ML20090G673
| ML20090G673 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 05000000 |
| Issue date: | 04/30/1984 |
| From: | Leanne Flores, Jefferies W, Randles J DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20090G665 | List: |
| References | |
| DPC-NF-2010, NUDOCS 8407250158 | |
| Download: ML20090G673 (238) | |
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{{#Wiki_filter:_- l , - - - - - - - - - - - a L DUKE POWER COMPANY McGUIRE NUCLEAR STATION CATAWBA NUCLEAR STATION NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN DPC-NF-2010 April,1984 Nuclear Production Department Nuclear Engineering _ - - - - c.--m - -mm--,-r pa7EMoeng
N. l MCGUIRE NUCLEAR STATION CATAWBA NUCLEAR STATION j NUCLEAR PHYSICS METHODOLOGY F0". RELOAD DESIGN DPC-NF-2010 APRIL, 1984 L. H. FLORES W. G. JE!TERIES J. H. RANDLES D. E. B0RTZ R. H. CLARK DUKE POWER COMPANY NUCLEAR PRODUCTION DEPARTMENT NUCLEAR ENGLNEERING NUCLEAR DESIGN = . - . . . . . . . . . . _ .
STATEMENT OF DISCLAIMER This report was prepared by Duke Power Company (" Duke Power") for filing with the Unit ad States Nuclear Regulatory Commission ("USNRC") for the sole purpose of obtaining approval of Duke Power's PWR nuclear design methods at McGuire and Catawba. Duke Power makes no warranty or representation and assumes no obligation, responsibility, or liability with respect t> the contents of this report or its accuracy or completeness. Any use of or reliance on the report or the information contained in this report is at the sole risk of the party using or relying on it. 1
ABSTRACT This Technical Report describes Duke Power Company's Nuclear Design Methodology for the McGuire and Catawba Nuclear Station. The nuclear design process consists of mechanical properties used as nuclear design input, the nuclear code systi'm and methodology Duke Power intends to use to perform design calculations and to provide operational support, and the development of statistical reliability factors. 11
l TABLE OF CONTENTS
- 1. Introduction Page 1.1 Introduction 11 1.2 Definition of Terms 1-3
- 2. Fuel Description 2.1 Fuel Pellet 2-1 2.2 Fuel Rod 2-1 2.3 Fuel Assembly 2-2 2.4 Core Component Data 2-4 '-
- 3. Nuclear Code System 3.1 Introduction 3-1 3.2 Sources of Input Data 3-1 3.3 Cross Section Preparation 3-2 3.3.1 Fuel Calculations 3-2 3.3.2 Non-Fuel Calculations 3-3 3.4 PDQ97 Models 3-3 3.4.1 Colorset PDQ 7 Modeling 3-4 3.4.2 Quarter Core PDQ$7 Model 3-5 3.5 EPRI-N0DE-P Model 3-6
- 4. Fuel Cycle Design 4.1 Preliminary Fuel Cycle Design - Initialization 4-1 4.1.1 Review of Des.ign Basis Information 4-1 4.1.2 Determination of Cycle - Specific 4-1 Operating Requirements 4.1.3 Preliminary Loading Pattern and 4-1 -
Reload Region Determination iii
TABLE OF CONTENTS (Contd.) P.iULe 4.2 Final Fuel Cycle Design 4-2 4.2.1 Fuel Shuffle Optimization and Cycle 4-3 Depletion 4.2.2 Rod Worth Calculations 4-4 4.2.3 Fuel Burnup Calculations 4-7 4.2.4 Reactivity Coefficients and Deficits 4-8 4.2.5 Assessment of FFCD 4-14
- 5. Nodal Analysis Methodology 5-1 5.1 Purpose and Introduction 5-1 5.2 Fuel Cycle Depletion - Nodal Code 5-1 5.3 Rod Worth Analysis 5-2 5.3.1 Differential Rod Worth Analysis 5-2 5.3.2 Integral Rod Worth Analysis 5-3 5.4 Shutdown Margin Analysis 5-3 5.4.1 Shutdown Margin 5-3 5.4.2 Shutdown Boron Concentration 5-4 5.5 Rod Insertion Limit Assessment 5-5 5.5.1 Rod Insertion Limit-Criteria 5-5 5.5.2 Rod Insertion Limit - Nodal Analysis 5-5 5.6 Trip Reactivity Analysis 5-6 5.6.1 Minimum Trip Reactivity 5-7 5.6.2 Trip Reactivity Shape 5-7 5.7 Assessment of Nodal Analyses 5-7 iv
TABLE OF CONTENTS (Contd.)
- 6. Calculation of Safety Related Physics Parameters Pm 6.1 Reload Specific Safety Related Physics Parameters 6-1 6.2 Calculational Methodology 6-1 6.2.1 Power Distributions 6-1 6.2.2 Control Rod Worth 6-3 6.2.3 Kinetics 6-5 6.3 Comparison of Cycle Specific Safety Related Physics Parameters 6-6
- 7. 3-D Power Peaking Analysis 7.1 Power Peaking Criteria 7-1 7.2 CAOC Power Peaking Control 7-1 7.3 Power Peaking and Verification 7-2 7.4 Advanced Maneuverability 7-3 7.4.1 Analysis Procedure 7-4
- 8. Radial Local Analysis 8-1 S.1 Background 8-1 8.2 Comparison of PDQ97 to CASMO-2 at 8-1 Hot Full Power Condition 8.3 Comparisons of PDQC7 to Cold Criticals 8-2 8.4 Conclusion 8-2
- 9. Development of Core Physics Parameters 9.1 Startup Test Predictions 91 9.1.1 Critical Boron Concentrations and 9-1 Boron Worths 9.1.2 Xenon Worth and Defect 9-2 9.1.3 Rod Worths 9-2 v
i TABLE OF CONTENTS (Contd.) Page 9.1.4 Reactivity Coefficients 9-4 9.1.5 Power Distribution 9-5 9.1.6 Kinetics Parameters 9-5 9.2 Core Physics Report 9-5 10.0 Physics Test Comparisons 10.1 Introduction 10-1 10.2 Critical Boron Concentrations 10-2 10.2.1 Measurement Technique 10-2 10.2.2 Calculational Technique 10-2 10.2.3 Comparison of Calculated and 10-2 Measured Results 10.2.4 Summa ry 10-3 10.3 control Rod Worth 10-3 10.3.1 Measurement Techniques 10-3 10.3.2 Calculational Techniques 10-3 10.3.3 Comparisons of Calculated and 10-3 Measured Results 10.3.4 Summa ry 10-4 10.4 Ejected Rod Worths 10-4 10.4.1 Measurement Technique 10-4 10.4.2 Calculational Technique 10-4 10.4.3 Comparison of Calculated and 10-5 Measured Results 10.5 Isothermal Temperature Coef ficient 10-5 10.5.1 Hessurement Technique 10-5 10.5.2 Calculational Technique 10-5 vi e _- __
M.- s TABLE OF CONTENTS (Contd.) i 10.5.3 Ccaparison of Calculated and Measured Results 10-6 e 10.5.4 Summary 10-6 11.0 Power Distribution Comparisons 11.1 Introduction and Summary 11-1 11.1.1 Introduction 11 1 11.1.2 Summary 11-1 11.2 Measured Data 11-2 11.2.1 Measured .lssembly Power Data 11-2 11.2.2 Measurament System Description 11-2 11.3 EPRI-NODE-P Power Distribution 11-3
, 11.3.1 EPRI-NODE-P Model 11-3 11.3.2 Fuel Cycle Simulation 11-3 11.3.3 Radial Power Methodology 11-4 11.3.4 Assembly Peak Axial Power 11-5 Methodology 11.3.5 Conclusions- 11-6 11.4 PDQ$7 - Power Distribution Comparisons 11-6 11.5 Statistical Analysis 11-7 11.5.1 Observed Nuclear Reliability Factor Derivation 11-7 11.5.2 Normality Test Results 11-9 11.5.3 Observed Nuclear Reliability Factors (ONRF) for EPRI-NODE-P 11-9 11.5.4 Quantitative Comparison of EPRI-N0DE-P,to Measurement 11-10 11.5.5 Relative Percent Differences 11-11 11.5.6 Conclusions 11-12 vii I
hM
LIST OF TABLES Table M 2-1 Westinghouse System and Component Data Optimized 2-5 Fuel Assembly Design 2-2 Comparison of 17x17 Optimized Assembly and 17x17 2-8 Standard Assembly Design Parameters 3-1 Codes Employed for Cross Section Calculation by 3-8 Composition 4-1 Nuclear Design Basis Data for Reload Design 4-15 4 Shutdown Margin Calculation 4-16 4-3 Maximum FA Factors for Design DNB 4-17 4-4 Boron Parameters 4-18 5-1 BOC Trip Reactivity Calculation 5-9 6-1 Reload Safety Related Physics Parameters 6-7 6-2 Reload Safety Related Kinetics Parameters and Computer Codes 6-8 7-1 . Design Limit F 7-5 q 7-2 F qMargin to LOCA 7-6 8-1 Characteristics of 1/8th Assembly Simulations 8-4 8-2 Peak Pin Power Comparison 8-5 8-3 Statistical Summary of % Differences between PDQ97 8-6 and CASMO-2 1 Critical Boron Concentrations (PPMB) 9-6 9-2 Boron Worth (PCM/PPMB) 9-7 9-3 Core Physics Data 9-8 10-1. McGuire Critical Boron Concentrations at Hot Zero 10-7 Power, BOC 10 McGuire Hot Full Power Critical Boron Concentrations 10-8 10-3 McGuire Control Rod Worths at HZP, BOC 10-9 10-4 McGuire Control' Rod Worths at HZP, BOC - Boron End 10-10 Points viii
LIST OF IABLES (Contd.) Table Pg 10-5 McGuire Control Rod Worths at HZP, BOC - PDQ97 10-11 10-6 McGuire Ejected Rod Worths 10-12 10-7 McGuire Isothermal Temp. Coef. at HZP, BOC 10-13 11-1 McGuire Unit 1 Cycle 1 State Points 11-13 11-2 McGuire Unit 1 Cycle 1A State Points 11-14 11-3 Sequoyah Unit 1 Cycle 1 State Points 11-15 11-4 McGuire Unit 1 Cycle 1 and 1A and Sequoyah Unit 1, Cycle 1 State Points for PDQ97 Calculated and Measured Data 11-16 11-5 Difference Distribution Normality Tests 11-17 11-6 Calculated ORNFs and Associated Data 11-18 11-7 Difference, Means, and Standard Deviat....; for Assembly Radial Powers (C,M 1 1.0) 11-19 11-8 Difference, Means, and Standard Deviations for Assembly Peak Axial Powers (C,M 1 1.0) 11-20 11-9 Percent Difference Means (C,M 1 1.0) - Assembly Radial Powers 11-21 11-10 Percent Difference Means.(C,M 1 1.0) - Assembly Peak Axial Powers 11-22 i I t i ix
LIST OF FIGURES Figure Pa.ge 3-l' Nuclear Flow Chart 3-9 3-2 17x17 Assembly PDQ97 Colorset Geometry 3-10 3-3 PDQ97 Quarter Core Model Assembly Geometry 3-11 3-4 'PDQ97 Quarter Core 17x17 Geometry 3-12 7-1 Design Load Flow Maneuver 7-7 7-2 Reduced Temperature Return to Power 7-8 7-3 Power - AFD Operating Limits - Typical 7-9 8-1 Rod Power Comparison BOL HFP Case Number 1 8-7 2 Rod Power Comparison BOL HFP Case Number 2 8-8 8-3 Rod Power Comparison BOL HFP Case Number 3 8-9 8-4 Rod Power Comparison ~BOL HFP Case Number 4 8-10 18-5 Rod Power' Comparison BOL HFP Case Number 5 8-11 8-6 Rod Power Comparison BOL HFP Case Number 6 8-12 8-7 Rod Power Comparison BOL HFP Case Number 7 8-13 8-8 Rod Power Comparison BOL HFP Case Number 8 8-14 8-9. -Rod Power Comparison'BOL HFP Case Number 9- 8-15 8-10 Rod Power Comparison BOL HFP Case Number 10 8-16 9-1 _ Boron Letdown Curve HFP, ARO 9-9 9-2 Differential Boron Worth HZP, BOL- 9-10 ,. .9-3 Inverse Boron Worth HFP, ARO 9-11 G 9-4 Equilibrium Xenon Worth HFP 9-12 9-5 Xenon Reactivity Defect BOL 9-13 9-6 Integral Rod Worth HZP, BOL 9-14
.10-1 McGuire 1 Cycle 1 Boron Letdown Curves 10-14 i
10-2 McGuire 1 Cycle 1A Boron Letdown Curves 10-15 x
LIST OF FIGURES (Contd.) Figure Page 11-1 Instrumented Fuel Assemblies for McGuire and Sequoyah 11-23 11-2 McGuire Unit 1, Cycle 1 Control and Shutdown Bank Locations 11-24 11-3 McGuire Unit 1, Cycle 1 Core Loading Pattern 11-25 11-4 Sequoyah Unit 1, Cycle 1 Control and Shutdown Bank Locations 11-26 11-5 Sequoyah Unit 1, Cycle 1 Core Loading Pattern 11-27 11-6 McGuire 1 Cycle 1 Calculated vs. Measured Assembly 11-28 to to 11-30 Radial Powers 11-52 11-31 McGuire 1 Cycle 1 Calculated vs. Measured Asseembly 11-53 to to 11-55 Peak Axial Powers 11-77 11-56 McGuire 1 Cycle 1A Calculated vs. Measured Assembly 11-78 to to 11-60 Radial Powers 11-82 11-61 McGuire 1 Cycle IA Calculated vs. Measured Assembly 11-83 to to 11-65 Peak Axial Powers 11-87 11-66 Sequoyah I Cycle 1 Calculated vs. Measured Assembly 11-88 to to 11-72 Radial Powers 11-94 11-73 Sequoyah 1 Cycle 1 Calculated vs. Measured Assembly 11-95 to to 11-79 Peak Axial Powers 11-101 11-80 McGuire 1 Cycle 1 PDQ$7 Calculated vs. Measured 11-102 to to 11-83 Assembly Radial Powers 11-105 11-84 Sequoyah 1 Cycle 1 PDQ97 Calculated vs. Measured 11-106 to to 11-86 Assembly Radial Powers 11-108 xi
- 1. Introduction
, 1.1 Introduction A commercial Pressurized Water Reactor (PWR) is designed to hold a con-stant number of nuclear fuel assemblies which are generally identical me-chanically, but differ in the amount of fissile material content. During cycles subsequent to the initial cycle, fuel assemblies differ in burnup b as well. Refueling occurs at intervals of 6 to 18 months, depending on t the utility's operational requirements. At refueling, a predetermined number of irradiated fuel assemblies are discharged and the same number are -loaded as fresh (reload region) or possibly irradiated assemblies.
The' fuel management scheme determines the locations of all fresh and irradiated assemblies. This report describes some of the various aspects of nuclear design with principal emphasis placed upon development. of a core loading pattern and nuclear calculations performed to evaluate safety and operational parameters. The following sections provide detailed discussion, including descriptions, of design methods, analytical formulations, and calculational procedures involved in the various nuclear design tasks for the McGuirc and Catawba Nuclear Stations. The nuclear design is essentially a series of analytical calculations with the objective of designing the reload core in such a manner that the reactor can be operated up to a specified power level for a specified number of days within acceptable safety and operating limits. It consists of the development of the basic specifications of the reload region (fuel enrichment, number of assemblies, uranium loading, etc.); it
. sets forth the number and identity of each residual fuel assembly, selects the location of each fuel assembly in the core for the new fuel cycle,
' establishes the core characteristics. The nuclear design used in conjunction with the thermal hydraulic and safety analyses establishes the operating limits, control rod limits, and protection system setpoints.
t In arriving at the final nuclear design, the designer tries to meet the 1-1 ll
requirements imposed by the operational considerations, fuel economics considerations, and safety considerations. These requirements are called nuclear design criteria and are as follows:
- 1. Initial core excess reactivity will be sufficient to enable full power operation for the desired length of the cycle.
2 .- The fuel assemblies to be discharged at the end of the fuel cycle will attain maximum permissible burnup so that maximum energy extraction consistent with the fuel mechanical integrity criteria is achieved. ,
- 3. Values of important core parameters (moderator temperature coeffi-cient, Doppler coefficient, ejected rod worth, boron worth, control rod worth, maximum linear heat rate of the fuel pin at various eleva-tions in the core, and shutdown margin) predicted for the cycle are conservative with respect to their values assumed in the safety analy-sis of various postulated accidents, and if they are not conservative, acceptable reevaluation or reanalysis of applicable accidents is performed.
- 4. The power distributions within the reactor core for all possible (or permissible) core conditions that could exist during the operation of the cycle will not lead to exceeding the thermal design criteria of the fuel or exceeding the LOCA-limited peaking factors.
- 5. Fuel management will produce fuel rod power and burnup consistent with the mechanical integrity analysis ot the fuel rod.
The nuclear design process described in this report consists of mechanical properties used as nuclear design input, the nuclear code system and methodology Duke Power intends to use to perform design calculations and to provide operational support, and the development of statistical reliability factors. 1-2 l
1.2 Definition of Terms The terms and symbols used in this report will be consistent with those employed by Westinghouse for its nuclear engineering reports. Presented below are terms which will be needed throughout the text of the report: a/o atom percent ARI all rods in ARO all rods out ' axial offset PT -PB , where PT is the integrated power in (A.O.) PT+Pp the top half of the core, and P B s the integrated power in the bottom half of the core p delayed neutron fraction for group i g S,gg effective delayed neutron fraction in core BOL- beginning of life BP burnable poison BU fuel burnup c C Chemical shim boron concentration in the main coolant B CZP cold zero power EOL end of cycle life EQXE equilibrium xenon condition GWD/MTU Gigawatt days per metric ton of initial uranium metal,1 GWD/MTU is 1000 Mt'D/MTU HFP hot full power HZP hot zero power l l 1 _ _ _ _ _ _
- 5. delayed neutron importance factor AI or flux difference between the top and bottom Axial Flux' halves of the core; in.this reporc, AI is a Difference calculated value, rather than a difference (A.F.D.) between measured signals from the excore detectors T
'K(z). F normalized to the maximum value allowed at any chreheight A* . prompc neutron lifetime MOL' middle of cycle life MWD /MTU' measure of energy extracted per unit weight of initial uranium metal fuel; is equal to 1 megawatt times I day, divided by 1 metric ton of uranium pcm percent mille (a reactivity change that equals 10 5 ap) ppa ' parts per million by weight; which specifies the amount of chemical shim boron present by weight in the main coolant system radial local ratio of assembly maximum rod to assembly average x-y power RCCA rod cluster contrcl assembly; the type of control
~
rod assembly used in McGuire and Catawba. (All RCCA are full length absorbers for both plants.) p reactivity K1 -K 2 , where Kt and K 2are eigenvalues 3p obtained from two Calculations where only one parameter was varied shutdown amount of negative reactivity (p) by which a margin reactor core.is maintained in a HZP subcritical condition after a control rod trip step unit of control rod travel equal to 0.625 inch T m derator temperature; defined as the temperature MOD
' corresponding to the. average water enthalpy of the core T ,, resonance temperature of the fuel w/o weight percent 1-4
Power distributions will be quantified in terms of hot channel factors.
-These factors are a measure of the peak pellet power and the energy produced in the coolant. The factors are:
Power density thermal power produced per unit volume of the core (KW/ liter) Linear Power Density thermal power produced per unit length of active fuel ' (KW/ft) Average Linear Power Density total thermal power produced in the core divided by the total . active fuel length of all fuel rods in the core Local Heat Flux local heat flux on the cladding surface (BTU /ft2 /hr) Rod Power or Integral Power is the length integrated linear power density in one rod (KW)- Various hot channel factors are: T Fq, Heat Flux Hot Channel Factor, the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod flux, including conservatisms for fuel pellet and rod dimensional uncertainties. F , Nuclear Heat Flux Hot Channel, is defined as the maximum local fuel rod linear power density divided by the average linear power density, assuming nominal fuel rod and pellet dimensions. Fq , Engineering Heat Flux Hot Channel, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter. Combined statistically the net effect is a factor of 1.03 to be , applied to calculated KW/ft. FAH, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest in-tegrated power to the average rod power.
- It is convenient, for the purposes of discussion, to define subfactors T
of F9; however, design limits are set in terms of the total peaking factor. I 1-5
F = Total peaking factor or Heat Flux Hot Channel Factor _ Maximum KW/ft Average KW/ft without densification effects F = xF q
= max [Fg (Z) x P (Z)] x x where:
and F are defined above. q
= 95/95 Relability factor for Fq (Section 6.2.1.2)
Fg (2) = ratio of peak power density to average power density in the horizontal plane at elevation Z
.P(Z) = ratio of the power per unit core height in the horizontal plane at elevation Z to the average value of power per unit core height including densification allowance F = max [Fg (Z) x P(Z) x S(Z)] x xF q where S(Z) = the allowance made for densification effects at height Z in the core.
1-6
l I
- 2. Fuel Description The fuel is described as a composition of fuel assembly (material selection, I
fuel rod lattice), spacer grid (material selection, number of spacer grids and_ interface with internals), and fuel rod (rod dimensions, cladding type and dimensions, pellet density and dimensions, fuel stack height, fill gas pressure and composition). 2.1 Fuel Pellet The fuel pellets are right circular cylinders consisting of slightly en-riched uranium dioxide powder which has been compacted by cold pressing and then sintered to the required density. The ends of each pellet are dished slightly to allow greater axial expansion at the center of the pellets. 2.2 Fuel Rod The fuel rods consist of uranium dioxide ceramic pellets contained in slightly cold worked Zircaloy-4 tubing which is plugged and seal welded at the ends to encapsulate the. fuel. Void volume and clearances are provided within the rods to accommodate fission gases released _from the fuel, differential thermal expansion
-between the cladding and the fuel, and fuel density changes during irra-diation, thus, avoiding overstressing of the cladding or seal welds.
Shifting of the fuel within the cladding during handling or shipping prior to core loading is prevented by a stainless steel helical spring which bears on top of the fuel. During assembling of these rods, the pellets are stacked in the cladding to the required fuel height, the
- spring is then inserted into the top end of the fuel tube and the end plugs are pressed into the ends of the tube and welded. All fuel rods are internally pressurized with helium during the welding process in order to minimize compressive cladding stresses and prevent clad flatten-ing due to coolant operating pressure.
, 2-1
u 2.3 Fuel Assembly Each fuel assembly consists of 264 fuel rods, 24 guide thimble tubes and 1 instrumentation thimble tube all arranged within a supporting structure. The' fuel assembly structure consists of a bottom nozzle, top nozzle, guide and instrument thimbles and grids. The instrumentation thimble-is located in the center position and provides a channel for insertion of an incore neutron detector, if the fuel assem-bly is located in an instrumented core position. This tube is a constant diameter and is expanded at the top and midgrids to force the thimble and sleeve outward to a predetermined diameter, thus joining the two components. The guide-thimbles are structural members which provide channels for the i neutron absorber rods, burnable poison rods, neutron source or thimble plug assemblies. .Each thimble is fabricated from Zircaloy-4 tubing having j two'different diameters. The tube diameter at the top sections provides the annular area necessary to permit rapid control rod insertion during a reactor trip. The lower portion of the guide thimble is swaged to a
- - smaller diameter to reduce diametral clearances and produce a dashpot ac-
, tion near the end of the control rod travel during normal trip operation. Holes are provided on the thimble tube above the dashpot to reduce the rod drop time. The dashpot is closed at the bottom by means of an end plug which is provided with a small flow port to avoid fluid stagnation in the dashpot volume during normal operation.
.The bottom nozzle serves as the bottom structural element of the fuel as-sembly'and directs the coolant flow distribution to the assembly. The square nozzle is fabricated from Type 304 stainless steel and consists of a perforated plate and f'our angle legs with bearing plates. These legs form a plenum for the inlet coolant flow to the fuel assembly. The plate
, also prevents accidental ~ downward ejection of the fuel rods from the fuel assembly. The bottom nozzle is fastened to the fuel assembly guide thimbles by weld-locked screws which penetrate through the nozzle .tud mate with a threaded plug in each guide thimble.
; 2-2
The top nozzle assembly functions as the upper structural element of the fuel assembly and provides a partial protective housing for the rod clus-
-ter control assembly or other core components. It consists of an adapter plate, enclosure, top plate and pads. The springs and bolts are made of Inconel 718, whereas the top nozzle is made of Type 304 stainless steel.
The square adapter plate is provided with round penetrations and semi-circular ended slots to permit the flow of coolant upward through the top nozzle. Other round holes are provided to accept sleeves which are welded to the adaptor plate and mechanically attached to the thimble tubes. The top plate has a large square hole in the center to permit access for the control rods and the control rod spiders. Holddown springs are mounted on the top plate of the top nozzle assembly
.and are fastened in place by bolts and clamps located at two diagonally . opposite corners. On the other two corners integral pads are positioned which contain alignment holes for locating the upper end of the fuel assembly.
The fuel rods are supported at intervals along their length, by grid as-semblies which maintain the lateral spacing between the rods. Each fuel b rod is supported within each grid by the combination of support dimples and springs. The magnitude of the grid restraining force on the fuel rod 11s set high enough to minimize possible fretting, without overstressing the cladding at the point of contact between the grids and the fuel rods.
.The grid assemblies also allow axial thermal expansion of the fuel rods without imposing restraints sufficient to develop buckling or distortion of the fuel rods.
Two types of grid assemblies are used in each fuel assembly. Both types consist of individual slotted straps interlocked into an " egg-crate" arrangement. One type, with mixing vanes projecting from the edges of the straps into the coolant stream, is used in the high heat flux region of the fuel assemblies to promote mixing of the coolant. The second type, located at the ends of the assembly, does not contain mixing vanes on the internal straps. .The outside straps on all grids contain mixing vanes i which, in addition to their mixing function, aid in guiding the grids and 2-3
fuel assemblies past projecting surfaces during handling or during loading and unloading of the core. As identified in Table 2-2, the optimized fuel assembly (OFA) design uses a Zircaloy-4 material for the intermediate grids. This material is primarily chosen for its low neutron absorption properties. The material used for the end grids is Inconel-718, chosen for its corrosion resistance and high strength properties. This design is compared to the standard fuel assembly design which utilizes Inconel-718 for all grid assemblies. Because of the considerable design, engineering and testing needed to incorporate a new fuel design into a reload core, it is usually not con-sidered unless there is sufficient economic, engineering or regulatory justification. If, however, sufficient justification exists, the new fuel design is typically documented in a generic topical report and the reload report would reference this topical report. 2.4 Core Component Data The basic physical dimensions and materials of the fuel pellet, fuel rod, fuel assembly, rod cluster control assembly, burnable poison assembly, and neutron source assembly are used in the fuel cycle design, thermal-hydraulic design and fuel mechanical performance. Table 2-1 presents a summary of this data for the Westinghouse standard fuel assembly design. In addition, Table 2-2 is presented to identify the significant differ-ences between Westinghouse standard and optimized fuel assembly design. All data presented is intended as an example. 1 I I 2-4
TABLE 2-1 WESTINGHOUSE SYSTEM AND COMPONENT DATA OPTIMIZED FUEL ASSEMBLY DESIGN I
-Active Core Equivalent Diameter, in. 132.7 i- Core Average Active Fuel :leight, First Core, in. (cold dimensions) 144 Height-to-Diameter Ratio 1.09
-Total Cross-Section Area, fta 96.06 H2 0/U Molecular Ratio, Lattice (Cold) 2.73 Fuel Assemblies
- Number 193
' Rod Array 17 x 17 Rods per Assembly- 264 E
_ Rod Pitch, in. 0.496 Overall Transverse Dimensions, in. 8.426 x 8.426 Fuel Weight (as UO 2 ), Ibs. 204,200 Zircaloy Weight, lbs. (active core) 45,352
' Number of Grids per Assembly two - R type six - Z type Composition of grids two INC718 End Grids six ZIRC4 Spacer Grids Weight of Grids (Effective in Core) lbs. INC - 332 Zire - 2985 Number of Guide Thimbles per Assembly 24 Composition of Guide Thimbles Zircaloy 4 Diameter of Guide Thimbles (upper 0.442 I.D. x 0.474 0.D.
part), in. Diameter of Guide Thimbles (lower 0.397 I.D. x 0.429 0.D. part), in. Diameter of Instrument Guide Thimbles, in. 0.442 I.D. x 0.474 0.D. A 2-5
TABLE 2-1 (Continued) Fuel Rods Number Per Assembly 264 Outside Diameter, in. 0.360 Diametral Gap, in. 0.0062 Clad Thickness, in. 0.0225 Clad Material Zircaloy-4 Fuel Pellets Material UO2 Sintered Density (percent of Theoretical) 95 Diameter, in. 0.3088 Length, in. 0.510 Mass of UO2 Per Foot of Fuel Rod, lb/ft 0.334 Hybrid Rod Cluster Control Assemblies Neutron Absorber BC 4 B 4 C Diameter, in. 0.294 Density, lbs/in 3 0.064 Tip Material Ag-In-Cd Composition 80%, 15%, 5% AgInCd Diameter, in. 0.301 Length, in. 40 Density, lbs/in 3 0.367 Cladding Material Type 304, Cold Worked Stainless Steel Clad Wall Thickness, in. 0.0385 Number of Clusters Full Length 53 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight (dry), Ib. 90 2-6
TABLE 2-1 (Continued) Burnable Poison Rods Material Borosilicate Glass Outside Diameter, in. 0.381 Inner Tube, 0.D., in. .1815 Clad Material- Stainless Steel Inner Tube Material Stainless Steel Boron Loading (w/o 2B 30 in glass rod) 12.5 Weight of Boron - 10 per foot of rod, lb/ft .000419 Ag-In-Cd Rod Cluster Control Assemblies Neutron Absorber Ag-In-Cd Composition 80%, 15%, 5% Diameter, in. 0.341 Length, in. 142.0 Density, lbs/in3 0.367 Cladding Material. Type 304, Cold Worked Stainless Steel Clad Wall Thickness, in. 0.0185 Number of Absorber Rods- 24 per Cluster 2-7
TABLE 2-2 COMPARISON OF 17 x 17 OPTIMIZED ASSEMBLY AND 17 x 17 STANDARD ASSEMBLY DESIGN PARAMETERS Optimized - Pa ramete r Assembly Design Standard Design Fuel Ass'y. Length, in. 159.8 159.8 Fuel Rod Length, in. 151.6 151.6 Assembly Envelope, in. 8.426 8.426 Compatible with-core internals Yes Yes Compatible w/ Movable In-Core Detector Syatem Yes Yes ~ Fuel Tube Material Zircaloy-4 Zircaloy-4 Fuel Rod Clad OD, in. 0.360 0.374 Fuel Rod Clad Thickness, in. 0.0225 0.0225 Fuel / Clad Gap, mil 6.2 6.5 Fuel Pellet dia. in. 0.3088 0.3225 Relative UO 2/ Rod 0.92 1.0 Guide Thimble Material Zircaloy-4 Zircaloy-4 Guide Thimble OD, in. 0.474 0.482 Guide Thimble Wall Thickness,.in. 0.016 0.016 Spacer Grid Structural Mat'i. inner Zircaloy-4 Inconel-718 grid (6) Spacer Grid Structural Mat'l. end grid-(2) Inconel-718 Inconel-718 Grid Support for Fuel Rods 6 Point; 2 Springs + 6 Point; 2 Springs 4 Dimples + 4 Dimples Grid Height, inch, less 2.25 1.32 vanes / inner straps Grid Fabrication Method Laser welded joining Brazed joining of of interlocking interlocking stamped straps stamped straps Grid / Guide Thimble Attach. Thimbles bulged Thimbles bulged together with sleeves together with sleeves laser prewelded onto prebrazed onto grid grid straps straps 2-8
1
- 3. Nuclear Code System 3.1 Introduction Nuclear design calculations performed for Westinghouse reactors employ the .
EPRI-ARMP code system 1 and the CASMO-2 code a
. A summary description of each code is given in Appendix A. The ARMP/CASMO-2 code sequence has been reviewed and approved by the NRC for use in the design of reload cores for Oconee Nuclear Station by Duke Power 3 . Presented in this section will be a description of the sequence, cross section preparation and parameteriza-tion, PDQ974modeling procedures, and EPRI-N0DE-P5 modeling procedures.
^ The nuclear calculational system enables the nuclear engineer to numeri-cally model and simulate the reactor core. The system used by Duke Power for McGuire and Catawba is outlined in Figure 3-1. 3.2 ' Sources of Input Data The determination of nuclear fuel loading patterns and core physics char-L acteristics requires an accurate database consisting of: 1
- 1. Core operating conditions
- 2. Dimensional characteristics
- 3. Composite materials and mechanical properties
- 4. Nuclear cross sections The FSAR, supplemented by vendor reports and open literature,-is the pri-mary source of data for' items 1 to #3. These data are used as_ input to the cross section generators and core simulators. A secondary data source for the core simulators are estimates of fuel pellet volume-averaged tempera-tures which are calculated by a fuel performance code, such as COMETHE-7 L IIIKs. or' TACO-2 , as functions of power and burnup.
4
~
The cross sect' ion generators CASMO-2 and EPRI-CELLS use processed ENDF/B
. libraries unique to each code.
EPRI-CELL is a unit cell lattice code which is used to calculate few group cross sections for fuel and nonfuel composi'. ions as shown in Table 3-1. 3-1
CASMO-2 uses a processed version 8 of the ENDF/B-3 library. Group cross sections of o,, of, vo g, otr,. scattering kernels, resonance integrals, and
- fission product data are among the data contained in this library. The 69 group lib'rary is' divided into 14 fast, 13 resonance, and 42 thermal energy range groups. A~25 group version of this library is also used.
