ML20148D049

Describes Methods Used in Physics Model Development, Including Descriptions of Techniques Used in Generation of Input & Modeling of Core & Comparisons of Plant Data W/ Results Generated by Physics Model

ML20148D049
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	08/25/1978
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML20148D026	List: ML20148D020 ML20148D049
References
NUDOCS 7811020218
Download: ML20148D049 (61)

Text

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V BIG ROCK POINT PliYSICS METHODOLOGY ggpogy Consumers Power Ccrnpany Jackson, Michigan i

August 25, 1978 4,

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g TABLE OF CONTENTS Page Number ABSTRACT 1.0 Introduction 1

2.0 Description of Big Rock Point 2

j 30 Physics Model 3

31 Overview 3.

3 1.1 GR%KOverview 3

32 Assembly Physics Calculations 3

3 2.1 CASM%

4 3 2.2 PDQ7 5

33 GR%K Input Generation 6

3.h DescriptionofGRpKComputerProgram 9

3.h.1 GR%K Physics Model 9

3.h.1.a Source or Power Iteration 9

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3.h.1.a.1 Source Equations 10 3.4.1.a.2 Physics Parameters - K and M 12 gg 3.4.2 void Iteration 16 3.h.3 Control Rod Positioning and Power 17 Variation 3.h.h Fuel Burnup 17 35 Peaking Factor Algorithm 17 3.6 Fluxwire Calculations 19 37 Thermal Hydraulic Limits

-20 h.0 Verification of the Big Rock Point 23 Physics Model 4.1 BOC Cold Critical Control Rod Pattern 23 h.2 Axial Profiles

.23 h.3 Core Multiplication Factor Versus 2h Exposure 50 Summary and Conclusions 25

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i ABSTRACT This report describes the methods used in the Big Rock Point Physics Model developed by Consumers Power Company. Included in this r.eport are_ descriptions-j of.the techniques used.in generation of input and modeling of the core. - Com-parisons of plant data to results generated by the Physics Model are also in-cluded.

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1.0 INTRODUCTION

The purpose of this report is to describe the met ads used in the Big Rock Point Physics Model. ThemajorcompucntofthemodelisGRpK,athreedimensional

' boiling water reactor simulator code. GR%Khasbeenusedforreactor. physics work for the Big Rock Point (BRP) plant since 1972.

Features of GRgK-include calculation of reactivity inventory, peak heat flux, minimum critical heat flux ratio, maximum average planar linear heat generation rate, dryout times, critical control rod patterns and assembly exposure inventory.

Consumers Power Company first developed reload physics designs for Cycle 9 (1971) and has used GRpK since Cycle 10.

Included in these designs are loading pattern selection, cycle lifetime, cold critical control rod pattern,. rod and notch worths, shutdown margin, moderator temperature coefficient, void coefficient',

doppler coefficient, liquid poison worth, scram function, misloaded Assembly effects, and control rod withdrawal sequence for the entire cycle.

GR%K-is also used for core follow work. The fluxwire measured core power _ shape iscomparedwithGRpK'spowershape. Power distribution correction factors are then calculated for various discrete regions of the core and the maximum allow-able power level is determined. Plant personnel perform most of the core follow work.

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2.0 DESCRIPTION

OF BIG ROCK POINT Big Rock Point is a 2h0 MWt boiling water reactor located on the shore of Lake Michigan just North of Charlevoix, Michigan. Uranium dioxide as well as mixed oxide pellets contained in zircaloy tubing constitute the fuel. There are 8h assemblies of either 9 x 9 or 11 x 11 arrays with an active fuel height of 70 inches. Most assemblies contain removable corner rods. These positions con-tain either zircaloy experimental rods or cobalt target rods. Figure 2-1 shows the core layout, control rod positions and fluxvire positions.

32 B C cruciform control rods control the reactor. The 16 outer blades, groups g

A and B are strong (all poison tubes filled) and are used only for shutdown. The 16 inner blades are weak (40 inner poison tubes are empty). 8 or 10 of these weak blades are typically used to regulate the core at power and they are alternated each cycle. Power maneuvering is done entirely with control rods since Big Rock Point has external circulation and does not use flow control.

Big Rock Point is currently in its 15th refueling cycle.

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FIGURE 2-1 BIG ROCK POINT CORE CONFIGURATION N

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3 30 PHYSICS MoDEL 31 OVERVIEW The calculational sequence for Big Rock Point physics is diagrammed in Figure 3-1.

The primary. component of the sequence is the three dimensional reactor simulator program,GR%K. The remainder of the sequence primarily involves the generation of a

input for GR%K. The input consists of: assembly neutronics parameters as a func-i 4

tion of local operating conditions, physics features of each fuel. type, hydraulic parameters, technical specification and other limits, and the reactor. operating t

state.

l The neutronics parameters are derived from the results of assembly physics calcu-lations performed with such programs as CASMO, PDQ7 or XPOSE 3 These calcu-lations are performed either by Consumers Power or by the fuel vendor.

3 1.1 GRdK Overview The GR%K program is run using the appropriate option to compute the desired core parameter. Reactivity, power distribution and margin to thermal limits are computed in a ctraightforward manner by modeling the reactor. operating state being analyzed. Control rod worths and notch worths are computed by calculating reactivity at two different states (i.e., rod'in sequence and rod inserted).

Coefficients ~ (void' and -doppler).are computed with a special option that allows one feedback mechanism to be varied while the worth of the rest remain constant.

Fuel burnup is incremented by inputting a core average exposure and allocating it based on the computed power distribution.

3.2 ASSEMBLY PHYSICS CALCUIATIONS The purpose of assembly physics calculations in the calculational sequence is s

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O to determine the relationship between the physical state of the reactor and the neutronics parameters required in the GRpK physics model. These calculations are typically performed for a horizontal slice of a single fuel assembly, for which the region of solution is the area bounded by the centerlines of the water gaps surrounding the bundle. They are two dimensional with reflecting boundary condi.tions, hence the term " infinite lattice". The method of solution is few group fine mesh diffusion theory (PDQ), or transport theory (CASMO).

