ML20030A445
| ML20030A445 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/18/1960 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090520 | |
| Download: ML20030A445 (11) | |
Text
S E C T IO N 'IV g
f NUCLEAR CHARACTERISTICS OF THE REACTOR A.
PURPOSE AND SCOPE This section describes the characteristics of the nuclear process taking place in the reactor. This information supplements the general description of the process presented in Section III.
B.
GENERAL DESCRIPTION 1.
Basic Considerations The Big Rock high power, density reactor plant is a single cycle, forced circulation boiling water reactor. The initial core of the Big Rock reactor will be designed for a power density of 45 kw per liter at an operati'ng pressure of 1050 psia. An associated researc,h and development program is proposed to extend the reactors operating characteristics to larger power densities and greater pressures. Consistent with these objectives, the nuclear characteristics of the re-r.ctor will be such that throughout the entire operating history the following design objectives are met:
1.1 The negative Doppler, reactivity effect and prompt heating of the coolant will exert a great enough negative reactivi6y to prevent destructive damage to the reactor during a credible prcmpt excursion from the cold condition.
- 1. 2 The average reactivity coefficient of voids, which results from boiling of water in the coolant channels, will be neg-ative for the cold, critical reactor.
- 1. 3 The nuclear characteristics of the reactor core will be such that the core, the coolant flow loops, and the steam supply syrstem will be stable to nuclear and coupled nuclear-hyd:aulic perturbations except during restrictive modes of reactor operation.
l4 The reactor control system will be designed with suf-ficient excess control capacity auch that the safety of l
the reae. tor system will not be jeopardized 1n the event the most valuable control rod were to be inoperative.
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BIO i 09 05.2 0
IV-2 1.5 An entirely separate backup reactor control device will t
be designed to shut down the reactor during all credible contingencies of operation.
- l. 6 Ths nuclear characteristics of the reactor will be such that the reactor operator by proper manipulation of the control rods can maintain power distributions through-out reactor life which will prevent operation of regions of the core at power densities in excess of burnout limitations.
The ensuing discussion relates to the nuclear character-istics of the initial core of the Big Rock reactor.
2 The Big Rock 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 contributes both to the power output of the plant and to the fraction of delayed neutrons in the core.
I The lattice is heterogeneous, being made up of uraniurr. ox-ide rods approximately 0. 4 inch in diameter and spaced so as to produce an array with an over-all water-to-fuel volume ratio of 3. 2 to 1.
The lattice, in addition, is clumped, in that the uranium oxide rods are arranged in fuel element bundles of 64 rods each, and the core is composed of 120 of these bundles with additional strtictural material and water in the spaces between bundles.
The water which produces the moderation and also serves as the reactor coolant is normally at a saturation temperature of 550 *F at 1050 psia. The fuel rods, which are poor thermal conductors, are at considerably higher temperatures than the surrounding moderator and will support a temperature dis-tribution within themselves 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 reactivity effect which occurs without significant time lag, while heat conduction of water (and the consequent formation of steam voids) must await the transfer of heat through the fuel material.
IV-3 The neutron lifetime is approximately 5 x 10-5 seconds and
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determined primarily by the relatively high thermal neutron absorption of the lattice. This lifetime is characteristic of a light water moderated reactor.
2.
Features Specific to a Forced Circula. ion Boiling Water Reactor The active core is approximately five feet in diameter and six feet high. From the standpoint of the nuclear characterir'.ics, this core is of intermediate size in that, while the power dis-tribution is not determined entirely by core leakage (as would be the case for a small reactor), the power distribution is also not overly flexible, as would be the case for a very 1crge, light water core. Neve rthele s s, the dimensions of the core are con-siderably 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 distribution. 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 materials, and poisons in the core.
l i
Because the Big Rock reactor is a boiling reactor, it exhibits characteristics of non-uniformity and non-linearity. Specif-ically, the power distribution is a function of power because of the feedback through steam void formati~on; and the char-acteristics of the core are space dependent because of the I
non-uniform distribution of steam voids. This necessitates care in the discussion of the nucigar characteristics and I
makes any table of data which is derived from average core properties only indicative of the general situation.
In the rigorous sense, it is necessary to restrict descriptions of the Big Rock high power density reactor to the description of the individual operating states. It is not possible to calcu-late off-critical configurations and interpret the excess re-activity 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 critical condition. It is to be recognized, in addition, that the Big Rock 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 purpose of characterizing the reactors operating condition
IV-4 is approximately. 03 Ak.
