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{{#Wiki_filter:UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                       Revision:    24.0 D. C. COOK NUCLEAR PLANT                                                     Table: 14.1.0-1 UPDATED FINAL SAFETY ANALYSIS REPORT                                                      Page:       1 of 2 RANGE OF PLANT NOMINAL CONDITIONS USED IN SAFETY ANALYSES1 P a r a meter                          Ca se 1           Ca se 2         Ca se 3         Ca se 4       Ca se 5       Ca se 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 Temp er a tu r es, ° 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
{{#Wiki_filter:INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT ANALYSIS REPORT Revision: 24.0 Table: 14.1.0-1 Page:
A brief description of each case follows Table 14.1.0-1.
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).
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)..
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 eq ual to, or less than, the 2% uncertainty assumed in the safety analyses.
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.
Unit 2
UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                           Revision:    24.0 D. C. COOK NUCLEAR PLANT                                                           Table: 14.1.0-1 UPDATED FINAL SAFETY ANALYSIS REPORT                                                          Page:       2 of 2 P a r a meter                        Ca se 1           Ca se 2         Ca se 3         Ca se 4         Ca se 5           Ca se 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
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT ANALYSIS REPORT Revision: 24.0 Table: 14.1.0-1 Page:
  % 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).
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  
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).
% SG Tube Plugging 10 10 10 10 10 10 A BRIEF DESCRIPTION OF VARIOUS CASES LISTED Case 1 and 2:
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).
These parameters cases were used to support operation during mixed core cycles (Cycles 8 and 9).
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 3:
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).
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).
Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                    Revision:   19.1 D. C. COOK NUCLEAR PLANT Table: 14.1.0-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     1 of 4
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 19.1 Table: 14.1.0-2 Page:
1 of 4 Unit 2


==SUMMARY==
==SUMMARY==
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Initial Reactor Revised  NSSS              Vessel Computer       Moderator     Moderator                                          Vessel              Pressurizer Fault                                                              DNB        Thermal  Thermal            Average Codes        Temperature     Density  Doppler                                Coolant              Pressure Conditions                                                          Correlation  Design    Power          Temperature Utilized      (pcm/°F)     (K/gm/cc)                                         Flow                 (PSIA)
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Procedure Output 1            (°F)
Moderator Density (K/gm/cc)
(GPM)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
(MWt)
Reactor Vessel Coolant Flow (GPM)
Uncontrolled W-3 ANF RCCA Bank TWINKLE Withdrawal                        See Section                         WRB-2 FACTRAN                            N/A2       3 No       0     162,840   547.0     2037.0 4 from a                              14.1.1.2                            and THINC Subcritical W-3 V-5 Condition W-3 ANF RCCA               LOFTRAN N/A           N/A     N/A   WRB-2 V-       Yes     3600   366,400   581.3     2100.0 5 Misalignment        THINC 5
Vessel Average Temperature
1 Includes reactor coolant pump heat, if applicable.
(°F)
2 N/A - Not Applicable 3
Pressurizer Pressure (PSIA)
Zero Power Doppler Power Defect at BOL assumed to be - 1000 pcm.
Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition TWINKLE FACTRAN THINC See Section 14.1.1.2 N/A2 3
4 Core Pressure 5
W-3 ANF WRB-2 and W-3 V-5 No 0
For transition cycles, pressurizer pressure is 2250 psia.
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.
Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                            Revision:   19.1 D. C. COOK NUCLEAR PLANT Table: 14.1.0-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     2 of 4
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 19.1 Table: 14.1.0-2 Page:
2 of 4 Unit 2


