ML21125A521
Text
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revised: 28.0 D. C. COOK NUCLEAR PLANT Table: 14.1-1 Page: 1 of 2 UPDATED FINAL SAFETY ANALYSIS REPORT Unit 1 Design Power Capability Parameters Used in Non-LOCA Safety Analyses (MUR Power (Reduced Temperature (Return to RCS (Rerating)2 Uprate)1,2 and Pressure)1,2 NOP/NOT)1 Parameter Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Case 7 Case 8 NSSS Power, MWt 3,327 3,327 3,262 3,262 3,425 3,425 3,327 3,327 Core Power, MWt 3,315 3,315 3,250 3,250 3,413 3,413 3,315 3,315 RCS Flow, gpm/loop 83,200 83,200 83,200 83,200 88,500 88,500 83,200 83,200 Minimum Measured Flow, gpm/loop 84775 84775 84775 84775 91600 91600 84,775 84,775 RCS Temperatures, ºF Core Outlet 593.1 613.6 589.7 611.9 583.6 614.0 593.1 613.6 Vessel Outlet 588.2 609.1 586.8 609.1 580.7 611.2 588.2 609.1 1
The non-LOCA analyses are based on a Thermal Design Flow (TDF) of 83,200 gpm / loop and a Minimum Measured Flow (MMF) of 84,775 gpm / loop.
However, subsequent evaluations were performed to show that the following higher flows are also supported: 88,500 gpm / loop (TDF) and 90,725 gpm / loop (MMF).
2 Cook Unit 1 is not licensed to operate at the rerated conditions specified by Cases 5 and 6 with 30% steam generator tube plugging (SGTP) levels. However, several events that were previously performed using these conditions were subsequently evaluated to support the 30% SGTP program. Hence, the rerated conditions are also specified in this table for completeness.
Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revised: 28.0 D. C. COOK NUCLEAR PLANT Table: 14.1-1 Page: 2 of 2 UPDATED FINAL SAFETY ANALYSIS REPORT Unit 1 Design Power Capability Parameters Used in Non-LOCA Safety Analyses (MUR Power (Reduced Temperature (Return to RCS (Rerating)2 Uprate)1,2 and Pressure)1,2 NOP/NOT)1 Parameter Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Case 7 Case 8 Core Average 557.6 579.5 555.8 579.4 549.7 581.8 557.6 579.5 Vessel Average 553.7 575.4 553.0 576.3 547.0 578.7 553.7 575.4 Vessel/Core Inlet 519.2 541.7 519.2 543.5 513.3 546.2 519.2 541.7 Steam Generator Outlet 518.9 541.5 518.9 543.2 513.1 546.0 518.9 541.5 Zero Load 547.0 547.0 547.0 547.0 547.0 547.0 547.0 547.0 2,250 or 2,250 or 2,250 or 2,250 or 2,250 or 2,250 or RCS Pressure, psia 2,250 2,250 2,100 2,100 2,100 2,100 2,100 2,100 Steam Pressure, psia 618 765 595 749 603 820 618 765 Steam Flow (106 lb/hr total) 14.44 14.50 14.12 14.17 14.98 15.07 14.44 14.50 Feedwater Temp., ºF 437.4 437.4 434.8 434.8 442 442 437.4 437.4 SG Tube Plugging, % 30 30 30 30 10 10 30 30 Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 22.1 D. C. COOK NUCLEAR PLANT Table: 14.1-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 REACTOR TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN SAFETY ANALYSES 1 TIME LIMITING REACTOR TRIP POINT REACTOR TRIP FUNCTION DELAY ASSUMED IN ANALYSIS (sec)
Power range high neutron flux, high setting 118 percent 0.5 Power range high neutron flux, low setting 35 percent 0.5 Overtemperature T Variable, see Figure 14.1-1, -2, -3 & -4 8.0 2 Overpower T Variable, see Figure 14.1-1, -2, -3 & -4 8.0 3 High pressurizer pressure 2420 psig 2.0 Low pressurizer pressure 1825 psig 4 2.0 High pressurizer water level 100% NRS 2.0 Low reactor coolant flow (from loop flow 87 percent loop flow 1.0 detectors)
Undervoltage trip 5 1.5 Low-low steam generator level 0.0 percent of narrow range level span 2.0 High-High steam generator level 6: 82 percent of narrow range level span
- Turbine Trip 2.5
- Feedwater Isolation 11.0 1
The control rod scram time to dashpot is 2.4 seconds.
