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Palisades Nuclear Plant - Chapter 14 UFSAR, Rev. 31, Section 14.15-1 - Figure 14-22-2, Revision 25
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FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-1 Revision 21 Page 1 of I INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER PARAMETER Initial core power level, MWt Core inlet coolant temperature, OF Core mass flow rate, 106 Ibm/hr Reactor coolant system pressure, psia Steam generator pressure, psia Initial pressurizer liquid volume, ft 3 Steam generator level, ft above tube sheet* Lower core flowrate dispositioned in Reference 6.ASSUMED VALUE 2600.6 550.65 138.*2,110 770 800 31.74 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-2 Revision 21 Paae 1 of I SETPOINTS FOR THE STEAM GENERATOR TUBE RUPTURE SETPOINTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Parameter Setpoint Steam generator MSSV setpoint, psia AFW actuation on steam generator level AFAS signal generation, % NR SIAS setpoint, psia Shutdown cooling entry conditions:

Hot leg temperature, OF Pressurizer pressure, psia 1000 23.7 1605 300 270 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-3 Revision 21 Page 1 of 2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Time Setpoint (second) Event or Value 1.0 Tube rupture occurs ----32.9 Proportional heaters are fully energized, psia 2085 105.7 Backup heaters are energized, psia 2035 211.3 Heaters are de-energized on low level in the pressurizer, ft 558.6 703.7 Pressurizer pressure reaches low pressurizer pressure setpoint (TM/LP floor), psia 1700.704.8 Trip signal is generated 705.2 Trip breakers open 706.1 Turbine Valves begin to close 707.1 Turbine valves are completely closed 708.2 Loss of offsite power 714.8 Feedwater flow begins ramping down at a rate of 5%/second 715.9 SIAS setpoint is reached, psia 1605 720.3 MSSVs begin to open, psia 1000 725.8 Pressurizer empties 733.9 Safety Injection pumps reach full speed 735.0 Upper head void begins to appear 811.5 Safety Injection flow to RCS begins, psia 1237.7 995.0 Maximum upper head void fraction 0.271 1107.0 Minimum PCS pressure, psia 1107.8 1370.5 Upper head void disappears FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-3 Revision 21 Page 2 of 2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Time Setpoint (second) Event or Value 1372.0 Pressurizer begins to refill 1466.6 Low steam generator level signal for Auxiliary feedwater actuation, ft 25.7 1586.6 Auxiliary feedwater reaches the steam generators, Ibm/sec/SG 27.0 1800.0 Operator takes action, opens ADVs to initiate cooldown 3000.0 Operator isolates the affected SG, below setpoint loop temperatures, OF 525.0 13000.0 Operator initiates steaming the affected generator to avoid overfilling, percent SG wide range span 90 23300.0 Shutdown Cooling entry condition is reached, PCS pressure, psia/temperature, OF 270/300 28800.0 PCS pressure and temperature demonstrated to be stabilized, transient terminated.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-4 Revision 21 Page 1 of I INTEGRATED PARAMETERS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Parameter 0-2 hr 0-8 hr Integrated primary to secondary leak, Ibm 183,202 605,101 Integrated Steam release, Ibm a. Through affected SG ADV 37,382 313,736 b. Through affected SG MSSV 44,654 44,654 c. Through intact SG ADV 185,000 719,448 d. Through intact SG MSSV 44,645 44,645 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-5 Revision 28 Page 1 of I STEAM GENERATOR TUBE RUPTURE (SGTR) RADIOLOGICAL ANALYSIS-INPUTS AND ASSUMPTIONS Input/Assumption Value Core Power Level 2703 MWth Initial PCS Equilibrium Activity 1.0 [tCi/gm DE 1-131 and 100/E-bar gross Initial __PCS _EquilibriumActivity_

activity Initial Secondary Side Equilibrium Iodine Activity 0.1 ILCi/gm DE 1-131 Maximum pre-accident spike iodine concentration 40 jtCi/gm DE 1-131 Maximum equilibrium iodine concentration 1.0 ,tCi/gm DE 1-131 Duration of accident-initiated spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdown cooling and terminate 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months steam release 529,706 Ib,, for pre-accident iodine spike PCS Mass case 459,445 Ibm for concurrent iodine spike case 141,065 Ibm per SG (minimum mass used to SG Secondary Side Mass maximize concentration from tube leakage)Integrated Mass Release Table 14.15-6 Secondary Coolant Iodine Activity prior to 0 1 DE 1-131 accident Faulted SG (flashed tube flow) -Table Steam Generator Secondary Side Partition 14.15-11 Coefficients Faulted SG (non-flashed tube flow)- 100 Intact SG -100 Break Flow Flash Fraction Table 14.15-7 Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Control Room Ventilation System Time of manual control room normal 20 minutes intake isolation and switch to emergency mode Breathing Rates Offsite RG 1.183, Section 4.1.3 Control Room RG 1.183, Section 4.2,6 Control Room Occupancy Factor RG 1.183 Section 4.2-6 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-6 Revision 28 Page I of I SGTR RADIOLOGICAL ANALYSIS -INTEGRATED MASS RELEASES (1)Time Break Flow in Steam Release from Ruptured SG Steam Release from (hours) Ruptured SG (Ibm Unaffected SG (lb,")) (ibm)0 0,190417 24,011,15 0 0 0.196417 -0.5 37,111.85 44,654 53,574 0.5 -1.388889 81,281 22.152.3 109,629.6 1.388889 2 40,798 15,229.7 75,370.4 2- 3 (38889 64,773 75,485.6 145,983.5 3,638889-8 357.126 200,868.4 388,464.5 8- 7201 0 0 0 I Ih I wr1tc assIiumcd to te constant within time period FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-7 Revision 28 Paiae I of I SGTR RiADIOLOGICAL ANALYSIS -FLASliNG FRACTION FOR FLOW FROM BROKEN ITUBE (seconds)

Flashing Fraction 0 (HI I(0 707 I 0.065 736 0.031 859 0.023 1090 0.006 1800 0.006 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-8 Revision 28 Page 1 of I SuTR RADIOLOGICAL ANALYSIS-40 iCI/GM D.E. 1-131 ACTIVITIES Isq)(opeActivity

________________________________(p~Ci/gni)

Iodine-131 33.2194 lodine-I1 32 7.6660)Iodine- 1 33 ,34.4971 Iodine- 134 3.002 5 IodifC- 135 14,6932 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-9 Revision 28 Page I of I SGTR RADIOLOGICAL ANALYSIS -IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS Input Assumption Value Maximum Letdown [low 40 gpm Assumed Letdown Flow

44 gpm at 120F., 2060 psia\McxiImm Idicnti licd PtCS Leakage 10 gp11 Maximum Unidentified PC'S Leakage I gpm I'CS Mass 459,445 lb,, 1-131 Decay Constant 5.986968F-5 min I- 132 Decay Constant 0.005023 min I- 133 Decay Constant 0.000555 min 1-134 Decay Constant 0.013178 min I- 135 Decay Constant 0.O01 748 min* maximuml letdown 11ow plus 10% uncerlainty FSAR CHAPTER 14- SAFETY ANALYSIS TABLE 14.15-10 Revision 28 Page 1 of I S(ITR RAI)IOIICAL ANALYSIS -CON( URRENT (335 X) IODINE SPIKE APPEARANCE RATE Isotope Appearance Rate Time of Depletion (Ci/min) (hours)Iodine-131 58.0966961 8 Iodine- 132 79.8319317 8 Iodine- 133 90.13 10904 , Iodine- 134 74.0318685 8 Iodine- I 35 68.9790622 8

FSAR CHAPTER 14- SAFETY ANALYSIS TABLE 14.15-11 Revision 28 Page I of I SGTR RADIOLOGICAL ANALYSIS -AFFECTED STEAM GENERATOR WATER LEVEL AND DECONTAMINATION FACTORS FOR FLASHED FLOW Time Water Level Above U-Tubes Calculated I)econtamination Factor (seconds) (feet) Decontamination Factor Used in Analysis 0 0.0 (assumed)*

.0 1.0 707.1 0.0 (assumed) 1.0 ).0 736 0.11 1.002299 1.002299 859 0.55 1.045037 1.045037 1090 1 39 1 452436 1.452436 1800 3.97 1.467378 1.467378 5000 6,79 60.03443 1.467378 7200 9A3 38.01867 1.467378 13100 12.34 553073.5 58.16008 28800 15.16 58.16008 58.16008 It is coinservat i ey assumed that no scrtubbing o nc curs un til after the reactor trip ait 707. I seconds. Since tile U-tUIbes remain covered thiroughout the event, it is also conservatively assLurn ed that at the time o( trip the water level is just above the top of the U-tubes. lie time-dependent water level after the trip is a f'unctioi of the alllowahie prinlary to secondary leakage, broken ube h lw, and MI SSV/AI)V releases from the al'tkctcd steam generator.

To minimiie the water level available for scrubbilng, tile Iocatinn ot thie tlbe break is assumed to be at tie top of the L1-tubes.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.16-1 Revision 23 Page 1 of I EVENT

SUMMARY

FOR THE EOC HZP CONTROL ROD EJECTION EVENT Ejection of a Single Control Rod Core Power Reached VHP Trip Setpoint Core Power Peaked Core Average Rod Surface Heat Flux Peaked Minimum DNBR Occurred Scram Rod Insertion Begins VALUE 36.86% RTP 1,903%RTP 101.9% RTP see Table 14.1-5 TIME (sec)0.0 0.309 0.410 0.507 0.507 1.409 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.16-2 Revision 28 Page 1 of 2 CONTROL ROD EJECTION RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value Core Power Level 2703 MWth Core Average Fuel Burnup 39,300 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Maximum Radial Peaking Factor 2.04% DNB Fuel 14.7%% Fuel Centerline Melt 0.5%LOCA Source Term Table 14.22-3 Initial PCS Equilibrium Activity 1.0 ý.LCi/gm DE 1-131 and 100/E-bar gross Initial __PCS _EquilibriumActivity_

activity Initial Secondary Side Equilibrium Iodine 0.1 ýiCi/gm DE 1-131 Activity Release From DNB Fuel Section 1 of Appendix H to RG 1.183 Release From Fuel Centerline Melt Fuel Section 1 of Appendix H to RG 1.183 Steam Generator Secondary Side Partition 100 Coefficient Steam Generator Tube Leakage 0.3 gpm per SG Time to establish shutdown cooling 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months PCS Mass 432,976.8 Ibm minimum -141,065 Ibm (per SG)SG Secondary Side Mass Minimum mass used for SGs to maximize steam release nuclide concentration.

Particulate

-95%Chemical Form of Iodine Released to Elemental

-45%ContanmentElemental

-4.85%Containment Organic -0.15%Particulate

-0%Chemical Form of Iodine Released from SGs Elemental

-97 %Organic -3%Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Time of Control Room Ventilation System 20 minutes Isolation Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.16-2 Revision 28 Page 2 of 2 CONTROL ROD EJECTION RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS InputlAssum ption Value Containment Volume 1.64E+06 ft 3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.10% (by weight)/day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.05% (by weight)/day Aerosols -0 1 hr1 Containment Natural Deposition Coefficients Elemental Iodine -1.3 hr-1 Organic Iodine -None FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.16-3 eilsion 28 Pa2e I of I CONTROL ROD EJECTION RADIOLOGICAL ANALYSIS -STEAM RELEASE SG Steam Release (lb.,)0 1lO0 Sec 107,158.8 I 100 sec 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 31,336.8 0.5 hr 8 hr 1,007,100 ,8 hr 0 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-1 Revision 28 Page 1 of 1 SAMPLED LBLOCA PARAMETERS Phenomenological Time in cycle (peaking factors, axial shape, rod properties and burnup)Break type (guillotine versus split)Break size Critical flow discharge coefficients (break)Decay heat Critical flow discharge coefficients (surgeline)

Initial upper head temperature Film boiling heat transfer Dispersed film boiling heat transfer Critical heat flux Tmin (intersection of film and transition boiling)Initial stored energy Downcomer hot wall effects Steam generator interfacial drag Condensation interphase heat transfer Metal-water reaction Plant 1 Offsite power availability Core power and power distribution Pressurizer pressure Pressurizer liquid level SIT pressure SIT liquid level SIT temperature (based on containment temperature)

Containment temperature Containment volume Initial flow rate Initial operating temperature Diesel start (for loss of offsite power only)1 Uncertainties for plant parameters are based on plant-specific values with the exception of "Offsite power availability," which is a binary result that is specified by the analysis methodology.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-2 Revision 28 Page 1 of 2 PLANT OPERATING RANGE SUPPORTED BY THE LOCA ANALYSIS Event Operating Range 1.0 Plant Physical Description

1. 1 Fuel a) Cladding outside diameter 0.417 in b) Cladding inside diameter 0.367 in c) Cladding thickness 0.025 in d) Pellet outside diameter 0.360 in e) Pellet density 96.0% of theoretical f) Active fuel length 132.6 in g) Resinter densification

[2%]h) Gd 2 0 3 concentrations 2, 4, 6 and 8 w/o 1.2 RCS a) Flow resistance Analysis considers plant-specific form and friction losses b) Pressurizer location Analysis assumes location giving most limiting PCT (broken loop)c) Hot assembly location Anywhere in core d) Hot assembly type 15x15 AREVA NP e) SG tube plugging 15%2.0 Plant Initial Operating Conditions

2. 1 Reactor Power a) Nominal reactor power 2,565.4 MWt b) LHR 15.28 kW/ft'c) F' 2.042 2.2 Fluid Conditions a) Loop flow 130 Mlbm/hr < M < 145 Mlbm/hr b) PCS inlet core temperature 537 T _ 544 'IF 3 c) Upper head temperature

< core outlet temperature d) Pressurizer pressure 2,010 K P <_ 2,100 psia 4 e) Pressurizer liquid level 46.25% s: L _ 67.8%f) SIT pressure 214.7 < P 239.7 psia g) SIT liquid volume 1, 040 < V 1,176 ft 3 h) SIT temperature 80 < T 140 OF (coupled to containment temperature) i) SIT resistance (fL/D) As-built piping configuration j) Minimum ECCS boron 1,720 ppm 1 Includes a 5% local LHR measurement uncertainty, a 3% engineering uncertainty and a 0.5925% thermal power measurement 2 uncertainty.

Includes a 4.25% measurement uncertainty.

