ML20138D318
| ML20138D318 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/31/1993 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19303F306 | List: |
| References | |
| OPPD-NA-8302-NP-R03, OPPD-NA-8302-NP-R3, NUDOCS 9302170261 | |
| Download: ML20138D318 (68) | |
Text
m Table 2 NEUTRONICS DESIGN METHODS AND VERIFICATION OPPD-NA-8302-NP Rev. 03
- 1. Title Page Change the revision number and date.
- 2. All Pages Update the revision number.
- 3. All Pages - Change all references of Advanced Nuclear Fuels Corp.
to Westinghouse and ANF to H. __
- Change all references of CE or Combustion Engineering to ABB-CE or ABB-Combustion Engineering.
- 4. Page 111 Update Table of Contents with new page numbers and
- Replace QUIX with HERMITE where appropriate since HERMITE replaces QUIX as the axial shape analysis code.
- Revise Section 5.0.
- 5. Pages iv-vi Update List of Tables and List of Figures.
- 6. Page vii Update revision sheet.
- 7. Page 3 Replace QUIX with HERMITE.
- 8. Pages 3-4 Changed ROCS method utilized to Nodal Expansion _ Method (NEM) from Higher Order Difference Method (HOD).
- 9. Pages 4-5 Replace Section 2.2.3 describing the QUIX code with a description of HERMITE.
- 10. Page 6 Change reference to using ABB-CE computer codes on the Windsor, Connecticut mainframe computer to using workstation versions of the ABB-CE codes on the OPPD-Nuclear Engineering UNIX Workstation Network located on the Fort Calhoun Station site.
- 11. Page 6 Replace references to using cross-sections for ANF fuel with using cross-sections for H fuel.- Also added statement about using cross-sections for fuel assemblies containing hafnium flux suppression rods and for naturai uranium fuel assemblies.
- 12. Page 6 Replace QUIX with HERMITE.
- 13. Page 8 Replace Section 3.2.3 describing the QUIX code with a ,
description of HERMITE.
- 14. Page 9 Delete reference to PDQ-X._ I
.15 . Page 9 Reword last sentence of Section 4.1 for improved-clarity. .
- 16. Page 10_ Update description of method of MTC calculation to remain consistent with ABB-CE methodology and cases performed for HERMITE axial shape analysis.
- 17. Page 10 Update description of method of FTC calculation to remain consistent with ABB-CE methodology and cases performed for HERMITE axial shape analysis.
9302170261 93020S PDR ADOCK'05000285 P PDR
Table 2 NEUTRONICS DESIGN METHODS AND VERIFICATION j OPPD-NA-8302-NP Rev. 03
- 18. Page 10 Add FTC to MTC discussion for completeness.
- 19. Page 11 Delete Doppler uncertainty reference, as it-is contained.
in Reference 4-1.
- 20. Page 11 Change uncertainty on Beff and 1* to remain consistent with ABB-CE analysis methods.
- 21. Page 11 Several changes in Section 4.4 to. delete reference to Fxy and 2-D. All analyses are now being done in 3-D with the added processing power available from the workstation codes.
- 22. Page 11 Reference to combination of-. distortion factor with 1 dropped rod worth was celeted as it is more appropriate l to include in OPPD-NA-8303-P, Rev. 4, Section 5.3.7. '
- 23. Page 12 One additional change to delete Fxy and a thange to reference Reference 4-1 for biases and uncertainties.
- 24. Page 12 Last paragraph of Section 4.4 deleted as it is more appropriate to include in OPPD-NA-8303,- Rev._4 ' Section 5.3.7. (to maintain consistency with dange Number 22 above.)
- 25. Page 13 Moved the uncertainty statement from here to the end of the section and updated reference'to Reference 4-1.
- 26. Pages 13-14 Update method of calculation of axial thermal-hydraulic changes from HFP-HZP to remain consistent with ABB-CE applications methods and .is conservative with respect.to previous methods.
- 27. Page 14 Add reference for biases and uncertainties to-refer to Reference 4-1.
28 _Page 14 Update method for calculating the maximum differential CEA worth to remain consistent with ABB-CE-applications methods and is conservative with respect to previous-methods.
- 29. Page 15 Update the method for deriving limiting valves. for doppler reactivity insertion for the Steam Line Break Cooldown.
- 30. Page 16 Delete references to calculation' of Doppler reactivity insertion as this is described in OPPD-NA-8303-P, Rev.
L 4, Section 5.8.4.
- 31. Page 16 Last sentence of Section 4.8-was deleted as the discussion of Doppler uncertainty for the event is described in OPPD-NA-8303-P, Rev. 4, Section -5 8.4.
132. _-Page 16' _ Replace Section 4.10 describing the QUIX code with a.
description _of HERMITE.
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- Table 2 NEUTRONICS DESIGN METHODS AND VERIFICATION OPPD-NA-8302-NP Rev. 03
- 33. Page 17 Change Section 5.0 to reference the CEPAK to DIT conversion and the PDQ-X to MC conversion in previous revisions of OPPD-NA-8302-NP-A. Deleted PDQ-X reference and included reference of the replacement of _QUIX with--
HERMITE.
]
- 34. Page 17 Reduce Section 5.0 to two' sections: 1) The latest 1 benchmarking results since mid-Cycle 11, and 2) a summary of the ongoing verification program,
- 35. Page 18 Section 6.0: ,
- Upgrade Reference 2-3 to the most recent approved- j topical report. -i
- Add HERMITE topical reference to the end of Section '
2.0 References.
- Upgrade Reference 3-2 to the most recent approved topical report.
- 36. Page 19 - Upgrade Reference 4-5 to the most recentLapproved topical report.
- Delete two references, add two references _and renumber.
- 37. Pages 20-27 Update Section 5.0 Tables to reflect most recent -
benchmarking results.
- 38. Pages 28-57 Update Section 5.0 Figures to re;1ect most recent benchmarking results.
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i Lj Omaha Public Power District 'I Nuclear Analysis _
Reload Core Analysis Methodology 1
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Neutronics Design Methods And Verification
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OPPD-NA-8302-NP Rev. 03 J_anuary 1993- .. ,
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ABSTRACT This document is a Topical Report describing Omaha Public Power District's l reload core neutronics desig'n methods for application to Fort Calhoun Station
!- Unit No. 1.
The report addresses the District's neutronics design methodology and its application to the calculation of specific physics parameters for reload l- cores. In addition, comparisons of results obtained using this methodology to results from experimental measurements and independent calculations are provided.
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OPPD-NA-8302-NP,~Reve 03 1
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PROPRIETARY DATA CLAUSE This document is the property of Omaha Public Power District (0 PPD) proprietary information, indicated by brackets, has been removed. This information was developed by ABB-Combustion Engineering (ABB-CE) and Westinghouse Electric Corporation (H). The ABB-CE and Westinghouse information was purchased by OPPD under proprietary information agreements.
l OPPD-NA-8302-NP, Rev. 03 l 11 l
r TABLE OF CONTENTS Section Eagt
1.0 INTRODUCTION
..................................................... 1 2.0 BASIC PHYSICS MODELS ............................................. 2 2.1 Neutron Cross-Sections ..................................... 2 2.2 Di f f u s i on Th eo ry Model s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.1 MC ................................................. 3 2.2.2 ROCS ............................................... 3 2.2.3 HERMITE ............................................ 4 3.0 FORT CALHOUN PHYSICS MODELS ...................................... 6 3.1 Neutron Cross-Sections ..................................... 6 3.2 Diffusion Theory Models .................................... 6 4
3.2.1 MC ................................................. 7 !
1 3.2.2 ROCS ............................................... 8 ;
3.2.3 HERMITE............................................ 8 4.0 APPLICATION OF PHYSICS METHODS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1 Radi al Peaki ng Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.2 Reactivity Coefficients .................................... 9 l 4.3 Neut ron Ki neti cs Pa rumeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 4.4 Dropped CEA Data ........................................... 11 4.5 C E A Ej e c t i o n Da t a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.6 CEA Reactivity ............................................. 13 4.7 CEA Withdrawal Data ........................................ 14 4.8 Reactivity Insertion for Steam Line Break Cooldown ......... 15 4.9 Asymmetric Steam Generator Event Data ...................... 16 4.10 HERMITE Calculations ....................................... 16 5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION ...... 17 5.1 The Di st ri ct 's Ongoing Benchmarking Program . . . . . . . . . . . . . . . . 17 5.2 Summary .................................................... 17
6.0 REFERENCES
........................................................ 18 OPPD-NA-8302-NP, Rev. 03 iii
LIST OF TABLEJ Iable Tit 1e Eage 5-1 Unrodded HZP Critical Boron Concentrations . . . . . . . . . . . . . . . . . . . 20 5-2 Low Power Physics Isothermal Temperature Coef ficients . . . . . . . . 21 5-3 Comparison of Calculated and Measured Isothermal . . . . . . . . . . . . . 22 Temperature Coefficients 5-4 Comparison of Calculated and Measured Power Coefficients ..... 23-5-5 Cycle 12 CEA Worths .......................................... 24 5-6 Cycle 13 CEA Worths .......................................... 25 5-7 Cyc l e 14 C E A Wo rt h s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 OPPD-NA-8302-NP, Rey. 03 iv
LIST OF FIGURES 11111 EAM f.12s1 Cycle 11 Critical Boron Concentration vs Burnup . . . . . . . . . . . . . . 28 5-1 Cycle 12 Critical Baron Concentration vs Burnup .............. 29 5-2 Crcle 13 Critical Boron Concentration vs Burnup . . . . . . . . . . . . . . 30 5-3 31 5-4 Cyrie 14 Critical Gorer Concentration vs Burnup . . . . . . . . . . . . . .
Cycle 11 Magrated Radici Peaking (Fai) vs Burnup ............ 32 5-5 Cycle 12 Integrated Radial Pecki.ng (FaT ) vs Burnup ... ........ 33 5-6 Cycle 13 Integ.'ated Radial Peaking (FaT ) vs Burnup ............ 34 5-7 5-8 Cycle 14 Integrated Radial Peaking (FaI ) vs Burnup ............ 35 5-9 Cycle '.1 Pl anar Radi al Peaking (FxyT ) vs Burnup . . . . . . . . . . . . . . . . 36 Cycle 12 Planar Radial Peaking (Fxy )T vs Burnup . . . . . . . . . . . . . . . 37 5-10 Cycle 13 Pl anar Radial Peaking (Fxy )T vs Burnup . . . . . . . . . . . . . . . 38 5-11 Cycle 14 Planar Radial Peaking (Fxy )T vs Burnup . . . . . . . . . . . . . . . 39 5-12 Cycle 11 ROCS-CECOR RPD Comparison - Axially Integrated . . . . . . 40 5-13 0 13,809 MWD /T Cycle 12 ROCS-CECOR RPD Comparison - Axially Integrated . . . . . . 41 5-14 0 664 MWD /T Cycle 12 ROCS-CECOR RPD Comparisen - Axially Integrated . . . . . . 42 5-15 0 5,946 MWD /T Cycle 12 ROCS-CECOR RPO Comparison - Axially Integrated . . . . . . 43 5-16 010,941MWE,/T Cycle 13 ROCS-CECOR RPD Comparison - Axially Integrated . . . . . . 44 5-17 0719 MWD /T Cycle 13 ROCS-CFCOR RPD Comparison - Axially Integrated . . . . . . 45 5-18 0 7,408 MWD /T OPPD-NA-8302-NP, Rev. 03 v
r LIST OF FIGURES Fioure Title h 5-19 Cycle 13 ROCS-CECOR RPD Comparison - Axially Integrated .. . . . . 46 0 15,249 MWD /T 5-20 Cycle 14 ROCS-CECOR RPD Comparison - Axially Integrated .. . . . . 47 0 1,549 MWD /T 5-21 Cycle 14 ROCS-CECOR RPD Comparison - Axially Integrated ...... 48 0 3,152 MWD /T 5-22 Cycle 11 ROCS-CECOR Comparison - Normalized Axial Power . . . . . . 49 0 13,809 MWD /TCoreAvg.
