ML20154H635

Rev 2 to Nuclear Analysis Reload Core Analysis Methodology, Neutonic Design Methods & Verification

ML20154H635
Person / Time
Site:	Fort Calhoun
Issue date:	04/30/1988
From:	OMAHA PUBLIC POWER DISTRICT
To:
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ML20153E168	List: ML20153E164 ML20153E171 ML20154H635 ML20154H655
References
OPPD-NA-8302-NP-R02, OPPD-NA-8302-NP-R2, NUDOCS 8809220078
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Omaha Public Power District Huclear Analysis (

Reload Core Analysis Methodology '

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Neutronics Design Methods And Verification I t

i CPPD.NA 8302.NP  ;

Rev 02 April 1988 l

Copy No,

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l ABSTPACT s ,

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This document is a Topical Report describing Omaha Public Power District's re- '

) load core neutronics design methods for application to Port Calhoun Station f

. Unit No 1. i 1

The report addresses the District's neutronics design methodology and its oppli- l t

cation to the calculation of specific physics parameters for reload cores- In  :

a addition, comparisons of results obtained using this methodology to results I from experimental sessurssents and independent calculations are provided, ,

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OPPD NA 8302 NP, Rev. 02 i L i

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Table of contents Section Eagg h

1.0 INTRODUCTION

1 2.0 BASIC PHYSICS MODELS 1 2.1 Neutron Cross-Sections 2 2.2 Diffusion Theory Models 3 2.2.1 MC 3 l

2.2.2 ROCS 4 2.2.3 QUlX 5 3.0 FORT CALHOUN PHYSICS MODELS 7 3.1 Neutron Cross-sections 7 3.2 Diffusion Theory Models 8 3.2.1 MC 8 3.2.2 ROCS 9 3.2.3 QUIX 9 4.0 APPLICATION OF PHYSICS METHODS 10 4.1 Radial Peaking Factors 10 4.2 Reactivity Coefficients 11 4.3 Neutron Kinetics Patameters 12 4.4 Dropped CEA Data 12 4.5 CEA Ejection Data 14 4.6 CEA Reactivity 14 4.7 CEA Withdrawal Data 16 4.8 Reactivity Insertion for Steam Line Break Cooldown 17 4.9 Asymmetric Steam Generator Event Data 18 4.10 QUIX Calculations 19 5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUM STATION 19 5.1 ROCS MC. Generated Albedos 20 5.2 Planar Radial Peaking Factors 21 5.3 Integrated Radial Peaking Factors 22 OPPD NA 8302 NP. Rev. 02 11

Table of Contents-(Continued) i

. i I, Section Eggg.

l 5.0 VERIFICATI0*! 0F NEUTRONICS M09E1.S FOR FORT CAMOUN STATION 5.4' The District's Ongoing Benchmarkin5 Program 22 5.5 Sununary 23

6.0 REFERENCES

24 i

APPENDIX A i 4

CEPAX/DIT Verification Program APPENDIX B '

Ongoing Benchmarking Pro 5 ram Historiesl Inforn.acion I

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I LIST OT TABLES I TABLE TITLE EAGE 5-1 Unrodded HZP Critical Boron Cencentrations Calculations 26 52 Low Power Physics Isothermal Temperature Coefficients 27 5 'a Comparison of Calculated and Measured Isorhermal Temper. 28 ature Coefficiants 54 Comparison of Calculated and Measured Power coefficients 30 5-5 Cycle 11 CEA k'orths 31 I

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OPPD NA 8302 NP, Rev. 02 iv,

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i OMAHA PUBLIC ?O'n'ER DISTRICT NEUTRON.CS DESIGN METHODS AND VERIFICATION i

REVISION DATE 00 September 1983 01 November 1986 02 April 1988 i

OPPD NA 8302 NP Rev. 02 Y

Omaha Public Power District Reload Core Analysis Methodology Neutronics Design Methods and Verification

1.0 INTRODUCTION

The District's neutronics design calculation methods are described along with results obtained when these methods are compared to experimental measurements and independent calculations. The discussion of calcula-tional methods includes descriptions of the basic computer codes and pro-cedures for applying these codes. Comparison of the calculations to measurements and independent calculations are performed using the same codes nd computational methods used in the Fort Calhoun reload core de-

• sign efforer The basic physics models, supplied by Combustion Engineer-ing (CE), ar> described in Section 2.0. Section 3.0 describes the Dis-trict's application of these models to the Fort Calhoun reactor. Sec-tion 4.0 presents the application of these physics models to the reload core analysis. Sec. tion 5.0 discusses the District's latest verification program whl-h includes the recent cycle by cycle comparisons of District calculated data to measured data and data from independent calculations.

Section 6.0 contains the individual references. Appendix A discusses the CEPAK to DIT cross-sections Verification Program, and Appendix B pre-sents the historical information, collected from beginning of core life, from previous ongoing benchmarking programs mentioned in previous neu-tronics methodology submittals.

2.0 BASIC PHYSICS MODEl.S The District's neutronics design analysia for the Fort Calhoan core is based on a combination of multi grvup neutron spectrum csiculations, which provide cross sections cppropriately averaged ove: a few broad energy groups and few group one , two and three- dimens onal diffusion theory calculations, which result in integral and differential reactivity effects and power distributions. Calculations are performed with the aid of computer programs embodying analytical procedures and fundamental nuclear data consistent with the current State of the Art.

s CPPD NA 8302 NP. Rev. 02 Page 1 of 31  ;

2.0 BASIC PHYSICS MODELS (Continued) 2.1 Neutron Cross-sections The data base for both fast and thermal neutron cross sections is de-rived from ENDF/B-IV with changes recommended by the cross section Evaluation Working Group (Reference 2 1). These recommendations con- i sist of changes to the shielded resonance of U238, and the Watt fission spectrums of U 235 and Pu239, and changes in A for U 235 and Pu239 Few group cross sections, for subregions of the core

• hat are represented in spatial diffusion calculations, (e.g..

fuel pin cells, moderator channels, structural member cells, etc.)

i are calculated by the DIT lattice program. These cross sections are generated as a function of fuel temperature and moderator tempera-ture to accommodate the temperature feedback routines within the diffusion theory models.

The DIT code performs all the functions of the traditional transport methods which attempt to represent the complexities of the PWR fuel assembly geometry, including neutron energy spectrum interactions in the fuel, control rods, control rod locations (water holes), burn-able absorber rods, and incore flux detectors. The essential fea.

ture of DIT, which distinguishes it from the traditional methodology is that the spectrum spatial averaging procedures are based on calcu-lations in two dimensional geometry. Hence few approximations to

! the geometry representation are necessary. The use of nodal trans-l port theory has made it feasible to retain discrete pin geometry in both the fine and broad energy group calculat.ons. A more complete description of the DIT procedures for generating few. group neutron cross sections can be found in References 2 2 and 2 3.

Previously, the District utili:ed the CEPAK program to produce few-group cross sections. These cross sections were also generated as l i

functions of faal and moderator temperature. Comparisons of calcu- l lated and measured data reported in Section 5.0 include calculations l performed using the CEPAX program.

OPPD NA 8302.NP. Rev. 02 t Page 2 of 31 ,

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2.0 BASIC PHYSICS MODELS (Continued) 2.1 Neutron Cross-Sections (Continued)

The CEPAK program is the synthesis of a number of computer codes, many of which were developed at other labora :ories, e.g. , FORM, THERMOS and CINDER. These programs are interlinked in a consistent way with inputs from differential cross section data from an exten-4 sive library. A description of the CEPAK procedures used to gener-ate few-group neutron cross sections can be found in Reference 2 4 i

2.2 Diffusion Theerv Models The diffusion theory models package used to calculate core physics

! parameters for Fort Calhoun Station consist of the MC, ROCS, and I QUIX computer codes. The MC (fine mesh) and ROCS (coarse mesh) codes can be executed in one, two or three dimensions to calculate static and depletion dependant parameters such as reactivities.

flux, nuclide and power distributions and CEA worths. The QUIX code j

is executed in one dimension to calculate axial power distributions and CEA worth, Previously, the District utilized the PDQ X code to produce tradi-tional two dimensional fine mesh core design information. The PDQ X i

l program was an extension of PDQ-7 and HARMONY programs mentioned in Reference 2 5. Comparisons of calculated data between PDQ X and MC are reported in Section 5.0.

2.2.1 gg i

j The MC program is a fine mesh method used to solve the two-group neutron diffusion equation. MC uses the 3 D coarse j mesh analysis (ROCS) to recover local informatton on power.

! burnup and flux by performing fine mesh, imbedded diffusion I

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2.0 BASIC PHYSICS MODEl.S (Continued)

, 2.2 Diffusion Theorv Model (Continued) 2.2.1 lig (Continued) theory calculations within the coarse mesh nodes. The capa-bilities of MC offer a more computationally efficient alter-native to conventional fine mesh diffusion theory computer t codes (i.e., PDQ X) which in practice are limited to 2 D f core analyses (Reference 2 6). MC also eliminates the PDQ-X problem of representing large gradients in the vicinity of j CEA guide tubes (water holes) and burnable absorber pins.

