JP2971638B2 - Technetium separation and recovery method - Google Patents

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 a waste liquid generated in a facility for reprocessing spent nuclear fuel.

[0002]

2. Description of the Related Art Spent nuclear fuel reprocessing facilities use a nitric acid aqueous solution and an organic solvent to separate fission products and transuranium elements in spent nuclear fuel into high-level waste liquid.
Uranium and plutonium are recovered as products. In such a spent nuclear fuel reprocessing facility, a uranyl nitrate solution, a plutonium nitrate solution, a low-level waste solution, and various other solutions are handled in addition to the high-level waste solution. Technetium, one of the fission products, is present in spent nuclear fuel, but because of its complex chemical behavior, technetium is handled in spent nuclear fuel reprocessing facilities as described above and more or less in the solution. Contains technetium.

[0003] This technetium has an extremely long half-life, which may affect the environment when high-level waste liquid is disposed of as a vitrified product. There is a need to separate and collect technetium in these solutions because of the possibility of being transferred into the gas phase in various processes and the fact that it can be effectively used as a valuable element that does not exist in nature.

[0004]

This technetium is:
Extremely long half-life has the potential to affect the environment when disposing of high-level waste liquid as vitrified matter. It is desired to separate and collect technetium in these solutions because of the possibility of migration, and the fact that it can be effectively used as a valuable element that does not exist in nature.

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

In the case of plutonium nitrate solution,
It is known that an insoluble plutonium polymer is formed when the concentration of free acid in a solution is low (for example, Chemical Engineering for Fuel Reprocessing and Radioactive Waste Management in Nuclear Chemical Engineering Vol. (See page 150, published by Nikkan Kogyo Shimbun on December 29, 1958.) When such a polymer is formed, not only may the process piping be clogged, but also there is a risk of critical safety due to the unexpected accumulation of plutonium.Therefore, when technetium in a plutonium nitrate solution is separated and recovered by a membrane, etc. It is necessary to keep the free acid concentration in the solution below the concentration at which the plutonium polymer is formed.

In order to cope with such a problem, 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 solution generated in a spent nuclear fuel reprocessing facility or the like. It is an object of the present invention to provide a method for separating and recovering technetium without using technetium.

[0008]

That is, in the method of separating and recovering technetium according to the present invention, a solution containing technetium is brought into contact with a recovery solution via an anion exchange membrane, and technetium existing in the form of anions is subjected to diffusion dialysis. And is separated and recovered in a recovery liquid.

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 via an anion exchange membrane, and technetium existing in the form of anions is separated into the recovery solution by diffusion dialysis. After the recovery, the technetium-containing recovery solution is subjected to a reduction treatment to convert technetium to an oxide and separated and recovered from the recovery solution.

Further, in the method for separating and recovering technetium according to the present invention, a solution containing technetium is brought into contact with a recovery solution via an anion exchange membrane, and technetium existing in an anion form is diffused into the recovery solution by diffusion dialysis. When separating and recovering, an acidic solution is used as the recovering liquid.

[0011]

In the method for separating and recovering technetium of the present invention, a solution containing technetium, such as a radioactive waste liquid generated in a used reprocessing facility, is brought into contact with the recovered solution via an anion exchange membrane to perform diffusion dialysis. Technetium, which forms an anion called pertechnetate ion (TcO ₄ ⁻ ), can be selectively separated and recovered from fission products other than uranium, plutonium, and technetium that form cations. The technetium thus separated in the recovery liquid is converted from pertechnetate ions to technetium oxide (TcO ₂ ) by performing a reduction treatment such as an electrolytic reduction treatment, and precipitated or filtered as an oxide solid. For example, it can be separated and collected from the collected liquid.

Further, 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 can be recovered without reducing the free acid concentration in the solution to be treated to a free acid concentration such that an insoluble plutonium polymer is formed by making the recovery solution in contact with the anion exchange membrane an acidic solution. It can be separated and collected inside.

