ML21294A100
| ML21294A100 | |
| 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: ML21294A100 (51) | |
Text
SSES .. FSAR APPENDIX 15E INITIAL CORE FOR SSES UNITS 1 AND 2 NON-LIMITING EVENTS Rev. 53, 04/99 15E-1
SSES-FSAR Table Rev. 55 TABLE 15E.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNITS 1 AND 2 NON-LIMITING EVENTS (VALUES ARE FOR THE INITIAL CORES ONLY)
Maximum Core Maximum Maximum Maximum Average Number of Maximum Dome Vessel Steam line Surface Heat Valves - Duration Neutron Pressure Pressure Pressure Flux % of Frequency 1st of (1)
Section Figure Description Flux psig psig psig Initial CPR Category* Blowdown Blowdown 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 15.1.1 Loss of Feedwater Heater See Text and Appendices 15C and 15D for current cycle limits 15.1.2 Feedwater Controller Failure See Text and Appendices 15C and 15D for current cycle limits 15.1.3 Pressure Regulator Failure - Open See Text and Appendices 15C and 15D for current cycle limits 15.1.4 Inadvertent Opening of Safety or Relief See Text Valves 15.1.6 RHR Shutdown Cooling Malfunction See Text 15.2 INCREASE IN REACTOR PRESSURE 15.2.1 Pressure Regulator Failure - Closed See Text 15.2.2 15E.2.2-1 Generator Load Reject - Bypass On 281.7 1154 1179 1153 109.7 0.11 a 16 9 sec 15.2.2 Generator Load Reject- Bypass Off See Text and Appendices 15C and 15D for current cycle limits 15.2.3 15E.2.3-1 Turbine Trip - Bypass On 167.2 1143 1167 1132 101.4 0.09 a 16 8 sec FSAR Rev. 64 Page 1 of 3
SSES-FSAR Table Rev. 55 TABLE 15E.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNITS 1 AND 2 NON-LIMITING EVENTS (VALUES ARE FOR THE INITIAL CORES ONLY EXCEPT AS NOTED)
Maximum Core Maximum Maximum Maximum Average Number of Maximum Dome Vessel Steam line Surface Heat Valves - Duration Neutron Pressure Pressure Pressure Flux % of Frequency 1st of (1)
Section Figure Description Flux psig psig psig Initial CPR Category* Blowdown Blowdown 15.2.3 Turbine Trip - Bypass Off See Text and Appendices 15C and 15D for current cycle limits 15.2.4 15E.2.4-1 Inadvertent MSIV Closure (3) 144.3 1250 1281 12501 100.8 0.11 a 12 16 sec estimated 15.2.5 15E.2.5-1 Loss of Condenser Vacuum 167.5 1140 1165 1131 101.3 <0.09 a 13 20 sec 15.2.6 15E.2.6-1 Loss of Auxiliary Power Transformer 104.5 1145 1160 1140 100.1 0.0 a 16 16 sec 15.2.6 15E.2.6-2 Loss of All Grid Connections 107.2 1140 1161 1130 100.1 0.0 a 13 17 sec 15.2.7 15E.2.7-1 Loss of All Feedwater Flow (3) 102.0 1040 1081 1029 102.0 0.0 a 0 0 sec 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 15E.3.1-1 Trip of One Recirculation Pump Motor 103.6 1015 1053 998 100.0 0.0 a 0 0 sec 15.3.2 15E.3.2-1 Trip of Both Recirculation Pump Motors 103.5 1113 1127 1109 100.1 0.0 a 10 28 sec 15.3.3 Seizure of One Recirculation Pump See Text and c (Single Loop Operation) Appendices 15C and 15D for current cycle limits 15.3.4 Recirculation Pump Shaft Break See Text c 15.4 REACTIVITY AND POWER ANOMALIES 15.4.1.1 RWE - Refueling See Text b FSAR Rev. 64 Page 2 of 3
SSES-FSAR Table Rev. 54 TABLE 15E.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 AND 2 NON-LIMITING EVENTS (VALUES ARE FOR THE INITIAL CORES ONLY)
Maximum Core Maximum Maximum Maximum Average Number of Maximum Dome Vessel Steam line Surface Heat Valves - Duration Neutron Pressure Pressure Pressure Flux % of Frequency 1st of (1)
Section Figure Description Flux psig psig psig Initial CPR Category* Blowdown Blowdown 15.4.1.2 RWE - Startup See Text b 15.4.2 RWE - At Power See Text a 15.4.3 Control Rod Maloperation See Subsections 15.4.1 and 15.4.2 15.4.4 15E.4.4-1 Startup of Idle Recirculation Loop 323.4 973 988 967 134.9 (2) a 0 0 15.4.5 Recirculation Flow Controller Failure See Text and Appendices 15C and 15D for current cycle limits 15.4.7 Misplaced Bundle Accident See Text 15.5 INCREASE IN REACTOR INVENTORY 15.5.1 15E.5.1-1 (3) 112 1039 1079 1028 112 0.18 a 0 0 Inadvertent HPCI Pump Start 15.5.3 BWR Transients That Increase Reactor See Sections Coolant Inventory 15.1 and 15.2
- Frequency a = Moderate b = Infrequent c = Limiting Faults Notes I
- 1. CPRs values in this Table are for the initial cores for Units 1 and 2 and are non-limiting. SEE TEXT.
- 2. Event initiated from low power levels - Initial core MCPR safety limit 1.06 is not violated
- 3. The event was re-analyzed for 3952 MWt rated power. The results show that the vent is non-limiting and is therefore reported in this Appendix.
