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{{Adams
#REDIRECT [[L-2016-198, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 14, Safety Analysis, Appendix 14A Thru Appendix 14G]]
| number = ML16330A239
| issue date = 10/29/2016
| title = Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 14, Safety Analysis, Appendix 14A Thru Appendix 14G
| author name =
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number =
| contact person =
| case reference number = L-2016-198
| package number = ML16330A191
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 69
}}
 
=Text=
{{#Wiki_filter:APPENDIX 14A
 
TURKEY POINT PLANT UNIT 3  CYCLE 28 RELOAD CHARACTERISTICS AND PARAMETERS
 
14A-i  Revised 03/11/2016 C28 TABLE OF CONTENTS Section Title Page
 
==1.0  INTRODUCTION==
AND SUMMARY  ........................................................ 14A-1
 
1.1 Introduction  ............................................................................... 14A-1  1.2 General Description  .................................................................. 14A-1
 
Appendix A  Turkey Point Unit 3 Cycle 28..........
................
..................
..................
... 14A-A1 Core Operating Limits Report (COLR)
 
14A-ii Revised 03/11/2016 C28 LIST OF TABLES Table Title Page 14A-1 Fuel Assembly Design Parame ters  ......................-----------.14A-2 Turkey Point Unit 3 - Cycle 28 14A-2 Kinetics Characteristics..............
.................................................................... 14A-3 Turkey Point Unit 3 - Cycle 28 14A-3 Shutdown Requirements and Margins  .........................................................14A-4 Turkey Point Unit 3 - Cycles 27 and 28
 
LIST OF FIGURES
 
Figure  14A-1 Reference Core Loading Pattern  .................................................................. 14A-6 Turkey Point Unit 3 Cycle 28 14A-2 Burnable Absorber Locations
......................................................................... 14A-7 Turkey Point Unit 3 Cycle 28
 
14A-iii Revised 03/11/2016 C28C28
 
==1.0 INTRODUCTION==
and SUMMARY
 
===1.1 Introduction===
This report presents reload characteristics and parameters associated with Turkey Point Unit 3 Cycle 28. The Cycle 28 core is a full core with 15x15 Upgrade fuel assemblies in Regions 28, 29 and 30. 1.2 General Description The Turkey Point Unit 3 reactor core is comprised of 157 fuel assemblies arranged in the core loading pattern configuration shown in Figure 14A-1.
All fuel assemblies have axial blankets at both the top and bottom of the fuel stack to reduce neutron leakage and to improve uranium utilization. Regions 28, 29 and 30 fuel assemblies are 8 inch long, with 2.6 w/o enriched UO 2 annular pellets. The design parameters for the Cycle 28 core are provided in Table 14A-1.
The Cycle 28 core uses a single type of burnable absorber, the Westinghouse Integral Fuel Burnable Absorber (IFBA) rods composed of ZrB 2 coated fuel pellets with 2.2125 mg 10B/in. The active absorber length is 120 inches. Their locations in the core are shown in Figure 14A-2. 
 
The core design parameters for Cycle 28 are as follows:
Parameter Current Licensing Basis
 
Core Power (MW t)    2644  Pressurizer Pressure (psia)    2250  Core Inlet Temperature 1 (F)    549.2  Core Inlet Temperature 2 (F)    550.2  Thermal Design Flow (gpm)    260,700  Minimum Measured Flow (gpm)    270,000 Average Linear Power Density (kW/ft)  6.714
: 1. Based on Thermal Design Flow.
: 2. Based on Minimum Measured Flow.
 
The core kinetics characteristics and shutdown requirements and margins are provided in Tables 14A-2 and 14A-3, respectively. The Core Operating Limits Report (COLR) for Cycle 28 is provided in Appendix A.
 
14A-1 Revised 03/11/2016 C28C28C28C28C28 Table 14A-1 Fuel Assembly Design Parameters Turkey Point Unit 3 - Cycle 28
 
14A-2  Revised 03/11/2016 Region  28A 28B 28C 29A 29B 29C 29D 29E 30A 30B 30C Enrichment 1 (w/o U235)  3.797 3.797 4.210 3.987 4.196 4.393 4.393 4.939 3.900 4.100 4.500 Density1 (% Theoretical) 95.65 95.65 95.46 95.49 95.63 95.81 95.81 95.81 95.50 95.50 95.50 Number of Assemblies 8 4 16 5 24 20 4 16 20 20 20 Approximate Burnup at  Beginning of 30,986 33,139 37,533 25,963 26,183 22,180 24,308 25,690 0 0 0 Cycle 28 (MWD/MTU) 2  Fuel Type Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade  Upgrade Upgrade Upgrade  Upgrade Upgrade Number of IFBA/Assembly 100 116 32 148 148 48 100 148 8@64 12@148 148 16@48 4@80 Total Number of IFBA 800 464 512 740 3552 960 400 2368 2288 2960 1088 Fuel Rods/Region Axial Blankets (AB) 3  YES YES YES YES YES YES YES YES YES YES YES AB Annular Pellets  YES YES YES YES YES YES YES YES YES YES YES AB Enrichment (w/o) 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 ZIRLOTM Cladding NO NO NO NO NO NO NO NO YES YES YES 1  As-built values for burned regions and design values for fresh region      2  Based on an assumed Cycle 27 burnup of 20,226 MWD/MTU (LW)    3  Axial blankets in all regions are 8 inch long. C28 Table 14A-2 KINETICS CHARACTERISTICS TURKEY POINT UNIT 3 - Cycle 28
 
Moderator Temperature Current Limit Cycle 28 Coefficient (pcm/°F)
: a. Most positive +5.0 ( 70% RTP)  +1.1 (HZP, 541°F,    (linear ramp to 0 at  2000MWD/MTU), linear ramp      100% RTP)  to 0 at 100% RTP
: b. Most negative  32.2  Doppler Coefficient (pcm/°F) -2.9 to -1.0 -1.97 to -1.22 Most Negative to Least Negative Delayed Neutron Fraction, eff 0.0044 to 0.0075 0.0048 to 0.0064  Minimum to Maximum Maximum Differential Rod
<100 58.3 Worth of Two Banks Moving Together at HZP (pcm/in)
Shutdown Margin (pcm)
: a. BOC 1000* 3512 1770**  b. EOC 1770 2073
* MODES 1 through 4 with at least 1 RCP running
** MODE 4 without RCPs running and MODE 5
 
14A-3 Revised 03/11/2016 C28C28 Table 14A-3 Shutdown Requirements and Margins Turkey Point Unit 3 - Cycles 27 and 28 Cycle 27            Cycle 28 BOC  EOC  BOC  EOC Control Rod Worth (%)  All Rods Inserted Less  5.69  5.72  5.98  5.95 Worst Stuck Rod (1) Less 7%    5.29  5.32  5.56  5.53 Control Rod Requirements (%)  Reactivity Defects (Doppler,  1.98  3.52  2.05  3.46 TAVE, Void, and Redistribution)
Rod Insertion Allowance (RIA)  ---    ---    ---    --- RCCA Repositioning Allowance (see note)
(2) Total Requirements  1.98  3.52  2.05  3.46 Shutdown Margin (1) - (2) (%) 3.31  1.80  3.51  2.07 Required Shutdown Margin (%) 1.00  1.77  1.00  1.77
 
Note: Additional margin to accommodate a 22 &deg;F cooldown is not available for the EPU cycles. The RIA term is already included in the Reactivity Defects term in the methodology used to compute the shutdown margin.
 
14A-4 Revised 03/11/2016 C28C28C28 Table 14A-4 DELETED   
 
14A-5 Revised 03/11/2016
 
Figure 14A-1 Reference Core Loading Pattern Turkey Point Unit 3 Cycle 28
 
14A-6 Revised 03/11/2016 C28 Figure 14A-2 Turkey Point Unit 3, Cycle 28 Burnable Absorber and Source Rod Locations
 
TYPE TOTAL
## I (Total number of fresh IFBA Rods) ----------6336
 
14A-7 Revised 03/11/2016 C28
 
Appendix A Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR)
 
14A-A1 Revised 03/11/2016 C28 Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR)
 
==1.0 INTRODUCTION==
 
This Core Operating Limits Report for Turkey Point Unit 3 Cycle 28 has been prepared in accordance with the requirements of Technical Specification 6.9.1.7. 
 
