EP2695167A1 - Procédé de décontamination de radionucléides présents dans des matériaux carbonés et/ou graphitiques irradiés avec des neutrons - Google Patents

Procédé de décontamination de radionucléides présents dans des matériaux carbonés et/ou graphitiques irradiés avec des neutrons

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Publication number
EP2695167A1
EP2695167A1 EP12722670.2A EP12722670A EP2695167A1 EP 2695167 A1 EP2695167 A1 EP 2695167A1 EP 12722670 A EP12722670 A EP 12722670A EP 2695167 A1 EP2695167 A1 EP 2695167A1
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EP
European Patent Office
Prior art keywords
graphite
carbon
binder
radionuclides
materials
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Withdrawn
Application number
EP12722670.2A
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German (de)
English (en)
Inventor
Dirk Vulpius
Werner Von Lensa
Rahim Nabbi
Eric Mauerhofer
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Forschungszentrum Juelich GmbH
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Forschungszentrum Juelich GmbH
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Publication of EP2695167A1 publication Critical patent/EP2695167A1/fr
Withdrawn legal-status Critical Current

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Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/28—Treating solids
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/28—Treating solids
    • G21F9/30—Processing

Definitions

  • the invention relates to methods for the decontamination of radionuclides from neutron irradiated carbon and / or graphite materials, which consist of a heterogeneous compound of crystalline filler materials and microcrystalline amorphous binder materials.
  • Radiotoxic or radionuclides which are formed by neutron activation of impurities contained in graphite and / or by neutron activation of the ambient atmosphere and / or by nuclear fission.
  • radiotoxic agents present a particular disposal problem when they are volatile, such as tritium ( 3 H), or persistent and volatile such as 36 C 1 , or emit particularly penetrating radiation, such as 60 Co.
  • the carbon contained in the graphite is activated with the atomic weight 13 itself to radiocarbon ( 1 C).
  • the capture cross-section for this reaction is only 0.0014 barn, but is not negligible because of the significant concentration of 1.07% of l3 C.
  • 13 C is generated by neutron capture of 12 C (0.0035 barn) during irradiation and is thus increasingly contributing to the formation of 14 C, too.
  • radiocarbon is produced by neutron activation of nitrogen at atomic weight 14.
  • I4 represents N with 99.64% majority proportion in naturally occurring nitrogen.
  • the cross section for the neutron capture with subsequent emission of a proton is 1, 93 barn.
  • radiocarbon is formed via the 0.038% naturally occurring oxygen isotope with atomic weight 17 via neutron capture (0.257 barn) with subsequent emission of an alpha particle.
  • nitrogen and oxygen may be either part of the reactor atmosphere or part of chemical bonds in reactor materials in the neutron field.
  • Radiocarbon differs only by its atomic weight from the stable isotopes of carbon. It therefore basically has the same chemical behavior as the other carbon isotopes. Furthermore, it is implemented in biological processes such as stable carbon and not recognized as a foreign substance. Its release into the biosphere is therefore to be avoided. For this reason, the limits for the disposal and potential release of radiocarbon are extremely stringent. For example, the KONRAD repository requires a l4 C release from the waste container of ⁇ 1% of the 14 C inventory if the activity limit per waste package is to be fully utilized.
  • the radiotoxic agents and in particular 1 C form only a few ppm (parts per million) of the total mass of graphite.
  • the radiotoxic agents according to the prior art knowledge, are distributed more or less homogeneously over their entire volume, so that the entire volume is considered to be radioactive waste, which in some countries can be termed intermediate-level waste (ILW) or the half-life of 5730 years as long-lived low-level waste (LLLW). Therefore, in view of scarce and expensive repository capacities, there is a need to selectively remove and concentrate such portion of the radiotoxins from the graphite so that the remaining graphite can be classified as either low-level waste (LLW) or even cleared and reused.
  • ILW intermediate-level waste
  • LLLW long-lived low-level waste
  • the invention describes a method for the decontamination of neutron irradiated carbon and graphite materials, such as nuclear graphite and coal or even high purity graphite, the materials may be in the form of solid pieces, granules or powder.
