JPH0986936A - Recovery of technetium - Google Patents

Abstract

PROBLEM TO BE SOLVED: To separate technetium in a simple operation from a nitric acid solution containing uranium generated from spent nuclear fuel reprocessing plants. SOLUTION: Either ammonia gas, ammonia water or sodium hydroxide is added to a uranium-contg. nitric acid solution to form ammonium biuranate or sodium biuranate precipitate which is then subjected to solid-liquid separation. In this case, as uranium and technetium transfer to the solid phase and liquid phase, respectively, the liquid is electrolytically reduced to effect electrodeposition of the technetium which is then separated.

Description

Detailed Description of the Invention

[0001]

TECHNICAL FIELD The present invention relates to a nitric acid solution containing uranium generated from a spent nuclear fuel reprocessing facility, or a uranium enrichment facility, a conversion processing facility, and a spent nuclear fuel reprocessing facility in a nuclear fuel cycle. And a method for recovering technetium from technetium-containing waste liquid generated from facilities or equipment such as facilities using radioisotopes.

[0002]

2. Description of the Related Art In the process of reprocessing spent nuclear fuel by the Purex method, uranium and plutonium are recovered by dissolving the spent nuclear fuel in nitric acid and by a solvent extraction method using a TBP / n dodecane solvent. . There are many kinds of fission products in the spent fuel, and among them, technetium has complicated behavior and is said to be a nuclide that is difficult to separate and purify from uranium and plutonium in the solvent extraction process. . The reason is that technetium forms compounds such as UO ₂ (NO ₃ ) (TcO ₄ ) in a solution of spent nuclear fuel dissolved in nitric acid, and is extracted on the solvent side (TBP / n dodecane) together with uranium. This is because it becomes easier. The solvent from which uranium was extracted was washed with nitric acid and then back-extracted from the solvent with dilute nitric acid.
By repeating such extraction / washing / back-extraction operations, fission products such as technetium and uranium are separated / purified.

In the nuclear fuel cycle, the uranium thus recovered is reused as nuclear fuel after being processed in a conversion facility, an enrichment facility, and a reconversion processing facility. In this case, if the separation of uranium and technetium is insufficient in the above-mentioned reprocessing step, technetium will remain in each step of the nuclear fuel cycle, and technetium will be contained in the waste liquid generated from the above facility. It becomes necessary to remove technetium from these waste liquors. As a method for removing technetium from the waste liquid, there are a precipitation method and an adsorption method.

[0004]

As described above, since the cycle of extraction, washing and back-extraction is repeated by the solvent extraction method in the reprocessing step, the process becomes complicated,
The cost of separation and purification must be high.

When removing technetium from the waste liquid, a ferrite precipitation method is generally used as a precipitation method.
In a waste liquid containing no reducing agent or the like, technetium is usually present in the form of TcO ₄ ⁻ , and when a ferrite precipitation method is applied to this waste liquid, TcO ₄ ⁻ is produced due to the reducing power when ferrite is formed. There is reduced to TcO _2, so that the TcO ₂ is taken in the ferrite. However, when producing ferrite,
There is a condition that the concentration of Fe ²⁺ needs to be a certain value or more, and the pH at the time of ferrite formation is also limited to the range of 6 to 8. Further, in the ferrite precipitation method, a large amount of used ferrite is discharged as radioactive waste.

[0006] Although adsorption method is performed by utilizing an ion-exchange or activated carbon adsorption, the principle is, TcO in the effluent ₄ ^-
It is intended to adsorb ions by anion exchange reaction. However, when anions such as silver nitrate coexist in the waste liquid, the competitive exchange reaction with these anions significantly reduces the adsorption performance of technetium. Also in the adsorption method, used resin and activated carbon are discharged as secondary waste.

The present invention has been made in view of the problems of the conventional technetium recovery method as described above, and it is possible to obtain uranium from a uranium nitrate solution containing technetium generated in a solvent extraction step of a reprocessing facility. An object of the present invention is to provide a method for separating technetium from technetium by a simple operation and preventing technetium from remaining in uranium collected in a reprocessing facility. Another object of the present invention is to provide a technetium recovery method for efficiently removing technetium that migrates to the waste liquid side in each facility of the nuclear fuel cycle and for reducing the amount of secondary waste generated as much as possible.

[0008]

Means for Solving the Problems A method for recovering technetium according to the present invention is a method in which technetium and uranium are extracted into a solvent in a solvent extraction step of a reprocessing facility, and technetium is back-extracted from the solvent together with uranium by dilute nitric acid. A process of adding ammonia gas, aqueous ammonia or sodium hydroxide to a nitric acid solution containing uranium to form a precipitate of ammonium diuranate or sodium diuranate, and solid-liquid separation of the precipitate to obtain a solid. The process of transferring uranium to the liquid side and technetium to the liquid side.