3 Sampled range of +7 OF includes both operational tolerance and measurement uncertainty.

4 Based on representative plant values, including measurement uncertainty.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-2 Revision 28 Page 2 of 2 PLANT OPERATING RANGE SUPPORTED BY THE LOCA ANALYSIS Event Operating Range 3.0 Accident Boundary Conditions a) Br eak location Cold leg pump discharge piping b) Br eak type Double-ended guillotine or split c) Break size (each side, relative to CL 0.05 <_ A _ 0.5 full pipe area (split)pipe) 0.5 < A < 1.0 full pipe area (guillotine) d) Wo rst single-failure Loss of one ECCS pumped injection train e) Off site power On or Off f) LPSI flow Minimum flow g) HPSI flow Minimum flow h) ECCS pumped injection temperature 100 IF i) HPSI delay time 30 (w/ offsite power)40 seconds (w/o offsite power)j) LPSI delay time 30 (w/ offsite power)40 seconds (w/o offsite power)k) Co ntainment pressure 14.7 psia, nominal value I) Con tainment temperature 80 < T < 140 OF m) Containment spray/fan cooler delays 0/0 seconds FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-3 Revision 28 Page 1 of I STATISTICAL DISTRIBUTION USED FOR PROCESS PARAMETERS Operational Parameter Uncertainty Parameter Range Distribution Core Power Operation

(%) Uniform 1000 -100.5 Pressurizer Pressure (psia) Uniform 2,010 -2,100 Pressurizer Liquid Level (%) Uniform 46.25 -67.8 SIT Liquid Volume (ft 3) Uniform 1,040 -1,176 SIT Pressure (psia) Uniform 214 7 -239.7 Containment/SIT Temperature (0 F) Uniform 80 -140 Containment Volume' (xl0' ft 3) Uniform 1.64 -1.80 Initial Flow Rate (Mlbm/hr)

Uniform 130- 145 Initial Operating Temperature

(°F) Uniform 537-544 SIRWT Temperature (0 F) Point 100 Offsite Power Availability 2 Binary 0,1 Delay for Containment Sprays (s) Point 0 Delay for Containment Fan Coolers (s) Point 0 HPSI Delay (s) Point 30 (w/ offsite power)40 (w/o offsite power)LPSI Delay (s) Point 30 (w/ offsite power)40 (w/o offsite power)1 Uniform distribution for parameter with demonstrated PCT importance conservatively produces a wider variation of PCT results relative to a normal distribution.

Treatment consistent with approved RLBLOCA evaluation model 2 (Reference 5).No data are available to quantify the availability of offsite power. During normal operation, offsite power is available.

Since the loss of offsite power is typically more conservative (loss in coolant pump capacity), it is assumed that there is a 50 percent probability the offsite power is unavailable.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-4 Revision 28 Page 1 of I

SUMMARY

OF MAJOR PARAMETERS FOR THE LIMITING PCT CASE 6.0 % Gad Rod Core Average Burnup (EFPH) 7,381.22 Core Power (MWt) 2,572.79 Hot Rod LHR, kWlft 14.60 Total Hot Rod Radial Peak (F r) 2.040 Axial Shape Index (ASI) 0.1602 Break Type Guillotine Break Size (ft 2/side) 3.339 Offsite Power Availability Not Available Decay Heat Multiplier 1.01073 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-5 Revision 28 Page 1 of I

SUMMARY

OF HOT ROD LIMITING PCT RESULTS 15 x 15 AREVA NP Fuel Type w/o Gd 2 03 Case Number 22 PCT Temperature 1,740 OF Time 27.2 s Elevation 2.151 ft Metal-Water Reaction Oxidation Maximum Total Oxidation 0.59%< 0.01%

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-6 Revision 28 Page 1 of I CALCULATED EVENT TIMES FOR THE LIMITING PCT CASE Event Time (sec)Break Opened 0 PCP Trip 0 SIAS Issued 0.6 Start of Broken Loop SIT Injection 14.9 Start of Intact Loop SIT Injection (loops 1B, 2A and 2B, respectively) 17.1, 17.1 and 17.1 Beginning of Core Recovery (Beginning of Reflood) 27.2 PCT Occurred 27.2 Start of HPSI 40.6 LPSI Available 40.6 Broken Loop LPSI Delivery Began 40.6 Intact Loop LPSI Delivery Began (loops 1B, 2A and 2B, respectively) 40.6, 40.6 and 40.6 Broken Loop HPSI Delivery Began 40.6 Intact Loop HPSI Delivery Began (loops 1B, 2A and 2B, respectively) 40.6, 40.6, 40.6 Broken Loop SIT Emptied 50.7 Intact Loop SIT Emptied (loops 1B, 2A and 2B, respectively) 50.8, 54.6 and 53.1 Transient Calculation Terminated 300 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-7 Revision 28 Page 1 of I CONTAINMENT HEAT SINK DATA Total Material Heat Sink Surface Area (if 2) thickness (f)t)1 Containment Dome and Upper 0.0208 Carbon steel liner; no Wall 69,630.20 coatings 4.2625 Concrete; no coating Carbon steel liner; no 2 Containment Wainscot 0.0208 oating 2,20010 coating 4.2625 Concrete; no coatings 3 Containment Floor Slab 1.5 Concrete, no paint 7,567.80 0.0208 Carbon steel; no paint 15.971 Concrete; no paint 4 Containment Sump Slab 0.0156 Stainless steel 380.10 1.5 Concrete; no coating 0,0208 Carbon steel; no paint 28.3 Concrete; no coating 5 Reactor Cavity Slab 380.10 0.0208 Stainless steel 1.4792 Concrete; no coating 6 Lower Biological Shield 243.4 (Inner 0.015625 Stainless steel; no paint surface of cylindrical shape) 7,9167 Concrete; no coating Internal Concrete with Carbon 0.0208 Carbon steel Steel Liner Plate 2,048.40 1 3.8958 Concrete; no coating 8 Internal Concrete with Stainless 0.0417 Stainless steel Steel Liner Plate 4,712370 1 2,4083 Concrete; no coating Carbon steel liner; no Internal Concrete with Decking 2,672.90 0.004 coating 2.4833 Concrete; no coatings 10 Internal Concrete 62,870.90 1.708 Concrete; no coating 11 Gravel Pit 384,50 4.208 Concrete; no coating 12 Equipment Tanks and Heat 18,011.00 0.0364 Carbon steel; no paint Exchangers 13 Miscellaneous Equipment 18,344.80 0.0112 Carbon steel; no coating 14 Polar Crane 8,241.50 0.1258 Carbon steel; no coating 15 Ductwork plus Electrical Panels 31,127.50 0.0026 Carbon steel; no coating 16 Grating 16,812.20 0.00692 Carbon steel; no coating 17 quarter inch Structural Steel 35,812.90 0.0217 Carbon steel; no coating 18 half inch Structural Steel 48,705.20 0.0433 Carbon steel; no coating 19 Sump Strainer and Piping 3,750.00 0.00645 Stainless steel FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.1-8 Revision 28 Page 1 of I CONTAINMENT INITIAL AND BOUNDARY CONDITIONS Parameter Parameter Value Containment free volume range, 1.64E+06 to 1.80E+06 Initial relative humidity 100.0 %Initial compartment pressure, psia 14.7, nominal value Initial compartment temperature, 'F 80 _< T < 140 Containment spray time of delivery, sec 0.0 Containment spray flow rate, lb/sec 576.7 Containment spray temperature, OF 40.0 Fan cooler heat removal as a function of Temp Heat Removal temperature (OF) (BTU/sec)284 -196242.0 264 -157899.0 244 -118137.0 224 -82197.0 204 -53190.0 184 -32475.0 164 -19533.0 144 -11559.0 124 -6735.0 104 -3831.0 35 0.0 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.2-1 Revision 24 Page 1 of I SYSTEM PARAMETERS AND INITIAL CONDITIONS USED IN THE PALISADES SBLOCA ANALYSIS Palisades Parameter Analysis Value Primary Heat Output, Mwth 2580.6 Primary Coolant Flow, gpm 341,400 Operating Pressure, psia 2060 Inlet Coolant Temperature, OF 544 SIT Pressure, psia 215 SIT Fluid Temperature, OF 100 Steam Generator Tube Plugging, % 15 SG Secondary Pressure, psia 763 SG Main Feedwater Temperature, OF 439.5 SG Auxiliary Feedwater Temperature, OF 120 HPSI Fluid Temperature, OF 100 Reactor Scram Low Pressure Setpoint (TM/LP floor), psia 1585 Reactor Scram Delay Time on TM/LP, s 0.8 Scram CEA Holding Coil Release Delay Time, s 0.5 SIAS Activation Setpoint Pressure, psia 1450 HPSI Pump Delay Time on SIAS, s 40 Main Steam Safety Valve Setpoint Pressure, psia MSSV-1 1029.3 MSSV-2 1049.9 MSSV-3 1070.5 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.2-2 Revision 28 Page 1 of I PCT RESULTS OF THE PALISADES SBLOCA ANALYSIS Break Size (ft 2)POT (OF)0.04 0.05 0.06 0.08 0.10 0.15 1296 1451 1479 1734 1654 1356 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.2-3 Revision 28 Page 1 of 1 SEQUENCE OF EVENTS FOR THE PALISADES SBLOCA EVENT Event Time (s)Break in Cold Leg 2B opened 0.0 Pressurizer Pressure reached TM/LP setpoint 16.98 Reactor scram 18.28 Loss of off-site power 18. 28 MFW terminated

18. 28 Turbine tripped 18. 28 Pressurizer pressure reaches SIAS setpoint (1450 psia) 24.86 Minimum SG level reaches AFAS setpoint (23.7% span) 25 HPSI pump ready for delivery 6 4.86 Cold Leg pressure reaches HPSI shutoff head (1200.7 psia) 96 Motor-driven AFW delivery begins 14 5 Loop seal in Cold Leg 1 B cleared 282 Break uncovered 300 PCT occurs 1690 SIT discharge begins 1690 Reactor vessel mass inventory reaches minimum value 1698 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.2-4 Revision 28 Page 1 of 1 SBLOCA ANALYSIS CALCULATION RESULTS Peak Cladding Temperature Temperature (IF) 1734 Time (s) 1690 Elevation (ft) 10.2 Metal-Water Reaction Local Maximum (%) 2.0 Elevation of Local Maximum (ft) 10.2 Total Core Wide (%) <1.0 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.3-1 Revision 21 Page 1 of 2 MAXIMUM STRESSES, PRESSURES AND DEFLECTIONS IN CRITICAL REACTOR INTERNALS FOLLOWING A MAJOR LOSS OF COOLANT ACCIDENT Structural Component Core Barrel Failure Mode and Loading Condition Tension -Axial Load Buckling -External Pressure Tension -Internal Pressure Bending -Transverse Load Shear -Transverse Load Bending -Axial and Transverse Load Deformation

-Axial and Transverse Load Location of Failure Middle Section of Core Barrel Upper Portion of Core Barrel (Arch)Middle Section of Core Barrel Failure Condition(a) 54,000 psi Ap = 572 psi 54,000 psi 54,000 psi 32,400 psi 54,000 psi Allowable Condition(b) 29,300 psi Ap = 381 psi 29,300 psi 43,950 psi 17,580 psi 32,230 psi Calculated Condition 3,200 psi Ap = 380 psi 26,750 psi 22,510 psi 7,710 psi 70,310 psi Lower Core Support Beam Flange Junction of Flange to Web Lower End of Shroud Control Rod Shrouds 1 st Row (Near Nozzle)Center of Shroud Defl = 0.76" Defl = 0.51" Defl > 0.51" (a) The figures in this column represent the estimated stress, pressure or deflection limits at which the component will no longer perform its function.(b) The figures in this column represent the allowable stress, pressure or deflection limits in accordance with the design bases established in Chapter 3 of this FSAR.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.3-1 Revision 21 Page 2 of 2 MAXIMUM STRESSES.

PRESSURES AND DEFLECTIONS IN CRITICAL REACTOR INTERNALS FOLLOWING A MAJOR LOSS OF COOLANT ACCIDENT Structural Component Control Rod Shrouds 2nd Row Failure Mode and Loadinq Condition Bending -Axial and Transverse Load Deformation