  • the specific, selective decontamination of the releasable radionuclides is here understood to mean the extensive removal of problematic radionuclides for disposal, which can be released from the waste by leaching operations or outgassing in the short to long term.
  • any remaining in graphite residuals of radionuclides are not considered problematic if they are so firmly built into the crystalline structure that they are not removable or releasable without their complete destruction.
  • the partially decontaminated graphite obtained in this way can either be disposed of or fed to new applications.
  • the released radionuclides since they are produced in an enriched form, can be stored in a volume-saving manner or recycled or reused as valuable materials. This applies in particular to the radionuclides tritium and radiocarbon, for which there are numerous scientific and technical applications.
  • the present invention relates to a method for the decontamination of releasable radionuclides from neutron irradiated carbon and / or graphite materials, which consist of a heterogeneous compound of crystalline fillers.
  • materials and microcrystalline-amorphous binder materials comprising mechanical and / or chemical treatment methods, the mechanical treatment methods comprising the steps breaking, shaking and sieving and the chemical treatment methods comprise a supply of activation energy, which is characterized in that the mechanical treatment methods and / or the supply of the amount of activation energy is adjusted so that only the microcrystalline regions of the binder material and / or the entire porous body of amorphous carbon components of the carbon and / or graphite material are attacked and dissolved, and that there is a subsequent fractionation of reaction products, which were separated from the carbon and / or graphite materials by the mechanical and chemical treatment processes.
  • the method according to the main claim is based on the common idea, on the reaction of the neutron irradiated carbon and / or graphite material to be decontaminated
  • Nuclear graphite is a technical product made from a filler material and a binder material.
  • the filler material may, for. B. natural graphite or petroleum coke exist.
  • binder material z As coal tar or Phenoiharz use.
  • Filler material and binder material are mixed together and formed into a shaped part. This is fired under atmospheric nitrogen, the binder material pyrolyzing. After further impregnation and firing processes, the molded part is then graphitized at up to 3000 ° C., with partial formation of crystalline graphite structures in the pyroxylated binder material occurring in some cases. Through this process, filler material and binder material are combined to form a solid but still heterogeneous graphitic body.
  • the invention is based on the recognition that filler and binder materials do not combine with one another in such a way that a homogeneous solid body is formed. Rather, a product with partially different density and strength arises.
  • the filler material especially if it consists of crystalline natural graphite, has a higher density and strength than the binder material.
  • the inventors have recognized that the majority of the releasable radionuclides in neutronenbestrahltem nuclear graphite (z. B. 3 H, l4 C, 36 C1, 60 Co, 90 Sr 137 Cs, among others,) is in the amorphous and microcrystalline proportions of the former binder material.
  • This binder material is pyrolyzed and graphitized during graphite production and forms the basis for the pronounced pore system of nuclear graphite.
  • the former binder material consists of microcrystalline graphite components and amorphous carbon components.
  • the binder regions behave more or less like activated carbon and adsorb impurities (eg nitrogen) not only during production but also during irradiation in nuclear plants by adsorption from the air and / or the cooling gas. Farther these areas have a higher chemical reactivity [1].
  • recoil reactions in the nuclear activation process play a role in the final incorporation of the activation products into the neighboring structures.
  • the specific, selective decontamination of neutron irradiated carbon and graphite materials according to the invention now means to remove those releasable radionuclides which are located in such binding sites.
  • radionuclides are seen which, because of their genesis, are incorporated directly into the graphite lattice of the filler particles or otherwise have no access to the outer surface or inner pore system of the graphite body. Such radionuclides can not be mobilized without complete destruction of the graphite lattice (that is, the entire graphite body).
  • Radionuclides which are in loose chemical or physical bonds and are accessible from the outside can be removed from the nuclear graphite according to the invention by targeted chemical or electrochemical reactions with or without the assistance of mechanical influences (grinding, breaking, shock waves, ultrasound, etc.).
  • the type of chemical, mechanical or electrochemical reaction is to be chosen so that only the surface of the graphite body, that is the outer (macroscopic) and inner (microscopic) is attacked.
