ML20212N027

Rev 1 to Reload Core Analysis Methodology Overview

ML20212N027
Person / Time
Site:	Fort Calhoun
Issue date:	11/30/1986
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19292G906	List: ML20212N015 ML20212N021 ML20212N027 ML20212N032
References
OPPD-NA-8302-NP-R01, OPPD-NA-8302-NP-R1, NUDOCS 8703120319
Download: ML20212N027 (266)

Text

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O q Omaha Public Power District Nuclear Analysis Reload Core Analysis Methodology i

Neutronics Design Methods And Verification ,

OPPD-NA-8302-NP Rev. 01 O

November 1986 Copy No.

O hoj3120319870309 p ADOCK 05000285 PDR

J A8STRACT This document is a Topical Report describing Omaha Public Power District's reload core neutronics desigr. methods for application to Fort Calhoun Station l Unit No. 1.

The report addresses the District's neutronics design methodology and its application to the calculation of specific physics parameters for reload cores.

< In addition, comparisons of results obtained using this methodology to results from experimental measurements and independent calculations are provided.

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i Proprietary Data Clause This document is the property of Omaha Public Power District (0 PPD) and contains i proprietary information, indicated by brackets, developed by Combustion Engi-neering (CE) and Exxon Nuclear Company, Inc. (ENC). The CE and ENC information l was purchase'd by OPPD under proprietary information agreements.

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Table Of Contents

'T . Section JPage V

1.0 INTRODUCTION

1 2.0 BASIC PHYSICS MODELS 1 2.1 Neutron Cross-Sections 1 2.2 Diffusion Theory Models 3 2.2.1 PDQ-X 3 2.2.2 ROCS 4 2.2.3 QUIX 4 3.0 FORT CALHOUN PHYSICS MODELS 6 3.1 Neutron Cross-Sections 6 3.2 -Diffusion Theory Models 7 3.2.1 PDQ-X 7 3.2.2 ROCS, 8 3.2.3 QUIX 9 4.0 APPLICATION OF PHYSICS METHODS 9 4.1 Radial Peaking Factors 9 4.2 Reactivity Coefficients 10

) 4.3 Neutron Kinetics Parameters 11 4.4 Dropped CEA Data 12 4.5 CEA Ejection Data 13 4.6 CEA Reactivity 14 4.7 CEA Withdrawal Data 16 4.8 Reactivity Insertion for. Steam Line Break Cooldown 16 4.9 Asymmetric Steam Generator Event Data 17 4.10 QUIX Calculations 18 5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION 18 5.1 Core Reactivity 19 5.2 Power Distributions 20 5.2.1 Radial Power Distributions 20 5.2.2 Axial Power Distributions 21 5.3 Reactivity Coefficients 21 5.4 CEA Reactivity Worth 22 5.5 Comparisons to Critical CEA Positions Following a Reactor Trip 22 O

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Table of Contents (Continued)

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5.0 VERIFICATION OF NEUTR.0NICS MODELS FOR~ FORT CALHOUN STATION

'(Continued) 5.6 Comparisons to ' Independent Radial Power Distribution Calculations -

23 5.7 The District's Ongoing Benchmarking Program' 23 5.8 Sumary 23

6.0 REFERENCES

54 APPENDIX A Cycle 8 Radial Power Distribution Comparisons 56 APPENDIX 8 Axial Pc.<er Distribution Comparisons 92 APPENDIX C-Cycle 9 and 10 Verification Program Additions 112 O

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I ' OMAHA PUBLIC_ POWER DISTRICT NEUTRONICS DESIGN METHODS AND VERIFICATION i ,

i REVISION DATE ,

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' Omaha Public Power District Reload Core Analysis Methodology b

v Neutronics Design Methods and Verification

1.0 INTRODUCTION

The District's neutronic design calculation methods are described along with results obtained when these methods are compared to experimental measurements and independent calculations. The discussion of calculational methods includes descriptions of the basic computer codes and procedures for applying these codes. Comparison of the calculations to measurements and independent calculations are performed using the same codes and com-putational methods used in the Fort Calhoun reload core design efforts.

The basic physics models, supplied by Combustion Engineering (CE), are described in Section 2.0. Section 3.0 describes the District's application of these models to the Fort Calhoun reactor. Section 4.0 presents the application of these physics models to the reload core analysis. Section 5.0 discusses the comparisons of District calculated data to measured operating data from Fort Calhoun Station and data from independent calcula-tions. Section 6.0 contains the individual references.

, 2.0 BASIC PHYSICS MODELS The District's neutronics design analysis for the Fort Calhoun core is based on a combination of multi-group neutron spectrum calculations, which provide cross-sections appropriately averaged over a few broad energy groups and few-group I , 2- and 3- dimensional diffusion theory calcula-tions, 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 i

consistent with the current State-of-the-Art.

2.1 Neutron Cross-Sections The data base for both fast and thermal neutron cross-sections is derived from ENDF/B-IV with changes recommended by the cross-section OPPD-NA-8302-NP Rev. 01 Page 1 of 259

2.0 BASIC PHYSICS MODELS (Continued)

_p i l-V 2.1 Neutron Cross-Sections (Continued) evaluation working group (Reference 2-1). These recommendations consist of changes to the shielded resonance of U 238 , and the Watt fission spectrums of U 35 and Pu 239 , and changes in- for U 235 and 239 Pu . Few group cross-sections, for subregions of the core that are represented in spatial diffusion calculati'ans, (e.g., fuel pin cells, mcderator channels, structural member cells, etc.) are calculated by the DIT latice program. These cross-sections are generated as a function of fuel temperature and moderator temperature to accommodate l 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), burnable o absorber rods, and incore flux detectors. The essential feature of DIT, which distinguishes it from the traditional methodology is that the spectrum spatial averaging procedures are based on calculations in two-dimensional geometry. Hence few approximations to the geometry representation are necessary. The use of nodal transport theory has made it feasible to retain discrete pin geometry in both the fine and broad energy group calculations. 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 utilized the CEPAK program to produce few-group cross-sections. These cross-sections were also generated as functions of fuel and moderator temperature. Comparisons of cal-culated and measured data reported in Section 5.0 include calculations performed using the CEPAK program.