The EPRI-CELL library is derived from the ENDF/B-4 library 10 The 97 en-ergy groups are divided into 62 fast groups and 35 thermal groups. 3.3 Cross Section Preparation In order to model the neutronics of a reload core, it is necessary to gen-erate a set of cross sections for use in a diffusion theory code. Two cross section generators are currently used at Duke Power: Table 3.1 shows the core materials or compositions which are parameterized by CASMO-2 and EPRI-CELL. Inputs which are provided to both codes are: lattice materials and geom-etry, temperatures for fuel, clad, and moderator, effective resonance temperature, fuel enrichment, soluble boron concentration, number of de-pletion steps, length of depletion steps, etc. 3.3.1 Fuel Calculations Calculations for fuel regions employ fixed fuel and moderator tempera-tures for the cell depletion. Restart calculations are performed at 4 various burnups to parameterize fuel cell cross sections at varying moderator and fuel temperatures. The output of EPRI-CELL and CASMO-2 consists of sets of broad group cross sections which characterize the regions of interest. Cross sec-tions are then formatted into PDQ$7 tableset structure using either NUPUNCHER11 (1-dimensional parameterization), or MULTIFIT 12 (2 and 3 dimensional parameterization or g-factors). Cross sections from 13 CASMO-2 are similarly formatted using CHART. 3-2
3.3.2 Non-Fuel Calculations Cross. sections for empty control rod guide tubes, reflector, instrument thimble, and the water gap are calculated with either EPRI-CELL or CASMO-2. Separate cross section sets are generated for various modera-tor temperatures. Strong absorbers such as RCCA and BP require reaction rate matching to i obtain diffusion theory egivalent cross sections. Calculations using CASMO-2 are performed for these strong absorbers where first a transport theory method determines absorption rates, and then a series of diffu-sion theory iterations are performed to calculate a g-factor such that the absorption rates agree between both types of flux solutions. These g-factors are then incorporated in the tabulated cross sections. RCCA cross sections are evaluated at BOL HFP conditions, while BP cross sections are evaluated with an HFP depletion calculation. In both types of g-factor calculations, lattices with expected core tv-erage enrichments are used. Core baffle cross sections are also calcu-lated with CASMO-2. A lattice geometry is employed, with the baffle material density modified to reflect real versus modeled thickness in the quarter core PDQ97 discrete pin model. 3.4 PDQ97 Models The PDQ97 few group diffusion-depletion code is employed for core modeling. Two different models are used currently. The first is the assembly colorset model, which is used for calculating k,and Ma data for EPRI-NODE-P 3-D simulations. The second model is the quarter-core model, which is used for X-Y power distribution calculations and for normalization of EPRI-NODE-P radial power distributions. Aspects which are common to both PDQ97 models are:
- 1. Discrete pin representation
- 2. Two group cross sections
- 3. Mixed Number Density thermal group constants 3-3
}
- 4. Improved Removdl Treatment removal cross sections
- 5. . Microscopic cross.section parameterization for uranium, plutonium,
' burnable absorber, soluble boron, xenon, samarium, and lumped fission products
- 6. Thermally expanded geometry - pin pitch and assembly pitch
-3 .'4.1 .Colorset PDQ97 Modeling The colorset PDQ$7 model' consists typically of four quarter assemblies arranged such that a representative neutron spectrum is obtained.
Figure 3-2 shows a typical colorset geometry. To accommodate asymmetric burnable poison rod loadings, full or half assembly geometries are used. The EPRI-ARMP PWR Procedures 14 are used for modeling, and most of the conventions and guidelines are employed. Fuel types are determined according to enrichment and BPRA loading. Km and Ma data for each fuel type are calculated by performing the follow-ing operations: I~. BOL Cases - 0 MWD /MTU A. k,and Mr .unrodded vs. Teod (Inlet, Average, Outlet) B. . k, and M 2 - rodded vs. Teod (Inlet, Average, Outlet) C. Boron worth D. Doppler worth II. Depletion Data - Exposure dependent data A. -Nominal HFP depletion at constant T eod' fuel
-B. Branch cases from depletion
- 1. Boron worth
- 2. Control rod worth
- 3. Equilibrium Xenon worth
- 4. Doppler worth
- 5. Moderator temperature worth In.the above PDQ97 branch calculations, only one parameter is varied, j allowing a partial derivative of reactivity with respect to that para-meter to be calculated.
3-4
The parameterization procedure involves approximately 150-200 cases, depending on the number of depletion steps.
- The output from the PDQ97 colorset cases is written to PDQ97 integral files which in turn are processed by the linking codes EPRI-FIT 15 and SUPERLINK i s to yield B-constant data for EPRI-N0DE-P.
~3.4.2 Quarter Core PDQ97 Model Two-dimensional X-Y core simulations are performed with a discrete pin PDQ97 model. Assembly average and maximum pin powers are calcu-lated, along with critical boron concentrations and other reactivity parameters. Moderator and doppler feedbacks are incorporated in this m odel.
The geometry employed utilizes thermally expanded dimensions. Figure 7 3-3 shows a geometry of a fuel assembly and water gaps. Figure 3-4 shows the complete quarter core mesh layout.
' The plane of solution used in quarter core analyses is the axial midplane or the six foot level of the active fuel. Moderator and doppler feedbacks are employed as described in Reference 17.
The depletion calculation is used to determine burnup dependent parameters. The soluble boron concentration is modified at each timestep such that the reactor is approximately critical. Timesteps are taken using point depletion so that the core average exposure advances by: 150, 500, 1000, 2000, ..., N
- 2000 MWD /MTU until end of life (EOL)-is reached. EOL is typically defined as that exposure where the critical boron at hot full power, equilibrium xenon conditions is 10 ppmB.
PDQ97 depletion calculations are used to determine the following
. parameters:
- 1. Assembly. average and maximum pin powers
'2. Core reactivity 3-5
1 i
. 1
- 3. Nuclide' reaction rates: Fission and absorption
- 4. 'Nuclide inventories
- 5. Neutron flux distributions Other calculations performed with the quarter core model may include:
'1. RCCA b'ank worths
- 2. Boron and xenon worths
- 3. Power deficits
'4. Moderator and doppler temperature coefficients Cases 2,.3, and 4 are usually performed with a nodal code; however, these are shown to demonstrate the quarter core model's flexibility.
3.5 EPRI-NODE-P Model
'EPRI-NODE-P is the nodal code employed for three-dimensional analyses and reactivity studies. A' summary description of EPRI-NODE-P is given in Appendix A. ' Typical calculations which are performed with the Duke Power
- EPRI-NODE-P model are:
~1. Full core ejected rod worths
- 2. Power deficits
- 3. Differential rod worths
- 4. Axial xenon transients
- 5. -Three-dimensional power distributions, etc.
The quarter core model uses one radial node per assembly and eighteen axial nodes. .- Each unique combination of enrichment and BPRA loading comprises a sepa-rate fuel type. The fuel type is parameterized by sets of fitting coeffi-cients which determine reactivity due to control rods, exposure, soluble boron, xenon, etc. Doppler and moderator feedbacks are explicitly treated. EPRI-NODE-P radial power distributions are normalized near the beginning of cycle. Assembly average powers are adjusted to match quarter core ! PDQ97' calculations with radial albedoes - Ha and an internal leakage fac-tor gH. The axial power distribution, is adjusted using vertical leakage 3-6
factors a ydetermined from comparisons of calculated and measured axial 4 power distributions from benchmark core follow calculations. Sections 5, 6, 7, and 9 discuss indepth calculational precedures of EPRI-NODE-P. Sections 10 and 11 address benchmarking cf EPRI-NODE-P and PDQ97 calculations to measured power and reactivity data. f 1 3-7
g TABLE 3-1 Codes Employed for Cross Section Calculation by Composition
- 1. EPRI-CELL
. v.
- a. Uraaium Fuel
- b. Empty Control Rod Guide Tube / Instrument Tube
- c. Reflector:.3
- d. Water Gap e
- 2. CASMO-2~
s
- a. Burnable Poison Rod Assembly
- b. Gadolinia doped Urauium Fuel
'c. ' Control Rod - AgInCd or.B4C' s d. Baffle s x'>
\
'% I s
Y
\
A s - , 3-8 5
Figure 3-1 NUCLEAR FLOW CHART FOR EPRI-ARMP EPRI-CELL u NUPUNCHER or : 1 MULTIFIT
; CPM or CASMO :
y u PDQ07 COLOR SET 2-D 1/4 CORE PDQ07 U EPRI-SHUFFLE EPRI-FIT U SUPERLINK 1F EPRI-NODE : u Desired 3-D Information 9 9 3-9
1 FIGU RE 3-2 17X17 ASSEMBLY PDQ07 COLORSET GEOMETRY 1 1 1 1 1 2 0 2 4 6 8 0 2 4 6- 8 0 IT ' F F G G C C IT p , 2 l p G G G C C C l
- - - ~
4- FUEL ASSEMBLY REGir l G C
- ~
6 C
/ G.
bOOOOO'" Color No.1 Color No. 2 O
/
l
'10 WATER GAP
< Color No. 4 Color No. 3 r f
~'
f 'T B B G G 14 - ' i - B 1 G
- - """ ~ ~
16 B G B G G G 18 ' IT B B G .G IT 20 . , b LEG E4Q'. IT INSTRUMENTTU8E C GulDE TUBE WITH CONTROL ROD G EMPTY GulOE TUBE : B GUIDE TUBE WITH BURNABLE POISON ROD F FUEL ROD 3-10
FIGURE 3-3 PDQ07 QUARTER CORE MODEL ASSEMBLY GEOMETRY
~
a G G G G G G G G G G z 9 0 7G E G IT G G w
- s s
G G G G G G G !. G G G l-1r C WATER GAP : l l l FUEL R EGION ; i i-L LEGEND; IT: INSTRUMENT TUBE G: CONTROL ROD GUIDE TU8E i 1 t 3-11
FIGU RE 3-4 PDQ07 QUARTER CORE 17X17 GEOMETRY 1 1 1 1 ZERO CURRENT 6 8 0 2 4 6 BOUNDARY 6 4 3 2 0 9 3 5 m - I QUARTER - ASSEMBLY w
~
I M N N ONE-HA LF - ASSEMBLY Mi
~
FULL / ASSEMBLY _ m M
- BAFFLE R EF LECTOR 1
4 ZERO FLUX BOUNOARY 1'.19 i
i
- 4. Fuel Cycle Design 4.1 Preliminary Fuel Cycle Design - Initialization To commence the design of a reload, an initialization procedure is used.
Core operation requirements along with planned changes in reactor primary or secondary systems are assembled. A preliminary loading pattern is de-signed which meets operational requirements. Physics data from the pre-liminary design are compared with core operating requirements to determine the adequacy of the reload design. Likewise, physics data are compared to Technical Specifications to verify that the preliminary design will con-form to existing limits. 4.1.1 Review of Design Basis Information The preliminary design procedure requires assembly of design basis information which in turn will determine the cycle's operational capabilities. Typical design-basis data are shown on Table 4-1. Table 4-1 and other pertinent nuclear design data are assembled and reviewed for consistency with previous sets of design basis data. 4.1.2 Determination of Cycle-Specific Operating Requirements Design basis data-from Table 4-1 uniquely determines expected operat-ing requirements and capabilities. For instance, a longer than annual cycle may require a low leakage loading pattern and the use of burnable absorber rods. A larger energy requirement than can be provided by nor-mal operation with a given reload enrichment may require a planned power coastdown at end of cycle. Similarly, other design bases will govern the rest of the cycle-specific operational characteristics. 4.1.3 Preliminary Loading Pattern and Reload Region Determination
' The purpose of a preliminary Joading pattern analysis is to determine the uranium and sep' a rative work requirements to meet a desired cycle .
lifetime. The cycle lifetime is either estimated by the BOL excess
. reactivity (p, c) or by a reload core depletion with a coarse mesh PDQ$7 or a nodal (EPRI-NODE-P) model. If the number of new fuel assemblies and their enrichment is known, this analysis will yield an estimate of the cycle lifetime.
4-1
When uranium requirements are to be determined, an iterative series of BOL calculations are performed in which:
- 1. A loading pattern is developed with a reasonable radial power distribution.
- 2. The BOL excess reactivity (p, ,) is calculated when p ec is defined below.
- 3. If p is sufficient to meet design cycle burnup, the chosen fixed exc enrichment and number of assemblies is used.
- 4. If p,, is not sufficient, the enrichment / number of assemblies are modified and Step 1 is again performed.
The analysis of BOL excess reactivity uses the following definitions of P exc K' -1 p E exc " K,'ff e where: K',gg is the HFP, EQXE, equilibrium SM, at BOL at BOL and 0 PPMB. is a fairly linear fun tion of burnup. lor out-in reloads, p exc Therefore, for this type of reload, an estimate of cycle life-time can be made with a high degree of confidence. However, for cores which use low leakage loading patterns, the cycle lifetime is usually confirmed by using either a coarse mesh PDQ97 model or a N0DE-P model depletion of the reload core. 4.2 Final Fuel Cycle Design Having determined the number and enrichment of the fuel assemblies during the PFCD, the final fuel cycle design (FFCD) coacentrates on optimizing the placement of fresh and burned assemblies and burnable poison assem-blies (if any) to result in an acceptable fuel cycle design. It must meet the following d'esign criteria with appropriate reductions to account for calculational uncertainties: 4-2
- 1. Fg - See Table 4-3.
- 2. Moderator Temperature Coefficient must meet Technical Specification requirements for all operating modes.
- 3. Maximum pellet burnup i 50,000 MWD /MTU.
- 4. Shutdcwn Margin must meet Technical Specification requirements for all operating modes.
- 5. Maximum linear rod power 1 12.9 Kw/ft at 102% power.
- 6. Ejected rod Fq and worth to be within limits of FSAR.
- 7. Dropped rod maximum Fg to be within F3AR limits.
During the FFCD, nuclear calculations are performed to generate these parameters for input to fuel mechanical performance analyses, thermal and thermal-hydraulic analyses, and maneuvering analyses. 4.2.1 Fuel Shuffle Optimization and Cycle Depletion
'BOC power distribution calculations are performed using combinations of EPRI-SHUFFLE and PDQ97. Initial runs start with the fuel shuffle scheme developed in the PFCD and the shuffle scheme (fuel assembly loading pattern) is modified to minimize power peaking. This is accom-plished by a trial and error type search until an acceptable BOC power distribution results.
The reload design's "burnup window" is assessed at BOC to ensure that preliminary safety criteria are met. That is, in the design of cycle N, effects of tolerances in core burnup achieved during cycle N-1 are examined. At the lower cycle N-1 core burnup, the moderator temperature coefficient is calculated to ensure that Technical Specification limits are met. At the higher cycle N-1 burnup, the power distribution is cal-culated to verify that power peaking is not excessive. This analy, sis ensures that the design of cycle N is adequate at BOC, given.a prede-termined tolerance on the core burnup for cycle N-1. The cycle is then depleted in steps corresponding to 0, 150, 500, 1000, 2000, 4000... MWD /MTU to verify that power peaking versus burnup remains acceptable. The shuffling variations include rearranging the location 4-3
of the burned or fresh fuel assemblies, BP placement, and rotation of the spent fuel assemblies. These calculations are typically performed assuming quarter core symmetry. The shuffle pattern determined by optimizing power distribution may later need to be modified based upon results obtained in the remaining nuclear calculations. 4.2.2 Rod Worth Calculations Control rods serve several functions in the McGuire reactor. The pri-mary function is to provide adequate shutdown capability during normal and accident conditions. They are also used to maintain criticality during power maneuvers and to maintain the Axial Flux Difference (AFD) within Technical Specification limits. Since the presence of control rods influences both power distributions and criticality, it is neces- . sary in many calculations to evalute not only the reactivity effect but also the perturbation that a given rod configuration has on the power distribution. McGuire and Catawba are typically operated in the ARO or feed and bleed mode. All RCCA have full length absorber rods. During full power oper-ation, control bank D is typically inserted about six inches (215 steps withdrawn) in the active core. Bank D is used to control power during load follow maneuvers and, in conjunction with banks B and C, to achieve criticality during startup. Calculations of control rod worth and power peaking (Fq ) are used in the safety analysis of the reload core. The calculations discussed in sub-sequent sections include the following:
- 1. Control Rod Worths
- 2. Shutdown Margin
- 3. Ejected Rod Analysis
- 4. Dropped Rod Analysis 4-4
4.2.2.1 Control Rod Worths RCCA bank locations in McGuire and Catawba usually are fixed and do not change from cycle to cycle. The worth of each control bank (A, B, C, D) is calculated in quarter core geometry using either PDQ97 or
.EPRI-NODE at BOC and EOC, at HFP and HZP. The total rod worth (ARI) is calculated at BOC, EOC, and any-limiting burnup at HZP only for use in the shutdown margin calculation.
4.2.2.2 Shutdown Margin Searches for the highest worth stuck rod are performed at BOC, EOC, or any limiting burnup for HZP conditions using full core EPRI-NODE and/or PDQ97 calculations. Table 4-2 summarizes the results of a shutdown margin calculation. The total rod worth described in section 4.2.2.1 is shown as Item 1. Item 2 is the worth of the highest stuck rod. The total worth reduced by the stuck rod worth is shown as the net worth (Item 3). A calcula-tional uncertainty of 10% is subtracted ofi in Step 4, and Step 5 shows the available rod worth. The required rod worth is calculated next in Steps 6-9. The power deficit obtained by running an EPRI-N0DE or PDQ97 cases at HFP and HZP (using constant Boron and Xenon) and subtracting the reactivities is shown as item 6. This reactivity insertion accounts for Doppler and Moderator deficits. The maximum allowable inserted rod worth, item 7, is obtained from the allowable rod insertion and the integral rod worth curve for that insertion (generated by EPRI-NODE). This ac-counts for the maximum allowed rod insertion at HFP. An axial flux redistribution occurs when the power level is reduced from HFP to HZP. This-redistribution causes an increase in reactivity. If Item 6 is calculated using-a 3-D code such as EPRI-NODE, no additional penalty is' required here. However, if Item 6 was calculated using 2-D PDQ97, an additional reactivity penal.ty is assessed as Item 8. The sum of
'these required worths (Item 9) is the total required worth.
4 4-5
an \ a
'The. shutdown margin is shown as Item 10 and is defined as the total available worth minus the total required worth. For McGuire and
-Catawba, the Technical Specification requirements are 1.3% ap for T 1 00*F and 1.0% ap for TAVG < 00*F.
AVG 4.2.2.3 Rod Insertion Limit Verification As part of the reload design procedare, the current Rod Insertion
~
Limits are' verified for applicability in the reload core. (See Section 5.4.) 4.2.2.4 Ejected Rod Analysis The Final Safety Analysis Reportissis (FSAR) presents two limiting criteria for the ejected rod accident: hot channel factor (F q) and
. reactivity insertion. The accident has been analyzed at HFP and HZP conditions at'BOL and EOL.
Ejected rod calculations are performed on a Unit-specific basis to verify that reactivity insertions and hot channel factors do not exceed original FSAR accident analysis values. Calculational limits are established by reducing FSAR values of reac-tivity insertion 10% and hot channel factor by 15%. i. To verify that the ejected rod parameters are within calculational
' limits, ejected rod calculations are performed at BOC and EOC or at other limiting times in cycle life at both HFP and HZP.
The calculation of ejected rod parameters is accomplished using full core two dimensional pin mesh PDQ97 or EPRI-N0DE calculations. The HZP ejected rod calculations are performed with control banks B and C at their insertion limits in the core and with bank D fully inserted. Single rods in banks D, C, and B are removed in subsequent cases and the worth of the ejected rod is calculated by subtracting the reac-tivities of the cases before and after the rod was removed. The fuel 4-6
..,c , , - - - ,,-,,,,--,,_,.,,,.,,,,,,,,,n, , , , - , , . , - - , , , - - - - ,,-------w-,,,, - - - , , - , + - - - - - - ---,w - ---
and moderator temperature is held constant and equal to the HZP moder-ator temperature for these calculations. The highest worth calcu-lated by the above procedure is the worst ejected rod at HZP. If the ejected - rod worth exceeds the calculational limit, rod insertion
~ limits are revised.
The HFP ejected rod worths are performed in a similar manner to the 4 HZP calculations with the exceptions that only bank D is inserted at the HFP insertion limit and that the fuel temperature and moderator temperatures correspond to those of HFP conditions. The HFP ejected E rod worths are performed without thermal feedback to be conservative. If-the ejected rod worth exceeds the calculational limit, rod inser-
. tion limits are revised.-
.A parallel analysis, addressing F q, is performed at the same time as
.the rod worth analyses.
4.2.2.5 Dropped Rod Analysis > The calculation of the rod drop peaking factor is required to deter-
.mine the DNBR resulting from the. rod drop. Full core calculations using_EPRI-NODE are performed with thermal-hydraulic feedback.
Search cases are run where single RCCA are: inserted until the maximum N Fg and rod worth have been determined. RL factors, determined from PDQ97. discrete pin calculations, are combined with nodal power data from EPRI-NOEE with the NUC-MARGINS Code. NUC-MARGINS then calculates Fg using the above data and additional factors to account for conser-vatism, tilt, and other parameters which would affect the value of Fg. This dropped rod F and worth are used as input to the accident analy-sis evaluation. 4.2.3 ; Fuel Burnup Calculations One of the current design criteria is that maximum pellet burnup is
< 50,000 MWD /MTU. This criterion is confirmed during the final fuel
. cycle design. Depletion calculations from 2-D quarter core pin mesh PDQ97'models yield core, assembly average, and single fuel rod burnups.
From these values a maximum ratio of single rod to assembly average 4-7
burnup can be calculated for each assembly. This data is then used in conjunction with 3-D EPRI-N0DE depletion calculations (where the axial burnup distribution is calculated) to arrive at a local burnup limit which can be compared to the design limit of 50,000 MWD /MTU.
- Generally, the assembly average burnups are in the 36,000 MWD /MTU range and sufficient margin to the 50,000 MWD /MTU limit exists to make the detailed calculation described above unnecessary.
As a result of DOE and EPRI extended burnup programs, along with fuel - assembly design improvements, future relaxation of the burnup limit is = expected. L L 4.2.4 Reactivity Coefficients and Defects Reactivity coefficients define the reactivity insertion for small changes in reactor parameters such as moderator temperature, fuel tem-perature, and power level. These parameters are input to the safety analysis and used in modeling the reactor response during accidents and M transients. Whereas reactivity coefficients represent reactivity ef-I fects over small changes in reactor parameters, reactivity defects usually apply to reactivity inserted from larger changes typical of HFP = to HZP. An example of a reactivity deficit is the power defect from HFP to HZP used in the shutdown margin calculation. A different way of [ e looking at the terms is that the coefficient when integrated over a r given range yields the defect, or the coefficient is the partial deriv-ative of reactivity with respect to one specific parameter. = Coefficients of reactivity are calculated using PDQ@7 or EPRI-NODE. p First a nominal case is established at some reference conditions. Then one parameter of interest is varied up and/or down by a fixed amount in
; another calculation and the resulting change in core reactivity divided
, by the parameter change is the reactivity coefficient. p 4.2.4.1 Doppler Coefficient The Doppler Coefficient (DC) is the change in core reactivity produced
= by a small change in fuel temperature.
m. m E 4-8 IIE M M M EMEMMmm m a.
m -
)
The major component of the Doppler Coefficient arises from the behav-ior of the Uranium-238 and Plutonium-240 resonance absorption cross sections. As the fuel temperature increases, the resonances broaden increasing the chance that a neutron will be absorbed and thus de-creasing the core reactivity. If Case.1 represents the reference case with an effective fuel tem-perature Ti (and K1 effective) and Case 2 represents a second case where the fuel temperature has been increased or decreased by approxi-
- mately 50'F and is T2 , (and K2 effective) the Doppler Coefficient is mathematically calculated from'the following equation:
~
eff eff a D* eff* eff x Os = 4(pcm)/'F Ti-T2 ) In the final fuel cycle design, both HFP and HZP Doppler Coefficients are calculated with either EPRI-NODE or PDQ97. 4.2.4.2 Moderator Temperature Coefficient The Moderator Temperature Coefficient (MTC) is the change in reac-tivity produced by a small change in moderator temperature. In McGuire and Catawba, the average core moderator temperature increases linearly as power is escalated from 0 to 100% HFP. At 100% HFP, the core average moderator temperature is approximately 592*F. Therefore, for accident and transient analyses it is necessary to know the moder-ator temperature coefficient over a range of moderator temperatures for- CZP to HFP. 4 These analyses can be performed with either EPRI-NODE and/or PDQ97 by , effecting a change in the core average moderator temperature. Cases are run with the moderator temperature at 5*F above and at the refer-i ence temperatures. If these cases and the resulting K s are effective identified as Case 1
- (TMOD 1
, K ,gg) and Case 2 (TMOD 2 i
K*e gg), the 4-9
. .--. .. -. = ._ _ -. . - - - . . . . .. . - . . . - . . - . . - . . - . - - - . .
p: moderator temperature coefficient is calculated from the following equation: eff eff x 10s = 3p(pc )foy G Taod * !ff *Nff
< (TMODt - TMOD2 )
Since the reload core is designed with a predetermined flexibility (burnup window), the MTC is verified to be within its design limit at the short end of the wind 6w as well as at the nominal BOC burnup.
'.4'2.4.3 ' Isothermal Temperature Coefficient The fractional change in reactivity due to a small change in core tem-perature is defined as the isothermal temperature coefficient (ITC) of reactivity. This is equal to the sum of the moderator and Doppler ~
temperature coefficients and may be explicitly calculated at HZP for isothermal conditions (Tg=TMOD) by setting both the fuel and moder-ator temperatures approximately 5'F above and at the reference HZP moderator temperature. This calculation may be performed with PDQ97 and/or EPRI-NODE.
'4.2.4.4 Power Coefficient and Power Defect The power coefficient of reactivity is the core reactivity change re-sulting from an incremental change in core power level. The power de-feet is usually the total reactivity change associated with a power level change from HZP to HFP.
The power coefficient is defined by the following equation:
~
eff eff a p= K1 gf x Ka ff x 10s = ap(pcm)/% power Pt-P2 where: K1 gg is K-effective for the core at power Pt (%) Ka gg is K-effective for the core at power Pa (%) 4-10
-- . . _ . .- - . _ - _ _ . ___ ~
-Neglecting second order effects this equation is equivalent to the following:
a ATMOD ,y ATFUEL P=aTMOD AP D AP where: a THOD is the moderator temperature coefficient and a I' D the Doppler temperature coefficient. Since the power coefficient should include flux redistribution effects resulting from axial variations in burnup and isotopics as well as non-uniform fuel temperature distributions, it should be performed using a 3-D simulator with thermal hydraulic feedback. If the calcu-lation is performed using a 2-D model then it should be corrected for the 3-D effects. A typical power coefficient calculation for HFP would proceed in the following manner: The HFP case is run using EPRI-NODE and the core K,gg is calculated (K1 ). Then a second EPRI-NODE case is run with the 'II' core power level reduced 5% while holding everything else constant. The K,gg from this case, Ka,g , is used along with the results _ from the reference case to calculate the power coef ficient: K,gg :K,gg u p= K ,ggx Ka ,ggx 105= ap(pca) 1 p3 - p, IN i The power defect is calculated for use in the shutdown margin calcu-
'lation (see Section 4.3.2.2) and is the reactivity change from HZP to MFP. This calculation should be performed in three dimensions to satisfactorily model the axial flux redistribution, however, a two dimensional calculation may be performed and corrected for this flux redistribution phenomenon.
4-11
Two EPRI-N0DE or PDQ97 runs are made to calculate the power defect. The first is made at 100% HFP and the second at HZP. These calcula-tions are usually performed at B0C and EOC. Both HFP and the HZP cases should have the equilibrium xenon concen-tration corresponding to HFP. The power defect is calculated from the following equation:
~
eff eff Power Defect = x 105 = ap(pcm) K1 xK2eff HFP + HZP eff where: K effective is core Keffective at HZP and K2 ff e is core Keffective at HFP. 4.2.4.5 Miscellaneous Coefficients For reload design, certain coefficients of reactivity are not routine-ly calculated. These include moderator density coefficient, moderator pressure coefficient, and moderator void coefficient. These coeffi-cients are calculated in an analogous manner by varying the appropri-ate core reactivity parameters. 4.2.4.6 Boron Related Parameters Critical boron concentrations for varioca core conditions during cycle lifetime are calculated using PDQ97 and/or EPRI-N0DE. Table 4-4 lists conditions that critical boron concentrations and boron worths are calculated. In addition to these, an ARO critical boron letdown curve is generated for HFP EQXE. 4.2.4.7 Xenon Worth The HFP equilibrium xenon worth is calculated at BOC (4 EFPD) and at EOC. 4-12
Calculations using either PDQ97 or EPRI-NODE are performed for HFP equilibrium xenon conditions. If PDQ97 is being used, a second no xenon case is run by either zeroing out the xenon number density or zeroing out the xenon cross section. If EPRI-N0DE is being used, the power level on the xenon card can be set to zero and the time in hours set to 40.0 to obtain a no xenon concentration. The difference in reactivities between the equilibrium and no xenon cases is the xenon worth. The xenon worth is used primarily for plant operation, i.e., startup after trip, rather than as a safety parameter. 4.2.4.8 Kinetics Parameters The kinetics behavior of the nuclear reactor is often described in terms of solutions to the Inhour equation for six effective groups of delayed neutrons. Transient and accident analyses often involve ki-netic modeling of the reactor core. The rate of change in power from a given reactivity insertion can be calculated by solving the kinetics equations if the six group effective delayed neutron fractions, the six group precursor decay constants, and the prompt neutron lifetime are known. The computer codes used to calculate these parameters are PDQ@7 and DELAY. The method employed here is identical to that reviewed and ap-proved by the NRC in Reference 20 for Oconee Reload Design. PDQ07 is used to obtain spatially averaged isotopic fission rates as a function of burnup, and DELAY calculates kinetics parameters and then uses these parameters to solve the Inhour equation and thereby relate the stable reactor period to the reactivity insertion. This information is also needed for startup physics testing. Calculations are per-formed at BOL and EOL. The sum of the six groupg p effective, ef-fective, for the new reload cycle is compared to those values used in the FSAR. 4-13
'E I 1
1 4.2.5 Assessment of FFCD.
. 'Once the FFCD calculations are performed, the resultant data are as-sessed for validity and consistency with core operation requirements and safety limits.
.The validation assessment consists of two different methods:
- 1. Comparison of chlculations for reasonable agreement with previous data for similar conditions (i.e. , comparison of BOL stuck rod
_ worths for. cycle N+1 to cycle N when both cycles have similar reload designs). -
- 2. Checks of assumptions, intermediate calculations for code inputs, code input data and output are performed to insure calculational accuracy.
The assessment of consistency with requirements is also a two step.
. procedure:
- 1. The calculated results must be within bounding values as determined by the Final Safety Analysis Report, or be proven to be non-limiting by way of a separate safety analysis or safety evaluation.
- 2. Calculated results must also show that core operational requirements are satisfied.
t A complete discussion of requirements for documentation and quality assurance for aafety related calculations is presented in Section 4.8 of the Duke Power Administrative Policy Manual for Nuclear Stations 21, 4 4-14
Table 4-1 Nuclear Design Basis Data For Reload Design
- 1. Power operation mode: load follow or base load.
- 2. Vessel internal or core component modifications.
.3. Expected minimum and maximum cycle burnups.
- 4. Feed enrichment (if already contracted for).
-5. Number and design of feed assemblies.
- 6. RCS hydraulic conditions.
4-15
o Table 4-2 Shutdown Martin Calculation BOC, % ap Available Rod Worth
- 1. Total rod worth, HZP 6.46
- 2. Maximum stuck rod, NZP -1.39
- 3. Net Worth 5.07
- 4. Less 10% uncertainty .51
'5. Total available worth 4.56 Required Rod Worth
- 6. Power defect, HFP to HZP .88
- 7. Max allowable inserted rod worth .36
- 8. Flux redistribution .63
- 9. Total required worth 1.87
- Ol' . Shutdown Margin (total avail. worth minus total required worth) 2.69 NOTE: Required shutdown margin is 1.30% ap for T,y >200'F.
4-16
Table 4-3 , Maximum Fg Factors for Design DNB Assembly Type Formulation 1
-17 x 17 Standard Design (Mixed 0FA and STD) 1.49 (1 + .3 (1-P))
- 17 x 17 Optimized Design (Mixed 0FA and STD) 1.49 (1 + .3 (1-P))
17 x'17. Optimized Design (All 0FA) 1.49 (1 + .3 (1-P))
+'
1 Note:- The value of 1.49 is a Technical Specification limit on the measured rg. P is the normalized core power and is 1.0 at full power. For reload design purposes, the calculational limit for Fg N is 1.435. 4-17
. ,. , , - - . - - - , . , - - - . - - - - . , - - . . , - - - , , , - - - , , - , - - , _ , ,. .- a.n,, . - ,_ _ , , . . , - - - - - . . , , , - - , , - - - . -
Table 4-4 Boron Parameters Critical Boron - ppa HZP, ARO, BOC, No Xenon HZP, Bank D inserted, BOC, No Xenon I HZP, Bank D + C inserted, BOC, No Xenon HFP, ARO, EQXE vs exposure Boron Worth - pps/7,ap HFP, EQXZ, ARO vs. exposure HZP, NOXE, ARO vs. exposure Boron Worth Versus Baron Concentration - HZP, NOXE BOC EOC 4-18
- 5. Nodal Analysis Methodology 5.1 Purpose and Introduction Nodal analysis allows for modeling of the reactor core in three-dimensions and for performing calculations which because of either code restraints or economic restraints cannot be performed by any other means. Examples of nodal code capabilities include:
- 1. Calculations which need a three-dimensional geometry such as differential rod worths, axial xenon transients and three-dimensional power distributions.
- 2. Calculations which need a full-core geometry such as stuck and ejected rod worths.
