ML21125A338

5 to Updated Final Safety Analysis Report, Chapter 14, Tables

ML21125A338
Person / Time
Site:	Palisades
Issue date:	04/14/2021
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
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FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-1 Revision 35 Page 1 of 1 SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS ACCEPTANCE CRITERIA VALUE REFERENCE HTP DNB Correlation 95/95 Limit 1.141* 21 Fuel Centerline Melt 21.000 kw/ft 50

• Different DNB correlations and applicable 95/95 limits apply depending on operating conditions. A mixed core penalty is also applied to these limits. See Reference 10.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-3 Revision 26 Page 1 of 1 TRIP SETPOINTS FOR ANALYSIS OF PALISADES REACTOR AT 2565.4MWt TRIP SETPOINT VALUES DELAY TIME (sec)

Low Reactor Coolant Flow 93% of Tech Spec flow 0.8 High Pressurizer Pressure 2,277 psia 0.8 Low Pressurizer Pressure 1,750 psia 0.8 Low Steam Generator Pressure 485 psia 0.8 Low Steam Generator Level1 23.7% 0.8 Thermal Margin/Low Pressure2 = P f(TH,TC) 0.8 Variable High Power Trip <23.1%3 above power 0.6 with a 113.4% maximum and a 36.35% minimum 1

Narrow Range Level 2

The TM/LP trip setpoint is based on pressurizer pressure (P) setpoint, varying as a function of the maximum cold leg temperature (TC), the measured power, and the measured axial shape index.

3 Used for fast transient. For slow transient, it is 20.2% above power.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 1 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 (b) 15.1.2 Increase in Feedwater Flow Bounded 15.1.3 (b) 15.1.3 Increase in Steam Flow 14.10

1. Transient Response Bounded Ref. 29

2. MDNBR Reanalyzed Ref. 61

3. FCM/LHR Reanalyzed Ref. 61 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve

1. Power Operation (Mode 1) Bounded 15.1.3

2. Startup (Mode 2) Bounded 15.4.1 15.1.5 Steam System Piping Failures 14.14 Inside and Outside of Containment

1. Transient Response Bounded Ref. 36

2. MDNBR Reanalyzed Ref. 61

3. LHR Reanalyzed Ref. 61

4. Reactivity Verification Reanalyzed Ref. 61

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 2 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load 14.12

1. Primary Over-pressurization Bounded Ref. 38

2. Secondary Over-pressurization Bounded Ref. 30

3. MDNBR Non-Limiting Ref. 18 15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Valves (MSIVs) Bounded 15.2.1 15.2.5 Steam Pressure Regulator Failure Not Applicable; BWR Event 15.2.6 Loss of Nonemergency AC Power to the Station Auxiliaries

1. Short-term Consequences Bounded 15.3.1

2. Long-term Consequences Bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow 14.13

1. Primary Over-pressurization Bounded 15.2.1

2. Minimum Steam Generator Inventory Bounded Ref. 9

3. MDNBR Bounded 15.3.1

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 3 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment

1. Cooldown Bounded 15.1.5

2. Heatup Bounded 15.2.1

3. Long-term Cooling Bounded 15.2.7 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow 14.7

1. Transient Response Bounded Ref. 11(a)

2. MDNBR Reanalyzed Ref. 61 15.3.2 Flow Controller Malfunction Not Applicable 15.3.3 Reactor Coolant Pump Rotor Seizure 14.7

1. Transient Response Bounded Ref. 11(a)

2. MDNBR Reanalyzed Ref. 61 15.3.4 Reactor Coolant Pump Shaft Break Bounded 15.3.3 14.7

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 4 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal From a Subcritical or Low Power Startup Condition 14.2.1

1. Transient Response (Mode 2 /Mode 3) Bounded Ref. 27(a)/Ref. 5

2. MDNBR Non-Limiting Ref. 27 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 14.2.2

1. Transient Response Bounded Ref. 13(a)

2. MDNBR Reanalyzed Ref. 61

3. FCM/LHR Reanalyzed Ref. 61 15.4.3 Control Rod Misoperation 15.4.3.1 Dropped Control Bank/Rod 14.4

1. Transient Response Bounded Ref. 11(a)

2. Half-Scram Transient Bounded Ref. 7

3. MDNBR Reanalyzed Ref. 61

4. FCM/LHR Reanalyzed Ref. 61 15.4.3.2 Dropped Part-Length Control Rod Bounded 15.4.3.1 14.6 15.4.3.3 Malpositioning of the Part-Length Control Group Not Applicable

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 5 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.4.3.4 Statically Misaligned Control Rod/Bank 14.6

1. MDNBR Reanalyzed Ref. 61

2. FCM/LHR Reanalyzed Ref. 61 15.4.3.5 Single Control Rod Withdrawal 14.2.3

1. Transient Response Bounded Ref. 15.4.2

2. MDNBR Reanalyzed Ref. 61

3. FCM/LHR Reanalyzed Ref. 61 15.4.3.6 Core Barrel Failure Bounded 15.4.8 14.5 15.4.4 Start-Up of an Inactive Loop Not Credible 14.8 15.4.5 Flow Controller Malfunction Not Applicable; No Flow Controller 15.4.6 CVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant Reanalyzed Ref. 61 14.3 15.4.7 Inadvertent Loading and Operation Not Part of of a Fuel Assembly in an Licensing Basis Improper Position

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 6 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.4.8 Spectrum of Control Rod Ejection Accidents 14.16

1. Transient Response Bounded Ref. 38

2. MDNBR Reanalyzed Ref. 61

3. FCM/LHR Reanalyzed Ref. 61

4. Deposited Enthalpy Reanalyzed Ref. 61

5. Radiological Analysis Bounded Ref. 40 and 41 15.4.9 Spectrum of Rod Drop Accidents Not Applicable; (BWR) BWR Event 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 Inadvertent Operation of the ECCS That Increases Reactor Coolant Inventory

1. Primary Over-pressurization Bounded 15.2.1

2. Reactivity Bounded 15.4.6 15.5.2 CVCS Malfunction That Increases Reactor Coolant Inventory

1. Primary Over-pressurization Bounded 15.2.1

2. Reactivity Bounded 15.4.6

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 7 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve

1. Short-term Consequences Bounded Ref. 11(a)

2. Long-term Consequences Bounded 15.6.5

3. MDNBR Reanalyzed Ref. 61 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment Bounded 15.6.5 14.23 15.6.3 Radiological Consequences of Steam Generator Tube Failure Bounded Ref. 41 14.15 15.6.4 Radiological Consequences of a Not Applicable; Main Steam Line Failure Outside BWR Event Containment 15.6.5 Loss of Coolant Accidents Resulting From a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary

1. SBLOCA Bounded Ref. 53 14.17.2

2. LBLOCA Bounded Ref. 51 14.17.1 UO2 Rods Bounded Ref. 51 8% Gadolinia Rods/6% Gadolinia Rods/

2% Gadolinia Rods/4% Gadolinia Rods Bounded Ref. 51

3. Radiological Consequences Bounded Ref. 12 14.22

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-4 Revision 35 Page 8 of 8 DISPOSITION OF EVENTS

SUMMARY

FOR PALISADES CYCLE 28 Bounding Updated SRP Event Event or FSAR Designation Name Disposition Reference Designation 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Deleted(c) Ref. 41 14.21 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) Deleted(c) Ref. 41 14.20 15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures Bounded Ref. 41 14.20 15.7.4 Radiological Consequences of Fuel Handling Accidents Bounded Ref. 23 14.19 15.7.5 Spent Fuel Cask Drop Accidents Bounded Ref. 24 14.11

a. PTSPWR2 system response analysis is given reference for this event.

b. Deleted from the FSAR.

c. This section of the Standard Review Plan has been deleted.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-5 Revision 35 Page 1 of 3 CYCLE 28

SUMMARY

OF RESULTS FOR STANDARD REVIEW PLAN CHAPTER 15 EVENTS (Reference 61)

SRP FSAR EVENT MDNBR PEAK(2) LHR (kW/ft) MAXIMUM PRESSURE (psia) 15.1.3 14.10 Increase in Steam Flow 1.314 17.43 Note5 15.1.5 14.14 Steam System Piping Failures 1.250 17.55 Note3 Inside and Outside of Containment 15.2.1 14.12 Loss of External Load Note5 Note5 26274 15.2.7 14.13 Loss of Normal Feedwater Note5 Note5 Note5 15.3.1 14.7.1 Loss of Forced Reactor Coolant Flow 1.206 Note5 Note5 15.3.3 14.7.2 Reactor Coolant Pump Rotor Seizure 1.237 Note5 Note5 15.4.1 14.2.1 Uncontrolled Control Bank Withdrawal Note5 Note5 Note5 15.4.2 14.2.2 Uncontrolled Control Bank Withdrawal 1.286 17.88 Note5 at Power 1 Not used.

2 The calculated FCM limit for Cycle 28 was bounded by the Technical Specification limit of 21.000 kW/ft, which was used as the actual Cycle 28 limit for UO2 rods. The centerline melt temperature limit is 4,687oF. These values preclude melting of any fuel rod (with or without gadolina).

3 This is a depressurization event. Limit is not challenged by this event.

4 The maximum secondary side pressure for the secondary side over pressurization case is 1,063.3 psia.

5 Limit is not challenged by this event.

6 Case with minimum margin to DNBR 95/95 limit is reported.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-5 Revision 35 Page 2 of 3 CYCLE 28

SUMMARY

OF RESULTS FOR STANDARD REVIEW PLAN CHAPTER 15 EVENTS (Reference 61)

SRP FSAR EVENT MDNBR PEAK(7) LHR (kW/ft) MAXIMUM PRESSURE (psia) 15.4.3 Control Rod Misoperation 14.4.1 -Dropped Rod 1.168 18.55 Note9 14.4.2 -Dropped Bank8 14.2.3 -Single Rod Withdrawal 1.338 18.90 Note9 14.6 -Statically Misaligned Control Rod 1.250 17.35 Note9 15.4.6 14.3 CVCS Malfunction Resulting in Adequacy of Shutdown Margin is demonstrated Decreased Boron Concentration 15.4.8 14.16 Control Rod Ejection Less than 4306 oF Peak Fuel Note9 HTP limit Centerline Temperature

<12% Fuel Failures Predicted 15.6.1 Note10 Inadvertent Opening of a PWR 1.454 Note9 Note9 Pressurizer Pressure Relief Valve 7 The calculated FCM limit for Cycle 28 was bounded by the Technical Specification limit of 21.000 kW/ft, which was used as the actual Cycle 28 limit for UO2 rods. The centerline melt temperature limit is 4687oF. These values preclude melting of any fuel rod (with or without gadolina).

