RU2560119C1 - Waste nuclear fuel processing method - Google Patents

Abstract

FIELD: power industry.

SUBSTANCE: in the offered method the tablets of oxidic spent nuclear fuel destroyed at cutting of fuel elements are dissolved with heating in water solution of iron nitrate (III) at the molar ration of iron to uranium in the fuel 1.5-2.0:1, the formed deposit of the hydroxy iron salt with undissolved nuclear fuel fission products are separated by filtering, and from the obtained subacidic solution the uranyl peroxide is precipitated by consecutive adding to the solution with mixing of disodium salt of ethylene diamine tetraacetic acid. Then the obtained heterogeneous system is held at least 30 minutes and after separation washing with acid and water the uranyl peroxide deposit of peroxide is subjected to solid-phase recovery at heating by processing by its alkaline solution of a hydrazine hydrate in water at double-triple molar excess of hydrazine with reference to uranium, with subsequent separation of obtained hydrated uranium dioxide UO₂·2H₂O, its washing by HNO₃ solution with the concentration 0,1 mol/l, water and drying. While the deposit of the hydroxy iron salts with fission products, mother solution of the stage of sedimentation of peroxides from residue of fission products, waste of alkaline and washing solutions are supplied to the waste collector for their subsequent processing.

EFFECT: improvement of environmental safety and decrease of waste amount.

9 cl

Description

The invention relates to the field of nuclear energy, in particular to the processing of spent nuclear fuel (SNF), and can be used in the technological scheme of processing, including MOX fuel, since the extraction of the remaining amounts of U and Pu from SNF for the preparation of new fuel is the main the goal of a closed nuclear fuel cycle, which the country's nuclear power industry is oriented towards. Currently, it is relevant to create and optimize new, low-waste, environmentally friendly and economically feasible technologies that would ensure reprocessing of spent nuclear fuel from both operating and generation 3 and 4 reactors using fast neutrons operating on mixed oxide uranium-plutonium fuel (MOX fuel )

Known methods for processing spent nuclear fuel using fluorine or fluorine-containing chemical compounds. The resulting volatile fluoride compounds of the components of the nuclear fuel go into the gas phase and are distilled off. During fluorination, uranium dioxide is converted to UF ₆ , which vaporizes relatively easily, unlike plutonium, which has lower volatility. Typically, in the reprocessing of spent nuclear fuel by this method, spent nuclear fuel is fluorinated, not all uranium contained in it is extracted, but only its necessary amount, thereby separating it from the rest of the reprocessing fuel. After that, the evaporation regime is changed and SNF is also removed from the residual in the form of vapors and a certain amount of plutonium contained in it.

[RF patent No. 2230130, C22B 60/02, publ. 01/19/1976]

The disadvantage of this technology is that in this method of SNF processing, gaseous, aggressive and environmentally toxic chemical compounds are used. Thus, the technology is environmentally unsafe.

One of the close in fact to the claimed method is a known method, as claimed in US Pat. RF №2403634, (G21C 19/44, publ. 10.11.2010), according to which SNF regeneration includes the stage of fuel dissolution in nitric acid solution, the stage of electrolytic regulation of valency, with the restoration of Pu to trivalent state and maintaining the pentavalent state of Np, the stage of hexavalent extraction uranium extracting agent in an organic solvent; a precipitation step with oxalic acid, leading to the co-precipitation of minor actinides and fission products remaining in the nitric acid solution as an oxalate precipitate; a chlorination step to convert the oxalate precipitate into chlorides by adding hydrochloric acid to the oxalate precipitate; a dehydration step to produce synthetic anhydrous chlorides by dehydration of chlorides in a stream of argon gas; and the step of electrolysis in a molten salt with the dissolution of anhydrous chlorides in the molten salt and the accumulation of uranium, plutonium and minor actinides at the cathode due to electrolysis.

The disadvantage of this method of SNF processing is its multi-stage nature and complexity in implementation, since it includes electrochemical stages that are energy-intensive, require special equipment and the process at high temperature, especially when working with salt melts.

There is also a method according to which spent nuclear fuel is processed purely pyrochemically using a molten salt of uranium or plutonium, after which the separated components of nuclear fuel are reused. When pyrochemical reprocessing of spent nuclear fuel, its induction heating in the crucible and its cooling by supplying refrigerant to the crucible are used.

