ML17352B005

Nuclear Physics Methodology for Reload Design of Turkey Point & St Lucie Nuclear Plants.

ML17352B005
Person / Time
Site:	Saint Lucie, Turkey Point
Issue date:	01/31/1995
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17352B002	List: ML17352B000 ML17352B004 ML17352B005
References
NF-TR-95-01, NF-TR-95-1, NUDOCS 9501270179
Download: ML17352B005 (170)

Text

NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN OF TURKEY POINT & ST. LUCRE NUCLEAR PLANTS NF-TR-95-01 0'2QG3'ARY 1995 FLORIDA POWER Ec LIGHT COMPANY NUCLEAR FUEL SECTION Z(JMO BEACH, FLOR1DA g5pi27pi q +5pii7 pgR ADOCK pgppp2+p pg P

ABSTRACT This document describes the nuclear design methodology employed by Florida Power & Light Company (FPL) to analyze the core design characteristics necessary to support a fuel reload for Turkey Point Units 3 and 4 and St. Lucie Units 1 and 2. This methodology, including all computer programs used, was obtained from Westinghouse Electric Corporation. Calculations were performed using this methodology and the results compared to operating data from Turkey Point and St Lucie. The quality of the comparisons demonstrates FPL's ability to perform reload core design for FPL's nuclear units.

0 TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

AND CONCLUSIONS 1.1 OB JECTIVE

1.2 BACKGROUND

.3 SCOPE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.4 CONCLUSION

S 2.0 PHYSICS METHODOLOGY ..... .. ~.......... ~........ ~.... 5

~ ~ ~

2.1 CROSS SECTION LIBRARY ~ .. ~..... ~ ~ 5 2.2 LATTICE MODELING IN PHOENIX-P ~ ~ 6 2.2.1 FUEL CELL MODEL ... ~.......... 7 2.2.2 DISCRETE ABSORBER MODEL ......... ~ ~ 7 2.2.3 STRUCTURAL CELL MODEL ~ ~ 9 2.3 BAFFLE-REFLECTOR MODELING 9

~ ~

2.4 THREE-DIMENSIONAL NODAL MODEL ~ ~ 9 2.5 ONE-DIMENSIONAL DIFFUSION THEORY MODEL 10 3.0 PHYSICS MODEL APPLICATIONS ....... ~..................

~ - 11 3.1 CORE POWER DISTRIBUTIONS AT STEADY STATE C ONDITIONS 11 3.1.1 POWER DISTRIBUTIONS 11 3.1.2 POWER PEAKING 12 3.1.3 FUEL DEPLETION 12 3.2 AXIAL POWER DISTRIBUTION CONTROL LIMITS .......... 13 3.3 CORE REACTIVITY PARAMETERS 14 3.3.1 MODERATOR TEMPERATURE COEFFICIENT 15 3.3.2 DOPPLER COEFFICIENTS 15 3.3.3 TOTAL POWER COEFFICIENT 16 3.3.4 ISOTHERMAL TEMPERATURE COEFFICIENT 17 3.3.5 BORON REACTIVITY COEFFICIENT ................ 17 18 3.3.6 XENON AND SAMARIUM WORTH 3.3.7 CONTROL ROD WORTH 18.

3.3.8 NEUTRON KINETICS PARAMETERS 19 3.4 CORE PHYSICS PARAMETERS FOR TRANSIENT ANALYSIS NPUT ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 20 4.0 PHYSICS MODEL VERIFICATION TURKEY POINT UNITS ~ ~ 21 4.1 CYCLE DESCRIPTIONS ~ ~ 21 4.2 ZERO POWER PHYSICS TESTS ~ ~ 23 4.2.1 CRITICAL BORON CONCENTRATIONS ~ ~ 24 4.2.2 TEMPERATURE COEFFICIENTS ~ ~ 24 4.2.3 CONTROL ROD WORTH ~ ~ 24 4.2.4 DIFFERENTIAL BORON WORTH ~ ~ 25

TABLE OF CONTENTS (CONTINUED)

SECTION PAGE 4.3 POWER OPERATION o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 26 4.3.1 BORON LETDOWN CURVES 26 4.3.2 POWER PEAKING FACTORS 27 4.3.3 RADIAL POWER DISTRIBUTIONS 28 4.3.4 AXIAL POWER DISTRIBUTIONS AND AXIAL OFFSETS . 28 4.4 S UMMARY ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 29 5.0 PHYSICS MODEL VERIFICATION ST. LUCIE UNITS 73 5.1 CYCLE DESCRIPTION 73 5.2 ZERO POWER PHYSICS TESTS 74 5.2.1 CRITICAL BORON CONCENTRATION 75 5.2.2 MODERATOR TEMPERATURE COEFFICIENT 75 5.2.3 CONTROL ROD WORTH 75 5.2.4 DIFFERENTIAL BORON WORTH 75 5.3 POWER OPERATION 5.3.1 BORON LETDOWN CURVES 5.3.2 AXIAL POWER DISTRIBUTIONS

................ .. ~

76 76 76 5..4

SUMMARY

76 6..0 REFERENCES 102 APPENDIX A WESTINGHOUSE COMPUTER CODES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 04 A.1 FIG HTH ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 04 A.2 PHOENIX-P ....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 05 1 06 A.3 ANC ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

A.4 APOLLO ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 07 3.3.

LtST OF TABLES TABLE PAGE 4.1-1 Turkey Point Unit 4 Fuel Specification ....................... 3p 4.2-1 Turkey Point Unit 4 HZP Physics Test R eview Criteria ....... ~ . ~...

~ ~ ~ ~ ~ . ~ ~ .. ~ ~ . ~ ~............. 31 4.2-2 Turkey Point Unit 4 Critical Boron Concentration Comparison Between Measurement and Predictions for Cycles 12, 13, and 14 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 32 4.2-3 Turkey Point Unit 4 Moderator and Isothermal Temperature Coefficient Comparison Between Measurement and Prediction for Cycles 12, 13, and 14 ...... ~ ~ ~ ~ ~ 33 4.2-4 Turkey Point Unit 4 Control Rod Worth Comparison Between Measurement and Prediction for Cycles 12, 13, and 14 ...... 34 e

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

4.2-5 Turkey Point Unit 4 HZP Differential Boron Worth Comparison Between Measurement and Prediction for Cycles 12, 13, and 14 ................ ...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 35 4.3-1 Turkey Point Unit 4 Cycles 12, 13, and 14 Boron Letdown Comparison Between Measurement and Prediction............................... 36 4.3-2 Turkey Point Unit 4 Cycles 12, 13, and 14 Power Peaking Factor (F~) Comparison Between Measurement and Prediction ....................... 37 4.3-3 Turkey Point Unit 4 Cycles 12, 13, and 14 Power Peaking Factor (F~) Comparison Between Measurement and Prediction ....................... 3S 4.3-4 Turkey Point Unit 4 Cycles 12, 13, and 14 Axial Offset Comparison Between Measurement and Prediction .......................... 39

e LIST OF TABLES (CONTINUED)

TABLE PAGE 6.2-1 St. Lucie Unit 1 Critical Boron Concentration Comparison Between Measurement and Predictions for Cycles Oy 1 1 and l2 y s ~ ~ ~ ~ 78 5.2-2 St. Lucie Unit 1 Moderator Temperature Coefficient Comparison Between Measurement and Prediction for Cycles 10, 11; and 12 ......................... 79 6.2-3 St. Lucie Unit 1 Control Rod Worth Comparison Between Measurement and Prediction for Cycles 10, 11, and 12 ......................... 80 6.2% St. Lucie Unit 1 HZP Differential Boron Worth Comparison Between Measurement and Prediction for Cycles 1 0, 11, and 12 o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 81 5.3-1 St. Lucie Unit 1 Cycles 10, 11, and 12 Boron Letdown Comparison Between Measurement and Prediction ............................... 82

LIST OF FIGURES FIGURE PAGE 4.1-1 Turkey Point Unit 4 Cycle 12 Core Loading d Pattern ....................................... 40 4.1-2 Turkey Point Unit 4 Cycle 13 Core Loading d Pattern ......................................

I 41 4.'I-3 Turkey Point Unit 4 Cycle 14 Core Loading d Pattern ....................................... 42 4.2-1 Turkey Point Unit 4 Cycle 12 Measured versus Predicted Control Bank C Integral Rod W0 rth ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 43 4.2-2 Turkey Point Unit 4 Cycle 13 Measured versus Predicted Control Bank A Integral Rod Worth o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

4.2-3 Turkey Point Unit 4 Cycle 14 Measured versus Predicted Shutdown Bank B Integral Rod Worth e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 46 4.3-1 Turkey Point Unit 4 Cycle 12 Boron Letdown Comparison Between Measurement and Prediction ....................'.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 46 4.3-2 Turkey Point Unit 4 Cycle 13 Boron Letdown Comparison Between Measurement a nd Prediction ................................... 47 4.3-3 Turkey Point Unit 4 Cycle 14 Boron Letdown Comparison Between Measurement and Prediction .......... ~................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 48 4.3-4 Turkey Point Unit 4 Cycle 12 F~

Comparison Between INCORE and ANC ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 49 4.3-5 Turkey Point Unit 4 Cycle 13 F~

Comparison Between INCORE and ANC e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 50 4.3-6 Turkey Point Unit 4 Cycle 14 F~ '

Comparison Between INCORE and ANC ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 61

LIST OF FIGURES (CONTINUED)

FIGURE PAGE 4.3-7 Turkey Point Unit 4 Cycle 12 F~ Comparison Between INCORE and ANC ............. ~...............

~ 52 4.3-8 Turkey Point Unit 4 Cycle 13 F~ Comparison Between INCORE and ANC .. ~ ~.......................... 53 4.3-9 Turkey Point Unit 4 Cycle 14 F~ Comparison Between INCORE and ANC ......... ~................... 54 4.3-10 Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE and ANC -2320 MWD/MTU ................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 55 4.3-11 Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE and ANC - 6975 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 56 4.3-12 Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE and ANC - 118'12 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 57 4.3-13 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and ANC - 2440 MWD/MTU ................. ~................ 58 4.3-14 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and ANC - 6678 MWD/MTU s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 59 4.3-15 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and ANC -12316 MWD/MTU ............................... 60 4.3-1 6 Turkey Point Unit 4 Cycle 14 Radial Power Distribution Comparison Between INCORE and ANC -600 MWD/MTU ...... ~................. ~ ~ ~ ~ ~ ~ 61 4.3-17 Turkey Point Unit 4 Cycle 14 Radial Power Distribution Comparison Between INCORE and ANC - 6836 MWD/MTU .................................. 62 VI.

0 0

LIST OF FIGURES (CONTINUED)

FIGURE PAGE Turkey Point Unit 4 Cycle 14 Radial Power Distribution Comparison Between INCORE and ANC - 10704 MWD/MTU ... ~..... .. ~..... ..

~ ~ ~ ~ ~ ..... 63 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 7620 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ 64 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 9458 MWD/INTU ...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ 65 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 11812 MWD/MTU ........... ~.......... ~........ 66 Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE and ANC - 2440 MWD/MTU 67 Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE and ANC - 6678 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 68 Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE and ANC - 12316 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 69 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Gomparison Between INGORE and ANC - 600 MWD/MTU ................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 70 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE and ANC - 6836 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 71 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE and ANC - 10704 MWD/MTU ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 72

LIST OF FIGURES (CONTINUED)

FIGURE PAGE St. Lucie Unit 1 Cycle 10 Core Loading d Pattern e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 84 St. Lucie Unit I Cycle 11 Core Loading Pattern .......... ~......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 85 St. Lucie Unit 1 Cycle 12 Core Loading d Pattern ............... ~ .. ~..... ~........ ~..... 86 St. Lucie Unit 1 Cycle 10 Measured versus Predicted Reference Bank Integral Rod Worth e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 87 St. Lucie Unit 1 Cycle 11 Measured versus Predicted Reference Bank Integral Rod Worth ~ ~ ~ ~ 's ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 88 St. Lucie Unit 1 Cycle 12 Measured versus Predicted Reference Bank Integral Rod Worth 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 89 St. Lucie Unit 1 Cycle 10 Boron Letdown Comparison Between Measurement a nd Prediction .. ~....................... ~............. 90 St. Lucie Unit 1 Cycle 11 Boron Letdown Comparison Between Measurement J

a nd Prediction ........... ~...... ~..... ~ ~ . ~ ~....... ~... 91 St. Lucie Unit 1 Cycle 12 Boron Letdown Comparison Between Measurement a nd Prediction ............................. ~..... ~.... 92 St. Lucie Unit 1 Cycle 10 Axial Power Distribution Comparison Between INPAX and ANC -372 MWD/MTU ................................... 93 St. Lucie Unit 1 Cycle 10 Axial Power Distribution Comparison Between INPAX and ANC -6904 MWD/MTU ................................. 94

e FIGURE LIST OF FIGURES (CONTINUED)

PAGE 5.3-6 St. Lucie Unit 1 Cycle 10 Axial Power Distribution Comparison Between INPAX and ANC - 15718 MWD/MTU . ~ ~...... ~..... ~ ~ .'............... 96 5.3-7 St. Lucie Unit 1 Cycle 11 Axial Power Distribution Comparison Between INPAX and ANC - 186 MWD/MTU .. . ~ ~ ~ ~ ~ ~ .. ~ ~ ~ ~ ~ ~.... ~............. 96 6.3-8 St. Lucie Unit 1 Cycle 11 Axial Power Distribution Comparison Between INPAX and ANC -6721 MWD/MTU .................................. 97 5.3-9 St. Lucie Unit 1 Cycle 11 Axial Power Distribution Comparison Between INPAX and ANC - 12118 MWD/MTU . ~... .. ~ ~............... ~........ 98 Cycle 12 Axial Power

~ ~

5.3-10

~ St. Lucie Unit 1

~

Distribution Comparison Between INPAX and

~

ANC - 625 INWD/MTU ~ ~ ~ ~.... -

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ..

~ ~ ~ ~ ~ ~...........

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 99 5.3-11 St. Lucie Unit 1 Cycle 12 Axial Power Distribution Comparison Between INPAX and ANC - 6620 MWD/MTU .............................. 100 5.3-12 St. Lucie Unit 1 Cycle 12 Axial Power Distribution Comparison Between INPAX and ANC - 13320 MWD/MTU ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~.... ~.......

~ 101

1.0 INTRODUCTION

AND CONCLUSIONS This report describes the physics methods used by Florida Power & Light Company (FPL) to analyze the core characteristics for our four Pressurized Water Reactors (PWR). It includes a summary description of the Westinghouse computer programs and methodology as applied by FPL to model the Turkey Point and St. Lucie Nuclear Power Station cores. Comparisons between predictions and operating data are provided as a demonstration of FPL's qualifications to use the Westinghouse methodology to perform reload design calculations for the Turkey Point and St. Lucie nuclear units.

