Method for flattening the curve of evolution of heat in a fast reactor core

The core of a fast reactor having at least three core regions is made up of vertical fuel elements each having an axial passage of sufficient diameter to permit the flow of molten fissile material in the event of a power excursion. The diameter decreases from the central region of the core to the periphery, the decrease being proportional to the reduction in neutron flux density in order to maintain the integral of conductivity at a substantially constant value.

BACKGROUND OF THE INVENTION 
This invention relates to a method for flattening the curve which is 
representative of the evolution of heat within a nuclear reactor core in 
the radial direction and in the axial direction of said core. The 
invention also relates to a nuclear reactor core for the application of 
said method. 
In the more exact terms, the present invention is intended to ensure that 
the heat flux released by a nuclear reactor core in the radial directions 
and in its axial direction is made as uniform as possible. 
In order to gain a clearer understanding of the problem, reference will be 
made to FIG. 1 of the accompanying drawings in which is shown 
diagrammatically the curve of distribution of neutron flux within a 
nuclear reactor core. 
More precisely, said core is concerned with the case in which the 
enrichment of the fissile material constituting the reactor core is 
homogeneous. 
In this figure, the neutron flux is plotted as ordinates along the axis Oy 
and the distance from the axis of the reactor core is plotted on the axis 
of abscissae Ox. It is considered in this figure that the reactor core has 
symmetry of revolution about its axis Oy. Said neutron flux is found to 
decrease substantially from the axis to the periphery of the reactor core. 
The zone a represents the fissile region of the core and the zone c 
represents the fertile blanket. As is already known, this phenomenon is 
due to the fact that fast neutrons exhibit a strong tendency to escape at 
the periphery of the reactor core. 
The curve of evolution of heat has substantially the same shape in the case 
in which the enrichment is the same throughout the reactor core. 
It is clearly an advantage to obtain a flatter curve of evolution of heat. 
FIG. 2 shows a known method employed precisely for flattening the curve 
which is representative of the evolution of heat. In this figure, the 
reactor core is shown diagrammatically in vertical half-section. 
The axis Ox represents the distance with respect to the vertical axis Oy of 
the reactor core. In this embodiment, the central portion of the reactor 
core a is constituted by fissile material which has a degree of enrichment 
E.sub.1. The central zone a is surrounded by an annular peripheral zone 
having a thickness b. In this zone, the fissile material has a degree of 
enrichment E.sub.2 which is higher than E.sub.1. Finally and in accordance 
with known practice in breeder reactors, provision is made for a second 
annulus having a thickness c which constitutes the radial fertile blanket. 
There is also shown in this drawing the curve I which gives the neutron 
flux as a relative value, that is, as a percentage of the maximum flux 
along the axis of the reactor core. 
The full-line curve I just mentioned is a continuous curve. 
The higher degree of enrichment within the zone having a thickness b simply 
has the effect of slowing-down the fall in neutron flux without, however, 
raising it to any considerable extent. This change is indicated by the 
point of inflexation A. 
The evolution of heat is represented by the broken-line curve II. It is 
apparent that this non-continuous curve is constituted by three portions. 
With the scales adopted as ordinates, the first portion II.sub.a follows 
the curve which gives the neutron flux whereas in the first peripheral 
zone, the portion of curve II.sub.b is located distinctly above the 
neutron-flux curve I, this being clearly due to the increased degree of 
enrichment of the fissile material in this zone. Finally, the third 
portion II.sub.c corresponds to a very slight evolution of heat within the 
radial fertile blanket as a result of a very weak enrichment (natural 
uranium or more generally depleted uranium). 
It is thereof readily apparent that this core structure permits a certain 
flattening of the radial curve of evolution of heat. 
Moreover, there was described in U.S. Pat. No. 3,932,217 granted Jan. 13, 
1976 to Pierre Charles Cachera and assigned to Electricite de France, a 
method for achieving enhanced safety of a fast reactor which consists in 
employing safety fuel elements in which is formed a central passage or 
axial flow duct of large diameter on the order of 10/15 of the diameter of 
the fuel element and having a sufficient diameter to ensure that the fuel 
which may be caused to melt within the central portion of the element 
under the action of an accidental power excursion is capable of flowing by 
gravity rapidly into the lower blanket placed as a catchpot without 
impairing the fuel can. Each safety element has a stack of fissile fuel 
pellets in the can with each pellet having this axial flow duct and also 
having a lower member of refractory material with an axial duct at the top 
of substantially the same diameter as the duct in the fissile portion. 
