LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Table 15.1-1 Through 15.1-6
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TABLE 15.1-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS Guaranteed Nuclear Steam Supply System thermal power output The Engineered Safety Features design rating {maximum calculated turbine rating) Thermal power generated by the reactor coolant pumps (nominal) Guaranteed Core Thermal Power 1 of 1 SGS-UFSAR 3471 MWt 3577 MWt 12 MWt 3459 MWt Revision 19 November 19, 2001 I I
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- TABLE 15.1-2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED Reactivity Coefficients Assumed Moderator Moderator Computer Temperature(1) Density(1) Faults Codes Utilized CONDITION II Uncontrolled RCC Assembly Bank TWINKLE, FACTRAN + 5 x 10-S Withdrawal from a Subcritical THING Condition Uncontrolled RCC Assembly Bank LOFTRAN 0 and 0.52 Withdrawal at Power RCC Assembly Misalignment THING, ANC, 0 LOFTRAN Uncontrolled Boron Dilution NA NA NA Partial Loss of Forced Reactor LOFTRAN 0 Coolant Flow THING, FACTRAN Start-up of an Inactive Reactor Coolant Loop Loss of External Electrical Load LOFT RAN 0 and 0.52 and/or Turbine Trip Loss of Normal Feedwater LOFT RAN NA 1 of 4 SGS-UFSAR Initial NSSS Thermal Power Output Assumed Dopoler{2) Lower Lower and upper Upper NA Upper Lower and Upper NA Revision 21 December 6, 2004 (MWt) 0 3431 {3) 2058.6 and 343.1 3411 (4} 0 and 3423 3431 3423 {DNB Cases) 3491.5 {Pressure Cases) (5) 3491.5 (5)
TABLE 15.1-2 (Cant)
- Reactivity Coefficients Assumed Moderator Computer Temperature( 1) Faults Codes Ulilized (1\krFJ CONDITION II {cont'd) Loss of Offsite Power to the LOFTRAN Plant Auxiliaries Excessive Heat Removal Due to LOFT RAN Feedwater System Malfunctions Excessive Load Increase LOFT RAN Accidental Depressurization of LOFTRAN the Reactor Coolant System Accidental Depressurization of LOFT RAN the Main Steam System
- Inadvertent Operation of ECCS LOFT RAN During Power Operation 2 of4
- SGS-UFSAR Moderator Density(1) (1\k/gm/cc} NA 0.52 0 and 0.52 0 Function of Moderator Density (See Sec 15.2.13) (Fig 15.2-41) 0 Initial NSSS Thermal Power Output Assumed Doppler{2} {MWtl NA 3491.5 {5} Lower 0 and 3411 {6) Lower 341 1{6} Upper 3411{6} Fig. 15.4-49 0 ( Subcritical) Lower 3491.5 (5} Revision 21 December 6, 2004 I I TABLE 15.1-2 (Cant) Reactivity Coefficients Assumed Moderator Moderator Computer Temperature( 1) Density(1) Faults Codes Utilized (t:\kfOF) (Ak/gm/cc} CONDITION Ill Loss of Reactor Coolant from NOTRUMP, SBLOCTA Small Ruptured Pipes or from Cracks in Large Pipe which Actuate Emergency Core Cooling Inadvertent Loading of a Fuel PHOENIX-P, ANC NA Assembly into an Improper Position Complete Loss of Forced Reactor LOFTRAN 0 Coolant Flow THINC, FACTRAN Waste Gas Decay Tank Rupture NA NA Single RCC Assembly Withdrawal ANC, THINC NA at Full Power PHOENIX-P CONDITION IV Major rupture of pipes containing SATAN Function of reactor coolant up to and BASH Moderator including double-ended rupture coco Density {See of the largest pipe in the LOCBART Section 15.4.1) Reactor Coolant System (Loss of Coolant Accident) 3 of4 SGS-UFSAR Initial NSSS Thermal Power Doppler{2) NA Upper NA NA Function of Fuel Temp. (See Section 15.4.1) Revision 25 October 26, 201 0 Output Assumed (MWt) 3479 3216-3563 (4) 3431 3577 3423 3579 TABLE 15.1-2 (Cant) Reactivity Coefficients Assumed Moderator Moderator Computer Temperature(1} Density(1) Faults Codes Utilized (AkJOFl (Ak/qm/ccl Doppler(2) CONDITION JV (cant) Major Secondary System Pipe LOFTRAN, THINC Function of Fig.15.449 Rupture, up to and Including Moderator Double-Ended Rupture (Rupture Density (See of a Steam Pipe} Section 15.2.13) (Fig. 15.4-50 Unit 1) (Fig. 15.4-48 Unit 2) Steam Generator Tube Rupture NA NA NA NA Single Reactor Coolant Pump LOFTRAN 0 Upper Locked Rotor and Reactor THINC, FACTRAN Coolant Pump Shaft Break Fuel Handling Accident NA NA NA Rupture of a Control Rod TWINKLE, FACTRAN -0 pcm/°F BOL Consistent Mechanism Housing (RCCA PHOENIX*P -26 pcm/oF EOL with lower Ejection) limit shown on Fig 15.1-5 NOTES: (1} Only one is used in an analysis, i.e., either moderator temperature or moderator density coefficient. (2) Reference Figure 15.1-5 for Doppler power coefficients. See UFSAR Section 4.5 for the applicable station reload analysis. (3) Cases are considered at 3 different initial power levels-100%, 60%, and 10%. (4) Core power is assumed in the analysis . . (5) Analysis is performed at 102% of an NSSS power of 3423 MWt which is equivalent to 100.6% of 3471 MWt. (6) No pump heat is assumed in the analysis. (7) Analysis is performed at 102% of a core power of 3411 MWt which rs equivalent to 100.6% of 3459 MWt. 4of4 SGS-UFSAR Revision 20 May6,2003 Initial NSSS Thermal Power Output Assumed {MWtl 0 (Subcritical) 3577 3431 3600 I 0 and 3479 (7)
'<, ( *: TABLE 15.1-3 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSIS Trip Function Power range high neutron flux, high setting Power range high neutron flux, low setting Overtemperature AT Overpower !J.T High pressurizer pressure Low pressurizer pressure Low reactor coolant flow (from loop flow detectors} Undervoltage trip Turbine trip Low-low steam generator level ...... ---High steam generator level of feedwater pumps and closure of feedwater system va:ves, and turbine trip Underfrequency trip Loss c: offsite power time de:..ay N87SS: Limiting Trip Point Assumed In Analyses 118 percent 35 percent Variable, see Figure 15.1-1 Variable, see Figure 15.1-1 2425 psig 1825 psig 87-percent loop flow 68 percent nominal Not Applicable 0 percent of Narrow Range Level Span 73 percent of Narrow Range Level Span 53.9 Hz Not Applicable Time Delay (sec) 0.5 0.5 -7. 0 { 1) (Ref. 21) { 2) 7 . 0 ( 1 ) ( Ref . 21 ) 2.0 2.0 1.0 1.5 1.0 2.0 2.0 0.6 1.5 (3) . ' ': i.me delay (including R':'D response time and trip circuit channel elec:ronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint at the channel sensor until the rods begin to drop. t See Reference 21, Section 15.1.10. E":-om rod motion 1 of 1 SGS-UFSAR Revision 18 April26, 2000 I TABLE 15.1-4 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT -POWER RANGE NEUTRON FLUX CHANNEL -BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRUMENTATION ERRORS Nominal Setpoint (percent of rated power) 109 Calorimetric Errors in the Measurement of Secondary System Thermal Power: Variable Feedwater temperature Feedwater pressure (small correction on enthalpy) Steam pressure (small correction on enthalpy) Feedwater flow Accuracy of Measurement of Variable (Percent Error) +0.5 +0.5 +2 +1.25 Estimated Effect on Thermal Power Determination (Percent Error) 0.3 1.25 Assumed calorimetric error (percent of rated power) 2 Axial power distribution effects on total ion chamber current Estimated error (percent of rated power) 3 Assumed error (percent of rated power) 5 Instrumentation channel drift and setpoint reproducibility Estimated error (percent of rated power) 1 Assumed error (percent of rated power) 2 Maximum overpower trip point assuming all individual errors are simultaneously in the most adverse direction (percent of rated power) 118 1 of 1 SGS-UFSAR Revision 6 February 15, 1987 TABLE 15 .. 1-5 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998 TABLE 15.1-6 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998