-Axial and Transverse Load Bending -Transverse Load Bending -Axial Load Location of Failure Lower End of Shroud Center of Shroud Center of Beam Junction of Flange and Barrel Cylinder Failure Condition(a) 54,000 psi Defl = 0.76" 54,000 psi 54,000 psi Allowable Condition(b) 32,230 psi Defl = 0.51" 43,950 psi 43,950 psi Calculated Condition 28,090 psi Defl -0.279" 12,980 psi 40,630 psi Upper Grid Beam Upper Structure Flange (a) The figures in this column represent the estimated stress, pressure or deflection limits at which the component will no longer perform its function.(b) The figures in this column represent the allowable stress, pressure or deflection limits in accordance with the design bases established in Chapter 3 of this FSAR.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.17.3-2 Revision 30 Page 1 of I ASYMMETRIC LOADS ANALYSIS -REACTOR VESSEL INTERNAL COMPONENT STRESS MARGINS Component Core Support Barrel Lower Support Structure Upper Guide Structure Location Percent Margin (%)*Upper Flange Upper Cylinder Center Cylinder Support Columns Beams Core Support Plate Grid Beams 6 7 11 2 3 13 1*Percent margin is computed as (Saiow -S.Ic) (100%) / Sallow, where S,,,, is the calculated component stress and Sallow is the ASME Code allowable stress.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Page 1 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft 2)1. Containment Wall and Dome 69,630.2 Carboline 3912 Carbo Zinc 11 Carbon Steel Liner Air Gap Concrete 2. Containment Wainscot 2,200.2 Phenoline 305 Carbo Zinc 11 Carbon Steel Liner Air Gap Concrete 3. Containment Floor Slab 7,567.8 Phenoline 305 Carboline 195 Concrete Air Gap Carbon Steel Air Gap Concrete 4. Containment Sump Slab 380.1 Stainless Steel Air Gap Concrete Air Gap Carbon Steel Liner Air Gap Concrete 5. Reactor Cavity Slab (Note 1) 380.1 Stainless Steel Air Gap Concrete Air Gap Unibestos Stainless Steel FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Page 2 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKSlSOURCES HEAT SINK SURFACE AREA (ft 2)6. Lower Biological Shield (Note 2) 417.8 Stainless Steel Air Gap Concrete 7. Internal Concrete 61,337.5 Phenoline 305 Carboline 195 Concrete 8. Internal Concrete with Carbon Steel Liner Plate 2,048.4 Stainless Steel Wool Carbon Steel Air Gap Concrete 9. Internal Concrete with Stainless Steel Liner 4,712.7 Plate Stainless Steel Air Gap Concrete 10. Internal Concrete with Decking (Note 3) 2,672.9 Carbon Steel Air Gap Concrete Carboline 195 Carboline 305 11. Gravel Pit 375.1 Phenoline 305 Carboline 195 Concrete/Gravel Mixture 12. Structural Steel Adjacent to the Liner Plate 30,609.3 Carboline 3912 Carbo Zinc 11 Carbon Steel 13. Structural Steel 41,628.4 Carbo Zinc 11 Carbon Steel FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Page 3 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKSlSOURCES HEAT SINK SURFACE AREA (f t 2)14. Polar Crane 7,044 Carboline 3912 Carbo Zinc 11 Carbon Steel 15. Pressurizer Quench Tank (Note 4) 679 Carbon Steel Carbo Zinc 11 16. Safety Injection Tanks (Note 5) 4,098.4 Stainless Steel Carbon Steel Carbo Zinc 11 17. Clean Waste Receiver Tanks (Note 6) 9,255.6 Carbon Steel Carbo Zinc 18. Clean Waste Receiver Tank Skirts (Note 7) 3,577.2 Carbon Steel Carbo Zinc 11 19. Shield Cooling Surge Tank (Note 8) 112.2 Carbon Steel Carbo Zinc 11 20. Deleted 21. Letdown Heat Exchanger 101.8 Phenoline 305 Carbo Zinc 11 Carbon Steel 22. Shield Cooling Heat Exchanger 25 Carbo Zinc 11 Carbon Steel Water FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Paae 4 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK 23. Head Lift Rig and Containment Air Coolers Phenoline 305 Carbon Steel 24. Electrical Panels Carbo Zinc 11 Carbon Steel SURFACE AREA (ft 2)14,308.2 2,141.4 25. Refueling Stainless 26. Grating Carbon 27. Ductwork Carbon Mast and Grapple Steel Steel Steel 1,371.1 14,369.4 24,463.3 35,539.2 12,441.8 10,378.8 28. PCS Metal Wall #1 Reactor Vessel and Internals 29. PCS Metal Wall #2 Reactor Vessel and Internals 30. PCS Metal Wall #3 Reactor Core FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Page 5 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKSlSOURCES Notes: 1 The reactor cavity slab heat conductor is in contact with the containment atmosphere on both sides.2 The lower biological shield heat conductor is a tube. While the surface area specified above represents the outside surface area, only the inside surface area is in contact with the containment atmosphere.

3 The internal concrete with decking heat conductor is in contact with the containment atmosphere on both sides.4 The pressurizer quench tank heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

5 The safety injection tanks heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

6 The clean waste receiver tanks heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

7 The clean waste receiver tank skirts heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

8 The shield cooling surge tank heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-2 Revision 21 Paae 1 of I LOCA ANALYSIS ENGINEERED SAFEGUARDS EQUIPMENT ALIGNMENT D/G 1-2 Failure Equipment Operated D/G 1-1 Failure Equipment Operated Containment Sprays LPSI HPSI Containment Air Coolers Component Cooling Water Service Water P-54B & P54C P-67B P-66B P-52A & P-52C P-7B P-54A P-67A P-66A VHX-1, VHX-2 & VHX-3 P-52B P-7A & P-7C FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-3 Revision 25 Paqe 1 of I LOCA INITIAL CONDITIONS Containment Free Volume Containment Temperature Containment Pressure Relative Humidity SIRW Tank Temperature 1.64 x 106 ft 3 145 0 F 15.7 Psia 30%100OF FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-4 Revision 29 Page 1 of I CONTAINMENT BUILDING RESPONSE TO LOCA DOUBLE ENDED GUILLOTINE BREAK IN A HOT LEG Case DIG 1-2 Failure D/G 1-1 Failure Peak Pressure (Psi5)54.2 54.2 Time (Sec)13.2 13.2 The peaks for both cases are the same because they occurred so early in the transient that the differences in safeguards equipment used had not yet taken effect.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.1-5 Revision 29 Page 1 of I LOCA ANALYSIS PARAMETER ASSUMPTIONS Initial Containment Air Temp Initial Containment Pressure Relative Humidity CCWHX Tube Fouling Coefficient Service Water Temperature DIG 1-2 (RCF) Failure Data 145 0 F 15.7 Psia (1.0 Psig)30%.001 hr-ft 2-OF/BTU 85 0 F ECCS Injection Flow pre-RAS (1 HPSI, 1 LPSI pump)ECCS Injection Flow post-RAS (1 HPSI pump)1 SW Pump Flow Rate to CCWHXs 1 CCW Pump Flow Rate to SDCHXs 2 CS Pump Flow Rate to Containment (pre RAS)2 CS Pump Flow Rate to Containment (post RAS-HLI)Post-RAS Spillage after Initiation of Hot Leg Injection ECCS Injection Flow after Initiation of Hot Leg Injection 3,471 gpm 705 gpm 4,214 gpm 4,480 gpm 2,472 gpm 1,684 gpm 328 gpm 273 gpm D/G 1-1 (LCF) Failure Data ECCS Injection Flow pre-RAS (1 HPSI, 1 LPSI pump)ECCS Injection Flow post-RAS (1 HPSI pump)2 SW Pump Flow Rate to CCWHXs 2 SW Pump Flow Rate to 3 Containment Air Coolers 1 CCW Pump Flow Rate to SDCHXs 1 CS Pump Flow Rate to Containment (pre RAS)1 CS Pump Flow Rate to Containment, 1 header (pre RAS)1 CS Pump Flow Rate to Containment (post RAS-HLI)Post-RAS Spillage after Initiation of Hot Leg Injection ECCS Injection Flow after Initiation of Hot Leg Injection 3,443 gpm 703 gpm 4,286 gpm 1, 600 gpm/Air Cooler 4,480 gpm 1,781 gpm 1,233 gpm 788 gpm 308 gpm 279 gpm FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-1 Revision 29 Page 1 of I INITIAL CONDITIONS FOR THE MSLB CONTAINMENT ANALYSIS Parameter Containment Free Volume, ft 3 Initial Containment Temperature, OF Initial Containment Pressure, psig Initial Containment Humidity, %Containment Spray Water Temperature, OF Main Feedwater Regulating Valve Closure Time, sec Main Steam Isolation Valve Closure Time, sec Assumed Value 1.64 x 106 145.0 1.0*30 100.0 22 2* Zero power cases assumed 1.5 psig FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-2 Revision 23 Page 1 of I INITIAL CONDITIONS FOR THE MSLB CONTAINMENT ANALYSIS Power- and Case-Dependent Parameters for CONTRANS Code Power Power Cold Leg S/G Pressure PCS Flow Case % MWTh* Temp, 'F psia Rate, lbm/hr#102% 102 2600.6 550.65 770.0 144.6x10 6 75% 75 1917.5 548.70 784.0 144.6xl 0 6 0% 0 20.0 539.00 900.0 144.6x10 6 EEQ 102 2600.6 550.65 770.0 144.6x10 6 This power level includes an assumed contribution of 20 MWTh from the primary coolant pumps.# Lower PCS flowrate dispositioned in Reference

25.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-3 Revision 29 Page 1 of I MSLB CONTAINMENT ANALYSIS RESULTS Power Level Case Description Peak Pressure 53.5 Limiting Pressure -Relay 5P-7 Failure w/Open MSIV Bypass Valves 0%

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-4 Revision 27 Page 1 of 5 MSLB ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft 2)1. Containment Wall and Dome 69,630.2 Carboline 3912 Carbo Zinc 11 Carbon Steel Liner Air Gap Concrete 2. Containment Wainscot 2,200.2 Phenoline 305 Carbo Zinc 11 Carbon Steel Liner Air Gap Concrete 3. Containment Floor Slab 7,567.8 Phenoline 305 Carboline 195 Concrete Air Gap Carbon Steel Air Gap Concrete FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-4 Revision 27 Page 2 of 5 MSLB ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft 2)4. Containment Sump Slab 380.1 Stainless Steel Air Gap Concrete Air Gap Carbon Steel Liner Air Gap Concrete 5. Reactor Cavity Slab (Note 1) 380.1 Stainless Steel Air Gap Concrete Air Gap Unibestos Stainless Steel 6. Lower Biological Shield (Note 2) 417.8 Stainless Steel Air Gap Concrete 7. Internal Concrete 61,337.5 Phenoline 305 Carboline 195 Concrete 8. Internal Concrete with Carbon Steel Liner Plate 2,048.4 Stainless Steel Wool Carbon Steel Air Gap Concrete 9. Internal Concrete with Stainless Steel Liner Plate 4,712.7 Stainless Steel Air Gap Concrete FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-4 Revision 27 Page 3 of 5 MSLB ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft 2)10. Internal Concrete with Decking (Note 3) 2,672.9 Carbon Steel Air Gap Concrete Carboline 195 Carboline 305 11. Gravel Pit 375.1 Phenoline 305 Carboline 195 Concrete/Gravel Mixture 12. Structural Steel Adjacent to the Liner Plate 30,609.3 Carboline 3912 Carbo Zinc 11 Carbon Steel 13. Structural Steel 41,628.4 Carbo Zinc 11 Carbon Steel 14. Polar Crane 7,044 Carboline 3912 Carbo Zinc 11 Carbon Steel 15. Pressurizer Quench Tank (Note 4) 679 Carbon Steel Carbo Zinc 11 16. Safety Injection Tanks (Note 5) 4,098.4 Stainless Steel Carbon Steel Carbo Zinc 11 17. Clean Waste Receiver Tanks (Note 6) 9,255.6 Carbon Steel Carbo Zinc 18. Clean Waste Receiver Tank Skirts (Note 7) 3,577.2 Carbon Steel Carbo Zinc 11 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-4 Revision 27 Page 4 of 5 MSLB ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft 2)19. Shield Cooling Surge Tank (Note 8) 112.2 Carbon Steel Carbo Zinc 11 20. Deleted 21. Letdown Heat Exchanger 1.01.8 Phenoline 305 Carbo Zinc 11 Carbon Steel 22. Shield Cooling Heat Exchanger 25 Carbo Zinc 11 Carbon Steel Water 23. Head Lift Rig and Containment Air Coolers 14,308.2 Phenoline 305 Carbon Steel 24. Electrical Panels 2,141.4 Carbo Zinc 11 Carbon Steel 25. Refueling Mast and Grapple 1,371.1 Stainless Steel 26. Grating 14,369.4 Carbon Steel 27. Ductwork 24,463.3 Carbon Steel FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.2-4 Revision 27 Page 5 of 5 MSLB ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES Notes: 1 The reactor cavity slab heat conductor is in contact with the containment.

atmosphere on both sides.2 The lower biological shield heat conductor is a tube. While the surface area specified above represents the outside surface area, only the inside surface area is in contact with the containment atmosphere.

3 The internal concrete with decking heat conductor is in contact with the containment atmosphere on both sides.4 The pressurizer quench tank heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

5 The safety injection tanks heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

6 The clean waste receiver tanks heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

7 The clean waste receiver skirts heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

8 The shield cooling surge tank heat conductor is a tube. The surface area specified above represents the outside surface area, which is the carbo zinc 11 side and which is in contact with the containment atmosphere.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.3-1 Revision 28 Page 1 of 1 REACTOR CAVITY GEOMETRIC FACTORS Volume of Cavity 6,653 ft3 Volume of Sump 1,364 ft 3 Mass of Upper Seal 3,000 lb Refueling Pool Seal Breaks and Begins To Lift at 5.8 Psi Total Forward Loss Flow Area Coefficient (ft2) (ft2)Refueling Pool Seal Before Breaking Away 4.77 0.57 After Broken Away 82.23 1.42 Annulus Around Coolant Pipes 24.2 1.45 30-Inch Access Tube 4.75 2.37 6 Pipes Into Sump 10.1 1.19 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.3-2 Revision 28 Page 1 of I GEOMETRY AND PEAK PRESSURES IN STEAM GENERATOR COMPARTMENTS Steam Generator Compartment North South Volume 55,210 62,090 Vent Area (ft2)1,043.3 1,091.3 Peak Pressure (Psi)24.8 22.4 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.18.3-3 Revision 28 Page 1 of I DIFFERENTIAL PRESSURES AT VARIOUS LOCATIONS Calculated Design Pressure Pressure (Psi) (Psi)1 Maximum Uplift Differential Pressure Across the Reactor Cavity Floor for a 42-Inch Pipe Double-Ended Rupture Outside the Reactor Cavity 0.4 7.3 2. Maximum Differential Pressure Across the Primary Shield Walls Due To a Break of a 42-Inch Pipe Within the Reactor Cavity 52.4 72 3. Maximum Differential Pressure Across the Primary Shield Walls Due To a Break of a 30-Inch Pipe Within the Reactor Cavity 67.7 72 4. Maximum Differential Pressure Across Secondary Shield Walls of the North Steam Generator Compartment Due To a 42-Inch Pipe Double-Ended Rupture Within the Compartment 24.8 31 5. Maximum Differential Pressure Across the Secondary Shield Walls of the South Steam Generator Compartment Due To a 42-Inch Pipe Double-Ended Rupture Within the Compartment 22.4 27 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.19-1 Revision 28 Paqce 1 of 1 FUEL HANDLING ACCIDENT (FHA) RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value Core Power Level Before Shutdown 2703 MWth Core Average Fuel Burnup 39,300 MWD/MTU Discharged Fuel Assembly Burnup 39,300 -58,900 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Maximum Radial Peaking Factor 2.04 Number of Fuel Assemblies Damaged 1 fuel assembly Delay Before Spent Fuel Movement 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months FHA Source Term for a Single Table 14.19-2 Assembly Water Level Above Damaged Fuel 22.5 feet minimum Assembly Elemental

-252 Iodine Decontamination Factors Organic -1 Overall -183.07 Noble Gas Decontamination Factor 1 Elemental