This arrangement makes it possible to ensure enhanced reactor safety since 
there is obtained at the time of an accidental power excursion of 
reduction of the quantity of fissile material within the central region of 
the core in which the neutron flux has the highest intensity, thus 
resulting in a decrease in reactivity which automatically puts an end to 
said power excursion. 
BRIEF DESCRIPTION OF THE INVENTION 
This invention is precisely directed to a method for flattening the curve 
of evolution of heat within a nuclear reactor core which again makes it 
possible to retain the method of safety described in the foregoing but 
also provides advantages of an economic order. 
The method adopted for flattening the curve of evolution of heat of fuel 
elements within the core of a liquid metal cooled fast neutron reactor is 
distinguished by the fact that the reactor core has at least three regions 
constituted by vertical fuel elements of constant enrichment and of the 
same outer diameter, said fuel elements having an axial passage whose 
diameter decreases from the central region of the core to the periphery at 
which the neutron flux decreases so as to produce in each section of the 
fuel an integral of conductivity in the vicinity of the maximum value 
adopted or in other words to ensure that the temperature at the limit of 
the central passage is at all points in the vicinity of the maximum 
temperature adapted for the fuel material (for example 2,250.degree. C. in 
the case of a mixed UO.sub.2 -PUO.sub.2 oxide). Thus the wall of said 
central passage will have in vertical cross-section a profile which 
corresponds to an isothermal curve. It would therefore be possible to 
maintain this structure under irradiation since it is known that the fuel 
tends to be rearranged towards said isothermal internal profile during 
inpile irradiation. 
With reference to the term "integral of conductivity", it is considered as 
universal (See "UO.sub.2 Properties affecting performances" by M. F. Lyons 
et al., page 8). These terms may be defined in the following manner. 
In the fissible material, the thermal conductivity is a function of the 
temperature .lambda.(T). 
At the hottest point of the fissile material (the center of the pellet or 
the edge of the hole), a very high temperature T.sub.0 obtains. On the 
periphery of the pellet, the fissile material, is at its lower temperature 
T.sub.1. 
By definition, one calls "integral of conductivity" the integral 
##EQU1## 
wherein T is the temperature within the fissile material. 
In a fuel pin, the integral of conductivity is defined only along a line 
passing through the axis of the pin, said line connecting the center of 
the pellet (or the edge of the central hole) to the outer edge of the 
pellet. 
The present invention is also concerned with a fast reactor core having at 
least three regions, said core being essentially constituted by an 
assembly of vertical fuel elements of constant enrichment and of the same 
outer diameter, each fuel element having an axial passage of sufficient 
diameter to permit the flow of molten fissile material, the diameter of 
each axial passage of the fissile portions of the fuel elements which are 
located at the center region of the reactor core being larger than that of 
the fuel elements located in core regions adjacent to the periphery of 
said reactor core, the diameter of axial passage of all fuel elements 
which are located at the same distance from the axis of the reactor core 
being of the same value at the same height. 
Moreover, in each fuel element, the diameter of the axial passage 
progressively decreases from the center of the fuel element to its lower 
end (as shown in FIG. 3). 
The novel method of flattening of the curve of evolution of radial heat is 
employed with the use of a constant enrichment, thereby securing the 
advantages of reduction in the void percentage at the periphery, slight 
reduction in flux at the periphery which in turn has the effect of 
reducing the neutron flux and therefore of improving the breeding gain, 
simplification of cheking operations during manufacture of the fuel. 
Simplification of inspection and checking during fuel manufacture makes it 
possible to adopt a reactor core which has at least three fuel zones.

DESCRIPTION OF PREFERRED EMBODIMENTS 
As stated earlier, the method consists in forming the nuclear reactor core 
by means of fuel elements of constant enrichment and of the same outer 
diameter, the axial canal diameter of at least the central fuel elements 
being sufficient to permit the flow of molten fissile material. This canal 
diameter varies, however, according to the position of the fuel element 
within the reactor core. 
In more precise terms, the reactor core comprises at least three sections 
having fuel elements with different axial diameters, the diameter of the 
axial passage decreasing from the center to the periphery of the reactor 
core either in the axial or radial direction, the reduction in diameter 
being proportional to the reduction in neutron flux density so as to 
maintain the integral of conductivity at a substantially constant value. 
