JPH07140293A - Recovery method of transuranic element from high level radioactive waste liquid - Google Patents

Abstract

PURPOSE:To provide a method of recovering transuranic elements efficiently without requiring an Li reducing process, without using cadmium for a positive electrode, without requiring a denitration process and a multistage extraction process and without generating a larger quantity of solid-liquid waste. CONSTITUTION:Oxalic acid 2 of concentration not less than twice the concentration required to recover alkaline earth metal elements, transuranic elements and rare earth elements in high level radioactive waste liquid is added in the state of solid to the high level radioactive waste liquid and matured at the temperature in a range of about 60 deg.C to 100 deg.C so as to precipitate the transuranic elements 9, rare earth elements 8 and alkaline earth metal elements 10 as oxalate, thus separating these elements from alkaline metal and white metal elements. The precipitated oxalate is transformed into a chloride in an atmosphere containing chlorine, and the chloride is dissolved in molten salt and subjected to electrolytic purification using an insoluble positive electrode formed of material existing stably until chlorine generating potential and an insoluble negative electrode so as to deposit the transuranic elements 9 to the negative electrode for separative recovery.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for recovering transuranium elements from high-level radioactive liquid waste, and more particularly to plutonium, americium, and curium from high-level radioactive liquid waste generated from a fuel reprocessing facility of a nuclear fuel cycle. Method for recovering various transuranium elements.

[0002]

2. Description of the Related Art Conventionally, as a method for treating high-level waste liquid generated from a PUREX method reprocessing plant, a vitrified body such as plutonium, americium, and curium, which has a long half-life and is highly toxic, which does not separate transuranic elements is used. Disposal in the deep ground is considered. Therefore, if the hazard coefficient of the vitrified body containing transuranium elements such as plutonium, americium, and curium is controlled to be equal to the hazard coefficient of natural uranium, it is 10
A management period of ⁵ years or more is required.

However, from high-level waste liquid to plutonium,
9 transuranic elements such as americium and curium
When 9.9% is recovered and vitrified and stored, it takes a control period of the order of 10 ² years to have the same hazard coefficient as natural uranium. Therefore, from the viewpoint of shortening the storage period of vitrified solids and exposure control, it is desired to develop a method for separating and disposing transuranic elements such as plutonium, americium, and curium from high-level waste liquid.

Wet methods and dry methods are known as methods for recovering transuranium elements such as plutonium, americium, and curium from high-level waste liquid. The wet method is a method of recovering transuranium elements from a high-level waste liquid by combining a solvent extraction method, a cation exchange separation method and a precipitation method. The dry method is a method in which high-level waste liquid is denitrated by microwaves to be converted into oxides, then converted to chlorides, and then electrolytically refined in a molten salt.

As the wet method, a method in which a solvent extraction method using an organic solvent and a cation exchange separation method are used in combination is known (Japanese Patent Laid-Open No. 61-30799). That is, high-level waste liquid is brought into contact with tributyl phosphate (TBP) to separate uranium / plutonium into the organic phase, denitration with formic acid, and then DID
Americium and curium are separated by extraction with PA and further purification by cation exchange separation. In these steps, there are problems that the process itself cannot be made compact because a multi-step operation is required and a large amount of organic liquid waste generated from the solvent extraction step is generated.

The wet method includes a precipitation method as well as a solvent extraction method and an ion exchange method. The precipitation method is a method for rapidly recovering transuranium elements such as plutonium, americium, and curium from high-level waste liquid. An example is the coprecipitation method in which an excess amount of rare earth element is added to a high-level liquid waste, and the transuranic element is precipitated along with the precipitation of the rare earth element (see the column for analytical chemistry).

However, the method of recovering transuranium elements such as plutonium, americium, and curium by coprecipitation has a problem that a new salt waste is produced. Therefore, as a method of recovering transuranium elements such as plutonium, americium, and curium, there is an oxalate precipitation method in which oxalic acid is added to high-level waste liquid to precipitate and separate transuranium elements such as plutonium, americium, and curium. Attention has been paid.

[0008] L. Cecille et al. Of Ispra Research Institute of JRC reported that trivalent transuranium elements present in high-level liquid waste are
It is reported that the oxalate can be recovered in a yield of 99.0 to 99.8% (IAEA-SM-246 / 2).
2). These experiments were carried out at nitric acid concentrations of 0.5 to 1.2 mol / l. It is considered that the nitric acid concentration after intermediate storage of the high-level liquid waste is about 2.0 mol / liter, so it is considered that the high recovery rate is achieved by lowering the nitric acid concentration of the liquid waste.