The inputs to the assembly physics calculations consist of the location, com-position and temperature of all the important materials within the assembly; such as fuel, burnable poison and inert rods; flow channel; control rods and coolant. Dimensions and compositions of assembly components are acquired from the vendor drawings and design reports. Structural materials are assumed to be at the coolant temperature, while the assembly average fuel and cladding temperatures are computed using a one dimensional heat transfer modei of a fuel rod as in the program GAFEX.13 The important outputs from the assembly physics calculations are: the infinite neutron multiplication factor, Kgg; the migration area, M ; the one-group flux; the ratio of kappa over nu; the fluxwire absorption factor and the local fuel pin power distribution.

The following paragraphs describe the computer programs used for assembly physics calculations.

3 2.1 CASMO CASMO is a multigroup two dimensional transport theory code for static and burnup calculations of BWR and FWR fuel assemblies. The code handles a geometry con sisting of cylindrical fuel rods of varying composition in a diagonally symmetric s

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square pitch array, with allowance for gadolinium and other burnable poisons, in-core instrument channels, water gaps and cruciform control rods. CASMO incorporates libraries of multigroup neutron cross-sections and other nuclear parameters so that the calculation is a one-step process from the users stand-point. The program performs spectrum calculations on the various compositions within the assembly and generates few group cross-sections for each. These are then used in the two-dimensional neutronics calculation, which consists of a variation of transport theory called transmission-probabilities. Using the computed fluxes and few group microscopic cross-sections CASMO depletes the fuel isotopes stepwise out to the projected lifetime of the assembly.

3 2.2 PDQ7 pT Q

PDQ7 solves the few-group neutron diffusion problem with burnup in one to three i

dimensions. Cross-section input for PDQ7 is generated using programs such as HAMMER

, EPRI-CELL or XPOSE. In the development of the cross sdctions a standard " pin cell" approach is used for interior fuel pins, while an extra l

region of water is added to the cell to account for the presence of water gaps for fuel pins on the outside row of the assembly. Diffusion theory " equivalent" I

cross sections are found for crntrol rods and lumped absorbers by adjusting absorption cross sections to match reaction rates computed with a transport theory model such as CASMO. Cross sections for water gaps and structure are derived from an exterior fuel pin spectrum. PDQ7 is used for special studies 5

that require a code with n ore flexability than CASMO such as 3D calculations

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asymmetric bundles and multi-assembly configurations.

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33 GR%K INPUT GENERATION In order to compute a three dimensional reactor power distribution, the physics algorithm in GR%K requires three dimensional distributions of the basic physics parameters K M and the ratio </v.

These gg, parameters in turn are dependent on the local operating state:

power, steam voids, control rods and the burnup history of the fuel. These dependencies are computed separately from GR%K using assembly cell physics calculations simulatingvariouslocalconditions,andtheresultsareinputtoGRgKinthe form of coefficients to polynomial functions.

For instance, to compute the dependence of K on local steam void fractions, gg assembly physics calculations are performed at three discrete void fractions, usually 0%, 25%, and 50% steam voids. The three values of K res h ng from gg these calculations are then fit to a simple polynomial in moderator density through the solution of 3 simultaneous equations resulting in:

K

= B (1 + B U + B

}

(}

gg g

7 10 s

Where U is the moderator der.sity relative to saturation, and B, B and B g

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thesolvedcoefficientswhicharetheninputtoGR%K.

Since K is also a function of control rods and the dependency on control rods and steam voids are interrelated (non-separable), six more assembly physics calculations are performed resulting in two more equations similar to (1),

one for one adjacent control rod, and another for two adjacent control rods.

For partially roded nodes, GR%K interpolates between the two appropriate evaluated equations.

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5 The reactivity worth of other reactor operating parameters such as fuel burnup, burnable poison concentration, doppler and equilibrium xenon concentration is also computed with assembly physics calculations. These worths are expressed in terms of AK/K vs nodal power, exposure or void weighted exposure, where Kl-K aK/K-M x1 is We M er m e" valm of K,, and K is the " perturbed" value where the K1 2

perturbation is due to changes in the operating parameter, i.e.,

doppler, xenon etc. The calculated value of K is then modified by a series of multipliers 00 of the fonn (1 - 6K/K) for each factor affecting reactivity. This procedure is repeated for nach individual node in the calculation resulting in the three di-mensional array of physics parameters required for the neutronics calculation.

1 Table I is a compilation of the physics parameters evaluated for each fuel type in the generation of GR%K input. Column 1 is the desired physics parameter, columns 2 and 3 indicate what physical parameters are considered in the calcu-lation of the physics parameter, columns 14 and 5 indicate the degree of poly-nomial used in GR%K to describe the relationship, and column 6 shows the number of assembly physics calculations required to generate the coefficients. Many of the assembly calculations are used for more than one purpose, so the total number of calculations performed is less than the total indicated by adding column 6.

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TABLE I Input for Power Operation Neutronics Parameter Functional Dependence Degree of Fit Minimum Number Lf A

B A

B Assembly Cale"'_etions M

Steam Voids 2

3 K

Control Rods Steam Voids 1

2 9

9 AK/K) Doppler Power Steam Voids 1

1 3

. AK/K) Equilibrium Sm Power 0

2 Xenon Cross Sections Burnup 2

3 BK/K) Xenon Xenon Concentration 2

3 6K/K) Initial Burnable Poison Control Rods Steam Voids 1

2 9

AK/K) Burnup Burnup Void-Weighted Burnup 2

2 30 Fraction of B.P. Remaining Burnup 2

6 K/v Burnup 2

10 Input for Zero Power M

Temperature 2

3 K

Control Rods Temperature 1

2 18 9g AK/K)PeakSamarium Burnup O

2 6K/K) Initial Burnable Poison Control Rods Temperature 1

2 18 Xenon, Burnup, Fraction of Burnable Poison remaining and K/v are the same as above.

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3.4 DESCRIPTION

OF' THE GR%K COMPUTER PROGRAM l

The following is a description of the calculations performed in the GR%K com -

- puter program. GRgKincorperatestheFLARE physics model along with a themal-hydraulic feedback model, a local peaking factor calculation, and routines which evaluate margin to thermal limits.

3.4.1 GRdK Physics Model GR%K calculates a nodal power density for a three dimensional core geometry.

The code is based on modified one group diffusion theory using the infinite multiplication factor (Kgg) and migration area (M ) as the basic physics inputs.

An albedo at the core surface simulates the reflector so that only mesh points within the active fuel region are considered. Each assembly.is represented by one horizontal and nine vertical mesh points (nodes).

The complete iterative calculation consists of four levels:

1.

Source or Power Distribution, 2.

