WO2007065836A1

WO2007065836A1 - Nuclear fuel containing a burnable poison and method of manufacturing such a fuel

Info

Publication number: WO2007065836A1
Application number: PCT/EP2006/069091
Authority: WO
Inventors: Fabrice Mazaudier; Gabriel Pham
Original assignee: Commissariat A L'energie Atomique
Priority date: 2005-12-05
Filing date: 2006-11-30
Publication date: 2007-06-14
Also published as: FR2894374B1; FR2894374A1

Abstract

The subject of the present invention is a nuclear fuel based on uranium oxide, which contains a consumable poison formed by boron, the boron being dispersed in the fuel in the form of glass/titanium diboride, the glass containing 7% boron by weight. The subject of the present invention is also a method of manufacturing the fuel according to the present invention, in which after sintering the pellets, a reduction step has been omitted.

Description

NUCLEAR FUEL COMPRISING A CONSUMABLE POISON AND A METHOD OF MANUFACTURING SUCH A FUEL

DESCRIPTION OF THE TECHNICAL FIELD AND PRIOR ART

The present invention relates to a nuclear fuel based on uranium oxide comprising boron, as a consumable neutron poison and to a process for the production in the form of pellets of such a fuel.

Nuclear fuel pellets of this type can be used in nuclear reactors such as pressurized water reactors.

The fuel which is used in the core of a nuclear reactor, has the role of providing energy in the form of heat by fission of the nuclides (uranium, plutonium, thorium, ...) which it contains, under the neutron action. In pressurized water reactors (PWR), this fuel is in the form of uranium dioxide pellets

(OU ₂ ). However, for the sake of using high energy value radioactive elements from reprocessing plants, part of the fuel present in the reactor core (up to 30% of the total fuel present in the core) can be MOX fuel (Mixed Oxide fuel) comprising a mixture of uranium oxide and plutonium oxide (the plutonium oxide content in the fuel can be between 3% and 7.5% depending on the location in the core). If the use of MOX has the double advantage of making it possible to limit the quantities of plutonium to be stored and to save uranium whose resources are exhaustible, the addition of plutonium in the fuel has the disadvantage of increasing reactivity of the heart because the ²³⁹ Pu of plutonium oxide is more reactive than the ²³⁵ U of uranium oxide, in particular, during the "critical in terms of reactivity" phases of starting the reactor, d '' or power variation.

Limitation of reactivity is ensured by materials comprising neutron-absorbing nuclides. The control and command bars which contain boron, in the form of B ₄ C and a ternary absorbent alloy Silver-Indium-Cadmium (AIC), constitute the first means of regulating, or even eliminating (in case of emergency) the neutron flux in which all the fuel rods bathe.

Adding boric acid H ₃ BO ₃ to the primary circuit is a second way of regulating reactivity.

However, these two means of reactivity control have limits. The addition of boron in the elements of metallic structures poses problems of resistance of important materials (embrittlement of the structures) and the quantities of soluble boron in the coolant are also limited.

There is a third means of controlling the neutron flux by which a consumable neutron poison (eg Gd ₂ O ₃ , Eu ₂ O ₃ ) is introduced directly into the fuel. This method appears as particularly interesting because it allows self-regulation of the production of neutrons. However, there are risks of disturbing the physical behavior of the pastilles in the heart.

Several elements of the rare earth family such as Gd, Er, Eu, Dy, Hf are a priori good potential candidates to be incorporated into the fuel. The choice between these different elements is the result of an optimization study between different criteria:

• The first of these criteria is the absorption capacity which is expressed by the value of the neutron absorption cross section expressed in barn (1 barn = ICT ²⁴ cm ² ). In general, isotopes like ¹⁵⁵ Eu, ¹⁵⁷ Gd, ¹⁶⁴ Dy, have a neutron efficiency well suited to a domain of the neutron spectrum resulting in numerous peaks (or resonances) for particular values of the kinetic energy of the neutrons. In particular, gadolinium is an excellent consumable neutron poison. In the presence of thermal neutrons, the two isotopes ¹⁵⁵ Gd and ¹⁵⁷ Gd have very high capture sections.

• The second criterion to take into account concerns the exhaustion kinetics which, for high combustion rate objectives, must be spread over time.

• The third criterion to take into account concerns the adaptation to the neutron spectrum and more precisely the way in which this cross section varies as a function of the kinetic energy of the neutrons. This adaptability plays an important role in the stability of steering.