2 Total time delay (including RTD time response, and trip circuit, 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 channel electronics delay presented in the UFSAR Table 14.1-2 (Unit 1) or Table 7.2-6 (Unit 2). An evaluation has been performed (Reference 11) that demonstrates that the analyses remains bounding, given that the total 8.0 second time delay in the above table is satisfied.
3 Overpower T reactor trip was assumed in the steamline break mass/energy release outside containment calculations.
4 A value of 1845 psig is used in LOCA analyses.
5 No explicit value assumed in the analysis. Undervoltage trip setpoint assumed reached at initiation of analysis.
6 Time delay between High-High steam generator level and Turbine Trip is 2.5 seconds. The Turbine Trip subsequently causes a Reactor Trip.
UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 14.1-3 Page: 1 of 4 UPDATED FINAL SAFETY ANALYSIS REPORT Summary of Initial Conditions and Computer Codes Used Reactiv ity Coefficients Assumed Rev ised Initial NSSS V essel Computer Moderator Reactor V essel Pressuriz er Fault Moderator DNB Th ermal Th ermal Av erage Codes Temperature Density Doppler Coolant Flow Pressure Conditions Correlation Design Power Output Temperature Utiliz ed (pcm / °F) (K/gm/cc) (gpm) 1 (psia)
Procedure (MWt) (°F)
Uncontrolled RCCA B ank TWINK LE W-3/WRB -1 Withdrawal FACTRAN Refer to Section 14.1.1 Min 2 See Section No 0 146,432 547 2033 from a THINC 14.1.1 Subcritical Condition Uncontrolled Min 3,327 575.45 RCCA B ank LOFTRAN + 5 .54 and WRB -1 Y es 1996 339,100 564.58 2,100 Withdrawal at Max 4 333 549.93 Power3 1
The non-LOCA analyses are based on a Thermal Design Flow (TDF) of 83,200 gpm / loop and a Minimum Measured Flow (MMF) of 84,775 gpm / loop.
However, subsequent evaluations were performed to show that the following higher flows are also supported: 88,500 gpm / loop (TDF) and 90,725 gpm / loop (MMF).