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 con-dition of the reactor. For a small reactor with constant statistical weights these distinctions are not necessary.
The power distribution displays a relatively low peak to average which is caused largely by the self-flattening ef-fects of the void distribution which result from the com-pensating effect of the steam voids at the top of the reactor core and the control rods in the bottonn of the reactor core.
It is expected that power distributions adequate to meet full power performance will be achieved in the Big Rock reactor.
Neve rtheles s, continuous programming of the control rod pattern is expected to be used in order to optimize the core power distribution. The rod programs will be determined partially in advance of operation and partially from the use of the in-core flux monitor system described in Section III.
In this way the operator will be able to trim the power dis-tribution for most efficient operation, depending on the read-ings of the core instruments and using as a guide the calcu-lations previously supplied to him. The Big Rock reactor is expected to operate dynamically, to a good approximation as a one node reactor, and consequently, a very large portion of the operating experience in boiling reactors to date is ex-pected to apply to this plant. Because of the long time con-stant of the fuel and the high operating pressure, no diffi-culties are expected due to dynamic instability. The derivative of the reactivity with respect to steam formation (dk/dv) is expected to be kept well within that which can be tolerated for stable plant performance.
C.
EFFECT OF VARIABLES ON REACTIVITY 1.
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 r-mode rated. If it is too much under-moderated, the value of dk/dv will be too large and provide a large feedback loop gain which will contribute to system instability. If the lattice is lese under-moderated, the moderator temperature coefficients in the cold shutdown condition may be positive.
In the Big Rock Plant the moderator temperature coefficient in the cold shutdown condition is approximately
+1. 5x10-5
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(Ak/k)/ *F in the absence of control rods. The presence of control rods will tend to make this coefficient negative when we
IV-5 the reactor is in the cold condition. The coefficient will be-f come increasingly negative as the reactor is brought up to the operating conditions, and is equal to -10. 2x10-5 (Ak/k)/ *F at operating temperature with no voids in the core.
The void coefficient within the steam generating area of the lattice is at all times negative and along with the moderator temperature coefficient, will become increasingly negative as reactor temperature is increased. At full power operating conditions, the average void coefficient will be approximately
-0. 3 (Ak/k)/ void fraction, corresponding to a reactor power coefficient of reactivity of -6x10-4 (Ak/k)/MWe.
The Doppler coefficient is a direct function of lattice design and of fuel temperature in the Big Rock reactor; this coef-licie nt is approximately -0. 4x10-5 (Ak/k)/ 'F, in the cold shutdown condition and, as such, provides a major part of the inherent safety against startup accidents. The Doppler effect is, to some degree, dependent upon the fuel temper-ature and the shape of the temperature distribution through a fuel element. The over-all Doppler coefficient becomes more negative as the operating temperature is increased.
Estimated fuel temperature coefficients of reactivity due to the Doppler phenomena at the various steady-state operating conditions are listed in Table IV-1.
At rated operating conditions the fuel temperature varies from somewhere under 700'F at the surface to as high as about 3b40'F at the pellet center. The Doppler effect can be calcu-lated in terms of an average intermediate fuel iimperature under these conditions. During the startup accident or any rapid change in power, however, the heat conduction mechanism in the fuel pellet will be relatively flat. Under these conditions the Doppler effect can remove approximately 2 percent in re-activity before any fuel melting is experienced.
Cons equently, the Doppler effect exerts a stronger influence during a start-up accident than would be expected from examination of its normal operating performance.
Because of the necessity to provide suitable regulation in the steam formation process, the introduction of steam voids must reduce reactivity, and consequently, the lattice is designed for a negative void coefficient, as mentioned earlier. This, how-ever, implies that the pressure coefficient of the reactor must be positive. Extensive operating experience with other boiling reactors shows that this is a satisfactory condition, primarily
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IV-6 because of the inertia which the large saturated water system presents to pressure changes, and the consequent ease which pressure variations may be controlled, either manually or by the use of external pressure control equipment.
A summary of the reactivity coefficients of the Big Rock re-actor in its initial operating condition is presented in Table IV-1.
