KR100465742B1 - Composite nuclear fuel material and method of manufacture of the material - Google Patents

Abstract

The present invention relates to a composite fuel material and a method of manufacturing the same.

SUMMARY OF THE INVENTION An object of the present invention is a composite fuel comprising a ceramic matrix which is inert during irradiation, and nuclear fuel particles having a high resistance to the progression of the fissure upon irradiation and having a high retention of volatile fission products. To provide the material.

The above object is achieved using a method that allows generating some micron spacing between the particles and the matrix.

Description

Composite nuclear fuel material and method of manufacture of the material}

The present invention relates to a nuclear fuel material, and more particularly to a nuclear fuel material having a high resistance to crack propagation during irradiation and having a high retention of volatile fission products and a method for producing the composite fuel material. will be.

Numerous problems arise during the use of nuclear fuel, especially with regard to the mechanical resistance of the fuel during irradiation. During irradiation, local stresses are created due to fuel expansion and the release of volatile fission products, causing fuel fuel fractures to rupture the fuel sheath. The interaction between material and fuel cladding due to internal pressures and cracks in the fuel cladding should be limited. As a result, the quality of the fuel used is of paramount importance and any change that is currently made in terms of improving the performance of the fuel should seek to limit the interaction of the cladding and the fuel.

In the document entitled 'Artie Energetique B811 3620-11 / 5.12', the fuel pellets are made of UO ₂ powder and U ₃ O ₈ is added to improve the solid form of the raw pellets. The raw pellets are prepared at high pressure so that the UO ₂ / U ₃ O ₈ particles are fixed by the addition of zinc stearate. This document also discusses the possibility of adding the product of porosity to the fuel. The obtained raw pellet is then sintered at high temperature.

Patent application FR-A-2 706 066 describes a nuclear fuel containing UO ₂ with improved retention properties of fission products. In this document, fuel includes metals such as Cr or Mo that can trap oxygen from fission, forming oxides.

However, none of the above documents simultaneously solves the problems caused by nuclear fuel expansion and the release of volatile fission products upon irradiation and generating local stresses inside the material.

It is an object of the present invention to accurately provide a novel fuel material which solves the above-mentioned problems, and a method of producing the fuel material.

This new type of composite fuel material comprises a ceramic matrix which is inert upon irradiation and the nuclear fuel particles are dispersed at intervals of 1 to 10 μm between the matrix and the particles and have a coefficient of thermal expansion lower than the coefficient of thermal expansion of the fuel particles. .

This matrix is inert upon irradiation and has a reduced effective neutron absorption cross-sectional area and a reduced expansion upon irradiation.

The matrix with the above characteristics is, for example, a ceramic such as spinel MgAl ₂ O ₄ , an oxide such as MgO magnesia, or yttrium oxide Y ₂ O ₃ . Preferably, the matrix used is spinel MgAl ₂ O ₄ .

Although the prior art mentions materials such as beryllium oxide, aluminum oxide, and zirconium oxide, these materials have a number of disadvantages. Beryllium oxide is a neutron reducer, and aluminum oxide and zirconium oxide have an expansion effect upon spinning. In addition, zirconium oxide is dissolved in UO ₂ at a temperature of 1200 ° C. or higher (phase change). These drawbacks result in no fuel material having the required properties.

Particles of the nuclear fuel is made of some form of, in general, the oxide particles are used as particles of a mixed oxide containing UO ₂ as UO _2, UO _2, or -PuO _2, UO ₂ -ThO _2. In addition, various additives may be added, such as known additives used to adjust particle size to improve retention of fission products.

Preferably, the particles of nuclear fuel used in the present invention are UO ₂ particles.

The particle size is chosen by sintering and then obtaining a gap of about 2-3 μm between the particle and the matrix. Generally, particles of 70-230 μm size are used, preferably 90-120 μm. In order to obtain the material of the present invention with the required properties, the proportion of UO _{2 in} the fuel material generally accounts for 20-40% of the total volume of the fuel material.

