ML20078D965
| ML20078D965 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/30/1983 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19268E058 | List: |
| References | |
| OPPD-NA-8302-NP, NUDOCS 8310050246 | |
| Download: ML20078D965 (115) | |
Text
{{#Wiki_filter:.. . . . _ _ _ _ . a d Omaha Public Power District Nuclear Analysis Reload Core Analysis I i Neutronics Design Methods and Verification OP PD-NA-8302- NP September 1983 1 E
+ 4
)
8310350246 830926 PDR ADOCK 05000205 PDR p
ABSTRACT This document is a Topical Report describing Omaha Public Power District's reload core neutronics design methods for application to Fort Calhoun Station 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. 5 i
Proprietary Data Clause This document is the property of Omaha Public Power District (0 PPD) and contains proprietary infonnation, indicated by brackets, developed by Combustion Engineering (CE) and Exxon Nuclear Company, Inc. (ENC). The CE and ENC infonnation was purchased by OPPD under proprietary infonnation agreements. 11 -
. . _ . . . ~ _
r Table of Contents Section Page
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 PD0-X 3 2.2.2 ROCS 4 2.2.3 qulX 4 3.0 FORT CALHOUN PHYSICS MODELS 6 3.1 Neutron Cross-Sections 6 3.2 Diffusion Theory Models 7 3.2.1 PD Q-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.d 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 L T t iii
Table of Contents (Continued)
/
Section Pace 5.0 VERIFICATION OF NEUTRONICS 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 Summary 23
6.0 REFERENCES
53 APPENDIX A Cycle 8 Radial Power Distribution Canparisons APPENDIX B Axial Power Distribution Canparisons
~
I L I l g iv 1 -,...:,-,-.,:.,,_
Omaha Public Power District Reload Core Analysis Methodology 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 ex-perimental 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 cal-culations to measurements and independent calculations are performtj using the same codes and canputational methods used in the Fort Cal-houn reload core design efforts. The basic physics models, supplied by Combustion Engineering (CE), are described in Section 2.0. Sec-tion 3.0 describes the District's application of these models to the Fort Calhoun reactor. Section 4.0 presents the application of these physics nodels 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-tio ns . Section 6.0 contains the individual references. 2.0 BASIC PHYSICS MODELS The District's neutronics design analysis for the Fort r-lhoun core is based on a combination of multi-group neutron spectrum calcula-tions, which provide cross-sections appropriately averaged over a few broad energy groups and few-group 1, 2- and 3- dimensional diffusion theory calculations, which result in integral and differential reac-tivity effects and power distributions. Calculations are performed with the aid of conputer programs enbodying analytical procedures and fundamental nuclear data 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 reconmended by the 1
. .. - .7
2.0 BASIC PHYSICS MODELS (Continued) + 2.1 Neutron Cross-Sections (Continued) cross-section evaluation working group (Reference 2-1). These recmmendations consist of changes to the shielded resonance of U233, and the Watt fission spectrums of U235 and Pu239, and changes in A for U235 and Pu239 Few group cross-sections, for subregions of the core that are represented in spatial diffu-sion calculations, (e.g., fuel pin cells, moderator 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 accomodate the temperature feedback routines within the diffusion theory models. The DIT code perfonns 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 absorber rods, and incore flux detectors. The essential feature of DIT, which distin-guishes it frm the traditional methodology is that the spec-trum spatial averaging procedures are based on calculations in two-dimensional gemetry. Hence few approximations to the gem-etry representation are necessary. The use of nodal transport theory has made it feasible to retain discrete pin gemetry in both the fine and broad energy group calculations. A more cm-plete description of the DIT procedures for generating few-f group neutron cross-sections can be found in References 2-2 and 2-3.
.5 Previously, the District utilized the CEPAK program to produce few-group cross-sections. These cross-sections were also gener-l- ated as functions of fuel and moderator temperature. Capa ri-sons of calculated and measured data reported in Section 5.0 in-l I
clude calculations performed using the CEPAK program. 2
i I 2.0 BASIC PHYSICS MODELS (Continued) i ! 2.1 Neutron Cross-Sections (Continued) l The CEPAX 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 programa, are interlinked in a consistent way with inputs from differential cross-section data from an extensive library. A description of the CEPAK pro-cedures 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 phys-ics parameters for Fort Calhoun Station consist of the PDQ-X, ROCS, and QUIX computer codes. The PD0-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 and CEA worths. The QUIX code is executed in one dimension to calculate axial power distribu-tions and CEA worth [ ]. 2.2.1 PD 0-X The PDQ-X program is an extension of PDQ-7 and HARMONY programs (Reference 2-5) to include the following op-tional capabilities: (1) Fuel temperature feedback in the two-dimensional geometry option, 3 (2) Fuel and moderator feedback in the three-dimen-sional geometry option, (3) Poison content criticality searches, and , (4) Spatial feedback on the power distribution with I fuel and moderator temperature in the 1-dimen-sional geometry option. 3
2.0 BASIC PHYSICS MODELS (Continued) l 2.2 Diffusion Theory Model (Continued) l l 2.2.1 PDQ-X (Continued) l l PDQ-X enploys macroscopic (static) or microscopic (depletion) 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 cen-tered higher order finite difference fonnulation. It incorporates closed channel thermal hydraulic modeling into its evaluation of the interaction of neutron flux effects and the macroscopic physical and thermal prop-i erties of distributed materials. Because of its nodal structure and course mash, ROCS is more efficient than PDQ-X for evaluating a core's static and depletion de-pendent properties. ROCS also employs macroscopic (static) or microscopic (depletion) cross-sections gen-erated by the methods described in Section 2.1. A
~
more canplete description of the ROCS program is found in References 2-3 ard 2-6. 2.2.3 OUIX The QUIX program is a one-dimensional (axial) represen-tation of the core used to determine static and time 1 dependent reactivities and power distributions at sel-ected stages of depletion. This program solves the neutron flux and associated eigenvalue in problems con-taining up to 140 distinct regions or. conpositions with variable mesh. intervals. The macroscopic cross-section distributions, fission product yields, and
.4
~ . _, _ - .. . _ .. -_ _ _ _ _ _ _ _
2.0 BASIC PHYSICS MODELS (Continued) 2.2 Diffusion Theory Model (Continued) 2.2.3 QUIX (Continued) xenon and baron microscopic cross-sections required as input to QUIX are obtained fron either a one-dimension-al PDQ-X calculation or a three-dimensional ROCS calcu-lation. Local power density (fuel tenperature) feed-back is included by modifying the point wise macroscop-ic absorption and removal cross-sections. The change in cross-sections is represented by a function of the difference between the local axial power density and the referenced power density. Moderate density feed-back is included by relating changes in the macroscop-ic absorption and removal cross-sections to the local hydrogen number density which is calculated from en-thalpy at each axial segment. These cross-section functions are generated in such a way that the fuel and moderator tenperature coefficients calculated by QUIX are equal to or conservative with respect to the fuel and moderator tenperature coefficients calculated by ROCS. The axial reflector cross-sections input to QUIX are detennined in such a way that the steady state axial power distribution geaerated by QUIX matenes the axial power distribution generated by ROCS. Details of the above treatments are given in Reference 2-7. In addition to the eigenvalue problem, QUIX will perfonn four types of searches to obtain a specific ., eigenvalue, viz. , a uni fona poi son search, buckling search, CEA region boundary search, and a moderator density dependent poison search. The unifann poison search assumes an axially constant macroscopic absorp-tion cross-section whereas the moderator density de-pendent poison search assumes a distributed macroscop-5
2.0 BASIC PHYSICS MODELS (Continued) 2.2 Diffusion Theory Model (Continued) 2.2.3 OUIX (Continued) ic absorption cross-section dependent upon the axial moderator density. The moderator density dependent search is used to simulate the reactivity effects of + the soluable boron in the reactor coolant. Through the use of rod shadowing factors, shape an-nealing factors and shape index biases, the QUIX pro-gram has the capability of simulating excore detector response expected during normal operation. The proce-dures used for these simulations are described in Ref-erence 2-8. 3.0 FORT CALHOUN PHYSICS MODELS The District utilizes the basic CE physics models described in Sec-tion 2.0 to model the Fort Calhoun reactor core. The computer codes which embody these basic physics models are maintained on the CE com-puter system at Windsor, Connecticutt. The District accesses these ! computer codes through a time sharing systen. CE maintains all docu-mentation and quality assurance programs related to these computer codes. The following paragraphs discuss the specifics of the Fort Calhoun models. 3.1 Neutron Cross-Sections The two-group neutron cross-sections utilized in the ROCS and i 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 fuel assembli es. The cross-sections have been generated for the District by CE and are based on information supplied by the District. 6
3.0 FORT CALHOUN PHYSICS MODELS (Continued) 3.1 Neutron Cross-Sections (Continued) The cross-sections utilized to nodel the Fort Calhoun reactor are in the fona of universal table sets. The two-group cross-sections are generated as functions of enrichment, fuel temper-ature, moderator temperature, burnup and in the case of shimmed fuel assemblies, B4C shim number density. The table sets are applicable over a fuel temperature range from roon tenperature to 1800 K and a moderator tenperature range from room tenpera-ture 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 ther-mal 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 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 PD0-X The District's PDQ-X model is a two-group, two-dimen-sional 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 assem- ' bly and the core shroud, the core shroud, the water i [ gap between the core shroud and the core barrel, the l core barrel and a portion of the water gap between the l core barrel and the thermal shield. The model i s rep- - , resentative of the core between 20% and 80% of full core height. 7
r 3.0' FORT CALHOUN PHYSICS MODELS (Continued) 3.2 Diffusion Theory Models (Continued) i 3.2.1 PD0-X (Continued) J The PDQ-X model is used to simulate the expected mode of operation in the cycle being analyzed. This calcu-lation results in material distributions.and radial peaking factors which are used in the safety analysis and setpoint generation. 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 thennal 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 dicusses the impact of operation with a time averaged lead bank insertion of [ ]. 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.
