ML17156A688

Qualification of Steady-State Core Physics Methods for BWR Design & Analysis.

ML17156A688
Person / Time
Site:	Susquehanna
Issue date:	07/31/1988
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17156A689	List: ML17156A688 ML18040A903
References
PL-NF-87-001-A, PL-NF-87-1-A, NUDOCS 8807280374
Download: ML17156A688 (297)

Text

PL-NF-87-001-A B)z

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Pennsylvania Power 8 Light Company 88072S0374 880719 PDR AGQCK 05000387 '*-,

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April 28'988 Docket Nos. 50-387/388 Nr. Harold W. Keiser Senior Vice President-Nuclear Pennsylvania Power and Light CdIIpany 2 North Ninth Street Al l entown, Pennsyl vani a 18101

SUBJECT:

TOPICAL PEPORT PL-NF-87-001, "OUALIFICATION OF STEADY STATE CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS" (TAC NOS. 65171 AND 65172)

RE: SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 The staff has completed action on your March 31, 1987 request for review of Topical Report PL-NF-87-001 related to BWR Steady-State Core Physics Methods.

Our consultant, Brookhaven National Laboratory (BNL) reviewed your report and provided a Technical Evaluation Peport (TER) outlining its reviews and conclusions. The staff has reviewed the RNL TER and has prepared the enclosed safety evaluation.

Based nn our review, we have concluded that the sub.ject Topical Report is acceptable for the purpose of licer sing actions on Susquehanna Steam Electric Station, Units 1 and 2.

Sincerely, WIa 1 ter R. Butler, Director Pro.',ect Directorate. I-2 Division of Reactor Pro,iects I/II Office of Nuclear Peactor Regulation

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Enclosure:

Safety Evaluation cc w/enclosure See next page

Mr. Harold W. Keiser Susquehanna Steam Electric Station Power 5 Light

'ennsylvania Company Units 1 5 2 CC:

Jay Silberg, Fsq. Mr. W. H. Hirst, Manager Shaw, Pittman, Potts A Trowbridge Joint Generation 2300 N Street N.W. Projects Department Washington, D.C. 20037 Atlantic Electric P.O. Box 1500 Bryan A. Snapp, Esq. 1199 Black Horse Pike Assistant Corporate Counsel Pleasantville, New lersey 08232 Pennsylvania Power 5 Light Company 2 North Ninth Street Regi ona 1 'dmi ni stra tor, Reoi on I Allentown, Pennsyl vani a 18101 U.S. Nuclear Regulatory Commission 475 Allendale Road Mr. E. A. Heckman King of Prussia, Pennsylvania 19406 Licensing Group Supervisor Pennsylvania'Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. F. I. Young Resident Inspector P.O. Box 52 Shickshinny, Pennsylvania 18655 Mr. R. J. Benich Services Project Manager General Electric Company 1000 First Avenue King of Prussia, Pennsylvania 19406 Mr. Thomas M. Gerusky, Director Bureau of Radiation Prntection Resources Commonwealth of Pennsylvania P. 0. Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III

'Allegheny Elec. Coorperative, Inc.

212 Locust Street P.O. Box 1266 Harrisburg, Pennsylvania 17108-1266

~p,g REgy Ip UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE 0 y*p%

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGllLATION RELATING TO LICENSING TOPICAL REPORT PL-NF-87-001, REV.O

" UALIFICATION OF STEADY STATE CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS" PENNSYLVANIA POWER 8I LIGHT COMPANY SUS UEHANNA, UNITS 1 AND 2 DOCKET NOS. 50-387 AND 50-388

1. 0 INTRODUCTION By letter dated March 31, 1987, the Pennsylvania Power and Light Company (the licensee) requested approval of Topical Report PL-NF-87-001, Rev. 0, for the purpose of its use in licensing actions for the Susquehanna Steam Electric Station (SSES) Units 1 and,2.

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The report describes the qualiIication of the CPM-2 lattice physics and SIMULATE-E three-dimensional nodal core simulator programs for the steady state design and analysis of boiling water reactors (BWRs). These programs are part of the Advanced Recycle Methodology Program (ARMP) developed by the Electric Power Research Institute (EPRI) for steady state analyses of light water reactors. Brief descriptions of the CPM-2 and SIMULATE-E programs are presented along with comparisons to measurements from operating BWRs and experimental criticals. The results of selected PD07 calculations for uniform lattice criticals and single fuel bundles are also presented. These programs and associated methodologies are used by the licensee for plant operations support, various fuel cycle and safety related calculations, and to provide necessary neutronics input data to transient 'analyses for the two unit Susquehanna Steam Electric Station.

2.0

SUMMARY

OF TOPICAL REPORT The SIMULATE-E three-dimensional code is used by the licensee to model the coupled neutronic and thermal-hydraulic behavior of the Susquehanna Unit 1 and 2 BWR

cores. The required nuclear data are generated by the CPN-2 program which models the BWR fuel bundle and its environment (by-pass channel, cruciform control rod, etc.) in two-dimensions.

2. 1 Descri tion of the CPN-2 Pro ram CPN-2 is a modified version of the CPM (Collision Probability Module) code developed in Sweden by AB Atomenergi/Studsvik for the analysis of PWR and BWR fuel assemblies. The modeling combines fine group spectrum calculations for sub-regions of the assembly (e.g. fuel pin-cells), with a multigroup transport calculation for a partially homogenized, hetrogeneous assembly in two-dimensional (xy) geometry. The code is distributed by EPRI, ard is identical to the original CPN except for the input module which has been improved to make the program more "user friendly." Since these modifications ('as well as those made by the licensee in their implementation and use of CPM-2) did not affect the neutronics calculations, all the original benchmarking of CPN by EPP1/Studsvi k is applicable to CPN-2 as well.

The calculational sequence for a typical BWR assembly involves three basic steps, with the spatial and energy detail becoming successively coarser as larger regions of the assembly are considered. These steps are termed the micro-group, macro-group, and two-dimensional assembly calculations. Cruciform control rods are treated via a special subroutine, and the depletion of gado-linia bearing fuel pins requires an auxiliary calculation with the NiCBURN code.

2.2 CPN-2 uglification The accuracy/adequacy of various aspects of CPM-2 and its models (e.g. nuclear data, treatment of control rods and gadolinia) is demonstrated by comparisons to measured results from power reactors and experimental configurations.

Comparisons of eigenvalues (k'eff ff), pin power/fission rate distributions, and concentrations versus burnup are presented.

'sotopic Some of these results were

generated by the licensee, while others were taken from the EPRI/Studsvik benchmarking of the original version of CPM.

Pin-cell calculations simulating 14 room temperature uniform lattice critical experiments were performed by PPP~L to assess the accuracy of the CPM-2 reactivity calculation (based on the measured buckling). Eight of the configurations contained U02 fuel and the fuel for the remaining 6 contained 2.0 weight percent PuO> in natural uranium. CPM-2 slightly underpredicted (by about 0.5> k) the k eff for the U02 criticals, and overpredicted the ff multiplication factor for the remaining criticals, resulting in an averaqe k

eff of I.COOS with a standard deviation of 0.0072 considering all criticals.

ff The accuracy of the CPN-2 calculation of the rod-wise power distribution was evaluated by comparisons to the gamma-scan measurements performed at squad Cities Unit 1 at the end of Cycle 2. Two 7x7 M02 and three U02 bundles (one 8x8 and two 7x7) were considered in the comparisons. Burnup and void operating history data were obtained for each bundle-elevation from a SIMULATE-E simulation. These data were used in CPN-2 bundle calculations to arrive at the CPM-2/SIMULATE-E predicted statepoints corresponding to the measured data. The comparisons showed generally good agreement between measurement and prediction (average = 4.0l) with CPM-2 tending to overpredict the peak rod power.

The results of the EPRI/Studsvik benchmarking of the original CPM code to uniform lattice criticals, small core critical experiments performed at the KRITZ facility, and -measured concentrations of uranium and plutonium isotopes from Yankee and Saxton spent fuel are also presented. These comparisons show generally reasonable agreement between CPM predicted and measured quantities.

,".3 Descri tion of SINULATE-E The EPRI distributed SIMULATE-E three-dimensional coupled neutronics/

thermal-hydraulics core simulator program is used by PPAL in their steady state core analyses. The thermal-hydraulics calculations use an EPRI developed void correlation and the FIBMR methodology developed by Yankee Atomic Electric

Company. The methodology employed for the neutronics calculations may be selected by the user from several available options; PPSL uses the Modified Coarse Mesh Diffusion Theory (PRESTO) option. Two-group macroscopic cross sections for each fuel type are determ'.ned by CPM-2 as a function of fuel exposure, void history, moderator, fuel and control conditions, and xenon concentration. After processing by NORGE-B2, they are input to SIMULATE-E along with radial and axial albedos applied at the core-reflector interfaces.

Normal'ization of the model to match plant operating data is performed via adjustment of several input data parameters. Separate mndels are created at hot operating and cold conditions. The licensee has made a number of charges to the code, including the ability to calculate the Critical Power Ratio (based on the Advanced Nuclear Fuels Corporation, formerly EXXON Nuclear, XN-3 critical heat flux correlation), and linear heat generation rate and average planar heat generation rate thermal limits evaluations. These changes have not resulted in any changes to the basic neutronics or thermal-hydraulics calculations.

2.4 . SIMULATE-E (}uglification The qualification of the SIMULATE-E program is based on simulations of the first two cycles of (juad Cities Unit-1 (gC-1) and Peach Bottom Unit-2 (PB-2),

and of the first two-plus and one-plus cycles (i.e., from BOL to approximately early 1987) of Susquehanna Units 1 and 2, respectively. Comparisons of SIMULATE-E predicted values were made to hot and cold multiplication factors (k ff) and power and flow distributions. The accuracy of the predicted power distributions was evaluated based on comparisons to TIP detector readings, and to results from gamma-scans. Power and flow distributions were compared to

.results from. the on-line core monitoring system.

The k ff comparisons for the Susquehanna units considered 257 hot operating condition steady-state statepoints, and 39 (3 local and 36 in-sequence) cold critical statepoints. These comparisons indicated that the ability of the PPSL SIMULATE-E hot and cold models to predict k ff depends on the core average eff exposure and the gadolinia loading. There is a nearly constant bias between the hot and cold predictions, with the hot k ff consistently lower. Using this

data, the licensee generates hot and cold cycle-dependent target critical core k

ff curves for use in the core follow, and shutdown margin and control rod eff worth analyses of individual cycles.

The power distribution comparisons utilized all available TIP sets from both Susquehanna units and considered nodal and axially averaged (radial) quantities. Asymmetries in the measured data were quantified by considering symmetric nodal or radial TIP readings to provide an estimate of the measurement uncertainties associated with each TIP set. Nodal RMS errors tend to be in the 4-6$ range, with differences near the middle of cycle and end of cycle power coastdown in the 6-95 range. The average nodal and radial RMS errors considering all 82 TIP sets are 5.74 and 2.58 percent, respectively.

The corresponding average asymmetries based on 44 TIP sets are 5.22 and 2.55 percent, respectively.

Four core average axial power distribution and three-bundle flow comparisons are also presented, considering one statepoint per Susquehanna unit/cycle.

These. comparisons are made to data produced by the on-line Core Monitoring System (CMS) to demonstrate consistency of the results. (The bF. process computer Pl program was used for the first cycle of both units, with the ANF POMERPLEX CMS used in all subsequent cycles). These comparisons showed good agreement between the SIMULATE-E and CYS results.

Comparisons tn measured data from the first two cycles of squad Cities Unit I (gC-1) were also performed. In addition to hot reactivity and TIP data similar to that from the Susquehanna units, the OC-I measurements included 33 cold critical configurations (22 local) from Cycle-l, and bundle gamma scan measurements from the end of cycles (EOC) one and two.

The gC-I hot critical comparisons showed a similar trend versus exposure to that observed earlier; however, the relatively low gadolinia loading in gC-1 resulted in the absence of the bowl-shaped gadolinia component in the variation. The large cold critical data base served to augment the earlier analyses. The gC-1 cold critical comparisons were used to confirm that there

is no significant bias between SIMULATE-E f< for in-sequence predictions of k eff and local critical configurations.

The gC-I based power distribution comparisons considered 15 TIP sets from Cycle I and 13 sets from Cycle 2, along with gamma scan data from 31 and 89 bundles at EOCl and EOC2, respectively. The nodal and radial RNS differences from the TIP comparisons are roughly twice as large as those observed ~or thp Susquehanna comparisons. The EOCI gamma scan data consisted of measuring the axial peak to bundle average La-140 activities and served to benchmark the SIMULATE-E calculation of the axial peaking factor. The resulting difference was 1"; ( =25) with the agreement for controlled bundles considerablv better than for uncontrolled. The EOC2 gamma scar. data is much more extensive and permits comparisons of individual bundle axial La-140 activity distributions, as well as radial, nodal and peak to average comparisons. Peripheral and mixed oxide bundles were not included in the radial and nodal comparisons and the top and bottom six inches were eliminated from the nodal corn'parisors. The peak-to average comparisons resulted in an averaqe di erence o. about 0.2~ ( =1.5f) with a maximum difference of about 4%. The average standard deviation from the individual bundle gamma scans was 6.3X with more than 85> o< the individual bundle 's in the 5-8~ range. The standard deviation from the radial and nodal gamma scan comparisons were about 2'A and 5.5', respectively. The ouot~d measurement uncertainty for the qamma scans was 31,.

The final qualification of SIMULATE-E presented in the report consists of power distribution comoarisons to TIP measurements and data from the GE Pl process computer for Peach Sottom Unit 2 {PB-2) cycles 1 and/or 2. The level nf agreement with measured TIP data from these cnmparisons is reasonable and consistent with that observed earlier. The purpose of the PB-2 simulations was tn generate input for the analysis of the turbine trip tests performed near the end of Cycle 2, including an accurate representation of the initial conditions.

The non-steady state operation that preceded these tests required an accurate modeling of non-equi librium xenon distributions and concentrations. Comparisons of the predicted core average axial power distributions just prior to the

three tests (top peaked, middle peaked and slightly bottom peaked) to data from the process computer showed good agreement.

2.5 Descri tion of PD 7 The geometry in the CPM-2 lattice physics code is limited to representing an individual fuel assembly. In some applications, however, a multiple assembly calculation is required, and for these applications PPSL uses the general purpose.PD(7 code. The program solves the few group diffusion .theory equation based on the finite difference spatial approximation in one, two, or three dimensions. While up to five energy groups are permitted (including two overlapping) thermal energy groups, the licensee generally utilizes four groups with a single thermal group. Microscopic or macroscopic cross section data may be employed; PP8L typically uses macroscopic data from CPM-2 and processed with the COPHIN code.

. 6 6II6 The PP5L qualification of PD(7 consisted of analyzing the same ur iform lattice criticals used in the benchmarking of CPM-2, along with comparisons to CPM-2 assembly calculations for typical controlled and uncontrolled BWR fuel bundles.

The uniform lattice calculations modelled the critical core configurations in one-dimensional cylindrical geometry with an explicit accounting of the radial reflector and a bucl ling correction to account for axial leakage.

Reasonable agreement was obtained with the CPM-P calculated k +f s, 0.9972 eff versus 0.9951 and 1.0076 versus 1.0144 for the UOand mixed oxide lattices, respectively.

The PD07 single fuel assembly calculations modelled each pin-cell explicitly, and used shielding factors derived by comparison to CPM-2 results, for gadolinia bearing fuel pins and control rods. Two separate fuel bundles from the initial core loading of the Susquehanna units were selected fnr the .

comparisons. The results showed generally good agreement between CPM-2 and

PD(7 for the bundle k 's and rod-wise power distributions with maximum errors of about 4$ and ?,. for uncontrolled and controlled bundles, respectively.

3. 0 EVALUAITON The CPM-2 and SIMULATE-E programs were developed by EPRI for the steady state analyses of LWRs. The licensee plans to use these codes for plant operations support, various fuel cycle and safety related calculations, and to provide necessary neutronics input data to transient analyses for the two BWR units at the Susquehanna Steam Electric Station.

The present review considered the information presented in the topical report and additional information provided by the licensee in a letter dated February 17, 1988. The review considered the qualification of the FIBWR thermal-hydraulics methodology only in its role as an integral part of the SIMULATE-E program. The performance of FIBWR as a stand-alone thermal-hydraulics code, and the validity/applicability of the ANF XN-3 CHF corre1ation were considered to be outside the scope of this review.

The methodologies (not including the qualification presented in this report) embodied in the CPN-2 and SIMULATE-E programs have been previously reviewed and found acceptable for steady state nuclear core design analyses of plants other than Susquehanna, and are representative of current practice.

The primary role of CPN-2 within the PPSL calculational sequence for BWR analyses is to provide nuclear data (basically two-group cross sections) to the SIMULATE-E core simulator program. The benchmarking of SIMULATE-F via comparisons to measurements from operating BWRs therefore serves as the ultimate, though somewhat indirect, qualification of CPM-2. However, PP8L and EPRI/Studsvik have performed a number of comparisons to measured data from experimental configurations and operating BWRs to test various aspects of the CPM/CPh1-2 neutronics calculation methodology and nuclear data.

Comparisons to uniform lattice cold criticals and KRITZ small core criticals provide an integral test of the ability of CPM-2 4o predict reactivity (multiplication factors). Comparisons to measured rod-wise gamma scan data for selected assemblies from an operating BWR, and to measured rod-wise fission rate distributions from KRITZ experiments, serve as. a qualification of the treatment of neutron transport and other aspects of the modelling in the hiqhly heterogeneous environments of real RWR fuel bundles and reactor cores.

Finally, comparisons of calculated uranium and plutonium isotopic concentrations were made to data from the destructive analysis of spent fuel rom the Yankee and Saxton reactors. The level of agreement between CPM-2 calculated and measured quantities is reasonable, and typical of that observed with currently accepted methods. In addition, CPM-2 tends to overestimate the local peaking factor in an assembly, implying a generally conservative prediction of the linear heat generation rate.

The benchmarking of the SIMULATE-E program consisted of simulations of several cycles of operation of three BWRs including all available data from PPAL's Susquehanna units starting at beginning of Cycle-1 (BOCI).

The hot reactivity comparisons involved more than five op~rating cycles (almost 300 statepoints) for cores containing a variety of BWR fuel bundle designs.

The calculated hot k ff exhibited a bias relative to the measured critical eff k

ff which was consistent in magnitude with that observed for accented eff three-dimensional core simulator codes. The observed variation led to the development of a correlation which is a bowl shaped function of gadolinia loading and a roughly linear function of exposure. This "target" k ef< f< is used to predict the critical core k eff for a particular unit-cycle.

ff The cold critical comparisons considered 47 insequence and 25 local configurations. The results showed a similar variation in the predicted cold critical k eff ff to that observed for hot conditions; the cold critical keff "target" for use with SIMULATE-E is therefore obtained bv adding a constant bias to the hot correlation. In addition, the results showed no significant differences between the k ff for local and in-sequence criticals, thereby

10 demonstrating the ability of SIMULATE-E to perform shut-down margin ca 1 cul ati ons.

The benchmarking of the SIMULATE-E calculation of power distributions considered measured TIP detector readings and gamma scans, and data from plant core monitoring systems. The albedos and other adiustable parameters were determined durinq model normalization to operating data from Susquehanna Unit l Cycles 1 and 2, and remained unchanged for all subsequent simulations.

The comparisons for the 82 TIP sets covering more than three cycles of operation of the two Susquehanna units yielded average nodal and radial RMS differences of 5.7 and 2.6 percent, respectively. The estimated errors in the TIP measurements were determined by considering symmetric detector readings, and were of the same order. The TIP comparisons for Duad Cities and Peach Bottom yielded higher differences, i.e., nodal and radial RNS errors considering all TIP sets of about 10 and about 5 percent, respectively, for Outed Cities, and somewhat lower for Peach Bottom.

h The comparisons to the squad Cities qamma scan measurements at EOCl and EOC2 further demonstrated the ability of SIMULATF.-E to calculate power distributions.

The axial peak to average was predicted to within about 11 with a standard deviation of 1-2%, and the standard deviations from the radial and nodal comparisons were about 2 and about 5 percent, respectively. The peripheral bundles were not included in these comparisons, and in addition the top and bottom six inches were not considered in the nodal comparisons. The quoted uncertainty for the gamma scan measurement is 3.0X.

Comparisons of core average axial power distributions to results from the AE P1 or ANF POl<ERPLEX core monitoring systems for the Susquehanna units and Peach Bottom Unit-2 (PB-2) near EOC2, though limited, showed good agreement.

The PB-2 comparisons considered the effects of non-equilibrium xenon and included top, middle and bottom peaked axial power distributions. Three bundle flow distributions from the Susquehanna core monitoring systems were also compared to results generated by SIMULATE-E with generally good agreement. The

power distribution comparisons of SIMULATE-E to measured data showed generally reasonable agreement and were consistent with, the levels of agreement observed with accepted methods. The larger differences observed in the Puad Cities and Peach Bottom comparisons are partially due to the SIMULATE-E models not being specifically normalized for these simulations. The generally good agreement, however, provides reasonable confidence that SIMULATE-E can be used for predictive calculations for the Susquehanna units.

The limited comparisons of PD(7 to results from uniform lattice criticals and CPM-2 single assembly calculations showed reasonable agreement. Thc comparisons were based on the use of 4 energy group cross sections from CPM-2.

The licensee notes that while it does not intend to perform three-dimensional calculations with PDg7, it may use the program for various two-dimensional analyses including independent verification of calculations, calculations of non-standard configurations such as partially loaded cores, and in the development of future model improvements for SIMULATE-E. Appropriate qualification by the licensee of the use of PDg7 for configurations larger than multiple bundle arrays is recommended.

4.0 CONCLUSION

S The CPM-2 and SIMULATE-E codes were developed under the sponsorship of the Electric Power Research Institute and are part of the presently recommended procedures for BWR analyses similar to those intended for application to Susquehanna Units 1 and 2. The benchmarking of the codes by the licensee relative to measurements from operating reactors and experimental configurations resulted in agreement typical of that observed with accepted methods. The comparisons of PD(7 to results from uniform lattice criticals and CPM-2 single assembly calculations also showed reasonable agreement. The staff therefore concludes that the CPM-2/SIMULATE-E methodology, and the use of PDg7 for auxiliary calculations represent an acceptable approach for analyses performed by the licensee in support of license applications and operation of the two BWR reactors at the Susquehanna Steam Electric Station.

The staff recommends that appropriate qualification be made by the licensee of the use of PO(7 for configurations larger than multiple bundle arrays, if such configurations are considered for calculation by PDg7. The staff also recommends continued comparisons of calculated physics parameters with measured data from future physics startup tests and reactor fuel cycles.

Principal Contributor: D. Fieno

PL-NF-87-001-A Issue Date: July, 1988 QUALIFICATION OF STEADY STATE CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS PL-NF-87-001 Revision 0 March 1987 Principal Engineers Andrew Dyszel Kenneth C. Knoll Contributing Engineers John H. Emmett Eric R. Jebsen Chester R. Lehmann Anthony J. Roscioli Robert M. Rose John P. Spadaro William J. Weadon Approved: Date: 3/31/87 Jo M. Kulick Supervisor-Nuclear Fuels Engineering Date: 3/31/87 Je e S. Stefanko

.-Nuclear Fuels a stems Engineering

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LEGAL NOTICE This topical report represents the efforts of Pennsylvania Power G Light Company (PPsL) and reflects .the technical capabilities of its nuclear fuel management personnel. The information contained herein is completely true and accurate to the best of the Company's knowledge. The sole intended purpose of this report and the information contained herein is to provide a technical basis for PPGL's qualification to perform steady state core physics analyses of the Susquehanna SES reactors. Any use of this report or the information by anyone other than PP&L or the U.S. Nuclear Regulatory Commission is unauthorized. With'regard to any unauthorized use, Pennsylvania Power s Light Company and its officers, directors, agents, and employees make no warranty, either expressed or implied, as to the accuracy, completeness, or usefulness of this report or the information, and assume no liability with respect to its use.

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ABSTRACT This topical report presents the benchmarking analyses which demonstrate the validity of Pennsylvania Power 6 Light Company's (PPGL's) analytical methods as well as PPaL's qualification to perform steady state core physics calculations for reload design and licensing analysis applications.

PPGL's steady state core physics methods are based mainly on the computer codes provided by the Electric Power Research Institute. These codes include:

the MICBURN gadolinia fuel pin depletion code; the CPM-2 assembly lattice depletion code; and the SIMULATE-E three-dimensional core simulation code.

The benchmarking analyses contained in this topical report include comparisons of PPsL's CPM-'2 fuel pin and assembly calculations to uniform lattice critical experiments and to gamma scan measurements taken from the Quad Cities Unit 1 reactor. Extensive benchmarking of PPGL's SIMULATE-E models is also presented, including comparisons to measured neutron flux data (i.e.,

Traversing In-core Probe data) and criticals from all available Susquehanna SES cycles, two cycles of Quad Cities Unit 1, and two cycles of Peach Bottom Unit 2; the SIMULATE-E models are also benchmarked against, gamma scan measurements from Quad Cities Unit 1. PPGL's calculations with the industry standard diffusion theory code PDQ7 are also included in this topical report.

In total, the benchmarking results compare very favorably to the measured data, and thus demonstrate PPGL's qualifications to perform steady state core physics calculations for reload design'nd licensing analysis applications.

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ACKNOWLEDGEMENTS The authors gratefully acknowledge the expert stenographic work provided by Ms. Evelyn Lugo and Ms. Sandra K. Lines, and the excellent graphics prepared by Mr. Francis E. Grim and Ms. Denise S. Showalter, all of whose efforts have contributed to the quality and timely completion of this topical report.

The authors also acknowledge the efforts of Mr. Rocco R. Sgarro for his licensing reviews and coordination with the NRC.

In addition, the consulting reviews and recommendations provided by Dr. Jack R. Fisher and Mr. Rodney L. Grow of Utility Resource Associates, and Mr. Edward D. Kendrick, Dr. Antonio Ancona, and Mr. Demitrios T. Gournelos of Utility Associates International are greatly appreciated.

I QUALIFICATION OF STEADY STATE CORE PHYSICS METHODS FOR BWR DESIGN AND ANALYSIS TABLE OF CONTENTS Section Page 1.0 Introduction 2.0 Lattice Physics Methods 2.1 Description of CPM-2 8 2.2 Uniform Lattice Criticals 19 2.3 Quad Cities Pin Power Distribution Comparisons 24 2.4 EPRI Benchmark Evaluations 38 3.0 Core Simulation Methods 49 3.1 Description of SIMULATE-E 50 3.2 Susquehanna SES Units 1 and 2 Benchmark 54 3.2.1 Hot Critical Core Reactivity Comparisons 56 3.2.2 Cold Critical Core Reactivity Comparisons 57 3.2.3 Traversing In-core Probe Data Comparisons 59 3.2.4 Core Monitoring System Comparisons 65 3.3 Quad Cities Unit 1 Cycles 1 and 2 Benchmark 140 3.3.1 Hot Critical Core Reactivity Comparisons 141 3.3.2 Cold Critical Coze Reactivity Comparisons 141 3.3.3 Traversing In-core Probe Data Comparisons 142 3.3.4 Gamma Scan Comparisons 143 3.4 Peach Bottom Unit 2 Cycles 1 and 2 Comparisons 185 4.0 Special Applications with PDQ7 195 4.1 Description of PDQ7 196 4.2 Uniform Lattice Criticals 198 4.3 Comparisons to CPM-2 201 5.0 Summary and Conclusions 206 6.0 References 209 Amendments

1. Response to NRC Request For Additional Information 213

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LIST OF TABLES Table Number Title Page General Design and Operating Features of the Susquehanna SES Reactors 2.1.1 Sixty-nine Group Energy Boundaries for the CPM and 12 MICBURN Cross Section Library 2.1.2 Energy Group Structure for"Macro-Group and Two-Dimensional .13 Calculations 2.1.3 Heavy Nuclide Chains 14 2.1.4 Fission Product Chains l5 2.1.5 Modifications to ENDF-B/III Data for CPM-2 Cross Section 16 Library 2.2.1 TRX Uniform Lattice Critical Test Data 20 2.2.2 ESADA Uniform Lattice Critical Test Data 21 2.2.3 CPM-2 Results for TRX Criticals 22 2.2.4 CPM-2 Results for ESADA Criticals 23 2.3.1 Assemblies Used in Rod to Rod Gamma Scan 27 2.3.2 Quad Cities Unit 1 End of Cycle 2 -- Summary of 28 Normalized LA-140 Activity Pin Comparisons 2.3.3 Quad Cities Unit. 1 End of Cycle 2 Peak La-140 Activity 29 Comparisons 2.4.1 EPRI-CPM Results from the TRX Critical Benchmarking 2.4.2 EPRI-CPM Results from the ESADA Critical Benchmarking 41 2.4.3 EPRI Isotopic Comparisons to Saxton Data 3.2.1 Measured Core Operating Parameters for SIMULATE-E Core 67 Reactivity Calculations 3.2.2 Summary of the Susquehanna SES Benchmarking Data Base 3.2.3 Susquehanna SES Hot Critical Core K-effective Data 69

'I 3.2.4 Susquehanna SES Target vs. SIMULATE-E Calculated Critical 79 Core K-effective Statistics 3.2.5 Susquehanna SES Unit 2 Cycle 2 Core K-effective 80 Sensitivity to Measured Core Operating Data

LIST OF TABLES (continued)

Table Number Title ~Pa e 3.2.6 Susquehanna SES Calculated Cold Xenon-Free Critical Core K-effectives 3.2.7 Susquehanna SES Cold Minus Hot Critical Core K-effective 83 3.2.8 Susquehanna SES Unit 1 Cycle 1 TIP Response Comparisons 85 3.2.9 Susquehanna SES Unit 1 Cycle 2 TIP Response Comparisons 86 3.2.10 Susquehanna SES Unit 1 Cycle 3 TIP Response Comparisons 87 3.2.11 Susquehanna SES Unit 2 Cycle 1 TIP Response Comparisons 88 3.2.12 Summary of Susquehanna SES TIP Response Comparisons 89 3.2.13 Summary of Susquehanna SES TIP Response Asymmetries 90 E

3.3.1 Quad Cities Unit 1 Cycle 1 Calculated Cold Xenon-Free 148 Core Critical K-effectives 3.3.2 Quad Cities Unit 1 Cycle 1 In-Sequence Versus Local 149 Critical Comparison 3 3.3, Summary of Quad Cities Unit 1 Cycles 1 and 2 TIP 151 Response Comparisons 3.3.4 Quad Cities Unit 1 EOC 1 Gamma Scan Comparisons-- 152 Uncontrolled Bundles 3.3.5 Quad Cities Unit 1 EOC 1 Gamma Scan Comparisons 153 Controlled Bundles 3.3.6 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 154-Peak to Average La-140 Activities 3.3.7 Quad Cities Unit 1 EOC 2 Individual Bundle Comparisons 156 4.1.1 Energy Group Structure Used in PDQ7 Calculations 197 4.2.1 PDQ7 Results for TRX Criticals 199 4.2.2 PDQ7 Results for ESADA Criticals 200

LIST OF FIGURES Figure Number Title Page Susquehanna SES Units 1 and 2 Core 1.2 Typical Core Power vs. Core Flow 1.3 PPaL Steady State Core Physics Methods Computer Code Flowchart 2.1.1 Calculational Flow in CPM-2 17 2.1.2 Example of BWR Cell Geometry in the 2-D Calculation 18 2 ~ 3.-1 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 30 Normalized LA-140 Pin Activities -- Assembly ID: GEB159--

93 Inches from Bottom of Core Cities Unit 2.3.2 Quad 1 EOC 2 Gamma Normalized LA-140 Pin Activities Comparisons Scan Assembly ID: GEB161 31 56 Inches from Bottom of Core 2.3.3 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 32 Normalized LA-140 Pin Activities Assembly ID: GEH002 21 Inches from Bottom of Core Cities Unit 2.3.4 Quad 1 EOC 2 Gamma Scan Normalized LA-140 Pin Activities Comparisons Assembly ID: GEH002 33 93 Inches from Bottom of Core 2.3.5 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 34 Normalized LA-140 Pin Activities Assembly ID: CX0672 21 Inches from Bottom of Core 2.3.6 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 35 Normalized LA-140 Pin Activities -- Assembly ID: CX0672 87 Inches from Bottom of Core 2.3.7 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 36 Normalized LA-140 Pin Activities Assembly ID: CX0214--

51 Inches from Bottom of Core 2.3.8 Quad Cities Unit 1 EOC 2 Gamma Scan Comparisons 37 Normalized LA-140 Pin Activities Assembly ID: CX0214 129 Inches from Bottom of Core 2.4.1 Fission Rate Comparison for an o Sx8 BWR Assembly of the Plutonium Island Type T=245 C 2.4.2 Fission Rate Comparison for a 15x15 PWR Mixed Oxide 0 44 Assembly with Water Holes and Absorber Rods T=245 C

LIST OF FIGURES (continued)

Figure Number Title Page 2.4.3 Fission Rate Comparison for a 14x14 PWR Mixed Oxide 45 Assembly Surrounded By UO 2

Assemblies T=240 0 C 2.4.4 EPRI-CPM Comparison to Yankee PU-239/PU-240 Isotopic 46 Ratios 2.4.5 EPRI-CPM Comparison to Yankee PU-240/PU-241 Isotopic 47 Ratios 2.4.6 EPRI-CPM Comparison to Yankee PU-241/PU-242 Isotopic 48 Ratios 3.1.1 BWR Fuel Assembly Bypass Flow Paths 53 3.2.1 SIMULATE-E Hot and Cold Critical Core K-effectives vs. 91 Core Average Exposure 3.2.2 SIMULATE-E Hot Critical Core K-effective vs Core Thermal 92 Power 3.2.3 SIMULATE-E Hot Critical Core K-effective vs Total Core 93 Flow 3.2.4 SIMULATE-E Hot Critical Core K-effective vs Core Inlet 94 Subcooling 3.2.5 SIMULATE-E Hot Critical Core K-effective vs Dome Pressure 95 3.2.6 SIMULATE-E Hot Critical Core K-effective vs Critical 96 Control Rod Density 3.2.7 Target and SIMULATE-E Calculated Hot Critical Core 97 K-effectives vs. Core Average Exposure 3.2.8 Susquehanna SES Units 1 and 2 Core TIP Locations 98 3.2.9 Susquehanna SES Relative Nodal RMS of TIP Response 99 Comparisons 3.2.10 Susquehanna SES Unit 1 Cycle 1 Average Axial TIP Response 3.00 Comparison 1.490 GWD/MTU Cycle Exposure 3.2.11 Susquehanna SES Unit 1 Cycle 1 Radial TIP Response 101 Comparisons -- 1.490 GWD/MTU Cycle Exposure 3.2.12 'usquehanna SES Unit 1 Cycle 1 Individual TIP Response 102 Comparisons 1.490 GWD/MTU Cycle Exposure

LIST OF FIGURES (continued)

Figure Number Title Page 3.2.13 Susquehanna SES Unit 1 Cycle 1 Average Axial, TIP 103 Response Comparison 5.918 GWD/MTU Cycle Exposure 3.2.14 Susquehanna SES Unit 1 Cycle 1 Radial TIP Response 104 Comparisons 5.918 GWD/MTU Cycle Exposure 3.2.15 Susquehanna SES Unit 1 Cycle 1 Individual TIP Response 105 Comparisons -- 5.918 GWD/MTU Cycle Exposure 3.2.16 Susquehanna SES Unit 1 Cycle 1 Average Axial TIP 106 Response Comparison 11.617 GWD/MTU Cycle Exposure 3.2.17 Susquehanna SES Unit 1 Cycle 1 Radial TIP Response 107

-Comparisons 11.617 GWD/MTU Cycle Exposure 3.2.18 Susquehanna SES Unit 1 Cycle 1 Individual TIP Response 108 Comparisons 11.617 GWD/MTU Cycle Exposure 3.2.19, Susquehanna SES Unit 1 Cycle 2 Average Axial TIP Response 109 Comparison -- 0.200 GWD/MTU Cycle Exposure 3.2.20 Susquehanna SES Unit 1 Cycle 2 Radial TIP Response 110 Comparisons 0.200 GWD/MTU Cycle Exposure 3.2.21 Susquehanna SES Unit 1 Cycle 2 Individual TIP Response Comparisons -- 0 '00 GWD/MTU Cycle Exposure 3.2.22 Susquehanna SES Unit 1 Cycle 2 Average Axial TIP Response 112 Comparison 2.587 GWD/MTU Cycle Exposure 3.2.23 Susquehanna SES Unit 1 Cycle 2 Radial TIP Response 113 Comparisons 2.587 GWD/MTU Cycle Exposure 3.2.24 Susquehanna SES Unit 1 Cycle 2 Individual TIP Response 114 Comparisons 2.587 GWD/MTU Cycle Exposure 3.2.25 Susquehanna SES Unit 1 Cycle 2 Average Axial TIP Response 115 Comparison 4.638 GWD/MTU Cycle Exposure 3.2.26 Susquehanna SES Unit 1 Cycle 2, Radial TIP Response 116 Comparisons 4.638 GWD/MTU Cycle Exposure 3.2.27 Susquehanna SES Unit 1 Cycle 2 Individual TIP Response 117 Comparisons 4.638 GWD/MTU Cycle Exposure 3.2.28 Susquehanna SES Unit 1 Cycle 3 Average Axial TIP 118 Response Comparison -- 0.178 GWD/MTU Cycle Exposure

LIST OF FIGURES (continued)

Figure Number Title Page 3.2.29 Susquehanna SES Unit 1 Cycle 3 Radial TIP Response 119 Comparisons 0.178 GWD/MTU Cycle Exposure 3.2.30 Susquehanna SES Unit 1 Cycle 3 -Individual TIP Response 120 Comparisons <<-. 0.178 GWD/MTU Cycle Exposure 3.2.31 Susquehanna SES Unit 1 Cycle 3 Average Axial TIP 121 Response Comparison 2.228 GWD/MTU Cycle Exposure 3.2.32 Susquehanna SES Unit 1 Cycle 3 Radial TIP Response 122 Comparisons -- 2.228 GWD/MTU Cycle Exposure 3.2.33 Susquehanna SES Unit 1 Cycle 3 Individual TIP Response 123 Comparisons -- 2.228 GWD/MTU Cycle'xposure 3.2.34 Susquehanna SES Unit 2 Cycle 1 Average Axial TIP 124 Response Comparison -- 0.387 GWD/MTU Cycle Exposure 3.2.35 Susquehanna SES Unit 2 Cycle 1 Radial TIP Response 125 Comparisons 0.387 GWD/MTU Cycle Exposure 3.2.36 Susquehanna SES Unit 2 Cycle 1 Individual TIP Response 126

,Comparisons 0.387 GWD/MTU Cycle Exposure 3.2.37 Susquehanna SES Unit 2 Cycle 1 Average Axial TIP 127 Response Comparison 5.249 GWD/MTU Cycle Exposure 3.2.38 Susquehanna SES Unit 2 Cycle 1 Radial TIP Response 128 Comparisons -- 5.249 GWD/MTU Cycle Exposure 3.2.39 Susquehanna SES Unit 2 Cycle 1 Individual TIP Response Comparisons 5.249 GWD/MTU 'Cycle Exposure 129 3.2.40 Susquehanna SES Unit. 2 Cycle 1 Average Axial TIP Response Comparison 12.050 GWD/MTU Cycle Exposure 130 3.2.41 Susquehanna SES Unit 2 Cycle 1 Radial TIP Response 131 Comparisons -- 12.050 GWD/MTU Cycle Exposure 3.2.42 Susquehanna SES Unit 2 Cycle 1 Individual TIP Response 132 Comparisons 12.050 GWD/MTU Cycle Exposure 3.2.43 Susquehanna SES Unit 1 Cycle 1 SIMULATE-E vs. GE Process 133 Computer Core Average Axial Power Distribution .

3.2.44 Susquehanna SES Unit 1 Cycle 2 SIMULATE-E vs. POWERPLEX 134 Core Average Axial Power Distribution ~

LIST OF FIGURES (continued)

Figure Number Title Page 3.2.45 Susquehanna SES Unit 1 Cycle 3 SIMULATE-E vs. POWERPLEX 135 Core Average Axial Power Distribution 3.2.46 Susquehanna SES Unit 2 Cycle 2 SIMULATE-E vs. POWERPLEX 136 Core Average Axial Power Distribution 3.2.47 Susquehanna SES Unit 1 Cycle 1 SIMULATE-E vs. GE Process 137 Computer Bundle Flows at 1.490 GwD/MTU 3.2.48 Susquehanna SES Unit 1 Cycle 3 SIMULATE-E vs. POWERPLEX 138 Bundle Flows at 0.178 GWD/MTU 3.2.49 Susquehanna SES Unit 2 Cycle 2 SIMULATE-E vs. POWERPLEX 139 Bundle Flows at 0.583 GWD/MTU 3.3.1 Quad Cities Unit 1 Core TIP Locations 158 3.3.2 SIMULATE-E Hot Critical Core .K-effective vs. Core Average 159 Exposure 3.3.3 Quad Cities Unit 1 Cycle 1 SIMULATE-E Hot and Cold 160 Critical Core K-effectives 3.3.4 Cities Unit 1 Average Axial TIP Response Quad Comparison 2.2391 Cycle GWD/MTU Core Average Exposure 161 3.3.5 Quad Cities Unit 1 Cycle 1 Radial TIP Response 162 Comparisons 2.239 GWD/MTU Core Average Exposure 3.3.6 Quad Cities Unit 1 Cycle 1 Individual TIP Response 163 Comparisons <<- 2.239 GWD/MTU Core Average Exposure

' ' Cities Unit 1 Average Axial TIP Response Quad 7.3961 Cycle 3 164 Comparison GWD/MTU Core Average Exposure 3.3.8 Quad Cities Unit 1 Cycle 1 Radial TIP Response 165 Comparisons -- 7.396 GWD/MTU Core Average Exposure 3.3.9 Quad Cities Unit 1 Cycle 1 Individual TIP Response Comparisons 7.396 GWD/MTU Core Average Exposure 3.6'6 3.3.10 Cities Unit 2 Average Axial TIP Response Quad Comparison 7.5321 Cycle GWD/MTU Core Average Exposure 167 3.3.11 Quad Cities Unit 1 Cycle 2 Radial TIP Response 168 Comparisons 7.532 GWD/MTU Core Average Exposure

LIST OF FIGURES (continued)

Figure Number Title ~Pa e 3.3.12 Quad Cities Unit 1 Cycle 2 Individual TIP Response 169 Comparisons 7.532 GWD/MTU Core Average Exposure 3.3.13 Quad Cities Unit 1 Cycle 2 Average Axial TIP Response 170 Comparison 13.198 GWD/MTU Core-Average. Exposure 3.3.14 Quad Cities Unit 1 Cycle 2 Radial TIP Response 171 Comparisons 13.198 GWD/MTU Core Average Exposure 3.3.15 Quad Cities Unit 1 Cycle 2 Individual TIP Response 172 Comparisons 13.198 GWD/MTU Core Average Exposure 3.3.16 Quad Cities Unit 1 EOC 1 Gamma Scan Comparison 173 Normalized Axial La-140 Activity Bundle Location 23,10 3.3.17 'uad Cities Unit 1 EOC 1 Gamma Scan Comparison 174 Normalized Axial La-140 Activity Bundle Location 55,40 3.3.18 Quad Cities Unit 1 EOC 1 Gamma Scan Comparison 175 Normalized Axial La-140 Activity 31 Bundle Average 3.3.19 Quad Cities Unit 1 EOC 2 Radial Gamma Scan Comparison 176 3.3.20 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 177 Bundle ID: CX0662 3.3.21 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 178 Bundle ID: CX0399 3.3.22 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 179 Bundle ID: CX0231 3.3.23 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 180 Bundle ID: CX0297 .

