ML21294A088
| ML21294A088 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2021 |
| From: | Talen Energy, Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21294A245 | List:
|
| References | |
| PLA-7935 | |
| Download: ML21294A088 (33) | |
Text
SSES-FSAR Text Rev. 55 APPENDIX 15D SUSQUEHANNA STEAM ELECTRIC STATION UNIT 2 FINAL SAFETY ANALYSIS REPORT -
CYCLE SPECIFIC DATA 15D.1 Appendix D Contents 15D.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 15D.1.2-1.
15D.1.2 References 15D.1.2-1 ANP-3741(P), Revision 2, Susquehanna Unit 2 Cycle 20 Reload Licensing Analysis, Framatome Inc., June 2019. (General Reference per NEI 98-03)
SSES-FSAR Table Rev. 67 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Maximum Maximum Maximum Maximum Maximum Core Number of Neutron Dome Vessel Steam line Average Valves - Duration Flux,% of Pressure Pressure Pressure Surface Heat Frequency 1st of Section Figure Description1 Rated psia psia psia Flux, % CPR Category Blowdown 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 0.14 Moderate 0 0 sec 15.1.2 15D.1.2-1 Feedwater Controller Failure (100% Power, ~350 1236 1236 1233 ~125 A11 - 0.34 Moderate 14 4 sec 108 Mlbm/hr, Maximum Allowable Scram A10 - 0.35 estimate Time) 15.1.3 15D.1.3-1 Pressure Regulator Failure - Open 100 1051 1094 1043 100 A11 - 0.01 Moderate 2 See Text A10 - 0.01 15.1.4 Inadvertent Opening of Safety or Relief See Text Moderate Valves 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 Moderate Appendix 15E 15.2.2 15D.2.2-1 Generator Load Reject- Without Bypass ~400 1246 1272 1260 ~125 A11 - 0.44 Moderate 14 10 sec (100% Power, 108 Mlbm/hr, Maximum A10 - 0.43 estimate Allowable Scram Time) 15.2.3 Turbine Trip - Bypass Operable See Text and Moderate Appendix 15E 15.2.3 15D.2.2-1 Turbine Trip - Without Bypass ~400 1246 1272 1260 ~125 A11 - 0.44 Moderate 14 10 sec (100% Power, 108 Mlbm/hr, Maximum A10 - 0.43 estimate Allowable Scram Time) 15.2.4 Inadvertent MSIV Closure See Text and Moderate Appendix 15E FSAR Rev. 70 Page 1 of 4
SSES-FSAR Table Rev. 67 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Maximum Maximum Maximum Maximum Maximum Core Number of Neutron Dome Vessel Steam line Average Frequenc Valves - Duration Flux,% of Pressure Pressure Pressure Surface Heat y 1st of Section Figure Description1 Rated psia psia psia Flux, % CPR Category Blowdown Blowdown 15.2.5 Loss of Condenser Vacuum See Text and Moderate Appendix 15E 15.2.6 Loss of Auxiliary Power Transformer See Text and Moderate Appendix 15E 15.2.6 Loss of All Grid Connections See Text and Moderate Appendix 15E 15.2.7 Loss of All Feedwater Flow See Text and Moderate Appendix 15E 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 Moderate Appendix 15E 15.3.2 Trip of Both Recirculation Pump Motors See Text and Moderate Appendix 15E 15.3.3 15.D.3.3-6 Seizure of One Recirculation Pump 0.33- A10 Limiting (Single Loop Operation) 67 959 973 958 67 0.77 - A11 Fault 15.3.4 Recirculation Pump Shaft Break See Text Limiting Fault FSAR Rev. 70 Page 2 of 4
SSES-FSAR Table Rev. 67 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Maximum Maximum Maximum Maximum Maximum Core Number of Neutron Dome Vessel Steam line Average Valves - Duration Flux,% of Pressure Pressure Pressure Surface Heat Frequency 1st of Section Figure Description1 Rated psia psia psia Flux, % CPR Category Blowdown Blowdown 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 See Text Note 5 Note 5 Note 5 Note 5 0.21 Moderate Operable 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 Moderate Appendix 15E 15.4.5 Recirculation Flow Controller Failure(3) See Text Note 5 Note 5 Note 5 Note 5 0.17 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 FSAR Rev. 70 Page 3 of 4
SSES-FSAR Table Rev. 67 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Maximum Maximum Maximum Maximum Maximum Core Number of Neutron Dome Vessel Steam line Average Frequenc Valves - Duration Flux,% of Pressure Pressure Pressure Surface Heat y 1st of Section Figure Description1 Rated psia psia psia Flux, % CPR Category Blowdown Blowdown 15.5 INCREASE IN REACTOR INVENTORY 15.5.1 Inadvertent HPCI Pump Start See Text and A11 - 0.52 Moderate
@ limiting Power Appendix 15E A10 - 0.51 15.5.3 BWR Transients That Increase Reactor See Sections Coolant Inventory 15.1 and 15.2 Notes
- 1. Unless otherwise stated, the plant initial condition listed in this table for transients is: 102% 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 analyses are initiated from low power/low flow conditions.