The PDQ X diffusion theory formulation does not provide the correct flux levels for fuel adjacent to water holes or burn- l able absorbers, in fact, when PDQ X makes adjustments to ab-sorption and removal cross sections necessary to obtain cor-rect reaction rates for non fuel cells, results are made worse. By using the imbedded nodal calculational technique.

MC removes the need to use PDQ X type adjustments.

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hC employs macroscopic (static) and microscopic (depletion)

cross section data generated by methods described in Section

. 2.1.  :

2.2.2 EQC1 f

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The ROCS program is a coarse mesh two group solution of the l [

4 neutron diffusion equation based upon a mesh centered higher [

order finite difference formulation. It incorporates closed .

channel thermal hydraulic modeling into its evaluation of l

the interaction of neutron flux effects and the macroscopic t i

j physical and thermal properties of distributed materials.

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Because of MC dependency on ROCS for input information and l I ,

I t l OPPD NA 8302 NP. Rev. 02 Page 4 of 31 i

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' 2. 0 BASIC PHYSICS MODELS (Continued) 2.2 Diffusion Theorv Model (Continued)  ;

2.2.2 EQC.1 (Continued) the ROCS coarse mesh nodal structure, ROCS is more efficient i 1

than MC for evaluating a core's static and depletion depen-4 dent properties. ROCS also employs macroscopic (static) and  !

microscopic (depletion) cross sections generated by the methods described in Seccion 2.1. A more complete descrip-tion of the ROCS program is found in References 2 3 and 2 7. l 2.2.3 Q.U13 The QUIX program is a one dimensional (axial) representation  !

of the core used to determine static and time dependent reac-

j. civities and power distributions at selected stages of deple-tion. This program solves the neutron flux and associated

]

eigenvalue in problems containing up to 140 distinct regions or compositions with variable mesh intervals. The macro-

) scopic cross section distributions, fission product yields, and xenon and boron microscopic cross sections required as -

input to QUIX are obtained from a three dimer.tional ROCS

calculation. Local power density (fuel temperature) feed-f back is included by modifying the point wise macroscopic

. absorptien and removal cross se:tions, The change in cross- l l sections is represented by a function of the difference be-i i 4

j tween the local axial power density and the referenced power  ;

density. Moderator density feedback is included by relating {

changes in the macroscopic absorption and removal crocs sec.  !

tions to the local hydrogen number density which is calculat-ed from enthalpy at each axial segment. These cross sec- j tions are generated in such a way that the fuel and moder-l ator temperature coefficients calculated by QUIX are equal J  !

OPPD NA 8302 NP Rev 02 [

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2.0 BASIC PHYSICS MODELS (Continued) ,

2.2 Diffusion Theorv Model (Continued) 2.2.3 QH13 (Continued) '

to or conservative with respect to the fuel and moderator temperature coefficients calculated by ROCS. The axial reflector cross sections input to QUIX are determined in .

such a way that the steady state axial power distribution generated by QUIX matches the axial power distribution gen-ersted by ROCS. Details of the above trea:ments are given in Reference 2 8. -

In addition to the eigenvalue problem, QUIX will perform four types of searches to obtain a specific eigenvalue, viz. , a uniform poison search, bucklin5 search, CEA region boundary search, and a moderator density dependent poison ,

search. The uniform poison search assumes an axially con-stant macroscopic absorption cross section whereas the mod-erator density dependent poison search assumes a distributed macroscopic absorption cross section dependent upon the axial moderator density. The moderator density dependent search is used to simulate the reactivity effects of the soluble. boron in the reactor coolant.  !

E Through the use of rod shadowing factors, shape annealing factors and shape index biases, the QUIX program has the capability of simulating excore detector response expected during normal operation. The procedures used for thase i

' t simulations are described in Reference 2 9.  :-

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3.0 FORT Caul 0UN PHYSICS MODELS The District utill:es the basic CE physics models described in Section 2.0 to model the Fort Calhoun reactor core. The computer codes which embody these basic physics models are maintained on the CE computer system at Windsor, Connecticut. The District accesses these computer codes through a time sharing system. CE maintains all documentation and quality assur-ance programs related to these computer codes. The following paragraphs discuss the specifics of the Fort Calhoun models.

3.1 Neutren Cross Sections The two group neutron cross sections utiliced in the ROCS and MC models of the Fort Calhoun reactor core are generated using the DIT code. Cross sections have been generated for unshimmed ANF and CE fuel assemblies and shimmed ANT and CE fuel assemblies. The cross-sections have been generated for the District by CE and are based on information supplied by the District.

The cross-sections. utilized to model the Fort Calhoun reactor are in the form of universal table sets. The two group cross sections are generated as functions of enrichment, fuel temperature, moderator temperature, burnup and in the case of shimmed fuel assemblies, B4 C shim number density. The table sets are applicable over a fuel temperature range from room tempera *.ure to 1800'K and a mod-erator *emperature range from room temperature to 600*K. The fine l mesh table sets include explicit treatment of the pin cells immed-iately around the CEA guide tube (water hole) to properly account for the peaking of thermal flux in these water holes. Therefore, no corrections need be applied to the pin powers produced by the fine mesh model.

OPPD NA 8302 NP, Rev. 02 Page 7 of 31 0______ . __

3.0 FORT CAL.HOUN PHYSICS MODEl.S (Continued) 3.2 Diffusion Theorv Models The District utilizes the MC, ROCS and QUIX models described in Sec-tion 2.0. The District utilizes both two dimensional and three-dimensional ROCS /MC models. The QUIX model is a one dimensional model.

3.2.1 tiG The District's MC model is a two group fine mesh model which

, generates its solutions based upon a two dimensional deple-tion in the x-y plane. MC uses imbedded fine mesh calcula-tions within coarse mesh ROCS nodes to produce solutions in the axial (z) dimension. Each fuel pin cell and shim pin cell is represented by a sin 51e mesh point. The model also  !

. includes explicit representation of the CEA guide tube i (water holes). The macroscopic cross section to node assign-J ment is located in a geometry file which also provides shim  ;

loadings and uranium metal weight for depletion calcula-tions. The model is representative of the core between 15%

and 85% of full core height.

[

The MC model is used to simulate the expected mode of oper-ation in the cycle being analyzed. This calculation results in material distributions and radial peakin5 factors which l are used in the safety analysis and seepoint generation.

Unlike PDQ X, MC utilizes exposure and geometry information l.

t to create a library of precalculated coefficients for the I

incore monitoring system. I The mode c.! operation at the Fort Calhcun reactor is base i loaded operation. Base loaded operation consists of reactor l-i ,

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E 3.0 FORT CALHOUN PHYSICS MODELS (Continued) 3.2 Diffusion Theorv Models 3.2.1 tiQ (Continued)-

operation at or very near rated thermal power throughout the cycle. The lead CEA bank insertion is held to a minimum.

Historically the lead CEA bank at Fort Calhoun has been in-sorted less than 5% of the time whenever the reactor is at a steady power level. Reference 3-1 discusses the impact of operation with a time averaged lead *ack insertion. Due to its dependency on ROCS coarse mesh intormation, the model must be depleted in the same number and magnitude of ROCS time steps, which are typically depleted in time steps of 1,000 MWD /hTU.

-In order to use MC beginning with the District's Cycle 12 Reload License Submittal, MC prediction calculations were

, benchmarked against PDQ X prediction calculations for Cycles to and 11. An overview of the benchmarking results are re-ported in Section 5.0, 3.2.2 EQ.Q.S.

The District utilizes a three dimensional and a two dimen. f sional two group ROCS model. The boundary conditions are [

derived in accordance with the methodology discussed in Reference 3 2. t i

3.2.3 Q.V.13 j The District utilizes a one hundred and twenty five axial node QUIX model. The data for the QUIX model is obtained l from the three dimensional ROCS calculations.

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4.0 APPLICATION OF PHYSICS METHODS Previous sections have focused on the reactor physics models utilized by j the District to model the Fort Calhoun reactor. In this section, calcu-

• 1ations of the various core parameters used in the safety analysis are '

described. The main parameters considered are the radial peaking factors (FR and Fxy), the moderator temperature coefficient, the fuel temper- l ature or Doppler coefficient, the neutron kinetics parameters, CEA drop  !

t data, CEA ejection data, CEA scram reactivity, reactivity insertion for l the steamline break cooldown, radial peaking data for the asymmetric steam  ;

generator event, and axial power distributions.

4.1 Radial Peakinz Factorg The radial peaking factors FR and Fxy, are calculated using the MC and 3-D ROCS models. Values of FR and Fxy for both unrodded and rodded core configurations are obtained directly from the MC [

power distribution. Since the MC model utilizes a pin power correc-  !

tion edit implicitly accounting for the peaking of the thermal flux i in the CEA guide tubes (water holes) no correction is required to ,

the peaking factors calculated by MC. Unlike PDQ X, the values of F

xy and FR for unrodded and redded cores are reported as core peakin5 edits. MC reads in the 3 D ROCS power histories, calculates F

xy for the plane of depletion and then synthesizes the calcula- l tion plane by plane to obtain the maximum core F xy. MC also  !

