RU2598943C1 - Method of producing solid plutonium dioxide solution in uranium dioxide matrix - Google Patents

Abstract

FIELD: radiochemical industry; nuclear power engineering.

SUBSTANCE: invention relates to and is aimed at obtaining a mixed dioxide (U,Pu)O₂, which can be used for production of mixed uranium-plutonium MOX fuel for VVER-1000 reactors and fast neutron reactors (BN-600, BN-800) at nuclear power stations. Method of producing solid plutonium dioxide solution in uranium dioxide matrix includes: reacting of nitrate complexes of uranium and plutonium having their relative content in the solution of 95-70 and 5-30 wt% correspondingly with hydrazine hydrate in the molar ratio of - hydrazine hydrate: uranium, plutonium - equal to 2 in order to produce a mixed amorphous compound of uranium and plutonium, maintenance of the mixed amorphous compound of uranium and plutonium in mother solution at the temperature of 80-90 °C during at least 3.5 hours to obtain a fine powder sediment of homogeneously mixed hydrated uranium and plutonium dioxides, separation of the sediment from mother solution and its heating to the temperature of 280-300 °C till obtaining the target product.

EFFECT: invention provides cost-effective, simple and less energy consuming method of producing solid plutonium dioxide solution in a uranium dioxide matrix.

1 cl, 6 dwg, 2 ex

Description

The invention relates to the radiochemical industry and nuclear energy and is directed to the production of mixed dioxide (U, Pu) O ₂ , which can be used for the manufacture of nuclear mixed uranium-plutonium MOX fuel from VVER-1000 reactors and fast neutron reactors (BN-600, BN-800) nuclear power plants.

To prepare pelletized MOX fuel, a mechanical mixture of UO ₂ and PuO ₂ powders is used. The MIMAS process (France) ["Advanced MIMAS process" is most tested for these purposes. Auteurs: DUCROUX R .; COUTY Y .; LEROUX JC Editeur SFEN. Conférence: International nuclear conference on recycling, conditioning and disposal, Nice, FRA, 1998-10-25]. It includes two main stages of powder preparation:

- joint grinding of powdered oxides of uranium and plutonium with the formation of a concentrate with a plutonium content in the mixture up to 25-30%;

- dry dilution of the specified concentrate with uranium dioxide to the final required plutonium content.

The main disadvantage of the MIMAS process and other methods based on the mixing of dry powders of uranium and plutonium oxides is the difficulty in obtaining the most homogeneous compositions, which leads to a decrease in the amount of burnable fractions and the incomplete dissolution of spent nuclear fuel during its reprocessing.

This disadvantage can be avoided if the joint deposition of uranium and plutonium from a solution is carried out with a further conversion of the obtained compounds to mixed dioxide of uranium and plutonium (U, Pu) O ₂ .

There is a method according to which mixtures of oxides are obtained by precipitation from solutions of mixtures of ammonium diuranate and plutonium hydroxide, followed by filtration, drying, calcination and reduction with hydrogen [Samoylov AG Fuel elements of nuclear reactors. - M .: Energoatomizdat, 1985, p. 64].

The disadvantages of this method include the following:

- the complexity of the process;

- the need to use high temperatures;

- the use of hydrogen for recovery.

In another method [RU 2282590 C2, IPC C0G 43/00] a technical solution is proposed that makes it possible to obtain mixed uranium and plutonium dioxide with a homogeneous distribution of actinides in powder granules by the following operations:

- preliminary reduction of uranium to U (IV) by introducing into the solution of the reducing agent - hydrozonium ions [N ₂ H ₅ ] ⁺ ;

- stabilization of uranium in the oxidation state of IV by complexing agents - diethylenetetraaminopentaacetic or nitriloacetic acids, which also form complexes with Pu (IV);

- co-precipitation of uranium and plutonium using a special water-ethanol medium, adding up to 30 (vol)% ethanol to the solution and creating a pH of 7.5 with concentrated ammonia;

- drying and calcining the precipitate at a temperature of more than 650 ° C in an inert atmosphere.

The disadvantages of this method include:

- the complexity of the process;

- the need for preliminary reduction of uranium, i.e. the presence of a separate stage of the process;

- the use of a special water-ethanol medium for coprecipitation;

- the need for an inert atmosphere and the use of high temperatures.

