JPH05100085A - Separating/recovering method for technetium - Google Patents

Abstract

PURPOSE:To separate and recover the technetium in a solution such as a radioactive waste liquid generated in a reprocessing facility of the nuclear spent fuel. CONSTITUTION:An radioactive waste liquid 1 incorporating technetium is brought into contact with a recovery liquid 3 via an anion exchange membrane 2 to perform the diffusing/dialyzing process, and the technetium in the radioactive waste liquid 1 is separated and recovered into the recovery liquid 6. The recovery liquid 6 is reduction-processed to dialyze technetium as an oxide, and it is separated and recovered from the recovery liquid 6.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for separating and recovering technetium from a solution such as waste liquid generated in a reprocessing facility for spent nuclear fuel.

[0002]

2. Description of the Related Art At a spent nuclear fuel reprocessing facility, a nitric acid aqueous solution and an organic solvent are used to separate fission products and transuranium elements in the spent nuclear fuel into a high-level waste liquid,
Recover uranium and plutonium as products. In such a spent nuclear fuel reprocessing facility, in addition to the high-level waste liquid, uranyl nitrate solution, plutonium nitrate solution, low-level waste liquid, and various other solutions are handled. One of the fission products, technetium, is present in the spent nuclear fuel, but technetium exhibits complicated chemical behavior, so it is handled in the spent nuclear fuel reprocessing facility as described above and more or less in the solution. Contains technetium.

Since this technetium has an extremely long half-life, there is a possibility that it will have an impact on the environment when a high-level waste liquid is disposed of as a vitrified substance. It is desired to separate and recover technetium from these solutions because it may be transferred to the gas phase in various steps and can be effectively utilized as a valuable element that does not exist in nature.

[0004]

[Problems to be Solved by the Invention]
Due to its extremely long half-life, it may have an environmental impact when disposing of high-level waste liquid as vitrified solids, and due to the presence of volatile chemicals, it may enter the gas phase in various processes of reprocessing facilities. It is desired to separate and collect technetium in these solutions because of the possibility of migration and the effective utilization as a valuable element that does not exist in nature.

The present invention has been made under the above circumstances, and an object thereof is to provide a method for separating and recovering technetium in a solution such as a waste liquid generated in a reprocessing facility for spent nuclear fuel.

By the way, in the case of plutonium nitrate solution,
It is known that an insoluble plutonium polymer is produced when the concentration of free acid in the solution is low. (For example, Kazuse Kiyose, "Nuclear Chemical Engineering IV Volume IV Chemical Reprocessing and Radioactive Waste Management Chemical Engineering" Showa December 29, 58, published by Nikkan Kogyo Shimbun, page 150). When such a polymer is formed, it not only clogs the process pipe, but also may cause a critical safety problem due to the unexpected accumulation of plutonium.Therefore, when technetium in plutonium nitrate solution is separated and recovered by a membrane, etc. Therefore, it is necessary to prevent the free acid concentration in the solution from falling below the concentration at which the plutonium polymer is produced.

In order to deal with such a point, the present invention further reduces the concentration of free acid in the solution to be treated when separating and recovering technetium in a solution such as a waste liquid generated in a reprocessing facility for spent nuclear fuel. It is an object of the present invention to provide a method for separating and recovering technetium without using any method.

[0008]

That is, the method for separating and recovering technetium of the present invention comprises contacting a solution containing technetium with a recovery solution through an anion exchange membrane, and diffusively dialysis the technetium existing in the form of anion. It is characterized in that it is separated and recovered in a recovery liquid by.

Further, in the method for separating and recovering technetium of the present invention, a solution containing technetium is brought into contact with a recovery solution through an anion exchange membrane, and technetium existing in the form of an anion is separated into the recovery solution by diffusion dialysis. After the recovery, the recovery liquid containing technetium is reduced to convert the technetium into an oxide, which is separated and recovered from the recovery liquid.