2 Minimum Doppler power defect (pcm / %power) = -9.55 + 0.00104Q where Q is in MWt.
3 Multiple power levels, Tavg, and reactivity feedback cases were ex amined.
4 Max imum Doppler power defect (pcm / %power) = -19.4 + 0.002Q where Q is in MWt.
5 Value used in the DNB analysis is performed at the MUR power uprate program max imum Tavg of 575.4°F.
Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 14.1-3 Page: 2 of 4 UPDATED FINAL SAFETY ANALYSIS REPORT Summary of Initial Conditions and Computer Codes Used Reactiv ity Coefficients Assumed Rev ised Initial NSSS V essel Computer Moderator Reactor V essel Pressuriz er Fault Moderator DNB Th ermal Th ermal Av erage Codes Temperature Density Doppler Coolant Flow Pressure Conditions Correlation Design Power Output Temperature Utiliz ed (pcm / °F) (K/gm/cc) (gpm) 1 (psia)
Procedure (MWt) (°F)
RCCA LOFTRAN N/A7 NA NA WRB -1 Y es 3,250 339,100 576.35 2,100 Misalignment 6 THINC Uncontrolled 3,425 N/A N/A N/A N/A N/A N/A N/A N/A N/A B oron Dilution 0 Loss of Forced LOFTRAN Reactor FACTRAN + 5 N/A Max WRB -1 Y es 3,270 339,100 576.35 2,100 Coolant Flow6 THINC Locked Rotor LOFTRAN + 5 N/A Max N/A N/A 3,335 332,800 581.4 2,317 (Peak Pressure)
Locked Rotor LOFTRAN
+ 5 N/A Max N/A N/A 3,335 332,800 581.4 2,033 (Peak Clad Temp) FACTRAN LOFTRAN Locked Rotor FACTRANT + 5 N/A Max WRB -1 Y es 3,270 339,100 576.3 2,100 (Rods-in-DNB )6 HINC 6
An uprated core power of 3315 MWt (NSSS power of 3327 MWt) is supported via an evaluation that addresses a reduction in power uncertainty from 2% to approx imately 0.3%.
7 N/A - Not Applicable Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 14.1-3 Page: 3 of 4 UPDATED FINAL SAFETY ANALYSIS REPORT Summary of Initial Conditions and Computer Codes Used Reactiv ity Coefficients Assumed Rev ised Initial NSSS V essel Computer Moderator Reactor V essel Pressuriz er Fault Moderator DNB Th ermal Th ermal Av erage Codes Temperature Density Doppler Coolant Flow Pressure Conditions Correlation Design Power Output Temperature Utiliz ed (pcm / °F) (K/gm/cc) (gpm) 1 (psia)
Procedure (MWt) (°F)
Loss of Max E lectrical Load Y es 3,327 339,100 576.35 2,100 LOFTRAN + 5 .54 and WRB -1 and/or Turbine No 3327 332,800 581.4 2033 Min Trip8 Loss of Normal LOFTRAN + 5 N/A Max N/A N/A 3,409 332,800 548.9 2,317 Feedwater9 E x cessive Heat Removal9 Due 3,327 576.35 to Feedwater LOFTRAN N/A .54 Min WRB -1 Y es 339,100 2,100 0 547 System Malfunction E x cess Load Max Increase9 LOFTRAN N/A 0 and .54 and WRB -1 No 3,425 366,400 578.7 2,100 Incident Min 8
Minimum and Max imum reactivity feedback cases were ex amined 9
Values presented were used in the rerating analysis. Subsequent evaluations support the 30% SGTP parameters presented as Cases 1 and 2 of Table 14.1-1 Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 27.0 D. C. COOK NUCLEAR PLANT Table: 14.1-3 Page: 4 of 4 UPDATED FINAL SAFETY ANALYSIS REPORT Summary of Initial Conditions and Computer Codes Used Reactiv ity Coefficients Assumed Rev ised Initial NSSS V essel Computer Moderator Reactor V essel Pressuriz er Fault Moderator DNB Th ermal Th ermal Av erage Codes Temperature Density Doppler Coolant Flow Pressure Conditions Correlation Design Power Output Temperature Utiliz ed (pcm / °F) (K/gm/cc) (gpm) 1 (psia)
Procedure (MWt) (°F)
Loss of Offsite Power (LOOP)9 LOFTRAN + 5 N/A Max N/A N/A 3,409 332,800 548.9 2033 to the Station Aux iliaries Rupture of a LOFTRAN See Figure See Figure Steam Pipe (at N/A W-3 N/A 0 332,800 547 2,100 THINC 14.2.5-1 14.2.5-2 hot z ero power)
Rupture of a LOFTRAN Least Steam Pipe (at VIPRE N/A 0.54 WRB -1 Y es 3,327 339,100 575.4 2,250 Negative10 full power) ANC Rupture of a Control Rod TWINK LE See Section 3,335 332,800 581.4 Drive N/A Min N/A N/A 2,033 FACTRAN 14.2.6 0 146,432 547 Mechanism Housing 10 Least negative power defect (pcm / %power) = 9.55 + 0.0355Q , where Q is in %power.
Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 18.2 D. C. COOK NUCLEAR PLANT Table: 14.1-4 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS POWER RANGE NEUTRON FLUX SETPOINT AND ESTIMATED ERROR INSTRUMENTATION ALLOWANCES:
ERRORS: (% of rated power)
(% of rated power)
Nominal Setpoint 109 -
Calorimetric Error 21 0.31 Axial power distribution effects on 5 3 total ion chamber current Instrumentation channel drift and 2 1.0 setpoint reproducibility Maximum overpower reactor trip point assuming all individual errors 119 -
are simultaneously in the most adverse direction 1
MUR power uprate uses reduced calorimetric error allowance. The sum of the change in Rated Thermal Power defined in the Technical Specifications and the MUR reduced calorimetric error allowance is equal to, or less than, the original +2% value supported by the safety analyses.
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 16.6 D. C. COOK NUCLEAR PLANT Table: 14.1-5 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 DONALD C. COOK NUCLEAR PLANT UNIT 1 SGTP PROGRAM INPUT ASSUMPTIONS FOR RCS VOLUMES1 INITIAL CONDITIONS INPUT ASSUMPTIONS 0% SGTP 30% SGTP Reactor Vessel (ft3) 4643 4643 Steam Generators (ft3 - Total) 43572 3442 (2)
Reactor Coolant Pumps (ft3 - Total) 324 324 Loop Piping (ft3 - Total) 1185 1185 Surge Line Piping (ft3) 43 43 Pressurizer (ft3) 1834 1834 Total RCS Volume (ft3) (Ambient Conditions) 12,386 11,471 Total RCS Volume (ft3) 12,758 11,815 (Hot Conditions includes 3% for thermal expansion) 1 The volumes presented in this table represent the reactor coolant system volumes from the Westinghouse IMP database SEC-LIS-4428-C3. Note that the volumes documented in this table are slightly different than those used in the Westinghouse analysis of record, but have been evaluated for acceptability as documented in letter AEP-99-485.
2 The SG tube volume is assumed to be 762 ft3/SG (3048 ft3 total). The increase in SG tube plugging from 0% to 30% results in a total reduction in SG tube volume of approximately 915 ft3. The reduction between the SG tube volume and SG tube plugging is assumed to be a linear relationship; e.g. at 15% SGTP, total volume reduction is 0.15*(3048 ft3) = 457.2 ft3.
UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 14.1-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 DONALD C. COOK NUCLEAR PLANT UNIT 1 SGTP PROGRAM INPUT ASSUMPTIONS FOR STEAM GENERATOR SECONDARY MASS INITIAL CONDITIONS1 INPUT ASSUMPTIONS 0% SGTP 30% SGTP Cases Original Design Low Temp High Temp Case Case A1 Case A2 Steam generator secondary side mass 106,506 106,7992 112,1923 (Total lbs/SG) 1 Initial conditions are presented for SGTP levels of 0% (Original Design) and 30% to bound the range of SGTP levels.