  • the method makes use of the fact that there is an accumulation of radionuclides on the said outer and inner surfaces.
  • 13 C converts to l4 C by a ⁇ , ⁇ reaction with a cross section of 0.0014 barn for thermal neutrons.
  • 14 N transforms in l4 C by an ⁇ , ⁇ reaction with a thermal cross section of 1.93 barn.
  • l4N concentrations of 300 ppm (0.03%) are found, while the natural l3C content of carbon is 1.07%. In spite of the about 3-fold lower concentration of 14 N to 13 C, under the given conditions the
  • Concentration of oxygen in nuclear graphite can be very different. For example, on graphite surfaces in contact with atmospheric oxygen, oxygen concentrations are found to be 3%. For coal rock, oxygen concentrations in the percentage range were also measured inside. For this general consideration, therefore, an average oxygen content in nuclear graphite of 1% is assumed. The proportion of 17 0 in natural oxygen, however, is very low at 0.038%. This results in a 13 C / I 7 0 ratio of about 2800, that is, the l 7 0 concentration is about 2800 times lower than the 13 C concentration. Thus, the probability that I4 C is formed from 17 0, is still about 15 times lower than from l3 C. This means that the l 4 C-
  • Carbon and Graphitwerksto ff is produced.
  • Other nitrogen contents may also be present in the binder material in the form of heterocyclic organic compounds (eg in coal tar).
  • nitrogen and carbon are extremely inert in elemental form, nitrogen still forms covalent bonds on graphite surfaces, as confirmed by quantum mechanical calculations [2, 3]. Nitrogen is therefore on Graphite surfaces as a chemisorbed film. This can also be detected by means of secondary ion mass spectroscopy on the basis of depth profiles, as can be seen in FIG.
  • the pore system is lined with chemisorbed nitrogen in nuclear graphite, then it is likely that radiocarbon therefrom will also be on the surface or in near-surface regions of the pore system; because despite the relatively high recoil energy of the 14 C atom from the 1 N nuclear reaction of a maximum of 41.4 keV, which breaks all chemical bonds, the l4 C atom does not move so far from its origin due to numerous collisions with lattice atoms in that it is inside a graphite crystal lite. In addition, there is a low probability that it will find a free lattice site there to be firmly bound into the graphite lattice.
  • soot is finely divided microcrystalline amorphous carbon having a specific surface area of 10 to 1000 m 2 / g
  • soot is an ideal adsorbent for gases (similar to activated carbon).
  • a surface covered with soot such as the highly porous carbon and graphite materials increases the adsorption capacity of these materials for gases and other impurities many times over. The occupation of these surfaces with soot should therefore be strictly avoided.
  • the method according to the invention discriminates between the releasable radionuclide and the remaining waste in several respects and thus that only the releasable radionuclides can be selectively separated from the remaining waste.
  • Decontamination always has the goal to concentrate the radionuclide as much as possible and as little harmless material as possible, eg. As the crystalline filler particles to deduct from the waste.
  • the radionuclide is bound with a lower binding energy in the microcrystalline to amorphous binder material than in the crystalline regions. This binding energy can be chemical or physical in nature.
  • the binder area acts more like activated carbon to absorb various contaminants.
  • the binding of the radionuclide in the binder to the bond in the crystalline regions may have been selectively altered, in particular weakened.
  • the activation energy can now for example be specifically chosen so that just these changed bonds are solved to the binder. Then, the radionuclide is selectively removed from the binder while essentially being spared in the crystalline regions.
  • the invention is based essentially on exploiting the differences between the pyrolyzed and (partially) graphitized binder regions and the crystalline filler aggregates, such as, for example:
  • Exemplary embodiments are described below. They are based on methods that can be used to separate the fractions with the releasable radionuclides early and, if necessary, to apply specific treatments. Also should pretreatment stages, such. As the separation of tritium or chlorine in order to avoid isotopic mixtures in post-treatment steps where these elements occur in higher concentrations (eg hydrogen in electrolysis, chlorine in the case of volatilization of immobile carbides).
