RU2562809C1 - Method of regeneration of nuclear fuel powders from fuel elements and dispersion compositions based on aluminium and aluminium alloys - Google Patents

Abstract

FIELD: power engineering.

SUBSTANCE: invention is designed for the process of extraction of powder particles of nuclear fuel during processing of inferior and unradiated fuel elements of dispersion type with shells and core matrix from aluminium or its alloys at the stage of their manufacturing. The process of dissolution of inferior fuel elements and wastes from aluminium and/or its alloys is carried out in caustic soda solution at temperature of 15-75°C with caustic soda concentration of up to 30%. Separated and washed powder of nuclear fuel is treated in nitric acid solution with concentration of 7-25% for 2-60 seconds at temperature of 15-40°C, washed in deionised water at temperature of 15-95°C, dried at temperature of up to 165°C with residual pressure of less than 2 kPa for more than 5 minutes, and mechanical treatment is carried out in atmosphere of inert gas or nitrogen on sieves with cell size equal to and/or less than average maximum permissible and equal to and/or more than average minimum permissible size of powder particles.

EFFECT: possibility to repeatedly use regenerated powder of nuclear fuel in manufacture of fuel elements.

4 cl, 1 tbl, 5 dwg

Description

The invention relates to the nuclear industry and is intended for the process of extracting powdered particles of nuclear fuel during the processing of substandard and non-irradiated dispersion-type fuel elements with shells and a core matrix of aluminum or its alloys at the stage of their manufacture, as well as waste in the form of chips, scraps, bast marks in the manufacturing process of such cores and fuel elements.

The recoverable nuclear fuel particles are powder or granules from compounds of uranium or its alloys. Most often, the material of the dispersed nuclear fuel particles is uranium dioxide, which has recently been replaced by uranium silicides and uranium-molybdenum alloys in most of the dispersion-type fuel elements for research reactors.

U-9% MO alloy nuclear fuel powder has been used in the fuel composition in fuel elements (fuel elements) for the reactor of the first nuclear power plant since 1954, U ₃ Si ₂ compounds have been used in plate-type dispersion fuel elements for research reactors for more than 20 years, however, data on the extraction of nuclear fuel particles of this type from non-irradiated and substandard fuel rods at the stage of their manufacture for reuse, as well as the results of the operation of fuel rods with re-extracted nuclear fuel, are absent. The information sources contain data on the processing of already irradiated fuel elements with nuclear fuel of this type. According to these data, the main processing processes are aimed at separating the irradiated fuel particles from the cladding and matrix of the fuel elements and their further chemical processing. The processes of extraction and processing of nuclear fuel powder given in the literature, as a rule, do not provide for its reuse. The nuclear fuel powder, separated from the shells and the fuel matrix of aluminum and its alloys, enters a chemical redistribution, including further dissolution and subsequent extraction purification of the solutions, followed by separation of uranium from fission fragments and impurities, with the repetition of a complete redistribution of both chemical and metallurgical obtaining again nuclear fuel powder in the form of uranium-based alloy particles.

The reduction of the stages of the technological process for the processing of substandard fuel elements and waste is of great economic importance due to the high cost of both production and uranium itself.

A known method of separating an aluminum alloy shell from a core of uranium or a uranium-molybdenum alloy by electrochemical dissolution of a fuel cladding in 7-12 normal nitric acid, where a core of uranium or a uranium alloy is used as a cathode (see UK patent No. 1129698 from 09.10. 1968 IPC C23B 3/02).

The disadvantage of this technology is the impossibility of using it to extract dispersed fuel in the form of powder or granules from fuel rods and their waste, since, as a result of electrochemical dissolution, in addition to dissolving the cladding, the fuel core or fuel particles dissolve over unequal dissolution times.

Known technologies for removing shells from aluminum and its alloys by dissolving them in an aqueous solution of sodium hydroxide without additives or with the addition of sodium nitrate or nitrite at temperatures up to 100 ° C, followed by separation of the resulting solution from the fuel. When using alkaline solutions with a concentration of up to 30% NaOH, the losses of uranium and its alloys, such as U-Mo and U-Si, are insignificant, which leads to the use of this method in the processing of irradiated fuel elements with a shell made of aluminum and aluminum alloys.