I utilizes the 3 D ROCS power information to calculate F Rbased on j the axial integration of the planar power distribution obtained from ROCS. The uncertainties for the radial peaking factors are given in j Reference 4 1. j 1 i The physics models are used to calculate the expected values of TR  !

j t and Fxy. The actual values of TR and Fxy used in the safety "

l analysis are chosen to be conservatively high with respect to those f

j anticipated during the core life. I t

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4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.2 Reactivity Coefficienr1 The ROCS models are used to calculate the moderator temperature coefficient (MTC) and the fuel temperature coefficient (FTC). The MTC is defined as the change in reactivity per degree change in '

moderator temperature. Calculationally, the MTC at a temperature of T eod is determined by running three calculations; one at T eod' one at T mod &

10'F and one at Tmod 10'F. The MTC at a temperature of Tmod is the average of the two calculated values.

The reactivity change is calculated with the ROCS model by varying  :

the inlet temperature while holding all other parameters such as the '

fuel temperature and nuclide concentrations constant.

The FTC or Doppler coefficient is defined as a change in reactivity per degree change in the effective fuel temperature. The effect of l fuel temperature upon resonance neutron energy absorption is account-l ed for in the ROCS /MC models by means of powe.' feedback options. l ,

,. The representation of the variation in the few group cross sections f

with fuel temperature involves two main segments. The first is to represent the variation in cross section with fuel temperature, the second is to relate fuel temperature to reactor power density. The .

first portion is included in the basic methods employed to generate l the few group cross sections. The second portion requires establish >

ment of correlations between fuel temperature (i.e., effective fuel temperature to be used in generation of cross sections) and the reac. '

tor power density. The relationship between fuel temperature and l reactor power density employs direct fits to FATES (Reference 4 2)

fuel data. This method results in the fuel temperature correlation l 1

, for each fuel type which is both local power density and fuel expo- 1 l

sure dependcnt.  ;

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4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.2 Reactivity Coefficients (Continued)  !

The reduction in reactivity resulting from an increase in effective fuel temperature is determined by ROCS. Typically, a temperature interval of 50'F is used to determine this coefficient. t The physics models are used to calculate the expected values of the MTC throughout the cycle. The actual values of the MTC used in the j safety analysis are chosen to conservatively bound exp6cted values  ;

of the MTC. The measurements of the MTC made during the operation i J

of the reactor include uncertainties to assure that the actual HTC does not exceed the values used in the safety analysis. A fifteen '

percent uncertainty is applied to the Doppler coefficient when it iJ  !

l used in the safety analysis calculations. .

4.3 Neutron 1(inetics Parameters l

} The neutron kinetics parameters A, A and the neutron lifetime, t

! 1*, are calculated using Combustion Engineering's ROCS computer '

code. The technique utilized co calculate the kinetics para.

a meters and the neutron lifetime f.s based on first order perturbation ;

i theory. Decalls of the perturbation approach are discussed in I References 4.3 and 4.4 I 4.4 Droened CEA Data

(

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t The neutronics data unique to the dropped CEA analysis are the  ;

j valuee of FR and F xy following the drop of a CLA and the reac. (

i civity worth of the dropped CEA. The values of TR and Fxy in-

crease due to a large azimuthal tilt caused by the drop of a CEA and  !

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l 4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.4 Dronned CEA Data (Continued) 9 4

occur on the side of the ccre opposite the dropped CEA. Because the

~

maximus FR and Fxy occur far away from the dropped CEA, the  !

intra assembly power distribution is not perturbed. Therefore, the  !

"pose drop" value of Fg and F xy can be calculaced by multiplying the "pre drop" values of FR and F xy by the ratio of the assembly power after and before the drop of the CEA. This ratio is the  ;

4 distortion factor. The distortion factor ia defined as the ratio of the assembly RPD at a given power level and time in core life >

containing a dropped CEA to the same assembly RPD without a dropped CEA.

F The distortion factor and dropped CEA reactivity worth can be cal- I culated'using the 2 D or 3 D ROCS model. The 2 D ROCS cal'eulations yield the Fxy distortion factor as a function of CEA bank inser-

tion (i.e., ARO, Bank 4 In, Banks 4+3 In) and power level. The 3-D j Fg distortion factor is calculated for a specific CFA insertion

) and power level. Sufficient margin exists at the lower power levels

, such that F Rdependent DNBR calculations do not adversely effect i

opercting margin. The "post drop" value of FR using the 3 D FR l distortion factor is calculated by multiplyin5 the "prn drop" value l

t of FRf r the particular CEA insertion and power level by the 3 D f

FR distortion factor.

The 2 D and 3 D ROCS "post drop" power distributions.are calculated j with fuel temperature and moderator temperature feedback. The cal-culations assume that the core average Axial Shape Index (ASI) is l being controlled within the "constant ASI" limits in accordance with I the Fort Calhoun Operating Manual.

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8 r 4.0- APPLICATION OF PHYSICS METHODS (Continued) t 4.4 Droceed CEA Data (Continued)

An uncertainty is applied to the reactivity worth of the dropped CEA based on the verification contained in Reference 4 5. ,

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4.5 CEA E4ection Data I L

r The neutronics data unique to the CEA ejection analysis are the f

values pre ejected and post ejected radial peaking factors and the reactivity worth of the ejected CEA. The maximum past ejection radis1 peaking factor and maximum ejected CEA reactivity worths are calculated for the maximum CEA insertion allowed by the PDIL at HTP [

i and HZP. The neutronics parameters are calculated using HFP and HZP l

2 D ROCS and MC models. The post ejection radial peaking factor is

[

i calculated by multiplying the 2 D ROCS post ejection assembly RPD by '

the corresponding pin to box ratio from MC. The ejected CEA reactiv-icy worth is obtained directly from ROCS calculations. ROCS post ejection power distributions are calculated without moderator or i fuel temperature feedback. ,

,\

The post ejectior, va?.ue of F q which is obtained using MC, is cal- ,

culated by multiplying the post ejection value of Fxy by the max-imum value of F , the azimuthal tilt allowance, the augmentation j factor, the engineering heat flux factor, the fuel densification l j factor, and the F quncertainty documented in Reference 4 1. An l

uncertainty is applied to the ejected CEA worth.  ;

f I

l 4.6 CEA Reactivity I I I i

q The CEA reactivity calculations done in a reload core safety analy-

! sis are the calculation of the total reactivity of CEA's inserted 1

r i L 4

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4.0 APPLICATION OF PHYSICS METHODS.(Continued) 4.6 CEA Reactivitv (Continued) into the core during a reactor trip (CEA scram reactivity), the generation of the scram reactivity curves, and the calculation of required shutdown margin. -

The CEA scram reactivity worth at HZP is calculated by obtaining the net worth for all CEA's between the HZP PDIL CEA position and the fully inserted position and subtracting the worth of the highest worth stuck CEA. These calculations are done using the ROCS model.

An uncertainty is applied to the HZe CEA scram reactivity vo.th.

The HZP CEA scram reactivity for the CEA ejection transient is calculated in a similar fashion except that the worth of the ejected and highest stuck worth CEA's are subtracted from the net worth.

The scram CEA worth at HFP is calculated by obtaining the HTP net ,

worth for all CEA's between the HFP PDIL CEA position und the fully inserted position, subtracting the worr.h of the highast worth stuck CLA and subtracting the moderator void collapse allowance. The ther-mal hydraulic axial gradient reduction allowance, and the loss of worth between HFP and HZP are also subtracted from the HFP net worth for the scram CEA worth to be used in all transients except the four pump loss of flow event and tha steam lins break incident. These are not applied to the four pump loss of flow scram CEA worth be-cause the closest approach to the SAFDL during the four pump loss of flow event occurs prior to significant CEA insertion. These allow-ances are not applied to the steam line break (SLB) incident HFP CEA scram worth because the HFP SLB reactivity insertion curvas implicit.

ly account for these effects.