1.1 OBJECTlVE The objective of this report is to demonstrate FPL's competence to perform reload design analyses for our four nuclear power plants. To this end, extensive design calculations have been performed for Cycles 12, 13 and 14 of Turkey Point Unit 4 and the results are compared to actual plant operating data herein. Unit 4 was chosen for its wide variety of assembly and poison types, its transition to axial blanketed fuel, its large number of reinserted fuel assemblies, vessel flux reduction features (e.g., Hafnium inserts at the periphery), and its low leakage fuel management. Design calculations have also been performed for St. Lucie Unit 1, Cycles 10, 11, and 12 and a limited set of results have been compared to actual plant operating data. Unit 1 was chosen for comparison because of its use of Gadolinium burnable poisons, axial blankets and vessel flux reduction features in the core design.

1.2 BACKGROUND

FPL has determined that in-house capability to design reload cores for our units would provide the following benefits:

~ Improved control over the design, yielding more control of the decision process,

~ Improved optimization of the design, allowing better fuel utilization and economics, and

~ A better understanding of the design, leading to more comprehensive evaluations of core safety.

Various physics methodologies were reviewed to determine which best satisfied FPL's needs. FPL decided to use the Westinghouse approach, one of our NSSS vendors and present fuel supplier for Turkey Point. The Westinghouse methodology provided four important advantages:

A physics methodology which included extensive written procedures (METCOM) which documented in step by step fashion core design calculational practices.

A training program which provided hands on experience by utilizing METCOM and performing actual calculations on the computer workstation to ensure that the FPL engineers understood the Westinghouse methodology.

A physics methodology previously reviewed and generically approved by the NRC for all PWR applications, and An agreed upon process under which FPL engineers would perform the calculations related to the reload physics analysis process independently of Westinghouse for Turkey Point Unit 3, Cycle 14 with Westinghouse providing Quality Assurance of all calculations.

The purpose of this effort was to demonstrate the ability of FPL to perform the required analysis and to use lessons learned to improve the implementation prior to operating independently from Westinghouse.

Implementation of the above decision required entering into a technology exchange agreement with Westinghouse Electric Corporation. This agreement also provides FPL the ability to upgrade codes and methods to be consistent with any revisions developed by Westinghouse. The relevant computer programs and associated methodology of Westinghouse's Commercial Nuclear Fuel Division have been transferred to FPL. A description of the applicable physics models is provided in the next chapter while the computer programs themselves are discussed in Appendix A. The computer programs and procedures (METCOM) are incorporated into the FPL Quality Assurance Program.

Training of FPL personnel in the Westinghouse methods was performed during 1993 utilizing the Nuclear Core Design Training Center approach provided by Westinghouse. FPL individuals were trained in areas ranging from Loading Pattern Scoping, Cross-Section Development, Loading Pattern Generation, Safety Analysis Models and Analysis, Nuclear Design IYlodels and Analysis, to the development of Core Follow Analysis. In all, 14 FPL individuals were trained by Westinghouse in these areas representing well over 5500 manhours of training. Ongoing training by Westinghouse has also been provided, a recent two day training session reviewed modifications to METCOM and provided technical interactions between FPL personnel and Westinghouse designers.

SCOPE FPL has performed in-house core design calculations and core follow analysis for Turkey Point for many cycles. Core follow results obtained during Unit 4 Cycles 12, 13, and 14 provide ample data with which to compare predicted power distributions, predicted boron letdown curves, and fuel depletion calculations. In addition, the startup physics measurements conducted during the startup of each cycle provide an

0 additional source of valid data for evaluating the physics model predictions of critical boron concentrations, control rod worth, and temperature coefficients. Detailed comparisons of the predictions and measurements are presented in Section 4.

FPL has also performed in-house core design calculations and core follow analysis for the St. Lucie Units. Comparisons between measurements and predictions for St. Lucie Unit 1 Cycles 10, 11, and 12 are presented in Section 5 using Westinghouse methodology.

All methods used to generate the results detailed in this report (computer programs and model development) are standard licensed methods used by the Westinghouse Commercial Nuclear Fuel Division. Therefore, the calculational uncertainties (e.g., see Reference 1) associated with the methods are unchanged and do not require re-quantification. In addition, the methods utilized to process measured data (e.g., see Reference 2) for Turkey Point are also standard to Westinghouse such that measurement uncertainties do not require re-determination by FPL.

1.4 CONCLUSION

S This report describes the use of the Westinghouse methodology as applied by FPL to model the Turkey Point Unit 4 and St. Lucie Unit 1 cores.

Calculations were performed for Cycles 12, 13, and 14 for Turkey Point Unit 4 and the results were compared to actual operating data. Assemblies from Turkey Point Unit 4, Cycles 9, 10, and 11 were also modeled to establish the appropriate axial burnup distribution's. Calculations were performed for Cycles 10, 11, and 12 for St. Lucie Unit 1 as described in Section 5. The results from these comparisons demonstrate FPL's understanding of the methodology and show that FPL can apply the METCOM procedures and computer codes during the performance of future reload design analyses for FPL nuclear units.

0 2.0 PHYSICS METHODOLOGY This section describes the Westinghouse codes and methodology used by FPL to perform design calculations for reload cores. The major features associated with each model are discussed, as is the interaction between models. This methodology was also used to obtain the results presented in Section 4 and Section 6. Descriptions of the individual computer codes used are provided in Appendix A.

Lattice physics parameters for unit assemblies and baffle-refiector cross sections are calculated with PHOENIX-P (Reference 3 and 11), a two-dimensional multi-group transport theory code. Fuel and clad temperatures are generated with the FIGHTH (Reference 9 and 10) code. The three-dimensional advanced nodal code ANC (Reference 8) is used to predict reactivity, power distributions, and other relevant core characteristics. In addition, APOLLO (Reference 12), a one-e dimensional diffusion theory code is available to calculate differential control rod worth and axial power distributions for the heat flux hot channel factor (F~)

synthesis to establish operational limits. The cross section library, as well as PHOENIX-P, nodal, and diffusion theory models are discussed in the following sections.

The models described here are representative of current Westinghouse practices.

FPL's calculational capabilities are anticipated to evolve in parallel with Westinghouse's through planned implementation of the technology exchange agreement between the two corporations.

2.1 CROSS SECTION LIBRARY The PHOENIX-P computer program's nuclear cross section library contains microscopic cross section data based on a 42 energy group structure derived from ENDF/B-V files. This cross section library was designed to properly capture integral properties of the multigroup data during the 0 group collapse in order to accurately model important resonance parameters, and to provide the overall accuracy of reactivity predictions necessary for core design. In addition, this library has been developed in a manner consistent with current Westinghouse methodologies and accumulated core design experience. The development and benchmarking of the PHOENIX-P library are described in Reference 3.

For gadolinium, the cross-sections are obtained from the Criticality Safety CSRL-V 227 group ENDF/B-V library. Resonance effects are added by the NITAWL-S code using Nordheim treatment. The 227 groups are subsequently collapsed to the PHOENIX-P 42 group structure using the XSDRN-PM transport theory cell code.

2.2 LATTICE MODELING IN PHOENIX-P In PHOENIX-P, the fuel, discrete absorbers, and structural components within a single fuel assembly are represented in their exact lattice configuration. Discontinuity factors, pin factors, and homogenized two-group microscopic cross sections are generated as a function of burnup for input to ANC. For isotopes and materials represented explicitly in ANC, microscopic cross sections are generated, including xenon, samarium, soluble boron, water density, and burnable absorbers. To obtain constants for rodded assemblies, branch calculations are performed at selected burnups.

A three region cylindrical cell description for each cell within the lattice is allowed in PHOENIX-P. Principles of material preservation are employed to construct three region cell representations, since most lattice cells consist of more than three subregions. The outer region (third region) of each cell, defined by the fuel pin pitch, has a common composition in all cells in a given lattice configuration. Grids are modeled by smearing the grid material uniformly over this common outer region. Grids are only

smeared in the active fuel region. The sections following describe the various types of cell models.

2.2.1 FUEL CELL MODEL The fuel pellet outer radius defines the innermost region of a fuel rod cell.

The middle region is defined by the clad outer diameter and incorporates the pellet-clad gap. Appropriate number densities are specified for the uranium isotopes and oxygen for fresh fuel. Isotopic information for burned fuel, including decay chains, is obtained from previous depletion calculations of fresh fuel. For fuel pellets with integral fuel burnable absorber (IFBA) zirconium diboride coating, the coating material is smeared into the clad region rather than being explicitly installed as a coating on the surface of the pellet. PHOENIX-P corrects for the reactivity effect of modeling the absorber as smeared into the clad instead of on the pellet.

2.2Z DISCRETE ABSORBER MODEL A. BURNABLE ABSORBER RODS Turkey Point has used two types of discrete burnable absorber (BA) rods: Wet Annular Burnable Absorbers (WABAs) and Pyrex glass.

The cell representation for the two BA types is significantly different.

The WABA contains moderator material in the central region, while the Pyrex BA is voided in the central region. The surface area of the absorber material must be preserved in addition to the quantity of material.

Since a fast neutron can pass through the absorber region of a WABA, become thermalized in the inner region, and be absorbed, both the inner and outer surfaces of the absorber are important.

Region 1 of the cell is therefore defined as moderator material with an outer radius equivalent to the BA pellet inner radius. Region 2 is

defined as pure pellet material with an outer radius equal to the outer radius of the pellet. The inner WABA cladding, inner pellet-clad gap, outer pellet-clad gap, outer cladding, guide tube,,and sleeve materials are all smeared into the moderator region in order to preserve material quantities.

For Pyrex absorbers, the inner gap, inner clad and pellet absorber material are smeared into the first region with a radius equivalent to the pellet outer radius. Region 2 is made up of the absorber outer clad, moderator, guide tube and sleeve volumes, and materials. The small volume of moderator between the outer clad and the guide tube is modeled as if it were outside the guide tube. This is a minor approximation, since the zircaloy guide tube material is nearly transparent to neutrons.

For gadolinium, PHOENIX-P uses 42 group microscopic cross sections for the gadolinium isotopes as a function of Gd-165 and Gd-167 depletion along with lattice and other geometry specific aspects to produce appropriately weighted two group, homogenized cross-sections for ANC.

CONTROL RODS Control rod cells are modeled in a manner similar to Pyrex BA cells, except that the dimensions and material in the. pellet region are different. Resonance calculations are performed by PHOENIX-P for the Ag-In-Cd control rod material. For St. Lucie, control rods are modeled as five regions consisting of 84C absorber, clad, moderator, guide tube, and moderator.

C. HAFNIUM ABSORBERS Hafnium absorber rod cells are modeled in a manner similar to Pyrex BA cells, except that the dimensions and material in the pellet region are different. Hafnium rods decrease the power and thereby the fast fluence in core locations close to the reactor pressure vessel weld.

This reduction is required for pressurized thermal shock (PTS) considerations.

2.2.3 STRUCTURAL CELL MODELS Certain cells, known as structural cells, contain neither a strong absorber or material that is depletable. Examples of these include guide tubes, instrument tubes, water displacer rods, and stainless steel rods. These can typically be represented with three regions or less and do not require special neutronic considerations. Sleeve volume is preserved by calculating an effective guide tube thickness that equates to the total 0 sleeve volume.

2.3 BAFFLE-REFLECTOR MODELING Baffle-refiector cross sections are generated by performing a one-dimensional slab calculation with PHOENIX-P. Such a model is developed by using a series of fuel cells approximating two fuel assemblies,. the assembly/baffie gap, baffle, refiector, core barrel, thermal pad (on the fiats),

and moderator. A homogenized set of cross sections for ANC is obtained, Iepresenting the spectrum variations existing between the fuel assemblies, baffle, and reflector.

2.4 THREE-DIMENSIONAL NODAL MODEL Homogenized cross sections, discontinuity factors, and pin factors are generated on a cycle specific basis using PHOENIX-P depletion calculations. These parameters are then used to model the three-dimensional core in ANC. A fuel assembly consists of four radial nodes.

0 In order to obtain an accurate pin power recovery solution, the burnup gradient within each node is represented in ANC. A burnup gradient algorithm matches nodal corner and surface average burnups.

Explicit representations of axially heterogeneous features such as axial blankets and burnable absorbers are made using the variable axial mesh capability in ANC. Typically, 24 axial mesh intervals produce accurate axial power distributions. To account for spectrum effects induced by variable length burnable absorbers and fuel burnup gradients, axial zoning of the burnup dependent cross sections is employed. Burnable absorber history effects are also taken into account by using appropriate sets of fuel cross sections.

The three-dimensional ANC calculational results can be used to predict peaking factors, critical boron concentrations, core power distributions, control rod worth, and reactivity coefficients. This model can also be collapsed to two dimensions for those calculations (e.g., determination of the highest worth stuck rod) where a three-dimensional representation is not required.

2.5 ONE-DIMENSIONAL DIFFUSION THEORY MODEL A three-dimensional ANC model can be collapsed radially to generate a one-dimensional APOLLO model. The cross sections are flux and volume weighted, and a burnup and elevation dependent radial buckling search is performed to normalize the APOLLO model to ANC. The one-dimensional diffusion theory model is used for calculations where additional detail is desired in the axial direction. To this end, the axial mesh is redefined to comprise 40 or more axial intervals. APOLLO can be used to generate integral and differential control rod worth curves, determine control rod insertion limits, and analyze axial power distributions in order to establish limits on axial offset during power operation.

3.0 PHYSICS MODEL APPLICATIONS The physics methodology discussed in Section 2 was developed in order to provide reliable analytical predictions in the following four major areas:

Core power distributions at steady state conditions, Axial power distribution control limits, Core reactivity parameters, and Core physics parameters for transient analysis input Often more than one model may be used to perform a specific analysis. The preferred model depends upon a number of considerations including the degree of accuracy desired and the specific applications.

e 3.1 CORE POWER DISTRIBUTIONS AT STEADY STATE CONDITIONS The prediction of steady-state core power distributions is fundamental to the design, analysis, and surveillance of nuclear reactor cores. Accurate prediction of core power distributions leads to confidence in developing and optimizing core loading patterns, ensuring compliance with Technical Specification limits, and determining fuel assembly burnups and isotopic inventories.

3.4.1 POWER DISTRIBUTIONS Global core power distributions are obtained as a function of burnup from three-dimensional ANC depletion calculations. Calculations are also performed at selected burnups for various power levels and control rod configurations. Peak rod powers and hot channel factors are generated by pin power reconstruction within ANC using rod-by-rod power distributions from single assembly two-dimensional PHOENIX-P fine mesh spectrum calculations.

3.1.2 POWER PEAKING

~ ~

Local power peaking is monitored to ensure that the peak pellet power and the total energy content within each coolant channel remain within Technical Specification and/or fuel design limits. The factors used to measure local power peaking include:

~ the heat flux hot channel factor, F~, defined as the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod heat flux,

~ the nuclear enthalpy rise hot channel factor, F~, defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power, and

~ the planar radial power peaking factor, F~(Z), defined as the ratio of the peak power density to the average power density in the horizontal plane at elevation z.