2 For Tavg of 553°F.
3 For Tavg of 576°F.
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 14.1-7 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 DONALD C. COOK NUCLEAR PLANT UNIT 1 SGTP PROGRAM INPUT ASSUMPTIONS FOR REACTOR COOLANT SYSTEM PRESSURE DROP(1)
Initial Conditions Input Assumptions 0% SGTP Pressure 30% SGTP Pressure Drop, psi Drop, psi Reactor Vessel, including nozzles (psi) 47.21 44.26 Loop Piping (psi) 5.14 4.55 Steam Generator (psi) 43.23 53.58 Total (psi) 95.58(1) 102.39(1)
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 14.1-8 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 ECCS INJECTION TO RECIRCULATION SWITCHOVER MODEL FOR THE CONTAINMENT RESPONSE ANALYSIS TIME AND RWST DELIVERED VOLUME AFTER RWST EARLIEST SWITCHOVER SETPOINT IS REACHED TIME AFTER RWST SWITCHOVER DELIVERED EVENT SETPOINT IS VOLUME REACHED (gal)
(sec)
Earliest switchover setpoint is reached 0 280,000 Stop RHR / CTS pumps 0 280,000 Start RHR / CTS pumps 300 285,975 CCP and SI suction realignment to RHR discharge 1707 314,000 (minimum RWST delivered)
UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 TIME SEQUENCE OF EVENTS TIME ACCIDENT EVENT (sec)
Uncontrolled RCCA Bank Withdrawal At Full Power Initiation of uncontrolled RCCA bank Case A (high insertion rate, max withdrawal at a high reactivity 0 feedback) insertion rate (80 pcm/sec)
Power range high neutron flux high 4.9 trip signal initiated Rods begin to fall into core 5.4 Minimum DNBR occurs 5.8 Initiation of uncontrolled RCCA bank Case B (small insertion rate, max withdrawal at a small reactivity 0 feedback) insertion rate (4 pcm/sec)
Overtemperature T reactor trip signal 321.2 initiated Minimum DNBR occurs 322.1 Rods begin to fall into core 323.2
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 SEQUENCE OF EVENTS FOR LOSS OF FLOW AND LOCKED ROTOR ACCIDENTS TIME ACCIDENT EVENT (sec)
All pumps lose power and begin coasting down, Complete Loss of Flow 0.0 undervoltage trip signal generated Rods begin to drop 1.50 Minimum DNBR occurs 3.40 One operating pump loses power and begins coasting Partial Loss of Flow 0.0 down Low reactor coolant flow trip setpoint reached in 1.74 faulted loop Rods begin to drop 2.74 Minimum DNBR occurs 3.90 Locked Rotor One pump rotor seizes 0.0 Low reactor coolant flow trip setpoint reached in 0.04 faulted loop Rods begin to drop 1.04 Time at which minimum DNBR is predicted to occur 2.6 Maximum RCS pressure occurs 3.20 Maximum clad temperature occurs 3.49 UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 28 D. C. COOK NUCLEAR PLANT Table: 14.1.6-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 Parameters Used for the Radiological Consequence Analysis of a Locked Rotor Event Parameter Value Core Power Level 3480 MWt Fuel Clad Failure 11%
Core Fractions Released from Damaged Rods I-131 0.08 Other Halogens 0.05 Kr-85 0.10 Other Noble Gases 0.05 Alkali Metals 0.12 Fuel Rod Peaking Factor 1.65 Core Release Fraction Multiplier for High Burnup Fuel 1.0104 Secondary Coolant Limit for Normal Operation 0.1 µCi/gm D.E. I-131 Primary Coolant Mass 466,141.5 lbm Secondary System Mass 97,515.7 lbm/SG (minimum) 161,000 lbm/SG (maximum)
Primary-to-Secondary Leak Rate 1 gpm to all steam generators Steam Generator Steam Release 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 460,000 lbm 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1,256,000 lbm 8 - 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 1,347,000 lbm Partition Coefficients Iodines 100 Alkali Metals 500 Nobles Gases 1 Duration of Intact SB Tube Uncovery After Reactor Trip 40 minutes Intact Tube Leakage Flashing Fraction During Uncovery 0-400 seconds: 8%
400-900 seconds: 6%
900-1700 seconds: 5.5%