[RF patent No. 2226725, G21C 19/46, publ. 01/19/2009]

Pyrometallurgical technologies do not lead to the formation of large quantities of liquid radioactive waste (LRW), and also provide compact equipment placement, however, they are very energy-intensive and technologically complex.

Also spent fuel reprocessing methods include:

(1) a method comprising oxidizing uranium with gaseous chlorine, nitrogen oxides, sulfur dioxide in a dipolar aprotic solvent or a mixture thereof with a chlorine-containing compound [RF patent No. 2238600, G21F 9/28, publ. 04/27/2004];

(2) a method for dissolving materials containing metallic uranium, comprising oxidizing metallic uranium with a tributyl phosphate-kerosene mixture containing nitric acid [US Pat. No. 3,288,568, G21F 9/28, publ. 12/10/1966];

(3) a method for dissolving uranium, comprising oxidizing metallic uranium with a solution of bromine in ethyl acetate when heated [Larsen R.P. Dissolution of uranium metal and its alloys // Analit. chem. V. 31. Issue 4. P. 545-549].

The disadvantages of these methods include increased fire hazard systems and the limited scope of their use.

A widespread SNF reprocessing technology is the Purex process (which we took as a prototype), in which SNF containing uranium, plutonium, and fission products (PD) of nuclear fuel are dissolved in strongly acidic solutions of nitric acid when heated to 60-80 ° C. After that, actinides are removed from the nitric acid solution by the organic phase containing tributyl phosphate in kerosene or another organic solvent. The following are the technological stages associated with the separation of uranium and plutonium and their purification from PD. The Purex process is described, for example, in the monograph The Chemistry of the Actinide and Transactinide Elements, 3rd Edition, Edited by Lester R. Morss, Norman M. Edelstein and Jean Fuger. 2006, Springer, pp. 841-844.

The specified SNF reprocessing process is multi-stage and is based on the use of environmentally hazardous environments:

(1) nitric acid (6-8 mol / L) as a solvent for spent nuclear fuel at 60-80 ° C and forming aggressive gaseous products during reactions with its participation;

(2) since the acidity of the solution after completion of the dissolution is approximately 3.5 mol / L with respect to nitric acid, this inevitably leads to the use of extraction to extract U (Pu) with organic solvents;

(3) the use of organic solvents that are toxic, combustible, flammable, explosive and often unstable to radiation leads to the formation of large amounts of waste together with aqueous LRW (up to 7-12 tons per 1 ton of reprocessed SNF).

The present invention is the creation of an innovative, low-waste, environmentally friendly and economically feasible technology for reprocessing spent nuclear fuel.

The problem is solved using a new method for processing spent nuclear fuel, characterized in that the oxide spent nuclear fuel pellets destroyed during the fueling of fuel elements are dissolved during heating in an aqueous solution of iron (III) nitrate with a molar ratio of iron to uranium in the fuel equal to 1.5-2, 0: 1, the precipitate of the basic salt of iron with undissolved fission products of nuclear fuel is separated by filtration, and a feather is precipitated from the obtained weakly acidic solution, containing mainly uranyl nitrate uranyl oxide by sequentially feeding ethylene diamine tetraacetic acid disodium salt into a solution with stirring in a molar excess of uranium equal to 10% and a 30% hydrogen peroxide solution taken in a 1.5-2-fold molar excess of uranium at a temperature not higher than 20 ° C, the resulting heterogeneous system is maintained for at least 30 minutes, and after separation and washing with acid and water, the precipitate of uranyl peroxide is subjected to solid-phase reduction by heating by treating it with an alkaline solution of hydrazine hydrate in water e with a 2-3-fold molar excess of hydrazine with respect to uranium, followed by separation of the obtained hydrated uranium dioxide UO ₂ · 2H ₂ O, washing it with HNO ₃ solution with a concentration of 0.1 mol / l, water and drying, while the precipitate basic salts of iron with fission products, the mother liquor of the precipitation stage of peroxides with residues of fission products, waste alkaline and wash solutions are sent to the waste collection for subsequent processing.

Typically, SNF is dissolved in a temperature range of 60-90 ° C for no more than 5-10 hours using an aqueous solution of iron (III) nitrate with a pH of 0.2 to 1.0.

