RU2357311C2 - Method of processing spent nuclear fuel - Google Patents

Abstract

FIELD: chemistry.

SUBSTANCE: present invention relates to chemical and radiochemical industry and can be used for recycling spent nuclear fuel, mainly dispersion fuel elements with compositions (cermets) of the type Be - Ube₁₃, Be - UO₂, and returning enriched uranium into the fuel cycle, extracting (recycling) uranium and beryllium from waste products of uranium-beryllium compositions. In the method of processing spent uranium-beryllium nuclear fuel, spent fuel elements and fuel assemblies are broken down to pieces, sunk into water and dissolved through addition of gaseous hydrogen fluoride into the solution. After completion of the dissolution process, solid parts are separated and pulp is separated from beryllium fluoride solution and washed. The pulp is heated and treated with steam. If necessary, the obtained uranium dioxide undergoes further purification.

EFFECT: invention simplifies the method of processing spent nuclear fuel, mainly compositions of the type Be - Ube₁₃, Be - UO₂, and increases safety of carrying out the process as well.

4 cl

Description

The invention relates to the field of chemical and radiochemical industry and can be used for regeneration of spent nuclear fuel, mainly dispersion fuel elements with compositions (cermets) of the type Be - UBe ₁₃ , Be - UO ₂ , and the return of enriched uranium to the fuel cycle, extraction (disposal) uranium and beryllium from waste products of uranium-beryllium compositions.

Known methods for processing irradiated metallic uranium (in order to extract the plutonium accumulated in it) by dissolving in nitric acid, followed by extraction extraction and purification of uranium and plutonium from fission products (PD). (Emelyanov V.S., Evstyukhin A.I. Metallurgy of nuclear fuel. M: Atomizdat, 1968, p. 343; Zemlyanukhin V.I. et al. Radiochemical processing of nuclear fuel of nuclear power plants. - M.: Energoatomizdat, 1989, p. .54).

A known method of processing uranium-aluminum nuclear fuel by dissolving it in dilute nitric acid, with mercury nitrate as a catalyst, at the boiling point of the solution (about 110 ° C) (V. Zemlyanukhin and other Radiochemical processing of nuclear fuel of nuclear power plants. - M. : Energoatomizdat, 1989, p. 58-59). The disadvantage of this method is the use of a large amount of mercury, which intensifies corrosion, complicates the processing and storage of waste, and also increases the likelihood of the formation of stabilized emulsions that complicate the subsequent extraction cleaning process.

The above methods are of little use for the processing of uranium-beryllium compositions, since when the fuel is dissolved, joint solutions of uranium and beryllium are formed. A large amount of beryllium in solutions with highly enriched uranium greatly complicates the fulfillment of nuclear safety conditions.

The closest analogue of the proposed method is a method for dissolving the cores of uranium-zirconium fuel elements in hydrofluoric acid (Zemlyanukhin V.I. et al. Radiochemical processing of nuclear fuel of nuclear power plants. - M .: Energoatomizdat, 1989, p. 59). When dissolving such fuel, strict control over the concentration of hydrofluoric acid and the temperature of the solution is necessary. Excess of free fluoride ions causes a loss of uranium in the precipitate in the form UF ₄ · _^3/4 H ₂ O, and their reduction - hydrolysis of zirconium salts.

The disadvantage of this method is that a large amount of beryllium in solutions with highly enriched uranium extremely complicates the fulfillment of nuclear safety conditions of processes.

In addition, the drawback of this method is the fact that a large excess of free fluoride ions causes a loss of uranium in the precipitate in the form UF ₄ · _^3/4 H ₂ O. To hold the uranium in the solution it is recommended to transfer the hexavalent state by introducing oxidizing process.

The disadvantage of this method is that with a further separation of the elements by extraction, a large number of solutions are formed that must be disposed of.

The objective of the proposed technical solution is to simplify the method of processing spent nuclear fuel, mainly compositions of the type Be - UBe ₁₃ , Be - UO ₂ , as well as increasing safety during the process.

The technical effect of the processing method is that the fuel components are transferred into different phases convenient for their further separation: uranium was released as a solid compound, and beryllium in solution.

The above task is achieved by the fact that in the method of processing spent uranium-beryllium nuclear fuel, spent fuel rods and fuel assemblies with cores from Be - UBe ₁₃ and Be - UO ₂ compositions are chopped into pieces, immersed in water, and hydrogen fluoride gas is supplied therein, maintaining a certain temperature, and at the end of the dissolution process, the solid parts are separated, the pulp is separated from the beryllium fluoride solution, washed, the pulp is heated, steam is treated and, if necessary, the separated uranium dioxide will be added. Flax cleaning.

Hydrogen fluoride gas is supplied to maintain its concentration in the solution (1-3)%.

When dissolved, the temperature of the solution is maintained no more than 70 ° C.

When gaseous hydrogen fluoride is introduced into the water, dilute (~ 1-3%) hydrofluoric acid is formed. Beryllium of fuel cores, including UBe ₁₃ intermetallic acid, reacts with acid to form beryllium fluoride well soluble in water, and UBe ₁₃ intermetallic uranium interacts mainly with water to form insoluble dioxide (~ 95%), and partially with acid to form sparingly soluble tetrafluoride (~ 5%). At the same time, the product of these exothermic reactions is the evolution of gaseous hydrogen. The supply of hydrogen fluoride to the solution is continued throughout the entire process of dissolution of spent uranium-beryllium nuclear fuel, the rate of dissolution of spent uranium-beryllium nuclear fuel is controlled by changing its flow rate and its supply is stopped when approaching the dissolution time of the cores, when the concentration of hydrofluoric acid should be close to zero.

