RU2400846C1 - Dissolution method of off-grade and/or spent fuel - Google Patents

Abstract

FIELD: power industry. ^ SUBSTANCE: invention can be used for re-processing of off-grade fuel assemblies of fuel elements and assemblies, which have gone through nuclear fuel cycle in the reactor. Dissolution of off-grade and spent nuclear fuel is characterised by use of water solution of ferrum salt as dissolution; at that, dissolution is carried out at pH 1.0-1.4, during 3-5 hours and molar ratio of fuel to ferrum salt equal to 1:2.1-2.5. As ferrum salt there used is ferric chloride or ferric nitrate. ^ EFFECT: achieving the best ecological, economical and social effects owing to eliminating the use of concentrated solutions of nitric acid, organic solvents and simplifying the further re-processing of spent fuel.

Description

The present invention relates to methods for dissolving actinide oxides, which are the basis of oxide nuclear fuel, and can be used for processing substandard fuel assemblies of fuel elements (TVEL) and assemblies that have passed the nuclear fuel cycle (NFC) in the reactor.

The development of the country's nuclear energy industry is focused on a closed nuclear fuel cycle, which, among other urgent tasks, involves the extraction of its main components (U and Pu) from spent nuclear fuel (SNF) for reuse. This significantly increases the economic feasibility of the nuclear fuel cycle. It is urgent to develop and optimize alternative concepts for the currently used process for the processing of spent nuclear fuel (Purex process), which ultimately allows you to achieve the best environmental, economic and social effects.

Non-aqueous methods for reprocessing spent nuclear fuel are known, in particular, a method for regenerating spent nuclear fuel based on exposure to 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. A method for such processing of spent nuclear fuel by evaporating the resulting fluoride compounds of the constituent fuel components is described, for example, in "JAERY-M6592 (1976)." During fluorination, uranium dioxide contained in SNF is converted to UF ₆ , which vaporizes relatively easily, unlike plutonium, which has lower volatility. Typically, in the reprocessing of spent nuclear fuel by the proposed method, the spent fuel is not fluorinated in order to extract all the wounds contained in it, but fluorinated in order to extract only the necessary amount of uranium from it, separating it from the rest of the reprocessed spent fuel. After extracting a certain amount of uranium from it, the evaporation mode is changed and some quantity of plutonium contained in it is also removed in the form of vapors [RF patent No. 2230130, С22В 60/02, 01/19/1976]. Thus, in this method of SNF processing, gaseous, aggressive and environmentally toxic chemical compounds are used.

In another method, spent nuclear fuel is reprocessed pyrochemically using a molten salt of uranium or plutonium, after which the separated components of the nuclear fuel are reused. In the pyrochemical reprocessing of spent nuclear fuel, its induction heating in the crucible and its cooling are applied by supplying a refrigerant to the crucible. Pyrometallurgical technologies do not lead to the formation of large quantities of liquid radioactive waste, and also provide a more compact placement of equipment, which occupies many times less production space compared to existing water technologies. At the same time, they are very energy-intensive and technologically sophisticated [RF patent No. 2226725, G21C 19/46, 01/19/2009].

Non-aqueous methods for dissolving SNF also include:

(1) a method comprising oxidizing uranium in a solvent using dissolved gases as an oxidizing agent, for example, chlorine, nitrogen oxides, sulfur dioxide, and as a solvent, a dipolar aprotic solvent or a mixture thereof with a chlorine-containing compound [RF patent RF No. 2238600].

(2) a method for dissolving materials containing metallic uranium, comprising oxidizing metallic uranium with a tributyl phosphate-kerosene mixture containing nitric acid [US Patent No. 3288568, 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, No. 4, p.545-549].

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

The most widely used spent nuclear fuel reprocessing technology we have taken as a prototype is the Purex process, in which fuel containing uranium and plutonium is dissolved in concentrated nitric acid at a temperature of 60-80 ° С. After that, they are extracted from the nitric acid solution with an organic phase containing tributyl phosphate or another extractant in kerosene or some other organic solvent. The following are the technological stages associated with the separation of uranium and plutonium and their purification from fission products of nuclear fuel that arise during the operation of a nuclear reactor. The Purex method of spent nuclear fuel reprocessing is described, for example, in The Chemistry of the Actinide and Transactinide Elements Third Edition, Edited by Lester R. Morss, Norman M. Edelstein and Jean Fuger 2006, Springer p.841-844 .