5-23 Cycle 12 ROCS-CECOR Comparison - Normalized Axial Power . . . . . . . 50 0664 MWD /TCoreAvg.
5-24 Cycle 12 ROCS-CECOR Comparison - Normalized Axial Power ...... 51 0 5,946 MWD /T Core Avg.
5-25 Cycle 12 ROCS-CECOR Comparison - Normalized Axial Power . .. . . . . 52 0 10,941 MWD /TCoreAvg.
5-26 Cycle 13 ROCS-CECOR Comparison - Normalized Axial Power .. . . . . 53 0 719 MWD /T Core Avg.
5-27 Cycle 13 ROCS-CECOR Comparison - Normalized Axial Power . .. . . . 54 0 7,408 MWD /T Core Avg.
5-28 Cycle 13 ROCS-CECOR Comparison - Normalized Axial Power . . . . . . . 55
- O 15,249 MWD /T Core Avg.
l 5-29 Cycle 14 ROCS-CECOR Comparison - Normalized Axial Power .. . . . . 56 0 1,549 MWD /T Core Avg.
l 5-30 Cycle 14 ROCS-CECOR Comparison - Normalized Axial Power ....... 57 03,152 MWD /TCoreAvg.
OPPD-NA-8302-NP, Rev. 03 vi l
OMAHA PUBLIC POWER DISTRICT NEUTRONILS DESIGN METHODS AND VERIFICATION Revision p_ata 00 September 1983 01 November 1986 02 April 1988 03 January 1993 E OPPD-NA-8302-NP, Rev. 03 vii
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l OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY NEUTRONICS DESIGN METHODS AND VERIFICATION
1.0 INTRODUCTION
l This document describes the District's neutronics design calculation methods l along with results obtained by comparing experimental measurements and -l independent calculations. The discussion o' calculational methods-includes descriptions of the basic computer codes and procedures for applying these codes. Comparisons of the calculational methods to experimental measurements and independent calculations used the same codes _and computational methods used in the Fort Calhoun reload core design efforts.
Section 2.0 describes the basic physics models supplied by ABB-Combustio_n:
Engineering (ABB-CE). Section 3.0 details the District's application of these models to the Fort Calhoun reactor. Section 4.0' presents the application of these physics models to the reload core analysis. Section 5.0 discusses the District's latest verification _ program that includes ~ the recent cycle-by-cycle comparisons of District calculated data to_ measured data and j- data from independent calculations. Section 6.0 contains the individual-L references, i ,
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L OPPD-NA-8302-NP, Rev. 03 Page 1 of 57~
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2.0 BASIC PHYSICS MODELS The District's neutronics design analysis for the Fort Calhoun core is based upon a combination of multi-group neutron spectrum calculations. These calculations provide cross-sections appropriately ' averaged over a few broad energy groups and few-group one , two- and three-dimensional diffusion theory calculations that result in integral and differential reactivity effects and power distributions. Computer programs generate calculations embodying analytical procedures and fundamental nuclear data consistent with the most. ,
sophisticated computer programs currently available.
2.1 Neutron tross-Sections Derivation of the data base for both fast and thermal neutron cross-sections is from ENDF/B-IV with changes recommended by the Cross-Section Evaluation Working Group (Reference 2-1). These recommendations consist of changes to the shielded resonance of_ U238, and the Watt fission spectrums of U235 and Pu2 39, and changes in A for U235 and Pu239 The DIT lattice program calculates the few group cross-sections for subregions of the core represented in-spatial diffusion calculations (e.g., fuel pin cells, moderator channels,-
structural member cells, etc.). These cross-sections'are generated as l
a function of fuel temperature and moderator temperature to accommodate' '
the temperature feedback routines within the diffusion theory models.
The DIT code performs all functions of the traditional transport methods that attempt to represent the complexities of the PWR fuel assembly geometry, including neutron energy spectrum interactions in the fuel, control rods, control rod locations (water holes), burnable absorber rods, and incore flux detectors. The essential feature of DIT that distinguishes it from the traditional methodology is that the, spectrum spatial averaging procedures.are based on calculations in '
two-dimensional geometry. Therefore, few approximations.to the geometry representation are necessary. The use of nodal transport theory has made it feasible to retain discrete pin geometry .in both the fine and broad energy group calculations. A complete description of.
the DIT procedures for generating few-group neutron cross-sections is .
in References 2-2 and 2-3.
OPPD-NA-8302-NP, Rev. 03 Page 2 of 57
2.0 BASIC PHYSICS M0CELS (Continued) 2.2 Diffusion Theorv Models The diffusion theory models package used to calculate core physics parameters for Fort Calhoun Station consist of the MC, ROCS, and HERMITE computer codes. The MC (fine mesh) and RCCS (coarse mesh) g codes can be executed in one, two or three dimensions to calculate static and depletion dependent parameters such as reactivity, flux, nuclide and power distributions and CEA worths. The HERMITE code is executed in one dimension to calculate axial power distributions and l
CEA worth [ ]. l 2.2.1 MC The MC program is a fine mesh method used to solve the two group neutron diffusion equation. MC uses the 3-D coarse mesh analysis (ROCS) to recover local information on power, burnup and flux by performing fine mesh, imbedded diffusion theory calculations within the coarse mesh nodes. The capabilities of MC offer a more computationally ef ficient alternative to conventional fine mesh, dif fusion theory computer codes (i .e., PDQ-X) which, in practice, are limited to 2-D core analyses (Reference 2-6). MC also eliminates the PDQ-X problem of representing large gradients near CEA guide tubes (water holes) and burnable absorber pins.
The PDQ-X diffusion theory formulation does not provide the correct flux levels for fuel adjacent to water holes or burnable absorbers. When PDQ-X makes adjustments to absorption a: J removal cross-sections necessary to obtain correct reaction rates for non-fuel cells, results are made worse. By using the imbedded nodal calculational technique, MC removes the need to use PDQ-X type adjustments.
MC employs macroscopic (static) and microscopic (depletion)'
cross-section data generated by methods described in Section 2.1.
2.2.2 R0f1 The ROCS program is a coarse mesh two group solution of the neutron diffusion equation based upon a mesh centered, Nodal- l OPPD-NA-8302-NP, Rev. 03 Page 3 of S7
2.0 BASIC FriYSICS MODELS (Continued) 2.2.2 ROIS (Continued)
Expansion Method (NEM). It incorporates closed channel thermal g hydraulic modeling into its evaluation of the interaction of neutron flux effects and the macroscopic physical and themal properties of distributed materials. Because of MC degndency on ROCS for input information and the ROCS coarse mesh nodal structure, ROCS is more efficient than MC for evaluating a core's static and depletion dependent properties. ROCS also employs macroscopic (static) and microscopic (depletion) cross-sections generated by the methods described in Section 2.1.
A complete description of the ROCS program is in References 2-3 and 2-7.
2.2.3 HERMITE The HERMITE program is a space-time kinetics computer code which was developed by ABB-CE for the analysis of design and off-design transients in large PWR's. The three-dimensional, four group, time-dependent neutron diffusion equation is solved by a finite element method and includes feedback effects of fuel temperature, coolant temperature, coolant density and control rod motion. The heat conduction equation in the pellet, gap and clad is solved by a finite difference method. Continuity and energy conservation equations are solved for the coolant enthalpy and density.
In the HERMITE thermal-hydraulic calculation, the core is modeled
! as a collection of closed, parallel flow channels. Each flow channel is defined by a portion of the cross sectional area of the core. This cross sectional area may represent a part of a fuel assembly, a whole fuel assembly, or a group of fuel l assemblies and corresponds to a specified region of the horizontal plane of the neutronics mesh structure.
1 l In one- or three-dimensional problems where the axial dimension is explicitly modeled, the core is divided into axial nodes which consist of one or more finite element mesh intervals. The power OPPD-NA-8302 -NP, Rev. 03 l Page 4 of 57
2.0 BASIC PHYSICS MODELS (Continued) 2.2.3 HERMITE (Continued) generated in each axial node of each channel is determined during the neutronics calculation. Additional information on the structure of the HERMITE model can be found in Reference 2-10.
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OPPD-NA-8302-NP, Rev. 03 Page 5 of 57
3.0 FORT CALH0UN PHYSICS MODELS The District uses the basic ABB-CE physics models described in Section ?.0 to model the Fort Calhoun reactor core. The computer codes (in executable form only) that embody these basic physics models are maintained on the OPPD Nuclear Engineering UNIX Workstation Network located at the Fort Calhoun j Administration Building. The District has obtained access from ABB-CE to use l I
worsstation versions of the ABB-CE physics computer codes. AdB-CE maintains all documentation and quality assurance programs related to these workstation computer codes. The following paragraphs oiscuss the specifics of the Fort Calhoun models.
3.1 Neutron Cross-Sections The two-group neutron cross-sections used in the ROCS and MC models of the Fort Calhoun reactor core are generated using the DIT code.
Cross-sections have been generated for unshimed W and ABB-CE fuel assemblies as well as shimed ABB-CE fuel assemblies and W IFBA fuel assemblies. Cross-sections have also been generated by ABB-CE for fuel containing the full-length hafnium flux suppression rods and for the H natural uranium fuel assemblies. The cross-sections have been generated for the District by ABB-CE and are based on information supplied by the District.
The cross-sections used to model the Fort Calhoun reactor are in the form of universal table sets. The two group cross-sections are generated as functions of enrichment, fuel temperature, moderator temperature, burnup and, in the case of shimed fuel assemblies, B C4 shim number density. The table sets are applicable over a fuel temperature range from room temperature to 1800 K and a moderator temperature range from room temperature to 600 K. The fine mesh table sets include explicit treatment of the pin cells imediately around the CEA guide tube (water hole) to properly account for the peaking of thermal flux in these water holes. Therefore, no corrections need be applied to the pin powers produced by the fine mesh model.
3.2 Diffusion Theorv Models The District uses the MC, ROCS and HERMITE models described in Section 2.0. The District uses both two-dimensional and three-dimensional l
ROCS /MC models. The HERMITE model is a one-dimensional model. l OPPD-NA-8302-NP, Rev. 03 Page 6 of 57
3.0 FORT CALHOUN PHYSICS MODELS (Continued) 3.2 Diffusion Theorv Models (Continued) 3.2.1 tic The District's MC model is a two-group fine mesh model that generates its solutions based upon a two-dimensional depletion in the x y plane. MC uses imbedded fine mesh calculations within
' coarse mesh ROCS nodes to produce solutions in the axial (z) dimension. Each fuel pin cell and shim pin cell is represented by a single mesh point. The model also includes explicit representation of the CEA guide tube (water holes). The macroscopic cross-section to node assignment is located in a geometry file that also provides shim loadings and uranium metal weight for depletion calculations. The model is representative of the core between 15% and 85% of full core height.
