ML20082P243
| ML20082P243 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 07/31/1991 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20082P236 | List: |
| References | |
| NUDOCS 9109100248 | |
| Download: ML20082P243 (117) | |
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NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN VlltGIL C. SUMMElt NUCLEAR STATION
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VIRGIL C. SUMMER NUCLEAll STATION NUCLE All PilYSICS METIIODOLOGY FOlt itELOAD DESIGN-JULY 1991 SOUTH CAltOLINA ELECTitlC & GAS COM PANY N UCLEAll OPEll ATI'ONS DIVISION J ENKINSVILLE, SOUTil C AltOLINA
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AHSTilACT This document summarizes the nuclear design methodology employed by SCE&G to perform reli ud core design analyses for VCSNS. This methodology, including all computer programs used, was obtained in its entirety from Westinghouse Electric Corporation. Calculations were performed using this methodology and the results compared to operating data from VCSNS. The quality of the comparisons demonstrates SCE&G's ability to perform reload core design for VCSNS.
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TAHl E OF('ONTENTS Section Pan 1.0 INTitODUCTION ANI) CONCI.USIONS 1
1.1 Objective 1
1.2 llackground 1
1.3 Scope 2
1.4 Conclusions 2
2.0 l'IlYSICS M ETilOI)Ol.OGY 5
2.1 Cross Section Library 5
2.2 PilOENIX-P Lattice Modeling 6
2.2.1 Fuel Cell Model 6
2.2.2 Discrete Absorber Models 7
2.2.3 Structural Cell Models 8
2.3 llame-Reflector Modeling 8
2.4 Three. Dimensional Nodal Model 8
2.5 One-Dimensional Diffusic
'eory Model 9
3.0 PilYSICS MODEl. Al'PLICATIONS 11 3.1 Core Power Distributions at Steady State Conditions 11 3.1.1 Power Distributions 11 3.1.2 Power Peaking 12 3.1.3 Fuel Depletion 12 3.2 Axial Power Distribution Control Limits 12 3.3 Core lleactivity Parameters 13 3.3.1 Moderator Temperature Coefficient 14 3.3.2 Doppler Temperature Coefficient 15 3.3.3 Total Power Coefficient 15 3.3.4 Isothcanal Temperature CoefGeient 16 3.3.5 Baron Reactivity Coefficient 16 3.3.6 Xenon and Samarium Worths 17 3.3.7 Control Rod Worths 17 3.3.8 Neutron Kinetics Parameters 18 1
TAIliR 01' CONTENTS acontinued)
Section l' age 3.4 Core Physics Parameters for Transient Analysis input 18 3.4.1 Shutdown Margin 19 3.4,2 Trip llenctivity 19 3.4.3 ControlItod Misalignment 20 3.4.4 Boron Dilution 20 3.4.5 Controlllod Bank Withdrawal 21 r
3.4.6 ControlIlod Ejection 21 3.4.7 Steamline lireak 21 4.0 PilYSICS MODEl, VElt!FICATION 23 4.1 Cycle Descriptions 23 4.2 Zero Power Physics Tests 24 4.2.1 Critical Baron Concentrations 24 I
4.2.2 Isotherrnal Ternpc.mtnre Coemeients 25 4.2.3 ControlItod Worths 25 4.2.4 Differential Boron Worths 26 4.3 Power Operation 26 4.3.1 Boron Letdown Curves 27 4.3.2 Power Peaking Factors 27 4.3.3 Itadial Power Distributions 28 4.3.4 Axial Power Distributions and Axial Offsets 28 4.4 Summary 29 5.0 ItF.FEltENCES 99 6.0 APPENDIX 101 6.1 FIGIITil 102 6.2 PHOENIX.P 102 6.3 ANC 103 6.4 APOLLO 104 ii
l.lST O F TAlilES Tv.ble Pace 4.1-1 V.C. Summer Nuclear Station Fuel Specification 31 4.2-1 V.C. Sununer Nuclear Station llZP Physics Test Iteview Criteria 32 4.2-2 V.C. Summer Nuclear Station Critical Baron Concentration Comparison Between Measurement and Prediction for Cycles 3,4. and 5 33 4.2-3 V.C. Sununer N uclear Station Isothermal Temperature Coefficient Comparison Between Measurement and Prediction for Cycles 3,4, and 5 34 4.2-4 Y.C. Summer Nuclear Station Control Rod Worth Comparison Between Measurement and Prediction for Cycles 3,4, and 5 35 4.2 5 V.C. Summer Nuclear Station IIZP Differential Baron Worth Comparison Between Measurement and Prediction for Cycles 3,4, and 5 36 4.3 1 V.C. Summer Nuclear Station Cycle 3 Boron Letdown Comparison Between Measurement and Prediction 37 4.3 2 V.C. Summer Nuclear Station Cycle 4 Baron Letdown Comparison Between Measurement and Prediction 38 4.3 3 V.C. Summer Nuclear Station Cycle 5 Boron Letdown Comparison Between Measurement and Prediction 39 4.3-4 V.C. Summer Nuclear Station Cycle 3 Power Peaking Factor Comparison Between Measurement and Prediction 40 iii
i I,lST OF TA files (continued)
Table l'a ge 4.3 5 V.C. Sununer Nuclear Station Cycle 4 l'ower l'eaking Factor Comparison lietween Measurement and l'rediction 41 4.3 6 V.C. Sununer Nuclear Station Cycle 5 l'ower I)eaking Factor Comparison 11etween Measurement and l>rediction 42 4.3 7 Y.C. Summer Nuclear Station Cycle 3 Axial Offset Comparison lietween Measurement and l'rediction 43 4.3-8 V.C. Summer Nuclear Station Cycle 4 Axial Offset Comparison lietween Measurement and l'rediction 44 4.3 9 V.C. Sununer Nuclear Station Cycle 5 Axial Offset Comparison Iletween Measurement and l'rediction 45 a
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4.1-1 V.C. Sununer Nuclear Station Cycle 3 Core Loading 47 4.1 2 V.C. Sununer Nuclear Station Cycle 4 Core leading 48 4.1 3 V.C. Sununer Nuclear Station Cycle 5 Core Loading 49 4.2 1 V.C. Sununer Nuclear Station Cycle 3 Measured versus l'redicted IIANK 11 Integral flod Worth 50 4.2 2 V.C. Sununer Nuclear Station Cycle 4 Measured versus Predicted ll AN K Il Integral llod Worth 51 4.2-3 V.C. Sununer Nuclear Station Cycle 5 Measured versus Predicted il AN K 11 Integral llod Worth 52 4.3 1 V.C. Summer Nuclear Station Cycle 3 lloron Letdown Comparison lletween Measurement and Predictwn a3 4.3-2 V,C. Summer Nuclear Station Cycle 4 lloron Letdown Comparison lletween Measurement and Prediction 54 4.3 3 V.C. Sununer Nuclear Station Cycle 5 lloron Letdown Comparison lletween Measurement and Prediction 55 4.3-4 V.C. Sununer Nuclear Station Cycle 3 F Delta-ll Comparison lietween INCORE and ANC 56 4.3-5 V.C. Summer Nuclear Station Cycle 4 F Delta Il Comparison 13etween INCORE and ANC 57 4.3 6 V.C. Sununer Nuclear Station Cycle 5 F-Delta.ll Comparison Iletween INCOllE and ANC 58 v
1,lST OF FIGUllES
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Figure l'Aig 4.3-7 V.C. Summer Nuclear Station Cycle 3 Fq Comparison lietween INCollE and ANC 59 4.3 8 V.C. Summer Nuclear Station Cycle 4 Fq Comparioun lletween INCOltE and ANC 60 4.3 9 Y.C, Summer Nuclear Station Cycle 5 Fq Comparison lietween INCOllE and ANC 61 4.310 V.C. Summer Nuclear Station Cycle 3 Itadial Power Distribution Comparison lietween INCOllE and ANC for Alap FCFht 03 007 62 4.311 V.C. Sununer Nuclear Station Cycle 3 Radial Power Distribution Comparison Between INCOllE and ANC 3
for Slap FCF5103-014 61
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4.312 V.C. Summer Nuclear Station Cycle 3 Itadial Power Distribution Comparison lletween INCollE and ANC for hian FCFM-03 017 64 4.313 V.C. Sununer Nuclear Station Cycle 3 lladial Power Distribution Comparison lietween INCOllE and ANC for hlap FCFht 03 025 65 4.3-14 V.C. Summer Nuclear Station Cycle 3 Itudial Power Distribution Comparison lietween INCOllE and ANC for Map FCFM 03 031 66 4.3 15 V.C. Summer Nuclear Station Cycle 3 Itadial Power Distribution Comparison lletween INCOllE and ANC for h!ap FCFht 03 037 67 vi l
s 1.lST OF FIG UlmS teontinued)
Figure hge 4.316 V.C. Sununer Nuclear Station Cycle 4 Itadial Power Distribution Comparison 11etween INCOllE and ANC for Map FCFM 04 005 68 4.3-17 V.C. Summer Nuclear Station Cycle 4 Itadial Power Distribution Comparison Between INCOllE and ANC for Map FCFM 04-010 69 4.3-18 V.C. Summer Nuclear Station Cycle 4 Itadial l'ower Distribution Comparison lletween INCollE and ANC for Map FCFM-04-012 70 4.319 V.C. Sununer Nuclear Station Cycle 4 Itadiall'ower L)istribution Comparison lietween INCOllE and ANC for Map FCFM-04 016 71 4.3 20 V.C. Summer Nuclear Station Cycle 4 Itadial Power Distribution Comparison Between INCOllE and ANC for Map FCFM-04-021 72 4.3-21 V.C. Summer Nuclear Station Cycle 4 fladial Power Distribution Comparison lletween INCOllE and ANC for Map FCFM-04-026 73 4.3-22 V.C. Summer Nuclear Station Cycle 5 Itadial Power Distribution Comparison 13etween INCOItE and ANC for Map FCFM 05 006 74 4.3 23 Y.C. Summer Nuclear Station Cycle 5 Itadial Power Distribution Comparison Between INCOllE and ANC for Map FCFM-05-012 75 vii
9 LIST OF FIGUllES (continued)
Figure Page 4.3-24 V.C.Sununer Nuclear Station Cycle 5 Radial Power Distribution Comparis.on lietween INCOllE and ANC for Map FCFM 05 015 76 4.3 25 V.C. Sumnaer Nuclear Station Cycle 5 Itadial Power Distribution Comparison Between INCOllE und ANC for Map FCFM-05-018 77 4.3 26 V.C. Summer Nuclear Statim Cydu 5 ttwial Power Distribution Comparison Betwen INCOltE and ANC for Map FCFM-05-020 78 4.3 27 V.C. Summer Nuclear Station Cycle 5 8adial Power Distribution Comparison Between INC0!tE and ANC for Map FCFM 05-022 79 4.3-28 V.C. Summer Nuclear Station Cycle 3 Axial Power Distribution Comparison Between INCOllE and ANC for Map FCFM-03 007 80 4.3 29 V.C. Summer Nuclear boation Cycle 3 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM-03-014 81 4.3 30 V.C. Sununer Nuclear Station Cycle 3 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM 03 017 82 4.3-31 V.C. Summer Nuclear Station Cycle 3 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM 03-025 83 l
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1,lST OF FIG UllF.S (continued)
Ficure Pane 4.3 32 V.C. Summer Nuclear Station Cycle 3 Axiall'ower Distribution Comparison lietween INCollE and ANC for Map FCFM-03 031 84 1
4.3 33 V.C. Summer Nuclear Station Cycle 3 Axial Power Dit*.ribution Comparison Detween INCOllE and ANC
- i for Ma p FCFM-03-037 85 4.3 34 V.C. Summer Nuclear Station Cycle 4 Axial Power DistributionComparison Between INCollE and ANC i
l for Map FCFM.04 005 8 41
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4.3-35 V.C.Sununer Nuclear Station Cycle 4 Axial Power Distribution Comparison lietween INCollE and ANC for Map FCFM-04 010 87 4.3 36 V.C. Summer Nuclear Station Cycle 4 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM 04-013 88 4.3 37 V.C. Summer Nuclear Station Cycle 4 Axial Power Distribution Comparison Iletween INCollE and ANC for Map FCFM 04-016 89 4.3 38 V.C. Summer Nuclear Station Cycle 4 Axial Power -
Distribution Comparison Between INCORE and ANC for Map FCFM 04 021 90 4.3 39 V.C. Summer Nuclear Station Cycle 4 Axial Power Distribution Comparison Between INColtE and ANC for Map FCFM-04 026 91 IX
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.IJST OF FIGUllRS teontinued)
Finure
,P;ge 4.3-40 V.C. Sununer Nuclear Station Cycle 5 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM-05-006 02 1
4.3 41 V.C. Sutarner Nuclear Station Cycle 5 Axial Power Distribution Comparison Between INCOllE and ANC for Map FCFM-05 012 93 4.3 42 V.C. Summer Nuclear Station Cycle 5 Axial Power
)
Distribution Comparison Between INCORE and ANC i
for Map FCFM 05 015 94 1
4.3 43 V.C. Sununer Nuclear Station Cycle 5 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM 05-018 95 4.3 44 V.C. Summer Nuclear Station Cycle 5 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM.05 020 96 4.3-45 V.C. Summer Nuclear Station Cycle 5 Axial Power Distribution Comparison Between INCORE and ANC for Map FCFM-05 022 97 l
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l.0 INTitOI)UCTION ANI)CONCl.USIONS This report briefly describes the PWit physics methods used by South Carolina Electrie and Gus Company (SCE&G). It includes a sununary descriptior, of the Westinghouse computer programs and methodology as applied by SCE&G to model the V. C Surnmer Nuclear Statino (VCSNS) core. Comparisons between predictions and operating data are provided us a demonstration of SCE&G's qualiGeations to use the Westinghouse meti.odology to perform reload d,; sign calculations for VCSNS 1,1 Oll.J ECTIVE The objective of this report is to demonstrate SCE&G's capability to perform reload design analyses for VCSNS. To this end, design calculations have been performed for Cycles 3,4, and 5 and the results are compared to actual plant operating data herein.
1.2 I! ACKGitOUND SCE&G has realized that in-house capability to design reload cores for VCSNS would provide the following beneGts:
A better understanding of the design, yielding more control of the decision
- process, An optimization of the design, allowing a greater involvement in planning, and Better quality control of the design, leading to more comprehensive evaluations of core safety.
Various physics methodologies were reviewed to determine which best satisGed SCE&G's needs. SCE&G decided to use the approach of Westinghouse, our NSSS vendor and present fuel supplier The Westinghouse methodology offers three distinct advantages:
A systematic physics methodology developed for and previously applied to a large number of PWil designs, including VSCNS and similar plants, 1
A physics methodology previously reviewed and approved by the NHC, and A physics methodology that is already cornpatible with the other analytical design processes (e.g., thermal hydraulics and safety analyses) being used for VCSNS.
Implementation of the above decision involved entering into a technology exchange agreement with Westinghouse Electric Corp. The relevant methodology and associated computer programs of the Westinghouse Commercial Nuclear Fuel Division have been transferred to SCE&G. A description of the physics models is provided in the next chapter while the computer programs themselves are discussed in the Appendix.
1.3 SCOl'R 1
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SCE&G has performed in house design calculations and core follow analysis for i
VCSNS since its initial startup. Core follow results acquired during Cycles 3,4, and 5 provide reliable data to which predicted power distributions, predicted boron letdown curves, and fuel depletion calculations may be compared. In addition, the startup physics measurements made during the startup of each cycle provide reliable data for evaluating the physics model predictions of critical boron concentrations, control rod worths, and temperature coefficients. Detailed comparisons are presented in Section 4.
All methods employed (model development, computer programs, ete ) to generate the results contained in this report are standard licensed methods used by the Westinghouse Commercial Nuclear Fuel Division. Therefore, the calculational uncertainties (e.g., see Reference 1) associated with the methods are unchanged and their requantification is unnecessary. Similarly, the methods utilized to l
process measured data (e.g., see Reference 2), are also standard to Westinghouse l
such that measurement uncertainties do not require redetermination.
1.4 CONCI.USIONS l
This report summarizes the Westinghouse methodology used by SCE&G to model the VCSNS core. Calculations were performed for Cycles 3,4, and 5 and the 2
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results are compared to actual operating data. Cycles 1 ar.d 2 were also modeled to establish the appropriate burnup distributions. The results demonstrate that SCE&G understands the methodology and can apply it correctly during the performance of future reload design analyses for the V. C. Summer Nuclear Station.
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2.0 PIIVSICS METilODOLOGY This section brie 0y describes the Westinghouse methodology used by SCE&G to perform design calculations for reload cores. The major features associut.d with each model are discussed as well as the interaction between models. This methodology was also used to obtain the results presented in Section 4.
Descriptions of the individual computer codes used are provided in the Appendix.
Detailed discussions are contained in the Westinghouse documentation referenced here and in the Appendix.
Lattice physics parameters for unit assemblies and bafDe-reDector cross sections are calcuhited with the two-dimensional multi-group transport theoty code, PilOENIX P Fuel and clad temperrtures are generated with the FIGilTil code.
The core is modeled in three dimensions with the advanced ra dal ende, ANC, which is used to predict reactivity, power distributtns, and other relevant core characteristics. In addition, the one dimensional diffusion theory code, APOLLO, is used to calculate differential control rod worths and axial power distributions for the heat nux hot channel factor (Fy) synthesis to eatablish operational limits. The cross section library as well as PI(OENIX.P. nodal, and diffusion theory models are discussed in the following sections.
The models described here are representative of current Westinghouse practice.
SCE&G's calculational capabilities are expected to evolve la parallel with Westinghouse's through continued implementation of the technology exchange agreement between the two organizations.
2.1 CitOSS SF.CTION Lillit AltY The nucleta cross section library ased by the PIIOENIX P computer progrnm contains microscopic cross section data based on c. 42 energy group structure which has been derived from ENDFiUN Gles. The PilOENIX.P cross section library was designed to properly capture integral properties of the multigroup dato during the group collapse, enabiing accurate modeling of impc.rtant resonance parameters, and to provide the overall accuracy of reactivity predictions required for core design. It has been developed in a manner consistent with existing Westinghouse methodologies and accumulated 5
experience in core design, The generation and benchmarking of the PHOEN!X-P library are described in detail in Reference 3.
2.2 PliOENIX P 1.ATTICE PiODEl.ING In PIIOENIX-P, the fuel, discrete absorbers, and structural components within a single fuel assembly at : represented in their. exact lattice configuration.
Homogenized two-group microscopic cross sections, discontinuity factors, and pin factors are generated as a function of burnup for input to ANC. Microscopic cross sections are generated for isotopes and materials represented explicitly in ANC, These include xenon, samarium, soluble boron, water, and burnable absorbers.
Branch calculations are performed at selected burnups to obtain constants for rodded assemblies.
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PHOENIX P allows a three region cylindrical cell description for each cell within the lattice. Since most lattice cells consist of more than three subregions, material preservation principles are employed to construct a three region cell representation. The third or outer region of each cell, defined by the fuel pin pitch, has a common composition in all cells in a given lattice problem. The grids are modeled by smearing the grid material uniformly over this common outer region.- Only grids in the active fuel are smeared. The following sections describe the different types of cell models.
2.2.1 FUEliCEl.l. MODEl.
The innermost region of a fuel rod cell is defined by the fuel pellet outer radius.
The second region is defined by the clad outer diameter and includes the gap.
For fresh fuel, the c.ppropriate number densities are specified for the uranium L
isotopes and oxygen, For burned fuel, isotopic information for the depletion and i
decay el Hs modeled in PHOENIX-P is obtained from previous depletion calculatior.s. Fuel pellets with integral fuel burnable absorber (IFBA) are not explicitly modeled with a coating on the pellet; rather, the absorber material is smeared into the clad region. PHOENIX-P corrects for the effect upon reactivity of modeling the absorber in the clad instead of on the pellet.
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2,?. 2 lilSCiti? TIC AllSOltilitit MODICI.S
- TitNABLE ABSOltl3 Ell 1(ODS Two types of discrete burnable absorber (BA) rods have been used at VCSNS: Pyrex glass and wet annular burnable absorber (walla). The Pyrex BA is voided in the central region while the walla contains moderator material, hence the cell represent 9 tion for the two IIA types is significantly different. Also, for llA's, the surface area of the absorber material must be preserved in addition to the amount of material.
Fer a Pyrex BA, the void, inner clad, and BA pellet material are smeared imo the first region with a radius equal to the BA pellet outer radius.
Ile gion 2 comprises the llA outer clad, gap, guide tube and sleeve volumes,
- 3nd materials. Note that the small volume ef moderator between the llA outer clad and the guide tube is modeled as if it is outside the guide tube.
Since the zircaloy guide tube material is nearly transparent to neutrons, this is a minor approximation.
in the case of walla's, both the inner and outer surfaces of the absorber are important since a fast neutron can pass through the absorber, become thermalized in the inner region, and be absorbed. Therefore, region 1 of the cell is defined as moderator material with an outer radius equivalent to the BA pellet inner radius and region 2 as pure pellet material with an outer radius equivalent to the BA pellet outer radius. Tb inner clad, inner gap, outer gap, outer clad, guide tube, and sleeve materials are smeared into the moderator region to preserve materials.
B. CONTROLItODS Control rod cells are modeled the same as Pyrex B A cells; the only distinction is the dimension of and material in the pellet region.
PHOENIX P performs resonance calculations for the Ag-In-Cd control rod material.
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2.2.3 STit UCTU lt A l. C E l.1, M OI)El.S Structural cells are cells that contain neither a strong absorber nor material that is depletable Then include guide tubes, instrumentation tubes, water displacer rods, and stainless steel rods. Typically these can be represented with three or fewer regions and do not require any special neutronic considerations.
Sleeves are accounted for by calculating an effective guide tube thickness that preserves the sleeve volume.
2.3 IIAFFIE itEFIECTOlt MOI)RI.ING A one-dimensional slab calculation is performed with PHOENIX-P to generete baffle-reflector cross sections for ANC. The model consists of a series of fuel cells approximating two fuel assemblies, assembly /bafTle gap, baffle, reflector, core barrel, thermal pad (on the flats), and moderator. A set of homogenized cross L
sections for ANC is obtained which reflect the complex spectrum variation which exists between the fuel assemblies, baffh, and reflector.
2.4 TilItEE.IllM ENSIONAI, NOI) Al. MOI)RI, The homogenized cross sections, discontinuity factors, and pin factors generated on a cycle specific basis with PHOENIX-P' depletion calculations are used to model the three-dimensional core in ANC. Each fuel assembly is represented by four radial nodes. To obtain an accurate pin power recovery solution, the burnup gra "ent within each node is represented in ANC A burnup gradient algorithm
- node corner and surface average burnups.
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i.rolly heterogeneous features such as axial blankets and part length burnable absorbers are explicitly represented using the variable axial mesh capability in l
- ANC. Generally,20 to 24 axial mesh intervals produce accurate axial power i
distributions. Axial zoning of burnur dependent fuel cross sections is used to account for spectrum effects induced by axial burnable absorber and fuel burnup gradients. Previous cycle burnable absorber history effects are also accounted for by using different sets of fuel cross sections.
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Three-dimensional ANC calculations are used to predict core power distributions, peaking factors, critical baron concentrations, control rod worths, and reactivity coefficients. The three-dimensional model can also be collapsed to two-dimensions for certain calculations (e.g., selection of the highest worth stuck rod) where a three-dimensional representation is not necessary.