For steady state conditions, these are obtained from three-dimensional ANC calculations using pin power reconstruction. For maneuvering and transient xenon conditions, a three-dimensional, one-dimensional, synthesis technique (see Section 3.2) may be used.

3.1.3 FUEL DEPLETION Three-dimensional fuel depletion calculations are performed with ANC.

Rod-by-rod burnup distributions are obtained from the ANC depletions.

Specific fuel nuclide inventories are obtained from two-dimensional single assembly PHOENIX-P depletion calculations.

0 Q32 AXIALPOWER DISTRIBUTION CONTROL LIIHITS The axial power distribution is primarily affected by control rod position, xenon, burnup, and temperature distributions. Axial power distribution control limits are used to ensure that thermal limits are not violated during power level changes, control rod motion, and the resulting xenon redistributions. This is accomplished by maintaining the axial flux difference within acceptable boundaries. Axial flux difference, b,l, is defined as the difference between the upper and lower excore detector signals.

Axial power distribution control limits for Turkey Point are determined using Westinghouse's Relaxed Axial Offset Control (RAOC) calculational procedure (Reference 4). The RAOC calculational procedure begins by defining "provisional" hl limits which are wider than the expected LOCA limits (or, alternately, the RAOC hl limits from the previous cycle may be used if it is desired only to verify their acceptability). Xenon transient simulations are performed. with the one-dimensional APOLLO code at various burnups and for different power levels, constrained by the provisional hl limits and power dependent rod insertion limits. A library of axial xenon shapes is constructed at each burnup. Next, axial power shapes are generated with APOLLO for all possible combinations of xenon shapes, power levels, and rod insertions. These axial shapes .are synthesized with height dependent planar radial power distributions from three-dimensional ANC calculations. Imposition of the LOCA F~ limits for normal operation then defines the allowable hl limits (or verifies that the previous cycle's limits are acceptable) for the cycle. The axial power shapes corresponding to cases within the hl limits are checked against thermal hydraulic constraints from Loss of Flow Accident simulations and the peak power and DNB limits for accident conditions.

For normal operations, more restrictive h,l limits are developed if either the F~ limits or thermal hydraulic constraints are exceeded. For accident conditions, analyses are performed to verify that all design limits are met.

lf necessary, trip setpoints may be revised and/or the RAOC Cg Ijmjts tightened. Therefore, the RAOC procedure provides axial power shape information which is used to verify that all design limits are met. The RAOC dl limits are placed in the Turkey Point Core Operating Limit Report and apply during plant operation.

3.3 CORE REACTWITY PARAMETERS The core reactivity is affected by changes in the reactor which occur during operation as the result of fuel depletion and abnormal or accident conditions. Reactivity coefficients quantify the rate of reactivity change to be expected in response to changes in power, moderator or fuel temperatures, and soluble boron concentration. Reactivity defects refer to the integral of the corresponding reactivity coefficient between two reactor statepoints with all other variables remaining constant. Xenon, samarium, and control rod worth are also typically required to fully define the change in reactivity between two core configurations. In addition, neutron kinetics parameters are needed to describe the time dependent behavior of the core.

Quantification of these effects are needed: (a) to provide input to safety analyses, (b) to provide guidance to the reactor operators, and (c) to ensure compliance with Technical Specifications. Therefore, the physics models described in Section 2 are used to calculate reactivity coefficients, reactivity worth, and kinetics parameters as a function of core burnup, moderator temperature, and power level.

3.3.'I MODERATOR TEMPERATURE COEFFICIENT The moderator temperature coefficient (MTC) is defined as the change in reactivity per degree change in moderator temperature. The effect of concomitant changes in moderator and soluble boron densities are included. The MTC is sensitive to the values of the moderator density, moderator temperature, soluble boron concentration, fuel burnup, and the presence of control rods and/or burnable absorbers which reduce the required soluble boron concentration and increase the leakage of the core.

The MTC may be positive or negative depending on the magnitude of change of the individual components of this coefficient.

The MTC is calculated using the ANC core model described in Section 2.4 by varying the inlet temperature around a reference temperature. The moderator temperature coefficient is analyzed for various reactor conditions, from hot zero power (HZP) to hot full power (HFP), for various boron concentrations and control rod positions, and at various cycle burnups. The moderator temperature defect is also obtained using data from the ANC core model.

3.3.2 DOPPLER COEFFICIENTS The Doppler temperature coefficient is defined as the change in reactivity per degree change in effective fuel temperature. The effective fuel temperature accounts for the spatial variation in fuel temperature throughout the core. The Doppler power coefficient represents the corresponding change in reactivity per percent change in reactor power.

These coefficients are primarily a consequence of the Doppler broadening of U-238 and Pu-240 resonance absorption peaks which increases the effective resonance absorption cross section of the fuel with increasing fuel temperature.

0 The Doppler power coefficient is normally calculated using the ANC core model by varying the reactor power level about a reference power (which in turn varies the fuel temperature) while holding the product of the power level and the enthalpy rise constant. The FIGHTH code provides effective fuel temperatures, which account for spatial variations 'in temperature within the pellet, as a function of power level and burnup. The Doppler coefficient is analyzed at different power levels and for various cycle burnups. Doppler reactivity defects can also be obtained using the ANC model by varying the reactor power at various times in life, while holding the product of the power level and the enthalpy rise constant.

At hot zero power, the Doppler temperature coefficient may be calculated by subtracting the moderator temperature coefficient from the isothermal temperature coefficient (ITC), provided ITC is explicitly calculated '(see Section 3.3.4).

3.3.3 TOTAL POWER COEFFICIENT The total power coefficient is defined as the change in reactivity per percent change in core power level. This coefficient represents the combined effect of moderator temperature and fuel temperature changes for an associated change in core power level.

The total power coefficient is calculated using the ANC core model by varying the core power level around a reference value while allowing the inlet temperature to change in accordance with the inlet program for the plant. The power coefficient is analyzed at different power levels and at various times in core life. The power defect is also obtained using the ANC model by varying the reactor power.

3.3.4 ISOTHERMAL TEMPERATURE COEFFICIENT The isothermal temperature coefficient (ITC) is defined as the change in reactivity per uniform degree change in core temperature. The ITC is the temperature coefficient directly measured during startup physics testing.

The ITC can be calculated by summing the moderator temperature coefficient and the Doppler temperature coefficient. Alternately, the ITC may be calculated explicitly using the ANC core model by varying both the moderator temperature and the fuel temperature about a uniform reference temperature.

The isothermal temperature defect (ITD) refers to the change in reactivity between hot zero power temperatures and temperatures below hot zero power. ITDs are needed as a function of temperature and burnup for various rod patterns to establish shutdown boron concentration requirements. ITDs are calculated with the ANC model using cross sections generated with PHOENIX-P at specific temperatures between hot zero power and 68'F.

3.3.5 BORON REACTMTY COEFFICIENT The boron reactivity coefficient, also referred to as the differential boron worth, is defined as the change in reactivity per ppm change in the soluble boron concentration. The inverse of the boron reactivity coefficient is referred to as the inverse boron worth. It provides a means of determining the change in soluble boron concentration necessary to compensate for a given reactivity change. The magnitude of the boron reactivity coeffiiclent depends primarily on the soluble boron concentration, the moderator temperature, control rod insertion, and the presence of burnable absorbers.

The boron reactivity coefficient is calculated using the ANC core model by perturbing the boron concentration in both directions about a reference Boron worths are calculated Q value and computing the reactivity change.

as a function of boron concentration, power level, temperature, burnup, and control r'od configuration.

3.3.6 XENON AND SAMARlUII WORTH The fission products Xe-135 and Sm-149 possess large thermal absorption cross sections. Knowledge of the concentrations and reactivity worth of these isotopes as well as the changes which occur in response to plant maneuvers is crucial to reactor control. Since Xe-135 is also produced by iodine decay, it initially builds up and then decays following a reduction in power or shutdown. Sm-149 is a stable isotope produced by promethium decay. Following a reactor shutdown, its concentration increases. Upon restart it gradually returns to its equilibrium value.

Equilibrium xenon and samarium worth are calculated with the ANC core model at various power levels and core burnups. Changes in their worth and axial fluctuations in isotopic concentrations during transient operation are obtained using the ANC and/or APOLLO models.

3.3.7 CONTROL ROD WORTH Control rod worth refers to the reactivity difference between two control rod configurations. The total control rod worth, trip reactivity shape (i.e.,

the inserted rod worth versus rod position), integral and differential worth of individual banks, and worth of individual rod cluster control assemblies (e.g., stuck, ejected, and dropped rods) are determined as required for startup physics testing, plant operations, and input to safety analyses.

Control rod worths are analyzed for all normal and many abnormal control rod configurations as a function of burnu, power level, and moderator temperature. Total rod worth and the integral worth of individual rod banks and rod clusters are calculated using the ANC core model.

Differential rod worths are obtained with the ANC and/or APOLLO models.

3.3.8 NEUTRON KINETICS PARAMETERS Neutron kinetics parameters, which include delayed neutron fractions, decay constants, and the prompt neutron lifetime, are required as input to the plant reactivity computer and to various safety analyses. These parameters are also input to the Inhour equation to generate core reactivity as a function of startup rate and period. The kinetics parameters are evaluated at hot full power and hot zero power conditions for various cycle burnups and control rod configurations.

The PHOENIX-P cross section library contains delayed neutron fractions and decay constants for fissionable nuclides for each of the six delayed neutron energy groups. The core averaged delayed neutron fractions are obtained by weighting the delayed neutron fractions for each group by the regionwise fraction of fissions in each isotope and the regionwise power and volume weighting in the core. The core average decay constants are calculated in a similar manner. The fraction of fissions in each isotope are obtained from single'ssembly PHOENIX-P calculations. Regionwise power sharings for various core conditions are obtained using the ANC core model. A delayed neutron importance factor (to account forspectrum differences between delayed and prompt neutrons) is used to calculate an effective core average delayed neutron fraction.

The prompt neutron lifetime also depends upon the core composition (fuel enrichment, burnup, absorbers, etc.). Single assembly PHOENIX-P calculations provide the neutron lifetime for the fuel in each core region.

The core average value is determined through a power and volume weighting process.

CORE PHYSICS PARAMETERS FOR TRANSIENT ANALYSIS INPUT The physics models described in Section 2 are used to generate key input parameters for various safety analyses. These key safety parameters include reactivity coefficients, control rod worth, and limiting power distributions during both normal operations and accidental transients.

Reference 5 provides a detailed description of how these parameters are calculated for Turkey Point.

4.0 PHYSICS MODEL VERIFICATION TURKEY POINT UNITS Core physics model verification typically includes comparisons of predictions to plant startup and operating data. Turkey Point Units 3 & 4 are currently in their fourteenth and fifteenth cycles of operation, respectively. In this section, predictions made using the physics methodology described in Section 2 are compared to zero power physics test measurements and at power operating data for Turkey Point. For St. Lucie, this data is presented in Section 5.

As stated in Section 1, the methods employed to generate the predictions reported in this section are standard licensed and NRC approved methods used by Westinghouse's Commercial Nuclear Fuel Division. The comparisons reported herein provide additional verification of the predictive capabilities of this methodology; however, their primary purpose is to demonstrate FPL's ability to perform design calculations for the Turkey Point Units 3 & 4.

e Turkey Point Units 3 & 4 are similar in design. Each reactor is a closed cycle pressurized light water moderated and cooled system, which uses slightly enriched uranium dioxide fuel. Each unit is currently designed to produce 2200 MWt core power. The reactor core consists of 167 fuel assemblies. Turkey Point Units 3 and 4 core and fuel assembly designs are essentially identical, both utilizing a low leakage core design. Each fuel assembly consists of a 16x16 array of 204 fuel rods, 20 guide thimbles, and one instrument thimble. The Turkey Point Unit 4 Cycles 12, 13, and 14 were selected for core physics model verification, since each of these cycles has different design attributes which provide an opportunity to model different design features.

4.1 CYCLE DESCRIPTIONS Turkey Point Unit 4 Cycle 12 began operation on June 11, 1989, and shutdown on November 24, 1990 after 406 effective full power days (EFPD),

corresponding to a cycle burnup of 12441 megawatt days per metric ton

(MWD/MTU). Turkey Point Unit 4 Cycle 12 was fueled with two different fuel designs. The burned fuel of Regions 9B, 11B, 12C, 13A, 138, and 13C are the familiar Low Parasitic Fuel (LOPAR) design. Regions 12A, 12B, 13D, and 13E and the fresh regions 14A, 14B, 14C, and 14D are of the Westinghouse Optimized Fuel Assembly (OFA) design. The core loading pattern for Cycle 12, including the assembly locations, the number of integral Fuel Burnable Absorbers (IFBAs), the number of Wet Annular Burnable Absorbers (WABAs), and the locations of control banks are shown in Figure 4.1-1. The core also contains part-length hafnium rods.

These rods decrease the power and thereby the fast fluence in core locations close to the reactor pressure vessel weld. This reduction is required for pressurized thermal shock (PTS) considerations. There are 240 hafnium rods in the core. They are 36 inches long and positioned slightly below the core midplane. Figure 4.1-1 gives the core locations for the hafnium rods. A quarter core representation is used since the core is symmetric.

Turkey Point Unit 4 Cycle 13 began operation on October 27, 1991 and shutdown on April 10, 1993 after 441 EFPD, corresponding to a cycle burnup of 13433 MWD/MTU. Turkey Point Unit 4 Cycle 13 was also fueled with both LOPAR and OFA fuel assembly designs. In Cycle 13, sixteen assemblies from earlier cycles were re-inserted. Special modeling of these re-inserted assemblies was necessary to account for their loss in reactivity due to the excessive time that the re-inserted assemblies resided in the spent fuel pool. The core loading pattern for Cycle 13, including the assembly locations, the number of IFBAs, the number of WABAs, and the locations of control banks are shown in Figure 4.1-2. The Cycle 13 core also contains the part-length hafnium rods as in Cycle 12.

Turkey Point Unit 4 Cycle 14 began operation on May 26, 1993 and was shutdown on October 2, 1994 after 454 EFPD, corresponding to a cycle 0 burnup of 13793 MWD/MTU. Turkey Point Unit 4 Cycle 14 was fueled entirely with assemblies of OFA design, and the fresh fuel of Region 16 introduced axial blankets into Unit 4. Axial blankets consist of a nominal six inches of natural UO, pellets at the top and bottom of the fuel pellet stack to reduce neutron leakage and to improve uranium utilization. The core loading pattern for Cycle 14, including the assembly locations, the number of lFBAs, the number of WABAs, and the locations of control banks are shown in Figure 4.1-3. The Cycle 14 core also contains the part-length hafnium rods as in Cycles 12 and 13. Fuel batch characteristics for Cycles 12, 13, and 14 are summarized in Table 4.1-1.