Void Distribution, 3

control Rod Positioning and Power Variation, and 4.

Fuel Burnup or Thermal Limits Evaluation.

A description of each level follows. An explanation of the equations solved also appear.

3.h.l.a source or Power Iteration The neutron source at each node is ' calculated as a function of:

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K at that point, 2.

the neutron source at the six neighboring points, and 3

a transport kernel.

The transport kernel is a measure of the probability that a neutron born at node m is absorbed at node 1 and is a function of migration area and node spacing. K is calculated at each node and includes the following effects:

gg 1.

Presence or absence of one or two adjacent control rods. These control rods can be strong or weak in any combination.

2.

Local moderator density or coolant temperature.

3 Power dependent xenon and Doppler reactivity.

4.

Local fuel exposure.

5 Presence of burnable poisons.

6.

Equilibrium or peak samarium.

M is also calculated at each node as a function of local moderator density or i

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temperature. The initial values of X and M are based on a flat power dis-tribution and are updated thereafter following each void iteration.

3.h.1.a.1 Source Equations Source at code 1 is defined by:

A (3-3) 8 g

1 oog where A

absorption rate at 1*

=

g g

S,W g gg (3 h)

+ S W or A

=

• ' note: a single subscr t i or m will be used interchangeably with the sub-script ijk*

t

11

/V vhere.W

= probability that a neutron born in node m will be absorbed in node 2 add the prime indicates summation over the six nearest neighbors.

Combining these equations and dividing K by A (the eigenvalue) yields:

K W

oog m mg s

(3-5) 1 A+K (6-QW 9

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vhe:e A = eigenvalue ot= albedo g is non-zero at the boundaries of the core (<6 neighbors) only and is handled o

separately for the top reflector, bottom reflector, and for the sides of each peripheral fuel assembly.

A is recalculated after each iteration based on a solution to the neutron balance summed over the entire core:

3 Source + Inleakage - Outleakage Absorption

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S,W -)

S S

+

W (n - - a g

t)

(3-6) g g

g y,

t m

L t

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S g q

A K

• • L where ng = number of external nodes adjacent to node t, both ng and o are g

zero for all internal nodes, and the sums are over all nodes in the reactor.

Equations 3-5 and 3-6 are the basic equations solved during the source iteration.

The transport kernel is defined as:

.s....

E W g'*=

(3-7) 2r t I

where Mg

= migration area of node i r

= mesh spacing.

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A more detailed analysis of these equations are presented in Section 3 of reference h.

Several source iterations are performed per void iteration. After each series of source iterations the nodal power distribution is computed from the neutron source by multiplying by the ratio of kappa (the effective energy released per fission) over nu (the average neutron yield per fission). A polynomial function ofc/9 versus burnup is included in GRgK for each fuel type.

3.4.1.a.2 Physics Parameters - K and M g

Two basic neutronics parameters are used to calculate the reactor fission source distribution: K,, the infinite multiplication factor, and M, the neutron mi-Ol gration area. A three dimensional array of these qur,ities is generated by the program using algebraic functions that describe the dependence of K and gg 2

M on local reactor conditions. The coefficients of these functions are input for each fuel type and are derived from the results of assembly physics cal-culations performed over the range of expected local operating conditions. The K

and M arrays are reevaluated after each void iteration and used as input for the next series of cource iterations.

There are two reactor conditiont normally simulated separately using GR%K:

zero power and power operation. For zero power calculations, M is calculated as a quadratic function of isothermal reactor tempersture ranging from ambient condi-tions up to the operating saturation temperature. For power operation calcula-tions, M is computed as a function of in-channel steam voids ranging from zero varieswithvoidfraction,controlO to core exit void fraction at full power. K9 e

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'v rod configuration, burnable poison concentration, fuel exposure, fuel tempera-ture (nodal power) and xenon and samarium concentrations.

For zero power calculations K, is required as a function of reactor temperature g

for the following combinations of control rods inserted into a node:

1. ' no control' rods, 2.

one weak control rod, 3

one strong control rod, 4.

two weak control rods, 5

two strong control rods, and 6.

one weak and one strong control rod.

The core locations of the strong versus weak rods do not change and therefore have been built into the program. Since the strong control rods are withdrawn before significant power levels are achieved, K

's for power operation are gg computed as a function of steam voids for the following fractions of control only:

1.

no control rods, 2.

one weak control rod, and 3

two weak control rods.

If the fuel type in question contains burnable poisons, coefficients to polynomials

/K, due to the burnable poison as a function of '

are input that describe AK l

9 control density and moderator condition in the same manner as K Since the gg burnable poison typically does not cover the entire active length of the assembly, parameters are input that describe the axial position of the poison.

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.... m

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Punctions are also provided that describe the reactivity defect due to xenon as a function of nodal power. The equilibrium xenon-135 and iodine-135 concentra-tions are calculated for each node according to:

x) & b P' S (d

+e f

g 7

(34) eqt X

4 e; ${f P.S g (3-9)

I

=

eqt A I Where P is the reactor power and A is a decay constant. Thequantities$[f)'7, the iodine yield at rated power in an average node; $ b e X, the xenon yield; f

& c,X, the xenon rbsorption, rate, are calculated as a function of fuel and burnup based on input coefficients for each fuel type. Transient xenon after shutdown is calculated from the following:

I

-AXt

-AIt}

^I '9A(*

AXt (3-10)

Xtt eqL X

e

+

=

A Ay I

Where t is the time after shutdown. The reactivity worth in dK/K of xenon is then computed as a function of the calculated xenon concentration using input coefficients.

The program also accounts for the reactivity defect due to samarium. For power operation calculations an equilibrium value is used, while for zero power calcu-lations either the peak samarium defect is used or zero defect depending on whether the assembly has been previously burned or is fresh fuel.

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1 The reactivity defect due to doppler is computed as a function of local power and steam voids. The two basic relationships used to calculate this function are:

1.

the variation of effective fuel temperature with LHGR, and 1

2.

the reactivity defect from doppler as a function of fuel temperature at i

i different void conditions.

GR%K calculates the doppler defect as a linear function.of power multiplied by a linear function of steam voids as specified by input coefficients for each 1

fuel type.