The MC model is used to simulate the expected modt of operation in the cycle being analy7ed. The calculations result in material distributions and radial peaking factors that are used in the safety analysis and setpoint generation. Unlike PDQ-X, MC uses exposure and geometry information to create a library of pre-calculated coefficients for the incore monitoring system.
The mode of operation at the Fort Calhoun reactor is base loaded operation. Base loaded operation consists of reactor operation at or very near rated thermal power throughout the cycle. The lead CEA bank insertion is held to a minimum.
Historically the lead CEA bank at Fort Calhoun has been inserted less than 5% of the time whenever the reactor is at a steady power level. Reference 3-1 discusses the impact of operation with a time averaged lead bank insertion of [ ) . Due to its dependency on ROCS coarse mesh information, the model must be depleted in the same number and magnitude of ROCS time steps, which are typically depleted in time steps of 1,000 MWD /MTV.
To use MC beginning with the District's Cycle 12 Reload Lice v a Submittal, MC prediction calculations were benchmarked aga ost OPPD-NA-8302-NP, Rev. 03 Page 7 of 57
3.0 FORT CALHOUN PHYSICS MODELS (Continuad) 3.2 Diffusion Theorv Models (Continued) 3.2.1 MC (Continued)
PDQ-X prediction calculations for Cycles 10 and 11. An overview of the benchmarking results is reported in Section 5.0.
3.2.2 RDCi The District uses a three-dimensional and a two-dimensional two-group ROCS model . [
] The two-dimensional model is representative of the core between 20% and 80% of full core height. [
] The boundary conditions are derived in accordance with the methodology discussed in Reference 3-2.
3.2.3 HERMITE The District uses a 125 axial node HERMITE model. The data for the HERMITE model is obtained from the three-dimensional ROCS calculations.
OPPD-NA-8302-NP, Rev. 03 Page 8 of 57
4.0 APPLICATION OF PHYSICS METHODS Previous sections have focused on the reactor physics models used by the District to model the Fort Calhoun reactor. In this section, calculations of the various core parameters used in the safety analysis are described. The primary core parameters considered are the radial peaking factors (Fa and i Fxj), the moderator temperature coefficient, the fuel temperature or Doppler coefficient, the neutron kinetics parameters, CEA drop data, CEA ejection j data, CEA scram reactivity, reactivity insertion for the steamline break cooldown, radial peaking data for the asymetric steam generator event, and axial power distributions. Appropriate biases and uncertainties that apply to the parameters are listed in Reference 4-1.
4.1 Radial Peakina Factors The radial peaking factors, Fa and Fxy, are calculated using the MC and 3-D ROCS models. Values of Fa and Fxy for both unrodded and rodded core configurations are obtained directly from the MC power distribution.
Since the MC model uses a pin power correction edit implicitly account.ng for the peaking of the thermal flux in the CEA guide tubes (water holes) no correction is required to the peaking factors calculated by MC. The values of Fxy and Fa in MC for unrodded and g rodded cores are reported as core peaking edits. MC reads in the 3-D ROCS power histories, calculates Fxy for the plane of depletion and then synthesizes the calculation plane by plane to obtain the maximum core Fxy. MC also uses the 3-D ROCS power information to calculate Fa based on the axial integration of the planar power distribution obtained from ROCS. The measurement uncertainties for the radial peaking factors are given in Reference 4-2. l The physics models are used to calculate the expected values of Fn and Fxy. The actual values of FR and Fxy used in the safety analysis are chosen to conservatively bound those anticipated during the core life. E 4.2 Reactivity Coefficients The ROCS models are used to calculate the moderator temperature coefficient (MTC) and the fuel temperature coefficient (FTC). The MTC i defined as the change in reactivity per degree change in moderator OPPD-NA-8302-NP, Rev. 03 Page 9 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued)
I 4.2 Reactivity Coefficients (Continued) temperature. Calculationally, the MTC at a temperature of Tnod is determined by performing two calculations; one at Tnog(Tin +15 F) a i one at Tnod(Ti n -15 F) . The MTC at a temperature of Tmod is the reactivity change divided by the temperature change over this interval. The reactivity change is calculated with the ROCS model by varying the inlet temperature while holding all other parameters such as the fuel temperature and nuclide concentrations constant.
The FTC or Doppler coefficient is defined as a change in reactivity per degree change in the effective fuel temperature. The effect of fuel temperature upon resonance neutron energy absorption is accounted for in the ROCS /MC models by means of power feedback options. The representation of the variation in the few-group cross-sections with fuel temperature involves two main segments. The first segment represents the variation in cross-section with fuel temperature; the second segment relates fuel temperature to reactor power density. The first portion is included in the basic methods employed to generate the few-group cross-sections. The second portion requires establishment of correlations between fuel temperature (i.e., effective fuel temperature to be used in generation of cross-sections) and the reactor power density. The relationship between fuel temperature and reactor power density employs direct fits to FATES (Reference 4-3) fuel data. This g method results in the fuel temperature correlation for each fuel type that is both local power density and fuel exposure dependent.
The reduction in reactivity resulting from an increase in effective fuel temperature is determined by ROCS. Typically runs are made at 160,130, 70 and 40 percent power to get a series of points of fuel temperature versus reactivity while holding mo ator temperature and density and nuclide concentrations constant. n,ese points are then fitted to a linear curve to determine the FTC at any fuel temperature.
The biases and uncertainties for the FTC are given in Reference 4-1.
The physics models are used to calculate the expected values of the MTC and FTC throughout the cycle. The actual values of the MTC and FTC l OPPD-NA-8302-NP, Rev. 03 Page 10 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.2 Reactivity Coefficients (Continued) used in the safety analysis are chosen in conservatively bound expected values of these parameters. The measurements of the MTC made during g the operation of the reactor include uncertainties to assure that the actual MTC does not exceed the values used in the safety ana?ysis.
4.3 Neutron Kinetics Parameters The neutron kinetics parameters p, A and the neutron lifetime,1*, are !
calculated using ABB-CE's ROCS computer code. The technique used to- -j calculate the kinetics parameters and the neutron lifetime is based on-first order perturbation theory. Details of the perturbation approach are discussed in References 4-4 and 4-5 [ j
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4.4 Drooned CEA Data L
! The neutronics data unique to the dropped CEA analysis are tne values of Fa following the drop of a CEA and the reactivity worth of the dropped CEA. The values of Fa increase due to a -large azimuthal tilt caused by the drop of a CEA and occur on the side of_the core' opposite-the dropped CEA. Because the maximum Fa occurs far away from the g dropped CEA, the intra-assembly power distribution is not perturbed.
Therefore, the " post drop" value of Fa can be calculated by multiplying the ' pre-drop" values of Fa by the ratio of the assembly power after and before the drop of the CEA. This ratio is the distortion factor. -The distortion factor is defined as the ratio of the assembly RPD at a given power level and time in core life containing a dropped CEA to the same assembly RPD without a dropped CEA.
The distortion factor and dropped CEA reactivity worth are calculated-
, using the 3-D ROCS model. The 3-D Fa distortion factor is calculated for a specific CEA insertion and power level. The " post drop" valu'e of Fg using the 3-D Fg distortion factor is calculated by multiplying the OPPD-NA-8302-NP, Rev. 03 Page 11 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.4 Q.rapp_ed CEA Data (Continued)
" pre-drop" value of Fa for the particular CEA insertion and power level by the 3-D Fa distortion factor. The 3 J ROCS " post drop" power distributions are calculated with fuel temperature and moderator temperature feedback. The calculations assume that the core average Axial Shape Index (ASI) is being controlled within the " constant ASI" limits in accordance to the Fort Calhoun Operating Manual.
Biases and uncertainties associated with *be calculation of the distortion factors for the CEA Drop analy31s are located in Reference 4-1.
4.5 CEA Eiection Data The neutronics data unique to the CEA ejection analysis are the values pre-ejected and post-ejected radial peaking factors and the reactivity worth of the ejected CEA. The maximum post ejection radial peakir.g factor and maximum ejected CEA reactivity worths are calculated for the maximum CEA insertion allowed by the PDIL at HFP and HZP. The neutronics parameters are calculated using HFP and HZP 2-0 ROCS and MC models. The post ejection radial peaking factor is calculated by multiplying the 2-D ROCS post ejection assembly RPD by the corresponding pin to box ratio from MC. The ejected CEA reactivity worth is obtained directly from ROCS calculations. ROCS post ejection power distributions are calculated without moderator or fuel temperature feedback.
The post ejection value of Fqwhich is obtained using MC, is calculated by moltiplying the post ejection value of Fxy by the maximum value of F2 , the azimuthal tilt allowance, the augmentation factor, tne engineering heat flux factor, the fuel densification factor, and the F q uncertainty documented in Reference 4-2. Biases and uncertainties for the ejected CEA worth are located in Reference 4-1.
OPPD-NA-8302-NP, Rev. 03 Page 12 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.6 CEA Reactivity The CEA reactivity calculations done in a reload core satety analyses are the calculation of the total reactivity of CEA's inserted into the core during a reactor trip (CEA scram reactivity), the generation of the scram reactivity curves, and the calculation of required shutdown margin.
The CEA scram reactivity worth at HZP is calculated by obtaining the net worth for all CEA's between the HZP PDIL CEA position and the fully inserted position and subtracting the worth of the highest worth stuck CEA. These calculations are done using the ROCS model. The HZP CEA g scram reactivity for the CEA ejection transient is calculated in a similar fashion except that the worth of the ejected and highest stuck worth CEA's are subtracted from the net worth.
The scram CEA worth at HFP is calculated by obtaining the HFP net worth for all CEA's between the HFP PDIL CEA position and the fully inserted position, subtracting the worth of the highest worth stuck CEA and subtracting the moderator void collapse allowance. The thermal hydraulic axial gradient reduction allowance and the loss of worth between HFP and HZP are also subtracted from the HFP net worth for the scram CEA worth to be used in all transients except the four pump loss of flow event and the steamline break incident. These are not applied to the four pump loss of flow scram CEA worth because the closest approach to the SAFDL during the four pump loss of flow event occurs prior to significant CEA insertion. These allowances are not applied to the steamline break (SLB) incident HFP CEA scram worth because the HFP SLB reactivity insertion curves implicitly account for these effects.
The axial thermal hydraulic reactivity effects for HFP to HZP transients are accounted for by measuring the change in reactivity resulting from collapsing a 3-D distributed moderator temperature to a 3-D flat moderator temperature profile. This is done by setting the power in a ROCS case to HZP from a HFP input file and allowing only 0 PPD-NA-8302-NP, Rev. 03 Page 13 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.6 CEA Reactivity (Continued) fuel temperature to provide feedback. In the next case the core average moderator temperature from the HFP case is input to the HZP case as the inlet temperature and the fuel temperature are frozen, only allowing the moderator density and temperature to provide feedback.
The bounding values at both BOC and E0C are used for calculation of scram worths for transient analyses.