TABLE IV - 1 REACTIVITY COEFFICIENTS Cold Hot
- Ope rating Moderator Temperature Coef. (Ak/k)/ *F
+ 1. 5x 10- 5**- 10. 2x10-5 Void Coefficient, (Ak/k)/ void fraction
.02
- O.15
-0.3 Reactor Power Coefficient, (Ak/k)/Mwe
-6x10-4 Fuel Temperature (Doppler) Coef (ak/k)/T -0.4x10-5
- 0.6x10-5
-0.8x10-5
- No Voids
- With no control rods 2.
Irradiation Dependence of Core Characteristics As fuel irradiation progresses, a number of the core charac-teristics undergo a significant change. The reactivity of the fuel changes as the initial fuel is burned and as plutonium is produced. Because of the initial enrichment level involved and because of the plant conversion ratio of approximately
. 45, the reactivity in the Big Rock reactor always decreases with exposure.
In the initial operating condition, the effective delayed neutron fraction, including the contribution of U-238 fissions, is 0. 007, which, with approximately 3 percent reactivity represented by the steam void distribution results in a reactivity in dollars of $4.15 in steam voids. Because of the smaller delayed neu-tron fraction which is characteristic of plutonium, the delayed neutron fraction in the reactor will change with irradiation.
When the large delayed neutron yield due to fast fission of U-238 is also included, the effective delayed neutron fraction at a fuel irradiation of 10,000 MWD / ton is expected to be 0. 006 i
The presence of plutonium is also expected to affect the temper-ature and void coefficients of the reactor. However, because of
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IV-7 the enrichment level of the Big Rock reactor, the buildup of plutonium in the fuel will have a negligible or slightly nega-tive effect on these coefficients (this contrasts with the pos-itive 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 presence of the expected quantities of plutonium.
D.
FUEL CYCLE AND CONTROL 1.
Control of the Clean Core The objectives of control system in the reactor are to provide sufficient reactivity variation to permit the plant to be brought to the various desired operating conditions at the will of the operator. Specifically, this means the plant must be suitably shut down in the cold condition and yet must have enough re-activity to permit operation at the desired power level for a period of time sufficient to correspond to the necessary re-fueling interval. It is important to note, however, that the full reactivity swing of the initial fuel loading need not be controlled by the mechanical control system of the reactor.
Indeed, it is inefficient to do so.
In the Big Rock reactor plant, the mechanical control rod system is designed to pro-vide reactivity control in excess of that ' required for all operating purposes and for the fuel burnup reactivity changes experienced between refueling intervals for equilibrium fuel cycling. However, the necessity for starting the core in-itially with a full core of fresh fuel introduces a large re-activity swing when long fuel irradiation is attempted from the first. Mechanical control system of the Big Rock reactor is not designed to handle this initial reactivity swing without the aid of additional reactivity controlling devices.
Conse-quently, it is expected that an auxiliary means of control will be used to provide control during the first several fuel cycling intervals, but will no longer be required once an equilibrium refueling pattern has been established.
Temporary poison is expected to take the form of either stainless steel channels or burnable poison. Stainless steel channels may be substituted for part or all of the zircaloy channels in the initial core as fuel irradiation progresses and as additional reactivity is needed, the reactor may be shut down and a fraction of the stainless steel channels will be replaced with zircaloy channels. After several such op-(
erating periods, the stainless steel channels will have been removed from the core entirely, and the reactivity swing
IV-8 remaining will be within the range of the equilibrium core re-g activity changes and can, therefore, be controlled by the me-chanical control system.
An alternate to the use of stainless steel channels is burnable poisons.
Burnable poisons may be located uniformly or non-uniformly within the reactor core to provide the necessary supplement to the mechanical control system. The burnable poisons will be utilized in such a manner that they will not adversely affect the reactivity coefficients of the reactor or the power distribution throughout reactor lifetime.
Thirty-two uniformly spaced control rods of cruciform shape with a 7-7/8" blade width are used for normal reactivity con-trol. With the inclusion of temporary poison in the form of stainless steel channels or burnable poisons, no regions with-out adequate control exist.
Because of the reduced diffusion length in water at room temperature and because the reactivity of the core reaches its maximum in the vicinity of room temperature, the ab-solute strength of the control system must be set depending upon the characteristics of the cold reactor. The present I
control system will provide a shutdown margin of approx-imately 3 percent in k when used in conjunction with the temporary poisons mentioned above. A similar shutdown margin is available in the equilibrium core when temporary poisons may not be necessary.