The Technical Specifications (TS) affected by this report are listed below with the section and page for each one of the TS addressed in this COLR document. Section  Technical Specification Page 2.1  2.1.1 Reactor Core Safety Limits  14A-A3 2.2  2.2.1 Reactor Trip System Instrumentation Setpoints 14A-A3-14A-A4 2.3  3.1.1.1 Shutdown Margin Limit for MODES 1, 2, 3, 4  14A-A4 2.4    3.1.1.2 Shutdown Margin Limit for MODE 5 14A-A4 2.5  3.1.1.3 Moderator Temperature Coefficient 14A-A5 2.6  4.1.1.3 MTC Surveillance at 300 ppm 14A-A5 2.7  3.1.3.2 Analog Rod Position Indication System 14A-A5 2.8  3.1.3.6 Control Rod Insertion Limits  14A-A5 2.9  3.2.1  Axial Flux Difference  14A-A5 2.10  3.2.2  Heat Flux Hot Channel Factor F Q(Z) 14A-A5 2.11  3.2.3  Nuclear Enthalpy Rise Hot Channel Factor  14A-A6 2.12  3.2.5  DNB Parameters 14A-A6 Figure  Description A1  Reactor Core Safety Limit - Three Loops in Operation 14A-A7 A2  Required Shutdown Margin vs Reactor Coolant Boron Concentration 14A-A8 A3  Turkey Point Unit 3 Cycle 28 Rod Insertion Limits vs Thermal Power 14A-A9 A4  Axial Flux Difference as a Function of Rated Thermal Power 14A-A10 
 
14A-A2 Revised 03/11/2016 C28C28C28 Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR) 2.0 Operating Limits The cycle-specific parameter limits for the specifications listed in the Introduction are presented below and listed sequentially by Technical Specification (TS). These limits have been developed using the NRC-approved methodologies specified in TS 6.9.1.7.
 
2.1 Reactor Core Safety Limits - Three Loops in Operation (TS  2.1.1)
 
  - Figure A1 (page 14A-A7)
In Modes 1 and 2, the combination of Thermal Power, reactor coolant system highest loop average temperature and pressurizer pressure shall not exceed the limits in Figure A1.
2.2 Reactor Trip System Instrumentation Setpoints (TS  2.2.1)
NOTE 1 on TS Table 2.2-1 Overtemperature T  - 1 = 0s, 2 = 0s Lead/Lag compensator on measured T  - 3 = 2s  Lag compensator on measured T  - K1 = 1.31  - K2 = 0.023/F  - 4 = 25s, 5 = 3s Time constants utilized in the lead-lag compensator for Tavg  - 6 = 2s  Lag compensator on measured Tavg  - T  583.0 F Indicated Loop Tavg at RATED THERMAL POWER
  - K3 = 0.00116/psi  - P'  2235 psig Nominal RCS operating pressure
  - f1(I) = 0 for q t - qb between - 18% and + 7%
.      For each percent that the magnitude of q t - qb exceeds - 18%,
the T Trip Setpoint shall be automatically reduced by 3.51% of its value at RATED THERMAL POWER; and For each percent that the magnitude of q t - qb exceeds + 7%, the T Trip Setpoint shall be automatically reduced by 2.37% of its value at RATED THERMAL POWER.
Where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q t + qb is total THERMAL POWER in percent of RATED THERMAL POWER.
 
14A-A3 Revised 03/11/2016 C28 Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR)
NOTE 2 on TS Table 2.2-1 Overtemperature T  The Overtemperature T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel, 0.2% T span for the Pressurizer Pressure channel, and 0.4% T span for the f(l) channel. No separate Allowable Value is provided for Tavg  because this function is part of the T value. NOTE 3 on TS Table 2.2-1 Overpower T  - K4  = 1.10  - K5  0.0/F  For increasing average temperature
  - K5 = 0.0/F For decreasing average temperature
  - 7  0 s  Time constants utilized in the lead-lag compensator for Tavg  - K6 = 0.0016/F For T > T"
  - K6 = 0.0  For T  T"  - T"  583.0F Indicated Loop Tavg at RATED THERMAL POWER  - f2 (I) = 0  For all I  NOTE 4 on TS Table 2.2-1 Overpower T  The Overtemperature T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel. No separate Allowable Value is provided for Tavg  because this function is part of the T value.
2.3 Shutdown Margin Limit for MODES 1, 2, 3 and 4 (TS  3.1.1.1)
  - Figure A2 (page 14A-A8) 2.4 Shutdown Margin Limit for MODE 5 (TS  3.1.1.2)
  -  1.77% k/k   
 
14A-A4 Revised 03/11/2016 C28 Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR) 2.5 Moderator temperature coefficient (MTC)  (TS  3.1.1.3)
  - + 5.0 x 10
-5 k/k/F  BOL, HZP, ARO and from HZP to 70% Rated Thermal Power (RTP)  - From 70% RTP to 100% RTP the MTC  decreasing linearly from < + 5.0 x 10
-5 k/k/F  to < 0.0 x 10
-5 k/k/F    - Less negative than - 41.0 x 10
-5 k/k/F  EOL, RTP, ARO
 
2.6 Moderator temperature coefficient (MTC) Surveillance at 300 ppm (TS  4.1.1.3)
  - Less negative than - 35.0 x 10
-5 k/k/F (-35 pcm/F) Within 7 EFPD of reaching equilibrium boron concentration of 300 ppm. The Revised Predicted near - EOL 300 ppm MTC shall be calculated using the algorithm contained in WCAP-13749-P-A: Revised predicted MTC = Predicted MTC + AFD Correction - 3 pcm/F If the Revised Predicted MTC is less negative than the SR 4.1.1.3.b 300 ppm surveillance limit and all the benchmark criteria contained in the surveillance procedure are met, then an MTC measurement in accordance with SR 4.1.1.3.b is not required to be performed. The neutronics methods used with WCAP-13749-P-A are those described in WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
2.7 Analog Rod Position Indication System (TS  3.1.3.2)
  - Figure A3 (page 14A-A9)    The All Rods Out (ARO) position for all shutdown Banks and Control Banks is defined to be 228 steps withdrawn.
 
2.8 Control Rod Insertion Limits (TS  3.1.3.6)
 
  - Figure A3 (page 14A-A9)    The control rod banks shall be
 
limited in physical insertion as specified in Figure A3 for ARO =228 steps withdrawn.
 
2.9 Axial Flux Difference (TS  3.2.1)
 
  - Figure A4 (page 14A-A10)
 
14A-A5 Revised 09/01/2016 C28C28C28C28C28 Turkey Point Unit 3 Cycle 28 Core Operating Limits Report (COLR) 2.10 Heat Flux Hot Channel Factor F Q(Z)  (TS  3.2.2)
  - [FQ]L = 2.30    - K(z) = 1.0 For 0'  z  12' where z is core height in ft 2.11 Nuclear Enthalpy Rise Hot Channel Factor  (TS  3.2.3)
  - FHRTP = 1.600 PFH  =  0.3
 
2.12 DNB Parameters  (TS  3.2.5)
    - RCS Tavg < 585.0 oF 
  - Pressurizer Pressure > 2204 psig
 
14A-A6 Revised 03/11/2016 C28C28 Figure A1 Reactor Core Safety Limit - Three Loops in Operation
 
14A-A7 Revised 03/11/2016
 
Figure A2 Required Shutdown Margin vs Reactor Coolant Boron Concentration
 
14A-A8  Revised 03/11/2016
 
FIGURE A3 Turkey Point Unit 3 Cycle 28 Rod Insertion Limits vs Thermal Power ARO = 228 Steps Withdrawn, Overlap = 100 Steps
 
14A-A9  Revised 03/11/2016 C28 FIGURE A4 Axial Flux Difference as a Function of Rated Thermal Power Turkey Point Unit 3 Cycle 28
 
14A-A10  Revised 03/11/2016 C28
 
APPENDIX 14B
 
TURKEY POINT PLANT UNIT 4 CYCLE 29 RELOAD CHARACTERISTICS AND PARAMETERS
 
14B-i Revised 07/21/2016 C28 TABLE OF CONTENTS
 
Section Title Page 
 
==1.0  INTRODUCTION==
AND SUMMARY  ......................................................................14B-1
 
1.1 Introduction  ................................................................................................14B-1  1.2 General Description  ...................................................................................14B-1 Appendix A  Turkey Point Unit 4 Cycle 29  ...........
................
...............
................
................
.......14B-A1  Core Operating Limits Report (COLR)
 
14B-ii      Revised 07/21/2016 C28 LIST OF TABLES
 
Table Title Page
 
14B-1 Fuel Assembly Design Parameters................................................................................ 14B-2  Turkey Point Unit 4 - Cycle 29
 
14B-2 Kinetics Characteristics.................................................................................................. 14B-3  Turkey Point Unit 4 - Cycle 29 14B-3 Shutdown Requirements and Margins........................................................................... 14B-4  Turkey Point Unit 4 - Cycles 28 and 29
 
LIST OF FIGURES Figure  14B-1 Reference Core Loading Pattern.................................................................................... 14B-6  Turkey Point Unit 4 Cycle 29 14B-2 Burnable Absorber Locations........................................................................................ 14B-7 Turkey Point Unit 4 Cycle 29
 