-99.85%Chemical Form of Iodine In Pool Org an -0.15%Organic -0. 15%Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Time of Control Room Ventilation 20 minutes System Isolation Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 Elemental iodine -94%FHB Ventilation Filter Efficiencies Organic iodine -94%1 Noble gas -n/a FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.19-2 Revision 28 Paize 1 of 2 FUEL HANDLING ACCIDENT RADIOLOGICAL ANALYSIS -SOURCE TERM Nuclide Activity Nuclide Activity Nuclide Activity (Curies) I (Curies) (Curies)Co-58 Co-60 Kr-85 Kr-85m Kr-87 Kr-88 Rb-86 Sr-89 Sr-90 Sr-91 Sr-92 Y-90 Y-91 Y-92 Y-93 Zr-95 Zr-97 Nb-95 Mo-99 Tc-99m Ru- 103 Ru- 105 Ru- 106 Rh-105 Sb- 127 Sb- 129 Te- 127 Te- I 27m Te- 129 Te- I 29m Tc- 13 lm Te- 132 1-131 1-132 1-133 1-134 O.O000E+00 0.0000E+00

0. 1052E+05 0. 1 174E+03 0.16471--05

0. 18 19E404 0.7020E-+06 0.8456E+05 0.2679E+05 0.44531E+OI 0.8623E+05 0.9107E 106 0.3229E403 0.4137E+05 0.12101+07 0.1684E306 0.1248EF+07 0.8264E 06 0.7956E t06 0.12 16E+07 0.5426E+03 0.5771E106 0.39581' 406 0.6450E4+05 0.1176[E+03 0.73441E+05

0. 12221- 105 0.2383E405 0.3637E+05 0.3690F 105 0.6852E +06 0.64241+06 0.70601+06 0.30191E 06 0.208713-09 1-135 Xe-133 Xe-135 Cs-134 Cs- 136 Cs-137 Ba- 139 Ba- 140 La- 140 La- 141 La- 142 Ce- 141 Ce- 143 Co- 144 Pr- 143 Nd-147 Np-239 Pu-238 Pu-239 Pu-240 Pu-241 Am-24 I Cm-242 Cm-244 1-130 Kr-83m Xe-138 Xe-131in Xe-133m Xe- 135m Cs- 138 Cs-I 34m Rb-88 Rb-89 Sb- 124 Sb-125 0.8949E t04 0.1298E+07 0.8201 E405 0.2034E+06 0.5284F+05

0. 1001Of06 0.4861 E-04 0.11301+07 0.1235E +07 0.2730E+03 0.5794E-03 0.1 168E+07 0.4100E f 06 0.1014E +O07 0.1071E +07 0.42111-+06 0.1023E+08 0.4494E + 04 0.3578E1-03 0.5406E1 03 0. 1522F 06 0.18971E+03 0.5649E +05 0.1339E +05 0.254613+04 0.3727E1300 0.0000E-f 00 0.8276E +04 0.34031E+05

0. 1434E1 '04 0.0000E +00 0.5 122E 100 0.4804E 4 0I 0.00001E 100 0.1663 E + 04 0.1566k +05 Sb- 126 Te-131 Te-133 Te- 134 Te- I25m Te-133m Ba- 141 Ba- 137m Pd- 109 Rh- 106 Rh-103m Tc-101 Eu-I154 Eu-155 Eu-156 La- 143 Nb-97 Nb-95m Pm- 147 Pm-148 Pm- 149 Pm- 151 Pm- 148m Pr- 144 Pr- 144m Sm-153 Y-94 Y-95 Y-91m Br-82 Br-83 Br-84 Am-242 Np-238 Pu-243 0.9900E+03 0.8307E+04 0.2034E- 10 0.2217E-14 0.3417F+04 0.12 13E-09 0.OOOOE+00 0.1041 E+06 0.2825E+05 0.5771 E+06 0. 1097E+07 0.OOOOE+00 0.1246E+05 0.8442E+04 0.19351E 06 0.0000E+00 0.1692E+06 0.8748E104 0.1296E+06 0.1 659E+06 0.2481 E+06 0.5012E- 05 0.2899E+05 0.10 155i+07 0.1217E +05 0.2171E+06 0.0000E+00 0.OOOOE+00 0.1 702E+05 0.2060E+04 0.88331-01 0.0000E+00 0.1 138E+05 0.2238E+06 0.568 1E403 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-1 Revision 28 Page 1 of I MHA SEQUENCE OF EVENTS FOR THE DOSE CONSEQUENCE ANALYSIS Time (minutes)

Event/Action t = 0.0 Release of radionuclides to the containment atmosphere starts and the containment atmosphere begins leaking at the T.S. leak rate limit. Loss of Off-Site Power occurs. CHP and CHR signals are generated.

The control room is depressurized.

Control room inleakage occurs at the base infiltration rate.t = 1.0 Full spray flow is delivered to the containment atmosphere by the Containment Spray System. Removal of particulate and elemental iodine species begins at this time. No credit is taken for the removal of organic iodine species.t = 1.5 The control room is pressurized to > 1/8 " H20 and running in the E-HVAC mode with one train operational due to the loss of one safety train. Control room unfiltered inleakage past the normal intake isolation dampers and the smoke purge dampers begins.t = 19.0 The initial SIRWT inventory is depleted and containment spray suction is aligned to the containment sump. Leakage from ESF components and via the SIRWT begins. This assumes runout flows on 2 HPSIs, 2 LPSI's, 3 Containment Spray Pumps, minimum inventory of the SIRWT, and a containment backpressure of 55 psig.t = 150.9 The elemental iodine decontamination factor reaches 200 at this time.t = 203.1 The aerosol iodine decontamination factor reaches 50 at this time.t = 600.0 Containment spray flow is conservatively assumed to be terminated.

However SIRWT leakage is assumed to continue as if the CSS pumps continued to operate.t = 1440.0 The containment design leak rate is assumed to decrease to one-half (t = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) the T.S. leakrate.t = 43200 Low Population Zone (LPZ) doses are integrated over the interval (t= 30 days) from the initiation of the incident to 30 days. Site Boundary (SB) doses are integrated over the worst 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.Control Room doses are integrated over the interval (t = 30 days)from the initiation of the incident to 30 days.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-2 Revision 28 Paqe 1 of 3 MAXIMUM HYPOTHETICAL ACCIDENT I LOSS OF COOLANT ACCIDENT RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value Release Inputs: Core Power Level 2703 MWth Core Average Fuel Burnup 39,300 MWD/MTU Fuel Enrichment 3.0 -5.0 w/o Initial POS Equilibrium Activity 1.0 p.Ci/gm DE 1-131 and 100/E-bar gross Initial __PCSEquilibriumActivity_

activity Core Fission Product Inventory Table 14.22-3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.10% (by weight)/day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.05% (by weight)/day MHA release phase timing and duration Table 14.22-4 Core Inventory Release Fractions (gap RG 1.183, Sections 3.1, 3.2, and Table 2 release and early in-vessel damage phases)ECCS Systems Leakaqe (from 19 minutes to 30 days)39,054 ft.3 Sump Volume (minimum)0.053472 ft 3/min ECCS Leakage (2 times allowed value)Calculated

-0.03 to 0.06 Flashing Fraction Used for dose determination

-0.10 97% elemental, 3% organic Chemical form of the iodine released from the ECCS leakage 2 (current design basis)Iodine Decontamination Factor No credit taken for dilution or holdup FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-2 Revision 28 Page 2 of 3 MAXIMUM HYPOTHETICAL ACCIDENT I LOSS OF COOLANT ACCIDENT (MHA/LOCA)

RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value SIRWT Back-leakage (from 19 minutes to 30 days)Sump Volume ECCS Leakage to SIRWT (2 times allowed value)Flashing Fraction (elemental iodine assumed to be released into tank space based upon partition factor)SIRWT liquid/vapor elemental iodine partition factor Elemental Iodine fraction in SIRWT 292,143 gallons (minimum valve for ECCS leakage, maximizes sump iodine concentration) 430,708 gallons (maximum value for SIRWT backleakage to be consist with assumption of minimum water level in SIRWT)7.2 gpm until 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after RAS, then 0.0125 gpm 0% based on temperature of fluid reaching SIRWT Table 14.22-9 Table 14.22-8 Initial SIRWT Liquid Inventory (minimum at 4,144 gallons time of recirculation) 4,144_gallons Release from SIRWT Vapor Space Table 14.22-10 Removal Inputs: Containment Aerosol/Particulate Natural Deposition (only credited in unsprayed 0.1/hour regions)Containment Elemental Iodine Wall 2.3/hour Deposition Containment Spray Coverage >90%Spray Removal Rates: Elemental Iodine 4.8/hour Time to reach DF of 200 2.515 hours0.00596 days 0.143 hours 8.515212e-4 weeks 1.959575e-4 months Aerosol 1.8/hour (reduced to 0.18 at 3.385 hours0.00446 days 0.107 hours 6.365741e-4 weeks 1.464925e-4 months )Time to reach DF of 50 3.385 hours0.00446 days 0.107 hours 6.365741e-4 weeks 1.464925e-4 months FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-2 Revision 28 Page 3 of 3 MAXIMUM HYPOTHETICAL ACCIDENT I LOSS OF COOLANT ACCIDENT (MHA/LOCA)

RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value Spray Initiation Time 60 seconds (0.016667 hours)Control Room Ventilation System Table 14.24-1 Time of automatic control room 90 seconds isolation and switch to emergency mode Control Room Unfiltered Inleakage 16 cfm Transport Inputs: Containment Leakage Release Containment closest point ECCS Leakage Plant stack SIRWT Backleakage SIRWT vent Personnel Dose Conversion Inputs: Atmospheric Dispersion Factors Section 2.5.5.2 Offsite Tables 14.24-2 and 14.24-3 Onsite Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-3 Revision 28 Page 1 of 2 MHA/LOCA SOURCE TERM Nuclide Curies Nuclide Curies Co-58 Co-60 Kr-85 Kr-85m Kr-87 Kr-88 Rb-86 Sr-89 Sr-90 Sr-91 Sr-92 Y-90 Y-91 Y-92 Y-93 Zr-95 Zr-97 Nb-95 Mo-99 Tc-99m Ru- 103 Ru- 105 Ru- 106 Rh-105 Sb-i127 Sb- 129 Te- 127 Te- 127m Te- 129 Te- I29m Te-131m Te- 132 1-131 1-132 1-133 1-134 1-135 Xe- 133 0.OOOOE+00 O.0000E+00

0. 1052E 4 07 0.1948E+08 0.3756E+08 0.5286E+08 0.1959E+06 0.7213E+08 0.8458E4+07 0.8874E+08 0.9557E+08 0.8737E+07 0.9264E+08 0.9596E+08 0.1 101E+09 0.1236E+09 0.1206E+09 0.1249E+09 0.1368E+09

0. 1 198E4+09 0.1260E+09 0.9451+/-E+08 0.5794E+08 0.8741 E+08 0.9111 E+07 0.2568E+08 0.9047E+07 0.1223E+07 0.2528E+08 0.3772E+07 0.11 13E+08 0.10481E+09 0.7483E+08 0.10681E+09 0.14621E+09 0.1602E+09 0.1372E+09 0.1466E+09 Pu-239 Pu-240 Pu-241 Am-241 Cm-242 Cm-244 1-130 Kr-83m Xe-138 Xe-131m Xe- 133m Xe-135m Cs-138 Cs- 134m Rb-88 Rb-89 Sb- 124 Sb-125 Sb- 126 Te- 131 Te- 133 Te- 134 Te-125m Te-133m Ba- 141 Ba-137m Pd- 109 Rh-106 Rh- 103m Tc-101 Eu- 154 Eu- 155 Eu- 156 La- 143 Nb-97 Nb-95m Pr- 147 Pr- 148 0.3558E+05 0.5406E+05 0.1 522E+08 0. 1884E+05 0.5669E+07 0.5943E+06 0.3743E+07 0.9119E+07 0.1211 E+09 0.8346E+06 0.4659E+07 0.2999E+08 0.1340E+09 0.4920E+07 0.5369E+08 0.6895 E+08 0.1702E+06 0.1 567E+07 0.1107E+06 0.6601-E+08 0.8639E+08 0.1220E+09 0.3413 E+06 0.5406E+08 0.1188E+09 0.1043E+08 0.3327E+08 0.6285E+08 0.1135 E+09 0.1261 E+09 0.1247E+07 0.8448E+06 0.2023E+08 0.1 108 E+09 0.12 16F+09 0.8835E+06 0.1292E+08 0.2144E+08 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-3 Revision 28 Page 2 of 2 MHA/LOCA SOURCE TERM Nuclide Curies Nuclide Curies Xe-135 0.4692E+08 Pm- 149 0.4541E+08 Cs- 134 0.2037E+08 Pm- 151 0.1 606E+08 Cs-136 0.5873E+07 Pm-148m 0.2999E+07 Cs- 137 0.1 1001E+08 Pr- 144 0.1025E+09 Ba- 139 0.1307E+09 Pr-I 44m 0.1 224E 1-07 Ba- 140 0.1 260E+09 Sm- 153 0.4423E+08 La- 140 0.1299F+09 Y-94 0.1 105F+09 La- 141 0. 1193E+09 Y-95 0.1183E+09 La- 142 0. 1 156E+09 Y-91 m 0.5151E+08 Ce- 141 0.1212E+09 Br-82 0.5282E+06 Ce- 143 0.11 15E+09 Br-83 0.9102E+07 Ce- 144 0. 1020E+09 Br-84 0. 1591 E+08 Pr- 143 0.111 IE+09 Am-242 0.9062E4 07 Nd- 147 0.47701+08 Np-238 0.4306E+08 Np-239 0.1830E+ 10 Pu-243 0.4690E+08 Pu-238 0.3927E+06 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-4 ReJsion 28 Page 1of I MHAILOCA RELEASE PHASES Phase Onset Duration Gap Release 30 seconds 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Early In-Vessel 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 1.3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months s* From Regulatory Guide 1.183, Table 4 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-5 Revision 28 Paue 1 of I MHA/LOCA TIME DEPENDENT SIRWT PH Time (hous SIRWT pH (hours)0.3167 4.500 0.50 4.508 1.3167 4.544 1.3167 4.544 2.00 4.544 4.00 4.545 8.00 4.546 16.00 4.548 24.00 4.550 48.00 4.557 72.00 4.563 96.00 4.570 120.00 4.576 144.00 4.583 168.00 4.589 192.00 4.595 240.00 4.607 288.00 4.618 336.00 4.630 384.00 4.64 I 432.00 4.651 528.00 4.672 624.00 4.692 720.00 4.711 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-6 Revision 28 Page 1 of I MHA/LOCA TIME DEPENDENT SIRWT TOTAL IODINE CONCENTRATION Time SIRWT Iodine Concentration (hours) (gm-atom/liter) 0.3167 0.OOE+00 0.50 9.60E-07 1.3167 4.82E-06 1.3167 4.82E-06 2.00 4.84F-06 4.00 4.90E-06 8.00 5.02E-06 16.00 5.25E-06 24.00 5.48E-06 48.00 6.16E-06 72.00 6.82E-06 96.00 7.46E-06 120.00 8.08E-06 144.00 8.68E-06 168.00 9.26E-06 192.00 9,83E-06 240.00 1.09E-05 288.00 1.20E-05 336.00 1.29E-05 384.00 1.39E-05 432.00 1.48E-05 528.00 1.64E-05 624.00 1.79E-05 720.00 1.93E-05 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-7 Revision 28 Page 1 of I MHA/LOCA TIME DEPENDENT SIRWT LIQUID TEMPERATURE Time (hr) Temperature