[0013]

Embodiments of the present invention will be described below with reference to the drawings. The same parts are denoted by the same reference numerals throughout the drawings. Example 1 FIG. 1 shows a method for separating and recovering technetium according to the present embodiment. The radioactive waste liquid 1 containing technetium is subjected to diffusion dialysis treatment with an anion exchange membrane 2 to obtain a radioactive waste liquid 1.
The method comprises a diffusion dialysis step 4 for separating the technetium in the recovery liquid 3. In the 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 comes into contact with the recovery liquid 3 via the anion exchange membrane 2. Then, technetium forming an anion called pertechnetate ion is separated from the radioactive waste liquid 1 to the recovery liquid 3 side. After the diffusion dialysis step 4, the radioactive waste liquid 1 is used as the post-treatment radioactive waste liquid 5 from which technetium has been removed.
The recovery liquid 3 is recovered as a technetium-containing recovery liquid 6, respectively.

FIG. 2 schematically shows the principle of separating technetium in the radioactive waste liquid 1 by the anion exchange membrane 2 and the structure of the diffusion dialysis apparatus 11, wherein the diffusion dialysis apparatus 11 is installed vertically. An anion exchange membrane 2 and two cells A and B separated by the anion exchange membrane 2. Although each cell is divided into five in the figure, this is for the purpose of explaining the movement of 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 dialysis apparatus 11 from below, and the recovery liquid 3 is introduced into the cell B from above, the radioactive waste liquid 1 is referred to as pertechnetate ion. Technetium forming anions is subjected to diffusion dialysis from the radioactive waste liquid 1 having a high technetium concentration to the recovery liquid 3 having a low technetium concentration via the anion exchange membrane 2 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 1
Remains inside. Therefore, the radioactive waste liquid 1 flowing from the lower part of the cell A decreases the technetium concentration as it comes into contact with the recovery liquid 3 in the cell B, but the effluent 1 still contains cations such as uranium, plutonium and fission products, Leads to. On the other hand, the recovered liquid 3 flowing 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 recovered liquid 6.

Here, a simulated solution having the composition shown in Table 1 was prepared as a radioactive waste liquid, and the above-described diffusion dialysis treatment was performed using 1.0 liter of the simulated solution as a recovery liquid and 1.0 liter of pure water. Tables 2 and 3 show the composition of the simulated solution and the composition of the recovered solution after the treatment. Incidentally, (metastable nuclide technetium generating from ⁹⁹ Mo, half-life 6.04H) In this experiment ^99m Tc as Technetium is used, the value of technetium from Table 1 to Table 3 is a value obtained by attenuation correction. Further, in this simulated solution, cesium and strontium were added as representatives of fission products.

[0017]

[Table 1]

[0018]

[Table 2]

[0019]

[Table 3]

As is clear from these tables, the simulated solution that has passed through the A-cell after the diffusion dialysis treatment has the same cesium and strontium concentration as before, but has a nitric acid concentration of 1.0 mol / liter. The concentration decreased to 0.1 mol / liter, and the technetium concentration fell below the detection limit. The recovered solution after diffusion dialysis that passed through the B cell had a nitric acid concentration of 0.9 mol / liter and a technetium radioactivity concentration of 10 microCuries / liter, which was the same as the simulated solution before the treatment.
One liter of solution was obtained.

As described above, according to the technetium separation and recovery method of this embodiment, technetium which forms an anion called pertechnetate ion in the waste liquid is replaced by uranium, plutonium and technetium other than uranium and plutonium which form cations. Can be selectively separated and recovered from fission products of.

Example 2 The method for separating and recovering technetium according to the present example does not significantly reduce the free acid concentration in the radioactive waste liquid 1 during the diffusion dialysis step 4 as the recovery liquid 3 in the processing step shown in FIG. An acidic solution was used.

FIG. 3 schematically shows the behavior of the cation, the pertechnetate anion and the free acid in the diffusion dialysis step 4 of the present embodiment. In FIG. 3, the diffusion dialysis device 11
As described in Example 1, the radioactive waste liquid 1 containing technetium is introduced into the A cell from below, and the acidic recovery liquid 3 'is introduced into the B cell from above. Diffusion dialysis device 1
In 1, the radioactive waste liquid 1 flowing from below the A cell comes into contact with the acidic recovery liquid 3 ′ through the anion exchange membrane 2, so that technetium forming anions in the radioactive waste liquid 1 has a lower technetium concentration. The uranium, plutonium and fission products forming cations in the radioactive waste liquid 1 are diffused and dialyzed against the acidic recovery liquid 3 ′ and cannot remain in the anion exchange membrane 2 and remain in the radioactive waste liquid 1. Thus, the radioactive waste liquid 1 flows out of the upper part of the A cell as the post-treatment 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
The technetium concentration is gradually increased as it comes into contact with the radioactive waste liquid 1 in the cell, and is recovered as a technetium-containing acidic recovery liquid 6 '. When the concentration of the free acid in the radioactive waste liquid 1 is higher than the concentration of the free acid in the acidic recovery liquid 3 ′, the free acid moves to the acidic recovery liquid 3 ′ side. It does not decline.