The generation of the scram reactivity curves uses the methodology discussed in Reference 4-7. l The calculation of the required shutdown margin is only performed at HZP since the shutdown margin at power is controlled by the PDIL. The available HZP shutdown margin is equivalent to the HZP CEA scram reactivity. Biases and uncertainties for CEA Reactivity are listed in
' Reference 4-1, 4.7 CEA Withdrawal Data The reactor core physics data unique to the CEA withdrawal analysis 'is the maximum differential CEA worth. This is the maximum amount of j reactivity at any time in core life that can be added to the core per inch of CEA motion. [
]
This calculation produces a more conservative estimate of the expected maximum differential worth than previous methods.
This maximum differential worth is then combined with the maximum CEA withdrawal rate of 46 inches / minute to arrive at the maximum reactivity insertion rate.
OPPD-NA-8302-NP, Rev. 03 Page 14 of 57
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.8 Reactivity Insertion for Steam Line Break Cooldown The reactor core physics data unique to the steam line break transient analysis is the reactivity insertion due to the cooldown of the moderator. There are two sources of this reactivity insertion. The first is the positive reactivity insertion due to the increasing density of the moderator as the cocidown progresses. The second is the reactivity insertion due to the Doppler coefficient as the effective fuel temperature changes.
Reactivity insertions due to the moderator density increase and the Doppler coefficient are both calculated using a full core ROCS model.
The axial leakage or buckling is adjusted such that the moderator temperature coefficient calculated by the ROCS model corresponds to the most negative Technical Specification limit. The reactivity insertion calculations are performed with all CEA's except the most reactive CEA inserted in the core.
The moderator density reactivity insertion curve for the hot zero power
( steamline break case is calculated by successively lowering the inlet temperature of the ROCS model from 532 F and allowing only moderator temperature feedoack in the model. The calculations typically result in a curve of reactivity insertion versus moderator temperature from a hot zero power temperature of 532 F to 212 F.
The Doppler reactivity insertion for the hot zero power case bounds the value of the reactivity insertion cticulated by ROCS for the FTC. The fuel temperature feedback in the mode' allows the production of a curve of Doppler reactivity as a function of fuel temperature. All zero power calculations are performed assuming there is no decay heat and no credit is taken for local voiding in the region of the stuck CEA.
The moderator density reactivity insertion curve for the full power case is calculated by decreasing the power level and core average average coolant temperature from full power to the hot zero power inlet temperature and then successively lowering the inlet temperature as in OPPD-NA-8302-NP, Rev. 03 Page 15 of 57
I 4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.8 Egantivity Insertion for Steam Line Break Cooldown (Continued) l the hot zero power case. Only moderator temperature feedback is used in the ROCS model. 5 Since the moderator reactivity insertion curve corresponds to an MTC that is at the Technical Specification limit, no additional uncertainty is added to this curve. l 4.9 Asymmetric Steam Generator Event Data l
The reactor core physics data unique to the asymmetric steam generator event [
] For the range of temperatures considered, the intra-assembly peaking does not vary as the inlet temperature is changed. [
l 1
l
[
i 3
4.10 HERMITE Calculations l
The District uses the HERMITE model to perform various axial shape analyses related to the generation of the reactor protective system j setpoints. The HERMITE calculations are carried out according to the methodology discussed in Reference 4-7.
l OPPD-NA-8302-NP, Rev. 03 Page 16 of 57
1 I
5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION l The District has performed extensive verification of the neutronics models used in the reload core analyses. The results of the previous District verification efforts were reported in References 5-1, 5-2 and 5-3. These effo'rts consisted of: 1) using cross-sections produced by CEPAK and DIT and ,
confirming the District's ability to use the models with DIT cross-sections, and 2) the benchmarking of the fine-mesh, imbedded nodal diffusion theory code, MC. The verification of replacing QUIX with HERMITE is documented in Reference 5-4. Extensive verification of the use of DIT cross-sections was also done by ABB-CE and reported in Reference 5-5.
5.1 The District's Onacino Benchmarkina Program The data reported in this section consists of historical and current information for the District's ongoing benchmarking program.- The historical data, collected through mid-Cycle 11, is reported in Reference 5-3. The verification program has been updated through mid-Cycle 14. Tables within Section 5 have been updated to include only recent cycle by cycle information (Cycles 11, 12, 13 and 14).
Verifications of program segments include information consisting of startup physics testing predictions, reactor testing analysis and a core follow effort. This program will continue to provide verification data in the future.
5.2 Summary The District has an ongoing neutronics methodology verification program. The results of this verification program for previous cycles demonstrate the ability of the District to use the neutronics methods described in this document.
OPPD-NA-8302-NP, Rev. 03 Page 17 of 57
6.0 REFERENCES
Section 2.0 References 1
2-1 ENDF-313 " Benchmark Testing of ENDF/B Data for Thermal Reactors, I
Archival Volume," July,1981.
2-2 A. Jonsson, J. R. Rec and U. N. Singh, " Verification of a Fuel Assembly Spectrum Code Based on Integral Transport Theory," Trans.
Am. Nucl . Soc. , 28, 778 (1978). ,
2-3 CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," j Ap ri l , 1983.
2-4 System 80 PSAR, CESSAR, Vol. I, Chapter 4.3.3, Amendment No. 3, June 3, 1974.
2-5 W. R. Cadwell, "PDQ-7 Reference Manual," WAPD-TM-678, January,1968.
2-6 S. F. Grill, A. Jonsson and M. W. Crump, "Recent Developments in the ROCS /MC Code for Retrieving Local Power Information in Coarse-Mesh Analysis," CNS/ANS International Conference on Numerical Methods in Nuclear Engineering, Montreal, Canada, September 6-9, 1983.
2-7 T. G. Ober, J. C. Stark, I. C. Richard and J. K. Gasper " Theory, Capabilities, and Use of the Three Dimensional Reactor Operation and Control Simulator (ROCS)," Nucl . Sci . Eng. , 64,-605, (1977).
2-8 System 80 PSAR, CESSAR, Vol. 1, Appendix 4A, Amendment No. 3, June 3, 1974.
' 2-9 CENPO-199-P, Revision 1-P-A, "CE Setpoint Methodology," January 1986.
2-10 CENPD-188-A, "HERMITE, A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," March, 1976.
Section 3.0 References 3-1 CENPD-199-P, Revision 1-P-A, "CE Setpoint Methodology," January 1986.
3-2 CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design,"
April , 1983.
l OPPD-NA-8302-NP, Rev. 03 Page 18 of 57 1
I
6.0 REFERENCES
(Continued) ]
Section 4.0 References L
l 4-1 CE-CES-129 Rev.1-P, " Physics Biases and Uncertitinties," U. Decher -
1 ABB Combustion Engineering Nuclear Power,' August 2,1991.
4-2 CENPD-153, Revision 1-P-A," INCA /CECORPowerPeakingUncertainty,"
! May, 1980.
4-3 " Development and Verification of a Fuel Temperature Correlation for g Power Feedback and Reactivity Coefficient Application," P. H. Gavin and P. C. Rohr, Trans. A Nucl . Eqq. 30, p. 765,1978.
l i
4-4 A. F. Henry, " Computation of Parameters Appearing in the Reactor g.
Kinetic Equations," WAPD-142, December 1955.
4-5 R. W. Hardie, W. W. Litke, Jr. , " PERT-V, A Two Dimensional Perturbation [
Code for Fast Reactor Analysis," BNWL-1162.
4-6 CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design,"
April , 1983.
4-7 CENPD-199-P, Revision 1-P-A, "CE Setpoint Methodology," January 1986.
Section 5.0 References l.
- i. 5-1 CEN-242-(0)-P, OPPD "cgonses to NRC Questions on Fort Calhoun Cycle 8, l February 18, 1983.
! 5-2 " Reload Core Analysis Methoualogy, Neutronics Design Methods and Verification," OPPD-NA-8302-P-A, Rev. 01.
5-3 " Reload Core Analysis Methodology, Neutronics' Design Methods and l Verification," 0 PPD-NA-8302-P-A, Rev. 02.
a-4 CENPD-188-A, "HERMITE, A Multi-Dimensional Space-Time Kinetics Code for -
PWR Transients," March,1976.
5-5 CENPD-153-P, Revision 1-P-A, " INCA /CECOR Power Peaking Uncertainty,"
May 1980.
OPPD-NA-8302-NP, Rev. 03 Page 19 of 57
TABLE 5-1 Unrodded HZP Critical Boron Concentrations cycle feaSured Dpm 3-D ROCS-DLI
~ ~
12 1500 13 1568 14 1182 OPPD-NA-8302-NP, Rev. 03 Page 20 of 57-
TABLE 5-2 Low Power Physics Isothermal Temperature Coefficients Measured ITC ROCS-DIT ITC Cycle (x 10-4 Ao/ F) (x 10-4 Ao/ F) 12 0.24 13 0,32 14 -0.09 _ _
l OPPD-NA-8302-NP, Rev. 03 Page 21 of 57
TABLE 5-3 Comparison of Calculated and Measured l Isothermal Temperature Coefficients ,
l BOC Critical Boron Calculated Percent of Concentration Measured ITC ROCS-DIT ITC Cycle Rated Power (com) ix 10-4 Ao/ F) (x 10-4 Ao/ F) 12 93 1050 -0.53 13 94 1113 -0.51 14 90 768 -0.87 _ _
E0C Critical Boron Calculated Percent of Concentration Measured ITC ROCS-DIT-ITC Cycle Rated Power (opm) (x 10-4 Ao/ F) (x 10-4 Ao/ F) 12 95 309 -1.79 13 95 325 -1.69 NOTE: Full Rated Power = 1500 MWt OPPD-NA-8302-NP, Rev. 03 Page 22 of 57
TABLE 5-4 Comparison of Calculated and Measured Power Coefficients Critical Boron Calculated Burnup Percent of Concentration Measured PC ROCS-DIT PC frcle MWD /MTU Rated Power 1922). (x 10-4 Ao/ F) (x 10-4 Ao/ F) 12 425 93 1050 -1.42 12 9691 95 309 -1.63 13 373 94 1113 -1.26 13 10694 95 325 -1.51 14 355 90 768 -1.55 _
I NOTE: Full Rated Power = 1500 MWt
't 4
OPPD-NA-8302-NP, Rev. 03 Page 23 of 57
_ _ . .__.__________________________--_-_----_1 -
TABLE 5-5 Cycle 12 CEA Worths Calculated 3-D ROCS-DIT Gr_qqa Measured (%Agl _M A 1.92 B 1.52 4 0.61 3 0.56 2 0.80 1 0.63 Total 6.04 __ ;
l l
l u
l l
l l
l l
I-l I
I l --
i l
l OPPD-NA-8302-NP, Rev. 03 Page 24 of 57
TABLE 5-6 Cycle 13 CEA Worths Calculated 3-D ROCS-DIT Grsg Measured (%Aoi _ %Ap)_
(
A 1.80 8 1.56 4 0.45 3 0.73 2 1.12 l
1 0.81 Total 6.47 __ __
l OPPD-NA-8302-NP, Rev. 03 Page 25 of 57
TABLE 5-7 Cycle 14 CEA Worths Calcul 'ed 3-D ROCS-DIT
~'
Etqua Measured (Mal _ %Ap),
(
A 1.60 B 1,64 4+3 1.25 2+1 1.53-Total 6.00 OPPD-NA-8302-NP, Rev. 03 Page 26 of 57
This Page Intenti_onally left-Blank
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CYCLE 11 C R ~~ CA 30 RO \ C0 \ C E \~~ RK 0 \ vs 3U R \ U 3 BORON CONCENTRATION (ppm) 1300 1200- ,
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0 1000 2000 3000 4000 5000 6000 -7000' 8000 9000 10000 11000 12000 13000 14000 BURNUP (MWD /MTU).