8 Bounded by dropped rod.

9 Limit is not challenged by this event.

10 This event is not part of the Licensing Basis analyses for Palisades

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-5 Revision 35 Page 3 of 3 CYCLE 28

SUMMARY

OF RESULTS FOR STANDARD REVIEW PLAN CHAPTER 15 EVENTS (Reference 61)

SRP FSAR EVENT DISPOSITION/RESULTS 15.6.2 14.23 Radiological Consequences of the Failure of Small Note11 Lines Carrying Primary Coolant Outside Containment 15.6.3 14.15 Radiological Consequences of Steam Generator Tube Failure Note11 15.6.5 14.17 Loss of Coolant Accidents Resulting from a Spectrum Note11, 10CFR50.46(b) 14.22 of Postulated Piping Breaks Within a Reactor Coolant acceptance criteria are met 15.7.3 14.20 Postulated Radioactive Releases due to Liquid-Containing Note11 Tank Failures 15.7.4 14.19 Radiological Consequences of Fuel Handling Accidents Note11 15.7.5 14.11 Spent Fuel Cask Drop Accidents Note11 11 Radiological consequences acceptance criteria are met.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-6 Revision 33 Page 1 of 4

SUMMARY

OF RADIOLOGICAL CONSEQUENCES OF THE CHAPTER 14 EVENTS FSAR SECTION SCENARIO DESCRIPTION OFFSITE DOSES AND LIMITS CONTROL ROOM HABITABILITY Exclusion Time to E-HVAC Pre- Post- Low Control Offsite Area [min] Control Room Cask Drop Isolation Isolation Population Room Dose Limits (5) Boundary Dose Limits (6)

Scenario  % Filter  % Filter Zone CRE Inleakage Dose

[rem] (0-2 hrs) [rem]

Bypass Bypass [rem] [cfm] [rem]

[rem]

Section 14.11: 30 Day Decay 0 min CASK DROP IN THE SPENT w/Charcoal N/A 10 TEDE 6.3 2.04 0.25 TEDE 5 1.37 FUEL POOL Filters 100 cfm (Section 14.24 for CRH) 30 Day Decay 0 min w/Charcoal N/A 17.5 TEDE 6.3 2.78 0.35 TEDE 5 1.99 Filters 100 cfm 90 Day Decay Not Required (3)

No Charcoal 100 100 TEDE 6.3 0.08 0.01 TEDE 5 1.67 Filters 100 cfm Exclusion Time to E-HVAC Low Control Offsite Area [min] Control Room Population Room Section 14.14: Dose Calculation Assumptions Dose Limits (5) Boundary Dose Limits (6)

Zone CRE Inleakage Dose STEAM LINE BREAK [rem] (0-2 hrs) [rem]

[rem] [cfm] [rem]

[rem]

(Section 14.24 for CRH) 20 min Break Outside the Containment TEDE 25 0.67 0.20 TEDE 5 3.34 20 cfm Exclusion Time to E-HVAC Low Control Offsite Area [min] Control Room Population Room Section 14.15: Iodine Spike Assumptions Dose Limits (5) Boundary Dose Limits (6)

Zone CRE Inleakage Dose STEAM GENERATOR TUBE [rem] (0-2 hrs) [rem]

[rem] [cfm] [rem]

RUPTURE WITH A LOSS OF [rem]

OFFSITE POWER 20 min Previous Iodine Spike TEDE 25 0.99 0.22 TEDE 5 3.79 100 cfm (Section 14.24 for CRH) 20 min Generated Iodine Spike TEDE 2.5 1.17 0.21 TEDE 5 3.48 100 cfm Exclusion Time to E-HVAC Low Control Offsite Area [min] Control Room Population Room Dose Calculation Assumptions Dose Limits (5) Boundary Dose Limits (6)

Section 14.16: Zone CRE Inleakage Dose

[rem] (0-2 hrs) [rem]

CONTROL ROD EJECTION [rem] [cfm] [rem]

[rem]

EVENT Auto Switch (4)

Containment Release TEDE 6.3 2.60 0.68 TEDE 5 1.27 (Section 14.24 for CRH) 20 cfm 20 min Steam Generator/ADV Release TEDE 6.3 2.70 0.43 TEDE 5 1.00 20 cfm

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-6 Revision 33 Page 2 of 4

SUMMARY

OF RADIOLOGICAL CONSEQUENCES OF THE CHAPTER 14 EVENTS FSAR SECTION SCENARIO DESCRIPTION OFFSITE DOSES AND LIMITS CONTROL ROOM HABITABILITY Exclusion Time to E-HVAC Low Control Offsite Area [min] Control Room Population Room Dose Calculation Assumptions Dose Limits (5) Boundary Dose Limits (6)

Zone Dose

[rem] (0-2 hrs) CRE Inleakage [rem]

[rem] [rem]

[rem] [cfm]

Section 14.19:

FUEL HANDLING INCIDENT One fuel bundle fails 2 days after shutdown, 20 min TEDE 6.3 2.20 0.28 TEDE 5 4.04 No Charcoal Filtration 100 cfm (Section 14.24 for CRH)

One fuel bundle fails 2 days after shutdown, 20 min TEDE 6.3 2.02 0.25 TEDE 5 3.68 10% Filtered Release 100 cfm One fuel bundle fails 2 days after shutdown, 20 min TEDE 6.3 1.31 0.17 TEDE 5 2.22 50% Filtered Release 100 cfm Low Time to E-HVAC Site Control Offsite Dose Population [min] Control Room Dose Boundary Room Section 14.21.1: Dose Calculation Assumptions Limits (1) Zone Limits (2)

(60 min) Dose GAS DECAY TANK RUPTURE [rem] (60 min) CRE Inleakage [rem]

[rem] [rem]

[rem] [cfm]

(Section 14.24 for CRH)

Thyroid 75 0.000 0.000 20 min Thyroid 30 0.000 22 Hr Decay Tank Inventory Whole Body 6 0.415 0.067 85 cfm Whole Body 5 0.036 Low Time to E-HVAC Site Control Offsite Dose Population [min] Control Room Dose Boundary Room Iodine Spike Assumptions Limits (1) Zone Limits (2)

(60 min) Dose

[rem] (60 min) CRE Inleakage [rem]

[rem] [rem]

[rem] [cfm]

Section 14.21.2:

VOLUME CONTROL TANK Thyroid 75 1.59 0.258 20 min Thyroid 30 0.277 RUPTURE Equilibrium Iodine Whole Body 6 0.081 0.013 85 cfm Whole Body 5 0.005 (Section 14.24 for CRH) Thyroid 75 63.77 10.32 20 min Thyroid 30 11.086 Previous Iodine Spike Whole Body 6 0.081 0.013 85 cfm Whole Body 5 0.005 Thyroid 75 3.69 0.597 20 min Thyroid 30 0.922 Generated Iodine Spike Whole Body 6 0.081 0.013 85 cfm Whole Body 5 0.005

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-6 Revision 33 Page 3 of 4

SUMMARY

OF RADIOLOGICAL CONSEQUENCES OF THE CHAPTER 14 EVENTS FSAR SECTION SCENARIO DESCRIPTION OFFSITE DOSES AND LIMITS CONTROL ROOM HABITABILITY Exclusion Time to E-HVAC Low Offsite Area [min] Control Room Control Population Offsite Dose Calculation Assumptions Dose Limits (5) Boundary Dose Limits (6) Room Zone

[rem] (0-2 hrs) CRE Inleakage [rem] Dose

[rem]

[rem] [cfm] [rem]

Containment Leakage 10.48 1.53 1.31 Section 14.22: ESF Leakage 2.28 1.74 1.58 PALISADES MAXIMUM HYPOTHETICAL ACCIDENT (MHA) SIRWT Leakage 0.00 0.00 0.00 Non-SIRWT Shine N/A N/A 0.30 SIRWT Shine N/A N/A 0.81 Auto Switch (4)

Total Dose TEDE 25 12.76 3.27 TEDE 5 4.00 16 cfm Exclusion Time to E-HVAC Low Control Offsite Area [min] Control Room Population Room Iodine Spike Assumptions Dose Limits (5) Boundary Dose Limits (6)

Section 14.23: Zone Dose

[rem] (0-2 hrs) CRE Inleakage [rem]

SMALL LINE BREAK [rem] [rem]

[rem] [cfm]

OUTSIDE CONTAINMENT 20 min Generated Iodine Spike TEDE 2.5 0.41 0.05 TEDE 5 0.53 100 cfm

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.1-6 Revision 33 Page 4 of 4

SUMMARY

OF RADIOLOGICAL CONSEQUENCES OF THE CHAPTER 14 EVENTS Notes:

(1) The acceptance criteria for the offsite dose limits are given in 10CFR Part 100 Section 11 (Reference 42). It states that the public 'would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine' following a stipulated fission product release. In addition, the terms 'well within and small fraction of' are utilized in the acceptance criteria in Chapter 15 of the Standard Review Plan (SRP 15) (Reference 43). The two terms correspond to 25% (75 rem thyroid, 6 rem whole body) and 10% (30 rem thyroid, 2.5 rem whole body) of the radiation levels given in 10 CFR Part 100 respectively.

(2) The acceptance criteria for the control room dose limits are given in 10CFR Part 50 Appendix A, General Design Criterion 19 (GDC 19) (Reference 44). It states that the doses will be less than '5 rem whole body or its equivalent to any part of the body, for the duration of the accident.' SRP 6.4 (Reference 45) interprets the limits of GCD 19 of being 5 rem whole body, 30 rem thyroid, and 30 rem skin dose.