Furthermore, the method for separating and recovering technetium of the present invention comprises bringing a solution containing technetium into contact with a recovery solution through an anion exchange membrane, and converting technetium existing in the form of anion into the recovery solution by diffusion dialysis. When separating and collecting, an acidic solution is used as a recovering liquid.

[0011]

In the method for separating and recovering technetium of the present invention, a solution containing technetium such as radioactive waste liquid generated in a used reprocessing facility and a recovery solution are contacted through an anion exchange membrane for diffusion dialysis, Technetium forming an anion called pertechnetate (TcO ₄ ⁻ ) can be selectively separated and recovered from fission products other than uranium, plutonium and technetium forming cations. Further, the technetium thus separated in the recovered liquid is converted from pertechnetate ion to technetium oxide (TcO ₂ ) by performing reduction treatment such as electrolytic reduction treatment, and is precipitated or filtered as an oxide solid. It can be separated and recovered from the recovery liquid by, for example,

Furthermore, in the case of a plutonium nitrate solution generated in a spent nuclear fuel reprocessing facility, it is known that an insoluble plutonium polymer is produced when the concentration of free acid in the solution is low. The technetium is recovered by using an acidic solution as the recovery solution in contact with the anion-exchange membrane without lowering the free acid concentration in the solution to be treated to such a level that the insoluble plutonium polymer is produced. It can be separated and recovered.

[0013]

Embodiments of the present invention will be described below with reference to the drawings. Note that the same parts are denoted by the same reference symbols throughout the drawings. Example 1 FIG. 1 shows a method for separating and recovering technetium according to this example, in which a radioactive waste liquid 1 containing technetium is subjected to a diffusion dialysis treatment with an anion exchange membrane 2 to obtain a radioactive waste liquid 1.
It comprises a diffusion dialysis step 4 for separating the technetium therein into a recovery solution 3. In this diffusion dialysis step 4, the radioactive waste liquid 1 containing radioactive elements such as fission products such as uranium, plutonium and technetium generated in the used reprocessing facility is contacted with the recovery liquid 3 through the anion exchange membrane 2. Then, technetium forming anions called pertechnetate ions is separated from the radioactive waste liquid 1 on the side of the recovery liquid 3. After the diffusion dialysis step 4, the radioactive liquid waste 1 is the treated liquid radioactive liquid 5 from which technetium has been removed,
The recovered liquid 3 is recovered as a technetium-containing recovered liquid 6, respectively.

FIG. 2 schematically shows the principle of separating technetium in the radioactive liquid waste 1 by the anion exchange membrane 2 and the construction of the diffusion dialysis apparatus 11 thereof. The diffusion dialysis apparatus 11 is installed vertically. The anion exchange membrane 2 and the two cells A and B separated by the anion exchange membrane 2. In addition, although each cell is divided into five in the figure, this is for explaining the movement of the ions in the solution in five stages for convenience.

When the radioactive waste liquid 1 containing technetium is introduced into the cell A of the diffusion dialyzer 11 from below and the recovery liquid 3 is introduced into the cell B from above, it is called pertechnetate ion in the radioactive waste liquid 1. The technetium forming anions is subjected to diffusion dialysis through the anion exchange membrane 2 from the radioactive waste liquid 1 having a high technetium concentration to the recovery liquid 3 side having a low technetium concentration to form cations in the radioactive waste liquid 1. Uranium, plutonium and fission products cannot pass through the anion exchange membrane 2 and the radioactive waste liquid 1
Remains inside. Therefore, the radioactive waste liquid 1 flowing in from the lower part of the cell of A lowers the technetium concentration as it comes into contact with the recovered liquid 3 in the cell of B, but the upper part of the cell still contains cations such as uranium, plutonium and fission products. Leading to. On the other hand, the recovery liquid 3 flowing in from the upper part of the cell B is gradually increased in technetium concentration as it comes into contact with the radioactive waste liquid 1 in the cell A, and is recovered as the technetium-containing recovery liquid 6.

