Text

9 to Updated Final Safety Analysis Report, Appendix 15A - Tables
	ML19189A273
Person / Time
Site:	Catawba
Issue date:	04/06/2017
From:	; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML17103A316	List: ML17103A307; ML19189A258; ML19189A259; ML19189A260; ML19189A261; ML19189A262; ML19189A263; ML19189A264; ML19189A265; ML19189A266; ML19189A267; ML19189A268; ML19189A269; ML19189A270; ML19189A271; ML19189A272; ML19189A273; ML19189A274; ML19189A275; ML19189A277; ML19189A278; ML19189A279; ML19189A280; ML19189A281; ML19189A282; ML19189A283; ML19189A284; ML19189A285; ML19189A286; ML19189A287; ML19189A288; ML19189A289; ML19189A290; ML19189A291; ML19189A292; ML19189A293; ML19189A294; ML19189A295; ML19189A296; ML19189A298; ML19189A299; ML19189A300; ML19189A301; ML19189A302; ML19189A303; ML19189A304; ML19189A305; ML19189A306; ML19189A307;
References
	CNS-17-016
	Download: ML19189A273 (114)
	v • d • e

Catawba Nuclear Station UFSAR Appendix 15A. Tables Appendix 15A. Tables

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

(22 OCT 2001)

Table 15-1. Deleted Per 1992 Update

Catawba Nuclear Station UFSAR Table 15-2 (Page 1 of 3)

(09 OCT 2016)

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. Unit 1 long term core cooling

2. Unit 2 long term core cooling

3. Unit 1 short 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

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

(09 OCT 2016)

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 Deleted Per 2010 Update 15.4.4 Startup of an Inactive 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. Steam generator overfill

3. DNB analysis

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

(09 OCT 2016)

FSAR Section Description of Transient Summary of Cases Analyzed 15.6.5 Loss of Coolant Accident

1. DECLG CD=1.0, Reference Transient

2. 1.5 inch SBLOCA (Unit 1)

3. 2 inch SBLOCA (Unit 1)

4. 3 inch SBLOCA (Unit 1)

5. 4 inch SBLOCA (Unit 1)

6. 1.5 inch SBLOCA (Unit 2)

7. 2 inch SBLOCA (Unit 2)

8. 3 inch SBLOCA (Unit 2)

9. 4 inch SBLOCA (Unit 2)

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

(09 OCT 2016)

Table 15-3. Summary of Computer Codes and Methodologies Used in Accident Analyses Computer Code or Methodology Transient Numbers1 Analyzed with that Computer Code or Methodology NOTRUMP 15.6.5 LOCTA-IV 15.6.5 LOTIC 15.6.5 WCOBRA/TRAC 15.6.5 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, 15.1.3, 15.1.5, 15.2.7, 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 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3a,d, 15.4.8, 15.6.1, 15.6.3 Deleted Per 2006 Update BWU 15.1.3 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 Deleted Per 2006 Update SIMULATE-3K 15.4.8 Note:

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

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

(09 OCT 2016)

Table 15-4. Summary of Input Parameters for Accident Analyses Using Computer Codes FSAR Section Case Identifier (refer to Table 15-2)

Moderator Temperature Coefficient (pcm/°F)

Moderator Density Coefficient

(%k/k/g/cc)

Doppler Coefficient (pcm/°F)

Initial Core Output (MWt)

RCS Flow (gpm) Note 19 Vessel Tavg (°F)

Pzr Press.

(psia)

Pzr Liquid Inventory (ft3 or %)

Feedwater Temperature

(°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 5 561 2208 16%

Note 17 15.1.5 1

Note 9 NA

-3.5 0

371,796 561 2198 16%

Note 17 15.1.5 2

Note 9 NA

-3.5 0

371,796 561 2198 16%

Note 17 15.1.5 3

NA Note 20

-1.2 3469 390,000 587.5 2250 46%

445 15.1.5 4

NA Note 21

-1.2 3469 388,000 585.1 2250 46%

443 15.2.3 1

NA Note 6

-0.9 3479 381,420 591.5 2280 64%

443 15.2.3 2

NA Note 6

-0.9 3479 420,000 591.5 2310 64%

443 15.2.6 Note 14 NA Note 14 3479 373,596 594.8 2250 61.5%

440 15.2.7 1

NA Note 6

-0.9 3479 381,420 589.1 2208 46%

440 15.2.7 2

NA Note 6

-0.9 3479 376,530 594.8 2208 51%

440 15.2.7 3

NA Note 6

-0.9 3469 384,000 585.1 2250 46%

443 15.2.8 1

NA Note 6

-0.9 3479 373,596(U2)

Note 7(U1) 594.8(U2) 589.1(U1) 2208 52.5%(U2) 46%(U1) 445(U2) 440(U1) 15.2.8 2

NA Note 6

-0.9 3469 388,000 (U1) 390,000 (U2) 585.1(U1) 587.5(U2) 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 375,552 589.1 2310 64%

443

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

(09 OCT 2016)

FSAR Section Case Identifier (refer to Table 15-2)

Moderator Temperature Coefficient (pcm/°F)

Moderator Density Coefficient

(%k/k/g/cc)

Doppler Coefficient (pcm/°F)

Initial Core Output (MWt)

RCS Flow (gpm) Note 19 Vessel Tavg (°F)

Pzr Press.

(psia)

Pzr Liquid Inventory (ft3 or %)

Feedwater Temperature

(°F) 15.3.3 2

NA Note 6

-0.9 3469 388,000 585.1 2250 46%

442 15.3.3 3

NA Note 6

-0.9 3469 388,000 585.1 2250 46%

442 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%

335.7 15.4.2 2

NA Note 6 Note 4 273 375,669 563.8 2250 37%

333.0 15.4.2 3

NA Note 6 Note 4 1734.5 384,120 571 2250 31%

382.3 15.4.2 4

NA Note 6 Note 4 3399.6 384,120 584.5 2250 45.4%

438.3 15.4.2 5

NA Note 6 Note 4 3469 388,000 585.1 2250 46%

440.6 15.4.3a, b

NA Note 6

-0.9 3469 384,000 585.1 2250 46%

443 15.4.3c NA NA NA 3469 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

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

(09 OCT 2016)

FSAR Section Case Identifier (refer to Table 15-2)

Moderator Temperature Coefficient (pcm/°F)

Moderator Density Coefficient

(%k/k/g/cc)

Doppler Coefficient (pcm/°F)

Initial Core Output (MWt)

RCS Flow (gpm) Note 19 Vessel Tavg (°F)

Pzr Press.

(psia)

Pzr Liquid Inventory (ft3 or %)

Feedwater Temperature

(°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 14 NA Note 14 3479 373,599 581.1 2310 64%

440 15.6.3 2

Note 14 NA Note 14 3479 Note 12 571.1 2310 53.3%

440 15.6.3 3

NA Note 8

-0.90 3469 388,000 585.1 2250 46%

442 15.6.5 1

NA Note 11 Note 11 344518 Note 16 587.5 2250 55%

442 15.6.5 (Unit 1) 2 NA Note 11 Note 11 3479 Note 16 585.1 2250 55%

442 15.6.5 (Unit 1) 3 NA Note 11 Note 11 3479 Note 16 585.1 2250 55%

442 15.6.5 (Unit 1) 4 NA Note 11 Note 11 3479 Note 16 585.1 2250 55%

442 15.6.5 (Unit 1) 5 NA Note 11 Note 11 3479 Note 16 585.1 2250 55%

442 15.6.5 (Unit 2) 6 NA Note 11 Note 11 3479 390,000 587.5 2250 55%

442 15.6.5 (Unit 2) 7 NA Note 11 Note 11 3479 390,000 587.5 2250 55%

442 15.6.5 (Unit 2) 8 NA Note 11 Note 11 3479 390,000 587.5 2250 55%

442 15.6.5 (Unit 2) 9 NA Note 11 Note 11 3479 390,000 587.5 2250 55%

442 Notes:

1. The assumed feedwater temperature varies inversely proportional to the overfeed percentage.

2. -0.9 pcm/°F at HFP to -1.20 pcm/°F at HZP

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

(09 OCT 2016)

FSAR Section Case Identifier (refer to Table 15-2)

Moderator Temperature Coefficient (pcm/°F)

Moderator Density Coefficient

(%k/k/g/cc)

Doppler Coefficient (pcm/°F)

Initial Core Output (MWt)

RCS Flow (gpm) Note 19 Vessel Tavg (°F)

Pzr Press.

(psia)

Pzr Liquid Inventory (ft3 or %)

Feedwater Temperature

(°F)

3. -1.04 pcm/°F at HFP to -1.325 pcm/°F at HZP.

4. -1.2 pcm/°F at HFP to -1.5 pcm/°F at HZP.

5. 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.

6. The most positive MTC (most negative MDC) allowed by the Technical Specifications was used.

7. 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.

8. The Catawba 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. An RCS flow of 390,000 gpm is assumed. The results of this transient are not sensitive to RCS flow.

13. Deleted

14. The results of this transient are not sensitive to reactivity feedback assumptions.

15. Deleted Per 2013 Update.

16. 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 the reduced RCS flow.

17. Main feedwater temperature is 60ºF. Auxiliary feedwater temperature is 32ºF.

18. Analysis was originally performed at 3445 MWt (3411 MWt 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 15-82.

19. Evaluations of all accidents for RCS Flows of 384,000 gpm (U1) and 387,000 gpm (U2) concluded that there would be no impact on meeting the relevant acceptance criteria due to the reduced RCS flow.

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

(09 OCT 2016)

FSAR Section Case Identifier (refer to Table 15-2)

Moderator Temperature Coefficient (pcm/°F)

Moderator Density Coefficient

(%k/k/g/cc)

Doppler Coefficient (pcm/°F)

Initial Core Output (MWt)

RCS Flow (gpm) Note 19 Vessel Tavg (°F)

Pzr Press.

(psia)

Pzr Liquid Inventory (ft3 or %)

Feedwater Temperature

(°F)

20. Based on MTC = -17 pcm/°F

21. Based on MTC = - 24 pcm/°F

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

(22 OCT 2001)

Table 15-5. Deleted Per 1992 Update

Catawba Nuclear Station UFSAR Table 15-6 (Page 1 of 1) 09 OCT 2016)

Table 15-6. 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 15.4.7 Note 2 15.4.8 2.2 15.6.1 2.2 15.6.2 Note 2 15.6.3 2.2,(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 the dose and overfill analyses.

2. Reactor trip was not necessary to analyze transient.

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

(22 OCT 2001)

Table 15-7. 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 Note 2 0.5 Power range high neutron flux, low setting 116.1%

0.5 Overtemperature T Variable see Figure 15-1 1.5 Overpower T Variable see Figure 15-1 1.5 High pressurizer pressure Note 2 2.0 Low pressurizer pressure Note 2 2.0 Low reactor coolant flow (from loop flow detectors) 83.5% loop flow 1.0 Undervoltage trip Note 3 1.5 Low-low steam generator level Note 2 2.0 Safety injection Not applicable

2.0 Notes

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.

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

(22 OCT 2001)

Table 15-8. Deleted Per 1992 Update Table 15-9. Deleted Per 1992 Update Table 15-10. Deleted Per 1992 Updaate Table 15-11. Deleted Per 1992 Update

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

(24 APR 2006)

Table 15-12. Fission Product Radioactivity Levels in the Reactor Core*

Noble Gases Tellurium Group Radioisotope Core Activity (Curies)

Radioisotope Core Activity (Curies)

Kr83m 1.56E+07 Se81 5.84E+06 Kr85m 3.40E+07 Se83 7.04E+06 Kr85 1.07E+06 Se83m 7.97E+06 Kr87 6.96E+07 Se84 2.91E+07 Kr88 9.79E+07 Se87 2.11E+07 Kr89 1.25E+08 Sb127 9.65E+06 Xe131m 1.43E+06 Sb128 1.65E+06 Xe133m 6.72E+06 Sb128m 1.59E+07 Xe133 2.08E+08 Sb129 3.43E+07 Xe135m 4.51E+07 Sb130 1.15E+07 Xe135 6.65E+07 Sb130m 4.76E+07 Xe137 1.98E+08 Sb131 8.38E+07 Xe138 1.98E+08 Sb132m 4.95E+07 Te127m 1.58E+06 Halogens Te127 9.51E+06 Core Activity Te129 3.27E+07 Radioisotope (Curies)

Te129m 6.63E+06 Br83 1.55E+07 Te131 8.69E+07 Br85 3.41E+07 Te132 1.49E+08 Br87 5.56E+07 Te133 1.22E+08 I130 2.96E+06 Te133m 1.01E+08 I131 1.04E+08 Te134 2.12E+08 I132 1.52E+08 I133 2.15E+08 I134 2.47E+08 I135 2.06E+08

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

(24 APR 2006)

Noble Gases Tellurium Group Alkali Metals Alkali Earth Metals Core Activity Core Activity Radioisotope (Curies)

Radioisotope (Curies)

Rb86 2.08E+05 Sr89 1.03E+08 Rb88 1.00E+08 Sr90 9.31E+06 Rb89 1.33E+08 Sr91 1.66E+08 Rb90 1.25E+08 Sr92 1.69E+08 Cs134 2.09E+07 Sr93 1.83E+08 Cs136 5.60E+06 Ba139 2.00E+08 Cs137 1.26E+07 Ba140 1.88E+08 Cs138 2.09E+08 Ba141 1.82E+08 Cs139 1.96E+08 Ba142 1.78E+08