Similar attempts are being made in Japan. Kobayashi et al. From JAERI, reported that high-level simulated waste liquid (nitric acid concentration 2.
Formic acid was added to 0 mol / liter to denitrate (pH 0.
5), by the method of adding oxalic acid,
It has been reported that 9.9% of simulated transuranic elements can be recovered (JAERI M 88-168).

As a method of reducing the nitric acid concentration, a method of adding formic acid has been considered, but there is a problem associated with a process of filtering the precipitated element after adding formic acid.

From the above, in consideration of the amount of waste and the simplicity of the process, 99% transuranium elements such as plutonium, americium and curium can be directly extracted from the high level waste liquid by the oxalate precipitation method without denitration. ．
It is necessary to develop a method for recovering about 9%.

However, although the precipitation method can recover transuranic elements, it cannot separate rare earth elements having similar chemical properties to those of transuranic elements. Therefore, in order to remove the rare earth element from the transuranium element recovered by the precipitation method, a method in which the recovered oxalate precipitate is made into a solution and then separated by a solvent extraction method or an ion exchange method is being studied.

In the dry method, all of the high-level waste liquid is denitrified by microwaves to form oxides, which are then converted into chlorides, and then KCl.
A method of dissolving in a molten salt of LiCl and adding a reducing agent such as Li to the molten salt to extract it in a cadmium bath, and then performing electrolysis using the cadmium bath as an anode to recover the transuranium element on the cathode. Known (Central Research Institute of Electric Power Industry, General Report, T15, 1990). Compared with the wet method, the efficiency of separating rare earth elements that have similar chemical properties to transuranium elements is low, but compared to the wet method, the process itself does not have multiple stages and liquid waste is not generated. There is.

However, in the currently proposed dry method, a large amount of chlorine is required for chloride conversion in order to convert the high-level waste liquid into chloride after converting all of the high-level waste liquid into oxide, and before performing electrolysis. There are problems that a large amount of Li as a reducing agent is necessary to reduce the transuranium element in the molten salt of KCl-LiCl, and that a large amount of cadmium used as the anode is also necessary. Therefore, it is desired to develop a method for recovering transuranium elements that does not use cadmium for the Li reduction step or the anode.

From the above, it is necessary to develop a method for separating and recovering the transuranium element and the rare earth element from the molten salt in consideration of the amount of waste and the simplicity of the process.

[0016]

SUMMARY OF THE INVENTION The object of the present invention is to solve the above-mentioned problems of the prior art, such as plutonium, americium, and curium from high-level radioactive waste liquid generated from a fuel reprocessing facility of a nuclear fuel cycle. It is an object of the present invention to provide a method for efficiently recovering various transuranium elements.

[0017]

In order to achieve the above object, the transuranic element recovery method according to the present invention is a nuclear fuel reprocessing facility in which alkaline earth metals, transuranic elements and rare earth elements are dissolved together with nitric acid. A method for recovering transuranium element from a high-level radioactive waste liquid, comprising the alkaline earth metal element, transuranic element in the high-level radioactive waste liquid,
And oxalic acid at a concentration more than twice the concentration required for recovering the rare earth element are added to the high-level radioactive waste liquid in a solid state, and the high-level radioactive waste liquid to which oxalic acid is added is about 60 ° C to about 100 ° C. Aging at a temperature in the vicinity temperature range, transuranic elements, rare earth elements and alkaline earth metal elements are precipitated as oxalates to separate them from alkali metal elements and white metal elements, and the precipitated oxalates are chlorinated. Is converted into chloride under an atmosphere containing, and the chloride is dissolved in a molten salt, and an insoluble anode formed of a material that exists stably up to the chlorine generation potential and an insoluble cathode are used. Electro-refining is carried out to deposit the trans-uranium element on the cathode for separation and recovery.

Further, pure water is added to the high-level radioactive waste liquid to adjust the concentration of nitric acid contained in the high-level radioactive waste liquid to 1.0 to
It is characterized in that the oxalic acid is added after the amount is adjusted to 2.0 mol / liter.

The chlorine-containing atmosphere is formed of hydrogen chloride, carbon tetrachloride, chlorine gas, sulfur chloride, phosgene, phosphorus pentachloride, ammonium chloride, or thionyl chloride, or a combination thereof. It is characterized by

The molten salt used in the electrolytic refining is a mixed salt of potassium chloride and lithium chloride, a mixed salt of lithium chloride and sodium chloride, a mixed salt of calcium chloride and potassium chloride, or lithium chloride, potassium chloride and barium chloride. And a mixed salt of calcium chloride and calcium chloride.

The anode in the electrolytic refining is
It is characterized in that it is formed of any one of carbon, tantalum, gold, platinum, molybdenum, tungsten, titanium nitride, and titanium carbide.

