ML21294A086
| ML21294A086 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2021 |
| From: | Talen Energy, Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21294A245 | List:
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| References | |
| PLA-7935 | |
| Download: ML21294A086 (41) | |
Text
SSES-FSAR Text Rev. 54 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-3625(P), Revision 0, Susquehanna Unit 1 Cycle 21 Reload Licensing Analysis, AREVA Inc., December 2017. (General Reference per NEI 98-03)
SSES-FSAR Table Rev. 65 TABLE 15C.0-1 RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Maximum Core Maximum Maximum Maximum Maximum Average Number of Neutron Flux Dome Vessel Steam line Surface Valves - Duration
% of Pressure Pressure Pressure Heat Frequency 1st of Section Figure Description1 Rated psig psig psig 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 0.12 Moderate 0 0 sec 15.1.2 15C.1.2-1 Feedwater Controller Failure 222 1247 1268 1257 118 0.27 Moderate 14 4 sec (100% Power, 108 Mlbm/hr, Max Allowable estimate Scram Time) EOC RPT Operable 15.1.3 15C.1.3-1 Pressure Regulator Failure - Open 102 1106 1129 1106 103 0.01 Moderate 2 See Text 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 15C.2.2-1 Generator Load Reject- Without Bypass 258 1263 1287 1306 117 0.27 Moderate 14 10 sec (100% Power, 108 Mlbm/hr, max allowable estimate Scram Time) EOC RPT Operable 15.2.3 Turbine Trip - Bypass Operable See Text and Moderate Appendix 15E 15.2.3 15C.2.2-1 Turbine Trip - Without Bypass 258 1263 1287 1306 117 0.27 Moderate 14 10 sec (100% Power, 108 Mlbm/hr, Max Allowable estimate Scram Time) EOC RPT Operable 15.2.4 Inadvertent MSIV Closure See Text and Moderate Appendix 15E FSAR Rev. 69 Page 1 of 4
SSES-FSAR Table Rev. 65 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Maximum Core Maximum Maximum Maximum Maximum Average Number of Neutron Flux Dome Vessel Steam line Surface Valves - Duration
% of Pressure Pressure Pressure Heat Frequency 1st of Section Figure Description1 Rated psig psig psig 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 15C.3.3-1 Seizure of One Recirculation Pump Limiting
& (Single Loop Operation) 67 1035 1070 1035 67 0.33 Fault 15C.3.3-3 15.3.4 Recirculation Pump Shaft Break See Text Limiting Fault FSAR Rev. 69 Page 2 of 4
SSES-FSAR Table Rev. 65 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Maximum Core Maximum Maximum Maximum Maximum Average Number of Neutron Flux Dome Vessel Steam line Surface Valves - Duration
% of Pressure Pressure Pressure Heat Frequency 1st of Section Figure Description1 Rated psig psig psig 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 0.22 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 0.34 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 FSAR Rev. 69 Page 3 of 4
SSES-FSAR Table Rev. 65 TABLE 15C.0-1 (Contd)
RESULTS
SUMMARY
OF TRANSIENT EVENTS UNIT 1 (TYPICAL)
Maximum Core Maximum Maximum Maximum Maximum Average Number of Neutron Flux Dome Vessel Steam line Surface Valves -
% of Pressure Pressure Pressure Heat Frequency 1st Duration of Section Figure Description1 Rated psig psig psig Flux,% CPR Category Blowdown Blowdown 15.5 INCREASE IN REACTOR INVENTORY 15.5.1 Inadvertent HPCI Pump Start See Text and 0.39 Moderate (at 60% power) Appendix 15E 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: 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.