• The fourth criterion to take into account concerns the easy availability of the element so as not to require an additional enrichment operation.

The different elements Gd, Er, Eu, Dy, Hf each have advantages.

• U0 ₂ pellets doped with rare earths (Gd, Er, Eu, Dy) in the form of oxide are now used industrially. By reaction of successive captures, they gradually transform into stable and non-capturing isotopes without modification of the chemical composition.

• Europium has an effective cross section of high neutron absorption. The advantage of this element lies in its reaction products which are themselves good absorbents, which increases their lifespan.

• Erbium, which has a lower absorption cross-section than Gd or Eu, is an element suitable for very long life cycle management. On the other hand, it participates in a slight degradation of the thermal conductivity of the fuel.

• The incorporation of neutron poisons in the form of rare earth oxides and hafnium in UO ₂ does not present any particular difficulties: these compounds are introduced and dissolve at high temperature in the material to form solid solutions fluoritics. The pellets are thus produced by reducing sintering at 1700 ^° C.

The various elements Gd, Er, Eu, Dy, Hf do not, however, meet all the criteria set out above.

On the other hand, the isotope ¹⁰ B (natural boron is a mixture of the isotopes ¹⁰ B (19.8% _m ) + ¹¹ B (80.2% _m )), has, compared to the other absorbents, an effectiveness in a very broad spectrum, from fast neutrons to thermal neutrons, which prompts its use for piloting PWRs. The capture reaction, of type (n, α), is given below:

¹⁰ B + ¹ Ti -> ⁷ Li + ⁴ He + 2.6 MeV.

The isotope ¹⁰ B makes it possible to comply with all the criteria described above, since its absorption efficiency is good, the enrichment process is industrially controlled and its behavior under irradiation is well known (the effective section of absorption of nuclide ⁷ Li produced by absorption of a neutron, is negligible).

There are currently several types of process for incorporating boron into the fuel.

A first type of process consists of coating the pellets or covering them with a film. Document US 4560575 describes the coating of fuel pellets (UO ₂ , ThO ₂ or PuO ₂ ) with BN. The fuel pellet is exposed to a gas stream of boron trichloride and anhydrous ammonia at a temperature of about 600-800 ^° C. At these temperatures, the problem of corrosion of uranium oxide by hydrochloric acid does not arise. A boron nitride layer is formed by the product of the reaction of boron trichloride with anhydrous ammonia on the fuel pellet: BCl ₃ + NH ₃ → BN + 3 HCl.

Document WO 9508827 proposes a nuclear fuel comprising boron particles and / or a boron compound coated with a film retaining boron. The film can be metallic (alloy or simple body), or ceramic (carbide, nitride, silicide or oxide).

There are also methods for direct incorporation of boron, in particular document US 3431329 describes a method for incorporating ZrB ₂ . To avoid oxidation of ZrB ₂ , particles of UH ₃ are added to the powder mixture. The oxygen produced by the reduction of UO ₂ preferentially causes the oxidation of UH ₃ and the formation of water vapor with the hydrogen released. ZrB ₂ is stable under these conditions.

However, the methods described above do not seem to take account of the volatility of boric anhydride B ₂ O ₃ from 1000 ^° C. Thus, unlike the isotopes absorbing rare earths which can be directly introduced in the form of oxides ( ex: Gd ₂ O ₃ , Er ₂ O ₃ ), boron cannot be introduced directly into the UO ₂ powder intended for sintering: boric anhydride which forms on contact with UO ₂ , melts at 45O ⁰ C to volatilize at 1000 ^° C., that is to say at a temperature below that usually reached when sintering raw uranium oxide pellets.

Document FR2 814 584 proposes a process for the direct incorporation of boron into the fuel with stabilization of the latter, in order to limit the losses of boron during the manufacture of the pellets.

In document FR 2 814 584, the nuclear fuel based on UO ₂ comprises a consumable poison comprising boron dispersed in the fuel in the form of metal boride and glass containing boron. The addition of glass makes it possible to solve the problem of the formation of boric anhydride B ₂ O ₃ from 1000 ^° C. Thanks to the association of a glass containing boron with a boride such as titanium boride TiB ₂ , the zirconium boride ZrB ₂ and the hafnium boride HfB ₂ , a vitreous phase is formed around this boride, protecting it from oxidation. In addition, the glass promotes the distribution of the neutron absorbing phase at the grain boundaries and gives the pellet a certain plasticity at high temperature, which is an advantage under the conditions of use in the reactor.

