JP4901359B2 - Radioactive waste disposal method - Google Patents

Description

  The present invention relates to a radioactive waste processing method and processing apparatus, and more particularly to a radioactive waste processing method and processing apparatus containing uranium, a transuranium element, a radionuclide, or a compound thereof.

As a method of separating and recovering uranium and other radionuclides from uranium and uranium compounds containing uranium and uranium compounds generated from facilities handling uranium, etc., uranium is dissolved with organic acid as a method for separating and recovering uranium and other radionuclides. And a processing method is performed (see, for example, Patent Documents 1 and 2). In this method, since a decomposable organic acid is used as a medium for dissolving uranium from a fluoride-based uranium waste, the amount of neutralized salt generated when hydrochloric acid or the like is used can be reduced. In this method, after dissolving uranium using an organic acid, uranium hydroxide (UO ₂ (OH) ₂ ) or uranium peroxide (UO ₄ ) is used to efficiently precipitate and separate uranium. A method of precipitation and separation / recovery is used.
JP 2004-20251 A (2nd page, lines 2-7) JP 2005-127926 A

However, in the above method, the organic acid itself is in a reduced state as compared with normal water, and the recovery rate is reduced during separation due to complex formation between the organic acid and uranyl ions (UO ²⁺ ). There is a fear. In addition, the remaining organic acid finally becomes a secondary waste source. Furthermore, when uranium eluted with an organic acid is precipitated as uranyl hydroxide (UO ₂ (OH) ₂ ) or uranium peroxide (UO ₄ ), depending on the addition conditions of the organic acid or oxidizing agent, Due to segregation of formation, a desired amount of precipitate may not be obtained.

  The present invention has been made in consideration of the above-described circumstances, and effectively separates uranium, transuranium elements or other radionuclides from uranium, transuranium elements, other radionuclides or radioactive waste containing the compounds thereof. An object of the present invention is to provide a processing method and a processing apparatus for recovered radioactive waste.

In order to achieve the above object, a method for treating a radioactive waste according to one embodiment of the present invention includes a method for treating uranium, a transuranium element, a uranium element, a transuranium element, a radionuclide, or a radionuclide or a compound thereof. While detecting the oxidation-reduction potential in the oxidation-reduction reaction in which the uranium, transuranium element or other radionuclide obtained in the dissolution-separation step is reacted with an oxidizing agent, A precipitate generating step for generating a precipitate by increasing the addition amount of the oxidizing agent and / or increasing the addition rate when the detected potential is within a predetermined range in which segregation of precipitate formation does not occur; and the precipitate generating step And a separation step of separating the precipitate obtained.

  According to the present invention, a radioactive waste treatment method and treatment for effectively separating and recovering uranium, transuranium element or other radionuclide from uranium, transuranium element, other radionuclide or radioactive waste containing the compound thereof. Equipment can be provided.

  EMBODIMENT OF THE INVENTION Below, the form for implementing this invention is demonstrated based on drawing. The present invention is not limited to these embodiments.

(First embodiment)
FIG. 1 is a flowchart showing the procedure of the radioactive waste processing method according to the first embodiment of the present invention. FIG. 2 is a diagram schematically illustrating a main configuration of an example of a radioactive waste treatment apparatus used in this embodiment. As shown in FIG. 2, the radioactive waste treatment apparatus includes a precipitation reaction tank 1, an oxidation-reduction potential measuring device 2, a stirring blade 3, and an oxidant addition line 4 for supplying an oxidant from an oxidant storage tank 10. Yes. In FIG. 2, the waste liquid treated in the precipitation reaction tank 1 is transferred from the waste liquid extraction line 5 to the solid-liquid separator 7 by the transfer pump 6, separated into solid and liquid, and processed in the post-treatment tank 8. In addition, the waste liquid after being separated into a solid and a liquid can be returned to the precipitation reaction tank 1, and treatment such as decomposition of the organic acid in the waste liquid can be performed in the precipitation reaction layer.