- 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.
FSAR Rev. 70 Page 4 of 4
SSES-FSAR Table Rev. 66 TABLE 15D.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 2
- 1. Thermal Power Level, MWT 3952 (100%)
Rated Value 4031 (102%)
Analysis Value
- 2. Steam Flow, Mlbs/hr 16.624 (At 100% Power and 100 Mlbs/hr)
- 3. Maximum Core Flow, Mlbs/hr 108.0(3)
- 4. Feedwater Flow Rate, Mlbs/hr 16.592 (At 100% Power and 100 Mlbs/hr)
- 5. Feedwater Temperature,F 403.3 (At 100% Power and 100 Mlbs/hr)
- 6. Vessel Dome Pressure, psig 1035.7 (At 100% Power and 100 Mlbs/hr)
- 7. Vessel Core Pressure, psig at Channel Exit 1047.4 (At 100% Power and 100 Mlbs/hr)
- 8. Turbine Bypass Capacity, % Rated 21.5%
- 9. Core Coolant Inlet Enthalpy, BTU/lb 523.6 (At 100% Power and 100 Mlbs/hr)
- 10. Turbine Inlet Pressure, psia 976.3 ATRIUM-10
- 11. Fuel Types ATRIUM-11 I t-----+------------
- 12. Core Average Gap Conductance, BTU/hr-ft2-F 500 to 1700(1)
- 13. Core Leakage Flow,% 10%(2)
- 15. MCPR Safety Limit See Table 15D.0-3
- 16. Doppler Coefficient See Note 4 FSAR Rev. 70 Page 1 of 3
SSES-FSAR Table Rev. 66 TABLE 15D.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 2
- 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 15D.0-5
- 21. Jet Pump Ratio 2.1
- 22. Safety Relief Valve Capacity (16 Valves) 87%
Percent of Rated Steam Flow
- 23. Relief Function Delay, sec 0.1
- 24. Relief Function Response, sec 0.15 25-a. Relief Mode Set Points for Safety/Relief Valves, psig 2 @ 1106 3 @ 1136 4 @ 1116 3 @ 1146 4 @ 1126 25-b. Safety Mode Set Points for Safety/Relief Valves, psig 2 @ 1175 6 @ 1195 8 @ 1205
- 26. Number of Valve Groups Simulated 3 High Flux Trip, % Rated
- 27. 122 Analysis set point
- 28. High Pressure Trip, 1105 Analysis Set Point, psig
- 29. Vessel Level Trips, High Level (L8) 54 Nominal Setpoints (L4) 30 Inches Above(+), Below (-) Low Level (L3) 13 Dryer Skirt Bottom, Low Low Level (L2) -38 (See Note 5) Low Low Low Level (L1)-129
SSES-FSAR Table Rev. 66 TABLE 15D.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 2
- 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 2.
- 5. Analytical limits for level setpoints include drift and uncertainty allowances.
FSAR Rev. 70 Page 3 of 3
SSES-FSAR Table Rev. 64 TABLE 15D.0-3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT (ALL FUEL)
UNIT 2 MCPRSL for MCPRSL for Two Loop Single Loop Operation Operation Refer to TS 2.1.1.2 FSAR Rev. 69 Page 1 of 1
SSES-FSAR Table Rev. 61 TABLE 15D.0-4 UNIT 2 MINIMUM MCPR REQUIREMENT FOR SINGLE LOOP OPERATION MCPR Safety 1.11 Limit ATRIUM-10 1.44 Minimum MCPR Requirement Typical 1.88 ATRIUM-11 Minimum MCPR Requirement*
(Based on Analysis of Pump Seizure Accident in Single Loop Operation)
MINIMUM MCPR REQUIREMENT FOR TWO LOOP OPERATION MCPR Safety 1.08 Limit ATRIUM-10 1.30 Minimum MCPR Requirement Typical 1.37 ATRIUM-11 Minimum MCPR Requirement*
(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 two-loop and single-loop MCPR safety limits of 1.08 and 1.11, respectively.
