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: 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 pgppp2+p q +5pii7 pgR ADOCK 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

1.0 INTRODUCTION

AND CONCLUSIONS 1.1 OBJECTIVE

1.2 BACKGROUND

.3 SCOPE

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

1.4 CONCLUSION

S PAGE 2.1 2.2 2.3 2.4 2.5 CROSS SECTION LIBRARY........

~.. ~.....

LATTICE MODELING IN PHOENIX-P 2.2.1 FUEL CELL MODEL... ~..........

2.2.2 DISCRETE ABSORBER MODEL.........

2.2.3 STRUCTURAL CELL MODEL BAFFLE-REFLECTOR MODELING THREE-DIMENSIONAL NODAL MODEL........

ONE-DIMENSIONAL DIFFUSION THEORY MODEL 2.0 PHYSICS METHODOLOGY.....

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

~

~

~

~

~

~

~

~

~

~

~

~

~

~

5 5

6 7

7 9

9 9

10 3.0 PHYSICS MODEL APPLICATIONS.......

~ - ~..................

11 3.1 3.2 3.3 3.4 CORE POWER DISTRIBUTIONS AT STEADY STATE CONDITIONS 3.1.1 POWER DISTRIBUTIONS 3.1.2 POWER PEAKING 3.1.3 FUEL DEPLETION AXIALPOWER DISTRIBUTION CONTROL LIMITS..........

CORE REACTIVITY PARAMETERS 3.3.1 MODERATOR TEMPERATURE COEFFICIENT 3.3.2 DOPPLER COEFFICIENTS 3.3.3 TOTAL POWER COEFFICIENT 3.3.4 ISOTHERMAL TEMPERATURE COEFFICIENT 3.3.5 BORON REACTIVITY COEFFICIENT................

3.3.6 XENON AND SAMARIUM WORTH 3.3.7 CONTROL ROD WORTH 3.3.8 NEUTRON KINETICS PARAMETERS CORE PHYSICS PARAMETERS FOR TRANSIENT ANALYSIS NPUT

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

11 11 12 12 13 14 15 15 16 17 17 18 18.

19 20 4.1 4.2 CYCLE DESCRIPTIONS ZERO POWER PHYSICS TESTS 4.2.1 CRITICAL BORON CONCENTRATIONS 4.2.2 TEMPERATURE COEFFICIENTS 4.2.3 CONTROL ROD WORTH 4.2.4 DIFFERENTIAL BORON WORTH 4.0 PHYSICS MODEL VERIFICATION TURKEY POINT UNITS

~

~

~

~

~

~

~

~

~

~

~

~

~

~

21 21 23 24 24 24 25

TABLE OF CONTENTS (CONTINUED)

SECTION PAGE 4.3 4.4 POWER OPERATION o

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

4.3.1 BORON LETDOWN CURVES 4.3.2 POWER PEAKING FACTORS 4.3.3 RADIAL POWER DISTRIBUTIONS 4.3.4 AXIALPOWER DISTRIBUTIONS AND AXIALOFFSETS S UMMARY

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

26 26 27 28 28 29 5.0 PHYSICS MODEL VERIFICATION ST. LUCIE UNITS 5.1 CYCLE DESCRIPTION 5.2 ZERO POWER PHYSICS TESTS 5.2.1 CRITICAL BORON CONCENTRATION 5.2.2 MODERATOR TEMPERATURE COEFFICIENT 5.2.3 CONTROL ROD WORTH 5.2.4 DIFFERENTIAL BORON WORTH 73 73 74 75 75 75 75 5.3 POWER OPERATION................................

76 5.3.1 BORON LETDOWN CURVES.................

76 5.3.2 AXIALPOWER DISTRIBUTIONS................

~..

76 5..4

SUMMARY

76 6..0 REFERENCES 102 APPENDIX A.1 A.2 A.3 A.4 A

WESTINGHOUSE FIGHTH PHOENIX-P.......

ANC

~

~

~

~

~

~

~

~

~

~

~

~

APOLLO.........

COMPUTER CODES

~

~

~

~

~

~

~

~

~

~

1 04

~

~

~

~

~

~

~

~

~

~

1 04

~

~

~

~

~

~

~

~

~

~

1 05

~

~

~

~

~

~

~

~

~

~

~

1 06

~

~

~

~

~

~

~

~

~

~

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 Review 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 e

4.2-5 Turkey Point Unit 4 Control Rod Worth Comparison Between Measurement and Prediction for Cycles 12, 13, and 14......

Turkey Point Unit 4 HZP Differential Boron Worth Comparison Between Measurement and Prediction for Cycles 12, 13, and 14................

~......

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

34

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 y and l2 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 L

d oading Pattern.......................................

40 4.1-2 Turkey Point Unit 4 Cycle 13 Core L

d I

oading Pattern......................................

41 4.'I-3 Turkey Point Unit 4 Cycle 14 Core L

d oading Pattern.......................................

42 4.2-1 Turkey Point Unit 4 Cycle 12 Measured versus Predicted Control Bank C Integral Rod 0rth

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

W

~

~

~

~

~

~

~

43 4.2-2 4.2-3 Turkey Point Unit 4 Cycle 13 Measured versus Predicted Control Bank A Integral Rod Worth o

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 ANC -2320 MWD/MTU................

and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

55 4.3-11 4.3-12 Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE ANC - 6975 MWD/MTU Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE ANC - 118'12 MWD/MTU and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

56 and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 9458 MWD/INTU......................

~

~

~

~

~

~

~

~

64

~

~

~

~

~

~

~

~

~

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 ANC - 2440 MWD/MTU and 67 Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE ANC - 6678 MWD/MTU Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE ANC - 12316 MWD/MTU and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

68 and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

69 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Gomparison Between INGORE ANC - 600 MWD/MTU.................

Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE ANC - 6836 MWD/MTU Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE ANC - 10704 MWD/MTU and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

70 and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

71 and

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

72

LIST OF FIGURES (CONTINUED)

FIGURE PAGE St. Lucie Unit 1 Cycle 10 Core L

d oading Pattern e

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

84 St. Lucie Unit I Cycle 11 Core Loading Pattern.......... ~.........

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

85 St. Lucie Unit 1 Cycle 12 Core L

d oading 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 orth

~

~

~

~ 's

~

~

~

~

~

~

~

~

~

~

~

W

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 and 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 LIST OF FIGURES (CONTINUED)

FIGURE 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 5.3-10 St. Lucie Unit 1 Cycle 12 Axial Power

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

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 ofthe decision

process,

~

Improved optimization ofthe 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:

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

Atraining 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 forthe 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.

Allmethods 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 ofthe 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 forTurkey 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 forunit 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 ofcurrent 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 forthe two BAtypes 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 ofthe 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 ifit 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 of84C 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.

0 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 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 forANC 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 offuel 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 forall possible combinations ofxenon

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 ifeither 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 LimitReport 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 ofthe corresponding reactivity coefficient between two reactor statepoints with all other variables remaining constant.

Xenon, samarium, and control rod worth are also typically required to fullydefine 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 ofthe 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 ofthe 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 ofburnable absorbers.

The boron reactivity coefficient is calculated using the ANC core model by perturbing the boron concentration in both directions about a reference Q

value and computing the reactivity change.

Boron worths are calculated as a function of boron concentration, power level, temperature,

burnup, and control r'od configuration.

3.3.6 XENON AND SAMARlUIIWORTH 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 fullpower and hot zero power conditions forvarious 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 offissions 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 ofthe predictive capabilities ofthis methodology; however, their primary purpose is to demonstrate FPL's ability to e

perform design calculations for the Turkey Point Units 3 & 4.

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 fullpower 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 ofthese re-inserted assemblies was necessary to account fortheir 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. Alldifferences 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. Alldifferences 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.

4.2.3 CONTROL ROD WORTH Control rod worths were measured by the Rod Swap Technique.

First, the t

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 ofboron 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 fluxshapes 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 fullpower 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 RADIALPOWER 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.

Allcomparisons 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 forcore 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.