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4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.6 CEA Reactivity (Continued) l t

The moderator void collapse allowance is 0.0% op at BOC and 0.1%

as at EOC. The thermal hydraulic axial gradient reduction allow- I

, ance is 0.2% ap at BOC and 0.4% op at EOC. An uncertainty is applied to the HTP CEA scram reactivity worth. The HFP scram '

reactivity for the CEA ejection transient is calculated in a similar fashion except that the worth of the ejected and highest stuck worth CEA's are subtracted from the net worth. All CEA worth calculations assume the ASI is being controlled within the "constant ASI" limits in accordance with the Fort Calhoun Operating Manual. -

t I

The generation of the scram reactivity curves utilizes the methodol-  !

ogy discussed in Reference 4 6. l The calculation of the required shutdown margin is only performed at HZP since the shutdown margin at power is controlled by ths PDIL.  !

j. The avail 61e HZP shutdotm margin is equivalent to the HZP CEA scram 4

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reac r.ivi ty.  :

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4.7 fJ,A 'Jithdyaval Data I '

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The reactor core physics data unique to the CEA withdrawal analysis l is the maximu:n differential CEA worth. This is ':he maximuzs amount 7

of reactivity at any time in core life that can be added to the core l por inch of CEA motion. h n the maximum differential CEA worth is  !

j combined with the maximum CEA withdrawal race of 46 inches / minute, a ,

j conservative withdrawal rate expressed in tap /see is obtained  !

t and used as input to the CEA withdrawal analysis, i l

a

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.t, 4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.7 CEA Withdrawal DgI.g (Continued)

The maximum differential CEA worth is obtained for the sequential withdrawal of the CEA banks from the HZP PDIL to an all rods out con.

.,,. dition. The 3-D ROCS model is utilized to-calculate this parameter.

The calculations are performed assuming that the reactor is being controlled within the "constant ASI" limits in accordance with the Fort Calhoun Operating Manual.

4.8 Reactivity Insertion for Steam Line Break Cooldown f

The reactor core physics data unique to the steam line break trans-  !

isnt analysis is the reactivity insertion due to the cooldown of the moderator. There are two sources of this reactivity insertion. Thc f

first-is the positive reactivity insertion due to the increasing j density of the moderator as the cooldown progresses. The second is i i

the reactivity insertion due to the Doppler coefficient as the effee-l tive fuel temperature changa.s.

i

Reactivity insertions due to the moderator density increase and the '

j Deppler coefficient are both calculated using a full core ROCS i model. The axial leakage or buckling is adjusted such that the mod- i

erator temperature coefficient calculated by the ROCS modtal corre- l

] sponds to t.he most negative Technical Specification limit, The reac- (

civity insertion calculations are performed with all CEA's except i j the most reactive CEA inserted in the core, i

The moderator density reactivity insertion curve for the hoe zero i j power steam line break case is calculated by successively lowering i the inlet temperature of the ROCS model from 532'T and allowing only l moderator temperature feedback in the model. The calculations typi.  ;

}

cally result in a curve of reactivity insertion vs. moderator temp. [

l erature from a hoe zero power temperature of 532'T to 212*F. E f

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4.0 APPLICATION OF PHYSICS METh0DS (Continued) 4.8 Reactivity Insertion for Steam Line Break Cooldown (Continued)

The Doppler reactivity insertion curve for the hoe zero power case is also calculated by steadily decreasing the inlet temperature of the ROCS model. The fuel temperature feedback in the model allows the production of a curve of Doppler reactivity as a function of  ;

fuel temperature. All zero power calculations are performed assum-ing there is no decay heat and no credit is taken for local voiding in the region of the stuck CEA.

1 The moderator density reactivity insertion curve for the full power case is calculated by decreasing the power level and core average I average coolant temperature from full power to the hoe zero power inlet temperature and then successively lowering the inlet tempera-

] ture as in the hoc zero power case. Only moderator temperature feed. i back is utilized in the ROCS model. The Doppler reactivity inser.

j tion curve is calculated by a similar otocedure utilizing the fael temperature feedback in the model.

Since tha moderator reactivity instreion curve corresponds to an MTC which is ac the Technical Specifiestion limit, no additional uncer- -

j tainty is added ,to this curve. A fiftsen percett uncertainty ic applied to the Doppler reactivity insertion curve.

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4.9 Asymmetric Steam Generator Event Data f

i For the range of temperatures considered, the intra assembly peakutg [

does not vary as the inlet temperature is changed.

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4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.10 OUIX Calculations t

The District utilizes the QUIX model to perform various axial shape  !

analyses related to the generation of the reactor protective system setpoints. The QUIX calculations are carried out in accordance with j the methodology discussed in Reference 4 6. l I

l' 5.0 VERIFICATION OF NEUTRONICS MODEI.S FOR FORT CALHOUN STATION  !

i l The District has performed extensive verification of the neutronics models I used in the reload core analyses. The results of the previous District verification efforts were reported in References 5-1 and 5 2. These j efforts consisted of utilizing cross sections produced by CEPAX and DIT ,

t i

and confirming the District's ability to use the models with DIT cross. '

l sections. Extensive verification of the use of DIT cross sections was f also done by Combustion Engineering (CE) and reported in Referenea 5 3. l

} In order to present the most recent verification efforts in this arction.

l the previously nontionod verificatton programs have besn argsnited into an l i i appendix. The verifica. tion cf CE?A4 to DIT cross sections, censisting of j t

aredictions, of overall core reactivity, power distributions, reactivity [

t s

coefficients, CL\ worth and xenon reactivity, along with the appropriate j figures and tables, is reported in Appendix A.

4 i The most current neutreni:s vartfication program performed by the District {

1 involved the benchmarking of the fine mesh, imbedded nodal diffusion [

} theory code. MC. with PDQ X. The purpose of this program was to replace PDQ X with the more cost efficient MC code and demonstrate the District's I ability to effectively use MC models.

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5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) l This verification is in addition to the comprehensive verification of the methods done by CE and described in References 5 4 and 5 5. It is not the intention of the District to repeat CE's verification effort which in-ciudes a statistical assessment of the adequacy of the uncertainties used l

by both CE and the District. Rather, it is the District's intent to demon-strate that the District can adequately model the Fort Calhoun core and that the results of the District's verifier *an effort are consistent with those reported in References 5 4 and 3 5.

The District's verification using 40 in4 c.edictions made for cycles 10 and 11. Benchmarking ras 5 .. the predictions of ROCS /

MC generated albedo boundary v.h i tadial peaking factors (Fxy) and integrated radial peak. ,, The results of the verification effort include data lot e s X and MC predictions.

5.1 ROCS MC Cenerated Albedos c

rior to tha Cycle 12 R4 actor Physics Reload Analysis, tne Fort ca.1.houn Station core boundary conditions for ROCS model were gener-ated by CE using the CEFLACS computer coce. The MC code has the capability of calculating coarse mesh (ROCS) boundary albedos. The MC generated albedos are fed into ROCS prior to performing fuel de.

pletion.

A comparison of the MC generated albedos against the CEFLACS gener-ated albedos is necessary to confirm the validity of the MC gener-ated albedos. These radial power distribution differences between the MC generated albedos and CEFLACS generated albedos are within 3%. The differences at the periphery are consistently higher than tha ones for the inner assemblies. This behavior is mainly due to the direct effect imposed c.1 the peripheral assemblies by the virtue ,

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OPPD NA 8302 NP Rev. 02 Page 20 of 31

5.0 VERIFICATI0!i 0F 11EUTR0!!ICS MODELS FOR FORT CAUiOU!! STATIO!! (Concinued) 5.1 R0CS MC-Cenerated Albedos (Continued) of their location. The low neutron radial leakage core loading patterns employed ty the District result in a low neutron flux and power distribution at the periphery which in turn calculate higher peripheral power distribution differences. The magnitude of these

, differences is consistent with those reported by CE in Reference 5 5.

5.2 Planar Radial Peakine Factors MC is a 3 D code that performs 2 D depletion calculations. The third dimension is generated by synthesis with ROCS. Planar radial f peaking factors (Fxy) are thus calculated for the plane of inter. p est and then synthesized axially in order to determine the maximum i planar radial peaking factor in the core. [

The difference between the MC calculated and PDQ calculated Fxy's  ;

do not follow a unified pattern throughout the core. The planar [

radial peaking factor distributions are influenced by several fac. ,

cor:s, some of which are working in opposite directions. Combustion Engineering, in their independent, review of the District's bench-marking analysis, stated that the results of the District'.s berch. ,

marking of MC are consistent with similar work done at CE on other plants.

The planar radial peaking factors predicted by MC for the plane of depletion compara closely to PDQ predictions. MC predictions of corevide peaking factors, as expected, produce slightly more con-servative results than PDQ midplane predictions.

OPPD flA 8302 !iP, Rev. 02 Page 21 of 31

5.0 "ERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) 5.3 Intecrated Radial Peakine Factors MC performs the calculations of the integrated radial peaking fsc.

tors (FR) for all assemblies. The MC. calculated FR 's are com-pared to PDQ RF 's which are calculated by multiplying each assem.

l bly pin to box raeio by its corresponding relative power density from ROCS core edits. The calculation of the integrated radial peaking factors in MC is performed by multiplying the assembly axial integral of the planar radial power distribution.

The fluctuations between MC. calculated and PDQ. calculated FR 's cannot be represented by a uniform pattern throughout the core.

Similar to the planar radial peaking factors, the integrated radial peaking factor variations are influenced by a number of factors.

However, the overall findings show acceptable results. Combustion Engineering,, in their independent review of the District's MC benchmarking analysis, s steo that the results of the District's banchmarking of MC e.re cera45: eat wi';h aimilar work done at CE on other plants.

The FR 's predicted by MC for the plane of depletion compare close+

ly to PDQ predictions.

