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 FSAR Rev. 69 15D-1 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 FSAR Rev. 70 Page 1 of 4 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux,% of Rated Maximum Dome Pressure psia Maximum Vessel Pressure psia Maximum Steam line Pressure psia 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 0.14 Moderate 0
0 sec 15.1.2 15D.1.2-1 Feedwater Controller Failure (100% Power, 108 Mlbm/hr, Maximum Allowable Scram Time)
~350 1236 1236 1233
~125 A11 - 0.34 A10 - 0.35 Moderate 14 4 sec estimate 15.1.3 15D.1.3-1 Pressure Regulator Failure - Open 100 1051 1094 1043 100 A11 - 0.01 A10 - 0.01 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 15D.2.2-1 Generator Load Reject-Without Bypass (100% Power, 108 Mlbm/hr, Maximum Allowable Scram Time)
~400 1246 1272 1260
~125 A11 - 0.44 A10 - 0.43 Moderate 14 10 sec estimate 15.2.3 Turbine Trip - Bypass Operable See Text and Appendix 15E Moderate 15.2.3 15D.2.2-1 Turbine Trip - Without Bypass (100% Power, 108 Mlbm/hr, Maximum Allowable Scram Time)
~400 1246 1272 1260
~125 A11 - 0.44 A10 - 0.43 Moderate 14 10 sec estimate 15.2.4 Inadvertent MSIV Closure See Text and Appendix 15E Moderate
SSES-FSAR Table Rev. 67 FSAR Rev. 70 Page 2 of 4 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux,% of Rated Maximum Dome Pressure psia Maximum Vessel Pressure psia Maximum Steam line Pressure psia Maximum Core Average Surface Heat Flux, %
CPR Frequenc y
Category Number of Valves -
1st Blowdown Duration of Blowdown 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 15.3.3 15.D.3.3-6 Seizure of One Recirculation Pump (Single Loop Operation) 67 959 973 958 67 0.33-A10 0.77 - A11 Limiting Fault 15.3.4 Recirculation Pump Shaft Break See Text Limiting Fault
SSES-FSAR Table Rev. 67 FSAR Rev. 70 Page 3 of 4 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux,% of Rated Maximum Dome Pressure psia Maximum Vessel Pressure psia Maximum Steam line Pressure psia Maximum Core Average Surface Heat Flux, %
CPR Frequency Category Number of Valves -
1st Blowdown Duration of 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 Operable See Text Note 5 Note 5 Note 5 Note 5 0.21 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 0.17 Moderate 15.4.7 15.4.7 Misplaced Bundle Accident Rotated Bundle Accident See Text See Text Note 5 Note 5 Note 5 Note 5 Note 5 Note 5 Note 5 Note 5 See Text See Text Infrequent Infrequent
SSES-FSAR Table Rev. 67 FSAR Rev. 70 Page 4 of 4 TABLE 15D.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 2 (TYPICAL)
Section Figure Description1 Maximum Neutron Flux,% of Rated Maximum Dome Pressure psia Maximum Vessel Pressure psia Maximum Steam line Pressure psia Maximum Core Average Surface Heat Flux, %
CPR Frequenc y
Category Number of Valves -
1st Blowdown Duration of Blowdown 15.5 INCREASE IN REACTOR INVENTORY 15.5.1 Inadvertent HPCI Pump Start
@ limiting Power See Text and Appendix 15E A11 - 0.52 A10 - 0.51 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: 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.
SSES-FSAR Table Rev. 66 FSAR Rev. 70 Page 1 of 3 TABLE 15D.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 2
- 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
- 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 1700(1)
- 13.
Core Leakage Flow,%
10%(2)
- 14.
Required MCPR Operating Limit See Unit 2 COLR (FSAR section 16.3 - TRMs)
- 15.
MCPR Safety Limit See Table 15D.0-3
- 16.
Doppler Coefficient See Note 4 I
t-----+------------
SSES-FSAR Table Rev. 66 FSAR Rev. 70 Page 2 of 3 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)
Percent of Rated Steam Flow 87%
- 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 4 @ 1116 4 @ 1126 3 @ 1136 3 @ 1146 25-b. 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. 66 FSAR Rev. 70 Page 3 of 3 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.
SSES-FSAR Table Rev. 64 FSAR Rev. 69 Page 1 of 1 TABLE 15D.0-3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT (ALL FUEL)
UNIT 2 MCPRSL for Two Loop Operation MCPRSL for Single Loop Operation Refer to TS 2.1.1.2
SSES-FSAR Table Rev. 61 FSAR Rev. 70 Page 1 of 1 TABLE 15D.0-4 UNIT 2 MINIMUM MCPR REQUIREMENT FOR SINGLE LOOP OPERATION MCPR Safety Limit 1.11 ATRIUM-10 Minimum MCPR Requirement 1.44 Typical ATRIUM-11 Minimum MCPR Requirement*
1.88 (Based on Analysis of Pump Seizure Accident in Single Loop Operation)
MINIMUM MCPR REQUIREMENT FOR TWO LOOP OPERATION MCPR Safety Limit 1.08 ATRIUM-10 Minimum MCPR Requirement 1.30 Typical ATRIUM-11 Minimum MCPR Requirement*
1.37 (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.
