Text

Rev 0 to, Comm Ed Topical Benchmark of BWR Nuclear Design Methods
	ML20065Q853
Person / Time
Site:	Dresden, Quad Cities, LaSalle
Issue date:	11/30/1990
From:	Keffer J, Oster W, Schmidt R, Wieging J; NUCLEAR FUEL SERVICES, INC.
To:	;
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ML17202U908	List: ML20065Q853;
References
	NFSR-0085, NFSR-0085-R00, NFSR-85, NFSR-85-R, NUDOCS 9012180099
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{{#Wiki_filter:emf'5 iNuclear Fuel Services { ) COMMONWEALTH EDISON COMPANY TOPICAL BENCHMARK OF BWR NUCLEAR DESIGN METHODS BY JOHN W. KEFFER WILLIAM H. OSTER RANDALL R. SCHMIDT JOAN E. WlEGING "ove= Sea,' = .1 _._ pm p.:,:o <:,so.:,4j, Commonwealth Edison Company

I s ATTACHMENT-COMMONWEAL TH EDISON COMPANY TOPICAL REPORT BENCHMARK OF BWR NUCLEAR DESIGN METHODS (NFSR-0085, REVISION 0) o

1 NFSR-0085 Revision 0 Commonwealth Edison Company Topical Benchmark of BWR Nuclear Design Methods by John W. Keffer William H. Oster Randall R. Schmidt Joan E. Wieging 'I N/ Prepared By: ,/. k [Odt'- / O/81 %iR 14 9).c4 %t-m % y uu i %. d-3 Reviewed By: Nuclear Design Supervist.r Approved By: +C' - ///2//so Nuclear Fuel Servi @ Manager 'Date Nuclear Fuel Services Commonwealth Edison Company 72 W. Adams St., Room 922E Chicago, Illinois 60603

NFSR-0085 Revision 0 Statement of Disclaimer This document was prepared by Commonwealth Edison Company for filing with the United States Nuclear Regulatory Commission for the sole purpose of obtaining approval of Commonwealth Edison Company's BWR Nuclear Design Methods. Commonwealth Edison Company makes no warranty or representation and assumes no obligation, responsibility, or liability with respect to the contents of this report or its accuracy or completeness. Any use of or reliance on this report or the information contained in this report is at the sole risk of the party using or relying on it. 5 ii

NFSR-0085 Revision 0 Abstract This topical report summarizes the nuclear analysis methods employed by the l Commonwealth Edison Company (Edison) in support of reload design for its Boiling Water Reactors (BWRs). The nuclear analysis methods are based on the General Electric neutronic design computer codes, TGBLA and PANACEA, which have previously been reviewed and approved by the NRC. The results of an extensive benchmark program are presented to demenstrate Edison's ability to independently perform the nuclear analyses required for the licensing, operation, testing, and surveillance of a BWR reload cycle. The benchmark included a total of 18 unit-cycles, and comparisons were made to measured critical conditions and core power distributions. i l iii l \\

NFSR-0085 Revision 0 Table of Contents Page List of Tables vi List of Figures vii 1. Introduction and overview 1-1 1.1 Introduction 1-1 1.2 Overview of NFSR-0085, Revision 0 1-2 1.3 Scope of Analyses 1-2 1.4 Vendor Interactions 1-3 2. Summary and Conclusions 2-1 3. Neutronic Methodology 3-1 3.1 Lattice Physics Codes 3-1 3.2 Core Simulator Code 3-2 4. Neutronic Methods Validation 4-1 4.1 Validation of Lattice Physics Model 4-1 4.2 Validation of 3D Core Simulator 4-1 4.2.1 Hot Critical Eigenvalues 4-2 4.2.2 Cold Critical Eigenvalues 4-2 4.2.3 TIP Results 4-3 5. References 5-1 iv

i i NFSR-0085 Revision 0 i Table of Contents, Continued Page Appendices: 2 Appendix A - Fuel Bundle Nomenclature A-1 Appendix B - Statistical Basis for TIP Results B-1 Appendix C - TIP Traces-for Quad Cities Station Unit 1 C-1 Appendix D - TIP Traces for Quad Cities Station Unit 2 D-1 j Appendix E - TIP Traces for Dresden Station Unit 3 E-1 Apperdix F - TIP Traces for LaSalle County Station Unit 1 F-1 Appendix G - TIP Traces for LaSalle County Station Unit 2 G-1 9 ) 1 ( l j. v _ _ _ _.. _. ~. _. _ _. _ _ _ _ _. _... _ _ _..., _ _ _. _

1 4 irSR-0085 Revision 0 List of Tables Table Page l 1-1 Neutronic Parameters Required for Operation, 1-4 Testing, and Surveillance 2-1 Summary of Benchmark Results 2-2 3-1 Basic Neutronic Computer Codes 3-3 4.2-1 Rated Reactor Core Parameters 4-5 t 4.2-2 Fuel Loading Summary 4-6 4.2-3 Cold Critical Eigenvalue Results - 4-8 Quad Cities Station Units 1 and 2 j 4,2-4 Cold Critical Eigenvalue Results - 4-9 l Dresden Station Unit 3 4.2-5 Cold Critical Eigenvalue Resulte - 4-10 LaSalle County Station Units 1 and 2 4 4.2-6 TIP Re Quad Cities Station Units-1 and 2 4-12 1 4.2-7 TIP Resu. Dresden Station Unit 3 4-13 4.2-8 T!P Results - LaSalle County Station Units 1 and 2 4-14 l 4 l vi 9-m-1-y- ---y p 9_gye,u.w,g ,,,. _ _,q.e.yw. ..,,,y ,.,.y eiw., ,p.y% g + p. i, y.ycay,r-y

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i l NFSR-0085 i Revision 0 list of Figures l l Figure Page 4.2-1 Quad Cities 1 Cycle 7 - Pot Eigenvalues, Power, 4-15 and Flow 4.2-2 Quad Cities 1 Cycle 8 - Hot Eigenvalues, Power, 4-16 and Flow 4.2-3 Quad Cities 1 Cycle 9 - Hot Eigenvalues, Power, 4-17 and Flow 4.2-4 Quad Cities 1 Cycle 10 - Hot Eigenvalues, Power, 4-18 and Flow 4.2-5 Quad Cities 2 Cycle 7 - Hot Eigenvalues, Power, 4-19 and Flow 4.2-6 Quad Cities 2 Cycle 8 - Hot Eigenvalues, Power, 4-20 and Flow 4.2-7 Quad Cities 2 Cycle 9 - Hot Eigenvalues, Power, 4-21 and Flow 4.2-8 Quad Cities 2 Cycle 10 - Hot Eigenvalues, Power, 4-22 and Flow 4.2-9 Dresden 3 Cycle 8 - Hot Eigenvalues, Power, and Flow 4-23 4.2-10 Dresden 3 Cycle 9 - Hot Eigenvalues, Power, and Flow 4-24 4.2-11 Dresden 3 Cycle 10 - Hot Eigenvalues, Power, and Flow 4-25 4.2-12 Dresden 3 Cycle 11 - Hot Eigenvalues, Power, and Flow 4-26 4.2-13 LaSalle 1 Cycle 1 - Hot Eigenvalues, Power, and Flow 4-27 4.2-14 LaSalle 1 Cycle 2 - Hot Eigenvalues, Power, and Flow 4-28 4.2-15 LaSalle 1 Cycle 3 - Hot Eigenvalues, Power, and Flow 4-29 4.2-16 LaSalle 2 Cycle 1 - Hot Eigenvalues, Power, and Flow 4-30 4.2-17 LaSalle 2 Cycle 2 - Hot Eigenvalues, Power, and Flow 4-31 i i vii 1 1

s l NFSR-0085 l j Revision 0 List of figures, Continued Figure Page 4.2-18 LaSalle 2 Cycle 3 - Hot Eigenvalues, Power, and Flow 4-32 4.2-19 Quad Cities Station - Summary of Hot Critical 4-33 Eigenvalues 4.2-20 Dresden Station - Summary of Hot Critical-Eigenvalues 1-34 4.2-21 LaSalle County Station - Summary of Hot critical 4-35 Eigenvalues a 1 I 3 l viii