2.5 ONE.1)lM ENSION Al.1)lFFUSION TilEOltY MOI)RI.
The three-dimensional ANC model is radially homog;enized to generate a one-dimensional APOl.LO model. Cross sections are flux and volume weighted, and a burnup and elevation dependent radial buckling search is performed to normalize the APOLLO model to ANC. The axial mesh is redefined to comprise 40 or more axial intervals. The one-dimensional difTusion theory model is u.ed for calculations where additional detail is desirable in the axial direction. These include generation of differential and integral control rod worth curves, determination of control rod insertion limits, and analysis of axial power distributions to establish limits on axial offset during power operation.
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i 3.0 l>llVSICS MODEl, Al)l*l lC ATIONS The physics methodology discussed in Section 2 was developed in order to provide reliable analytical predictions in the following four major areas:
Core power distributions at steady state conditions.
Axial power distribution control. limits.
Core reactivity parameters, and Core physics parameters for transient analysis input.
Often more than one model may be used to perform a speciGe analysis. The preferred model depends upon a number of considerat ons including the degree of i
accuracy desired, the specific applications, and the required resources.
The references made in this section refer to the specific models described in Section 2. SCE&G will continue to upgrade the methodology used such that it remains current with the latest approved calculational techniques being employed by Westinghouse.
3.1 COltE l'OWElt I)lSTitillUTIONS AT STR ADY STATE CONI)lTIONS The prediction of steady state core power distributions is fundamental to the design, analysis, and surveillance of nuclear reactor cores. Accurate prediction of core power distributions leads to confidence in developing and optimizing core loading patterns, ensuring compliance with Technical Specification limits, and determining fuel assembly burnups and isotopic inventories.
3.1.1 l'OWElt DISTitillUTIONS Global core power distributions are obtained as a function of burnup from three-dimensional ANC depletion calculations. Calculations are also performed at selected burnups for various power levels and control rod conGgurations. Peak rod powers and hot channel factors are generated by pin power reconstruction within ANC using rod-by-rod power distributions from 11
single assembly two-dimensional PilOENIX,P fine mesh spectrum calculations.
3.1.2 l'OWElt l'E AKING Local power peaking is monitored to ensure that the peak pellet power and the total energy content within each coolant channel remain within Technical Specification and /or fuel design limits. The factors used to measure local power peakinginclude:
- the heat Oux hot channel factor, F,defin.
iximum local heat y
flux on the surface of a fuel rod divided by the,erage fuel rod heat Oux,
- the nuclear enthalpy rise hot channel factor, F u, defined as the ratio of 3
the integral of linear power along the rod with the highest integrated power to the average rod power, and
- the planer radial power peaking factor, Fu(Z), defined as the ratio of the peak pin power density to the average pin power density in the horizontal plane at elevation z.
For steady state conditions, these are obtained from three-dimensional ANC calculations using pin power reconstruction. For maneuvering and transient xenon conditions, a three-dimensional, one-dimensional synthesis technique (see Section 3.2) is used.
3.1.3 FUEL DEPl.ETION Three-dimensional fuel depletion calculations are performed with ANC. Itod-by-rod burnup distributions are obtained from the ANC depletions; specine fuel nuclide inventories are obtained from two-dimensioral single assembly PIIOENIX-P depletion calculations.
3.2 AXI Al POWEll DISTitillUTION CONTitOI.1.lMITS The axial power distribution is primarily affected by contcol rod position, xenon and burnup distributions, and temperature. Axial power 6:stribution control limits are used to ensure that thermal limits are not violated during power level 12
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changes, control rod motion, and the resulting xenon redistributions. This is accomplished by maintaining the axial nux diiTerence within acceptable boundaries. Axial Oux difference, al, is denned as the difference between the upper and lower excore detector signals, divided by the sum of these two signals.
Axial power distribution control limits are determined using Westinghouse *s llelaxed Axial Offset Control (llAOC) calculational procedure titeference 4). The RAOC calculational procedure begins by defining tentative al limits which are wider than the expected LOCA limits tor, alternately, the ItAOC al limits from the previous cycle may be used if it is desired only to verify their acceptability).
Xenon transient simulations are performed with the one dimensional ? ?OLLO code at various burnups and for different power levels, constrained by the tentative al limits and power dependent rod insertion limits. A library of axial xenon shapes is constructed at each burnup. Next, axial power shapes are generated with APOLLO for all possible combinations of xenon shapes, power levelo, and rod indertions. These axial shapes are synthesized with height dependent planer radial power distributions from three-dimensional ANC calculations. Imposition of the LOCA kw!ft limits for normal operation then defines the allowable al limits (or veriGes that the previous cycle's limits are acceptable) for the cycle. The axial power shapes corresponding to cases within the al limits are checked against thermal hydraulic constraints from loss of Gow accident simulations and the peak power and DNillimits for accident conditions.
For normal operations, more restrictive al limits are developed if either the kw!ft limits or thermal hydraulic constraints are exceeded. For accident conditions, analyses are performed to verify that all design limits are met if necessary, trip setpoints may be revised and/or the RAOC al limits tightened. Therefore, the RAOC procedure provides axial power shape information which is used to verify that all design limits are met. The itAOC al limits are placed in the Core Operating Limit Iteport and apply during plant operation.
3.3 COllE ItE ACTIVITY P AItA METEllS The core reactivity is affected by changes in the reactor which occur during operation as the result of fuel depletion, operator actions, and abnormal or accident conditions. Reactivity coefTicients quantify the rate of reactivity change 13
to be expected in response to changes in power, moderator or fuel temperatures, and soluble boron concentration. Itenctivity defects refer to the integral of the corresponding reactivity coefficient between two reactor statepoints with all other variables remaining constant. Xenon, samariun. and control rod worths are also typically required to fully define the change in reactivity between two core configurations. In addition, neutron kineties parameters are needed to describe the time dependent behavior of the core.
Quantification of these effects is needed for input to safety analy -
a well as to provide guidance to the reactor operators and ensure compliance with Technical Specifications. Therefore, the physics models described in Section 2 are used to calculate reactivity coefBcients, reactivity worths, and kinetics parameters as a function of core burnup, moderator temperature, and power level.
3.3.1 MODEll ATOlt TEM PRil ATUltE COEFFICIENT The moderator temperature coefGeient (MTC) is defined as the change in reactivity per degree change in moderator temperature. The effect of concomitant changes in moderator and soluble boron densities are included.
The MTC is sensitive to the values of the moderator density, moderator temperature, soluble boron concentration, fuel burnup, and the presence of control rods and/or burnable absorbers which reduce the required soluble baron concentration and increase the leakage of the core. The MTC may be positive or negative depending on the magnitude of change of the individual components of this coefficient.
The MTC is calculated using the ANC core model described in Section 2A by varying the moderator temperature around a reference temperature. The moderator temperature coefncient is analyzed for various reactor conditions, from hot zero power (HZP) to hot full power (IIFP), for various boron concentrations and control rod positions, and at various eyele burnups. The moderator temperature defect is also obtained using the ANC core model.
14
3.3.2 dol'I't.Elt TEMl' Ell ATUllE COICFFICIENT The Doppler temperature coemeient is defined as the change in reactivity per degree change in effective fuel temperature. The effective fuel temperature accounts for the spatial variation in fuel temperature throughout the core. The Doppler power coemeient represents the corresponding change in reactivity per percent change in reactor power. These coefficients are primarily a consequence of the Doppler broadening of U 238 and Pu-240 resonance absorption peaks which increases the effective resonance absorption cross scetion of the fuel with increasing fuel temperature.
The Doppler power coemeient is normally calculated using the ANC core model by varying the reactor power level about a reference power (which in turn varies the fuel temperature) while holding the product of the power level and the enthalpy rise constant. The Doppler power coemeient in then converted to a temperature coemeient using a power / temperature relationship obtained from FIGIITil calculations. The FIGitTH code provides effective fuel temperatures, which account for spatial variations in temperature within the pellet, as a function of power level and burnup. The Doppler coemeient is analyzed at different power levels and for various cytle burnups. Doppler reactivity defects can also be obtained using the ANC model by varying the reactor power at various times in life, while holding the product of the power level and the enthulpy rise constant.
At hot zero power, the Doppler temperature coemeient may he calculated by subtracting the moderator temperature coefficient from the isothermal temperature coemeient, provided the latter has been calculated explicitly (see Section 3.3 A).
3.3.3 TOTAL. I'OWElt CORFFICIENT The total power coemeient is defined as the change in reactivity per percent change in core power level. It represents the combined effect of moderator temperature and fuel temperature changes for an associated change in core power level.
15
The total power coefficient is calculated using the ANC core model by varying the core power level around a reference value while allowing the inlet temperature to change in accordance with the inlet program for the plant. The power cociTicient is analyzed at different power levels and at various times in core life. The power defect is also obtained using the ANC model by varying the reactor power.
3.3.4 ISOTilEllM A1, TEM PEll ATUltR COEFFICIENT The isothermal temperature coefficient (ITC) is defined as the change in reactivity per uniform degree change in core temperature. Normally calculated only at hot zero power, the ITC is the temperature coefficient directly measured during startup physics testing. The ITC can be calculated by summing the moderator temperature coeflicient and the Doppler temperature coefficient.
Alternately, the ITC may be calculated explicitly using the ANC core model by varying both the moderator temperature and the fuel temperature about a uniform reference temperature.
. The isothermal temperature defect (ITD) refers to the change in reactivity between hot zero power temperatures and temperatures below hot zero power.
ITDs are needed as a function of temperature and burnup for various rod patterns to establish shutdown boron concentration requirements. They are calculated with the ANC model using cross sections generated with PHOENIX-P at specific temperatures between hot zero power and 68*F.
3.3.5 llOltON llE ACTIVITY COEFFICIENT The boron reactivity coefficient, also referred to as the differential boron worth, is defined as the change in reactivity per ppm change in the soluble baron concentration. The inverse of the boron reactivity coefficient is referred to as the inverse boron worth. It provides a means of determining the change in soluble boron concentration necessary to compensate for a given reactivity change. The magnitude of the boron reactivity coefiicient depends primarily on the soluble boron concentration, the moderator temperature, control rod insertion, and the presence of burnable absorbers.
16 l'
?.
l l
The baron reactivity coeilicient is calculated using the ANC core model by
. perturbing the boron concentration in both directions about a reference value and computing the reactivity change. Iloron worths are calculated as a function of boron concentration, power level, temperature, burnup, and control rod configurationi l
3.3.6 XICNON AND SAM AIllUM WOllTilS 1
The iission products Xe 135 and Sm-149 possess large thermal absorption cross sections. Knowledge of the concentrations and reactivity worths of these isotopes as well as the changes which will occur in response to plant maneuvers is crucial to reactor control. Since Xe-135 is also produced by iodine decay, it initially builds up and then decays following a reduction in power or shutdown.
Sm-149 is a stable isotope produced by promethium decay. Following a reactor shutdown,its concentration increases. Upon restart it gradually returns to its equilibrium value.
r Equilibrium xenon and samarium worths are calculated with the ANC core model at various power levels and core burnups. Changes in their worth and axial fluctuations in isotopic concentrations during transient operation are obtained using the ANC and/or APOLLO models.
3.3.7 CON 101. ItOD WOllTIIS Control rod worth refers to the reactivity difference between two control rod configurations. The total control rod worth, trip reactivity shape (i.e., the l
inserted rod worth versus rod position), integral and difTerential worths of individual banks, and worths ofindividual rod cluster control assemblies (e.g.,
stuck, ejected, and dropped rods) are determined as required for startup physics testing, plant operations, and input to safety analyses.
(
Control rod worths are analyzed for all normal and many abnormal control rod configurations as a function of burnup, power level, and moderator I
temperature. Total rod worths and the integral worths ofindividual rod banks and rod clusters are calculated using the ANC core model. Differential rod worths are obtained with the ANC and/or APOLLO models.
17
3.3.8 NEUTitON EINETICS PAllAMETEllS Neutron kinetics parameters, which include de'ayed neutron fractions, decay constants, and the prompt neutron lifetime, are required as input to the plant reactivity computer and to various safety analyses. The parameters are also input to the Inhour equation to generate core reactivity as a function of startup rate and period. The kinetics parameters are evaluated at hot full power and hot zero power conditions for various cycle burnups and control rod configurations.
The PIIOENIX P cross section library contains delayed neutron fractions and decay constants for iissionable nuclides for each of the six delayed neutron energy groups. The core averaged delayed neutron fractions are obtained by weighting the delayed neutron fractions for each group by the regionwise l
fraction of fissions in each isotope and the regionwise power sharing in the core.
The core average decay constants are calculated in a similar manner. The fraction of fissions in each isotope are obtained from single assembly PHOENIX-P calculations. Regionwise power sharings for various core conditions are obtained using the ANC core model. A delayed neutron importance factor (to account for spectrum differences between delayed and prompt neutrons) is used to get an effective core average delayed neutron fraction.
The prompt neutron lifetime also depends upon the core composition (fuel enrichment, burnup, absorbers, etc.).
Single assembly PIIOENIX-P calculations provide the neutron lifetime for the fuel in each core region. The core average value is determined through a power and volume weighting process.
3.4 COllE PilYSICS PAItAMETEllS FOlt TitANSIENT ANA1.YSIS INPUT The physics models described in Section 2 are used to generate key input parameters for various safety analyses. Reference 5 provides a detailed-description of how these parameters are calculated and how they.are utilized 18
i l
l during the safety evaluation process for a reload. core. This section briefly discusses how these physics parameters are determined.
3A.1 SIIUTDOWN M AltGIN Shutdown margin is defined as the amount of negative reactivity by which the core is suberitical in a particular shutdown condition. At hot zero power conditions following a reactor trip, shutdown margins are calculated using the ANC core model. The highest worth control rod is assumed to remain fully withdrawn while xenon and soluble boron concentrations remain unchanged.
Doppler and moderator defects, flux redistribution, rod insertion prior to trip, and calculational uncertainties are accounted for. Shutdown margins are evaluated following trips from hot full power and hot zero power conditions as a j
function of burnup.
Shutdown margins are also evaluated for temperatures below hot zero power.
In' this case, no credit is taken for xenon. The minimum allowed shutdown requirement at a particular time in core life is typically determined by either the steamline break or boron dilution accident analysis.
3A.2 Titil' It E ACTIVITY The trip reactivity worth is defined as the control rod worth available for insertion at the time of reactor trip. It is determined with the ANC core model.
The highest worth control rod is assumed to be stuck in its fully withdrawn position and the remainder of the rods are assumed to be at the rod insertion limit. Calculations are made as a function of burnup and the most limiting (i.e.,
smallest) value is used.
The-trip reactivity shape refers to the amount of reactivity insertion as a function of rod position following a trip. It is determined using the ANC and/or APOLLO models. The most limiting shape corresponds to the minimum rate of inserted rod worth as a function of rod position, and generally results from the most bottom skewed axial power distribution. Power distributions are calculated as a function of burnup to identify the most bottom skewed power profile.
19
3.4.3 CONTROL. ROD MISAl.IGNMRNT A thermal-hydraulic analysis is required to show that the departure from nucleate boiling ratio (DNBR) criteria is not violated in the event of a misaligned control rod. Misalignment accidents include:
a.
dropped control rods, b.
statically misaligned control rods, and c.
a single rod withdrawal.
For core configurations with dropped control rods, control rod worths and nuclear enthalpy rise hot channel factors (Faii) are determined using t ae ANC core model. Rods dropped from both the unrodded state and with contral rods at the rod insertion limit are considered. This information is used to confirm that DNBR limits will not be violated during the traasient following a dropped control rod event.
Statically misaligned control rods are also analyzed using the ANC core model.
The maximum value of Fati s determined as a function of power level to ensure i
that DNBR limits will not be violated. The extreme cases of static misalignment are represented by the single dropped rod discussed above and by a rod being fully withdrawn while the remaining rods in the same bank are at the rod insertion limit. Calculations for the latter type of core configuration also provide Fati and pin power census data to verify that less than 5% of the rods in the core will violate the DNBR limit before the plant trips during a single rod withdrawal accident.
- 3.4.4 HORON Dil.UTION Baron dilution refers to the inadvertent admissian of unborated water into the
- Reactor Coolant System (RCS), diluting the boron concentration and causing a
. loss of shutdown margin. Limiting values of the initial baron concentration and boron worth to be used in the analysis are determined with the ANC core model fci refueling, cold and hot shutdown, hot standby, startup, and at power conditions. For cold and hot shutdown and hot standby conditions, the analyses 20
establish shutdown baron concentrations necesuary to preserve required operator action times at higher RCS boron concentrations.
3.4.5 CONTitOI. ROl) II A NK WITilDR A W Al.
]
The withdrawal of control rod banks is analyzed to confirm that the maximum differential rod worth does not exceed the limiting values assumed in _ safety analyses. At full power, the control banks are withdrawn in their normal sequence. To simulate worst case rod bank withdrawal from suberitical conditions, each combination of two banks adjacent in the withdrawal sequence are moved with 100% overlap at hot zero power. Calculations are performed with the ANC and/or APOLLO models taking into account the effects of adverse axial xenon distributions i
L l
3A.6 CONTitOI. ROD EJECTION l
The rod ejection accident is analyzed to confirm that limits on peak clad temperature and maximum energy deposition in the fuel are not exceeded and that the reactor remains suberitical following the transient. Ejected rod worths and heat Oux hot channel factors (F.m) are determir,?d using the ANC core model. Calculations are performed for various power levels at the beginning and end of cycle with control rods inserted to the rod insertion limit. Also, trip reactivity is evalutted for a combination of two control rods unavailable: one stuck and one ejected.
3A.7 STE AMl.1NE RREAK Key input parameters to the thermal-hydraulic and transient analyses of a hypothetical steamline break are determined with the ANC core model. These include temperature and power coefficients of reactivity, control rod and soluble boron worths, location of the most limiting (from a DNBR standpoint) stuck rod, kinetics parameters, and shutdown margin. The break is conservatively assumed to occur at the end of cycle with the plant suberitical at the hot zero power temperature. After statepoint parameters (inlet temperatures, pressure, now, and thermal power) characteristic of the time at which the maximum power occurs during the transient are determined, radial 21
and axial power distributions, enthalpy rise hot channel factors, and reactivity information are calculated The assumed reactor conditions are representative of a non uniforni coolant inlet temperature distribution with the most limiting control rod stuck out of the core and all other rods fully inserted. This information is then used to perform a thermal hydraulic evaluation of DNillt.
22
4.0 PilYSICS MODEL VERIFICATION Core physics model verification typically includes comparisons of predictions to plant startup and operating data. The V. C. Summer Nuclear Station (VCSNS)is currently in its sixth cycle of operation. In this section, predictions made using the physics methodology described in Section 2 are compared to zero power physics test measurements and at power operating data accumulated over the past three cycles. As stated in Section 1, the methods employed to generate the predictions reported in this section are standard licensed methods used by Westinghouse's Commercial Nuclear Fuel Division. The comparisons reported herein provide additional verification of the predictive capabilities of this methodology: however, their primary purpose is to demonstrate SCE&G's ability to perform design calculations for VCSNS.
4.1 CYCL.E DESCillPTIONS Cycle 3 of VCSNS began operation on December 14, 1985, and shutdown on March 6.1987, after 411 effective full power days (EFPD) corresponding to a cycle burnup of 15792 h!WD'MTU. The core loading pattern for Cycle 3, including the locations and number of part length pyrex burnable absorbers and the locaFons of control rod banks, is shown in Figure 4,1-1.
A quarter core representation is used since the core loading is symmetric. Four Vantage-5 Demon stration Assemblies containing Integral Fuel Burnable Absorbers tIFBAs), initially loaded in Cycle 2, were returned to the core.
Cycle 4 of VCSNS began operation on June 6,1987, and shutdown on September 16,1988, after 424 EFPD corresponding to a cycle burnup of Iti2ti4 MWD'MTU.
The core loading pattern for Cycle 4, including the locations and number of wet annular burnable absorbers (W AB As), is shown in Figure 4.1-2.
The four Vantage-5 Demonstration Assemblies were also returned to the core for Cycle 4.
Cycle 5 of VCSNS began operation on December 26,1988, and shutdown on March 23,1990, after 346 EFPD corresponding to a cycle burnup of 13670 MWD'MTU. The core loading pattern for Cycle 5, including the locations and number ofIFBAs,is shown in Figure 4.1-3. Fuel batch characteristics for Cycles 3,4, and 5 are summarized in Table 4.1-1.
23 filk
l l
4,2 ZRitO l'OWElt I'llYSICS TESTS After each refueling at VCSNS, startup physics tests are conducted to verify that the nuclear characteristics of the core are consistent with design predictions.
While the reactor is maintained at hot zero power (IIZP) conditions, the following physics parameters are measured:
1 Critical boron concentrations, lsothermal temperature coefficient.
Control rod worths, and Differential bcron worth.
Table 4.21 contains the zero power physics test review criteria, which represent the maximum expected deviation between predicted and measured values for each parameter.
The following sections briefly describe the measurement and calculational techniques and summarize the results of the zero power physics tests for Cycles 3, 4, and 5. Small changes in core reactivity were measured by feeding the signal from the spare power range neutron detector into a reactivity computer which solves the point kinetics equation. The computer output was plotted on a strip chart recorder. All predictions were made with the three-dimensional ANC model described in Section 2.4.
4.2.1 ClllTIC Al. HOttON CONCENTit ATIONS Critical baron concentrations were measured by acid-base titration of reactor coolant samples taken under equilibrium conditions. Samples were taken with all rods out ( ARO), that is, fuPy withdrawn from the core, and with Bank B, the reference bank (see Section 4.2.3), fully inserted. Critical boron searches were performed with the three-dimensional ANC model for these core configurations to obtain the predicted concentrations. The measured and predicted eritical baron concentrations are compared in Table 4.2-2. All differences are within the + 50 ppm review criteria.
24
f 4,2.2 ISOTilF.ltM A1.TICMl'ICit ATUltl? COlsFFICIRNTS isothermal temperature coefficients ilTCsi were measured by making small changes in the reactor coolant system temperature and determining the corresponding change in reactivity with the reactivity computer. ITCs were predicted by uniformly varying the core temperature by +yF about the 11Z1' temperature in the ANC model. The moderator temperature is varied directly; Doppler elTects on reactivity are determined using fitting coefficients obtained from FIGilTil calculations. The measured and predicted ITCs are compared in Table 4.2 3. All differences are well within the review criteria of13 pem '"F.
4.2.3 CONTitOI. ItOD WOltTilS Control rod worths were measured by the Itod Swap Technique. First, the worth of the reference bank (the bank of highest worth) was measured by boron dilution. Stepwise bank insertion was used to maintain criticality and differential worths were obtained from the reactivity computer response. The differential worths were sununed to provide the integral wsth of the reference bank. Then, maintaining the boron concentration at a constant value, critical configurations were established with each remaining bank fully inserted and the reference bank partially withdrawn. The integral worth of each inserted bank was determined from the critical position of the reference bank after the exchange by applying analytical corrections to account for the effect of the inserted bank on the partial integral worth of the reference bank. This procedure is described in detail in Iteference 6.
The ANC model was used to predict the individual control rod bank worths as well as to generate the corrections used to infer the measured worths. The measured and predicted worths are compared in Table 4.2-4: all differences are within the review criteria listed in Ta, 4.2 1.
