ML20095H496
Text
McGuire Nuclear Station UFSAR Appendix 15A. Tables Appendix 15A. Tables
McGuire Nuclear Station UFSAR Table 15-1 (Page 1 of 1)
(14 OCT 2000)
Table 15-1. Deleted Per 1993 Update
McGuire Nuclear Station UFSAR Table 15-2 (Page 1 of 2)
(09 OCT 2015)
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
McGuire Nuclear Station UFSAR Table 15-2 (Page 2 of 2)
(09 OCT 2015)
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
McGuire Nuclear Station UFSAR Table 15-3 (Page 1 of 1)
(09 OCT 2015)
Table 15-3. Summary of Computer Codes and Methodologies Used in Accident Analyses Computer Code or Methodology Transient Numbers 1Analyzed with that Computer Code 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 CASMO-4/SIMULATE -3 MOX 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, 15.4.3, 15.4.4, 15.4.6, 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.
McGuire Nuclear Station UFSAR Table 15-4 (Page 1 of 5)
(22 APR 2017)
Table 15-4. Summary of Input Parameters for Accident Analyses Using Computer Codes FSAR Section Case ID MTC (pcm/°F)
Moderator Density Coefficient
(%k/k/g/cc)
Doppler Coefficient (pcm/°F)
Initial Core Output (MWt)
RCS Flow (gpm)
Vessel Tavg (°F)
Pzr Press.
(psia)
Pzr Liquid Inventory
(%)
Feedwater Temp. (°F) 15.1.2 1
-51 NA
-1.2 3469 388,000 585.1 2250 64 443 15.1.2 2
Note 9 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 15.2.6 Note 14 NA Note 14 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 15.3.1 NA Note 6
-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
McGuire Nuclear Station UFSAR Table 15-4 (Page 2 of 5)
(22 APR 2017)
FSAR Section Case ID MTC (pcm/°F)
Moderator Density Coefficient
(%k/k/g/cc)
Doppler Coefficient (pcm/°F)
Initial Core Output (MWt)
RCS Flow (gpm)
Vessel Tavg (°F)
Pzr Press.
(psia)
Pzr Liquid Inventory
(%)
Feedwater 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 15.4.2 2
NA Note 6 Note 4 273 375,669 563.8 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, b
NA Note 6
-0.9 3469 388,000 585.1 2250 46 443 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 15.4.8 3
Note 10 Note 10 Note 10 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
McGuire Nuclear Station UFSAR Table 15-4 (Page 3 of 5)
(22 APR 2017)
FSAR Section Case ID MTC (pcm/°F)
Moderator Density Coefficient
(%k/k/g/cc)
Doppler Coefficient (pcm/°F)
Initial Core Output (MWt)
RCS Flow (gpm)
Vessel Tavg (°F)
Pzr Press.
(psia)
Pzr Liquid Inventory
(%)
Feedwater 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 15.6.3 2
Note 6 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 15.6.5 5
NA Note 11 Note 11 3479 Note 19 585.1 2250 55 442
McGuire Nuclear Station UFSAR Table 15-4 (Page 4 of 5)
(22 APR 2017)
FSAR Section Case ID MTC (pcm/°F)
Moderator Density Coefficient
(%k/k/g/cc)
Doppler Coefficient (pcm/°F)
Initial Core Output (MWt)
RCS Flow (gpm)
Vessel Tavg (°F)
Pzr Press.
(psia)
Pzr Liquid Inventory
(%)
Feedwater 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.
- 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.
McGuire Nuclear Station UFSAR Table 15-4 (Page 5 of 5)
(22 APR 2017)
FSAR Section Case ID MTC (pcm/°F)
Moderator Density Coefficient
(%k/k/g/cc)
Doppler Coefficient (pcm/°F)
Initial Core Output (MWt)
RCS Flow (gpm)
Vessel Tavg (°F)
Pzr Press.