NFSR-0085 Revision 0 Section 1 - Introduction and Overview 1.1 Introductign This report summarizes the nuclear analysis methods employed by l Commonwealth Edison Company (Edison) in support of reload analysis for its l Boiling Water Reactors (BWRs) as well as the benchmark data which resulted l from the use of these methods. This report demonstrates Edison's capability to independently perform the steady-state neutronic analysis portions of the reload design process. This encompasses the neutronic analyses required for the the steady-state licensing, operation, testing, and surveillance of a BWR reload cycle. This benchmarking analysis includes generation of hot and cold critical eigenvalue data and a comparison of the predicted to measured Transverse incore Probe (TIP) readings. The hot and cold critical eigenvalue data demonstrates that the reactivity of the reactor is predictable and consistent. Accurate prediction of the TIP readings demonstrates the ability to predict the core power distribution. Results from the following units and cycles are summarized in this report: Quad Cities Station Units 1 and 2, Cycles 7 through 10; Dresden Station Unit 3, Cycles 8 through 11; LaSalle County Station Units 1 and 2, Cycles 1 through 3. This database includes fuel designs-from 7x7 to 9x9 fuel pin arrays, various water rod configurations, axially dependent lattice designs throughout the enriched portion of the assembly, and both General Electric (GE) and Advanced Nuclehr Fuels (ANF) fuel product lines. The design codes TGBLA and PANACEA, which were developed by GE, form the methodology basis for this benchmarking effort. Extensive design participation training was provided to five Edison engineers at the GE facilities in San Jose, California. This training included the performance of the full scope of neutronic calculations required for a reload design, use of the appropriate computer programs, and training in the acceptability and limitations of these computer programs for calculating neutronic parameters. Each training assignment lasted approximately one year. The overall neutronic design process employed by Edison is based on the approach described in GE Document NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel, Reference 1. Edison's fuel vendors will continue to perform the balance of plant transient and accident analyses. In the future, a separate report will be submitted suppor+.ing the use of Edison methods for performing these categories of calculations. Page 1-1 i w- ? rr

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t NFSR-0085 Revision 0 1.2 Overview of NFSR-0085. Revision 0. A summary of the results and conclusions reached from this benchmark effort is included in Section 2. These results demonstrate Commonwealth Edison's proficiency in the use of the GE neutronic code package. Section 3 includes a short description of the neutronic methodology used for this benchmark analysis. This methodology has already been approved by the NRC in GESTAR II, Reference 1, and by review of the GE topical on neutronic methods, Reference 2. Detailed results from the benchmark effort are in Section 4. The key parameters for the benchmark are hot and cold critical eigenvalue data, which demonstrate that the reactivity of the core can be reliably predicted; and comparisons of calculated to measured TIP data, which demonstrate that the core power distribution can be reliably predicted. Comparisons were made to eighteen unit-cycles, demonstrating the adequacy of the methods. 1.3 Scoce of Analyses Edison will perform the neutronic analyses required for the steady-state licensing, operation, testing, and surveillance of a BWR reload cycle. Specifically, the basic nautronic and processing computer codes described in Section 3 will be emp yed to generate lattice physics data, determine assembly loading pattern,, and calculate neutronic parameters for steady-state operation, testing, and surveillance. Table 1-1 provides examples of the type of neutronic parameters which are required for reactor operation, testing, and surveillance. Edison will evaluate these parameters for each BWR reload cycle for which it performs the neutronic design, regardless of the fuel vendor. An exception to this is the process computer information, which is the cycle specific input data for the onsite core monitoring code. Edison will provide the process computer information -only to those plants which are supplied with GE fuel. Edison will not currently be providing process computer information for non GE-supplied reloads in order to maintain compatibility with the critical power correlation and steady-state neutronics models used in the core monitoring software for non GE-supplied reloads. Therefore, Edison is not reque". ting NRC approval for application of the TGBLA/ PANACEA code package for process computer input for non GE-supplied reloads. The methodology and core conditions employed by Edison to perform tia neutronic design and licensing analyses are identitil to those employea by l l Page 1-2 l

NFSR-0085 Revision 0 General Electric (Reference 1). Edison's capability to perform the scope of analyses listed in Table 1-1 using the GE methodology is justified by the benchmark results outlined in Section 2 and detailed in Section 4 of this report. Until NRC approval is obtained, Edison will continue to i perform the majority of the;e analyses in parallel with the fuel vendor's analyses of record to ensure proficiency with the code package is maintained. J To summarize, Edison is requesting approval to calculate neutronic l parameters required for steady-state operation, testing, and surveillances, examples of which are li ted in Table 1-1, for all Edison BWRs, and this request is supported by the results of this report. 1.4 Vendor Interactions The fuel vendor will continue to be responsible for all balance of plant Anticipated Operational Occurrence analyses, such as pressurization-transients, and Loss Of Coolant Accident analyses. The basepoint for the evaluation of each cycle will be transmitted from Edison to the vendor for these analyses. Page 1-3

NFSR-0085 Revision 0 l Table 1-1 Neutronic Parameters Required for Operation, Testing, and Surveillance .l The following neutronic parameters are examples of the type of calculations required for reactor operation, testing, and surveillance. 1. Calculation of "R", the additional Shutdown Margin required at the beginning of the cycle to ensure the Minimum l Shutdown Margin requirement is met throughout the cycle. 2. Shutdown margin calculations for special conditions, such as out-of-service control rods. 3. Hot excess reactivity. 4. Data required for operation, such as target rod patterns, hot reactivity anomalies and core axial power distribution. 5. Data required for startup, such as control rod worths and criticality predictions. 1 6. Process Computer Input. This will be provided only to those units using a TGBLA/ PANACEA-based core monitoring code and will not be provided by Edison to those units using non GE-supplied fuel as part of the scope of this topical. l Page 1-4

.~.._-.. - 4 NFSR-0085 Revision 0 Section 2 - Summary and Conclusions The TGBLA/ PANACEA code package shows good agreement with the measured performance parameters of the benchmark data summarized in this report and the results are comparable to those calculcted by GE, which were reported in the GE acutronics methods topical, Reference 2. Specifically: 1. The predicted TIP response for the Edison BWRs shows good agreement with measured TIP data. 2. The calculated hot operating core eigenvalue is consistent between reactor types and shows little cycle exposure dependence. 3. The calctilated cold critical eigenvalue is consistent and predictable. Table 2-1 summarizes the benchmark results from the last two cycles for each unit. Bassd on the results, it is concluded that: 1. Commonwealth Edison has demonstrated proficiency with the TGBLA/ PANACEA code package and that Edison can perform in an acceptable manner all neutronic analyses required for the licensing, operation, testing, and surveillance of a BWR raload cycle; and 2. Commonwealth Edison is justified in its application of calculational uncertainties identical to those applied by GE. l l i i l Page 2-1 j

NFSR-0085 i Revision 0 Table 2-1 Summary of Benchmark Results Parameter Quad Cities Dresden laSalle Number of Unit-Cycles 4 2 4 Nodal TIP Power, 7.95 7.39 7.03 % Standard Deviation Radial TIP Power, 4.22 3.74 3.98 i % Standard Deviation Hot Critical Eigenvalue: Mean 1.0003 1.0041-0.9985 % Standard Deviation 0.14 0.17 0.21 Cold Critical Eigenvalue: Mean 1.0031 1.0062 1.0007 % Standard Deviation 0.21 0.17 0.25 i ( Page 2-2