1700 seconds-40 min: 4%
Iodine Chemical Form Elemental 97%
Organic 3%
Particulate 0%
Unit 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 28 D. C. COOK NUCLEAR PLANT Table: 14.1.6-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 Parameters Used for the Radiological Consequence Analysis of a Locked Rotor Event Parameter Value Release Location Offsite Unit 1 Main Steam Enclosures Onsight Unit 2 PORVs/MSSVs Offsite Breathing Rates 0-8 hours 3.5E-04 m3/sec 8-24 hours 1.8E-04 m3/sec 24-720 hours 2.3E-04 m3/sec Control Room Parameters Volume 50,616 ft3 Normal Ventilation Makeup Flow Rate 880 cfm Emergency Ventilation Makeup Flow Rate 880 cfm Emergency Ventilation Recirculation Flow Rate 4520 cfm Emergency Ventilation Filter Efficiency1 Elemental Iodine 94.05%
Organic Iodide 94.05%
Particulates 98.01%
Delay to Switch to Emergency Mode 20 minutes (manual)
Unfiltered Inleakage 40 cfm Occupancy Factors 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.6 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.4 Breathing Rate 3.5E-04 m3/sec 1
Includes 1% filter bypass leakage Unit 1
UFSAR REVISION 30.0 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 SEQUENCE OF EVENTS FOR LOSS EXTERNAL ELECTRICAL LOAD TIME CASE EVENT (sec)
Minimum Feedback with Pressure Loss of external electrical load 0.0 Control OTT trip setpoint reached 13.1 Peak RCS pressure occurs 13.1 Rods begin to drop 15.1 Minimum DNBR occurs 16.6 Maximum Feedback with Pressure Loss of external electrical load 0.0 Control Minimum DNBR occurs 0.0 Peak RCS pressure occurs 9.3 Low-low steam generator level trip 35.2 setpoint reached Rods begin to drop 37.2 Minimum Feedback without Pressure Loss of external electrical load 0.0 Control Minimum DNBR occurs 0.0 High pressurizer pressure trip setpoint 8.6 reached Rods begin to drop 10.6 Peak RCS pressure occurs 11.6 Maximum Feedback without Pressure Loss of external electrical load 0.0 Control Minimum DNBR occurs 0.0 High pressurizer pressure trip setpoint 9.2 reached Rods begin to drop 11.2 Peak RCS pressure occurs 12.1 UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 16.6 D. C. COOK NUCLEAR PLANT Table: 14.1.9-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEEDWATER ACCIDENT EVENT TIME (sec)
Loss of Normal Feedwater Main feedwater flow stops 10.0 Low-low steam generator water level trip 35.7 signal initiated Rods begin to fall into core 37.7 Auxiliary Feedwater Pumps Start and 95.7 Supply the Steam Generators Cold Auxiliary Feedwater is Delivered to the Steam Generators (MFW purged)
Steam Generators #1 and #4 311.5 Steam Generators #2 and #3 1027 Peak water level in pressurizer occurs 6426 Core decay heat plus RCP heat decreases to auxiliary feedwater heat removal ~6426 capacity UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.10-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 SEQUENCE OF EVENTS FOR EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTION TIME CASE EVENT (sec)
One main feedwater control 0.0 Feedwater System Malfunctions: Excessive valve fails fully open feedwater flow at full power to a single steam Hi-hi steam generator water 40.4 generator (Automatic Rod Control) level signal generated Turbine trip occurs due to hi-hi 42.9 steam generator water level Minimum DNBR occurs 43.0 Reactor trip occurs due to 44.9 turbine trip Feedwater isolation achieved 51.4 One main feedwater control 0.0 Feedwater System Malfunctions: Excessive valve fails fully open feedwater flow at full power to a single steam Hi-hi steam generator water 40.5 generator (Manual Rod Control) level signal generated Turbine trip occurs due to hi-hi 43.5 steam generator water level Minimum DNBR occurs 43.5 Reactor trip occurs due to 45.0 turbine trip Feedwater isolation achieved 51.5 UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 18.1 D. C. COOK NUCLEAR PLANT Table: 14.1.10-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 SEQUENCE OF EVENTS FOR EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTION TIME CASE EVENT (sec)