(psia)
Pzr Liquid Inventory
(%)
Feedwater 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
McGuire Nuclear Station UFSAR Table 15-5 and 15-6 (Page 1 of 1)
(14 OCT 2000)
Table 15-5. Deleted Per 1992 Update Table 15-6. Deleted Per 1992 Update
McGuire Nuclear Station UFSAR Table 15-7 (Page 1 of 1)
(09 OCT 2015)
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.
McGuire Nuclear Station UFSAR Table 15-8 (Page 1 of 1)
(14 OCT 2000)
Table 15-8. Trip Points and Time Delays to Trip Assumed in Accident Analyses Trip Function Limiting Trip Point Assumed in Analysis Time Delays (Seconds)
Power range high neutron flux, high setting Note2 0.5 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 detectors) 83.5% loop flow 1.0 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.
McGuire Nuclear Station UFSAR Table 15-9 (Page 1 of 1)
(14 OCT 2000)
Table 15-9. Deleted Per 1992 Update
McGuire Nuclear Station UFSAR Table 15-10 (Page 1 of 1)
(13 OCT 2018)
Table 15-10. Reactor Core Iodine and Noble Gas Source Terms Gap Release Fractions Nuclide Core Inventory¹ (Curies)
Locked Rotor2 Gap Release Fractions Rod Ejection2 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
McGuire Nuclear Station UFSAR Table 15-11 (Page 1 of 1)
(14 APR 2005)
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.
McGuire Nuclear Station UFSAR Table 15-12 (Page 1 of 4)
(13 OCT 2018)
Table 15-12. Environmental Consequences Rem Total Effective Dose Equivalent (TEDE)
Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room 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 (2.5) 0.08 (2.5) 3.74 (5.0)
Locked Rotor 15.3.3 Loss of Offsite Power 1.73 (2.5) 0.20 (2.5) 2.91 (5.0)
Offsite Power Available 1.68 (2.5) 0.17 (2.5) 1.69 (5.0)
Rem Total Effective Dose Equivalent (TEDE)
Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room Rod Ejection 15.4.8 2.52 (6.3) 0.67 (6.3) 4.05 (5.0)
McGuire Nuclear Station UFSAR Table 15-12 (Page 2 of 4)
(13 OCT 2018)
Rem Total Effective Dose Equivalent (TEDE)
Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room 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)
McGuire Nuclear Station UFSAR Table 15-12 (Page 3 of 4)
(13 OCT 2018)
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)
Accident FSAR Section Rem Total Effective Dose Equivalent (TEDE)
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)
McGuire Nuclear Station UFSAR Table 15-12 (Page 4 of 4)
(13 OCT 2018)
Rem Total Effective Dose Equivalent (TEDE)
Accident FSAR Section Exclusion Area Boundary Low Population Zone Control Room Fuel Handling Accident Inside Containment 15.7.4.1 3.25 (6.3) 0.29 (6.3) 3.86 (5.0)
Deleted row per 2017 update Dropped Weir Gate Inside SFP Building 15.7.4.3 6.16 (6.3) 0.56 (6.3) 3.25 (5.0)
Accident FSAR Section 2-hr Dose at 2500 ft. Exclusion Area Boundary 30 day Dose at 29000 ft. Low Population Zone Exclusion Area Boundary Low Population Zone Control Room Tornado Generated Missile Accident 15.10.3 0.72 (25.0) 0.71 (25.0) 1.85 (5.0)
Accident FSAR Section 2-hr Dose at 2500 ft. Exclusion Area Boundary Whole Body Thyroid Cask Drop Accident 15.7.4.5 0.01 0.2 (2.5)
(30.0)
McGuire Nuclear Station UFSAR Table 15-13 (Page 1 of 3)
(09 OCT 2015)
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 Full Power All main feedwater control valves fail fully open 0 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 time only) 260 Inadvertent Opening of a Steam Generator Relief or Safety Valve Inadvertent opening of one main steam safety valve 0
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 maintained at hot zero power Break occurs 0
Operator manually trips reactor 0
Pressurizer level goes offscale low 12 SI actuation on low pressurizer pressure 21 Criticality occurs 22 Steam line isolation on low steam line pressure 24 Main feedwater flow ceases 33 SI pumps begin to deliver unborated water to RCS 38
