ML20030A347

Chapter 8 to Final Hazards Summary Rept for Big Rock Point, Nuclear Thermal & Hydraulic Characteristics

ML20030A347
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Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/14/1961
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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NUDOCS 8101090361
Download: ML20030A347 (20)

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SECTION 8 I

THERMAL. & HYDRAULIC CijA RACTERISTICS

'I

NUCLEAR, r

l

8.1 INTRODUCTION

b.

8.1.1 Core Design IObjecdves 1

i The initial core of the Big Rock Point reactor is designed E

for an average power density of 45 kw per liter of active I

core at an operating pressure of 1050 psia to produce 157 The life of the fuel is designed for 10.000 MWD per Mwt.

short ton of contained elemental uranium. Design of the initial core is in accordance with the criteria indicated in 1

the following paragraphs for achieving these stated objec-tives. ' Operating conditions as described in Section 10 of this report will be compatible with the capabilities of this g

initial core design.

8.1. 2 Core Design Limitations t

The present core power level is limited to'157 thermal

m. n watts; the description of the physical composition How-core is given in Section 4. 2 of this report.

of 4- _.11 components of the plant are designed for a systcm evei.

pressure of 1500 psia and a power of 240 thermal mega-4 The physi _ cal _ composition and esablishment of watts.

nuclear, thermal and hydraulic characteristics of the ~

subse'quent cor'e[(capable ofl pro'ducing_230_1heatmalmgga -

watts) is to be determined at. a future date depending upon the successful achievement of the objectives of the assd-

~-

The initial

{

ciated research ano development program.

phase of testing in that program is presented in Section 10 of this report.

g THERMAL AND HYD.RAULIC CHARACTERISTICS OF CORE

8. 2

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Design Criteria for Initial Core Load 8.2.1 I

The thermal and hydraulic design criteria afi eting core b

design and operation are maximum fuel temperm re, thermal

8. 2.1.1 burnout and hydraulic stability.

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8. 2.1. 2 The-aaximum fuel temperature criterion used in present design practice is to avoid mel'ing of the UO2 and thereby

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avoid possible effects upon the integrity of the enclosing stainless steel jackets. Thus, a limit is placed upon the ture of the hottest fuel rod at maximum fuel center tempe

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over-power cce citions to avoid fuel melting.

The minimum burnout margin assumed acceptable in the l

8. 2.1. 3 design of the Big Rock Point reactor was 1. 5 at maximum permitted overpower conditions.

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00/oy036

i Section 8 Page 2

8. 2.1. 4 The hydraulic stability of the reactor is demonstrated by a frequency response-phase shift analysis.

8.2.2 Powor Distributions

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8. 2. 2.1 The power distributions, as described by the power peaking factors are the major pararmters affecting the fuel tempera-ture and burnout margin. The present maximum total design power peaking.(heat flux) factor of 4.1 is made up of the prod-uct of axial, radial, local, and overpower contributions as shown in Table 8.1.

TABLE 8. I p~

DESIGN DATA FOR INITIAL CORE (a) Reference Rating Gross Electric Power - MW 50 Reactor Power - MW 157 Reactor Pressure, psia 1050 Turbine Pressure, - psia 1015 Feedwater Temp.

• F 346 (b) Fuel & Core Assembly Fuel Material UO2 235 Initial Enrichment - % U

3. 2 Total Weight of UO in Core - lbs 19,300 2

Standard Fuel Rod Diameter - in.

O.388 Standard Fuel Pellet Diameter - in.

O.345 Corner Fuel Rod Diamet er - in.

O.350 Corner Fuel Pellet Diameter - in.

O.283

(

Fuel Rod Clad Material 304 Stainless Steel Clad Thickness, Standa rd Rod, - in.

0. 019 Clad Thickness, Corne r Rod - in.

0. 0 31 Standard Fuel Rod Active Length - in.

70 k

Section 8 Page 3 Segmente.d Corner Rod, Active Length (4 rods / bundle) in.

59 Number of Std. Fuel Rods Per Bundle 13 2 Number of Corner Fuel Rods Per Bundle 12 Number of Fuel Bundles with Channels in Core 56 Equiv. Core Diameter - in.