==SUMMARY==
==SUMMARY==
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Initial Reactor Revised        NSSS              Vessel Computer       Moderator    Moderator                                                 Vessel              Pressurizer Fault                                                            DNB        Thermal      Thermal            Average Codes      Temperature     Density    Doppler                                      Coolant              Pressure Conditions                                                        Correlation  Design        Power            Temperature Utilized      (pcm/°F)     (K/gm/cc)                                                 Flow                 (PSIA)
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Procedure      Output 1              (°F)
Moderator Density (K/gm/cc)
(GPM)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
(MWt)
Reactor Vessel Coolant Flow (GPM)
Uncontrolled         N/A         N/A           N/A       N/A       N/A         N/A           3600    N/A       N/A         N/A Boron Dilution        N/A         N/A           N/A       N/A       N/A         N/A             0      N/A       N/A         N/A Loss of Forced   LOFTRAN                                           W-3 ANF Reactor          FACTRAN           +5           N/A       Max 6 WRB-2 V-       Yes           3608   366,400   581.3 7     2100.0 (5)
Vessel Average Temperature
Coolant Flow      THINC                                                5 Locked Rotor (Peak           LOFTRAN           +5           N/A       Max (6)   N/A         N/A           3680   354,000   585.4       2312.6 Pressure)
(°F)
Locked Rotor LOFTRAN (Peak Clad                         +5           N/A       Max (6)   N/A         N/A           3680   354,000   585.4       2037.4 FACTRAN Temp) 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.
Pressurizer Pressure (PSIA)
Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                      Revision:   19.1 D. C. COOK NUCLEAR PLANT Table: 14.1.0-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     3 of 4
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 19.1 Table: 14.1.0-2 Page:
3 of 4 Unit 2


==SUMMARY==
==SUMMARY==
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Initial Reactor Revised  NSSS              Vessel Computer   Moderator   Moderator                                          Vessel              Pressurizer Fault                                                    DNB        Thermal  Thermal            Average Codes    Temperature   Density  Doppler                                Coolant              Pressure Conditions                                                  Correlation  Design    Power          Temperature Utilized    (pcm/°F)   (K/gm/cc)                                           Flow                 (PSIA)
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Procedure Output 1            (°F)
Moderator Density (K/gm/cc)
(GPM)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
(MWt)
Reactor Vessel Coolant Flow (GPM)
Locked Rotor   LOFTRAN (Rods-in-       FACTRAN         +5         N/A       Max (6) WRB-2         Yes     3608   366,400   581.3       2100.0 DNB)              THINC Loss of Normal         LOFTRAN         0         N/A       Max (6)   N/A         N/A     3680   354,000   585.4       2312.6 Feedwater Loss of Offsite Power (LOOP)
Vessel Average Temperature
LOFTRAN         0         N/A       Max (6)   N/A         N/A     3680   354,000   541.4       2312.6 to the Station Auxiliaries See Rupture of a   LOFTRAN     See Figure                       W-3 ANF N/A       Figure                 NO        0    354,000    547.0      2100.0 Steam Pipe        THINC      14.2.5-1                        W-3 V-5 14.2.5-2 Unit 2
(°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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                Revision:   19.1 D. C. COOK NUCLEAR PLANT Table: 14.1.0-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     4 of 4
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 19.1 Table: 14.1.0-2 Page:
4 of 4 Unit 2


==SUMMARY==
==SUMMARY==
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Initial Reactor Revised        NSSS                Vessel Computer       Moderator    Moderator                                                     Vessel              Pressurizer Fault                                                            DNB        Thermal      Thermal              Average Codes    Temperature     Density    Doppler                                          Coolant                Pressure Conditions                                                        Correlation    Design        Power            Temperature Utilized      (pcm/°F)   (K/gm/cc)                                                     Flow                 (PSIA)
OF INITIAL CONDITIONS AND COMPUTER CODES USED REACTIVITY COEFFICIENTS ASSUMED Fault Conditions Computer Codes Utilized Moderator Temperature (pcm/°F)
Procedure      Output 1              (°F)
Moderator Density (K/gm/cc)
(GPM)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output 1 (MWt)
(MWt)
Reactor Vessel Coolant Flow (GPM)
Rupture of a Control Rod TWINKLE       See Section                                                     3660 10  354, 000    585.4 Drive                                          N/A         ,
Vessel Average Temperature
8 9 N/A         N/A                                         2037.4 (4)
(°F)
FACTRAN          14.2.6                                                            0    162, 840    547.0 Mechanism Housing Rupture of Feedwater       LOFTRAN           N/A         .54       Max (6)     N/A         N/A           3680   354, 000   585.4       2162.6 Pipe 8
Pressurizer Pressure (PSIA)
Full Power Doppler Power defect at BOL and EOL assumed to be -966 pcm and -893 pcm respectively.
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.
9 Zero Power Doppler only Power defect at BOL and EOL assumed to be -965 pcm and -849 pcm, respective.
10 Core thermal power.
10 Core thermal power.
Unit 2
UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                  Revision:   19.1 D. C. COOK NUCLEAR PLANT Table: 14.1.0-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     1 of 1
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 19.1 Table: 14.1.0-3 Page:
1 of 1 UNIT 2


