EP2695167A1

EP2695167A1 - Method for decontaminating radionuclides from neutron-irradiated carbon and/or graphite materials

Info

Publication number: EP2695167A1
Application number: EP12722670.2A
Authority: EP
Inventors: Dirk Vulpius; Werner Von Lensa; Rahim Nabbi; Eric Mauerhofer
Original assignee: Forschungszentrum Juelich GmbH
Current assignee: Forschungszentrum Juelich GmbH
Priority date: 2011-04-06
Filing date: 2012-03-30
Publication date: 2014-02-12
Also published as: DE102011016272A1; WO2012136191A1

Abstract

The invention relates to a method for decontaminating radionuclides from neutron-irradiated carbon and/or graphite materials, which are composed of a heterogeneous combination of crystalline filler materials and microcrystalline-amorphous binder materials, comprising mechanical and/or chemical treatment methods, wherein the mechanical treatment methods comprise the steps of fracturing, shaking, and sifting and the chemical treatment methods comprise supplying activation energy, characterized in that the mechanical treatment methods and/or the supply of the amount of activation energy is set in such a way that only the microcrystalline regions of the binder material and/or the entire porous body of the amorphous carbon fractions of the carbon and/or graphite material are acted on and dissolved, and that reaction products that were separated from the carbon and/or graphite materials by the mechanical and chemical treatment methods are subsequently fractionated.

Description

Method for the decontamination of radionuclides from neutron irradiated

Carbon and / or graphite materials

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.

State of the art

Carbon and graphite materials, such. As nuclear graphite and coal, are contaminated when used in nuclear reactors with various 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. These 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.

In addition, 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. In addition, 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. In addition, 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. In addition, 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.

From DE 10 2004 036 631 Al a method for the selective removal of radiocarbon via a chemical reaction at higher temperatures is known. In this case, a gaseous corrosion medium is placed on the outer and inner surfaces of the graphite, wherein only a few percent of the graphite corrode, but a large proportion of the radio-carbon is released. The method exploits the knowledge that most of the radiocarbons are concentrated near internal surfaces in the pore system of graphite, because it was thought that the radiocarbon was essentially activated by activation of adsorbed nitrogen or oxygen in the pore system of graphite is formed and therefore preferably oxidized.

The prior art endeavors to remove irradiated nuclear graphite as large amounts of radionuclides with the lowest possible mass losses. There is no distinction between "releasable" and "non-releasable" levels of radionuclides. This is unfavorable in setting the decontamination target and evaluating the decontamination success of corresponding methods. In the final disposal or reuse of nuclear graphite, however, it is crucial to know the difference between "releasable" and "non-releasable" proportions of radionuclides, because only the releasable, that is loosely bound proportions of radionuclides are considered to be problematic for humans and the environment , This requirement is also reflected in the acceptance conditions for the KONRAD repository and constitutes an additional aim of the invention by aiming at fractionation and separation of the fraction in the waste, which have both low residual fractions of ¹⁴ C and release rates of less than 1%. This proportion could easily be stored under KONRAD conditions after further conditioning. The higher contaminated fraction can be further treated by methods described in the prior art and either decontaminated to the extent that it also corresponds to the conditions of acceptance or z. B. by conversion into carbides the criterion of lower release rates below 1% corresponds.

In addition, however, it has now been recognized that the production of carbon and graphite materials for nuclear applications operates under conditions which, due to the presence of nitrogen during the impregnation operations, the graphitization process and the subsequent long cooling phase, result in the formation of chemisorbed nitrogen compounds and thus especially favor high levels of releasable radiocarbon. Task and solution

It is therefore an object of the invention to provide a method which allows a specific, selective decontamination of the releasable radionuclides from neutron irradiated carbon and graphite materials.

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 (non-releasable 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.

This object is achieved by methods according to the main and secondary claim. Further advantageous embodiments will be apparent from the dependent claims.

Subject of the invention

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

1. on the use of specific chemical and physical properties of the pyrolyzed or graphitized additives (filler) in relation to the binder,

2. from the different local concentrations of activatable elements (eg nitrogen) in the aggregates and binder materials in the production of nuclear graphites and coal rocks and

3. to selectively influence the different chemical or physical bonds of radionuclides compared to natural isotopes in order to remove the highest possible proportion of releasable radionuclides from the waste matrices of the neutron-irradiated carbon and graphite materials.