The cathode in the electrolytic refining is
Iron, tantalum, gold, platinum, molybdenum, tungsten,
Alternatively, it is characterized by being formed of either carbon.

The cathode in the electrolytic refining is
It is characterized in that it has a receiving portion formed of an insulator for receiving the precipitate of transuranium element in the lower portion.

[0024]

[Function] The concentration of oxalic acid in the high-level radioactive liquid waste is at least twice the concentration required to recover the alkaline earth metal element, transuranium element and rare earth element, that is, at least twice the chemical equivalent. It is added to the high-level radioactive waste liquid in the state, and the high-level radioactive waste liquid containing oxalic acid is aged at a temperature in the temperature range of about 60 ° C to about 100 ° C. Can be precipitated with. For this reason, the process can be simplified as compared with the conventional method in which after denitration with formic acid, the precipitate is filtered and oxalic acid is added to recover the transuranium element.

Since this precipitated oxalate is converted into chloride under an atmosphere containing chlorine, the chlorine required for conversion is higher than that in the conventional case where the high level radioactive waste liquid is converted into oxide and then converted into chloride. Can be reduced, and KCl-LiCl can be used before electrorefining or electrolysis.
In order to reduce the transuranium element in the molten salt, it is possible to avoid the need for a large amount of Li as a reducing agent.

Further, as in the conventional method, the high-level waste liquid is completely denitrated by microwaves to form an oxide, which is then converted into a chloride, which is then dissolved in a molten salt of KCl-LiCl to reduce Li and the like. The agent is added to the molten salt and extracted in a cadmium bath.After that, electrolysis is performed using the cadmium bath as the anode to recover the transuranium element on the cathode.Unlike the conventional case, the anode in electrorefining has a chlorine generation potential. Cadmium can be prevented from being used as the material of the anode by forming it with a stable material.

As a result, no Li reduction step is required, no cadmium is used for the anode, no denitration process or multiple extraction steps are required, and a large amount of solid liquid waste is not generated. Moreover, transuranic elements can be efficiently recovered.

[0028]

Embodiments of the present invention will be described below with reference to the drawings. FIG. 1 is a block diagram for explaining a group separation method for high-level waste liquid in the transuranium element recovery method according to the present invention. The high-level waste liquid 1 contains fission products such as alkali metals, alkaline earth metals, platinum group, rare earth elements, and transuranium elements such as plutonium produced during operation. Here, the alkali metal is, for example, cesium, the alkaline earth metal is, for example, strontium or barium, the platinum group is, for example, molybdenum, technetium, ruthenium, rhodium, palladium,
The rare earth element is, for example, cerium, europium, praseodymium.

In FIG. 1, first, oxalic acid 2 is added to the high-level waste liquid 1 to generate an oxalate salt. Next, the alkali metal element 5 and the platinum group element 6 are recovered in the filtrate 4 by the oxalate precipitation method 3, and the rare earth element 8, the transuranium element 9, the alkaline earth metal element are contained in the precipitate 7. Collect 10.

Next, the recovered precipitate 7 is converted to chloride 11
I do. Next, the rare earth element 8, the alkaline earth metal element 10, and the transuranium element 9 are separated by the electrolytic refining method 12.

The operating principle of the oxalate precipitation method 3 will be described in detail below. The transuranium element 9 exists in the state of plutonium (III), americium (III), etc. in the solution, and forms an oxalate by the reaction of oxalic acid with the formula (1).

2Pu ³⁺ +3 (COOH) ₂ → Pu ₂ (C ₂ O ₄ ) ₃ + 6H ⁺ (1) 2Am ³⁺ +3 (COOH) ₂ → Am ₂ (C ₂ O ₄ ) ₃ + 6H ⁺ Also, transuranium Rare earth elements, which have similar chemical properties to those of elements, are lanthanum (III), cerium (III),
It exists in a state such as neodymium (III) and forms an oxalate salt by the reaction of the formula (2) with oxalic acid.

2La ³⁺ +3 (COOH) ₂ → La ₂ (C ₂ O ₄ ) ₃ + 6H ⁺ 2Ce ³⁺ +3 (COOH) ₂ → Ce ₂ (C ₂ O ₄ ) ₃ + 6H ⁺ (2) 2Nd ³⁺ +3 (COOH) ₂ → Nd ₂ (C ₂ O ₄ ) ₃ + 6H ⁺ Similarly, the reaction formulas for forming cesium oxalate and strontium oxalate are shown in formula (3).

2Cs ⁺ + (COOH) ₂ → Cs ₂ (C ₂ O ₄ ) + 2H ⁺ (3) Sr ²⁺ + (COOH) ₂ → Sr (C ₂ O ₄ ) + 2H ⁺ Cesium oxalate, Strontium oxalate, Solubility of rare earth and transuranium oxalates in pure water (Robert C.Weast.Ph.D. Et al., CRC Handbookof Chem
istry and Physics, 66th, B68-B161 (1985)) and solubility products (see Chemistry column) are shown below.