This section addresses the role of a nodal code in performing cycle deple-tions, generating rod worth data, determining shutdown margins and shut-down boron concentrations, setting control rod insertion limits, and determining trip reactivity worths and shapes. A nodal code is also used to calculate many of the startup test parameters and core physics parameters described in Section 9 of this report. The nodal code used for McGuire and Catawba analyses is EPRI-N0DE-P. (See descriptions in Section 3 and Appendix A). This code was approved by NRC for use in Oconee reload design in Reference 4. EPRI-N0DE-P can be run with either a quarter-core or a full-core geometry. The McGuire and Catawba models utilize one radial node per assembly and twelve to eighteen axial nodes. EPRI-NODE-P radial powers are normalized to the two-dimensional PDQ97 assembly powers near the beginning of each cycle. 5.2 Fuel Cycle Depletion - Nodal Code A fuel cycle depletion is performed for each cycle using nodal analysis. The nodal radial powers are normalized to the two-dimensional quarter-core PDQ97 powers at HFP conditions with equilibrium xenon and samarium at approximately 1000 MWD /MTU (25 EFPD) into the cycle. The nodal core 5-1
i m model is then depleted from BOC to EOC in steps corresponding to 0, 150, 500,.1000, 2000, 4000.... MWD /MTU. This depletion is performed in the critical boron search mode, si;n nominal rod insertion (usually 215 SWD) and equilibrium xenon. , I History files are saved at each burnup step throughout the cycle 1 depletion. These history files contain records of the power, exposure and xenon concentration for each node in the core. As a result of the nodal core depletion, the following data is obtained:
- 1. Two and three-dimensional power distributions at each burnup step.
- 2. A boron letdown curve, i.e., critical boron concentrations as a function of burnup.
- 3. Axially-dependent parameters such as offset or axial flux difference as a function of burnup.
- 4. Assembly exposures as a function of core-averaged burnup.
- 5. History files at approximately every 2000 MWD /MTU throughout the cycle saved for later use. .
. 5.3 Rod Worth Analysis Nodal analysis is used to calculate various rod worths which require three-dimensional capabilities. These calculations include differential rod worths and integral rod worths.
5.3.1 Differential Rod Worth Analysis
, Differential rod worths are calculated as a function of rod insertion.
The differential rod worth is defined as the change it reactivity asso-ciated with a small change in rod position. This rod worth is deter-mined by running two EPRI-N0DE-P cases at different rod insertions with all other parameters held constant (power, burnup, xenon, boron) and then by dividing the reactivity difference by bank height difference. 5-2
Differential rod worths for the control banks are calculated at HZP and HFP, at BOC and EOC, and at no xenon, equilibrium xenon, and peak xenon conditions. The rod banks are inserted both sequentially and in 50% overlap. 5.3.2 Integral Rod Worth Analysis Integral rod worths are defined as the integral of the differential rod worth data. Integral rod worths are determined using EPRI-N0DE-P by summing up the reactivities resulting from the differential rod worth analysis. Total integral rod worths for a rod bank can be calculated either with a two-dimensional or three-dimensional code by subtracting the. reactivities resulting from cases where the rod bank is out and then in (other parameters held constant). However, in order to get the inte-gral rod worth as a function of rod position, i.e., the shape of the rod worth curve, the three-dimensional nodal code is used. Integral rod worth calculations for the control banks are performed at HZP and HFP, at BOC and EOC, and at no xenon, equilibrium xenon, and peak xenon conditions. The rod banks are inserted sequentially with 50% overlap. The total rod worth (ARI) is calculated at BOC, EOC, and any limiting burnup at HZP for use in the shutdown margin calculation. 5.4 Shutdown Margin Analysis
-5.4.1 Shutdown Margin Shutdown margin calculations are performed using EPRI-N0DE-P. Section 4.2.2.2 describes the general shutdown margin methodology. Table 4-2 summarizes the results of a shutdown margin calculation.
EPRI-NODE-P is used specifically to calculate:
- 1. The total rod worth (ARI) at HZP, BOC, and EOC (Item I in Table 4-2). This worth is determined by running EPRI-N0DE-P cases at ARO and ARI (with constant boron and xenon) and subtracting the reactivities.
5-3
- .~_ - - . - . -_.
n M
- 2. The maximum stuck rod worth at HZP, BOC, and EOC (Item 2 in Table 4-2). EPRI-NODE-P utilizes its full-core capabilities in determin-
- iag the worst case stuck rod. The worth of the stuck rod is deter-mined by subtracting the reactivities between two EPRI-NODE-P cases, one with ARI, the other with ARI and the stuck rod out.
1 3. The power deficit from HFP to HZP, at BOC and EOC (Item 6 in Table 4-2. 'This deficit is determined by running EPRI-NODE-P cases at HFP and NZP (with constant boron and xenon) and subtracting the , creactivities. This reactivity insertion accounts for Doppler and Moderator deficits, and for axial flux redistribution.
- 4. The maximum allowable inserted rod worth at HFP, BOC, and EOC (Item 7 in Table 4-2). This worth is obtained by reading the integral rod worth curve at the rod insertion limits (See Section 5.3.2).
5.4.2 Shutdown Boron Concentration The shutdown boron concentration is another parameter that is determined using.three-dimensional nodal analysis. Since the shutdown margin is determined based on the worst case stuck rod out of the core with ali other rods in, the full-core capability of EPRI-NODE-P is needed. EPRI-NODE-P is first used to determine the worst case stuck rod by cal-culating the worth of various rods in the core. After the worse case stuck rod is determined, an EPRI-NODE-P boron search case is performed at the ARI-stuck rod out conditions. This boron concentration is ad-
-justed based on boron worth results until the core reactivity reflects the appropriate margin (1.3% Ap for temperatures greater than 200*F, 1.0% Ap for temperatures less than or equal to 200'F). The resulting boron concentration is the shutdown boron concentration required for 1
the conditions modeled in the nodal code. This calculated boron
. concentration is conservatively increased by 100 ppe.
A shutdown boron concentration can be determined for any moderator tem-perature provided the input cross sections remain valid. Typical aver-age moderator temperatures for which shutdown boron concentrations are provided are 68'F, 200*F, 500*F, and the HZP average moderator tempera-ture (approximately 557'F). 5-4 _ _---___-_u._--_____.._
5.5 Rod Insertion Limit Assessment Control rod i.nsertion limits define how deep the control rods may be in-serted into the core during normal operation as a function of the power level. It is a Technical Specification requirement that the rods never be inserted deeper than the established limits. This analysis is usually
- a. verification that the Rod Insertion Limits from cycle N-1 are adequate for cycle N.
5.5.1 Rod Insertion Limit-Criteria The control rod insertion limits are determined based on:
- 1. Maintaining the required minimum shutdown margin throughout the cycle life,
- a. 1.3% Ap for T > 200*F
- b. 1.0% Ap for T 5 200*F
- 2. Maintaining the maximum calculated enthalpy rise peaking factor to:
AH $ 1.435 (1 + .3[1-P].). F
- 3. The worst case consequences of a Rod Ejection, Rod Drop, or Rod Misalignment accident being acceptable, i.e., verifying that re-activity insertions and hot channel factors (F q) don't exceed the currently approved accident analysis values.
5.5.2 Rod Insertion Limit - Nodal Analysis Determining cont rol rod insertion limits involves an iterative process based on satisfying the above criteria. This process begins with inser-tion limits from the previous cycle. The first requirement for insertion limits is that of satisfying the reactivity constraints, i.e., maintaining the required shutdown margin. The insertion limits from the previous cycle, along with integral rod worth curves for control bank: in 50% overlap for the current cycle, are used to calculate the maximum allowable inserted rod worth for input into the shutdown margin calculation. The shutdown margin is 5-5
calculated at BOC, EOC, anli any limiting burnup in order to determine if-the control rod insertion limits are acceptable. If the shutdown n.:. n .. margin criteria is not; satisfied, the insertion limits are adjusted until satisfactory margin is obtained. ,
-The insertion limits also have to satisfy the peaking factor cons traint's . The nodal powers are synthesized with discrete pin PDQ97 pin powers to give values of [g at various power levels from HZP to HFP.' 'The* values of [g are then compared to the Technical Specification limits. .If the Technical Specification limits are not satisfied,.the control rod insertion limits are adjusted until satis-factory values of [g are obtained. (It may be found that to satisfy the Technical > Specification limits on [g the loading pattern scheme needs to be altered. See Section 4.2).
~
p In addition to satisfying reactivity and peaking factor constraints
'during normal operation, the control rod insertion limits may need to b-be modified based on the worst case consequences of an ejected RCCA, 4
a dropped RCCA or a statically misaligned RCCA. Evaluations are per-formed with"the nodal, code to identify the worst case rod configuration
-during a withdrawal or misalignment event, that is, to identify the single-RCCA wh'ich produces the maximum [g (control rods held at inser-tion limits). The results of the three-dimensional nodal analysis with
'i these worst case rod configurations are compared to the design criteria associated with each event. The acceptability of the control rod inser-,
tion limits is dependent on the criteria being satisfied. (Sections
-4.2.2.3 and 4.2.2.4 address spec!fically ejected rod and dropped rod 4 analyses.)
\
5.6 Trip Reactivity Analysis The minimum trip reactivity and the. shape of the trip reactivity insertion ci$rve(inserte.drodworthasafunctfonofrodposition)arebothgener-
- 'ated using nodal analysis. These parameters are needed to perform the safety analysis for a loss of flow accident or an uncontrolled RCCA bank,
' ~ withdrawal or\ ejection event at power.
., . r. 1 5-6
% i
._ .. - _ n. _ _ ._ , _ , . _ . _ _ _ _ , _ _,__ _ .
3 5.6.1 Minimus Trip Reactivity The minimum trip reactivity is the minimum amount of reactivity avail-
.able to be inserted into the core in the event of a reactor trip. It is evaluated for each reload core to ensure that the previously set <
limits are still valid. The minimum trip reactivity at or near full power is calculated by sub-tracting the entire rod insertion allowance and the difference between the control rod requirements at 90% FP and 100% FP from the minimum available N-1 rod worth (most reactive rod stuck out of core). The rod insertion allowance is the amount of reactivity associated with the con-trol rod insertion limits. It is the difference in reactivity between
~
an ARO case and one with control rods at their insertion limits. The minimum trip reactivity calculation is performed at both BOC and EOC. A sample BOC calculation is shown in Table 5-1. 5.6.2 Trip Reactivity Shape The shape of the trip reactivity insertion curve defines the inserted rod worth as a function of rod position. The most limiting shape is the one which defines the minimum inserted rod worth as a function of rod position. This most limiting shape is evaluated each reload cycle to ensure that the values for the minimum inserted rod worth vs. rod position used in the safety analysis are still applicable. The most limiting trip reactivity shape typically corresponds to the most bottom-skewed axial power shape. HFP axial power distributions are examined from BOC to EOC,'with control rods at the full power rod insertion limits and the most reactive rod stuck out of the core. After the most limiting power shape is found, the N-1 control rods are inserted into the core in a stepwise manner. The results of this insertion yield the minimum inserted rod ' worth vs. rod position curve. 5.7 Assessment of Nodal Analyses Once the nodal calculations are performed, the resultant data are assessed for validity and consistency with core operation requirements and safety limits. 5-7
i 1
, l The validation assessment consists of two different methods: ;
- 1. Comparison of calculations'for reasonable agreement with previous data for similar conditions.
- 2. Checks of assumptions,' intermediate calculations for code inputs, code input'dataland' output are performed to insure calculational accuracy.
The assessment of consistency with requirements is also a two step
. procedure:
~1. .The calculated results must be within bounding values as determined by the Final Safety Analysis Repor't, or be proven to be non-limiting by wa'y o'f a separate safety analysis. or safety evaluation.
L w rg.' f 2. , Calculated results must.also show that core operational requirements are1 satisfied.
.A complete discussion of requiremente,for documentation and quality assur-
' ance for safety related calculations ,is presented in Section 4.8 of the Duke Power Administrative Policy Manual for' Nuclear Stations.21
_/ _, j' ( 'l p # A
/-
.g~,
J .p-
/
-l) / ,
./
l A 5-8
Table 5-1 BOC Trip Reactivity Calculation BOC Worths BOC Worths at 100% FP at 90% FP CRA Requiraments (% ap) (% Ap) Power Defect
- 2.12 1.96 Rod Insertion Allowance 1.13 1.13 Total CRA Requirement 3.25 3.09 BOC Worths Trip Reactivity (% Ap)
Minimum Available N-1 Rod Worth 6.18 Rod Insertion Allowance -1.13 Difference between CRA Requirements
>at 100% FP and 90% FP .16 Minimum Trip Reactivity 4.89
- The Power Defect includes doppler, variable T MOD and redistribution effects.
5-9
- 6. Calculation of Safety Related Physics Parameters 6.1 Reload Specific Safety Related Physics Parameters With a reload of fresh fuel, a reactor core's physics characteristics are altered in three major areas:
- 1. Power distribution
- 2. Control rod worths
- 3. Kinetics Each of the above has its own subset of specific parameters which are addressed individually. Also, non-fuel-related changes, such as a re-vised T,y program, are accounted for in calculations of cycle specific safety parameters. Table 6.1 lists these parameters.
This section identifies safety parameters which are examined during a reload analysis for the McGuire and Catawba cores and outlines analysis methods for determining values. 6.2 Calculational Methodology 6.2.1 Power Distributions Core power distributions are calculated in two dimensions with PDQ97 and in three dimensions with EPRI-NODE. Radial local assembly factors are derived from PDQ$7 calculations and used with EPRI-NODE nodal powers to verify that the reload core operates within design Fq and F versus power limits. AH 6.2.1.1 Radial Power Peaking PDQ97 calculations are performed for several different operating conditicas:
- 1. ARO Nominal Depletion at KFP EQXE
- 2. Control Bank D inserted - HFP
- 3. HZP Sequential bank insertions: BOL and EOL
- 4. HFP Xe, Se, or soluble boron variations from nominal 6-1
Conditions 1, 2, and 3 are primarily power peaking calculations; j while condition 4. calculations provide reactivity data for the , reactor's Operator Aid Computer (OAC). 22 code post processes PDQ97 data files to produce: The'PDQEDIT
-1. Maps of assembly power, burnup, and isotopics
- 2. Summaries of core average, maxima, and minima
-3. Assembly radial local factors
- 4. Data for the offline measured power program - (See Section 11.2.2) 6.2.1.2 Total Power Peaking The quarter core EFRI-NODE model, described in Section 3, is used with the analysis techniques of Section 5 to evaluate a broad spec-trum of power distributions. A full core model is used for evaluating non-symmetric power distributions, such as the dropped rod configura-tion. ,
Like the PDQ97 model, the EPRI-NODE model has moderator, doppler, and xenon feedbacks. Nodal powers ere multiplied by the respective assem-bly radial local factor to yield F as formulated in equation 6-1: Fq= Max (F g x RLg_
) (6-1)
Where: RLg = Ra' dial local for assembly A. NODE' F i,A = Nodal power calculated at axial location i and for assembly A. Reliability factors as described in Section 11 are applied to PDQ97 and EPRI-NODE calculations such that with a 95 percent confidence-level, 95 percent of the calculated powers will be greater than or equal to measured powers. These factors are defined as: F = 95/95 Reliability factor for F g
= 95/95 Reliability factor.for F q 6-2
An additional multiplier, F q, is applied to F to q account for manu-facturing tolerances (See Section 1.2). Power distributions which are used to verify the reload design employ the following formulations: C F =F g xF (6-2) Fq= xF qxF (6-3) 6.2.2 Control Rod Worth Individual and bank RCCA worths vary with each reload. Safety parameters which pertain to RCCA worth are:
- 1. Shutdown margin
- 2. MasJmum differential bank worth
- 3. Ejected RCCA worth
- 4. Dropped RCCA worth
- 5. Trip reactivity Control rod worths are evaluated with either PDQ97 or EPRI-N0DE.
Sections 4 and 5 describe the calculational procedure for the shutdown margin. 6.2.2.1 Shutdown Margin There are two shutdown margins which are required:
- 1. 1.3% AK/K for T,y > 200'F
- 2. 1.0% AK/K for T < 200'F avg -
The 1.3% shutdown margin is based on an anal. sis of a steam-line break accident at EOL and HZP conditions. The 1.0% margin is an industry standard. The shutdown margin is evaluated at BOL, EOL, and at any intermediate burnup where the margin is seen to be more limiting. (See Section 4.2.2.2.) i l 6-3
6.2.2.2 Maximum Differential Rod Worth The uncontrolled RCCA withdrawal accident requires an estimate of the maximum differential bank worth at HZP. Two control banks are assumed withdrawn simultaneously with no overlap. The maximum worth (pem/ inch)'is determined by a series of EPRI-NCDE calculations simulating two-bank withdrawal for various pairs of sequential RCCA banks, e.g., C and C . B C Calculations are performed at BOL and EOL. When the maximum differential bank worth has been determined, a 10% conserva-
.tism factor is then applied. This conservative worth is compared to the FSAR value.
6.2.2.3 Ejected Rod Worth The ejected RCCA worth is evaluated using EPRI-NODE with a full core model. Calculations are performed at BOL and EOL and at HFP and HZP conditions. In all calculations ~, control banks D, C, and B are inserted at their respective insertion limits, depending on core power. Due to the short duration of this accident, adiabatic conditions are assumed - no moderator or doppler feedbacks. The ejected rod is simu-lated by fully withdrawing an inserted rod from a rodded location and calculating the resultant reactivity and power distribution. The ejected rod worth is then: ej R Ap (pcm) = KA ~ A
* (~}
ej R
, K K A A
, Where:
-A~- Initial condition, e.g., HFP and BOL R - Denetes' control banks at insertion limits ej - Denotes ejected-rod K - K,ff calculated by EPRI-NODE 6-4
.=- - -
'Although Doppler feedback is not used in the power distribution calcu-lation, the Doppler Coefficient is an input to the safety evaluation
-of this accident.
6.2.2.4 LDropped Rod Worth
~The' dropped RCCA is modeled with a full core EPRI-NODE calculation.
It is assumed that no trip signal is generated, and therefore all
,other RCCA remain at their current positions. Search cases are.per-formed ~1n which single RCCAs are fully inserted into the core. :
Calculations are-performed at full power and equilibrium xenon
-with Doppler and moderator feedbacks. The results of these cases are examined for both reactivity insertion and power peaking.
- 6.2.2.5 Trip Reactivity Trip reactivity insertion and the shape of the trip curve are calcu-lated with EPRI-NODE. (See Section 5.5.)
A full core model is used to evaluate:
- 1. N-1 reactivity insertion. rate at constant acceleration.
- 2. Net reactivity insertion at trip (N-1).
The N-1 trip assumes that the most reactive control rod is stuck. The calculated trip worth is then conservatively reduced by the 100 to 90% power defect. 6.2.3 Kinetics
- Kinetics parameters of each reload design are evaluated to verify
- thatfdesign limits of the FSAR are met. Table 6-2 lists these safety parameters and the codes which are used to evaluate them. Reactivity coefficients and their respective calculational procedures have been discussed in Sections 4 and 5.
The-DELAY code, which was reviewed and approved by the NRC in Reference 4, uses data extracted from PDQ97 calculations to derive 8 , Ag, and A*. 1 The prompt neutron lifetime is calculated using the formulation: 6-5
--w-P4m, - s-_ w - - - ,e--
2 p= 1 + (6-5) I bl 2 b2 I
)
Where: 1 Vi
= 8.448 x 10" + ca (B10)
V2 = 2.2 x 105 kl * "h! 1 2 * (k/b1) X (Vh/Ka) K,ff = K1+K2 (For a two group formulation with group 2 being Mixed Number Density) An isotopic fission rate weighting procedure is used to derive the delayed neutron fractions, g, and the precursor half-lives, Ag . Reactor period and reactivity versus doubling time are calculated using the Inhour equation. DELAY calculations are performed at BOL, MOL, and EOL. Calculated
' data from DELAY are also used for measuring rod worths during startup.
6.3 Comparison of Cycle Specific Safety Related Physics Parameters After the safety parameters have been calculated for the reload' core, a comparison to FSAR or other current safety analysis values is performed. If the comparison shows that the safety related physics parameters are conservative with respect to the current safety analysis limits, then no additional safety analyses are required. However, if a safety related parameter is non-conservative, then a safety evaluation is performed to determine the need for a new safety analysis, or the core is redesigned to yield conservative safety parameters. 6-6
TABLE 6-1 Reload Safety Related Physics Earameters
- 1. Power Distribution A. Fuel assembly and fuel rod powers (Two-Dimensional, nominal conditions) - FAH B. Maximum local rod power (F )
- 1. Nominal operation - for Rod Insertion Limits
- 2. Transient - load follow, axici xenon redistribution, rod ejection
- 2. Control Rod Worths A. Individual Bank Worth B. Stuck Rod Worth C. Shutdown Margin D. Dropped Rod Worth E. Differential Worth of two banks in 100% overlap at HZP F. Trip reactivity
- 3. Kinetics A. Moderator-Temperature Coefficient B. Doppler Power Coefficient C.. Boron Worth D. .pg - Effective and A g
E. Prompt Neutron Lifetime F. Time-dependent Reactivity 6-7
TABLE 6-2 Reload Safety Related Kinetics Parameters and Computer Codes 4 Parameter Computer Code
- 1. Moderator Temperature Coefficient EPRI-N0DE or PDQ97
- 2. Doppler Only Power Coefficient EPRI-NODE or PDQ97
- 3. Total Rod Worth EPRI-N0DE or PDQ97
- 4. Maximum Differential Rod Worth of EPRI-NODE Iwo Banks Moving Together
- 5. Ejected Rod Worth EPRI-NODE
- 6. Dropped Rod Worth EPRI-N0DE
- 7. Prompt Neutron Lifetime. A* DELAY
- 8. Delcyed Neutron Fraction, S DELAY 6-8
- 7. 0-) Power Peaking Analysis 7.1 Power Peaking Criteria Once a' loading pattern has been developed, as in Section 4, using two-dimensional analyses, power peaking behavior is examined with three-dimensional analyses. The three-dimensional analysis is used to esta-blish operational power-axial flux difference limits. These limits are such that the initial conditions used in the ECCS and Loss-of-Flow transients are satisfied. Currently, the LOCA F limits and DNB limits q
of these transients are determined by Westinghouse. Similarly, the three dimensional analysis is used to establish power peaking information for setting RPS limits through the OTAT and OPAT trip functions. The fuel limits to which the design calculations are compared are centerline fuel melt and DNBR. 7.2 CAOC Power Peaking Control Local power peaking (F q ) is controlled with the Constant Axial Offset Control (CAOC) procedure 23 in Westinghouse reactors. This method min-imizes power peaking by maintaining the A0 within a band about a target value. The band width is typically +3 and -12% for reload cores. The target A0 is measured at approximately HFP, ARO EQXE conditions. The band width does not vary; however, the target A0 changes approximately
-linearly with exposure. Technical specifications allow only short dura-tion (less than one hour in twenty-four) exceptions to A0 limits. In this way, xenon ma1 distributions are minimized.
' Xenon maldistributions, combined with RCCA bank motions during power maneuvers, are the primary cause of severe F s. Typical power maneuvers q
used at McGuire and Catawba are:
- 1. ' Minimum boron ~ duty - Power maneuver is accomplished primarily by con-trol bank motion. Boron requirements are minimized to accommodate bank motion and A0 control. See Figure 7-1 for comparative plots of RCCA position, boron concentration, reactor power, and A0 during this maneuver.
7-1
- 2. Maximum return to power - A0 is maintained near appropriate upper (lower) limits such that bank motion for power change will result in the A0 staying within its band. Near the end of the cycle, when boration/ dilution abilities are limited, the core average moderator
-temperature is reduced, thus giving a net positive reactivity inser-tion to increase power. When full power is reached, the inlet tem-perature is returned to its programmed value. Figure 7-2 illustrates system parameters during this transient.
During the above power maneuvers, the A0 is controlled using the CAOC strategy.
~
7.3 Power Peaking and Verification Verification of a reload design, using 3-D calculations, consists of simulating a series of power maneuvers and evaluating the margin to the design F where: q (yDesign ,pT) Q Q Margin - Design 7 Q and: FDesign = design limit Fq. F = calculated F with all allowances and 9 conservatismh Power maneuvers are performed at BOL and EOL, usually with three separate power levels from 100% power and subsequent power ascension. Load follow maneuvers are simulated, using a series of EPRI-NODE calcul-
'ations. Initial HFP conditions are assumed. Core power is reduced via a
_ step change in control bank location, along with a simultaneous step power change. A series of one-hour tiresteps are used to calculate and update xenon, Iodine, and core power distributions. Core power is maintained at a reduced level until peak xenon conditions are reached, typically 6 to 9 hours. Figure 7.1 qualitatively displays important core parameters during the power maneuver. 7-2
The method ofqF calculation is identical to that for the LOCA total peak, as described in Reference 3. It is assumed that assembly radial local factors are_ constant during the transient. Allowances are made for:
- 1. -Model uncertainty
- 2. Engineering hot channel factor-
- 3. Power spike 4
- 4. Quadrant tilt
- 5. Others, as applicable The nodal powers and PDQ$7 radial local factors are combined in the NODE Utility Code 24 to calculate margins to the LOCA F . Core height-dependent q
LOCAqF limits are presented in Table 7-1. Fq margins are evaluated at:
- 1. Transient initiation - control bank insertica
- 2. Peak xenon - after bank withdrawal
- 3. Xenon undershoot F
q margins for the power maneuver in Figure 7-1 are shown on Table 7-2, where the McGuire 2 Cycle 1 core was analyzed. 7.4 Advanced Maneuverability To improve the core's power flexibility, the CAOC " bands" from Section 7.2 can be widened.by an extensive power peaking analysis. This analysis is performed using EPRI-NODE and methods similar to the " rods-out" maneuver-ing analysis of Sections 5 and 7 in Reference 3 are performed. Unlike the Oconee methodology, penalty functions, f(AI), are used in the reactor trip system functions to prevent CFM, or DNB problems. These
' functions are parameterized to cover a wide range of operating conditions, including skewed axial power shapes. A more complete description of these functions, OTAT and OPAT, can be found in the Bases for Section 2.0 section of the McGuire Units 1 and 2 Technical Specifications.
The primary design criterion in developing the widened power - AFD operating band is that the core can be operated such that the design LOCA Kw/ft' limits are not violated. The resultant operating limits for a typical' reload core are shown in Figure 7-3. 7-3
. - - - . ~ - . .
l 7.4.1 Analysis Procedure ' Several assumptions are made in deriving the power - AFD window:
- 1. Control bank insertion as a function of power is limited to the Rod Insertion Limits.
- 2. Azimuthal-and. radial power transients are effectively dampened.
(However, allowance for quadrant power tilt is made.)
- 3. Axially skewed power distributions are limited to Condition I (normal) operation which might maldistribute the core xenon distribution. Condition II events - rod ejection, rod drop, etc., are analyzed separately in the accident analysis.
- 4. Xenon distributions are core and cycle specific. The radial xenon distribution is consistent with assembly radial power.
-At various times in reactor life, axial xenon transients are induced by either load follow transients or return to power from an equilibrium xenon state at a lower power level. Core xenon conditions include, but are not limited to: equilibrium, maximum, and minimum states.
The above states are used as input to control bank scan calculations using EPRI-NODE. The nodal powers are converted, using PDQ97 along with other factors ('F q
, quadrant tilt, FR , etc.), to F qwith appropriate conservatisms and allowances. From the control bank scan calculations at various power levels - down to 50% full power, a flyspeck plot of margins (see Section 7.3) then determines the power - AFD operating limits.
e 7-4
TABLE 7-1 Design Limit F q Fraction of McGuire McGuire Catawba Core Height First Core Transition Cores
- All Cores 00_
2.32 2.15 2.32 0.50 2.32 2.15 2.32 0.90 2.18 2.02 2.18 1.0 1.50 1.40 1.50
- These limits apply to cores with a mixture of 17x17 Standard and 17x17
- Optimized fuel assemblies. These values are subject to change pending any I
. future LOCA reanalyses.
7-5
. _. . ..- = - _ - --.. , . . ___
I TABLE 7-2 F Margin to LOCA q Power Level Minimum Margin Time (hours) (% Full Power) To LOCA (%) 1-12 100 18.27 l
-13 50 54.69 14 50 54.17 15~ 50 53.49 16 50 53.76 17 50 52.14 50 51.63
.19 50 51.27 20 50 51.08
-21 100 14.37
.22 100 15.74 23 100 16.80 24 100 16.75 25 100 15.75 26 100 15.26 27 100 15.17 28- 100- 15.29 29 100 15.62 30 100 16.14 31 100 16.78 32 100 17.39 33 100 17.99 s
. I i
( ^.. v 7-6 er * -+ce - - , w.iw -.y-am - -- -ee* vem e- --w- w - w ww
" e Design Lord Follow Mansuvar
=
99
. 19<
'REaC704
- pooge the .
4
' GO' 30' Pts FeDa 733 730-7t0<
Boa 0se CosscimfAATIO8e crease 700 Ma 05 570< Or te 12= 40menalitt0 assoog CDesCE 4f A Afl0fe 1.0 s SS 4 s i
. 2
- - - - - - - - - - -------_-n-
[: Is e. 2-4
' 00EALAfECE N 4
-4<
-10
.II.. - -. - - -----,- f.
is la a e4 < 80 < N< asamous to LOCa Fatiu ; 33 N< 28 N< y IS . . . . , , , , , , . 9 33 94 10 3 23 2e SG N N N 33 3e finns peoump 7-7
Figure 7-2 REDUCED TEMPERATURE RETURN TO POWER 100-REACTOR POWER (%) 75-50 [ 10 g T ave (OF) V RCCA 228-POSITION ( SWD) 113- / 1 I I I I O 10 20 30 40 50 60 TIME (Minutes) 7-8
FIGURE 7-3 POWER - AXIAL FLUX DIFFERENCE OPERATING LIMITS-TYPICAL (-15,102) ^
; (6,102)
- 90
- 80
- 70 RESTRICTED REGION
- 60
(-36,50) e - 50 i(30,50)
- 40
- 30
- 20
- 10
, i i i i i i i i
-50 -40 -30 -20 -10 0 10 20 30 40 50 AXIAL FLUX DIFFERENCE,% 7-9
- 8. Radial Local Analysis 8.1 ' Background
.The' local radial is an important factor in fuel cycle design because of its significant influence on LOCA and DNB analysis. The premise for performing this analysis is to evaluate the ability of PDQ97 to predict the radial local. The radial local is defined as the ratio of the maximum pin power, to the assembly average planar (x y) power.
. Duke Power. Company currently uses two computer codes to calculate radial local factors, PDQ$7 and CASMO-2. PDQ97 is a 1, 2, or 3 dimensional two neutron energy group diffusion theory code, whereas CASM0-2 is a 2-dimen-sional1multigroup transport theory code, which utilizes transport probabi-lities in.the solution of_-the' transport equation. The 2-dimensional'PDQ97 code is the primary calculational tool used to model reactor cores (for additional 'information concerning the use of this code, refer to Section 3.4).
E LEnergy_and burnup~ dependent Mixed Number Density (MND) cross sections used jf by PDQ$7 are developed in accordance with ARMP14 procedures. CASMO-2 is used primarily to' generate multigroup constants (i.e., control rod and burnable absorber cross sections), and as a benchmark code.
- 3.2 Comparison of PDQ97 to CASMO-2-at Hot Full Power Condition cThe predictive capabilityfof PDQW7 was assessed by performing a series of eighth assembly calculations using both~PDQ97_and CASMO-2. A typical 3 Westinghouse 17x17, 3.2 w/o Uranium-235 optimized fuel assembly was modeled using these codes.
~
All simulations .were performed at beginning of life (BOL), hot full power
'(HFP), no xenon conditions,=for at this time severe pin power' peaking is most prominent. Simulations were performed for a variety of burnable
?
'8-1 .
__n _____.____ - _ _ _ _ _ _ _ _
absorber loadings and soluble boron concentrations. The enrichment, burn-able absorber loadings, and boron concentration of each case investigated i
~
are representative of future'McGuire and Catawba reloads. Table 8-1 con-
-tains a summary of the cases that were investigated.
Figures'8-1-through 8-10 contain 1/8 assembly pinvise power comparisons between'PDQ97 and CASM0-2. Results from these comparisons indicate that 1PDQ97 conservatively overpredicts the maximum CASMO-2 pin power. This
- overprediction ranges from 0.86% to 2.26%. PDQ97 also correctly identi-fies-the location of the CASMO-2 maximum pin power. Comparisons between PDQ97.and CASMO-2 maximum pin powers for each case are tabulated in Table 8-2.
The global pre'dictive capability of PDQ97 was assured by performing a statistical analysis over all pins in the problem and for pins with powers greater than or equal to 1.000. The average, and the average absolute differences along with respective standard deviations, are presented in Table 8-3 for all cases investigated. 8.3 Comparisons of PDQW7 to Cold Criticals. The ability of PDQ97 to predict pin powers at cold conditions was by per-forming a series of simulations based on the B&W uranium criticals. In all simulations, PDQ97 conservatively and accurately predicted the maximum pin power. For additional specifics concerning the comparisons of PDQ97 to the B&W uranium criticals, refer to Reference 3. 4 8.4~ Conclusion , b Comparisons between PDQ97 and CASM0-2 at HFP conditions indicate that PDQ97. conservatively predicts the maximum pin power within an assembly over a wide range of burnable absorber loadings and soluble boron concen-trations. PDQW7 comparisons to B&W cold criticals indicate that PDQ97 4 also conservatively predicts ma:imum pin powers. This conservatism
' demonstrated by PDQ97 can be directly attributed to the use of MND thermal 8-2
cross sections3 . Therefore, in light of the conservatism that was demon-strated by PDQ97 over a wide range of conditions, it is not necessary to apply an uncertainty factor to the PDQ97 predicted radial local. 5 P G
- 8-3
TABLE 8-1 Characteristics of 1/8th Assembly Simulations
-ENRICHMENT PURNABLE ABSORBER BORON CONCENTRATION
- CASE. W/0 U-235 LOADING (PPMB) 1 3.2 0 0 2 3.2 0 950 3 3.2 4 0 4- 3. 2. 4 950 5 3.2 12 0 6 3.2 12 950
-7
- 3.2 16 0
-8 3.2 16 950 9 3.2 20 0
, 10 3.2 20 950 1-d 8-4
1 1 TABLE 8-2 Peak Pin Power Comparison PDQ$7 CASMO DIFFERENCE % DIFFERENCE CASE PEAK PIN POWER PEAK PIN POWER PDQ97-CASMO (P-C)/C
'l 1.053 1.042 0.011 1.056 2 1.051 1.039 0.012 1.155 3 1.055 1.046 0.009 0.860 4 1.053 1.043 0.010 0.959 5 1.152 1.131 0.021 1.857 1.137 1.119 0.018 1.609 7 1.188 1.163 0.025 2.150
.8 1.170 1.149 0.021 1.828 9 1.178 1.152 0.026 2.257 10 1.164 1.140 0.024 2.105 8-5
TABLE 8-3
-Statistica'l Summary of Percent Differences between PDQ97 and CASMO-2 For Pins in Assembles with Powers Greater Than or Equal to 1.000.
