ML20138J791

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Rev 0 to Methods for Analysis of BWRs Lattice Physics
ML20138J791
Person / Time
Site: Oyster Creek
Issue date: 11/13/1985
From: Bond G, Fu H, Furia R
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20138J788 List:
References
TR-020, TR-020-R00, TR-20, TR-20-R, NUDOCS 8512180041
Download: ML20138J791 (83)
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Text

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i Page 1 TR 020 l-i I.

I i Methods for the Analysis of Boiling Water Reactors Lattice Physics TR 020 -

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(REV. 0) l l

BA NO: 335430 H. Fu .

R. V. Furia AUTHORS DATE July 25. 1985 APPROVALS:

  • 0 /l-AT ad.C NUCLEAR ANALYSIS &'" FUELS' DIRECTOR DATE GPU Nuclear 100 Interpace Parkway Parsippany, New Jersey 07054 8512180041851gg9 pog ADOCK 050 pg P

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Page 2 ABSTRACT This report describes the computer code, CPM, used at GPUN for BWR fuel lat-tice calculations. It demonstrates GPUN's lattice physics modeling capabil-ity by comparing the CPM calculated results with plant measured data and higher order calculations.

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Page 3 Acknowledgement The authors would like to express their appreciation to Mr. M. Bcg and Mr.

C. B. Mehta for their assistance in the benchmarking of Haten Cycle 1 gamma scan data. The authors also would like to thank Electric Power Research Institute and Exxon Nuclear Corporation for permitting us to use the results in their published reports.

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Page 4 Table of Contents Section Page 1.0 Introduction 8 2.0 Description of CPM 10 2.1 Introduction 10 2.2 Nuclear Data Library 14 2.3 Calculations on Unit Cells 15 2.3.1 The Resonance Treatment 15 2.3.2 The Micro Group Calculation 16 2.3.3 The Macro Group Calculation 18 2.4 Treatment of Cadolinia in Fuel 19 2.5 Treatment of Cruciform Control Rod 21 2.6 The Two Dimensional Calculation 21 2.7 The Burnup Calculation 23 3.0 Verification 38 3.1 GPUN Verification 38 3.1.1 Comparison with Oyster Creek Pin Gamma Scan 38 3.1.2 Comparison with Monte Carlo Calculations 40 3.1.3 Comparison with Hatch-1 Pin Gamma Scan 40 3.2 EPRI Benchmarking 42 3.2.1 Hot Critical Experiments in KRITZ 42 3.2.2 Cold Critical on Uniform Lattice 43 3.2.3 Yankee and Saxton Isotopics 43 4.0 Summary 79 5.0 References 81 Total 6a

a Page 5 List of Tables Table Title Page 2.1 Energy Boundaries in the 69 Group Library 26 2.2 Table of Nuclides 27 3.1 Summary of CPM Calculated and Oyster Creek La-140 Gamma Scan Measurement Comparison 45 l

l 3.2 Comparison of CPM Calculated and Monte Carlo Calculated K-Infinity 46 3.3 Summary of CPM Calculated and Hatch-1 La-140 Gamma Scan Measurement Comparison 47 3.4 CPM Results for TRX Oriticals 48 3.5 CPM Results for ESADA Criticals 49 3.6 Isotopic composition in Saxton: Comparison between CPM and Experiment 50 l

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Page 6 List of Figures Figure Title Page 2.1 Flow Diagram of CPM 29 2.2 Flowchart of the Resonance Calculation 30 2.3 ' Flowchart of Micro-Group Calculation 31 2.4 Example of Geometry in Macro-Group Calculation 32 2.5 Flowchart of MICBURN 33 2.6a Actual Control Rod Configuration 34 2.6b Representation of Control Rod Wing In CPM 34 2.7 Example of BWR Cell Geometry in the 2D Calculation 35 2.8 Heavy Nuclide Chains 36 2.9 Fission Product Chains .

37 3.1-3.2 Comparison of CPM Calculated and Measured Pin Power 51-52 Distribution for Oyster Creek, Axially Averaged 1976 La-140 Gamma Scan 3.3-3.4 Comparison of CPM Calculated and Measured Pin Power 53-54 Distribution for Oyster Creek, Axially Averaged 1977 La-140 Gamma Scan 3.5-3.9 Comparison of CPM Calculated and Monte Carlo 55-59 Calculated Pin Power Distribution 3.10-3.21 Comparison of CPM Calculated and Hatch-1 La-140 60-71 Gamma Scan Measured Pin Power Distribution 3.22. Fission Rate Distribution in an 8x8 BWR Box of the Pu Island Type T=245'c 72 3.23- Fission Rate Distribution in a 15x15 PWR MO: Assembly 73 with Water Holes and Absorber Rods, T=245'C b

- . . = . = __ ,

Page 7 List of Figures (Cont'd)

Figure Title Page 3.24 Fission Rate Distribution in a 14x14 PWR MO: Assembly 74 Surrounded by UO Assemblies, T=240'c 3.25 CPM Calculated K-Effective vs. Measured Buckling 75 3.26 Yankee. The Isotopic Ratio Pu-239/Pu-240 76 Comparison Between EPRI-CPM and Experiment 3.27 Yankee. The Isotopic Ratio Pu-241/Pu-242 77 Comparison Between EPRI-CPM and Experiment 3.28 Yankee. The Isotopic Ratio Pu-240/Pu-241 78 Comparison Between EPRI-CPM and Experiment t

, + , - - - . , - , ., - , - , . . , , , . . , ,,. , - - . ~ - -

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Page 8 1.0 Introduction This report describes the lattice physics method in use at GPUN.

The method utilizes the computer code, CPM (Reference 1), which is a multigroup two-dimensional collison probability code for burnup calculations on BWR and PWR assemblies. The CPM code was de-veloped by EPRI and is included in the EPRI Advanced Recycle Meth-odology Program. It has been benchmarked with experimental re-suits from critical facilities and power reactors. The benchmark work showed good agreement between the code calculated values and the experimental results. Details of the comparisons are re-ported in Reference 2. Because of its accuracy, CPM has been used as an analytical benchmarking tool for other lattice physics codes.

Additional work has been performed at GPUN to demonstrate the ac-curacy of the code and GPUN's capability to model fuel lattices using CPM. The CPM calculated pin power distributions are com-pared with the Oyster Creek gamma scan data and Monte Carlo calcu-lations (References 3-6). The CPM calculated critical reactivity and pin power distribution for an unexposed bundle are compared to those obtained from Monte Carlo calculations. The CPM calculated pin power distributions for an exposed bundle are compared to gamma scan data at Oyster Creek and Hatch-1. All these comparisons show

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Page 9 l excellent agreement and form the basis of confidence for the GPUN lattice physics method. The measured gamma scan data have not been corrected for core flux tilt effects caused by control rod and fuel exposure. Therefore, an apparent power shift is observed between the measured and calculated pin power distribution. This is a common phenomenon when modelling a finite 2-dimensional lattice with an infinite lattice. Because the lattice physics results are used to provide inputs to the 3-dimensional reactor simulator model, the adequacy of the lattice physics model is further demonstrated by the agreement between core simulation results and plant operating data. This comparison is provided in ,

the GPUN report on its 3-D steady state core simulation model (Reference 14).

Section 2 of this report briefly describes the methodology of

. CPM. The verification work performed by GPUN and EPRI are ex-plained in Section 3. The description of EPRI's work is taken J

from References 1 and 2. The summary of GPUN's lattice physics model and its basis of confidence are given in Section 4.

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Page 10 2.0 Description of CPM The contents of this section are condensed from EPRI CPM manual (Reference 1).

-2.1 Introduction to CPM CPM is a multigroup two-dimensional collision probability code for burnup calculations on BWR and PWR assemb. . The code handles a geometry consisting of cylindrical fuel rods of varying composition in a square pitch array with allowance for fuel rods loaded with gad-olinium, burnable absorber rods, cluster control rods, in-core in-strument channels, water gaps, boron steel curtains and cruciform control rods in the regions asparating fuel assemblies. EPRI-CPM was originally developed by AB Atomenergi and later was modified and in-cluded in the EPRI Advanced Recycle Methodology Program.

The flow of calculations in the CPM program (EPRI-CPM) is shown in the simplified diagram illustrated in Figure 2.1. In the first part mac-roscopic group cross sections are prepared for the succeeding micro group calculations. A 69 group data library, essentially based on ENDF/B-III, is an integer part of-the code and most macroscopic cross sections are directly calculated from the' densities, geometries etc.

given in the input. The effective cross sections in the resonance region for the important resonance absorbers are calculated in a special module. Use is made of the equivalence theorem which relates tabulated effective resonance integrals for each resonance absorber in each' resonance group to the particular heterogeneous problem. The ,

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( Page 11 equivalence expression is derived from suitably defined rational ap-proximations for the fuel self collision probability. The resonance integrals obtained from the equivalence theorem are used to calculate effective absorption and fission cross sections. The screening ef-fact between different pins is considered by the use of Dancoff fac-i tors.

The resonance region is defined to lie between 4 eV and 9118 eV.

Resonance absorption above 9118 eV is assumed to be unshielded. The 1 eV resonance in Pu-240 and the 0.3 eV resonance in Pu-239 are ade-quately covered by the concentration of thermal groups around these resonances and are consequently excluded from the special resonance treatment.- Four nuclides, namely U-235, U-236, U-238 and Pu-239, are treated as resonance absorbers.

The cross sections thus prepared are used in a series of micro-group t j calculations to obtain detailed neutron energy spectra to be used for energy condensation and spatial homogenization of elementary. pin c

cells. The method of collision probabilities is used and applied to a simplified cluster geometry usually consisting of three or four regions.

4 The micro-group calculation is fast and is repeated for each type of pin in the assembly, so that individual spectra are obtained for pins containing fuel of different enrichment. To provide micro-group spectra for condensation and homogenization of an absorber pin cell, a micro-group calculation is carried out in four or five annular

Page 12 regions. The innermost one is the absorber material. It is sur-rounded by the canning and water and the last region represents the surrounding fuel pins-by use of. smeared out cross sections, which are provided by the previous micro-group calculations on fuel pins. An analogous procedure is used to determine micro-group spectra for water holes within the assembly.

When gadolinium is used as burnable absorber in fuel rods, the ini-tially homogeneously distributed Gd is burnt in a complicated way.

The microscopic burnup of Gd in fuel rods is calculated outside CPM 4 by the MICBURN code (Reference 7), which provides effective cross sections versus burnup for the Gd homogenized over the fuel rod. The j effective cross sections are used as input to CPM.

After the micro-group calculation either of two options may be chosen.

