JP2938287B2 - Treatment of radioactive liquid waste - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a radioactive waste liquid generated in the decontamination treatment of a nuclear power plant or an eluent generated in the process of eluting a used ion exchange resin used for treating the radioactive waste liquid. The present invention relates to a method for treating a radioactive waste liquid which is reduced in volume and solidified.

[0002]

2. Description of the Related Art In many cases, radioactive effluents generated from nuclear power plants contain radionuclides in the effluents in the form of metal complex salts. For example, a decontamination treatment waste liquid of a plant system includes a chelating agent represented by EDTA (ethylenediaminetetraacetic acid) as a component of a decontamination agent used at the time of decontamination.
Organic acids and inorganic acids, mainly citric acid and oxalic acid,
It contains a reducing agent and other corrosion inhibitors, and such radioactive waste liquid is treated with a radionuclide (metal component: M) such as cobalt, manganese, or iron and a complex with a chelating agent ( ^Yn- ). To form a metal complex salt (for example, MY ^2- ).

[0003] As a method for treating such radioactive liquid waste, nuclides are mixed in the form of ions or metal complex salts by a treatment method known as a coprecipitation treatment method, a coagulation sedimentation method, an evaporation concentration method, a reverse osmosis membrane filter method, or the like. After the radioactive waste liquid is concentrated, the concentrated liquid with a high radioactivity level is solidified by a cement solidification method, etc., and stored as radioactive solid waste at the plant premises. For low-radioactive-level treatment liquids), solidify and dispose of them as low-radioactive-level waste, or remove radioactive substances by chemical treatment so as not to affect the surrounding environment, and then remove rivers, etc. Conventionally, a treatment method for releasing the ash to the air has been practiced.

In the treatment of the above-mentioned decontamination treatment waste liquid, a treatment method of adsorbing radioactive nuclides to the ion exchange resin by an ion exchange method using a cation exchange resin _{and an} anion exchange resin as a mixed bed type is widely adopted. ing. The anion exchange resin does not contain a radionuclide with a medium-long half-life and has a low radioactivity level, whereas the cation-exchange resin has a radionuclide with a medium-length and a half-life having a high radioactivity level (for example, Co-6).
A cation nuclide component such as 0). In this ion-exchange method, the ion-exchange resin used in the treatment system for radioactive waste liquid is generated in large quantities as secondary waste after use. Moreover, the spent ion exchange resin used for the treatment of the radioactive waste liquid as described above, particularly the cation exchange resin, has a high radioactivity concentration as described above. It is required to convert to a stable form and to reduce the volume as much as possible, and as a processing method therefor, the following method is conventionally known.

(1) The radioactive waste resin is directly solidified with a solidifying material such as cement, asphalt, plastic, or the like, and is stored and stored at the site as a high radioactive solid waste. (2) Reduce the volume by treating the radioactive waste resin by direct incineration, thermal decomposition, wet oxidative decomposition, or the like. (3) After the radionuclide is eluted from the waste resin using an eluent, the low-activity resin is mineralized by, for example, incineration, and the eluate containing one radionuclide is subjected to a reverse osmosis membrane after neutralization. After a concentration treatment such as a filter method or an evaporative concentration method, the concentrate is sealed in an inorganic substance such as cement and solidified.

[0006]

However, the above-mentioned conventional method for treating radioactive waste liquid has the following problems. That is, the coprecipitation method and the coagulation method are effective methods for a waste liquid in which a chelating agent is not present. However, in a waste liquid containing a chelating agent, a radionuclide combines with the chelating agent to form a strong metal complex salt. Inhibits coprecipitation and aggregation. For this reason, in order for the coprecipitation method and the coagulation method to function effectively, a treatment step for breaking the bond between the radionuclide and the chelating agent is required in the treatment process. According to the method, depending on the components of the decontamination agent, it may be difficult to treat.In addition, the amount of ion exchange resin used is large, and the used ion exchange resin is discharged in large quantities as secondary waste. Processing and storage management are troublesome.