62.5 Approximate Number of Channels. S. S.

28 Approximate Number of Channels, Zr-2 28 Channel Wall Thickness (Zr-2) - in.

0.100 Channel Wall Thickness (SS) - in.

0.075 Channel Inside Width (Zr-2) - in.

6.54 Channel Inside Width (SS) - in.

6.57 Moderator to Fuel Area Ratio

2. 7 (c) Centrol Rods Type of Drive Hydraulic locking piston Number of Control Rods 32 Poison Material in Rods B C Powder in 4

S. S. tub e s Pitch (square array) - in.

10.466 Active Length - in.

68 Shape Cruciform Width - in.

11 -1 / 2 Blade Thickness - in.

5/16 Sheath Thickness - in.

1/16 Number B C Filled, S. S. Tubec Per Blade 11 6 4

B C Tube Diameter - in.

0.175 4

B C Tube Wall Thickness - in.

0.020 4

Section 8 Page 4 (d) Heat Transfer & Fluid Flow Average Heat Flux at Rated - Btu /hr-ft2 110 x 103 3

447 x 10 Maximum Heat Flux at Overpower -

g 2

4720 Total Heat Transfer Area - it Avg. Fuel Center Temperature at Rated Power

• F 1500 Maximum Fuel Center Temperature at Rated Powe r

• F 3600 Maximum Fuel Center Temperature at Overpower

• F 4400 I

Peaking Factors Axial 1.55 Radial (incl. possible variations) 1.25 Local (incl. water slab, intercontrol rod, & mfg. effects) 1.65' Overpower (incl. stea.n flow and tran-sient effects) 1.27 l

Maximum Peaking Factor at Overpower 4.1 l

l Active Core Power Density - kw/1 45 1

Minimum Burnout Ratio (full power, steady state)

2. 3 Minimum Burnout Ratio (overpower)

1. 8 Core Exit Steam Flow Rate - Ibs/hr

5. 6 x 105 Average Steam Quality, Core Exit, %

49 Steam Quality Entering Drum, %

5. 2 Coolant Saturation Temperature

'F 550.5 i

2 Coolant Flow Area - in / channel 26 Av. Liquid Velocity at Channel Inlet-ft/sec

5. 6 6

Total Reci rculating Flow Rate - lbs/hr

11. 6 x 10 Hydraulic Diameter - in.

O.52 i

(

Section 8 Page 5

0. 2. 2. 2 In calculations of burnout margin, hot channel factors are used to obtain the maximum channel enthalpy rise and exit

,~

quality. The hot channel factor is thus a ratio of the maxi-mum enthalpy rise in the hot channel at overpower to the core average enthalpy rise at rated power. The hot channel factor represents regional or channel power peaking rather than rod locals which are covered by the power peaking factor s.

As such, the total hot channel factor of 2. 3 was derived from the power peaking values, reduced to con-sider the regional or average heating effects as well as reduced hot channel flow.

8.2.3.

Core Flow and Pressure Drop 8.2.3.1 The total coolant flow through the reactor is 11. 6 x 106

~

lb/hr. Fuel bundle coolant flows from the inlet diffuser, through openings of the guide tube support, up through the guide tube and orifice, and through the fuel channel.

8. 2. 3. 2 Core bypass leakage flow is provided to remove heat generated between f:uel channels, in the water reflector, and between thermal shield and reactor vessel wall.

t 8.2.3.3 Flow orifices are used to increase the single phase pres-sure drop of the core. The effect of this single phase pres-sure drop is to make individual channel flow less dependent on power, thus improving burnout ratio in the core, and improving parallel channel stability. Selective zone flow orifi'cing could also be used to control the flow dis-tribution to follow the core power distribution with the initial or subsequent cores.

8. 2. 3. 4 The total pressure drop through the u re in 157 Mwt operation at 1050 psia is approximately 9 psi, which approximately 3 psi is across the flow orifices and the remainder across the fuel bundle.

I 8.2.4 Material Temperatures 8.2.4.1 The maximum local UO2 center temperature is about 4400*F when the reactor is at maximum overpower heat flux. This temperature is conservatively based upon back-calculations for temperatures which were deduced from similar fuel operated in a reactor until onset of fuel center void formation associated with melting. The value for this temperature has been conservatively chosen to be 5000*F.

i...

i Secti:n 8 Page 6 Rev 1 (3/12/62) 8.2.4.2 The maximum fuel jacket temperature in the hottest fuel rod at the peak cverpower condition is less than 700 F.

8.2.4.3 Heat generated in the boron carbide and stainless steel structure of the control rods is transferred to the reactor coolant which flows through the holes in the sheath of the control rods past the individual poison tubes, and also by conduction through the sheaths. The resultant temperature of the 7 uon carbide in the hottest poison tube at overpower,

averaged over the length of the tube, is 700 F.