3.4 DLt, District's Qugg h pehturkine Proer n The data reported in this section consisco of both historical and current information for the District's ongoing benchmarking pro.

gram. The historical data, collected through cycle 10 and reported in Reference 5 2. is located in Appendix 8. The verification pro.

gram has been updated through mid Cycle 11. Tables within Section i have been updated to include only recent cycle by cycle information (Cycles 9, 10 and 11).

OPPD.NA.8302.NP. Rev. 02 Page 22 of 31

i 5.0 "ERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) ,

l 5.4 The District's Onreine Benehmarkine Prerram (Continued) i Both verification of program segments include information consisting of startup physics testing predictions, reactor testing analysis and !

a core follow effort. This prograr, will continue to provide verifi- !

cation data in the future.

5.5 so--arv l The District has an ongoing neutronics methodology verification pro- !

r gram. The results of this verification program for previous cycles j demonstrate the ability of the District to utilize the neutronics [

methods described in this document. f ll i

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6.0 REFERENCES

Seccion 2.0 References 21 ENDF 313 "Renchmark Testing of ENDF/B Data for Thermal Reactors, Archival Volume," July, 1981.

22 A. Jonsson, J. R. Rec and U. N. Singh, "Verificatinn ' e Fuel Assembly Spectrum Code Based on Integral Transport Th. '"

Trans. Am. Nucl. Soc., 28, 778 (1978),

23 CENPD 226 P "The ROCS and DIT Computer Codes for Nuclear Design,"

December, 1981, 24 System 80 PSAR, CESSAR, Vol. I. Chapter 4.3.3, Amendment No. 3.

June 3, 1974, 25 W. R. Cadve11. "PDQ 7 Reference Manual " WAPD TM 678. January, 1968, 26 S. F. Crill A. Jonsson and M. W. Crump. "Recent ivelopments in the ROCS /MC Code for Retrieving Local Power Information in Coarse.

Mesh Analysis," CNS/ANS International Conference on Numerical Methods in Nuclear Engineering, Montreal, 'anada. Septemoer 6 9, 1983.

27 T. C. Ober, J. C. Stark. I. C. Richard and J. K. Casper "Theory. l Capabilities, and Use of the Thrae Dimensional Reactor Operation and Centrol Simulator (ROCS)," Nucl. Sci. Eng., 64, 605, (1977).

28 System 80 PSAR, CESSAR, Vol.. 1, Appendix 4A, Amendment No. 3. June 3, 1974 {

29 CENPD 199 P, Revision 1 P, "CE Seenoint Methodology," Ap.tt 1982. [

112.1.b.u L 0 P e fe rene e s 31 CZhPD 199 P, Revision 1 P, "CE Set;:ir.c Me :hodology," April 1982.

4 32 CENPD 226 P, "The ROCS and DIT Computer Codes for br. clear Design,"  !

December, 1981.

Section 4.0 References -

41 CENPD 153, Revision 1 P A, "Evaluation of '<^"ttainties in the Nuclear Power Peaking Measured by the Self r red Pixed In Core Detector System," May, 1980.

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OPPD.NA 8302 NP. Rev. 02 Page 24 of ?1 i

l 5.0 LETERENCES (Continued)

Section 4.0 References (Continued) 4-2 "Development and Verification of a Fuel Temperature Correlation  !

for Power Feedback and Reactivity Coefficient Applicat; '," P. H.

Cavin and P. C. Rohr, Trann- 6L. HusL.12s . 30, p . 7 6 5, '78.

43 A. F. Henry, "Computation of Parameters Appearing in the Reactor e Kinetic Equations " WAPD 142 December 1955. l 44 R. W. Hardie W. W. Licke, Jr. , "PERT.V. A Two Dimensional Portur.

bacion Code for Fast Reactor Analysis," BNWL 1162.

45 CENPD 226 P, "The ROCS and DIT Computer Codes for Nuclear Design "

December, 1981.

46 CENPD 199 P. Revision 1 P, "CE Setpoint Methodology," April 1982.

Section 5.0 Refereneen J

51 CEN 242-(0).P. OPPD Responses to NRC Questions on Fort Calhoun cycle 8. February 18, 1983.

52 "Reload fa':e Analysis Methodology, Neutronics Design Methods and Verification " OPPD NA 8302 P, Rev. 01.

53 CENPD 153 P, "INCA /CECOR Power Peaking Uncertainty," May, 1980.

i 54 CENPD 226 P, "The ROCS and DIT Computer Codes for Nuclear Design," l l December, 1981.

55 S. T. Crill, A. Jonsson and M. W. Crump. "Recent Developments in the ROCS /MC Code for Retrieving Local Power Information in Coarse.

, Mesh Analysis," CNS/ANS International Conference on Numerical Methods in Nuclear Engineering, Montreal, Canada, September 6 9,

1983.

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TAB 1.E 5 1 Unrodded HZP Critical Boron Concentrations Calculations 3.D Measured ROCS C1ql.g eem f,.Qin 9 1518 10 1474 11 1502 .. ,

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OPPD NA 8302.NP, Rev. 02 Page 26 of 31 l

TABL2 5 2 Low Power Physics Isothermal Temperature Coefficients Boron DIT Concentration Measured ROCS Cy,q1g (eem) (ar /* r) (ar/'r)

~ ~

9 1457 0.30 x 10*4 10 1457 0.23 x 10*'

11 1496 0.20 x 10*' ,

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OPPD NA 8302 NP, Rev, 02 Page 27 of 31

_ , , . . . , _ - . _ - _ , _ _ - - - - - . - . _ - _ _ _ . . . , - _ _ . . _ _ . ~ . , --.__.v. _ - - . - - - - - _ , - ---

TABLE 5 3 Comparison of Calculated and Measured Isother:nal Temperature coefficients BOC Calculated '

Critical Boron Measured DIT ROCS Percent of Concentration ITC ITC ,

Q ela Rated Power (rem) (x100 Ar/'F) (x10*Ar/'F) 9 94 1036 0.39 '

10 95 1017 0.48 11 93 1073 0.43 .' .

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TABLE 5 3 Comparison of Calculated and Measured Isothermc1 Temperature Coefficients (Continued)

EOC Calculated Critical Boron Measured DIT ROCS Percent of Concentration ITC ITC Cvele Rated Power (eem) (x100 3e/'F) (x100 Ar/*Fi 9 96 300 1.46 10 95 302 1.53 11 95 301 1,62 t

NOTE: Full Raced Power - 1500 MWe r I

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OPPD NA 8302 NP, Rev. 02 Page 29 of 31

TABLE 5 4 Comparison of Calculated and Measured Power Coefficients DIT ROCS Percent Critical Measured Calculated of Boron Power Power Burnup Rated Conc. Coeff. Coeff.

,Q1clg !f' *D /MTU Power (eem) (ae/t Power) (ac/t Pover)

~ ~

9 420 94 1036 1.64 x 10

9 9663 96 300 1.57 x 10

10 583 95 1017 1.24 x 10

10 9261 95 302 1.40 x 10*'

11 433 93 1073 0.95 x 10

11 9765 95 301 1.52 x 10*'

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NOTE: Full Rated Power - 1500 M'.'t '

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l OPPD NA 8302 NP, F.sv. 02 Page 30 of 31

TABLE 5 5 Cycle 11 CEA k'orths Calculated DIT ROCS 3.D ECB!2 (Saei (too)

~ ~

4 0.76 3 0.46 2 0.97 1 0.58 Total (4+3+2+1) 2.77

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l OPPD NA.8302.NP, Rev. 02 Page 31 of 31

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OMAHA PUBLIC POWER DISTRICT

!GCLEAR ANALYSIS i RELOAD CORE ANALYSIS METHODOLOGY t CEPAX/DIt VERITICATION PROGRAM .

l OPPD NA 8302 NP APPENDIX A Rev. 02 l i

April 1988 f l

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.c Table of Contents r

f SECTION FJWE A.0 VERIFICATION OF NEUTRONICS MODE 1.S FOR FORT 1 CAillot'N STATION A.1 Core Reactivity 1 ,

A.2 Power Distributions 2 ,

A.2.1 Radial Pover Discributions 2 t

A.2.2 Axial Power Distributions 3 A.3 Reactivity Coefficients 4 A.4 CEA Reactivity k' orth 4 A.$ Comparisons to Critical CEA Positions 4 Following a Reactor Trip I

l A.6 Comparison to Independent Radial Power 5 Distribution Calculations ,

REFFRENCES 6

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OPPD NA 8302 NP. App. A. Rev. O.