The distance between the matrix and the particles is generated by the expansion volume for the volatile fission products and, in part, controlled by a part of the UO ₂ fuel expansion upon irradiation limits the local stress generated during the use of nuclear fuel, and UO ₂ particles and Delay damage to the matrix.

In addition, the nuclear fuel material of the present invention has better thermal conductivity than that of UO ₂ single. The fuel temperature reached while the reactor is being irradiated is reduced along the thermal gradient obtained between the center of the pellet and the surroundings, making it desirable for the retention of fission products, and the heat dissipation due to the diffusion of fission products to UO ₂ . Is activated.

In addition, the nuclear fuel material of the present invention has excellent resistance to crack propagation inside the pellets. The gap present between the fuel particles and the matrix results in the bending of the cracks generated during the irradiation.

In the present invention also the following steps;

a) mixing particles of a nuclear fuel precursor having a diameter within the aforementioned particle size range with a powder of material to form a ceramic matrix inert upon irradiation,

b) pressing the mixture to form a shape, and

c) sintering the compacted mixture in a reducing atmosphere comprising H ₂ O, under conditions such that the crystal lattice volume of the nuclear fuel precursor is reduced during the heat treatment, for example by heat treatment.

It provides a method for producing a nuclear fuel material comprising a.

If the fuel should be UO ₂ , the precursor that can be used for the volume reduction of this characteristic crystal lattice is either U ₃ O ₈ or U ₃ O ₈ -UO ₂ mixture.

While the tetragonal U ₃ O ₈ oxide is converted into cubic structured uranium dioxide UO ₂ , the crystal lattice volume of the fuel is reduced between the matrix and the particles to form a few micron gap.

While mixing the precursor particles with the matrix powder, it is essential to remove large lumps of matrix powder that may form during the mixing, as a way to optimize the performance during compression and then to increase the density.

The mixing step is most important as it determines the final homogeneity of the composite fuel. This should be done slowly in order not to produce additional particulates.

For example, it may be mixed for at least 30 minutes by mechanical stirring, using a blade mixer or a turbula mixer set at a low speed of 20 rpm.

Cold pressing of the mixture, in particular, mixture cooling pressing into pellet formation, can be carried out using conventional methods. For example, it can be produced using hydraulic pressure, preferably twin effect press.

In addition, a double compression cycle with an intermediate adjustment device will limit the shortcomings related to the poor distribution of the mixture in the pressure matrix. At this pressure, the first pressing step can be performed at a pressure of 30 to 50 MPa, and the second step is performed at a pressure of 300 to 350 MPa. Each step includes a pressure rise cycle, a pressure hold cycle, and a pressure drop cycle.

For example, it is strongly required to smooth the compression matrix using a lubricant, zinc stearate in the form of an aerosol prior to compacting the powder.

In addition, in order to limit the separation of particles in the matrix powder, which lowers the homogeneity of the fuel pellets, the transfer time of the powder mixture composed of the matrix and the fuel particles from preparation to compression should be short enough. Any distribution defects at the manufacturing stage become defects that cannot be reabsorbed defects during sintering.

Sintering is carried out under the conditions (temperature, time, atmosphere) of converting the precursor to the fuel and suitably raising the density of the fuel and the matrix both materials.

In order to convert from precursor to fuel, a damp reducing atmosphere must be used for U ₃ O ₈ particles. It is preferably composed of hydrogen humidified, for example, by about 2% water volume, to allow sufficient reduction of U ₃ O ₈ to UO ₂ at temperatures of up to 600 ° C. Humidification during sintering activates the diffusion of cations, increasing the density of both materials. The reducing atmosphere described above also consists of a mixture of humidification with hydrogen and an inert gas such as argon.

When pure humidified hydrogen is used or a humidified hydrogen argon mixture is used, the partial pressure ratio of the pH ₂ / pH ₂ O ratio is 40/60, preferably 50 zones.