Allstartup 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 9B 11B 12A 12B 12C 13A 13B 13C 13D 13E 14A 14B 14C 14D 9

9B 11A 13B 13C 13D 13E 14A 14B 14C 14D 15A 15B 15C 15D 13E 14A 14B 14C 14D 15A 15B 15C 15D 16A 16B 16C 16D 16E HUMBER OP ASSEMBLIES 1

8 8

12 8

4 4

4 24 32 28 4

8 12 8

1 8

4 4

8 24 28 4

8 12 16 12 16 8

25 4

8 12 16 12 16 16 16 4

8 8

INITIAL ENRICEQG&T w/o U-235 3.40 3.40 2.60 3.45 3.00 3.00 3.10 3.10 3.20 3.40 3.40 3.40 3.80 3.80 3.30 3.40 3.10

3. 10

3. 10 3.20 3.40 3.40 3.40 3.80 3.80 3.60 3.60 4.00 4.00 3.40 3.40 3.40 3.80 3.80 3.60 3.60 4.00 4.00 3.60 3.60 3.60 4.00 4.00 BOC BURNUP mo/mv 16080 29279 22862 28354 24975 18285 17489 17592 18769 15354 0

0 0

0 27419 29845 26325 28073 31944 32654 27752 16549 16232 13179 13952 0

0 0

0 32717 31222 31768 24027 30454 17883 17489 15004 15055 0

0 0

0 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:

Moderator Temperature Coefficient Isothermal Temperature Coefficient

+2. 0 pcm/'F Control Rod Bank Worths:

Reference Bank Worth "Swap" Worths Differential Boron Worth

~10'o i15: or 100 pcm whichever is greater

~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 CONFXGURATXON CRXTXCAL BORON CONCENTRATXON (PPM)

MEASURED PREDXCTED DXFFERENCE M

P (M-P) 12 12 13 13 0"

ARO BANK C in ARO BANK A in ARO BANK SB in 1538 1399 1554 1401 1698 1552 1584 1428 1560 1408 1691 1544

-46

-29

-6

-7 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 M

PRED1CTED DIFFERENCE P

(M-P) 12 0.92 0.58 0.34 13 0.24 0.06 0.18 0

0.26

1. 13

-0.87 CYCLE BANK ISOTHERMAL TEMPERATURE COEFFICIENT (PCM/ F)

CONFIGURATION 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 CONFZGURATZON COKZROL ROD WORTH (PCM)

MEASURED M

PREDZCTED D1FFERENCE(0)

P

( (M-P) /P) *100 12 13 BANK D BANK C(1)

BANK B BANK A BANK SB BANK SA TOTAL(2)

BANK D BANK C BANK B BANK A(1)

BANK SB BANK SA TOTAL(2)

BANK D BANK C BANK B BANK A BANK SB (1)

BANK SA TOTAL(2) 691 1314 375 1177 1180 1000 5737 641 1022 435 1232 1183 826 5338 636 1093 435 1086 1173 1052 5475 718 1365 380 1204 1202 1017 5886 682 992 457 1275 1233 836 5475 661 1172 480 1102 1195 1094 5704

-3. 76

-3. 74 1

~ 3 1

-2. 24

-1. 83

-1.67

-2.53

-6.09 2.97

-4. 88

-3.41

-4. 03 "1.17

-2.51

-3. 78

-6. 74

-9.37

-1.45'1.84

-3.84

-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 CONFIGURATION DIFFERENTIAL BORON WORTH (PCM/PPM)

MEASURED PREDICTED DXPPERENCE (8)

M P

((M-P) /P) *100 12 13 Average Over Bank C insertion Average Over Bank A insertion 9.45 8.05 8.78 8.34 7.63

-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 MND/MTV CRXTXCAL BORON CONCENTRATXON (PPM)

PREDXCTED DIFFERENCE P

(M-P) 12 0-0 150 2000 2320 3000 4020 4940 5890 6975 8000 10000 11184 12000 12441 150 1000 2000 2440 3224 4888 6678 8265 8754 10608 12316 150 600 1000 1830 2521 3428 5000 5986 8148 8995 9871 10704 12000 1437 1111 1020 986 954 867 815 744 657 567 379 277 200 161 1082 1014 938 924 862 750 610 473 436 276 121 1213 1189 1135 1103 1056 991 869 792 601 520 453 362 218 1426 1124 1006 989 950 882 812 733 637 546 361 250 174 133 1104 1027 953 926 869 739 591 458 418 258 109 1212 1166 1142 1098 1043 980 858 775 587 510 428 349 226 11-

-13 14

-3

-15 3

11 20 21 18 27 26 28

-22

-13

-15

-2

-7 11 19 15 18 18 12 1

23-7 5

13 11 11 17 14 10 25 13-8

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 mo/MTU F~ (MAX)

MEASURED PREDZCTED DXFFERENCE M

P

( (M-P) /P) *100 0

12 13 150 4945 5860 6890 7620 8363 9082 9458 10323 11121 11812 150 244 0.

3224 4888 6678 8265 8754 10608 12316 600 1830 2521 3428 5986 6836 7659 8143 8995 9871 10704 1.475 1.492 1.499 1.498 1.493 1.502 1.511 1.509 1.511 1.504 1.508

1. 557

1. 438 1.442 1.447 1.519 1.509 1.511 1.541 1.545 1.468 1.462 1.485 1.465 1.471 1.472 1.478 1.476 1.482 1.494 1.496 1.416 1.456 1.459 1.456 1.471 1.481 1.486 1.489 1.487 1.484 1.479 1..486

1. 440

1. 437 1.435 1.455

1. 477 1.491 1.507 1.508 1.420 1.421 1.427 1.427

1. 440 1.452 1.460

1. 4'64

l. 464 1.461 1.462 4.17 2.47 2.74 2.88 1.51 1.41 1.70 1.35 1.60 1.35 2.01 4.77

-0.14 0.35 0.84 4.39 2.16 1.34 2.26 2.45 3.38 2.89 4.06

'2. 66 2.15 1.38 1.23 0.82 1.23 2.26 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 mO/MTU (MAX)

PREDXCTED DXFFERENCE P

( (H-P) /P) +100 12 150 4945 5860 6890 7620 8363 9082 9458 10323 11121 11812 1.920 1.692 1.709 1

~ 718 1.722 1.709 1.699 1.724 1.713 1.692 1.688

l. 874

1. 632

1. 649 1.644 1.657 1.672 1.672 1.671 1.661 1.652 1.644 2.39 3.67 3.63 4.50 3.92 2.21 1.61 3.17 3.13 2.42 2.67 14 150 2440 3224 4888 6678 8265 8754 10608 12316 600 1830 2521 3428 5986 6836 7659 8143 8995 9871 10704 1.773 1.642 1.614 1.664 1.719 1.726 1.732 1.762 1.755 1.815 1.745 1.825 1.730 1.725 1.742 1.741 1.741 1.735 1.729 1.732 1.756 1.634 1.621 1.596 1.617 1.650 1.669 1.681 1.671 1.682 1.677 1.686 1.672 1.681 1.688 1.699 1.702 1.701 1.701 1.696 0.97 0.49

-0.43 4.26 6.31 4.60 3.77 4.82 5.02 7.91 4.06 8.24 3.42 2.62 3.20 2.47 2.29 2.00 1.65 2.12 4.3-4 T&&EY POXNT UNZT 4 CYCLE 12,13 AND 14 AXIAL OFFSET COMPARXSON MMZllitlltlMMRROIIMMMMZlIMO ZRMOZCI'ZOM CYCLE CYCLE BURNUP mWD/MTV AZXAL OFFSET (8)

PREDXCTED P

DIFFERENCE M-P 12 13 14 150 4945 5860 6890 7620 8363 9082 9458 10323 11121 11812 150 2440 3224 4888 6678 8265 8754 10608 12316 600 1830 2521 3428 5986 6836 7659 8143 8995 9871 10704