The glass used in document FR 2 814 584 containing boron is a borosilicate having a mass boron content of 4%.

The manufacturing process includes the steps of preparing a mixture of nuclear fuel powder, metal boride and boron-containing glass, shaping the powder into raw pellets, sintering the pellets at a temperature between 1200 ⁰ C and 1300 ⁰ C and from reduction under hydrogen of the pellets at a temperature between 800 ^° C. and 1000 ^° C.

However, an improvement in the homogeneity of the distribution of boron in the pellets is sought.

In addition, the process described above is long, and the reduction step can cause mechanical weakening of the pellets.

Finally, the cost price of the process is high, in particular, due to the steps of heating to very high temperatures.

It is therefore an object of the present invention to offer a nuclear fuel comprising a consumable poison, having a good homogeneity of distribution of boron and whose boron maintenance rate is high, and meets the specifications for fuels nuclear for PWR.

According to the present invention, the maintenance rate is understood to mean the ratio between the amount of boron in the fuel when it is ready for use and the amount of boron introduced into the mixture at the start of fuel fabrication.

It is also an object of the present invention to provide a simplified manufacturing process for a nuclear fuel comprising a consumable poison and more economically profitable

STATEMENT OF THE INVENTION

The inventors unexpectedly discovered that a nuclear fuel comprising, among other things, a glass with 7% of boron associated with diboride of titanium, made it possible to obtain a very good distribution of boron throughout the fuel, while having a boron content very close to that of the raw mixture.

Surprisingly, compared to the state of the art, a very small increase in the amount of boron in the glass introduced into the fuel, associated with titanium diboride makes it possible to homogenize significantly the boron in the fuel.

The previously stated aims are then achieved by a nuclear fuel based on uranium oxide, for example uranium oxide or MOX, of a glass comprising 7% by mass of boron and titanium diboride, and by a process not comprising a reduction step under a hydrogen atmosphere.

In fact, the inventors have observed a marked reduction in the overstoichiometry in the sintered pellets produced from the fuel according to the invention, the ratios amount of oxygen to the amount of uranium measured being close to 2. Thus the reduction step , which aims to reduce this overstoichiometry, is no longer necessary. The elimination of this step has the advantage of reducing the fragility of the pellets.

In addition, found in the sintered pellets according to the present invention, almost 90% of the amount of boron introduced into the mixture. Thus the nuclear fuel according to the present invention additionally offers a high boron retention rate. limiting to a minimum the oxidation of the boride into B ₂ O ₃ liable to vaporize.

The present invention therefore mainly relates to a nuclear fuel based on uranium oxide, comprising a consumable poison formed by boron, the boron introduced into the fuel being contained in titanium diboride and in glass, the glass comprising 7% by mass of boron.

The fuel according to the invention preferably has a mass content of boron of

500 ppm.

The glass advantageously has a glass transition temperature approximately equal to 500 ^° C.

The glass can comprise from 20 to 25% by mass of B ₂ O ₃ and from 65 to 70% by mass of silicon oxide SiO ₂ .

The nuclear fuel can be uranium oxide or a mixed uranium-plutonium oxide UO ₂ -PuO ₂ .

The present invention also relates to a method for manufacturing nuclear fuel pellets according to the present invention, comprising the steps

a) mixture of nuclear fuel powders, titanium diboride and glass comprising

7% by mass of boron and of homogenization of the mixture obtained by grinding,

b) shaping the mixture in the form of raw pellets by compression, c) sintering the raw pellets under a weakly oxidizing gas at a temperature of 1200 to 1300 ^° C., then optional rectification, the pellets then being ready to be used.

The method according to the invention may also include a step of granulating the mixture obtained at the end of step a), during which wafers with a diameter greater than that of the desired pellets, are produced by pressing under a pressure of 1 'order of 80 MPa, then crushed and forced through a sieve.

Sintering step c) may include a first temperature rise phase at a speed of 2.5 ° / min, a second stage phase of 4 hours at a temperature between 1200 ^° C. and 1300 ^° C., and a third phase decrease in temperature by 5 ° / Min until reaching room temperature.

BRIEF DESCRIPTION OF THE DRAWINGS

The present invention will be better understood with the aid of the description which follows and of the appended drawings, in which:

FIG. 1 is a table grouping together the compositions of the different samples studied,

FIG. 2 represents micrographs of a fuel according to the invention (BORE 3) and of two other fuels (BOREl and BORE 2),

- Figure 3 is a graphical representation of thermogravimetric measurements on the fuel according to the invention and on two others mixtures comprising 7% glass and a metal boride other than TiB2,

- Figure 4 is a representation of a thermal sintering cycle implemented in a method according to the present invention.