4.2 ZERO POWER PHYSlCS TESTS After each refueling at the Turkey Point units, startup physics tests are conducted to verify that the nuclear characteristics of the core are consistent with design predictions. While the reactor is maintained at hot zero power (HZP) conditions, the following physics parameters are measured:

Critical boron'oncentrations, isothermal temperature coefficient, Control rod worth, and Differential boron worth Table 4.2-1 contains the zero power physics test review criteria, which represent the maximum expected deviation between predicted and measured values for each parameter.

P The following sections briefly describe the measurement and calculational techniques and summarize the results of the zero power physics tests for Turkey Point Unit 4, Cycles 12, 13, and 14. Small changes in core reactivity were measured by feeding the signal from a power range neutron

detector into a reactivity computer which solves the point kinetics equation. The computer output was plotted on a strip chart recorder. All predictions were made with the three-dimensional ANC model described in Section 2.4.

4.2.'I CRITICAL BORON CONCENTRATIONS Critical boron concentrations were measured by acid-based titration of reactor coolant samples taken under equilibrium conditions. Samples were taken with all rods essentially out and with the reference bank (see Section 4.2.3) inserted. Critical boron searches were performed with the three-dimensional ANC model for these core configurations to obtain the predicted concentrations. The measured and predicted critical boron concentrations are compared in Table 4.2-2. All differences are within the

+60 ppm review criteria.

4.2.2 TEIIPERATURE COEFFICIENTS

~ ~

Isothermal temperature coefficients (ITCs) were measured by making small changes in the reactor coolant system temperature and determining the corresponding change in reactivity with the reactivity computer. ITCs were predicted by uniformly varying the core temperature by +6'F about the HZP temperature in the ANC model. The moderator temperature is varied directly; Doppler effects on reactivity are determined using fitting coefficients obtained from FIGHTH calculations. The measured and predicted ITCs and Moderator Temperature Coefficients (MTCs) are compared in Table 4.2-3. All differences are well within the review criteria of +2 pcml'F. The measured Moderator Temperature Coefficient is obtained by subtracting the Doppler Coefficient from the measured ITC.

t 4.2.3 CONTROL ROD WORTH Control rod worths were measured by the Rod Swap Technique. First, the worth of the reference bank (the bank of highest worth) was measured by 0 boron dilution. Stepwise bank insertion was used to maintain criticality and differential worth were obtained from the reactivity computer response. The differential worths were summed to provide the integral worth of the reference bank. Then, maintaining the boron concentration at a constant value, critical configurations were established with each remaining bank fully inserted and the reference bank partially withdrawn.

The integral worth of each inserted bank was determined from the critical position of the reference bank after the exchange by applying analytical corrections to account for the effect of the inserted bank on the partial integral worth of the reference bank. This procedure is described in Reference 6.

The ANC model was used to predict the individual control rod bank worth as well as to generate the corrections used to infer the measured worth.

The measured and predicted worth are compared in Table 4.2Q; all differences are within the review criteria listed in Table 4.2-1. Measured and predicted reference bank integral rod worth shapes are compared in Figures 4.2-1 through 4.2-3.

4.2.4 DIFFERENTlAL BORON WORTH Measured differential boron worths were obtained by dividing the measured reference bank worth (see Section 4.2.3) by the difference between the critical boron concentrations measured with all rods out and with the reference bank inserted. The differential boron worth does not change significantly over this range of boron concentration. Boron worths were predicted by varying the boron concentration by +25 ppm about the HZP all rods out critical 'boron concentration in the ANC model. The measured and predicted boron worth are compared in Table 4.2-5. All differences are well within the +15% review criteria.

POWER OPERATION In support of the Turkey Point Technical Specification requirements, the core power distribution is measured at least once every 31 EFPD using the in-core instrumentation system. Neutron flux measurements made by movable in-core fission chambers are combined with analytically determined power to reaction rate ratios using the computer program INCORE (Reference 2) to infer (i.e., "measure"), a three-dimensional power distribution. The power to reaction rate ratios are generated with the three-dimensional ANC model using cross sections derived from PHOENIX.

INCORE is a data analysis code written to process information obtained by in-core instrumentation. INCORE synthesizes measured axial flux shapes and theoretical elevation dependent X-Y power distributions to obtain a power distribution throughout the core.

In this section, measured data obtained from INCORE is compared to predictions made with the three-dimensional ANC Model. Included are:

Power peaking factors, F~ and F~,

Average assembly radial power distributions, Core average axial power distributions, and Axial offset Also, measured and predicted boron letdown curves are compared. Boron letdown refers to the reduction of the all rods out (ARO), hot full power (HFP) criticai boron concentration as a function of core burnup.

4.3.1 BORON LETDOWN CURVES Reactor coolant system boron concentrations are measured daily regardless of power level or control rod bank insertion. Critical boron concentrations measured at or very close to hot full power all rods out equilibrium xenon and samarium conditions are compared to the predicted 0 boron letdown curves for Cycles 12, 13, and 14 in Figures 4.3-1 through 4.3.3. The predicted curves were obtained from design depletions with the three-dimensional ANC model.

'fable 4.3-1 compares measured and predicted critical boron concentrations at the time of INCORE power distribution measurements.

The measured concentrations were corrected to hot full power all rods out equilibrium xenon and samarium conditions in accordance with the Turkey Point units surveillance procedures. The predicted concentrations were obtained by performing critical boron searches with the ANC model at the specified burnups of the measurements. The mean difference between measured and predicted critical boron concentrations for all three cycles is 9 ppm with a standard deviation of 13 ppm.

4.3.2 POWER PEAKING FACTORS

~ ~

The nuclear enthalpy rise hot channel factor (F~) and the heat flux hot channel factor (F~) were measured using the INCORE code, as discussed above. Predicted peaking factors were obtained from three-dimensional ANC calculations performed for core conditions similar to those at the time of the measurements. Power peaking factors measured during Cycles 12, 13, and 14 are compared to predicted values in Figures 4.3-4 through 4.3-9

, and in Tables 4.3-2 and 4.3-3. For F~, the mean difference between the measured and predicted values for the three cycles is 2.02% with a standard deviation of 1.27%; for F~ the mean difference is 3.33% with a standard deviation of 1.86%. Regarding the F~ comparisons, it is noted that spacer grid effects are inherent in the measured values but the grids are not explicitly modeled in ANC. The magnitude of this effect can be seen from Figures 4.3-19 through 4.3-27.

0 4.3.3 RADIAL POWER DISTRIBVTIONS

~ ~

Core power distributions were measured with the INCORE code, as discussed above. The measured power distributions are typically referred to as flux maps. INCORE also produces predicted power distributions at the burnup of the flux map by interpolating between power distributions generated using the three-dimensional ANC model at specific burnups during a depletion calculation. Since the core is loaded symmetrically, ANC depletion calculations are performed assuming quarter-core reflective symmetry for Cycles 12 and 13, and rotational symmetry for Cycle 14. The predicted power distributions are expanded to full core for comparison to the measured distributions.

Figures 4.3-10 through 4.3-18 compare measured and predicted assembly relative power distributions at selected burnups for Cycles 12, 13, and 14.

All comparisons are for the hot full power all rods out condition since this is the normal mode of operation for the Turkey Point units. The mean absolute difference between measured and predicted assembly relative powers is less than .021 and the standard deviation is less than .023 for these comparisons.

4.3.4 AXIALPOWER DISTRIBUTIONS AND AXIALOFFSETS Measured core average axial power distributions from each of the flux maps discussed in the previous section are compared to predicted axial distributions in Figures 4.3-19 through 4.3-27. The predicted distributions were obtained from three-dimensional ANC calculations performed for core conditions similar to those. at the time of the flux maps. Note that since the grid straps are not modeled explicitly in the ANC model, no depressions are seen at the grid locations in the predicted distributions.

This difference coupled with the normalization of both measured and predicted axial power distributions to unity causes the measured relative power to appear slightly higher between grid locations.

Axial offset refers to the percent difference between the relative power in the top half of the core and that in the bottom half of the core divided by the sum of these two relative powers. Axial offsets measured using the INCORE code are compared to predicted values from ANC calculations for core conditions similar to those at the time of the measurements in Table 4.3X. The mean difference between measured and predicted values for Cycles 12, 13, and 14 is 0.66% with a standard deviation of 1.54%.

4.4

SUMMARY

In this section, predictions made using Westinghouse's reload core design methodology are compared to zero power physics test measurements and at-power operating data from Turkey Point Unit 4, Cycles 12, 13, and 14.

In all cases, the predictions agree well with the measurements. All startup test predictions are within the review criteria listed in Table 4.2-1 ~

Predicted critical boron concentrations at power are within 50 ppm of the measured values, and the predicted power distributions are close to the measured values, as evidenced by Figures 4.3-10 through 4.3-27. The excellent agreement between the predictions and the measurements reported here demonstrates FPL's capability to apply the Westinghouse licensed methodology to reload core design for Turkey Point Units 3 and 4.

TURKEY POXNT UNXT 4 CYCLE 12, 13 AND 14 FUEL SPECXPXCATXON HUMBER OP INITIAL BOC ASSEMBLIES ENRICEQG&T BURNUP w/o U-235 mo/mv 9B 1 3.40 16080 11B 8 3.40 29279 12A 8 2.60 22862 12B 12 3.45 28354 12C 8 3.00 24975 13A 4 3.00 18285 13B 4 3.10 17489 13C 4 3.10 17592 13D 24 3.20 18769 13E 32 3.40 15354 14A 28 3.40 0 14B 4 3.40 0 14C 8 3.80 0 14D 12 3.80 0 9 8 3.30 27419 9B 1 3.40 29845 11A 8 3.10 26325 13B 4 3. 10 28073 13C 4 3. 10 31944 13D 8 3.20 32654 13E 24 3.40 27752 14A 28 3.40 16549 14B 4 3.40 16232 14C 8 3.80 13179 14D 12 3.80 13952 15A 16 3.60 0 15B 12 3.60 0 15C 16 4.00 0 15D 4.00 0 13E 8 3.40 32717 14A 25 3.40 31222 14B 4 3.40 31768 14C 8 3.80 24027 14D 12 3.80 30454 15A 16 3.60 17883 15B 12 3.60 17489 15C 16 4.00 15004 15D 4.00 15055 16A 16 3.60 0 16B 16 3.60 0 16C 4 3.60 0 16D 8 4.00 0 16E 8 4.00 0 0

TABLE 4.2-1 TKGGCEY POINT UNIT 4 CYCLE 12, 13 AND 14 HZP PHYSICS TEST REVIEW CR1TERIA PAEVLMETER REVIEW CRITERIA Critical BoronConcentration ~50 ppm Temperature Coefficients: +2. 0 pcm/'F Moderator Temperature Coefficient Isothermal Temperature Coefficient Control Rod Bank Worths:

Reference Bank Worth ~10'o "Swap" Worths i15: or 100 pcm whichever is greater Differential Boron Worth ~15'o

TABLE 4. 2-2 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 CRITICAL BORON CONCENTRATION COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CRXTXCAL BORON CONCENTRATXON (PPM)

CONF XGURATXON MEASURED PREDXCTED DXFFERENCE M P (M-P) 12 ARO 1538 1584 -46 12 BANK C in 1399 1428 -29 13 ARO 1554 1560 -6 in 0"

13 BANK A 1401 1408 -7 ARO 1698 1691 BANK SB in 1552 1544 Acceptance Criteria is +50 ppm

~ 32

TABLE 4.2-3 TUEQ<EY POINT UNIT 4 CYCLE 12, 13 AND 14 MODERATOR AND ISOTHERMAL TEMPERATURE COEFF ICXENT COMPARISON BETWEEN MEASUREMENT AND PREDXCTXON CYCLE BANK MODERATOR TEMPERATURE COEFFICIENT (PCM/ F)

CONFIGURATION MEASURED PRED1CTED DIFFERENCE M P (M-P) 12 0.92 0.58 0.34 13 0.24 0.06 0.18 0.26 1. 13 -0.87 0 CYCLE BANK ISOTHERMAL TEMPERATURE COEFFICIENT (PCM/ F)

CONF IGURATION MEASURED PREDICTED DIFFERENCE M P (M-P) 12 -0.88 -0.63 -0.25 13 -1.66 -1.65 -0.01

-1.44 -0.57 -0.88 Acceptance Criteria is +2 pcm/'F e

TABLE 4.2-4 TURKEY POINT UNIT 4 CYCLE 12, 13 AND 14 CONTROL ROD WORTH COMPARISON BETWEEN MEASUREMENT AND.PREDICTION CYCLE BANK COKZROL ROD WORTH (PCM)

CONF ZGURATZON MEASURED PREDZCTED D1FFERENCE(0)

M P ( (M-P) /P) *100 12 BANK D 691 718 -3. 76 BANK C(1) 1314 1365 -3. 74 BANK B 375 380 1 31

~

BANK A 1177 1204 -2. 24 BANK SB 1180 1202 -1. 83 BANK SA 1000 1017 -1.67 TOTAL (2) 5737 5886 -2.53 13 BANK D 641 682 -6.09 BANK C 1022 992 2.97 BANK B 435 457 -4. 88 BANK A(1) 1232 1275 -3.41 BANK SB 1183 1233 -4. 03 BANK SA 826 836 "1.17 TOTAL (2) 5338 5475 -2.51 BANK D 636 661 -3. 78 BANK C 1093 1172 -6. 74 BANK B 435 480 -9.37 BANK A 1086 1102 -1.45'1.84 BANK SB (1) 1173 1195 BANK SA 1052 1094 -3.84 TOTAL (2) 5475 5704 -4.01 Acceptance Criteria is +15. or 100 pcm which ever is greater (1) Reference Bank - Acceptance Criteria is +10%

(2) Sum of all measured banks within +7%

TABLE 4.2-5 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 HZP DIFFERENTIAL BORON WORTH COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK DIFFERENTIAL BORON WORTH (PCM/PPM)

CONFIGURATION MEASURED PREDICTED DXPPERENCE (8)

M P ((M-P) /P) *100 12 Average Over Bank C insertion 9.45 8.78 7.63 13 Average Over Bank A insertion 8.05 8.34 -3.47 14 Average Over Bank SB insertion 8.56 8.13 5.29 TABLE 4.3-1 TIMEY POINT UNIT 4 CYCLE 12, 13 AND 14 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE CYCLE BURNUP CRXTXCAL BORON CONCENTRATXON (PPM)

MND/MTV PREDXCTED DIFFERENCE P (M-P) 12 0 1437 1426 11-150 1111 1124 -13 2000 1020 1006 14 2320 986 989 -3 3000 954 950 4020 867 882 -15 4940 815 812 3 5890 744 733 11 6975 657 637 20 8000 567 546 21 10000 379 361 18 11184 277 250 27 0- 12000 12441 150 1000 2000 2440 200 161 1082 1014 938 924 174 133 1104 1027 953 926 26 28