Because of the particular difficulty in removing radiocarbon from waste matrices of neutron-irradiated carbon and graphite materials, the methods become particularly detailed with the example of decontamination of nuclear graphite (almost completely graphitized high-purity graphite) and of coal-stone (incompletely graphitized engineering graphite).

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. As 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. Thus, the filler material, especially if it consists of crystalline natural graphite, has a higher density and strength than the binder material. These material properties are also retained in the graphitized nuclear graphite: both with the naked eye and under the microscope can be seen the original filler material as zones of greater density and the original binder material as zones of lower density (see Figure 1).

Furthermore, one sees in the zones of the binder material a widely branched pore system, which is partly responsible for the difference in strength and density. This difference in density and the differences in the sizes of the crystals and their mechanical bridges can now be used to separate filler and binder material again by suitable methods, despite previous graphitization. This separation is advantageous since it has been recognized that radionuclides which have contaminated the nuclear graphite by neutron irradiation are located primarily in the zones of the original binder material, ie in the porous microcrystalline-amorphous structures, set (see Figure 2).

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. Chemically speaking, the former binder material consists of microcrystalline graphite components and amorphous carbon components. The surface of the microcrystalline graphite portions as well as the entire porous body of the amorphous carbon moieties with their reactive centers (corners and edges on the crystals, unbonded bonding sites on carbon atoms of the amorphous portion) form the preferred sites of attachment of radioactive and nonradioactive impurities in nuclear graphite, the loading These places with impurities already largely happens during the production of these materials by the existing atmosphere (mainly nitrogen) and by impurities in the binder material (eg coal tar). It should be noted that a large number of chemical bonds in the aggregates are decomposed or formed during the heating phase. But even during the cooling phase after the graphitization or pyrolysis process successively various chemical equilibria are passed, which are quasi 'frozen' or 'conserved' upon further cooling on the surfaces of the crystalline and amorphous structures. The more finely crystalline or amorphous these structures are, the more reactive they react in the attachment of non-carbonaceous elements (eg oxygen, nitrogen, hydrogen) or chemical compounds (eg hydroxyl groups). A particularly high proportion of amorphous structures in the binder areas occurs, for. For example, in the carbon-based matrix of ball fuel elements (so-called A3 matrix), in (non-graphitized) coal rock and highly irradiated graphites (by radiation-induced defects) on. Because of these structures, 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]. In addition, 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. In contrast, 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).

Those 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.

The described situation will be illustrated by the example of the formation and distribution of radiocarbon ( ^l4 C): For the formation of ^l4 C by thermal neutrons there are, as already explained, three main sources: the

Carbon isotope ^l3 C, the nitrogen isotope ¹ N and the oxygen ^{isotope l7} 0. First, the activation ^behavior of ¹⁴ N versus ^l3 C is considered:

¹³ C converts to ^l4 C by a η, γ reaction with a cross section of 0.0014 barn for thermal neutrons. ¹⁴ N transforms in ^l4 C by an η, ρ reaction with a thermal cross section of 1.93 barn. The means that the probability that a ^l4 C atom is formed from a ^l4 N atom is about 1400 times higher than from a ^l3 C atom. On the other hand, in nuclear graphite, ^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

Formation of ¹⁴ C from ^l4 N still about 40 times higher than from ^l3 C.

A similar observation can be made for ^l7 0: ¹⁷ 0 is transformed into ^l C by an η, α-reaction with a thermal cross section of

0.257 barn around. This means that the probability that a ¹⁴ C atom is formed from an O atom is about 180 times higher than from a C atom. 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-

Proportion from ¹⁷ 0 approx. 1/15 of the ^{, 4} C-portion from ¹³ C is. The decisive starting nuclide for the formation of ¹⁴ C in nuclear graphite is therefore ^{, 4} N. This situation is considered below.

^{! 4} N, which is 99.6% contained in natural nitrogen, is incorporated into the nucleus graphite mainly by the nitrogen atmosphere, under which the

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). Although 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.

If 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. Rather, ¹⁴ C will be located near defect sites close to the surface, as is known for neutron-induced radiation defects in graphite. [4] This part of the ¹⁴ C inventory in neutron-irradiated nuclear graphite can be determined from the graphite by a judicious choice of reaction conditions, as shown below remove. ^{However, 4} C atoms, which by chance have found a free lattice site and thus are firmly bound into the graphite lattice, can not be removed in this way, but due to their strong bond also do not pose a problem in disposal Probability are those ¹ C atoms, which have arisen from ¹³ C atoms, which are distributed homogeneously over the entire graphite body according to the natural isotopic composition of carbon, as shown in Figure 4. It should be ^noted that additional ^l3 C is formed by neutron activation of C.