3.3.24 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison-- 181 Bundle ID: CX0717 3.3.25 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 182 Bundle ID: CX0378 3.3.26 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison 183 Bundle ID: CX0150 3.3.27 Quad Cities Unit 1 EOC 2 Gamma Scan Comparison-- 184 Bundle ID: GEH029

LIST OF FIGURES (continued)

Figure Number Title Page 3.4.1 Peach Bottom Unit 2 Cycles 1 and 2 Relative Nodal RMS 187 of TIP Response Comparisons 3.4.2 Peach Bottom Unit 2 Cycle 1 Average Axial TIP Response 188 Comparison 11.133 GWD/MTU Core Average Exposure 3.4.3 Peach Bottom Unit 2 Cycle 1 Radial TIP Response 189 Comparisons 11.133 GWD/MTU Core Average Exposure 3.4.4 Peach Bottom Unit 2 Cycle 1 Individual TIP Response 190 Comparisons 11.133 GWD/MTU Core Average Exposure 3.4.5 Peach Bottom Unit 2 Cycle 2 -- Average Axial TIP Response 191 Comparison -- 13.812 GWD/MTU Core Average Exposure 3.4.6 Peach Bottom Unit 2 Cycle 2 Radial TIP Response 192 Comparisons 13.812 GWD/MTU Core Average Exposure 3.4.7 Peach Bottom Unit 2 Cycle 2 Individual TIP Response 193 Comparisons 13.812 GWD/MTU Core Average Exposure 3.4.8 Peach Bottom Unit 2 End of Cycle 2 Core Average Axial 194 Power Distributions 4.3.1 CPM-2 vs. PDQ7 Pin Power Distribution Comparison GE 202 Initial Core High Enriched Fuel Type Uncontrolled 4.3.2 CPM-2 vs. PDQ7 Pin Power Distribution Comparison -- GE 203 Initial Core High Enriched Fuel Type Controlled 4.3.3 CPM-2 vs. PDQ7 Pin Power Distribution Comparison GE 204 Initial Core Medium Enriched Fuel Type Uncontrolled 4.3.4 CPM-2 vs. PDQ7 Pin Power Distribution Comparison GE 205 Initial Core Medium Enriched Fuel Type -- Controlled

I I

l

1.0 INTRODUCTION

Pennsylvania Power & Light Company (PP&L) operates the two unit Susquehanna Steam Electric Station (SES) near Berwick, Pennsylvania. Both of the Susquehanna SES reactors are General Electric Company Boiling Water Reactor (BWR) -4 product line reactor systems; each has a rated thermal power output of 3293 Megawatts. The general core design and operating features are given in Table 1.1, Figure 1.1, and Figure 1.2.

The purpose of this report is to describe the steady state core physics methods used by PP&L for BWR core analysis and to provide qualification of the analytical methodologies which will be used to perform safety related analyses in support of licensing actions. This report will satisfy the guidelines in Reference l.

PP&L's steady state core physics methods are based on the Electric Power Research Institute (EPRI) code package (Reference 2), as depicted in the flowchart contained in Figure 1.3. The main computer codes are the CPM-2/PP&L (hereafter referred to as CPM-2) fuel bundle lattice physics depletion code and the SIMULATE-E/PP&L (hereafter referred to as SIMULATE-E) three-dimensional core simulation code. Both of these codes represent state-of-the-art techniques for reactor analysis and are described further in Sections 2.1 and 3.1, respectively. The MICBURN/PP&L code (hereafter referred to as MICBURN) provides a detailed representation of the depletion of a single gadolinia (Gd 0 ) bearing fuel pin; the NORGE-B2/PP&L code (hereafter referred to as NORGE-B2) provides a nuclear cross section data link from CPM-2 into SIMULATE-E as well as the POWERPLEX core monitoring system. The PDQ7 code, linked to CPM-2 via the COPHIN program, is an industry standard diffusion theory simulation used by PP&L for special applications. TIPPLOT provides

.plotting and statistical analysis capabilities. The RODDK-E/PP&L code is used to determine control rod worth for shutdown margin analyses and to estimate core shutdown margin.

PP&L utilizes the above mentioned cOdes and associated methodologies for plant operations support applications (e.g., core follow analyses, development of

target control rod patterns, predictions of startup critical rod patterns, operating strategy evaluations, etc.), independent design verification calculations, reload fuel/core design analyses, safety analyses, and core monitoring system data bank updates. The steady state core physics methods described in this report are also used to develop the necessary neutronics data input to PPGL's transient analyses.

The qualification of PPGL's steady state core physics methods is based largely on comparisons of calculated core parameters to measured data from the Susquehanna SES Units l and 2, Peach Bottom Unit 2, and Quad Cities Unit 1 reactors. All of the model preparation and benchmarking calculations represent work performed by PPGL. The computer codes and the calculations supporting this work are documented, reviewed, and controlled by formal procedures which are encompassed within PPGL's nuclear quality assurance program.

TABLE 1 1 GENERtG DESIGN AND OPERATING FEATURES OF THE SUS UEZGQlNA SES REACTORS Reactor Type/Configuration: BWR-4/2 Loop Jet Pump Recirculation System Rated Core Power: 3,293 HW Thermal 6

Rated Core Flow: 100x10 ibm/hr Reactor Pressure at Rated Conditions: 1020 psia Number of Fuel Assemblies: 764 Number of Control Rods: 185 Number of Traversing Zn-core Probe Locations: 43

FIGURE 1.1 SUSQUEHANNA SES UNlTS 1 AND 2 CORE 59 57 55 ++++++++

53

++++++++

43 41 39 37 35 33

'27 25 23 ++++++++

21

++++++++

++++++++

++++++++

3 1

000204060810 12 14 16 18 2022 24 262830323436384042444648505254565860 X

+ Control Rod Location

~ Traversing In core Probe l ocation

FIGURE 1.2 TYPICAL GORE POWER VS CORE FLOW 120 110

/

100

/

APRM SORY~:

90 APRM 100% Xe ROD BLOCK ROD LINE r'

80'5 r

ROD BLOCK MONITOR I /

7o V

Ol

/

//

6o ~T O

/

/

/

I O 50 I ~

I O I I

I I

40 30 20 ~ ~

NAT CIRC 2-PUMP MIN FLOW!

10 0

0 10 20 I

30 40 50 I

60 I

70 80 90,;00 TOTAL CORE FLOW, 5 RATED

FIGURE 1.3 PP&L STEADY STATE CORE PHYSICS METHODS COMPUTER CODE FLOWCHART MICBURN Gd Depletion CPM-2 Lattice Physics COP HIN Date Link NORGE-B2 Data Link POWER PLEX PDQ Core Monitoring System Diffusion Theory SIMULATE-E 3-D Simulation TIPPLOT RODDK-E Statistical Analysis Shutdown Margin TRANSIENT ANALYSIS

2.0 LATTICE PHYSICS METHODS The lattice physics methods currently in use at PPGL are based on the CPM-2 and MICBURN computer codes which were originally developed by EPRI as part of the Advanced Recycle Methodology Program (Reference 2). CPM-2 is used at PPGL to calculate the two energy group cross sections for input to SIMULATE-E and POWERPLEX. The code is also used to provide detector model response data which is used by SIMULATE-E to determine calculated Traversing In-core Probe (TIP) responses. The calculated TIP responses are routinely compared to measured TIP data to assess nodal model accuracy and to provide the Rod Block Monitor (RBM) simulation employed for certain safety analyses (e.g., Rod Withdrawal Error) .

A full description of CPM-2 is provided in Reference 3 but is also summarized in Section 2.1. Sections 2.2 and 2.3 provide comparisons to both uniform lattice critical and reactor operating data. Several uniform lattice critical calculations were performed at PPGL to determine the accuracy of the reactivity calculation. Additional comparisons have been made to pin gamma scan measurements from the Quad Cities Unit 1 reactor to benchmark the pin power distribution.

In addition to PPGL calculations, EPRI sponsored extensive benchmarking of the code (Reference 4) which was performed during the original development of EPRI-CPM (Reference 5). Further development at S. Levy (under EPRI contract) vastly simplified the required user input. This modified version of the computer program is distributed by EPRI as CPM-2. The improvements in the input module greatly reduce the possibility of input errors since only physical dimensions and design values are required for input. CPM-2 generates all required number densities and determines appropriate thermal expansions.

Only the input module was changed leaving the neutronics calculations identical to the original EPRI-CPM. Further modifications have been made at PPGL to have the code conform to our computer system operational requirements as well as to provide additional calculational outputs. These modifications have not resulted in any changes to the neutronics calculation. Therefore, all EPRI benchmarking on the original EPRI-CPM remains applicable to the version of CPM-2 used at PPGL. Section 2.4 summarizes the EPRI benchmarking results.

2.1 Descri tion of CPM-2 The CPM-2 computer code was developed for analysis of both BWR and PWR fuel assemblies. The code performs a two-dimensional calculation which permits explicit modeling of fuel pins, water rods, a fuel channel, wide and narrow water gaps, control elements, and in-core instrumentation tubes. The neutronics calculation solves the integral neutron transport theory equation by the method of collision probabilities.

Figure 2.1.1 presents the normal calculational flow for a BWR fuel assembly.

The calculation consists of four basic parts. The resonance calculation is performed first to determine effective microscopic cross sections in the resonance region. The micro-group calculation is performed next for each different type of pin cell and the resulting detailed energy group spectra are then used to collapse the 69 energy group cross sections into several broad groups. The macro-group calculation uses these broad group cross sections to determine the neutron spectra across an assembly converted to one-dimensional cylindrical geometry. This spectra is used to further reduce the number of I

energy groups to be used in the final two-dimensional calculation.

The resonance calculation is used to provide effective cross section data in the resonance region between 4 eV and 9118 eV. All resonance absorption above this limit is treated as unshielded. The large resonances in Pu-240 at 1.0 eV and in Pu-239 at 0.3 eV are adequately treated in the detailed thermal spectra calculation by the larger number of thermal groups around each of these resonances. The nuclides treated in the resonance calculations are U-235, U-236, U-238 and Pu-239.

The resonance calculation makes use of tabulated resonance integrals for a homogeneous mixture. These are converted to correspond to the heterogeneous geometry through use of the equivalence theorem. The nuclear data library contains tables of the homogeneous integrals for the resonance nuclides as a function of fuel temperature and potential scattering cross section. The fuel temperature used is the effective Doppler temperature for the mixture. Fuel collision probabilities used during the resonance integral evaluation are approximated using the Carlvik approximation (Reference 6). Once effective

resonance cross sections are calculated for absorption and fission, they are modified to correct for resonance overlap. Dancoff correction factors are then calculated for each pin and used to correct the effective cross sections to account for the effects of rod shadowing.

The micro-group calculation is performed in 69 energy groups shown in Table 2.1.1 for each different type of pin in the assembly being modeled. Pins are differentiated by type (i.e., water rod, fuel rod, absorber rod, etc.). Fuel rods are further differentiated by fuel material, enrichment, pellet or rod dimensions, etc. Each micro-group calculation models a single pin in one-dimensional cylindrical geometry. For fuel pins, separate regions are used for fuel, cladding, and moderator. An extra region is placed around the pin cell and is used to account for the fuel channel wall and the water gaps.

For absorber and water rods, separate regions are included for 'the absorber or water region, cladding, and moderator. A buffer region consisting of homogenized average fuel cells with a thickness of 2.5 mean free paths is placed around the absorber cell. This is used to provide a reasonable neutron spectrum incident on the non-fuel cell. The micro-group calculation is used to provide a detailed energy spectrum which is used to collapse the 69 group cross sections to fewer groups averaged over each pin cell. This is necessary since a two-dimensional calculation in 69 energy groups is not practical.

When homogenizing cross sections over a pin cell for an absorber pin, the average cross sections will result in an overestimation of the thermal flux in subsequent homogeneous calculations. This will cause a corresponding overestimation of the absorber worth. For absorber pin cells, two calculations are performed. The first calculation uses the heterogeneous geometry as previously discussed. The second calculation is for a homogenized absorber pin cell. Correction factors are calculated for each energy group as the ratio of the heterogeneous problem flux to that of the homogeneous problem. These factors are used to correct the two-dimensional fluxes in the final calculation so that reaction rates and reactivity are conserved.

Following the micro-group calculation, a macro-group calculation is performed.

In this calculation, the fuel assembly is converted to one-dimensional cylindrical geometry. Each concentric row of pins, starting from the assembly

center and proceeding outward, occupies one cylindrical shell. The fuel channel wall, water gap and control rod (if present) also occupy one shell each. This calculation is performed in 25 energy groups using the collapsed cross section data from the micro-group calculation. The energy group structure is given in Table 2.1.2. This calculation is used to determine the energy spectra in each region to further collapse the cross section data. By performing this calculation, fewer energy groups are necessary in the two-dimensional calculation because the effects of the water gaps are taken into account.

The final two-dimensional calculation in CPM-2 solves the integral transport equation in X-Y Cartesian coordinates using the method of collision probabilities. This calculation is used to determine the multigroup flux across the assembly, local pin power distribution, and the assembly eigenvalue. The pin cells, channel wall, water gaps and control rod are represented. Diagonal symmetry is assumed as shown in Figure 2.1.2. The calculation is performed in the five energy groups shown in Table 2.1.2 using cross section data collapsed from the macro-group calculation. Collapsed two group cross section data averaged over the fuel assembly are then used in SIMULATE-E and POHERPLEX. Few group cross section data can also be determined over specified regions to provide input to PDQ7.

For fuel rods that contain gadolinia, special calculations are performed with MICBURN (Reference 7) to account for the spatial shielding of the absorber.

This calculation is used to provide effective microscopic cross sections for gadolinia in 69 energy groups for use in CPM-2. MICBURN models only the burnable absorber pin cell. The gadolinia fuel rod is usually modeled using ten mesh points to provide sufficient detail to calculate the radial flux distribution. These fluxes are expanded to 20 radial zones for the actual gadolinia depletion. From the calculation, effective gadolinia cross sections are obtained for use in CPM-2. These are tabulated as a function of the fraction of Gd-155 plus Gd-157 remaining in the pin.

The fuel depletion algorithm in CPM-2 utilizes a predictor-corrector methodology. In the predictor step, the fluxes from the two-dimensional calculation from timestep t are used to deplete the nuclide inventories to n 1 10-

timestep tn . A new flux calculation at timestep tn is performed using the predicted nuclide inventory. Once these fluxes are known, the depletion t

chains are reevaluated from timestep n 1 to tn (i.e., corrector step). The final number densities used at timestep tn are the average of the results from the predictor and corrector steps. The primary heavy nuclides plus 22 fission products are explicitly tracked. The remaining fission products are tracked using two pseudo-isotopes which are used to represent non-saturating and slowly saturating fission products. The list of heavy nuclides tracked in CPM-2 is provided in Table 2.1.3 and the fission products are shown in Table 2.1.4.

The nuclear data library (Reference 8) used in CPM-2 was developed and benchmarked with the original EPRI-CPM program (Reference 4). The data library was generated from ENDF/B-III data with modifications based on benchmarking studies. The sixty-six elements shown in Tables 2.1.3 and 2.1.4 are represented in 69 energy groups. These are divided into 27 fast and 42 thermal groups. The energy group structure was defined with a significant number of energy groups around the 0.3 eV Pu-239 and 1.0 eV Pu-240 resonances.

This permits treatment of these resonances during the thermal group calculation without the need for a specific resonance calculation.

Several modifications were made to the ENDF/B-III data library based on extensive EPRI benchmarking (Reference 8). The principal modification to the library is a uniform reduction of the U-238 microscopic absorption cross sections in the resonance region based on Hellstrand's measurements on isolated rods (Reference 9). This modification for U-238 is within the data uncertainties in the ENDF-B/III data. Other modifications are listed in Table 2.1.5. The reduction of the Pu-240 absorption cross section was necessary to account for shielding of the higher energy resonances which is not treated in the resonance calculations.

11

TABLE 2 1 1 SIXTY-NINE GROUP ENERGY BOUNDARIES FOR THE CPM & MICBURN CROSS SECTION LIBRARY Energy Energy Energy Group Boundary ~GZOU Boundary ~GUOU Bo~dazar MeV eV eV 1 10.00 24 48.052 52 0.280 2 6.0655 25 27.700 53 0.250 3 3.679 26 15.968 54 0.220 4 2.231 27 9.877 55 0.180 5 1.353 28 4.00 56 0.140 6 0.821 29 3.30 57 0.100 7 0.500 30 2.60 58 0.080 8 0.3025 31 2.10 59 0.067 9 0.183 32 1.50 60 0.058 10 0.1110 33 1.30 61 0.050 ll 0.06734 0.04085 34 1.15 62 0.042 12 35 1.123 63 0.035 13 0.02478 36 1.097 64 0.030 14 0.01503 37 1. 071'.045 65 0.025 38 66 0.020 39 1.020 67 0.015 40 0.996 68 0.010

--eV 41 0.972 69 0.005 42 0.950 0.0 15 9118.0 43 0.910 16 5530.0 . 44 0.850 17 3519.1 45 0.780 18 2239.45 46 0.625 19 1425.1 47 0.500 20 906.898 48 0.400 21 367.262 49 0.350 22 148.728 50 0.320 23 75.501 51 0.300 Resonance region consists of groups 15 through 27.

Source: M.

of Edenius, et. al., "The EPRI-CPM Data EPRI CCM-3, November, .1975.

Library," Part II, Chapter 4 12-

TABLE 2.1.2 ENERGY GROUP STRUCTURE FOR MACRO-GROUP AND TWO-DIMENSIONAL CALCULATIONS Macro-Group Calculation 2-D Group Calculation Fine Energy Macro Energy

~Grou $ $ Boundaries Group ~Grou $ Boundaries MeV --eV--

1-2 10.0 - 3.679 '1 1-8 10.0 x 10 6 - 5.530 X 10 3

3-4 3.679 - 1.353 2 9-17 5.530 x 10 6.25 x 10 5 1.353 0.821 3 18-20 6.25 x 10 1.80 x 10 6 0.821 0.500 4 21-22 1.80 x 10 1 5.00 x 10 2 7-8 0.500 0.183 5 23-25 5.00 x 10 0.0 9-10 0.183 - 0.06734 11-12 0.06734 0.02478 13-15 0.02478 0.005530

'-eV--

16-18 5530 1425.1 19-21 1425.1 148.728 22-25 148.728 15.968 I-'8 15.968 9.877 I 26 27 28-31 32-35 9.877 4.00 1.50 4.00 1.50 1.097 36-38 1.097 1.020 39-45 1.020 0.625 46-48 0.625 0.350 i 19 21 49-51 52-54 55-57 58-60 0.350 0.280 0.180 0.080 0.280 0.180 0.080 0.050 r 61-63 64-66 67-69 0.050 0.030 0.030 0.015 25 0.015 0.0 I

13

TABLE 2.1.3 Heavy Nuclide Chains

1. U235 ~ U236 ~ Np237 ~ Pu238

2. U238 ~ Pu239 ~ Pu240 ~ Pu241 ~ Pu242 ~ Am243 ~ Cm244 (25%)

3. U238 ~ Pu239 ~ Pu240 ~ Pu241 ~ Am241 ~ Am242m ~ Am243 ~ Cm244 (75%)

4. U238 ~ Pu239 ~ Pu240 ~ Pu241 ~ Am241 ~ Cm242 ~ Pu238 (n,2n)

5. U238 ~ Np237 ~ Pu238 Source: A. Ahlin, et. al, "The Collision Probability Module EPRI-CPM,"

Part II, Chapter 6 of EPRI CCM-3, November, 1975.

14-

TABLE 2 1.4 FISSION PRODUCT CHAINS

1. Kr83

2. Rhl03

3. Rh105

4. Ag109

5. Xe131

6. Cs133 ~ Cs134

7. Xe135 ~ Cs135

8. Nd143

9. Nd145 (52.77%)

10. Pm147 ~ Pm148 ~ Sm149 ~ Sm150 ~ Sm151 ~ Sm152 ~ Eu153 ~ Eu154 ~ Eu155 (47.23%)

11. Pm147 ~ Pm148m.~ Sm149 ~ Sm150 ~ Sml51 ~ Sm152 ~ Eu153 ~ Eu154 ~ Eu155

12. Pm147 ~ Sm147

14. Slowly-Saturating Fission Products Source: A. Ahlin, et. al, "The Collision Probability Module EPRI-CPM,"

Part II, Chapter 6 of EPRI CCM-3, November, 1975.

15

TABLE 2.'1.5 MODIFICATIONS TO ENDF-B/III DATA FOR CPM-2 CROSS SECTION LIBRARY Nuclide Cross Section Modification U-238 Increased by 8% for groups 1 through 5 cr and vE Increased by 4.5% for groups 1 through 5 a,g'g Reduced by 30% for group 4 a,g'+g Reduced by 20% for group 5 RI Resonance integral reduced by 0.3 1 RI TO p

where T the group lethargy width a the group potential scattering cross section RI g

Resonance integral from ENDF-B/III data RI effective group resonance integral in CPM library PG-240 a a

Reduced by 50% for. groups 16 through 27

FIGURE 2.1.1 CALCULATIONALFLOW IN CPM-2 INPUT RESTART FILE RESONANCE DATA CALCULATION LIBRARY CALC MACROSCOPIC M!CHURN CROSS SECTIONS I

MICRO GROUP CALC 69 ENERGY GROUPS CONDENSE TO 25 MACRO GROUPS HOMOGENIZE TO MACRO REGIONS MACRO GROUP, CALC IN ANNULAR GEOMETRY CONDENSE.TO 5 GROUPS CALC CROSS SECT FOR 2-0 REGIONS 2-D ASSEMBLY CALCULATION CALC FEW GROUP CONSTANTS AND REACTION RATES BURNUP CORRECTOR ZERO BURNUP BURNUP P R ED I CTO 8 END 17

FIGURE 2.1.2 EXAMPLE OF BWR CELL GEOMETRY IN THE 2-D CALCULATION STEEL CONTROL ROD WIDE WATER GAP FUEL PIN CELL INNER WATER GAP CHANNEL NARROW WATER GAP IN-CORE DETECTOR

2.2 Uniform Latter.ce Criticals One method to determine the accuracy of the reactivity calculation in CPM-2 is through comparison to uniform lattice critical measurements. The test assembly contains fuel pins with a single enrichment moderated by water at room temperature and atmospheric pressure. A sufficient number of fuel pins is added to the assembly until .criticality is achieved. Radial and axial buckling, which are input to the CPM-2 analyses, are determined from the measured data.

The uniform lattice critical experiments chosen for analysis were obtained from the Westinghouse TRX (Reference 10) and ESADA (Reference 11) criticals.

The TRX criticals that were analyzed by PP&L with CPM-2 are the eight UO 2

experiments. The rod enrichment for all eight experiments was 1.3 weight percent U-235 with UO pellet densities of 7.52 g/cm 3 for two measurements, 7.53 g/cm 3 for three measurements, and 10.53 g/cm 3 for the remaining three.

Water-to-metal ratios varied from 3.0 to 5.0. The conditions are summarized in Table 2.2.1. Six of the ESADA criticals were analyzed by PPGL. All of these contained 2.0 weight percent PuO in natural uranium. Zn four experiments, eight. percent (by weight) of the plutonium was Pu-240; in the remaining two, twenty-four percent (by weight) of the plutonium was Pu-240. A summary of the conditions is given in Table 2.2.2.

The CPM-2 calculated assembly K-effectives are provided in Tables 2.2.3 and 2.2.4 for the TRX and ESADA criticals, respectively. The CPM-2 calculated K-effectives for the ESADA criticals have been corrected by -0.4% ak to account for the presence of spacers in the core. An additional correction for self-shielding of the plutonium grains has not been included. This correction varies from -0.05% to -0.45% ~k. The calculated average K-effective from all criticals is 1.0005 with a standard deviation of, 0.0072.

TABLE 2.2.1 TRX UNIFORM LATTICE CRITICAL TEST DATA Pellet Experiment U-235 Densi)y Diameter Water to Critical Number Identification (wt. 4) (g/cm ) (in.) Metal Ratio of Fuel Rods TRX1 1.3 7.53 0.601 1269+3 TRX2 1.3 7.53 0.601 1027+3 TRX3 1.3 7.53 0.601 987+3 TRX4 1.3 7.52 0.388 3045+3 TRX5 1.3 7.52 0.388 2784+3 TRX6 1.3 10.53 0.383 2173+3 TRX7 1.3 10.53 0.383 3.6 1755+3 TRX8 1.3 10.53 0.383 1575+3 Source: J. R. Brown, et.al., "Kinetic and Buckling Measurements on Lattices of Slightly Enriched Uranium or UO Rods In Light Water," WAPD-176, January, 1958.

20

TABLE 2.2.2 ESADA UNIFORM LATTICE CRITICAL TEST DATA Lattice Experiment Pu-240 Pitch Critical Number Identification (wt:. %) (in) of Rods ESADA 1 0.69 514 ESADA 3 0.75 321 ESADA 4 0.9758 160 ESADA 6 1.0607 152 ESADA 12 24 0.9758 247 ESADA 13 24 1.0607 243 Source: R. D. Learner, et. al. "PuO -UO Fueled Critical Experiments,"

WCAP-3726-1, July, 1967.

21

Table 2.2.3 CPM-2 RESULTS FOR TRX CRITICALS Experimental Experiment Material chuckling CPM-2 Identification (m ). K-effective TRX1 28.37 0.9934 TRX2 30.17 0.9958 TRX3 29.06 0.9942 TRX4 25.28 0.9939 TRX5 25.21 0.9934 TRX6 32.59 0.9974 TRX7 35.47 0.9970 TRX8 32.22 0.9960 Average K-effective = 0.9951 Standard Deviation = 0.0016 22-

TABLE 2 2 4 CPM-2 RESULTS FOR ESADA CRITICALS Experimental Experiment Pu-240 Material Buckling CPM-2 Identification (wt. 4) (m ) K-effective*

ESADA 1 69.6 1.0026 ESADA 3 90.0 1.0004 ESADA 4 8 104.72 1.0129 ESADA 6 8 98.4 1.0116 ESADA 12 24 79.5 1.0101 ESADA 13 24 73.3 1.0077 Average K-effective = 1.0076 Standard Deviation = 0.0050

• All CPM-2 calculated K-effectives have been adjusted by -0.4% ~k to account for spacer worth.

23

2.3 uad Cities Pin Power Distribution Com arisons Additional verification of CPM-2 performed by PP&L includes comparisons to the pin gamma scan measurements from Quad Cities Unit 1 at the end of Cycle 2 (Reference 12). In 1976 under EPRI sponsorship, General Electric performed detailed gamma scan measurements at Quad Cities. These measurements included pin-by-pin gamma scan measurements for six separate assemblies which included three mixed oxide (MO ) and three UO bundles (see Table 2.3.1) . Each bundle was. disassembled and scanned at eight separate axial locations. No tie rods or spacer capture rods from any bundle were scanned and only nine rods from bundle GEB161 were scanned. The measured La-140 intensities were corrected to correspond to activity. at shutdown. The practical accuracy of the reported data including measurement uncertainty and measurement method bias is approximately 3.0$ (Reference 12, Section 4.3).

The gamma scan data itself is a measure of La-140 gamma activity. During reactor operation, La-140 is produced both as a fission product and by Ba-140 decay. Since the half-life of Ba-140 is approximately 13 days and that, of La-140 is approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, the distribution of the Ba-140 and La-140 concentrations will be representative of the power distribution integrated over the last two to three months of reactor operation. After shutdown, the only source of La-140 is from decay of Ba-140. Because the half-life of La-140 is short with respect to Ba-140, after about ten days the decay rate of La-140 is controlled by the decay of Ba-140. Therefore, the relative measured La-140 activities are compared to the relative calculated Ba-140 concentrations, and the La-140 concentration does not need to be calculated.

The local power distributions calculated by CPM-2 were converted to relative Ba-140 concentrations prior to the comparison. The SIMULATE-E code was used to calculate the exposure and void history conditions for each bundle and axial elevation for which measurement data existed. The CPM-2 calculated relative Ba-140 concentrations for each pin were then determined for each of these conditions.

Bundle GEB162 was located on the core periphery. Consequently, a steep neutron flux gradient existed across the bundle. In CPM-2, a zero current 24

boundary condition is assumed to exist. This is reasonable for interior bundles but will cause large errors for peripheral bundles, particularly for those pins adjacent to the reflector region. Because peripheral bundles are low power bundles and do not operate close to thermal limits, high accuracy is not necessary. Therefore, comparisons to the GEB162 bundle are not included in the results.

A pin comparison is defined as a comparison between the relative measured and calculated La-140 activities for all scanned pins at a specific axial location within a given bundle. For each comparison, the calculated and measured La-140 activities are normalized to 1.0 based on the number of pins for which there were measurements. Samples of these pin comparisons are presented in Figures 2.3.1 through 2.3.8. A difference between the measured and calculated normalized La-140 activities for each pin is calculated as:

i

e. = c.

i - m.i where i

m. = the normalized measured La-140 activity for fuel pin i,

c. = the normalized calculated La-140 activity for fuel pin i.

The standard deviation for each pin comparison is calculated as:

N g(e. - e) 100 x

N-1 M where M = the average of the normalized measured data for the comparison

= 1.0 for all comparisons due to normalization, e = the average difference between the measured and calculated normalized La-140 activities

= 0.0 for all comparisons due to normalization, N = the number of pins in the comparison.

A summary of the standard deviations for each of the comparisons is given in Table 2.3.2. The average standard deviation for all comparisons is 4.00%. Zf only UO bundles are compared, the average standard deviation is only 3.37%.

25

Assuming the standard deviations are due to a combination of independent.

measurement and calculational uncertainties, the calculational standard deviation can be determined from the following equation:

2 2 2 6 0 + c total calc meas where c

total1 = total standard deviation from the comparisons 6

calc

= calculational standard deviation, and a = measurement standard deviation meas Assuming a measurement accuracy of 3.0%, the calculational standard deviation is 2.6% for all bundles or 1.5% for UO bundles only.

The CPM-2 code is also used to calculate the Local Peaking Factor (LPF) for each lattice type. The LPF is the ratio of the'maximum pin power in a six-inch segment to the average pin power in the same six-inch segment, of a fuel assembly. An accurate calculation is important because the local peaking factor is input to the core monitoring system and SIMULATE-E and is used to determine the linear heat. generation rate. Because La-140 activity is proportional to the pin power distribution, an estimate of the LPF can be made from the gamma scan measurements. A comparison between the measured and calculated ratios of the peak pin La-140 activity to average pin La-140 activity is presented in Table 2.3.3. The average difference from all of the comparisons is 2.49%. If only the UO fuel bundles are included, the average difference becomes 0.98%. As shown in Table 2.3.3, most of the CPM-2 calculations result in an overestimation of peak La-140 activity, and, therefore, conservatively estimate the linear heat generation rate.

26-

TABLE 2.3.1 ASSEMBLIES USED IN ROD TO ROD GAMMA SCAN Assembly Location Number of Identification Bundle ~(x. ) Rods Scanned GEB159 7x7 MO Center Design 31/32 40 GEB162 7x7 M02 Peripheral Design 5,48 40 GEB161 7x7 M02 Center Design 29,32 9 GEH002 8x8 UO Reload Core Design 13,36 55 CX0672 7x7 UO Initial Core Design 15,36 40 CX0214 7x7 UO Initial Core Design 33,34 40 27

Table 2.3.2 QUAD CITIES UNIT 1 END OF CYCLE 2

SUMMARY

OF NORMALIZED LA-140 ACTIVITY PIN COMPARISIONS ASSEMBLY AXIAl CALCULATED VOID CALCULATED STANDARD ID ELEVATION ( IN) HISTORY (X) BURNUP (GWD/MTU) DEV (X)

GEB159 15 2.9 11.53 3.94 GEB159 21 8.0 11.89 4.21 GEB159 51 39.1 10.24 3.79 GEB159 56 43.5 10.16 4.42 GEB159 87 58.7 9.88 4.38 GEB159 93 60.7 9.86 4.77 GEB159 123 67.9 7.79 5.89 GEB159 129 68.8 6.43 5.90 GEB161 15 2.9 11.59 2.61 GEB161 21 8.'1 11.94 2.44 GEB161 51 39.3 10.25 4.48 GEB161 56 43.6 10.17 4.81 GEB161 87 58.7 9.89 5.81 GEB161 93 60.8 9.87 6.24 GEB161 123 68.0 7.79 7.53 GEB161 129 68.9 6.43 7.84 GEH002 15 2.8 10.74 2.97 GEH002 21 7.7 11.09. 2.38 GEH002 51 37.4 9.88 2.45 GEH002 56 41.7 9.75 2.27 GEH002 87 57.0 9.50 2.59 GEH002 93 59.1 9.46 2.68 GEH002 123 66.7 7.53 2.15 GEH002 129 67.9 6.27 2.25 CX0672 15 0.3 15.86 5.24 CX0672 21 3.4 17.42 5 ~ 02 CX0672 51 26.0 20.25 3,51 CX0672 56 31:0 19.56 3.62 CX0672 87 50.5 19.07 3.41 CX0672 93 53.0 18.90 3.76 CX()672 123 61.1 14.97 3.54 CX0672 129 61.9 12.65 3.55 CX0214 15, 1.8 16.01 4.08 CX0214 21 4.3 17.50 5.04 CX0214 51 29.3 19.54 2.87 CX0214. 56 34.0 19.47 3.60 CX0214 87 51.1 19.45 2.90 CX0214 93 53.7 19.09 3.83 CX0214 123 63.4 14.78 3.61 CX0214 129 64.2 12.58 3.54 OVERALL AVERAGE: 4.00 STANDARD DEVIATION: 1.40 U02 BUNDLE AVG: 3.37 STANDARD DEVIATION: 0.88 M02 BUNDLE AVG: 4.94 STANDARD DEVIATION: 1.52 28-

TABLE 2 3 3 QUAD CITIES UNIT 1 END OF CYCLE 2 PEAK LA-140 ACTIVITY COMPARISONS Axial Assembly Elevation Measured Peak Calculated Peak Difference Identification (IN) La-140 Activity* La-140 Activity* (@)

GEB159 15 1.137 1.249 9.85 21 ., 1.115 1.186 6.37 51 1.116 1.166 4.48 56 1.099 1.156 5.19 87 1.103 1.136 2.99 93 1.102 1.135 2.99 123 1.134 1.172 3.35 129 1.159 1.202 3.71 GEB161 15 1.115 1.139 2.15 21 1.110 1.137 2.43 51 1.107 1.158 4.61 56 1.089 1.160 6.52 87 1.094 1.170 6.95 93 1.110 1.171 5.50 123 1.131 1.194 5.57 129 1.172 1.212 3.41 GEH002 15 1.103 1.133 2.72 21 1.099 1.129 2.73 51 1.110 1.124 1.26 56'7 1.100 1.116 1.45 1.118 1.113 "0.45 93 1.119 1.120 0.09 123 1.135 1.139 0.35 129 1.135 1.147 1.06 CXO672 15 1.106 1.124 1.63 21 1.080 1.119 3.61 51 1.097 1.096 -0.09 56 1.098 1.100 0.18 87 1.096 1.094 -0.18 93 1.071 1.092 1.96 123 1.088 1.111 2.11 129 1.101 1.119 '.63 CX0214 15 1.108 1.125 1.53 21 1.078 1.115 3.43 51 1.091 1.103 1.10 56 1.066 1.097 2.91 87 1.126 1.093 -2.93 93 1.123 1.092 -2.76 123 1.131 1.119 -1.06 129 1.114 1.127 1.17 Average Difference: 2.49%

Average Difference (UO Bundles Only): 0.98%

Average Difference (M02 Bundles Only): 4.75%

• Peak La-140 Activity = ratio of the peak pin La-140 activity to the average pin La-140 activity in an axial segment of a fuel bundle.

29-

FIGURE 2.3.1 QUAD CITIES UNIT 1 EOC 2 GANESA SCAN COhPARISION NORhM IZED LA-140 PIN ACTIVITIES ASSEhSLY ID: GEB159 93 IN. FROM BOTTOM OF CORE Wide Wide Gap 1.075 1.068 1.017 1.053 1.093 h/eas 1.111 1.101 1.063 1.090 1.104 Calc 0.036 0.033 0.046 0.037 0.011 Ca 1 c-Meas 1.050 .968 1.042 1.093 1.082 1.003 1.010 1.101 0.994 1.040 1.122 1.110 1.012 1.005 0.051 0.026 .002 0.029 0.028 0.009 .005

l. 014 .860 .929 . 938 .848 1.040 0.822 0.875 0.888 0.786 0.026 .038 .054 .050 .062 1.021 1.079 .951 .976 1.090 1.086 1.063 1.122 0.875 0.909 0.980 1.088 0.042 0.043 .076 .067 . 110. 0.002 1,060 .925 1.002 .504 1.071 1.110 0.888 0 '09 0.564 1.024 0.050 .037 .093 0.060 .047 1.064 .993 .821 1.045 1.063 .740 1.092 1.090 1.012 0.786 0.980 1.024 0.726 1.135 0.026 0.019 .035 .065 .039 .014 0.043 1.102 1.006 1 '64 1.056 1.048 1.104 1.005 1.088 1.135 1.119 0.002 .001 0.024 0.079 0.071 VOID LEVEL (X): 60.7 BURNUP (GWD/hKU): 9.86 STANDARD DEVIATION: 4.77Ã (40 PINS)

X indicates either tie rod or spacer capture rod (not measured) 30

FIGURE 2.3.2 QUAD CITIES UNIT 1 EOC 2 GNM SCAN COMPARISION NORMALIZED LA-140 PIN ACTIVITIES ASSEMBLY ID:'EB161 56 IN. FROM BOTTOM OF CORE Wide Wide Gap 1.016 1.077 Meas 1.068 1.102 Calc 0.052 0.025 Calc-Meas 1.031 1.089 1.055 1.160 0.024 0.071

.980 .955 0.913 0.933

.067 .022 1.004 0.978

.026 1.040 0.978 -FCO404

.062 Q +Q Q

~~c 7gc

.808 0.812 0.004 VOID LEVEL (K): 43.6 BURNUP (GWD/MTU): 10.17 STANDARD DEVIATION: 4.81K ( 9 PINS)

X indicates either tie rod or spacer capture rod (not measured) 31

FIGURE 2.3.3 QUAD CITIES UNIT 1 EOC 2 GAhM SCAN COMPARISION NORM IZED LA-140 PIN ACI'IVITIES ASSEMBLY ID: GEH002 21 IN. FROM EYZIQM OF CORE Wide Wide Gap 0.996 1.032 1.053 1.030 1. 070 1. 045 Meas 0.969 1.011 1.013 1.007 1. 049 1.036 Calc

.027 .021 .040 .023 .021 .009 C-M

1. 033 0.995 1.078 1.038 1.020 1.032 1.099 1.013 1.011 0.969 1.068 1.027 1.018 1 '36 1.081 1 ~,008

,022 .026 .010 .011 .002 0.004 .018 .005 1.074 1.011 0.956 0.936 0.961 0. 994 1.,068 0.955 0.943 0.934 0.951 0.975

.006 .056 .013 .002 .010 .019 1.027 1.026 0.948 0.916 0.940 0.929 0.969 1.061 1.013 1.027 0.943 0.922 0.931 0.927 0.963 1.074

.014 0.001 .005 0.006 .009 .002 .006 0.013

1. 026 1.039 0. 934 0 '49 0.935 0.937 1.038 1.007 1.018 0. 934 0.931 0.937 0.954 1.064

,019 .021 0.000 .018 0.002 0.017 0.026 1.047 0.957 0.927 0.931 0.918 0.961 1.036 0.951 0.927 0.937 0.936 0.972

.011 .006 .000 0.006 0.018 0.011 1.054 1.079 0.992 0.955 0.932 0.939 0.979 1.088 1.049 1.081 0.975 0.963 0.954 0.972 0.990 1.129

.005 0.002 .017 0.008 0.022 0.033 0.011 0.041 1.028 0.995 1.034 1.011 1.072'.129 0.960 1.036 1.008 1.074 1.064 1 039

~

0.008 0.013 0.040 0 '53 :r>rOCO8 0.057 0.079 VOID LEVEL (X): 7.7 BURNUP (GWD/MHJ): 11.09 STANDARD DEVIATION: 2. 38/ (55 PINS)

X indicates either tie rod or spacer capture rod (not measured)

W indicates water rod 32

FIGURE 2.3.4 QUAD CITIES UNIT 1 EOC 2 GAhMA SCAN COhPARISION NORhM IZED LA-140 PIN ACTIVITIES ASSEMBLY ID: GEH002 93 IN. FROM BOTTOM OF CORE Wide Wide Gap 1.085 1 '98 1.106 1.048 1.117 1.106 Meas 1.104 1.112 1.077 1.062 1.109 1.120 Calc 0.019 0.014 .029 0.014 .008 0.014 C-M 1.101 1.014 1.119 1.070 1.048 1.061 1.083 1.029 1.112 1.022 1.102 1.043 1.023 1.036 1.083 1.040 0.011 0.008 .017 .027 .025 .025 0.000 0.011 1.105 1.015 0.967 0. 934 0. 934 0.979 1.102 0.961 0.933 0.913 0. 924 0.946

.003 .054 .034 .021 .010 .033 1.079 1.060 0.945 0.915 0.920 0.912 0.937 1.062 1.077 1.043 0.933 0.895 0.891 0.882 .0. 917 1.052

.002 .017 .012 .020 .029 .030 .020 .010 1.063 1.009 0.913 0.902 0.902 0.915 1 '19 1.062 1.023 0.913 0.891 0.880 0.897 1.031

.001 0.014 .000 .011 .022 .018 0.012 1.040 0.924 0.897 0.913 0.876 0.910 1.036 0.924 0.882 0.880 0.874 0.910

.004 0.000 .015 .033 .002 0.000 1.092 1.075 0.969 0.898 0.889 '.895 0.913 1.039 1.109 1.083 0.946 0.917 0.897 0.910 0.929 1.100 0.017 0.008 .023 0.019 0.008 0.015 0.016 0.061 1.090 1.000 1. 015 1.002 1.035 0.954 1.120 1.040 1.052 1.031 ee4~c~e 1.100 1.046 0.030 0.040 0.037 0.029 gcQQQ>f+ 0.065 0 '92 VOID LEVIK (X): 59.1 BURNUP (GWD/hKU): 9.46 STANDARD DEVIATION: 2.68/ (55 PINS)

X indicates either tie rod or spacer capture rod (not measured)

W indicates water rod 33

FIGURE 2.3 '

QUAD CITIES UNIT 1 EOC 2 GAhQlA SCAN COMPARISION NORhM IZED IA-140 PIN ACTIVITIES ASSEMBfY ID: CX0672 21. IN. FROM BOTTOM OF CORE Wide Wide Gap 1.049 1.001 1.012 1.032 0. 938 Meas 0.958 0.909 0.977 0.997 0.940 Calc

.091 .092 .035 .035 0.002 Ca 1 c-Meas 0.971 1.007 0.911 1.066 1.059 1.050 0.981 0.909 0.932 0.902 1.039 1.046 1.071 0.979

.062 .075 .009 .027 .013 0.021 .002 0.946 1.053 0.991 0.986 1.009 0.902 1.006 0.987 0.994 1.034

.044 .047 .004 0.008 0.025 1.031 1. 074 1.002 0.981 1.032 1.001 0.977 1.039 0.987 0.977 0.991 1.098

.054 .035 .015 .004 .041 0.097 1.080 0. 984 0.971 0.914 1.010 1.046 0.994 0.977 0.981 1.020

.034 0.010 0.006 0.067 0.010 1.024 1.078 1 '29 0.990 0.996 0.983 1 '17 0.997 1.071 1.034 0.991 1.020 1.046 1.119

.027 .007 0.005 0.001 0.024 0.063 0.102 0.905 0.940 1.020 0.905 0.940 0.979 1.119 1.011 0.035 0.039 0.099 0.106 VOID lKVEf (X): 3.4 BURNUP (GWD/hfZU): 17.42 STANDARD DEVIATION: 5.02K (39 PINS)

X indicates either tie rod or spacer capture rod (not measured) 34

FIGURE 2.3.6 QUAD CITIES UNIT 1 EOC 2 GAhMA SCAN COMPARISION NORhM IZED IA-140 PIN ACTIVITIES ASSEMBIY ID CX0672 87 IN. FROM BOTlQM OF CORE Wide Wide Gap 1.067 1.000 0.972 1.040 0.993 h/eas 1.077 0.998 1.033 1.051 1.026 Calc 0.010 .002 0.061 0.011 0.033 Cal c-h1eas 0.992 0.983 0.919 1.062 1 '62 1.035 0.981 O.S98 0.966 0.923 1.028 1.028 1.053 1.013 0.006 .017 0.004 .034 .034 0.018 0.032 0.936 1.029 0.994 0.982 1.018 0.923 0.986 0.953 0.953 0.992

.013 .043 .041 .029 .026 1.016 1.054 0 '76 0.956 0.981 1.055 1.033 1.028 0.953 0.918 0.926 1.072 0.017 .026 .023 .038 .055 0.017 1.075 0.942 0. 948'.918 0.909 0.992 1.028 0.953 0.914 0.956

.047 0.011 .030 0.005 .036 1.012 1.036 .0.990 0.969 0.938 0.985 1.096 1.051 1.053 0.992 0.926 0.956 0.985 1.094 0.039 0.017 0.002 .043 0.018 .000 .002 0.984 0.932 1.049 1.007 1.036 1.026 1.013 1.072 1.094 1.038 .