As rare earth elements, lanthanum, cerium,
Neodymium value reported, solubility 10 ^-5 g / 100 ml
, The solubility product is 10 ^-27 (mol / liter)
^It is reported as ^{m + n} . On the other hand, the solubility product of plutonium, which is a type of transuranium element, is 4-5 × 10 ^−22.
And the solubility product of the rare earth element is almost the same order.

The solubility of strontium oxalate is 5.1 × 10 ^-3 g / 100 ml, and the solubility product is 1.6 × 10 ^-7.
(Mol / liter) ^{m + n,} which is relatively close to that of rare earth elements and transuranium elements.

On the other hand, the solubility of cesium oxalate is 282 g / 100 ml, which is higher than those of rare earth elements and transuranium elements. Therefore, when oxalic acid is added to the solution, the oxalates of rare earth elements, transuranium elements, and strontium oxalates, which have a relatively low solubility and solubility product of oxalates, and alkaline earth metal elements, such as strontium oxalate, which have a relatively low solubility product. Will precipitate.

FIG. 2 shows iron, assuming an actual high-level waste liquid,
The results of investigating the effect of precipitation aging temperature using a high-level simulated waste liquid prepared with 15 elements of nickel, chromium, sodium, rubidium, cesium, strontium, barium, zirconium, tellurium, molybdenum, ruthenium, rhodium, platinum, plutonium Show. The nitric acid concentration was diluted 2-fold to 1 mol / liter, and the amount of oxalic acid added was 3 mol per liter of high-level simulated waste liquid.

The reaction between plutonium and oxalic acid in the solution is carried out according to the equation (4).

2Pu ³⁺ + 3C ₂ O ₄ ^2- → Pu ₂ (C ₂ O ₄ ) ₃ (4) Further, the dissociation reaction of oxalic acid in a solution is represented by the formula (5).

H ₂ C ₂ O ₄ → HC ₂ O ₄ ⁻ + H ⁺ (5) HC ₂ O ₄ ⁻ → C ₂ O ₄ ^{2 −} + H ⁺ The solubility of oxalic acid (anhydrous) in pure water is 0. It is relatively insoluble at 97 mol / liter and 3.41 mol / liter at 60 ° C (see RIKEN), but the solubility tends to increase with increasing temperature. Therefore, when the precipitation aging temperature is increased, the oxalate ion in the solution is increased, the equilibrium of formula (4) is shifted to the right, and plutonium oxalate is obtained. Therefore, the precipitation aging temperature is 60 to
When the temperature is 90 ° C., the recovery rate of transuranium element can be set to 99.9%.

FIG. 3 shows the results of examining the effect of nitric acid concentration using a high-level simulated waste liquid prepared from five elements, cesium, strontium, ruthenium, cerium, and plutonium, assuming an actual high-level waste liquid.

In the experiment with a nitric acid concentration of 1 mol / liter, a high recovery rate of 99.9% can be obtained without increasing the precipitation aging temperature, but in the experiment with a nitric acid concentration of 2 mol / liter, the precipitation aging temperature was increased by 99%. A recovery rate of .9% is obtained, and the recovery rate of transuranium element tends to decrease as the nitric acid concentration increases. This is because when the acid concentration of the solution increases, the dissociation equilibrium of oxalic acid as shown in formula (5) shifts to the left and the production of oxalate ion is inhibited, so that the production of plutonium oxalate is less likely to occur. Responsible.

When the nitric acid concentration is reduced, plutonium oxalate is easily produced, but when the acid concentration is reduced, IV-valent plutonium may be hydrolyzed to form a polymer. Generating condition of plutonium polymers NO ₃ ^- the ratio of plutonium, NO ₃ ^- / plutonium, 0.8
Stable formation in the range of 4.0 (The chemistry of Plu
tonium Sol-Gel Process, Radiochimica Acta 25, p139
-148 (1978)). Therefore, it is desirable that the nitric acid concentration is 1 mol / liter or more in consideration of safety regardless of the plutonium concentration.

The composition of the coexisting elements of the high-level liquid waste changes depending on the burnup, cooling period, history, etc. of the liquid waste. Zirconium, tellurium, molybdenum, etc. have the property of being easily hydrolyzed, and when denitration is performed to reduce the nitric acid concentration to pH 0.5 (corresponding to a nitric acid concentration of about 0.3 mol / liter), 99% or more of zirconium and tellurium, 90% or more of molybdenum precipitates (JAERI-M 89-16
8). Since these elements may be mixed into the transuranium element, it is desirable that the nitric acid concentration be as high as possible.