McGuire Nuclear Station UFSAR Table 15-13 (Page 2 of 3)
(09 OCT 2015)
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 hot zero power Break occurs 0
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 pressure 24 Main feedwater flow ceases 32 SI pumps begin to deliver unborated water to RCS 53 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
McGuire Nuclear Station UFSAR Table 15-13 (Page 3 of 3)
(09 OCT 2015)
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
McGuire Nuclear Station UFSAR Table 15-14 (Page 1 of 1)
(30 NOV 2012)
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
McGuire Nuclear Station UFSAR Table 15-15 (Page 1 of 1)
(24 APR 2014)
Table 15-15. Deleted per 2014 Update
McGuire Nuclear Station UFSAR Table 15-16 (Page 1 of 2)
(30 NOV 2012)
Table 15-16. Time Sequence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (Sec)
- 1. Maximum Secondary System Pressure Case Turbine Trip 0.0 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 Case Turbine Trip, loss of main feed flow 0.0 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 feedwater pumps is started 60 Core decay heat decreases to auxiliary feedwater heat removal capacity
~ 600 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 reached 56.4 Minimum DNBR occurs 58.0 Rods begin to drop 58.4
- 2. Long-Term Core Cooling Case Main feedwater flow stops 0.01
McGuire Nuclear Station UFSAR Table 15-16 (Page 2 of 2)
(30 NOV 2012)
Accident Event Time (Sec)
Pressurizer PORVs begin cycling 38.2 Low-low steam generator level reactor trip reached 56.6 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 to auxiliary feedwater heat removal capacity
~1749 Feedwater System Pipe Break Feedwater line break to SG B 0
Safety injection on high containment pressure 10.1 Reactor trip on high containment pressure SI 10.1 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 deliver flow 70 SG B boiled dry 100 Core decay heat decreases to auxiliary feedwater heat removal capacity
~ 1750 Auxiliary feedwater to faulted generator isolated 1800 End of simulation 3000
McGuire Nuclear Station UFSAR Table 15-17 (Page 1 of 1)
(11 NOV 2006)
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 Flow Coastdown begins 0.0 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 Flow All operating pumps lose power and begin coasting down 0.0 Reactor coolant pump undervoltage trip point reached 0.0 Rods begin to drop 1.5 Minimum DNBR occurs 3.4 Reactor Coolant Pump Shaft Seizure (Core Cooling Capability for Offsite Power Maintained)
Rotor on one pump locks 1.0 Low flow reactor trip setpoint reached 1.08 Rods begin to drop 2.08 Minimum DNBR occurs 3.5 Reactor Coolant Pump Shaft Seizure (Core Cooling Capability for Offsite Power Lost)
Rotor on one pump locks 1.0 Low flow reactor trip setpoint reached 1.08 Rods begin to drop 2.08 Minimum DNBR occurs 3.9 Reactor Coolant Pump Shaft Seizure (Peak RCS Pressure)
Rotor on one pump locks 1.0 Low flow reactor trip setpoint reached 1.06 Rods begin to drop 2.06 Maximum RCS pressure occurs 5.6
McGuire Nuclear Station UFSAR Table 15-18 (Page 1 of 1)
(09 OCT 2015)
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)
(lbm) 2.95E+05
- b. Total Steam Release from the Intact Steam Generators (LOOP)
(lbm) 8.85E+05
- c. Total Steam Release from the Faulted Steam Generator (OPA)
(lbm) 3.42E+05
- d. Total Steam Release from the Intact Steam Generator (OPA)
(lbm) 1.02E+06
- 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, elemental/organics) 99, 98
- 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
McGuire Nuclear Station UFSAR Table 15-19 (Page 1 of 3)
(09 OCT 2015)
Table 15-19. Time Sequence of Events for Incidents which Cause Reactivity and Power Distribution Anomalies Accident Event Time (sec.)
Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition (Core Cooling Capability)
Initiation of uncontrolled rod withdrawal from 10-9 of nominal power 0.0 Power range high neutron flux low setpoint reached 11.2 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 from a Subcritical or Low Power Startup Condition (Peak RCS Pressure)
Initiation of uncontrolled rod withdrawal from 10-9 of nominal power 0.0 Power range high neutron flux low setpoint reached 11.2 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 at Power (Peak RCS Pressure)
Initiate Bank Withdrawal 0.0 High Pressure Reactor Trip Setpoint Reached 12.3
McGuire Nuclear Station UFSAR Table 15-19 (Page 2 of 3)
(09 OCT 2015)
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 OTT Reactor Trip Setpoint Reached 39.2 Control Rod Insertion Begins 40.7 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature Initiation of pump startup 0.1 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 (manual rod control)
Dilution begins Reactor trip setpoint reached 0
Operator terminates dilution
< 998 b) Dilution during power operation (automatic rod control)
Dilution begins Rod insertion limit alarm setpoint reached 0
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 setpoint reached 0.056
McGuire Nuclear Station UFSAR Table 15-19 (Page 3 of 3)
(09 OCT 2015)
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 setpoint reached 0.272 Peak nuclear power occurs 0.323 Rods begin to fall into core 0.772
McGuire Nuclear Station UFSAR Table 15-20 (Page 1 of 1)
(13 OCT 2018)
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,
°F 3327 1621 2818 1299 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
McGuire Nuclear Station UFSAR Table 15-21 (Page 1 of 1)
(11 NOV 2006)
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, elemental/organics) 99, 98.05
- 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
McGuire Nuclear Station UFSAR Table 15-22 (Page 1 of 1)
(14 OCT 2000)
Table 15-22. Deleted Per 1998 Update.
McGuire Nuclear Station UFSAR Table 15-23 (Page 1 of 1)
(13 OCT 2018)
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 Safety valve opens 0.1 Low pressurizer pressure reactor trip setpoint reached 22.9 Rods begin to drop 24.9 Minimum DNBR occurs 25.4 Steam Generator Tube Rupture (Dose Analysis)
Double ended tube rupture occurs 0.1 Manual reactor trip 1200 Loss-of-offsite power occurs 1200 Steamline PORV on ruptured SG fails open 1201 2 pump/2 train maximum safety injection begins 1212 Operators isolate CA flow to the ruptured SG 1290 Operators identify ruptured SG and close ruptured SG MSIV 2100 Operators close failed open steam line PORV 3362 Operators begin RCS cooldown with operable SG PORVs 5754 Operators close operable steam line PORVs 6325 Operators open pressurizer PORV to depressurize RCS 6850 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
McGuire Nuclear Station UFSAR Table 15-24 (Page 1 of 1)
(22 APR 2017)
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, elementals/organic) 99, 98
- 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
McGuire Nuclear Station UFSAR Table 15 15-32 (Page 1 of 1)
(14 OCT 2000)
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
McGuire Nuclear Station UFSAR Table 15-33 (Page 1 of 1)
(22 APR 2017)
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 (MWth) 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
- c. Control Room Volume (ft3) 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
- 9.
Annulus Ventilation Data
- a. VE fan flow (cfm) 7200
- b. VE Iodine filter efficiency (% particulates, elemental and organic) 98, 91
- 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
McGuire Nuclear Station UFSAR Table 15-34 (Page 1 of 1)
(10 OCT 2009)
Table 15-34. Deleted Per 2009 Update
McGuire Nuclear Station UFSAR Table 15-35 (Page 1 of 2)
(13 OCT 2018)
Table 15-35. Source Term Inventory and Gap Fractions Assumed for Fuel Handling and Tornado Missile Accidents Nuclide Assembly Inventory (curies)
Gap Release Fractions1 Gap Inventory (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
McGuire Nuclear Station UFSAR Table 15-35 (Page 2 of 2)
(13 OCT 2018)
Nuclide Assembly Inventory (curies)
Gap Release Fractions1 Gap Inventory (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.