("F)0.3167 100.0 0.50 100.0 1.3167 100.0 1.3167 100.0 2.00 100.0 4.00 100.5 8.00 101.3 16.00 102.4 24.00 103.2 48.00 104.7 72.00 105.0 96.00 105.0 120.00 104.9 144.00 104.8 168.00 104.8 192.00 104.7 240.00 104.6 288.00 104.6 336.00 104.5 384.00 104.5 432.00 104.5 528.00 104.4 624.00 104.4 720.00 104.4 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-8 Revision 28 Pai-e I of I MHA/LOCA Time Dependent SIRWT Elemental Iodine Fraction Time (hr) Elemental Iodine Fraction 0.3167 0.OOE+00 0.50 2.02E-02 1.3167 7.93E-02 1.3167 7.93E-02 2.00 7.95E-02 4.00 8.02E-02 8.00 8.16E-02 16.00 8.42E-02 24.00 8.68E-02 48.00 9.38E-02 72.00 1.OOE-0O 96.00 1.06E-0 I 120.00 1. I1 E-01 144.00 1.15E-01 168.00 I. 19E-0 1 192.00 1.23E-01 240.00 1.29E-0 1 288.00 1.34E-01 336.00 1.38E-01 384.00 1.41E-01 432.00 1.44E-01 528.00 1.47E-01 624.00 1.49E-0 I 720.00 1.49E-0 I FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-9 Revision 28 Page 1 of I MHA/LOCA TIME DEPENDENT SIRWT PARTITION COEFFICIENT Elemental Iodine Partition Time (hr) Coefficient 0.3167 45.65 0.50 45.65 1.3167 45.65 1.3167 45.65 2.00 45.65 4.00 45.21 8.00 44.53 16.00 43.61 24.00 42.95 48.00 41.74 72.00 41.50 96.00 41.50 120.00 41.58 144.00 41.66 168.00 41.66 192.00 41.74 240.00 41.82 288.00 41.82 336.00 41.89 384.00 41.89 432.00 41.89 528.00 41.97 624.00 41.97 720.00 41.97 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.22-10 Revision 28 Page 1 of I MHA/LOCA ADJUSTED RELEASE RATE FROM SIRWT Time Adjusted Iodine Release Rate (hours) (cfm)0.3167 9.1718E-04 1.3167 I. 1922E-05 8.00 1.2895E-05 24.00 1.4921 E-05 72.00 1.7737E-05 168.00 1.9907E-05 240.00 2.1376E-05 336.00 2.2501 E-05 432.00 2.3366E-05 624.00 2.3737E-05 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.23-1 Revision 28 Page 1 of 1 SMALL LINE BREAK OUTSIDE OF CONTAINMENT RADIOLOGICAL ANALYSIS -INPUTS AND ASSUMPTIONS Input/Assumption Value PCS Equilibrium Activity 1.0 DE 1-131 and 1 00/E-bar gross PCS__EquilibriumActivityactivity Break Flow Rate 160 gpm Break Temperature 135 0 F Break Pressure 35 psia Time required to isolate break 60 minutes Maximum equilibrium iodine concentration 1.0 pCi/gm DE 1-131 Iodine appearance rate for concurrent Table 14.23-2 iodine spike (500x)Iodine fraction released from break flow 10%Auxiliary building ventilation system filtration None Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Control Room Ventilation System Time of manual control room normal 20 minutes intake isolation and switch to emergency mode Breathing Rates Offsite RG 1.183 Section 4.1.3 Onsite RG 1.183 Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.23-2 Revision 28 Paze 1 of 1 SMALL LINE BREAK OUTSIDE OF CONTAINMENT RADIOLOGICAL ANALYSIS -CONCURRENT (500 X) IODINE SPIKE APPEARANCE RATE Appearance Rate Isotope (Ci/min)Iodine-131 86.7114868 Iodine-132 119.152137 Iodine- 133 134.524016 Iodine-134 110.495326 Iodine- 135 102.953824 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.24-1 Revision 28 Pane 1 of 3 TIME DEPENDENT CONTROL ROOM PARAMETERS (For TID-14844 based analyses.)

X/Q X/Q Containment Releases SIRWT Releases (Ventilation Stack/Aux Bldg)Time Breathing Occupancy Normal Emergency Normal Emergency Interval Rates Factors Intake Intake Intake Intake[m 3 Is] [sIm 3] [s/mi 3] [s/mi 3] [s/mi 3]0 -8 hr 3.470x10' 1.0 7.72x10-4 2.56xl 0-4 1.32x10-2 6.35xl 0-4 8 -24 hr 1.750x10" 4 1.0 4.55x10-4 1.51x104 7.78x10-3 3.74x10 4 1 -4 days 2.320x10' 0.6 2.90x10 4 9.60x10-5 4.95x10-3 2.38x10 4 4 -30 days 2.320x10-4 0.4 1.27x10-4 4.22x10-5 2.18x10_3 1.05x10-4 Atmospheric Dispersion Coefficient for Unfiltered Air Inleakage

= same as normal intake BOUNDING CR-HVAC FLOWS Emergency Mode Total Filtered Flow Emergency Mode Fresh Air Make-up Flow Emergency Mode Recirculation Flow Emergency Mode Unfiltered Inleakage Flow Normal Mode Fresh Air Make-up Flow Base Infiltration Leak Rate (Depressurized)

= 2827.2 cfm= 1413.6 cfm= 1413.6 cfm= 16 cfm (1)= 660.0 cfm= 384.2 cfm CR-HVAC FILTER EFFICIENCIES CR-HVAC Emergency Mode Charcoal Filter Efficiencies

= 99% for iodine and particulates

= 0% for noble gas (1) See specific events for actual Control Room envelope unfiltered inleakage assumed.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.24-1 Revision 28 Paqe 2 of 3 TIME DEPENDENT CONTROL ROOM PARAMETERS (For TID-14844 based analyses.)

ACCIDENT TIMING SCENARIOS Event Abbreviation FSAR SRP Accident Section Section Scenario t Cask Drop Accident SFCD 14.11 15.7.5 1 Main Steam Line Break MSLB 14.14 15.1.5 2 Steam Generator Tube Rupture SGTR 14.15 15.6.3 2 Control Rod Ejection CRE 14.16 15.4.8 3,2t Loss of Coolant Accident LOCA 14.17 15.6.5 3 Fuel Handling Accident FHA 14.19 15.7.4 1 Liquid Waste Incident LWI 14.20 15.7.2* 1 Gas Decay Tank Rupture GDTR 14.21.1 15.7.1* 1 Volume Control Tank Rupture VCTR 14.21.2 15.7.3 1 Small Line Break Outside Containment SLBOC 14.23 15.6.2 1 Maximum Hypothetical Accident MHA 14.22 15.6.5 3 t The four types of accident scenarios (1-4) are described below.1 The Control Rod Ejection has two release scenarios, an induced LOCA and a S/G-ADV release. The accident scenario type for these release scenarios are listed respectively, in the table above.* The section has been deleted from the Standard Review Plan, however, it remains part of the licensing basis for Palisades.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.24-1 Revision 28 Page 3 of 3 TIME DEPENDENT CONTROL ROOM PARAMETERS (For TID-14844 based analyses.)

The CR-HVAC flow mode, flow rates, and the time that these items change following accident initiation, are important parameters for determining control room radiological consequences.

The time to CR-HVAC emergency mode of operation is particularly important, and depends mainly on whether a Loss of Offsite Power (LOOP) occurs coincident with an accident and whether a Containment High Pressure (CHP) or Containment High Radiation signal (CHR) is generated at accident initiation.

Events that do not generate a CHP or CHR are collectively referred to as "Non-CHP/CHR Events;" whereas those that do, are referred to as "CHP/CHR Events." Four different accident scenarios result from the combination of these two items and encompass most FSAR Chapter 14 events: 1. Non-CHP/CHR Events Without a LOOP 2. Non-CHP/CHR Events With a LOOP 3. CHP/CHR Events With a LOOP 4. CHP/CHR Events Without a LOOP Note: No FSAR Chapter 14 events utilize scenario 4.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.24-2 Revision 28 Page I of I CONTROL ROOM ATMOSPHERIC DISPERSION (XIQ) FACTORS FOR AST ANALYSIS EVENTS Release 2 hr 2-8 hr 8-24 hr 1-4 days 4-30 days Receptor Release Point Receptor Point /QXIQ XQXQx Pair X/Q XQX A Containment Normal Intake 9.16E-03 7 17E-03 2 68E-03 2 07E-03 1.57E-03 Closest Point 'B'B Containment Emergency 7.26E-04 6 18E-04 2 47E-04 1 77E-04 1.30E-04 Closest Point Intake Normal Intake C SIRWT Vent N aIk 9.57E-02 7 59E-02 2 87E-02 2 19E-02 1.65E-02'B'D SIRWT Vent Emergency 9.66E-04 7 92E-04 3 13E-04 2 20E-04 1.64E-04 Intake E Plant Stack Normal Intake 5.29E-03"1:

.8 9E-03(" 15 1E-03{" .1 3E-03"I 8.41E-04("'B'F Plant Stack Emergency 8.32E-04 7 69E-04 2 83E-04 2 15E-04 157E-04 Intake G Closest ADV Normal Intake 9.95E-0311

.9 6E-03, .2 7E-03 2 13 9E-031 2 1 1.80E-031"'A'H Closest ADV Emergency 736E-04 6 42E-04 2 43E-04 1 75E-04 1 28E-04 Intake I Closest SSRV Normal Intake 1.24E-021

-- -'A'J Closest SSRV Emergency 796E-04 Intake K Containment Normal Intake 125E-02 9 83E-03 3 62E-03 2 86E-03 2.28E-03 Equipment Door 'B'L Containment Emergency 7.32E-04 6 13E-04 2 45E-04 1 75E-04 1.29E-04 Equipment Door Intake M Feedwater Area Normal Intake 2.20E-02 1 75E-02 7 1OE-03 5 24E-03 3.87E-03 Exhauster V-22A 'A'N Feedwater Area Emergency 8.65E-04 7 56E-04 2 81 E-04 2 04E-04 1.47E-04 Exhauster V-22A Intake (1) bounding XIQ values used for F IIA, SLBOC and SFCD (2) bounding X/Q values used for SGTR FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.24-3 Revision 28 Paue 1 of I RELEASE-RECEPTOR POINT PAIRS ASSUMED FOR AST ANALYSIS EVENTS Event MHA Normal Intake & Unfiltered Emergenc, Intake Inleakage Containment Leakage A B ECCS Leakage E F SIRWT Backleakage C D FHA Containment Release K L FHB Release E F SFCD Filtered Release E F Unfiltered Release K L MSLB Break Release M N MSSV/ADV Release G H I&G J&l-!SGTR Initial release %ia SSRVs Initial release via SSRVs switching to ADVs switching to ADVs CRE Containment Leakage A B I& G J&H Secondar%

Side Release Initial release via SSRVs Initial release via SSRVs switching to ADVs s itching to ADVs SLBOC E F FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.1-3 Revision 21 PALISADES SCRAM CURVE 1.0 0 q)0 z.75.5.25.0.0.5 1.0 1.5 2.0 2.5 Time (seconds)Note: Time measured from the point at which the control rod drive clutch receives the signal to release the control rods.3.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-1 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP REACTIVITY INSERTION CURVE 1. .4 o 2.0.0 10.0 20.0 80.0 40.0 50.0 60.0 Time (seconds)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-2 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP REACTIVITY FEEDBACKS.1-.0 0 S 4.-4 0.J-2-S 13 0 Dk-Doppler 0 0 Dk-ModeMtADr

..o I. °...I I D10 20.0 30.0 400 60.Time (seconds)60.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-3 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP TOTAL REACTIVITY

.9.75.8 0.45.3.15.0.0 10.0 2 30.0 40.0 50.0 Time (secnds)Total Reactivity 60.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-4 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP POWER AND HEAT FLUX 80n 60O'4 0 101 V N~ZW 0-MO.0 l1.0 20.0 30.0 40.0 5 Time (seconds)670.0 80.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-5 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP SYSTEM PRESSURE 2150.0 2100.0@U 2=2050.0 I .l * *

I I , .l ., .I , ,.0 100 2 s0 .4U 50 Time (seconds)0 70.0 80 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.1-6 Revision 21 CONTROL ROD WITHDRAWAL INCIDENT HZP INLET ENTHALPY 5425 5r.5 5m5.5m2.0 laD 2 30.0 40.0 50.0 rime (seconds)6000 80.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-1 Revision 21 REACTIVITIES FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER 4.0 2.0 0......................................................................

