Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
  • G21C3/42—Selection of substances for use as reactor fuel
  • G21C3/58—Solid reactor fuel Pellets made of fissile material
  • G21C3/62—Ceramic fuel
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C21/00—Apparatus or processes specially adapted to the manufacture of reactors or parts thereof
• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
  • G21C3/42—Selection of substances for use as reactor fuel
  • G21C3/58—Solid reactor fuel Pellets made of fissile material
  • G21C3/62—Ceramic fuel
  • G21C3/623—Oxide fuels
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
  • Y02E30/00—Energy generation of nuclear origin
  • Y02E30/30—Nuclear fission reactors

Definitions

• the invention relates to a method for preparing a porous nuclear fuel comprising uranium, possibly plutonium and possibly at least one minor actinide implementing steps that do not involve pulverulent compounds of these elements.
• this process can be applied in the recycling of minor actinides via the incorporation of these minor actinides into the abovementioned fuel, which is intended to be used to constitute nuclear reactor nuclear rods or to enter the formation of transmutation targets, in order to carry out nuclear transmutation experiments in particular to better understand the transmutation mechanism of these minor actinide elements.
• This process can find, more broadly, simply application in the manufacture of porous fuels comprising uranium.
• minor actinide means actinide elements other than uranium, plutonium and thorium, formed in reactors by successive neutron captures by standard fuel nuclei, minor actinides being americium, curium and neptunium.
• pressurized water reactors operating with uranium-based fuels generate fission products, some of which are in the form of gases, as well as heavy elements: minor actinides.
• the latter formed by successive neutron captures of the fuel nuclei, are mainly isotopes of neptunium, americium and curium. They are at the origin of a strong emission a and a release of helium gas in large quantity.
• the new uranium-based fuels intrinsically have, because of these phenomena occurring during use or during storage for the fuels incorporating, as soon as they are manufactured, a quantity not negligible minor actinides, a stable level of porosity under irradiation, which allows the evacuation of these fission gases and decay helium without physical degradation of the fuel.
• minor actinides are separated during the treatment of spent fuel, uranium and plutonium, and are then incorporated at a higher level in fuel elements that are distinct from the elements.
• Standard fissile fuel reactor Fuel elements comprising minor actinides may consist, for example, of cover elements disposed at the periphery of the core of a reactor. This recycling pathway makes it possible in particular to avoid degrading the characteristics of the reactor core with non-standard fuels incorporating minor actinides by concentrating the recycling problems generated by these actinides on a reduced flow of material.
• minor actinides are mixed at a low level and are distributed almost uniformly throughout the standard reactor fuel elements. To do this, during spent fuel treatment, uranium, plutonium and minor actinides are treated together to form oxides, which are then used in the manufacture of said fuels.
• the recommended porosity rate for such fuels must be of the order of 14 to 16%, just as the porosity must be an open porosity, so as to facilitate the release of the helium produced and to avoid the phenomena of fuel swelling subsequent to the self-irradiation induced by the production of minor actinides.
• the present inventors set themselves the goal of proposing a process for preparing a porous fuel comprising uranium, which does not have the drawbacks inherent in the use of organic porogenic agents, namely the degradation of these agents from the stage of mixing fuel precursors, this innovative process through the use of inorganic pore-forming agents, which allow, among other things, a control of the porosity, both in quantitative terms and in qualitative terms (especially in terms of pore sizes and pore characteristics).
• the invention relates to a method of manufacturing a porous fuel comprising uranium, possibly plutonium, and optionally at least one minor actinide comprising successively the following steps:
• a) a step of compacting a mixture comprising a first type of agglomerates comprising uranium oxide in the form of uranium dioxide UO 2 , optionally plutonium oxide, and optionally at least one oxide of a minor actinide, and a second type of agglomerates comprising uranium oxide in the form of triuranium octoxide U 3 O 8 , optionally plutonium oxide, and optionally at least one oxide of an actinide minor;
• the second type of agglomerates mentioned above may be prepared prior to compaction step a) by a succession of specific operations which will be explained in more detail below.
• first type of agglomerates or “agglomerates of the first type”
• second type of agglomerates or “agglomerates of the second type” will be used interchangeably.
• the compaction step is performed on a mixture comprising a first type of agglomerate comprising oxide uranium in the form of uranium dioxide U0 2 , optionally plutonium oxide, and optionally at least one oxide of a minor actinide, and a second type of agglomerates comprising uranium oxide in the form of U 3 O 8 triuranium octoxide, optionally plutonium oxide, and optionally at least one oxide of a minor actinide.
• the oxide of a minor actinide may be americium oxide, such as AmO 2 , Am 2 O 3 , or curium oxide, such as that Cm0 2 , Cm 2 0 3 , neptunium oxide, such as Np0 2 and mixtures thereof.