NFSR-0085 Revision 0 Section 3 Neutronic Methodoloov The methodology used for the benchmark analysis for Edison's BWRs is based on the General Electric (GE) proprietary codes TGBLA, for lattice physics calculations, and PANACEA, for the three-dimensional (30) core simulation. The GE code package was written for the VAX computer system and was installed on a VAX system at Edison. After installation on the Edison system the code package eas validated and verified against a standard set of sample problems to ensure the results were consistent with those generated by GE. For reference purposes, the major codes in the GE code package are listed in Table 3-1. The NRC has reviewed and approved the methodology of this code package. This approval and a detailed discussion of the methodology can be found in References 1 and 2. 3.1 lattice Physics Codes The GE codes TGBLA and GEllB (References 3 and 4, respectively) are used to generate the required lattice physics data for the GE core simulator code PANACEA (Reference 5). The lattice physics code TGBLA generates the required lattice physics data for the benchmarking effort. TGBLA assumes an infinite lattice, or zero current, bundle configuration and uses a combination of diffusion and transport theory to determine the lattice characteristics. TGBLA calculates the rod-by-rod thermal spectra by the leakage-dependent integral transport method. Leakage iterations between diffusion theory and thermal spectrum calculations are carried out to generate thermal broad group difrusion parameters. Thermal broad-group neutron cross-sections are calculated for homogenized fuel rod cells using a condensed sixteen coarse group thermal cross-section library. In the epithermal and fast energy range, the level-wise resonance integrals are calculated by an intermediate resonance approximation in which the intermediate resonance parameters are fuel rod temperature dependent. Cross-sections for the fast and epithermal region are calculated using a 68 fine group cross section library. Two-dimensional, coarse mesh, broad group, diffusion theory calculations are then used to determine the nodal flux and power distributions in the BWR lattice. GEllB is a data manipulation code which accesses the cross-section and reactivity output from TGBLA and calculates the reactiv:ty fits required for PANACEA. In this manner the results of the TGBLA calculations are Page 3-1 l

l NFSR-0085 i Revision 0 reduced to libraries of lattice reactivities, relative rod powers, and few group cross sections as a function of instantaneous void, exposure, exposure-weighted void history, control state, and fuel and moderator temperature for use in the PANACEA calculations. 3.2 Core Simulator Code PANACEA is a three-dimensional, coarse mesh, one group, coupled nuclear / thermal-hydraulics computer program for analyzing a BWR core. The coarse mesh width is on the same order of magnitude as the fast neutron mean free path. When coupled with the one group assumption, this mesh width is considered adequate as the global neutron flux shape is primarily determined by the diffusion of fast neutrons. A seven-point difference scheme is used to solve the one group diffusion equation. At the boundary the difference equation is modified to include reflector effects. PANACEA uses k-infinity fits to calculate nodal core reactivity. These fits are quadratic in void history, and table look-up in exposure. LaGrangian interpolation is used to calculate intermediate exposure values. There is a different set of base k-infinities for both the hot and cold conditions. These base k-infinities are corrected for controlled conditions, using a controlled to uncontrolled ratio; for instantaneous void effects; for various neutron poisons; and for Doppler effects. Doppler is represented using an effective fuel temperature, which is based on the nodal power. l Page 3-2 l l

1 NFSR-0085 Revision 0 Table 3-1 Basic Neutronic Computer Codes (sde Name Descriotion TGBLA Macro-and Microscopic Lattice Cross Section Generator .GELIB Processing Code to Prepare TGBLA Data for Use by PANACEA PANACEA Three Dimensional Spatially Dependent, One Group Core Simulator i i Page 3-3 l . ~

l NFSR-0085 Revision 0 Section 4 - Neutronic Methods Validation The methods validation data for the Edison BWR calculational methods are contained in this section. Comparisons between calculations and measurements are presented for Quad Cities Units 1 and 2, Dresden Unit 3, and LaSalle County Station Units 1 and 2. 4.1 Validation of lattice Physics Model i Edison used the lattice physics code TGBLA to create the neutronic inputs for the downstream core simulator validation. The methodology in TGBLA has been previously approved by the NRC for use on BWRs. This approval is documented in the GE Licensing Topical, Reference 2. As part of this topical, GE validated the core power distribution predictions based on global gamma scan measurements from Quad Cities Unit 1, Cycles 2, 4, and 5; Hatch Unit 1, Cycles 1 and 3; and Millstone Unit 1, Cycle 7. The isotopic burnup was validated against measurements from Quad Cities 1 Cycle 2. The ultimate test of the lattice physics model is demonstrated by the validation of the Edison units using historical cycles. This process demonstrates Edison's capability to determine the neutronic parameters for current fuel product lines and core loadings. Details of this validation are contained in Section 4.2. 4.2 Validation of 30 Core Simulator The PANACEA reactor core simulator code is verified by comparing calculated and measured reactor parameters. Core follow calculations were performed for the Quad Cities, Dresden, and LaSalle reactors and the results were compared to measured data. Pertinent reactor core parameters at rated operating conditions for the units are given in Table 4.2-1. The fresh fuel loaded in the cycles of interest is summarized in Table 4.2-2. j Appendix A describes the nomenclature for the bundle names in Table 4.2-2. It should be noted t'st the initial core state conditions, namely the nodal exposure and void history data, evaluated for the Quad Cities and Dresden Units were provided by GE and were generated using GE's GENESIS methods. These methods have since been replaced with the TGBLA/ PANACEA l GEMINI methods. The use of data which were generated using GENESIS l methods and supplied by GE as the initial core conditions for this l benchmark effort results in anomalous effects due to the use of two discrete methods for the early transitional cycles. These anomalous effects show up as high TIP standard deviations and eigenvalue l Page 4-1 1 ,,._,,_-__.,,,.-y

F4 s NFSR-0085 Revision 0 uncertainties until the transitional effects of the initial core conditions disappear; these effects also are seen in the transitional cycles modelled by GE with their GEMINI methods. This is illustrated by the large decrease in calculated versus measured TIP standard deviations from the early transitional cycles to the more recent cycles. 4.2.1 dat Critical Eioenvalues The hot critical eigenvalue results as a function of cycle exposure are shown in Figures 4.2-1 through 4.2-4 for Quad Cities Unit 1: Figures 4.2-5 through 4.2-8 for Quad Cities Unit 2; Figures 4.2-9 through 4.2-12 for Dresden Unit 3; Figures 4.2-13 through 4.2-15 for LaSalle County Unit 1; and Figures 4.2-16 through 4.2-18 for LaSalle County Unit 2. The hot critical data have not been corrected for reactivity biases associated with the effects of channel bow, crud, incore instrumentation, and fuel assembly spacers. Hot critical eigenvalues should be consistent and predictable as a function of cycle exposure to enable the engineer to develop adequate projections of the critical eigenvalue for upcoming cycles. The hot critical eigenvalues developed as part of this benchmark effort are shown to be a strong function of plant type and fuel product line being loaded but are consistent within these parameters. A comparison of Figures 4.2-19 through 4.2-21 demonstrate this. The hot critical eigenvalues from the last two cycles of each unit in the database used for this benchmark are shown as a function of exposure on these figures. Quad Cities data is contained on Figure 4.2-19; Dresden on Figure 4.2-20; and LaSalle on 4.2-21. The trends demonstrated in these figures will be used to pr(dict the hot critical eigenvalue for future cycles. 4.2.2 Cold Critic,1 Eioenvalues The cold crit: cal eigenvalue results are shown in Table 4.2-3 for Quad Cities Station; in Table 4.2-4 for Dresden Station; and in Tsbie 4.2-5 for LaSalle County Station. All cold critical data are from in-sequence, xenon-free startups. The cold critical data have not been corrected for reactivity biases associated with the effects of channel bow, crud, incore instrumentation, and fuel assembly spacers. Corrections have been made for reactor period and temperature at the time of criticality, as the lattice physics data were generated assuming steady-state Page 4-2 m........ -" - - - - ~ - ' ' - - - - " - - ^ ' ' " - - " ' - - - - '

