JP2008139265A - Method and apparatus for treating radioactive waste - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a radioactive waste treatment method capable of easily and efficiently separating and recovering radionuclides from radioactive waste containing radionuclides such as uranium and fluorine substances as well as of reducing the amount of generation of secondary waste accompanying the treatment. <P>SOLUTION: The radioactive waste treatment method comprises: a process for preparing an eluate in which uranium is eluted by contacting uranium containing waste with water or weakly acidic water with not more than 0.1 mol of acid concentration; and a process for separating and recovering the uranium from the eluate. <P>COPYRIGHT: (C)2008,JPO&INPIT

Description

  The present invention relates to a method and apparatus for treating radioactive waste, and particularly, the amount of secondary waste generated by the treatment is small, and the above-mentioned radionuclide is easily obtained from attached radioactive waste containing radioactive nuclides such as uranium and fluorine components. Further, the present invention relates to a radioactive waste processing method and a processing apparatus that can be efficiently separated and recovered.

  As a method for separating and recovering uranium and other radionuclides from waste containing uranium and uranium compounds generated from uranium handling facilities, or waste containing these uranium components, the following various treatment decontamination methods are available. The proposal has been put into practical use.

That is, for decontamination of radioactively contaminated metal waste, a method of decontamination with sulfuric acid as a decontamination solution (for example, see Patent Document 1), a method of decontamination with an organic acid (for example, Patent Literature) 2) has been proposed. In addition, as a method for separating uranium from radioactive liquid waste, a method of repeating the extraction step after eluting U and heavy metal components with a nitric acid-based liquid (for example, see Patent Document 3) has been reported. Furthermore, as a method for treating waste containing fluoride-based uranium, a method of treating with hydrochloric acid or a method of treating with an organic acid (for example, Patent Document 4) has been reported.
JP 2000-199800 A Japanese Patent Laid-Open No. 9-113690 Japanese Patent Laid-Open No. 5-80195 JP 2004-20251 A

  The treatment methods disclosed in Patent Documents 1 and 2 described above are intended to decontaminate metal waste to which uranium has adhered. In the treatment methods disclosed in these documents, the secondary waste generated with the treatment is not particularly mentioned in the case of Patent Document 1, for example, in the case of treating with sulfuric acid. Since a neutralization treatment is usually required, a large amount of neutralized salt containing uranium can be assumed to be generated as secondary waste, and there is a problem that a great deal of labor and treatment costs are required for the treatment. is there.

  On the other hand, in the case of the treatment method disclosed in Patent Document 2, an organic acid is used for decontamination of radioactive metal waste, and the metal component eluted in the decontamination waste liquid is recovered with an ion exchange resin, and the recovered ion is recovered. The exchange resin is regenerated with sulfuric acid. For this reason, it is necessary to treat the used sulfuric acid, and it is usually assumed that a large amount of neutralized salt is generated as a secondary waste by the neutralization treatment. Similarly, the treatment requires a great deal of labor and treatment costs. There is a difficult point.

  Further, the radioactive waste liquid treatment method disclosed in Patent Document 3 provides a technique for separating uranium from the radioactive waste liquid. Nitric acid is added to the acidic waste liquid, and uranium is extracted by a solvent extraction method using tributyl phosphoric acid (TBP). The method of separating is adopted. In this case, the nitric acid-based solvent extraction method is excellent in the separation performance of uranium, but has a problem that the waste liquid treatment is very difficult. In other words, nitrate-based nitrogen has strict effluent standards according to the Water Pollution Law, and even if neutralization is carried out, storing a large amount of nitrate is regulated by the Fire Service Act, which is virtually impossible. Therefore, it is necessary to recover nitric acid from the waste liquid at a very high recovery rate. For this purpose, a large-scale nitric acid recovery device is required, and there is a problem that the installation cost and the operation cost of the processing device become high. In addition, tributyl phosphate generated as a waste solvent has a high boiling point of 289 ° C., and is difficult to recover by evaporation under normal conditions. Decomposition is required for disposal. Moreover, since decomposition products such as calcium phosphate are generated by this decomposition treatment, it has been pointed out that these treatment products become a large amount of secondary waste.