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2.0 BASIC PHYSICS MODELS (Continued)

. /~N O 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 laboratories, e.g., FORM, THERMOS and CINDER. These programs are interlinked in a consistent way with inputs from differential cross-section data from an extensive library.

A description of the CEPAK procedures used to generate few-group neutron cross-sections can be found in Reference 2-4.

2.2 Diffusion Theory Models The diffusion theory models package used to calculate core physics parameters for Fort Calhoun Station consist of the PDQ-X, ROCS, and QUIX computer codes. The PDQ-X and ROCS codes can be executed in one, two and three dimensions to calculate static and depletion dependent parameters such as reactivities, flux, nuclide and power distributions q and CEA worths. The QUIX code is executed in one dimension to calcu-V late axial power distributions and CEA worth [

3 2.2.1 PDQ-X The PDQ-X program is an extension of PDQ-7 and HARMONY programs (Reference 2-5) to include the following optional capabilities:

(1) Fuel temperature feedback in the two-dimensional geometry option, (2) Fuel and moderator feedback in the three-dimensional geometry option, (3) Poison content criticality searches, and O,

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2.0 BASIC PHYSICS MODELS (Continued) g O 2.2 Diffusion Theory Model (Continued) 2.2.1 PDQ-X(Continued) ,

(4) Spatial feedback on the power distribution with fuel and moderator temperature in the 1-dimensional geometry option.

PDQ-X employs macroscopic (static) or microscopic (deple-tion) cross-section data generated by methods described in Section 2.1, 2.2.2 ROCS The ROCS program is a course mesh 2-group solution of the neutron diffusion equation based upon a mesh centered higher order finite difference formulation. It incorporates closed channel thermal hydraulic modeling into its evaluation of O

v the interaction of neutron flux effects and the macroscopic physical and thermal properties of distributed materials.

Because of its nodal structure and course mesh, ROCS is more efficient than PDQ-X for evaluating a core's static and depletion dependent properties. ROCS also employs macro-scopic (static) or microscopic (depletion) cross-sections generated by the methods described in Section 2.1. A more complete description of the ROCS progran is found in Refe-rences 2-3 and 2-6, 2.2.3 QUIX The QUIX program is a one-dimensional (axial) representation of the core used to determine static and time dependent reactivities and power distributions at selected stages of depletion. This program solves the neutron flux and asso-4 ciated eigenvalue in problems containing up to 140 distinct 4

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2.2 Diffusion Theory Model -(Continued) 2.2.3 QUIX (Continued) t _ regions or compositions with variable mesh: intervals'. The I

. macroscopic cross-section distributions, fission product yields, and xenon and boron nicroscopic cross-sections l

required.as input to QUIX are obtained from'either a one-dimensional PDQ-X calculation or a three-dimensional ROCS calculation. Local power density (fuel temperature) feed-back is included by modifying the point wise macroscopic 5 absorption and removal cross-sections. The change in cross-sections is represented by a function of the dif-

., ference between the local axial power density and the referenced power density.. Moderator density feedback is (

included by relating changes in the macroscopic absorption-

{ and removal cross-sections to the local hydrogen number

{ density which is calculated from enthalpy at each axial i

segment. These cross-section functions are generated in such a way that the fuel and moderator temperature coeffi-

, cients calculated by QUIX are equal to or conservative with j respect to the fuel and moderator temperature coefficients i

calculated by ROCS. The axial reflector cross-sections l input to QUIX are determined in such a way that the steady

state axial power distribution generated by QUIX matches the l axial power distribution generated by ROCS, Details of the above treatments are given in Reference 2-7.

j In addition to the eigenvalue problem, QUIX will perform l four types of searches to obtain a specific eigenvalue,

! viz., a uniform poison search, buckling search, CEA region boundary search, and a moderator density dependent poison l

j search. The unifonn poison search assumes an axially I

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2.0 BASIC PHYSICS MODELS (Continued)

'd 2.2 Diffusion Theory Model (Continued) 2.2.3 QUIX (Continued) constant macroscopic absorption cross-section whereas the moderator density dependent poison search assumes a dis-tributed macroscopic absorption cross-section dependent upon the axial moderator density. The moderator density de-pendent search is used to simulate the reactivity effects of the soluble boron in the reactor coolant.

Through the use of rod shadowing factors, shape annealing factors and shape index biases, the QUIX program has the capability of simulating excort detector response expected during normal operation. The procedures used for these simulations are described in Reference 2-8.

pd 3.0 FORT CALHOUN PHYSICS MODELS The District utilizes 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 rtodels.

3.1 Neutron Cross-Sections The two-group neutron cross-sections utilized in the ROCS and PDQ-X models of the Fort Calhoun reactor core are generated using the DIT code. Cross-sections have been generated for unshimmed ENC and CE fuel assemblies and shimmed ENC and CE fuel assemblies. The cross- l sections have been generated for the District by CE and are based on k information supplied by the District.

OPPD-NA-8302-NP Rev. 01 Page 6 of 259

3.0 FORT CALHOUN PHYSICS N00ELS (Continued)

O 3.1 Neutron Cross-Sections (Continued)

The cross-sections utilized to model the Fort Calhoun reactor are in the fonn 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, 84C shim number density. The table sets are applicable over a fuel temperature range from room temperature to 1800 K and a moderator temperature range from room temperature to 800*K. The fine mesh table

, sets include explicit treatment of the pin cells immediately 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.

3.2 Diffusion Theory Models The District utilizes the PDQ-X, ROCS and QUIX models described in N Section 2.0. The PDQ-X model is a fine mesh two-dimensional model.