Therefore, in order to recover the transuranium element from the high-level waste liquid, it is desirable not to denitrate and reduce the nitric acid concentration. However, considering the efficiency of recovering the transuranium element, it is desirable to lower it as much as possible.

From the above, the nitric acid concentration is 1-2 mol /
Make liter, add oxalic acid, and set precipitation aging temperature to 60
99.9% from high level waste liquid by heating up to ~ 90 ° C
It is possible to recover the above transuranium elements and further prevent the inclusion of coexisting elements.

Table 1 shows a conventional example of the recovery rate of recovering plutonium by adding oxalic acid to 20 ml of a high-level simulated waste liquid having a nitric acid concentration of 2.0 mol / liter and the experimental results of the present invention.

[0049]

[Table 1] From the above, by adding oxalic acid directly to the high-level waste liquid and adjusting the precipitation aging temperature to 80 ° C., about 99.9% of transuranic element can be recovered without denitration.

Table 2 shows the amount of high-level waste liquid generated from the Barnewell reprocessing plant. High-level waste liquid flow rate, the amount of high-level waste liquid generated is 1 than the operating time per day
It can be calculated as 8 m ³ .

[0051]

[Table 2] Table 3 shows a comparison of the addition amount of formic acid and the addition amount of oxalic acid in the conventional example and the example of the present invention in the process of recovering transuranium element from the high-level simulated waste liquid.

[0052]

[Table 3] * Simulated waste liquid composition and concentration are different from the present invention.

** Estimated weight from specific gravity of 1.22 g / cm ³ Also, Table 4 shows equipment and capacity of each process calculated from the amount of high-level waste liquid generated per day. As a conventional example, a process of denitrifying a high-level waste liquid with formic acid and then adding oxalic acid to recover the transuranium element is shown.
Further, as an example of the present invention, a method of recovering transuranic elements without denitration by adding oxalic acid directly to a high-level waste liquid and setting the precipitation aging temperature to 80 ° C. is shown. In the conventional example, the amount of formic acid and the amount of oxalic acid added to the high-level waste liquid were calculated with reference to the report of Ispra.

[0054]

[Table 4] As is clear from Table 3, when the examples of the present invention and the conventional examples are compared, almost the same simulated transuranium element recovery rates are achieved. However, when comparing the amount of the reagent added per liter of high-level waste liquid, the number of examples of the present invention is reduced as compared with the conventional example. Further, when the equipment and capacity of each process are calculated from the high-level waste liquid generation amount per day, the number of processes can be reduced to 5 steps in the embodiment of the present invention, whereas the conventional example has 8 steps.

Further, comparing the installed capacities, it is 118 m ³ (excluding the filter) in the conventional example, whereas it is 86 m ³ (excluding the filter) in the embodiment of the present invention. Therefore, if the facility cost is simply calculated by comparing the facility capacities, the cost of the embodiment of the present invention can be expected to be reduced by about 30% as compared with the conventional example.

FIG. 4 shows a block diagram of a conventional example (a) and the present invention (b) regarding a method for recovering transuranium elements by carrying out potentiostatic electrolysis using a KCl-LiCl molten salt. In the conventional example (a), after the chloride 13 of transuranic element generated in the chloride conversion step is dissolved in the molten salt 14, cadmium 15 is added to prepare the bath 16. After that, zero-valent metal lithium 17 is added to the cadmium 15 anode to reduce 18 the elements in the bath. After that, electrolysis 19 is performed to recover the transuranium element 21 as a precipitate 20. Elements other than transuranium element 21 are present in bath 22.

In the present invention (b), chloride 1 is added to molten salt 14.
Bath preparation 16 is carried out by adding 3. Next, electrolysis 19 is performed using a carbon anode and an iron cathode, and the transuranium element 21 is recovered in the precipitate 20. Elements other than transuranium element 21 are present in bath 22.

Next, the working principle of the electrolytic refining method for performing the electrolysis 19 will be described in detail below. The transuranium element exists in the molten salt in the form of plutonium (III), americium (III), neptunium (III), etc., and produces plutonium, americium, and neptunium metal by the reaction of the formula (6) on the cathode.

Pu ³⁺ + 3e ⁻ → Pu Am ³⁺ + 3e ⁻ → Am (6) Np ³⁺ + 3e ⁻ → Np Further, a rare earth element having a chemical property similar to that of transuranium is lanthanum ( III), cerium (III),
Exists in a state such as neodymium (III) and on the cathode (7)
The reaction of the formula produces lanthanum, cerium and neodymium metals.