Fiaure 5-2 C R ~~ CA _ 30 R0 \ CO \ C E \~~ RA~~ 0 \ vs BU R N U 3 CYCLE 12 BORON CONCENTRATION (ppm) 1200 1100.
Y%
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5 900 g A h*
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0 1000 -2000 3000 4000 5000- 6000 7000 8000 9000. 10000 11000 12000 13000 14000.-
BURNUPL(MWD /MTU)
(
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Figure 5-3 .
CYCLE 13 CRITICAL BORON CONCENTRATION vs BURNUP g
BORON CONCENTRATION (ppm) 1200 .,
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. %., a n*!'
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Figure 5-4 ;
CYCLE 14 CRITICAL BORON CONCENTRATION vs BURNUP 900 .
A A A g
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Figure 5-6'- .
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CYCLE '11 .
\~~EGRA~~E J RAJ. A_ 3 EA< .\ G vs 3U R N U? l i
RAD PEAKING FACTOR-(FRT) 1.9 PREDICTED
-ACTUAL. -
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.~ , , -_ _ _ _ _ _ . .
Figure 5-6 CYCLE 12 LV AX V U V \~~EG RA~~E J RAJ A_ 3EA< NG vs 3U R \ U?
1.90 RADIAL PEAKING FACTOR (Fr)
PRE 0lCTED (HFP)
ACTUAL
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Figure 5-7 ,
4 CYCLE 13 MAXIMUM INTEGRATED RADIAL PEAKING vs BURNUP a RADIAL PEAKING FACTOR (Fr) g ACTUAL
~
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$ 1.75
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Figure 5-9' CYCLE 11 3
_ A\AR RAJ A_ 3 A< \G vs 3JR\ J3 L
. RAD PEAKING FACTOR (FxyT)
- 1.9
-- PREDICTED 4
ACTUAL l- 1.85 3
- 1:
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.3
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BURNUP (MWD /MTU)- t
- , , . - . ,m.- . . .. ;,,-._'.,... . .. ._a__________________.___ . _ _ _ . . _ _ _ __.- _ _ _ _ _ _
Figure 5-10 CYCLE 12 V AX. ViU V 3 _A\ AR RAJ A_ 3EA< \ G vs 3U R \ U 3
, ,g g RADIAL PEAKING FACTOR (Fxy)
--- PREDICTED (HFP)
ACTUAL 1.85 1.80
.aI-
? T 1.75 h
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--_____ __ _p 1.60 1.55 1.50 '
0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 12000 13000 14000 BURNUP (MWD /MTU) e
Figure 5-11 ,
CYCLE 13 MAXIMUM PLANAR RADIAL PEAKING vs BURNUP RADIAL PEAKING- FACTOR (Fxy)
ACTUAL 1.85 .---=-
PREDICTED (HFP) 1.80 o 1.75 4 '
?
j E 1.70
%b
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- i. 1.40 - - - - - - - - - - - -
- O 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 12000 13000 14000 15000 16000 BURNUP (MWD /MTU)
Figure 5-12 ,
i -
CYCLE 14 1.90 slAXIMUM Fxy vs. BURNUP ;
1.85
- r i
- - 1 ;
, 8 ' .80
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TECH SPEC ACTUAL
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-11 .12 13 14 CYCLE 14 Rt1RNilP fx 1 non MWD /MTII) w&.pJ m- g-+%+ e r-- w w-- 9 ,-m - g-W- . __ . _ _ _ _ _ 2_ _ _ . . _ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _
l Figure 5-13 I
OMAHA INSI TYP8 CY11 BOX13 603 WINST 02.Y ROCS-CECOR COAPARISON AX] ALLY INTEGRATE 1
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- o e o e 027s OM e o e e e-2.201*
e 99ee OM/100 e OJe 10126L 1.180s 102e OL/10325K/104 .938' e .000o OL/10524L 106 OJs10723M/108 O e
- .009e *- e 1.343e o e o a 8 .762* e-1 .575 015e e .000* e .054*
- e .000* e 4.012e e 28Je110 e
.000* OM
- 111e* OM/112 e 1 OK/11327Le114
- OM/115 OL*116 OK/117 OM/11
.000* *
- e
.000e e
e
. . 167
.008* e e e e
e e
e e 662 e e e e e e e
e e e 123 eOM/124 e OL/125 e OM/126e OMe 127 OM 128 OJe129e e o e e e e e e e e e e e e e e e e e
- OL 130* OK/131e OJ 132eOL 133*e e e e e e e e e e e BASE CASE VALUES CORP CASE VALUES AVGa 1.00000ST0= .25262 MAXIMUM VALUE 1.34317AT 23 AVG = 1.00000STDa .24908 MINIMUM VALUE IS MAXIMUM VALUE 1.39705AT 8 42432AT 1 MINIMUM VALUE IS 46792AT 11 NUuBER OF SAMPLES 23 Of GREES OF FREEDOM 22 ABSOLUTE DIFFERENCE AVG =
PCT DIFFERENCES
.00000$1D= .03280 VAxtuuM VALUE .00004AT 6 AVG =
.29888STD= 4.18328 MINIMUM VALUE IS .04586AT 5 MAXIMUM VALUE 13.68315AT 1 MINIMUM VALUE IS -3.83068AT 10 OPPD-NA-8302-NP, Rev. 03 Page 40 Of 57
l i
i I
Figure 5-14 l i
i I
l CMAHA INSTCY12 TYPE664 DOXWD/TINST ONLY ROCS-CECOR COMPARISON AXIALLY INTEGRATED !
BASE CASE ABS Olff PCT O!FF OM 1 OK 2 CK 3 OM 4e e e e e e e e e e e e e e e .
OJe 5 ON 6 ON e
e e
e e
e 7ee ON/ 8ee OM/ 9ee ON/ 10 e e
ONe 11 ON 12.OJe 13e e e e e e e e e i e e e e e t e OJe 14e ON 15e ON/ 16 e OL 17 2M/ 18 ON/ 19* OM/ 20 000* e OL 2104/ 22*ON 23 1Je 24 e e e
- 387*
e e e .000*
- e e e e e e e .000e e e e
e e .037e e
e-9.Si4e e ON 25 GN/ 26 OL/ 27 EM 28 OM/ 29 4L/ 30 OM/ 31 3M 32 OL/ 33 ON/ 34 ON 35' e 1.326e e 1.329e .952s e 1.319e e
- e * .023e * - e e e-1,021. e ,000e e ,013e e e
'o e-1. 718e .567+ e .048* * .957e e e
ON 36 Ot. e e 10M 47 e e e
37ee OM 38e OM/ 39 e OL 40.ON/ 41 e OL 42 e OM/ 43
- ON 46*
e e e e e . .. e e 446e *
- e e *
- e OM 48
- e e * * * *
.009 ON/ 49 1,022e e-2.013e
- DM/ 142 50 OM/ 51 OLe52 OM 53,943*
e 8L 541.188e 7M 55 OL 56 e OM/e 57 OM/ 58 ON/ 59e o OK 60e e . e e . .027e .002* *e e e e e i e
e e e 1.907 e e e 2.884e .207* *
- e
'11K 61 '
e 17M/900* 62 ON/ 6316L/ 64 ON/ 6515L 66 OL e .288e e * .934e .010e OK 73 .007e e .018e e
e 982e 6714L e -
.974e68 ON/ 6913L/
e .931 70.ON/ e 7112M/
.890e 72e 3.480*
.011e e .003e . .022e e
e .829
- 1.949e e-1.114e * .317
- 2.342e e .017 OK 74 e 1.902e e e
- ON/ 75
- OM/ 76
- OM/
OM 86e e e e - .006*
- e e e e
= e e o e-1,01Se
.286e'
.013e e e OM 87 e
. .571* e-1.057+ . . e e e 22N e 1.151*88 OL 89 *OM 90 *OM/ 91*OL 92 ON/ e 93 *OL 94 OM/
- 95 *OM . 96
- OL 9721N 98e o e 022e o e e- * *
- e 1.892e e e a e e e e e e 1.145.*
031 e
ON =
- 2.722e e 99o ON/100 e OL/10126M e 1.315e 102 OM/10325L/104 e .956e OM/10524M e 1.303* 106 OL/10723N/108 ON 109e e e e e .008e e .003* ** 1.3278 e
- - e e .631 e .005* -
- e 344 . 420e e-1,024
.811e e 28Je110 e
372* ON 111 e ON/112 e OL*11327M/114 1.148* ON/115
- OM/116 e OL 117 e
ON/118 ON 119 OJe120s
.022e e e * .016e e e o e e e e-5.920e * *
- 1.423e e e e e e e e CJe121.ON 122 04123 eON/124 eOM/125e ON/126e ON 127e ON 128 OJe129 e e e e e e e e .
e e e
. e e
. . e e . . e OM 130' OK 131 OK 132 OM 133e e e . e e e e e e e
l BASE CASE VALUES COMP CASE VALUES AVG = 1.00000STD= .30607 'AVC= 1.00000STD=
MAXIMUM VALUE 1.32914AT 5 .30759 MAXIMUM VALUE 1.30830AT S MINIM S VALUE IS .28809AT 11 MIN! MUM VALUE IS .29812AT 11 NUMBER OF SAMPLES 27 DECREES OF FREEDOM 26 ABSOLUTE DIFFERENCE PCT DIFFERENCES AVC= .00000STD= '.01803 MAXIMUM VALUE AVG = .22552STD= 2.72936 MINIMUM VALUE IS.03116AT .03681AT 211 MAX) MUM VALUE 3.47994AT 11 MINIMUM VALUE IS- ~9.51426AT 1 OPPD-NA-8302-NP, Rev. 03 Page-41 of 57
l J
Figure 5-15 OMAHA CY12 INST. TYPE 5946 BOX MWD / INST ONLY ROCS-CECOR COMPARISON AX] ALLY INTEGRATED BASE CASE ABS 01FF PCT O!FF 4 l
OM 1 OK 2 OK 3 OM ~ 4e e e e e e e e e e e e e e e e OJe 5 ON 6 ON 7 ON/ 8 eOM/ 9 ON/ 10 eON 11 eON 12 0Je 13e e e e e e e e ,
e o e e . . e e e e e e .e e e e e e e
OJe 14e ON 15 ON/ e 16 OL . 17 2M/ 18 ON/ 19 OM/ 20 e OL. 21 ON/ 22 ON 23 IJe 24*
- e e e e 000, e. e e -397 e
e e
e e
e . .000 e e e e .