==SUMMARY==
==SUMMARY==
OF INITIAL CONDITIONS AND COMPUTER CODES USED: SEPARATE FULL VANTAGE 5 CORE ANALYSES Reactivity Coefficients Assumed Revised                  Reactor      Vessel Computer    Moderator      Moderator                                  Initial NSSS                            Pressurizer DNB        Thermal                    Vessel      Average Fault Conditions       Codes     Temperature     Density    Doppler                        Thermal Power                            Pressure Correlation  Design                  Coolant Flow Temperature Utilized    (pcm/°F)     (K/gm/cc)                                 Output) (MWt)1                             (PSIA)
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)
Procedure                  (GPM)          (°F)
Moderator Density (K/gm/cc)
Uncontrolled Rod 3608                      581.3 Cluster Assembly                         N/A3         .54       Max4 LOFTRAN                                            WRB-2         Yes         2165       366,400       567.6       2100.0 Bank Withdrawal At                        +5          N/A        Min5 361                      550.4 Power 2 Loss of Electrical Load                 N/A           .54       Max(4)
Doppler DNB Correlation Revised Thermal Design Procedure Initial NSSS Thermal Power Output) (MWt)1 Reactor Vessel Coolant Flow (GPM)
LOFTRAN                                            WRB-2         Yes         3600       366,400       581.3       2100.0 or Turbine Trip 6                        +5          N/A        Min(5)
Vessel Average Temperature
Excessive Heat Removal Due to                           N/A           .54       Min(5)   WRB-2         Yes         3600       366,400       581.3       2100.0 LOFTRAN Feedwater System                        N/A           .54       Min(5)   WRB-2        Yes            0        366,400      547.0      2100.0 Malfunction N/A            0        Min(5)
(°F)
Excess Load Increase    LOFTRAN                                            WRB-2         Yes         3600       366,400       581.3       2100.0 N/A          .54      Max(4) 1 Includes reactor coolant pump heat, if applicable.
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.
2 Multiple power levels, Tavg, and reactivity feedback cases were examined.
3 N/A - Not Applicable 4
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).
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).
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.
6 Minimum and maximum reactivity feedback cases were examined.
UNIT 2
UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                  Revision:     21.2 D. C. COOK NUCLEAR PLANT Table: 14.1.0-4 UPDATED FINAL SAFETY ANALYSIS REPORT                                  Page:       1 of 1 RPS TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN NON-LOCA SAFETY ANALYSES Limiting Trip Nominal         Point Assumed In Trip Function                                                              Time Delay Setpoint              Analysis (seconds)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 21.2 Table: 14.1.0-4 Page:
Power range high neutron flux, high setting           109%                   118%               0.5 Power range high neutron flux, low setting             25%                   35%               0.5 See              Variable, see Overtemperature T                                                                               8.0 1 Table 2.2-1       Figures 14.1.0-5,6 Variable, see Overpower T                                       in Tech Spec                                 8.0  (1)
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)
Figures 14.1.0-5,6 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 90% loop flow         87% loop flow             1.0 (From loop flow detectors) 2905 volts Undervoltage trip volts each bus                                             NA 2 1.5 each bus Underfrequency trip                                   57.5 Hz                 57 Hz               0.6 21% of narrow         0.0% of narrow Low-low steam generator level                                                                    2.0 range span            range span 1
Power range high neutron flux, high setting 109%
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.
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.
2 No explicit value assumed in the analysis. Undervoltage reactor trip setpoint assumed reached at initiation of analysis.
UNIT 2
UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                      Revision:         19 D. C. COOK NUCLEAR PLANT Table: 14.1.0-5 UPDATED FINAL SAFETY ANALYSIS REPORT                                      Page:         1 of 1 ESF ACTUATION SETPOINTS AND TIME DELAYS TO ACTUATION ASSUMED IN NON-LOCA SAFETY ANALYSES Limiting Nominal            Actuation          Time Delay ESF Actuation Function Setpoint       Setpoint Assumed         (Seconds)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
In Analyses Safety Injection (SI) 27 w/offsite
19 Table: 14.1.0-5 Page:
  - Low pressurizer pressure                                   1815 psig           1700 psig power 1 37 w/o offsite power 2 27 w/offsite
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)
  - Low steamline pressure                                     600 psig             344 psig power (1) 37 w/o offsite power (2)
Safety Injection (SI)  
Auxiliary Feedwater (AFW) 21% of narrow        0.0% of narrow
- Low pressurizer pressure 1815 psig 1700 psig 27 w/offsite power 1 37 w/o offsite power 2  
  - Low-low steam generator water level                                                                     603 range span          range span 67% of narrow         82% of narrow High-high steam generator Level Turbine Trip                                                             2.5 range span           range span Steamline Isolation on low steam line pressure                 NA4                 NA (4)               115 Feedwater Line Isolation on high-high steam               67% of narrow         82% of narrow 11 6 generator water level                                        range span          range span 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.
- Low steamline pressure 600 psig 344 psig 27 w/offsite power (1) 37 w/o offsite power (2)
Auxiliary Feedwater (AFW)  
- 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.