SSES-FSAR Table Rev. 65 FSAR Rev. 69 Page 1 of 1 TABLE 15D.0-5 AVERAGE SCRAM INSERTION TIMES UNIT 2 Refer to COLR Section 5.0
SSES-FSAR NIMS Rev. 56 TABLE 15 D. 1. 1-1 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING UNIT 2 TIME, SECONDS 0
2
- 40 (estimate)
==60 (estimate) 600 (estimate)
FSAR Rev. 59 EVENT Initiate a 100°F temperature reduction into the feedwater system.
Initial effect of unheated feedwater starts to raise core power level and $learn flow, (Transport delay in feedwater piping is neglected).
APRM high neutron flux alarm sounds.
Reactor variable settle into new steady state, (below Scram trip point).
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 Page 1 of 1
SSES-FSAR Table Rev. 66 FSAR Rev. 70 Page 1 of 1 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
= Maximum Allowable Exposure
= EOC
SSES-FSAR Table Rev. 57 FSAR Rev. 70 Page 1 of 1 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.
SSES-FSAR Table Rev. 66 FSAR Rev. 70 Page 1 of 1 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
SSES-FSAR Table Rev. 60 FSAR Rev. 70 Page 1 of 1 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 1.81 Minimum CPR (ATRIUM-10)
Minimum CPR (ATRIUM-11)
Note: Figures include a 0.5 second null transient
SSES-FSAR Table Rev. 59 FSAR Rev. 70 Page 1 of 1 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 2.50 Minimum CPR (ATRIUM-10)
Minimum CPR (ATRIUM-11)
SSES-FSAR NIMS Rev. 58 FSAR Rev. 64 Page 1 of 1 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.
SSES-FSAR Table Rev. 57 TABLE 15D.4.5-1 SEQUENCE OF EVENTS FOR RECIRCULATION FLOW CONTROLLER FAILURE TIME. SECONDS 0
-220
-220
-230 UNIT2 EVENT Master Flow Controller fails initiating a slow run-up of both reactor recirculation pumps Two relief valves open at 1120.7 psia.
Reactor high flux scram (analyticat setpoint, 122%).
Two relief valves reseat at 1045. 7 psi a.
This sequence of events is for the event initiated from:
INITIAL CONDITIONS Power
=
Flow
=
Bypass
=
Exposure
=
FSAR Rev. 59 69%
60M lbs/hr Inoperable EOC Page 1 of 1
SSES-FSAR NIMS Rev. 57 FSAR Rev. 68 Page 1 of 1 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.
SSES-FSAR Table Rev. 57 TABLE 15D.4.9-1 SEQUENCE OF EVENTS FOR CONTROL ROD DROP ACCIDENT UNIT 2 MATE ELAPSED TIME 0
second second seconds 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.
Control rod becomes unstuck and drops to the drive position at the nominal measured velocity plus three standard deviations.
Reactor goes on a positive period and initial power increase is terminated by the Doppler effect.
APRM 120% power signal scrams the reactor Scram terminates the accident.
FSAR Rev. 61 Page 1 of 1 APPROXI
SSES-FSAR Table Rev. 65 FSAR Rev. 70 Page 1 of 1 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
Key Parameters 400 i=====~ ____ _,_ ____ __,_ ____ -+-------+
-1 Core Power 300
-g hj 200 a:
--ecore Heat Flux
-acoreFlow
- - -4 Vsl Exit Stm Flow
--5 Feed Flow
-100 --------.--------------------------
0.0 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 15D.1.2-1-1, Rev. 67
en 0,)
..c Sensed Water Level 600 i====~-------------------t--------i
-1 Sensed Level 590 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
c?
- u; s Q)
Cl)
Cl)
Q) 0..
Vessel Dome Pressure 1400-i=====~=-----'----__,,-----+------;-
-1 Steam Dome Pressure
--e Lower Plenum Pressure 1300 1200 1100 1000-----------........ ---------------- 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-----+---+-,-~----------....----~--....... --------1 0.0 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 15D.1.2*1*5, Rev. 66
al
~
'l5 50
~
~
Q) a.
0 Key Parameters
-50 +---~--~--~~--~--~---+---~-----+
2 C
ai ijj
.J 8
- e 0 a.
E "
ti.i
~
0.0 5.0 10.0 15.0 20.0 Time(s)
Downcomer Water Level 560.0 540.0 520.0 l-1CNTRLVAR-,!;2!l 500.0 i----,----,----,----,----,----,----,--------,.
