ML20309A766

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2 to Updated Final Safety Analysis Report, Chapter 15, Appendix 15A, Tables
ML20309A766
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Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 10/08/2020
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Duke Energy Carolinas
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Office of Nuclear Reactor Regulation
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RA-19-0424
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McGuire Nuclear Station UFSAR Appendix 15A. Tables Appendix 15A. Tables

McGuire Nuclear Station UFSAR Table 15-1 (Page 1 of 1)

Table 15-1. Deleted Per 1993 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-2 (Page 1 of 2)

Table 15-2. Summary of Accidents Analyzed With Computer Codes FSAR Section Description of Transient Summary of Cases Analyzed 15.1.2 Increase in Feedwater Flow 1. full power

2. zero power 15.1.3 Increase in Steam Flow 1. manual rod control, most negative moderator coefficient
2. automatic rod control, most negative moderator coefficient 15.1.4 Accidental Depressurization of Main Steam System 15.1.5 Steam Line Break 1. offsite power maintained at hot zero power
2. offsite power lost at hot zero power
3. CFM at hot full power
4. DNB at hot full power 15.2.3 Turbine Trip 1. peak RCS pressure
2. peak Main Steam System pressure 15.2.6 Loss of Offsite Power 15.2.7 Loss of Normal Feedwater 1. short term core cooling
2. long term core cooling 15.2.8 Feedwater Line Break 1. long term core cooling
2. short term core cooling 15.3.1 Partial Loss of Flow 15.3.2 Complete Loss of Flow 15.3.3 Locked Rotor 1. peak RCS pressure
2. core cooling with offsite power maintained
3. core cooling with offsite power lost 15.4.1 Zero Power Rod Bank Withdrawal 1. core cooling
2. peak RCS pressure (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-2 (Page 2 of 2)

FSAR Section Description of Transient Summary of Cases Analyzed 15.4.2 At Power Rod Bank Withdrawal 1. bank withdrawal from 10% power core cooling

2. bank withdrawal from 8% power peak RCS pressure
3. bank withdrawal from 50% power core cooling
4. bank withdrawal from 98% power core cooling
5. bank withdrawal from 100% power core cooling 15.4.3 Control rod misoperation
a. Dropped rod(s)
b. Dropped rod bank
c. Misaligned rod
d. Single rod withdrawal 15.4.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature 15.4.7 Misloaded Assembly 1. Region 1 Region 3
2. Region 1 Region 2
3. Region 2 in center
4. Region 2 in periphery 15.4.8 Rod Ejection 1. BOL, full power
2. BOL, zero power
3. EOL, full power
4. EOL, zero power
5. BOL, full power peak RCS pressure 15.6.1 Accidental RCS Depressurization 15.6.3 Steam Generator Tube Rupture 1. Thermal-hydraulic input to dose analysis
2. DNB analysis 15.6.5 Loss of Coolant Accident 1. DECLG CD=1.0, Reference Transient
2. 1.5 inch SBLOCA
3. 2 inch SBLOCA
4. 3 inch SBLOCA
5. 4 inch SBLOCA (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-3 (Page 1 of 1)

Table 15-3. Summary of Computer Codes and Methodologies Used in Accident Analyses Computer Code or Transient Numbers 1Analyzed with that Computer Code Methodology or Methodology WLOP, W-3S 15.1.5 RETRAN-02 15.1.2, 15.1.3, 15.1.4, 15.1.5, 15.2.3, 15.2.6, 15.2.7, 15.2.8, 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3a, b, d, 15.4.4, 15.4.8, 15.6.1, 15.6.3 VIPRE-01 15.1.2, 15.1.5, 15.2.7(1), 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3a, b, d, 15.4.8, 15.6.1, 15.6.3 SCD 15.1.2, 15.1.3, 15.2.7, 15.2.8, 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3a, b, d, 15.6.1, 15.6.3 WRB-2M 15.1.2, 15.2.7(1), 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3, 15.4.8, 15.6.1, 15.6.3 Deleted Per 2006 Update Deleted Per 2008 Update CASMO 3/SIMULATE-3P or 15.1.2, 15.1.3, 15.1.4, 15.1.5, 15.2.3, 15.2.6, 15.2.7, 15.2.8, CASMO-4/SIMULATE -3 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3, 15.4.4, 15.4.6, MOX 15.6.1, 15.6.3 NOTRUMP 15.6.5 WCOBRA/TRAC 15.6.5 LOCTA-IV 15.6.5 LOTIC 15.6.5 Deleted Per 2006 Update SIMULATE-3K 15.4.8 Note:

1. Transients are numbered according to the cases listed in Table 15-2.

(09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-4 (Page 1 of 5)

Table 15-4. Summary of Input Parameters for Accident Analyses Using Computer Codes Moderator Density Doppler Initial Core RCS Pzr Liquid FSAR Case MTC Coefficient Coefficient Output Flow Vessel Pzr Press. Inventory Feedwater Section ID (pcm/°F) (%k/k/g/cc) (pcm/°F) (MWt) (gpm) Tavg (°F) (psia) (%) Temp. (°F) 15.1.2 1 -51 NA -1.2 3469 388,000 585.1 2250 64 443 9

15.1.2 2 Note NA -3.5 0 382,000 557 2250 34 70 15.1.3 1 -51 NA -1.2 3469 388,000 585.1 2250 64 443 15.1.3 2 -51 NA -1.2 3469 388,000 585.1 2250 64 443 15.1.4 Note 9 NA -3.5 0 Note 16 561 2208 16 Note 15 15.1.5 1 Note 9 NA -3.5 0 371,796 561 2198 16 Note 15 15.1.5 2 Note 9 NA -3.5 0 371,796 561 2198 16 Note 15 15.1.5 3 NA Note ²¹ -1.2 3469 388,000 585.1 2250 46 443 15.1.5 4 NA Note ²¹ -1.2 3469 388,000 585.1 2250 46 443 15.2.3 2 NA Note 6 -0.9 3479 420,000 589.1 2310 64 440 15.2.3 1 NA Note 6 -0.9 3479 373,596 589.1 2280 64 440 14 14 15.2.6 Note NA Note 3479 373,596 589.1 2250 55 440 15.2.7 1 NA Note 6 -0.9 3469 388,000 585.1 2250 46 443 15.2.7 2 NA Note 6 -0.9 3479 379,464 589.1 2208 46 440 15.2.8 1 NA Note 6 -0.9 3479 Note 18 589.1 2208 46 440 15.2.8 2 NA Note 6 -0.9 3469 388,000 585.1 2250 46 443 6

15.3.1 NA Note -0.9 3469 388,000 585.1 2250 46 440 15.3.2 NA Note 6 -0.9 3469 388,000 585.1 2250 46 443 15.3.3 1 NA Note 6 -0.9 3479 379,464 589.1 2310 64 443 15.3.3 2&3 NA Note 6 -0.9 3469 388,000 585.1 2250 46 442 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-4 (Page 2 of 5)

Moderator Density Doppler Initial Core RCS Pzr Liquid FSAR Case MTC Coefficient Coefficient Output Flow Vessel Pzr Press. Inventory Feedwater Section ID (pcm/°F) (%k/k/g/cc) (pcm/°F) (MWt) (gpm) Tavg (°F) (psia) (%) Temp. (°F) 15.4.1 1 NA Note 6 Note 4 0 299,613 557 2250 16 NA 15.4.1 2 NA Note 6 Note 4 0 371,796 557 2310 34 NA 15.4.2 1 NA Note 6 Note 4 347 384,120 559.8 2250 19 336 6 4 15.4.2 2 NA Note Note 273 375,669 563.8 2250 37 333 15.4.2 3 NA Note 6 Note 4 1734 384,120 571.0 2250 31 382 15.4.2 4 NA Note 6 Note 4 3399.6 384,120 584.5 2250 45.4 438 15.4.2 5 NA Note 6 Note 4 3469 388,000 585.1 2250 46 440 15.4.3a, NA Note 6 -0.9 3469 388,000 585.1 2250 46 443 b

Deleted Row per 2017 Update 15.4.3c NA NA NA 3411 388,000 590.8 2250 NA NA 15.4.3d NA Note 6 Note 4 3469 388,000 585.1 2250 46 440 15.4.4 -51 NA -1.2 1735 272,747 574.8 2208 30.4 372 15.4.7 1 NA NA NA 3493 NA NA NA NA NA 15.4.7 2 NA NA NA 3493 NA NA NA NA NA 15.4.7 3 NA NA NA 3493 NA NA NA NA NA 15.4.7 4 NA NA NA 3493 NA NA NA NA NA 15.4.8 1 Note 10 Note 10 Note 10 3479 371,796 589.1 2203 46 NA 15.4.8 2 Note 10 Note 10 Note 10 68 290,000 561 2203 16 NA 10 10 10 15.4.8 3 Note Note Note 3479 371,796 589.1 2203 46 NA 15.4.8 4 Note 10 Note 10 Note 10 68 290,000 561 2203 16 NA (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-4 (Page 3 of 5)

