ML21125A517
Text
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT ANALYSIS REPORT Revision: 24.0 Table: 14.1.0-1 Page:
1 of 2 Unit 2 RANGE OF PLANT NOMINAL CONDITIONS USED IN SAFETY ANALYSES1 Parameter Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 NSSS Power, Mwt 3600 3600 3600 3600 3600 3600 Core Power, Mwt 35882 35882 35882 35882 35882 35882 RCS Flow,(gpm/loop) 88,500 88,500 88,500 88,500 88,500 88,500 Minimum Measured Flow, (total gpm) 366,400 366,400 366,400 366,400 366,400 366,400 RCS Temperatures, °F Core Outlet 613.5 585.8 618.4 618.2 585.8 585.7 Vessel Outlet 610.2 582.3 615.2 615.0 582.3 582.2 Core Average 579.5 550.1 584.8 584.9 550.1 550.1 Vessel Average 576.0 547.0 581.3 581.3 547.0 547.0 Vessel/Core Inlet 541.8 511.7 547.3 547.6 511.7 511.8 Steam Generator Outlet 541.6 511.4 547.1 547.4 511.4 511.5 Zero Load 547.0 547.0 547.0 547.0 547.0 547.0 RCS Pressure, psia 2250 2250 2250 2100 2250 2100 1 A brief description of each case follows Table 14.1.0-1.
2 The Best Estimate Large Break (LB) LOCA analyses with RHR cross-ties open support plant operation with a core power at 3468 MWt (plus 0.34% uncertainty).
The SBLOCA analysis with the High Head SI cross-tie valves open supports plant operation up to a core power of 3600 MWt (plus 0.34% uncertain)..
Evaluations have been performed to support the Measurement Uncertainty Recapture (MUR) power uprate, where the sum of the power uprate and the revised, reduced calorimetric power uncertainty remains equal to, or less than, the 2% uncertainty assumed in the safety analyses.
UFSAR Revision 30.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT ANALYSIS REPORT Revision: 24.0 Table: 14.1.0-1 Page:
2 of 2 Unit 2 Parameter Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Steam Pressure, psia 780.4 587.0 820.0 820.0 587.0 587.0 Steam Flow, (106 lb/hr total) 15.98 15.90 16.0 16.0 15.9 15.9 Feedwater Temp.,°F 449.0 449.0 449.0 449.0 449.0 449.0
% SG Tube Plugging 10 10 10 10 10 10 A BRIEF DESCRIPTION OF VARIOUS CASES LISTED Case 1 and 2:
These parameters cases were used to support operation during mixed core cycles (Cycles 8 and 9).
Case 3:
These parameters incorporate a core power level of 3588 MWt, an NSSS power level of 3600 MWt (which includes 12 MWt for reactor coolant pump heat), an average steam generator tube plugging level of 10%, RCS pressure of 2250 psia, and an upper bound vessel average temperature of 581.3°F. This parameter case was used to support high RCS temperature and high RCS pressure operation for a full VANTAGE 5 core (Cycle 10 and beyond).
Case 4:
These parameters incorporate the same features as case 3, except the RCS pressure is 2100 psia. This parameter case was used to support high RCS temperature and low RCS pressure operation for a full VANTAGE 5 core (Cycles 10 and beyond).
Case 5:
These parameters incorporate the same features as case 3, except the lower bound vessel average temperature is 547°F. This parameter case was used to support low RCS temperature and high RCS pressure operation for a full VANTAGE 5 core (Cycles 10 and beyond).
Case 6:
These parameters incorporate the same features as case 5, except the RCS pressure is 2100 psia. This parameter case was used to support low RCS temperature and low RCS pressure operation for a full VANTAGE 5 core (Cycles 10 and beyond).
UFSAR Revision 30.0
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1 of 4 Unit 2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
Reactor Vessel Coolant Flow (GPM)
Vessel Average Temperature
(°F)
Pressurizer Pressure (PSIA)
Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition TWINKLE FACTRAN THINC See Section 14.1.1.2 N/A2 3
W-3 ANF WRB-2 and W-3 V-5 No 0
162,840 547.0 2037.0 4 RCCA Misalignment LOFTRAN THINC N/A N/A N/A W-3 ANF WRB-2 V-5 Yes 3600 366,400 581.3 2100.0 5 1 Includes reactor coolant pump heat, if applicable.