In a treatment method such as an evaporative concentration method or a reverse osmosis membrane filter method, it is basically possible to concentrate a radionuclide in the form of a metal complex salt (for example, MY ²⁻ ). When treating a radioactive liquid waste having a high radioactivity level, there are the following problems. That is, in general, in the evaporative concentration method and the reverse osmosis membrane filter method, a neutralization treatment by pH adjustment is necessary as a pretreatment. Therefore, when the radioactive waste liquid is concentrated, sodium sulfate generated by the neutralization treatment is contained in the concentrated liquid. Since a large amount of sodium nitrate is contained, and these components are finally transferred to the waste solid as it is and remain there, it is difficult to reduce the volume of the waste solid. Even if the concentrated solution treated by the evaporative concentration method or reverse osmosis membrane filter method is finally solidified with cement, asphalt, plastic, etc., the solidified radioactive waste contains a chelating agent that affects the safety of disposal. Will remain included. (Note: In the United States, the conditions for accepting radioactive waste disposal sites stipulate the allowable amount of chelating agents contained in waste.) Before forming a solid, the radionuclide and chelating agent are complexed. Combustion and high-temperature decomposition methods can be applied to separate and remove the chelate components from the concentrated liquid. However, for concentrated liquids with high radioactivity levels, high levels of chelate components are included in the combustion and pyrolysis exhaust gases. In order to emit radioactivity, an exhaust gas treatment apparatus is required to have a large radioactivity removal treatment capacity, which makes the equipment complicated and large. In addition, even after combustion and thermal decomposition, organic components such as chelating agents remain as residues, and sodium sulfate, sodium nitrate, etc. generated by the neutralization treatment and decomposition residues also remain in the solidified form, so radioactive waste Cannot be reduced sufficiently.

Further, when the used ion exchange resin having a high radioactivity level is treated by the above-mentioned waste resin treatment method, the following problems remain. In other words, in the method of solidifying radioactive waste resin directly with cement or the like, the volume of solidified material is enormous because the volume of the waste resin is not reduced, and its storage management is troublesome. There is concern about ensuring stability during storage for a long period.

Further, in the treatment method of reducing the volume of radioactive waste resin by direct incineration, the exhaust gas and the radioactive concentration in the furnace accompanying the incineration increase, and the exhaust gas treatment system becomes complicated and large.
The handling of incineration residues is troublesome, and the amount of residue generated does not always achieve the expected volume reduction effect depending on the properties of the waste resin and the incineration conditions. Also, in the thermal decomposition method and the wet oxidative decomposition method, a large amount of decomposition residue is generated as in the incineration method.

[0010] Furthermore, the method of eluting radionuclides from waste resin using an eluent to reduce the radioactivity of the waste resin can be achieved by disposing the waste resin after elution as low radioactive waste.
Although there is an advantage that the amount of high radioactive waste generated can be significantly reduced compared to the above-mentioned treatment method, the eluate separated from the waste resin is concentrated by reverse osmosis membrane filter method, evaporative concentration method, etc. Treatment, neutralization process (pH adjustment) during the process,
A large amount of generated sodium sulfate, caustic soda and other components migrate into the concentrate, resulting in an increase in the amount of solidified material with high radioactivity. In addition, the concentrate was finally solidified with cement, asphalt, plastic, etc. Even so, the solidified radioactive waste will still contain chelating agents that affect the safety of disposal.

In addition, co-precipitation method,
The use of a coagulation treatment method is also conceivable, but these enrichment treatment methods are effective when the radionuclide in the waste liquid exists in the ionic form as described above. In order for the nuclide to combine with the chelate component to form a strong metal complex salt and to inhibit coprecipitation and coagulation, a treatment for breaking the bond between the nuclide and the chelate component to decompose the form of the metal complex salt is required.

SUMMARY OF THE INVENTION The present invention has been made in view of the above points, and has as its object to solve the above-mentioned problems and to rationally treat radioactive waste liquids for various radioactive waste liquids containing the metal complex salt described above. It is an object of the present invention to provide a processing method in which a highly radioactive waste liquid generated by the treatment is changed into an inorganic and stabilized solidified product, and the amount of the generated waste is reduced as much as possible.

[0013]

In order to achieve the above-mentioned object, in the treatment method of the present invention, a radioactive waste liquid is subjected to an electrodialysis treatment to separate and recover components such as an eluent from the liquid. The treatment is performed through a treatment step, a second treatment step of concentrating waste liquid after collecting the eluent, and a third treatment step of heat-melting and solidifying the concentrated liquid generated in the preceding step. .

[0014] The second treatment step of concentrating the radioactive waste liquid is performed by adding a pH adjuster, a coprecipitant, and a flocculant to the waste liquid to remove nuclides contained in the eluate by hardly dissolving metal hydroxide. Co-precipitation / aggregation treatment step of solid-liquid separation in place of the material, and membrane separation step of membrane-treating the co-precipitation / aggregation treatment supernatant. In this step, the pH of the waste liquid is preferably adjusted. 10
Adjust to the range of ~ 13.5.