1 STANDARD CASE 5* ABS (6) DEVIATION (5) S.D [ ABS (5)] 1 0.4450 0.5566 0.5524 0.4339 2 0.4657 0.5554 0.5627 0.4697
- 3. 0.2151 0.3470 0.4098 0.3010
-4 0.2109~ 0.3595 0.4215 0.2987 5 0.8916 1.0620 0.8705 0.6396 6 0.7936 0.9548 0.7733 0.5503 l7 0.9321. 1.1509 1.0832 0.8311 8 0.8057 1.0241 0.9635 0.7107
'9 0.7130 0.8202 0.8885 0.7851 10 0.6458 0.7530 0.8109 0.7069 Statistical Summary of Percent Differences between PDQ97 and CASMO-2 For All Pins Within An Assembly STANDARD CASE' 5*- ABS (5) DEVIATION (5) S.D { ABS (5)]
-1. .0.0030 0.6395 0.7867 0.4463 2 0.0066 0.6572 0.8119 0.4648 3 -0.0255 0.4606 0.6328 0.4281 4 -0.0066 0.4520 0.6280 0.4298
.5 0.0616 1.0682 1.2511 0.6310 6 0.0394 0.9801 1.1449 0.5713 7' O.0585 0.9416 1.2120 0.7499 8 0.0398 0.8436 1.0926 0.6819 9 0.0268 0.8776 1.1696 0.7604 10' O.0293 0.8059 1.0732 0.6972
- NOTE: D = [(PDQ57 - CASMO-2)/CASMO-2] *100 N
D= I D./N i=1 1 8-6
, Figura 8-1 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 1 PDQ-7 CASMO-2 PPMB 0 0 NUMBER BA 0 0 0.0 K-INFINITY 1.3479 1.14R6 0.0
- MAX. ROD POWER 1.053 1.042 1.020 1.002 1.024 1.009 1.021 1.002 1.002 1.024 1.009 1.010
- 0. 0 1.024 1.026 o, o 0.0 1.025 1.027 o,o 1.020 1.002 1.004 1.033 1.023 1.023 1.009 1.013 1.037 1.045 1.018 1.000 1.002 1.033 1.042 0.O CASMO-2 1.021 1.007 1.011 1.038 1.053 0.O PDQ-7 0.0 1.018 1.020 0.0 1.030 1.014 0.976 0.0 1.017 1.019 0.0 1.037 1.010 0.975 1.008 0.990 0.990 1.011 0.986 0.970 0.959 0.955 1.006 0.993 0.994 1.006 0.989 0.964 0.947 0.940 0.986 0.985 0.985 0.985 0.979 0.972 0.966 0.967 0.981 0.982 0.979 0.979 0.980 0.973 0.961 0.953 0.954
, 0.971
. 8-7
. s Figura 8-2 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 2 ' CASMO-2 PDQ-7 PPMB 950 950 NUMBER BA 0 0 0.0 K-INFINITY 1 2122 1.2077 0.0
- MAX. R0b POWER 1.051 1.039 1.017 1.000 1.021 1.007 i N
'1,018' 1.000 1.000 1 1.021 1.007 1.008 '
v - O.0 % 1.021 1,'J23 00 0.0 1.022 , 1 024 o,o 1 1.'018 - 1.000 1.002 1.030 1.020 1.021 1.007 1.011 1.035 1.042 1.016 0.999 1.001 1.030 1.039 0.0 CASMO-2 1;020 1.006 1.010 1.035 1.051 00 PDQ-7 00 1.017 1.019 0.0 1.029 1.014 0.979 00 1.016 1.018 00 1.036 1.011 0.977 1.007 0.990 0.990 1.011 0.988 0.973 0.963 0.960 1.006 0.993 0.994 1.007 0.991 0.966' O.950 0.945 0.987 0.987 0.986 0.987 0.982 0.975 0.971 0.972 0.986 0.983 0.980 0.980 0.982 0.975 0.964 0.957 0.959 0.976 I C O
Figure 8-3 CASMO-2 AND PDQ-7 ROD POWER COMPARISON EOL HFP NO KENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 3 PDQ-7 CASMO-2 PPMB 0 0 NUMBER BA --' 4 '4 0.0 K.-INFINITY ~1.2978 1.2956
- 0. 0
- MAX. ROD POWER 1.055 1.046 4
1.011 0.988
~ 1.013 0.991 1.006 0.975 0.933 1.003 0.972 0.915
- 0. 0 0.989 0.893 0. 0
- 0. 0 0.977 0.890 0. 0 1.011 0.980 0.940 0.905 0.964 1.009 0.979 0.925 0.908 0.963 1.022 0.998 0.989 1.011 1.032 0. 0 CASMO-2 1.027 1.006 0.992 1.008 1.040 0. 0 PDQ-7 0.0 1.034 1.033 0.0 1.046' 1.036 1.005 0.0- 1.036 1.034 0. 0 1.055 1.039 1.011
- 1.033 1.014 1.015 1.037 1.014 1.000 0.992 0.990 1.037 1.024- 1.023 1.036 1.022 1.000 0.987 0.983 1.015- 1.015 1.015 1.016 1.011 1.005 1.002 1.004 1.019 1.018 1.015 1.014 1.016 1.010 1.001 0.996 0.999 1.018
. 8-9 . _ _ _ _ . - _ - . _ _ . _-
Figure 8-4 CASMO-2 AND PDQ-7 ROD POWER COMPARISON' BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 4 PDQ-7 CASMO-2 PPMB __ .950 _950 NUMBER BA 4 4
- 0. O K-INFINITY 1 1757 1.167'1
- 0. O
- MAX. ROD POWER 1.053 1.043 j
1.012 - 0.990j- ~ 1.014 0.993
,e ~
s1.007: 5 0.977 0.937, ' 1.00'l 0.975 C '. 919 . -
'0. Q 2
0.9 61 0.898 'o,o O.07 0.980 0.8'J6 '0.0 I.011 0.981 0.943 0.909 0.965
!!1.010, 0.98i 0512. 0.965 '
- ~ .
0.92{ [
=-
j~ , ,
. 021'--
~
0.998 0.989. 1.010 1.030 ' O.0 CASMO-2
- 1. fC6 / 1.006 0.993 Os 1.009,. 1.039 00 PDQ-7
} o e~ , w
,,, =r s 9*
,0. 0, .
1.031 1.031 0.0 1.043 - 1.034 1.005 0.0 1.033 1.'032 0.0 1.053 1.03L '1.010 J- ,
'1.031 i.013 ' 1.'013 ~. 035 1.013 1.000 0.992 0.991 1.035 1.022 If021 1.034 1.020 0.999 0.987 0.983 e ,, /
y ?', y ', 1.'013 1.013 1.013 1. 01S - 1.010 . 1.005 1.002 1.005 1.020 1.016- 1.013 1.013 1.015 ; 1.009 1.001 0.996 1.000 1.018 __ _ _ M9 W 4 M.
? H f er e J
{
-f
./
/( [
, 8-10' <
Figure 8-5 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 5 PDQ-7 CASMO-2 PPMB 0 0 NUMBER BA 12 12 0.0 K-INFINITY 1.2073 1.2029 00
- MAX. ROD POWER 1.152 1.131 1.131
- 1.108 1.152 1.132 1.123 1.099 1.088 1.144 1.123 1.109 00 1.109 1.093 0.0 0.0 1.123 1.102 0.0 ,
1.093 1.063 1.036 1.019 0.9'46 1.109 1.081 1.045 1.014 0.937 1.076 1.038 0.980 0.905 0.864 0.0 CASMO-2 1.085 1.047 0.970 0.906 0.855 0.0 PDQ-7 0.0 1.043 0.930 0.0 0.860 0.876 0.932 0.0 1.041 0.934 0.0 0.849 0.867 0.908 1.055 1.020 0.967 0.907 0.930 0.950 0.973 1.062 0.996 1.028 0.956 0.906 0.910 0.934 0.959 0.989 1.035 0.026 1.004 0.984 0.983 0.992 1.007 1.024 1.046 1.048 l 1.032 1.000 0.976 0.973 0.984 1.000 1.023 1.054 '
. 8-11 _____
Figura 8-6 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 6 PDQ-7 *CASMO-2 PPMB 950 950 NUMBER BA 12 12 0.0 . K-INFINITY 1.1010 1.0941 J 0.0
- MAX. ROD POWER 1.137 1.119 )
1.119 1.097 1.137 1.118 1.112 1.090 1.080
-1.130 1.110 1.099 00 1.100 1.086 0.0 0.0 1.112 1.094 0.0 1.086 1.058 1.032 1.018 0.949 1.100 1.074 1.041 1.013 0.941 1.070 1.035 0.981 0.909 0.871 0.0 CASMO-2 1.079 1.044 0.970 0.911 0.864 0.0 PDQ-7 00 1.041 0.933 0.0 0.868 0.884 0.938 00 1.039 0.937 0.0 0.858 0.876 0.915 1.052 1.019 0.969 0.912 ?36 0.956 0.977 0.999 1.059 1.027 0.958 0.912 0.917 0.940 0.964 0.992 1.034 1.026 1.005 0.987 0.987 0.996 1.010 1.025 1.049 1.044 1.030 1.001 0.980 0.978 0.988 1.004 1.024 1.054 t
l
, 8-12
u Figure 8-7 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 7 PDQ-7 CASMO-2 PPMB 0_ . O NUMBER BA _6 16 00 K-INFINITY "I~.1629 1.1576
- 0. 0
- MAX. ROD POWER 1.188 1.163 1.163 1.138 1.188 1.166 1.150 1.125 1.111 1.173 1.152 1.137 0.0 1.123 1.110 0.0 0.0 1.138 1.124 0.0 1.072 1.051 1.040 1.035 0.966 1.080 1.065 1.052 , 1.034 0.962 0.950' O.975 0.966 0.915 0.885 0.0 CASMO-2 0.952 0.962 0.953 0.918 0.879 00 PDQ-7
- 00. 0.909 0.904 0.0 0.882 0.905 0.966 00 0.906 0.902 0.0 0.874 0.900 0.946 0.922 0.949- 0.948 0.916 0.955 0.983 1.010 1. 's 0.919 0.934 0.933 0.918 0.937 0.971 1.001 1.6 0.992 0.993 0.995 0.996 1.009 1.027 0.989 1.046 1.066 1.093 0.993 0.993 0.993 1.004 1.023 1.046 1.071 1.105
. 8-13
- Figure 8-8 CASMO-2 AND PDQ-7 ROD POWER COMPARISLN BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 8 PDQ-7 CASMO-2 PPMB 950 950 NUMBER BA 16 16 0.0 K-INFINITY 1.0655 1.0581 . 0. 0
- MAX. ROD POWER 1.170 1.149 1.149 1.126 1.170 1.130 1.138 1.114 1.102 1.157 1.137. 1.125
. O. 0 1.113 1.102 0.0 0.0 1.127 1. 115- 0.0
-1.067 1.047 1.037 1.033 0.968 1.074 1.059 1.048 1.033 0.965
'O.952- 0.976 0.968 0.919 0.890 0.O CASMO-2 0.953 0.963 0.955 0.923 0.886 00 PDQ-7 0.0- 0.913 0.909 0.0 0.888 0.910 0.969 0.0 0.911 0.907 0.0 0.881 0.906 0.950 0.927 0.953 0.953 0.921 0.959 0.986 1.011 1.035 0.925 0.939 0.938 0.923 0.943 0.974 1.002 1.033 0.995 0.996 0.997 0.999 1.011 1.027 1.045 1.064 1.088 0.993 0.996 0.997 0.996 1.006 1.024 1.044 1.068 1.099 i
_ 8-14
Figure 8-9 CASMO-2 AND PDQ-7 ROD POWER COMPARISON BOL HFP NO XENON 3.2 W/0 U-235 OPT 17x17 FA CASE NUMBER 9 PDQ-7 CASMO-2 PPMB 0 0 NUMBER BA __ 20 20 0.0 00 K-INFINITY '1.1206 1.1148
- MAX. ROD POWER ~ 1.178 1.152 1.152 1.121 1.178' 1.148 1.132 1.094 1.036 1.151 1.112 1.034 O. 0 1.084 0.967 o. 0 0.0 1.087 0.977 0.0 1.061 1.026 0.971 0.899 0.903 1.064 1.033 0.958 0.898 0.877 0.952 0.972 0.951 0.891 0.873 0.0 CASMO-2 0.956 0.959 0.931 0.887 0.865 0.0 PDQ-7 0.0 0.923 0.916 0. 0 0.898 0.929 0.998 0.0 0.924 0.915 0. 0 0.891 0.928 0.984
=
0.948 0.976 0.975 0.943 0.985 1.017 1.048 1.077 0.950 0.964 0.962 0.948 0.971 1.010 1.046 1.084 1.025 1.027 1.029 1.031 1.045 1.065 1.087 1.109 1.136 1.027 1.030 1.031 1.032 1.046 1.069 1.095 1.124 1.160 i
. 8-15
Figura 8-10 CASMO-2 AND PDQ-7 ROD POWER COMPARISON i BOL HFP NO XENON
~
l 4 3.2 W/O U-235 OPT 17x17 FA CASE NUMBER 10 PDQ-7 CASMO-2
- i. 1
, ,, s . PPMB 950 950 0.0 NUMBER BA 20 20 0.0 . K-INFINITY 1.0315 1.0238 i
- MAX. ROD POWER 1.164 1.140 1.140 1.112 1.164 1.136 1.123 1.087 1.032 1.140 1.103 1.029 00 1.078 0.967 0.0 00 1.082 0.977 0.0 1.058 1.025 0.973 0.905 0.909 1.062 1.031 0.960 0.904 0.883 0.954 0.974 0.954 0.897 0.880 0.0 CASMO-2 0.958- 0.962 0.936 0.894 0.873 0.0 PDQ-7 00 0.927 0.921 00 0.903 0.932 0.998 i
00 0.928 0.920 00 0.898 0.932 0.984 0.951 0.978 0.978 0.946 0.986 1.016 1.045 1.072 0.953 0.967 0.966 0.951 0.973 1.010 1.043 1.077 1.025 1.027 1.029 1.031 1.044 1.062 1.082 1.102 1.028 1.127 1.031 1.031 1.031 1.044 1.065 1.089 1.115 1.149 1 i
- 9. DEVELOPMENT OF CORE PHYSICS PARAMETERS
'Upon completion of the Final Fuel Cycle Design, both PDQ97 and EPRI-N0DE depletions, boron concentrations and worths, power distributions, etc.
-have been generated primarily for HFP and.some HZP conditicas. The pur-pose of this stage of developing core physics parameters is to provide additional calculations to supplement those already performed. The re-
, sults of these calculations are used for startup test predictions and core physics parameters throughout the cycle.
-9.l' Startup Test Predictions After each' refueling, . the reactor undergoes a startup test program aimed at-verifying that the reactor core is correctly loaded and to verify reac-tor behavior is as predicted by the nuclear simulators which were used in generating the data used in the plant's safety analysis.
, 9.1.1 Critical Boron Concentrations and Boron Worths EPRI-NODE and/or PDQ97 may.be used to calculate critical boron concen-trations and boron worths at a variety of rod configurations, at HZP and HFP, as a function boron concentration, at different xenon concen-
.trations, and.at different times in the. fuel cycle. EPRI-NODE is capa-ble of critical boron searches and when critical boron concentrations are desired, it is usually run in this mode. An alternative method ic
.to correct the input boron concentration to the critical boron concen-tration using a calculated boron worth and the calculated reactivity.
PDQ97 is'usually run in.this manner to determine. critical boron concentrations. LTable.9-1 shows some of the critical boron calculations normally per-
' fonned ' for startup physics tests. These calculations are performed after the sequential insertion of each control or shutdown bank and are sometimes referred to as boron endpoints.
Critical boron concentrations at HZP and HFP with all rods out are also calculated as a function of cycle burnup. Figure 9-1 illustrates the form in which these results are displayed. These curves are referred to as boron letdown curves. l 9-1
_ _~. - _
. ~.
The boron worths are usually calculated by running two identical cases except that the soluble boron concentration is different. The difieren-tial boron worth is-calculated by subtracting the reactivities and divi-
. ding by the boros difference. Differential boron worths are usually quoted in PCM/PPMB. The inverse boron worth is the inverse of the dif-
^
-ferential-boron worth and is usually quoted in PPMB/%ap.
Table 9-2 shows the soluble boron worths usually performed for startup physics tests. Similar.t.o critical boron concentrations, these worths are calculated with sequential bank insertions. Differential boron worth (or inverse boron worth) can also be calculated as a function of boron concentration and as a function of cycle burnup. Figures 9-2 and 9-3 show the form in which results of these types of calculations are displayed. 9.1.2 Xenon Worth and Defect
-Xenon worth is calculated as a function of cycle burnup using either PDQ97 or EPRI-NODE. 'The nominal HFP depletion cases with equilibrium xenon are used as input to a second set of cases where the xenon concen-tration is. set to zero (or the xenon cross sections are set to zero).
The difference in reactivities between the equilibrium xenon and no xenon cases equals the equilibrium xenon worth at HFP. The results are displayed-in a format similar to Figure 9-4. Xenon worth can also be eresented as a function of power level. Worths presented in this manner are usually referred to as the equilibrium xenon reactivity defect and are quoted in either pcm or %Ap. Figure 9-5 shows the'results of a xenon defect calculation. 9.1.3 Rod Vorths-9.1.3.1 Group Worths The worth of the shutdown and control banks are calculated at BOC HZP for use in the zero power physics testing. The rod banks are sequen-tially inserted or withdrawn from the EPRI-NODE calculation assuming no control rod overlap. The bank worth is the difference in reac-tivity between the fully inserted case and the fully withdrawn case. i 9-2
Integral rod worth curves are calculated at BOC HZP for control banks B, C and D. The rod banks are fnserted both sequentially and with 50% overlap. Figure 9-6 shows the form in which these results are
. displayed.
Control bank worths with sequential insertion and integral rod worth curves with 50% overlap are calculated at HFP equilibrium xenon both at BOC and EOC. 9.1.3.2 ~ Stuck Rod Worth The maximum worth of a single control rod stuck out of the reactor core at HZP is calculated during the final fuel cycle design (Section 4.2.2.2). The worth of the stuck rod is used by the site engineers in the reactivity balance procedures to guarantee shutdown margin. If the stuck rod worth is to be measured during the startup test pro-gram, then a recalculation of the worth is performed rimulating the test conditions. This worth would then be provided as a startup test prediction. 9.1.3.3 Dropped Rod Worth The maximum worth of a single control rod dropped into the reactor core is calculated during the final fuel cycle design (Section 4.2.2.4). If_this parameter is to be measured during the startup test program, then a recalculation of the worth is performed simulat-ing the test conditions. This worth would then be provided as a startup test prediction.
.9.1.3.4 Ejected Rod Worth The maximum ejected control rod worth is calculated during the final
. fuel cycle design (Section 4.2.2.3). If this parameter is to be mea-sured during the startup test program, then a recalculation of the worth is performed simulating the test conditions. This worth would then be provided as a startup test prediction.
i 9-3
a e 9.1.4 Reactivity Coefficients
- 9.1.4.l' HZP Coefficients At HZP the isothermal temperature coefficient is measured by varying the average moderator temperature approximately 5*F, taking data once equilibrium is reached, then returning the temperature back to its original value, taking data and establishing equilibrium. The calcu-lations used for predicting the isothermal temperature coefficient j
.should'be run at 557*F and 562*F using either EPRI-NODE or PDQ97.
L
.Th'e resulting reactivity change is then divided by the 5*F temperature
' ~
change to yield the HZP isothermal temperature coefficient. The Doppler or fuel temperature coefficient at HZP can be calculated by varying the fuel temperature while maintaining the moderator tem-perature constant at 557'F. The resulting reactivity change divided by the change in fuel temperature is the Doppler coefficient at HZP. The predicted moderator coefficient is calculated by subtracting the Doppler coefficient from the isothermal coefficient. It is compared to the (inferred) measured moderator coefficient obtained by sub-tracting the predicted Doppler. coefficient from the measured isother-mal coefficient. Alternately, the moderator temperature coefficient can also be explic-itly calculated. 9.1.4.2 HFP Coefficients Both a temperature coefficient of reactivity and a power Doppler coef-ficient of reactivity are calculated at HFP. The temperature coeffi-cient is calculated by running one equilibrium HFP case at BOC (4EFPD EPRI-NODE or PDQ97) and a second case which has lowered the moderator temperature 5*F. The difference in reactivity divided by the tempera-ture change is the temperature coefficient. To calculate the power Doppler coefficient, a third case is performed where the power level is reduced to 95% FP. All other parameters are kept at the HFP equili-brium values. The difference in reactivity between the HFP and the 95% FP cases divided by 5% FP is the power Doppler coefficient. 9-4
9.1.5 Power Distribution Power distributions, both assembly radial and total peaking factors, are measured at various power levels as identified in the test proce-dures for McGuire/ Catawba reload startups. Calculations using EPRI-NODE are run at these power levels and nominal conditions to provide predicted power distributions for comparison. 9.1.6 Kinetics Parameters Kinetics parameters are calculated using the methodology and codes as discussed in Section 4.3.3.8. These parameters include the six group 1 I
$ effective and A , total effective and A*, and reactivity versus positive and negative doubling times. These kinetics parameters are generated for both BOC HZP and BOC HFP conditions with ARO. A second set of delayed neutron data is generated at EOC.
9.2 Core Physics Report The purpose of the core physics report is to document the predicted be-havior of the reactor -core as a function of burnup and power level. It is intended to be used for operator guidance and to aid the site engineer. Portions of the information included will reiterate data found in the nuclear design report and the startup test prediction report, however, much data not needed for these reports is useful to the operator and site engineers. This report will include sufficient information to calculate reactivity balance throughout the cycle. Table 9-3 lists items typical of what will be calculated for this report. Any additional calculations will be per-formed using- either EPRI-NODE or PDQ%7. 9-5
Table 9-1 CRITICAL BORON CONCENTRATIONS (PPMB) HZP, NOXE, $EFPD ARO Bank D in Banks D + C in Banks D + C + B in Banks D + C + B + A in Banks D + C + B + A + SE '" Banks D + C + B + A + SE+SD " Banks D + C + B + A + SE+8D+ C I" I" Banks D + C + B + A + SE+8D+ C+ B Banks D + C + B + A + SE+ D+8C+ B A HFP, N0XE, PEFPD ARO HFP, EQXE, 4EFPD ARO Bank D in HFP,EQXE,EOC ARO 4 4 9-6
Table 9-2 l l BORON WORTH (PCM/PPMD)
-HZP, NOXE, 9EFPD ARO Bank D in*
Banks D + C in Banks D + C + B in Banks D + C + B + A in ARI
-HFP, EQXE, 4 EFPD ARO HFP, EQXE, EOC ARO
. j ..
- Note: When bank worths are determined using interchange (swap) with a reference control bank, the toron worth with the reference bank only inserted is evaluated in place of sequential insertions.
9-7
's Table 9-3 Core Physics Data A. Critical Boron Concentrations
- 1. ARO HFP versus Burnup
- 2. ARO HZP versus Burnup B. Shutdown Boron Concentrations required for shutdown with highest worth rod stuck out (NoXe)
- 1. HZP versus Burnup
- 2. 500*F, 200*F and 68'F versus Burnup C. Differential Boron Worth HFP, HZP versus Burnup I
D. Power Distributions from the Cycle Depletion
- E. Rod Worths BOC, EOC, HFP and HZP F. Xenon Worth versus Power Level G.- Xenon Worth versus Burnup H Reactivity Coefficients versus Temperature, Power Level and Burnup 4
9-8
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10.0 Physics Tests Comparisons 10.1 Introduction This section presents measurement and calculational techniques and com-parisons of calculated and measured results for some key core physics parameters. The physics parameters include hot zero power (HZP) and hot full power (HFP) critical boron concentrations, HZP control rod worths , and ejected rod worths, and HZP isothermal temperature coefficients. The measured data is from the McGuire Nuclear Station, Unit 1 Cycles 1 and 1A, and Unit 2, Cycle 1. (Broken hold down springs on some Burnable Poison rods were found during an outage on McGuire Unit 1 at 191.5 EFPD. During this outage, 94 of 96 Burnable Poison Rod Assemblies were removed from the core. Cycle 1A is the continuation of Cycle 1 but without the Burnable Poison Rods.) The measurment techniques discussed are those currently used at the station. The HZP measurments were taken at beginning-of-cycle (BOC) during the Zero Power Physics Testing.- The HFP boron concentration measurements were taken at various time steps throughout the cycles. All calculations were performed with EPRI-NODE-P. The comparisons of calculated and measured results present the means of the differences between the measured and calculated data and the corre-sponding standard deviations. The mean and standard deviation are de-fined as follows: IX i Mean = x = Standard * ~ *i) Deviation
- n-I where. xg= value for the ibb observation n = number of observations.
10-1
s. 10.2 Critical Boron Concentrations
-10.2.1 . Measurement Technique Critical. boron concentrations are measured at HZP and HFP by an acid-
~
base titration of a reactor coolant system sample.
-The mea'surement uncertainty for critical boron concentrations is due to-(1)-error in the titration method and (2) error due to differences between the sample concentration and the core average concentration.
-Based on conservative estimates of these errors, the total uncertainty associated with the critical boron concentration measurements is less than 20 ppab.
-10.2.2 Calculational Technique Critical boron concentrations are calculated at HZP and HFP using EPRI-NODE-P in the boron search mode. Since the search does not yield anexactly'criticalvalue,fixedboronrunsusingEPRI-NODE-Parealso made to calculate a boron worth, which is then used to correct the cal-culated boron concentration to exactly critical.
10.2.3 Comparison of Calculated and Measured Results .
.10.2.3.1 Hot Zero Power Comparison The' calculated and measured critical boron concentrations at HZP and BOC for McGuire Unit 1, Cycles 1 and 1A, and Unit 2, Cycle 1 are com-pared in Table 10-1. Each entry corresponds to a different control rod position. The mean of the differences for these three cycles was found to be -7 ppab with a standard deviation of 16-ppsb.
10.2.3.2 Hot Full Power Comparison The calculated and measured critical boron concentrations at HFP for McGuire Unit 1, Cycles 1 and 1A, are compared in Table 10-2. The mean of the differences for these cycles is -41 ppsb with a standard deviation of 11 ppab. 10-2
4 The data displayed in Table 10-2 can be visualized better by examin-ing plots of soluble boron concentration as a function of burnup. These boron letdown curves are shown in Figures 10-1 and 10-2. + 10.2.4 Summary The comparison between EPRI-NODE-P and measured critical boron concen-trations at HZP and HFP indicate EPRI-NODE-P can adequately predict soluble boron concentrations. 4- ' 10.3 Control Rod Worth 10.3.1 -Measurement Techniques Individual control rod bank worths are measured by the boron swap tech-nique. This technique involves a continuous decrease in boron concen-tration together with an insertion of the control rods in small, dis-crete steps. The change in reactivity due to each insertion is deter-mined from reactivity computer readings before and after the insertion. The worth of each rod bank is the sum of all the reactivity changes for that bank. Measured bank worths in ppmB can be determined independent i of the reactivity computer by using the measured boron endpoints.
- 10.3.2 Calculational Techniques-Individual and total controlling rod bank worths in terms of reactivity are' calculated by making two EPRI-NODE-P runs. The first is a boron search run with the rod bank (s) out. The boron concentration found in this run is then used in a fixed boron run with the rod bank (s) in.
The difference in reactivity between these two runs with constant boron concentration is the rod bank (s) worth. Bank wor *.hs were also calculated using the calculated Boron endpoints. These bank worths are in terms of ppmB.
' 10.3.3 ' Comparison of Calculated and Measured Results A comparison of calculaced and measured control rod worths in terms of reactivity is shown in Table 10-3. This table compares the worths of control banks: D, C, B, and A and shutdown banks: E, D, and C at HZP
>. 10-3
- +- -- - . - , . . - , _ , , r - - + - - - -- ,- ,_ ,, - ,v., - - - - - - - - . , - . - - , - - - - , . - , - - , - . - - - - , ,
T1 1 I 1 1 1 and BOC for McGuire Unit 1, Cycles 1 and 1A, and McGuire Unit 2, Cycle i
- 1. A comparison of calculated and measured control rod worths in terms. of ppmB is shown in Table 10-4. This table also compares the worths of control banks: D, C, B, and A and shutdown banks: E, D, and C at HZP and BOC for McGuire Unit 1, Cycles 1 and 1A, and McGuire Unit 2 Cycle 1. Table 10-5 is a comparison of PDQ$7 calculated and measured control rod worths.
PDQ97 calculated bank worths agree well to measured with an average difference of 2.7% and a standard deviation of 3.3%. EPRI-NODE-P calculated bank worths similarly agreed well with an average difference of -4.5% and a standard deviation of 5.1%. Rod worths calculated using boron endpoints also agreed well, with an average difference of -2.2% and a standard deviation of 7.9%. 10.3.4 Summa ry The comparisons between the calculated and measured control rod worths at HZP indicate that EPRI-NODE-P can adequately predict control rod worths. Tables 10-3 and 10-4 indicate consistent agreement using either reactivity or boron endpoint measurement techniques. 10.4 Ejected Rod Worths Ejected rod worth is defined here as the measured worth of the worst case ejected rod- No error adjustments have been included. 10.4.1 Measurement Technique Ejected rod worths are measured by boron swap. The boron swap method is similar to the method used to measure control rod worth. It in-volves maintaining criticality by varying the boron concentration to compensate for the ejection of the ,orst case rod. The control rod positions are held constant. As was done for control rod worth, the ejected rod worth is determined from the reactivity computer. 10.4.2 Calculational Techniques Ejected rod worths are calculated using EPRI-NODE-P to simulate boron swap. A boron search run is first performed to determine the critical boron concentration at the rod group position. The boron concentration 10-4
as calculated in the EPRI-NODE-P run should be corrected for exact criticality. Using this corrected boron concentration and a constant rod group position, the reactivity is determined with the worst case rod first in and then out. The ejected rod worth is the difference in reactivity between the worst case rod in and out. 10.4.3 Comparison of Calculated and Measured Results A comparison of calculated and measured ejected rod worth for McGuire Unit 1, Cycle 1, is given in Table 10-6. 10.5 Isothermal Temperature Coefficients
-The isothermal temperature coefficient is defined as the change in reac-tivity per unit change in moderator temperature at hot zero power, i.e.,
"T
- ATSE 10.5.1 Measurement Techniques The isothermal temperature coefficient is measured by executing an average moderator temperature ramp to +5*F and then a ramp down to the initial equilibrium critical conditions. During each change, reactivity is measured on the reactivity computer and other pertinent data is measured. After each change, steady state conditions are established. The isothermal temperature coefficient is determined as the change in reactivity between plateaus divided by the change in tem-perature. Since two different temperature ramps are executed, two coefficients can be determined. The reported isothermal' temperature coefficient is an average of these two coefficients.
10.5.2 Calculational Technique The isothermal temperatur'e coefficient at HZP is calculated using EPRI-NODE-P. Two cases with the same boron concentration and rod positions but different moderator temperatures are run. The isothermal temperature coefficient is the difference in reactivity between the two cases divided by the difference in the moderator temperatures. 10-5
gw -
.,4 x >
g
; - ) y \_, :
% (;
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[ l '\. t *h 1 =/._ M. t 3 q g' 3 - 1 i .t ,, . (10.5.3;;,Com,parison of Calculated an,d Measured Results
- A comparison of calculated and measured isothermal temperature coeffi-s t
, cients at HZP and BOC for McGuire' Unit 1, Cycles 1 and 1A, and Unit 2,
-t , Cycle.1 is presented in Table 10-7. The mean of all the differences N i' was found.to be 1.38 pcm/*F with a standard deviation of 1.87 pcm/*F.
\. i 10.5.4 Summary I
The comparison between calculated and measured isothermal temperature 8
i- :) coefficients indicates that EPRI-NODE-P is a good predictor of isother-i
- q. mal temperature coefficients.
_('- 1 \ 1 .s
' , 'Q- y.
1 t b , ( k L. ( lt b t; i " t
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Table 10-1 MCGUIRE CRITICAL BORON CONCENTRATIONS AT HOT ZERO POWER, BOC Critical Boron Conc. PPM Unit Cycle Calculated Measured Difference 1 1 1301 1310 -9 1242 1248 -6 1123 1128 -5 1033 1029 4 972 967 5 888 891 -3 822 819 3 728 723 5 1 1A - 1269 1310 -41 1200 1242 -42 1090 1125 -35 2 1 1280 1295 -15 1221 1217 4 1101 1097 4 1002 997 - 5 944 938 6 861 860 1 788 791 -3 691 694 -3 Mean ---- ----
-6.6 - Standard Deviation ---- ----
15.7 Difference = Calculated - Measured b 10-7
. . = m-- --
y,-h.'3 .A v .- .
.4 ..