Moderator Density Doppler Initial Core RCS Pzr Liquid FSAR Case MTC Coefficient Coefficient Output Flow Vessel Pzr Press. Inventory Feedwater Section ID (pcm/°F) (%k/k/g/cc) (pcm/°F) (MWt) (gpm) Tavg (°F) (psia) (%) Temp. (°F) 15.4.8 5 Note 10 Note 10 Note 10 3479 371,796 589.1 2310 64 443 15.6.1 0.0 NA -0.9 3469 388,000 587.5 2250 46 445 15.6.3 1 Note 6 NA -1.2 3479 373,596 581.1 2310 64 440 6

15.6.3 2 Note NA -0.9 3469 388,000 585.1 2250 46 442 15.6.5 1 NA Note 11 Note 11 344520 Note 19 587.5 2250 55 442 15.6.5 2 NA Note 11 Note 11 3479 Note 19 585.1 2250 55 442 15.6.5 3 NA Note 11 Note 11 3479 Note 19 585.1 2250 55 442 15.6.5 4 NA Note 11 Note 11 3479 Note 19 585.1 2250 55 442 19 15.6.5 5 NA Note 11 Note 11 3479 Note 585.1 2250 55 442 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-4 (Page 4 of 5)

Moderator Density Doppler Initial Core RCS Pzr Liquid FSAR Case MTC Coefficient Coefficient Output Flow Vessel Pzr Press. Inventory Feedwater Section ID (pcm/°F) (%k/k/g/cc) (pcm/°F) (MWt) (gpm) Tavg (°F) (psia) (%) Temp. (°F)

Notes:

1. Deleted per 1998 update.
2. -0.9 pcm/°F at HFP to -1.20 pcm/°F at HZP
3. -1.04 pcm/°F at HFP to -1.325 pcm/°F at HZP
4. -1.20 pcm/°F at HFP to -1.50 pcm/°F at HZP.
5. Deleted per 1998 update.
6. The most positive MTC (implemented as a least positive or most negative MDC) allowed by the Technical Specifications was used.
7. Deleted per 1998 update.
8. The McGuire Technical Specification limit for the moderator temperature coefficient (MTC) is based on a +7 pcm/°F MTC from 0 to 70% of nominal power, ramping to 0 pcm/°F at full power. Sensitivity studies have shown that a 0 pcm/°F MTC at a full power condition conservatively bounds the combinations of power and MTC permitted by the Technical Specifications.
9. Refer to Figure 15-17.
10. Refer to Section 15.4.8.2.2.
11. The moderator density and Doppler effects on reactivity during LOCA transients are accounted for in the evaluation models as described in Section 15.6.5 and the associated references.
12. Deleted per 1998 update.
13. Deleted per 1998 update.
14. The results of this transient are not sensitive to reactivity feedback assumptions.
15. Main feedwater temperature is 60°F. Auxilliary feedwater temperature is 32°F.
16. An RCS flow of 390,000 gpm x 0.99 - 2.2% is assumed. The analysis results are always bounded by results in Section 15.1.5.

Therefore, the analysis was not re-analyzed with 388,000 gpm flow.

17. Deleted Per 2012 Update.
18. An RCS flow of 390,000 gpm - 2.2% is assumed. The analysis was evaluated and the reduced flow has negligible impact on the analysis.
19. An RCS flow of 390,000 gpm is assumed. An evaluation of a change to 388,000 gpm concluded that there would be no impact on meeting the relevant acceptance criteria due to reduced RCS flow.

(22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-4 (Page 5 of 5)

Moderator Density Doppler Initial Core RCS Pzr Liquid FSAR Case MTC Coefficient Coefficient Output Flow Vessel Pzr Press. Inventory Feedwater Section ID (pcm/°F) (%k/k/g/cc) (pcm/°F) (MWt) (gpm) Tavg (°F) (psia) (%) Temp. (°F)

20. Analysis was originally performed at 3445 MWt (3411 plus 1% for conservatism). However, 1% for heat balance error was also added into the analysis, so it remains bounding for the MUR (3479 MWt). An MUR uprate evaluation was performed at 3469 MWt (101.7% of 3411 MWt) plus 0.3% uncertainty to derive the PCT penalty included in Table 15.61.
21. Based on MTC = -24 pcm/°F (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-5 and 15-6 (Page 1 of 1)

Table 15-5. Deleted Per 1992 Update Table 15-6. Deleted Per 1992 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-7 (Page 1 of 1)

Table 15-7. Rod Drop Times Used in FSAR Analyses FSAR Section Drop Time to Dashpot (sec) 15.1.2 2.2 15.1.3 Note 2 15.1.4 Instantaneous 15.1.5 Instantaneous for hot zero power 2.2 for hot full power 15.2.3 2.2 15.2.6 2.2 15.2.7 2.2 15.2.8 2.2 15.3.1 2.2 15.3.2 2.2 15.3.3 2.2 15.4.1 2.2 15.4.2 2.2 15.4.3 2.2 15.4.4 Note 2 15.4.6 Note 1 15.4.7 Note 2 15.4.8 2.2 15.6.1 2.2 15.6.2 Note 2 15.6.3 2.2, Note 1 15.6.5 (small break) 2.2 15.7 (all sections) Note 2 Notes:

1. Results of transient are not sensitive to rod drop time. For FSAR Section 15.6.3, this note only applies to the dose analysis.
2. Reactor trip was not necessary to analyze transient.

(09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-8 (Page 1 of 1)

Table 15-8. Trip Points and Time Delays to Trip Assumed in Accident Analyses Time Limiting Trip Point Assumed in Delays Trip Function Analysis (Seconds)

Power range high neutron flux, high Note2 0.5 setting Power range high neutron flux, low setting 116.1% 0.5 Overtemperature T Variable see Figure 15-1 1.51 Overpower T Variable see Figure 15-1 1.51 High pressurizer pressure Note2 2.0 Low pressurizer pressure Note2 2.0 Low reactor coolant flow (from loop flow 83.5% loop flow 1.0 detectors)

Undervoltage trip Note3 1.5 Low-low steam generator level Note2 2.01 Safety injection Not applicable 2.0 Note:

1. Time delay from the indicated parameter satisfying the trip condition until the beginning of rod motion. The delays due to RTD response (T trips only) and electronic signal filtering are accounted for by explicit modeling.
2. The numerical setpoint assumed for this trip function varies depending on the accident being analyzed. The values used are given in the descriptions of the various accidents.
3. A value for this trip setpoint is not explicitly modeled. However, an actual trip setpoint of less than 68% of nominal bus voltage, adjusted for uncertainty and margin, may invalidate the delay time to trip assumed in the analysis.

(14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-9 (Page 1 of 1)

Table 15-9. Deleted Per 1992 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-10 (Page 1 of 1)

Table 15-10. Reactor Core Iodine and Noble Gas Source Terms Gap Release Fractions Core Inventory¹ Locked Rotor2 Rod Ejection2 Nuclide (Curies) Gap Release Fractions Gap Release Fractions I-130 2.52E+04 5% 10%

I-131 7.52E+05 8% 10%

I-132 1.11E+06 5% 10%

I-133 1.60E+06 5% 10%

I-134 1.86E+06 5% 10%

I-135 1.52E+06 5% 10%

Kr-83m 1.27E+05 5% 10%

Kr-85m 2.85E+05 5% 10%

Kr-85 7.31E+03 10% 10%

Kr-87 5.86E+05 5% 10%

Kr-88 8.29E+05 5% 10%

Kr-89 1.07E+06 5% 10%

Xe-131m 9.63E+03 5% 10%

Xe-133m 4.88E+04 5% 10%

Xe-133 1.57E+06 5% 10%

Xe-135m 3.20E+05 5% 10%

Xe-135 4.14E+05 5% 10%

Xe-137 1.48E+06 5% 10%

Xe-138 1.52E+06 5% 10%

Rb-86 1.68E+03 12% 12%

Rb-88 8.48E+05 12% 12%

Rb-89 1.13E+06 12% 12%

Rb-90 1.07E+06 12% 12%

Cs-134 1.91E+05 12% 12%

Cs-136 4.16E+04 12% 12%

Cs-137 9.15E+04 12% 12%

Cs-138 1.59E+06 12% 12%

Cs-139 1.51E+06 12% 12%

Br-83 1.27E+05 5% 10%

Br-85 2.85E+05 5% 10%

Br-87 4.72E+05 5% 10%

Note:

1. Based on fuel assembly burnup to 62,000 MWD/MTU
2. Based upon Regulatory Guide 1.183 (13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-11 (Page 1 of 1)

Table 15-11. Normal Reactor Coolant Specific Activities for Iodine and Noble Gas Isotopes Nuclide Specific Activity1 (µCi/g)

I-131 0.66 I-132 0.24 I-133 1.1 I-134 0.16 I-135 0.58 Xe-131m 1.9 Xe-133m 3.1 Xe-133 281.0 Xe-135m 0.7 Xe-135 6.3 Xe-138 0.7 Kr-83m 0.0.