The District utilizes both a two-dimensional and three-dimensional ROCS model. The QUIX model is a one-dimensional model.

3.2.1 PDQ-X The District's PDQ-X model is a two-group, two-dimensional fine mesh model in which each fuel pin cell and shim pin cell is represented by a single mesh point. The model includes explicit representation of the CEA guide tube (water holes), the CEA's, the interassembly water gaps, the water gap between the outer fuel as'sembly and the core shroud, the core shroud, the water gap between the core shroud and the core barrel, the core barrel and a portion of the water gap between the core barrel and the thermal shield. The model is representative of the core between 20%

and 80% of full core height.

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3.0 FORT CALHOUN PHYSICS MODELS (Continued) 3.2 Diffusion Theory Models (Continued) 3.2.1 PDQ-X(Continued)

The PDQ-X model is used to simulate the expected mode of operation in the cycle being analyzed. This calculation results in material distributions and radial peaking factors which are used in the safety analysis and setpoint genera-tion. The mode of operation at the Fort Calhoun reactor is base loaded operation. Base loaded operation consists of reactor 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 inserted less than 5% of the time whenever the reactor is at a steady state power level. Reference 3-1 discusses the impact of operation with a time averaged lead p bank insertion of [ J. The model is typically depleted in time steps of 1,000 MWD /MTU.

3.2.2 ROCS The District utilizes a three-dimensional and a two-di-mensional two-group ROCS model. [

] The two-dimensional model is representative of the core between 20% and 80% of full core height. [

] The boundary conditions are derived in accordance with the methodology discussed in Reference 3-2.

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3.0 FORT CALHOUN PHYSICS MODELS (Continued)

-p 3.2 Diffusion Theory Models (Continued) 3.2.3 -QUIX The District utilizes a one hundred and twenty-five axial node QUIX model. The data for the QUIX model is obtained from the three-dimensional ROCS calculations.

4.0 APPLICATION OF PHYSICS METHODS Previous sections have focused on the reactor physics models utilized by the District to model the Fort Calhoun reactor. In this section, calcu-lations of the various core parameters used in the safety analysis are described. The main parameters considered are the radial peaking factors (FRand Fxy), the moderator temperature coefficient, the fuel temperature or Doppler coefficient, the neutron kinetics parameters, CEA drop data, CEA q ejection data, CEA scram reactivity, reactivity insertion for the steamline V break cooldown, radial peaking data for the asymmetric steam generator event, and axial power distributions.

4.1 Radial Peaking Factors The radial peaking factors, F Rand Fxy, are calculated with the PDQ and 3-D ROCS models. Values of Fxy for both unrodded and rodded core configurations are obtained directly from the PDQ power distribution.

Since the cross-sections utilized by the PDQ model implicitly account for the peaking of the thermal flux in the CEA guide tubes (water holes) no correction is required to the peaking factors calculated by PDQ. The values of F for the unrodded core are obtained by multi-R plying the integrated assembly powers from ROCS by the pin to box ratio obtained from PDQ. The value of RF f r various rod configura-tions is derived by multiplying the assembly normalized planar power for each ROCS plane by the pin to box ratio for the rod configuration

,A in that plane and summing these values for all planes and dividing by the number of planes.

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'4.0 APPLICATIONOFPHYSICSMETHODS(Continued) o

- 4.1 Radial Peaking Factors (Continued)

The uncertainties for the radial peaking factors are given in Refe-rence 4-1.

The physics models are used to calculate the expected values of FR and used in the safety analysis are F

xy. The actual values of FR and Fxy chosen to be conservatively high with respect to those anticipated during the core life.

4.2 Reactivity Coefficients 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 noderator temperature. Calculationally, the MTC at a temperature of Tmod is determined by running three calculations; one at Tmod, one at Tmod +

U(~'N 10 F and one at Tmod -10'F.

The MTC at a temperature of Taod 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 fuel temperature upon resonance neutron energy absorption is accounted for in the ROCS and PDQ models by means of power feedback options.

The representation of the variation in the few group cross-sections with fuel temperature involves two nain 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 the few-group cross-sections. The second portion requires establish-ment of correlation between fuel temperature (i.e., effective fuel 0 PPD-NA-8302-NP Rev. 01 Page 10 of 259

l 4.0 APPLICATION OF PHYSICS METHODS (Continued) p V 4.2 Reactivity Coefficients (Continued) temperature to be used in generation of cross-sections) and the reactor power density. The relationship between fuel temperature and reactor power density employs direct fits to FATES (Reference 4-2) fuel data. This method results in the fuel temperature correlation for each fuel type which is both local, power density and fuel exposure dependent.

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.

The physics models are used to calculate the expected vaiues of the MTC throughout the cycle. The actual values of the MTC used in the safety analysis are chosen to conservatively bound expected values of p the MTC. The measurements of the MTC made during the operation of the

\- reactor include uncertainties to assure that the actual MTC does not exceed the values used in the safety analysis. A fifteen percent uncertainty is applied to the Doppler coefficient when it is used in the safety analysis calculations.

4.3 Neutron Kinetics Parameters The neutron kinetics parameters g, x and the neutron lifetime, g*, are calculated using Combustion Engineering's ROCS computer code. l The technique utilized to calculate the kinetics parameters and the neutron lifetime is based on first order perturbation theory. Details of the perturbation approach are discussed in References 4.3 and 4.4.