Here, a simulated solution having the composition shown in Table 1 was prepared as a radioactive waste liquid, and 1.0 liter of this simulated solution was used as a recovery liquid to perform the above diffusion dialysis treatment using 1.0 liter of pure water. The simulated solution composition and the recovered solution composition after the treatment at this time are shown in Tables 2 and 3, respectively. In this experiment, ^99m Tc (a metastable nuclide of technetium produced from ⁹⁹ Mo, half-life 6.04h) was used as technetium, and the values of technetium in Tables 1 to 3 are values after attenuation correction. Further, in this simulated solution, cesium and strontium were added as typical fission products.

[0017]

[Table 1]

[0018]

[Table 2]

[0019]

[Table 3]

As is clear from these tables, the simulated solution that had passed through the A cell and had been subjected to the diffusion dialysis treatment had the same cesium and strontium concentrations as those before the treatment, but a nitric acid concentration of 1.0 mol / liter or more. The concentration decreased to 0.1 mol / liter, and the technetium concentration fell below the detection limit. The recovered solution after diffusion dialysis treatment that passed through the B cell had a nitric acid concentration of 0.9 mol / liter and a technetium radioactivity concentration of 10 microcurie /
There was a liter of solution.

As described above, according to the method for separating and recovering technetium of the present embodiment, technetium forming an anion called pertechnetate ion in the waste liquid is converted into cations other than uranium, plutonium and technetium. Can be selectively separated and recovered from fission products of.

Example 2 In the technetium separation and recovery method of this example, the free acid concentration in the radioactive waste liquid 1 in the diffusion dialysis step 4 as the recovery liquid 3 in the treatment step shown in FIG. It uses an acidic solution.

FIG. 3 schematically shows the behavior of cations, pertechnetate anions and free acids in the diffusion dialysis step 4 of this example. In FIG. 3, the diffusion dialysis device 11
In the same manner as described in Example 1, the radioactive waste liquid 1 containing technetium is introduced into the cell A from the bottom, and the acidic recovery liquid 3'is introduced into the cell B from the top. Diffusion dialyzer 1
In 1, the radioactive waste liquid 1 flowing in from the lower part of the A cell comes into contact with the acidic recovery liquid 3 ′ through the anion exchange membrane 2, so that the technetium forming anions in the radioactive waste liquid 1 has a lower technetium concentration. Uranium, plutonium and fission products which form cations in the radioactive waste liquid 1 by diffusion dialysis to the acidic recovery liquid 3'cannot permeate the anion exchange membrane 2 and remain in the radioactive waste liquid 1. In this way, the radioactive waste liquid 1 flows out from the upper part of the A cell as the treated radioactive waste liquid 5 from which technetium has been separated and removed. The acidic recovery liquid 3 ′ flowing from the upper part of the cell B is
The technetium concentration is gradually increased as it comes into contact with the radioactive waste liquid 1 in the cell, and the technetium-containing acidic recovery liquid 6'is recovered. When the concentration of free acid in the radioactive waste liquid 1 is higher than the concentration of free acid in the acidic recovery liquid 3 ', it migrates to the acidic recovery liquid 3'side, but falls below the free acid concentration of the acidic recovery liquid 3'. It never drops.

Here, a simulated solution having the composition shown in Table 4 was prepared as a radioactive liquid waste, and 1.0 liter of this simulated solution was used as a recovery solution and 1.0 liter of nitric acid solution having the same nitric acid concentration of 3.0 mol / liter as the simulated solution was used for diffusion dialysis. Processed.
Table 5 shows the composition of the simulated solution and the composition of the recovered solution after the treatment.
And shown in Table 6. In this experiment, ^99m Tc (a metastable nuclide of technetium produced from ⁹⁹ Mo, half-life 6.04h) was used as technetium, and the values of technetium in Tables 4 to 6 are values after attenuation correction. Further, in this simulated solution, cesium and strontium were added as typical fission products.

