ML20138E742
| ML20138E742 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/22/1995 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20136C539 | List:
|
| References | |
| FOIA-96-485 NUDOCS 9506280347 | |
| Download: ML20138E742 (8) | |
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ATTACHMENT 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT NF-TR-95-01
" NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN OF TURKEY POINT & ST. LUCIE NUCLEAR PLANTS" i
FLORIDA POWER AND LIGHT COMPANY j
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1.0 INTRODUCTION
In a letter of January 17,1995 (Ref.1), from T. F. Plunkett to the U.S.
Nuclear Regulatory Commission (NRC), Florida Power and Light Company (FPL) submitted Topical Report NF-TR-95-01 (Ref. 2) entitled " Nuclear Physics i
Methodology for Reload Design of Turkey Point & St. Lucie Nuclear Plants" for NRC review (Ref. 2). The report describes the methodology used by FPL to analyze the core design characteristics for Turkey Point Units 3 and 4 and St.
Lucie Units 1 and 2.
The methodology was obtained from Westinghouse Electric Corporation and calculations using this methodology were performed by FPL and the results compared to operating data from Turkey Point and St. Lucie.
However, since Westinghouse is the nuclear steam supply system (NSSS) vendor and present fuel supplier for only the Turkey Point units, only approval for FPL use of the methodology for reload design calculations for Turkey Point Units 3 and 4 was requested at this time.
2.0 TOPICAL REPORT SUMARY The report describes the use of Westinghouse methodology as applied by FPL to analyze the core characteristics of the Turkey Point and St. Lucie nuclear plants. Startup physics measurements as well as core follow results for Turkey Point Unit 4 during cycles 12,13, and 14 are used to compare critical boron concentrations, temperature coefficients, control rod worth, differential boron worth, power peaking factors, and radial and axial power distributions. Comparisons between measurements and predictions of critical boron concentration, moderator temperature coefficients (MTCs), control rod worth, differential boron worth, and axial power distributions for St. Lucie Unit 1 Cycles 10, 11, and 12 are also presented.
3.0 TOPICAL REPORT EVALUATION FPL has entered into a technology exchange agreement with the Commercial Nuclear Fuel Division of Westinghouse through which the relevant physics design methodology and associated computer programs have been obtained. A WSAG 2205 ZM
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training program was initiated which included hands-on experience performing actual calculations to ensure that the FPL engineers understood the Westinghouse methodology. Some of the reload physics calculations were
.i performed by FPL independently of Westinghouse with Westinghouse providing quality assurance of all calculations. All of the methods employed and described in this topical report (including model development, computer programs, measured data processing, etc.) are NRC-approved standard Westinghouse methods and reflect current practices.
3.1 Computer Codes PHOENIX-P is a two-dimensional multigroup transport theory code (Ref. 3) which has been qualified and approved (Ref. 4) for use in calculating pressurized water reactor (PWR) lattice physics parameters and determining neutronics i
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input for the two-group diffusion theory code, ANC.
PH0ENIX-P uses a 42-energy group cross section set derived from the standard ENDF/8-V cross section library.
ANC is an approved Westinghouse three-dimensional two-group diffusion theory nodal code (Ref. 5) which was also qualified for use with PH0ENIX-P by l
Reference 4.
The code is based on coarse mesh nodal (4 nodes per assembly) diffusion theory using the non-linear nodal expansion method, with coupled thermal-hydraulic and Doppler feedback. The code includes the following modeling capabilities: solution of the two-group neutron diffusion equation, equivalence theory cross section homogenization, cross section depletion, explicit baffle / reflector modeling, and a rod power recovery model.
The two-group model solves the neutron diffusion equation in three dimensions, with assembly homogenization.
In order to preserve the flux and current continuity at nodal interfaces, ANC uses flux assembly discontinuity factors 4
that are obtained from the PH0ENIX-P two-dimensional detailed lattice analysis. ANC also employs flux ~ discontinuity carryction factors to combine the global (nodal) flux shape and the assembly heterogeneous flux distribution for the rod power recovery model. The use of an explicit baffle / reflector cross section representation eliminates the need for user-supplied albedoes, normalization, or other adjustment at the core / reflector interface.