McGuire Nuclear Station UFSAR Table 15-36 (Page 1 of 1)
(14 OCT 2000)
Table 15-36. Deleted Per 1992 Update
McGuire Nuclear Station UFSAR Table 15-37 (Page 1 of 1)
(30 NOV 2012)
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
McGuire Nuclear Station UFSAR Table 15-38 (Page 1 of 1)
(14 OCT 2000)
Table 15-38. Deleted Per 1998 Update.
McGuire Nuclear Station UFSAR Table 15-39 (Page 1 of 1)
(22 APR 2017)
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
McGuire Nuclear Station UFSAR Table 15-40 (Page 1 of 1)
(14 OCT 2000)
Table 15-40. Parameters Used to Evaluate LOCA During Lower Containment Pressure Relief
- 1. REACTOR COOLANT RADIOACTIVITY INVENTORY BASES
- 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
McGuire Nuclear Station UFSAR Table 15-41 (Page 1 of 1)
(09 OCT 2015)
Table 15-41. Deleted Per 2015 Update
McGuire Nuclear Station UFSAR Table 15-42 (Page 1 of 1)
(30 NOV 2012)
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
McGuire Nuclear Station UFSAR Table 15-43 and 15-44 (Page 1 of 1)
(14 OCT 2000)
Table 15-43. Deleted Per 2001 Update Table 15-44. Deleted Per 2001 Update
McGuire Nuclear Station UFSAR Table 15-45 (Page 1 of 1)
(14 OCT 2000)
Table 15-45. Minimum Injected ECCS Flows Assumed in LBLOCA Analyses.
One Train Operational RCS Pressure (psia)
High-Head SI (gpm)
Intermediate-Head (gpm)
Low-Head SI (gpm) 14.7 285 420 2600 50 280 410 1800 75 280 410 1225 100 275 405 500 125 275 400 0
McGuire Nuclear Station UFSAR Table 15-46 (Page 1 of 2)
(14 APR 2005)
Table 15-46. Parameters for Post-LOCA Subcriticality Analysis Volume Grouping Boron Concentration (ppm)
Low Head Safety Injection (LHSI) Discharge to Intermediate Head Safety Injection (IHSI) and High Head Safety Injection (HHSI) suction (Valve NI136B to Valves NI332A & NI333B)
RWST minimum3 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 isolation MOVs) 3501 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 (Valve ND58 to Valves NI332A & NI333B)
(LHSI Discharge to Valves ND58 & NI136B) variable2 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 isolation MOVs) variable2
McGuire Nuclear Station UFSAR Table 15-46 (Page 2 of 2)
(14 APR 2005)
Volume Grouping Boron Concentration (ppm)
Note:
- 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.
McGuire Nuclear Station UFSAR Table 15-47 and 15-48 (Page 1 of 1)
(14 OCT 2000)
Table 15-47. Deleted Per 1998 Update Table 15-48. Deleted Per 1998 Update
McGuire Nuclear Station UFSAR Table 15-49 (Page 1 of 1)
(05 APR 2011)
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.
McGuire Nuclear Station UFSAR Table 15-50 (Page 1 of 1)
(13 APR 2008)
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 (psia) 3 Injecting Lines (gpm) 1 Spilling Line (gpm) 3 Injecting Lines (gpm) 1 Spilling Line (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.
McGuire Nuclear Station UFSAR Table 15-51 (Page 1 of 1)
(13 APR 2008)
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
McGuire Nuclear Station UFSAR Table 15-52 (Page 1 of 1)
(14 OCT 2000)
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
McGuire Nuclear Station UFSAR Table 15-53 (Page 1 of 3)
(24 APR 2014)
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 (TS) limit and maximum baseload FQ PCTPD
- 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
McGuire Nuclear Station UFSAR Table 15-53 (Page 2 of 3)
(24 APR 2014)
Parameter Initial Transient Uncertainty or Bias 2.2 Fluid Conditions
- a.
Tavg Nominal Tavg = 587.5°F (Catawba Unit 2)
PCTIC3
- 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 units)
PCTIC
- h.