i 1 NFSR-0085 Revision 0 i conditions cnd a moderator and fuel temperature of 20 degree C. As stated in Section 4.2.1 for the hot critical eigenvalues, the l cold critical eigenvalues must also be consistent and predictable as a function of exposure, as these are required to calculate core subcriticality. Cold. critical eigenvalues have a larger degree of scatter as a function of exposure than hot critical eigenvalues, but also are consistent and predictable as a function of plant type and i fuel being loaded in the core. Therefore, core suberiticality can be assured under all conditions. i 4.2.3 TIP Results Measured and calculated Traversing Incore Probe (TIP) data have been compared and are summarized in this section. The TIP standard deviations were calculated over the entire axial length of the core, which is 24 nodes. A detailed discussion of the numerical basis for the TIP standard deviations and asymmetries is included in Appendix B. The TIP standard deviations and asymmetries for each statepoint are shown in Table 4.2-6 for Quad Cities Station; in Table 4.2-7 for Dresden Station; and in Table 4.2-8 for LaSalle County Station. TIP asymmetry date is included to provide a method for evaluating the adequacy of the TIP standard deviations. As discussed previously, there are high TIP standard deviations in the early transitional cycles for Quad Cities and Dresden due to the t initial core conditions reflecting the transition from the earlier GE GENESIS models to the current GEMINI models. The TIP standard deviations for the last two cycles of each plant are acceptably low, as nodal TIP standard deviations are less than 10% and radial standard deviations are less than 6% for all_statepoints. These values demonstrate that the core power distribution is being adequately predicted by the Edison models. i h Page 4-3

~. i i 1 i NFSR-0085 I Revision 0 2 Detailed TIP results are contained in the appendices as follows: i l Unit Cvele Accendix Fiaure l QCl 7 C C-1 QCl 8 C C-2 001 9 C C-3 l QCl 10-C C-4 i QC2 7 D D-1 l QC2 8 0 D-2 QC2 9 D D-3 QC2 10 0 0-4 DR3 8 E E-1 DR3 9 E E-2 DR3 10 E E-3 4 DR3 11 E E-4 LSI 1 F F-1 LSI 2 F F-2 LSI 3 F F-3 LS2 1 G G-1 LS2 2 G G-2 LS2 3 G G-3 i l 1 l l Page 4-4 1 4 .___m-m--.-. - -,---., s -. - - - - - - ~, -.. ~. - -. ~ ~., ,-,y.,--..--.-.~w,-.. --,,..--~-y--<- ..v.a =w-

i NFSR-0085 Revision 0 Table 4.2-1 Rated Reactor Core Parameters Ouad Cities Station Units 1 and 2 Thermal Power, MWt 2511 Core Flow, Mlb/hr 98.0 Inlet Subcooling, Btu /lbm 524 Core Midplane Pressure, psia 1035 Total Assemblies in Core 724 Average Power Density, kw/l 41 Dresden Station Units 2 and 3 Thermal Power, MWt 2527 Core Flow, Mlb/hr 98.0 1 l Inlet Subcooling, Btu /lbm 524 Core Midplane Pressure, psia 1035 Total Assemblies in Core 724 Average Power Density, kw/l 41 LaSalle County Station Units 1 and 2 Thermal Power, MWt 3323 Core Flow, Mlb/hr -108.5 l Inlet Subcooling, Btu /lbm 528 l Core Midplane Pressure, psia 1035 Total Assemblies in Core 764 Average Power Density, kw/1 50 Page 4-5 a

NFSR-0085 Revision 0 Table 4.2-2 Fuel loading Summary Mn11 Cycle Fuel Tvoe loadqd Number loaded 001 7 GE6-P80RB265-6G2.0-80M-145 64 GE6-P80RB265-6G3.0-80M-145 160 8 GE78-P80RB265-6G3.0-80M-145 116 GE7B-P80RB283-7G4.0-80M-145 80 9 GE7B-P80RB299-7G4.0-80M-145 144 GE78-P8DRB282-7G3.0-80M-145 72 10 GE88-P8DQB300-7G4.0-80M-145 120 GE88-P80QB300-9G4.0-80M-145 80 QC2 7 GE78-P80RB265-6G3.0-80M-145 204 8 GE7P 3DRB282-7G3.0-80M-145 72 GE7 8DRB283-7G4.0-80M-145 104 9 GE78-P80RB299-7G3.0-80M-145 88 GE7B-P8DRB299-7G4.0-80M-145 64 10 GE8B-Pb708300-9G3.0-80M-145 92 GE8B-P8D98316-7G4.0-80M-145 72 DR3 8 ANF-P8 DEB 269-5G3.0-80M-145 224 9 ANF-P8 DEB 203 5G3.5-80M-145 184 10 ANF-P9DNB313-8G4.0-80M-145 160 ANF-P9DNB313-9G4.0-80M-145 16 i 11 ANF-P90NB313-9GZ-80M-145 72 ANF-P9DNB313-9GZl-80M-145 96 Page 4-6

i i NFSR-0085 Revision 0 I Table 4.2-2, Continued i Fuel Loading Sumary Unit. [y.qlg Fuel Tvoe loaded Number Loaded i LSI 1 CCIB071-NOG-100M-150 92 8CIB176-4GZ-100M-150 240 BCIB219-4GZ-100M-150 432 3 l 2 GE78-P8CRB299-6G3.0-100M-150 232 ? 3 GE88-P8 COB 301-8GZ-100M-150 112 GE8B-POCQB320-9GZ-100M-150 112 l LS2 1 8CIB071-N0G-100M-150 92 8CIB176-4GZ-100M-150 240 3 8CIB219-4GZ-100M-150 412 ? 2 GE78-P80RB299-6G3.0-100M-150 224 3 GE8B-P8CQB300-6G3.0-100M-150 144 GE88-P8CQB320-7GZ-100M-150 96 q l i l Page 4-7

NFSR-0085 Revision 0 Table 4.2-3 Cold Critical Eigenvalue Results Quad Cities Station Units 1 and 2 Cycle Cold

Exposure, Critical Unit Cycle mwd /St Eigenvalue-QCl 7

0 1.0058 0C1 7 1211 1.0070 001 7 3927 1.0049 0C1 8 0 1.0041 001 8 4143 1.0015 QCl 8 7115 0.9999 001 9 0 1.0037 001 9 3400 1.0018 001 9 3713 1.0016 0C1 10 0 1.0067 001 10 2123 1.0056 0C1 10 8166 1.0035 QC2 7 0 1.0070 QC2 7 1033 1.0075 QC2 7 4713 1.0043 QC2 8 0 1.0050 QC2 8 4064 1.0037 i QC2 8 5538 1.0025 QC2 9 0 1.0040 002 9 3083 1.0021 QC2 9 3759 1.0016 i 002 10 0 1.0053 QC2 10 316 1.0035 QC2 10 3462 1.0028 QC2 10-6019 0.9986 Page 4-8

NFSR-0085 Revision 0 Table 4.2-4 Cold Critical Eigenvalue Results Dresden Station Unit 3 Cycle Cold

Exposure, critical Unit Cycle NWd/St Eigenvalue DR3 8

0 1.0056 DR3 3 2815 1.0023 DR3 9 0 1.0068 DR3 9 0 1.0076 DR3 9 841 1.0063 DR3 9 1107 1.0065 DR3 9 3888 1.0040 DR3 9 6072 1.0033 DR3 10 0 1.0078 DR3 10 1179 1.0066 DR3 10 2508 1.0057 DR3 10 4353 1.0032 DR3 11 0 1.0086 DR3 11 4139 1.0055 i l Page 4-9 l

NFSR-0085 ) Revision 0 Table 4.2-5 Cold Critical Eigenvalue Results LaSalle County Station Units 1 and 2 i Cycle Cold