The specified process of repartition of spent nuclear fuel is multi-stage, based on the use of environmentally unsafe environments:

(1) nitric acid (6-8M), as a SNF solvent, operating at temperatures of 60-80 ° C and forming aggressive gaseous products during reactions with its participation;

(2) toxic, combustible, flammable, and explosive organic solvents in extraction processes, moreover, often unstable to radiation that destroys the extractant, which worsens the process or leads to its shutdown.

As a result, the production of the target products (compounds of uranium and plutonium) is inevitably accompanied by the formation of large volumes of strongly acidic aqueous and toxic organic, in addition highly radioactive waste solutions (RAW) (150-700 l / ton of fuel containing up to 15-20% fissile elements (U and Pu) and their fission products (PD); 3-5 m ³ of medium activity solutions).

Therefore, it is relevant to create and optimize new, low-waste, environmentally friendly and economically feasible technologies that would ensure reprocessing of spent nuclear fuel, both operating and generation 3 and 4 reactors using fast neutrons operating on mixed oxide uranium-plutonium fuel (MOX), which ultimately will achieve the best environmental, economic and social effects.

The objective of the present invention is to develop a new alternative approach to the processing of substandard and spent nuclear fuel, which allows to achieve better environmental, economic and social effects by eliminating the use of concentrated solutions of nitric acid, organic solvents and simplifying the subsequent processing of spent nuclear fuel.

The problem is solved by the method of dissolving substandard and spent nuclear fuel, characterized by using an aqueous solution of an iron (III) salt as a solvent, while dissolving is carried out at a pH of 1.0-1.4 and a molar ratio of fuel to iron salt of 1: 2, 1-2.5.

Preferably, iron (III) chloride or nitrate is used as the iron salt. To obtain a uranium content of 50-60 g / l, a 25-30% aqueous solution of an iron (III) salt is usually used.

Solutions of nitrate or ferric chloride (III) combine oxidizing (redox vapor Fe (III) / Fe (II)) and complexing (NO ₃ ^- or Cl ^- ion) properties. The dissolution of the oxides takes place at [H +] ~ 0.1 M, (pH ~ 1), at room temperature. The main advantage of this approach to the dissolution of actinide oxides is that after the dissolution of the oxides, the acidity of the solution makes it possible to use the most suitable known and well-developed recovery methods, such as peroxide, fluoride, carbonate or oxalate methods. This, in turn, allows us to abandon the extraction stage of the separation of uranium and plutonium from the solution, which must be used for strongly acidic solutions of nitric acid, which leads to subsequent problems associated with the disposal of the resulting organic waste solutions.

The following examples illustrate the solubility of aqueous solutions of nitrate or ferric chloride (III) with respect to UO ₂ , a solid solution of PuO ₂ in UO ₂ (MOX fuel), and uranium dioxide containing cesium, strontium and technetium.

Example. one

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. This procedure guaranteed the preparation of the crystalline structure of UO ₂ . Samples of UO ₂ were added to aqueous solutions of ferric chloride (FeCl ₃ × 6H ₂ O) or iron nitrate (Fe (NO ₃ ) ₃ × 9H ₂ O) in centrifuge tubes. The suspensions were mixed and the uranium content in the solution was periodically monitored by alpha spectrometry after separation of the heterogeneous phases by centrifugation. The results of determining the dependence of the degree of dissolution of UO ₂ on the molar ratio of UO ₂ to FeCl ₃ × 6H ₂ 0 or Fe (NO ₃ ) ₃ × 9H ₂ O in their solutions with pH 1 are presented in Table 1. The volume of solutions is 5 ml, weighed portions of UO _{2 are} practically the same for all cases - 61.0 ± 1.0 mg or 0.26 mmole UO ₂ temperature ~ 22 ° C, * - in percent of the sample taken.

The data in table 1 show that the dissolution of UO ₂ passes strictly with stoichiometry of the reaction

Therefore, it is sufficient to have in the solution no more than 10% of an excess amount of FeCl ₃ × 6H ₂ O or Fe (NO ₃ ) ₃ × 9H ₂ O with respect to their amounts determined by stoichiometry of the redox reaction (1) with UO ₂ .