In carrying out the processing method,
There are the following embodiments. (1) In the above-mentioned membrane separation step, membrane treatment is performed using a hollow fiber membrane filter of an ultrafiltration membrane. (2) In the third processing step of heating, melting and solidifying the concentrated liquid,
The solidification treatment is performed using a microwave heating melting solidification method.

[0016]

[Action] Taking the used ion exchange resin used in the treatment system of radioactive waste liquid as an example, first use an eluent (if the waste resin is a cation exchange resin, use sulfuric acid H ₂ SO ₄ and an anion When caustic soda NaOH is used for the exchange resin), radionuclides, such as Co, Mn, and Fe, adsorbed on the waste resin are eluted from the waste resin and transferred into the eluted waste liquid.

Here, the generated elution waste liquid is used as the first
Most of the eluent components are separated and recovered from the eluent containing the radionuclide by electrodialysis in the treatment step, and the concentration of the eluent contained in the remaining eluate waste containing the nuclide is greatly reduced. The eluted components recovered in this processing step are used again as an eluent. In this electrodialysis treatment, an ion exchange membrane (fraction molecular weight: about 100 to 300) is used.
Use This is to minimize the migration of the metal complex salt (having a molecular weight of 300 or more) bound to the radionuclide and EDTA to the recovery liquid side of the eluent.

On the other hand, the remaining eluate containing a radionuclide such as Co, Mn, and Fe is neutralized in the second treatment step to adjust the pH to a range of 10 to 13.5, and then a coprecipitant, By adding the coagulant, nuclides in ionic form are also precipitated as soluble metal hydroxide (colloidal) and concentrated. Further, the supernatant is subjected to a membrane treatment in the next step to solid-liquid separate the remaining metal hydroxide in the liquid. In addition, if a hollow fiber membrane filter of an ultrafiltration membrane is adopted in this membrane treatment step,
Sulfuric acid H ₂ SO ₄ , sodium sulfate Na ₂ SO ₄ , caustic soda NaOH, sodium nitrate NaNO ₃ , which are added and generated in the process of elution and neutralization of waste resin, pass through the filter membrane and are backwashed (concentrated solution) And is processed in a low-radiation processing system together with the processing solution that has passed through the membrane.

Then, the concentrated waste liquid (including the backwash liquid) containing the radionuclide obtained in the above-mentioned concentration step is heated and melt-solidified in the subsequent third treatment step to be solidified. By employing a microwave heating device in this step, the volume of the concentrated waste liquid can be efficiently reduced by heating, and the amount of radioactive transfer to the exhaust gas treatment system can be reduced. In addition, the separation and recovery of the eluent, and the p
Since the separation process of H adjuster, coprecipitant, etc. is performed, only the radionuclide hydroxides and other radioactive substances represented by cladding can be processed in the final heat melting and solidification process, resulting in high radiation. The amount of solids generated at the functional level is minimized.

The waste resin whose radioactivity is reduced by the elution treatment and whose radioactivity is reduced is subjected to volume reduction by incineration and direct solidification after storage and storage as low radioactive waste.

[0021]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of a treatment method according to the present invention will be described below with reference to FIGS. 1 and 2 for an elution waste liquid containing a radionuclide generated in a process of eluting a used ion exchange resin. 1 is a processing step diagram, and FIG. 2 is a system flow diagram of a processing apparatus based on FIG.

First, a highly radioactive waste resin (spent ion exchange resin used for treating a radioactive waste liquid generated from a nuclear power plant) is charged into the resin elution tank 1 by a fixed amount, and then the resin elution tank 1 is charged. When an eluent (acidic eluent such as sulfuric acid is used when the waste resin is a cation exchange resin and alkaline eluent is used when the waste resin is an anion exchange resin) is passed through the eluent, The radionuclide elutes to generate an eluent. This eluent is once transferred to the waste liquid tank 2. It should be noted that the radioactive nuclide-removed waste resin which has been reduced in radioactivity is taken out of the resin elution tank 1 and stored, then reduced in volume by incineration, solidification, etc., and disposed of as low radioactive waste.

