RU2576530C1 - Method of cleaning uranium products from treatment of spent nuclear fuel from ruthenium - Google Patents

Abstract

FIELD: chemistry.

SUBSTANCE: disclosed method includes cleaning nitric uranium products from the treatment of spent nuclear fuel from ruthenium, containing up to 300 g/l of uranium and 40 g/l of nitric acid, on a solid-phase catalyst in the presence of a hydrazine nitrate reducing agent with the concentration of up to 10 g/l.

EFFECT: achieving the decontamination factor of nitric acid uranium solutions from ruthenium isotopes higher than 85.

5 cl, 1 dwg, 2 tbl, 2 ex

Description

The invention relates to radiochemical technology and can be used to purify ruthenium isotopes from nitric acid uranium solutions in spent nuclear fuel (SNF) processing technology.

Ruthenium isotopes are formed as a result of fission of uranium and plutonium nuclei in the reactors of nuclear power plants, submarines, ships. Most of the radioactive isotopes of ruthenium are short-lived, but the two main radioactive isotopes - ruthenium-103 and ruthenium-106 - have fairly long half-lives (39.3 and 373.6 days, respectively). Ruthenium isotopes make up 30% of the total mass of all fission fragments, the nuclei of which capture neutrons and break the chain reaction, create radiation levels that are significantly higher than permissible. The extremely high migration ability of ruthenium isotopes creates the risk of radioactive contamination. The problem of cleaning is the decontamination of equipment, work clothing, etc. - from radio ruthenium also has its own specifics. Depending on the chemical state of radio ruthenium, it can either be easily washed and removed, or it can be deactivated with great difficulty.

In nitric acid solutions, ruthenium exists in the form of various nitrate, nitro, and mixed nitratonitro complexes of trivalent ruthenium nitrosyl (RuNO) ³⁺ . They are combined with hydroxy and aqualigands and have the general formula [RuNO (NO ₃ ) _x (NO ₂ ) _y (OH) _z (H ₂ O) _5-xyz ] ^3-xyz . Ruthenium nitrosyl nitrate complexes are described by the general formula [RuNO (NO ₃ ) _x (H ₂ O) _5-x ] ^3-x , where x = 3.2.1.0; and nitro complexes - by the formula [RuNO (NO ₂ ) _x (H ₂ O) _y (OH) _z ] ^3-xz , where x + y + z = 5.

Despite the fact that the issues of separation of the main amount of fragmentation elements during the processing of spent nuclear fuel have been completely resolved, the problem of separation of radio-ruthenium does not have a complete solution and has not been actualized earlier due to its small amount in the spent nuclear fuel received for processing and a rather short half-life of the main isotopes . The polyvalency and variety of forms of radio-ruthenium (including polymer and colloidal) affects the distribution coefficients and may reduce the efficiency of the extraction of the target product in the Purex process.

Existing methods of purification from radio-ruthenium, such as ion-exchange, precipitation, extraction, biological purification, oxidative distillation, do not allow to obtain high purification coefficients of technological products due to the variety and complexity of various radio-ruthenium compounds in nitric acid solutions. Modeling a process with stable ruthenium is complicated by the lack of the ability to obtain identical forms. For this reason, a single mechanism for the separation of ruthenium from nitric acid media, which is characteristic of the spent fuel reprocessing technology, has not yet been proposed. A partial solution to the problem is to increase the exposure time of spent nuclear fuel before radiochemical reprocessing.

The prior art method for the purification of uranium hexafluoride from ruthenium compounds, which consists in organizing the contact of the initial solution with an inorganic sorbent - ferric fluoride. As a result, radio ruthenium is selectively extracted from the solution [Patent RU 2068287, A62D 001/00, 10.27.1996]. The disadvantages of this method are: the need to transfer uranium to hexafluoride, providing a vacuum (56.5 mm Hg), which complicates the instrumentation of the process, increasing the hydrodynamic resistance of the column while ensuring the height of the bulk layer of a granulometrically inhomogeneous sorbent (400-1100 mm) , a relatively low degree of purification (not more than 73.2%), the probability of contamination of the target product with iron compounds.

Closest to the claimed method is a method for processing irradiated nuclear fuel [Patent RU 2295167, G21C 19/46, 03/10/2007] (prototype), including the operation of purifying uranium reextracts (up to 300 g / l) from ruthenium at the stage of precipitation of ammonium polyuranates with ammonia in the presence of thiocarbonate diamide. After filtering, the radio-ruthenium is separated together with the mother liquor. The disadvantages of this method include the presence of a filtering operation, which is lengthy (limiting) in the purification of uranium from ruthenium, in addition, the use of a paper filter complicates the technological adaptation of this method; The disadvantage is the introduction of an additional reagent (diamide thiocarbonic acid). The ongoing processes of coprecipitation (trapping by sediment) and adsorption of radio ruthenium on the surface of the solid phase do not allow to achieve uranium purification with obtaining purification factors of more than 5.

The objective of the invention is to develop a technologically suitable method for purifying technological uranium nitrate products from spent nuclear fuel reprocessing from radio-ruthenium.