The separated uranyl peroxide should be washed with a 0.05 mol / L HNO ₃ solution, and its solid-phase reduction should be carried out with a 10% aqueous hydrazine hydrate solution at pH 10 at 60-90 ° C for 10-15 hours.

Advantageously, hydrated uranium dioxide is dried at 60-90 ° C.

It is possible to conduct the process in two bifunctional devices connected in series, the design of which provides for a filtration unit and the possibility of changing the spatial orientation of the devices by 180 °, the first of which is used to dissolve and collect the waste of the process, and the second to precipitate uranium peroxide, its solid-phase reduction and isolation target product.

The technical result of the method is achieved by the fact that at all stages of SNF reprocessing, the fuel components (UO ₂ with a content of up to 5 wt.% ²³⁹ Pu) - U (Pu), solvent (iron nitrate), precipitating (hydrogen peroxide) and reducing reagents are in different phases convenient for their further separation. At the stage of dissolution, uranium passes into solution, and the bulk of the solvent reagent is released in the form of a solid compound. At the stage of deposition of peroxide and its solid-phase reductive conversion into uranium dioxide, the target product is in solid form and is easily separated from the liquid phase.

The proposed method is as follows.

Uranium dioxide tablets (UO ₂ with a content of up to 5 wt.% ²³⁹ Pu) destroyed during fueling of fuel elements are immersed in water containing iron (III) nitrate and dissolved by heating to 60-90 ° C. The resulting solution containing U (Pu) and the pulp of the basic iron salt formed upon dissolution are separated. After removing the solution from U (Pu), the precipitate of the main iron salt — the iron salt with PD — Mo, Tc, and Ru (~ 95%) and partially Nd, Zr, and Pd (~ 50%) — remains in the waste collection.

Hydrogen peroxide is added to the separated solution with U (Pu), and at room temperature, uranyl peroxide is precipitated, with plutonium also precipitating. The coefficient of purification of the target product from PD is about 1000. After completing the precipitation, the precipitate is separated from the slightly acidic mother liquor, which remains in it PD and Fe (III) nitrate are sent to the waste collector with a basic salt precipitate. The solution from washing the sludge of mixed peroxide is also sent to the waste collector. Next, solid-phase reduction of the formed peroxide is carried out after the introduction of hydrazine hydrate with stirring with a stream of nitrogen at 80-90 ° C and hydrated U (Pu) dioxide is obtained. The separated alkaline solution is transported to a waste collector. The dioxide precipitate is washed with a small volume of 0.1 M HNO ₃ , then with distilled water, which is also sent to the waste collection. The obtained target product is dried in a stream of heated nitrogen at 60-90 ° C and discharged from the apparatus.

Weakly acidic and slightly alkaline aqueous waste solutions that are collected as SNF is processed in the waste collector are removed by evaporation, and the iron in them is precipitated in the form of hydroxide together with cations of 2-, 3- and 4-valent PD. A solid product from iron compounds with PD included in their phase is the only waste in the inventive method for processing spent nuclear fuel. Evaporated water can be condensed and returned if necessary to the process.

SNF can be processed in a bifunctional special apparatus (apparatus), the design of which provides for a filtration unit (UV), a shirt capable of supplying a coolant and carrying out the dissolution process at a temperature of ≤90 ° C in the reaction mixture, and the ability to change the spatial orientation by 180 ° apparatus.

The process is conducted, as a rule, in two series-connected bifunctional devices as follows.

When the filtration unit of the device is located in the upper part, the apparatus is designed to dissolve SNF. The resulting solution containing U (Pu) and the pulp of the basic iron salt formed during the dissolution of SNF are separated. To do this, the device is turned 180 °, while the UV is in the bottom. Filtration is carried out by applying excess pressure to the internal volume of the apparatus, or by connecting it to a vacuum line. After filtering and removing the solution from U (Pu), a device with a precipitate of an iron salt and PD (Mo, Tc, and Ru (~ 95%) and partially Nd, Zr, and Pd (~ 50%)) is rotated through 180 ° to the position where UV is located in the upper part, and then the device performs the function of a collection of waste solutions.