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4.0 APPLICATION OF PHYSICS METHODS (Continued)

V 4.4 Dropped CEA Data The neutronics data unique to the dropped CEA analysis are the values of F and F R xy following the drop of a CEA and the reactivity worth of the dropped CEA. The values of FR and Fxy increase due to a large azimuthal tilt caused by the drop of a CEA and occur on the side of the core opposite the dropped CEA. Because the maximum RF and Fxy occur far away from the dropped CEA, the intra-assenbly power distri-bution is not perturbed. Therefore, the " post drop" value of FR and F

xy can be calculated 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 distortion factor. The distortion factor is defined as the ratio of the assembly RPD from a radial power distribution at a given power level and time in core life containing a dropped CEA to assembly RPD from a radial power distribution at the _

same power level and time in core life without a dropped CEA. [

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The distortion factor and dropped CEA reactivity worth can be calcu-lated using the 2-D or 3-D ROCS model. The 2-D ROCS calculations yield the F xy distortion factor as a function of CEA bank insertion (i.e., AR0, Bank 4 In, Banks 4+3 In) and power level. [

.] The 3-D F R distortion

_ factor is calculated for a specific CEA insertion and power level. _

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, 4.0 ' APPLICATION OF PHYSICS METHODS (Continued)

V 4.4 Dropped CEA Data (Continued)

] Sufficient margin exists at the lower power levels

[ .] for the F dependent DNBR calculations does not adversely effect operating R

margin. The " post drop" value of RF using the 3-D F R distortion factor is calculated by multiplying the " pre-drop" value of RF f r the particular CEA insertion and power level by the 3-D RF distortion factor.

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

/- Becausethe[

] of the dropped CEA during the fuel cycle and because of the

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-] documented in Reference 4-5, [ ] uncertainty is applied to the distortion factor. A[ ] uncertainty is applied to the reactivity worth of the dropped CEA based on the verification contained in Reference 4-5.

4.5 CEA Ejection Data The neutronics data unique to the CEA ejection analysis are the value of F xy following the ejection of a CEA and the reactivity worth of the ejected CEA. The maximum post ejection value of F and maximum ejected CEA reactivity worths are calculated for the maximum CEA insertion allowed by the PDIL at HFP and HZP. The physics parameters are calculated using a HZP 2-0 ROCS model or a HZP 2-0 PDQ model. The post ejection value of F xy is obtained directly from the PDQ calcu-lation. The post ejection value of F ,is x obtained from the 2-D ROCS OPPD-NA-8302-NP Rev. 01 Page 13 of 259

I 4.0 APPLICATION OF PHYSICS METHODS (Continued)

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b 4.5 CEA Ejection Data (Continued) calculation by multiplying the 2-D ROCS post ejection assembly RPD by a conservatively high pin to box ratio. The ROCS methodology is conservative with respect to the more exact PDQ n.ethod. The ejected l CEA reactivity worth is obtained directly from either calculation. l Both the PDQ and ROCS post ejection power distributions are calculated without moderator or fuel temperature feedback.

The post ejection value of Fq is calculated by multiplying the post ejectia value of F by the maximum value ofz F , the azimuthal tilt allowance, the augmentation factor, the engineering heat flux factor, the fuel densification factor, and the F uncertainty q

documented in Reference 4-1. A[ ] uncertainty is applied to the ejected CEA worth.

4.6 CEA Reactivity C

The CEA reactivity calculations done in a reload core safety analysis are the calculation of the total reactivity of CEA's inserted 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. A[

]uncertaintyisappliedtotheHZPCEAscramreactivityworth.

The HZP CEA scram reactivity for the CEA ejection transient is calcu-lated in a similar fashion except that the worth of the ejected and highest stuck worth CEA's are subtracted from the net worth, p The scram CEA worth at HFP is calculated by obtaining the HFP net worth for all CEA's between the HFP PDIL CEA position and the fully OPPD-NA-8302-NP Rev. 01 Page 14 of 259

4.0 . APPLICATION OF PHYSICS METHODS (Conti7ued) .

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/ 4.6 CEA Reactivity (Continued) inserted position, subtracting the vorth of the highest worth stuck CEA and subtracting the moderator vcid collapse allowance. The thermal hydraulic axial gradient redection 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 the steam line break incident. These are not applied to the four pump loss of flow scram CEA worth because the closest approach to the SAFDL during the four pump loss of flow event occurs prior to significant CEA insertion. These allowances are not applied to the steam line break (SLB) incident HFP CEA scran worth because the HFP SLB reactivity insertion curves implicitly account for these effects.

The moderator void collapse allowance is 0.0% an at 80C and 0.1% an at E0C. The thermal hydraulic axial gradient reduction allowance is 0.2%

an at B0C and 0.4% an at E0C. A[ ] uncertainty is applied to the HFP 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.

The generation of the scram reactivity curves utilizes the methodology discussed in Reference 4-6.

The calculation of the required shutdown margin is only performed at HZP since the shutdown margin at power is controlled by the PDIL. The available HZP shutdown margin is equivalent to the HZP CEA scram reactivity.

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4.0 ' APPLICATION OF PHYSICS METHODS (Continued) d 4.7 CEA Withdrawal Data The reactor core physics data unique to the CEA withdrawal analysis is the maximum differential CEA wor ~th. This is the maximum amount of reactivity at any time in core life that can be added to the core per inch of CEA motion. When the maximum differential CEA worth is com-bined with the maximum CEA withdrawal rate of 46 inches / minute, a conservative withdrawal rate expressed in %Ap/sec is obtained and used as input to the CEA withdrawal analysis.

The maximum differential CEA worth is obtained for the sequential withdrawal of the CEA banks from the HZP PDIL to an all rods out condition. The 3-D ROCS model is utilized to calculate this para-meter. The calculations are performed assuming that the reactor is being controlled within the " constant ASI" limits in accordance with the Fort Calhoun Operating Manual.

C

\ 4.8 Reactivity Insertion for Steam Line Break Cooldown The reactor core physics data unique to the steam line break transient analysis is the reactivity insertion due to the cooldown of the moderator. There are two sources of this reactivity insertion. The first is the positive reactivity insertion due to the increasing density of the moderator as the cooldown progresses. The second is the reactivity insertion due to the Doppler coefficient as the effec-tive fuel temperature thanges.

Reactivity insertions due to the moderator density increase and the Doppler coefficient are both calculated using a full core ROCS model.