(3) 'Not Required' - Control Room CRHVAC emergency mode not needed to maintain control room doses below specified limits. Specifically, the 30 day integrated TEDE doses do not exceed limits assuming no emergency filtration.

(4) 'Auto Switch' - Automatic Switchover to CRHVAC emergency mode following CHP or CHR and Diesel sequencing due to a loss of Offsite Power. Total time to CRHVAC emergency mode is assumed to be 1.5 minutes (Reference 41)

(5) The acceptance criteria for the offsite doses are given in RG 1.183 (Reference 54), which are based upon 10CFR Part 50 Section 67. RG 1.183 and 10CFR 50.67 do not provide specific acceptance criteria for either the cask drop in the spent fuel pool or the small line break outside containment. The acceptance criteria for the cask drop in the spent fuel pool are taken as those of the fuel handling accident. The acceptance criteria for the small line break outside containment are taken to be a small fraction (i.e., 10%) of the 10CFR 50.67 limits.

(6) The acceptance criteria for the control room doses are given in 10CFR Part 50 Section 67.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.2.1-1 Revision 23 Page 1 of 1 EVENT

SUMMARY

FOR THE UNCONTROLLED BANK WITHDRAWAL FROM A LOW POWER EVENT Event Value Time(sec)

Bank Withdrawal Begins 0.00 Letdown Valve Open 0.00 Pressurizer Spray Activates 0.01 Low VHP Trip Setting reached 51.0 Peak Power Level 1815.7 Mwt 51.61 Peak Core Average Heat Flux 56,630 Btu/hr-ft2 52.59 Peak Pressurizer Pressure see Table 14.1-5 65.95

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.2.2-1 Revision 23 Page 1 of 1 EVENT

SUMMARY

FOR THE UNCONTROLLED ROD BANK WITHDRAWAL EVENT FROM POWER EVENT VALUE TIME (sec)

Start Rod Withdrawal 0.00 Letdown Flow Valve Open 0.00 Reactor Scram (TM/LP Trip) 24.82 Turbine Stop Valve Closed 25.00 Peak Power Level 2900.9 MWt 25.38 Peak Pressurizer Pressure 2267.1 psia 27.00 Steam Line Safety Valves Open 29.50 Peak Steam Dome Pressure 1039.6 31.69

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.7.1-1 Revision 21 Page 1 of 1 EVENT

SUMMARY

FOR THE LOSS OF FORCED REACTOR COOLANT FLOW EVENT VALUE TIME (sec)

Initiate Four-Pump Coastdown 0.00 Letdown Flow Valve Opens 0.00 Pressurizer Spray Actuates 0.60 Reactor Scram (Low Flow) 1.64 Turbine Stop Valve Closed 1.80 Peak Power Level 2686.7 MWt 2.20 Minimum DNBR Predicted to Occur see Table 14.1-5 2.91 Peak Core Average Temperature 579.01F 3.37 Peak Pressurizer Pressure 2127.78 psia 4.70 Peak Steam Dome Pressure 963.39 psia 6.64

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.7.2-1 Revision 21 Page 1 of 1 EVENT

SUMMARY

FOR THE REACTOR COOLANT PUMP ROTOR SEIZURE EVENT VALUE TIME (sec)

Primary Coolant Pump Rotor Seizes 0.0 Pressurizer Spray Actuates 0.20 Reactor Scram (Low Flow) 1.13 Turbine Stop Valve Closed 1.40 Peak Power Level 2743.1 MWt 1.68 Minimum DNBR Predicted to Occur see Table 14.1-5 1.76 Peak Core Average Temperature 579.3 °F 2.09 Peak Pressurizer Pressure 2145.19 psia 3.82 Peak Steam Dome Pressure 986.34 psia 6.13

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.10-1 Revision 26 Page 1 of 1 SEQUENCE OF EVENTS FOR EXCESS LOAD LIMITING MDNBR CASE Event Time (sec) Value 30% step increase in steam flow (event initiator) 0.0 ---

Auctioneered core power signal reaches VHPT 22.99 110.4% of RTP setpoint Scram insertion begins 24.08 ---

MDNBR occurred 24.2 See Table 14.1-5

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-1 Revision 32 Page 1 of 1 POSTULATED CASK DROP ACCIDENTS Impact Cask Cask Drop Height Velocity @

Location Weight(Tons) Air(ft) Water(ft) Total(ft) Impact(in/sec) Penetration(in) Impact on Structure STRUCTUAL Cask Loading 97.5 4 35.6 39.6 498.2 Crush depth The Pressure transferred Area of foam=17.15 to the slab is less than P(Critical) for the slab MTC drop into 97.5 16 --- 16 385.2 Crush depth The Pressure transmitted washdown pit of foam=22.96 to the slab is less than P(Critical) for the slab MTC drop onto 93.5 5.83 --- 5.83 232.6 --- MTC remains intact VCC in Track Alley & LDS remains intact MTC seismic --- --- --- --- --- --- The loaded MTC/MSB overturn in pool will not tip over &

loading area damage fuel RADIOLOGICAL MTC Tip over 97.5 --- --- --- --- --- Failure of all 73 onto West 11x7 assemblies & doses fuel rack within applicable 10 CFR 50.67 limits Drop of the loaded 93.5 5.83 --- 5.83 232.6 --- The MTC/MSB will not MTC on to VCC fail. No radiological release.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-2 Revision 28 Page 1 of 4 Table 14.11-2 Spent Fuel Cask Drop Radiological Analysis - Source Terms*

Listed source term is for a single assembly.

30 Day 90 Day Nuclide Decay Decay (Curies) (Curies)

Co-58 0.0000E+00 0.0000E+00 Co-60 0.0000E+00 0.0000E+00 Kr-85 0.6563E+04 0.6493E+04 Kr-85m 0.0000E+00 0.0000E+00 Kr-87 0.0000E+00 0.0000E+00 Kr-88 0.0000E+00 0.0000E+00 Rb-86 0.3205E+03 0.3450E+02 Sr-89 0.5045E+06 0.2213E+06 Sr-90 0.5178E+05 0.5157E+05 Sr-91 0.1431E-16 0.0000E+00 Sr-92 0.0000E+00 0.0000E+00 Y-90 0.5184E+05 0.5159E+05 Y-91 0.6805E+06 0.3344E+06 Y-92 0.0000E+00 0.0000E+00 Y-93 0.3964E-15 0.0000E+00 Zr-95 0.8884E+06 0.4637E+06 Zr-97 0.1774E-06 0.0000E+00 Nb-95 0.1146E+07 0.7766E+06 Mo-99 0.6714E+03 0.1816E-03 Tc-99m 0.6469E+03 0.1749E-03 Ru-103 0.6102E+06 0.2118E+06 Ru-105 0.0000E+00 0.0000E+00 Ru-106 0.2925E+06 0.2611E+06 Rh-105 0.5545E+00 0.3056E-12 Sb-127 0.3350E+03 0.6807E-02 Sb-129 0.0000E+00 0.0000E+00 Te-127 0.8360E+04 0.5490E+04 Te-127m 0.8207E+04 0.5606E+04 Te-129 0.1138E+05 0.3301E+04 Te-129m 0.1748E+05 0.5071E+04 Te-131m 0.5959E-02 0.2118E-16 Te-132 0.1664E+04 0.4757E-02

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-2 Revision 28 Page 2 of 4 Table 14.11-2 Spent Fuel Cask Drop Radiological Analysis - Source Terms*

30 Day 90 Day Nuclide Decay Decay (Curies) (Curies)

I-131 0.5355E+05 0.3038E+03 I-132 0.1715E+04 0.4902E-02 I-133 0.5526E-04 0.7976E-25 I-134 0.0000E+00 0.0000E+00 I-135 0.0000E+00 0.0000E+00 Xe-133 0.3313E+05 0.1194E+02 Xe-135 0.0000E+00 0.0000E+00 Cs-134 0.1016E+06 0.9612E+05 Cs-136 0.6310E+04 0.2638E+03 Cs-137 0.6483E+05 0.6459E+05 Ba-139 0.0000E+00 0.0000E+00 Ba-140 0.2450E+06 0.9480E+04 La-140 0.2819E+06 0.1091E+05 La-141 0.0000E+00 0.0000E+00 La-142 0.0000E+00 0.0000E+00 Ce-141 0.6297E+06 0.1752E+06 Ce-143 0.3033E+00 0.2220E-13 Ce-144 0.8358E+06 0.7222E+06 Pr-143 0.2630E+06 0.1226E+05 Nd-147 0.7146E+05 0.1663E+04 Np-239 0.2042E+04 0.1581E+02 Pu-238 0.2174E+04 0.2199E+04 Pu-239 0.2968E+03 0.2968E+03 Pu-240 0.2972E+03 0.2972E+03 Pu-241 0.8941E+05 0.8870E+05 Am-241 0.1070E+03 0.1260E+03 Cm-242 0.2428E+05 0.1882E+05 Cm-244 0.2743E+04 0.2726E+04 I-130 0.5151E-13 0.0000E+00 Kr-83m 0.0000E+00 0.0000E+00 Xe-138 0.0000E+00 0.0000E+00 Xe-131m 0.2962E+04 0.1204E+03 Xe-133m 0.5377E+01 0.3038E-07 Xe-135m 0.0000E+00 0.0000E+00

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-2 Revision 28 Page 3 of 4 Table 14.11-2 Spent Fuel Cask Drop Radiological Analysis - Source Terms*