Here, a simulated solution having the composition shown in Table 4 was prepared as a radioactive waste liquid, and diffusion dialysis was performed using 1.0 liter of the simulated solution and 1.0 liter of a nitric acid solution having the same nitric acid concentration of 3.0 mol / l as the simulated solution as a recovery solution. Processing was performed.
Table 5 shows the simulated solution composition and the recovered solution composition after the treatment.
And Table 6. Incidentally, (metastable nuclide technetium generating from ⁹⁹ Mo, half-life 6.04H) In this experiment ^99m Tc as Technetium is used, the value of technetium from Table 4 to Table 6 is a value obtained by attenuation correction. Further, in this simulated solution, cesium and strontium were added as representatives of fission products.

[0025]

[Table 4]

[0026]

[Table 5]

[0027]

[Table 6]

As is clear from these tables, the simulated solution that has passed through the inside of the A cell after the diffusion dialysis treatment has the same cesium and strontium concentration as before, but the technetium concentration is below the detection limit. Was. The nitric acid concentration is unchanged since it is the same concentration as the recovered solution. Also,
In the recovered solution after the diffusion dialysis treatment that passed through the B cell, the cesium and strontium concentrations were 50 ppm or less, and the technetium radioactivity concentration was 10 microcuries / liter, which was the same as the simulated solution before the treatment.

As described above, according to the technetium separation / recovery method of this embodiment, technetium which forms an anion called pertechnetate ion is converted into a cation without lowering 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. For this reason, generation | occurrence | production of the insoluble plutonium polymer by the free acid concentration fall in radioactive waste liquid can be suppressed.

Example 3 As shown in FIG. 4, the method for separating and recovering technetium in this example is the electrolytic dialysis step 4 described in Example 1 and the technetium-containing recovery solution 6 after the diffusion dialysis step 4 is electrolytically reduced. Electrolytic reduction step 21 to be treated, and technetium oxide solid 22 generated by the electrolytic reduction
And a filtration step 24 for separation from the filter step 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 comes into contact with the recovery liquid 3 via the anion exchange membrane 2 in the diffusion dialysis step 4. Then, technetium present as pertechnetate ion (TcO ₄ ⁻ ) from the radioactive waste liquid 1 is separated to the recovery liquid 3 side. After the diffusion dialysis step 4, technetium (VII) which is separated from the radioactive waste liquid 1 and dissolved in the recovery liquid 6 as pertechnetate ion,
In the next electrolytic reduction step 21, it is reduced to IV valence and precipitates as an oxide solid (TcO ₂ ). The recovery liquid 6 that has produced the technetium oxide solid is separated into an oxide solid 22 and a filtrate 23 in a filtration step 24, and technetium is recovered as the oxide solid 22.

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

By the way, as a method of recovering technetium, an oxide of technetium (T
A common method is to produce cO ₂ ) and coprecipitate it with iron hydroxide. However, since iron hydroxide is also generated in such a method, a step of further separating technetium from iron hydroxide is required.

Here, 1 liter of the simulated solution having the composition shown in Table 1 was used as the radioactive waste liquid, and 1 liter of water was used as the recovery liquid for diffusion dialysis treatment, and 1 liter of the technetium-containing recovery liquid having the composition shown in Table 3 was electrolyzed. When the reduction process is performed and Fe
Table 7 shows the amount of each solid generated when the reduction treatment was performed by the coprecipitation method. In the Fe coprecipitation method, 1 liter of the technetium-containing recovery solution having the composition shown in Table 3 was added to 0.1 M of Fe (II).
Was added. In this experiment, technetium
^99m Tc5 ⁹⁹ Tc to microcuries as carrier
(Half-life 2.1 ⁴ × 10 ⁵ years) A solution to which 5 microcuries were added was used.