FSAR Rev. 64 Page 3 of 3
SSES-FSAR TABLE 15E.0-2
. INPUT.PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS
- UNITS 1 AND 2 INITIAL CYCLES
- 1. Thermal Power Level, Mwr Rated Value (NBR) 3293
-Analysis Value (104.4% NBR) 3439
- 2. Steam Flow, Mlbs/hr
- Analysis Value (105% NBR) 14.153
- 3. Core Flow Mlbs/hr 1 100.0
- 4. Feec:twater Flow Rate, lbs/sec Analysis Value 3921
- 5. Feedwater Temperature, °F 386.9
- 6. Vessel Dome Pressure, psig 1020.0
- 7. Vessel Core Pressure. psig 1030.0
- 8. Turbine bypass Capacity, %NBR 25%
- 9. Core Coolant Inlet Enthalpy, BTU/lb 521 .1
- 10. Tur~ine I.nlet Pressure, psig 960.0
- 11. Fuel Lattice 8x8
- 12. Core Average Gap Conductance, BTU/sec-ff-°F 0.1744
- 13. Core Leakage Flow,% 9.85
- 14. Required MCPR Operating Limit Not Applicable to current cycles
~CPRs are non-limiting '
- 15. MCPR Safety Limit 1.06
- 16. Doppler Coefficient0 Nominal EOC-1 (-)cents/°F 0.2255 Analysis Data 0.2142 Rev. 53. 04/99 Page 1 of 3
SSES-FSAR
~
- TABLE-15E.0-2 (Cont'd)
INP_UT PARAMETERS ANO INITIAL CONDITIONS FOR TRANSIENTS UNITS 1 AND INITIAL CYCLES
- 17. Void Coefficient.,.
Nominal EOC-1 (-)cents/°F 7.48 Analysis Data for Power Increase Events 12.0 Analysis Data for Power Decrease Events 6.61
- 20. Control Rod Drive Speed, See Figure 15E.0-2 Position Versus Time
- 21. Jet Pump Ratio, M 1.84
- 22. Safety Relief Valve Capacity,% NBR@ 1091 psig 99.0 Manufacturer Crosby Number installed 16
- 23. Relief Function Delay, sec 0.4
- 24. Relief Function* Response, sec <0.15
- 25. Set Points for Safety/Relief Valves, psig 2@1110 3@1140 4@1120 3@1150 4(@1130
- 26. Number of Valve Groups Simulated 5
- 27. High Flux Trip, %NBR Analysis set point 125.3
- 28. High Pressure Scram Set Point, psig 1071
- 29. Vessel Level Trips, Inches Above(+), Below(-) Dryer Skirt High Level (L8):S 58.7 Bottom, Tech Spec settings (L4)~ 30 Low Level (L3)~ 12.5 Low Low Levet (L2)~ -38
- 30. APRM Thermal Trip Set Point,%NBR 1125.0 Rev. 53, 04/99 Page 2 of 3
SSES-FSAR TABLE 1SE.0-2 (Cont'd)
INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS .
UNITS -1 AND 2 - INITIAL CYCLES
- 31. Recirculation Pump Trip Delay! sec 0.175
- 32. Recirculation Pump Trip Inertia time constant for Analysis*, sec 4,5
- The inertia time constant is defined by the* expression:
t= (2xJon)/QTo Where: t = Inertia time constant, sec Jo= Pump Motor Inertia, lb-ft2 n = Rated pump speed, rps g = gravitational constant ft/sec2 To = Pump shaft torque, lb-ft
...... Parameters used in REDY only. ODYN values are cal~ulated within the code for equilibrium cycle conditions.
Rev. 53, 04/99 Page 3 of 3
SSES-FSAR TABLE 15E.1.1-1 OPERATOR ACTIONS WHEN REACTOR SCRAM IS INCURRED
- Place Mode switch to shutdown
- Confirm all rods have inserted and power is decreasing
- Insert neutron monitoring detectors
- Confirm main turbines trip and main generator lockout.
- Monitor and control reactor pressure
- Monitor and control reactor water -level
- Cool down the reactor per standard procedure (if required)
- Monitor entry condition parameters for other Emergency Operating Procedures Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 15E.1.4-1 SEQUENCE OF EVENTS FOR INADVERTENT SAFETY RELIEF VALVE OPENING TIME (SEC.) EVENT 0 Opening of 1 safety relief valve, reaches full flow and remains open throughout the event.
1,200 Reactor scrammed on high suppression pool temperature .
Technical Specification limit of 11 o*F.
Closure of all MSIVs.
Two loops of RHR suppression pool cooling placed into service.
5,200 Reactor depressurization initiated on high suppression pool temperature Technical Specification limit of 12o*F.
59,200 Reactor depressurized to 14.7 psia, terminating blowdown through safety relief valve.
Rev. 54, 10/99 Page 1 of 1
SSES-FSAR TABLE 15E.1.4-2 SAFETY RELIEF VALVE OPENING EVENT ACTIVITY ABOVE SUPPRESSION POOL (curies)
Isotope Realistic Design Basis 1-131 . 1.35E-Q3<1> 5.40E-01 132 3.64E-02 6.19E+00 1-133 1.90E-02 3.81E+00 r-134 9.27E-02 1.54E+01 1-135 2.23E-02 5.87E+00 Kr-83m 1.44E+01 2.93E+00 Kr-85m 2.54E+00 5.25E+00 Kr-85 7.94E-03 1.72E-02 Kr-87 8.73E+00 1.72E+01 Kr-88 8.73E+00 1.72E+01 Kr-89 5.40E+01 1.12E+02 Xe-131m 6.19E-03 1.29E-02 Xe-133i-n 1.19E-01 2.50E-01 Xe-133 3.33E+00 7.06E+00 Xe-135m 1.11E+01 2.24E+01 Xe-135 9.53E+00 1.89E+01 Xe-137 6.19E+01 1.29E+02 Xe-138 3.65E+01 7.67E+01
- 1. 1.35E-03 = 1.35 x 1Q-03 Rev. 54, 10/99 Page 1 of 1
SSES-FSAR TABLE 15E.1.4-3 SAFETY RELIEF VALVE OPENING EVENT ACTIVITY RELEASED TO THE ENVIRONS (curies)
Isotope Realistic Design Basis 1
1-131 1.31 E-05( > 5.24E-03 1-132 3.35E-05 5.71 E-03 1-133 1.46E-04 2.92E-02 1-134 . 1.52E-06 2.52E-04 1-135 9.B0E-05 2.58E-02 Kr-83m 7.31 E-01 1.48E-01 Kr-85m 7.18E-01 1.48E+00 Kr-85 7.94E-03 1.72E-02 Kr-87 1.10E-01 2.17E-01 Kr-88 1.20E+00 2.37E+00 Kr-89 1.65E-44 3.43E-44 Xe-131m 6.07E-03 1.27E-02 Xe-133m 1.0BE-01 2.35E-01 Xe-133 3.20E+00 6.91E+00 Xe-135m 1.68E-03 4.42E-01 Xe-135m 5.37E+00 1.27E+01 Xe-137 4.13E-36 8.62E-36 Xe-138 1.12E-07 2.36E-07
- 1. 1.31 E-05 = 1.31 x 10'°5 Rev. 54. 10/99 Page 1 of 1
SSES-FSAR TABLE 15E.1.4-4 SAFETY RELIEF VALVE OPENING EVENT OFFSITE RADIOLOGICAL DOSES (rems)
Total Whole Thyroid Source Terms Body Gamma Inhalation Realistic 1.12E-Q6<1> 2.65E-08 Design Basis 2.43E-06 7.16E-06
- 1. 1.12E-06 = 1.12 x 10-08 Rev. 54, 10/99 Page *1 of 1
SSES-FSAR TABLE 15E.1.6-1 SEQUENCE OF EVENTS FOR INADVERTENT RHR SHUTDOWN COOLING OPERATION APPROXIMATE ELAPSED TIME EVENT 0 Reactor at states 8 or D (of Appendix 15A) when RHR shutdown cooling inadvertently activated.