0.042 0.081 0.023 0.087 0.002 VOID IZVEI (/): 50.5 BURNUP (GWD/hfQJ): 19.07 STANDARD DEVIATION: 3.41K (40 PINS)

X indicates either tie rod or spacer capture rod (not measured) 35

FIGURE 2.3.7 QUAD CITIES UNIT-'-1 EOC 2 GAhMA SCAN COhPARISION NORhNLIZED LA-140 PIN ACTIVITIES ASSEMBLY ID: CX0214 51 IN. FROM BOTTOM OF CORE Wide Wide Gap 1.051 0. 993 1 022

~ 0.971 habeas 1.021 0.959 1.005 0.989 Calc

.030 .034 .017 0.018 Calc-Meas 0.960 0. 984 0.933 1.054 1. 054 1.057 0.957 0.959 0,950 0.918 1.033 1.036 1.059 0.998

.001 .034 .015 .021 .018 0.002 0.041 0.941 1.024 1.001 0. 984 1.002 0.918 0.974 0.973 0.976 1.014

.023 .050 .028 .008 0.012 0.999 1.054 0.985 0. 948 1.010 1.035 1.005 1.033 0.973 0.951 0.961 1.083 0.006 .021 . 0.003 .049 0.048 012'.993 1.030 0.980 0.948 0.958 1.036 0.976 0.951 0.952 0.991 0.006 .017 .029 0.004 0.033 0.984 0.988 0.992 0.951 1.027 1.091 1.023 1.014 0.961 0.991 1.018 1.103 0.039 0.026 .031 0.040 .009 0.012 0.967 0.934 1.084 1.073 0.982 0.989 0.998 1.083 1.103 1.027 0.022 0.064 .001 0.030 0.045 VOID LEVEL (X): 29.3 BURNUP (GWD/hKU): 19.54 STANDARD DEVIATION: 2. 87K (38 PINS)

X indicates either tie rod or spacer capture rod (not measured) 36-

FIGURE 2.3.8 QUAD CITIES UNIT 1 EOC 2 GAhMA SCAN COMPARISION NORhM,IZED EA-140 PIN ACTIVITIES ASSEMBLY ID: CX0214 129 IN. FROM BOP%M OF CORE Wide Wide Gap 1.114 1.036 1.064 1.056 0.996 Meas 1.127 1.021 1.064 1.087 1.039 Calc 0.013 .015 0.000 0.031 0.043 Calc-Meas 1.026 1. 034 0.960 1.058 1.054 1.060 1.009 1.021 0.990 0.922 1.044 1.040 1.073 1.022

.005 .044 .038. .014 .014 0.013 0.013 0.969 1.005 0.938 0.927 0.970 0.922 0.988 0.937 0.931 0.976

.047 .017 .001 0.004 0.006 1,093 1.062 0.939 0.890 0.957 1.020 1.064 1.044 0.937 0.882 0.894 1.067

.029 .018 .002 .008 .063 0.047 1.046 0.956 0.906 0.891 0.919 1.040 0.931 0.882 0.874 0.921

~ 006 .025 .024 .017 0.002 1.050 1.047 0.956 0.967 0.931 0.969 1.006 1 '87 1.073 0.976 0.894 0.921 0.957 1.096 0.037 0.026 0.020 .073 .010 .012 0.090 1.062 0. 991 1.092 1.030 0.946 1.039 1.022 1.067 1.096 1.032

.023 0.031 .025 0.066 0.086 VOID IZVEL (X): 64.2 BURNUP (GWD/hfQJ): 12.58 STANDARD DEVIATION: 3.54/ (40 PINS)

X indicates either tie rod or spacer capture rod (not measured) 37

2.4 EPRI Benchmark Evaluations During the original development of EPRI-CPM, benchmarking calculations were performed against both uniform lattice critical tests and power reactor operating data. These include:

- hot critical data from the Kritz reactor, cold uniform lattice critical data from the TRX and ESADA criticals, and

- isotopic comparisons based on the post irradiation analysis of Yankee and Saxton spent fuel.

All calculations were made using the current version of the CPM cross section library and are documented in Reference 4. The results of those benchmarking comparisons are reported in this section.

Four experiments from the high temperature Kritz facility were modeled with EPRI-CPM to compare fission rates. The first three experiments involved one BNR and two PNR fuel lattices. All three lattices contained both mixed oxide and uranium oxide pins. The system temperature for these three experiments 0 0 was 245 C (473 F). The fourth experiment was a uniform lattice critical utilizing 1.35% enriched UO rods. Critical data was taken at 20 0 C and 210 0 C 0 0 (68 F and 410 F). Details concerning the experiments and calculations are given in Reference 4.

Measured and calculated fission rates for the first three Kritz experiments are reproduced from Reference 4 and shown in Figures 2.4.1 through 2.4.3. The fission rates were normalized so that the average 'of all measured pins was 1.0. In the third experiment, the UO and mixed oxide assemblies were normalized separately. The respective eigenvalues for each lattice are shown on the appropriate figures. The only results from the Kritz uniform lattice criticals were the eigenvalues. The calculated eigenvalues were 0.997 and 0.993 for the 20 0C and the 210 0 C criticals, respectively.

38-

Part of the EPRI-CPM benchmarking also included calculations of uniform lattice criticals from both the TRX and ESADA critical experiments. The results from these calculations are presented in Tables 2.4.1 and 2.4.2. The results reported by EPRI for the TRX criticals include a correction factor based on comparisons of the EPRI-CPM results to five group radial PDQ calculations. This correction is on the order of 0.003 to 0.004 dk. Removing this adjustment from the EPRI-CPM results would provide excellent agreement between the original EPRI-CPM benchmarking and the PPSL CPM-2 calculations presented in Section 2.2. The results reported by EPRI for the ESADA criticals include correction factors to account for the presence of the spacers and the self-shielding of the plutonium grains. The spacer correction used was -0.4% ~k for all cases. This adjustment was also made to the CPM-2 calculations presented in Section 2.2. The shielding correction applied in the EPRI-CPM results varied between -0.05% bk to -0.45% bk. Specific details concerning the exact correction for each experiment was not available. The magnitude of this correction is consistent with the difference between the EPRI-CPM and the PP&L CPM-2 calculations.

Isotopic comparisons were also performed using both Yankee (Reference 13) and Saxton (Reference 14) isotopic data. The results from the Yankee comparisons are shown in Figures 2.4.4 through 2.4.6. All calculations show good agreement between the calculated ratios and measured data. The Pu-241/Pu-242 ratio is slightly overpredicted (approximately 3%) at end of life (30 GWD/MTU). The results from the Saxton comparisons are given in Table 2.4.3.

39-

TABLE 2 4.1 EPRI~M RESULTS FROM THE TRX CRITICAL BENCHMARKING Hexagonal Lattice Pellet Experiment Pitch Diameter B (experyental) EPRI~M Identification (in) (in) (m ) K-effective TRX1 0.868 0.601 28.4 0.997 TRX2 0.929 0.601 30.2 0.999 TRX3 0.989 0.601 29.1 0.998 TRX4 0.613 0.388 25.3 0.998 TRX5 0.650 0.388 25.2 0.997 TRX6 0.613 0.383 32.6 1.000 TRX7 0.650 0.383 35.5 1.000 TRX8 0.711 0.383 34.2 1.000 Average K-effective = 0.999 + 0.001 Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

40-

TABLE 2 4 2 EPRI~M RESULTS FROM ESADA CRITICAL BENCEiMARKING Lattice Boron 2

Experiment Pitch Concentration (experimental) EPRI-CPM

' 'IH B (m ) K-effective ESADA1,2 8% Pu-240 0.69 0 69.1 0.999 ESADA3 8% Pu-240 0.75 90.0 1.000 ESADA4,5 8% Pu-240 0.9758 105.9 1.008 ESADA6 8% Pu-240 1.0607 98.4 1.010 ESADA7 8% Pu-240 1.380 50.3 0.997 ESADA8 8% Pu-240 0.69 261 62.6 1.004 ESADA9 8% Pu-240 0.9758 261 83.7 1.002 ESADA10 8% Pu-240 0.69 526 58.3 1.002 ESADA11 8% Pu-240 0.9758 526 63.1 0.999 ESADA12 24% Pu-240 0.9758 79.5 '.004 ESADA13 24% Pu-240 1.0607 73.3 1.002 Average K-effective = 1.002 + 0;004 Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

41

TABLE 2.4 3 EPRI ISOTOPIC COMPARISONS TO SAXTON DATA Measured Nuclide Percent Concentration Measurement Difference*

Nuclide (Atom 4) Uncertaint (%) (*)

U-234 0.00465 28.7 15.9 U-235 0.574 0.9 -0.3 U-236 0.0355 5.6 2.8 U-238 99.386 Pu-238 0.109 2.2 -11.4 Pu-239 73.77 -0.3 PU-240 19.25 0.2 1.6 Pu-241 6.29 0.3 0.4 Pu-242 0.579 0.9 -16.0

• Percent Difference = ca lc-mea s x 100 meas Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

FIGURE 2.4.1 FISSION RATE COMPARISON FOR AN Sx8 BWR ASSEMBLY 0

OF THE PLUTONIUM ISLAND TYPE T=245 C WIDE GAP 0 UO, RODS

+1.9

+0.7 +0.1

-0.5 ~+~.0 +2.0 MO, RODS I

O

+0.6 l +2.9 (+0.1 +0.9 z

I

-0.6 -1.6 +0.6 I J

-0.5 -0,5 i+0.8 +0.6i +0.5 -1.1

-2.8 -0.6 -1.3 -1.0 ~2o7 NARROW GAP CPM ex This figure shows P x 100 for all measured rod positions.

exp Experimental Uncertainty (la) in MO rods: + 1.4%

Experimental Uncertainty (lc) in UO rods: + 0.7%

2 Fission Rate in MO rods relative to UO rods: + 1.6%

Calculated keff ff 1.001 Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

43

FIGURE 2'.4.2 FISSION RATE COMPARISON FOR A 15xl5 PWR MIXED OXIDE ASSEMBLY 0

WITH WATER HOLES AND ABSORBER RODS T~245 C" CENTRAL WATER HOLE ABSORBER ROD

+1.0 MO, RODS I

+3.1 -0.1 +1.2 303 -1.3 -0.6

+2.2 -0.4 -0.8 -0.4

+1.1 -2.2 +1.4 -3.3

+3.7 -2.0 +1.7 +0.3 -3 9 -1.6

+0.7 +2.3 -0.7 +0.2 ex This figure shows CPM P

x 100 for all measured rod positions.

exp Experimental Uncertainty (lc): + 1.4%*

Calculated,k eff ff = 0.999

• Not inc3,uding geometric uncertainties.

Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

44

FIGURE 2.4.3 FISSION RATE COMPARISON FOR A 14xl4 PWR MIXED OXIDE ASSEMBLY 0

SURROUNDED BY UO ASSEMBLIES T=240 C 2

CENTER

+1 8 HIGH ENRICHED ~

MO, RODS

~

-0.3 CL

+2.1 WATER HOLES O

+1.0 +0.9 X

-1.7

+1.2 +0.7I -1.6 -0.6

+0.9 LOW ENRICHED <<3 7 MO, RODS

+O.B ENR UO, RODS -0.8

+1.4 -0.4

-O.B P - P This figure shows CPM exp x 100 for all measured rod positions.

exp The fission rate was normalized separately for each type of assembly.

The average fission rate in each MO assembly'elative to the rate in the UO assemblies predicted by DIXP was 1.9% lower than the measured ratio.

Experimental uncertainty (la) for each type of fuel separately: + 0.8%

Experimental uncertainty (la) for the average fission rate in MO rods relative to UO rods: + 1.4%

Calculated keffff = 0.997 Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

45-

FIGURE 2.4.4 EPRI-CPM COMPARISON TO YANKEE PU-239/PU-240 ISOTOPIC RATIOS 9.0

~ ~

~ w

~~

7.0 O

C ID Ps lL CO m

cv .0 n.

~ l 4.0

~

~

~

t ~

3.0 0.0 5.0 10.0 15.0 20.0 .0 5

F. P. vol. wgt. number density x 10 10 20 30 MWd/kgU o Measured Data EPRI-CPM Results Source: M'. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

FIGURE 2.4.5 EPRI-CPM COMPARISON TO YANKEE PU-240/PU-241 ISOTOPIC RATIOS 8.0

~ ~

~ t 0

cv 4.0 tL

~ ~

~~ ~

0.0 10.0 15.0 20.0 25.0 30.0 F. P. vol. wgt number density ~ 10

0. 10 20 30 MWd/kg U

~ Measured Data EPRI-CPM Results Source: M. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

FIGURE 2..4.6 EPRI-CPM COMPARISON TO YANKEE PU-241/PU-242 ISOTOPIC RATIOS 10.0

~ Oy

~y

~ ~

9.0 8.0 O

4 7.0

~0 c4 IL cv 6.0 ~ ~

~ ~

0 5.0 4.0 0.0 5.0 10.0 15.0 20.0 30.0 F. P. vol. wgt.number density x 105 10 20 30 MWd/kg U

~ Measured Data EPRI-CPM Results Source: E. Edenius, "EPRI-CPM Benchmarking," Part I, Chapter 5 of EPRI CCM-3, November, 1975.

48-

3.0 CORE SIMULATION METHODS The three-dimensional nodal simulation code used by PPGL is the SIMULATE-E (Reference 15) computer program distributed by EPRI. This code has been used to provide the steady state operations support at PPGL and will be utilized for reload core design and licensing, analyses. The code is used to calculate core reactivity, power and flow distributions, thermal limits, and Traversing In-core Probe (TIP) response. A'ull description of the SIMULATE-E methodology is contained in Reference 15. A brief summary is presented in Section 3.1.

SIMULATE-E has been benchmarked by PPsL against extensive reactor operating data. The Susquehanna SES benchmarking includes comparisons to hot and cold critical data, TIP measurements, and core monitoring system calculations.

These comparisons are presented in Section 3.2. Comparisons have also been made to the Quad Cities Unit 1 hot and cold critical data, TIP measurements, and end of Cycles 1 and 2 gamma scan data. The Quad Cities comparisons are presented in Section 3.3. Comparisons were also made to Peach Bottom Unit 2 Cycles 1 and 2 data. The Peach Bottom Unit 2 reactor was modeled primarily to prepare input to the transient analysis of the three turbine trip tests.

Section 3.4 presents comparisons to several TIP sets through both cycles and to the core monitoring system power distributions taken prior to each turbine trip test.

3.1 Descri tion of SIMULATE-E The SIMULATE-E computer program was written to perform three-dimensional analyses of light water reactors. The code combines both neutronics and thermal hydraulics calculations. The neutron balance equation is solved using response matrix techniques developed by Ancona (Reference 16) . The response matrix parameters are determined using the PRESTO option (Reference 17). The thermal hydraulics module contains the EPRI void correlation (Reference 18) and the FIBWR (Reference 19) code to determine axial voiding and flow distribution. The neutronics and thermal hydraulics are solved iteratively until a consistent solution is achieved.

The reactor core is modeled as an array of cubic nodes each containing a homogenized portion of a fuel assembly. For the Susquehanna SES BWRs, each fuel assembly is modeled using 25 axial nodes, thus resulting in six inch nodes describing the 150 inch active fuel region. Albedo boundary conditions are used to account for the reflector zones, thus eliminating the need to explicitly model the reflector.

The neutronics calculation requires the solution of the neutron balance equation for each node. This balance equation is first recast in terms of response matrix parameters which describe how a neutron interacts with adjacent nodes. Several options exist in SIMULATE-E which can be used for determination of the response matrix parameters. The option used by PPGL is the Modified Coarse Mesh Diffusion Theory (MCMDT) also referred to as the PRESTO option (Reference 17). This option calculates the various transmission probabilities using node average fluxes. The MCMDT option calculates the node center and node surface fluxes using Fick's Law. The node average flux is then determined as a weighted average of the surface and center fluxes. The weighting factors were developed through model normalization. Once the node average fluxes are determined, the various transmission probabilities can be evaluated and the neutron balance equation is solved.

Nodal cross section data are input to SIMULATE-E in two groups for each different lattice type. If axial zoning of fuel is present (either due to enrichment or gadolinia content), separate lattice types are assigned.

Cross section dependencies include:

fuel exposure void history (i.e., exposure-weighted relative moderator density) relative moderator density (hot only) control rod presence fuel temperature (hot only) control rod history xenon concentration moderator temperature (cold only)

II The effect of each dependency is calculated utilizing CPM-2. The final cross section data tables are prepared for SIMULATE-E using NORGE-B2 (Reference 20).

The radial, top, and 'bottom reflector regions are not modeled explicitly.

Instead, these regions are taken into account by use of albedo boundary conditions. Radial albedos are calculated using the ABLE (Reference 21) program developed by Science Applications International for EPRI. The top and bottom albedos were determined based on comparison to plant data during model normalization. Different albedo boundary conditions are used for cold and hot conditions.

Several of the input data parameters used by SIMULATE-E require adjustment to match plant operating data. This normalization process was performed using Susquehanna SES Unit 1 Cycles 1 and 2 data. All parameters changed in this fashion were held constant for all other calculations including the Quad Cities and Peach Bottom calculations.

The thermal hydraulics calculations use the FIBWR methodology (Reference 19) developed by Yankee Atomic, Electric Company. This calculation determines total core pressure drop and core bypass flow. The pressure drop calculation determines the frictional pressure drop, local (i.e., form) losses, acceleration (i.e., momentum change) pressure drop, and elevation head. The core bypass flow calculation allows for modeling the flow paths shown in Figure 3.1.1. FIBWR as a stand-alone code has been bpnchmarked by Yankee Atomic Electric Company against. data from Vermont Yankee and the Frigg Loop tests (see Reference 22).

51

During installation at PPGL minor code modifications have been made to adapt the SIMULATE-E code to PPGL computer operational requirements.

In addition, code changes were made by PPGL which include:

Critical Power Ratio (CPR) evaluations utilizing the Advanced Nuclear Corporation (formerly Exxon Nuclear Company) XN-3 critical heat 'uels flux correlation (Reference 23)

Linear Heat Generation Rate (LHGR) and Average Planar Linear Heat Generation Rate (APLHGR) thermal limits evaluations calculation of Axial Exposure Ratio error corrections provided by EPRI These changes, with the exception of error corrections, have not resulted in any change to the neutronics or thermal hydraulics calculations'.

- 52

FIGURE 3.1.1 BWR FUEL ASSEMBLY BYPASS FLOW PATHS TOP OF CORE 2CHI Spacer height ~ HFSGw2HET HFSG 1

Note: Bottrxrr entry peripheral fue'I supports are welded into the core support plate.

For these bundles, path numbers 1,2,5 and 7 do not exist.

HFSGn ZUHB Channel 8

Lower tie plate Spring plug g Core 6 support 2GEO la 2

Botton of core In-core Fuel Support, guide tube Shroud Control rod guide tube Core length < 2CHI +

fuel length + 2GEO 1. Control rod guide tube fuel support

2. Control rod guide tube-core support plate Fuel length ~ 2UHA + 2HET + 2UHB 3. Core support plate in-core guide tube

4. Core support plate-shroud

5. Control rod guide tube--drive housing 7 6. Fuel support--lower tie plate

7. Control rod drive cooling wate~

drive housing B Channel lower tie plate

9. Lower tie plate holes

10. Sp~ing plug-core support Source: B. J. Gitnick, "FIBWRr A Steady-State Core Flow Distribution Code for Boiling Water Reactors; Computer Code User's Manual,"

EPRI NP 1924 CCMr Julyr 1981 53

3.2 Sus uehanna SES Units 1 and 2 Benchmark Comparisons of SIMULATE-E calculations to observed data from the Susquehanna SES operating reactors provide a direct means of qualifying the accuracy of SIMULATE-E. Two directly measurable sets of parameters against which comparisons can be made for Susquehanna SES consist of the core critical K-effective state point data (hot and cold) and the Traversing In-core Probe (TIP) neutron flux measurements. This type of benchmarking validates the overall BWR analysis process from lattice physics to three-dimensional simulation.

For core critical K-effective comparisons, the measured steady state core operating parameter's listed in Table 3.2.1 provide the necessary input for a SIMULATE-E calculation. This data is also used to model the accumulation of core history through multiple depletions (i.e., core follow) . These calculations assume constant core conditions during a short time period, usually less than one week.

The coze critical calculations at steady state conditions are used to qualify SIMULATE-E's capability to predict core reactivity throughout a cycle. Design analyses, such as cycle length, shutdown margin, hot excess reactivity, rod withdrawal error, misloaded bundle, standby liquid control system worth, and control rod drop, require the prediction of the core reactivity throughout a cycle. Because the SIMULATE-E hot and cold models differ, separate hot and cold critical K-effective comparisons are performed to determine the individual uncertainties for the above analyses. For the hot critical core K-effective comparisons, reactivity calculations rely on the statepoint parameters listed in Table 3.2.1. The cold critical core reactivity benchmark involves reactiv'ity calculations for all cold xenon-free criticals for the Susquehanna SES cores. The hot and cold K-effective comparisons are used to establish the target critical core K-effective and to assess the uncertainty in reactivity predictions.

TIP comparisons test the ability of SIMULATE-E to calculate the neutron flux in a local 'region between four fuel assemblies. The TIP measurements used in the comparisons are six-inch collapsed detector signals. These are

- 54

\

synthesized from one-inch axial data that are averaged by the core monitoring system through a trapezoidal averaging technique. For Cycle 2 and beyond of both units, the core monitoring system, POWERPLEX (Reference 24), also corrects the TIP measurements for any axial shift in the measurements. A Gaussian smoothing procedure compares measured neutron flux dips to the expected dip locations, based on fixed LPRM and spacer locations, and corrects the axial alignment of the one-inch data.

The Susquehanna SES core operating histories from Unit 1 Cycles 1, 2, and part of Cycle 3 and Unit 2 Cycle 1 and part of Cycle 2 are contained in the benchmark data base. The two units share identical core geometry and rated core conditions. The Cycle 1 operating cores of both units contain the same General Electric Sx8 fuel design and core loading patt'em. Zn addition, both Cycle 1 operating strategies have extended cycle operation via bottom burn spectral shift and coastdown. Cycle 2 of Unit 1 operated with a small 192 Exxon Sx8 bundle reload core that experienced a cycle length less than half of Cycle 1 and less than any planned future cycle. Cycle 3 of Unit 1 was loaded with 296 Exxon Sx8 bundles and Cycle 2 of Unit 2 was loaded with 324 Exxon 9x9 bundles. The anticipated equilibrium batch size for the planned eighteen month cycles is 240 Exxon 9x9 bundles. At the time this report was prepared, Unit 1 was in its third cycle of operation and Unit 2 was in its second cycle of operation. Therefore, the benchmark data base only includes the first third of Unit 1 Cycle 3 operation and the early portion of Unit 2 Cycle 2 operation. Table 3.2.2 summarizes the total Susquehanna SES benchmarking data base included in this report.

For all Susquehanna SES hot comparisons, unly steady state data have been used. This requires core conditions to remain constant over a period of time to allow the xenon concentration to equilibriate. This requires no rod pulls or significant change in core thermal power, flow, or feedwater temperature within approximately three days prior to the data point. For the cold comparisons, sufficient time at zero power is required to allow for xenon decay. 1n addition, a reactivity adjustment is made for the reactor period.

- 55

3.2.1 Hot Critical Core Reactivit Com arisons The results of the SIMULATE-E core follow calculations that are based on measured core operating data for the Suscpxehanna SES cores form the hot critical core reactivity data base. These calculations result i:n a total of 257 steady state core K-effective comparisons for various core operating conditions. Table 3.2.3 shows a complete list of the steady state core data and its corresponding calculated hot critical core K-effective tabulated by unit and cycle.

Figures 3.2.1 through 3.2.6 show plots of hot critical core K-effective versus core average exposure, core thermal power, total core flow, core inlet subcooling, dome pressure, and critical control rod density, respectively.

These figures provide information on the dependencies and biases inherent in SIMULATE-E. It is apparent that the critical K-effective varies with core average exposure. The K-effective from Cycle 1 of both units exhibits a bowl-shaped trend up to 7000 MWD/MTU cycle exposure, at which point gadolinia content is substantially depleted. The available data from Cycle 3 also exhibits the same trend. For this cycle, the initial core average gadolinia content was almost the same as for the Cycle 1 cores. Unlike Cycles 1 and 3, the K-effective from Cycle 2 of Unit 1 exhibits a very flat trend throughout the entire cycle. Cycle 2 of Unit 1 contains approximately one-fourth the initial gadolinia content of either Cycle 1 or Cycle 3 of Unit 1. These trends suggest a dependency on gadolinia loading. After the gadolinia has essentially burned out in Cycle 1, the critical core K-effective increases slightly with exposure. Therefore, the hot critical core K-effective exhibits a linear dependence on exposure coupled with a bowl-shaped trend due to gadolinia depletion.

PPGL has developed a method which correlates the hot critical core K-effective data to the core average exposure and gadolinia content. Using this correlation, target critical core K-effective curves are generated for each cycle. Figure 3.2.7 shows the comparison of the target critical core K-effective curves and the SIMULATE-E calculated core K-effective for each unit and cycle. Table 3.2.4 shows the mean difference and standard deviation between the target and SIMULATE-E calculated critical core K-effective for

each unit and cycle. The overall results show very good agreement with the target K-effective. For Unit 2 Cycle 2 only three data points are included in the data base. These data yield a mean difference of 0.00186 ~K from the target which is larger than two times the standard deviation of the data base (i.e., 2a is 0.00122 4K). It is anticipated that the Unit 2 Cycle 2 SIMULATE-E calculated core K-effective will follow the target but will be offset by a constant bias. More recent core follow calculations, which are not included in this report, support this expected trend. The offset is likely due to differences between Sx8 and 9x9 fuel designs. PPaL continues to perform hot critical core K-effective calculations as part of the routine core follow analyses, and the results are used to further improve the accuracy of the target critical core K-effective. On a whole, the use of the correlation provides a good assessment of critical core reactivity and can be used for design and analysis of future .cycles.

An important consideration in evaluating reactivity results is the measurement uncertainty in the core operating conditions. Because measured core operating data (i.e., the parameters listed in Table 3.2.1) are entered into SIMULATE-E, the calculated core reactivity is affected by any errors in the measured inputs. The changes in core reactivity from measurement uncertainty primarily depend on two core modeling phenomena, the void and Doppler coefficients of reactivity. As these coefficients change with core life and designs, the sensitivity of SIMULATE-E to measurement uncertainties changes. Table 3.2.5 shows measurement uncertainties based on Reference 25 and their effects on core reactivity for Susquehanna SES Unit 2 Cycle 2. The total K-effective sensitivity due to the measurement uncertainties is 0.00151 bK. SIMULATE-E calculations of hot critical core K-effective for the 257 data points analyzed by PPGL result in a standard deviation which is less than this sensitivity.

3.2.2 Cold Critical Core Reactivit Com arisons The accuracy of the SIMULATE-E calculation of core shutdown margin and control rod worths at cold conditions depends on its ability to predict cold core reactivity for different core designs, core average exposures, and control rod configurations. Shutdown margin calculations also rely on the accuracy of the modified coarse mesh diffusion theory prediction of large neutron flux 57

gradients characteristic of one-rod-out configurations. Local Critical tests exhibit flux gradients simila'r to shutdown margin calculations. Therefore, the qualification of the SIMULATE-E code requires benchmarking to both local and in-sequence cold xenon-free criticals. The Susquehanna SES benchmarking data base contains three local and 36 in-sequence criticals as shown in Table 3.2.2. In addition to the Susquehanna SES cold critical benchmarking calculations, comparisons to the Quad Cities Unit 1 Cycle 1 local and in-sequence criticals were performed to further support the validation.

Section 3.3.2 describes the Quad Cities Unit 1 Cycle 1 benchmarking analyses.

Table 3.2.6 contains results of the Susquehanna SES cold xenon-free criticals.

As shown, the criticals were performed at temperatures between 100 and 212 0 F and at various core average exposures. The core K-effective in Table 3.2.6 includes a reactor period correction which is typically less than 0.001.

Figure 3.2.1 shows these results together with those of the hot benchmark.

The calculated cold critical core K-effectives consistently follow the hot, critical core K-effective with a constant bias throughout exposure. This trend indicates that the cold methods and models also depend on core avexage exposure and gadolinia content. A bias between cold and hot core K-effectives has been reported by others and has been investigated in Reference 26.

A method that accounts for the core average exposure and gadolinia content dependencies results in an accurate assessment of the cold xenon-free critical core K-effective and its uncertainty. Figure 3.2.1 indicates that a bias exists between the hot and cold SIMULATE-E calculated core critical K-effectives. Reference 26 supports this observation. This bias is constant and is not exposure or gadolinia dependent. Therefore, the target cold critical core K-effective is determined by adding a bias to the hot critical core K-effective.

Table 3.2.7 shows the results for the Susquehanna SES in-sequence and local cold criticals. The mean difference between the SIMULATE-E hot and cold calculated core K-effective for all 39 criticals is 0.00671 and the standard deviation is 0.00111. The mean difference between the hot target core K-effective curve and the SIMULATE-E cold calculated core K-effective for all 39 criticals is 0.00659 and the standard deviation is 0.00137. These two standard deviations are small and are typical of the calculated core K-effective variation for criticals at a given exposure. For example, the standard deviation of the Unit 2 Cycle 1 zero exposure calculated cold critical K-effectives is 0.00099. Table 3.2.7 also shows that the cold to hot K-effective bias for the local critical tests is not significantly different than the bias for the in-sequence criticals.

As in the hot reactivity benchmark, PPSL will continue to perform cold critical core K-effective comparisons to be used for periodic updating of the target cold critical core K-effective. The use of this target cold critical core K-effective provides a good assessment of critical core reactivity and can be used for design and analysis of future cycles.

3.2. 3 Traversing In-core Probe Data Comparisons Comparisons to measured TIP data provide an assessment of how well SIMULATE-E calculates the core power distribution. The TIP detectors are located in the water gap corner opposite the control rod between four fuel assemblies as shown in Figure 3.2.8. SIMULATE-E calculates a TIP response for each six-inch axial segment at each radial TIP location by power weighting input detector response functions as follows:

M ER = Q R. P.

where M = the number of bundles around a TIP detector (for all plants modeled, M=4),

R, = the detector response function, j

P. = the SIMULATE-E calculated nodal power.

j The detector response, R., is a functional relationship which can be expanded j'o:

j UNC CT CT EV U where F = the base componentof the detector response for an uncontrolled node, G = the fraction of the node which is controlled, G = 0 node is uncontrolled, G = l node is fully controlled, F = the correction to the base component for a fully controlled node, F F = the correction to the base response to account for moderator density.

F , F and F are functions of exposure and void history (i.e., exposure-weighted relative moderator density). F U is a function of the relative moderator .density and void history. The detector model in SIMULATE-E assumes that the detector response from each assembly is not affected by the other three.

The detector response functions are generated using calculated data from CPM-2. In CPM-2, a small amount of U-235 is placed in the water gap corner opposite the control blade. The local fission rate is calculated in this region for different conditions of exposure, void history, control state and relative moderator level. This data is formulated into a polynomial fit determined for each separate lattice type.

Both nodal and radial (i.e., axially integrated) TIP comparisons have been made to the Susquehanna SES TIP data. For the nodal comparisons, the six-inch averaged measured data is compared to the calculated nodal TIP response. This provides an assessment of the accuracy of the nodal power distribution which affects calculated margin to operating limits such as the Linear Heat Generation Rate (LHGR) limit. For the radial comparisons, the average of all TIP measurements at a radial location is compared to the average of the calculated values at that radial location. This provides an assessment of the accuracy of the radial (i.e., bundle average) power distribution which affects calculated margin to operating limits such as Critical Power Ratio (CPR).

Prior to making the comparisons, the calculated data is normalized to the measured data. Each of the calculated nodal detector responses is multiplied by a normalization factor. The factor is calculated as:

TNF = T/ER where T = the average of all measured TIP responses in a given TIP set, ER = the average of all calculated TIP responses in a given TIP set.

A TIP set is defined as a complete set of TIP measurements from the entire core. For Susquehanna SES a TIP set consists of 24 measurements at each of the 43 radial locations in the core for a total of 1032 measurements. In each of the comparisons presented in this section, all radial TIP locations and all axial elevations have been included.

For the nodal comparisons, the difference between calculated and measured data is determined as:

e k,m

=ER k,m

-Tk,m where ER = the calculated TIP response for axial elevation, k, and radial location, m, Tk k,m

= the measured TIP response for axial elevation, k, and radial location, m.

The Root Mean Square (RMS) of the differences for each radial TIP location is calculated as:

K 2

RNS Z'k, m K-1 ~

where K = the number of axial TIP measurements (i.e., 24) at a radial TIP location.

The relative RMS of the differences for, each TIP set is calculated as:

g RMS 100 RMS nod where M = the number of radial TIP locations (i.e., 43 for Susquehanna SES) .

For the radial comparisons, a similar RMS is calculated. First, the calculated and measured individual TIP readings are axially averaged as follows:

K ER m

QER /K K

T m QT /K where ER m

= the average of the calculated TIP responses at a given radial location, m, T

m

= the average of the measured TIP responses at a given radial location, m.

The difference between the calculated and measured radial TIP response in percent is:

(ER - T )

m in e x 100 T

The RMS of the differences for all TIPs for a given TIP set is calculated as:

Z'.'MS radi al M-1 An estimate of the TIP measurement uncertainty can be determined by calculating the nodal and radial TIP response asymmetries. During A-sequences and all-rods-out core configurations, the control rod pattern is eighth-core 62-

mirror symmetric. In addition, the fuel loading patterns for all of the Susquehanna SES cycles have been designed to be eighth-core symmetric. Under these conditions, a line of symmetry exists along the TIP locations as shown in Figure 3.2.8. For the TIPs not located directly on this symmetry line, there will be a symmetric TIP having nearly the same neutron flux conditions.

These symmetric TIP pairs should give the same measurements except for exposure asymmetries which can add approximately 1% nodal asymmetry.

To calculate the nodal asymmetry, the nodal difference for each symmetric TIP pair, n, is calculated as:

e = T T k, n k,ml k,m2 where k,ml and Tkk,m2 the six-inch detector measurements at axial location, k, Tk =

and symmetric TIP locations m1 and m2.

The RMS of the nodal differences in percent is:

100 ASY n

x 1 K-1 (Tm1 + Tm22) where T

ml and T m2

= the average measured TIP response for symmetric TIP locations ml and m2.

The average nodal asymmetry is calculated as the arithmetic average of the symmetric pair asymmetries:

Q ASY nod N where N = the number of symmetric TIP pairs (i.e., 19 for Susquehanna SES) .

63

The radial TIP response asymmetry is calculated using the relative difference between the axially-averaged .TIP measurements for each symmetric pair, n.

This difference is calculated as:

T - T ml m2 x

D 100 n +Tm2 )

(T ml The .mean absolute asymmetry is calculated as:

The results of the TIP response comparisons separated by unit and cycle are reported in Tables 3.2.8 through 3.2.11. These include comparisons to all available steady state TIP sets. No TIP data have been excluded from the comparison. An overall summary of the results from the comparisons is given in Table 3.2.12. A summary of the asymmetries averaged by unit and cycle is

,given in Table 3.2.13 which shows the nodal and radial asymmetries for Unit 2 Cycle 1 are approximately 2% worse than the asymmetries for Unit 1 Cycle 1.

This larger TIP response asymmetry indicates larger measurement uncertainty for Unit 2 Cycle 1 and also explains why the TIP response comparisons for Unit 2 Cycle 1 tend to be worse than for Unit 1 Cycle 1 even though the core loadings were identical.

The nodal results from the TIP response comparisons are also displayed versus core average exposure in Figure 3.2.9a. No definite trends with exposure are evident. When the data is displayed versus fraction of cycle length as in Figure 3.2.9b, a trend is apparent. The results in the middle of the cycle tend to be worse than at the beginning of the cycle or end of full power. For the end-of-cycle power coastdown, the relative RMS from the TIP response comparisons increases. This is expected because core operating parameter measurement uncertainties increase for lower power conditions. In addition, the SIMULATE-E model is developed primarily based on full power operating conditions. When the cross section tables are developed, dependencies are included to correct for Doppler and instantaneous relative moderator density.

- 64

The uncertainties in these corrections increase as conditions deviate from full power. Therefore, the corresponding RMS from the TIP response comparisons will also increase. The results even for the end of cycle power coastdown comparisons are still good. The Unit, 1 end of Cycle 1 RMS was just over 6% at approximately 81% of rated power, and the Unit 2 end of Cycle 1 RMS was less than 8% at approximately 71% of rated power. Several of the comparisons for the middle and end of Unit 2 Cycle 1 exhibit approximately 8%

RMS which is larger than expected. During these TIP measurements, there were suspected problems with some of the TIP machines; this is supported by the larger nodal asymmetries experienced for these TIP sets. Overall, the results from the nodal TIP response comparisons are quite good with an average RMS of 5.75%.

Graphical results of the TIP response comparisons are included for each unit and cycle. Due to the large number of TIP sets and TIP locations within a TIP set, figures of TIP response comparisons are presented for beginning, middle, and end of cycle. For each exposure point, core average axial, radial, and four individual TIP response comparisons are included. The individual TIP response comparisons in the figures were selected along a line from the core periphery to the center as shown in Figure 3.2.8. The same four TIP locations are always shown. These comparisons are shown in Figures 3.2.10 through Figure 3.2.42.

3.2.4 Core Monitoring System Com arisons The ability of SIMULATE-E to accurately calculate power distributions is demonstrated in Sections 3.2.3, 3.3.3, and 3.3.4. The purpose of this section is to provide a comparison of the SIMULATE-E calculated power and flow distributions to those of the on-line Core Monitoring Systems (CMS). Four axial power comparisons and three bundle flow comparisons are presented. The data were taken from one point in the Susquehanna SES Unit 1 Cycles 1, 2, and 3, and 'Unit 2 Cycle 2. This selection provides a good mix regarding thermal hydraulic and neutronic differences in design. Although these comparisons do not represent a validation of the SIMULATE-E models, they demonstrate consistency with the systems used to monitor the core. The CMS for Cycle 1 of both units is the General Electric Company Process Computer P1 program; for

- 65

the reload cycles of both units, the CMS is the ANF (formerly Exxon Nuclear Company) POWERPLEX CMS.

Figures 3.2.43 through 3.2.46 show the core average axial power. distribution comparisons. These figures show good agreement, and indicate consistency between the independent core analysis methods for axial power distribution determination.

Figures 3.2.47 through 3.2.49 shOw the core flow distribution comparisons.

These figures show excellent agreement between the SIMULATE<<E and CMS calculated bundle flows for the three comparisons. The effects of peripheral and central orificing for the core combinations of GE SxS and Exxon SxS, GE SxS and Exxon 9x9, and all GE SxS are accurately modeled.

- 66

TABLE 3 2 1 MEASURED CORE OPERATING PARAMETERS FOR SIMULATE-E CORE REACTIVITY CALCtKATIONS Hot Core Operating Condition Core Thermal Power Total Core Flow Core Inlet Subcooling Core Pressure Control Rod Pattern Cold Core Condition Core Moderator Temperature Reactor Period Control Rod Pattern 67

TABLE 3.2 2

SUMMARY

OF THE SUSQUEHANNA SES BENCEBQLRKING DATA BASE Number of Number of Number of Cold Unit a cle TIP Co arisons Core Criticals Core Criticals U1C1 31 87 U1C2 47 U1C3 23 10 U2C1 32 97 13*

U2C2 None All 82 257 39

• Includes three local criticals.