[0025]

[Table 4]

[0026]

[Table 5]

[0027]

[Table 6]

As is clear from these tables, the simulated solution that had passed through the A cell and had been subjected to the diffusion dialysis treatment had the same cesium and strontium concentrations as those before the treatment, but the technetium concentration was below the detection limit. It was The nitric acid concentration does not change because it is the same as the recovered liquid. Also,
The concentration of cesium and strontium in the recovered liquid after the diffusion dialysis treatment that passed through the B cell was 50 ppm or less, and the technetium radioactivity concentration was 10 microcurie / liter, which was the same as that of the simulated solution before treatment.

As described above, according to the method for separating and recovering technetium of the present embodiment, technetium forming an anion called pertechnetate ion is formed into a cation without decreasing the concentration of nitric acid in the waste liquid. Uranium,
It can be selectively separated and recovered from fission products other than plutonium and technetium. Therefore, it is possible to suppress the generation of an insoluble plutonium polymer due to a decrease in the concentration of free acid in the radioactive waste liquid.

Example 3 As shown in FIG. 4, the method of separating and recovering technetium of this example is such that the diffusion dialysis step 4 described in Example 1 and the technetium-containing recovery liquid 6 after this diffusion dialysis step 4 are electrolytically reduced. The electrolytic reduction step 21 for treating and the technetium oxide solid 22 produced by the electrolytic reduction treatment are filtered to obtain a filtrate 2
And a filtration step 24 for separating from 3.

In the above technetium separation and recovery method,
The radioactive waste liquid 1 containing radioactive elements such as fission products such as uranium, plutonium and technetium generated in the used reprocessing facility is contacted with the recovery liquid 3 through the anion exchange membrane 2 in the diffusion dialysis step 4. The technetium existing as pertechnetate ions (TcO ₄ ⁻ ) in the radioactive waste liquid 1 is separated into the recovery liquid 3 side. After the diffusion dialysis step 4, technetium (VII) separated from the radioactive waste liquid 1 and dissolved in the collected liquid 6 as pertechnetate ion is
In the next electrolytic reduction step 21, it is reduced to an IV value and becomes an oxide solid (TcO ₂ ) and precipitates. The recovered liquid 6 that has produced the oxide solid of technetium is separated into the oxide solid 22 and the filtrate 23 in the filtration step 24, and the technetium is recovered as the oxide solid 22.

Thus, the reduction step 2 according to this embodiment
In 1, the technetium was converted from VII value by electrolytic reduction method.
It is reduced to IV valence to generate technetium oxide (TcO ₂ ), which is separated in the filtration step 24.

By the way, as a method of recovering technetium, a reduction treatment with Fe (II) is carried out to obtain an oxide of technetium (T
A general method is to produce cO ₂ ) and further coprecipitate it with iron hydroxide. However, since iron hydroxide is also produced by such a method, a step of further separating technetium from iron hydroxide is required.

Diffusion dialysis treatment was performed using 1 liter of the simulated solution having the composition shown in Table 1 as the radioactive waste solution and 1 liter of water as the recovery solution, and 1 liter of the obtained technetium-containing recovery solution having the composition of Table 3 was electrolyzed. When reduction processing and Fe
Table 7 shows the amount of solids generated when the reduction treatment was carried out by the coprecipitation method. In the Fe coprecipitation method, Fe (II) 0.1 M was added to 1 liter of the technetium-containing recovery solution having the composition shown in Table 3.
Was added. Also, in this experiment, as technetium
^99m Tc 5 microcurie ⁹⁹ Tc as carrier
(Half-life 2.1 ⁴ × 10 ⁵ years) The one to which 5 microcurie was added was used.

[0035]

[Table 7]

As shown in Table 7, the solid amount obtained by the technetium separation and recovery method of this example is 3.89 × 10 ^-4.
g, which is almost equal to the amount of TcO ₂ converted from the technetium content in Table 3. That is, ⁹⁹ specific activity of specific activity ^{1.7 × 10 -2 Ci / g ...} 99m Tc of Tc is ^{5.3 × 10 6 Ci / g ...} 99 Tc amount of 5μCi respectively from ,, because it is the 2.94 × 10 ^{- The} amount of ^99m Tc of ⁴ g ... 5 μCi is 1.89 × 10 ⁻¹⁷ g, and the amount of oxide TcO _{2 of} these technetium is calculated as 3.89 × 10 ⁻⁴ g. Therefore, it can be said that the solid obtained by the electrolytic reduction treatment consists only of pure technetium oxide. On the other hand, the solid amount 8.98 g generated upon reduction treatment by the Fe coprecipitation method is 0.1 M Fe (O
H) ₂ amount, mostly due to the reducing agent.