• the plutonium oxide may be in the form of Pu0 2 and / or Pu 2 0 3 .
• the agglomerates of the first type and the agglomerates of the second type advantageously have a spherical shape, this form being particularly suitable in the context of the invention, since it allows an easy filling of the molds and a distribution in these molds, in which can take place the compacting step.
• these agglomerates can be called spherules.
• the agglomerates of the second type may, in particular, be in the form of spheres having a mean diameter greater than 50 ⁇ m, preferably ranging from 100 to 1200 ⁇ m.
• the compacting step may be carried out by means of a press, which will apply a pressure to the mixture of agglomerates placed in a mold, the shape corresponds to the shape that one wishes to assign to the porous fuel, this form being conventionally that of a pellet.
• the pressure applied is adjusted according to the desired microstructure and the dimensions of the agglomerates.
• the compacting step may consist in applying to the mixture of agglomerates a pressure ranging from 100 to 1200 MPa, preferably from 300 to 600 MPa.
• the process of the invention may comprise a step of preparation of the agglomerates of the first type and / or agglomerates of the second type and, in particular, a step of preparation of the agglomerates of the second type.
• a cation exchange resin comprising carboxylic groups, this resin being constituted by resin-exchange resin beads; cations comprising carboxylic groups, whereby uranium in uranyl form and optionally plutonium and / or at least one minor actinide in cation form remain attached to the resin;
• the inventors have been able to note, surprisingly, that the resulting agglomerates have a conserved spherical shape relative to the initial cation exchange resin beads, despite a significant shrinkage. in size by a factor of about 1.5.
• This property is particularly interesting in the context of the invention, because it allows to control later the porosity of the fuel prepared in accordance with the method of the invention.
• the first step is to prepare a charge solution to be passed over a cation exchange resin comprising carboxylic groups.
• This feed solution when it contains only uranium in the form of a hydroxylated uranyl complex, can be prepared by introducing a predetermined quantity of oxide uranium UO 3 or optionally U 3 O 8 , in a nitric acid solution, said quantity being fixed so as to form a hydroxylated uranyl nitrate complex of formula UO 2 (NO 3 ) 2 -x (OH) x with x l 1, for example a 25% hydrolysed uranyl nitrate complex of formula U0 2 (N0 3 ) 1.5 (OH) O , s.
• This filler solution when it further comprises plutonium and / or at least one minor actinide in the form of plutonium nitrate (for example, Pu (III)) and / or nitrate of at least one minor actinide can be prepared as follows:
• uranium oxide UO 3 or optionally U 3 O 8 into said first solution, said quantity being fixed so as to form a hydroxylated uranyl nitrate complex of formula UO 2 ( NO 3 ) 2 -x (OH) x with x l 1, for example a 25% hydrolysed uranyl nitrate complex of formula UO 2 (NO 3) 1.5 (OH) 0.5;
• a step of mixing the resulting solution preferably at room temperature, optionally followed by a filtration step.
• the feed solution may be prepared by introducing into a first solution comprising nitrate of said actinide element and / or plutonium and already uranyl nitrate or nitric acid, a predetermined quantity of trioxide of uranium so to get the amount of desired uranium and a hydroxylated uranyl nitrate complex of formula UO 2 (NO 3 ) 2- x (OH) x with x l 1.
• uranyl cation be found in the form of hydroxylated uranyl nitrate complex, since it has been demonstrated that the presence of this complex is the driving force for the exchange between the resin and the cations present in the solution of charge.
• the presence of this complex in the charge solution makes it possible in particular to cause the concomitant ion exchange of uranyl cations and actinide and / or plutonium cations with the protons of the cation exchange resin, during the passage of the charge solution over the -this.
• This predetermined quantity of uranium trioxide to be introduced into the first solution is set so that the molar ratio between the number of moles of nitrate ions and the number of moles of uranium is less than 2.
• the resins used are conventionally in the form of polymer beads incorporating exchangeable groups, in our case carboxylates bearing H + protons.
• the resins used in the context of the invention may be resins resulting from the (co) polymerization of (meth) acrylic acid or acrylonitrile with a crosslinking agent, especially divinylbenzene (DVB).
• the selected cation exchange resin may be subjected to one or more treatment stages before the charge solution is passed, among which may be mentioned:
• a calibration step wet, so as to isolate the desired particle size fraction, for example, a fraction ranging from 600 to 800 microns; at least one washing step by carrying out a basic and acidic treatment cycle with ammonia and nitric acid followed by a rinsing step with demineralized water;
• a shape-sorting step so as to eliminate the broken or nonspherical particles, this step being able to be performed on an inclined table.
