CN116265618A - Molten salt electrolysis method for treating uranium-containing material - Google Patents
Molten salt electrolysis method for treating uranium-containing material Download PDFInfo
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- C—CHEMISTRY; METALLURGY
- C25—ELECTROLYTIC OR ELECTROPHORETIC PROCESSES; APPARATUS THEREFOR
- C25C—PROCESSES FOR THE ELECTROLYTIC PRODUCTION, RECOVERY OR REFINING OF METALS; APPARATUS THEREFOR
- C25C3/00—Electrolytic production, recovery or refining of metals by electrolysis of melts
- C25C3/34—Electrolytic production, recovery or refining of metals by electrolysis of melts of metals not provided for in groups C25C3/02 - C25C3/32
-
- C—CHEMISTRY; METALLURGY
- C25—ELECTROLYTIC OR ELECTROPHORETIC PROCESSES; APPARATUS THEREFOR
- C25C—PROCESSES FOR THE ELECTROLYTIC PRODUCTION, RECOVERY OR REFINING OF METALS; APPARATUS THEREFOR
- C25C7/00—Constructional parts, or assemblies thereof, of cells; Servicing or operating of cells
- C25C7/005—Constructional parts, or assemblies thereof, of cells; Servicing or operating of cells of cells for the electrolysis of melts
-
- C—CHEMISTRY; METALLURGY
- C25—ELECTROLYTIC OR ELECTROPHORETIC PROCESSES; APPARATUS THEREFOR
- C25C—PROCESSES FOR THE ELECTROLYTIC PRODUCTION, RECOVERY OR REFINING OF METALS; APPARATUS THEREFOR
- C25C7/00—Constructional parts, or assemblies thereof, of cells; Servicing or operating of cells
- C25C7/02—Electrodes; Connections thereof
- C25C7/025—Electrodes; Connections thereof used in cells for the electrolysis of melts
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- Chemical & Material Sciences (AREA)
- Engineering & Computer Science (AREA)
- Chemical Kinetics & Catalysis (AREA)
- Electrochemistry (AREA)
- Materials Engineering (AREA)
- Metallurgy (AREA)
- Organic Chemistry (AREA)
- Electrolytic Production Of Metals (AREA)
Abstract
The invention relates to a molten salt electrolysis method for treating uranium-containing materials, and belongs to the field of uranium metallurgy. The method is implemented by adopting a molten salt electrolytic tank, wherein the tank body of the molten salt electrolytic tank is divided into a first chamber, a second chamber and a lower chamber, and a first molten salt electrolyte and an anode are placed in the first chamber; a second molten salt electrolyte and a cathode are disposed within the second chamber. And connecting the anode with the positive electrode of the power supply, connecting the cathode with the negative electrode of the power supply, and electrifying for electrolysis to enable the anode to perform oxidation reaction and generate uranium ions, so that reduction reaction occurs on the surface of the cathode and a metal uranium product is obtained. The method has the advantages of short flow, strong impurity removal capability, high product purity and the like.
Description
Technical Field
The invention belongs to the field of uranium metallurgy, and particularly relates to a molten salt electrolysis method for treating uranium-containing materials.
Background
Nuclear energy is a clean and efficient new energy, the type of a nuclear power station in the current and the future twenty-three years is mainly a thermal neutron reactor, and the working principle is as follows: u-235 with the content of 1.8-4.5% is used as nuclear fuel, the U-235 undergoes nuclear fission reaction under the action of thermal neutrons and generates a large amount of heat energy, and the heat energy is converted into electric energy through multiple steps. Nuclear fuels can be divided into three categories by material type:
(1) a metal type nuclear fuel. Uranium metal and uranium alloy are used as main materials, the application history is long, and the method has the advantages of high fissionable atomic density, high thermal conductivity and the like, but has relatively poor irradiation resistance and very active chemical properties.
(2) Ceramic type nuclear combustionAnd (5) material. Including oxide, carbide and nitride types, among which are UO 2 The fuel is the ceramic nuclear fuel with the most application at present, and has the characteristics of no phase change at high temperature, good chemical stability and poor thermal conductivity. The UC fuel and the UN fuel have the advantages of high heat conductivity, high density and the like, but have active properties or high manufacturing cost.
(3) A dispersion type nuclear fuel. Refers to a nuclear fuel formed by uniformly dispersing a fuel phase (particles containing fissionable nuclides) in a matrix phase (non-fissionable material). Common fuel phases include intermetallics of uranium with aluminum, beryllium, oxides, nitrides, carbides, etc. of uranium.
As the fission reaction proceeds, some fission products (neutron poisons) are generated in the nuclear fuel, which have a strong absorption of neutrons, so that the reactor can no longer sustain the progress of the nuclear fission chain reaction, and thus have to be discharged from the reactor and replaced with new fuel, which is also called spent fuel, and has a strong radioactivity.
The ratio of metal elements in spent fuel is about: -94% uranium, -4% fission products and-1% plutonium. The fission products are of a wide variety and very complex, including: minor actinides (Np, am, cm), long-lived fissile species (e.g., tc-99, I-129, se-79, zr-93, cs-135), lanthanides (e.g., la, ce, pr, nd), noble metals (e.g., ru, pd, ag), alkali and alkaline earth metals (e.g., cs, sr, ba), rare refractory metals (e.g., zr, mo).
In order to dispose of such spent fuel, which is potentially harmful to humans and the environment, while recovering a large amount of uranium and plutonium therein, which can be recycled or recycled, a number of aqueous and dry post-treatment processes have been proposed.