68-

TABLE 3.2.3 BUS EHANMA SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT=1 CYCLE-"1-CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- DONE CONTROL ROO CALCULATED EXPOSURE EXPOSURE POWER POWER FLOW COOLING PRESSURE DENSITY CORE CASE ( GWD/MTU ) lGWD/WTU) (WTH) l%) (%) I BTU/LBH) (PSIA) l%) K-EFFECTIVE 1 0.221 0.221 1432 43 54 23.8 974 20.4 0.99184 2 0.836 0.836 3250 99 98 23.7 1001 12.6 0. 99142 3 1.490 1.490 3280 100 100 23.6 1005 13.9 0.98987 4 1.596 1.596 3278 100 88 23.6 1002 13.6 0.98665 5 1.736 1.736 3291 100 97 24.3 1001 14,0 0.98919 6 1.758 1.758 3296 100 98 24.2 1001 14.1 0.98886 7 l. 799

1. 908 1.799 3291 100 99 23.8 1001 14.1 0.98938 8 ~

1.908 3293 100 98 24.0 1000 14.1 0.98960 9 2.070 2.070 3293 100 97 24.2 994 14.1 0.98884 2.706 2.706 ll 10 12 2.906 2.975 2.906 2.975 3281 32S9 3291 100 100 100 94 98 97 25.0 24.2 24.4 1000 999 999 14.8 15.0 15.0 0.98937 0.98990 0.98988 13 3.116 3.116 3291 100 96 24.7 999 15,0 0.99009 14 3.367 3.367 3292 100 98 24.2 1004 15. 9 0.98971 15 3.517 3.517 3289 100 98 24.2 1002 15. 9 0.99020 16 3.663 3.663 3292 100 96 24.5 1002 15. 9 0.99042 17 3.776 3.776 3290 100 96 24.6 1001 15.9 0.99058 18 3.836 3.836 3293 100 95 24.S 1003 15.9 0.99061 19 3.918 3.918 3298 100 98 24.0 1000 16.0 0.99080 20 4.036 4.036 3290 100 97 24.3 1003 16.0 0.99100 21 4.193 4.193 3290 100 96 24.5 1002 16.0 0.99116 22 4.31S 4.318 3296 100 98 24.2 1003 16.1 0.99138 23 4.506 4.506 3288 100 96 24.5 1003 16.1 0.99163 24 4.517 4.517 3289 100 97 24.4 1004 16.1 0.99176 25 5.061 5.061 3290 100 99 23.8 1005 17.6 0.99254 26 5.070 5.070 3288 100 99 23.9 1005 17.6 0.99242 27 5.347 5.347 3281 100 97 24.3 1002 18.0 0.99219 28 5.410 5.410 3294 100 98 24.0 1002 17.9 0.99267 29 5.463 5.463 3291 100 99 23.8 1002 17.9 0.99294 30 5;580 5.580 3294 100 99 23.7 1002 17,8 0.99350 31 5.614 5.614 3295 100 99 23.9 1002 17.8 0.99358 32 5.650 5.650 3287 100 99 23.8 1002 17.8 0.99367 33 5.855 5.855 3293 100 99 23 ' 1001 17.0 0.99362 34 5.918 5.918 3289 100 98 24.1 1001 16.7 0.99362 35 6.087 6.087 3286 100 96 24.3 1000 16.4 0.99430 36 6.241 6. 241 3288 100 98 23.9 1001 16.4 0.99437 37 6.436 6.436 3265 99 96 24 ' 999 16.3 0.99454 38 6.563 6.563 3286 100 99 23.8 999 16.3 0.99463 39 6.716 6.716 3283 100 98 24.1 999 15.0 0.99460 40 6.723 .6.723 3290 100 98 24.0 999 15.0 0.99460

TABLE 3.2.3 ( CONTINUE D )

SUSQJEHAWA SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT"-1 CYCLE=l CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB DONE CO)(TROL ROD CALCULATED EXPOSURE EXPOSURE POWER POHER FLOW COOLING PRESSURE DENSITY. CORE CASE (GHD/HTU) (GHD/HTU) IWfH) (%) (/) ( BTU/LBH) (PSIA) (%) K-EFFECTIVE 41 6.893 6.893 3282 100 96 24.4 998 14.6 0.99472 4R 7.000 7.000 3285 100 99 23.7 998 14.6 0.99516 43 7.155 7. 155 3291 100 97 24.5 1008 14.5 0. 99471 7.235 7. 235 3276 99 94 25.1 1007 14.7 0.99396 45 7.365 7.365 3285 100 96 24.5 1006 14.3 0.99490 46 7.638 7.638 3R73 99 95 24.7 993 13.0 0. 99410 47 7.670 7.670 3288 100 96 24 ' 993 13.0 0. 99411 4S 7.'840 7.84io 3284 100 97 24.3 992 12.6 0 ~ 99494 49 7.899 7.899 3288 100 96 24.7 992 12.4 0.99490 50 8.013 8.013 3289 100 94 24.9 992 12.0 0.99502 51 52 8.164 8.308 S.164 8.308 3301 3290 100 100 94 97 24.8 R4.2 987 991 ll.7 11.4 0.99484 0.99537 53 8. 341 8.341 3293 100 96 24.5 991 11.3 0.99536 54 8.481 8.481 3288 100 96 990 10.4 0.99544 55 8.51S 8.518 3286 100 98 24.0 990 10.4 0.99550 56 8.587 8.587 3286 100 99 23.5 990 10.4 0. 99561 57 8.60R 8.60R 3284 100 99 23.5 990 10.4 0.99553 58 8.658 8.658 3287 100 99 23.7 991 10.2 0.99550 59 8.968 8.968 3283 100 98 24.4 1005 8.6 0.99614 60 8.992 8.992 3283 100 98 24.4 1005 8.6 0.99591 61 9.116 9. 116 3287 100 99 23.7 993 S.R 0.99569 62 9.287 9. 287 3285 100 96 24.3 990 7.5 0.99552 63 9.796 9. 796 3269 99 99 23.6 1002 5.4 0.99628 64 9.909 9,909 3279 100 96 24.4 1002 4.6 0.99650 65 9.979 9.979 3284 100 99 23.7 1002 4.6 0.99665 66 10.092 10.092 3288 100 93 25.5 1002 R.7 0.99696 67 10.139 10.139 3282 100 94 R5.0 1002 R.7 0.99708 68 10.288 10.288 3278 100 96 24.4 1001 2.4 0.99712 69 10.301 10.301 3281 100 97 24.3 1001 2.4 0.99707 70 10.324 10.324 3287 100 98 24. 1 1002 2.4 0.99695 71 10.463 10.463 3285 100 99 23.8 1001 2.3 0.99700 72 10.589 10.589 3293 100 93 25.5 1002 1.1 0.99685 73 10.653 10.653 3294 100 95 25.0 1002 0.0 0.99724 74 10.688 10.688 3290 100 97 1001 0.0 0.99732 75 10.757 10.757 3291 100 100 23.6 1001 0.0 0.99752 76 10.770 10.770 3284 100 100 R3.6 1001 0.0 0.99746 77 10.828 10.828 3235 98 100 23.3 999 0.0 0.99745 78 10.933 10.933 3202 97 100 23.1 999 0.0 0.99675 79 11.022 11.02R 3112 94 100 RR.6 996 0.0 0.99713, 80 11.083 11.083 3060 93 100 22. 2 993 0.0 0.99712

TABLE 3.2.3 (CONTI)i)ED)

SUSQUEHA))NA SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT=1 CYCLE=l CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- OOHE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POHER POHER FLOH COOLING PRESSURE OE))SITY CORE CASE ( Gtl0/Nll 1 (GHD/HTU) lHHTH) (%) (%) ( BTU/LBN) (PSIA) )%) K-EFFECTIVE 81 82 11.153 11.217 ll. 153 11.217 2991 2943 91 89 100 100 21.8 21.6 992 990 0.0 0.0

0. 99749 0 '9742

'483 11.259 11.332 11.259 11.332 2897 2834 88 86 100 99 21.3 21.0 988 986 0.0 0.0 0.99761 0 '9770 85 11.464 11.464 2776 84 99 20.8 987 0.0 0 '9718 86 11.542 11.542 2714 82 99 20.6 992 0.0 0.99746 87 11.617 11.617 2669 81 100 20.6 1014 0.0 0.99806

TABLE 3.2.3 (CONTINJED)

SUSQUEHA)4'JA SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT=1 CYCLE=R CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POWER POWER FLOW COOLING PRESSURE DENSITY CORE CAGE ( Q1D/HTU ) t QID/t1lU ) (tl4l H) (%) l%) t BTU/LBH) (PSIA) l%) K-EFFECTIVE 88 0.200 9. 634 3271 99 25.0 998 4.2 0.99654 89 0.268 'J. 70R 3286 100 96 R4.4 998 4.2 0.99688 90 0.345 9. 780 3285 100 97 24.2 996 4.3 0.99650 91 0.406 9.841 3289 100 100 R3.5 996 4.3 0.99725 92 0.559 9. 994 3292 100 97 24. 1 996 4.1 0.99726 93 0.725 10.160 3296 100 96 24.4 996 4 1

~ 0.99727 94 0.789 10.224 3295 100 95 24.8 1002 4.1 0.99733 95 0.915 10.350, 3290 100 94 25.3 100R 4.1 0.99732 96 0.962 10.398 3294 100 93 R5.4 1002 4.1 0.9972R 97 1.R48 10.684 3292 100 97 24.3 1001 4.1 0.99712 98 1.331 10.767 3291 100 96 24.6 1000 4.0 0.99758 99 1.451 10.8S7 3R95 100 94 25.1 1000 4.1 0.99707 100 1.528 10.965 3293 100 98 24.0 1001 7.2 0.99669 101 1.661 11.098 3292 100 97 24.4 1001 7.2 0.99688 102 1.803 11 24io 3293 100 95 24.9 1001 7.2 0.99703 103 1.866 11.303 3293 100 94 25.2 1001 7.R 0.99696 104 1.931 11.368 3294 100 97 24.4 1001 7.4 0.99707 105 2.066 11.504 3R91 100 97 24.5 1001 6.6 0.99733 106 2.227 11.665 3R93 100 95 25.0 1001 6.6 0.99734 107 2.381 11.819 3291 100 96 24.8 1000 6.7 0.99733 108 2.415 11.854 3291 100 95 24.9 1000 6.7 0.99727 109 2.500'.587 11.939 3288 100 96 Z4.6 1000 6.8 0.99733 ill 110 112 2.642 R.784 12.026 12.081 12.224 3291 3290 3286 100 100 100 96 96 96 24.7 24.7 R4.6 1000 1000 1000 6.8 6.8 6.8 0.99723 0.99721

0. 9973R 113 2.903 12.343 3292 100 96 24.7 1000 6.8 0.99710 114 3.039 12.479 3299 100 97 24.5 1000 6.8 0.99702 115 3 '97 1Z.538 3290 100 96 24.5 999 6.8 0.99695 116 3.323 12.764 3294 100 97 24.5 1000 7.5 0.99603 117 3.439 12.880 3292- 100 99 23.8 999 7.5 0.99636 118 3.605 13.047 3288 100 99 23.9 999 7.3 0.99683 119 3.688 13.130 3292 100 99 23.9 999 7.R 0.99674 120 3.727 13.169 3292 100 99 23.7 998 7.R 0.99688 121 3.877 13.320 3289 100 98 24.0 1001 7.1 0.99636 122 3.902 13.345 3293 100 99 23.9 1001 7.1 0.99626 123 4.014 13.457 3291 100 99 23.8 1001 6.8 0.99651 124 4.075 13>518 3285 100 99 23.7 1000 '6. 7 0.99679 125 4.403 13.847 3290 100 97 24.4 1000 4.2 0.99697 126 4.513 13.957 3291 100 96 24.8 1000 2.2 0.99708 127 4.598 14.042 3286 100 98 24.2 999 2' 0.99711

TABLE 3 ~ 2. 3 l CONTINUED )

SUSQUEHA)QA SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT=1 CYCLE=2 CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POWER POllER FLOW COOLING PRESSURE DENSITY CORE CASE lG)l0/MTU) lGll0/HTU) l tklTH ) l%) l%) l BTU/LBH) l PSIA) l/) K-EFFECTIVE 128 4. 638 14.082 3290 100 99 23;9 1000 2.2 0.99713 129 4.775 14.220 3286 100 100 23.5 999 1.9 0.99717 130 4.881 14.326 3223 98 100 23.2 996 1.9 0.99721 131 4.953 14.398 3290 100 98 24.2 999 2.0 0.99'726 132 5.038 14.484 3206 100 95 24.8 999 0.2 0.99718 133 5.128 14.574 3292 100 98 24.0 999 0.2 0.99722 134 5. 175 14.621 3285 100 99 23.7 999 0.2 0.99716

TABLE 3.2.3 lCOtlTINUED)

SUSqUEHA)t)A SES HOT CRITICAL'ORE K-EFFECTIVE DATA

"--"-UNIT>>l CYCLE=3 CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE PONE R POHER FLOH COOLING PRESSURE DENSITY CORE CASE l GHD/HTlJ ) ~

l GHD/tlTU ) lHNTH) l%) l%) l BTU/LBH) lPSIA) l/) K-EFFECTIVE 135 0.178 B. 160 3294 100 97 24.4 1002 7.7 0.99368 136 0.286 8.268 3290 100 98 24.1 1001 7.7 0.99377 137 0.423 8.405 3288 100 98 24.2 1000 7.7 0.99374 138 0.543 8.525 3293 100 97 24.4 1000 7.7 0.99378 139 0.771 8.753 3292 100 95 24.9 1001 7.'8 0.99302 140 0.867 8.849 3292 100 97 24.3 1000 8.0 0. 99313 141 0.925 .8.907 3293 100 99 23.8 1000 8.4 0.99307 142 0.967 8.949 3288 100 98 24.2 1000 8.4 0.99286 143 1.084 9.066 3291 100 95 25. 1 1004 7.7 0.99285 144 1.180 9.162 3288 100 94 25.4 1003 7.7 0.99283 145 1.290 9.272 3291 100 96 24.8 1003 8.0 0.99270 146 1.410 9.392 3291 100 25.2 1003 8.0 0.99272 147 1.442 9.424 3292 100 94 25.3 1003 8.0 0.99272 148 1.602 9.584 3292 100 94 25.4 1002 8.1 0.99258 149 1.722 9.704 3292 100 94 25.3 1002 8.2 0.99252 150 1.867 9.849 3293 100 93 25.5 1002 8.3 0.99251 151 1.967 9.949 3287 100 93 25.5 1001 8.4 0.99256 152 2.063 10. 045 3293 100 98 24. 1 1002 9.8 0.99315 153 2.228 10.210 3293 100 96 24. 6 1001 9.8 0.99331 154 2.334 10.316 3292 100 97 24.5 1001 9.9 0.99324 155 2.431 10.413 3289 100 96 24.8 1001 9.9 0.99340 156 2.567 10.549 3294 100 25.2 1001 9.9 0.99343 157 2.782 10.764 3295 100 96 24.8 1000 10.7 0.99344

TABLE 3.2.3 (CONTIQJED )

SUSQUEMAttlA SES MOT CRITICAL CORE K-EFFECTIVE DATA UNIT=2 CYCLE=l CYCLE CORE AVERAGE PERCENT TOTAL CORE QS- DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POHER POWER FLOW COOLING PRESSURE DENSITY CORE CASE (GHO/HTU) (GHD/NTlJ) JHWTM) (%) L%) ( BTlJ/LBN ) t PSIA) l%) K-EFFECTIVE 158 0.131 0.131 1278 39 43 26.8 947 21.8 0.99106 159 0.387 0.387 2347 71 98 18.2 972 16.8 0.99040 160 0.487 0.487 2341 71 98 18.2 971 16.8 0.99082 161 0. 759 0. 759 3170 96 99 23.2 999 13.4 0.98969 162 0. 976 0.976 3282 100 98 23.9 1000 13.6 0.98914 163 1.117 1.117 3288 100 93 25.6 1006 13.1 0.98893 164 1.284 1.284 2654 81 72 28.4 985 14.7 0.98846 165 1.446 1.446 3290 100 95 25.2 1020 13.2 0.98895 166 1.549 1.549 3297 100 96 24.7 1004 13.2 0.98892 167 1.641 1.641 3293 100 96 24.6 1004 13.2 0.98899 168 169

l. 768 1.863 1.768 1.863 3292 100 97 24 4 1004 13.4 0.98902 3290 100 96 24.6 1002 13.4 0.98902 170 1.933 1.933 3288 100 97 24.3 1002 13.5 0.98917 171 2,004 2.004 3293 100 96 24.8 1002 13.5 0.98892 172 2.092 2.092 3293 100 97 24.3 1002 13.7 0.98885 173 2.168 2.168 3292 100 96 24.7 1002 13.7 0.98887 174 2.263 2.263 3293 100 96 24.S 1002 13.9 0.98883 175 2.391 2.391 3293 100 25.2 1002 13 ~ 9 0.98881 176 2.615 2.615 3294 100 98 24.2 1003 15.0 0.98831 177 2.717 2.717 3288 100 98 24.2 1002 15.0 0.98S76 178 2.78S 2.788 3286 100 97 24.4 1002 15.0 0.98892 179 2.868 2.868 3288 100 97 24.5 1002 15.0 0.98905 180 2.906 2.906 3289 100 96 24.6 1002 15.0 0.98913 181 2.999 2.999 3294 100 96 24.8 1002 15.0 0.98924 182 3.117 3.117 3295 100 96 24.7 1002 15.2 0.98935 183 3.264 3.264 3290 100 95 25.1 1002 15.0 0.98912 184 3.392 3.392 3286 100 95 24.8 999 15.0 0.98970 185 3.661 3.661 3286 100 96 24.6 997 15.8 0.99004 186 3.882 3.882 3285 100 93 25.4 1000 15.S 0.99012 187 3.983 3.983 328S 100 94 25.2 1000 15.8 0.99053 188 4.114 4.114 3284 100 94 25.2 999 15.8 0.99099 189 4.575 4.575 3290 100 96 24.5 998 16.4 0.99154 190 4.668 4.668 3291 100 97 24 4 999 16.4 0.99185 191 4.759 4.759 3286 100 96 24.5 1002 16.4 0.99203 192 4.869 4.869 3291 100 95 24.7 1001 16.8 0.99224 193 4.963 4.963 3289 100 96 24.7 1001 16.8 0.99245 194 5.066 5.066 3291 100 98 24.1 1001 17.7 0.99280 195 5.193 5.193 3292 100 99 23.8 1002 17.8 0 '9289 196 5.249 5.249 3293 100 99 23.6 1001 17.8 0.99300 197 5.357 5.357 3288 100 97 24.3 1001 17.6 0.99328

TABLE 3.2.3 ( CONTINUED )

SUSQUEHA)t8L SES HOT CRITICAL CORE K-EFFECTIVE DATA UNIT-"2 CYCLE 1 --<< -

CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POWER POWER FLOH COOLING PRESSURE DENSITY CORE CASE (GHD/HTll) (GHD/HTU) (tSTH) (%) (%) ( BTU/LBH) (PSIA) (%) K-EFFECTIVE 198 5.523 5.523 3292 100 97 R4.3 1005 17.6 0.99357 199 5.616 5.616 3290 100 98 24.1 100R 17.6 0.9936R 200 5.726 5.726 32.92 100 99 23.9 1002 17.6 0.99376 201 5.832 5.832 3284 100 99 23.8 1002 17.6 0.99394 202 5.935 5.935 3291 100 98 24.1 1004 17 ~ 1 0.99371 R03 6.'028 6.028 3291 100 99 23.9 1004 17.1 0.99383 204 6.122 6.122 3RSB 100 97 24.3 1003 16.8 0.99393 205 6.216 6.216 3286 100 98 R4.0 1004 16.8 0. 99412 206 6.318 6.318 3287 100 99 23.9 1003 16.8 0.99407 207 6.494 6.494 3288 100 98 24.R 1008 16.8 0.99400 208 6.5'75 6.575 3294 100 96 24.8 1009 16.4 0.99385 209 6.752 6.752 3295 100 99 24.0 1009 16.4 0. 99440 210 6.817 6.817 3292 100 97 24.3 1001 16.1 0.99416 211 6.893 6.893 3294 100 98 24.1 100R 16.1 0.99436 212 6.924 6.924 3292 100 98 24.0 1002 16. 1 0.99429 213 7.047 7.047 3287 100 97 24.2 1001 14. 9 0.99433 214 7.140 7.140 3287 100 93 25.4 1001 14.4 0.99401 215 7.313 7.313 3R94 100 97 24.3 1001 14.4 0.99448 216 7.396 7.396 3293 100 99 23.8 1001 14.4 0,99456 217 7.561 7.561 "

2671 81 71 R9.9 978 14. 7 0.99331 218 7.70R 7.70R 3288 100 98 24.1 1005 13.6 0.99396 219 7 '79 7.779 3293 100 96 24.7 1004 13.2 0.99385 RRO 7.842 7.842 3289 100 98 R4.1 1004 13.2 0 ~ 99I418 221 7.98R 7.982 3292 100 98 24.2 1004 12.8 0.99459 222 8.100 8.100 3293 100 95 R4.9 1003 12.6 0. 99443 223 8,178 8.178 3283 100 97 24.4 1003 12.6 0. 99459 224 8.390 8.390 3287 100 98 R4.1 1005 1R.7 0. 99454 RR5 8.596 8.596 3293 100 95 24.8 1002 12.3 0.99380 226 8.767 8.767 3288 100 99 23.7 1003 12. 0 0.99451 227 8.880 8.880 3292 100 94 25.1 1003 S. 7- 0.99501 228 8.973 8.973 3292 100 97 1002 8.7 0.99498 229 9.053 9.053 3295 100 99 23.8 1003 8.6 0.99534 230 231 232 9.275 9.412 9.539 9.275 9.412 9.539 3290 3289 3293 100 100 100

'996 96 24.5 23.8 24.5 1002 1002 1002 7.4 7.3 6'

0 ~ 99511 0.99501 0.99515 233 9.666 9.666 3285 100 99 23.6 1001 6.1 0.99524 234 9.835 9.835 3288 100 94 25.1 1001 4.5 0.99529 235 9.986 9.986 3290 100 98 24.0 1002 0.99525 236 10.067 10.067 3262 99 99 R3.5 1001 0.99525 237 10.192 10.192 3284 100 100 23.6 1003 3.6 0.995R7

TABLE 3.2+3 (CONTI)NEO)

SUSQUEHAN SES HOT CRITICAL CORE K-EFFECTIVE DATA lNIT=R CYCLE=l CYCLE CORE AVERAGE PERCEt)T TOTAL CORE SUB- OOt)E CONTROL ROO CALCULATED EXPORJRE EXPOSURE POHER POHER FLOH COOLING PRESSURE DENSITY CORE CASE )GHO/t)TU) )GHD/HTU) )NPH) )%) t%) t BTlJ/LBH ) (PSIA) t%) K-EFFECTIVE 238 10.351 10.351 3293 100 97 24.3 1004 R.9 0.99537 239 10.467 10.467 3290 100 100 23.5 1003 2.7 0.99543 24IO 10.635 10.635 3279 100 97 24.2 1002 R.R 0.99601 241 10.675 10.675 3280 100 93 25.4 1002 0.0 0.99613 242 10.789 10.789 3288 100 97 R4 ~ 2 1003 0.0 0.99601 243 10.851 10.851 3285 100 99 23.7 1002 0.0 0.99581 244 11.007 11.007 3163 96 100 22.8 998 0.0 0.99607 245 11.109 11.109 3085 94 100 22.3 995 0.0 0.99622 246 11.28R 11.282 3016 92 100 22.3 1007 1.7 0.99555 247 11.333 11.333 R979 90 100 2R.O 1007 1.7 0.99553 24S 11.436 11.436 2858 87 100 21.4 1006 1.7 0+99625 249 11.517 11.517 R7SS 85 100 21.0 1006 1.7 0.99653 250 11.642 11.642 2685 82 100 20.4 1005 1.7 0.99671 251 11.824 11.824 2575 78 100 19.8 1001 3.4 0.99621 252 11.915 11.915 2478 75 100 19.1 998 3.4 0.99667 253 11.984 11.984 2404 73 100 18.7 996 3.4 0.99707 254 12.050 12.050 2350 71 100 18.3 994 34 0.99706

TABLE 3.2.3 )CONTINUED)

SUSQUEHARSL SES HOT CRITICAL CORE K-EFFECTI DATA

-- - l5IT=2 CYCLE=2 CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB DONE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POHER PO)lER FLOH COOLING PRESSURE DENSITY CORE CASE lGND/HTU) lGHO/))TU) l tOITN ) UZI )%) l BTU/LB)l) [PSZAI l/) K-EFFECTIVE 255 0.310 8.003 3290 100 96 24.4 1000 8.3 0.99563 256 0.430 8'123 3292 100 96 1000 8.3 0.99558 257 0.583 8.276 3294 100 96 24.4 1000 8.3 0.99525

TABLE 3.2.4 SUSQUEHANNA SES TARGET VS SIMULATE-E CALCULATED CRITICAL CORE K-EFFECTIVE STATISTICS Number Mean Standard of Observations Difference* Deviation Ujcl 87 0.00035 0.00059 U2C1 97 -0.00026 0.00050 U1C2 47 -0.00020 0.00046 U2C2 0.00186 0.00023 Ulc3 23 0.00015 0.00032 All 257 0.00002 0.00061

• Mean Difference is the average difference of the SIMULATE-E calculated K-effective minus the target K-effective.

TABLE 3 2 5 SUSQUEHANNA SES UNIT 2 CYCLE 2 CORE K-EFFECTIVE SENSITIVITY TO MEASURED CORE OPERATING DATA UNCERTAINTIES Initial Conditions Core Thermal Power 3293 MW Total Core Flow 100x10 6 ibm/hr Core Inlet Subcooling 24 Btu/ibm Pressure 1000 psia Measured Measurement Standard Deviation* Core K-effective Sensitivity Parameter (*) (ax)

Core Thermal Power 1.8 0.00097 P

Total Core Flow 2.5 0.00098 f

Core Inlet Subcooling 5.2 0.00061 Pressure 0.5 0.00006 pres 1/2 Total 4 b, + E f + + 0.00151 <K p DHS pres

• Source: "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NED0-10958-A, January, 1977.

80-

TABLE 3 2 6 SUSQUEHANNA SES C2KCULATED COLD XENON-FREE CRITICAL CORE K-EI."FECTIVES UNIT 1 CYCLE 1 Core Average Cycle Control Rod Core Calculated Exposure . Exposure Density Temperature Core Case (GWD/MTU) (Gm/mo) (*) (DEG F) K-Effective 0.000 0.000 74 101.8 1.00045 0.000 0.000 74 105.9 1.00027 0.000 0.000 74 122.5 0.99914 0.000 0.000 73 141.0 0.99985 0.000 0.000 74 120.0 1.00040 0.958 0.958 72 200.0 0.99698 1.490 1.490 73 186.0 0.99674 5.110 5.110 74 182.5 0.99877 5.185 5.185 74 164.0 0.99821 UNIT 1 CYCLE 2 Core Average Cycle Control Rod Core Calculated Exposure Exposure Density Temperature Core Case (GWD/NTU) (GND/NTU) (~) (DEG F) K-effective 10 9.434 0.000 73 157.1 1.00512 ll 12 9.434 9.434 0.000 0.000 71 158.1 1.00498 71 180.4 1.OO466 13 9.434 0.000 68 205.8 1.00359 14 9.434 0.000 68 211.1 1.00341 UNIT 1 CYCLE 3-Core Average Cycle Control Rod Core Calculated Exposure Exposure Density Temperature Core Case (GWD/MTU) (GWD/mv) (*) (DEG F) K-effective 15 7.982 0.000 75 174.2 1.00110 16 7.982 0.000 75 175.8 1.00112 17 7.982 0.000 75 190.3 1.00067 18 7.982 0.000 74 189.9 1.00086 19 7.982 0.000 74 195.4 1.00067 20 7.982 0.000 74 202.2 1.00046 21 7.982 0.000 74 206.2 1.00029 22 8.612 0.630 74 170.5 1.00128 23 8.612 0.630 74 156.3 1.00171 24 10.601 2.619 74 209.4 0.99950 81

TABLE:3.2.6 (continued)

SUSQUEHANNA SES CALCULATED COLD XENON-PREE CRITICAL CORE K-EPI. ECTIVES UNIT 2 CYCLE 1 Core Average Cycle Control Rod Core Calculated

. Exposure Exposure Density Temperatures Core Case (GWD/RZtJ) (GWO/mV) (*) (DEG F) K-effective 25 0.000 0.000 74 111.4 0.99827 26* 0.000 0.000 98 117.0 0.99706 27* 0.000 0.000 98 118.8 0.99696 28* 0.000 0.000 98 119.7 0.99835 29 0.000 0.000 74 120.7 0.99756 30 0.000 0.000 75 136.0 0.99569 31 0.158 0.158 74 162.0 0.99806 32 0.847 0.847 73 163.0 0.99639 33 0.976 0.976 74 115.0 0.99746 34 0.976 0.976 73 161.5 0.99688 35 2.391 2.391 73 207.0 0.99426 36 8.390 8.390 74 158.0 1.00272 37 11.208 11.208 58 195.0 1.00429 UNIT 2 CYCLE 2 Core Average Cycle Control Rod Core Calculated Exposure Exposure Density Temperature Core Case (cwo/mu) (cd/mo) (*) (DEG F) K-effective 38 7.693 0.000 75 133.0 1.00084 39 7.693 0.000 75 139.5 1.00083

• Local Criticals TABLE 3.2 7 SUS UEHANNA SES COLD MINUS HOT CRITICAL CORE K-EFFECTIVE UNIT 1 CYCLE 1 Core Average Cycle Core K cold K.

hot K cold hot Exposure Exposure Temperature (CWO/MTU) (GWD/MTU) (DEG F) calc calc calc target 0.000 0.000 101.8 0.00802 0.00765 0.000 0.000 105.9 0.00784 0.00747 0.000 0.000 120.0 0.00797 0.00760 0.000 0.000 122.5 0.00671 0.00634 0.000 0.000 141.0 0.00742 0.00705 0.958 0.958 200.0 0.00656 0.00769 1.490 1.490 186.0 0.00741 0.00796 5.110 5.110 182.5 0.00612 0.00631 5.185 5.185 164.0 0.00547 0.00567 UlC1 Average: 0.00706 0.00708 U1C1 Standard Deviation: 0.00090 0.00079 UNIT 1 CYCLE 2 cold Khot Core Average Cycle Core K K Exposure Exposure Temperature cold hot (GWD/MTU) (CWO/MTU) (DEG F) calc calc calc target 9.434 0.000 157.1 0.00811 0.00786 9.434 0.000 158.1 0.00797 0.00772 9.434 0.000 180.4 0.00765 0.00740 9.434 0.000 205.8 0.00658 0.00633 9.434 0.000 211.1 0.00640 0.00615 U1C2 Average: 0.00734 0.00710 U1C2 Standard Deviation: 0.00080 0.00080 UNIT 1 CYCLE 3 Core Average Cycle Core K K K Exposure Exposure Temperature cold hot cold hot (CWO/MTU) (MWO/MTU) (DEG F) calc calc calc target 7.982 0.000 174.2 0.00672 0.00665 7.982 0.000 175.8 0.00674 0.00667 7.982 0.000 189.9 0.00648 0.00641 7.982 0.000 190.3 0.00629 0.00622 7.982 0.000 195.4 0.00629 0.00622 7.982 0.000 202.2 0.00608 0.00601 7.982 0.000 206.2 0.00591 0.00584 8.612 0.630 156.3 0.00843 0.00961 8.612 0.630 170.5 0.00800 0.00918 10.601 2.619 209.4 0.00610 0.00766 U1C3 Average: 0.00670 0.00705 UlC3 Standard Deviation: 0.00084 0.00134 83

TABLE 3.2.7 (continued)

SUS UEKLNNA SES COLD MINUS HOT CRITICAL CORE K-EFFECTIVE UNIT 2 CYCLE 1 Core Average Cycle Core K cold hot. K cold hot Exposure Exposure Temperature (MWD/MTU) (MWD/MTU) (DEG F) calc calc calc target 0.000 0.000 111.4 0.00659 0.00547 0.000* 0.000 117.0 0.00538 0.00426 0.000* 0.000 118.8 0.00528 0.00416 0.000* 0.000 119.7 0.00667 0.00555 0.000 0.000 120.7 0.00588 0.00476 0.000 0.000 136.0 0.00401 0.00289 0.158 0.158 162.0 0.00692 0.00615 0.847 0.847 163.0 0.00687 0.00689 0.976 0.976 115.0 0.00812 0.00820 0.976 0.976 161.5 0.00754 0.00762 2.391 2.391 207.0 0.00539 0.00511 8.390 8.390 158.0 0.00802 0.00744 11.208 11.208 195.0 0.00819 0.00782 U2C1 Average: 0.00653 0.00587 U2C1 Standard Deviation: 0.00129 0.00164 UNIT 2 CYCLE 2 Core Average Cycle Core K 1C K 'hot Exposure Exposure Temperature cold hot cold (MWD/MTU) (MWD/MTU) (DEG F) calc calc calc target 7.693 0.000 133.0 0.00472 0.00548 7.963 0.000 139.5 0;00471 0.00547 U2C2 Average:, 0.00472 0.00547 U2C2 Standard, Deviation: 0.00001 0.00001 Overall Average: 0.00671 0.00659 Overall Standard Deviation: 0.00111 0.00137

• Local Criticals 84-

3 2.8 SUS UKQLNNA SES UNIT 1 CYCLE 1 TIP RESPONSE COMPARISONS Nodal Radial Cycle Control Nodal TIP Radial TIP Exposure Rod RMS Asymmetry RMS Asymmetry Date (CWO/MTU) Sequence (~) (~) (~) (*)

12/16/82 0.221 B2 5.09 2.78 02/07/83 0.836 A2 4.03 2.60 1.58 1.18 04/04/83 1.490 B2 5.04 1.70 06/09/83 1.799 B2 4.97 1.79 08/10/83 2.706 A1 5.12 4.26 1.62 1.58 08/19/83 2.906 A1 5.21 4.26 1.63 1.58 09/13/83 3.367 Bl 5.62 1.71 10/03/83 3.836 B1 5.46 1.74 10/18/83 4.193 B1 5.62 1.91 11/01/83 4.517 B1 5.60 1.91 12/01/83 5.070 A2 5.93 4.72 1.81 1.60 04/03/84 5.410 A2 6.12 4.96 1.96 1.74 04/12/84 5.614 A2 6.14 5.16 1.85 1.63 04/26/84 5.918 B2 5.72 1.98 05/24/84 6.563 B2 5.80 1.97 05/31/84 6.716 Al 5.82 4.89 1.89 1.47 06/08/84 6.893 A1 5.38 5.02 1.88 1.75 06/25/84 7.235 Al 5.00 4.60 1.94 1.63 07/24/84 7.638 Bl 4.75 1.85 08/02/84 7.840 Bl 4.61 1.87 08/16/84 8.164 B1 4.53 1.69 08/24/84 8.341 A2 4.53 4.73 2.13 1.60 08/30/84 8.481 A2 4.57 4.84 2.14 1.77 09/04/84 8.602 A2 4.52 4.83 2.22 1.93 11/30/84 10.288 B2 4.73 1.91 12/13/84 10.589 B2 4.80 1.70 12/16/84 10.653 ARO 4.68 1.67 12/21/84** 10.770 ARO 4.96 4.45 1.72 1.57 01/10/85** 11.083 ARO 5.93 4.56 1.73 1.62 02/01/85** 11.464 ARO 6.07 4.72 1.62 1.64 02/08/85** 11.617 ARO 6.03 4.80 1.76 1.74

• Reactor conditions for this TIP set: 60% of rated flow 40% of rated power

• • End of cycle power coastdown data 85-

TABLE 3 2 9 SU UEEGLNNA SES.UNIT 1 CYCLE 2 TIP RESPONSE COMPARISONS Nodal Radial Cycle Control Nodal TIP Radial TIP Exposure Rod RMS Asymmetry RMS Asymmetxy Date (CWO/MTU) ~ee ence (e1 (*) (e) (4) 06/24/85 0.200 Al 4.79 3.64 2.52 2.24 07/03/85 0.406 Al 4.89 3.67 2.72 2.40 07/19/85 . 0.789 Al 4.76 3.28 08/08/85 1.248 B1 5.78 2.86 08/20/85 1.528 B1 5.17 2.70 09/06/85 1.931 Bl 5.97 2.75 09/12/85 2.066 A2 6.42 3.57 2.57 2.40 09/27/85 2.415 A2 5.58 3.75 2.73 2.55 10/04/85 2.587 A2 5.48 3.80 2.72 2.56 10/23/85 3.039 A2 4.55 3.74 2.70 2.51 11/15/85 3.323 B2 5.04 3.09 12/12/85 3.877 Al 4.75 3.02 01/14/86 4.638 A1 4.99 4.37 2.64 2.49 86-

TABLE 3.2 10 SUS UEMHNA UNIT SES 1 CYCLE 3 TIP RESPONSE COMPARISONS Nodal Radial Cycle Contxol Nodal TIP Radial TIP Exposure Rod RMS Asymmetxy RMS Asymmetxy Date (GWD/MTU) ~ee ence (a) (~) (*) (~)

's/os/86 0.178 A1 5.16 3.41 2. 74 2.47 07/03/86 0.925 Al 6.06 4.34 4.14 3.58 07/10/86 1.084 B1 5.68 2.80 08/20/86 2.063 A2 8.12 3.58 2.82 2.55 08/27/86 2.228 A2 8.71 3.84 2.89 2.69 09/10/86 2.567 A2 9.03 6.28 3.75 5.13 87

SU VEZGLNNA SES UNIT 2 CYCLE 1 TIP RESPONSE COMPARISONS Nodal Radial Cycle Control Nodal TIP Radial TIP Exposure Rod RMS Asymmetry RMS Asymmetry Date (GWO/MTU) Sequence (a) (*) (~) (~)

07/23/84 0.131 A2 7.05 5.24 2.82 1.34 09/12/84 0.387 A2 5.37 5.09 2.58 2.23 10/08/84 0.759 A2 4.73 5.13 2.30 2.18 01/16/85 1.117 A2 4.76 5.71 2.20 2.34 02/07/85 1.446 B2 5.51 2.58 03/07/85 2.092 B2 5.43 2.64 03/20/85 2.391 B2 5.58 2.68 04/04/85 2.615 Al 5.65 5.79 2.31 2.51 04/15/85 2.868 Al 5.75 6.35 2.58 2.87 05/15/85 3.392 Al 5.93 7.30 2.75 2.88 06/10/85 3.882 B1 5.79 2.44 06/20/85 4.114 Bl 5.79 2.60 08/01/85 4.869 A2 6.61 6.56 2.76 2.61 08/12/85 5.066 A2 7.83 6.18 2.59 2.23 08/20/85 5.249 A2 7.80 7.78 3.82 3.58 09/09/85 5.726 A2 7.70 8.53 3.99 4.03 10/01/85 6.216 B2 7.81 5.18 10/18/85 6.575 B2'2 5.84 3.04 10/28/85 6.817 5.51 2.68 11/19/85 7.313 Al 7.55 9.04 5.96 5.96 12/17/85 7.779 Bl 4.92 3.08 01/30/86 8.596 A2 4.94 6.30 2.99 2.86 02/19/86 9.053 A2 5.75 7.83 4.56 5.12 03/06/86 9.412 A2 6.99 9.58 6.35 6.76 03/12/86 9.539 B2 5.12 2.37 03/25/86 9.835 B2 5.56 2.55 04/04/86 10.067 B2 5.92 2.68 04/29/86 10.635 B2 6.02 2.36 05/15/86 11.007* ARO 7.21 5.78 2.46 2.91 06/23/86 11.282* B2 6.56 2.38 07/11/86 11.642* B2 6.81 2.45 08/08/86 12.050* B2 7.81 3.44

• End of cycle power coastdown data.

88

TABLE 3.2.12

SUMMARY

OF SU UEHANNA SES TIP RESPONSE COMPARISONS Average Average Nodal Radial .

Number RMS RMS Unit a cle of TIP Sets (*) (~)

Ulcl 31 5.24 1.86 U1C2 13 5.24 2.79 U1C3 7.13 3.19 U2C1 32 6.17 3.07 Overall Average 82 5.74 2.58 89

TABLE 3 2 13

SUMMARY

OP SUS UEHANNA SES TIP RESPONSE ASYMMETRIES Average Average Nodal Radial Number Asymmetry Asymmetry Unit a cle of TIP sets (~) (~)

Ulcl 16 4.59 1.63 U1C2 3.79 2.45 U1C3 4.29 3.28 U2C1 16 6.76 3.28 Overall Average 44 5.22 2.55 90

FIGURE 3.2. I SIMULATE-E HOT AND COLD CRITICAL CORE K-EFFECTIVES VS CORE AVERAGE EXPOSURE 1.01 k

1.00-

v j +

I O

0;""':

,~o::..:Legend o U1C1 HOT 0:9S I mi cI U2C1HOT U1C2HOT lC O .. 'i7 U2C2HOT O

U1C3HOT 0.98-- U1C1COLD U2C1COLD U1C2COLD U2C2COLD

+ U1C3COLD 0 1 2 3 4 5 6 7 S 9 10 11. 12 13 14 15 COBE AVERAGE EXPOSURE (GWD/MTU)

FIGURE 3.2.2 SIMULATE-E HOT CRITICAL CORE K-EFFECTIVE VS CORE THERMAL POWER 1.01 1.00 LLI I

I-O lLI U 0 0.99-I UJ LL' 0O .. .........

~ Legend

~ --.".-0 U1C1 0.98 "" "" cI U2C1 U1C2 U2C2 U1C3 0.97 50 60 66 70 76 80 86 90 100 106 CORE THERMAL POWER (% OF RATED)

FIGURE 3.2.3 SIMULATE-E HOT CRITICAL CORE K-EFFECTIVE VS TOTAL GORE FLOW 1.01-100

--" 0"--

0 I IP O

0 0.99 I 0 0

CC 0O Legend ...':,.......

-" -"-.-0 U1C1 0.98 ..""""'0 U2C1 U1C2

""v U2C2 o U1C3 0.97-40 50 60 70 80 90 100 TOTAL CORE FLOW (% OF RATED)

FIGURE 3.2.4 SIMULATE-E HOT CRITICAL CORE K-EFFECTIVE VS CORE INLET SUBCOOLING 1.01 1.00 o o .:

. Booooo:.o o LU C3., O...CI.....:.

8 I

C3 UJ 0 99-I hC CC 0O ..... Legend,

- -.-"o UlC1 0.98- 0 U2C1 U1C2 U2C2 o U1C3 0.97 15 16 17 18 19 20 21 22 23 24 25 26 27. 28 29 30 CORE INLET SUBCOOLING (BTU/LBM)

FIGURE 3.2.5 SIMULATE-E HOT CRITICAL CORE K-EFFECTIVE VS DOME PRESSURE 1.01 1.00-od'oze,"

jjjo" I-O

~ I" ---.

LIJ 02..:...;.

D 0.99 I

UJ  : 0 CC O ....... Legend C3 I

~ -" -o U1C1 0.98- "-.a U2C1 U1C2 U2C2 0 U1C3 0.97 940 950 960 970 . 980 990 1000 1010 1020 1030 DOME PRESSURE {PSIA)

FIGURE 3.2.6 SIMULATE-E HOT CRITICAL CORE K-EFFECTIYE VS CRITICAL CONTROL ROD DENSITY 1.01 I ~

1.00 ~ C 1 ~ ~

LLI 0

a Pen,:

0 I?I jp:4k~:,

gg: Qj

@'P, jP Q::c5: cD~

I- "HD'lf."--gg' O::: I.::.:.:.

cl O ..Q...........,........~.........

IJJ

O 0.99-I IJJ CC 0O ...,.;:.:.... Legend::...-....::....:...:, .....,:.. .:

s ~ --:-""o UlC1 0.98- ""i"""0 U2C1 UlC2 U2C2 o U1C3 0.97-2 4 6 8 10 12 14 16 18 20 22 0

CRITICAL CONTROL ROD DENSITY (%)

hag ~ ae eRl

FIGURE 3.2.7 TARGET AND SIMULATE-E CALCULATED HOT CRITICAL CORE K-EFFECTIVES VS CORE AVERAGE EXPOSURE 1.01-1.00- .--'---'-.-.:.----'---.'- U1C2 TARGET:""'---.'-.--.'" -'"-.:.

LIJ U1C1 and U2C1 TARGET '..-.:---:----.

I O

r ~

0 0.9S-I U1C3 TARGET

--:"--'-.: U2C2 TARGET:."-.:-----:.-.--'.-"--: I \ r UJ CC 0

O Legend ..,:.....

~ ~

~ -o U1C1

0. C 0 U2C1 U1C2 U2C2 o U1C3 0.97 .

0 3 4 5 6 7 8 9 10 11 12 13 14 15 CORE AVERAGE EXPOSURE (GWD/MTU)

FJGURE 3.2.8

)Q)

SUSQUEHANNA SES UNITS 1 AND 2 CORE TIP LOCATIONS LINE OF TIP SYMMETRY 59 57 55 53 51 49

+

47 45 43 41 39 +++++

37 35 +++++

33 31 +++++ +

29 27 +++++

25 23 +++++

21

++++ ++

++ +

++

++

00 02 04 0608 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 4 042 444648505254565860 X

Control Rod Location Location For Individual TIP Response Comparisons

~ Traversing In-core Probe Location

FIGURE 3.2.9 SUSQUEHANNA SES RELATIVE NODAL RMS OF TIP RESPONSE COMPARISONS 10 FIGURE 3.2.9a V) 8 m Q.p lZ Q Q Qg 6 0~".'-"-~--- ----'.-"--O -"-

Cl 0

LLI p

Qp Q

00::

Q 0 O. 0

~: .

g, 0

~+p Q.

b.

0 4 ~ ".O- 'I Legend l~

0 U1C1 UJ CL" 2 i...b, U1C2 CL o U1C3 I-Q U2C1 0

0 2 4 6 8 10 12 14 16 CORE AVERAGE EXPOSURE (GWD/MTU) 10 FIGURE 3.2.9b V) 8 Q

Q CL" Q'G 6 C "..". """"Cl'"'""b;"""0 OO D Q.'00~ QQO 0 0 0 QQO ~

Q 0Z 00

.: b, QO 4 ~ "--"O.

Legend I~

0 U1C1 >0 D (4

>0 LLI CC 2 U1C2  ! ' 4 Qw Ow CL I-UlC3 ~

as i aoQ Q

wo- i i LJ U2C1 0 I 0 0.2 0.4 0.6 0.8 FRACTION OF CYCLE LENGTH 99

FIGURE 3.2.10 SUSQUEHANNA SES UNIT 1 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 1.490 GWD/MTU CYCLE EXPOSURE 180 180 140 120 '"" "" o'"<>" o" g + + g" 0+

I- q () Q zD 100 0

Ill +

zz 0 80 Co Q.