14B-iii Revised 07/21/2016 C28C28C28C28C28
 
==1.0 INTRODUCTION==
and SUMMARY
 
===1.1 Introduction===
This report presents reload characteristics and parameters associated with Turkey Point Unit 4 Cycle 29. The Cycle 29 core is a fullcore with 15x15 Upgrade fuel assemblies in Region 28, 29, 30, and 31.
1.2 General Description The Turkey Point Unit 4 reactor core is comprised of 157 fuel assemblies arranged in the core loading pattern configuration shown in Figure 14B-1.
All fuel assemblies have axial blankets at both the top and bottom of the stack to reduce neutron leakage and to improve uranium utilization. Regions 28, 29, 30, and 31 fuel assembly blankets are 8 inches long, with Natural UO 2 annular pellets in Region 28 and 2.6 w/o enriched UO 2 annular pellets in the other regions. The design parameters for the Cycle 29 core are provided in Table 14B-1.
The Cycle 29 core uses a single type of burnable absorber, the Westinghouse Integral Fuel Burnable Absorber (IFBA) rods composed of ZrB 2 coated fuel pellets with 2.2125 mg 10B/in. The active absorber length is 120 inches. Their locations in the core are shown in Figure 14B-2.
The core design parameters for Cycle 29 are as follows:
Parameter Current Licensing Basis Core Power (MW t)    2644  Pressurizer Pressure (psia)    2250  Core Inlet Temperature 1 (&deg;F)    549.2  Core Inlet Temperature 2 (&deg;F)    550.2  Thermal Design Flow (gpm)    260,700 Minimum Measured Flow    270,000  Average Linear Power Density 3 (kW/ft)  6.714  1. Based on Thermal Design Flow. 2. Based on Minimum Measured Flow.
The core kinetics characteristics and shutdown requirements and margins are provided in Tables 14B-2 and 14B-3, respectively. The Core Operating Limits Report (COLR) for Cycle 29 is provided in Appendix A.
 
14B-1 Revised 07/21/2016 C28C28C28C28C28 Table 14.B-1 Fuel Assembly Design Parameters Turkey Point Unit 4 Cycle 29 Region 28C 29B 29C 29D 29F 30A 30B 30C 30D 30E 30F 30G 31A 31B 31C Enrichment 1  (w/o U235) 4.006 4.009 4.009 4.405 4.405 3.807 3.807 3.807 4.199 4.400 4.400 4.400 3.900 4.100 4.400 Density1 (%  Theoretical) 95.67 95.88 95.88 95.71 95.71 95.65 95.65 95.65 95.05 95.91 95.91 95.91 95.50 95.50 95.50 Number of Assemblies 1 4 4 8 8 8 4 20 8 8 8 8 12 36 20 Approximate Burnup  at BOC 29 23,088 33,351 33,387 36,575 36,555 25,160 24,969 24,780 24,181 21,437 22,062 23,725 0 0 0 (MWD/MTU) 2      Fuel Type Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade Upgrade    Number of  IFBA/Assembly 116 100 148 32 148 80 116 148 116 48 64 80 4@16 8@148 4@100 20@116 12@148 8@16 8@32 4@80    Total Number of IFBA 116 400 592 256 1184 640 464 2960 928 384 512 640 1248 4496 704 Fuel Rods/Region
 
Axial Blankets  (AB) 3 YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES AB Annular Pellets YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES AB Enrichment (w/o) NAT U 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.6 Optimized ZIRLO TM Cladding NO NO NO NO NO NO NO NO NO NO NO NO YES YES YES Notes 1. As built values for burned regions and design values for fresh regions. 2 Based on assumed Cycle 28 burnup of 19,292 MWD/MTU (Long Window). 3. Axial blankets in all regions are 8 inches long.
 
14B-2 Revised 07/21/2016 C28 Table 14B-2 KINETICS CHARACTERISTICS TURKEY POINT UNIT 4 - Cycle 29 Moderator Temperature Coefficient (pcm/&deg;F)  Current Limit Cycle 29
: a. Most positive +5.0 (70% RTP) +2.2 (HZP, 541 &deg;F, 2000    (linear ramp  MWD/MTU), linear ramp    to 0 at 100% RTP)  rate to 0 at 100% RTP
: b. Most negative  -41  -36.2 Doppler Coefficient (pcm/&deg;F) -2.9 to -1.0 -1.94 to -1.22 Delayed Neutron Fraction eff (%) 0.44 to 0.75 0.48 to 0.65
 
Maximum Differential Rod 100  57.8 Worth of Two Banks Moving Together at HZP (pcm/in)
Available Shutdown Margin (%)  a. BOC 1.00*  3.197 1.77**  b. EOC 1.77  1.865
* MODES 1 through 4 with at least 1 RCP running ** MODE 4 without RCPs running and MODE 5
 
14B-3 Revised 07/21/2016 C28C28C28C28C28 Table 14B-3 Shutdown Requirements and Margins  Turkey Point Unit 4 - Cycles 28 and 29 Cycle 28 Cycle 29    BOCEOCBOC EOCControl Rod Worth (%)          All Rods Inserted Less Worst Stuck Rod 5.78 5.79 5.47 5.68      (1)  Less 7% 5.37 5.38 5.09 5.29      Control Rod Requirements (%)          Reactivity Defects (Doppler, TAVE, Void, and Redistribution) 1.97 3.41 1.89 3.42      Rod Insertion Allowance (RIA)  RCCA Repositioning Allowance (see note) --- --- --- ---     
(2) Total Requirements 1.97 3.41 1.89 3.42 Shutdown Margin (1) - (2) (%) 3.41 1.97 3.20 1.87 Required Shutdown Margin (%) 1.00 1.77 1.00 1.77
 
Note: Additional margin to accommodate a 22 &deg;F cooldown is not available for EPU cycles. The RIA term is already included in the Reactivity Defects term in the methodology used to compute the shutdown margin.
 
14B-4  Revised 07/21/2016 C28 Figure 14B-1 Turkey Point Unit 4, Cycle 29 Reference Core Loading Pattern 14B-5  Revised 07/21/2016 C28 Figure 14B-2 Turkey Point Unit 4, Cycle 29 Burnable Absorber and Source Rod Locations
 
14B-6  Revised 07/21/2016 C28
 
Appendix A Turkey Point Unit 4 Cycle 29 Core Operating Limits Report (COLR)
 
14B-A1 Revised 07/21/2016 C28 Turkey Point Unit 4 Cycle 29 Core Operating Limits Report 1.0 Introduction This Core Operating Limits Report for Turkey Point Unit 4 Cycle 29 has been prepared in accordance with the requirements of Technical Specification 6.9.1.7. The Technical Specifications (TS) affected by this report are listed below with the section and page for each one of the TS addressed in this COLR document. Section Technical Specification Page  2.1  2.1.1  Reactor Core Safety Limits 14B-A3  2.2  2.2.1  Reactor Trip System Instrumentation Setpoints 14B-A3-14B-A4  2.3  3.1.1.1  Shutdown Margin Limit for MODES 1, 2, 3, 4 14B-A4  2.4  3.1.1.2  Shutdown Margin Limit for MODE 5 14B-A4  2.5  3.1.1.3  Moderator Temperature Coefficient 14B-A5  2.6  4.1.1.3  MTC Surveillance at 300 ppm 14B-A5  2.7  3.1.3.2  Analog Rod Position Indication System 14B-A5  2.8  3.1.3.6  Control Rod Insertion Limits 14B-A5  2.9  3.2.1  Axial Flux Difference 14B-A5  2.10  3.2.2  Heat Flux Hot Channel Factor F Q(Z)  14B-A5  2.11  3.2.3  Nuclear Enthalpy Rise Hot Channel Factor 14B-A6  2.12  3.2.5  DNB Parameters  14B-A6  Figure  Description A1    Reactor Core Safety Limit - Three Loops in Operation 14B-A7 A2    Required Shutdown Margin vs Reactor Coolant Boron Concentration 14B-A8 A3    Turkey Point Unit 4 Cycle 29 Rod Insertion Limits vs Thermal Power 14B-A9 A4    Axial Flux Difference as a Function of Rated Thermal Power  14B-A10     
 
14B-A2 Revised 07/21/2016 C28C28C28 Turkey Point Unit 4 Cycle 29 Core Operating Limits Report 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in the Introduction are presented below and listed sequentially by Technical Specification (TS). These limits have been developed using the NRC-approved methodologies specified in TS 6.9.1.7.
2.1 Reactor Core Safety Limits - Three Loops in Operation (TS  2.1.1)
- Figure A1(page 14B-A7)
In Modes 1 and 2, the combination of Thermal Power, reactor coolant system highest loop average temperature and pressurizer pressure shall not exceed the limits in Figure A1. 2.2 Reactor Trip System Instrumentation Setpoints (TS 2.2.1)  NOTE 1 on TS Table 2.2-1 Overtemperature T  - 1 = 0s, 2 = 0s  Lead/Lag compensator on measured T - 3 = 2s  Lag compensator on measured T - K1 = 1.31  - K2 = 0.023/F - 4 = 25s, 5 = 3s  Time constants utilized in the lead-lag compensator for Tavg - 6 = 2s  Lag compensator on measured Tavg - T  583.0 F  Indicated Loop Tavg at RATED THERMAL POWER
- K3 = 0.00116/psi 
- P'  2235 psig  Nominal RCS operating pressure
- f1(I) = 0 for q t - qb between - 18% and + 7%
.      For each percent that the magnitude of q t - qb exceeds - 18%,    the T Trip Setpoint shall be automatically reduced by 3.51% of its value at RATED THERMAL POWER; and    For each percent that the magnitude of q t - qb exceeds +7%, the T Trip Setpoint shall be automatically reduced by 2.37% of its value at RATED THERMAL POWER.
Where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q t + qb is total THERMAL POWER in percent of RATED THERMAL POWER. 
 