-22

-13

-15

-2 3224 862 869 -7 4888 750 739 11 6678 610 591 19 8265 473 458 15 8754 436 418 18 10608 276 258 18 12316 121 109 12 150 1213 1212 1 600 1189 1166 23 1000 1135 1142 -7 1830 1103 1098 5 2521 1056 1043 13 3428 991 980 11 5000 869 858 11 5986 792 775 17 8148 601 587 14 8995 520 510 10 9871 453 428 25 10704 362 349 13 12000 218 226 -8

TABLE 4.3-2 TUEL(:EY POINT UNIT 4 CYCLE 12, 13 AND 14 POWER PEAKING FACTOR (F~) COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE CYCLE BURNUP F~ (MAX) mo/MTU MEASURED PREDZCTED DXFFERENCE M P ( (M-P) /P) *100 12 150 1.475 1.416 4.17 4945 1.492 1.456 2.47 5860 1.499 1.459 2.74 6890 1.498 1.456 2.88 7620 1.493 1.471 1.51 8363 1.502 1.481 1.41 9082 1.511 1.486 1.70 0 9458 10323 11121 11812 1.509 1.511 1.504 1.508 1.489 1.487 1.484 1.479 1.35 1.60 1.35 2.01 13 150 1. 557 1 ..486 4.77 244 0. 1. 438 1. 440 -0.14 3224 1.442 1. 437 0.35 4888 1.447 1.435 0.84 6678 1.519 1.455 4.39 8265 1.509 1. 477 2.16 8754 1.511 1.491 1.34 10608 1.541 1.507 2.26 12316 1.545 1.508 2.45 600 1.468 1.420 3.38 1830 1.462 1.421 2.89 2521 1.485 1.427 4.06 3428 1.465 1.427 '2. 66 5986 1.471 1. 440 2.15 6836 1.472 1.452 1.38 7659 1.478 1.460 1.23 8143 1.476 1. 4'64 0.82 8995 1.482 l. 464 1.23 9871 1.494 1.461 2.26 10704 1.496 1.462 2.33 TABLE 4. 3-3 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 POWER PEAKING FACTOR (Fg) COMPARISON BETWEEN MEASUR229"NT AND PREDICTION CYCLE CYCLE BURNUP (MAX) mO/MTU PREDXCTED DXFFERENCE P ( (H-P) /P) +100 12 150 1.920 1.692

l. 874 2.39 4945 1. 632 3.67 5860 1.709 1. 649 3.63 6890 1 718

~ 1.644 4.50 7620 1.722 1.657 3.92 8363 1.709 1.672 2.21 9082 1.699 1.672 1.61 9458 1.724 1.671 3.17 10323 1.713 1.661 3.13 11121 1.692 1.652 2.42 11812 1.688 1.644 2.67 150 1.773 1.756 0.97 2440 1.642 1.634 0.49 3224 1.614 1.621 -0.43 4888 1.664 1.596 4.26 6678 1.719 1.617 6.31 8265 1.726 1.650 4.60 8754 1.732 1.669 3.77 10608 1.762 1.681 4.82 12316 1.755 1.671 5.02 14 600 1.815 1.682 7.91 1830 1.745 1.677 4.06 2521 1.825 1.686 8.24 3428 1.730 1.672 3.42 5986 1.725 1.681 2.62 6836 1.742 1.688 3.20 7659 1.741 1.699 2.47 8143 1.741 1.702 2.29 8995 1.735 1.701 2.00 9871 1.729 1.701 1.65 10704 1.732 1.696 2.12 4.3-4 T&&EY POXNT UNZT 4 CYCLE 12,13 AND 14 AXIAL OFFSET COMPARXSON MMZllitlltlMMRROIIMMMMZ lIMO ZRMOZCI'ZOM CYCLE CYCLE BURNUP AZXAL OFFSET (8) mWD/MTV PREDXCTED DIFFERENCE P M-P 12 150 -2.46 -2.57 -0.11 4945 -2.55 -1.85 -0.70 5860 -2.99 -1.96 -1.03 6890 -3.05 -2.13 -0.92 7620 -1.60 -2.28 0.68 8363 -1.42 -2.24 0.82 9082 -0.87 2 ~ 11 1.24 9458 -1.66 -2.06 0.40 10323 -3.09 -1.96 1 13

~

11121 -2.04 -1.93 -0.11 11812 -2.23 -1.93 -0.30 13 150 F 00 0..34 6.66 2440 2.34 -1.42 3.76 3224 0.93 -1.79 2.72 4888 -0.41 -2.25 1.84 6678 -0.66 -2.45 1.79 8265 -1.51 -2.59 1.08 8754 -1.99 -2.50 0.51 10608 -1.99 2 23

~ 0.24 12316 -1.61 -1.86 0.25 14 600 3.62 1.85 1.77 1830 1.19 0.56 0.63 2521 3428

-1 '0

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~ 0

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FIGURE 4.1-2 TURKEY POINT UNIT 4, CYCLE 13 LOADING PATTERN 48 ASSEMBLY FEED 6 5 4- 2 9B 15B 13E 15A 13B 15A 'j3C D 8 WABA SB 0 WABA 0 8 WABA 20 HF 15B 13E 14D 14A 15A 9 15C 13D 8 WABA A 6 WABA SA 20 HF 13E 11A 15B SB C 4 WABA 15A 14A 14A 14D 15C 13E 0 WABA A SB 8 WABA 13B 15A 11 A 14A 15D LEGEND 0 6 WABA C RCCA Bank or 15A 9 15B 15C Removable BP 8 WABA SA 4 WABA 8 WABA HF a PTS Hafnllkn Ateocber R a Re4naerted Aeeocnbly 15C 14C 13E 15A-3.6 w/o No IFBA B 15B -3.6 w/o 32 IFBA 15C-4.0wlo No IFBA 13C 'f 3D 150 -4.0w/o 88 IFBA 20 HF 20 HF 28 Assemblies 3.6 Wt.%

20 Assemblies O 4.0 Wt.%

FIGURE 4.1N TURKEY POINT UNIT 4, CYCLE 14 LOADING PATTERN 52 ASSEMBLY FEED 6 5. 4- 2.

H 14A 158 'I6B 14A 'f68 14D 14B D SB WABA D 4 WABA 20 HF' 15A 16C 15A 14C 16A 15B 14A 8 WABA A 6 WABA SA 20 HF 15B 15A 14A 16A 14D 16B 15C SB 16 WABA C 4 WABA B E 16B 14C 16A 15A 15C 16E 13E 8 WABA A 6 WABA SB 14A 16A 14D 15C 15D LEGEND D 6 WABA C 16B 158 16B 16E 14A 4 WABA SA 4 WABA 14O 16D 15C 13E 16A - 3.6 w/o No IFBA B 16B -3.6 w/o 32 IFBA 16C -3.6 w/o 64 IFBA 14B 14A 16D-4.0w/o 16 IFBA 20 HF 20 HF 16E -4.0 w/o 48 IFBA 36 Assemblies @3.6 Wt.%

16 Assemblies 4.0 Wt.%

FX~& 4.2-1 raaZZV XOXmm VNXm 4 CrCLZ 12 IISUIIEO IIERSUS SSEORUSEO BANK C XNTEGRAL ROD WORTH 1400 1300 1200 PREDICTED 1100 g 1000 O

900 g

D4 soo A

g 700 600 500 400 300 200 100 0

0 40 80 120 160 200 240 ROD POSITION (STEPS WXTHDRAWN)

FXGURE 4.2-2 TURKEY POXNT UNXT 4 CYCLE 13 MEASURED V1MSU8 PREDXCTED BANK A XNTEGRAL ROD WORTH 1400 1300 MKR,SURED 1200 PREDICTED 1100 g 1000 O

Qc 900 g

800 A

700 600 500 400 300 200 100 0

0 40 80 120 160 200 240 ROD POSITION (STEPS WITHDRAWN)

FXGURE 4 2<<3 TURKEY POXNT UNXT 4 CYCIsZ 3 4 EEIIEPEEP VEIIPEP PEEPECPE EQLSK SB XNTEGRAL ROD WORTH 1400 1300 1200 MKR.SWED PREDXCTED 1100 g 1000 U

900 g

coo A

g 700 600 500 400 300 200 100 0

0 40 80 120 160 200 240 ROD POSXTXON (STEPS NXTHDRAWN)

FIGURAL 4-3-1

!VUXKEY POINT UNXT 4 CYCLE 12 BORON LETDOWN COMPARISON BETWEEN AND PREDICTXON 1600 1SOO 1400 MFJLBU&~23 1300 PREDICTED

> 1200 g

0 H1100 g

~ 1000 g

900 0

O 800 0g 0

700 600 500 400 300 200 100 0

0 2000 4000 6000 8000 10000 12000 14QQO CORE AVERAGE BURHUPg MWD/MTU

PXGURE 4 3-2 TURKEY POXNT UNXT 4 CYCLE 13 BORON LETDOWN COMPARXSON BETWEEN AND PREDXCTXON 1600 1500 1400 MEASURED 1300 PREDXC TED L 1200 O 1100 f, 1000 800 0

0 700 600 H

500 400 300 200 100 2000 4QOO 600Q 800Q 1QOOO 12000, 14000 CORE AVERAGE BURRKJPg MWD/MTU 0

FXGURE 4 3-3 TURKEY POXNT UNXT 4 CYC~ 14 BORON LETDOWN COMPARXSON BETWEEN AND PREDXCTXON 1600 1500 1400 1300 PREDXCTED PI 1200 8

O 1100

~1000

~

g 0

800 0

700 600 500 400 300 200 100 0

0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP i MWD/HTU e

FXQURPi 4.3-4 maZZm POXNT ONXT 4 CrCZa t2 F DELTA H COMPARXSON BETWEEN XNCORE AND ANC 2.0 1.9 1.8 XNCORE 1.7 1.6 1.4 1.3 1.2 1.0 0 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUP, MWD/MTU FXCt9& 4.3-5 TURKEY PQXNT UNXT 4 CYCLE 13 F DELTA H COMPARXSON BETWEEN XMCQRE AND ANC 2.0 1.9 XNCORE 1.8 1.7 1.6 0!'.3 1.0 0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURHUPi MWD/MTU 5Q e

FXGUE& 4 3-6 mnmxm POXNT amXT 4 CYCLE 14 F DELTA H COMPARXSON BETWEEN XNCORE AND AMC 2.0 1.9 1.8 INCORE 1.7 1.6 1.4 1.3 1.2 1.0 0 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUPg MWD/MTU

FXGURE 4 3-7 TURKEY POXNT UNXT 4 CYCLE 12 FQ COMP2QLESON BETWEEN XNCORE AND AMC XNCORE h h 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUPg MWD/MTU FXCaaE 4.3-8 TURKEY POXNT UNXT 4 CYCLE 13 FQ COMPARXSON BETWEEN XNCORE AND AMC XNCORE d

2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP, MWD/MTU

- S3-

FXCRGtE 4 3-9 TUESDAY POXNT UNXT 4 CYCLE 14 FQ COMPARXSON BETWEEN XNCORE AHD ANC 2 50 2.25 XNCORE 2.00 1.25 1.00 0 2000 4000 6000 8000 10000 12000 CORE AVEEVRGE BURWJPg MWD/MTU I.

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FIGURE 4.3-12 TURKEY POINT UNIT 4, CYCLE 12 RADIALPOWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC 15 14 13 12 10 9 8 7 6 0267 0.325 G267 I 0266 G 319 0266 I 1.88% 0.38%

0.3S% I OA10 0.704 1.058 Q847 1.038 G6S7 Q393 I 0.410 0.697 1.049 Q 849 1.049 0.697 0.410 I P 0.86% %24% -1.05% -1A3% 4.15%

0.00% 1.00% I 0.492 1.158 1.301 1.084 1296 1.075 1Z63 1.111 0.474 'I 0.495 1.139 1275 1.075 1.316 1.075 lZ75 1.139 0.495 I N

%.61% 1.67% 204% 0.84% -1.52% 0.00% %.94% -2A6% <24%

I OA92 1.175 1.170 1.140 .1.370 1.048 1.369 1.138 1.165 1.135 0.475 I OA95 .1.158 1.150 1.126 1.377 1.033 1.377 1.126 1.150 1.15S 0.495 I M

%.61% 1A7% 1.74% 124% %.51% lA5%  %.58% 1.07% 1.30% -1 99% I 0.397 I.140 1.162 1.108 1.383 1.115 1.147 1.112 1.374 1 107

~ l.151 1.128 G398 I 0.411 1.139 1.150 1.123 1.385 1.1N 1.142 1.108 1.384 1.123 1.150 1.139 OA10 I

-3.41% 0.09% 1.04% -1.34% %.14% 0.54% OA4% 0.36% %.72% -1.42% 0.09%  %.97% -2.93% I 0.683 1261 I.117 1.366 1.073 1.163 1.159 1.152 1.05S 1247 1.126 1Z77 Q691 I 0.697 1.275 1.126 1.383 1.056 1.153 1. 145 1.152 1.0M 1.383 1.126 1275 0.697 I

-2.01% -1.10% -123% 1.61% 0.87% 122% 0.00% 0.19% -2.60% 000% 0.16% %.86%

0265 1.038 1.079 1.373 1.120 1.158 l.383 1.079 1.347 l. 143 1.092 1.361 1.085 1.046 0267 0.266 1.049 1.075 1.377 1 108

~ 1.151 1.361 1.071 1.361 1.151 1.107 1.377 1.075 1.049 0266

-1.05% 0.37% %29% 1.0S% 0.61% 1.62% 0.75% -1.03% %.70% -1.36% -1.16% 0.93% %29% 0.38%

0.327 0.879 l.324 1.060 1.156 1.155 1.089 1.147 1.09S 1.153 1.146 1.043 1.325 0.846 0.324 0.319 O.S50 1.315 1.033 1.143 1.145 1.071 1.109 1.071 1.145 1.143 1.032 1.315 0.849 0.319 H 2.51% 3.41% 0.68% 2.61% 1.14% O.S7% 1.6S% 3.43% 2.52% 0.70% 0.26% 1.07% 0.76% %.35% 1.57%

0.273 1.077 1.101 1.378 1.117 1.168 1.397 1.091 1.390 1.168 1.118 1.362 1.054 1.036 0267 0.266 1.049 1.075 1.377 1.107 1.151 1.361 1.071 1.361 1.151 1.107 1.377 1.075 1.049 0266 2.63% 2.67% 2.42% 0.07% 090% 1.48% 2.65% 1.87% 2.13% 1.48% 0.99% -1.5% -1.95% -1.24% 0.38%

0.695 1.281 1.129 1.365 1.065 1.167 1.158 1.174 1.075 1.38S 1.130 1070 0.687 0.697 1.275 1.126 1.383 1.056 1.152 1.145 1.152 1.056 1.383 1.126 1275 0.697 C.29% OA7% 0.27% -1.30% 0.85% 1.30% 1.14% 1.91% 1.80% 0.36% N.39% -1.43%