0> -2.0-4.0 TOT0L.-6.0 0S 10 15 20 25 30O3 TItC, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-2 Revision 21 REACTOR POWER LEVEL FOR UNCONTROLLED BANK WITHDRAWAL FULL POWER 400O 3500 3M0 ismo Iow 00 0 5 10 is 30 2 30 3 TI ME. SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-3 Revision 21 CORE AVERAGE HEAT FLUX FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER-- 20.0 18.0 I 16.0 14.0 S 12.0 ,-J Lai C.9 SE .0 6.0 4.0 0 10 Is 20 35 Tr ItE, SCC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-4 Revision 21 PRESSURIZER PRESSURE FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER 2600 24W 0 S 10 IS 20 25 30 35 TI ME, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-5 Revision 21 PRESSURIZER LIQUID LEVEL FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER 15 14 13-J 12..J I'10 5 10 is 20 25 30 35 TIME, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-6 IP,,-ic~inn "34 Rc~,iair~n

'74 PCS MASS FLOW RATE FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER x 50.0..- .-I 4S.0 CC)40.0 35.0 30.0 0 5 i0 is 20 TIME, SEC 25 i5 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-7 Revision 21 PCS TEMPERATURES FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER 625 sm 0 S 10 Is 20 TIME SIX 35 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-8 Revision 21 SECONDARY PRESSURE FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER STEAM GENERATOR al-44---STErM GENERATOR w2 I110!100 a, 0 S 18 is m TIIM, SC JU FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.2.2-9 Revision 21 S/G LIQUID LEVEL FOR UNCONTROLLED BANK WITHDRAWAL AT FULL POWER 0.0-2.0-4.0-6.0 STEJ11 GENERATOR

.*-)-- STERNI GENERATOR

@2-0.0-10.0 a 5 10 I5 20 T I ME, SEC 35 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.1-1 Revision 21 PRIMARY COOLANT SYSTEM MASS FLOW RATE FOR LOSS OF FORCED FLOW 39.0 34.0 ca EID C&3 L-.2 30.0*-'z 1-22.0 18.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 TItME, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.1-2 Revision 21 REACTOR POWER LEVEL FOR LOSS OF FORCED FLOW uw.0 u~w.o Ma:.0.0 0.0 1.0 3.0 3.0 4.0 5.0 9.0 7.0 TIME, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.1-3 Rgkuicir~n

)4 CORE AVERAGE HEAT FLUX FOR LOSS OF FORCED FLOW (-)c.'J F-.1.~t I-.~1 (I-CD tJ It.0 15.6 13.2 I -.-I --10.6 8.4 h n 0.0 1.0 2.0 3.0 4.0 TIME, SEC 5.0 6.0 7.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.1-4 Rouvicinn 91 PRESSURIZER PRESSURE FOR LOSS OF FORCED FLOW m m~m mm I 2a0.0 am.0 M I C 2m.0 IM.0.18MI.0 I 0.0 100.0 II A A a A 1.0 2.0 3.0 4.0 TIME, SC S.0 6.0 7.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.1-5 Revision 21 PRIMARY COOLANT SYSTEM TEMPERATURES FOR LOSS OF FORCED FLOW 6w.0 sm.0 I Mn.0 V5.0 mn................

PCs AVERAGE--COLD LED--- HOT LEG---------------

CORE INLET..........................

....... ...............

-----------------------

,5.0 5w.0 0.0 1.0 2.0 3.0 TRIfE, SEC 4.0 5.0 l.O 7.0 r-AK LCHAP I ER 14 -SAFETY ANALYSIS FIGURE 14.7.2-1 Revision 21 PRIMARY COOLANT SYSTEM MASS FLOW RATE FOR REACTOR COOLANT PUMP ROTOR SEIZURE lb I=*o " 0.0 LI am ----------------------1"6 .0 2A &@ CA0 GA6.7.TIME, 3M FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.2-2 Revision 21 REACTOR POWER LEVEL FOR REACTOR COOLANT PUMP ROTOR SEIZURE 3r500.0 m .-w=! J --w -v- v-30M0.0 2m.0 15ZW.0 Im.O 0.0 0.0 1.0 3.0 3.0 4.0 5.0 4.0 7.0 TMSEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.2-3 Revision 21 CORE AVERAGE HEAT FLUX FOR REACTOR COOLANT PUMP ROTOR SEIZURE N , 12. .0 50.0 0.0 I.0 2.0 3.0 4.0 5.0 6.0 7.0 T I W, SEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.2-4 Revision 21 PRESSURIZER PRESSURE FOR REACTOR COOLANT PUMP ROTOR SEIZURE mo.0 luo.0 am.. .e1 .0 3.0 4. .1. .TIME, sEC FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.7.2-5 Revision 21 PRIMARY COOLANT SYSTEM TEMPERATURES FOR REACTOR COOLANT PUMP ROTOR SEIZURE=.a 6.0~I a mS.o W05.0..............

... R..WW__COLD Us----CORE IMML;.. ...... .*... ...... .........

..... ..... °o... °* ........ .........°..........

° .........

--q.°....o...........

°..°......

- ooo..5w5.0 5m.O 0.0 1.0 2.0 3.0 4.0 TIME, SEC S.o 6.0 7.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.10-1 Revision 25 POWER COMPARISONS

-EXCESS LOAD 130 Actual Reactor Power Decalibrated NI Power....- Decalibrated Thermal Po er 120 --VHPT setpoint 100 90 0 5 10 15 20 25 30 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.10-2 Revision 25 PCS COOLANT TEMPERATURE

-EXCESS LOAD 620 610 600 590 Y- 580 c2 570 a)E E 560 I-550 540 530 520___ Thot------.----

T cold..................0 5 10 15 Time (s)20 25 30 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.10-3 Revision 25 PRESSURIZER PRESSURE -EXCESS LOAD 2200 2100 2000 C'1900 1800 L 0 5 10 15 Time (s)20 25 30 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.10-4 Revision 25 PRESSURIZER COLLAPSED LIQUID LEVEL -EXCESS LOAD 70 60 CL UI).-0 50 40 ý30 -20 0 5 10 15 Time (s)20 25 30 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.10-5 Revision 25 COMPONENTS OF REACTIVITY

-EXCESS LOAD a)0.5 0.4 0.3 0.2 0.1 0.0-0.1-0.2-0.3-0.4-0.5 0 5 10 15 20 25 Time (s)30 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.11-1 Revision 23 PARTIAL OPERATING FLOOR PLAN EL 649'-0" Withheld under 10 CFR 2.390 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.12-1 Revision 23 REACTOR POWER LEVEL FOR LOSS OF EXTERNAL LOAD EVENT 0 a_3000.0 2500.0 2000.0 1500.0 1000.0 500.0..0.0 2.5 5.0 7.5 10.0: Time (sec)12.5 : 1:5.0 17.5 20.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.12-2 Revision 23 PRIMARY PRESSURES FOR LOSS OF EXTERNAL LOAD EVENT 2800.0 2600.0 C-2400.0 2200.0 2000.0.0 2.5 5.0 7.5 10.0 12.5 Time (sec)15.0 17.5 20.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.12-3 Revision 23 PRESSURIZER LIQUID VOLUME FOR LOSS OF EXTERNAL LOAD EVENT 080.. .. .. ..................

'..... .........

.. .......* *", .. "

...........

..........

"... ....... ".... -........ "S1040.01.020.0.-1000.0 980.0.0. 2.5 .. 7.5: 1o0.61 .12,5 1 5.0 1. 5 .20:0..: " '.. "" " T im e _ :." : : :i ""

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.12-4 Revision 23 PRIMARY COOLANT SYSTEM TEMPERATURES FOR LOSS OF EXTERNAL LOAD EVENT FIGURE 14.12-4 Revision 23 u)0l..E)cu 620,0 600.0 580.0 560.0 540;0 520.0"--,--a Cold Leg 0-----o Averageo I--' Hot Leg 111111111111

..............o 2.5 5.0 7.5 10o0 Time (sec)12.5 15,0 17.5 20.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.12-5 Revision 23 SECONDARY PRESSURES FOR LOSS OF EXTERNAL LOAD EVENT.1 00 ,0:, ;. .,'.1000.0 0 900.0'I)' 800.0 700.0 o-o Steam Generator 1*-' Steam Generator 2.0 2.5 5.0 7.5 10.0 12.5 15.0 175 20.0 Time: (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-1 Revision 24 Reactor Power, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 120.0 100.0 I I I I I I I I H 0 0~0 I)Uy 80.0 60.0 40.0 20.0.0 111111 I I lull,, I

I I I I 6000.0 7000.0 8000.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-2 Revision 24 Primary Coolant System Loop Temperatures, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled cin 0 Fy 600.0 590.0 580.0 570.0 560.0 550.0 540.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-3 Revision 24 Primary Coolant System Loop Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled E 0 0 0-J 9000.0 7500.0 6000.0 4500.0 3000.0 1500.0 I I I I I I I I I I I I I I I I I S ' I I I I I I I-Loop 1A o-o Loop 2B-Loop 1A o-e Loop 2B I I i I I I 0 hi~~ i I.0 I I I I I I I I I I I.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 Time (sec)8000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-4 Revision 24 Pressurizer Pressure, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 2600.0 C U)o 2400.0 U)U)2200.0 0_N U)u) 2000.0 1800.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-5 Revision 24 Cl)E 0 cn C-U)N n3 Pressurizer Spray Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 6 0 .0 , I I I I I I I I I I I I I-I I I 40.0 20.0.0 ,-20.0 II.0 1000.0 2000.0 3000.0 4000.0 Time (sec 5000.0 6000.0 7000.0 8000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-6 Revision 24 Pressurizer SRV Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 15.0 O9 c/)E U-Ln U)N:3 U1)U_10.0 5.0.0-5.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-7 Revision 24 Pressurizer Level, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 70.0 U)U)-J U)N:3 C/)U)U)0~60.0 50.0 40.0 30.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-8 Revision 24 SG Auxiliary Feedwater Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 30.0 E 0-4 Q)25.0 20.0 15.0 10.0 5.0.0 I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 1, I11,1S L L 1 1 I I I I I I I I I I o-o o-S SG-21 i i i i i i i i i i i i i i i i i i i i i i i i i.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-9 Revision 24 SG Dome Pressure, LNFF Analysis with Off-Site Power Available and Steam "Dump System Disabled 1100.0 0D U)0~0)c-a)CD Ei 0~ai)4-5 cf)1000.0 900.0 800.0 o-o SG-2 700.0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-10 Revision 24 SG Liquid Mass Inventory, LNFF Analysis with Off-Site Power Available and Steam Dump System Disabled 0 0~-4 Q)CD a)150000 125000 100000.75000 50000 25000 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 7000.0 8000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-11 Revision 24 Reactor Power, LNFF Analysis with Off-Site Power Available and Steam Dump System Available Q)0 0L 0 (-)oy 120.0 100.0 80.0 60.0 40.0 20.0.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTEP 1A -QACCETV A LEK... ...-I-, 1010FIGURE 14.13-12 Revision 24 Primary Coolant System Loop Temperatures, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 580.0 ..-Vessel Outlet o-o Average SVessel Inlet L§ 560.0 a)Q) 540.0 E a.)(I-(9 r: 520.0 5 0 0 .0 , ....., .....t.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-13 Revision 24 Primary Coolant Svstem Loop Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Available U)E-Q 0 0-L U)C-)9000.0 7500.0 6000.0 4500.0 3000.0 1500.0I-I Q--o Loop 1A-Loop 2B-Loop 1A'--o Loop 2B.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-14 Revision 24 Pressurizer Pressure, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 0)U)Q)U)-N U)U)aD 3000.0 2750.0 2500.0 2250.0 2000.0 1750.0 1500.0..............I ...I ............................-.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-15 Revision 24 Pressurizer Spray Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 60.0 E 0 cf)N ci)L.40.0 20.0.0 111111 1111111111111111111111111111111111111111111

-20.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-16 Revision 24 Pressurizer SRV Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 15.0 U)E-o 0 V.)N U)0 0_10.0 5.0.0 I I I I H---- RV-1039-oRV-1040RV-1041 A p 0 A 0 p 0-5.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-17 Revision 24 Pressurizer Level, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 70.0-J q)N C')(I)0 L.0~60.0 50.0 40.0 30.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-18 Revision 24 SG Auxiliary Feedwater Flow, LNFF Analysis with Off-Site Power Available and Steam Dump System Available (J)E 0 U-a-, ci, U_°U-j_30.0 25.0 20.0 15.0 10.0 5.0 a0--o SG-2 0-3SG-2.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-19 Revision 24 SG Dome Pressure, LNFF Analysis with Off-Site Power Available and Steam Dump System Available 1100.0 U)U)U)L-0 I)E 1000.0 900.0 800.0 700.0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-20 Revision 24 SG Liquid Mass Inventory, LNFF Analysis with Off-Site Power Available and Steam Dump System Available U)0~-J 0 (-4 E c/)150000 125000 100000 75000 50000 25000 0.0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 3500.0 4000.0 4500.0 5000.0 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-21 Revision 24 Reactor Power, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 120.0 100.0 I ...I .I .I ...I 0 0~Q)080.0 60.0 40.0 20.0.0.0 2000.0 4000.0 6000.0 Time (sec)8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-22 Revision 24 U-u)U-r, Primary Coolant System Loop Temperatures, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 600.0 I , , , I , , I-Vessel Outl(o-o Average 590.0 Vessel Inlet 580.0 570.0 560.0 550.0 540.0 i I I I I I.0 2000.0 4000.0 6000.0 8000.0 Time (sec)10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-23 Revision 24 Primary Coolant System Loop Flow, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 9000.0 U)E 0 C-0J 0 b_c_0 r)7500.0 6000.0 4500.0 3000.0 1500.0 Loop 1A o-o Loop 2B A -Loop 1A 0 -o Loop 2B i I I I I I I I I II I I I.0.0 2000.0 4000.0 6000.0 Time (sec)8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-24 Revision 24 Pressurizer Pressure, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 2600.0 U)U)Q)N U)U)n)2400.0 2200.0 2000.0 1800.0.0 2000.0 4000.0 6000.0 Time (sec)8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-25 Revision 24 Pressurizer Level, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 70.0 0)N C/)Cl)n)60.0 50.0 40.0 I I I I I I I I I I I I I I I I I I I 30.0.0 2000.0 4000.0 6000.0 Time (sec).8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-26 Revision 24 SG Auxiliary Feedwater Flow, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 30.0 I I I I I I (Ti E 0 x 25.0 20.0 15.0 10.0 5.0.0 0 3 ----0 0 E) 0 1- E) 0 o-a SeG-I1 o--o SG-2 I I I I , I , I i I i I , i.0 2000.0 4000.0 6000.0 Time (sec)8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-27 Revision 24 SG Dome Pressure.

LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 1100.0 I I I I I I I I I I I I I I I U)U)U)(D 0 CD E 4-U, 1000.0 900.0 800.0 SG-1 SG-2 I , , I , i 700.0 I i , I , , I ,.0 2000.0 4000.0 6000.0 Time (sec)8000.0 10000.0 12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.13-28 Revision 24 SG Liquid Mass Inventory, LNFF Analysis without Off-Site Power Available and Steam Dump Systems Disabled 150000 Ea U)U)0)40 a-125000 100000 75000 50000 25000 0.0 2000.0 4000.0 6000.0 8000.0 10000.0 Time (sec)12000.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-1 Revision 24 Break Flow Rates During LHR-Limitinq Transient 6000 5000 C.)E cc 0 U-e Cu CIO 4000 3000 2000 1000 0 0 50 100 150 200 250 300 350 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-2 Revision 24 Steam Generator Pressures During LHR-Limiting Transient cu.F a-U)cL u, c: I.-a.-0U E U, 1100 1000 900 800 700 600 500 400 300 200 100 0 0 50 100 150 200 250 300 Time (sec)350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-3 Revision 24 Steam Generator Heat Transfer Rates During LHR-Limitinq Transient cu I-Cu c-a)2500 2000 1500 1000 500 0 0 50 100 150 200 250 300 350 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-4 Revision 24 Steam Generator Secondary-Side Total Fluid Inventories During LHR-Limiting Transient 250000 E 2.0 0 U)0 U)M a)c E 0 U)200000 150000 100000 50000 0 0 50 100 150 200 250 300 Time (sec)350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-5 Revision 24 Core Inlet Temperatures During LHR-Limiting Transient 600 ,--o Unaffected Sector--A Affected Sector 550 500 -e -e-i)Cu 450 E I--, 400 0 o 350 300 250 ....0 50 100 150 200 250 300 350 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-6 Revision 24 Core Inlet Flow Rates Durinq LHR-Limitinci Transient 20000 I, 9---- -c- --e -- & ---e (3'- ---e- -& -- --e - ---0-- --ý(15000 E a)10000 0 a)0 5000 A A A. A---- Stuck Rod Region Rest of Affected Sector-e Unaffected Sector-13 ---& --El ----B ---Eý- ---EJ- --El- --0 ---a --El ---9 ---L3-- --4D- --El- --G 0 0 50 100 150 200 Time (sec)250 300 350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-7 Revision 24 Pressurizer Pressure During LHR-Limiting Transient I.-U, U, a)N=3 cl)U)2200 2000 1800 1600 1400 1200 1000 800 600 0 50 100 150 200 250 300 350 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-8 Revision 24 Pressurizer Liquid Level Durinq LHR-Limitinq Transient cu C,)ci)-J C,)a,):3 L..CA 70 60 50 40 30 20 10 0 0 50 100 150 200 250 300 350 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-9 Revision 24 Total HPSI Flow Rate Durinl LHR-Limitinq Transient 50 40ý2 0 U)0/U-0m I"1-30 20 10 0 0 50 100 150 200 250 300 Time (sec)350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-10 Revision 24 Reactivity During LHR-Limiting Transient 10 5 C.)cu 0-5-10L 0 50 100 150 200 250 300 Time (sec)350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-11 Revision 24 Reactor Power During LHR-Limiting Transient 30 25 a)4--0-0 0~CL.0.4-cu 20 15 10 5 0 0 50 100 150 200 250 300 Time (sec)350 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-12 Revision 24 Break Flow Rates During DNBR-Limiting Transient 6000 5000 E.0 cu~U-U)4000 3000 2000 1000 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-13 Revision 24 Steam Generator Pressures During DNBR-Limiting Transient cu 0n U)1100 1000 900 800 700 600 500 400 300 200 100 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-14 Revision 24 Steam Generator Heat Transfer Rates During DNBR-Limiting Transient 2500 cu-0-cu I-a)"-E2 a)U)2000 1500 1000 500 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-15 Revision 24 Steam Generator Secondary-Side Total Fluid Inventories During DNBR-Limiting Transient 250000 E 0 4D-a 0 (D U)0 U)au CU 200000 150000 100000 50000 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-16 Revision 24 Core Inlet Temperatures During DNBR-Limiting Transient 600 550 u-0 Q-E a)0 0 500 450 400 350 300 250 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-17 Revision 24 Core Inlet Flow Rates During DNBR-Limitinq Transient 20000 cj)E.0 0 0 15000 10000 5000 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-18 Revision 24 Pressurizer Pressure During DNBR-Limitinq Transient 2200 ....2000 1800 CD 1600 a)(I)a,)0 1400 N n 1200 a)1000 800 600 0 100 200 300 400 500 600 700 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-19 Revision 24 Pressurizer Liquid Level During DNBR-Limiting Transient CL (U 4--0n (U 70 60 50 40 30 20 10 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-20 Revision 24 Total HPSI Flow Rate During DNBR-Limitinq Transient 50 40 cl)E Cu 0 LL U)-30 20 10 0 L 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-21 Revision 24 Reactivity During DNBR-Limiting Transient 10 5 C.)0-5-10 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.14-22 Revision 24 Reactor Power During DNBR-Limiting Transient 30 25'4--0 0-0~a-)wV 20 15 10 5 0 0 100 200 300 400 500 600 Time (sec)700 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-1 Revision 21 SGTR WITH LOAC: CORE POWER vs TIME FIGURE 14.15-1 Revision 21 ill t20 100-80-60 40 20 300 600 900 1200 1500 1800-rME, 5ECON0S FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-2 Revision 21 SGTR WITH LOAC: CORE COOLANT TEMPERATURE vs TIME FIGURE 14.15-2 Revision 21 660 6~I I b.I.'U, I I 800 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-3 Revision 21 SGTR WITH LOAC: PRIMARY COOLANT SYSTEM PRESSURE vs TIME FIGURE 14.15-3 Revision 21 2200 i 2000 1800 1600 1400 1200 1000 TIME. SECONOS FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-4 Revision 21 SGTR WITH LOAC: STEAM GENERATOR PRESSURE vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-5 Revision 21 SGTR WITH LOAC: TUBE LEAK FLOW RATE vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-6 Revision 2.1 SGTR WITH LOAC: INTEGRATED TUBE LEAK FLOW vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-7 Revision 21 SGTR WITH LOAC: PRESSURIZER LIQUID VOLUME vs TIME 900 750 600 I a-I N N 450 300 Lso FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-8 Revision 21 SGTR WITH LOAC: AFFECTED STEAM GENERATOR SAFETY VALVE (MSSV)FLOW RATE vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-9 Revision 21 SGTR WITH LOAC: AFFECTED STEAM GENERATOR SAFETY VALVE (MSSV)INTEGRATED FLOW vs TIME 60000 50000 -4 30000-20000 t 0000-300 I I I 300 600 900 1200 L500 1800 T (ME -'ECONOS FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-10 Revision 21 SGTR WITH LOAC: STEAM GENERATORS LIQUID MASS vs TIME 360000 300000 240000-AFFECTED SO L20000 UNAFFECTED SG 60000 300 600 900 TIME. SECONDS 1200 Lc-rjc 1800 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-11 Revision 21 SGTR WITH LOAC: CORE POWER vs TIME-11 1 I i I I I 100 S80 I 60-Io N 40-20 5000 10000 15000 20000 25000 30000 TIME. SECONOS FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-12 Revision 21 SGTR WITH LOAC: CORE COOLANT TEMPERATURES vs TIME FIGURE 14.15-12 Revision 21 U.'la FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-13 Revision 21 SGTR WITH LOAC: PRIMARY COOLANT SYSTEM PRESSURE vs TIME I 2500 2000 1500 t1000~1~ I I I *I__ I 1 -t 50o 5000 10000 L5000 20000 TIME. SECONOS 25000 30000 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-14 Revision 21 SGTR WITH LOAC: STEAM GENERATORS PRESSURE vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-15 Revision 21 SGTR WITH LOAC: PRESSURIZER LIQUID VOLUME vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-16 Revision 21 SGTR WITH LOAC: TUBE LEAK FLOW RATE vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-17 Revision 21 SGTR INTEGRATED LEAK FLOW vs TIME W.-uI Q00 600 500 400 300 200 100 0 INTEGRATED LEAK FLOW I a CThaueande)

TIME, seconads'Uw.c'190 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-18 Revision 21 SGTR ADV FLOW RATE vs TIME ADV FLOW RATE u I.-300 210 260 240 220 200 160 140 120 100 soo so 40 40 20 0 0 4 a 12 16 20 24 28 CThousands)

TIME, seconas FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-19 Revision 21 SGTR INTEGRATED ADV FLOW vs TIME INTEGRATED ADV FLOW 800 700 600 500 400 300 200 100 0 Affolte~ S/GG II ] I I I I I J I I I I I I I 0 4 6 12 is Cfhousands)

T I ME. econas 20 24 2e FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-20 71 Ri~viQinn 91 SGTR PCS SUBCOOLING vs TIME FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.15-21 Revision 21 SGTR HPSI FLOW RATE vs TIME HPSI FLOW RATE so 50 40 30 20 w ILI 4 B a a CThaueandin)

TIME, secondS 20 24 28 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.16-1 Revision 23 CONTROL ROD EJECTION, EOC HZP CASE: CORE POWER.60000 50000?10000.030000&20000 10000 oo0..*I.. -0 5 io .Time (8)1.5 20 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.16-2 Revision 23 CONTROL ROD EJECTION, EOC HZP CASE: CORE AVERAGE HEAT-FLUX-BASED LHR 8.-Core Ave. LHR 6.4 4.83.2 1.8.0 0 5 10 15 20 Time (3)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.16-3 Revision 23 CONTROL ROD EJECTION, EOC HZP CASE: TOTAL CORE REACTIVITY 2.o-a Total 0-1.0C-2.-4.0 5 10 15 20 Time (W)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-1 Revision 28 SCATTER PLOT OF OPERATIONAL PARAMETERS One-Sided I I Break Area orn m .*N im me *(ft lside)0.0 1.0 2.0 3.0 4.0 5.0 BumI Time oooo ooo*o * * *****oooooo (hours)0.0 5000.0 10000.0 15000.0 Core Power Lem m mm Om (MW 2565.0 2570.0 2575.0 2580.0 2585.0 LHGR L a =00,1011111am am m (KWtft)12.0 13.0 14.0 15.0 16.0 A S I U P1o1IM i-0.4 -0.3 -0.2 -0.1 00 0.1 0.2 03 0.4 Pressurizer

'Pressure m oo 01 0o11,m m Om 0 (psia)2000.0 2020.0 2040.0 2060.0 2080.0 2100.0 Pressurizer Liquid Level mlO el qo nm (%)40.0 50.0 60.0 70.0 RCS (Tcold)Temperature m m m omilm m m mm.('F) I m '536.0 538.0 540.0 542.0 544.0 TotalII Loop Flow 011111 m moo Nmm. e Om (Mlb/hr)130.0 135.0 140.0 145.0 SIT Liquid [Volume m mm O oem Om 1000.0 1050.0 1100.0 1150.0 1200.0 SIT Pressure

om mumo m om (psia)210.0 220.0 230.0 240.0 Containment Volume mB mc m O mD (ft3)1.60e+06 1.65e+06 1.70e+06 1.75e+06 1.80e+06 SIT ' 'Temperature mm ummui om Om .me. m em (F)80.0 100.0 120.0 140.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-2 Revision 28 PCT VERSUS PCT TIME SCATTER PLOT FROM TRANSIENT CALCULATIONS PCT vs Time of PCT 2000 1800 1600 1400 -U ZI.-1200 1000 lip 800 -M E 0 Split Break ]D Guillotine Break 600 400 ' ' '0 100 200 300 400 500 Time of PCT (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-3 Revision 28 PCT VERSUS BREAK SIZE SCATTER PLOT FROM TRANSIENT CALCULATIONS PCT vs One-sided Break Area 2000,tI 1800 F Li 1600 F* .El[]Li LI D EPI 0 El~1400 LILI Li Li LL 0 0-0 1200 F U U 1000 F U U U U if m 800 600 F* Split Break EL Guillotine Break 400 L 0.0 1.0 2.0 3.0 Break Area (ft2/side) 4.0 5.0 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-4 Revision 28 TOTAL OXIDATION VS. PCT SCATTER PLOT FROM TRANSIENT CALCULATIONS I Total Oxidation vs PCT 0.10 0 Split Break l Guillotine Break 0.08 k 0.06 F X 0 0.04 k 0.02 V[]LID, 0.00 40 I =0 600 800 1000 1200 1400 1600 1800 2000 PCT (°F)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-5 Revision 28 MAXIMUM OXIDATION VERSUS PCT SCAlTER PLOT MAXIMUM OXIDATION VERSUS PCT SCATTER PLOT FROM TRANSIENT CALCULATIONS Maximum Oxidation vs PCT 1.0 0 Split Break El Guillotine Break 0.9 k L]0.8 0.7 0.6 k El El C.o 0.5 x 0 0.4 E 0.3 V El Li1 r]EU li F-1 mill*Ei Eli El EU 0.2 0.1 0.0 400 600 800 1000 1200 PCT (F)1400 1600 1800 2000 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-6 Revision 28 PEAK CLADDING TEMPERATURE FOR THE LIMITING CASE PCT Trace for Case #22 PCT = 1739.1 OF, at Time = 27.15 s, on 6% Gad Rod 2000 1500 U-E 1000 0~C-500 0 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-7 Revision 28 BREAK FLOW FOR THE LIMITING CASE Break Flow 80 r--- -, ----Vessel Side--- Pump Side Total 60 E u-0)(U40 20 44t" 0-20 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-8 Revision 28 CORE INLET MASS FLUX FOR THE LIMITING CASE Core Inlet Mass Flux 1000 500-- Hot Assembly---- Surround Assembl---Average Core Outer Core cI E.0 x ca C 0-500 100 200 0 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-9 Revision 28 CORE OUTLET MASS FLUX FOR THE LIMITING CASE Core Outlet Mass Flux 1000 500 E-o xn U)0-500-1000 0 100 200 Time (s)300 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-10 Revision 28 VOID FRACTION AT PCS PUMPS FOR THE LIMITING CASE Pump Void Fraction 1.0 0.8 0.6 0.2 0.4 I,:~;Broken Loop 1 Intact Loop 2 Intact Loop 3 Intact Loop 4 0.2 0.0 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-11 Revision 28 ECCS FLOW (INCLUDES SIT, HPSI, AND LPSI) FOR THE LIMITING CASE ECCS Flows 3000 2000.0 LL 1000 0 0 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-12 Revision 28 UPPER PLENUM PRESSURE FOR THE LIMITING CASE Upper Plenum Pressure 3000 2000 5D a-1000 0 0 100 200 Time (s)300 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-13 Revision 28 COLLAPSED LIQUID LEVEL IN THE DOWNCOMER FOR THE LIMITING CASE Downcomer Liquid Level 30 1 Sector 1 (broken)Sector 2---- Sector 3 Sector 4 Average 20 10 0 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-14 Revision 28 COLLAPSED LIQUID LEVEL IN THE LOWER PLENUM FOR THE LIMITING CASE Lower Vessel Liquid Level 10 8 6 75-J 4 2-0 1 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-15 Revision 28 COLLAPSED LIQUID LEVEL IN THE CORE FOR THE LIMITING CASE Core Liquid Level 15 10 ci, (D-j 5, O 0 100 200 Time (s)300 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-16 Revision 28 CONTAINMENT AND LOOP PRESSURES FOR THE LIMITING CASE Containment and Loop Pressures 100 90 80 70 SG Outlet (primary Upper Plenum Downcomer Inlet (j~0 C,)(0 0 0~60 50 40 30 20 10 0 0 100 200 300 Time (s)