14B-A3 Revised 07/21/2016 C28 Turkey Point Unit 4 Cycle 29 Core Operating Limits Report NOTE 2 on TS Table 2.2-1 Overtemperature T  The Overtemperature T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel, 0.2% T span for the Pressurizer Pressure channel, and 0.4% Tspan for the f(I) channel. No separate Allowable Value is provided for Tavg because this function is part of the T value.
NOTE 3 on TS Table 2.2-1 Overpower T  - K4  = 1.10 - K5  0.0/F  For increasing average temperature
- K5 =  0.0/F For decreasing average temperature
- 7  0 s Time constants utilized in the lead-lag compensator for Tavg - K6 = 0.0016/F For T > T"
- K6 = 0.0 For T  T"  - T"  583.0F Indicated Loop Tavg at RATED THERMAL POWER
- f2 (I) = 0  For all I  NOTE 4 on TS Table 2.2-1 Overpower T The Overpower T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel. No separate Allowable Value is provided for Tavg because this function is part of the T value.
2.3 Shutdown Margin Limit for MODES 1, 2, 3 and 4 (TS  3.1.1.1)
- Figure A2 (page 14B-A8) 2.4 Shutdown Margin Limit for MODE 5 (TS  3.1.1.2)
- > 1.77 % k/k   
 
14B-A4 Revised 07/21/2016 C28 Turkey Point Unit 4 Cycle 29 Core Operating Limits Report 2.5 Moderator temperature coefficient (MTC)  (TS  3.1.1.3)
- < + 5.0 x 10
-5 k/k/F  BOL, HZP, ARO and from HZP to 70% Rated Thermal Power (RTP)
  - From 70% RTP to 100% RTP the MTC  decreasing linearly from < + 5.0 x 10
-5 k/k/F  to < 0.0 x 10
-5 k/k/F      - Less negative than - 41.0 x 10
-5 k/k/F  EOL, RTP, ARO 2.6 Moderator temperature coefficient (MTC) Surveillance at 300 ppm (TS  4.1.1.3)
- Less negative than - 35.0 x 10
-5 k/k/F (-35 pcm/F) Within 7 EFPD of reaching equilibrium boron concentration of 300 ppm.
The Revised Predicted near - EOL 300 ppm MTC shall be calculated using the algorithm contained in WCAP-13749-P-A: Revised predicted MTC = Predicted MTC + AFD Correction - 3 pcm/F If the Revised Predicted MTC is less negative than the SR 4.1.1.3.b 300 ppm surveillance limit and all the benchmark criteria contained in the surveillance procedure are met, then an MTC measurement in accordance with SR 4.1.1.3.b is not required to be performed. The neutronics methods used with WCAP-13749-P-A are those described in WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores,"  June 1988 2.7 Analog Rod Position Indication System (TS  3.1.3.2)
- Figure A3 (page 14B-A9)
The All Rods Out (ARO) position for all shutdown Banks and Control Banks is defined to be 229 steps withdrawn.
2.8 Control Rod Insertion Limits (TS  3.1.3.6)
- Figure A3 (page 14B-A9)
The control rod banks shall be limited in physical insertion as specified in Figure A3 for ARO = 229 steps withdrawn.
2.9 Axial Flux Difference (TS  3.2.1)
- Figure A4 (page 14B-A10)
 
14B-A5 Revised 09/20/2016 C28C28C28C28 Turkey Point Unit 4 Cycle 29 Core Operating Limits Report 2.10 Heat Flux Hot Channel Factor F Q(Z)  (TS  3.2.2)
- [FQ]L = 2.30  - K(z) = 1.0  For 0' <  z < 12' where z is core height in ft 2.11 Nuclear Enthalpy Rise Hot Channel Factor  (TS  3.2.3)
- FHRTP = 1.600  PFH  =  0.3 2.12 DNB Parameters  (TS  3.2.5)
- RCS Tavg < 585.0 oF  - Pressurizer Pressure > 2204 psig 
 
14B-A6 Revised 07/21/2016 C28 Figure A1 Reactor Core Safety Limit - Three Loops in Operation 14B-A7 Revised 07/21/2016
 
Figure A2 Required Shutdown Margin vs Reactor Coolant Boron Concentration
 
14B-A8 Revised 02/24/2015
 
Figure A3 Turkey Point Unit 4 Cycle 29 Rod Insertion Limits vs Thermal Power ARO = 229 Steps Withdrawn, Overlap = 101 Steps
 
14B-A9 Revised 07/21/2016 C28 Figure A4 Axial Flux Difference as a Function of Rated Thermal Power Turkey Point Unit 4 Cycle 29
 
14B-A10 Revised 07/21/2016 020406080100120-50-40-30-20-1001020304050Percent of RATED THERMAL POWER (%)Axial Flux Difference (%)(-9,100)(+6,100)(-30,50)(+20,50)UNACCEPTABLE OPERATIONUNACCEPTABLE OPERATIONACCEPTABLE OPERATIONC28
 
APPENDIX 14 C TURKEY POINT UNITS 3 AND 4 UPDATED FSAR   
 
MODIFICATION OF THE TURBINE RUNBACK SYSTEM
 
THIS APPENDIX HAS BEEN ENTIRELY DELETED 
 
FLORIDA POWER AND LIGHT COMPANY
 
14C-1 Rev. 12  5/95 
 
APPENDIX 14D
 
FLORIDA POWER AND LIGHT COMPANY
 
TURKEY POINT UNITS 3 AND 4
 
DELETED IN ITS ENTIRETY REFER TO CHAPTER 9, SECTION 9.5 AND CHAPTER 14, SECTION 14.2.1.3
 
HIGH DENSITY SPENT FUEL STORAGE RACKS
 
14D-1 Revised 09/29/2005
 
APPENDIX 14E
 
FLORIDA POWER AND LIGHT COMPANY
 
TURKEY POINT UNITS 3 AND 4
 
DELETED IN ITS ENTIRETY REFER TO CHAPTER 9, SECTION 9.5 AND CHAPTER 14, SECTION 14.2.1.3
 
SPENT FUEL STORAGE FACILITY MODIFICATION
 
SAFETY ANALYSIS REPORT
 
14E-i      Revised 09/29/2005 APPENDIX 14F  ENVIRONMENTAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT
 
This appendix contains the original licensing basis LOCA dose analysis. This
 
analysis has been replaced with a revised analysis that can be found in
 
Section 14.3.5.
 
The results of analyses described in this section demonstrate that the
 
amounts of radioactivity released to the environment in the event of a
 
loss-of-coolant accident (which has an exceedingly low probability of
 
occurrence) are substantially less than the guidelines specified in 10 CFR
 
100. In summary, the computed thyroid dose values are (using the release
 
assumption of TID-14844):
 
North Boundary    South    Low Population    Exclusion Radius  Boundary    Distance
 
Integrated Dose        4164 ft        5582 ft      5 miles 0-2 hour dose, rem  93      65    9  0-31 day dose, rem  109      75    10
 
Loss-of-Coolant Accident
 
The loss-of-coolant accident has the potential for the highest off site
 
doses, compared to all other accidents. The loss of coolant accident may
 
result in a significant amount of clad rupture; however, since the fuel does
 
not melt, only a limited quantity of fission products are released. If it is
 
assumed that all the rods fail and that all the fission products in the gap
 
spaces were released, the total release from the core would be less than 5%
 
of the saturation quantities of the radioactive iodines and noble gases. 
 