30 Day 90 Day Nuclide Decay Decay (Curies) (Curies)

Cs-138 0.0000E+00 0.0000E+00 Cs-134m 0.0000E+00 0.0000E+00 Rb-88 0.0000E+00 0.0000E+00 Rb-89 0.0000E+00 0.0000E+00 Sb-124 0.5599E+03 0.2806E+03 Sb-125 0.8790E+04 0.8446E+04 Sb-126 0.1278E+03 0.4525E+01 Te-131 0.1341E-02 0.4765E-17 Te-133 0.0000E+00 0.0000E+00 Te-134 0.0000E+00 0.0000E+00 Te-125m 0.1902E+04 0.1940E+04 Te-133m 0.0000E+00 0.0000E+00 Ba-141 0.0000E+00 0.0000E+00 Ba-137m 0.6132E+05 0.6110E+05 Pd-109 0.1499E-10 0.0000E+00 Rh-106 0.2925E+06 0.2611E+06 Rh-103m 0.5502E+06 0.1908E+06 Tc-101 0.0000E+00 0.0000E+00 Eu-154 0.6312E+04 0.6229E+04 Eu-155 0.4225E+04 0.4129E+04 Eu-156 0.2550E+05 0.1649E+04 La-143 0.0000E+00 0.0000E+00 Nb-97 0.1911E-06 0.0000E+00 Nb-95m 0.6589E+04 0.3441E+04 Pm-147 0.1206E+06 0.1162E+06 Pm-148 0.4604E+04 0.3221E+03 Pm-149 0.3193E+02 0.2179E-06 Pm-151 0.3097E-02 0.1665E-17 Pm-148m 0.1557E+05 0.5685E+04 Pr-144 0.8358E+06 0.7222E+06 Pr-144m 0.1003E+05 0.8666E+04 Sm-153 0.5430E+01 0.2827E-08 Y-94 0.0000E+00 0.0000E+00 Y-95 0.0000E+00 0.0000E+00

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-2 Revision 28 Page 4 of 4 Table 14.11-2 Spent Fuel Cask Drop Radiological Analysis - Source Terms*

30 Day 90 Day Nuclide Decay Decay (Curies) (Curies)

Y-91m 0.9094E-17 0.0000E+00 Br-82 0.1919E-02 0.1012E-14 Br-83 0.0000E+00 0.0000E+00 Br-84 0.0000E+00 0.0000E+00 Am-242 0.1237E+02 0.1236E+02 Np-238 0.1195E+02 0.6211E-01 Pu-243 0.2743E-06 0.2743E-06

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.11-3 Revision 32 Page 1 of 1 Spent Fuel Cask Drop 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 w/o Number of Fuel Assemblies Damaged 73 Cases 1 & 2 - 30 days Delay Before Cask Drop Case 3 - 90 days Source Terms See Table 14.11-2 Water Level Above Damaged Fuel Assembly 23.4 feet minimum Elemental - 285 Iodine Decontamination Factors Organic - 1 Overall - 200 Noble Gas Decontamination Factor 1 Elemental - 99.85%

Chemical Form of Iodine In Pool Organic - 0.15%

Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Cases 1 & 2 - 0 seconds Time of Control Room Ventilation System Isolation Case 3 -

Cases 1 & 2 - 0 seconds Time of Control Room Filtered Makeup Flow Case 3 -

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 14.12-1 Revision 23 Page 1 of 1 EVENT

SUMMARY

FOR LOSS OF LOAD Primary Side Pressurization EVENT VALUE TIME (Sec)

Turbine Trip 0.00 Reactor scramed on high pressure trip 7.0 Pressurizer safety valves open 9.7 Peak PCS pressure 2626.9 psia 10.2 Pressurizer safety valves close 11.1 Lowest bank of main steam line 11.2 safety valves opened Secondary Side Pressurization EVENT VALUE TIME (Sec)

Turbine Trip 0.0 Loss of Main Feedwater 0.1 Steam Line Safety Valves Open 3.2 Reactor Scram (High Pressurizer Pressure) 7.6 Peak Power Level 108% 8.1 Pressurizer PORVs Open 9.0 Peak Secondary Side Pressure 1063.25 psia 13.7

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.13-1 Revision 33 Page 1 of 1 INITIAL CONDITIONS FOR THE LOSS OF NORMAL FEEDWATER ANALYSIS Parameter Value Core Power, MWt (with uncertainty) 2580.6 Core Inlet Temperature, F 544 Pressurizer Pressure, psia 2060 Pressurizer Liquid Level, % 62 PCS Loop Flow Rate, gpm 341,400 Primary Coolant Pump Heat, MW 16.3 Steam Generator Pressure, psia 770.7 Steam Generator Liquid Level, % Narrow Range 63.9 Steam Generator (A) Secondary Total Mass, lbm 133,961 Steam Generator (B) Secondary Total Mass, lbm 133,971 Main Steam Flow, lbm/hr 11.354 x 106 MFW Temperature, F 440.7 Steam Generator Blowdown per Steam Generator, lbm/hr 30,000

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.13-2 Revision 33 Page 1 of 1 SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEEDWATER FLOW ANALYSIS WITH OFFSITE POWER AVAILABLE AND STEAM DUMP SYSTEM DISABLED TIME EVENT (sec) 0 Total loss of main feedwater 22.8 Auxiliary feedwater actuation signal on low steam generator water level 23.6 Reactor trip signal on low steam generator water level 23.7 Main turbine trip 24.1 Control rods begin to drop 26.0 Early maximum post-trip PCS average temperature (571.5F) 28.0 Early maximum pressurizer level (65.7%)

142.8 Motor-driven AFW pump starts (120 seconds after AFAS) 1200.0 AFW flow controller increased to maximum setting by operator 1500.0 PCPs tripped by operator 3120.0 Late maximum PCS average temperature (570.2oF) 3148.0 Late maximum pressurizer level (67.2%)

4238.0 Minimum steam generator liquid inventory occurs in Steam Generator A (8,515 lbm) 8000.0 End of calculation

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.13-3 Revision 33 Page 1 of 1 SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEEDWATER FLOW ANALYSIS WITH OFFSITE POWER AVAILABLE AND STEAM DUMP SYSTEM AVAILABLE TIME EVENT (sec) 0 Total loss of main feedwater 22.8 Auxiliary feedwater actuation signal on low steam generator water level 23.6 Reactor trip signal on low steam generator water level 23.7 Main turbine trip 24.1 Control rods begin to drop 22.0 Maximum post-trip PCS average temperature (571.4F) 26.0 Maximum pressurizer level (64.7%)

142.8 Motor-driven AFW pump starts (120 seconds after AFAS) 1200.0 AFW flow controller increased to maximum setting by operator 1500.0 PCPs tripped by operator 1712.0 Minimum steam generator liquid inventory occurs in Steam Generator A (8,268.4 lbm) 5000.0 End of calculation

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.13-4 Revision 33 Page 1 of 1 SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEEDWATER FLOW ANALYSIS WITHOUT OFFSITE POWER AVAILABLE AND STEAM DUMP SYSTEM DISABLED TIME EVENT (sec) 0 Total loss of main feedwater 22.8 Auxiliary feedwater actuation signal on low steam generator water level 23.6 Reactor trip signal on low steam generator water level 23.7 Main turbine trip 24.1 Control rods begin to drop 26.0 Maximum post-trip PCS average temperature (571.5F) 28.0 Maximum pressurizer level (65.6%)

34.6 PCPs automatically tripped (11 seconds after reactor trip) 142.8 Motor-driven AFW pump starts (120 seconds after AFAS) 7342.0 Minimum steam generator liquid inventory occurs in Steam Generator A (10,618 lbm) 12000.0 End of calculation

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.13-5 Revision 33 Page 1 of 1 RESULTS

SUMMARY

FOR LOSS OF NORMAL FEEDWATER CASE MINIMUM STEAM MAXIMUM GENERATOR LIQUID MASS PRESSURIZER LEVEL (lbm) (%)

SG A / SG B / Total Off-site Power Available 8,515.0 / 8,549.0 / 17,064.0 67.2 (Steam Dump System Disabled)

Offsite Power Available 8,268.0 / 8,374.0 / 16,642.0 64.7 (Steam Dump System Available)

Offsite Power Unavailable 10,618.0 / 10,644.0 / 21,262.0 65.6 (Steam Dump System Disabled)

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-1 Revision 24 Page 1 of 3 MAIN STEAM LINE BREAK INPUT PARAMETERS AND ASSUMPTIONS Parameter Biasing Value Break location Limiting Main steam line, at steam generator outlet Break flow area (choked at Maximum 2 steam generator integral flow 18 5 in 4 8 restrictor)

Break flow model Conservative Moody critical flow (at steam generator integral flow restrictor)

Break fluid conditions Conservative Steam-only restriction Steam generator tube plugging Minimum No plugged tubes Full-power moderator Most-negative analysis limit -35 pcm/ F temperature coefficient Scram worth Cycle 17 minimum analysis value 5022 pcm for full-power cases Technical Specification minimum 2000 pcm shutdown margin for hot zero power cases Shutdown control rod Maximum peaking Most reactive control rod stuck configuration out of the core Low Steam Generator Pressure Minimum analysis value 485.0 psia Main Steam Isolation Signal and scram setpoint Containment High Pressure Main Conservative Not credited Steam Isolation Signal and scram setpoints Main Steam Isolation Signal Maximum delay 1.0 second signal processing Main Steam Isolation Valve Maximum delay 5.0 seconds after Main Steam closure Isolation Signal Main feedwater termination Maximum delay 22.0 seconds after Main Steam Isolation Signal Main feedwater temperature Minimum Reduced to 32.1 F when turbine trips Main feedwater flow Maximum Increased 20% when break occurs (full-power cases)

Main feedwater delivery Limiting All to affected steam generator when break occurs Auxiliary feedwater actuation Minimum delay When break occurs Auxiliary feedwater temperature Minimum 32.1 F Auxiliary feedwater flow rate Maximum analysis value 200 gpm / Steam Generator

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-1 Revision 24 Page 2 of 3 MAIN STEAM LINE BREAK INPUT PARAMETERS AND ASSUMPTIONS Parameter Biasing Value Auxiliary feedwater delivery Limiting All to affected steam generator Auxiliary feedwater termination Maximum delay for operator 1800.0 seconds after break action occurs Single active failure Worst 1 of 2 High Pressure Safety Injection pumps required to be in service fails Initial boron concentration in Minimum 0 ppm safety injection lines 3

Total safety injection line purge Maximum 78 ft volume Safety Injection and Refueling Technical Specification Minimum 1720 ppm Water Tank boron concentration Safety Injection and Refueling Technical Specification Minimum 40.0 F Water Tank temperature Low Pressurizer Pressure Safety Minimum analysis value 1450.0 psia Injection Signal setpoint Safety injection actuation (if Maximum delay 30.0 seconds after Safety primary coolant system pressure Injection Signal with offsite power is less than High Pressure Safety available Injection Pump shutoff head) 40.0 seconds after Safety Injection Signal with loss of offsite power High Pressure Safety Injection Minimum Degraded flow curve for single pump flow pump High Pressure Safety Injection Analysis value 1200.7 psia pump shutoff head Turbine control valves position Limiting Increased to fully open when break occurs (full-power case)