Kr-85m 2.1 Kr-85 8.8 Kr-87 1.2 Kr-88 3.7 Kr-89 0.0 Note:

1. Reactor coolant concentrations at equilibrium assuming Technical Specification Iodine activity.

(14 APR 2005)

McGuire Nuclear Station UFSAR Table 15-12 (Page 1 of 4)

Table 15-12. Environmental Consequences Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Low Control Room Boundary Population Zone Main Steam Line Break 15.1.5 Pre-Existing Iodine Spike 0.23 0.03 2.12 (25.0) (25.0) (5.0)

Coincident Iodine Spike 0.25 0.08 3.74 (2.5) (2.5) (5.0)

Locked Rotor 15.3.3 Loss of Offsite Power 1.73 0.20 2.91 (2.5) (2.5) (5.0)

Offsite Power Available 1.68 0.17 1.69 (2.5) (2.5) (5.0)

Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Low Population Control Room Boundary Zone Rod Ejection 15.4.8 2.52 0.67 4.05 (6.3) (6.3) (5.0)

(13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-12 (Page 2 of 4)

Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Low Population Control Room Boundary Zone Instrument Line Break 15.6.2 Pre-Existing Iodine Spike 1.12 0.10 0.45 (2.5) (2.5) (5.0)

Coincident Iodine Spike 0.41 0.04 0.14 (2.5) (2.5) (5.0)

Steam Generator Tube Rupture 15.6.3 Pre-Existing Iodine Spike 4.78 0.67 2.97 (25.0) (25.0) (5.0)

Coincident Iodine Spike 2.08 0.42 1.61 (2.5) (2.5) (5.0)

Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room Loss of Coolant Accident 15.6.5 12.25 2.23 4.86 (25.0) (25.0) (5.0)

Accident FSAR Section Rem Total Effective Dose Equivalent (TEDE)

(13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-12 (Page 3 of 4)

Exclusion Area Boundary Low Population Zone Control Room Waste Gas Decay Tank Failure 15.7.1 0.25 0.02 0.01 (0.5) (0.5) (5.0)

Liquid Storage Tank Failure 15.7.2 1.89 0.17 0.61 (2.5) (2.5) (5.0)

Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room Cask Drop in Pit 15.7.4 0.01 0.0009 0.0006 (6.3) (6.3) (5.0)

(13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-12 (Page 4 of 4)

Rem Total Effective Dose Equivalent (TEDE)

Accident FSAR Section Exclusion Area Boundary Low Population Control Zone Room Fuel Handling Accident Inside Containment 15.7.4.1 3.25 0.29 3.86 (6.3) (6.3) (5.0)

Deleted row per 2017 update Dropped Weir Gate Inside SFP Building 15.7.4.3 6.16 0.56 3.25 (6.3) (6.3) (5.0) 2-hr Dose at 2500 ft. Exclusion Area 30 day Dose at 29000 ft. Low Boundary Population Zone Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room Tornado Generated Missile Accident 15.10.3 0.72 0.71 1.85 (25.0) (25.0) (5.0) 2-hr Dose at 2500 ft. Exclusion Area Boundary Accident FSAR Section Whole Body Thyroid Cask Drop Accident 15.7.4.5 0.01 0.2 (2.5) (30.0)

(13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-13 (Page 1 of 3)

Table 15-13. Time Sequence of Events for Incidents Which Cause an Increase In Heat Removal By The Secondary System Accident Event Time (sec.)

Excessive Feedwater Flow at All main feedwater control valves fail fully open 0 Full Power Over power T setpoint reached 53.2 Reactor trip occurs due to overpower T 54.7 Turbine trip occurs due to reactor trip 54.9 Minimum DNBR occurs 55.0 Excessive Increase in Secondary Steam Flow Manual Reactor Control 10% step load increase 0 Equilibrium conditions reached (approximate 260 time only)

Inadvertent Opening of a Inadvertent opening of one main steam safety 0 Steam Generator Relief or valve Safety Valve Pressurizer empties 102 Low pressurizer pressure setpoint reached 211 Return to Criticality 254 Borated water reaches core 329 Low steam line pressure setpoint reached NA Steam Line Isolation NA Subcriticality achieved 418 Steam System Piping Failure

1. With offsite power Break occurs 0 maintained at hot zero Operator manually trips reactor 0 power Pressurizer level goes offscale low 12 SI actuation on low pressurizer pressure 21 Criticality occurs 22 Steam line isolation on low steam line 24 pressure Main feedwater flow ceases 33 SI pumps begin to deliver unborated water to 38 RCS (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-13 (Page 2 of 3)

Accident Event Time (sec.)

Peak heat flux occurs 119 NV injection lines purged of unborated water 119 One train of SI fails 119 Subcriticality achieved 166 Pressurizer level returns onscale >200

2. With offsite power lost at Break occurs 0 hot zero power Operator manually trips reactor 0 Pressurizer level goes offscale low 12 SI actuation on low pressurizer pressure 21 Offsite power lost 21 Reactor coolant pumps begin to coast down 21 Main feedwater pumps trip 21 Criticality occurs 22 Steam line isolation on low steam line 24 pressure Main feedwater flow ceases 32 SI pumps begin to deliver unborated water to 53 RCS NV Injection lines purged of unborated water 137 One train of SI fails 137 Pressurizer level returns onscale 182 Peak heat flux occurs 224 Subcriticality achieved 242 Deleted per 2015 update
3. CFM at hot full power Break occurs 0 High flux trip setpoint reached 12.7 Reactor trip occurs due to high flux trip 13.2 Peak reactor power occurs 13.3 Turbine Trip occurs due to reactor trip 13.5 Loss of offsite power occurs on turbine trip 13.5 RCPs trip due to loss of offsite power 13.5
4. DNB at hot full power Break occurs 0 (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-13 (Page 3 of 3)

Accident Event Time (sec.)

OPT trip setpoint reached 11.6 Reactor trip occurs due to OPT trip 12.1 Peak reactor power occurs 12.3 Turbine trip occurs due to reactor trip 12.4 Loss of offsite power occurs on turbine trip 12.4 RCPs trip due to loss of offsite power 12.4 MDNBR occurs 13.2 (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-14 (Page 1 of 1)

Table 15-14. Parameters for Main Steam Line Break Dose Analysis

1. Failed fuel (%) 0
2. Iodine spike values for each case
a. Pre-existing spike 60
b. Coincident spike 500
3. Control Room Data
a. Control room volume (ft3) 107,000
b. Control room pressurization (cfm) 1800
c. In-leakage before pressurization (cfm) 625
d. In-leakage after pressurization (cfm) 210
e. Control room filter efficiencies (% particulates, elementals/organic) 99, 98
4. Partitioning fraction 0.01
5. Iodine fractions (% elemental, organic) 97, 3
6. Maximum primary to secondary leak rate (gpd) 389
7. aLetdown flow (gpm) 125
8. Reactor coolant system leakage (gpm) 11
9. Total steam release from the faulted steam generator (lbm) 2.34E+05
10. Total steam release from the intact steam generators (lbm) 1.85E+06 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-15 (Page 1 of 1)

Table 15-15. Deleted per 2014 Update (24 APR 2014)

McGuire Nuclear Station UFSAR Table 15-16 (Page 1 of 2)

Table 15-16. Time Sequence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (Sec)

Turbine Trip

1. Maximum Secondary System Pressure Turbine Trip 0.0 Case Pressurizer PORVs lift 4.3 Steam Safety Valves lift 6.7 Overtemperature T setpoint reached 13.8 Control rod insertion begins 15.3 Peak secondary system pressure occurs 18.4
2. Maximum Primary System Pressure Turbine Trip, loss of main feed flow 0.0 Case High pressurizer pressure setpoint reached 5.6 Control rod insertion begins 7.6 Steam Safety Valves lift 8.0 Pressurizer Safety Valves lift 8.2 Peak primary system pressure occurs 8.7 Loss of Non-Emergency AC Power Main feedwater flow stops 0.1 Power lost to control rod gripper coils 0.1 Reactor coolant pumps begin to coastdown 0.1 Rods begin to drop 0.6 Peak water level in pressurizer occurs 3 Flow from two motor driven auxiliary 60 feedwater pumps is started Core decay heat decreases to auxiliary ~ 600 feedwater heat removal capacity Loss of Main Feedwater
1. Short-Term Core Cooling Case Main feedwater flow stops 1.0 Pressurizer PORVs begin cycling 24.6 Low-low steam generator level reactor trip 56.4 reached Minimum DNBR occurs 58.0 Rods begin to drop 58.4
2. Long-Term Core Cooling Case Main feedwater flow stops 0.01 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-16 (Page 2 of 2)

Accident Event Time (Sec)

Pressurizer PORVs begin cycling 38.2 Low-low steam generator level reactor trip 56.6 reached Rods begin to drop 58.6 Steam safety valves lift 60.1 Auxiliary feedwater flow on 116.6 Core decay heat plus pump heat decreases ~1749 to auxiliary feedwater heat removal capacity Feedwater System Pipe Break Feedwater line break to SG B 0 Safety injection on high containment 10.1 pressure Reactor trip on high containment pressure 10.1 SI Reactor coolant pumps tripped 10.1 Turbine trip on reactor trip 10.2 Steam line isolation on turbine trip 10.2 Safety injection terminated 70 Motor-driven auxiliary feedwater pumps 70 deliver flow SG B boiled dry 100 Core decay heat decreases to auxiliary ~ 1750 feedwater heat removal capacity Auxiliary feedwater to faulted generator 1800 isolated End of simulation 3000 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-17 (Page 1 of 1)

Table 15-17. Time Sequence of Events for Incidents Which Cause a Decrease in Reactor Coolant System Flow Accident Event Time (Sec)