0.392 1.120 1.155 1.105 1.353 1.107 1.160 1.122 1.359 1.104 1.161 1.156 OA09 0.411 1.139 1.150 1.123 1.384 1.108 1.141 1.10S 1.384 1.123 1.150 1.139 0.410

<.62% -1.67% OA3% -160% -2.24% 4.N% 1.67% 126% -1.81% -1.69% 0.96% lA9% %24%

0.475 1.137 1.142 1.114 1.365 1.073 1.361 1.113 1.145 1.146 'OA84 0.495 1.158 1.150 1.126 1.377 1.032 1.376 1.126 1.150 1.157 0.495 D

<.04% -1.81% 4.70% -1.07% 4.87% 3.97% -1.09% -1.15% C.43% -2.22%

0.471 1.113 1.254 1.083 1.342 1.055 1252 1.119 0.475 0.495 1.139 I.274 1.075 1.315 1.075 1274 1.138 0.495 INCORE 4.85% -2.28% -1.57% 0.74% 2.05% -1.86% -1.73% -1.67% <.04%

0.391 0.679 1.061 0.853 l.044 0.674 0.394 ANC 0.410 0.697 1.048 0.849 1.048 0.697 0.410 4.63% -2.58% 124% 0.47% 4.38% -3.30% -3.90%

% DIFFERENCE 0.269 0.327 0.269 0266 0.319 0266 Mean Absolute Difference 0.013 1.13% 251% 1.13%

Standard Deviation 0.009 BURNUP = 11812 MWD/MTU POWER LEVEL = 1HP%%d D BANK AT 228 STEPS

FIGURE 43-13 TURKEY POINT UNIT 4, CYCLE 13 RADIALPOWER DISTRIBUlION COMPARISON BETWEEN INCORE AND ANC 15 13 12 10 2 1 0251 0273 G251 I I 0244 Q272 G244 I I 2.87% 0.37% 2.87% I OA24 1.119 0.907 1.1CO GSO& OA31 I I OA16 0.814 1.146 0.928 1.146 G814 OA16 I I 1.92% 1.11% -2.3N -22N 4.01%  %.98% 3.61% I I OA57 1.089 1277 1.015 1.326 Q971 1252 1.109 I I OA47 1.106 1292 1.010 1342 1.010 1292 1.106 OA47 I I 224% -1.54% -1.16% 0.50% -1.19% -3.8N  %.10% Q27% 3.5S% I I Q452 1.070 1.104 1.014 1282 Q991 1265 I.IXI3 1.115 1.0S6 I OA48 1.080 1.105 G989 1288 0.998 1288 0.989 1.105 1.080 I I O.89%  %.93% %.09% 2.53% NA7% %.70% -1.79% 1A2% 0.90% 0.56% I G409 1.051 l.139 1270 1205 1223 1.310 1233 1.197 I Z75 1.106 1.082 OAIS OA17 1.107 1.106 125B 1.187 1214 1292 1214 1.187 125S 1.106 1.107 OA17

-1.92% -5.0N 2.98% 0.95% 1.52% 0.74% 1.39% 1.57% 0.84% 1.35% 0.00% -226% 024%

a799 1263 1.034 IZI6 1235 1.315 1.072 1204 1231 I203 1.008 1257 GS04 0.815 1293 0.992 1.188 1222 1.300 1.051 1.300 1222 1.188 G992 1293 Q815

-I 96% -2.32% 423% 1.15% 200% 0.31%

0257 G244 1~ 133 1.147 1.031 1.011 I.393 1290 2.36%

1237 1215 1.06%

1.30 I 1298 I.M5 1.051 1.316 1.M9 1.044 1.051 0.74% 126%

IM4 129S l~

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1~1 Q994 1.117 1290 1.011 1,147 G250 G244 J 5.33% -122% 1.98% 0.78% 1.S1% 023% 1.33% -3.S7% %.67% %.31% 1.07% -225% -1.68% -2.62% 2AN 272 2.94%

0.932 0.929 1.33S 1.343 0.32% 4.37%

1.011 Q999 129K 1.303 1292 0.85%

1.067 1.048 1.81%

1.324 1.363 1.040 1.053

-2.8N -123% -323%

1.319 1.363 I.o&2 1.048 iX&6 G 966 1292 0.999 1.34% -2.01% -3.30%

1.343 1243 Q929 Q919 Q00% -1.08%

a272 G272 GOO%

H 0257 Q244 5.33%

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1254 1290 1.007 1.011 1.125 1 ~ 147

-1.92%

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G 0.833 1292 1.024 1214 1240 1.322 1.070 1295 l229 1.198 0.989 12&2 Q821 0.815 12R3 0.992 1.188 1222 1.300 1.051 1.m 1222 1.188 0.992 1293 Q815 221% %.08% 323% 2.19% lA7% 1.69% 1.81%  %.38% 0.57% 0.84%  %.30% -2A0% G74%

OA33 1.112 1.128 1284 1203 1214 1291 1226 1207 12Bb 1.105 1.085 OA24 0.417 1.107 1.106 1258 1.187 1214 1292 1214 1.187 1258 1.106 1.107 OA17 3.84% 0.45% 1.99% 2.07% 1.35% Q00% %.08% 0.99% 1.68% 223% %.09% -1 99% 1.68%

OA72 1.103 1.11 I G977 1251 1.005 1297 1.018 1.103 1.0S7 Q459 OA48 1.080 1.105 0.989 1288 0.998 1288 0.9S9 1.105 1.079 OA48 D 5.36% 2.13% 0.54% -121% -2.87% 0.70% 0.70% 2.93% %.18% 0.74% 2.46%

OA71 1.084 1243 1.001 1.332 1.020 1272 1.077 OA58 INCORE OA47 1.106 1292 1.010 1.342 1.010 1291 1.106 OA47 5.37% -I 99% -3.79% 4.75% Q99% -1 A7% -2.62% 2A&%

aSo7 1.116 0.921 1.149 0.814 OA15 0.416 0.814 1.146 0.92S 1.146 0.814 OA16

'L DIFFERENCE 0.96% -2.&2% %.75% 026% Q00% %24%

0252 0281 0261 0244 G272 Q244 Mean Absolute Difference Q015 328% 3.31% 6.97%

Standard Devlatlon Q012 BURNUP = 2440 MWD/MTU POWER LEVEL = 99,4'L D BANKAT 215 STEPS

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RGURE 43-36 TURKEY POINT UNIT 4, CYCLE 14 RADIALPOWER DISTRIBU11ON COMPARISON BETWEEN INCORE AND ANC 15 13 12 10 7 6 3 2 I

I I

Q227 Q236

<.81%

0242 G251

-3.$ %

I I

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I I I

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I G349 0.798 1.085 0.7S9 1.085 0.797 a351 I I I I M52 GSOI 1.080 0.795 18386 0.803 M53 I I I I  %.85% N.37% OA&% N.75% %.09% N.75% %.57% I I I I Q390 1.050 1252 1279 1.373 1214 1254 1459 OAOI I I I I G391 1.057 1259 1230 1215 1234 12&2 1.059 0.392 I I I 426% N.66% %.56% 3.98% 4.41% -1.62% 4.63% QIXI% 2.30% I I I Q390 Q947 1242 1.090 1290 1.011 1283 1.072 1207 0.936 OA I 1 I 0.392 0.945 1240 1.081 1270 1.004 1273 1.084 1241 G945 G391 I M

%.51% Q21% 0.16% 0.83% 1.57% 0.70% Q79% -1.11% -2.74% %.95% 5.12% I Q338 1.077 1255 1261 1286 1.17S 1297 1.154 1274 1.154 1.050 I 0.353 1.059 1241 1246 1267 1.166 1.310 1.169 1271 1240 1457 Q352 I

<25% 1.70% 1.13% 120% 1.50% 1.03% N.99% -128% 024% W.94% %.66% 4.55%

I G773 l.313 1.110 1295 1.026 1218 1257 1219 1.019 1243 l.040 1. I BI 0.750 I G803 1262 1.084 1271 1.018 1225 1263 1227 1.018 1266 1.081 1259 Q801 I x

-3.74% 4.04% 2A0% 1.89% 0.79% %.57% QA8%  %.65% 0.10% -1.82% <.79% %20% %.37%

1.067 1254 1.329 1.191 1219 1282 1236 1293 1225 1.168 1220 1.152 1.045 1.086 1234 1273 1.169 1228 1285 1233 1285 1225 1,166 1270 1230 1.080

-1.75% 1.&2% 4A0% 1.88%  %.73% %23% G24% a&2% a00% 0.17% -3.94% 434% -324%

.7 l. 1.023 1.368 1285 1261 0.987 1238 1265 1288 0.941 1236 G 764 G249 1 0.795 1.315 1.004 1.310 1263 1233 0.961 1233 1263 1.310 1.II4 1.315 0.795 G251 0.80% %25%  %.15% 1.S9% 4A3% 1.74% 227% 2.71% OA1% 0.16% -1.68% %27% %.01% %.90%

0248 1.132 1256 1287 I.172 1248 1.353 1267 1.321 1270 1.181 1252 1205 1.069 Q233 0236 1.080 1230 1270 1.166 1225 1285 1233 1285 1227 1.169 1273 1234 1.086 Q237 5.08% 4.81% 2.11% 1.34% G51% 1.88% 529% 2.76% 2,80% 3.50% 1.03% -1.65% -2.35% -157% -1.69%

0.828 1294 1.091 ~~ 1273 1.038 1237 1290 1279 1.086 1.324 1.105 1261 Q780 0.801 1259 1.081 1267 1.018 1227 1263 1225 1.018 1271 1.084 12&2 Q803 3.37% 278% 0.93% OA7% 1.96% 0.81% 2.14% 4.41% 6.68% 4.17% 1.94% %.08% -2.86%

0.360 1.085 1249 1239 12&0 1,170 1.323 1.197 1.304 1278 1239 1.041 0.338 a352 1,057 1240 1246 1271 1.169 1.310 1.166 1266 1246 1241 1.059 G 353 227% 265% Q73% N.56% %.87% 0.09% 0.99% 266% 3.00% 2.57% %.16% -1.70% 425%

OA02 0 902 1.183 1.035 1279 1.021 1254 1.083 1272 0.933 G374 0.391 0.945 1241 1.084 1273 1.004 1270 1.081 1240 0.945 G392 2.81% 4.55% <.67% 4.52% 0.47% 1.69% -126% 0.19% 2.58% -127% 4$  %

0.386 1.031 1230 1216 1.374 1221 123 I 1.025 Q377 INCORE 0.392 1.059 1262 1234 1.314 12M 1259 1.057 0.391

-1.53% -2.64% -2.54% -1 A&% 4.57% 4.73% -222% -3.03% -3.5S%

Q346 0.792 1.082 0.753 1.032 0.777 Q340 0.353 0.803 1.086 0.794 1.080 0.801 0.352

'L DIFFERENCE -1 98% -1.37% 4.37% -5.16% AA4% -300% -3A1%

0235 0241 G237 0251 Mean Absolute Difference 0.021 4.84% -3 98%

Standard Deviation 0.019 BURNUP = 600 MWD/MTU POWER LEVEL = 18K D BANKAT 228 SIEPS FIGURE 4.3-17 TURKEY POINT UNIT 4, CYCLE 14 RADIALPOWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC 15 13 12 10 0248 0245 122%

0268 0265 1.13%

0249 0246 122%

I I

I I

I R

0.37) 0.797 1.069 0.784 1.067 0.792 0.370 I I Q 370 0.790 1.055 0.781 1.058 0.791 0.370 I I 027% 0.89% 1.33% 0.38% 0.85% 0.13% 0.00% I I OA09 ).092 ) 290 1.196 1.355 1.179 ) 282 1.091 OA03 I I OA03 1.084 1281 1.188 1.336 1.189 1282 1.084 OA03 I I

)A9% 0.74% 0.70% 0.67% 1A2% %.84% 0.00% 0.65% Q00% I I OA04 0.918 1.177 ).069 1.370 ).021 1.37) 1.956 1.168 0.90) 0.398 I OA03 0.910 1.170 1.067 1.359 1.011 1.360 1.069 1.170 0.910 0.403 I 025% 0.88% 0.60% 0.19% 0.81% 0.99% 0.81% %28% N.) 7% 4.99% -124% I 0.362 1.057 1 172

~ 1.188 1.346 1.158 1.384 1.167 1.369 1.177 1.148 1.066 0.36) I 0.370 1.084 1.170 1.188 1.339 1.152 1.375 1.154 1.342 1.188 1.170 1.084 Q370 I

-2.16% -1.57% 0.17% 0.00% 0.52% 0.52% 0.65% 1.13% 2.01% Z.93% -1.88% -1.66% -2.43%

I 0.790 1275 1.067 ).356 0.999 1.176 1.209 1.197 1.020 1.339 1.054 1269 0.787 I 0.791 1282 1.069 1.342 1.010 1.183 1204 1.184 1.010 '1.339 ).O67 1281 0.790 I x

%.) 3% %.55% %.) 9% 1.04% -).09% N.59% OA2% 1.10% 0.99% Q00% -122% N.94% %.38%

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.268 0.783 l.336 ).022 1.393 120) 1.183 0.949 1.185 1.182 1.361 0.996 ).363 0.786 0268

.265 0.781 ).336 1.011 1.375 1204 1.187 0.937 1.)S7 1204 1.375 1.011 1.336 0.780 0265 1.13% 0.26% 0.00% ).09% 1.31% %25% %.34% 128% 4.)7% -1.83% -1.02% -)A8% 2.02% 0.77% 1.13%

0.247 1.063 1.)85 1.367 1.165 1.1S9 1.373 ).183 1.363 1.178 ).138 1.351 1.180 1.070 0248 0.245 1.055 1.188 1.359 1.152 1.183 1.363 1.187 1.363 1.184 1.154 1.360 1.189 1.058 0246 0.82% 0.76% C.25% 0.59% 1.13% 0.51% 0.73%  %.34% 0.00% 4.5)% -1.39% %.66%  %.76% 1.13% 0.81%

0.792 ).2SO ).953 1.357 1.016 1.182 1.178 1.175 1.017 1.329 1.043 )269 0.788 0.790 1.281 1.067 1.339 1.010 1.184 1204 1.1S3 1.010 1.342 1.06S 1282 0.791 0.25% %.08% %.37% 1.34% 0.59% 4.)7% -2.16% %.68% 0.69% %.97% -2.34% -1.01% 4.38%

0.367 ).OS4 1. I70 1.178 1.345 1.142 ).377 1.153 1.339 1.168 1.153 ).074 0.366 0.370 1.084 1.170 1.188 1.342 1.154 1.375 1.152 1.339 1.188 1.170 1.084 0.370 4.81% 0.00% 0.00% %.84% 022% -'. 04% 0.'15% O.N% 0.00% -1.68% -)A5% 4.92% -1.0S%

0.404 0.906 1. ) 62 1.058  !,382 1.034 1.3S7 1.950 1.)58 0.901 0.402 OA03 0.910 1.170 1.958 ).3M 1.011 1.359 1.067 1.170 0.910 0.403 D 0.25%  %.68% C.94% ).62% 2.27% 2.06% %.66% -1.03% 4.99%  %.25%