The main reactions of this process are:

2Al + 2NaOH + 2H ₂ O → 2NaAlO ₂ + 3H ₂

2Al + NaOH + NaNO ₂ + H ₂ O → 2NaAlO ₂ + NH ₃

Using a large excess of sodium nitrate, you can avoid the release of ammonia:

2Al + 2NaOH + 3NaNO ₃ → 2NaAlO ₂ + H ₂ O + 3NaNO ₂

It was found that the emission of explosive gases is minimal when the process of dissolution of aluminum proceeds according to the reaction:

20Al + 17NaOH + 21NaNO ₃ → 20NaAlO ₂ + 18NaNO ₂ + 3NH ₃ + 4H ₂ O

(see. Processing of nuclear fuel. Edited by S. Stoler and R. Richards. M., Atomizdat 1964 p. 42;; Ageenkov A.T., Nenarozhomov E.A., Savelyev V.F., Yastrebov A.B. Preparation of irradiated nuclear fuel for chemical processing. M., Energoatomizdat, 1982, pp. 44-45).

The sodium aluminate formed is an unstable compound and may precipitate. In order to prevent this from happening, the molar ratio in the solution of caustic soda and aluminum is 1.65: 1. A thin diffusion layer of the interaction between uranium and aluminum formed on the surface of particles during temperature exposure during the manufacture of a fuel rod can partially dissolve in sodium hydroxide solution, forming uranium oxide on the surface.

These methods are developed and used for the processing of irradiated fuel rods with a shell of aluminum or its alloys in order to separate aluminum with impurities from nuclear fuel, fission fragments and radioactive decay products obtained in the process of using fuel as a part of fuel rods in a nuclear reactor, followed by processing of spent fuel and do not provide for the reuse of fuel unchanged, and, therefore, do not provide for the preparation of its surface.

A known method of processing the surface of uranium to prepare it for plating, ensuring surface uniformity, the absence of oxide and acid films (see US patent No. 2836548 from 05/28/1958, IPC B21 3/16). The method includes treating the surface of uranium for one minute in an aqueous solution of 40-55% nitric acid, subsequent processing with a 30-40% hydrochloric acid solution until it appears black on the surface, with repeated processing in a 40-55% nitric acid solution for several seconds until black is eliminated and the surface of the uranium is grayed out. This method is not acceptable for processing compositions containing molybdenum, that is, uranium-molybdenum, in the composition of fuel alloys, since both molybdenum and uranium are actively dissolved in a solution of hydrochloric acid. Therefore, the surface of the fuel particles during surface treatment in this way is corroded, becomes loose, highly developed, which is unacceptable.

The closest in technical essence and the achieved result to the claimed method of extracting nuclear fuel powder from a dispersion composition for reuse in the manufacture of fuel elements is the method implemented in the technology of dissolving fuel elements and their waste from aluminum and its alloys in 10% sodium hydroxide solution at temperature 75-100 ° C with the separation of fuel from the solution, washing the separated fuel with water until neutral, drying the fuel in air at a temperature of 25 ° C for 24 hours or in an argon stream at a temperature of 40-50 ° C for 2-3 hours and with fuel quality control (see Gibadulin R.Kh., Popov V.V., Troyanov V.M. Technological studies on the use of uranium-molybdenum fuel for research reactors with reduced uranium enrichment // fuel assemblies, fuel elements and zirconium materials for nuclear reactors /: Collection of reports of the Seventh Russian Conference on Reactor Materials Science, Vol. 2., Part 2. Dimitrovgrad September 8-12, 2003. - Dimitrovgrad, 2004. - p. 104-118. Publishing House SSC RF RIAR. 2004, RU / ISBN 5-94831-033-7; research report “Preliminary selection and justification of the optimal composition of an alloy based on uranium and fuel composition of a fuel element for IR”. Topic Code 05832 SSC RF IPPE Inv. No. 9935, 1999, with. 63-64, 69-70) is a prototype.