Example. 2

The kinetics of dissolution of ^{238 + 235} UO ₂ in solutions of FeCl ₃ × 6H ₂ O and Fe (NO ₃ ) ₃ × 9H ₂ O with pH ~ 1 at a molar ratio of UO ₂ to FeCl ₃ × 6H ₂ O and Fe (NO ₃ ) ₃ × 9H ₂ O equal to 1: 2.1 at ~ 22 ° C is shown in Fig.1. The volume of solutions is 1 ml, weighed portions of UO ₂ are 100.0 + 1.7 mg each. The amounts of FeCl ₃ × 6H ₂ O and Fe (NO ₃ ) ₃ × 9H ₂ O in solutions are 220 mg and 329 mg, respectively.

Here, for comparison, the kinetics of dissolution of uranium dioxide in 6MHNO ₃ + 0.1MHF is presented. This mixture of acids is most often used for preparative purposes to dissolve actinide oxides.

The data obtained, presented in figure 1, show that the dissolution of UO ₂ in all types of solutions occurs at almost the same rate. Moreover, if in a solution of a mixture of acids 6MH NO ₃ + 0.1M HF quantitative dissolution of 100 mg UO _{2 is} achieved in 3 hours, then over the same period of time, the same amount of uranium dioxide in 1 ml of solutions of nitrate or chloride of Fe (III) is dissolved in 90% After 5 hours, when shaking in a LAB PU-02 mixing device, uranium dioxide is completely dissolved in them. The uranium concentration obtained in solutions (100 mg / ml) is quite comparable with the uranium concentration (300 mg / ml) achieved by dissolving SNF in technological strong nitric acid solutions (6-8M HNO ₃ , t = 60-80 ° C).

Example. 3

A portion of the industrial fuel tablet — a solid solution of PuO ₂ (4.6 wt.%) In UO ₂ (MOX fuel), ground to a powder state, was introduced into solutions of ferric chloride or iron nitrate in centrifuge tubes. Suspensions were shaken in a LAB PU-02 mixing device. The solutions were periodically monitored for the content of ²³⁹ Pu by alpha spectrometry on an Alpha-Analist spectrometer (Kanbera firm) after separation of heterogeneous phases by centrifugation. The oxidation state of plutonium in solutions was determined by spectrophotometric method. The uranium content after dissolution of the MOX fuel sample was also found by spectrophotometry (Unicam UV-340 spectrophotometer), since it is difficult to reliably determine ^238-233 U by radiometry at ~ 5% by weight of ²³⁹ Pu in the solution. For this, an aliquot of the solution was treated with zinc amalgam and the dissolved U (VI) was reduced to U (IV). Using the molar extinction coefficient for the absorption band of U (IV) at 650 nm, equal to 82 M ^-1 × cm ^-1 , the concentration of uranium in the solution was determined.

The result of the dissolution of MOX fuel in solutions of chloride and iron nitrate with a pH of ~ 1 is presented in Table 2. Samples of MOX fuel are almost identical for both cases - 116.5 ± 3.5 mg The molar ratio of MOX to FeCl ₃ × 6H ₂ 0 (245 mg) and Fe (NO _{3) 3} × 9H ₂ O (366 mg) in 1: 2.1 solutions. The volume of solutions is 1 ml, temperature ~ 22 ° С.

Thus, the results presented in Table 2 show that a solid solution of PuO ₂ (4.6 wt.%) In UO ₂ dissolves as well as pure uranium dioxide in solutions containing Fe (III) nitrate or chloride.

In figure 2. The spectrum of the solution after dissolution of MOX in iron nitrate is presented. As spectrophotometric measurements show, uranium enters the solution in the form of UO ₂ ²⁺ , and plutonium is reduced to Pu (III) by Fe (II) ions formed during the oxidation of U (IV) by iron (III). In Fe (III) chloride solutions, completely similar spectra of U (IV) and Pu (III) are obtained.