FSAR Rev. 69 Page 4 of 4
SSES-FSAR Table Rev. 64 TABLE 15C.0-2 INPUT PARAMETERS AND INITIAL CONDITIONS FOR TRANSIENTS UNIT 1
- 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 1047.4 at Channel exit (At 100% Power and 100 Mlbs/hr)
- 8. Turbine bypass Capacity, % Rated 21.5%
- 9. Core Coolant Inlet Enthalpy, BTU/lb 523.6(2)
(At 100% Power and 100 Mlbs/hr)
- 10. Turbine Inlet Pressure, psia 976.3
- 11. Fuel Types ATRIUM-10
- 12. Core Average Gap Conductance, BTU/hr-ft2-°F 500 to 1600(1)
- 13. Core Leakage Flow,% 10%(2)
- 15. MCPR Safety Limit See Table 15C.0-3
- 16. Doppler Coefficient See Note 4 FSAR Rev. 69 Page 1 of 3
SSES-FSAR Table Rev. 64 TABLE 15C.0-2 (Contd)
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 87%
(16 Valves)
Percent of Rated Steam Flow
- 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 3 @ 1136 4 @ 1116 3 @ 1146 4 @ 1126 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 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 (-) Dryer Skirt Bottom, (L3) 13 Low Level (See Note 5) (L2) -38 Low Low Level Low Low Low Level (L1)-129
SSES-FSAR Table Rev. 64 TABLE 15C.0-2 (Contd)
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.
FSAR Rev. 69 Page 3 of 3
SSES-FSAR Table Rev. 63 TABLE 15C.0-3 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT (ALL FUEL)
UNIT 1 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. 58 TABLE 15C.0-4 UNIT 1 MINIMUM MCPR REQUIREMENT FOR SINGLE LOOP OPERATION MCPR 1.08 1.09 1.10 1.11 1.12 1.13 Safety Limit Minimum 1.41 1.42 1.43 1.44 1.45 1.46 MCPR Requirement (Based on Analysis of Pump Seizure Accident in Single Loop Operation)
MINIMUM MCPR REQUIREMENT FOR TWO LOOP OPERATION MCPR 1.07 1.08 1.09 1.10 1.11 Safety Limit Minimum 1.29 1.30 1.31 1.32 1.33 MCPR Requirement (Based on Analysis of Pump Seizure Accident in Two Loop Operation)
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 64 TABLE 15C.0-5 AVERAGE SCRAM INSERTION TIMES UNIT 1 Refer to COLR Section 5.0 FSAR Rev. 69 Page 1 of 1
SSES-FSAR Table Rev. 56 TABLE 15C.1.1-1 SEQUENCE OF EVENTS FOR LOSS OF FEEDWATER HEATING UNIT 1 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 steam flow, (Transport delay in feedwater piping is neglected).
~40 (estimate) APRM high neutron flux alarm sounds.
~60 (estimate) Reactor variables settle into new steady state, (below Scram trip point).
600 (estimate) 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.
FSAR Rev. 60 Page 1 of 1
SSES-FSAR Table Rev. 64 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.
22.660 L8 vessel level setpoint trips main turbine and feedwater pumps.
22.730 Reactor scram trip actuated from main turbine stop valve position switch.
22.760 Bypass Valves actuated 22.835 Recirculation pump trip (RPT) actuated by stop valve position switch.
25.269 Second group of safety/relief valves activate due to high pressure. (First group out of service) 25.709 Third group of safety valves activate.
Initial Conditions:
Power = 100%
Flow = 108 Mlbs/hr Bypass = Operable RPT = Operable Scram Time = Maximum Allowable Exposure = EOC FSAR Rev. 69 Page 1 of 1
SSES-FSAR Table Rev. 56 TABLE 15C.1.3-1 SEQUENCE OF EVENTS FOR PRESSURE REGULATOR FAILURE - OPEN UNIT 1 I TIME, SECONDS EVENT 0 Initial conditions, maximum limit on steam flow to turbine.
0.2 Main turbine bypass valves full open 10.82 Main steamline isolation trip occurs.
11.63 Initiation of scram trip signal, 0.06 seconds after the Main steam isolation valves reach 85% open position.
15.50 Pressure in reactor vessel reaches a minimum and starts to increase.
15.82 MSIV's are fully closed.
48 (est) Relief valves at lowest setting start to cycle to remove decay heat.
FSAR Rev. 60 Page 1 of 1
SSES-FSAR Table Rev. 64 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.