8 J
3.0 FORT CALHOUN PHYSICS MODELS (Continued) 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 calcelations. 4.0 APPLICATION OF PHYSICS TETHODS Previous sections have focused on the reactor physics models utilized by the District to model the Fort Calhoun reactor. In this section, calculations of the various core paraneters used in the safety analysis are described. The main parameters considered are the radial reaking factors (FR and Fxy), the moderator temperature coefficient, the fuel temperature or Doppler coefficient, the neutron kinetics parameters, CEA drop data, CEA ejection data, CEA scram reactivity, reactivity insertion for the steamline break cooldown, radial peaking data for the asymmetric steam generator event, and axial power distributions. 4.1 Radial Peaking Factors The radial peaking factors, FR and 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 distributions. 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 re-quired to the peaking factors calculated by PDQ. The values of . FR for the unrodded core are obtained by multiplying the inte-grated assembly powers from ROCS by the pin to box ratio ob-
'tained from PDQ. The value of FR for 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 in that plane and summing these values for all planes and dividing by the number of planes.
9
4.0 APPLICATION 0F PHYSICS METHODS (Continued) 4.1 Radial Peaking Factors (Continued) The uncertainties for the radial peaking factors are given in Reference 4-1. The physics models are used to calculate the expected values of FR and Fxy. The actual values of FR and Fxy used in the safety analysis are chosen to be conservatively high with espect to those anticipated during the core life. 4.2 Reactivity Coefficients The ROCS models are used to calculate the moderator temperature coefficient (MTC) and tne fuel temperature coefficient (FTC). The MTC is defined as the change in reactivity per degree change in moderator temperature. Calculationally, the MTC at a tenperature of Tmod is detennined by running three calcula-tions; one at Tmod, one at Tmod + 10 F and one at Tmod -10*F. The MTC at a tenperature of Tmod is the average of the two calculated values. The reactivity change is calculated with the ROCS model by varying the inlet tenperature 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 reac-tivity 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 irodels by means of power feedback options. The representation of the variation
~
in the few group cross-sections 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 the few-( - group cross-sections. The second portion requires establish-ment of correlations between fuel tenperature (i.e., effective 1 10
r 4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.2 Reactivity Coefficients (Continued) fuel 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 detemined by ROCS. Typically, a temperature interval of 50 F is used to determine this coeffi-cient. 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 safety analysis are chosen to conservatively bound expected values of 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 tne Doppler coefficient when it is used in the safety anal-ysis calculations. 4.3 fleutron Kinetics Parameters The neutron kinetics parameters 8,1, and the neutron lifetime, f.*, are calculated using Combustion Engineering's BETAF compu-ter code. The technique utilized to calculate the kinetics parameters and the neutron lifetime is based on first order per- i turbation theory. Details of the perturbation approach are dis-cussed in References 4.3 and 4.4. The computer program, BETAF, uses data from integral files created by direct and adjoint flux solution PDQ calculations. [ l l 11
4.0 APPLICATION OF PHYSICS METHODS (Continued) i 4.4 Dropped CEA Data The neutronics data unique to the dropped CEA analysis are the
- values of FR and Fxy following the drop of a CEA and the reac-tivity worth of the dropped CEA. The values of FR and Fxy in-crease 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 Fg and Fxy occur far away from the dropped CEA, the intra-assembly power distribution is not perturbed. Therefore, the " post drop" value of FR and Fxy can be calcu-lated by multiplying the " pre-drop" values of FR and Fxy by the ratio of the assembly power after and before the drop of the CEA. This ratio is the distortion factor. The distortion fac-tor 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 asseably RPP from a radial power distribution at the same power level and time in core life with-out a dropped CEA. [.
]
The distortion factor and dropped CEA reactivity worth can be calculated using the 2-D or 3-D ROCS model. The 2-D ROCS calcu-lations 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 FR distortion factor is calcu-lated for a specific CEA insertion and power level. [-
12
1 l l 4.0 APPLICATION OF PHYSICS METHODS (Continued) j l l 4.4 Dropoed CEA Data (Continued)
] Sufficient margin exists at the lower power levels [
! ] for the FR dependent DNBR calculations does not adversely effect operating margin. The " post drop" value of FR using the 3-D FR distortion factor is calculated by multiplying the " pre-drop" value of FR for the particular CEA insertion and power level by the 3-D FR distortion factor. The 2-D and 3-D ROCS " post drop" power distributions are calcu-lated with fuel temperature and moderator taperature feedback. The calculations 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. Because the [
.] of the dropped CEA during the fuel cycle and be-cause of the [
.] documented in Refer-ence 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 Refer-ence 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 -> F7y 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 Fxy is obtained directly fra the PDQ calculation. The post ejection value of Fxy is obtained from the 2-D ROCS calculation 13
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.5 CEA Ejection Data (Continued) by multiplying the 2-D ROCS post ejection assembly RPD by a con-servatively high pin to box ratio. The ROCS methodology is con-servative with respect to the more exact PDQ method. The eject-ed CEA reactivity worth is obtained directly from either calcu-lation. Both the PDQ and RCCS post ejection power distribu-tions are calculated without moderator or fuel tenperature feed-back. The post ejection value of Fqis calculated by multiplying the post ejection value of Fxy by the maximum value of F z , the azi-muthal tilt allowance, the augmentation factor, the engineering heat flux factor, the fuel densification factor, and the Fq uncertainty documented in Reference 4-1. A[ ] uncer-tainty is applied to the ejected CEA worth. 4.6 CEA Reactivity 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 reactiv-ity), the generation of the scram reactivity curves, and the calculation of required shutdown margin. The CEA scram reactivity worth at HZP is calculated by obtain-ing the net worth for all CEA's between the HZP PDIL CEA posi-tion and the fully inserted position and subtracting the worth of the highest worth stuck CEA. These calculations are done using the ROCS model. A[ ] uncertainty is applied to the HZP CEA scram reactivity worth. The HZP CEA scran reac-tivity for the CEA ejection trt...;ient is calculated in a simi-lar 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 HFP net worth for all CEA's between the HFP PDIL CEA position and 14
4.0 APPLICATION OF PHYSICS METHODS (Continued) l 4.6 CEA Reactivity (Continued) the fully inserted position and subtracting the worth of the highest worth stuck CEA. The themal hydraulic axial gradient reduction allowance, the moderator void collapse allowance, and the loss of worth between HFP and HZP are alsc 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 scram worth because the SLB reactivity insertion curves are calcu-lated with the CEA's fully inserted. The moderator void collapse allowance is 0.0% Ap at BOC ard 0.1% ap at E0C. The themal hydraulic axial gradient reduction allowance is 0.2% ao at BOC and 0.4% Ap at E0C. A [ ten per-cent] 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 th.t the worth of the ejected and highest stuck worth CEA's are subtracted fran 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 meth-odology discussed in Reference 4-6. 5 The calculatian of the required shutdown margin is only per-fomed at HZP since the shutdown margin at power is controlled by the PDIL. The available HZP shutdown margin is equivalent l to the HZP CEA scram reactivity. l l 15
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.7 CEA Withdrawal Data
.)