For analytical purposes the amount of radioactive fission products that could
 
be released from the core have been calculated according to the fundamental
 
assumptions given in Reference 1 (TID 14844). This calculational model has
 
been widely used in evaluating the capability of PWR containment systems in
 
the event of the core melt down. However, it should be pointed out that no
 
accident of this magnitude has been described for these units; in fact, an
 
accident of this magnitude is not considered credible.
 
14F-1 Rev. 10  7/92 The TID 14844 model assumes that 50% of the total core iodine inventory is released, and that one half of this amount becomes plated out onto surfaces
 
within the containment. The remaining one half, or 25% of the total core
 
iodine inventory, is assumed to be in the containment atmosphere and
 
available for leakage. As a function of time the charcoal filter system
 
collects and retains the iodine, and thereby the amount of iodine available
 
for leakage is substantially reduced. 
 
The TID 14844 model also assumes that 100% of the total core noble gas
 
inventory and 1% of the total core solid fission product inventory are
 
released into the containment. 
 
Core Inventory of Iodines and Noble Gases
 
The total core inventory was calculated on the basis of the reactor having
 
been operated as follows: (1) 2300 MW(t), (2) 625 days of full-power
 
operation to produce 1-129 and the stable isotopes, and (3) except for I-129,
 
full-power operation to reach the saturation inventory of the radioactive
 
isotopes. Table 14F-1 gives information on the major iodine isotopes
 
computed for the Turkey Point core, based on data given in TID 14844. Table
 
14F-2 gives information on the major noble gas isotopes. 
 
Iodines and Noble Gases in Containment Atmosphere
 
The amount of noble gases in the containment atmosphere at time zero
 
(according to the TID 14844 model) is the total amount listed in Table 14F-2.
 
These gases are assumed to be completely mixed in the atmosphere, and
 
available for leakage. 
 
The amount of iodine in the containment atmosphere at time zero (according to
 
the TID 14844 model, 25% of total) adds up to the following:
 
Total of I-127 and I-129  2,550 grams, stable Total of I-131, I-132, I-133, I-134 and 1-135      152 grams, radioactive  Total Iodine in Containment  2,702 grams 
 
14F-2 Rev. 10  7/92 The iodine, when released from the core, has been observed by those working in the field to be essentially composed of elemental iodine with little more
 
than a trace of organic iodides. Upon reaching the containment, and as a
 
function of time, some of the elemental iodine reacts with organic materials
 
to form organic iodides, typified by methyl iodide. Also, some hydrogen
 
iodide is formed.
 
The percentage of the iodine in the containment atmosphere that becomes
 
converted into methyl iodide is not precisely known. The best evidence
 
indicates that the value lies between an infinitesimal amount and 5%. It is
 
stated in Reference 2 that "Although there is only a small amount of
 
information available on which to base a judgement, a value of 10% for
 
organic (nonremovable) iodides in the total available for leakage is
 
considered very conservative...". For dose calculations the elemental iodine
 
was taken as 95% and the methyl iodide as 5%.
 
With respect to iodine cleanup, the dose calculations are based on the
 
removal that occurs only in the charcoal filter units and the 50% plateout
 
previously mentioned. That is a conservative assumption since cleanup will
 
also be achieved as follows:
: 1. Some iodine will be deposited on particles in the atmosphere. Some of these particles will be entrained by the containment borated spray
 
water. The remainder of the particles will be collected in the HEPA
 
filters. 
: 2. Based on information given in Reference 3, and companion reports, the elemental iodine (and iodides other than organic) in the atmosphere may
 
be effectively cleaned up by the containment spray water. This cleanup
 
by the water is not permanent (since no iodine retaining agent is
 
added) in that the iodine will seek an equilibrium distribution between
 
the water and the air in accordance with its partition factor.
 
14F-3 Rev. 10  7/92 Iodine Cleanup With Emergency Containment Filter Units The capability of the emergency containment filter units to collect elemental
 
iodine and methyl iodide is indicated by a "decontamination factor" (DF),
 
which in turn depends upon a "removal constant" (). Removal constants were computed on the basis of the equation and numerical values given in Table
 
14F-3. The following removal constants were computed:
 
Number of Filter              Elemental Iodine          Methyl Iodide
 
Units Operating a                    b        3 (total installed)                  3.53                      2.74 
 
2 (minimum safeguards)              2.35                      1.83 
 
The general decontamination factor equation is given in Table 14F-4. With
 
the use of this equation the following decontamination factors were
 
calculated, based on the iodine in the containment being composed of 0.95
 
elemental iodine and 0.05 methyl iodide:
 
2 Filter Units            3 Filter Units 
 
Operating                Operating
 
Time period DF DF 0-2 hours                            4.68                      6.97 
 
2-12 hours                        > l00*                    > l00* 
 
12 hours - 31 days                > l00*                    > l00*
 
Containment Assumptions
 
The containment design leak rate is 0.25% per day (2.9 x 10
: 8) fraction/sec) at the design pressure of 59 psig. In the event of a loss-of-coolant
 
accident the containment pressure will rise to some value less than 59 psig,
 
and will then decrease to near atmospheric pressure due to the action of the
 
containment sprays and emergency containment coolers.
* These values were arbitrarily limited in order to obtain a finite number in the dose calculations.
 
14F-4 Rev. 10  7/92 For the dose calculations the pressure of the containment was assumed toremain at 59 psig for the entire length of the period, and thereby the leak 
 
rate was taken as a fixed value of 0.25% per day. This assumption tends to
 
be very conservative, particularly for the "12 hours-31 days" period.
 
Atmospheric Dispersion Model
 
For calculational purposes, the pressurized air-steam mixture in the
 
containment was assumed to leak out at the established leak rate given above.
 
This leakage from the containment becomes dispersed into the atmosphere and
 
the dose rate to an individual at any specific location is a function of
 
source concentration, time, distance, and atmospheric dispersion.
 
Dilution multipliers (x/Q), which reflect relative concentrations of
 
radioactivity in the atmosphere as a function of distance from the
 
containment, were calculated in accordance with equations and meteorological
 
conditions given in Tables 14F-5 and 14F-6.
 
No credit was taken for the building wake effect for either the "2-12 hours"
 
period or the "12 hours-31 days" period. This introduces some conservatism
 
near the site boundary, but the error diminishes with distance. The values
 
of y and z were taken from Reference 4.
 
The dilution multiplier values (in seconds/cubic meter) for the stated
 
conditions at various locations are tabulated below:
 
North Boundary    South    Low Population    Exclusion Radius  Boundary    Distance
 
Time period        4164 ft        5582 ft      5 miles 
 
0-2 hours      154 x 10
-6  108 x 10
-6    15.0 x 10
-6  2-12 hours      108 x 10
-6  66 x 10
-6      6.5 x 10
-6  12 hours - 31 days    4.32 x 10
-6    2.64 x 10
-6    0.24 x 10
-6   
 
14F-5 Rev. 10  7/92 Thyroid Dose Computations The thyroid doses for various time periods were calculated according to the
 
equation and values given in Table 14F-7. 
 
The following values were obtained:
 
North Boundary    South    Low Population    Exclusion Radius  Boundary    Distance
 
Integrated Dose        4164 ft        5582 ft      5 miles 0-2 hour dose, rem  93      65    9 
 
0-31 day dose, rem  109      75    10 
 
These values demonstrate that the amount of radioactivity that would be
 
released to the environment in the event of a loss-of-coolant accident give
 
dose values that are substantially less than the guidelines specified in 10
 
CFR 100.
 
Several parameter studies were performed in order to indicate the change in
 
thyroid dose values that would result in the event of a deviation in an 
 
original assumption. For example, it was found that the doses remain almost
 
unaffected in case of filter unit fan failure after a brief period of time.
 
The above given dose values were based on two filter units operating
 
continuously for the duration of the accident. The principal cleanup occurs
 
within the first two hours; in fact, within this period of time the iodine
 
concentration will be reduced to less than 2% of the original concentration. 
 
14F-6 Rev. 10  7/92 After two hours, the filter units serve to continue cleaning the air of residual amounts of iodine. The following tabulation illustrates the
 
insensitivity of the dose values due to equipment malfunction after two
 
hours.
 
Dose at Exclusion Radius, rem
 
Classification  Condition      0-2 hours  0-31 days
 
Normal  Two filter units operating  93  109    31 days or longer. 
 
Abnormal  One filter unit operating  93  110 31 days or longer. Second filter unit operating for first 2 hours only. 
 
Abnormal  Two filter units operating  93  111    first 2 hours only.
 
In case a filter unit does fail after operating for a period of time, the
 
radioactive decay heat is absorbed by the borated water spray system to the
 
filters, thereby holding the collected iodine within the charcoal. 
 