8.2.5 Thermal Burnout 8.2.5.1 Thermal turnout can be described as rapid fuel jacket melt-down with poor surface heat transfer conditions. Burnout can thus occur when so much steam is generated at some point on a heat transfer surface that it completely blankets the surface, thus increasing resistance to the passage of heat.

l 8.2.5.2 The burnout margin at a point is the ratio of heat flux suf-l ficient to cause burnout at the point to the local generated heat I

flux. The burnout correlations used in the design of the Big l

Rock Nuclear Plant are based on extensive out-of-pile tests carried out in an electrically heated loop. The test data and basis for correlation are given in GEAP 3892, " Burnout Limit Curves for Boiling Water Reactors," E. Janssen and S. Levy.

The burnout correlation for 1000 psia is:

(q/A)BO = 0.705 + 0.237 G

( x3 x

(q/A)BO = 1.634 - 0.270 G - 4.710 x xy ( x (x2 (q/A)BO = 0.605 - 0.164 G - 0.653 x x2 <

• l

[

xi = 0.197 - 0.108 G l

x, = 0. 254 - 0. 026 G 2 x 10-6 (q/A)BO = burnout heat flux, Btu /hr-ft 2 x 10-6 G

= mass flow rate,1b/hr-ft x

= quality, weight fraction steam i

This correlation is based on the data obtained over a range of flow path hydraulic diameters and is valid for hydraulic diameters i

less than 0.6 inch.

At pressures above or below 1000 psia, the burnout heat flux is

{

obtained by adding the following pressure dependent term to the heat fluxes given by the above correlation:

2 440 (1000 - P)

Btu /hr-f t where P is the system pressure in psia.

8.2.5.2.1 Data is given in GEAP 3892 to substantiate this hydraulic diameter effect and pressure effect on burnout heat flux.

8.2.5.3 The present minimum burnout margin at hot channel overpower I

conditions has been calculated to be about 1.8 at 1050 psia, 157 Mwt.

Secti:n 8 Page 6A (3/12/62) 8.2.5.4 During power operation the minimum burnout ratio will be determined as follows:

1 1.

Apply an overpower and transient allowance to the heat fluxes determined from wire data or in-core chamber readings that have been calibrated in terms of detailed power distribution data.

2.

Determine total recirculation flow, reactor inlet enthalpy from feedwater temperature and power level from plant instrumentation. Adjust for overpower condition.

3.

Determine flow to the potentially limiting fuel bundles and quality at potentially limiting points. Finally, calculate o

burnout heat flux and burnout ratio at potentially limiting points using precalculated curves and nomograms.

4.

Note the location at which the burnout ratio is a minimum and monitor continuously by means of in-core chambers or by observation of control rod motion and total core power change. Maintenance of the minimum burnout ratios in excess of 1. 5 will satisfy the burnout margin limit.

8.2.5.5 During the initial plant startup, basic experimental power distribution measurements will serve to verify calculated power distributions. In-core ion chambers will be calibrated by comparison with flux wires and calculated and/or measured power distributions.

4 2

s t

'k

Section 8 Page 7 8.'2. 6 '

Hydraulic Stability

8. 2. 6.1 In operation, the onset of hydraulic and nticlear instability can be sensed.and predicted (by observation of the flow and nuclear instrumentation) well before divergent oscillations occur. Methods for sensing and predicting instability experims -tally have been demonstrated on the boiling water reactors, BORAX, SPERT,. EBWR, VBWR, AND Dresden.

i 8' 2' 6. 2 Based on this reactor experience, and heat transfer loop j

i operation, methods of hydrodynamic analysis have been f

developed for reactor application. This hydrodynamic analysis is based on the physical concepts of momentum inte rchange, conservation of energy, and continuity of mass rather than steam void versus quality correlations or steam slip assumptions.

8.2.6.3 The stability criteria' is based on a freque tcy response-phase shift analysis tnat must yield a positive phase margin for both the hydraulic loop response and the coupled nuclear-hydraulic loop response.

This analys i s i s based on the differential equations of reactor kinetics, fuel heat transfer, single and two-phase flow, and thermodynamics. Transfer functions are developed from these differential equations in order to perform this frequency--phase shift anaylsis.

Application of this method of analysis to Dresden and VBWR have demonstrated that when tlIis criterion is satisfied, no tendency toward instability occurs.

