Note: Descriptions are shown in the official language in which they were submitted.
~ 1 74983
LOW COOLANT VOID REACTIVITY FUEL BUNDLE
FIELD OF THE INVENTION
This invention relates to a fuel bundle design for a heavy water moderated
and cooled reactor and in particular, to a fuel bundle that provides control of the
reactivity upon coolant voiding in a reactor such as a CANDU type nuclear reactor.
BACKGROUND OF THE INVENTION
The reactor core of a heavy water moderated and cooled nuclear reactor such
10 as a CANDU type reactor contains a plurality of fuel channels each of which
contains a plurality of fuel bundles placed end-to-end inside a pressure tube. Each
fuel bundle contains a set of fuel pins of fissionable material. When the reactor is in
operation, heavy water coolant flows over the fuel bundles to cool the fuel and
remove heat from the fission process. The coolant also acts to transfer the heat to a
steam generator that drives a turbine to produce electrical energy. Because the
coolant circuit is pressurized, the coolant does not boil significantly and the density,
temperature and volume of the coolant in the fuel channels remain generally
constant with time.
When there is a breach in the coolant circuit such as in a loss of coolant
20 accident, the coolant starts to boil and coolant voiding occurs. In a heavy water
cooled reactor, the rate of neutron multiplication increases rapidly when the coolant
in the fuel channel is lost. This phenomenon is due to positive coolant void
reactivity and is an undesirable occurrence. Reactivity is a measure of the ability of a
reactor to multiply neutrons. Positive coolant void reactivity is the increase in
reactivity due to voiding of the coolant.
2~ 74983
The problem of positive coolant void reactivity is especially relevant in a
major loss of coolant accident which results in a high rate of discharge of coolant
from the reactor core and consequently a high rate of increase in the reactor power
level. In anticipation of such a situation, it has been necessary to provide a
shutdown system to detect abnormal situations in the reactor and respond by
initiating steps to arrest the increased rate of neutron multiplication without
damaging the fuel. The importance of such a shutdown ~y~leln is accentuated by the
fact that the coolant is no longer available to cool the fuel. Heavy water reactors
such as CANDU type reactors have been equipped with two equally effective and
10 totally redundant shutdown ~y~lel~s.
Although shutdown systems are an accepted means of ensuring safe
operation of the reactor, an uncontrolled increase in the rate of neutron
multiplication is an inherently unstable behaviour of the reactor. Furthermore,
licensing authorities are becoming increasingly concerned about the potential
problem of coolant void reactivity and the safety concerns which it generates. It has
been suggested by some licensing authorities that the peak value of the positivecoolant void reactivity be limited to the prompt critical reactivity which is the
reactivity at which the cycle time of neutron multiplication starts to drop rapidly.
One possible means to reduce the value of the positive coolant void reactivity
20 is to configure a fuel bundle specifically intended to minimize void reactivity on
loss of coolant. A number of approaches have been considered in the prior art.
The first approach is to add burnable poisons to the fuel bundle. An example
of this approach was disclosed in Canadian patent application No. 2,097,412 of the
inventor Ardeshir R. Dastur filed May 31,1993 and published December 1,1994. In
the fuel bundle described in the application, the creation of negative reactivity is
achieved by the judicious distribution of burnable neutron poisons and of fissile
2 1 74983
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material amongst the fuel pins. The design uses enriched uranium for producing
an increase in neutron multiplication in the outer region of the fuel bundle (where
the thermal neutron flux tends to decrease upon a decrease in coolant density) and
depleted uranium in a central region of the fuel bundle (where the thermal neutron
flux tends to increase upon a decrease in coolant density). Neutron absorbing
material is mixed with the depleted uranium to absorb thermal neutrons in the
central region. The rate of neutron absorption in the mixture of the fertile material
and the absorber in the central region increases as the mixture redistributes the
neutron flux across the bundle on coolant voiding and thereby produces negative
10 reactivity when the coolant density decreases. One of the principal disadvantages of
this approach is the requirement for the use of enriched fuel as opposed to natural
uranium. Natural uranium fuel is the predominant fuel of CANDU type reactors.
The increased need for U235 enriched fuel represents a significant reduction in the
resource utilisation advantage that CANDU type reactors have over light water
reactors. In addition, U235 fuel is not as accessible as natural uranium in manycountries of the world and in particular, in many of the countries where CANDU
reactors are currently used. This reduced accessibility is a disadvantage in and of
itself and also results in significantly increased fuelling costs.