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

(24 APR 2006)

Noble Metals Lanthanides Core Activity Core Activity Radioisotope (Curies)

Radioisotope (Curies)

Mo99 1.97E+08 Y90 9.66E+06 Mo101 1.76E+08 Y91 1.34E+08 Mo102 1.70E+08 Y91m 9.72E+07 Tc99m 1.74E+08 Y92 1.51E+08 Tc101 1.76E+08 Y93 1.23E+08 Tc104 1.46E+08 Y94 1.90E+08 Ru103 1.72E+08 Y95 1.94E+08 Ru105 1.25E+08 Zr95 1.78E+08 Ru106 6.37E+07 Zr97 1.78E+08 Ru107 7.57E+07 Nb95 1.79E+08 Rh103m 1.72E+08 Nb95m 1.98E+06 Rh105 1.12E+08 Nb97 1.78E+08 Rh106m 4.02E+06 La140 1.98E+08 Rh107 7.59E+07 La141 1.81E+08 Pd109 4.67E+07 La142 1.82E+08 Pd111 7.22E+06 La143 1.79E+08 Pd112 3.24E+06 Nd147 6.93E+07 Nd149 4.04E+07 Cerium Group Nd151 2.16E+07 Core Activity Pm147 1.75E+07 Radioisotope (Curies)

Pm148 1.88E+07 Ce141 1.73E+08 Pm148m 2.96E+06 Ce143 1.79E+08 Pm149 6.64E+07 Ce144 1.32E+08 Pm151 2.18E+07 Ce145 1.21E+08 Sm153 5.73E+07 Ce146 9.35E+07 Sm156 2.75E+06 Np237 4.23E+01 Eu154 9.87E+05 Np238 5.04E+07 Eu155 3.86E+05 Np239 2.32E+09 Eu156 3.17E+07

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

(24 APR 2006)

Cerium Group Core Activity Core Activity Radioisotope (Curies)

Radioisotope (Curies)

Eu157 3.66E+06 Np240 7.00+06 Pr142 7.74E+06 Pu236 7.32E+01 Pr143 1.56E+08 Pu238 4.29E+05 Pr144 1.33E+08 Pu239 3.74E+04 Pr144m 1.86E+06 Pu240 5.16E+04 Pr145 1.21E+08 Pu241 1.45E+07 Pr146 9.40E+07 Pu242 2.97E+02 Pr147 7.22E+07 Pu243 5.62E+07 Am241 1.75E+04 Am242m 1.14E+03 Am242 8.95E+06 Am243 4.41E+03 Am244 2.45E+07 Cm242 5.13E+06 Cm244 9.41E+05

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

(24 OCT 2004)

Table 15-13. Reactor Coolant Specific Activities for Iodine and Noble Gas Isotopes Nuclide Specific Activity1 (µCi/g)

I-131 2.5 I-132 0.9 I-133 4.0 I-134 0.6 I-135 2.2 Xe-131m 2.3 Xe-133m 17.5 Xe-133 278.2 Xe-135m 0.49 Xe-135 7.4 Xe-138 0.66 Kr-83m 0.47 Kr-85m 2.1 Kr-85 7.5 Kr-87 1.3 Kr-88 3.7 Kr-89 0.0 Note:

1. Reactor coolant concentrations at equilibrium assuming 1 percent failed fuel.

Catawba Nuclear Station UFSAR Table 15-14 (Page 1 of 2)

(09 OCT 2016)

Table 15-14. Total Effective Dose Equivalents (TEDEs - Rem) Following Design Basis Events Design Basis Event UFSAR Section Exclusion Area Boundary Low Population Zone Boundary Control Room Main Steam Line Break 15.1.5.3 Pre-existent iodine spike 0.13 0.03 Accident initiated iodine spike 0.14 0.13 Locked Rotor Accident 15.3.3.3 1.63 0.31 1.56 Rod Ejection Accident 15.4.8.3 4.75 3.37 2.70 Instrument Line Break 15.6.2 Pre-existent iodine spike 0.60 0.09 0.80 Accident initiated iodine spike 0.17 0.02 0.18 Steam Generator Tube Rupture (1) 15.6.3.3 Pre-existent iodine spike 1.19 0.28 1.32 Accident initiated iodine spike 0.61 0.19 0.81 Loss of Coolant Accident 15.6.5.3 8.79 3.78 3.31 Waste Gas Tank Rupture 15.7.1 0.27 0.04 0.10 Liquid Storage Tank Rupture 15.7.2 0.99 0.14 1.00 Fuel Handling Accident (3,4) 15.7.4 1.76 2.59 Weir Gate Drop (4) 15.7.4 2.68 4.24 Fuel Cask Drop (4) 15.7.5 0.006 0.001 Notes:

1) A supplemental analysis of the steam generator tube rupture is reported in Section 15.6.3.3. It is the basis for a set of conditions cited in Facility Operating License Amendment 159/151. The license conditions set limits of 0.46 µCi/gm Dose Equivalent Iodine-131 (DEI) for equilibrium iodine specific activity in the reactor coolant and 26 DEI for transient iodine specific activity in the reactor coolant. Thyroid radiation doses are calculated in this supplemental analysis and reported in Table 15-74.

Catawba Nuclear Station UFSAR Table 15-14 (Page 2 of 2)

(09 OCT 2016)

2) This footnote is reserved for future use.

3) Section 15.7.4 reports the analysis of fuel handling accident in containment and a fuel handling accident in the fuel building. The radiation doses calculated for offsite locations and in the control room take the same values.

4) TEDEs at the boundary of the Low Population Zone (denoted as the LPZ) are not listed for the fuel handling accidents, weir gate drop, or fuel cask drop. The regulatory acceptance criteria in 10 CFR 50.67 and Regulatory Guide 1.183 for radiation doses at the Exclusion Area Boundary and LPZ are the same. In addition, for the fuel handling accidents and weir gate drop, the TEDEs at the LPZ are significantly lower than the TEDEs at the Exclusion Area Boundary.

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

(09 OCT 2016)

Table 15-15. 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

Overpower 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 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 Peak heat flux occurs 119 NV injection lines purged of unborated water 119 One train of SI fails 119

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

(09 OCT 2016)

Accident Event Time (sec.)

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

3. CFM at hot full power Break occurs 0

High flux trip setpoint reached 12.8 Reactor trip occurs due to hight flux trip 13.3 Peak reactor power occurs 13.5 Turbine trip occurs due to reactor trip 13.6 Loss of offsite power occurs on turbine trip 13.6 RCPs trip due to loss of offsite power 13.6

4. DNB at hot full power Break occurs 0

OPT trip setpoint reached 11.6 Reactor trip occurs due to OPT trip 12.1 Peak reactor power occurs 12.3

Catawba Nuclear Station UFSAR Table 15-15 (Page 3 of 3)

(09 OCT 2016)

Accident Event Time (sec.)

Turbine trip occurs due to reactor trip 12.4 Loss of offsite power occurs on trubine trip 12.4 RCPs trip due to loss of offsite power 12.4 MDNBR occurs 13.2

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

(22 OCT 2001)

Table 15-16. Deleted Per 1992 Update

Catawba Nuclear Station UFSAR Table 15-17 (Page 1 of 2)

(17 OCT 2013)

Table 15-17. Parameters for Postulated Main Steam Line Break Offsite Dose Analysis

1. Data and assumptions pertaining to the radioactive source term

a. Equilibrium reactor coolant DEX specific activity (µCi/gm -

Note 1) 670

b. Equilibrium reactor coolant DEI specific activity - concurrent iodine spike (µCi/gm - Note 2) 1

c. Concurrent iodine spike multiplier for the equilibrium reactor coolant iodine appearance rate 500

d. Maximum reactor coolant DEI specific activity-pre-existent iodine spike (µCi/gm - Note 2) 60

e. Steam generator secondary side DEI specific activity (µCi/gm) 0.1

f. Main condenser iodine scrubbing efficiency (%)

100

g. Iodine composition fractions (%)

Diatomic iodine 97 Organic iodine compounds 3

2. Data and assumptions pertaining to transport and release of radioactivity

a. Offsite power available?

No

b. Initial steam release from the faulted steam generator (lbm)

Unit 1 (0-600 sec) 234,000 Unit 2 (0-400 sec) 165,000

c. Primary-to secondary leak rate (gpd per SG) 150

d. Limiting time spans for tube bundle uncovery in the intact steam generators (hr)

(Unit 1) 0-2.5 (Unit 2) 0-2.5

e. Iodine partition fraction for steam releases 0.01

f. Integrated steam release from the intact steam generators (lbm, 0-5.5 hr)

Unit 1 0- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.35 E+05 0-5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 1.85E+06 Unit 2 0- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.22E+05 0- 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 1.81E+06

3. Dispersion data

a. /Q at exclusion area boundary (sec/m3) 4.78x10-4

b. /Q at low population zone (sec/m3)

Catawba Nuclear Station UFSAR Table 15-17 (Page 2 of 2)

(17 OCT 2013) 0- 8 hr 6.85x10-5 8-24 hr 4.00x10-5 24-96 hr 2.00x10-5 96-720 hr 7.35x10-6

4. Dose conversion data

a. Source of dose coefficients Deep dose equivalent coefficent FGR 12 Committed effective dose equivalent coefficient FGR 11

5.

Total Effective Dose Equivalents (TEDEs, Rem)

a.

Exclusion Area Boundary Main steam line break with pre-existent iodine spike 0.13 Main steam line break with concurrent iodine spike 0.14

b.

Boundary of the Low Population Zone Main steam line break with pre-existent iodine spike 0.03 Main steam line break with concurrent iodine spike 0.13 Notes:

1) DEX denotes Dose Equivalent Xenon-133. The isotpic activities corresponding to this value are presented in Table 15-83.

2) DEI denotes Dose Equivalent Iodine-131. The isotopic activities corresponding to this value are presented in Table 15-84.

Catawba Nuclear Station UFSAR Table 15-18 (Page 1 of 3)

(17 OCT 2013)

Table 15-18. 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 Case Turbine Trip, loss of main feed flow 0.0 Pressurizer PORVs lift 3.3 Steam Safety Valves lift 6.3 Overtemperature T setpoint reached 12.9 Control rod insertion begins 14.4 Peak secondary system pressure occurs 14.8

2. Maximum Primary System Pressure Case Turbine Trip, loss of main feed flow 0.0 High pressurizer pressure setpoint reached 4.8 Control rod insertion begins 6.8 Pressurizer Safety Valves lift 7.0 Peak primary system pressure occurs 7.5 Steam Safety Valves lift 8.1

3. Deleted

4. Deleted 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 coast down 0.1 Rods begin to drop 0.6 Peak water level in pressurizer occurs 4

Flow from two motor driven auxiliary feedwater pumps is started 60 Feedwater lines are purged and cold auxiliary feedwater is delivered to four steam generators 195 Core decay heat decreases to auxiliary feedwater heat removal capacity

~ 1200 Loss of Normal Feedwater Flow

1. Unit 1 Long-Term Core Cooling Case Main feedwater flow stops 1.0 Pressurizer PORVs begin cycling 40.2

Catawba Nuclear Station UFSAR Table 15-18 (Page 2 of 3)

(17 OCT 2013)

Accident Event Time (Sec)

Low-Low steam generator level reactor trip reached 57.7 Rods begin to drop 59.7 Steam safety valves lift 61.1 Auxiliary feedwater flow on 117.7 Core decay heat plus pump heat decreases to auxililary feedwater heat removal capacity

~1290

2. Unit 2 Long-Term Core Cooling Case Main feedwater flow stops 0.0 Pressurizer sprays on 10.69 Pressurizer PORVs begin cycling 27.92 Low-low steam generator level reactor trip reached 44.89 Rods begin to drop 46.89 Steam safety valves lift 49.47 Pressurizer sprays off 50.80 Auxiliary feedwater flow on 104.89 Core decay heat plus pump heat decreases to auxiliary feedwater heat removal capacity

~230

3. Unit 1 Short-Term Core Cooling Case Main feedwater flow stops 0.0 PORVs begin cycling 24.6 Low-low steam generator level reactor trip reached 57.0 Minimum DNBR occurs 58.9 Rods begin to drop 59.0 Feedwater System Pipe Break (Unit 2 Analysis)

Feedwater line break to SG B 0

Safety injection on high containment pressure 10.10 Reactor trip on high containment pressure SI 10.10 Turbine trip on reactor trip 10.3 Steam line isolation on hi-hi containment pressure 15

Catawba Nuclear Station UFSAR Table 15-18 (Page 3 of 3)

(17 OCT 2013)

Accident Event Time (Sec)

Reactor coolant pumps tripped 15 Safety injection terminated 70 Motor-driven auxiliary feedwater pump delivers flow 75.3 Auxiliary feedwater to faulted generator isolated 120 SG B boiled dry 130 Core decay heat decreases to auxiliary feedwater heat removal capacity

~1655 End of simulation 2000 Feedwater System Pipe Break (Unit 1 Analysis)

Feedwater line break to SG B 0

Safety injection on high containment pressure 10.05 Reactor trip on high containment pressure SI 10.05 Reactor coolant pumps tripped 10.05 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

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

(21 OCT 2010)

Table 15-19. Deleted Per 2010 Update

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

(09 OCT 2016)

Table 15-20. 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 1.47 Rods begin to drop 2.47 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) (U1)

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 (Core Cooling Capability for Offsite Power Lost) (U2)

Rotor on one pump locks 0.0 Low flow reactor trip setpoint reached 0.0 4

Rods begin to drop 1.0 4

Minimum DNBR occurs 2.9 Reactor Coolant Pump Shaft Seizure (Peak RCS Pressure)

Rotor on one pump locks 0.0 Low flow reactor trip setpoint reached 0.07 Rods begin to drop 1.07 Maximum RCS pressure occurs 4.7

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

(22 OCT 2001)

Table 15-21. Deleted Per 1992 Update

Catawba Nuclear Station UFSAR Table 15-22 (Page 1 of 3)

(21 OCT 2010)

Table 15-22. Parameters for the Postulated Locked Rotor Accident

1.