- 0-10 min. Slow rise in reactor power.
+/-10 min. Operator may take action to limit power rise. Flux scram will occur if no action is taken.
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.2-1 SEQUENCE OF EVENTS FOR FIGURE 15E2.2-1 GENERATOR LOAD REJECTION, BYPASS ON TIME, SEC EVENT
(-)0.015 (approx.) Turbine-generator detection of loss of electrical load.
0 Generator lockout relays act to initiate turbine control fast valve closure.
0 Turbine-generator PLU trip initiates main turbine bypass system operation.
0.016 Fast control valve closure (FCV) initiates scram trip 0.016 Fast control valve closure (FCV) initiates a recirculation pump trip (RPT).
0.07 Turbine control valves closed.
0.11 Turbine bypass valves start to open.
1.145 Group 1 relief valves actuated.
1.160 Group 2 relief valves actuated.
1.356 Group 3 relief valves actuated.
1.520 Group 4 relief valves actuated.
1.789 Group 5 relief valves actuated.
FSAR Rev. 61 Page 1 of 1
SSES-FSAR TABLE 15E.2.3-1 SEQUENCE OF EVENTS FOR TURBINE TRIP WITH BYPASS OPERABLE FIGURE 15E.2.3-1 TIME, SEC EVENT 0 Turbine trip initiates closure of main stop valves.
0 Turbine trip initiates bypass operation.
0.01 Main turbine stop valves reach 90% open position and initiate reactor scram trip.
0.01 Main turbine stop valves reach 80% open position and initiate a recirculation pump trip (RPT) 0.1 Turbine stop valves closed.
0.1 Turbine bypass valves start to open to regulate pressure.
1.7 Group 1 relief valves actuated.
1.9 Group 2 relief valves actuated.
2.2 Group 3 relief valves actuated.
2.4 Group 4 relief valves actuated.
2.6 Group 5 relief valves actuated.
4.3 L8 vessel level set point trips feedwater pumps.
10.1 Group 1 relief valves close.
44.8 Vessel water level decreases to L2 vessel level set point. \
75 (est.) HPCI/RCIC flow enters vessel not simulated).
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.4-1 SEQUENCE OF EVENTS FOR MSIV CLOSURE, FIGURE 15E.2.4-1 TIME, SEC EVENT 0 Initiate closure of all main steam line isolation valves (MSIV).
0.30 MSIVs reach 90%* open.
0.36 MSIV position trip scram initiated.
- Group 1 safety valves assumed to be out of service 3.7 Group 2 safety valves open.
3.8 Group 3 safety valves open.
~20 All safety valves reclose (estimate).
23.0 Group 1 pressure relief valves reopen.
29.0 Group 1 pressure relief valves reclose.
36.0 Group 1 pressure relief valves reopen.
40.0 Group 1 pressure relief valves reclose.
Valves opening based on safety settings is conservative
- Changed to 85% with no significant impact on transient results.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR TABLE 1SE.2.5-1 TYPICAL RATES OF DECAY FOR CONDENSER VACUUM CAUSE
- ESTIMATED VACUUM DECAY RATE I
- 1. Failure or Isolation of < 1 inch Hg/minute Steam Jet Air Ejectors I
- 2. Loss of Sealing Steam to ~1 to 2 inches Hg/minute Shaft Gland Seals I
- 3. Opening of Vacuum Breaker Valves
~2 to 12 inches Hg/minute I
- 4. Loss of One or More -4 to 24 inches Hg/minute Circulating Water Pumps I Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 15E.2.5-2 LOSS OF CONDENSER VACUUM SEQUENCE OF EVENTS FOR FIGURE 15E.2.5-1 TIME. SEC EVENT
-0.0 (est} Initiate simulated loss of condenser vacuum at 2 inches of Hg per second.
0.0 (est) Low condenser vacuum main turbine trip actuated.
0.0 (est) Low condenser vacuum feedwater trip actuated.
-0.01 (est) Main turbine trip initiates reactor scram.
0.01 (est) Main turbine trip initiates recirculation pump trip (RPT)
.1 (est) Turbine stop valve closes .
1.7 Group 1 relief valves set points actuated.
1.9 Group 2 relief valves set points actuated.
2.2 Group 3 relief valves set points actuated.
2.4 Group 4 relief valves set points actuated.
6.5 Low condenser vacuum initiates main steam line isolation valve closure.
6.5 Low condenser vacuum initiates bypass valve closure.
21.5 Group 1 relief valves close.
23.5 Vessel water level decreases to L2 vessel level set point.
53 (est) HPCI/RCIC system flow enters vessel (not included in simulation) .
90+ Relief valves cycle as required on pressure.