  Next, the radioactive waste processing method according to this embodiment will be described with reference to FIGS. In the method for treating radioactive waste according to this embodiment, for example, uranium is dissolved from a radioactive waste 11 containing uranium and / or a uranium compound (hereinafter referred to as “uranium”) with an organic acid solution (S1), and a solid ( Separation into solid waste 12) and liquid (uranium-containing organic acid solution 13) (S2), a redox reaction in a redox reaction in which the uranium obtained by the dissolution / separation step and the uranium obtained by this dissolution / separation step are reacted with each other. A precipitate generating step (S3) for generating (forming) a precipitate by adjusting the addition amount of the oxidant while detecting the potential, and the precipitate obtained by the precipitate generating step is a solid (solid (uranium)). A separation step (S4) for separating the waste 14) into a liquid (organic acid solution 15). In FIG. 1, an alkaline earth metal hydroxide is added to the carbonate ions generated in the organic substance decomposition step (S5) and the organic substance decomposition step contained in the solution (for example, formic acid solution) 15 remaining after the separation step. And adding an alkaline earth metal carbonate production step (S6) for producing a hardly soluble carbonate, and a cement solidified body producing step (S7) for kneading a radioactive waste liquid containing the hardly soluble salt with the cement. Solidified waste 16 is produced.

In the dissolution step (S1), uranium, transuranium elements having a higher atomic number than uranium, transuranium elements such as plutonium, other radionuclides or other radioactive nuclides, or radioactive wastes containing the compounds thereof from an organic acid solution with uranium, transuranium elements or other radionuclides Dissolve. In the dissolution process, the radioactive waste to be dissolved by the organic acid is, for example, uranium waste containing (for example, adhering) uranium and / or uranium compounds generated from uranium handling facilities or facilities transported between uranium handling facilities. Things. Specific examples of uranium waste include bricks, diatomaceous earth, ore waste, sludge, chemical trap material (NaF), sediments mainly composed of CaF ₂ , Al ₂ O, to which uranium and / or uranium compounds are attached. _3, wastes such as UF ₄ and the like.
Among these radioactive wastes, so-called fluoride-based uranium wastes such as sodium fluoride (NaF) and calcium fluoride (CaF ₂ ), which are mainly used as chemical trapping agents for uranium exhaust gas, are more effective with organic acids. Therefore, it is a suitable object.

In the dissolution step, examples of the organic acid used include formic acid and oxalic acid. The concentration (% by weight) of the organic acid used and the amount used can be appropriately determined according to the type of radioactive waste to be dissolved. In addition, the conditions for dissolving the radioactive waste with the organic acid, for example, the temperature, the dissolution time, and the like can be appropriately determined according to the type of the radioactive waste to be dissolved.
Of these organic acids, formic acid is preferred because it can be easily decomposed by ozone or hydrogen peroxide when remaining.

  In the solid-liquid separation step (S2), unnecessary residues are separated from the uranium-containing organic acid solution dissolved in the dissolution (S1) by, for example, a solid-liquid separator. For example, in the case of fluoride-based uranium waste, sodium fluoride, calcium fluoride, and the like are separated as a solid. The separated solid is then subjected to a predetermined treatment if necessary, and then disposed of according to the purpose.

Next, the precipitate generation step (S3) detects the redox potential in the redox reaction in which the uranium in the solution obtained after the solid is separated in the solid-liquid separation step and the oxidizing agent are detected. Based on the results, the amount of oxidant added is adjusted to produce a precipitate. The reaction between uranium (containing organic acid solution) and the oxidizing agent is performed by the following two formulas UO ₂ (OOCH) ₂ + 3H ₂ O ₂ → UO ₄ + 2CO ₂ + 4H ₂ O
UO ₂ (OOCH) ₂ + 2H ₂ O ₂ → UO ₂ (OH) ₂ + 2CO ₂ + 2H ₂ O
It is assumed that this is due to at least one of the following, and it is assumed that a uranyl hydroxide (UO ₂ (OH) ₂ ) precipitate or a uranium peroxide (UO ₄ ) precipitate is generated.
Here, simply adding a large amount of an oxidizing agent causes segregation of precipitate formation (segregation of reaction) and the like, and a sufficient amount of uranyl hydroxide (UO ₂ (OH) ₂ ) precipitate or uranium peroxide ( UO ₄ ) A precipitate may not be obtained. In this precipitate formation reaction, the oxidation-reduction potential is set to a predetermined potential or higher, and by adjusting the addition amount of the oxidizing agent at this oxidation-reduction potential, segregation of precipitate formation does not occur, and a sufficient amount of uranium can be produced. A precipitate (uranium hydroxide (UO ₂ (OH) ₂ ) precipitate or uranium peroxide (UO ₄ )) can be obtained.

The redox potential of the above redox reaction is preferably 400 mV or more. When the oxidation-reduction potential is less than 400 mV, there is a possibility that a sufficient amount of uranium precipitate (uranium hydroxide (UO ₂ (OH) ₂ ) precipitate or uranium peroxide (UO ₄ )) cannot be obtained. is there.