• the washing step mentioned above is intended to clean the resin of any presence of synthetic residues.
• the attachment of an ammonium group by proton neutralization reaction of the carboxylic groups allows swelling of the resin favorable to better access of the pores to the wash water.
• the passage of nitric acid then makes it possible to replace the ammonium groups with H + protons to restore the carboxylic groups.
• the resin optionally treated if necessary, is then advantageously moistened generously and placed in a column to form a bed of resin particles for receiving the charge solution.
• the operation of passage of the charge solution on the resin conventionally consists in letting it flow, by percolation, through the bed and recovering at the outlet of the bed an eluate.
• the resin comprising carboxylic groups progressively exchanges its protons against the uranyl cations and the cations of the actinide element and / or plutonium.
• the pH of the eluate decreases sharply when starting exchange with the resin in proton form (i.e. comprising carboxylic groups -COOH). It then gradually goes back to find the pH value of the input charge, which means that the exchange is complete and the resin is saturated with metal cations. It is thus possible to stop the passage of the charge solution on the resin. In other words, one proceeds, conventionally, passing over the resin of the filler solution until an eluate having a concentration identical to that of the filler solution.
• the eluate recovered during the process may be subjected to a recycling step, for example, by adjusting the acidity of this eluate by adding nitric acid, by dissolving optionally in the solution of the uranium oxide and by supplementing with a solution of nitrate of actinide and / or lanthanide if necessary, so as to constitute a new solution of charge, intended to be passed on the resin.
• the resin is subjected to a thermal treatment operation, in a medium comprising oxygen, whereby spherical-shaped agglomerates comprising uranium oxide in the form of triuranium octoxide U 3 are obtained.
• O 8 optionally plutonium oxide and / or an oxide of at least one minor actinide.
• This heat treatment operation is conventionally carried out at a temperature and time effective to obtain the formation of triuranium octaoxide U 3 O 8 , optionally plutonium oxide and / or oxide of at least one minor actinide.
• This effective temperature and duration can be easily determined by those skilled in the art by simple tests until the desired phases are obtained, these phases being detectable by simple analysis techniques, such as X-ray diffraction.
• this heat treatment operation can be carried out at a temperature ranging from 600 to 1400 ° C. for a duration ranging from 1 to 6 hours.
• the agglomerates of the first type can be prepared by reduction of agglomerates comprising uranium oxide in the form of Triuranium Octoxide U 3 O 8 optionally in combination with plutonium oxide and one or more oxides at least one minor actinide, said agglomerates may be prepared beforehand by the implementation of a succession of operations i), ii) and iii) as defined above.
• This reduction can consist in applying to said agglomerates a temperature and time effective to obtain agglomerates of the first type, namely agglomerates comprising uranium oxide in the form of uranium dioxide UO 2 , optionally plutonium oxide. and optionally at least one oxide of a minor actinide.
• This effective temperature and duration can be easily determined by those skilled in the art by simple tests until the desired phases are obtained, these phases being detectable by simple analysis techniques, such as X-ray diffraction.
• this reduction can be carried out at a temperature ranging from 600 to 1000 ° C. for a duration ranging from 1 to 12 hours.
• the method of the invention may further comprise a dry mixing step of said agglomerates of the first type and of the second type, this dry mixing step being carried out before the compaction step and after the possible step for preparing said agglomerates.
• This mixing step consists in bringing the agglomerates of the first type and the second type into contact in appropriate proportions according to the desired stoichiometry and aims to obtain, in particular, a homogeneous mixture, for example by means of a roller stirrer. , a turbula mixer or an oscillating stirrer. This mixing step will be carried out with the necessary care, in order to avoid damage to the agglomerates and in particular to break them.
• the reduction step b) is carried out, which can be carried out by passing a stream comprising a reducing gas at a temperature ranging from 600 to 1000 ° C for a period ranging from 1 to 12 hours, this reduction step having the function of reducing all or part of the triuranium octoxide U 3 O 8 to uranium dioxide UO 2 by means of which there is concomitant formation of a porosity generated by the mesh size reduction between that of U3O8 and that of UO2.
• the reduction step b After the reduction step b), it may be implemented a sintering step, the purpose of which is to consolidate the fuel obtained at the end of the process, and in particular to densify it.
• the sintering step can be carried out by heating at a temperature ranging from 1000 to 1900 ° C for a period ranging from 1 to 12 hours.
• the above-mentioned reduction step and the sintering step can be carried out during a single thermal cycle, the reduction step taking place during the temperature rise up to 1000 ° C while the step of sintering takes place above 1000 ° C (from 1000 to 1900 ° C as mentioned above).