A tabulation of the reactivity con' trol requirements for the Big Rock reactor is given in Table IV-2.
During the ma-jority of core lifetimes, adequate excess reactivity will be available for override of the xenon transient.
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IV-9 TABLE IV - 2 REACTIVITY CONTROL REQUIREMENTS Initial Equilibrium Core Core Initial enrichment, percent U-235
- 3. 9
- 3. 9 Design irradiation, MWD /t 10,000 10,000 Discharge enrichment, percent U-235
- 2. 8
- 2. 8 Total Pu concentration, percent Pu 0.43 0.43 Control Reactivity Balance Temperature
.01
.01 Voids
.03
.03 Xe and Sm poisoning
.035
.035 Fuel cycle and maneuvering
. 16
.055 Shutdown margin
.03
.03 Total Ak
.265
. 16 Control Strength Available Control elements
. 20
.16*
Temporary poison
.065 none Total Ak
.265
. 16 Backup Emergency Control, Sodium Pentaborate 20 Boron atoms per cc of core coolant 2x10 Minimum control strength (diffused), Ak
- 0. 26
- Minimum which will be allowed due to burnup of control rod material.
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IV-10.
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- 2.
' Refueling The Big Rock reactor is designed to be refueled on a progres-
^
sive partial batch schedule, although the detailed refueling pla'n. has not yet been specified and probably will not be en-L tirely determined until after operation has begun. It is ex-pected that a plan similar to.the following will be used. The initial core will be operated six months, at which time part of the stainless steel channels (provided stainless steel channels are used as the temporary poison) in the core will be removed and replaced with zircaloy. This will be done approximately uniformly across the core. During succeeding
+-
refueling intervals, the reme.inder of the stainless steel chan-l nels will be removed, until the core no longer has reactivity control which is capable of removal, and insertion of clean fuel becomes necessary.
j 3
It is anticipated that an inward progression refueling schedule may be used. In this case, the central section of the core will be removed and the' remainder of the core will be moved pro-g res sively inward. A fresh batch of fuel corresponding to ap-proximately 20 to 30 percent of the core will be inserted near the periphery of the core. This form of fuel progression is
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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 j
MWD /t, averaged over the first core. which will correspond to equilibrium irradiation somewhat longer once equilibrium refueling has been established. The reactivity of the suc-ceeding batches of fuel which are added to the core may be increased to something in excess of the reactivity of the initial core to correspond to the additional control capacity which is available as a result of partial burnout of the re-actor core. In the event that higher reactivity fuel is 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, and to avoid local criticality p roblem's.
3.
Off-standard Control Conditions The control system of the Big Rock reactor has been designed with various off-standard situations in mind. Specifically, the
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effects of stuck control rod mechanisms are considered.
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IV-11
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In a light water reactor containing appreciable excess reac-tivity, the cold clean core can contain a large number of critical masses. 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 tc, consider local criticality caused by inoperative control mechanisms. The Big Rock reactor has sufficient excess shutdown capacity that it can be satis-factorily brought ' elow 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 con-ditions, of course, the control system is stronger and the reactivity of the core is somewhat less.
Consequently, the stuck rod situation is less severe and several adjoining con-trol rods may be lost without endangering this shutdown capability. If several such control rods are stuck from the core while hot, shutdown can be achieved with the back-up safety system (liquid poison) to permit reduction of the reactor temperature to a sufficient level that repairs can be made. Consequently, stuck control rods usually do not con-stitute a sa fety hazard during power operation in the hot con-dition.
k The local comrol characteristics of the lattice also dictate a startup accident configuration, in which several control rods are withdrawn surrounding a central inserted control rod. This configuration increases the statistical weight in the remaining inserted control rod and makes its withdrawal produce a large change in reactivity.
Consideration of a pessimistic configuration combi'ral with the maximum possible speed of rod withdrawal will yield a maximum reactivity in-sertion rate which will be used in determining the startup accident.
4.
Special Operating Considerations The large number of critical masses in the reactor core dic-tates special procedures at some conditions of plant operation.
For example, the loading of the core or the insertion of new fuel requires treatment of each leuc 1 region of freshly in-serted fuel as a special critical experiment.
In the same fashion, normal plant operation will involve bring-ing a local region of the core to criticality and then opening up the critical region as more reactivity or power generating area
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is required, until eventually the entire core is contributing to the power production. Operational procedures will be designed with these characteristics in mind.
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