8. 2. 7 Reactor Design Data The inter-related thermal and hydraulic characteristics of the core are summarized in Table 8.1, (physical arrangement and dimensions are discussed in Section 4. 2). The reference core describeo meets the design objectives within the criteria.

8. 3 NUCLEAR CHARACTERISTICS OF THE REACTOR

8. 3.1 Design Criteria The nuclear characteristics of the reactor will be such that throughout the entire operating history, the following design criteria are met:

Section 8 Page 8

8. 3.1.1 The negative Doppler reactivity effect and moderator tempera ture reactivity effect willinsert enough negative reactivity to prevent destructive damage to the reactor vessel during a credible powe r excursion.

8. 3.1. 2 The average reactivity coefficient of voids will be negative for all conditions.

8. 3.1. 3

• ibe reactor control system is designed with sufficient excess control capacity such that the safety of the reactor system will not be jeopardi zed in the event the most valuable control rod were to be inoperative in the fully withdrawn condition.

8. 3.1. 4 The power distribution will be sufficiently flat to permit operation.

of the ccre at rated power and overpower without exceeding es+.ablished temperature and heat flux limitations.

8.3.2 Basic Considerations

8. 3. 2.1 The ensuing discussion relates to the nuclear characte istics of the initial core of the Big Rock Point reactor.

8.3.2.2 The Big Rock Point high power density reactor is a light water moderated low enrichment, uranium oxide fuel reactor. The use of light water produces a neutron spectrum such that the majority of fissions-from which power is derived are produced by thermal neutrons. Because of the presence of U-238, approximately six percent of the total power is pro'duced by the direct fissioning of this material by fast neutrons. This increases the f rar. tion of delayed neutrons in the core. In a bundle irradiated to an average exposure of 10,000 MWD /T, approximately 31% of the power is produced by Pu239 fissions.

8.3.2.3 The lattice is heterogeneous, being made up of uranium oxide rods as described in Section 4. 2 and detailed in Table 8.1.

All fuel rods contain an initial enr'ichment of 3. 2% U2,-

weight fraction of total uranium. The corner rods are of reduced size to reduce the local power peaking factor.

Fifty-six of these bundles are used in the 157 MWt core.

8.3.2.4 The core is controlled by means of 32 cruciform control rods arranged as indicated in Drawing 197E853. The neutron absorbing material in these rods is boron carbide as discussed in Section 4. 3.

The control system has sufficient shutdown capacity to maintain the core suberitical with any single con-trol rod fully withdrawn. Stainless steel channels are employed

Section 8 Page 9 in the initial loading to supplement the control rod strength and satisfy the shutdown ma rgin. The reactivity status of the equilibrium (design exposure) core will make use of the steel channels unnecessay.

l

-8.3.2.5 The water which produces the moderation and also serves

[

as the reactor coolant is normally at saturation temper-l ature of 550*F at 1050 psia although tests may be run

[

using pressures up to 1500 psia in connection with the l

program described in Section 10 of this report.

The fuel rods, which are poor thermal conductors, are at considerably higher temperatures than the surrounding moderator and will support a temperature within them-selves of several thousand degrees F.

For this reason, the Doppler effect, which is dependent directly on fuel temperatures, is of considerable importance to plant dynamics and safety, principally because this is a reac-tivity effect which occurs as rapidly as fuel temperature rises while heat conduction to water (and the consequent formation of steam voids) must await the transfer of heat through the fuel material.

8.3.2.6 The neutron lifetime is calculated to be 4 x 10-5 seconds and determined primarily by the relatively high thermal neutron absorption of the lattice. This lifetime is characteristic of a light water moderated reactor.

8.3.3 Features Specific to This Reactor 8.3.3.1 The active core is approximately five feet in diameter and six feet high. From the standpoint of the nuclear charac-I teristics, this core is of intermediate size in that, while the power distribution is not controlled by cor neutron leakage (as would be the case for a small reac-or), the power distribution is also not overly flexible, as would be the case for a very large, light water core. Never-theless, the dimensions of the core are considerably in excess of the neutron migration length in the core material, and hence, make necessary the consideration of the effects of control rod and steam void patterns on the power dis tribution. Furthermore, the ability of the flux distribution to vary with operating states makes calibration of the control rods a relatively meaningless procedure, since the calibration is affected very strongly by the position of steam voids, control e

materials, and poisons in the core.