2 N/A - Not Applicable 3 Zero Power Doppler Power Defect at BOL assumed to be - 1000 pcm.
4 Core Pressure 5 For transition cycles, pressurizer pressure is 2250 psia.
UFSAR Revision 30.0
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2 of 4 Unit 2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
Reactor Vessel Coolant Flow (GPM)
Vessel Average Temperature
(°F)
Pressurizer Pressure (PSIA)
Uncontrolled Boron Dilution N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 3600 0
N/A N/A N/A N/A N/A N/A Loss of Forced Reactor Coolant Flow LOFTRAN FACTRAN THINC
+5 N/A Max 6 W-3 ANF WRB-2 V-5 Yes 3608 366,400 581.3 7 2100.0 (5)
Locked Rotor (Peak Pressure)
LOFTRAN
+5 N/A Max (6)
N/A N/A 3680 354,000 585.4 2312.6 Locked Rotor (Peak Clad Temp)
LOFTRAN FACTRAN
+5 N/A Max (6)
N/A N/A 3680 354,000 585.4 2037.4 6 Maximum Doppler power coefficient (pcm/%power) = -19.4 + 0.002Q, where Q is in MWt (see Figure 14.1.0-1) 7 For Transition Cycles, Vessel Average Temperature is 576°F.
UFSAR Revision 30.0
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3 of 4 Unit 2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
Reactor Vessel Coolant Flow (GPM)
Vessel Average Temperature
(°F)
Pressurizer Pressure (PSIA)
Locked Rotor (Rods-in-DNB)
LOFTRAN FACTRAN THINC
+5 N/A Max (6)
WRB-2 Yes 3608 366,400 581.3 2100.0 Loss of Normal Feedwater LOFTRAN 0
N/A Max (6)
N/A N/A 3680 354,000 585.4 2312.6 Loss of Offsite Power (LOOP) to the Station Auxiliaries LOFTRAN 0
N/A Max (6)
N/A N/A 3680 354,000 541.4 2312.6 Rupture of a Steam Pipe LOFTRAN THINC See Figure 14.2.5-1 N/A See Figure 14.2.5-2 W-3 ANF W-3 V-5 NO 0
354,000 547.0 2100.0 UFSAR Revision 30.0
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4 of 4 Unit 2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
Reactor Vessel Coolant Flow (GPM)
Vessel Average Temperature
(°F)
Pressurizer Pressure (PSIA)
Rupture of a Control Rod Drive Mechanism Housing TWINKLE FACTRAN See Section 14.2.6 N/A 8, 9 N/A N/A 3660 10 0
354, 000 162, 840 585.4 547.0 2037.4 (4)
Rupture of Feedwater Pipe LOFTRAN N/A
.54 Max (6)
N/A N/A 3680 354, 000 585.4 2162.6 8 Full Power Doppler Power defect at BOL and EOL assumed to be -966 pcm and -893 pcm respectively.
9 Zero Power Doppler only Power defect at BOL and EOL assumed to be -965 pcm and -849 pcm, respective.
10 Core thermal power.
UFSAR Revision 30.0
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1 of 1 UNIT 2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED: SEPARATE FULL VANTAGE 5 CORE ANALYSES Reactivity Coefficients Assumed Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Moderator Density (K/gm/cc)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output) (MWt)1 Reactor Vessel Coolant Flow (GPM)
Vessel Average Temperature
(°F)
Pressurizer Pressure (PSIA)
Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power 2 LOFTRAN N/A3
+5
.54 N/A Max4 Min5 WRB-2 Yes 3608 2165 361 366,400 581.3 567.6 550.4 2100.0 Loss of Electrical Load or Turbine Trip 6 LOFTRAN N/A
+5
.54 N/A Max(4)
Min(5)
WRB-2 Yes 3600 366,400 581.3 2100.0 Excessive Heat Removal Due to Feedwater System Malfunction LOFTRAN N/A N/A
.54
.54 Min(5)
Min(5)
WRB-2 WRB-2 Yes Yes 3600 0
366,400 366,400 581.3 547.0 2100.0 2100.0 Excess Load Increase LOFTRAN N/A N/A 0
.54 Min(5)
Max(4)
WRB-2 Yes 3600 366,400 581.3 2100.0 1 Includes reactor coolant pump heat, if applicable.