0.392 1.052 ).243 1.178 ).360 1295 1.293 1.089 0.405 INCORE O.403 1.084 1.282 1.189 1.336 1.187 1.2S) 1.084 0.403

-2.73% -295% -3.04% 4.93% ).80% 1.60% 0.94% 0.46% 0.50%

ANC 0.356 0.767 1.054 0.792 1.076 0.805 0.374 0.370 0.791 1.058 0.780 1.055 0.790 0.370

'/ DIFFERENCE -3.78% -3.03% 4.38% 1.54% 1.99% ).90% 1.08%

0.239 0.271 0251 0246 0265 0245 Mean Absolute Dttterence O.ON -2.85% 226% 245%

Standard Devlation- 0.007 BURNUP = 6836 MWD/MTU POWER LEVEL = 100% D BANKAT 228 STEPS

FIGURE 43-) 8 TURKEY POINT UNIT 4, CYCLE 14 RADIALPOWER DISIRIBUIIONCOMPARISON BETWEEN INCORE AND ANC 15 14 13 12 10 9 8 7 0265 Q260 9291 0285 1.92% 2.11%

9266 Q26)

).92%

I I

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G391 QSOO 1.057 Q793 ).060 QSI 0.387 I G3M 0.797 1.050 a794 1.052 0.797 0.388 I 0.77% 0.3S% 0.67% %.)3% 0.76% G38% %26% I OA37 1.104 ) Z74 1.159 ) 2)0 1.167 ) 475 I 094 OA23 I OA24 )A)90 )272 1.169 1.320 1.170 ) 272 ).090 OA24 I 3.07% )2S% 0.16% %.86% %.76% %26% G24% G37% %24%

I 1.172 ).060 1.369 1.015 1279 1.071 1.170 OA23 I OA24 1.159 1.068 ) 272 )AX)8 1.373 ).ON 1.159 0.921 OA24 I M 0.94% 1.12% W.75% %22% O.N% OA4% 0.19% 0.95% %24% I G379 1.073 1.167 1.184 ).361 1.143 )275 1.156 1.391 1.184 ).16) 1.087 Q383 0.388 ),090 1.159 1.179 1.359 1.143 1.369 1.144 1.361 1.179 1.159 ).090 L388 I

-2.32% -1.56% Q69% OA2% 0.15% GOO% OA4% 1.05% 220% OA2% 0.17% %28% -) % $ I Q797

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I. I47 1.169 1.369 1.372 1.155 1.143 1.05%

I,165 1.165 1.354 1.354 1.146 1.162 1.351 1.354 0.00% 0.00% -1.38% 422% NA3% %.70%

1.161 1.166 1.136 1.144 1.372 1.373

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Q265 G261 1.53%

G 0.785 ) 243 1.052 1.388 1.016 1.149 1.134 1.148 1.018 1.352 1.04S ) 260 0.791 0.797 ) 272 1.068 1.359 1.012 1.166 1.177 1 165

~ 1.012 1.361 1.069 )272 0.797

-1.51% -228% -).50% 2.13% OA0% -)A6% -3.65% -)A6% D.59% %.66% -1.96% N.94% N.75%

0.377 1.066 I.149 1.) 51 1.355 1.115 1.363 1.138 1.359 1.171 I.IS I 1.087 MN Q388 ).090 1.159 1.179 1.361 1.144 1.143 1.358 1.179 1.159 ).090

-2.84% -2X% -2.37% NA4% -2.53% CA4% NA4% 0.07% %.68% Q.N% %2$ %

OA)6 0.904 1.135 1.039 1.389 1.036 IA09 1.061 1.139 0.919 OA30 OA24 0.921 1.159 1.959 1.373 1.008 1.371 1.158 Q921 OA23 D

-).89% -1.85% -2.07% -2.81% 1.17% 2.78% 2.77% -1.64% 422% 1.65%

0.417 1.067 )245 I.) 60 ).339 1.187 ) 280 1.075 OA25 INCORE OA24 1.090 1.272 1.170 1.320 1.)N )Z72 1.090 OA23

-1.65% -2.11% -2.12% I A4% 1.54% 0.63% -1.38% 0.47%

0.377 0.781 1.050 0.800 1.066 Q820 0.380 0.388 0.797 1.052 0.793 1.050 0.796 a3ss

% DIFFERENCE -2.84% -2.0)% %.)9% G88% ).52% 3.02% -2.06%

0256 Q261 0285 Q260 Mean Absolute Dttference QOO9 -1.92% 1.75% 1.92%

Standard Devhtlon 0.008 BURNUP = 10704 MWD/MTU POWER LEVEL = 1K8 D BANKAT 228 SIEPS ZxeVRE 4.3-19 TURKEY POXNT UNXT 4 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2.00 1.75 INCORE 1.50 g

1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHT~ INCHES BIBQCUP=7620 MWD/HTU POWER LEVEL~100'4 D BARR AT 228 STEPS t

FXG&& 4 3-20 TKGQCEZ'OXNT UNXT 4 CYCZiE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2.00 1.75 INCORE 1.50 0N 1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOH TOP 2QPZAL HEIGHT~ INCHES BURNUP~9458 MWD/HTU POWER LEVEL<100% D BANK AT 228 STEPS rXGmu" 4.3-21 TUEGKlY POXNT UNXT 4 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2 00 1.75 INCORE 1.50 1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOH TOP AXXAL HEIGHT~ INCHES BVRNUP~11812 HWD/MTU POWER LEVEL~100R D SACR AT 228 STEPS FIGURE 4 3-22 TURKIC POXNT UNIT 4 CYCLE 13 AXIAL POWER DXSTRXBUTXON COMPARISON BETWEEN INCORE AND ANC

2. 00 XNCORR 1.75 0g '.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEIGHTg XNCHES BVRNUP=2440 HWD/MTU POWER LEVEL~100% D BANK AT 228 STEPS Pxemm 4.3-23 TXGGCEY POXNT UNXT 4 CYCLE 13 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN% XNCORE AND ANC 2.00 INCORE 1.75 1.50 g

1.25 Pq 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHTs INCHES BURNUP=6678 MWD/MTU POMER LEVZZsm100% D BAHR AT 228 STEPS

-68

0 FxeURE 4.3-24 TUEQCEY POXNT UNXT 4 CYCLE 13 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2.00 1.75 INCORE 1.50 g0 1.25 Pq 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHTg INCHES BMQwWP~12316 MWD/HTU POWER L1PGW~100% D BANK AT 228 STEPS FXemm 4 3-25 TURKEY POXNT UNXT 4 CYCLE 14 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2.00 1.75 INCORE 1.50 g

1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHTg INCHES BVRNVP~600 MWD/MTU POWER ZaEVZLs100Ss D BANK AT 228 STEPS 0

0

FXGURE 4 3-26 TKGQCEY POXNT UNXT 4 CYCLE 2.4 AXXAL POWER DXSTRXBUTXON COMPARXSON BZrWZZN XNCORZ AND ANC 2.00 1.75 INCORE 1.50 g

O 1.25 Pq 1.00 N

g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOH TOP AXXAL HEIGHT~ INCHES BUEQCUP=6836 HWD/HTU POWER LEVEL~1008 D BASK AT 228 STEPS 0

I PXQURE 4.3-27 TUEQCEeY POXNT UNXT 4 CYCLE 14 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 2.00 1.75 XNCORE 0g 1.25 pq 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEIGHTg INCHES BURHUP~10704 MWD/MTU POWER LEVELm100Ss D RhHK AT 228 STEPS 0 5.0 PHYSICS MODEL VERIFICATION ST. LUCIE UNITS Core physics model verification for St. Lucie will include comparisons between measurement and predictions for St. Lucie Unit 1. St. Lucie Unit 1 is currently in its thirteenth cycle of operation. In this section, predictions made using the physics methodology described in Section 2 are compared to zero power physics test measurements and at power operating data. As stated in Section 1, the methods employed to generate the predictions reported in this section are standard licensed methods used by Westinghouse's Commercial Nuclear Fuel Division. The purpose of these comparisons is to demonstrate FPL's competence to use these methods to analyze the core configurations found at the St. Lucie Units.

St. Lucie Units 1 8 2 are similar in design. St. Lucie Unit 1 is a Combustion O Engineering (CE) reactor with a thermal rating of 2700 MW. The core consists of 217 assemblies of the CE 14x14 design. St. Lucie Unit 2 is also a CE reactor with a thermal rating of 2700 MW. The core for St. Lucie Unit 2 consists of 217 assemblies of the CE 16x16 design. The St. Lucie Unit 1 Cycles 10, 11, and 12 were selected for the core physics model verification due to the greater complexity in modelling the design features utilized in St.

Lucie Unit 1. These design features include axial blankets, Gadolinium burnable absorbers, and Vessel Fluence Reduction Assemblies (initiated in Cycle 11) which contain uranium tails and Hafnium absorbers placed in the guide tubes.

5.1 CYCLE DESCRIPTIONS St. Lucie Unit 1 Cycle 10 began operation in April 1990 and shutdown in October 1991 after a 477 Effective Full Power Days (EFPD) cycle. Cycle 10 consisted of debris resistant fuel (long end cap design) with an active fuel length of 134.06 inches. All fuel utilized axial blankets. The 0

core loading pattern for Cycle 10, including a description of the fresh fuel and the locations of control rods are shown in Figure 5.1-1. A quarter core representation is used since the core is symmetric.

St. Lucie Unit 1 Cycle 11 began operation in December 1991 and shutdown in March 1993 after a 442 EFPD cycle. Cycle 11 consisted of debris resistant fuel with an active fuel length of 136.7 inches. Vessel Fluence Reduction Assemblies (VFRA) on the core periphery were introduced in Cycle 11. The VFRA assemblies utilized uranium tails and Hafnium absorbers to reduce peripheral power. All fuel with the exception of VFRA utilized axial blankets. The core loading pattern for Cycle 11, including a description of the fresh fuel and the locations of control rods are shown in Figure 5.1-2.

St Lucie Unit 1 Cycle 12 began operation in June 1993 and shutdown in October 1994 after a 463 EFPD cycle. Cycle 12 consisted of debris resistant fuel with an active fuel length of 136.7 inches. All fuel utilized axial blankets with the exception of the VFRA . The core loading pattern for Cycle 12, including a description of the fresh fuel and the locations of control rods are shown in Figure 5.1-3.

5.2 ZERO POWER PHYSICS TESTS After each refueling at the St. Lucie Units, startup physics tests are conducted to verify that the nuclear characteristics of the core are consistent with design predictions. While the reactor is maintained at hot zero power (HZP) conditions, the following physics parameters are measured; Critical Boron Concentrations, Moderator Temperature Coefficient, Control Rod Worth, and Differential boron worth

. 2.1 CRITICAL BORON CONCENTRATION Table 5.2-1 provides the comparisons between HZP critical boron concentrations measurements and predictions for Cycles 10, 11, and 12.

The values represent all rods out (ARO) and reference bank in conditions. As shown, excellent agreement is demonstrated for each case with all differences well within the +60 ppm review criteria.

5.2.2 MODERATOR TEMPERATURE COEFFICIENT Table 5.2-2 provides the comparisons between HZP Moderator Temperature Coefficient measurements and predictions for Cycles 10, 11, and 12. Again, excellent agreement is demonstrated with all differences being well within the review criteria of +2 pcml'F.

5.2.3 CONTROL ROD WORTH Table 5.2-3 provides the Control Rod Worth comparisons between measurement and prediction for Cycles 10, 11, and 12. In all cases, the agreement is within criteria with exceptional agreement being achieved for Cycles 11 and 12. Figures 5.2-1, 6.2-2 and 5.2-3 show the integral rod worth comparisons for the Reference Bank. The predicted rod worth and integral worth were calculated at the exact conditions which were present during the measurement. Excellent agreement is observed between measured and predicted integral worth.

5.2.4 DIFFERENTIAL BORON WORTH Table 5.2P provides the Differential boron worth comparisons between measurement and predictions for Cycles 10, 11, and 12. Both the measured and predicted values are obtained using the worth of the Reference Bank in pcm divided by the change in boron concentration from ARO to Reference Bank inserted. All differences are well within the expected performance.

6.3 POWER OPERATION 6.3.1 BORON LETDOWN CURVES Reactor coolant system boron concentrations are measured daily at the plant. Critical boron concentrations measured at or very close to hot full power all rods out equilibrium xenon and samarium conditions are compared to the predicted boron letdown curves for Cycles 10, 11, and 12 in Figures 6.3-1, 6.3-2 and 5.3-3. The predicted curves were obtained from design depletions with the three-dimensional ANC model. Table 6.3-1 shows the difference in ppm between measurement and ANC at various cycle exposures. The mean difference between measured and predicted critical boron concentration for all three cycles is 3 ppm with a standard deviation of 15 ppm.

5.3.2 AXIALPOWER D)STRIBUTIONS Measured core average axial power distributions from Beginning-of-Cycle (BOC), Middle-of-Cycle (MOC) and End-of-Cycle (EOC) obtained with the incore monitoring code INPAX (Reference 13) using incore detector "snapshots" were compared to predicted axial distributions in Figures 6.3% through 5.3-12. The predicted distributions were obtairied from three-dimensional ANC calculations performed for core conditions similar to those at the time of the "snapshots". Overall, the comparisons show excellent agreement between measured and predicted axial power distributions.

5.4

SUMMARY

ln this section, predictions made using Westinghouse's reload core design methodology are compared to zero power physics test measurements and at power operating data from St. Lucie Unit 1, Cycles 10,11, and 12. In all cases, the predictions agree very well with the measurements. The excellent agreement between the predictions and the measurements reported here demonstrates FPL's capability to apply the Westinghouse licensed methodology to perform reload core design for the St. Lucie Units.