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.5-3 TRIPS SIGNALS ASSOCIATED WITH LOSS OF CONDENSER VACUUM VACUUM PROTECTIVE ACTION INITIATED (INCHES OF HG) 27 to 28 Normal Vacuum Range 21.7 Main Turbine Trip (Stop Valve Closure) 10.2 Main Steam Line Isolation Valve (MSlV) Closure 7 Mainsteam Turbine Bypass Valves Closure Reactor Feed Pump Turbine Trip (Stop Valves 17.4 Closure)
FSAR Rev. 56
SSES-FSAR TABLE 15E.2.6-1 LOSS OF AUXILIARY POWER SEQUENCE OF EVENTS FOR FIGURE 15E.2.6-1 TIME, SEC EVENT 0 Loss of auxiliary power transformer occurs.
0 Recirculation system pump motors are tripped.
0 Condensate booster pumps are tripped.
0 Condenser circulating water pumps tripped.
2.0 Closure of main steamline isolation valves initiated.
2.0 Reactor scram initiated.
4.0 Feedwater turbines tripped off.
4.4 Group 1 Safety/Relief valves actuated.
21 Group 1 Safety/Relief valves closed.
32 Initiate HPCI and RC(C operation (not simulated).
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.6-2 LOSS OF ALL GRID CONNECTIONS SEQUENCE OF EVENTS FOR FIGURE 15E.2.6-2 TIME, SEC EVENT
(-)0.015 (approx.) Loss of Grid causes turbine-generator to detect a loss of electrical load.
0 Control valve fast closure.
0 Turbine-generator trip initiates main turbine bypass system operation.
0 Recirculation system pump motors are tripped.
0 Fast control valve closure (FCV) initiates a reactor scram trip.
0 Initiation of standby AC power systems.
0.1 Turbine bypass valves open.
0.15 Turbine control valves closed.
1.2 Group 1 relief valves actuated.
1.4 Group 2 relief valves actuated.
1.5 Group 3 relief valves actuated.
1.7 Group 4 relief valves actuated.
2.0 MSIV's start to closure.
4.0 Feedwater turbines tripped off.
18.7 Group 1 safety relief valves close.
37.2 Initiate Containment Isolation, HPCI and RCIC operation, (L2) (not simulated).
FSAR Rev. 61 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.7-1 LOSS OF FEEDWATER FLOW SEQUENCE OF EVENTS FOR FIGURE 15E.2.7-1 TIME, SEC EVENT 0 Trip of all feedwater pumps initiated.
5.0 Feedwater flow decays essentially to zero.
5.1 Recirculation pumps runback, low feed water flow 6.8 Vessel water level (L3) trip initiates scram trip.
52.8 Vessel water level (L2) trip initiates recirculation pump system trip.
52.8 Vessel water level (L2) trip initiates containment isolation.
82.8 Vessel water level (L2) trip initiates RCIC operation -
(30 sec. delay) (HPCI not simulated)
The MSIVs will not close until water level reaches L1. Water level is not expected to reach L1 during this event since RCIC initiates at L2.
SRVs Do Not Open FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.9-1 SEQUENCE OF EVENTS FOR FAILURE OF RHR SHUTDOWN COOLING TIME EVENT 0 Reactor is operating at 102% of rated thermal power when LOP transient occurs initiating plant shutdown 0 Concurrently loss of one division of power occurs 10 min. Controlled depressurization initiated (100°F/hr) and continues until vessel pressure reaches approximately 115 psia 15 min. Operators initiate suppression pool cooling 140 min. When vessel pressure reaches 115 psia, a failure in a shutdown cooling suction valve prevents operation of normal shutdown cooling 170 min. Operator initiates core spray for use with alternate shutdown cooling. Operator opens ADS valves to achieve continuous core flow 6.5 hrs Peak suppression pool temperature is attained FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 54 TABLE 15E.2.9-2 INPUT PARAMETERS FOR EVALUATION OF FAILURE OF RHR SHUTDOWN COOLING Core Thermal Power (MWt) 4031 Initial RPV Pressure (psia) 1050 Initial Vessel Temperature (°F) 550 Suppression Pool Temperature (°F) 90 Suppression Pool Liquid Volume (ft3) 115,810 Service Water Temperature (°F) 97 RHR Heat Exchanger K-value (Btu/sec-°F) 317.5 RHR Pool Cooling Flow Rate (gpm) 9750 Core Spray (1 Loop) Flow Rate (gpm) 7900 FSAR Rev. 64 Page 1 of 1
SSES-FSAR TABLE 15E.3.1-1 SEQUENCE OF EVENTS FOR TRIP OF ONE RECIRCULATION PUMP TIME, SEC EVENT 0 Trip of one recirculation pump initiated.
5.7 Diffuser flow decreases significantly in the tripped loop.
30.0 Core flow stabilizes at new equilibrium conditions.
42.0 Power level stabilizes at new equilibrium conditions Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 15E.3.1-2 SEQUENCE OF EVENTS FOR TRIP OF TWO RECIRCULATION PUMPS TIME, SEC EVENT 0 Trip of both recirculation pumps initiated.
4.0 Vessel water level (LB) trip initiates turbine trip.
4.0 Feedwater pumps are tripped off.
4.0 Turbine trip initiates bypass operation.
4.0 Turbine trip initiates re~ctor scram trip.
7.0 Group 1 pressure relief valves open .
12.0 Group 1 pressure relief valves closed .
46.0 L2 vessel level set point initiate HPCI and RCIC.
76 HPCI/RCIC flow enter vessel (not simulated).
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 15E.4.4-1 SEQUENCE OF EVENTS ABNORMAL STARTUP OF IDLE RECIRCULATION PUMP FOR FIGURE 15E.4.4-1 TIME, SECOND EVENT 0 Start pump motor. I 9.0 Startup loop flow starts to increase significantly. I 10.0 Reactor high flux scram initiated. l 11.0 Vessel level reaches (L4) Low Level Alarm. I 23.5 Vessel level reaches (L3) Low Level Scram Trip. 1*
35.0 Diffuser flows and pressures begin to stabilize. I 50.0 Vessel level begins to stabilize. l I
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR Table Rev. 2 TABLE 15E.5.1-1 SEQUENCE OF EVENTS FOR INADVERTENT STARTUP OF HPCI Time, sec. Event 0.0 Simulate HPCI cold water injection 1.0 Full flow established for HPCI No reactor scram No recirculation pump trip No SRV flow 50 Reactor variables settle into new steady state FSAR Rev. 64 Page 1 of 1
160.---------------------------..