Accumulator liquid volume Nominal (7106 gal, McGuire units)
PCTIC
- 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
McGuire Nuclear Station UFSAR Table 15-53 (Page 3 of 3)
(24 APR 2014)
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)
Mc Guire Nuclear Station UFSAR Table 15-54 (Page 1 of 1)
(14 OCT 2000)
Table 15-54. Best-Estimate Large Break LOCA - Overall Results Component Blowdown Peak (°F)
First Reflood Peak
(°F)
Second Reflood Peak
(°F)
PCT50%
<1256
<1384
<1512 PCT95%
<1548
<1692
<2028
McGuire Nuclear Station UFSAR Table 15-55 (Page 1 of 3)
(22 APR 2017)
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 H
F < 1.67 [see Note 2]
d)
Hot assembly average linear heat rate PHA 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
McGuire Nuclear Station UFSAR Table 15-55 (Page 2 of 3)
(22 APR 2017)
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
McGuire Nuclear Station UFSAR Table 15-55 (Page 3 of 3)
(22 APR 2017)
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.
Mc Guire Nuclear Station UFSAR Table 15-56 (Page 1 of 1)
(14 OCT 2000)
Table 15-56. Rod Census Used in Best-Estimate large Break LOCA Analysis Rod Group Power Ratio (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
McGuire Nuclear Station UFSAR Table 15-57 (Page 1 of 1)
(30 NOV 2012)
Table 15-57. Deleted Per 2012 Update
McGuire Nuclear Station UFSAR Table 15-58 (Page 1 of 2)
(10 OCT 2009)
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 5%
Early In-Vessel 95%
Gap 5%
Early In-Vessel 35%
Gap 5%
Early In-Vessel 25%
Gap 0%
Early In-Vessel 5%
Gap 0%
Early In-Vessel 2%
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (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
McGuire Nuclear Station UFSAR Table 15-58 (Page 2 of 2)
(10 OCT 2009)
Noble Metals Cerium Group Lanthanides Release Fractions Release Fractions Release Fractions Release Fractions Release Fractions Gap 0%
Early In-Vessel 0.25%
Gap 0%
Early In-Vessel 0.05%
Gap 0%
Early In-Vessel 0.02%
Gap 0%
Early In-Vessel 0.02%
Gap 0%
Early In-Vessel 0.02%
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (Curies)
Nuclide Inventory (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
McGuire Nuclear Station UFSAR Table 15-59 (Page 1 of 1)
(10 OCT 2009)
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
McGuire Nuclear Station UFSAR Table 15-60 (Page 1 of 1)
(10 OCT 2009)
Table 15-60. Isotopic Inventory of the Dry Cask Drop Accident Isotope Ci/Assembly Ci/Cask Release Chemical Form ISG-5 Release Fraction to Cask Cask Release Fraction Release from Pool 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
McGuire Nuclear Station UFSAR Table 15-61 (Page 1 of 1)
(09 OCT 2015)
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 Factor Burndown 15 74 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 Assessments 2086 77 Small Break LOCA; NOTRUMP Analysis of Record PCT (2-inch break) [See Table 15-49]
1323 75 PCT Assessments None 0
74 Current Licensing Basis SBLOCA PCT Including Assessments 1323 74
McGuire Nuclear Station UFSAR Table 15-62 (Page 1 of 1)
(30 NOV 2012)
Table 15-62. Dose Equivalent Iodine-131 (DEI-131)
Isotope Concentration
(µCi/gm)
FGR No. 11, Table 2.1 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
McGuire Nuclear Station UFSAR Table 15-63 (Page 1 of 1)
(30 NOV 2012)
Table 15-63. Dose Equivalent Xenon-133 (DEX-133)
Isotope Concentration
(µCi/gm)
FGR No. 12, Table III.1 DCFs (Sv-s/Bq-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
McGuire Nuclear Station UFSAR Table 15-64 (Page 1 of 1)
(24 APR 2014)
Table 15-64. Deleted Per 2014 Update