As described above, according to the method for separating and recovering technetium of the present embodiment, radioactive effluent containing technetium is subjected to diffusion dialysis treatment to transfer technetium into the recovery liquid,
Furthermore, by electrolytically reducing the recovered liquid containing technetium, it is possible to suppress the generation of waste due to reducing agents,
Further, technetium can be selectively reduced even in the presence of other anions, and finally technetium can be separated and recovered from the radioactive waste liquid as pure technetium oxide.

Example 4 The technetium separation and recovery method of this example is the same as in Example 4 except that an acidic solution is used as the recovery liquid 3 to separate and recover technetium.

According to this example, by using an acidic solution as the recovery liquid, as described in Example 2, the generation of insoluble plutonium polymer due to the reduction of the free acid concentration in the radioactive waste liquid by the diffusion dialysis treatment was performed. In addition to being able to suppress it, by subjecting the recovered liquid 6 ′ after the diffusion dialysis treatment to electrolytic reduction treatment, it is possible to selectively reduce technetium and precipitate it as an oxide as described above. Finally, pure technetium oxide can be separated and recovered from the radioactive liquid waste without generating wastes such as polymers and reducing agents.

[0040]

As is apparent from the above description, according to the present invention, a solution containing technetium as pertechnetate ions is brought into contact with a recovery solution using an anion exchange membrane to perform a diffusion dialysis treatment. , Technetium can be separated from other radioactive elements existing as cations and transferred into a recovery liquid for separation and recovery. Further, in the case of this diffusion dialysis treatment, by using an acidic solution as a recovery liquid, even when the solution containing technetium is an acidic solution in which plutonium coexists, a decrease in free acid concentration such that an insoluble plutonium polymer is generated is reduced. Technetium can be separated and recovered in the recovery liquid without causing it. Furthermore, by subjecting the recovered liquid after the diffusion dialysis treatment to a reduction treatment, it is possible to separate and recover the technetium oxide solid from the recovered liquid.

[Brief description of drawings]

FIG. 1 is a flow chart showing a method for separating and recovering technetium of Example 1 of the present invention.

FIG. 2 is a diagram schematically showing an apparatus structure in a diffusion dialysis step of Example 1 and movement of ions in a solution.

FIG. 3 is a diagram schematically showing an apparatus structure in a diffusion dialysis step of Example 2 and movement of ions in a solution.

FIG. 4 is a flow chart showing a method for separating and recovering technetium of Example 3 of the present invention.

[Explanation of symbols]

 1 ………… Radioactive waste liquid 2 ………… Anion exchange membrane 3 ………… Recovered liquid 3 ′ …… Acid recovery liquid 5 ………… Radioactive waste liquid after treatment 6 ………… Technetium-containing recovery liquid 6 ′ …… Technetium-containing acidic liquid Recovery liquid 11 ………… Diffusion dialyzer 11 22 ………… Technetium oxide solid 23 ………… filtrate

Claims (3)

[Claims] 1. A technetium separation characterized in that a solution containing technetium is brought into contact with a recovery solution through an anion exchange membrane, and technetium existing in the form of an anion is separated and recovered in the recovery solution by diffusion dialysis. Recovery method. 2. A solution containing technetium is brought into contact with a recovery solution through an anion exchange membrane, and technetium existing in the form of an anion is separated and recovered in the recovery solution by diffusion dialysis, and then the recovery solution containing technetium. A method for separating and recovering technetium, which comprises converting technetium into an oxide by reduction treatment and separating and recovering from the recovery liquid. 3. The method for separating and recovering technetium according to claim 1, wherein the recovery liquid is an acidic solution.