1.0 0.8 0.6 0.4 0.2 0.0 0.0 5.0 10.0 Tlme(s 16.0 Valve Stem Position
//
-t T\\JWJ)9 Coo!rol Val.'9
~TurtmeStopV:if,?
--31.!MtSteaml!ci'ationV~
I T 20.0
-0.2 0.0 5.0 10.0 Tlme(s) 15.0 20.0 Pressure 1400 1200 1000
'iii a 800 i 600
- a.
400 200 0 0.0 5.0 10.0 15.0 20.0 Time (s)
Flow 6000.0 4000.0 2000.0 f--1/4--S--s---~-- S---5--6---5-- 1 i 0.0
__,_ _ _,_ _ __, __ _,_i.., __, __ *--+-->-~_,__,._..,
-2000.0 +----~--;----,----,----,----,-----,-----\\-
0.0 6.0 10.0 15.0 20.0 Time(s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT PRESSURE REGULATOR FAILED OPEN TYPICAL OF UNIT 2 FIGURE 15D.1.3-1, Rev 0
600 -1 Core Power
--£Core Heat Flux 500
--acore Flow
- - -4 Vsl Exit Sim Flow
-5 Feed Flow 400 "O
300 Q) 1i:l a::....
0 200 C:
Q) 8 mo Q) a..
\\
\\
\\
0
\\
-100
-200 0.0 1.0 Key Parameters
,..... -,t'
...,~
I.... __,......,.
,,,,,--i-... _..... '
I 2.0 3.0 Time (s)
FSAR REV. 70 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT 4.0 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 Time (s)
FSAR REV. 70 3.0 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT 4.0 SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure lSD.2.2-1-2, Rev. 66
'iii 8
~
- J
(/)
(/)
Q)
Q..
Steam Dome and Plenum Pressure 1400i:=======~:---_,_--~--------+----------t
-1 Steam Dome Pressure
--e Lower Plenum PressJre 1300 1200 1100 1000 +------,-----;----..-----,.-------.----....----....-------+
0.0 1.0 2.0 Time (s)
FSAR REV. 70 3.0 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT 4.0 SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure 15D.2.2-1-4, Rev. 66
~
E
.0 Q) cu a::
5 a:
SRV Flow 4000ir====:=,::::===~~--'----'---~--~--~--~---;
-1 Total SRV Flo 3000 2000 1000 0+--------------------.-----....-----,----
0.0 1.0 2.0 Time (s)
FSAR REV. 70 3.0 SUSQUEHANNA STEAM ELECTRIC STATION UNIT2 FINAL SAFETY ANALYSIS REPORT 4.0 SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS (TYPICAL)
Figure 15D.2.2-1-5, Rev. 66
150 100 0::
'5 50 C: @
- a.
- 0. --
-60 ~
o.o 580.0 570.0 6
iii a,
-' 560.0 550.0 Key Pan;nheters i
-icorePo-,v,er
~~~~\\Aw;
~-Corafto,i/
.-~~VsJ.&tstrnAa.v
~"-"--
-':::~~~--
I I
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I:/
\\ :,,,
\\~
i 6.0 Downcomer Water Level 4.0 6.0 Time(s) 8.0 J
10.0 m
'iii El, fl!
"' £ Pressure 1200 1150 1100 1050 1000 950 2.0 4.0 6.0 8.0 Time (s)
Flow 6000.0 --~-----+--~--+---~---~----~--,.
4000.0 2000.0 0.0
-1VesselBi:it$leamAow
~Twbinefb.Y
-Jfeed.valerffow
---ITotalSRV-
--5TurbtneB I
Row
-~1 i
~-
-2000.0 +--~----+--~--+---~---~--+---~--+
0.0 2.0 4.0 6.0 8.0 10.0 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
I 0
C: =
Key Parameters 80 t----'-----;--------';=_,==a;Co<a=o=aP=..,=,=-~-----t
eeoretteatAo~
- .tCore FkJw
- - -4 Vs! Exit Stm Row
---6FO&<IF":!;_
60
_.a---G--3-------&---/4 -!t----3----s--u--
0-1----,----.-------.--------,-----+
o.o 2.0 4.0 6.0 Time(sl Sensed Water Level 585 't=_=,,,,,,,.,,,=""s=....,.=,,s-------'-----;----'---=-----i 680
! 675 I
570 565+--------------i----------t 0.0 2.0 4.0 6.0 Time Isl I
.0 =
Steam Dome and Plenum Pressure 1020 t,__,==;ae,s~""'*m""0o',ccmss, Prs=...,.,~,,,----'--------'----'-----'----t
--21.o.Ya, Plenum Pressure 1000 940 \\----~-~-----~--~---,-----!
0.0 2.0 4.0 6.0 Time(s}
SRV Flow
-1 TolaJ SRV Flud 0
~ 0+-+-1--1f--+-+-+--1--+--+---t-+->->-+-1--1--
cr:
I 0
-1.J-------------,----,-----~
M
~
~
M 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