FIG. 3 is a vertical sectional view of a fuel element which is partly 
identical with that shown in the patent cited earlier. 
Said fuel element comprises an external metallic can 12, a portion 14 of 
fissile material in which is formed a hollow axial passage 16. The fuel 
element is provided at the top portion thereof with a solid upper blanket 
18 and at the bottom portion thereof with a lower blanket 20 which 
constitutes a first catchpot 22. 
Beneath said lower blanket 20, provision is made for a fission-gas 
expansion chamber 24 and finally for a catchpot 20 of refractory material 
located at the extreme lower end and designed to form a molten core 
catchpot which is placed beneath the catchpot 22 and constitutes a "second 
line of defense". 
In accordance with will-known practice, the fissile zone 14 can be formed 
by a stack of pellets of enriched uranium oxide or of mixed oxide of 
plutanium and uranium (or thorium). The oxide may be replaced by any other 
chemical compound of U and Pu which melts freely. 
The differences between the fuel element shown in FIG. 3 and the fuel 
element described in the patent cited in the foregoing 
(N.degree.3,932,217) lies in the fact that the diameter of the axial 
passage 16 progressively decreases between the central zone of the fuel 
assembly represented in the figure by the dashed line 28 and the lower 
portion of the fissile zone of the fuel element represented by the dashed 
line 30. 
Since the zone 14 is formed by a stack of hollow sintered pellets, the 
variation in diameter can be achieved in a non-continuous manner at the 
outset at each change of pellet. The opening formed in the pellets can 
also be frusto-conical. Any small surface irregularities along the axial 
passage will in any case tend to disappear during irradiation. 
There is shown opposite to the fuel element the continuous curve III which 
corresponds to the neutron flux in the axial direction of the fuel 
element, that is, in the axial direction of the reactor core. 
For the reasons set forth in the foregoing, the flux curve falls very 
rapidly towards the upper end and the lower end of the fissile zone of the 
fuel element to a value of approximately one-half the value attained at 
the center. 
The corresponding evolution of heat is represented by the broken-line curve 
IV. 
Said curve IV is in fact made up of four portions. The portions IV.sub.a 
and IV.sub.b correspond to the upper and lower blankets of the fuel 
element. The evolution of heat in these portions is clearly very slight 
since the enrichment is of small value. 
The portion IV.sub.c corresponds to the evolution of heat in the upper half 
of the fissile zone of the fuel element. Scales can be chosen for this 
curve so as to ensure that said curve of evolution of heat coincides with 
the neutron flux curve within said zone. 
On the contrary, in the region IV.sub.d which corresponds to the lower half 
of the fissile zone of the fuel element, it is apparent that the 
progressive reduction in diameter of the axial passage permits a 
substantial increase in evolution of heat with respect to that which would 
have taken place without this modification of the axial passage. 
Said variation in diameter of the central passage is calculated so as to 
ensure that the integral of conductivity is practically retained, which is 
necessary in order to maintain the maximum temperature of the fuel below a 
predetermined temperature (2,250.degree. C., for example, in the case of 
UO.sub.2 -PuO.sub.2). 
When the coolant which usually consists of liquid sodium but can also be 
gas under pressure flows upwards within the reactor curve, the lifting of 
core IV.sub.d shown in FIG. 3 is highly advantageous. 
It would also be possible to contemplate the symmetrical reduction in 
diameter of the axial passage when passing from the central region of the 
fissile zone of the fuel element to the upper end of this latter. Such a 
modification would assuredly offer an advantage from an economic viewpoint 
but is not adopted in the preferred embodiment shown in FIG. 3 since this 
arrangement could give rise to doubtful intrinsic safety of the fuel 
element in the event of a power excursion which results in partial 
melt-down of fuel. In point of fact, the fuel which might melt in the 
upper portion of the fissile zone would be liable to increase the 
reactivity by flowing zones of higher neutron flux located at the 
mid-height of the reactor core and thus to reduce the reactivity drop 
which would be expected as a result of melt-down and of gravitational flow 
of part of the fuel which is present in the central and bottom zones of 
the reactor core. 