,l' r ,a/ / ' ** 'l,1 <
,e 1 , y, .- .js-, , ,
y / -
' // ..: Table 10-2
' ' I, ' '
- r. MCCUIRC 1 CYCLES 1-1A i !",.. , , HOTFULLPOWEy/ CRITICAL BORON CONCENTRATIONS
- Critical' Boron Conc. PPM Difference Unitf Er?D Calculated Measured PPM 1 -
24.6 860 880 -20 1 34,4 J 846 865 -19
- 1 5).2 / 838 864 -26
+
49.2: 823 862 -39
.82.2 761 801 -40
? 9 07.'4 , 745 790 -45
. 99.0 729 771 -42
. 101.2 724 762 -38 126.0 667 724 -57
, 154.4 600 650 -50 180.'7 531 591 -60 1A 203.7 782 831 -49 217.5 713 751 -38 227.8 673 719 -46 232.9 653 696 -43 238.5 631 677 -46 255.2 . 566 615 -49 279.8 473 511 -38 300.6 395 434 -39 330.8 281 318 -37 Mean ---- ----
-41.1
!- Standard Deviation ---- ---- 10.5 10-8
Table 10-3 MCGUIRE CONTROL ROD WORTHS AT HOT ZERO POWER, BOC Rod Worth (PCM) Unit / Cycle Bank Calculated Measured Difference (PCM) Difference (%) 1/1 CD 606 669 -63 -9.4 CC 1217 1250 -33 -2.6 CB 925 996 -71 -7.1 CA 654 695 -41 -5.9 SE -884 840 44 5.2 SD 668 755 -87 -11.5 SC 961 1011 -50 -4.9 1/1A CD 685 712 -27 -3.8 CC 1100 1038 62 6.0 2/1 CD 604 664 -60 -9.0 CC 1224 1283 -59 -4.6 CB 1004 1105 -101 -9.1 CA 618 678 -60 -8.8 SE 862 853 9 1.1 SD 738 771 -33 -4.3 SC 992 1026 -31 -3.0 Mean ----
--- . -37.6 -4.5 Standard Deviation ---- ----
43.8 5.1 Difference (pcm) = Calculated-Measured Difference (%) = Calculated-Measured x 100 Measured 10-9
s r.. Table 10-4 MCGUIRE CONTROL ROD WORTHS AT HOT ZERO POWER, BOC USING BORON ENDPOINTS Rod Worth (PPM) Unit / Cycle Bank Calculated Measured Difference (PPM) Difference (%) 1/1 CD- 59 62 -3 -4.8 I CC 119' 120 -1 -0.8 i CB 90 99 -9 -9.1 CA 61 62 -1 -1.6 SE 84 76 8 10.5 SD 66 72 -6 -8.3 SC 94 96 -2 -2.1 1/1A CD 69 68 1 1.5 CC 110 117 -7 -6.0 2/1 CD 59 78 -19 -24.4 CC 120 120 0 0.0 CB 99 100 -1 -1.0 CA 58 59 -1 -1.7 SE 83 78 5 5.4 SD 73 69 4 5.8 SC 97 97 0 0.0 Mean ---- ----
-2.0 -2.2 Standard Deviation ---- ----
6.3 7.9 10-10
Table 10-5 MCGUIRE PDQ$7 CALCULATED ROD WORTHS VS MEASUPID R0D WORTHS AT HZP, BOC Rod Worth (PCM) Unit / Cycle Bank Calculated Measured Difference (PCM) Difference (%) 1/1 D 644 669 -25 -3.7 C 1214 1250 -36 -2.9 B 962 996 -34 -3.4 1/1A D' 667 712 -45 -6.3 C 1088 1038 50 4.8 2/1 D 637 664 -27 -4.1 C 1261 1283 -22 -1.7 B 1090 1105 -15 -1.4 A 638 678 -40 -5.9 Mean ---- ----
-22 -2.7 Standard Deviation ---- ----
28 3.3 Difference (pcm) = Calculated-Measured Difference (%) = Calculated-Measured x 100 Measured y 4 10-11 3 r .
l m Table 10-6
.- MCGUIRE 1 CYCLE 1
"~
-EJECTED ROD WORTHS Worth (PCM)
Cycle Location Calculated Measured Difference (pen) 11 .D-12 406 432 -26 0
+
i
?
l 4 10-12
Table 10-7 MCGUIRE ISOTHERMAL TEMPERATURE CGEFFICIENTS AT HOT ZERO POWER, BOC Control Rod Temp. Coeff., (pcm/*F) Unit / Cycle Configuration Calculated Measured Difference (pcm/*F)
-1/1 ARO -1.03 -0.57 -0.46 Din -2.09 -2.02 -0.07 C & D in -6.03 -5.86 -0.17 B,C & D in -6.08 -6.83 0.75 A,B,C,& D in -9.37 -9.72 0.35 1/1A- 'ARO -4.51 -1.13 -3.38 Din -5.86 -1.98 -3.88 C & D in -9.76 -4.83 -4.93 2/1 ARO -2.34 -1.41 -0.93 Din -3.54 -2.73 -0.81 C & D in -7.70 -6.07 -1.63 Mean --- --- ---
-1.38 Standard Deviation --- --- ---
1.87
-Difference = Calculated-Measured T
4 t 10-13
\ l FIGURE 10-1 MCGUIRE 1 CYCLE 1 BORON LETDOWN CURUES 900 g : N_ ._ 0 3 U ! 800 P
- i N 8
R 700 N N 3 3 H : H 600. S 7 c C - P : p -
~
M ~ 500-j.................................. ......... ......... ......... ......... ......... ..... 0 20 40 60 80 100 120 140 160 180 BURHUP (EFPD)
*= CALCULATED X= MEASURED -
FIGURE 10-2 i MCGUIRE 1 CYCLE 1-A BOROH LETDOWN CURUES 900 , 0 : N L U 800 : B h % g 7" y . s a Sal x \ s R t l 0 *** j
^
C 300. ' N 33 P : j200...........................j.........,.........,.........,........., 200 220 240 260 280 300 320 340 i BURNUP (EFPD) i
*= CALCULATED X= MEASURED
,11. 0 ' POWER DISTRIBUTION C0tiPARISONS INTRODUCTION AND
SUMMARY
.11.1.1. -Introduction
-The current nuclear code employed by Duke Power Company for three d'imensional assembly power calculations is EPRI-NODE-P. This code has been benchmarked against McGuire Unit 1 Cycle 1 and part of Cycle IA.
It has also been benchmarked against TVA's Sequoyah Unit 1 Cycle 1, This work encompassed. derivation of measured power distributions for the above cycles, simulations of the above cycles using EPRI-NODE-P, development of fitting procedures for the calculated assembly peak 1. axial powers, and development of a statistical basis for estimating the calculational. accuracy of EPRI-NODE-P. 11.1.2 Summary A data base consisting of McGuire Unit 1 Cycle 1 and part of Cycle IA, and TVA's Sequoyah Unit 1 Cycle 1, measured and EPRI-NODE-P calculated fuel assembly powers was assembled. Calculated and measured powers
' were statistically combined to derive 95/95 Observed Nuclear Relia-bility Factors (ONRF) for EPRI-NODE-P. ONRF's were calculated for both assembly radial powers and assembly peak axial powers. The assembly radial power is defined as the ratio of assembly average power to core average power. The assembly peak axial power is defined as the maximum assembly x-y planar average power along the fuel assembly length relative to the core average power. Fq is then the product of the assembly radial local and peak average power (see equation 6-1).
ORNFs of 1.03 for the radial powers and 1.06 for the peak axial powers were determined. i 11-1
11.2 MEASURED DATA 11.2.1 Measured Assembly Power Data The measured power data base comprises assembly power data from McGuire Unit 1, Cycle 1 and part of Cycle IA, and TVA's Sequoyah Unit 1, Cycle 1. All measured assembly power data are directly ',caceable to signals from the incore detector system. 11.2.2 Measurement System Description The incore detector systems at McGuire and Sequoyah consist of 6 movable miniature fission chambar neutron detectors. The detectors are inserted into the bottom of the reactor vessel and driven up through the core to the top. They are then slowly withdt.swn through the core. Incore flux maps are obtained by taking voltage signal readings from the detectors as they are withdrawn through the core. This data is then stored on the plant computer. The detectors travel inside thimbles that are located in the Instrument There are 58 instrumented assem-
~
Guide Tube of the fue1 assemblies. blies out of a total of 193 fuel assemblies. There are 61 voltage signals recorded axially along each of instrumented fuel assemblies. The instrumented fuel assemblies are shown on Figure 11-1. The detectors are inter-calibrated by inserting each detector into one reference (calibration) fuel assembly. After each flux map the detec-tor signals are processed by Shanstrom Nuclear Associates Code for
- Operating Reactor Evaluation (SNA-CORE)23 SNA-CORE uses the 58 x 61 array of signals to calculate peaking factors, (radial powers and assembly peak axial powers) for each of the 193 assemblies. The 193 radial powers and assembly peak axial powers are then averaged into eighth core or quarter core, depending on the cycle. These peaking factors then make up the measured data base. All power measurements were taken at approximately equilibrium xenon conditions. Tables 11-1,
'11-2, and 11-3 show the selected reactor state points.
11-2
~. -.
11.3. EPRI-NODE-P POWER DISTRIBUTION COMPARISONS
'11.3.1 EPRI-NODE-P Model The primary three-dimensional nuclear code employed at Duke Power is EPRI-NODE-P. This code is used for all maneuvering analyses, core follow, and physics test data where three-dimensional core power distributions are required. In this section, comparisons of measured and EPRI-NODE-P calculated values will be shown for both radial powers and assembly peak axial powers. Comparisons were performed on a total of 37 reactor state points covering McGuire Unit 1, Cycle 1 and part of 1A, and Sequoyah Unit 1, Cycle 1.
McGuire Unit 1, Cycle 1 and Sequoyah Unit 1, Cycle I were modeled using eighth core symmetry. McGuire Unit 1, Cycle 1A was modeled using quarter core symmetry. Each fuel assembly was modeled with one radial and 12 equidistant axial nodes. The active stack height was set at 144 inches. ' Control rods could be positioned continuously in this model. Simulations of the McGuire and Sequoyah cores were performed using
. methods described in Section 3.5 and 5.2.
11.3.2 Fuel Cycle Simulations Using the EPRI-NODE-P model described in section 11.3.1, McGuire Unit 1, Cycles 1 and part of 1A, and TVA's Sequoyah Unit 1, Cycle I were depleted using thermal and hydraulic feedbacks. The depletions were performed in a core follow mode, utilizing critical boron searches at each exposure step. McGuire Unit-1, Cycle 1 operated until 191.5 EFPD. Control and shut-down Fsnk locations are shown on Figure 11-2. The core loading pattern is shown on Figure 11-3. During this time the unit was operated mostly at the 50% and 75% power plateaus because of power limitations imposed by steam generator flow-impingement problems. The EPRI-NODE-P radial powers were normalized to PDQ97 depletion at 25 L. EFPD for McGuire Unit 1, Cycle 1. There were 25 state points for this cycle. These are shown on Table 11-1. Figures 11-6 to 11-30 show 11-3
comparisons of calculated and measured radial powers. Figure 11-31 to 11-55 show comparisons of calculated and measured assembly peak axial powers. The data used for McGuire Unit 1, Cycle 1A was through 250 EFPD. Control and shutdown bank locations are the same as those for McGuire Unit 1, Cycle 1. The core loading pattern for cycle 1A was the same as the loading pattern for Cycle 1 excapt all but 2 burnable poison rods were removed. The two that remained were in core locations H-3 and H-13. The unit was operated mostly at 100% power during this time after the steam generator flow impingement problem was corrected. The EPRI-NODE-P radial powers were normalized to PDQ97 depletion at 257 EFPD for McGuire Unit 1, Cycle IA. There were 5 state points for the part of this cycle that was used. These are shown on Table 11-2. Figures 11-56 to 11-60 show comparison of calculated and measured radial powers. Figures 11-61 to 11-65 show comparisons of calculated and measured assembly peak axial powers. TVA's Sequoyah Unit 1, Cycle 1 operated until the end of cycle which lasted 390 EFPD. Control and shutdown bank locations are shown on Figure 11-4. The core loading pattern is shown on Figure 11-5. The EPRI-NODE-P radial powers were normalized to PDQ$7 depletion at 25
-EFPD for Sequoyah Unit 1, Cycle 1. There were 7 state points for this cycle. These are shown on Table 11-3. Figures 11-66 to 11-72 show comparison of calculated and measured radial powers. Figures 11-73 to 11-79 show comparison of calculated and measured assembly peak axial' powers.
11.3.3: Radial Power Methodology The radial powers are radial peaking factors. Therefore, the radial peaking factors from SNA-CORE are compared directly to the normalized radial powers-(P(I,J)) from EPRI-N0DE-P. 11-4
11.3.4 Assembly Peak Axial Power Methodology The assembly peak axial powers are peaking factors. There are 61 assembly axial powers for each fuel assembly calculated by SNA-CORE. Of these 61 assembly axial powers, the maximum is chosen for the
" measured" assembly peak axial power. The EPRI-NODE-P model calculated 12 nodal axial powers per assembly. The assembly peak axial power could not be compared directly to the maximum nodal power.
Therefore, the nodal axial powers were curve fit using the following equation: 3 P (z) = I A Sin (nnz) + B,Cos (nnz) n=1 Where: A , B, = Fourier series coefficients z = normalized vertical axis variable n = Fourier sequence number The 12-level node powers were fit, yielding 61 assembly axial powers for each assembly at each state point. The assembly peak axial power was then selected from the 61 calculated assembly axial powers and the 12 nodal powers. 11.3.5 Conclusions EPRI-NODE-P yielded consistently good power distributions when compared to measured power distributions. This conclusion applies for both radial and assembly peak axial power comparisons. Although the conclu-sions in this section are qualitative, quantitative statistical results of these comparisons will be shosu in Section 11.5. 11.4 PDQ97 - POWER DISTRIBUTION COMPARISONS Radial power distributions from the PDQ97 depletions of McGuire Unit 1,
~ Cycle 1, Cycle IA, and Sequoyah Unit 1, Cycle I were compared to measured radial power distributions from SNA-CORE at various burnups.
The PDQ97 model employed a 2-dimensional geometry with two neutron energy groups. (For additional information concerning the use of this code, refer to Section 3.4). All power distributions from PDQ97 11-5
were performed at hot full power all rods out. Table 11-4 compares the state points of the measured data to that of PDQ97. Figures 11-80 to 11-86 show the comparisons of the radial powers. I 11.5. STATISTICAL ANALYSIS 11.5.1 Observed Nuclear Reliability Factor Derivation This section will address quantitatively statistics arising from Section 11.3. Normal distribution theory will be used in deriving calculational uncertainties. In deriving the calculational uncertainty for EPRI-N0DE-P, the algebra-ic difference between a calculated and a measured value forms a normal-ly distributed (refer to Section 11.5.2) random variable. The difference variable is defined: Dg =C g -M g (11-1) where: D is the i difference; 1<i<N th C is the i calculated value (radial or assembly peak axial power) M is the l' measured value (radial or assembly peak axial power) The mean of the difference as defined in equation'11-2 is: 5 = 5 -' E (11-2) n where: C = (I Cg ) + n (11-2a) i=1 n H =-(I M ) + n (11-2b) i=1 n 6 = (I Dg) + n (11-2c) i=1 a = number of observations in sample 11-6
Now a one sided upper bound factor is derived by employing One Sided Upper To.lerance Limit (OSUTL) methodology. For a normal random vari-able X with a sample mean X and standard deviation S, the OSUTL of X is defined by:
'OSUTL(X) = 5 + K x S (11-3) n where: 5 = (I X.) 1 n (11-4) e i=1 n _
S = [(I (X. - X)2) + (n-1)] (11-5) i=1
- In equation 11-3, K is the one-sided tolerance factor. Equation 11-3 is formulated such that a predetermined proportion of the population (P) is b'elow the OSUTL with a confidence factor (a)as. K is explicitly dependent on n,-P, and a.
Following industry practice, P = 95% and a = 95%. 4 The OSUTL is given for D by: OSUTL(D) = 6 + K x S(D) (11-6) C is a deterministic variable and does not have an OSUTL per se, but a reasonable upper limit to C can be defined by: UL(C) = 5 + OSUTL(D) (11-7) i n UL(C) = M + 6 + K x S(D) (11-7a) If one substitutes equations 11-2 into equation 11-7 you.obtain the
' following:
UL(C) = 5 + C - 5 + K x S(D) (11-8) or UL(C).= C + K x S(D) (11-8a) I 11-7
From equation (11-8a), it is more obvious that the upper limit is a function of the calculated parameter. Also, it is obvious that the standard deviation being associated with the calculated limit is that of the difference distribution. This means that any error in the l measurement of the radial or assembly peak axial power as well as any calculational error will be included in the UL(C) parameter. While equation 11-7a and 11-8a are valid, the definition of D E C ! M (equation 11-2) lead 3 to UL(C) being smaller if the measured parameter is underpredicted. The conservative solution to this is to subtract 5 in equation 11-7a instead of adding it. This would yield the follow-ing equation: UL(C) = 5 - 5 + K x S(D) (11-9) Finally, the Observed Nuclear Reliability Factor (ONRF) is defined as the quotient of UL(C) from equation 11-9 and the mean of the measurements: UL(C) ONRF = 5 (11-10) or, M - 5 + K x S(D) ONRF = _ (11-11) M The ONRF from equation 11-11 will be used as a multiplicative factor applied to EPRI-NODE-P calculated powers such that: ONRF x C > M (11-12) for 95% of the population and with a confidence factor of 95%. Sepa-rate ONRF's are derived for radial and assembly peak axial powers. This procedure was employed in Reference 3 to statistically evaluate ORNFs for EPRI-NODE-P as part of the Oconee Reload Design Methodology. 11.5.2 Norma? ity Test Results In analyzing the normality of the difference distributions, C,M data were grouped into the following categories: 11-8
- 1) reactor cycle: McGuire 1, Cycle 1; McGuire 1, Cycle IA; Sequoyah 1, Cycle 1
- 2) grouped. cycles: All reactor cycles combined
- 3) type: radial powers or assembly peak axial powers 1ThedifferencedistributionswereanalyzedforInormalityusingtheD'
< _ test from ANSI N15.15 _1974.27- Using the engineering judgement that only peaking factors greater than the core average are the area of concern,'pairsofC,Mwherebothare}1.0 will be treated. Table 11-5 displays the normality test'results. The level of significance was chosen to'be .05. Therefore, the D' statistic must be between the
.025 and .975 percentage point D' values for normality. Here, 3 out of
- 4. assembly-radial power distributions were normal and 4 assembly peak axial power distributions were normal. The remainder of the difference distributions yielded D'. statistics that were close to the critical values and.were therefore classified as nearly normal.
- 11.5.3. Observed Nuclear Reliability Factors (ONRF) for EPRI-NODE-P 1-In-this subsection the statistical treatment developed in Section 11.5.1 will be utilized to develop ONRF's (F3 and F ) for McGuire Unit
, 1, Cycle 1 and part of Cycle IA, and TVA's Sequoyah Unit 1, Cycle 1, combined. All pairs of C, M > 1.0 from all 37 state points of McGuire Unit 1, t Cycle 1 and part of Cycle IA, and Sequoyah Unit 1. Cycle 1, were obtained. The procedure was applied to radial powers and repeated for assembly peak axial powers. The variables shown in equation 11-11 were ! then derived and the ONRF's calculated. As an example, for radial ORNF (F ): LS = 1.131 D = 0.002 S(D) = 0.020 N = 846 K = 1.7343 (N = 846, 95%/95%)
. 11-9
- . - - ~ - , - - . , - . - - - , - - - - , , , , , , . . - - --,w ,-
Therefore, the ONRF would be: ONRF = 1.131 - 0.002 + (1.7343 x 0.020) (11-13) 1.131 ONRF = 1.029 (11-13a) Table 11-6 shows the calculated ORNF's and the data used to calculate them. 11.5.4 Quantitative Comparisons of EPRI-NODE-P to Measurement By analyzing the variable D as defined in equation 11-1, the accuracy of EPRI-NODE-P can be assessed. Four important statistical prcperties of D are discussed. 6 is the mean of'the differences between EPRI-NODE-P and measured assembly powers. For McGuire Unit 1, Cycle 1 and part of 1A, and Sequoyah Unit 1, Cycle 1 6 is 0.002 for radial powers and -0.031 for assembly peak axial powers. The above means were derived from all pairs of C, M > 1.0 from all 37 state points. Subsequent statistics are also derived from this consideration. S(D), the standard deviation of the differences, indicates the spread of the values of D about 6. For the above cycles, S(D) for radial powers is 0.020. S(D) for assembly peak axial powers is 0.028. 7-1The'mean of the absolute differences. ABS (D) and its standard deviation can be combined to give limits on this variable. 95% confidence limits on the means were given by: t(.05,n) x S (ABS (D))
~ ABS (D)U,L = ABS (D) i /n (11-14)
Eqaation 11-14 yields ABS (D)g,g = 0.018 0.001 11-10
for radial powers for C, M pairs 1 1.0 for all 37 state points and:
-ABS (D)U,L = 0.036 i 0.001 for essembly peak axial powers for all C, M pairs 1 1.0 for all 37 state points.
Tables 11-7 and 11-8 present summary D statistics for radial and assembly peak axial powers, respectively, where C,M 1 1.0-for all pairs considered. 11.5.5 Relative Percent Differences The relative percent difference between EPRI-NODE-P calculated values and measured values will be defined: x 00
% Diff = M (11-15)
This section will address relative percent differences derived from: a) the sample mean b) the mean of the absolute value Since negative percent differences represent calculational nonconser-vatisms, the minimum values will be more important. Relative percent differences for all'C,M 1 1.0 will be discussed. Combining data.for McGuire Unit 1, Cycle 1, and part of Cycle IA, and Sequoyah Unit 1, Cycle 1, the following results were obtained. The average percent difference was 0.167 and the absolute 1.555 for radial powers. Also, the average percent difference was -2.195 and the absolute 2.392 for assembly peak axial powers. Table 11-9 shows summary data for percent differences derived from calculated and measured radial powers. Values are , resented by cycle
-and for all cycles combined. Table 11-10 is similar to Table 11-9 and provides data for assembly peak axial power percent differences.
11-11
t
- 11.5.6 Conclusions A statistical analysis of EPRI-NODE-P calculated and plant measured i power distributions has been performed. The resulting ONRF's for all C, M pairs > 1.0 for all 37 state points are:
Radial ONRF (F ) Assembly Peak Axial ONRF ( ) 1.03 1.06 These values while based upon calculations and measurements performed on-McGuire Unit 1, Cycles 1 and part of IA, and Sequoyah Unit 1, Cycle 1 are applicable to all McGuire and Catawba units for the following reasons:
- 1. McGuire, Catawba, and Sequoyah have identical incore 4
detector systems.
- 2. All units are manufactured by the same vendor and use similar fuel.
- 3. Calculations for all units were performed using the same calculational methods and procedures. Similarly, all calculations performed for McGuire and Catawba will use the same calculational methods and procedures.
As an additional verification of the conservatism in the 1.03 radial and.1.06 assembly peak axial'ONRF's, all calculated maximum radial powers were multiplied by 1.03 and compared to measured. Similarly all calculated assembly peak axial powers were multiplied by 1.06 and compared to measured. 29 out of 843 (3.4%) radial powers exceeded the 1.03 x maximum calculated radial power. 43 out of 1038 (4.1%) assembly-peak axial powers exceeded the 1.06 x maximum calculated assembly peak axial power. Therefore, the 1.03 radial factor was satisfactory for
-the' entire population. The 1.06 assembly peak asial factor was also satisfactory for the entire population.
I l e 11-12 , l
~;.
Table 11-1 MCGUIRE UNIT 1 CYCLE 1 STATE POINTS Control Bank D Axial Offset
'Psint #. EFPD Power (%) Position (Steps) (Meas /Cale)(%)
-1 1.28 30 213 -4.67/-4.78
- 2. 5.27 30 170 -10.68/-9.20 3 7.70 48 200 -7.59/-6.83
, . 4 11.42 48 164 -11.90/-11.07 5 37.10 50 186 -8.76/-7.70
.6' 41.59 50 201 -5.56/-6.30
-7 48.75 50 201 -6.27/-6.01
'8 59.37 50 201 -5.06/-5.83
~9 75.38 50 198 -6.10/-5.86
-10 '80.46 75 213 -8.57/-6.94 11' 91.54 75 213 -7.41/-6.75 12 104.47 50 215 -4.07/-3.58 13 112.05 50 215 -1.57/-3.43 14 115.69 75 217 -5.61/-6.52 15 118.71 50 180 -8.60/-7.50 16 122.15 75 215 -5.58/-6.36 17 '130.59 75 215 -7.58/-6.17
- 18 135.44 75 215 -5.77/-5.99 19 139.82 50 180 -8.43/-6.82 20 141.52 50 215 -0.54/-2.52 21 146.01 75 215 -4.80/-5.86 22 150.19- 50 215 -0.70/-2.32 23 162.76 .50 215 -4.80/-2.33 24 173.34 50 215 -0.29/-2.27 25 . 185.58 50 215 -0.45/-2.24 11-13
/r Table 11-2 MCGUIRE UNIT 1 CYCLE 1A STATE POINTS Control Bank D Axial Offset Point # EFPD Power (%) Position (Steps) (Meas / Calc)(%)
1 - 198.66 90 217 0.73/-0.93
.2. 217.53 100 209 1.35/-5.05
.3 223.35 100 211 -3.51/-4.92
, -4 236.23 100 211 -3.44/-4.89
-5 249.75 100 221 -2.51/-3.77 i
1 4 Iv 4 1
.-I' 11-14
Table 11-3 SEQUOYAH UNIT 1 CYCLE 1 STATE POINTS 4 Control Bank D Axial Offset
?Piint #- EFPD Power (%) Position (Steps) (Meas / Calc)(%)
1 71.82 100 200 -7.31/-9.01 2- 101.62 100 218 -4.36/-6.19 3 133.29 100' 216 -3.95/-5.60
~4 166.04 100 210 -2.68/-5.51 5 231.70 '100 216 -1.36/-3.77
-6 290.04~ 100 216
-1.51/-3.40 7 .378.92- 100 222 -1.43/-2.86 I,. '
e 1 11-15
Table 11-4 Mr.GUIRE UNIT 1 CYCI.ES 1 AND 1A AND SEQUOYAH UNIT 1 CYCLE 1 STATE POINTS FOR PDQS7 CAI/;ULATED AND MEASURED DATA PDQ$7 Calculated' Measured Control Bank D Control Bank D Paint # Unit Cycle ;Burnup Position (Steps) Power (%) Burnup Position (Steps) Power (%) 1 M1 1 52.2 228 100 48.8 201 50 2 M1 1 104.4 228 100 104.5 215 50 3 M1 1 156.7 228 '100 150.2 215 50- . 4 MI 1A 208.9 228 100 198.7 217 90 5 S1 1 103.6 228 100 101.6 218 100 -. 6 S1 1 155.5 228 100 133.3 210 100 7 SI 1 362.7 228 100 378.9 222 100 F
--__u-
y
' Table 11-5
. DIFFERENCE DISTRIBUTION NORMALITY TESTS FOR C,M > 1.0 - 5% LEVEL OF SIGNIFICANCE Assembly Radial Powers Unit / Cycle! N D'(P=.025) D' D'(P=.975) Remarks M1/C1 510 3215.0 3274.7 3275.0 Normal M1/ CIA- 190 725.9 746.0 748.1 Normal S1/C1 146 487.6 491.9 504.6 Normal
'All Combined. 846 6886.7 7000.9 6986.2 Nearly normal Assembly Peak Axial Powers Unit / Cycle N D'(P=.025) D' D'(P=.975)' Remarks
-M1/C1 642 4546.4- 4586.3 4621.7 Normal M1/ CIA 220 904.9 922.9 930.5 No rmal SI/Cl- 176 646.4 646.4 666.9 Normal
- All Combined 1038 9345.5 9379.5 9489.8 Normal 11-17
T .s Table 11-6 Calculated ORNFs and Associated Data Assembly Radial Power ORNF (F ) 5= = 1.131-5-
= -0.002 S(D) = 0.020 N = 846
- K '= 1.7343 (N=846, 95%/95%)
ORNF (F R) = 1.029 Assembly Peak Axial Power ORNF ( ) 5 ~ = .1' . 375 - 5 = -0.031 S(d) ='O.028 N- = 1038-
- K_ =.1.7259 (N=1038, 95%/95%)-
ORNF_( ) = 1.058 r - . 18
( +
\
t s , Table 11-7 DIFFERENCE MEANS AND STANDARD DEVIATIONS FOR ASSEMBLY RADIAL POWERS (C,M >1.0) Unit / Cycle N_ D S(6) ABS (D) S(ABS (6) M1/C1 510 -0.001 0.019 0.017 0.008 N1/ CIA 190 -0.001 0.025 0.023 0.010 SI/C1 146 0.013 0.012 0.014 0.010 All Combined 846 0.002 0.020 0.018 0.010 l l l 11-19
Table 11-8 DIFFERENCE MEANS AND STANDARD DEVIATIONS FOR ASSEMBLY PEAK AXIAL POWERS (C,M >1.0) Unit / Cycle N_ D S(6) ABS (6) S(ABS (5)) M1/C1 642 -0.029 0.027 0.032 0.023 N1/ CIA 220 -0.039 0.033 0.041 0.029 S1/C1 176 -0.028 0.026 0.031 0.023 All Combined 1038 -0.031 0.028 0.036 0.025 11-20
. Table 11-9 PERCENT DIFFERENCE MEANS (C,M > 1.0) - ASSEMBLY RADIAL POWERS Unit / Cycle .Mean % Difference e
Mean Absolute % Difference MI/C1 -0.058 1.452 M1/ CIA 0.007 2.043
= SI/C1 1.163 1.281 All: Combined 0.167 1.555 t
% 9 %
$~
f:. i-l 4 , 11-21
Table 11-10 PERCENT DIFFERENCE MEANS j (C,M >'l'.0) - ASSEMBLY PEAK AXIAL POWERS Unit / Cycle Mean % Difference Mean Absolute % Difference
M1/C1 -2.001 2.196
- M1/ CIA -2.838 3.031 S1/C1' -2.099 2.310
- All Combined -2.195 2.392 i
( t O
, e
h i 0 '
- s s
a p-i 11-22 s, I & , _ ,
Figure 11-1 Instrumented Fuel Assemblies McGuire and Sequoyah X 1 X X 2 X X X 3 l X X I X X X -X X S x X x x i 6 X X X X X
': 7 X X X X ~8' X X X X X. X X X y -
Y
<;9- X X X X
-10 X X X 11
.X X X X l X N21-X X X
$3 X X X X L $4 X X X X 5- X X
- i i A l
I R P 'N M L K J H G F E D C B A i (( m '
V Figure 11-2 Control and Shutdown Bank Locations McGuire 1 Cycle 1 X 1
-2 S B C B S A A 3 S D B B C 4- S D S A E A 5 S S c p 6 B C A C B 7 S 8 B- B
- 8. C- S A D A S E E W.- Y 9' S B
8 B 10' B C A C B , 'll- S D C 12 - - S D S A E A
'13 S 8 C -B B D
'14 S B C B S A A 15-
<< t a
'R P N M' L K J H G F E D C B A
Eigure 11-3
, Core Loading Pattern McGuire 1 Cycle 1 X
1 10 10 to C C C C C C C 2 9 12 20 19 12 9 C C C A C A C A C C C 3 9 20 16 16 16 20 9 C C B A B A B A B A B C C 4 20 20 16 16 20 20 C B B B A B A B A B B B C 12 20 16 16 16 20 12 5 C C A B A B A B A B A B A C C 6 10 16 16 20 20 16
-C 16 10 A B A B A B A B A B A B A C 7' _ 20 16 20 20 20 16 20 C C A B A B A B A B A B A C C 8- 10 16 16 20 20 16 10 C A B A B A B A B A B A 16'I B A C w Y 9- _ 20 16 20 20 20 16 20 C C A B A B A B. A B A B A C C 10- 10 16 16 20 20 -16 16 10 C A B A B A B A B A B A B A C 01: 12 20 16 16 16 20 12 C C A B A B A B A B A B A C C 32 20 20 16 16 20 20 C B B B A B A B A B B B C 33 - 9 20 16 16 16 20 9 C C B A B A B A B A B C C l'
04 ~ 9 12 20 19 12 9 C C C A C A C A C C C 1 l 10 10 10 05 C C C C C C C R P N M L K J H G F E D C B A A Region 1 (2.1 w/o) C Region 3 (3.1 w/o) 5 Region 2 (2.6 w/o) Number indicates number of burnable poison rods m _ _ _ _ - _ _ - _ _ _____ _
Figure 11-4 Control and Shutdown Bank Locations Sequoyah l Cycle 1 4 8 1 2 S B C B S A S S S S D B n C 4- S D D D SA A
-5 Sc D 6 .C A C B B
-7 S B B
'8 .C D A D A D C Y
f 'V 9' Sg S 3 LIO B C A C B l 11 - S ' 8 D C l [12 S D D D S A A 13 S 8 C B B D
'14 ~ S B C B S 9- A I
. IS l t 2 R .P N. M L K J H G F E D C B A 9
Figure 11-5 Core Loading Pattern Sequoyah 1 Cycle 1 X l' 10 10 10 C C C C C C C 9 9 12 12 9 9 2 C C B A B A B A B C C i 9 16 16- 20 16 16 9 C C B A B A B A B A B C C i
, 4- 16 16 16 16 16 16
'C B A B A B A B A B A B C 9 16 20 20 20 16 9
-5 C A B A A B B B A B A B A B C
- 6 10 16 20 20 20 20 16 10 C A B A B A B A B A B A B A C 7
12- 16 20 12 20 16 12 C B A B A B A A A B A B A B C 0 A A B A A A A A B A A g. 12 16 20 12 20 16 12 C B A- B A B A A A B A B A B C 10 10 16 20 20 20 20 16 10 _C_ A B A B A B A B A B A B A C 11 9 16 20 20 20 16 9 C B A B A <B A B A B A B A B C 12 16 16 16 16 16 16 C B A B A B A B A B A B C
-9 16 16 20 16 16 9 13 ~ C C B A B A B A B A B C C 1
84 9 9 12 12 '9 9 C C B A T A B A B C C 10 10 10 85- C C C C C C C
'R P N M .L K J H G F E D C B A Region 1 (2.1 w/o)
[ Region 3 (3.1 w/o) Y Region 2 (2.6 w/o) Number indicates number of burnable poison rods
Figure 11-6 MCGUIRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 1.28 EFPD 301FP CONTROL BANK D AT 213 STEPS WITHDRAUN 8 F E D C B A i H seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseessessessessess e 1.03 e .94
- 1.11 e 1.10
- 1.21
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- 1.06 * . 73*
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- 1.10 e 1.12
- 1.20
- 1.09
- 1.05 e .73
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- 1.18 e 1.11
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- 1.14
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- 12
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i 11-28
~~
Figure 11-7 MCSUIRE-1 CY-1 ASSENBLY RABIAL POWERS CALCULATED VS NEASURED 5.27 EFPD 301FP CONTROL BANK D AT 170 3TEPS WITHDRAUN N 8 F E D C B A seeeeeeeeeeeeeeeeeeeeeeec++eeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**ees
* .97 * .93
- 1.i2
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- 1.08
- 1.01
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- 9
- 1.06
- 1.03 + 1.18
- 1.16
- 1.17
- 1.05 * .89 *
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,
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- CALCULATED 13 * .98 * .48
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Figure 11-8 l MC6blRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 7.70 EFPS 481FP CONTROL BANK D AT 200 STEPS WITHDRAUN E D C B A N 8 F seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeaseeeeeeeeeeeeeeeeeeeeeecesseseese**e
* .97
- 1.14
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Figure 11-9 MCGUIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCULATER VS MEASURED 11.42 EFPD 481FP CONTROL BANK B AT 164 STEPS WITHDRAUN H 6 F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeenseeeeesesees***e**** e .99 .* .97
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- 1.19
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*
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- 1.13 * .90 * .83
- 12
- 1.12 * .88 * .82 *
*
- e e seeeeeeeeeeeeeeeeeeeeeessesewee
* .99 * .47
- CALCULATED 13 * .94 * .44 e MEASURED e e e esseeeeeeeeeese$ese$e
'11-31
Figure 11-10 MCSUIRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 37.10 EFPD 501FP CONTROL BANE B AT 184 STEPS WITHDRAWN N 8 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese+eesseesses
- 1.07
- 1.03 e 1.17 e 1.16
- 1.23
- l.10
- 1.04 * .71
- 8 e 1.10 e 1.07
- 1.18 e 1.18 e 1.23
- 1.10
- 1.03 * .70 e e e e o e e e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eo***
- 1.14
- 1.08 e 1.22
- 1.15
- 1.17 e 1.00 * .77
- 9
- 1.15
- 1.11 e 1.21
- 1.17
- 1.14 * . 1.01 e .76
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eeeeeeeeeeeeeeeee*ese***ees e 1.20 e 1.15
- 1.19
- 1.07
- 1.01 * .44
- 10 e 1.21
- 1.19
- 1.18
- 1.09 * .98 * .45
- e e e e e *
- esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eeeeee e 1.19 e 1.05 e 1.09 e .98 * .54 e 11 e 1.20
- 1.05 e 1.06 e .97 * .55
- e e e e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeee**eeseeeeeeeee***ese***
- 1.17 * .92 * .81
- 12
- 1.15 e .90 e .81
- e e e e seeeeeeeeeeeeeeeeeeeeeeessessee e .98 * .47
- CALCULATED 13 e .94 * .46
- MEASURED e e e eteeeeeeeeeee$e444 44 O
L1-32
i Figure Il-11
'MCSUIRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 41.59 EFPD 501FP CONTROL BANK D AT 201 STEPS WITHDRAWN H 8 F E D C B A seeeeeeeeesseesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesoseseee************ees
- 1.10 e 1.04
- 1.18 e 1.16
- 1.22
- 1.10
- 1.03 * .70 e 8
- 1.10 * '1.08 e 1.18
- 1.21
- 1.22
- 1.11
- 1.01 * .70
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessese**
e 1.15
- 1.08
- 1.22
- 1.15 e 1.16
- 1.00 * .76
- 9
- 1e15
- 1.12
- 1.21
- 1.18
- 1.14
- 1.01 * .75
- e e e e i e * *
*eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesse***
e 1.20 e 1.15
- 1.19
- 1.07
- 1.00 * .44
- 10
- 1.19
- 1.19
- 1.18
- 1.09 * .96 o .44
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee***ese*** -
- 1.19
- 1.06 e 1.09 * .97 * .54
- 11 e 1.19
- 1.08
- 1.07 * .96 e .54
- e e *
- e o seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*ese
- 1.20 * .93 * .81 e 12 e 1.18 * .91 * .80
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* .98 * .47
- CALCULATED 13 * .94 * .46
- MEASURED e *
- esseeeeeeeeee*1eeee4e 11-33
Figure 11-12 NCSUIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCUL ATED VS MEASURED 48.75 EFPS 501FP CONTROL BANK D AT 201 STEPS WITHDRAUN H S F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.11 e 1.05
- 1.18 e 1.17 e 1.22 e 1.10
- 1.03 * .70
- 4 8
- 1.11
- 1.09
- 1.19
- 1.21
- 1.22
- 1.12
- 1.01 * .70 *
- e-
- e e e *
- e seeeeeeeeeeeeeeeeeeeeasseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.15
- 1.09
- 1.22 e 1.15
- 1.16
- 1.00 * .76
- 9
- 1.15
- 1.12 e 1.21
- 1.19
- 1.14
- 1.01 * .75
- e e e- e e o e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.20
- 1.14 e 1.19
- 1.07 e 1.00 e .44
- 10
- 1.19 e 1.19 e 1.18 e 1.09 * .96 * .44 *
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.19
- 1.07
- 1.09 * .97 * .54 6 11 e 1.19
- 1.08
- 1.07 * .95 * .54
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeses
- 1.20 * .93 * .81
- 12
- 1.18 * .91 * .80
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeese
- j. * .98 * .47
- CALCULATED 13 * .94 e .46
- MEASURED e
- e eeeeeeeeeeeeeeeeeeeee 1]-34
Figure 11-13 MCSUIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCLILATED VS MEASURED 59.37 EFPS 501F? CONTROL BANK D AT 201 STEPS WITHDRAWN H S F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.12
- 1.07
- 1.10 e 1.17 e 1.22
- 1.10 e 1.02 * .69
- 8 e 1.12
- 1.10
- 1.19
- 1.21
- 1.22
- 1.12 e 1.01 e .70
- e e e e e
- e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesseees
- 1.16
- 1.10
- 1.22
- 1.16
- 1.15
- 1.00 * .75
- 9
- 1.16
- 1.13
- 1.21
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* * *
- e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**es**e*
- 1.20
- 1.16
- 1.19
- 1.07 * .99 * .63
- 10
- 1.19
- 1.19 e 1.17
- 1.09 * .94 * .44 *
*
- e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.19 e 1.07 e 1.08 * .97 * .54
- 11
- 1.18
- 1.07 e 1.06 * .95 e .54 *
- e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.19 * .93 * .80
- 12
- 1.17 * .92 * .80 e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeee.