Scram signal processing Maximum delay 0.8 second Turbine trip Maximum delay At scram signal Initial reactor power Bounding 2565.4 MW at full power 1.0 W at hot zero power Initial pressurizer pressure Nominal 2060.0 psia Initial pressurizer level Programmed 57.0% of span at full power 42% of span at hot zero power Initial primary coolant system Maximum at full power 544.0 F cold leg temperature 532.0 F

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-1 Revision 24 Page 3 of 3 MAIN STEAM LINE BREAK INPUT PARAMETERS AND ASSUMPTIONS Parameter Biasing Value Initial primary coolant system Technical Specification minimum 341,400 gpm total flow rate value minimum measurement uncertainty Core bypass flow rate Maximum analysis value 3% of primary coolant system total Initial reactor vessel upper head Analysis assumption Close to initial primary coolant temperature system hot leg temperature Initial steam generator inventory Analysis value 141,065 lbm / steam generator at full power 210,759 lbm / steam generator at hot zero power Initial steam generator pressure Analysis value 820.4 psia at full power 889.1 psia at hot zero power

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-2 Revision 24 Page 1 of 1 STEAM LINE BREAK SEQUENCE OF EVENTS DURING LHR-LIMITING TRANSIENT (HZP, OFFSITE POWER AVAILABLE)

Event Time After Break (sec)

Reactor is at EOC all-rods-in-except-most-reactive-rod HZP condition. 0.0 Double-ended guillotine break in main steam line at steam generator outlet occurs.

Low Steam Generator Pressure ESF signal (485 psia) initiates steam 13.8 generator isolation.

MSIVs are fully closed (6.0 sec delay). 19.8 Low Pressurizer Pressure ESF signal (1450 psia) initiates SI. 20.4 Shutdown worth has been fully overcome by moderator and Doppler 32.2 feedback (0.0 $ total reactivity).

Credited HPSI pump is running at rated speed (30.0 sec delay) and 50.4 begins filling SI lines with borated water.

Peak post-scram power (26.92% of rated) and peak LHR occur. 158.0 Borated water has filled SI lines and begins to enter PCS cold legs. 158.7 Borated water front has passed through core. Power begins to 160.0 decrease noticeably.

Affected steam generator begins to dry out. 243.0 Affected-sector core inlet temperature begins to increase. Power 248.0 begins to drop to decay-heat level.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-3 Revision 24 Page 1 of 1 OVERALL CORE CONDITIONS AT TIME OF PEAK LHR Parameter Value Time, sec 158.0 Reactor Power, MW 690.70 Core Outlet Pressure, psia 863.16 Core Inlet Temperature, °F Affected Sector 380.21 Unaffected Sector 489.31 Core Inlet Flow Rate, lbm/sec Stuck Rod Region 5,489.9 Rest of Affected Sector 14,564 Unaffected Sector 18,316

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-4 Revision 24 Page 1 of 1 STEAM LINE BREAK SEQUENCE OF EVENTS DURING DNBR-LIMITING TRANSIENT (HZP, LOSS OF OFFSITE POWER)

Event Time After Break (sec)

Reactor is at EOC all-rods-in-except-most-reactive-rod HZP condition. 0.0 Double-ended guillotine break in main steam line at steam generator outlet occurs. Offsite power is lost.

Low Steam Generator Pressure ESF signal (485 psia) initiates steam 12.4 generator isolation.

MSIVs are fully closed (6.0 sec delay). 18.4 Low Pressurizer Pressure ESF signal (1450 psia) initiates SI. 26.3 Shutdown worth has been fully overcome by moderator and Doppler 49.7 feedback (0.0 $ total reactivity).

Credited HPSI pump is running at rated speed (40.0 sec delay) and 66.3 begins filling SI lines with borated water.

Borated water has filled SI lines and begins to enter PCS cold legs. 170.1 Borated water front has passed through core. 172.0 Peak post-scram power (13.23% of rated) and MDNBR occur. Power 190.0 begins to decrease.

Affected steam generator begins to dry out. 517.0 Affected-sector core inlet temperature begins to increase. Power 592.0 begins to drop to decay-heat level.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-5 Revision 24 Page 1 of 1 OVERALL CORE CONDITIONS AT TIME OF MDNBR Parameter Value Time, sec 190.0 Reactor Power, MW 339.35 Core Outlet Pressure, psia 821.76 Core Inlet Temperature, °F Affected Sector 314.56 Unaffected Sector 466.81 Core Inlet Flow Rate, lbm/sec Stuck Rod Region 768.61 Rest of Affected Sector 992.17 Unaffected Sector 661.63

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-6 Revision 28 Page 1 of 1 Main Steam Line Break (MSLB) Radiological Analysis - Inputs and Assumptions Input/Assumption Value Core Power Level 2703 MWth Core Average Burnup 39,300 MWD/MTU Radial Peaking Factor 2.04 0.5% DNB Fuel Damage 0% Fuel Centerline Melt Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdown cooling and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> terminate steam release Time for PCS to reach 212oF and terminate 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SG tube leakage PCS Mass 432,977 lbm Maximum (Hot Zero Power) - 210,759 lbm (used for faulted SG to maximize release)

SG Secondary Side Mass Minimum (Hot Full Power) - 141,065 lbm (used for intact SG to maximize concentration)

Release from Faulted SG Instantaneous Steam Release from Intact SGs Table 14.14-7 Secondary Coolant Iodine Activity prior to 0.1 Ci/gm DE I-131 accident Steam Generator Secondary Side Partition Faulted SG - none Coefficients Intact SGs - 100 Atmospheric Dispersion Factors Offsite Section 2.5.5.2 Onsite Tables 14.24-2 and 14.24-3 Control Room Ventilation System Table 14.24-1 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 Factors RG 1.183 Section 4.2.6

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.14-7 Revision 28 Page 1 of 1 MAIN STEAM LINE BREAK (MSLB) RADIOLOGICAL ANALYSIS - INTACT SG STEAM RELEASE RATE Time Intact SG Steam Release (hours) (lbm) 0-8 800,000 8 - 720.0 0

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-1 Revision 21 Page 1 of 1 INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER PARAMETER ASSUMED VALUE Initial core power level, MWt 2600.6 Core inlet coolant temperature, °F 550.65 Core mass flow rate, 106 lbm/hr 138.*

Reactor coolant system pressure, psia 2,110 Steam generator pressure, psia 770 Initial pressurizer liquid volume, ft 3 800 Steam generator level, ft above tube sheet 31.74

• Lower core flowrate dispositioned in Reference 6.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-2 Revision 21 Page 1 of 1 SETPOINTS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER Parameter Setpoint Steam generator MSSV setpoint, psia 1000 AFW actuation on steam generator level AFAS signal generation, % NR 23.7 SIAS setpoint, psia 1605 Shutdown cooling entry conditions:

Hot leg temperature, °F 300 Pressurizer pressure, psia 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 3 558.6 703.7 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, lbm/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, °F 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, °F 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 1 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, lbm 183,202 605,101 Integrated Steam release, lbm

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 1 STEAM GENERATOR TUBE RUPTURE (SGTR) RADIOLOGICAL ANALYSIS

- INPUTS AND ASSUMPTIONS Input/Assumption Value Core Power Level 2703 MWth 1.0 Ci/gm DE I-131 and 100/E-bar gross Initial PCS Equilibrium Activity activity Initial Secondary Side Equilibrium Iodine Activity 0.1 Ci/gm DE I-131 Maximum pre-accident spike iodine concentration 40 Ci/gm DE I-131 Maximum equilibrium iodine concentration 1.0 Ci/gm DE I-131 Duration of accident-initiated spike 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdown cooling and terminate 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> steam release 529,706 lbm for pre-accident iodine spike PCS Mass case 459,445 lbm for concurrent iodine spike case 141,065 lbm 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 Ci/gm DE I-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 1 of 1 SGTR RADIOLOGICAL ANALYSIS - INTEGRATED MASS RELEASES (1)

Break Flow in Steam Release from Time Steam Release from Ruptured SG Ruptured SG Unaffected SG (hours) (lbm)

(lbm) (lbm) 0 - 0.196417 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 - 3.638889 64,773 75,485.6 145,983.5 3.638889 - 8 357,126 200,868.4 388,464.5 8 - 720 0 0 0 (1)

Flowrate assumed to be constant within time period

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-7 Revision 28 Page 1 of 1 SGTR RADIOLOGICAL ANALYSIS - FLASHING FRACTION FOR FLOW FROM BROKEN TUBE Time Flashing Fraction (seconds) 0 0.110 707.1 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 1 SGTR RADIOLOGICAL ANALYSIS - 40 CI/GM D.E. I-131 ACTIVITIES Activity Isotope (Ci/gm)

Iodine-131 33.2194 Iodine-132 7.6660 Iodine-133 34.4971 Iodine-134 3.0025 Iodine-135 14.6932

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-9 Revision 28 Page 1 of 1 SGTR RADIOLOGICAL ANALYSIS - IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS Input Assumption Value Maximum Letdown Flow 40 gpm Assumed Letdown Flow

• 44 gpm at 120F, 2060 psia Maximum Identified PCS Leakage 10 gpm Maximum Unidentified PCS Leakage 1 gpm PCS Mass 459,445 lbm I-131 Decay Constant 5.986968E-5 min-1 I-132 Decay Constant 0.005023 min-1 I-133 Decay Constant 0.000555 min-1 I-134 Decay Constant 0.013178 min-1 I-135 Decay Constant 0.001748 min-1

• maximum letdown flow plus 10% uncertainty

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-10 Revision 28 Page 1 of 1 SGTR RADIOLOGICAL ANALYSIS - CONCURRENT (335 X) IODINE SPIKE APPEARANCE RATE Appearance Rate Time of Depletion Isotope (Ci/min) (hours)

Iodine-131 58.0966961 >8 Iodine-132 79.8319317 >8 Iodine-133 90.1310904 >8 Iodine-134 74.0318685 >8 Iodine-135 68.9790622 >8

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.15-11 Revision 28 Page 1 of 1 SGTR RADIOLOGICAL ANALYSIS - AFFECTED STEAM GENERATOR WATER LEVEL AND DECONTAMINATION FACTORS FOR FLASHED FLOW Time Water Level Above U-Tubes Calculated Decontamination Factor (seconds) (feet) Decontamination Factor Used in Analysis 0 0.0 (assumed)* 1.0 1.0 707.1 0.0 (assumed)* 1.0 1.0 736 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 9.43 38.01867 1.467378 13100 12.34 553073.5 58.16008 28800 15.16 58.16008 58.16008 It is conservatively assumed that no scrubbing occurs until after the reactor trip at 707.1 seconds. Since the U-tubes remain covered throughout the event, it is also conservatively assumed that at the time of trip the water level is just above the top of the U-tubes. The time-dependent water level after the trip is a function of the allowable primary to secondary leakage, broken tube flow, and MSSV/ADV releases from the affected steam generator. To minimize the water level available for scrubbing, the location of the tube break is assumed to be at the top of the U-tubes.