• FIG. 1 represents a photograph obtained by optical microscopy of U 3 O 8 spherules obtained according to example 1.
• FIG. 2 represents a photograph obtained by optical microscopy of UO 2 spherules obtained according to example 1.
• FIG. 3 represents a graph illustrating the thermal cycle applied during the reactive sintering step in the context of Example 1 and the comparative example.
• FIG. 4 represents a photograph obtained by light microscopy of the pellets obtained at the end of example 1.
• FIG. 5 represents a photograph obtained by light microscopy of the pellets obtained at the end of the comparative example.
• This example illustrates the preparation of a porous uranium oxide fuel comprising UO 2 according to the process of the invention.
• This preparation includes:
• an acid-deficient uranyl nitrate feed solution obtained by dissolving 35 g of uranium trioxide UO 3 in 1 L of 260 mM uranyl nitrate solution was prepared in the first stage.
• a partially hydrolysed uranyl solution corresponding to the UO 2 (NO 3 ) 1.3 (OH) O 7 formulation is obtained.
• the final concentration of uranium is 400 mM and the pH value is raises to 3.4, which constitutes sufficient conditions for a cationic exchange on a carboxylic resin.
• this previously prepared solution was passed, with a flow rate of 2 ml / min on a 1.8 cm 2 section column comprising a bed of carboxyl type cation exchange resin type IMAC HP 335 Dow Chemicals, particle size range 630-800 ⁇ m and equivalent to 40 g of dry resin in proton form.
• the resin thus dried is then subjected to a heat treatment consisting in calcining it in air at 800 ° C. for 4 hours with a temperature rise of 1 ° C./min, whereby the resulting product is in the form of spherules which, after analysis by X-ray diffraction, show the presence of a phase U 3 0 8 orthorhombic structure.
• spherules have an average particle diameter, measured by optical microscopy, of 425 ⁇ m.
• step c) The mixture resulting from step c) is subjected to compaction at 400 MPa using a 5 mm diameter tri-shell matrix with stearic acid lubrication of the matrix and the pistons.
• step e) Reduction and sintering steps
• the mixture thus pressed is subjected to a reduction step and a sintering step under 4% hydrogenated argon according to a thermal cycle illustrated in FIG. 3 attached.
• the reduction step takes place during the temperature rise up to 1000 ° C, while the sintering step as such takes place at 1750 ° C for 4 hours.
• the pellet whose geometric density reaches 83% of the theoretical density of UO 2, has a high level of open percolating porosity and homogeneously distributed.
• This example illustrates the preparation of a uranium oxide UO 2 fuel only from spherules U0 2.
• This preparation includes:
• an acid-deficient uranyl nitrate feed solution obtained by dissolving 35 g of uranium trioxide UO 3 in 1 L of 260 mM uranyl nitrate solution was prepared in the first stage.
• a partially hydrolysed uranyl solution corresponding to the UO2 (NO3) 1.3 (OH) 0.7 formulation is obtained.
• the final concentration of uranium is of 400 mM and the pH value is 3.4, which constitutes sufficient conditions for a cationic exchange on a protonated carboxylic resin.
• this previously prepared solution was passed at a flow rate of 2 ml / min on a 1.8 cm 2 section column comprising a bed of carboxyl type cation exchange resin of the IMAC HP 335 type of the company.
• Dow Chemicals, particle size range 630-800 ⁇ m and equivalent to 40 g of dry resin in proton form.
• the resin thus dried is then subjected to a first heat treatment of calcining in air at 800 ° C. for 4 hours with a temperature rise of 1 ° C./min, whereby the resulting product is in the form of spherules, which , after analysis by X-ray diffraction, display the presence of a U 3 O 8 phase of orthorhombic structure.
• spherules have a diameter Particle size, measured by optical microscopy, 425 ⁇ m.
• the geometric density of the raw pellet determined by weighing and measurement of the dimensions (diameter and height respectively measured using a profilometer and a comparator) is estimated at 56% of the theoretical density of the oxide of uranium UO 2 (which is 10.95 g / cm 3 according to JCPDS sheet 00-041-1422).
• Sintering step During this step, reactive sintering of the lozenge under argon is carried out hydrogenated at 1750 ° C for 4 hours, according to a thermal cycle identical to that of Example 1 (this thermal cycle being shown in Figure 3 attached).
• the pellet obtained has a porosity of nearly 7% (this porosity being determined geometrically).
• the pellet whose geometric density reaches 93% of the theoretical density of UO 2 has a low porosity rate.

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• 2011
  • 2011-08-22 FR FR1157443A patent/FR2979469A1/fr not_active Withdrawn
• 2012
  • 2012-08-21 KR KR1020147007697A patent/KR102084425B1/ko not_active Expired - Fee Related
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