,017e e e .000s e e e e-OM 25 6N/ 26 OL/ 27 SM1.283.OM/ 28 29 e
e 1.346* *
- AL/ 30.OM/ 31 3M 32 OL/ 33 948 1.269e e ON/ 34e ONe 35 -4. 23 7,.
e e e .0D6 . ,033e e .010e e 021, e -e e '
e e 409e e-2.578e e 1.099 e-1.622 . - e *-
y e ON 36ee OL 37 eOM 38.OM/ 39 OL 40
- ON/ 41
10M 47 e- e e e e * * *
- e 491 e e . . e
- 1.154e .930.7M 55 1.156e OL 56 e OM/ 57 OM/ 58.ON/ 59 -
.e OK 00e e
e 018e e e e .027. .015e e
- e e e t.563 . . e e 2.881.-1.337* * ,
e
-11K 61- 1 e . .338 '
e 17M/
. .97562.ON/ 6316L/9348 64 ON/ 6515L960s 60 OL 6714L e- .957*- 68 ON/ 6913L/ 70 ON/ 71.2M/ 72e . 006 * .-
OK' 73e 001e e 024 e .935e .966e 1.786e-e e .003e e .000, e .023e- * .008 OK 74 e .112e e 2.558e e - e .011e
- 2.433 ON/ 75 *OM/ 76 *OM/ 7720L 78 7919L 80
.812e .*
- e 1.149.OM 364e
- OM BGe * * * -
- 948 OM 81181. e 1.137e82 OM/ 83e OM/ 84 e ON/ 85* .e e e e .e-1 012,e e 008e * .011e *
- OM 87 e
e
.034 e .877 e .953* * *
e 95-OLe 92 ON/ 93 OL 94 OM/ 95 OM 96 OL 9721N 98, e e e e e e e.
e 008e o e e e 1- 173e
- .696 e o e e e e: e 'e- .006.-
e e e- -* e *
- ON 109e e
e 99 ON/100 e
e e
e OL/10126M e 1.268*
e-
- 102
.955e 003 OM/10325L/104
- 1.254e OM/10524M
- 1.360 e 106 OL/107 e e e e -1,019.e 531 e
'e
- .004* .008.- - *-
.303 . .294 -.
e .Stae e 28Je110.ON e .305 111* ON/112
- OL 11327M/114* '1.160s ON/115 *'
OM/116
- OL 117 ON/118 ON 119 OJe120e e
e .DC4e . e e 012e e e * * *
- e-1.163* e e e 1.035 e e . .
e e o e
e OJe 121 ON 122.ON 123 ON/124 OM/125 e
e e e
o e
e o
e e
e
- e. ON/126 ON 127 ON 128.OJe12 e . .
e e a
. e .- e e .. .
e . * . 'e e e e .. e e BASE CASE VALUES COMP CASE VALUES- +
AVG = 1 MAXIMUM,0000051D=- .28942 AVG = 1.0000051D= .28810 VALUE 1.35951AT 23 MAXIMUM VALUE 1.35192AT 6=
MINIMUM VALUE 15 .33785AT 11-MINIMUM VALUE 15. 434388AT 11 NUMBER OF SAMPLES 27 DEGREES OF. FREEDOM 26 ABSOLUTE DIFFERENCE PCT DIFFERENCES-AVG = . 00'XX)510= .01463 MAXIMUM VALUE .02679AT 8 AVG = - .03166ST0= 1.64416 MINIMUM VALUE 15 .03308AT 5 MAXIMUM VALUE 2.88097AT 8 MINIMUM VALUE IS -4. 23 722AT -
.1 0 PPD-NA-8302-NPe Rev..03' -
Page 42 Of 57
- a.- = - - . . : .- - - _ _ _ _ _ - _ = _ _ _ _ - _ _ = _ _ _ .
1 1
l Figure 5-16
{
I l
CMAHA CY12 10941 WDINST ONLY ROCS-CECOR CCMPARISON AXI ALLY INTEGRATED INST TYPE BOX BASE CASE ABS OlFF PCT DIFF OM 1 OK 2 OK 3 OM 4*
e e e e e
. . . e e e e e e e OJe 5 ON 6 ON 7e ON/ 8 eOM/ 9 ON/ 10 ON 11 ON 12 OJe 13e e e e *
- e e e e e e e e e e e e e e e e e e e e e e e OJe 14e ON 15e 04/ 16e OL 17 e2M/,000 18,ON/ 19e OM/ 20e OL 21e ON/ 22e ON 23e IJo 24*
e .358*
e e
e e e . .000 . e o e e .043*
e e e .000e e e e e e11.995e e ON 25e6N/ 1.260* 26 OL/ 27*SM 28 OM/ 29* 4L/ 30 OM/ 31 3M 32 OL/ 33 eON/ 34 ON 35 1.261* .971e
- 1.241e e e e
e .081e * .054e e .009e e .035e e e e l e e 6.316e e-4.250s e .904e e-2.800* e e o I
e ON 36e OL 37 eOM 38.CM/ 39 OL 40 eON/ 41 .OL 42 OM/ 43 eOM 44eOL 45 eON 46ee 10M 47e e e e e e e * *
- e OM 48 e . E50, e e e e e e e e e e e e e
.025 ON/ 49e W/
e-4.547* 1.205* 50 OM/ 51* OL 52eOM 53eSL 54 7M 55 OL 56 OM/ 57 OM/ 58 ON/ 59ee o
.904e 1.137e e e e
- OK 60s e .035' e e e .055e .014e e e e 11K 61 e e e-2.886e e s e 6.121e-1.262e e
- e . .344e e
66 003*
e 17M/00062,ON/ 6316L/ e 64 ON/ 6515L.924,OL
,961s e 6714L 68
.923eON/ 6913L/ 70
.967e ON/ 7112M/ 72e OK 73e .000e .002 e e e 1.061e .883 a e na6e e .037e e .00Se e .038 OK 74 e
e
. .000, e .169e
- 3.843*
- 3.998* e .4888 e-3.566e o e ON/ 75 e OM/ 76 e OM/ 7720L 78 OM 7919L 80 OM 8118L 82 OM/ 83 OM/ 84 *ON/ 85e e e
a 1.133e e .915e e 1.126*
- e
e 22N 88 OL 89 eOM 90 OM/ 91 OL 92 eON/ 93 eOL 94 OM/
e 1.200e 95eOM 96eOL 9721N 98a e e .026 e e e e
e 1.205 -e e-2.188, e e e o
e e o e e * .030s '
e e e e e-2.493e
- ON 99 ON/100 OL/1012CM 102 OM/10325L/104 e = 1.241s OM/10524M 106 OL/10723N/108 ON 109*
e e e . -
e 982*
- 1.228* e 1.290e e
. ,019e . 020e e 072* e o e o e-2,034e 749e e-1.953e e-1.646e e 5.543, e 28Je110 e .346* ON 111* ON/112e OL 11327M/114 ON/115 OM/116 OL 117 ON/118 ON 119 OJe120e e 1.197e e * * *
- e
. .054e e *
- 028e e e e e e e e15.750s e a e-2.315e o e e e e s e
e OJe121 e e
ON 122 e ON 123.ON/124 . OM/125.ON/126 e ON 127 e ONe 128 OJe129' j e e e e a e e e e . . e e e e e e e-e OM 130 OK 131 OK 132 OM 133e e e e e e e e e e 1 e e e s -
BASE CASE VALUES COMP CASE VALUES AVGa 1.00000STD- .28701 AVG = 1.00000STD= .27687 MAXIMUM VALUE 1.28986AT 23 MAXIMUM VALUE 1. 36137AT 6 MINIMUM VALUE IS .34599AT 28 MINIMUM VALUE IS .39061AT 11 .
l NUMBER OF SAMPLES 26 DEGREES OF FREEDOM 25 ABSOLUTE DIFFERENCE PCT DIFFERENCES AVG = .00000STD. .03860 MAXIMUM VALUt .08087AT 6 AVG = .743685TD= 5.10131
! MINIMUM VALUE 15 .05360AT 5 MAXIMUM VALUE 15.74960AT 28 MINIMUM VALUE IS -4.54708AT 10 l
l OPPD-NA-8302-NP, Rev. 03 Page 43 Of 57
Figure 5-17 'I OMAHA CY13 719 WWD/TINST ONLY ROCS-CECOR CCWARISON AX1 ALLY' INTEGRATED -
INST TYPE BOX BASE CASE I ABS OlFF l PCT DIFF I OM/ 1 OL 2 OL 3 OM/ dee ;
e e e e
. . e e e e e e . .
.OM/ 5 OP 12 OM/ 13e 6eON - 7 eOP/ 8.ON/. 9.OP/ 10 ON -
- e. -11.OP e e e e e e e e e e -e e e e e e e e e e e -e OM/ 14e OP/ 15o ON/ 16e OP/ 17e.2N/ 18 OP/ 19 eON/ 20 eOP/ 21 e-ON/ 22e OP/ 231M/ 24e i e
e , e e e
.000e 000, e- e e e e .360e 'l e .011e e e e e . .000e e- e - e e . e-2.990.--
y OP 25 6N/ 26 ON 27 SM 28 ON 29 4M/ 30 ON . 31 3M .32 ON -33 ON/ 34 OP 35e ,
e e 1-151e e 1.187e- e 1.084e e 1.171 e - e e e e e .025 * .026e e 011e e .010e e e e i l
e e 2.193 .-2.1838 e 1.013e * .853e *- 'e 'e ON 36 eOP/ 37 eOM 38 ON/ 39 OM 40 OP/ 41 OM- 42 ON/ 43 OM - 44 OP/ 45 ON . 46*
e e e e e e. e- e e -e -
- i 10M/ 47 e e e- e _e e e' e e e OM/ 48 - i
.292*
- e e . . . e e e .- e e. e4
- .023 OP/ 49e 9N/ 50 ON 51*OM 52*OP/ 53e8M 54 7P/ 55 OM 56 eON 57 ON/ 58 *OP/ 59e*- .e 1 e-7.780s 1,050s 1.317e
- e- i OL 60s
- 1..021 252.e *
- e .012e .037 * * *- 11L 81 e e e 1.700s e e e-1.119e 2.816e *- e e e- .mo e 68 17N/941* 62 OP/ 6316M/ 64.OP/ 6515M 66 OM/ 6714M .022e --
e e 1.075 e 1.062e
- 1.060 OP/ ' 6913M/ 1.076 70.OP/ 7112N/ 72*
. .941e-6.654e OL 73e .010e e .020e e .024. *
,022,e
- 010e -* .010 OL -74 ' !
e e
e-1.047e
- 1.875 e-2.284e e-2.050 e 1.785 e-1.099e e 78 e e e OP/ 75 ON/ 76.ON 7720M1,273.OP/ 7919M -- 80 OP/ 8118M 82 ON 83 *ON/ 84
- OP/ 86e e 1.052e e 1.254* e-- e. ,
OM/ 86e e e e .013e .
- 015e .- 001* *~ ,e- ~ 0M/ 87 e e . e e-1.041e. *-1,387e -
- - .Ute e e e -o-e 22N 88 e e e 960.OP/ 89 OM 90e ON/ 91e- OM 92 OP/ 93 OM .e 94 ON/ 95 e OM ie96 OP/e9721N .eas -98 e e
,005e
.477 e
e e
e e
e e ,
e e
e o
e e
e.
e e- .876 We -
OP 99 ON/100 ON 10126M 102 ON 10325M/104 ON 10524M 106 ON 10723N/108 OP 100, e o e e 1.163e e 1.081e e 1.138s e 149e e e e * * .002e e .014e e 023e e 1 028 e e e e' .198 e 1.317e e 1.998e e 2.442 .