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

SUMMARY

FOR THE EOC HZP CONTROL ROD EJECTION EVENT VALUE TIME (sec)

Ejection of a Single Control Rod ------- 0.0 Core Power Reached VHP Trip Setpoint 36.86% RTP 0.309 Core Power Peaked 1,903%RTP 0.410 Core Average Rod Surface Heat Flux Peaked 101.9% RTP 0.507 Minimum DNBR Occurred see Table 14.1-5 0.507 Scram Rod Insertion Begins ------- 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 1.0 Ci/gm DE I-131 and 100/E-bar Initial PCS Equilibrium Activity gross activity Initial Secondary Side Equilibrium Iodine 0.1 Ci/gm DE I-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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PCS Mass 432,976.8 lbm minimum - 141,065 lbm (per SG)

Minimum mass used for SGs to SG Secondary Side Mass maximize steam release nuclide concentration.

Particulate - 95%

Chemical Form of Iodine Released to Elemental - 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 Input/Assumption Value 3

Containment Volume 1.64E+06 ft Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.10% (by weight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.05% (by weight)/day Aerosols - 0.1 hr-1 Containment Natural Deposition Coefficients Elemental Iodine - 1.3 hr-1 Organic Iodine - None

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.16-3 Revision 28 Page 1 of 1 CONTROL ROD EJECTION RADIOLOGICAL ANALYSIS - STEAM RELEASE SG Steam Release Time (lbm) 0 - 1100 sec 107,158.8 1100 sec - 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 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 34 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) Gd2O3 concentrations 2, 4, 6 and 8 w/o 1.2 RCS a) Flow resistance Analysis considers plant-specific form and friction losses Analysis assumes location giving most limiting PCT b) Pressurizer location (broken loop) c) Hot assembly location Anywhere in core d) Hot assembly type 15x15 Framatome 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/ft1 c) FrT 2.042 2.2 Fluid Conditions a) Loop flow 130 Mlbm/hr M 145 Mlbm/hr b) PCS inlet core temperature 537 T 544 F3 c) Upper head temperature < core outlet temperature d) Pressurizer pressure 2,010 P 2,100 psia4 e) Pressurizer liquid level 46.25% L 67.8%

f) SIT pressure 214.7 P 239.7 psia g) SIT liquid volume 1,040 V 1,176 ft3 h) SIT temperature 80 T 140 F (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 uncertainty.

2 Includes a 4.25% measurement uncertainty.

3 Sampled range of +7 F 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 34 Page 2 of 2 PLANT OPERATING RANGE SUPPORTED BY THE LOCA ANALYSIS Event Operating Range 3.0 Accident Boundary Conditions a) Break location Cold leg pump discharge piping b) Break 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) Worst single-failure Loss of one ECCS pumped injection train e) Offsite power On or Off f) LPSI flow Minimum flow g) HPSI flow Minimum flow h) ECCS pumped injection temperature 100 F 30 (w/ offsite power) i) HPSI delay time 40 seconds (w/o offsite power) 30 (w/ offsite power) j) LPSI delay time 40 seconds (w/o offsite power) k) Containment pressure 14.7 psia, nominal value l) Containment temperature 80 T 140 F m) Containment spray/fan cooler delays 0/0 seconds

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.17.1-3 Revision 28 Page 1 of 1 STATISTICAL DISTRIBUTION USED FOR PROCESS PARAMETERS Operational Parameter Uncertainty Parameter Range Distribution Core Power Operation (%) Uniform 100.0 - 100.5 Pressurizer Pressure (psia) Uniform 2,010 - 2,100 Pressurizer Liquid Level (%) Uniform 46.25 - 67.8 3

SIT Liquid Volume (ft ) Uniform 1,040 - 1,176 SIT Pressure (psia) Uniform 214.7 - 239.7 Containment/SIT Temperature (°F) Uniform 80 - 140 1 6 3 Containment Volume (x10 ft ) Uniform 1.64 - 1.80 Initial Flow Rate (Mlbm/hr) Uniform 130 - 145 Initial Operating Temperature (°F) Uniform 537 - 544 SIRWT Temperature (°F) Point 100 Offsite Power Availability2 Binary 0,1 Delay for Containment Sprays (s) Point 0 Delay for Containment Fan Coolers (s) Point 0 30 (w/ offsite power)

HPSI Delay (s) Point 40 (w/o offsite power) 30 (w/ offsite power)

LPSI Delay (s) Point 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 (Reference 5).

2 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 1

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, kW/ft 14.60 T

Total Hot Rod Radial Peak (Fr ) 2.040 Axial Shape Index (ASI) 0.1602 Break Type Guillotine Break Size (ft2/side) 3.339 Offsite Power Availability Not Available Decay Heat Multiplier 1.01073

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.17.1-5 Revision 34 Page 1 of 1

SUMMARY

OF HOT ROD LIMITING PCT RESULTS 15 x 15 Framatome Fuel Type w/o Gd2O3 Case Number 22 PCT Temperature 1,740 F Time 27.2 s Elevation 2.151 ft Metal-Water Reaction Oxidation Maximum 0.59%

Total Oxidation < 0.01%

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.17.1-6 Revision 28 Page 1 of 1 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 1 CONTAINMENT HEAT SINK DATA 2 Total Material Heat Sink Surface Area (ft )

thickness (ft)

Containment Dome and Upper Carbon steel liner; no 1 0.0208 Wall 69,630.20 coatings 4.2625 Concrete; no coating Carbon steel liner; no 2 Containment Wainscot 0.0208 2,200.20 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 1.5 Concrete; no coating 380.10 0.0208 Carbon steel; no paint 28.3 Concrete; no coating 5 Reactor Cavity Slab 0.0208 Stainless steel 380.10 1.4792 Concrete; no coating 243.4 (Inner 6 Lower Biological Shield 0.015625 Stainless steel; no paint surface of cylindrical shape) 7.9167 Concrete; no coating Internal Concrete with Carbon 7 0.0208 Carbon steel Steel Liner Plate 2,048.40 3.8958 Concrete; no coating Internal Concrete with Stainless 8 0.0417 Stainless steel Steel Liner Plate 4,712.70 2.4083 Concrete; no coating Carbon steel liner; no 9 Internal Concrete with Decking 0.004 2,672.90 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 Equipment Tanks and Heat 12 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 1 CONTAINMENT INITIAL AND BOUNDARY CONDITIONS Parameter Parameter Value Containment free volume range, ft3 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 1 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, °F 544 SIT Pressure, psia 215 SIT Fluid Temperature, °F 100 Steam Generator Tube Plugging, % 15 SG Secondary Pressure, psia 763 SG Main Feedwater Temperature, °F 439.5 SG Auxiliary Feedwater Temperature, °F 120 HPSI Fluid Temperature, °F 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 1 PCT RESULTS OF THE PALISADES SBLOCA ANALYSIS Break Size (ft2) PCT (°F) 0.04 1296 0.05 1451 0.06 1479 0.08 1734 0.10 1654 0.15 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 64.86 Cold Leg pressure reaches HPSI shutoff head (1200.7 psia) 96 Motor-driven AFW delivery begins 145 Loop seal in Cold Leg 1B 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 (F) 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 Failure Mode and Failure Allowable Calculated Component Loading Condition Location of Failure Condition(a) Condition(b) Condition Core Barrel Tension - Axial Middle Section of 54,000 psi 29,300 psi 3,200 psi Load Core Barrel Buckling - External Upper Portion of p = 572 psi p = 381 psi p = 380 psi Pressure Core Barrel (Arch)

Tension - Internal Middle Section of 54,000 psi 29,300 psi 26,750 psi Pressure Core Barrel Lower Core Support Bending - Transverse Beam Flange 54,000 psi 43,950 psi 22,510 psi Load Shear - Transverse Junction of Flange 32,400 psi 17,580 psi 7,710 psi Load to Web Control Rod Shrouds Bending - Axial and Lower End of Shroud 54,000 psi 32,230 psi 70,310 psi 1st Row Transverse Load (Near Nozzle)

Deformation - Axial Center of Shroud Defl = 0.76" Defl = 0.51" Defl > 0.51" and Transverse Load (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 Failure Mode and Failure Allowable Calculated Component Loading Condition Location of Failure Condition(a) Condition(b) Condition Control Rod Shrouds Bending - Axial and Lower End of Shroud 54,000 psi 32,230 psi 28,090 psi 2nd Row Transverse Load Deformation - Axial Center of Shroud Defl = 0.76" Defl = 0.51" Defl - 0.279" and Transverse Load Upper Grid Beam Bending - Transverse Center of Beam 54,000 psi 43,950 psi 12,980 psi Load Upper Structure Bending - Axial Junction of Flange 54,000 psi 43,950 psi 40,630 psi Flange Load and Barrel Cylinder (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 1 ASYMMETRIC LOADS ANALYSIS - REACTOR VESSEL INTERNAL COMPONENT STRESS MARGINS Component Location Percent Margin (%)*