The filtered solution with U (Pu) is fed into a second apparatus of the same design at the position when the UV is located at the top of the device. Hydrogen peroxide was added to the solution, and U (Pu) peroxide was precipitated at room temperature. After completing the deposition, the device is turned 180 ° and filtration separation is carried out through the bottom of the apparatus. The resulting peroxide remains on the filter in the apparatus, and the mother liquor with dissolved PD (purification factor of about 1000) and residual Fe (III) nitrate is sent to the first apparatus with the precipitation of the basic salt, which has become a waste collector.

The device is turned into the UV position in the upper part and the peroxide precipitate from the filter in the apparatus is washed off with a small amount of hydrazine hydrate-containing water to form a pulp in which the peroxide is converted by hydrazine by solid-phase reduction into hydrated U (Pu) dioxide at 80-90 ° C.

After completing solid-phase reduction and obtaining hydrated dioxide U (Pu), the apparatus is transferred to a position in which it performs the filtering function. The separated alkaline solution is sent to the first apparatus with a precipitate of basic salt, which has become a waste collector. The dioxide precipitate is washed with a small volume of 0.1 M HNO ₃ , then with distilled water, which is also sent to the waste collection. A device with a precipitate of hydrated U (Pu) O ₂ · nH ₂ O is rotated 180 ° to the position when the UV is located in the upper part. Next, the apparatus is dried the target product at 60-90 ° C by applying a stream of nitrogen, and upon completion of drying, the drug is unloaded from the apparatus.

The following examples illustrate the effectiveness of using aqueous weakly acidic solutions of Fe (III) nitrate (chloride) for dissolving oxide SNF with simultaneous separation of U (Pu) at this stage from a portion of the PD, followed by their separation from the residues of PD during peroxide deposition of U (Pu) from the resulting solution . Further solid-phase reductive conversion of the peroxide first to hydrated, and then to crystalline dioxide U (Pu) increases the efficiency of the proposed method.

Example

A powdered sample of uranium dioxide ( ^{238 + 235} UO ₂ ) was preliminarily calcined at 850 ° C in an argon atmosphere with 20% hydrogen for 8 hours.

A tablet or powder of ceramic nuclear fuel containing uranium and 5 wt.% Plutonium weighing 132 g is immersed in an aqueous solution of iron (III) nitrate with a volume of 1 l with a pH of at least 0.2 at a concentration of Fe (NO ₃ ) ₃ in water from 50 up to 300 g / l and dissolved by heating to 60-90 ° C with a molar ratio of Fe (III) to fuel as 1.5 to 1.

The pH and the uranium content in the solution are monitored and the tablets continue to dissolve until the uranium content in the successive samples is changed. As a result of the dissolution process, a solution is obtained containing predominantly uranyl nitrate and having a pH value of ≤2, and a precipitate of the basic iron salt. It takes no more than 5-7 hours for the quantitative dissolution of the samples taken.

The resulting nitrate solution is separated from the pulp by filtration, for example, using a cermet filter. The precipitate of the basic iron salt remaining on the filter is washed with water and sent to the waste collector along with the wash water.

To a weakly acid solution of separated uranyl nitrate at a temperature of ≤20 ° C add 60 ml of a 10% solution of the disubstituted sodium salt of EDTA (Trilon-B), mix for 10 minutes. A complex compound of white uranyl precipitates in solution.

While stirring, to the resulting suspension, add 50 ml portions at intervals of 1-1.5 min, 300 ml of a 30% solution of hydrogen peroxide (H ₂ O ₂ ) also at a temperature of ≤20 ° C to obtain uranyl peroxide, with which also quantitatively plutonium co-precipitates.

Stirring is then carried out for at least 20-30 minutes to better ripen the precipitate of uranyl peroxide, which has a light yellow color.

The precipitate of uranyl peroxide is separated by filtration from the mother liquor, which is sent to a waste collector. The precipitate is washed with 0.25 L of 0.05 M HNO ₃ , the washing solution is sent to the waste collection.

The washed precipitate of uranyl peroxide is first transferred into a suspension with a 10% aqueous alkaline solution of hydrazine hydrate in water, and the solution has a pH value of ~ 10.

With stirring and heating the suspension to 80 ° C, uranyl peroxide transforms into hydrated dioxide UO ₂ · H ₂ O during solid-phase reduction of U (VI) with hydrazine to U (IV).