-2.46

-2.55

-2.99

-3.05

-1.60

-1.42

-0.87

-1.66

-3.09

-2.04

-2.23 F 00 2.34 0.93

-0.41

-0.66

-1.51

-1.99

-1.99

-1.61 3.62 1.19

-1'0

-0.13

-1.36

-1.64

-1.93

-2.08

-2.10

-1.70 2 ~ 12

-2.57

-1.85

-1.96

-2.13

-2.28

-2.24 2 ~ 11

-2.06

-1.96

-1.93

-1.93 0..34

-1.42

-1.79

-2.25

-2.45

-2.59

-2.50 2

~ 2 3

-1.86 1.85 0.56 0.17

-0.60

-1.88

-2.08

-2.20 2

~ 1 7

-2. 09

-1.93

-1.82

-0.11

-0.70

-1.03

-0.92 0.68 0.82 1.24 0.40 1

~ 1 3

-0.11

-0.30 6.66 3.76 2.72 1.84 1.79 1.08 0.51 0.24 0.25 1.77 0.63

-1.47 0.47 0.52 0.44 0.27 0.09

-0.01 0.23

-0.30

~

0

~

~

~ l

~

~

~

~

~

~

~

I ~

~

~

~

~ I

~

~

~

~ I w

FIGURE 4.1-2 TURKEY POINT UNIT4, CYCLE 13 LOADINGPATTERN 48 ASSEMBLYFEED 6

5 4-2 9B D

15B 13E 8 WABA SB 15A 13B 0 WABA 0

15A 8 WABA

'j3C 20 HF 15B 8 WABA 13E 14D 14A A

15A 6 WABA 9

SA 15C 13D 20 HF 13E SB 11A C

15B 4 WABA 15A 0 WABA 14A 14A A

14D SB 15C 8 WABA 13E 13B 0

15A 11 A 6 WABA C

14A 15D LEGEND 15A 8 WABA 9

SA 15B 4 WABA 15C 8 WABA RCCA Bank or Removable BP HF a PTS Hafnllkn Ateocber R a Re4naerted Aeeocnbly 13C 20 HF 15C 14C B

'f3D 20 HF 13E 15A-3.6 w/o No IFBA 15B -3.6 w/o 32 IFBA 15C-4.0wlo No IFBA 150 -4.0w/o 88 IFBA 28 Assemblies 3.6 Wt.%

20 Assemblies O 4.0 Wt.%

FIGURE 4.1N TURKEY POINT UNIT4, CYCLE 14 LOADINGPATTERN 52 ASSEMBLYFEED 6

5.

4-2.

H 14A D

158 SB

'I6B WABA 14A D

'f68 14D 4 WABA 14B 20 HF' E

15A 15B SB 16B 8 WABA 14C A

16A 6 WABA 16C 15A 8 WABA 15A 14A 14C A

16A 6 WABA 15A SB 15C 16A 14D 16 WABA C

15B SA 16B 15C 4 WABA B

16E 13E 14A 20 HF 14A D

16A 14D 6 WABA C

15C 15D LEGEND 16B 4 WABA 158 SA 16B 4 WABA 16E 14A 14O 14B 20 HF 16D 15C B

14A 20 HF 13E 16A - 3.6 w/o No IFBA 16B -3.6 w/o 32 IFBA 16C -3.6 w/o 64 IFBA 16D-4.0w/o 16 IFBA 16E -4.0 w/o 48 IFBA 36 Assemblies @3.6 Wt.%

16 Assemblies 4.0 Wt.%

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

g 900 D4 soo A

g 700 600 500 400 300 200 100 0

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

1400 FXGURE 4.2-2 TURKEY POXNT UNXT 4 CYCLE 13 MEASURED V1MSU8 PREDXCTED BANK A XNTEGRAL ROD WORTH 1300 1200 1100 MKR,SURED PREDICTED 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) 1400 FXGURE 4 2<<3 TURKEY POXNT UNXT 4 CYCIsZ 3 4 EEIIEPEEP VEIIPEP PEEPECPE EQLSK SB XNTEGRAL ROD WORTH 1300 1200 1100 MKR.SWED PREDXCTED 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) 1600 FIGURAL 4-3-1

!VUXKEY POINT UNXT 4 CYCLE 12 BORON LETDOWN COMPARISON BETWEEN AND PREDICTXON 1SOO 1400 1300

> 1200 g0H1100 g

g

~ 1000 900 0O g

800 0

0 700 600 500 MFJLBU&~23 PREDICTED 400 300 200 100 0

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

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

0 700 600 H

500 MEASURED PREDXCTED 400 300 200 100 2000 4QOO 600Q 800Q 1QOOO

12000, 14000 CORE AVERAGE BURRKJPg MWD/MTU 0

1600 FXGURE 4 3-3 TURKEY POXNT UNXT 4 CYC~ 14 BORON LETDOWN COMPARXSON BETWEEN AND PREDXCTXON 1500 1400 1300 PI 1200 8O 1100

~~1000 g

800 0

0 700 600 500 PREDXCTED 400 300 200 100 e

0 0

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

2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURHUPi MWD/MTU 5Q e

2.0 FXGUE& 4 3-6 mnmxm POXNT amXT 4 CYCLE 14 F DELTA H COMPARXSON BETWEEN XNCORE AND AMC 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-

2 50 FXCRGtE 4 3-9 TUESDAY POXNT UNXT 4 CYCLE 14 FQ COMPARXSON BETWEEN XNCORE AHD ANC 2.25 XNCORE 2.00 1.25 1.00 0

2000 4000 6000 8000 10000 12000 CORE AVEEVRGE BURWJPg MWD/MTU I.

0

~ ~

~ ~

o

~ ~

~ ~

0>>

~

~

~

~

~

~

~

a

~

a

~

~

I

~ ~

0 I 'I

~

~

~

~t I

I

~

~ " '

I >>

~

~

~

I ail I ~

~ '.

~

~

I>>>> I I

~ ~

~ Vg I>> ~RRI555 I >>I

~

I I

~

~

I

~ '.

~

~

I I>>'.

>>I >>

I I

~

~

I

~

~ '

~

> ~

~

I>> ~

~

~

> ~

. Ill.

I

~ >>'.

~ ~

I ~

g ~

I I'.

I I ~il>>

I

~

J ~

il>>

~ ~

~ >> ~

I I

~ '.

I

~

I

~

~

I I

~

~

I

~

I ~ I I

~

~ ~

I

~ '

~ '

~ ~

I

~ '.

~

~ '.

I ~ ~

I" ~

I

~

I '

I>> '.Rl

~ ~

I>> './ ~

~

I

~ ~ '.

~ ~ ~

~ >>>>

I ~

~ 4

~ ~

I ~ ~

. 'gg

/ ~

~ ~

~ ~

~ >>

~ I

~"

I

~

~

~ ~

I

~

~

I ~ ~

00

~ ~ ~

I

~ ~

~

~

0'

~ ~

I I

~

~ 0',

~ ~ '.

~ ' '.

I

~ I ~

I

~

~ ~

I" ~

I>> I

~ ~

~ ~

I ~ ~

I ~ I

~ ~

I

~

~

~

~ '>>

I

~ I '.

~ ~ '

~ '.

I

~ ~

~ >> ~

~ I

~ ~

~

~ I I>> ~

~

~

~ ~

I ~ >>

I>>

~ ~

I e ~ ~

I e

~

~

~

I ~

~ >>

I

~ ~

i

~

~ >>I

~ ~

I

~ '.

~

~ ~ '.

~ ~

~ ~ ~

~ '.

I

~ ~ '.

~ I '.

~ I

~ I I>>

I>> ~

I ~

I ~

I

~

~

~ '.

~

~

~ '.

I

~

I

~

~

I

~

~

I

~

I '

I ~ ~

I

~

~ ~

I

~

I >> ~ '.

I

~

I I

I I

~

I I

~

~

I ~ ~ '.

~

~ ~

~ 0=

~

o I I

~

I

~

~ ~ '.

0

~ ~

~

~ o>> ~

~ ~

~

~

~

~

~

~

~ ~

I ol

~

>>I I

~

'. ll~ '.

I I

~

~

~

~

I ~ ~

~ ~ ~

~

~

~

~

~ ~

I ~

I

~

I I

o ~ '.

I o ~ '.

I

~

~

I

~

~ '

I I'

~

~"

~ ~

~/

~

~

~

~

I ~

~

~ o I

~ ~

I

~ ~ o I

~ ~

I

~ ~ '.

I lo'.

I

~ ~

I

~

I

~

~

~"'.

I

~ ~ ~

I ~ ~

~ ~

~ ~

~ ~

~>>'

I

~

~ ~

~ ~

I>>o I

o ~

~ ~

~ ~

I I

I

~ I '.

I

~ ~ ~

I

~

~ o

~ '

I

~ ~

I>>

lo ~

~ ~

I I'.

~

o '.

I>>l

~V I

~

I ~ ~

~ ~ ~

~ ~

I ~ ~

~ ~

I I ~

I

~ ~

l>>o

~ ~

~ ~

lo>>

I I

~

~

I

~

o ~

~

~ o'.

I

~ ~

I ~

~ ~

I

~ ~ '.

I

~

I

~

~ Io'.

~ '

~ ~ '.

I>>

I ~

I ~ ~

I

~

o

~

~

oo

~ ~

~ '

~

I

~ ~ '

~

~>>o l>>o I

~

lo'

~ o I'

~

ohio I

~

~

I ~

~ ~

~

~

I

~ o'

~

~ ~ '

I

~ ~

I

~ ~

~ ~ '.