0.000 Turbine control valves closure on GLR (Generator Load Reject) 0.070 Scram initiated; TCV (Turbine Control Valve) fast closure (Trip oil pressure low) 0.175 A&B RPT: Turbine Control Valve fast closure Group 1 safety valves out of service.
1.96 Group 2 safety valves activate due to high pressure.
2.20 Group 3 safety valves activate due to high pressure.
Initial Conditions Power: 100% Flow: 108 Mlbs/hr Bypass: Inoperable Scram: Maximum Allowable RPT: Operable FSAR Rev. 69 Page 1 of 1
SSES-FSAR Table Rev. 59 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.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 57 TABLE 1SC.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 1.84 Minimum CPR FSAR Rev. 62 Page 1 of 1
SSES-FSAR Table Rev. 57 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.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev . 56 TABLE 15C.4.5-1 SEQUENCE OF EVENTS FOR RECIRCULATION FLOW CONTROLLER FAILURE UNIT 1 TIME, SECONDS EVENT 0 Master Flow Controller fails initiating a slow run-up of both reactor recirculation pumps
~220 Reactor high flux scram (analytical setpoint, 122%).
~220 Two relief valves open at 1120. 7 psia.
~230 Two relief valves reseat at 1045.7 psia.
This sequence of events is for the event initiated from:
Power = 69%
Flow = 60 Mlbs/hr Bypass = Inoperable Exposure = EOC FSAR Rev. 60 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. 57 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 Banked 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.
<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.
FSAR Rev. 63 Page 1 of 1
SSES-FSAR Table Rev. 63 TABLE 15C.4.9-2 CONTROL ROD DROP ACCIDENT UNIT 1 (TYPICAL)
Cycle Exposure, GWD/MTU BOC Control Rod Sequence B Rod Group 1 Dropped Rod Location 14-47 Dropped Rod Worth 10.24 mk (from 00 to 48)
Number of Fuel Rods with Fuel Enthalpy above 170
<2000 cal/gm Peak Deposited Enthalpy, cal/gm 191.3 FSAR Rev. 69 Page 1 of 1
Vessel Steam and Feedwater Flow Rates 140 120 i ' ',
100 I -,,,
80 ' ' ',
- Steam Flow ',
60 Percent Rated 40
Feedwater Flow n./\.
v~
20 0
-20
-40
-60 0 5 10 15 20 25 30 Time, Seconds SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15C.1.2-1-1, Rev. 65 FSAR Rev. 69
Water Level, inches 80 70 60
~~
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Water Level, inches 50 ....
40 .... /
30 ~ - , - Water Level, inches 20 10 0
0 5 10 15 20 25 30 Time, Seconds SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15C.1.2-1-2, Rev. 65 FSAR Rev. 69
Core Power and Heat Flux, % Rated 250 200
- Power, % Rated
Heat Flux, % Rated Percent Rated 150 IP I \
100 - \.
50 \.
' ~,
0 0 5 10 15 20
' 25 30 Time, Seconds SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15C.1.2-1-3, Rev. 65 FSAR Rev. 69
Vessel Dome Pressure, psia 1,300 1,250 /"-.-...
I .........