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.
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.
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
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.
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.
6 Feedwater Line isolation total delay time includes valve closure time and electronics and sensor delay time.
Unit 2
UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                      Revision: 20.2 D. C. COOK NUCLEAR PLANT Table: 14.1.0-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page:     1 of 4 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Reactor Trip             ESF Actuation Fault Conditions                                                                      Other Equipment       ESF Equipment Functions                Functions 14.1.1     Uncontrolled RCCA bank withdrawal             Power range high flux             NA                      NA                  NA from a subcritical condition (low setpoint) 14.1.2   Uncontrolled RCCA bank withdrawal at           Power range high flux,             NA              Pressurizer safety          NA power                      overtemperature delta-T,                               valves, steam high pressurizer pressure,                         generator safety valves high pressurizer level 14.1.3               RCCA misalignment 14.1.4               (including rod drop) 14.1.5         Uncontrolled Boron Dilution             Source range high flux             NA              Low insertion limit          NA power range high flux                               annunciators for overtemperature delta-T                                 boration 14.1.6.1     Partial and complete loss of forced       Low flow, undervoltage             NA           Steam generator safety         NA reactor coolant flow                  underfrequency                                        valves 14.1.6.2     Reactor coolant pump shaft seizure                 Low flow                     NA               Pressurizer safety         NA valves, steam (locked rotor) generator safety valves 14.1.7   Startup of an inactive reactor coolant loop               -                        -                      -                    -
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 20.2 Table: 14.1.0-6 Page:
1 1
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)
This cannot occur in Modes 1 and 2 as restricted by the Cook Nuclear Plant Unit 2 Technical Specifications.
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)
Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                    Revision:     20.2 D. C. COOK NUCLEAR PLANT Table: 14.1.0-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page:       2 of 4 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Reactor Trip         ESF Actuation Fault Conditions                                                                  Other Equipment       ESF Equipment Functions              Functions 14.1.8 Loss of external electric load or turbine High pressurizer pressure Steam generator lo-lo       Pressurizer     Auxiliary Feedwater trip                    overtemperature delta-T,          level          safety valves, steam        System lo-lo steam generator                        generator safety valves level 14.1.9         Loss of normal feedwater             Steam generator lo-lo   Steam generator lo-lo Steam generator safety Auxiliary Feedwater level, manual              level            valves, pressurizer       System safety valves 14.1.10 Feedwater system malfunctions that result   Power range high flux,      High-high steam      Feedwater isolation          NA in an increase in feed water flow       (low and high setpoints),   generator level-steam generator lo-lo   produced feedwater level (Intact steam   isolation and turbine generators)                trip 14.1.11         Excessive load increase             Power range high flux,           NA                  Pressurizer              NA overtemperature delta-T,                         safety valves, steam overpower delta-T                          generator safety valves 14.1.12   Loss of offsite power to the station       Steam generator lo-lo   Steam generator lo-lo     Steam generator     Auxiliary Feedwater Auxiliaries                              level                level            valves, pressurizer       System safety valves 14.2.4       Steam generator tube failure           Reactor Trip System       Engineered Safety   Steam generator safety   Emergency Core Features Actuation    and/or relief valves, Cooling System, System          steamline stop valves  Auxiliary Feedwater System, Emergency Power System Unit 2
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 20.2 Table: 14.1.0-6 Page:
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                                  Revision:     20.2 D. C. COOK NUCLEAR PLANT Table: 14.1.0-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page:       3 of 4 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Reactor Trip           ESF Actuation Fault Conditions                                                                Other Equipment       ESF Equipment Functions              Functions 14.2.5         Rupture of a Steam Line           SIS, low pressurizer     Low pressurizer      Feedwater isolation  Auxiliary Feedwater pressure, manual           pressure low     steamline stop valves    System, Safety compensated steamline                           Injection System pressure, high containment pressure, manual Inadvertent opening of a steam generator           SIS              Low pressurizer      Feedwater isolation  Auxiliary Feedwater relief or safety valve                                       pressure, low     steamline stop valves   System, Safety compensated                                Injection System steamline pressure 14.2.6 Spectrum of RCCA ejection accidents     Power range high flux,           NA                    NA                    NA high positive flux rate 14.2.8     Feedwater system pipe break         Steam generator lo-lo     High containment      Steamline isolation  Auxiliary Feedwater level, high pressurizer     pressure, steam        valves, feedline      System, Safety pressure, SIS      generator lo-lo water isolation, pressurizer Injection System level, low        self-actuated safety compensated steamline      valves, steam pressure      generator safety valves Unit 2