The axial leakage or buckling is adjusted such that the moderator temperature coefficient calculated by the ROCS model corresponds to the most negative Technical Specification limit. The reactivity insertion calculations are performed with all CEA's except the most reactive CEA inserted in the core.

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I 4.0 APPLICATION OF PHYSICS METHODS (Continued)

- v 4.8 Reactivity Insertion for Steam Line Break Cooldown (Continued)

The moderator density reactivity insertion curve for the hot zero power steam line break case is calculated by successively lowering the inlet temperature of the ROCS model from 532 F and allowing only moderator temperature feedback in the model. The calculations _

typically result in a curve of reactivity insertion vs. moderator temperature from a hot zero power temperature of 532 F to 212 F.

The Doppler reactivity insertion curve for the hot zeto 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 assuming there is no decay heat and no credit is taken for local voiding in the region of the stuck CEA.

pJ The moderator density reactivity insertion curve for the full power case is calculated by decreasing the power level and core average average coolant temperature from full power to the het zero power inlet temperature and then successively lowering the inlet temperature as in the het zero power case. Only moderator temperature feedback is utilized in the ROCS model. The Doppler reactivity insertion curve is calculated by a similar procedure utilizing the fuel temperature feedback in the model.

Since the moderator reactivity insertion curve corresponds to an MTC which is at the Technical Specification limit, no additional uncer-tainty is added to this curve. A fifteen percent uncertainty is applied to the Doppler reactivity insertion curve.

4.9 Asymmetric Steam Generator Event Data The reactor core physics data unique to the asymmetric steam generator V event [are plots of AF RT vs. AT IN and AF T vs. AT IN. These plots are OPPD-NA-8302-NP Rev. 01 Page 17 of 259

4.0' APPLICATION OF PHYSICS METHODS (Continued) 4.9 Asymmetric Steam Generator Event Data (Continued) calculated using the ROCS model.] For the range of temperatures considered, the intra-assembly peaking does not vary as the inlet jamperatureischanged. [

i k

(

3 4.10 QUIX Calculations .

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 the methodology discussed in Reference 4-6.

5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN 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 5-I. This effort utilized cross-sections produced by CEPAK. The nethodology discussed in this report utilizes cross-sections produced by DIT. In order to demonstrate the District's ability to utilize the models with the DIT cross-sections, n additional verification was undertaken.

J OPPD-NA-8302-NP Rev. 01 Page 18 of 259 i

S

-5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued)-

This verification is in addition to the extensive verification of these methods done by Combustion Engineering (CE) and reported in Reference 5-2.

' It is not the District's intent to repeat CE's extensive verification i effort which includes a statistical assessment of the adequacy of the uncertainties used by both CE and the District. .Rather, it is the Dis-trict's intent to demonstrate that the District can adequately.model the-Fort Calhoun core and that'the results of the District's verification L effort are consistent with those reported in Reference 5-2.

. The District's verification using DIT cross-sections utilizes data recorded '

i for Cycles 6, 7 and 8. Benchmarking was performed for the prediction of -

overall core reactivity, power distributions, reactivity coefficients, CEA worth and Xenon reactivity. The.results of the verification efforts l

include data for both CEPAK and DIT cross-sections.

. The verification uses experimental data from the' Fort Calhoun reactor and -

i f- independent calculations performed by CE and Exxon Nuclear Company, (ENC).

\ Experimental data is obtained from startup tests and core follow programs.

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

5.1 Core Reactivity The analysis of predicted reactivity for Fort Calhoun Station utilizes studies of startup tests and plant data obtained during operation at power. The parameter used to measure reactivity is the critical boron

! concentration.

Comparisons between measured and calculated critical boren concentra-tions for the unrodded HZP core are presented in Table 5-1. The i

results using the DIT cross-sections are consistent with those re-

!- ported in Reference 5-2.

i O OPPD-NA-8302-NP Rev. 01 Page 19 of 259

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

5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued)

U 5.1 Core Reactivity (Continued)

Operating plant data has been analyzed and evaluations of core reac-tivity predictions carried out. The measured and calculated full power and unrodded core critical boron concentrations for Cycles 4, 5, 6, 7 and 8 are shown in Figures 5-1 through 5-5. There is little difference between calculated curves utilizing either CEPAK or DIT cross-sections for Cycles 6, 7, and 8. Results for the operating plant data comparisons demonstrate the District's ability to calculate core reactivity.

5.2 Power Distributions Extensive comparisons of power distributions have been performed for Fort Calhoun and other CE reactors. These comparisons are contained in References 5-2 and 5-3. The data given for Fort Calhoun in Refe-rence 5-3 were supplied by the District.

5.2.1 Radial Power Distributions The District has performed comprehensive core follow calcu-lations since the start of Cycle 3 in 1976. Table 5-2 summarizes the results of comparisons between the axially integrated assembly power as calculated by ROCS and that measured by CECOR using the self-powered rhodium detector for Cycles 3, 4, 5, 6, 7 and 8. These comparisons are only performed for instrumented assemblies because CECOR calcu-4 lates the power for non-instrumented assemblies using coupling coefficients derived from the physics codes. The instrumented assembly powers are calculated by a method independent of the predictive code. Sample comparisons for Cycle 8 are included in Appendix A of this document.

O OPPD-NA-8302-NP Rev. 01 Page 20 of 259

5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) p V 5.2 Power Distributions (Continued) 5.2.1 Radial Power Distributions (Continued)

~

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

5.2.2 Axial Power Distributions The District has benchmarked the ROCS code against CECOR measured. axial power distributions. Appendix B contains comparisons of core average and selected assembly axial power distributions for Cycles 5 through 8.

1 The District has benchmarked the QUIX code against measured data by comparing the QUIX calculated ASI and the CECOR V(3 measured ASI during an axial oscillation test performed during Cycle 8 power ascension testing. The lead bank CEA's remained in the core during the entire test. The result of the comparison is shown in Figure 5-6.

i The comparisons demonstrate the District's capability to calculate axial power distributions using both ROCS and QUIX.