l The reactor core physics data unique to the CEA withdrawal anal-ysis is the maximum differential CEA worth. This is the maxi-mum 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 cambined with the maximum CEA with-drawal 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 sequen-tial withdrawal of the CEA banks from the HZP PDIL to an all rods out condition. The 3-D ROCS model is utilized to calcu-late this parameter. The calculations are perfomed 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 The reactor core physics data unique to the steam line break transient analysis is the reactivity insertion due to the cool-down of the moderator. There are two sources of this reactiv-ity insertion. The first is the positive reactivity insertion due to the increasing density of the moderator as the cooldown progres ses. The second is the reactivity insertion due to the Doppler coefficient as the effective fuel temperature changes. 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 tenperature coefficient calculated by the ROCS model corresponds to the most negative Technical Specifi-cation limit. The reactivity insertion calculations are per-fomed with all CEA's except the most reactive CEA inserted in the core. 16
I 4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.8 Reactivity Insertion for Steam Line Break Cooldown (Continued) I 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 inser-tion vs. moderator temperature fran a hot zero power tenpera-ture of 532*F to 212*F. The Doppler reactivity insertion curve for the hot zero power case is also calculated by steadily decreasing the inlet temper-ature 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. The moderator density reactivity insertion curve for the full power case is calculated by decreasing the power level and core average coolant temperature from full power to the hot zero power inlet tenperature and then successively lowering the inlet temperature as in the hot zero power case. Only mod-erator tenperature 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 addi-tional uncertainty 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 event [ 17
.- .. . ~ ~ - -
4.0 APPLICATION OF PHYSICS METHODS (Continued) 4.9 Asymmetric Steam Generator Event Data (Continued)
] For the range of temperatures considered, the intra-assembly peaking does not vary as the inlet temperature is changed. [
]
]
4.10 QUIX Calculations The District utilizes the QUIX model to perfonn various axial shape analyses related to the generation of the reactor protec-tive 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 neu'.ronics . models used in the reload core analyses. The results of the previous . District verification efforts were reported in Reference 5-1. This effort utilized cross-sections produced by CEPAK. The methodology 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, additional verification was undertaken. 18
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 exten-sive verification effort which includes a statistical assessment of the adequacy of the uncertainties used by both CE and the District. Rather, it is the District's intent to demonstrate that the District can adequately model the Fort Calhoun core and that the results of the District's verification effort are consistent with those reported in Reference 5-2. The District's verification using DIT cross-sections utilizes data recorded for Cycles 6, 7 and 8. Benchmarking was perfomed for the prediction of overall core reactivity, power distributions, reactiv-ity coefficients, CEA worth and Xenon reactivity. The results of the verification efforts include data for both CEPAK and DIT cross-sec-tions. The verification uses experimental data fran the Fort Calhoun reactor and independent calculations perfomed by CE and Exxon Nuclear Com-pany,(ENC). 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 con-putations. 5.1 Core Reactivity The analysis of predicted reactivity for Fort Calhoun Station utilizes studies of startup tests and plant data obtained dur-ing operation at power. The parameter used to measure reactiv- i ity is the critical boron concentration. Comparisons between measured and calculated critical boron con-centrations for the unrodded HZP core are presented in Table 5-1. The results using the DIT cross-sections are consistent with those reported in Reference 5-2. i 19
5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) 5.1 Core Reactivity (Continued) Operating plant data has been analyzed and evaluations of core reactivity predictions carried out. The measured and calculat-ed 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 cmparisons demon-strate the District's ability to calculate core reactivity. 5.2 Power Distributions Extensive conparisons of power distributions have been per-formed for Fort Calhoun and other CE reactors. These conpari-sons are contained in References 5-2 and 5-3. The data given for Fort Calhoun in Reference 5-3 were supplied by the Dis-trict. 5.2.1 Radial Power Distributions 1 The District has performed conprehensive core follow calculations since the start of Cycle 3 in 1976. Table 5-2 summarizes the results of conparisons be-tween the axially integrated assembly power as calcu-lated 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 in-strumented assemblies because CECOR ' calculates the , power for non-instrumented assemblies using coupling coefficients derived fran the physics codes. The in- ; strumented assembly powers are calculated by a method l independent of the predictive code. Sample conpari-sons for Cycle 8 are included in Appendix A of this ! document. ! 20
.-_---_____.--_______________._____--____--__.---_r--------_---.-__
qq..y o .: ' s
-t \. . _
- s', ;\'
5.0 VERIFICAT!DN 05 NEUTRONICSLM0DELS FOR FORT CALH0UN STATION (ContinuedF I . [ , , . s 3 5.2 Power Dis'tribution's (Continded) S' .
. . 5.2.1 Radial Power Distribution (Continued)
The'extensf ve comparison betwer.- the calculated and b
- measured Eadial power distrfbdti,$n.;
o verif tes the capa-
. bility of the District to calculate these power distri-butions. j a
n .
'S.2.2 #
Arial Power Distributions ' 2
\ q\ %
r [ >3 l - -
!+ i s x a 3 i TheDistricthasbenchmarl[edtheROCScodeagainst j , . CECOR measured axial ' power distVibutions. /ipper. dix B contains comparisons of core average and selected ,
1 l
~
assemblyjaxial power, distributfons"for Chles 5' . through 8. 4
,11 N i '
i t .; , 4
, lhe District has benchaarked the QUIX code agejnst
, 4
*- / measured ' data by comparing the QllIE calculated ASI and .
^
3 ' '( the CECOP measured ASI dur ing a6arIal hscillation '
^
1$ test performed during Cycle 8 power l ascension testing. 4 p ( , The lead bank CEA's remained trAthe" core during the
, -p entire test., The result of the comparison is shown in s , s 3 .j < Figure 5-6. ~ ~!
' '; The .,- comparisons de'aicnst' A} rate the Cistrict's capcbility 4
- 1 -
' ~
, t g to, cai<,ulate , axial power distributions:usf ng b$th ROCS 4 ; andQUIX./ .
,., r j
> 1 l x <
r
.t .- n ./ 1 '\ .
t . - r t , l -x 5.3 . Reactivity Coefficients _, ,
,.l )
, 3. , , .
c l s. O' ' l + fThe cappbility ofJthe District / s ROCS model to predict. the Iso-l rs, 7 .t . a theraa? Temperature Coefficient (ITC) and,the- Power Coefficient t ,~ n . ? ~
~?
s
'PC) has been benchmarked against physics test,s conducted at 3 ;rort -Calhoun for all%)reting cycles. Table 5-3 shows the com-
?;
m a 1 .s: n %
- t >
.l ,
,[ 7 , ,. . >
' I '
,s
"/ '~(! .
, I, ,,
^.
>X , *i,
- f. '
p, e 21' b',.' ~'I17 4 I
- 7U'; &
tg
, ,1% , . - ., < . - - - , a '
N3 %X - - < - - - - - - --
5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) 5.3 Reactivity Coefficients (Continued) parison between calculated and measured ITC's for zero power startup testing at the beginning of the cycle. Also included are calculations perfomed by ENC, using XTG. The camparison 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 re-sults reported in Reference 5-2. 5.4 CEA Reactivity Worth The District has extensively benchmarked the ROCS code against mesured and independently calculated values of CEA reactivity worth. Tables 5-6 through 5-13 show the results of this benen-marking ef fort. CE perfomed the PDQ calculations for. Cycles 1, 2 and 4. ENC perfonned the XTG calculations for Cycles 6, 7 and 8. The District perfomed all 2-0 and 3-D ROCS calcula-tions and the Cycle 5 PDQ calculations. These results demon-strate 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 Positions Fol10 wino a RE_ctor Trip
'Another' measure of the ability of the 3-0 ROCS model to accur-ately predict reactivity changes is its ability to . predict the critical boron concentration and CEA position following a reac- N tor trip. A study of this: type was done_ for criticalities-dur-i ng the recovery froa -a reactor. trip -for Cycle 2. - ' Thi s study .
showed that the maximum reactivity error between measured crit-
.ical parameters and calculated parameters was [ _] ap. This
-demonstrates the ability of ~ the District's; ROCS model to accur-ately model the power defect and xenon buildup and decay.