DETAILED PRESENTATION OF PARTICULAR EMBODIMENTS

The nuclear fuel according to the present invention is manufactured using a nuclear material based on uranium oxide, a glass comprising 7% by mass of boron and titanium diboride TiB ₂ .

Thereafter the glass used will be designated VOIB.

The nuclear fuel can be uranium oxide or a mixed uranium oxide-plutonium oxide compound, MOX type.

The increase in the mass percentage of boron in the glass makes it possible to increase the glass / boride ratio, for the same boron content introduced into the nuclear fuel. Indeed, the glass providing more boron, it is possible to reduce the amount of boride in the form of TiB2 introduced into the mixture. This is particularly interesting, since on the one hand the glass does not lose its boron in temperature, and on the other hand during sintering the glass softens and is distributed between the grains and coats the boride, protecting it from oxidation. Thus the increase in the glass / boride ratio makes it possible to effectively protect the boride and promotes the maintenance of the boron. The mass content of boron in nuclear fuel is between 100 ppm and 1000 ppm, and advantageously equal to 500 ppm.

VOIB glass used as a mixture has thermomechanical characteristics compatible with the process for manufacturing fuel pellets, and with use in a nuclear reactor.

The glass used in mixture with nuclear fuel based on uranium oxide has a resistance to relatively oxidizing atmospheres. In addition, it is fluid enough to flow within the grains and to coat the boride, however avoiding causing a disintegration of the pellet in the manufacturing temperature range of use of the pellets, in particular between 800 ^° C. and 1200 ⁰ C.

VOIB glass has the characteristics reproduced in the first line of Table I below, in the second line are indicated the characteristics of the Duran ® glass used in the prior art.

Table I

VOIB glass in particular has a mass percentage of B ₂ O ₃ of 13% and a mass percentage of silicon oxide SiO ₂ of 81%.

The glass chosen has a sufficiently low viscosity between 500 ^° C. and 1000 ^° C., typically 10 ¹³ ' ³ poises at 500 ^° C., to be sufficiently liquid between 500 ° C and 1000 ⁰ C to coat the boride in solid form or boric anhydride before its vaporization, and to distribute itself between the grains so as to have a homogeneous distribution.

In Table II below are reproduced the features glass temperatures võib according to the present invention and, for comparison, those of glass Duran ^® used.

Table II

By way of comparison and in order to show the advantage of nuclear fuel according to the present invention, the description will also set out, in addition to measurements relating to the fuel according to the present invention designated by BORE 3, measurements of the characteristics of other BORE mixtures 1 and BORE 2 comprising borosilicate glass and a uranium oxide control not comprising glass.

The compositions of the BORE 1, BORE 2 and BORE 3 mixtures are grouped in the table in FIG. 1.

For this, the different measures will allow:

To evaluate the degree of maintenance of boron in the various fuels manufactured and its reproducibility, and • to assess the homogeneity of the distribution of boron in the various tablets produced, and from one tablet to another.

We will first describe the means and methods of analysis used to perform the characterization.

The mass compositions in the various elements used before applying the process for manufacturing the pellets were measured.

The characteristics of the pellets at the end of the manufacturing process to comply with the specifications, i.e. the total mass loss during the process, the ratio of oxygen atoms to the ratio of uranium atoms 0 / U, the dimensional characteristics of the pellets (diameter and height) and the density, were also measured.

The mass of each pellet (raw then sintered) is measured for example using a Mettler balance of type AT250 (precision 0.1 mg) and given, to the nearest 0.01 g.

The diameter of each pellet is measured at three points, top (Dl), middle (D2), and bottom (D3) using a Zygo 1202B laser micrometer (precision 1 μm), the diameters being given at 0, 01 mm close. The average diameter is determined using the following formula:

The height h of each tablet is measured using the SONY μ-mate device (precision 1 μm). The height of each column of ten tablets is given to the nearest 0.1 mm. The geometric density is calculated with an accuracy of ICT ² g / cm ³ .

A certain number of pellets are analyzed for exhaustive determination of the banal and capturing impurities of the UO ₂ used and for the development of the dilution process by mixing-grinding in ethanol medium.