5.3 Reactivity 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 5-3 shows the comparison between calculated O OPPD-NA-8302-NP Rev. 01 Page 21 of 259 l

t

5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) 73

(

V) 5.3 Reactivity Coefficients (Continued) and measured ITC's for zero power startup testing at the beginning of the cycle. Also included are calculations performed by ENC, using XTG. The comparison of measured and calculated ITC's for "at power" conditions is shown in Table 5-4. The comparison of measured and calculated Power Coefficients is shown in Table 5-5. 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 5-2.

5.4 CEA Reactivity Worth The District has extensively benchmarked the ROCS code against measured l and independently calculated values of CEA reactivity worth. Tables 5-6 through 5-13 show the results of this benchmarking effort. CE performed the PDQ calculations for Cycles 1, 2 and 4. ENC performed

\ the XTG calculations for Cycles 6, 7 and 8. The District performed all 2-D and 3-D ROCS c'alculations and the Cycle 5 PDQ calculations.~

These results demonstrate the District's capability to calculate CEA worths and the results using DIT cross-sections are consistent with the results reported in Reference 5-2.

5.5 Comparisons to Critical CEA Positiens Following a Reactor Trip I Another measure of the ability of the 3-D ROCS model to accurately 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 from a reactor trip for Cycle 2. This study showed that the maximum reactivity error between measured critical parameters and calculated parameters was [ ] ap . This demonstrates the ability of the District's ROCS model to accurately model the power defect and xenon r buildup and decay.

OPPD-NA-8302-NP Rev. 01 Page 22 of 259

5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALH0UN STATION (Continued) 5.6 Comparison to Independent Radial Power Distribution Calculations Comparisons between the District's ROCS model calculations and ENC XTG model calculations of the HFP radial power distributions have been performed. Figures 5-7 through 5-12 show comparisons between CEPAK-ROCS model and XTG model calculations for the beginning and end of Cycle 6, 7 and 8. Figures 5-13 through 5-18 show. comparisons between DIT-ROCS model and XTG model calculations for the beginning and end of Cycle 6, 7 and 8. The comparisons show good agreement between the independent models.

5.7 The District's Ongoing Benchmarkino Program Much of the data reported in this section was drawn from the Dis-trict's ongoing benchmarking program. This program includes startup physics testing predictions, reactor testing analysis and a core p follow effort. The verification program has been updated for Cycles 9 and 10. Tables are updated within Section 5 which contain cycle by cycle data. New tables and figures for Cycles 9 and 10 are located in Appendix C. This program will provide additional verification data in the future.

5.8 Summary The District has an ongoing neutronics methodology verification program. The results of this verification program for previous cycles demonstrate the ability of the District to utilize the neutronics methods described in this document.

. O OPPD-NA-8302-NP Rev. 01 Page 23 of 259

4 t

Table 5-1

( Unrodded HZP Critical Boron Concentrations Calculations 3-D* 2-D* 3-D Measured ROCS ROCS ROCS PDQ i Cycle ppm (CEPAK) (CEPAK) (DIT) (CEPAK) XTG 1 933 - -

2 1240 - -

. 3 1000 - -

,. 4 1027 - - - -

5 1242 - - - -

6 1230 - -

7 1241 -

l 8 1240 - -

9 1518 - - - -

l 10 1474 - - - -

1

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

3

'l l

t l

l l:

l O OPPD-NA-8302-NP Rev. 01 Page 24 of 259 l

w- - - - _--,--v.

.mn_,_. _ _ _ . , _ _ _ _ , _ . . _ _ , . . , , . , , , , , , _ _ _ _ _ , , , _ _ _ , , , _ _

! Table-5-2 Summary of Comparisons of Measured and Calculated L' -

Integral Assembly Relative Power Densities Cycle .CEPAK DIT Nominal Burnup Power Cycle (MWD /MTU) (%) (%)  % of Full Power 3 80 -- 45 3 177.5  ! - 60 3 510 - 95 3 800 I - 100 3 1000 - 100 3 1428 - 100 l

3 2510  ! - 100

3 3100 - 100 3 3500 - 100 3 4000 - 100 3 4500 - 100 4

3 5200 - 100 3 5800 - 100 3 6400 - 100 3 7200 - 90 3 7715 - 80 4

4 200 - 100 4 1000 - 100 4 2000 - 100 i

4 3000 - 100 4 4000 - 100 4 5000 - 100 V 4 6000 ~

- 100 OP'PT-NA-8302-NP Rev. 01 Page 25 of 259 i

Table 5-2 Sumary of Comparisons of Measured and Calculated

-(~')

v' Integral Assembly Relative Power Densities (Continued)

Cycle CEPAK DIT Nominal Burnup Power Cycle (MWD /MTU) (%) (%)  % of Full Power 4 7000

~'! -

100 4- 8200 }

100 5 300 -

100 5 1000 i -

100 5 2000 -

100 5 3000 -

100 5 4000 , -

100 5 5000 4 100 5 6000 -

100 6 50 66 4 -.