Partial Loss of Forced Reactor Coolant Coastdown begins 0.0 Flow Low flow reactor trip setpoint reached 1.5 Rods begin to drop 2.5 Minimum DNBR occurs 3.3 Complete Loss of Forced Reactor Coolant All operating pumps lose power and begin 0.0 Flow coasting down Reactor coolant pump undervoltage trip 0.0 point reached Rods begin to drop 1.5 Minimum DNBR occurs 3.4 Reactor Coolant Pump Shaft Seizure Rotor on one pump locks 1.0 (Core Cooling Capability for Offsite Power Maintained)

Low flow reactor trip setpoint reached 1.08 Rods begin to drop 2.08 Minimum DNBR occurs 3.5 Reactor Coolant Pump Shaft Seizure Rotor on one pump locks 1.0 (Core Cooling Capability for Offsite Power Lost)

Low flow reactor trip setpoint reached 1.08 Rods begin to drop 2.08 Minimum DNBR occurs 3.9 Reactor Coolant Pump Shaft Seizure Rotor on one pump locks 1.0 (Peak RCS Pressure)

Low flow reactor trip setpoint reached 1.06 Rods begin to drop 2.06 Maximum RCS pressure occurs 5.6 (11 NOV 2006)

McGuire Nuclear Station UFSAR Table 15-18 (Page 1 of 1)

Table 15-18. Parameters for Locked Rotor Dose Analysis

1. Data and assumptions used to estimate radioactive sources from postulated accidents.
a. Failed fuel for Loss of Offsite Power (LOOP) scenario (%) 6
b. Failed fuel for Offiste Power Available (OPA) scenario (%) 1.5
c. Reactor core inventory Table 15-10
d. Concurrent Iodine Spiking Factor 335
e. Iodine fractions (% elemental, organic) 97, 3
f. Reactor turbine and trip (minutes) 0
2. Data and assumptions used to estimate activity released
a. Total Steam Release from the Faulted Steam Generator (LOOP) 2.95E+05 (lbm)
b. Total Steam Release from the Intact Steam Generators (LOOP) 8.85E+05 (lbm)
c. Total Steam Release from the Faulted Steam Generator (OPA) 3.42E+05 (lbm)
d. Total Steam Release from the Intact Steam Generator (OPA) 1.02E+06 (lbm)
e. Control room volume (ft³) 107,000
f. Control room pressurization (cfm) 1800
g. Control room in-leakage before pressurization (cfm) 500
h. Control room in-leakage after pressurization (cfm) 210
i. Control room filter efficiencies (% particulates, 99, 98 elemental/organics)
j. Steam generator partitioning fraction 0.01
3. Dispersion data
a. Distance to exclusion area boundary (m) 762
b. Distance to low population zone (m) 8850
c. X/Q at exclusion area boundary (sec/m³) 9.0E-04
d. X/Q at low population zone (sec/m³) 8.0E-05
4. Dose data Table 15-12 (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-19 (Page 1 of 3)

Table 15-19. Time Sequence of Events for Incidents which Cause Reactivity and Power Distribution Anomalies Accident Event Time (sec.)

Uncontrolled RCCA Bank Withdrawal Initiation of uncontrolled rod withdrawal 0.0 from a Subcritical or Low Power from 10-9 of nominal power Startup Condition (Core Cooling Capability)

Power range high neutron flux low 11.2 setpoint reached Peak nuclear power occurs 11.3 Rods begin to fall into core 11.7 Peak heat flux occurs 12.0 Minimum DNBR occurs 12.0 Peak average fuel temperature occurs 12.2 Uncontrolled RCCA Bank Withdrawal Initiation of uncontrolled rod withdrawal 0.0 from a Subcritical or Low Power from 10-9 of nominal power Startup Condition (Peak RCS Pressure)

Power range high neutron flux low 11.2 setpoint reached Peak Nuclear Power occurs 11.3 Rods begin to fall into core 11.7 Peak RCS Pressure 13.9 Uncontrolled RCCA Bank Withdrawal at Power (Core Cooling Capability)

Initiate Bank Withdrawal 0.0 Pressurizer Sprays Full On 7.3 Pressurizer PORVs Full Open 24.4 High Flux Trip Setpoint Reached 42.6 Pressurizer Safety Valves Lift 42.9 Control Rod Insertion Begins 43.1 Uncontrolled RCCA Bank Withdrawal Initiate Bank Withdrawal 0.0 at Power (Peak RCS Pressure)

High Pressure Reactor Trip Setpoint 12.3 Reached (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-19 (Page 2 of 3)

Accident Event Time (sec.)

Pressurizer Safety Valves Lift 14.0 Control Rod Insertion Begins 14.3 Peak Pressure Occurs 14.8 Single RCCA Withdrawal Initiate RCCA Withdrawal 0.0 Pressurizer Sprays Full On 2.2 RCCA Completely Withdrawn 4.2 OT T Reactor Trip Setpoint Reached 39.2 Control Rod Insertion Begins 40.7 Startup of an Inactive Reactor Coolant Initiation of pump startup 0.1 Pump at an Incorrect Temperature Pump reaches full speed 10.1 Peak heat flux occurs 15.5 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 1.

a) Dilution during power operation Dilution begins -

(manual rod control)

Reactor trip setpoint reached 0 Operator terminates dilution < 998 b) Dilution during power operation Dilution begins -

(automatic rod control)

Rod insertion limit alarm setpoint 0 reached Operator terminates dilution <1554

2. Dilution during startup Dilution begins -

Reactor trip setpoint reached 0 Operator terminates dilution < 998 Deleted Per 2008 Update.

Rod Cluster Control Assembly Ejection

1. Beginning of Cycle, Full Power Initiation of rod ejection 0.0 Power range high neutron flux high 0.056 setpoint reached (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-19 (Page 3 of 3)

Accident Event Time (sec.)

Peak nuclear power occurs 0.083 Rods begin to fall into core 0.556

2. End of Cycle, Zero Power Initiation of rod ejection 0.0 Power range high neutron flux low 0.272 setpoint reached Peak nuclear power occurs 0.323 Rods begin to fall into core 0.772 (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-20 (Page 1 of 1)

Table 15-20. Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident Time in Cycle Beginning Beginning End End Power Level, % 102 0 102 0 Ejected rod worth, $ 0.19 1.32 0.26 1.45 Delayed neutron fraction, % 0.56 0.56 0.47 0.47 Fq after rod ejection 4.90 19.60 4.84 20.78 Number of operational pumps 4 3 4 3 Max. fuel pellet average temperature, 3327 1621 2818 1299

°F Max. fuel center temperature, °F 5068 2062 4326 1640 Max. clad temperature, °F 798 741 1296 768 Max. fuel stored energy, cal/gm 147 61 132 47

% Failed fuel <22 <22 <22 <22 (13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-21 (Page 1 of 1)

Table 15-21. Parameters for Rod Ejection Accident Dose Analysis

1. Data and assumptions used to estimate radioactive sources from postulated accidents
a. Failed fuel (%) 22
b. Reactor core inventory Table 15-10
c. Iodine fractions (% elemental, organic) 97, 3
d. Reactor and turbine trip (minutes) 0
2. Data and assumptions used to estimate activity released
a. Control Room volume (ft3) 107,000
b. Control Room pressurization (cfm) 1800
c. Control room in-leakage before pressurization (cfm) 500
d. Control room in-leakage after pressurization (cfm) 210
e. Control room filter efficiency (% particulates, 99, 98.05 elemental/organics)
f. Steam generator iodine partitioning fraction 0.01
3. Dispersion data
a. Distance to exclusion area boundary (m) 762
b. Distance to low population zone (m) 8850
c. /Q at exclusion area boundary (sec/m3) 9.0E-04
d. /Q at exclusion area boundary (sec/m3) 0-8 hours 8.0E-05 8-24 hours 5.2E-06 1-4 days 1.7E-06 4-30 days 3.7E-07
4. Dose data Table 15-12 (11 NOV 2006)

McGuire Nuclear Station UFSAR Table 15-22 (Page 1 of 1)

Table 15-22. Deleted Per 1998 Update.