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 20.2 Table: 14.1.0-6 Page:
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                            Revision:   20.2 D. C. COOK NUCLEAR PLANT Table: 14.1.0-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page:       4 of 4 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR FAULT CONDITIONS Reactor Trip       ESF Actuation Fault Conditions                                                        Other Equipment         ESF Equipment Functions          Functions 14.3 Loss of coolant accidents resulting from   Reactor Trip System Engineered Safety  Service Water System      Emergency Core the spectrum of postulated piping breaks                       Features Actuation  Component Cooling        Cooling System, within the reactor coolant pressure                             System         Water System steam     Auxiliary Feedwater boundary                                                          generator safety and/or System, Containment relief valves        Heat Removal System, Emergency Power System Unit 2
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 20.2 Table: 14.1.0-6 Page:
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                Revision:     18 D. C. COOK NUCLEAR PLANT Table: 14.1.1-1 UPDATED FINAL SAFETY ANALYSIS REPORT                                Page:     1 of 1 TIME SEQUENCE OF EVENTS Accident                                   Event                               Time (sec)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
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 Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                  Revision:   16.1 D. C. COOK NUCLEAR PLANT Table: 14.1.2B-1 UPDATED FINAL SAFETY ANALYSIS REPORT                                  Page:       1 of 1 TIME SEQUENCE OF EVENTS (FULL VANTAGE 5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.1.2B-1 Page:
Accident                                           Event                       Time (sec)
1 of 1 Unit 2 TIME SEQUENCE OF EVENTS (FULL VANTAGE 5 CORE)
Uncontrolled RCCA Bank Withdrawal At Full Power Initiation of uncontrolled RCCA bank withdrawal at Case A: (high insertion rate a high reactivity insertion                                 0 max feedback) rate (80 pcm/sec)
Accident Event Time (sec)
Power range high neutron flux high trip signal 5.8 initiated Rods begin to fall into core                               6.3 Minimum DNBR occurs                                       6.4 Case B: (small insertion rate,     Initiation of uncontrolled RCCA bank withdrawal at 0
Uncontrolled RCCA Bank Withdrawal At Full Power Case A: (high insertion rate max feedback)
max feedback)              a small reactivity insertion rate (4 pcm/sec)
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)
Overtemperature T reactor trip signal initiated         314.5 Minimum DNBR occurs                                     316.2 Rods begin to fall into core                             316.5 Unit 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                Revision:   30.0 D. C. COOK NUCLEAR PLANT Table: 14.1.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT                    Page:     1 of 1 TIME SEQUENCE OF EVENTS Accident                                   Event           Time (sec)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 30.0 Table: 14.1.5-1 Page:
1 of 1 Unit 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
Uncontrolled Boron Dilution
Uncontrolled Boron Dilution
: 1. Dilution during Refueling     Dilution begins                         0 Shutdown margin lost                 1848
: 1. Dilution during Refueling Dilution begins 0
: 2. Dilution during startup       Dilution begins                         0 Shutdown margin lost                 2100
Shutdown margin lost 1848
: 2. Dilution during startup Dilution begins 0
Shutdown margin lost 2100
: 3. Dilution during full power operation
: 3. Dilution during full power operation
: a. Automatic reactor control Dilution begins                         0 Shutdown margin lost                 2760
: a. Automatic reactor control Dilution begins 0
: b. Manual reactor control   Dilution begins                         0 Overtemperature T reactor trip       90 Shutdown margin lost                 2760 Unit 2
Shutdown margin lost 2760
: b. Manual reactor control Dilution begins 0
Overtemperature T reactor trip 90 Shutdown margin lost 2760 UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                        Revision:   16.1 D. C. COOK NUCLEAR PLANT Table: 14.1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT                            Page:     1 of 1 TIME SEQUENCE OF EVENTS Time Accident                                 Event (sec)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.1.6-1 Page:
Loss of Forced Reactor Coolant Flow Four loops in operation, four pumps coasting down All operating pumps lose power and begin 0.0 coasting down Reactor coolant pump under-voltage trip 0.0 point reached 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 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                            Revision:   18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.6-2 UPDATED FINAL SAFETY ANALYSIS REPORT                            Page:     1 of 1 TIME SEQUENCE OF EVENTS Accident                                       Event             Time (sec)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 18.1 Table: 14.1.6-2 Page:
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 0.02 setpoint reached Rods begin to drop                   1.02 Time at which minimum DNBR is 2.2 predicted to occur Maximum RCS pressure occurs           3.10 Maximum clad temperature occurs       3.60 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                                    Revision:     18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT                                      Page:       1 of 1 TIME SEQUENCE OF EVENTS (FULL VANTAGE 5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 18.1 Table: 14.1.8-1 Page:
Accident                                         Event                       Time (sec)
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