On the other hand, the eluent containing radionuclides transferred to the waste liquid tank 2 is led to the electrodialyzer 3 in the first treatment step, where the eluent components are separated and recovered from the eluate. That is, the inside of the tank of the electrodialysis apparatus 3 incorporates an ion exchange membrane (an eluent is sulfuric acid and an anion exchange membrane (anion membrane) A is used in the illustrated example) as a dialysis membrane. While the eluent is flowing in the left chamber in the tank, water is flowed in the right chamber. As a result, ions such as SO 4 2− are transferred from the left chamber to the right chamber at the boundary of the anion exchange membrane A, and are collected in the collection tank 4. On the other hand, radionuclides (Co2 +, Mn2 +, F
Since e2 +) is a cation, it does not pass through the ion exchange membrane and flows out of the left chamber as it is to return to the waste liquid tank 2. Thus, the separation of the sulfuric acid and the nuclide in the eluent is performed.
The recovered liquid recovered from the right chamber into the recovery tank 4 is used again as an eluent after replenishing sulfuric acid corresponding to the unrecovered amount.

Next, in the second treatment step, the eluent containing the nuclides recovered in the waste liquid tank 2 is transferred to a coprecipitation / aggregation treatment apparatus 5, where a pH adjuster (eg, NaOH, Ca (OH)) is used.
₂ ) was added to adjust the pH to a range of 10 to 13.5, and a coprecipitant (eg, FeSO ₄ ) was quantitatively added, mixed and stirred, and allowed to stand. A coprecipitation treatment and a coagulation sedimentation treatment are carried out so as to separate solid-liquid into (metal hydroxide of the nuclide component and clad) and the supernatant. Then, the supernatant is passed through a hollow fiber membrane filter 6 of an ultrafiltration membrane in the next membrane separation step, where the colloidal nuclide component and clad mixed in the supernatant are captured, and the filter is filtered. Nuclide components trapped in the backwash solution (concentrated solution)
And the treated liquid that has passed through the filter is treated as a low radioactive waste liquid. When the concentration of the waste liquid in the above-mentioned coprecipitation / aggregation treatment step is low, the membrane separation treatment may be performed by directly passing the coprecipitation treatment liquid through the hollow fiber membrane filter without performing the aggregation / sedimentation separation operation. It is possible.

Next, the sediment concentrate of the coprecipitation / coagulation treatment device 5 and the backwashing solution of the hollow fiber membrane filter 6 which have been concentrated and separated in the previous step are transferred to the microwave heating device 7 in the third treatment step. Here, an evaporation-drying-decomposition-melting-solidification treatment is performed. As a result, the concentrated liquid evaporates to reduce its volume, and if necessary, a solidifying material such as glass is added to obtain an inorganic and stabilized solidified substance with high radioactivity. The exhaust gas generated in this process is processed in an exhaust gas processing system. In addition, the processing system shown in FIG. 2 can perform an automatic operation in accordance with a waste resin processing process.

Next, a simulation test liquid equivalent to the radioactive elution waste liquid obtained by actually eluting the used ion exchange resin used for the treatment of the radioactive waste liquid in a nuclear power plant was prepared and used as a sample according to the present invention. The results of evaluation tests on the treatment methods performed by persons and the like are described below. First, assuming that a 1N (standard) sulfuric acid aqueous solution was used as a waste resin elution waste solution, the simulation test solution had a sulfuric acid concentration of 49 g / l, and a radionuclide component of Co and Mn ion concentrations of 5 mg / l. The Fe ion concentration was adjusted to 500 mg / l.

In the electrodialysis membrane processing apparatus, the anion exchange membrane (molecular weight cut off 300, effective membrane area 400 c
m ² ), the simulated test solution was dialyzed (acid recovery treatment) using an apparatus having a processing volume of 10 l per time. As a result, the recovery rate of sulfuric acid H ₂ SO ₄ was 90%, and the C
o, Mn ions are 5% or less (0.25 mg / l or less), Fe
The result of the ion was 10% or less (50 mg / l or less). The sulfuric acid concentration of this recovered solution was about 44 g / l,
-H The use of ₂ SO ₄ eluents are possible reuse as by eluent in adding additional sulfuric acid to about 5g. On the other hand, the remaining eluate containing nuclides has a sulfuric acid concentration of about 5 g.
/ L, Co, Mn ion concentration was 4.8 mg / l, and Fe ion concentration was 450 mg / l.

Next, the eluate after the above-mentioned acid recovery treatment was transferred to a coprecipitation / aggregation treatment device, and the pH value was adjusted to 12 by adding caustic soda NaOH as a pH adjuster. 100 mg / l of ferrous iron (FeSO ₄ )
After adding, stirring and mixing, the mixture was allowed to stand to separate into a concentrated precipitate and a supernatant, and the supernatant was passed through a hollow fiber membrane filter (fraction molecular weight: 150,000) of an ultrafiltration membrane to be subjected to membrane treatment. .