22-
5.0 VERIFICATION OF NEUTRONICS MODELS FOR FORT CALHOUN STATION (Continued) 5.6 Comoarison to Independent Radial Power Distribution Calcula tions , Comparisons between the District's ROCS model calculations and ENC XTG model calculations of the HFP radial power distribu-tions have been performed. Figures 5-7 through 5-12 show com-parisons 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 model s.
5.7 The District's Ongoing Benchmarking Program Much of the data reported in this section was drawn from the District's ongoing benchmarking program. This program includes startup physics testing predictions, reactor testing analysis and a core follow effort. 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. l 1 i l 23 _ _ _ ~ . __. _ _
c 1 Table 5-1 Unrodded HZP Critical Boron Concentrations Calculations i 3-D* 2-D* 3-D Measured ROCS ROCS ROCS PDQ Cycle ppm (CEPAK) (CEPAKl (DIT) (CEPAKl XTG 1 933 - 2 1240 - 3 1000 -
- 1 4 1027 - - -
5 1242 - - - 6 1230 - - 7 1241 - - - 8 1240 - -
- A 20 ppm bias has been applied to these calculations.
.24-
Table 5-2 l Summary of Comparisons of Measured and Calculated ,1 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 - 100 3 1000 - 100 3 1428 - 100 3 2510 j
- 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 25
1 Cycle CEPAK DIT Nominal j Burnup Power , Cycle (MWD /MTU) (%) (%) % of Full Pcwer 4 7000 ! - 100 4 8200 - 100 5 300 - 100 5 1000 - 100 5 2000 - 100 5 3000 - 100 5 4000 - 100 5 5000 - 100 5 6000 , 100 i - __ 6 50 66 6 500 - 100 6 1000 - 100 6 2000 90 6 3000 - 65 6 4000 75 6 5000 - 75 6 5800 75 6 6500 100 6 7500 - 50 6 8500 95 6 9500 95 6 10500 95 l
- 7 135 70 I l l 7 500 -
100 l 7 1000 - 100 l l 7 2000 100 7 3000 - 100 26
Cycle CEPAK DIT Nominal Burnup Power I Cycle (MWD /MTU) (%) (%) % of Full Power 4000 100 7
,- 100 i 7 5000 _
6000 . 100 7
- 100 7 7000 i
100 7 8000 100 7 9725 50 45 8 250 - 100 8 1000 100 8 2000 100 8 h 5 27
! Table 5-3 Low Power Physics Isothennal Temperature Coefficients Boron CEPAK* DIT ROCS XTG Concentration Measured ROCS (Ap/ F) Cycle (ppm) (Ap/*F) (Ap/ F) (Ap/*F) 1 9 93 0.26
- 10-4 - -
1 l 1 854 -0.11
- 10-4 - -
l 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 - -
5 1213 0.23
- 10-4 7 1213 0.12
- 10-4 8 1240 0.16
- 10-4 -- .-
' Calculated results were biased by 0.20
- 10-4 Ap/*F Y
Table 5-4 Comparison of Calculated and Measured Isothemal Temperature Coefficient B0C Cal culated* Cal culated Critical Boron Measured CEPAK-ROCS OIT-ROCS Percent of Concentration TC ITC Cycle Rated Power (ppm) (*10{ap/*F) (X104 ap/ F) (X10{TCap/ F) 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 _ _t E0C 1 75(1) 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 - i 6 96(2) 307 -1.51 . 7 95(2) 192 -1.85 I (1) Full Rated Power = 1420 MWt (2) Full Rated Power = 1500 MWt l !
- B0C calculated results were biased by 0.20
- 10-4 ap/oF and E0C calcuiated results were biased by 0.40
- 10-4 ap/*F.
20
Table 5-5 Conparison of Calculated and Measured Power Coefficients CEPAK-ROCS DIT-ROCS Percent Measured Calculated Calculated of Critical Power Power Power Burnup Rated Boron Coeff. Coeff. Coeff. Cycle MWD /MTV Power Conc. (Ap/% Power) (ap/% Power),19/% Powerl 2 10877 46(1) 104 -1.95 x 10-4 3 157 46(1) 720 -1.47 x 10-4 3 1513 90(1) 535 -1.12 x 10-4 3 4183 90(1) 309 -1.31 x 10-4 3 7208 90(1) 62 -1.48 x 10-4 4 267 92(1) 690 -1.04 x 10-4 4 4690 94(1) 288 -1.12 x 10-4 4 8027 95(1) 44 -1.10 x 10-4 5 426 93(1) 876 -1.05 x 10-4 5 6815 94(1) 296 -1.25 x 10-4 6 400 95(1) 848 -1.11 x 10-4 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) 191 -1.57 x 10-4 8 79(2) 817 -1.18 x 10-4 (1) Full Rated Power = 1420 MWt (2) Full Rated Power = 1500 MWt l 30
Table 5-6 Cycle 1 CEA Worths Cal culated Cal culated Cal culated CEPAK-ROCS CEPAK-ROCS CEPAK-PDQ Measured 3-D 2-D 2-0 G roup (%Ao) (%Ao) (%Ao) (%Ao) 4 0.58 > i 3 0.57 ' i 2 2.01 A 3.06 ! B 2.10 i Total (4+3+2+A+B) 8.32 1 Table 5-7 Cycle 2 CEA Worths Calculated Calculated Calculated CEPAK-ROCS CEPAK-ROCS CEPAK-PDQ Measured 3-D 2-0 2-D Group (%Ap) (%Ap) (%Ap) (%Ap) 4 0.65 3 0.41 2 1.67 _ _ l 1 '0.95 - l Total .* (4+3+2+1) 3.68 _ _ l 31
a Table 5-8 Cycle 3 CEA Worths Cal culated Cal culated CEPAK-ROCS CEPAK-ROCS Measured 3-D 2-D Group (%Ap) (%Ao) (%Ap) 4 0.74 3 0.59 2 1.96 , 1 0.80 - Total (4+3+2+1) 4.09 - Table 5-9 Cycle 4 CEA Worths Cal culat ed Cal culated CEPAK-ROCS CEPAK-ROCS Measured 3-D 2-0 Group (%Ap) (%Ao) (%Ap) 4 0.63 l, 3 0.63 2 1.90 __ _ 1 0.92 - Total -, (4+3+2+1) 4.05 - 1 i i 32
Table 5-10 Cycle 5 CEA Worths Cal culated Cal culated Cal culated CEPAK-ROCS CEPAK-ROCS CEPAK-PDQ Measured 3-D 2-D 2-D G roup (%Ao) (%Ao) (%Ao) (%Ao) 4 0.57 3 0.67 2 1.40 .-. - 1 0.99 - - Total (4+3+2+1) 3.63 - - Table 5-11 Cycle 6 CEA Worths Calculated Calculated Calculated Calculated XTG CEPAK-ROCS DIT-ROCS DIT-ROCS Measured 3-D 2-D 3-D Group (%ao) (%Ao) (%Ao) (%Ao) (%Ao)
~~
4 0.52 - 3 0.66 - 2 1.57 - 1 0.93 - Total - -, (4+3+261) 3.68 33
Table 5-12 Cycle 7 CEA Worths Cal culated Cal culated Cal culated Cal culated XTG CEPAK-ROCS DIT-ROCS DIT-ROCS Measured 3-D 2-D 3-D G roup (%Ao) (%Ap) (%Ao) (%Ap) (%ao) 4 0.49 - 3 0.47 - i 2 1.65 - 1 0.70 - Total - - (4+3+2+1) 3.27 _ __ _ - _ _ Table 5-13 , Cycle 8 CEA Worths Cal culated Calculated Calculated Calculated XTG CEPAK-ROCS DIT-ROCS DIT-ROCS Measured 3-D 2-D 3-0 Group (%ap) (%Ap) (Do) (%Ap) (%Ap) ( - 3 0.63 2 0.99 1 1.00 Total (4+3 +2 +1 ) 3.20 ... 34
CYCLE 4 CRITICAL BORON CONCENTRATION vs BURNUP FIGlE 5-1 , BORON CONCENTRATIM (ppa) 900 850 800 750 700 650
. 600 <
550 500 450 1 400 i I 300 : 250 200 i 150 100 50 0 0 1000 2000 -3000 4000 5000 6000 7000 8000 9000 1
d 4 CYCLE 5 CRITICAL BORON CONCENTRATION vs BURNUP FIGURE 5-2
,_ BORON CONCENTRATION (ppa) 950 -
900 850 800 750 I 700 650 600 a
! cn 550 500 q
- 450 400 l 350 300
! 250 200 , 150 100 2 0 0 1000 2000 3000 4000 5000 6000 70G 8000 9000 10000 11000 BURNUP WWD/T)
CYCLE 6 CRITICAL BORON CONCENTRATION vs BURNUP FI6t E 5-3 BORON CONCONTI(N (ppa) g _ 900 . i BE0 800 750 700 650 , 600 t tj 550 500 450 4 400 .