Measured and predicted reference bank integral rod worth shapes are compared in Figures 4.21 through 4.2 3.
25 I
l l
4.2.4 DIFFEltENTI Al. llOltON WOltTilS hieasured differential boron worths were obtained by dividing the measured reference bank worth (see Section 4.2.3) by the difference between the critical boron concentrations measured with all rods out and with the reference bank inserted. The differential boron worth does not change significantly over this range of boron concentration. Boron worths were predicted by varying the baron concentration by 15 ppm about the IIZP all rods out critical baron 2
concentration in the ANC model. The measured and predicted boron worths are compared in Table 4.2-5. All differences are well within the il5Ce review criteria.
4.3 POWElt OPEll ATION In support of VCSNS Technical SpeciGeation requirements, the core power distribution is measured at least once every 31 EFPD using the incore instrumentation system. Neutron flux measurements made by movable incore fission chambers are combined with analytically determined power to reaction rate ratios using the computer program INCORE (Reference 2) to infer, i.e.,
" measure," a three-dimensional power distribution. The power to reaction rate ratios are generated with the three-dimensional ANC model using cross sections derived from PHOENIX-P.
In this section, measured data obtained from INCORE is compared to predictions made with the three-dimensional ANC Alodel. Included are; e
Power peaking factors, F and Fati, q
+
Average assembly radial power distributions, Core average axial power distributions, and e
e Axial offset.
Also, measured and predicted boron letdown curves are compared. Boron letdown refers to the reduction of the all rods out hot full power critical boron concentration as a function of core burnup.
26
i 4.3.1 IlOllON l.RTI)OWN CUltVI:S Reactor coolant system boron concentrations are measured daily regardless of I
power level or control rod bank insertion. Critical boron concentrations measund at or very close to hot full power all rods out equilibrium xenon and samarium conditions are compared to the predicted baron letdown curves for Cycles 3. 4, and 5 in Figures 4.31 through 4.3.3. The predicted curves were obtained from design depletions with the three dimensional ANC model.
Tables 4.31 through 4.3 3 compare measured and predicted critical boron concentrations at the time ofINCOftE power distribution measurements. The measured concentrations were corrected to hot full power all rods out
. equilibrium xenon' and samarium conditions in accordance with VCSNS.
l surveillance procedures. The predicted ceneentrations were obtained by.
performing critical baron searches with the ANC model at the specific burnups of the measurements. The mean difference between measured and predicted critical baron concentrations for all three cycks is 12 ppm with a standard deviation of 19 ppm.
4.3.2 l'OWRit l'E ARING FACTOltS The nuclear enthalpy rise hot channel factor (Fato and the heat flux hot channel factor (F ) were measured using the INCORE code, as discussed above.
y Predicted peaking factors were obtained from three dimensional ANC calculations performed for core conditions similar to those at the time of the measurements. Power peaking factors measured during Cycles 3,4, and 5 are compared to predicted valuesin Figures 4.3 4 through 4.3-9 and in Tables 4.3-4 through 4.3 6.
For Fali, the mean difference between the measured and predicted values for the three cycles is 0.58% with a standard deviati m of 1.117c; for F the mean difference is 2.077c with a standard deviation of 1...%.
q Regarding the F comparisons,it is noted that spacer grid effecto are inherent u
in the measured values but the grids are not explicitly modeled in ANC. The-magnitude of this effect can be seen from Figures 4.3 28 through 4.3-45.
27
i 4.3.3 ItADI Al. I'OWElt 1)lSTitillUTIONS Core power distributions were measured with the INCollE code, as discussed above. The measured power distributions are typically referred to as flux maps.
INCORE also produces predicted power distributions at the burnup of the flux map by interpolating between power distributions generated using the three-dimensional ANC model at specific burnups during a depletion calculation.
Since the core is loaded symmetrically, ANC depletion calculations are performed assuming quarter-core rotational symmetry. The predicted power distributions are expanded to full core for comparison to the measured distributions.
Figures 4.310 through 4.3 27 compare measured and predicted assembly relative power distributions at selected burnups spanning Cycles 3. 4. and 5.
All comparisons are for the hot full power all rods out condition since this is the normal mode of operation at VCSNC The mean absolute difference between measured and predicted assembly relative powers is less than.014 and the standard deviation is less than.017 for these comparisons.
4.3.4 AX1 Al. l'OWElt 1)lSTitillUTIONS AND AXI Al. OFFSETS Measured core overage axial power distributions from each of the flux maps discussed in the previous section are compared to predicted axial distributions in Figures 4.3 28 through 4.3 45. The predicted distributions were obtained from three dimensional ANC calculations performed for core conditions similar to those at the time of the Oux maps. Note that since the grid straps are not modeled explicitly in the ANC model, no depressions are seen at the grid locations in the predicted distributions. This difference coupled with the normalization of both measured and predicted axial power distributions to unity causes the measured relative power to appear slightly higher between grid locations.
Axial offset refers to the percent difference between the relative power in the top half of the core and that in the bottom half of the core. Axial offsets measured using the INCORE code are compared to predicted values from ANC calculations for core conditions similar to those at the time of the 28
measurements in Tables 4.3 7 through 4.3 9. The mean dilTerence between measured and predicted values for Cycles 3,4 und 5 is (1.79 with a standard deviation of 0.9%.
4.4 SUM M Al(Y In this section, predictions made using Westinghouse's reload core design methodology are compared to zero power physics test ineasurements and at power operating data from Cycles 3,4, and 5 of VCSNS. In all cases, the predictions agree very well with the measurements. All startup test predictions are within the review criteria listed in Table 4.2-1. Predicted critical horon concentrations at power are well within 50 ppm of the measured values, and the predicted power distributions are quite close to the measured, as evidenced by Figures 4.3-10 through 4.3 45. The excellent agreement between the predictions and the measurements reported here demonstrates SCE&G's capability to apply the Westinghouse licensed methodology to reload core design for VCSNS.
29
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4 d
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n
TABLE 4.1-1 V.C. SUMMER NUCLEA A STATION FUEL SPICIFIC 4 TION initial BOC Number of Ennchment Burnup Cycle Batch Assemblies w/o U 235 MWD /MTU 3
2 5
2.61 20549 3
40 3.11 20632 4
40 3.44 10876 4A(a) 4 3.42 11127 5
68 3.84 0
4 4
21 3.44 25833 4Ata) 4 3.42 29743 5
76 3.84 16395 6
56 3.60 0
5 3
1(b) 3.11 24855 4
4 3.44 18877 5
40 3.84 26033 6
48 3.60 18909 7Ata) 36 3.79 0
7B(a) 28 4.20 0
(a) Vantage-5 Fuel Design (b) Discharged at EOC2 31
V.C. SU M M N CLEA STATION HZP PHYSICS TEST REVIEW CRITERIA li Parameter Review Critena Critical Boron 150 ppm Concentrations Isothermal Temperature 13 pcmrF Coefficient Control Rod Bank Worths:
Reference Bank Worth 110%
" Swap" Worths 1 15% or 100 pcm, whichever is greater Dif f erential Boron Worth 1 15 %
32 l
l l
TABLE 4.2-2 V.C. SUMMER WUCLEAR STATION CRITICAL DORON CONCENTRATION COMPARISON BETWEEN MEASUREMENT AND PRE 01CTION FOR CYCLES 3,4 AND 3 Critical Boron Concentrati n (ppm)
Bank C_ycle Configuration M
P (M P) 3 ARO 1840 1837 3
BANK B in 1635 1648
-13 4
ARO 1831 1807 24 BANK B in 1650 1628 22 5
ARO 1969 2013
-44 BANK B in 1821 1868
-47 33
TABLE 4.2 3 V.C. SUMMER NUCLEAR STATlON ISOTHERMAL. TEMPERATURE COF FIDE NT COMPARISON BETWEEN MEASUREMENT AND PREDICTION FOR CYCLES 3.4 AND 5 ITC (ptmf F)
Bank Cycle Confiauration M
P (M P) 3 ARO 1.95 1.09
-0.86 4
ARO 2.89 2.23
-0.66 5
ARO
+ 3.15
+ 2.35 0.80 1
N t
9 0
34 q
ulimmuMq tum uumf us i
TABLE 4.2 4 V.C. SUMMLR NUCLL AR STATION CONTROL 8100 WORTH COMPARISON BETWEEN MEASUREMENT AND PR[DICllON f 081 CYCLES 3,4 AND 5 Control Rod Worth (Dcm)
(.MF - )qi,
.P Bank M (a)
,,_ P Cycle Configuration 3
Bonk D 878 8 967.0 9,12 Bank C 738 4 854.0 13.54 Bank B (b) 1393 5 1398.0
-0 32 Bank A 605.0 587.0 3.07 BankSB 780.2 902.0
-13.50 BankSA 1204.4 1184.0 1.72 4
Bank D 1089.6 1070.5 1.78 Bank C 845.9 816.5 3.60 Bcnk B (b) 1353.9 1324.2 2.24 Bank A 443.7 425 2 4.35 Bank 5B 882.1 869.2 1 48 BankSA 1134.1 1124.4 0.86 5
Bank D 932.0 1003.7
-7.14 Bunk C 654.2 727.4
-10.06 Bank B (b) 1077.9 1125.6
-4.24 Bank A 626.5 621.9 0.74 BankSB 811.4 910.9
-10.92 BankSA 956.5 991.7
-3.55 (a) Measured Rod Worths were determined using " Rod Swap" Methodology (b) Reference Bank j
l 35 4
TABl.E 4.2 5 V.C. SUMMER NUCLEAR $TATION HZP DIFf ERENTIAL BORON WORTH COMPARISON BETWEEN MEASUREMENT AND PREDICTION FOR CYCLES 3,4 AND 5 Differential Boron Wortn (pcm/ ppm)
M-P Bank
( T )g; Cycle Confiouration M
P 3
Average Over 6.80 7.40 8,11 B ANK B Insertion 4
7.48 7.40 1.08 Average Over B ANK B Insertion 5
7.28 7.76 6.19 Average Over B ANK B Insertion i
36
TABLE 4.31 V.C. 5UMMER NUCLEAR 5 TAT 10N CYCLE 3 BORON LETDOWN COMPARISON DETWEEN MCA$llREMENT AND PREDICTION Critical Boron Cor. centration (ppm)
Cycle Burnup MWD /MTU M
P (M P) 617 1202 1233
-31 1194 1190 1197 7
1839 1155 1162 7
2341 1144 1132 12 2860 1105 1102 3
3365 1081 1071 10 3920 1045 1034 11 4758 1006 979 27 4988 989 965 24 5558 952 923 29 6345 909 870 39 7111 850 815 35 7606 805 779 26 8126 763 741 22 8689 713 696 17 9389 675 646 29 l
9646 637 626 11 10624 568 548 20 11123 520' 505 15 12435 410 396 14 l
12608 403 382 21 13581 312 296 16 i
14057 268 255 13 L
14867 198 183 15 l-37 l
l
~
TA8LE 4.3 2 V.C. SUMMER NUCLE AR STATION CYCLE 4 BORON LETDOWN COMPARISON 6ETWEEN MEASUREMENT AND PREDICTION Critical Boron Concentration (nom)
Cycle Burnup MWD /MTU M
P (M P) 397 1176 1211 35 1232 1161 1152 9
2007 1099 1108
-9 3164 1060 1034 26 4154 984 968 16 5106 916 900 16 6398 814 803 11 7548 733 714 19 8608 663 631 32 9783 561 537 24 10850 468 447 21 12192 352 329 23 12784 301 277 24 13818-193 183 10 14970 96
-77 19 l
38 I
l
TABLE 4.3 3 V.C.$Uf 4MER NUCLEAR STATION CYCLE 5 BORON LE1DOWN COMPARISON BETWEEN ME ASUREMENT AND PREDICTION Critical Boron Concentration (opm)
Cycle Burnup MWD /MTU M
P (M P) 423 1440 1480
-40 2201 1356 1384
-28 3260 1261 1303
-42 4090 1224 1230
-6 5064 1135 1136 1
6174 1031 1021 10 7438 903 887 16 8168 828 810 18 9354 709 687 22 10539 588 565 23 11647 488 453 35 12793 373 338 35 39
TA8LE 4.3 4 V.C.5UMMIR NUCLEAR STATION CYCLE 3 POWER PE AKING F ACTOR COMPARISON BETWEEN MEASUREMENT AND PREDICTION F As (Max)
Fn(Max)
Cycle 4
.P )%
( M.P )%
M Burnup T
M P
P MWD /MTU M
P 617 1.403 1.401 0.14 1.732 1.719 0.76 1194 1 403 1.395 0.57 1.712 1.709 0.18 1839 1.397 1.392 0.36 1.701 1.685 0.95 2341 1.393 1.393 0.00 1.687 1.672 0.90 2860 1.387 1.391 0.29 1.675 1.658 1.03 3365 1.384 1.391
-0.50 1.669 1.648 1.27 3920 1.382 1.390
-0.58 1.668 1.642 1.58 4758 1.381 1.388
-0.50 1.645 1.637 0.49 4988 1.382 1.385
-0.22 1.640 1.628 0.7/.
5558 1.380 1.385 0.36 1.640 1.631 0.55 6345 1.379 1.381
-0.14 1.629 1.620 0.56 7111 1.376 1.380
-0.29 1.614 1.615
-0.06 7606 1.376 1.379
-0.22 1.628 1.615 0.80 8126 1.379 1.378 0.07 1.629 1.616 0.80 8689 1.373 1.377
-0.29 1.638 1.618 1.24 9389 1.377 1.376 0.07 1.632 1.625 0.43 9646 1.377 1.376 0.07 1.635 1.630 0.31 10624 1.388 1.379 0.65 1.635 1.624 0.68 11123 1.384 1.380 0.29 1.636 1.619 1.05 12435 1.388 1.384 0.29 1.636 1.624 0.74 12608 1.390 1.385 0.36 1.658 1.625 2.03 13581 1.389 1.382 0.51 1.618 1.619
-0.06 14057 1.383 1.381 0.14 1.649 1.616 2.04 14867 1.384 1.379 0.36 1.632 1.606 1.62 40
TABLE 4.3 5 V.C. SUMMER NUCLEAR STATION CYCLE 4 POWER PE AKING FACTOR COMPARISON BETWEEN MEASUREMENT AND PREDICTION F s(Max)
Fn(Max) 3 Cycle Burnup (MP)%
( M P )q; MWD /MTU M
P P
M P
T~
397 1.388 1.394
-0 43 1.566 1 494 4.82 1232 1.382 1.387
-0.36 1.580 1.508 4.77 2007 1.385 1.38' O.29 1.589 1.514 4.95 3164 1.392 1.388 0.29 1.603 1.542 3.96 4154 1.401 1.393 0.57 1.607 1.559 3.08 5106 1.406 1.396 0.72 1.622 1.572 3.18 6398 1.411 1.404 0.50 1.632 1.588 2.77 7548 1.414 1.408 0.43 1.644 1.604 2 49 8608 1.419 1 413 0 42 1.637 1.610 1.68 9783 1.410 1.413
-0.21 1.654 1.609 2.80 10850 1.416 1.414 0.14 1.646 1.602 2.75 12192 1.407 1.409 0.14 1.625 1.579 2.91 12784 1.408 1.406 0.14 1.627 1.568 3.76 13818 1.398 1.398 0.00 1.609 1.556 3.41 14970 1.388 1.388 0.00 1.586 1.543 2.79 41
TA8LE 4.3 6 V.C. 5UMMER NUCLEAR STATION CYCLE 5 POWER PEAKING F ACTOR COMPARISON BETWEEN MEASUREMENT AND PREDICTION 1
F3s (Ma x)
Fo (Ma x)
Cycle Burnup
( M-P )og (TM P yo,6 MWD /MTU M
P P
M P
l 423 1.464 1.438 1.81 1.926 1.863 3.38 2201 1.494 1.456 2.61 1.887 1.818 3.80 3260 1.491 1.447 3.04 1.847 1.769 4.41 4090 1.496 1.436 4.18 1.841 1.743 5.62 5064 1.470 1.421 3.45 1.771 1.730 2.37 6174 1.459 1.406 3.77 1.759 1.716 2.51 7438 1.419 1.393 1.87 1.735 1.692 2.54 8168 1.411 1.387 1.73 1.700 1.675 1.49 9354 1.395 1.377 i.31 1.685 1.652 2.00 10539
. 1.381 1.366 1.10 1,703 1.636 4.10 11647-1.375 1.362' O.95 1.653 1.628 1.54 12793 1.373 1.358 1,10 1.641 1,624 1.05 r
J P
42
.mr-y.-.
r..',,,
-e
.--,a-.m_-..-.3.._,w,,,.,..v ww.-
r
.=
TABLE 4.3-7 V.C. $UMMER NUCLEAR STATION CYCLE 3 AXl AL Of FSET COMPARI$0N BETWEEN MEA $UREMENT AND PREDICTION Axial Of fset (%)
i Cycle Burnup MWD /MTU M
P (M P) 617
-0.43 0.88 0.45 1194 0.13 1.51 1.38 1839 1.22
-2.06 0.84 2341 1.57 2.45 0.88 2860 1.27 2.73 1.46 3365 1,31 3.06 1.75 3929
-2.59 3.39 0.80 4758
-2,15
-3.87 1.72 4988 2.27
-3.86 1,59 5558 3.10
-4.18 1.08 6345 2.42 4.27 1.85 7111
-2.37
-4.49
- 2. i 2 7606
-3.24
-4.52 1.28 8126 3.20 4,54 1.34 8689
-3.92 4.52 0.60 9389 3.29
-4.39 1.10 9646 3.57
-4.22 0.65 10624
-3.12
-3.82 0.70 11123 2.98 3.57 0.59 12435 2.81
-2.91 0.10 12608
-4.10 2.79
-1.31 13581
-2.41 2.52 0.11 14057
-3.40
-2.41
-0.99 14867 2.97
-2.19 0.78 l
l l^
~
43
~_
TABLE 4.3 8 V.C. 5UMMER NUCLEAR STATION CYCLE 4 AXIAL OFF5ET COMPARISON BETWEEN MEASUREMENT AND PREDICTION Anial Offset (%)
Cycle Burnup MWD /MTU M
P (M P) l 397 2.30 0.17 2.47 1232 0.72 1.22 1.94 2007
-0.59 1.90 1.31 3164 2.29 2.79 0.50 4154
-2.56
-3.31 0.75 5106
-2.63 3.66 1.03 6398
-3.41
-3.93 0.52 7548
-3.67
-4 19 0.52 8608 3.31
-4.22 0.91 9783
-4.05
-3.91
-0.14 10850 3.56
-3.42 0.14 12192 3.38 2.85
-0.53 12784
-3.32 2.61
-0.71 13818 3.20 2.46 0.74 14970 2.76
-2.39 0.37 44
_,n.,
,--..--w
TA8LE 4.3 9 V.C. $UMMER NUCLEAR STATION CYCLE $
AXtAL OFF$CT COMPARISON BETWEEN MEASUREMENT AND PREDICTION Axial Offset (o )
o Cycle Burnup MWD /MTU M
P (M P) 423 3.82 3.15 0.67 2201 2.62 0.41 2.21 3260 0.25
-1.13 1.38 4090 1 67 1.98 0.31 5064
-1.58 2.80 1.22 6174 2.38
-3.18 0.80 7438 2.68 3.33 0.65 8168
-1.78 3.38 1.60 9354
-2.29 3.46 1.17 10539
-4.05 3.38
-0.67 11647 2.68 3.29 0.61 17793 2.89
-3.30 0.41 45
t i
i t
o I
i i
?
r E
l
- 1'his page intentionally bisnk."
i 7
6 tr a
3 I
r
.I I
P i
I 1
l l
46
.._._z..._______.__.,_._-
FIGURE 4.1-1 V.C. 5UMMER NUCLE AR STATION CYCLE 3 CORE LOADING 4
H G
F E
D C
B A
1 5
4 3
4 4A 4
2 16 40(a) 3 C
D 5
3 5
3 5
4 5
4 9
5 3
5 4
5 5
16 20 24 10 C
D B
A 3
3 5
3 5
5 3
11 20 16 4
SB 4
5 4
5 4
3 12 24 16 8
C 4A 4
5 5
3 13 40(a) 24 4
5A 4
5 5
3 14 12 D
A 2
4 LEGEND BATCH
- WABA's RCCA(b)
(a) Depleted IFBA Rods (b) Control Rod Banks (D,C,B,A) or Shutdown Banks (SB,5A) 47
FIGURE 4.1-2 V.C. SUMMER f3UCLE AR STATION CYCLE 4 CORE LOADING H
G F
E D
C B
A 4
5 5
5 4
5 5
4 16 20 8
C D
5 4
6 5
6 5
6 5
9 16 20 20 4
6 4
6 5
6 5
10 20 16 12 C
D B
A 5
5 6
4A 6
6 5
11 16 40(a) 8 8
SB 4
6 5
6 4
5 12 20 8
B C
5 5
6 6
5 13 20 12 8
5 6
5 5
14 4
D A
4 5
5 BATCH I
- WAB A's RCCA(b)
(3) Depleted IFBA Rods (b) Control Rod Banks (D C,B,A) or Shutdown Banks (SB,5A) 48
F!GURE 4.1-3 V.C. SUMMER NUCLEAR STATION CYCL E 5 CORE LOADING H
G F
E D
C B
A 3
7A 5
5 5
78 4
6 64 64 8
C D
7A 5
7A 6
6 6
78 6
7A 6
7A 6
7A 78 10 64 80 80 C
D B
A 5
6 7A 5
7A 78 5
11 80 80 SB 5
6 6
7A 5
5 12 80 B
C 78 6
7A 78 5
13 64 80 4
78 78 5
14 64 D
A 6
6 LEGEND 15 r7ATCH
- IFB A's RCCA(a)
(a) Control Rod Banks (D.C,B, A) or Shutdown Rod Banks (SB,5A) 49
4 FIGURE 4.21 v.C. SUMMER NUCLEAR STATION CYCLE 3 MEASURE D VE R5US PREDICTED BANK B INTEGRAL ROD WORTH 1,400 i % -
1,300
\\
1,200 MEASURED x
\\
Pe _
1,100
\\
N 1,000
\\
e G
300
\\
800 L
W 700 L
O R
T 600 H
N?=
500 P
C 400 300 200 s9
,k.
100 0
1 0
20 40 60 80 100 120 140 160 180 200 220 240 B ANK POSITION, STEPS WITHDRAWN 50
FIGURE 4.2-2 V.C, SUMMER NUCLEAR STATION CYCLE 4 MEASURLD VER$US PREDICTED BANK B INTEGRAL ROD WORTH i
1,400 I
a x
1,300 '*w"*-
s 1,200 s
MEASURED x
l I
1.100 "N
i l
"x" N
1 000 f
hi ",',
G 900
\\\\
R A
i",
800 L
W 700 N
O N "5 R
N T
600 4
H N*g 500 P
N C
400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 B ANK POSITION, STEPS WITHDR AWN 51
i J
FIGURE 4.2 3 V.C. SUMMER NUCLEAR STATION CYCLE 5 MEASURED VERSUS PREDICTED BANK B INTEGRAL ROD WORTH 1,400 1,300 1,200 MEASURED x
PREDICTED N
1,000 sg' G
900 A
N 800 W
700 0
R T
600
\\
500 P
C 400 M
h 300 200
\\
100
\\\\
0
=
0 20 40 60 80 100 120 140 160 180 200 220 240 B ANK POSITION, STEPS WITHDRAWN 52
FIGURE 4.3 1 V.C. SUMMER NUCLEAR STATION CYCLE 3 BORON LETDOWN COMPARISON i'
BETWEEN MEASUREMEN1 AND PREDICTION 1,600 i
1,500 1,400 Measured x
1,300 Predicted
-Ea, 1,200
- 2 1,100 g
1,000 5
900 Z0 800 i
v
]g 700-3 a
600 E
500 cc*
4
. h, 9M 300
\\,
200 100 0
0
- 2,000 - 4,000 6,000 8,000
_10,000 12,000 14,000 17,000 CORE AVERAGE BURNUP, MWD /MTU 53 uv>
,m.,
..m-m..
y,
-n
.,y..
