Plastic and nonmetallic article shaping or treating: processes – Shaping or treating radioactive material
Reexamination Certificate
1999-10-01
2001-04-24
Jenkins, Daniel J. (Department: 1742)
Plastic and nonmetallic article shaping or treating: processes
Shaping or treating radioactive material
C419S019000
Reexamination Certificate
active
06221286
ABSTRACT:
BACKGROUND OF THE INVENTION
In the use of nuclear fuels based on oxide, particularly uranium oxide, one of the problems caused is due to the release of fission gases in the element during the operation of the reactor, because these fission products must be kept in the fuel element, particularly in the actual fuel pellets, so as to limit the internal pressure of the sheaths or cans and the interaction of the fission products with the latter.
Therefore, at present, the burn-up of nuclear elements is limited to 50 GWj/t of U in order not to exceed the threshold beyond which the release of fission gases becomes significant.
However, operators of electronuclear reactors, particularly pressurized water reactors (PWR) wish to optimize control of the nuclear fuel by increasing burn-up of the uranium dioxide pellets contained in the rods in order to achieve minimum values of 60 to 70 GWj/tU.
Research carried out up to now for obtaining such an improvement has used procedures for increasing the size of the uranium dioxide grains, because ii: has been found that the gas quantity released by an irradiated large grain fuel is less than that released by an irradiated small grain fuel. Use has also been made of procedures for forming precipitates in the nuclear fuel in order to anchor the fission gases on said precipitates.
In order to obtain an increase in the size of the uranium dioxide grains, it is possible to add additives such as iO, Nb O, Cr O, Al O, V O and MgO to the uranium dioxide powder subject to fritting or sintering in order to activate its crystal growth, provided that the sintering takes place under a wet hydrogen atmosphere so that the added oxide quantity remains in solution in the uranium dioxide and is not reduced to a metallic element. The use of such additives for obtaining a large grain microstructure is, e.g. described by Killeen in Journal of Nuclear Materials, 88, 1980, pp 177-184, Sawbridge et al in Report CEGB RD/B/N 4866, July 1980 and Radford et al in Scientific Paper 81-7D2-PTFOR-P2, 1981. However, the use of certain additives of this type can lead to an increase in the diffusion coefficients of cations and fission gases in the uranium dioxide, which is unfavorable for the retention of the fission products and does not make it possible to take full advantage of the large grain microstructure.
Another procedure for improving the retention rate of nuclear fuels consists of dispersing in the uranium dioxide grains nanoprecipitates of a second phase for ensuring the anchoring of the fission products on such second phase. Nanoprecipitates of this type can consist of magnesium oxide inclusions, as described by Sawbridge et al in Journal of Nuclear Materials, 95, 1980, pp 119-128 and in FR-A-2 026 251.
SUMMARY OF THE INVENTION
The present invention makes use of a method different from that described hereinbefore for improving the retention rate of fission products in a nuclear fuel . This method consists of trapping the oxygen atoms released by the fission of the uranium and/or plutonium atoms, so as to maintain the O/U (Th, Pu) or O/M stoichiometry with M=U+Pu or U+Th or U+Pu+Th of the fuel at 2 and thus prevent a rise in the diffusion coeffients in the fuel and a reduction of its thermal conductivity, the increase in the diffusion coefficients is a mechanism leading to the accumulation of fission products at the grain boundaries, followed by release of the fission products. In the same way, a reduction of the thermal conductivity of the fuel is prejudicial, because it has the effect of increasing the temperature of the fuel for the same linear power and consequently both reducing the solubility of the fission products and favoring their diffusion.
The invention also relates to a process for improving the retention of fission products within a ceramic nuclear fuel based on UO
2
, ThO
2
and/or PuO
2
, which consists of including in the ceramic nuclear fuel at least one metal able to trap oxygen by forming an oxide having a free formation enthalpy at the operating temperature T of the nuclear reactor below the free formation enthalpy at the same temperature T of the superstoichiometric oxide or oxides of formulas (U, Th)O
2+x
and/or (U, Pu)O
2+x
in which x is such that 0<x≦0.01.