La ³⁺ + 3e ⁻ → La Ce ³⁺ + 3e ⁻ → Ce (7) Nd ³⁺ + 3e ⁻ → Nd Similarly, the alkali metal element and the alkaline earth element are cesium (I) and strontium in the molten salt. It exists in a state such as (II), and produces cesium and strontium metal by the reaction of the formula (8) on the cathode.

Cs ⁺ + e ⁻ → Cs (8) Sr ²⁺ + 2e ⁻ → Sr The selectivity of these reactions is determined by the redox potential. The redox potential varies depending on the valence of an element and between elements. Taking neptunium as an example of transuranium elements, neptunium (III) / neptunium (0)
The redox potential of is based on the silver / silver chloride reference electrode
It is 1.311V. Taking the example of neodymium as the rare earth element believed to migrate during oxalate precipitation, the redox potential of neodymium (III) / neodymium (0) is -2.097 V with reference to a silver / silver chloride reference electrode. is there.

Since neptunium has a higher potential than neodymium, the electrolytic potential is based on the silver / silver chloride reference electrode.
At 1.311 V, only neptunium is selectively deposited and neodymium remains in the salt. In addition, since the free energy for chloride formation of strontium, which is considered to be transferred to cesium and molten salt, is more stable than that of rare earth elements, the redox potential may be baser than that of rare earth elements. Therefore, it is easier to separate from the transuranium element than the rare earth element, and strontium remains in the salt. By utilizing such a principle and performing potentiostatic electrolysis, transuranic elements can be separated from other elements, especially rare earth elements having a small potential difference and difficult to separate.

FIG. 5 shows a conventional example (a) and the present invention (b) as a method of recovering transuranium elements by carrying out potentiostatic electrolysis using a molten KCl-LiCl salt and a reaction formula at the cathode and anode. .

In the conventional example (a), the cadmium anode 23 is arranged below the molten salt 14, and the iron cathode 24 is arranged in the molten salt. Since the operating temperature is 500 ° C, cadmium (melting point 320.9 ° C, boiling point 765 ° C) exists as a liquid. The transuranium element 25 to be recovered is a cadmium anode 2
It exists in 3 in 0 valence. When electrolysis is started, the transuranium element 25 existing in the cadmium anode 23 is eluted and exists as trivalent ions 26 in the molten salt 14, and is deposited at the cathode 24.

In the present invention (b), since the carbon anode 27 is used, the anode conventionally arranged in the lower portion is the molten salt 14
It exists in a solid state inside. Further, as the cathode 24, the iron cathode 24 is arranged in the molten salt 14 as in the conventional example. The transuranium element 25 existing as trivalent ions 26 in the molten salt 14 is deposited on the cathode 24.
In 4, the chlorine ions 28 are deprived of electrons to generate chlorine 29.

In the present invention (b), the chloride of the transuranium element produced in the chloride conversion step is dissolved in the molten salt in the molten salt, and then electrolysis is performed using a carbon anode and an iron cathode. Therefore, it is possible to recover the transuranium element from the molten salt without the need for the reduction step, the lithium necessary for the reduction step, and the need for cadmium which is present in a large amount in the anode.

In Table 5, the amount of molten salt used in the experiment, the amount of simulated transuranium element (neodymium), the amount of rare earth element (cerium),
The amount of anode and cathode and the amount of lithium used as a reducing agent are shown. Since the experiment was performed cold, neodymium was selected as the simulated transuranium element and cerium was selected as the rare earth element.
The addition amount of neodymium was 1 wt% to the molten salt, and the addition amount of cerium was 4 wt% to the molten salt. The amount of lithium added was such that neodymium and cerium could all be reduced.

[0068]

[Table 5] In the conventional example, 500 g (58 cm ² ) of cadmium used as an anode when treating the same amount of transuranium element, and 1.51 g (3 cm ² ) of lithium necessary for the reduction process.
You will need it. The total capacity of these is 61 cm ² , which occupies 30% of the electrolytic bath. Therefore, the electrolytic cell capacity of the present invention is reduced by about 30% as compared with the conventional example. From this, if the electrolytic cell construction cost is simply calculated from the comparison of the installed capacities, the embodiment of the present invention is 3 times as compared with the conventional example.
A cost reduction of 0% can be expected.

Another embodiment of the present invention will be described below with reference to the drawings. FIG. 6 shows the cathode shapes of the conventional example (a) and the present invention (b) for a method of recovering transuranium elements by carrying out potentiostatic electrolysis using a KCl-LiCl molten salt. In the conventional example (a), the iron cathode has a cylindrical shape 30, and it is difficult to recover the transuranium element because the crystal growth is difficult and the crystal growth may easily drop from the cathode. According to the present invention (b), however, the shape of the cathode is such that it is difficult for the crystal to grow and the crystal grows easily from the cathode by providing a sink made of the insulator 31 below the cathode of the conventional cylindrical type 30. Transuranic elements that may fall can be easily recovered.