- 350
. 300 1 1 3 5.JO s 150
, 100 1 50 0 -
., 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 j Bl H I P S W O/I)
CYCLE 7 CRITICAL BORON CONCENTRATION vs BURNUP FI6 LEE 5-4 BORONCONCENTRATION(ppal 950 900 85G 800 750 700 !
- I 600 f
M 550 ! 500 2 I 400 350 300 250 200 150
- 100 50 0
1 4 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 i N M/I} d
CYCLE 8 CRITICAL BORON CONCENTRATION vs BURNUP FIGl E 5-5 _BOAM CGCENTRATIGi (ppm) 900 850 800 750 700 650 600 g 550 500 450 400 350 i 300 l 250 200 150 100 50 0 0 1000. 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 RINF $00/T)
BENCHMARK OF QUIX ASI VS. CECOR ASI Figure 5-6 0.20 - - - - - 0.15 i 0.10 i E 0.05 i
$ $5 E
I 2~ 0.00 [
- -0.05
-o.10 j -0.15 1
1
-0.20 -
) Tite thours) i .i
1 l l J i Asser.bly.;PDCroaris:n Cc.ana Public F.nier District Ficure rc ;o c:a Cao i . iarJ n.. cn,ej rn . .a .:. v i v e. e.61 ccf iu >. L : o-:
- s wel,. o t:: t ; 00 e.n..:
. . :.o 00. .i 5-7 .
41
i 1. l h M e unno.nuwea-ana ,ygggpygge;;;,7cy ,,; ,, d?
i t i l S 4 AssemblyEPOC0:parison OmahaPublicPowerDistrict Figure 800 7 HFP AR0 and- 900 opa E0ron FortCalhounStation-Unitflo.! 5-9 l t 43
l t Assembly BPD Comparison EOC 7 0:ahaPublicPowerDistrict Figure 10,000WD/MTUfiFPAB0-ispaBarca fortCalhounStation-UnitNo.i 5-10 44
_ a f I I i
~ - , . - - . __
m l l l Asse:blyRP0 Comparison F 11c Power District figure 8008hFPARDand-770coaBaron FartCalhounStation-UnitNo.1 5-11
1 l l I i Assecoly RPO C0:Darisan E0C 8 - 9000ED/NTUHFPAs0-i003 Baron FortCalhoaStaton-UnitNo.1 f2 46
l . l l i 4 P er 0istrict Figure Assemb1yRPDCo BBC6HFPARDand Boren0pfm# FortCalounStaYon4ni 47~
1 i I W t i
- l l
. I
~
~
/ - J
.s
't 5 #
./+
l AssemblyRPDComparisonE0C6 OmahaPublicPower' District
, Figure'.
I 10 000 MWD /NTU HFP AR0 20 ppm Baron Feht Calhoun Station-Unit No 1 5-14 l
- \r, ;,
e aa
.- ., . - - .. a _ ._ . . _ . 21.. .:
,r t
.. \,, , .
~ <
, ~
i 4 1 l , I 1 l x .
- m. ,
~
l ' 1 % A 4 9 b 1 r A l .,
'4 s- ,
AssemblyRPDComparison OmahaPublic'PowerDistrict Figure 80C 7 HFP ARD and 900 ppm Baron FortCalhounStation-UnitNo.i 5-15
.t- <
. .~ _
49 i.
l 1 ( 4 7 f f Assembl OmahaPublicPowerDistrict Figure 10000 MWD MTU HFP /yRPDComparisonE0C7 AR0 i ppm Baron FortCalhounStation-UnitNo.i 5-16 50
. ._ . . _ __ . _ _ _ . _- __ . _ . . _ - . . _ . _ _ - ~ _ _ . _ _ . _ _ _ _ .
4 i ! l i a ~ l 1 i AssemblyRPOComparison OmahaPublicPowerDistrict Figure 80C 8 HFP ARO and 770 ppm Baron FortCalhounStation-UnitNo.i 5117 51
_-; .m.p an -.. - m a -.e-I r i i Assemb RPDComparisonEDC8
* *
- ru +e a i 99 8eron rJ"a"jllliggge;jj,stejgt i , ejgge
- . - . - eu _ - - -.- .
6.0 REFERENCES
Section 2.0 References 2-1 ENDF-313, " Benchmark Testing of ENDF/B Data for Thennal Reactors, 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 Conputer 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 Manual," WAPD-TM-670, 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, Revision 1-P, "CE Setpoint Methodology," 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 Verification 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. 53
6.0 REFERENCES
Section 4.0 References (Continued) 4-5 CENPD-226-P, "The ROCS and DIT Conputer Codes for Nuclear Design," Decenber,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. ( 4 i 4
-g n
! i l 3 54 I i
{ APPENDIX A l CYCLE 8 RADIAL POWER DISTRIBUTION COMPARISONS T
I l l CEPAK OMAHA CY8 45 PERCENT 5-( INST ONLY ROCS-CECOR COMPARISON LEVEL 1 Bot 8 m
&~
i l 1 i I I 4 4 i I l - i t t i i i 1 i I h esp " l I l L' _ y- ,-e ' g-- yy-g-t 9 T t+--T-175 W #"" 9D TF ' "i 'Y"'F T # '
5a- as n. - .,e a- m., --m , a-. m -., I CEPAK S- COR C IPARISON l J' I i I-I s
) .i
- I l
""* ' ~
r~ p-- ,, _
CEPAK
> GMAHA CY8 45 PERCENT e ( INST ONLY ROCS-CECOR COMPARISON LEVEL 3 80C 8 .
t 4 4 I + . i i e >
\
i I m
CEPAK , OMAHA CY8 45 PERCENT ( INST ONLY ROCS-CECOR COMPRRISOM LEVEL 4 BOC. 8 1 I l
\
'gu e w , y -4 yr& = g- r-- m,, -, -y &>- * =-e- -
+.-,,4 -ww e n, m3. -w e- :e= w~ w ,
CEPAK OMRHR CY8 45 PERCENT { INST ONLY ROCS-CECOR COMPARISON RXIRLLY INTEC-RATED Boc 8 4 h N
(. CEPAK
'i l
t
- OMAHA CY8 99 PERCENT INST ONLY ROCS CECOR COMPARISO N LEVEL 1. 1000 Mwo/r e
I i i 4 1 4 i i s. l s t Y
- , . , + . , s y,--- .- m.,,.. . . , . , - . . , _ . , _ , - -, . - - . , . r. .n , , - . ,- e, ,
( CEPAK 0MRHR CY8 99 PERCENT INST. ONLY ROCS-CECOR COMPARISO N LEVEL 2 1000 Mwon 1
,f _
't 4 f i . I i i
.i l
. . __ _ , , . . -_- . = . . , , . , . , _ . , _ . . . _ . _ _ . , . - , _ . . _ . , _ _
4 CEPAK t
. OMAHA CY8 45 PERCENT
-{- -INST ONLY ROCS-CECOR COMPARISON LEVEL 3 80C 8 .
i N f e I i i 1
\
l [ .'