The water method post-treatment process is represented by PUREX flow, after spent fuel is sheared and acid-dissolved, uranium and plutonium are selectively extracted by an organic extractant, and then UO is prepared respectively 2 And PuO 2 The raffinate containing highly reflective elements requires further processing. The method is widely adopted at present, but the extraction process flow is complex, the equipment scale is large, and the organic extraction is carried outThe agent is not resistant to irradiation, and a large amount of organic waste liquid which is difficult to treat is generated. These disadvantages result in the inability of the water-based post-treatment processes such as the PUREX process to accommodate future post-treatment requirements of spent fuel.
The dry method is always regarded as a candidate technology for the post-treatment of the next generation of spent fuel, and has the advantages of capability of treating different types of spent fuel, radiation resistance, low critical risk of high-temperature inorganic salt, compact equipment structure and the like.
The dry post-treatment process comprises a volatilization method, an oxide electrodeposition method, a fused salt electrolysis method and a melt extraction method, wherein the United states proposes the most representative and application potential of the fused salt electrolysis method, and the dry post-treatment process mainly comprises the following four steps: head end treatment, electrolytic reduction, electrolytic refining, molten salt purification and waste treatment.
Head end processing: cutting spent fuel, removing cladding, oxidizing and volatilizing and treating waste gas. UO (UO) 2 Through this step, it is mostly converted into U 3 O 8 。
Electrolytic reduction: cathode-accepting TiO 2 Inspiring the fused salt electro-deoxidation technology (namely the famous FFC-Cambridge method), the oxide ceramic spent fuel can also adopt the fused salt electro-deoxidation technology to prepare metallic uranium, namely uranium oxide is taken as a cathode, inert metallic platinum is taken as an anode, and LiCl-Li at 650℃ is taken as an anode 2 The O molten salt system is electrolyte, and oxygen ions in the cathode uranium oxide are removed and coarse metal uranium is generated through electro-deoxidation and in-situ lithium thermal reduction in the electrolysis process. In the process, the deep impurity removal function is not provided, and fission products still remain in cathode product metal uranium.
Electrolytic refining: the coarse metal uranium is used as anode to contain UCl 3 Firstly, stainless steel is used as a cathode for electrified electrolysis, a crude metal uranium anode is subjected to oxidation and dissolution reaction and uranium ions, plutonium ions and other fission product ions (including minor actinide ions, rare earth ions and the like) are generated, and the reduction potential of the uranium ions is corrected in the molten salt, so that solid metal uranium with higher purity is preferentially deposited on the surface of the stainless steel cathode; as electrolysis proceeds, plutonium ions and fission product ions are continuously enriched in the molten salt, at which time the cathode is replaced toAnd (3) liquid metal cadmium, uranium ions and plutonium ions in the molten salt enter the liquid cadmium through underpotential deposition, the liquid cadmium is then extracted, and cadmium is removed through a vacuum distillation method, so that a uranium-plutonium mixed metal product is obtained.
Finally, the molten salt containing minor actinides and rare earth elements, which are considered "neutron poisons", need to be further processed.
It can be seen that the electrolytic refining process is not only cumbersome but also has the following drawbacks: (1) the solubility of the liquid cadmium cathode to uranium and plutonium is extremely low, and the alloying rate of the uranium is slow, so that the uranium can form a solid metal shell on the surface of the liquid cadmium in the electrolysis process, and then solid dendrites grow upwards; (2) the liquid cadmium cathode has the same dissolution effect and underpotential deposition effect on fission products such as rare earth elements, minor actinides and the like, and the selectivity of the refining process is greatly reduced; (3) liquid cadmium needs to be subjected to vacuum distillation frequently, so that the energy consumption is high, and the cadmium has extremely toxic property.
Based on this, a new method and apparatus for treating uranium-containing materials by dry process are needed to solve the above problems.
Disclosure of Invention
The invention aims to provide a method for treating uranium-containing materials by a fused salt electrolysis method, which is used for producing metal nuclear fuel or is used for post-treatment of spent fuel so as to solve the problem that rare earth metal impurities and minor actinide impurities are difficult to separate in the prior art.
In order to achieve the above purpose, the present invention adopts the following technical scheme:
the invention provides a molten salt electrolysis method for treating uranium-containing materials, which is suitable for a molten salt electrolysis cell, wherein the molten salt electrolysis cell comprises an electrolysis cell body and an insulating partition plate; the electrolytic tank body is provided with an upper chamber and a lower chamber, and an insulating partition plate is arranged in the upper chamber so as to divide the upper chamber into a first chamber and a second chamber;
the lower chamber is filled with liquid alloy, the first chamber is filled with a first molten salt electrolyte, and an anode is arranged in the first molten salt electrolyte; a second molten salt electrolyte is contained in the second chamber, and a cathode is arranged in the second molten salt electrolyte; the first molten salt electrolyte and the second molten salt electrolyte are separated by the insulating separator, both of which are in contact with the liquid alloy; wherein the anode comprises uranium-containing material; the liquid alloy is an alloy of uranium metal and auxiliary metal;
the method comprises the following steps: and connecting the anode with the positive electrode of the power supply, connecting the cathode with the negative electrode of the power supply, and electrifying for electrolysis to enable the anode to perform oxidation reaction and generate uranium ions, and performing reduction reaction on the surface of the cathode to obtain a metal uranium product.
The uranium-containing material is crude metal uranium or/and uranium compounds. The crude metal uranium or/and uranium compounds can be taken from raw materials before nuclear fission reaction to produce metal uranium nuclear fuel; the spent fuel after the nuclear fission reaction can also be taken for a dry post-treatment stage.
The uranium compound is composed of uranium and a nonmetallic element, wherein the nonmetallic element comprises one or more of oxygen, carbon and nitrogen.
Preferably, the uranium compound is one or more of uranium oxide, uranium carbide, uranium nitride, uranium oxycarbide, uranium oxynitride, uranium carbonitride and uranium oxycarbonitride.