Another approach has been to replace the fuel material in the central pins
20 with graphite. The primary disadvantage of this approach is that the reduction in
positive coolant void reactivity is limited. In addition, the use of graphite in the
fuel bundle raises the possibility of the rapid release of neutron kinetic energy that is
stored in the graphite. Furthermore, this approach achieves a reduced fuel exit
burnup and leads to a reduction in the rated power of the bundle.
Another approach has been to replace the central fuel pins with a core of non-
displaceable scattering material. Such a concept is described by M. H. M. Roshd et al.
in Transactions of the American Nuclear Society, 1977 Annual Meeting, June 16 - 17,
2 1 7~983
~ .
1977. In particular, calculations based on conventional 61 and 37 pin CANDU fuelbundles modified by removing the central pins and replacing them with air filled or
D2O filled aluminum cans showed that a large reduction in coolant void effect can
be achieved.
Some prior art fuel pin geometries adopt a relatively close packed
arrangement of fuel pins in an effort to increase efficiency or improve energy
output. For example, in U. S. Patent No. 3,179,571 Schabert et al. there is described a
fuel unit having a geometric arrangement of differently sized fuel pins in an
annular area about a central coolant filled tube. While this arrangement is stated to
10 reduce the self-shielding effect and provide a uniform distribution of coolant, the
design has not been optimized to reduce coolant void reactivity.
While some of the prior art designs have been successful in significantly
reducing the coolant void reactivity without requiring the addition of either
enriched fuel or burnable poisons as materials in the fuel pins, such designs
generally suffer a penalty in fuel exit burnup and maximum linear element rating.
Fuel exit burnup is a measure of the amount of fissile material consumed (or power
produced) before the fuel element is removed from the reactor. Maximum linear
power rating is a measure of the fuel centreline temperature at the bundle poweroutput limit. Accordingly, while reductions in coolant void reactivity can be
20 acieved with prior designs, there is a need for a fuel bundle which uses natural
uranium fuel and which controls coolant void reactivity without incurring an
unacceptable penalty in fuel exit burnup maximum linear power rating and withoutmaking unacceptable demands on the fuel handling system.
2 1 74 983
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SUMMARY OF l~IE INVENTION
The present invention is based on the discovery that coolant void reactivity
in heavy water cooled and moderated reactors is dependant on the mean chord
length of the coolant in the subchannels between the fuel pins and that the
reactivity can be decreased by reducing the mean chord length. In the present
invention, the fuel bundle geometry provides reduced mean chord length and
reduced coolant void reactivity.
The fuel bundle design contains a central area of neutron scattering material
that does not void from the central area upon a loss of coolant accident. In an
10 annular area about the central region, a plurality of fuel pins are disposed with
coolant flow subchannels therebetween. The fuel pins can advantageously be
disposed in concentric rings, each ring having an equal number of fuel pins with the
diameter of the fuel pins being graduated from one ring to another such that theoutermost ring has the largest fuel pins and the innermost ring has the smallest fuel
pins.
In accordance with the present invention there is provided a fuel bundle for a
heavy water cooled and moderated reactor comprising a central cylindrical regioncontaining a neutron scattering material which does not void from said central
region upon a loss of coolant accident and an annular region co-axially disposed20 about said central region containing a plurality of fuel pins of elongated cylindrical
shape disposed in parallel spaced relation with coolant flow subchannels
therebetween, said fuel pins being uniformly disposed in said annular region in
equal numbers in a plurality of concentric rings, said fuel pins in each concentric
ring being smaller in diameter than the fuel pins in the adjacent outer ring anddisposed on radial lines midway between the radial lines on which adjacent fuel
pins in adjacent rings are disposed.
2~ 74983
-
In accordance with another aspect of the present invention, the difference in
radii of adjacent rings is less than the sum of the radius of each fuel pin in one
adjacent ring and the radius of each fuel pin in the other adjacent ring.
The above and other aspects of the present invention will become apparent
from the following description taken with the accompanying drawings.