,,.v.4,iv.,
,4,+-
, -._w m -y w m,<
p-y vy r.<i-e sv
FIGURE 4.3 2 V.C. SUMMER NUCLEAR STATION CYCLE 4 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION 1,600 1,500 1,400 Measured x
1,300 Predicted
\\
1,200 S
2 1,100
\\
S h
1,000 4
Y w
900 Eo 800 l' '
v 3
700 8
k' a
600 b
t-500 p
400 300 N,,
200 100 s
0 S.
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 17,000 CORE AVERAGE BURNUP, MWD /MTU 54
l FIGURE 4.3 3 V.C. SUMMER NUCLE AR STATION CYCLE 5 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION 1,600 1,500 1,400
~~ '
Measured x
=
1,300 Predicted
^
k 1,200 k
S 2
1 h
,100 1.000 5
3(
3 900
\\
2
.l 8
800
-h
=
700 b
600 a
0 35 t-500 400 lb'N-300 200 100
[
0 0
2,000 4,000 6,000 8,000 10,000 11,000 14,000 17,000 CORE AVERAGE BURNUP, MWD /MTU 55 l
l
FIGURE 4.3-4 V.C. SUMMER NUCLEAR STATION CYCLE 3 F-DELTA-H COMPARISON BETWEEN INCORE AND ANC 1.9 INCORE x
ANC 1.8 i
y 1.7 F
1.6 3
+-
l D
E L
1.5 I
T
{
A
- = - + - -
H 14
_ A.__
^. x >
r.
y 1.3 1.2 1
1.1 1
4 0
2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18.000 CORE AVERAGE BURNUP, MWD /MTU 56
FIGURE 4.3 5 V.C. SUMMER NUCLEAR STATION CYCLE 4 F DELTA HCOMPARIson BETWEENINCORE AND ANC 2
l I
\\
1.9 i
INCORE x
l ANC 1.8 l
l 1.7 F
1.6 D
E L
1.5 I
T A
=
1 I4 H
g f
I 1.3 1.2 M
1.1 gg6 0
2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 CORE AVERAGE BURNUP, MWD /MTU 57
FIGURE 4.3 6 V.C. SUMMER NUCLEAR STATION CYCLE 5 F DELTA H COMPARISON BETWEENINCORE AND ANC 2
1.9 INCORE x
1.8 ANC 1.7 F
1.6 D
E L
1.5 T
h-A H
14 N wY
=
x I
1.3 f
1.2 1.1 i
1 0
2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 l
CORE AVERAGE BURNUP, MWD /MTU l
l 58 i
l.
PGURE 4.3-7 V.C. SUMMER NUCL EAR STATION CYCLE 3 Fq COMPARISON BETWEEN INCORE AND ANC l
2.5 INCORE x
2.25 ANC 2
Fo 1.75 g N %><
w; %
m a
1.5 1.25 1
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 CORE AVERAGE BURNUP, MWD /MTU 59
FIGURE 4.3 8 V.C. SUMMER NUCLEAR STATION CYCLE 4 F A COMPARISON BETWEENINCORE AND ANC 2.5 INCORE x
2.25 ANC 2
Fo 1.75
=
=
x
=
x
'N 1;
y 1.5 1.25 1
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 CORE AVERAGE BURNUP, MWD /MTU 60
FIGURE 4.3 9 V.C. 5UMMER NUCLEAR STAT 10N CYCLE S Fq COMPARISON BETWEENINCORE AND ANC 2.5 INCORE x
2.25 ANC 2
i
=
N s
\\
Fo 1.75
\\
=
F w
1.5 1.25 1
0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000 18,000 CORE AVERAGE BURNUP, MWD /MTU 61
1 I
FIGURE 4.310 V.C. SUMMER NOCLEAR STATION CYCLE 3 RADI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 03 007 R
P N
M L
K J
H G
F E
D C
B A
0 427 04020 0 835 1
0 419 0395 0 422 1.91 1 77 3 08 0 415 0.924 1 010 1.027 1.041
- O 967 0.425 2
0 412 0 926 1.0'12 1.019 1.038 0 929 0413 0 73
.O.22 0 78 0 79 2.22 4 09 2 91 0 424 1.066 1.140 1.239 1.206 1.257 1.164 1.079 0427 3
0 420 1.055 1.135 1.238 1.203 1.242 1138 1.057 0420 0 95 1 04 0 44 0 06 0.25 1.21 2 28 2 08 1.67 0.426 0874 1.168 1.261 1.207 1.257 1.223 1.270 1.191 0 871 0426 4
0 420 0 864 1.180 1.266 1.215 1.264 1.219 1.269 1.182 0864 0 420 1 90 1.16 0 68 0 39 0 C6 0.55 0.33 0.08 0 76 0.8 i 1.43 0.425 1.088 1.197 1.021 1.186 1.017 1.056 1.030 1.184 1.011 1.186 1.070 0 418 5
0413 1.057 1.1 u l1.015 1.190 1.037 1077 1045 1196 1.015 1.180 1,055 0 412 2.91 2 93 1.2'i 0.59
-0 17
-1.93
-1.95
-1.44 1.00
-0.39 0.51 1.42 1.46 0 938 1.149 1.277 1.201 0.983 1205 1 227 1.207 0 969 1188 1.278 1.150 0 937 6
0.9?9 1 138 1.269 1 196 0 994 1.239 1 264 1.247 0 994 1.190 1.266 1.135 0 926 0 97 0 97 0 63 0 42
-1.11
-2 74
-2 93 3.21
-2 52
-0.17 0 95 1.32 1.19 0.422 1.034 123b 1.213 1.040 1.232 1.025 1 237 1022 1.200 1025 1.222 1.236 1.025 0.417
'7 0 422 1.038 1.242 1.219 1.045 1.24't 1.054 1 276 1.054 1.239 1.037 1215 1.238 1.032 0.419 0 00
-0.39
-0.56 0 49 0.48
-1.20 2.75 3.06
-3.04
-3.15
-1.16 0.58
-0 16
-0.68 0.48 0 395 1.015 1.199 1.247 1.054 1.235 1.241 0 940 1.252 1.241 1.0". i 1.269 1.200 1 010 0.392 g
0.395 1.019 1.203 1.264 1 077 1.264 1 276 0 976 1.276 1.264 1.07.
1.264 1.203 1 019 0 395 0.00
-0.39 0 33 1.34
-2.14 2.29
-2.74
-2.77
-1.88 1.82
-0.09 g 'O
-0.25
-088
-0.76 0.419 1.038 1.i:9 1.207 1.015 1.221 1.055 1.265 1.042 1.234 1.038 1.212 1.23a 1.034 0.422 9
0.419 1.032 1.238 1.215 1.037 1.239 1.054 1.276 1.054 1.247 1.045 1.219 1.242 1.038 0.422 0.00 0 58 0.89
-0.66 2.12
-1.45 0.09 0 86
-1.14
-1.04
-0.67 0.57
-0.24
-0.39 0.00 0.943 1.156 1.263 1.174 0.986 1.236 1.249 1.234 0.998 1.204 1.271 1.153 0 941 4*-
0.926 1.135 1.266 1.190 0.994 1.247 1.264 1.239 0.994 1.196 1.269 1.138 0 929 1.84 1.85
-0.24
-1,34
-0.80
-0.88
-1.19
-0.40 0 40 0 'i7 0.16 1.32 1.29 0.416 1.066 1,187 1.614 1.198 1.040 1.071 1.036 1.203 1.022 1.187 1.070 0.424 11 0.412 1.055 1.180 1.015 1.196 1.045 1.077 1.037 1.190 1.015 1.182 1057 0 413 0E7 1.04 0.59
-0.10 0.17
-0.48
-0.56
-0.10 1.09 0.69 0.42 1.23 2.66 0.420 0.864 1.180 1.272 1.212 1.256 1.221 1.288 1.197 0 869 0 423 12 0.420 0.864 1.182 1.269 1.219 1.264 1.215 1.266 1.180 0 864 0 420 0.00 0.00
-0.17 0.24
-0.57
-0.63 0.49 1.74 1.44 0.58 0.71 0.420 1.056 1.140 1.242 1.210 1.256 1.163 1075 0 423 0.420 1.057 1.138 1.242 1.203 1.238 1.135 1.055 0.420 13 0.00 0.09 0 18 0.00 0.58 1.45 2 47 1.90 0 71 0 412 0 936 1.64 1.029 1.047 0 949 0424 0 413 0 929 1.038 1.019 1.032 0 926 0 412 14
-024 0.75 0 58 0 98 1.45 2 48 2 91 0.429 0 402 0 427 INCORE Mean Abso'ute Difference = 0 0107 0 422 0.395 0.419 ANC 15 Standard Deviation = 0 0142 1.66 1.77 1.91
% DIFFERCNCE BURNUP = 1839 MWD /MTU POWER LEVEL = 100.0%
D B ANK AT 228 STEPS 62 l
l l
FIGURE 4.3-11 V.C. SUMMER NUCLEAR STATION CYCLE 3 RADIAL POWER DISTRIBUTION COMPARISON BETWE8iNINCORE AND ANC FOR MAP FCFM-03 014 R
P N
M L
K J
H G
F E
D C
B A
0.420 0 394 0.426 1
0 413 0 387 0 415 1.69 1 81 2.65 0.418 G.911 1.024 0.952 1.043 0.950 0 430 2
0 415 0.917 1017 0.974 1.022 0 919 0 416 0.72
-0.65 0.69 0 82 2.05 3.37 3.37 0429 1.070 1.160 1.199 1.143 1.218 1.209 1.100 0 432 3
0.426 1.063 1.168 1.198 1.139 1.201 1.170 1.065 0.427 0.70 0.66
-0.68 0 08 0.35 1 42 3.33 3.29 1.17 0.431 0 876 1.215 1.246 1.235 1.215 1.251 1.258 1.221 0 875 0433 4
0 427 0.870 1.211 1.258 1.241 1.220 1.245 1.260 1.214 0.870 0.426 0.94 0.69 0.33
-0.95
-048
-0 41 0 48
-0.16 0.58 0 57 1.64 0.422 1.080 1.223 1.029 1.228 1.014 1036 1.018 1.221 I.020 1.219 1.080 0 422 5
0.416 1.065 1.214 1.025 1.233 1.030 1.052 1.037 1.239 1.025 1.211 1.063 0.415 1.44 1.41 0.74 0.39
-0 41
-1.55 1.52
-1.83
-1.45
-0.49 0 '10 1.00 1.69 0.922 1.174 1.265 1.245 0.995 1.235 1.213 1.239 0.985 1.226 1.266 1.185 0.931 6
0.919 1.170 1.260 1.239 1.002 1.261 1.238 1.269 1.002 1.233 1.258 1.168 0.917 0.33 0.34 0.40 0 48
-0.70
-2.06 2 02
-2.36
-1.70
-0.57 0.64 1.46 1.53 0.412 1.018 1.196 1.241 1.034 1.259 1.029 1.272 1.030 1.238 1.012 1.230 1.193 1.014 0.412 7
0.415 1.022 1.201 1.245 1 037 1.269 1050 1.298 1.050 1.261 1.030 1.241 1.198 1.017 0.413
-0.72
-0.39
-0.42
-0.32
-0.29
-0.79
-2.00
-2.00
-1.90
-1.82
-1.75
-0.89
-042
-0.29
-0.24 0.388 0.971 1.138 1.207 1.034 1.217 1.276 0962 1.290 1.231 1.053 1.210 1.135 0.971 0 387 8
0 387 0 974
. 139 1.220 1.052 1238 1.298 0 976 1.298 1.238 1.052 1220 1.139 0 974 0.387 0.26
-0.31
-0.09
.l 07 1,71 1.70
-1.60 1.43
-0.62
-0.57 0.10
-0.32
-0.35 0.31 0.90 0 422 1.028 1.211 1.237 1.012 1.246 1.052 1.294 1.048 1.267 1.M6 1.236 1.188 1.014 0415 9
04 3 1 017 1.198 1.241 1.030 1.261 1.050 1.298 1.050 1.269
- .037 1.245 1.201 1.022 0 415 2.18 1 03 1.09 0 32 1.75
-1.19 0.19
-0 31
-0.19 0.16
-0.19
-0.72
-1.08
-0.78
-0.48 0 936 1.191 1.257 1.218 0.995 1 263 1.231 1.256 1.004 1.241 1.258 1.16I.
0 921 10 09K 1 168 1258 1.233 1.002 1.26J l238 1.261 1 002 1.239 1.260 1.170 0.919 2.07 1.97 0 Os 1.22, -0.70 0 47 0 57
-0 40 0.20 0 16
-0.16
-0.51 0.22 9.419 1.074 1.217 1 020 1.240 1.035 1050 1.018 1.237 1.027 1.217 1.067 0.417 11 0.415 1.063 1.211 1.025 1.239 l037 1052 1.030 1.233 1.025 1.214 1.065 0.416 0.96 1.03 0.50
-0.49 0.08
-0 19
-0.19
-1.17 0.32 0.20 0.25 0.19 0.24 0 426 0.867 1.208 1.261 1.242 1.217 1.250 1.277 1.227 0.876 0.431 12 0426 0.870 1.214 1.260 1.245 1.220 1.241 1 258 1.211 0 870 0 427 0.00
-0.34
-0 49 0.08
-0.24 0 25 0 73 1.51
!.32 0 69 0.94 0 426 1.061 1.171 1.207 1147 1.219 1.189 1.083 0430 0 427 1.065 1.170 1.201 1 139 1.198 1.168 1 063 0 426 13 0.23
-0.38 0 u9 0.50 0 70 1.75 1.80 1.88 0 94 O414 0.916 1.019 0.983 1.034 0.933 0.426 0 416 0 919 1022 0 974 1017 0.917 0 415 14
-0.48
-033
-029 0 92 1.67 1.74 2.65 0414 0.391 0.416 INCORF Mean Absolute Difference = 0.0088 0 415 0 387 0413 ANC 15 Standard Deviation = 0.0117
-0 24 1.03 0.73
"<c DIFFERENCE BURNUP = 3920 MWD /MTU POWER LEVEL = 99.0%
D B ANK AT 228 STEPS 63 l
- - _ - - _ _ _ - _ _ -. _ - - -. -. - - _ - - - - - - - ~ _ - - - - - - - - -
I FIGURE 4.3-12 V.C. SUMMER NUCLEAR STATION CYCLE 3 RADI AL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP-FCFM-03 017 R
P N
M L
K j
H G
F E
D C
B A
O420 0 394 0 423 1
0 411 0 386 0 414 2.19 2 07 2 17 0424 0.905 1.018 0.960 1.033 0.932 0 428 2
0.418 0 910 1.010 0 952 1.015 0.912 0 418 1.44
-0.55 0.79 0.84 1.77 2.19 2.39 0437 1.080 1.181 1.174 1.105 1.188 1.216 1.090 0.437 3
0.431 1.065 1.187 1175 1.104 1178 1.190 1.066 0 431 1.39 1 41
-0.51
-0.09 0 09 0.85 2 18 2 25 1.39 0.436 0877 1.232 1.242 1.254 1.192 1.263 1.245 1.232 0 876 0437 4
0.431 0.873 1.227 1.252 1.258 1.196 1.261 1.253 1.229 0.873 0 431 1.16 0.46 0 41
-0.80 0.32
-0.33 0.16
-064 0.24 0.34 1.39 0 421 1.073 1.233 1.031 1.256 1.015 1.027 1.020 1.254 1.028 1.232 1.080 0.424 5
0 418 1.066 1.229
.030 1.260 1.027 1040 1.033 1.265 1.030 1.227 1.065 0 418 0.72 0.66 0.33 0.10
-0.32 1.17
-1.25
-1.26 0.87 0 19 0 41 1.41 1.44 0.910 1.187 1.253 1.267 1.003 1.254 1.202 1.260 0.997 1.257 1.256 1200 0.919 6
0 912 1.190 1.253 1.265 1.007 1.271 1.220 1.279 1007 1.260 1.252 1.187 0.910
-0.22
-0.25 0,00 0.16
-040
-1.34
-1.48
-1.49
-0 99
-0 24 0.32 1.10 0.99 0.409 1.009 1.171 1.257 1.032 1.274 1.030 1.287 1.031 1.256 1015 1.258 1.169 1.005 0.408 7
0.414 1.015 1.178 1.261 1.033 1.279 1046 1.306 1.046 1.271 1.027 1.258 1.175 1.010 0.411
-1.21
-0.59
-0.59
-0.32
-0.10
-0.39
-1.53 1.45
-1.43
-1.18
-1.17 0.00
-0.51
-0.50 0.73 0.386 0.947 1.100 1.184 1.024 1.203 1.292 0.964 1.300 1.214 1.040 1196 1.099 0.948 0.384 g
0.386 0.952 1.104 1.196 1.040 1.220 1.306 0 974 1.306 1.220 1.040 L196 1.104 0.952 0.386 0.00
-0 53
-0.36
-1.00
-1.54 1.39
-1.07
-1.03 0.46
-0.49 0.00 0.00
-045 0.42
-0.52
- 0.417 1.018 1.184 1.253 1.011 1.2f 9 1.047 1.303 1.044 1.278 1.035 1.263 1,171 1.009 0.411 9
0.4 n 1 010 1.175 1.258 1.027 1.271 1.046 1.306 1.046 1.279 1.033 1.261 1.178 1.015 0 414 1.46 0 79 0.77 0.40
-1.56
-0.94 0.10
-0.23
-0.19
-0.08 0 19 0.16
-0.59
-0.59 0.72 0922 1.204 1.250 1.249 1.001 1.277 1.215 1.273 1.010 1.272 1.255 1.193 0 917 10 0 910 1.187 1.252 1.260 1,007 1.279 1.220 1.271 1007 1.265 1.253 1.190 0.912 1.32 1.43
-0.16
-0.87
-0.60
-0.16
-041 0.16 0.30 0.55 0 16 0.25 0.55 0.421 1.073 1.235 1.032 1.264 1.030 1.036 1.022 1.260 1.032 1.231 1.069 0.421 1,'
0418
'. 065 1.227 1.030 1.265 1.033 1 040 1,027 1.260 1.030 1.229 1.066 0.418 O.72 0 75 0.65 0.19
-0.08
-0.29 0 38
-0.49 0.00 0.19 0 16 0.28 0.72 0.432 0.874 1.232 1.251 1.235 1.191 1.268 1.265 1.237 0 874 0 432 12 0.431 0.873 1.229 1.253 1.261 1.196 1.258 1.252 1.227 0.873 0.431 0.23 0.1 :
0.24 0.16
-0.48
-042 0.79 1.04 0.81
-0.11 0.23 0.432 1.069 1.188 1.172 1.101 1.188 1.201 1.078 0 432 0.431 1.066 1.190 1.178 1.104 7.175 1.137 1.065 0 431 13 0.23 0.28
-0 17
-0.51
-0.27 1.11 1.18 1.22 0.23 0.420 0 914 1.018 0 932 1.020 0.919 0 427 0.418 0 912 1 015 0 952 1.010 0 910 0 418 14 0.48 0 22 0.30 0.00 0.99 0.99 2 15 0.415 0.386 0.411 INCORE Mean Absolute Difference = 0.0065 0 414 0 386 0 411 ANC 15 Standard Deviation = 0.0085 0.24 0.00 0.00
% DIFFERENCE BURNUP = 5558 MWD /MTU. POWER LEVEL = 99.8%
D B ANK AT 228 STEPS 64
l RGURE 4.3-13 V.C. 5UMMER NUCLEAR STATION CYCLE 3 RADIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 03-025 R
P N
M L
K J
H G
F E
D C
B A
O409 0390 0.426 1
0 415 0 390 0 417
-145 0.00 2.16 0428 0.889 0.999 0 919 1.020 0.920 0.433 2
0.424 0.899 1005 0.925 1.008 0.900 0 424 0.94
-1.11 0.60
-0.65 1.19 2.22 2 12 0.442 1.068 1.210 1.137 1.051 1.149 1.231 1.079 0 449 3
0.440 1.064 1.219 1.143 1.054 1.145 1220 1 064 0 440 0 45 0.38
-0 74 0 52
-0.28 0 35 0.90 1,41 2.05 0.444 0.878 1.248 1225 1.276 1 153 1.285 1.230 1.255 0883 0447 4
0 440 0.876 1.248 1.236 1.285 1.162 1.288 1.237 1.250 0 876 0.440 0.91 0.23 0.00 0.89
-0,70
-0.77
-0.23
-0.57 0 40 0.80 1.59 0,427 1.072 1.252 1.035 1.292 1.010 1.010 1.061 1.288 1.030 1.249 1.077 0 430 5
0 424 1.064 1.250 1.036 1.297 1.021 1.022 1.026 1.302 1036 1.248 1.064 0 424 0 71 0 75 0.1 fi
-0.10 0.39
-1.08
-1.17
-097
-1.08
-058 0 08 1 22 1.42 0 907 1.229 1.241 1.304 1.008 1.270 1173 1.273 0999 1.297 1.245 1.232 0907 5
0 900 1 220 1.237 1.302 1.013 1.285 1.187 1.291 1013 1.297 1236 1.219 0 899 0.78 U.74 0.32 0 15 0.49
-1.17
-1.18
-1.39
-1.38 0.00 0.73 1.07 0.89 0.426 1 021 1.148 1.292 1.029 1.290 1.025 1.299 1.022 1.268 1.019 1.296 1.144 1.000 0 412 7
0.417 1.008 1145 1.288 1.026 1.291 1.036 1 313 1036 1.285 1 021 1285 1.143 1.005 0.415 2 16 1.29 0.26 0.31 0.29
-0.08
-1.06
-1.07
-135
-1.32 0.20 0.86 0 09 0.50 0 72 0.399 0.93fi 1.060 1.156 1.013 1.177 1.303 of 8 1.307 1 181 1.028 1.169 1.056 0 923 0.390 8