4 28M/110.OP/111 e- .351
- ON/112
- OP/11327N/114 e.1.259e OP/115.ON/11G.OP/117.ON/118.OP/119 -
e; OM/120. - "
e .003e e e e =. 015 e ~e e~ - * 's- e e- '
e .738* * *
- 1.192e e s -* -e e' e' e OM/121 e OP 122 e ON 123 e OP/124.ON/125 OP/126 .e ON 127.OP1128.OM/129
. e e o e . . . e e-e e e . ., .- e e e e
.OM/130.OL 131.OL 132 e OM/133 e e e . .- --
i e e e e e. ,
BASE CASE VALUES ' COMP CASE VALUES AVG =- 1.000005T0- .3005) AVGa 1.00000STD=- .30915-MAXIMUM VALUE. 1,31699AT 7 MAXIMUM VALUE 1.35406AT* 1- :t MINIMUM VALUE IS 29201AT 10 MINIMUM VALUF.15 .26929AT to' ~v NUuBER OF SAMPLES 27 DEGREES OF FREEDOM ABSOLUTE DIFFEREKCE ' PCT DIFFERENCES AVG = .000005TD- .01849' IAVGa .53090STD= I2.52424 .
~ MAXIMUM VALUE .03709AT 7 MAXIMUM VALUE - 2,81591AT 7 MINIMUM VALUE IS - .02591AT. '5 MINIMUM VALUE-IS -1.77998AT 10 <
OPPD-NA-8302-NP, Rev. 03-Page_44;0f 57
.. - . . = = ..a. .-. .=. - -
w.
-= . . .. .=- . .- . . _ . . - . - ~~
rigure 5-18 OMAHA CY13 7408 INST TYPE BOXMwD/ INST ONLY ROCS-Cf COR COMPARISON AXI ALLY INTEGRATED BASE CASE ABS DIFF PCT DIFF e
OM/ 1 OL 2 OL 3e OM/ de e e o e e e e e e e +
- OM/ 5 OP 6 ON e
e e
e e
e 7e OP/ 8 eON/ 9 OP/ e 10 DN e 11 OP e 12 OM/ e 13e e e e
. e e e e e e
- e e e o
e e e
OM/ 14 OP/ e e e e
e e 15e ON/ 16 OP/e 17 e
e e
2N/ 18 OP/
.000*
- 19 ON/ 20 OP/ 21 ON/ 22 OP/ 231M/ 24e e e e e .000* * . . .
=
e -
393e OP e . .000* e e e e-1 005e 26.ON 27* bM 28 o 25e6N/ 225e e
e 1.155 1.132.ON 29 e4M/ 1.060s30 ON 31 3M 32 e 1.120 ON 33 eON/ 34 OP 35e e
035* * .019e e .013
- e e e
e 2.992 e-1.701 .007 e e ON 36 eOP/ 37 OM
- 1.270* * .617e *
- e e
10M/ 47 e a
e
. 38 ON/ 39 OM 40 OP/e 41 OM 42 ON/ 43 eOM 44 OP/ 45 ON 46e e
o e e e e e e e e .340e e o e e e . . e e e
-o e e e e e *
- OM/ 48
,021 e-6.270e OP/ 49 94/ 50 ON 51 e 1.261
.022e o e e
JD00s 1.359e a e o e e e
e e e 1,783 , 000.e .037e *
- e e e e
.000 2.748e . .
e 11L 61 e e 402e OL 73 17N/ .00562 OP/ 6310M/ 64 OP/0 7 7e 6515M 66 OM/ 6714M e 1.061e 68 OP/ 6913M/ 027*70 OP/ 711 e
1.025e
- 1. 056 e
- 1.
- .018e e 1.077e e ,000,-6.765e
- * .511e e 1,672 a .018e . e .013e e e-1.695e e-1.628,018.e e .000 74 e e 09/ 75 eON/ 76 ON 7720M 78 OP/ 7919M 80 e
1.219e e .000.OL e
- 1.234e 82 e
OM/ 86 * * * .019e e
1.074.OP/
.014e 8118M1.221.04
- e 83 ON/ e 84 eOP/ 85e*
e e e
- -1.531* e .006e e e e e-1.340. * .468e e e OM/ 87 e e e 22N e .000s88 OP/
- 89 OM
- 90 DN/
- e e .000* *
- e e a e .
- 1.007e e e e e e
. ,003*
OP e e e e . 275e -
e e 99e DN/100 e e e
ONe 10126M1.121e 102 . 1.069eON 10325M/104 e 1.100* ON 10524M e .000e 106 eON 10723N/108 O e - 008* * .005, e e * * .728e e .013e e
.000e e e .454e e 1.186* e 000* *
- 28M/110 OP/111 ON/112 OP/11327N/114 OP/115 ON/116 OP/117 ON/118 OP/1
.386e e e l}D?e e
- e 1.271e e e e e e
- e
.439e e e e
.012 . e e e e
- OM/121 OP e
.9733 e e e e e e e
e e
122 ON 123 o e OP/124 o ON/125 e OP/126
- ON ,
127 OPe 128 OM/129s e e .
e o e e e e . . e e e e e . . e
.OM/130e OL 131e OL 132eOM/133'e e e e e e e e e e e BASE CASE VALUES COMP CASE VALUES AVG = 1.00000STD= 30357 MAX] MUM VALVE 1. 358 75A T 7 AVG = 1.00000STD= .31154 MIN] MUM VALUE IS MAX] MUM VALUE I,39609AT 7
.33954AT 10 MINIMUM VALUE 15 .31825AT 10 NUMBER OF SAMPLES 23 DEGREES OF FREEDOM 22 ABSOLUTE DIFf ERENCE AVG =
PCT OlffERENCES
.00000STD= .01794 MAXIMUM VALUE .03734A1 7 AVG = .435585TD= 2.37961 MIN] MUM VALUE 15 . 02719AT 11 MAXIMUM VALUE 2,99213AT 6 MINIMUM _ VALUE 15 -6. 76507AT 11 OPPD-NA-8302-NP, Rev. 03
.Page 45 Of 57
Figure 5-19 OMAHA CY13 15249 MWDINST ONLY ROCS-CECOR COMPARISON Ax! ALLY INTEGRATE 0 INST TYPE BOX BASE CASE ABS OlFF PCT OlFF OM/ 1 OL 2 OL 3 OM/ 4e e e e e e e e e e e o e e e e OM/ 5 OP 6 ON 7e OP/ 8 ON/ 9 OP/ 10 DN 11 OP 12 OM/ 13e e e e e e e e e e e e e e e e o
. . e e e e e e . e o
. e e OM/ 14 OP/ 15 ON/ 16 OP/ 17 e e e e 2N/
.00018.OP/ 19 ON/ 20 OP/e 21 ON/e 22 OP/e 231M/ 447e 24e e e e e e e e .000e e e e e e ,002e e e .000,
. e e e e e .559 OP 25e6N/ 26 ON 27 SM 28 ON 29 4M/ 30 ON 31 3M 32 ON 33 ON/ 34 OP 35 1.154
'o e
e 1.094*
- 1.040s e 1.086e e *
- e e .029e e .019e e .007* * .012*
- e e e e-1.778e
- 2.525' e .629e e-l.109* *
- e e
ON 36 eOP/ 37 eOM 38 eON/ 39 eOM 40 OP/ e 41 OM e
42 ON/
e 43 OM e
. 44 OP/
e 45 ON 46e e e e e s e o e e e e
- OM/ 48 10M/402 e 47.s e e e o e e e e e e e e e
.013 OP/ 49e9N/
e-3.226* 50 ON 51 eOM 52 eOP/ 53 84 .000 1.238e 54 7P/ 55 1.323,OM 56 ON 57 ON/ 58 OP/ 59ee
- e o e e 06 60* e .020e e o e .000e . 06 7 e o e e 11L 61 e e e 1.624e e e e .000* 5.082e e o e e 486 e
- *17N/
1.085*62 OP/ 6316M/ 64 OP/ 6515M 66 OM/ 6714M 68 OP/ 6913M/ 70 OP/ 7112N/
e 1.037e e 1.079e
- 1.081e e 1.045e e 72e .026 000e-5.290*
OL 73e 005e e .009e e .011e e .013e e .002* * .000 OL 74 e e 481* * .893e e .975 .-1.180s e .156* * .000e e e
e CP/ 75e ON/ 76 e ON 7720M 78 OP/ 7919M 80 OP/ 8118M 82 ON 83 *ON/ 84 OP/ 85 e
e
- 1.192e e 1.084*
- 1.179e
- e e OM/ 86e e * * .024*
- 016e * .012* *
- OM/ 87
- e e e e-2.037e e-1.490,
- e e .988s *
000* *
. 000, e e o e e e 1.040e e o e e e e .000s e e e . . . e e e . 008.e
.786 OP 99 ON/100 ON 10126M 102 ON 10325M/104 ON 10524M 106 ON 10723N/108 OP 109e e e o
e 1.094 e 1.053*
- 1.068e e .000* e e * .019e .006e a e .006e e .000
- e-1.768e e .607* e .606e e .000, e 28M/110 440e CP/111e ON/112* OP/11327N/114 OP/115eON/116 OP/117 ON/118 OP/119 OM/120e e 009e
- e e 1.252 e a e e e e e e e e * *
- 2.004e o e e 006.e 491 e e e e e e e
OM/121e OP 122e ON 123e OP/124 ON/125 OP/126 ON 127 OP 128 OM/129 e e e e e e e e e e . . e e e e e e e e . . . . . e e
OM/130e OL 131 OL 132 OM/1'33 e e e e e e e e e e e e e BASE CASE VALUES COMP CASE VALUES AVG = 1.00000ST0= .27178 AVG = 1.000005T0= .27818 MAXIMVM VALUE 1.32317AT 7 R4xlMUM VALUE 1.39041AT 7 MIN!WUM VAlliE-15~~~740176AT f0 MINIMUW VALUE IS ,38882AT 10 NUMBER OF SAMPLES 23 OEGREES OF FREEDOM 22 ABSOLUTE OlFFERENCE PCT OlFFERENCES AVC= .00000STD= .02044 max 1 MUM VALUE .06724AT 7 AVC= 20060$T0= 2.09842 MINIMUM VALUE IS .02569AT max! MUM VALUE 5.08165AT 7 11 MINIMUM VALUE IS -5. 28971 AT 11 OPPD-NA-8302-NPe Rev. 03' Page 46 Of 57 l
Figure 5-20 ROCS-CECOR Comparison At 1549 MWD /MTU Axially Integrated u be u. - Assembly Nurrter - Asserrdy Type - Detector Number D.DDDD - CECOR HPD E EEEE - Absolute Or!!erence F.FFFF - Percent Dinerence i N 2 N' 3 N 4 N s w e na i es o #a e Pt so m tt et it m i3 4 t4 w is ha is P It ni to N a to e mN r1 mi 22 e u ao 24 wi 0.9729 0.2057 0 0331 - 0.0146 3.4C2? -7.0977
~~
zs as es e e as M M Pt & 3 M* m F/ 4 St 16 R P/ 3 u f5 14 P r. At
., 1.1546 1.3004 1.2554 1.2463 0.0012 - 0.0125 0.0235 0.0416 0.1039 - 0.9612 1.8719 3 3379 J6 Pt 31 h4 se Pt 19 #/ 40 N 48 hs 42 N O Pt 44 Pt 46 h4 4e Pt 47 N to 44 N O3293 l
-0 0332 4. a2 s. N . si a sa N au w P, . .m, s. N ,, m aN =m