The process of reducing U (VI) to U (IV) is monitored by periodic sampling of a suspension containing not more than 50 mg of solid suspension. The precipitate was dissolved in a mixture of 4M HCl with 0.1M HF, and the first spectrum of the solution was recorded. Then the solution is treated with an amalgam and the second spectrum of this solution is recorded. In this case, all uranium in solution must be completely reduced to U (IV). Thus, if the first and second spectra coincide, then the solid-phase reduction process is completed. Otherwise, the conversion of peroxide to uranium dioxide is continued. The process is completed in 10-15 hours.

The obtained hydrated uranium dioxide is separated by filtration from an alkaline solution (volume ~ 0.6 l), the solution is sent to the waste collection. The precipitate of hydrated uranium dioxide is washed on a filter with 0.25 L of 0.1 M HNO ₃ to neutralize the alkali remaining in the volume of the precipitate, then with the same volume of water to remove traces of acid from the volume of the precipitate, controlling the pH of the last wash water. Wash solutions are sent to the waste collection.

The results of analyzes of the mother liquor and uranium peroxide indicate that the degree of deposition of uranium is not less than 99.5%, and the iron content in the selected peroxide does not exceed 0.02 wt.%.

Washed from traces of alkali, the precipitate of uranium peroxide is dried, for example, a stream of nitrogen heated to 60-90 ° C and discharged from the apparatus in the form of a powder.

The result is at least 131.3 g of uranium dioxide.

In aqueous solutions collected in the waste collection with a slightly alkaline reaction, iron residues in the form of amorphous hydroxide are released. The heterogeneous suspension is evaporated, while almost complete removal of water is achieved. A wet or dry solid product, which is mainly iron compounds, is the only waste in the inventive method for processing ceramic oxide fuel using solutions of iron (III) nitrate.

The inventive method allows to simplify the reprocessing of spent nuclear fuel and to exclude the formation of LRW in comparison with the Purex process.

New significant and distinctive features of the proposed method (in comparison with the prototype) are:

- the use of aqueous weakly acidic solutions of Fe (III) nitrate to dissolve oxide SNF, which were not previously used for this. Without significant deterioration in the dissolving power, iron nitrate can be replaced with Fe (III) chloride;

- unlike the prototype, there is no special stage with the introduction of ferrous sulfate into the system to reduce Pu (IV) to Pu (III). In the inventive method, upon dissolution of oxide uranium and mixed fuel, uranium (IV) oxidizes Fe (III) to uranium (VI), and the Fe (II) cations formed in this process reduce Pu (IV) to Pu (III), and the actinides quantitatively transform into a solution in the form of their nitrates;

- in the inventive method, it is not necessary to introduce acid to dissolve SNF, since the medium used has an acidity due to the hydrolysis of iron (III) nitrate, and depending on its concentration from 50 to 300 g / l, the pH in the range from 1 to 0.3 ;

- in the inventive method, after dissolving the fuel, the acidity of the resulting solutions will be ≤0.1 M (for uranium 100-300 g / l), while in the Purex process strongly acidic ~ 3M HNO ₃ solutions are formed, which inevitably leads to extraction and the formation of a large number of organic and aqueous LRW;

- low acidity after the SNF is dissolved by the present method allows one to abandon the extraction extraction of fuel components with organic solutions, simplify the organization of the SNF reprocessing process and eliminate LRW in comparison with the Purex process technology;

- in the inventive method, the process of dissolving the fuel is completed by obtaining a solution containing U (Pu) and a precipitate of the basic iron salt, in an amount of ~ 50% of the initial content of iron (III) nitrate;

- fission products, such as Mo, Tc and Ru (~ 95%) and partially from Nd, Zr and Pd (~ 50%), are separated from uranium already at the stage of SNF dissolution and are concentrated in the resulting precipitate of the basic iron salt. This is also an advantage of the proposed method for dissolving spent nuclear fuel in comparison with the Purex process;

- structural materials of fuel cladding shells and phases formed from PD in the SNF matrix in the form of light metal (Ru, Rh, Mo, Tc, Nb) and gray ceramic inclusions (Rb, Cs, Ba, Zr, Mo) do not dissolve in the slightly acidic solutions used; . Therefore, slightly acidic will be less contaminated with the components of the dissolved shells and PD, in contrast to 6-8 M HNO ₃ in the Purex process;