For example, uranium oxide includes UO 2 ，U 3 O 8 ，UO 3 The method comprises the steps of carrying out a first treatment on the surface of the The uranium carbide comprises UC, U 2 C 3 ，UC 2 The method comprises the steps of carrying out a first treatment on the surface of the Uranium nitride includes UN, U 2 N 3 ，UN 2 The method comprises the steps of carrying out a first treatment on the surface of the Uranium oxycarbide to UC 1-x O x (0<x<1) For example UC 0.5 O 0.5 (i.e. U) 2 CO or UO-UC); uranium oxynitride as UN 1-x O x (0<x<1) The method comprises the steps of carrying out a first treatment on the surface of the Uranium carbonitride UC y N 1-y (0<y<1) The method comprises the steps of carrying out a first treatment on the surface of the Uranium oxy-carbon to UC x O y N 1-x-y (0<x+y<1)。
Preferably, the uranium containing material anode may be connected to the positive electrode of the power source by a current collector. The current collector may be selected from metals having a lower activity and a higher melting point than metallic uranium, such as stainless steel, platinum, tungsten, molybdenum. The polar current collector is a basket for holding uranium-containing materials, or is a bar material directly inserted into the uranium-containing materials, or is a wire material for binding the uranium-containing materials, and can be applied to the invention in any shape capable of forming conductive contact with the uranium-containing materials.
Uranium compounds containing carbon or nitrogen have good electrical conductivity and undergo oxidation reactions upon contact with the anode current collector. For higher uranium oxides, e.g. UO 3 The upper limit of the content of the carbonaceous conductive agent is to set the molar ratio of the carbon element to the oxygen element in the uranium compound to 1, that is, n (C): n (O). Ltoreq.1, in order to improve the conductivity as much as possible and to avoid carbon powder precipitation side reactions due to excessive carbon.
For some carbon-rich uranium compounds (e.g. UC), the method comprises adding a certain proportion of oxygen-rich uranium compounds (e.g. U 3 O 8 、UO 2 ) And further sintering treatment, a large amount of carbon powder can be prevented from being precipitated in the electrolysis process.
In the anode, the oxidation reaction of crude metallic uranium or some typical uranium compounds occurring during electrolysis is:
U-ne - →U n+ n＝3～6
UO 2 +C-ne - →U n+ +CO/CO 2 ↑
UC+UO 2 -ne - →U n+ +CO/CO 2 ↑
UC 0.5 O 0.5 -ne - →U n+ +CO↑
UN-ne - →U n+ +N 2 ↑
UC 0.25 O 0.25 N 0.5 -ne - →U n+ +CO↑+N 2 ↑
the crude metal uranium is low-purity metal uranium or waste uranium metal, and the waste uranium metal mainly refers to metal uranium spent fuel.
Preferably, the crude metallic uranium or the uranium compound is produced from uranium oxide by a fused salt electro-deoxidation method or a high-temperature thermal reduction method.
PreferablyThe fused salt electro-deoxidation method uses uranium oxide as a cathode or a mixture of uranium oxide and uranium carbide or/and carbon powder as a cathode, and LiCl-Li 2 O-based fused salt or CaCl 2 And (3) taking CaO-based molten salt as an electrolyte, adopting an inert anode or a graphite anode, and electrolyzing at 650-1000 ℃ to obtain a cathode product which is crude metal uranium or/and uranium oxycarbide. Regarding the preparation of the cathode sheet by the molten salt electro-deoxidation method, the general flow is as follows: and adding a certain amount of binder into uranium oxide and a mixture thereof, uniformly mixing, pressing and sintering to form the cathode. To reduce LiCl-Li 2 O-based fused salt or CaCl 2 Melting point of CaO-based fused salts, optionally with the addition of other chloride salts, such as KCl, naCl.
Preferably, the high-temperature thermal reduction method comprises metallothermic reduction and carbothermic reduction, wherein one or more of active metals lithium, sodium, calcium and magnesium are used as reducing agents, or carbon powder or/and uranium carbide are used as reducing agents, and uranium oxide is reduced at the temperature of more than 500 ℃ under non-oxidizing atmosphere to prepare crude metal uranium or uranium compounds. The products with different phase compositions, such as UC, can be obtained by controlling the factors such as the C/O ratio, the temperature, the atmosphere, the time and the like in the mixture 1-x O x The value of x in the mixture is influenced by the C/O ratio in the mixture, and nitrogen-containing uranium compounds such as uranium oxy-carbon can be obtained in a nitrogen atmosphere. Of course, crude metallic uranium may also be produced by the metallothermic reduction of uranium halides.
It is worth mentioning that, for uranium oxide raw materials containing rare earth oxide impurities, in the fused salt electro-deoxidation method or the high-temperature thermal reduction method, the rare earth elements are more active and more oxygen-philic, so that the rare earth oxide is more difficult to reduce than uranium oxide and minor actinide oxides, and certain rare earth oxide impurities still remain in the reduction product.
In the anode, active metals (alkali metals, alkaline earth metals, rare earth metals, and transuranic metals) in the fission products are dissolved into the molten salt by chemical dissolution reaction or anodic oxidation to generate chlorides. And part of rare earth oxide, zirconium and noble metal cannot be electrochemically dissolved and remain in the anode scrap. Thus, a preliminary separation of uranium from fission products, in particular rare earth elements with "neutron toxicity", can be achieved by anodic reactions.
Preferably, the first molten salt electrolyte or the second molten salt electrolyte is composed of an alkali metal halide or/and an alkaline earth metal halide, and a halide of uranium is dissolved.