Table 6 shows the results of experiments using the conventional example and the cathode of the present invention. The current efficiency was 4.2% to 1 when tested with the conventional electrolytic cell structure using cadmium as the anode.
It rose to 7.5%. Also, when a simulated transuranium element recovery experiment was conducted in the electrolytic cell using carbon of the present invention as an anode, the current efficiency increased to 41.9% and the recovery rate was 10.
It rose about twice.

[0071]

[Table 6] next. Another embodiment of the present invention will be described below with reference to the drawings. FIG. 7 shows the cathode shapes of the conventional example (a) and the present invention (b) for a method of performing constant potential electrolysis using a KCl-LiCl molten salt to recover the transuranium element. In the conventional example (a), the iron cathode has a cylindrical shape 30, and it is difficult to recover the transuranium element because the crystal growth is difficult and the crystal growth may easily drop from the cathode. However, according to the present invention (b), the shape of the cathode is such that by providing a sink made of the insulating material 31 below the flat iron cathode 32, it is difficult for the crystal to grow and the crystal grows and easily falls from the cathode. Possible transuranic elements can be easily recovered.

FIG. 8 shows a layout of the cathode and anode of the conventional example (a) and the present invention (b) in the electrolytic cell for electrolysis using the cathode of the present invention. In the conventional example (a), electrolysis is performed using a columnar anode 33 and a flat plate cathode 32, and a flat plate cathode (lower insulation) 34 in which a sink made of an insulator 31 is provided below. Since the electrolytic solution has a certain resistance, the current flowing through the electrolytic solution tends to take a path as short as possible. In the conventional example (a), only the lower part of the cylindrical anode 33 is involved in the reaction, so that the electrolytic capacity per unit time is limited.

However, in the present invention (b), the flat plate type anode 35 is used.
Therefore, the distance from the cathode can be kept constant, and the electrode area involved in the reaction can be increased. Therefore, the electrolysis capacity per unit time can be increased.

Table 7 shows the electrolysis time in the electrolytic cell for electrolysis using the anode of the conventional example and the anode of the present invention. In order to make the cost of producing the electrolytic cell the same, electrolysis was performed using the same size (capacity) of the anode. To treat 1 kg of transuranium element, 100 liters of electrolytic cell capacity is required (diameter 0.2
5 m, height 0.25 m). If a cubic anode with a side of 0.05 m and a height of 0.2 m is used, the area of the anode used for electrolysis is 0.002 m ² in the conventional example. On the other hand, in the present invention, the anode area used for electrolysis is 0.0
It will be 1 m ² . Comparing the present invention with the conventional example, if the current per unit area is constant, the current will be 5 times, and the electrolysis time required to recover 1 kg of transuranium element will be 1/5 time, so the electrolysis time will be greatly increased. It can be shortened.

It is also possible to perform electrolysis with a columnar or rectangular parallelepiped anode instead of the cubic anode.

[0076]

[Table 7]

[0077]

As described above, according to the constitution of the present invention, the present invention has the following effects. (1) Plutonium, americium,
Since transuranic elements such as curium are collected and disposed of, the management period after vitrification can be shortened to the order of hundreds to thousands of years. (2) DIDPA or TBP generated by a conventional transuranic element recovery method such as plutonium, americium, and curium, which has a solvent extraction method using an organic solvent as a main step.
Since it does not generate organic liquid waste such as, it is possible to reduce the main process and the associated waste treatment process. (3) Formic acid is added because it does not require a denitration step as compared with the conventional transuranium element recovery methods such as plutonium, americium, and curium, in which formic acid is added to the high-level waste liquid and then denitration is performed, and then oxalic acid is added. The process is compact and unnecessary. (4) Compared with the conventional molten salt electrolysis method using cadmium as an anode, since carbon is used as an anode, cadmium waste is not generated and the amount of waste can be greatly reduced. (5) Compared to the conventional molten salt electrolysis method that requires a reduction step using lithium, a reduction step is not required, so that it is not necessary to add lithium or the like, and waste can be greatly reduced. (6) The transuranic element, which is difficult to recover, can be easily recovered by using the crucible type metal electrode, and the electrolysis efficiency can be increased.

[Brief description of drawings]

FIG. 1 is a block diagram for explaining a group separation method for high-level waste liquid in an embodiment of a transuranium element recovery method according to the present invention.

FIG. 2 is a graph showing the effect of precipitation aging temperature on the transuranic element recovery rate.

FIG. 3 is a graph showing the influence of nitric acid concentration on the transuranic element recovery rate.