~

~ ~'.

I

~ '

~ ~

I ~ '.

~

~

~ ~

I ~

I o I ~

~

~

ol

~

~

I ~

I

~ ~

~

~

~ ~

I I

~ ~

I I ~ '

~

I

~

~>>I I>>'

~

I ~

~

~ o ~

~

~

~ ~

I o'.

~

I ~ '.

~ i '.

I

~ ~

~

s

=

=

I I

ol Ia

~

~ ~

~

~ -

~

~

o I ~ '.

~

o

~

~

g

~

FIGURE 4.3-12 TURKEYPOINT UNIT4, CYCLE 12 RADIALPOWER DISTRIBUTIONCOMPARISON BETWEEN INCORE ANDANC 15 14 13 12 10 9

8 7

6 0.397 0.411

-3.41%

OA92 OA95

%.61%

I.140 1.139 0.09%

0.683 1261 0.697 1.275

-2.01%

-1.10%

0.492 0.495

%.61%

1.175

.1.158 1A7%

1.162 1.150 1.04%

I.117 1.126 0267 0.325 G267 0266 G319 0266 0.3S%

1.88%

0.38%

OA10 0.704 1.058 Q847 1.038 0.410 0.697 1.049 Q849 1.049 0.00%

1.00%

0.86%

%24%

-1.05%

1.158 1.301 1.084 1296 1.075 1.139 1275 1.075 1.316 1.075 1.67%

204%

0.84%

-1.52%

0.00%

1.170 1.140

.1.370 1.048 1.369 1.150 1.126 1.377 1.033 1.377 1.74%

124%

%.51%

lA5% %.58%

1.108 1.383 1.115 1.147 1.112 1.123 1.385 1.1N 1.142 1.108

-1.34%

%.14%

0.54%

OA4%

0.36%

1.366 1.073 1.163 1.159 1.152 1.383 1.056 1.153

1. 145 1.152

-123%

1.61%

0.87%

122%

0.00%

G6S7 0.697

-1A3%

1Z63 lZ75

%.94%

1.138 1.126 1.07%

1.374 1.384

%.72%

1.05S 1.0M 0.19%

Q393 0.410 4.15%

1.111 1.139

-2A6%

1.165 1.150 1.30%

1 ~ 107 1.123

-1.42%

1247 1.383

-2.60%

0.474 0.495

<24%

1.135 1.15S

-1 99%

l.151 1.150 0.09%

1.126 1.126 000%

0.475 0.495 1.128 G398 1.139 OA10

%.97%

-2.93%

1Z77 Q691 1275 0.697 0.16%

%.86%

I I

I I

I P I'I I N I

I I M I

I I

I I

I 0265 0.266 0.327 0.319 2.51%

0.273 0.266 2.63%

1.038 1.079 1.049 1.075

-1.05%

0.37%

0.879 l.324 O.S50 1.315 3.41%

0.68%

1.077 1.101 1.049 1.075 2.67%

2.42%

0.695 1.281 0.697 1.275 C.29%

OA7%

0.392 1.120 0.411 1.139

<.62%

-1.67%

0.475 0.495

<.04%

1.373 1.377

%29%

1.060 1.033 2.61%

1.378 1.377 0.07%

1.129 1.126 0.27%

1.155 1.150 OA3%

1.137 1.158

-1.81%

1.120 1.158 l.383 1.079 1.347 l.143 1 ~ 108 1.151 1.361 1.071 1.361 1.151 1.0S%

0.61%

1.62%

0.75%

-1.03%

%.70%

1.156 1.155 1.089 1.147 1.09S 1.153 1.143 1.145 1.071 1.109 1.071 1.145 1.14%

O.S7%

1.6S%

3.43%

2.52%

0.70%

1.117 1.168 1.397 1.091 1.390 1.168 1.107 1.151 1.361 1.071 1.361 1.151 090%

1.48%

2.65%

1.87%

2.13%

1.48%

1.365 1.065 1.167 1.158 1.174 1.075 1.383 1.056 1.152 1.145 1.152 1.056

-1.30%

0.85%

1.30%

1.14%

1.91%

1.80%

1.105 1.353 1.107 1.160 1.122 1.359 1.123 1.384 1.108 1.141 1.10S 1.384

-160%

-2.24% 4.N%

1.67%

126%

-1.81%

1.142 1.114 1.365 1.073 1.361 1.113 1.150 1.126 1.377 1.032 1.376 1.126 4.70%

-1.07%

4.87%

3.97%

-1.09%

-1.15%

1.092 1.107

-1.36%

1.146 1.143 0.26%

1.118 1.107 0.99%

1.38S 1.383 1.104 1.123

-1.69%

1.145 1.150 C.43%

1.361 1.377

-1.16%

1.043 1.032 1.07%

1.362 1.377

-1.5%

1.130 1.126 0.36%

1.161 1.150 0.96%

1.146 1.157 1.085 1.046 0267 1.075 1.049 0266 0.93%

%29%

0.38%

1.325 0.846 0.324 1.315 0.849 0.319 0.76%

%.35%

1.57%

1.054 1.036 0267 1.075 1.049 0266

-1.95%

-1.24%

0.38%

1070 0.687 1275 0.697 N.39%

-1.43%

1.156 OA09 1.139 0.410 lA9% %24%

'OA84 0.495

-2.22%

H D

INCORE ANC 0.471 0.495 4.85%

1.113 1.254 1.083 1.342 1.055 1252 1.139 I.274 1.075 1.315 1.075 1274

-2.28%

-1.57%

0.74%

2.05%

-1.86%

-1.73%

0.391 0.679 1.061 0.853 l.044 0.674 0.410 0.697 1.048 0.849 1.048 0.697 4.63%

-2.58%

124%

0.47%

4.38%

-3.30%

1.119 1.138

-1.67%

0.394 0.410

-3.90%

0.475 0.495

<.04%

% DIFFERENCE Mean Absolute Difference Standard Deviation 0.013 0.009 0.269 0.327 0.269 0266 0.319 0266 1.13%