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l LIST OF TABLES  :

TABLE TITLE EASE A1 Unrodded HZP Critical Borots Concentrations Calculations 7 A2 Summary of Comparisons of Measured and Calculated Integral 8 Assembly Relative Power Densities A3 Low Power Physics Isothermal Temperature coefficients 11 A4 Comparison of Calculated and Measured Isothermal 12 Temperature Coefficients A5 Comparison of Calculated and Measured Power coefficient 14 A6 Cycle 1 CEA Vorths 15 A7 Cycle 2 CEA Vorths 15 A8 Cycle 3 CEA Vorths 16 A9 Cycle 4 CEA Vorths 16 A 10 Cycle 5 CEA Vorths 17 A 11 Cycle 6 CEA Vorths 17 A 12 Cycle 7 CEA Vorths 18 A 13 Cycle 8 CEA Vorths 18 OPPD NA 8302 NP, App. A, Rev. 02 11

A.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT Caul 0UN STATION The District has performed extensive verification of the neutronics models used in the reload core analyses. The results of the previous District verification efforts were reported in Reference A 1. This eff ort utili::ed l cross sections produced by CEPAX. The methodoiogy discussed in this re.

port utilizes cross sections produced by DIT. In order to demonstrate the District's ability to utilite the models with the DIT cross sections, addi-tional verification was undertaken.  !

This verification is in addition to the extensive verification of these methods done by Combustion Engineering (CE) and reported in Reference A 2.

It is not the District's intent to repeat CE's extensive verification ef-fort which includes a statistical assessment of the adequacy of the uncer-tainties used by both CE and the District. Rather, it is the District'- .

intent to demonstrate that the District can adequately model the Fort Cal-houn core and that the results of the District's verification effort are  ;

i consistent with those reported in Reference A 2.  !

The District's verification using DIT cross sections utili:es data record-ed for Cycles 6, 7 and 8. Benchmarking was performed for the prediction '

of overall core reactivity, power distributions, reactivity coefficients.

CEA vorth and Xenon reactivity. The results of the verification efforts include data for both CEPAK and DIT cross sections.

The verification uses experimental, data from the Fort Calhoun reactor and I independent calculations performed by CE and Exxon Nuclear Company (ESC). f Experimental data is obtained from startup tests and core follow programs.  ;

Calculational data is obtained from startup predictions, special analysis of startup tests or design lifetime computations.

A.1 Core Penetivity i  ;

The analysis of predicted reactivity for Fort Calhoun Station util-ites studies of starcup tests and plant data obtained during opera- l tion at power. The parameter used to measure reactivity is the l

critical boron concentration.

OPPD NA 8302 NP App. A. Rev. 02 i i Page 1 of 18 I

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A.0 ERITICATION OF NEUTRONICS MODELS FOR FORT Call 10VN STATIO!1 (Continued) l A.1 (' ore Reactivity (Continued) l Comparisons between measured and calculated critical boron concentra-tions for the unrodded HZP core are presented in Table A 1. The re.

l suits using the DIT cross sections are consistent with those reported i in Reference A 2.

Operating plant data has been analyzed and evaluations of core reac-tivity predictions carried out. There is little difference between calculated curves utilizing either CEPAK or DIT cross sections for Cyclas 6, 7, and S. Results for the operating plant data comparisans demonstrate the District's ability to calculate core reactivity.

A.2 Power Distributions Extensive comparisons of power distributions have been performed for Fort Calhoun and other CE reactors. These comparisons are contained I in References A 2 and A 3. The data given for Fort Calhoun in Refer-ence A 3 vere supplied by the District.

I A.2.1 Radial Power Distributions The District has performed comprehensive core follow calcula-tions since the start of Cycle 3 in 1976. Table A 2 summar.

izes the results of comparisons between the axially integrat- l ed assembly power as calculated by ROCS and that measured by CECCR using the self powered rhodiurn detector for Cycles 3.

4, 5. 6, 7 and 8. These comparisons are only performed for  ;

instru:nented assemblies because CECOR calculates the power i for non instrumented assemblies using coupling coefficients i ,

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OPPD NA 8302 NP. App. A. Rev. 02 i Page 2 of 1S

A.0 ERIFICATION OF NEUTRONICS MODELS FOR FORT Caul 0UN STATION (Continued) l A.2 fnver Distributions (Continued)

A.2.1 Radial Power Distributions (Continued) derived from the physics codes. The instrumented assembly powers are calculated by a method independent of the predic-  :

tive code. Sample comparisona for Cycle 8 are included in this document.

The extensive comparison between the calculated and measured radial power distributions verifies the capability of the Dis-trict to calculate these power distributions.

A.2.2 Axial Power Distributiens l The District has benchmarked the ROCS code & gainst CECOR mea-sured axial power distributions. This document contains com-parisons of co.re average and selected assembly axial power i distributions for Cycles 5 through 8.

The District has benchmarked the QUIX code against measured  !

data by comparing the QUIX calculated ASI and the CECOR mea-sured ASI during an axial oscillation test performed during Cycle 8 power ascension testing. The lead bank CEA's re-mained in the core during the entire test.

I the comparisons demonstrate the District's capability to cal-culate axial power distributiona using both ROCS and QUIX.

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OPPD NA 8302 NP App. A. Rev. 02 l Page 3 of 1S  !

A.0 '!ERITICATION OF NEUTRONICS MODELS FOR FORT CAL!!OUN STATION (Continued) l A.3 Renetivity Coefficients The capability of the District's ROCS model to predict the Isothermal Temperature Coefficient (ITC) and the Power Coefficient (PC) has been benchmarked against physics tests conducted at Fort Calhoun for all operating cycles. Table A 3 shows the comparison between calculated l ,

and measured ITC's for zero power startup testing at the beginnin5 of the cycle. Also included are calculations performed by ENC. using XTC. The comparison of measured and calculated ITC's for "at power" (

conditions is shown in Table A 4. The comparison of measured and calculated Power Coefficients is shown in Table A 5. In all cases, the ROCS code accurately predicts the behavior of the core and the results using the DIT cross sections are consistent with results reported in Reference A 2. l A.4 CEA Reactivity Vorth l

The District has extensively benchmarked the ROCS code against mea-sured and independently calculated values of CEA reactivity worth.

Tables A 6 through A 13 show the results of this benchmarking effort. { '

CE performed the PDQ calculations for Cycles 1, 2 and 4 ENC per-  ;

formed the XTC calculations for Cycles 6, 7 and 8. The District per- '

formed all 2 D and 3 D ROCS calculations and the Cycle 5 PDQ calcula-tions. These results demonstrate the District's capability to calcu-late CEA worths and the results using DIT cross sections are consis-  !

tent with the results reported in Reference A 2.

1 I

I A.5 Co?carisons to Critical CEA Positions Follovine s Reacter Trio 4 l

Another measure of the ability of the 3 D ROCS model to accurately l predict reactivity changes is its ability to predict the critical  !

boron concentration and CEA position following a reactor trip. A

study of this type was done for criticalities during the recovery t i

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OPPD NA-8302 NP App. A, Rev. 02 Page 4 of 18 I

l A,0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN 07.*'70N (Continued) l A,5 Corearisons en Critical CEA Poritions Followine a Reactor Trio l (Continued) from a reactor trip for Cycle 2. This study demonstrates the ability of the District's ROCS model to accurately model the power defect and 7 xenon buildup and decay. {

A.6 Cetearison to Indecendent Radial Power Distributien Calculations l Comparisons between the District's ROCS model calculations and ENC i t

XTC model calculations of the HTP radial power distributions have been performed. The comparisons show good agreement between the independent models.

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l REFERENCES t

A1 CEN 242-(0).P. OPPD Responses to NRC Questions em Fort Calhoun Cycle 8.  !

February 18, 1983. '

A2 CEN 226 P, "The ROCS and DIT Computer Codes for Nuclear Design," [

December, 1981.  !

A3 CENPD 153 P, "INCA /CECOR Power Peaking Uncertainty," May, 1980.

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Table A 1 L*nrodded HZP Critical Boron Concentrations Calculations 3 D* 2 D* 3-D Measured ROCS ROCS ROCS PDQ h erm (CEPAM) (C EPA)3 /M ( C E PA)') KIQ _

1 933 2 1240 3 1000 4 1027 ,

5 1242 6 1230 7 1241 8 1240

• A 20 ppm bias has been applied to these calculations.

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[ OPPD NA 8302 !n, App. A. Rev. 02 '

Page 7 of 1S I

Table A 2 Su==ary of Comparisons of Measured and Calculated Integral Assembly Relative Power Densities Cycle CEPAX DIT Seminal Burnup Power CZalg ( M'.'D /MTU) (t) 111 n of Full Power 3 80 45 3 177.5 60 3 510 95 3 800 100 3 1000 100 3 1428 100 3 2510 100 3 3100 100 3 3500 100 3 4000 100 3 4500 100 3 5200 100 3 5800 100 3 6400 100 3 7200 90 3 7715 80 4 200 100 4 1000 100 4 2000 100 4 3000 100 4 4000 100 4 5000 100 4 6000 100 OPPD NA 8302 NP, App. A. Rev. 02 Fage 8 of 18

Table A 2 Sunaary of Comparisons of Meatured and Calculated Integral Assembly Relative Power Densities (Continued) ,

Cycle CEPAK DIT Nominal Burnup Power M (MVD/MTU)

(t) m n of Full Power 4 7000 100 4 8200 100 5 300 100 5 1000 100  ;

5 2000 100  ;

5 3000 100 ,

5 4000 100 l

5 5000 100 5 6000 100 6 50 66 6 500 100 6 1000 100 6 2000 90 6 3000 65 6 4000 75 6 5000 73 6 5800 75 6 6500 100 6 7500 50 6 8500 95 6 9500 95 6 10500 95 7 135 , ,.