  Examples of the oxidizing agent used in this step include hydrogen peroxide and ozone. Of these, hydrogen peroxide is easy to handle, and ozone is a gas, so the amount of waste generated can be reduced.

  Further, the addition amount and the addition rate of the oxidizing agent are appropriately determined according to the type of radionuclide contained in the organic acid (for example, the type of radioactive waste dissolved in the organic acid (for example, type of uranium waste)). be able to. For example, the addition amount of the oxidizing agent can be determined to be 3 times or more equivalent of uranium.

  Next, this process will be further described. First, formic acid is added to a uranium-containing organic acid (for example, formic acid) solution to adjust the initial pH to 1.5. This is because when the pH is high (for example, 2 or more), uranium precipitates may not be easily generated. Next, an oxidizing agent (for example, hydrogen peroxide) is added to the uranium-containing formic acid solution. By the addition of the oxidizing agent, the decomposition of formic acid occurs predominantly and the redox potential increases. Thereafter, when a predetermined oxidation-reduction potential (for example, 400 mV) is reached, the generation of uranium precipitates becomes more dominant than the decomposition of formic acid, and the uranium precipitates are generated well. At this time, the oxidation-reduction potential is a value within a predetermined range. Here, the uranium precipitate can be more effectively generated by increasing the addition amount of the oxidizing agent and / or increasing the addition rate. The increase in the addition amount of the oxidizing agent and / or the increase in the addition rate can be appropriately determined according to the concentration of uranium in the organic solution. The formation of uranium precipitates at this oxidation-reduction potential does not cause segregation of precipitate formation, and a desired amount of precipitates can be obtained satisfactorily.

In addition, when the addition of the oxidizing agent is continued after the generation of the uranium precipitate, the decomposition reaction of formic acid in the solution is restarted. After the formic acid decomposition reaction is resumed, the redox potential fluctuates (eg, rises). Therefore, for example, by detecting a change (for example, an increase) in the oxidation-reduction potential with the oxidation-reduction potential measuring device 2 installed in the precipitation reaction tank 1, it can be determined that the generation of the uranium precipitate has been completed.
For example, the fluctuation of the oxidation-reduction potential is detected by calculating an inclination angle (for example, a value close to 0) of the increase of the oxidation-reduction potential at the time of uranium precipitate formation, and the inclination angle becomes equal to or greater than a predetermined threshold value. Can be determined as the end of the generation of uranium precipitate. This predetermined threshold can be appropriately determined according to the concentration of the organic acid (formic acid), the addition rate of the oxidizing agent, and the like.

  In the solid-liquid separation step (S4), the uranium precipitate obtained in the precipitate generation step is separated by, for example, a solid-liquid separator. By this solid-liquid separation, the uranium precipitate can be treated as a solid waste. Further, when uranium in the uranium precipitate is reused, it can be reused after a predetermined treatment.

In the organic matter decomposition step (S5), the organic matter (for example, formic acid) remaining in the organic acid (formic acid) solution after separating the solid waste in the solid-liquid separation step is converted into an oxidizing agent (for example, hydrogen peroxide). Add and disassemble. When hydrogen peroxide is added to the formic acid solution, decomposition of formic acid occurs and the redox potential increases. The decomposition of formic acid is the following reaction,
HCOOH + H ₂ O ₂ → CO ₂ + 2H ₂ O
Is done.

  When the decomposition of the organic substance in the organic substance decomposition step ends, the redox potential varies (for example, the increase in the redox potential stops). Therefore, for example, the oxidation-reduction potential measuring device 2 can determine whether or not the decomposition of the organic matter has been completed by detecting the change in the oxidation-reduction potential, for example, the stop of the increase in the oxidation-reduction potential. For example, the change in the oxidation-reduction potential is detected by calculating the inclination angle of the increase in the oxidation-reduction potential at the time of decomposition of the organic substance, and determining that the decomposition of the organic acid is completed when the inclination angle becomes a predetermined threshold value or less. can do. This predetermined threshold value can be appropriately determined according to the type of organic matter to be decomposed, the addition rate of the oxidizing agent, and the like. For example, this threshold value (inclination angle) can be determined as 0 °.

  The redox potential measuring device 2 for detecting the redox potential is installed in the post-treatment tank 8 when the organic matter is decomposed in the post-treatment tank (for example, organic acid decomposition layer) 8. Further, when returning the solution containing the organic acid to the precipitation reaction tank 1 after separating the uranium precipitate, the oxidation-reduction potential measuring device 2 installed in the precipitation reaction tank 1 can be used.