2 Multiple power levels, Tavg, and reactivity feedback cases were examined.
3 N/A - Not Applicable 4 Maximum Doppler Power coefficient (pcm/%power) = -19.4 + 0.002Q, where Q is in MWt (see Figure 14.1.0-1).
5 Minimum Doppler power coefficient (pcm/%power) = -9.55 + 0.00104Q, where Q is in MWt (see Figure 14.1.0-1).
6 Minimum and maximum reactivity feedback cases were examined.
UFSAR Revision 30.0
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1 of 1 UNIT 2 RPS TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN NON-LOCA SAFETY ANALYSES Trip Function Nominal Setpoint Point Assumed In Analysis Limiting Trip Time Delay (seconds)
Power range high neutron flux, high setting 109%
118%
0.5 Power range high neutron flux, low setting 25%
35%
0.5 Overtemperature T See Table 2.2-1 Variable, see Figures 14.1.0-5,6 8.0 1 Overpower T in Tech Spec Variable, see Figures 14.1.0-5,6 8.0 (1)
High pressurizer pressure 2385 psig 2428 psig 2.0 Low pressurizer pressure 1950 psig 1907 psig 2.0 High pressurizer water level 92% of span 100% span 2.0 Low reactor coolant flow (From loop flow detectors) 90% loop flow 87% loop flow 1.0 Undervoltage trip volts each bus 2905 volts each bus NA 2 1.5 Underfrequency trip 57.5 Hz 57 Hz 0.6 Low-low steam generator level 21% of narrow range span 0.0% of narrow range span 2.0 1 Total time delay (Including RTD time response, trip circuit, and channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall. The time delay assumed in the analysis supports the response time of the RTD time response, trip circuit delays, and the channel electronics delay presented in the UFSAR Table 14.1.0-4. An evaluation has been performed (Reference 9) that demonstrates that the analyses remains bounding given that the total 8.0 second time delay in the above table is satisfied.
2 No explicit value assumed in the analysis. Undervoltage reactor trip setpoint assumed reached at initiation of analysis.
UFSAR Revision 30.0
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19 Table: 14.1.0-5 Page:
1 of 1 Unit 2 ESF ACTUATION SETPOINTS AND TIME DELAYS TO ACTUATION ASSUMED IN NON-LOCA SAFETY ANALYSES ESF Actuation Function Nominal Setpoint Limiting Actuation Setpoint Assumed In Analyses Time Delay (Seconds)
Safety Injection (SI)
- Low pressurizer pressure 1815 psig 1700 psig 27 w/offsite power 1 37 w/o offsite power 2
- Low steamline pressure 600 psig 344 psig 27 w/offsite power (1) 37 w/o offsite power (2)
- Low-low steam generator water level 21% of narrow range span 0.0% of narrow range span 603 High-high steam generator Level Turbine Trip 67% of narrow range span 82% of narrow range span 2.5 Steamline Isolation on low steam line pressure NA4 NA (4) 115 Feedwater Line Isolation on high-high steam generator water level 67% of narrow range span 82% of narrow range span 11 6 Feedwater Line Isolation on low steam line pressure NA (4)
NA (4) 8 (6) 1 Emergency diesel generator starting and sequence loading delays NOT included. Offsite power available.
Response time limit includes opening of valves to establish safety injection (SI) path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valves open, then VCT valves close) is included.
2 Emergency diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valve close) is included.
3 For Loss of Normal Feedwater and Loss of Offsite Power to Station Auxiliaries occurrences, the delay time assumed is 60 seconds from the initiation of the signals. For Feedwater Line Break event, the delay time assumed is 600 seconds (10 minute operator action delay) from the initiation of the break.