Core Support Barrel Upper Flange 6 Upper Cylinder 7 Center Cylinder 11 Lower Support Structure Support Columns 2 Beams 3 Core Support Plate 13 Upper Guide Structure Grid Beams 1

• Percent margin is computed as (Sallow - Scalc) (100%) / Sallow, where Scalc 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 (ft2)

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 SINKS/SOURCES HEAT SINK SURFACE AREA (ft2)

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 SINKS/SOURCES HEAT SINK SURFACE AREA (ft2)

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 Page 4 of 5 LOCA ANALYSIS CONTAINMENT BUILDING HEAT SINKS/SOURCES HEAT SINK SURFACE AREA (ft2)

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

28. PCS Metal Wall #1 35,539.2 Reactor Vessel and Internals

29. PCS Metal Wall #2 12,441.8 Reactor Vessel and Internals

30. PCS Metal Wall #3 10,378.8 Reactor Core

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.1-1 Revision 27 Page 5 of 5 LOCA 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 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 Page 1 of 1 LOCA ANALYSIS ENGINEERED SAFEGUARDS EQUIPMENT ALIGNMENT D/G 1-2 D/G 1-1 Failure Failure Equipment Equipment Operated Operated Containment Sprays P-54B & P54C P-54A LPSI P-67B P-67A HPSI P-66B P-66A Containment Air Coolers VHX-1, VHX-2 & VHX-3 Component Cooling Water P-52A & P-52C P-52B Service Water P-7B P-7A & P-7C

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.1-3 Revision 25 Page 1 of 1 LOCA INITIAL CONDITIONS Containment Free Volume 1.64 x 106 ft3 Containment Temperature 145°F Containment Pressure 15.7 Psia Relative Humidity 30%

SIRW Tank Temperature 100°F

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.1-4 Revision 29 Page 1 of 1 CONTAINMENT BUILDING RESPONSE TO LOCA DOUBLE ENDED GUILLOTINE BREAK IN A HOT LEG Peak Pressure Time Case (Psig) (Sec)

D/G 1-2 Failure 54.2 13.2 D/G 1-1 Failure 54.2 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 1 LOCA ANALYSIS PARAMETER ASSUMPTIONS Initial Containment Air Temp 145°F Initial Containment Pressure 15.7 Psia (1.0 Psig)

Relative Humidity 30%

CCWHX Tube Fouling Coefficient .001 hr-ft2-°F/BTU Service Water Temperature 85°F D/G 1-2 (RCF) Failure Data ECCS Injection Flow pre-RAS (1 HPSI, 1 LPSI pump) 3,471 gpm ECCS Injection Flow post-RAS (1 HPSI pump) 705 gpm 1 SW Pump Flow Rate to CCWHXs 4,214 gpm 1 CCW Pump Flow Rate to SDCHXs 4,480 gpm 2 CS Pump Flow Rate to Containment (pre RAS) 2,472 gpm 2 CS Pump Flow Rate to Containment (post RAS-HLI) 1,684 gpm Post-RAS Spillage after Initiation of Hot Leg Injection 328 gpm ECCS Injection Flow after Initiation of Hot Leg Injection 273 gpm D/G 1-1 (LCF) Failure Data ECCS Injection Flow pre-RAS (1 HPSI, 1 LPSI pump) 3,443 gpm ECCS Injection Flow post-RAS (1 HPSI pump) 703 gpm 2 SW Pump Flow Rate to CCWHXs 4,286 gpm 2 SW Pump Flow Rate to 3 Containment Air Coolers 1, 600 gpm/Air Cooler 1 CCW Pump Flow Rate to SDCHXs 4,480 gpm 1 CS Pump Flow Rate to Containment (pre RAS) 1,781 gpm 1 CS Pump Flow Rate to Containment, 1 header (pre RAS) 1,233 gpm 1 CS Pump Flow Rate to Containment (post RAS-HLI) 788 gpm Post-RAS Spillage after Initiation of Hot Leg Injection 308 gpm ECCS Injection Flow after Initiation of Hot Leg Injection 279 gpm

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.2-1 Revision 29 Page 1 of 1 INITIAL CONDITIONS FOR THE MSLB CONTAINMENT ANALYSIS Assumed Parameter Value Containment Free Volume, ft3 1.64 x 106 Initial Containment Temperature, °F 145.0 Initial Containment Pressure, psig 1.0*

Initial Containment Humidity, % 30 Containment Spray Water Temperature, °F 100.0 Main Feedwater Regulating Valve Closure Time, sec 22 Main Steam Isolation Valve Closure Time, sec 2

• Zero power cases assumed 1.5 psig

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.2-2 Revision 23 Page 1 of 1 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 2600.6 550.65 770.0 144.6x106 75% 75 1917.5 548.70 784.0 144.6x106 0% 0 20.0 539.00 900.0 144.6x106 EEQ 102 2600.6 550.65 770.0 144.6x106

• This power level includes an assumed contribution of 20 MWTh from the primary coolant pumps.

1. Lower PCS flowrate dispositioned in Reference 25.

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.2-3 Revision 29 Page 1 of 1 MSLB CONTAINMENT ANALYSIS RESULTS Power Peak Pressure Case Description Level (psig)

Limiting Pressure - 0% 53.5 Relay 5P-7 Failure w/Open MSIV Bypass Valves

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 (ft2)

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 (ft2)

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 (ft2)

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 (ft2)

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

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 ft3 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 1 GEOMETRY AND PEAK PRESSURES IN STEAM GENERATOR COMPARTMENTS Steam Generator Volume Vent Area Peak Pressure Compartment (ft3) (ft2) (Psi)

North 55,210 1,043.3 24.8 South 62,090 1,091.3 22.4

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.18.3-3 Revision 28 Page 1 of 1 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 Page 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 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%

Noble gas - n/a

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.19-2 Revision 28 Page 1 of 1 FUEL HANDLING ACCIDENT RADIOLOGICAL ANALYSIS - SOURCE TERM Nuclide Activity Nuclide Activity Nuclide Activity (Curies) (Curies) (Curies)

Co-58 0.0000E+00 I-135 0.8949E+04 Sb-126 0.9900E+03 Co-60 0.0000E+00 Xe-133 0.1298E+07 Te-131 0.8307E+04 Kr-85 0.1052E+05 Xe-135 0.8201E+05 Te-133 0.2034E-10 Kr-85m 0.1174E+03 Cs-134 0.2034E+06 Te-134 0.2217E-14 Kr-87 0.1647E-05 Cs-136 0.5284E+05 Te-125m 0.3417E+04 Kr-88 0.4302E+01 Cs-137 0.1100E+06 Te-133m 0.1213E-09 Rb-86 0.1819E+04 Ba-139 0.4861E-04 Ba-141 0.0000E+00 Sr-89 0.7020E+06 Ba-140 0.1130E+07 Ba-137m 0.1041E+06 Sr-90 0.8456E+05 La-140 0.1235E+07 Pd-109 0.2825E+05 Sr-91 0.2679E+05 La-141 0.2730E+03 Rh-106 0.5771E+06 Sr-92 0.4453E+01 La-142 0.5794E-03 Rh-103m 0.1097E+07 Y-90 0.8623E+05 Ce-141 0.1168E+07 Tc-101 0.0000E+00 Y-91 0.9107E+06 Ce-143 0.4100E+06 Eu-154 0.1246E+05 Y-92 0.3229E+03 Ce-144 0.1014E+07 Eu-155 0.8442E+04 Y-93 0.4137E+05 Pr-143 0.1071E+07 Eu-156 0.1935E+06 Zr-95 0.1210E+07 Nd-147 0.4211E+06 La-143 0.0000E+00 Zr-97 0.1684E+06 Np-239 0.1023E+08 Nb-97 0.1692E+06 Nb-95 0.1248E+07 Pu-238 0.4494E+04 Nb-95m 0.8748E+04 Mo-99 0.8264E+06 Pu-239 0.3578E+03 Pm-147 0.1296E+06 Tc-99m 0.7956E+06 Pu-240 0.5406E+03 Pm-148 0.1659E+06 Ru-103 0.1216E+07 Pu-241 0.1522E+06 Pm-149 0.2481E+06 Ru-105 0.5426E+03 Am-241 0.1897E+03 Pm-151 0.5012E+05 Ru-106 0.5771E+06 Cm-242 0.5649E+05 Pm-148m 0.2899E+05 Rh-105 0.3958E+06 Cm-244 0.1339E+05 Pr-144 0.1015E+07 Sb-127 0.6450E+05 I-130 0.2546E+04 Pr-144m 0.1217E+05 Sb-129 0.1176E+03 Kr-83m 0.3727E+00 Sm-153 0.2171E+06 Te-127 0.7344E+05 Xe-138 0.0000E+00 Y-94 0.0000E+00 Te-127m 0.1222E+05 Xe-131m 0.8276E+04 Y-95 0.0000E+00 Te-129 0.2383E+05 Xe-133m 0.3403E+05 Y-91m 0.1702E+05 Te-129m 0.3637E+05 Xe-135m 0.1434E+04 Br-82 0.2060E+04 Te-131m 0.3690E+05 Cs-138 0.0000E+00 Br-83 0.8833E-01 Te-132 0.6852E+06 Cs-134m 0.5122E+00 Br-84 0.0000E+00 I-131 0.6424E+06 Rb-88 0.4804E+01 Am-242 0.1138E+05 I-132 0.7060E+06 Rb-89 0.0000E+00 Np-238 0.2238E+06 I-133 0.3019E+06 Sb-124 0.1663E+04 Pu-243 0.5681E+03 I-134 0.2087E-09 Sb-125 0.1566E+05

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.22-1 Revision 28 Page 1 of 1 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 " H2O 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 HPSI's, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 Page 1 of 3 MAXIMUM HYPOTHETICAL ACCIDENT / LOSS OF COOLANT ACCIDENT (MHA/LOCA)

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 PCS Equilibrium Activity 1.0 Ci/gm DE I-131 and 100/E-bar gross activity Core Fission Product Inventory Table 14.22-3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.10% (by weight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 Leakage (from 19 minutes to 30 days) 39,054 ft.3 Sump Volume (minimum) 0.053472 ft3/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 / 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 292,143 gallons (minimum valve for ECCS leakage, maximizes sump iodine concentration) 430,708 gallons (maximum value for SIRWT backleakage to be consist with ECCS Leakage to SIRWT (2 times allowed assumption of minimum water level in value) SIRWT)