0 390 0.925 1.054 1.162 1.022 1.187 1.313 0 968 1.313 1.187 1.022 1.162 1.054 0.925 0 390 2.31 1.19 0.57
-0.52
-0.88
-0 84 0.76
-1.03
-0.46
-0.51 0 59 0.60 0.19
-0 22 0.00 0 424 1.018 1.153 1.285 1.012 1.280 1.042 1312 1.033 1.289 ' l.025 1.282 1.141 1.009 0.418 9
0 415 1.005 1.143 1.285 1.021 1.285 1036 1.313 1.036 12wl 1.026 1.288 1.144 1.008 0417 2 17 1.29 0.87 0.00
-0.88
-0 39 0.58
-0.08
-029
-0.15
-0.10
-047 0.35 0.10 0.24 0.913 1.237 1.234 1.283 1.007 1.290 1.183 1.286 1.013 1.304 1.228 1222 0.909 10 0.899 1.219 1.236 1297 1.013 1.291 1.187 1.285 1.013 1.302 1.237 1.220 0 900 1.56 1.48
-0. l t,
-1.08
-0 59
-0 08 0 34 0 08 0.00 0.15
-0.73 0 16 1.00 0.426 1.071 1.249 1.024 1.300 1.025 1.020 1.022 1.299 1.034 1.244 1.066 0 427 11 0.424 1.064 1.248 1.030 1.302 1.026 1.022 1 021 1.297 1.036 1.250 1.064 0424 0.47 0.66 0.08 1.16
-0.15
-0.10
-0.20 0.10 0.15
-0.19
-0.48 0.19 0.71 0.439 0.871 1.237 1.231 1.284 1.157 1.289 1.242 1 252 0.872 0 441 12 0 440 0 876 1.250 1.237 1.288 1 162 1.285 1 236 1.248 0 876 0440
-0 57
-1 04 0 49 0.31
-0.43 0.31 0.49 0 32
-046 0.23
_-0,23 0 440 1064 1.200 1.144 1.052 1.145 1.224 1.071 0441 0 440 1.064 1.220 1.145 1.054 1.143 1.219 1.064 0 440 13 0 00 0.00 0 00
-009
-0.19 0.17 0 41 0.66 0 23 0424 0 911 1 020 0.929 1.005 0.902 0429 0 424 0 900 1.008 0 925 1.005 0 899 0 424 14 0 00 1.22 1.19 0 43 0.00 0 33 1.18 0.427 0 394 0 412 INCORE Mean Absolute Dif ference = 0 0060 0 417 0 390 0 415 ANC 15 Standard Deviation = 0 0076 2 40 1.03
-0.72
% DIFFERENCE BURNUP = 8689 MWD /MTU POWER LEVEL = 100.0%
D BANK AT 228 STEPS 65 1
1
FIGURE 4.3-14 V.C. 5UMMER NUCLEAR STATION CYCLE 3 RADI AL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP FCFM 03 031 R
P N
M L
K J
H G
F E
D C
B A
3 0433 0409 0434 1
0.424 0 401 0426 2.12 2 00 1.88 o439 0.888 1.014 0924 1.027 0914 0 440 2
u 431 0.898 1.010 0 920 1.013 0.899 0.431 1.86
-1.11 0 40 0.43 1.38 1 67 2.09 0.454 1.074 1.230 1.121 1.027 1.130 1.241 1.076 0.460 3
0 449 1.064 1.236 1.129 1.031 1.130 1.236 1 064 6449 1.11 0.94
-0 49
-071
-0 39 0.00 0 40 1.13 2 45 0.453 0 884 1.261 1.214 1.285 1.133 1.290 1.212 1.259 0886 0456 4
0 443 0 881 1.256 1223 1295 1.143 1 297 1.223 1.257 0881 0 449 0.89 0.34 0 -10
-0.74
-0.77
-0 87 0 54
-0.90 0.16 0.57 1 56 0432 1.066 1.260 1.038 1.308 1.003 1.000 1.007 1.299 1.027 1.253 1.075 0.439 5
0 431 1.064 1.257 1.036 1.310 1 016 1012 1.020 1.314 1.036 1.256 1.064 0 431 0.23 0.19 0.24 0 19
-0.15
-1.28
-1.19
-1.27
-1.14
-0.87
-0.24 1.03 1.86 0.893 1.229 1.221 1.319 1.009 1.272 1153 1.274 1.000 1.308 1.228 1.250 0 910 6
0.899 1.?36 1.223 1.314 1.012 1.284 1.165 1.289 1.012 1.310 1.223 1.236 0 898
-0.67
-0 57
-0.16 0.38
-0.30, 0 93 1.03
-1.16
-1.19
-0 15 0.41 1.13 1.34 0.439 1.021 1.124 1.297 1.022 1.289 l'.017 1.297 1.014 1.270 1.013 1.306 1.130
.006 0.422 7
0.426 1.013 1.130 1.297 1.020 1.289 1.026 1.307 1.026 1.284 1.016 1.295 1.129 1.010 0 424 3.05 0.79
-0.53 0.00 0 20 0.00
-0.88
-0.77
-1.17 1.09
-0.30 0.85 0.09
-0.40
-0.47 0.413 6.927 1.031 1.135 1.001 1.154 1.299 0.953 1.302 1.159 1.016 1.150 1.033 0.919 0.402 8
0.401 0.920 1.031 1.143 1.012 1.165 1.307 0 960 1.307 1.165 1.012 1.143 1.031 0.920 0 401 2.99 0.76 0.00
-0.70
-1,09
-0.94
-0.61
-0.73
-0.38
-0.52 0.40 0.61 0 19
-0.11 0.25 0.437 1.030 1.143 1.298 1.005 1.278 1.031 1.306 1.022 1.286 1.014 1.286 1.120 1.012 0 427
'424 1.010 1.129 1.295 1.016 1.284 1.026 1.307 1.026 1.289 1.020 1 297 1.130 1.013 0 426 g l 3.07 1.98 1.24 0.23
-1.08
-0.47 0.49
-0.08
-0.39
-0.23 0.59
-0.85
-0.88
-0.10 0.23 0.917 1.261 1.223 1.297 1.001 1.287 1.160 1.281 1.011 1.311 1.211 1.230 0.906 10 0.898 1.236 1.223 1.310 1.012 1.289 1.165 1.284 1.012 1.314 1.223 1.236 0.899 2.12 2.02 0.00
-0.09
-0.49
-0.16
-0.43
-0.23 0.10
-0.23
-0.98
-0,49 0.78 0 438 1.080 1.265 1.028 1.313 1.017 1,009 1.012 1.310 1.031 1.248 1.059 0.431 11 0.431 1.064 1.256 1.036 1.314 1.020 1.012 1.016 1.310 1.036 1.257 1.064 0.431 1.62 1.50 0.72
-0.77
-0.08
-0.29
-0.30
-0.39 0.00
-0 48
-0.72
-0.47 0.00 0.454 0.882 1.249 1.219 1.291 1.137 1.295 1.226 1.257 0.876 0.448 12 0.449 0.881 1.257 1.223 1.297 1143 1.295 1.223 1.256 0.881 0.449 1.11 0.11
-0.64
-0.33
-0.46
-0.52 0.00 0.25 0.08
-0.57
-0.22 0.452 1.067 1.238 1.127 1.031 1.131 1.243 1.067 0 448 0.449 1.064 1.236 1.130 1.031 1.129 1.236 1.064 0 449 13 0.67 0.28 0.16
-0.27 0.00 0.18 0.57 0.28
-0.22 0.432 0.912 1.026 0.928 1.013 0.903 0 435 0.431 0.899 1.013 0.920 1.010 0.898 0.431 14 0.23 1.45 1.28 0.87 0.30 0.56 0.93 0 438 0.407 0.425 INCORE Mean Absolute Dif ference = 0.0065 0.426 0.401 0 424 ANC 15 Standard Deviation = 0.0081 2.82 1.50 0.24
% DIFFERENCE BURNUP = 11123 MWD /MTU POWER LEVEL = 99.9%
D BANK AT 228 STEPS 66 l
FIGURE 4.3-15 V.C. SUMMER NUCLEAR STATION CYCLE 3 RADIAL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP-FCFM 03-037 R
P N
M L
K J
H G
F E
D C
B A
0454 0 433 0.452 1
0 440 0 420 0 442 3.18 3 10 2 26 0 450 0 889 1030 0935 1.043 0 915 0449 2
0.441 0.902 1.025 0.930 1.028 0.903 0 441 2.04 1.44 0.49 0 54 1.46 1.33 1.81 0 463 1 069 1.227 1.108 1.012 1.114 1.241 1.072 0 471 3
0.461 1.065 1.244 1.122 1.022 1.123 1.245 1.065 0 461 0 43 0.38
-1.37
-1.25 0.98
-0.80
-032 0.66 2 17 0 464 0687 1.255 1.196 1.283 1.118 1.279 1.188 1.254 0 891 0 467 4
0.461 0.886 1.255 1.207 1.296 1.131 1.297 1.207 1.256 0.886 0 461 0.65 0.11 0.00
-0 91
-1.00
-1.15
-1.39
-1.57
-0.16 0.56 1.30 0.440 1.064 1.254 1.030 1.307 0.996 U.991 0.996 1.296 1.023 1.250 1.077 0.450 5
0.441 1065 1.256 LO32-1.309 1.010 1.006 1013 1.3 n 1.032 1.255 1.065 0 441
-0.23
-0.09
-0 16
-0.19
-015
-139
-1.49
-1.68
-1 14
-0.87 O 40 1.13 2 04 0.894 1.233 1204 1319 1.008 1.267 1.138 1.273 1.002 1.311 1.212 1.262 0.917 6
0.903 1.245 1.207 1.311 1008 1.276 1.145 1.280 1.008 1.309 1.207 1.244 0.902 l
-1.00
-0.96 0.25 0.61 0 00
-0.71
-0.61
-0.55 0.60 0.15 0.41 1.45 1.66 0.456 1.030 1.109 1.293 1.017 1.282 1.005 1.2bd 1.009 1.273 1.014 1.305 1.122 1.019 0.438 7-0.442 1.028 1.123 1.297 1 013 1.280 1.015 1.292 1.015 1.276 1.010 1.296 1.122 1.025 0.440 3.17 0.19
-1.25
-031 0.39 0.16 0 99
-0.46
-0.59
-0.24 0 40 0.69 0.00
-0.59
-0 45 0.433 0 933 1.014 1.120 0.997 1.136 1.287 0 950 1.295 1.146 1.016 1.137 1.002 0 927 0.421 g
0.420 0.930 1.022 1 131 1.006 1.145 1.292 0 952 1.292 1145 1.006 1 131 1.002 0 930 0 420 3.10 0.32
-0.78 0 97
-0.89
-O 79
-0.39
-0 21 0.23 0 09 0.99 0.53 0.00
-0.32 0.24 0.454 1.042 1.131 1.297 1.000 1.269 1.015 1.292 1.016 1.284 1.009 1.278 1.108 1.024 0 442 g
0.440 1.025 1.122 1.29t>
1.010 1.276 1.015 1.292 1.015 1.280 1.013 1.297 1.123 1.028 0 442 3.18 1.66 0.80 0.08
-0 99
-0 55 0 00 0 00 0.10 0.31
-0 39
-1.46
-1.34
-0.39 0.00 0.922 1.271 1.208 1,296 1.001 1.282 1.145 1.281 1.011 1.310 1.192 1.234 0.909 10 0 902 1.244 1.207 1.309 1.008 1.280 1.145 1.276 1.008 1.311 1.207 1.245 0 903 2.22 2.17 0 08 0.99
-0.69 0 16 0.00 0.39 0.30
-008
-1.24
-0.88 0.66 0.449 1.084 1.266 1.023 1.308 1.013 1.006 1.013 1.311 1.028 1.247 1.059 0.441 11 0.441 1.065 1.255 1.032 1.311 1.013 1.006 1.010 1.309 1.032 1.256 1.065 0.441 1,81 1.78 0.88 0.81
-0.23 0 00 0 00 0.30 0.15
-0.39
-0 72 0.56 0.00 0.467 0889 1.246 1.201 1.292 1.125 1.300 1.216 1.261 0.885 0.461 0.461 0.886 1.256 1.207 1.297 1.131 1.296 1.207 1,255 0 886 0.461 1.30 0 34
-0 80 0 50
-0 39
-0.53 0.31 0.75 0 48 0.11 0.00 0.465 1.072 1.249 1.114 1.018 1.124 1.258 1.075 0461 0.461 1.065 1.245 1.123 1.022 1.122 1.244 1.065 0.461 13 0 87 0.66 0 32
-0.80 0 39 0.18 1.13 0 94 0.00 g 0444 0917 1.033 0.933 1.027 0.911 0 449 0.441 0.903 1028 0 930 1.025 0 902 0 441 14 0 68 1.55 0 49 0 32 0.20 1 00 1.81 0453 0.426 0 441 INCORE Mean Absolute Difference = 0.0071 0 442 0 420 0.440 ANC 15 Standard Deviation = 0.0089 2 49 1.43 0 23
% DIFFERENCE BURNUP = 13581 MWD /MTU POWER LEVEL = 100.0%
D BANK AT 228 STEPS 67
FIGURE 4.3-16 V.C. SUMMER NUCLEAR STATION CYCLE 4 RADI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 04 005 R
P N
M L
K J
H G
F E
D C
B A
U.391 0.384 0 394 1
0 393 a 386 0 394
-051
-0.52 0 00 0 442 0741 1.056 0 972 1.059 0743 0 441 2
0 440 0 739 1.052 0969 1.054 0 739 0 440 0 45 0 27 0.38 0.31 0 47 0 54 0.23 U.463 0.984 1204 1264 1.222 1.260 1.198 0980 0 463 3
0 aci 0 979 1.202 1.265 1.225 1263 1.201 0 982 0464 0.43 0 51 0.17
-0 08
-0.24 0.24 0 25
-0 20 0.22 0.472 0.735 1194 1.220 1.230 1 050 1.224 1.215 1.197 0.731 0.461 4
0 464 0.729 1.196 1.228 1.238 1057 L233 1 224 1197 0 729 0 461 1.72 0.82 0.17
-065
-065
-0 66 0 73 0 74 0 00 0.27 0,00 0.453 1 011 1.210 0.926 1.258 1 270 1.178 1.274 1.256 0 926 1.204 0 981 0 434 5
0 440 0 982 L197 0 923 1.265 1.292 1,198 1.291 1.264 0 923 1 196 0 979 0.440 2.95 2.95 1.09 0 33
-0 55
-I 70 1 67
-1.32
-0.63 0 33 0 67 0 20
-1.36 0.749 1.216 1.232 1,264 1.161 1.193 1.200 1.198 1.157 1.271 1.239 1218 0.748 g
0739 1.201 1.224 1.264 1.171 1.215 1.224 1.217 1.171 1.265 1.228 1.202 0739 1.35 1.25 0 65 0.00
-0 85
-1.81 1.96
-1.56
-1.20 0.47 0.90 1.33 1 22 0 389 1.054 1.2'i7 1.227 1.277 1.199 0 977 1.182 0 978 1.194 1.290 1.256 1 287 1.073 0.401 7
0.394 1.054 1.263 1.233 1.291 1.217 0995 1.208 0 995 1.215 1.292 1238 1.265 1052 0.393
-1.27 0 00 0.32
-0.49
-1.08 1.48 1.81 2.15
-1.71 1.73
-0.15 1.45 1.74 2 00 2.04 0 381 0.968 1.228 1,050 1.178 1 202 1.183 u925 1.193 1.210 1.206 1.073 1 243 0 985 0 393 g
0.386 0.969 1.225 1.057 1.198 L224 1.208 0948 1.208 1.224 1.198 1057 1.225 0 969 0 386
-1.30
-0.10 0.24
-0.66 1 67
-1.80
-2 07 2 43
-1 24
-114 0.67 1.51 1.-17 1.65 1.81 0 389 1.059 1.284 1.236 1.273 1.LO7 1.011 1.202 0.988 1.210 1.283 1.225 1.264 1.067 0398 g
0.393 1.052 1.265 1.238 1.292 1.215 0.995 1.208 0 995 1.217 1.291 1.233 1.263 1.054 0 394 1.02 0.67 1.50 0 16
-1 47
-066 1.61
-0.50
-0.70
-058
-062 0 65 0.08 1.23 1.02 0 753 1.223 1.233 1.261 1.175 1.211 1.208 1.210 1.173 1.265 1.200 1.192 0.750 10 0.739 1.202 L228 1.265 1.171 1.217 1.224 1.215 1.171 1.264 L224 1.201 0.739 1.89 1.75 0 41
-0 32 0 34
-049
-1.31 0 41 0.17 0 08
-l96
-0.75 1.49 0.444 0.990 1.204 0 924 1.271 1.286 1.192 1.287 1.271 0 921 1.182 0 975 0.440 11 0.440 0.979 1196 0.923 1.264 1.291 1198 1.292 1.265 0 923 1197 0982 0 440 0.91 1.12 0.67 0.11 0 55
-039
-0 50 0.39 0.47
-0.22
-1.25
-0.71 0.00 0 462 0729 1.195 1.224 1.224 1 049 1.243 1.241 1.204 0 723 0 460 12 0 461 0 729 l 197 1 224 1 233 1 "57 1 238 3 228 Ll96 0 729 U 464 0.22 0 00
-0.17 0 90 0 73 0 76 0 40 1.06 0.67
-082 0 86 0464 0 983 1.203 1.270 1239 1.290 1 232 0988 0457 0464 0 982 1.201 1.263 1.225 1.265 1.202 0.979 0 461 13 0.00 0.10 0 17 0 55 1.14 1.98 2 50 0.98
-o 87 0440 0 740 1060 0.982 1 075 0759 0.449 0440 0.739 1.054 0969 1.052 0739 0 440 14 0 00 0 14 9 57 1.34 2 19 2.71 2 05 0 395 0392 0 406 INCORE Mean Absolute Difference = 0 0088 0.394 0 386 0.393 ANC 15 Standard Deviation = 0 0115 0 25 1 55 3.31
% DIFFERENCE BURNUP = 1232 MWD /MTU POWER LEVEL = 99.8%
D B ANK AT 228 STEPS 68
l FIGURE 4.3 17 V.C. 50MMER NUCLEA R STATION CYCLE 4 RADI AL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP-FCfM 04 010 R
P N
M L
K J
H G
F E
D C
B A
0 3t(4 0378 0388 1
0 386 0381 0 387 0 52 0 79 0.26 0.445 0.724 1.020 0 936 1 031 0732 0 445 2
0 441 0 723 1.024 0 936 1.026 0 723 0 441 0.91 0 14 0.20 0 00 0 49 1.24 0.91 0468 0.996 1207 1.219 1.240 1.216 1.205 0.991 0 470 3
0 464 0 986 1.206 1.224 1246 1.222 1.205 0.988 0467 0.86 1.01 0 08
-0 41
-0 43 0 33 0 00 0.30 0 64 0472 0.730 1.201 1.196 1265 1.033 1.255 1185 1.202 0 731 0 4ti7 4
0.467 0 726 1.199 1.201 1272 1.040 1267 1.198 1.200 0 726 0464 1.07 0 55 0.17
-0 42 0 55
-0.67 0.95 1.09 0 17 0 69 0 65 0.447 1.002 1.206 0915 1.297 1.250 1.151 1 252 1.295 0 918 1.208 0 992 0,438 5
0 441 0 988 1.200 0.913 1 301 1.266 1167 1.265 1.300 0 913 1.199 0 986 0 441 1.36 1 42 0 50 0.22
-0.31
-1.26
-1.37
-1.03 0 38 0 55 0.75 0.61 0 68 0.726 1.210 1.199 1.299 1 162 1.257 1196 1.262 1.162 1.310 1210 1.222 0.732 6
0 723 L205 L198 L300 Llio L272 1 214 1 274 1170 1.301 1.201 L206 0 723 0.41 0.41 0 08
-0.08
-0.68 1.18
-1.48 0 94 0 68 0.69 0.75 1.33 1.24 U.384 1.023 1.220 1.26u 1.254 1.259 0999 1.247 1.002 1 259 1.268 1286 1.242 1.043 0 393 7
0.387 1 026 1.222 1.267 1.265 L274 1.015 1.270 1015 1 272 1 266 1272 1224 1.024 0.386
-0 78
-0 29
-0 16
-0 55 0 87
-1.18 1.58
-1 81 1.28
-1.02 0 16 1 10 1 47 1.86 1.81 0.377 0933 1.244 1.033 1.153 1.198 1.248 0952 1258 1.203 1.174 1.051 1.263 0 952 0.388 g
0.381 0.936 1.246 1.040 1167 1.214 1270 0 974 1270 1.211 1.167 1.040 1.246 0 936 0 381
-1.05
-0.32 0 16
-067
-1.20
-1.32 1.73
-2.26
-0.94 0 91 0.60 1.06 1.36 1.71 1 84 0 383 1.027 1.232 1.269 1.253 1.269 1.031 1.266 1009 1.271 1259 1.257 1.221 1.038 0 391 g
0386 1024 1.224 1.272 1.266 1.272 1.015 1 270 1.015 1.274 1.265 1 267 1.222 1026 0 387 0.78 0.29 0.35
-0.24
-1.03
-0.24 1.58 0 31 0 59 0.24 0 47 0 79
-0.08 1.17 1.03 0730 1.216 1.204 1.299 1.176 1 275 1 205 1270 1.173 1 301 I176 1.191 0.732 10 0.723 1 206 1201 L301 Ll70 1274 L214 1.272 1.170 1.300 1 198 1.205 0 723 0.97 1.00 0 25
-0 15 0 51 0.08
-0 74 0 16 0.26 0 08 l.84
-1.16 1.24 0 444 0 993 1 204 0 915 1.307 1 264 1164 1,262 1.306 v910 1185 0 978 0 439 11 0.441 0 986 L199 0 913 L300 L265 1167 1 266 L301 0 913 1.200 0 988 0.441 0.68 0 71 0 42 0.22 0 54
-008 0 26 0.32 0.38
-0 33
-1.25
-1.01 0 45 0.466 0 727 1.199 1196 1.260 1 03..