251%

1.13%

BURNUP = 11812 MWD/MTU POWER LEVEL= 1HP%%d D BANKAT228 STEPS

FIGURE 43-13 TURKEYPOINT UNIT4, CYCLE 13 RADIALPOWER DISTRIBUlIONCOMPARISON BETWEEN INCORE ANDANC 15 0257 G244 5.33%

G409 OA17

-1.92%

a799 0.815

-I 96%

1 ~ 133 1.147

-122%

13 Q452 OA48 O.89%

1.051 1.107

-5.0N 1263 1293

-2.32%

1.031 1.011 1.98%

12 OA57 OA47 224%

1.070 1.080

%.93%

l.139 1.106 2.98%

1.034 0.992 423%

I.393 1290 0.78%

OA24 OA16 1.92%

1.089 1.106

-1.54%

1.104 1.105

%.09%

1270 125B 0.95%

IZI6 1.188 2.36%

1237 1215 1.S1%

10 0.814 1.11%

1277 1292

-1.16%

1.014 G989 2.53%

1205 1.187 1.52%

1235 1222 1.06%

1.30 I 1298 023%

0251 0244 2.87%

1.119 1.146

-2.3N 1.015 1.010 0.50%

1282 1288 NA7%

1223 1214 0.74%

1.315 1.300 1.15%

I.M5 1.051 1.33%

0273 Q272 0.37%

0.907 0.928

-22N 1.326 1342

-1.19%

Q991 0.998

%.70%

1.310 1292 1.39%

1.072 1.051 200%

1.316 1.M9

-3.S7%

G251 G244 2.87%

1.1CO 1.146 4.01%

Q971 1.010

-3.8N 1265 1288

-1.79%

1233 1214 1.57%

1204 1.300 0.31%

1.044 1.051

%.67%

GSO&

OA31 G814 OA16

%.98%

3.61%

1252 1.109 1292 1.106

%.10%

Q27%

I.IXI3 1.115 0.989 1.105 1A2%

0.90%

1.197 IZ75 1.187 125S 0.84%

1.35%

1231 I203 1222 1.188 0.74%

126%

IM4 l~

129S 1015

%.31%

1.07%

OA47 3.5S%

1.0S6 1.080 0.56%

1.106 1.106 0.00%

1.008 G992 1.61%

1~1 1290

-225%

1.082 1.107

-226%

1257 1293

-2.78%

Q994 1.011

-1.68%

2 1

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I OAIS OA17 024%

GS04 Q815

-1.35%

1.117 G250 1,147 G244 J

-2.62%

2AN 272 2.94%

0257 Q244 5.33%

0.932 0.929 0.32%

I ~ l53 1.147 0.52%

0.833 0.815 221%

OA33 0.417 3.84%

1.33S 1.011 1.343 Q999 4.37%

129K 1.026 1288 1.011 12K lA8% %.16%

1292 1.024 12R3 0.992

%.08%

323%

1.112 1.128 1.107 1.106 0.45%

1.99%

OA72 1.103 OA48 1.080 5.36%

2.13%

1.303 1292 0.85%

1238 1215 1.89%

1214 1.188 2.19%

1284 1258 2.07%

1.11 I 1.105 0.54%

1.067 1.048 1.81%

1.309 1298 0.85%

1240 1222 lA7%

1203 1.187 1.35%

G977 0.989

-121%

1.324 1.363

-2.8N 1.04&

1.051 1.322 1.300 1.69%

1214 1214 Q00%

1251 1288

-2.87%

1.040 1.053

-123%

1.335 1.369

-2A8%

1.070 1.051 1.81%

1291 1292

%.08%

1.005 0.998 0.70%

1.319 1.363

-323%

1.048 1.051 1295 1.m

%.38%

1226 1214 0.99%

1297 1288 0.70%

I.o&2 iX&6 G966 1.048 1292 0.999 1.34%

-2.01%

-3.30%

129 I 1202 1254 1297 1215 1290 4A&% -1.07%

-2.79%

l229 1.198 0.989 1222 1.188 0.992 0.57%

0.84%

%.30%

1207 12Bb 1.105 1.187 1258 1.106 1.68%

223%

%.09%

1.018 1.103 1.0S7 0.9S9 1.105 1.079 2.93%

%.18%

0.74%

1.343 1243 Q00%

1.007 1.011 12&2 1293

-2A0%

1.085 1.107

-1 99%

Q459 OA48 2.46%

Q919 a272 Q929 G272 H

-1.08%

GOO%

1.125 G255 1 ~ 147 G244 G

-1.92%

4.51%

Q821 Q815 G74%

OA24 OA17 1.68%

D INCORE

'L DIFFERENCE OA71 OA47 5.37%

1.084 1.106

-I 99%

0.416 0.96%

1243 1292

-3.79%

aSo7 0.814 1.001 1.010 1.116 1.146

-2.&2%

1.332 1.342 4.75%

0.921 0.92S

%.75%

1.020 1.010 Q99%

1.149 1.146 026%

1272 1.077 OA58 1291 1.106 OA47

-1 A7% -2.62%

2A&%

0.814 OA15 0.814 OA16 Q00%

%24%

Mean Absolute Difference Standard Devlatlon Q015 Q012 0252 0244 328%

0281 G272 3.31%

0261 Q244 6.97%

BURNUP = 2440 MWD/MTU POWER LEVEL= 99,4'L D BANKAT215 STEPS

~ ~

~

>> ~

a

~

~ )

~

~

~

~ 0

~

Q

~

I

~

I I

~

I

~ wl f5aNRNNNER

~I'.

I

~

I

~ '

~ I>>

~ I>>

~

I I ~ I'.

I r I

I ~ >>

~ ~

I

~

I

~

~ >>I

~ I I

~

I >>'

It ~

I ~

I

~

I ~

I

~ '

I

~ 'I I i I

I I

. II~

. I,

I

~

~ '.

I >>I I

~

I I

~

~

~ ~

II~

~ I>>

I

~

I

~

~

~

I ~ II'

~

I

~

~ I iI>>

~ ~

I I'

~

~ ~I

~

~

I

~ '

I I

~ ~ '.

~ I>>

I

~

I I I'.

~ I It ~

~ II' I '.

I

~ I III

~

~

II ~

~

~ '.

I

~

~ ~ '.

I

~

~

~

I I

I ~

II>>

I

~ o ~

I ~ I'

~ '

~I r ~

~

~ I

~

~

I I>>

I

~

I I

~ ~ '

I

~

~ I' '.

I

~

~

~

~ ~'I.

I I

~ >>'

a

~ I

~

~

I

~ ~

~ ~

I

~ ~ I

~

~

~

I>> ~

~

>> ~

~ ~

I

~ >>'.

I ~ ' '.

~ ~

I/>>

~ ~

I

~ '.

~

I I'.

>> ~

~ ~

~ ~

I

~

I I

~ ~ ~

~ ~

I ~

I ~

I I '.

I I

~ '.

I I'.

I I>>'.

I I

~ ~

I >>; ~

I ~

~ I I

~ ~

I

~ I ~

I ~ ~

~

~ ~ ~

~ ~

~

IP ~

I ~

~

~ ~

~

I

~ '

~

~

~

~

~

~ ~ '.

I

>> ~ '.

I >>I>>

I

~

~ I I

~

I

~

I .

I

~ ~ '.

I

~ I I

~

~

= -

a I I o

~ e

~

~ C

~

e I

I I II I

~

~ I ~

I '.

~

~ I '.

~ ~

I ~

~ ~

~

s

~ ~

~

~

o

~

~

i s

~

~

~ ~

s

~

s I

~

I

~ '

I I>>

~

~

I

~

I

~ ~

I

~

~

I

~

I

~

I 'V I

~ ~ o I

~

III.

~ Io.

I

~

I ~ o

~

~

I ~

~ ~

I ~

~ >>

I

~ o I" ~

~

~

I

~ '

~

~

~

~ I

~ C ~

~ ' '.

I o I

~

I

~

~

~

~ ~

I I I'.

I

~

~ '

~ '.

I

~

~

I

~ ~

I ~

I ~

I ~ ~

I ~

I

~ ~ o I

~ ~

~

~ s

~ ~ '

~

~ ~

Io ~

~ ~

~/

I ~

I ~

~

~

I ~

I

~

~

I

~

~

I ~

Io'.

I

~ ',

I I

~

I

~ ~

I'

~ I

~ I

~ ~

I s

~

I

~

I o

~

~ ~

I

~

~ IV I ~

I

~

I

~

~ ~

I

~

I I

~ '

~ ~

I

~

s

~

~

~

I

~

~ ~ o

~

~ '.

~

o ~

~>> ~

~ C ~

~ ~

~ ~

~>>'.

I

~N

~ ~

I ~

I o

~

I ~ ~

~

~ ~

I o

~

~

~

II>>

I ~ ~

I

~

~ ~

I

~

~ ~ '

I

~ C ~

~ ~

~

o

~

I

~

I

~ ~

I ~

~ I I

~

~

~

~

I o ~

I

~

~

I

~

~ ~

I ~ o ~

I

~

~

~ >>

~ I IV I ~ V I>>

~ ~ ~

I ool

~ ~

lo I oV I

s ~

I s ~

I s'.

~

~ o I>>

I ~

I

~

o ~ '.

I

~

t>> ~

~

~

~ ~

I ~

~ ~

~

~

I

~ ~

I

~ '.

~ I ~ '.

I

~ .

I

~ '.

I'.

I

~

~

~

~

I '.

I

~

~ ~

~ ~

I s ~

~ '

~

s I

~ ~

I

~

I

~ ~

I o ~ '

~

lo I

~

I

~

I

~

~ V

~

~

~ >>

~ IV IIV

~ >>

~ '.

I

~

I

~

I

~

~ I '.

I Io ~

I o

~

I I

~

I

~ ~

I I ~

~ ~

I

~

I oV I

~

I

~ ~

I

~ ~ '

~

I

~

~ ~

I

~ ~

I ~

~

~ ~

~ ~

~

~

I I/

~ ~

RGURE 43-36 TURKEYPOINT UNIT4, CYCLE 14 RADIALPOWER DISTRIBU11ON COMPARISON BETWEEN INCORE ANDANC 15 13 12 10 7

6 3

2 I

I I

I I

I I

I I

1 0.80%

Q338 0.353

<25%

G773 G803

-3.74%

1.067 1.086

-1.75%

.7 0.795

%25%

Q390 G391 426%

Q390 Q947 0.392 0.945

%.51%

Q21%

1.077 1255 1.059 1241 1.70%

1.13%

l.313 1.110 1262 1.084 4.04%

2A0%

1254 1.329 1234 1273 1.&2%

4A0%

l.