For each production, the boron content is determined on two tablets. A whole tablet is used to determine the average boron content. Four to six sections of the 2 ^nd pellet are analyzed separately in order to assess the homogeneity of the boron in the pellet. The determination of boron is supplemented by that of the counter-elements associated with it in the additives, ie Si, Al, Na, K and Ti.

The analysis takes place in three stages. First, the pellet or the section of sintered pellet is dissolved in an HNO ₃ / HC1 mixture, a mixture of sulfuric acid and hydrofluoric acid can also be used. Then, a liquid-liquid extraction of uranium is carried out using a mixture TBP (tributylphosphate of formula (C ₄ H ₉ O) ₃ PO) + dodecane (Ci ₂ H ₂ 6) • An analysis of the different phases (except that containing uranium) is carried out by ICP-AES (Atomic Emission Spectrometry-Inductive Coupled Plasma Source) which allows the determination of elements at high contents as well as in the form of traces. The principle of identification and quantification is based on the spectroscopic analysis of the lines emitted by the excited atoms in a vapor of the sample. Table III below summarizes the theoretical mass and volume compositions actually used for the various mixtures to reach the target boron content of 500 ppm / U.

Table III

BORE 1 BORE 2 BORE 3

Mass loss / After co-grinding 0.7 o, 7 o, 4 initial mass

(%) After granulation 2.8 1, 6 o, 7

The mass losses due to the various operations (sieving / forcing after co-grinding in ethanol and after granulation) were estimated. The table in FIG. 1 groups together the estimation of the mass losses for the different samples BORE 1, BORE 2 and BORE 3. It is assumed that these losses do not modify the contents used, in other words that there is no no selective loss of any of the additives.

The O / U ratios of the different samples were determined by thermogravimetry after sintering. Table IV below summarizes the values obtained (ND = Not determined). For the determination of this ratio, we assume that the mass gain of the pellets during thermogravimetry comes only from the oxidation of UO _{2 + x} to U ₃ O ₈ : the different boron additives are not assumed oxidize. Table IV

After the sintering operation, the inventors observed, surprisingly, a marked reduction in the over-echiometry with ratios 0 / U close to 2. For example for BORE 3, the ratio 0 / U is equal to 2.025 ± 0.001, we are therefore very close to stoichiometry.

Up to now, it was understood that a reduction step after sintering had to be carried out to get closer to stoichiometry.

However, the results show that, contrary to what is currently practiced, this reduction step is not necessary for the fuel according to the present invention.

The process according to the present invention therefore requires less energy and less time.

The density of UO _{2 + x} increases with over-echiometry and can be calculated, for 0 <x <0.25, using the following relation:

p (UO _{2 + X} ) = 10.957 + 1.4829.x (experimental formula determined from multiple experimental measurements carried out in the laboratory)

Table V below makes it possible to compare the superstoichiometry of the fuel according to the present invention with that of a raw control, of a sintered control. It appears that the fuel according to the present invention has a very low superstoichiometry corresponding to a 0 / U 2.025 ratio (the 0 / U ratio to stoichiometry being 2), and therefore a reduction step to approach the stoichiometry appears to be not being necessary.

Table V

The theoretical density (p _C aic in the table in FIG. 6) of the mixtures is calculated with a mixing law taking into account the quantities actually used and the over-echiometry of the UO ₂ .

• Raw tablets: UO ₂ , 15 + glass (table 2) + boride (table 3) + zinc stearate

(p = 1.13 g. cm ^"3 ).

• Sintered pellets: U0 _{2 + x} (x = 0.018; 0.025 or 0.030) + glass + boride.

In Table VI below, the measurements of the characteristics of the nuclear fuel according to the present invention are grouped together with those of the fuels by way of comparison BORE 1 and BORE 2. The measurements were carried out on the raw pellets and the sintered pellets . Table VI

The measurements of the contents of B, Si, Al, K, Na and Ti in the BORE 3 pellets are grouped in Table VII below.

Table VII

Elements B (ppm / U) Si (ppm / U) Al (ppm / U) Ti (ppm / U)

BORE 3 manufacturing

Sintered tablet

101 ± 11 97 ± 11 723 ± 57 whole 449 ± 50

Section 1 399 ± 40 34 ± 9 102 ± 11 608 ± 57

Section 2,461 ± 50,214 ± 23,124 ± 23,813 ± 57

Section 3 505 ± 40 229 ± 23 122 ± 17 820 ± 57

Section 4 446 ± 50 192 ± 23 114 ± 12 758 ± 57

Average sections

453 167 116 750 frits

Raw tablet 506 705 73 764 For the BORE 3 sample which corresponds to the fuel according to the present invention, the boron content after sintering is relatively close to that before sintering (raw pellet).