  • the method according to the invention essentially comprises chemical or mechanical treatment methods or a combination of both treatment methods.
  • an activation energy is supplied to the neutron-irradiated carbon and graphite material, wherein the amount of the supplied activation energy is adjusted so that only the microcrystalline regions of the binder material and / or the entire porous body of the amorphous carbon components of the carbon and / or graphite material attacked and will be dissolved and thus the microcrystalline amounts of binder and / or bound to the entire porous body of the amorphous carbon contents of the carbon and / or graphite material only radionuclides g surface on the upper, released.
  • the amount of activation energy supplied should be adjusted so that preferably the microcrystalline regions of the binder material and / or the entire porous body of the amorphous carbon components of the carbon and / or graphite material react and the crystalline region does not react or only very much later.
  • a corrosion medium eg, oxygen
  • the carbon atoms of the graphite are attacked. These are more loosely bound in the binder than in the filler particles (because they are crystalline and larger).
  • radionuclides such. B. I4 C released.
  • the key to this is that naturally in the binder, which is preferably attacked, the radionuclides are present in increased concentration. This is also true because of the other surface to volume ratio.
  • the corrosion medium preferably attacks the microcrystalline constituents. Along with this, the mechanical stability is considerably reduced.
  • Radionuclides that are bound there, for example, by
  • a halogen or halogen ions are transferred from the corrosion medium into at least one gaseous reaction product.
  • the reaction of the corrosion medium with the radionuclide may be a redox reaction in which the corrosion medium is reduced and the radionuclide is oxidized.
  • the hydrogen, the hydrogen ions, the oxygen, the oxygen ions, the halogen and / or the halogen ions can be presented as such. They can be presented, for example, in low concentration in an inert gas stream. But they can also be formed when supplying the activation energy from the corrosion medium.
  • a peroxide, water / steam, an acid, a caustic, carbon dioxide, a complexing agent, a halogen compound, a hydrocarbon, a halohydrocarbon and / or other pyrolyzable and / or dehydratable reagent may be used as Corrosion medium can be selected.
  • the corrosion medium can be used in liquid form or in the form of solid pieces, as granules or powder.
  • the corrosion medium preferably contains at least one binary metal oxide, sulfate, nitrate, hydroxide, carbonate, hydride and / or halosuccinimide.
  • the hydrogen, the hydrogen ions, the oxygen, the oxygen ions, the halogen and / or the halogen ions can react in particular in the nascent state (see electrolysis) with the radionuclide. They are particularly reactive in this state.
  • Another form of supply of activation energy may be by the electrolysis of neutron-irradiated carbon and graphite materials.
  • the electrolysis uses various different properties of binder and filler areas, such. B. the different porosity between filler and binder area with regard to penetration and wetting by the electrolyte, the chemical reactivity of the binder components and the mechanical stability.
  • the conditions of the electrolysis are preferably chosen such that disintegration of the microcrystalline and amorphous binder regions occurs.
  • the electrolysis of neutron-irradiated carbon and graphite materials is a universal process by which in principle a wide range of radionuclides can be removed from carbon and graphite materials.
  • the electrolysis z. B. in dilute nitric acid, wherein the graphite to be decontaminated as anode (positive pole) is connected in a DC circuit.
  • graphite nitrate a known graphite intercalation compound which hydrolyzes immediately in the aqueous medium to form graphite oxide (graphite with varying proportions of hydroxyl groups).
  • graphite oxide graphite with varying proportions of hydroxyl groups.
  • This disintegration during the electrolysis can be mechanically z. B. support by ultrasound, so that the filler particles are early and undestroyed dissolved out of the graphite or Kohleste in alliance. It has been proven that the carbon oxides occurring in this process and especially the carbon dioxide have a particularly high proportion of radiocarbon,
  • the resulting carbon dioxide is enriched with radiocarbon and can be used, for example, by reaction with alkaline absorbents (eg alkalis), for example by conversion to barium carbonate as the radiocarbon source, or converted to other stable compounds (eg carbides) for disposal.