FSAR CHAPTER 14 -SAFETY ANALYSIS CORE EFFECTIVE FLOODING RATE DELETED in Revision 28 FIGURE 14.17.1-17 Revision 28 FSAR CHAPTER 14 -SAFETY ANALYSIS CORE COLLAPSED LIQUID LEVEL FIGURE 14.17.1-18 Revision 28 DELETED in Revision 28 FSAR CHAPTER 14 -SAFETY ANALYSIS CORE QUENCH LEVEL FIGURE 14.17.1-19 Revision 28 DELETED in Revision 28 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-20 Revision 28 PCT-NODE HEAT TRANSFER COEFFICIENT DELETED in Revision 28 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.1-21 Revision 28 PEAK CLADDING AND RUPTURE LOCATION CLADDING TEMPERATURE FOR THE LIMITING CASE DELETED in Revision 28 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-1 Revision 28 Break Mass Flow Rate (Limitina Case)1D:089W8 240cd2007 20:59 0.08 ft2 Break 1500 1000 10 LJL 500 0 2500 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-2 Revision 28 Primary and Secondary Pressures (Limiting Case)ID:08968 240ct2007 20:59:00 PaLM5.0.08ft2_DMX 2400 2200 2000 1800 1600 1400 1200 (0 1000 0.08 ff2 Break 0 500 1000 1500 2000 2500 lime (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-3 Revision 28 Normalized Reactor Power (Limitinq Case)ID:08968 240ct2007 20:59:00 paLM5_0.08ft2_DMX 120 110 100 90 80 S70 60 50 0.2 50 0 40 30 20 10 0.08 ft2 Break 0 500 1000 1500 2000 2500 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-4 Revision 28 Total HPSI Mass Flow Rate (Limitina Case)ID:089W8 24Oct2007 20:59:00 paLM5._0.Ofi2_DMX 100 80 0.08 ft2 Break U.60 40 20 0 0 500 1000 1500 2000 2500 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-5 Revision 28 Total SIT Mass Flow Rate (Limitinq Case)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-6 Revision 28 Loop Seal Void Fractions (Limitinq Case)ID:08968 240ct2007 20:59:00 paL M5_0.08ft2_DMX 1.00 0.90-0 LO&Lo 0.800--*L--A LO 0.70-0.60._2 ,,E 0.50 0.40 0.30 0.20 0.10 0.00 0.08 ft2 Break op 1A op 1B op 2A op 2B (broken)lure --0-'-0 500 1000 Time (sec)1500 2000 2500 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-7 Revision 28 Break Void Fraction (Limitinq Case)ID:08968 240ct2007 20:59:00 PaL M5_0.08f2_DMX 1.00 0.90 0.80 0.70-0.60.2 ,,f 0.50 0.40 0.30 0.20 0.10 0.00 0.08 fM2 Break IF...-F--.. I a 500 1000 1500 2000 2500 Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-8 Revision 28 Reactor Vessel and PCS Mass Inventories (Limitinq Case)D:08968 240ct2007 20:59:00 paLM5_0.08fi2_DMX 300000--0 PC 250000 .......-.

R%200000 150000 100000 50000 0 0 500 0.08 ft2 Break 1000 1500 2000 Time (sec)2500 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-9 Revision 28 Hot Channel Collapsed Level (Limitincq Case)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-10 Revision 28 Fluid and Claddinq Temperatures (Limitinq Case)ID:08968 240ct2D07 paLM5_0.8f2_DMX 1800 1600 1400 L 1200 E 1000 800 600 400 0 500 0.08 ft2 Break Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-11 Revision 28 SG Narrow Range Liquid Levels (Limiting Case)ID:08968 240ct2007 20:59.00 paLIM5_0.08ft2_DMX 100 80 z 60,-J S40 20 0.08 ft2 Break Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-12 Revision 28 AFW Flow Rates (Limitinq Case)ID:08968 240ct2007 20;59:00 PaLMS._0.08MDMX 50 40 30 20 10-1-10 Lr 0.08 ft2 Break Time (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.17.2-13 Revision 28 Total MSSV Flow (Limiting Case)ID:089W8 240c12007 20:59:00 paLM_5_0.08ft2_DMX 0.08 ft2 Break 800 600 400 200--0 SG-1.........

s SG-2 0 500 Imoo 100 2000 2500 lime (sec)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.18.1-1 Revision 29 LOCA CONTAINMENT PRESSURE PROFILE 60 50 _____ _____ ____40-* 30 20 10 -0 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 Time (sec)-DIG 1-1 Failure ----DIG 1-2 Failure FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.18.1-2 Revision 29 LOCA CONTAINMENT TEMPERATURE PROFILE 300 280 260 240-220 S200 180 160 140 120 100 1/////// // I//E-01 1.E+00 1.E+01 1.E+02 1.E+03 Time (sec)1.E+04 1.E+05 1.E+06 1.E+07 I -DIG 1-1 Failure ---- DIG 1-2 Failure -,EEQ Pro~file FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.18.2-1 Revision 29 MSLB CONTAINMENT RESPONSE MAXIMUM PRESSURE PROFILE I Corlainmrnnt Prbssure PRI 1 .1 CL 1 10 100 loo 1o+0o4 le+006 Tlme (sec)GOTI4W 7.2943kI Jam2"),GlO 1,52t1S FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.18.2-2 Revision 31 MSLB CONTAINMENT RESPONSE ENVIRONMENTAL QUALIFICATION PROFILE 400 EU I.E+00 1.E+01 1.E+02 1.+E03 1.E+04 1.E+05-EEO Limit-- -. -P-7 (0%)....SIS-6 (102%)--- -0% 1-1 EOG Fagure with LOOP------ 5P-7 (102%)...... 0% 1-2 EDG Failure with LOOP-... SIS-6 (0%)

FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.22-1 Revision 25 PALISADES CONTAINMENT HYDROGEN ANALYSIS I Deleted per FSAR-2479 FSAR CHAPTER 14 -SAFETY ANALYSIS FIGURE 14.22-2 Revision 25 CONTAINMENT TEMPERATURE FOR H2 GENERATION I Deleted per FSAR-2479 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-1 Revision 21 Page 1 of I INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER PARAMETER Initial core power level, MWt Core inlet coolant temperature, OF Core mass flow rate, 106 Ibm/hr Reactor coolant system pressure, psia Steam generator pressure, psia Initial pressurizer liquid volume, ft 3 Steam generator level, ft above tube sheet* Lower core flowrate dispositioned in Reference 6.ASSUMED VALUE 2600.6 550.65 138.*2,110 770 800 31.74 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-2 Revision 21 Paae 1 of I SETPOINTS FOR THE STEAM GENERATOR TUBE RUPTURE SETPOINTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Parameter Setpoint Steam generator MSSV setpoint, psia AFW actuation on steam generator level AFAS signal generation, % NR SIAS setpoint, psia Shutdown cooling entry conditions:

Hot leg temperature, OF Pressurizer pressure, psia 1000 23.7 1605 300 270 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-3 Revision 21 Page 1 of 2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Time Setpoint (second) Event or Value 1.0 Tube rupture occurs ----32.9 Proportional heaters are fully energized, psia 2085 105.7 Backup heaters are energized, psia 2035 211.3 Heaters are de-energized on low level in the pressurizer, ft 558.6 703.7 Pressurizer pressure reaches low pressurizer pressure setpoint (TM/LP floor), psia 1700.704.8 Trip signal is generated 705.2 Trip breakers open 706.1 Turbine Valves begin to close 707.1 Turbine valves are completely closed 708.2 Loss of offsite power 714.8 Feedwater flow begins ramping down at a rate of 5%/second 715.9 SIAS setpoint is reached, psia 1605 720.3 MSSVs begin to open, psia 1000 725.8 Pressurizer empties 733.9 Safety Injection pumps reach full speed 735.0 Upper head void begins to appear 811.5 Safety Injection flow to RCS begins, psia 1237.7 995.0 Maximum upper head void fraction 0.271 1107.0 Minimum PCS pressure, psia 1107.8 1370.5 Upper head void disappears FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-3 Revision 21 Page 2 of 2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Time Setpoint (second) Event or Value 1372.0 Pressurizer begins to refill 1466.6 Low steam generator level signal for Auxiliary feedwater actuation, ft 25.7 1586.6 Auxiliary feedwater reaches the steam generators, Ibm/sec/SG 27.0 1800.0 Operator takes action, opens ADVs to initiate cooldown 3000.0 Operator isolates the affected SG, below setpoint loop temperatures, OF 525.0 13000.0 Operator initiates steaming the affected generator to avoid overfilling, percent SG wide range span 90 23300.0 Shutdown Cooling entry condition is reached, PCS pressure, psia/temperature, OF 270/300 28800.0 PCS pressure and temperature demonstrated to be stabilized, transient terminated.

FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-4 Revision 21 Page 1 of I INTEGRATED PARAMETERS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Parameter 0-2 hr 0-8 hr Integrated primary to secondary leak, Ibm 183,202 605,101 Integrated Steam release, Ibm a. Through affected SG ADV 37,382 313,736 b. Through affected SG MSSV 44,654 44,654 c. Through intact SG ADV 185,000 719,448 d. Through intact SG MSSV 44,645 44,645 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-5 Revision 28 Page 1 of I STEAM GENERATOR TUBE RUPTURE (SGTR) RADIOLOGICAL ANALYSIS-INPUTS AND ASSUMPTIONS Input/Assumption Value Core Power Level 2703 MWth Initial PCS Equilibrium Activity 1.0 [tCi/gm DE 1-131 and 100/E-bar gross Initial __PCS _EquilibriumActivity_

activity Initial Secondary Side Equilibrium Iodine Activity 0.1 ILCi/gm DE 1-131 Maximum pre-accident spike iodine concentration 40 jtCi/gm DE 1-131 Maximum equilibrium iodine concentration 1.0 ,tCi/gm DE 1-131 Duration of accident-initiated spike 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdown cooling and terminate 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months steam release 529,706 Ib,, for pre-accident iodine spike PCS Mass case 459,445 Ibm for concurrent iodine spike case 141,065 Ibm per SG (minimum mass used to SG Secondary Side Mass maximize concentration from tube leakage)Integrated Mass Release Table 14.15-6 Secondary Coolant Iodine Activity prior to 0 1 DE 1-131 accident Faulted SG (flashed tube flow) -Table Steam Generator Secondary Side Partition 14.15-11 Coefficients Faulted SG (non-flashed tube flow)- 100 Intact SG -100 Break Flow Flash Fraction Table 14.15-7 Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Control Room Ventilation System Time of manual control room normal 20 minutes intake isolation and switch to emergency mode Breathing Rates Offsite RG 1.183, Section 4.1.3 Control Room RG 1.183, Section 4.2,6 Control Room Occupancy Factor RG 1.183 Section 4.2-6 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-6 Revision 28 Page I of I SGTR RADIOLOGICAL ANALYSIS -INTEGRATED MASS RELEASES (1)Time Break Flow in Steam Release from Ruptured SG Steam Release from (hours) Ruptured SG (Ibm Unaffected SG (lb,")) (ibm)0 0,190417 24,011,15 0 0 0.196417 -0.5 37,111.85 44,654 53,574 0.5 -1.388889 81,281 22.152.3 109,629.6 1.388889 2 40,798 15,229.7 75,370.4 2- 3 (38889 64,773 75,485.6 145,983.5 3,638889-8 357.126 200,868.4 388,464.5 8- 7201 0 0 0 I Ih I wr1tc assIiumcd to te constant within time period FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-7 Revision 28 Paiae I of I SGTR RiADIOLOGICAL ANALYSIS -FLASliNG FRACTION FOR FLOW FROM BROKEN ITUBE (seconds)

Flashing Fraction 0 (HI I(0 707 I 0.065 736 0.031 859 0.023 1090 0.006 1800 0.006 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-8 Revision 28 Page 1 of I SuTR RADIOLOGICAL ANALYSIS-40 iCI/GM D.E. 1-131 ACTIVITIES Isq)(opeActivity

________________________________(p~Ci/gni)

Iodine-131 33.2194 lodine-I1 32 7.6660)Iodine- 1 33 ,34.4971 Iodine- 134 3.002 5 IodifC- 135 14,6932 FSAR CHAPTER 14 -SAFETY ANALYSIS TABLE 14.15-9 Revision 28 Page I of I SGTR RADIOLOGICAL ANALYSIS -IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS Input Assumption Value Maximum Letdown [low 40 gpm Assumed Letdown Flow

44 gpm at 120F., 2060 psia\McxiImm Idicnti licd PtCS Leakage 10 gp11 Maximum Unidentified PC'S Leakage I gpm I'CS Mass 459,445 lb,, 1-131 Decay Constant 5.986968F-5 min I- 132 Decay Constant 0.005023 min I- 133 Decay Constant 0.000555 min 1-134 Decay Constant 0.013178 min I- 135 Decay Constant 0.O01 748 min* maximuml letdown 11ow plus 10% uncerlainty FSAR CHAPTER 14- SAFETY ANALYSIS TABLE 14.15-10 Revision 28 Page 1 of I S(ITR RAI)IOIICAL ANALYSIS -CON( URRENT (335 X) IODINE SPIKE APPEARANCE RATE Isotope Appearance Rate Time of Depletion (Ci/min) (hours)Iodine-131 58.0966961 8 Iodine- 132 79.8319317 8 Iodine- 133 90.13 10904 , Iodine- 134 74.0318685 8 Iodine- I 35 68.9790622 8