6 500 -

100 6 1000 -

100 6 2000 90 6 3000 -

65 6 4000 75 6 5000 -

75 W 5"l 6 5800 75 6 6500 100 6 7500 -

50 6 8500 95 I

6- 9500 95 6 10500 95

, 7 135 ~ - ~

70 OPPC-NA-8302-NP Rev. 01 j Page 26 of 259

Table 5-2

) Summary _of Comparisons of Measured and Calculated

's Integral Assembly Relative Power Densities (Continued)

Cycle CEPAK DIT Nominal Burnup Power Cycle (MWD /MTU) (%) (%)  % of Full Power 7 500 -- 100 l

7 1000 -

100 7 2000 100 7 3000 - 100 7 4000 100

~

7 5000 - - 100 7 6000 100 7 7000 - 100

~ ~

7 8000 t 100 4' 7 9725 100 8 50 45 8 250 - 100 8 1000 100 8 2000 100 l

i i

O  ;

OPPD-NA-8302-NP Rev. 01 Page 27 of 259 l

Table 5-3 Low Power Physics Isothermal Temperature Coefficients Boron CEPAK* DIT Concentration Measured ROCS ROCS XTG Cycle (ppm) (An/ F) (Ap/*F) (Ap/ F) ( Ao/ F) 1 993 0.26

• 10 4 - -

1 854 -0.11

• 10 4 - -

2 1240 0.41

• 10 4 - -

2 1198 0.32

• 10 4 - -

2 1164 0.09

• 10 4 - -

3 1000 -0.078

• 10 4 - -

4 1020 0.14

• 10 4 - -

5 1228 0.20

• 10 4 - -

~

6 1213 .0.23

• 10 4 7 1213 0.12

• 10 4 8 1240 0.16

• 10 4 - _.

- -~~~

9 1457 0.30

• 10 4 - -

10 1457 0.23

• 10 4 -

• Calculated results were biased by 0.20

• 10 4 Ap/ F OPPD-NA-8302-NP Rev. 01 Page 28 of

_ _- . ._. . . _ - _ . ~ . . . -_ _ .. __

~

4 Table 5-4, '-

l Comparison of Calculated.and Measured

. Isothermal Temperature Coefficient

[ 80C Calculated

• Calculated Critical Boron Measured CEPAK-ROCS DIT-ROCS

<- Percent of Concentration ITC ITC ITC Cycle Rated Power (ppm) (*104ap/ F) (X104ap/ F) (X104ap/*F) 1 2 69(I) 927 -0.28 --

i 3 46(1) 720 -0.41  ! -

4 92(1) 690 -0.42 -

5 93(I) 876 -0.19 -

~

6 95(I) 848 -0.46  ;

7 96(2) 817 -0.52 ~

. 8 79(2) 817 -0.84 i

9 94(2) 1036 -0.39 - i O 10- 95( ) 1017 -0.48 -

'I i

1

O OPPD-NA-8302-NP Rev. 01 Page 29 of 259

-Table 5-4 Comparison of Calculated and Measured Isothermal Temperature Coefficient

(~'i

-U (Continued)

E0C Calculated.

Calculated

• Critical Boron Measured CEPAK-ROCS DIT-ROCS Percent of Concentration 'ITC ITC .

ITC Cycle Rated Power (ppm) (*104ap/ F) (X104ap/ F) (X104ap/ F) 1 75(I) 239 -0.98 -

2 46(1) 104 -1.62 -

3 90(1) 62 -1.65 -

4 95(1) 44 -1.41 -

5 94(1) 296 -0.97 -

6 96(2) 307 -1.51 7 95(2) 192 -1.85 8 95(2) 292 -1.86 -

l 9 96(2) 300 -1.46 -

10 95(2) 302 -1.53 -

- w.

I J

(1) Full Rated Power = 1420 MWt l (2) Full Rated Power = 1500 MWt

• B0C calculated results were biased by 0.20

• 10 4 ap/'F and E0C calculated results were biased by 0.40

• 10 4 ap/ F.

OPPD-NA-8302-NP

Rev. 01 Page 30 of 259

Table 5-5 Comparison of Calculated and Measured Power Coefficient CEPAK-ROCS DIT-ROCS

{v')- Percent Measured Calculated Calculated of Critical Power Power Power Burnup Rated Boron Coeff. Coeff. Coeff..

Cycle MWD /MTU Power Conc. ( Ap/% Power) ( Ap/% Power) ( Ap/% Power) 2 10877 46(I) 104 -1.95 x 10 4 -

3 157 46(I) 720 -1.47 x 10 4 -

3 1513 90(I) 535 -1.12.x 10 4 -

3 4183 90(I) 309 -1.31 x 10 4 -

3 7208 90(I) 62 -1.48 x 10 4 -

4 267 92(1) 690 -1.04 x 10 4 -

4 4690 94(I) 288 ~ -1.12 x 10 4 -

4 8027 95(I) 44 -1.10 x 10 4 ,

5 426 93 II) 876 -1.05 x 10 4 -

5 6815 94(I) 296 -1.25 x 10 4  ;

6 400 95(II 848 -1.11 x 10 4 T 6 6467 96(2) 307 -1.45 x 10 4 .

7 450 96(2) 817 -0.98 x 10 4 7 6900 95(2) 283 -1.30 x 10 4 -

7 7800 95(2) 192 -1.57 x 10 4 I 8 459 79(2) 817 -1.18 x 10 4 _.

8 6150 95(2) 292 -1.70 x 10 4 -

9 420 94(2) 1036 -1.64 x 10 4 -

9 9663 96(2) 300 -1.57 x 10 4 -

10 583 95(2) 1017 -1.24 x 10 4 -

10 9261 95(2) 302 -1.40 x 10 4 -

(1) Full Rated Power = 1420 MWt (2) Full Rated Power = 1500 MWt OPPD-NA-8302-NP Rev. 01 Page 31 of 259

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

Table 5-6 Cycle 1 CEA Worths p.