The fuel depletion model uses macroscopic cross sections to account for fuel exposure without tracking the individual nuclide concentrations. ANC can be used to calculate the three-dimensional pin-by-pin power distribution in a manner that accounts for individual pin burnup and spectral effects. ANC also calculates control rod worth and moderator, Doppler, and xenon and samarium feedback effects.
APOLLO is a Westinghouse one-dimensional axial two-group diffusion theory code (Ref. 6), currently under NRC review, which uses radially homogenized flux and volume weighted cross sections from the three-dimensional ANC model. The one-dimensional APOLLO model is normalized to the three-dimensional ANC model results by performing an elevation-dependent radial buckling search at each burnup step (Ref. 7). APOLLO is an advanced version of the approved PANDA code (Ref. 8) which was also described in the October 1984 meeting between Westinghouse and the NRC (Ref. 9).
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FIGHTH=is a Westinghouse computer code derived from previous LASER (Ref. 10) and REPAD. (Ref.11) models and which has been accepted (Ref.12) for i.
predicting steady-state fuel rod temperatures for low-enriched sintered U0 i
fuel rods. Thiscodeiscurentlyusedonlyforcalculatingfuelandcladding l
effective temperatures for input to the PHOENIX-P code as a function of j
burnup,, linear heat generation rate, moderator temperature and flow rate.
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The standard Westinghouse INCORE. computer code (Ref. 13) is used to process F
the neutron flux measurements made by the movable incore fission chambers to determine the core power distribution, as required by the Turkey Point Technical Specifications (TS). The measured flux values are combined with power-to-reaction rate ratios analytically generated with the PHOENIX-P/ANC models in order to infer a " measured" three-dimensional power distribution.
This standard Westinghouse technique allows use of the previously established measurement uncertainties. Since all methods employed are stated to be 1
standard licensed methods, the Westinghouse calculational uncertainties (Ref.
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- 14) for the nuclear hot channel factors are also used by FPL.
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-FPL has used the Westinghouse methodology package described above to perform i
design calculations for Cycles 12, 13 and 14 of Turkey Point Unit 4.
Unit 4 was chosen because of Its wide variety of assembly and burnable absorber i
types, its transition to axial blanketed fuel, its large number of reinserted l
fuel assemblies, vessel flux reduction features (e.g., hafnium inserts at the periphery), and its low-leakage fuel management. An evaluation of these comparisons is presented below for the key PWR physics parameters to be i
i generated by the licensee.
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3.2 Critical Baron Concentrations I
Critical boron concentrations were measured at hot zero power (HZP) conditions
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by acid-based titration with all rods out (ARO) and with the reference bank (the bank of highest worth) fully inserted. The FPL ANC three-dimensional model~ predictions of critical boron concentration were compared to zero power startup test measurements for Cycles 12, 13 and 14 of Turkey Point Unit 4.
The results from the HZP comparisons qualify the model for predicting the
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critical boron concentration and reactivity for beginning-of-cycle (BOC),
xenon-free conditions. Six measurements from the three cycles of startup tests are included. All differences are within the 150 ppm review criterion.
4 3.3 Isothermal T==erature Coefficients The isothermal temperature coefficient (ITC) is defined as the change in reactivity due to an incremental change in the core average moderator and fuel temperature.
ITCs were measured by making small changes in the reactor coolant system (RCS) temperature and determining the corresponding change in reactivity with the plant reactivity computer. FPL used the three-dimensional ANC model to calculate the ITC by unifomly varying the moderator temperature by 15 *F about the HZP temperature and by determining the Doppler (fuel) temperature effect using the fitting coefficients from the FIGHTH calculations. The measured and predicted ITCs and MTCs compared within the review criterion of 12 pca F from the three cycles of operation. Note that 1 pcm is equivalent to Ix10'y* percent Ak/k.
The measured MTC is obtained by
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j subtracting the Doppler coefficient from the measured ITC.
3.4 Control Rod Worths
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Control rod worth is the reactivity difference'(pcm) between two different control rod configurations.- The worth of the reference bank (the bank of highest worth) was measured by boron dilution, using step-wise bank insertion j
and summing the differential worths obtained from the reactivity computer..