  • alkaline absorbents eg alkalis
  • barium carbonate as the radiocarbon source
  • other stable compounds eg carbides
  • tritium can also go into solution.
  • the electrolysis of irradiated nuclear graphite should be carried out as possible after separation of the tritium.
  • the electrolysis can be used with appropriate choice of voltage for the triggering of ion migrations.
  • the preferred electrolyte is sodium perchlorate, since it does not form any complexes that interfere with electromobility with the contaminants. This process can also be used to accelerate leaching.
  • the magnitude of the electrical voltage or of the electric field in the solid is dimensioned such that it is at least sufficient to overcome the activation energy for the diffusion (> 1 eV).
  • the result is the following procedure: The irradiated nuclear graphite voltage is applied in a certain amount, whereby a displacement of the radionuclide compounds enforced in the field direction and thus a spatial concentration of these compounds is achieved (for the purpose of chemical release / separation).
  • the result of the method consists in the positive influence on the migration or diffusion of radiocarbon compounds by an externally applied electric field of certain field strength.
  • Another embodiment is the surface electrolysis z.
  • B the spent fuel pellets of high temperature reactors on the ball-pile principle. It is well known that radiocarbon deposited preferentially on the spherical surface and in near-surface layers because, despite the high purity of the graphitic matrix material, the concentration of chemisor physisorbed nitrogen was highest there. Selective anodic electrolysis of the upper layers of the fuel element sphere releases the l4 C contained therein in concentrated form as l4 C0 2 . This can be collected with sodium hydroxide solution or another alkaline absorbent and processed further. The electrolysis can for example be adjusted so that only the surface is removed by 1 to 2 mm. Furthermore, by monitoring the 14 C concentration, for example via a scintillation detector in the online process, the electrolysis can then be stopped when the 14 C concentration decreases.
  • Mechanical treatment methods :
  • One embodiment of the mechanical treatment method relates to the mechanical comminution of neutron-irradiated carbon and graphite materials.
  • the binder areas are significantly weaker in their mechanical strength than the filler components.
  • the mechanical cohesion of the microcrystalline regions of the binder material and / or the entire porous body of the amorphous carbon fractions of the neutron-irradiated carbon and / or graphite material can be destroyed and the Fractionate ingredients so that filler and binder areas can be separated from one another using different crystal sizes / grain sizes
  • the mechanical crushing is carried out to the maximum grain size of the filler particles, which can vary depending on the graphite or carbon type.
  • the binder particles are smaller due to their lower breaking strength.
  • the finely crystalline portion of the binder reacts preferably with the grinding, breaking or sieving.
  • Subsequent sighting and post-treatment procedures e.g. B. with corrosive agents may be advantageous. This also leads to the preferential release of contaminants of the binder and to the surfaces of the filler crystallites. These reactions may, for. B. in a rotating reaction vessel, in a mixer or in a fluidized bed reactor.
  • a thermal pretreatment is advantageous in which certain contaminants / radionuclides are partially expelled by pyrolysis of hydrocarbons (eg tritium) or by slight corrosion.
  • An upstream thermal pretreatment or corrosion especially at temperatures between 350 to 900 ° C., in particular below 500 ° C. (chemical oxidation range, see FIG. 5), can increase the mechanical stability of the waste material. targeted weaken in the binder areas with minor mass losses or degrees of corrosion and support the subsequent mechanical comminution and fractionation.
  • a removal of the outer layers can also be achieved by other methods such.
  • mechanical trimming, brushing, laser ablation or - particularly advantageous - oxidation with oxygen accomplish.
  • An oxidation with oxygen can be carried out, for example, at temperatures of more than 350 to ⁇ 900 ° C or carried out above 900 ° C. Different areas can be attacked by the choice of the oxidation temperature:
  • Boundary layer diffusion area (/> 900 ° C) [5]
  • the oxygen can penetrate deep into the pores of the graphite and thereby oxidize the binder, which leads to a disintegration of the graphite and thus to an exposure of the filler particles (see FIG. 5).
  • the binder and volatile with oxygen radionuclides such. B. released radiocarbon.