A known method of obtaining a solid solution of plutonium dioxide in a matrix of uranium dioxide [RU 2446107 C1, IPC C0G 43/025]. This method involves the interaction of nitrate solutions of uranium and plutonium with a relative content in the solution of 95 ÷ 70 and 5 ÷ 30 wt. %, respectively, with hydroxylamine, which leads to the reduction of plutonium to the trivalent state and the coprecipitation of uranium and plutonium in the form of a homogeneous mixture of uranium sulphate hydroxylamine with plutonium hydroxide, and further decomposition of the resulting precipitate in air at 200-300 ° C.

The disadvantages of this method are:

- isolation of a solution of intermediate multivalent compounds of uranium and plutonium;

- low thermal stability of the resulting mixture of oxides in air.

The problem to which the invention is directed, is to develop an economically feasible relatively simple and less energy-intensive method for producing a solid solution of plutonium dioxide in a matrix of uranium dioxide (U, Pu) O ₂ suitable for producing MOX fuel by reducing the number of process steps and its temperature.

To solve this problem, a method of obtaining a solid solution of plutonium dioxide in a matrix of uranium dioxide involves the interaction of nitrate complexes of uranium and plutonium with a relative content of 95-70 and 5-30 wt. % respectively with hydrazine hydrate at a molar ratio of hydrazine hydrate: uranium, plutonium, equal to 2, to obtain a mixed amorphous compound of uranium and plutonium, holding the mixed amorphous compound of uranium and plutonium in the mother liquor at a temperature of 80-90 ° C for at least 3.5- 5 hours to obtain a precipitate of finely divided powder of homogeneous mixed hydrated uranium and plutonium dioxide, separating the precipitate from the mother liquor and heating it to a temperature of 280-300 ° C until the formation of the target product.

In a particular embodiment, the exposure of the amorphous compound of uranium and plutonium in the mother liquor and the heating of the fine powder is carried out with convective heat input.

The presence of plutonium in the uranium dioxide matrix stabilizes the crystal structure so much that even after a solid solution of plutonium dioxide in the uranium dioxide matrix is heated to 800 ° C, only one homogeneous mixture of two (U, Pu) O ₂ dioxides is identified.

The choice of the relative content of uranium and plutonium in the solution is due to the fact that the plutonium content in the MOX fuel is from 5 to 30 wt. %

A twofold molar excess of hydrazine hydrate is necessary for the complete transfer of uranium and plutonium to a fine powder of homogeneous mixed hydrated uranium dioxide and plutonium, a further increase in the molar excess of hydrazine hydrate does not affect the technical result, and an increase in the molar excess of hydrazine hydrate is not economically feasible.

At temperatures below 80 ° C, there is no quantitative transition of a mixed amorphous compound of uranium and plutonium into a powder of a homogeneously mixed hydrated dioxide of uranium and plutonium; at temperatures above 90 ° C, the process is not economically feasible.

At temperatures below 280 ° C, there is no quantitative transition of a homogeneously mixed hydrated uranium and plutonium dioxide to a solid solution of plutonium dioxide in a uranium dioxide matrix (Fig. 5 and Fig. 6); at temperatures above 300 ° C, the process is not economically feasible.

The essence of the invention is illustrated by the following illustrations.

In FIG. 1 shows the spectrum of the initial solution of U (VI) and Pu (VI) in 0.1 mol / L HNO ₃ diluted 50 times.

In FIG. Figure 2 shows the spectrum of a solution of a mixed compound of U and Pu isolated from a suspension maintained at T = 90 ° C for 3.5 hours and dissolved in a mixture of 6 mol / L HNO ₃ and 0.01 mol / L HF.

In FIG. Figure 3 shows the spectrum of solution U obtained by dissolving its compound isolated from a suspension aged at T = 90 ° C for 2 days in a mixture of 6 mol / L HNO ₃ with 0.1 mol / L HF.

In FIG. Figure 4 shows the spectrum of a Pu solution obtained by dissolving its compound isolated from a suspension aged at T = 90 ° C for 3.5 hours in a mixture of 6 mol / L HNO ₃ with 0.1 mol / L HF.