Another example is the sensitivity of the system to the methyl iodide
 
content, since it cannot be established at this time precisely what fraction
 
of the iodine will be in the methyl iodide form. Calculations were made to
 
examine the variation in the 0-2 hour dose at the north boundary that would
 
occur if the methyl iodide content in the containment atmosphere varied from
 
0% to as much as 15%.
 
Methyl Iodide, Fraction 
 
                          .00
          .05
          .10
        .15 DF                      4.75        4.68        4.62        4.56 Dose, rem                92          93          94          95 
 
For the calculations it was assumed that two filter units were operating with
 
a  of 2.35 for elemental iodine and a  of 1.83 for methyl iodide, as given earlier. One concludes from the above that the exact amount of methyl iodide
 
does not need to be known since the total dose varies very little. 
 
14F-7 Rev. 10  7/92 A third example is the sensitivity of the system to unfilterable iodide. The concept of an unfilterable form of airborne iodine is hardly consistent with
 
any physical model of filtration. It is possible, but not reasonable, on the
 
basis of a thorough examination of the data (refer to references given in
 
Reference 5), that some forms of iodine might be removed at very low
 
efficiencies. It is a simplified approach to the calculations to assume that
 
there is a form of iodine which is "unfilterable," or will be removed at zero
 
percent efficiency, even though this does not agree with experimental data. 
 
In order to show sensitivity, calculations were made on the assumption of
 
varying amounts of unfilterables to determine the variation in the 0-2 hour
 
dose at the north and south boundaries, and the 0-31 day dose at a distance
 
of 5 miles, with the unfilterable iodine varying in concentration from zero
 
to 15% of the iodine concentration in the containment atmosphere. The
 
results, with 2 filter units operating, were as follows:
 
Fraction of Iodine that is Unfilterable:
Integrated  Dose
                    .00
        .05
        .10
        .15 0-2  hr Dose, rem, North Boundary    92        108        125        142 0-2  hr Dose, rem, South Boundary    64        76        88        100 
 
0-31 day Dose, rem, at 5 Miles        10        16        22        28 
 
In reviewing the results computed on this basis, it is seen that the doses
 
are all much less than 300 rem, even with the unfilterable content being 15%.
 
Although the applicant does not believe that this calculational model is the
 
proper one to use, it should be noted that the calculated dose values are
 
low.
 
Short-term Thyroid Doses at Beach and Scout Camps
 
The maximum thyroid doses have also been considered for areas within the site
 
boundary temporarily occupied by the public assuming the TID-14844 accident
 
analysis model. These areas are the Turkey Point Beach at 2000 feet, the
 
Girl Scout Camp at 2300 feet and the Boy Scout Camp at 2900 feet from the nearest containment structure. The respective /Q values at these distances, considering the volume source correction, are 3.2 x 10
-4, 2.8 x 10
-4 and 2.3 x 10-4 sec/M3 for the period of 0 to 2 hours following the postulated LOCA.
 
14F-8 Rev. 10  7/92 By selection of a very conservative value of 59 psig maximum containment pressure for the leakage driving function over the entire initial two hours,
 
the effective maximum containment leak rate is 0.25% / day. The resultant
 
maximum two hour thyroid dose at the indicated locations, generated from an
 
initial 95% elemental iodine and 5% methyl iodide atmospheric constituency,
 
are:
 
Turkey Point Beach 190  rem Girl Scout Camp 170  rem Boy Scout Camp 138  rem
 
These values point out the requirement for the site evacuation procedure to
 
be implemented within the initial 2 hour period, which will be provided and
 
followed. 
 
Whole Body Dose Computations
 
Whole body doses resulting from the accident were also computed. The major
 
contribution is the dose from immersion in the plume. The direct radiation
 
dose from the containment is insignificant due to the shielding provided by
 
its walls. 
 
Direct doses were calculated assuming immersion in a semi-infinite cloud
 
containing a uniform distribution of the gas isotopes which have leaked from
 
the containment. Cloud concentrations assumed were those actually calculated
 
at the centerline of the plume. 
 
The following whole body doses from the passing cloud were computed:
 
North Boundary    South    Low Population    Exclusion Radius  Boundary    Distance
 
Integrated Dose        4164 ft        5582 ft      5 miles 0-2 hour dose, rem  3.1  2.2  0.4  0-31 day dose, rem  5.2  3.5  0.6  These values are small compared to the guidelines specified in 10 CFR 100.
 
14F-9 Rev. 10  7/92 Radiological Assessment of Containment Purge The radiological doses due to a postulated loss of coolant accident presented
 
in the proceeding analyses assumed that there was no containment purging
 
occurring at the onset of the accident. Discussed herein are the results of
 
an analysis performed to determine the incremental radiological dose at the
 
site boundary and low population zone assuming the purge valves are fully
 
open when the accident initiates and close upon receipt of signal as
 
designed. These incremental doses, when added to those previously presented
 
in Section 14.3.5, provide a maximum set of doses for a LOCA with containment
 
purge. The results of this evaluation are presented in the following tables:
(6)
THYROID DOSE (rem)
Increment Due
 
Location LOCA To Purging Total 
 
Site Boundary -              93                    10                103 
 
(0-2 hour)
 
Low Population Zone -        9                      1                  10 
 
(0-2 hour)
 
WHOLE BODY (rem)
 
Increment Due
 
Location LOCA To Purging Total 
 
Site boundary -            3.1                  .002                3.1 
 
(0-2 hour) 
 
Low Population Zone -      .4
                  .0002                .4 (0-2 hour)
 
The major assumptions which were used in the evaluation of the incremental
 
dose are listed below:
: 1. The containment purge valves are closed 5 seconds after the containment high pressure signal is transmitted. There is a 2.7 second delay before
 
14F-10 Rev. 10  7/92 the increased containment pressure is detected which results in a total of 7.7 seconds for valve closure (8 seconds was conservatively assumed).
: 2. Radioactive releases via the purge valves during closure is from the Reactor Coolant System only.
: 3. The primary coolant iodine activity corresponds to the maximum limit of 30 Ci/gm Dose Equivalent.
: 4. It is conservatively assumed during the initial 8 seconds that 5O% of the blowdown (worst FSAR case) from the break flashes and becomes
 
homogeneously mixed in the containment atmosphere. All of the iodine in
 
the flashed steam is assumed to become airborne.
: 5. The flow through the purge valves is assumed to be a mixture of steam and water. Frictionless flow through the valves is assumed.
: 6. FSAR meteorology is assumed.
: 7. Standard TID 14844 methodology was used to calculate the incremental doses.
The results clearly indicate that the anticipated dose caused by a LOCA with
 
containment purging at the onset of the accident is well within the limits of
 
10 CFR 100.
 
14F-11 Rev. 10  7/92 References
: 1. J. J. DiNunno, F. D. Anderson, R. E. Baker, and R. L. Waterfield, Calculation of Distance Factors for Power and Test Reactor Sites, USAEC Report TID-14844, March 23, 1961.
: 2. Supplemental Safety Evaluations by the Division of Reactor Licensing, United States Atomic Energy Commission, in the Matter of Florida Power
 
and Light Company, Turkey Point Units 3 & 4, July 12, 1968.
: 3. Nuclear Safety Program Annual Progress Report for Period Ending December 31, 1967, Oak Ridge National Laboratory, ORNL-4228, April 1968.
: 4. W. F. Hilsmeier and F. A. Gifford, Jr., Graphs for Estimating Atmospheric Dispersion, Report ORO-545, Weather Bureau Research Station,
 
Oak Ridge, Tenn., August 23, 1962.
: 5. Supplement No. 14 to Application for Licenses, re Florida Power & Light Company, Turkey Point Units 3 & 4, USAEC Docket Nos. 50-250, 50-251,
 
March 14, 1968.
: 6. R. E. Uhrig (FPL) letter #L-79-346, to A. Schwencer (NRC), dated December 13, 1979, "Containment Purge".
 