8.3.3.2 Because the Big Rock Point reactor is a boiling water reactor, it exhibits characteristics of non-uniformity and non-linearity.

Specifically, the power distribution is a function of power because of the feedback through steam void formation; and 4

I Section 8 Page 10 the chs.cteristics of the core are space dependent because of t.e non-uniform distribution of steam voids. This necessi-tates care in the discussion of th,e nuclear characteristics and makes any table of data which is derived from average core properties only indicative of the general situation.

8.3.3.3 In the rigorous sense, it is necessary to restrict descriptions I

of the Big Rock Point high power density reactor to the descrip-tion of the individual operating states. It is not possible to calculate off-critical configurations and interpret the excess reactivity as a quantity to be compensated for by the insertion of control poisons. Consequently, discussion of reactivity changes must always be based on real operating states in an exactly critica.1 condition. It is to be recognized, in addition, that the Big Rock Point reactor is a forced circulation reactor.

To some extent, therefore, the core power distribution may be varied by insertion of various patterns of orifices in the bottom of the coolant inlet to the reactor core. The reactivity in voids for purposes of arriving at an end of life reactivity balance is 0.019 oke ff. Earlier in life, higher neutron. leakage can cause the reactivity in voids to be as high as 0. 03 okeff. The reactivity it.

voids depends very stro,ngly on the void and power distribution.

For a reactor as large as this one, the reactivity in voids must always be expressed in such a fashion that it applies to a specific operating condition of the reactor. For a small reactor with constant statistical weights, these distinctions are not necessary.

8.3.3.4 The power distribution displays a relatively. low peak-to-average value which is due partly to the self-flattening effect of the void distribution and partly to the judicious use of the control rod system. Continuous programming of the control rod pattern will be used in or2er to maintain the desired distri-bution of power in the core. Control rod programs will be determined partially by meais of calculations performed in advance of operation, and partially from readings obtained from the in-core flux instrumentation.

8. 3. 3. 5 Calculations performed to date indicate that the Big Rock Point reactor will operate dynamically, to a godd approximatica, as a i

one mode reactor, and consequently, a very large portion of the operating experience in boiling water reactors to date is applicable to this plant..Because of the long time constant of the fuel and the high operating pressure, no difficulties are expected due' to dynamic instability. The derivative of the reactivity with respect to void content (dk/dv) will be kept to a value well within that

Section 8 Page 11 which can be tolerated for stable plant performee. Some of the important nuclear parameters of the Big Rock Point reactor core are summarized in the following table.

TABLE 8. 2 NUCLEAR PARAMETERS Zircaloy-2 Channels 0% Void 0% Void 20% Void 68'F 550'F 550'F 2

{

Age, cm 33.7 52.5 64 9 2

2 L

Thermal diffusion area, cm

2. 6

5. 3 6.1 p

Total resonance escape probability O. 81 0.76 0.73 p28 Probability of escaping 238 0.88 0.84

0. 81 resonance capture in U k.

Infinite lattice multiplication factor:

Uncontrolled 1.248 1.265 1.235 Controlled 1.006 0 921 0.871 Stainless Steel Channels 2

T Age, cm 32.2 49 9 61.5 L

Thermal diffusion area, cm2

2. 2

4. 5

5. 2 2

(

p Total resonance escape probability O. 81 0.76 0.73 p28 Probability of escaping 238 resonance capture in U 0.88 0.84

0. 81 k.

Infinite lattice multiplication factor:

Uncontrolled 1.131 1.153 1.120

(

Cont rolled O.970 0.889 0.836 8.3.4 Reactivity Coefficients In order that steam formation be a stable process in a boiling reactor, the lattice must be established such that the core is unde rmode rated. However, if the reactor is greatly under-i moderated, the large negative void coefficient will produce a f

Section 8 Page 12 feedback effect which may contribute to system instabilitfes, The water-to-fuel ratio in the Big Rock Point reactor has been selected so as to give satisfactory reactivity coefficients.

8. 3, 4.1 Temperature Coefficient From the values listed in Table 8. 2, it is apparent that the r.eactivity change with temperature is dependent upon the positic.ns of the control rods. The temperature coefficient is thus expected to vary as the control rods are repositioned during the operation of the reactor.

At temperatures of the order of 20 *C, the effect of tempera-ture increase is usually an increase in k of the lattice.

This effect is opposed by an increase in the leakage due to the increase in the migration area with temperature. The net effect is a keff variation which may be initially. slightly positive.