-7?-

TABLE 5.2-1 ST. LVCZE UNZT 1 CYCLE 10,11 AND 12 HZP CRZTZCAL BORON CONCENTRATZON COMPARZSON BETWEEN MEASUE%2KENT AND PREDZCTZON CYCLE BANK CRITICAL BORON CONCENTRATION (PPM)

CONFIGURATION MEASURED PREDICTED DIFFERENCE M (M-P) 10 ARO 1598 1609 -11 10 BANK A in 1477 1496 -19 ARO,. 1393 1396 -3 in -2 0"

BANK A 1279 1281 ARO 1419 1427 BANK A in 1303 1307 Acceptance Criteria is +50 ppm

TABLE 5.2-2 ST. LUCXE UNZT 1 CYCLE 10,3.1 AND 3.2 HZP MODERATOR TEMPERATURE COEFFXCXENT CDNPIIRESDN BRPIIERN RPBSDRESSSR IIRD PREDSCS'EDN CYCLE BANK MODERATOR TEMPERATURE COEFFXCXENT (PCM/ F)

CONFXGURATXON

'EASURED PREDXCTED DXFFERZKCE M P (M-P) 10 4.41 5.70 -1.25

2. 56 2.57 -0.01 1.54 2.18 -0. 64 0 Acceptance Criteria is +2 pcm/'F

TABLE 5. 2-3 ST. LUCIE UNIT 1 CYCLE 10,11 AND 12 CONTROL ROD WORTH COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CONTROL ROD WORTH (PCM)

CONF XGURATXON MEASVRED PREDXCTED DXFFERENCE(8)

M P ( (M-P) /P) *100 10 BANK 7 516 4,80 7 ..50 BANK 6 367 404 -9.16 BANK 5 374 430 -13.02 BANK 4 584 631 -7.45 BANK 3 368 420 -12.38 BANK 2 789 848 -6.96 BANK 1 746 822 -9.25 BANK B 543 584 -7.02 BANK A(1) 1015 1014 0.10 TOTAL (2) 5302 5635 -5.91 BANK 7 590 522 13. 03 BANK 6 & B 715 774 -7 62

~

BANK 5 Ec 3 850 882 -3. 63 BANK 4 738 699 5.58 BANK 2 791 795 -0.50 BANK 1 808 806 0.25 BANK A(1) 1136 1106 2.71 TOTAL (2) 5628 5583 0.81 12 BANK 7 654 573 14. 14 BANK 6 Ec 3 929 915 1.53 BANK 5 Sc B 509 550 -7.45 BANK 4 824 809 1.85 BANK 2 699 740 -5.54 BANK 1 759 771 -1.56 BANK A(1) 1099 1136 -3.26 TOTAL (2) 5473 5495 -0.40 Acceptance Criteria is ~15: or 100 pcm which ever is greater (1) Reference Bank - Acceptance Criteria is +10.

(2) Sum of all measured banks within ~10%

0 TABLE 5.2-4 ST. LUCIE UHIT 1 CYCLE 10,11 AND 12 HZP DIFFERENTIAL BORON NORTH COMPARISON BETWEEN MEASUEKKENT'AND PREDICTION CYCLE BANK DXFFERENTXAL BORON WORTH (PCM/PPM)

CONF XGURATXON MEASVRED PREDXCTED DXFFERENCE (0)

M P ( (M-P) /P) *100 10 Average Over Bank A insertion 8.39 8.97 -6.50 Average Over Bank A insertion 9.96 9.62 3.53 12 Average Over Bank A insertion 9.47 9.47 0.00 TABLE 5.3-1 ST. LUCXE UNIT 1 CYCLE 10,11 AND 12 BORON LETDOWN COMPARXSON BETWEEN MEASUR229"NT AND PREDZCTXON CYCLE'YCLE BURNUP CRITICAL BORON CONCENTRATION (PPM) mn/MTU MEASURED PREDICTED DIFFERENCE M P (M-P) 10 139 1160 1178 -18 278 1130 1162 -32 696 1090 1117 -27 1392 1050 1074 -24 2784 990 987 3 4176 915 910 5 5568 845 840 5 6960 780 772 8352 9744 720 660 709 633 ll 27 8

11136 560 545 15 12528 440 438 2 13920 320 314 6 15312 190 187 3 15947 129 129 0 13.6 952 968 -12 278 939 953 -14 679 903 909 -6 1359 857 862 -3 2718 787 768 19 4078 687 681 6 5437 611 599 12 6796 528 517 11 8155 455 440 15 19514 372 359 13 10874 291 276 15 12233 195 176 19 13592 82 63 19 14404 13 -7 20 Acceptance Criteria is g50 ppm

TABLE 5. 3-1 (CONTZNUED)

ST.- LUCXE UNXT 1 CYCLE 10,11 AND 12 BORON LETDOWN COMPARXSON BETWEEN MEASUREMENT AND PREDXCTXON CYCLE CYCLE BURNUP CRITICAL BORON CONCENTRATION (PPM) mWV/MTU MEASURED PREDICTED DIFFERENCE M P (M-P) 12 132 961 991 -30 265 940 976 -36 661 920 934 -14 1324 892 892 0 2648 805 807 2 3972 743 729 14 5296 672 658 14 6614 596 589 7 7944 537 525 12 9268 443 453 10 10592 390 369 21 0 11916 13240 13723 287 175 131 273 164 122 14 11 9

Acceptance Criteria is +50 ppm

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P,XGURE 5.3-4 ST LUCXE UNIT 1'YCZE 10 AXIAL POWER DXSTRXBUTXON COMPARISON BETWEEN XNPAX AND ANC 2.00 1.75 1.25 g 1.00 N

g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEXGHT~ XNCHES BUEQgUP 372 MWD/MTU POWER LEVEL 1 00~o

-'93-

FIGURE 5.3-5 ST LUCXE UNIT 2. CXCLE 10 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AHD ANC 2.00 1.75 1.50 0g 1.25 Pg 1.00 N

g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEIGHT~ XNCHES BURHUP 6 g 9 04 MHD/MTU POWER LXMKt 1 00 o FIGURE 5.3-6 ST LUCIE UNIT 1 CYCLE 10 AXXAL POWER DISTRIBUTION COMPARISON BETWEEN XNPAX AND ANC 2.00 1.75 1.50 1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHTg INCHES BURNUP=15, 718 MWD/MTU POWER L&lEL=100>

PXGURE 5.3-7 ST LUCXE UNXT 1 CYCLE 11 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 2 00 1.75 1 50 0g 1.25

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g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOH TOP AXXAL HEXGHTi XNCHES BUEMJP=185 MWD/MTU POWER LEDS L=100i FIGURE 5 3-8 ST LUCXE UNXT 1 CYCLE 11 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 2 00 1.75 1 ~ 50 g

0 1.25 Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEIGHT, XNCHES BKKUP=6 721 MWD/KZU POWER LEVEL=100<

FICUS 5.3-9 ST LUCXE UNIT 1 CYCLE 11 AXIAL POWER DISTRIBUTION COMPARISON BETWEEN XNPAX AND ANC 2.00 1.75 4

Pg 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEXGHTi XNCHES BUEMJP=12, 188 MWD/MTU POWER LEVEL=100'98-

FXQURE 5-3-10 ST LUCXE UNXT 1 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 2 00 1.75 1.50 0g 1.25 Pq 1.00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP 2QCXAL HEIGHT, INCHES BUEQGJP 625 MWD/MTU POWER LEVEL=1 0 0 ~o FXGURE 5.3-11 ST LUCXE UNXT 1 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETNPNN XNPAX AND ANC 2 00 1 75 1.50 I 1.25 H

g 1-00 g 0.75 0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXXAL HEIGHTi INCHES BUR5KJP 6 g 620 MWD/HTU POWER LEVEL 1 00io 100

PXGURE 5. 3-12 ST LUCXE UNXT 1 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 2.00 1.75 ANC 1.50 O

1.25 d

Pg 1.00

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6.0 REFERENCES

1. Langford, F.L. and Nath, R.J., "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308-L, April 1969, and Spier, E.M. and Nguyen, T.G., "Update to WCAP-7308-L-P-A (Proprietary), Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.

2. Meyer, C.E. and Stover, R.L, "INCORE Power Distribution Determination in Westinghouse Pressurized, Water Reactors," WCAP-8498, July 1975.

3. Nguyen, T.Q., et al, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A (Proprietary), June 1988.

Miller, R.W., et al, "Relaxation of Constant Axial Offset Control/FQ 0 , Surveillance Technical Specification," WCAP-10216-P,-A (Proprietary),

June 1983.

5. Bordelon, F.M., et al, "Westinghouse Reload Safety Evaluation Methodology," WCAP-9272-P-A (Proprietary), July 1985.

6. Camden, T.M., et al, "Rod Bank Worth Measurements Utilizing Bank P

Exchange," WCAP-9863-A (Proprietary), May 1982.

7. Camden, T.M., et al, "PALADON-Westinghouse Nodal Computer Program," WCAP-9485 (Proprietary) and WCAP 9486, December 1978 and Supplement 1, WCAP-9485-A (Proprietary) and WCAP-9486-A (Non-Proprietary), September 1981 ~

Liu, Y.S., et al, "ANC: A Westinghouse Advanced Nodal Computer

~

8. ~

Code," WCAP-10965-P-A (Proprietary), December 1985.

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9. Poncelet, C.G., et al, "LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS," WCAP-6073, April 1966.

10. Olhoeft, J.E., "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962.

Harris, A.J., et al, "A Description of the Nuclear Design Analysis Programs for Boiling Water Reactors," WCAP-10106-P-A (Proprietary),

June 1982.

12. Barry, R.F., et. al, "The PANDA Code," WCAP-7048-P-A (Proprietary) and WCAP-7757-A, January 1975.

13. Correll, G.R., et al, "INPAX-II: A Reactor Power Distribution Monitoring Code," Exxon Nuclear Company, XN-NF-83-09(p), March 1983.

~ ~ ~

14. Morita, T., et al, "Power Distribution Control and Load Following

~

~

Procedures - Topical Report," WCAP-8385, September 1974.

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APPENDlX A This section describes the primary Westinghouse computer programs used by FPL to perform the required reload core design calculations for Turkey Point and St. Lucie. These codes are used in a manner similar to that outlined in Section 3 of Westinghouse's licensed reload methodology topical report (Reference 6).

Although the codes described in this appendix are not specifically addressed in the topical, two of the codes, FIGHTH and APOLLO (Reference 12), contain the same basic methodology as the licensed versions. The updated code versions include engineering enhancements (e.g., editing improvements, minor modeling improvements, and larger problem size capabilities) relative to the original code versions. The updated code versions were described at a meeting between the NRC Core Performance Branch and Westinghouse's Nuclear Fuel Division at the October 1984, at which time the differences between the original and updated code versions were discussed. The NRC concurred that the updated code O versions were essentially the same as the original versions, employing the same fundamental solution algorithms as the original versions.

The two major remaining codes, PHOENIX-P and ANC incorporate significant improvements to the methodologies discussed at the 1984 Westinghouse/NRC meeting. PHOENIX-P is a two-dimensional multigroup lattice code which does not rely on the spatial/spectral interaction assumptions inherent in the previous methodology. ANC is an advanced version of the PALADON code (Reference 7) incorporating nonlinear nodal expansion, equivalence theory (for cross section homogenization), and a pin power recovery model. The topical reports (References 3 and 8) qualifying PHOENIX-P and ANC for use in reload core design have been approved by the NRC.

A.1 FIGHTH The FIGHTH code computes effective temperatures in low enriched, sintered UO, fuel rods for specified values of burnup, linear heat

generation rate, moderator temperature, and flow rate. Resulting fuel and clad temperatures are used as input for the PHOENIX-P code. FIGHTH accounts for the radial variation of the heat generation rate, thermal conductivity, and thermal expansion in the fuel pellet; elastic deflection in the cladding; and pellet-clad gap conductance. The pellet-gap conductance is dependent upon the type of initial fill gas, the hot open gap dimensions, and the fraction of the pellet circumference over which the gap is effectively closed due to pellet cracking. References 9 and 10 provide a description of the basis of the FIGHTH program.

PHOENIX-P PHOENIX-P is a two-dimensional multigroup transport theory code used to calculate lattice physics parameters for PWR core modeling. In PHOENIX-P, the detailed spatial flux and energy distribution solution is divided into two major steps. In step one, a two-dimensional fine energy group nodal solution which couples individual subcell regions (pellet, clad, and moderator) as well as surrounding pins, is obtained. PHOENIX-P uses a Carlvik's collision probability approach and heterogeneous response fluxes to preserve the heterogeneity of the pin cells and their surroundings. The nodal solution provides a detailed and accurate local flux distribution. This distribution is then used to spatially homogenize the pin cells into fewer groups.

ln the second step of the solution process, PHOENIX-P solves for the angular flux distribution using a standard S'discrete ordinates calculation.

This technique utilizes group-collapsed and homogenized cross sections obtained from the first step of the solution. The S" fluxes are then utilized to normalize the detailed spatial and energy nodal fluxes. These normalized nodal fiuxes are used to compute the reaction rates and power distributions used to deplete the fuel and burnable absorbers. A standard B1 calculation is used to evaluate the critical spectrum of the fundamental

-105-

0 mode and to provide an improved fast diffusion coefficient for the core spatial codes.

PHOENIX-P employs a 42 energy group library which has been derived primarily from ENDF/B-V files. The PHOENIX-P cross section library was designed to correctly capture integral properties of the multi-group data during the group collapse, in order to properly model significant resonance parameters. The library contains all the neutronic data necessary for modeling fuel, fission products, cladding and structural, coolant, and control/burnable absorber materials present in most PWRs.

A detailed discussion of the methodology and models incorporated in PHOENIX-P may be found in References 3 and 11.

ANC ANC is an advanced multidimensional nodal methods program used to predict core reactivity parameters, power distributions, detector thimble fluxes, and other important core characteristics. ANC uses the nodal expansion method to solve the two-group diffusion equations. Partial currents and average neutron fluxes for the nodes are determined from continuous homogeneous neutron flux profiles by employing fourth order polynomial expansions for each of the x, y, and z directions across the node. Discontinuity factors are used to adjust the homogeneous cross-sections in order to preserve the nodal surface fluxes and currents that would be obtained from an equivalent heterogeneous model. In addition, ANC contains a pin-power recovery algorithm which couples the analytic solution of the two-group diffusion equations with the pin power information from PHOENIX-P. ANC is able to accurately reconstruct the results of fine mesh models using these methods. A detailed description of the methodology employed in ANC is contained in Reference 8.

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ANC is capable of performing either two or three-dimensional calculations with a wide variety of options. The code can handle geometries ranging from octant to full core and supports various symmetries. Feedback mechanisms make adjustments to the macroscopic cross sections to account for any changes in fuel temperature or moderator density. Xenon and samarium buildup and decay are modeled in addition to fuel and burnable absorber depletion. Typical applications of ANC include:

~ Differential and integral control rod worth,

~ Axial and radial power distributions,

~ Reactivity coefficients,

~ Critical core configurations,

~ Shutdown margins, and

~ Fuel and burnable absorber loading patterns.

A.4 APOLLO APOLLO is based on a one-dimensional two-group algorithm utilizing steady state diffusion theory solved via the finite difference method.

Normally, an APOLLO model is generated by radially homogenizing a three-dimensional ANC model. APOLLO is an advanced version of the PANDA code, described in Reference 12. Cross sections are flux and volume weighted over each mesh interval and a burnup and elevation dependent radial buckling search is performed to normalize the APOLLO model to ANC. APOLLO is used for applications which require a finer mesh in the axial direction than ANC, as a relatively high number of mesh points are available. Applications typically include:

~ Axial power distributions, including F~ synthesis,

~ Differential and integral control rod worth, Trip reactivity curves,

~ Load follow evaluations, and

~ Control rod insertion limits.

-107-

The algorithms used in APOLLO account for space dependent feedback effects due to xenon, samarium, rod position, boron, fuel temperature, and water density.

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