When using this method, the process of dissolution of aluminum at temperatures of 80-100 ° C is violent, difficult to control, with the release of a large amount of hydrogen, which when mixed with air forms explosive mixtures. It is also possible to deposit sodium aluminate in the foam layer on the walls of the reactor, which, insoluble in sodium hydroxide and acid solutions, can enter and precipitate on the rough surfaces of the extracted nuclear fuel particles, which contaminates the powder of the extracted nuclear fuel. The precipitate of sodium aluminate consists of particles less than 5 microns in size, which are combined into larger agglomerates, covering small particles of fuel, and, forming strong lumps after drying, settle on the surface of nuclear fuel particles.

When using this method, a solution of caustic soda remains in the micropores on the surface of nuclear fuel particles having cracks, which are clogged there by hydrogen, which is difficult to remove by washing with water. Caustic soda remaining in the pores can migrate to the surface of the fuel during operation of the fuel rod in the reactor due to the elevated temperature and the effects of fuel fission products and interact with the core matrix material and the fuel cladding, damaging them. In addition, a 10% sodium hydroxide solution is not enough to completely remove from the surface of the nuclear fuel a thin diffusion layer of interaction with the distorted crystal structure of the nuclear fuel material. This layer with a thickness of less than 1 μm and a variable composition of UO _{2, у} -Al _x , where x varies from 2 to 4, is formed as a result of the mutual diffusion of atoms of uranium, oxygen, aluminum and impurities contained in the aluminum alloy during the temperature-time interaction in the manufacturing process of the fuel rod.

The given temperature-time regimes of drying the nuclear fuel powder in air at 25 ° C for 24 hours or in an argon stream at a temperature of 40-50 ° C for 2-3 hours are long and do not provide complete physical removal from the surface and pores of the regenerated particles and chemically adsorbed water molecules.

Due to the strong impact of the loads on the particles of nuclear fuel in the manufacture of fuel elements, some particles retain their size, but crack, which is undesirable for their subsequent use. The chemical effect of the sodium hydroxide solution does not allow them to be rejected.

An object of the invention is to provide a method for the regeneration of dispersed nuclear fuel from fuel rods and dispersion compositions based on aluminum and aluminum alloys, which makes it possible to extract nuclear fuel powder from substandard and non-irradiated dispersion type fuel rods with shells and a core matrix of aluminum or its alloys at the stage of their manufacture, and also from the waste generated in the manufacturing process of such cores and fuel rods, clean the surface of the extracted powder particles from impurities and the product in the interaction, while maintaining the maximum particle source in the chemical and phase composition, as well as possible, the surface morphology, to carry out grading of nuclear fuel particles for reuse for making fuel elements.

The problem is achieved due to the fact that in the proposed method for the regeneration of nuclear fuel from fuel rods and dispersion compositions based on aluminum and aluminum alloys, including dissolving the cladding and matrix of fuel rods in a solution of caustic soda, separating the powder of nuclear fuel, washing the separated powder with water until neutral, drying the powder in an argon atmosphere, according to the invention, dissolving the cladding and matrix of fuel elements with a solution of sodium hydroxide is carried out at a temperature of 15-75 ° C with a concentration of up to 30%, after the process of the creation of aluminum and / or its alloys in a solution of caustic soda and separation of the powder of nuclear fuel, the washed powder of nuclear fuel is treated in a solution of nitric acid with a concentration of 7-25% for 2-60 seconds at a temperature of 15-40 ° C, washed in deionized water at a temperature 15-95 ° C, subsequent drying is carried out at temperatures up to 165 ° C with a residual pressure of less than 2 kPa for more than 5 minutes, then subjected to machining on sieves with a mesh size equal to and / or less than the average maximum and equal to and / or more than the average mi imalno allowable size of the powder particles in an inert gas atmosphere or nitrogen.

These features are significant and interconnected with the formation of a stable set of essential features sufficient to obtain the desired technical result.