  • the reaction rate is very low (see FIG. 6). This reaction is applicable to graphites in which the radionuclide to be removed is distributed throughout the graphite body and has to be concentrated by the process.
  • the radionuclide to be removed is concentrated in a certain area of the graphite body - such as.
  • the radiocarbon on the surface of the spent fuel pellets of high temperature reactors - it is advantageous in view of the effectiveness of the method to increase the oxidation temperature to get into the pore diffusion region or the boundary layer diffusion region.
  • a radio-carbon-enriched combustion gas because the Enrichment already exists in the material and does not have to be accomplished by the process.
  • the non-releasable fraction of radionuclides are very firmly bound in the graphite, their removal for disposal is not required because z. B. the release rates are below 1%.
  • the graphite can even be used to chemisorb radionuclides on its inner surfaces and thus fix them securely for long times. This may possibly be supported by measures to close the pore system.
  • a corrosive agent eg pure oxygen
  • a corrosive agent can also be adsorbed before heating in the interior, ie in the pore system, of the carbon and / or graphite materials. This is particularly the case when radionuclides such as tritium and radiocarbon are to be obtained in high enrichment in order to supply them to the recycling cycle.
  • both the proportions of previously chemisorbed nitrogen and thus of the 14 C which has also accumulated there as well as fractions of deposited corrosion media eg oxygen, C0 2 , H 2 , H 2 0
  • a heating of the carbon and / or graphite materials in an inert gas atmosphere (eg N 2 or Ar) or in a vacuum at a temperature range between 500 and 1500 ° C can be carried out, so that only a part of the radionuclides are removed from the carbon and / or graphite material, but this in a particularly high enrichment.
  • FIGS. 9 and 10 show this state of affairs on the basis of measurement series carried out. Shown is the proportion of released tritium or radiocarbon versus the released (corroded) portion of total carbon ( 12 C and 13 C).
  • the reaction media used were inert gas (N 2 or Ar) or water vapor at different temperatures. It can be seen that the enrichment ratio 3 H / total carbon or 14 C / total carbon in the case of heating tests in inert gas (N or argon) is greatest.
  • the shape of the sample is decisive: In massive graphite samples, the release of radionuclides is higher than in the case of milled ones, in which the pore system with the reactive gases contained therein is largely destroyed.
  • This method is made possible by the oxygen adsorbed especially in the binder material of the carbon and / or graphite material and / or other adsorbed corrosion media (eg water vapor) which preferentially react with the radionuclides present in concentrated form in reactive form and on the surfaces.
  • adsorbed corrosion media eg water vapor
  • oxidizing functional groups eg hydroxyl, carbonyl, carboxyl, ether or epoxy groups
  • the functional groups are split off from their binding in graphite at temperatures above 500 ° C and can react directly with radionuclides such as tritium and radiocarbon, which are finely distributed in the binder system, to form gaseous (easily releasable) reaction products.
  • radionuclides such as tritium and radiocarbon
  • the proportion of these functional groups can even be increased by the carbon and / or graphite material is converted by suitable methods before the thermal treatment in oxidized carbon and / or graphite material (graphite oxide).
  • suitable methods may be purely chemical (treatment with strong oxidizing media such as hydrogen peroxide) or electrochemical (anodic oxidation in dilute acids).
  • the released in these reactions in the form of gaseous compounds radionuclides such. B. 14 CO can then also be supplied to the recycling cycle due to their high concentration.
  • the thermal energy kB corresponding to the temperature T. Boltzmann constant) must correspond at least to this activation energy.
  • the activation energy can be reduced by a catalyst.
  • a temperature between 350 to 1500 ° C, preferably selected between 500 and 1300 ° C. In this temperature range, the reaction takes place depending on the nature of the orrosion medium with a practical reaction rate.
  • Concentration e.g., pure oxygen
  • Concentration may be added to it at the inner
  • the penetration depth of the corrosive medium can be adjusted via the suitable choice of the process temperature and the process duration.
  • This may possibly also be a precursor for a mechanical aftertreatment (eg in rotating drums), because this corrosion treatment at high temperatures causes a weakening of the binder or disintegration of the microcrystalline and amorphous binder regions, which coincides with the contamination profile.