1,200 I Dome Pressure, psia -
1,150
- Dome Pressure, psia I
1,100 1,050 1,000 0 5 10 15 20 25 30 Time, Seconds SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15C.1.2-1-4, Rev. 65 FSAR Rev. 69
Core Flow % Rated 120 j
110 100 90 l w\
Core Flow, % Rated 80 70 \
60 \
50 40 0 5 10 15 20 25
" 30 Time, Seconds SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA FEEDWATER CONTROLLER FAILURE, MAXIMUM DEMAND, WITH HIGH WATER LEVEL TRIP (TYPICAL)
Figure 15C.1.2-1-5, Rev.65 FSAR Rev. 69
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FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PRESSURE REGULATOR FAILURE STEAM FLOW AT 130% OF STEAM FLOW AT 3441 MWt FIGURE 15C.1.3-1-3, Rev 55 AutoCAD: Figure Fsar 15C_1_3_1_3.dwg
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FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PRESSURE REGULATOR FAILURE STEAM FLOW AT 130% OF STEAM FLOW AT 3441 MWt FIGURE 15C.1.3-1-4, Rev 55 AutoCAD: Figure Fsar 15C_1_3_1_4.dwg
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1 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 800 0 2 4 6 8 10 12 14 16 TIME (SEC)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PRESSURE REGULATOR FAILURE STEAM FLOW AT 130% OF STEAM FLOW AT 3441 MWt FIGURE 15C.1.3-1-5, Rev 55 AutoCAD: Figure Fsar 15C_1_3_1_5.dwg
Vessel Steam and Feedwater Flows 120 100
-~ ,,, Feedwater Flow
-- ,/
I -\
80
% Rated 60 40 20
)
' \
I V f /\ ( \ I V'
I\
V
,r-...
\
/'
0 r
~ 1~ r--- Steam Flow
-20 1fv1
~
-40
-60
-80 0 1 2 3 4 5 6 Time, Seconds FSAR REV.69 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE 15C.2.2-1-1, Rev 64
Water Level, Inches 40 35 / --.........
"- /
30 ~ ......
~
Water Level, Inches 25 .......
20 ~ ....
15 10 5
0 0 1 2 3 4 5 6 Time, Seconds FSAR REV.69 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE 15C.2.2-1-2, Rev 64
Core Power & Heat Flux 300 250 200 Core Power
% Rated 150 Heat Flux 100 50 0
0 1 2 3 4 5 6 Time, Seconds FSAR REV.69 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE 15C.2.2-1-3, Rev 64
Vessel Dome Pressure 1,300 1,250
/ -
1,200 /
Vessel Dome Pressure, psia 1,150 I/
1,100 I 1,050 J
1,000 950 900 0 1 2 3 4 5 6 Time, Seconds FSAR REV.69 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE 15C.2.2-1-4, Rev 64
Core Flow 120 110 I\
100
--'\
\
90 \
v-'\
% Rated
~
80
~ r---.._
70 60 "~ -..........__
50 40 0 1 2 3 4 5 6 Time, Seconds FSAR REV.69 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 TYPICAL FINAL SAFETY ANALYSIS REPORT SUSQUEHANNA GENERATOR LOAD REJECT WITHOUT BYPASS AND TURBINE TRIP WITHOUT BYPASS FIGURE 15C.2.2-1-5, Rev 64
SSES-FSAR APPENDIX C UNIT 1 I '.
g sot----:l-\\__:::::t-----
] 60 ',
- ~;pt---
- =::::::p~::Z-~v::'.:~:: :===i:::--=-*-=---=-----=-*=* -====~--~
I Core ~ow i
~
~ 40+---\..,,...-+/-'-_.----------+-----+----+----+-----+------1 lNeutron Auxl 20+----+-----1-------1-----+----+----+----+------1 2 4 6 8 10 12 14 16 TTme (sec.)
210 I I Iopera<ing LOOp Jet t'ump I 160
- l-----
Flow {% Rated)
C:
/
Cl) I
~
~ 60
~ '--i I 10
. :se,zure LOOp Jet t'ump Flow {ll: Rated)
\ I 1<ec1rc. t'ump ,l}P"=
{,C lnltlal I
-40 4 5 a 10 12 14 16
' TTme (sec.)
rt)
Cl) 150 r~eam ~,owl
.s::
0 Rated)
C:
v 100 A
- --- *---~
reedwater Aowl
{% Rated)
~ --- -- ---- ----.........--- -****
'- 50 0
"C Inmer LeVel, ,ncnes Above Battom I '\ /
'iii of Separator Skirt !Dome Pressure!
3 0 Q.
<I 0 -50 ~ r ~ Change {psi) I
~
...__; ~ r---.. /
I
...,C: -100
~
~ -150 0 2 4 6 8 10 12 14 16 TTme (sec.)