Thus, a solution of nitric acid not only neutralizes possible residues of sodium hydroxide in the pores, but also cleans the surface of fuel particles of possible contaminants and a diffusion layer formed from the interaction between uranium, oxygen and aluminum on the surface of the particles during temperature exposure during the manufacture of a fuel element. At the same time, a new protective oxide layer with a thickness of less than 0.5 μm is formed on the surface of the particles. The concentration of nitric acid and the time of exposure to the powder of nuclear fuel depend on whether the powder was exposed to temperature exposure, the time and temperature of exposure.

Thus, particles of nuclear fuel powder extracted from the estimates of the powder mixture after pressing the porous core preform, which have not undergone heat treatment operations, are processed in a solution of nitric acid with a minimum concentration of 7% at room temperature only to neutralize sodium hydroxide residues, since in this case they have there is a minimal oxide layer on the surface. The temperature range of 15-40 ° C is due to the fact that with an increase in the process temperature, the solubility of molybdenum in a uranyl nitrate solution decreases, a precipitate forms, and the dissolution rate of uranium sharply increases with increasing temperature, and the process becomes unstable and difficult to control.

The use of treatment in a solution of nitric acid in place with subsequent treatments by washing, drying and sieving allows you to use the regenerated-reverse powder of nuclear fuel repeatedly.

Repeated washing in deionized - distilled - water at a temperature of 15-95 ° C until a neutral reaction ensures the removal of residual nitric acid solution. The best results in washing and defragmenting particles are achieved when washing in an apparatus with ultrasonic action through the water on the particles being washed.

The drying temperature of the particles of nuclear fuel is due to the requirement of its excess during degassing over the operating temperature of the product with fuel. As a rule, the surface temperature of dispersion fuel fuel rods in operation does not exceed 100 ° C. In this case, the upper temperature limit is limited by the temperature at which active oxidation of the nuclear fuel powder begins, when nitrous oxide - uranium oxide - begins to form on the surface of particles from uranium alloys. For uranium and its alloys at temperatures below 165 ° C, the oxide film mainly consists of a dense layer of uranium dioxide and has the protective ability to slow down the mutual diffusion of uranium and aluminum during heat treatments, while its thickness, as a rule, should not exceed 1.2 microns. In addition, the selected temperature-time drying mode is selected from the condition that it does not affect the phase composition of the nuclear fuel material.

Drying at a residual pressure of less than 2 kPa ensures the removal of physically adsorbed water molecules at room temperature of 15 ° C, since at this pressure water boils at a temperature below 15 ° C, which significantly reduces the drying time. Drying the extracted nuclear fuel powder at a temperature of 100-165 ° C ensures the desorption of water from micro- and mesopores and prevents the formation of an uncontrolled oxide layer.

The mechanical treatment of the extracted powder on sieves with a mesh size equal to or less than the average maximum allowable and equal to or greater than the average minimum allowable powder size not only selects the desired powder fraction, but also ensures defragmentation of agglomerated and broken particles. Between sieves with a mesh size equal to the average maximum allowable and equal to the average minimum allowable particle size of the powder are sieves with mesh sizes between them. This ensures high-quality defragmentation of the destroyed and agglomerated particles and reduces the load on the sieve with a minimum allowable mesh size.

The essence of the method is illustrated by graphic materials.

In FIG. 1 shows a pellet made of U-9% Mo alloy: initial state before starting the manufacture of a fuel rod core.

In FIG. 2 shows a granule from U-9% Mo alloy extracted from the core of the IRT-3M fuel rod.

In FIG. 3 shows granules from U-9% Mo alloy fraction 40-160 μm, extracted from IRT-3M fuel rod and processed according to the proposed method (1) and in 40% nitric acid for 6 minutes at a temperature of 50 ° C (2).

In FIG. 4 shows a particle of an U ₃ Si ₂ alloy extracted from an aluminum fuel rod matrix before processing on sieves.