, .... . _ . . . . , , . . , , . . . . .. -, , - _ . - - . - _ - - , - - , m, , , , - . , , - ~ . - , -
( CEPAK OMRHR CY8 99 PERCENT INST ONLY ROCS-CECOR COMPAR. ISO N LEVEL 4 1000 Mworr de l 5
\
\
g
[ CEPAK 'c OMAHR CY8 99 PER. CENT - INST ONLY ROCS-CECOR COMPARISO N AXIALLY INTEGRATED loco Mwo/T , e M )' ll J l - I i i-e I i O d I a d
- .w . , - , 4y, -
~'Tew- r '-Y rF-= r c-v' r --e- p w=rT'Tw +-v---V rw e vr" W 't-d'- *** -
- 9'~ wr" # y- Y e'* " - *
( CEPAK
! OMRHR CY8 99 PERCENT INST ONLY ROCS-CECOR COMPRRISO N LEVEL i 2.000 Mwo/T 1
4 i i- ! i
+
i ( i .i i e I k 1.
\
i E D. l
'g .
t n
( CEPAK OMAHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPRRISO N LEVEL 2 2000 gwoj7 l k a dumm -- - l k . I t [' l
^ ' ' ' , . ~. , .. .-----,,y,,- - . , , -,n.,.v-.--. , ,-, ,. , y
{ CEPAK OMAHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPARISO N LEVEL 3 Ecco nwo/T (
\
t
CEPAK OMRHR CY8 99 PE.RCENT INST ONLY ROCS-CECOR COMP AR I SO N ' N LEVEL 4. 1000 mwo/T
.j i
9
\
I . o
i 4 ( CEPAK OMRHR CY8 99 PERCENT _ INST ONLY ROCS-CECOR COMPRRISO N RXIRLLY INTEGRATED z.000 stvo/r G A l s' 1 I 4 4 ; i i i
'S l
r
t t. DIT > i t , f 0MAHA CY8 45 PERCENT INST ONLY ROCS-CECOR COMPRRISOM LEVEL 1 45 mic o ) r k
! i:
i-
-. 0 J
h
}
l l 5 I i 4 e 1 ll. [ i
)
i 9
. . , . ,, , - , , - - . . - - - . ,+,,,,,.,,.,,,,,4,-,,-. ,,..,,n,,,, , , , , - --l-,- , , , -
s s. sm.. u -.. m - ,- S 4 e > A - - ~ i DIT N9'HA CY8 45 PERCENT INS 7 N f LEVEL- h0CS-CECOR COMbARISON I 45 mwo/r 1 t e
/
/
4 i $ l
- (. , ,
- t t
\..-
I l i l I i P
r l DIT l l OMRHR CY8 45 PERCENT INST ONLY ROCS-CECOR COMPRRISOM LEVEL 3 45 mwn/r f h
\
\
\
\
\
I
. i l
DIT OMAHA CY8 45 PERCENT INST ONLf ROCS-CECOR COMPRRISOV LEVEL 4 45 mtoolr 2
~!
N N t 8
DIT OMAHA CY8' 45 PERCENT i !NST ONLY ROCS-CECOR COMPARISON AXIRLLY INTEGRATED 45 mtuo/r i 4 l l .i c i 4 '. /
- .1 e
i i i h s
\,
9
, . , , -_.,.....m.m -, , . . . - - . . ,y . , p u r.. , , ., . , _.,,,___-e,, n.+ -- , ,,r_ __ m,. . - - --- - -
.y-
r , ; . -s.i., t !
' l
, ':-- , wa
~
+
1 r. DIT OMRHR CY8 79 PERCENT INST ONLY ROCS-CECOR 'COMPRRISOM LEVEL 1 Mo mW Dif ( i
/
r _ ______m - - _ . . . _ .
s e f ? i DIT , OMAHA CY8 'I9 PERCCNT INST ONLY- ROCS-CECOR COMPARISOM LEVEL 2 &,30 mwo)r, t $ t 1 f 1, ;
\
k e
,~
b A # g g*
, r r 5 i
) .
i i E o
.Y I '
i , I f l 9 9 e
+
s w I w 4
r, ,
f s
\,. ,.
n - ;- . i . .. 4.' j 4.# ,I' f n f f ek , l o- _4 r ,
- y. -
.~
~ .a . ,
' l,
;/- w
.. e
_______.._m___________1'. -.
- 8
~
DIT OMAHA'CY8 79 PERCENT INST ONLY ROCS-CECOR COMPARISON LEVEL 3 B.30 mto of r . 1 h r
, --c ,- = -, em e ,-- 3 e = w- # y,- + - , + +~+,=,----,y ,-,,,as p g---- - - , -, -- -e -
DIT l OMAHA CY8 79 PERCENT INST ONLY ROCS-CECOR COMPC.RISCM LEVEL 4 Aao mwolr l
\,
c DIT OMAHA CY8 79 PERCENT INSr.ONLY ROCS-CECOR COMPARISON RXIRLLY I N T EC-R A T E D M o moo /T i i
DIT " OM.HA 4 CY8- 99 PERCENT INST ONLY ROCS-CECOR COM"Pq'RISOV LEVEL 1 1000 meo/T } l l ]; 1 M f J 9
\
I i i l 1
,, .. . . - , . - - , , - _ . , . . . - - _ . . , _ . . . . . _ - - . , . . . . . - _ , , , , ,, , , , , - - .-4. .. . y
r DIT GMAHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPARISON LEVEL 2 1000 M W olT- ', < S r i l _ , .., , , , , _ _ _ _ - . - , _ . . . _ _ _ _ , _ , . _ . , _ , , - - . . . _ = . _ . . . - . . _ . _ . . . . . . . _ - , . _ , , . . , _ . . _ . ~ . , - . . . .
DIT OMRHR CY8 99 PERCENT INST ONLY ROCS-CECCR COMPRRISOU LEVEL 3 tereo moolv f i i f1 4 4 i I J i 4 1 1 1 J I
-\ .
I i s k 4
-v,+ . w2,r-, - - , , - .-a---e-+- e4 -re v w -+we ,,-v- r e-W-,+rv+<* -wm-w-- r t w ~+ += _
DIT l OMAHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPRRISOM l.EVEL 4 tooo mtooJT l . 1 4 1 1 a l
<r< - v, ,, , n ,w., ~--,v- -v-w n .n-w, e--v-~ -aw rr e-- --w- ~.+<r,- , - - rv. ~- - - - -w gr -w-~ r - ,n-- - - - - - - - -
DfT OMRHR CY8 99 PERCENT INST ONLY ROCS-CECOR COMPRRISON AXIRLLY INTEGRRTED teco (hwoJT f a e i f A I i l i 1 4 S y m - h -- -- r w -e -r-we -
-g e m-e., .-w, w- w~ -,m----+- % ,---,, mew-rs* e ,
- w ev - m . , -- m wey-+,
i DIT OMRHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPARISOM LEVEL- 1 aow mwo/r 1 9 l 4 1 I l l P J l 4-L f 4
-k
\
s . \
.\
i-i i d
- ,-aw - yw, e g., - - , . -w,- <p- =- e w er . - ,+'e-'=reee--ww -g-=----' +f. * - g w'e t + e -
Y-g- - --=eet+-r-ne v4e --
DIT
- GMAHR CY8 99 PERCENT INST ONLY -ROCS-CECOR COMPARISDN LEVEL 2 2o00 mu>olT I
l. i }. i l i l-4
\
I s 1 4
\
\
l i
- DIT r
- GMAHA CY8 99 PERCENT l INST ONLY ROCS-CECOR COMPARISON LEVEL 3 8000 mwolv 1
I m 0 _se" h i ,r 1 J 1 5
\
I
\ \
i i i
4 l DIT OMAHA CY8 99 PERCENT ' INST ONLY ROCS-CECOR COMP 9RISON LEVEL 4 9,doo mue lT t 4 ( 1 1 1 i a 1 i 1 l i
>\
s , r
