_ __________
- _***c..*/~* ------------'-
o.o 2.0 4.0 Time s SRV Flow - SLPS_A11_67.2P _52F
- -* 1 Total SRV Fiow 0
E
" 0 cc 0
G:
0
-1 0.0 2.0 4.0 Time s FSAR REV. 71 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT SINGLE LOOP OPERATION TYPICAL OF UNIT 1 FIGURE 15C.3.3-6, Rev 0 6.0 6.0