* .97 * .44
- CALCULATED l
13 e .95 * .46
- NEASURED s e e seeeeeeeeeeeeeeeeeees I
i
Figure 11-14 t NC6UIRE-1 CV-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 75.38 EFPD 501FP CONTROL BANK D AT 198 STEPS WITHDRAWN H 8 F E D C B A seeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.12 e 1.09
- 1.19
- 1.18
- 1.21 e 1.11 e 1.02 * .49
- 8
- 1.12
- 1.11
- 1.18
- 1.21
- 1.20
- 1.13
- 1.00 * .49
- e o e e e e e *
- eseeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees.
- 1.17
- 1.12
- 1.22
- 1.16
- 1.15
- 1.00 * .75
- 9 e 1.15
- 1.14
- 1.20
- 1.19
- 1.13
- 1.02 * .74 *
.* e e * * *
- e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.21 e 1.17
- 1.18
- 1.07 * .98 * .63
- 10
- 1.19
- 1.20
- 1.17
- 1.10 * .96 * .64 *
*
- e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.19 e 1.08
- 1.08 * .94 * .53
- 11
- 1.18 e 1.01
- 1.06 e .95 * .54 *
- e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese
- 1.18 * .93 * .80
- 12
- 1.14 * .92 * .80
- e e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
* .94 * .46
- CALCULATED 13 * .94 * .46 e NEASURED e e e seeeeeeeeeeeeeeeeeeee s
e 11-36
__ _. ~. Figure 11-15 MCGUIRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED YS MEASbRED 80.44 EFPB 751FP CONTROL BANK D AT 213 STEPS WITHDRAWN N 6 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee****eeseeeeeeeeeeeeeeeeees e 1.17
- 1.11
- 1.20 e 1.19 e 1.21
- 1.10
- 1.01 * .48
- O e 1.14
- 1.14 e 1.20
- 1.22 e 1.20
- 1.11
- 1.00 * .70
- e e e e e e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18
- 1.13 e 1<22
- 1.17
- 1.14 * .99 * .74 *
, 9 e 1.17 e 1.15
- 1.21 e 1.19
- 1.12
- 1.01 * .74
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.21 e 1.18
- 1.18 e 1.07 e .98 * .63
- 10
- 1.19
- 1.19
- 1.17
- 1.09 * .95 * .64 *
- e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 e 1.09 e 1.07 e .95 * .53
- 11
- 1.18
- 1.09 e 1.05 * .95 e .53 e o e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.19 * .93 e .79
- 12
- 1.17 * .92 e .80
- e e e e seeseeeeeeeeeeeeeeeeeeeeeeeeees a .95 e .46
- CALCULATED 13 e .94 * .44
- MEASURED e e e seeeeeeeeeeeeeeeeeee*
11-37
Figure 11-16 4 NC8UIRE-1 CV-1 ASSEMBLY RABIAL POWERG CALCULATED VS NEASURED 91.54 EFPD 75ZFP CONTROL BANK D AT 213 STEPS WITHDRAWN N 8 F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee................. e 1.17
- 1.12 e 1.20
- 1.19
- 1.20
- 1.11
- 1.00 * .68 e 8 e 1.15
- 1.15 e 1.20
- 1.22
- 1.19
- 1.11 * .99 e .70 *
* * *
- e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeet. ...
e 1.19
- 1.14
- 1.22
- 1.17
- 1.14 * .99 * .73
- 9
- 1.17
- 1.16
- 1.20
- 1.19
- 1.11
- 1.01 * .73
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee,
- 1.21 e 1.18
- 1.18
- 1.08 e .97 * .43
- 10
- 1.19
- 1.21 e 1.14
- 1.10 e .94 e .43
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees.
- 1.19
- 1.09
- 1.07 e .95 e .53
- 11
- 1.18
- 1.11
- 1.05 * .95 * .53
- e e e e e e l seeeee.eeeeeeeeeeeeee..e..eeeeeeeee......ee.. ee.ee
- 1.10 * .94 e .79
- 12
- 1.17 e .93 * .79
- e e e 4
! eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .94 * .46
- CALCULATED 13 e .93 * .44
- NEASURED l
i e e e seeeeeeeeeeeeeeeeeese l l 1 11-36
l Figure 11-17 l l 1 l NC6UIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCULATEB VS MEASURED 104.47 EFP3 50EFP CONTROL BANK D AT 215 STEPS WITHDRAWN H 8 F E D C 3 A esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesteeseeeeeeeeeeeeeeeeeeeeeeeeeeseeeseesseees
- 1.16
- 1.12
- 1.19 e 1.18
- 1.19
- 1.10
- 1.00 e .48
- 8
- 1.15
- 1.14
- 1.19
- 1.22 e 1.19
- 1.12 * .99 * .69 *
*
- e e e *
- e e e*eeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eese**e
- 1.17
- 1.13
- 1.20
- 1.16 e 1.13
- 1.00 * .74
- 9 e 1.17
- 1.15
- 1.19
- 1.19
- 1.11 e 1.02 * .73
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeee
- 1.20
- 1.18
- 1.17 e 1.08 e .97 * .43
- 10
- 1.18
- 1.19
- 1.16 e 1.10 * .95 * .44 *
- a e e e *
- eeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**ese**ee '
- 1.18
- 1.10 e !.07 * .96 * .53
- 11
- 1.16
- 1.10
- 1.06 * .95 e .53 *
- e
- e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseesse e 1.20 * .95 * .80
- 12
- 1.18 * .94 e .80 *
- e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeese
* .96 * .46
- CALCULATED 13 e .94 * .46
- MEASURED e *
- eseeeeeeeeeeee***esee 4
s 6
Figure 11-18 l l MC8UIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCULATED VS MEASURED l l 112.05 EFPB 501FP CONTROL BANK D AT 215 STEPS WITHDRAWN H 8 F E 8 C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.eeeeeeeeeeeeeeeesseees.**esese e' 1.16
- 1.12
- 1.19
- 1.18 e 1.19 e 1.10
- 1.00 * .68
- 8
- 1.14
- 1.14 e 1.18
- 1.21
- 1.19
- 1.12 * .99 * .70
- e e e e e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee***eseseeeeeeeesseeeeeeeeses
- 1.17
- 1.14 e 1.20
- 1.16 e 1.13
- 1.00 * .74
- 9
- 1.16
- 1.15
- 1.19
- 1.19 e 1.11
- 1.01 * .74
- e e e e a e e 4 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.19 6 1.18
- 1.17
- 1.08 * .97 * .63 e 10 e 1.18 e 1.20
- 1.15
- 1.10 e .95 * .44
- e e e e e e e seeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.18
- 1.10 e 1,07 * .97 * .53
- 11
- 1.17
- 1.10
- 1.06 * .95 * .53
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese
- 1.20 e .96 * .80
- 12
- 1.18 * .94 e .80
- e e e e f
seeeeeeeeeeeeeeeeeeeeeeeeeeeees e .96 * .46 e CALCULATED 13 e .95 * .47 e MEASURED e e e seetteeeeeeeeeeeeeeee i t 11-40*
I-Figure 11-19 i j i NC8UIRE-l CY-1 ASSEMBLY RABIAL POWERS CALCULATED VS MEASURED 115.49 EFPS 751FP CONTROL BAMM B AT 217 STEPS WITHDRAUN N O F E D C 3 A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeee*****
- 1.10 e 1.14
- 1.20 e 1.19
- 1.19 e 1.11
- 1.00 e .48 '-
O
- 1.14
- 1.14
- 1.20 e 1.22
- 1.19
- 1.12 * .99 * .49
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee>eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 e 1.15
- 1.21
- 1.17
- 1.13
- 1.00 * .73
- f e 1.10
- 1.14
- 1.19 e 1.19 e 1.11
- 1.01 * .73
- e e e o e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.20 e 1.18 e 1.17 e 1.08 e .97 e .43 e 10 e 1.19
- 1.20
- 1.15 e 1.10 e .95 * .44
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.18
- 1.10
- 1.07 * .95 e .53 e 11 e 1.17 e 1.10 e 1.05 e .95 e .53
- e
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees*****
- 1.18 e .95 * .79 e 12
- 1.17 e .94 * .80
- e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ,
* .94 e .44
- CALCULATED 13 e .94 e .44
- MEASUREB e e e eseeeeeeeeeeeeeeeeees I
l i 11-41
Figure 11-20 NC8UIRE-1 CY-1 ASSEMBLY RADIAL POWERS CALCULATED VS MEASURED 118.71 EFPS 501FP CONTROL BANK B AT 180 STEPS WITHDRAUN N 8 F E B C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees l e 1.09 e 1.11 e 1.19
- 1.20 e 1.20
- 1.12 e 1.02 e .70 e '
8 e 1.08 e 1.13 e 1.18
- 1.23 e 1.19 e 1.14 e 1.01 e .72 *
*
- e e e e e e e I eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees* eeeeee e 1.17
- 1.15 e 1.21 e 1.17 e 1.14 e 1.02 * .75
- 9 e 1.15
- 1.14
- 1.19 e 1.20
- 1.12 e 1.04 e .75 e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseseeeeeeeeeeeeeeeee
- 1.20 e 1.18
- 1.17
- 1.09 * .98 * .43
- 10 e 1.18 e 1.20 e 1.15 e 1.11 e .96 * .45
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17 e 1.08 e 1.04 e .94 e .53
- 11 e 1.14 e 1.09
- 1.04 * .94 * .54
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.12 e .93 * .79 e 12 e 1.10 e .92 * .79 e e e o e eseeeeeeeeeeeeeeeeeeeeeeeeeeee.
l * .94 e .44
- CALCULATED 13 e .92 e .44 e HEASURED e e e ,
$$4seeeeeeeeeeeeeeeee l'
l l J 11-42
t i i Figure 11-21 MC8UIRE-1 CV-1 ASSEN8LY RASIAL POWERF, CALCULATED VS MEASURED 122.15 EFPS 751FP CONTROL BANK B AT 215 STEPS WITHDRAWN N 8 F E 3 C 3 A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessessee**ee
- 1.17 e 1.14
- 1.20
- 1.19 e 1.19 e 1.11
- 1.00 o .68
- 8
- 1.15 e 1.14
- 1.19
- 1.22 e 1.18 e 1.12 e .99 * .70
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19
- 1.15
- 1.20 e 1.17 e 1.13 e 1.00 * .73
- f
- 1.17
- 1.17
- 1.18
- 1.19 e 1.10
- 1.02 * .73
- e e e e e e e e eeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.20 e 1.18 e 1.17 e 1.08 e .97 e .43 e 10 e 1.18
- 1.21
- 1.15
- 1.10 * .94 e .44 *
*
- e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18
- 1.10
- 1.07 * .96 * .53 e 11 e 1.17
- 1.11 e 1.05 * .95 * .53
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.18 e .95 e .7f e 12 e 1.17 * .94 e .79
- e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees
* .94 * .44
- CALCULATED 13 * .93 * .47
- MEASURED e e e seeeeeeeeeeeeeeeeeeee
_ _ . . _ , . . _ - . _ _ .__.- __ _. _ _ _ _ _ _ 015 1---
FIGURE 11-22 NC8UIRE-1 CV-1 ASSEMBLY RASIAL P99ERS CALCULATED VS NEASURED 130.59 EFPS 751FP CCNTROL BANK 3 AT 215 STEPS WITHDRAWN N 8 F E 3 C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeees e 1.17 e 1.15 e 1.19 e 1.19 e 1.10 e 1.11
- 1.00 * .68
- 8 e- 1.14 e 1.14 e 1.18 e 1.22 e 1.17 e 1.13 e .99 e .70
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeee**
e 1.19 e 1.15
- 1.20 e 1.17
- 1.12
- 1.00 e .73
- f
- 1.17 e 1.17 e 1.18
- 1.18
- 1.10
- 1.01 * .73 e ;
e e e o e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.20 e 1.18
- 1.14
- 1.00 e .97 e .43
- 10 e 1.18 e 1.20
- 1.14
- 1.10 * .f4 * .64
- e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17
- 1.11
- 1.07 e .f4 e .53
- 11
- 1.17
- 1.11 e 1.05 e .95 e .53
- e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.18 e .f4 e .79 e 12
- 1.17 * .95 * -80
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .f4 e .44 e CALCULATED 13 e .f4 e .47 e NEASURED e e e eeeeeeeeeeeeeeeeeeese 11-46
' Figure 11-23 NCSUIRE-1 CY-1 ASSEMBLY RABIAL PSUERS CALCULATED VS MEASURED 135.44 EFPS 751FP CONTROL DAMM B AT 215 STEPS WITHDRAUN H 8 F E 3 C B A eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesse**eesseeeeeees
- 1.17
- 1.15
- 1.19
- 1.19
- 1.18
- 1.11
- 1.00 * .48
- 8 e 1.16 e 1.17
- 1.19 e 1.22
- 1.17 e 1.12 * .99 e .70
- e o e e e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18 e 1.14 e 1.20 e 1.17 e 1.12
- 1.00 e .73
- 9 e 1.17
- 1.17
- 1.18 e 1.19 e 1.10 e 1.02 e .73
- l e a o w e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee !
e 1.08 e .97 * .43
- 1 1.19 e 1.18
- 1.14
- 10
- 1.18 e 1.20 e 1.15 e 1.10 e .94 * .44 * )
e e o e e o e ' eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeees e 1.17 e 1.11 e 1.07 e .94 * .53 e 11 e 1.14
- 1.11 e 1.05 e .95 e .53 e e o e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee***eveees
- 1.18 e .96 * .79 *
- 12 e 1.17 * .95 * .00 *
- e e 4 eseeeeeeeeeeeeeeeeeeeeeeeeeeeee e .95 e .44
- CALCULATED 13 e .94 * .47 e MEASURED e e e seeeeeeeeeeeeeeeeeeee i
I I l i 11-45
Figure Il-24 I 9
+
1 NC8UIRE-1 CY-1 ASSENBLY RABIAL POWERS CALCULATED Vs NE?*URED J 139.82 EFPS 501FP CONTROL BANK B AT 180 STEPS UITHDRAUN N 8 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeesteeeeeeeeeeeeeeeeeeeeeeeeeeerseeeeeeeeeeeeeeeeeeeee
- 1,09 e 1.12
- 1.14 e 1.19 e 1.19
- 1.12
- 1.01 * .70
- 8 e 1.08 e 1.14 e 1.18 e 1.23 e 1.19 e 1.14 e 1.01 * .71
- e e e e e e e o e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17 e 1.15
- 1.20 e 1.17 e 1.13 e 1.03 e .75
- 9
- 1.15
- 1.17 e 1.19 e 1.20 e 1.12
- 1.04 e .75
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 a 1.18 e 1.16 e 1.09 * .90 * .44
- 10 e 1.18 e 1.20 e 1.14 e 1.11 e .96 * .45 e e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.14
- 1.09 e 1.04 e .97 e .54
- 11
- 1.15 e 1.09 e 1.04 e .95 e .54 e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.11 e .94 e .79 e 12
- 1.11 * .93 e .79 *
. e e e e j eeeeeeeeeeeeeeeeeeeeeeeeeeeeees e .94 * .44
- CALCULATED 13 * .92 * .46 e NEASURE8 e e *
......e..e...ee.....e l
l t-i m. ( l i 11-46
, - - - , - - . - , - - , - , , - - - , - . - . . _ _ - - . - - - - - -_r _ _ _..----__.----
Figure 11-25 MC6UIRE-1 CV-1 ASSEMBLY RADIAL POWERS CALCULATE 5 VS NEASURED 141.52 EFPS 54XFP CONTROL BANK D AT 215 STEPS WITHDRAUN H 8 F E 9 C 3 A , eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.16 e 1.13 e 1.18
- 1.18 e 1.17 e 1.11 e 1.00 * .49 e 8 e 1.15
- 1.14 e 1.18 e 1.21
- 1.17 e 1.12 e .99 * .70
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17
- 1.14
- 1.19
- 1.14 e 1.12 e 1.01 * .74
- 9 e 1.14
- i.14 e 1.17
- 1.18
- 1.10 e 1.02 * .74
- e e o e e e e e
- seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.18
- 1.18 e 1.14 e 1.08 * .97 * .43
- 10
- 1.17 e 1.20
- 1.14
- 1.10 * .95 * .44 *
*
- e e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeene
- 1.17 e 1.11
- 1.07 e .97 e .53 e 11
- 1.16
- 1.11 e 1.04 e .94 * .53
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.19 * .97 e .80 e 12
- 1.17 * .96 e .41
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees
* .94 * .47 e CALCULATED 13 * .95 e .47
- MEASURED e e e seeeeeeeeeeeeeeeeeeee i
i l i l 11-47
- Figure 11-26 NCSUIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCULATED VS MEASURED 144.01 EFPB 751FP CONTROL BANK 3 AT 215 STEPS WITHDRAWN N 0 F E B C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17 e 1.15 e 1.19 e 1.19
- 1.17
- 1.11
- t.00 e .48
- 8 e 1.16
- 1.18 e 1.18 e 1.22
- 1.17
- 1.12 * .98 e .70
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.10 e 1.16
- 1.15
- 1.14
- t.12
- 1.01 * .74
- 9
- 1.17
- 1.10 e 1.18 e 1.19 e 1.09
- l.02 0 .73 *
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.19 e 1.18
- 1.16
- 1.09 * .97 * .43
- 10
- 1.17 e 1.21
- 1.14 e 1.10 * .94 e .44
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese
$ l.17 e 1.11
- 1.07 * .96 * .53
- 11
- 1.14
- 1.12
- 1.04 e .95 e .53
- e e e e e e seeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeee****ee**e
- 1.18 e .94 * .79 e 12
- 1.14 e .95 e .79 e e o e e seeeeeeeeeeeeeeeeeeeeeeeesseees
- * .95 * .44
- CALCULATED ~
13 0 .93 e .47 e MEASURED e e e eeeeeeeeeeeeteseseote 11-48
l i Figure 11-27 NC8UIRE-l CY-1 ASSENBLY RABIAL POWERS CALOULATES YS MEASUREB 150.19 EFPS 501FP CONTROL BANN 3 AT 215 STEPS WITHDRAWN N 8 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.16 e' 1.14 e 1.17 e- 1.18 e 1.17
- 1.11
- 1.00 * .49 e 8 e 1.14 e 1.16
- 1.17 * -1.21
- 1.14
- 1.12 * .99 * .71
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.17
- 1.15 e 1.18
- 1.14 e 1.12 e 1.01 * .74
- 9 e 1.i5 e 1.14 e 1.16
- 1.19 e 1.10 e 1.03 * .74
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18
- 1.17
- 1.15 e 1.09 e .97 * .43
- 10
- 1.14 e 1.20
- 1.14
- 1.11 e .95 e .45 *
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.14 e 1.11 e 1.07 * .?7 * .54
- 11
- 1.14 e 1.12 e 1.05 e .94 * .54 -e o e e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 e .98 e .80 e 12
- 1.17 e .97 * .80 *
*
- e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee
* .97 * .47 e CALCULATES 13 e .95 e .48
- NEASURED e e e seeeeeeeeeeeeeeeeeeet 11-49 ,
Figure 11-28 i MCSUIRE-1 CY-1 ASSEMBLY RASIAL POWERS CALCULATED VS NEASURES 162.76 EFPB 501FP CONTROL BANK 3 AT 215 STEPS WITHORAWN F E D C 3 A N 8 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeee**e
- 1.15
- 1.14 e 1.17 e 1.18
- 1.14 e 1.11
- 1.00 * .69
- e 1.17 e 1,17
- 1.21 e 1.16 e 1.13 * .?? e .70
- 8 e 't.14 e e e o e e e e e
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.16 e 1.15
- 1.17 e 1.16 e 1.11
- 1.02 * .74 e 9
- 1.15
- 1.17 e 1.16 e 1.18 e 1.09
- 1.02 e .74
- o e e e *
- e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.17
- 1.17 e 1,15 e 1.09 e .97 e .44 e 10
- 1.16 e 1.20 e 1.13 e 1.11 e .f4 e .45 e e e e o e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.16 e 1.12 e 1.07 e .97 e .54
- 11 e 1.15
- 1.12 e 1.05 e .96 e .54 e e e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 e .98 * .00
- 12 e 1.17 e .97 e .00
- e e e e l eeeeeeeeeeeeeeeeeeeeeee**eeeees
* .97 e .47 e CALCULATED l 11 e .95 * .48
- NEA80 RED i
e e e seeeeeeeeeeeee***eeee L 11 ,50
Figure 11-29 l l MCOUIRE-1 CY-1 ASSEMBLY RABIAL POWERS CALCULATED VS MEASURED 173.34 EFPB 501FP CONTROL BANK B AT 215 STEPS WITHDRAUN 8 F E D C B A N eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.15
- 1.14 e 1.14
- 1.17
- 1.15 e 1.11
- 1.00 * .49
- 8 e 1.13
- 1.17
- 1.16
- 1.21 e 1.14 e 1.13 * .99 e .71
- e e e e e e e e e
eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.14
- 1.15 e 1.17
- 1.14
- 1.11
- 1.02 e .74
- 9
- 1.14
- 1.17
- 1.15 e 1.18 e 1.09
- 1.03 * .74 *
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.17 e 1.17 e 1.14
- 1.09 e .97 * .44
- 10 e 1.15 6 1.19 e 1.12 e 1.11
- e94 * .45
- e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.15 e 1.12
- 1.07 * .98 * .54 e 11 e 1.14 e 1.12 e 1.05 e .94 e .54 e e e e o e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18 e .99 e .81
- 12 e 1.14 e .98 e .81
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees
* .97 e .47
- CALCULATED 13 e .95 e .49 e MEASURED e e e esoteeeeeeeeeeeeeeeee i
i 11*51 +
l Figure 11-30 i l l MCSUIRE-1 CV-1 ASSEMBLY RASIAL POWERS CALCULATED VS NEASURED 185.58 EFPS 501FP CONTROL BANK 3 AT 215 STEPS WITHDRAUN 6 F E D C B A N esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeee++aee****eeeeee******eseeeee********eeeeee***** e 1.15
- 1.14 e 1.14 e 1.17 e 1.15
- 1.11 e 1.00 e .70
- O e 1.13 e 1.14 e 1.15 e 1.20 e 1.14
- 1.12 e .?? * .72
- e e a e e o e e e 1 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- l.16 e 1.15 e 1.14
- 1.15 e 1.11 e 1.03 e .75
- 9 e 1.14 e 1.16 e 1.14 e 1.18 e 1.09 e 1.04 e .74 e e e e e e e e e
j e e e e e e e eee e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e
- e 1.14
- l.17 e 1.14 e 1.09 e .97 e .44
- l 10 e 1.14
- 1.19 e 1.12
- 1.11 e .95 e .44 e e
e e e o e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessseeeeeeeeeeee*es
, e 1.15 e 1.12 e 1.07 e .f8 e .54
- l 11 e 1.14 e 1.13 e 1.04 e .97 e .54
- l e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeesse*******************
e 1.18 e .99 e .01 e 12 e 1.17 * .99 * .81
- e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee l e .97 e .48 e CALCULATED 13 e .96 e .49 e NEASURED e e e eeeeeeeeeeeeee**eeeee l
l l .l I 11-52
Figure 11-31 NCGUIRE-1 CV-1 ASSEMBLY PEAN AXIAL POWER - CALCULATED VS MEASURED 1.28 EFPD 30ZFP CONTROL BANK B AT 213 STEPS WITHIRAWN N 8 F E D C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41
- 1.27 e 1.50 e 1.48 e 1.43 e 1.44 e 1.43 e .99 e 8 e 1.40 e 1.32 e 1.50
- 1.52 e 1.42 e 1.47 e 1.42
- 1.00 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.45
- 1.34
- 1.60 e 1.50 e 1.59
- 1.34
- 1.09
- f
- 1.43
- 1.38
- 1.57 e 1.54
- 1.55 e 1.40
- 1.08
- e e e e e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.54 e 1.49
- 1.41
- 1.43 e 1.41 e .?! e 10
- l.55
- 1.54 e 1.59 e 1.49 e 1.37 * .92 e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.40 e 1.42
- 1.52
- 1.38 * .77
- 11 e 1.40
- 1.45 e 1.52 e 1.37 e .78
- e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.74 e 1.29 e 1.18
- 12
- 1.71
- 1.29
- 1.18 e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.45 * .48
- CALCULATED 13 e 1.39 e .47 e NEASUREB e e e eeeeeeeeeeeeeeeeeeeee e
4 11-53
Figure 11-32 MC8UIRE-1 CY-1 ASSEMBLY PEAN AXIAL POWER - CALCULATER VS MEASURED 5.27 EFP8 301FP CONTROL BANN 3 AT 170 STEP 5 WITH8RAWN N 8 F E B C 3 A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.50
- 1.33
- 1.54 e 1.53 e 1.48 e 1.50 e 1.47 e 1.01
- 8 e 1.48 e 1.40 e 1.58 e 1.61 e 1.71 e 1.55 e 1.47 e 1.03 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.50 e 1.39 e 1.65
- 1.55
- 1.63
- 1.40
- 1.11
- 9
- 1.51 e 1.43 e 1.65 e 1.42 e 1.62 e 1.45 e 1.12
- e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.61
- 1.54 e 1.45 e 1.47
- 1.44 * .93
- 10 e 1.42 e 1.41 e 1.44 e 1.54 e 1.42 e .95 e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**ee e - 1.45 e 1.44 e 1.54 e 1.41 * .79 e 11 e 1.47 e 1.50 e 1.57 e 1.40 e .80 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.81 e 1.33 e 1.20
- 12 e 1.76
- 1.33 e 1.21
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.47 * .49
- CALCULATES 13 e 1.43 e .49 e MEASURED e e e seeeeeeeeeeeeeeeeeees 11-54
Figure 11-33 l l i MC6UIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS NEASURED 7.70 EFPB 40XFP CONTROL BANN 3 AT 200 STEPS WITHDRAUN N 6 F E D C 3 A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.44 e 1.33
- 1.54
- 1.52 e 1.45 e 1.47 e 1.42 * .97 e 8
- 1.47 e 1.38 e 1.54
- 1.58 e 1.44
- 1.50 e 1.40 e .99 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.49
- 1.38
- 1.43
- 1.52 e 1.58 e 1.35
- 1.04
- 9
- 1.50
- 1.43
- 1.42
- 1.50 e 1.57
- 1.39
- 1.07 e e e e e o e o e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.59
- 1.52 e 1.42
- 1.43 e 1.39 * .89
- 10 e 1.58 e 1.57 e 1.41
- 1.49 e 1.34 * .91
- e e e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.41
- 1.42
- 1.51 e 1.35 * .74
- 11 e 1.42 e 1.45
- 1.51
- 1.35 e .77
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.71
- 1.27 e 1.14
- 12
- 1.71
- 1.28
- 1.14
- e e e
- seeeeeebeeeeeeeeeeeeeeeeeeeeeee
- 1.39 * .44
- CALCULATED 13 e 1.37 e .44
- MEASURED e e e seeeeeeeeeeeeeeeeeeee 11-55
Figure 11-34 O MCOUIRE-1 CY-1 ASSEMBLY PEAM AXIAL POWER - CALCULATED YS MEASURED 11.42 EFPD 481FP CONTROL PANK D AT 144 STEPS WITH9RAWN H 6 F E D C D A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.54
- 1.38 e 1.59 e 1.57 4 1.70
- 1.51
- 1.45 * .99
- 8 e 1.54
- 1.46
- 1.44 e 1.44
- 1.73 9 1.57 e 1.44
- 1.01
- e e e
- e e e o e eeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.55
- 1.43
- 1.47
- 1.57
- 1.43 * -'~? e 1.09 e 9
- 1.57
- 1.51 e 1.70
- 1.65
- 1.63
- 1.85
- 1.10 e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeees
- 1.e4
- 1.56
- 1.44 e 1.47 e 1.42 * .91 e 10 e 1.47
- 1.45
- 1.68
- 1.55 e 1.41 * .94 *
- e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.47 e 1.54 e 1.38 * .77
- 11
- 1.49
- 1.52
- 1.56
- 1.39 * .80
- e o e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeevseeeees
- 1.76 e 1.30
- 1.14
- 12
- 1.75
- 1.32 e 1.19 e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41 * .47 e CALCULATED 13 e 1.40 e .47 e NEASURED e *
- 4eeeeeeeeeeeeeeeeeeee r
i l 11-56
Figure 11-35 l l l NCSUIRE-1 CY-1 ASSEMBLY PEAN AXIAL POWER - CALCULATE 3 VS MEASURED 37.10 EFPD 50!FP CONTROL BANK B AT 184 STEPS WITHDRAWN N 8 T E D C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.52
- 1.40 e 1.57 e 1.55 e 1.43 e 1.44 e 1.37 e .93
- 8
- 1.58
- 1.50 e 1.44
- 1.42 e 1.47 e 1.50 e 1.40 e .95
- e e o e e o e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.53 e 1.44 e 1.42
- 1.53
- 1.54 e 1.33 e 1.01 e f
- I.61 e 1.53 e 1.66
- 1.60 e 1.54 e 1.38 e 1.03 e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.60 e 1.54
- 1.59
- 1.43 e 1.33 e .85
- 10
- 1.47
- 1.44 e 1.63 e 1.50 e 1.34 e .89 e e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.59 e 1.43
- 1.44 e 1.30 e .72
- 11 e 1.44 e 1.48 e 1.48 e 1.33 e .75
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.45 e 1.25
- 1.09 e 12
- 1.64 e 1.26
- 1.13 e e o e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.32 e .43 e CALCULATED 13
- 1.32 e .44
- MEASURED e e e seeeeeeeeeeeeeeeeeees
. 11-57
Figure 11-36 NC6UIRE-1 CY-1 ASSEMBLY PEAN AXIAL POWER - CALCULATES V5 NEASURED 41.59 EFPD 501FP CONTROL BANK B AT 201 STEPS WITHDRAUN H 6 F E 3 C 3 4 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.50 e 1.37
- 1.55 e 1.53 e 1.40 e 1.44 e 1.35 e .92 e B
- 1.55
- 1.47
- 1.41 e 1.44
- 1.45 e 1.51 e 1.37 e .95 e o e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.52
- 1,43 e 1.40 e 1.52 e 1.52 e 1.31 e 1.00 e f
- 1.57 e 1.52 e 1.44 e 1.40
- 1.53
- 1.37 e 1.01 *
- e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.58
- 1.52 e 1.57 e 1.41 e 1.31 e .84 e 10
- 1.62 e 1.41 e 1.40
- 1.48
- 1.30 e .84 e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeves e 1.57
- 1.42 e 1.44 e 1.20 * .71 e 11
- 1.42
- 1.48
- 1.45
- 1.29 * .73
- e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.62
- 1.24
- 1.07 e 12
- 1.64
- 1.25 e 1.09 e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.30 e .42
- CALCULATED 13
- 1.29 e .42 e MEASURED e e e eteeeeeeeeeeeeeeeeete I
f 11-58
Figure 11-37 I !~ I l NC6UIRE-t CY-1 ASSEMBLY PEAN AXIAL POWER - CALCULATED VS NEASURED 48.75 EFPD 501FP CONTROL BANN D AT 201 STEPS WITH8RAWN H 6 F E D C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.50
- 1.40 e 1.54 e 1.53
- 1.50 e 1.44 e 1.33 * .91
- 8
- 1.54 e 1.50 e 1.42 e 1.44
- 1.44
- 1.51 e 1.34 * .f4 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.51 e 1.43 e 1.59 e 1.51 e 1.50 e 1.30 * .99
- f
- 1.57
- 1.52
- 1.44
- 1.40 e 1.53
- 1.34 e 1.01 *
- e o e e e e e
, eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.57
- 1.52 e 1.55
- 1.40 e 1.29 * .83
- 10
- 1.42
- 1.41 e 1.59 e 1.47 e 1.30 e .84
- e o e e e e e eeeeeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.54 e 1.41
- 1.42 e 1.27 e .70 e 11 e 1.41
- 1.47 e 1.44 e 1.20 e .72 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeoseeeeeeeeeeeeeeeeeeeeees
- 1.40
- 1.23
- 1.04
- 12 e 1.43
- 1.25
- 1.09
- e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.20 * .41
- CALCULATES 13
- 1.29 e .42
- MEASURED e e e eeeeeeeeeeeeeeeeeeeee 11-59