14F-12 Rev. 10  7/92 TABLE 14F-1  IODINE ISOTOPES AND THEIR ESTIMATED QUANTITIES FOR A FULL CORE INVENTORY AT TIME ZERO
 
Isotope              Half-Life Grams Curies
 
I-127              Stable                      2,040        0
 
I-129              1.72 x 10 7 years            8,170          ~ 0 I-131              8.05 days                  452            57.7 x 10 6 I-132              2.4 hours                  8.25            87.5 x 10 6 I-133              20.8 hours                  109.7          129.5 x 10 6 I-134              52.5 minutes                5.35            151.3 x 10 6 I-135              6.68 hours                  31.9            117.2 x 10 6
Lumping all radioactive isotopes into an I-131 equivalent                                      109 x 106 
 
Rev. 10  7/92 TABLE 14F-2  NOBLE GAS ISOTOPES AND THEIR ESTIMATED QUANTITIES FOR A FULL CORE INVENTORY AT TIME ZERO
 
Isotope                        Half-Life Curies
 
Kr-83m                          114 minutes                      10.6 x 10 6
Kr-85                          10.76 Years                      0.83 x 10 6
Kr-85m                          4.36 hours                      25.5 x 10 6
Kr-87                          78 minutes                      47.3 x 10 6
Kr-88                          2.77 hours                    64.3 x 10 6
Xe-131m                        12.0 days                        0.46 x 10 6
Xe-133m                        2.3 days                        3.08 x 10 6
Xe-133                          5.27 days                        128.4 x 10 6
Xe-135m                        15.6 minutes                    41.5 x 10 6
Xe-135                          9.13 hours                      32.0 x 10 6 
 
Rev. 10  7/92 TABLE 14F-3  EQUATION FOR REMOVAL CONSTANT
                            = n v e m 60 V
 
  =  removal constant, per hour n  =  number of filter units operating
 
v  =  atmosphere flow through each filter unit, cu ft/min
 
e  =  charcoal filter efficiency, fraction
 
m  =  atmosphere mixing factor, fraction
 
V  =  free volume of containment, cu ft
 
Elemental iodine      Methyl iodide
 
a              b     
 
v                                        37,500                37,500
 
e                                          0.9                    0.7
 
m                                          0.9                    0.9
 
v                                      1.55 x 10 6            1.55 x 10 6 
 
Rev. 10  7/92 TABLE 14F-4 GENERAL DECONTAMINATION FACTOR EQUATION
 
DF =                                    1                               
 
Fa =      filterable elemental iodine, fraction of total iodine in containment
 
atmosphere.
 
Fb =      filterable methyl iodide, fraction of total iodine in containment atmosphere.
 
Fc=      unfilterable iodine and iodide; engineering tests indicate no components to be unfilterable; therefore, this is assumed to
 
be zero.
 
t1=      time of operation prior to the period under consideration, hours.
 
t2=      time of operation during the period under consideration, hours.
 
Rev. 10  7/92 F + t e - 1 eF + t e - 1 eFc2bt-t-b2at-t-a2b1b2a1a TABLE 14F-5  DILUTION MULTIPLIER EQUATIONS
 
Time period
 
X =  concentration, curies/cu. meter
 
Q =  source strength, curies/second
 
~ =  average wind speed, meters/second
 
~i =  wind speed for condition i, meters/second y=  horizontal dispersion parameter, meters z=  vertical dispersion parameter, meters zi= vertical dispersion parameter for condition i, meters c =  building shape factor (selected as 0.5)
 
A =  cross-sectional area of building normal to wind (1750 sq meters)
=  sector size, radians x =  distance from source, meters
 
f =  fraction of time wind blows in sector
 
Fi=  fraction of time condition i exists
 
Rev. 10  7/92 0-2 hours                      cA) +  (
1 = Q        zy  2-12 hours x  /2  1 = Q      z  12 hours - 31 days              x  /2  F  f = Q        ziii TABLE 14F-6  METEOROLOGICAL CONDITIONS
 
Time period Condition
 
0-2 hours                          Stability category, Pasquill F;
 
Wind speed, 2 meters/sec;
 
Wind direction, unvarying.
 
2-12 hours                          Stability category, Pasquill F;
 
Wind speed, 2 meters/sec;
 
Wind direction,10 degree sector.
 
12 hours - 31 days                  Wind direction, 22.5 degree sector;
 
Wind blowing in this sector 25% of
 
the time with the following
 
variable conditions:
 
Stability  Wind speed
 
Fraction category meters/sec
 
                                      .25      F            2
 
                                      .50      D            5
 
                                      .25      C            4
 
Rev. 10  7/92 TABLE 14F-7  THYROID DOSE EQUATION AND SPECIFIC VALUES
_    x 1    DCF            Dose ( in rem)  = t BLA    Q    DF     
 
t = time period, hours
 
B = breathing rate, cu. meters/hour
 
L = reactor building leak rate, per second
 
_
A = average inventory of equivalent I-131 available for leakage
 
assuming no filter unit cleanup during the period, curies
 
x    = atmospheric dilution multiplier, seconds/cu. meter Q
 
DF = iodine decontamination factor for the period; that is, the ratio
 
of iodine without cleanup to iodine with cleanup
 
DCF = dose conversion factor for I-131, rem/curie
 
0-2 hours 2-12 hours 12 hours - 31 days t                    2                  10                732
 
B                    1.25                1.00              .834
 
L                    2.9 x 10
-8          2.9 x 10
-8        2.9 x 10
-8 _
A                    26.17 x 10 6        23.04 x 10 6        5.24 x 10 6
DF (2 units)        4.68                100                100       
 
DCF                  1.48 x 10 6          1.48 x 10 6        1.48 x 10 6    x                    Refer to tabulation given in  paragraph "Atmospheric Q                    Dispersion Model".
 
Rev. 10  7/92 APPENDIX 14G  HISTORICAL DISCUSSION OF CONTAINMENT PRESSURE TRANSIENT MARGINS ASSOCIATED WITH CONTAINMENT STRUCTURAL PRESSURE OF 59 PSIG
 
INTRODUCTION
 
This appendix contains the original FSAR discussion of the containment design
 
pressure margins associated with the original containment structural
 
capability pressure of 59 psig. Since the original containment structural
 
capability pressure of 59 psig has been replaced with the licensed design
 
basis pressure (55 psig) approved by the Atomic Energy Commission (AEC)
 
during the operating license stage, this discussion is of historical
 
importance only and does not apply to the current licensed containment design
 
pressure or to the basis for calculating the minimum required prestress
 
forces for the containment post-tensioning system. Refer to the engineering
 
evaluation contained in Reference 1.
 
BACKGROUND
 
The licensed containment design basis pressure of 55 psig was established
 
during the very early stages of plant licensing and has carried through to
 
current licensing documents. The PSAR and FSAR indicated that a 55 psig
 
reference containment design pressure was conservatively established for the
 
design basis (29-inch double-ended pipe break) loss-of-coolant accident
 
(LOCA), based on a 49.9 psig calculated peak pressure plus a 10% safety
 
margin; and the structural proof test was conducted at 115% design pressure
 
to check structural integrity. Refer to PSAR Sections 5.4.1.a and 12.2.3
 
(Reference 2), and to original (1970) FSAR Section 5.1.1, (Reference 3).
 
Other LOCA study cases, assuming partial safeguards availability, were also
 
considered. These study cases did not constitute licensed design basis
 
accident scenarios, but rather provided an indication of potential
 
containment performance requirements beyond-the-licensing-basis for purposes
 
of establishing conservative design margins for the containment structures.
 
14G-1 Rev. 11  11/93 These scenarios were developed in response to Atomic Energy Commission (AEC) questions, and to address uncertainties as to the availability of primary
 
system accumulators. As a result, some of these other cases assumed partial
 
safeguards operation with no core cooling, which were conditions that are
 
beyond the required postulation of a single active or passive failure. Refer
 
to PSAR Supplement 2, Questions 1.0 and 3.0 (Reference 4). For instance, the
 
AEC requested that a "no-core-cooling" case be considered, in which partial
 
safeguards equipment, operating on diesel power, introduced all the safety
 
injection water directly into the sump. This case resulted in a maximum
 
pressure of 58.5 psig. However, the value of 55 psig came about as the
 
result of the design basis analysis which assumed that partial safeguards
 
equipment, operating on diesel power, provided core cooling by having 2/3 of
 
the safety injection water flow paths reach the core.
 
To accommodate these hypothetical, beyond-the-licensing-basis scenarios, the
 
containment structure was designed with additional margins to withstand a
 
pressure of 59 psig; however, the licensed design basis LOCA analysis
 
calculated peak pressure was 49.9 psig, and "55 psig [was] considered as
 
nominal structural design pressure, thus allowing a margin of 10% over the
 
calculated peak accident pressure."  Refer to original 1970 FSAR, Section
 
5.1.1 - Reference 3).
 
CONTAINMENT MARGIN EVALUATIONS
 
Evaluation of the capability of the containment and associated cooling
 
systems to absorb energy additions without exceeding the containment design
 
pressure requires consideration of two periods of time following a postulated
 
large area rupture of the reactor coolant system.
 
The first period is the blowdown phase. Since blowdown occurs too rapidly
 
for the containment cooling systems to be activated, there must be sufficient
 
energy absorption capability in the free volume of the containment (with due
 
credit for energy absorption in the containment structures) to limit the
 
resulting pressure below design.
 
14G-2 Rev. 11  11/93 The second period is the post-blowdown period where the containment cooling systems must be able to absorb any postulated post-blowdown energy additions
 
and continue to limit the containment pressure below design.
 