1.023 1.315 1.004

%.15%

1.S9%

G349 M52

%.85%

1.050 1.057 N.66%

1242 1240 0.16%

1261 1246 120%

1295 1271 1.89%

1.191 1.169 1.88%

1.368 1.310 4A3%

Q227 Q236

<.81%

0.798 1.085 GSOI 1.080 N.37%

OA&%

1252 1279 1259 1230

%.56%

3.98%

1.090 1290 1.081 1270 0.83%

1.57%

1286 1.17S 1267 1.166 1.50%

1.03%

1.026 1218 1.018 1225 0.79%

%.57%

1219 1282 1228 1285

%.73%

%23%

1285 1261 1263 1233 1.74%

227%

0242 G251

-3.$%

0.7S9 0.795 N.75%

1.373 1215 4.41%

1.011 1.004 0.70%

1297 1.310 N.99%

1257 1263 QA8%

1236 1233 G24%

0.987 0.961 2.71%

1.085 0.797 18386 0.803

%.09%

N.75%

1214 1254 1234 12&2

-1.62%

4.63%

1283 1.072 1273 1.084 Q79%

-1.11%

1.154 1274 1.169 1271

-128%

024%

1219 1.019 1227 1.018

%.65%

0.10%

1293 1225 1285 1225 a&2%

a00%

1238 1265 1233 1263 OA1%

0.16%

a351 M53

%.57%

1459 1.059 QIXI%

1207 1241

-2.74%

1243 1266

-1.82%

1.168 1,166 0.17%

1288 1.310

-1.68%

OAOI 0.392 2.30%

I I

I I

I I

I I

I I

I I

I I

I I

0.936 OA I 1 G945 G391

%.95%

5.12%

1.154 1.050 1240 1457 W.94%

%.66%

Q352 4.55%

l.040

1. I BI 0.750 1.081 1259 Q801

<.79%

%20%

%.37%

1220 1.152 1.045 1270 1230 1.080

-3.94% 434%

-324%

0.941 1236 G764 1.II4 1.315 0.795

%27%

%.01%

%.90%

I I

R I

I I

I I

I I

I I M I

I I

I I

I x G249 G251 0248 0236 5.08%

1.132 1.080 4.81%

1256 1287 1230 1270 2.11%

1.34%

I.172 1.166 G51%

1248 1.353 1267 1225 1285 1233 1.88%

529%

2.76%

1.321 1270 1285 1227 2,80%

3.50%

1.181 1.169 1.03%

1252 1205 1.069 1273 1234 1.086

-1.65%

-2.35%

-157%

Q233 Q237

-1.69%

0.828 0.801 3.37%

0.360 a352 227%

INCORE

'L DIFFERENCE 1294 1.091

~~ 1273 1259 1.081 1267 278%

0.93%

OA7%

1.085 1249 1239 1,057 1240 1246 265%

Q73%

N.56%

OA02 0 902 1.183 0.391 0.945 1241 2.81%

4.55%

<.67%

0.386 1.031 0.392 1.059

-1.53%

-2.64%

Q346 0.353

-1 98%

1.038 1237 1290 1.018 1227 1263 1.96%

0.81%

2.14%

12&0 1,170 1.323 1271 1.169 1.310

%.87%

0.09%

0.99%

1.035 1279 1.021 1.084 1273 1.004 4.52%

0.47%

1.69%

1230 1216 1.374 1262 1234 1.314

-2.54%

-1 A&%

4.57%

0.792 1.082 0.753 0.803 1.086 0.794

-1.37%

4.37%

-5.16%

1279 1.086 1.324 1225 1.018 1271 4.41%

6.68%

4.17%

1.197 1.304 1278 1.166 1266 1246 266%

3.00%

2.57%

1254 1.083 1272 1270 1.081 1240

-126%

0.19%

2.58%

1221 123 I 1.025 12M 1259 1.057 4.73%

-222%

-3.03%

1.032 0.777 Q340 1.080 0.801 0.352 AA4% -300%

-3A1%

1.105 1261 Q780 1.084 12&2 Q803 1.94%

%.08%

-2.86%

1239 1.041 0.338 1241 1.059 G353

%.16%

-1.70%

425%

0.933 G374 0.945 G392

-127% 4$ %

Q377 0.391

-3.5S%

Mean Absolute Difference Standard Deviation 0.021 0.019 0235 0241 G237 0251 4.84%

-3 98%

BURNUP = 600 MWD/MTU POWER LEVEL= 18K D BANKAT228 SIEPS FIGURE 4.3-17 TURKEYPOINT UNIT4, CYCLE 14 RADIALPOWER DISTRIBUTIONCOMPARISON BETWEEN INCORE ANDANC 15 0.362 0.370

-2.16%

0.790 0.791

%.) 3%

13 OA04 OA03 025%

1.057 1.084

-1.57%

1275 1282

%.55%

12 OA09 OA03

)A9%

0.918 0.910 0.88%

1 ~ 172 1.170 0.17%

1.067 1.069

%.) 9%

10 0248 0245 122%

0.37) 0.797 1.069 Q370 0.790 1.055 027%

0.89%

1.33%

).092

)290 1.196 1.084 1281 1.188 0.74%

0.70%

0.67%

1.177

).069 1.370 1.170 1.067 1.359 0.60%

0.19%

0.81%

1.188 1.346 1.158 1.188 1.339 1.152 0.00%

0.52%

0.52%

).356 0.999 1.176 1.342 1.010 1.183 1.04%

-).09%

N.59%

0268 0265 1.13%

0.784 0.781 0.38%

1.355 1.336 1A2%

).021 1.011 0.99%

1.384 1.375 0.65%

1.209 1204 OA2%

0249 0246 122%

1.067 0.792 1.058 0.791 0.85%

0.13%

1.179

)282 1.189 1282

%.84%

0.00%

1.37) 1.956 1.360 1.069 0.81%

%28%

1.167 1.369 1.154 1.342 1.13%

2.01%

1.197 1.020 1.184 1.010 1.10%

0.99%

0.370 0.370 0.00%

1.091 1.084 0.65%

1.168 1.170 N.)7%

1.177 1.188 Z.93%

1.339

'1.339 Q00%

OA03 OA03 Q00%

0.90) 0.398 0.910 0.403 4.99%

-124%

1.148 1.066 1.170 1.084

-1.88%

-1.66%

I I

I I

I I

I I

0.36)