DIT ! OMAHA CY8 99 PERCENT INST ONLY ROCS-CECOR COMPARISON AXIALLY I N T EGR A T E D acoO tmoofr l i : j l l
, . - ~ ~
i d i i
\
i I s 4
\
1 7 t'
f APPENDIX B AXIAL POWER DISTRIBUTION COMPARIS0NS i l t
4 CECORC D ) AND ROCSC 1 NORMAL.TZED AXIAL POWER PLOT FT. CAL. + 0 MWD /MTU CORE AVG LECOR AXIAL PEAK / AVERAGE 1.1972 LOCATION 70 PCT FROM BOTTOM ASI= .0239 ROCS AXIAL PEAX/ AVERAGE 1.2141 LOCATION 54 PCT FROM BOTTOH ASI= .0577 PLOT NO 1 FILE A256435 g,4 _ _ PLOT NO EFILE C5ASACBC00 .. 1.0 NORMALIZEO
. POWER
.8
.6
.4
~
j {*" 5 5 3 5 'T B 5 5 O 10 20 30 40 50 80 70 80 SO 100 PERCENT HEIGHT FROM BOTTOM OF ACTIVE CORE Cycle 5 . I
CECOR( a ) AND ROCSC ) NORMALIZED AXIAL POWER PLOT FT. CAL. 3K MWO/T CORE AVG ~ BOTTOM ASI= .0034 CECCR AXIAL PEAX/ AVERAGE 1.1121 LOCATION 28 PCT FROM ROCS AXIAL PEAK / AVERAGE 1.0956 LOCATION 68 PCT FROM BOTTOM ASI= .0594 ,l PLOT NO 1 FILE A266334
- PLOT NO EFILE C50EPSTS03 __
1.4 I 'I
~
1.2 1.0 NORMALIZED P0tSER
.8
.6
.4
.2 i i i i i i j I
10 20 30 40 50 60 70 80 SO 150 0 PERCENT HEIGHT FROM BOTTOM OF ACTIVE CORE Cycle 5 - N
CECOR( o ) AND ROCSC ) NORMALIZED AXIAL POWER PLL FT. CAL. 3K MWD /T ASSEMBLY 78 - CECOR AXIAL PEAK / AVERAGE 1.1205 LOCATION 74 PCT FROM BOTTOM ASI= .0166 ROCS OXIAL PEAK /AVERACE 1.1029 LOCATION 77 PCT FROM BOTTOM ASI= .0341 PLOT NO 1 FILE A266334 1.4
- PLOT NO EFILE C50EPSTS03 __
- . ___ l 1.2 1.0 NORMALIZE 0 POWER i
.8
, .6
.4 l
.2 I
I - I ; i g I I l I . 30 40 50 60 70 60 90 100 i 0 10 20 PERCENT HEIGHT FROM BOTTOM OF ACTIVE CORE Cycle 5
CECOR(a) AND ROCSC ] NORMALIZED AXIAL POWER PLOT FT. CAL. 3K MWD /T ASSEMBLY 98 CECOR AXIAL PEAK / AVERAGE 1.1325 LOCATION 74 PCT FROM BOTTOM ASI= .0174 ROCS AXIAL PEAK / AVERAGE 1.1223 LOCATION 33 PCT FROM BOTTOM ASI= .0794 FLOT NO 1 FILE A266334 1,4 __ PLOT NO 2 FILE CSDEPSTS03 __ 1.2 1 1.0 10RMALI2E0 POWER
.8 ,
.6 4
.2 -
! ! I I l l [ g i O 10 20 30 40 50 60 70 80 90 100 PERCENT HEIGHT FROM BOTTOM OF ACTIVE CORE Cycle 5
CECORCe ) AND ROCSC ) N0ftMALIZED AXIAL POWER PLOT FT. CAL. 3K MWD /T ASSEMBLY 108 - CECOR AXIAL PEAK / AVERAGE 1.1299 LOCATION 28 PCT FROM BOTTOM ASI= .0125 ROCS AXIAL PEAK / AVERAGE 1.0935 LOCATION 64 PCT FROM BOTTOM RSI= .0588 PLOT NO 1 FILE A266334 1.4 - PLOT NO 2 FILE C50EPSTS03 _-. l 1.a 1.0 CRMALIZED POMEP
.8
.6 4
- .a -
i i
! I I I I I I l 0 10 20 30 40 50 60 70 80 90 100 PERCENT HEIGHT FROM BOTTOM OF RCTIVE CORE Cycle 5
CECOR(4) AND ROCS ( ) . NORMALIZED RXIRL POWER PLOT FT. CAL.: SOC 6 66 PERCENT CORE AVG CECOR RXIRL PERK /RVERROE 1.1932 LOCATION S6 PCT FROM B0TTON RSI= .00S93 ROCS RXIAL PERK /RVERAGE 1.2022LOCRTION 66 PCT FRON BOTTOM RSl= .01929 PLOT NO IFILE R308936 14 _ _ PLOT NO 2FIL E C6PWRRSN06 _-
~ - --
1.2 1.0 NORMALIZED POWER 4 8 6 i j 4 1
.2 g i I l I I I I i 0 10 20 30 40 50 60 70 80 90 100 PERCENT HEIGHT FROM BOTTON OF RCTIVE CORE Cycle 6
.h
CECOR(*) ANC ROCS ( ) N"RMALIZED RXIRL PCNER PLOT . FT.CRL.: 5000 MWD /T C C R E A V C-
.C;CC
( ? '. Ci: 0% in;_ PF 0;./O'/f 005: ; .CS30LCC O T ICtl 2SPLi . C C.'i 001 Tf:3 601:. P O. .X .i lli. " ? .'K.in'ii. P E 0! ! .0 9 5 2; er 0 : 1 Cil 26 PLT : T.c3 80il r:;l 001:. 0720 . P. . CC
- i. t i40 5
.r .t : .- : . . . '.3, ., ;. .;
l
. <. .r P i C1 rin 2i :i. .: C(ESPCPGC2 ,-
t l .__
. . t.
. s t. iu. . ...., s.. ..n.. .-
P C..:~ R 0 G i
. 4
. t 6
. i.
l
; 6 i i i . . . .
C .0 20 30 30 50 6C 70 S t. u9 100
~
PERCE*;T HE;CHT FRCM DDTTf;M Cf ACTI'li' CDP.E
. Cycle 6
CECOR(*) AND ROCS ( ) NORMALIZED RXIAL POWER PLOT FT. CAL..: 9500 MWD /T ChtLE. C. CORE AVG CECOR RXIAL PEAX/RVERAGE 1 1177 LOCATION 24 PCT FROM BOTTOM RSI: 0405 ROCS RXIRL PEAK /RVERROE 1 1313 LOCATION 15 PCT FROM BOTTCM ASI: 1128 PLOT NO IFILE A349903 3,4 _ PLOT HC 2 FILE CY6CFDP812
- T 1
1.2 - 1.0 NORMALIZED PGWER 8
.6
.4 2
3 I I I I I I I I O 10 20 30 40 50 60 70 80 90 100 PERCENT HEIGHT FROM BOTTOM OF RCTIVE CORE Cycle 6
. . . .._v--- -- ~*-
CECOR.( & ). AND- ROCS C ). NORMALIZED AXIAL POWER. P E.0 T E T . C AL ..:- 1.3 5 MW D / T CORE AVG CECOR.RXIRL PERK /RVERAGE.l.1709LOCRTION 56 PCT FROM 80iTON ASl= 01920 - ROCS RXIRL PERK /RVERROE 1 1943 LOCATION 66 PCT FROM BOTTOM RSl= 02295 ' PLOT N0 IFILE.R364966001 : g,4 __ PLOT NO 2EILE.CY7135PT02 __
~
l . 3 1.2
- d
.I I
10 NORMALIZED
~
1 POWER .l 8 1
.G 4
2 i i i i i i i i i 0- 10' 20- 30- 40~ 50 - 60 70 - 80 90 100 4 PERCfNT HEIGHT FROM [,TTON OF RCTIVE CORE @) l Cycle 7
9 CECOR.(&) AND ROCS (- ) NORMALI' ZED RXIAL POWER _ PLOT e - ET. CAL.: 500 MWD /T' CORE AVG CECOR. AXIAL PEAK / AVERAGE 1 1409 LOCATION 32 PCT ~FROM BOTTON ASI= .02053 !' ROCS AXIAL PEAK /RVERAGE 1 1634LOCRT10N 59 PCT'FR0H BOTTON RSI= .06(14 i' .- PLOT N0 IFILE A366076001 , g ,, 4 __ PLOT NO 2F.lLE CY7CFDPB04 -- 4 . o
- l. 2 j
. s.
i Y .
- l. 0
- NORMALI' ZED POWER i
8 , o ; -
.4 .,
'9 :
21 . 1
.)