-10 082 1.0241 12507 1.5389 e,0 N/ -0.0270 0 % 49 - 0.0145 es w ti
-2.C365 1.9189 - 0.9422 0.3970 a2 Pt ti o ha 64 P/ u a na oc P/ t t, s1 Mt a n 14 e na io PI i3 it m ra P/ 12 - 0.0697 0.9478 12729 12882 12870 1.2701 Faaed -17.557 r3 w - 0.0771 0.0105 - 0.0126 -0 0114 0.0133 Detector 74 w
-8.1346 0.8249 -0.9781 -0 8858 1.0472 Level is a2 to N r; M is N ao to n4 al*Ti"Ti n n4 ea N is u5 H N MM 1.1137 42532 1.1020 se N - 0.0280 0.0215 -0.0163 e7 N
! - 2.51/1 1 7156 -1.4791
( se Pi 2z as R4 no et ei Pt e2 N/ e as mw as Pt w Pt er ne os P/ 21 j 0.7339 0.7458
.1--i.0032 - 0.0151
-0.4300 - 2 0247 WW ftt 400 P tiel R$ 102 Pi 26 103 N6 804 P/ 25 lub M tu6 P/ 24 607 15 las P 23 109 R1 12598 12019 1.2619 Failed O K81 0."J170 0.0260 Detector ,
22305 1.3472 2.0604 level 110 Ni 28 til H2 112 P 113 R7 114 N 1, 116 A3 lie N 867 R7 116 P (19 m 120 N/
02043 1.0072 l
- 0.0132 -0.0012
- 6.461 4 -0 1191 let w 172 ha 123 P/ t% R2 '26 Pt 126 F.2 427 P/ tJe R2 - 120 w 130 N i34 w '32 w 134 N
't. ROCS Case Values: Absolute Differences:
Average = 0.9974 Average = -0.0061 Standard Deviabon = 0.4322 Standard Deviation = 0.0317 -
Maximum Value = 1.5244 at Detectc / Maximum Value = 0.0416 at Detector 3 Msimum Value = 0.1911 at Deteck,r 28 Minimum Value = -0.0771 at Detector 1; CECOR Ca:e Values:- Percentage Differences:
Average = 0.9974 Avvage = -1.8477 Standard Deviabon = 0.4281 - Standard Deviabon = 4.9234 Maximum Value = 1.5389 at Detector 7 Maximum Value = 3.4022 at Detector 2 Minimum Value = 02043 at Detector i.4 Minimum Value = -17.5567 at Detector 11 OPPD-NA-8302-NP, Rev. 03 Page 47 of 57-
Figure 5-21 ROCS-CECOR Comparison At 3152 MWD /MTU Axially Integrated u sa cc - Assembly Number - Assembly Type - Detector Number D DDDD - CECOR RPD E.EEEE - Absolute Difference F.FFFF - Percont Differerce i N 2 w s N e N 6 N 4 R2 1 Pt 4 R2 9 Pt 60 M7 il PI 12 RI 13 4 la N 16 H2 16 P t7 R7 le N 2 19 M 20 N F1 R1 22 P 23 iS 24 4 1 0.3889 02002
-0 0039 - 0.0118
- 0.3944 -5.8941 24 Hl 16 P 4 21 h6 28 Pt 6 29 M 10 Pt 4 31 m n Pt 3 At 4 J4 P J6 Al 1.1238 1 2838 12685 1.2308 0 0094 - 0.0175 0.0097 0.0295 o 0 8364 -1.3631 0.7647 2.3852 n PI 31 k4 M VI 59 Pt 40 N et h6 42 N 43 Pl 44 Pt 46 he 46 Pl 47 N to 44 N 0.3172
- 0.0158 4e n2 wN e si n> s2 N s3 94 se et e $6 he i ss N 61 4 66 N $9 R2
- 4.9811 1.0110 1.2451 1.5802 60 Nr -0.0042 0.0171 -0.0272 et N si
- 0.4154 1.3734 -1.7213 0.3813 62 et 17 63 R3 64 Pt s6 e as sa Pt u, aw e er $. e ns 70 Pi i3 ri so 72 Pt 12 -0.0233 0.9125 1 3.834 12836 12867 1.2848 Failed -6.1107 r3 N - 0.0150 0.0219 - 0.0157 - 0.0188 0.0206 Detector i4 N
-1.6438 1.7064 -12231 -1.4611 1.5956 Level 16 R2 16 N r7 Rs in N 20 is as ~ so P/ i9 at ne a2 N to s3 m se N as At 1.1152 12498 1.1086 se N - 0.0243 0.0124 - 0.0177 uN
- 2.1790 0.9922 - 1.5966
, se P/ 2e 49 he sa Pt vt Pi 92 N/ $3 he M Ni a Pt 96 Pt 97 As n P/ 2t l 0.7172 0.7336 l 0.0139 -0.0025 l 1.9381 -0 3408 l W9 hl 100 P 101 h5 102 Pt 26 10J h6 104 Pt 2t 106 M i;6 P/ 24 10/ m tas P 23 10D R$
12461 1.2754 12475 Failed 0.0202 0.0028 0.0188 Detector 1.0211 0.2195 1.5070 Level 610 Ni as ist rG 112 P 113 R1 114 N 21 116 R3 116 N tli R? 114 P 119 R2 IJD N/ '
O.1987 1.0012
-0 0103 - 0.0162
-5.1837 -1.6181 12% Ni 122 h2 123 Pt 124 R2 126 P/ 126 R2 121 Pt 12e Rz 129 N 130 N 1J1 Nt 132 N 133 N ROCS Case Values; Absobte DMerences Avera0e = 0,9972 Average = -0.0018 Standard Deviaton = 0.4399 Standard Devianon = 0.0174 Maximum Value = 1.5530 at Detector 7 Maximum Value = 0.0295 at Detector 3 Minimum Value = 0.1884 at Detector 28 Mriimum Va!ue = -0.0272 at Detector 7 CECOR Case Values: Percmtago Differences:
Average = 0.9972 Avenkge = -0.8200 Standard Deviation = 0.4372 Standard Devianon = 2.4896 Maximum Value = 1.5802 at Detector 7 Maximum Value = 2.3852 at Detector 3 Mstimum Value = 0.1987 at Detector 28 Mriimum Value = -6.1107 at Detector 11 I- OPPD-NA-8302-NP, Rev. 03 Page 48-Df 57
Figure 5-22 .
ROCS-CECOR AX A_ S- A3E 'COM 3AR SO\
~
CYCLE 11 CORE AVERAGE = 13,809 MWD /T ROCS AXIAL PEAK / AVERAGE = 1.3225 AT 87% FROM BOTTOM, ASI = .0688 CECOR AXIAL PEAK / AVERAGE = 1.1370 AT 70% FROM BOTTOM, ASI = .0363 NORMAllZED POWER CECOR 1.3 ROCS ' '
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Figure 5-23 ,
CYCLE 12 ROCS-CECOR AX A_ S-A?E COV 3AR SO\'
CORE AVERAGE = 664 MWD /T ROCS AXIAL PEAK / AVERAGE = 1.1829 AT 56% FROM BOTTOM, ASI = .0803 CECOR AXIAL PEAK / AVERAGE = 1.1620 AT 62% FROM BOTTOM, ASI = .0060 NORMAllZED POWER 14 CECOR 1.3
--- ROCS
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Figure 5-24 -
l ROCS-CECOR AX A_ S-A3E COV3AR SON CYCLE 12 CORE AVERAGE, 5946 MWD /T ROCS AXIAL PEAK / AVERAGE = 1.0818 AT 25% FROM BOTTOM, ASI = .0980
.0223 CECOR AXIAL PEAK / AVERAGE = 1.0985 AT 28% FROM BOTTOM ASI =
NORMALIZED POWER 1.4 CECOR 1.3 ROCS o 1.2 5
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Figure 5-25 g ROCS-CECOR AX A_ S- A3E COV 3AR SON CYCLE 12 CORE AVERAGE 10941 MWD /T ROCS AXIAL PEAK / AVERAGE. = 1.0920 AT 16% FROM BOTTOM. ASI = .1036 CECOR AXlAL PEAK / AVERAGE = 1.1245 AT 64% FROM BOTTOM, ASI = - 0172 .
1
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Figure 5-26 ,
r -
ROCS-CECOR AX A_ S- AJE COV 3AR SO\
CYCLE 13
' CORE AVERAGE, 719 MWD /T
' ROCS AXIAL PEAK / AVERAGE = 1.1583 AT 67% FROM BOTTOM, ASI = .0461 -
CECOR AXIAL PEAK / AVERAGE = 1.1428 AT 62% FROM BOTTOM, ASI = .0099 NORMAllZED POWER 1.4 CECOR 1.3 ROCS g 1.2 j*} "
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Figure 5-27 g ,
ROCS-C ECO R AX A_ S- A3 E CO V 3AR SO\
CYCLE 13 CORE AVERAGE 7408 MWD /T ROCS AXIAL PEAK / AVERAGE = 1.1043 AT 80% FROM BOTTOM. ASI = .0610 CECOR AXIAL PEAK / AVERAGE = 1.0947 AT 72% FROM BOTTOM, ASI = .0167 NORMAllZED POWER 1.4 CECOR 1.3
.----- ROCS '
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Figure 5-28 ,
4 ROCS-CECOR AX A_ S- A3E CONA3AR SON CYCLE 13 CORE AVERAGE.15.249 MWD /T ROCS AXIAL PEAK / AVERAGE = 1.1942 AT 86% FROM BOTTOM, ASI = .0067 CECOR AXIAL PEAK / AVERAGE = 1.1255 AT 78% FROM BOTTOM, ASI = .0408 3, NORMALIZED POWER ,
CECOR 13 ROCS
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Figure 5-29.
ROCS-CECO R AXIAL SHA? E CO V PAalSON CYCLE 14 CORE AVERAGE,1549 MWD /MTU ROCS AXIAL PEAK / AVERAGE = 1.1692 AT 50% FROM BOTTOM, ASI = 0.00686 CECOR AXIAL PEAK / AVERAGE = 1.2012 AT 36% FROM BOTTOM, ASI = 0.04669 1.4 NORMAllZED POWER CECOR ROCS 021 _-
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Figure 5-30 g
ROCS-CECOR AXIAL SHAP E CO V PA llSON CYCLE 14 COPE AVERAGE,3152 MWD /MTU ROCS AXtAL PEAK / AVER. AGE = 1.1619 AT 50% FROM BOTTOM, ASI = 0.00271 CECOR AXtAL PEAK / AVERAGE = 1.1854 AT 34% FROM BOTTOM. ASI = 0.04435 g,4 NORMAllZED POWER CECOR ROCS 1.2 k
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