FSAR Rev. 70 Page 1 of 1
SSES-FSAR Table Rev. 65 TABLE 15D.0-5 AVERAGE SCRAM INSERTION TIMES UNIT 2 Refer to COLR Section 5.0 FSAR Rev. 69 Page 1 of 1
SSES-FSAR NIMS Rev. 56 TABLE 15 D. 1.1-1 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING UNIT 2 TIME, SECONDS EVENT 0 Initiate a 100°F temperature reduction into the feedwater system.
2 Initial effect of unheated feedwater starts to raise core power level and $learn flow, (Transport delay in feedwater piping is neglected).
- 40 (estimate) APRM high neutron flux alarm sounds .
==60 (estimate) Reactor variable settle into new steady state, (below Scram trip point).
600 (estimate) Operator begins to reduce power (reduce core flow and/or insert normal sequence control rods) to restore plant operation within normal power-flow conditions.
The above times are estimates. This event is a relatively slow transient and the analysts was performed as a series of steady-state calculations FSAR Rev. 59 Page 1 of 1
SSES-FSAR Table Rev. 66 TABLE 15D.1.2-1 SEQUENCE OF EVENTS FOR FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND UNIT 2 (TYPICAL)
TIME, SECONDS EVENT 0 Initiate simulated failure of 130% upper limit on feedwater flow.
13.79 L8 vessel level setpoint trips main turbine and feedwater pumps.
17.92 Reactor scram trip actuated from main turbine stop valve position switch.
17.95 Bypass Valves actuated 18.03 Recirculation pump trip (RPT) actuated by stop valve position switch.
~21 Activation of safety/relief valves.
Initial Conditions:
Power = 100%
Flow = 108 Mlbs/hr Bypass = Operable RPT = Operable Scram Time = Maximum Allowable Exposure = EOC FSAR Rev. 70 Page 1 of 1
SSES-FSAR Table Rev. 57 TABLE 15D.1.3-1 SEQUENCE OF EVENTS FOR PRESSURE REGULATOR FAILURE - OPEN UNIT 2 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 the reactor 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.
FSAR Rev. 70 Page 1 of 1
SSES-FSAR Table Rev. 66 TABLE 15D.2.2-1 SEQUENCE OF EVENTS FOR GENERATOR LOAD REJECTION WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS UNIT 2 (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.
0.001 Turbine control valves (TCV) close on GLR, (Generator Load Reject) 0.080 Initiate scram on TCV fast closure (Trip oil pressure-low).
0.100 Turbine control valves closed.
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 Time RPT: Operable FSAR Rev. 70 Page 1 of 1
SSES-FSAR Table Rev. 60 TABLE 15D.3.3-1 PUMP SEIZURE ACCIDENT FROM TWO LOOP OPERATION SEQUENCE OF EVENTS UNIT 2 TIME, SEC EVENT 0.0 Single Pump Seizure was Initiated
~1.0 Jet Pump Diffuser Flow Reverses in Seized Loop 1.30 Minimum CPR (ATRIUM-10) 1.81 Minimum CPR (ATRIUM-11)
Note: Figures include a 0.5 second null transient FSAR Rev. 70 Page 1 of 1
SSES-FSAR Table Rev. 59 TABLE 15D.3.3-2 PUMP SEIZURE ACCIDENT FROM SINGLE LOOP OPERATION SEQUENCE OF EVENTS UNIT 2 TIME, SEC EVENT 0.0 Single Pump Seizure was Initiated N/A Jet Pump Diffuser Flow Reverses in Seized Loop 1.84 Minimum CPR (ATRIUM-10) 2.50 Minimum CPR (ATRIUM-11)
FSAR Rev. 70 Page 1 of 1
SSES-FSAR NIMS Rev. 58 TABLE 15D.4.2-1 SEQUENCE OF EVENTS - RWE IN POWER RANGE UNIT 2 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.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 57 TABLE 15D.4.5-1 SEQUENCE OF EVENTS FOR RECIRCULATION FLOW CONTROLLER FAILURE UNIT2 TIME. SECONDS EVENT 0 Master Flow Controller fails initiating a slow run-up of both reactor recirculation pumps
-220 Two relief valves open at 1120.7 psia.