  • the released radionuclides eg 14 C, metallic radionuclides
  • chemical and mechanical treatment methods can be combined.
  • first the carbon and / or graphite materials under "mild" conditions ie at temperatures below 700 ° C, for example, under an oxygen atmosphere over a period of, for example be heated for about a few hours, then an electrolysis is carried out.
  • the electrolysis can, for example, additionally be supported by treatment with ultrasound.
  • the remaining after the separation of the binder material crystalline graphite can be supplied to the recycling cycle and reused.
  • FIG. 2 Scanning electron micrograph in FIG.
  • Fig. 3 14 N depth profile (as CN ⁇ ion) of irradiated (a) and unirradiated (b) AVR reflector graphite (AVR - Häyer remplisreaktor GmbH, Jülich)
  • FIG. 4 l3 C depth profile of irradiated (a) and unirradiated (b) AVR reflector graphite (AVR - Häyer remplisreaktor GmbH, Jülich)
  • FIG. 5 Schematic representation of the reaction areas of graphite oxidation [5]
  • FIG. 6 Kinetic representation of the reaction areas of graphite oxidation [6]
  • FIG. 7 Scanning electron micrograph of untreated irradiated AVR fuel matrix (A3 starting material)
  • FIG. 8 Scanning electron micrograph of disintegrated irradiated AVR fuel element matrix (graphite oxide)
  • the electrolysis conditions were the following:
  • Table 1 shows radionuclide release and measurable residual activity in the electrolysis of a graphite sample of an irradiated AVR fuel.
  • the fallen graphite flakes were filtered from the electrolyte solution, washed with distilled water and dried for 9 hours in a drying oven at 80 ° C. The same procedure was followed with the graphite particles still adhering to the graphite sample, which were removed by gentle scratching with a spatula.
  • the resulting reaction product was identified as "graphite oxide", a non-stoichiometric graphite compound with varying proportions of oxygen-containing functional groups (mainly hydroxyl groups) containing 60% of the starting material of the graphite sample.
  • the nuclide-specific activity of the electrolytic solution, the graphite oxide and the graphite electrode residue was determined by gamma spectrometry (60Co, 137Cs, etc.) or liquid scintillation measurement (3H, 14C, 90Sr - after combustion of a sample liquor and subsequent radiochemical separation) ,
  • Radionuclide release was determined by determination of radioactivity (Bq / g) before and after electrolysis. It indicates by what percentage a reaction product is less active than the starting material. This statement clearly illustrates the decontamination success of the process.
  • the measured release of activity is not a release of radionuclides from a homogeneous solid, but a release of radionuclides from parts of a heterogeneous solid.
  • the releasable portion of the radionuclides of irradiated carbon and graphite materials is in the regions of the binder material. This is preferably attacked by the electrolysis and the radionuclides therein are selectively released. As solid reaction products remain those parts of the nuclear graphite, which are less affected by the electrolysis. These are the filler particles. However, these have from the outset less activatable components and thus a significantly lower radioactivity than the binder material.
  • the radionuclide release thus means a preferential release of radionuclides the binder material leaving the less contaminated constituents of the nuclear graphite.
  • the values for I4 C in Table 1 mean that z. B. only about 13% of the total activity of l4 C in the remaining filler particles in graphite electrode residue.
  • the graphite oxide consists essentially of the detached filler particles and has 21% of the mean starting activity at l4 C. This may mean that even minor residues of the binder adhere to the graphite oxide particles, because they were only exposed to electrolysis for a short time.
  • Table 1 Radionuclide release and residual activity in the electrolysis of a graphite sample of an irradiated AVR fuel assembly
  • the graphite oxide was then subjected to a thermal treatment of 6 hours at 900 ° C under N 2 . With elimination of water (pyrolysis of the hydroxyl groups), the graphite oxide changed back into graphite.
  • Figures 7 and 8 show scanning electron micrographs of untreated irradiated AVR fuel assembly matrix (A3 starting material) and disintegrated, ie treated by electrolysis, irradiated AVR fuel element matrix (graphite oxide from the filler particles).