BRIEF DESCRIPIION OF THE DRAWINGS
In drawings which illustrate embodiments of the invention,
FIG. 1 is a cross-sectional view of a 37-pin fuel bundle design of the prior art;
FIG. 2 is a cross-sectional view of a 42-pin fuel bundle design of the present
10 invention;
FIG. 3 is a portion of a cross-sectional view of the fuel bundle design of FIG. 2
showing the geometry of the subchannel;
FIG. 4 is a cross-sectional view of a 36-pin fuel bundle design of the present
invention; and
FIG. 5 is a cross-sectional view of a 30-pin fuel bundle design of the prior artwhich includes a non-displaceable central scattering core of graphite.
21 74983
-
DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT
The fuel bundle design commonly used in CANDU reactors is the 37-pin
design shown in FIG. 1. The fuel bundle, generally designated by reference numeral
8, contains a set of thirty-seven equally sized like fuel pins, one of which is
designated by reference numeral 10, disposed in three concentric rings 12, 14, 16
about a central fuel pin 18. The fuel material in each of the fuel pins is generally
natural uranium in the form of uranium dioxide pellets (not shown). The diameterof fuel pins 10 is 13.08 mm. A plurality of fuel bundles 8 arranged end-to-end in
pressure tube 20 is encased within c~l~n~1ria tube 22. Gas, typically helium or carbon
10 dioxide, is present within the annular space 24 between pressure tube 20 and
calandria tube 22 to thermally insulate pressure tube 20 from calandria tube 22 and
the heavy water moderator which flows in the space outside calandria tube 22.
Heavy water coolant is contained within pressure tube 20 and fills the subchannels
26 between the fuel pins. When the reactor is in operation, the coolant flows over
the fuel bundle to cool the fuel and remove heat from the fission process.
Minimizing coolant void reactivity was not a predominant consideration in
developing the prior art 37-pin fuel bundle design.
The fuel bundle of the present invention and the manner in which it
minimi7es coolant void reactivity can best be understood by first considering how
20 coolant void reactivity arises in prior art fuel bundles such as the 37-pin design of
Fig. 1. In a nuclear reactor, the relative rates of neutron production and neutron
absorption determine the rate of neutron multiplication. Loss of coolant produces
positive coolant void reactivity because it causes a loss of neutron scattering from
the lattice and consequently an alteration in the speed of neutrons which has a
direct effect on the rates of neutron production and neutron absorption.
Fast neutrons, such as fission neutrons which have not been slowed down by
2 l 74983
-
the moderator, can produce fission in all uranium and plutonium isotopes
including U238. The minimum neutron speed required to produce such fission
corresponds to an energy of about 1 Mev. The natural uranium fuel used in
CANDU type reactors is comprised of less than approximately 0.73 percent U235 and
Pu239 isotopes. Most of the rem~inc1er (approximately 99 percent) consists of U238.
Because relatively few U235 and Pu239 nuclides are present compared with U238, the
probability of a fast neutron colliding with these nuclides and producing fission
before it collides with a nuclide of scattering material and slows down below anenergy level of 1 Mev, is small. In contrast there is a much greater probability that a
10 fast neutron will produce fission in U238 because of the large number of U238nuclides present. Fission of U238 by fast neutrons contributes almost 6 percent to
energy production and 2.2 percent to neutron multiplication in CANDU type
reactors.
There are two types of fissions by fast neutrons: (i) fissions of U238 in a fuelpin that are produced by fast neutrons that were born in the same fuel pin, and (ii)
fissions of U238 in a fuel pin that are produced by fast neutrons that were born in
other fuel pins. The rate of U238 fission produced by external fast neutrons is
relatively small because the fast neutron must travel through the heavy water
coolant without collision to maintain sufficient speed to cause fission in U238. Even
20 a single collision of a neutron in heavy water is sufficient to slow a 1 Mev neutron
to an energy that can no longer produce fission in U238. When the coolant has
voided, there is therefore an increased probability of a fast neutron reaching another
pin without collision and accordingly, the rate of type (ii) U238 fission referred to
above increases on loss of coolant. The resulting increase in the rate of neutron
multiplication accounts for approximately one third of the coolant void reactivity in
CANDU type reactors.
21 74983
-
The abundance of U238 in natural uranium fuel is responsible for another
major source of positive coolant void reactivity. In particular, the increase in
reactivity is a result of the effect of the coolant on the rate of neutron absorption.