The sintering temperature and time are chosen in terms of the fuel and the material used as the matrix. For example, a temperature of 1640-1740 ° C. may be used. For example, for spinel MgAl ₂ O ₄ , the sintering temperature is 1650 ° C.

로 저하시킴을 연속적으로 포함할 수 있다.During sintering, the treatment cycle is allowed to raise the temperature at a rate of about 100 to 300 ° C./h up to the sintering temperature, maintain the sintering temperature for one hour, and maintain the temperature at a rate of 150 to 350 ° C./h

Lowering may be included.

According to the present invention, with the above method, it is possible to obtain a nuclear fuel material using a precursor of a ceramic matrix and a nuclear fuel which is inert during irradiation, and between the fuel particles and the matrix, which is produced as the precursor is converted into fuel and the particle size is reduced. It is characteristic in that gaps occur.

For example, the precursor of the particles because it is U ₃ O ₈ of orthorhombic structure, U ₃ O ₈ is UO ₂ particles having a small crystal lattice volume of the cubic structure than is obtained. The conversion from U ₃ O ₈ to UO ₂ is carried out at a low temperature of 600 ° C. or lower.

Before the process of densification of the UO ₂ particles and spinel matrix mixture begins, a gap occurs at the interface between the matrix and the UO ₂ particles due to a reduction in the underlying crystal lattice size in the region of 21% of the total volume.

During sintering, both components have similar densification kinetics, and since the densification of both components is initiated simultaneously, this gap is partially recovered during sintering.

Precursor particles include the following steps;

1) pressing the precursor powder to obtain a raw material compact,

2) grinding the raw material compact to obtain fine particles,

3) screening particulates,

4) spheroidizing the selected fine particles with precursor particles, and

5) removing by selecting precursor particles whose diameter is not within the above particle size range

Can be obtained by setting in the process.

When the fuel precursor is U ₃ O ₈ , it can be obtained by calcining the UO ₂ powder at a temperature of 450 to 500 ° C. This calcination is for example carried out in an alumina or Inconel vessel.

When the precursor described above was a mixture of UO ₂ and U ₃ O ₈ powders, the U ₃ O ₈ powder was prepared as before, and the amount of required UO ₂ powder was added.

The next preparation step for the particles of the U ₃ O ₈ fuel precursor is, for example, compressing the U ₃ O ₈ powder or the mixture of UO ₂ -U ₃ O _{8 to} 100 MPa to obtain a raw material compact, and the oscillating granulator ( Crushing the above-mentioned raw material compact to obtain fine particles, selecting the fine particles, spheroidizing the selected fine particles into particles, and the diameter of the particles is not within the above particle size range. By removing particles by screening.

The fine particles obtained by grinding the raw compacts of U ₃ O ₈ should be circular without sharp corners.

The ground fines are then sorted. Particulates are screened using a screen, preferably stainless steel. The mesh of the screen used is adapted to the particle size or the size of the required particle, providing for volume reduction due to changes in crystal structure during heat treatment.

The increase in spacing obtained after sintering depends in particular on the size of the fuel precursor particles.

The selection is carried out, for example, to select particles between 100 and 300 μm, preferably between 120 and 160 μm. In the case of screening at 120 to 160 mu m, the diameters of the particles obtained after sintering will be between 90 and 120 mu m.

For example, in a spherical vessel located at the center of gravity of the mixer enclosure, spheroidization is performed using a mixer or a turbula mixer or the like for at least 20 hours.

Other features and advantages of the present invention will become apparent with reference to the following examples, which are clearly non-limiting and given the purpose shown with reference to the appended drawings.

In this embodiment, the composite fuel material is prepared comprising a spinel matrix MgAl ₂ O ₄ in which particles of UO ₂ fuel are dispersed.

The first step in the preparation of the composite fuel material is to prepare precursor particles of UO ₂ fuel, which are U ₃ O ₈ particles.

The UO ₂ powder described above is calcined for 2 hours in an air at a temperature of 500 ° C. to obtain a good U ₃ O ₈ oxide powder.