Since technetium does not migrate to ammonium biuranate or sodium diuranate precipitates,
By performing solid-liquid separation, uranium and technetium are separated.

As a solid-liquid separation method, it is preferable to carry out electrolytic reduction using a filtrate containing technetium as an electrolytic solution. By this electrolytic reduction, technetium can be reduced and deposited on the cathode and recovered. The filtrate is released to the environment as a technetium-free waste solution. It is preferable that the temperature of the electrolytic solution in the electrolytic reduction is set from room temperature to the boiling point or less, and the PH value of the electrolytic solution is set between 0 and 14. A particularly preferable temperature of the electrolytic solution is in the range of 50 to 60 degrees Celsius.

In the technetium recovery method according to the present invention, the waste liquid to be subjected to electrolytic reduction is not limited to the above-mentioned filtrate, but a conversion facility, a concentration facility, a reconversion processing facility, a facility using radioactive isotopes in a nuclear fuel cycle. Also, waste liquid generated from medical facilities, etc. is also applicable.

[0012]

BEST MODE FOR CARRYING OUT THE INVENTION In the separation / purification process of a spent nuclear fuel reprocessing facility, technetium in a nitric acid solution of uranium exists as a 7-valent pertechnetate ion (TcO ₄ ⁻ ). When ammonia gas, aqueous ammonia or sodium hydroxide solution is added to the solution for neutralization, a precipitate of ammonium diuranate or sodium diuranate is produced according to the following reaction. 2UO ₂ (NO ₃ ) ₂ + 6NH ₃ + 3H ₂ O → (NH ₄ ) ₂ · U ₂ O ₇ ↓ + 4NH ₄ NO ₃ (1) 2UO ₂ (NO ₃ ) ₂ + 6NaOH → Na ₂ · U ₂ O ₇ ↓ + 4NaNO ₃ + 3H ₂ O (2)

In these reactions, technetium is the T
Since cO ₄ ⁻ remains in the solution as it is, technetium can be separated from uranium by solid-liquid separation of a precipitate of ammonium diuranate or sodium diuranate. On the other hand, since technetium is a relatively noble element electrochemically, when electrolysis is performed using the above-mentioned filtrate as the electrolytic solution, TcO ₄ ⁻ is reduced at the cathode and deposited on the electrode. In this way, technetium can be removed from the filtrate.

In this case, the pH value of the electrolytic solution may be in the range of 0 to 14, and the process control is easy. However, when increasing the technetium reduction rate, the pH value is preferably in the range of about 1 to 13. The temperature of the electrolytic solution can be set in the range from room temperature to the boiling point, but in the case of increasing the technetium reduction rate, it is preferably set in the range of 50 to 60 degrees Celsius.

The technetium reduced and deposited on the cathode is converted into a cathode by changing the polarity of the cathode, and is again converted into Tc.
The electrode can be reused because it can be removed from the electrode after being oxidized to O ₄ ⁻ . Therefore, the used electrode is not discharged as waste, and the amount of secondary waste generated can be reduced.

[0016]

EXAMPLE In accordance with the present invention, an experiment was conducted to recover technetium from a uranyl nitrate solution containing technetium. The results will be described below. [Example 1] Ammonia was added to a uranyl nitrate solution containing technetium to adjust the pH to 7, thereby forming a precipitate of ammonium diuranate. After solid-liquid separation was performed, the ammonium biuranate precipitate was washed with dilute aqueous ammonia, and this washing liquid was added to the filtrate. The results shown in Table 1 were obtained when the precipitation of the ammonium diuranate and the radioactivity measurement of technetium in the filtrate were carried out.

(Table 1) Technetium radioactivity during ammonium diuranate precipitation and in filtrate Ammonium diuranate precipitation (Bq) 150 In filtrate (Bq) 19,800

As described above, 1% or less ([150 ÷ (150 + 19800)] <0.01) of technetium migrated into the ammonium diuranate precipitation, 99% or more migrated to the filtrate side, and further uranium. Since it was not detected in the filtrate, it was confirmed that technetium was efficiently separated.

Example 2 The pH value of the technetium-containing filtrate (0.2 M, NH ₄ NO ₃ property) of Example 1 was adjusted to 0.
To Pt mesh (surface area = 135 cm ² ) as the cathode and Pt plate (surface area =
51 cm ² ) as the anode, and cathode current density = 30 m
After conducting constant current electrolysis at A / cm ² for 1 hour, the concentration of technetium remaining in the electrolytic solution was measured. The temperature of the electrolytic solution was kept at 60 degrees Celsius. The result is as shown in FIG.

