ML22340A166
Text
I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 29 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 14.2.5-1 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS R ERE PP OO RR TT Page: 1 of 1
LIMITING STEAMLINE BREAK STATEPOINT DOUB LE ENDED RUPTURE INSIDE CONTAINMENT WITH OFFSITE POWER AVAILABLE
Time Pressure HeatFlux Inlet Temp Flow Boron Reactivity Density (sec) Psia Fraction Frac PPM Percent GM/CC
Cold °F H °F
118.4 600.77 0.173 334.1 448.9 1.0 1.19 0.030 0.856
UN IT 2 I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 29 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 14.2.5-2 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS R ERE PP OO RR TT Page: 1 of 1
TIME SEQUENCE OF EVENTS
Accident Event Time (sec)
Rupture of a Steamline
- 1. Inside Containment With Steam line ruptures 0.0 Offsite Power available Low steamline pressure setpoint reached 0.26 Feedwater Isolation (All loops) 8.26 Steamline Isolation (Loops 2, 3 and 4) 11.26 Pressurizer empties 13.8 SI flow starts 27.26 Criticality attained 22.6 Boron from SI reaches core 38.4 Peak heat flux attained 118.4 Core becomes subcritical 121.0
- 2. Inside Containment Without Steam line ruptures 0.0 Offsite Power available Low steamline pressure setpoint reached 0.26 Feedwater Isolation (All loops) 8.26 Steam Isolation 11.26 Pressurizer empties 15.4 Criticality attained 27.4 SI flow starts 37.26 Boron from SI reaches core 52.0 Peak heat flux attained 299.7 Core becomes subcritical ~ 309
UNIT 2 UFSAR Revision 31.0
IND IANAMICH IG AN P OW ER Revision: 27.0 D. C. COOK NUCL EAR P L ANT Table: 14.2.6-1 UP D ATED FINAL SAFETY ANAL Y SIS R EP OR T Page: 1 of 1
P a r a meter s Used in th e Ana l y sis o f th e R o d Cl u ster Co ntr o l Assemb l y Ej ectio n Accident Time in Cy cl e Accident P a r a meter s H Z P H FP H Z P H FP B eg inning B eg inning End End Power Level (% ) 0 102 0 102 Ej ected Rod Worth (% k) 0.75 0.15 0.89 0.19 Delayed Neutron Fraction (% ) 0.50 0.50 0.40 0.40 Feedback Reactivity Weighting 2.071 1.30 3.621 1.30 Trip Reactivity (% k) 2. 4. 2. 4.
FQ before Rod Ej ection 2.50 2.50 2.36 2.50 FQ after Rod Ej ection 12. 7.0 25.0 7.3 Number of Operational Pumps 2. 4. 2. 4.
R esu l ts Maximum Fuel Pellet Average Temperature (°F) 3439 4268 3630 4159 Maximum Fuel Center Temperature (°F) 3922 4983 4009 4910 Maximum Fuel Stored Enthalpy (cal/ gm) 145.6 188.6 155.3 182.8 Fuel Melt in Hot Pellet, % 0 < 10 0 < 10
Unit 2 I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 16.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 14.2.8-1 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS R ERE PP OO RR TT Page: 1 of 2
TIME SEQUENCE OF EVENTS
Accident Event Time (sec)
Main Feedwater Line Rupture (With Power)
Main feedwater line rupture occurs 10.0 Low-low steam generator water level trip signal initiated 16.0 Rods begin to fall into core 18.0 SIS low pressurizer pressure setpoint reached 78.0 Feedwater isolation (Loops 2, 3, 4) 86.0 SIS flow starts 106.0 SIS low steamline pressure setpoint reached in two loops 239.8 Steamline isolation (All loops) 250.8 Auxiliary feedwater starts to deliver to intactsteam generators 610.0
Steam generator safety valve setpoint reached in intact steam generators 910.0
Core decay heat plus RCP heat decreases to auxiliary feedwater heat removal capacity ~1500.0
Pressurizer safety valve setpoint reached Never reached
Main Feedwater Line Rupture (Without Power)
Main feedwater line rupture occurs 10.0 Low-low steam generator water level trip signal initiated 16.0 Rods begin to fall into core 18.0 RCS pumps begin to coastdown 20.0 SIS low steamline pressure setpoint reached in two loops 150.6 Feedwater isolation (Loops 2,3,4) 158.6 Steamline isolation (All loops) 161.6 SIS flow starts 189.0
UNIT 2 I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 16.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 14.2.8-1 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS R ERE PP OO RR TT Page: 2 of 2
TIME SEQUENCE OF EVENTS
Accident Event Time (sec)
Auxiliary feedwater started to deliver to intact steam 610.0 generators Steam generator safety valve setpoint reached in intact steam 668.0 generators Core decay heat decreases to auxiliary feedwater heat ~1200.0 removal capacity Pressurizer safety valve setpoint reached Never reached
UNIT 2