ML24159A195

Updated Final Safety Analysis Report (Ufsar), Revision 32, Chapter 14, Table 14.0-1 (Unit 2)

ML24159A195
Person / Time
Site:	Cook
Issue date:	05/30/2024
From:	Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
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ML24159A261	List: AEP-NRC-2024-20, Updated Final Safety Analysis Report (Ufsar), Revision 32, Chapter 3, Reactor (Unit 2) ML24159A163 ML24159A164 ML24159A165 ML24159A166 ML24159A167 ML24159A168 ML24159A169 ML24159A170 ML24159A171 ML24159A172 ML24159A173 ML24159A174 ML24159A175 ML24159A176 ML24159A177 ML24159A178 ML24159A179 ML24159A180 ML24159A181 ML24159A182 ML24159A183 ML24159A184 ML24159A185 ML24159A186 ... further results
References
AEP-NRC-2024-20
Download: ML24159A195 (1)

Text

I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 18.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 14.0-1 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

OCCURRENCES EVALUATED FOR VANTAGE 5 FUEL

Fault Conditions Evaluated Accident FSAR Section 14.1.7 Startup of an Inactive Reactor Coolant Loop Fault Conditions Analyzed Accident FSAR Section 14.1.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical Condition 14.1.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power 14.1.3 Rod Cluster Control Assembly Misalignment

14.1.4 Rod Cluster Control Assembly Drop

14.1.5 Uncontrolled Boron Dilution 14.1.6 Loss of Reactor Coolant Flow (including Locked Rotor Analysis)1 14.1.8 Loss of External Electric Load or Turbine Trip(1), 2,

14.1.9 Loss of Normal Feedwater3,4 14.1.10 Excessive Heat Removal due to Feedwater System Malfunctions 14.1.11 Escessive Load Increase Incident 14.1.12 Loss of Offsite Power to the Station Auxiliaries (3), (4)

14.2.4 Steam Generator Tube Rupture

14.2.5 Rupture of a Steam Line 14.2.6 Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) 14.2.8 Major Rupture of a Main Feedwater Pipe (3), (4)

14.3.1 Major Reactor Coolant Pipe Ruptures Loss of Reactor Coolant From Small Ruptured Pipes 14.3.2 From Cracks in Large Pipes which Actuates the Emergency Core Coolant System

1 Reanalyzed to support PSV setpoint tolerance of +3%.

2 Reanalyzed to support MSSV setpoint tolerance of +3%.

3 Evaluated to support MSSV setpoint tolerance of + 3%.

4 Evaluated to support PSV setpoint tolerance of + 3%.

UNIT 2