The sublayer metals contained in the fuel elements (sodium, calcium), and in the spent fuel, fission products during the dissolution of the fuel cores are distributed between the beryllium fluoride solution and solid uranium compounds in accordance with the solubility of their fluorides.

An increase in the concentration of hydrofluoric acid by more than 3% and a temperature of more than 70 ° C leads to an increase in the proportion of uranium tetrafluoride, its solubility in an acid solution, and, accordingly, to an increase in the loss of uranium with a solution of beryllium fluoride.

Upon dissolution of the cores Be - UO ₂ , uranium dioxide practically does not interact with dilute hydrofluoric acid and precipitates.

Thus, in the dissolution of the Be - UBe ₁₃ and Be - UO ₂ compositions, the final products in both cases are pulp - a solution of beryllium fluoride containing solid uranium compounds.

Heated pulp to 500-800 ° C.

When the pulp is heated to less than 500 ° C, not all fluorine is released from the pulp; heating to more than 800 ° C is impractical.

New significant features of the proposed method (compared with the prototype) are:

- the use of dilute hydrofluoric acid to dissolve spent nuclear fuel (compositions Be - UBe ₁₃ , Be - UO ₂ ) and maintain its minimum concentration during the dissolution process, which immediately leads to the conversion of beryllium and uranium into easily separable phases (beryllium goes into solution, and uranium into solid compounds);

- the separation of solid uranium compounds from a solution of beryllium fluoride, which virtually eliminates beryllium and fluorine from the further process of purification of enriched uranium from fission products, makes it possible to use, for example, standard equipment made of stainless steels in extraction processes;

- the use of gaseous hydrogen fluoride for the continuous production of hydrofluoric acid and maintaining its constant low concentration during the dissolution of SNF allows one to drastically reduce the volume of solutions, simplify the organization, conduct, and control of the rate of dissolution of the fuel core cores.

Spent fuel assemblies or fuel rods with Be - UBe ₁₃ or Be - UO ₂ cores containing 70-90% beryllium are cut into pieces. In a nuclear-safe, corrosion-resistant, hydrofluoric acid sealed apparatus-solvent, water is poured, pieces of fuel elements are loaded, and gaseous hydrogen fluoride is fed into the water. The onset of dissolution of the cores by the resulting hydrofluoric acid is evidenced by the evolution of hydrogen and an increase in solution temperature. By controlling the flow of hydrogen fluoride, the concentration of hydrofluoric acid is maintained at a level of (1-3) percent. The temperature of the solution is not more than 70 ° C, for which water is passed through the cooling jacket of the apparatus. The evolved hydrogen with water vapor enters the cooler-condenser, the condensate drains into the solvent apparatus, and the hydrogen is sent for combustion with atmospheric oxygen. The approximation of the end of the process of dissolution of the cores of the fuel rods indicates a decrease in temperature in the apparatus and the flow of hydrogen fluoride is stopped. As a result of dissolution of the cores, a pulp of uranium dioxide (~ 95%) and tetrafluoride (~ 5%) is formed in a solution of beryllium fluoride. Other components of fuel elements and fuels (sublayers of sodium or calcium, PD), in accordance with their chemical properties, form mainly fluorides, of which insoluble accompany solid uranium compounds, and soluble go into solution, thus partially purifying uranium from some PD. The material of the cladding of the fuel elements (steel EI, EP) during the dissolution of the cores of beryllium compositions practically does not corrode.

The resulting pulp is drained from the apparatus, the sections of the cladding of the fuel rods are washed from the remnants of the pulp and removed. The collected pulp is processed - solid uranium compounds, namely uranium dioxide mixed with tetrafluoride and insoluble PD fluorides, are separated from a solution of beryllium fluoride and soluble PD fluorides. The precipitate is washed from a solution of beryllium fluoride, the washing solutions are returned to the next operation of dissolving the fuel elements.

The precipitate of solid uranium compounds is loaded into the furnace retort, dried, heated to a temperature of 500-800 ° C, water vapor is passed through the retort, the fluoride impurities are “pyrohydrolyzed”, and fluorine is removed with steam in the form of hydrogen fluoride to obtain uranium dioxide contaminated with PD with an admixture of ~ 1% beryllium and less than 0.1% fluorine, sent for further purification by known methods, for example, extraction or gas fluoride.

Claims (4)

1. A method of processing spent uranium-beryllium nuclear fuel, including chopping into pieces of spent fuel elements and fuel assemblies and dissolving them in hydrofluoric acid, characterized in that the pieces are immersed in water, hydrogen fluoride gas is supplied therein, maintaining a certain temperature, and at the end of the process the solids are separated, the pulp is separated from the beryllium fluoride solution, washed, the pulp is heated, steamed and, if necessary, the separated uranium dioxide is subjected. linen cleaning. 2. The method according to claim 1, characterized in that gaseous hydrogen fluoride is supplied to maintain its concentration in the solution of 1-3%. 3. The method according to claim 1, characterized in that when dissolved, the temperature of the solution is maintained no more than 70 ° C. 4. The method according to claim 1, characterized in that the pulp is heated to 500-800 ° C.