80 40 20 0 I 0 1 2 3 4 6 6 7 8 9 10 11 1213 141516 17 18 192021222324 CORE AXIAL NOOE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 100

FIGURE 3.2.11 SUSQUEHANNA SES UNIT 1 CYCLE 1 RADIAL TIP RESPONSE COMPARISONS 1.490 GWD/MTU CYCLE EXPOSURE I

61 I 59

57. -1.58 0.73 1.56 -2.14 55 53 51 4.9 0.53 1.5 -0.07 -0.53 2.49 1.15 47 45 43 + -0.55 -0.20 41 ss + ++ 0.89 23 0.8 37

+++

33 29

+++ 3.26 -1.65 -1.91 -0.15 19 I

+ + +

25 21

+++ 3.89 16 -1. 63 -2.46 -0.24 -0.81 1.6 3 I

I 19 I, I I

-3.47 -0.13 -0.81 17 15 13 I

1.77 I

++ 0 I

0. 03 11 9

+ -0.24 -0.57 0.33 -1.12 2.00 7

l I

I

.I I

I I

I I I I I I I I I 000204060810 12 14 16 18 2 022 24 26 28303 234363840 424 4 464850 52 54 56 58 60 Diff = [(Calc - Mess)iCore Avg TIP Response] X 100%

101

FIGURE 3.2.12 SUSQUEHANNA SES UNIT 1 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 1.490 GWD/MTU CYCLE EXPOSURE MONITOR LOCAllOH 80>30 MONITOR LOCAllOH 4b>bj 140 140 lee lee

<<0 0

0 1$ 0 0 C

X obbo b 0 I 0 0 140 "0 L ls 140 I

00 ~ ~ $ ~4 0

4~

I l-.o--

4

>>--- so T

t' 0

0 ee C 40 4 40 44 I ~ $ 4 4 0 1 4 0 10<<4$ 1~ I41$ 10ITI414$ $ $ 14444$ 4 I 4 4 ~ ~ ~ 'I ~ ~ 10 11 1414 Ii I ~ I ~ IT Ii 14$ 4 $ 14444$ 4 CORP. AXIALHOOR CORK AXIALHODR t NCA$VNSOht Sects<<0$ t <<$ $ 4UNCO Ttt Nseto<<4$

OALOVLATCO ht $ $ 44ONSS 0 OAIOUMTCDht NsstoNSS

~ ooNTNVL sos tosmo<< ~ co<<TNVL $ 00 tosmoN MONITOR LOCATION e0,00 MONITOR LOCATOH 32.80 144 1$ 0 lee Iee 140 140 0

0 0 0 4 V 0~80 II 0

6 I I TOI 0 is

>40 tto 1$ 0 2:

~0 ~0 C.

a 0 0

0 40 44>>

I 44 0

~ I 4 $ 4 4 4 T 4 ~ 14111414141$ 1411141$ $ 4$ 14$ 44$ 4 0 \ ~ $ 4 ~ ~ 1 ~ ~ 14 41141444>4 I ~ IT IL I~ 4441144414 CORK AXIALNOOK CORK AXIALHOOR t NCASVIICO Tlt IlsstONss t NcACUIIco ht scstoNec o OALOVLATCOTll IICsto<<0C o OALOVLATCDTlt SCetONSS N oo<<TNOL Aoo tosmo<< ~ OO<<hloL soo toshIO<<

- 102

FIGURE 3.2.13 SUSQUEHANNA SES UNIT 1'CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 6.918 GWD/MTU CYCLE'XPOSURE 180 180 140

+ 0 5 0 Q 120 .0..

I- 0 0

D +

100 Q

""0" III +

Z Z

80 V)

CL 60 0

40 0

20 0 l 0 1 2 3 4 5 6 7 8 9 10 1112131415161718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 103

FIGURE 3.2.14 SUSQUEHANNA SES UNIT t CYCLE 1 RADIAL TIP RESPONSE COMPARISONS 5.918 GWD/MTU CYCLE EXPOSURE 61 59 57 -2.75 -0.93

++ 1.22 -5.29 55 S3 51

++

49 47 0.07 .23 -0.70 -0.76

++ 0.5 45 43 ++ ++

41 39 37

++ 2.5

+ ++

I 0.6 1.49

+

40 35 I 33 2.3 79 2.2 -0.01 0.5 0 31 29 27 1.91

++

25 23

'7 a

iI 40

++ 36 .14 19 17 2.0 9 2.4 3 -2. 67 -0.22 -0.86 I

15 13 I

9 78 .17 -0 .76 -3.26 95 I

1 I I I I 0002 04060810 12 14 16 18 2Q 22 24 26 28 30 32 34 36 3840 42 44 464850 52 54 56 58 60 Diff = [(Calc - Meas}/Core Avg TIP Responsej X 100%

104

FIGURE 3.2.15 SUSQUEHANNA SES UNIT 1 CYCLE 1 INDIYIDUALTIP RESPONSE COMPARISONS 5.9'I8 GWD/MTU CYCLE EXPOSURE MONITOR LOCATION dIL$ $ MONITOR LOCATION 4$ ,$ $

ISS ISS lls f

IIS --- 4- od, 4

I

~I tts L

r 0 0

5 lg 100 lc Lts 0  ??Nl 00- T I 4

IL C 00 I= se 0

I~ ~0 ts ts

~ t ~ l ~ 7 ~ lt 11 ll ll lt Il 10 17 IS 10 le ll lt ll lt 0 ~ t ~ 4 ~ ~ 7 ~ ~ 10 11 Il 10 lt ll IS 17 I~ 10 ts tl tt ll lt CO CORe AXIALNooe f NNAsvaso ll~ aespoNss f NNAsvaso TlpllpNsspoNse 0 OAIovMTtoTI ~ aeepoaee 0 OALovMTSO asepoase

~ CONTNOL aoe poeITION ~ OONTNOL aoo posITION MOHITOA LOCATIOH 00,$ $ MONITOR LOCATIOH $ $ ,$ $

Its lts Ise Ne t.h..e 4 vs 0 0 0

e Ite f0]f o

0 t L r

0 e ~

~

eef d lls se ~0 0

se ..L

~0 4" ts I I I l ~ t 0 ~ 7 ~ ~ 10111ll ~ %101 ~ 171 ~ Iltetlltlllt 0 I t 0 ~ S ~ 'I ~ ~ 10111tllltll'I~ 17 II 70 ts Il tt ll ll COIIe AXIAI.Nooe coee AxIALNooe o

IISAsvalo TI~ aeepoase CALOVLATSOTIP NSSPONee f0 vKAsvalo lipTlpalspoase OALOOLATSO asspoase

~ ooNTNOL aoo poelsoa ~ ooataoL aoo posITIVN 105

g ~l FlGURE 3.2.16 SUSQUEHANNA SES UNIT 1 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 11.617 GWD/MTU CYCLE EXPOSURE 180 1BQ 140 0 0 0

+ 0 0 ':

0 120 I- 0 zD 0

100 ILl "Z 0 0

+

80 ~ +

CO 0

Q.

eo 0

+

40 20 0

0 1 2 3 4 6 B 7 8 9 10 11 1213 14 1616 17 18192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE 106-

FIGURE 3.2.'17 SUSQUEHANNA SES UNIT 1 CYCLE 1 RADIAL TIP RESPONSE COMPARISONS 11.6 17 GWD/MTU CYCLE EXPOSURE I

61 59 57 -1.89 -2.88 0.84 -2.01 55 53 51

-0.47 -2.18

++

49 47 0.21 0.01 1.66

++

43

++ -0.36 86 2.0 2.0 4.3

++

41 39

++

37 35 33 -0,88 -0.10 31 29 20 60 1.82 3.64 1.81

++

27 25 0.5 -0.39 -0.24 .70 3.4 -2.45 -0.08 23 21 19

.17 1.08 -1.08 -0.80 0.2 15 13 9

7 5

3

'2.57 45 -0 28 -0.88 Y1 I 0002 04 0608 10 12 14 16 18 20 22 24 26 28 30 32 34 3638 40 42 44 46 4850 52 54 56 58 60 X

Diff = [(Calc - Mess)/Core Avg TIP Responsej X 100%

107

flGURE 3.2.18 SUSQUEHANNA SES UNIT 1 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 18.617 GWD/MTU CYCLE EXPOSURE MOIKTOR LOCATION KK,KS MONITOR LOCATION CC,SK IN IN r Nl Ite g

Z Ile 0 4

-K Ig IN r ee o 4 +

~

0 j le o

0 r

0

~ w I~

~ ~ ~ I I I ~ I ~ ~ 14 II II IS 11 le 14 lf Illetell tttttl ~ I 1 ~ ~ I ~ I 4 ~ 14 11 It 14 11 'N 14 lf I ~ Il te tl tl tl II CORE AXIALNOOK COILK AXIALNODE NIARURRO Tlo 4 Reoooeo 4 NRARVRIOTIP RSSIONRR 0 OAIOVLAIROIIR Ra40aeo 0 OALOVLATCOTIR Rterooeo

~ CORIROL ROO RoemON ~ coof RVL 100 toemoN MOIKTOR LOCATION 40,8K Ne II~

11 ~ 1N 4 0 og J4 "r e 0 o 0

+ 0 0 0 0+

lie ee 0

o r'-t o Z

~ I 0 0

4 0

~ I '0 ' L l.

+

0 te

~ ~ I I ~ ~ ~ I ~ ~ 14 11 IIIS It lllelf I~ lel4llttttlt I 1 I ~ ~ I I I ~ le 11111I II tell If I ~ 1SNIIISIIII CORE AXIALNOOK CORK AXIALNOOK r NCASURC0 TIR RCSRORSR + NCARURCOTIRRCteoNRR 0 OALOVLAfCo TIR RRRSORSR 0 OALOIRATCOTIR RNtoRSR

~ 001TROL 100 eoemoN ~ CONTROL 100 ROQllON 108

FIGURE 3.2.19 SUSQUEHANNA SES UNIT 1 CYCLE 2 AVERAGE AXIALTIP RESPONSE COMPARISON 0.200 GWD/MTU CYCLE EXPOSURE 9O

,80 70 Bo I-zD 6 o ~ + i ~ o eo " "+:""""'""

...................."':."""""'"" i.... ... .

"'.""@'"O')

..........+ ...Q...

LQ zz 40 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Q ~ ~ ~ ~

CO 0 CL 3O 20 10 0

0 1 2 3 4 5 6 7 8 9 10 1112131415161718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE 109

FIGURE 3.2.20 SUSQUEHANNA SES UNIT 1 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 0.200 GWD/MTU CYCI E EXPOSURE 61 59 57 -2.65 -2.51 4.50 -3.40 55 53 51

-1.86 t+

49 47 45 05 -2.27

++ .20 43 I 41 3 -0 33 -1;65 -0 .82 27 39 37

+

I 35 I 33 2.88 -0.30 90 58 2.6 .92 6.5 6 31 29 27 25 20 54 2.4 -3.31 15 4.96 4.4 9 23 I I 21 19 17 -0 .13 41 3.7 0.4 3

+ -0 .77 -0.20 15 I  !

I 1 3 I 11 9 -2.74 97 06 7

I I

3 Y i I 00 02 0< 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 X

Diff = [(Calc - Mess)/Core Avg TIP Response] X 100%

- 110-

FIGURE 3.2.21 SUSQUEHANNA SES UNIT 1 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 0.200 GWD/MTU CYCLE EXPOSURE I$OIETOA LOCATIOII 45,$ $ $ IONTOA LOCATIOK 4' 4

O ~

$ o 0 0 0 X 0$

ss

+

. + 4OO ~ 0 N

X X

ss

~

$ ~ $ og Q cs S ~ ~ ~ T S S 'N 11 ls N N 'N N IT I ~ IS SS Sl TS SS SI 0 I S ~ ~ ~ T ~ NIIISNNNNITNI~ Ssslsssssl CORE AXIALIIOOK CORK AXIALNOOK

+ NSASSSKOTlt SSSSTNNS + NSASVASO Tlt SlstONSS lit 0 CAIOVLAIKS AsstONSO lit 0 0 LCVLATTO ASStONSC

~ CONTNOL NOO SOSITlDN 0 CONTNOL 000 tOSITNN IKONITOR LOCAtlON OO,$ $ LIOIIIIORLOCAllOII $ $,$ $

~0

+)

l ss

~

) 4

-k~-4-0-$ ~rr-p

'0

~ gX 0

SS a~.~s

+ ss 0

0 TS Q

~ ~ S S ~ ~ 0 T ~ ~ ISIINNNNNITISISSSSISSSSSI S I S S ~ ~ 'T ~ ~ N ~ I IS IS N N N ll ls ll SO Sl SS Sl SI CORE AXIALIIOOE COIIK AZALIIOOE 0 ICKASVAKOTlt NKStOINK 0 llKASVAKOTN NKStONSK o CALOVIAIKS litNsstoNSK O OAIOVLATKO Tlt asstONSC

~ OONTNOL SOO SOSITlCW ~ OONTNOL 000 tONllON

FIGURE 3.2.22 SUSQUEHANNA SES UNIT 1 CYCLE 2 AVERAGE AXIALTIP RESPONSE COMPARISON 2.587 GWD/MTU CYCLE EXPOSURE 9O Bo 70 eo I-z Q C

LLI 60 i

zz Q

+ 4o V

CO Q F 0

so 20 10 0

0 1 2 S 4 6 6 7 8 9 10 11 1213 14 'I616 17 18 192021222324 CORE AXIAL NODE

+ MEASURED TIP RESPONSE Q CALCULATEDTIP RESPONSE 112-

FIGURE 3.2.23 SUSQUEHANNA SES UNIT 1 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 2.687 GWD/MTU CYCLE EXPOSURE 59 57 -3.57 -1.22 5.71 -1.33 55 53 51 I 49 -2.61 1.5 3 0.5 0.6 .75 47 I J

45 43 41 0.0 -2. 92 1.17 -1.05 -0.33 39 37 35 33 2.32 0.8 -2. 39 -3.13 -2.40 0.8 5.4 31 29 27 25- -1.85 -0

++

23 21 48 02 -2.01 5.9

++

19 17 05 2.6 73 4.38 -1.41 -0.16 -1.79 15 13 11 9

+ -3.15 0.7 -6.86 7

5-3 1

00 02040608 10 12 14 16 18

+

20 22 24 26 2830 32 i343638 40 42 44464850 52 54565860 X

Diff = [(Calc - Meas)/Core Avg TIP Response] X 100%

113

FIGURE 3.2.24 SUSQUEHANNA SES UNIT 1 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 2.687 GWD]MTU CYCLE EXPOSUR~

~ IONITOII LOCATION aL$ $ MONITOR LOCATION 4$ ,$ $

$ o 0

0 0 +f0 oool ae Ig se 4..$ . o .L..

g

'TI T ee 0 I o O.

C ae I

I I I I

~ s ~ ~ ~ r 0 I~ 11 ls>>>>>>>> lr>>>> te tl st aa el ~ 1 s ~ ~ ~ ~ r e ~ >> n ts Ie 1I Ia Ie tl ta>> sa st ss ae aI CORK AICALNOOK CORK AXQL NOOK

+ vsaava sot>> saeaooee 4 Neaeoeeo tie ete JONet 4 oalNAAtso tie sseaoees 0 oslolsstso tto eaaaoees

~ oooteol soo roeNNN ~ ooNININ. Noo roeltNN

$ IONITOR LOCATION $ $ .$ $

ee

+

0 + 0 lr 4 ~ 4s'Z 0

te se d X

0 Q u $ I~

~ ee

~ I t e ~ ~ s r ~ lail>>la>>>>>>lt>>leaetlsssesl ~ I t a ~ ~ r ~ ~ >> 11 ts >> 11 Ia Ie lr Ia COIIK AICAL NOOK COR! AXIALNOOK

+ vsAsvsto tie sseeoos$ + vtssvsso lie llssroees o OAIOIAAISOtie Steeooas o OAIOVLStto tte Itaaroees

~ ooeteoL aoo IoeotoN ~ OoetaoL Noo eo alttoN

- 114

FIGURE 3.2.25 SUSQUEHANNA SES UNIT 1 CYCLE 2 AVERAGE AXIALTIP RESPONSE COMPARISON 4.638 GWD/MTU CYCLE EXPOSURE 90 80 70 BO g)

I-zD 0 0 q) 0 C 0 4 ~ b $ + +

50 ill +

zz 40 0

Co Q.

0

~ 30 +

20 10 0

0 1 2 3 4 5 6 7 8 9 101112131415161718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 115

FIGURE 3.2.26 SUSQUEHANNA SES UNIT 1 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 4.638 GWD/MTU CYCLE EXPOSURE I

61 59 57 -3.56 -2.44 5.36 -2.44 55 53 51 49 2.30 .73 2.3 3053 3.68 -1.23 47 45 43 ++ -4.00 41 39 ++ -2.36 -1.59 0;51 37 35

-1. 01 -0.86 -2.30

++ J 33 31 0.83 20 28

++ 6.5 6 29 27 ++

25 23 21 I~

-2.94 .73 -2.05 -0.21 2.24

++

49 I

17 -0.46 0.8 -0.36 0.3 0,3 2 0.4 5 15 13 2.10 32 2.10 -4.56 000204060810 12 14 16 18 2022 24 26283032343638 4042 444648505254565860 X

Diff = [(Calc - Meas)/Core Avg TIP Response] X 100%

116

FIGURE 3.2.27 SUSQUEHANNA SES UNIT 1 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 4.638, GWD/MTU CYCLE EXPOSURE MONITOR LOCATION dd,dd MONITOR LOCAllON Sd,dd

~0

~0 I

ss ~0 J Sogf ~ ~ ~

0 s

gg ss 0 '+00 Of O 0 0 0 4 0 0 0 0 0

0 SL Q ss

~ SS ~0 0

0 ts I t S S S 0 T ~ ls 11 TS lt 11 IS lt If IS lt SS SI SS SS Sl 0 \ ~ ~ S ~ ~ T t 0 10 11 lt It 11 lt 10 lf lt lt tt Sl St SS SI CORE AXIALNOOK o NSASOSSO TI ~ PNSPONSS + NSASONSO TIP NSSPONSS 0 OALOOLATSO TIP NSSPONSS 0 OALOSIAftoTIP NSSPONSS

~ CONTSOL NON POSITION ~ CONTSOL 100 POSITION MONITOR LOCATION sd,dd MONITOR LOCATION dd,dd

~S ~S

~S Ts 5 +PL oOOO odogoo 0 .t.

LS t 0 ~0 0

f o g S~ ....0 ~~ (-0.,0. 0 0

0 ss 10 J I I I S 0 0 0 1 ~ ~ lt II.ISISll lsl~ fl ltlttttlSSSSSI CORK AXIALHOOK 0 I 0 t ~ S ~ T 0 ~ lt II flit ll I ltlf I

~ ~ I~ Sttl ltlttl CORK AXIALHOOK P NSASOSSO TIP IISSPONSS 0 I SASSSSO llP SSSPONSS o OuauuSSO TIP NSSPONSS o OALOOIATSO TIP NSSPONSS

~ OONTS OL %00 POSNION ~ OONTNOL SOO PotlflON 117

FIGURE 3.2.28 SUSQUEHANNA SES UNIT 1 CYCLE 3 AVERAGE AXIALTI P RESPONSE COMPARISON

. 0.178 GWD/MTU CYCLE EXPOSURE 90 80 70 BO " +" "0 I- 0 zD I 50 .

+

LLI 0 zz 40 0 M

0 30 20 10 0

0 1 2 3 4 5 6 7 8 9 10'11 121314151817 18182021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 118-

FIGURE 3.2.29 SUSQUEHANNA SES UNIT,1 CYCLE 3 RADIAL TIP RESPONSE COMPARISONS 0.178 GWD/MTU CYCLE EXPOSURE 61 59 57 0.54 -0.14 5.79 -0.83 55 53 51 49 -1.48 -2.08 -2.40 0.6 0.38 2,8 47 45 43 41 -2.68 1.55 -0.77 4.98 3.38 -3.25 0.08 39 37 35 33 I

28 -0.32 6.4 -3.09

+ +

++

01 31 29 27 25 0. 62 3.8 -4.95 36 -3.72 2.0 23 21 19 17 60 -1.62 2.2 3.4 29 -0.51 15 13 11 9 -2.37 25 99 7

5 I I 0002040608 10 12 14 16 18 20 22 24 26 2830323436384042 44464850 52 545658 60 X

Diff = [(Calc - Measj/Core Avg TIP Response] X 100%

119

FIGURE 3.2.30 SUSQUEHANNA SES UNIT 1 CYCLE 3 INDIVIDUALTIP RESPONSE COMPARISONS 0.178 GWD/MTU CYCL'E EXPOSURE MONITOR LOCAllON dlLdd MONITOR LOCATION aa,dd

~0 00

,4T 0

o +

4 4 4

00 pt4 C

Z tg 00 L 0 IN 4 Q I

~- $ 00 0

0

~0

~ 00 L Se I.

~0 I 0 ~ 0 ~ 0 ~ I~ tt ts'I ~ lt 10 10 tt 10 10 00 01 As es 04 I 0 0 ~ 0 ~ 7 ~ lett tale tt tete tt I~ 10 te Sl 1~ 00 SI CORS AIDALNODS CORS.AXIAL NODS 4 vaA00000 tti 00000000 + vsA00000 tta 0 0000000 4 OAAOUIAtaott ~ Naasovea 4 OAIOUAAtaa tto 0 0000000

~ Oovta OA 000 000ttvvt ~ CoataoL 000 000IIION MONITOR LOCATION ~$

00

~0 4...0 .4 t 4 4 o ja 4 0 00 4 doo 0 +" -4"4-d-9' To4

'4 5 00 4

+ 0 O te

) te le 4

~~

I~

I 0 0 0 ~ ~ 7 ~ ~ I~ 11 ts 10 It Ia 10 It I ~ I~ test as sect I 0 ~ 0 0 ~ 0 ~ ~ lett tats v tete It 10100001tltltl CORS AXIALNODS CORS AXIALNODS 4 vaAaaaaa Tla assaovaa 4 VSASUaaa lat IICS00000 o OAIOUIAtao tt0 aaaroaea o OAIOUIAtaa Its Itaaaoaaa

~ OONIN OL 000 eosttvvl ~ OONtllOl 000 000ttION 120

FIGURE 3.2.31 SUSQUEHANNA SES UNIT 1 CYCLE 3 AVERAGE AXIALTIP RESPONSE COMPARISON 2.228 GWD/MTU CYCLE EXPOSURE 9O eo 70

+

Bo

.: 0" 0 I- ~ 0 0 +

D z 0 50 LU zz o o o o 0

+ + q + 0 40 Co 0

0 3O '

20 10 0

0 1 2 3 4 5 6 7 8 9 10 1112131415161718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 121

FIGURE 3.2.32 SUSQUEHANNA SES UNIT 1 CYCLE 3 RADIALTIP RESPONSE COMPARISONS 2.228 GWD/MTU CYCLE EXPOSURE S9 S7 0.84 1.58 6.61 -1.24 55 I

53 51 49 -0.94 -3.19 -0 54 3.0 2.5 3.89 47 45 43 41 -2.97 0.85 52 3.2 -0.10 0.83 39 37 35 33 1.8 -2:14 -0. 06 4.4 -1.02 31 29 27 25 0.61 3.3 -2.87 -3.98 -1.19 1.99 23 21 19 17 1.6 -5.43 -2.78 0.6 -6.42 03 1S 13 11 9 .44 -4.29 56 -3.65 7

5 3

1 00 02 04 06 08 '10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 X

Diff = [(Calc - Mess)/Core Avg TIP Response] X 100'/o 122

FIGURE 3.2.33 SUSQUEHANNA SES UNIT 1 CYCLE 3 INDIVIDUALTIP RESPONSE COMPARISONS 2.228 GWD/MTU CYCLE EXPOSURE MONITOR LOCATION SS.SS MONITOR LOCATION 44SS ee I

I 1

ee I

I I

Te 1 Te - a, j +S I 0 I

~4 o

K l

ee Vl I 0 t 4 0 4 4 4 I g 0 5>> 'i- r 0 O

X

$ >> T r -"'-

0 f 0 I

0 24 -$ L r -

I I I I I

I I

>> 1 Ie I I 7 t I I 4

~ 1 ~ ~ L 4 ~ T ~ 4 NI 11 le I~ If 14 14 If 'I ~ Ie 24 N 22 ee ef 1 4 ~ ~ 4 ~ 2 ~ 4 Ie 11 14 I~ >> le le IT le te N 21 2224 N CO RK AXIALNODS COllh AXIALNOOC veAevheo lit 0 eetohee + NNAOVNeellt heetONeh 0 CAICOLATeo lit Neetoheo 0 OALOVLATee Tlt heetoheh

~ CONlNOL 400 t041TtON ~ CONTNOL NOO tOemON MONITOR LOCATION OO,SS MONIIOR LOCATION SWISS

~~I

~4 Te o o

+ 0 0 oo4 S 0 0

I 0 0 ~

~4 o 4 0 0 ee ooe C 0

C 0 Q>>

~ ee Ie+ le

~ 1 4 4 ~ ~ ~ T ~ ~ 10 11 Ie Ie 12 I~ I~ \f Ie 14 24 11 12 22 ef 1 4 ~ ~ 4 4 T ~ ~ I~ 11 1$ Ie 12 Ie 'I ~ Il 12 I~ 24 21 22 2$ $4 COhK AXIALNODS C Oh8 AXIALNO 1 IITAOVheOTI~ heetOI>>e 0 CALOVLATNOlitheetoheh

+ tfeAOVNTO ltt heetohea o OALCVLAT!OTlt heetoheh 0 ooNTNCL 100 toemoN ~ OOKfhOL NOO toe ITION

- 123

FIGURE 3.2.34 SUSQUEHANNA SES UNIT 2 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 0.387 GWO/MTV CYCLE EXPOSURE 180 180 140

+

+ ..o...4......~...............:,....e.

co 120 0 I- 0 R

100

'll 0

K ~

+

X 80 CO L,

ea 20 0

0 1 2 3 4 5 e 7 8 8 1011 1213141516 1718182021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE

- 124

FIGURE 3.2.36 SUSQUEHANNA SES UNIT 2 CYCLE 1 RADIAL TIP RESPONSE COMPARISONS 0.387 GWD/MTU CYCLE EXPOSURE 61 59 57 0.09 3.47 3.60 1.23 55 53 51 49 -1.06 4.76 .41 1.89 1.28 47 45 43 41 0.6 0.5 0.88 -5.24 -1.56 -2.47 0.62 39 37 35 33 4.5 -0.55 -0.89 -1.98 -1.80 -2.08 91 I

31 29 27 25 29 -0.40 -1.33 -5.22 0.7 2.2 4.3 23 21 19-17 20 22 -0.07 -0.76 -3.68 -0 ~ 05 2.9 -2.93 15 13 11 9 10 -0 09 -1.85 13 -3.21 7

5 3

1 00 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 X

Diff = [(Calo - Mess)/Core Avg TIP Response] X 100%

125

I FIGURE 3.2.36 SUSQUEHANNA SES UNIT 2 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 0.387 GWD/MTU.CYCLE EXPOSURE MONITOR LOCATTON 4OPS I ~

I 0

I 0

4 ~ T T 0

te I I 0 1 t ~ ~ ~ ~ 1 0 ~ Te 11 Te N Te N le lT I ~ N te tl tete ee ~ 1 t ~ ~ ~ 0 1 0 ~ lt 11NNNleleTT Nl~ tetlttleel CORE AXIALNODE CORE AXIALNODE

+ vtAell tee TN 0 teoooee 4 TNAe vote TN ttetovee o DAIOOIATteTN atetooea 0 OAIOIAATCDTV et etovec

~ covleoc too toelTTDO ~ oovleoe eoo toeITlo4I MONITOR LOCAllON 40,5$ MONITOR LOCAllON 02,$ 3 t ++aJ 0 o 4 0

4' +

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ee J J. ~ ee

~ ~ t e 0 ~ e 1 I

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CORE AXIAI.NODE I

NllltltllleNlTIANtetlttteel e 1 t e ~ ~ ~ 1 ~ ~ Ie 11 Ie lt N CORE AXIALNODE 14 le 11 11 It te 11 tt te el I

t VtAeette Tlt 1 tttOINe + NEAOVACO Tle teetOINC 0 OALDVCATKD M tttKHNO o OAIOVCATXD~AettOTNC

~ OOVTDCN. DOD tetITloe ~ oovTDDL too toeITlov

- 126-

FIGURE 3.2.37 SUSQUEHANNA SES UNIT 2 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 5.249 GWD/MTU CYCLE EXPOSURE 180 1eo 140 120 I-z 100 0

+ 0 LLI 0 Q

z + + 0 0

80 + g}

Co CL Q

eo 40 20 0

0 1 2 3 4 6 e 7 8 9 10 11 1213141616 1718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE 127

FIGURE 3.2.38 SUSQUEHANNA SES UNIT 2 CYCLE 'f RADIAL TIP RESPONSE COMPARfSONS 5.24S GWD/MTU CYCLE EXPOSURE 61 59 57 i i

-2.63 2.68 1.22 -0.92 55 53 51 49 -4.77 0.7 5.03 -0.95 3.39 .73 47 45 43 41 -3.29 0.97

+ 8.3 -2.87 0 .29 -2.33

+3 07 39 37 35 ++ I I

-0 .17 -0.65 33 31 ++ 1.2 8 8 35

~ 0.8 .17 29 27 ++

25 23

-3.71 6.32 43 -2.98

++ 0. 82 19 I

17 .27 5.9 -0.18 -1.78 1.2 2.8 -4.87 15 I I

13 11 4.9 -0.67 -6.45 -2. 98 -4.53 7

5 3

v I I I I I 0020 6 08 10 12 14 16 18 20 22 24 26 28303234 363840 4244 46485052 54 56 58 60 Diff = [(Calc - Mess)/Core Avg TlP Response] X 100%

128-

FIGURE 3.2;39 SUSQUEHANNA SES UNIT 2 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 6.249 GWD/MTU CYCLE EXPOSURE MONITOR LOCAllON 4IL$$ MOIQTQR LOCAllON 4$ ,$ $

100 NS NS

+ ~ $ $ 4+ 4 Itt C ,'0I 0 at C 100 I

~ 0. 0 T g. 00 .S..

0 0- C..

4

~ 00 I I 4 O I J 00 I 1 .

I I

o 1 I 1 0 ~ 0 ~ ~ 0 ~ ~ a 11 lt a 10 aa Tf 'a lt 00 tl tt St a 0 1 t ~ 0 0 ~ T ~ ~ <<TITSISSINNITNISSSSISSS ~ SI CORK A)QAL NODC COAQ A)QAL NODS

+ NSASUSSD Tlo SSSOONK + NSASUSSO TIS IISSSONSS 4 AkISSAATSDTI~ NSSSONK 0 OAIOULAItoTlo Ntttot 0 0

~ CONT SOL SOO SotfAON ~ CONTSOL NOD SOSITION MONITOR LOCAllON 40,$ $ MONITOR LOCAllON $$ ,$ $

100

<<0 100 0

o40 0 4 110

~ 4 f L

Z 0 0 0 0

100 0

w'$-'f' 4

0

~0 ~ +~I ~0 00 0

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00 00 I I I 0 I 0 0 ~ ~ ~ 1 0 ~ I~ ll lt lt 10 <<a 11 1010 Sttlttttti ~ 0 0 ~ 0 ~ T ~ ~ 'l0 11 10 lt 10 a % IT N 10 St tl St tt Sl COR4 ATQAL NODE CORK NQAL NODS t SNASUSCD TID SCSSONK 4 NCASUSCD Tlt DCSSONK o OAIOUIATCDTID IICSDONK 4 OAIJUIATCDTID IICSJONK

~ DONT SOL 000 SOSNION ~ OOWIOL NOO SONllON 129

FIGURE 3.2.40 SUSQUEHANNA SES UNIT 2 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 12.050 GVYD/MTU CYCLE EXPOSURE 180 180 140 + o 0

o o o t 0

0 120 J l-zD 100 Ill zz Q

+

80 j CO IL Bo 40 0

+

20 0 1 2 3 4 5 B 7 8 S 10 11 1213 14 151B 17 18 192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE o CALCULATED TIP RESPONSE j.30

FIGURE 3.2.4'f SUSQUEHANNA SES UNIT 2 CYCLE 1 RADIAL TIP RESPONSE COMPARISONS 12.050 GWD/MTU'CYCLE EXPOSURE I I 61 59 57 -3.23 -1.66 0.34 -2.11 55 53 51 49 -4.02 -3.01 0.89 -1.52 9.21 -0.71 47 45 43 41 -4.34 50 2.9 64 6.2 3.30 -0.06 39 37 35 33 -0.18 68 4.8 0.4 6.6 -1.88 1.46 31 29 27 25 1.45 77 3.42 -0.81 .51 23 21 19 17 -7 -1.16 -0.93 7.4 -2.00 15 13 11 9 2.8 99 0.7 2.4 -2.08 7

5 3

1 y

000204060810 12 14 16 18 20 22 24 2628303234363840 42 444648505254565860 X

Diff = [(Cele - Meas)/Core Avg TIP Response] X 100%

131

FIGURE 3.2.42 SUSQUEHANNA SES UNIT 2 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 12.050 GWD/MTU CYCLE EXPOSURE leOHITOR LOCATIOH 4$ ,$ $

'14$

+

4 o 4 oo,>

4$

~ ts 4 S

<<f gee) L D'.

so 4 I to I7 e

+

0

~ eo ~ oo

+

d I ~ ~ $ ~ 7 ~ Te Tl T$ 1$ 14 le Te 11 T<<4 te $1 tt te $ 4 ~ 1 t ~ 4 e ~ t e ~ Tohltlt 14 1$ 1<< lt 4 1$ te tl tt te $4 CORK AXIALHOOK CORK ANALHOOK 4 Neoeoeeo lit Neetoese + NeAe seto lit TNetoose 0 OlIOOLNTOD he etetONse o OAtowATto h t acetoeee

~ OONTOOL NOO toelhON ~ OONTNOL NOO to¹hDN MOHITOR LOCAnOH 40,$ $ IIOHITOllLOCATIOH $ $ ,$ $

14$ 1$ $

+

4 4 0 T 0

3 0 4

0 4 0

~o ~o o

)

~ ss 4$

0 0

+

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~ I t $ ~ e o 1 ~ ~ lehT<<1$ 14&41144$ <<tlttt$ $4 ~ I t ~ $ ~ 7 ~ IS h It ll 14 1$ 1<< IT 4 Is t<<11 tt $ $ $ 4 CORK AXIALHOOK CORK AXIALHOOK

+ lltA$I¹$ 0 Tl ~ eeseoese + NeAtsetO lit Ntstoest o OAllelttes llt Nestohoe o OAA004ATtO Tlt Neetoete

~ OONT$ 0L NOO te¹llON ~ ODNTNOL NDO $ 0¹hON 132-

FIGURE 3.2.43 SUSQUEHANNA SES UNIT 1 CYCLE 1 SIMULATE-E VS GE PROCESS COMPUTER CORE AVERAGE AXIALPOWER DISTRIBUTION 1.5 w10

+

O

)

I SIMULATEE 0.5 GE Process Corn~uter 0.0 2 3 4 5 6 7 8 9 10 11 12 BOTTOM TOP Cycle Average Exposure = 1.490 GWD/MTU Core Power Level = 99.6% of rated Total Core Flow = 100 Mlbm/hr Reactor Pressure = 1005 psla Core Inlet Subcooling = 23.6 Btu/Ibm 133

FIGURE 3.2A4 SUSQUEHANNA SES UNIT t CYCLE 2 SIMULATE-E VS POWERPLEX CORE AVERAGE AXIALPOWER DISTRIBUTION 1.5 CL 1.0 LIJ O

CL

)

LIJ I

LIJ SIMULATE-E 0.5 POWER PLEX 0.0 3 5- 7 9 11 13 15 17 19 21 23 25 BOTTOM TOP Cycle Average Exposure = 2.587 GWD/MTU Core Power Level = 99.9% of rated Total Core Flow = 95.8 Mlbm/hr Reactor Pressure = 1000 psla Core Inlet Subcooling = 24.7 Btu/Ibm l34

FIGURE 3.2.45 SUSQUEHANNA SES UNIT 1 CYCLE 3 SIMULATE-E VS POWERPLEX CORE AVERAGE AXIALPOWER DISTRIBUTION 1.5 1.0 LLI O

CL /

LIJ

/

/ SIMULATEE CL 0.5

/ POWER PLEX 0.0 3 5 7 9 11 13 15 17 19 21 23 25 BOTTOM TOP Cycle Average Exposure = 0.178 GWD/MTU Core Power Level = 100% of rated Total Core Flow = 96.9 Mlbm/hr Reactor Pressure = 1002 psia Core Inlet Subcooling = 24.4 Btu/Ibm 135

FIGURE 3.2.46 SUSQUEHANNA SES UNIT 2 CYCLE 2 SIMULATE-E VS POWERPLEX CORE AVERAGE AXIAL POWER DISTRIBUTION 1.5 1.0 LLJ

/

O. /

CL LJ

/

I t SIMULATE-E LJJ 0.5 I POWER PLEX I

0.0 1 3 5 7 9 11 13 15 17 19 21 23 25 BOTTOM TOP Cycle Average Exposure = 0.583 GWD/MTU Core Power Level = 100% of rated Total Core Flow = 96.2 Mlbm/hr Reactor Pressure = 1000 psia Core Inlet Subcooling = 24.4 Btu/Ibm l36-

FIGURE 3.2.47 SUSQUEHANNA SES UNIT 1 CYCLE I SIMULATE-E VS GE PROCESS BUNDLE FLOWS AT 1.490 GWD/NITU 'O 0.120

0. 122 0.002 0.119 0.120 PROC COMP 0.121 0.122 0.002 0.002 SIMULATE-E 0.117 0.119 0.118 DIFFERENCE 0.121 0.120 0.122 0.004 0.001 0.004 Units are Mlbrn/hr 0.119 0.131 0.131 0.118 0.121 0:135 0.133 0.121 Average Difference: 0.001 0.002 0.004 0.002 .0.003 Standard Deviation: 0.002 0.119 0.132 0.130 0.118 0.117 0.121 0.133 0.135 0.119 0.121 0.002 0.001 0.005 -. 0.001 0.004 0.118 0.119 0.117 0.116 0.116 0.118 0.120 0.121 0.119 0.119 0.117 0.120 0.002 0.002 0.002 0.003 0.001 0.002 0.116 0.117 0.118 0.117 0.116 0.117 0.117 0.119 0.118 0.120 0.117 0.119 0.118 0.120 0.003 0.001 0.002 0.0 0.003 0.001 0.003 0.116 0.130 0.131 0.118 0.117 0.130 0.130 0.118 0.119 0.135 0.132 0.120 . 0.117 0.134 0.131 0.119 0.003 0.005 0.001 0.002 0.0 0.004 0.001 0.001 0.115 0.130 0.129 0.115 0.116 0.130 0.130 0.118 0.120 0.119 0.131 0.134 0.117 0.119 0.131 0.134 0.118 0.119 0.004 0.001 0.005 0.002 0.003 0. 001 0.004 0.0 0.001 0.113 0.115 0.114 0.114 0.115 0.117 0.119 0.119 . 0.122 0.128 0.116 0.117 0.115 0.117 0.115 0.119 0.118 0.119 0.122 0.127 0.003 0.002 0.001 0.003 0.0 0.002 -0.001 0.0 0.0 -0.001 0.111 0.112 0.112 0.112 0.113 0.114 0.119 0.121 0.131 0.136 0.070 0.115 0.114 0.116 0.114 0.114 0.116 0.119 0.121 0.126 0.135 0.068 0.004 0.002 0.004 0.002 0.001 0.002 0.0 0.0 -0.005 -0.001 -0.002 0.114 0.118 0.117 0.115 0.116 0.119 0.124 0.125 0.135 0.068 0.115 0.118 0.119 0.116 0.117 0.120 0.124 0.127 0.134 0.068 0.001 0.0 0.002 0.001 0.001 0.001 0.0 0.002 -0.001 0.0 0.121 0.124 0.125 0.123 0.123 0.126 0.135 0.138 0.069 . 0.070 0.120 0.123 0.123 0.121 0.122 0.126 0.130 0.136 0.068 0.068

-0.001 -0.001 -0.002 -0.002 -0.001 0.0 -0.005 -0.002 -0.001 -0.002 0.128 0.129 0.130 0.130 0.130 0.132 0.141 0.069 0.129 0.129 0.130 0.130 0.131 0.134 0.140 0.068 0.001 0.0 0.0 0.0 0.001 0.002 0.001 -0.001 0.069 0.069 0.069 0.069 0.069 0.070 0.070 0.068 0.068 0.068 0.068 0.068 0. 068 0.068

-0.001 -0.001 -0.001 -0.001 -0.001 -0.002 -0.002 137

FIGURE 3.2.48 SUSQUEHANNA SES-UNIT 'I CYCLE 3 SIMULATE-E VS POWERPLEX BUNDLE FLOWS AT 0.178 GWD/MTU 0.114 0.116 0.002 0.114 0.119 POWER PLEX 0.115 0. 123 0.001 0.004 SIMULATE-E 0.118 0.117 0.119 DIFFERENCE 0.121 0.122 0.120 0.003 0.005 0.001 Units are Mlbm/hr 0.116 0.119 0.118 0.119 0.117 0.122 0.118 0.122 0.001 0.003 0.0 0.003 Average Difference: 0.001 Standard Deviation: 0.002 0.120 0.118 0.121 0.117 0.120 0.123 . 0.121 0.123 0.121 0.120 0.003 0.003 0.002 0.004 0.0 0.121 0.131 0.130 0.122 0.121 0.130 0.121 0.133 0.130 0.124 0.120 0.132 0.0. 0.002 0.0 0.002 -0.001 0.002 0.124 0.129 0.131 0.119 0.122 0.127 0.128 0.126 0.133 0.133 0.121 0.124 0.130 0.128 0.002 0.004 0.002 0.002 0.002 0.003 0.0 0.119 0.121 0.120 0.119 0.118 0.121 0.119 0.119 0.119 0.123 0.119 0.122 0.118 0.122 0.118 0.122 0.0 0.002 0.003 '0.001 0.0 0.001 -0.001 0.003 0.119 0.115 0.118 0.114 0.117 0.115 0.119 0.116 0.120 0.121 0.119 0.121 0.118 0.120 0.119 0.121 0.119 0.119 0.002 0.004 0.003 0.004 0.003 0.004 0.002 0.003 -0.001 0.117 0.120 .'0.120 0.117 0.117 0.120 0.120 0.120 0.121 0.124 0.117 0.123 0.119 0.119 0.116 0.122 0.120 0.122 0.121 0.124 0.0 0.003 -0.001 0.002 -0.001 0.002 0.0 0.002 0.0 0.0 0.119 0.117 0.120 0.114 0.118 0.118 0.122 0.118 0.125 0.130 0.064 0.121 0.121 0.122 0.117 0.120 0.121 0.124 0.122 0.127 0.132 0.063 0.002 0.004 0.002 0.003 0.002 0.003 0.002 0.004 0:002 0.002 -0.001 0.117 0.119 0.118 0.118 0.119 0.121 0.123 0.126 0.129 0.064 0.116 0.120 0.117 0.120 0.118 0.122 0.122 0.127 0.131 0.063

-0.001 0.001 -0.001 0.002 -0.001 0.001 -0.001 0.001 0.002 -0.001 0.120 0.117 0.120 0.117 0.121 0.120 0.128 0.131 0.064 0.065 0.121 0.118 0.121 0.119 0.122 0.122 0.129 0.132 0.063 0.064 0.001 0.001 0.001 0.002 0.001 0.002 0.001 0.001 -0.001 -0.001 0.123 0.126 0.124 0.127 0.126 0.128 0.136 0.064 0.123 0.125 0.124 0.126 0.126 0.128 0.133 0.063 0.0 -0.001 0.0 0.001 0.0 0.0 -0.003 -0.001 0.064 0.064 0.064 0.064 0.064 0.064 0.065 0.062 0.062 0.062 0.062 0.062 0.063 0.064

-0.002 -0.002 -0.002 -0.002 -0.002 -0.001 -0.001 138

FIGURE 3.2.49 SUSQUEHANNA SES UNIT 2 CYCLE 2 SIMULATE-E VS POWERPLEX BUNDLE FLOWS AT 0.583 GWD/MTU 0.118 0.116

-0.002 0.119 0.119 POWER PLEX 0.117 0.123

-0.002 0.004 SIMULATE-E 0.116 0.123 0.121 DIFFERENCE 0.119 0.122 0.124 0.003 -0.001 0.003 Units are Mlbm/hr 0.119 0.117 0.121 0.117 0.118 0.121 0.119 0.121

-0.001 0.004 -0.003 0.004 Average Difference: 0.001 Standard Deviation: 0.003 0.117 0.123 0.120 0.121 0.118 0.121 0.121 0.122 0.120 0.121 0.004 -0.002 0.002 -0.001 0.003 0.124 0.130 0.133 0.120 0.122 0.126 0.121 0.133 0.132 0.122 0.121 0.129

-0.003 0.003 -0.001 0.002 -0.001 0.003 0.120 0.132 0.130 0.122 0.118 0.128 0.124-0.122 0.132 0.132 0.120 0.121 0.128 0.128 0.002 0.0 0.002 -0.002 0.003 0.0 0.004 0.119 0.117 0.121 0.115 0.118 0.115 0.118 0.113 0.119 0.120 0.119 0.119 0.117 0.119 0.117 0.118 0.0 0.003 -0.002 0.004 -0.001 0.004 -0.001 0.005 0.113 0.118 0.114 0.116 0.112 0.116 0.112 0.116 0.113 0.118 0.117 0.118 0.116 0.117 0.115 0.117 0.116 0.118 0.005 -0.001 0.004 0.0 0.005 -0.001 0.005 0.0 0.005 0.116 0.115 0.118 0.111 0.115 0.113 0.117 0.113 0.118 0.121 0.116 0.120 0.118 0.116 0.115 0.118 0.117 0.118 0.119 0. 125 0.0 0.005 0.0 0.005 0.0 0.005 0.0 0.005 0.001 0. 004 0.112 0.118 0.115 0.115 0.111 0.117 0.115 0.119 0.121 0.132 0.064 0.116 0.118 0.119 0.115 0.116 0.117 0.120 0.119 0.125 0.130 0.063 0.004 0.0 0.004 0.0 0.005 0.0 0.005 0.0 0.004 -0.002 -0. 001,;

0.117 0.113 0.117 0.112 0.117 0.115 0.120 0.122 0.130 0.064 0'.116 0.117 0.115 0.117 0.116 0.119 0.120 0.126 0.129 0.063

-0.001 0.004 -0.002 0.005 -0.001 0.004 0.0 0.004 -0.001 -0.001 0.116 0.117 0.116 0.118 0.117 0.122 0.124 0.133 0.064 0.065 0.118 0.116 0.119 0.117 0.121 0.121 0.127 0.131 0.063 0.064 0.002 -0.001 0.003 -0.001 0.004 -0.001 0.003 -0.002 0.001 0.001 0.125 0.123 0.126 0,125 0.128 0.129 0.137 0.065 0.122 0.124 0.123 0.126 0.125 0.130 0.134 0.064 0.003 0.001 -0.003 0.001 0.003 0.001 -0.003 0.001 0.064 0.064 0.064 0.064 0.064 0.065 0.065 0.062 0.062 0.062 0.062 0.063 0.063 0.064

-0.002 -0.002 -0.002 -0.002 -0.001 -0.002 -0.001 139-

3.3 uad Cities Unit 1 Cycles 1 and 2 Benchmark An additional demonstration of the SIMULATE-E calculational accuracy was performed by comparing SIMULATE-E results to measurements from the Quad Cities Unit 1 Cycles 1 and 2 cores. After the end of Cycles 1 and 2, gamma scan measurements of selected fuel bundles were taken. This provides an excellent measurement of the power distribution averaged over the last two to three months of each cycle's operation. This technique for measuring the power distribution is not prone to the types of errors that are typical of TIP measurements. Reported accuracy of the gamma scan measurements, combining measurement uncertainty and measurement method bias, is approximately 3%

(Reference 12), whereas TIP uncertainty for reload cores is typically 5.1%

(Reference 25).