  Thus, the end of decomposition of the organic matter can be easily determined by detecting the redox potential.

  In the carbonate generation step (S6), an alkaline earth metal hydroxide is added to the carbonate ions generated in the organic matter decomposition step (S5) to generate a hardly soluble alkaline earth metal carbonate.

In the organic substance decomposition step, the organic acid (formic acid) decomposed by blowing hydrogen peroxide or ozone mainly becomes water and carbon dioxide. This carbon dioxide is dissolved in water and carbonate ions are present in a large amount in the solution. An alkaline earth metal hydroxide is added to the solution to produce an alkaline earth metal carbonate.
In the case where a cement solidification process as described later is performed on an acidic waste liquid such as a carbonic acid solution, it is necessary to perform a neutralization process. When an alkali metal hydroxide such as sodium hydroxide is used as the neutralizing agent, the neutralization reaction proceeds by the following reaction.
NaOH + H ₂ CO ₃ → NaHCO ₃ + H ₂ O
After this, when the hydration reaction of the cement material proceeds, the following reaction CaO (cement material) + NaHCO ₃ → CaCO ₃ + NaOH
As a result, an alkali metal hydroxide is formed again. On the other hand, when an alkaline earth metal hydroxide is used as the neutralizing agent, the following reaction Ca (OH) ₂ + H ₂ CO ₃ → CaCO ₃ (precipitation) + 2H ₂ O
As shown in FIG. 1, a slightly soluble alkaline earth metal carbonate is formed.

  Examples of the alkaline earth metal hydroxide include calcium hydroxide and magnesium hydroxide.

  The amount of the alkaline earth metal hydroxide to be reacted with the carbonate ion is preferably 1 equivalent or more of the carbonate ion, but even when it is not more than 1 equivalent of the carbonate ion, it is almost equal to 1 equivalent of the carbonate ion. In contrast, the amount of carbonate ions remaining hardly interferes with the cement hydration reaction. This is presumably because the alkaline earth metal hydroxide in the cement to be kneaded thereafter forms a carbonate salt that is hardly soluble with carbonate ions.

  The carbonate generation step (S6) includes not only the carbonate ions generated in the organic matter decomposition step (S5) but also the carbonic acid generated by the decomposition of the organic acid (formic acid) in the precipitate generation step (S3). Alkaline earth metal hydroxide is added to the ions, that is, carbonate ions present in the solution after separation of the uranium precipitate in the solid-liquid separation step (S4), and the hardly soluble alkaline earth Metal carbonates can also be produced.

  As described above, by using the alkaline earth metal hydroxide, it is possible to avoid the production of the hardly soluble alkaline earth metal carbonate during neutralization and the hindering of the hydration reaction of the cement material. Therefore, stable solidified product can be achieved by using alkaline earth metal hydroxide as the neutralizing agent. Further, it is possible to avoid the alkali metal hydroxide being regenerated again after the cement is solidified, as in the case of using the alkali metal hydroxide as a neutralizing agent. When contact is made with groundwater after disposal of waste, this alkali metal hydroxide is also readily soluble and excessively alkalizes groundwater, but alkaline earth metal hydroxides are alkaline earth Since only metal carbonate is generated, it is possible to avoid changing the liquidity of groundwater more than necessary.

  In the cement solidified body producing step (S7), the solution (waste liquid) in which the poorly soluble alkaline earth metal carbonate is produced in the carbonate producing step and the cement-based material are kneaded and cured, Perform solidification processing.

As the cement-based material, various cements such as blast furnace cement and fly ash cement can be used in addition to normal Portland cement and various Portland cements mainly composed of CaO, SiO ₂ and Al ₂ O ₃ . Among these, those containing Portland cement and / or blast furnace cement are preferable because the quality after cement solidification is good. Moreover, aggregates, such as sand and gravel, can also be added as needed.

  The blending ratio of the solution containing the alkali metal carbonate obtained by the above-described carbonate production step and the cementitious material can be appropriately determined depending on the concentration of the solution used, the type of the cementitious material, and the like. Further, the kneading conditions for kneading these materials, for example, the kneading temperature, the kneading time, etc. can be appropriately determined depending on the concentration of the solution used, the type of cementitious material, and the like. The kneading time can be, for example, 10 minutes or longer. The kneading temperature can be, for example, room temperature.

  In addition, the time for curing the kneaded material obtained by kneading can be appropriately determined depending on the type of cementitious material used. For example, it may be 2 days or more.