Data pertaining to the radioactive source term

a.

Percent fuel with clad failure (%)

Deleted Per 2006 Update Deleted Per 2006 Update Offsite power available 1

Offsite power lost 5.4

b.

Offsite power available?

No

c.

Core isotopic inventory Table 15-73

d.

Iodine composition fraction (%)

Elemental iodine 97 Particulate iodine 0

Organic iodine compounds 3

e.

Fission product gap fraction (%)

Section 15.0

2.

Data and assumptions pertaining to transport and release of radioactivity

a.

Fraction of gap inventory released to the reactor coolant (%)

100

b.

Initial primary-to-secondary leak rate (gpd per SG) 150

c.

Time spans of SG tube bundle uncovery (seconds)

Offsite power available S/G 1 895 S/G 2 46 S/Gs 3 & 4 a piece 9087 Total for Unit 1 19,115 Offsite power lost S/G 1 617 S/G 2 0

S/Gs 3 & 4 a piece 3073 Total for Unit 2 6763

d.

SG iodine partition fraction 0.01

e.

Duration of post accident unit cooldown (hours) 6

f.

Total steam released (lbm)

Unit 1 0-2 hours 748,330

Catawba Nuclear Station UFSAR Table 15-22 (Page 2 of 3)

(21 OCT 2010) 2-6 hours 989,069 Unit 2 0-2 hours 736,434 2-6 hours 966,906

g.

Power level (MWt) 3479.

3.

Dispersion data

a.

/Q at exclusion area boundary (sec/m3) 4.78x10-4

b.

/Q at low population zone (sec/m3) 6.85x10-5

c.

/Q for the control room (sec/m3, Note 2)

Releases from the outboard SG doghouse vents 0 -

2 hr 7.14x10-3 2 -

8 hr 4.05x10-3 8 -

10 hr 2.24x10-3 10 -

24 hr 1.81x10-3 24 -

96 hr 1.24x10-3 96 -

720 hr 7.26x10-4 Releases from the inboard SG doghouse steam vents 0 -

2 hr 2.27x10-3 2 -

8 hr 1.74x10-3 8 -

10 hr 1.02x10-3 10 -

24 hr 8.70x10-4 24 -

96 hr 7.14x10-4 96 -

720 hr 5.74x10-4

4.

Dose conversion data

a.

Method of conversion from activity to dose R.G. 1.183

b.

Source of dose conversion factors Whole body radiation dose coefficient FGR 12 Origin radiation dose coefficient FGR 11

5.

Data pertaining to the control room and Control Room Area Ventilation System Table 15-41

Catawba Nuclear Station UFSAR Table 15-22 (Page 3 of 3)

(21 OCT 2010)

6.

Total effective dose equivalents (Rem)

a.

Exclusion Area Boundary 1.63

b.

Low Population Zone 0.31

c.

Control Room 1.56 Note on Table 15-22

1) The values for the control X/Qs shown here correspond to both control room outside air intakes being open during the event.

Catawba Nuclear Station UFSAR Table 15-23 (Page 1 of 4)

(09 OCT 2016)

Table 15-23. 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 Valve Lifts 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 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

Catawba Nuclear Station UFSAR Table 15-23 (Page 2 of 4)

(09 OCT 2016)

Accident Event Time (sec.)

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

<1034 b) Dilution during power operation (automatic rod control)

Dilution begins Rod insertion limit alarm setpoint reached 0

Operator terminates dilution

<1611

2. Dilution during startup Dilution begins Reactor trip setpoint reached 0

Operator terminates dilution

<1034

3.

a) Dilution during hot standby (BDMS operable)

Dilution begins 0

BDMS setpoint reached 2355 Dilution source isolated 2380 Borated water reaches core 2657 b) Dilution during hot standby (BDMS inoperable)

Dilution begins 0

High-flux-at-shutdown alarm setpoint reached 5771 Operator terminates dilution 6671

4.

a) Dilution during hot shutdown without an RCP in operation (BDMS operable)

Dilution begins 0

BDMS setpoint reached 2021 Dilution source isolated 2046

Catawba Nuclear Station UFSAR Table 15-23 (Page 3 of 4)

(09 OCT 2016)

Accident Event Time (sec.)

Borated water reaches core

<2262 b) Dilution during hot shutdown without an RCP in operation (BDMS inoperable)

Dilution begins 0

High-flux-at-shutdown alarm setpoint reached 5594 Operator terminates dilution

<6494 c) Dilution during cold shutdown with an RCP in operation (BDMS operable)

Dilution begins 0

BDMS setpoint reached 2134 Dilution source isolated 2159 Borated water reaches core

<2397 d) Dilution during cold shutdown witth an RCP in operation (BDMS inoperable)

Dilution begins 0

High-flux-at-shutdown alarm setpoint reached 5923 Operator terminates dilution

<6823

5.

a) Dilution during cold shutdown without an RCP in operation (BDMS operable)

Dilution begins 0

BDMS setpoint reached 2029 Dilution source isolated 2054 Borated water reaches core 2278 b) Dilution during cold shutdown without an RCP in operation (BDMS inoperable)

Dilution begins 0

High-flux-at-shutdown alarm setpoint reached 5634 Operator terminates dilution

<6534 c) Dilution during cold shutdown with an RCP in operation (BDMS operable)

Dilution begins 0

BDMS setpoint reached 2143 Dilution source isolated 2168 Borated water reaches core

<2418 d) Dilution during cold shutdown witth an RCP in operation (BDMS inoperable)

Dilution begins 0

High-flux-at-shutdown alarm setpoint reached 5969 Operator terminates dilution

<6869 Rod Cluster Control Assembly Ejection

1. Beginning of Life, Full Power Initiation of rod ejection 0.0 Power range high neutron flux high setpoint reached 0.056

Catawba Nuclear Station UFSAR Table 15-23 (Page 4 of 4)

(09 OCT 2016)

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

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

(22 OCT 2001)

Table 15-24. Deleted Per 1992 Update

Catawba Nuclear Station UFSAR Table 15-25 (Page 1 of 1)

(09 OCT 2016)

Table 15-25. Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident Time in Cycle Beginning Beginning End End Power Level, %

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 4.75 19.60 4.84 20.78 Number of operational pumps 4

3 4

3 Max. fuel pellet average temperature, °F 3313 1626 2850 1316 Max. fuel center temperature, °F 5021 2064 4353 1646 Max. clad temperature, °F 794 738 1296 755 Max. fuel stored energy, cal/gm 146 61 133 48

% Failed fuel

<50

Catawba Nuclear Station UFSAR Table 15-26 (Page 1 of 6)

(17 APR 2012)

Table 15-26. Parameters for the Postulated Rod Ejection Accident

1.

Data pertaining to the radioactive sources term

a. Percent fuel with clad failure 40

b. Percent fuel melted 0

c. Core isotopic inventory Table 15-73

d. Iodine composition fraction (%)

Elemental iodine 97 Particulate iodine 0

Organic iodine compounds 3

e. Fission product gap fraction (%)

Alkali metals 12 All other fission products 10

f. Fractions for source term release to be containment (%)

Noble gas containment release fraction 100 Iodine containment release fraction 100

g. Fractions for source term release to the containment sump (%)

Noble gas containment release fraction 0

Iodine containment release fraction 100

h. Fractions for source term release to the reactor coolant (%)

Noble gas containment release fraction 100 Iodine containment release fraction 100

2.

Data and assumptions pertaining to transport and release of radioactivity

a. Containment volume (cu. ft. - Note 1) 1,016,454

b. Containment compartment volume (cu. ft.)

Lower compartment (Note 2) 346,895 Upper compartment 669,559

c. Containment leak rate (% air mass per day) 0 hr t 24 hr 0.3 t > 24 hr 0.15

d. Bypass leakage fraction (% containment leak rate) 7

e. Annulus volume (cu. ft. - Note 3) 484,090

f. Annulus ventilation total airflow rate (cfm))

8,100

g. Annulus ventilation response time (sec) 23

Catawba Nuclear Station UFSAR Table 15-26 (Page 2 of 6)

(17 APR 2012)

h. Annulus ventilation filter efficiencies and bypass fraction (%)

Elemental iodine 92 Particulate iodine forms 95 Organic iodine compounds 92 Bypass airflow fraction 1

i.

Containment sump volume (cu. ft.)

80,160

j.

Rate of ESF leakage in the Auxiliary Building (gpm) 1 Initially filtered 0.5 Initially unfiltered 0.5

k. Auxiliary Building Ventilation System filter efficiencies (%)

Elemental iodine 92 Particulate iodine forms 95 Organic iodine compounds 92 Bypass airflow percent 1

l.

Time at which the VA System is aligned to all rooms with ESF equipment (hours) 72

m.

Rate of ESF backleakage to the Refueling Water Storage Tank (gpm) 10 (Note 7)

n.

Iodine partition fractions for ESF backleakage to the RWST 0 - 2 hr 0.000 2 - 8 hr 3.135x10-6 8 - 10 hr 1.266x10-5 10 - 24 hr 2.332x10-5 24 - 96 hr 8.910x10-4 96 - 720 hr 2.415x10-2

o.

Characteristics of natural deposition of fission products in containment Natural deposition time constants (hr-1) 0.00 - 0.50 hr 0.02801 0.50 - 1.80 hr 0.05713 1.80 - 3.80 hr 0.06502 3.80 - 11.80 hr 0.09151 11.80 - 13.80 hr 0.09146 13.80 - 22.22 hr 0.09146 22.22 - 27.78 hr 0.03770 27.78 - 33.33 hr 0.02770

Catawba Nuclear Station UFSAR Table 15-26 (Page 3 of 6)

(17 APR 2012) 33.33 - 720.00 hr 0.00000 Natural deposition decontamination factor limits 0.00 - 0.50 hr 1.0134 0.50 - 1.80 hr 1.0944 1.80 - 3.80 hr 1.3220 3.80 - 11.80 hr 1.3220 11.80 - 13.80 hr 3.9270 13.80 - 22.22 hr 3.9270 22.22 - 27.78 hr 8.2920 27.78 - 33.33 hr 8.2920 33.33 - 720.00 hr 1.0000

p. Offsite power available?

Yes (Note 4)

q. Initial primary to secondary leak rate (gpd per SG - Note 5) 150

r. Time spans of SG tube bundle uncovery (seconds)

Unit 1 S/G 1 751 S/G 2 662 S/Gs 3 & 4 a piece 2458 Total for Unit 1 6329 Unit 2 S/G 1 2425 S/G 2 2605 S/Gs 3 & 4 a piece 810 Total for Unit 2 6650

s. SG iodine partition fraction 0.01

t.

Duration of post accident cooldown (hr) 6

u. Total steam released (lbm)

Unit 1 0-2 hours 748,330 2-6 hours 989,069 Unit 2 0-2 hours 736,434 2-6 hours 966,906

Catawba Nuclear Station UFSAR Table 15-26 (Page 4 of 6)

(17 APR 2012)

v. Power level (MWt) 3479.

3.

Dispersion data

a. /Q at exclusion area boundary (sec/m3) 4.78x10-4

b. /Q at low population zone (sec/m3) 0- 8 hr 6.85x10-5 8-24 hr 4.00x10-5 24-96 hr 2.00x10-5 96-720 hr 7.35x10-6

c.

Control Room /Q values (sec/m3, Note 8)

Releases from the steam generator doghouse steam vents (Note 9) 0 - 2 hr 7.14x10-3 2 - 8 hr 4.05x10-3 8 - 10 hr 2.24x10-3 10 - 24 hr 1.81x10-3 24 - 96 hr 1.24x10-3 96 - 720 hr 7.26x10-4 Containment leakage and ESF leakage in the Auxiliary Building 0 - 2 hr 1.04x10-3 2 - 8 hr 8.82x10-4 8 - 10 hr 4.14x10-4 10 - 24 hr 3.68x10-4 24 - 96 hr 2.67x10-4 96 - 720 hr 1.87x10-4 ESF backleakage to the FWST 0 - 2 hr 1.26x10-3 2 - 8 hr 9.78x10-4 8 - 10 hr 5.30x10-4 10 - 24 hr 4.10x10-4 24 - 96 hr 2.68x10-4 96 - 720 hr 1.88x10-4

4.