The use of such a metallic additive consequently makes it possible to maintain the O/U (Th or Pu) or O/M ratio defined hereinbefore of the nuclear fuel at a value of 2 and in this a way to avoid an increase in the diffusion coefficients, which remain at a low value, and a reduction in the thermal conductivity of the fuel. Thus, a high fission product retention rate is obtained.
This procedure can be combined with known methods of increasing the size of the UO
2
and/or PuO
2
and/or ThO
2
grains and forming precipitates for anchoring the fission gases, which is very interesting and makes it possible to improve the performance characteristics of the fuel.
In the case of uranium dioxide-based fuels, the free formation enthalpy of the superstoichiometric oxide UO
2+x
with 0<x≦0.01 can be expressed in oxygen potential and calculated on the basis of the law of Lindemer and Besmann, as described in Journal of Nuclear Materials, 130, 1985, pp 473-488. In this case and as is indicated on p 480 thereof, the oxygen potential &Dgr;G(O
2
) of the above-defined superstoichiometric oxide can be expressed in J/mole according to the following formula:
−360 000+214 T+4 RTLn[2
x
(1−2x)/(1−4
x
)
2
]
in which R is the molar constant of the gases, T is the temperature in Kelvins and x is as defined hereinbefore.
Moreover, for uranium dioxide-based fuels, the metal included in the fuel must be able to form an oxide having an oxygen potential defined by the formula: &Dgr;G(O
2
)=RT Ln (pO
2
) in which R is the molar constant of the gases, T the reactor operating temperature and p(O
2
) the partial oxygen pressure, equal to or below the above-estimated value for UO
2+x
in accordance with the Lindemer and Besmann law. Examples of suitable metals are Cr, Mo, Ti, Nb and U.
The invention also relates to a fuel for nuclear reactors comprising a ceramic material based on UO
2
, ThO
2
and/or PuO
2
in which is dispersed at least one metal able to trap oxygen and having the characteristics given hereinbefore.
According to the invention, the oxide-based ceramic material can be constituted by UO
2
, ThO
2
PuO
2
or mixtures thereof, the mixed oxide UO
2
—PuO
2
or UO
2
—ThO
2
, mixed oxides based on UO
2
and other oxides such as oxides of rare earths or mixed oxides based on PuO
2
.
Preferably, the ceramic material is based on UO
2
and the dispersed metal is able to form an oxide having an oxygen potential below the oxygen potential of UO
2+x
, as described herinbefore.
In general, to obtain a burn-up of 60 GWj/t
−1
: the dispersed metal represents 0.1 to 2% by weight of the fuel material. Preferably, the metal is chromium and represents 0.1 to 1 or better still 0.2 to 0.5% by weight of the fuel material.
Moreover, the fuel material can also comprise additives such as TiO
2
, Nb
2
O
5
, Cr
2
O
3
, Al
2
O
3
, V
2
O
5
and MgO, in order to increase the size of the fuel grains and/or aid the anchoring of the fission products, as well as other additives, e.g., SiO
2
, in order to improve other properties.
The fuel material according to the invention can be prepared by conventional sintering or fritting processes by adding to the ceramic material powder to be sintered the metal, either in metallic form, or in the form of an oxide or oxygenated compound.
In the first case, after shaping the powder by cold compression, sintering is carried out in a dry hydrogen atmosphere e.g. having a water content below 0.05 volume % so as not to oxidize the metal.
In the second case, if it is wished to simultaneously obtain a size increase of the UO
2
, ThO
2
and/or PuO
2
grains, use is made of an oxide or oxygenated compound quantity which may or may not exceed the solubility limit of the oxide or oxygenated compound in
Dehaudt Philippe
Peres Veronique
Framatome
Jenkins Daniel J.
Pollock, Vande Sande & Amernick, R.L.L.P.
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