For this operation, UO ₂ powder is fed to the alumina vessel. The height of the powder layer is 3 cm or less, so that complete oxidation can occur throughout the entire volume of the powder. Thus, U ₃ O ₈ raw material powder is obtained.

Thereafter, the U ₃ O ₈ raw material powder is pressed into a tablet-type at a pressure of 100 MPa using a dual effect hydraulic pressure. The raw material compact is then pulverized slowly to obtain circular fine particles without sharp corners using an oscillating granulator.

The particles are screened using a stainless steel screen with mesh pores of 125 to 160 μm. The particles obtained after sorting are then spheroidized. The spheroidization is carried out in a turbula mixer for 20 hours in a spherical vessel located in the gravity center of the mixer enclosure. After spheroidization, fine particles that are not within the particle size range (125-160 μm) are removed by screening.

The second step in the manufacture of the composite fuel material is to prepare the composite material.

In the first step of producing the composite fuel material, the U ₃ O ₈ particles described above are mixed with the matrix powder. The matrix powder described above is spinel MgAl ₂ O ₄ . In order to optimize the performance during densification after compaction, large lumps of spinel powder formed during mixing are removed.

Thereafter, mixing of the U ₃ O ₈ particles with the MgAl ₂ O ₄ spinel matrix is performed for 30 minutes by mechanical sterling using a blade mixer set at a low speed of at least 20 rpm.

After the first step at 50 MPa comprising a pressure rise cycle of 3 seconds, a pressure hold cycle of 4 seconds, and a pressure drop cycle of 3 seconds, a pressure rise cycle of 2 seconds, a pressure hold cycle of 4 seconds, and a pressure of 3 seconds With the second step at 300 MPa including the drop cycle, the mixture is then compressed using dual effect hydraulic pressure. The compressive matrix is smoothed with zinc stearate aerosol.

The compacted mixture of spinel matrix / U ₃ O ₈ precursors is then heat treated in a sintering oven.

This heat treatment is carried out for 1 hour at a reducing atmosphere of 1650 ° C. of H ₂ + 2% H ₂ O, raising the temperature at a rate of 150 ° C./h up to a sintering temperature of 1650 ° C., and maintaining this temperature for 1 hour. And lowering the above temperature to room temperature at a rate of 300 ° C./h.

The final density of the pellets in this example is above 94%, which is the theoretical density.

The spinel matrix and UO ₂ particles are very dense, as shown in the metal surface enlarged view of FIG. 1. The final shrinkage of the pellets described above is about 21%. This is greater than the contraction of the UO ₂ spinel complex in full contact between the matrix and the particles. This recovers some of the initial spacing created between the matrix and the particles. Therefore, the densification reaction rate of the composite material upon sintering is matched with that of the spinel matrix.

Thus, the presence of the particles does not interfere with the densification reaction of the material. Although the densification of both components separated begins at the same time, at the beginning of sintering, UO ₂ particles have a faster densification rate than MgAl ₂ O ₄ spinels. The phenomenon of partially recovering the gap between the matrix and the particles is explained by the explanation ratio by increasing the density of spinel from 1300 ° C or higher. At this temperature, the reaction rate of uranium dioxide is slowed while the sintering rate of spinel is not yet at its maximum.

Contact between the particles and the matrix during polishing of a sample of the fuel material according to the invention when the removal process of removing particles remaining after sectioning having a diameter shorter than the actual diameter of the U ₃ O ₈ particles is observed. This lack is explained. The more this removal is done, the thicker the spacing produced is and the larger the size of the initial U ₃ O ₈ particles.

The warpage phenomenon of the crack is shown in FIG. 2. The crack passes around the particle at the interface level with low resistance. Branching points and cracks in the cracks can be visualized to indicate reinforcement of the material.

Stress relief at the interface appears as reducing the proportion of UO ₂ particles cracked inside the composite material.

The composite fuel material according to the present invention includes a ceramic matrix which is inert during irradiation and nuclear fuel particles having high resistance to crack propagation during irradiation and high retention of volatile fission products, thereby preventing local stress from being formed. This prevents the fuel cladding tube from rupturing due to the fuel cracking material.