Exposure, Critical Unit Cycle mwd /St Eigenvalue LSI 1

0 1.0041 LSI 1 0 1.0035 LSI 1 0 1.0041 LSI 1 0 1.0050 LSI 1 485 1.0037 LSI 1 574 1.0036 LSI 1 574 1.0038 LSI 1 574 1.0040 4 LSI 1 897 1.0028 LSI 1 1781 1,0015 1.Sl 1 1781 1.0021 LSI 1 2125 'l.0014 LSI 1 2595 1.0006 LSI 1 2595 1,0005 LS1 1 6002 0.9969 LSI 1 7902 0.9978 i LSI 1 8872 0.9966 LSI 1 9321 0.9960 LSI 2 0 1.0033 LS1 2 0 1.0032 LSI 2 2664 1.0002 LSI 2 3477 0.9979 LSI 3 0 1.0020 LSI 3 182 1.0007 LS2 1 0 1.0041 LS2 1 0 1.0040 LS2 1 0 1.0379 LS2 1 843 1.0029 'LS2 1 2665 1.0010 LS2 1 4042 1.0000 LS2 1 5715 0.9984 LS2 1 8198 0.9978 LS2 1 8310 0.9984 Page 4-10

NFSR-0085 Revision 0 Table 4.2-5, Continued Cold Critical Eigenvalue Results LaSalle County Station Units 1 and 2 t Cycle Cold

Exposure, Critical Unit Cycle mwd /St Eigenvalue LS2 2

0 1.0038 LS2 2 5182 0.9963 LS2 3 0 1.0011 LS2 3 3884 0.9986 I 1 l Page 4-11

l NFSR-0085 Revision 0 Table 4.2-6 TIP Results Quad Cities Station Units 1 and 2 Cycle Exposure, Standard Deviations i Unit Cycle mwd /St Radial Nodal l 0C1 7 96 5.4 13.2 0C1 7 3146 6.2 17.1 Q01 7 6362 5.1 14.6 QCl 8 521 6.0 12.7 001 8 3931 6.2 12.4 QCl 8 7370 5.8 9.0 QCl 9 599 4.7 8.4 001 9 3526 5.0 8.0 0C1 9 7664 4.8 9.1 0C1 10 1026 5.1 8.1 001 10 4347 5.1 8.1 - QC1 10 6415 4.7 7.7 QC2 7 273 10.0 15.9 QC2 7 3970 9.5 16.0 3 QC2 7 5496 9.n 17.1 QC2 8 630 4.8 9.6 QC2 8 4195 4.5 7.1 QC2 8 7327 4.5 8.0 QC2 9 129 3.5 6.9 QC2 9 3245 4.3 6.9 QC2 9 5542 3.4 7.3 QC2 10 891 3.0 8.7 QC2 10 3315 3.2 8.0 QC2 10 6974 3.5 7.9 i i l Page 4-12 w,- w

_ _ _ _. _ _ _ _ _. _ ~ NFSt-0085 Revision 0 Table 4.2-7 TIP Results Dresden Station Unit 3 Cycle Exposure, Standard Deviations Unit Cycle mwd /St Radial Nodal DR3 8 1001 5.45 9.69 DR3 8 4048 6.38 10.42 DR3 8 6148 6.37 8.72 DR3 9 314 6.05 10.11 DR3 9 3247 5.91 8.21 DR3 9 4630 6.83 9.79 DR3 10 1287 3.21 7.00 DR3 10 4855 3.34 6.67 DR3 li 166 4.33 7.65 DR3 11 3237 4.15 8.18 l l Page 4-13 l

NFSR-0085 Revision 0 Table 4.2-8 TIP Results LaSalle County Station Units 1 and 2 Cycle Exposure, Standard Deviations Unit Cycle mwd /St Radial Nodai LSI 1 1001 2.01 6.58 i LS1 1 1605 1.94 7.37 LSI 1 4934 3.34 6.22 LSI 1 6293 3.33 6.96 LSI 1 9395 3.63 6.42 LSI 1 10173 3.39 6.79 LSI 2 1032 4.33 7.75 LS1 2 3375 3.30 6.38 LSI 2 5406 3.16 6.18 LSI 3 1098 3.85 6.44 LSI 3 4167 3.50 6.12 LSI 3 7493 4.12 6.10 LS2 1 704 1.73 9.82 LS2 1 3200 2.69 6.84 LS2 1 9269 2.68 6.51 LS2 2 657 5.11 7.87 LS2 2 3257 3.60 8.94 LS2 2 6732 4.36 6.39 LS2 3 194 3.08 7.59 LS2 3 4090 3.77 6.68 LS2 3 6837 5.36-7.31 ( i l Page 4-14 )

i 2 Quad Cities 1 Cycle 7 ^ Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 g 1.005 7e...i 100 {

gy-7, k= +==-gp q=7a< ip -- ^ -

\\ s L ~.: ~ ~ = 1 s e 1.000 ~4- + + -- - + - - - - - - - + 80 t 8 g 1 0.995 +- - - ?- - -+ - - ~ 60 i i i i i 0.990 40 O 1 2 3 4 5 6 7

a

-n Cycle Exposure, GWD/STU na 88 Eigenvalue -- Thermal Power - - - - Flow.

Quad Cities 1 Cycle 8 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 I 100 "E i 1.006 , yy'( fyj .;==z y-=:~ p==ux ~;, - >; 7'==~, - - m ,,',i.*,

c. i v

,,n y ~ y ~ m 1.000 --i ~ - 80 0.995 60 i 0.990 40 n2 0 2 -4 6 8

a

-w l Cycle Exposure, GWD/STU r e-88 o Eigenvalue Thermal Power

Flow i

T ---"'T' - + ' -. r' '5 g I-i1

Quad Cities 1 Cycle 9 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120' f,, ,p<~ = g y ;f ,r= z. ji=====;..r== => yew 'ey=-x 100 j 1.00G y s_/ g sj = ~ u ~ M - --- 80 1.000 + + 60 0.995 4-- - - - ~ - - - - - i i i i 0.990 40 0 2 4 6 8 10 {3. Cycle Exposure, GWD/STU E! 88 m o Eigenvalue Thermal Power

Flow

W Quad Cities 1 Cycle 10 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 $r' 100 i 2' 1.005 - r -- r . -,2-T T ti- /- = 7,'N _,,,f"ig' 0'> j, 7 g .,. s s

, //

1 1.000 + 80 0.995 + -+ 60 I i 0.990 40 0 2 4 6 8 10 @'3 Cycle Exposure, GWD/STU EZ 88* o Eigenvalue --- Thermal Power

Flow

=

Quad Cities 2 Cycle 7 Hot Eigenvalues, Power, and Flow a Eigenvalue % of Rated 1.010 120 i =---y.:7_ a ; _ jece-100 { i 7 1.005 w ce f ______ /_ - N 2 4. ? 7 1.000 t "- - i - + 80 1 I 0.995 i' - - - 60 i i 0.990 40 0 1 2 3 4 5 6 24. <w Cycle Exposure, GWD/STU

P E8

'a Eigenvalue Thermal Power

Flow t

m Quad Cities 2 Cycle 8 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 2 2 1.005 pre- . : _ y,;, _ _ _ _ :_ ~ ~ ;'; -, . = ; r-q. c;' ' ;. 100 t i g' h ~ 1.000 + + g-80 0.995 ~ + 60 a i i 0.990 40 I 0 2 4 6 8 {3 Cycle Exposure, GWD/STU Eg 88 O Eigenvalue Thermal Power

Flow

ki t 20wa,8" m2 n8 o , y.. - ~ 0 0 2 0 0 0 0 1 1 8 6 4 7 d e ta R wf w o

Y, 6

o o

3,,-

- s l l F F s 9d n c, 5 ea U l T r c S e yr z. / w e .f D o Cw ts P W l 2 o 4G a m P e r s r e u h e s s c o T i t e f: 3p i u x Cl

c E

a e d v r 'c an O 2 y e C u ue i ;_ - la Qig v s n E e g i t e E g: o u 1 i l H av n e g i E O 0 5 0 5 0 1 0 0 9 9 0 0 0 9 9 1 1 1 0 0 y;;.4-i

,b l e =:na 8u. x* n 8, , (., ~ & 0 0 2 0 0 0 0 1 1 8 6 4 0 d 1 e ta R wf o w o o l lF .5 F [ .i i 8 ~ ~ 0d [ 1 n a U e T r l c r, S e / w ye D o Cw 3,,,' P 6W - l l o lr G a 2 's m P s e r s s' e ru h s es o T ~ e i i p t u x i ?1,j ; i 4E Cl .- ',5 .i' a e e dv yi,- l c a n y e C u e s', u l g a t Qi vn E w i i 2 qe e 1 g i t e E o u .) H a l vn e g

.';, l i

E t S : 0 0 5 0 5 0 1 0 0 9 9 0 0 0 9 9 1 1 1 0 0

j.,h" i

P Dresden 3 Cycle 8 Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 I J 1.005 , - - - - ~ - - + =a,.-- - _ _ _ _ _us!