(14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-23 (Page 1 of 1)

Table 15-23. Time Sequence of Events For Incidents Which Cause A Decrease In Reactor Coolant Inventory Accident Event Time (sec)

Inadvertent Opening of a Pressurizer Safety valve opens 0.1 Safety Valve Low pressurizer pressure reactor trip 22.9 setpoint reached Rods begin to drop 24.9 Minimum DNBR occurs 25.4 Steam Generator Tube Rupture Double ended tube rupture occurs 0.1 (Dose Analysis)

Manual reactor trip 1200 Loss-of-offsite power occurs 1200 Steamline PORV on ruptured SG fails 1201 open 2 pump/2 train maximum safety injection 1212 begins Operators isolate CA flow to the ruptured 1290 SG Operators identify ruptured SG and close 2100 ruptured SG MSIV Operators close failed open steam line 3362 PORV Operators begin RCS cooldown with 5754 operable SG PORVs Operators close operable steam line 6325 PORVs Operators open pressurizer PORV to 6850 depressurize RCS Break flow terminated 6931 (DNB Analysis) Double ended tube rupture occurs 1.0 Reactor trip/turbine trip on oTT 319.0 Reactor coolant pumps lost 319.0 MDNBR occurs 320.9 (13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-24 (Page 1 of 1)

Table 15-24. Parameters for Steam Generator Tube Rupture Dose Analysis

1. Failed fuel (%) 0
2. Reactor and turbine trip (minutes) 20
3. Iodine spike values for each case
a. Pre-existing spike 60
b. Coincident spike 335
4. Control Room Data
a. Control room volume (ft3) 107,000
b. Control room pressurization (cfm) 1800
c. In-leakage before pressurization (cfm) 500
d. In-leakage after pressurization (cfm) 210
e. Control room filter efficiencies (% particulates, 99, 98 elementals/organic)
5. Partitioning fraction (steam generator/condenser) 0.01,0.15
6. Iodine fractions (% elemental, organic) 97, 3
7. Maximum primary to secondary leak rate (gpd) 389
8. aLetdown flow (gpm) 125
9. Reactor coolant system leakage (gpm) 11
10. Total steam release from the ruptured steam generator (lbm) 2.40E+05
11. Total steam release from the intact steam generator (ibm) 1.75E+05 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15 15-32 (Page 1 of 1)

Table 15-25. Deleted Per 1998 Update Table 15-26. Deleted Per 1998 Update Table 15-27. Deleted Per 1996 Update Table 15-28. Deleted Per 1996 Update Table 15-29. Deleted Per 1996 Update Table 15-30. Deleted Per 1996 Update Table 15-31. Deleted Per 1998 Update Table 15-32. Deleted Per 1998 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-33 (Page 1 of 1)

Table 15-33. Parameters for Minimum Safeguards (Design Basis) LOCA Dose Analysis

1. Data and assumptions used to estimate radioactive source from postulated accidents
a. Power Level (MW th) 3479
b. Failed fuel (%) 100
2. Iodine Species Breakdown (% particulate, elemental, organic)
a. Containment Model 95, 4.85, 0.15
b. ECCS Model 0, 97, 3
3. Data and assumptions used to estimate activity released
a. Containment Free Volume (including ice condensers)
1. Upper containment volume (ft3) 827,000
2. Lower containment volume (ft3) 391,000
3. Total containment free volume (ft3) 1,120,000
b. Annulus Volume - half of the volume credited (427,000 ft3) 213,000 3
c. Control Room Volume (ft ) 107,000
d. Containment Leak Rate (percent of containment volume per day)
1. 0 t 24 hrs 0.3
2. t > 24 hrs 0.15
e. Bypass Leakage Fraction 0.07
f. VE start time (seconds) 39
g. Annulus vacuum established (seconds) 71
4. Equipment Hatch Release (sscm) 500
5. ECCS back-leakage to the Auxiliary Building (gpm) 0.9
6. ECCS back-leakage to the FWST (gpm) 10
7. Control Room In-leakage Data (One train of VC)
a. Time of control room pressurization (seconds) 30
b. Control room in-leakage before pressurization (cfm) 625
c. Control room in-leakage after pressurization (cfm) 210
8. Control Room Ventilation Data
a. VC fan flow (cfm) 1800
b. VC iodine filter efficiency (% particulates, elemental and 99, 98 organic)
9. Annulus Ventilation Data
a. VE fan flow (cfm) 7200
b. VE Iodine filter efficiency (% particulates, elemental and 98, 91 organic)
10. Spray Removal Data
a. NS Start Time (minutes) 80
b. Auxiliary Spray Start Time (minutes) N/A
c. Spray credit ceases (hours) 24
d. Spray Decontamination Factors
1. Particulate 50
2. Elemental 200
3. Organic (Spray credit not taken for organic Iodine) N/A
11. Doses Table 15-12 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-34 (Page 1 of 1)

Table 15-34. Deleted Per 2009 Update (10 OCT 2009)

McGuire Nuclear Station UFSAR Table 15-35 (Page 1 of 2)

Table 15-35. Source Term Inventory and Gap Fractions Assumed for Fuel Handling and Tornado Missile Accidents Assembly Inventory Gap Release Gap Inventory Nuclide (curies) Fractions1 (curies)

Br-83 1.31E+05 0.05 6.55E+03 Br-85 2.99E+05 0.05 1.50E+04 Br-87 4.95E+05 0.05 2.48E+04 I-130 3.95E+04 0.05 1.98E+03 I-131 8.09E+05 0.08 6.47E+04 I-132 1.18E+06 0.05 5.90E+04 I-133 1.67E+06 0.05 8.35E+04 I-134 1.95E+06 0.05 9.75E+04 I-135 1.60E+06 0.05 8.00E+04 Kr-83m 1.32E+05 0.05 6.60E+03 Kr-85m 2.98E+05 0.05 1.49E+04 Kr-85 7.48E+03 0.10 7.48E+02 Kr-87 6.15E+05 0.05 3.08E+04 Kr-88 8.69E+05 0.05 4.35E+04 Kr-89 1.12E+06 0.05 5.60E+04 Xe-131m 1.24E+04 0.05 6.20E+02 Xe-133m 5.20E+04 0.05 2.60E+03 Xe-133 1.65E+06 0.05 8.25E+04 Xe-135m 3.62E+05 0.05 1.81E+04 Xe-135 4.12E+05 0.05 2.06E+04 Xe-137 1.55E+06 0.05 7.75E+04 Xe-138 1.59E+06 0.05 7.95E+04 Rb-86 2.54E+03 0.12 3.05E+02 Rb-88 8.89E+05 0.12 1.07E+05 Rb-89 1.18E+06 0.12 1.42E+05 Rb-90 1.12E+06 0.12 1.34E+05 Cs-134 2.06E+05 0.12 2.47E+04 Cs-136 5.92E+04 0.12 7.10E+03 Cs-137 9.23E+04 0.12 1.11E+04 Cs-138 1.66E+06 0.12 1.99E+05 (13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-35 (Page 2 of 2)

Assembly Inventory Gap Release Gap Inventory Nuclide (curies) Fractions1 (curies)

Cs-139 1.58E+06 0.12 1.90E+05 Note:

NRC Assumption in Regulatory Guide 1.183 For fuel pins which exceed the rod power/burnup criteria of Footnote 11 in RG 1.183, the gap fractions from RG 1.183 are increased by a factor of 3 for Kr-85, Xe-133, Cs-134 and Cs-137, and increased by a factor of 2 for I-131, and other noble gases, halogens and alkali metals (References 1 and 2). A maximum of 25 fuel rods, per assembly, shall be allowed to exceed the rod power/burnup criteria of Footnote 11 in RG 1.183, in accordance with the license amendment request submitted by letter dated July 15, 2015.

(13 OCT 2018)

McGuire Nuclear Station UFSAR Table 15-36 (Page 1 of 1)

Table 15-36. Deleted Per 1992 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-37 (Page 1 of 1)

Table 15-37. Parameters for Postulated Instrument Line Break Accident Analysis

1. Failed fuel (%) 0
2. Isolation of Instrument Line (minutes) 30
3. Iodine spike values for each case
a. Pre-existing spike 60
b. Coincident spike 335
4. Control Room Data
a. Control room volume (ft3) 107,000
b. Control room pressurization (cfm) 1800
c. In-leakage before pressurization (cfm) 625
d. In-leakage after pressurization (cfm) 210
e. Control room filter efficiencies (% particulates, elementals/organic) 99, 98
5. Partitioning fraction 0.1
6. Iodine fractions (% elemental, organic) 97, 3
7. aAssumed ILB flow rate (gpm) 150
8. Reactor coolant system leakage (gpm) 11 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-38 (Page 1 of 1)

Table 15-38. Deleted Per 1998 Update.

(14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-39 (Page 1 of 1)

Table 15-39. Parameters Used to Evaluate Tornado Missile Impact On Spent Fuel

1. Meteorology
a. Offsite atmospheric dilution for tornado conditions 8.1E-5 s/m3
2. Spent Fuel Radioactivity Bases
a. Number of fuel assemblies damaged 38
b. Conservative case maximum assembly inventory See Table 15-35
c. Decay period
1) 8 Assemblies 16 days
2) 30 Assemblies 295 days
3. Iodine Partition Factor 200
4. Effective Iodine Composition Fractions
a. Elemental Iodine 57%
b. Organic Iodine 43%
5. Ventilation Credit Assumed
a. Duration VF is in filter mode or secured 27 days
b. Time required to start VC 30 minutes
c. Rate of Unfiltered Control Room Inleakage
1. Pre-pressurization 500 cfm
2. Pressurization 210 cfm
d. VC Air Flow Rate 1800 cfm (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-40 (Page 1 of 1)

Table 15-40. Parameters Used to Evaluate LOCA During Lower Containment Pressure Relief

1. REACTOR COOLANT RADIOACTIVITY INVENTORY BASES
a. Iodine concentrations 60 µCi/gm Thyroid Dose Equivalents of I-131
b. Noble gas concentrations 100/µCi/gm
c. Reactor coolant mass 447,274 lbm
2. RADIOACTIVITY RELEASE BASES
a. Lower containment air mass (for dilution) 37,360 lbm Basis: Active volume = 368,000 ft3 Temperature = 250ºF
b. Lower containment mass release 19 lbm Basis: LOCA overpressure = 12 psig VQ valve isolation = 4 sec
c. Filtration None (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-41 (Page 1 of 1)

Table 15-41. Deleted Per 2015 Update (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-42 (Page 1 of 1)

Table 15-42. Input Parameters Used in the SBLOCA Analyses Parameter Value used Core power (mwt) 3479 Total peaking factor, FQ 2.7 ( 4 ft), 2.5 (> 4 ft)

Hot rod enthalpy rise peaking factor (FH) 1.67 K(z) limit 1.0 ( 4 ft), 0.9259 (> 4 ft)

Power shape See Figure 15-137 Fuel assembly array 17x17 RFA Nominal cold leg accumulator water volume (ft3/accumulator) 950 Nominal cold leg accumulator tank volume (ft3/acumulator) 1363 Minimum cold leg accumulator gas pressure (psia) 570 Cold leg accumulator temperature (°F) 125 Pumped safety injection flow See Table 15-50 Pumped safety injection temperature (°F) 110 Nominal vessel average temperature (°F) 585.1 Pressurizer pressure (psia) 2250 RCS flow (gpm/loop) 97,500 Steam generator tube plugging (%) 5 Pressurizer low pressure safety injection setpoint (psia) 1715 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-43 and 15-44 (Page 1 of 1)

Table 15-43. Deleted Per 2001 Update Table 15-44. Deleted Per 2001 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-45 (Page 1 of 1)

Table 15-45. Minimum Injected ECCS Flows Assumed in LBLOCA Analyses.