- acidity ≤0.1 M of the resulting solutions with a uranium concentration of 100-300 g / l is optimal for the precipitation of uranium (VI) and plutonium (IV) peroxides. Hydrogen peroxide is preferred, since it transfers uranium to the U (VI) state, which is required for quantitative precipitation;

- when U (Pu) peroxide is precipitated from solution, quantitative separation of U from almost all PD and iron residues in solution is achieved (purification coefficient ~ 1000);

- a new and original solution in the claimed method is the process of solid-phase reduction in an aqueous suspension of U (Pu) peroxide with hydrazine hydrate at 90 ° C to hydrated U (Pu) O ₂ × nH ₂ O, followed by drying of the target product at 60-90 ° C and unloading from the apparatus,

- weakly acidic and slightly alkaline aqueous solutions, the waste accumulated as SNF is processed in the waste collection is removed by evaporation, and the iron in it is deposited in the form of hydroxide together with cations of 2-, 3- and 4-valent PD. A solid product from iron compounds with PD included in their phase is the only waste in the inventive method for processing oxide SNF.

Claims (9)

1. A method of processing spent nuclear fuel, characterized in that the oxide spent nuclear fuel pellets destroyed during the fueling of fuel elements are subjected to dissolution when heated in an aqueous solution of iron (III) nitrate with a molar ratio of iron to uranium in the fuel equal to 1.5-2.0 : 1, the precipitate of the basic iron salt with undissolved fission products of nuclear fuel is separated by filtration, and uranyl peroxide is precipitated from the obtained weakly acidic solution containing predominantly uranyl nitrate after an oleaginous feed into the solution with stirring of disodium salt of ethylenediaminetetraacetic acid in a molar excess with respect to uranium equal to 10% and a 30% solution of hydrogen peroxide taken in a 1.5-2-fold molar excess with respect to uranium at a temperature of no higher than 20 ° C, the resulting heterogeneous system is kept for at least 30 minutes and after separation and washing with acid and water, the precipitate of uranyl peroxide is subjected to solid-phase reduction by heating by treating it with an alkaline solution of hydrazine hydrate in water at a 2–3 molar a surplus of hydrazine with respect to uranium, followed by separation of the obtained hydrated uranium dioxide UO ₂ · 2H ₂ O, washing it with HNO ₃ solution with a concentration of 0.1 mol / l, water and drying, while the precipitate of basic salts of iron with fission products, mother the solution of the precipitation stage of peroxides with the remains of fission products, the waste of alkaline and wash solutions are sent to the waste collection for subsequent processing. 2. A method of processing spent nuclear fuel according to claim 1, characterized in that the dissolution of spent nuclear fuel is carried out at 60-90 ° C. 3. A method of processing spent nuclear fuel according to claim 1, characterized in that an aqueous solution of iron (III) nitrate with a pH from 0.2 to 1.0 is used to dissolve the fuel. 4. The method of processing spent nuclear fuel according to claim 1, characterized in that the dissolution of spent nuclear fuel is not more than 5-10 hours. 5. A method of processing spent nuclear fuel according to claim 1, characterized in that the precipitate of uranyl peroxide is washed with a solution of HNO ₃ with a concentration of 0.05 mol / L. 6. A method of processing spent nuclear fuel according to claim 1, characterized in that the solid-phase reduction is carried out with a 10% aqueous solution of hydrazine hydrate at pH 10. 7. A method of processing spent nuclear fuel according to claim 1, characterized in that the solid-phase reduction is carried out at 60-90 ° C for 10-15 hours. 8. The method of processing spent nuclear fuel according to claim 1, characterized in that the drying of hydrated uranium dioxide is carried out at 60-90 ° C. 9. A method of processing spent nuclear fuel according to any one of paragraphs. 1-8, characterized in that the process is conducted in two series-connected bifunctional devices, the design of which provides for a filtration unit and the possibility of changing the spatial orientation of the devices by 180 °, the first of which is used to dissolve and collect waste from the process, and the second to deposit peroxide uranyl, its solid-phase reduction and isolation of the target product.