The technical result of the invention is to achieve ratios of purification of uranium nitrate from ruthenium isotopes of more than 85. Purification ratios are obtained after analysis of residual amounts of the ruthenium-106 isotope.

To achieve the specified technical result in a method for purifying technological uranium products from spent nuclear fuel reprocessing from ruthenium, the purification is carried out by separating radio mercury into the solid phase to a solid-phase catalyst in dynamic mode in a column-type thermostatic apparatus of continuous operation in the presence of a reducing agent.

In a particular case, a platinum catalyst supported on a VP-1AP anion-exchange resin with a mass content of platinum up to 2% is used as a solid-phase catalyst.

In a particular case, hydrazine nitrate with a concentration of up to 10 g / l is used as a reducing agent, since it has reducing properties sufficient for catalytic activation and is suitable for radiochemical production, since it is not a salt-forming agent; it can be completely destroyed on a solid-phase catalyst with the formation of products that do not pollute solutions.

In the particular case of catalytic purification, the diameter / height ratio of the bulk catalyst layer in the catalytic column is 1: 5, which makes it possible to achieve maximum catalyst performance.

In the particular case, the process is conducted at a temperature of 68-70 ° C. This allows you to achieve maximum ratios for cleaning solutions from ruthenium.

The possibility of implementing the proposed method is confirmed by studies on the manufactured laboratory catalytic column depicted in FIG. 1. The column is a vertical thermostatic apparatus with a lower solution flow, having a laminar flow zone, a catalysis zone, a gas separation zone. The catalysis zone is cut off by mesh partitions. The prepared catalyst is poured through the upper discharge hatch and sealed by the upper screen. The pore volume of the granular catalyst layer in this case is 5.1-12.5% of the bulk volume. The catalysis zone is a vertical cylindrical column filled with a catalyst in the ratio "diameter / height" 1: 5. The catalyst is a porous support uniform in particle size distribution (anion exchange resin with a grain size of 0.1-0.7 mm) having an active surface area of 2.0-35.0 m ² / g, with an ultrafine homogeneous platinum layer deposited, which is a catalyst quantitative reduction of anionic forms of radio ruthenium on the surface of the catalyst.

The proposed method is implemented in the following sequence: the catalyst is prepared by impregnating the VP-1AP resin with an alkaline solution of platinum hydroxide, the prepared catalyst is placed in a column, the column is thermostated. By means of a metering pump, the initial solution is fed to the catalytic column, and a cleaning process is carried out. The solution after the catalytic column is collected in portions and analyzed for the content of isotopes of ruthenium, nitric acid, hydrazine nitrate, uranium (4+).

At least 70% radio-ruthenium removal is carried out during catalyst regeneration in the activation mode by passing a 3M solution of nitric acid through a granular layer at a temperature of 78 ° C and a flow rate of 5-7 kb / h. The remaining amount of radio ruthenium (up to 30%) is recovered during pyrochemical catalyst regeneration.

Example 1

A platinum catalyst was prepared deposited on VP-1AP anion-exchange resin with a mass content of platinum up to 2%. The initial solution was fed to a catalytic column (see table 1).

The contact time of the initial solution with the catalyst is 55-70 s. The process temperature is 68 ° C. The flow rate of the initial solution is 7 columns. r / h The experiment was carried out in the above sequence.

The residual volume specific activity of the solution for ruthenium-106 after purification was less than 793 Bq / L, the degree of extraction of ruthenium was 98.82%, and the purification coefficient was more than 85. The experimental results are presented in table 1.

Example 2

The process of purification from ruthenium was carried out analogously to example 1, however, the concentration of uranium in the initial solution was 70 g / l, nitric acid - 5 g / l.

The residual volume specific activity of the solution for ruthenium-106 after purification was less than 327 Bq / L, the degree of extraction of ruthenium was 99.51%, and the purification coefficient was more than 206. The experimental results are presented in table 2.

The proposed method has the following advantages over the prototype: the continuity and high productivity of the process, the ability to isolate radio ruthenium concentrate, a high degree of extraction of ruthenium from the solution - more than 99%, high cleaning coefficients - more than 85.

The technical result of the invention, namely the achievement of coefficients of purification of nitric acid uranium solutions from ruthenium isotopes of more than 85, allows us to judge the possibility of introducing this invention (method) into the technology for processing spent nuclear fuel for purification from radioradiation of products in the process of uranium regeneration.

Claims (5)

        1. The method of purification of technological uranium products of the spent nuclear fuel reprocessing from ruthenium, characterized in that the purification is carried out by separating the radio-mercury in the solid phase to the solid-phase catalyst in dynamic mode in a column thermostatic apparatus of continuous type in the presence of a reducing agent.         2. The method according to p. 1, characterized in that as a solid-phase catalyst using a platinum catalyst deposited on an anion exchange resin VP-1AP with a weight content of platinum up to 2%.         3. The method according to p. 1, characterized in that as a reducing agent use hydrazine nitrate with a concentration of up to 10 g / L.         4. The method according to p. 1, characterized in that when cleaning the ratio "diameter / height" of the bulk catalyst layer in a thermostatic apparatus of continuous type column type is 1: 5.         5. The method according to p. 1, characterized in that the process is conducted at a temperature of 68-70 ° C.

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