Q370

-2.43%

1.054 1269 0.787

).O67 1281 0.790

-122%

N.94%

%.38%

I I

R I

I I

I I

I I

I I

I I

I I

I I x 0.249 O246 1.22%

.268

.265 1.13%

1.958 1.058 0.95%

0.783 0.781 0.26%

).)88 1.189

%.08%

l.336

).336 0.00%

).368 1.360 0.59%

).022 1.011

).09%

1.153 1.166 1.346 1.154 1.184 1.363 4.09%

-1.52%

-125%

1.393 120) 1.183 1.375 1204 1.187 1.31%

%25%

%.34%

1.199 1.187 1.01%

0.949 0.937 128%

1.377 1.180 1.363 1.183 1.03%

%25%

1.155 1.152 0.26%

1.185 1.182 1.361 1.)S7 1204 1.375 4.)7%

-1.83%

-1.02%

).357 1.178 1.957 1.359 1.187 1.055

%.15%

%.76%

l.14%

0.996

).363 0.786 1.011 1.336 0.780

-)A8%

2.02%

0.77%

0248 0245 122%

0268 0265 1.13%

0.247 0.245 0.82%

1.063 1.)85 1.055 1.188 0.76%

C.25%

0.792

).2SO 0.790 1.281 0.25%

%.08%

0.367

).OS4 0.370 1.084 4.81%

0.00%

0.404 OA03 0.25%

1.367 1.359 0.59%

).953 1.067

%.37%

1. I70 1.170 0.00%

0.906 0.910 1.165 1.1S9 1.373

).183 1.152 1.183 1.363 1.187 1.13%

0.51%

0.73%

%.34%

1.357 1.016 1.182 1.178 1.339 1.010 1.184 1204 1.34%

0.59%

4.)7%

-2.16%

1.178 1.345 1.142

).377 1.188 1.342 1.154 1.375

%.84%

022%

-'. 04%

0.'15%

1. )62 1.058

!,382 1.034 1.170 1.958

).3M 1.011

%.68%

C.94%

).62%

2.27%

1.363 1.178

).138 1.363 1.184 1.154 0.00%

4.5)%

-1.39%

1.175 1.017 1.329 1.1S3 1.010 1.342

%.68%

0.69%

%.97%

1.153 1.339 1.168 1.152 1.339 1.188 O.N%

0.00%

-1.68%

1.3S7 1.950 1.)58 1.359 1.067 1.170 2.06%

%.66%

-1.03%

1.351 1.180 1.070 1.360 1.189 1.058

%.66%

%.76%

1.13%

1.043

)269 0.788 1.06S 1282 0.791

-2.34%

-1.01%

4.38%

1.153

).074 0.366 1.170 1.084 0.370

-)A5% 4.92%

-1.0S%

0.901 0.402 0.910 0.403 4.99%

%.25%

0248 0246 0.81%

D INCORE ANC

'/ DIFFERENCE 0.392 O.403

-2.73%

1.052

).243 1.178

).360 1.084 1.282 1.189 1.336

-295%

-3.04%

4.93%

).80%

0.356 0.767 1.054 0.792 0.370 0.791 1.058 0.780

-3.78%

-3.03%

4.38%

1.54%

1295 1.293 1.089 1.187 1.2S) 1.084 1.60%

0.94%

0.46%

1.076 0.805 0.374 1.055 0.790 0.370 1.99%

).90%

1.08%

0.405 0.403 0.50%

Mean Absolute Dttterence Standard Devlation-O.ON 0.007 0.239 0.271 0246 0265

-2.85%

226%

0251 0245 245%

BURNUP = 6836 MWD/MTU POWER LEVEL= 100%

D BANKAT228 STEPS

FIGURE 43-) 8 TURKEYPOINT UNIT4, CYCLE 14 RADIALPOWER DISIRIBUIIONCOMPARISON BETWEEN INCORE ANDANC 15 14 13 12 10 9

8 7

G379 0.388

-2.32%

Q797

)AI%

OA24 0.94%

1.073

),090

-1.56%

)269

)272

%24%

OA37 OA24 3.07%

1.167 1.159 Q69%

I.ON 1.0N 0.00%

G391 G3M 0.77%

1.104

)A)90

)2S%

1.172 1.159 1.12%

1.184 1.179 OA2%

1.3N

)2b)

Q59%

QSOO 0.797 0.3S%

)Z74

)272 0.16%

).060 1.068 W.75%

).361 1.359 0.15%

1.021 1.012 0.89%

0265 Q260 1.92%

1.057 1.050 0.67%

1.159 1.169

%.86%

1.369

)272

%22%

1.143 1.143 GOO%

1,166 1.165 Q09%

9291 0285 2.11%

Q793 a794

%.)3%

)2)0 1.320

%.76%

1.015

)AX)8 O.N%

)275 1.369 OA4%

I.)76 1.177

%.08%

9266 Q26)

).92%

).060 1.052 0.76%

1.167 1.170

%26%

1279 1.373 OA4%

1.156 1.144 1.05%

1.176 1.166 Q86%

QSI 0.797 G38%

)475

)272 G24%

1.071

).ON 0.19%

1.391 1.361 220%

).028 1.012 1.5S%

0.387 0.388

%26%

I 094

).090 G37%

1.170 1.159 0.95%

1.184 1.179 OA2%

OA23 OA24

%24%

0.921

).16) 1.159 0.17%

)271 1.064 1.358

).658 G96%

%.37%

OA23 OA24

%24%

1.087

).090

%28%

1280

)272 0.63%

Q383 L388

-)$%

Q805 Q797 1.00%

I I

R I

I I

I I

I I

I I M I

I I

I I x Q270

).ON 1.165 9261 1.052 1.170 A5%

1.62%

NA3%

1.387 1.373 1.02%

1.146 1.167 1.144 1 ~ 166 0.17%

0.09%

1.361 1.354 0.52%

1. 166 1.362 1.162 1.354 0.34%

0.59%

1.163 1.165 N.)7%

1. 14) 1.378 1.143

)272

%.17%

G44%

1.173 1.169 034%

l.072 Q2N 1.050 Q260 J 2.10%

3A6%

G794 1.320 1AO%

%.13%

N.76%

1.008 0.89%

1.374

)269 0.37%

I,)73 1.177

%.34%

1.161 1.162

%.09%

0.937 0.929 0.86%

1.159 1.162

%26%

1.166 1.177

%.93%

&69 1.015 1.369 1.008 0.00%

0.69%

).339 1.320

)A4%

G804 G295 0.793 G285 H 1.39%

3.51%

025S 1.041 I. I47 0260 1.0%

1.169

%.77%

%.86%

-1.88%

0.785

)243 0.797

)272

-1.51%

-228%

0.377 1.066 Q388

).090

-2.84% -2X%

OA)6 OA24

-).89%

INCORE

% DIFFERENCE 1.369 1.372 1.052 1.068

-).50%

I.149 1.159 0.904 0.921

-1.85%

0.417 OA24

-1.65%

1.155 I,165 1.143 1.165 1.05%

0.00%

1.388 1.016 1.359 1.012 2.13%

OA0%

1.) 51 1.355 1.179 1.361

-2.37%

NA4%

1.135 1.039 1.159 1.959

-2.07%

-2.81%

1.067

)245 1.090 1.272

-2.11%

-2.12%

0.377 0.781 0.388 0.797

-2.84%

-2.0)%

1.354 1.354 0.00%

1.149 1.166

-)A6%

1.115 1.144

-2.53%

1.389 1.373 1.17%

I.)60 1.170 1.050 1.052

%.)9%

1.146 1.351 1.162 1.354

-1.38% 422%

1.134 1.148 1.177 1 ~ 165

-3.65%

-)A6%

1.363 1.138 MN 1.143 CA4% NA4%

1.036 IA09 1.008 1.371 2.78%

2.77%

).339 1.187 1.320 1.)N IA4%

1.54%

0.800 1.066 0.793 1.050 G88%

).52%

1.161 1.166 NA3%

1.018 1.012 D.59%

1.359 1.358 0.07%

1.061

)280

)Z72 0.63%

Q820 0.796 3.02%

1.136 1.372 1.144 1.373

%.70%

%.07%

1.352 1.04S 1.361 1.069

%.66%

-1.96%

1.171 I.IS I 1.179 1.159

%.68%

Q.N%

1.139 0.919 1.158 Q921

-1.64% 422%

1.075 OA25 1.090 OA23

-1.38%

0.47%

0.380 a3ss

-2.06%

1.163 1.170

%.60%

)260

)272 N.94%

1.087

).090

%2$ %

OA30 OA23 1.65%

).058 Q265 1.052 G261 G 1.52%

1.53%

0.791 0.797 N.75%

D Mean Absolute Dttference Standard Devhtlon QOO9 0.008 0256 Q261

-1.92%

0285 Q260 1.75%

1.92%

BURNUP = 10704 MWD/MTU POWER LEVEL= 1K8 D BANKAT228 SIEPS 2.00 ZxeVRE 4.3-19 TURKEY POXNT UNXT 4 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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

2.00 FXG&& 4 3-20 TKGQCEZ'OXNT UNXT 4 CYCZiE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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 2 00 rXGmu" 4.3-21 TUEGKlY POXNT UNXT 4 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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

2. 00 FIGURE 4 3-22 TURKIC POXNT UNIT 4 CYCLE 13 AXIAL POWER DXSTRXBUTXON COMPARISON BETWEEN INCORE AND ANC 1.75 XNCORR 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 2.00 Pxemm 4.3-23 TXGGCEY POXNT UNXT 4 CYCLE 13 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN% XNCORE AND ANC 1.75 INCORE 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

2.00 FxeURE 4.3-24 TUEQCEY POXNT UNXT 4 CYCLE 13 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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 2.00 FXemm 4 3-25 TURKEY POXNT UNXT 4 CYCLE 14 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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 AXIALHEIGHTg INCHES BVRNVP~600 MWD/MTU POWER ZaEVZLs100Ss D BANK AT 228 STEPS 0

0

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

2.00 I

PXQURE 4.3-27 TUEQCEeY POXNT UNXT 4 CYCLE 14 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNCORE AND ANC 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.

O St. Lucie Units 1 8 2 are similar in design.

St. Lucie Unit 1 is a Combustion 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.

Allfuel 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 CONFIGURATION CRITICAL BORON CONCENTRATION (PPM)

MEASURED PREDICTED DIFFERENCE M

(M-P) 10 10 0"

ARO BANK A in ARO,.