The K,g vs exposure and void-weichted exposure equation is of the following form:

l

.(3-11)

I gEg' 66 22 g (1.0 - B23 t ) - 3 B

E 24 1 (l'O - 325 i )

l E

=

Y Y

g Burnup B26 +

i i

where f

Eg = exposure at node t Vg = void history at node 1 (product of exposure and voids) l The coefficients for equation 3-11 are calculated by a two dimensional least I

squares fit of computed K versus burnup out to at least 1.5 times the expected j

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averaBe discharge exposure for three different void conditions (usually 0, 25 and 50%). For burnable poison fuel, a polynomial function is provided that describes l

the burnup-dependent behavior of the reactivity worth of the, poison.

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3.4.2 Void Iteration The void model in FLARE has been entirely replaced by a more sophisticated ther-mal hydraulics model. It requires the core active coolant flow, coolant inlet enthalpy, reactor vessel outlet pressure, total reactor power, and the nodal relative power distribution calculated in the source loop. Spacer loss coeffi-cients, core inlet friction factors, hydraulic diameters, wetted perimeters and thermal hydraulic model selections are required user input data. These parameters are obtained from analyses and test data from both the vendor and Consumers Power. The reactor core is hydraulically modeled with closed flow channels each containing one fuel assembly.

An iterative solution technique is used where the total core active coolant flow is uniformly apportioned to each flow channel in the reactor core. The total pressure gradient, made up of components due to friction, momentum, elevation, i

and local losses, through each flow channel is then calculated using'this esti-mate.

The average pressure gradient for the reactor core is then calculated and each channel flow rate is modified as follows:

\\

Wnew (I,J) = Wold (I,J)

(3-12) p y) where: Wnew (I,J) = New estimate of flow rate in flow channel (I,J).

Wold (I,J) = Previous e, stimate of flow rate in flow channel (I,J).

Pave = Average pressure gradient for the reactor core.

P(I,J) = Pressure gradient of flow channel (I,J).

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./U Af ter the channel flow rates are modified they are summed and normalized to the known total core active coolant flow. This new estimate of channel flow rates is then used to recompute the flow channel pressure gradients. The pro-1 cedureisrepeatedforeachpower/flowiteration.

Nodal steam void fraction is calculated from nodal quality using the correlation presented as equations 3-17 of reference 4 and is converted to relative modera-tor density. This is needed to calculate K and M for the next series of g9 source iterations. The void and source iterations are continued until a con-verged power and void distribution is attained.

3.h.3 control Rod Positioning and Power Variation GR%K has a control loop which will automatically adjust control rod groups one e

notch at a time in any given input sequence until the core reactivity equals a pre-set value. An automatic search for the reactor power required f6r criti-cality at any given control rod position is also available.

3.4.4 Fuel Burnup GR%K updates nodal fuel burnups using the computed power distribution, the assembly weights as input by fuel type, and the core average exposure increment for the burnup step.

3.5 PEAKING FACTOR AIDORIThM Thepeakingfactorcorrelation'sfunctioninGR%Kistofindpeakpinpowhsfor each assembly, which are necessary inputs to the MCHFR, MCPR and heat flux cal-culations. The code first takes the coarse global power shape and finds a 7

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e w

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polynomial function describing the axial power shape in each node via a SPLINE curve fit. These functions are evaluated at four points per node. This method describes in finer detail the axial power variations that occur over the length of the assembly and results in a computed power for every two inches of core height. These will be referred to as expanded or sub-nodal powers.

The local fuel pin power is then obtained by multiplying the expanded nodal powers by a local peaking factor. These local peaking factors are calculated by combining the gross horizontal power tilt across an assembly at each axial sub-node with the beginning of life (BOL) assembly infinite lattice local power distribution. The infinite lattice local power distributions are derived by means of two dimensional assembly physics calculations for each fuel type, and are input for three assembly control conditions:

1.

uncontrolled, 2.

singly weak controlled, and 3

doubly weak controlled.

The program will orient the local power distributions for controlled assemblies based on which corners of the assembly are adjacent to control rods. If a sub-node contains a control rod tip, the controlled power distribution is chosen.

For the sub-node immediately above the control rod tip, the following equation is used to calculate the pin powers:

PP = 0.h7 x PP + 0.M x PP y

c where PP = pin power u = uncontrolled c = controlled r

7 19

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C.

This equation is conservatively based on a three dimensional physics-calculation of.the power peaking at the tip of a control rod. E i

The gross horizontal power tilt across each assembly at each sub-node is calcu-lated by fitting the one group neutron flux in the assembly with the flux in the assembly on either side to a quadratic equation. This equation is' solved at the center line of eaca w of fuel pins in both the X and Y directions and the result is multiplied by the infinite lattice locals to yield total local peaking factors. For assemblies on the edge of the core a straight line inter-polation is performed with the one adjacent assembly.

j i

i The flux is calculated from the sub-nodal power and polynomial equations that l

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describe the variation of the flux to power ratio ( s' ) as a function of steam voids, control and fuel exposure. The coefficients for these equations are derived from the assembly physics calculations and are input to GR/K.

i 1

After the nodal local pin powers are calculated, the program selects the highest The highest value between the maximum infinite lattice local power and the j

one.

maximum total local power including tilt is then chosen as the local peaking fac-tor for that sub-node and is later used in the calculation of peak heat flux.

3.6 FLUXWIRE CALCUIATIONS GRpK includes a fluxwire activation computation for comparison of the calculated core power shape with fluxwire measurements. Coefficients to functions that

' describe the ratio of the neutron activation in a copper fluxwire located in

..p-the corner of an assembly to the assembly power as a function of steam voids, b

r 7

.,,,e

.n.,en--

~m-

+-+-a----

20 control rod position, and assembly exposure are input for each fuel type. This data is obtained from two dimensional assembly physics calculations. The results from the four assemblies adjacent to the fluxwire position are averaged and the axial shape is printed out. The total activations for the eight fluxvires are normalized and printed out also. The fluxvire positions can be seen in Figure 2-1, 37 THERMAL HYDRAULIC LIMITS A thermal limit calculation to determine the peak heat flux and the minimum critical heat flux ratio (MCHFR) in the core is performed by GR%K. Local peak heat fluxes for each sub-node are calculated based on total core power, normalized power shape, local peaking factors and the heat transfer area for each fuel type. Hot channel flow reduction and enthalpy rise factors are included in this calculation. The synthesized Hench-Levy correlation is then applied to yield critical heat flux and MCHFR. Another thermal limit found is the minimum critical power ratio (MCPR) as calculated by the Exxon Nuclear XN2 correlation. 7 MCPR is expected to replace MCHFR as a technical specification limit in the near future, and at that time MCHFR will no longer be calculated.