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 12.2 reached Peak pressurizer pressure occurs                     14.2 Rods begin to drop                                   12.5 Minimum DNBR occurs                                 16.0
: 1. With pressurizer control (min fdbk)
: 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 53.7 trip point reached Rods begin to drop                                   55.7 Minimum DNBR occurs                                     1
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
: 3. Without pressurizer control (min fdbk)     Loss of electrical load                               0.0 High pressurizer pressure reactor trip point 7.3 reached Peak pressurizer pressure occurs                     9.3 Rods begin to drop                                   9.0 Minimum DNBR occurs                                   (1)
: 2. With pressurizer control (max fdbk)
: 4. Without pressurizer control (max fdbk)     Loss of electrical load                               0.0 High pressurizer pressure reactor trip point 7.4 reached Rods begin to drop                                   9.4 Peak pressurizer pressure occurs                     9.5 Minimum DNBR occurs                                   (1) 1 DNBR never decreases below its initial value.
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
UNIT 2
: 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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                            Revision:   16.1 D. C. COOK NUCLEAR PLANT Table: 14.1.9-1 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS Time Accident                               Event (sec)
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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 115.7 Supply the Steam Generators 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 4800 heat removal capacity UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                          Revision:     17 D. C. COOK NUCLEAR PLANT Table:14.1.10B-1 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
Time Accident                                   Event (sec)
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:
Feedwater System Malfunctions:
Excessive feedwater flow at full power to a single steam generator (Manual Rod Control)
Excessive feedwater flow at full power to a single steam generator (Manual Rod Control)
One main feedwater control valve fails fully 0.0 open Hi-hi steam generator water level signal 30.2 generated Turbine trip occurs due to hi-hi steam 32.7 generator water level Minimum DNBR occurs                                 34.0 Reactor trip occurs due to turbine trip             34.7 Feedwater isolation achieved                       41.2 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                          Revision:     17 D. C. COOK NUCLEAR PLANT Table:14.1.10B-2 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
Time Accident                                   Event (sec)
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:
Feedwater System Malfunctions:
Excessive feedwater flow at full power to a single steam generator (Automatic Rod Control)
Excessive feedwater flow at full power to a single steam generator (Automatic Rod Control)
One main feedwater control valve fails fully 0.0 open Hi-hi steam generator water level signal 30.1 generated Turbine trip occurs due to hi-hi steam generator 32.6 water level Minimum DNBR occurs                                 33.0 Reactor trip occurs due to turbine trip             34.6 Feedwater isolation achieved                       41.1 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                          Revision:     17 D. C. COOK NUCLEAR PLANT Table:14.1.10B-3 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
Time Accident                                   Event (sec)
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:
Feedwater System Malfunctions:
Excessive feedwater flow at full power to all four steam generators (Manual Rod Control)
Excessive feedwater flow at full power to all four steam generators (Manual Rod Control)
All four main feedwater control valves fail fully 0.0 open Hi-hi steam generator water level signal 31.5 generated Turbine trip occurs due to hi-hi steam generator 34.0 water level Minimum DNBR occurs                                 34.5 Reactor trip occurs due to turbine trip             36.0 Feedwater isolation achieved                         42.5 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                          Revision:     17 D. C. COOK NUCLEAR PLANT Table:14.1.10B-4 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
Time Accident                                   Event (sec)
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:
Feedwater System Malfunctions:
All four main feedwater control valves fail Excessive feedwater flow at full                                                       0.0 fully open power to all four steam generators (Automatic Rod Control)
Excessive feedwater flow at full power to all four steam generators (Automatic Rod Control)
Hi-hi steam generator water level signal 31.8 generated Turbine trip occurs due to hi-hi steam 34.3 generator water level Minimum DNBR occurs                               35.5 Reactor trip occurs due to turbine trip           36.3 Feedwater isolation achieved                     42.8 UNIT 2
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 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                        Revision:   16.1 D. C. COOK NUCLEAR PLANT Table:14.1.11B-1 UPDATED FINAL SAFETY ANALYSIS REPORT                          Page:     1 of 1 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:14.1.11B-1 Page:
Time Accident                                   Event (sec)
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS (FULL V-5 CORE)
Accident Event Time (sec)
Excessive Load Increase
Excessive Load Increase
: 1. Manual reactor control (Min fdbk)     10% step load increase                     0.0 Equilibrium conditions reached           160.0
: 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
: 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
: 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 UNIT 2
: 4. Automatic reactor control (Max fdbk) 10% step load increase 0.0 Equilibrium conditions reached 70.0 UFSAR Revision 31.0