[0035]

[Table 7]

As shown in Table 7, the amount of solids obtained by the method of separating and recovering technetium of this example was 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, the specific activity of ⁹⁹ Tc is 1.7 × 10 ⁻² Ci / g... The specific activity of ^99m Tc is 5.3 × 10 ⁶ Ci / g. Therefore, the ⁹⁹ Tc amount of 5 μCi is 2.94 × 10 ^{− The} amount of ^99m Tc of ⁴ g... 5 μCi is 1.89 × 10 ⁻¹⁷ g..., And the amount of technetium oxide TcO ₂ 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, 8.98 g of the solid generated during the reduction treatment by the Fe coprecipitation method was 0.1 M Fe (O
H) ₂ equivalents, mostly due to the reducing agent.

As described above, according to the technetium separation / recovery method of the present embodiment, the radioactive waste liquid containing technetium is subjected to diffusion dialysis to transfer technetium into the recovered liquid.
In addition, the effluent containing technetium is electrolytically reduced to reduce the generation of waste due to reducing agents, etc.
In addition, technetium can be selectively reduced even in the presence of other anions, and technetium can be finally separated and recovered from the radioactive waste liquid as a pure technetium oxide.

Embodiment 4 The technetium separation / recovery method of this embodiment is the same as that of Embodiment 4, except that an acidic solution is used as the recovery liquid 3, to separate and recover technetium.

According to this embodiment, by using an acidic solution as the recovery solution, as described in the second embodiment, the generation of the insoluble plutonium polymer due to the decrease in the concentration of the free acid in the radioactive waste liquid by the diffusion dialysis treatment is prevented. As described above, technetium can be selectively reduced and precipitated as an oxide by performing an electrolytic reduction treatment on the recovered solution 6 'after the diffusion dialysis treatment, and the insoluble plutonium can be suppressed during the treatment process. Finally, pure technetium oxide can be separated and recovered from the radioactive waste liquid without generating wastes such as polymers and reducing agents.

[0040]

As is apparent from the above description, according to the present invention, a diffusion dialysis treatment in which a solution containing technetium as pertechnetate ion is brought into contact with a recovery solution using an anion exchange membrane is performed. In addition, technetium can be separated from other radioactive elements existing as cations, transferred to a recovery liquid, and separated and recovered. In addition, in the case of the diffusion dialysis treatment, by using an acidic solution as the recovery liquid, even when the solution containing technetium is an acidic solution in which plutonium coexists, a decrease in the free acid concentration that generates an insoluble plutonium polymer is prevented. Technetium can be separated and recovered in the recovery liquid without causing the recovery. Furthermore, by subjecting the collected solution after the diffusion dialysis treatment to a reduction treatment, it can be separated and collected from the collected solution as a technetium oxide solid.

[Brief description of the drawings]

FIG. 1 is a flowchart showing a method for separating and recovering technetium according to 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 illustrating an apparatus structure in a diffusion dialysis step of Example 2 and movement of ions in a solution.

FIG. 4 is a flowchart showing a method for separating and recovering technetium according to a third embodiment 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 acid Recovered liquid 11: Diffusion dialysis device 11 22: Technetium oxide solid 23: Filtrate

──────────────────────────────────────────────────続 き Continued on the front page (58) Field surveyed (Int.Cl. ⁶ , DB name) G21C 19/46 G21F 9/06 G21F 9/12

Claims (3)

(57) [Claims] 1. Separation of technetium, wherein a solution containing technetium is brought into contact with a recovery solution via an anion exchange membrane, and technetium existing in the form of anions is separated and recovered in the recovery solution by diffusion dialysis. Collection method. 2. A technetium-containing solution is brought into contact with a recovery solution via an anion exchange membrane, and technetium existing in the form of anions is separated and recovered in the recovery solution by diffusion dialysis. Technetium is converted to an oxide by subjecting the technetium to an oxide, and the technetium is separated and recovered from a recovery solution. 3. The method for separating and recovering technetium according to claim 1, wherein the recovery liquid is an acidic solution.