140 120
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~
0 0::::
w DESIGN FLOW CONTROL LINE 0 80 TYPICAL POWER FLOW a.. DERATE LINE
...J
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I 60 I-40 s:0 0 w
w
...J LL a..
Cl) w a..
0::: ~
0 ::J CAVITATION REGION () a..
0 0 20 NATURAL (PUMP) CIRCULATION w w I- I-
~
<(
0:::
MINIMUM PUMP SPEED J' H H' 0
0 40 60 80 100 120 CORE FLOW (% rated)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TYPICAL POWER/FLOW MAP FIGURE 15E.0-1, Rev 54 AutoCAD: Figure Fsar 15E_0_1.dwg
1 0 0 - - - - - - - - - - - - - - - - - - - - - - - - - 45 SCRAM REACTIVITY VERSUS TIME SUSQUEHANNA, EOC-1 CONTROL ROD DRIVE VERSUS TIME FEBRUARY 09, 1977 90 40 80 C CRD POSITION IN PERCENT 35 1 NOMINAL SCRAM CURVE IN (-$)
2 SCRAM CURVE USED IN ANALYSIS 70 30 60 ,......_
th 25 ~
'-' 5 z 50 i=
(.)
0 <(
i= w ci5 a:::
0 20 c..
40 15 30 10 20 10 5 ol..L.~~---1-...L...-.....I._Jo 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 TIME (sec)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SCRAM POSITION AND REACTIVITY CHARACTERISTICS FIGURE 15E.0-2, Rev 54 AutoCAD: Figure Fsar 15E_0_2.dwg
1 NEUTRON FLUX 1 VESS cl PRES RISE (PSI) 2 PEAK FUEL CENTER TEMP 2 STM LINE PRES RISE (PSI) 3 AVE. SURFACE HEAT FLUX 3 SAFE' -Y VALVE FLOW(%)
ISO. t - - t - - t - - - - t - - - - - - - - - + 4 FEEDWATER FLOW 300.1------L-----l...---~ 4 RELIEF VALVE FLOW(%)
5 VESEL STEAM FLOW 5 BYPA ,S VALVE FLOW(%)
6 TURB NE STEAM FLOW (%)
6' w
~
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200.~---~-----1-----1-------+----
0 1-z w
0 w
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II. 6. 8. 8.
"* 6.
TIME (SEC.) TIME (SEC.)
1 LEVEL (INCH-REF-SEP-SKIRT) 1 VOID R cACTIVITY 2 W R SENSED LEVEL (INCHES) 2 DOPPL cR REACTIVITY 3 N R SENSED LEVEL (INCHES) 3 SCRArl REACTIVITY 200. I. 4 TOTAL '<EACTIVITY 4 CORE INLET FLOW (%)
5 DRIVE FLOW 1 (%)
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TIME (SEC.) TIME (SEC)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECTION, WITH BYPASS ON FIGURE 15E.2.2-1, Rev 54 AutoCAD: Figure Fsar 15E_2_2_1.dwg
l NEUT ON FLUX l E CPSIJ 2 AVE FUEL TEMP ~°F*10) 2 [SE CPS[)
3 AVE URFACE HE FLUX 3 SE (PSI) 4 FEED ATER FLOW II ow C,
5 VESS L STEAM F ow ow w 150 250 I-a O::'.
LL 0
100 150 I-z w
u O::'.
w Q_ 50 50 0 -50 0 10 20 TlME 1 CORE INLET SUB OOLINB 2 W R ENSED LEV L( nlCHFS) 3 N R ENSED LEV L([NCHES) 4 CORE INLET FLO (,., .....
5 ORIV FLOW 1 ( J 150 0 120 w
t-er O::'.
lJ..
0 80 100 I-z w
CJ O::'.
LLJ 50 a.. 1.10 II I
0
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- - __l___~-:.k...1 ..,.-:-=.
2-, - 3:, " 0 50 ;'5 100 TIME (SEC) CORE FLOW (%)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TURBINE TRIP, TRIP SCRAM, BYPASS AND RPT-ON FIGURE 15E.2.3-1, Rev 54 AutoCAD: Figure Fsar 15E_2_3_1.dwg
Closure of All MSIVs 160 140 Core Power
"'C 120
.....ca (1) 100
~ 80
.....C:
(1) 60
(.)
(1) 40 Q. Steam Flow 20 0
-20 0 1 2 3 4 5 6 7 Time (Seconds)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT CLOSURE OF ALL MSIVs FIGURE 15E.2.4-1, Rev 55 AutoCAD: Figure Fsar 15E_2_4_1.dwg
1 NEUT OH FLUX 1 VE!IS L PRES R[ E CP8rl 2 AVE FUEL TEMP (°F*lO) 2 sr11 [NE ~RES [SE CPS[)
3 AVE URFACE HE r FLUX 3 TURB HE PRES R SE CPSD LI FEE LI REU F ilALVE F ow 5 YES ow S SAFE Y VALVE ow a L60 269 6 BYPA S FLOW w
I-a ex I.L..
C 168 L00 I-z w
u
~
w se 60 a..
0 -6e e ae ~0 CSEC) l CORE INLET SUB OOUNG l 2 W R ENSED LE\' LC INCHES) 2 IJ)(
3 N R ENSED LEV LC INCHES)
LI CORE INLET FLO s DRIV FLOW 1 ( )
(l:J ,,.,
160 0 120 w
I-er
°'
~
0 80 100 I-z.
IJJ
<J a:
UJ 60 0..