=e-T

,--,-P~~------ 100 e 2 m e i c 8 'V! ~ 7 l 1.000 -l 4 + 2 -- '- -~ - 80

i8j, it 8

f 0.995 -F +- ' i- - 60 i a a O.990 40 0 2 4 6 8 F M, - w, ' v. O o ?- y Q-Q ,..n._.....;... yg 3 s e co o 't i a CL n v> c g o y g cn c. M 3 X "U N W g s @q) _3 L'_= o-L Q i o C> .. f... gC 3 C L Q t ),, c .O LL,l 3 .!.$..p.. e W >m

o 1

g 8 l l b en W i i o o to o to o t-- o o c) cn O. O. o. 0) c) r r-e o o Page 4-24

1 Dresden 3 Cycle 10 Hot Eigenvalues, Power, and Flow-Eigenvalue % of Rated ^ 1.010 120 _e a' 1.005 -r 100 'i -l '--+8 ~.., i c -~.;.,,.'~L.),_ ;, ~~. y, e 4 6 sf -t s r 1.000 --3 -i- ' - + - - 80 s i, i. i, s t is e cs O.995 + + ". +;, 60 8, u, r i j 0.990 40 l mz 0 2 4 6 8 D -w Cycle Exposure, GWD/STU 1A 38 o Eigenvalue Thermal Power

Flow t

S Dresden 3 Cycle 11 Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 --l--- 5 100 g-y 1.005 t - y;., 7r+- 3 my 7.__.___ O / .i %? \\1

'~,

A _.. \\. p g m g 8-- '-'*4s p_ f., 0 / 4 i ts/ .c ~ L \\ .e i m i t N s. l 1.000 fl~ il- + 80 '\\

i

!i l k' i, i 0.995 ', t- -f - 60 '.,l L 0.990 40 nz O 2 4 6 8 ea -n Cycle Exposure, GWD/STU na 88 o . Eigenvalue Thermal Power - --- Flow -

t LaSalle 1 Cycle 1 Hot Eigenvalues, Power, 'and Flow Eigenvalue % of Rated 1.010 120 1 1 2 2 1.005 -i pa==d:-:n x,- , ', - -,; ' n ; -^-r-~V2 100 O lj a n (' / ig p / j ,t s s ti Y* s- . 's i sy -; \\ i 4 s '15 Is

i

\\ ~ ;' t t\\ (s s

g f

i sI s \\, l4 s,lI g) f 9 h l1 ~ e, 0 I \\1 s Sy: p i 1.000 -' F -+ "- '7 - - - - 80 1 I i i 0.995 -f- + 60 i i i i 0.990 40 0 2 4 -6 8 10 12 'a 0 m -n Cycle Exposure, GWD/STU ra 88 o '" Eigenvalue Thermal Power

Flow i

LaSalle 1 Cycle 2 Hot Eigenvalues, Po.wer, and Flow Eigenvalue % of Rated 1.010 120 t y, = iry- - - 100 3 I 1.005 t + + a - e., ~ ', = s., i f> ;' 4,' t -n. ,,+~~2 M i -I, Y,!~ ~ - + - - 80 ^ 1.000 ..8 ,1 l .i .f g( i 1, i' 'i N 0.995 +, h m ",- ~+- = = ~ - - 60 t i i i i 0.990 40 0 1 2 3 4 5 6 7. $3 Cycle Exposure, GWD/STU E5 88 o '" Eigenvalue - Thermal Power -- -- Flow

i 1 ,lll', l\\i1!liil 1 P za=a8* w%'8o l e 4= ". ~L m e 0 0 2 0 0 0 0 1 1 8 6 4 0 d 1 e ta R wf o w o o l l F F l' 8 .J d . rj6i = l> t ; l, = n 3a 7 i,' 3~ U T r S e er f / w e D o l c yw i, 6W P i. r i.< ;,T i ',. l G a Co m P ,e r r e 1 u h s s e o T e p l ~- x lalu .:t 4E Sa e a v lil lc L n y e e ' cp'i,,i",- C u ?* l . l, a t g N:;\\, - ', ' i v i n E e

t 2

g ..v i t e E o u f'., l H a i v c_' ne g i E 0 0 5 0 5 0 1 0 0 9 9 0 0 0 9 9 1 1 1 0 0 ,gm #Ae .l i 4 i 2,

i LaSalle 2 Cycle 1 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 2 ,s :, y 7 - - - - 100 3 A 1.005 s #.:! -- i n.

4 -

--L++ - -=v " + :r a ii = w T 'q l \\\\

i. ' ' i l t

b ',) F- ~ ~ - 80 - ~ h 7' ' - ~ 1.000 ,f If

- i l

60 0.995 -ih* + ( I i i i l 0.990 40 mz O 2 4 6 8 10 12

?

Cycle Exposure, GWD/STU Eo? 88 o Eigenvalue Thermal Power

Flow

~

LaSalle 2 Cycle 2 ~ Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 t z' e i 1.005

7-er -e v

100 4 4- ~ ', ; = =a,- % n: r w=s -== s

;r -

z,<.. y ?.... = s i i (.' i, v s i O

5. l

". 8' 4 ':'

f

',.ls 7 s 'i;, ~ s 1.000 -+ Y ^t - n 80 p, i e I 0.995 - - - + - 60 .f i i 0.990 40 mz*3 O 2 4 6 8 10 1 Cycle Exposure, GWD/STU Eg =' 8 O Eigenvalue Thermal Power

Flow

LaSalle 2 Cycle 3 Hot Eigenvalues, Power, and Flow Eigenvalue % of Rated 1.010 120 Y A 1.005 ~ -i~-bF--- 'x - 4 ' ~ h 5' ' i' ' ' ' J' ^v==d' 100 k .{ e,l l' g ~ ~ = u L, l 'j; ~ 'n y 1.000 4 ~ + - - - -- 80 ^ -~ 0.995 -+- 60 i i i i i 0.990 40 0 1 2 3 4 5 6 7 E3 Cycle Exposure, GWD/STU EI 8 8.. ,u Eigenvalue Thermal Power

Flow

Quad Cities Station Summary. of Hot Critical Eigenvalues 1.010 i I 1.005 /+ + p++ +++ a[t. 4,. 4 ~E. [+ + Moqg-t 9-1.000 + o dD D 9 O a CD w g9 G i 0.995 l i j l i i 0.990 0 2 4 6 8 10 Cycle Exposure, GWD/STU Unit 1 Cycle 9 + Unit 1 Cycle 10 kh Ci' Unit 2 Cycle 9 0 Unit 2 Cycle 10 g8 => m o

t Dresden Station Summary of Hot Critical Eigenvalues 1.010 k -t b tF:+ +4 I j+ +-H-4 1.005 + 3 k.y++g ; y y y g a 1.000 t A i b 0.995 r .l i 0.990 O 2 4 6 8 10 Cycle Exposure, GWD/STU EM <m ~