The ratio of calculated to allowable maximum average planar linear heat generation rate (MAPLHGR) for each sub-node of every assembly is found based on total core power, normalized power shape and the limits for MAPLHGR input by fuel type as a function of exposure. The maximum fraction of MAPLHGR limit for each assembly and fuel type is then determined as shown by equations 3-lh and 3-15

(

}

Z=

(3-lk) 137.1441 (PIH) where Z = maximum allowable radial times axial for each quarter node F = active fuel length.

,n a

21 l'O B = number of fuel rods per assembly

.X = MAPLHGR limit at each node based on exposure of that node and a 1

linear interpolation between input MAPLHGR limits PTH = core thermal power

• 9

• 1 g,

MW 84 assemblies where.96 = ECCS gamma smearing factor.

Fraction of MAPLHGR limit = f (3-15) where P = actual radial times axial.

GRpK computes the duration of time it takes each assembly to uncover during a loss-of-coolant accident and compares this to the time calculated in the LOCA O

i analysis. Actual assembly dryout. times are calculated by using the core thermal I

power, the radial power distribution and the Modified Armand steady state void fraction. These times are compared to the dryout times assumed in the LOCA analysis corresponding to the MAPIEGR limit which is most closely approached in each assembly. The following set of-equations are used to calculate dryout times. 9

1. 8 at =

'(3-16) q, 3

where 6t = time to boiling transition 9f = saturated liquid density at initial operating pressure h

= enthalpy of vaporization at initial operating pressure fg

q.,

g lV(1 - a )

O V

f =~ i itial steady state power generation in the bundle where V.= active coolant volume n

.V.

p.~,k

\\"

t

.I

22 O

ag = initial steady state void fraction averaged over entire fuel rod length (as calculated by the Modified Armand model)

Equation 3-12 reduces to:

7920.7168 at =

q,,.

(3-17)

MAPLHGR limits are from the Technical Specifications as derived from LOCA analyses.

9 a

0 i

s

]

9 A9

e i

FIGURE 3-1 BRP PHYSICS CALCULATIONAL SEQUENCE PHYSICAL PROPERTIES PELLET,- ASSEMBLY, CORE V

V CASMO VENDOR DATA OR 2

K,.,M

, PEAKING XPOSE FACTORS, ETC.

If ifV INPUT PREPARATION CURVE, FITTING, ETC v

REACTOR OPERATING GROK STATE, POWER, ROD U

CORE SIMULATION POSITION, BURNUP

-> POWER DISTRIBUTION

--> MARGIN TO THERMAL LIMITS

-> ROD & NOTCH WORTHS

-> COEFFICIENTS

-> lNCREMENTAL BURNUP

-> REACTIVITY O

4 e

w

.~

9

.----g-.me g--,,w-e-a y

g-4

,e.,

p.

gvy,

_wu, w

---,__,._m__

~

23

(

4.0 VERIFICATION OF THE BIG ROCK POINT PHYSICS MODEL The results achieved from the BRP Physics Model are examined in this section.

Comparisons with reactor operating data are made for the BOC cold critical control rod patterns and fluxwire measurements. Also, the calculated core multiplication factor versus core burnup is presented. Data is presented from the last two operating cycleo as the physics models used in the analysis of those cycles most closely represent the current procedures as outlined in this report.

4.1 BOC COLD CRITICAL CONTROL ROD PATTERN BOC 14 and 15 actual and predicted cold critical control rod patterns are shown in Figures.4-1 and 4-2.

Comparisons of actual to predicted cold critical rod patternsshowthatGR/Kpredictedcorrectlythetotalnotcheswithdrawnatthe

['T beginning of cycle 15 and was within three notches for cycle 14.

%.l 4.2 AXIAL PROFILES The ability of GR%K to predict axile power profiles is verified by comparing actual fluxwire results to profiles generated by GRpK. An explanation of the techniques used by GR/K to generate power profiles at the.various fluxwire loca-tions is given in Section 3.6.

Comparisons of fluxvire profiles to those of GR%K are shown in Figures 4-3 to 4-34. Included in these figures are compari-sons for BOC 14, EOC 14, BOC 15 and MOC 15 These Figures show consistantly good agreement between the calculated and measured shapes.

O

24 43 core MULTIPLICATION FACTOR.VERSUS EXPOSURE Figure 4-35 shows the calculated core multiplication factor as a function of exposure-for cycles 14 and 15 This figure indicates that the physics mcdel for Big Rock Point is consistant in its representation over the wide range of l

. core conditions existing during the course of a cycle, including variations in.

control rod density, burnable poison concentrations, void density, reactor power, and fuel exposure. The low calculated reactivity at the beginning. of each cycle is attributed to samarium and -xenon nonequilibrium conditions.

O t

4 W

0 s) s as-

--y

-e e..v.g+-,

y e

, - -. = - -

e-e--

,, w e.- e rp e..w,.e~.

~.- ~ ~..

,.3-%-

FIGURE 4-1 ACTUAL VS PREDICTED BOC 14 COLD CRITICAL

?

ROD POSITIONS (23=OUT) 9 A

B C

D E

F 1

5 7

5 6

ij 2

6 0

0 0

0 5

l.

f 3

5 0

0 0

0 7

.4 7

0 O

0 0

5 5

5 0

0 0

0 6

6 6

5 7

5 O

PREDICTED CRITICAL TOTAL NOTCHES WITHDRAWN = 92 A

8 C

D E

F 1

5 7

5 7

2 7

0 0

0 0

5 i

l 3

5 0

0 0

0 7

4 7

0 0

0 0

5 5

-5 0

0 0

0 7

..l l 6

6 5

7 5

! 1 I

I I;

i ACTUAL CRITICAL TOTAL NOTCHES WITHDRAWN = 95

FIGURE 4-2 ACTUAL VS PREDICTED BOC 15 COLD CRITICAL ROD POSITIONS (23=OUT)

A B

C D

E F

1 5

5 4

5 2

5 0

0 0

0 5

l 3

4 0

0 0

0

'S 4

5 0

0 0

0 4

5 5

0 0

0 0

5 6

5 4

5 4

PREDICTED CRITICAL TOTAL NOTCHES WITHDRAWN = 81 A

B C

D E

F 1

5 5

4 5

)

2 5

0 0

0 0

5 1

3 4

0 0

0 0

5 4

5 0

0 0

0 4

5 4

0 0

0 0

5 l

6 5

4 5

5 O

ACTUAL CRITICAL TOTAL NOTCHES WITHDRAWN = 81 y

,. e -

u-

..wi,

+

0 9'

p FIGURE 4-3 NORMAllZED AXIAL ' PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76 GROK._

b LOCATION NO 1 is 1.6 1,4 x

0 0

asee2gOsagogg O!

n a

60 ti O O d

LO go 6-

=

g O

o

.0.8 O

O O

O o

06 0o o

O 0.4 g

0.2 i

e i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

FIGURE 4-4 NORMALIZED AXIAL PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76 GROK-b LOCATION NO 2 1.8 i

1.6 h {OS QO 1.4 00no x

3 oO 5

g.