  On the other hand, a hydrochloric acid treatment technology for fluoride uranium waste has also been proposed. That is, a method in which uranium waste is dissolved in hydrochloric acid, uranium is precipitated from the solution as uranium peroxide, and a method of removing it with a chelate resin is used in combination. When carrying out the uranium precipitation treatment, a masking agent is added, but this masking agent may eventually become a large amount of secondary waste. In addition, the above-mentioned chelate resin, which has been tried to be used in the present invention, is generally difficult to elute and regenerate. When the resin is repeatedly reused in order to reduce the amount of waste resin, a large amount of regenerated waste liquid containing uranium is obtained. May occur and the amount of uranium waste may increase. Therefore, the resin cannot be repeatedly used, a large amount of waste resin is generated, and the detoxification treatment requires a great cost. Furthermore, since the treatment is based on hydrochloric acid, the final wastewater needs to be neutralized, and there is a problem that a large amount of neutralized salt is generated.

  Furthermore, the said patent document 4 is disclosing the technique which processes fluoride type | system | group uranium waste with an organic acid type | system | group. Since this method uses a decomposable organic acid as a medium for dissolving waste, the amount of neutralized salt generated can be significantly reduced compared to the other methods described above. However, in the above hydrochloric acid treatment, it is assumed that the fluoride waste is completely dissolved by the acid, and even if the non-contaminated part is mixed in the waste, the entire treatment object is dissolved. Therefore, it is pointed out that the concentration of dissolved components in the solution is excessively increased more than necessary, and as a result, it becomes a secondary waste source.

  As described above, the known decontamination and treatment methods for uranium wastes generate a large amount of secondary waste such as neutralized salts, waste ion exchange resins, and waste solvents originating from chemicals used in the treatment. There are challenges. In addition to secondary waste originating from pharmaceuticals, there is also an increase in the amount of secondary waste generated due to dissolution of non-contaminated parts that do not need to be treated by dissolving the waste with acid. It is.

  The present invention has been made to solve these problems. Particularly, the amount of secondary waste generated by the treatment is small, and the above-mentioned radioactive waste is deposited from the attached radioactive waste containing radionuclides such as uranium and fluorine components. It is an object of the present invention to provide a radioactive waste processing method and processing apparatus capable of easily and efficiently separating and recovering nuclides. Specifically, the present invention provides a method and a processing apparatus for separating and recovering uranium from radioactive waste to which uranium and uranium compounds are attached, and fluoride waste such as sodium fluoride, which is mainly a trap agent for uranium exhaust gas. For the purpose.

  In order to achieve the above object, the method for treating radioactive waste according to the present invention provides a effluent from which uranium is eluted by contacting the waste adhering uranium with water or weakly acidic water having an acid concentration of 0.1 mol or less. And a step of separating and recovering uranium from the eluate.

  Water or weakly acidic water is used as the medium for eluting the uranium component. Since the solubility of the uranium component in water is good, it is possible to elute the uranium component by using water as the elution medium.

  In particular, by using weakly acidic water as the elution medium, it is possible to prevent uranium components and the like in the eluate from being precipitated in the ion exchange resin tower. The weakly acidic water is an acid solution having a low degree of ionization, and specifically needs to be a weakly acidic solution having an acid concentration of 0.1 mol / liter or less. If the acid concentration is within the weakly acidic range of 0.1 mol or less, it is possible to effectively prevent the uranium component from being precipitated inside the ion exchange resin tower or the like during the treatment step.

  On the other hand, in the case of a strongly acidic solution having an acid concentration exceeding 0.1 mol, even the non-contaminated portion in the waste is dissolved, so the amount of secondary waste produced as a by-product due to a reduction in processing efficiency. Will also increase. Therefore, the acid concentration of the weakly acidic water to be used is defined as 0.1 mol or less, preferably 0.05 mol or less, and more preferably 0.01 mol or less.