In one of them the micro-group spectra are directly used to obtain broad group cross sections for smeared pin cells for the succeeding two-dimensional calculations. In the other one, one-dimensional cal-culations on a cylindricalized assembly are performed. These macro group calculations are made in a maximum of 25 energy groups. This intermediate step before the two-dimensional calculation allows some consideration to be taken to the effect of the difference in water gap thickness on opposite sides of a BWR assembly. The solution from

' these. calculations gives new neutron fluxes to be used for the energy condensation of cross sections for the two-dimensional calculation.

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--,-e , c.w, --.c,- -c -e --em..y -~~..,w- --- e- p-- m-- -r m -- + ' * ' -

. Page 13 In case a cruciform control rod is present, smeared cross sections for the blades are generated by use of neutron fluxes obtained from a heterogeneous treatment of the geometry.

.The data generated in the previous steps constitute the input to the two-dimensional routine. This module which is based on an accurate collision probability method is one of the key modules in CPM. It gives the eigenvalue and the flux distribution in a BWR or a PWR as-sembly. The calculation is performed in a maximum of 12 energy groups for a geometry containing a maximum of 10x10 pin cells plus water gaps. Diagonal symmetry is required.

A fundamental buckling mode is used for modifying the results obtained from the transport calculation to include effects of leakage. This calculation can 'be carried out either in the diffusion theory approx-imation or by use of the B1-approximation.

The isotopic depletion as a function of irradiation is calculated for each fuel pin and for each region containing a burnable absorber.

The burnup chains, with the isotopes linked through absorption and decay, are linearized and 22 separate fission products, 2 pseudo fis-sion products and 14 heavy nuclides are treated.

A predictor-corrector approach is used for the burnup calculation.

For each burnup step the depletion is calculated twice, first using the spectra at the start of the step and then, after a new spectrum f

,- -n- _ . , - . . , , , , ,

Page 14 calculation, using the spectra at the end of the step. Average number densities from these two calculations are used as start values for the next burnup step.

The CPM code has been benchmarked against hot critical experiments in the KRITZ reactor, TRX and ESADA criticals on uniform lattice at room temperature, and post irradiation investigations of exposed fuel from the Yankee and Saxton reactors (Reference 2). The benchmark work shows good agreement between experimental results and dPM calculated results. Section 3.2 gives the details of the benchmark.

4 2. 2 - Nuclear Data Library The EPRI-CPM library contains microscopic cross sections in 69 energy groups for 66 elements. It is essentially generated from the ENDF/B-III file. The group boundaries are distributed in a manner shown in Table 2.1. There are 14 groups in the fast region (above 9 kev), 13 groups between 9 kev and 4 eV and 42 groups below 4 eV. The distribution in energy of these 42 groups contains--concentration of groups around the 0.3 eV resonance of Pu-239 and the 1 eV resonance of Pu-240.-

The nuclides available in the current library are listed in

Table 2.2. -Each nuclide is identified by a nuclide identification number. This number is in general chosen so that the first figures are equal to the ' atomic number of the nuclide. The last three figures are equal to the atomic weight of the nuclide if data are l given for a special isotope of the element, otherwise the last j l

figures are 000. j

Page 15 Table 2.2 lists the temperature at which cross sections are available. Data for other temperatures are determined by interpola-tion or extrapolation. Table 2.2 also indicates the nuclides with all relevant cross sections tabulated and the nuclides with only ab-sorption cross section tabulated.

The library is cosposed of six subfiles containing such quantities as group cross sections for different temperatures, resonance integrals for different potential scattering cross sections and temperatures and fission product yields and decay constants.

The library has undergone comprehensive tests. Group cross sections have been checked against various compilations and other sources and a number of criticality tests have been performed. The library used in EPRI-CPM is CPMLIB3 which modified the original 69 group library based on the experiences obtained in the benchmarking of CPM.

2.3 Calculations on Unit Cells 2.3.1 The Resonance Treatment The resonance energy region is defined to lie between 4 eV and 9118 eV. ' Resonance' absorption above 9118 eV is regarded as being un-shielded. The 1.0 eV resonance in Pu-240 and the 0.3 eV resonance in Pu-239 are adequately covered by the concentration of thermal groups around these resonances and are consequently excluded from the special resonance treatment. In the present version, four nuclides -

U-235, U-236, U-238 and U-239 - are treated as resonance absorbers.

l Page 16 The equivalence theorem is used to relate the heterogeneous problem to an equivalent homogeneous problem. Figure 2.2 shows a flowchart of the resonance calculation. It can be interpreted in the following way.

The fuel to fuel collision probability is expressed as a sum of rationals of effective potential'eross section. The resonance integral is determined from '

the fuel collision probability in accordance with the equivalence theorem. The effective resonance inte-grals are then obtained by interpolation from tables of homogeneous resonance integrals in the data library.- The homogeneous resonance integrals are tabulated with potential cross section and temperature as parameters. The interpolation is based on a T[Tandop I

dependence. A first order correction for the interaction associated with the presence of several nuclides in the same material is in-cluded. The spatial dependence of the resonance cross sections within the assembly is taken care of by Dancoff factors, which are calculated separ-(

ately for each fuel pin.

2.3.2 The Micro Group Calculation A-solution of the transport equation with 69 energy groups and many space points, simultaneously in space and energy, is not feasible

- - , . , - - . - --y

Page 17 because of computer limitations. Therefore, one has to separate par-tially the energy dependence from the space dependence. In CPM, the transport calculation starts with a number of microgroup calculations in the library group structures. A micro-group calculation is made for each type of pin in the fuel assembly. Control rod and boron curtains are neglected in the micro-group calculations. A flowchart is shown in Figure 2.3.

Each micro-group pin cell calculation provides a solution to the space - energy transport equation for isotopic scattering by means of collision probability calculations in 69 energy groups and in a sim-plified, cylindrical geometry consisting of four or five regions.

For fuel pin cells, four regions representing fuel, cladding, cool-ant,and the water gap and channel in a BWR cell. The absorber pin cell on water ho'le is survounded by a circular buffer zone, which is 2.5 mean free paths thick in the highest group below 0.625 eV. Data for the buffer zone are flux and volume averaged cross sections of the average pin cell.

The micro-group calculation provides 69 group spectra which are used for energy condensation and spatial homogenization of the elementary pin cells. Thus, broad group cross sections are determined for smeared pin cells. The micro-group calculation is fast and is re-peated for each different type of pin in the assembly, so that indiv-idual spectra are obtained for pins containing fuel of different ini-tial enrichment. During depletion, pins which were originally iden-tical will develop slightly different nuclide inventories. A single

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Page 18
average spectrum based on the average nuclide inventory will be cal-

, culated for this group of pins, but the macroscopic cross sections i

used for each pin in the two-dimensional calculation will reflect its own nuclide inventory.

2.3.3 The Macro-Group Calculation I

The macro-group calculation may be performed for pin cells or as an

intermediate step for BWR assemblies. The calculation is carried out in annular geometry. The user specifies the number of groups up to 25 groups and the group structures in the macro-group calculation.

Macroscopic cross sections are obtained by condensation and homogen-i ization of pin cells using the fluxes from the micro-group calcula-tions with flux and volume weighting. In pin cell geometry, no spa-tial smearing is necessary before the macro-group calculation, since i

) it only serves'the-purpose of giving a more detailed flux distribu-tion in the cell compared to the micro-group calculation.

l 'The two dimensional geometry of the fuel assembly is cylindricalized i to preserve the mean chord of the water gaps and the volumes of each

pin cell layer of the fuel zone. The macro-group calculation is made twice, first for the geometry with the wider water gap as the outer-most annulus and thereaf ter for the corresponding geometry with the j_ narrow water gap. An example of the geometry used is shown in Figure 2.4.

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Page 19 The macro-group calculation'is optional but is recommended for most geometries with water gaps. The calculations are fast and generally allow the' succeeding two dimensional calculation to be made in fewer energy groups than would be needed without the macro-group calcula-tion.

2.4 Treatment of Gadolinia in Fuel When gadolinium is used as burnable absorber in fuel rods, the gado-linium which is initially homogeneously distributed is burnt in a 1

, complicated way. The microscopic burnup-of gadolinium in fuel rods is calculated outside CPM by the code MICBURN (Reference 7), which provides effective cross sections, homogenized over the fuel rod,for the gadolinium as a function of burnup. The effective cross sec-tions are used as input.to CPM and the flux in a rod containing gado-linium is then calculated in the same way as for absorber rods.

A flowchart of MICBURN is shown in Figure 2.5. The geometry ant material composition are obtained from the card input, and microscopic cross sections are read from the CPM library. Macroscopic cross sec-tions are obtained in a straightforward way except for the heavy

</ nuclides for which a special' resonance treatment similar to that in the CPM program is used to calculate effective cross sections in the resonance groups.

The gadolinia pin is first divided into equidistant radial micro regions. Up to 100 regions may be used, but 20 is a typical number.

Each micro region defines a burnup region. The transport equation is l

-.-----,- - , = . , - . . . - - - . . . , .

Page 20

.olved using collision probabilities for a number of macro regions, each consisting of one or more micro regions. The boundaries and the number of the micro regions are the same at all burnup steps. The number of macro regions is also kept constant but the boundaries are changed automatically at each burnup step so that the macro regions are distributed in an efficient way for the transport calculation.

The gadolinia pin is surrounded by a buffer zone consisting of a uni-form pin cell lattice which is defined by the user. The main trans-port calculation is carried out in an annular geometry where the gad-olinia pin is surrounded by can, moderator and buffer zone. Before the main transport calculation can be carried out the buffer zone has to be homogenized. For this, a special transport calculation is made using collision probabilities in three regions, fuel, can and modera-tor, defining the uniform' lattice of the buffer zone. The flux dis-J tribution obtained from this " buffer zone calculation" is used to calculate cross sections for the homogenized buffer zone.

The main transport calculation is then carried out using integral transport theory. It is made in the library group structure using collision probabilities in the flat source approximation. The mesh is determined by the macro regions. At most 20 macro regions may be <

used.

In the burnup calculation, the whole gadolinium chain from Gd-154 to Gd-158 is taken into account. In addition, the burnup of heavy nuc-lides and buildup of fission products is calculated. The buffer zone T

T

Page 21 is-also burnt. At each burnup step effective cross sections for the gadolinia.are homogenized over the fuel rod, parameterized as a func-

{ tion ~of gadolinium' depletion, and written on a file for use in the main CPM calculation.