FSAR REV.65
- 100% Rated Core Power is SUSQUEHANNA STEAM ELECTRIC STATION Assumed to be 3952 MWth UNIT 1 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT TWO LOOP OPERATION 4031 MWth*/108Mlbm/HR CORE RESPONSE FIGURE 15C.3.3-1, Rev 59 AutoCAD: Figure Fsar 15C_3_3_1.dwg
SSES-FSAR APPENDIX C UNIT 1
- 1. 7 *-*-*-*-*-*-*-*-*-*--------------*-*-*-*-**-*-*-*-*--------*------ *-*-*-*-*-*-*-*-*-*-------*------- *-*-*-*-**-*-*-*-*------------~----*--------* --**--***-*-*-*---~--
)i 1.6 +------+------+------+------+------+-----+------+----;/
1.5 +-------+-------+------+-------+-------+------+-----t----<
8
~
I 1.4 /
I D..
ii J
- 8 1.3 + - - - - - - - + - - - - - - - + - - - - - + - - - - - - - + - - - - - - - + - - - - - , . . _ - - - - - l
~ ,l,....-o--o-= ~
= --o---o--c=~ - ~~
1.2 + - - - - - - - - + - - - - - - ~ . . - - - - - - - - - - - - - - - - ....,.._-..n.....---+--..r-,,'-/--n------+-------,
r--n__ __ _.~ ~
1.1 -l----+----+-----+---==-=~,:i:::!::::::Q:::::i-___--+------+-----1 I Minimum CPR of 1.11 att= 1.82 sec I 1.0 -+-------+-------+------+-------+-------+-----+-------I 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 Time (sec.)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT TWO LOOP OPERATION 4031 MWth/108Mlbm/HR TYPICAL CPR RESPONSE FIGURE 15C.3.3-2, Rev 59 AutoCAD: Figure Fsar 15C_3_3_2.dwg
SSES-FSAR APPENDIX C UNIT 1 60
~ s: _ \;': '
~
" ~ ..
Heat Flux o-----+-----+------+------1
~ 4: \ \ v - -----+------+--------+-,---------+------+- ----+-
~ 3u+:--~:,r1-----t----+---=~,===.--t----+-----+-------1
":: -~ , ICore Flowl
- I "I Neutron Flux I 10+-----+-----+------+-------1----+-----+-----+-------1 0+-------------------1----------------~
0 2 4 6 8 10 12 14 16 Time (sec.)
14 12 -- --,_\
100
'.. IActive JetRated)
(%
Pump Flow I
- -c -......_____
~
V 1=Cl) 6 -
~
e
~ 40 c--
20 K ------ !Inactive Jet Pump Flow I
(% Rated 0
// ------ I Recirc. Pump Speed I
-20 (% Initial)
-40 0 2 4 6 8 10 12 14 16 Time (sec.)
!Steam Flowl 7 - I (% Rated) I I
1.c
~
6- ~ ~-*- **-- ... _____ I I ****.-************
C u
C 5,O
_,.,..,,- "" ~~ --.
IFeedwater Flow1
% Rated) I 4 -- Water Level, inches Above Bottom 1 of Separator Skirt I 3*~-
2:o
~...
~
0 g
ll 0
-1 0
---~ I Dome Pressure Change (ps0I
~
~ -20
-30 0 *)
4 8 10 12 14 16 Time (sec.)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT
- 100% Rated Core Power is PUMP SEIZURE ACCIDENT Assumed to be 3952 MWth SINGLE LOOP OPERATION 2652 MWth*/52Mlbm/HR CORE RESPONSE FIGURE 15C.3.3-3, Rev 1 AutoCAD: Figure Fsar 15C_3_3_3.dwg
SSES-FSAR APPENDIX C UNIT 1 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 Time (set.)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 FINAL SAFETY ANALYSIS REPORT PUMP SEIZURE ACCIDENT SINGLE LOOP OPERATION 2652 MWth/52 Mlbm/HR TYPICAL CPR RESPONSE FIGURE 15C.3.3-4, Rev 1 AutoCAD: Figure Fsar 15C_3_3_4.dwg