The remaining banks were then individually fully inserted, while holding boron concentration constant, and withdrawing the reference bank to maintain criticality. The integral worth of each inserted bank is inferred from the equivalent worth of the reference bank measured critical position, corrected i
for the presence of the inserted bank. This is consistent with the Westinghouse Rod Swap Technique (Ref. 15), which was approved by the NRC in i.
1982. The three-dimensional ANC model was used for the prediction of the individual control rod bank worths and was compared by FPL with the BOC zero-power startup measurements for three operating cycles of Turkey Point Unit 4.
ia All relative differences were within the st review criteria of *10% on the
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reference bank worth and il5% (or 100 pcm) on the swapped rod worths. The ANC model is also used to generate the analytical correction factors which account 1-for the effect of the inserted bank on the partial integral worth of the reference bank.
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'3.5 Differential Baron Worths
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Measured differential boron worths (pcm/ ppm) were inferred by dividing the L
measured reference bank worth by the difference between the critical boron a
concentrations with ARO and the reference bank inserted. The three-dimensional ANC model was used to calculate the worth of a 125 ppm change about the HZP measured ARO critical boron concentration. The measured and l
predicted boron worths from the three Turkey Point cycles were compared by FPL. ' All relative differences were within the test review criterion of 115%.
3.6 Boron Letdown Curves Critical boron concentrations from measured hot full power (HFP), equilibrium xenon and samarium conditions were compared to the three-dimensional ANC model i
predicted boron letdown curves for the three cycles of Turkey Point Unit 4 stated above. These at-power comparison results, corrected for control rod insertion and for deviations' from the full-power, equilibrium xenon and
' samarium conditions, are used as estimates of the model uncertainty for all equilibrium power conditions with thermal feedback. There are a' total of 31 measurements from three operating cycles, taken at the time of INCORE power distribution measurements. The mean difference between neasured and predicted critical boron concentrations for all three operating cycles is 9 ppe, with a
. standard deviation of 13 ppe, and well within the test review criterion of 150 PPs.
3.7 Power Peakina Factors Measured values of the primary power peaking factors, the heat flux hot i
channel factor (F ) and the nuclear enthalpy rise hot channel factor (FL),
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5 were inferred using the Westinghouse INCORE code.
The predicted power peaking factors were obtained from the three-dimensional ANC model results at the closest depletion interval. For F and predicted values for 31 measur, the mean difference between the measured ed statepoints over the three operating cycles was 3.33% with a standard deviation of 1.86%. For FL, the mean difference is 2.02% with a standard deviation of 1.27%. These are within the Westinghouse uncertainty values stated in Reference 14.
3.8 Radial Power Distributiors The measured radial power distributions are inferred by the INCORE code, after the flux map measurements are performed using the moveable incore neutron flux detector system. The predicted power distributions are interpolated from the three-dimensional ANC depletion step results at HFP, ARO operating conditions.
The mean. absolute difference between measured and predicted assembly relative powers is less than 0.021 with a standard deviation less than 0.023.
3 3.9 Axial Power Distributions and Axial Offset 1
i A total of nine axial power distribution measurements from the above flux maps over the three cycles were compared to the three-dimensional ANC model predicted values at similar depletion points. The measured axial offset (AO),
defined as the percent difference between the relative power in the top half of the core and that in the bottom half of the core, is also inferred by INCORE and is compared with the predicted values from ANC at 31 flux map
-statepoints. The mean difference between measured and predicted values for the three cycles is 0.66% with a standard deviation of 1.54%.
4.0 SUMARY AND CONCLUSIONS FPL has performed benchmarking for three cycles of operating data from Turkey Point Unit 4 using currently accepted Westinghouse reload design methodologies. The benchmarking effort consisted of detailed comparisons of the calculated physics parameters with the measurements obtained from the Turkey Point PWR. The results demonstrate that the Turkey Point plant-specific agreement is within the Westinghouse determined uncertainty analysis for the stated PWR physics parameters. This effort also demonstrates the capability of FPL to use the Westinghouse computer program package for application to the Turkey Point units. FPL intends to use these methods for steady-state PWR core physics reload design applications and safety analysis inputs.