Specifically, the alkali metal halide is one or more of LiCl, naCl, KCl, rbCl, csCl, liF, naF, KF, rbF, csF; the alkaline earth metal halide is MgCl 2 ，CaCl 2 ，SrCl 2 ，BaCl 2 ，MgF 2 ，CaF 2 ，SrF 2 ，BaF 2 One or more of the following. Alkali metal halides and alkaline earth metal halides serve as supporting electrolytes and serve to dissolve uranium halides, and both alkali metal ions and alkaline earth metal ions are more difficult to reduce than uranium ions. Lithium salts, sodium salts, potassium salts, magnesium salts and calcium salts are preferable in view of price cost and ease of handling.
Preferably, the uranium halide is UCl n Or/and UF n (3.ltoreq.n.ltoreq.6). These uranium halides are used to provide uranium ions in either dissociated or complexed states. I.e. comprising U in dissociated or complexed form 6+ 、U 5+ 、U 4+ And U 3+ One or more ions of (a) and (b). For higher uranium halides, it may be present/added in the form of complex salts.
Further, the first molten salt electrolyte and the second molten salt electrolyte may be the same or different in composition.
The cathode is one of materials such as metallic uranium, molybdenum, tungsten, platinum and the like which are difficult to carry out alloying reaction with the metallic uranium.
Preferably, the auxiliary metal comprises one or more of Fe, co, ni, cr, mn. These several auxiliary metals have a high solubility for uranium and can form alloys with a melting point below 1000 c, for example 70at% u-Fe alloy remains liquid at 800 c, 78.5at% u-Mn alloy has a melting point of only 716 c, 77at% u-Ni alloy has a melting point of 740 c, or perhaps lower.
Plutonium, neptunium, americium, curium and other minor actinides are of similar nature to uranium and can be dissolved in the liquid alloy by substitution of uranium atoms. However, the solubility of rare earth metals in the auxiliary metals is small, for example, the melting temperature of 1-50 at% La-Fe alloy is above 1000 ℃, and other rare earth metals are similar. Furthermore, the rare earth element differs so much from the actinide element in nature that it is difficult to form a miscible liquid alloy, for example, la has a solubility in U of 1.3at% at 1150 ℃, which means that rare earth metal impurities are difficult to dissolve into the liquid alloy. Therefore, the invention can effectively separate rare earth metal impurities which are difficult to separate by other methods.
Further, the temperature in the electrolytic tank is 700-1000 ℃.
Further, the cathode current density is 0.01-2.0A/cm 2 Or controlling the anode current density to be 0.01-1.5A/cm 2 。
Furthermore, the power-on electrolysis mode is not limited, and can be arbitrarily selected in the modes of voltage transformation, constant voltage, current transformation and constant current.
Preferably, the electrified electrolysis mode is constant voltage, constant current and unidirectional pulse.
The invention provides a molten salt electrolysis method for treating uranium-containing materials, which comprises the following steps:
in the electrified state, in the first chamber, uranium-containing material anodes undergo oxidation reaction to be consumed, and uranium ions are generated and enter a first molten salt electrolyte; uranium ions in the first molten salt electrolyte undergo a reduction reaction at the interface of the first molten salt electrolyte and the liquid alloy in the lower chamber and generate uranium atoms into the liquid alloy, and the reaction equation is as follows:
U n+ +ne - u (liquid alloy)
Metal impurities more reactive than uranium (e.g., alkali metal, alkaline earth metal, and rare earth ion impurities) oxidize out of the anode and enter the first molten salt electrolyte in ionic form, but are more difficult to reduce and enter the liquid alloy than uranium ions. In particular, the liquid alloy also exhibits "resistance" to "neutron poison" rare earth elements even if the rare earth ions are reduced to rare earth metals at the interface due to concentration buildup and the like, the rare earth metals float on top of the liquid alloy due to low solubility, are difficult to enter the interior of the liquid alloy, and are more difficult to migrate through the liquid alloy to the second chamber. Such impurities are effectively trapped in the first molten salt electrolyte. For metallic impurities and their compounds (e.g., fe) that are slightly more inert than uranium, oxidation reactions generally do not occur and are difficult to enter into the molten salt, even if under anodic polarization, oxidation reactions occur and enter into the first molten salt electrolyte, which is then reduced to metal at the liquid alloy interface and dissolved into the liquid alloy.
In the second chamber, uranium atoms in the liquid alloy in the lower chamber undergo oxidation reaction at the interface of the liquid alloy and the second molten salt electrolyte and generate uranium ions to enter the second molten salt electrolyte, and uranium ions in the second molten salt electrolyte undergo reduction reaction at the surface of the cathode to generate metallic uranium products, wherein the reaction equations are respectively as follows:
u (liquid alloy) -ne - →U n+
U n+ +ne - →U
For impurities entering the liquid alloy, plutonium, other actinides and other impurity atoms which are more inert than uranium have low concentration and low electrochemical activity, so that oxidation reaction is difficult to occur and the impurities enter a second molten salt electrolyte, namely the impurities are continuously enriched in the liquid alloy, and the purity of a cathode product metallic uranium is not influenced; in which impurity atoms more reactive than uranium are difficult to precipitate at the cathode due to a difference in deposition potential even if they are oxidized at the interface to generate cations and enter the second molten salt electrolyte.
Therefore, the metal impurities or fission products which are more active or more inert than uranium can be effectively blocked through multiple electrochemical reaction interface actions, molten salt dissolution dilution actions and selective dissolution actions of liquid alloy, and metal uranium products with higher purity can be obtained. Particularly has good interception effect on rare earth elements, and can separate out a metal uranium product with the rare earth element content not more than 1% at the cathode, thereby meeting the requirements of nuclear fuel for fast neutron reactors.
Along with the long-term operation of the molten salt electrolysis bath, when the impurity content of the liquid alloy is continuously enriched and reaches a certain degree, the liquid alloy needs to be purified.