When the properties of the treatment liquid obtained by this membrane separation treatment were examined, the following results were obtained. Co, Mn ion concentration: 0.01 mg / l or less Fe ion concentration: 0.05 mg / l or less NaSO ₄ concentration: 7.2 g / l NaOH concentration: 0.4 g / l Therefore, coprecipitation treatment-membrane separation treatment The decontamination coefficient DF of nuclide components by Co is about 480 or more for Co and about 1000 for Fe.
The above is obtained.

On the other hand, a concentrated concentrate obtained by coprecipitation / aggregation treatment,
The backwash liquid (concentrated liquid) of the hollow fiber membrane filter was heated and melt-solidified using a microwave heating device.
The components and amounts of the concentrated liquid to be treated here are shown below. Co ion concentration: about 480 mg / l Mn ion concentration: about 480 mg / l Fe ion concentration: about 45000 mg / l NaSO ₄ concentration: 7.2 g / l NaOH concentration: 0.4 g / l Assuming that 100 m ^{3 is} treated annually, the results of trial calculation of the amount of solidified matter generated by the above treatment are as follows. However, NaSO ₄ , NaO
In calculating the concentration of H, a trial calculation was performed assuming two cases where the concentration ratio of the concentrated solution (precipitate solution, backwash solution) was 50 times and 100 times.

Solid content of Co: 480 g Solid content of Mn: 480 g Solid content of Fe: 45 kg Solid content of NaSO ₄ : 7.2 kg (concentration ratio 100 times,
Generation amount 1 m ³⁾ 14.4 kg (concentration rate 50 times, the amount 2m ³⁾ solid content of the NaOH: 0.4 kg (concentration rate 100 times,
Generated amount 1m ³ ) 0.8kg (concentration ratio 50 times, generated amount 2m ³ ) However, NaSO ₄ and NaOH are not decomposed by microwave heating treatment (heating temperature of 100 ° C or less) and migrate into the solidified body. Was calculated.

Next, as a comparative example of the processing method of the present invention,
The same simulated test solution as described above was treated by the conventional treatment method shown in FIG. 3, and the properties of the eluent generated during the treatment process and the components and amounts of solidified substances were investigated and estimated. The result is as follows. (1) Amount of NaSO ₄ generated by pH adjustment (neutralization): (a) When the amount of NaSO _{4 in} the eluate is 49 g / l, the concentration of NaSO ₄ becomes 70.6 g / l.

(B) The amount of NaSO _{4 in} the eluent is 90%
When 4.9 g / l is recovered, the NaSO ₄ concentration is 7.9.
It becomes 1 g / l. (When the acid recovery process in the eluent is performed on the upstream side) (2) Assuming that the concentration ratio in the concentration process is 100 and the decontamination coefficient DF of the concentration method is 100 or more,
The properties of the processing solution and the concentrated solution are as follows.

Properties of treatment liquid: Co ion concentration: about 0.05 mg / l or less Mn ion concentration: about 0.05 mg / l or less Fe ion concentration: about 5 mg / l or less NaSO ₄ concentration: 0.76 g / l or less [ Case of the above (a)] 0.071 g / l or less [Case of the above (b)] Properties of concentrated solution: Co ion concentration: about 500 mg / l Mn ion concentration: about 500 mg / l Fe ion concentration: about 50,000 mg / l NaSO ₄ concentration: 7060 g / l [case (a) above] 706 g / l [case (b) above] Further, if the concentration of the eluent is assumed to be 100 m ³ per year and the concentration magnification is 100 times, the concentration treatment is carried out. Of processing solution by
Since m ³ and the amount of the concentrated solution are 1 m ³ , the components and amounts of the concentrated solution to be subjected to the next solidification treatment are as follows.

The amount of Co component: of 7060kg [said (a) of the case] 706kg [see (b): an amount of about 0,5Kg Mn components: an amount of about 0,5Kg Fe component: an amount of about 50 kg NaSO ₄ components Case) In order to solidify the concentrated solution directly with cement or the like, the concentrated solution is 1 m ³ , and when the concentrated solution is solidified by microwave heating, the water content of the concentrated solution is 1 m ^3.
Evaporates and is removed from the components of the solidified body.