In FIG. 5 shows particles from an alloy U ₃ Si ₂ extracted from an aluminum matrix of a fuel rod after processing on sieves.

The method is as follows.

Substandard fragments of the fuel composition in the form of fuel clippings, press residues or other turns are sorted and loaded into separate for each type of turns porous baskets of material that does not interact with sodium hydroxide and a solution of nitric acid, then lowered into a stainless steel reactor made in nuclear -safe performance, in sodium hydroxide solution with a solution concentration of 5-30%. In the process of etching aluminum control the temperature and the upper level of the solution. If the temperature rises to 75 ° C, the basket is removed from the solution or the flow of cooling water into the reactor cooling jacket is increased. After the dissolution of aluminum is complete, the basket with the extracted nuclear fuel powder is washed with deionized distilled water until neutral and lowered for several seconds in a reactor with a solution of nitric acid. After that, the powder is washed with deionized water until neutral. In the process, the molybdenum content in the used nitric acid solution is controlled, which should not exceed 25 g / l.

Next, the powder is loaded into a container, loaded into a vacuum oven, where the necessary vacuum is created to a residual pressure of 2 kPa, heated to the required temperature and kept for more than 5 minutes. The dried powder container is filled with an inert gas. In a box with an inert gas atmosphere, the powder is machined on sieves with a mesh size equal to or less than the average maximum allowable and equal to or greater than the average minimum allowable powder size. The obtained powder of the required fraction is monitored for compliance with the required parameters in terms of chemical and phase composition, then it is packaged in an airtight container in an inert gas atmosphere.

It is possible to carry out the process in reactors with the separation of the nuclear fuel powder from the solutions by the decantation method instead of moving chemical powder treatments in baskets through operations.

The practical implementation of the method is illustrated by the following examples.

Example 1

Extraction of nuclear fuel particles in the form of granules from an U-9% Mo alloy with a fraction of + 40-160 μm from the aluminum matrix of the IRT-ZM fuel rod.

The dissolution of the core fragments of a fuel rod in a 10% sodium hydroxide solution was carried out at a temperature of:

start of process 20 ° C;

stopping the dissolution process 75 ° C;

with a ratio of caustic soda to aluminum 2.5-3.0: 1 to complete dissolution of aluminum.

During the dissolution process, control of the solution level and concentration of caustic soda was carried out, with periodic addition of caustic soda solution to the reactor, as well as interruption of the process by removing the basket of soluble material from the solution in order to reduce the temperature of the process.

After dissolution of aluminum, granules from U-9% Mo alloy were separated from sodium hydroxide solution and washed three times until neutral in deionized - distilled water with a solid to liquid ratio of 1: 5-8. Next, the granules were separated from the wash water and treated with a solution of nitric acid with a concentration of 250 g / l (22%) at a temperature of 15-30 ° C for 45 seconds, washed three times until neutral in distilled water at a solid to liquid ratio of 1: 7-10 until a neutral reaction, pH = 6-7, was separated from the wash water and dried - degassed at a residual pressure of 5 × 10 ^-3 mm RT. Art. at a temperature of 120-140 ° C for 20 minutes. This temperature is selected from the condition that the temperature of the onset of active oxidation of granules from U-9% Mo alloy with a fraction of + 40-160 μm is 200-218 ° C.

The granules were degassed in a nuclear-safe valve container, in which it is possible to remove the container from the drying oven and fill it with an inert gas - argon without depressurization.

Next, mechanical processing of granules in an argon atmosphere was performed on sieves with mesh sizes of 160 μm, 125 μm, 100 μm, 71 μm, 40 μm with a vertical vibration amplitude of the sieve of up to 0.5 mm with a frequency of 90-180 1 / min and a horizontal amplitude of 3 mm with a frequency of 120-300 1 / min.

A certain amount of particles of nuclear fuel powder in the form of granules extracted from an aluminum matrix has a surface deformation in the form of dents from interaction with other particles formed during the manufacture of a fuel element, and slices formed during the cutting of substandard fuel elements. These types of deformation of the particle surface reduce the mobility of particles in the fuel mixture during pouring and pressing, that is, it reduces the possibility of separation of the dispersion fuel mixture, therefore, they have a positive effect on the process.