Figure 11-38 NC6UIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATES VS MEASURED 59.37 EFPS 501FP CONTROL BANK 3 AT 201 STEPS WITH8AAUN N 8 F E 8 C 8 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.49 e 1.41
- 1.53 e 1.52 e 1.54
- 1.43 e 1.31 * .89
- 8 e 1.54 e 1.47 e 1.59 e 1.40 e 1.61 e 1.48 e 1.33 e .92 e e e e e e e e e e l esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.51 e 1.44 e 1.57 e 1.50 e 1.48 e 1.29 * .?7
- f
- 1.55 e 1.50 e 1.40 e 1.56 e 1.50 e 1.33 * .99
- e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.55 e 1.51
- 1.53 e 1.39
- 1.27 * .82 e 10
- 1.58 e 1.57 e 1.55
- 1.44
- 1.27 * .84 e e e e e e s e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.54 e 1.41 e 1.40 e 1.25 e .69 e 11 e 1.57 e 1.43 e 1.41 e 1.24 e .71 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.58 e 1.22 e 1.04 e 12 e 1.40 e 1.23
- 1.07
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.24 e .40
- CALCULATES 13 e 1.27 e .41 e MEASURES e o e eteeeeeeeeeeeeeeeeece a
e E 11-60
l Figure 11-39 l l NC8UIRE-1 CY-1 ASSENILY PEAK AXIAL POWER - CALCULATED VS MEASURED 75.38 EFPD 501FP CONTROL BANK 3 AT 198 STEPS WITMDRAWN H 8 F E D C 3 A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.48 e 1.41 e 1.51 e 1.51
- 1.53
- 1.41
- 1.28 * .87 e 8 e 1.51 e 1.44 e 1.54 e 1.57 e 1.55 e 1.44 e 1.29 * .90
- e o e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.49 e 1.43 e 1.54 e 1.48
- 1.45 e 1.27 * .95
- 9 e 1.51
- 1.48 e 1.54 e 1.53
- 1.44
- 1.31 * .f4
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.53
- 1.50 e 1.50
- 1.37 e 1.24 * .00
- 10 e 1.54
- 1.55 e 1.51
- 1.42 e 1.24 * .83 *
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.51 e 1.40 e 1.37 e 1.23 e .48 e 11 e 1.53 e 1.42 e 1.38 e 1.24 e .49 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.54 e 1.21 e 1.02 e 12
- 1.54
- 1.22 e 1.05 e e e e e seereeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.23 * .59
- CALCULATED 13
- 1.24 * .41
- NEASURED e e e seeeeeeeeeeeeeeeeeeee 11 61
Figure 11-40 NC6UIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASUREB i 80.44 EFPD 751FP CONTROL BANN 8 AT 213 STEPS WITHDRAWN ' H 8 F E 9 C B A seesesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.49
- 1.42 e 1.51 6 1.50
- 1.51 e 1.39
- 1.26 e .04 e 8
- 1.54
- 1.45 e 1.55 e 1.57
- 1.54
- 1.44
- 1.29 e .90
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
-* 1.49 e 1.43 e 1.52 e 1.47
- 1.43
- 1.25 e .93 e f e 1.53
- 1.40 e 1.54 e 1.53 e 1.45
- 1.30 e .94 *
*
- e o e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.52
- 1.48
- 1.48
- 1.34
- 1.22 * .79
- 10 e 1.53 e 1.53 e 1.50 e 1.41 e 1.23 e .82
- e
- e e e
- e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee****eeeeeee e 1.49 e 1.38
- 1.35
- 1.21 e .47
- 11 e 1.51 e 1.40
- 1.34 e 1.22 * .48
- e e o e e e l eeeeees.eeeeeeeeeeeeeeeeeeeeceeeeeeeeeeeeeeeeeeeeee e 1.52
- 1.~ t t s 1.00 *
! 12 e 1.53
- 1.00
- 1.04
- e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeeen
- 1.20 * .58
- CALCULATED 13 e 1.22 * .40 e MEASURED e e e seeeeeeeeeeeeeeeeeese I
4
.- w- - - - - - . , _ . . . . . _ _ , s _
Figure 11-41 l ( l NCGUIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURES 91.54 EFPD 751FP CONTROL BANN 3 AT 213 STEPS WITHDRAWN H 6 F E D C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.47
- 1.41
- 1.49
- 1.48
- 1.48 e 1.38 e 1.25 * .85
- 8
- 1.50
- 1.47 e 1.52
- I.56 e 1.50 e 1.41
- 1.24 * .89
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.47
- 1.43
- 1.50 e 1.45
- 1.41
- 1.24 e .92 e f
- 1.50
- 1.48
- 1.52
- 1.51
- 1.41
- 1.28 * .93 *
*
- e e o e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.50
- 1.47 e 1.44
- 1.34
- 1.21 e .78 e 10
- 1.51 e 1.53 e 1.47
- 1.39
- 1.20 e .50
- e e e e e
- e seeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.47 e 1.37 e 1.33 e 1.19 e .44 * ,
11
- 1.49
- 1.40
- 1.33
- 1.20 * .47 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.50 e 1.19 e .99 e 12
- 1.50 e 1.19
- 1.01 e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.19 * .57 e CALCULATED 13
- 1.19 * .59
- MEASURED e e e j eeeeeeeeeeeeeeeeeeeee 4
11-63
Figure 11-42 NC8UIRE-1 CY-1 ASSEMBLY PEAK AXIAL P00ER - CALCULATE 3 VS REASURED 104.47 EFPD 501FP CONTROL BANK 3 AT 215 STEPS WITHDRAUN 3 H 8 F E 3 C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee, e 1.43 e 1.38 e 1.44
- 1.45
- 1.44 e 1.35 e 1.22 e .84
- 8 e 1.47 e 1.44
- 1.49 e 1.53 e 1.48
- 1.40 e 1.24 e .86
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.43 e 1.39 e 1.45 e 1.42 e 1.37 e 1.23 e .90 e f
- 1.46 e 1.45 e 1.49 e 1.48
- 1.39 e 1.27 e .92 e e e e e e e e o eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.45 e 1.44 e 1.41
- 1.32 e 1.18 e .77 e to e 1.46 e 1.48 e 1.44 e 1.38 e 1.19 e- .79 e e e o e e e o seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.43 e 1.34 e 1.31 e 1.18 * .45
- 11 e 1.44 e 1.38 e 1.33 e 1.19 e .44 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.47 e 1.18 e .98 e 12 e 1.49
- 1.18
- 1.00
- e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.18 e .57
- CALCULATED 13 e 1.18 * .58 e MEASURED e e e seeeeeeeeeeeeeeeeeeee i
t i 11-64
Figura 11-43 l MCSUIRE-1 CV-1 ASSEMBLY PEAN AXIAL POWER - CALCULATEP V8 NEASURED 1 112.05 EFPB 50!FP CONTROL BANN 3 AT 215 STEPS WITHDRAUN i H 8 F E 3 C 9 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41 e 1.38 e 1.43 e 1.43 e 1.42 e 1.34 e 1.20 e .83
- 8
- 1.44 e 1.42
- 1.47 e 1.51 e 1.44 e 1.39 e 1.23 e .84
- e e o e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.42 e 1.39 e 1.43 e 1.41 e 1.35 e 1.22 e .90 e f e 1.44 e 1.43 e 1.47 e 1.47
- 1.37 e 1.26 e .91
- e e e o e e e e seeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.43 e 1.42 e 1.40 e 1.31 e 1.17 e .77 e 10 e 1.44 e 1.48 e 1.43 e 1.34 e 1.18 e .79
- e e e e e e e esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.41 e 1.35
- 1.29 e 1.18 e .45 e 11 e 1.45
- 1.37
- 1.32 e 1.19 e .64 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.46
- 1.18 e .97 e 12
- 1.48
- 1.18 e 1.00
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.17 * .57
- CALCULATES 13 e 1.19 e .59 e NEASURED e e e eeeeeeeeeeeeeeeeeeeee 4
11-65
Figure 11-44 l l NCSUIRE-l CY-1 AlltN8'.Y PEAM AXIAL PCWER - CALCULA1tB VS NEASURED 115.49 EFP4 753FP CONTROL BANN 3 AT l'17 SitPS WITHSRAWN N 8 F E B C I A esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.45
- 1.40 e 1.45 e 1.45 e 1.44 e 1.35 e 1.21 e .83 e :
8 e 1.47 e 1.44 e 1.48 e 1.50 e 1.44 e 1.30 e 1.22 e .84 e ! e o e e e o e e e ; seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.44 e 1.41 e 1.44
- 1.42 e 1.37 e 1.23 e .90 e f e 1.44 e 1.44 e 1.47 e 1.44 e 1.37 e 1.25 e .90
- e e e e o e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.45 e 1.44 e 1.42 e 1.32 e 1.10 * .77 e 10 e 1.44 e 1.47 e 1.42 e 1.35 e 1.17 e .70
- e e e o e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.43 e 1.34 e 1.30 e 1.17 * .45 e 11 e 1.43 e 1.35 e 1.30 e 1.17 * .45 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.10 e .97 e 12 e 1.47
- 1.17 e .99
- e o e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.14 * .54 e CALCULATED ;
11 e 1.17 e .50 e HEASUREB e e e seeeeeeeeeeeeeeeeeees i G 11-66
Figuro 11-45 NC6UIRE-1 CY-1 ASSEMBLY PEAN AXIAL POWER - CALCULATES VS MEASURES 118.71 EFPS 501FP CONTROL SANN 8 AT 100 STEPS UlfMBRAUN ! H 8 F E 8 C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee 7 e 1.44
- 1.41 e 1.44 e 1.44 e 1.45 e 1.37 e 1.23 e .85
- 8 e 1.48 e 1.47 e 1.50 e 1.54 e 1.48 e 1.43 e 1.24 e .89 e e o e o e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.45 e 1.42 e 1.47 e 1.44 e 1.38 e 1.25 * .92 e f
- 1.47 e 1.47 e 1.49 e 1.50 e 1.40 e 1.30 e .94 e o e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.45 e 1.43 e 1.34 e 1.20 e .78 e 10 e 1.48 e 1.52 e 1.45 e 1.40 e 1.20 e .82
- e o e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.44 e 1.37 e 1.32 e 1.20 * .44 e 11 e 1.47 e 1.41 e 1.34 e 1.21 e .A8 e o e e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.49
- 1.20 * .99
- 12 e 1.51 e 1.22 e 1.02
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.20 e .58
- CALCULATED 13 e 1.21 * .40
- NEASURED e e e eeeeeeeeeeeeeeeottete SUL-(h.R - - - _ , . _ _ - -. . - _ - - _ _ - _ . . _ . . _ - _ _ _ _- _
Figura 11-46 NCSUIRE-1 CY-1 Al8EN9LY PEAR AXIAL POWER - CALCULATES VS MEASURE 8 122.15 EFPS 751FP CONTROL BANN 3 AT 215 STEPS WITHDRAWN i N 8 F t 3 C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.40 e 1.44 e 1.44 e 1.42 e 1.34
- 1.21 e .83
- 8 e 1.45 e 1.44 e 1.44 e 1.49 e 1.43 e 1.38 e 1.21 * .84 e e e e e e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.43 e 1.44 e 1.44 e 1.41 e 1.34 e 1.22 e .89
- t e 1.44
- 1.44 e 1.45 e 1.45 e 1.35 e 1.24 e .90 e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.43 e 1.40 e 1.32 e 1.17 * .74 e 10 e 1.44 e 1.47 e 1.40 e 1.35 e 1.15 * .78 e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.42 e 1.35 e 1.29 e 1.17 * .45 e il e 1.42 e 1.34 e 1.28 e 1.14 e .45 e i
e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeee e 1.45 e 1.18 e .94 e 12 e 1.45 e 1.17 * .98 e e e o e seeeeeeeeeeeeeeeeseeeeeeeeeeeee e 1.14 * .54 e CALCULAft3 13 e 1.*5 * .58 e NEASUNED e e e eeeeeeeeeeeeeeeeeeeee I
- _ - _. _ , _ _ _ ___ _ _.- __. __ __ .11-68._ _ _ __ ____ _ _ _ ___._.,_. _ _ _ .
Figuro 11-47 t NCSUIRE-l CY-1 A88tN8LY PEAR AXI AL P0Utt - CALCULATES VS MEAluttB 130.59 EFPS 753FP CONTROL BANN 3 AT 215 STEPS UlTN8RAUN N O F E B C 3 A l ee....eeeeeeeeeeeee...........e e..........................eeeee.........e e eeee
- 1.43 e 1.39 e 1.43 e 1.43 e 1.41 e 1.34 e 1.20 e .83
- I
- 1.45 e 1.43 e 1.44 e 1.49 e 1.42
- 1.38
- 1.22 e .86 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.42 e 1.39 e 1.43
- t.40 e 1.34 e 1.22 e .09 e f e 1.44 e 1.44 e 1.45 e 1.43 e 1.34 e 1.25 e .90
- e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.43 e 1.42 e 1.39 e 1.31 e 1.16 e .74
- 10 e 1.45 e 1.47 e 1.39 e 1.35 e 1.14 e .7? e o e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41 e 1.35 e 1.29 e 1.14 . .44
- 11 e 1.42 e 1.34
- 1.29 e 1.17 * .44 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.17 e .f4 e 12 e 1.44 e 1.18 e .99 e e e e e
, eeeeeeeeeeeeeenseeeeeeee.eeeeee e 1.15 e .54 . CALCULATED 13
- 1.17 e .50 e NEASURED e e e eeeeeeeeeteteeeeeeeee 11-69
Figure 11-68 i NC40!RL-1 CV-1 ASSEN8LY PEAN AXIAL POWER - CALCULATES VS MEASURES 135.44 (FPS .75tFP CONTROL BANN 9 AT 215 litPS WITNDRAWN . N 8 F t 3 C 3 A , eeeeeeee....e ........ee...e... .ee...eeeee.e.eee.....eee.....eeeeeeee ee.eeeeee
. 1.42
- 1.39 e 1.42 e 1.42 e 1.40 e 1.33 e 1.19 . .82 e r I e 1.44
- 1.43 e 1.45 e 1.48
- 1.41 e 1.34 e 1.20 * .05 * ,
. e e . . .
e e .
....e eeeeee...eeeeeeee...... .......e ee...e eeeeeeeeeeeeeeee......e........ ...
e 1.41 e 1.39 e 1.42 e 1.40
- 1.34
- 1.22 e .89
- r 9
- 1.43 e 1.42
- 1.43
- 1.43
- 1.34 e 1.23 e ett e
. e . . e e . .
................eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee....................ee !
e 1.42 e 1.42 e 1.30 e 1.31
- l.14 e .74
- l 10 e 1.42 e 1.45 e 1.30 e 1.33 e 1.14 . .77 * -
e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee. e 1.40 e 1.34 e 1.20 e 1.14 * .44 e 11 e 1.40 e 1.34 e 1.27 e 1.15 e .44 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee.......eeeeee. eeeee e 1.44 e 1.17 e .94 e 12 e 1.43 e 1.16 e .97
- e e o e eeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.15 e .54 e CALCULAftB 13 e 1.15 . .57 e NEASURED e o e e eeeeeeeeeeeeeeee...
f f I i 1 1 11-70 :
l Figuro 11-49 NCSUIRf-l CY-1 AlttNSLY PEAN AXI AL POWER - CALCULAf tB VS MEASuttl 139.02 EFPS 501FP CONTROL BANN 8 AT 100 litPS WITNDRAUN M 0 F t 3 C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.42 e 1.30 e 1.42 e 1.43 e 1.40 e 1.34 e 1.21 e .04
- I e 1.40 e 1.44 e 1.40 e 1.51 e 1.45 e 1.41 e 1.24 e .00 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41 e 1.39 e 1.42 e 1.41 e 1.35 e 1.24 e .91 e f e 1.47 e 1.44 e 1.47 e 1.47 e 1.30 e 1.20 * .92 e e e e e e e e e ,
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ! e 1.42 e 1.42 e 1.39 e 1.32 e 1.17 e .77 e 10 e 1.47 e 1.49 e 1.43 e 1.30 e 1.19 e .00
- r e o e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ,
- 1.40 e 1.35 e 1.29 e 1.19 e .45 e l
11 e 1.44 e 1.41 e 1.32 e 1.20 * .47 e ! e . e e e e l seeeeeeeeeeeee..eee.eeeeeee.eee.ee.eeeeeeeeeeeeeese e 1.45 e 1.19 e .90 e ! 12 e 1.50
- 1.21 e 1.01 e i e e e e t seeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.10 e .57 e CALCULATED 13 e 1.20 e .40 e NEASURED e e e teeeeeeeeeeeeeeeeeete e
i i 11-71
Figuro 11-50 l I NCOUIRE-1 CY-1 AlltNDLY PEAR AXIAL POWER - CALCUL4Tt3 VS REASUNED 141.52 IFPB 30tFP CONTROL BANN 8 AT 215 STEPS WITHDRAWN N 8 F I 9 C 3 A eeeeeeeeeeeeeeeeeeeeeeeee.eeeee.eeeeeeeeeeeeeeeeee.eeeeeeeeeeeeeeeeee.eee.eeee.ee ; e 1.37 e 1.35 e 1.37 e 1.30 e 1.34 e 1.30 e 1.14 * .01 e 8 e 1.43 e 1.42 e 1.44 e 1.40 e 1.42 e 1.37 e 1.20 e .05 e e e e e s e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.37 e 1.35 e 1.30 e 1.34 e 1.30 e 1.20 * .07 e ; 9 e 1,47 e 1.42 e 1.44 e 1.44 e 1.34 e 1.24 * .90 *
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ,
e 1.30 e 1.30 e 1.35 e 1.28 e 1.13 * .75 e ! 10 e 1.43 e 1.44 e 1.39 e 1.34 e 1.14 e .70 e o e e e e e e i eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee,
- 1.34 e 1.32 e 1.25 e 1.14 e .43 e 11 e 1.41 e 1.35 e 1.29 e 1.17 e .44 e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.41 e 1.14 * .94 e 12 e 1.45 e 1.18 e .99 e e e e e
- seeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.14 e .54 e CALCULATED ;
13 e 1.1/ e .59 e NEASURE8 e e e esteeeeeeeeeeeeeeeete t 11-72
. _ . _ _ . . _ _ . _ _ _ , _ . _ _ , _ _ . . _ _ _ _ _ , _ _ . _ _ _ . ~ . - _ - _ _ _ _ - _ _ _ _ -
Figura 11-51 t i RCSUIRt 1 CV-1 488tN9LY PEAR AXIAL POWER - CALCULATE 9 V8 NEASURED 144.01 IFPI 75IFP CONTROL BANN 9 AT 215 STEPS WITN944WN N O F t 3 C 8 A l eeeeeeeeeeeeeeeeeeeeeeeeeeeees.eeeeeeeeeeeeeeeeeeeeeeeeeeee..eeeeeeeeeeeeeee..ese e 1.41 e 1.34 e 1.40 e 1.41 e 1.30 e 1.32 e 1.19 e .02 e' ; I e 1.43 e 1.44 e 1.43 e 1.47 e 1.44 e 1.34 e 1.19 e .45 e e e e e e o e e e eeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeees e 1.40 e 1.38 e 1.40 e 1.39 e 1.32 e 1.21 * .44
- f e 1.42 e 1.43 e 1.42 e 1.43 e 1.32 e 1.23 e .99
- e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.44 e 1.44 e 1.37 e 1.30 e 1.15 e .74 e 10 e 1.41 e 1.45 e 1.37 e 1.33 e 1.14 e .70 e e o e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.38 e 1.34 e 1.17 e 1.14 e .44 e '
11 e 1.39 e 1.35 e 1.24 e 1.15 e .44 e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ,.
- 1.43 e 1.17 e .95 e 12 e 1.42
- 1.14 e .97 * !
e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee :
- 1.14 e .54 4 CALCULATE 8 '
13 e 1.14 e .50 e REASURES , e e e eteeeeeeeeeeeeeeeeet, 5 E i
_ -._._ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ ~ _ _ _ _ _ _ Figure 11-52 e RCSUIRE-l CV-l 488tNDLY PEAM AXI AL PSUER - CALCULAf t3 VS mtASWRtB 1 150.19 EFPS 543FP CONTROL BANW 8 AT 215 litPS WITN8RAWN l N 8 F E 8 C 8 A , \ 1 e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee,
- 1.34 e 1.34 e 1.34 e 1.38
- 1.35 e 1.29 e 1.14 e .81 e 8 e 1.39 e 1.44 e 1.44 e 1.44 e 1.38 e 1.35 e 1.18 e .84 e o e e o e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.34 e 1.34 e 1.34 e 1.35 e 1.29 e 1.19 e .87
- 9 e 1.34 e 1.39 e 1.39 e 1.41 e 1.30 e 1.22 e .88 e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee>eeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.34 e 1.37 e 1.34 e 1.27 e 1.13 e .74 e le e 1.39 e 1.43 e 1.35 e 1.32 e 1.13 e .77 e e e e e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.35 e 1.31 e 1.24 e 1.14 e .43 e ,
il e 1.37
- 1.34 e 1.25 e 1.15 e .44 e i e e e e s e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee 3 1.40 e 1.14 * .f4 e '
12 e 1.41 e 1.14 e .97 e ! e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.13 e .35
- CALCULAft3 13 e 1.14 e .50 e NEA894(8 e e e eeeeeeeeeeeeeeee4.cee t
11 14
Figure 11-53 MCSUIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 162.76 EFP8 50ZFP CONTROL BANK B AT 215 STEPS WITHDRAWN N 8 F E D C 3 A seeescoeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees.eeeeeeeeeeeeeeeeeeee....
- 1.35
- 1.33
- 1.35
- 1.36 e 1.33
- 1.28
- 1.15 e .80
- 8
- 1.39 *- 1.39 e 1.39
- 1.43 e 1.34
- 1.33 e 1.17 e .83 e e e e e e e e e e eseesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.34 e 1.33 e 1.35
- 1.34
- 1.28 e 1.18 * .86
- 9
- 1.37
- 1.38 e 1.38
- 1.39
- 1.29 e 1.21 e .87 *
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeees, seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.35
- 1.36
- 1.32
- 1.26
- 1.12 * .74
- 10
- 1.37 e 1.42
- 1.34
- 1.31 e 1.12 e .77
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.34
- 1.30
- 1.23
- 1.13 * .63 e 11 e 1.36 e 1.32
- 1.24 e 1.14 e .44
- e e e e e e see.eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesse
- 1.38
- 1.15 * .93
- 12
- 1.40
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- 1.12 * .55
- CALCULATED 13
- 1.14 * .58 e MEASURED e e e
$$eeeeeeeeeeeeeeeeeee
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=
.a,-
i Figure 11-54 l l MC6UIRE-1 CY-1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURES 173.34 EFPD 50ZFP CONTROL BANK B AT 215 STEP 8 UITHDRAUN H 8 F E D C 3 A \ eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.35
- 1.33
- 1.33 e 1.35
- 1.32
- 1.28
- 1.14 * .80 e 8
- 1.36 e 1.37
- 1.37
- 1.42
- 1.34 e 1.34 e 1.17 e .84 e e e e e o e
- e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.33
- 1.33 e 1.33 e 1.33
- 1.27
- 1.18 * .84 e 9
- 1.35
- 1.38 e 1.34 e 1.39 e 1.28 e 1.22 * .88 e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.34
- 1.35 e 1.31
- 1.26 e 1.11 * .74
- 10
- 1.35
- 1.41
- 1.33
- 1.31
- 1.12 * .77. e e e e e e o e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.32 e 1.30
- 1.23
- 1.13 * .42
- 11 e 1.34 e 1.33 e 1.24 e 1.14 * .44
- e e e e *
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.38
- l.15 * .93
- 12 e 1.39
- 1.17 * .94
- e o e e seeeeeeeeeeeeeeeeeeeeeeeee**ees
- e 1.12 * .55
- CALCULATED L 13 e 1.14 e .50
- MEASURED
! e e e l ....................e l l l 11-76
Figure 11-55 4 NC8UIRE-1 CY=1 ASSEMBLY PEAM AXIAL POWER - CALCULATED US MEASURED 185.58 EFPB 501FP CONTROL BANK B AT 215 STEPS WITHDRAWN H 6 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.34 e 1.32 e 1.32 e 1.34 e 1.30 e 1.27
- 1.13 * .80
- 8 e 1.35
- 1.37
- 1.35
- 1.40 e 1.33 e 1.32 e 1.16 * .84
- e e e e e e e e 's eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.32
- 1.32
- 1.32 e 1.32 e 1.26 e 1.18 e .86
- 9
- 1.34 e 1.36 e 1.34 .e 1.38 e 1.27
- 1.21 * .87
- e e e e e e e ,e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesseeeeeeeeeeesesseee e 1.32 e 1.34 e 1.30 e 1.25 e 1.11 * .74
- 10 e 1.34 e 1.39 e 1.31 e 1.30 e 1.11 e .77
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessessessesse e 1.31 e 1.30 e 1.22 e 1.13 e .62
- 11 e 1.33 e 1.32 e 1.24
- 1.14 * .44 *
- e o e e
- eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeee e 1.37 e 1.15 * .93 e 12
- 1.39 e 1.17 * .96 e e e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.12 * .55
- CALCULATED 13
- 1.13 * .58 e MEASURED e e o eseeeeeeeeeeeeeeeeeee 11-77
Figure 11-56 MCSUIRE-1 CY-14 ASSEMBLY RABIAL POWER CALCULATER Y8 MEASURED 198.46 EFPS 90ZFP CONTROL BANK 3 AT 217 STEPS WITHDRAWN N 6 F E 3 C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.12 e 1.20 e 1.11 e 1.16
- 1.07 e 1.13 e 1.00 * .76
- 8
- 1.00 e 1.22
- 1.08 e 1.18 e 1.05
- 1.18 * .99 * .79
- e e e o e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.20 e 1.12 e 1.19
- 1.09 e 1.15 e 1.07 e 1.14 e .76 *
? e 1.21
- 1.08
- 1.20
- 1.05
- 1.18 e 1.04 e 1.17 * .76
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.10 e 1.18 e 1.09
- 1.16 e 1.08 e 1.13 e .96 e .70 e 10 e 1.06 e 1.19 e 1.04
- 1.19 e 1.04 e 1.15 e .93 * .71
- e o e e e e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeneseeeeeeeeeeeeeeeeeees e 1.13 e 1.07 e 1.15 e 1.10
- 1.19
- 1.04
- 1.02 * .54
- 11
- 1.14 e 1.04
- 1.17 e 1.07 e 1.21 e 1.03
- 1.03 e .54
- e e e e e e e o e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.01 e 1.11 e 1.04
- 1.18 e 1.18 e 1.12 * .82 e 12 e .99
- 1.13 e 1.04
- 1.20 e 1.16 e 1.13 * .83
- e s e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .96 e 1.00
- 1.09 e 1.04 e 1.11 e 1.01 e .54
- 13 * .98 * .98
- 1.13 e 1.02 e 1.11 e 1.01 e .57
- e e e e e e e e eeeeeeeeeeeeetseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .92
- 1.07 * .93
- 1.00 e .81 e .56
- 14 * .91 e 1.11 e .90 e 1.00 e .81 e .57
- e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
* .71 * .72 e .47 * .53 e CALCULATEB 15 * .73 * .72 * .47 e .52 e NEASURED e e o e e seeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee 11-78
Figure 11-57 l-l NC8UIRE-1 CV-14 ASSEMBLY RADIAL POBER CALCULATED VS MEASURED 217.53 EFPB 1001FP CONTROL BANK 3 AT 209 STEPS WITHDRAWN N 8 F E 3 C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessesseseeeeeeeeeeeeeeeeeeeeeeessees****eeeese e 1.10 e 1.19
- 1.10
- 1.15 e 1.07 e 1.13 e 1.00 * .74
- 8 e 1.04 e 1.21
- 1.08
- 1.18
- 1.05 e 1.17 * .99 * .79 e o e o e e o e e e seeeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeee e 1.19 e 1.11 e 1.18 e 1.08 e 1.15
- 1.07 e 1.13 e .74
- f e 1.20
- 1.07
- 1.19
- 1.05
- 1.17
- 1.04
- 1.17 * .77
- e e o e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eeee**e
- 1.09 e 1.17 e 1.09
- 1.14 e 1.00 e 1.12 o .94 e .71
- 10 e 1.04 e 1.18 e 1.05 e 1.17 e 1.05
- 1.15 * .94 * .72
- e e e e e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.13 e 1.07
- 1.15
- 1.09 e 1.17 e 1.05 e 1.02 * .55
- 11 e 1.14 e 1.04
- 1.17 's 1.07
- 1.19 e 1.02 e 1.0L e .55 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee****es****
e 1.02 e 1.11
- 1.04
- 1.17
- 1.15
- 1.10 * .82
- 12
- 1.00
- 1.13
- 1.04
- 1.19
- 1.13
- 1.11 * .83
- e e e e e e e 4 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeesese e .98 e 1.01
- 1.10 e 1.04
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- 1.02 e 1.10 e 1.00 e .58
- e e e e e *
- e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessesseese e .94
- 1.09 e .94 e 1.01 e .82 e .54 e 14 * .93 e 1.12 * .92
- 1.02 * .82 e .57 e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
* .73 e .74 * .49 e .54
- CALCULATER 15 e .74 e .74 e .71 * .54 e NEASURE3 e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees I
L nn _52/R
Figure 11- 58 MCSUIRE-1 CY-1A ASSEMBLY RADIAL POWER CALCULATED VS MEASURED 223.35 EFPS 100!FP CONTROL BANK 3 AT 211 STEPS WITHDRAWN N 8 F E D C B A eseeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.09 e 1.18
- 1.10 e 1.15
- 1.07 e 1.13
- 1.00 e .77
- 8 e 1.06 e 1.21 e 1.08 e 1.18
- 1.05
- 1.18 * .?? e .79 e e e e e o e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18 e 1.11 e 1.17
- 1.08 e 1.14
- 1.07 e 1.13 * .77
- 9 e 1.20 e 1.07
- 1.19 e 1.05
- 1.17
- 1.05
- 1.17 e .77
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e -1.09
- 1.17
- 1.09 e 1.15
- 1.08
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- 10 e 1.07
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- e e e e e *
- e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeevesessee
- 1.13
- 1.07
- 1.14
- 1.09
- 1.17
- 1.05 e 1.02 * .55
- 11
- 1.17
- 1.04 e 1.17 e 1.07
- 1.19
- 1.02 e 1.02 * .55 *
*
- e e e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.02
- 1.11
- 1.04 e 1.17
- 1.14
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- 1.19
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- 1.13
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- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseseeese e .95 e 1.09 * .75 e 1.01 * .82 * .57 e 14 e .93
- 1.12 e .92
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- e e e e e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
* .74 * .74 * .49 e .54 e CALCULATEB 15 * .74 * .74 e .71 * .54 e NEASUREB e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee J
11 - 80 i
i Figure 11- 59 NCSUIRE-1 CY-14 ASSEMBLY RADIAL POWER CALCULATED VS NEASURED 234.23 EFPB 1001FP CONTROL BANK B AT 211 STEPS WITHDRAWN N 8 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.09
- 1.17 e 1.09
- 1.14
- 1.07
- 1.13 e 1.00 e .77
- 8 e 1.04 e 1.21 e 1.08 e 1.17 e 1.05 e 1.17 e .99 * .80
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.17 e 1.10 e 1.14
- 1.07
- 1.14 e 1.07 e 1.13 e .77 e f e 1.20
- 1.07 e 1.18 e 1.05 e 1.14 e 1.04
- 1.17 e .77
- e e e o e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.08 e 1.16 e 1.08 e 1.15 e 1.07
- 1.12 * .97 e .71
- 10
- 1.04
- 1.18
- 1.05
- 1.17 e 1.05
- 1.14 e .f4 e .73 e e
- e e e e o e e seeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.13 e 1.04 e 1.14 e 1.08
- 1.17
- 1.05
- 1.02 * .55
- 11
- 1.16 e 1.04 e 1.16 e 1.04
- 1.18 e 1.02 e 1.02 e .56
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.03 e 1.11 e 1.04 e 1.16 e 1.15
- 1.10 * .83
- 12 e 1.01 e 1.13
- 1.04 e 1.18
- 1.13 e 1.10 * .82
- o e e
- e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.00 e 1.02 e 1.10 e 1.04
- 1.10
- 1.00 e .57 e 13 e 1.00 * .??