FIG. 4 is a block diagram of a conventional example (a) and the present invention (b) regarding a method of recovering transuranium elements by performing potentiostatic electrolysis in a molten salt.

FIG. 5 is a diagram showing a constitutional view of a conventional example (a) and the present invention (b) and a reaction formula at an anode / cathode of a method for recovering transuranium element by performing potentiostatic electrolysis in a molten salt.

FIG. 6 shows a method for recovering transuranium element by potentiostatic electrolysis in molten salt, and FIG. 5 shows the cathode shapes of the conventional example (a) and the present invention (b) by potentiostatic electrolysis in molten salt. Regarding the method of recovering the conventional method (a) and the present invention (b)
Sectional views (a1), (b1) and plan views (a2), (b2) showing the cathode shape of FIG.

FIG. 7 is a sectional view (a1), (b1) and a plan view (a) showing arrangements of an anode and a cathode of a conventional example (a) and the present invention (b) in an electrolytic cell for electrolysis using the cathode of the present invention.
2), (b2).

FIG. 8 is a cross-sectional view (a), (b) showing the arrangement of the cathode and the anode of the conventional example (a) and the present invention (b) in the electrolytic cell for electrolysis using the cathode of the present invention.

[Explanation of symbols]

 1 High-level waste liquid 2 Oxalic acid 3 Oxalate precipitation method 4 Filtrate 5 Alkali metal element 6 Platinum group element 7 Precipitate 8 Rare earth element 9 Transuranium element 10 Alkaline earth metal element 11 Chloride conversion 12 Electrolytic refining 13 Chloride chloride Material 14 Molten salt 15 Cadmium 16 Preparation of bath 17 Lithium 18 Reduction 19 Electrolysis 20 Precipitate 21 Transuranium element 22 Bath 23 Cadmium anode 24 Cathode 25 Transuranium element (0 valent) 26 Transuranium element (trivalent) 27 Carbon anode 28 Chlorine Ion 29 Chlorine 30 Cylindrical cathode 31 Insulator 32 Flat cathode 33 Cylindrical anode 34 Flat cathode (lower insulation) 35 Flat anode

Claims (7)

[Claims] 1. A method for recovering transuranic elements from a high level radioactive waste liquid of a nuclear fuel reprocessing facility in which alkali metals, alkaline earth metals, transuranium elements, rare earth elements and white metal elements are dissolved together with nitric acid. , Adding oxalic acid in a solid state to the high-level radioactive waste liquid at a concentration twice or more the concentration required to recover the alkaline earth metal element, transuranium element and rare earth element in the high-level radioactive waste liquid, The high-level radioactive liquid waste added with oxalic acid is aged at a temperature in the temperature range of about 60 ° C to about 100 ° C to precipitate the transuranium element, the rare earth element and the alkali metal earth element as an oxalate salt, and then the alkali metal Separated from elements and white metal elements, the precipitated oxalate is converted to chloride under an atmosphere containing chlorine, and the chloride is dissolved in molten salt. Then, electrolytic refining is performed using an insoluble anode formed of a material that stably exists up to the chlorine generation potential and an insoluble cathode, and transuranic elements are deposited on the cathode and separated and recovered. A method for recovering transuranium elements. 2. The concentration of nitric acid contained in the high-level radioactive liquid waste is 1.0 to 2.0 by adding pure water to the high-level radioactive liquid waste.
The method for recovering transuranic element according to claim 1, wherein the oxalic acid is added after the mol / liter. 3. The atmosphere containing chlorine is formed of hydrogen chloride, carbon tetrachloride, chlorine gas, sulfur chloride, phosgene, phosphorus pentachloride, ammonium chloride, or thionyl chloride, or a combination thereof. The transuranic element recovery method according to claim 1 or 2, characterized in that. 4. The molten salt in the electrolytic refining is a mixed salt of potassium chloride and lithium chloride, a mixed salt of lithium chloride and sodium chloride, a mixed salt of calcium chloride and potassium chloride, or lithium chloride, potassium chloride and barium chloride. 4. The transuranic element recovery method according to claim 1, wherein the transuranic element recovery method is any one of mixed salts of calcium chloride and calcium chloride. 5. The anode in the electrolytic refining is carbon, tantalum, gold, platinum, molybdenum, tungsten,
The transuranic element recovery method according to any one of claims 1 to 4, wherein the transuranic element recovery method is formed of either titanium nitride or titanium carbide. 6. The cathode in the electrolytic refining is formed of any one of iron, tantalum, gold, platinum, molybdenum, tungsten, or carbon. The method for recovering transuranium element described in. 7. The cathode in the electrolytic refining has a receiving portion formed below it with an insulator for receiving a precipitate of transuranium element. The method for recovering transuranium element according to Crab.