  By such cement solidification treatment, it is possible to form a solidified body that suppresses the occurrence of breathing and has high compressive strength.

  As described above, according to the present embodiment, in order to stably generate the uranium precipitate, the uranium precipitate is generated by adjusting the addition amount of the oxidizing agent while detecting the oxidation-reduction potential. Therefore, a uranium precipitate can be generated more efficiently. Therefore, uranium can be separated and recovered more effectively.

  Next, specific examples and comparative examples of the present invention will be shown respectively, and the effects of the present invention will be described in more detail by comparing these examples and comparative examples.

Example 1
The composition of the treatment liquid in which uranium is eluted from the waste containing uranium or uranium and fluorine, which is the premise of the present invention, with an organic acid is not constant when the generation source is different or the treatment history is not the same. For this reason, in this example, an experiment was performed by creating a simulated solution.
A solution obtained by dissolving 50 g of a formic acid solution as an organic acid and dissolving 1 g of a mixture of uranyl nitrate and calcium fluoride in a predetermined mixing ratio was used. The solution was adjusted to pH 1.5 using formic acid. Thereafter, the formation of a uranium precipitate (uranium hydroxide and / or uranium peroxide) was confirmed while hydrogen peroxide was added. From the time when the redox potential reached 400 mV, good formation of uranium precipitate was obtained. Hydrogen peroxide was further added at this oxidation-reduction potential. As a result, a uranium precipitate of 90% or more was obtained by adding 3 times equivalent hydrogen peroxide to uranium.

(Example 2)
The same procedure as in Example 1 was performed except that 5 g of a mixture of uranyl nitrate and calcium fluoride having the same mixing ratio as in Example 1 was used. From the time when the oxidation-reduction potential reached 400 mV, good precipitation of uranium oxide was obtained. In a solution in which 5 g of a mixture of uranyl nitrate and calcium fluoride having a predetermined blending ratio is dissolved, only about 50% of a precipitate is obtained by adding 3 times the equivalent amount of hydrogen peroxide to uranium, and more than 90% Further recovery of uranium required the addition of hydrogen peroxide. The amount of hydrogen peroxide added was finally 30 times equivalent to the amount of uranium.
This is considered to be due to the predominance of the formic acid decomposition reaction among the uranium precipitate formation reaction and formic acid decomposition reaction, which are thought to occur by the addition of hydrogen peroxide to the uranyl nitrate and calcium fluoride mixture. .

  Next, detection (monitoring) of a change in oxidation-reduction potential due to the addition of an oxidizing agent in the precipitation formation reaction will be described. FIG. 3 is a diagram showing a change in redox potential when hydrogen peroxide is added to 50 ml of a 0.1 M formic acid solution. First, the initial pH of the solution required for the precipitation treatment was adjusted to pH = 1.5 using formic acid. This is because if the initial pH is 2 or more, it is considered that the formation of a good precipitate of the uranium compound may not be obtained. For reference, the initial pH adjustment (pH = 1.5) is described not only for the addition of formic acid, which is an organic acid, but also when hydrochloric acid is used.

  As shown in FIG. 3, when the solution contains only formic acid, the oxidation-reduction potential is initially low. The addition of hydrogen peroxide increases the redox potential slowly. Furthermore, if the addition of hydrogen peroxide is continued, the tendency for the oxidation-reduction potential to increase is moderated when the oxidation-reduction potential is around 400 mV. In this way, when formic acid is used, the oxidation-reduction potential increases slowly, so by detecting (monitoring) the oxidation-reduction potential, it is possible to identify the oxidation-reduction potential that forms uranium precipitates well. Can do. FIG. 3 is a simulation performed for a system that does not contain uranium, but the same tendency is observed for a system that contains uranium.

  Thus, when a formic acid solution is used, the reaction depends on the oxidation-reduction potential in the solution in order to stably produce uranium oxide, so the amount of hydrogen peroxide added can be adjusted while detecting the oxidation-reduction potential. By adjusting, it becomes possible to determine the precipitation conditions and to suppress segregation of precipitate formation.

Example 3
In this embodiment, an alkaline earth metal hydroxide is added to carbonate ions to form a hardly soluble alkaline earth metal carbonate, and a solution containing the alkaline earth metal carbonate and cement are kneaded. Next, a method for performing cement hardening treatment for curing will be described with reference to FIG. FIG. 4 is a diagram showing a specific procedure for solidifying the carbonate ion-containing solution (waste liquid) after separation of the uranium precipitate.