As temperature is increased, k increases less rapidly and the leakage effect produces a net negative coefficient.

The leakage portion of the coefficien? is dependent on the buckling in the region of interest. At the start of life, when control rods induce internal leakage, the effect is to make the coefficient more negative than at the end of life when the only leakage is through external boundaries.

In Tr.ble 8. 3 the start of life value is comp.osed of a positive component in the steel channel region, and a negative component in the zircaloy channel region. In the end of life case, the buckling change produced by withdrawal of control rods has caused the coefficient in a zircaloy channel core to change from an initial negative value of 41:. x 10-4 one ' keff per ' F to a pcsitive =e indicated it. Se tab! e.

Table 8. 3 contains the basic coefficients for cores containi g either 56 zircaloy channels or 56 stainleos steel channels.

)f these, only the end of life zircaloy case vrould be approected in practice. The difference between the initial core and end of cycle data is prod'. ced e *.irel b. the diff e re nce in r:cution lea ka ge f r om the j n t-t r

• ce ' _ca d r gc,

In the initial core, a smaller region can be critical.

m

Section 8 Page 13 TABLE 8. 3 TEMPERATURE COEFFICIENT (Ak eff /kc f per *F; r

t 68'F 3 20 *F 550*F Steel channels:

Clean initial core

+ 0. 4 x 10 - 4

-0. 7 x 10 -4

-1. 3 x 10 - 4 End of cycle

+0. 6 x 10 - 4

+ 0.1 x 10 - 4

-0. 7 x 10 -4 Zircaloy channels:

Clean initial core

-0.1 x 10-4

-1. 4 x 10 - 4

- 2. 5 x 10 -4

-4 End of cycle

+ 0. 3 x 10 - 4

0. 0

-0. 6 x 10 8.3.4.2 Void Coefficient In any rise in power level resulting in formation of voids, the voids produced are confined to the regions of the core within the fuel channels. The calculated coefficients are based on this assumption.

The operating reactor contains regions with control rods as well as unrodded regions. The void coefficient may thus be expected to vary from point to point in the core as well as varying due to control rod movement. ' The calculated void coefficient is based on the unrodded region since it is expected that voids would first be formed in these regions. The leakage from such a region was assumed to be sufficient to make the effective multipli-stion factor unity in the region. As in the case of the temperature coefficient, it is anticipated that the i

void coefficient will become 1 ess negative as the fuel is depleted due to the removal of control rods from the core.

The coefficients inthe clean core and at the end of an operat-ing cycle are shown in Tabl e 8. 4.

The difference between f

initial core ar.d er d cf c ycle data ie p-od:ced ertirely by the diff erence.-

in neutron leakage from tne just-critical loadi gs.

h the initial core, a smaller region can be critical.

(

l

Section 8 Page 14 TABLE 8. 4 VOID COEFFICIENT (Akeff /keff pcr Unit _ Void) 68'F 550*F 0% Void 20 to 50% Void Steel Channels:

Clean initial core

-0.18

-0.17 End of cycle

-0.09

-0.14 Zircaloy Channels:

Clean initial

-0,24

-0.23 End of cycle

-0,05

-0.12 8.3.4.3 Doppler Coefficient 238 The temperature broadening of the U resonance peaks pro-duces an important effect in low enrichment reactors called the Doppler effect. Because this effec 6 agends upon the thermal motion of the fuel atoms, it pro < duces an instantaneous reduction in reactivity whenever the fuel temperature rises. The Doppler coefficient is essentially independent of irradiation and thus remains constant over the operating period.

The rate of change of reactivity with fuel temperature increases as the fraction of neutrons captured in the resonances increases.

l The Doppler coefficient thus tends to be higher at operating con-ditions due to the re' duction in water. This effect is counteracted i

by the fact that the Doppler coefficient 'decreas'es as the fuel l

temperature increases. The net result is that the Doppler coefficient remains relatively constant from cold to operating I

conditions as is seen in Table 8. 5.

(

TABLE 8. 5

• DOPPLER COEFFICIENT ( Ak f.7) x

(

Fuel Temperature Moderator Akm / h per

• F

(

68'F 68 *F, 0 Voids

-1. 47 x 10-5 13 7.3 *F 550 *F, O Voids

-1. 0 3 x 10 -5 13 23 *F 550 *F, 20% Voids

-1.15 x 10-5

(

• Coefficients apply to k, assuming a given constant fuel tempera-ture over the lattice.