BANK A in ARO BANK A in 1598 1477 1393 1279 1419 1303 1609 1496 1396 1281 1427 1307

-11

-19

-3

-2 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 0

1.54 Acceptance Criteria is

+2 pcm/'F 2.18

-0. 64

TABLE 5. 2-3 ST.

LUCIE UNIT 1 CYCLE 10,11 AND 12 CONTROL ROD WORTH COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CONFXGURATXON CONTROL ROD WORTH (PCM)

MEASVRED M

PREDXCTED DXFFERENCE(8)

P

( (M-P) /P) *100 10 12 BANK 7 BANK 6 BANK 5 BANK 4 BANK 3 BANK 2 BANK 1 BANK B BANK A(1)

TOTAL(2)

BANK 7 BANK 6

& B BANK 5 Ec 3 BANK 4 BANK 2 BANK 1 BANK A(1)

TOTAL(2)

BANK 7 BANK 6 Ec 3 BANK 5 Sc B

BANK 4 BANK 2 BANK 1 BANK A(1)

TOTAL(2) 516 367 374 584 368 789 746 543 1015 5302 590 715 850 738 791 808 1136 5628 654 929 509 824 699 759 1099 5473 4,80 404 430 631 420 848 822 584 1014 5635 522 774 882 699 795 806 1106 5583 573 915 550 809 740 771 1136 5495 7..50

-9.16

-13.02

-7.45

-12.38

-6.96

-9.25

-7.02 0.10

-5.91

13. 03

-7

~ 62

-3. 63 5.58

-0.50 0.25 2.71 0.81

14. 14 1.53

-7.45 1.85

-5.54

-1.56

-3.26

-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 CONFXGURATXON DXFFERENTXAL BORON WORTH (PCM/PPM)

MEASVRED PREDXCTED DXFFERENCE (0)

M P

( (M-P) /P) *100 10 Average Over Bank A insertion 8.39 8.97

-6.50 12 Average Over Bank A insertion Average Over Bank A insertion 9.96 9.47 9.62 9.47 3.53 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 mn/MTU CRITICAL BORON CONCENTRATION (PPM)

MEASURED PREDICTED DIFFERENCE M

P (M-P) 10 139 278 696 1392 2784 4176 5568 6960 8352 9744 11136 12528 13920 15312 15947 1160 1130 1090 1050 990 915 845 780 720 660 560 440 320 190 129 1178 1162 1117 1074 987 910 840 772 709 633 545 438 314 187 129

-18

-32

-27

-24 3

5 5

8ll 27 15 2

6 3

0 13.6 278 679 1359 2718 4078 5437 6796 8155 19514 10874 12233 13592 14404 952 939 903 857 787 687 611 528 455 372 291 195 82 13 968 953 909 862 768 681 599 517 440 359 276 176 63-7

-12

-14

-6

-3 19 6

12 11 15 13 15 19 19 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 mWV/MTU CRITICAL BORON CONCENTRATION (PPM)

MEASURED PREDICTED DIFFERENCE M

P (M-P) 0 12 132 265 661 1324 2648 3972 5296 6614 7944 9268 10592 11916 13240 13723 961 940 920 892 805 743 672 596 537 443 390 287 175 131 991 976 934 892 807 729 658 589 525 453 369 273 164 122

-30

-36

-14 0

2 14 14 7

12 10 21 14 11 9

Acceptance Criteria is +50 ppm

~

I

~

~

I

~

~ I

~

~

I

~

~

I

~

~ ~

~ a

~

~

~

~ ~

C"

~

~

~

I

~ ~

C>

I

~

~

I ol C~

~

~

~ I;

~

~ ) 9:

~

~

~

I I

~

~ ~

~

~

~

~ )

~

~

ri (ii

~

(ri

~

~

I

~

~

I

~

~

I

~

~ I;

~

~

~

~

~

~

0

)

~ ~

r

~

~ ~

C"

~

~

~

I

~

~

C>>;

~

~

I

~

~ ~

C" I

~

~

I

1100 FXeaRZ 5-2-3.

ST LVCXE VNXT 1 CYCLE 10 lllHHIED VIHIBUS PRROXCTZO REFERENCE BANK XNTEGRAL ROD WORTH 1000 900 MEASURED

.PREDICTED 800 8O 700 0

8"'oo 300 200 100 0

0 40 80 120 ROD POSITION (INCHES WITHDRAWN) g P

0

~

I 0

Q4 0

l

~

~

~

.MWlMR

- SM MMMM.. EM

~~i~

~~%M

~~~K~~~~

~~~~R~

~

0 I

~

3l

(

" ~

Dl

~ ~

i~

,MR%MMMMM

~~1~~

.MMIMS

- SM MMES.. SM

~~h~

MME&MMMM

~~~K~~~~

MMML%M

~~~~RM~~

~~~~~ik~~

I s

~

I

~

~

~ ~

~

I I

Dl I8

~

~

DlI "EEEEEEEEkkakkakk raaaaarkrkrrkkrr "kakakakkrkkkraak aaraaaaaaaanrrr aarkkrkaEEEEEEEE waaakkkkrnrarra EHEEEEEEEEEEEEEE

,awi%aaaaaraaaaaaa

"'kawaraar~~

~mark

.arab+%ra)

~ Eaaa "kkrrk~era

., arkr

.Eaarkhi%IIIaaaasNEE EEEEEEE%aakkkrkr EEkkkrahKSEEEEEER "akkaarraakrkkrrr kkkkkkkkrhiararak "aaaarkkaraaakkar anrarnrrraana EEEEEEEEEEREREEE

,naaararraraL)raa EEEEEaakrkkkkr ra EEEEEEEEEEEEEERE akaraaaarrkarrrk kkrkkkaaraaaarar k

EEEEEEEEEEEL%

akraaakaaaarkraa E

EEEEEEkkkkka kkkkrkkkkraa

~

~

I

1200 PXGURE 5-3-2 ST LUCXE UNXT 1 CYCLE 11 BORON LETDOWN COMPARXSON BETWEEN AND PREDXCTXON 1100 1000 PcA 900 80 g

800 VO0 MEASURED PREDICTED 600 0

0 500 OH 400 l4O 300 200 100 0

0 2000 4000 6000 8000 10000 12000 14000 CORE AVMGLGE BUEQGJPg MND/MTU 1200 ZXeaZE 5.3-3 ST LUCXE UNXT 1 CYCLE 12 BORON LETDOWN COMPARXSON BETWEEN AND PREDXCTXON 1100 1000 Pr 900 0

800 woo MFASURFD PREDXCTED g

600 0

C40 Pl 500 OH 400 U

300 200 100 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUPg MWD/MTU

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

2.00 FIGURE 5.3-5 ST LUCXE UNIT 2. CXCLE 10 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AHD ANC 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 2.00 FIGURE 5.3-6 ST LUCIE UNIT 1 CYCLE 10 AXXAL POWER DISTRIBUTION COMPARISON BETWEEN XNPAX AND ANC 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>

2 00 PXGURE 5.3-7 ST LUCXE UNXT 1 CYCLE 11 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 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 BOTTOH TOP AXXAL HEXGHTi XNCHES BUEMJP=185 MWD/MTU POWER LEDS L=100i 2 00 FIGURE 5 3-8 ST LUCXE UNXT 1 CYCLE 11 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 1.75 1 ~ 50 g0 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<

2.00 FICUS 5.3-9 ST LUCXE UNIT 1 CYCLE 11 AXIAL POWER DISTRIBUTION COMPARISON BETWEEN XNPAX AND ANC 1.75 Pg 1.00 4

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-

2 00 FXQURE 5-3-10 ST LUCXE UNXT 1 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETWEEN XNPAX AND ANC 1.75 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 2QCXAL HEIGHT, INCHES BUEQGJP 625 MWD/MTU POWER LEVEL=1 0 0 ~o 2 00 FXGURE 5.3-11 ST LUCXE UNXT 1 CYCLE 12 AXXAL POWER DXSTRXBUTXON COMPARXSON BETNPNN XNPAX AND ANC 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

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

1.25 Pg 1.00 d

~ 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 BURNVP=13, 320 MWD/HTU POWER LWlEL=100io

-101-

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.

0 Miller, R.W., et al, "Relaxation of Constant Axial Offset Control/FQ

, 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

~

8.

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

~

~

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

-102-

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.

-103-

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 O

code versions were discussed.

The NRC concurred that the updated code 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 fillgas, 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 fluxdistribution. 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 fluxdistribution 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.

Astandard B1 calculation is used to evaluate the critical spectrum ofthe 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.

-106-

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.

-108-