T Unit 3 Cycle 10

+ Unit 3 Cycle 11 r8 om O a 4

4' LaSalle County Station Summary of Hot Critical Eigenvalues i 1.010 1.005

+&+q,\\

i 5 1.000 ~ Ni,+#h*,++ ggg,b65Y p+% %qpop ~ + o t e a (p**f*+, t.+.... + 0.995 i i' 0.990 O 2 4 6 8 10 Cycle Exposure, GWD/STU %'35 Unit 1 Cycle 2 + Unit 1 Cycle 3 -}y -o O Unit 2 Cycle 3 + Unit 2 Cycle 2 _ - 1

NFSR-0085 Revision 0 Section 5 - References 1. GE Proprietary Document NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel", as supplemented. 2. GE Proprietary Document NEDE-30130-P-A, " Steady-State Nuclear Methods", April 1985. 3. R. T. Chiang and M. Yamamoto, GE Proprietary Document NEDE-30002P, "TGBLA Lattice Physics Methods", December 1982. 4. GE Document NEDE-25337, Revision 2, "An Introduction to the GELIB System", October 1982. 5. C. L. Martin and F. Rahnema, GE Proprietary Document NELF-20884P, " PANACEA - BWR Core Simulator", March 1988. l o l Page 5-1

NFSR-0085 Revision 0 Appendix A Bundle Nomenclature The bundle nomenclature used in this report is best described by an example. Bundle Name - GE88-P80Q8300-7G4,0-80M-145 Where: GE8B - Fuel Product Line Identifier Options: ANF = All ANF Fuel Product Lines l GExB = GE Fuel Product Line x, with or without Barrier (B) l P - Pre-pressurized fuel. Options: P - Pressurized Greater than 1 atm. No Entry - Unpressurized (1 atm) 8 - Lattice Array Size Options: 7 = 7x7 Lattice Array 8 - 8x8 Lattice Array 9 - 9x9 Lattice Array 0 - Lattice Type Options: D = D-lattice (Non Uniform Water Gaps) C = C-lattice (Uniform Water Gaps) Q - Lattice Product Line Options: I = GE Initial Core Fuel l R = GE Reload Fuel (Prior to GE8) Q = GE8 Reload Fuel E = ANF 8x8 1 Water Rod Reload Fuel N = ANF 9x9 2 Water Rod Reload Fuel B - Indicates this is a Bundle Name, Not a Lattice Name 300 - Average enrichment of bundle, w/o V235

100 Page A-1

NFSR-0085 Revision 0 7G4.0 - Cadolinia Loading 1 7 - Maximum Number of Gadolinia pins in any layer 4.0 - Gadolinia Concentration. This is set to "Z" if the bundle has axial ~fy varying gadolinia. 80M - Channel Thickness 80M - 80 mil thickness (Quad Cities ~and Dresden) 100M - 100 mil thickness (LaSalle) 145 - Fuel Length, inches 144 or 145 - Quad Cities and Dresden 150 - LaSalle l l l Page A-2

NFSR-0085 l Revision 0 l Appendix B Statistical Basis for TIP Results l Several standard' deviations are presented in_the. text. The bases for these calculations are discussed in this appendix. Nodal Standard Deviation The nodal standard deviations are based on the nodal percent differences for each TIP reading in the core. The nodal percent differences are calculated as follows: Nodal Perce.,t Difference - ( C - M ) / Mbar

100 where:

C - Calculated TIP Reading M - Measured TIP Reading Mbar - Average of all TIP Readings Since the measured data are normalized such that the core average value is one, Mbar will always be equal to one and the above equation simplifies to the difference between the calculated and measured values. The nodal standard deviations were calculated using the nodal percent differences calculated in this way for the entire axial length of the core, or 24 nodes. This is more conservative than using 20 node data, as the low power nodes in the core top and bottom are more difficult to predict, but are given equal weighting in the calculation. Radial Standard Deviation The radial standard deviations are based on the radial percent differences for each TIP string in the core. The radial percent differences'are the differences between the summations of the calculated TIP readings and the measured TIP readings for each TIP string. Again, since the measured data are normalized such that the core average value is one, the normalization factor l Mbar will always be equal to one. l i The radial standard deviations were calculated using the radial pe~rcent l differences calculated in this w&y for the entire acial length of the core, or i 24 nodes. 1 Page B-1 I w a ~ < ~ -w

NFSR-0085 Revision 0 Appendix C TIP Traces for Quad Cities Station Unit 1 A comparison of calculated versus measured core axial TIP traces for Quad Cities Station Unit 1 is included in this appendix. Figures included are: Figure C Quad Cities Station Unit 1 Cycle 7 Results Figure C Quad Cities' Station Unit 1 Cycle 8 Results Figure C Quad Cities Station Unit 1 Cycle 9 Results Figure s 4 - Quad Cities Station Unit 1 Cycle 10 Results l I Page C-1 ~ - - - < -,,-..,n-

-1 NFSR-0085 Revision.0 1 Figure C-1. Quad Cities Station Unit 1 Cycle 7 Results IPC V01 0 - - VIA!'JED TPt vel 0 -- MIA$tAtti - r - -c -,At t e, catca 2. 0 - - * -,-- i--i --, -- r- -* - r- -i - + - i, - - cauurto, 2. 0 < - r - - - * - - t -- + - - F -,- -, - -,.--r--r--c-,--, .:.. !..,!..,l...l.. !..l..!..,l.e,l..,!..!. !.,!..,! .. l..l.,l..,l..,l.. l..!.. -l.,l..,l...!.....,F..,l..,l i,g I,3 . r. i i i i ! : : : ! ! : I -:-/. - : - :- :-- :-- +:--*: - :. --:-- :-- -- +! ,... l..l..l.. l...l...l..l.. l.. l...l...l.......... E" 1,6 - al.5

+--+------a--+--+--------*

2 ....' !..!.. l...!...!...l..l..l...l...l l l l l =j.4 ni4 i ..... i i i .e 'g< 1,2 ) ....i I.p.. i +... .i.... ;.. p...;.. 7,..;. 1 .. p.p..;..q..;...p..;. ..;..;...p..p.9 3 i g e i i i i i. 1.2 --*- r-t,- --t.- - ;---l--t. --*- r- : ?- ;. .. /.p.7,.3 i I i i i,, ......;.. p..p..;. 3 .....p..q. 3 _t,g _ t,a = I i i i.. = ...... i i i i i i i i r:0.8 - --F - t. -i. -- t. -- F --F --r--l--i. -- t. - '-i.--i 9.8 - --F-i--i. -- t. --! --F--F-i.--. 5--r--r-t,-! g i I i i i 0 -- F--F -+ -i,-- t, -- F --F - t-i.--i. -- t. - F '\\ - t. -i 20.6 - F--F--l--i--i--F--F--F-i-- f--t-. -F-i--i he,.5 -... i = ..... i......, 4 .p..p..;. 4. 4..p..p. +..!. 4..p..p..p. -4 E0.4 -F --h- +- 4 -- 4 -- F--F --h -i-- 4 -- t -- F - i--4 ! ! l ! ! ! l l ! l ! l'\\- - ! ... !..:r...,..,!..,!.. !,..!..!...!..,!.. !..!..l.. '.,l 0.2 o,2 r --r-- i--, -- t -- r- -r - r,- - t - - r - r - '\\...?..,,l r r,, , r, \\ \\ l l l l: l l l ; l I l l l l l l l l l l 0,0 o,a o to : 32 40 50 to to as so ;to 11012013a 140 iso e is to ao 4a sa to le la la tas its 123 130 140 150 INCHt! ABOYt 90f1R Of AtilVI TVtt IRCE3180Y180TTW Of ACilY! FVtt ilCURE01C7CAI 01-03-83 106.3 WO/VT FIGURE 01C7CA1 07-14-13 3468.2 lit 0M ) LOT OF CORI AVERACE DATA )LOTOFConEAYERACEDATA 1.000 ll!A3. STRIIC AVC. = 1.000 1.000 li!A3. STRING AVC = 1.000 lALC $! RING AVC. lALC. STRING AVC. = = lATAPOINISAREPLANARAVERACES lATA POINTS ARE PLANAR AVIIACES plPREA0lNOSAIIALLYNORMAll2ED: NO IIP RIAllNCS AIIALLY NCRWAllIED: NO CECO PROPt!ETARY W5JX3 JO908289 07/11/10 lECO PROPRIETARY U3JX3 J0806261 07/11/90 TPC V91 0 - - WEA3UED I - CALCUu ID 2.0 --r.---i- '--t--r--r- .--i-,.- r--r-!-,- q,g... !...l..,l..,l..,!.. l..l. -l -,l..,!...l..l. -l.,l..,l i i i i... .i .i i i I i..,, a1 26 - - - - - - - - - + - - - - - - - - - * - - + - - + - - - - - - - - * .. l..... ;...l.....p.. p;..;.....;;...;.. ;..;l l l l l m,4 7 M i i i i.... I i. I e I n;'2 y -g- .y c - 1. 0 " -/ t 1-,, --r--r-t-e, -i---% t-t-i. = 8.' I I f i t 1% i $n,3 ,..p....q..t.. p..p..p. q...;.. p..y g.,..j..q . i.. .. i i i.. - n,o.5 ,F --F-i,- -i,-- t. -- F --F--b-.l--i,-- F --F--h. i i --i h0.4 - F--h -+ - t--f --h -+-+-4 --i --F--h i--4 i i. , i i i i f , i e i i i i i i i i i 0.2 --- E --h ---4 -- t-- f --h--l-- 4-- 4-- f--h--h - s--4 l l l l l l 1 l l i l t l 0.0 0 10 23 30 40 50 10 70 la to 104110120130140154 ICit3 AB0Yt ICITW of AcilYE TVtt I 'lCURE 01 C7 CA1 02-15-64 7012.5 WD/WT ' LOT OF CORI AVERACE DATA 1.000 WEA3. STRING AVC = 1.000 lALC. STRING AVC. = l lATA POINTS ARE PLANAR AVERACES I IIP READINC3 AIIALLY NORMAll2[0: NO