For nonmetallic impurities, free carbon, oxygen-containing compounds and dropped anode slag in the first chamber are not in contact with the second molten salt electrolyte and the cathode in the second chamber in space, so that the problem of high impurity content of C, O and the like in a metallic uranium product and the problem of short circuit of carbon powder contact with a conductive agent caused by the problem can be avoided.
After multi-batch electrolysis, the first molten salt electrolyte is mainly enriched with fission products such as alkali metal, alkaline earth metal, rare earth metal and the like, and the fission products can be extracted through the existing purification treatment process, such as molten salt/molten alloy extraction, molten salt electrodeposition, aqueous solution dissolution and other purification treatment processes. The liquid alloy is mainly enriched with plutonium and minor actinides, these fission products can be extracted by molten salt electrolysis or melt extraction processes, such as the molten salt electrolysis process: the liquid alloy is taken as an anode to carry out molten salt electrolysis, uranium, plutonium and minor actinides in the liquid alloy are extracted and deposited into a cathode, and mixed actinide metal products are changed into low radionuclides with medium and short service lives or more stable by transmutation, or the low radionuclides are mixed with the metal uranium products to be directly used as fast neutron reactor nuclear fuel.
The invention has adaptability to the anode slag dropping/breaking phenomenon in the electrolytic process. For anode slag/blocks of metallic uranium or uranium alloy, uranium atoms in the anode slag/blocks can be dissolved into liquid alloy, and the effect of electrolytic refining can still be achieved by means of electrochemical interface reaction of the second chamber. For anode slag/lumps of uranium compounds, which float at the interface of the first molten salt electrolyte and the liquid alloy, a reduction reaction occurs in turn to uranium atoms and dissolve into the liquid alloy, e.g. UO 2 The following reactions occur at the interface:
UO 2 -4e - u (liquid alloy) +2O 2-
Dissociated O 2- Enters the first molten salt electrolyte and migrates toward the anode, and then oxidation reaction occurs to generate gas evolution. Therefore, the invention has no phase structure for uranium compoundsThere are stringent requirements, i.e. not limited to solid solution forms only, but also near solid solutions or composite materials.
The invention has adaptability to the dropping phenomenon of products in the electrolysis process. The dropping of the product is generally attributable to chemical factors, which are the disproportionation reaction of metal uranium ions in molten salt to produce elemental metal, or physical factors, which are the dropping of the metal product at the cathode due to weak bonding. In the prior art, the dropped product is settled at the bottom of the molten salt electrolysis tank so as to be difficult to recycle, and even if recycled, the product is poor in quality and needs to be further subjected to a purification process. The product powder falling from the molten salt or the cathode can be absorbed by the liquid alloy at the bottom of the molten salt electrolytic tank, then the product powder continuously participates in the reaction according to the electrochemical mechanism, and finally the metal product is separated out from the cathode.
The invention also has the following beneficial effects on the basis of inheriting the advantages of the dry post-treatment process:
(1) Continuous production and short process. The anode is utilized to directly produce the metal uranium by a fused salt electrolysis method, and the one-step fused salt electrolysis operation is equivalent to a multi-step separation process in the existing treatment process, so that complicated vacuum distillation operation is not required. The continuous operation of molten salt electrolysis can be realized by arranging a plurality of anodes in the molten salt electrolysis tank to realize the timely taking out of the residual anode and the timely adding of the new anode, or by adopting a continuous feeding mode into the basket type anode, and the production efficiency is high.
(2) The electrolytic raw material is wide and the processing capability is wide. The method can generate a metal uranium product by utilizing a natural uranium compound, can also carry out electrolytic refining on crude metal uranium, and is more suitable for treating various spent fuel in metal type, ceramic type, dispersion type and the like.
(3) The purity of the product is high, and the byproducts are easy to treat. Impurity ions with different electrochemical behaviors can be effectively controlled in molten salt or liquid alloy, multiple electrochemical interface reactions and the dissolution of different melts ensure the purity of cathode product metal, and simultaneously, transuranics (plutonium and minor actinides) and rare earth elements are respectively enriched in the liquid alloy and the first molten salt electrolyte, so that the reduction ofThe pressure of the subsequent purification treatment is reduced; the gases evolved by anodic electrolysis being non-corrosive, e.g. CO 2 And N 2 Is nontoxic and harmless, and CO is converted into CO after oxidation/combustion 2 。
(4) The operation adaptability is strong. In the electrolysis process, the soluble anode continuously provides metal uranium ions for molten salt so as to maintain the concentration balance of ionic or atomic metal uranium in each melt in the molten salt electrolysis tank. The dropped residue can still participate in the reaction at the interface of the liquid alloy or be blended into the liquid alloy and then participate in the reaction at the interface, thereby improving the yield and correspondingly adopting solid solution or composite anode. Uranium powder which is disproportionated and separated out in the electrolyte or falls from the cathode is dissolved in the liquid alloy and continuously participates in the reaction, so that the loss rate is reduced. The liquid alloy has strong depolarization effect, for polyvalent metal, the incomplete reduction of high-valence metal uranium ions and current consumption caused by the incomplete reduction are relieved in the first chamber, and low-valence metal uranium ions are easier to generate in the second chamber, so that the low-valence metal uranium ions are easily separated out as coarse-grained metal products on the surface of the cathode.
Drawings
In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the drawings required for the description of the embodiments will be briefly described below, and it is obvious that the drawings in the following description are only some embodiments of the present invention, and other drawings may be obtained according to these drawings without inventive effort for a person skilled in the art.
FIG. 1 is a schematic diagram of a molten salt electrolysis apparatus according to the present invention;
reference numeral 1: 1-a current collector; 2-uranium-containing material; 3-a first molten salt electrolyte; 4-liquid alloy; 5-an electrolytic tank body; 6-a second molten salt electrolyte; 7-metallic uranium; 8-cathode; 9-insulating spacers.