2.5- Treatment of cruciform Control Rod A cruciform control rod usually consists of a central steel region and four control rod wings. The. hub is homogeneous but the wings of the blade contain tubes of~ absorbing material enclosed within a per-forated metal sheet as shown in Figure 2.6a. Before the two dimen-sional calculation can be carried out on a cell including a control rod, homogenized cross sections must be obtained for these strongly 4

heterogeneous wings.

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,The wings are homogenized in such a way as to maintain their black-ness. The homogenization is performed on the assumption that the a

wing consists of two regions - the cylindrical absorber material and i

a homogeneous mixture'made up of the absorber tube walls, the sur-rounding metal sheet, and any water between the two. The relative flux' in the two regions shown in Figure 2.6b is obtained for each 2-D

-energy group by a collision probability calculation which includes the effect of an anisotropic surface current.

2.6 The Two-Dimensional Calculation A.two dimensional calculation is required to determine the flux,

-local power distribution, and bundle reactivity in a BWR or PWR assembly. For this calculation, CPM uses a multigroup integral

Page 22 transport routine based on the method of collision probabilities which are evaluated in two dimensional cartesian geometry.

Typical geometry of a BWR assembly treated by the two-dimensional routine is shown in Figure 2.7. In a BWR cell, the wing of the cruciform control rod is homogenized as described in Section 2.5.

The central steel region and end plates may be treated as separate regions. The box wall and the inner water gap is homogenized prior 3

to the two-dimensional calculation if requested by the user. The condensation of the 69 group spectra to the number of groups used for the two-dimensional assembly calculations is either made in one step (for PWR assembly) or vit macro group calculations explained in Section 2.3.3.

Diagonal assembly symmetry is assumed and the 2D calculation is made on one triangular half of the assembly cell. The triangular cell is placed in a rectangular XY-coordinate system so that the two smaller sides of the triangle coincide with the coordinate axis. The cell is divided into macro meshes by lines drawn parallel to the X and Y-axis. These macro meshes will be rectangular, except along the hypotenuse of the triangle where other figures will be formed.. They are determined by the material regions formed by pin cells, box wall, cruciform control rods, etc. The macro meshes are further divided into micro meshes as specified by the user. The lines run completely through the system. A flat flux region usually consists of one micro mesh but may also consist of several meshes, e.g. composite regions are formed along lines introduced by the cruciform control rod and the boron curtain.

Page 23 I ' Sets of equidistant parallel lines are drawn across the assembly at a number of equally spaced angles. The lines are reflected at the j boundary a number of times, sufficient to allow the neutrons to be tracked a specified optical length. The track intercepts and region numbers associated with the intercepts for each line are written as a

. record to a file. This track intercept file is read to the primary memory for each energy group. Collision probabilities are then cal-culated utilizing the information in the track intercept record and the group constants. In. order to improve the accuracy of the colli-sion probabilities the track intercepts are renormalized using the ratio between the exact region volume and the volume which is calcu-lated numerically from the track intercepts. Finally, the collision probability matrix is balanced by adding a correction term to the diagonal element making the row sum equal to unity.

A fundamental buckling mode is used for modifying the infinite lat-tice results obtained from the transport calculation to include the effects of leakage. This calculation may be carried out either in diffusion theory or by use of the Bi-leakage method.

2.7 The Burnup Calculation s

The burnup calculation is carried out in two partial steps. Going from the time to_t to-to, first a " predictor" step is made using the fluxes obtained from the spectrum calculation at tn_t. The predictor step provides predicted number densities at to where, after the cross sections are updated, a new spectrum calculation i

Page 24 gives fluxes to be used in a " corrector" step. The final number den-sities at to are then given by the average value of the results from the predictor and corrector steps. This method is very effi-cient and makes it possible to take much longer burnup steps than is usually the case in cell burnup codes.

The basic burnup chains, with the isotopes linked through absorption and decay, are linearized and the differential burnup equations are solved by a fast analytical treatment. Five standard linear chains are treated for the heavy nuclides as shown in Figure 2.8. Twenty two individual and two pseudo-fission products are included in four-teen linear fission product chains shown in Figure 2.9. The individ-ually treated fission products account for about 90% of the total fission product absorption. Boron in boron steel curtains or in burnable poison rods is also burnt, as is gadolinium in burnable ab-sorber rods.

i The isotopic depletion is calculated separately in each fuel pin and

in each burn absorber region. If xenon is not included in the input, an estimate of the equilibrium xenon number density is made at zero burnup. In the following burnup, xenon number density is calculated by solving the first order differential burnup equations of the lin-1 i earized chains.

The fluxes obtained from the transport calculation are normalized so l

that the total source in the cell is equal to unity. Before deple-i tion, the fluxes are renormalized so that the average power density l

is equal to that given in the input.

. . _ . . _ = _ ~ _ _. . _._.

Page 25 The depletion of burnup absorber is calculated in the same way as for nuclides taking part in the ordinary burnup chains. The effective l cross section for gadolinium is obtained by linear interpolation in the cross section tables generated by MICBURN using effectivs Gd fraction as the parameter. The effective Gd fraction at each burnup level is calculated by dividing the Gd-155 and Gd-157 number density by the initial Gd-155 and Gd-157 number density.

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Page 26 TABLE 2.1 EEERGY BOUNDARIES IN THE 69 GROUP LIBRARY Group- Energy Group Energy MeV eV 1' 10.0 -6.0655 36 1.097-1.071 2 6.0655 -3.679 37 1.071-1.045 3 3.679 -2.231 38 1.045-1.020 4 2.231 -1.353 39 1.020-0.996 .

5 1.353 -0.821 40 0.996-0.972 6 0.821 -0.500 41 0.972-0.950

.7 0.500 -0.3025 42 0.950-0.910 8 0.3025 -0.183 43 0.910-0.850 9 0.183 -0.1110 44 0.850-0.780 10 0.1110 -0.06734 45 0.780-0.625 11 0.06734-0.04085 46 0.625-0.500

-12 0.04085-0.02478 47 0.500-0.400 13- 0.02478-0.01503 48 0.400-0.350

14. 0.01503-0.009118 49 0.350-0.320 50 0.320-0.300 eV 51 0.300-0.280 15 9118.0 -5530.0 52 0.280-0.250

~16 5530.0 -3519.1 53 0.250-0.220' 17 3519.1 -2239.45 54 0.220-0.180 18 2239.45 -1425:1 -55 0.180-0.140 19 1425.1 - 906.898 56 0.140-0.100 20 906.898- 367.262 57 0.100-0.080 21 367.262 .148.728 58 0.080-0.067 22 148.728- 75.501 59 0.067-0.058 23 75.501- 48.052 60 0.058-0.050 24 48.052- 27.700 61 0.050-0.042 25 27.700- 15.968 62 0.042-0.035 26- 15.968- 9.877. 63 0.035-0.030 27 9.877-' 4.00 64 0.030-0.025 28 4.00 - 3.30 65 0.025-0.020 29 3.30 - 2.60 66 0.020-0.015 30 2.60-- 2.10 67 0.015-0.010

'31 2.10 - 1.50 68 0.010-0.005 32 1.50 - 1.30 69 0.005-0 33 '1.30 - 1.15

.34- .1.15 - 1.123

~35 1.123- 1.097-1

t Page 27

TABLE 2.2 TABLE OF NUCLIDES No . , Nuclide ' Identification NINA* Temperatures (K)*

number 1 H- 1001 0 296, 350, 400, 450,

. 500, 600 2 D 1002 0 296 3 B 5000 0 300 4 B-10 5010 0 293.6 5 C 6000 0 300 6 _0(free) 8000 0 296, 400, 600 7 0(not'to be used) 8001 0 296, 500, 800, 1200

8. Al 13000 0 300 9 Si 14000 0 300 10 Cr 24000 0 300 11~ Mn 25000 1 300

! 12 Fe 26000 0 300 13 Ni 28000 0 300 3

14 Cu-63 29063 1 300 t 15 Ag(shielded) 47000 0 300

, ~ 16 - Cd(shielded) 48000 0 300 17 In(shielded) 49000 0 300 18- Gd-154 64154 1 300 19 Gd-155 64155 1 293.6

'20 Gd-156- 44156 1 300

21. 'Gd-157 - 64157 '1 293.6 22- Gd-158 64158 1 300 23- Dy-164 66164 1 -293.6 24 Lu-176 71176 1 293.6

-25 Kr-83 36083 1 293.6 l 26 Rh 103 45103 1 293.6 ,

27 Rh-105 45105 1 293.6 4

28 ~Ag-109 47109 1 293.6 29 .Xe-131 54131 1 293.6

~30 Xe-135 -54135 1 293.6 31' Cs-133 55133 1 293.6 32 Cs-134 55134 1 293.6 ,.

.33 Cs-135 55135 1 293.6 34 Nd-143 60143 1 293.6 i 35 Nd-145 60145 1 293.6 4

36 Pm-147 61147 1 293.6

-37' Pm-148 61148 1 293.6 38 Pm-148m 61248 1 293.6

.39 Sm-147 62147 1 293.6

?40 -Sm-149 62149 1 293.6 41 Sm-150 62150 1 293.6 42 Sm-151 62151 1 293.6 L

t

~ <

  • w , y --,.'-- ,y,.v,w w . .-y,. ~, ------~7.-- , ,, ,---y,----- - , ,-y. , ,, . . - - - - - - .- - - - - - .----- -

Page 28 TABLE 2.2 (Contd.)

TABLE OF NUCLIDES No'. Nuclide Identification NINA 1 Temperature (K)*

number 43 Sm-152- 62152 1 293.6 44 Eu-153 63153 1 293.6

'45 Eu-154 63154 1 293.6

'46 Eu-155 ' 63155 1 293.6 47' NSFP 401 1 293.6 48' 'SSFP' 402 1 293.6 49 U-234 92234 0 300 50 U-235 92235 0 300

'51- U-236 92236 0 300 52 U-238 92238 0 300 53 Np-237 93237 0 300

.54- Pu-238 94238 0 300 55 Pu-239 94239 0 300 56 Pu-240 94240 0 300, 600, 1200 57 Pu-241- 94241 0 300 58 Pu-242 94242 0 300 59 Am-241 95241 0 300

_: 60 Am-242m 95242 0 300 61 Am-243 95243 0 300

62 Cm-242 96242 1 300

!s 63- -Cm-244 '96244 1 300 64 1/v-abs 1 1 293.6 65 Zr-2 302 0 300, 600, 1200

1. NINA = 0 all relevant cross sections tabulated i

~= 1 only absorption cross sections tabulated

2. Temperatures for which thermal cross sections are tabulated in subfile 4 of the library. The program interpolates and extrapolates to other

, . temperatures.