Data pretaining to the control room and Control Room Area Ventilation System Table 15-41

5.

Dose conversion data

Catawba Nuclear Station UFSAR Table 15-26 (Page 5 of 6)

(17 APR 2012)

a. Method of conversion to dose R.G. 1.183

b. Source of dose conversion factors External dose FGR 12 Inhaled dose FGR 11

6.

Total Effective Dose Equivalents (Rem)

a.

Exclusion area boundary SG releases 2.44 Containment leakage 2.30 Total ESF leakage 0.53 Total (Note 10) 4.75

b.

Low population zone SG releases 0.39 Containment leakage 2.43 Total ESF leakage 2.12 Total (Note 10) 2.82

c.

Control room SG releases 1.31 Containment leakage 1.39 Total ESF leakage 1.22 Total (Note 10) 2.70 Notes on Table 15-26

1) The containment volume and compartment volumes do not include the volume of the ice condenser.

2) The lower compartment volume includes the volume of the dead ended compartments.

3) It is assumed that the airflow returned to the annulus by the Annulus Ventilation System mixes with the air in half of the volume of the annulus or 242,045 cu. ft.

4) For the purposes of calculation of steam releases, offsite power is assumed to be available and the main coolant pumps in operation. Steam and radioactivity are assumed to be released to the environment via the steam generator power operated relief valves. This is consistent with assumed unavailability of the condenser dump valves and the modulated atmospheric dump valve which is consistent with assumed loss of offsite power.

5) Time dependent primary to secondary leak rates were calculated with the use of RETRAN02.

6) The radiation doses for a rod ejection accident postulated at Unit 2 are presented given that they are higher than the radiation doses for a rod ejection accident postulated at Unit 1.

7) This is the limiting value set for rate of ESF backleakage to the FWST. For the AST analysis of the design basis LOCA, this rate was set to 20 gpm.

8) Both control room outside air intakes are assumed to be open with a 60/40 split in airflow.

Catawba Nuclear Station UFSAR Table 15-26 (Page 6 of 6)

(17 APR 2012)

9) These values correspond to the limiting release point on a outboard steam generator doghouse.

10) The higher of the radiation dose constituents from containment and ESF leakage is added to the radiation dose constituent from steam generator releases.

Catawba Nuclear Station UFSAR Table 15-27 (Page 1 of 1)

(22 OCT 2001)

Table 15-27. Deleted Per 1995 Update

Catawba Nuclear Station UFSAR Table 15-28 (Page 1 of 1)

(24 APR 2006)

Table 15-28. Time Sequence of Events For Inadvertent Opening of a Pressurizer Safety Valve Event Time (sec)

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

Catawba Nuclear Station UFSAR Table 15-29 (Page 1 of 2)

(17 OCT 2013)

Table 15-29. Parameters for Postulated Instrument Line Break Offsite Dose Analysis

1.

Data and assumptions pertaining to the radioactive source term

a.

Reactor coolant DEX specific activity (µCi/gm - Note 1) 670

b.

Equilibrium Reactor coolant DEI specific activity - concurrent iodine spike (µCi/gm - Note 2) 1

c.

Concurrent iodine spike multiplier for the equilibrium reactor coolant iodine appearance rate 335

d.

Maximum reactor coolant DEI specific activity pre-existent iodine spike (µCi/gm - Note 2) 60

e.

Iodine composition fractions (%)

Diatomic iodine 97 Organic iodine compounds 3

2.

Data and assumptions pertaining to acttivity release

a.

Break flow rate (gpm referenced at standard conditions) 150

b.

Break flow iodine partition fraction 0.1

c.

Time to isolate break (minutes) 30

3.

Dispersion data

a.

/Q at Exclusion Area Boundary (sec/m3) 4.78x10-4

b.

/Q at the boundary of the Low Population Zone (sec/m3) 0 hr 8 hr 6.85x10-5 8 hr - 24 hr 4.00x10-5 24 hr - 96 hr 2.00x10-5 96 hr - 720 hr 7.35x10-6

c.

Control room /Q (sec/m3) 0 hr - 2hr (Note 3) 1.74x10-3 0 hr - 2hr (Note 4) 1.04x10-3 2 hr - 8hr (Note 3) 1.47x10-3 2 hr - 8hr (Note 4) 8.82x10-4 8 hr - 10hr (Note 3) 6.90x10-4 8 hr - 10hr (Note 4) 4.14x10-4 10 hr - 24 hr (Note 5) 3.74x10-4 24 hr - 96 hr 2.67x10-4 96 hr - 720 hr 1.87x10-4

Catawba Nuclear Station UFSAR Table 15-29 (Page 2 of 2)

(17 OCT 2013)

4.

Data pertaining to the control room and Control Room Area Ventilation System Table 15-41

a.

Time to start redundant pressurized filter train - failure of the on-line pressurized filter train (minutes) 30

5.

Dose Data

a.

Source of dose coefficients Deep Dose Equivalent coefficients FGR 11 Committed Dose Equivalent and Committed Effective Dose Equivalent (CEDE) coefficients FGR 12

b.

Source of CEDE coefficients for defining the DEI source term FGR 11

c.

Total effective dose coefficients (TEDEs, Rem)

Exclusion Area Boundary Pre-existent iodine spike 0.60 Concurrent iodine spike 0.17 Boundary of the Low Population Zone Pre-existent iodine spike 0.09 Concurrent iodine spike 0.02 Control room (Note 6)

Pre-existent iodine spike 0.80 Concurrent iodine spike 0.18 Notes:

1) The term DEX denotes Dose Equivalent Xenon-133. The isotopic activities corresponding to this value are presented in Table 15-83.

2) The term DEI denotes Dose Equivalent Iodine-131. The isotopic activities corresponding to this value are presented in Table 15-84.

3) This control room X/Q value is used for the instrument line break scenarios with one control room outside air intake initially closed.

4) This control room X/Q value is used for the instrument line break scenarios with failure of the on-line Control Room Area Ventilation pressurized filter train. Both control room outside air intakes initially are open.

5) Regardless of the scenario, both control room outside air intakes are open beginning at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the initiating event.

6) The limiting scenario includes failure of the on-line Control Room Area Ventilation pressurized filter train.

Catawba Nuclear Station UFSAR Table 15-30 (Page 1 of 1)

(22 OCT 2001)

Table 15-30. Deleted Per 1990 Update

Catawba Nuclear Station UFSAR Table 15-31 (Page 1 of 3)

(17 OCT 2013)

Table 15-31. Parameters for Postulated Steam Generator Tube Rupture Dose Analysis

1.

Data and assumptions pertaining to the radioactive source term

a.

Reactor coolant DEX specific activity (µCi/gm - Note 1) 670

b.

Equilibrium Reactor coolant DEI specific activity - concurrent iodine spike (µCi/gm - Note 2) 1

c.

Concurrent iodine spike multiplier for the equilibrium reactor coolant iodine appearance rate 335

d.

Maximum reactor coolant DEI specific activity pre-existent iodine spike (µCi/gm) 60

e.

Steam generator secondary side DEI specific activity (µCi/gm) 0.1

f.

Preaccident main condenser iodine scrubbing efficiency (%, Note

3) 100

g.

Iodine composition fractions (%)

Diatomic iodine 97 Organic iodine compounds 3

2.

Data and assumptions pertaining to activity released

a.

Post accident condenser scrubbing efficiency (%, Note 4) 85

b.

Main Feedwater isolation At unit trip

c.

Steam generator iodine partition factor 100

d.

Integrated break flow (lbm, Note 5) 355,000

e.

Integrated steam releases from the ruptured SG (lbm, Note 5)

To the condenser (before trip) 1,110,000 To the atmosphere (after trip) 259,000

f.

Integrated steam releases from the intact SGs (lbm, Note 5)

To the condenser (before trip) 3,330,000 To the atmosphere (after trip) 1,850,000

g.

Intact steam generator tube leak rate (gpm per steam generator) 0.104

3.

Dispersion data

a.

/Q at Exclusion Area Boundary (sec/m3) 4.78x10-4

b.

/Q at the boundary of the Low Population Zone (sec/m3.)

0 hr 8 hr 6.85x10-5 8 hr - 24 hr 4.00x10-5 24 hr - 96 hr 2.00x10-5 96 hr - 720 hr 7.35x10-6

Catawba Nuclear Station UFSAR Table 15-31 (Page 2 of 3)

(17 OCT 2013)

c.

Control room /Q (sec/m3 Releases from the ruptured steam generator (Note 6) 0 hr - trip (Note 7) 1.74x10-3 0 hr - trip (Note 8) 1.04x10-3 trip - 2 hr Note 7) 1.19x10-2 trip - 2 hr (Note 8) 7.14x10-3 2 hr - 8 hr (Note 7) 6.75x10-3 2 hr - 8 hr (Note 8) 4.05x10-3 8 hr - 10 hr (Note 7) 3.74x10-3 8 hr - 10 hr (Note 8) 2.24x10-3 10 hr - 24 hr (Note 9) 1.81x10-3 24 hr - 96 hr 1.24x10-3 96 hr - 720 hr 7.26x10-4 Releases from the intact steam generators (Note 6) 0 hr - trip (Note 7) 1.74x10-3 0 hr - trip (Note 8) 1.04X10-3 trip - 2 hr (Note 7) 6.49x10-3 trip - 2 hr (Note 8) 3.89x10-3 2 hr - 8 hr (Note 7) 4.18X10-3 2 hr - 8 hr (Note 8) 2.51x10-3 8 hr - 10 hr (Note 7) 2.38x10-3 8 hr - 10 hr (Note 8) 1.43x10-3 10 hr - 24 hr (Note 7) 1.18x10-3 24 hr - 96 hr 8.89x10-4 96 hr - 720 hr 6.25x10-4

4.

Data pertaining to the control room and Control Room Area Ventilation System Table 15-41

5.

Dose Data

a.

Source of dose coefficients Deep Dose Equivalent coefficients FGR 12 Committed Dose Equivalent and Committed Effective Dose Equivalent (CEDE) coefficients FGR 11

b.

Source of CEDE coefficients for defining the DEI source term FGR 11

c.

Total effective dose coefficients (TEDEs, Note 11)

Catawba Nuclear Station UFSAR Table 15-31 (Page 3 of 3)

(17 OCT 2013)

Exclusion Area Boundary Pre-existent iodine spike 1.19 Concurrent iodine spike 0.61 Boundary of the Low Population Zone Pre-existent iodine spike 0.28 Concurrent iodine spike 0.19 Control room Pre-existent iodine spike 1.32 Concurrent iodine spike 0.81 Notes:

1) DEX denotes Dose Equivalent Xenon-133. The isotopic activities corresponding to this value are presented in Table 15-83.

2) DEI denotes Dose Equivalent Iodine-131. The isotopic activities corresponding to this value are presented in Table 15-84.

3) The efficiency of the condenser in scrubbing iodine from the steam flow into it is set to this value (100%) before the initiating event to increase to a maximum the iodine specific activity in the Main and Auxiliary Feedwater Systems and associated condensate grade sources.

4) The value is taken in the calculations of fission product releases between the initiating event and trip of the affected nuclear unit.

5) This value is taken for the steam generator tube rupture scenario yielding the limiting offsite TEDE (the TEDE at thr Exclusion Area Boundary following a steam generator tube rupture at Unit 2 with the power operated relief valve for the ruptured steam generator failed open).

6) Before unit trip, fission products are released from the unit vent stack. After trip, fission products are released from the steam vents on the roofs of the steam generator doghouses.

7) One control room air intake is open.

8) Two control room air intakes are open.

9) From this time onwards, both control room intakes are open.

10) The ruptured steam generator and one intact steam generator presumably vent through steam vents on the outboard steam generator doghouse. The remaining two intact steam generators presumably vent through steam vents on the inboard steam generator doghouse.

11) The limiting TEDEs at offsite locations are associated with a steam generator tube rupture scenarios at Unit 2 with the power operated relief valve for the ruptured steam generator failed open. The limiting TEDEs in the control room for a steam generator tube rupture are associated with a steam generator tube rupture at Unit 2 with a closed control room outside air intake.

