ML22340A147

1 to Updated Final Safety Analysis Report, Chapter 14.2, Figures 14.2.5-1 to 14.2.7-6 (Unit 1)

ML22340A147
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: ML22340A147 (1)

Text

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-1 Sheet 1 of 1

Title:

Variation of Reactivity with Core Temperature at 1050 psia for the End of Life Rodded Core with One Control Rod Assembly Stuck (Zero Power) for the Steamline Break Double Ended Rupture Event UFSAR Revision 31.0

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-2 Sheet 1 of 1

Title:

Doppler Power Feedback for the Steamline Break Double Ended Rupture Event UFSAR Revision 31.0

moo 400 I

\\

1

\\

n I

0 5

10 15 20 25 30 35 40 45 50 Cold Leg Safety injection [Ibmkec]

DONAWC.COOK NUCLEARPLANT UNIT1 FIGURE 14.2.5-3 Safety Injection Flow Supplied By One Charging Pump For The Suamhnc Break Double Ended Ruprurc Event JULY 1997 UFSAR Revision 31.0

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-4 Sheet 1 of 1

Title:

Nuclear Power and Core Heat Flux vs. Time for the Steamline Break Double Ended Rupture Event

[Downstream of the Flow Restrictor with Offsite Power Available]

UFSAR Revision 31.0

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-5 Sheet 1 of 1

Title:

Core Average Temperature and RCS Pressure vs. Time for the Steamline Break Double Ended Rupture Event

[Downstream of the Flow Restrictor with Offsite Power Available]

UFSAR Revision 31.0

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-6 Sheet 1 of 1

Title:

Pressurizer Water Volume vs. Time for the Steamline Break Double Ended Rupture Event

[Downstream of the Flow Restrictor with Offsite Power Available]

UFSAR Revision 31.0

UNIT 1 Revision: 23 Change

Description:

UCR-1969 AMERICAN ELECTRIC POWER COOK NUCLEAR PLANT NUCLEAR GENERATION GROUP BRIDGMAN, MICHIGAN UFSAR Figure: 14.2.5-7 Sheet 1 of 1

Title:

Reactivity and Core Boron Concentration vs. Time for the Steamline Break Double Ended Rupture Event

[Downstream of the Flow Restrictor with Offsite Power Available]

UFSAR Revision 31.0

INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised: 27.0 Chapter:

14 Sheet:

1 of 1 UFSAR Figure: 14.2.5-8 Change

Description:

UCR-2054, Rev. 0 Unit 1

Title:

Nuclear Power and Core Heat Flux vs Time for the Rupture of a Steam Pipe from Hot Full Power Event -

(0.89 ft2 Break, Largest Break Size to Trip on OPT)

UFSAR Revision 31.0

INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised: 27.0 Chapter:

14 Sheet:

1 of 1 UFSAR Figure: 14.2.5-9 Change

Description:

UCR-2054, Rev. 0 Unit 1

Title:

Pressurizer Pressure and Mixture Volume vs Time for the Rupture of a Steam Pipe from Hot Full Power Event - (0.89 ft2 Break, Largest Break Size to Trip on OPT)

UFSAR Revision 31.0

INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised: 27.0 Chapter:

14 Sheet:

1 of 1 UFSAR Figure: 14.2.5-10 Change

Description:

UCR-2054, Rev. 0 Unit 1

Title:

Vessel Inlet Temperature and SG Pressure vs Time for the Rupture of a Steam Pipe from Hot Full Power Event - (0.89 ft2 Break, Largest Break Size to Trip on OPT)

UFSAR Revision 31.0

INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised: 27.0 Chapter:

14 Sheet:

1 of 1 UFSAR Figure: 14.2.5-11 Change

Description:

UCR-2054, Rev. 0 Unit 1

Title:

Loop Steam Flow and T Signal vs Time for the Rupture of a Steam Pipe from Hot Full Power Event -

(0.89 ft2 Break, Largest Break Size to Trip on OPT)

UFSAR Revision 31.0

DONALD C. COOK NucLEARPuNT UNIT1 FIGURE 14.26 1 Nudur Power vs.Tim For TIE Rod Ejection Evenf Hot Zero Power. EIKI Of Life JULY 1997 UFSAR Revision 31.0

- II Fuel Centdim DONALD C. COOK NUCLEARPLANT UNIT1 FIGURE 14.2.62 Fuel CeatatiDt, Fuel Avenge, and Clad Outer Surfrce Tc~

vs.Trmc For l-be Rod I

Ejection Evmt, Hot Zero Power, End Of Lik JULY 1997 UFSAR Revision 31.0

DONALD C. COOK NIT-PLANT UNlTl L

Nuclear Power vs.Tim For Tk Rod Ejection Event.

Hot FuIl Power, End Of Life JULY 1997 UFSAR Revision 31.0

I clad outer surfpa r -

DONALDCCOOK Nu-m UNIT1 FIGURE 14.264 cIentm+, F\\wl Avenge, and cladoutcrsurf~T~

m.Time For Tbc Rod Ejectioa Event, Hot Full Power, End Oftife JULY 1997 UFSAR Revision 31.0

UFSAR Revision 31.0

UFSAR Revision 31.0

UFSAR Revision 31.0

UFSAR Revision 31.0

UFSAR Revision 31.0

UFSAR Revision 31.0