Fig. 2 is a flow chart of a method for processing uranium-containing material by molten salt electrolysis according to the present invention.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the technical solutions of the present invention will be described in detail below. It will be apparent that the described embodiments are only some, but not all, embodiments of the invention. All other embodiments, based on the examples herein, which are within the scope of the invention as defined by the claims, will be within the scope of the invention as defined by the claims.
FIG. 1 is a schematic view of a molten salt electrolysis apparatus according to the present invention. Referring to fig. 1, the molten salt electrolysis apparatus includes an electrolysis cell body 5 and an insulating separator 9. The electrolytic tank body 5 has an upper chamber and a lower chamber, and an insulating partition 9 is provided in the upper chamber to partition the upper chamber into a first chamber and a second chamber.
Placing a liquid alloy 4 in the lower chamber, placing a first molten salt electrolyte 3 in the first chamber, and arranging an anode in the first molten salt electrolyte 3; the anode comprises a uranium-containing material 2 and a current collector 1, the uranium-containing material 2 being immersed in a first molten salt electrolyte 3, wherein the current collector 1 is designed as a basket.
Placing a second molten salt electrolyte 6 in the second chamber and disposing a cathode 8 in the second molten salt electrolyte 6; the first molten salt electrolyte 3 and the second molten salt electrolyte 6 are separated by the insulating separator 9, and the first molten salt electrolyte 3 and the second molten salt electrolyte 6 are both in contact with the liquid alloy 4; the liquid alloy 4 is an alloy of metal uranium and auxiliary metal, and the liquid alloy 4 is used for constructing an electrochemical reaction interface of uranium ions/uranium atoms and is used as a delivery medium of the uranium atoms.
The anode is connected with the positive electrode of a power supply, the cathode 8 is connected with the negative electrode of the power supply and electrified, the temperature is 800-1000 ℃, and the anode current density is 0.01-1.5A/cm 2 Or controlling the cathode current density to be 0.01-1.5A/cm 2 The method comprises the steps of carrying out a first treatment on the surface of the In the electrolytic process, uranium-containing material 2 undergoes oxidation reaction and uranium ions are generated, and metallic uranium 7 products are separated out from the surface of cathode 8.
In addition to the cell body 5 shown in fig. 1, the structure of the cell body 5 may be designed in various forms, such as a U-shaped cell. The shape of the cell body 5 may be varied, and for example, the bottom of the cell is not limited to a round bottom, but may be a trapezoidal bottom or a flat bottom. The molten salt electrolysis cell capable of achieving physical separation of the first molten salt electrolyte 3 and the second molten salt electrolyte 6 and conduction of the liquid alloy can be applied to the method of the present invention.
Fig. 2 is a flow chart of a method for processing uranium-containing material by molten salt electrolysis according to the present invention. Referring to fig. 1 and 2, the uranium-containing material 2 includes a plurality of materials such as crude metal uranium, uranium oxide, uranium carbide, uranium nitride, and uranium oxycarbide, where the crude metal uranium, the uranium oxide, and the uranium nitride may be directly obtained from spent fuel, and the crude metal uranium and uranium compounds may be synthesized from the uranium oxide by a molten salt electro-deoxidation method or a high-temperature thermal reduction method.
In the molten salt electrolytic tank (with a baffle plate) shown in fig. 2, the crude metal uranium or uranium compound is added into a current collector net bag, electrolysis is conducted, and a metal uranium 7 product is obtained at a cathode. And after multiple and long-time electrolysis, taking out and purifying the molten salt electrolyte or the liquid alloy when the metal impurities/fission products in the molten salt electrolyte or the liquid alloy are enriched to a certain concentration, so as to obtain various fission products with smaller volume and easy treatment. And returning the purified molten salt and the purified liquid alloy to the molten salt electrolysis process.
Fig. 2 is only one flow, and all the flows of treating uranium-containing materials by using the molten salt electrolysis tank disclosed by the invention through a molten salt electrolysis method are within the protection scope of the invention.
Example 1
Preparing crude metal uranium by a fused salt electro-deoxidation method: u is set to 3 O 8 Uniformly mixing with binder polyvinyl alcohol in a ball mill, tabletting, and sintering at 1200 deg.C under argon atmosphere for 12 hr to obtain U 3 O 8 Test pieces. In this way U 3 O 8 The test piece is used as a cathode, the metal platinum is used as an anode, and LiCl-Li is used as a cathode 2 O(Li 2 O2 wt%) molten salt is used as electrolyte, and crude metal uranium is obtained after constant-pressure electrolysis for 10 hours at 650 ℃.
The fused salt electrolysis method for treating the crude metal uranium comprises the following steps: in a molten salt electrolysis cell as shown in fig. 1, a first molten salt electrolyte is contained in a first chamber, and the first chamber comprises CaF with the molar ratio of 35:60:5 2 -NaF-UF 4 Is filled withA platinum wire mesh basket (anode current collector) of the coarse metal uranium is arranged in the first molten salt electrolyte; the second chamber contains a second molten salt electrolyte which is MgCl with the mol ratio of 15:82:3 2 -NaCl-UCl 3 And a platinum wire was inserted as a cathode. The lower chamber of the molten salt electrolytic tank is filled with a liquid alloy which comprises a U-Ni alloy with the U content of 77at percent, and the liquid alloy can ensure that the first molten salt electrolyte and the second molten salt electrolyte are not contacted. Electrifying and electrolyzing at 950 ℃ under the argon atmosphere condition, and controlling the initial cathode current density to be 0.01A/cm 2 And (5) after electrolysis for 12 hours, taking out a cathode metal uranium product. After analysis, the purity of uranium was 99.9%.