The resulting fuel granules obtained as a result of the method were analyzed and reused in the manufacture of fuel rods, which later became part of the fuel assemblies that successfully passed reactor tests in a research reactor with an average burnup of the U-235 isotope ~ 50.3% and post-reactor studies. The maximum burnup reached 58.6%.

A comparative result of the quality and morphology of the surface of the granule in the initial state before the manufacture of the fuel rod core and the granule from the U-9% Mo alloy extracted from the aluminum matrix of the core according to the proposed method is shown in FIG. 1, FIG. 2 respectively.

Quantitative results are shown in the table.

Example 2

Extraction of particles of the U ₃ Si ₂ compound with a particle size of less than 160 μm from cut blanks of plate type fuel rods with dimensions of not more than 100 × 120 × 3 mm in an aluminum matrix and shells from an AmG2 aluminum alloy.

The dissolution of the shells of aluminum alloy and the core of a fuel rod made of aluminum in a 10% sodium hydroxide solution was carried out at a temperature of 20-50 ° C with a sodium hydroxide / aluminum ratio of 2.5-3.0: 1 until aluminum was completely dissolved. During the dissolution process, control of the solution level and concentration of caustic soda was carried out with periodic addition of caustic soda solution to the reactor. After aluminum was dissolved, the particles of the compound U ₃ Si ₂ were separated from the sodium hydroxide solution and washed three times until neutral in deionized water at a solid to liquid ratio of 1: 7-10.

Then the particles were separated from the wash water and treated with a solution of nitric acid with a concentration of 150 g / l (14%) at a temperature of 15-30 ° C for 60 minutes, washed three times in deionized water with a solid to liquid ratio of 1: 7-10 until neutral pH = 6-7, was separated from the wash water and dried in an argon atmosphere at a temperature of 90-140 ° C for 30 minutes. The U ₃ Si ₂ alloy particle extracted from the aluminum matrix according to the proposed method, before processing on the sieves, is shown in FIG. 4. In order to separate the destroyed particles into fragments, the particles were machined on sieves with mesh sizes of 160 μm, 125 μm, 100 μm, 71 μm, 40 μm in an argon atmosphere, with a vertical vibration amplitude of the sieve of up to 1 mm with a frequency of 90-180 1 / min and a horizontal amplitude of 3 mm with a frequency of 120-300 1 / min. U ₃ Si ₂ alloy particles extracted from an aluminum matrix according to the proposed method and processed on sieves are shown in FIG. 5.

Thus, the data presented show that, as a result of the implementation of the method, nuclear fuel powders regenerated from fuel elements and dispersion compositions based on aluminum and aluminum alloys meet the requirements and can be reused in the production of dispersion type fuel elements.

Claims (4)

1. A method of regenerating nuclear fuel powders from fuel elements and dispersion compositions based on aluminum and aluminum alloys, comprising dissolving the shell and matrix of the fuel element in a sodium hydroxide solution, separating the nuclear fuel powder, washing the separated powder with water until neutral, drying the powder in an argon atmosphere characterized in that the washed nuclear fuel powder is treated in a solution of nitric acid with a concentration of 7-25% for 2-60 seconds at a temperature of 15-40 ° C, washed in days onized water at a temperature of 15-95 ° C, dried at a temperature of up to 165 ° C and carry out mechanical processing on sieves. 2. The method according to p. 1, characterized in that the dissolution of the shell and matrix of the fuel element is carried out with a solution of caustic soda at a temperature of 15-75 ° C. 3. The method according to p. 1, characterized in that the powder is dried at a residual pressure of 2 kPa for more than 5 minutes. 4. The method according to p. 1, characterized in that the mechanical treatment of the dried nuclear fuel powder is carried out in an atmosphere of inert gas or nitrogen on sieves with a mesh size equal to and / or less than the average maximum allowable and equal to and / or more than the average minimum allowable particle size powder.

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