  In the above radioactive waste processing method, uranium adheres to the waste and contains fluorine. In the separation and recovery step, uranium is separated and recovered from the eluate and fluorine is separated and recovered. It is preferable to do. According to this method, uranium and fluorine can be effectively recovered from fluoride-based uranium waste.

  Furthermore, in the said radioactive waste processing method, it is preferable to isolate | separate the uranium from the eluate which eluted the said uranium by making a slightly soluble compound process. Thus, the uranium compound can be easily separated by an ordinary solid-liquid separation device such as a filter by subjecting uranium to a hardly soluble compound.

  Moreover, in the said radioactive waste processing method, it is preferable to use together the uranium separation by the said sparingly soluble compound formation process, and the uranium separation by an ion exchange resin process in the said separation-and-recovery process. The final separation efficiency of uranium can be further increased by using uranium separation by a hardly soluble compounding treatment having a high separation efficiency and uranium separation by an ion exchange resin treatment having a high separation accuracy.

  Furthermore, in the method for treating radioactive waste, it is preferable that the poorly soluble compound of uranium generated by the hardly soluble compounding treatment is uranyl hydroxide or uranium dioxide. If the hardly soluble compound is uranyl hydroxide or uranium dioxide, the solubility in water is small, so that the uranium concentration in the eluate can be effectively reduced. Moreover, solid-liquid separation is easy, and a uranium component can be recovered with a high yield.

  In the radioactive waste treatment method, the ion exchange resin used in the ion exchange resin treatment is preferably a chelate resin. This chelate resin reacts with the ions in the eluate to form a chelate compound having a cyclic structure, and exhibits an ion exchange action. Uranium remaining in a small amount in the eluate after the process can be efficiently removed to the ppb level.

  Furthermore, in the above-mentioned radioactive waste processing method, a ligand (complexing agent) that forms a complex anion by binding to uranium is added to the eluate in the previous stage of uranium separation by the ion exchange resin treatment, It is preferable to separate the uranium-containing component by treating the produced complex anion with an anion exchange resin. The ligand is a molecule, ion, or group bonded to the central atom of the chelate compound or coordination compound. In the present invention, formic acid that forms a complex with uranium is a ligand (complexing agent). Used as.

  After uranium is treated as a poorly soluble compound and separated as a precipitate in the previous step, the complexing agent such as formic acid is added to the eluate after the precipitation separation, and uranium generates complex anions and anions. It can be adsorbed on the exchange resin and effectively separated. By this separation step, uranium in the eluate can be removed to a concentration of ppb level.

Furthermore, in the said radioactive waste processing method, it is preferable to isolate | separate by carrying out the poorly soluble compound process as a said fluorine separation-and-recovery method. As the poorly soluble compound of fluorine, there is a form such as calcium fluoride (CaF ₂ ) generated by adding calcium to the eluate and neutralizing it. Since the solubility of calcium fluoride in water is extremely small, the concentration of fluorine in the eluate can be greatly reduced by producing a precipitate of calcium fluoride. Further, the fluorine concentration in the eluate can be further reduced by adding a fluorine removal step using an ion exchange resin after the treatment process.

  Moreover, in the said radioactive waste processing method, it is preferable to implement the said hardly soluble compound conversion process of fluorine simultaneously with the hardly soluble compound conversion process of uranium. In other words, when calcium is added to the uranium eluate first and neutralization is subsequently carried out, it is possible to simultaneously produce uranium hydroxide and calcium fluoride, which are hardly soluble compounds, and to simultaneously separate uranium and fluorine. Thus, the configuration of the processing facility can be simplified.

  Furthermore, in the said radioactive waste processing method, it is preferable that the waste is a uranium exhaust gas treatment agent to which uranium adheres, and the uranium exhaust gas treatment agent from which uranium components are eluted and separated is reused. In this way, by using a uranium exhaust gas treatment agent as waste to which uranium adheres, and elution and separation of uranium components by the above-described treatment method, it becomes possible to recycle and reuse the used uranium exhaust gas treatment agent. .