1.2 og%sa m

o e

a s

0 0

d 1.0 3

0 O

o O

b Q

. 0.a 2

8 0

0.6 8

[

0.4.-

O O

, 0.2 e;

1 t

1

.40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH).

8.

e.

FIGURE 4-5 NORMALIZED AXIAL ' PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76 GROL.

8 LOCATION NO 3 1.8 1.6 1.4 5

SgO 6

e 1'2 6hoo O

00 O

me o

oo@

4 O

6 g

.10 0

O' b

oO O

_0.8 O

O o

a o O

o

. 0.6 O

o o

O 0.4 6

0.2

.e i

f 1

I t

i i

i 40 60.

80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/'2 INCH)

l FIGURE 4-6 NORMAllZED AXIAL PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWlRE DATA OF 9-28-76 G R O K...

b LOCATION NO 4 1.8 1.6 1A OOOhog x

g 1.2 6

0 O

0 O o E

o 06 4

O bo gj

.1.0 o

OO e

O 0

. 0.8 h

O O

go 0.6 -a O

g O

o O

0.4 o

0.2.

o h

40 60.

80 10 0 120 14 0 160 180 FLUXWtRE UNITS (1/2 INCH)

.l FIGURE 4-7 NORMAllZED AXIAL PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76

)

GROK_

O LOCATION NO 5 1.8 l

I.6 1.4 E

'2

• *0 O

^

6 og

.t0 0

=

go e

.0.8 6

g O0 O

0 0.6 O

6 O OA O

0.2

~

1

' ~

't I

i t

1 i

i I

-40 60-80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

=

FIGURE 4-8 NORMALIZED AXIAL PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76 GROK_

d i

LOCATION NO 6 1.8

')

1.6 1.4 x

b 1.2 66 63ghQOo6@c g

6 0000 0

0 h

d t0 o0 O

e O

o g

g gO

. 0.8 O

O 0.6 O

l 0.4 O

l l

g.

0.2 r

i i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

1

/

FIGURE 4 -9 NORMAllZED AXIAL PROFILE BIG ROCK POINT l

BOC 14 FLUXWIRE O

FLUXWIRE DATA

'~

~

GROK-d LOCATION NO 7 1.8 1.6 1A O

O x

O =E

^So 0^^8cena no o

5-O 0 00 bgO 10 60 O

c 60 0

. 0.8 3

8 0

O o

O 0.6 b

a 0.4 O

0.2 Pd 40; 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH) s.

FIGURE 4 -10 NORMALIZED AXf AL PROFILE BIG ROCK POINT BOC 14 FLUXWIRE O

FLUXWIRE DATA OF 9-28-76 GROK_

d LOCATION NO 8 t,

12 1.6 j

1A 6h x

21.2 Q

bo g'

O hogggg$b o

O g

Q bO Oo O

~to' E

6 OQ O

g o

o O

o

_0.8 O

O 6

0 O

-)

0.6 6

g O

o.4 6

0.2 i

i i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

i FIGURE 4-11 NORMALIZED AXIAL PROFILE BIG ROCK POINT l

EOC 14 i

FLUXWIRE O

FLUXWIRE DATA OF 4-18-77 GROL o

LOCATION NO 1 i

... l.8 1

1.6 1.4 88806 s

a#

O

(

u.

1.2 -

geO m>

e 0e a0 a

't0 E

bO o'O 60 oO oO 0.8 oO O

60 O

0.6 O

o O

O M

6 o

O 0.2 i

i i

e 40 60-80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH) i

l FIGURE 4-12 NORMALIZED AXIAL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FluxwlRE DATA

~~

GROK O

LOCATION NO 2 1B 1.6 1.4 OggOooD x

0 3'

a66 8

g 0

s 200 60 o

g to o

3 e

a g

60

. 0.8 l

6O O

60 6

0.6 1

O O

O o

04

(

l 0.2 o

e c

(,

40' 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

i Q

FIGURE 4-13 NORMALIZED AXIAL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FLUXWIRE DATA OF 4-18-77 GROK.:

O LOCATION NO 3 1.8 1.6 j

1.4 Bob D$

O OQ y

ob O

u.

1.2 Jw OO bO 6g@DO 60 gj

. l.0 0

oO 0

60 o

.0.8 no 3 0 O

0.6 O

O e

O 0.4 0.2 i

i i

i n

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

=

FIGURE 4 -14 NORMALIZED AX1AL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FtuxwlRE DATA r

OF 4-18-77 GROK.

8 LOCATION NO 4 1.8 1.6 1.4 bo y

O gh d

1.2 -

@ehO s

O E

6 8

Ei

.LO a

x a

b O

l 0.8 oO O

l OO o

0.6 O

6 O

o o

0.4 D

0.2

~

f f

1 t

I f

a g

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH) i

FIGURE 4 -15

,q NORMALIZED AXIAL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FLUXWIRE DATA OF 4-18-77 GROK._

d LOCATION NO 5 1.8 14 i

1.4 Be@a 80 e@

x o

a a

O:'

neO 2

~

b h

O I

d

.1.0 m

O o

bO o

0.a 60 O

O O o

60 0.6 O

.o j

O o

I 0.4 D

' O.2 1

i i

i

-40 60 80 10 0 120 14 0 -

160 180 j,,'

i FLUXWIRE UNITS (1/2 INCH)

s FIGURE 4 -16 NORMAllZED AXIAL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FLUXWIRE DATA 4~