UFSAR Revision 31.0 INDIANA MICHIGAN POWER                            Revision:   16.1 D. C. COOK NUCLEAR PLANT Table: 14.1.12-1 UPDATED FINAL SAFETY ANALYSIS REPORT                              Page:     1 of 1 TIME SEQUENCE OF EVENTS Time Accident                                 Event (sec)
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.1.12-1 Page:
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 117.0 Supply the Steam Generators Cold Auxiliary Feedwater is Delivered to the Steam 534.0 Generators Core decay heat decreases to auxiliary feedwater heat
1 of 1 UNIT 2 TIME SEQUENCE OF EVENTS Accident Event Time (sec)
                                                                                      ~800.0 removal capacity Peak water level in pressurizer occurs                       1406.0 UNIT 2}}
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 31.0}}

Latest revision as of 11:57, 27 November 2024

1 to Updated Final Safety Analysis Report, Chapter 14.1, Tables 14.1.0-1 to 14.1.12-1 (Unit 2)
ML22340A175
Person / Time
Site: Cook  
Issue date: 11/30/2022
From:
Indiana Michigan Power Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML22340A137 List: ... further results
References
AEP-NRC-2022-62
Download: ML22340A175 (1)
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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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)

Auxiliary Feedwater (AFW)

- 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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

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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 31.0

IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table: 14.1.9-1 Page:

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 31.0

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

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 31.0

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 31.0

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 31.0

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 31.0

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 31.0

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 31.0

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