  In addition, the processing apparatus for carrying out the above-mentioned radioactive waste processing method is a method of contacting the waste adhering uranium with water or weakly acidic water having an acid concentration of 0.1 mol or less to dissolve uranium from the effluent. And a means for preparing and a means for separating and recovering uranium from the eluate.

  According to the method for treating radioactive waste according to the present invention having the above-described configuration, since neutral water or weakly acidic water is used as an elution medium for eluting uranium components, it dissolves even to the non-contaminated portion of the waste. It is possible to preferentially elute uranium while suppressing the amount of non-contaminating components dissolved, suppressing the mixing of unnecessary components into the processing solution, and secondary disposal resulting from them. It becomes possible to greatly reduce the amount of generated matter.

  Next, an embodiment of a method for treating radioactive waste according to the present invention will be specifically described based on the following examples with reference to the accompanying drawings.

  FIG. 1 shows a typical flow chart showing the features of the radioactive waste processing method according to the present invention. In the method for treating radioactive waste according to the present embodiment, water or a very weak acid such as uranium or waste containing uranium and fluorine, for example, a dilute acid having a concentration of 0.1 mol or less, more preferably a concentration of 0.1. It consists of a uranium elution step in which an elution medium composed of 01 mol or less of weakly acidic water is acted to elute uranium and separate it from insoluble matter, and a separation step to separate uranium or uranium and fluorine from the eluate.

  When water or a very weak acid is added to uranium or waste containing uranium and fluorine, dissolution of non-contaminating components in the waste is suppressed, and adhering uranium can be preferentially eluted. For example, FIG. 2 shows the dissolution rate of uranium adhering to sodium fluoride (NaF) pellets commonly used as a treatment agent for exhaust gas containing uranium, together with the dissolution rate of NaF itself.

  Specifically, when 20 mL of pure water (pH 6.9) is added to 1 g of sodium fluoride pellets to which uranium has adhered and stirred for 2 hours at room temperature, the adhering uranium is dissolved in almost 100% water (in the elution medium). did. On the other hand, when 1 g of sodium fluoride pellets not adsorbing uranium was dissolved under the same conditions, the dissolution rate of sodium fluoride was 45%. Undissolved 55% sodium fluoride can be separated from uranium by solid-liquid separation and can be regenerated as pellets.

  In the next separation step, uranium and fluorine were chemically separated from the eluate thus obtained, and the uranium and fluorine concentrations in the solution were below the wastewater standard value. After separating uranium and fluorine, wastewater can be discharged to the environment.

  In the conventional treatment method, all waste containing uranium is dissolved with acid, so a large amount of secondary waste is generated by dissolving even non-contaminated parts that do not need to be treated. In the treatment method of the example, since neutral water or weakly acidic water is used as an elution medium, it is possible to preferentially elute uranium components while suppressing the amount of dissolution of non-contaminating components, It became possible to suppress the mixing of unnecessary components into the treatment solution and reduce the amount of secondary waste generated due to them.

  For example, when sodium fluoride pellets of the above-mentioned uranium exhaust gas treating agent are treated, only uranium is eluted with water or the like as in this example, compared with the case where the entire amount of uranium components including sodium fluoride is dissolved. In this case, it was possible to reduce the elution amount of sodium fluoride in the treatment solution to half or less. As a result, it has been found that the total amount of recovered materials containing uranium and fluorine generated in the latter stage is greatly reduced, and the disposal cost of radioactive waste can be greatly reduced.

  In the treatment method of this embodiment, salts such as sodium remaining in the solution are released into the environment. However, since the salt concentration can be reduced, the load on the environment can be reduced. In the case of a conventional treatment method using an inorganic acid, since neutralized salts derived from inorganic acid are also contained in the treatment liquid, the salt concentration in the discharged water becomes very high. In the treatment method of the example, the neutral salt derived from the reagent is unnecessary. In addition, even when an extremely thin acid is used, if a decomposable organic acid is used, the salt content derived from the reagent can be reduced to zero. Thus, by significantly reducing the amount of salinity in the discharged water, the influence on the environment can be reduced especially when drainage is discharged into the river.