A significant number of cold critical tests was performed during Cycle 1. The available cold data include both in-sequence and local criticals. In-sequence criticals are typical of normal reactor startups with withdrawn control rods uniformly dispersed throughout the core. Local criticals involve withdrawal of 'a few control rods (usually from two to four) in a localized area of the core producing very peaked neutron flux gradients.

In addition to the gamma scan and cold critical data, hot reactivity statepoint and TIP measurement data are also presented in this section.

The Quad Cities Unit 1 core (Figure 3.3.1) is slightly smaller than the Susquehanna SES cores (Figure 3.2.8), containing 724 versus 764 fuel assemblies, and its rated core thermal power is approximately 25% less than that of the Susquehanna SES units. For the Quad Cities initial cycle, the entire core consisted of General Electric Company (GE) 7x7 fuel with a low gadolinia loading. This contrasts the Susquehanna SES cores where a relatively high gadolinia loading was present in the 8x8 fuel. The Quad Cities reload fuel for Cycle 2 consisted of only 23 GE 7x7 fuel assemblies, 36 GE Sx8 fuel assemblies, and five mixed oxide test assemblies. The GE reload fuel contained a small gadolinia loading.

- 140

3.3.1 Hot Critical Core Reactivity Com arisons The purpose for benchmarking the hot critical core K-effective for Quad Cities is to determine if any major differences in results and trends exist between Susquehanna SES and Quad Cities. Because the Quad Cities core contains mainly 7x7 fuel and lower gadolinia content, the benchmark provides a good contrast to the Susquehanna SES benchmark and a test of the steady state methodology.

Figure 3.3.2 shows the Quad Cities Unit 1 Cycles 1 and 2 calculated hot critical core K-effectives with those of Susquehanna SES. Although Quad Cities results show more variation, a linearly increasing trend is present.

This trend is consistent with the Susquehanna SES results and supports the exposure dependency of the SIMULATE-E calculated critical core K-effective.

No bowl-shaped trends are evident in the Quad Cities results. This trend is attributed to the lower gadolinia loading in Quad Cities versus Susquehanna SES. The large variation in K-effective is possibly due to the inclusion of data that does not meet the steady state criteria defined in Section 3.2 for Susquehanna SES data. The measured core operating parameters used 'as input to SIMULATE-E are contained in Reference 27. As evident from Figure 3.3.2, the Susquehanna SES data essentially forms a continuous line of data as a result of a very detailed SIMULATE-E depletion calculations; however, the Quad Cities K-effectives are quite sparse.

3.3.2 Cold Critical Core Reactivity Com arisons The benchmark of the SIMULATE-E calculated cold critical K-effective to the Quad Cities Unit 1 Cycle 1 cold xenon-free in-sequence and local criticals provides qualification of PPaL's cold methodology and models to perform shutdown margin calculations. Comparisons to the large local critical database (22 local criticals) test PPGL's calculation of rod worths in large local flux gradient locations that are typical of shutdown margin calculations. PPGL's approach in benchmarking to the Quad Cities cold criticals is to compare the calculated in-sequence critical K-effectives (ll total) to the local critical K-effectives. Table 3.3.1 presents the Quad Cities Unit. 1 Cycle 1 calculated cold critical .K-effectives which have been corrected for reactor period. Comparing local to in-sequence critical results 141

demonstrates the capability to.calculate the same core K-effective for critical conditions with both; peaked and uniform neutron flux distributions.

The local critical K-effectives are compared to the average of the in-sequence critical K-effectives at the same exposure. Table 3.3.2 shows the results of the comparisons. The average difference between the K-effectives is 0.00007 and the standard deviation equals 0.00064. Both of these values are well within the uncertainty in predicting the Susquehanna SES cold critical core K-effective (i.e., standard deviation equal to 0.00137). This demonstrates that no bias exists between in-sequence and local critical calculations.

An additional test of PPGL's methods involves demonstrating that the same observed bias between hot and cold critical core K-effective for Susquehanna SES also exists between hot and cold critical core K-effective for Quad Cities. Figure 3.3.3 shows the hot and cold critical core K-effectives.

Despite the variation in and lack of hot critical core K-effective data, the difference between the calculated hot and cold K-effectives is similar to that of the Susquehanna SES data.

3.3.3 Traversing In-core Probe Data Com arisons Although the primary reason for the development of the Quad Cities model is to perform the gamma scan comparisons, some TIP data is available for comparison from Reference 27 and 28. This includes 15 TIP sets from Cycle 1 and 13 TIP sets from Cycle 2. A TIP set contains 24 axial measurements taken at each of the 41 radial TIP locations. Radial TIP detector locations are shown in Figure 3.3.1.

The SIMULATE-E code was used to calculate the TIP responses for each of the 28 TIP sets. As described in the Susquehanna SES TIP response comparison section, the SIMULATE-E calculated TIP responses are renormalized so that the core average calculated TIP response is the same as the core average measured TIP response. The average RMS of the differences between the SIMULATE-E calculated and measured TIP responses for each TIP response comparison is calculated as described in Section 3.2.3. Results from the nodal and radial comparisons are given in Table 3.3.3. Comparisons have been reported for all TIP sets with the exception of Case 16. Correct measured TIP response data is

- 142-

unavailable for this case. Although several of the other TIP sets were taken before the core had time to reach an equilibrium xenon distribution due to control rod position, power or flow changes, they have been included in the comparison.

Figures 3.3.4 through 3.3.15 present. representative TIP response comparisons for Cycles 1 and 2. For two exposure points in each cycle, core average axial, radial, and four individual TIP response comparisons are included. The individual TIP response comparisons in the figures, were selected along a line from the core periphery to the core center as shown in Figure 3.3.1. The same four TIP locations are always shown.

3.3.4 Gamma Scan Comparisons At the end of Cycles 1 and 2 gamma scan measurements were taken. The available Cycle 1 data (Reference 29) consist of axial peak to bundle average La-140 activities for 31 fuel bundles, individual axial traces from two fuel bundles, and the axial trace from the average of the 31 individual traces.

Use of this data is primarily limited to benchmarking the axial peaking factor. The Cycl'e 2 data (Reference 12) are much more extensive. A total of 89 fuel bundles were scanned. Of these, 71 were located in one octant of the core, providing measurement data for most of the fuel bundles in that octant.

The remaining 18 fuel bundles were chosen in other octants to check for asymmetries. Seventy-three of the bundles were scanned at 12 axial locations at approximately twelve-inch intervals. The remaining 16 bundles were scanned at 24 axial locations at approximately six inch intervals. The reported measured activity was corrected to correspond to activity just after shutdown.

The practical accuracy of the reported data including measurement uncertainty and measurement method bias is approximately 3% (Reference 12, Section 4.3).

As previously discussed in Section 2.3, the gamma scan data itself is a measure of La-140 gamma activity. During reactor operation, La-140 is produced both as a fission product and by Ba<<140 decay. Since the half-life of Ba-140 is approximately 13 days and that of La-140 is approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, the distribution of the Ba-140 and La-140 concentrations will be 143-

representative of the core power distribution integrated over the last two to three months of reactor operation. After shutdown, the only source of La-140 is from decay of Ba-140. Because the half-life of La-140 is short with respect to Ba-140, after about ten days the decay rate of La-140 is controlled by the decay of Ba-140. Therefore, the relative measured La-140 activities are compared to the relative calculated Ba-140 concentrations, and the La-140 concentration does not need to be calculated.

The SIMULATE-E code was used to calculate the nodal Ba-140 concentrations at the end of both cycles. At the end of Cycle 1, the peak to average Ba-140 concentration was calculated for each of the 31 fuel bundles. Of these, 17 were uncontrolled and 14 were partially controlled. The calculated and measured peak to average data for the uncontrolled and controlled fuel bundles is shown in Tables 3.3.4 and 3.3.5, respectively. The average difference for all 31 fuel bundles is 1.2% with a standard deviation of 2.1%. These results demonstrate excellent agreement to the measured axial peaking factor.

Three axial traces from Cycle 1 are also available from Reference 29. The measured and calculated La-140 activities for each trace are normalized to 1.0 prior to the comparison. Figure 3.3.16 shows the comparison for the uncontrolled bundle, and Figure 3.3.17 shows the comparison for the controlled bundle. Figure 3.3.18 shows the comparison for the axial 31 bundle average La-140 activities. The measured data for these plots were only available in graphical form from Reference 29. Therefore, no statistics are computed from the comparisons, but the figures demonstrate the ability of SIMULATE-E to calculate axial power shape.

More extensive gamma scan measurements were taken at the end of Cycle 2. The data supplied in Reference 12 allow for radial, nodal, peak to average, and bundle (axial) comparisons. For the radial and nodal comparisons, the peripheral bundles have been eliminated. These bundles are low in power and, consequently, of no concern from a thermal limits perspective. For the nodal comparisons the top and bottom six inches have also been eliminated. These nodes are low in power and are, consequently, of little importance from a safety standpoint. The mixed oxide bundles have also been eliminated from the nodal and radial comparisons since they are atypical of Susquehanna SES reload fuel.

<< 144-

Prior to making any comparison, the measured and calculated data were normalized such that the core average relative activity was 1.0. However, for the calculated data only the nodes for which there were measured data were used in the normalization process.

The comparisons are based on the mean difference between calculated and measured normalized La-140 activities. This difference is calculated as:

i

e. = c.-m.

i. i where

c. = the normalized calculated La-140 activity,

~

i

m. = the normalized measured La-140 activity.

The subscript i denotes either the average activity for the bundle for the radial comparisons or the nodal activity for the nodal comparison. The standard deviations for the comparisons are calculated as:

N a(s) =

P (e. i e)

X 100 N-1 M where M = the average of the normalized measured data for the comparison

= 1.0 for all comparisons due to normalization, e = the average difference between the measured and calculated normalized La-140 activities

= 0.0 for all comparisons due to normalization, N = number of La-140 activities for the comparison.

The radial comparisons were obtained by averaging the nodal La-140 activities for each bundle. The results from the comparisons are shown in Figure 3.3.19.

The standard deviation of 1.82% reported on the figure was calculated for those bundles included in the octant shown in the figure. If the additional 11 bundles from the other octants are included in the comparison, the standard deviation becomes 1.92%. Based on the comparisons, no significant deviation 145

in the radial power shape is apparent indicating .SIMULATE-E will provide an accurate assessment of the Critical Power Ratio. The standard deviation from the nodal comparisons is 5.45%. Assuming a 3.0% measurement uncertainty, the calculational standard deviation is 4.55%.

The SIMULATE-E calculated peak to average La-140 activity was compared to the measured data. The percent difference for each assembly is calculated as:

e.'

c.-m, i

i m x 100 where c = the calculated peak to average La-140 activity for fuel bundle i,

m. = the measured peak to average La-140 activity for fuel bundle i.

The results of the comparisons are shown in Table 3.3.6. The average difference is -0.2% with a standard deviation of 1.5%. These comparisons included all assemblies and accounted for all axial nodes. The results indicate excellent agreement for the axial peaking factor and are consistent with the Cycle 1 results.

The results from the individual bundle comparisons are shown in Table 3.3.7.

These comparisons are also reported for every bundle and included all axial nodes. For each bundle, the average difference between the calculated measured nodal activities is calculated as:

K e

Z'>>,.

n K where e

k,n

= the difference between the measured and calculated normalized nodal La-140 activities for bundle n, and axial node k, K = number of axial nodes in the bundle for which measurements were taken.

- 146

The standard deviation for each fuel bundle is:

K g(e -e) 100 0

n K-1 Figure 3.3.20 shows the fuel assembly with the best axial agreement (Bundle CX0662). Although this particular bundle is. located on the core periphery, it exhibits excellent agreement for all axial locations. The worst comparison is shown in Figure 3.3.21 (Bundle CX0399). The calculated average difference of 12.2% is mostly due to differences in the top and bottom nodes.'owever, the calculated La-140 activity in the center section of the bundle still agrees well with the measured data. Figures 3.3.22 through 3.3.27 show example comparisons which are more typical of the rest of the assemblies. Most of the calculated difference is due to nodal comparisons at the top and bottom of the core. Different top and bottom albedos could have eliminated much of this error. As discussed in Section 3.1, the albedos, which were developed as a result of the Susquehanna SES model normalization, were also used in the Quad Cities calculations. It is expected that due to different core and fuel designs for Quad Cities, the top and bottom albedos would differ from the Susquehanna SES values. Although the Susquehanna SES albedos were utilized in the Quad Cities calculations, the SIMULATE-E model provides an accurate calculation of the power distribution, This supports the use of the SIMULATE-E model to predict power distributions for fuel designs other than those in the normalization database.

- 147

TABLE 3.3.1 QUAD CITIES UNIT 1 CYCLE 1 CALCULATED COLD XENON-FREE 4 CORE CRITICAL K-EFFECTIVES Core Average Core Reactor Number of Calculated Exposure Temperature Period Controlled Local (L) or Core (GWD/RZU) (DEG F) (sec) Notches 4

In-secpxence (I) ff K-e ective 0.0 . 152 180 6390 I 0.99314 0.0 160 60 8400 L 0.99289 0.0 159 75 8404 L 0.99207 0.0 159 160 6344 I 0.99288 0.0 161 150 6324 I 0.99284 0.0 160 50 8404 L 0.99192 0.0 159 32 8402 L 0.99209 0.0 158 78 8402 L 0.99281 0.0 158 90 8402 L 0.99324 0.0 159 41 8392 L 0.99257 0.0 157 65 8402 L 0.99270 0.0 157 125 6336 I 0.99304 0.0 160 65 6318 I 0.99278 0.0 159 332 8392 L 0.99366 0.0 159 245 8392 L 0.99353 0.0 160 38 8392 L 0.99247 0.0 158 42 8416 L 0.99247 0.0 159 39 8392 L 0.99252 0.0 158 39 8402 L 0.99268 0.0 158 169 8394 L 0.99082 0.0 155 120 6412 I 0.99171 2.866 163 300 6698 I 0.99590 3.748 70 .43.7 8438 L 0.99847 3.748 75 47.5 8428 L 0.99840 3.748 77 280 8430 L 0.99760 3.748 108 54 8426 L 0.99718 3.748 120 300 7118'378 I 0.99818 3.748 120 157 L 0.99790 3.748 '125 140 8378 L 0.99812 3.748 178 181 6830 I 0.99730 4.938 120 45 6936 I 0.99829 6.911 182 100 8394 L 1.00026 6.911 179 300 6514 I 1.00041 148

TABLE 3 3 2 QUAD CITIES UNIT 1 CYCLE 1 IN-SEQUENCE VERSUS LOC2LL CRITICAL COMPARISON ~

Core Average Control Rods Core Reactor Exposure Withdrawn Temperature Period In-secpxence Minus (GWD/MTU) and Position (DEG P) (sec) Local K-ef fective f 0.0 38,11 9 48 42,11 8 48 160 60 -0.00016 0.0 38,11 48 38,15 44 159 75 0.00066 0.0 46,19 8 48 46,23 8 44 160 50 0.00081 0.0 46,19 8 48 50,19 8 46 159 32 0.00064 0.0 50,23 8 48 50,19 8 46 158 78 -0.00008'.0 46,23 8 48 50,23 9 46 158 90. -0.00051 26,31 8 48 0.0 26,35 8 48 159 41 0.00016 30,31 8 08 0.0 18,11 8 48 22,11 8 46 157 65 0.00003 26,27 8 48 0.0 26,31 8 48 159 332 -0.00093 30,31 8 08 26,27 8 48 0.0 30,27 8 48 159 245 -0.00080 30,31 9 08 26,23 8 48 0.0 26,27 8 48 160 38 0.00026 30,27 8 08 22,39 8 48 0.0 22,35 9 24 158 42 0.00026 26,35 8 08 26,39 8 48 0.0 22,39 8 48 159 39 0.00021 26,35 8 08 149

TABLE 3.3.2 (conti.nued)

QUAD CITIES UNIT 1 CYCLE 1 IN-SEQUENCE VERSUS LOCAL CRITICAL COMPARISON Core Average Control Rods Core Reactor Exposure Withdrawn Temperature Period In-secgxence Minus (GWD/MTU) and Position (DEG P) (sec.) Local K-effecti.ve 42,39 9 48 0.0 42,35 9 38 158 39 0.00005 38,35 8 08 38,39 8 48 0.0 38,35 9 48 158 169 0.00191 34,35 9 06

3. 748 26,11 9 38 22,11 8 20 70 43.7 "0.00073 3.748 26,11 9 48 22,11 9 20 75 47.5 -0.00066 3.748 22,11 6 48 22,15 6 18 77 280 0.00014 3.748 50,27 6 48 50,23 9 22 108 54 0.00056 26,15 9 48 3.748 22,11 8.48 120 157 -0.00016 18,15 6 22 14,27 9 48 3.748 10,23 I 48 125 140 -0.00038 14,19 8 22 22,15 I 48 6.911 22,11 6 48 182 100 0.00015 26,11 8 06 Average = 0.00007 Standard Deviation = 0.00064 150

TABLE 3 3 3

SUMMARY

OP QUAD CITIES UNIT 1 CYCLES 1 AND 2 TIP RESPONSE COMPARISONS Core Average Nodal Radial Case Exposure RMS RMS Number Date (CWO/MTU) (a) (*)

Cycle 1 1 6/29/72 0. 27,2 9.43" 5.43 2 8/30/72 0.712 8.85 5.67 3 9/11/72 0.882 8.26 5.80 4 11/01/72 1.470 10.43 5.72 5 12/26/72 2.239 8.38 5.61 6 3/08/73 3.190 9.09 .5. 79 7 5/16/73 3.836 9.61 6.12 8 6/06/73 4.074 9.87 6.46 9 7/19/73 4.737 9.84 5.91 10 8/30/73 5.301 10.72 5.87 11 11/01/73 6.031 13.84 5.36 12 12/11/73 6.558 11.11 5.80 13 12/29/73 6.807 9.23 5.63 14 2/13/74 7.396 11.42 4.97 15 3/05/74 7.659 11.72 5.58 16* 3/26/74 7.980 Cycle,l Average 10.12 5.71 Cycle 2 17 7/26/74 7.303 12.55 4.38 18 8/15/74 7.532 10.18 4.85 19 9/12/74 7.964 8.89 4.25 20 10/23/74 8.423 10.30 4.63 21 11/18/74 8.789 8.08 4.66 22 12/11/74 9.141 7.80 4.80 23 4/03/75 10.173 8.07 4.94 24 6/19/75 11.238 7.92 4.42 25 8/08/75 11.935 8.79 5.00 26 10/20/75 12.896 8.16 5.29 27 11/13/75 13.198 8.55 4.76 28 12/19/75 13.611 11.65 4.54 29 12/31/75 13.741 12.73 5.03 Cycle 2 Average 9.51 4.73 Combined Average 9.84 5.26

• Correct measured TIP response data is unavailable.

151

TABLE 3 3 4 QUAD CITIES UNIT 1 EOC 1 GAMMA SCAN COMPARISONS UNCONTROLLED BUNDLES Measured Calculated Core Peak to Average Peak to Average Difference Location La-140 Activi La-140 Activity (*)

39,58 1'. 271 1.270 -0.1 41,58 1.212 1.239 2.2 41,56 1.224 1.218 -0.5 17,48 1. 287. 1.289 0.2 55,42 1.185 1.244 5.0 57,42 1.191 1.260 5.8 57,40 1.245 1.257 1.0 07,34 1.176 1.214 3.2 09,32 1.148 1.194 4.0 07,26 1.170 1.227 4.9 09,24 1.186 1.234 4.0 31,26 1.354 1.329 -1.8 47,18 1.250 1.259 0.7 23,10 1.178 1.178 0.0 25,08 1.239 1.224 1~2 31,10 1.172 1.182 0.9 33,08 1.221 1.235 1.1 Avexage Difference = 1.7%

Standard Deviation = 2.3%

- 152-

TABLE 3.3.5 QUAD CITIES UNIT 1 EOC 1 GAMMA SCAN COMPARISONS CONTROLLED BUNDLES Measured Calculated Core Peak to Average Peak to Average Difference Location La-140 Activity La-140 Activity (*)

39,56 1.282 1.284 0.2 17,50 1.609 1.631 1.4

'15,48 1.280 1.307 2.1 55,40 1.269 1.279 0.8 09,34 1.418 1.394 -1.7 07i32 1.322 1.332 0.7 09,26 1.366 1.398 2.3 07,24 1.231 1.256 2.0 49,18 1.602 1.625 1.4 47,16 1.283 1.305 1.7'1.2 25,10 1.358 1.342 23,08 1.251 1.247 -0.3 33,10 1.385 1.350 -2. 5 31,08 1.369 1.373 0.3 Average Difference = 0.5%

Standard Deviation = 1.5%

153-

TABLE 3.3.6 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISONS PEAK TO AVERAGE LA-140 ACI'IVITIES BUNDLE LOCATION PEAK TO AVERAGE DIFFEfKNCE ID (X Y) MEASURED CALC <r.)

CX0214 (33,34) 1.1923 1.2004 0.7 GEB159 (31,32) 1.1379 1.1365 -0.1 CX0575 (31,34) 1.1937 1.1881 -0.5 CX0588 (33,32) 1.1842 1.1973 1.1 CX0420 ( 7,32) 1.2420 1.2620 1.6 CX0052 (15,32) 1.1990 1.1966 -0.2 CX0287 (23,34) 1.1871 1.1856 -0.1 CX0378 (17,42) 1.2089 1.2120 0.3 GEH023 ( 9,40) 1.2008 1.1887 -1.0 CX0150 ( 7,42) 1.3108 1.2733 -2.9 CX0440 ( 9,42) 1.2586 1.2200 -3.1 CX0351 ( 7,40) 1.2714 1.2417 2~3 CX0453 (23,32) 1.1975 1.1945 -0.2 CX0723 (17,40) 1.2028 1.1979 -0.4 CX0015 <15,42) 1.1894 1.1776 -1.0 CX0316 (15,40) 1.2173 1.2001 -1;4 CX0498 (25,34) 1.1892 1.1869 -0.2 CX0044 ( 7,34) 1.2445 1.2457 0.1 CX0327 ( 9,34) 1.2285 1.2303 0.1 CX0106 ( 9,32) 1.2341 1.2511 1.4 CX0165 (25,32) 1.1932 1.1991 0.5 CX0306 (15.34) 1.2006 1.1936 '>>0.6 CX0660 (17,34) 1.1856 1.1745 -0.9 CX0310 (27,34) 1.1866 1.1903 0.3 CX0523 ( 3,36) 1.3468 1.3509 0.3 CX0093 (13,40) 1.2187 1.1839 -2.9 CX0297 (23,38) 1.1982 1.2008 0.2 CX0611 ( 3,40) 1.3924 1.3674 -1.8 CX0024 (15, 46) 1.1872 1.2023 1.3 CX0225 (21.32) 1.1864 1.1768 -0.8 CX0617 ( 9,46) 1.3086 1.2762 -2.5 CX0231 (15.38) 1.2181 1.1889 -2.4 CX0585 (19.36) 1.2103 1.1764 -2.8 CX0631 ( 5,38) 1.3263 1.3051 -1.6 CX0186 (19.42) 1.2039 1.2106 0.5 CX0332 (11,44) 1.2169 1.2093 -0.6 CX0161 (19,38) 1.2237 1.2032 -1.7 CX0100 (13,46) 1.2225 1.2170 -0.4 GEH022 ( 9,36) 1.1944 1.1921 -0.2 GEH029 (13,44) 1.1660 1.1759 0.9 CX0281 (21,36) 1.1844 1.1688 -1.3 CX0399 ( 9,38) 1.2548 1.1978 -4.5 CX0396 (11,40) 1.2054 1.1825 -1.9 CX0198 ( 5,36) 1.2875 1.2798 -0.6 CX0393 (11,36) 1.2028 1.1983 -0.4 GEH002, (13.36) 1.1651 1.1789 1.2 GEB132 (17,36) 1.1589 1.1546 -0.4 GEB160 (31,30) 1.1353 1.1360 0.1

- 154-

3.3.6 (continued)

QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN (X)%'ARISONS PEAK TO AVERAGE LA-140 ACTIVITIES BUNDLE LOCATION PEAK TO AVERAGE DIFFERENCE ID (X Y) MEASURED CAID (X)

GEB161 (29,32) 1.1335 1.1364 0.3 GEB158 (29,30) 1.1302 1.1354 0.5 CX0494 ( 7,48) 1.3490 1.3528 0.3 CX0490 ( 5,46) 1.3495 1.3452 -0.3 CX0174 ( 7.46) 1.3333 1.3344 0.1 CX0683 ( 1,32) 1.3311 1.3556 1.8 CX0520 ( 3,32) 1.3116 1.3402 2.2 CX0394 (11,32) 1.2263 1.2398 1.1 CX0137 ( 5,32) 1.2785 1.2884 0.8 CX0482 (27,32) 1.1802 1.1830 0.2 CX0717 (19,32) 1.1563 1.1638 0.6 CX0682 ( 1,40) 1.4128 1.4320 1.4 GEH008 (13,48) 1.2107 1.2051 -0.5 GEB123 (17,44) 1.1664 1.1740 0.6 GEB149 (21,40) 1.1861 1.1857 -0.0 CX0719 ( 9,50) 1.3328 1.3279 -0.4 CX0672 (15,36) 1.1984 1.1817 -1.4 CX0362 (13.34) 1.2144 1.2070 -0.6 GEB105 (25,36) 1.1673 1.1497 -1.5 CX0546 ( 9,52) 1.3590 1.3700 0.8 CX0553 ( 5,44) 1.3583 1.3118 -3.4 CX0662 ( 3,42) 1.3827 1.3960 1.0 CX0643 ( 1,34) 1.3357 1.3612 1.9 CX0397 (13,38) 1.2103 1.1840 2~2 CX0286 ( 9,48) 1.3266 1.2986 -2.1 CX0191 (11,50) 1.2933 1.2899 -0.3 CX0057 (13,32) 1.2233 1.2261 0.2 CX0124 (17,10) 1.2137 1.2194 0' CX0414 (47,38) 1.1473 1.1833 3.1 CX0412 (37,48) 1.1852 1.1909 0.5 CX0384 (23.14) 1.1946 1.1909 -0.3 CX0318 (13,24) 1.1558 1.1842 2.5 CX0401 (47,24) 1.2168 1.1825 -2.8 CX0398 (23,48) 1.1941 1.1882 -0.5 CX0359 (37,14) 1.1634 1.1884 2.1 CX0711 (49,10) 1.2911 1.3174 2.0 CX0096 ( 9,18) 1.2007 1.2434 3.6 CX0622 (47, 6) 1.3119 1.3566 3.4 CX0445 (41,18) 1.2010 1.2007 -0.0 GEB162 ( 5,48) 1.3518 1.3518 0.0 CX0162 (17,32) 1.1777 1.1698 -0.7 AVERAGE DIFFERENCE: -0. 2r.

STANDARD DEVIATION: 1.5r.

155

TABLE 3.3.7 UAD CITIES UNIT 1 EOC 2 INDIVIDUALBUNDLE GAMMA SCAN COMPARISIONS STANDARD BUNDLE LOCATION AVERAGE DEVIATION ID (X Y) DIFFERENCE (X)

CX0 546 ( 9,52) 0.007 4.34 CX0719 ( 9.50) ,0.016 4.95 CX0191 (11,50) 0.017 5.37 GEB162 ( 5,48) 0.027 4.28 CX0494 ( 7,48) 0.014 5.41 CX0286 ( 9,48) 0.032 6.11 GEH008 (13,48) -0.020 7.46 CX0398 (23,48) 0.045 6.32 CX0412 (37,48) -0.007 5.74 CX0490 ( 5,46) 0.038 5.11 CX0174 ( 7,46) 0.021 5.93 CX0617 ( 9,48) 0.033 6.30 CX0100 (13,46) 0.029 7.55 CX0024 (15,46) 0.013 5.73 CX0553 ( 5,44) 0.036 5.87 CX0332 (11,44) 0.016 6.65 GEH029 (13,44) -0.026 7.42 GEB123 (17.44) 0.027 8.73 CX0662 ( 3,42) 0.003 3.98 CX0150 ( 7,42) 0.005 . 6.68 CX0440 ( 9,42) 0.020 7.49 CX0015 (15,42) 0.005 7.17 CX0378 (17,42) -0.012 7.92 CX0186 (19,42) 0.007 6.55 CX0682 ( 1.40) 0.012 3.83 CX0611 ( 3,40) 0.005 5.22 CX0351 ( 7,40) -0.003 6.50 GEH023 ( 9,40) -0.054 8.18 CX0396 (11,40) -0.031 6.75 CX0093 (13,40) -0.010 7.11 CX0316 (15,40) 0.007 7.05 CX0723 (17,40) -0.016 6.92 GEB149 (21,40) -0.006 7.71 CX0631 ( 5,38) -0.001 6.50 CX0399 ( 9,38) -0.012 12.17 CX0397 (13,38) 0.008 6.17 CX0231 (15,38) -0.009 6.33 CX0161 (19.38) -0.003 5.79 CX0297 (23.38) 0.007 5.78 CX0414 (47,38) -0.035 5.41 CX0523 ( 3,36) 0.012 4.65 CX0198 ( 5,36) 0.008 5.96 GEH022 ( 9,36) -0.074 7.27 CX0393 (11.36) -0.007 6.42 GEH002 (13,36) -0.045 7.70 CX0672 (15,36) -0 '04 6.59 GEB132 (17,36) -0.033 7.80 CX0585 '(19,36) 0.007 6.64 CX0281 (21,36) 0.009 6.33 156

TABLE 3.3.7 (continued)

UAD CITIES UNIT 1 EOC 2 INDIVIDUAl,BUNDLE GAMMA SCAN COMPARISIONS STANDARD BUNDLE LOCATION AVERAGE DEVIATION ID (X Y) DIFFERENCE (X)

GEB105 (25,36) -0.023 7.50 CX0 643 ( 1,34) 0.043 4.24 CX0 044 ( 7.34) -0.002 5.77 CX0327 ( 9,34) -0.008 6.29 CX0362 (13,34) -0.011 6.14 CX0306 (15,34) -0.022 5.97 CX0660 (17,34) 0.005 .5.86 CX0287 (23,34) -0.005 7.41 CX0498 (25,34) 0.005 6.13 CX0310 (27,34) 0.005 6.16 CX0575 (31,34) 0.003 6.80 CX0214 (33,34) -0.004 5.92 CX0683 ( 1,32) 0.045 4.37 CX0520 ( 3,32) 0.016 4.90 CX0137 ( 5,32) 0.012 5.15 CX0420 '( 7.32) 0.014 5.52 CX0106 ( 9,32) .-0.002 4.99 CX0394 (11.32) -0.016 5.88 CX0057 (13,32) -0.007 5.69 CX0052 (15,32) -0.012 6.43 CX0162 (17,32) -0.015 5.66 CX0717 (19,32) -0.003 5.82 CX0225 (21,32) -0.002 5.31 CX0453 (23,32) -0.009 6.47 CX0165 (25,32) -0.006 5.40 CX0482 (27.32) -0.007 6.56 GEB161 (29,32) -0.034 8.22 GEB159 (31,32) -0.026 8.46 CX0588 (33,32) 0.015 7.37 GEB158 (29,30) -0.038 7.89 GEB160 (31.30) -0.018 7.77 CX0318 (13,24) -0.023 5.56 CX0401 (47,24) -0.001 6.78 CX0096 ( 9.18) -0.020 4.91 CX0445 (41,18) 0.018 7.70 CX0384 (23,14) 0.004 6.72 CX0359 (37,14) 0.034 4.88 CX0124 (17,10) 0.013 5.62 CX0711 (49,10) 0.030 5.15 CX0622 (47, 6) 0.033 3.67 157

FIGURE 3.3.1 QUAD CITIES UNIT I CORE TlP LOCATIONS LINE OF TIP SYMMETRY 59 57 55 53 51 49 47 45 43 41 39 37 35 33 31 29 27 25 23 21 0002 04 0608 10 12 14 16 18 20 22 24 26 28 30 32 34 3638 40 42 44 46 4850 52 54 56 58 60 X

Control Rod Location Location For Individual TIP Response Comparisons

~ Traversing In-core Probe Location

FIGURE 3.3.2 SIMLUATE-E HOT CRITICAL CORE K-EFFECTIVE VS CORE AVERAGE EXPOSURE 1.01 e

e 1.00 s s 'I s s I s e

e

~

~ ~ ~

e l

I-O UJ

~ e e

@ e e

~~o0'eee r I

UJ 0 99

~:::::: ~

~:

~

"0 Legend fC U1C1 HOT.

O " .'" "".'" 0 O '. U2C1 HOT U1C2'OT 0.98 U2C2 HOT o U1C3 HOT

~ QC1C1 HOT

~ QC1C2 HOT l r l 0.97-0 1 2 3 4 5' 7 8 10'1 12 13 14 15 CORE AVERAGE EXPOSURE (GWD/MTU)

FIGURE 3.3.3 QUAD CITIES UNIT 1 CYCLE 1 SIMULATE-E HOT AND COLD CRITICAL CORE K-EFFECTIVES 1.01 I f I

0 0

0 * ~

1.00 N...:.......',..... x .

I IJJ 0 ~

~

~

~ ~

~

I I-o o IJJ LL U ......................:..0...:...O.

hC I

0.99-

o  :' ':..:.

0 LLI lC ......:.... 0 :. Legend 0

O C I

o QC1C1 HOT x QC1C1 COLD 0

0.88

""'.97 I

t 0 1 2 3 4 6 6 7 10 CORE AVERAGE EXPOSURE (GWD/MTU)

FIGURE 3.3.4 QUAD CITIES UNIT 'I'CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 2-239 GWD/MTU CORE AVERAGE EXPOSURE 160 160 140 120 I-K + 0 0 0 + 0 D ~

+

100 ..0..

tQ R +

e so U

Co CL eo 40 20 0

0 1 2 8 4 6 6 7 S 9 10 1112181415161718192021222824 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE

>> 161

FIGURE 3.3.5 QUAD CITIES UNIT 1 CYCLE 1 RADIALTIP RESPONSE COMPARISONS 2.239 GWD/MTU CORE AVERAGE EXPOSURE 61 59 57 -3.21 -0.23 -2.86 55 53 51 49 -4.34 6.34 -0.58 -10.36 6.36 2.26 47 45 43 ++

41 39

-8. 33

++ .61 13 14.

+++

37 35 ++ ++

33 31 29 2.7

++ 5.4

++ I 1.3 27 25 2.68 7.8 -6. 66 -1. 17 -5.21 23 21 19 I

++ 4 85 -10.58 17 15 13 1.68

++ .59 87

-3.66 .30 -0. 80 -2.41 00 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 5456586 0 X

Diff = [(Calc - Meas)/Core Avg Tlp Response] X 100%

- 162

FIGURE 3.3.6 QUAD CITIES UNIT 1 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 2.239 GWD/MTU GORE AVERAGE EXPOSURE IlOHIIOR LOCAllOH dd,dd MOHIlOR LOCATIOH 44,dd

'll

~I ttt 0 0

a..e +

IOO too rr s'~> r O ~0

';KZ Vo s

~0 a

0 I a 0 O.

~o .i. C 0 0

to I

I l I o t ~ ~ ~ t ~ 1 ~ ~ tt Tl 1t I ~ lt lo a TT lo lt to ol oo to RI ~ ~ ~ ~ 4 t ~ T t ~ 10 m to It tt tt Io IT lt a to a tt to tt CORK ALGAL HOOK CORK AXIALHOOK 0 Votatttomr RttrONOR 0 taAWMO mr RurORu 0 AAICIRATRDmr Rot rotot o INLCOIATtoTI ~ Ruroutt

~ CoNTRoL Roo roamoN ~ coNTRCL Roo rottmoN MONITOR LOCA1IOH 40,5$ IIOIKTOR LOCATIOH dK,SO IOO too 0

tto I o Oo 0 0 t tto 4 0 g 0 L 0 0 5 ao 0

o 5 eo 0

r ~o 0

a oo TO T~

~ I ~ o ~ t ~ T ~ ~ lt 11 lt 'lo V 1t 10 IT lt lt tt tt tt tt tt ~ I 0 o ~ t ~ T ~ ~ 10 lt lo I ~ 1I 1 ~ 'I ~ 1T 1t I ~ tt tl tt oo tt CORK AXIALKOOK CORK AXIALHOOK

+ vtAtutto mr RttroNot 0 VRAtutto mr Rttrouot o AALOutATCOTlr Rttrouot o CALOutATROmr Rttroeu

~ CCNTRCL Roo rovmoN v ooNTRDL Roo tovmoN

- 163

RGURE 3.3.7 QUAD CITIES UNIT 1 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 7.398 GWD/MTU CORE AVERAGE EXPOSURE 180 180 140 120 ~ ~ ~ ~ ~ ~ l ~ ~ ~ ~ ~ ~ ~

I-R D q o0d'e

~: g 4 4 g ~ 0 +

+

100 0 y +

LU + 0 +

K 80 CO Q.

60 40 20 0

0 1 2 3 4 5 6 7 8 9 10 1112131415161718192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 164

FIGURE 3.3.8 QUAD CITIES UNIT 'f CYCLE t RADIAL TIP RESPONSE COMPARISONS 7.396 GWD/MTU CORE AVERAGE EXPOSURE 61 59 57 55 ++ -5.23 -1.15 -1.54 53 51

++ ++

49 47 45

-7. 61 5.14

++ .69 5;2

+++

43 41 -4 99 5.9 2 -2.00 94 2.90 6.0 39 37 35 ++ -3.54 -3.52 33 31 3.4

++ 5.3

+

2.6 29 27 ++ ++

25 23

. 21 4.89 2.08

++ 9.4

++ 69 19 17 2.5 -7 -3.59 20 05 -4.54 52 15 13 11 9 .10 -10 5 -0.89 0.05 7

5 3

1 000204060810 12 14 16 18 20 22 24 26 28303234363840 42 4446 485052 54565860 X

Diff = [(Calo - Meas)/Core Avg TIP Response] X 100%

- 165

FIGURE 3.3.S QUAD CITIES UNIT 1 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 7.396 GWD/MTU CORE AVERAGE EXPOSURE MONITOR LOCATION NL,$$ MONITOR LOCATTON 4$ ,$$

I to 100 7

X L

Z 00$ S 8404 0 I

~ Ito 6 0+ gQ 0 o 0' 0 I 0

0 t 01 o ~ b 0 T T'

~ ~ t 0 0 ~ ~ T ~ ~ TC ll ltltltlololtlolttotlttCCCI ~ ~ t ~ 0 ~ ~ T ~ 0 TO ll 0 10 TI It IO lf It 10 to tl tt tt tt CORK AXIALNODE CORK ANALNODE 0 oclcuoto litoctfootc + WAOWCO Tlt Oottouoo 0 OALOQAICO M IICCPOOCC 0 OALOOIATCOTlt IlottooK

~ COOTCOL 000 tooOIOO ~ COCTOOL 000 totmoo MONITOR LOCATION 41,$ $ MONTTOR LOCATION $ $ .$ $

100

~ 10 Tto L ooT+ oooI 0 oo4 5 Ioo t

I 0

eo I

I I

I

~ 1 ~ 0 ~ 0 ~ T ~ ~ 10 11 lt It 11 It Io 11 It 10 tt tl Ct tt 01 ~ ~ 0 0 ~ C ~ T 0 ~ IO Il It It IT I~ IO It I0 I ~ Ct ~I tl tt M CORE AXIALNODE COIIK AXIALNOOK

~ tlcAcuccDllt hcotolltc + OCACIOND TIt IICttoCCC 0 OALOIAATCO ~ I Cttolltt

~ ooutcoL coo toolcloN O OALOOLATCOnt aaetoaao

~ OOOTCOL 000 000OIOO 166

FIGURE 3.3.10 QUAD CITIES UNIT 1 CYCLE 2

'VERAGE AXIALTIP RESPONSE COMPARISON T.532 GWD/MTU CORE AVERAGE EXPOSURE 180 160 140 0

0 0

120

'I-R 0 100 ILI

+

z 6 80 CO 0

CL 0 BO 0 40 20 0 1 2 3 4 6 B 7 8 9 10 11 1213 14 1616 17 18192021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE

- 167

FIGURE 3.3.11 QUAD CITIES UNIT 1 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 7.532 GWD/MTU CORE AVERAGE EXPOSURE 61 59 57 -3.46 -1.98 -4.09 55 53 51 +++

49 47 45

-5.98 5.6

+++ -5.64 0.9 43 41 .97 12 3.6 -1.49 5.4 4.5 39 +

37 35 ++

33 31 29 0.6 6.31 -1.02

++

27 25 3.04 -0.33 -0.16 -5.76 23 21 I 19 I

17 1.2 2\ 22 3.3 5.89 -10.22 15 13 I

9 .10 32 3.4 -0. 62 -4.32 7

5 3

1

+ +

I I 00 02 04 0o" 08 10 12 14 16 18 2022 24 2628303234363840 424446 ~8 50525 4565860 X

Diff = [(Calc - Meas)/Core Avg TlP Response] X 100%

168

FIGURE 3.3.12 QUAD CITIES UNIT 1 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 7.532'GWD/MTU CORE AVERAGE EXPOSURE QONllOfl IOCAllON 4l,EE lee

~ ce 0

+

o II 0 0 +

5 L Ig Ies +

1$ $

NI 1 o + t 4 00 I

~e 0 I ~e 0 g p o

lI 0 0 4 I 0 ~

C 0 el b o

~e o o 0

t )

I I

~ ~ l ~ i e ~ T ~ Te ll 1$ 1 ~ li 1$ le IT 1$ 1$ le $ 1 $$ $ $ N ~ 1 s e i ~ ~ T $ ~ 1$ 111$ 1$ 111$ 141$ 1~ Telelllllsli CORE AXIALNODE CORE AXIALNODE 0 NNAewso ne Neseosss + IINAssesone ssseosss I'ALosMTso n ~ esseosos 0 OALOOMTCO ne lisle~

~ CONTNOL Noo eoslnON ~ coNTNoL eoo eoetnos IlONIIOR LOCAllON 40,$ E lle 0

~$0

+

0 4 0

Ies 0

0 o 0

~e loge eo

~ ee

~ I l ~ C l ~ T ~ ~ 1 ~ ll Tl 'll li 1$ I~ 1$ 1$ Te $ $ $ 1 $ $ $$ $ $ ~ s e i ~ ~ T ~ ~ Te 11 Tl 1$ li ll Ts TT I ~ le le ~ I ll ll li CORE AXIAL NODE CORE AXIALNODE

+ NNAswlso ill'sseoNss + IITAsssso ne 1$ $$ 0Nss o OAIOSLATSOnf $$ $ $ 0NSS 0 OAICAILATTOne 11$ $ $ 0~

~ OONnCOL 000 KWTICNI ~ OONINOL $ 00 eoelnON 169

FIGUAE 3.3.13 QUAD CITIES UNIT 1 CYCLE 2 AVERAGE AXIALTIP RESPONSE COMPARISON 13.1S8 GND/MTU CORE AVERAGE EXPOSURE 180 160 140 120 I- 0 0 Z

D LI 100 0 O....