One main feedwater control valve Feedwater System Malfunctions: Excessive 0.0 fails fully open feedwater flow at full power to all four steam Hi-hi steam generator water level generators (Automatic Rod Control) 42.7 signal generated Minimum DNBR occurs 45.0 Turbine trip occurs due to hi-hi 45.2 steam generator water level Reactor trip occurs due to turbine 47.2 trip Feedwater isolation achieved 53.7 One main feedwater control valve Feedwater System Malfunctions: Excessive 0.0 fails fully open feedwater flow at full power to all four steam Hi-hi steam generator water level generators (Manual Rod Control) 42.4 signal generated Turbine trip occurs due to hi-hi 44.9 steam generator water level Minimum DNBR occurs 42.5 Reactor trip occurs due to turbine 46.9 trip Feedwater isolation achieved 53.4 UNIT 1
UFSAR REVISION 30.0 INDIANA MICHIGAN POWER Revision: 16.6 D. C. COOK NUCLEAR PLANT Table: 14.1.12-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 TIME SEQUENCE OF EVENTS FOR LOSS OF ALL AC POWER TO STATION AUXILIARIES ACCIDENT EVENT TIME (sec)
Loss of All AC Power to AC power is lost 10.0 Station Auxiliaries Main feedwater flow stops 10.0 Low-low steam generator water level trip 35.7 signal initiated Rods begin to fall into core 37.7 Reactor coolant pumps begin to coastdown 39.7 Auxiliary Feedwater Pumps Start and 115.7 Supply the Steam Generators Cold Auxiliary Feedwater is Delivered to the Steam Generators (MFW purged)
Steam Generators #1 and #4 331.5 Steam Generators #2 and #3 1047 Peak water level in pressurizer occurs 3980 Core decay heat decreases to auxiliary
~3980 feedwater heat removal capacity UNIT 1
UFSAR REVISION 30.0 INDIANA AND MICHIGAN POWER Revision: 28.0 Table: 14.1.13-D. C. COOK NUCLEAR PLANT 1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Potential Turbine-Generator Missiles Unit 1 (GE) 1 Weigh t Failure Speed (Rpm) V elocity & Kinetic Energy after Part (Name) Cross Sectional Area Av erage (Pounds) (and % of 180 0 ) Leav ing Turbine Casing Min. (Sq . Ft.) Max . (Ft/Sec) (Ft-lbs)
Vane of Last Stage B ucket 54 3,130 (174%) 0.033 1.6 1,170 1 x 106 120ºF Segment of Last Stage Wheel 8,264 3,190 (177%) 5.16 8.43 11.70 409 21.5 x 106 Unit 2 (B B ) 2 Weigh t Failure Speed (Rpm) V elocity & Kinetic Energy Part (Name) Cross Sectional Area Av erage (Pounds) (and % of 180 0 ) after Leav ing Turbine Casing Min. (Sq . Ft.) Max . (Ft/Sec) (Ft-lbs)
Vane of Last Stage B ucket 168 3,040 (169%) 0.106 3.64 1,135 3.4 x 106 120º Segment of Nex t-to-Last Disc (Disc 3) 8,360 3,170 (176%) 5.8 13.2 20.6 551 39.4 x 106 120º Segment of Disc 1 13,350 3,170 (176%) 9.55 15.08 20.60 634 83.3 x 106 120º Segment of Disc 2 12,100 3,170 (176%) 7.24 13.92 20.60 574 61.8 x 106 120º Segment of Disc 4 16,600 3,170 (176%) 11.70 15.70 19.70 595 91.3 x 106 1
The postulated turbine missile information in this table is for the removed General E lectric low pressure turbines as this analysis bounds other rotating elements.
The current missile analysis for the Unit 1 low pressure turbines is based on the missile probability analysis discussed in Section 1.4. 7.
2 The postulated turbine missile information in this table is for the removed B rown B overi low pressure turbine as these analyses are also included in structural design criteria shown in Table 5.1-1. The current missile analysis for the Unit 2 low pressure turbines is based on missile probability analysis discussed in Section 1.4.7.