Although neutrons of all speeds are absorbed by U238, the probability of neutron
collision increases as the neutrons slow down and eventually reach the energy level
of the heavy water moderator. The rate of slow (approximately 0.5 ev) neutron
absorption in U238 is orders of magnitude higher than the rate of fast (1 to 2 Mev)
neutron absorption. Between these two limits, the relationship between absorption
rate and neutron speed is monotonic except at a relatively narrow energy band of
10 about 100 kev. At this so called "resonance energy band" the rate of neutron
absorption by U238 is disproportionately high. In fact, over 25 percent of the total
neutron absorption occurs within this energy band. In the lattice of the reactor, fast
neutrons are slowed down by collisions in the moderator and to some extent by
collisions in the coolant. The collisions in the coolant cause a significant number of
neutrons to be shifted into the resonance energy band which, in turn, promotes
their absorption. Upon loss of coolant, this shift no longer occurs and the
absorption rate of U238 neutrons is therefore reduced. This reduction in the U238
absorption rate accounts for another approximately one third of the positive coolant
void reactivity in CANDU type reactors.
The remaining third of coolant void reactivity production in CANDU type
reactors is a result of isotopes other than U238, for instance U235 and Pu239, and the
higher actinides and fission products.
It has been discovered that the amount of coolant void reactivity produced
depends on the probability that a neutron will collide with a nuclide of the coolant
when the neutron is travelling between fuel surfaces. This probability is
determined by the number of nudides of the coolant that a neutron will encounter
between fuel surfaces. The latter parameter is called the coolant chord length. The
21 74983
-
mean chord length of the coolant is the mean distance that a neutron must travel
through the coolant between fuel surfaces. It is directly proportional to the volume
of coolant in the subchannel divided by the surface area of the three fuel pins in the
subchannel. It depends on the shape of the subchannel, the volume of the coolant
in the subchannel and on the angular direction in which the neutrons are
travelling.
The fuel bundle design of the present invention exploits the above described
behaviour of neutrons to minimi7e coolant void reactivity. In FIG. 2, a 42-pin fuel
design of the present invention is shown. Pressure tube 32 contains a plurality of
10 fuel bundles, one of which is designated by reference numeral 30. The fuel bundles
are arranged end-to-end along the length of the fuel channel (not shown). Gas,
typically helium or carbon dioxide, is present within the annular space 34 between
pressure tube 32 and c~l~n~lria tube 28. Calandria tube 28 is surrounded by
moderator.
Fuel bundle 30 has two concentric rings of elongated cylindrically shaped fuel
pins, two of which are ~lesign~ted by reference numeral 44 in FIG. 2. Fuel pins 44 are
disposed in parallel spaced relationship in concentric outer ring 36 and inner ring 38
about a central solid cylinder 40 of neutron scattering material. A zirconium alloy
sheath 46 surrounds the perimeter of cylinder 40. Heavy water coolant flows
20 through the subch~nnels 42 about fuel pins 44 between pressure tube 32 and central
cylinder 40. Fuel pins 44 and central cylinder 40 are held together by endplates (not
shown) at either end of the fuel bundle.
In the 42-pin design, outer ring 36 and inner ring 38 each contain twenty-one
fuel pins. Outer ring 36 has a pitch circle radius of approximately 44.7 mm and
inner ring 38 has a pitch circle radius of approximately 34.7 mm. The pitch circle
radius is defined as the radius of the circle which passes through the centres of the
11 2174983
fuel pins comprising the ring. Each of fuel pins 44 in outer ring 36 has a diameter of
11.40 mm and each of fuel pins 44 in inner ring 38 has a diameter of 9.40 mm. Fuel
pins 44 are clad in a zirconium alloy sheath (not shown) having a thickness of
approximately 0.3 mm to 0.4 mm. Central cylinder 40 has a diameter of 55.3 mm
and sheath 46 has a thickness of at least 0.3 mm. Fuel pins 44 contain pellets of
natural uranium dioxide as fuel material. Alternatively, slightly enriched uranium
dioxide, plutonium dioxide mixed with uranium dioxide or unreprocessed spent
light water reactor fuel (referred to as DUPIC) may alternatively be used.
The geometric arrangement of the pins is shown in detail in FIG. 3. The
10 centres of outer ring fuel pins 44A and 44B are disposed along radial lines 48 and 50
respectively. The centre of inner ring fuel pin 52 is disposed along radial line 54
which is midway between radial lines 48 and 50. Consequently, the distance
between pin 52 and pin 44A (which distance is designated reference numeral 56) is
equal to the distance between pin 52 and pin 44B (which distance is designated
refelel,ce numeral 58). Angle 60 is 17.143 degrees and angle 62 is half thereof.