3. -Fission-products not separately treated are represented by two pseudo fission products, one non saturating (NSFP) and one slowly saturating '

(SSFP).

Figure 2.1 FLOW DIAGRAM OF CPM

> Input

, m;

_ _ _ __ y Restart file n-- -_---J t

sonance calculation 4 Data library Y

Macroscopi p_______q cross sections <  ; MICBURN I g_______g t

Micro group calculation 69 gr. max 5 regions V

Condense to macro groups (max 25 groups)

Homogenize to macro regions y

y Condense to max 12 groups Macro group calculation in Calc. cross sections annular geometry for 2-D regions y -

Condense to max 12 groups Calc. cross sectioins for 2D-regions

%u *

.E Control rod calc.

CROCOP Y

2-dimensional collision probability calculation V

>l Fundamental mode calc. l t

Few group constants Reaction rates 6

Y Burnup corrector V Zero burnup  ;

P-C Number densities j Y

Burnup a predictor Y

l End l

Figure 2.2 FLOWCHART OF THE RESONANCE CALCULATION

FLURES Collision probabilities v

RESPAR Parameters for the equivalence theorem

<r RESABS Effective cross sections for uniform pin cell lattice

$. pin cell assembly v

DANSO Dancroff factors 1r DANCOR Position dependent effective cross section v

STOP

Figure 2.3 l FLOWCHART OF MICRO-GROUP CALCULATION i Define average fuel pin cell Geometry y Fuel pin cells Collision probabilities for "

cylindricalized pin cell Can Fuel V

_a j Correction of CP for the 5 influence of water gap o.

3 a V Can Solve the neutron balance

% Fuel k eq in 4 regions and

_~ 69 energy groups Coolant R

Homogenized V Gap Determine various types of pin cells V ,

Absorber rods and water holes Calculate cross sections for buffer zone Homogenized

, buffer zone i

$g f V Collision prcbabilities for Coolant Sj

  • cylindricalized absorber iE pin cell (or water hole) Can 37 surrounded by buffer zone k$

_o Absorber

-c V

RE Solve neutron balance eq l

in up to 5 regions and l 69 energy groups V

Homogenization of pin cells i and condensation to macro groups

l Figure 2.4 EXAMPLE OF GEOMETRY IN MACRO-GROUP CALCULATION i

O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O O u On pin O 0 0 0 0 0 0 0 @ PuO:-pin

_ 0 0 0 0 0 0 0 0 O O O 010 0 0 0

/

Wide water gap

\

Narrow water gap 1

Homogenized L t Outer water 9 PuO control rod PuO gap UO Box UO Box Homogenized fuel layers Ouier water Homogenized fuel layers gap inner water gap inner water gap

Figure 2.5 FLOWCHARY OF MICBURN Input v

Resonance Data library calculation tape v

Macroscopic cross sections ir Pin cell calc.

Buffer zone v

Homogenization

  • of buffer zone v

Define macro regions o

Transport calculation v

Flux distribution in micro regions v

Burnup calculation u

End

l Figure 2.6a ACTUAL CONTROL ROD CONFIGURATION entra t c re 7 Sta niess steel Stainless steel N V/ N///////////

N\vHHMMMMH V////////// M M HA

\/

Water V///N /////// H B.c 11 -

Figure 2.6b REPRESENTATION OF CONTROL ROD WING IN CPM st I and water esomer @.c)

V '

j jj

Figure 2.7 EXAMPLE OF BWR CELL GEOMETRY IN THE 2D CALCULATION Steel C

>/ ontrol s

rod Wide water gap

'N

's y ..e, ,,n ..,,

~

'N Inner water gap Box wall Narrow water

/

1111111111111111.1111111,11,1,,,,,,,,;,;,

/ gap ri m m u sssssssssssis ,ssi, \

\

Boron curtain

Page 36 FIGURE 2.8 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'+ P240 + Pu241 + Am241 + Cm242 + Pu238 (n 2n)
5. -U238 + Np237 + P238

.-.e.-. - . _ - , - ,,, -, - , . ._ -- , _ _ _

Page 37 FIGURE 2.9 FISSION PRODUCT CHAINS

1. Kr83~
2. Rh103
3. Rh105

'4 . Ag109'

5. Xe131

'6. Cc133 + Cs134

7. Xe135 + Cs135

-8. Nd143

9. Nd145 (52.77 %)
10. Pm147 + Pm148 + Sm149 + Sm150 + Sm151 + Sm152 +

Eu153 + + Eu154 + Eu155 (47.23 %) -

i 11. Pm147 + Pm148 + Sm149 + Sm150 + Sm151 + Sm152 +

Eu153 + Eu154 + Eu155

12. Pm147 + Sm147
13. NSFP (non saturating F.P.)
14. SSFP (slowly saturating F.P.)

i I

k I

,~

L

Page 38 3.0- Verification The EPRI-CPM has been benchmarked against selected series of experi-mental results from critical facilities and power reactors (Reference 2). The overall agreement between CPM and experiments ver-ify the following aspects of the code:

. nuclear data library : CPMLIB3

. neutron transport. treatment

. control rod treatment

. gadolinium treatment

. pin cell calculation

. two dimensional fuel bundle calculation

. depletion calculation To augment EPRI's benchmarking work, GPUN has performed calculations to compare CPM results to measured reactor data and to calculations made by higher order methods. The adequacy of CPM is. further demonstrated by comparing the CPM / NODE-B core analysis results with Oyster Creek-operating data. Section 3.1 describes the verification work at GPUN and Section 3.2 describes the EPRI benchmarking work. The 3D simulator analysis results will be reported in a separate topical report.

3.1 GPUN Verifications 3.1.1 Comparison with Oyster Creek Pin Gamma Scan The local power distributions calculated by CPM are verified by compar-ison to fuel rod gamma scan measurements made at Oyster Creek. At the

Page 39 end of Oyster Creek cycle 5 and cycle 6 two Oyster Creek fuel assemblies were scanned. The measurements were performed by removing the fuel rod from the fuel assemblies and measuring the La-140 activity, over the entire length of the rods. The gamma scan data then represents an axially-averaged measurement. Unfortunately, gamma scan at various axial heights were not available. The measured data are documented in References 3 and 4.

A single assembly fuel depletion calculation was performed for each of the two measured fuel assemblies using CPM. The fuel depletion calculations were performed using the actual control rod history and void history at three axial heights for each assembly. The averaged results were used for comparison.

The ratio of measured versus calculated pin power for the two assem-blies are given in Figures 3.1 through 3.4. Because of our proprietary agreement with Exxon Nuclear Corporation, the measured and calculated pin power values are not shown in these figures. Results of these com-parisons are summarized in Table 3.1. The standard deviations for the mean ratio are within the standard deviation of the measured data as shown in Table 3.1. The worst comparison is for UD4070 measured in 1977 but the standard deviation (1.004 + .103) is still comparable with the 13.4% from measurement.

Page 40 3.1.2 Comparison with Monte Carlo Calculations The CPM code has been benchmarked with higher order Monte Carlo calcu-lations (Reference 5). The Monte Carlo calculations was performed by thel fuel vendor based on the general Monte Carlo code developed by Battelle Northwest (Reference 8). A comparison of K-infinity values are shown in Table 3.2. The local power distribution comparison are shown in Figures 3.5 through 3.9. The standard distribution for power distribution is 1.8% for all cases studied. The agreement in K-infinity values is also good as can be seen in Table 3.2. The average percentage difference is 0.36 with a standard deviation of 0.98.

3.1.3 Comparison with Hatch-1 Pin Gamma Scan Besides Oyster Creek pin gamma scan data, the local power distributions calculated by CPM are also compared with the gamma scan data at the end of Cycle 1 of Hatch-1. This demonstrates our ability to model other reactors adequately and enhances the confidence of the lattice physics model.

The pin gamma scan measurements of four Hatch-1 fuel assemblies were taken at the end of C"ycle 1. The experimental results are published in Reference 6. Following the fuel bundle design information and core performance data (Reference 15), the operating histories of these four bundles were obtained and modeled with CPM at three axial heights cor-responding to measurement locations. The axial exposure and void data for these four bundles at the end of the cycle are from N0DE-B depletion

Page 41 4

The resulted pin power distributions are compared of Hatch-1 cycle 1.

with the measured ones. The comparison results are summarized in Table 3.3. Detail pin power distribution comparisons are shown in Figures 3.10-3.21.

The agreement between measured and calculated pin power is generally better for the 129 inches cases than the 15 inches cases. This is probably due to the differences in void history and rod history. The l collision probability can be calculated more accurately at higher

. voids, as in the 129 inches cases, due to the longer neutron mean free

.s i path. During Hatch cycle 1 oper.Ition none of the four bundles examined had a significant control rod history at 129 inches above the bottom of the fuel while a more. complicated rod history existed for each of the four bundles at 15 inches above the bottom of the fuel. While the CPM' calculations duplicated the bundle control rod history, the effect of surrounding control rods were not included. The control rod history for the last few months of the cycle was held at 15.7%, which is significant and would contribute to the flux tilting, especially in the lower part of the core. The overall standard deviation between the measured and calculated pin power distribution is 5.3% which is higher than the 2% experimental uncertainty given in Reference 3.6. However, the comparison is still very good considering the operating history of

- Hatch-1: cycle 1 and the fact that the lattice physics model is not capable of handling power tilting effects due to surrounding assemblies.

Page 42 3.2 EPRI Benchmarking The benchmarking work described in this section was performed by EPRI and documented in Reference 2.

3.2.1 Hot Critical Experiments in KRITZ Four different experiments in the high temperature critical facility KRITZ at Studsvik, Sweden were selected for the benchmarking of CPM.

The study includes BWR systems, PWR systems, and pin cell lattices using enriched-UO or mixed oxide fuel. The temperature for BWR and PWR lattices was about 245'C, while the temperature for uniform pin cell lattice were 20*C and 210*C. Details of experiments and calcula-tions are reported in References 2 and 9.

Measured and calculated fission rate distributions are compared in Figs. 3.22-3.24. The figures show the deviation in precentage between theory and experiment in all measured pins. The fission rates were normalized so that the average value in the measured pins in the assem-bly is equal to unity. The predicted reactivities for the critical system studies are also listed in Figs. 3.22-3.24. The calculated values are all close to unity. The agreement with the experimental value is generally better than could be expected from the uncertainty in the tracroscopic calculation and in the experiments themselves.

Page 43 The uniform pin cell lattic studied is a lattice of 1.35% enriched UO rods. The K-effective was 0.997 at 20*c and 0.993 at 210*c. The high temperature value is slightly lower than the value for the cold lattice. This is a tendency that has been observed by many other lat-tice physics code.