An extreme case is the ¹³ C contained in the carbon black. B. from carbon monoxide during the cooling phase of the annealing and graphitization process at temperatures below 1000 ° C according to the back reaction of the Boudouard equilibrium

2 CO ^ C0 ₂ + C This soot settles on the inner and outer surfaces of the carbon / graphite material. If it is activated ¹³ C content to ^l4 C, is this ¹⁴ C again on the inner and outer surfaces of the carbon / graphite material before and is easily released in accordance with the above, because it with a recoil energy from the η, γ- Activation reaction of a maximum of 2.60 keV is not incorporated into the internal structures of the graphite crystallites.

The soot presents another problem: since 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.

It should be noted that - in addition to nitrogen - other elements such. For example, oxygen, hydrogen and even carbon atoms, etc. on graphite surfaces can be chemisor- phoresed [3],

Other contaminants in graphite are due to chemical contaminants in the unirradiated material. For example, the tritium is due to lithium and Borverunreini ments. There is also evidence that the pore system such impurities as near-surface deposits in the form of so-called "hotspots" (= high concentration accumulations) has.

It has been recognized that 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. In many cases, 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. Thus, the binder area acts more like activated carbon to absorb various contaminants. If, for example, the radionuclide has been formed in situ by a nuclear reaction or has been exposed to neutron irradiation in its place in the waste, 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.

Against this background, 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:

• Grain sizes and surfaces to volume ratios of crystallites

• Density / porosity

• Mechanical properties and stability / strength

• Chemical reactivity including binding energies of impurities

• adsorption or desorption

• Electrical / thermal conductivity / ion migration

These differences are z. T. are mutually linked or conditional upon each other, but they can be used to develop the process of the present invention for fractionating the binder regions and filler particles for the purpose of separating the releasable radionuclides from the binder material and the non-releasable radionuclides from the filler material.

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.

Chemical treatment methods:

The different chemical reactivity of binder and filler materials is a starting point for the development of the method according to the invention.

In the chemical treatment method, 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.

In the context of the invention, for example, the addition of a corrosive medium, the performance of an electrolysis, the supply of thermal energy, ie. H. for example, heating or oxidation.

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. When adding 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). With the attack of the corrosion medium to the carbon atoms of the graphite simultaneously associated 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.

When a corrosion medium is added, a reaction initially takes place with the microcrystalline or amorphous regions of the binder material.

Radionuclides that are bound there, for example, by

• hydrogen or hydrogen ions,

• oxygen or oxygen ions and / or

• A halogen or halogen ions are transferred from the corrosion medium into at least one gaseous reaction product.

In particular, 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. For example, 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.

Preferably, the electrolysis z. B. in dilute nitric acid, wherein the graphite to be decontaminated as anode (positive pole) is connected in a DC circuit. This results in the formation of 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). By this process, as well as by simultaneously occurring oxidation processes due to the formation of native oxygen, it is according to recent findings preferably to a disintegration of the filigree and amorphous binder areas. 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.

Before electrolysis, it may be advantageous to carry out a thermal pretreatment at temperatures in the range of 700 to 1300 ° C. For example, in the electrolysis of irradiated nuclear graphite, tritium can also go into solution. Here, it is advantageous to separate the tritium prior to electrolysis by preheating to use the high degree of enrichment of the baking gases and to avoid larger amounts of weakly enriched tritium as radioactive waste. It should also be noted that in the electrolysis of tritium-containing water at the cathode tritiated, which is not or very difficult to separate from the proportionately far greater amount of ordinary hydrogen, whereby the entire cathode gas would become radioactive waste. Therefore, the electrolysis of irradiated nuclear graphite should be carried out as possible after separation of the tritium.

It is generally advantageous to carry out the electrolysis of irradiated nuclear graphite in separate electrode compartments connected to a diaphragm to prevent graphite particles from passing from the anode to the cathode. Reaction of nascent hydrogen with graphite produces methane, which contains portions of radiocarbon and thus can lead to further undesired contamination of the cathode gas.

In granulated waste molds, it is proposed to carry out the diaphragm as a container for the granules and the electric current through an electrode or to supply a plurality of electrodes in the interior of this container or via the surface-standing granules with which or which the electrically conductive granules are in contact in the electrolyte. Here, too, timely fractionation into binder and filler fractions can be accelerated by the simultaneous use of ultrasound. This application of ultrasound is also possible with massive graphite or coal wastes immersed in an electrolyte.