Example 2
Preparing uranium oxide by a carbothermal reduction method: u is set to 3 O 8 Uniformly mixing carbon powder and binder polyvinyl alcohol in a ball mill for 2 hours, tabletting, and sintering at 1200 ℃ for 8 hours under the vacuum degree of 10Pa to obtain UC 0.5 O 0.5 Test pieces.
The molten salt electrolysis method for treating uranium oxycarbide comprises the following steps: in a molten salt electrolytic cell as shown in fig. 1, a first molten salt electrolyte is contained in a first chamber, and the composition of the first molten salt electrolyte is NaF-KF-UF with the molar ratio of 90:4:6 4 Filled with UC as described above 0.5 O 0.5 Placing a stainless steel mesh basket (anode current collector) of the test piece in the first molten salt electrolyte; the second chamber contains a second molten salt electrolyte with a composition of NaF-KF-UF in a molar ratio of 90:4:6 4 And a tungsten rod was inserted as a cathode. The lower chamber of the molten salt electrolytic tank is filled with a liquid alloy which comprises a U-Fe-Co ternary alloy with the U content of 78at% and the Fe content of 15at%, and the liquid alloy can ensure that the first molten salt electrolyte and the second molten salt electrolyte are not contacted. Electrifying and electrolyzing at 1000 ℃ under the argon atmosphere condition, and controlling the initial cathode current density to be 1A/cm 2 And (5) after electrolysis for 6 hours, taking out a cathode metal uranium product. After analysis, the purity of uranium was 99.8%.
Example 3
Preparing uranium oxynitride by a carbothermal reduction method: the simulated spent fuel is mixed multi-metal oxide, wherein the metal element types and the proportion are as follows: 93% U,0.5% La,1% Ce,0.5% Pr,1% Nd,1% Cs1% Ba,1% Mo,1% Zr. Uniformly mixing the simulated spent fuel, uranium carbide and binder polyvinyl alcohol in a ball mill for 2 hours, tabletting, and sintering for 8 hours at 1200 ℃ in a nitrogen atmosphere to prepare nitrogen-doped uranium oxide, namely uranium oxynitride UC 0.5-0.5x O 0.5-0.5x N x Test pieces.
The molten salt electrolysis method for treating the uranium oxynitride carbon comprises the following steps: in a molten salt electrolytic cell as shown in fig. 1, a first molten salt electrolyte is contained in a first chamber, and the composition of the first molten salt electrolyte is KCl-LiCl-UCl with the molar ratio of 80:12:8 3 Filled with the uranium oxynitride UC 0.5-0.5x O 0.5-0.5x N x A platinum wire mesh basket (anode current collector) of the test piece is arranged in the first molten salt electrolyte; the second chamber contains a second molten salt electrolyte which comprises NaCl-KCl-UCl with the mol ratio of 46:48:6 4 And a zirconium rod was inserted as a cathode. The lower chamber of the molten salt electrolytic tank is filled with a liquid alloy which comprises a U-Mn alloy with the U content of 78.5at percent, and the liquid alloy can lead the first molten salt electrolyte and the second molten salt electrolyte not to be contacted. Electrifying and electrolyzing at 780 ℃ under the argon atmosphere condition, and controlling the initial cathode current density to be 1.5A/cm 2 And (5) after 10 hours of electrolysis, taking out a cathode metal uranium product. After analysis, the purity of uranium was 99.6%.
Example 4
Preparing crude metal uranium by a fused salt electro-deoxidation method: the simulated spent fuel is mixed multi-metal oxide, wherein the metal element types and the proportion are as follows: 93% U,0.5% La,1% Ce,0.5% Pr,1% Nd,1% Cs,1% Ba,1% Mo,1% Zr. Uniformly mixing the simulated spent fuel and the binder polyvinyl alcohol in a ball mill for 2 hours, tabletting, and sintering for 10 hours under the argon atmosphere at 1400 ℃ to prepare the simulated spent fuel test piece. The simulated spent fuel test piece is used as a cathode, graphite is used as an anode, and LiCl-Li is used 2 O(Li 2 O3 wt%) molten salt is used as electrolyte, and crude metal uranium is obtained after constant-pressure electrolysis for 8 hours at 700 ℃.
The fused salt electrolysis method for treating the crude metal uranium comprises the following steps: in the molten salt electrolytic tank shown in the attached figure 1, a first molten salt electrolyte is contained in a first chamber, and the composition of the first molten salt electrolyte is NaCl-UCl with the molar ratio of 90:10 3 Stainless steel filled with the above crude metal uraniumA steel mesh basket (anode current collector) is arranged in the first molten salt electrolyte; the second chamber contains a second molten salt electrolyte which is composed of KCl-UCl with the molar ratio of 93:7 3 And a molybdenum rod was inserted as a cathode. The lower chamber of the molten salt electrolytic tank is filled with a liquid alloy which comprises a U-Fe alloy with the U content of 70at percent, and the liquid alloy can ensure that the first molten salt electrolyte and the second molten salt electrolyte are not contacted. Electrifying and electrolyzing at 850 ℃ under the argon atmosphere condition, and controlling the initial cathode current density to be 0.1A/cm 2 And (5) after 10 hours of electrolysis, taking out a cathode metal uranium product. After analysis, the purity of uranium was 99.7%.
And supplementing the crude metal uranium into the stainless steel mesh basket, carrying out second batch of molten salt electrolysis in the same molten salt electrolysis tank under the same conditions, and taking out a cathode metal uranium product after 10 hours of electrolysis. After analysis, the purity of uranium was 99.7%.