Flashing Fraction (elemental iodine assumed to be released into tank space 7.2 gpm until 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RAS, then based upon partition factor) 0.0125 gpm SIRWT liquid/vapor elemental iodine partition 0% based on temperature of fluid factor reaching SIRWT Elemental Iodine fraction in SIRWT Table 14.22-9 Table 14.22-8 Initial SIRWT Liquid Inventory (minimum at 4,144 gallons time of recirculation)

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 <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br /> Aerosol 1.8/hour (reduced to 0.18 at 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />)

Time to reach DF of 50 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.22-2 Revision 28 Page 3 of 3 MAXIMUM HYPOTHETICAL ACCIDENT / 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 16 cfm Control Room Unfiltered Inleakage 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 0.0000E+00 Pu-239 0.3558E+05 Co-60 0.0000E+00 Pu-240 0.5406E+05 Kr-85 0.1052E+07 Pu-241 0.1522E+08 Kr-85m 0.1948E+08 Am-241 0.1884E+05 Kr-87 0.3756E+08 Cm-242 0.5669E+07 Kr-88 0.5286E+08 Cm-244 0.5943E+06 Rb-86 0.1959E+06 I-130 0.3743E+07 Sr-89 0.7213E+08 Kr-83m 0.9119E+07 Sr-90 0.8458E+07 Xe-138 0.1211E+09 Sr-91 0.8874E+08 Xe-131m 0.8346E+06 Sr-92 0.9557E+08 Xe-133m 0.4659E+07 Y-90 0.8737E+07 Xe-135m 0.2999E+08 Y-91 0.9264E+08 Cs-138 0.1340E+09 Y-92 0.9596E+08 Cs-134m 0.4920E+07 Y-93 0.1101E+09 Rb-88 0.5369E+08 Zr-95 0.1236E+09 Rb-89 0.6895E+08 Zr-97 0.1206E+09 Sb-124 0.1702E+06 Nb-95 0.1249E+09 Sb-125 0.1567E+07 Mo-99 0.1368E+09 Sb-126 0.1107E+06 Tc-99m 0.1198E+09 Te-131 0.6601E+08 Ru-103 0.1260E+09 Te-133 0.8639E+08 Ru-105 0.9451E+08 Te-134 0.1220E+09 Ru-106 0.5794E+08 Te-125m 0.3413E+06 Rh-105 0.8741E+08 Te-133m 0.5406E+08 Sb-127 0.9111E+07 Ba-141 0.1188E+09 Sb-129 0.2568E+08 Ba-137m 0.1043E+08 Te-127 0.9047E+07 Pd-109 0.3327E+08 Te-127m 0.1223E+07 Rh-106 0.6285E+08 Te-129 0.2528E+08 Rh-103m 0.1135E+09 Te-129m 0.3772E+07 Tc-101 0.1261E+09 Te-131m 0.1113E+08 Eu-154 0.1247E+07 Te-132 0.1048E+09 Eu-155 0.8448E+06 I-131 0.7483E+08 Eu-156 0.2023E+08 I-132 0.1068E+09 La-143 0.1108E+09 I-133 0.1462E+09 Nb-97 0.1216E+09 I-134 0.1602E+09 Nb-95m 0.8835E+06 I-135 0.1372E+09 Pm-147 0.1292E+08 Xe-133 0.1466E+09 Pm-148 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 0.1606E+08 Cs-136 0.5873E+07 Pm-148m 0.2999E+07 Cs-137 0.1100E+08 Pr-144 0.1025E+09 Ba-139 0.1307E+09 Pr-144m 0.1224E+07 Ba-140 0.1260E+09 Sm-153 0.4423E+08 La-140 0.1299E+09 Y-94 0.1105E+09 La-141 0.1193E+09 Y-95 0.1183E+09 La-142 0.1156E+09 Y-91m 0.5151E+08 Ce-141 0.1212E+09 Br-82 0.5282E+06 Ce-143 0.1115E+09 Br-83 0.9102E+07 Ce-144 0.1020E+09 Br-84 0.1591E+08 Pr-143 0.1111E+09 Am-242 0.9062E+07 Nd-147 0.4770E+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 Revision 28 Page 1 of 1 MHA/LOCA RELEASE PHASES Phase Onset Duration Gap Release 30 seconds 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Early In-Vessel 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

• From Regulatory Guide 1.183, Table 4

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.22-5 Revision 28 Page 1 of 1 MHA/LOCA TIME DEPENDENT SIRWT PH Time 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 4.641 432.00 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 1 MHA/LOCA TIME DEPENDENT SIRWT TOTAL IODINE CONCENTRATION Time SIRWT Iodine Concentration (hours) (gm-atom/liter) 0.3167 0.00E+00 0.50 9.60E-07 1.3167 4.82E-06 1.3167 4.82E-06 2.00 4.84E-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 1 MHA/LOCA TIME DEPENDENT SIRWT LIQUID TEMPERATURE Time (hr) Temperature (oF) 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 Page 1 of 1 MHA/LOCA Time Dependent SIRWT Elemental Iodine Fraction Time (hr) Elemental Iodine Fraction 0.3167 0.00E+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 1.00E-01 96.00 1.06E-01 120.00 1.11E-01 144.00 1.15E-01 168.00 1.19E-01 192.00 1.23E-01 240.00 1.29E-01 288.00 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-01 720.00 1.49E-01

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.22-9 Revision 28 Page 1 of 1 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 1 MHA/LOCA ADJUSTED RELEASE RATE FROM SIRWT Time Adjusted Iodine Release Rate (hours) (cfm) 0.3167 9.1718E-04 1.3167 1.1922E-05 8.00 1.2895E-05 24.00 1.4921E-05 72.00 1.7737E-05 168.00 1.9907E-05 240.00 2.1376E-05 336.00 2.2501E-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 1.0 Ci/gm DE I-131 and 100/E-bar gross PCS Equilibrium Activity activity Break Flow Rate 160 gpm Break Temperature 135oF Break Pressure 35 psia Time required to isolate break 60 minutes Maximum equilibrium iodine concentration 1.0 Ci/gm DE I-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 Page 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 Page 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 3 3 3 3 3

[m /s] [s/m ] [s/m ] [s/m ] [s/m ]

0 - 8 hr 3.470x10-4 1.0 7.72x10

-4 2.56x10

-4 1.32x10

-2 6.35x10

-4 8 - 24 hr 1.750x10-4 1.0 4.55x10

-4 1.51x10

-4 7.78x10

-3 3.74x10

-4 1 - 4 days 2.320x10-4 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 = 2827.2 cfm Emergency Mode Fresh Air Make-up Flow = 1413.6 cfm Emergency Mode Recirculation Flow = 1413.6 cfm Emergency Mode Unfiltered Inleakage Flow = 16 cfm (1)

Normal Mode Fresh Air Make-up Flow = 660.0 cfm Base Infiltration Leak Rate (Depressurized) = 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 Page 2 of 3 TIME DEPENDENT CONTROL ROOM PARAMETERS (For TID-14844 based analyses.)

ACCIDENT TIMING SCENARIOS Event Abbreviation FSAR SRP Accident Section Section Scenario 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 3,2 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 The four types of accident scenarios (1-4) are described below.

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 32 Page 1 of 1 CONTROL ROOM ATMOSPHERIC DISPERSION (X/Q) FACTORS FOR ALTERNATE SOURCE TERM ANALYSIS EVENTS Release -

Receptor 0-2 hr 2-8 hr 8-24 hr 1-4 days 4-30 days Release Point Receptor Point Pair X/Q X/Q X/Q X/Q X/Q Containment Normal Intake A 9.16E-03 7.17E-03 2.68E-03 2.07E-03 1.57E-03 Closest Point B Containment Emergency B 7.26E-04 6.18E-04 2.47E-04 1.77E-04 1.30E-04 Closest Point Intake Normal Intake C SIRWT Vent 9.57E-02 7.59E-02 2.87E-02 2.19E-02 1.65E-02 B

Emergency D SIRWT Vent 9.66E-04 7.92E-04 3.13E-04 2.20E-04 1.64E-04 Intake Normal Intake (

E Plant Stack 5.29E-03 1) 3.89E-03(1) 1.51E-03(1) 1.13E-03(1) 8.41E-04(1)

B Emergency F Plant Stack 8.32E-04 7.69E-04 2.83E-04 2.15E-04 1.57E-04 Intake Normal Intake G Closest ADV 9.95E-03(2) 7.96E-03(2) 3.27E-03(2) 2.39E-03(2) 1.80E-03(2)

A Emergency H Closest ADV 7.36E-04 6.42E-04 2.43E-04 1.75E-04 1.28E-04 Intake Normal Intake I Closest SSRV 1.24E-02(2) - - - -

A Emergency J Closest SSRV 7.96E-04 - - - -

Intake Containment Normal Intake K 1.25E-02 9.83E-03 3.62E-03 2.86E-03 2.28E-03 Equipment Door B Containment Emergency L 7.32E-04 6.13E-04 2.45E-04 1.75E-04 1.29E-04 Equipment Door Intake Feedwater Area Normal Intake M 2.20E-02 1.75E-02 7.10E-03 5.24E-03 3.87E-03 Exhauster V-22A A Feedwater Area Emergency N 8.65E-04 7.56E-04 2.81E-04 2.04E-04 1.47E-04 Exhauster V-22A Intake (1) bounding X/Q values used for FHA, SLBOC and SFCD (2) bounding X/Q values used for SGTR

FSAR CHAPTER 14 - SAFETY ANALYSIS TABLE 14.24-3 Revision 28 Page 1 of 1 RELEASE-RECEPTOR POINT PAIRS ASSUMED FOR AST ANALYSIS EVENTS Event Normal Intake & Unfiltered MHA Emergency 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&H SGTR Initial release via SSRVs Initial release via SSRVs switching to ADVs switching to ADVs CRE Containment Leakage A B I&G J&H Secondary Side Release Initial release via SSRVs Initial release via SSRVs switching to ADVs switching to ADVs SLBOC E F