As described above, it was confirmed that technetium was reduced and deposited on the cathode to remove technetium from the electrolytic solution. The pH value of the electrolytic solution can be set to any pH value within the range of 0 to 14. As is clear from FIG. 1, the pH value of the electrolytic solution is preferably in the range of 1 to 13 in order to deposit technetium in a shorter time.

[Example 3] A pH value of a technetium-containing filtrate similar to that in Example 2 was adjusted to 7 to prepare an electrolytic solution.
Under the same electrolysis conditions as in Example 2, constant current electrolysis was performed for 1 hour with the temperature of the electrolytic solution being 25 to 60 degrees Celsius. The results of measuring the concentration of technetium remaining in the electrolytic solution after electrolysis are shown in FIG.

As described above, even when the temperature of the electrolytic solution is 25 ° C., the removal rate of 95% or more can be obtained by setting the electrolysis time to 1 hour. The technetium removal rate increases as the temperature of the electrolyte increases, but no significant difference is observed in the technetium removal rate when the temperature of the electrolyte is 50 degrees Celsius and 60 degrees Celsius. Therefore, the temperature of the electrolyte is 5 degrees Celsius.
It can be seen that the technetium removal rate stops increasing at 0 degrees.

Example 4 For a uranyl nitrate solution containing technetium, a NaOH solution was added to adjust the pH to 7
Was adjusted to produce a precipitate of ammonium diuranate. Filtrate after solid-liquid separation (0.2 M, NaNO ₃
Property) is adjusted to a range of 0 to 14 and subjected to constant current electrolysis for 1 hour under the same electrolysis conditions as in Example 2,
The concentration of technetium remaining in the electrolytic solution was measured. The result is shown in FIG. The technetium removal rate shows almost the same tendency as in Example 1. Therefore, it was confirmed that there is no difference in the effect of the present invention regardless of whether the electrolyte is NaNO ₃ or NH ₄ NO ₃ .

[Example 5] In Examples 2 to 4, technetium was electrodeposited using the filtrate of ammonium heavy uranate or sodium diuranate as an electrolytic solution. Waste liquid of concentration is generated.
Therefore, the electrolytic solution was NaNO ₃ salt, the salt concentration and pH were changed, and technetium was electrodeposited under the same electrolysis conditions as in Example 2. The results are shown in Table 2.

(Table 2) NaNO ₃ salt concentration, pH and technetium removal rate NaNO ₃ concentration (M) pH technetium removal rate (%) 0.2 0 70 0.2 1 95 95 0.2 7 99 0.2 10 95 0.2 13 87 0.2 0.2 14 77 1.25 0 70 1.25 14 70 6.0 0 18 6.0 6.0 1 54 6.0 6.0 7 49 6.0 14 34

As described above, although the removal rate of technetium decreases as the salt concentration increases, the technetium removal rate similar to that when the salt concentration is 0.2 M can be obtained by further lengthening the electrolysis time.

[0027]

As described above, according to the technetium recovery method of the present invention, technetium can be efficiently removed from the uranium-containing nitric acid solution generated from the spent nuclear fuel reprocessing facility. In addition, technetium can be removed from technetium-containing waste liquid generated from facilities and equipment such as spent nuclear fuel reprocessing facilities, nuclear fuel conversion processing facilities, and radioactive isotope use facilities. It can be significantly reduced.

[Brief description of drawings]

FIG. 1 is a graph showing the relationship between the technetium removal rate and the pH of an electrolytic solution.

FIG. 2 is a graph showing the relationship between the technetium removal rate and the temperature of the electrolytic solution.

FIG. 3 is a graph showing the relationship between the technetium removal rate and the pH of the electrolytic solution.

Front page continued (72) Inventor Koji Kuno 2600 Ishigami Sojuku, Tokai-mura, Naka-gun, Ibaraki Prefecture Sumitomo Kinzoku Mining Co., Ltd.

Claims (4)

[Claims] 1. A process of adding ammonium gas, aqueous ammonia, or sodium hydroxide to a nitric acid solution containing uranium to form a precipitate of ammonium diuranate or sodium diuranate, and the solidification of the precipitate. A method for recovering technetium, comprising the steps of liquid separation and transferring uranium to the solid side and technetium to the liquid side. 2. The method for recovering technetium according to claim 1, wherein the liquid containing technetium is electrolytically reduced, and technetium is electrodeposited on the cathode to separate technetium from the liquid. 3. The temperature of the electrolytic solution in the electrolytic reduction is set from room temperature to the boiling point or less, and the PH value of the electrolytic solution is set to be 0 to 14. The method for recovering technetium according to 2. 4. The method for recovering technetium according to claim 3, wherein the temperature of the electrolytic solution is set in a range of 50 to 60 degrees Celsius.