70 r i

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Table A 2 Surnary of Comparisons of Measured and Calculated Integral Assembly Relative Power Densities (Continued)

Cycle CEPAK DIT Nominal Burnup Power M (WD/MTU) (4) m n of Tull Power

~

7 500 100 7 1000 100

/ 2000 100 7 3000 100 ,

7 4000 100 7 5000 100 7 6000 100 7 7000 100 7 8000 100 7 9725 100 8 50 45 8 250 -

100 8 1000 100 8 2000 _ 100 1

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Table A3 1.ov Power Physics Isothorrnal *ct perature coe fficients

'oron CCPAR* DIT

oncontration  :!easured ROCS ROCS XTC

< ,. e l n < nm i t ar re rs rac /' es rao/*rs rar e'es 1 i

1 993 0.26 x 10*'

1 854 0.11 x 10*4 2 1240 0.41 x 10*'

2 1198 0.32 x 10*'

2 1164 0.09 x 10*'

3 1000 0.078 x 10*'

4 1020 0.14 x 10*'

5 1228 0.20 x 10*'

6 1213 0.23 x 10*'

. h 7 1213 0.12 x 10*.* l l

8 1240 0.16 x 10*' '

" Calculated results were biased by 0.20

• 10*' as/'T '

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1 OPPD !!A 3302-!;P. App. A. Rev. U2 Page 11 of 1S l

Table A 4 Comparison of Calculated and Measured isothermal Temperature coefficients BOC Calculated

• Calculated Critical Boron Measured CEPAX ROCS DIT. ROCS Percent of Concentration TC ITC ITC Cycle Rated Power (ppm) (x10{as/'T) (x10as/

• T) (x10'aa/*T)  :

l 1

~

2 69(1) 927 0.28 -

3 46(1) 720 0.41 4 92(1) 690 0.42 5 93(1) 876 0.19 6 95(1) 848 0.46 '

7 96(2) 817 0.52 8 79(2) 817 0,84 i

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j 1

1 I

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OPPD NA 8302 NP, App. A, Rev. 02 -

Page 12 of 18  !

l

Table A 4 Comparison of Calculated and Measured Isothermal Temperature Coefficients (Continued)

EOC t

Calculated

• Calculated f Critical Boron Measured CEPAX. ROCS DIT. ROCS Percent of Concentration iTC JTC ,ITC Cycle Rated Power (ppm) (x10*as/'T) (x10*as/'T) (x10'as/'F) 1 75(1) 239 0.98 2 46(1) 104 1.62 3 90(1) 62 L 65 4 95(1) 44 1.41 .

5 9'4 (1) 296 0.97 6 96(2) 307 1.51 7 95(2) 192 1.85 . -

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(1) Tull Rated Power - 1420 MVt (2) Tull Rated Power = 1500 MVt

• BOC calculated results were biased by 0.20 x 10*0 as/'T and EOC l calculated results were biased by 0.40 x 10*' as/'T.  !

l 1 i

l 0 PPD.NA 8302.NP. App. A. Rev. 02 Page 13 of 18

Table A 5 l t

Cornparison of Calculated and Measured Power Coefficients I

CEPAX ROCS DIT. ROCS Percent Measured Calculated Calculated  :

of Critical Power Power Power l Burnup Rated Boron Coeff. Coeff. Coeff.

CXClf,  %'D /MTU fptyfI Cone. (Ar/t Poverl (Ar/t Pover) (Ar/t Pover) l 2 10877 46(1) 104 1.95 x 10 '

3 157 46(1) 720 1.47 x 10 '

3 1513 90(1) 535 1.12 x 10

3 4183 90(1) 309 1.31 x 10*'

3 7208 90(1) 62 1.48 x 10

4 267 92(1) 690 1.04 x 10

4 4690 94(1) 288 1.12 x 10

4 8027 95(1) 44 1.10 x 10

5 426 93(1) 876 1.05 x 10*'

5 6815 94(1) 296 1.25 x 10

6 400 95(1) 848 1.11 x 10 l j 6 6467 96(2) 307 1.45 x 10 '

7 450 96(2) 817 0.98 x 10 6 7 6900 95(2) 283 1.30 x 10 '

7 7800 95(2) 192 1.57 x 10 '

8 459 79(2) 817 1.18 x 10 ,

b i

i (1) Pull Rated Power - 1420 .Wt  !

(2) Pull Rated Power - 1500 .Wt l

i

, 1 1 I OPPD.SA.8302.NP. App. A. Rev. 02 Page la of 15 i

l l

Table A 6 i I

Cyc13 1 CEA Vorths Calculated Calculated Calculated CEPAK ROCS CEPAK ROCS CEPAK PDQ Measured 3D 2D 2D h (tar) (%Ar) ISAr) (%Ar)

. ~ i 4 0.58 3 0.57 2 2,01 A 3.06 8 2.10 Total (4+3+2+A+5) 8.32

• i l

Table A 7 Cycle 2 CEA Vorths Calculated Calculated Calculated CEPAK ROCS CEPAK ROCS CEPAK PDQ Measured 3D 2D 2D h (tar) (6As) (tari (tAri  ;

4 0.65 I

3 0.41 2 1.67 [

1 0.95 '

Total (4+3+2+1) 3.68 OPPD NA 8302 NP. App. A, Rev. 02 Page 15 of 18

Table A 8 Cycle 3 CEA Vorths Calculated Calculated CEPAA. ROCS CEPAX ROCS Measured 3.D 20 CI.cn (gari (ear) (gar)

~ ~

4 0.74 3 0.59 2 1.96 1 0.80 Total (4+3+2+1) 4.09 Table A 9 Cycle 4 CEA Vorths Calculated calculated CEPAX ROCS CEPAX. ROCS Measured 3.D 20 Greur (gar) (gar) tear) 4 0.63 3 0.60 2 1.90 1 0.92 Total (4+3+2+1) 4.05 .

OPPD.NA 3302.NP. App. A. Rev. 02 Page 16 of 18

v Table A 10 I Cycle $ CEA Worths Calculated Calculated Calculated CEPAK. ROCS CEPAK. ROCS CEPAX.PDQ Measured 3D 2D 2.D  ;

0,Im ftar) (tar) (tar) (tar) ,

t 4 0.57  ;

3 0.67  !

2 1.40  !

1 0.99 i

Total r (4+3*2+1) 3.63 _

j f

Table A.11 l ,

Cycle 6 CEA Vorths (

i Calculated Calculated Calculated Calculated '

XTC CEPAK. ROCS CEPAX PDQ CEPAK.PDQ Measured 3.D 2D 3.D I h /

• A .a ) r n a.-) (tari (tar) (t!;s ,

i 4 0.52 I 3 0.66 1 1

2 1.57  !

1 0.93 l Total (4+3+2+1) 3.68  !

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CPPD.NA.8302 NP App. A. Rev. 02 i Page 17 of 15 [

Table A 12 Cycle 7 CEA Vorths Calculated Calculated Calculated Calculated XTG CEPAX ROCS CEPAX.?DQ CEPAX PDQ ,

Measured 3.D 2.D 3.D i C.IE12 - ' .h f'Ari (4Ari _ f4Ari - (% Ari -

l I "

7 4 s.41 3 0.47 I 2 1,63 1 0.66 Total (4+3+2+1) 3.27 l_ ,

Table A.13 C/cle 8 CEA '.'orths l Calculated Calculated Calculated Calculated XTC CEPAX. ROC 4 CEPAX PDQ CEPAX.PDQ f Measured 3.D 2D 3.D Q.tZ12 (4Ari f9Ari (9Ari . flari (tAri

~

4 0.58 '

3 0.63 ,

s 2 0.99 t 1 1.00 Totsi i (a+3+2+1) 3.20 ~

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L OPPD.NA 8302.NP. App A. Rev. 02 Page 18 of 18 l

OMAHA PUBLIC POER DISTRICT NUCLEAR ANALYSIS ONGOING BENCIO!ARKING PROGRAM HISTORICAL INFOP.MATION OPPD NA 3302 NP APPENDIX B Rev. 02 April 1988 s

t LIST OF TABLES TABLE TITLE EAGE B1 Unrodded HZP Critical Boron Concentrations Calculations 1 B2 Low Power Physics Isothermal Temperature coefficients 2 B3 Comparison of Calculated and Measured Isothermal 3 Temperature Coefficient ,

B4 Comparison of Calculated and Measured Power Coefficient 5 B5 Cycle 1 CEA Worths 6 B6 Cycle 2 CEA Worths 6 B7 Cycle 3 CEA Worths 7 B8 Cycle 4 CEA Worths 7 B9 Cycle 5 CEA Worths 8 B-10 Cycle 6 CEA Worths 8 B ll Cycle 7 CEA Worths 9 B 12 Cycle 8 CEA Worths 9 B 13 Cycle 9 CEA Worths 10 B-14 Cycle 10 CEA Worths 10

. t I

I OPPD NA 8302 NP, App. B. Rev. 02 i

Table B 1 Unrodded ilZP Critical Boron Concentrations Calculations 3 D* 2-D* 3D Measured ROCS ROCS ROCS PDQ Cvele rm ICf26D, (CEPAFD (DIT) (CEPAV) KlQ 1 933 2 1240 3 1000 4 1027 5 1242 6 1250 7 1241 8 1240 9 1518 10 1474

~

• A 20 pp.n bias has been applied to these calculations.