  Further, as in this embodiment, by using water or an extremely thin acid as an elution medium, it is possible to reduce the cost of the chemical reagent as compared with a conventional treatment method using an acid at a high concentration.

  In addition, when the waste to be treated is valuable, such as sodium fluoride pellets, it is possible to collect and reuse the insoluble components after uranium elution, which also reduces the operating cost of the treatment equipment. Become.

  The processing flow of 2nd Embodiment of the processing method of the radioactive waste based on this invention is shown in FIG. In the method of this example, uranium and fluorine in the eluate are precipitated and separated from the eluate by changing the form of a poorly soluble compound having low solubility. In the processing flow shown in FIG. 3, the fluorine separation step is performed after the uranium separation step. However, these steps can be performed simultaneously. Moreover, it is also possible to add the fluorine removal process by an ion exchange resin as needed after the process of making a fluorine insoluble compound.

Examples of the poorly soluble uranium compound include uranium dioxide (UO ₂ ) and uranyl hydroxide (UO ₂ (OH) ₂ ). Since it is considered that uranium exists mainly in the form of uranyl ions (UO ₂ ²⁺ ) in the eluate, uranyl hydroxide is generated by neutralizing this eluate. Since the solubility product (Ksp) of uranyl hydroxide is 1.5 × 10 ⁻²⁰ , the concentration of uranium dissolved in pH 7 water is 0.36 mg / L. Therefore, by generating this precipitate, the uranium concentration in the eluate can be reduced to 0.36 mg / L. Uranium dioxide can be obtained by reducing uranyl ions to a pH of weak acid or higher. The solubility of uranium dioxide in water is 0.81 mg / L at room temperature, and the uranium concentration is 0.71 mg / L. Therefore, by producing this compound, the uranium concentration in the solution can be reduced to 0.71 mg / L.

Examples of the poorly soluble fluorine compound include calcium fluoride (CaF ₂ ). Calcium fluoride is produced by adding calcium to the eluate and neutralizing it. The solubility of calcium fluoride in water is 16 mg / L at room temperature, and the fluorine concentration is 7.8 mg / L. Therefore, by producing a calcium fluoride precipitate, the fluorine concentration in the solution can be reduced to 7.8 mg / L. Moreover, the fluorine concentration in the solution is further reduced by adding a fluorine removal step using an ion exchange resin to the subsequent stage.

  By appropriately selecting the uranium and fluorine precipitation generating operation, it is possible to carry out the separation of uranium and fluorine simultaneously or sequentially. When calcium is first added to the uranium eluate and subsequently neutralized, uranyl hydroxide and calcium fluoride are simultaneously generated, and uranium and fluorine can be separated simultaneously. Also, after neutralizing the eluate first to produce uranyl hydroxide first and separating uranium, adding calcium to the filtrate produces calcium fluoride, which makes it possible to separate uranium and fluorine separately. Become.

Based on legal regulations such as the Nuclear Reactor Regulation Law and the Mine Safety Law, the uranium drainage standard value outside the facility is ²³⁸ U concentration of 2 × 10 ⁻² Bq / cm ³ in the strictest chemical form. . When this radioactivity concentration is converted into elemental uranium concentration, it becomes 1.6 mg / L. In either case of the two uranium sparingly soluble compounding steps, the concentration of uranium in the solution can be reduced to the conversion reference value or less, so that the waste liquid can be discharged outside the facility in one step.