3.2.2 Cold critical on Uniform Lattices The benchmarking of CPM against cold criticals on uniform lattices was on the results from TRX and ESADA experiments (References 10 and 11).

The calculation using the experimental values as the geometrical buck-ling. Results are listed in Tables 3.4 and 3.5 and in Fig. 3.25.

, These provide a basis of confidence in the pin cell treatment used by l

CPM.

l 3.2.3 Yankee and Saxton Isotopics Isotopic composition calculated by CPM have been compared with spent fuel isotopic data obtained from Yankee and Saxton (References 12 and 13). These provide a basis for confidence in pin cell treatment and l depletion calculation.

Isotopic ratios for plutonium in Yankee are compared in Figs.

3.26-3.28. The abscissa is a fission product number density assuming a l

yield of unity and volume averaged over the pin cell. The correspond-ing burnup in MWD /kgU is also shown. The dots are experimental results and the line the CPM results.

l w___________.

Page 44 The agreement between calculated and experimental isotopic ratios is good. The calculated ratios Pu-239/Pu-240 and Pu-240/Pu-241 are within the scatter of the experimental results and the ratio Pu-241/Pu-242 is slightly overpredicted (by about 3 % at 30 MWD /kgU).

Calculated and measured isotopic compositions for Saxton (pellet, rod MY, zone 6) are compared in Table 3.6. The agreement is good for the most important uranium and plutonium isotopes as well as for americium and curium. The concentrations of Np-237, Pu-238 and Pu-242 are under-estimated.

(

0 i

1

,_ -m,___, --, . ,r .. _ _ _-,-- - , m--,, - - , - , ,,--. - -_ -- - _ _ - . - - - - .

_ , . ~

Page 45 TABLE 3-1

,~

SUPffARY OF CPM CALCULATED AND OYSTER CREEK La-140 GAMMA SCAN MEASUREMENT COMPARISON Exposure Mean Standard Y~r Assembly ~ .h MWD /MTU Ratio

  • Deviation 1976 UD3109 8x8. 3800 1.000 0.034 UD4070 8x8 3900 1.001 0.047 Average 1.001 0.041 Experiment 0.056 1977 UD3109 8x8 12000 1.000 0.031 UD4070 8x8 11500 1.004 0.104 Average 1.000 0.093

. Experiment 0.134 i

CRatio = Gamma Scan Measured / CPM Calculated l

l l

I

Page 46 TABLE 3.2 COMPARISON OF CPM CALCULATED AND MONTE CARLO CALCULATED K-INFINITY K-Infinity Cn e Description Monte Carlo CPM  % Difference

  • 0% Voids, without Gd20s, No Rods 1.3073 0.0018 1.2978 0.73 32% Voids, without Gd20s, No Rods 1.2889 ! 0.0018 1.2782 0.83 64% Voids, without Gd203, No Rods 1.2590 2 0.0017 1.2427 1.29 32% Voids, with Gd203, No Rods 1.2111 1 0.0010 1.1948 1.35 32% Voids, without Gd20s, With Rods 0.970 0.0024 0.9673 0.28 celd, without Gd20s, No Rods 1.289 0.0024 1.3025 -1.04 cold, without Gd203, with Rods 1.100 0.0030 1.1098 -0.89

% difference mean 0.36

% difference standard deviation 0.98

  • % Difference = (Monte Carlo - CPM)/(Monte Carlo)*100

Page 47 TABLE 3-3

SUMMARY

OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASUREMENT COMPARISON Inches From Exposure Mean Standard Assembly _ h Bottom of Fuel MWD /MTU Ratio

  • Deviation EX393 7X7 15 9576 1.004 0.068 63 13184 1.000 0.046 129 10401 1.000 0.030 EX373 7X7 15 11333 0.989 0.086 63 12763 1.005 0.045 129 10268 1.000 0.048 RX141 7X7 15 11239 1.001 0.079 63 12131 1.001 0.050 129 10483 0.998 0.027 HX161 7X7 15 9874 1.001 0.044 63 14145 1.000 0.039 129 10450 1.000 0.032 Average 1.000 0.053 Experiment 0.02 cRatio = Gamma Scan Measured / CPM Calculated

Page 48 TABLE 3.4 CPM RESULTS FOR TRX CRITICALS Hexagonal lattice Pellet B*(exp) pitch diameter No. (in) (in) (m-*) k.rr 1 .868 .601 28.4 .997 2 .929 .601 30.2 .999 3 .989 .601 29.1 .998 4 .613 .388 25.3 .998 5 .650 .388 25.2 .997 6 .613 .383 32.6 1.001 7 .650 .383 35.5 1.000

8 .711 .383 34.2 1.000

Page 49 TABLE 3.5 CPM RESULTS FOR ESADA CRITICALS Exp.. Fuel Type Lattice Boron B* (exp) k.cs pitch conc.

(in) (ppm) (m-*)

l', 2 8% Pu-240 .69 0 69.1 .999 3 8% Pu-240 .75 0 90.0 1.000 4,5 8% Pu-240 .9758 0 105.9 1.008 6 8% Pu-240 1.0607 0 98.4 1.010 7 8% Pu-240 1.380 0 50.3 .997 8 8% Pu-240 .69 261 62.6 1.004 9 8% Pu-240 .9758 261 83.7 1.002 10 8% Pu-240 .69

  • 526 58.3 1.002 11 8% Pu-240 .9758 526 63.1 .999 12 24% PU-240 .9758 0 79.5 1.004 13 24% PU-240 1.0607 0 73.3 1.002 i

i l

Page 50 TABLE 3.6 ISOTOPIC COMPOSITION IN SAXTON COMPARISON BEIVEEN CPM AND EXPERIMENT Nuclide Experiment Experimental CPM-exp . tao Uncertainty exp Atom %

U-234 .00465 28.7 +15.9 U-235 .574 .9 - 0.3 U-236 .0355 5.6 + 2.8 U-238 99.386 0 0 Pu-238 .109 2.2 -11.4 Pu-239 73.77 0 - 0.3 Pu-240 19.25 .2 + 1.6 Pu-241 6.29 .3 + 0.4 Pu-242 .579 .9 -16.0 Atom Ratios Np-237/U-238 1.14'10-* 15 -26.4 Pu-239/U-238 .04383 .7 + 0.2 Pu-238/Pu-239 1.75'10-* .4 - 9.8 Am-241/Pu-239 .0123 15 -10.6 Cm-242/Pu-239 1.05'10-* 10 0 Cm-244/Pu-239 1.09 10-* 20 0

Page 51 Figure 3.1 COMPARISON OF CPM CALCULATED AND MEASURED PIN POWER DISTRIBUTION FOR OYSTER CREEK, AXIALLY-AVERAGED 1976 La-140 GAlWIA SCAN wide map wida 1.030 1.031 1.011 0.949 0.976 Ratio.

gap-1.024 1.063 1.051 1.014 1.005 0.994 0.994 1.037 1.044 1.034. 1.001 0.999 l

-0.993 1.037 0.970 0.953 0.993 1.019 0.988 0.993 0.961 1.004 0.950 0.940 0.943 i

l Fuel Assembly: UD3109 I

Distance Above Bottom of' Fuel: 27 inches Average Exposure: 3800 MWD /MTU

. Average Void History: 38% i Ratio = Gamma Scan Measured / CPM calculated

Page 52 Figure 3.2 COMPARISON OF CPM CALCULATED AND MEASURED PIN POWER DISTRIBUTION FOR OYSTER CREEK, AXIALLY-AVERAGED 1976 La-140 GAMMA SCAN wide gap Old3 1.007 0.971 1.024 1.016 1.107 Ratio gap 0.999 1.019 1.027 1.029 0.989 1.068 1.035 1.024 1.031 0.967 0.964 1.050 1.041 1.025 1.027 1.041 1.032 1.001 0.988 0.934 0.891 0.965 0.917 0.941 0.958 0.932 Fuel Assembly: UD4070 Average Exposure: 3900 MWD /MTU Average Void History 38%

Ratio = Gamma Scan Measured / CPM Calculated

c' Page 53 Figure 3.3 COMPARISON OF CPM CALCULATED AND MEASURED PIN POWER DISTRIBUTION FOR OYSTER CREEK, AXIALLY-AVERAGED 1977 La-140 GAMMA SCAN wide gap wida 1.020 0.976 1.005 0.983 1.026 Ratio 8P 1.025 1.027 1.000 0.935 l

l i

Fuel Assembly: UD3109 Average Exposure: 12000 MWD /MTU Average Void History: 27%

Ratio = Gamma Scan Measured / CPM calculated

Page 54 Figure 3.4 COMPARISON OF CPM CALCULATED AND MEASURED PIN POWER DISTRIBUTION FOR OYSTER CREEK, AXIALLY-AVERAGED 1977 La-140 GAMMA SCAN wide gap wida 0.899 1.027 0.813 0.972 0.907 Ratio

-g:p 0.977 0.879 '0.929 1.111 0.980 0.890 0.990 1.093 1.130 1.109 1.054 1.160 1.226 0.911 0.963 1.051 0.917 1.119 1.092 0.864 1.030 1.145 1.080 0.874 1.010 0.921 Fuel Assembly: UD4070 Average Exposure: 11500 MWD /MTU Average Void History: 24%

Ratio = Gamma Scan Measured / CPM calculated i

. =_ .. - .