~

GROK-d LOCATION NO 6 j

1.8 1.6 1.4 o2228e a

bg a

b 1.2 g

s O

002 o@a 5

O O

g t0 g

cc oO O

_ 0.8 n o a

60 O

0.6 O

o o

O O

0.4 g

+e 0.2 e

i i

i i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

p FIGURE 4 -17 NORMALIZED AXIAL - PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FLUXWIRE DATA OF 4-18-77 i

GROK_

b LOCATION NO 7 ~

_ 18 1.6 l

l l

1.4. -

- i gGS 20 0

n 1.2 3

V 0

g 6

0 30 oo e

oe

i to 6

O 600 O

60 6

0.8 OO O

O O bo o

0.6

~

O O

o O

o 0.4

- g 0.2 O

V 40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/'2 INCH) u

-n---se-------

4 9-g FIGURE 4-18 NORMALIZED AXIAL PROFILE BIG ROCK POINT EOC 14 FLUXWIRE O

FLUXWIRE DATA OF 4-18-77 GROK_

d LOCATION NO 8 I

1.8 16 O

O 1.4 -

b O

aQ x

60 s

n O

g 1.2 -

u.

!,n

@o

?8 o

o O

60 O g

. 0.8 O

O O O g

a 0.6 O

O O

o O <-

g o, _e a Art =

0.2 i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

3 FIGURE 4-19 w/

NORMALIZED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWIRE O

FLUXWIRE DATA.

OF 12-19-77 GROL o

LOCATION NO 1 1.8 1.6 1.4 3

pu 1.2 l

oho O

SSS8@@@gg d

_t0 6

o m

O QO o

o O

O

..a 3

o j

0 O

O O

0.6

-Q O

o 0.4 4

0.2.

O V-i i

i 40-60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH).

FIGURE 4 20 NORMAll2ED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWIRE O

FLUXWIRE DATA -

O F 12-19-77.

GROK_

d LOCATION NO 2 12 1.6 1.4 Ob ti o

g gj

.t0 o

o b O oo 0.a 60 O

o O

O 0.6 0

o 3

O 0 0 0.4

~

O 01 -

i i

i e

i 40 60 80 10 0 120 14 0 160 180 i

FLUXWIRE UNITS (1/2 INCH)

- - - - - +

-wgw

---p.

r w-y wg e--

.mb-----

.-u---~-

FIGURE 4 21 V

NORMALIZED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWlRE

_O_

FLUXWIRE DATk OF 12-19-77 GROK.

8 LOCATION NO 3

.1.8 1.6 1.4 000 6 60 5

0 00 ~

60 eo

~

s O

628 8666 g

60 6

60 6o 0000 6

d

. l.0 O

60 06 O

Oo

. 0.8

- o On 0

O 0.6 O

0.4 o

02 -

O 40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/'2 INCH)

g FIGURE 4-22 NORMAllZED AXIAL PROFILE BlG ROCK POINT BOC 15 FLUXWlRE O

FLUXWIRE DATA O F 12-19-77 GROK_

d LOCATION NO 4 l

.1.8 1.6 t

i f

1.4 x

O 39862 ggRed an el 2

'2 -

E OO 4

O 6

y to 9

9 o

O

.0.8 o

O O

O o

i O

)

0.6 O

oO

)

0.4 o

0.2 e

I f

f i

f f

I I

I 40 60 80 10 0 120 14 0 160 180

)

FLUXWIRE UNITS (1/2 INCH)

o FIGURE 4-23 NORMAllZED AXIAL PROFILE i

BIG ROCK POINT BOC 15 FLUXWIRE O

FluxwlRE DATA OF 12-19-77 GROK_

d LOCATION N0 5 l

1B.

1.6 1.4 x

of ggg20e e*

l 9

0a22sso*

O 6

Gj

.1.0 o

bo

.b x

60 60 0.8 O

l 6

S O

O.6 o

O O

0

\\

OA

~

0.2 O

I I

9 1

i i

3 l

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

FIGURE 4-24 NORMAllZED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWIRE O

FLUXWIRE DATA OF 12-19-77 OK_

d LOCATION NO 6 1.8 1.6 1,4 x

5 gg g

g,

.1.2 g

obO b

E Qo 3

o 0

8 m

b O

_0.a O

O S

0 0,6 0.4

. 0.2 i

i i

l 40-60 80 10 0 120 14 0 160 180 FLUXWIRE UtilTS (1/2 INCH) t

g a

/~].

FIGURE 4-25

'V NORMALIZED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWIRE O

FLUXWIRE DATA

~

GROK~

d LOCATION NO 7 1.8 1.6 1,4

- E o e@8 M

u.

.1.2 -

g@OO 00 g

oc 60

.1.0 w

cc o O o

o g

. 0.8 O

O

.60 00-O 0.6 oO o

0.4 i

0.2 -

t

' 'w i

t t

t e

i g

e 40 60 80 10 0.

120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH) v.

FIGURE 4-26 NORM AllZED AXIAL PROFILE BIG ROCK POINT BOC 15 FLUXWIRE O

FtuxwlRE DATA OF 12-19-77 GROK_

d LOCATION NO 8 18 1.6 1.4 0

x gdO 6

g 1.0 0

8 600 b

.0.8 O

0 3

8 O

0.6 b

O O

o 0.4 01 -

i i

40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

(--

i FIGURE 4-27 k/

NORMALIZED AXIAL PROFILE BIG ROCK POINT MOC 15 FLUXWIRE O

FLUXWIRE DATA OF 4-3-78 GROK.

8 LOCATION NO 1

. ts.

1.6 1.4 E

QOOOO o

Q 12 -

a 3Q g

a 60 00 q

08 l

bO d

.t0 O

oO o

O

. 0.a 30 o

S

~

O 0.6 n

O 6

O O

o, o

0.2 i

i i

i i

e

,. 40 ~,

60 80 10 0 120 14 0 160 180

...y M~

FLUXWIRE UNITS (1/'2 INCH)

,4 4

. -, - -, - ~

~,<r

'=

~

?-

FIGURE 4 -28 g

' NORMALIZED AXIAL PROFILE BIG ROCK POINT MOC 15 FLUXWIRE O

FLUXWIRE DATA OF 4-3-78 GROK_

d LOCATION NO 2

.1.8 1.6 1.4 E. i.2 0

og g

ggOD 3

W O

bho 0

10 ~

g oO 06 O

O9

. 0.a 9

o o

0.6 g g

8 0.4 O

o 0.2

.40 60 80.