FIG. 1 shows a metal surface enlarged view (magnification 500) of UO ₂ particles having a diameter of 90 to 120 μm dispersed in a MgAl ₂ O ₄ spinel matrix, 1 to 1 between the matrix and the UO ₂ particles. An interval of 2 μm is shown.

FIG. 2 shows an enlarged view of a metal surface (magnification 50) of UO ₂ particles having a diameter of 90 to 120 μm dispersed in a MgAl ₂ O ₄ spinel matrix, and warpage of cracks due to the spacing between the matrix and the particles. Indicates.

Claims (19)

Inert upon irradiation, comprising a ceramic matrix in which particles of nuclear fuel are dispersed at intervals of 1 to 10 μm between, Wherein said matrix has a coefficient of thermal expansion lower than that of said fuel particles. The composite fuel material of claim 1, wherein the matrix is a material selected from spinel, magnesia, and yttrium oxide. The composite fuel material of claim 2, wherein the spinel is MgAl ₂ O ₄ . The composite fuel material of claim 1, wherein the dispersed fuel particles are particles of UO ₂ or mixed oxides containing UO ₂ . The composite fuel material of claim 1, wherein the particles dispersed in the matrix have a diameter between 70 μm and 230 μm. The composite fuel material of claim 4, wherein the particles dispersed in the matrix have a diameter of 70 μm to 230 μm. The composite fuel material of claim 1, wherein the particles dispersed in the matrix have a diameter of 90 μm to 120 μm. 5. A composite fuel material according to claim 4, wherein UO2 comprises 20-40% of the total volume of the fuel material. A method for producing a nuclear fuel material according to claim 1, comprising the following steps; a) mixing precursor particles of nuclear fuel having a diameter within the above particle size range with a powder of material to form a ceramic matrix inert upon irradiation, b) pressing the mixture to form a shape, and c) sintering the compacted mixture in a reducing atmosphere comprising H ₂ O, under conditions such that the crystal lattice volume of the nuclear fuel precursor is reduced during heat treatment. Method of producing a nuclear fuel material, characterized in that it comprises a. 10. The method of claim 9, wherein the reducing atmosphere comprises H ₂ and H ₂ O. 10. The method of claim 9, wherein the reducing atmosphere is a mixture of argon, H ₂ and H ₂ O. The method of producing a nuclear fuel material according to claim 10 or 11, wherein the partial pressure ratio of pH ₂ / pH ₂ O is 40/60. 10. The method of claim 9, wherein the diameter of the particles of the fuel precursor is between 100 and 300 [mu] m. 10. The method of claim 9, wherein the diameter of the particles of the fuel precursor is between 120 and 160 micrometers. 10. The method of claim 9, wherein the precursor is a U ₃ O ₈ or UO ₂ -U ₃ O ₈ mixture. 10. The method of claim 9, wherein the pressing to form the shape comprises the following two steps; A first compression step carried out at a pressure of 30 to 50 MPa, and Second crimping step carried out at a pressure of 300 to 350 MPa Method for producing a nuclear fuel material characterized in that carried out. 17. Nuclear fuel material according to any one of claims 9, 10, 11, 13, 14, 15 and 16, wherein sintering is carried out at a temperature of 1640-1700 ° C. Manufacturing method. The method of claim 9, wherein the precursor particles are the following steps; 1) compressing the precursor powder to obtain a raw material compact; 2) grinding the raw material compact to obtain fine particles, 3) selecting the fine particles, 4) spheroidizing the selected fine particles into precursor particles, and 5) removing by selecting precursor particles whose diameter is not within the above particle size range A method for producing a nuclear fuel material, characterized in that obtained by setting in the process. The method according to claim 18, wherein the precursor powder is U ₃ O ₈ , and U ₃ O ₈ is obtained by calcining UO ₂ at a temperature of 450 ° C. to 500 ° C. in air.

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