Page 55 Figure 3.5 COMPARISON OF CPM CALCULATED AND MONTE CARLO CALCULATED PIN POWER DISTRIBUTION wide map wid] 1.065

_ gap 1.042 1.022 0.919 1.018 XMC 0.892 1.031 CPM '

1.030 0.987 Ratio 1.167 1.019 0.889 1.147 1.046 0.922 1.017 0.974 'O.964 1.098 0.970 0.872 0.849 1.094 0.993 0.880 0.842 1.004 0.977 0.991 1.008 1.097 1.010 0.886 0.841 0.840 1.098 0.998 0.885 0.848 0.854 0.999 1.012 1.001 0.992 0.984 1.146 1.050 0.947 0.902 0.911. 0.947 1.158 1.061 0.942 0.901 0.906 0.960 0.990 0.990 1.005 1.001 1.006 0.986 0.934 1.072 1.088 1.000 1.049 1.104 1.106 0.907 _1.069 1.097

  • 1.047 1.052 1.110 1.088 1.030 1.003 0.992 1.012 0.997 0.995 1.017 0% Void-without Cd203 - Unrodded Mean Ratio  : 1.000 Standard Deviation: 0.016

-. , _ _ . _ _ . ~ _

Page 56 Figure 3.6 COMPARISON OF CPM CALCULATED AND MONTE CARLO CALCULATED PIN POWER DISTRIBUTION wide gap wid3 1.141 g:p 1.088 1.049 0.942 1.071 XMC 0.924 1.059 CPM 1.019 1.011 Ratio 1.203 1.068 0.940 1.177 1.064 0.926 1.022 1.004 1.015 1.130 1.001 0.872 0.825 1.115 1.001 0.874 0.825 1.013 1.000 0.998 1.000 1.121 1.007 0.881 0.840 0.819 1.116 1.002 0.875 0.827 0.829 1.004 1.005 1.007 1.016 0.988 1.182 1.064 0.912 0.856 0.848 0.937 1.177 1.065 0.931 0.881 0.882 0.935 1.004 0.999 0.980 0.972 0.961 1.002 0.944 1.076 1.076 0.994 0.998 1.083 1.082 ,

0.923 1.072 1.087

  • 1.027 1.028 1.087 1.070 1.023 1.004 0.990 0.968 0.971 0.996 1.011 32% Void-without Gd203 - Unrodded Mean Ratio  : 1.001 Standard Deviation: 0.019 i

r . - . - - , . - - - - - - , , - - ~ , . _ - - - . - - - - - - - . . - -

~

Page 57 Figure 3.7 COMPARISON OF CPM CALCULATED AND MONTE CARLO CALCULATED PIN POWER DISTRIBUTION wide map wida 1.184

. gip' 1.134 1.044 0.974 1.116 XMC 0.955 1.092 CPM 1.020 1.022 Ratio 1.222 1.101 0.950 1.207 1.091 0.940 1.012 1.009 1.011 1.119 1.000 0.863 0.816 1.136 1.017 0.877 0.818 0.985 0.983 0.984 0.998 1.138 0.999 0.853 0.793 0.191 1.131 1.011 0.871 0.812 0.807 1.006 0.988 0.979 0.977 0.980 1.201 1.083 0.920 0.848 0.840 0.908 1.190 1.069 0.922 0.861 0.856 0.905 1.009 1.013 0.998 0.985 0.981 1.003 0.968 '

1.095 1.070 0.989 0.984 1.040 1.035 0.932 1.070 1.072 ' 1.003 0.997 1.052 1.038 l1.039 l 1 023 0.998 0.986 0.987 0.989 0.997 64% Void-without Gd203 - Unrodded Mean Ratio  : 1.000 Standard Deviation: 0.018 1

Page 58 s.

Figure 3.8 COMPARISON f'F CPM CALCULATED AND MONTE CARLO CALCULATED PIN POWER DISTRIBUTION wide man wida. 1.216.

3:p 1.177 1.033 1.006 1.139 XMC 0.997 1.137 CPM 1.009 1.002 Ratio 1.255- 1.117 0.962 1.265 1.134 0.974 0.992 0.985 0.988 1.173 1.047 0.898 0.833~

1.191- 1.056 0.900' 0.822 0.985- 0.991 0.998 1.013 1.188 1.049 0.873 0.759 0.755 l'.184 1.042 0.867 0.761 0.765 1.003 1.007 1.007 0.997 0.987 1.255 1.114 0.884 0.380 0.813 0.898 1.243 1.099 0.887 0.400 0.808 0.916 1.010 1.014 0.997 0.950 1.006 0.980 1.015 1.129 1.086 0.969 0.988 1.058 1.082 0.972 1.110 1.080 ' O.971 1.002 1.091 1.089 1.044 1.017 1.006 0.998 0.986 0.970 0.994 32% Void-with Gd203 - Unrodded Mean Ratio  : 0.999 Standard Deviation: 0.018

Page 59 Figure 3.9 COMPARISON OF CPM CALCULATED AND MONTE CARLO CALCULATED PIN POWER DISTRIBUTION wide gap L eida - 0.446-grp 0.443 1.007 0.435 0.653 KMC 0.437 0.653 CPM 0.995 1.000 Ratio 0.620 0.778 0.846 0.628 0.780 0.827 0.987 0.997 1.023 0.672 0.843 0.883 0.928 0.688 0.831 0.882 0.939 0.977 1.014 1.001 0.988 0.755 0.944 0.981 1.001 1.089 0.758 0.923 0.965 1.015 1.082 0.996 1.023 1.017 0.986 1.006 0.966 1.096 1.079 1.105 1.181 1.323 0.952 1.087 1.098 1.132 1.195 1.308 1.015 1.008 0.983 0.976 0.988 1.011 0.961 1.188 1.324 1.341 1.406 1.527 1.542 0.902 1.189 1.335 - 1.357 1.423 1.548 1.547 1.065 0.999 0.992 0.988 0.988 0.986 0.997 32% Void-without Gd203 - Rodded Mean Ratio  : 1.000 Standard Deviation: 0.018 .

l Page 60 i

l Figure 3.10 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION  !

wide gap wida 0.442 0.457 0.602 0.871 1.032 Gamma Scan g:p 0.427 0.461 0.596 0.848 0.991 CPM 1.035 0.991 1.010 1.027 1.042 Ratio 0.443 0.658 0.698 0.917 1.001 1.092 1.072 0.461 0.639 0.690 0.895 0.979 1.113 1.072 0.960 1.029 1.011 1.024 1.023 0.981 1.000 0.664 0.914 0.972 1.011 1.165 0.690 0.913 0.973 1.056 1.137 0.962 1.002 0.999 0.953 1.025 0.575 0.909 0.935 1.070 1.191 1.297 0.596 0.895 0.973 1.119 1.219 1.378 0.964 .1.015 0.961 0.957 0.977 0.941 1.035 1.069 1.112 1.136 1.183 0.979 1.056 1.119 1.140 1.281 1.058 1.013 0.994 0.997 0.924 0.965 1.198 1.232 1.211 1.214 1.307 1.159 0.848 1.113 1.137 1.219 1.281 1.371 1.265 1.137 1.077 1.084 0.994 0.943 0.953 0.916 1.259 1.216 1.351 1.169 1.197 0.991 1.072 -

1.378 1.265 1.370 1.271 1.135 0.981 0.924 0.874 Fuel Assembly: HX393 Distance Above Bottom of Fuel: 15 inches Exposure: 9576 MWD /MTU Void History: 0%

l

Page 61 Figure 3.11 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAP 99L SCAN MEASURED PIN POWER DISTRIBUTION wide aan wide 1.142 1.016 1.023 1.052 1.027 Gamma Scan gap 1.048 0.967 0.997 1.066 1.062 CPM 1.090 1.050 1.026 0.987 0.967 Ratio 1.003 1.028 0.984 1.111 1.074 1.087 0.942 0.967 0.997 0.935 1.063 1.057 1.092 1.010 1.037 1.031 1.052 1.045 1.016 0.995 0.933 0.987 1.052 0.996 0.952 0.973 0.935 0.989 0.964 0.961 0.980 1.055 1.064 1.033 0.990 0.993 0.991 1.129 1.004 0.909 0.886 1.038 0.997 1.063 0.964 0.919 0.963 1.074 0.994 1.062 1.041 0.989 0.920 0.966 1.078 ~0.971 0.923 0.899 0.914 1.057 0.961 0.919 0.889 0.964 1.020 1.010 1.004 1.011 0.948 1.094 1.101 1.007 0.949 0.927 0.960 0.854 1.066 1.092 0.980 0.963 0.964 1.004 0.949 1.026 1.008 1.028 0.985 0.961 0.956 0.900 1.090 0.997 1.046 0.874 0.910 1.062 1.010 '

1.074 0.949 1.022 1.026 0.987 0.974 0.921 0.891 Fuel Assembly: HX393 Distance Above Bottom of Fuel: 63 inches Exposure: 13184 MWD /MTU Void History: 39%

i 1

E Page 62 m

e Figure 3.12 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.206 1.041 1.026 1.086 1.118 Gamma Scan gap 1.148 1.027 1.032 1.114 1.139 CPM 1.051 1.014 0.995 0.975 0.982 Ratio 1.018 1.060 0.957 1.078 1.069 1.083 0.971 1.027 1.023 0.939 1.063 1.055 1.094 1.034 0.992 1.037 1.019 1.014 1.013 0.990 0.939 s

0.938 1.030 0.948 0.940 0.926 0.939 1.000 0.939 0.931 0.929 r 0.999 1.030 1.010 1.010 0.997 g.

1.046 1.092 0.972 0.879 0.946 1.045 1.032 1.063 0.939 0.869 0.918 1.060 1.014 1.027 1.035 1.011 1.031 0.986 1.075 0.936 0.906 0.868 0.904 1.055 0.931 0.869 0.862 0.911

< 1.019 1.006 1.042 1.007 0.993 1.080 1.098 0.950 0.940 0.897 0.964 0.844 L: 1.114 1.094 0.929 0.918 0.911 0.959 0.933 0.969 1.003 1.023 1.024 0.985 1.005 0.905 1.130 1.004 1.081 0.879 0.969 1.139 1.034 '

1.060 0.933 1.039 0.992 0.971 1.020 0.942 0.933 Fuel Assembly: HX393 Distance Above Bottom of Fuel: 129 inches v Exposure: 10401 MWD /MTU Void History: 67%

E E

m em-

Page 63 -

Figure 3.13 COMPARIS0N OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION

_a wide -

gap wide 0.349 0.393 0.592 0.924 1.230 Gamma Scan i gap 0.414 0.452 0.586 0.832 0.975 CPM _

0.843 0.869 1.010 1.110 1.261 Ratio _

0.364 0.552 0.641 0.896 1.050 1.216 1.222 -

0.452 0.625 0.684 0.884 0.970 1.103 1.068 I 0.805 0.883 0.937 1.013 1.082 1.102 1.144 7 0.632 0.858 0.980 1.086 1.292 -

0.684 0.906 0.974 1.061 1.153 '

O.924 0.947 1.006 1.023 1.120 -

[ 0.492 0.586 0.802 0.884 0.946 0.974 1.114 1.133 1.258 1.479 1.232 1.383 -

0.840 0.907 0.971 0.983 1.021 1.069 -

0.939 1.035 1.110 1.196 1.297 0.970 1.061 1.133 1.169 1.296 0.968 0.975 0.980 1.023 1.001 0.766 1.020 1.154 1.184 1.279 1.387 1.276 0.832 1.103 1.153 1.232 1.296 1.382 1.276 _

0.921 0.925 1.001 0.961 0.987 1.004 1.000 0.988 0.975 1.063 1.068 1.352 1.217 1.368 j 1.383 1.276 1.376

i 1.013 0.995 0.977 0.954 0.994  ;