- 1.13 e 1.01 e 1.10 e .?? * .59
- e s e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*******
e .96
- 1.10 e .95 e 1.01 e .82 e .57
- 14 e .94
- 1.12 e .93
- 1.02 e .82 e .58 e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e .75 e .75 e .70 e .55 e CALCULATE 3 15 * .77 * .75 * .72 * .55 e NEASUREB e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees 11-81 I JL _
Figure 11- 60 l l NC8UIRE-1 CY-14 ASSEMBLY RABIAL POWER CALCULATED VS MEASURED 249.75 EFPS 100XFP CONTROL BANK D AT 221 STEPS WITHDRAUN N 8 F E B C B A seeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.10 e 1.17
- 1.08
- 1.13
- 1.04
- 1.12
- 1.00 * .77 * .
l 8 e 1.07 e 1.20
- 1.04
- 1.15
- 1.05
- 1.14 * .99 * .80
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesseeeeeee e 1.17 e 1.09
- 1.15 e 1.07
- 1.13 e 1.04
- 1.13 e .77
- f
- 1.19 e 1.04
- 1.17
- 1.05
- 1.17
- 1.04
- 1.15 e .77
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- 1.08 e 1.15 6 1.07 e 1.14 e 1.07 e 1.12 * .97 * .72
- 10 e 1.04
- 1.17
- 1.05
- 1.17
- 1.05 e 1.14 * .93 e .72
- e e e e e e e e o eseeeeeeeeeeeeeeeeeeeweseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee oo e 1.12 e 1.04
- 1.13
- 1.08 e 1.17
- 1.05
- 1.02 * .54
- 11
- 1.17
- 1.04 e 1.15
- 1.05
- 1.18 e 1.03
- 1.03 e .55 *
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.03 e 1.11 e 1.04 e 1.17
- 1.17 e 1.10 e .83 e 12
- 1.03 e 1.14 e 1.02 e 1.17 e 1.14 e 1.11 e .83
- e e e e e e e e esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessees
- 1.01
- 1.03 e 1.10 e 1.05
- 1.10
- 1.00 e .58
- 13
- 1.00 e 1.00
- 1.14 e 1.02
- 1.09
- 1.00 * .59 *
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .f4
- 1.10 e .f4
- 1.02 * .83 * .58
- 14 * .95 e 1.13 e .94 e 1.03 e .81 e .58
- e e e o e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e - .74 e .74 * .71 e .55 e CALCULATES 15 * .78 * .75 e .73 e .54 e NEASURED e e e a e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
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l l 11-82
Figure 11-61 NC8UIRE-1 CY-IA ASSEMBLY PEAK AXIAL POWERS CALCULATED VS MEASURED l l 198.66 EFPB 901FP CONTROL BANN D AT 217 STEPS WITHDRAWN N 8 F E D C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee***ese e 1.34
- 1.41 e 1.30
- 1.35
- 1.25
- 1.29
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- 8 e 1.34 e 1.51
- 1.33
- 1.44
- l'.27
- 1.41 e 1.19 * .94
- e e e e *
- e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.41
- 1.32
- 1.38 e 1.27 e 1.33
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- 1.33
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*
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- eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesse4ese***
- 1.29
- 1.37
- 1.28
- 1.35
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- 10
- 1.31
- 1.46 e 1.30
- 1.45
- 1.29
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* * * * * * *
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- 1.32
- 1.25
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- 1.18
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- 1.11
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- 1.19
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- 13
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- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeestseese e 1.04
- 1.22
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- 1.08
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- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees.
* .81 * .82 * .76 * .60
- CALCULATED 15 * .87 * .86 * .83 * .63
- HEASURED e o e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese 11-83
Figure 11-62 MCOUIRE-1 CY-1A ASSEMBLY PEAK AXIAL POWERS CALCULATED VS NEASURED 217.53 EFPD 100%FP CONTROL BANK D AT 209 STEPS WITHDRAWN
)
H 6 F E D C B A eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesssesseesteses
- 1.33
- 1.43
- 1.30
- 1.37 e 1.26
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- 8
- 1.31
- 1.48 e 1.33 e 1.42
- 1.26
- 1.39
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- 1.42
- 1.31
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- 1.29
- 1.39 e 1.28
- 1.37 e 1.27 e 1.33
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- 10
- 1.30
- 1.44
- 1.27
- 1.42
- 1.26
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11-84
Figure 11-63 NC6UIRE-1 CY-1A ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 223.35 EFPD 1001FP CONTROL BANK D AT 211 STEPS WITHDRAUN H 6 F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseesseessesseessessessessessessese***ese**es****
- 1.32
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8 e 1.33
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- 1.24
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- 11
- 1.42
- 1.27
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- 1.25 e 1.23 * .66
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- 1.19
- 1.31 e 1.24
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- e e e e e e
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- 1.14
- 1.18
- 1.29
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- 1.10
- 1.29 e 1.10
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M 11-85 .....__ _ _
Figure 11-64 MC6UIRE-1 CY-1A ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 236.23 EFPD 1001FP CONTROL BANK D AT 211 STEPS WITHDRAWN H 8 F E D C B A seeesseseseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.29
- 1.39
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- 1.27
- 1.41 e .93
- e e e e o e *
- e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.26
- 1.36
- T.26 e 1.34
- 1.25 e 1.32
- 1.12 * .83
- 10 e 1.28
- 1.42
- 1.27
- 1.41 e 1.27 e 1.38 e 1.12 * .87 *
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessesseessesseesseseeeeeeeeeeeeeeeeeeeese*
- 1.32
- 1.23
- 1.34
- 1.26
- 1.38
- 1.23
- 1.20 * .64
- 11
- 1.41
- 1.26
- 1.40
- 1.29
- 1.45 e 1.24
- 1.22 * .66
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18
- 1.30
- 1.24
- 1.38
- 1.38
- 1.31 * .97
- 12
- 1.21
- 1.36
- 1.25 e 1.44
- 1.41
- 1.34 e .98
- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesse
- 1.15
- 1.18
- 1.29
- 1.22
- 1.30 e 1.18 * .67 e 13
- 1.19
- 1.19
- 1.36
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- e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeestsesse
- 1.11
- 1.30 e 1.10
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- 14
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eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese
* .88 * .88 * .82 e .64
- CALCULt.fEP 15 * . 91 e .89 * .86 e .45 e MEASURED '
e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese 11-86
Figure 11-65 N 5 h i i i MCGUIRE-1 CY-1A ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 7 249.75 EFPD 100ZFP CONTROL BANK D AT 221 STEPS WITHDRAWN H 8 F E D C B A r sessesseeeeeeeeeeeeeeeeeeeeeeeeessessessessessessessessessessessesse***eessessess i e 1.26 e 1.35
- 1.24 e 1.31
- 1.21
- 1.29
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8 e 1.26
- 1.40 e 1.24
- 1.36
- 1.25
- 1.38 e 1.18 * .95
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secesseesseesseossesseessessessessessesseessesseessessessessessesseessessessesses
- 1.35
- 1.25
- 1.33 e 1.22
- 1.30
- 1.22
- 1.32 e .89
- 9 e 1.41
- 1.26
- 1.39
- 1.24
- 1.39
- 1.23
- 1.37 e .91
- a e e e e e o e e
- _
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e e
- e e e e e
- eesseeeeeeeeeeeeeeeeeeeeeeeeeeeessessessessesseesseesseessesseeessesseessesessess e 1.29
- 1.21
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- l 11
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- 1.24
- 1.34 e 1.24 e 1.41
- 1.22 e 1.20 e .65
- e e e e e e e e e sesseeeeeeeeeeeeeeeeeeeeeeeeeeeese.eesseessessesseesseessessesse*eresseessessesse 2
- 1.17 e 1.28
- 1.21
- 1.35 e 1.35
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- v5
- 12
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- 1.38 e 1.37
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e e e e e e e e -- essessessessessessesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeessesseessessesse**es e 1.15
- 1.17
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- 13
- 1.18
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- 1.22 e 1.30
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- 1.29
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- 14
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eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessessesse h 5 11-87
Figure 11-66 SEQUOYAH-1 CV-1 ASSEMBLY RADIAL POWER CALCULATED VS HEASURED 71.82 EFPD 100(Z)FP CONTROL BANK D AT 200 STEPS WITHDRAUN F E D C B A H 0 eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseseesseseeeeeee
- 1.14 e 1.09 . 1.19 e 1.12
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- esseeeeeeeeeeeeeeeeeeeeeee....eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee********
e 1.18 e 1.12 e 1.20 e 1.16
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- 9
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e 1.20 e 1.13
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- 1.19 e 1.13
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11 e 1.18 e 1.13 e 1.08 e .92 * .56
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- 12
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* .98 * .48
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(' 11-88
Figure 11-67 u 5 5 n. SEGUOYAH-1 CY-1 ASSEMBLY RADIAL POWER CALCULATED VS MEASURED . 101.42 EFPB 100(1)FP CONTROL BANK D AT 218 STEPS WITHDRAUN F D C B A : H 6 E eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.17 e 1.11
- 1.20 e 1.15
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- 8
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- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese 3
+ 1.19
- 1.15 e 1.21
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l seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.20 e 1.15
- 1.18
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10
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!
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- 1 11
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E 11-89
I I i I Figure 11-68 F
. 1
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SEQUOYAH-1 CY-1 ALSEMBLY RADIAL POWER CALCULATED VS HEASURED k 133.30 EFPD 100(11FP CONTROL BANK D AT 216 STEPS WITHDRAUN F D C B A H 6 E seesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*o**
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e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessessesseeeeeeeeeeeeeeeeeeeeeessessessessese***
. 1.19
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9
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eseeeeeeeeeeeeeeeesse.eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**se**ee*** '
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- CALCULATED 13 4 1.01 * .51
- MEASURED -
e e e _ . esseeeeeeeeee4essso4e e r 11-90 l
Figure 11-69 SEQUOYAN-1 CV-1 ASSEMBLY RADIAL POWER CALCULATED VS MEASURED 144.04 EFPD 100(%)FP CONTROL BANK D AT 210 STEPS WITHDRAUN N 6 F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeeeeeeeeeeeeeeeeeeeeeeeeeeee. ...............
- 1.15 e 1.13 e 1.20
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- 8
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- 1.19
- 1.18
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- 1.20
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13
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11-91
L. Figure 11-70
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e SEQUDYAH-1 CY-1 ASSEMBLY RADIAL POWER CALCULATED VS MEAEURED J r 231.70 EFPD 100(%)FP CONTROL BANK D AT 214 STEPS WITHDRAWN 3 H 6 F E D C B A / eseeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesessesseseees
- 1.11
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e o e o e e e e e seesesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**seesseeeeeeeeeeeeeeeessessesseesseseos
- 1.16
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11-92 1 M
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Figure 11-72 5 4 a 1 SEGUOYAH-1 CY-1 ASSEMBLY RADIAL POWER CALCULATED VS HEASURED j 378.92 EFPD 100(1)FP CONTROL BANK D AT 222 STEPS WITHDRAWN j a H 6 F E D C B A ] 4 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees ...eeeeeee..ees.o.e ..e.e. g -
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E e 11-94
.s Figure 11-73 e . ...;
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u r
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s '.. SEQUDYAH 1 CYCLE 1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED EN .. 71.82 EFPD 100(%)FP CONTROL BANK D AT 200 STEPS WITHDRAUN , W'tr . i!; g ..- .-. a H G F E C D D E j r}g,j, eeeeeeeeeeeeseseseeeeseeoeeeseeeseeeeeeeeeeoeeeseeeseeseseseseeseseesesseessesses ;.g>. T c
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9
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- 1.52
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- 1.25 * .60
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- MEASURED
- e
- esseeeeeeeese4e4eeses 11-95
Figure 11-74 SEQUDYAH I CYCLE 1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 101.62 EFPD 100(1)FP CONTROL BANK D AT 218 STEPS WITHDRAWN F E D C B A H 6 esseeeeeeeeeeeeeeeeeeeeeeeeasessesseeessessessessessessessessessesses**essssseses e 1.44
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Figure 11-75 SEQUOYAH 1 CYCLE I ASSEMBLY PEAK AXIAL POWER - CALCULATED VS NEASURED 133.30 EFPD 100(%)FP CONTROL BANK D AT 216 STEPS WITHDRAUN H 6 F E D C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee ssee.*esee....eees e 1.40
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- 1.43
- 1.39
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- 1.42
- 1.33 e 1.17 * .79 *
* * * *
- e ,
eseesseeessesseesseessessessessessessessessessesseessessesses
- 1.40
- 1.36
- 1.26
- 1.08 * .62
- 11
- 1.42
- 1.38
- 1.29
- 1.10 * .46
- e e e e e e sessessessessessesseessessessessesseessesessee*es:ee
- 1.30
- 1.18 * .95
- 12
- 1.34
- 1.22
- 1.03 *
* *
- e seesessesseesseeeeeeessesee44e*
- 1.14 * .56
- CALCULATED 13
- 1.23 * .62
- NEASURED e *
- etesessessesse*4*e4:*
~ .
11-97
Figure 11-76 SEQUOYAH 1 CYCLE 1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 166.04 EFPD 100(%)FP CONTROL BANK D AT 210 STEPS WITHDRAUN H G F E D C B A sessesseessessessessessesseesseeeeeeeeeeeeeeeeee**eseeeeeeeeeeesese**********se*e
- 1.37
- 1.33
- 1.40
- 1.39
- 1.38
- 1.28
- 1.15 * .81
- 8 e 1.37
- 1.29
- 1.39
- 1.38 e 1.39
- 1.29
- 1.17 * .84
- e e o e e e e *
- eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesesseeeeeeeeeeeeeeeeeeeeeeeeeeee******
- 1.39
- 1.40 e 1.40
- 1.39 e 1.30 e 1.17 * .86
- 9
- 1.38
- 1.37
- 1.42 e 1.40
- 1.32
- 1.20 * .90 e e
- e e e e e
- esseeesseessessessesseessessessesseeeeeeeeeeeeeeeeeeeeeeeee**ese*******
- 1.40
- 1.38 e 1.36 e 1.28
- 1.11 * .74
- 10
- 1.40 e 1.36
- 1.38 e 1.30 e 1.13 * .78
- e e e e e e e eseesseesseeeeeeeeeeeeeeeeeeeeessessessesse**essees*es****es*
- 1.37
- 1.34
- 1.24
- 1.06 * .62
- 11
- 1.36
- 1.34
- 1.25 e 1.08 * .66
- e *
- e
- e essessesessesseesseessesessessessessesses***ese**+e
- 1.28
- 1.17 * .93
- 12
- 1.30
- 1.19
- 1.00
- sesseesseessesse*** esses *******
- 1.12 * .56
- CALCULATED 13
- 1.19 * .41
- MEASURED e e e see*eassessese**essee
-G.
I 11-98
Figure 11-77 SEQUDYAH 1 CYCLE 1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 231.70 EFPD 100(I)FP CONTROL BANK D AT 216 STEPS WITHDRAWN H G F E D C B A. eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesse*****e**se** e 1.27 e 1.26
- 1.32
- 1.35
- 1.31
- 1.24
- 1.12 * .80
- 8
- 1.28
- 1.24
- 1.31 e 1.34 e 1.31 e 1.26
- 1.13 e .83
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**essesses e 1.31 e 1.35 e 1.33 e 1.34 e 1.25
- 1.15 * .85
- 9
- 1.30 e 1.33 e 1.33
- 1.35
- 1.26
- 1.17 * .80 *
* * * *
- e e e esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese**ee**e
- 1.33 + 1.34
- 1.29 e 1.26 e 1.08 e .74
- 10
- 1.33 e 1.33
- 1.30
- 1.27
- 1.10 * .77
- e e e e e e
- esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*essesse
- 1.30 e 1.30 e 1.19
- 1.05 * .62
- 11
- 1.30
- 1.30
- 1.21
- 1.06 * .65 *
- e e e e e esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**es**e*
- 1.23 e 1.16 e .92 e 12
- 1.25
- 1.18 * .98 *
- e e e seesseessessessessessessesessee
=
- 1.10 * .56 e CALCULATED 13 e 1.15 e .62
- MEASURED e e e eeseessessese$e$s*e$e e
11-99
1 Figure 11-78 SEQUOYA4 1 CYCLE i ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED 292.04 EFPl. 100(1)FP CONTROL BANK D AT 216 STEPS WITHDRAWN H 0 F E D C B A seeeeeeeeeeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeesse e 1.23 e 1.22 e 1.28 e 1.32
- 1.27 e 1.26
- 1.12 * .82
- 8 e 1.24 e 1.23 e 1.28 e 1.33 e 1.29 e 1.26 e 1.14 e .85
- e e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.26 e 1.31 e 1.29
- 1.31
- 1.23 e 1.16 e .87
- 9
- 1.28
- 1.32 e 1.31 e 1.34
- 1.24 e 1.18 e .09 *
- e o e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
- 1.29 e 1.31 e 1.26
- 1.25 e 1.08 * .76
- 10
- 1.31 e 1.32
- 1.28
- 1.27
- 1.11 * .79
- e e e e e o e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeeee e 1.27 e 1.28 e 1.18 e 1.06 e .64 e 11 e 1.28
- 1.30
- 1.20
- 1.07 * .66 e e e
- e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeessees
- 1.21 e 1.17 * .92 e 12 e 1.24 e 1.19 * .97
- e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.10 e .57
- CALCULATED 13 e 1.15 e .63
- HEASURED e e e eseeeeeeeeeeeeeeteese
=
11-100
I-Figure 11-79
=
5= 7 SEQUDYAH 1 CYCLE 1 ASSEMBLY PEAK AXIAL POWER - CALCULATED VS MEASURED : 378.92 EFPD 100(%)FP CONTROL BANK D AT 222 STEPS WITHDRAWN 7-- H G F E D C B A ;_ eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee4e4eeee4eeese ;
- 1.19
- 1.20 e 1.24 e 1.30 e 1.26
- 1.27
- 1.15 e .86
- k;'
8 e 1.22
- 1.22
- 1.27 e 1.34
- 1.28 e 1.29
- 1.17 e .90
- e
- e e e e e e e ,,
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseseeeeees e 1.23 e 1.19
- 1.26 e 1.30 e 1.23 e 1.19 e .89
- 9
- 1.26 e 1.33 e 1.30
- 1.34 e 1.25
- 1.22 * .93
- e e o e
- e e e e.o o e ee.eeeeeee..e.e.ee.e e e e ee.e ees.e ee...e eee.e.ee e e..eee eees.4.e4 e e.
- 1.26 e 1.30
- 1.25
- 1.26
- 1.11 * .80 e -
10 e 1.29 e 1.33
- 1.28
- 1.30 e 1.14 e .84 e e e e e e e e ..
esseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese.seeeeeeeeeeeeee4 4eseese ; I e 1.26
- 1.28
- 1.20
- 1.09 e .67 4 11 e 1.28 e 1.32 e 1.23 e 1.11 e .69 +
e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee44eeeee4e
- 1.22 e 1.19 e .94 e 12
- 1.26
- 1.24 e 1.00 e j
- e e e eseseeeeeeeeeeeeeeseessessee4e*
- 1.11 e .59
- CALCULATED 13 e 1.19 e .67 e MEASURED e e e seeeeeeeeeeeesse4eee4 c
11-101
Figure 11-80 . MCSUIRE-l CY-1 P9057 CALCULATE 9 VS NEASURED ASSEMBLY RADIAL POWERS ) P9887 - 52.2 EFPD VS. CORE MEAS - 48.8 EFPB F E B C 3 A H 8 seee.ee......ee.ee.e..e.ee..ees.o.ee..e...ee..eee..ee.ee.seee.eee.....e.e..seeee. :l
?~
e 1.14 e 1.10 e 1.20 e 1.20
- 1.23 e 1.12
- 1.01 e .68
- 8
- 1.12
- 1.09 e 1.l*
- 1.21 e 1.23 e 1.12 e 1.01 * .71
- e e e e e e e o seeeeeeeeeeeeeeeeeeeeeeeeeee.4eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.18 e 1.12 e 1.23 e 1.18 e 1.15 e 1.00 e .74 *
.75
- f e 1.15 e 1.13 e 1.22
- 1.19
- 1.14
- 1.02 e e e e e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee++e
- 1.22 e 1.19 e 1.20
- 1.09 * .97 e .63
- 10 e 1.20 e 1.19 e 1.18
- 1.10 * .97 * .44
- o e o e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee.eee
- 1.20
- 1.08
- 1.08 e .94 * .53 e 11
- 1.19
- 1.08 e 1.07 e .94 * .54
- e e e e e e
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.22 e .92 e .50 e 12 e 1.18 * .92 * .81
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeees e .94 e .44 e CALCULATED 13 * .95 e .44 e MEASURED e e e seeeeeeeeeeeeeeeeeeee m
M 11-102 3
3 .g~~ -
~ ~
n: , l s ,. 'p"; Figure,11-81.i- ." yi x s ;,..m L 7 \- , , g
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v . . 4 l > . s ,,- ~
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- s. 1 ~- , - -
, . .s ,~ '
~- _ _ _ .
s MCSUIRE-1 CY-f, -P9057 CALCULATER (18 AEASUACE ASSEMBLY RABIAL POWERS
, s ,
P3087 - 104.4 CFPS VS. CORE REA8 - 104.5 EFPS , H 8 - F g ,. 3 C 3 A eseeeeeeeeeeeeeeeeeeeesseesseeessesseesseseeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeessee
.. 1.17
- 1.13
- 1.1 */ * -1.20' *, 1.20 e 1.12 e 1.00 e .69
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1.19
- 1.12
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- e
*; e e ~e e e e e
eeeeeeeeseeeeee4heeeeeeeescibeseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.1 C
- 1.14 *~ 1. 2'l o' 1.18
- 1.13
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- e
.74
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*, "1.20 * ~ i.19 e, 1.18 e 1.09 . .96 e .44 e 10
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- 1.11 * .94 * .44 *
~
e e e e e e e
' 'eeee nee e e e e e r. e s s e e e e e e e e e eee e eee e e e e e e e e e e e e e e e e
- 1.19
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,18 c~ 1.17 e _1.11
- 1.07 * .94 e .53
- s
* *
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eesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.21 * .94 e .80 e 12 6 1.18 e .94 * .80 *
- i e e e seseeeeeeeeeeeeeeeeesseessessee
* .96 e .47
- CALCULATER 13 e .f4 * .47
- MEASURED
- e o e seeeeeeeeeeeeeeeeeeee s
11-103
~ ' ' ' ' ' ' ' ' . ,a - - , , - , , ~_.__
Figure 11-82 NC8UIRE-1 CY-1 P9057 CALCULATED VS NEASURED ASSEMBLY RADIAL POWERS PSG87 - 154.7 EFPB VS. CORE MEAS - 150.2 EFPD H 8 F E D C B A seeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeesseeee***ese***esesse e 1.17
- 1.15 e 1.18
- 1.19
- 1.17
- 1.12
- 1.00 e .70
- 8
- 1.15
- 1.17
- 1.17
- 1.21
- 1.17 e 1.13
- 1.00 e .71
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee******e e 1.17
- 1.15
- 1.18 e 1.17 e 1.12
- 1.02 * .75
- 9 e 1.15 e 1.17 e 1.17 e 1.19
- 1.10 e 1.03 e .74
- e o e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee***eesee ,
e 1.18 e 1.18
- 1.14 e 1.10 e .94 e .45
- 10
- 1.14 e 1.20
- 1.14
- 1.11 * .95 * .45
- e o e o e *
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese***eese e 1.17 e 1.11 e 1.07 * .95 * .55
- 11 e 1.16
- 1.12 e 1.04 e .96 o .54
- e e e e e
- eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*******
- 1.20 e .96 * .81
- 12 e 1.17 e .97 e .81 *
- e e e seeeeeeeeeeeeeeeeeeeeee*****ese e .96 e .49
- CALCULATE 3 13 e .95 * .48
- MEASURES e e
- seeeeeeeeeeeee*******
11-104
Figure 11-83 l l 'NCSUIRE-1 CV-1A P3887 CALCULATES VS MEASURED ASSEMBLY RADIAL POWERS l PB887 - 200.9 EFPB VS. CORE MEAS - 198.7 EFPB ! H 8 F E 3 C B A seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.10
- 1.22 e 1.00 e 1.17 e 1.05
- 1.16 * .99 * .78
- Oe 1.08
- 1.23 e 1.09
- 1.19 e 1.04 e 1.18 * .99 e .79
- e o e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.22
- 1.09 e 1.20 e 1.06
- 1.17
- 1.05
- 1.18 * .74
- f e 1.22
- 1.08
- 1.20 e 1.04
- 1.18
- 1.05
- 1.17 e .74
- e o e o e e o e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.07
- 1.19 e 1.04 e 1.18 e 1.04 e 1.14 * .95 e .72 e 10 e 1.07
- 1.19 e 1.06 e 1.1f e 1.04
- 1.14 e .93 * .72 e e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
- 1.14
- 1.04 e 1.17
- 1.08
- 1.22 e 1.04 e 1.03 e .54
- 11 e 1.17 e 1.04 e 1.18 e 1.08 e 1.22
- 1.04 e 1.03 e .54
- e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e .98 e 1.12 e 1.04 e 1.21 e 1.20 e 1.15 * .83
- 12 e .99 e 1.13 e 1.04
- 1.21 e 1.17 e 1.13 * .83
- e e e e e e e e l e..e..........e.e...e.ee....e..ee.....e.ee..ee.eee.ee....e.e..eee.eeese 1 e .94 * .98 e 1.12 e 1.03 e 1.14 e 1.02 e .58
- 13 e .99 * .98 e 1,14
- 1.02 e 1.12 e 1.01 e .58 *
, e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee
* .91
- 1.11 * .91
- 1.01 e .82 * .58 e ,
14 e .91
- 1.11 * .91
- 1.01 e .82 o .57
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eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee : e .73 * .72 e .49 e .53
- CALCULATED '
15 * .74 * .72 * .70 * .53
- MEASURED l e s e e e l
eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese l 1 l 11-105 l I i
. - , - , - - - - --_m - . _ _ _ + - . - -
T
- Figure 11-84 SE808YAN-1 CV-1 PDGf7 CALCULATES VS MEASURED ASSEMBLY RADIAL POBERS PD887 - 103.6 EFPB VS. CORE MEA 8 - 101.6 EFPB H 8 F E 8 C B A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese e 1.14 e 1.08 e 1.17 e 1.13 e 1.18 e 1.08
- l.01 e .71
- 8 e 1.14 e 1.07 e 1.17
- 1.13 e 1.17 e 1.07 e 1.00 * .71
- e e o e e e e e e
eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseeeee e 1.17
- 1.12 e 1.19 e 1.17 e 1.14 e 1.02 * .77
- 9 e 1.16 e 1.12 e 1.19 e 1.17 e 1.13 e 1.01 e .77 e e e e o e e *
- eeeeeeeeeeeeeesseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.18 e 1.13
- 1.18 e 1.10 * .97 * .45
- 10 e 1.18
- 1.13 e 1.18 e 1.10 e .98 e .66 e e e e e e e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.18 e 1.15 e 1.09 * .91 * .55
- 11 e 1.18
- 1.14 e 1.00 e .92 * .54 e e e o e e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.13 e 1.00 * .84 e 12
- 1.11
- 1.00 * .85
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeee.
- 1.01 e .50
- CALCULATED 13
- 1.02 * .51 e NEASURED e e e eeeeeeeeeeeeeeeeeeeee 11-106
Figure 11-85 SE000YAN-1 CV-1 P9057 CALCULATE 8 VS MEASURED ASSEMBLY RABIAL POWERS P8087 - 155.5 EFPB VS. CORE MEAS - 133.3 EFPB 8 F E D C 3 A H soeeeeeeeeeeeeeeeeeeoosseeeeeeeeeeeeeeeeeeeeeeeessseeeeeeeeeeeeeeeeeeeeeeeeeeeees e 1.17 e 1.11 e 1.18 e 1.16
- 1.17 e 1.09 e 1.00 * .71
- 8 e 1.13
- 1.09
- 1.18
- 1.14
- 1.15 e 1.08 e .99 e .71
- e e e e e e e e e
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee*** eses e 1.18 e 1.14 e 1.19
- 1.18 e 1.12 e 1.02 * .74
- 9 e 1.17 e 1.14 e 1.19 e 1.18 e 1.12 e 1.01 e .77 e e e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.19 e 1.16 e 1.17 e 1.10 * .94 e .45
- 10 e 1.19
- 1.16
- 1.17
- 1.10 * .f4 e .44
- e e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees e !.17 e 1.15 e 1.07 e .90 * .55
- 11 e 1.16
- 1.14 e 1.04 e .92 e .56
- e e e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeees
.e 1.11 e 1.00 * .82 e 12
- 1.08 e 1.00 * .84 e e o e o eeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e .99 e .50
- CALCULATED 13 e 1.00 * .52
- MEASURED e e e eeeeeeeeeeeeeeeeeeeee 1
11-107 9
- - - - - - - - - - , - - - . _ , , - - _ _ . , ,-,--e -
-_,,,e a,------,--...m. - - , - . _ - . ,e- ---,e - - , - , , , - - - - - - , - - - - , . . . . - - , - ,
- <_ _: s ._
Figure 11-86 SEGUOYAH-1 CV-1 P3087 CALCULATED VS MEASURES ASSEMBLY RADIAL POWERS PDG87 - 342.7 EFPS VS. CORE MEAS - 378.9 EFPD H 8 F E D C 3 A eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeseesseessesseesse e 1.04 e 1.04 e 1.09 e 1.15
- 1.11
- 1.13
- 1.01 * .77
- 8 e 1.03 e 1.04 e 1.09
- 1.14 e 1.09
- 1.10
- 1.01 e .78
- e e e e e e e e e
seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeses e 1,07
- 1.14
- 1.11 e 1.14 e 1.09 e 1.04 * .80 e 9 e 1.07
- 1.14 e 1.11 e 1.15 's 1.08 e 1.05 * .81
- e e e o e o e e eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee e 1.11 e 1.15 e 1.11 e 1.12 * .98 * .72
- l 10
- 1.11
- 1.14
- 1.10 e 1.12 * .99 * .73
- e e o e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese
- 1.11 e 1.15 e 1.04 e .95 * .40 e I 11 e 1.10 e 1.13 e 1.05 e .94 * .41
- l e e e e e e eseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeses l e 1.08 e 1.05 e .85 e 12
- 1.07
- 1.07 e .87
- e e e e seeeeeeeeeeeeeeeeeeeeeeeeeeeeee
! e 1.01 e .57
- CALCULATED 13
- 1.03 e .59 e MEASURED e e
- toteeeeeeeeeeeeeeeses t
I i L 11-108 i l l l l
- 12. References .
~
- 1. Nuclear Associates International Corp. , " Advanced Recycle Methodology Program System Documentation", CCM-3, (EPRI Confidential), September, 1977.
- 2. Studsvik Energiteknik AB, "CASMO-2 A Fuel Assembly Burnup Program,"
Studsvik/NR-81/3, 1981.
- 3. Duke Power Company, "Oconee Nuclear Station Reload Design Methodology,"
NFS-1001, Rev. 4, June 1981.
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. March 1977.
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EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 5, October 1978.
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Studsvik/K2-81/491, 1981.
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EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 2, April 1976.
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EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 8, October 1975. 12-1 P e
- w - m --- - - .- - - - - - - ,--.~-.w v w - - ----, ,, . -, , - ,- , , - - - , ----
- 12. Duke Power Company, "NULTIFIT User Documentation," (Proprietary),
February 1983.
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1 l i
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~
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p
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March 1982.
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12-2
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