This invention refers to the tailoring of the density of nuclear
fuel pellets so that the local microscopic density of the fuel in a pellet is greater
than the overall density of the pellet, the method being based on prefabricated
The industrial fabrication of oxide nuclear fuel for light water
reactors or pressurised water reactors is in general achieved by precipitating
feed solutions which result in powders, by pressing these powders to become green
pellets and by thermally treating the pellets at elevated temperature under a
As conventional procedures exercise little control over the powder
generation, additional powder processing is required. In particular, the necessary
milling step results in large quantities of fine powders, which deposit on the
walls of the glove box and on the surfaces of the processing equipment in that
box. If mixed oxide fuel (MOX) is to be prepared, these deposits can cause a radiation
hazard to the personnel. Furthermore, the consequences of these deposits will
be exacerbated in future, as older Pu material which has been subjected to numerous
irradiation cycles will be used which will have an isotopic distribution such that
the radiation dose will be higher.
Most present fabrication routes lead to fuel pins consisting of pellets
with relatively uniform density (circa 95% of the theoretical value). Since the
present irradiation time to which the fuel can be exposed in the reactor, is determined
by the swelling of the fuel and the consecutive interaction with the cladding,
it should be useful to tailor the density of the pellets in order to reduce the
swelling effect, thereby extending the irradiation period of the fuel pins in
Following traditional methods, tailored porosities could be achieved
by adding pore formers to the powder, by pressing the mixture and by pyrolysing
the pore formers. The main disadvantage of this method lies in the pyrolysis which
requires specific atmospheric conditions and must be complete before sintering
of the pellets takes place. In addition, the pyrolysis is never complete, thus
leaving potentially detrimental residues.
It is thus un object of the present invention to propose a method
for reducing the overall density of the pellets to 90% or even less while still
maintaining a high local density (about 95%). The high local density is desired
in order to avoid sintering effects during irradiation of the fuel in the reactor.
This method should further avoid the use of pore formers and their subsequent pyrolysis.
This object is achieved by the method as defined in claim 1. Preferred
embodiments of this method are defined in the dependent claims.
The invention will now be described in more detail by means of some
According to a first embodiment, fuel pellets are fabricated based
uniquely on uranium. Uranyl-nitrate solutions are adjusted to a viscosity of about
0.02 to 0.1 Pas (Pascal sec) by adding trace amounts of long chain organic thickeners
such as Methocel, Dextran, Polyvinyl alcohol. The feed solution is dispersed into
droplets of selected size, which are then collected in a hydroxide bath. Due to
the network formed by the long chain organic polymers, precipitation then occurs
within the original droplets, so that nearly spherical beads are formed.
After removal of the nitrate salts and organic polymers, the water
content of the beads is reduced by an azeotropic distillation in a suitable immiscible
organic solvent or by the application of microwaves to the beads submerged in
a liquid which preferably does not absorb microwave energy.
The dimensions of the precursor beads are determined by the solution
properties and the type and operating conditions of the droplet disperser. Preferably
the beads should have diameters not smaller than 20 µm (to avoid dust problems)
and not larger than 300 µm (for good pressing properties). The precursor beads
obtained by the above described process are free flowing and dust-free; they do
not require any other physical treatment (e.g. milling) prior to pressing. Thus
the build-up of dusts in the glove box and consequently harmful radiation levels
are avoided. In addition, the beads produced in this way exhibit excellent sintering
A first predetermined quantity of precursor beads is then thermally
treated under an oxidising atmosphere (e.g. air, CO2) at a temperature
below 1150°C, preferably below 900°C, while a second predetermined quantity of
such beads is treated at the same temperature under a reducing atmosphere (e.g.
Ar/H2). During this treatment uranyl hydroxides in the beads of the
first quantity are converted into U3O8, while the beads of
the second quantity are converted to UO2.
It is well known that the density of U3O8 (ρ
= 8300 kg/m3) is lower than that of UO2 (ρ = 10960 kg/m3).
The next step consists in obtaining green pellets by compressing
a mixture of the beads of said first and said second quantity. These pellets are
then thermally treated under a reducing atmosphere (e.g. Ar/H2) at temperatures
less than 1100°C, preferably less than 900°C until the U3O8
beads in the pellets have converted into UO2 thereby reducing their
volume and creating voids. The pellets are then heated up to a temperature above
1600°C and regions corresponding to the precursor beads sinter to high density,
to give a backbone ensuring the mechanical stability of the pellets and at a later
stage in the reactor itself. The presence of the voids, however, hinders sintering
between neighbouring beads in the pellets. Thus, the voids remain after sintering
and pellets with high local densities, but with lower overall density are obtained.
In the case of MOX fuel the invention can be applied in the same
way: The Pu material should be present preferably in the beads of the second quantity,
so that the stoichiometry is significantly below 2.0 and mixed with U308
of the first quantity. The Pu content of the second quantity should be higher than
in the final product fuel and in proportion to the dilution achieved by mixing
with uranium. As the theoretical density of U3O8 of the first
quantity and the Pu containing material of the second quantity are both less than
the density of the final product, the thermal treatment applied after pellet pressing
results in a material with locally high density but a low overall density. The
distribution of the Pu is not homogeneous throughout the pellet,but this is not
detrimental to the mechanical stability during irradiation in the reactor.
The invention is not restricted to the methods as described above.
Thus, the size and number of the voids can be further controlled by selecting different
sizes for the precursor beads of both quantities: Smaller size of the beads of
the first quantity increases the number of voids, larger size of these beads increases
the size of the voids.
The invention therefore provides nuclear fuel pellets having a reduced
overall density which enhances stability against swelling during service in the
reactor, but having a backbone structure of high density which avoids further
sintering of the pellets during service. As a result, a fuel pin containing such
pellets supports a higher burn-up rate than conventionally equipped fuel pins.