{C0 PROP?IETARY V5Jt3 J0306789 07/11/10 Fage C-2

E NFSR-0085 Revision 0 Figure C-2 Quad Cities Station Unit 1 Cycle 8 Results IP: V01 0 - UCAIMED IPC 701 0 -. litAlatl 2,0... r..r. y. 4.. +.. p.. r..,...,.,. 1,.. r.C.u t.uu.,.tt o. 2.0 --F --*- i--i-- +.- r--F-r. -i.-- +. tu tul Af{0,. c 1

r. - r. - i- - i-i i i i i i i i i... i i i

. i ...'.4..:. 4..;.4 4 4. 4. 4 4..;...;. 4. 4 t,3.4 4 4. 4. 4..'.4 4... 4..?..L.4 4. 4 t,a. t=1.6 - ;* - i 1 : -- i

i 11 i ; i i
1 ;

21,6

*--+-------a~",.".,-*-i

- - - - - - + - - - - - - - - - - - - - + - - - - -, - -

E I

I i 4 i i i i t i i i E i l i 5 4 i. I i. i. l E14 ...i. =j 4...i.. i i i i i W -i .i d.2 I ' ' ' i * > i i ' i i i i i i i ' i - - ' d-I 5 '2 l 1 1

d--+--+--*-*-*--d

--- --*--+-- -- - --* ....i..l...t.. ;,.. x%. 7 / ,; i l l l ! .p.q..;..;..p.....;.....; d.., i ; i ! 1 l l AN S - t,o. ..ii,,, 7 ,.%\\..p.g..;..j -t,g 6..p.g..q p p.. = i,,. = i i i i i i.... p..p.g..q..t...p..p..;..q. 4..p. .p.g. 4 go g...,...p..:.. a.. t... p..p..p. g.. t... t... y go. .q..q l ....i,, i. ~ 0.6 - F --F-i.--i. --i. -- F --F-i.--i--i. -- t. --F- -i.--i. 10,6 F--F-F-i,--f--F--F--F-i,--f--t-F-, -F-i g i. 1 = i i i i.., i i I i .g,4. 4. 4. 4 4. 4..p..p.4 4. 4.. F,.4,..- 4 gg,4 4..p. 4..p.4.. 4...p..p..p..p.4.. 4,..p. 4 n. 4 l i i, e i i. 4--l-- i-- + 4--i-- I-- 4. -- +-4--i -} t-- i, t i 4 i I i 0,2 - --h --l -+ - +-- i ! --i.--+ -4. -- f --l --l-M--4 0.2 i i i i i,. i 0.0 - - - L J~ J - ! - L - L 4-J L - '- - L A J i i i i i i i i i 0.0 - L-L 4-J- !- L-L- L 4 - t-L 4---- J 0 10 10 30 40 H 80 70 to se 100 til 120130140150 0 10 20 10 40 Il 10 il 10 10 100 110 120 130 140 150 INet3 MOVI toitz 0F ACTIVE FUEL IN3t3 itM 10TTW of Atilf! TVil ilCL'RE 01 C8 cal 09-25-64 574.7 Iff0/WT FIGUREClC8 cal 04-2H5 4332.7itr0/W1 PLOT Of CCRE AVERACE DATA PLOTOFCOREAVERACECATA "LC, STRING AVC, 1,000 WEAS. SillNC AVC, = 1,000 lALC,titlNCAVC, 1,000 VIA3. $TRIIC AVC. = 1,000 = = PATAPOINisARIPLANARIVERAlts )ATA P0lN13 ARE PLANAR AVIRACES IIP REA0lNC1 AllALLY NORilAlllED: NO TIP REA0lNCS AIIALLY NORWALIZED: X0 lECO Pt0PtlEintY

3Jts J00082tl 07/1t/10 lECO PROPRIETARY W3JXS J0904169 07/11/10 TPC m 0 - - htA3UtEl 1 - tutUuftt, 2.0

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NFSR-OJ' 85 Revision 0 Appendix 0 TIP Traces for Quad Cities Station Unit 2 A comparison of calculated versus measured core axial TIP traces for Quad Cities Station Unit 2 is included in this appendix. Figures included are: Figure D Quad Cities Station Unit 2 Cycle 7 Results Figure D Quad Cities Station Unit 2 Cycle 8 Results Figure D Quad Cities Station Unit 2 Cycle 9 Results Figure D Quad Cities Station Unit 2 Cycle 10 Results a Page D-1

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NFSR-0085 Revision 0 Appendix E TIP Traces for Dresden Station Unit 3 A comparison of calculated versus measured core axial TIP traces for Dresden Station tinit 3 is included in this appendix. Figures included are: Figure E Dresden Station Unit 3 Cycle 8 Results Figure E Dresden Station Unit 3 Cycle 9 Results Figure E Dresden Station Unit 3 Cycle 10 Results figure E Dresden Station Unit 3 Cycle 11 Results i P Page E-1 e
NFSR-0085 Revision 0 Figure E-1 Dresden Station Unit 3 Cycle 8 Results IP: YD1 0
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NFSR-0085 Revision 0 Appendix F TIP Traces for LaSalle County Station Unit 1 A comparison cf calculated versus measured core axial TIP traces for LaSalle County Station Unit 1 is included in this appendix, figures included are: Figure F LaSalle County Station Unit 1 Cycle 1 Results Figure F LaSalle County Station Unit 1 Cycle 2 Results Figure F LaSalle County Station Unit 1 Cycle 3 Results i Page F-1
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NFSR-0085 Revision 0 Figure F-3 LaSalle County Station Unit 1 Cycle 3 Results tP: Vit
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f4FSR-0085 Revision 0 Appendix G TIP Traces for LaSalle County Station Unit 2 A comparison of calculated versus measured core axial TIP traces for LaSaile County Station Unit 2 is included in this appendix. Figures included are: Figure G LaSalle County Station Unit 2 Cycle 1 Results Figure G LaSalle County Station Unit 2 Cycle 2 Results Figure G LaSalle County Station Unit 2 Cycle 3 Results 5 E s Page G-1
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