In FIG. 5 shows the data of synchronous thermal analysis of hydrated dioxides of uranium and plutonium.

In FIG. Figure 6 shows the X-ray diffraction pattern of a solid solution of plutonium dioxide in a matrix of uranium dioxide (U, Pu) O ₂ after it was heated in air to 800 ° C: UO ₂ (1), PuO ₂ (2).

Examples of the method

Example 1

To the initial aqueous solution of uranium and plutonium nitrates in 0.1 mol / L HNO ₃ (Fig. 1) is added hydrazine hydrate N ₂ H ₅ OH in a molar ratio of N ₂ H ₅ OH: (U, Pu) equal to 2. In solution an amorphous suspension of yellow-gray color is formed. After 3.5 hours of heating it at a temperature of 90 ° C, the suspension passes from a bulk amorphous poorly settling state into a rapidly settling black precipitate.

The resulting precipitate of uranium and plutonium is separated from the mother liquor. To establish the forms of the presence of uranium and plutonium in the sediment, part of the precipitate was dissolved in nitric acid. As can be seen from the data in FIG. 2, after dissolution, only U (VI) and Pu (III) ions are present in the solution. Their appearance can be explained by the occurrence in the acid solution of the redox reaction Pu (IV) + U (IV) = Pu (III) + U (VI). This unequivocally proves that in the solid phase formed after heating the suspension and converting it to well settling, the black precipitate of uranium and plutonium are in the 4+ oxidation state, since only in this case the specified reaction can occur when dissolved in an acidic solution. Specially performed experiments separately with uranium and with plutonium showed that under similar conditions in solutions of uranium (Fig. 3), as well as plutonium (Fig. 4) are in a 4+ oxidation state. Thus, the black precipitate of the mixed compound of uranium and plutonium is a mixture of their hydrated dioxides.

A synchronous thermal analysis of the precipitate separated from the mother liquor was carried out (Fig. 5). After conducting a synchronous thermal analysis with heating of the sample to 800 ° C, an X-ray diffraction pattern of the obtained product was taken (Fig. 6). According to synchronous thermal analysis, a decrease in the weight of the analyzed sample occurs in the temperature range from ~ 80 ° C to ~ 280 ° C.

Further heating of the compound to 800 ° C leads to a slight change in its weight. The DTA curve (differential thermal analysis) shows that the crystallization of a solid solution of uranium dioxide and plutonium after the loss of hydrated water occurs in two stages with maximum exothermic effects at 175 and 225 ° C.

The result of the x-ray phase analysis shown in FIG. 6 clearly indicates that in the sample of a mixture of oxides heated to 800 ° C, only one actinide dioxide phase is identified - (U, Pu) O ₂ or a homogeneous mixture of two dioxides.

Example 2

The method of obtaining a solid solution of plutonium dioxide in a matrix of uranium dioxide is carried out, as in example 1, the process is carried out at 80 ° C for 5 hours with convective heat supply until a rapidly precipitating black precipitate is formed. The precipitate is separated from the mother liquor and calcined in air at a temperature of 300 ° C.

Thus, a simple and effective method has been developed for obtaining a solid solution of plutonium dioxide from a nitric acid solution of uranium and plutonium in a matrix of uranium dioxide, which can be used in the production of MOX fuel.

Claims (2)

1. The method of obtaining a solid solution of plutonium dioxide in a matrix of uranium dioxide, comprising the interaction of nitrate complexes of uranium and plutonium with a relative content in solution of 95-70 and 5-30 wt.%, Respectively, with hydrazine hydrate with a molar ratio hydrazine hydrate: uranium, plutonium, equal to 2 to obtain a mixed amorphous compound of uranium and plutonium, holding the mixed amorphous compound of uranium and plutonium in the mother liquor at a temperature of 80-90 ° C for 3.5-5 hours to obtain a fine powder precipitate homogeneously mixed hydrated uranium and plutonium dioxide, separating the precipitate from the mother liquor and heating it to a temperature of 280-300 ° C until the formation of the target product. 2. The method according to p. 1, characterized in that the exposure of the amorphous compounds of uranium and plutonium in the mother liquor and the heating of the fine powder is carried out with convective heat supply.

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