And supplementing the crude metal uranium into the stainless steel mesh basket, carrying out third batch molten salt electrolysis in the same molten salt electrolysis tank under the same conditions, and taking out a cathode metal uranium product after 12h of electrolysis. After analysis, the purity of uranium was 99.6%.
Comparative example
The crude metallic uranium prepared in example 4 was used as a raw material for molten salt electrolysis.
The comparative example adopts a common fused salt electrolytic tank (namely a single-chamber electrolytic tank without liquid alloy), and the fused salt is KCl-UCl with the molar ratio of 93:7 3 The stainless steel mesh basket (anode current collector) filled with the coarse metal uranium is placed in molten salt, and the molybdenum rod cathode is inserted into the molten salt. Electrifying and electrolyzing at 850 ℃ under the argon atmosphere condition, and controlling the initial cathode current density to be 0.1A/cm 2 And (5) after 10 hours of electrolysis, taking out a cathode metal uranium product. After analysis, the purity of uranium was 99.5%.
And supplementing the crude metal uranium into the stainless steel mesh hanging basket, carrying out second batch molten salt electrolysis in a single-chamber electrolytic tank under the same conditions, and taking out a cathode metal uranium product after 10 hours of electrolysis. After analysis, the purity of uranium was 99.2%.
And supplementing the crude metal uranium into the stainless steel mesh hanging basket, carrying out third batch molten salt electrolysis in a single-chamber electrolytic tank under the same conditions, and taking out a cathode metal uranium product after 12 hours of electrolysis. After analysis, the purity of uranium was 98.5%.
The foregoing is merely illustrative of the present invention, and the present invention is not limited thereto, and any person skilled in the art will readily recognize that variations or substitutions are within the scope of the present invention.
Claims (10)
1. A molten salt electrolysis method for treating uranium-containing materials, which is characterized by being suitable for a molten salt electrolysis tank, wherein the molten salt electrolysis tank comprises an electrolysis tank body and an insulating partition plate; the electrolytic tank body is provided with an upper chamber and a lower chamber, and an insulating partition plate is arranged in the upper chamber so as to divide the upper chamber into a first chamber and a second chamber;
the lower chamber is filled with liquid alloy, the first chamber is filled with a first molten salt electrolyte, and an anode is arranged in the first molten salt electrolyte; a second molten salt electrolyte is contained in the second chamber, and a cathode is arranged in the second molten salt electrolyte; the first molten salt electrolyte and the second molten salt electrolyte are separated by the insulating separator, both of which are in contact with the liquid alloy; wherein the anode comprises uranium-containing material; the liquid alloy is an alloy of uranium metal and auxiliary metal;
the method comprises the following steps: and connecting the anode with the positive electrode of the power supply, connecting the cathode with the negative electrode of the power supply, and electrifying for electrolysis to enable the anode to perform oxidation reaction and generate uranium ions, and performing reduction reaction on the surface of the cathode to obtain a metal uranium product.
2. The molten salt electrolysis process of claim 1, wherein the uranium containing material is crude metallic uranium or/and uranium compounds.
3. The molten salt electrolysis process of claim 2, wherein the uranium compound is composed of uranium and nonmetallic elements including one or more of oxygen, carbon, and nitrogen.
4. A molten salt electrolysis process for the treatment of a uranium containing material according to claim 3, wherein the uranium compound is one or more of uranium oxide, uranium carbide, uranium nitride, uranium oxycarbide, uranium oxynitride, uranium carbonitride, uranium oxycarbonitride.
5. The molten salt electrolysis process for the treatment of uranium containing material according to any one of claims 2 to 4, wherein the crude metallic uranium or uranium compound is produced from uranium oxide by a molten salt electro-deoxidation process or a high temperature thermal reduction process.
6. The molten salt electrolysis method for treating a uranium containing material according to claim 5, wherein the molten salt electro-deoxidation method is carried out by using uranium oxide as a cathode or a mixture of uranium oxide and uranium carbide or/and carbon powder as a cathode, and LiCl-Li 2 O-based fused salt or CaCl 2 The CaO-based molten salt is taken as electrolyte, an inert anode or a graphite anode is adopted, electrolysis is carried out at the temperature of 650-1000 ℃, and the cathode product is crude metal uranium or/and uranium oxycarbide;
the high-temperature thermal reduction method comprises metallothermic reduction and carbothermic reduction, wherein one or more of active metals lithium, sodium, calcium and magnesium are used as reducing agents, or carbon powder or/and uranium carbide are used as reducing agents, and uranium oxide is reduced at the temperature of more than 500 ℃ in a non-oxidizing atmosphere to prepare crude metal uranium or uranium compounds.
7. A molten salt electrolysis method for processing a uranium containing material according to any of claims 1 to 4, wherein the first molten salt electrolyte or the second molten salt electrolyte is composed of alkali metal halides or/and alkaline earth metal halides and uranium dissolved halides.
8. Molten salt electrolysis process for the treatment of uranium containing material according to claim 7A method characterized in that the alkali metal halide is one or more of LiCl, naCl, KCl, rbCl, csCl, liF, naF, KF, rbF, csF; the alkaline earth metal halide is MgCl 2 、CaCl 2 、SrCl 2 、BaCl 2 、MgF 2 、CaF 2 、SrF 2 、BaF 2 One or more of the following; the halide of uranium is UCl n Or/and UF n (3≤n≤6)。
9. A molten salt electrolysis process for the treatment of a uranium containing material according to any of claims 1 to 4, wherein the auxiliary metal includes one or more of Fe, co, ni, cr, mn.
10. The molten salt electrolysis method of treating a uranium containing material according to any one of claims 1 to 4, wherein an operating temperature within the molten salt electrolysis cell is 700 to 1000 ℃; the cathode current density is 0.01-2.0A/cm 2 Or controlling the anode current density to be 0.01-1.5A/cm 2 。
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