..... ~"
e 0 0 60 75 1,0 CORE FLOW C:, )
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF CONDENSER VACUUM AT 2 INCHES PER SECOND FIGURE 15E.2.5-1, Rev 54 AutoCAD: Figure Fsar 15E_2_5_1.dwg
1 NEUTRON FLUX 1 VESSEL F'RES RISE (PSI) 2 PEAK FUEL CENTER TEMP 2 STM LINE PRES RISE (PSI) 3 AVE. SURFACE HEAT FLUX 300. t - - - - - t - - - - - - t - - - - - - - + 3 TURBINE l~RES RISE (PSI) 150. 4 FEEDWATER FLOW 4 CORE INLET SUB (BTU/LB) 5 VESSEL STEAM FLOW 5 CORE AVlc. VOID FRAC (%)
6 TURBINE STEAM FLOW (%)
0 w
I-
<( 200.f------+-----f------+-----t----
O'.
IL 0
I-z w
0 O'.
w
~o.
~
- 20. 30. 40.
TIME (SEC) 1 LEVEL (INCH-REF-SEP-SKIRT) 1 NEUTRON FLUX 2 W R SENSED LEVEL (INCHES) 2 SURFACE HEAT FLUX 200. ____________ ------t 3 N R SENSED LEVEL (INCHES) 120. t - - - - - + - - - - - - l f - - - - - - l - - - - - t - - - - -
4 CORE INLET FLOW (%)
5 DRIVE FLOW 1 (%)
0 w
,------ ----- ~ 80. ------t------t ---~---1----1----+----
IL 0
1-z w
0
- 0. ffi 40.
~
-100. ~~.................., _____
- 0. 10. 2-.- 30. 40.
TIME (SEC)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF AUXILIARY POWER TRANSFORMER FIGURE 15E.2.6-1, Rev 54 AutoCAD: Figure Fsar 15E_2_6_1.dwg
1 NEUTRON FLUX 1 VESSEL Pl tES RISE (PSI) 2 PEAK FUEL CENTER TEMP 2 STM LINE F'RES RISE )PSI) 150.t-----t---- ____ 3 4
AVE. SURFACE HEAT FLUX FEEDWATER FLOW 200. ----~-----i---- 3 TURBINE F RES RISE (PRI) 4 CORE INLET SUB (BTU/LB) 5 VESSEL STEAM FLOW 5 CORE AVE VOID FRAC (%)
6 TURBINE E TEAM FLOW (%)
0 UJ
~
IL 0
f-z UJ
~
UJ e:,
-100._~~~~~-----="c-----~----~--~~
- 10. 20. 30. 40. o. 10. 20. 30. ~ -
TIME (SEC) TIME (SEC) 1 LEVEL (INCH-REF-SEP-SKIRl) 1 NEUTRON FLUX 2 WR SENSED LEVEL (INCHES) 2 SURFACE HET FLUX 3 N R SENSED LEVEL (INCHES) 100. 120.
4 CORE INLET FLOW(%)
5 DRIVE FLOW 1 (%)
0 UJ f-50 <( 60.
0::
IL 0
f-z
~ UJ
()
- o. 0::
UJ 110.
e:,
-50.
- o. 10. 2D. 30. 40.
-* a.o. .
TIME (SEC)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF ALL GRID CONNECTIONS FIGURE 15E.2.6-2, Rev 54 AutoCAD: Figure Fsar 15E_2_6_2.dwg
Loss of Feedwater Flow 50 25 - ~ Analytic L4 - Low Level Alarm Level 25.3 inches 0
"1 Analytic L3 - Low Ill a, -25 IDowncomer Water Level I Level 8.0 inches
.c u
.5
-50 v-- - ......._ .,, Low Level I
Analytic L2 - Low
~
a,
...a,
"\ J
\J - ~
-58 inches
.. -75 Nominal L 1- Low
~
-g -100 Low Low Level a,
Ill C
a, ti) -125 *
/ -129 inches
-150 J Top of Active Fuel I
/'
-175
-200 0 200 400 600 800 1000 1200 1400 Time (Seconds)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF FEEDWATER SENSED WATER LEVEL FIGURE 15E.2.7-1, Rev 55 AutoCAD: Figure Fsar 15E_2_7_1.dwg
Loss of Feedwater Flow 120 110 100 ~ V Jcore Flow I
\
90 \'¥ "C
80 I
l \ \
CD (U 70 \
a:
....C 60 \
Q)
(J CD 50 \ ~-
0.
40 '
.,,.--HHeat Flux I \
30 20 ~ Power II
\
k'-- v-10 0
0 25 50 75 100 125 150 175 Time (Seconds)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF FEEDWATER CORE POWER, HEAT FLUX AND CORE FLOW FIGURE 15E.2.7-2, Rev 1 AutoCAD: Figure Fsar 15E_2_7_2.dwg
Loss of Feedwater Flow 110 r,.
100
\i Vessel Steam Flow I 90 i l
80 !
'C sca 70
\
l l
a:
C 60 l l'
Cl) 50 \
~ I\
Cl) 40 ! Feedwater Flow :
- a. \
1_ _.,,,,
30
~\ RCIC FLOW I 20 10
'\
l I 0
V'--- -/ -- ~-~* ....,., ... _
I I I I I I I I I 0 25 50 75 100 125 150 175 200 225 250 Time (Seconds)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT LOSS OF FEEDWATER FEEWATER FLOW, STEAM FLOW AND RCIC FLOW FIGURE 15E.2.7-3, Rev 1 AutoCAD: Figure Fsar 15E_2_7_3.dwg
100 psi, 330-F TO 14.7 psi, 125"F I 1050 psia, 550°F TO 100 psi, 330"F .. I.. NO OFF SITE POWER C2 CS LOOP A ADS VALVE RHR LOOP A
~ (DIVISION 1 FSAR REV.65 AutoCAD: Figure Fsar 15E_2_9_1_1.dwg
~
FIGURE 15E.2.9-1-1, Rev 55 1., AVAILABLE)
LOSS OF ~~
OFF-SITE c_,
DEPRESSURIZE NORMAL SUSQUEHANNA STEAM ELECTRIC STATION POWER AUTOMATIC VESSEL VIA SHUTDOWN TRANSIENT RELIEF INITATED ADS/RELIEF VALVE VALVE ACTUATION P=1050 psia ACTUATUION T=550"F +RHR SUPPRESSION ADS/RHR COOLING LOOPS POOL COOLING ()_
1/0-~ C1 0-1,, .--C_S_L_O_O_P_B....,
UNITS 1 & 2 7 ,(' ADS VALVE
..... (/) ~fs-----1 RHR LOOP B 01 (b (b (b (DIVISION 2 is.:>., AVAILABLE)
<0~*
.....le-,
FINAL SAFETY ANALYSIS REPORT I~
(,.I
....... 01 0
-, fT1 .