The boron retention rate is very high since it is around 90% on average (ratio between the content in the whole sintered tablet and the boron content in the raw tablet).

On reading Table VII, it can be seen that the boron content measured on different sections varies between 399 ± 40 and 505 ± 40, the boron is therefore distributed relatively homogeneously throughout the tablet.

In addition, it appears that the nuclear fuel according to the present invention (BORE 3) offers a very good boron maintenance rate, since it is of the order of 90%.

For comparison, tables VIII.1 and VIII.2 below group together the measurements on BORE 1 and BORE 2 materials.

Table VIII.1

Elements B (ppm / U) Si (ppm / U) Al (ppm / U) Ti (ppm / U)

BORE 1 manufacturing

Sintered tablet

200 ± 30 <10 76 ± 10 50 ± 10 whole

Section 1 340 ± 40 130 ± 15 170 ± 20 500 ± 50

Section 2 280 ± 30 40 ± 5 100 ± 15 400 ± 40

Section 3,250 ± 30 36 ± 5 97 ± 15,360 ± 40

Section 4 240 ± 50 56 ± 10 80 ± 20 380 ± 80

Section 5 200 ± 20 39 ± 5 30 ± 10 210 ± 40

Average sections

262 60 95 370 frits

Raw tablet 508 885 100 920 Table VIII.2

Elements B (ppm / U) I Si (ppm / U) | A1 (ppm / U) I Ti (ppm / U)

BORE 2 manufacturing

Sintered tablet

106 ± 15 20 ± 5 23 ± 2 <2 whole

Section 1,400 ± 50 <10 71 ± 15 4 ± 1

Section 2 195 ± 40 <15 63 ± 20 3 ± 1

Section 3 332 ± 50 <15 76 ± 15 5 ± 2

Section 4 242 ± 50 <15 69 ± 15 4 ± 1

Section 5 416 ± 50 <15 69 ± 15 4 ± 1

Section 6 440 ± 40 <10 60 ± 20 <20

Average sections

338 <15 68 <20 frits

Raw tablet 498 1058 100 traces

It is observed that, unlike the fuel according to the present invention, the distribution of boron in the pellet is very variable. Indeed, there is a great dispersion between the contents measured in the different sections. For BORE 1, the boron content can vary between 200 ± 20 and 340 ± 40 between two sections; for BORE 2, the boron content varies from 195 ± 40 to 440 ± 40.

In addition, the dosages indicate a very low maintenance of boron between 20% and 80% approximately.

In FIG. 2, one can see a photograph obtained in optical microscopy of BORE 3 and by way of comparison of the photographs of BORE 1 and BORE 2.

The sintered samples are cut transversely using a diamond wire saw. To limit the heating of the sample, the wire is constantly cooled under running water. After having coated them in an epoxy resin, the pellets are first polished under water using a Triefus polisher and polishing cloths with fixed abrasive particles (SiC) of decreasing particle size (400, 800 and 1200 corresponding particles with an average diameter of 35, 21.8 and 15.3 μm respectively). The last step is a diamond paste polishing to obtain a mirror polish. The samples are washed with ultrasound between each step of changing the polishing cloth.

To reveal the microstructure of the sample, a chemical attack is carried out at room temperature with a solution containing 20 volumes of H ₂ O, 2 volumes of H ₂ O ₂ and 1 volume of H ₂ SO ₄ . The attack time varies according to the sample (from 3 to 10 minutes). Observations under the optical microscope at different magnifications are made on the samples (at the edge and in the center of the pellets) before and after attack. The observations at low magnifications of the non-attacked samples aim to assess the distribution of pores within the pellets. Those with high magnifications of the attacked samples make it possible to locate the porosity and to determine the size of the grains.

The BORE 3 pellet has fairly irregular grains and clearly less angular than those of a uranium oxide pellet: there is a tendency to spheroidization of the grains of the borated pellets. Compared to the control tablet, the grain size of the borated tablets is not very different: there is a slight increase in size for the addition of boride + glass mixture (BORE 3).

In general, the grain sizes on the edges of the pellets seem smaller than in the center because of the different physical conditions which characterize these areas.

(higher density, changes in different temperatures, ...).

A relatively large intergranular porosity is observed, distributed homogeneously within the pellets.

The fuel according to the present invention therefore fulfills the criteria required for a nuclear fuel.