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

(17 APR 2012)

Table 15-32. Input Parameters Used in the SBLOCA LOCA Analyses (Unit 2)

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 15-282 Fuel assembly array 17x17 RFA Nominal cold leg accumulator water volume (ft3/accumulator) 1050 Nominal cold leg accumulator tank volume (ft3/accumulator) 1356 Minimum cold leg accumulator gas pressure (psia) 570 Cold leg accumulator temperature (°F) 125 Pumped safety injection flow See Table 15-57 Pumped safety injection temperature (°F) 110 Nominal vessel average temperature (°F) 587.5 Pressurizer pressure (psia) 2250 RCS flow (gpm/loop) 97,500 Steam generator tube plugging (%)

10 Pressurizer low pressure safety injection setpoint (psia) 1715

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

(22 OCT 2001)

Table 15-33. Deleted Per 1994 Update Table 15-34. Deleted Per 1994 Update Table 15-35. Deleted Per 1995 Update Table 15-36. Deleted Per 1994 Update Table 15-37. Deleted Per 1995 Update

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

(15 NOV 2007)

Table 15-38. Small Break LOCA Time Sequence of Events (Unit 2) 1.5 inch (sec) 2 inch (sec) 3 inch (sec) 4 inch (sec)

Start 0

0 0

0 Reactor Trip signal 249 91 24 14 ESFAS signal 261 101 34 24 ECC delivery 293 133 66 56 Loop seal cleared N/A N/A 489 298 Core uncovery N/A 2019 900 669 Cold leg accumulator injection N/A N/A N/A 919 RWST low level 1214 1207 1198 1180 Peak cladding temperature occurs N/A 2736 1724 1032 Core recovery N/A 4095 2768

>2000

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

(15 NOV 2007)

Table 15-39. Small Break LOCA Results Fuel Cladding Data (Unit 2) 1.5 inch 2 inch 3 inch 4 inch Peak cladding temperature 1 °F N/A 1054 1164 1243 Time of PCT (sec)

N/A 2736 1724 1032 PCT Location, (ft)

N/A 11.25 11.50 11.25 Maximum local ZrO2 (%)

N/A 0.04 0.10 0.08 Maximum local ZrO2 Location, (ft)

N/A 11.25 11.25 11.25 Total core-wide average ZrO2 (%)

N/A 0.00 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 Deleted Per 2006 Update Notes:

1 There is no core uncovery for the 1.5 inch case

Catawba Nuclear Station UFSAR Table 15-40 (Page 1 of 8)

(09 OCT 2016)

Table 15-40. Radiological Consequences of the Design Basis Loss of Coolant Accident -

Data and Results Parameter Value Notes Data and assumptions pertaining to the source term (References 19 and 20)

Power level (MWth) 3479 1

Fuel pins assumed to fail (%)

100 Activity release time span Gap release phase (min) 0.5 - 30 Early in-vessel release phase (hours) 0.5 - 1.8 Activity release fraction in the gap phase (%)

3 Noble gases 5

2 Halogens 5

Alkali metals 5

Activity release fraction in the early-in vessel phase (%)

3 Noble gases 95 2

Halogens 35 Alkali metals 25 Tellurium metals 5

Alkali earth metals (barium and strontium) 2 Noble metals 0.25 Cerium group 0.05 Lanthanides 0.02 Composition of iodine in containment atmosphere (%)

Diatomic iodine 4.85 Organic iodine compounds 0.15 Particulates (iodide salts) 95 Composition of iodine released from all ESF leakage (%)

Diatmoic iodine 97 Organic iodine compounds 3

Particulates (iodide salts) 0 Core Isotopic Inventory Cf. Table 15-12

Catawba Nuclear Station UFSAR Table 15-40 (Page 2 of 8)

(09 OCT 2016)

Parameter Value Notes Data and assumptions pertaining to activity transport in containment and release Lower compartment volume (cu.ft.)

346,895 4, 5 Upper compartment volume (cu.ft.)

669,559 5

Containment leak rate (La, mass % per day) 0.3 6

Containment bypass leak rate (% La) 7 6

Containment Air Return (VX) fan start time (minutes) 10 VX fan airflow rate (cfm per fan) 40,000 Time to begin credit for Spray washout (minutes) 80 7

CSS washout time constant Time Span (sec)

CSS time constant (sec-1)

Start End I2 CsI (8) 9 0.0 4800 0.00 0.00 4800 7800 0.23 9.51 7800 30000 0.23 0.95 30000 40000 0.23 0.95 40000 50000 0.23 0.95 50000 60000 0.23 0.95 60000 70000 0.22 0.95 70000 80000 0.20 0.95 80000 86400 0.20 0.95 86400

- 2592000 0.00 00.0 10 Data pertaining to the annulus and Annulus Ventilation (VE) System Annulus volume (cu.ft.)

484,090 11 AVS filter efficiency (%)

Diatomic iodine 92 Organic iodine compounds 92 Particulate fission products 95 AVS airflow bypass fraction (%)

1 Time at which the annulus pressure falls to -

0.25 w.g.

12

Catawba Nuclear Station UFSAR Table 15-40 (Page 3 of 8)

(09 OCT 2016)

Parameter Value Notes One AVS train in operation 41.4 sec 13 Both AVS trains in operation 30.5 sec 14 AVS Exhaust and Recirculation airflow rates Table 6-75 Data and assumptions pertaining to ESF System Leakage in the Auxiliary Building Containment sump volume (cu.ft.)

79,000 ESF Leak Rate in the Auxiliary Building (gpm)

Filtered 0.5 Initially unfiltered 0.5 15 Iodine partition fractions for ESF leak in the Aux Bldg Time Span (hours)

Partition fraction Filtered leak, one ABFVES train unavailable 16 0

2 0.100 2

72 0.025 72 720 0.010 Filtered leak, all ABFVES trains in operation 17 0

2 0.100 2

72 0.031 72 720 0.010 Initially unfiltered leak, one ABFVES train unavailable 18 0

2 0.010 2

72 0.010 72 720 0.010 Initially unfiltered leak, all ABFVES trains in operation 19 0

2 0.013 2

72 0.010 72 720 0.010 Initially unfiltered leak, all ABFVES trains in operation, RHRS or CSS HX failure 20 0

2 0.100 2

72 0.027 72 720 0.010

Catawba Nuclear Station UFSAR Table 15-40 (Page 4 of 8)

(09 OCT 2016)

Parameter Value Notes ABFVES carbon bed filter efficiency for iodine absorption (%)

21 Diatomic iodine 92 Organic iodine compounds 92 ABFVES airflow bypass percent 1

Data pertaining to ESF backleakage to the Refueling Water Storage Tank (RWST)

Rate of ESF backleakage to the RWST (gpm) 10 Iodine partition fraction for ESF back leakage to the RWST Time span (seconds)

Iodine partition fraction Design basis LOCA with no heat exchanger failure 0

2160 0

2160 3000 1.022x10-6 3000 3600 1.022x10-6 3600 4200 1.071x10-6 4200 4800 1.071x10-6 4800 6000 1.001x10-6 6000 7200 9.280x10-7 7200 28800 9.139x10-7 28800 36000 1.082x10-6 36000 86400 6.056x10-7 86400 345600 2.154x10-6 345600 2592000 2.238x10-6 Design basis LOCA with RHRS or CSS Heat Exchanger failure 0

2160 0

2160 3000 1.049x10-6 3000 3600 1.049x10-6 3600 4200 1.111x10-6 4200 4800 1.111x10-6 4800 6000 1.044x10-6 6000 7200 9.723x10-7 7200 28800 1.099x10-6 28800 36000 1.586x10-6 36000 86400 1.179x10-6

Catawba Nuclear Station UFSAR Table 15-40 (Page 5 of 8)

(09 OCT 2016)

Parameter Value Notes 86400 345600 8.273x10-6 345600 2592000 1.230x10-5 Atmospheric dispersion factors (/Qs) for transport of fission products to offsite locations

/Q at the Exclusion Area Boundary (sec/m3) 4.78x10-4 22

/Q at the boundary of the Low Population Zone 0 - 8 hr 6.85x10-5 8 - 24 hr 4.00x10-5 24 - 96 hr 2.00x10-5 96 - 720 hr 7.35x10-6

/Q for transport of fission products to the control room outside air intakes (23, 24)

Release from the unit vent stack (sec/m3) 25 0 - 2 hr 1.74x10-3 26, 27 0 - 2 hr 1.04x10-3 26, 28 2 - 8 hr 1.47x10-3 26, 27 2 - 8 hr 8.82x10-4 26, 28 8 - 10 hr 6.90x10-4 27 8 - 10 hr 4.14x10-4 28 10 - 24 hr 3.68x10-4 24 - 96 hr 2.67x10-4 96 - 720 hr 1.87x10-4 Release from the RWST vent (sec/m3) 29 0 - 2 hr 1.92x10-3 26, 27 0 - 2 hr 1.26x10-3 26, 28 2 - 8 hr 1.48x10-3 26, 27 2 - 8 hr 9.78x10-4 26, 28 8 - 10 hr 7.40x10-4 27 8 - 10 hr 4.86x10-4 28 10 - 24 hr 4.10x10-4 24 - 96 hr 2.86x10-4 96 - 720 hr 1.87x10-4 Airflow imbalance in the VC outside air intakes 60/40

Catawba Nuclear Station UFSAR Table 15-40 (Page 6 of 8)

(09 OCT 2016)

Parameter Value Notes Control Room and CRAVS PFT Data Table 15-41 Coefficients for conversion from activity to radiation dose Inhaled CDE coefficients FGR-11 30 DDE coefficients FGR-12 31 Offsite breathing rates (m3/sec) 0 - 8 hr 3.5x10-4 8 - 24 hr 1.8x10-4 24 - 720 hr 2.3x10-4 Control room breathing rate (0 - 720 hr) 3.5x10-4 Control room occupancy factors 0 - 24 hr 1.0 24 - 96 hr 0.6 96 - 720 hr 0.4 Limiting Offsite Radiation Doses (Rem, 32)

Exclusion Area Boundary TEDE 8.79 TEDE at the boundary of the Low Population Zone (LPZ) 3.78 Limiting Control Room Radiation Doses (Rem, 32)

Control room TEDE 34 Fission Product buildup in the Control Room 2.56 Direction radiation from external sources 0.75 Total 3.31 Notes:

1) The assumed power level (3479 MWt) is 102% of original licensed rated thermal for the reactor (3411 MWt).

2) All noble gases are assumed to be retained in the containment atmosphere, none are assumed to be in the containment sump.

3) Release of all fission products in the radioactive source term to the containment atmosphere pursuant to RG 1.183 is simulated. Release of tellurium and iodine isotopes to the containment sump only is simulated. This is equivalent to the Staff position that only iodine may be released from ESF leakage to the environment.

4) The ice condenser is not modeled as a containment volume.

5) The volume of the lower compartment includes the dead-end compartments.

Catawba Nuclear Station UFSAR Table 15-40 (Page 7 of 8)

(09 OCT 2016)

6) The containment leak rate and the containment bypass leak rate are partitioned into two constituents: one for the lower compartment and one for the upper compartment. The partitioning is in proportion with compartment volume.

7) The control room operators start the Containment Spray System (CSS) after they begin the transfer of the Emergency Core Cooling System to cold leg to recirculation. It is assumed that they start the CSS at 80 minutes after the initiating LOCA.

8) CsI (cesium iodide) is the notation for all fission products in particulate form. The NRC Staff takes the position that CSS washout time constants for fission products in particulate form should be reduced to 1/10 of the calculated value if the decontamination factor reaches 50.

9) The spray washout time constant for organic iodine compounds (e.g., CH3I) is 0.

10) It is assumed that the CSS System is turned off at 1 day after the initiating event.

11) Credit is taken for mixing of Annulus Ventilation System (AVS) recirculation airflow with air in half the volume of the annulus (242,045 cu.ft.). No credit is taken for mixing of containment leakage to the annulus directly in the annulus airspace. Rather, it is assumed that containment leakage to the annulus flows directly to the AVS Exhaust louvers.

12) The calculation of the annulus drawdown time takes into account the effect of the differences in hydrostatic gradients inside the annulus and outside the reactor building. The outside air temperature is set to the 99th percentile low temperature (1 percentile temperature).

13) The limiting single failure with the effect of loss of one AVS train is the Minimum Safeguards failure.

14) Scenarios in which both AVS trains are assumed to be in operation include the design basis LOCA with failure of a AVS pressure transmitter - the transmitter gives a false high indication of annulus pressure. The effected AVS train continues to operate but in the Exhaust Mode at times when it should operate in the Recirculation Mode.

15) It is assumed that the operators will align the Class 1E Exhaust filter trains of the Auxiliary Building Ventilation System ABFVES within 3 days following the initiating event.

16) The limiting scenario associated with these iodine partition fractions are the design basis LOCA with Minimum Safeguards failure.

17) This scenario encompasses all design basis LOCA scenarios for which the failure does not affect the ABFVES System.

18) The limiting scenario associated with these iodine partition fractions are the design basis LOCA with Minimum Safeguards failure.

19) The limiting scenarios include the design basis LOCA with either a AVS pressure transmitter failure or a closed control room air intake isolation valve.

20) The limiting failure is the design basis LOCA with failure of flow of cold water flow to either a RHRS or CSS Heat Exchanger.

21) No particulate fission products are assumed to be entrained in ESF Systems leakage.

22) This /Q is taken for the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period of maximum release of radioactivity to the environment.

23) The control room outside air intakes are the receptors for outside airflow drawn by the pressurized filter trains of the Control Room Area Ventilation System (CRAVS). The control

Catawba Nuclear Station UFSAR Table 15-40 (Page 8 of 8)

(09 OCT 2016) room /Qs for unfiltered inleakage to the control room are based on the intake penetrations being the limiting entry point for unfiltered inleakage to the control room.

24) The values for 10-720 hr are based on transport of fission products with dispersion to two CRAVS outside air intakes.

25) The unit vent stack is taken to be the release point for (1) containment leakage that is filtered by the Annulus Ventilation System, (2) containment bypass leakage, and (3) ESF leakage in the Auxiliary Building.

26) The 0-2 hr control room /Q is taken for the 2 hr period of maximum releases. The 2-8 hr control room /Q is taken for the remainder of the 0-8 hr time span. All releases were predicted to be at a maximum over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time span within 0-8 hr.

27) One CRAVS outside air intake is taken to be open.