  On the other hand, for fluorine, the drainage standard value by the water pollution method is 15 mg / L in the sea area and 8 mg / L outside the sea area. By carrying out the above-described step of forming a fluorine-insoluble compound, the concentration of fluorine in the eluate can be reduced to a level equal to or lower than the standard value of drainage to the outside of the sea area, so that the waste liquid can be discharged out of the facility. However, when discharging to the outside of the sea area, the solubility of calcium fluoride is close to the drainage standard value, so it is desirable to add a fluorine removal step with a resin in the latter stage in anticipation of safety margin. Specifically, it is possible to remove fluorine to the ppb level by carrying out the ion exchange resin treatment.

  Since the uranium eluate can be discharged in this way, the necessity to store and dispose of radioactive waste liquid is eliminated, and the disposal cost is reduced. Further, when the separation of uranium and fluorine is performed at the same time, since the precipitation separation operation is performed once, the configuration of the processing apparatus and the operation operation are simplified and the cost is directly reduced. On the other hand, when fluorine is separated after first separating uranium, the amount of radioactive waste can be greatly reduced because the fluorine recovered material becomes non-uranium waste. Therefore, it is possible to reduce the overall processing cost by selecting the simultaneous separation or sequential separation of uranium and fluorine according to the uranium and fluorine contents in the processing object and the processing equipment scale. .

  FIG. 4 shows a processing flow of the third embodiment of the present invention. In the radioactive waste processing method according to the present embodiment, the uranium separation step consists of two steps, the first uranium separation step by the hardly soluble compounding treatment and the second uranium separation step by the ion exchange resin. Are combined. As the ion exchange resin, a resin capable of adsorbing cations is used.

In the first uranium separation step, the uranium component is separated as an insoluble substance by converting uranium into a poorly soluble compound, but since a second separation step is provided in the subsequent stage, a compound having a higher solubility than the emission standard value. Uranium can also be separated in shape. In addition to uranyl hydroxide and uranium dioxide described in Example 2, it is also possible to separate in the form of uranium peroxide (UO ₄ ) produced by adding hydrogen peroxide under acidic conditions.

  When the uranium peroxide is formed, the solubility in water is 6.1 mg / L at room temperature, and the uranium concentration is 4.8 mg / L. In the second separation step, it is preferable to remove uranium remaining in the eluate after the first separation step with a chelate resin or the like having a high adsorption capacity for polyvalent cations. The chelating resin has high uranium adsorption performance, and uranium can be efficiently removed to the ppb level.

  In this way, when uranium components and the like are eluted from the waste using weak acid, it is possible to generate uranium peroxide precipitate by adding hydrogen peroxide to the eluate to be the first uranium separation step. is there. Then, as the second uranium separation step, uranium is adsorbed by separating it from sodium monovalent cation with a chelating resin that is excellent in adsorptivity of polyvalent cations that have aminophosphate groups that can be used under acidic conditions. By doing so, the uranium concentration in the eluate can be reduced to the ppb level. By the above separation step, it becomes possible to cope with drainage from a facility having a stricter management standard than the above-mentioned drainage standard value with a sufficient margin.

  In the 1st uranium separation process, when uranyl hydroxide or uranium dioxide is produced by neutralization and separated, a chelate resin with an amidoxime group that has excellent uranium adsorption in the neutral region is used in the 2nd separation process. Is possible. Even when this chelate resin is used, it is possible to reduce the uranium concentration in the eluate to the ppb level, and it is possible to discharge the waste water having a safety margin with respect to the discharge standard. Furthermore, since the above chelate resin easily elutes uranium by adjusting the pH of the eluate to acidic or alkaline, it is easy to reuse the resin repeatedly by elution regeneration, and ion exchange is a secondary waste. It is possible to reduce the amount of resin generated.

  FIG. 5 shows a processing flow of the fourth embodiment of the present invention. In the radioactive waste processing method according to the present embodiment, the uranium separation step is composed of two stages, the first uranium separation step by the hardly soluble compound and the second uranium separation step by the ion exchange resin. It is configured by combining. Also, before the second uranium separation step, a complexing agent (ligand) for generating uranium and complex anions is added, and uranium is separated using an ion exchange resin capable of adsorbing anions. It is configured.