As a final treatment of the electrolytic decontamination of irradiated nuclear graphite, recirculation of the graphite oxide formed in graphite by baking in an inert gas stream at about 900 ° C. is recommended. Associated with this is a further removal of tritium and radiocarbon similar to the initially mentioned annealing process by 10 to 50% of the residual activity of the graphite oxide.

The electrolysis can be used with appropriate choice of voltage for the triggering of ion migrations. Thus, even before the onset of native sour and hydrogen formation at the electrodes, ionic contaminants are transferred to the electrolyte. In this case, 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.

With the application of an electrical voltage another effect is used: namely the influence of self-diffusion of ions of the radiocarbon compounds and other active nuclide compounds, which form a relatively weak bond after formation upon irradiation and subsequent migration in crystalline and amorphous graphite and beyond an unsaturated Assume electron configuration or slightly ionized form. The strength of the (self-) diffusion is essentially determined by the ratio of the activation energy to the temperature. Since the activation energy for the (self-) diffusion in crystalline and amorphous structures in the range of 1 eV, it can be assumed that a migration or displacement of these compounds - in contrast to the temperature-induced isotropic diffusion - by an externally applied voltage is promoted or accelerated in a certain direction. The outer field causes - in the train of migration - also a polarization of the compounds.

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. Thus, by known methods with high-energy sound pulses / shock waves or else by progressive breaking, grinding and sieving, 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. In addition to the release of z. B. radiocarbon from any volatile compounds from the closed pore system, 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.

Here, too, if necessary, 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. As 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:

It is known that the oxidation of nuclear graphite with oxygen can be divided into three temperature-dependent ranges:

the chemical range (t <500 ° C),

the pore diffusion region (500 ° C <t <900 ° C) and the

Boundary layer diffusion area (/> 900 ° C) [5],

In the chemical field, 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). At the same time concentrated in the binder and volatile with oxygen radionuclides such. B. released radiocarbon. In this area, however, 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.

However, if the radionuclide to be removed is concentrated in a certain area of the graphite body - such as. 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. Thus, one obtains in a relatively short time a radio-carbon-enriched combustion gas, because the Enrichment already exists in the material and does not have to be accomplished by the process.

Since 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%. On the contrary, 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.

The inventors have recognized that in the event that the pore system of the carbon and / or graphite materials already contains reactive gases, the addition of a corrosive agent from the outside can be completely dispensed with. Furthermore, however, a corrosive agent (eg pure oxygen) 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. Because of the high surface area to volume ratio of the binder particles, 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 Thus, according to the invention, 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.

Furthermore, 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. Also oxidizing functional groups (eg hydroxyl, carbonyl, carboxyl, ether or epoxy groups), the unirradiated carbon and / or graphite material and in particular in the nuclear reactor with neutrons irradiated carbon and / or graphite material naturally contained, may be involved in this reaction. 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.

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). Such 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.

Which temperature is advantageous or necessary, inter alia, also depends on the activation energy of the concrete used corrosion medium. The thermal energy kB corresponding to the temperature T. = Boltzmann constant) must correspond at least to this activation energy. In this case, the activation energy can be reduced by a catalyst. Advantageously, 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.

In an advantageous embodiment, the corrosion medium in high

Concentration (e.g., pure oxygen) may be added to it at the inner

Surfaces adsorbed under variation of the deposition and diffusion conditions (temperature and pressure). Upon subsequent heating under inert gas, a higher part of the radio-carbon is then released than is the case only when previously stored under normal pressure, room temperature and air. Also by a suitable selection of the pressure and temperature conditions, the process of accumulation of corrosion media can be accelerated.

According to the known corrosion profile in region II of FIG. 6, 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) are released in

Form of a gaseous Reaktionsprodikt s released and can be converted as already explained in solution, e.g. by use of alkaline adsorbents.

In particularly advantageous embodiments of the method, chemical and mechanical treatment methods can be combined. For example, it may be advantageous that 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. Furthermore, it may be advantageous to heat the carbon and / or graphite materials under mild temperatures and then to connect a mechanical treatment by, for example, breaking, grinding, ultrasound, or electrical discharges to generate shock waves.

The remaining after the separation of the binder material crystalline graphite can be supplied to the recycling cycle and reused.