1.276 1.211 1.204 0720 0462 12 0 464 0726 1.200 1 198 1.267 1040 1 272 1.201 1.199 0726 0467 0 43 0.14 0 08
-0 17 0 55
-058 0 31 0.83 0 42
-083
-1.07 0 470 0 997 1,211 1 224 1255 1 241 1.230 0 991 0460 0 467 0988 1.205 1.222 1246 1.204 1206 0986 0464 13 0 64 0.91 0.50 0 16 0 72 1.39 1.99 0 51
-0 86 0445 0 728 1.031 0 946 1.040 0 737 0 446 0 441 0723 1.026 0 936 1 024 0723 0 441 14 0 91 0 69 0 49 1.07 1.56 1.94 159 g 0389 0.366 0.395 (NCORE Mean Absolute Difference = 0 0073 0 387 0381 0.386 ANC 15 Standard Deviation = 0.0092 0 52 1 31 2.33
% DIFFERENCE BURNUP = 4154 MWD /MTU POWER LEVEL = 100.0%
D B ANK AT 228 STEPS 69
FIGURE 4 3-18 V.C. SUMMER NUCLEAR STATION CYCLE 4 RADIAL POWER DISTRIBbTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 04 012 R
P N
M L
K J
H G
F E
D C
B A
0.385 0 381 0 389 1
0 386 0 383 0 387 0.26 0 52 0.52 0446 0 717 1.013 0925 1.018 0724 0 448 2
0 443 0 716 1 008 0 922 1.010 0.716 0.443 0.68 0.14 0.50 0 33 0.79 1.12 1.13 0.*71 0 994 1 205 1.196 1.255 1.195 1.203 0993 0.475 l 3
0 46.,
0.987 1.203 1.20 t 1 259 1.199 1.203 0 989 0 471 2
0 64 0.71 0.17
-0 42
-0 32
-033 0.00 0 40 0 85 0.474 0 729 1.200 1.179 1.282 1.026 1.275 1.172 1.200 0 732 0 472 4
0 471 0 726 1.194 1183 1.291 1.034 1.286 1.180 1195 0 726 0 468 0.64 0 41 0.50 0.34 0.70
-0.77
-0.86
-0 68 0.42 0 83 0 85 0 447 0.998 1.202 0 911 1.304 1.233 1.134 1.237 1.314 0.911 1.205 0 995 0442 5
0 443 0989 U 95 0 906 L307 1.251 1.151 1250 1.316 0.906 1.194 0 987 0.443 0.90 0.91 0 59 0.55
-0 23
-1.44 1.48
-104
-0 15 0 55 0 92 0 81 0.23 0 717 1.205 1.182 1.320 1.162 1.291 1.194 1.297 1.163 1 326 1.192 1.218 0 724 6
0.716 1.203 L180 L316 Ll68 1.306 1.211 1308 1168 1.317 1.183 1.203 0.716 0.14 0.17 0.17 0.30 0.51
-1.15
.l.40
-0.84
-0.43 0.68 0.76 1.25 1.12 l
6.384 1.009 1.201 1.280 1.236 1.291 1.015 1.288 1.018 1.294 1.252 1.304 1.215 1.023 0.392 7
0387 1.010 U 99 L286 1250 1.308 LO28 1.307 1028 1.300 1.251 1.291 1.201 1,008 0.386
-0,78
-0.10 0 17 0.47
-1.12
.l.30 1.26
-1.45
-0 97
-0.92 0.08 1 01 1.17 1.49 1.55 0.370 0.920 1.262 1,027 1.136 1.193 1.287 0 975 1.298 1 202 1 156 1.044 1.274 0 934 0 588 g
0.383 0.922 1.259 1 034 1.151 1.211 1.307 0.992 1.307 1.211 1.151 1.034 1.259 0.922 0383 1.04 0.22 0.24
-0.68 1.30
-1.49 1.53
-1.71
-0.69
-0 74 0 43 0.97 1.19 1,30 1.31 0.383 1.010 1.208 1.2M6 1237 1.304 1.049 1.313 1.025 1.306 1.239 1.272 1.195 1.020 0.390 g
0.386 1.008 1.201 1.291 1.251 1.306 1.028 1.307 1.028 1,308 1.250 1286 1.199 1.010 0.387 l
0.78 0.20 0 58
-0.39
-1.12
-0.15 2 04 0 46
-0,29 0 15 0 88
-1.09
-0.33 0.99 0.78 l
l 0.718 1208 1.182 1.313 1.175 1.311 1.206 1303 1.171 1.311 1.154 1.186 0.723 10-0.716 1.203 1.183 1.317 1.168 1.308 1.211 L306 1.168 1316 L180 L203 0.716 0.28 U.42
-0 08
-0.30 0.60 0.23 0 41
-0,23 0 26
-0 38
-2 20 1.41 0 98 0.446 0.994 1.197 0.903 1325 1.252 1.151 1.244 1.319 0.902 1.182 0981 0.444 11-0 443 0 987 1.194 0.906 1.316 1.250 1.151 1.251 1.317 0 906 1195 0 989 0.44'l 0.68 0.71 0.25
-0 33 0 318 0 I fi 0 00
-0.56 0.15
-0.44
-1.09
-0.81 0.23 0.472 0.727 1.189 1.179 1.282 1.029 1.290 1.186 1.196 0.722 0.469 12 0 468 0 726 1.195 1.180 1.286 1 034 1.291 1.183 1.194 0 726 0 471 0.85 0.14
-0.50
-0 06 0.31
-0.48
-0 08 0 25 0.17
-0.55 0 42 0475 0999 1.212 1.205 1.271 1.213 1.219 0.989 0.466 1
0 471 0989 1.203 1.199 1259 1.201 1.203 0 987 0 468 13 0.85 1 01 0.75 0.50 0.95 1.00 1.33 0.20
-0 43 l
l 0448 0.723 1.018 0933
- .020 0 726 0.446 1
0443 0 716 LO10 0 922 1.008 0.716 0 443 1.13 0 98 0.79 1.19 1.19 1.40 0.68 0 391 0.389 0 395 INCORE Mean Absolute Difference = 0 0069 0387 0 383 0.386 ANC 15 Standard Deviation = 0 0086 1.J3 1 $7 2.33
% DIFFERENCE BURNUP = 6398 MWD /MTU POWER LEVEL = 99.8%
D BANK AT 228 STEPS 70
FIGURE 4.3-19 V,C. SUMMER NUCLEAR STATION CYCLE 4 RADIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 04-016 R
P N
M L
K J
H G
F E
D C
B A
0390 0 388 0.393 1
0.390 0 389 0.391 0.00
-0.26 0 51 0452 0710 1 002 0918 1.010 0 724 0455 2
0449 0713 0 998 0 915 1.000 0 713 0449 0 67
-042 0 40 0.33 1 00 1.54 1 34 0479 0999 1.195 1.175 1.264 1.182 1219 1.008 0483 3
0475 0 990 1.201 1 183 1.271 1.I81 1.200 0 992 0477 0 84 0 91 0 50
-0 68
-055 0 08 1.58 1 61 1 26 0 482 0 732 1.194 1.159 1.292 1 019 1.289 1.159 1.196 0 731 0 478 4
0.477 0728 1.190 1.167 1.303 1.029 1.299 1164 1.192 0728 0475 1.05 0 55 0 34
-0 69 O 84 0 97
.0 77
-043 0 34 0 41 0.63 0 453 1.003 1.199 0 905 1.323 1 215 1.118 1.218 1 321 0 904 1 895 0 996 0.449 I
5 0449 0 992 1.192 0 90:
1325 1 236 1.138 1.235 1.325 Os01 1 190 Om 0 449 1
0.89 1.11 0 59 0 44
-0 15 1 70
-l76 1 38 0.30 0 33 0 42 0 61 0 00 0 716 1.205 1.168 1330 1.156 1.311 1184 1 313 1.151 1.327 1.170 1.217 0 724 6
0 713 1.200 1.164 1 325 1.163 1328 1.203 1.330 1 163 1.325 1167 1.201 0.713 0 42 0 42 0,34 0 38 0 60
-1.28
-1.58
-1.28
-1.03 0.15 0 26 1 33 1.54 0.389 1.005 1.186 1298 1.225 1.315 1.021 1.309 1.021 1 310 1227 1322 1.197 1.011 0 394 7
0.391 1 000 1 181 1.299 1.235 1.330 1 035 1.332 1.035 1328 1.236 1303 1.183 0 998 0 390 0.51 0 50 0 42
-008
-0 81
-113
.I 35
-1.73
-1 35
-1.36
-0 73 1 46 1.18 1.30 1 03 0.394 0920 1.278 1.024 1.122 1.185 1,308 0 982 1.321 1.193 1.143 1.043 1 286 0.925 0392 g
0389 0915 1271 1.029 1138 1.203 1 332 1003 1332 1.203 1.138 1 029 1.271 0915 0389 1.29 0 55 0.55
-049 1.41
-1.50
-l 80 2 09
-0 83
-0.83 0 44 1.36 1.18 1 09 0.77 0.397 1.013 1.200 1.304 1.221 1328 1 060 1.338 1.032 1329 1.226 1.28(i 1.171 1003 0 388 9
0 390 0 998 1183 1.303 1 236 1.328 1 035 1 332 1.035 1 330 1 235 1 299 1 181 1.000 0 391 1.79 1.50 1.44 0.08
-1.21 0.00 2 42 0.45
-029 0 08 0 73
-1.00
-0 85 0 30
-0.77 0 724 1 219 1.169 1.320 1.170 1.333 1197 1.319 1 161 1314 1135 1.161 0 689 10 0.713 1 201 1.167 1.325 1.163 1.330 1.203 1.328 1.163 1.325 1.164 1.200 0 713 1.54 1.50 0.17
-0.38 0 60 0.23
-0.50
-0 68 0 17
-0.83
-2 49 3.25 3 37 0.455 1.005 1.199 0.895 1.336 1.238 1.138 1.204 1.307 0.888 1.17(i 0980 0 448 gj 0.449 0 990 1.190 0.901 1.325 1.235 1.138 1.236 1.325 0.901 1 192 0 992 0 449 1.34 1.52 0 76 0 67 0.83 0.24 0 00
-2 59
-1.36
-1.44
-1.34 1,21
-0.22 0.481 0.730 1.182 1.164 1.291 1.020 1296 1.168 1.189 0724 0 47(i gg 0475 0 728 1.192 1.164 1.299 1.029 1.303 1167 1.190 0 728 0 477 1.26 0.27
-084 0.00
-0 62 0 87
-0 54 0.09
-008
-0.55
-0 21 0.483 1.004 1.213 1.192 1.293 1.20fi 1.224 0996 0473 0 477 0.992 1.200 1.181 1.271 1.183 1.201 0990 0 475 13 1.26 1.21 1.08 0 93 1.73 r 94 1.92 0.61
-042
- p. ___
0454 0.721 1.011 0 938 10 9 1728 0 456 0 449 0.713 1.000 0.915 9.a98 0.713 0 449 14 1.11 1.12 1.10 2 51 2.10 2.10 1 56
~
0.395 0 398 0 400 INCORE Mean Absolute Difference = 0.0093 0 391 0.389 0 390 ANC 15 Standard Deviation = 0.0118 1.02 2.31 2 56
% DIFFERENCE BURNUP = 8608 MWD /MTU POWER LEVEL = 1001%
D B ANK AT 228 STEPS 71
FIGURE 4.3-20 V.C. SUMMER NUCLEAR STATION CYCLE 4 RADI AL POWER DISTRIBUTION COMP ARISON BETWEENINCORE AND ANC FOR MAP FCFM-04 021 R
P N
M L
K j
H G
F E
D C
B A
0409 0 411 0 413 1
0 407 0 409 0 40s 0.49 0 49 1.23 0475 0720 1.008 0 928 1 018 0.739 0476 2
0465 0 722 1.000 0 921 1.002 0723 0 465 2 15
-028 0.80 0 76 1 60 2.21 2 37 0504 1.026 1.198 1.161 1.283 1.170 1.230 1031 0 501 0 494 1.005 1 200 1.165 1.285 1.164 1.202 1 007 0496 3
2 02 2 09
-0 33 0 34
-0 16 0.52 2.33 2 38 1 01 0.504 0 750 1.199 1.135 1.296 1.013 1.298 1.134 1.193 0 745 0 500 4
0.496 0.745 1.191 1 147 1306 1.022 1.303 1.145 1.193 0 745 0.494 1.61 0A7 0 67
-1.05
-0.77 0 88
-0 38
-0.96 0 00 0.00 1.21
~
0.470 1.018 1.197 0 898 1.311 1.181 1.091 1.184 a.307 0 896 1193 1.018 0 472 5
0.465 1.007 1.193 0.897 1.319 1.206 1114 1.205 1319 0 897 1.191 1.005 0 465 1.08 1.09 0.34 0.11
-0El 2 07
-2 06 1.74
-0.91
-0.11 0 17 1.29 1.51 0.725 1.207 1.147 1.320 1.132 1.306 1.156 1.309 1 126 1.315 1.147 1.226 0.740 6
0.72s 1.202 1.145 1.319 1.142 1328 1.179 1.330 1.142 1319 1.147 1.202 0.722 0.28 0 42 0.17 0.08
-0E8
-1.66 1 95
-1.58 1.40
-0.30 d.00 2.00 2 49 0.407 1.009 1.171 1.299 1.169 1.311 1.013 1.312 1014 1.309 1.194 1.324 1.182 1.017 0.413 7
0.408 1.002 1.164 1.303 1 205 1.330 1.031 1337 1.031 1.328 1.206 1306 1165 1000 0 407
-0.25 0.70 0 60
-0.31
-1.33
-1.43
-1.75
-1.87
-1.65
-143
-1.00 1,38 1.46 1.70 1.47 0.415 0.927 1.295 1.015 1.092 1.156 1 313 0.982 1.325 1,168 1.118 1.035 1.304 0 933 0.415 8
0.409 0.921 1.285 1.022 1.114 1.179 1.337 1 004 1.337 1.179 1.114 1.022 1 285 0 921 0 409 1.47 0 65 0.78
-0 68
-1.97 l 95 1.80
-2 19
-0 90 0.93 0.36 1.27 1 48 1.30 1.47 0.414 1.018 1.184 1.305 1184 1.323 1.054 1.341 1.026 1.328 1.194 1.287 1.153 1.007 0.406 9
0.407 1.000 1.165 1.306 1.206 1.328 1 031 1337 1.03!
1.330 1.205 1.303 1.164 1.002 0 408 1.72 1.60 1 63 0 08
-1.82 4.38 2 23 0.30 i.0 48
-0.15 0.91
-1.23 0 95 0.50 0.49 0 735 1.224 1 141 1.298 1140 1.331 1.159 1.316 1.138 1.304 1.113 1157 0.695 10 0.722 1.2G2 1.147 1.319 1.142 1 330 1.179 1.328 1142 1.319 1145 1.202 0723 1.80 1.83
-052 1.59 0 18 0 08
-085
-0.90 0 35
-1.14
-2 79
-3 74
-3 87 0 474 1.024 1 195 0 874 1.322 1.208 1.113 1 971 1.300 0 883 1180 0 999 0.470 11 0.465 1.005 1191 0.897 1319 1.205 1 114 1.206 1.319 0.897 1.193 1007 0 465 1.94 1 89 0 34
-2 56 0.23 0 25 0 09
-2.90
- 1. 4 1.56 1.09 0 79 1.08 0 503 0742 1.160 1.13tl 1.295 1 013 1.295 1.145 1.191 0 744 0 499 12-0.494 0 745 1.19'1 11,5 1.303 1 022 1306 1.147 1.191 0 745 0496 1,82
-040
-2 77
-079 0 61
-088
-0.84
-0.17 0.00
-0 13 0 60 0.507 1 031 1.223 1.178 1 314 1.190 1.227 1 012 0497 0 496 1.007 1.202 1164 1 285 1.165 1 202 1.005 0 494 13 2.22 2 as I 75 1 20 2.26 2.15 2 08 0.70 0 61 0476 0739 1 U25 0950 1.024 0738 0 471 0465 0 723 1 002 0921 1 000 0 722 0465 14 2 37 2.21 2 30 3.15 2 40 2.22
- .29 hjh Ojff 0jfh INC RE Mean Absolute Difference : 0 0119 15 standard Deviation = 0 0148 2 21 3.18 3 19
% DIFFERENCE BURNUP = 12192 MWD /MTU POWER LEVEL = 99.9%
D BANK AT 226 STEPS 72 1
FIGURE 4 3-21 V.C. SUMMER NUCLEAR STATION CYCLE 4 RADI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 04 026 R
P N
M L
K J
H G
F E
D C
B A
h434 0439 0436 1
0 428 0 434 0 429 1 40 1 15 1 63 h496 0 733 1.021 0943 1.033 0 755 0 495 2
0484 0 738 1014 0 937 1015 0738 0 484 2 48
-0 68 0 69 0 ti4 1,77 2.30 2 27 052e 1.048 1.194 1.150 1.280 1.161 1 231 1 046 n 518 3
0515 1022 1 203 1157 1.287 1157 1.203 1024 0 518 2.52 2 54
-075
-0 61
-054 0 35 2.33 2.34 0 00 0.528 0 776 1.210 1.125 1 280 1.006 1.283 1.133 1.196 0765 0522 4
1.518 0 766 1.195 1.135 1 293 1 017 1.291 1.134 1.196 0 766 0 515 1.93 1.31 1.26 0,88 1 01 1 08
-062
-0 09 0 00
-013 1.36 0.492 1.041 1.204 0.901 1.293 1.154 1.072 1.156 1.293 0897 1.195 1.035 0496 5
0 484 1.024 1.196 0 899 1.301 1.182 1098 1 181 1 301 0 899 1.195 1022 0484 1.65 1 66 0 67 0 22 0 61
-2 37 2 37 2.12 O 61
-0.22 0.00 1 27 2.48 0.743 1.210 1.138 1.303 1.112 1.279 1.130 1.281 1.108 1.296 1.135 1.2?8 0.761 g
0.738 1.203 1.134 1,301 1.122 l_305 1155 1.307 1.122 1.301 1.135 1203 0 738 0.68 0.58 0 35 0.15
-089
-l99 2 16
-1.99
-1.25
-0.38 0.00 2 08 3.12 1.4 !J 1.022 1.163 1.286 1.162 1.286 0.998 1 288 1.000 1.285 1.168 1.300 1.171 1.027 0.435 7
tc 29 1.015 1.157 1.291 1.181 1.307 1019 I317 1019 1.305 1.182 1.293 1.157 1.014 0 428 0 47 0.69 0 52
-0.39
-1.61
-1.61 2.06
-2 20
-1.86
-1.53
-1.18 0.54 1.21 1.28 1.64 0 441 0.9 h3 1.296 1.011 1.077 1.134 1.294 0 973 1.304 1.144 1.102 1.022 1.300 0 946 0.439 8
0.434 0.937 1.087 1.017 1.098 1.155 1.317 0 995 1.317 1.155 1.098 1.017 1.287 0 937 0 434 1.61 0.64 0.70
-0 59
-1.91
-1.82
-1.75
-2.21 0.99
-095 0.36 0.49 1 01 0.96 1.15 0.435 1.031 1.177 1.093 1.162 1.299 1.038 1.315 1.013 1.303 1171 1.270 1.142 1.014 0 425 g
0428 1.014 1.157 1.293 1.182 1.305 1.019 1.317 1.019 1.307 1.181 1.291 1.157 1.015 0A29 1,64 1.68 1.73 0.00
-1.69
-0 46 1.86
-0 15
-0.59
-0 31
-0.85
-1.63
-1 30
-0.10 4 33 0.750 1.222 1.133 1.286 1.121 1.305 1.144 1.293 1.116 1.287 1.103 1.159 0 711 10 0.738 1.203 1.135 1.301 1 122 1.307 1.155 1.305 1.122 1.301 1.134 1203 0.738 1.63 1.58
-0.18
-1.15
-0.09
-0.15
-0 95
-0.92
-0.53
-1.06
-2.73
-3.66 3.66 0.494 1.042 1.201 0.88?
1.305 1.183 1 097 1.150 1.281 0887 1.188 1.021 0 495 11 0.484 1022 1.195 0899 1.301 1.181 1.098 1.182 1.301 0.899 1.196 1.024 0 484 2.07 1.96 0 50
-2.00 0.31 0.17
-0.09
-2.71
-1.54
-1.33
-0 67
-0.29 2.27 0.527 0.766 1.170 1.128 1.284 1.006 1.282 1.132 1.195 0 769 0 525
- 2 0.515 0.766 1.196 1.134 1.291 1.017 1.293 1.135 1.195 0.766 0 518 2.33 0,00
-2 17
-053
-054
-0 89
-0.85
-0.26 0.00 0.39 1.35 0.530 1.049 1.225 1.169 1.314 1.180 1226 1.033 0 523 0.518 1.024 1.203 1.157 1.287 1.157 1.203 1.022 0 515 13 2.32 2 44 1.83 1.04 2.10 1.99 1.91 0.78 l_55 0.496 0.756 1.040 0.965 1.036 0 753 0 489 0.484 0 738 1015 0.937 1.014 0 738 0 484 14 2.48 2.44 2.46 2.99 2.17 2 03 1.03 0.439 0.447 0.441 INCORE Mean Absolute Difference = 0 0121 0 429 0434 0 428 ANC 15 Standard Deviation = 0.0149 2.33 3 00 3 04
% DIFFERENCE BURNUP = 14970 MWD /MTU POWER LEVEL = 100.0%
D B ANK AT 228 STEPS 73 l
FIGURE 4.3-22 V.C. SUMME81 NUCLEAR STATION CYCLE 5 RADI AL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP FCFM-05-006 R
P N
M L
K j
H G
F E
D C
B A
0 368 0407 0375 1
0 372 0 412 0.371 l.08
-1.21 1.08 0444 0976 1 047 0 873 1.061 1.016 0.470 2
0 459 0 979 1.044 0 872 1.043 0 980 0 459
-3.27
-0.31 0.29 0.11 1.73 3.67 2.40 0.398 1.110 1.222 1.146 1.303 1.156 1.245 1.148 0.414 3
0 407 1.128 1.227 1149 1.301 1.150 1.229 1.132 0 409
-2.21
-1.60
-0 41
-0 26 0 15 0 52 1.30 1.141 1 22 0.406 0 716 1.108 1.100 1114 1.100 1.113 1106 1.142 0.737 0412 4
0.409 0.727 1.127 1.112 1.118 1.105 1119 1.115 1.132 0.727 1.407 0.73
-1.24 1.69
-l ud
-036
-045 0 54
-0 81 0.88 1.38 1.23 0.467 1.155 1.133 1 049 1.246 1.103 1 068 1100 1.240 1.055 1.147 1.148 0458 5
0.459 U 32 U 32 1.057 1 253 1.124 1290 1.125 1 255 1057 1127 1128 1.459 1.74 2 03 0 09
-0 76 0.56
-1 87 2 02 2 22 1.20
-0.19 1 77 1.77
-0.22 0.992 1.244 1.116 1 250 1.173 1.284 1100 1.271 1.160 1.260 1.128 1.254 0.997 6
0 980 1.229 1.115 1255 1.186 1.309 1.129 1.309 1.186 1.253 1.112 1 227 0.979 1.22 1.22 0 09 0 40
-1.10
-1.91
-2 57 2.90 2.19 0 56 1.44 2.20 1.64 0.360 1 042 1.159 1.120 1.118 1.296 1.128 1.278 1124 1.273 1.110 1.145 1.177 1.076 0.381 7'
0 371 1.043 1.150 U 19 1.125 1.309 1.159 1.314 1159 1 309 1 124 1118 1.149 1.044 0 372
-2.96
-0.10 0.78 0 09 0 62
-0.99
-2 67
-2 74
-3 02
-2 75
-1.25 2 42 2 44 3 07 2.42 0.399 0.8U 1.317 1 097 1.066 1.104 1 287 1.022 1.287 1.103 1.086 1133 1.337 0893 0 420 g
0412 0 812 1.301 1105 1 090 1.129 1 314 1 054 1.314 1.129 1.090 1.105 1301 0 872 0 412
-3.16 0 46 1 23
-0.72
-2 20
-2 21 2 05 3 04 2 05 2.30
-0.37 2 53 2.77 2 41 1 94 0.361 1.046 1165 1.113 1.099 1.299 1.176 1.306 1.138 1.290 1.116 1138 1.176 1.075 0 380 g
0.372 1.044 1.149 1.118 1.124 1.309 1.159 1.314 1.159 1.309 1125 1 119 1.150 1.043 0.371
-2 96 0.19 1.39
-0.45
-2.22
-076 1.47
-0 61
-1.81 1.45
-0.80 1.70 2.26 3.07 2.43
=
0 993 1.245 1.099 1.225 1.173 1.316 1.119 1 307 1.184 1264 1.112 1.258 1.013 10 0 979 1 227 1.112 1 253 1 186 1 309 1.129 1.309 1.186 1 255 1115 1 029 0 980 1 43 1.47
-1 17
-2 23
-1.10 0 53
-0.89
-0 15
-0 17 0 72
-0.27 2.36 3.37 0.451 1.126 1.116 1028 1.252 1.126 109J 1.130 1.P68 1.059 1.133 1 147 0.470 11 0.459 1.128 1.127 1.057 1.255 1.125 1 090 1.124 1.253 1.057 1.132 1.130 0 459
-0.44
-0.18 0 98 2.74
-0.94 0 09 0 00 0.53 1.20 0.19 0.09 1.33 2.40 0.398 0.709 1.104 1.100 1.123 1 109 1134 1.127 1.139 725 0409 Jg 0.407 0.727 1.132 1.115 1.119 i.105 1.118 1.112 1.127 0 727 0 409
-2 21 0.48
-2 47
-1.35 0 36 0 36 1 43 1.35 1 06
-0.28 0 00 0 403 1.127 1.228 1.163 1.323 1.172 1.255 1.140 0.407 0.409 1.132 1.229 1.150 1.301 1149 1.227 1.128 0 407 13
-1.47 0 44 0.08 1.13 i.e9 2 00 0.28 i 06 0 00 0.457 0.978 1.056 0.883 1.068 1.001 0.468 0.459 0 980 1.043 0.872 1.044 0 979 0 459 14
-0 44
-0.20 1.25 1.25 2.30 2.25 1.96 0.370 0.415 0.379 INCORE Mean Absolute Difference = 0.0134 0.371 0 412 9.372 ANC 15 Standard Deviation = 0.0168
-0.27 0.73 1.88
% DIFFERENCE BURNUP = 423 MWD /MTU POWER LEVEL = 99.9%
D B ANK AT 230 STEPS 74
!l
FIGURE 4.3-23 V.C. SUMMER NUCLEAR STATION CYCLE 5 RADI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 05 012 R
P N
M L
K J
H G
F E
D C
B A
o372 0 411 0377 1
0 380 0 420 0 379
-2 11
-2 14
-0 53 0450 0 9 tid 1059 0.862 1.066 0985 o 4ti4 2
0459 0 967 1.057 0.863 1056 0.967 0 459
-1.96
-0.41 0.19
-0 12 0.95 1.86 1.09 0.414 1.119 1.262 I til I.289 1.114 1.265 1133 0 421 3
0 418 1.124 1.259 1.113 1.285 1.114 1 261 1.127 0420
-096
-044 0.24
-018 0.31 0 00 0 32 0 53 0.24 0 415 0 744 1.183 1 096 1.066 1.041 1.065 1.092 1.200 0 755 0 42t, 4
0 420 0 75?