28) Both CRAVS outside air intakes are taken to be open.

29) The vent of the RWST is the release point for iodine released from the airspace following a design basis LOCA with assumed ESF System backleakage to the RWST.

30) Federal Guidance Report 11, cf. Reference 98.

31) Federal Guidance Report 12, cf. Reference 99.

32) All radiation doses given include the combined effects of containment leakage and ESF leakage both inside the Auxiliary Building and to the Refueling Water Storage Tank.

33) The DF for spray washout of particulates was calculated to reach 50 at the beginning of this interval.

34) The constituents to the control room TEDE from external sources are associated specifically with fission products accumulating on ventilation filters and fission products inside containment (cf. Section 1.8.1.19).

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

(24 APR 2006)

Table 15-41. Data Pertaining to the Control Room and Control Room Area Ventilation (VC)

System.

Parameter Design Basis Value Control room volume (cu.ft.)

117,920 Lower bound VC PFT total airflow to the control room (cfm) 3,500 Lower bound VC PFT recirculation airflow through the control room (cfm) 1,500 Rate of unfiltered inleakage to the control room (cfm) 100 VC PFT efficiency for removal of fission products and bypass airflow fraction (%)

Diatomic iodine 98.1 Organic iodine compounds 98.1 Particulate fission products 99 Bypass airflow fraction 0.05 Flow split in the VC outside air intakes (%/%)

60/40 Note: Unless otherwise noted, this data is used in the calculations of radiation doses for all design basis accidents with postulated releases of radioactivity releases to the environment.

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

(17 OCT 2013)

Table 15-42. Parameters for Postulated Liquid Storage Tank Rupture Evaluation (UFSAR 15.7.2)

1.

Data and assumptions pertaining to the radioactive source term

a.

Tank ruptured Recycle holdup tank

b.

Tank volume 112,000 gallons

c.

Tank specific activities Noble gases Table 15-83 Iodine Table 15-84

d.

Iodine chemical composition fractions Diatomic iodine 0.97 Organic iodine compounds 0.03

2.

Data and assumptions used to estimate activity released

a.

Iodine partition fraction 0.1

b.

Release location Unit vent stack

3.

Dispersion data

a.

/Q for the EAB (0-2 hr value) 4.78E-04 sec/m3

b.

/Q for the LPZ (0-8 hr value) 6.85E-05 sec/m3

c.

/Q for the control room (0-2 hr value) 1.74E-03 sec/m3

4.

Data and assumptions pertaining to the control room

a.

Number of intakes open 1

b.

Control room volume 117,920 cu. ft

c.

Rate of unfiltered inleakage into the control room 2,100 cfm

d.

Status of ventilation pressurized filter trains (UFSAR 15.7.2.2)

Off

5.

Data and assumptions pertaining to the control room

a.

Method of calculation Regulatory Guide 1.183

b.

Source of dose coefficients External Federal Guidance Report 12 Inhaled Federal Guidance Report 11

c.

Post accident radiation doses Table 15-14

Catawba Nuclear Station UFSAR Table 15-43 (Page 1 of 1)

(22 OCT 2001)

Table 15-43. Parameters for Postulated Refueling Water Storage Tank Rupture Accident Conservative Realistic

1.

Data and assumptions used to estimate radioactive source from refueling water storage tank rupture Accident is not evaluated for realistic case

a.

Power level (MWt) 3565.

b.

Percent of fuel defected (%)

1.

c.

Release of activity by nuclide Table 15-44

2.

Pertinent data and assumptions used to estimate activity released

1. 100% of maximum tank activities released directly to discharge canal

2. No radiological decay credit is taken from tank to nearest surface water intake

3.

Concentration data

a.

Distance to nearest surface water intake 5.6 km

b.

Dilution factor 2.0E+4

Catawba Nuclear Station UFSAR Table 15-44 (Page 1 of 1)

(27 MAR 2003)

Table 15-44. Activities for Postulated Refueling Water Storage Tank Rupture Accident HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED Isotope Tank Activity (µCi/ml)

Resultant concentration at nearest surface water intake (µCi/ml)

Fraction of 10 CFR 20 limit Sr 89 4.1E-10 8.0E-14 2.7E-8 Sr 90 2.7E-11 5.3E-15 1.8E-8 Y 90 2.1E-11 4.1E-15 2.0E-12 Y 91 7.7E-10 1.5E-13 5.0E-9 Zr 95 8.9E-11 1.7E-14 2.8E-10 Nb 95 7.9E-11 1.5E-14 1.5E-10 Tc 99M 2.1E-8 4.1E-12 1.4E-9 Mo 99 2.2E-8 4.3E-12 1.1E-7 I 131 1.1E-7 2.1E-11 7.0E-5 I 132 1.8E-9 3.5E-13 4.4E-8 I 133 5.9E-12 1.2E-15 1.2E-9 Te 132 1.7E-9 3.3E-13 1.6E-8 Cs 134 1.1E-3 2.1E-7 2.3E-2 Cs 136 1.3E-3 2.5E-7 4.2E-3 Cs 137 7.8E-4 1.5E-7 7.5E-3 Ba 137M 7.3E-4 1.4E-7 4.7E-2 Ba 140 2.9E-10 5.7E-14 2.8E-9 La 140 2.9E-10 5.7E-14 2.8E-9 Ce 144 6.5E-11 1.3E-14 1.3E-9 Pr 144 6.5E-11 1.3E-14 4.3E-9 Xe 131M 6.7E-11 1.3E-14 6.5E-11 Xe 133 6.5E-13 1.3E-16 6.5E-13 Mn 54 5.0E-11 9.8E-15 9.8E-11 Co 58 1.4E-9 2.7E-13 3.0E-9 Co 60 1.8E-10 3.5E-14 1.2E-9 Fe 59 4.8E-11 9.4E-15 1.9E-10 Cr 51 4.4E-10 8.6E-14 4.3E-11 Total FMPC 8.2E-2

Catawba Nuclear Station UFSAR Table 15-45 (Page 1 of 2)

(09 OCT 2016)

Table 15-45. Activities in Highest Inventory Discharged Assembly for Postulated Fuel Handling Accidents Nuclide Assembly Inventory (Curies)

Gap Fraction 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+04 I-131 8.09E+05 0.05 6.47E+04 I-132 1.18E+06 0.05 5.09E+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 0.10 1.49E+04 Kr-85 7.48E+03 0.05 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 7.05E+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

Catawba Nuclear Station UFSAR Table 15-45 (Page 2 of 2)

(09 OCT 2016)

Nuclide Assembly Inventory (Curies)

Gap Fraction Gap Inventory (Curies)

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

1. The gap fractions are from Table 3 of NRC Regulatory Guide RG 1.183

2. For those fuel rods which exceed the rod power/ burnup criteria of Footnote 11 in RG 1.183, the gap fractions shown above for Kr-85, Xe-133, Cs-134, and Cs-137 are tripled while the gap fractions shown above for all other noble gases and alkali metals, and for all halogens are doubled. A maximum of 25 fuel rods per fuel assembly shall be allowed to exceed the rod power/ burnup limit of Footnote 11 in RG 1.183 in accordance with the license amendment request submitted July 15, 2015.

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

(09 OCT 2016)

Table 15-46. Parameters for Postulated Fuel Handling Accident (Note 1)

1.

Data and assumptions used to estimate radioative source from postulated accident

a.

Power Level (MWt) 3479

b.

Decay time before the initiating event (days) 3

c.

Gap fractions (%)

Kr-85 Other noble gases I-131 Other iodine radioisotopes 10 5

8 5

Note 2

d.

Number of fuel assemblies impacted 1

e.

Damage to each fuel assembly impacted All rods ruptured

f.

Other pertinent assumptions R.G. 1.183

2.

Data and assumptions used to estimate release of radioactivity from the pool Note 3

a.

Effective pool decontamination factor Noble gases Iodines 1

200

b.

Composition fractions of iodine forms leaving the pool

(%)

Diatomic iodine Organic iodine compounds 57 43

c.

Time profile for release to the environment Release profile time constant (hr-1)

Decaying exponential 2

Note 4

3.

Offsite dispersion data

a.

EAB (0-2 hr) X/Q (sec/m3 4.78x10-4

b.

LPZ (0-8 hr) X/Q (sec/m3 6.85X10-5 Note 5

4.

Data for transport of radioactivity with dispersion to the control room outside air intakes

a.

Number of intakes assumed openn 1

b.

Control room X/Q (sec/m3 1.74x10-3 Note 5

5.

Data pertaining to the control room and Control Room Area Ventilation (VC) System

a.

Control room volume (cu. ft.)

117,920

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

(09 OCT 2016)

b.

VC Pressure Filter Train airflow rates to the control room Total airflow rate Recirculation airflow rate VC PFT efficiencies Diatomic iodine Organic iodine compounds 3,500 1,500 99 95

6.

Data for conversion of activity to dose

a.

Coefficients for committed effective dose equivalents FGR-11

b.

Coefficients for deep dose equivalent FGR-12

c.

Other data for conversion of activity to dose R.G. 1.183

7.

Total Effective Dose Equivalent (TEDEs - Rem)

Note 6

a.

EAB TEDE 1.76

b.

Control room TEDE 2.59 Notes:

1. The data and assumptions listed in Table 15-46 are applied to the fuel handling accident inside containment and the fuel handling accident outside containment.

2. For those 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. A maximum of 25 fuel rods per fuel assembly shall be allowed to exceed the rod power/ burnup criteria of Footnote 11 in RG [Regulatory Guide] 1.183 in accordance with the license amendment request submitted by letter dated July 15, 2015.

3. The term pool denotes the spent fuel pool for fuel handling accidents outside containment and the reactor cavity for fuel handling accidents inside containment.

4. This time constant was selected to ensure that at 98% of the iodine taken to be released from the pool would be released to the environment within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

5. Only the values for the first time period (0-8 hr for the LPZ X/Q and 0-2 hr for the control room X/Q are listed as the release of radioactivity to the environment is essentially complete in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Cf. Note 4.

6. The LPZ TEDE is not listed as it is less than the EAB TEDE and compared to the same acceptance criterion (6.3 Rem).

Catawba Nuclear Station UFSAR Table 15-47 (Page 1 of 1)

(24 OCT 2004)

Table 15-47. Deleted Per 2004 Update

Catawba Nuclear Station UFSAR Table 15-48 (Page 1 of 2)

(24 APR 2006)

Table 15-48. 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)

(Valve ND28A to Valve NV813)

Refueling Water Storage Tank (RWST) to Valves RWST minimum3 FW28 & FW56 RWST to IHSI suction RWST minimum3 RWST to Valves NV252A & NV253B RWST minimum3 Normal Containment Spray Discharge RWST minimum3 Containment Spray Suction from RWST RWST minimum3 Valves NS43A & NS38B to Aux. Cont. Spray 3501 Valves FW28 and FW56 to LHSI Suction variable2 LHSI Suction from Sump 3501 LHSI Suction from Hot Legs 3501 Containment Spray Suction from Sump 3501 RCS variable2 LHSI Discharge to Cold Legs variable2 LHSI Discharge to IHSI and HHSI Suction variable2 (Valve NV813 to Valves NI332A & NI333B)

(LHSI Discharge to Valve ND28A)

(LHSI Discharge to Valve NI136B)

LHSI Discharge to B and C Hot Legs variable2 LHSI Discharge to Valves NS43A & NS38B 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 Valves NV252A & NV253B to HHSI Suction variable2

Catawba Nuclear Station UFSAR Table 15-48 (Page 2 of 2)

(24 APR 2006)

Volume Grouping Boron Concentration (ppm)

Notes:

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.

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

(24 APR 2006)

Table 15-49. Time Sequence of Events For Steam Generator Tube Rupture Event (Dose Analysis)

Time (sec)

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 identify ruptured SG and close ruptured SG MSIV 2100 Operators close failed open steam line PORV 2271 Operators begin RCS cooldown with operable SG PORV 3000 Operators close operable steam line PORV 3820 Operators open pressurizer PORV to depressurize the RCS 4410 Break flow terminated 4607 (DNB Analysis)

Double-ended tube reputere occurs 1.0 Reactor trip/turbine trip on OTT 319.0 Reactor coolant pumps lost 319.0 MDNBR occurs 320.9

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

(22 OCT 2001)

Table 15-50. Deleted Per 2001 Update Table 15-51. Deleted Per 2001 Update Table 15-52. Deleted Per 2001 Update Table 15-53. Deleted Per 2001 Update Table 15-54. Deleted Per 2001 Update Table 15-55. Deleted Per 2000 Update Table 15-56. Deleted Per 2000 Update

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

(15 NOV 2007)

Table 15-57. Minimum ECCS Flow Assumed in SBLOCA Analyses One Train Operational, Break Backpressure Equal to RCS Pressure 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 2007 Update.

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

(09 OCT 2016)

Table 15-58. Parameters for Postulated Weir Gate Drop

1. Data and Assumption

a. Decay time before the initiating event (days) 19.5

b. Number of fuel assemblies impacted 7

c. All other data and assumptions Table 15-46

2. Total Effective Dose Equivalent Note 1

a. EAB TEDE 2.68

b. Control room TEDE 4.24 Note:

1) The LPZ TEDE is not listed as it is less than the EAB TEDE and compared to the same acceptance criterion (6.3 Rem).