One Train Operational RCS Pressure High-Head SI Intermediate-Head Low-Head SI (psia) (gpm) (gpm) (gpm) 14.7 285 420 2600 50 280 410 1800 75 280 410 1225 100 275 405 500 125 275 400 0 (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-46 (Page 1 of 2)

Table 15-46. Parameters for Post-LOCA Subcriticality Analysis Volume Grouping Boron Concentration (ppm)

Low Head Safety Injection (LHSI) Discharge RWST minimum3 to Intermediate Head Safety Injection (IHSI) and High Head Safety Injection (HHSI) suction (Valve NI136B to Valves NI332A & NI333B)

Refueling Water Storage Tank (RWST) to Valve FW28 RWST minimum3 RWST to IHSI suction RWST minimum3 RWST to Valve NV223 RWST minimum3 Normal Containment Spray Discharge RWST minimum3 Containment Spray Suction from RWST RWST minimum3 LHSI Discharge to Aux. Cont. Spray (downstream of 3501 isolation MOVs)

LHSI Suction from Sump 3501 LHSI Suction from Loop C Hot Leg 3501 Containment Spray Suction from Sump 3501 RCS variable2 LHSI Discharge to Cold Legs variable2 LHSI Discharge to IHSI and HHSI Suction variable2 (Valve ND58 to Valves NI332A & NI333B)

(LHSI Discharge to Valves ND58 & NI136B)

LHSI Discharge to B and C Hot Legs variable2 Valve FW28 to LHSI Suction variable2 LHSI Mini-Flow variable2 IHSI Discharge to LHSI Discharge variable2 IHSI Discharge to Hot Legs variable2 IHSI Mini-Flow variable2 HHSI Discharge to Cold Legs variable2 Valve NV223 to HHSI Suction variable2 LHSI Discharge to Aux. Cont. Spray (upstream of variable2 isolation MOVs)

(14 APR 2005)

McGuire Nuclear Station UFSAR Table 15-46 (Page 2 of 2)

Volume Grouping Boron Concentration (ppm)

Note:

1. EOC Mode 4 RCS boron concentration.
2. "variable" indicates that the associated volume concentration is assumed equal to the RCS boron concentration, which is a function of burnup.
3. This boron concentration is equal to the cycle specific RWST minimum boron concentration specified in the Core Operating Limits Report. The analysis assumes RWST boron concentrations between 2475 and 2875 ppm.

(14 APR 2005)

McGuire Nuclear Station UFSAR Table 15-47 and 15-48 (Page 1 of 1)

Table 15-47. Deleted Per 1998 Update Table 15-48. Deleted Per 1998 Update (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-49 (Page 1 of 1)

Table 15-49. Small Break LOCA Results Fuel Cladding Data 1.5 inch 2 inch 3 inch 4 inch Peak cladding temperature 1 (°F) N/A 1323 1153 1208 Time of PCT (sec) N/A 3449 1986 1092 PCT location (ft) N/A 11.50 11.25 11.25 Maximum local ZrO2 (%) N/A 0.24 0.09 0.06 Maximum local ZrO2 location (ft) N/A 11.50 11.25 11.25 Total core-wide average ZrO2 (%) N/A 0.03 0.01 0.01 Hot rod burst time (sec) N/A N/A N/A N/A Hot rod burst location (ft) N/A N/A N/A N/A Note:

1. There is no core uncovery for the 1.5 inch case.

(05 APR 2011)

McGuire Nuclear Station UFSAR Table 15-50 (Page 1 of 1)

Table 15-50. Minimum ECCS Flow Assumed in SBLOCA Analyses (One Train Operational, Break Backpressure Equal to RCS Presure)

High-Head SI Intermediate-Head SI RCS Pressure 3 Injecting Lines 1 Spilling Line 3 Injecting Lines 1 Spilling Line (psia) (gpm) (gpm) (gpm) (gpm) 14.7 275 105 405 150 50 275 100 400 145 75 270 100 395 145 100 270 100 390 145 125 270 100 385 145 150 265 100 385 140 200 265 100 375 140 250 260 100 365 135 300 255 95 360 135 500 245 90 320 120 700 230 85 280 105 900 210 80 235 90 1100 195 75 175 65 1300 175 65 85 35 1450 160 60 0 0 1500 155 60 0 0 2310 0 0 0 0 Deleted Per 2008 Update.

(13 APR 2008)

McGuire Nuclear Station UFSAR Table 15-51 (Page 1 of 1)

Table 15-51. Small Break LOCA Time Sequence of Events 1.5 inch (sec) 2 inch (sec) 3 inch (sec) 4 inch (sec)

Start 0 0 0 0 Reactor trip signal 114 57 23 13 ESFAS signal 135 73 32 21 ECC delivery 167 105 64 53 Loop seal cleared N/A N/A 628 333 Core uncovery N/A 2378 993 703 Cold leg accumulator injection N/A N/A N/A 997 RWST low level 1211 1206 1199 1183 Peak cladding temperature occurs N/A 3449 1986 1092 Core recovery N/A 5122 2933 1971 (13 APR 2008)

McGuire Nuclear Station UFSAR Table 15-52 (Page 1 of 1)

Table 15-52. Large Break LOCA Time Sequence of Events for Reference Transient Event Time (seconds)

Break opening time 20 Safety injection signal 24 Accumulator injection begins 31 Pumped safety injection begins 56 Bottom of core recovery 58 Accumulators empty 62 Time of peak cladding temperature 286 (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-53 (Page 1 of 3)

Table 15-53. Key Large Break LOCA Parameters and Initial Transient Assumptions Parameter Initial Transient Uncertainty or Bias 1.0 Plant Physical Description

a. Dimensions Nominal PCTMOD1
b. Flow resistance Nominal PCTMOD
c. Pressurizer location Opposite broken loop Bounded
d. Hot assembly location Under limiting location Bounded
e. Hot assembly type 17x17 RFA with IFM Bounded
f. SG tube plugging level D5, maximum (10%) Bounded4 2.0 Plant Initial Operating Conditions 2.1 Reactor Power
a. Core average linear heat rate (AFLUX) Nominal power (3445 MWt)6 PCTPD2
b. Peak linear heat rate (PLHR) Derived from desired Tech Spec PCTPD (TS) limit and maximum baseload FQ
c. Hot rod average linear heat rate (HRFLUX) Derived from TS FH PCTPD
d. Hot assembly average heat rate (HAFLUX) HRFLUX/1.04 PCTPD
e. Hot assembly peak heat rate (HAPHR) PLHR/1.04 PCTPD
f. Axial power distribution (PBOT, PMID) Figure 15-244 PCTPD
g. Low power region relative power (PLOW) Minimum (0.2) Bounded4
h. Hot assembly burnup BOL Bounded
i. Prior operating history Equilibrium decay heat Bounded
j. Moderator Temperature Coefficient (MTC) Tech Spec Maximum (0) Bounded
k. HFP boron 800 ppm Typical (24 APR 2014)

McGuire Nuclear Station UFSAR Table 15-53 (Page 2 of 3)

Parameter Initial Transient Uncertainty or Bias 2.2 Fluid Conditions

a. Tavg Nominal Tavg = 587.5°F PCTIC3 (Catawba Unit 2)
b. Pressurizer pressure Nominal (2250 psia) PCTIC
c. Loop flow Minimum (97500 gpm) PCTMOD5
d. TUH Best Estimate 0
e. Pressurizer level Nominal (55% of volume) 0
f. Accumulator temperature Nominal (115°F) PCTIC
g. Accumulator pressure Nominal (631.5 psig, Catawba PCTIC units)
h. Accumulator liquid volume Nominal (7106 gal, McGuire PCTIC units)
i. Accumulator line resistance Nominal (McGuire Unit 2) PCTIC
j. Accumulator boron Minimum (McGuire units) Bounded 3.0 Accident Boundary Conditions
a. Break location Cold leg Bounded
b. Break type Guillotine PCTMOD
c. Break size Nominal (cold leg area) PCTMOD
d. Offsite power On (RCS pumps running) Bounded4
e. Safety injection flow Minimum Bounded
f. Safety injection temperature Nominal (85°F) PCTIC
g. Safety injection delay Max delay (17 sec) Bounded (24 APR 2014)