Example 4

The dissolution of UO ₂ containing ⁹⁹ Tc in the amount of 0.1 wt% by weight of the sample and uranium dioxide, with cesium and strontium labeled ¹³⁷ Cs and ⁸⁵ Sr (PD simulators), in amounts of 0.1 and 1 wt.%, Respectively, was carried out in conditions similar to examples 2 and 3. Samples of UO ₂ (Tc) 44.1 and 44.6 mg were dissolved in 2.5 ml of solutions containing 100 mg of FeCl ₃ × 6H ₂ O and 140 mg of Fe (NO ₃ ) ₃ × 9H ₂ O respectively. Samples of UO ₂ (Cs, Sr) 47.1 and 48.2 mg were dissolved in 2.5 ml of solutions containing 99 mg of FeCl ₃ × 6H ₂ O and 151 mg of Fe (NO ₃ ) ₃ × 9H ₂ O, respectively. The molar ratio of UO ₂ to FeCl ₃ × 6H ₂ O or to Fe (NO ₃ ) ₃ × 9H ₂ O in all cases was close to 1 to 2.1. Upon dissolution of samples of uranium dioxide with ⁹⁹ Tc, 20 ± 2% Tc passed into solutions of nitrate and chloride of Fe, and ~ 80 ± 2% of its amount remained in the form of an undissolved black precipitate. Thus, with the complete dissolution of UO ₂ containing technetium, the separation of uranium from the bulk of ⁹⁹ Tc is achieved already at the dissolution stage, in contrast to the Purex process, in which at the stage of extraction of uranium and plutonium from a nitric acid solution, technetium passes organic phase (table 3). Dissolution of uranium dioxide containing ¹³⁷ Cs and ⁸⁵ Sr in chloride or nitrate of Fe (III) is accompanied by their complete transition to the solution together with uranium (Table 4).

Chloric acid solutions (pH ~ 1) obtained by dissolving uranium dioxide containing ⁹⁹ Tc and ¹³⁷ Cs, ⁸⁵ Sr in them can be used to separate uranium from it and separate it from iron and fission products by peroxide deposition of uranium from these solutions to obtain uranyl peruranate.

In the future, to the combined mother liquor with iron and PD, you can first apply the hydroxide precipitation of iron together with fragmentation rare-earth elements, technetium and partially strontium and their separation from the solution. Then, ferrocyanide precipitation of cesium and strontium residues can be carried out from the remaining solution. Positive research results in this direction have already been obtained.

Thus, according to the invention, a new alternative approach to SNF reprocessing is developed, a new soft radiation-stable medium is found for dissolving oxide uranium and MOX fuel, UO ₂ can be dissolved at room temperature in aqueous weakly acidic solutions of FeCl ₃ × 6H ₂ O or Fe ( NO ₃ ) ₃ × 9H ₂ O with almost the same efficiency as in a strongly acidic medium - 6M HNO ₃ + 0.1M HF, while a solid solution of PuO ₂ (4.6 wt.%) In UO ₂ (MOX- fuel) dissolves in the same way as pure uranium dioxide in solutions containing nitrate or chlorine Horde Fe (III). Upon dissolution of the MOX fuel, uranium and plutonium go into solution in the form of U (VI) and Pu (III), respectively.

The method also allows the separation of the bulk of technetium remaining in the form of sediment.

The claimed approach to the dissolution of uranium oxides, substandard and spent oxide nuclear fuel and MOX fuel in a solution of Fe (III) nitrate or chloride provides the prerequisites for the development of further processing technology for the resulting solutions, which includes the stages:

(1) separation from technetium; (2) the separation of uranyl peruranate from the solution (in the presence of Pu, it will be released together with uranium); (3) the allocation from the remaining solution of iron and fission products for their immobilization in the form of matrices suitable for long-term environmentally safe storage. The described stages practically do not increase the volume of the solution obtained after dissolving UO ₂ , therefore, with this approach to SNF reprocessing, it is possible to achieve a significant reduction in the volumes of radioactive waste characteristic of the Purex process and reduce the risks that arise in the Purex process technology, which ultimately allows to achieve the necessary for the further development of the nuclear industry of the Russian Federation.

Table 1 Found in uranium solution in terms of UO ₂ The molar ratio of UO ₂ to FeCl ₃ × 6H ₂ O or Fe (NO ₃ ) ₃ × 9H ₂ O 1: 0.5 1: 1 1: 2 1: 5 mg % the weight* mg % the weight* mg % the weight* mg % the weight* fifteen 25 29th 48 61 one hundred 61 one hundred

Claims (2)

1. A method of dissolving substandard and / or spent nuclear fuel, characterized by using an aqueous solution of an iron (III) salt as a solvent, the dissolution being carried out at a pH of 1.0-1.4 and a molar ratio of fuel to iron salt of 1: 2 1-2.5. 2. The method according to claim 1, characterized in that the iron salt is chloride or iron (III) nitrate.

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