Margin - Blowdown Peak to Design Pressure
 
Point A in Figure 14G-1 corresponds to the internal energy at the end of a DE
 
break blowdown, 195 x 10 6 Btu. In order for the pressure to increase to design pressure (59 psig) the internal energy must be increased to 231 x 10 6 Btu (Point B). The allowed energy addition is therefore 36 x 10 6 Btu. Since energy transferred to the containment from the core is in the form of steam
 
the total transferred core energy corresponding to allowed energy addition is
 
as follows:
 
h fg                          921.9 Qcore  =          Qallowed  =  36 x 10 6 x          =  28.4 x 10 6 Btu                  h g                          1177.6
 
This allowable value of energy which could be transferred from the core to the containment without increasing the transient containment pressure to
 
design pressure can be compared to the energy stored in the reactor vessel
 
and transferred to the steam generator during blowdown for the double ended
 
break. The thick metal of the reactor vessel was not considered since a
 
negligible amount of this energy can be transferred in the short blowdown
 
time.
 
Stored in the core                  15.0 x 10 6  Btu Core internals Metal                0.3 x 10 6  Btu Transferred to Steam Generators      1.4 x 10 6  Btu                                   
 
16.7 x 10 6  Btu Thus, the containment has the capability to limit containment pressure below
 
design even if all of the available energy sources were transferred to the
 
containment at the end of blowdown. This would also include no credit for
 
14G-3 Rev. 11  11/93 energy absorption in the steam generator. For this to occur an extremely high core to coolant heat transfer coefficient is necessary. This would
 
result in the core and internals being completely subcooled and limit the
 
potential for release of fission products.
 
Additional Energy Added as Superheat
 
Line A to C on Figure 14G-1 represents a constant mass line extended into the
 
superheated region. Comparison of the energy addition allowable for the
 
superheated case relative to the saturated case shows a lesser ability of the
 
containment to absorb an equivalent amount of energy as superheat. An
 
addition of 8.5 x 10 6 Btu of energy after blowdown would cause the containment pressure to increase to design. The recombination of hydrogen
 
and oxygen from 9.6% Zr-H 2O reaction completed before the end of blowdown would be required to generate 8.5 x 10 6 Btu's of energy. For the case analyzed, the core was assumed to be in a subcooled state, and no Zr-H 2O reaction would be possible. In order for Zr-H 2O reaction to occur before the end of blowdown all of the stored initial energy must remain in the core. If
 
this occurred a blowdown peak containment pressure of only 44.2 psig would be
 
reached instead of 49.9 psig in the case analyzed. Lines D and E on Figure
 
14G-1 represent the superheat energy addition required to increase the
 
pressure to the design pressure and this corresponds to the hydrogen oxygen
 
recombination energy from a 15.8% Zr-H 2O reaction.
 
It is, therefore, concluded that the containment has the capability to absorb
 
the maximum energy addition from any loss-of-coolant accident without
 
reliance on the containment cooling system. In addition, a substantial
 
margin exists for energy additions from arbitrary energy sources much greater
 
than any possible.
 
Margin - Post Blowdown Energy Additions
 
The Safety Injection System is designed to rapidly cool the core and stop
 
significant addition of mass and energy to the containment.
 
14G-4 Rev. 11  11/93 However, the following cases are presented to demonstrate the capability of the containment to withstand post accident energy additions without credit
 
for core cooling.
 
Case 1 : Blowdown from a large area rupture with continued addition of the core residual energy and hot metal energy to the containment as
 
steam.
Case 2 : Same as Case I but with the energy addition from a maximum Zirconium - water reaction.
 
Figure 14G-2 presents the containment pressure transient for Case 1. For
 
this case the decay heat generated for a 2300 MWt core operated for an
 
infinite time is conservatively assumed. This decay heat is added to the
 
containment in the form of steam by the boiling off of water in the reactor
 
vessel. For this case injection water merely serves as a mechanism to
 
transfer the residual energy to the containment as it is produced. Injection
 
water is in effect throttled at the required rate.
 
In addition, all the stored energy in the core and internals which is
 
calculated to remain at the end of blow down is added in the same way during
 
the time interval between 12.7 and 36.5 seconds (corresponds to accumulator
 
injection time). Also all the sensible heat of the reactor vessel is added
 
as steam exponentially over 2000 seconds time interval.
 
The containment cooling system capability assumed in the analysis was one of
 
two available containment spray pumps and two of three available emergency
 
containment coolers. This is the minimum equipment available considering the
 
single failure criterion in the emergency power system, the containment spray
 
system and the fan cooler system.
 
The containment heat removal capability started at 60 seconds exceeds the
 
energy addition rate and the pressure does not exceed the initial blowdown
 
value. An extended depressurization time results due to the increased heat
 
load on the containment coolers.
 
14G-5 Rev. 11  11/93 It should be emphasized that this situation is highly unrealistic in that continued addition of steam to the containment after blowdown could not
 
occur. The accumulator and Safety Injection System acts to rapidly reflood
 
and cool the core.
 
Figure 14G-3 presents the containment pressure transient for Case 2. To
 
realistically account for the energy necessary to cause a metal-water
 
reaction, sufficient energy must be stored in the core. Storing the energy
 
in the core rather than transferring it to the coolant causes a decrease in
 
the blowdown peak.
 
The reaction was calculated using the parabolic rate equation developed by
 
Baker and assuming that the clad continues to react until zirconium oxide
 
melting temperature of 4800 oF is reached. An additional 10% reaction of the unreacted clad is assumed when the oxide melting temperature is reached. A
 
total reaction of 32.3% has occurred after 1000 seconds. If the reactions
 
were to be steam limited, they could result in a higher total reaction but at
 
a much later time. The reaction provided by the parabolic rate equation
 
therefore, imposes the greatest load on the containment cooling system.
 
As in Case 2, the residual heat and sensible heat is added to the containment
 
as steam. The energy from the Zr-H 2O reaction is added to the containment as it is produced. The hydrogen was assumed to burn as it entered the
 
containment from the break.
 
The blowdown peak was reduced to 44 psig and a peak pressure of 57.7 psig was
 
reached at 600 seconds. At this time the heat removal capability of the
 
containment cooling system assumed to be operating (one containment spray
 
pump and two fan coolers) exceeded the energy addition from all sources.
 
For comparison the containment pressure transients for Cases 1, 2 and the
 
double ended blowdown are replotted in Figure 14G-4. It is concluded that 
 
operation of the minimum containment cooling system equipment provides the
 
capability of limiting the containment pressure below its design pressure
 
with the addition of all available energy sources and without credit for the
 
cooling effect from the safety injection system.
 
14G-6 Rev. 11  11/93 DISCUSSION OF ENERGY SOURCES USED IN CASES 1 AND 2
 
The following is a summary of the energy sources and the containment heat
 
removal capacities used in the containment capability study. Figure 14G-5 
 
presents the rate of energy addition from core decay heat, Zr-H 2O reaction energy, and the hydrogen-oxygen recombination energy. The heat removal
 
capability for the partial containment cooling (one spray pump and two fan
 
coolers) is also presented. These heat removal values are for operation with
 
the containment at design pressure.
 
The integrated heat additions and heat removals for Cases 1 and 2 are plotted
 
in Figures 14G-6 and 14G-7, respectively. These curves are presented in a
 
manner that demonstrates the capability of the containment and the cooling
 
systems to absorb energy. The integrated heat removal capacity is started at
 
the internal energy corresponding to design pressure, while the integrated
 
heat additions begin from the internal energy calculated at the end of
 
blowdown for each case. The upper line on each curve is the containment
 
structures and containment cooling systems capability to absorb energy
 
additions without exceeding design pressure. The lower curve for each are
 
the energy addition curves, and since these energy additions are the maximum
 
possible with no credit for core cooling, there is more than adequate
 
capability to absorb arbitrary additions.
 
The curves in Figures 14G-8 and 14G-9 present the individual contribution of
 
the heat removal and heat addition source, respectively.
 
14G-7 Rev. 11  11/93 REFERENCES
: 1. Engineering Evaluation JPN-PTN-SENP-93-008,"No Significant Hazards Evaluation Related to Containment Design Pressure Technical 
 
Specification and UFSAR Changes," Revision 0, dated April 23, 1993.
: 2. Turkey Point Units 3 and 4 Preliminary Safety Analysis Report (PSAR),
Sections 5.4.1.a and 12.2.3, submitted by Application dated March 22,
 
1966. 
: 3. Turkey Point Units 3 and 4 (original) Final Safety Analysis Report (FSAR), Section 5.1.1, "Containment Structure Design Bases," Revision
 
4, dated August 12, 1970.
 
14G-8 Revised 05/14/2005}}

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