- J N
0 <D r+
(b I
!ll I N
R>
NOTES FOR FIGURE 1SE.2 .9 . 1-1 I ACTIVITY A I Initial pressure = 1050 psia initial temperature = 550°F For purposes of this analysis, the followinng worst-case conditions are assumed to exist.
(1) the reactor is assumed to be operating at 100% nuclear boiler rated steam flow; (2) a loss of power transient occurs; (3) a simultaneous loss of onsite power (Division 1 or Division 2), which eventually results in the operator not being able to open one of the RHR shutdown cooling line suction valves.
I ACTIVITY B I Initial system pressure = 1050 psia initial system temperature = 550°F Operator Actions During approximately the first 30 minutes, reactor decay heat is passed to the suppression pool by the automatic operation of the reactor relief valves. Reactor water level will be returned to normal by the HPCI and RCIC system automatic operation.
After approximately 10 minutes, it is assumed one RHR heat exchanger will be placed in the suppression pool cooling mode to remove decay heat. The operator will then initiate depressurization of the reactor vessel to control vessel pressure. Controlled depressurization procedures consisit of controlling vessel pressure and water level by using the ADS, RCIC and/or HPCI systems.
When the reactor pressure approaches 100 psig, the operator would normally prepare for operation of the RHR system in the shutdown cooling mode.
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NOTES FOR FIGURE 15E.2.9-1-1 FIGURE 15E.2.9-1-2, Rev 55 AutoCAD: Figure Fsar 15E_2_9_1_2.dwg
NOTES FOR FIGURE 1SE.2.9.1-1 IACTIVITY C 1 i (Division 1 fails, Division 2 available) (Figure 15E.2.9-5)
System pressure - 100psi System temperature - 330°F Operator Actions The operator establishes a closed cooling path as follows:
(1) One ADS valve (DC Division 2) is powered open; (2) Water is pumped from the suppression pool into the reactor vessel. The cooled suppression pool water picks up decay heat and flows out of the vessel through the open ADS valves and back to the suppression pool as shown in Figure 15E.2.9-5. The RHR B loop is used to cool the suppression pool as required.
(Division 2 fails, Division 1available) (Figure 15E.2.9-6)
System pressure - 100psi System temperature - 330°F Operator Actions The operator establishes a closed cooling path as follows:
(1) One ADS valve (DC Division 1) is powered open; (2) Water is pumped from the suppression pool and into the reactor vessel as shown in Figure 15E.2.9-6. The cooled suppression pool water picks up decay heat and flows out of the vessel through the open ADS valves and back to the suppression pool. The RHR loop is used to cool the suppression pool as required. Cold shutdown (P=14.7psia. T RPV=200°F is reached in approxiamately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NOTES FOR FIGURE 15E.2.9-1-1 FIGURE 15E.2.9-1-3, Rev 55 AutoCAD: Figure Fsar 15E_2_9_1_3.dwg
1050psia, 550'F TO 100 psi, 330'F 100psi, 330'F TO 147 psi, 125'F
.. 1..
DEPRESSURIZE NORMAL SHUTDOWN VESSEL TO COOLING MAIN CONDENSER (OFFSITE POWER) w
....I ID
- 5
~
<(
~
CONDENSER DEPRESSURIZE VESSEL w ~~ ADS/RELIEF VALVE 3'; NOT AVAILABLE VIA MANUAL RELIEF ALTERNATE PATH 0
Q. VALVE ACTUATION + RHR w SUPPRESSION POOL
!::: COOLING
(/)
FSAR REV.65 AutoCAD: Figure Fsar 15E_2_9_2.dwg FIGURE 15E.2.9-2, Rev 55 LL.
LL.
0 A D NORMAL SHUTDOWN NORMAL COLD SUSQUEHANNA STEAM ELECTRIC STATION COOLING SHUTDOWN (DIV. 1 & 2 POWER) SHUTDOWN INITIATED ACHIEVED F=1050 psia (VESSEL HEAD T=550'F REMOVED)
SUMMARY
OF PATHS wl-UNITS 1 & 2 1-Z ADS/RELIEF VALVE
-W
~!::: , o'<_ ATLERNATE PATH AVAILABLE TO ACHIEVE LL. (/)
oz LL.~
ol-H
(/)~
COLD SHUT DOWN C/lw 03';
FINAL SAFETY ANALYSIS REPORT
....J~I I DEPRESSURIZE VESSEL AUTOMATIC RELIEF VALVE VIA MANUAL RELIEF VALVE ACTUATION + RHR F FAILURE OF DIVISION 1.. I ADS/RELIEF VALVE ALTERNATE PATH ACTUATION SUPPRESSION POOL COOLING
1200 ~ - - - - - - - - - - - - - - - - - - - - - ~ 600
- - Reactor Vessel Pressure
\
I - - - - Reactor Vessel 1000 I Temperature 500 U:-
c,
\ CII e
- l 800 I 400 8.
2!
I I
,n :::s
,n \
I!!
e D..
I I ~
Cl) 600 300 i 1-
,n \\ ____ _ ai
,n
~... ---------- ----- ----- -*--- >
i
.su
-0
( .)
ca Cl) 0::
400 ---- -- --- - -- 200
~
Ill 200 100 o-i-~::=:=:==:;:==========::;:::::=============;=====~o 0 10 20 30 40 Time (hr)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT VESSEL TEMPERATURE AND PRESSURE VERSUS TIME (ACTIVITY C1 OR C2)
FIGURE 15E.2.9-3, Rev 55 AutoCAD: Figure Fsar 15E_2_9_3.dwg
LL c, 200 CD C
f
~
CD C.