(

Section 8 Page 15 8.3.4.4 Pressure Coefficient Any increase in reactor pressure will decrease the void content of the core, thus a negative void coefficient implies a positfve pressure coefficient. These two coefficients are related by the rate at which the void fraction changes with pressure. It is not possible to state a single rate of change of k with pressure unless the sequence of events causing the change in pressure is specified.

8.3.5 Reactivity and Control Requirements

8. 3. 5.1 The Big Rock Point reactor fuel is enriched to meet the reactivity requirements of 10,000 MWD /T discharge exposure averaged over the first core load. In addition, the core must have sufficient reactivity to allow for temperature and void effects and for the formation of xenon and samarium. The reactivity required for these effects is summarized in the following Table 8. 6.

l TABLE 8. 6 REACTIVITY REQUIREMENTS hk gg e

Temperature O.025

0. 019 Voids Xenon and Samarium

0. 0 31 Fuel depiction and maneuvering 0.127 Total ok required O.202 i

8.3.5.2 From the above table, it is seen that the initial reactivity require-ment is keff = 1. 202.

8.3.5.3 The control system is designed so that in the cold clean condition the reactor will remain in the suberitical condition even if the

(

strongest control rod wem completely removed from the core.

The satisfaction of this criterion will be demonstrated experi-(.

mentally. A margin of at least 1% Ak is included in the design to insure meeting this requirement, however, the 1% margin will not be experimentally demonstrated. From the following Table 8. 7, it is apparent that the control and reactivity require-ments are satisfied in the Big Rock Point reactor.

l Section 8 PageIb f

TABLE 8. 7 INITIAL REACTIVITY (COLD) keff All rods out, no steel channels 1.202 28 steel channels, 32 rods in 0.948 28 steel channels, 31 rods in, central rod out 0.965 28 steel channels, 31 rods in, outer rod out 0 963

8. 3. 5. 4 The reactivity worth of any one control rod depends on the flux distribution and thus is influenced by the positions of the other control rods. Specifically, the worth of a control rod may be increased by withdrawing only the adjacent rods since this condition produces an abnormally high flux peak in the vicinity of the central rod. This condition has been assumed in cal-culating the maximum rod worths tabulated below in Table 8. 8.

However, it is expected that the reactor will normally be operated so as to avoid this situation ar.d in general, the worth of any control rod will be considerably less than the values shown.

TABLE 8. 8 MOST PROBABLE MAXIMUM CONTROL ROD WORTH Ak/k Cold O.039 Hot (550 *F) no void O.042

(

Hot (550 *F) 20% void 0.030 i

8.3.6 Irradiation Dependence of Core Characteristics

8. 3. 6.1 During operation, the composition of the core undergoes a i

continuous change due to the depletion of U235 and the pro-1 duction of fission products and transuranium elements. At i

the same tiine, the control rods are moved so as to maintain ^

c riticality. At the enrichment level and water-to-fuel ratio selected for the Big Rock Point reactor, the reactivity of the f

core decreases with irradiatien necessitating a continual 1

~.-

l

Secti n 8 Pago 17 withdrawal of the control rods. The power distribution and consequently the void distribution are thus expected to change during the burnup cycle.

8.3.6.2 During fuel irradiation, the isotopic composition in a fuel rod changes in such a manner that the power distribution through the fuel rod is more strongly concentrated at the outside edge, and is thereby improved. The surface neutron resonance absorption in U-238 produces a concentration of the plutonium isotopes on the outside edge of the fuel pellet which increases the power generation in this area.

8.3.6.3

\\Vithin the fuel bundles, the rods near the corner of the bundle are irradiated to higher exposure than the other rods because j

of the higher neutron flux in this area. The refo re, the local j

powe r peaking in the fuel bundles may improve with lifetime due to the more rapid burnout of U-235 in the locations near the water gaps. The overall power distribution across the core also improves with exposure because of the more rapid burnup of fuel in the central, higher neutron flux regions.

This improvement is partially compensated for by removal of control rods which are used to flatten the power distribution.

The thermal neutron spectrum becomes more strongly dis -

torted from a Maxwellian due to the buildup of the plutonium isotopes with their low energy cross section resonances.

8.3.6.4 The reactivity coefficients become less negative with exposure primarily due to the removal of control rods from the core.