  20 ml (18) of a saturated calcium hydroxide solution was added to and mixed with 100 ml (17) of a carbonate ion concentration of 0.01 mol / L (S8). To this mixed solution, 250 g (19) of ordinary Portland cement was added as a cement-based material and kneaded for 10 minutes (S9). The kneaded material obtained by this kneading was allowed to stand for 2 days in an environment at a temperature of 20 ° C. (S10).

  When the sample produced as described above was visually confirmed after 2 days, it was confirmed that the cement solid waste 20 which did not generate the breathing which was excess water was obtained. Further, after one week, when the compressive strength of the sample was measured, it was confirmed that the sample had a strength of 1.5 MPa or more.

  Thus, the solidification which suppresses generation | occurrence | production of breathing and has a high compressive strength by the cement solidification after implementing said pre-processing about the carbonate ion which exists in the solution after isolation | separation of a uranium precipitate. Is possible.

It is a flowchart which shows the procedure of the processing method of the radioactive waste which concerns on the 1st Embodiment of this invention. It is a figure which shows typically the principal part structure of an example of the processing apparatus of the radioactive waste used for the 1st Embodiment of this invention. It is a figure which shows the relationship between oxidation-reduction potential and precipitation formation by an oxidizing agent. It is a figure which shows the specific procedure of the solidification process of the carbonate ion containing solution (waste liquid) after isolation | separation of a uranium deposit.

Explanation of symbols

  DESCRIPTION OF SYMBOLS 1 ... Precipitation reaction tank, 2 ... Redox potential measuring device, 3 ... Stirring blade, 4 ... Oxidizing agent (for example, hydrogen peroxide or ozone) addition line, 5 ... Waste liquid extraction line, 6 ... Transfer pump, 7 ... Solid-liquid separation 8 ... post-treatment tank, 10 ... oxidant storage tank, 11 ... fluoride-based uranium waste, 12 ... solid waste, 13 ... uranium-containing formic acid solution, 14 ... solid (uranium) waste, 15 ... formic acid solution , 16 ... Cement solid waste, 17 ... Carbonic acid-containing solution, 18 ... Saturated calcium hydroxide solution, 19 ... Normal portland cement, 20 ... Cement solid waste

Claims (9)

A dissolution and separation step of dissolving and separating the uranium, the transuranium element or the radionuclide from a radioactive waste containing uranium, a transuranium element, a radionuclide or a compound thereof with an organic acid solution;
While detecting the redox potential in the redox reaction in which the uranium, transuranium element or radionuclide obtained by the dissolution and separation step is reacted with an oxidizing agent, the detected potential does not cause segregation of precipitation formation. A precipitate generating step for increasing the addition amount of the oxidizing agent and / or increasing the addition rate to generate a precipitate in the case of
And a separation step of separating the precipitate obtained by the precipitate generation step.   2. The radioactive waste processing method according to claim 1, wherein the radioactive waste is fluoride-based uranium waste.   The method for treating radioactive waste according to claim 1 or 2, wherein the oxidation-reduction potential in the precipitate generation step is 400 mV or more when the organic acid solution is a formic acid solution.   The radioactive waste treatment method according to any one of claims 1 to 3, wherein the oxidizing agent in the precipitate generation step is hydrogen peroxide or ozone. A decomposition step of decomposing an organic substance remaining in the solution separated from the precipitate by the separation step by adding an oxide;
The method for treating a radioactive liquid waste according to any one of claims 1 to 4, further comprising a determination step of determining completion of the decomposition step by detecting a change in oxidation-reduction potential. Wherein the solution obtained by the separation step, of claims 1 to 4, characterized in that it comprises an alkaline earth metal carbonate product to obtain an alkaline earth metal carbonate by adding an alkaline earth metal hydroxides The radioactive waste processing method as described in any one of Claims. The radioactive waste processing method according to any one of claims 1 to 4 and 6, further comprising a cement solidified body generating step of kneading the solution obtained by the separation step and cement to generate a cement solidified body. An alkaline earth metal carbonate production step of adding an alkaline earth metal hydroxide to the solution obtained by the separation step or the solution obtained by the decomposition step to obtain an alkaline earth metal carbonate;
The radioactive waste processing method according to claim 5, further comprising: A solution obtained by the separation step or a solution obtained by the decomposition step, and cement.
A cement solidified body producing step for producing a cement solidified body by kneading
The processing method of the radioactive waste of Claim 5 or 8 provided with these.

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