I192 1105 1.071 1 047 1.071 1.107 1.196 0 752 0 418
.I 19
-1.06
-076
-0.81 0 47
-0 57
-0.56 1 36 0 33 0 40 0.48 0.459 1.132 1.202 1.079 1.305 1.067 1 020 1.066 1.286 1.070 1.205 1.136 0.455 5
0 459 Ll27 Ll" 1 077 3 301 3 088 1 040 1 089 L303 3 077 l 192 1 124 0 459 O 00 0 44 0 50 0 19 0.31
.l.93
-I92
-2 11
-130
-0 65 1.09 1 07
-0 87 O.968 1.060 1.105 1.313 1.172 1.311 1.f 35 1.303 1.156 1.302 1.110 1.270 0 962 6
0.967 1.261 1.107 1.303 1.176 1.327 1 li 3 1.327 1.176 1301-1.105 1.259 0.967 l
0.10 0 08 0.18 0.77 0.34
-1.21
- 1.t.3 1.81
-l 70 0.08 0 45 0.87
-0.52 0.373 1.060 1.122 1.078 1.095 1.331 1.127 1.325 1121 1.303
!.072 1.087 1.123 1.073 0.376 7
0.379 1.056 1.114 1.071 1.089 1.327 1.146 1.342 1.146 1.327 1088 1.071 1.113 1.057 0 380
-1.58 0.38 0:72 0.65 0.55-0.30
-1.66
-1.27
-2.18 1 81
-1.47 1.49 0.90 1.51 1.05 1
l 0.413 0.863 1.303 1040 1.019 1.083 1334 1.038 1330 1087 1.033 1 OG4 1.302 0 870 0 419 g
0.420 0.863 1.285 1.047 1.040 1.103 1342 1.053 1342 1.103 1.040 1.047 1.285 0,863 0 420
-1.67 0.00 1.40 0.67 2 02
-1.81
-0.60
-1.42 0 89
-1.45
-067 1.62 1.30 0 81 0,24 0.374 1.072 1.139 1.071 1.064 1.324 1.178 1.353 1.131 1.317 1.077 1.081 1.126 1.075 0382 g
0.380 1.057 1.113 1 071 1.088 1.327 1.146 1.342 1.146 1.327 1.089 1.071 1.114 1.056 0 379 1.58 1.42 2 34 0,60 2.21 0.23 2 79 0 82
-1.31
-0.75
-1.10 0 93 1.08 1.80 0.79 0.992 1.291 1.093 1.270 1.167 1.350 1.105 1.333 1.177 1.311 1.096 1.278 0 990 10 0 967 1.259 1.105 1.301 1R6 L327 1.103 1.327 1176 L303 1107 1.261 0.967 2 59 2.54
-1.09
-2.38
-0 77 1.73 0.18 0.45 0 09 0.61 0 99 1.35 2.38 0460 1.130 1.186 1.047 1.306 109fi 1.046 1.095 1.317 1 073 1.188 1.131 0 4ti7 11 c.459 1124 1.192 1.077 1.303 1.089 LO40 1.088 1.301 1.077 1.196 1.127 1.459 0.22 0.53
-050
-2.79 0.23 0 64 0.58 0 64 1.23
-0.37
-0.67 0.35 1.74 0.410 0.734 1.166 1.091 LO77 1.053 1184 1.114 1.198 0.744 0 416 12 0 418 0752 1196 1.107 1071 1.047 1.071 1.105 1.192 0 752 0.420 1.91 2.39 2.51
-1.45 0.56 0 57 1.21 0.81 0.50 1.06
-0 95 0 416 1.131 1.264 1127 1 306 1.131 1.278 1.124 0.413 0 426 1.127 1.261 1.114 1.285 1.113 1 259 1.124 0 418 13
-095 0.35 0.24 1.17-1.63 1.62 1.51 0 00
-1.20 0460 0 971 1.074 0873 1 078 0.982 0465 0 459 0 967 1.056 0863 1.057 0 967 0459 14 0 22 0 41 1.70 1.16 1.99 1.55 1.31 0379 0423 0.384 INCORE Mean Absolute Difference = 0 0100 0 379 0 420 0.380 ANC 15 standard Deviation = 0.0127 0 00 0.71 1.05
% DIFFERENCE BURNUP = 3260 MWD /MTU POWER LEVEL = 100.0%
D BANK AT 230 STEPS 75
FIGURE 4.3-24 V.C. SUMMER NUCLE AR STATION CYCLE 5 RADI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAPS!GM 05 015 R
P N
M L
K J
H G
F E
D C
B A
0396 0 441 0.401 1
0 402 0447 0 401
-1 49 1.34 0 00 0 465 0957 1081 0 878 1.092 0989 0.472 2
0469 0.969 1.083 0 884 1 082 0 968 0 469
-0.85
-1.24
-0 18
-0.68 0.92 2 17 0 64 0 434 1.123 1.266
- 1. 00 1.284 1.105 1.270 1.123 0 434 3
0.434 1 100 1.268 1109 1.287 1.109 1.269 1122 0436 0.00 0 27
-0 in
-0 81
-0 23
-0.36 0 08 0.09
-0.46 0435 0 773 1 216 1.095 1.052 1.027 1.047 1.075 1.212 0771 0 432 4
0.436 0 776 1.018 1.102 1.059 1.033 1.058 1.103 1.221 0 776 0 434
-023
-0.39
-0 16
-0 64
-0 66
-0 58
-1.04 2 54
-0 74 0.64
-046 0417 1.130 1 232 1 069 1.313 1.060 1.011 1 049 1.278 1.066 1215 1.117 0459 5
R4M U 22 1 221 1.084 1302 1.069 1 021 1.069 1.303 1084 1.218 1120 0.469 0.43 0 71 0.90 0.46 0 84 0 84 0 98
-1.87
-1.92
-1.66
-0 25
-027 2 13 0973 1.274 1107 1.319 1.157 1.300 1.069 1.284 1.129 1.292 1 093 1.266 0.961 6
0.968 1,269 1130 1.303 1.153 1.299 1075 1.299 1153 1.302 1.102 1.263 0.969 0.52 0 39 0 36 1 23 0.35 0 08
-0 56
-1.15
-2 08
-0 77
-0.82
-0.16
-0.83 0.369 1.083 1.118 1.071 1.087 1.318 1109 1.303 1.096 1.280 1.051 1067 1.109 1091 0.398 7
a401 1.082 1.109 1.058 1.069 1.299 1.114 1.308 1.114 1.299 1.069 1.059 1.109 1.083 0.402 2.99 0 09 0.81 1.23 1.68 1.46
-045
-0 38 1.62
-1.46 1 68 0.76 0.00 0.74 1.00 0.434 0880 1.304 1.037 1.019 1.073 1.313 1020 1.309 1.071 1.020 1.043 1.294 0881 0 442 8
0.447 0884 1.287 1.033 1.021 1 075 1.308 1.027 1 308 1 075 1 021 1 033 1.287 0 884 0.447 2.91
-0.45 1.32 0.39
-0.20 0 19 0.38 0 68 0.08
-0.37
-0.10 0 97 0 54
-0.34
-1.12 0.391 1.090 1.127 1 066 1.066 1.318 1.155 1 334 1.115 1.308 1.069 1.067 1.112 1.087 0 398 9
0.402 1.083 1.109 1.059 1069 1.299 1.114 1.308 1.114 1.299 1069 1058 1109 1.082 0 401 2 74 0 65 1 62 0 66
-0.28 1.46 3 68 1 99 0.09 0 69 0.00 u 85 0.27 0.46
-0.75 0.986 1.290 1 098 1292 1.159 1.329 1085 1.319 1.162 1 315 1087 1.274 0 971 10 0.969 1.268 1.102 1302 1.153 1 299 1.075 1.299 1153 1303 1103 1.269 0 968 1.75 1.74
-0 36
-0.71 0 52 2 31 0 93 1.54 0.78 0.92 1.45 0 39 0.31 0 469 1.125 1.217 1 068 1.317 1079 1029 1.081 1.313 1.075 1.204 1 115 0 468 jj 0469 1.120 1.218 1.084 1.303
- 069 1 021 1 069 1302 1084 1.221 1122 0.469 0.00 0.45
-008
-1.48 1.07 0 94 0.78 1 12 0 84 0 83
-1.39
-0.62
-0 21 0 428 0.765 1.208 1.093 1 061 1.036 1 070 1.107 1.218 0 760 0 427 12 0434 0.776 1 221 U O3 1 058 1.033 1.059 1.102 1218 0776 0 436
-1.38
-1.42 1.06
-0 91 0.28 0.29 1.04 0.45 0 00
-2 06
-2 06 0 429 1.106 1.260 1.116 1.299 1.117 1.279 1.115 0 425 0 436 1.122 1269 1109
- 1..' 87 1.109 1.268 1.120 0 434 13
-1.61 1.43
-0 71 0 63 0.93 0.72 0.87 0 45
-2.07 0 461 0.957 1.087 0884 1.097 0977 0473 0 469 0.968 1.082 0.884 1083 0 969 0,469 14
-1.71
-1.14 0.46 0 00 1.29 0 83 0 85 0.395 0 444 0.402 INCORE Mean Absolute Difference = 0.0085 0.401 0 447 0.402 ANC 15 stasndard Deviation = 0 0109
-1.50
-0.67 0.00
% DIFFERENCE BURNUP = 6174 MWD /MTU POWER LEVEL = 100.0%
D B ANK AT 230 STEPS 76
RGURE 4.3-25 V.C. SUMMER NUCLEAR STATION CYCLE 5 RADI AL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC FOR MAP FCFM 05 018 R
P N
M L
K J
H G
F E
D C
B A
u414 0 463 0 416 1
0 419 0 469 0 417 1 19 1.28 0 24 o 468 0955 1.097 0 897 1.105 0 979 0477 2
0477 0.968 1 097 0.901 1 096 0 968 0 476
-1.89 1.34 0 00 0.44 0.82 I.14 0 21 0 442 1.112 1 260 1.104 1.289 1.109 1.260 1.113 0 444 3
0 44f; 1.113 1.262 1.110 1 288 1.110 1.263 1.114 0 447
-0 90
-0 09
-0 10
-0 54 0 08
-0 09
-0 24
-018
-0 67 0445 0786 1.215 1.092 1.055 1.031 1.054 1 079 1.214 0.784 0.441 4
0 447 0.789 1.219 1.099 1.059 1.036 1059 1.100 1.221 0 789 0.446
-0 45
-0.38
-0.33
-0 64
-0.38 0 48
-047 1.91 0 57
-3 63
-1.12 0.480 1.12s 1.236 1 091 1.305 1.058 1014 1.046 1268 1065 1.217 1.105 0 463 5
0 476 1.115 1.221 1.083 1.292 1.065 1.020 1065 1.293 1.083 1.219 1.113 0 477 l
0.84 1 17 1.23 0 74 1.01
-0 66 0 59 l.78 1 93
.l.66 0.16 0 72
-2 94 1
0 975 1.271 1.107 1.312 1.147 1.283 0058 1.261 1.113 1.277 1.089 1.256 0,960 6
0 968 1.263
.100 1.293 1.141 1.281 1 066 1.281 1.141 1.292 1.099 1.262 0 968 0.72 0.63 034 1.47 0.53 0.16
-0.75
-1.56 2 45
-1.16
-0 91
-0 48
-0 83 0.402 1.095 1.121 1.071 078 1.297 1.097 1.281 1.077 1.257 1.040 1.059 1.107 1.102 0.415 7
0 417 1.096 1.110 1.059 1.065 1.281 1.100 1 286 1.100 1.281 1.065 1.059 1.110 1 097 0.419
-3.60 0.09 0 99 1 13 1.22 1.25 0.27
-0 39
-2.09
-1.87
-2.35 0.00
-0.27 0.46
-0.95 0.451 0.896 1.308 1.039 1.015 1.063 1.295 1.008 1.287 1,061 1.017 1.038 1.290 0 897 0 463 g
0.469 0.901 1.288 1.036 1.020 1.066 1.286 1.016 1.286 1.0tl6 1.020 1.036 1.288 0 90!
O469 3.84 0.55 1.55 0.29 0.49
-0 28 0.70
-0.79 0.03
-0.47
-0 29 0.19 0.16
-0.44 1.28 0.403 1.102 1.129 1.066 1.060 1.300 1.146 1315 1.102 1.290 1.064 1,061 1.107 1.096 0.413 g
0.419 1.097 1.110 1.059 1.065 1.281 1.100 1.286 1.100 1.281 1.065 1.059 1.110 1.096 0417
-3.82 0.46 1.71 0.66
-0.47 1.48 4.18 2.26 0.18 0 70
-0.09 0 !9
-0.27 0 00 0 96 0 984 1.282 1.096 1.284 1.15' l.317 1.080 1.303 1.153 1.306 1 084 1.261 0.966 10 0.968 1.262 1.099 1.292 1.141 1.281 1.066 1.281 1.141 1.293 1100 1.263 0 968 1.65 1.58 0.27 0 62 0 88 2 81 1.31 1.72 1 05 1 01
-145 0 16 0 21 0 479
.l.124 1.221 1.066 1312 1.083 1.036 1.081 1.308 1.075 1.207 1.109 0.476 11 0 177 1.113 1.219 1.083 1.293 1.065 1.020 1,065 1.292 1.083 1 221 1.115 0 476 0.42 0.99 0.16
--l.57 1.47 1.69 1.57 1.50 1.24
-0 74 1.15
-0.54 0.00 0.445 0.783 1.207 1.096 1.068
- 1. 14 4 1.071 1.104 1.218 0.776 0.440 12 0 440 0 789 1 221 1 100 1 059 1 430 1 059 l 099 l 219 0 789 0 447
-0.22
-0.76
-1.15 0 36 0 85 0.77 1.13 0.45 0 08
-1.65 1.57 0.440 1.090 1.254 1.124 1.307 1.119 1.271 1.102 0438 0.447 1.115 1.263 1.110 1.288 1.110 1.262 1.113 0 446
= 13
-1.57 2.24
-0.71 1.26 1.48 0.81 0 71
-0.99
-1.79 0 463 0.955 1.104 0.906 1.111 0 976 0476 0 476 0 968 1.096 0 901 1.097 0968 0477 14
-2.73
-1.34 0.73 0.55 1.28 0 83 0.21 Mean Absohate Difference = 0.0092 IN A C 15 Standard Oeviation = 0.0120 0 72 0.00 0.95
% DIFFERENCE BURNUP = 8168 MWD /MTU POWER LEVEL = 99.8%
D BANK AT 230 STEP 5 77
RGURE 4.3 26 V.C. SUMMER NUCLEAR STATION CYCLE 5 RADIAL POWER DISTRIBU TION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 05 020 R
P N
M L
K J
H G
F E
D C
B A
0433 0 487 0,436 1
0 438 0 493 0 436
-1.14
-1.22 0.00 0.466 0963 1.112 0919
- 1. I I 8 0.983 0489 2
0 486 0 968 1.108 0 919 0.106 0 967 0.485 0.00
-0 52 0 36 0.00 1.08 1.65 0.82 0 461 1.119 1.260 1.112 1.292 1.113 1.254 1.112 0460 3
0 458 1.107 1.253 1.112 1.285 1.112 1.253 1.108 0 aco 0.66 1.08 0 56 0.00 0.54 0.09 0.08 0.36 0.00 0.462 0.803 1.219 0.090 1059 1.038 1.058 1.076 1.213 0800 0 457 4
0.460 0.803 1 215 1.096 1.061 1.040 1.061 1.097 1.217 0 803 0 458 0.43 0.00 0 33 0.65
-0.19
-0.19
-0.28 l 91 0.33
-0.37
-0.22 0.489 1.121 1.231 1 uss 1.287 1 050 1.010 1.044 1.257 1.066 1.214 1.106 0.476 5
0.485 1.108 1.217 1.082 1.279 1.062 1.022 1063 1.279 1.082 1.215 1.107 0.486 0.82 1.17
- 1. i 5 0 55 0 63
-1.13
-1.17
-1.79 l 72 1.48
-008
-0.09 2 06
(
D Le 9 1.255 1 u99 1.293 1.131 1.257 1.048 1.244 1.108 1.267 1.087 1.252 0.965 6
0.9 37 1.253 1.097 1.279 1.130 1.262 1.059 1.262 1.130 1.279 1096 1.253 0 968 0/21 0. 1 11 0 18 1.09 0 09
-0 40
-1.04
-1.43
-l.95
-0.94
-0 82 0 08
-0.31 0.4?J 1.108 1.116 1.071 1.07d 1.279 1.080 1.257 1.0t19 1.243 1.041 1.047 1.102 1.107 0.435 7
0.436 1.106 1.112 1.0til 1.063 1.262 1 089 1264 1.089 1.262 1.062 1,061 1.112 1.108 0.438 1.38 0 18 0.36 0 94 1.41 1.n
-0.83 0 55
-l 84 1.51
-1.98
-1.32
-0.90
-0 09
-0,68 0 48ti 0 918 1.297 1.039 1 015 1.054 1.268 1.000 1.262 1.053 1.017 1 025 1 279 0 914 0 491 8
0.493 0 919 1.285 1.040 1.022 1.059 1 264 1 008 1.264 1.059 1.022 1 040 1.285 0 919 0.493
-1.42 0.11 0 93 0.10
-0 68
-047 0 32
-0,79
-016 0.57 0 49
.l.44
-047
-0.54
-0.41 0.431 1.119 1.13u 1.0ti6 1054 1.277 1133 1.287 1.085 1.265 1.057
! = 042 1.104 1.109 0.436 g
0.438 1.108 1.112 1.061 1.062 1.262 1.089 1.264 1.089 1.262 1.063 1 061 1.112 1.106 0.436 l.60 0.99 1.62 0 47
-0.75 1.19 4 04 1 82 0.37 0,24
-0.56
.l 79
-0.72 0 27 0.00 0 985 1.274 1.089 1.262 1.134 1.295 1 070 1.279 1.135 1.285 1.075 1.257 0 979 10 0.968 1.253 1.096 1.279 1.130 1.262 1.059 1.262 1.130 1.279 1.097 1.253 0.967 1.76 1 68
-. 0. 64 1.33 0.35 2 61 1 04 1.35 0 44 0.47 2.01 0 32 1.24 0 488 1.118 1,217 1.061 1.290 1.075 1.033 1.076 1.288 1.069 1.201 1.105 0.490 11 0 486 1.107 1.215 1.082 1.279 1063 1.022 1 062 1.279 1.082 1.217 1.108 0.485 0.41 0.99 0.16
-1.94 0.86 1.13 1.08 1.32 0.70
-1.20
-1.31
-0.27 1.03 0.457 0 795 1.200 1.084 1.066 1.045 1.073 1.097 1.211 0.788 0 454 12 0 458 0.803 1.217 1.097 1 061 1 040 1.061 1.096 1.215 0 803 0.460 0.22
.l.00 1.40
-1.19 0 47 0.48 1.13 0 09 0 33 1.87
-1.30 0 458 1.105 1.250 1.120 1.301 1.122 1.264 1.096 0 451 0 460 1108 1.253 1.112 1.285 1.112 1253 1.107 0 4h8 13
-0 43 0 27
-0.24 0 72 1.25 0.90 0 88
-0.99
-1 53 0 483 0.967 1.120 0.927 1.124 0 977 0 486 0 485 0 967 1,106 0.919 1.108 0 968 0486 14
-0 41 0 00 1.27 0.87 1.44 0.93 0 00 0 435 0 496 0.442 INCORE Mean Absolute Difference = 0.0084 0 436 0 493 0.438 ANC 15 Standard Deviation = 0.0110
-0.23 0.41 0.91
% DIFFERENCE BURNUP = 10539 MWD /MTU POWER LEVEL = 99.9%
D B ANK AT 230 STEPS 78 l
FIGURE 4.3 27 V.C. SUMMER NUCLE AR STATION CYCLE 5 RADIAL POWER DISTRIBUTIC N COMPARISON BETWEENIN20RE AND ANC FOR MAP FCFM 05-022 R
P N
M L
K J
H G
F E
D C
B A
0.455 0516 0 458 1
0 456 0 517 0455
-0 22
-0.19 0.66 E500 0.957 1.121 0937 1.132 0.988 0499 2
0.495 0.968 1.117 0937 1.115 0.968 0 495 1.01
-1.14 0 36 0 00 1.52 2 07 0.81 0.476 1.119 1.244 1.111 1.285 1.115 1.242 1105 0472 3
0 471 1.102 1.242 1.114 1.282 1 114 1.242 1.103 0 472 1.06 1.54 0.16 0.27 0 23 0 09 0 00 0 18 0 00 0 477 0 819 1.216 1.036 1.059 1 041 1.059 1067 1.202 0 811 0.471 4
0 472 0.815 1.208 1.093 1.065 1046 1.064 1.094 1.210 0 815 0.471 1.06 0.49 0.06
-0.64
-0.56
-048 0.47 2.47
-0.66 0.49 0.00 0.500 1.120 1.224 1.086 1.271 1,019 1.014 1.038 1.231 1.058 1.206 1.106 0 489 5
0.495 1.103 1.210 1.079 1.263 1.061 1.026 1.062 1263 1079 1.208 1.102 0.495 l
1.01 1.54 1.16 0.65 0.63
.l.13
-1.17 2 26
-2 53
-1.95 0,17 0.36
-1.21 9.973 1.248 1.098 1.276 1.120 1.237 1.043 1.222 1 062 1.244 1.082 1.244 0.972 6
0 968 1.242 1.094 1,263 1.121 1.244
- 1. 0P, 1.244 1 121 1.263 1.093 1.242 0.968 l
0 52 0.48 0.37 1 03
-0 09
-0.56
-1.14 1.77 2.59
-l io
-1.01 0.16 0.41 0.445 1.115
- 1. ! ! 8 1.072 1.072 1.255 1.070 1.236 1.056 1.218 1 032 1 056 1.110 1.123 0 456 7
0.455 1.115 1.114 1.064 1 062 1.244 1.080 1.243 1 080 1244 1.061 1.065 1.114 1.117 0.456
-1.20 0,00 0.36 0 75 0.94 0.68
-0 93 0.56 2.22 2.09
-2.73
-0.85
-0.36 0 54 0.00 0.507 0933 1.294 1043 1.017 1.047 1.243 0 991 1.236 1043 1.015 1.038 1282 0938 0518 8
0 517 0.937 1.282 1046 1.026 1.055 1.243 1.003 1.243 1055 1026 1.046 1.282 0 937 C.517 l.93
+ 0.4 3 0.94
-0.29
-0.88 0.76 0 00
-1.20
-0 56
-1.14
-1.07 0 76 0.00 0.11 0 19 0 446 1.126 1.131 1.068 1.051 1.255 1.117 1.259 1.069 1.239 1.052 1.058 1.112 1.122 0.456 g
0.456 1.117 1.114 1.065 1.061 1.244 1.080 1.243 1.080 1.244 1.062 1.064 1.114 1.115 0 455
-2.19 0.81 1.53 0 28
-0.94 0.88 3 43 1.29 1.02
-0 40
-094
-0.56
-0 18 0.63 0 22 0.987 1.265 1.008 1.248 1.124 1.271 1.060 1.252 1.120 1.266 1.075 1.24d 0.983 10 0 968 1.242 1.093 1.263 1.121 1.244 1.055 1.244 1.121 1.263 1.094 1242 0.968 I,96 1.85
-0.46
-1.19 0.27 2.17 0 47 0 64 0.09 0.24
-1.74 0 48 1.55 0.502 1.122 1.217 1.061 1.276 1.073 1.036 1.068 1.267 1.066 1.199 1.105 0502 11 0.495 1.102 1.208 1.079 1.263 1.062 1.026 1.061 1.263 1.079 1.210 1.103 0 495 1.41 1.81 0.75
-1.67 1.03 1.04 0 97 0.66 0.32
-1,20
-0.91 0.18 1.41 0 477 0.814 1.195 1.087 1.068 1.049 1.071 1.092 1.204 0 805 0 470 12 0417 0815 1.210 1.094 1.064 1.046 1.065 LU93 1.208 0.815 0 472 1.27
-0.12
-1.24 0 64 0.38 0 29 0.56 0 09
-033
-1.23 0 42 0 474 1.104 1.245 L123 1.299 1.124 1.255 1.094 0 467 0 472 1.103 1.242 I114 1.282 1.114 1.242 1.102 0 471 13 0 42 0.09 0 24 0 81 1.33 0.90 1.05
-0.73
-0.85 0 492 0 971 1.128 0 946 1.133 0 980 0 495 0495 0 968 1.115 0 937 1.117 0 968 0495 14 0.61 0.31 1.17 0 96 1.43 1.24 0 00 0 456 0 521 0.463 INCORE Mean Absolute Difference = 0.0087 0.455 0 517 0 456 ANC 15 standard Deviauon = 0.0114 022 0 77 1.54
% DIFFERENCE l
BURNUP = 12793 MWD /MTU POWER LEVEL = 100.1%
D BANK AT 230 STEPS l
l 79 l
RGURE 4.3 28 r
V.C. SUMMER NUCLEAR STATION CYCLE 3 AXIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP-FCFM 03-007 2
l 1.75 INCORE x
ANC N-1.5 2
0 R
M-L A-l.