4 Not Applicable 5 Steamline isolation total delay time includes valve closure time, and electronics and sensor delay. Technical Specifications require 8.0 second valve closure time.
6 Feedwater Line isolation total delay time includes valve closure time and electronics and sensor delay time.
UFSAR Revision 30.0
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1 of 4 Unit 2 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Fault Conditions Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 14.1.1 Uncontrolled RCCA bank withdrawal from a subcritical condition Power range high flux (low setpoint)
NA NA NA 14.1.2 Uncontrolled RCCA bank withdrawal at power Power range high flux, overtemperature delta-T, high pressurizer pressure, high pressurizer level NA Pressurizer safety valves, steam generator safety valves NA 14.1.3 RCCA misalignment 14.1.4 (including rod drop) 14.1.5 Uncontrolled Boron Dilution Source range high flux power range high flux overtemperature delta-T NA Low insertion limit annunciators for boration NA 14.1.6.1 Partial and complete loss of forced reactor coolant flow Low flow, undervoltage underfrequency NA Steam generator safety valves NA 14.1.6.2 Reactor coolant pump shaft seizure (locked rotor)
Low flow NA Pressurizer safety valves, steam generator safety valves NA 14.1.7 Startup of an inactive reactor coolant loop 1
1 This cannot occur in Modes 1 and 2 as restricted by the Cook Nuclear Plant Unit 2 Technical Specifications.
UFSAR Revision 30.0
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2 of 4 Unit 2 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Fault Conditions Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 14.1.8 Loss of external electric load or turbine trip High pressurizer pressure overtemperature delta-T, lo-lo steam generator level Steam generator lo-lo level Pressurizer safety valves, steam generator safety valves Auxiliary Feedwater System 14.1.9 Loss of normal feedwater Steam generator lo-lo level, manual Steam generator lo-lo level Steam generator safety valves, pressurizer safety valves Auxiliary Feedwater System 14.1.10 Feedwater system malfunctions that result in an increase in feed water flow Power range high flux, (low and high setpoints),
steam generator lo-lo level (Intact steam generators)
High-high steam generator level-produced feedwater isolation and turbine trip Feedwater isolation NA 14.1.11 Excessive load increase Power range high flux, overtemperature delta-T, overpower delta-T NA Pressurizer safety valves, steam generator safety valves NA 14.1.12 Loss of offsite power to the station Auxiliaries Steam generator lo-lo level Steam generator lo-lo level Steam generator valves, pressurizer safety valves Auxiliary Feedwater System 14.2.4 Steam generator tube failure Reactor Trip System Engineered Safety Features Actuation System Steam generator safety and/or relief valves, steamline stop valves Emergency Core Cooling System, Auxiliary Feedwater System, Emergency Power System UFSAR Revision 30.0
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3 of 4 Unit 2 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Fault Conditions Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 14.2.5 Rupture of a Steam Line SIS, low pressurizer pressure, manual Low pressurizer pressure low compensated steamline pressure, high containment pressure, manual Feedwater isolation steamline stop valves Auxiliary Feedwater System, Safety Injection System Inadvertent opening of a steam generator relief or safety valve SIS Low pressurizer pressure, low compensated steamline pressure Feedwater isolation steamline stop valves Auxiliary Feedwater System, Safety Injection System 14.2.6 Spectrum of RCCA ejection accidents Power range high flux, high positive flux rate NA NA NA 14.2.8 Feedwater system pipe break Steam generator lo-lo level, high pressurizer pressure, SIS High containment pressure, steam generator lo-lo water level, low compensated steamline pressure Steamline isolation valves, feedline isolation, pressurizer self-actuated safety valves, steam generator safety valves Auxiliary Feedwater System, Safety Injection System UFSAR Revision 30.0