As shown in FIG. 3, the fuel is provided in cross-section with a fissile 
zone, the lower portion of which is in the form of a nozzle. In the event 
of a fast power excursion, the temperature and the pressure of the gas 
located within the central passage 16 increase much faster than the 
temperature and the pressure of the fission gas which is present within 
the catchpot 22 and the chamber 24. The gas is thus impelled downwards, 
with the result that the fall of molten fuel under the action of gravity 
will be considerably accelerated, particularly at the level of the nozzle. 
FIG. 4 shows how the provision of three core regions formed of fuel 
elements having different canal diameter allows a flattening of the curve 
of evolution of heat by maintaining the integral of conductivity at a 
substantially constant value. 
More precisely, the upper curve of FIG. 4 represents on a different scale 
the curve of FIG. 1 illustrating the neutron flux distribution as a 
function of the distance from the axis or center line of the reactor core, 
when the enrichment of the fuel elements is constant throughout the 
reactor core. 
Moreover, it is shown at the lower portion of FIG. 4 that the reactor core 
is formed of three regions, called core 1, core 2 and core 3, located 
concentrically relative to the center line of the core, the fuel elements 
of each regions having the same outer diameter, whereas the fuel elements 
10a of the center region (core 1) have an axial passage or canal 16a of a 
greater diameter than the fuel elements 10b of the middle region (core 2), 
the latter elements 10b having an axial passage or canal 16b of a greater 
diameter than the fuel elements 10c of the outer region (core 3). In the 
described embodiment, the fuel elements 10c are even shown as solid 
elements having no axial passage. All the fuel elements of each core 
region are identical. 
The lower curve of FIG. 4 shows that the diameter of the passages of the 
fuel elements in each core region are not chosen at random. On the 
contrary, and according to the teachings of the invention, the reduction 
in diameter of the passages from the center region (core 1) until the 
outer region (core 3) is proportional to the reduction in the neutron flux 
density. Thanks to this specific distribution, FIG. 4 shows that the 
integral of conductivity at a given height in the core in the radial 
direction is maintained between two relatively close values I.sub.1 and 
I.sub.2. In other words, the integral of conductivity is maintained at a 
substantially constant value, which can be chosen as the maximum value 
adopted, thereby improving the economic characteristics of the reactor 
core. 
Although three core regions have been described it will be understood that 
the core can be divided in four or more core regions without departing 
from the scope of the invention. 
It is known that, in respect of equal power density in the fuel, the 
adoption of hollow fuel results in a reactor core of greater bulk and in 
higher neutron leakages. 
The slight overenrichment and the resultant drop in breeding gain was the 
disadvantage attached to the fuel described in the patent cited earlier. 
The type of fuel described in the present patent specification makes it 
possible to achieve a considerable reduction of neutron leakages, firstly 
by reducing the void percentage within the reactor core and especially at 
the periphery of this latter but also because the peripheral elements have 
no increased enrichment and by accepting within this zone a steeper 
downward flux curve and therefore a slightly lower value of neutron flux 
at the level of and across the surface which separates the reactor core 
from the lateral blankets. 
Furthermore, even if only the fuels which are located within the internal 
zone of the reactor core retain a central passage of sufficient diameter 
to permit flow of the molten fuel under the action of gravity, the safety 
claimed in the patent cited above is still ensured. The reason for this is 
that the counter-reaction to power excursion has a particularly marked 
effect in the case of fuels located at the center, at which the molten 
fuel flows from a zone of maximum neutron flux. 
Furthermore, at the time of a very fast power excursion, the melt-down 
process will begin with the central fuels. Let it be assumed that the same 
integral of conductivity has been retained at the mid-height of the 
reactor core, both at the center and at the periphery. The same maximum 
fuel temperature will then exist at all points (for example 2,250.degree. 
C., namely a margin of 500.degree. C. with respect to the melting point of 
the UO.sub.2 -PuO.sub.2 mixture which is estimated at 2,750.degree. C.). 
At the time of the power excursion, the value of neutron flux at the center 
will remain at each moment as at the outset approximately twice the value 
at the periphery. In consequence, the margin which is assumed by way of 
example to have the value of 500.degree. C. will be absored twice as 
quickly at the center as at the periphery. 
Thus, even in the case of a normal operating regime which is as 
"isothermal" as possible, the central passages of largest diameter must in 
fact be located at the center of the reactor core at which the melt-down 
process will begin in order to facilitate the discharge of molten fissile 
material.