Based on the analyses and results presented in the topical report, the NRC staff concludes that currently approved Westinghouse methodologies, as validated by FPL, can be applied to steady-state PWR reactor physics calculations for the Turkey Point reload design applications discussed in the above technical evaluation. The accuracy of this methodology has been demonstrated to be sufficient for use in design applications, including PWR reload physics' analysis, generation of transient analysis input data, startup predictions, plant reactivity computer inputs and Core Operating Limits Report (COLR) parameters -(axial flux differences, control rod insertion limits, and heat ~ flux hot channel factors).
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4 Application of the approved package is to be limited to the range of fuel assembly and core reload design parameters verified in the topical report.
Addition of new Westinghouse fuel designs would be acceptable without further i
review, if analyzed by currently approved methodologies. Future adoption by i
FPL of Westinghouse improved methodology which has been reviewed and approved
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by the NRC is also acceptable for COLR referencing. However, any change from the current fuel vendor, which also introduces different fuel designs or core operating strategies, may require further validation by the licensee since the j
approved computer codes and procedures have been qualified against a Westinghouse fuel design base.
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5.0 REFERENCES
(1) Letter from T. F. Plunkett (FPL) to Document Control Desk (NRC),
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Implementation of FPL Nuclear Physics Methodology for Calculations of Core Operating Limits Report Parameters, dated January 17, 1995.
(2) NF-TR-95-01, " Nuclear Physics Methodology for Reload Design of Turkey Point & St. Lucie Nuclear Plants," Florida Power & Light Company, January i
1995.
(3) WCAP-10106-P-A, "A Description of the Nuclear Design Analysis Programs for Boiling Water Reactors," June 1982.
(4)' WCAP-Il596-P-A, " Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," November 1987.
(5) WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Program,"
December 1985.
(6) WCAP-13524-P, " APOLLO - A One Dimensional Neutron Diffusion Theory Program," October 1992.
1 (7) WCAP-8903 (Proprietary), " Analysis of Elevated Dependent Power Peaking i
Factors," December 1976.
l (8) WCAP-7048-P-A and WCAP-7757-A (Non-proprietary), "The PANDA Code,"
January 1975.
(9) Letter from N. J. Liparulo (W) to R. C. Jones (NRC), " Presentation Material from the October 10, 1984'NRC/ Westinghouse Meeting on the Improved Versions cf Neutronics Codes Being Utilized (Non-proprietary),"
ET-NRC-92-3736, August 27, 1992.
(10) WCAP-6073, " LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS," April 1966.
(11) WCAP-2048, "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," July 1962.
(12) Letter from S. A. Varga (NRC) to D. L. Farrar (Ceco), accepting
" Commonwealth Ediscn Company Topical Report on Benchmark of PWR Nuclear
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l Design Methods," NFSR-0016, Rev. O, December 1983.
(13) WCAP-8498, "INCORE Power Distribution Determination in Westinghouse
,4 Pressurized Water Reactors," July 1975.
l-l (14) WCAP-7308-L, " Evaluation of Nuclear Hot Channel Factor Uncertainties,"
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April 1969, and " Update to WCAP-7308-L-P-A-(Proprietary), Evaluation of
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l Nuclear Hot Channel Factor Uncertainties," June 1988, i
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(15) WCAP-9863-P-A, " Rod Bank Worth Measurements Utilizing Bank Exchange," May l
1982.
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i ATTACHMENT 2 SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE FACILITY NAME Turkey Point Units 3 and 4 I
SUMMARY
OF REVIEW The SER involved a review of Topical Report NF-TR-95-01 which presented benchmarks of physics parameters calculated by Florida Power & Light Company for several operating cycles of Turkey Point using approved Westinghouse methodologies.
The review was conducted by the Reactor Systems Branch /DSSA/NRR.
Based on its review, the staff concludes that the Topical Report is acceptable for referencing by Florida Power & Light Company in licensing applications.
NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE - SAFETY ASSESSMENT /
OUALITY VERIFICATION The licensee addressed all aspects of the issues and presented a well-organized report.
This achievement indicated good interdepartmental communicationsand a technically qualified staff.
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1 AUTHOR:
L. Kono DATE:
5f/22195 i