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4 of 4 Unit 2 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Fault Conditions Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment 14.3 Loss of coolant accidents resulting from the spectrum of postulated piping breaks within the reactor coolant pressure boundary Reactor Trip System Engineered Safety Features Actuation System Service Water System Component Cooling Water System steam generator safety and/or relief valves Emergency Core Cooling System, Auxiliary Feedwater System, Containment Heat Removal System, Emergency Power System UFSAR Revision 30.0
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18 Table: 14.1.1-1 Page:
1 of 1 Unit 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Uncontrolled RCCA Withdrawal From A Subcritical Condition Initiation of uncontrolled RCCA withdrawal (63 pcm/sec) 0.0 High Neutron Flux Reactor Trip Setpoint (low setting) reached 12.2 Rods begin to fall into core 12.7 Minimum DNBR occurs 14.8 Peak Clad Average Temperature occurs 15.3 Peak Fuel Average Temperature occurs 15.6 Peak Fuel Centerline Temperature Occurs 16.0 UFSAR Revision 30.0
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1 of 1 Unit 2 TIME SEQUENCE OF EVENTS (FULL VANTAGE 5 CORE)
Accident Event Time (sec)
Uncontrolled RCCA Bank Withdrawal At Full Power Case A: (high insertion rate max feedback)
Initiation of uncontrolled RCCA bank withdrawal at a high reactivity insertion rate (80 pcm/sec) 0 Power range high neutron flux high trip signal initiated 5.8 Rods begin to fall into core 6.3 Minimum DNBR occurs 6.4 Case B: (small insertion rate, max feedback)
Initiation of uncontrolled RCCA bank withdrawal at a small reactivity insertion rate (4 pcm/sec) 0 Overtemperature T reactor trip signal initiated 314.5 Minimum DNBR occurs 316.2 Rods begin to fall into core 316.5 UFSAR Revision 30.0
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1 of 1 Unit 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Uncontrolled Boron Dilution
- 1. Dilution during Refueling Dilution begins 0
Shutdown margin lost 1848
- 2. Dilution during startup Dilution begins 0
Shutdown margin lost 2100
- 3. Dilution during full power operation
- a. Automatic reactor control Dilution begins 0
Shutdown margin lost 2760
- b. Manual reactor control Dilution begins 0
Overtemperature T reactor trip 90 Shutdown margin lost 2760 UFSAR Revision 30.0
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1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Loss of Forced Reactor Coolant Flow Four loops in operation, four pumps coasting down All operating pumps lose power and begin coasting down 0.0 Reactor coolant pump under-voltage trip point reached 0.0 Rods begin to drop 1.5 Minimum DNBR occurs 3.7 Four loops in operation, one pump coasting down Coastdown begins 0.0 Low flow reactor trip 1.28 Rods begin to drop 2.28 Minimum DNBR occurs 3.40 UFSAR Revision 30.0
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1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Single Reactor Coolant Pump Locked Rotor Four loops in operation, one locked rotor Rotor in one pump locks 0.00 Low reactor coolant flow trip setpoint reached 0.02 Rods begin to drop 1.02 Time at which minimum DNBR is predicted to occur 2.2 Maximum RCS pressure occurs 3.10 Maximum clad temperature occurs 3.60 UFSAR Revision 30.0
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1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL VANTAGE 5 CORE)
Accident Event Time (sec)
Loss of External Electric Load or Turbine Trip
- 1. With pressurizer control (min fdbk)
Loss of electrical load 0.0 Overtemperature T reactor trip point reached 12.2 Peak pressurizer pressure occurs 14.2 Rods begin to drop 12.5 Minimum DNBR occurs 16.0
- 2. With pressurizer control (max fdbk)
Loss of electrical load 0.0 Peak pressurizer pressure occurs 8.5 Low-low steam generator water level reactor trip point reached 53.7 Rods begin to drop 55.7 Minimum DNBR occurs 1
- 3. Without pressurizer control (min fdbk)
Loss of electrical load 0.0 High pressurizer pressure reactor trip point reached 7.3 Peak pressurizer pressure occurs 9.3 Rods begin to drop 9.0 Minimum DNBR occurs (1)
- 4. Without pressurizer control (max fdbk)
Loss of electrical load 0.0 High pressurizer pressure reactor trip point reached 7.4 Rods begin to drop 9.4 Peak pressurizer pressure occurs 9.5 Minimum DNBR occurs (1) 1 DNBR never decreases below its initial value.