10 0 120 14 0 160 180 FLUXWIRE, ONITS '(1/2 INCH)

FIGURE 4-29 NORMALIZED AXIAL PROFILE I

BIG ROCK POINT o

MOC 15 FLUXWIRE O

FLUXWIRE DATA OF 4-3-78 OK.

8 LOCATION NO 3 1.8 1

I O

I 1.4 O

OO y '

1.2 Q

OookO 6

i w

O

'w 6

2 O

h ObDa d

.t0 a

8 n

2 O

0.8 60 S

2 O

0.6 o

O O

o O

0.4 o

0.2 O'

I I

1 1

t i

e t

40

.60

-80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

[

s i

g FIGURE 4-30 NORMALIZED AXIAL PROFILE BIG ROCK POINT MOC 15 FLUXWlRE O

FLUXWIRE DATA

~~

GROK~

d LOCATION NO 4

.1.8

.1.6 1.4 g

oo O

9

^828ag2 SO 68 i

a0 Q

gi lo k

o OOg

.0.8 b O

@O 6

6 0

OO 0.4 b

0.2 i

i i

40 60 80 10 0

'120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

~!

C)Y FIGURE 4-31 NORMALIZED AXIAL PROFILE 1

B G ROCK POINT MOC 15 FLUXWIRE O'

FLUXWIRE DATA OF 4-3-78 GROK.

6 LOCATION NO 5 18 j

16 1.4 x

g b

e o 00 d

to o

30 o

4 60 m

O O O

O

. 0.8 k

0'0 i

0.6 OO 0.4 6

6 i

0.2 g

i t

t I

i a

f

.l 40-60 80_

10 0 120 14 0 160 180

~ FLUXWIAE UNITS (1/2. INCH)

i g

FIGURE 4-32 NORMAllZED AXIAL PROFILE BIG ROCK POINT

. AOC 15 FLUXWIRE O

FLUXWIRE DATA.

~~

GROK" d

LOCATION NO 6 1.8 1.6 1.4

~

g % sassss l'

s e

ago O

60 a

u O

wo

~

O O

O g

_0.8

- o O

O 0.6 0 M

0 O

0.4 O

0.2 I

1 1

1 e

i t

i a.

f 40-oO 80 10 0 120 14 0 160 180

~

FLUXWlRE UNITS (1/2 INCH) i,

I

(

FIGURE 4-33 NORMAllZED AX1AL PROFILE BIG ROCK POINT MOC 15 FLUXWIRE O

FLUXWIRE DATA GROK-O LOCATION NO 7 -

18 1.6

]

1.4 O

Oo 5

000ggOO O

O u.

1.2 o

6 V

0 0

O O

1.0 m

6 O-6 O O

o

. 0.8 OO k

On5

' O.6 QO O

0.4 O

O.2 y

i 40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

7

}

e j

g FIGURE 4-34 NORMALIZED AXlAL PROFILE BIG ROCK POINT MOC 15 FLUXWIRE O

FLUXWIRE DATA OF 4-3-78 GROL d

LOCATION NO 8 1

18 l

1.6 1.4 x

gdh

.i.e.

O oO o,

O 60 gO j

e o

l 1

O o

. 0.8 6

6 O.6 O

o 0.4 0.2 i

i 40 60 80 10 0 120 14 0 160 180 FLUXWIRE UNITS (1/2 INCH)

~.

(

4 FIGURE 4-35 i

CALCULATED CORE MU LTIPLIC ATION FACTOR VS BURNUP-I s

CYCLE 14 CYCLE 15 1.00 e.-,,.

I bg.99

.98 i

O 1

2 3

4 5

6 O

1 2

3 4

5 6

CYCLE BURNUP (GWD/ST) j h

I i

)

t

~

~e v+

6

^ ' -

)

l 25 50 C iRY AND CONCLUSIONS The reactor. physics methods employed at Consumers Power Company are very similar

- to methods used elsewhere in the industry. The computer models are or are de-rived from widely accepted codes which. are well tested and documented.

1

- Agreement with measured data has demonstrated the' accuracy and applicability of the methodology. Reactivity is consistently predicted at both cold and hot operating conditions, and power distributions agree well with the measurements, indicating that the various neutronic effects are being properly modeled.

i I

1

=

l l

l f

i

'I i

k 9

V

+-y

,-r..

s-4 e

,->ec e

e w.

-,.,a,

-gs--,,r-

---y,w

.e e

-,--,,wr g

--W

]

REFERENCES 1.

A. Ahlin and M. Edenius, "CASMO - The Fuel Assembly Burnup Program", AE-RF-76 4158,' ABATCMENERGI, Sweden (1976).

2.

W.R. Cadwell, "PDQ-7 Reference Manual", WAPD-TM-678 (1967).

3 F.'J. Skogen, "XPOSE - The Exxon Nuclear Revised LEOPARD", XN-CC-21, Revision 2, Exxon Nuclear Company (April 1975).

4.

D.L. Delp, et al, " FLARE - A Three Dimensional Boiling Water Reactor Simula-tor", GEAP h598 (July 1964).

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" Jersey. Nuclear Company Development Fuel Assemblies for Loading in the Consumers Power Company Big Rock Point Reactor Cycle 9", JN-70-1, Jersey Nuclear Company, (November 15, 1970), pp. 100-105, corrected via conversation between SVanyolkin-burg and GFPratt, Consumers Power Company, (August 29,1977).

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K. Galbraith and J. Jaech, "The XN-2 Critical Power Correlation", XN-75-34, Exxon Nuclear Company, (August 1, 1975).

/'

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A.A. Armand, "The Resistance During the Movement of a Two-Phase System in Horizontal Pipes", Translated by V. Beak, AERE Trans 828. Izvestiya Vsesojuznogo Teplotekhnicheskogo Instituta (1), pp 16-23, (19h6).

9

" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K", NED0-20566, (January,1976).

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J.E. Suich & H.C. Honeck, "The HAMMER System", DP-1064, January 1967.

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W.J. Eich et al, " Advanced Recycle Methodology Program System Documentation",

EPRI-RPil8-1,(October,1976).

13 K.P. Galbraith, "GAPEX, A Computer Program for Predicting Pellet to Cladding Heat Transfer Coefficients", XN-73-25, Exxon Nuclear Company (August 13,1973).

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