  According to the radioactive waste processing method according to the present example, as in Example 3, in the first uranium separation step, uranium is subjected to a hardly soluble compound treatment and separated as a precipitate. Next, when a complexing agent such as formic acid that binds to uranium to form a complex is added to the eluate after precipitation separation, uranium generates a complex anion that can be adsorbed on an anion exchange resin and separated. The By this separation step, uranium in the eluate is removed to the ppb level.

  Thus, according to the radioactive waste processing method according to the present embodiment, similarly to the third embodiment, by providing a two-stage uranium separation step, the tolerance is higher and safer than the drainage standard value. Drainage can be discharged. Further, since the anion exchange resin can be easily eluted and regenerated, the resin can be reused repeatedly. This anion exchange resin is generally considered to be reusable about 50 to 100 times, and it is possible to greatly reduce the amount of ion exchange resin generated as secondary waste.

  In addition, when an organic complexing agent such as formic acid is used as the complexing agent, it is possible to elute uranium using water as the elution medium, and in the eluent (eluent). Since the mixed organic matter can also be decomposed, there is no generation of secondary waste from the treatment process of the elution / regeneration waste liquid. In addition, since organic complexing agents remaining in the eluate after the uranium separation step can be easily removed by distillation recovery and decomposition treatment, calcium and trace amounts of alkali are removed in the subsequent fluorine separation step. It is possible to produce a calcium fluoride precipitate by adjusting the pH to be added. Therefore, if the organic complexing agent is used, the salt concentration in the wastewater is low, and the influence on the environment can be reduced.

The processing flow figure of the radioactive waste which shows the 1st Example of this invention. The table | surface which shows the difference of the dissolution rate of uranium and sodium fluoride in contrast. The processing flow figure of the radioactive waste which shows the 2nd Example of this invention. The processing flow figure of the radioactive waste which shows the 3rd Example of this invention. The radioactive waste processing flow figure which shows the 4th Example of this invention.

Claims (11)

Contacting waste with uranium with water or weakly acidic water with an acid concentration of 0.1 mol or less to prepare an eluate from which uranium is eluted, and a step of separating and recovering uranium from the eluate A method for treating radioactive waste. The uranium adheres to the waste and contains fluorine, while the separation and recovery step separates and recovers uranium from the eluate and separates and recovers fluorine. Radioactive waste disposal method. 3. The method for treating radioactive waste according to claim 1, wherein uranium is separated from the eluate from which uranium has been eluted by subjecting it to a hardly soluble compound. The method for treating radioactive waste according to claim 3, wherein uranium separation by the hardly soluble compounding treatment and uranium separation by an ion exchange resin treatment are used in combination in the separation and recovery step. The method for treating radioactive waste according to claim 3 or 4, wherein the poorly soluble compound of uranium generated by the hardly soluble compounding treatment is uranyl hydroxide or uranium dioxide. The method for treating radioactive waste according to claim 4, wherein the ion exchange resin used in the ion exchange resin treatment is a chelate resin. In the previous stage of uranium separation by the ion exchange resin treatment, a ligand that combines with uranium to form a complex anion is added to the eluate, and the complex anion produced is treated with the anion exchange resin to contain a uranium-containing component. The radioactive waste processing method according to claim 4, wherein: 3. The radioactive waste treatment method according to claim 2, wherein the fluorine is separated and recovered by subjecting fluorine to a hardly soluble compound treatment. 9. The radioactive waste treatment method according to claim 8, wherein the fluorine-insoluble compounding treatment is performed simultaneously with the uranium-insoluble compounding treatment. 2. The method for treating radioactive waste according to claim 1, wherein the waste is a uranium exhaust gas treatment agent to which uranium is adhered, and the uranium exhaust gas treatment agent from which uranium components are eluted and separated is reused. A means for preparing an eluate from which uranium is eluted by contacting waste with uranium attached with water or weakly acidic water having an acid concentration of 0.1 mol or less, and a means for separating and recovering uranium from the eluate are provided. A radioactive waste processing apparatus.

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