Fuel Assembly: HX373 i_ Distance Above

[ Bottom of Fuel: 15 inches _

Exposure: 11333 MWD /MTU i

Void History:

0%

T T

=

=

m i

e 2

i

r. =

J

Page 64 _

s l

Figure 3.14 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 0.411 0.484 0.622 0.886 1.140 Gamma Scan cap 0.434 0.468 0.599 0.850 1.014 CPM 0.948 1.035 1.039 1.042 1.124 Ratio 0.455 0.646 0.713 0.890 0.972 1.098 1.114 0.468 0.622 0.684 0.869 0.954 1.089 1.092 0.973 1.039 1.043 1.024 1.019 1.008 1.020 0.699 0.916 0.985 1.051 1.151 0.684 0.873 0.962 1.044 1.128

1.022 1.049 1.024 1.006 1.020 0.597 0.920 0.975 1.116 1.169 1.336 0.599 0.869 0.962 1.115 1.217 1.388 0.997 1.059 1.013 1.001 0.960 0.963 1.001 1.029 1.089 1.164 1.225 0.954 1.044 1.115 1.139 1.284 1.049 0.985 0.976 1.022 0.954 L 0.879 1.101 1.181 1.202 1.227 1.295 1.200 0.850 1.089 1.128 1.217 1.284 1.374 1.310 1.034 1.011 1.047 0.987 0.956 0.943 0.916 1.097 1.107 1.353 1.217 1.286 1.014 1.092 '

1.388 1.310 1.423 1.082 1.013 0.975 0.929 0.904 Fuel Assembly: HX373 Distance Above Bottom of Fuel: 63 inches Exposure: 12762 MWD /MTU Void History: 41%

MIM I El MI

Page 65

?

Figure 3.15 COMPARISON OF LPM CALCULATED AND HATCH-1 La-140 ..

GAMMA SCAN MEASURED PIN POWER DISTRIBUTION .

_ wide gap wide 1.202 1.050 1.067 1.117 1.090 Gamma Scan

_- gap 1.135 1.019 1.028 1.112 1.134 CPM 1.059 1.031 1.038 1.005 0.961 Ratio 2 1.076 1.106 0.973 1.112 1.062 1.076 0.976 1.019 1.019 0.935 1.062 1.054 1.094 1.034 a 1.056 1.086 1.041 1.047 1.007 0.983 0.944 0.986 1.048 0.940 0.931 0.940 5 0.935 0.999 0.938 0.931 0.927 3

1.054 1.049 1.002 1.000 1.014 1.083 1.113 0.986 0.882 0.917 1.011

._ 1.028 1.062 0.938 0.872 0.921 1.066 1.054 1.048 1.051 1.012 0.996 0.948 1.109 0.950 0.877 0.834 0.902 1.054 0.931 0.872 0.867 0.917 8

1.052 1.020 1.006 0.961 0.984 1.109 1.106 0.935 0.903 0.896 0.926 0.860 1.112 1.094 0.927 0.931 0.917 0.967 0.940 0.998 1.011 1.009 0.981 0.977 0.957 0.915 1

1.111 0.980 1.017 0.830 0.913 1.134 1.034 +

1.066 0.940 1.046 0.980 0.948 0.954 0.883 0.873 En

  • K-I_ Fuel Assembly: HX373 h

- Distance Above Bottom of Fuel: 129 inches

[ Exposure: 10268 MWD /MTU Void History: 70%

2 5

E M

E _.

Page 66 Figure 3.16 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.176 1.041 1.015 1.077 1.000 Gamma Scan gap 0.994 0.918 0.964 1.044 1.026 CPM 1.184 1.134 1.052 1.032 0.975 Ratio 1.030 1.080 0.992 1.130 1.098 1.082 0.937 0.918 0.976 0.918 1.068 1.065 1.104 0.991 1.122 1.106 1.080 1.058 1.031 0.980 0.946 0.999 1.143 1.055 0.993 1.005 0.918 1.005 0.978 0.979 1.005 1.088 1.136 1.078 1.014 1.000 1.045 1.140 1.006 0.913 0.941 1.011 0.964 1.068 0.978 0.946 0.992 1.091 1.084 1.067 1.028 0.965 0.949 0.927 1.090 0.949 0.927 0.914 0.932 1.065 0.979 0.946 0.923 0.996 1.024 0.969 0.980 0.990 0.936 1.038 1.081 1.019 0.931 0.894 0.941 0.827 1.044 1.104 1.005 0.992 0.996 1.035 0.944 0.994 0.979 1.014 0.939 0.898 0.909 0.876 0.951 0.895 0.997 0.816 0.888 1.026 0.991 -

1.091 0.944 1.099 0.927 0.904 0.914 0.864 0.880 Fuel Assembly: HX141 Distance Above Bottom of Fuel: 15 inches Exposure: 11239 MWD /MTU Void History: 0%

Page 67 Figure 3.17 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.157 1.018 1.079 1.170 1.141 Gamma Scan gap 1.146 1.029 1.038 1.099 1.082 CPM 1.010 0.989 1.039 1.065 1.054 Ratio 1.001 1.050 0.965 1.130 1.104 1.145 1.030 1.029 1.027 0.944 1.066 1.056 1.090 1.005 0.973 1.022 1.022 1.060 1.045 1.051 1.025 0.930 1.031 0.986 1.002 1.047 0.944 0.981 0.949 0.943 0.961 0.985 1.051 1.039 1.062 1.090 1.006 1.041 0.950 0.929 0.968 1.091 1.038 1.066 0.949 0.896 0.939 1.050 0.969 0.976 1.001 1.037 1.030 1.039 1.003 0.939 0.896 0.921 0.947 1.056 0.943 0.896 0.863 0.937 0.950 0.995 1.000 1.068 1.010 0.995 1.009 0.945 0.912 0.926 0.968 0.914 1.099 1.090 0.961 0.939 0.937 0.976 0.921 0.906 0.926 0.984 0.971 0.988 0.992 0.992 0.958 0.891 0.997 0.868 0.937 1.082 1.005 1.050 0.921 0.994 0.885 0.887 0.949 0.942 0.943 Fuel Assembly: HX141 Distance Above Bottom of Fuel: 63 inches Exposure: 12131 MWD /MTU Vold History: 407.

Page 68 Figure 3.18 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.137 0.978 1.016 1.109 1.134 Gamma Scan gap 1.133 1.019 1.029 1.113 1.133 CPM 1.003 0.960 0.987 0.996 1.001 Ratio 0.980 1.011 0.960 1.106 1.079 1.147 1.051 1.019 1.021 0.939 1.065 1.057 1.098 1.036 0.962 0.990 1.022 1.039 1.020 1.045 1.014 0.917 1.039 0.959 0.934 0.979 0.939 1.002 0.941 0.935 0.922 0.976 1.037 1.019 0.999 1.062 1.040 1.050 0.966 0.890 0.911 1.007 1.029 1.065 0.941 0.873 0.924 1.070 1.010 0.986 1.026 1.020 0.986 0.942 1.046 0.944 0.878 0.770 0.913 1.057 0.935 0.873 0.792 0.918 0.989 1.009 1.006 0.972 0.994 1.086 1.126 0.948 0.939 0.929 0.949 0.914 1.113 1.098 0.922 0.924 0.918 0.970 0.943 0.976 1.026 1.028 1.016 1.012 0.979 0.969 1.109 1.010 1.056 0.897 1.007 1.133 1.036 - 1.070 0.943 1.048 0.979 0.975 0.987 0.951 0.961 Fuel Assembly: HX141 Distance Above Bottom of Fual: 129 inches Exposure: 10483 MWD /MTU Void History: 66%

Page 69 Figure 3.19 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.091 0.965 1.010 1.001 0.894 Gamma Scan gap 1.052 0.952 0.994 1.070 1.041 CPM 1.037 1.013 1.016 0.935 0.859 Ratio 0.991 1.044 0.931 1.099 1.067 1.060 0.883 0.952 1.004 0.927 1.078 1.070 1.107 0.989 1.041 1.040 1.004 1.019 0.997 0.957 0.893 0.943 1.084 0.986 0.973 0.975 0.927 0.989 0.968 0.967 0.987 1.017 1.096 1.018 1.006 0.988 0.994 1.139 0.996 0.912 0.981 1.012 0.994 1.078 0.968 0.928 0.973 1.075 1.000 1.056 1.028 0.983 1.009 0.941 1.102 0.969 0.918 0.954 0.972 1.070 0.967 0.928 0.899 0.975 1.030 1.002 0.989 1.061 0.997 1.07: 1.124 1.047 0.968 0.947 1.033 0.929 1.07a 1.107 0.987 0.973 0.975 1.015 0.924 1.001 1.015 1.061 0.995 0.972 1.018 1.005 0.993 0.918 1.072 0.956 0.997 1.041 0.989 '

1.075 0.924 0.990 0.954 0.929 0.997 1.034 1.007 Fuel Assembly: HX169 Distance Above Bottom of Fuel: 15 inches Exposure: 9874 MWD /FtrU Void History: 0%

Page 70 Figure 3.20 COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION wide gap wide 1.036 0.930 1.016 1.023 0.976 Gamma Scan gap 1.025 0.954 0.985 1.055 1.055 CPM 1.010 0.975 1.031 0.969 0.927 Ratio 0.969 1.019 0.924 1.088 1.051 1.056 0.918 0.954 0.983 0.931 1.057 1.054 1.091 1.010 1.016 1.037 0.993 1.029 0.997 0.968 0.909 0.977 1.056 1.003 0.966 0.967 0.931 0.991 0.969 0.967 0.986 1.050 1.066 1.035 0.999 0.981 1.013 1.112 1.010 0.931 0.959 0.996 0.985 1.057 0.969 0.929 0.973 1.082 1.028 1.052 1.042 1.003 0.986 0.921 1.086 0.990 0.932 0.926 0.959 1.054 0.967 0.929 0.897 0.974 1.030 1.024 1.004 1.032 0.985 1.091 1.119 1.048 0.972 0.950 1.006 0.894 1.055 1.091 0.986 0.973 0.974 1.013 0.958 1.034 1.026 1.063 0.999 0.975 0.993 0.933 1.088 1.004 1.065 0.909 0.962 1.055 1.010

  • 1.082 0.958 1.030 1.031 0.994 0.985 0.949 0.934

\

Fuel Assembly: HX169 Distance Above Bottom of Fuel: 63 inches Exposure: 14145 MWD /MTU Void History: 38%

Page 71 Figure 3.21 .