-220 Reactor high flux scram (analyticat setpoint, 122%).
-230 Two relief valves reseat at 1045. 7 psi a.
This sequence of events is for the event initiated from:
INITIAL CONDITIONS Power = 69%
Flow = 60M lbs/hr Bypass = Inoperable Exposure = EOC FSAR Rev. 59 Page 1 of 1
SSES-FSAR NIMS Rev. 57 TABLE 15D.4.7-1 UNIT 2 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 90q or 180q 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. 68 Page 1 of 1
SSES-FSAR Table Rev. 57 TABLE 15D.4.9-1 SEQUENCE OF EVENTS FOR CONTROL ROD DROP ACCIDENT UNIT 2 APPROXI MATE ELAPSED TIME EVENT Reactor is operating at a 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 Bank Position Withdrawal Sequence (BPWS).
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.
second Reactor goes on a positive period and initial power increase is terminated by the Doppler effect.
second APRM 120% power signal scrams the reactor seconds Scram terminates the accident.
FSAR Rev. 61 Page 1 of 1
SSES-FSAR Table Rev. 65 TABLE 15D.4.9-2 CONTROL ROD DROP ACCIDENT UNIT 2 (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 FSAR Rev. 70 Page 1 of 1
Key Parameters 400 i=====~_____,_______,_____-+-------+
- 1 Core Power
--ecore Heat Flux
-acoreFlow
- - -4 Vsl Exit Stm Flow
--5 Feed Flow 300
-g hj 200 a:
-1000.0
.----- - - - - - - - - - - - - -20.0 10.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15D.1.2-1-1, Rev. 67
Sensed Water Level 600 i====~-------------------t--------i
- 1 Sensed Level 590 en 0,)
..c g 580 560 550 +--------r--------,-------.------+---------1 0.0 10.0 20.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15D.1.2-1-2, Rev. 66
Vessel Dome Pressure 1400-i=====~=-----'----__,,-----+------;-
- 1 Steam Dome Pressure
--e Lower Plenum Pressure 1300 c?
- u; s
Q) 1200 Cl)
Cl)
Q) 0..
1100 1 0 0 0 - - - - - - - - - - -........
o.o
10.0 20.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 1SD.1.2-1-4, Rev. 66
SRV Flow 3000-t=====:::;-------+-----~-----:------;
.-*~~~ !
00 2000 E
g Q)
'E a:
~
rr: 1000 0 -t-----+---+-,-~----------....----~--.. .
0.0 10.0 Time (s) 20.0
1 FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15D.1.2*1*5, Rev. 66
Key Parameters Pressure 1400 1200 1000 al
~ '"
'iii a 800
'l5
~
50 ;
~
Q) i a.
600 a.
400 0
200
-50 + - - - ~ - - ~ - - ~ ~ - - ~ - - ~ - - - + - - - ~ - - - - - + 0 0.0 5.0 10.0 15.0 20.0 0.0 5.0 10.0 15.0 20.0 Time(s) Time (s)
Downcomer Water Level Flow 6000.0 560.0 4000.0 2
C ai 540.0 2000.0 ijj
.J f--1/4--S--s---~-- S---5--6---5-- 1 i 520.0 0.0 __ ,_ _ _,_ _ __,__ _,_i.., __ , __ *--+-->-~_,__,._..,
l-1CNTRLVAR-,!;2!l 500.0 i - - - - , - - - - , - - - - , - - - - , - - - - , - - - - , - - - - , - - - - - - - - , . -2000.0 + - - - - ~ - - ; - - - - , - - - - , - - - - , - - - - , - - - - - , - - - - - \ -
0.0 5.0 10.0 16.0 20.0 0.0 6.0 10.0 15.0 20.0 Tlme(s Time(s)
Valve Stem Position 1.0 8
- e
"'0 0.8
//
- a. 0.6 E - t T\JWJ)9 Coo!rol Val.'9 ti.i"
~TurtmeStopV:if,?
0.4 --31.!MtSteaml!ci'ationV~ FSAR REV. 70
~
> 0.2 I SUSQUEHANNA STEAM ELECTRIC STATION T UNITS 1 AND 2 0.0 FINAL SAFETY ANALYSIS REPORT
-0.2 0.0 5.0 10.0 15.0 20.0 Tlme(s)
PRESSURE REGULATOR FAILED OPEN TYPICAL OF UNIT 2 FIGURE 15D.1.3-1, Rev 0
Key Parameters 600 - 1 Core Power
--£Core Heat Flux
--acore Flow 500 - - -4 Vsl Exit Sim Flow
- -5 Feed Flow 400 "O
Q) 300 1i:l a::
....0
-C:
Q) 8 200 Q) a.. mo \ ., ... I,~ .... __ ,......,.