McGuire Nuclear Station UFSAR Table 15-53 (Page 3 of 3)

Parameter Initial Transient Uncertainty or Bias

h. Containment pressure Minimum based on WC/T M&E Bounded
i. Single failure ECCS: Loss of 1 SI train Bounded
j. Control rod drop time No control rods Bounded 4.0 Model Parameters
a. Critical Flow Nominal (as coded) PCTMOD
b. Resistance uncertainties in broken loop Nominal (as coded) PCTMOD
c. Initial stored energy/fuel rod behavior Nominal (as coded) PCTMOD
d. Core heat transfer Nominal (as coded) PCTMOD
e. Delivery and bypassing of ECCS Nominal (as coded) Conservative
f. Steam binding/entrainment Nominal (as coded) Conservative
g. Noncondensable gases/accumulator nitrogen Nominal (as coded) Conservative
h. Condensation Nominal (as coded) PCTMOD Notes:
1. PCTMOD indicates this uncertainty is part of code and global model uncertainty.
2. PCTPD indicates this uncertainty is part of power distribution uncertainty.
3. PCTIC indicates this uncertainty is part of initial condition uncertainty
4. Confirmed by analysis
5. Assumed to be result of loop resistance uncertainty
6. Analysis was originally performed at 3445 MWt (3411 plus 1% for conservatism). However, 1% for heat balance error was also added into the analysis, so it remains bounding for the MUR (3479 MWt). AN MUR uprate evaluation was performed at 3469 MWt (101.7% of 3411 MWt) plus 0.3% uncertainty to derive the PCT penalty included in Table 15-61.

Sensitivity analysis concluded loss of offsite power is more limiting than assuming offsite power on (RCS pumps running)

(24 APR 2014)

Mc Guire Nuclear Station UFSAR Table 15-54 (Page 1 of 1)

Table 15-54. Best-Estimate Large Break LOCA - Overall Results Component Blowdown Peak (°F) First Reflood Peak Second Reflood Peak

(°F) (°F)

PCT50% <1256 <1384 <1512 95%

PCT <1548 <1692 <2028 (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-55 (Page 1 of 3)

Table 15-55. Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis Parameter Operating Range 1.0 Plant Physical Description a) Dimensions No in-board assembly grid deformation during LOCA + SSE b) Flow resistance N/A c) Pressurizer location N/A d) Hot assembly location Anywhere in core e) Hot assembly type Fresh 17X17 RFA f) SG tube plugging level 10% (Catawba 2) and 5% (McGuire and Catawba 1) 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Core avg linear heat rate Core power < 3445 MWt4 b) Peak linear heat rate FQ < 2.70 ( 4 ft), FQ 2.50 (> 4 ft) [see Note 1]

c) Hot rod average linear heat rate FH < 1.67 [see Note 2]

d) Hot assembly average linear heat rate P HA 1.67/1.04 [see Note 3]

e) Hot assembly peak linear heat rate FQHA < 2.7/1.04 ( 4 ft), FQ 2.50/1.04 (> 4 ft) [see Note 1]

f) Axial power dist (PBOT, PMID) Figure 15-243 g) Low power region relative power (PLOW) 0.2 PLOW 0.8 h) Hot assembly burnup 75000 MWD/MTU, lead rod i) Prior operating history All normal operating histories j) MTC 0 at HFP k) HFP boron Normal letdown l) Rod power census See Table 15-56 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-55 (Page 2 of 3)

Parameter Operating Range 2.2 Fluid Conditions a) Tavg 581.1 Tavg 593.9°F b) Pressurizer pressure 2190 PRCS 2310 psia c) Loop flow 97,500 gpm/loop d) TUH Current upper internals, Tcold UH e) Pressurizer level Normal level, automatic control f) Accumulator temperature 105 TACC 125°F g) Accumulator pressure 555 PACC 708 psig h) Accumulator volume 6790 VACC 7422 gal. (McGuire), 7550 VACC 8159 gal.

(Catawba) i) Accumulator fL/D Current line configuration j) Minimum accumulator boron 2275 ppm 3.0 Accident Boundary Conditions a) Break location N/A b) Break type N/A c) Break size N/A d) Offsite power Available or LOOP e) Safety injection flow Table 15-45 f) Safety injection temperature 58°F SI Temp 90°F, Reference 60 (covers a RWST temperature range of 70-100°F and component cooling water temperature down to 45°F) g) Safety injection delay 17 seconds (with offsite power) 32 seconds (with LOOP) h) Containment pressure Bounded -- see Figure 15-233 (22 APR 2017)

McGuire Nuclear Station UFSAR Table 15-55 (Page 3 of 3)

Parameter Operating Range i) Single failure Loss of one train j) Control rod drop time N/A Notes:

1. To account for fuel pellet thermal conductivity degradation, the allowed FQ peaking factor is subject to these normalization factors (interpolation allowed):

Hot Rod Average Burnup = 0 GWD/MTU, FQ normalization factor = 1.0 Hot Rod Average Burnup = 35 GWD/MTU, FQ normalization factor = 1.0 Hot Rod Average Burnup = 55 GWD/MTU, FQ normalization factor = 0.9 Hot Rod Average Burnup = 62 GWD/MTU, FQ normalization factor = 0.8

2. To account for fuel pellet thermal conductivity degradation, the allowed FH peaking factors are subject to these normalization factors (interpolation allowed):

Hot Rod Average Burnup = 0 GWD/MTU, FH normalization factor = 1.0 Hot Rod Average Burnup = 35 GWD/MTU, FH normalization factor = 1.0 Hot Rod Average Burnup = 55 GWD/MTU, FH normalization factor = 0.95 Hot Rod Average Burnup = 62 GWD/MTU, FH normalization factor = 0.9

3. To account for fuel pellet thermal conductivity degradation, the allowed PHA peaking factors are subject to these normalization factors (interpolation allowed; extrapolation beyond 59,615 MWD/MTU is acceptable, provided the individual fuel rod burnups remain withing the licensed limit of 62,000 MWD/MTU):

Assembly Average Burnup = 0 MWD/MTU, PHA normalization factor = 1.0 Assembly Average Burnup = 33,654 MWD/MTU, PHA normalization factor = 1.0 Assembly Average Burnup = 52,885 MWD/MTU, PHA normalization factor = 0.95 Assembly Average Burnup = 59,615 MWD/MTU, PHA normalization factor = 0.9

4. Analysis was originally performed at 3445 MWt (3411 plus 1% for conservatism). However, 1% for heat balance error was also added into the analysis, so it remains bounding for the MUR (3479 MWt). AN MUR uprate evaluation was performed at 3469 MWt (101.7% of 3411 MWt) plus 0.3% uncertainty to derive the PCT penalty included in Table 15-61.

(22 APR 2017)

Mc Guire Nuclear Station UFSAR Table 15-56 (Page 1 of 1)

Table 15-56. Rod Census Used in Best-Estimate large Break LOCA Analysis Power Ratio Rod Group (Relative to HA Rod Power)  % of Core 1 1.0 10 2 0.912 10 3 0.853 10 4 0.794 30 5 0.726 40 (14 OCT 2000)

McGuire Nuclear Station UFSAR Table 15-57 (Page 1 of 1)

Table 15-57. Deleted Per 2012 Update (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-58 (Page 1 of 2)

Table 15-58. Reactor Core Inventory and Release Fractions for LOCA Noble Gases Halogens Alkali Metals Tellurium Metals Ba, Sr Release Fractions Release Fractions Release Fractions Release Fractions Release Fractions Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel 5% 95% 5% 35% 5% 25% 0% 5% 0% 2%

Inventory Inventory Inventory Inventory Inventory Nuclide (Curies) Nuclide (Curies) Nuclide (Curies) Nuclide (Curies) Nuclide (Curies)

Kr83m 1.56E+07 Br83 1.55E+07 Rb86 2.08E+05 Sb127 9.65E+06 Sr89 1.03E+08 Kr85m 3.40E+07 Br85 3.41E+07 Rb88 1.00E+08 Sb129 3.43E+07 Sr90 9.31E+06 Kr85 1.07E+06 Br87 5.56E+07 Rb89 1.33E+08 Te127m 1.58E+06 Sr91 1.66E+08 Kr87 6.96E+07 I130 2.96E+06 Rb90 1.25E+08 Te127 9.51E+06 Sr92 1.69E+08 Kr88 9.79E+07 I131 1.04E+08 Cs134 2.09E+07 Te129 3.27E+07 Sr93 1.83E+08 Kr89 1.25E+08 I132 1.52E+08 Cs136 5.60E+06 Te129m 6.63E+06 Ba139 2.00E+08 Xe131m 1.43E+06 I133 2.15E+08 Cs137 1.26E+07 Te131 8.69E+07 Ba140 1.88E+08 Xe133m 6.72E+06 I134 2.47E+08 Cs138 2.09E+08 Te132 1.49E+08 Ba141 1.82E+08 Xe133 2.08E+08 I135 2.06E+08 Cs139 1.96E+08 Te133 1.22E+08 Xe135m 4.51E+07 Te133m 1.01E+08 Xe135 6.65E+07 Te134 2.12E+08 Xe137 1.98E+08 Xe138 1.98E+08 (10 OCT 2009)