1 I I I I I I I i . 0- 10' 20 30 40 - 50- 60 70 ' 80 - 90- 100 ' 4 '
~
'bh' PERCENT HEIGHT FROM Nthf TON OF ACTIVE CORE h -
Cycle 7 m _ _ _ _ _ _ _ . _ _ _ _ _ _
CECOR(&) AND ROCSC ) NORMALIZED AXIAL POWER PLOT FT. CAL.: 4000 MWD /T CORE AVG CECOR RXIRL PEAK /RVERROE 1 1063 LOCATION 2GPCT FROM BOTTOM ASI= 02550 ROCS RXIRL PERK / AVERAGE 1 0844 LOCATION 27 PCT FROM BOTTOM ASl= 09661 PLOT NO IFILE A378901001 g,4 __ PLOT NO 2 FILE C7ASBOEPO4 -- l.2 10 NORMALIZED POWER 8 6
.4 1
2 1 I I I I I I I I i 0 10 20 .30 40 50 00 70 80 90 100 PERCENT HEIGHT FROM 80TTOM OF RCTIVE CORE Cycle 7
. CECOR(& ) AND ROCSC 1 NORMALIZED RXIRL POWER PLOT FT.CRL : 8000 MWD /T -
CORE AVG
.00146 CECOR RX!AL PE96/RVER9CE 1 0761 LOCATION 74 PCT FROM SOTTOM RSI:
S0TICH RSI: 10118 ROCS 9X I9L PE9i'.'RVERAGE 1.08b5LCCA T ICN 19Pr! FROM PLOT NO IF!tE R392600001 g,4 _fLOT NO 2 FILE C7RSSSEP08 ... 12 10
, NORMRLIZED OWER
.e l l
.6
, .4 2/
- r. -
t
? 0 ,1d 20 30 40 .SC 60 .7C 8C 90 100
.' PERCENT HEIGHT FR0!' BOTTOM OF RCTIVE CORE Cycle 7
_-_ .~ . -- . . . - -.- - ._ - - . .- - - . .
]
l -
- CECOR(&) AND ROCSC ) NORMALIZED AXIAL POWER PLOT FT CAL.
- .45'MHD/T DIT CORE AVG
". CECOR'AXIRL PEAK /RVERAGE 1 182SLOCRTION G2 PCT FROF BOTTOM RSIr .01958 ' ROCS RX!Rt. PEAK /RVERRGE 1.2042LOCRTION 69 PCT FROM BOTTOM RSI: 0087G Pt GT N0 LF!lE R411839001 2 1 . ....
. f t 07- NO 2 FILE-CY80KBEN00
/ __
g '1. 2 . 1 G-09fiRLi2ED PCHER
- 6 t
I ,' t : .. G
.q --
F i l
! 4 .
1 4 1 . 2 - I 4 I i : r i . 4 G iG 20 .30 40 50 GG 70 80 90 100
- PERCENT HEIG!iT FROM BOTTOM OF'RCr!VE CORE .
p ,,. Cycle 8
=
x C$COR(& ) AND ROCSC ) NORMAL'IZED A'XIAL POWER PLOT . n FT. CAL. 15 MWD /T DIT 'RSSEMBLY 78-
.CECOR R.(IRL.' PEAK /RVERRGE 1.iG93LCCRTION OGPCT FROM BOTTOM RS! .03841 ,
~~- R O C S RXIRL PERK /RVERRGE 1 2040t.0CRTION 71 PCT FROM ~
80"0M RS!:: 00456 PLO' i10 IFilE R411839001 .;-
; , ,9 ' A Pt.CT NO 2F!t.E CY80KBEN00 .
u,- e
'N<.
.,s
<7
~ \
-. i . 2 , .
'. s i;0- ,
.N091'Al.!ZE0 T .,
~P0hER
.. s - ,
' ~
. 8
., .g 4
; .y .
1 ,
.G , 'l ~ ^ ' >
.A
.e .
~
.1 ,
- m. _
2 -. > s g ,,
+, ,
i i i i i ,, , i i . C 10 20 30 40- 50 GO 70 80 9b 100, PERCENr HE!051T FROM BOTTOM OF RCT!VE CORE Cycle 8 _ _ _ . .= - - - - .
CECOR(&) AND ROCSC ) NORMALIZED AXIAL POWER PLOT t-l FT CAL : 45 MHD/T CECOR AXIAL PERd/RVERAGE 1.1858t.CCATION DIT ASSEMBLY 98 62 PCT FROM BOTTOM ASI: .01085 - l ROCS RXIRt.. PEAK /RVERRGE 1. 21221.0C R T I C N G5 PCT FROM BOTT0t1 AS!= 03096 i Pl.0T NG IFIt.E R411839001
; , ,g .. P!.C T NC 2FIL.E CY80KBEN00 . .. '. % g i
1.G l- NORM A!. I ZEG l ecNER I +O
.G l
i r 4
+2 I ? I ' '
I I ' 1 0 10 20 30 40 50 GO 70 80 90 100 PERCEN! HEICiti FRCt180rr0N OF ACTIVE CORE Cycle 8
- _ - - _.- - _ . = . - _ - - _ _ - - _ _ -__-
_ CECOR(4) AND ROCS ( ) NORMALIZED AXIAL POWER. PLOT ET. CAL : ICCCMWD/ T D IT CORE RVG PECOR RXIAL PEAK /AVERADE 1.ilS7LOCAT10N 30 PCT FROM BOTTOM-ASI: '02560
.RCCS A'(IAL PEAK /AVERROE i.1055lGCATIGN 7iPCT FROM BCITOM ASir 03000 PLOT NC IF!LE'A415800001 '
-;;g -
_.PLCi NO 2F!LE CY81XBEN00 ..
~
i 12 I 1.0
- j. NCRMAl.!ZE0
-8'0HER
. t!
i } , .G 4 4
'2 i
! + i : : : i i i 0 10 20 30 40 50 00 70 80 90 100 PERCENT HE!GHT FRON BCTICM CF ACT!VE CORE i '
j Cycle 8 I
+
. CECOR(&) AND ROCSF ) NORMALIZED AXIAL POWER PLOT FT. CAL.: Icco MWD / T D I T ASSEMBLY 78 CECCR RXIRL PEAK /RVERAGE 1.0962LOCRTION 30 PCT FROM BOTTOM RSI: .01602 ROCS RXIRL PEAK /RVERAGE.1 1643LOCRT10N 73 PCT FROM BOTICH RS! .02858 PLOT NO 1 FILE R415800001 -
;,4 _ _ PLOT NC 2 FILE CYO1KBEN00 __
, i '2 i.0 NORMRL! ZED
-POWER 8
.0
.4 2
i 1 I I I I I I I I ! 0 10 20' 30 40 50 GO 70 80 90 ~00
, PERCENT HEIGHT FRCH BCTTOM CF RCT!VE CORE Cycle 8
. - - , . _ _ - ___ o
C E C O R.( & ) AND ROCS (~ ) NOR.MALIZED AXIAL POWER PLOT FT. CAL : Icoo MWD / T D IT ASSEMBLY 98 CECOR AX!RL PEAK /RVERRGE 1.lG15LOCRTION 30 PCT FROM 80TTOM ASI: .03905 ROCS RXIRL PERK /9VERAGE 1.1729LOCAT!0N G7 PCT FROM BOTTON RSI: 0502S PLOT NO IF!l.E R415800001 '
; , .9
.ALOT NO 2F!LE CY81KBEN00 .._
~
1.2 1.0 NORMRt.IZE0 r>utu R
.8
.G
.1 -
l 2 I 1 ,I ! ! ! I I I l 0 10 20 30 40 50 60 70 80 90 100 PERCENT HEIGHT FROM BOTTOM OF RCTIVE CORE Cycle 8 w- ,
l CECOR(&) AND ROCS ( ) NORMALIZED AXIRL POWER PLOT l~ FT. CAL.: 2000 MHD/T CORE AVG l' CECCR RXIRL PERK /RVERAGE 1 1055LOCRIIGN 30 PCT FRCl1 BCIICri RSI: 0034i ROCS RXIRL PEAK /RVERRGE 1.1389LOCRTION 73 PCT FRCt1 BCT TCt1 RSl = 01618 PLCT 110 IFILE R419000001
;,3 .__PLCT NO 0 FILE C(82KBENGC --.
. ~ _ _ _ . _..__ _ _ __,
i .2 l i
=0 NCRl!R1,! ZED PChER 8'
l
.0
.4 2
l I 1 I : I I I I 0 10 20 30 40 50 60 70 80 90 '00 PERCENT HEIG!!T FRCM BCITCM OF RCTIVE CCRE Cycle 8
.}}