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

(17 APR 2012)

Table 15-59. Input Parameters Used in the SBLOCA LOCA Analyses (Unit 1)

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 15-282 Fuel assembly array 17 x 17 RFA Nominal cold leg accumlator water volume (ft3/accumulator) 950 Nominal cold leg accumulator tank volume (ft3/accumulator) 1363 Minimum cold leg accumulator gas pressure (psig) 570 Cold leg accumulator temperature (°F) 125 Pumped safety injection flow see Table 15-57 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

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

(15 NOV 2007)

Table 15-60. Small Break LOCA Time Sequence of Events (Unit 1) 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

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

(15 NOV 2007)

Table 15-61. Small Break LOCA Results Fuel Cladding Data (Unit 1) 1.5 inch 2 inch 3 inch 4 inch Peak cladding temperature1 (°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 Deleted Per 2006 Update Notes:

1 There is no core uncovery for the 1.5 inch case

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

(22 OCT 2001)

Table 15-62. Deleted Per 2001 Update Table 15-63. Deleted Per 2001 Update Table 15-64. Deleted Per 2001 Update

Catawba Nuclear Station UFSAR Table 15-65 (Page 1 of 1)

(22 OCT 2001)

Table 15-65. Minimum Injected ECCS Flows Assumed in LBLOCA Analyses - One Train Operational RCS Pressure (psia)

High-Head SI (gpm)

Intermediate-Head SI (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

Catawba Nuclear Station UFSAR Table 15-66 and 15-67 (Page 1 of 1)

(22 OCT 2001)

Table 15-66. Deleted Per 2001 Update Table 15-67. Deleted Per 2001 Update

Catawba Nuclear Station UFSAR Table 15-68 (Page 1 of 1)

(22 OCT 2001)

Table 15-68. 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

Catawba Nuclear Station UFSAR Table 15-69 (Page 1 of 3)

(09 OCT 2016)

Table 15-69. Key Large Break LOCA Parameters and Initial Transient Assumptions.

Parameter Initial Transient Uncertainty or Bias 1.0 Plant Physical Description Dimensions Nominal PCTMOD 1

Flow resistance Nominal PCTMOD Pressurizer location Opposite broken loop Bounded Hot assembly location Under limiting location Bounded Hot assembly type 17x17 RFA with IFM Bounded SG tube plugging level D5, maximum (10%)

Bounded4 2.0 Plant Initial Operating Conditions 2.1 Reactor Power Core average linear heat rate (AFLUX)

Nominal - power (3445 MWt)7 PCTPD 2

Peak linear heat rate (PLHR)

Derived from desired Tech Spec (TS) limit and maximum baseload FQ PCTPD Hot rod average linear heat rate (HRFLUX)

Derived from TS FH PCTPD Hot assembly average heat rate (HAFLUX)

HRFLUX/1.04 PCTPD Hot assembly peak heat rate (HAPHR)

PLHR/1.04 PCTPD Axial power distribution (PBOT, PMID)

Figure 15-313 PCTPD Low power region relative power (PLOW)

Minimum (0.2)

Bounded4 Hot assembly burnup BOL Bounded Prior operating history Equilibrium decay heat Bounded Moderate Temperature Coefficient (MTC)

Tech Spec Maximum (0)

Bounded HFP boron 800 ppm Typical

Catawba Nuclear Station UFSAR Table 15-69 (Page 2 of 3)

(09 OCT 2016)

Parameter Initial Transient Uncertainty or Bias 2.2 Fluid Conditions Tavg Nominal Tavg= 587.5°F (Catawba Unit 2)

PCTIC 3

Pressurizer pressure Nominal (2250 psia)

PCTIC Loop flow Minimum (97500 gpm)

PCTMOD 5

TUH Best Estimate 0

Pressurizer level Nominal (55% of volume) 0 Accumulator temperature Nominal (115°F)

PCTIC Accumulator pressure Nominal (631.5 psig, Catawba units) PCTIC Accumulator liquid volume Nominal (7106 gal, McGuire units)

PCTIC Accumulator line resistance Nominal (McGuire Unit 2)

PCTIC Accumulator boron Minimum (McGuire units)

Bounded 3.0 Accident Boundary Conditions Break location Cold leg Bounded Break type Guillotine PCTMOD Break size Nominal (cold leg area)

PCTMOD Offsite power On (RCS pumps running)

Bounded6 Safety injection flow Minimum Bounded Safety injection temperature Nominal (85°F)

PCTIC Safety injection delay Max delay (17 sec)

Bounded Containment pressure Minimum based on WC/T M&E Bounded Single failure ECCS: Loss of 1 SI train Bounded Control rod drop time No control rods Bounded

Catawba Nuclear Station UFSAR Table 15-69 (Page 3 of 3)

(09 OCT 2016)

Parameter Initial Transient Uncertainty or Bias 4.0 Model Parameters Critical Flow Nominal (as coded)

PCTMOD Resistance uncertainties in broken loop Nominal (as coded)

PCTMOD Initial stored energy/fuel rod behavior Nominal (as coded)

PCTMOD Core heat transfer Nominal (as coded)

PCTMOD Delivery and bypassing of ECC Nominal (as coded)

Conservative Steam binding/entrainment Nominal (as coded)

Conservative Noncondensable gases/accumulator nitrogen Nominal (as coded)

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

7. Analysis was originally performed at 3445 MWt (3411 MWt 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 15-82.

Catawba Nuclear Station UFSAR Table 15-70 (Page 1 of 1)

(22 OCT 2001)

Table 15-70. 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

Catawba Nuclear Station UFSAR Table 15-71 (Page 1 of 3)

(09 OCT 2016)

Table 15-71. Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis Parameter Operating Range 1.0 Plant Physical Description Dimensions No in-board assembly grid deformation during LOCA +

SSE Flow resistance N/A Pressurizer location N/A Hot assembly location Anywhere in core Hot assembly type Fresh 17x17 RFA SG tube plugging level 10% (Catawba 2) and 5% (McGuire and Catawba 1) 2.0 Plant Initial Operating Conditions 2.1 Reactor Power Core avg linear heat rate Core power 101% of 3445 MWt 3

Peak linear heat rate FQ 2.70 ( 4 ft), FQ 2.50 (> 4 ft) [see Note 1]

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

Hot assembly average linear heat rate HA P

1.67/1.04 [see Note 4]

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

Axial power dist (PBOT, PMID)

Figure 15-312 Low power region relative power (PLOW) 0.2 PLOW 0.8 Hot assembly burnup 75000 MWD/MTU, lead rod Prior operating history All normal operating histories MTC 0 at HFP HFP boron Normal letdown Rod power census See Table 15-72

Catawba Nuclear Station UFSAR Table 15-71 (Page 2 of 3)

(09 OCT 2016)

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

Accumulator fL/D Current line configuration Minimum accumulator boron 2275 ppm 3.0 Accident Boundary Conditions Break location N/A Break type N/A Break size N/A Offsite power Available or LOOP Safety injection flow Table 15-65 Safety injection temperature 58° SI Temp 90°F, Reference 82 (covers a RWST temperature range of 70-100°F and component cooling water temperatures down to 45°F)

Safety injection delay 17 seconds (with offsite power) 32 seconds (with LOOP)

Containment pressure Bounded - - see Figure 15-302

Catawba Nuclear Station UFSAR Table 15-71 (Page 3 of 3)

(09 OCT 2016)

Parameter Operating Range Single failure Loss of one train 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. Analysis was originally performed at 3445 MWt (3411 MWt 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 15-82.

4. 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 within 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

Catawba Nuclear Station UFSAR Table 15-72 (Page 1 of 1)

(22 OCT 2001)

Table 15-72. 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

Catawba Nuclear Station UFSAR Table 15-73 (Page 1 of 1)

(24 APR 2006)

Table 15-73. Fuel Assembly Isotopic Radioactivity Levels (Alternative Source Term Analysis of the Catawba Design Basis Locked Rotor and Rod Ejection Accidents)

Noble Gases Halogens Alkali Metals Radioisotop e

Core Activity (Curies)

Radioisotop e

Core Activity (Curies)

Radioisotope Core Activity (Curies)

Kr83m 1.27E+05 Br83 1.27E+05 Rb86 1.68E+03 Kr85m 2.85E+05 Br85 2.85E+05 Rb88 8.48E+05 Kr85 7.31E+03 Br87 4.72E+05 Rb89 1.13E+06 Kr87 5.86E+05 I130 2.52E+04 Rb90 1.07E+06 Kr88 8.29E+05 I131 7.52E+05 Cs134 1.91E+05 Kr89 1.07E+06 I132 1.11E+06 Cs136 4.16E+04 Xe131m 9.63E+03 I133 1.60E+06 Cs137 9.15E+04 Xe133m 4.88E+04 I134 1.86E+06 Cs138 1.59E+06 Xe133 1.57E+06 I135 1.52E+06 Cs139 1.51E+06 Xe135m 3.20E+05 Xe135 4.14E+05 Xe137 1.48E+06 Xe138 1.52E+06

Catawba Nuclear Station UFSAR Table 15-74 (Page 1 of 1)

(24 OCT 2004)

Table 15-74. Parameters for the Steam Generator Tube Rupture Supplemental Offsite Dose Analysis

1.

Data pertaining to the radioactive source term

a.

Equilibrium reactor coolant specific activity (µCi/gm DEI) 0.46

b.

Transient reactor coolant specific activity(µCi/gm DEI) 26

2.

Data and assumptions pertaining to transport and release of radioactivity

a.

Power level (MWt) 3479.

b.

Condenser iodine scrubbing efficiency before unit trip (%)

85

c.

Time of unit trip (minutes after initiating event) 20

d.

Offsite power Lost at trip

e.

Time span of primary bypass (minutes after initiating event) 20-25

f.

Primary bypass fraction 0.12

g. Maximum flash fraction 0.14

h. Integrated break flow (lbm) 275,000

i.

Ruptured steam generator steam release after unit trip (lbm) 589,000

j.

Iodine partition fraction for steam releases 0.01

k. Intact steam generator tube leak rate (gpd per SG) 150

l.

Intact steam generator steam release after unit trip (lbm - all three SGs) 3,230,000

m.

Steam release rate before trip (lbm/min/SG) 64,110

3.

Dispersion data

a.

/Q at exclusion area boundary (sec/m3) 4.78x10-4

b.

/Q at low population zone (sec/m3) 6.85x10-5

4.

Dose conversion data

a.

Method of conversion from activity to dose R.G. 1.4

b.

Source of dose conversion factors Whole body radiation dose R.G. 1.109 Thyroid radiation dose ICRP 30

5.

Thyroid radiation doses at the exclusion area boundary (Rem, Note 1)

a.

Pre-existent iodine spike 57.0

b.

Concurrent iodine spike 22.0 Note:

1) Thyroid radiation doses at the exclusion area only are reported. They are limiting with respect to relative margin to the germane regulatory acceptance limits.

Catawba Nuclear Station UFSAR Table 15-75 (Page 1 of 2)

(18 APR 2009)

Table 15-75. Fission Product Radioactivity Levels in a Mixed Oxide Lead Test Assembly Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Noble Gases Tellurium Group Radioisotope LTA Activity (Curies)

Radioisotope LTA Activity (Curies)

Kr83m 6.25E+04 Se81 3.36E+04 Kr85m 1.17E+05 Se83 3.32E+04 Kr85 4.36E+03 Se83m 2.79E+04 Kr87 2.18E+05 Se84 1.01E+05 Kr88 2.93E+05 Se87 4.97E+04 Kr89 3.33E+05 Sb127 9.00E+04 Xe131m 8.93E+03 Sb128 1.63E+04 Xe133m 4.71E+04 Sb128m 1.32E+05 Xe133 1.35E+06 Sb129 2.82E+05 Xe135m 3.38E+05 Sb130 1.06E+05 Xe135 7.73E+05 Sb130m 2.97E+05 Xe137 1.23E+06 Sb131 5.11E+05 Xe138 1.06E+06 Sb132m 2.74E+05 Halogens Te127m 1.30E+04 Radioisotope LTA Activity (Curies)

Te127 Te129 8.21E+06 2.51E+05 Br83 6.25E+04 Te129m 4.78E+04 Br85 1.17E+05 Te131 6.00E+05 BR87 1.61E+05 Te132 1.04E+06 I130 2.82E+04 Te133 6.93E+05 I131 7.21E+05 Te133m 5.82E+05 I132 1.08E+06 Te134 1.04E+06 I133 1.39E+06 I134 1.48E+06 I135 1.31E+06 Alkali Earth Metals Radioisotope LTA Activity (Curies)

Alkali Metals Sr89 3.37E+05 Radioisotope LTA Activity (Curies)

Sr90 Sr91 3.25E+04 5.51E+05 Rb86 1.15E+03 Sr92 6.45E+05 Rb88 3.02E+05 Sr93 7.90E+05 Rb89 3.82E+05 Ba139 1.14E+06 Rb90 3.21E+05 Ba140 1.09E+06 Cs134 1.93E+05 Ba141 1.05E+06 Cs136 6.89E+04 Ba142 9.44E+05 Cs137 9.35E+04 Cs138 1.19E+06 Cs139 1.09E+06 Noble Metals Lanthanides Radioisotope LTA Activity (Curies)