  • A3 starting material untreated irradiated AVR fuel assembly matrix
  • disintegrated, ie treated by electrolysis, irradiated AVR fuel element matrix graphite oxide from the filler particles.
  • graphite oxide from the filler particles.
  • the binder material which in the present case constitutes only the thin boundary layer between the graphite particles, has been converted by the electrolysis process into gaseous reaction products (carbon dioxide).

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  • Solid-Sorbent Or Filter-Aiding Compositions (AREA)

Abstract

Procédé de décontamination de radionucléides présents dans des matériaux carbonés et/ou graphitiques irradiés avec des neutrons, qui sont constitués d'un composé hétérogène contenant des matériaux de charge cristallins et des matériaux de liant microcristallins-amorphes, ledit procédé comprenant des procédés de traitement mécaniques et/ou chimiques, les procédés de traitement mécaniques comprenant les étapes de broyage, secouage et criblage et les procédés de traitement chimiques comprenant un apport d'énergie d'activation. Ce procédé de décontamination est caractérisé en ce que l'on ajuste les procédés de traitement mécaniques et/ou l'apport de la quantité d'énergie d'activation de sorte que seules les zones microcristallines du matériau de liant et/ou la totalité du corps poreux des fractions de carbone amorphes du matériau carboné et/ou graphitique soient attaquées et décomposées, et en ce que l'on réalise ensuite un fractionnement des produits de réaction qui ont été séparés des matériaux carbonés et/ou graphitiques par les procédés de traitement mécaniques et/ou chimiques.
EP12722670.2A 2011-04-06 2012-03-30 Procédé de décontamination de radionucléides présents dans des matériaux carbonés et/ou graphitiques irradiés avec des neutrons Withdrawn EP2695167A1 (fr)

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DE201110016272 DE102011016272A1 (de) 2011-04-06 2011-04-06 Verfahren zur Dekontamination von Radionukliden aus neutronenbestrahlten Kohlenstoff- und/oder Graphitwerkstoffen
PCT/DE2012/000345 WO2012136191A1 (fr) 2011-04-06 2012-03-30 Procédé de décontamination de radionucléides présents dans des matériaux carbonés et/ou graphitiques irradiés avec des neutrons

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EP2695167A1 true EP2695167A1 (fr) 2014-02-12

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EP (1) EP2695167A1 (fr)
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FR2997543A1 (fr) * 2012-10-29 2014-05-02 Electricite De France Traitement thermique de dechets carbones, perfectionne par le choix des gaz injectes.
DE102013003847B3 (de) * 2013-03-07 2014-09-04 Forschungszentrum Jülich GmbH Fachbereich Patente Verfahren zur Dekontamination von Radionukliden aus neutronenbestrahlten Kohlenstoff- und/ oder Graphitwerkstoffen
US9793018B2 (en) * 2013-10-29 2017-10-17 Westinghouse Electric Company Llc Ambient temperature decontamination of nuclear power plant component surfaces containing radionuclides in a metal oxide
DE102014110168B3 (de) * 2014-07-18 2015-09-24 Ald Vacuum Technologies Gmbh Verfahren zur Dekontamination von kontaminiertem Graphit
GB202008809D0 (en) * 2020-06-10 2020-07-22 Univ Manchester Graphite decontamination
WO2024038283A1 (fr) * 2022-08-18 2024-02-22 Jacobs U.K. Limited Décontamination et régénération de graphite irradié
GB2621621A (en) * 2022-08-18 2024-02-21 Jacobs U K Ltd Decontamination and regeneration of irradiated graphite
CN116517526A (zh) * 2023-04-12 2023-08-01 安徽中核桐源科技有限公司 一种放射性同位素示踪剂及其制备方法

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DE102004036631B4 (de) * 2004-07-28 2013-02-21 Forschungszentrum Jülich GmbH Verfahren zur Behandlung einer mit Radiokarbon kontaminierten Keramik, insbesondere Reaktorgraphit
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DE102011016272A1 (de) 2012-10-11

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