'1he increase in control rod worth with an increase in void or temperature is an important effect producing a negative con-tri bution to the reactivity coefficients. The presence of plutonium also affects the temperature and void coefficients of the reactor. However, because of the enrichment level, the buildup of plutonium in t)he fuel has a small effect on the coefficients (this contrasts with the positive effect to be expected with the growth of plutonium in a reactor of lower or natural enrichment). Consequently, the safety of the reactor is in no way compromised by the pre sence of the expected quantities of plutonium.

8.3.6.5 In the initial operating condition, the effective delayed neutron fraction including the contribution of U-238 fissions is 0. 0070.

Because of the smaller delayed neutron fraction which is char-acteristic of plutonium, the delayed neutron fraction in the reactor will change with irradiation and at an average fuel irradiation of 10,000 MWD /T is expected to be about 0. 0053.

8.3.7 Fuel Cycle and Control _

8.3.7.1 Refueling The Big Rock Point reactor is designed to be refueled on a progressive partial batch schedule. It is anticipated that an inward progression refueling schedule will be used. In l

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S2cticn 8 Page 18 such a case, the central section of the core will be removed and the remainder of the core will be moved progressively inwa rd. A fresh batch of fuel corresponding to approximately 30 percent of the core will be inserted near the periphery of the core. This form of fuel progression is favorable to the reactor power distribution.

The inward progression of successive core refueling batches will be continued and eventually an equilibrium situation will be achieved. The present enrichment levels and refueling schedules are intended to achieve an irradiation of 10,000 MWD /T, averaged over the first core, which will correspond 1

to equilibrium irradiation somewhat longer once equilibrium refueling has been established. The reactivity of the succeed-ing batches of fuel which are added to the core may be increased to something in excess of the reactivity of the initial core to correspodd to the additional. control _ capacity which is available as a result of partial burnout of the reactor core. In the event that higher reactivity fuel _ used in succeeding batches which are added to the core, the fuel will be added in such a pattern as to avoid compromise of the design stuck control rod criteria.

Introduction of developmental fuel bundles of the research and development program will not necessarily follow the loading pattern given above, but the safety criteria will be obse rved.

8.3.7.2 Off-Standard Control Conditions The control system of the Big Rock Point reactor has been designed with various off-standard situations in mind. Spe-cifically, the effects of stuck control rod mechanisms are conside red.

In a light water reactor coptaining appreciable excess reactivity, the cold clean core can contain a large number of critical ma s s e s. It is, consequently, necessary to provide a reactivity controlling device for each portion of the core. This dictates a large number of strong control rods, but also makes it necessary to consider local criticality caused by inoperative control mechanisms. The Big Rock Point reactor has sufficient i

excess shutdown capacity that it can be satisfactorily brought below critical in the cold condition at the beginning of any fuel irradiation cycle with any one control rod stuck fully out of the core. In the hot or operating conditions, of course, the control system is stronger and the reactivity of the core is somewhat less. Consequently, the stuck rod situation

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is lesa severe and three oi four adjoining control rods may be lost without endangering this shutdowr. capability. If these control rods are stuck from the core while hot, shu t-1

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down can be achieved with the backup safety system (liquid poison) to permit holding the reactor suberitical indefinitely so that repairs can be made. Consequently, stuck control rods usually do not constitute a safety hazard during power operation in the hot condition.

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Se ction. 8 Page 19 The local control characteristics of the lattice also dictate a startup accident configuration, in which several control rods are withdraw.nt surrounding a central inserted control rod. This configuration increases the statistical weight in

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the remaining inserted control rod and makes its, withdrawal

'l produce a large change in reactivity. Consideration of a pessimistic configuration combined with the maximum possible speed of rod withdrawal yields a maximum reactivity in-serti'on rate which is used in determining the startup accident.

8.3.7.3 Special Operating Considerations

' The large number of critical masses in the reactor core 4

dictates special procedures at some conditions of plant i

operation. For exanple, the loading of the core or the insertion of new fuel requires consideration of the effect of each fuel addition on the core reactivity, both v?ith all control rods inserted and with the most important control rod withdrawn.. Refueling procedures are established to ensure that no fuel loading operation can jeopardize the ability of" the control system to maintain the reactor suberitical.

Normal plant startup will involve increases in local multipli-cation by control rod withdrawals distributed over the core.

In this manner the power distribution can be more easily adjusted to that desired at the normal operating pcwer level, and the. worth of indiv'idual control rods can be minimized.

1 Operational procedures are designed on the basis of these principles.

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