In the fuel bundle design of the present invention the reduced size and the
configuration of the subchannel between adjacent pins reduces the mean chord
length and decreases the probability of collisions occurring in the coolant.
Consequently, there is a decreased probability that during the slowing down process,
20 neutrons will be scattered into the resonance energy band. Therefore, under normal
operating conditions, the absorption rate of resonance energy neutrons in U238 is
low compared to the corresponding rate in standard fuel designs such as the 37-pin
fuel bundle of FIG. 1. On loss of coolant, there is normally a reduction in the
number of neutrons slowing down into the resonance energy band and hence a
decrease in the rate of absorption which leads to increased neutron multiplication.
However, in the fuel bundle design of the present invention, the reduced number
of neutrons being scattered into the resonance energy band which occurs on loss of
2 ~ 74 983
12
coolant produces significantly less decrease in the absorption rate of resonanceneutrons in U238 because the absorption rate of such neutrons in U238 is already low
under normal operating conditions. This minimal disparity in absorption rates
before and after loss of coolant minimi7.es the increase in neutron multiplication
and thus contributes to the reduction in coolant void reactivity.
The reduction in void reactivity of the design of the present invention is
attributable to a number of factors. The presence of central cylinder 40 provides a
centrally disposed scattering volume that does not void on loss of coolant
minimi7ing the disparity in neutron multiplication before and after loss of coolant.
10 In addition, in the present invention, the pins of the inner ring are nestled midway
between the adjacent pins of the outer ring which advantageously reduces the size
of the subchannels between the pins. As used herein, the term 'subchannel' refers
to the area deffned by a triangle whose apices are the centre of the fuel pin of one
ring and the centres of the adjacent fuel pins of an adjacent ring less the area of that
triangle occupied by the fuel pins. This close nestling of pins which results inreduced chord length and reduced subchannel volume is made possible by the use
of an equal number of smaller pins in the inner ring as in the outer ring and the
positioning of the inner pins on radial lines midway between the adjacent pins in
the outer ring.
As shown in FIG. 3, with the present invention, the difference in the pitch
circle radius of adjacent rings can be equal to or less than the sum of the radius of
pin 44A and the radius of pin 52 permitting a smaller sub~h~nnel volume than is
typical in prior designs. With this arrangement, the radially outermost limit of the
fuel pins in the inner ring can be equal to or greater than the radially innermost
limit of the fuel pins in the outer ring which allows the pins to be nestled and the
subchannel volume reduced while maintaining adequate spacing for
thermohydraulic and fuel engineering considerations. With the particular design of
~ 1 ~49PJ3
13
FIG. 2, the maximum value of the minimum distance between any two adjacent
fuel pins or between a fuel pin and the pressure tube or between a fuel pin and the
central sheath should be limited to approximately 2.0 mm. The minimum value of
the minimum distance is dictated by thermal hydraulic and other fuel engineeringrequirements and can be as low as approximately 0.5 mm.
In the present invention, the use of smaller pins in the inner ring is to be
contrasted with other prior art ~lesign~ such as Schabert et al. in which the fuel pins
in the inner ring are larger in diameter and therefore have a larger surface area than
the pins in outer rings. As noted above, mean chord length is dependent in part on
10 the surface area of the pins in contact with the coolant. The reduced average pin
size also decreases the probability of collisions of U238 within a fuel pin and
increases the probability of a fast neutron (such as an uncollided fission neutron)
escaping a fuel pin without collision. The reduced pin size is also a factor which
contributes to the size and shape of the subchannel between adjacent pins necessary
to decrease the probability of collisions occurring in the coolant. The close nestling
of the pins is also advantageous because it provides additional space in the centre of
the fuel bundle for neutron scattering material.
The placement of pins having a larger diameter in the outer ring is contrary
to conventional thinking because fuel located near the outside of the fuel bundle is
20 vulnerable to the highest operating temperature. Thus to n-inimi7e the
temperature increase in the fuel, fuel bundles are conventionally designed with the
smaller pins placed in the outer ring in order to accommodate a larger number ofpins. In the present invention, the placement of the larger fuel pins in the outside
ring does not impact unfavourably on fuel temperature because the neutron flux
level in the two rings is closer with this type of design. In addition, the overall size
of the fuel pins can advantageously be sufficiently small so as to not impact
unfavourably on fuel temperature.