By way of comparison only, the pellets of BORE 1 and BORE 2 have a relatively homogeneous mixed porosity.

Table IX below regroups the porosity values, the grain size as well as the relative density values based on the metrology results for the fuel according to the invention BORE 3. For comparison, the values of the fuel characteristics indicator, BORE 1 and BORE 2 are also indicated.

Table IX

It is observed that the porosity measured by image analysis and that deduced from the geometric density (metrology) for the BORE 3 fuel are very close.

The significant residual porosity observed can be explained by the inevitable partial oxidation of the boride in contact with U0 _{2 + x} which generated a certain volume of gaseous boric anhydride.

In addition, for comparison, FIG. 3 represents the gravimetric curves A, B and C produced respectively on mixtures comprising VOIB glass and a metal boride, such as TiB ₂ , SiB ₆ and AlBi ₂ . The thermal cycle is that indicated D.

Table X groups together the mass ratios between the VOIB glass and the metal boride for the mixtures mentioned above.

Paintings

* mass of boron corrected by the purity mentioned by the supplier

A significant loss of mass is observed from the start of heating for the V0IB / AlBi ₂ mixture (curve C).

In addition, the VOIB / TiB ₂ mixture according to the present invention appears to be more efficient than the VOIB / SiB ₆ mixture. Table XI below groups together the mass loss values for the mixture according to the invention and the VOIB / AIB12 and VOIB / SiB ₆ mixtures.

Table XI

TiB ₂ / VOIB SiB ₆ / VOIB AlB ₁₂ / VO IB mixtures

Δm (%) - 0.73 - 0.90 - 1.55

It therefore appears that the present invention relates to a combination between the boron content of the glass and the choice of metallic boride, that is to say titanium diboride.

The tests were carried out with uranium oxide, but comparable results can be obtained by using MOX as nuclear material.

We will now describe the process for manufacturing nuclear fuel pellets according to the present invention.

This process comprises the stages: a) mixing powders of nuclear fuel, titanium diboride and glass comprising 7% by mass of boron and homogenization of the mixture obtained by grinding;

b) shaping the mixture in the form of raw pellets by compression;

c) sintering the raw pellets under weakly oxidizing gas at a temperature of 1200 to 1300 ^° C.

It is also possible to provide for a step of rectification of the previously sintered pellets in order to bring their dimensions (diameter, height) back into the specification ranges. At the end of step c), the tablets are ready to be used. Advantageously, the method may include a step of granulating the mixture obtained at the end of step (a).

When forming the pellet, the compression can be of the cold compression type.

Prior to step a), the proportions of boride powder and of glass powder used are adjusted so that the crystallized compound (boride) introduces the desired amount of boron, for example 500 ppm, and that the glassy compound either in sufficient quantity to coat the latter.

We will now describe in detail the different steps of the method according to the present invention.

During step a), the different products

(nuclear fuel, for example uranium oxide or

MOX, metal boride and glass) are weighed beforehand and then mixed in a grinding jar. The grinding jar is advantageously made of WC tungsten carbide topped with balls of tungsten carbide, for example twenty, and of diameter 10 mm. A teflon seal, specifically designed to minimize the spread of doped UO ₂ powder, can equip the jars. As tungsten carbide is a very hard material, it allows efficient co-grinding without significant contamination of the jar.

The mixture can be homogenized by co ^¬ grinding in ethanol medium using a planetary mill for 2 hours. The mixture is then placed in an air oven at a temperature of 50 ^° C. for approximately 12 hours. The powder obtained is homogeneous in crystallized and glassy additives. The mixture is then forced through a 200 μm mesh opening sieve.

It is also possible to provide for mixing in several stages, in particular:

a first step, firstly mixing the boride and glass powders, for example with the planetary grinder in the alcoholic liquid phase,

- a second step adding, to the mixture obtained at the end of the first step, nuclear fuel powder.

During the granulation step, wafers with a diameter of approximately twice as large as the final target diameter are produced in a uniaxial compression press by applying a relatively low pressure of the order of 80 MPa, and at temperature ambient. The wafers obtained are then crushed in a mortar and forced through a sieve with a mesh opening of 200 μm. Granulation improves the flowability of the powder mixture and the filling of the press die during pelletizing.

0.3% by mass of lubricant is added

(zinc stearate (Zn (Ci ₈ H ₃₅ O ₂ ) 2)) • The homogenization of the powder / lubricant mixtures is carried out in a planetary mixer, for example rotating at a speed of 20 rpm ^"1. Granulation allows advantageously improve the flowability of the powder mixture and the filling of the press matrix during the shaping of the pellets.