Based on the above data, a trial calculation of the solidified body generation amount with the comparative example (conventional method) is as follows. Co of solid content: solid content of about 0.50 kg Mn: solid content of about 0.50 kg Fe: solid content of about 50kg NaSO _4: 7060kg [Case of (a) above] [wherein (a) Case] 706kg and As for the solid content of NaSO ₄ occupying the majority of the solidified product, the treatment method of the present invention described above and the above comparative example (conventional method) were compared with the method of the present invention / conventional example in a ratio of 7.2 kg / 7060kg ≒ 1/1000, 14.4kg / 70
60kg ≒ 1/500 7.2kg / 706kg ≒ 1/100, 14.4kg / 70
According to the treatment method of the present invention, the effect of reducing the volume of the solidified body by 50% is obtained as compared with the comparative example (conventional method), depending on the treatment conditions.
It can be seen that the improvement is up to 1000 times.

[0037]

As described above, a process for eluting radioactive nuclides from a radioactive waste liquid containing a metal complex salt generated in the decontamination treatment of a nuclear power plant or a spent ion exchange resin used in a treatment system for the radioactive waste liquid. According to the present invention, in the first treatment step, a waste liquid containing a radioactive nuclide component is subjected to an electrodialysis treatment, and a component such as an eluent is treated in the first treatment step. Is separated,
After the recovery, the waste liquid was co-precipitated / coagulated in a second treatment step, concentrated by membrane separation treatment, and further melt-solidified by microwave heating in a third treatment step.
The radioactive liquid waste containing high radionuclide can be treated by converting it into a reasonable and minimal amount of stable solidified product. As a result, it is possible to obtain a profit that greatly contributes to reducing the volume of highly radioactive waste stored and stored at the site of the nuclear power plant.

[Brief description of the drawings]

FIG. 1 is a process diagram of a method for treating a radioactive waste resin eluate according to an embodiment of the present invention.

FIG. 2 is a system flow diagram of a processing apparatus used for implementing the processing method shown in FIG. 1;

FIG. 3 is a process chart of a conventional method cited as a comparative example of the present invention.

[Explanation of symbols]

 Reference Signs List 1 resin elution tank 3 electrodialysis device 5 coprecipitation / coagulation treatment device 6 hollow fiber membrane filter 7 microwave heating device

──────────────────────────────────────────────────続 き Continuing on the front page (72) Inventor Yuichi Tsukamoto 3 in Myojin-cho, Tsuruga-shi, Fukui Power Reactor and Nuclear Fuel Development Corporation Inside the new conversion reactor Fugen Power Plant (72) Inventor Yasuo Oikawa Arata Tanabe, Kawasaki-ku, Kawasaki-shi, Kanagawa Prefecture No. 1-1 Fuji Electric Co., Ltd. (72) Inventor Seiji Mizukoshi 1-1, Tanabe Nitta, Kawasaki-ku, Kawasaki City, Kanagawa Prefecture Inside Fuji Electric Co., Ltd. No. 1-1 Fuji Electric Co., Ltd. (56) References JP-A-62-62297 (JP, A) JP-A-50-60700 (JP, A) (58) Fields investigated (Int. Cl. ⁶ , (DB name) G21F 9/06

Claims (4)

(57) [Claims] 1. A radioactive waste liquid containing a metal complex salt generated by a system decontamination treatment of a nuclear power plant, or an eluent containing a radionuclide generated in an elution process of a used ion exchange resin used for the treatment of the radioactive waste liquid, etc. A method for treating a radioactive waste liquid, wherein the radioactive waste liquid is subjected to an electrodialysis treatment to separate and recover components such as an eluent from the liquid.
, A second processing step of concentrating a waste liquid containing a radionuclide after recovering the eluent, and a third processing step of heating and melting the solidified liquid generated in the previous step to solidify it. In the method, the second treatment step comprises adding a pH adjuster, a coprecipitant, and a flocculant to the waste liquid, converting radionuclides contained in the eluate into hardly soluble metal hydroxide, and solid-liquid A method for treating a radioactive waste liquid, comprising: a coprecipitation / aggregation treatment step of separating; and a membrane separation step of subjecting the supernatant of the coprecipitation / aggregation treatment to a membrane treatment. 2. The method for treating radioactive waste liquid according to claim 1, wherein the waste liquid is adjusted to have a pH in the range of 10 to 13.5. 3. The method for treating radioactive waste liquid according to claim 1, wherein in the membrane separation step, membrane treatment is performed using a hollow fiber membrane filter of an ultrafiltration membrane. 4. The method for treating a radioactive waste liquid according to claim 1, wherein the solidification treatment is carried out using a microwave heating melting solidification method in the third treatment step.