Radioisotope LTA Activity (Curies)

Mo99 1.24E+06 Y90 3.42E+04 Mo101 1.19E+06 Y91 4.86E+05 Mo102 1.20E+06 Y91m 3.20E+05 Tc99m 1.11E+06 Y92 6.49E+05 Tc101 1.19E+06 Y93 5.44E+05

Catawba Nuclear Station UFSAR Table 15-75 (Page 2 of 2)

(18 APR 2009)

Noble Metals Lanthanides Radioisotope LTA Activity (Curies)

Radioisotope LTA Activity (Curies)

Tc104 1.16E+06 Y94 9.11E+05 Ru103 1.24E+06 Y95 9.78E+05 Ru105 1.04E+06 Zr95 8.90E+05 Ru106 6.94E+05 Zr97 1.05E+06 Ru107 6.71E+05 Nb95 8.81E+05 Rh103m 1.24E+06 Nb95m 9.88E+03 Rh105 9.62E+05 Nb97 1.06E+06 Rh106m 2.53E+04 La140 1.11E+06 Rh107 6.73E+05 La141 1.07E+06 Pd109 4.53E+05 La142 1.01E+06 Pd111 6.34E+04 La143 9.00E+05 Pd112 2.75E+04 Nd147 4.07E+05 Nd149 2.64E+05 Cerium Group Nd151 1.56E+05 LTA Activity Pm147 9.21E+04 Radioisotope (Curies)

Pm148 1.15E+05 Ce141 9.79E+05 Pm148m 1.85E+04 Ce143 9.06E+05 Pm149 3.90E+05 Ce144 6.43E+05 Pm151 1.56E+05 Ce145 6.22E+05 Sm153 4.66E+00 Ce146 5.13E+05 Sm156 2.36E+04 Np237 6.95E-02 Eu154 1.26E+04 Np238 1.03E+05 Eu155 3.99E+03 Np239 1.36E+07 Eu156 3.45E+05 Np240 4.60E+04 Eu157 3.48E+04 Pu236 2.40E-01 Pr142 5.81E+04 Pu238 2.08E+03 Pr143 8.55E+05 Pu239 1.29E+03 Pr144 6.49E+05 Pu240 9.77E+02 Pr144m 9.03E+03 Pu241 2.69E+05 Pr145 6.23E+05 Pu242 4.96E+00 Pr146 5.20E+05 Pu243 9.63E+05 Pr147 4.24E+05 Am241 4.40E+02 Am242m 3.56E+01 Am242 2.10E+05 Am243 8.28E+01 Am244 2.03E+05 Cm242 8.48E+04 Cm244 5.41E+03

Catawba Nuclear Station UFSAR Table 15-76 (Page 1 of 1)

(18 APR 2009)

Table 15-76. Fission Product Radioactivity Levels in a Mixed Oxide Lead Test Assembly (Fuel Handling Accident and Weir Gate Drop Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Noble Gases Halogens Radioisotope LTA Activity (Curies)

Radioisotope LTA Activity (Curies)

Kr83m 7.25E+04 Br83 7.22E+04 Kr85m 1.34E+05 Br85 1.33E+05 Kr85 1.39E+03 Br87 1.88E+05 Kr87 2.52E+05 I130 7.51E+03 Kr88 3.38E+05 I131 8.81E+05 Kr89 3.84E+05 I132 1.28E+06 Xe131m 1.07E+04 I133 1.64E+06 Xe133m 5.51E+04 I134 1.76E+06 Xe133 1.65E+06 I135 1.57E+06 Xe135m 3.92E+05 Xe135 6.80E+05 Xe137 1.48E+06 Xe138 1.26E+06 Alkali Metals Radioisotope LTA Activity (Curies)

Rb86 3.13E+02 Rb88 3.48E+05 Rb89 4.39E+05 Rb90 3.72E+05 Cs134 1.94E+04 Cs136 3.41E+04 Cs137 2.67E+04 Cs138 1.43E+06 Cs139 1.31E+06

Catawba Nuclear Station UFSAR Table 15-77 (Page 1 of 1)

(18 APR 2009)

Table 15-77. LOCA Release Fractions Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Release Fraction (%)

LEU Fuel MOX LTA Group (Elements)

Gap Release Phase Noble gases (Kr, Xe) 5 7.5 Halogens (Br, I) 5 7.5 Alkali metals (Rb, Cs) 5 7.5 Early In-vessel Release Phase Noble gases (Kr, Xe) 95 92.5 Halogens (Br, I) 35 52.5 Alkali metals (Rb, Cs 25 37.5 Tellurium group (Se, Sb, Te) 5 7.5 Alkali earth metals (Ba, Sr) 2 3

Noble metals (Mo, Tc, Ru, Rh, Pd) 0.25 0.375 Cerium group (Ce, Np, Pu) 0.05 0.075 Lanthanides (Y, Zr, Nb, La, Nd, Pm, Sm, Eu, Pr, Am, Cm) 0.02 0.03

Catawba Nuclear Station UFSAR Table 15-78 (Page 1 of 1)

(18 APR 2009)

Table 15-78. Non LOCA Gap Fractions Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Release Fraction (%)

LEU Fuel MOX LTA Group (Elements)

Locked Rotor Accident, Fuel Handling Accident, and Weir Gate Drop Kr-85 10 15 I-131 8

12 Other Noble gases (Kr, Xe) 5 7.5 Other Halogens (Br, I) 5 7.5 Alkali metals (Rb, Cs) 12 18 Rod Ejection Accident Noble gases (Kr, Xe) 10 15 Halogens (Br, I) 10 15 Alkali metals (Rb, Cs) 12 18

Catawba Nuclear Station UFSAR Table 15-79 (Page 1 of 1)

(18 APR 2009)

Table 15-79. Iodine Partition Fractions for Post LOCA ESF Leakage in the Auxiliary Building Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Iodine Partition Fraction (%)

Time span (hours)

All LEU Fuel With MOX LTAs Filtered leak, one VA train unavailable 0.0 - 2.5 2.5 - 72.0 72.0 - 720.0 0.100 0.022 0.010 0.100 0.022 0.010 Filtered leak, all VA trains available 0.0 - 2.5 2.5 - 72.0 72.0 - 720.0 0.100 0.028 0.010 0.100 0.028 0.010 Initially unfiltered leak, one VA train unavailable 0.0 - 2.9 2.9 - 72.0 72.0 - 720.0 0.010 0.010 0.010 0.010 0.010 0.010 Initially unfiltered leak, all VA trains available 0.0 - 2.9 2.9 - 72.0 72.0 - 720.0 0.013 0.010 0.010 0.014 0.010 0.010 Initially unfiltered leak, ND or NX NX failure, all VA trains available 0.0 - 2.9 2.9 - 72.0 72.0 - 720.0 0.100 0.024 0.010 0.100 0.024 0.010

Catawba Nuclear Station UFSAR Table 15-80 (Page 1 of 1)

(18 APR 2009)

Table 15-80. Iodine Partition Fractions for ESF Backleakage to the Refueling Water Storage Tank Following a Design Basis Rod Ejection Accident Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

Iodine Partition Fraction Time span (hours)

All LEU Fuel With MOX LTAs 0 - 2 0.000E+00 0.000E+00 2 - 8 3.135E-06 3.200E-06 8 - 10 1.266E-05 1.292E-05 10 - 24 2.332-05 2.379E-05 24 - 96 8.910E-04 9.069E-04 96 - 720 2.415E-02 2.433-02

Catawba Nuclear Station UFSAR Table 15-81 (Page 1 of 2)

(18 APR 2009)

Table 15-81. Effect of Operation of Unit 1 with four MOX LTAs on Post Accident Radiation Doses Currently, mixed oxide (MOX) fuel has been retired from use at Catawba Nuclear Station and any further reactor operation of MOX fuel may require a reanalysis of these values.

TEDE (Rem)

Design Basis Accident Scenario EAB LPZ CR Design basis locked rotor accident With all LEU fuel 1.52 0.30 1.19 Unit 1 in operation with four MOX LTAs (Note 1) 1.02 0.26 0.80 Regulatory limit 2.5 2.5 5

Design basis rod ejection accident With all LEU fuel 4.75 2.82 2.70 Unit 1 in operation with four MOX LTAs (Note 1) 3.96 2.75 2.32 Regulatory limit 6.3 6.3 5

6.05 3.08 2.22 Design basis LOCA With all LEU fuel Deleted Per 2007 Update Unit 1 in operation with four MOX LTAs (Note 2) 6.12 3.12 2.23 Regulatory limit 25 25 5

Design basis fuel handling accident With all LEU fuel 1.6 Note 3 2.3 Unit 1 in operation with four MOX LTAs (Notes 2, 4) 2.3 Note 3 2.1 Regulatory limit 6.3 6.3 5

Design basis weir gate drop With all LEU fuel 2.9 Note 3 3.5 Unit 1 in operation with four MOX LTAs (Notes 2, 4) 3.5 Note 3 3.3 Regulatory limit 6.3 6.3 5

Notes on Table 15-81

1.

A separate set of analyses of radiological consequences of the design basis locked rotor and rod ejection accident was completed for each of the two nuclear units at Catawba.

Catawba Nuclear Station UFSAR Table 15-81 (Page 2 of 2)

(18 APR 2009)

Only Unit 1 at Catawba is in operation with the four MOX LTAs. Therefore, radiation doses following the design basis locked rotor and rod ejection accidents involving the MOX LTAs are reported for Unit 1 only. For the locked rotor and rod ejection accidents involving an all LEU core, the limiting radiation doses are associated with Unit 2.

2.

One set of analyses of radiological consequences of the design basis LOCA, fuel handling accident, and weir gate drop that are bounding for both nuclear units at Catawba was completed.

3.

TEDEs at the LPZ are not reported for the design basis fuel handling accident and weir gate drop. In every case, they are bounded by the TEDEs at the EAB.

4.

For the fuel handling accidents and weir gate drop involving MOX LTAs, it was assumed that both control room outside air intakes were open. For the fuel handling accidents and weir gate drop involving all LEU fuel, it was assumed that only one control room intake was open.

Catawba Nuclear Station UFSAR Table 15-82 (Page 1 of 2)

(09 OCT 2016)

Table 15-82. 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 15-70]

2028 81 PCT Assessments Decay heat in Monte Carlo calculations 8

104 MONTECF power uncertainty correction 20 105 Safety Injection temperature range 59 82 Input error resulting in an incomplete solution matrix 25 106 Revised blowdown heatup uncertainty distribution 5

107 Vessel unheated conductor noding 0

108 Revised algorithm for average fuel temperature 0

108 Peak transient FQ = 2.7 in bottom third of core 0

109 Change from PAD 3.4 to PAD 4.0

-75 109 Fuel Thermal Conductivity Degradation with Peaking Factor Burndown 15 109 Revised Heat Transfer Multiplier Distribution

-85 111 HOTSPOT Clad Burst Strain Error 70 112 Unit 1 MUR Uprate to 101.7% of 3411MWt 16 113 Unit 1 Current Licensing Basis LBLOCA PCT Including Assessments 2086 113 Unit 2 Current Licensing Basis LBLOCA PCT Including Assessments 2070 112 Small Break LOCA; NOTRUMP Unit 1 Analysis of Record PCT (2-inch break) [See Table 15-61]

1323 110 PCT Assessments None 0

110 Unit 1 Current Licensing Basis SBLOCA PCT Including Assessments 1323 109 Unit 2 Analysis of Record PCT (4-inch break) [See Table 15-39]

1243 110

Catawba Nuclear Station UFSAR Table 15-82 (Page 2 of 2)

(09 OCT 2016)

PCT Assessments None 0

110 Unit 2 Current Licensing Basis SBLOCA PCT Including Assessments 1243 109

Catawba Nuclear Station UFSAR Table 15-83(Page 1 of 1)

(17 OCT 2013)

Table 15-83. Dose Equivalent Xenon-133 Noble Gas Specific Activities in the Reactor Coolant Radioisotope Specific Activity

(µCi/gm)

FGR No. 12, Table III.1 DCFs (Sv-s/(Bq-m3))

DEX Specific Activity

(µCi/gm)

KR-85M 2.06E+00 7.48E-15 9.88E+00 KR-85 7.52E+00 1.19E-16 5.74E-01 KR-87 1.34E+00 4.12E-14 3.54E+01 KR-88 3.71E+00 1.02E-13 2.43E+02 XE-131M 2.27E+00 3.89E-16 5.65E-01 XE-133M 1.75E+01 1.37E-15 1.54E+01 XE-133 2.78E+02 1.56E-15 2.78E+02 XE-135M 4.95E-01 2.04E-14 6.47E+00 XE-135 7.42E+00 1.19E-14 5.66E+01 XE-138 6.59E-01 5.77E-14 2.44E+01 Total DEX 6.70E+02

Catawba Nuclear Station UFSAR Table 15-84(Page 1 of 1)

(17 OCT 2013)

Table 15-84. Dose Equivalent Iodine-131 Noble Gas Specific Activities in the Reactor Coolant Radioisotope Specific Activity

(µCi/gm)

FGR No. 11, Table 2.1 DCFs (Sv/(Bq)

DEI Specific Activity

(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 Total DEI 1.00E+00

ML19189A273

Text

2.0 Notes

Navigation menu

Search