150 E
CD t-0 0
- a. 100 C:
0
,n
,n f
C.
C.
50 en 0 10 20 30 40 Time (hr)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SUPPRESSION POOL TEMPERATURE VERSUS TIME (90° SERVICE WATER TEMPERATURE)
CAPACITY C1 OR C2 FIGURE 15E.2.9-4, Rev 55 AutoCAD: Figure Fsar 15E_2_9_4.dwg
...-*----1**1----l+---..;..,POOL SUPPRESSION REACTOR J
--++---
AHR B 1-H-;;T-I EXCHANGER I
RHR SERVICE SERVICE WATERSYSTEM--+------1--i WATER DISCHARGE I I L ____ J FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT ACTIVITY C1 ALTERNATE SHUTDOWN COOLING PATH UTILIZING RHR LOOP B FIGURE 15E.2.9-5, Rev 54 AutoCAD: Figure Fsar 15E_2_9_5.dwg
REACTOR SUPPA ESSION
_____.,_*---*~------+11-----___:_P,ooL CORE SPRAY A AHR A 1-H-;;T -
l EXCHANGER I
AHR SERVICE SERVICE WATER SYSTEM ---t------"1 WATER DISCHARGE I I L ____ J FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT ACTIVITY C2 ALTERNATE SHUTDOWN COOLING PATH UTILIZING RHR LOOP A FIGURE 15E.2.9-6, Rev 54 AutoCAD: Figure Fsar 15E_2_9_6.dwg
l NEUT ON FLUX l VESS L PRES IU E (PSO 2 AVE FUEL TEl1P t°F*lO) 2 8Tt1 INE f"R:E:~ ISE (PSI) 3 ~VE IJRFACE HE FLUX 3 TURB NE PRES R SE CPSI) 4 FEED ATER FLOW II D[FF SER FUJI,! l 5 VESS L STERl'I F ow S DtFF SER FLOW 2 Cl 6 BYPA S FLOW w 1se "2S0 t-C!
Q:'.
LI...
0 100 160 t-z w
CJ QC w
...,. 50
- a. se \
\
~ v_
~ -
~
-1
~
. = :. . --
0 2:0
-60 e 20
,J 30 -
0 TlHE CSEC) 30
"" TJHE CSECJ l CORE INLET SUB DOLING 1 NEUT ON FLUX 2' W R l::NSEP LEV L.( tNC:HfS> 2 SURF CE HEAT F UX 3 N R ENSED LEV LC tNC11ES) 4 CORE INLET FLO (J:)
6 ORtV FLOM t t }
160 0 120 L&J I-a:
ct L'-
0 BIi z
w u
oc w 110 Cl.
0 e 10 ~e ae " 2.S: se 75 CORE FLOW CZ) 100 TINE CSEC)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TRIP OF ONE RECIRCULATION PUMP MOTOR FIGURE 15E.3.1-1, Rev 54 AutoCAD: Figure Fsar 15E_3_1_1.dwg
1 NEUT 1 VESS L PRES RI E CPSI) 2 AYE 3 AVE
~~EE1-¥~ttP I.JRFACE HE
(°F*lO)
T FLUX 2 sr11 INE PRES ISi~ (PSO S TURB HE PRES R SE C~St) 4 FEED ATER FLOW 4 DIFF SER FLOW 1
..... 5 VESS L STEAM F OM 5 DIFF SER FLOW 2 0 150 ~Se 1------1-----+--~6:....;:::B~r~P~A~S::.....;f~L~O~W.;_....,_____-I w
I-er
~
LL 0 100 I-z w
u
~
Lil 0.. 50 e -60
Q CCR~ tt-lLET Fl.0 (J.)
6 DR[V FLOW L ( )
C 120 LS0 w
~-
er
/);'.
LL 0 ae 190 I-z w
u 0::
4J 50 Q_ lt0 0 0 75 Hl0 FLO~ CO FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TRIP OF BOTH RECIRCULATION PUMP MOTORS FIGURE 15E.3.1-2, Rev 54 AutoCAD: Figure Fsar 15E_3_1_2.dwg
l NEUTRON ALLIX I VESSE_ F8ES RISE 1PSI1 2 PEAK FUEU CENTER TE~P 2 ST!-1 LINE !"RES H!SE ,PS! l 150. JAVE SURF~CE HERT FLJX 100. : TLJ9BI~E P.~~5 Ri5E rp5JJ 4 FEEuHRTE FLOW 4 OIFFJSER ;=L W ; ,
5 VESSEL S1EAM FLOW 5 OJFFJSER '.FLOW 2 t 6 TURBINE 9-ERM FLOW IZJ
r w
cr 100.i a:
LL 0
z w
u 50.
a::
u.J
~
~
o.o.
- 10. 20. 40.
TIME
!l LEVEL(!NdH-~Ef-SEP-SK!RT l NEU!TRCJ!ll FLUX I 2 WR SENSa ~EVELCINCHESJ I 2 SUAfACE HERT F ux 15:J. R * "VEL t INCHES) 120. 1 I
( Y. l
- 7. l 100. ci cU BO.
~
I...
- cl
- 50. z w
u 40.
w C: I-
~
0...
- o. 0. **
- o. . 50. 75. iJO.
CDRE FLCJW (1/.l FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT STARTUP OF IDLE RECIRCULATION LOOP PUMP FIGURE 15E.4.4-1, Rev 54 AutoCAD: Figure Fsar 15E_4_4_1.dwg
Inadvertent Startup of HPCI 120
~
I Power 110 v~ ------ ! /1
~
Heat Flux I Steam Flow I
,, 100 Jg
~
ca
'----f Core Flow I a:
C (I) u 90 .....
~
IFeed Flow Y I-(I)
Cl. 80 70 60 0 10 20 30 40 50 Time (Seconds)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT INADVERTENT HPCI PUMP START FIGURE 15E.5.1-1, Rev 3 AutoCAD: Figure Fsar 15E_5_1_1.dwg