OPPD NA 8302 NP. App, B, Rev. 02 Page 1 o f 10

Tabic B 2 Low Power Physics Isothermal Temperature Coefficients Boron CEPAK* DIT Concentration Measured ROCS ROCS XTC Cycle (remi fac/'F) (ac/'F) (ac/*F) (ae / 'FT 1 993 0.26 x 10*'

1 854 0.11 x 10*'

2 1240 0.41 x 10*'

2 1198 0.32 x 10*'

2 1164 0.09 x 10*'

3 1000 0.078 x 10*'

4 1020 0.14 x 10*'

5 1228 0.20 x 10*'

6 1213 0.23 x 10*'

7 1213 0.12 x 10*'

8 1240' O.16 x 10*'

9 1457 0.30 x 10*'

10 1457 0.23 x 10*'

• Calculated results were biased by 0.20 x 10*' op/'F OPPD NA 8302 NP, App. B, Rev, 02 Page 2 of 10

Table B 3 Comparison of Calculated and Measured Isothermal Temperature Coefficient BOC Calculated

• Calculated Critical Boron Measured CEPAK ROCS DIT ROCS Percent of Concentration .

ITC Cycle Rated Power (ppm) (x10{TC Ap/'F) (x10{TC ap/'F) (x104 40/'F) 1 -

2 69(1) 927 0.28 3 46(1) 720 0,41 4 92(1) 690 0.42 5 93(1) 076 0.19 6 95(1) 848 0,46 7 96(2) 817 0,52 8 79(2) 8 ). 7 0.84 9 94(2) 1036 0.39

• 10 95(2) 1017 0.48 _ _

OPPD ?JA 8302 tiP. App. B. Rev, 02 Page 3 of 10

N Table B 3 ,

Comparison of Calculated and Measured-Isothermal Temperature Coefficient ,

(Continued)

EOC  !

Calculated

• Calculated Critical Boron Measured CEPAX ROCS DIT ROCS Percent of Conc.entration JTC ITC Cycle Rated Power (ppm) (x10"ap/'F) (x10{TCap/'F) (x100 ap/*F) 1 75(1) 239 0.98 ,

2 46II} 104 1.62 3 90(1) 62 1,65 4 95(1) 44 1.41 5 94(1) 296 0.97 6 95(2} 307 1.51

'7 W' 2 ) 192 1.85 8 95(2} 292 1.86 l 9 96(2} 300 1.46 ,

I 10 95(2) 302 1.53 -

1

)

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(1) Full Rated Power - 1420 !Gt l

l l '(2) Full Rated Power - 1500 !Ne i I

• BCCcalculatedresultswerebiasedby0.20x10*fap/*Fand [

, EOC calculated results were biased by 0.40 x 10** ap/*F. l t '

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OPPD t1A 6302 tiP, App. B. Rev. 02 Page 4 of 10

l i .

Table 3-4 Coe.parison of Calculated and Measured Power Coefficient CEPAK ROCS DIT ROCS Percent Critical Measured Calculated Calculated of Boron Power Power Power Burnup Rated Conc. Coeff. Coeff. Coeff.

Cvele M'.JD /MTU Power (com) (or/$ Power) (or/* Poweri Ae/S Powari 2 10877 46(1) 104 1.95 x 10

3 157 46(1) 720 -1.47 x 10

3 1513 90(1) 535 1.12 x 10

3 4183 90(1) 309 -1.31 x 10*'

3 7208 90(1) 62 1.48 x 10

4 267 92(1) 690 1.04 x 10

4 4690 94(1) 288 1.12 x 10

4 8027 95(1) 44 1.10 x 10

5 426 93(1) 876 -1.05 x 10

5 6815 94(1) 296 1.25 x 10

6 400 95(1) 848 1.11 x 10

6 6467 96(2) 307 1.45 x 10

7 450 96(2) 817 0.98 x 10

7 6900 95(2) 283 1.30 x 10

7 7800 95(2) 192 1.57 x 10

8 459 79(2) 817 1.18 x 10

8 6150 95(2) 292 1.70 x 10

9 420 94(2) 1036 1.64 x 10

9 9663 96(2) 300 1.57 x 10

10 583 95(2) 1017 1.24 x 10

10 9261 95(2) 302 1.40 x 10 -

(1) Full Rated Power - 1420 L't (2) Full Rated Power - 1500 6't OPPD t'A 8302 NP, App. B, Rev. 02 Page 5 of 10

l Table B 5 Cycle 1 CEA Worths calculated Calculated Calculated CEPAK-ROCS CEPAK ROCS CEPAK PDQ Measured 3-D 2D 2-D Greco (too) (tae) (soo) (tao)

4. 0.58 3 0.57 2 2.01 A 3.06 B 2.10 Total (4+3+2+A+B) 8.32 -,

Table B 6 Cycle 2 CEA Worths Calculated Calculated Calculated CEPAK ROCS CEPAK ROCS CEPAK-PDQ MJasured 3D 2D 2D m

C (160) (tao) (tao) (46o)

~

4 0.65 3 0.41 2 1.67 1 0.95 Total (4+3+2+1) 3.68 _ ,

\

OPPD NA 8302 NP, App. B, Rev. 02 Page 6 of 10

Table B 7 Cycle 3 CEA Worths Calculated Calculated CEPAK ROCS CEPAK ROCS Measured 3D 2D h _Lu i (Sao) (noei

- 9 4 0.74 3 0.59 2 1,96 1 0.80 Total (4+3+2+1) 4.09 '

Table B 8 Cycle 4 CEA Worths Calculated Calculated CEPAK ROCS CEPAK ROCS Heasured 3D 2D Grour (460) (160) (S6e) 4 0.63 3 0.60 2 1.90 1 0.92 Total (4+3+2+1) 4.05 OPPD NA 8302 NP, App. B, Rev. 02 Page 7 of 10

l Table B 9 Cycle 5 CEA Worths j I

Calculated Calculated Calculated CEPAX-ROCS CEPAK ROCS CEPAK-PDQ Measured 3-D 2-D 2D Crono (*60) (46c) (Soc) (Soc) 4 0.57 3 0.67 2 1.40 1 0.99 Total (4+3+2+1) 3.63 Table B 10 Cycle 6 CEA Worths Calculated Calculated Calculated Calculated XTC CEPAK ROCS CEPAK-PDQ CEPAK PDQ Measured 3D 2D 3D Group (460) (Soc) /%Ac) (4Af) (%6e)

~

4 0.52 3 0.66 2 1.57 1 0.93 Total (4+3+2+1) 3.68 ~

OPPD NA 8302 NP. App. B, Rev. 02 Page 8 of 10

- .a______________.

'd l

Table B 11' Cycle 7 CEA Worths

)

. Calculated Calculated Calculated Calculated ,

XTG CEPAX ROCS l CEPAX PDQ CEPAX PDQ Measured 3D 2-D 3D '

GI.2W2 (%ool (%Aoi (%Aoi (%Aoi (460)  ;

,, ~1 4 0.49 3 0.47 i

.i 2 1.65 l 1 0.66 Total (4+3+2+1) 3.27 _, -

i Table B 12 [

l

. Cycle 8 CEA Worths  !

i Calculated Calculated Calculated Calculated  ;

XTG CEPAX ROCS CEPAX.PDQ CEPAK.PDQ Measured 3.D 2-D 3.D r Group (tao) (tao) (tao) (gao) (tao) .,

4 0.58 i 3 0.63 2 0.99 1 1.00 Total (4+3+2+1) 3.20 _  !

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i OPPD.tiA.8302.tiP App. B. Rev. 02 [

Page 9 of 10 j i

TABLE B 13 Cycle 9 CEA Worths Calculated DIT ROCS Measured 3D Group (too) (%Ao) 4 0.67 3 0.72 2 1.63 1 0.84 Total (4+3+2+1) 3.86 TiBLE B 14 Cycle 10 CEA Worths Calculated DIT ROCS Measured 3D G.I.222 (*ao) (*ao) 4 0.61 3 0.62 2 1.94 1 0.85 Total (4+3+2+1) 4.02 -

OPPD NA 8302 NP, App. B, Rev. 02 Page 10 of 10