\
\
\ ,, ,,,,,--i- ... _ ..... '
I 0
, ..... - ,t'
-100
-200 0.0 1.0 2.0 3.0 4.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure 15D.2.2-1-1, Rev. 67
Sensed Water Level 570t.==""""'~,:,:::::;---;------'----:-------------t----------,
/ - 1 Sensed Levell w 560 Q)
.c
(.)
C:
C, a3
~....
Q) rtl s 550 540 ------,----,-----.-------.----,-----.----"T"'"""-----1 0.0 1.0 2.0 3.0 4.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure lSD.2.2-1-2, Rev. 66
Steam Dome and Plenum Pressure 1400i:=======~:---_,_--~--------+----------t
- 1 Steam Dome Pressure
--e Lower Plenum PressJre 1300
'iii 8
~ 1200 -- - -- -----
- J
(/)
(/)
Q)
Q..
1100 1000 +------,-----;----..-----,.-------.----....----....-------+
0.0 1.0 2.0 3.0 4.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure 15D.2.2-1-4, Rev. 66
SRV Flow 4000ir====:=,::::===~~--'----'---~--~--~--~---;
- 1 Total SRV Flo 3000
~
E
.0 Q) cu 2000 a::
5 a:
1000 0+--------------------.-----....-----,----
0.0 1.0 2.0 3.0 4.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure 15D.2.2-1-5, Rev. 66
Key Pan;nheters Pressure 1200 150 i
-icorePo-,v,er
~"-"--
~~~~\Aw;
~-Corafto,i/ 1150
.-~~VsJ.&tstrnAa.v 100 -':::~~~-- -------------------- -- ---,----------
1100 m
- 0::
'5 C:
50 I
'iii El, fl! 1050
,.,@ I I
I £
- a. 1000 iI ,II I
- 0. -- - - ---+17 -
I:/
\
,,:, 950
\~
-60 ~
o.o 6.0 i
8.0 J
10.0 2.0 4.0 6.0 8.0 Time (s)
Downcomer Water Level Flow 6000.0 - - ~ - - - - - + - - ~ - - + - - - ~ - - - ~ - - - - ~ - - , .
-1VesselBi:it$leamAow
~Twbinefb.Y
-Jfeed.valerffow 580.0 ---ITotalSRV-
--5TurbtneB I Row 4000.0
-~1 i ~-
- 570.0 6
iii 2000.0 a,
-' 560.0 0.0 550.0
-2000.0 + - - ~ - - - - + - - ~ - - + - - - ~ - - - ~ - - + - - - ~ - - +
0.0 2.0 4.0 6.0 8.0 10.0 4.0 6.0 Time(s)
Time(s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT TWO LOOP OPERATION TYPICAL OF UNIT 2 FIGURE 15D.3.3-5, Rev 0
Key Parameters Steam Dome and Plenum Pressure 80 t----'-----;--------';=_,==a;Co<a=o=aP=..,=,=-~-----t 1020 t,__,==;ae,s~""'*m""0o',ccmss,Prs=...,.,~,,,----'--------'----'-----'----t
eeoretteatAo~ --21.o.Ya, Plenum Pressure
- .tCore FkJw
- - -4 Vs! Exit Stm Row
---6FO&<IF":!;_
60 1000
_.a---G--3-------&---/4 -!t----3----s--u--
0-1----,----.-------.--------,-----+ 940 \----~-~-----~--~---,-----!
o.o 2.0 4.0 6.0 0.0 2.0 4.0 6.0 Time(sl Time(s}
Sensed Water Level SRV Flow 585 't=_=,,,,,,,.,,,=""s=....,.=,,s-------'-----;----'---=-----i - 1 TolaJ SRV Flud 680 0 I
0 I
.0
=
C:
=
! 675
~
cr:
0+-+-1--1f--+-+-+--1--+--+---t-+->->-+-1--1--
I I 570 0 565+--------------i----------t 0.0 2.0 4.0 6.0 -1 M
.J-------------,----,-----~ ~ ~ M Time Isl Time(sl FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT SINGLE LOOP OPERATION TYPICAL OF UNIT 2 FIGURE 150.3.3-6, Rev 0