UFSAR Revision 30.0
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1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Loss of Normal Feedwater Main feedwater flow stops 10.0 Low-low steam generator water level trip signal initiated 55.7 Rods begin to fall into core 57.7 Two Motor-Driven Auxiliary Feedwater Pumps Start and Supply the Steam Generators 115.7 Cold Auxiliary Feedwater is Delivered to the Steam Generators 515.0 Peak water level in pressurizer occurs 4672 Core decay heat plus RCP heat decreases to auxiliary feedwater heat removal capacity 4800 UFSAR Revision 30.0
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17 Table:14.1.10B-1 Page:
1 of 1 UNIT 2 TTIIM MEE SSEEQQUUEENNCCEE OOFF EEVVEENNTTSS
((FFUULLLL VV--55 CCOORREE))
Accident Event Time (sec)
Feedwater System Malfunctions:
Excessive feedwater flow at full power to a single steam generator (Manual Rod Control)
One main feedwater control valve fails fully open 0.0 Hi-hi steam generator water level signal generated 30.2 Turbine trip occurs due to hi-hi steam generator water level 32.7 Minimum DNBR occurs 34.0 Reactor trip occurs due to turbine trip 34.7 Feedwater isolation achieved 41.2 UFSAR Revision 30.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
17 Table:14.1.10B-2 Page:
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
Accident Event Time (sec)
Feedwater System Malfunctions:
Excessive feedwater flow at full power to a single steam generator (Automatic Rod Control)
One main feedwater control valve fails fully open 0.0 Hi-hi steam generator water level signal generated 30.1 Turbine trip occurs due to hi-hi steam generator water level 32.6 Minimum DNBR occurs 33.0 Reactor trip occurs due to turbine trip 34.6 Feedwater isolation achieved 41.1 UFSAR Revision 30.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
17 Table:14.1.10B-3 Page:
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
Accident Event Time (sec)
Feedwater System Malfunctions:
Excessive feedwater flow at full power to all four steam generators (Manual Rod Control)
All four main feedwater control valves fail fully open 0.0 Hi-hi steam generator water level signal generated 31.5 Turbine trip occurs due to hi-hi steam generator water level 34.0 Minimum DNBR occurs 34.5 Reactor trip occurs due to turbine trip 36.0 Feedwater isolation achieved 42.5 UFSAR Revision 30.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
17 Table:14.1.10B-4 Page:
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
Accident Event Time (sec)
Feedwater System Malfunctions:
Excessive feedwater flow at full power to all four steam generators (Automatic Rod Control)
All four main feedwater control valves fail fully open 0.0 Hi-hi steam generator water level signal generated 31.8 Turbine trip occurs due to hi-hi steam generator water level 34.3 Minimum DNBR occurs 35.5 Reactor trip occurs due to turbine trip 36.3 Feedwater isolation achieved 42.8 UFSAR Revision 30.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:14.1.11B-1 Page:
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
Accident Event Time (sec)
Excessive Load Increase
- 1. Manual reactor control (Min fdbk) 10% step load increase 0.0 Equilibrium conditions reached 160.0
- 2. Manual reactor control (Max fdbk) 10% step load increase 0.0 Equilibrium conditions reached 40.0
- 3. Automatic reactor control (Min fdbk) 10% step load increase 0.0 Equilibrium conditions reached 160.0
- 4. Automatic reactor control (Max fdbk) 10% step load increase 0.0 Equilibrium conditions reached 70.0 UFSAR Revision 30.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.1.12-1 Page:
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Loss of Offsite Power to the Station Auxiliaries AC power is lost 10.0 Main feedwater flow stops 10.0 Low-low steam generator water level trip signal initiated 56.0 Rods begin to fall into core 58.0 Reactor coolant pumps begin to coastdown 58.0 Two Motor-Driven Auxiliary Feedwater Pumps Start and Supply the Steam Generators 117.0 Cold Auxiliary Feedwater is Delivered to the Steam Generators 534.0 Core decay heat decreases to auxiliary feedwater heat removal capacity
~800.0 Peak water level in pressurizer occurs 1406.0 UFSAR Revision 30.0