Show it:

FIG. 1: Scanning electron micrograph of a filler particle (large area in the middle of the figure) with surrounding binder material of an AVR coal test sample (AVR = Arbeitsgemeinschaft Versuchsreaktor Jülich)

FIG. 2: Scanning electron micrograph in FIG

Backscatter electron mode of the AVR coal sample of Figure 1 (white spots = impurities)

Fig. 3: ¹⁴ N depth profile (as CN ^~ ion) of irradiated (a) and unirradiated (b) AVR reflector graphite (AVR - Arbeitsgemeinschaft Versuchsreaktor GmbH, Jülich)

FIG. 4: ^l3 C depth profile of irradiated (a) and unirradiated (b) AVR reflector graphite (AVR - Arbeitsgemeinschaft Versuchsreaktor 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)

FIG. 9: ³ H release against ¹² C release (corrosion) under different conditions (MM = massive and MP = pulverulent graphite sample of the MERLI research reactor, FRJ-1)

FIG. 10: ^l4 C release against ¹² C release (corrosion) under different conditions (MM = massive and MP = powdery graphite sample of the MERLIN research reactor, FRJ-1, AVR = pulverulent graphite sample of the AV R experimental reactor)

Example:

An embodiment of the process according to the invention by electrolysis is described below by way of example:

The electrolysis conditions were the following:

Electrolyte: 5% HN0 ₃

Current: 0.5 A

Electrolysis time: 2 h

Table 1 shows radionuclide release and measurable residual activity in the electrolysis of a graphite sample of an irradiated AVR fuel.

As a graphite sample, a piece was taken from the spherical surface of an AVR fuel assembly (AVR = Arbeitsgemeinschaft Versuchsreaktor GmbH, Jülich) with a mass of 1.8 g. The electrolyte used was 5% nitric acid. The graphite sample was switched as an anode (positive pole) in a DC circuit. The cathode (negative pole) was an inactive (non-radioactive) graphite electrode. The current was set to 0.5 amps using a potentiostat. The Elektrol sedauer was 2 hours. During this time, the phitprobe on and it dropped graphite flakes from the graphite sample. After completion of the electrolysis, 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. By means of infrared spectroscopy, 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 Subsequently, 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.

It was recognized that 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. As has already been stated, 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

Contains 40% of the initial mass of the graphite sample

2 ⁾ Contains 60% of the initial mass of the graphite sample

The formation of one of the reaction products, the graphite oxide, was associated with a mass gain of 17% over the mass of the starting graphite. This is due to the binding of oxygen to the graphite lattice planes.

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). In both figures one can clearly see the flickering graphite particles, which are still connected by the binder in Figure 7 and separated (and chemically modified) in Figure 8 by the electrolysis. 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).

literature

[1] H.-K. Hinssen, K. Kuhn, R. Moormann, B. Schlögl, M. Fechter, M. Mitchell: Oxidation experiments and theoretical examinations on graphite materials relevant to the PBMR, Nucl. Closely. Of. 238 (2008) 3018-3025.

[2] R. Dovesi, C. Pisani, F. Ricca, C. Roetti; Chemisorption of periodic overlayers of atomic nitrogen on graphite, Surface Sei. 77 (1978) 409-415.

[3] T. Fromherz, C. Mendoza, F. Ruette: Chemisorption of atomic H _s C, N and O on a cluster model graphite surface, Mon. R. Astron. Soc. 263 (1993) 851-860.

[4] R.H. Telling, M.I. Heggie: Radiation Defects in Graphite, Phil. Mag. 87 (2007) 4797 ^ 1846.

[5] W. Delle, K. Koizlik, H. Nickel: Graphitic materials for use in nuclear reactors. Part 2: Polycrystalline graphite and fuel element matrix, Karl Thiemig AG, Munich, 1983, p. 133.

[6] B. Schlögl: Oxidation kinetics of innovative carbon materials with respect to heavy air intrusion perturbation in HTR 's and graphite disposal or processing, dissertation, RWTH Aachen University, 2009. The work leading to this invention has been funded under the Grant Agreement No 21 1333 under the Seventh Framework Program of the European Atomic Energy Community [FP7 / 2007-201 1].

Claims

Patent claims

A method of decontaminating releasable radionuclides from neutron irradiated carbon and / or graphite materials consisting of a heterogeneous compound of crystalline filler materials and microcrystalline amorphous binder materials, comprising mechanical and / or chemical treatment methods,

the mechanical treatment methods comprising the steps of breaking, shaking and sieving, and

the chemical treatment methods comprise a supply of activation energy,

characterized in that

the mechanical treatment processes and / or the supply of the amount of activation energy are adjusted such 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 are attacked and dissolved, and

that a subsequent fractionation of reaction products, which were separated by the mechanical and / or chemical treatment processes from the carbon and / or graphite materials.