21 74983
14
The central scattering volume of the present invention also reduces the
coolant mean chord length of the subchannels by reducing the track length of theneutrons in the coolant that are scattering between fuel pins. Collision of fastneutrons in the central scattering volume reduces the probability, on loss of coolant,
of fast neutrons (such as uncollided fission neutrons) escaping a first pin, reaching a
second pin without collision, and producing fission of U238 in the second pin. The
magnitude of this effect can be controlled by appropriate selection of the central
scattering material.
Central cylinder 40 preferably contains amorphous carbon as the neutron
10 scattering material. Amorphous carbon is suitable for the neutron scattering
material for several reasons. It is an effective neutron scattering material which
does not absorb neutrons to any appreciable degree. It can also be specifically
manufactured to have a density ranging from low to its theoretical density. In
general, a higher density of neutron scattering material results in a greater reduction
in coolant void reactivity. The ability to adjust the density allows a degree of control
over the level of coolant void reactivity realized by the fuel bundle. Amorphouscarbon is also suitable for the formation of the central cylinder because it is not
chemically poisonous and because it prevents the explosive release of stored
neutron kinetic energy during reactor operation. Reticulated silicon carbide is
20 another suitable material for the formation of the central cylinder because it shares
the characteristics of amorphous carbon outlined above.
Other scattering materials such as graphite and heavy water may also be used
although due consideration must be given to engineering constraints. For example,
while heavy water has scattering properties, provision would have to be made to
ensure that it does not void from the central core upon a loss of coolant accident. In
addition, the present invention may also include designs in which the neutron
scattering in the central region is provided substantially by the effect of the structural
21 74983
central cylinder 46 and the interior space that it defines. The central cylinder may
contain helium, carbon dioxide or other low density, non-corrosive and non-
neutron absorbing materials. Alternatively, the central cylinder may be solid
zirconium alloy.
While the present invention has been described as having 42 pins and the
specific geometry referred to above, a person skilled in the art will appreciate that
these details can be varied while remaining within the spirit and scope of the
present invention. For example, FIG. 4 shows a cross-sectional view of an
alternative 36-pin design. In this embodiment, there are eighteen pins per ring.10 The diameter of the pins in the outer ring 64 (one of which is designated reference
numeral 66) is 11.4 mm and the diameter of the pins in inner ring 68 (one of which
is designated reference numeral 70) is 11.0 mm. The pitch circle radii of outer ring
64 is 44.6 mm and the pitch circle radius of inner ring 68 is 34.5 mm. The thickness
of the zirconium alloy sheath (not shown) surrounding the fuel pins is between
about 0.3 mm and 0.4 mm. There is a central cylinder of amorphous carbon 72 or
other suitable neutron scattering material with a diameter of 54.6 mm and a
zirconium alloy sheath 74 having a thickness of at least 0.3 mm. Angle 73 is
approximately 20 degrees and angle 75 is approximately 10 degrees. The
arrangement of the pins is based on the same considerations as noted above with
20 respect to the 42-pin design. The m~nner in which this design controls coolant void
reactivity is based on the same principles as the 42-pin design. The 36-pin design
provides a higher power output and a marginally lower average fuel exit burnup
than the 37-pin design, especially when amorphous carbon is used.
Not only does the fuel bundle design of the present invention offer improved
coolant void reactivity over prior designs, it does so without significant penalty to
fuel burnup or maximum linear heat rating. For purposes of comparison, Fig. 5
represents the prior art 37-pin fuel bundle design shown in FIG.1 which has been
2 1 74983
16
modified by the use of a non-displaceable central scattering volume 40, as suggested
by Roshd et al. improve coolant void reactivity.
