ML22340A166

1 to Updated Final Safety Analysis Report, Chapter 14.2, Tables 14.2.5-1 to 14.2.8-1 (Unit 2)

ML22340A166
Person / Time
Site:	Cook
Issue date:	11/30/2022
From:	Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22340A137	List: AEP-NRC-2022-62, 1 to Updated Final Safety Analysis Report, Chapter 14.3, Reactor Coolant System Pipe Rupture (Loss of Coolant Accident) (Unit 1) ML22340A128 ML22340A129 ML22340A131 ML22340A132 ML22340A133 ML22340A134 ML22340A135 ML22340A136 ML22340A138 ML22340A139 ML22340A141 ML22340A142 ML22340A143 ML22340A144 ML22340A145 ML22340A146 ML22340A147 ML22340A148 ML22340A149 ML22340A150 ML22340A151 ML22340A152 ML22340A153 ML22340A154 ML22340A155 ML22340A156 ML22340A157 ML22340A158 ML22340A159 ML22340A160 ML22340A161 ML22340A162 ML22340A163 ML22340A164 ML22340A165 ML22340A166 ML22340A167 ML22340A168 ML22340A169 ML22340A170 ML22340A171 ML22340A172 ML22340A173 ML22340A174 ML22340A175 ML22340A176 ML22340A177 ML22340A178 ML22340A179 ... further results
References
AEP-NRC-2022-62
Download: ML22340A166 (1)

Text

IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:

29 Table: 14.2.5-1 Page:

1 of 1 UNIT 2 LIMITING STEAMLINE BREAK STATEPOINT DOUBLE ENDED RUPTURE INSIDE CONTAINMENT WITH OFFSITE POWER AVAILABLE Time (sec)

Pressure Psia HeatFlux Fraction Inlet Temp Flow Frac Boron PPM Reactivity Percent Density GM/CC Cold °F Hot °F 118.4 600.77 0.173 334.1 448.9 1.0 1.19 0.030 0.856 UFSAR Revision 31.0

IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:

29 Table: 14.2.5-2 Page:

1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)

Rupture of a Steamline

1. Inside Containment With Offsite Power available Steam line ruptures 0.0 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 Offsite Power available Steam line ruptures 0.0 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 UFSAR Revision 31.0

INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 27.0 Table: 14.2.6-1 Page:

1 of 1 Unit 2 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident Accident Parameters Time in Cycle HZP Beginning HFP Beginning HZP End HFP End Power Level (%)

0 102 0

102 Ejected 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 Ejection 2.50 2.50 2.36 2.50 FQ after Rod Ejection

12.

7.0 25.0 7.3 Number of Operational Pumps

2.

4.

2.

4.

Results 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 UFSAR Revision 31.0

IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.2.8-1 Page:

1 of 2 UNIT 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 intact steam 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 UFSAR Revision 31.0

IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.2.8-1 Page:

2 of 2 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)

Auxiliary feedwater started to deliver to intact steam generators 610.0 Steam generator safety valve setpoint reached in intact steam generators 668.0 Core decay heat decreases to auxiliary feedwater heat removal capacity

~1200.0 Pressurizer safety valve setpoint reached Never reached UFSAR Revision 31.0