McGuire Nuclear Station UFSAR Table 15-58 (Page 2 of 2)

Noble Metals Cerium Group Lanthanides Release Fractions Release Fractions Release Fractions Release Fractions Release Fractions Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel Gap Early In-Vessel 0% 0.25% 0% 0.05% 0% 0.02% 0% 0.02% 0% 0.02%

Inventory Inventory Inventory Inventory Inventory Nuclide (Curies) Nuclide (Curies) Nuclide (Curies) Nuclide (Curies) Nuclide (Curies)

Mo99 1.97E+08 Ce141 1.73E+08 Y90 9.66E+06 La140 1.98E+08 Eu155 3.86E+05 Tc99m 1.74E+08 Ce143 1.79E+08 Y91 1.34E+08 La141 1.81E+08 Eu156 3.17E+07 Tc101 1.76E+08 Ce144 1.32E+08 Y91m 9.72E+02 La142 1.82E+08 Pr143 1.56E+08 Ru103 1.72E+08 Np237 4.23E+01 Y92 1.51E+08 La143 1.79E+08 Pr144 1.33E+08 Ru105 1.25E+08 Np238 5.04E+07 Y93 1.23E+08 Nd147 6.93E+07 Pr144m 1.86E+06 Ru106 6.37E+07 Np239 2.32E+09 Y94 1.90E+08 Pm147 1.75E+07 Am241 1.75E+04 Rh103m 1.72E+08 Pu236 7.32E+01 Y95 1.94E+08 Pm148 1.88E+07 Am242m 1.14E+03 Rh105 1.12E+08 Pu238 4.29E+05 Zr95 1.78E+08 Pm148m 2.96E+06 Am242 8.95E+06 Pd109 4.67E+07 Pu239 3.74E+04 Zr97 1.78E+08 Pm149 6.64E+07 Am243 4.41E+03 Pu240 5.16E+04 Nb95 1.79E+08 Pm151 2.18E+07 Cm242 5.13E+06 Pu241 1.45E+07 Nb95m 1.98E+06 Sm153 5.73E+07 Cm242 9.41E+05 Pu242 2.97E+02 Nb97 1.78E+08 Eu154 9.87E+05 Pu243 5.62E+07 (10 OCT 2009)

McGuire Nuclear Station UFSAR Table 15-59 (Page 1 of 1)

Table 15-59. Assumptions Used for the Cask Drop Accident

1. Data and assumptions used to estimate radioactive source from postulated accidents
a. Number of assemblies ruptured 32
b. Percentage of pins breached (%) 100
c. Cask Free Volume (m3) 5.39
d. Respirable Fraction (%) 5
e. Percentage of particulate CRUD released to the Fuel Building (%) 30
2. Fuel Building Filtration Assumptions
a. HEPA filter particulate removal efficiency (%) 95
b. Charcoal filter volatile and I-129 removal efficiency (%) 90
3. /Q at Exclusion Area Boundary (sec/m3) 9.0E-04
4. Doses Table 15-12 (10 OCT 2009)

McGuire Nuclear Station UFSAR Table 15-60 (Page 1 of 1)

Table 15-60. Isotopic Inventory of the Dry Cask Drop Accident Release ISG-5 Release Cask Release Chemical Fraction to Release from Pool Isotope Ci/Assembly Ci/Cask Form Cask Fraction Area Mn-54 2.85 9.12E+01 Act. Prod. 7.83E-03 2.40E-01 1.71E-01 Fe-55 19.03 6.09E+02 Act. Prod. 7.83E-03 2.40E-01 1.14E+00 Co-60 19.03 6.09E+02 Act. Prod. 1.00E+00 2.40E-01 1.46E+02 Ni-63 373 1.19E+04 Act. Prod. 7.83E-03 2.40E-01 2.24E+01 Pu-238 2320 7.42E+04 Fines 3.00E-05 4.00E-02 8.91E-02 Pu-239 168 5.38E+03 Fines 3.00E-05 4.00E-02 6.45E-03 Pu-240 261 8.35E+03 Fines 3.00E-05 4.00E-02 1.00E-02 Pu-241 60700 1.94E+06 Fines 3.00E-05 4.00E-02 2.33E+00 Am-241 876 2.80E+04 Fines 3.00E-05 4.00E-02 3.36E-02 Cm-244 2300 7.36E+04 Fines 3.00E-05 4.00E-02 8.83E-02 H-3 206 6.59E+03 Gas 3.00E-01 8.00E-01 1.58E+03 Kr-85 3390 1.08E+05 Gas 3.00E-05 8.00E-01 2.60E+04 Sr-90 38400 1.23E+06 Volatile 3.00E-05 8.00E-02 2.95E+00 Y-90 38400 1.23E+06 Fines 3.00E-05 4.00E-02 1.47E+00 Ru-106 3140 1.00E+05 Volatile 2.00E-04 8.00E-02 1.61E+00 Rh-106 3140 1.00E+05 Volatile 2.00E-04 8.00E-02 1.61E+00 Sb-125 866 2.77E+04 Fines 3.00E-05 4.00E-02 3.33E-02 Te-125m 212 6.78E+03 Fines 3.00E-05 4.00E-02 8.14E-03 I-129 0.02 6.46E-01 Gas 3.00E-01 8.00E-01 1.55E-01 Cs-134 11200 3.58E+05 Volatile 2.00E-04 8.00E-02 5.73E+00 Cs-137 57600 1.84E+06 Volatile 2.00E-04 8.00E-02 2.95E+01 Ba-137m 54400 1.74E+06 Volatile 2.00E-04 8.00E-02 2.79E+01 Ce-144 1310 4.19E+04 Fines 3.00E-05 4.00E-02 5.03E-02 Pr-144 1310 4.19E+04 Fines 3.00E-05 4.00E-02 5.03E-02 Pm-147 13600 4.35E+05 Fines 3.00E-05 4.00E-02 5.22E-01 Eu-154 4560 1.46E+05 Fines 3.00E-05 4.00E-02 1.75E-01 Eu-155 2000 6.40E+04 Fines 3.00E-05 4.00E-02 7.68E-02 (10 OCT 2009)

McGuire Nuclear Station UFSAR Table 15-61 (Page 1 of 1)

Table 15-61. Summary of Licensing Basis LOCA PCT Results, Including PCT Assessments Description PCT (°F) Reference Best Estimate Large Break LOCA; CQD Analysis of Record PCT (Reflood 2) [See Table 15-54] 2028 52 PCT Assessments Decay heat in Monte Carlo calculations 8 69 MONTECF power uncertainty correction 20 70 Safety Injection temperature range 59 60 Input error resulting in an incomplete solution matrix 25 71 Revised blowdown heatup uncertainty distribution 5 72 Vessel unheated conductor noding 0 73 Revised algorithm for average fuel temperature 0 73 Peak transient FQ = 2.7 in bottom third of core 0 74 Change from PAD 3.4 to PAD 4.0 -75 74 Fuel Thermal Conductivity Degradation with Peaking 15 74 Factor Burndown MUR Uprate to 101.7% of 3411 MWt 16 74 Revised Heat Transfer Multiplier Distribution -85 76 HOTSPOT Clad Burst Strain Error 70 77 Current Licensing Basis LBLOCA PCT Including 2086 77 Assessments Small Break LOCA; NOTRUMP Analysis of Record PCT (2-inch break) [See Table 15- 1323 75 49]

PCT Assessments None 0 74 Current Licensing Basis SBLOCA PCT Including 1323 74 Assessments (09 OCT 2015)

McGuire Nuclear Station UFSAR Table 15-62 (Page 1 of 1)

Table 15-62. Dose Equivalent Iodine-131 (DEI-131)

Concentration FGR No. 11, Table 2.1 Isotope (µCi/gm) DCFs (Sv/Bq) DEI (µCi/gm)

I-131 7.56E-01 8.89E-09 7.56E-01 I-132 2.72E-01 1.03E-10 3.15E-03 I-133 1.21E+00 1.58E-09 2.15E-01 I-134 1.81E-01 3.55E-11 7.25E-04 I-135 6.65E-01 3.32E-10 2.49E-02 DEI 1.00E+00 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-63 (Page 1 of 1)

Table 15-63. Dose Equivalent Xenon-133 (DEX-133)

FGR No. 12, Table Concentration III.1 DCFs (Sv-s/Bq-Isotope (µCi/gm) m3) DEX (µCi/gm)

KR-85M 2.10E+00 7.48E-15 1.01E+01 KR-85 8.80E+00 1.19E-16 6.71E-01 KR-87 1.20E+00 4.12E-14 3.17E+01 KR-88 3.70E+00 1.02E-13 2.42E+02 XE-131M 1.90E+00 3.89E-16 4.74E-01 XE-133M 3.10E+00 1.37E-15 2.72E+00 XE-133 2.81E+02 1.56E-15 2.81E+02 XE-135M 7.00E-01 2.04E-14 9.15E+00 XE-135 6.30E+00 1.19E-14 4.81E+01 XE-138 7.00E-01 5.77E-14 2.59E+01 DEX 6.52E+02 (30 NOV 2012)

McGuire Nuclear Station UFSAR Table 15-64 (Page 1 of 1)

Table 15-64. Deleted Per 2014 Update (24 APR 2014)

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