D' Calculated Calculated Calculated CEPAK-ROCS CEPAK-ROCS CEPAK-PDQ Measured 3-D 2-D 2-D Group (%ao) (% Ap) (% Ap) l (%ao) 4 0.58 3 0.57 2 2.01 A 3.06 B 2.10 Total (4+3+2+A+8) 8.32 Table 5-7 1

Cycle 2 CEA Worths n Calculated Calculated Calculated CEPAK-ROCS CEPAK-ROCS CEPAK-PDQ Measured 3-D 2-D 2-D Group (%Ao) (%Ap) (%Ao) (%Ao) 4 -0.65 3 0.41 2 1,67 _ _

1 0.95 -

! Total (4+3+2+1) 3.68 -

4 O

OPPD-NA-8302-NP Rev. 01 Page 32 of 259

Table 5-8

. Cycle 2 CEA Worths I~'T Calculated Calculated l

' \/ -CEPAK-ROCS CEPAK-ROCS

)

Measured 3-D 2-D 1 Group: (%Ap) (%Ap) (%Ap) j i 4 0.74 J

3 0.59 E

l 2 1.96 1 0.80 -

1 Total (4+3+2+1) 4.09 -

i Table 5-9 t

Cycle 4 CEA Worths Calculated Calculated CEPAK-ROCS CEPAK-ROCS O. Group Measured

(%Ap) 3-0

'(%Ap) 2-D

(%Ap)

~

4 0.63

! 3 0.60 2 1.90 1 0.92 -

Total (4+3+2+1) 4.05 -

i i

i O l l OPPD-NA-8302-NP

Rev. 01 Page 33 of 259

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

k

- Table 5-10 i Cycle 5 CEA Worths-

/)

V Calculated CEPAK-ROCS Calculated

^ CEPAK-ROCS Calculated CEPAK-PDQ

-Measured 3-D- 2-0 2-D

-Group (%ao) (%6o) (%Ao) (%Ao) 4 0.57 l j s i . t 3 0.67' j'  ! ~I i.

I l

2 1.40 f _, _ _ . _

{

1 0.99  ! - -

1 Total I (4+3+2+1) 3.63 l

' Table 5-11 Cycle 6 CEA k'orths Calculated Calculated Calculated Calculated XTG CEPAK-ROCS CEPAK-PDQ CEPAK-PDQ

, Measured 3-D 2-D 3-D Group (%Ao) (%Ao) (%An) (%An) (%An) 4 0.52 -

! 0.66 -

3 2 1.57 -

1 0.93 -

! Total- r i . (4+3+2+1) 3.68 l

- -, - - . _ _ L - \

l l

l

!O

' OPPD-NA-8302-NP  :

Rev. 01 Page 34 of 259

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

\ _

Table 5-12 Cycle 7 CEA Worths

\ ; e,'

3 Calculated Calculated Calculated Calculated (d

Measured XTG CEPAK-ROCS 3-D CEPAK-PDQ 2-D CEPAK-PDQ 3-D Group (%Ao) , (%Ao) (%Ao) (% ao) (% t o) 4 0.49 -

3 0.47 -

2 1.65 -

1 0.66 -

P Total -

(4+3+2+1) 3.27 Table 5-13 Cycle 8 CEA Worths Calculated Calculated Calculated Calculated XTG CEPAK-ROCS CEPAK-PDQ CEPAK-PDQ Measured 3-D 2-D 3-D s Group (%30) (%ao) (%3 n) (%An) 1(% An)

~ ~ ~ ~

' - - ~ - ~

4 0.58 3 0.63 -

2 0.99 1 1.00 Total (4+3+2+1) 3.20

_ _ _ a

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6,0 REFERENCES Section 2.0 References 2-1 ENDF-313, " Benchmark Testing of ENDF/B Data for Thermal Reactors, I  ;

J Archival Volume," July, 1981.

2-2 A. Jonsson, J. R. Rec and U. N. Singh, " Verification of a Fuel Assembly Spectrum Code Based on Integral Transport Theory," Trans.

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

2-3 CENPD-226-P, "The ROCS and DIT Computer Codes for Nuclear Design,"

December, 1981.

2-4 System 80 PSAR, CESSAR, Vol. 1, Chapter 4.3.3, Amendment No. 3, June 3, 1974.

2-5 W. R. Cadwell, "PDQ-7 Reference fianual," WAPD-TM-678, January, 1968.

2-6 T. G. Ober, J. C. Stark, I. C. Richard and J. K. Gasper " Theory, Capabilities, and Use of the Three Dimensional Reactor Operation and Control Simulator (ROCS)," Nucl. Sci. Eng., 64,605,(1977).

2-7 System 80 PSAR, CESSAR, Vol. 1, Appendix 4A, Amendment No. 3, June 3, 1974.

2-8 CENPD-199-P, Revision 1-P, "CE Setpoint Methodology," April 1982.

Section 3.0 References

(]

3-1 CENPD-199-P, Revisfor 1-P, "CE Setpoint itethodology," April 1982.

3-2 CENPD-226-P, "The ROCS and DIT Computer Codes for Nuclear Design,"

December, 1981.

Section 4.0 References 4-1 CENPD-153, Revision 1-P-A, " Evaluation of Uncertainties in the Nuclear Power Peaking Measured by the Self-Powered Fixed In-Core Detector System," May, 1980.

4-2 " Development and \ erification of a Fuel Temperature Correlation for Power Feedback and Reactivity Coefficient Application," P. H. Gavin and P. C. Rohr, Trans. Am. Nucl. Soc. 30, p. 765, 1978.

4-3 A. F. Henry, " Computation of Parameters Appearing in the Reactor Kinetic Equations," WAPD-142, December 1955.

4-4 R. W. Hardie, W. W. Litke, Jr. , " PERT-V, A Two Dimensional Perturbation Code for Fast Reactor Analysis," BNWL-1162.

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6.0 REFERENCES

(Continu:d)

Section 4.0 References (Continued)

', , 4-5 CENPD-226-P, "The ROCS and DIT Computer Codes for Nuclear Design,"

5 . December, 1981.

4-6 CENPD-199-P, Revision 1-P, "CE Setpoint Methodology," April 1982.

Section 5.0 References 5-1 CEN-242-(0)-P, OPPD Responses to NRC Questions on Fort Calhoun Cycle 8, February 18, 1983.

5-2 CENPD-226-P, "The ROCS and DIT Computer Codes for Nuclear Design,"

! December, 1981.

5-3 CENPD-153-P, " INCA /CECOR Power Peaking Uncertainty," May, 1980.

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