2. The method according to claim 1, characterized in that at least one radionuclide, with which the neutron irradiated carbon and / or graphite materials are loaded, by a nuclear reaction, in particular nuclear fission or Neutronenaktivierung was formed.

3. The method according to any one of claims 1 to 2, characterized in that graphite, nuclear graphite or carbonaceous material are selected as neutron irradiated carbon and / or graphite materials.

4. The method according to any one of claims 1 to 3, characterized in that a ¹² C and / or ¹³ C-containing carbon and / or graphite material is selected.

Method according to one of claims 1 to 4, characterized in that in the form of solid pieces, as granules or powder present carbon and or graphite material is selected.

Method according to one of claims 1 to 5,

characterized,

that the radionuclides bound to a binder material used in the preparation of the matrix of the carbon and / or graphite material are removed.

, Method according to one of claims 1 to 6,

characterized in that 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 are comminuted by sonic pulses, shock waves, breaking, or grinding up to the maximum grain size of the filler material.

Method according to claim 7,

characterized in that

the resulting components of filler and binder material based on the different crystal size / grain size are fractionated from each other.

Method according to one of claims 7 to 8, characterized in that it is carried out in a rotating reaction vessel, in a mixer or in a fluidized bed reactor.

0. Method according to one of claims 7 to 9,

characterized in that

Before mechanical treatment, a thermal pretreatment, especially at temperatures of 350 to 900 ° C is performed.

11. The method according to any one of claims 1 to 6, characterized in that

an oxidation with oxygen at temperatures of more than 350 ° C and below 900 ° C is performed.

12. The method according to any one of claims 1 to 6,

characterized in that

a corrosion medium is supplied.

13. The method according to claim 12, characterized in that at least one binary metal oxide, sulfate, nitrate, hydroxide, carbonate, hydride and / or halogeno-succinimide is selected as the corrosion medium.

14. The method according to any one of claims 12 to 13, characterized in that a liquid corrosion medium is selected.

15. The method according to any one of claims 12 to 14, characterized in that a peroxide, water vapor, an acid, carbon dioxide, a complexing agent, a halogen compound, a hydrocarbon, a halohydrocarbon and / or other pyrolyzable and / or dehydratisierbares reagent selected as the corrosion medium becomes.

16. The method according to any one of claims 12 to 15,

characterized in that

the hydrogen, the hydrogen ions, the oxygen, the oxygen ions, the halogen and / or the halogen ions are formed in supplying the activation energy from the corrosion medium.

17. The method according to any one of claims 12 to 16, characterized in that the hydrogen, the hydrogen ions, the oxygen, the oxygen ions, the halogen and / or the halogen ions react in the nascent state with the radionuclide.

18. The method according to any one of claims 12 to 17, characterized in that a present in the form of solid pieces, as granules or powder corrosion medium is selected.

19. The method according to any one of claims 12 to 18, characterized in that water, a peroxide, an acid, an alkali, a complexing agent and or an electrolyte are selected as the corrosion medium.

20. The method according to any one of claims 1 to 6,

characterized in that

an electrolysis is carried out, wherein the conditions of the electrolysis are chosen so that there is a disintegration of the microcrystalline and amorphous binder regions.

21. The method according to claim 20,

characterized in that

the electrolysis is carried out in separate electrode compartments connected to a diaphragm.

22. The method according to any one of claims 20 to 21,

characterized in that

before the electrolysis, a thermal pretreatment at temperatures of 700 to 1300 ° C is performed.

23. The method according to any one of claims 1 to 22,

characterized in that

gaseous reaction products by alkaline absorbents, in particular alkalis, are collected.

24. The method according to any one of claims 1 to 6,

characterized in that

a thermal treatment at temperatures between 350 and 1500 ° C, preferably between 500 and 1300 ° C is performed.

25. The method according to claim 24,

characterized,

that the thermal treatment is controlled by the adjustment of the process temperature and process time so that there is a disintegration of the microcrystalline and amorphous binder regions.

26. The method according to any one of claims 24 to 25,

characterized,

a mechanical aftertreatment according to claims 7 to 9 is carried out.

27. The method according to any one of claims 1 to 26, characterized in that 1 ⁴ C and / or a metallic radionuclide is converted into at least one gaseous reaction product and converted into solution.

28. The method according to any one of claims I to 27,

characterized in that after separation of the binder material of the remaining crystalline graphite is reused.