+

p Q o e 0 0 9 ': + o Z +

Z 0

&0 (0

IL eo 40 20 0

0 1 2 3 4 6 6 7 8 9 10 11 1213 14161817 18182021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE o CALCULATED TIP RESPONSE 170

FIGURE 3.3.14 QUAD CITIES UNIT 1 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 13.198 GWD/MTU CORE AVERAGE EXPOSURE 61 59 57 -1.53 0.38 -2.00 55 53

+

51 49 -5.45 8.0 9.14 .15 -2,01 0.13 47 45 43 41 -7.62 90 3.8 -2.66 4.7 -1.53 39 37 35 33 5.0 2.6 5.84 3 072 .47 31 29 27 25 -0.16 -2.15 9.3

+++ 30 32 -4 09 23 21

+ + +

19-13'3 17 15 32 -6.15 3.74 .13 08 72 00 2.5 4.40 -0 93 3

1 Y

00 0204060810 12 14 16 18 2022 24 2628303234363840 424446485052 54565860 X

Diff = [(Caic - Mess)/Core Avg TiP Response] X 100%

171

FIGURE 3.3.16 r

QUAD CITIES UNIT 1 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 13.198 GWD/MTU CORE AVERAGE EXPOSURE MOMTOR LOCATION INL$$ MON ITOR LOCAllON 4$ ,$ $

la o

0 ls

+

a 0 ~ ooo e o a

+

r 0

a e s ~ T ~ ~ w TT>> Is TI >>>> IT>>>> as CI aa aa aa ~ ~ s 4 ~ ~ T ~ ~ >> Tl la ls Tl '>> ls TP I ~ '>> ea tl aa aa PI CO CORE AXIALNODE ICAsssso Tl ~ psapopss + IHAsssao Tlp saspol>>s 0 OAIOSLATCS llP SCSPOOSS 0 OA>>CSATCO Tar SCSPOSPS

~ COSTIIOL SOS POCITIOO ~ cocllKILsoo posllas MONITOR LOCATION 40,$ $ MONITOR LOCAllON Egin la la 0

p Tas P P'll 0 a>>s X L X

>>s lle 0 0 os ~ i

~0 0

t.

I a e s s s T ~ '>>ll>>lalCI >>if>>>>asalaaasa

~ 1 ~ s ~ ~ 'p ~ Is 11 ls ls II la Is IT Ia % as.aI aa aa aa CO hE AXIAlNODE CO E

+ ocACCI>>o lip cssposas + aacAsssso lip acapoces o oauwLATso Tlp ssaposss 0 oAIOoLATCO T>> sasposes

~ OOSTCOL OOS PONTlOS ~ ooslsoL aoo poNTlos

- 172-

FIGURE 3.3.16 QUAD CITIES UNIT 1 EOC 1 GAMMA SCAN COMPARISON NORMALIZED AXIAL LA-140 ACTIVITY BUNDLE LOCATION 23,10 1.8 e

1.6 1.4 ~ ~ e ~' ~ ~ v ~ ~

I 1.2 ~ ~ ~ e J ~ J J~ ~ ~ I O e

~

O e

1.0- ~ ~ ~ 'e e

//

~

0.8 /

/e e ~ ~ ~ ~ ~ ~ ~ ~ ~

/

e

/ e e $e

/

I- 0.6 . ~ .s

)

~ -~ .. e

'\

/ .'

\

....:;............:. Legend \

CC 0.4 ~ ~ r ~ ~ e GAMMA SCAN:

e SIMULATE-E 0.2- I ~ ~ ~ ~ ~ 'e e o.o $

2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 . 17 18 19 20 21 22 23 24 BOT1 OM AX1AL NODE TOP

FIGURE 3.3.17 QUAD CITIES UNIT 1 EOC 1 GAMMA SCAN COMPARISON NORMALIZED AXIAL LA-140 ACTIVITY BUNDLE LOCATION 66,40 1.8 1.6 1.4 >> >>

I- CONTROL: RQD 0 I: OSITIQN: ' 0 J I I I 8 J 1.2 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~

r>>

O 0 1.0- '>>

~

~

~

/

~

~

>> ~

y

~

I 0.8 >> / ~

Q ~

LlJ /

O.B

/

~ / >>

/

//

~

LLI >> >>

0.4 / Legend

/ // ~

GAMMA SCAN S I MU LATE-E 0.2 ~ ~ ~ ~ ~ ~ ~ ~ J ~

0.0 2 3 4 5 B 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 BOTTOM AXIAL NODE TOP

FIGURE 3.3.18 QUAD CITIES UNIT 1 EOC 1 GAMMA SCAN COMPARISON NORMALIZED AXIAL LA-140 ACTIVITY 31 BUNDLE AVERAGE 1.6

~ '

r r 1.4 ~ ~ ~ ~ ~ ~ ~

I 1.2 ~ ~ ~ ~ ~ ~ ~ ~ h ~ ~ Wh ~ ~ ~

r Q I V'

~t I 0 1.0 ~ ~

. r

\

P ~

~ ~

I ~ w 0.8 ~

~

~

~

LLI /'

0 1 0.6-

/

/ \

/ I UJ /

0.4 / I Legend ~ l GAMMA SCAN:

SIMULATE-E h

0.2 ~ 4 ~ ~ ~ ~ h ~ ~ h ~ ~ ~

0.0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 BOTTOM AXIAL-,-NODE TOP

FIGURE 3.3.19 QUAD CITIES UNIT 1 EDC 2 RADIAL GJQtSL SCAN COHPARISON 52 50 0.608 0.75?

0.624 0.774 0.016 0.017 0.579 0.737 1.175 0.593 0.769 l. 157

-0.018 0.014 0.032 0.674 0.817 1.015 1.057 X.XXX GAHHA SCAN 0.695 0.849 1.044 1.070 X.XXX SIHULATE-E

0. 021 0. 032 0.029 0.013 X.XXX DIFFERENCE 1.000 1.291 1.016 1.267

l. 273 1.300 0.016 -0.024 0.027 I STANDARD DEVIATION: 1.82/

42 0.828 0.949 1.092 1.120 1.093 0.834 0.969 1.098 1.109 1.100 0.006 0.020 0.006 -0.011 0.007 40 0.573 0.578 0.890 0.888 1.215 1.165 1.070 1.042 1.058 1.049 1.062 1.142 1.070 1.127

l. 234 1.229 0.005 -O.OOR -0.050 -0.028 -0.009 0.008 -0.015 -0.005 0.807 1.035 1.049 1.064 l. 059 1+026 1.034 0.807 1+025 1.057 1.05? 1.057 0.000 -0.010 0.008 -0.007 -0.002 0.008 36 0+680 0+839 1.274 1.0?RI 1.284 1.079 1.269 1.060 1.041 1. RRR 0.69R 0.848 1.204 1.06? I 1.R42 1.076 1.238 1.069 1.051 1.201 0.012 0.009 -0.070 -0.005I-0.042 -0.003 -0.031 0.009 0.010 -0.021 0.942 1.015 1.051 1.066 1.055 1.028 1.035 1.039 1.050 1.038 0.942 1.008 1.042 1 ~ 046 1.060 1.024 1.041 1.044 1.054 1.036 0.000 -0.007 -Oo009 -O.ORO 0.005 "0.004 0.006 0.005 0.004 -0.002 32 0 ~ 703 0 831 0 ~ 904 0 973 1 010 1 015 1 040 1 044 1 026 1 005 1 ~ 005 1 029 1 066 1.037

0. 719 0.843 0. 919 0. 972 0. 996 1. 009 1. 030 1. 031 1. 024 1. 005 0. 997 1. 024 1. 060 1.05R 0 ~ 016 0.012 0.015 -0.001 -0.014 -0.006 -0.010 -0.013 -0.002 0.000 -0.008 -0.005 -0.006 0.015 30 1 3 5 7 9 11 13 15 17 19 21 23 25 R7 29 31 33

FlGURE 3.3.20 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON A BUNDLE ID: CX0662 Ol Legend

~ ...'-" 0 =Measured o = Calculated I- ~

)

V C)

I O

Cl ll!

N I I CC ~

0 C) 0.0 12.0 24.0 3B.O 48.0 B0.0 72.0 84.0 9B.O 108.0 120.0 132.0 144.0 DlSTANCE FROM BOTTOM OF GORE (IN)

FIGURE 3.3.21 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: CX0399

~ ~ ~

Legend

~ -..'--. 0 =Measured 0- o = Calculated p v-I-

0 CD I

Q Q

LQ N

I

<o 0

\

Q 0.0 12.0 24.0 36.0 48.0 BO.O 72.0 84.0 SB.O 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

FIGURE 3.3.22 QUAD CITIES UNIT 1 EQC 2 GAMMA SCAN CQMPARISQN BUNDLE ID: CX0231 0

Legend

~ ...'--. o Measured o Calculated I- ~

)

Q C)

I g O O

lu N

0 Z

C)

O 0.0 12.0 24.0 3B.O 48.0 BO.O 72.0 84.0 9B.O 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

FIGURE 3.3.23 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: CX0297 Legend

.-----:-" o =Measured 0- o = Calculated

)I- ~ I P

O C) lI O

n M

N EC ~

0 z

0.0 12.0 24.0 S6.0 48.0 80.0 72.0 84.0 98.0 108.0 120.0 182.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

FIGURE 3.3.24 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: CX07'I7

~ ~ ~

Legend

~ ---.'.--. 0 =Measured o = Calculated I- IO

)I- r V

I O

1 Q

Ul N

K~

0 x

0.0 12.0 24.0 36.0 48.0 60.0 72.0 84.0 96.0 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

FIGURE 3.3.25 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: CX0378 Legend

~ ".-'.-" 0 =Measured o = Calculated L- u)

I-0 C)

I O

Q LU rl

<o 0

0.0 12.0 24.0 3B.O 48.0 60.0 72.0 84.0 9B.O 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

FIGURE 3.3.2B QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: CX0150 Legend

~ . ';" o = Measured 0 = Calculated l- ~

Q v'-

O a

g O T ~ ~ ~ ~ ~ ~ ~ ~

Q 0 LLI N

0 R

0.0 12.0 24.0 36.0 48.0 60.0 72.0 84.0 86.0 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF GORE (IN)

FIGURE 3.3.27 QUAD CITIES UNIT 1 EOC 2 GAMMA SCAN COMPARISON BUNDLE ID: GEH029 Legend

~ .""'.-" CI Measured 0 Calculated I- ~

) r' CO I

O T

0Ul t4 K

0Z g)

I I 0.0 12.0 24.0 3e.o 48.0 B0.0 72.0 84.0 88.0 108.0 120.0 132.0 144.0 DISTANCE FROM BOTTOM OF CORE (IN)

3.4 Peach Bottom Unit 2 C cles 1 and 2 Com arisons One specific application of PPGL's steady state core physics methods and models is to provide input to 'the transient analysis benchmarking of the Peach Bottom Unit 2 end of Cycle 2 turbine trip tests. In order to provide the necessary input, SIMULATE-E models of the Peach Bottom Unit 2 Cycles 1 and 2 cores were developed. These models were then used to simulate the Peach Bottom Unit 2 core depletion through Cycles 1 and 2. Comparisons to TIP measurements taken during Cycles 1 and 2 and to General Electric Company (GE) process computer Pl power distributions taken prior 'to the turbine trip tests assess the accuracy of the core depletion calculations.

Peach Bottom Unit 2 is a General Electric BWR-4 core that consists of 764 fuel assemblies with an active core height of 144 inches. The initial cycle contained 764 General Electric 7x7 fuel assembliesg Cycle 2 contained 576 initial-core fuel assemblies and 188 8x8 fresh fuel assemblies. Although reactor design and rated conditions are quite similar to Susquehanna SES, the Peach Bottom Unit 2 core loading pattern, fuel bundle design, inlet flow orifices, and core support plate bypass flow paths are significantly different. These design differences were taken into account in development of the Peach Bottom Unit 2 SIMULATE-E model. A more detailed description of the Peach Bottom Unit 2 core is found in Reference 30.

The average RMS of the differences between the SIMULATE-E calculated and measured TIP responses for each Peach Bottom Unit 2 TIP response comparison is calculated as described in Section 3.2.3. Figure 3.4.1 shows the RMS of the TIP response comparisons for Peach Bottom Unit 2 Cycles,l and 2. These comparisons are slightly worse than Susquehanna SES results but are still quite good. The Peach Bottom Unit 2 core operating data (Reference 30) used for modeling the core depletion was less detailed than the data used for Susquehanna SES. This lack of detailed data may be the cause of these slightly worse results.

Figures 3.4.2 through 3.4.4 show the end of Cycle 1 core average axial, radial, and four individual TIP response comparisons, respectively. Figures 3.4.5 through 3.4.7 present the same comparisons for end of Cycle 2. As shown 185

in these figures, the calculated TIP response agrees well with the measured data. These results therefore indicate that the SIMULATE-E models accurately'alculate three-dimensional core exposure, void history, and control history arrays for the end of each cycle.

As previously stated, the primary purpose for developing the Peach Bottom Unit 2 models was to generate the necessary transient analysis inputs (e.g.,

cross sections and kinetics parameters) . The end of Cycle 2 TIP response comparison indicates that the core history arrays have been accurately calculated. Because the turbine trip tests were performed over a span of a few weeks with a core power history plagued by nonsteady state operation, careful analysis of power maneuvers was required to adequately calculate the actual xenon concentration at the time of the tests. The accuracy of the calculated xenon concentration immediately prior to each turbine trip test can be assessed by comparing the SIMULATE-E calculated power distribution to the available GE process computer Pl power distribution (Reference 31) .

Figure 3.4.8 shows each axial power distribution comparison. The SIMULATE-E calculated power distributions are based on actual core conditions prior to the tests as reported in Reference 31. The figure shows the three different power distributions (i.e., top peaked, middle peaked, and slightly bottom that existed at the time of the three turbine trip tests. This I'eaked) indicates that the core conditions were considerably different for each test, and that the SIMULATE-E model is capable of calculating these differences.

Typical reload design and licensing applications do not require modeling the complexity of nonequilibrium xenon. Therefore, this benchmark provides a good test of PPGL's steady state physics models and methods in an application which is more difficult than the normal reload analyses.

186

FIGLIRE SA.1 PEACH BOTTOM UNIT 2 CYCLES 1 AND 2 RELATIVE NODAL RMS OF TIP RESPONSE COMPARISONS 12.0

~ ~ ~

11.0 10.0-I 9.0- "-

I CO

&.0-fL" 7.0-D '~ ~

0 60-.

Z UJ 5.0 4.0- ". J~

CC Legend 0 "

I- 3.0- ". PB2C1 PB2C2 2.0-1.0-0.0- '

0 3 4 5 6 7 8 10 11 12 13 14 CORE AVERAGE EXPOSURE (GWD/MTUj

FIGURE 3.4.2 PEACH BOTTOM UNIT 2 CYCLE 1 AVERAGE AXIALTIP RESPONSE COMPARISON 11.133 GWD/MTU GORE AVERAGE EXPOSURE 80 80

+++++ 0 0 0

70 +

0

+ 0 0

Bo 0

I- +

z 0 0

50 IU z

+ 0 4o J U

M CL so 20 10 0 I 0 1 2 S 4 5 6 7 8 8 10 11 121S 14 16 16 17 18 182021222324 CORE AXIALNODE

+ MEASURED TIP RESPONSE 0 CALCULATED TIP RESPONSE 188

FIGORE 3.4.3 PEACH BOTTOM UNIT 2 CYCLE 1 TIP RESPONSE COMPARISONS 'ADIAL 1l. I33 GWD/MTU CORE AVERAGE EXPOSURE 61 59 57 -3.03 1.67 3.81 -5.06 55 53 51 49 -3.30 -4.36 0.12 -1.27 -6. 44 47 45 43 41 -4.93 -0.20 9.5 3.4 0.06 -0 .43 39 37 35 ++

33 31 29

++ 11.53 -0.65 -3.11 1.0 27 25 -0.56 3.50 .16 -1.27 -3.79 4.7 23 21 19

++++

17 15

-0.89 -3.09 2.27

++++ -4 66 13

++ -0.09 -5 47 I

3.

++ -0.11 Y

1 I I I I'I I 000204060810 12 14 16 18 2022 24 2628303234363840 42 4446485052 54565860 X

Diff = [(Calc - Mess)/Core Avg TiP Responsej X 100%

189

FIGURE 3.4.4 I

PEACH BOTTOM UNIT 2 CYCLE 1 INDIVIDUALTIP RESPONSE COMPARISONS 11.133 GWD/MTU CORE AVERAGE EXPOSURE

~ IOHITOA LOCAllOH dIL$$ $ IOHIIOR LOCATIOH $ $ ,$ $

lee tse tee IIS Tes L

?

~

Iee g Iee gg 0 ~ 4 g

~s f

l's 0 0

0 + 0

+40 4 4 P

0

~ ~ s ~ ~ s 0 r ~ ~ te tt Te ts tt ts te tr es v se st ss es st ~ ~ S 0 ~ 0 r ~ ~ tenteteettete tr le ls le el ee ee sli CORK AXIALHOOK CO RK AXIALHOOK p NRAsveso lit NespoNss + NAstpteo Tlt Nssposss 0 INSOVIATSO Tl ~ RSSPONSS 0 OAIOVIATRD11P RssPONes N CONTROl ROO POSITION ~ CONTROL OOO toelllON MONTOll LOCAllOH 40,$ $ IIOHITOR LOCATIOH $ $ .$ $

tee lee tee P

P

?? 0 0 0 ~  ?

es ~0 0ee 4 0 0

o I 0 4

I I~ l g l

~ 1 ~ s ~ s ~ r s ~ le 11 te le 11 ts le lr te ts ee el se ss sl ~ 1 ~ s ~ ~ ~ 'I ~ e le lt ts le ll 1 ~ te tr le le ee el ee ee sl CORK AmAL HOOK CORK AXIALHOOK t

+ NaASVRm Tt RSSPO<<SS + NSASVRSO Ttt IleetONSS o OASOVlAISO Tlt RSPPONSS 0 CAlov lyso Tlt RsspoNes

~ CONTROL ROO PON110N ~ CCNTttoe Roo toNTICN 190

FIGURE 3.4.5 PEACH BOTTOM UNIT 2 CYCLE 2 AVERAGE AXIALTIP RESPONSE COMPARISON 13.812 GWD/MTU CORE AVERAGE EXPOSURE 180 160 140 120 ~ ~

I- tb K

D ..+.0 'P .g.

100 0 lU 0 K

Q .............

80 CO 0 CL eo

+

0 40 20 0

0 1 2 3 4 5 6 7 8 8 10 11 121314151617 18182021222324 CORE AXIALNODE

+ 'MEASURED TIP RESPONSE 0 CALCULATEDTIP RESPONSE 191

FlGURE 3.4.6 PEACH BOTTOM UNIT 2 CYCLE 2 RADIAL TIP RESPONSE COMPARISONS 13.812 GWD/MTU CORE AYERAGE EXPOSURE 61 59 57 -3.02 3.93 6.04 -9.96 55 53 51 49 -8.41 6.89 -3.78 3.24 5.34 47 45 43 41 -5. 20 0.4 -0. 35 -0.36 -0.93 -2.68 7.99 39 37 35 +++

33 31 2.01

+++ -2.70 -3.90 2.6 I

5.

29 27 ++ -4.07 -0.02 25 23

.19

++ 3.6 -3.21 3.9 21 19

-0 .13

++

17 15 2\ 1 2 -1.85 0.45

++ -3.26 -2.56 13

++

-0.21

++ -11.21 I I 0002 04 0608 10 12 14 16 18 2 022 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 5 4 565860 X

Diff = [(Calc - Meas)/Core Avg TIP Response] X 100%

192

FIGURE 3.4.7 PEACH BOTTOM UNIT 2 CYCLE 2 INDIVIDUALTIP RESPONSE COMPARISONS 43.812 GWD/MTU CORE AVERAGE EXPOSURE MONITOR LOCATION 50T$ $ MONITOR LOCATION 4$ T$ $

IW Na 1W

$ o Z 0 at IW 0 4

4

+ o

+ o 0 I a a ~ a 0 r ~ ~ >> 11 a la 11>>>> <<>>>> ae al as sa ai ~ a ~ 4 ~ ~ r ~ ~ >> 11 Is I ~ %>>>> Tr>>>> w sl as as ss CORE AXIALNODE CORE AXIALNODE

+ 1>>saoa>>I nr Naaroosa + 1>>saoo>>I nt Naarooaa 4 OAONLITciaonr IltaroNot o ossouulta nr Ntaroooa

~ CONTNOL Iioo to ainON

~ CONTNOL Noo roaNTON MOMTOR LOCATION 40,$ $ MONITOR LOCATION $ $ ,$ $

IW aa Tr L 4 0 4 4

+iog 44 4

X +

4 4 4

'll IW 4

~0 0 I

J.

~0

~ I ~ s ~ a ~ r a ~ ls 11 ls ls N>>>> it ls ia ss sl as ss ai s a ~ a ~ r ~ >> n Ts is N is I~ ir>> a ss si ss ss si CORE AXIALNODE COllE AXNLNODE 4 Wruuaau nr NaarONSa 4 Nahaislao nr oasroNso 4 CIIoosssao Tlt NaaroNaa o ossoosslaont Ncarooaa

~ CONTROL Iioo rOWloN ~ OONTIIOL NOO toainoN 193

FIGURE 3.4.8 PEACH BOTTOM UNIT 2 END OF CYCLE 2 CORE AVERAGE AXIALPOWER DISTRIBUTIONS 1.5 CL LLI 1.0 O

CL LIJ Legend I PI Data 0.5 SIMULATE-E LIJ CY 0.0 1 3 12 15 18 21 24 1.5 1.0 O

')I LIJ Legend Pf Data 0.5 SIMULATE-E LLI CL 0.0 I 1 3 12 15 18 21 24 1.5 CL 1.0 O

CL

)

LIJ Legend PI Dafa 0.5 SIMULATE-E CL 0.0 I 3 9 12 '15 18 21 24 BOTTOM AXIAL NODE TOP 194

4.0 SPECIAL APPLICATIONS WITH PD 7 Occasionally, applications require multiple assembly calculations. The lattice physics code CPM-2 is a single assembly code which is not capable of performing multiple bundle calculations. For these cases, the PDQ7 program is used. PDQ7 has been used for criticality analyses and to provide input to the three-dimensional nodal simulation codes.

To demonstrate PPsL's ability to use PDQ7, two sets of problems are presented.

The first set contains calculations of the uniform lattice criticals presented in Section 2.2 which were analyzed with CPM-2. The second set contains single fuel bundle calculations with both CPM-2 and PDQ7. For these cases, pin power distributions and assembly reactivities are compared.

195

4.1 Descri tion of PD 7 The PDQ7 computer program (Reference 32) was developed f'r fine mesh few group diffusion theory analysis. The program solves the neutron diffusion equation in one, two or three dimensions. Available options include rectangular, hexagonal, cylindrical or spherical geometries. A maximum of five energy groups are permitted. The mesh spacing is flexible allowing the user to define as much geometric detail as appropriate for the specific problem.

Cross sections for each problem may be input to PDQ7 as either macroscopic or microscopic data. At PPGL, this data would typically be CPM-2 generated macroscopic cross sections. For most applications, four group cross sections are used with energy boundaries as defined in Table 4.1.1.

- 196

TABLE 4.1.1 ENERGY GROUP STRUCTURE USED IN PDQ7 CALCULATIONS Energy Boundaries

~Gzou (eV) 1.0 x 10 7 8.21 x 10 5 5' -

2 8.21 x 10 5.53 x 10 5.53 x 10 3 - 0.625 0.625 - 0.0 197

4.2 Uniform Lattice Criticals The same uniform lattice criticals evaluated with CPM-2 in Section 2.2 were also analyzed with PDQ7. One>>dimensional cylindrical geometry was used to model each uniform lattice critical. The critical radius was defined to conserve the core cross sectional area and was determined from the critical number of pins. PDQ7 cross sections for the core region were obtained from CPM-2 pin cell calculations. The reflector cross sections were 'obtained from Reference 33. Because a radial reflector region was included in the PDQ-7 model, only an axial buckling term was required to account for the leakage.

As with the CPM-2 uniform lattice critical calculations presented in Section 2.2, the TRX and ESADA experiments were modeled with PDQ7. Tables 4.2.1 and 4.2.2 show the results of the PDQ7 calculations. The CPM-2 results from Section 2.2 are also included for comparison. The results from the TRX and ESADA calculations yield similar K-effectives.

- 198

TABLE 4 2.1 P 7 RESULTS FOR TRX CRITICALS Experimental Axial Experiment CPM-2 Material guckling PDQ7 Identification K-effective (m ) K-effective TRX1 0.9934 5.04 0.9969 TRX2 0.9958 5.12 0.9973 TRX3 0.9942 5.32 0.9954 0.9939 5.11 0.9961 TRX5 0.9934 5.26 0.9950 TRX6 0.9974 5.25 0.9996 TRX7 0.9970 5.25 0.9996 TRX8 0.9960 5.31 0.9978 Average K-effective 0.9951 0.9972 Standard Deviation 0.0016 0.0017 199-

TABLE 4 2 2 P 7 RESULTS FOR ESADA CRITICALS Experimental Axial Experiment CPM-2 Material chuckling PDQ7 Identification K-effective* (m ) K-effective*

ESADA 1 1.0026 8.56 1.0122 ESADA 3 1.0004 8.967 1.0158 ESADA 4 1.0129 9.466 1.0152 ESADA 6 1.0116 9.471 1.0133 ESADA 12 1.0101 9.436 1.0162 ESADA 13 1.0077 9.639 1.0140 Average K-effective 1.0076 1.0144 Standard Deviation 0.0050 0.0016

• All CPM-2 and PDQ7 calculated K-effectives have been adjusted by -0.4% ~K to account for spacer worth.

200

4.3 Com arisons to CPM-2 A second qualification of the use of PDQ7 at PPaL is through comparison to single assembly CPM-2 lattice physics calculations. To facilitate generation of the PDQ7 cross section data, the COPHIN (Reference 34) code was used.

Separate planar regions are defined for different fuel pin types, water rods and other regions (i.e., control rod, water gap, etc.). Fuel pin regions are grouped according to fuel pin enrichment and location. The mesh description is defined to explicitly model each pin and to conserve the volumes of each region.

When the fuel assembly being modeled contains gadolinia or a control rod, the neutron flux depression caused by the presence of the strong absorber can be reproduced using diffusion theory with a shielding factor. Without a shielding factor diffusion theory results in an overestimation of the neutron flux in the absorber region and a corresponding overestimation of the absorber worth. Shielding factors are developed and applied to the Group 4 (thermal) absorption and fission cross sections for gadolinia bearing fuel pins and the Group 3 and 4 absorption cross sections for control rods. These factors are derived by conserving the CPM-2 calculated absorption rate in the absorber.

The fuel assemblies chosen for the comparison are the Susquehanna SES initial core bundle designs. Two separate fuel designs were chosen for the analysis.

The results are shown in Figures 4.3.1 through 4.3.4. The agreement in power distribution for a single assembly is very good. The assembly eigenvalues (K-infinities) also agree well between the two codes, differing by less than 1 mk (or 0.1% bk) . This demonstrates that PPGL can perform accurate PDQ7 assembly calculations.

201-

FIGURE 4.3.1 CPM-2 VS PDQ7 PIN POWER DISTRIBUTION COMPARISON GE INITIALCORE HIGH ENRICHED FUEL TYPE UNCONTROLLED WIDE GAP WIDE 1.027 CPM-2 GAP 1.065 2.7 PDQ7 1.104 0.9 80  % DIFFERENCE 1.113 0.969 0.8 >>1.1 1.117 1.047 0.965 1.116 1.029 0.966

-0.2 -1.7 -0.9 1.147 1.076 0.860 0.114 1.140 1.060 0.875 0.116

-0.6 -1.6 1.7 0.9 1.124 1.022 0.842 0. S93 1 126

~ 1.022 0.862 1.001 0.2 0.0 2.4 0.8 1.074 1.029 0.110 0.893 0.992 0.948 1.081 1.037 0.109 0.901 0.973 0.923 0.7 0.8 -0.9 0.9 -1.9 -2.6 1.066 0.988 1.040 1.067 1.012 1.099 1.013 1.081 0.987 1.046 1.049 0.992 1.069 0.994 LINE OF 1.6 -0.1 0.5 -0.8 -2.0 -2.7 -1.9 SYMMETRY 0.997 1.069 1.086 1.146 1.179 1.148 1.132 1.060 1.033 1.084 1.089 1.143 1.166 1.140 1.134 1.073 3.6 1.4 0.4 -0.3 <<1.1 -0.7 0.2 2.2 CPM-2 K-INFINITY= 1.1428 PDQ7 K-INFINITY= 1.1426 202

FlGURE 4.3.2 CPM-2 VS PDQ7 PIN PO'tN'ER DISTRIBUTION COMPARISON GE INITlALCORE HIGH ENRICHED FUEL TYPE CONTROLLED WIDE GAP WIDE 0. 380 CPM-2 GAP 0.400 5.3 PDQ7 0.486 0.591  % DIFFERENCE 0.523 0.612 7.6 3.6 0.549 0.740 0.826 0.578 0.786 0.853 5.3 6.2 3.3 0.607 0.837 0.824 0.129 0.632 0.876 0.877 0.133 4.1 4.7 6.4 3.1 0.645 O.855 0.877 1.232 0.666 0.898 0.924 1.239 3.3 5.0 5.4 0.6 0.690 0.924 0.128 1.101 1.276 1.260 0.703 0.966 0.130 1.103 1.236 1.214 19 45 1.6 0.2 -3.1 -3.7 0.824 0.989 1.210 1.334 1.335 1.493 1.404 LINE OF 0.832 0.995 1.19 6 1.293 1.284 1.420 1.353 1.0 0.6 -1.2 -3.1 -3.8 -4.9 -3.6 SYMMETRY

. 0.952 1.166 1.300 1.465 1.573 1.577 1.586 1.487 0.973 1.173 1.288 1.433 1.519 1.524 1.539 1.465 2.2 0.6 -O.S 34 -3A -3.0 -1.5 CPM-2 K-INFINITY= 0.9623 PDQ7 K-INFINITY= 0.9615 203

FIGURE 4.3.3 CPM-2 VS PDQ7 PIN POWER DISTRIBUTION COMPARISON GE INITIALCORE MEDIUM ENRICHED FUEL TYPE UNCONTROLLED WIDE GAP WIDE 1.064 CPM-2 1.096 GAP 3.0 PDQ7 1.104 0.988  % DIFFERENCE 1.116 0.979 1.1 -0.9 1.120 1.084 0.921 1.121 1.067 0.915 0.1 -1.6 -0.7 1.082 1.032 0.847 0.139 1.078 1.018 0.854 0.139

-0.4 -1.4 0.8 0.0 1.080 1.039 0.907 0.124 1.079 1.026 0.908 0.126

-0.1 -1.3 0.1 1.6 1.122 1.096 0.962 0.885 0.794 0.900 1.122 1.078 0.952 0.889 0.809 0.898 0.0 -1.6 -1.0 0.5 1.9 -0.2 1.106 0.993 1.091 1.028 1.021 0.875 0.989 LINE OF 1.116 0.982 1.074 1.018 1.011 0.863 0.985 0.9 -1.1 -1.6 -1.0 -1.0 -1.4 -0.4 SYMMETRY 1.060 1.104 1.117 1.072 1.075 1.116 1.097 1.054 1.094 1.116 1.118 1.073 1.073 1.120 1.112 1.088 3.2 1.0 0.1 0.1 -0.2 0.4 1A 3.2 CPM-2 K-INFINITY= 1.1107 PDQ7 K-INFINITY= 1.1100 204

FIGURE 4.3.4 CPM-2 VS PDQ? PIN POWER DISTRIBUTION COMPARISON GE INITIALCORE MEDIUM ENRICHED FUEL TYPE CONTROLLED WIDE GAP WIDE 0.399 CPM-2 GAP 0.419 5.0 PDQ7 0.495 0.604  % DIFFERENCE 0.631 0.625 7.3 3.6 0.562 0.776 0.794 0.688 0.801 0.821 4.6 3.2 3.4 0.588 0.814 0.813 0.167 0.610 0.836 0.866 0.1B2 3.7 2.6 6.2 3.2 0.640 0.890 0.946 0.166 O.B64 0.897 0.969 0.169 2.2 0.8 2.5 2.6 0.742 1.014 1.061 1.076 1.028 1.208 0.743 1.028 1.0BO 1.076 1.033 1.192 0.1 1.4 -0.1 0.0 0.5 -1.3 0.869 1.006 1.266 1.290 1.361 1.1S7 1.381 LINE OF 0.868 0.998 1.244 1.266 1.313 1.163 1.349

-0.1 -0.7 -1.7 -1.9 -2.8 -2.8 -2.3 SYMMETRY 1.019 1.211 1.342 1.373 1A40 1.643 1.648 1.606 1.034 1.212 1.326 1.348 1.40B 1.609 1.521 1.600 1.6 0.1 -1.2 -1.8 -2.4 -2.2 107 -0.4 CPM-2 K-INFINITY= 0.9230 PDQ7 K-INFINITY= 0.9238 205

l 5.0

SUMMARY

AND CONCLUSIONS The analyses presented in this topical report demonstrate the validity of PPaL's analytical methods as well as PPGL's qualifications to perform steady state core physics calculations for reload design and licensing analysis applications.

The lattice physics qualification has been accomplished through comparison of the CPM-2 computer code results to various measurement data. Comparisons to 14 uniform lattice critical experiments yields an average K-effective of 1.0005 with a standard deviation of 0.0072. The average K-effective for the UO criticals is 0.9951 and the average K-effective for the plutonium criticals is 1.0076. The pin power distribution and hence local peaking factor calculation, has been benchmarked to the gamma scan data from Quad Cities Unit 1 which was taken at the end of Cycle 2. The average standard deviation from all of the comparisons is 4.0%. If only the UO bundles are considered, the average standard deviation reduces to 3.37%. this is close to the reported 3.0% practical accuracy of the data. The qualification of the lattice physics methods also relies on the original benchmarking of EPRI-CPM provided by EPRI. Because the neutronics methods in CPM-2 are identical to those in EPRI-CPM, this benchmarking remains valid for CPM-2. Some of the uniform lattice criticals analyzed in the EPRI benchmarking are the same experiments a's those analyzed by PPGL. After compensation was made for the correction factors applied to the EPRI-CPM results, the results from EPRI-CPM agreed very well with those from CPM-2.

The qualification of the core simulation methods not only demonstrates the accuracy of SIMULATE-E but also provides a demonstration of the entire steady state core physics methodology. The benchmarking results show that the calculated hot critical core K-effectives from SIMULATE-E can be accurately predicted by a correlation which considers both core gadolinia content and core average exposure. The mean difference between the SIMULATE-E calculated core K-effective and the correlation is only 0.00002 ak with a standard deviation of 0.00061 rlk. The cold critical core K-effective from SIMULATE-E can be accurately predicted by adding a constant bias of 0.00659 bk to the hot critical K-effective correlation. Comparisons of cold critical calculations 206-

to the target results in a standard deviation of 0.00137 ~k. In addition, there is no significant difference between the cold in-sequence and local critical calculations.

Comparisons of predicted TIP responses to measured TIP response data were performed as a means of assessing the accuracy of the SIMULATE-E power distribution calculation. Susquehanna SES nodal TIP response comparisons, which demonstrate the accuracy of the detailed power distribution, show an aver'age RMS of 5.74%. Radial TIP response comparisons were also performed in order to demonstrate the accuracy of the bundle power distribution, and the average RMS for Susquehanna SES is 2.58%. The same types of TIP response comparisons were also made for the first two cycles of Quad Cities. The average nodal TIP RMS is 9.84% and the average radial RMS is 5.26%.

Additionally, the SIMULATE-E power distribution calculations have been compared to the gamma scan measurements taken at the end of the first and second cycles of Quad Cities Unit 1. These measurements are representative of the core power distribution averaged over the last two to three months of operation. SIMULATE-E was used to calculate the nodal La-140 concentrations for comparison to the measured data. The results of the nodal comparisons, neglecting peripheral and axial end nodes, yield an RMS of 5.45%. For the radial comparison, neglecting peripheral bundles, an RMS of 1.92% was obtained. The axial peaking factor (on a nodal basis) was also compared to the measured gamma scan data. The average difference in the axial peaking factor was 1.2% with a standard deviation of 2.1% for Cycle 1 and -0.2$ with a standard deviation of 1.5% for Cycle 2.

This report also included SIMULATE-E calculations for Cycles 1 and 2 of Peach Bottom Unit 2. These calculations were performed in order to generate the neutronics input to PPGL's transient analysis methods benchmarking against the Peach Bottom end of Cycle 2 turbine trip tests. The predicted power distributions for each of the three turbine trip tests show excellent agreement to reported plant process computer data.

The PDQ7 computer program is used for special applications to perform mul+iple bundle criticality analyses and to augment nodal simulation code input. A demonstration of PPGL's use of the PDQ7 program includes comparisons to 207

uniform lattice critical experiments and pin power distribution calculations

'with CPM-2.

Xn conclusion,'he analysis results contained in this topical report demonstrate PP&L's qualifications to perform steady state core physics calculations. Extensive comparisons to measured data from Susquehanna SES, Quad Cities .Unit l, and Peach Bottom Unit 2 demonstrate the validity of the analytical methods as well as PPGL's capability to set up and properly apply the models. Comparisons to reactor designs other than PPGL's Susquehanna SES demonstrate PPGL's ability to extend the core modeling techniques developed for Susquehanna SES to other fuel and core designs.

PPGL is committed to maintaining a strong in-house core analysis capability and as part of that commitment, we continually evaluate the accuracy of our core simulation methods and make modeling improvements when appropriate.

Although PPGL's day-to-day core follow analyses are aimed primarily at plant operations support, the comparisons of *SIMULATE-E calculations (e.g., TXP response, K-effective, thermal margins) to the plant data also serve as a continuing methods benchmarking effort.

- 208

6.0 REFERENCES

1. NRC Generic Letter Number 83-11, "Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions," February 8, 1983.

2. "Advanced Recycle Methodology Program," EPRZ CCM-3, September, 1977.

3. D. B. Jones, "CPM-2 Computer Code User's Manual," Part II, Chapter 6 of EPRI NP-4574-CCM, February, 1987.

4. M. Edenius, "EPRI-CPM Benchmarking," Part 1, Chapter 5 of EPRI CCM-3, November, 1975.

5. A. Ahlin, et. al., "The Collision Probability Module EPRI-CPM," Part II, Chapter 6 of EPRZ CCM-3, November, 1975.

6. R. Stamm'ler, et. al., "Equivalence Relations For Resonance Integral Calculations," Journal of Nuclear Energy, Volume 27, page 885, 1973.

7. M. Edenius, A. Ahlin, "MICBURN: Microscopic Burnup Zn Gadolinia Fuel Pins," Part ZI, Chapter 7 of EPRZ CCM-3, November, 1975.

8. M. Edenius, et. al., "The EPRI-CPM Data Library," Part II, Chapter 4 of EPRI CCM-3, November, 1975.

9. L. Hellstrand, "Measurements of Resonance Integrals Reactor Physics in the Resonance and Thermal Regions," Proceedings of the National Topical Meeting, San Diego, CA, Volume II, page 157, February, 1966.

10. J. R. Brown, et. al., "Kinetic and Buckling Measurements on Lattices of Slightly Enriched Uranium or UO Rods In Light Water," WAPD-176, January, 1958.

11. R. D. Learner, et. al., "PuO - UO Fueled Critical Experiments,"

WCAP-3726-1, July, 1967.

209-

i'l

12. M. B. Cutrone and G. F. Valby, "Gamma Scan Measurements at Quad Cities Nuclear Power Station Unit 1 Following Cycle 2," EPRI NP-214, July, 1976.

13. R. J. Nodvik, "Supplementary Report. on Evaluation of Mass Spectrometric and Radiochemical Analysis of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, 1969.

14. R. J. Nodvik, "Saxton Core II Fuel Performance Evaluation," Part II WCAP-3385-56.

15. D. M. VerPlanck, "SIMULATE-E: A Nodal Core Analysis Program for Light Water Reactors," EPRI NP-2792-CCM, March, 1983.

16. A. Ancona, "Reactor Nodal Method Using Response Matrix Parameters," Ph.

D. Thesis Rensselaer Polytechnical Institute, 1977.

17. S. Borresen, "A Simplified, Coarse Mesh, Three-Dimensional Diffusion Scheme for Calculating the Gross Power Distribution in a Boiling Water Reactor," Nuclear Science and Engineering, Volume 44, pages 37-43, 1971.

18. G. S. Lellouche and B. A. Zolotar, "Mechanistic Model For Predicting Two-Phase Void Fraction For Water in Vertical Tubes," EPRI NP-2246-SR, February, 1982.

19. B. J. Gitnick, "FIBWR: A Steady-State Core Flow Distribution Code for Boiling Water Reactors; Computer Code User's Manual," EPRI NP-1924-CCM, July, 1981.

20. D. B. Jones and M. J. Anderson, "ARMP-02 Documentation: Part II, Chapter 12-NORGE-B2 Computer Code Manual," EPRI NP-4574-CCM, Part II-, Chapter 12, December, 1986.

21. B. L. Darnell, et. al., "SIMULATE-E: A Nodal Core Analysis Program for Light Water Reactors," EPRI NP-2792-CCM (Draft Revision), Appendix D, May, 1986.

- 210-

22.. A. F. Ansari, et. al., "FIBWR: A Steady-State Core Flow Distribution Code for Boiling Water Reactors," EPRI NP-1923, July, 1981.