During step b), the mixture obtained is pressed in the form of raw pellets having the desired dimensions. The pressing is carried out for example using a uniaxial compression press similar to that used in the granulation step, this operation also being carried out at ambient temperature. Pressing is carried out at a pressure between 350 MPa and 500 MPa.

The manufacturing process is called DCN (Double Normal Cycle) with a pelletizing pressure Ppastiiiage higher than that of granulation P _gr anuiation- The pressing step is similar to that practiced during the manufacture of combustible pellets in the state of technique for a pressurized water nuclear reactor, that is to say by uniaxial pressing under a pressure of 400 to 600 MPa. Consequently, this will not be described further.

In step c), the raw pellets are sintered under conditions favoring the maintenance of boron in the ceramic and avoiding or minimizing its oxidation to volatile boric anhydride.

Sintering is then carried out at a relatively low temperature, for example between 1250 ^° C. and 1300 ^° C., under an inert atmosphere, for example of argon in order to allow the flow of glass within the grains, which makes it possible to homogenize the distribution of boron on the one hand and the protection of the boride by coating it on the other hand, without any thermochemical degradation. The choice of a UO ₂ powder of suitable stoichiometry allows densification of the ceramic before 1000 ^° C., that is to say that the open porosity closes. Boron possibly oxidized under B ₂ O _{3 form} is then maintained in the closed pores of the pellet.

Thus low temperature sintering reduces the formation and volatilization out of the gaseous B ₂ O ₃ pellet.

An example of a thermal sintering cycle is shown in FIG. 4. It includes a phase E of temperature rise from ambient temperature to the bearing temperature, advantageously carried out with a temperature rise rate of 2.5 ° C. / min, a phase F of plateau for approximately 4 hours at a temperature of 125O ⁰ C, and a phase G of return to ambient temperature carried out with a temperature rise rate of 5 ° C / min.

When the dimensional specifications of the pellets are not met at the end of sintering, a step of rectifying the pellets is provided.

The boron retention rate in the fuel (around 90%) is significantly higher than that of the fuel manufactured with Duran® glass, with excellent homogeneity of distribution in the fuel, which is also much higher than that of the fuels of the art. prior. In addition, the pellets made with VOIB glass have good reproducibility.

The manufacturing process has been optimized.

A heat treatment step has been removed

(post-sintering reduction) and the fuel obtained does not exhibit any particular brittleness compared to a fuel not doped with boron.

Claims

1. Nuclear fuel based on uranium oxide, comprising a consumable poison formed by boron, the boron being dispersed in the fuel in the form of titanium diboride and glass, the glass comprising 7% by mass of boron.

2. Fuel according to claim 1, in which the mass boron content is 500 ppm.

3. Nuclear fuel according to claim 1 or 2, in which the glass has a glass transition temperature approximately equal to 500 ^° C.

4. Nuclear fuel according to one of claims 1 to 3, in which the glass comprises from 20 to 25% by mass of B ₂ O ₃ and from 65 to 70% by mass of silicon oxide SiO ₂ .

5. Fuel according to any one of the preceding claims, in which the fuel is uranium oxide.

6. Fuel according to any one of claims 1 to 5, wherein the fuel is a mixed uranium and plutonium oxide UO ₂ -PuO ₂ .

7. Method for manufacturing nuclear fuel pellets according to any one of the preceding claims, comprising the steps:

a) mixture of nuclear fuel powders, titanium diboride and glass comprising

7% by mass of boron and of homogenization of the mixture obtained by grinding,

b) shaping the mixture in the form of raw pellets by compression,

c) sintering the raw pellets under a weakly oxidizing gas at a temperature of 1200 to 1300 ^° C., then optional rectification, the pellets then being ready to be used.

8. The manufacturing method according to the preceding claim, comprising a step of granulating the mixture obtained at the end of step a), during which wafers of diameter greater than that of the desired pellets, are produced by pressing under pressure about 80 MPa, then crushed and forced through a sieve.

9. The manufacturing method according to claim 7 or 8, wherein step c) of sintering comprises a first phase of temperature rise at a speed of 2.5 ° / min, a second stage of 4 hours at a temperature between 1200 ⁰ C and 1300 ⁰ C, and a third phase of temperature reduction of 5 ° / Min until reaching an ambient temperature.