The fuel bundle ~1esign~ described above were subjected to WIMS and MCNP
simulations to evaluate coolant void reactivity, fuel exit burn-up and maximum
linear rating. The WIMS simulation code is available from the Reactor Shielding
Information Centre in Oakridge, Tennessee and is commonly used in conventional
reactor design. MCNP simulation is described by J.F. Briesmeister in A General
Monte Carlo Code For N-Particle Transport, LA-12625, 1988. The following fuel
designs were evaluated:
(i) 37-pin fuel bundle design shown in FIG. 1;
(ii) 42-pin fuel bundle design shown in FIG. 2;
(iii) 36-pin fuel bundle design shown in FIG. 4;
(iv) 34-pin fuel bundle design shown in FIG. 5; and
(v) 30-pin fuel bundle design shown in FIG. 6
The results of the evaluation are shown in Table 1 below. To conduct the
simulations, data was entered relating to the geometry, composition and operating
temperature of the lattice. The values given in each case are for equivalent reactor
operating conditions.
21 74983
17
TABLE 1
Void-Reactivity (mk) FuelExit Max. Linear
BetaWIMS MCNP Burnup Heat Rating
Design (mk) Fresh Equil Fresh Equil (MWd/Te) (kW/m)
42-Pin 6.18.7 5.6 8.3+.50 5.26208 58.4
36-Pin 5.910.1 65 9.4+.40 5.86527 60.2
37-Pin 5.9 17.2 13.8 16.5+.36 13.1 6690 60.7
30-Pin 59 12.7 8.7 12.1+.47 8.16~5 70A
All designs included in Table 1 use natural uranium in the form of uranium
10 dioxide as the fuel material. The scatterring material in the central core for the 36-
pin, 37-pin and 30-pin designs is amorphous carbon having a density of 1.6 g/cc.The values given for void reactivity are considered to be conservative because they
are expected to drop by between 0.5 and 1.0 mk due to the increase in neutron
leakage on coolant voiding which is not included in the calculations. The
maximum linear heat rating values are for a bundle power of 950kW. The MCNP
equilibrium void reactivity values have been inferred from the drop in void
reactivity with burnup as shown by WIMS.
The results in Table 1 show the clear superiority of the 42 and 36-pin designs
of the present invention over the standard prior art 37-pin design and as modified
20 with the inclusion of a central scattering core as represented by the 30-pin design. In
the case of the standard 37-pin design, all void reactivity values are substantially in
excess of the prompt critical value beta. While the 30-pin design which is a
21 74983
modification to the 37-pin design to replace the central 7 pins with a non-
displaceable scattering core does improve the void reactivity numbers, it introduces
a large penalty in the maximum linear heat rating value.
The 42 and 36 element designs of the present invention show extremely good
void reactivity numbers, without penalty in burn-up or maximum linear heat
rating. For example, the 42-pin design has an equilibrium coolant void reactivity
below prompt critical. This is also true of the 36-pin design of the present invention
when one takes into account the further 0.5 to 1 mk drop due to increased reactor
leakage on coolant voiding that is not reflected in Table 1. In contrast, neither the
10 37-pin or the 30-pin designs achieve below prompt critical void reactivity.
Moreover, in contrast to the 30-pin design that achieves some improvement
in void reactivity at the expense of maximum linear heat rating, both the 42-pin and
36-pin designs have maximum linear heat ratings below that of the standard 37-pin
design. In additon, there is only a slight fuel burnup penalty for the 42-pin and 36-
pin designs as compared to the 37-pin design. In the case of the 36-pin design, the
fuel burnup penalty is less than that of the 30-pin design. In fact, the 36-pin design is
superior to the 30-pin design in void reactivity, burnup and heat rating.
The present invention provides a fuel bundle which acts to control the
reactivity upon coolant voiding. As natural uranium may be used as the fuel
20 material the need for enriched fuel is elimin~ted. The fuel bundle is also
advantageous because there is little penalty of fuel burnup reduction or bundle
power rating reduction. Furthermore, the fuel bundle adopts the overall geometryof current CANDU fuel bundle designs and can therefore be used in currently
operating CANDU reactors.
While the invention has been described in connection with three specific
2 1 74983
19
embodiments thereof, numerous modifications and adaptations of those
embodiments are contemplated by the inventors and will occur to those skilled inthe art without departing from the spirit and scope of the invention as set forth in
the appended claims. For example, although the 42-pin and 36-pin designs described
here and shown in FIGS. 2 to 4 have two concentric rings of fuel pins, the present
invention could include a design having fuel pins arranged in three or more
concentric rings provided that each ring has an equal number of fuel pins and
provided the diameter of the fuel pins gets progressively larger from the innermost
ring to the outermost ring.