COMPARISON OF CPM CALCULATED AND HATCH-1 La-140 GAMMA SCAN MEASURED PIN POWER DISTRIBUTION E

b wide

$ gap wida 1.179 1.016 1.065 1.154 1.152 Gamma Scan

- gap 1.151 1.031 1.036 1.119 1.143 CPM 1.024 0.985 1.028 1.031 1.008 Ratio 1.034 1.051 0.972 1.099 1.061 1.142 1.036 1.031 1.025 0.942 1.066 1.057 1.096 1.038 1.003 1.025 1.032 1.031 1.004 1.042 0.998 0.952 1.070 0.945 0.945 0.940

(- 0.942 1.002 0.941 0.931 0.916 f 1.030 1.011 1.097 1.068 0.965 1.005 1.015 0.888 1.026 0.910 1.057 1.036 1.066 0.941 0.867 0.917 1.065 0.994 1.029 1.026 1.024 0.992 0.993 1.081 0.932 0.880 0.772 0.912 1.057 0.931 0.867 0.784 0.909 3 1.023 1.001 1.015 0.985 1.003

! 1.088 1.111 0.927 0.903 0.893 0.924 0.890 1.119 1.096 0.916 0.917 0.909 0.961 0.940 .

0.972 1.014 1.012 0.984 0.982 0.962 0.947

, 1.077 0.986 1.034 0.866 0.963

? 1.143 1.038 -

1.065 0.940 1.047 0.942 0.949 0.971 .1 0.922 0.920 e

o Fuel Assembly: HX169 Distance Above E Bottom of Fuel: 129 inches Exposure: 10450 MWD /MTU Void History: 69%

m M

5 b

F .

E E

Figure 3.22' FISSION RATE DISTRIBUTION IN AN 8x8 BWR BOX OF THE Pu ISLAND TYPE T=245 C Wide gap Wide gap 0 UO rods

+1.9 l c _._ _ _

+0.7 +0.1  ! I i i r--- J '- - -

-0.5 +2.0 MO: rods  !

I +1.0 I i  !

I I

+0.6 +2.9 g+0.1 +0.9 g

! . r--'

I I l-0.6 -1.6 +0.6 l L. _ _ ., ,. .J l i

-0.5 -0.5 l +0.6 +0.5 g + 0.5 -1.1 L ._ _ _ _ _ a

-2.8 -0.6 -1.3 -1.0 -2.7 The figure shows PCPM Pexp .100 for all rieasured rod positions Pexp Experimental uncertainty (1a) in MO 2 rods : 1.4 %

.. .. UOr  : 0.7%

" for the average fission rate in MO: rods relative to UO rods : 1.6 %

Calculated keff - 1.001

  • This figure is reproduced based on the data available in Reference 2.

Figure 3.23*

FISSION RATE DISTRIBUTION IN A 15x15 PWR MO: ASSEMBLY WITH WATER HOLES AND ABSORBER RODS T=245 C Central water hole Absorber rod

/ /

-y-yu

+1.0 MO: rods

+3.1 -0.1 +1.2

, -3.3 -1.3 -0.5 ___

+2.2 -0.4 -0.S -0.4 is// > -

~~- -~~

+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 The figure showsPCPM Pexp 100 for all measured rod positions Pexp Experimental uncertainty (1a) : 1.4%t Calculated keff - 0.999 tNot including geometrical uncertainties (compare text)

  • This figure is reproduced based on the data available from Reference 2.

Figure 3.24*

FISSION RATE DISTRIBUTICN IN A 14x14 PWR MO: ASSEMBLY SURROUNDED BY UO ASSEMBLIES T=240 C Center I

+1.8 High enriched MO: rods l

___ l

-0.3 h_- l v- 1

+2.1

/ l Water holes l g g 2

+1.0 +0.9 l

-J f

, , -- e

_ _- i e

-1.7 ~_ _- _ _r-

_ _- l ___-

I

+1.2 -0.5

+0.7 l -1.5

+1.5 l I

+0.9 Low enriched MO: rods -3.7 l l

+0.6 Enr. UO rods -0.8

+1.4 -0.4

-0.6 _-'- -

e a e e e The figure shows PCPM Pexp .100 for all measured rod positions.

Pexp The fission rate was normalized separately for each type of assembly. The average fission rate in the MO2 assembly relative to the rate in the UO2 assemblies predicted by DIXY was 1.9% lower than the measured ratio.

Experimental uncertainty (1a) for each type of fuel separately: 0.8%

/

Experimental uncertainty (la) for the average fission rate in MO2 rods relative to UO rods: 1 1.4 %

Calculated keff = 0.997

  • This figure is reproduced based on the data from Reference 2.

Figure 3.25*

CPM CALCULATED K-EFFECTIVE vs. MEASURED BUCKLING CPM CALCULATED K EFFECTIVE VS. MEASURED BUCKLING 1.02 , ,. , , , , , , , , ,

0:ERADA DATA

  1. :TRX DATA 1.01 -

o _

o un D o 0 h

u.

o o o 1.00 g

, ,s o o Q sa o 0.99 -

0.98 I I I I I I I I I I I O 20 40 60 80 100 BUCKLING (M -2)

'This figure is reproduced based on the data from Reference 2.

L

YANKEE. THE ISOTOPIC RATIO Pu-239/Pu-240 COMPARISON BETWEEN EPRI-CPM AND EXPERIMENT 9.0 ,

8.0

  • U-7.0 q

y.

+.

o j 6.0 ** -

S N

+

3 h

O 5 .0 5 *

.. s 4.0 w .

3.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 F.P. vol. wgt. number density x 10s 0 10 20 30 mwd /kgU

'This figure is reproduced based on the data from Reference 2.

Figure 3.27*

YANKEE. THE ISOTOPIC RATIO Pu-241/Pu-242 COMPARISON BETWEEN EPRI-CPM AND EXPERIMENT 10.0 ..

~

9.0

  • 1 8.0 S 7.0 E -

E 3

b N 6.0 .,

5 '..

5.0  :

is s' .

4.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 F.P. vol, wgt. number density x 13s I  :

0 10 20 30 mwd /kgU

  • This figure is reproduced based on the data from Reference 2.

Figure 3.28' YANKEE. THE ISOTOPIC RATIO Pu-240/Pu-241 COMPARISON BETWEEN EPRI-CPM AND EXFERIMENT 8.0 7.0

.+ ,

6.0 o *

s E 5.0 4
a.
  • B
  • sg 4.0 f
  • (.

3.0 2.0 ,

-. + -

'. 0.0 5.0 10.0 15.0 20.0 25.0 30.0 F.P. vol. wgt number <iensity x 10s

! I 1 1 0 10 20 30 mwd /kgU

  • This figure is reproduced based on the data from Reference 2.

Page 79 4.0 Summary The lattice physics code CPM is used in GPUN's steady state BWR anal-ysis methods to provide cross sections, peaking factors, delayed neu-tron fraction, and fuel constants and detector response factors for the 3D core simulator. The verification work performed by GPUN and EPRI provides the basis of onfidence in the CPM methodology and GPUN's lattice physicc model.

Comparisons of CPM calculated pin power distribution and the measured l gamma scan data for the two Oyster Creek and four Hatch-1 fuel assem-l l blies show excellent agreement. In general, the gamma scan comparison is better at the wide-wide corner than at the narrow-narrow corner.

CPM tends to underestimate the power at the wide-wide corner and over-estimate at the narrow-narrow corner. This is a result of the fact that the gamma scan data have not been corrected for core flux tilt l effects caused by control rod and fuel exposure. No attempt has been made to correct the measurement data for global power tilt because of the complexity involved. The CPM / Monte Carlo comparison also agree very well in pin power distribution. These comparisons form a sound basis for predicting power distribution and peaking factors. The CPM / Monte Carlo comparison also demonstrates the appropriateness of l transport theory, spectrum treatment, and strong absorber treatment l

l

Page 80 in CPM. The benchmarking work performed by EPRI demonstrates that the code can be expected to predict criticality reasonably. The EPRI work also demonstrates the adequacy of pin cell treatment, depletion cal-culation, and the nuclear data library.

i

l Page 81 l 5.0 References

1. EPRI, ARMP System Documentation, CCM-3, Part II.2, Chapter 6, "The Collision Probability Module EPRI-CPM," Sept. 1977.

i

2. EPRI, ARMP System Documentation, CCM-3, Part I.1, Chapter 5, "EPRI CPM Benchmarking," Sept. 1977.
3. Exxon Nuclear Company, XN-382, " Final Report Fuel Rod Examination of 8x8 Fuel Assemblies at Oyster Creek January 1976," May 1976.
4. Exxon Nuclear Company, XN-NF-77-49, "Non-destructive Examinatica of Exxon Nuclear Fuel at the Oyster Creek Reactor Spring 1977,"

Nov. 1977.

5. Exxon Nuclear Company, JN-72-1, "JNC Reload Fuel Assemblies for JCP&L Co. Oyster Creek Unit No. 1 (Type III Assemblies) Design Report," Feb., 1972.
6. EPRI NP-511, " Gamma Scan Measurements at Edwin I. Hatch Nuclear Plant Unit 1 Following Cycle 1," August, 1978.

Page 82

7. EPRI, ARMP System Documentation, CCM-3, Part II.2, Chapter 7, "MICBURN, Microscopic Burnup in Gadolinia Fuel Pins," Sept. 1977.
8. Battelle Northwest, BNWL - 1433, "BMC-1: The Battelle Monte Carlo Code," D. H. Thomsen and T.M. Traver, June 1970.
9. AB Atomenergi, Sweden, AE-RF-71-267, " Critical Experiments up to 245'C with H2O - Moderated - UOz - rod Lattices in KRITZ," ,

R. Persson, E. Blomsjo, and M. Edenius, 1971.

10 J. R. Brown et. al. , " Kinetic and Buckling Measurements of Lat-tices of Slightly Enriched Uranium or UO2 Rods in Light Water,"

WAPD-176, 1958.

11. R. D. Leame r e t . al . , Pu0 2 - UO Fueled Critical Experi-ments," WCAP-3726-1, 1967.

l

12. R. J. Nodvik, " Supplementary Report on Evaluation of Mass Spec-trometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086 (1969).

l l

Page 83

13. R. J. Nodvik, "Saxton Core II Fuel Performance Evaluation,"

Part II, WCAP-3385-56.

14. GPUN, Technical Data Report 437, "PSMS NODE-B (Offline IBM Ver-sion): Three Dimensional Reactor Simulator for Oyster Creek Core Analysis," R. V. Furia, June 1983.
15. EPRI NP-562, " Core Design and Operating Data for Cycle 1 of Hatch 1," January, 1979.

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