23. R. B. Macduff and T. W. Patten, "XN-3 Critical Power Correlation,"

XN-NF-512(P)(A) Revision 1 and Supplement 1, Revision 1, October 21, 1982.

24. S. W. Jones, et. al., "POWERPLEX Core Monitoring Software System Software Specification for the Susquehanna Steam Electric Station Susquehanna Units 1 and 2," XN-NF-83-35(P), Revision 1, August, 1986.

25. "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NED0-10958-A, January, 1977.

26. M. Edenius, "Studies of the Reactivity Temperature Coefficient in Light Water Reactors," AE-RF-76-3160, A. B. Atomenergi, 1976.

27. N. H. Larsen, et. al., "Core Design and Operating Data for Cycles 1 and 2 of Quad Cities 1," EPRI NP-240, November, 1976.

28. N. H. Larsen, "Core Design and Operating Data for Quad Cities 1 Cycle 3,"

EPRI NP-552, March, 1983.

29. G. R. Parkos, "BWR Simulator Methods Verification," NED0-20946A, January, 1977.

30. N. H. Larsen, "Core Design and Operating Data For Cycles 1 and 2 of Peach Bottom 2," EPRI NP-563, June, 1978.

31. L. A. Carmichael and R. D. Niemi, "Transient and Stability Tests at Peach Bottom Atomic Power Station Unit 2 at the End of Cycle 2," EPRI NP-564, June, 1978.

32. W. R. Cadwell, "PDQ7 Reference Manual," WAPD-TM-678, January, 1967.

211-

33. W. J. Eich, et. al., "Few Group Baffle and/or Reflector Constants for Diffusion Calculation Application," EPRI NP-3642-SR, August, 1984.

34. R. D. Mosteller and R. S. Borland, "COPHIN Code Description,"

EPRI NP-1385, April, 1980.

- 212

RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION 213

Pennsylvania Power 8 Light Company TWO NOrth Ninth Street ~ AllentOWn. PA 18101 ~ 215 I 770 5151 Harold W. Keiser Vice President-Nuclear Operations 215/770-7502 pEB >7 $ 88 Director of Nuclear Reactor Regulation Attention: Dr. W. R. Butler, Project Director

~

Project Directorate I-2 Division of Reactor Prdjects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO RAI ON CORE PHYSICS TOPICAL PLA-2983 FILES A7-8A, R41-2 Letter, M.C. Thadani to H.W. 'eference:

Keiser, "Request for Additional Information", dated January 11, 1988.

Dear Dr. Butler:

Attached please find PP&L's responses to the referenced staff questions on our topical report PL-NF>>87.-001, "Qualification of Steady State Core Physics Methods for BWR Design and Analysis."

Please be advised .that the schedule for the submittal of our remaining topical reports has been revised as follows:

Qualification of Transient Analysis July, 1988 Methods for BWR Design and Analysis Application of Reactor Analysis November, 1988 Methods for BWR,Design and Analysis Due to these delays in our planned completion dates, PP&L has also revised the first reload application of our in-house methods from Susquehanna SES Unit 1 Cycle 5 to Susquehanna SES Unit 2 Cycle 4 (planned startup: November 10, 1989). Accordingly, we are revising our request for your approval of PL-NF-87-001 from March, 1988 to July 5, 1988.

FILES A7-8A, R4)-2 PLA-2983 Dr. W. R. Butler Also attached for insertion into PL-NF-87-001 are replacement pages 51 and 208, which correct minor typographical errors, and replacement page 69 (Table 3.2.3), which provides corrected cycle and core average exposure values for Case 16, and the corrected cycle exposure value for Case 22.

Any questions on this submittal should be directed to Mr. R. Sgarro at (215) 770-7916.

Very tr ly yours, H. W. Keiser .

Vice President - Nuclear Operations Attachment cc: NRC Document Control Desk (original)

NRC Region I Mr. F. I. Young, NRC Resident Inspector SSES

@fr~~>>H~ C;- Thadani, NRC Prospect, Manager - Bethesda

Cross section dependencies include:

fuel exposure void history (i.e., exposure-weighted relative moderator density) relative moderator density (hot only) control rod presence fuel temperature (hot only) control rod history xenon concentration moderator temperature (cold only)

The effect of each dependency is calculated utilizing CPM-2. The final cross section data tables are prepared for SIMULATE-E using NORGE-B2 (Reference 20).

The radial, top, and bottom reflector regions are not modeled explicitly.

Instead, these regions are taken into account by use of albedo boundary conditions. Radial albedos are calculated using the ABLE (Reference 21) program developed by Science Applications International for EPRI. The top and bottom albedos were determined based on comparison to plant data during model normalization. Different albedo boundary conditions are used for cold 'and hot conditions.

Several of the input data parameters used by SIMULATE-E require adjustment to match plant operating data. This normalization process was performed using Susquehanna SES Unit 1 Cycles 1 and 2 data. All parameters changed in this fashion were held constant for all other calculations including the Quad Cities and Peach Bottom calculations.

The thermal hydraulics calculations use the FIBWR methodology (Reference 19) developed by Yankee Atomic Electric Company. This calculation determines total core pressure drop and core bypass flow. The pressure drop calculation determines the frictional pressure drop, local (i.e., form) losses, acceleration (i.e., momentum change) pressure drop, and elevation head. The core bypass flow calculation allows for modeling the flow paths shown in Figure 3.1.1. FIBWR as a stand-alone code has been benchmarked by Yankee Atomic Electric Company against data from Vermont Yankee and the Frigg Loop tests (see Reference 22).

- 51 "

I TABLE 3.2.3 EHAWA S S HOT CRI ICAL CORE K-EFFECTIVE DATA UNIT*1 CYCLE=1 CYCLE CORE AVERAGE PERCENT TOTAL CORE SUB- OOHE CONTROL ROD CALCULATED EXPOSURE EXPOSURE POWER POHER FLOW COOLING PRESSURE OENSITY CORE CASE ( GHD/NTU ) tGHD/HTU) (t%PH) (%) (%) l BTU/LBN) (PSIA) 0%) K-EFFECTIVE 1 O.ZRl O.R21 1432 43 54 23.8 974 20.4 0.99184 2 0.836 0.836 3250 99 98 23.7 1001 12.6 0.99142 3 1,490 1.490 3280 100 100 ~ 23.6 1005 13.9 0.98987 1.596 1.596 3278 100 88 23.6 1002 13. 6 0.98665 5 1.736 1,736 3291 100 97 24.3 1001 14.0 0.98919 6 1.758 1. 758 3296 100 98 24.R 1001 14.1 0.98886 7 1 ~ 799 1.799 3291 100 99 23.8 1001 14,1 0.98938 8 1.908 1. 908 3293 100 98 24.0 1000 14. 1 0.98960 9 2.070 2.070 3293 100 97 24.2 994 14. 1 0.98884 2.706 ll 10 1R 2.906 R.975 2.706 2.906 R.975 3281 3289 3291 100 100 100 98 97 25.0 24.2 1000 999 999 14.8 15.0 15.0 0.98937 0.98990 0.98988 13 3.116 3.116 3291 100 96 24.7 999 15.0 0.99009 14 3 '67 3.367 3292 100 98 24.2 1009 15. 9,. 0.98971 15 3.517 3.517 3289 100 98 R4.R 1002 15. 9 0.99020 16 3.663 3.663 3292 100 96 24.5 1002 15. 9 0.99042 17 3.776 3.776 3290 100 96 24.6 1001 15. 9 0.99058 18 3.836 3.836 3293 100 95 24.8 1003 15. 9 0.99061 19 3.918 3.918 3298 100 98 24.0 1000 16.0 Oo99080 20 4.036 4.036 3290 100 97 24.3 1003 16.0 0.99100 Zl 4.193 4.193 3R90 100 96 24.5 1002 16.0 0.99116 22 4.31S 4.318 3296 100 98 24.2 1003 16.1 0.99138 R3 4.506 4.506 3288 100 96 24.5 1003 16.1 0.99163 R4 4.517 4.517 3289 100 97 1004 16.1 0.99176 R5 5.061 5.061 3290 100 99 23.8 1005 17.6 0.99254 26 5.070 5.070 3288 100 99 23.9 1005 17. 6 0.99242 27 5.347 5.347 3281 100 97 24.3 100R 18.0 0. 99219 28 5.410 5.410 3Z94 100 98 24.0 1002 17. 9 0.99267 29 5.463 5.463 3291 100 99 23.S 1002 17.9 0.99294 30 5.580 5.580 3294 100 99 23.7 1002 17.8 0.99350 31 5.614 5.614 3295 100 99 23.9 1002 17 ' 0.99358 32 5.650 5.650 3287 100 99 23.8 1002 17.8 0.99367 33 5.855 5.855 3293 100 99 23.7 1001 17.0 0.9936R 34 5.918 5.918 3289 100 98 24.1 1001 16.7 0.99362 35 6.087 6.087 3286 100 96 24.3 1000 16.4 0.99430 36 6.241 6. 241 3288 100 98 23.9 1001 16.4 0.99437 37 6.436 6.436 3265 99 96 24.3 999 16.3 0.99454 38 6,563 6.563 3286 100 99 23.8 999 16.3 0.99463 39 6.716 6.716 3283 100 98 R4.1 999 15.0 0.99460 40 6.723 6.723 3290 100 98 Z4.0 999 15.0 0.99460

E I

uniform lattice critical experiments and pin power distribution calculations with CPM-2.

In conclusion, the analysis results contained in this topical report demonstrate PPGL's qualifications to perform steady state core physics calculations. Extensive comparisons to measured data from Susquehanna SES, Quad Cities Unit 1, and Peach Bottom Unit 2 demonstrate the validity of the analytical methods as well as PPaL's capability to set up and properly apply the models. Comparisons to reactor designs other than PP&L's Susquehanna SES demonstrate PPaL's ability to extend the core modeling techniques developed for Susquehanna SES to other fuel and core designs.

PP&L is committed to maintaining a strong in-house core analysis capability and as part of that commitment we continually evaluate the accuracy of our core simulation methods and make modeling improvements when appropriate.

Although PPaL's day-to-day core follow analyses are aimed primarily at plant operations support, the comparisons of SIMULATE-E calculations (e.g., TIP response, K-effective, thermal margins) to the plant data also serve as a continuing methods benchmarking effort.

208-

I l

CPM-2 Question 1 What are the bases for the depletion steps, spatial mesh, energy groups (macro and 2-D), convergence and other parameters used in production calculations with CPM-2/MICBURN?

~Res ense There are currently no specific EPRI guidelines available for development of MICBURN and CPM-2 input. The computer codes, however, have certain default settings with regard to iteration control, convergence accuracy, and energy group structure which were set by EPRI during the code development. These default values were used by PP&L for all calculations presented in PL-NF-87-001. No problems resulted from the use of the default iteration control inputs; the convergence criteria on the fundamental mode calculation is 1.0 x 10 -5 which is sufficient to provide consistent and accurate results.

Information on the energy group structure is presented in the response to Question 2.

The depletion step size used for the MICBURN calculations is set according to EPRI recommendations. These step sizes have been designed to limit the maximum gadolinia depletion to less than 4% of the initial amount for any given depletion interval. The depletion calculations are performed using 66 to 72 depletion steps.

The depletion step size in CPM-2 is set to provide smoothly varying cross section curves and lattice reactivity (see Figures C1.1 and C1.2 for examples). The timestep structure which is usually used for CPM-2 depletion calculations is:

0 ~ OR 0 1R 0 SR 1 ~ OR 1 SR 2 ~ OR 2 SR 3 ~ OR 3 ~ SR 4 ~ OR 4 SR 5 ~ OR 5 SR 6 .OR 6 ~ 5R 7 ~ OR 7 ~ 5R 8 ~ OR 8 ~ SR 9 ~ OR 9 SR 10 OR 12 ~ Sg 15 ~ OR 17 ~ Sg 20 ~ OR 22 ~ SR 25 ~ OR 27 ~ SR 30 ~ OR 35 ~ OR 40 ~ OR 45 OR SO.OR 55.0 GWD/MTU

For assemblies where the. gadolinia concentration is higher than 4 w/o additional CPM-2 timesteps are placed between 10.0 and 12.5 GWD/MTU. To evaluate the effects of the control rod presence, relative moderator density, fuel temperature, etc., restart calculations are performed at certain exposure points. These points are chosen so that the change in the cross section due to the change in the independent parameter (i.e. control rod presence, etc.)

is smoothly varying (see Figure C1.3 for an example). This change in cross section, not the absolute 'cross section, is used by SIMULATE-E. Sensitivity calculations have also been performed by PPGL to determine the effect of much finer timesteps on the CPM-2 results. The lattice reactivity from these sensitivity studies differed from the production calculations (coarser timesteps) by less than 0.001 ~K.

The spatial mesh used in MICBURN is somewhat finer than those recommended by the code developer. A total of 20 burnup (micro) regions and 10 flux (macro) regions are. used in the burnable absorber cell. A micro-region is defined as a homogenized material zone. A macro-region is composed of one or more micro-regions and is used for calculation of the flux. Figure C1.4 shows an example. This definition of zones within the fuel pin provides sufficient detail to accurately model the "onion skin" type depletion of a gadolinia pin.

The mesh spacing used in CPM-2 for the X and Y directions includes two meshes per pin cell, one mesh in the fuel channel wall, and two meshes in the water gap (bypass region). Sensitivity studies have been performed by PPsL in which the number of meshes in the pin cell has been increased from two to three.

This increase resulted in a maximum change in lattice reactivity of 0.005 ~K; typically, differences are much less. Using .this cross section data in the S1MULATE-E model has shown very little effect on the core power distribution and core K-effective. Additional work sponsored by EPRX has also examined differences between use of two versus three mesh points per pin cell.

Although these cases were limited to sub-assemblies (i.e., 3x3 fuel rod arrays), the resulting differences were quite small (i.e., less than 0.005 ~K) for varying gadolinia loading and void content and support the use of two mesh points per pin cell.

The sensitivity studies discussed above have been run to determine the effects of selected code inputs. The topical report PL-NF-87-001 provides a benchmark of the CPM-2 code with the Susquehanna SES model inputs and consequently an estimate of the code/model uncertainty.

FIGURE C1.1 THERMAL ABSORPTION CROSS SECTION BUNDLE ENRICHMENT: 2.19 W/0 4 GD 5 0.068 0.064 0.062 0.060 t I 0.048 0.048 0.044 ----:. Legend 0% VOID HISTORY p42 .: X 40% VOID HISTORY 0 70/o VOID HISTORY 0.040 0 10 15 20 25 30 35 40 45 50

. EXPOSURE (GWD/MTU)

FIGURE Ci.2 FUEL K-INFINITYVS EXPOSURE BUNDLE ENRICHMENT: 2.19 W/0 4 GD 5 1.2 I

I- 0 O.S zI

~ ~ ~ a ~ ~ ~

0.8 Legend 0% VOID HISTORY 0.7 "":" X 40% VOID HISTORY CI 70% VOID HISTORY O.B 0 6 10 16 20 26 30 36 40 46 60 EXPOSURE {Gwl3/MTU)

FIGURE C1.3 CHANGE IN SIGMA A-2 DUE TO CONTROL PRESENCE BUNDLE ENRICHMENT: 2.19 W/0 4 GD 5 0.016 Legend 0% VOID HISTORY 0.014 X 40% VOID HISTORY 0 70% VOID HISTORY 0.013 0.012 0.011 0.010 0 10 15 20 26 30 36 40 50 EXPOSURE (GWD/MTU)

Figure Cl.4 Definition of macro reg ions.

The figure:shows a case with 20 micro regions and 6 macro regions in the BA- pin.

N ~g5+ N ~57 BA- fuel Can Moderator Buffer zone QNop QNop hNop I

BNop I I

BNqp I I

I I I I I r 20Micro regions I I I ( I I I Il I I 6 Macro regions (+4 macro regions outside the BA-pin.)

rb Source: E. Edenius and A. Ahlin, "MZCBURN Microscopic Burnup in Gadolinia Fuel Pins," Part IZ Chapter 7 of EPRI CCM-3, September l977.

uestion 2 The 5 energy groups used for the 2-D calculations are somewhat coarse. Please comment.

~Res ense The use of five energy groups in the CPM-2 calculation is sufficient to accurately perform the two-dimensional calculation principally due to the method used to determine the five group cross sections. The CPM-2 calculation starts with a 69 energy group cross section library which was developed for general LWR analysis. For each two-dimensional calculation performed by CPM-2, micro-group and.macro-group calculations are performed which account for both the flux spectrum and the material present in the assembly. The

'I micro-group calculation is performed in 69 energy groups for each unique type of pin cell. Up to six separate calculations are permitted. If more than six unique pin types exist within a fuel lattice, similar pins must be averaged together. This micro-group calculation provides a detailed flux spectrum but does not account for the specific location of the pin. An extra region is used around each pin cell which does account for the effects of the presence of the bypass region and channel wall. The detailed energy spectrum is used to collapse the cross section data to 25 energy groups.

The macro-group calculation is performed in 25 energy groups and is a one-dimensional radial calculation for a regionally homogenized assembly (See Figure C2.1). Each row of fuel pins/water rods occupies a separate annular region starting at the center of the assembly proceeding outward. The channel wall, outer water gap, and control rod (if present) occupy separate regions.

This calculation, therefore, accounts for the. relative location of each material within the assembly. This is particularly important for fuel pins adjacent to the water gap. The 25 energy group fluxes calculated for each region are used to collapse the 25 group cross sections down to five for use in the final two-dimensional calculation. Since the flux spectrum used for this collapsing calculation already has the geometric effects factored into it, these five groups provide an accurate basis for the two-dimensional calculation whereas five group cross section data collapsed directly from pin cell cases might not.

Figure C2.1 Example of geometry in macro group calculation 00 00 00 0 00 00 00 0 UQ> pin 0 0 00 0 00 8 PuO> - pin 0 00 0 0 00 0 00 00 00 0 00 00 00 Wide water gap Narrow water gap homogenized outer water control rod gap UQ2 PuO>

Pu02 box UO 2 box homogenized fuel outer water homogenized fuel layers gap layers inner water gap inner water gap Source: A. Ahlin and M. Edenius, "The Collision Probability Module EPRl-CPM," Part IZ Chapter 6 of EPRZ CCM-3, September 1977.

uestion 3 How was the conversion from calculated power to Ba-140 concentrations performed for the CPM<<2 rod-wise comparisons to the Quad Cities gamma-scan results?

~Res ense Using the decay/production equation, the Ba-140 concentration can be calculated as:

NB(t) = t

<f n [SB(t)- XNB(t)1 dt n-1 where N

B (t) = the Ba-140 concentration at time t, S

B (t) = the Ba-140 production rate at time t,

= the Ba-140 decay constant.

Integrating and assuming S (t) is constant over each timestep gives:

N B

(tn ) = B n + N B

(tn 1)

B n e

- X<T (2) where tn = the end of timestep n, tn-1 = the beginning of timestep n.

Assuming that the average energy per fission is relatively constant over the time interval, S (tn ) can be approximated as:

S (t) =CY P(t) (3) where C = a unit conversion constant, Y

e

= the effective Ba-140 yield, P = the power density.

The Ba-140 concentrations are calculated by substituting Equation (3) into Equation (2) to obtain:

CY e

P(t n ) +

CY e

P(t n ) e X <T (4)

Since the final comparisons are made on a relative basis, relative Ba-140 concentrations are calculated as follows:

Rel. (t = 1 CY e

P(t n ) + CY e

P(t n ) -X~T N )

B tn-1 e (5) where N

B (tn ) = the average of the Ba-140 concentrations at the end of timestep n.

The power, P(t n ), is based on the CPM-2 relative pin power. The effective Yield, Ye', is calculated for each pin as follows:

M i = U-235, e

=

~ i i Y. F. U-238'u-239@ Pu 241 where Y = the Ba-140 yield for isotope i, i

F. = the fraction of fissions from isotope i.

The fission rates and hence the fraction of fissions from each isotope is calculated at each time tn by CPM-2. Equation (5) is solved to calculate relative N (tn ) for each rod by marching through the exposure points for each relative moderator density corresponding to 0%, 40%, and 70% void level. The SIMULME-E model calculation provides the exposure and void history for each axial plane for which measured data exists. These data are used to interpolate from the CPM-2 calculated data to determine the calculated relative Ba-140 distribution corresponding to the void history and exposure conditions at the location of interest.

guestion 4 The discussion in Section 2.3 needs more consistency, in references to measured and calculated values of power, and Ba-140 and La-140 activities in terms of what quantities are compared and their bases.

~Res oose The measured data used in the comparisons are the relative La-140 activities as reported in EPRI NP-214 "Gamma Scan Measurements at Quad Cities Nuclear Power Station Unit 1 Following Cycle 2" (Reference 12 in PL-NF-87-001). The calculated data used in the comparisons are the relative Ba-140 concentrations. These are derived from the CPM-2 calculated relative pin powers as presented in the, response to Question 3. The Ba-140 concentrations and activities are proportional to the La-140 concentrations and activities at any given time following shutdown from steady-state operation. The factor of proportionality significantly varies with time for the first week after shutdown, but after ten days, it remains essentially constant. Because all gamma scan measurements were taken following a shutdown period greater than ten days, the relative measured La-140 activities are compared to the calculated Ba-140 concentrations.

~nestion 5 Are the presently demonstrated accuracy and biases of CPM-2 calculations expected to hold for 9x9 and other advanced BWR bundle designs? Have any comparisons been made of CPM-2 to Monte Carlo calculations for 9x9 bundles of the type used in Susquehanna Unit 2? ~

~nes onse The accuracy and biases presented in PL-NF-87-001 are expected to hold for 9x9 and other advanced BWR bundle designs that. are similar to the 7x7, 8x8, and 9x9 fuel designs. Comparisons to the TRX, Kritz, and ESADA criticals show critical evaluations for a wide variety of fuel arrangements (i.e. varying pellet'diameters, pellet densities, water to metal ratios, and fuel rod pitches). Comparisons of CPM-2 to Monte Carlo calculations have not been made; however, the benchmarking presented in PL-NF-87-001 strongly supports the use of CPM-2 to model 9x9 fuel and other advanced BWR bundle designs similar to those presented.

uestion 6 Have any trends (biases) been observed in the accuracy of pin-power and LPF predictions vs. elevation, void history, exposure, control, etc.7

~Res ense The accuracy of the pin power distribution and local peaking factor does not appear to be correlated to exposure, void history, or elevation. This can be seen by examining the data from the Quad Cities gamma scan comparisons summarized .in Tables 2.3.2 and 2.3.3 of PL-NF-87-001. These data have been plotted in Figures C6.1 through C6.6. Overall, there does not appear to be any trend in the standard deviations of the pin comparisons relative to exposure, void history, or elevation. The interior mixed oxide bundles, GEB159 and GEB161, do show slightly increased standard deviations with increased elevations (i.e., void history). These bundle designs are not typical of expected fuel designs currently planned for use in Susquehanna SES.

It should also be noted that the calculated peak activity is normally high providing a conservative estimation of the local peaking factor.

In addition to the gamma scan comparisons performed at PPGL, EPRI sponsored benchmarking of the original EPRI-CPM. The results from these comparisons are consistent with the Quad Cities comparisons indicating CPM-2 calculations provide similar accuracy for different bundle designs.

Measured gamma scan data do not exist for any of the Susquehanna SES specific bundle designs which would permit direct comparison to pin powers. However, the TIP response comparisons presented in Section 3 of PL-NF-87-001 can be used to infer the accuracy of CPM-2. The TIP response model used in SIMULATE-E is developed based on CPM-2 calculations. These calculations require CPM-2 to predict a local fission rate at the detector location in the bypass region. If CPM-2 was unable to calculate accurate local peaking factors, it would also be unable to calculate accurate TIP response factors.

This would show up in the TIP response comparisons. The individual TIP response comparisons in Section 3 do not appear to contain any trends with control rod presence, exposure, void history (i.e., exposure-weighted relative moderator density), or relative moderator density. This agrees with the conclusions drawn from the comparisons to gamma scan data.

FIGURE C6.1 QUAD CITIES UNIT 1 END OF CYCLE 2 NORMALIZED La-140 ACTIVITY PIN COMPARISONS 0 GEB169 GEB161

~ GEH002 0

6 0 0":'egend '"

S CX0672

~ CX0214 I~

UJ Q

Cl CC CI cI Z 4 V) a ~ 4 6 8 10 12 14 16 18 20 22 CALCULATED BURNUP. (GWD/MTU)

mmmmmmmmmmwmwmmwm FIGURE C6.2 QUAD CITIES UNIT 1 END OF CYCLE 2 NORMALIZED La-140 ACTIVITY PIN COMPARISONS Legend P GEB169 GEB161

~ GEH002

~ CX0672 Z 6 ~ CX0214 0

l~

W a

Cl CC CI Z 4 I- p' 3 ""

~.:

~

0 10 20 . 30 40 60 60 70 CALCULATED VOID HISTORY (%)

FIGURE C6.3 QUAD CITIES UNIT 1 END OF CYCLE 2 NORMALIZED La-140 ACTIVITYPIN COMPARISONS Legend 0 GEB159 GEB161

~ GEH002

~ CX0672 8 ....~ CX0214 Z 0 0 0

l~

5 D

CL D

0 Z 4 (D

0 g I

~ ~

~ ~

20 40 60 80 100 120 140 ELEVATION (INCH)

W

FIGURE C6.4 QUAD CITIES UNIT 1 END OF CYCLE 2 PEAK La-140 ACTIVITYCOMPARISONS 10 Legend 0 GEB159 GE 8161

~ GEH002

~o 6 ~ "-.'- "" ~ CX0672 CX0214 0

c( 4 0

Q .~........:,

2 O

Z m.

-2 6 10 12 14 16 18 20 22 CALCULATED BURNUP (GWD/MTU)

FIGURE C6.5 QUAD CITIES UNIT 1 END OF CYCLE 2 PEAK La-140 ACTIVITY COMPARISONS 10 Legend 0 GEB159 GEB161

~ GEH002

~o 6 ~

r CX0872

~ CX0214 0

c( 4 0

UJ r

Q ~ ~

O O ~

' :r

..:..~............

Z g) 0-

<<2

-4 10 20 30 40 50 60 m m m m m m SPY'lY LJi iWekim m m m m m m

FIGURE C6.6 .

QUAD CITIES UNIT I END OF CYCLE 2 PEAK La-140 ACTIVITY COMPARISONS 10

~ ~ ~ ~ ~

~o 6 0 I 0 ~ ~

4 Cl 0 0 Cl CI Z

g) 0 Legend 0 GEB159 GEB161

---"-" ~ GEH002

~ CX0672 CX0214

-4 0 20 40 60 80 100 120 140 ELEVATION (INCH)

Question 7 How do the modifications to the ENDF/B-III nuclear data other than those noted for U-238 compare to the uncertainties in the basic data?

~Res ense The modification to the Pu-240 microscopic absorption cross sections is the only modification made to the ENDF/B-III cross section data other than those noted for U-238. This modification, as stated in Section 2.1 of PL-NF-87-001 and documented in Part II, Chapter 4 of EPRI CCM-3, "The EPRI-CPM Data Library," is a 50% reduction in the cross section, in the resonance energy region (i.e., energy groups 16 through 27). Although the accuracies of the ENDF/B-III data're not presented in the EPRI documentation, it is likely that this modification exceeds the uncertainties of the basic nuclear data. The modification, however, is required to compensate for the fact that Pu-240 is not treated as a resonance nuclide in CPM-2. The unmodified cross section would significantly overpredict the absorption in the resonance region.

Any modification to the Pu-240 microscopic absorption cross sections would affect the heavy nuclide concentration buildup with exposure. Table 2.1.3 of PL-NF-87-001 presents the heavy nuclide chains that include Pu-240. If the Pu-240 cross sections were inappropriately adjusted, the Pu-240, Pu-241, and Pu-242 concentrations would improperly accumulate with exposure. Table 2.4.3 and Figures 2.4.4 through 2.4.6 show comparisons of measured and calculated isotopic parameters. All calculations, which include the effect of the modified ENDF/B-III cross sections, show good agreement with measured data and provide indication that the concentrations are properly accumulating with exposure. This agreement therefore supports the acceptability of the modified Pu-240 microscopic absorption cross sections.

~tention 8 The Quad Cities-1 EOC2 gamma scan data are essentially representative of all rods out operation. What are the implications relative to the accuracy with which CPM-2 calculates individual rod powers for normal rodded conditions, and what assurance is there that any presently observed conservative trends (biases) are universal, and bounding?

~Res ense When performing safety analyses, generally only the limiting bundles are a concern. Therefore, it is normally only necessary to determine the uncertainty for uncontrolled conditions. The uncertainty is calculated from the Quad Cities Unit 1 end of Cycle 2 gamma scan comparisons. This uncertainty, however, can also be extended to cover the controlled configuration. Section 3 of PL-NF-87-001 contains comparisons made to operating data using the SIMULATE-E code. The cross section data and TIP response model are derived from CPM-2 calculated data. The results in Section 3, particularly the individual TIP response, do not show any increase in the standard deviation associated with the presence of a control rod (see PL-NF-87-001, Figures 3.2.15 and 3.2.36 for examples).

Reactivity comparisons from Susquehanna SES and Quad Cities cold critical evaluations also support these observations. The cold K-effectives from the local criticals and the K-effectives from the in-sequence criticals at the same exposure are not significantly different even though the control rod density is 98% for the local criticals and 74% to 75% for the in-sequence criticals. The Susquehanna SES and Quad Cities cold critical data is contained in Table 3.2.6 and Table 3.3.1 of PL-NF-87-001, respectively.

uestion 9 The CPM-2 comparisons to the 7 -scan data are influenced by the accuracy of the SIMULATE-E predictions of local effects (e.g. burnup, void, control history) for the scanned bundles/elevations. Have the SIMULATE-E local errors been considered to assure that the CPM-2 results are representative?

~Res ense When performing licensing calculations with SIMULATE-E, the local peaking factor which will be used for calculation of MCPR or LHGR will depend on the ability .of SIMULATE-E to predict nodal conditions. If the predicted conditions are incorrect, the calculated local peaking factor will be affected. The comparisons reported in PL-NF-87-001 include any additional uncertainties caused by the misprediction of the burnup or void history at the elevation of interest. These uncertainties will be taken into account in analyses which use SIMULATE-E to determine local peaking factor. The application of model uncertainties will be presented in detail in a topical report entitled "Application of Reactor Analysis Methods for BWR Design and Analysis".

SIMULATE-E Question 1 Does the data for assembly power peaking that is used in the calculation of fuel performance parameters (e.g. MLHGR, CPR) include all CPM-2 calculated statepoints (e.g. every burnup point and every nominal and off-nominal condition) or only a subset? If the latter, how are they selected to ensure conservatism?

~Res ense The CPM-2 based local and secondary peaking factors, which are required for the XN-3 critical power correlation, are used in the SIMULATE-E fuel thermal margin calculations. These peaking factors in SIMULATE-E are functions of void history (i.e., exposure-weighted relative moderator density), control rod presence, and fuel exposure. Although these peaking factors do not include some of the CPM-2 exposure state points and do not include a relative moderator density or control rod history dependence, the peaking factors are accurately represented in SIMULATE-E for all expected conditions. The peaking factors are not sensitive to these exclusions. Figure Sl.l'hows the local peaking factor values at 0% void history for three relative moderator densities (corresponding to 0%, 40%, and 70% void levels) and control rod history. The SIMULATE-E data agree well with all the CPM-2 data except for control rod histories past 5.0 GWD/MTU. Fuel assemblies with control rod histories approaching 5.0 GWD/MTU would have relatively low reactivity and would have significant margin to thermal limits. Since relative moderator density negligibly affects the local peaking factors as shown in Figure Sl.l and control rod histories for limiting bundles are less than 5.0 GWD/MTU, the effect of not considering these dependencies is insignificant.

FIGURE 81.1 UNCONTROLLED LOCAL PEAKING FACTOR DATA FOR 9XQ LATTICE AT 0'/0 VOID HISTORY 1.7 Legend 1.6 SIMULATE -E X CPM-2 0% VH 0 CPM-2 0% VH 1.5 TO 40V O

I- 0 CPM-2 0% VH O TO 70V 1.4 CPM-2 CONTROL HISTORY z

hC 13 C3 0 1.2 10 1e 20 26 30 40

uestion 2

a. What is the "flag" which signals the need for new normalization of the model adjustable input data parameters and/or radial and axial albedos?

b. How often are albedo/normalization parameter changes typically made?

c. What is the basis for performing the normalization when the code is used in a predictive mode for cores which differ significantly from those previously modeled?

~Res oese

a. Three major changes can affect the normalization parameters. A significant change in fuel design, core design, and/or calculational uncertainty will indicate that a new normalization should be performed.

In benchmarking the 7x7, 8x8, and 9x9 fuel bundle designs and the Quad Cities, Peach Bottom, and Susquehanna core designs, PPGL used the same set of normalization parameters for the various fuel and core designs. The benchmarking calculations completed to date show similar results between measured and calculated parameters and, therefore, support the use of the same normalization parameters for future Susquehanna SES fuel and core designs.

b. The adjustable albedo/normalization parameters have maintained consistency for all fuel and core designs as stated in response to Question 2a.

Changes have not been made and are not planned or expected to occur frequently. Future model enhancements may involve a change(s) in albedo/normalization parameters. For changes in models like this, benchmarking calculations would be performed to requalify or update the uncertainties in core reactivity and power distribution.

c. The presented TIP instrument response and core reactivity comparisons in PL-NF-87-001 are based on a consistent set of normalization parameters.

Using these comprehensive data that include a wide variety of fuel and core designs, PPGL developed a strong statistical data base to determine conservative margins for application to new core designs. The report

entitled "Application of Reactor Analysis Methods for BWR Design and Analysis" will present the use of these margins in Susquehanna SES safety analyses.

~uestion 3 Is the XN-3 correlation valid for 9x9 and other advanced design BWR bundles?

~Res onse The XN-3 correlation, developed by Advanced Nuclear Fuels Corporation (ANF),

formerly Exxon Nuclear Company, is valid for Sx8 and 9x9 fuel for the ranges of applicability specified in the associated Nuclear Regulatory Commission safety evaluations. Licensing Topical Report, XN-NF-734 (P) (A), "Confirmation of the XN-3 Critical Power Correlation for 9x9 Fuel Assemblies" describes the confirmation of XN-3 for the 9x9 fuel bundle design and is approved by the.

Nuclear Regulatory Commission. The original approval of the XN-3 Critical Power Correlation is provided in XN-NF-512(P)(A), "XN-3 Critical Power Correlation". This XN-3 correlation is used in the SIMULATE-E fuel performance evaluations. Sample SIMULATE-E test cases have been performed and documented to verify the correct implementation of the correlation. The XN-3 correlation is valid for the fuel bundle designs currently scheduled for

.loading into future Susquehanna SES cycles (i.e., 8x8 and 9x9 fuel bundle designs).

uestion 4 The TIP detector model in SIMULATE assumes that the response from each assembly is not affected by the presence of the other 3 surrounding the TIP.

Has this assumption been tested; is it adequate'Res ense PL-NF-87-001 states that the detector response from each assembly (i.e., R.)

j is not affected by the other three. However, the total detector response considers the effect of each supporting assembly power as follows:

M ER = R. P. (Section 3.2.3 of PL-NF-87-001)

M j j where ER = total detector response, M = number of bundles around a TIP detector (i.e. M=4),

R. = detector response contribution from assembly, j j, P. = SIMULATE-E calculated nodal power from assembly, j.

Each assembly power, P., is affected by the others through neutronic coupling the neutron balance equation. Therefore, the total detector response j'n 1

contribution from an assembly, R. P., implicitly takes into account the j

assembly powers. This methodology is validated through the TIP response j'ther comparisons presented in Section 3 of PL-NF-87-001. For example, Figures 3.2.12 and 3.2.15 in PL-NF-87-001 show three controlled and one uncontrolled TIP response comparisons. The controlled comparisons contain aggravated conditions of which one bundle is low in power and the other three are high in power. Major discrepancies would exist if the TIP response methodology is inadequate. As the figures show, excellent agreement for all three controlled TIP responses exist, and the results are very similar to the uncontrolled TIP response comparisons. This excellent agreement supports the TIP response model used in SIMULATE-E.

question 5 While it is true that peripheral assemblies and top and bottom axial nodes are generally low power and hence not of safety concern, eliminating them from the scan comparisons seems to remove a potentially valuable source of information on the accuracy/adequacy of albedo and reflector boundary condition dependencies. Please comment.

~Res onse The gamma scan data in EPRI NP-214 allow for radial, nodal, peak to average, and bundle (axial) comparisons. Peak to average and individual bundle (axial) comparisons utilize all the available gamma scan data. The radial comparisons utilize all the data with the exception of the mixed oxide and peripheral assembly data. For the nodal comparisons, the mixed oxide and peripheral assembly and top and bottom node data are eliminated. Table 3.3.7 of PL-NF-87-001 presents the individual bundle gamma scan comparisons for all bundles and nodes. The peripheral bundles in this table are:

CX0546 g GEB162 g CX0490 g CX0553 f CX0662 g CX0682 g CX0643 g CX0683 ~

Figure 3 '.20 of PL-NF-87-001 shows an axial comparison of a peripheral bundle. It is recognized that these comparisons directly assess the accuracy of the albedos used in SIMULATE-E. Comparisons of the peripheral assemblies and top and bottom axial node gamma scan results are slightly worse than the interior bundle gamma scan comparisons but are still quite good. The top and bottom albedos which are based on the Susquehanna SES data were used in the Quad Cities model. Due to different fuel and core designs, the top and bottom albedos would differ from the Susquehanna SES values. Although the Susquehanna SES albedos were utilized in the Quad Cities calculations, the SIMULATE-E model still provides an accurate calculation of the power distribution. Therefore, since the PPGL models were not normalized to the Quad Cities data and since the top and bottom nodes and peripheral bundles are low power regions of the core, the peripheral bundles were not included in the standard deviation calculation for the radial comparisons, and the peripheral bundles and top and bottom nodes were not included in the standard deviation calculation for the nodal comparisons.

uestion 6 Please explain why non-conventional definitions are used in the TIP and g -scan comparisons. For example, it is not obvious why T is used in the denominator for determining the differences in the radial TIP comparisons.

~Res oese In PL-NF-87-001, the differences and standard deviations for the TIP response and gamma scan comparisons are normalized with the average measured value, T, to express them in terms of a percentage of the core average. This approach results in a standard deviation expressed in units of percent. However, the result of the calculation is a standard deviation of the absolute differences.

Another method that could have been used involves conversion of the differences to a percentage of the measured value (i.e., by dividing by T ),

and then calculate the standard deviation of these percentage differences.

This second method, however, weights the differences for the detector locations with low readings (i.e., low power regions) more and the differences for detector locations with high readings (i.e., high power regions) less than the first method. Since the accuracy of the SIMULATE-E calculations in the high power regions is more important for thermal margin calculations, the first method is more appropriate.

An example is shown for a radial TIP response comparison to demonstrate the differences in the above approaches. The attached Figure S6.1 shows a radial TIP response comparison using the second method and Figure 3.2.29 of PL-NF-87-001 shows the radial TIP response comparison of the same data using the first method. Figure S6.2 shows the average of the measured TIP responses at each radial location (i.e., T ). Note that the use of the first method results in a higher 0 difference for the high measured values (e.g., TIP response at location 32-33 : 6e48% vs. 5.96%), and a lower % difference for the low measured values (e.g., TIP response at location 32-57 : 5.79% vs.

6.36%). For all the TIP response and gamma scan comparisons based on the first method, the location of the worse prediction (i.e. largest absolute difference) can be easily determined by finding the highest percent difference. Using the second method, the worse prediction is not necessarily at the location of highest percent difference. This is indicated in the

example where the TIP response calculation at 32-33 (i.e., a high power region) exhibits the worse absolute difference. The second method suggests that the TIP response calculation at 32-57 (i.e. a low power region) is worse.

FIGURE S6.1 SUSQUEHANNA SES UNIT 1 CYCLE 3 RADIAL TIP RESPONSE COMPARISONS 0.178 GWD/MTU CYCLE EXPOSURE 61 59 57 0.82 -0.14 6.36 -0.94 55 53 51 49 -2.03 -1.95 ~ 14 0.6 0.35 3.0 47 45 43 ++

41 39 37

-2.50

++ -0.75 3.4 -2.90 0.0 35 33 1.90 -0.28 -0.29 5.96 -2.80 0.43 31 29 27 ++ ++

25 23 21

++ 3.88 -4.33

+ +

1.26 .18 -3.29 2.2 19 17 1.0 -1.49 -1.53 2.33 3.39 -4.82 -0.77 15 13

+++

++ .13 -1.14 -1.86 3

1 Y

00 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 X

Diff = [(Calc - Mess)/Measured] X 100%

FIGURE S6.2 SUSQUEHANNA SES UNIT 1 CYCLE 3 MEASURED RADIAL TIP RESPONSE 0.178 GWD/MTU CYCLE EXPOSURE 61 59 57 30.77 44.60 42.58 41.57 55

,53 51 49 34.24 49 94 52.47 50.22 50.92 43.79 47 45 43 41 50 .19 48. 06 48.30 45.27 46.65 52.55 40.97 39 37 35 33 49. 40 46. 89 51".71 50. 87 46.67 51.77 44,74 31 29 27 25 50. 48 46. 72 53.61 50 90 49.51 52. 92 43.19 23 21 19 17 43. 82 50. 22 49.34 45. 47.20 51. 41 14 15 13 11 9 43. 62 52.12 51 19 50.01 34. 93 7

5 3

1 0002040608 10 12 14 16 18 20 22 24 26 283032 3436384042 44 46 4850 52 54565860 X

Core Average TIP Response = 46.82

uestion 7 Does PPaL intend to use PDQ-7 for applications significantly different from those for which benchmarking is provided in the report (e.g. core calculations)7

~Res ense PPaL does not intend to perform three-dimensional core statepoint or depletion calculations with PDQ-7. PPGL's primary intent i:s to use PDQ-7 for two-dimensional calculations to complement CPM-2 and/or SIMULATE-E for special applications (e.g., partially loaded core configurations and local criticality calculations). In some instances, PDQ-7 will be used as an independent verification of calculations. In addition, PPSL believes that future SIMULATE-E model improvements may be developed with the use of PDQ-7.

question 8 Do EPRI guidelines exist for the CPM-2 (cross section) COPHIN - PDQ-7 calculational path7 Are they followed by PPaL'?

~Res ense No EPRI guidelines currently exist for the CPM-2/COPHIN/PDQ-7 calculational path. The method used at PPGL is to use the CPM-2 macroscopic cross section data for fuel pins in the assemblies of interest. COPHIN assembles this data into cross section tables which are then used in PDQ-7. PPGL only uses PDQ-7 for special analyses that cannot be performed with CPM-2 and/or SIMULATE-E.

Each specific analysis will determine the particular manner in which the PDQ-7 model is developed.

GENERAL

/egestion Have the CPM-2/MICBURN, SIMULATE-E, FIBWR and the XN-3 correlation been reviewed and approved by the U.S. Nuclear Regulatory Commission?

~nes ense The neutronic methodology in CPM-2/MICBURN and SIMULATE-E, the thermal hydraulic methodology in SIMULATE-E (i.e., FIBWR), and the critical power methodology in SIMULATE-E (i.e., XN-3) have been reviewed and approved by the U.S. Nuclear Regulatory Commission as part of other topical reports. The neutronic methodology in CPM-2/MICBURN has been recently approved in the General Public Utilities Nuclear Corporation submittal of their lattice physics topical report. The SIMULATE-E methodology for the neutronic calculations, has also been approved in Yankee Atomic Electric Company's submittal of SIMULATE. The SIMULATE and SIMULATE-E neutronic methodologies are identical. With regard to the thermal-hydraulic methodology in SIMULATE-E (1 e., FIBWR), the FIBWR methodology has been approved for Yankee Atomic Electric Company. For XN-3, the U.S. NRC has approved the Exxon Nuclear Company (currently Advanced Nuclear Fuels) submittals XN-NF-512(P)(A) and XN-NF-734(P)(A)-

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