L' 1.25 k _.
Y xx i
/
A 1
N x
-A-x X
x
-l^
.75 3',_
.- P 9
n W-
,5 E --
x.
R
.25 0
0:
12 24 36' 48.
60 72 84 96.
108 120 132-144
. BOTTOM TOP AXIAL HEIGHT, Inches BURNUP = 1839 MWD /MTU POWER LEVEL = 100.0%
D BANK AT 228 STEPS 80
FIGURE 4.3-29 V,C. SUMMER NUCLEAR STATION CYCLE 3 AX1AL POWER DISTRIBUTION COMPARISON -
BETWEEN INCORE AND ANC FOR MAP-FCFM-03 014 2
INCORE x
ANC
.N-g
-O
~
R M
A-
~ L ~ 1.2 5.
.1 I
.Z e,
1 A
y x
+
x.
A N *
- X-
-l-A 75 P
=
- ~ O -
x
-- W
'5 E:
.R
.2 5 -
0 0'
.12 -
24-36 48 60
.72 84 96 108 120 132 144 BOTTOM'
D BANK AT 228 STEPS 81
.~
FIGURE 4.3 30 V.C. SUMMER NUCLEAR STATION CYCLE 3 AXI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP-FCFM-03-017 2
- 1. 5 INCORE x
ANC N
O 1.5 R
M A
L 1.25
-x x
- ==,,
N x.
3
=
\\
A
- J
.75 P
O W
~5 i
/
\\
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL HEIGHT, Inches BURNUP = 5558 MWD /MTU POWER LEVEL = 99.8%
D BANK AT 228 STEPS 82
l l
FIGURE 4.3 31 V.C. SUMMER NUCLEAR STATION CYCLE 3 AXI AL POWEA DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP-FCFM 03-025 2
1.75 INCORE x
ANC N
O 1.5 R
M A
L 1.25 I
Z E
f6 TAL,
x x xx,,
,,n,,,%
D f
x, A
X l
f
.75 5
P O
u W
'5 E
d
\\ u R
n
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXl AL HEIGHT, Inches BURNUP = 8689 NiWD/MTU POWER LEVEL = 100,0%
D BANK AT 228 STEPS 83
4'
'i FIGURE 4.3-32 V.C. SUMMER NUCLEAR STATION CYCLE 3 AXIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 03-031 2
1.75
' ~~
INCORE 4
ANC No 1.5 R
M A
L 1.25 I
Z E
^Y'
- h*****
D
^
t
" " * * " :f,
-g 3
.t A _ m
~
- A x
X I
=
g<
A
\\*
L
.75 P
O
~
W
'5 E
\\
R
.25 l
0 0
12 24-36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXtAL HEIGHT, Inches BURNUP = 11123 MWD /MTU POWER LEVEL = 99.9%
D BANK AT 228 STEPS 84
13 "
s FGURE 4.3 33 V.C. SUMMER NUCLEAR STATION CYCLE 3
- AXIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR M AP.FCFM-03-037 2
.5 INCORE x
ANC N
-l JO-1.5 -
>)
-q.
.M.
A.
L 1,25
- l.-
2-1
.xx p xx x xx x x j
^
I
,\\
A xx X
l
=
l A
t
.75 P
O.
- W--
'5' E-'
R=
i
.25
.I i
.n 0.-
12 24-36 48 -.
60 72 84 96 108-120 132 - 144 BOTTOM TOP
~ AXI AL HEIGHT, Inches BURNUP = 13581 MWD /MTU POWER LEVEL = 100.0%
- D BANK AT 228 STEPS 85 i
b q
cw y gutei+gyid--
A m.
w
-m.--,,.-._m m m
.___.,,,_________a
_m___m._______,.,___._______.e
l:
4/
FIGURE 4.3-34
'V.C. SUMMER NUCLEAR STATION CYCLE 4
)
AXI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC i
FOR MAP-FCFM-04 005 i
-2 i
i 1.75 INCORE x
ANC E
N-1.5
.O R
M A
L! 1.25 I --
t-12 E
>=* * *=' * * " " " "
=""xx=x-
- ,m,.
' _ ^
'D:
.j M-
,n
" Q x
x,_
g y
[
\\
l x
n
.A'
.75
.- L
.P.
- O
-- W -
J.5 E
-- R s 4
i!'
.2 5 -
0' 0:
12-24
.36 48 60 72 84 96-108 120 132 144 BOTTOM -
TOP' AXIAL HEIGHT, Inches BURNUP =.1232 MWD /MTU POWER LEVEL = 99.8%
D BANK AT 228 STEPS 86 i
i
=
FIGURE 4.3-35
~
V.C. SUMMER NU' LEAR STATION CYCLE 4 AXI AL POWER DISYRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP-FCFM-04-010
-2 1.75 INCORE x
ANC N-
=
1.5
-O R
M
-A
. L - 1.2 5 l
l
?
2
- x E-x x
xxxnxx m
-x XX x
g LA.
X.
-x g
- 75 P-n
.O W'
.5 -
-- E -
is R-l
.25 I'
l-0-
lE 0-12 24 36 48 60- ' 72 84 96 108 120 132 144-BOTTOM TOP p
AXIAL HEIGHT, Inches
' BU'RNUP = 4154 MWD /MTU POWER LEVEL = 100.0%
D BANK AT228 STEPS 87
o FIGURE 4.3-36
- V,C. SUMMER NUCLEAR STATION CYCLE 4 AX1 AL POWER DISTRIBUTION COMPARISON BETWEEN INCCRE AND ANC FOR MAP FCFM 04-012 2~
1.75 INCORE x
ANC N
L 1.5 O
R M
A.
L' 1,25 o
- I
~
uKturv y x x JCx a x*
v
.-[
xxx u=xx i
w A-
)
- X
. A 75 L.
T p-4
- o.
. W
.5 E-
- R-
.25 4
_ 0 12 24 36 48 60 84 96 108 120.132 144-BOTTOM TOP AXlAL HEIGHT, Inches
.BURNUP = 6398 MWD /MTU POWER LEVEL = 99.8%
D BANK AT 228 5TEPS 88 I
FIGURE 4.3-37 V.C. SUMMER NUCLEAR STATION CYCLE 4 AXIAL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP-FCFM-04 016 2
1.75 INCORE x
ANC
-N 1.5 O
R M^
l L 1.25 l
l 1
.xxx
[,F-x..,
, x xx,,
7 D-M *4;-
1
/
A X
i
\\
'73 L
~~
lt P
=
O W
.5 E
R
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXlAL HEIGHT, Inches BURNUP = 8608 MWD /MTU POWER LEVEL = 100.1%
D BANK AT 228 STEPS 89
FIGURE 4.3 38 V.C. SUMMER NUCLE AR ST ATION CYCLE 4 AXt AL POWER DISTRIBUTION COMF iRISON BETWEENINCORE AND ANC FOR MAP FCFM 04 021 2
1.75 INCORE x
ANC N
15 0
R M
A L 1.25 1
2 E
[, = " x
,x,u, xxo,,
C 1
I
.5 n
P O
l W
.5 I
E R
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXI AL HEIGHT, Inches BURNUP = 12192 MWD /MTU POWER LEVEL = 99.9%
D BANK AT 226 STEPS 90
FIGURE 4.3-39 V.C. $UMME R NUCLE AR ST All0N CYCLE 4 AXI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 04 026 2
1.75 INCORE x
ANC N
1.5 O
R M
A L 1.25 1
2 o
c=r.
- =a==
_..nx...
1 m
.3' X
=
a i
A
.75 L
=
P O
W
.5 E
R
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXt AL HEIGHT, Inches BURNUP = 14970 MWD /MTU POWER LEVEL = 100.096 D B ANK AT 228 STEPS 91
FIGURE 4.3 40 V.C. SUMMER NUCLEAR STATION CYCLE 5 AXI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP.FCFM-05 006 2
1.75 INCORE x
ANC 1.5 R
M A
~l' L 1.25 N
l f
n j
f
/~.,
O f
[*
A X
I A
.75 L
),
P
=
0 W
.5 E
R
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXI AL HEIGHT, inches BURNUP = 423 MWD /MTU POWER LEVEL = 99.9%
D B ANK AT 230 STEPS 92
FIGURE 4.3-41 V.C. SUMMER NUCLE AR STATION CYCLE 5 AXI AL POWER DISTRl8U110N COMPARISON BETWEENINCORE AND ANC F OR M AP-F CF M-05-012 2
1.75 INCORE x
ANC N
1.5 0
i R
M A
L 1.25 i
M-
_x f
N m,
O k*-
)
4.
A
=
X l
A
.75 y
P O
W
.5 i
I E
f R
)
.25 j
y 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AX1 AL HEIGHT, Inches BURNUP = 3260 MWD /MTU POWER LEVEL = 100.0%
D BANK AT 230 STEPS 93
l l
FIGURE 4.3 42 V.C. SUMMER NUCLEAR STATION CYCLE 5 AXI AL POWER DISTRIBUTION COMPARl50N BETWEENINCORE AND ANC FOR MAP f CFM 05 015 2
- 1. 5 lyCORE x
ANC 1.5 R
M A
L 1.25 Z
/ J ""W l
1_
,o b
~%
=
[
A X
r 75 L
P
]
O W
.5 E
R
.25 0
O 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXtAL HEIGHT, Inches BURNUP = 6174 M\\'st MTU POWER LEVEL = 100.0%
D BANK AT 230 STEPS 94
FIGURE 4.3-43 V.C. SUMMER NUCLEAR 51AllON CY CLE 5 AXI AL POWER Di$TRIDUTION COMPARISON bETWEENIW ORE ANU ANC FOR M AP tCF M 0%016 2
- 1. 5 INCORE x
ANC N
15 O
R M
A L 1,25 1
QL.......,
mn,===,
D "a
x i
p A
=
X l
=
A Y
.75 P
O W
.5 E
R
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXI AL HEIGHT, Inches BURNUP = 8160 MWD /MTU POWER LEVEL = 99.8%
D B ANK AT 230 STEPS 95
FIGURE 4.3-44 V.C. SUMMER NUCLEAR STATION CYCLF 5 AXI AL POWER DISTRIBUTION COMPARISON BETWEENINCORE AND ANC FOR MAP FCFM 05-020 2
1.75 INCORE x
ANC N
1.5 O
R M
A L 1.25 I
Z
""=
E
^
.x D
mg m
"n s
.75 L
P O
I W'
.5
/
\\
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXtAL HEIGHT, Inches BURNUP = 10439 MWD /MTU POWER LEVEL = 99.9%
D BANK AT 230 STEPS 96
1 I
FIGURE 4 3 45 V.C. SUMME R NUCLE AR ST ATION CYCLE 5 AXf AL POWE R DIS 1RIBUTION COMPARISON BETWEENINCORE AND ANC f OR MAP.FCIM 05 022 2
5 INCORE x
ANC N
1,5 O
R M
A L 1.25 1
I k
~.....,..
.X 3
,a x.
)
g i
\\
^
.75 L
P O
W
.5 E
R i,
.25 0
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXI AL HEIGHT, Inches BURNUP = 12793 MWD /MTU POWER LEVEL = 1001%
D BANK AT 230 STEPS 97
I i
"This page intentionally blan k."
98 1
5.0 ItEFEltENCES 1.
Langford, F. L. and Nath,11. J., " Evaluation of Nuclear llot Channel Factor Uncertainties," WC AP 7308.L, April 1969, and Spier, E. M. and Nguyen, T. G.,
" Update to WCAP-7308 L (Proprietary), Evaluation of Nuclear llot Channel Factor Unceruinties," March 1984.
2.
Meyer, C. E. and Stover,11. L.,"lNColtE Power Distribution Determination in Westinghouse Pressurized Water lleactors,"WCAP 8498, July 1975.
3.
Nguyen, T. Q., et al, " Qualification of the PilOENIX P/ANC Nuclear Design System for Pressurized Water llenetor Cores," WCAP.ll596.P-A (l'roprietary),
November 1987.
4.
Miller, P. W., et al, "Itelaxation of Constant Axial Offset Control /FQ Surveillance Technical Specification," WCAP-10216-P-A (Proprietary), June 1983.
5.
llordelon, F. M., et al," Westinghouse lleload Safe'y Evaluation Methodology,"
WCAP-9272-P-A (Proprietary), J uly 1985.
6.
Camden, T. M., et al, "Itod llank Worth Measurements Utilizing llank Exchange," WC AP-9863 A tProprietary), May 1982.
7.
Camden, T. M., et al, "PALADON-Westinghouse Nodal Computer Program,"
WCAP-9485 (Proprietary) and WCAP-9486, December 1978 and Supplement 1 WCAP-9485 A (Proprietary) and WCAP 9486 A (Non. Proprietary), September 1981.
8.
Liu, Y. S., et al, "ANC: A Westinghouse Advanced Nodal Computer Code,"
WC AP-10965-P-A (Proprietary), December 1985.
9.
Poncelet, C. G., et al,"LASEll-A Depletion Program for Lattice Calculations Based on MUFT and THEllMOS,"WCAP-6073, April 1966.
99
- 10. Olhoeft,J. E.,"The Doppler Effect for a Non-Uniform Temperature Distribution in iteactor l'uel Elements," WCAP 2048, July 1962.
I1. Ilarris, A.J., et al,"A Description of the Nuclear Design Analysis Programa for lloiling Water lleactors," WC AP-10100 P-A (Proprietary), J une 1982.
- 12. llarry,11. F., et al, "The PANDA Code," WCAP 7048 P-A (Proprietery) and WCAP-7757-A, January 1975.
100
6.0 APPENDIX This appendix describes the major Westinghouse computer programs used by SCE&G to perform reload core design calculations for VCSNS. The codes are used in a manner similar to that described in Section 3 of Westinghouse's licensed reload methodology topical report (lleference 5). Although the codes described in this appendix are not specifically mentioned in the topical, two of the codes, FIGllTli and APOLLO (lleference 12), contain the same fundamental methodology as the licensed versions. The " updated versions" provide engineering enhancements (e.g, larger problem size capabilities, editing improvemen'.s. and minor modeling improvements) relative to the original code versions. The updated code versions were described at a meeting between the Westinghouse Nuclear Fuel Division and the NRC Core Performance Ilranch in October 1984, at which time the ditTerences between the original and updated code versions were discussed. The NitC Core Performance llranch agreed that the updated code versions were fundamentally the same as the orginal versions, employing the same fundamental solution algorithms as the original versions.
The two remaining codes, PilOENIX-P and ANC, contain significant improvements to the methodoloi;ies discussed at the 1984 meeting between Westinghouse and the NRC. PIIOENIX-P is a two-dimensional multigroup lattice code which does nct rely on the spatial / spectral interaction assumptions inherent in the previous methodology. ANC is an advanced version of the PALADON code (Reference U incorporating the nonlinear nodal expansion method, the equivalence theory for cross section homogenization, and a rod power recovery model. Topical reports (References 3 and 8) qualifying PIlOENIX-P and ANC for use in reload core design have been approved by the NRC.
101
{
l
6.1 FIGilTil The FIGIITil code calculates efTective temperatures in low enriched, sintered
'J02 fuel rods for specined values of burnup, linear heat generation rate, moderator temperature and flow rate. The resultant fuel and clad temperatures are input to the PilOENIX-P code. The FIGilTil model accounts for the radial variation of heat generation rate, thermal conductivity, and thermal expansion in the fuel pellet; elastic denection in the cladding; and a pellet-clad gap conductance which depends on the kind of initial fill gas, the hot open gap dimensions, and the fraction of the pellet circumference over which the gap is effectively closed due to pellet cracking. Iteferences 9 and 10 provide a description of the basis of the FIGIITil program.
6.2 PilOENIX.P The PII0ENIX-P computer program is a two-dimensional multigroup transport theory code used to calculate lattice physics parameters for PWR core modeling.
In PilEONIX P, the solution for the detailed spatial Oux and energy distribution is divided into two major steps. In the first step, a two-dimensional One energy group nodal solution is obtained which couples individual subcell regions (pellet, clad, and moderator) as well as surrounding pins. PIIOENIX P uses a method based on the Carlvik's collision probability approach and heterogeneous response Ouxes which preserves the heterogeneity of the pin cells and their surroundings.
The nodal solution provides an accurate and detailed local Oux distribution which is then used to spatially homogenize the pin cells to fewer groups.
The second step in the solution process solves for the anrilar flux distribution using a standard S4 discrete ordinates calculation. This step is based on the group-collapsed and homogenized cross sections obtained from the Orst step of the solution, The S4 Duxes are then used to normalize the detailed spatial and energy nodal fluxes. The normalized nodal Duxes are used to compute reaction rates and power distributions and to deplete the fuel and burnable absorbers. A standard B1 calculation is employed to evaluate tue fundamental mode critical spectrum and to provide an improved fast diffusion coefficient for the core spatial codes.
102
l l
The PilOENIX-P code employs a 42 energy group library which has been derived mainly from ENDF/Il-V Gles. The PilOENIX-P cross section library was designed to properly capture integral properties of the multi-group data durinh group collapse, enabling preper modeling of 'mportant resonance parameters.
The library contains all neutronic data necessary for modeling fuel, fission products, cladding and structural, coolant, and control / burnable absorber materials present in PWits.
Detailed discussions of the methodology and models incorporated in PilOENIX-P are contained in Iteferences 3 and 11.
G.3 A NC ANC is a multidimensional nodal analysis program used to predict core reactiv;ty parameters, power distributions, detector thimble Guxes, and other relevant core characteristics. In ANC, the nodal expansion method is used to solve the two-group diffusion equations. With this method, the partial currents and average neutron fluxes for a node are determined from continuous J
RD%
. g homogeneous neutron flux profiles described by fourth order polynomial y
expansions for each of the x, y, and z directions acros.he node. Discontinuity
()
factors are used to modify the homogeneous cross sections to preserve the node surface fluxes and currents thet would be obtained from an cquivalent heterogeneous model.
ANC also contains a pin-power recovery model which couples the analytic solution to the two group diffusion equations with pin power information from PilOENIX P. ANC accurately reconstructs the results ofline mesh models using these methods. lleference 8 provides a detailed description of the methodology used in ANC.
ANC can be used to perform two or three-dimensional calculations with a wide variety of options. Geometries ranging from full core to octant are available with various symmetries. Feedback adjustments to macroscopic cross sections account for changes in fuel temperature and moderator density. Fuel and burnable absorber depletion and xenon and samarium buildup and decay are modeled.
Typical applications of the ANC code include the determination of:
Axial and radial power distributions, 103
e Differential and intergral control rod worths, e
Core reactivity coeflicients, o
Critical core configurations and shutdown snargins, and Fuel and barnable absorber loading patterns, G..t A P O I,l. O
)
APOLLO is a one.dirnensional two group steady state diffusion theory prograrn.
An APOLLO model is normally generated by radially homogenizing a three-dimensional ANC model. Cross sections are flux and volume weighted and a burnup and elevation dependent radial buckling search is performed to normalire the APOLLO model to ANC. Since a relatively large number of mesh points are permitted in APOLLO,it is used for applications which require a finer inesh in the axial direction. Typical applications include providing:
i l
Axial power distributions for Fq synthesis.
Differential and integral control rod worths, Trip reactivity curves, Load follow capability evaluations, and Control rod insertion limits.
Space dependent feedback effects due to xenon, samarium, rod position, boron, fuel temperature, and water density are accounted for in the calculations.
APOLLO is an advanced version of the PANDA code, which is described in Reference 12.
104
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